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{{#Wiki_filter:ENCLOSURE 1 REVISED EMERGENCY PLAN (E-PLAN)
{{#Wiki_filter:}}
IMPLEMENTING PROCEDURES (EPIPs)
REC/RPSS HANDBOOK Emergency Plan, Revision 58 (176 pages)
F3.17 - Core Damage Assessment, Revision 16 (52 pages)
F3.17.1 - Core Damage Determination, Revision 6 (21 pages)
REC/RPSS Handbook, Revision 26 (178 pages)
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT                            EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                REV:    58 Page i APPROVAL:
PCR #: 602000019099
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                    EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                        REV:    58 Page ii Record of Revision Date of        Revision Revision        Number Title Page i                                            March 2021          58 Record of Revision Page ii                              March 2021          58 Revise table 7 for 1(2) R70 and 71                      March 2021          58
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                                            EMERGENCY PLAN NUMBER:
E-PLAN EP                                  EMERGENCY PLAN                                                              REV:      58 Page 1 of 164 TABLE OF CONTENTS Section                                              Title                                                            Page TITLE PAGE ................................................................................................................... i RECORD OF REVISION.................................................................................................. ii TABLE OF CONTENTS .................................................................................................. 1 1.0      DEFINITIONS ....................................................................................................... 7 2.0      SCOPE AND PURPOSE .................................................................................... 11 3.0     
 
==SUMMARY==
......................................................................................................... 13 4.0      EMERGENCY CONDITIONS ............................................................................. 15 4.1    Classification System ............................................................................... 15 4.1.1  Notification of Unusual Event ....................................................... 15 4.1.2  Alert.............................................................................................. 16 4.1.3  Site Area Emergency ................................................................... 16 4.1.4  General Emergency ..................................................................... 17 5.0      ORGANIZATONAL CONTROL OF EMERGENCIES ......................................... 19 5.1    Normal Site Organization ......................................................................... 19 5.2    Normal Plant Organization ....................................................................... 19 5.3    Plant Emergency Organization ................................................................ 20 5.3.1  Direction and Coordination........................................................... 20 5.3.2  Plant Emergency Organization Coordinators ............................... 32 5.3.3  Plant Shift Organization ............................................................... 35 5.3.4  Plant Emergency Staff Augmentation Groups.............................. 38
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                                            EMERGENCY PLAN NUMBER:
E-PLAN EP                                  EMERGENCY PLAN                                                              REV:      58 Page 2 of 164 TABLE OF CONTENTS (Cont'd)
Section                                              Title                                                            Page 5.4    EOF Organization .................................................................................... 41 5.4.1  EOF Direction and Control ........................................................... 41 5.4.2  EOF Technical Support Group ..................................................... 43 5.4.3  EOF Radiation Protection Support Group .................................... 43 5.4.4  EOF General Support Staff .......................................................... 44 5.5    Recovery Organization ............................................................................ 45 5.6    Augmentation of Plant and EOF Emergency Organizations .................... 45 5.6.1  Offsite Support Response ............................................................ 45 5.6.2  Monticello Radiation Protection Group Support ........................... 46 5.6.3  Westinghouse Support ................................................................. 46 5.6.4  Local Support Services ................................................................ 47 5.7    Coordination with Governmental Response Organizations...................... 50 5.7.1  Minnesota Division of Homeland Security and Emergency Management (HSEM) .................................................................. 50 5.7.2  Minnesota Department of Health (MDH) ...................................... 50 5.7.3  Wisconsin Emergency Management ............................................ 51 5.7.4  Wisconsin Department of Health Services (DHS) ........................ 51 5.7.5  Goodhue, Dakota and Pierce County Sheriffs ............................. 52 5.7.6  Goodhue, Dakota, Pierce County and City of Red Wing Emergency Management ............................................................. 52 5.7.7  Prairie Island Indian Community .................................................. 52 5.7.8  Minnesota State Patrol ................................................................. 53 5.7.9  Minnesota Department of Transportation ..................................... 53 5.7.10 Canadian Pacific Railway-CP Railway (Soo Line) ....................... 53 5.7.11 Burlington Northern Santa Fe (BNSF) Railway ............................ 53 5.7.12 Department of the Army, Corps of Engineers, Lock & Dam
                        #3 ................................................................................................. 53 5.7.13 Nuclear Regulatory Commission (NRC)....................................... 54 5.7.14 Department of Energy (DOE) ....................................................... 55 5.7.15 Institute Of Nuclear Power Operations (INPO) ............................. 55
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                                        EMERGENCY PLAN NUMBER:
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Section                                          Title                                                            Page 6.0      EMERGENCY MEASURES ............................................................................... 57 6.1    Activation of Emergency Organization ..................................................... 57 6.1.1  Activation of Plant and EOF Organizations .................................. 57 6.1.2  Notification Scheme ..................................................................... 58 6.1.3  Communicators ............................................................................ 60 6.1.4  Authentication of Emergency Communications ............................ 64 6.2    Record Keeping ....................................................................................... 64 6.3    Summary of Site Response Actions ........................................................ 65 6.4    Assessment Actions................................................................................. 68 6.4.1  Dose Projections .......................................................................... 68 6.4.2  Radiological Surveys.................................................................... 73 6.5    Corrective Actions .................................................................................... 74 6.6    Protective Actions .................................................................................... 74 6.6.1  Evacuation and Sheltering ........................................................... 74 6.6.2  Use of Protective Equipment and Supplies .................................. 78 6.6.3  Contamination Control Measures ................................................. 84 6.7    Aid to Affected Personnel ........................................................................ 86 6.7.1  Emergency Personnel Exposure .................................................. 86 6.7.2  Decontamination and First Aid ..................................................... 87 6.7.3  Medical and Public Health Support .............................................. 88 6.7.4  Whole Body Counting Facilities ................................................... 89 7.0      EMERGENCY FACILITIES AND EQUIPMENT.................................................. 91 7.1    Emergency Control Centers..................................................................... 91 7.1.1  Technical Support Center (TSC) .................................................. 91 7.1.2  Operational Support Center (OSC) .............................................. 96 7.1.3  Emergency Operations Facility (EOF ........................................... 98 7.1.4  Control Room ............................................................................. 103
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                                          EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                                              REV:      58 Page 4 of 164 TABLE OF CONTENTS (Cont'd)
Section                                            Title                                                            Page 7.2    Communications .................................................................................... 104 7.2.1  Onsite Communications ............................................................. 104 7.2.2  Offsite Communications ............................................................. 105 7.2.3  Public Alert and Notification System (PANS) ............................. 109 7.3    Assessment Facilities ............................................................................ 110 7.3.1  Onsite Systems and Equipment ................................................. 110 7.3.2  Facilities and Equipment for Offsite Monitoring .......................... 131 7.4    Protective Facilities and Equipment ....................................................... 133 7.4.1  Assembly Points ........................................................................ 133 7.4.2  Operational Support Center ....................................................... 133 7.4.3  Emergency Operations Facility .................................................. 133 7.4.4  Mayo Clinic Health System ........................................................ 134 7.4.5  Red Wing Fire Station ................................................................ 134 7.4.6  Technical Support Center Emergency Locker............................ 134 7.5    First Aid and Medical Facilities .............................................................. 134 7.6    Damage Control Equipment and Supplies ............................................. 134 8.0      MAINTAINING EMERGENCY PREPAREDNESS ............................................ 135 8.1    Organizational Preparedness ................................................................ 135 8.1.1  Emergency Response Training .................................................. 135 8.1.2  Exercises, Drills, and Tests ........................................................ 136 8.2    Review and Updating of the Plan and Procedures ................................ 139 8.2.1  Organization of Plan................................................................... 139 8.2.2  Maintenance and Inventory of Emergency Equipment and Supplies ..................................................................................... 140
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                                          EMERGENCY PLAN NUMBER:
E-PLAN EP                                        EMERGENCY PLAN                                                      REV:      58 Page 5 of 164 TABLE OF CONTENTS (Cont'd)
Section                                                      Title                                                  Page 9.0      RECOVERY ..................................................................................................... 141 9.1    Investigation of Incidents ....................................................................... 141 9.2    Recovery Procedures ............................................................................ 141 9.3    Criteria for Resumption of Operations ................................................... 142 9.4    Long Term Recovery ............................................................................. 142 LIST OF ATTACHMENTS Attachment A ...................................................... Emergency Plan Implementing Procedures                    145 Attachment B ....................................................................... Summary of Emergency Supplies            150 Attachment C ......................................................... NUREG-0654/PI E-Plan Cross Reference                    154 LIST OF FIGURES Figure 1 Prairie Island Plant Emergency Organization ................................................ 22 Figure 2 Prairie Island EOF Organization.................................................................... 42 Figure 3 Prairie Island Onsite/Offsite Emergency Organization Interface ................... 61 Figure 4 HAB Communications ................................................................................... 63 Figure 5 Containment Dose Rate Versus Time ........................................................... 71 Figure 6 Plan View of TSC .......................................................................................... 93 Figure 7 Plan View of EOF Command Center ........................................................... 101
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                                          EMERGENCY PLAN NUMBER:
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Section                                            Title                                                            Page LIST OF TABLES Table 1      Guidance for Augmentation of Plant Emergency Organization ..................... 25 Table 2      Initial Protective Action Recommendation During a General Emergency ...... 79 Table 3 Recommended Protective Action to Avoid External and Internal Dose from Exposure to a Gaseous Plume ................................................................... 81 Table 4      Contamination Limits ..................................................................................... 84 Table 5      Prairie Island Site Communications Matrix .................................................. 108 Table 6      Seismographic Monitoring Devices ............................................................. 117 Table 7      Radiation Monitors ...................................................................................... 118 Table 8      Radiation Monitoring Instruments and Devices ........................................... 123 Table 9      Instruments Available for Monitoring Major Systems .................................. 124 Table 10 Offsite Meteorological Equipment ............................................................... 132 LIST OF ANNEX ANNEX A -Emergency Action Level Matrix (attached)
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                  EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                      REV:      58 Page 7 of 164 1.0    DEFINITIONS Listed below are some terms in this plan along with the definitions that should be applied to these terms when they are used in this plan.
1.1    Assessment Action - Actions taken during or after an accident to obtain and process information necessary to make decisions regarding emergency measures.
1.2    Corrective Actions - Emergency measures taken to terminate an emergency situation at or near the source in order to prevent or minimize a radioactive release, e.g., shutting down equipment, firefighting, repair and damage control, etc.
1.3    Emergency Action Level (EAL) - A predetermined, site-specific, observable threshold for a plant Initiating Condition (IC) that places the plant in a given emergency class. An EAL can be: an instrument reading; an equipment status indicator; a measurable parameter (onsite or offsite); a discrete, observable event; results of analyses; entry into specific emergency operating procedures; or another phenomenon which, if it occurs, indicates entry into a particular emergency class.
1.4    Emergency Class: - One of a minimum set of names or titles, established by the Nuclear Regulatory Commission (NRC), for grouping of normal nuclear power plant conditions according to (1) their relative radiological seriousness, and (2) the time sensitive onsite and off site radiological emergency preparedness actions necessary to respond to such conditions. The existing radiological emergency classes, in ascending order of seriousness, are called: Notification of Unusual Event (UE), Alert, Site Area Emergency (SAE), and General Emergency (GE).
1.5    Emergency Director (ED) - The Plant Manager or designee. This individual has overall responsibility and authority for managing the emergency effort within the plant. This person will also manage efforts external to the plant until the Emergency Operations Facility (EOF) Organization can relieve the ED of external tasks.
1.6    Emergency Manager (EM) - A designated member of site management. This person has the authority and responsibility for the management of (NSPM) Northern States Power Company - Minnesota overall response to an emergency. The EM will assume command and control at the Emergency Operations Facility and direct the NSPM response efforts.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                  EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                    REV:      58 Page 8 of 164 1.7    Emergency Planning Zones - a defined area around the plant to facilitate emergency planning by state and local authorities, to assure that prompt and effective actions are taken to protect the public in the event of a release of radioactive material. It is defined for:
1.7.1      Plume Exposure Pathway - a 10 mile radius around the plant where the principal exposure source is: (1) whole body exposure to gamma radiation from the plume and from deposited material; and (2) internal exposure from the inhaled radionuclides deposited in the body (Short Term Exposure).
1.7.2      Ingestion Exposure Pathway - a 50 mile radius around the plant where the principal exposure would be from the ingestion of contaminated water or foods such as milk or fresh vegetables (Long Term Exposure). The ingestion exposure pathway includes the plume exposure pathway.
1.8    Emergency Worker - Any individual involved in mitigating the consequences of an emergency situation and/or minimizing or preventing exposure to the offsite population. The emergency worker category includes emergency workers at the plant as well as individuals who are engaged in public service emergency activities -
firemen, policemen, medical support, and certain public officials. These are people who voluntarily place themselves as emergency workers.
1.9    Exclusion Area - The area surrounding the plant that is under direct Prairie Island Nuclear Generating Plant control. This includes the Corps of Engineering land north of plant and the islands located in the Mississippi River east of plant. It is sized such that any individual located on its boundary would not exceed 25 REM whole body or 300 REM thyroid from I-131 for two hours immediately following the design basis accident (approximately 2340 feet out to boundary).
1.10    Facility Activation - An Emergency Response Facility is activated when the minimum staff per Figures 1 and 2 is available and the facility is ready to assume its assigned Emergency Plan functions and relieve the on-shift staff of those functions.
Although the facility may be ready, the on-shift staff relief may be postponed in the interest of completing critical tasks prior to turnover.
1.11    Initiating Condition (IC): - One of a predetermined subset of nuclear power plant conditions when either the potential exists for a radiological emergency, or such an emergency has occurred.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                    REV:      58 Page 9 of 164 1.12    Northern States Power Company - Minnesota (NSPM) d/b/a Xcel Energy -
Operator of the Prairie Island Nuclear Generating Plant.
1.13    Protective Actions - Emergency measures taken before or after a release of radioactive materials in order to prevent or minimize radiological exposures to the population.
1.14    Protective Action Guides (PAG) - Projected dose to individuals, that warrants protective action prior to and/or following a radioactive release.
1.15    Recovery Actions - Actions taken after an emergency to restore the plant to normal.
1.16    Xcel Energy - Operating Utility of Northern States Power Company - Minnesota.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                              EMERGENCY PLAN NUMBER:
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                  EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                                      REV:      58 Page 11 of 164 2.0    SCOPE AND PURPOSE In accordance with license conditions, 10CFR Part 50, and NRC guidance, the Northern States Power Company - Minnesota (NSPM) has developed and implemented a radiological emergency response plan for the Prairie Island Nuclear Generating Plant (PINGP) and a joint off-site plan for the PINGP and the Monticello Nuclear Generating Plant. As asset owner NSPM, and Xcel Energy, the operating utility, retain all owner obligations.
This Emergency Plan is applicable to Prairie Island Nuclear Generating Plant (PINGP),
Units 1 and 2.
In any emergency situation at Prairie Island, the initial response to activate the Emergency Plan is accomplished by the plant staff and, if needed, immediate actions may be required by local support agencies. The plant, during initial stages of the emergency situation, must function independently coordinating both onsite and offsite activities. The augmented response organization will assume those tasks external (offsite) to the plant, thus allowing the plant staff to be responsible for all onsite activities. This plan covers the actions and responsibilities of the PINGP Emergency Organization and the Emergency Operations Facility Organization.
The purpose of the plan is to describe the following:
2.1    Organization and actions within the plant to control and limit the consequences of an accident.
2.2    Organization and actions controlling site and offsite activities in the event of an uncontrolled release of radioactive material. This includes notification of and coordination with required offsite support agencies.
2.3    Identifying and evaluating the consequences of accidents that may occur and affect the public and plant personnel.
2.4    Describing the protective action levels and actions that are required to protect the public and plant personnel in the event of an accident.
2.5    Consideration necessary for the purpose of reentry and short-term recovery.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                                      REV:      58 Page 12 of 164 2.6    Arrangements required for medical support in the event of injury.
2.7    Arrangements required for fire fighting support in the event of major fires requiring outside support.
2.8    The training necessary to assure adequate response to emergencies.
The Emergency Plan is dependent upon various standing plant operating, abnormal operating, emergency operating, plant safety, radiological control and security procedures and the Emergency Plan Implementing Procedures for the implementation of the plants response to the spectrum of emergency situations.
PINGP has procedures in place that implement on-site protective actions and personnel accountability during security events that are appropriate for plant and environmental conditions.
Coordination between plant, state, local and tribal authorities is defined in the Minnesota and Wisconsin state emergency operations plans, Goodhue, Dakota and Pierce county emergency plans and the Prairie Island Indian Communitys emergency plan. Goodhue, Dakota and Pierce Counties have, formulated for their respective areas, individual evacuation plans which are included in the respective state plans.
Monticello & Prairie Island (MT & PI) offsite response is detailed in the Corporate Nuclear Emergency Plan.
 
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==SUMMARY==
 
Abnormal events, both realized and potential, requiring emergency preparedness response are classified into four classes of Emergency Action Levels. The four levels of emergency classes, in increasing order of severity are:
3.1    Notification of Unusual Event (UE) 3.2    Alert 3.3    Site Area Emergency (SAE) 3.4    General Emergency (GE)
Each class requires specific immediate actions on the part of the plant staff in order to protect the public, plant personnel and property. As the severity level of the emergency increases, so does the response of the offsite agencies, in order to protect the public.
The lowest class (least severe) is the Notification of Unusual Event, and will be handled mainly by plant personnel, with only advisory notification to local and state authorities. The Alert Classification requires prompt notification of local and state authorities, which will place their various organizations in a standby mode. In both the Notification of Unusual Event and the Alert Classification, the plant staff is expected to restore the situation to normal without further or minimum involvement of offsite authorities. The two higher severity classes, the Site Area and General Emergency, (the General Emergency being the most severe), requires prompt notification of offsite authorities with immediate involvement of those organizations to assess the emergency situation and to implement the required protective actions for the general public.
During an Alert, Site Area, or General Emergency, Prairie Island Nuclear Generating Plant will automatically activate their site and offsite support emergency response organizations.
The normal site organization will staff the Plant Emergency Response Organization and the Emergency Operations Facility (EOF) Organization. The offsite organization will be staffed by members of the MT & PI Offsite Organization and be located at the Minnesota Emergency Operations Center. MT & PI Offsite Organization will communicate to the plant via the EOF Organization. The EOF Organization will support emergency response for the plant and relieve plant personnel of offsite activities who may be needed for plant activities.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                                      REV:      58 Page 14 of 164 When plant conditions stabilize and the potential for future degradation of plant conditions is small, the plant may terminate the emergency classification. If severe equipment or core damage has occurred, a transition to a recovery phase may be warranted. In general terms, an Unusual Event or Alert may be terminated without transition to Recovery while a Site Area Emergency or General Emergency will probably necessitate a planned transition to Recovery and the establishment of a Recovery Organization. The Recovery Organization will manage the overall recovery or post-accident outage plans as work is done to return the plant to a normal operational or shutdown status.
PINGP has and maintains the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an EAL has been exceeded. Upon identification of the appropriate emergency classification level the emergency condition will be promptly declared.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                    EMERGENCY PLAN NUMBER:
E-PLAN EP                                  EMERGENCY PLAN                                      REV:      58 Page 15 of 164 4.0    EMERGENCY CONDITIONS 4.1    Classification System Four Emergency Classification Levels (ECLs) are established, according to severity, taking into consideration potential as well as actual events in progress.
Initiating Conditions (ICs) are predetermined subset of plant conditions when either the potential exists for a radiological emergency, or such an emergency has occurred.
Emergency Action Levels (EALs) are plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions.
Annex A contains the Emergency Action Level (EAL) scheme as established by NEI 99-01, Revision 6.
It should be noted that various events could require a graded scale of response. A minor incident could increase in severity and advance to the next class of emergency. This Emergency Plan is constructed to provide for a smooth transition from one class to another.
4.1.1      Notification of Unusual Event (UE)
Notification of Unusual Events are events that are in process or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.
No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.
The purpose of the Notification of Unusual Event action level is to: (1) have the operating staff come to a state of readiness from the standpoint of emergency response in the event the handling of the initiating condition requires escalation to a more severe action level class; and (2) provide for systematic handling of unusual event information, i.e., to provide early and prompt notification of minor events which could lead to more serious consequences given operator error or equipment failure or which might be indicative of more serious conditions which are not yet fully realized.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                  EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                                      REV:      58 Page 16 of 164 4.1.2  Alert At the Alert action level, events are in process or have occurred which involve actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION.
It is the lowest level when some necessity for emergency planning and response offsite is necessary. Any radioactive release will be limited to a small fraction of the EPA Protective Action Guideline exposure levels.
The purpose of the Alert action level is to: (1) assure that emergency personnel are readily available to respond if the situation becomes more serious or to perform confirmatory radiation monitoring if required; and (2) provide offsite authorities current status information, i.e., early and prompt notification of minor events which could lead to more serious consequences given operator error or equipment failure or which might be indicative of more serious conditions which are not yet fully realized.
4.1.3  Site Area Emergency The Site Area Emergency action level describes events that are in process or have occurred which involve actual or likely major failure of plant functions needed for protection of the public or HOSTILE ACTION that result in intentional damage or malicious acts; (1) toward site personnel or equipment that could lead to the likely failure of or; (2) that prevent effective access to equipment needed for the protection of the public. It reflects conditions where significant offsite releases are likely to occur or are occurring but where a core melt situation is not expected although severe fuel damage may have occurred. Any radioactive releases are not expected to exceed the EPA Protective Action Guideline exposure levels except near the site boundary.
The purpose of the Site Area Emergency action level is to: (1) assure that response centers are manned; (2) assure that monitoring teams are dispatched; (3) assure that personnel required for evacuation of near-site areas are at duty stations if the situation becomes more serious; (4) provide current information for and consultation with offsite authorities; and (5) provide updates for the public through offsite authorities.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                                      REV:      58 Page 17 of 164 4.1.4  General Emergency The General Emergency action level describes events in process or have occurred which involve actual or imminent substantial core degradation or melting with the potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.
Radioactive releases can be reasonably expected to exceed the EPA Protective Action Guidelines exposure levels offsite for more than the immediate site area.
The purpose of the General Emergency class is to: (1) initiate predetermined protective actions for the public; (2) provide continuous assessment of information from licensee and offsite organization measurements; (3) initiate additional measures as indicated by actual or potential releases; (4) provide consultation with offsite authorities; and (5) provide updates for the public through offsite authorities.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                              EMERGENCY PLAN NUMBER:
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                  EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                    REV:      58 Page 19 of 164 5.0    ORGANIZATIONAL CONTROL OF EMERGENCIES 5.1    Normal Site Organization The normal site organization is comprised of the plant organization and several other site support organizations. The normal site organization can be accessed on the Prairie Island web page. Responsibilities and authorities of the various functional groups are delineated in plant Administrative Work Instructions.
5.2    Normal Plant Organization The normal plant operating crew is staffed and qualified to perform all actions that may be necessary to initiate immediate protective actions and to implement the emergency plan and is designated as the responsible group for such actions. The normal plant organization can be accessed on the Prairie Island web page.
The Plant Manager has overall responsibility for the safe, efficient operation of the plant and for compliance with operating license requirements. The Plant Manager SHALL select, train and supervise a qualified staff.
The Shift Manager reports directly to the Assistant Operations Manager who reports directly to the Operations Manager who reports directly to the Plant Manager. The Shift Manager is responsible for the direction and coordination of the Shift Supervisors on his/her shift to perform operations in accordance with the administrative controls and operating procedures. The Shift Manager coordinates activities with other plant groups as required to maintain the safe operation of the plant.
The Shift Supervisor reports to the Shift Manager. The Shift Supervisor is the single focal point for directing and coordinating the operations group, maintenance group and the plant security activities during his/her shift. The Shift Supervisor SHALL assume the primary management responsibility for the safe operation of the plant under all conditions during his/her shift. The responsibility and authority of the Shift Supervisor SHALL be to maintain the broadest perspective of operational conditions affecting the safety of the plant as a matter of highest priority at all times when on duty in the Control Room.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                  EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                      REV:      58 Page 20 of 164 5.3    Plant Emergency Organization A plant emergency organization is designated to augment the normal operating crew. Provisions have been made for rapid assignment of plant personnel to the plant emergency organization during emergency situations. The Prairie Island Plant Emergency Organization is shown in Figure 1.
Various areas of responsibility are assigned to segments of the plant staff during emergency situations as depicted in Table 1. Table 1 shows the personnel available on-shift and the capability for additional personnel within 60 minutes and 90 minutes of event declaration. Table 1 follows the guidance developed in accordance with 10 CFR 50 Appendix E. This staffing analysis is documented in F3-1.1, Emergency Plan On-Shift Staffing.
5.3.1    Direction and Coordination During the initial stages of an emergency condition at Prairie Island Nuclear Generating Plant, the Emergency Director has overall coordinating authority for Northern States Power Company - Minnesota (NSPM). The Emergency Director alone has the authority and responsibility to immediately initiate any emergency actions, including providing protective action recommendations to offsite authorities responsible for implementing offsite emergency measures.
The Shift Supervisor, of the affected unit, until properly relieved, SHALL remain in the Control Room at all times during accident situations, to direct the activities of control room operators. If necessary, the Shift Supervisor of the unaffected unit may function as an alternate Emergency Director backing up the Shift Manager. Twenty-four (24) hour coverage for the Emergency Director position is provided by the Duty Shift Manager who assumes the responsibility of the TSC Emergency Director at the onset of any emergency condition.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                              EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                                    REV:      58 Page 21 of 164 When the Technical Support Center (TSC) and Emergency Operations Facility (EOF) Organizations are activated, the Emergency Director (ED) and TSC staff will relieve the Emergency Director on shift of command and control functions as soon as practical and assume the responsibility for the management of NSPMs overall response to the emergency. The Emergency Director on shift can then direct the plants priorities for event responses. Upon activation of the EOF, responsibility for offsite functions of notification and protective action recommendations transfer from the TSC to the EOF Emergency Manager (EM). The transition of command and control functions is depicted below.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                                      EMERGENCY PLAN NUMBER:
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TSC Emergency Director Assembly Point                                      ERF Coordinator                                  Communicator Operations                          Radiological                                                      Work          Logistic Security                            TSC            Maintenance  Engineering Group                              Emergency                                                    Management      Support Group Leader                        Coordinator      Group Leader  Group Leader Leader                            Coordinator                                                      Leader        Leader Operations                                                TSC Offsite            REC                                OSC      Core Thermal              Status Board Group Leader                                            Coordinator Communicator        Assistant                        Coordinator    Engineer                  Keepers Assistant                                              Assistant Dose                                          Mechanical ENS                                                                                                                ERCS Projection                                        Engineer Communicator                                                                                                        Operator Specialist Field Team                                        Electrical                Record Log Communicator                                        Engineer                    Keeper HPN                                                                      Switchboard Communicator                                                                    Operator Minimum Staff Positions Positions listed are 60 minute responders
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                                  EMERGENCY PLAN NUMBER:
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TSC OSC Maintenance Coordinator Leader ERF Communicator Radiation                    Mechanical        Electrical    Instrument &
OPS Protection                  Maintenance      Maintenance        Control Advisor Coordinator                  Coordinator        Coordinator      Coordinator RP                  Chemistry                                                I&C        Rad Status Technicians            Technicians        Mechanics        Electricians Technicians  Communicator Status Board Keepers Minimum Staff Positions Positions listed are 60 minute responders unless otherwise noted on Table 1
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                              EMERGENCY PLAN NUMBER:
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                                              EMERGENCY PLAN NUMBER:
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Shift Supervisor (SRO):                                        Unit Supervisors                  2          -                -
Assessment of Control Room Reactor Operational Aspects                                            Reactor Operators (RO)            4          -                -
Auxiliary Operators                6          -                -
Notification/                        Notify State, local        Shift Emergency Communication                        and Federal                Communicator                      1          -                -
personnel &                Offsite Communicator              -          1                1 maintain communication    ENS Communicator                  -          1                1 Radiological Accident                Emergency Operations      Emergency Manager                  -          -                1 Assessment and                      Facility (EOF)            Emergency Director (TSC)          -          1                -
Support of                          Director Operational Accident Assessment                          Offsite Dose                Radiological Assessment                  Emergency Coordinator              -          1                -
RP Support Supervisor              -          -                1 Offsite Surveys            Radiation Protection              -          2                2 Specialist/Support Onsite Surveys              Radiation Protection Specialist    1          1                1 (out-of-plant)
In-Plant Surveys
 
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Radiochemistry              Chemistry Technician              1          1                -
Plant System                        Technical Support          Shift Technical Advisor            1          -                -
Engineering                                                      Core/Thermal Engineer (TSC)        -          1                -
Electrical                        -          1                -
Mechanical                        -          1                -
Repair and                          Repairs and                Mechanical Maintenance            -          1                -
Corrective Actions                  Corrective Actions Electrical Maintenance                        1                -
Instrument Control                -          -                1 Protective Actions                  Radiation                  Radiation Protection Specialist    1          1                1 (In-Plant)                          Protection:
: a. Access Control
: b. HP Coverage for repair, corrective actions, search and rescue, first-aid &
firefighting
: c. Personnel monitoring
: d. Dosimetry
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                                                    EMERGENCY PLAN NUMBER:
E-PLAN EP                                                      EMERGENCY PLAN                                                  REV:      58 Page 27 of 164 Table 1          Guidance for Augmentation of Plant Emergency Organization Capability for Additions Major Functional Area                    Major Tasks          Position Title or Expertise  On-Shift      60 min          90 min Fire Fighting                                                                                      Fire Brigade          Local Support per F5 Rescue Operations and First Aid                                                                                          2(1)            Local Support Site Access Control                      Security, firefighting      Security Personnel              As per and Personnel                            communications,                                            Security Accountability                          personnel                                                    Plan accountability TOTAL                              18            14              9 (1)
May be provided by shift personnel assigned other functions.
The above table was developed in accordance with 10 CFR 50 Appendix E. This staffing analysis is documented in F3-1.1, Emergency Plan On-Shift Staffing Analysis.
 
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                    EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                      REV:      58 Page 29 of 164 The Shift Manager SHALL be relieved of the Emergency Director responsibilities when the designated Emergency Director arrives onsite.
The Plant Manager or Designee SHALL be the designated Emergency Director and will be available with a pager on a twenty-four (24) hour basis.
When the Plant Manager is unavailable, (e.g., out of town), the designated Emergency Director responsibility will be passed onto another Plant Manager designee who is a member of senior plant management. Specific personnel assignments to the Emergency Director position are found in the MT & PI Nuclear Emergency Preparedness Telephone Directory.
The Shift Manager SHALL start the tasks assigned to the Emergency Director, (e.g., notification, activating onsite centers, etc.). These tasks SHALL be accomplished promptly and cannot wait for the designated individual to arrive at the plant site.
The Emergency Directors responsibilities are as follows:
A. Activation of onsite emergency organization -
: 1. Direct the activation of the onsite emergency response centers and monitor their habitability, and
: 2. Coordinate response of the plant onsite emergency organization.
B. Personnel accountability - During a plant evacuation the Emergency Director SHALL account for all personnel onsite within thirty minutes of the Site Area or General Emergency requiring the evacuation so that a search for missing personnel can be conducted. A continuous personnel accountability SHALL be maintained throughout the emergency. This responsibility may be delegated to a designated individual with assistance from the security force.
C. Radiological monitoring - The Emergency Director SHALL direct radiological monitoring of all personnel onsite and at the onsite assembly area, for contamination and/or excessive exposure. This responsibility may be delegated to the Radiation Protection Specialists or to a qualified operations member.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                                      REV:      58 Page 30 of 164 D. Exposure - The Emergency Director SHALL be responsible to authorize overexposures in excess of the normal limits (this responsibility may not be delegated).
E. Radiation Survey Teams - The Emergency Director SHALL direct the Radiation Survey Teams to obtain the necessary onsite and offsite samples and/or radiation surveys. This responsibility may be delegated to the Radiological Emergency Coordinator.
F. Offsite Dose Projections - The Emergency Director SHALL be responsible to project dose rates to the offsite population. This responsibility may be delegated to the Radiological Emergency Coordinator.
G. Protective Action - The Emergency Director SHALL be responsible for authorizing offsite Protective Action Recommendations (this responsibility may not be delegated and is relinquished to the Emergency Manager when the EOF is activated).
H. Notification - The Emergency Director SHALL be responsible to ensure that the necessary offsite notifications are initiated and completed. This responsibility may be delegated to the Shift Emergency Communicator (SEC). The SEC may designate offsite communications to a qualified Communicator.
: 1. Immediate (within 15 minutes)
The initial notification message to State, local and tribal authorities, from the plant, SHALL contain the following information:
a    Class of emergency b    Whether radioactivity is being released and in what form (liquid or gas) c    Potentially affected populace and area, if any d    Necessity of protective measures e    Brief description of the event Other information, i.e., meteorological data, etc., are available to these authorities via the follow-up notification messages.
 
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: 2. Subsequent Messages The plant will continue to provide updating information to offsite authorities. As soon as possible after the initial notification of an Alert, Site Area, or General Emergency, as much of the following information that is known and appropriate will be forwarded to offsite authorities:
a    Location of incident b    Name and telephone number of caller c    Date/time of incident d    Class of emergency e    Type of release (airborne, liquid, surface spill) and estimated duration f    Estimate of noble gas, iodine, and particulate release rates g    Prevailing weather conditions (wind speed, wind direction, temperature, atmospheric stability class, precipitation, if any) h    Actual or projected dose rates at site boundary i    Projected dose rate and integrated dose at 2, 5 and 10 miles and the Sectors affected.
j    Survey results of offsite dose rates or any surface contamination k    Plant emergency response actions in progress l    Request for onsite support from offsite support organizations m    Prognosis for worsening or termination of event based on plant information To provide ease in supplying the aforementioned information, a standardized form is used and incorporated into the implementing procedures.
I. Protracted Emergency Shift Coverage - The Emergency Director, with assistance from and coordination with other group Managers and Supervisors, SHALL ensure that work force requirements for all subsequent work shifts are determined and the necessary personnel are scheduled for the specific time period.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                    REV:      58 Page 32 of 164 5.3.2    Plant Emergency Organization Coordinators A. Technical Support Center Coordinator The Technical Support Center (TSC) Coordinator SHALL be responsible for the general activation, operation and coordination of activities in the Technical Support Center (TSC). Specific personnel assignments to the TSC Coordinator are found in the MT & PI Nuclear Emergency Preparedness Telephone Directory.
The responsibilities of the TSC Coordinator are:
: 1. Establish and verify radiological monitoring for the TSC;
: 2. Assist personnel performing the accountability check;
: 3. Coordinate activities of plant and non-plant personnel located in the TSC;
: 4. Periodically update personnel located in the TSC with appropriate information;
: 5. Maintain any necessary status boards;
: 6. Ensure technical guidance is provided to the Emergency Director and Control Room Operators on plant operations;
: 7. Establish or ensure that communications are established between all onsite emergency facilities and the EOF.
: 8. Ensure the Emergency Response Data System data link is established with the NRCs emergency center.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                                    REV:      58 Page 33 of 164 B. Operational Support Center Coordinator The Operational Support Center Coordinator SHALL be responsible for the general activation, operation, and coordination of activities in the Operational Support Center (OSC). Specific personnel assignments to the OSC Coordinator are found in the MT & PI Nuclear Emergency Preparedness Telephone Directory.
The responsibilities of the OSC Coordinator are:
: 1. Establish and verify radiological monitoring for the OSC and the Control Room;
: 2. Coordinate activities of plant personnel located in the OSC to support plant operations as requested by the Control Room and TSC.
: 3. Assist personnel performing the accountability check in the OSC and the Control Room.
: 4. Maintain the communication systems in the OSC. A person may be designated to act as a communicator.
: 5. Periodically update personnel located in the OSC with appropriate information.
: 6. Control the use of equipment located in the emergency locker.
C. Assembly Point Coordinator The Assembly Point Coordinator SHALL be responsible for the general operation of the assembly area. Specific personnel assignments to the Assembly Point Coordinator are found in the MT & PI Nuclear Emergency Preparedness Telephone Directory.
The responsibilities of the Assembly Point Coordinator are:
: 1. Verify that radiological monitoring has been established for the Assembly Point.
: 2. Coordinate activities of all personnel (plant and non-plant) located at the Assembly Point.
: 3. Assist the Emergency Director in performing the accountability check, as necessary.
 
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: 4. Maintain the communication systems. A person may be designated as the communicator, if necessary.
: 5. Control the use of equipment located in the Emergency Locker.
: 6. Update all personnel with appropriate information when directed by the Emergency Director.
: 7. Provide instructions to personnel when they are released from the assembly point for reentry or transport offsite.
D. Radiological Emergency Coordinator The Radiological Emergency Coordinator (REC) SHALL be responsible for radiological accident assessment, onsite and offsite.
The REC should report to the Technical Support Center when the TSC is activated. Upon activation of the EOF, the Radiation Protection Support Supervisor will assume responsibility for the offsite activities. The REC should transfer the responsibility for offsite accident assessment to the Radiation Protection Support Supervisor at the EOF. Specific personnel assignments to the Radiological Emergency Coordinator are found in the MT & PI Nuclear Emergency Preparedness Telephone Directory.
The responsibilities of the REC are:
: 1. Offsite dose assessment
: 2. Formulating offsite protective action recommendations
: 3. Offsite surveys
: 4. Onsite surveys
: 5. Chemistry
: 6. Radiochemistry
: 7. Onsite Radiation Protection for:
a    Access Control b    Damage control and repair c    Search and rescue d    First-aid e    Personnel monitoring and decontamination f    Dosimetry
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                  REV:      58 Page 35 of 164 5.3.3    Plant Shift Organization The following groups comprise the plants shift organization. Brief descriptions of their emergency responsibilities are included.
A. Operations Group The Operations Group consists of the Operations Manager, Asst.
Operations Manager, Shift Managers, Shift Technical Advisors, Shift Supervisors, and all operators.
The Operations Group SHALL have responsibility for:
: 1. Plant Operations and assessment of operational aspects of the emergency.
: 2. Short term damage control and repair for electrical, mechanical, and I&C equipment.
B. Security Group The Security Group consists of the Security Manager, the Security Staff, and the contract Security Force.
The Security Force SHALL:
: 1. Carry out the plant security and Access Control program.
: 2. Maintain strict personnel accountability onsite.
: 3. Assist communications efforts when necessary.
: 4. Assist in first aid treatment.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                  EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                    REV:      58 Page 36 of 164 C. Shift Manager The Shift Manager (SM) SHALL be onsite continuously. The Shift Manager SHALL assume overall coordination and control in the Control Room and provide direction as necessary to the Shift Supervisor.
The Shift Manager SHALL:
: 1. Assume the duties of the Emergency Director until relieved by the TSC Emergency Director. Portions of the E-Plan implementation may be delegated to other members of the plant staff as the condition of the plant dictate.
: 2. Assess the emergency condition, event evaluation, and safety related aspects of the plant.
D. Shift Technical Advisors Provide technical and engineering support in the area of accident assessment.
E. Shift Emergency Communicator (SEC)
The Shift Emergency Communicator (SEC) SHALL be onsite continuously. The SEC is responsible for initial notification to the offsite agencies and maintaining communications during emergency conditions. The SEC may designate offsite communications to a qualified Communicator.
: 1. When the EOF is activated, communications with the offsite agencies and personnel will be maintained by the EOF personnel.
NOTE:          2. As the emergency organization is activated, additional communicators from TSC support personnel should augment the plant staff to assist in communication efforts.
F. Fire Brigade The Fire Brigade should consist of:
: 1. Brigade Chief - Unit 1 Turbine Building APEO or as designated by the Shift Manager.
: 2. Assistant Chief - Any Qualified APEO.
 
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E-PLAN EP                                EMERGENCY PLAN                                      REV:      58 Page 37 of 164 Usually the APEO from the affected building SHALL fulfill the NOTE:          duties of the Brigade Chief in his absence.
: 3. Fire Fighters - BOP Operators.
: 4. Runner - As designated to accompany fire department, operate equipment, bring additional equipment to fire scene.
The Fire Brigade SHALL be responsible for firefighting and primary responders for Search and Rescue, as necessary.
The Red Wing Fire Department should provide emergency assistance and SHALL be called immediately on report of fire. Other plant personnel on site may be called on for emergency work or called to plant for emergency service.
G. Radiation Protection Specialist The Radiation Protection Organization consists of two Radiation Protection Specialists (RPS) onsite at all times. The RPS is responsible for conducting routine and special surveys, maintaining Access Control, writing RWPs and providing job coverage as required.
H. Chemistry Technician One Chemistry Technician is onsite at all times. The Chemistry Technician is responsible for chemistry, radiochemistry, dose assessments, and offsite dose projections. The Chemistry Technician is also cross-trained to support the Radiation Protection Specialist functions described in Section G above.
 
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E-PLAN EP                              EMERGENCY PLAN                                    REV:      58 Page 38 of 164 5.3.4    Plant Emergency Staff Augmentation Groups A. Maintenance Group The Maintenance Group consists of all mechanical maintenance personnel, all plant electricians and I&C Specialists. The onsite Emergency Organization includes the Maintenance Manager, who should report to the Technical Support Center (TSC); and the Maintenance Supervisors (mechanical, electrical and I&C), and designated Electricians who should report to the Operational Support Center (OSC). The mechanical, electrical and I&C maintenance staff in the OSC can be further augmented or decreased as emergency conditions dictate.
The Mechanical, Electrical, and I&C Maintenance Group SHALL have responsibility for:
: 1. Supporting the repair and corrective actions for the mechanical, electrical, and I&C systems in support of emergency response and recovery actions.
: 2. Supporting the Search and Rescue effort.
B. Radiation Protection Group and Chemistry Group The Radiation Protection and Chemistry Groups consists of the Radiation Protection Manager & Chemistry Manager and all members of the Radiation Protection and Chemistry Groups. Radiation Protection and Chemistry Managers and other designated group members should report to the Technical Support Center. Other Radiation Protection Specialists and Chemistry Technicians should report to the Operational Support Center.
The responsibilities of the Radiation Protection and Chemistry Groups are:
: 1. Offsite Dose Assessment
: 2. Offsite Surveys
: 3. Onsite Surveys
: 4. Chemistry
 
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: 5. Radiochemistry
: 6. Radiation Protection for:
a    Access Control b    Damage control and repair c    Search and rescue d    First aid e    Fire fighting f    Personnel monitoring and decontamination g    Dosimetry C. Engineering Group The Engineering Group consists of Systems, Programs, Design and Equipment Reliability.
Upon activation of the onsite emergency organization, Systems and Programs Engineering Managers and designated engineers assigned to the emergency organization should report to the Technical Support Center. Other designated engineers may be requested to further augment engineering support in the TSC.
The Engineering Group SHALL have responsibility for:
: 1. Providing technical support for plant system engineering on electrical/mechanical systems.
: 2. Providing technical support for operating radioactive waste control systems.
: 3. Providing core parameter analysis to determine current core status.
: 4. Providing plant parameter trending and analysis utilizing the Emergency Response Computer System (ERCS).
: 5. Projecting possible loss of key equipment and its consequences.
: 6. Providing technical support for emergency repairs and corrective actions on electrical/mechanical systems.
: 7. Update TSC staff of potential problems and developments.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                              EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                                  REV:      58 Page 40 of 164 D. Logistics Support Group The Logistics Support Group consists of Business Support Group (Administration Services and Document Control), Plant Services, and Site Materials.
Business Support Group SHALL supply logistical support in their area of expertise. Personnel in these areas may be called in to provide support for emergency response on an as needed basis.
Site Materials SHALL provide assistance in retrieving the parts necessary for an emergency response.
Plant Services SHALL support an emergency response by providing necessary assistance by the Nuclear Plant Service Attendants.
 
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E-PLAN EP                                EMERGENCY PLAN                                  REV:      58 Page 41 of 164 5.4    EOF Organization The EOF (Emergency Operations Facility) Organization consists of a Direction and Control Group and three subordinate groups. The EOF Organization is staffed by personnel from the sites Engineering and Project Management groups and Prairie Island Training Center staff. The Prairie Island EOF Organization is shown in Figure 2.
The EOF will be activated within 90 minutes of when an Alert, Site Area Emergency or General Emergency is declared.
5.4.1    EOF Direction and Control The Emergency Manager is responsible for overall direction and control of NSPMs emergency response effort. Designated members of management staff the Emergency Manager position in the EOF. Specific personnel assignments to the Emergency Manager position are found in the MT & PI Nuclear Emergency Preparedness Telephone Directory. The Emergency Manager relieves the Emergency Director of the following responsibilities:
A. Off-site dose projections and coordination and direction of the utility off-site radiological monitoring teams.
B. Authorization of offsite Protective Action Recommendations.
C. Communications with off-site authorities including Federal, State, Local and Tribal authorities and MT & PI Offsite executive management located at the Minnesota State Emergency Operations Center.
 
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E-PLAN EP                              EMERGENCY PLAN                                                REV:      58 Page 42 of 164 Figure 2        Prairie Island EOF Organization Emergency                        Emergency                                  Recovery Director                        Manager                                    Manager RP Support                        EOF                                Technical Support Supervisor (RPSS)                Coordinator                                Supervisor RPSS Assistant                      EOF                                Engineering Support State Liaison            Coordinator Assistant                            Team Lead Lead Electrical RPSS Assistant Security Coordinator                              Engineer Field Team and Dose Assessment Lead Mechanical Engineer Field Team                    Off-site Communicator                Communicator Status Board Keeper Rad Status Board                    ERF Keeper                  Communicator Administrative Dose Projection                Support Lead                              Trending Team Specialist                                                                Leader Administrative Support Staff ERCS Operator Count Room Chemistry Technician Event Status Board Keeper EOF Radiation Monitoring RP Specialist or                                                        Technical Corporate Chemistry Technician                                                        Communicator (EOF-JIC)
Field Teams and ENS Communicator Drivers Sample Couriers                                                        EOF Narrative Log Keeper Minimum Staff HPN                      Positions Communicator Positions listed are 90 minute responders
 
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E-PLAN EP                                EMERGENCY PLAN                                      REV:      58 Page 43 of 164 Other responsibilities of the Emergency Manager include:
A. Coordinate the emergency response efforts of other offsite support personnel assisting the plant organization.
B. Obtain and coordinate the services of outside consultants and vendors.
C. Advise utility management on matters related to emergency response efforts and needed resources to support the effort.
5.4.2    EOF Technical Support Group The EOF Technical Support Group consists of select personnel from the sites Engineering and Project Management groups and Training Center staff. The Technical Support Supervisor is staffed by senior personnel and reports to the Emergency Manager. The Technical Support Group is responsible for trending critical parameters, engineering evaluation in support of the TSC Engineering Group, technical assessment and advising the Emergency Manager on technical matters related to the event.
5.4.3    EOF Radiation Protection Support Group The EOF Radiation Protection Support Group is staffed by select personnel from the Training Center, plant Radiation Protection and Chemistry Groups and Emergency Plan Group. The Radiation Protection Support Supervisor position is staffed by senior personnel qualified in radiation assessment and reports to the Emergency Manager. The Radiation Protection Support Group includes plant Chemistry personnel for off-site dose projection and EOF Count Room operation and Nuclear Plant Service Attendants who function as sample couriers and drivers for off-site radiological monitoring teams. Radiation Protection Support Group responsibilities include:
A. Direction and coordination of the utility off-site radiological monitoring teams.
B. Off-site dose projection.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                  REV:      58 Page 44 of 164 C. EOF Count Room activation and operation.
D. EOF habitability, personnel monitoring and decontamination (as necessary).
E. Communications with state assessment groups on matters related to dose projections and off-site protective action recommendations.
F. Staffing the Health Physics Network (HPN) and communications with the NRC (as necessary).
The Radiation Protection Support Supervisor advises the Emergency Manager on matters related to actual or potential radiological impact on the environment, off-site protective action recommendations, and EOF habitability.
5.4.4    EOF General Support Staff The EOF General Support Staff consists of the EOF Coordinator, emergency communicators, administrative and security support personnel.
The EOF Coordinator position is staffed by senior Training Center or site Engineering and Project Management personnel and reports to the Emergency Manager. The EOF Coordinator is responsible for activation and operation of the EOF and assists the Emergency Manager with administrative duties. The emergency communicators, EOF Security Coordinator and Administrative Staff report to the EOF Coordinator. The emergency communicators are responsible for communications with offsite agencies as directed by the Emergency Manager. The Administrative Staff is responsible for emergency document control, recording and document distribution at the EOF. An EOF Coordinator Assistant is responsible for general logistics support and assisting the EOF Coordinator. The EOF Security Coordinator reports to the EOF Coordinator. Responsibilities of EOF Security include EOF access and dosimetry issuance to EOF personnel.
 
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E-PLAN EP                                EMERGENCY PLAN                                    REV:      58 Page 45 of 164 5.5    Recovery Organization The establishment of the Recovery Organization will be dependent upon the nature and severity of the event or plant conditions. In general terms, an Unusual Event or Alert may be terminated without establishing a special Recovery Organization while a Site Area Emergency or General Emergency will probably necessitate the establishment of a Recovery Organization. The Recovery Organization will manage the overall recovery or post-accident outage plans as work is done to return the plant to a normal operational or shutdown status.
The Recovery Manager is mainly responsible for management of the recovery phase and will perform their initial tasks as directed by the Emergency Manager.
The Recovery Manager will report to the Emergency Operations Facility and begin to prepare for the transition to Recovery, as necessary. If Recovery is imminent, the Recovery Manager will establish a recovery or post-accident outage organization following the sites plant event recovery protocols.
5.6    Augmentation of Plant and EOF Emergency Organizations 5.6.1    Offsite Support Response The emergency response plan for Prairie Island NGP is designed to be initially implemented independent of any offsite support. However, the onsite effort will be augmented with offsite support resources as described in the MT & PI Offsite Nuclear Emergency Plan.
It is the purpose of the offsite support organization to augment the onsite response effort with offsite support resources as soon as practical and as needed by the Prairie Island Site staff. Such areas of support include:
Government Agency Interface, Logistics Support, News Media Interface and Utility Executive Management Interface.
 
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E-PLAN EP                                  EMERGENCY PLAN                                  REV:      58 Page 46 of 164 5.6.2    Monticello Radiation Protection Group Support The Monticello Nuclear Generating Plant is located approximately 100 miles northwest of Prairie Island NGP. The Monticello Radiation Protection and Chemistry Groups are available for supporting the Prairie Island Radiation Protection Group with personnel and equipment during any emergency condition at Prairie Island.
5.6.3    Westinghouse Support Westinghouse emergency assistance is available on a twenty-four hour per day, seven day per week basis. Westinghouse will activate all appropriate features of the Westinghouse Emergency Response Plan to support the plant needs. When activated, the Westinghouse Emergency Response Plan becomes a functioning organization, comprised of individuals with unique technical, managerial and communication skills and experience, necessary to:
A. Make an early assessment of the situation B. Provide early assistance to the utility C. Mobilize appropriate Westinghouse critical skills and functions D. Initiate timely, accurate communications to involved and interested parties A Site Response Team may be dispatched to the site to obtain a first hand assessment of actual conditions and establish communications from the site to the Westinghouse response center, as deemed necessary.
 
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E-PLAN EP                              EMERGENCY PLAN                                    REV:      58 Page 47 of 164 5.6.4    Local Support Services A. Fire Fighting The Red Wing Fire Department will provide assistance in the event of a fire occurring at the plant. The duties and responsibilities of the Plant Fire Brigade, insuring complete coordination with the Fire Department, are covered in the Operations Manual, Section F5, Fire Fighting.
The Red Wing Fire Department will be the lead fire and Emergency Medical Service (EMS) agency for all emergencies. The Red Wing Fire Department maintains mutual aid agreements with other area ambulance and fire departments as specified in the City of Red Wing/Goodhue County Emergency Response Plan. These agreements provide that the City may call upon other resources to assist in responding to an emergency, including a Hostile Action Based (HAB) event. For a HAB event, Red Wing Fire Department will deploy a representative to the Incident Command Post dependent upon type, location, and scope of the incident, once scene safety is established.
The Red Wing Fire Department has various firefighting apparatus and water pumping equipment available for use. All Red Wing Fire Department apparatus can perform both fire fighting tasks, including rescue, and non-fire fighting tasks, including spraying to contain radiological releases and pumping water into the plant for refilling and cooling purposes. In all cases, such operations can begin once the radiological and security threats are mitigated to insure the safety of both plant personnel and fire fighters.
 
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E-PLAN EP                              EMERGENCY PLAN                                    REV:      58 Page 48 of 164 B. Hospital and Medical Support Medical support and treatment for non-radiological injuries is provided by the Mayo Clinic Health System, both of which are located in Red Wing, Minnesota. Radiological related injuries are treated at the medical center which is the primary treatment facility. Emergency plans have been prepared, and training of medical center personnel is accomplished on an annual basis.
Regions Hospital in St. Paul, Minnesota is designated as the definitive care center for Prairie Island Nuclear Generating Plant.
Regions Hospital may be used for radiation casualties, severe burn casualties, and other non-radiation injuries with use of an appropriate medical air transport service.
C. Ambulance Service The Red Wing Ambulance Service will provide service to the Prairie Island Nuclear Generating Plant. Training and participation in drills ensures that personnel involved in the transportation of radiation victims are knowledgeable in use of proper procedures and handling methods. Procedures are covered in the Operations Manual, Section F4, Medical Support and Casualty Care.
 
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E-PLAN EP                              EMERGENCY PLAN                                    REV:      58 Page 49 of 164 D. Local Law Enforcement For a Notification of Unusual Event (NUE) Security Condition and an Alert and Site Area Emergency Hostile Action Based (HAB) event at PINGP, the City of Red Wing Police Department is the lead law enforcement agency. For a HAB event, the Red Wing Police Department will set up an Incident Command Post (ICP) near the site. The pre-designated ICP locations have been identified; however, selection will depend on the incident. The City of Red Wing Police Department maintains the list of potential ICP sites and will be responsible for designating the site during a response and telling the other agencies responding to the location. Unified Command should be established and includes city, county, state, federal and utility expertise. Communication will be established between the Incident Commander and plant security and operations as soon as possible.
The Red Wing Police Department has the ability to request additional response resources from neighboring agencies (i.e. the primary source of additional resources will be the Goodhue County Sheriff's Office with the ability to request assistance from other neighboring agencies as necessary) to assist them in response to any Prairie Island contingency situation, including a HAB event.
The initial hostile action response goals are; maintain vital plant systems to prevent a release of radioactive materials, protection of on-site workforce, neutralizing the adversaries, and restoring plant operating conditions. Tactical operational priorities supported by Law Enforcemment include; securing a perimeter around the site, containment of vital areas, sweep and securing of vital areas, safe movement of critical workers on the site, neutralizing adversaries, protection/evacuation of the on-site workforce, and sweep of protected area and owner controlled area.
The Incident Command Post will support tracking resources and personnel at or near the site and the City of Red Wing/Goodhue County Emergency Operation Center (EOC) will support tracking resources and personnel off-site in accordance with the Radiological Emergency Plan. In the event that NSPM has declared a General Emergency as defined in the City/County Plan, the Goodhue County Sheriff's Office shall assume operational control over all emergency operations.
 
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E-PLAN EP                              EMERGENCY PLAN                                    REV:      58 Page 50 of 164 5.7    Coordination with Governmental Response Organizations 5.7.1    Minnesota Division of Homeland Security and Emergency Management (HSEM)
The Minnesota Division of Homeland Security and Emergency Management has the responsibility for notification and coordination of Minnesota State Agencies in the event of a major emergency at Prairie Island.
The MN HSEM is notified by Prairie Island NGP. In the event of an emergency situation at the plant, the MN duty officer will immediately call the MN Department of Health, the Governors Authorized Representative and other state agencies with emergency assignments to coordinate the implementation of any emergency procedures. The state agencies responsible for emergency procedures have established a system of twenty-four hour communications.
5.7.2    Minnesota Department of Health (MDH)
The Minnesota Department of Health (MDH) is responsible for providing radiological expertise in the State Emergency Operations Center in conjunction with the MN HSEM.
The Minnesota Department of Health will interpret data and participate in recommending protective actions to the Governors Authorized Representative.
 
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E-PLAN EP                                EMERGENCY PLAN                                    REV:      58 Page 51 of 164 5.7.3    Wisconsin Emergency Management The Wisconsin Emergency Management (WEM), has the responsibility for notification and coordination of Wisconsin state agencies in the event of a major emergency at Prairie Island NGP.
In the event of an emergency situation at the plant, Prairie Island NGP will notify the WEM duty officer who will notify the Wisconsin Department of Health Services (Radiation Protection Section) and other state agencies with emergency assignments, to coordinate the implementation of any emergency procedures. The state agencies responsible for emergency procedures have established a system of twenty-four hour communications.
5.7.4    Wisconsin Department of Health Services (DHS)
The Wisconsin Department of Health Services (DHS) is responsible to prevent exposure to ionizing radiation in amounts which are detrimental to health according to nationally accepted standards.
The Wisconsin DHS, Radiation Protection Section, is responsible for coordination of radiation response activities in the State of Wisconsin. In the event of an emergency at Prairie Island NGP, DHS, Radiation Protection Section will be concerned with monitoring the air and water about the plant to assure that the public is not exposed to levels of radioactive pollutants potentially detrimental to public health. DHSs facilities are located in Madison, Wisconsin.
 
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E-PLAN EP                                EMERGENCY PLAN                                      REV:      58 Page 52 of 164 5.7.5    Goodhue, Dakota and Pierce County Sheriffs The Sheriffs Departments will notify all necessary local emergency response groups in the event of an accident. The Sheriff is responsible for protection of the general public and can provide personnel and equipment for evacuation, relocation and isolation.
Goodhue County and the Sheriff also has agreements in place to request additional response resources from neighboring agencies, including resources needed to respond to a HAB event. For a HAB event, the Red Wing Police Department will set up an Incident Command Post (ICP) near the site. The Goodhue County Sheriff's Office Tactical Response Team will be the lead tactical response operations group coordinator and coordinate the tactical law enforcement response with Command. Goodhue County Sheriff's Office can request tactical team resources as needed from:
Minnesota State Patrol Special Response Team, Dakota County ERT, FBI SWAT and Washington County ERT.
5.7.6    Goodhue, Dakota, Pierce County and City of Red Wing Emergency Management The Goodhue, Dakota, Pierce County and City of Red Wing Emergency Management Organizations have the responsibility for notification and providing direction to residents in the event of a major emergency that affects their respective area of responsibility.
5.7.7    Prairie Island Indian Community The Prairie Island Indian Community has an Emergency Operations Plan that includes the description of tribal responsibilities during a nuclear plant declared event. The Prairie Island Nuclear Generation Plant conducts emergency notifications to the Treasure Island Casino security dispatch center who, in turn, notifies appropriate members of the Prairie Island Indian Community and their organization.
 
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E-PLAN EP                                EMERGENCY PLAN                                    REV:      58 Page 53 of 164 5.7.8    Minnesota State Patrol The Minnesota (MN) State Patrol has the responsibility to protect the general public by providing personnel and equipment to re-route traffic in the event of an emergency situation. Plans have been made for re-routing federal and state highways. Signs and equipment required for re-routing will be stored in the areas where they would be needed to facilitate highway closings. The MN Department of Transportation would be notified by the MN State Patrol to erect the signs.
5.7.9    Minnesota Department of Transportation The MN Department of Transportation will assist the MN State Patrol in blocking and re-routing traffic around the plant site. In addition to the necessary personnel; vehicles, signals, and barriers for setting up and maintaining detour routes are available.
5.7.10  Canadian Pacific Railway-CP Railway (Soo Line In an emergency situation, CP Rail will make every reasonable effort to expedite unblocking the road/railroad crossing near Prairie Island NGP.
The dispatcher will also provide routing assistance during an emergency at Prairie Island NGP.
5.7.11  Burlington Northern Santa Fe (BNSF) Railway The dispatcher will provide routing assistance during an emergency at Prairie Island NGP as per the Minnesota State emergency operations plan.
5.7.12  Department of the Army, Corps of Engineers, Lock & Dam #3 The Corps of Engineers at Lock & Dam #3 will be notified by the Minnesota Duty Officer of an emergency at Prairie Island NGP. The Lock and Dam personnel will notify all tows within radio range of impending evacuations and assist in evacuation of personnel at the Lock and Dam.
 
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E-PLAN EP                                EMERGENCY PLAN                                        REV:      58 Page 54 of 164 A complete description of response capabilities, organizational resources, activation plans, designations of NOTE:          emergency operations centers and letters of agreement are available in Minnesota and Wisconsins state emergency operations plans.
5.7.13  Nuclear Regulatory Commission (NRC)
The basic responsibilities of the NRC are to monitor, assess, and, if necessary, direct the utility to take actions to protect the health and safety of the public. For a radiological incident at a commercial power plant, the NRC is the Lead Federal Agency (LFA). The LFA is responsible for coordinating all Federal onscene actions. The NRC will coordinate Federal assistance to States and local organizations.
A principal role of the LFA is to assist the State in interpretation and analysis of technical information as a basis for making decisions about protective actions. This assistance will begin early in an incident from the NRC Operations Center in Rockville, MD, and later, from the utilitys emergency operations facility on scene. The NRC is an independent reviewer of the actions the utility is taking to correct the initiating and related problems. The NRC will assess actual or potential offsite impacts as well, and will make an independent evaluation of Protective Action Recommendations, if necessary. As the LFA, the NRC has the responsibility for coordinating the release of Federal information to the media and others. The NRC will conduct most public information activities from the utilitys Joint Information Center (JIC). The NRC also will keep the White House and Congress informed on all aspects of the event.
The NRC is responsible for giving the best possible advice at a given time to the States and will not limit its involvement to presenting a series of options.
The NRC also administers the Price-Anderson Act to ensure that the public that is affected by the event has adequate financial assistance to address most emergency needs.
 
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E-PLAN EP                                EMERGENCY PLAN                                  REV:      58 Page 55 of 164 5.7.14  Department of Energy (DOE)
Among its responsibilities as a support agency, DOE will coordinate the offsite radiological monitoring and assessment for the Lead Federal Agency (LFA) and the State during the initial phases of the emergency. It will maintain a common set of offsite radiological data and provide an appropriate interpretation of the data to the LFA and the State. DOE will manage the Federal Radiological Monitoring and Assessment Center (FRMAC), which is a multi-agency facility. DOE will conduct environmental monitoring, including air, ground, and water.
Their immediate objective is to rapidly dispatch a Radiological Assistant Program (RAP) Team to the scene to assess the hazard to the public and make recommendations to the authorities for the protection of the public.
The Planning Chief in the State EOC is the designated Minnesota authority to request RAP assistance, as stated in the Minnesota state plan, and the Wisconsin DH, Radiation Section, is the designated Wisconsin authority to request RAP assistance for Wisconsin, as stated in the Wisconsin state plan.
5.7.15  Institute Of Nuclear Power Operations (INPO)
INPO will coordinate requests from other utility INPO members and participants. They will notify NEI and EPRI of events, maintain an emergency resource capability and information on industry assistance capabilities coordinate the delivery of persons and materials under its Nuclear Power Plant and Transportation Agreements, and provide member communications to facilitate the flow of technical information about the emergency.
 
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E-PLAN EP                                EMERGENCY PLAN                                      REV:      58 Page 57 of 164 6.0    EMERGENCY MEASURES This section will describe the activation of the Emergency Organization. Various detailed and specific emergency measures that will be taken by the plant staff are further delineated in the plants emergency plan implementing procedures.
6.1    Activation of Emergency Organization 6.1.1    Activation of Plant and EOF Organizations The Shift Manager will be responsible for activating any part of the emergency organization. During the normal work week, the plant and training center public address systems will be used to activate the organizations. During the off-shift hours, activation of the emergency organizations will be accomplished using the ERO (Emergency Response Organization) Pager Network and the ERO Auto Dial System. Personal pagers are carried by the following personnel who are considered members of the emergency organization:
A. Radiation Survey Team Members B. Plant Operating Review Committee Members C. Maintenance Supervisors (Mechanical and Electrical)
D. I&C Supervisors E. Designated Engineers & Technical Personnel The ERO Pager Network is a personal pager system activated by a phone call. Upon receipt of a notification, it will be the responsibility of the supervisors to contact any additional personnel in their respective groups which may be required to report to the plant site, to staff the Technical Support Center, Operational Support Center and Emergency Operations Facility or to initiate offsite monitoring.
The ERO Auto Dial System is an automatic dialing telephone network with multiple outgoing telephone lines. When activated, it will call and deliver an emergency message to the Plant and EOF Organizations home telephones.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                    REV:      58 Page 58 of 164 The ERO Auto Dial System and ERO Pager Network are two notification system(s) used to activate the onsite emergency organization. One system is the backup of the other system. Both will be activated for ERO notification. Telephone numbers of all key emergency organization personnel are published in the MT & PI Nuclear Emergency Preparedness Telephone Directory.
If the event involves a credible security threat, EOF staff may be directed to staff the Backup EOF. In this case, the onsite ERO may be directed to the Red Wing Service Center until it is safe to staff the onsite OSC and TSC.
The Red Wing Service Center is to be used as the Alternative Facility during a security threat or event. The RWSC has communication links with the Control Room, EOF, and Security.
6.1.2    Notification Scheme When an abnormal condition is identified by the Operating Staff/Shift Supervisor, the Shift Supervisor will contact the Shift Manager and the Shift Emergency Communicator. An assessment of the safety significance will be performed, and a determination of the emergency classification will be made using the plants emergency plan implementing procedures.
Upon declaring an emergency condition, the Shift Manager will activate portions of the Emergency Plan as appropriate to respond to the declared emergency. During a Notification of Unusual Event, the Emergency Director position usually will not be staffed and the Shift Manager SHALL designate the Shift Emergency Communicator or other qualified communicator to make the necessary notifications of offsite state and local authorities. The Emergency Director position will be staffed during an Alert, a Site Area Emergency or General Emergency. The Shift Manager will assume the role as Emergency Director until relieved by the individual designated to relieve him.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                  EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                    REV:      58 Page 59 of 164 The Shift Manager/Emergency Director, will designate the Emergency Communicator or qualified designee to make notification calls to the following individuals or agencies, as detailed in the plants implementing procedures.
A. State of Minnesota HSEM B. State of Wisconsin Emergency Management C. Local Authorities (Wisconsin & Minnesota)
: 1. Dakota County Sheriff
: 2. Pierce County Sheriff
: 3. Goodhue County Sheriff D. Prairie Island Indian Community Representatives via Treasure Island Casino Security Dispatch Center E. Plant Manager (designated Emergency Director)
F. Emergency Manager G. Electric Utility System Operations Dispatcher H. NRC Resident Inspectors A more detailed call list of agencies and individuals, listing phone numbers, is included in the implementing procedures.
The Shift Manager/Emergency Director will ensure that the NRC Duty Officer is notified of the emergency by a qualified individual within one (1) hour of emergency declaration.
Eventually the Emergency Manager, in the Emergency Operations Facility (EOF), will relieve the Emergency Director of offsite communications and protective action recommendations. At that time offsite notification calls will be initiated by the EOF. The Prairie Island Onsite/Offsite Emergency Organization Interface is shown in Figure 3.
 
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Dedicated communicators will be assigned at each emergency operations center assuring a uniform transfer of information between segments of the onsite and offsite emergency response organizations. Initially, this responsibility rests with the Shift Emergency Communicator or qualified designee located in the Technical Support Center and subsequently with all backup communicators assigned these responsibilities.
Emergency Response Facilities such as the Technical Support Center, Operational Support Center, Control Room, Assembly Area and EOF will have dedicated communicators. Communicators will be assigned to specific communication duties, for example:
A.      ENS Hotline - licensed operator or designee B.      HPN Hotline - Radiation Protection personnel when requested by the NRC following facility activation.
C.      NRC Security Bridge - Security personnel when requested by the NRC following facility activation.
D.      Offsite State and Local Agency Notifications - Shift Emergency Communicator and Emergency Communicators E.      Survey Teams - Radiological Emergency Coordinator or Radiation Protection Support Supervisor and/or designee F.      Emergency Operating Centers - Operating Center Coordinators and/or designees G.      Others as deemed necessary
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                                                                                  EMERGENCY PLAN NUMBER:
E-PLAN EP                                                          EMERGENCY PLAN                                                                          REV:      58 Page 61 of 164 Figure 3            Prairie Island Onsite/Offsite Emergency Organization Interface TREASURE ISLAND NRC                              EMERGENCY DIRECTOR CASINO SECURITY OFFICE                            (EMERGENCY MANAGER)*
DISPATCH CENTER (LOCAL ISLAND RESIDENTS)
MISC OFF-SITE SUPPORT GROUPS XCEL ENERGY HEADQUARTERS WISCONSIN                                                    MINNESOTA COUNTY                                                      COUNTY SHERIFF                                                      SHERIFF COUNTY                                                                            COUNTY EMERGENCY GOVERNMENT                                                                EMERGENCY GOVERNMENT MUNICIPAL                                                                                    MUNICIPAL EMERGENCY GOVERNMENT                                                                          EMERGENCY GOVERNMENT WISCONSIN MINNESOTA EMERGENCY HSEM MANAGEMENT STATE                  GOVERNOR            STATE                STATE                GOVERNOR                STATE HEALTH                                    PATROL              HEALTH                                        PATROL DEPARTMENT                                                      DEPARTMENT SUPPORT                                                            SUPPORT MEDIA                                                      MEDIA AGENCIES                                                          AGENCIES
* EMERGENCY MANAGER assumes offsite responsibilities when the EOF Organization is activated .
 
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E-PLAN EP                              EMERGENCY PLAN                                  REV:      58 Page 63 of 164 Figure 4      HAB Communications Primary offsite authorities provide a 24 hour per day manning of communication links, as follows:
A. Wisconsin authorities
: 1. State of Wisc. (WEM) - State Patrol District 1 Dispatcher
: 2. Pierce County - Pierce County Sheriffs Dispatcher
 
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E-PLAN EP                                  EMERGENCY PLAN                                  REV:      58 Page 64 of 164 B. Minnesota authorities
: 1. State of Minnesota - Minnesota Duty Officer (MDO)
: 2. Goodhue County - Goodhue County Sheriffs Dispatcher
: 3. Dakota County - Dakota County Sheriffs Dispatcher C. Tribal Authorities - Treasure Island Security Dispatch 6.1.4      Authentication of Emergency Communications Communication, for the purpose of notifying offsite agencies that an emergency condition exists, SHALL be authenticated before offsite agency action is initiated. The authentications will be accomplished in accordance with the offsite agencies specific emergency plans.
6.2    Record Keeping It is the responsibility of all personnel involved in the emergency organization to ensure that accurate and complete records are maintained throughout the emergency situation. Emergency records may serve the following purposes:
6.2.1      Official documentation used to reconstruct the emergency for critique or analysis; 6.2.2      Check to ensure that necessary actions are completed during the course of an emergency; 6.2.3      Information and data collection during an emergency; and 6.2.4      Documentation of actions for legal purposes.
All activities performed by the operations staff SHALL be logged in the applicable reactor log. All other information and activities SHALL be maintained by the Emergency Director, Emergency Manager and various coordinators (e.g.,
individuals in charge of various emergency operating centers, radiation survey teams, etc.) for permanent plant records.
 
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E-PLAN EP                                EMERGENCY PLAN                                  REV:      58 Page 65 of 164 6.3    Summary of Site Response Actions Summarized below are the actions required by the site staff for each of the four emergency classifications. For each class of emergency, appropriate state, local, and tribal authorities will be notified. Depending on the emergency level classification, they will activate the segment(s) of their emergency organizations, according to their individual plans and based on the information received in the notification.
NOTIFICATION OF UNUSUAL EVENT
: 1. Promptly inform offsite authorities of unusual event status and the reason for the Unusual Event as soon as discovered.
: 2. Augment on-shift resources as needed.
: 3. Assess and respond to Unusual Event.
: 4. Terminate by contacting offsite authorities or
: 5. Escalate to a more severe class.
ALERT
: 1. Promptly inform offsite authorities of Alert status and reason for Alert as soon as discovered.
: 2. Augment resources by activating onsite Technical Support Center (TSC) and onsite Operational Support Center (OSC). The Emergency Operations Facility (EOF) and key offsite emergency organization personnel will be activated.
: 3. Assess and respond to the Alert condition.
: 4. Dispatch onsite and offsite survey teams and associated communications.
: 5. Provide periodic plant status updates to offsite authorities.
: 6. Provide periodic meteorological assessments to offsite authorities and, if any releases are occurring, dose estimates for actual releases.
: 7. Terminate by contacting offsite authorities.
or
 
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: 8. Escalate to a more severe class.
SITE AREA EMERGENCY
: 1. Promptly inform offsite authorities of Site Area Emergency status and reason for emergency as soon as discovered.
: 2. Augment resources by activating onsite Technical Support Center (TSC), onsite Operational Support Center (OSC) and Emergency Operations Facility (EOF).
: 3. Assess and respond to the Site Area Emergency.
: 4. If radiological or environmental conditions permit, evacuate onsite, nonessential personnel.
: 5. Dispatch onsite and offsite survey teams and associated communications.
: 6. Provide a dedicated individual for plant status updates to offsite authorities.
: 7. Make senior technical and management staff onsite available for consultation with NRC and State on a periodic basis.
: 8. Provide meteorological and dose estimates to offsite authorities for actual release via a dedicated individual.
: 9. Provide release and dose projections based on available plant condition information and foreseeable contingencies.
: 10. Terminate emergency class by contacting offsite authorities and initiate recovery phase.
or
: 11. Escalate to General Emergency class.
 
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E-PLAN EP                                EMERGENCY PLAN                                  REV:      58 Page 67 of 164 GENERAL EMERGENCY
: 1. Promptly inform offsite authorities of General Emergency status, appropriate offsite protective action recommendations and reason for emergency as soon as discovered.
: 2. Augment resources by activating onsite Technical Support Center (TSC), onsite Operational Support Center (OSC) and Emergency Operations Facility (EOF).
: 3. Assess and respond to General Emergency.
: 4. If radiological or environmental conditions permit, evacuate onsite, nonessential personnel.
: 5. Dispatch onsite and offsite survey teams and associated communications.
: 6. Provide a dedicated individual for plant status updates to offsite authorities.
: 7. Make senior technical and management staff onsite available for consultation with NRC and State on a periodic basis.
: 8. Provide meteorological and dose estimates to offsite authorities for actual releases via a dedicated individual.
: 9. Provide release and dose projections based on available plant condition information and foreseeable contingencies.
: 10. Terminate emergency class by contacting offsite authorities and initiate recovery phase.
 
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E-PLAN EP                              EMERGENCY PLAN                                  REV:      58 Page 68 of 164 6.4    Assessment Actions 6.4.1  Dose Projections Dose projections may be performed by using the standard dose projection program RASCAL (Radiological Assessment System for Consequence Analysis). Radioactive effluent release and meteorological data is procured from the Emergency Response Computer System (ERCS) and entered into RASCAL for real time dose assessments during inadvertent release of radioactive materials. The RASCAL program may be run from terminals that are located in the Control Room, TSC, EOF, and Backup EOF.
Meteorological data is stored and processed in the ERCS. The onsite 60 meter meteorological tower supplies the following:
A. Wind speed (10 and 60 meters)
B. Wind direction (10 and 60 meters)
C. Ambient Temperature D. DT (between 10 and 60 meters)
E. Rainfall A 22 meter backup meteorological tower is located near the EOF, and supplies the following:
A. Wind speed (22 meters)
B. Wind direction (22 meters)
 
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E-PLAN EP                                EMERGENCY PLAN                                    REV:      58 Page 69 of 164 Redundant instrumentation is provided on the onsite 60 meter meteorological tower, and may be designated as primary and secondary sensors. The 22 meter backup tower provides a set of tertiary sensors.
The ERCS continuously collects the meteorological data. Meteorological data from all three sets of instruments are displayed simultaneously as well as the calculated stability class (derived from the temperature readings). If all met data is unavailable, manual entry of met data may be made for accident calculations.
Surveillances and quality checks are performed on the meteorological tower equipment and data to ensure emergency responders will have access to representative onsite meteorological data. A daily review of a weeks trend of meteorological data is performed. The meteorological tower instruments are functionally tested monthly and calibrated at least annually.
Radiological effluent monitor data is also stored and processed in the ERCS. The effluent monitor reading, the calibration conversion factor and the vent flow rate result in a release rate. Effluent concentrations may also be manually entered into the computer if monitor data is not automatically available to the ERCS.
With meteorological and effluent release data available, calculations of offsite radiation dose, air concentration, ground deposition, and external dose rate from the plume can be made. Dose calculations are made for Total Effective Dose Equivalents (TEDE) and Thyroid Committed Dose Equivalents (Thyroid CDE). Results of all calculations can be printed in report format and, in most cases, displayed graphically. Isopleths can be displayed of any or all calculated outputs. Projected calculations take into account values of time of release and duration of release. The isopleth displayed is based on the assumption that the release continues for a predetermined duration time. This gives a display in which the plume overlays the region of potential highest dose.
The dose assessment computer allows quick accident dose calculations to be made, before any results from the Radiation Survey Teams are received. Radiation Survey Team results will be used to verify the dose calculations.
 
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E-PLAN EP                              EMERGENCY PLAN                                    REV:      58 Page 70 of 164 In case of potential release from the containment, the activity available for a release may be obtained from the containment high range dome monitors, as illustrated in Figure 5. The containment dome monitor reading and applicable calibration curve results in an activity available for release, and using an estimated release rate, an offsite dose calculation within the plume exposure pathway may be projected. The activity available in containment may also be obtained directly from sample analysis.
The containment dome monitors are also used as indicators for relative amounts of core damage, as illustrated in Figure 5. The indicated radiation levels in the containment gives an estimate of the gaseous radioactive concentrations in containment. Using the time after shutdown and the radiation levels, an estimate of the relative amount of core damage may be made. This must be used in a confirmatory sense, that is, as backup to other measurements of fission product release and other indicators.
 
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E-PLAN EP                              EMERGENCY PLAN                          REV:      58 Page 71 of 164 Figure 5    Containment Dose Rate Versus Time
 
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E-PLAN EP                              EMERGENCY PLAN                                    REV:      58 Page 72 of 164 The capability for remote interrogation of the meteorological data will be provided to NRC by either the Emergency Response Data System (ERDS) or direct telephone access to the individual responsible for making offsite dose projections. Implementing procedures will detail this activity.
A hand calculation methodology for offsite dose calculations is available in case of computer system and/or meteorological system failure.
Additionally, meteorological data may be obtained from local offsite locations. Atmospheric stability class and weather forecast information is available from the National Weather Service Twin Cities.
The capability to estimate the total offsite population dose (manrem) received during a release is available. The offsite dose assessment computer will supply the projected dose rates or doses at selected distances from the plant. Radiation Survey Team results may also be used to determine the offsite dose rates. Population distribution charts comprised of the geopolitical subareas are available. The Radiological Emergency Coordinator in the TSC or the Radiological Protection Support Supervisor in the EOF may determine the applicable dose rates in the geopolitical subarea and multiply dose rate times the exposure time, times the population in the geopolitical subarea of interest, thereby calculating an estimated total population dose.
The Emergency Director SHALL ensure that radiological information (both actual and potential) and recommendations for protective actions are transmitted to the offsite authorities. Upon activation of the EOF Organization, the responsibility for offsite accident assessment is transferred to the EOF. The EOF will serve as a base of operations for all site environmental surveillance, receipt and analysis of all field monitoring data, offsite dose projection and recommendations for offsite protective actions.
 
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E-PLAN EP                                EMERGENCY PLAN                                    REV:      58 Page 73 of 164 6.4.2  Radiological Surveys The Radiation Protection Group SHALL be responsible for all radiological surveys and personnel monitoring both onsite and offsite. The Emergency Director has the responsibility for directing all radiation safety during the emergency.
The Radiation Protection Specialists may be divided into two emergency Radiation Survey Teams. The teams are assigned offsite duties such as radiation surveys, air samples, or liquid sampling. The two offsite survey teams will conduct a search for the plume and obtain dose rates, and iodine, particulate or gaseous samples at pre-designated sample locations.
Plume exposure pathway maps with pre-designated sample locations are contained in the emergency survey kits. Additional duties onsite such as radiation surveys, sampling (airborne or liquid) and sample analysis using the equipment available onsite and/or the EOF Count Room facility are completed by other augmented personnel. Silver zeolite adsorbers are used to collect airborne iodine samples, both onsite and offsite. Silver zeolite adsorbers eliminate the problem of entrapped noble gases on the iodine adsorber, allowing a much lower detection sensitivity. Iodine samples may be analyzed in the EOF Counting Room.
The Radiation Survey Teams are activated via the ERO Auto Dial System and/or the ERO Pager Network or the telephone system. If the emergency occurs during normal working hours, the teams will be activated and respond within 10 minutes. If the emergency occurs during off hours, the first team will be activated and respond within sixty (60) minutes and the second team within ninety (90) minutes. Designated Emergency Lockers contain emergency survey kits, which include portable instruments, battery operated air samplers, liquid sampling equipment, and communication equipment.
 
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E-PLAN EP                                EMERGENCY PLAN                                      REV:      58 Page 74 of 164 6.5    Corrective Actions Certain actions may be taken by the Prairie Island staff during an emergency which may minimize the severity of the accident and lessen the amount of offsite releases.
These actions are outlined in the various standing plant abnormal operating, emergency operating, and plant safety procedures.
Repair and Damage Control is the responsibility of the Emergency Director and Shift Supervisors. During the onset of the emergency, plant operators are responsible for minor damage repair and control. Upon activation of the Plant Emergency Organization, equipment repair activities are the responsibility of the Maintenance Group, the I&C Group, the Electrical Group, and the Operations Group depending upon the extent and type of damage. Repair and damage control on radioactive or contaminated systems will be monitored by the Radiation Protection Group.
The Fire Brigade is composed of personnel in accordance with NRC requirements and is directed by a Fire Brigade Chief. Backup support is from the Red Wing Fire Department. All onsite Fire Brigade members are trained in the use of onsite fire fighting equipment and in proper fire fighting procedures. The Fire Brigade will be placed in action under the direction of the Brigade Chief.
6.6    Protective Actions 6.6.1    Evacuation and Sheltering In the course of an emergency at Prairie Island NGP when there is an actual or potential release of radioactive material to the environs in excess of normal operating levels, the Emergency Director SHALL be responsible to ensure that an assessment is made of the projected doses to persons onsite and offsite. Upon activation of the EOF, the Emergency Manager SHALL be responsible for ensuring that all assessments are made of the projected doses to the offsite population.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                  REV:      58 Page 75 of 164 The Protective Action Guides (PAGs), promulgated by the EPA, set dose guides for the offsite population. The Emergency Director also has the responsibility to ensure that protective actions are also taken to maintain exposure to onsite personnel within the PAGs. The Emergency Director or the Emergency Manager when the EOF is activated SHALL be responsible to recommend to the state and local authorities any protective actions for the offsite population whether the protective actions be based on predetermined Emergency Action Levels (EALs) or projected offsite dose assessment.
Plant Emergency Organization personnel fall into the category of Emergency Workers to which higher PAGs apply. The Emergency Director has the responsibility of maintaining doses within these PAGs.
A. Plant Site The primary protective measure for non-essential onsite personnel during a Site Area or General Emergency and possibly during an Alert, is evacuation to a suitable assembly area where the personnel can be monitored for contamination. The Emergency Director or Shift Manager, prior to ordering an evacuation SHALL determine the habitability of the assembly area (wind direction, magnitude of release, etc.).
If the normal onsite assembly area is determined to be uninhabitable, the Emergency Director will select a location farther from the plant site and designate the route to this location.
The Control Room operator will sound the evacuation alarm and announce the designated assembly area. If a location offsite is selected, the traffic route and area SHALL be announced.
Once non-essential personnel are accounted for and monitored for contamination, they may be released from the assembly area.
The evacuation routes from the assembly areas are limited to two directions: County 18 to Etter to Hwy 316 or County 18 to Hwy 61.
High water conditions may make the Etter route unusable, leaving only the County 18 to Hwy 61 route available. Prairie Island NGP vehicles and personal cars will be used to transport all personnel.
If conditions (meteorological or radiological) make land routes unavailable, evacuation by alternate means, (e.g., aircraft or watercraft) may be a viable alternative.
 
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E-PLAN EP                              EMERGENCY PLAN                                    REV:      58 Page 76 of 164 All non-essential personnel SHALL evacuate to the designated assembly area. The plant security force will assist in the evacuation by directing people to the proper assembly area. The Security Force SHALL direct employees to badge out of the Protected Area while exiting the Protected Area. The Security Force will perform an immediate check of the Protected Area to ensure that all personnel did indeed hear the evacuation alarm. The Security Force will perform a check of the Owner Controlled Area and warn all personnel of the evacuation in progress.
Radiation Survey Team Members, extra on-shift operators, group managers, Maintenance Supervisors, I&C Supervisors, Lead Maintenance Personnel and Station Electricians SHALL report to the Operational Support Center or the Technical Support Center, as applicable. Plant staff without emergency assignments SHALL evacuate to the designated assembly area. NRC Resident Inspector(s) may proceed to the Technical Support Center or Control Room.
Designated individuals SHALL complete an accountability check of personnel remaining within the Protected Area by verifying a list of personnel remaining in the Protected Area. The Emergency Director accepts responsibility for solving any discrepancies found during the accountability. The Emergency Director SHALL direct the necessary follow-up actions.
The Radiation Protection Group or qualified personnel SHALL monitor personnel at the assembly area for contamination, and any exposure determinations SHALL be completed, as conditions warrant. The emergency locker contains material necessary for decontamination of personnel under Radiation Protection Group supervision.
The assembly area SHALL remain in contact with the Emergency Director or designee via the telephone system or portable radio supplied in the emergency locker. The individual assigned as the Coordinator at the assembly area will be the contact point for all personnel.
The Emergency Director SHALL release non-essential personnel for departure from the site when conditions allow or demand this action.
The Emergency Director will designate the proper traffic routes to follow during the departure.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                                    REV:      58 Page 77 of 164 Onsite Protective Actions designed for protection of onsite personnel as described above may be inappropriate for a Hostile Action Event.
Alternate actions as described in NSIR/DRP-ISG-01 Section IV.F have been developed and proceduralized.
B. Offsite Areas The primary protective actions for the offsite population are sheltering or evacuation. The Emergency Director SHALL recommend the necessary protective actions to offsite authorities based on predetermined protective actions for a General Emergency Classification or results of offsite dose assessment. Upon activation of the EOF, the Emergency Manager SHALL be responsible for recommending protective actions for the offsite population. If protective actions are warranted prior to augmentation of state emergency response organizations, the Emergency Director SHALL recommend directly to county and tribal authorities the necessary protective actions. In both cases, total responsibility for carrying out the protective actions rests with offsite authorities. Prairie Island NGP SHALL make the recommendations and supply the required dose assessments.
C. Protective Action Guides (PAGs)
Table 2 and Table 3 provide guidelines and action levels to be used in the formulation of protective action recommendations for the offsite population and plant personnel.
The specific protective actions carried out by the offsite authorities are contained in their respective emergency plans.
D. Evacuation Time Estimates (ETE) - Plume Exposure EPZ.
Time estimates for evacuation of the plume exposure EPZ are referenced in an appendix to the Off-site Nuclear Emergency Plan and in the Plant Emergency Plan Implementing Procedure for making off-site protective action recommendations. PINGP and the States of Minnesota and Wisconsin use the ETE to develop pre-determined protective action recommendations.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                                    REV:      58 Page 78 of 164 6.6.2  Use of Protective Equipment and Supplies A. Onsite Respiratory Protection and Protective Clothing Protective clothing or respiratory protection for onsite personnel SHALL be as designated by the Radiation Protection Group or the Emergency Director.
Respiratory Protection will be used as necessary to reduce the inhalation of radioactive material. During emergency conditions, it may become impossible to maintain normal respiratory protection guidelines. An internal exposure program, whole body counting and/or bioassay program, SHALL be activated to ensure that all internal exposure is determined as assigned to the individual.
Respiratory equipment is stored in the OSC and TSC emergency lockers, Unit 1 695 Turbine Building Chem. Feed Station Area, Fire Brigade equipment room, and Access Control. Access Control is the main storage area for respiratory equipment. The respiratory equipment available is a combination of Self Contained Breathing Apparatus (SCBAs), and full face canister respirators.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                            EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                              REV:      58 Page 79 of 164 Table 2          Initial Protective Action Recommendation During a General Emergency The following situations require urgent actions by offsite officials. Conditions are based on Control Room indications with no dose projections required. The following protective action recommendations SHALL be made within 15 minutes.
Prerequisite: Plant Staff Detects GENERAL EMERGENCY
: 1. If wind is > 5 mph:        (1)      IF HAB Event concurrent with GE - Issue PAR to SHELTER ALL out to 2 Miles, Evacuate Downwind From 2 Miles to 5 Miles and Circle affected Subareas From 2 Miles to 5 Miles.
(2)      IF Rapidly progressing severe accident with all of the following:
This PAR is the initial after a GE has been declared AND There is LOSS of the containment barrier per the Emergency Action Levels AND Either of the following:
: a. Greater than or equal to Containment High Range Radiation Monitor Potential Loss Threshold (20%
Clad Damage) i.e. 1(2) R-48 or 49 reading > 20,000 R/hr OR
: b. An Offsite Dose Estimate indicates greater than PAGs at the site boundary is occurring or is likely to occur in an hour.
Issue PAR to Evacuate ALL out to 2 Miles, Evacuate Downwind From 2 Miles to 10 Miles and Circle affected Subareas From 2 Miles to 10 Miles.
(3)      IF Ongoing Rad release > EPA PAGs expected to be < 1 hour -
Issue PAR to SHELTER ALL out to 2 Miles, SHELTER Downwind From 2 Miles to 5 Miles, extend SIP 5-10 Mile if PAGS exceeded at 5 Mile and Circle affected Subareas From 2 Miles to 5 Miles.
(4)      Issue PAR to Evacuate ALL out to 2 Miles, Evacuate Downwind From 2 Miles to 5 Miles and Circle affected Subareas From 2 Miles out to 5 Miles.
(5)      Advise Remainder of Plume EPZ to Monitor EAS Broadcasts.
(6)      Continue with Step 2.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                            EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                                REV:        58 Page 80 of 164 Met Data from 22 meter tower OR wind < 5 mph OR unknown?      (1)        IF HAB Event concurrent with GE - Issue PAR to SHELTER ALL out to 2 Miles, Evacuate ALL From 2 Miles to 5 Miles and Circle ALL Subareas From 2 Miles to 5 Miles.
(2)        IF Rapidly progressing severe accident with all of the following:
This PAR is the initial after a GE has been declared AND There is LOSS of the containment barrier per the Emergency Action Levels AND Either of the following:
: a. Greater than or equal to Containment High Range Radiation Monitor Potential Loss Threshold (20% Clad Damage) i.e.
1(2) R-48 or 49 reading > 20,000 R/hr OR
: b. A RASCAL Dose Estimate indicates greater than PAGs at the boundary is occurring or is likely to occur in an hour.
Issue PAR to Evacuate ALL out to 10 Miles and Circle ALL Subareas From 2 Miles to 10 Miles.
(3)        IF Ongoing Rad release > EPA PAGs expected to be < 1 hour -
Issue PAR to SHELTER ALL out to 5 Miles and Circle ALL Subareas From 2 Miles to 5 Miles.
(4)        Issue PAR to Evacuate ALL out to 5 Miles and Circle ALL Subareas From 2 Miles out to 5 Miles.
(5)        Advise Remainder of Plume EPZ to Monitor EAS Broadcasts.
(6)        Continue with Step 2.
: 2. Continue with dose assessment throughout the emergency and revise initial Protective Action Recommendations in accordance with the protective action guidelines in Table 3.
The protective action recommendations described above are based on NOTES:          NRC Response Technical Manual, RTM-96, Vol. 1, Rev. 5, October 2002 and EPA 400-R-92-001, May 1992.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                      EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                        REV:      58 Page 81 of 164 Table 3 Recommended Protective Action to Avoid External and Internal Dose from Exposure to a Gaseous Plume PAGs for Early Phase Projected Doses Offsite Projected          Recommended Doses (mrem)                Protective Actions              Comments TEDE < 1000                No recommended                  The states of MN and WI may choose to Thyroid CDE < 5000          protective actions              implement sheltering or precautionary evacuation for the general public at their discretion.
TEDE  1000                Evacuate those sectors          Evacuation should be recommended in Thyroid CDE  5000          and distances where the          absence of local constraints. MN and WI PAG is exceeded. Use            may choose to shelter if evacuation were 0-2, 0-5 & 0-10 mile            not immediately possible due to offsite distances. Shelter for          constraints (severe weather, competing known impediments to            disasters or local traffic constraints).
evacuation and/or controlled puff release.
Notes: 1. TEDE = Total Effective Dose Equivalent, Thyroid CDE = Thyroid Committed Dose Equivalent
: 2. Based on EPA 400-R-92-001, May 1992
: 3. The Skin CDE PAG for evacuation of the general public is 50,000 mrem
: 4. Offsite projected doses include exposure from radioactive plume (external & internal) and 4 day exposure to ground contamination.
: 5. Known impediments to evacuation are conditions which make evacuation of the public impractical. Conditions include inclement weather (ice/snow storms where driving would be dangerous), and known impacts on the ability to execute public evacuations (severe damage to roads/infrastructures, etc.).
: 6. Controlled puff release exists when there is assurance that the release is short term (puff release) and the area near the plant cannot be evacuated before the plume arrives.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                      EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                        REV:      58 Page 82 of 164 Table 3 Recommended Protective Action to Avoid External and Internal Dose from Exposure to a Gaseous Plume PAGs for Emergency Workers TEDE Dose Limit (mrem)                            Activity                      Condition 5,000                            All emergency activities      This dose limit applies when a lower dose is not practicable through application of ALARA practices.
10,000                            Protecting valuable            Lower dose not practicable property 25,000                            Life saving or protection      Lower dose not practicable of large populations
>25,000                          Life saving or protection      Doses in excess of this dose limit of large populations          SHALL only be on a voluntary basis to persons fully aware of the risks involved.
Notes: 1. Based on EPA 400-R-92-001, May 1992
: 2. These are doses to non-pregnant adults from external exposure and intake during an emergency.
: 3. Exposures to the lens of the eye should be limited to 3 times the values listed and doses to the skin and/or extremities and any other organ should be limited to 10 times the values listed.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                  EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                                      REV:      58 Page 83 of 164 B. Radioprotective Drug The use of a stable iodine thyroid blocking agent, Potassium Iodide (KI), for plant staff and personnel assigned to onsite emergency operating centers is recommended in situations where airborne iodine concentrations have or could increase to unacceptable concentrations resulting in thyroid doses greater than 25 Rem (final recommendation by the Food and Drug Administration).
The Radiological Emergency Coordinator SHALL recommend the distribution of Potassium Iodide (KI). The Emergency Director SHALL then direct the distribution of Potassium Iodide (KI).
The Potassium Iodide (KI) tablets are stored in the TSC Emergency Locker, EOF Emergency Locker, and in the Field Survey Kits. The tablets will be distributed per the applicable implementing procedure.
C. Shielding All plant personnel, who are required to occupy the emergency operating centers, (i.e., Tech Support Center and the Control Room),
are protected from intense radiation fields and high airborne radioactivity levels by shielding and/or emergency air handling equipment.
All reactor coolant system sampling and radiochemical analysis may be completed using the shielded sampling system with reach rods in the hot sample room and a lead brick shielded work area in the hot cell area.
D. Offsite Areas There are no plans for the distribution of respiratory protective equipment and/or protective clothing for the general public. The distribution of thyroid blocking agents is the responsibility of the offsite officials. All Protective Actions to be taken for the general public are described in the offsite emergency plans.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                              EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                                    REV:      58 Page 84 of 164 6.6.3  Contamination Control Measures A. Onsite Areas The Emergency Director SHALL designate the Radiation Protection Group responsible for controlling or minimizing direct or subsequent internal exposure from radioactive materials deposited on the ground or other surfaces. The Radiation Protection Group SHALL be responsible for determining the extent of contamination in controlled and normally uncontrolled areas. During an emergency, guidelines to follow for contamination limits are shown in Table 4.
The Radiation Protection Group with assistance from the Security Force will establish new secondary access control points at the boundaries of the new controlled areas to ensure that all personnel entering the areas are properly badged and clothed.
The Radiation Protection group SHALL advise all personnel that contamination levels in some uncontrolled areas may significantly exceed normal levels. Without protective clothing, personnel will have to take precautions to avoid personal contamination. Limits for personal contamination will remain at the normal limits which will minimize the chance of ingestion of radioactive material.
Table 4    Contamination Limits LARGE AREA LOOSE SURFACE                                          FIXED WIPE Authorized Limits                                            (Beta-Gamma)        (Beta-Gamma)
(Beta-Gamma / Alpha) / 100 sq cm (Circle One)                                          (GM pancake probe)  (GM pancake probe)
NORMAL                < 100/10 dpm/100 cm2              < 100 ccpm          < 100 ccpm ELEVATED                < 1000/20 dpm/100 cm2              < 100 ccpm          < 100 ccpm EMERGENCY                < 5000/100 dpm/100 cm2              < 500 ccpm          < 500 ccpm Based on Manual of Protective Action Guides and Protective Actions for Nuclear Accidents, EPA 400-R-92-001, May 1992, Table 7-7.
Frisker response: 1 mR/hr  5000 cpm Cs 137.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                              EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                                    REV:      58 Page 85 of 164 Particular attention will be given to radioiodine contamination of the skin. Oxidizing agents, e.g., Beta dyne or Radiac Wash, are available in the decontamination kits to treat iodine skin contamination.
The Radiation Protection Group SHALL have the responsibility of controlling all onsite food and water supplies during the emergency.
Whenever a plant evacuation takes place involving radiological hazards onsite, all food and water supplies within the evacuation area may be considered contaminated and not for use.
Material decontamination SHALL be performed by the Nuclear Plant Service Attendants or designated personnel under supervision of the Radiation Protection Group. Procedures and equipment for material decontamination are listed in the Decontamination Procedures of the Operations Manual, Sections F-2 and D-13, and in the Radiation Protection Manual RPIPs.
Before any water or food can be consumed, the Radiation Protection Group will check and verify that the food itself and the eating surfaces are below the limits of Section F-2 of the Operations Manual (previously recorded). Random samples of food containers may be analyzed via the GEM Detector for low level contamination not detected by other methods.
During the recovery phase, all areas of the plant will be returned to the original low levels of surface contamination prior to their release for unrestricted entry.
B. Offsite Areas Contamination control in offsite areas is the responsibility of offsite officials with assistance from Prairie Island NGP. Required protective actions are delineated in Protective Actions guides and criteria are listed in the respective state emergency plans.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                              EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                REV:      58 Page 86 of 164 6.7    Aid to Affected Personnel The Emergency Director is responsible for the protection of personnel from exposure to radiation and contamination and arranging for treatment of radiologically induced or contaminated injuries. This responsibility may be delegated to the Radiation Protection Group.
6.7.1    Emergency Personnel Exposure The Prairie Island Radiation Protection Group has the necessary equipment and personnel required to provide continuous capability to control and determine radiation exposures of emergency organization personnel. The equipment consists of the following:
A. portable radiation detection instruments B. electronic dosimeters C. high and low range dosimeters D. DLRs E. extra high range dosimeters F. record keeping equipment Contractor and vendor representatives may also be present to assist in exposure control and augment the Radiation Protection Group capabilities.
In an emergency situation, all onsite personnel, some offsite support personnel and some local governmental emergency response personnel will be issued DLRs and/or SRDs. Exposure records will be maintained for all emergency response personnel issued dosimetry.
During accident situations, higher radiation exposures may be authorized by the Emergency Director in order to protect life and property. The emergency exposure guidelines established are based on the Environmental Protection Agencys PAGs for Emergency Workers, as listed in Table 3.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                  EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                    REV:      58 Page 87 of 164 Emergency workers (volunteers) may be allowed to exceed the 10CFR20 limits with specific authorization of the Emergency Director when performing activities to protect life and property.
In certain instances, it may be necessary to exceed 25 Rem exposure during lifesaving operations. All personnel involved SHALL be on a volunteer basis and will be advised of the effects of acute exposures and reasonable considerations of the relative risks.
In all circumstances, every effort SHALL be made to keep NOTE:          exposures within the annual limits of 10CFR20 (5 Rem Total Effective Dose Equivalent).
6.7.2  Decontamination and First Aid The Emergency Director SHALL delegate the responsibility for personnel decontamination to the Radiation Protection Group. Decontamination procedures and contamination limits are spelled out in the Radiation Protection Manual RPIPs and the Radiation Safety and Medical Support Sections of the Operations Manual, which SHALL be followed for both normal and emergency situations involving personnel injury and personnel contamination.
The primary decontamination facility is located at access control. Two showers and a double sink are located there. Special decontamination solutions are also available at access control.
When facilities at access control are not available, the assembly area emergency lockers contain equipment for personnel decontamination and personnel monitoring. Supplies include containers for liquid and solid waste. The decontamination kits contain oxidizing agents for decontamination of the skin due to radioiodines.
Decontamination operations at the assembly area will be confined to minor decontaminations because of limited resources. If necessary, individuals will be furnished with protective clothing and transported to alternate facilities. Contaminated clothing will be disposed of as radioactive waste.
The EOF has a decontamination shower with associated liquid retention system. Equipment for small decontaminations is also available along with personnel monitoring equipment.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                    EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                        REV:      58 Page 88 of 164 Contaminated individuals may be provided whole body counting analysis, as determined by the Radiological Emergency Coordinator. Whole body counting systems are located at PI & MT NGPs and/or mobile units which can be transported on or near the site.
Emergency First Aid will be applied to all injuries including contaminated injuries since contamination will not be life threatening whereas the lack of first aid could be life threatening.
First aid kits are located at the primary emergency centers in the plant.
The First Aid responsibility will be assigned to the Security Officer/EMT when they arrive on the scene. Selected members of the Security Force and plant staff are trained in Advanced First Aid and/or Emergency Medical Training (EMT).
The skill level of the staff is sufficient until offsite medical personnel arrive or until the victim is transported to the local hospital for further medical treatment.
The Operations Manual, Section F4, Medical Support and Casualty care, contains specific procedures for first aid situations complicated by contamination.
6.7.3  Medical and Public Health Support Medical support and treatment for radiological and non-radiological injuries is provided by the Mayo Clinic Health System located in Red Wing, Minnesota.
Mayo Clinic Health System has a staff of physicians and hospital personnel trained in the proper methods of contamination control. At least one physician has been offered special courses on the treatment of radiological injuries. Prairie Island NGP conducts yearly training sessions with hospital personnel assuring a knowledge of radiation and contamination control procedures.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                                    REV:      58 Page 89 of 164 Regions Hospital in St. Paul, Minnesota is designated as the definitive care center for Prairie Island Nuclear Generating Plant. Regions Hospital may be used for radiation casualties, severe burn casualties, and other non-radiation injuries with use of an appropriate medical air transport service. Medical definitive care centers are offered periodic radiological contamination control training by the Minnesota Division of Homeland Security and Emergency Management (HSEM) according to their plan.
Monitoring instruments and supplies are located at Mayo Clinic Health System to aid in radiation monitoring and contamination control (e.g.,
DLRs, SRDs, protective clothing, survey meters, etc.).
All casualties on site will be administered emergency First Aid and radiation casualties will be decontaminated to every extent possible prior to departure from the plant site to the hospital. Proper application of first aid will take precedence over decontamination efforts.
Transportation of radiation casualties from Prairie Island NGP will be provided by the Red Wing Ambulance Service. In addition to the Red Wing Ambulance, a plant vehicle could be used as an emergency vehicle for transportation of victims to the hospital.
Procedures to be used at the plant and at the hospital in treating victims of an accident involving radiation exposure and/or personnel contamination are established and delineated in Section F4 of the Plant Operations Manual, Medical Support and Casualty Care.
In addition Sacred Heart Hospital in Eau Claire, WI is prepared and will support request for assistance in response to an emergency at the Prairie Island Nuclear Plant. Sacred Heart Hospital will serve as a radiation accident receiving hospital and has a decontamination room and trauma treatment rooms with isolation capabilities.
6.7.4  Whole Body Counting Facilities A whole body counter is available at the Prairie Island NGP for determining the uptake of radioactivity. If this area becomes uninhabitable, the person may be transported to Monticello NGP where another whole body counter is available. Additional mobile whole body counters may be brought near or on the site if conditions make it a viable or necessary alternative.
 
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E-PLAN EP                              EMERGENCY PLAN                                  REV:      58 Page 91 of 164 7.0    EMERGENCY FACILITIES AND EQUIPMENT 7.1    Emergency Control Centers 7.1.1  Technical Support Center (TSC)
The Technical Support Center (TSC) is located across the Turbine Building from Units 1 & 2 Control Room. A plan view of the TSC is shown in Figure 6.
The Technical Support Center (TSC) will serve as a center outside the Control Room from which the plant management, technical, and engineering support personnel will:
A. Support the Control Room command and control functions B. Assess the plant status and potential offsite impact C. Coordinate emergency response actions The Technical Support Center has the following capabilities:
A. Working space for about twenty-five people on the main floor and working space for additional people on the other floor.
B. Shielding and ventilation cleanup system (PAC filter) to provide habitability under accident conditions.
C. An emergency locker containing monitoring equipment (radiation and airborne), respiratory protection equipment and thyroid blocking agent tablets.
D. Communication channels to all onsite and offsite emergency response centers (primary and backup).
E. A complete set of as-built drawings and other records such as plant layout drawings.
 
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E-PLAN EP                                EMERGENCY PLAN                      REV:      58 Page 93 of 164 Figure 6 Plan View of TSC
 
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E-PLAN EP                                EMERGENCY PLAN                                  REV:      58 Page 95 of 164 F. The capability to record and display the following:
: 1. Plant System Parameters a    Reactor Coolant System b    Secondary System c    ECCS System d    Containment
: 2. In-Plant Radiological Parameters a    Reactor Coolant System b    Containment c    Effluent Treatment d    Release Paths e    Area Monitors
: 3. Offsite Radiological Parameters a    Meteorology b    Offsite Radiation Levels The Technical Support Center SHALL be activated within 60 minutes when an Alert, Site Area or General Emergency is declared.
The Technical Support Center Coordinator SHALL be responsible for coordinating activities in the TSC. This individual SHALL be responsible for establishing the monitoring of direct radiation and airborne activity in the Technical Support Center. Communications SHALL be established between the TSC, OSC, Control Room and EOF.
If activation of the Technical Support Center occurs during normal work hours, instructions to report to the TSC will be received over the plant public address system.
If activation of the Technical Support Center occurs during the off duty hours, the Shift Manager SHALL designate the Shift Emergency Communicator to contact the Emergency Response Organization (ERO) by phone and/or ERO Pager Network and request them to report to the Technical Support Center.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                    REV:      58 Page 96 of 164 7.1.2    Operational Support Center (OSC)
The Operational Support Center will provide a center to assemble the necessary Operators, Radiation Protection Specialists, Instrument and Control, Electrical, Nuclear Plant Service Attendants, and Maintenance personnel to support the operations of the plant under emergency conditions without causing undue congestion in the Control Room.
The Operational Support Center is located in the New Administration Building.
The Operational Support Center will be activated within 60 minutes when an Alert, Site Area or General Emergency is declared.
The Operational Support Center Coordinator SHALL be responsible for the activation and coordination of activities in the OSC. The OSC Coordinator may designate a communicator to establish lines of communications between the Operational Support Center, the Control Room and the Technical Support Center.
If activation of the OSC occurs during a normal working day, instructions to report to the OSC will be received over the plant public address system.
Any Operations shift personnel on site that are not assigned to normal shift duty SHALL report to the OSC immediately. The following personnel will also report to the OSC if on site (additional personnel will be contacted as necessary):
A. Maintenance Supervisors (Mechanical and Electrical)
B. Designated Lead Station Electricians and Maintenance personnel C. Instrument and Control Supervisors D. Radiation Survey Team Members E. Nuclear Plant Service Attendants
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                  EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                    REV:      58 Page 97 of 164 If activation of the Operational Support Center occurs during off duty hours, the Shift Manager SHALL designate the Shift Emergency Communicator to activate the onsite emergency organization to establish an initial complement of support personnel to assist in the emergency (additional personnel will be contacted as necessary):
A. Maintenance Supervisors (Mechanical and Electrical)
B. Designated Lead Station Electricians C. Instrument & Control Supervisors D. Radiation Survey Team Members E. Designated Purchasing & Inventory Control Personnel F. Nuclear Plant Service Attendants Instrumentation is stored in the emergency locker which provides for monitoring both direct radiation and airborne radioactive contaminants.
An emergency locker located in the OSC contains all equipment necessary for reentry into the plant. This includes but is not limited to both waterproof and paper coveralls, respiratory protection (SCBAs),
dosimeters, radiation detection meters, air samplers, decontamination and first aid equipment.
Communication equipment (radio and telephone) is available for contacting designated sections of the emergency response organizations.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                                    REV:      58 Page 98 of 164 7.1.3    Emergency Operations Facility (EOF)
The Emergency Operations Facility (EOF) is a required emergency response facility located near the plant site to provide continuous coordination and evaluation of activities during an emergency having, or potentially having, environmental consequences. A plan view of the EOF Command Center is shown in Figure 7. The EOF will be activated within 90 minutes when an Alert, Site Area or General Emergency is declared.
The functions of the EOF will be:
A. Management of the overall NSPMs offsite emergency response in support of plant activities; B. Evaluate the magnitude and effects of actual or potential radioactive releases from the plant; C. Recommend appropriate offsite protective measures, in conjunction with the TSC personnel; D. Coordinate the offsite radiological monitoring during emergencies and recovery operations; E. Coordinate emergency response activities with those of local, State, Tribal, and Federal emergency response organizations; F. Provide current information on conditions potentially affecting the public to the NRC and to offsite emergency response agencies; G. Act as the post-accident recovery management center for both onsite and offsite activities, if necessary.
The EOF will be staffed by personnel from the Engineering and Projects Management groups and Prairie Island Training Center staff. Activation and various responsibilities within the EOF are described fully in Section F8 of the EOF Emergency Plan Implementing Procedures.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                  EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                      REV:    58 Page 99 of 164 The EOF has been constructed and designed in accordance with the guidance of NUREG 0696. The building has been designed to serve primarily as a Training Center on a regular basis with the capability for prompt conversion to the EOF function when required and, if needed, will serve as the Recovery Center.
The EOF is constructed in a manner which provides habitability in an accident situation. Shielding and ventilation treatment systems have been installed to maintain an acceptable environment. The EOF section of the training building is a concrete structure that contains sufficient shielding to exceed a protection factor of 5. The ventilation system has an emergency mode of operation that will pressurize the building through a High Efficiency Particulate Absolute (HEPA) filtration system.
The general layout of the buildings entrances and exits have been given consideration for operation of the building in an emergency mode.
Radiological monitoring and alarming are provided for the EOF portion of the building. Extensive communication equipment is installed in the building to provide primary and backup means of communication with outside agencies, offsite survey teams, TSC and the Control Room. The EOF portion of the building is served by a dual source power supply for those services necessary to make the EOF functional.
The EOF provides office space for each plant support group, key supervisors, state, local and tribal officials, and the NRC, as well as functioning as a command center. Each space is provided with furnishings necessary to perform routine office functions. The plant support groups and governmental representatives will perform their respective functions in these assigned offices. The command center is intended to function as a work space for the Emergency Manager, Radiation Protection Support Group, Technical Support Group, and for related critical communications.
These activities are assigned to this area, due to the high volume of activity and the importance of the information handled. Additionally, this area is the central area for displaying plant status, offsite survey status, conducting accident assessment and directing the activities of the offsite Emergency Response Organization.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                    REV:      58 Page 100 of 164 The EOF is supplied with the equipment necessary to fulfill its function as an offsite emergency response center. Radiation monitoring and decontamination equipment has been provided to supply offsite monitoring teams. Normal and emergency data acquisition is made available via the Emergency Response Computer System (ERCS). Office equipment such as facsimile machines, copy machines, microfiche readers, computers and printers connected to the Local Area Network are provided to facilitate administrative duties and technical reference work. General office supplies are stocked in adequate numbers. Operating procedures detailing the methods to activate the EOF, conduct routine administrative operations, surveys and accident assessment, analyze offsite survey samples, provide security and deactivate the Emergency Organization are developed and are available in the EOF. Other organizations procedures, plans and reference documents are also available to EOF personnel. If there is a need for expanded support facilities such as trailer space or communication hook-ups for vendors and support contractors, it may be provided at the EOF.
Because the EOF is located within the 10 mile EPZ, a Backup EOF exists in case an evacuation of the EOF is necessary. Equipment and facilities necessary to carry out this function are located at Xcel Energy corporate offices in downtown Minneapolis, Minnesota. A description of the Backup EOF facility is described in the Monticello & Prairie Island Offsite Nuclear Emergency Plan.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                                        EMERGENCY PLAN                          REV:      58 Page 101 of 164 Figure 7  Plan View of EOF Command Center
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                            EMERGENCY PLAN NUMBER:
E-PLAN EP                                        EMERGENCY PLAN                      REV:      58 Page 102 of 164 THIS PAGE IS INTENTIONALLY LEFT BLANK
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                  EMERGENCY PLAN NUMBER:
E-PLAN EP                                  EMERGENCY PLAN                                    REV:      58 Page 103 of 164 7.1.4    Control Room The Control Room SHALL be the initial onsite center of emergency control.
Control Room personnel must evaluate and effect control over the initial aspects of the emergency and initiate responses necessary for coping with the initial phases of an emergency until such time that the onsite emergency centers can be activated. These activities SHALL include:
A. Continuous evaluation of the magnitude and potential consequences of an incident B. Initial corrective actions All plant operations are controlled from here by the Shift Manager with direction from the management personnel located either in the Control Room or Technical Support Center.
The Control Room contains the necessary instrumentation (process and radiological) to evaluate all plant conditions. Habitability is maintained by shielding and the special ventilation system (PAC Filter), which is capable of operating in a cleanup or recycle mode.
All emergency equipment is supplied power from the emergency diesel generators with vital instrumentation powered from inverters connected to the storage batteries located in the battery rooms.
7.1.5    The Red Wing Service Center (RWSC) is to be used as an Alternative Facility during a hostile action or security event in the event that response to the site is unsafe. The RWSC will be used by TSC and OSC personnel until it has been determined that it is safe to return to the plant site. This facility is accessible in the event of an onsite Hostile Action and provides the ability to perform the following functions:
* Communication with the Control Room and onsite Security Forces
* Notification of offsite Emergency Response Organizations
* Engineering Assessment Activities including damage control team preparation and planning.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                  REV:      58 Page 104 of 164 7.2    Communications Various onsite and offsite communication systems are described in the following sections. Table 5 depicts the various communication links that may be established.
7.2.1    Onsite Communications All emergency operating facilities have at least two means of communications: (1) portable or installed radio systems; and (2) normal telephone communications.
The normal onsite communications during an emergency will be made via the plant telephone system with a public address system option. The telephone system is powered by noninterruptible power. The public address system includes about 175 loudspeakers located throughout the entire plant area.
A separate paging system has 20 handsets located at strategic plant areas.
At approximately 120 locations in the plant, jackboxes are located for the sound powered system. Each box contains six independent circuits for sound powered headsets. A jackbox is located in the Technical Support Center and Control Room.
The Control Room, Technical Support Center and EOF each have a multi-channel radio system console for communications. At least 50 portable radios are available for use throughout the plant during emergency conditions.
The plant evacuation alarm consists of a 125 VDC operated siren, manually started from the Control Room. This tone consists of a signal starting at approximately 600 cycles per second rising to a peak of approximately 1450 cycles per second, then returning slowly to the low value of 600 cycles per second and repeating. The Control Room operator can remove the siren tone for emergency voice communication over the loudspeaker PA system.
The plant fire alarm consists of a modulating signal interrupted continuously to give a Yip-Yip-Yip sound. This is activated manually from the Control Room.
During an emergency, designated individuals will be responsible for the communications at each of the emergency facilities, as delineated in Section 6.1.3.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                  REV:      58 Page 105 of 164 7.2.2    Offsite Communications Both normal and alternate communication links are provided to offsite agencies. Individuals designated to staff the offsite agency communication links are delineated in Section 6.1.3.
The Xcel Energy telephone network provides normal communications to offsite agencies through telephone lines via the Red Wing US West telephone Exchange, or via Xcel Energy fiber optic SONET communications network.
The Control Room, Technical Support Center and EOF have a dedicated Xcel Energy radio channel link to the Xcel Energy System Control Center, the Backup EOF, and the Minnesota HSEM Emergency Operating Center in St. Paul, Minnesota.
The Technical Support Center and EOF have a National Warning System (NAWAS) extension to the Wisconsin Emergency Management EOC at Madison, the Regional Warning Center at Eau Claire and the Pierce County EOC at Ellsworth, Wisconsin.
The Control Room, Technical Support Center and EOF each have a portable cellular phone and satellite phone for emergency communication use, as necessary.
The Technical Support Center has access to a computerized auto dial system used for notification of the sites Emergency Response Organization (ERO). This system consists of a telephone network of several outgoing telephone lines. When activated, it will call and deliver an emergency message to the plants emergency organizations home telephones.
The plant also has an Emergency Response Organization (ERO) Pager Network. Designated members of the sites emergency organization carry personal pagers which can be activated from the Technical Support Center, Control Room or alternate facility (RWSC). A special emergency code is displayed on the pager.
The Control Room, Technical Support Center and EOF have multi-channel radio system for communication with all Plant Radiation Survey Teams, Plant Operations Personnel, Plant Security Areas, county sheriffs, county EOCs, and Treasure Island Casino (Prairie Island Indian Tribe).
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                    REV:      58 Page 106 of 164 A telecopying network is set up between the TSC, EOF, state & county EOCs and Prairie Island Indian Tribe for the purpose of telecopying update information.
An emailing network is setup between the offsite agencies for the purpose of emailing the emergency notification form.
Auto ring lines link the Technical Support Center to the EOF and the Technical Support Center to the Minnesota State EOC.
Communication links are maintained with medical facilities, both fixed and mobile. The plant can update the hospital via the telephone network of the status of any injuries. Communication channels are provided between the hospital and the ambulance service via the radio system while the victim(s) are enroute.
The plant site also supports the NRCs Emergency Telecommunications System (ETS). The dial tone for the Prairie Island 106G PETS circuits are provided by Xcel Energys corporate communication network. The ETS provides for reporting emergencies and other significant events to the NRC, Incidence Response Center in Rockville, Maryland. Using the Xcel Energys private network should avoid the public switched network blockage anticipated during a major emergency.
The following NRC essential emergency communications functions will be provided by the ETS voice service.
A. Emergency Notification System (ENS): Initial notification by the licensee, as well as ongoing information on plant systems, status, and parameters. The ENS (Red Phone) is located in the Control Room, with extensions in the Technical Support Center (TSC) and EOF.
B. Health Physics Network (HPN): Communication with the licensee on radiological conditions (in-plant and off-site) and meteorological conditions, as well as their assessment of trends and need for protective measures on-site and off-site. NRC regional office or NRC Headquarters will announce their decision to establish the HPN link over the ENS. The HPN phones are located in the TSC and EOF.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                              EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                                    REV:      58 Page 107 of 164 C. Reactor Safety Counterpart Link (RSCL): Established initially with the base team, and then with the NRC site team representatives once they arrive at the site, to conduct internal NRC discussions on plant and equipment conditions separate from the licensee, and without interfering with the exchange of information between the licensee and NRC. This is the channel by which the NRC Operations Center supports NRC reactor safety personnel at the site. In addition, this link may also be used for discussion between the Reactor Safety Team Director and licensee plant management at the site. The RSCL phones are located in the TSC and EOF.
D. Protective Measures Counterpart Link (PMCL): Established initially with the base team, and then with the NRC site team representatives once they arrive at the site, to conduct internal NRC discussions on radiological releases and meteorological conditions, and the need for protective actions separate from the licensee and without interfering with the exchange of information between the licensee and NRC.
This is the channel by which the NRC Operations Center supports NRC protective measures personnel at the site. In addition, this link may also be used for discussion between the Protective Measures Team Director and licensee plant management at the site. The PMCL phones are located in the TSC and EOF.
E. Emergency Response Data System (ERDS) Channel: This dedicated computer network is a direct near real-time electronic data link between the plants on-site computer system and the NRC Operations Center that provides for the automated transmission of a limited data set of selected parameters. The plant activates the ERDS within one hour after declaring an emergency class of Alert, Site Area, or General Emergency. The ERDS supplements the existing voice transmission over the ENS.
F. Management Counterpart Link (MCL): Established for any internal discussions between the Executive Team Director or Executive Team members and the NRC Director of Site Operations or top level licensee management at the site. The MCL phones are located in the TSC and EOF.
G. Local Area Network (LAN) Access: Established with the base team and the NRC site team for access to any of the products or services provided on the NRC Operations Centers local area network. This includes technical projections, press releases, status reports, E-Mail, and various computerized analytical tools. The LAN access points are located in the TSC and EOF.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                                                                                                                                                                          EMERGENCY PLAN NUMBER:
E-PLAN EP                                                          EMERGENCY PLAN                                                                                                                                                                    REV:  58 Page 108 of 164 Table 5                        Prairie Island Site Communications Matrix Portable Cellular Telephone                                                                                                                                                                                    USNRC MCL RSCL PMCL LAN Xcel Metro Radio System                                              ERO Auto Dial System Plant Sound Powered Plant Phone Network                                                        Plant Page System                                                                        Plant Radio System                                                Personal Pagers Auto Ring Direct Phone System                                                                  Facsimile  USNRC/ENS  USNRC HPN Email                                                                                                  NAWAS                                                                                        Dialing Phone Control Room                        x                          x                x                            x                                      x                      x                        x                        x                                      x Tech Support Center          x      x                          x                x                            x                  x      x          x          x          x                        x                        x                x                      x                x Ops Support Center                  x                                                                          x                                                                                        x                                                                x Emerg Op Facility            x      x                                            x                            x                  x      x          x          x          x                        x                        x                                      x                x Backup EOF                  x      x                                            x                                                        x          x          x          x                                                  x Xcel/ System Ops                    x                                                                                                                                          x                        x                        x Monticello NGP                      x PI Plant Areas                      x                          x                                              x                                                                                        x                                                                x PI Monitoring Teams                                                                                                                                                                                      x                                                                x PI Indian Tribe              x      x                                                                                                      x                                                            x MN/HSEM-EOC                  x      x                                                                                                      x                                  x                                                  x Goodhue Co. Sheriff          x      x                                                                                                      x                                                            x Goodhue-Red Wing EOC        x      x                                                                                                      x                                                            x Dakota Co. Sheriff          x      x                                                                                                      x                                                            x Dakota EOC                  x      x                                                                                                      x                                                            x WI/WEM-EOC                  x      x                                                                                              x      x WI/WEM-Eau Claire                    x                                                                                              x Pierce Co. Sheriff          x      x                                                                                              x      x                                                            x Pierce EOC                  x      x                                                                                              x      x                                                            x Red Wing Police/Fire                x Red Wing Hospital                    x                                                                                                      x USNRC/HQ                            x                                                                                                      x          x          x                                                                                                                        x USNRC/REG III                        x                                                                                                                  x          x                                                                                                                        x USNRC/Resident Insp.                x                                                                          x                                                                                                                                                          x PI Emerg. Personnel                  x                                                                                                                                                                                                              x                      x MN/State Patrol                      x WI/State Patrol                      x                                                                                              x National Weather Service            x
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                  EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                      REV:    58 Page 109 of 164 7.2.3    Alert and Notification System (ANS)
Within the Plume Exposure Emergency Planning Zone (EPZ) there exist provisions for alerting and providing notification to the public. It is the responsibility of state and county governments to activate this system.
The plant maintains a basic fixed siren system for essentially 100%
coverage of the offsite population within 5 miles of the plant and population center coverage for the 5-10 mile zone. To reach persons not covered by these population center sirens, Homeland Security Emergency Management or the MN Duty Officer also activates the Integrated Public and Alert Warning System (IPAWS).
A special electronic siren is maintained near the Prairie Island Indian Community Center. The TSC has the capability to activate the siren with a special stutter tone at the declaration of a Site Area Emergency for the purpose of quickly notifying Prairie Islands Indian tribal leaders except during a Hostile Action Based (HAB) event. The siren would also be activated with the normal Alert tone by the Goodhue County Sheriffs Department during a General Emergency as part of the normal Public Alert and Notification System activation.
To supplement PANS, emergency alert radios have been installed in various commercial, institutional, and educational facilities in the 10-mile zone. These locations may harbor large groups of people during all or part of a day, justifying radio alert service, even though many of these facilities are already covered by state and county emergency warning plans. The emergency radios will either be activated by the National Weather Service or by the local county sheriffs dispatch office.
In the event of an emergency condition, alert and notification information will be relayed through established communication links described in the Minnesota and Wisconsin emergency response plans. Upon receiving notification of the emergency, offsite governments will, if necessary, activate public warning and information procedures which include the State Emergency Alert System (EAS). With this system, essentially 100% of the population in the 10 mile EPZ will be alerted within 15 minutes In the event a county primary siren activation system fails to operate, each county has a backup siren activation process on a separate activation system utilizing a different tower and controls for activation of the sirens.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                  REV:      58 Page 110 of 164 In conjunction with the siren system activation, the Integrated Public Alert Warning System (IPAWS) is also activated. This system is also used as a backup when the siren system or individual sirens are out of service.
7.3    Assessment Facilities The plant instrumentation and monitors perform indicating, recording and protective functions. The Reactor Protection System and associated plant instrumentation provide the ability to maintain plant safety from shutdown to full power operations and to monitor and maintain key variables such as reactor power, flow, temperature, and radioactivity levels within predetermined safe limits at both steady state conditions and during plant transients. Plant instrumentation and control systems also provide means to cope with abnormal operating conditions. The control and display of information of these various systems are centralized in the main Control Room. This instrumentation would provide the basis for initiation of protective actions.
7.3.1    Onsite Systems and Equipment A. Geophysical Phenomena Monitors
: 1. Meteorological Prairie Island has a 60 meter onsite meteorological tower located approximately 0.5 miles northwest of the plant. The tower is equipped with primary and secondary redundant sensors for the 10 and 60 meter temperatures, wind speeds, and wind directions powered by a primary and secondary power source. The following meteorological information is supplied by the tower:
a    Wind Direction (10 and 60 meter) b    Wind Speed (10 and 60 meter) c    Ambient Temperature d    T between 10 and 60 meter temperature indications e    Precipitation A 22 meter backup meteorological tower is located near the EOF. The backup meteorological tower provides the following:
All meteorological data is processed via the ERCS, and may be displayed in the Control Room, TSC, EOF, and Backup EOF.
Barometric pressure is also available in the Control Room.
 
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: 2. Seismic The Control Room has an installed earthquake detection system with a three step graded severity level of alarms:
a    Seismic Event - 3 percent vertical or horizontal acceleration b    Operational Basis Earthquake - 4 percent vertical or 6 percent horizontal acceleration (No equipment failure) c    Design Basis Earthquake - 8 percent vertical or 12 percent horizontal acceleration (possible equipment failure)
A visual and audible alarm will sound in the Control Room.
Upon activation, the accelerometers and accelerographs listed on Table 6 will be automatically recorded for future investigation.
: 3. Hydrologic River water level is available from two sources:
a    Indicators in Control Room which receive a signal from capacitance level probes located in several locations in the river water canals and the intake screenhouse.
b    Lock and Dam #3 (located about 1.6 miles SE) which would give essentially the same indication as at Prairie Island NGP.
B. Radiation Monitoring Equipment Onsite radiation monitoring equipment at Prairie Island NGP can be categorized into the following groups:
: 1. Process radiological monitoring system
: 2. Effluent radiological monitoring system
: 3. Airborne radioactivity monitoring system
: 4. Area radiation monitoring system
: 5. Portable survey and counting room equipment
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                                    REV:      58 Page 112 of 164 Table 7 lists all the area, process, and effluent monitors. Table 8 lists the general types of portable survey, count room, airborne monitoring and personnel monitoring equipment.
C. Process Monitors Adequate instrumentation monitoring capability exists to properly access the plant status during all modes of operation, i.e.,
instrumentation is available to the operator to determine plant status, aid in emergency classification determination, and aid in post accident assessment. Table 9 lists available instrumentation, ranges and their indicator locations.
D. Fire Detection The fire detection system consists of various types of detectors/flow devices throughout the main power building and in most of the outbuildings. Ionization, flame and thermal type fire detectors are located throughout safety related structures. Audible alarming is on the Control Room annunciator panel system for actuation or trouble.
The Control Room fire panel system will indicate zone location of the alarm. On receipt of the annunciator panel alarm, the fire panel is checked for location and operator assigned to effected area is called for immediate investigation.
Further details of the fire detection system are given in the plant safety procedures, Section F5 Appendix K, Fire Detection and Protection Systems.
E. Post Accident Liquid Sampling A post-accident liquid sampling system is installed at Prairie Island with associated procedures to provide the capability to obtain the following samples:
: 1. Sample of raw reactor water
: 2. Diluted samples of reactor water (boron, chloride, isotopic analysis, pH, etc.)
: 3. Dissolved gas sample for isotopic analysis (noble gases)
: 4. Dissolved hydrogen sample
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                    REV:    58 Page 113 of 164 The sampling system includes the following exposure reduction equipment:
: 1. Shielded sample lines and shielded drain lines in the Hot Sample Room.
: 2. Shielded sample panel which allows collections and analysis of a reactor coolant sample for hydrogen and isotopic analysis.
: 3. Shielded sample carriers for transporting samples to remote facilities (Hot Cell).
: 4. Remote analysis lab (Hot Cell) located on 695 elevation in the Turbine Building.
: 5. Shielded work area in the Hot Cell with an exhaust hood installed, which discharges through a PAC filter unit.
: 6. Remote counting labs with geometries for counting extremely high level radioactivity samples.
This system allows sample collection and analysis within the radiation exposure guidelines given in NUREG 0578.
F. Containment Air Sampling Following an accident, a containment air sample may be obtained, utilizing the gas analyzer to extract a sample via the Hydrogen Post LOCA System for determination of:
: 1. Hydrogen content
: 2. Isotopic analysis (noble gas)
All sampling will be completed within the exposure guidelines of NUREG 0578.
G. Shield Building Vent Sample The Shield Building Stack Hi-Range Monitor (located in the third floor of the turbine building) extracts a sample from the Shield Building stack and pumps it through a large sample chamber which houses the radiation detector. The hi-range detector reading is in mR/hr and is easily converted to mCi/cc via the applicable calibration curves.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                    EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                      REV:      58 Page 114 of 164 Prior to entering the sample chamber, the sample flow is directed through a particulate filter and a silver zeolite adsorber. The particulate filter and silver zeolite adsorber are manually removed and prepared for analysis in the counting labs.
Silver zeolite adsorbers eliminate the problem of entrapped noble gases on the iodine adsorber allowing a much lower NOTE:          detection sensitivity. In addition, air or N2 may be used to blow out the adsorber to further eliminate the entrapped noble gases.
In instances of monitor failure or offscale readings, procedures are available to allow the dose rate on the sample chambers to be measured using portable survey meters. The release concentration can then be calculated by converting the dose rate to concentration utilizing applicable calibration curves.
H. Containment High Range Area Monitors Two channels of Containment High Range Dome monitors are installed in the containments. Full scale reading on these monitors is 108 R/hr. This allows personnel to estimate the amount of activity in containment available for release and the severity of the accident from the applicable calibration curves.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                  EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                      REV:      58 Page 115 of 164 I. In-Plant Iodine Determination During emergency conditions, it will be necessary for emergency personnel to rapidly and accurately determine or estimate the airborne iodine activity in areas of the plant including all operating centers.
Samples for iodine activity are obtained with portable air samplers (AC and battery operated) and continuous air monitors (CAMs). The iodine is collected on silver zeolite adsorbers.
The use of silver zeolite adsorbers reduces the amount of noble gases entrapped on the adsorber. This reduces the NOTE:          minimum sensitivity level of iodine on the adsorber. In addition, air or N2 may be used to blow out the adsorber to further reduce the amount of entrapped noble gases.
The silver zeolite adsorbers may be analyzed using the GEM system in the onsite counting room or the EOF Counting Room. The adsorbers could also be analyzed with portable instrumentation.
The Control Room, Operational Support Center, Technical Support Center and EOF have continuous air monitors (CAMs) available to monitor the airborne iodine levels. A detector is continuously analyzing the activity (iodine) trapped on the carbon-impregnated filter paper.
This combination of equipment allows iodine determinations under all plant accident conditions.
An Iodine Monitoring program, acceptable to the NRC, was described in letters from L.O. Mayer, NSP, to Director of Nuclear Reactor Regulation, dated December 31, 1979, Lessons Learned Implementation and March 13, 1980, 1/1/80 Lessons Learned Implementation Additional Information.
 
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E-PLAN EP                              EMERGENCY PLAN                                        REV:      58 Page 116 of 164 J. Steam Line Monitors The steam line radiation monitor in conjunction with the ERCS (Emergency Response Computer System) computer will supply a value for noble gas activity released via the steam headers (steam dumps and safeties).
An alternate steam header release calculation procedure exists which allows the determination to be made with portable radiation equipment and applicable calibration curves. This will allow a backup method for release determination during instances of monitor failure.
Normally the air ejector discharge is routed to the Shield NOTE:          Building Exhaust stacks which are monitored by the low and high range stack radiation monitors.
K. Air Ejector Noble Gas Release Releases through the air ejectors are quantified via: (1) the installed air ejector radiation monitor and applicable calibration curves; (2) the Shield Building Exhaust Stack monitors (low and high range) and their applicable calibration curves; or (3) by local sample analysis.
 
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E-PLAN EP                              EMERGENCY PLAN                            REV:      58 Page 117 of 164 Table 6    Seismographic Monitoring Devices Triaxial Accelerometers (1)      Unit 1 Containment Low                                      32.5/210/697.5 (2)      Unit 1 Containment High                                    32.5/210/765.5 (3)      Aux Bldg Ground Floor                                      J.0/9.0/695 (4)      Unit 2 Containment High                                    29/95/765.5 Triaxial Accelerographs (1)      Aux Bldg Ground Floor                                      J.0/9.0/695 (2)      Aux Bldg Spent Fuel Pool                                    N.8/9.0/755 (3)      Aux Bldg Fan Floor                                          J.0/9.0/755 (4)      Unit 1 Containment Low                                      32.5/210/697.5 (5)      Unit 1 Containment High                                    32.5/210/765.5 (6)      Unit 2 Containment High                                    29/95/765.5 (7)      Unit 2 Containment Low                                      29/95/697.5 (8)      Turbine Building Ground Floor                              C.9/8.4/695 (9)      Turbine Building Operating Floor                            C.6/8.8/735 (10)    Screenhouse Low                                            C1.0/81.8/670 (11)    Screenhouse High                                            C1.0/81.8/695 (12)    Screenhouse Cooling Water Piping                            B1.9/91.5/680 (13)    Screenhouse Cooling Water Piping                            C1.5/91.7/692 (14)    Screenhouse Cooing Water Piping                            C1.5/81.3/692 (15)    Aux Bldg Chem & Vol Control Piping                          L.1/7.0/741 (16)    Aux Bldg Aux Feedwater                                      H.7/6.9/709
 
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E-PLAN EP                              EMERGENCY PLAN                              REV:      58 Page 118 of 164 Table 7      Radiation Monitors Area Monitors Radiation Channel No.              Detector/Location                          Instrument Range R-1                      GM/Control Room                            0.1-104 mr/hr 1-R-2                    GM/Containment Vessel Unit 1              0.1-104 mr/hr 2-R-2                    GM/Containment Vessel Unit 2              0.1-104 mr/hr R-3                      GM/Radiochem Lab                          0.1-104 mr/hr R-4                      GM/Charging Pumps Unit 1                  0.1-104 mr/hr R-5                      GM/Spent Fuel Pool                        0.1-104 mr/hr R-6                      GM/Hot Sample Room                        0.1-104 mr/hr 1-R-7                    GM/Incore Seal Table Unit 1                0.1-104 mr/hr 2-R-7                    GM/Incore Seal Table Unit 2                0.1-104 mr/hr R-8                      GM/Waste Gas Valve Gallery                0.1-104 mr/hr 1-R-9                    GM/Letdown HX Unit 1                      0.1-104 mr/hr 2-R-9                    GM/Letdown HX Unit 2                      0.1-104 mr/hr R-28                    Scint/New Fuel Pit                        1.0-105 mr/hr R-29                    Scint/Shipping Receiving                  0.1-104 mr/hr R-32                    Scint/Rad Waste Control Station            0.1-104 mr/hr R-33                    Scint/Rad Waste Bldg/2nd Floor            0.1-104 mr/hr R-36                    Scint/Charging Pumps Unit 2                0.1-104 mr/hr 1-R-48                  Ion Chamber/Containment Vessel Unit 1      1-108 R/hr 1R-49                    Ion Chamber/Containment Vessel Unit 1      1-108 R/hr 2R-48                    Ion Chamber/Containment Vessel Unit 2      1-108 R/hr 2R-49                    Ion Chamber/Containment Vessel Unit 2      1-108 R/hr
 
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E-PLAN EP                                EMERGENCY PLAN                                    REV:      58 Page 119 of 164 Table 7          Radiation Monitors Area Monitors Radiation Channel No.              Detector/Location                                Instrument Range 1-R-53                  SI Pump Area, Unit 1                              0.1-107 mr/hr 2-R-53                  SI Pump Area, Unit 2                              0.1-107 mr/hr 1-R-54                  CS Pump Area, Unit 1                              0.1-107 mr/hr 2-R-54                  CS Pump Area, Unit 2                              0.1-107 mr/hr 1-R-55                  Aux Bldg 695 East Area                            0.1-107 mr/hr 2-R-55                  Aux Bldg 695 West Area                            0.1-107 mr/hr 1-R-56                  Aux Bldg 695 West Area                            0.1-107 mr/hr 2-R-56                  Aux Bldg 695 East Area                            0.1-107 mr/hr 1-R-57                  Aux Bldg 715 East Area                            0.1-107 mr/hr 2-R-57                  Aux Bldg 715 West Area                            0.1-107 mr/hr 1-R-58                  Aux Bldg 715 West Area                            0.1-107 mr/hr 2-R-58                  Aux Bldg 715 East Area                            0.1-107 mr/hr 1-R-59                  Aux Bldg 715 Pent/Ltdn Area                      0.1-107 mr/hr 2-R-59                  Aux Bldg 715 Pent/Ltdn Area                      0.1-107 mr/hr 1-R-60                  Aux Bldg 735 North Area                          0.1-107 mr/hr 2-R-60                  Aux Bldg 735 North Area                          0.1-107 mr/hr 1-R-61                  A Stm Line Area                                  0.1-107 mr/hr 2-R-61                  A Stm Line Area                                  0.1-107 mr/hr 1-R-62                  Aux Bldg 755 East Area                            0.1-107 mr/hr 2-R-62                  Aux Bldg 755 West Area                            0.1-107 mr/hr NOTE:          All Area Monitors on this page have Ion Chamber type detectors.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                  EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                    REV:      58 Page 120 of 164 Table 7          Radiation Monitors Area Monitors Radiation Channel No.              Detector/Location                                Instrument Range 1-R-63                  Aux Bldg 755 West Area                            0.1-107 mr/hr 2-R-63                  Aux Bldg 755 East Area                            0.1-107 mr/hr 1-R-64                  Turb Bldg 735 North Area                          0.1-107 mr/hr 2-R-64                  Turb Bldg 735 North Area                          0.1-107 mr/hr R-65                    Oper Support Center                              0.1-107 mr/hr R-66                    D1 Dsl Gen Room                                  0.1-107 mr/hr R-67                    Inst and Control Shop                            0.1-107 mr/hr R-68                    Tech Support Center Rad                          0.1-107 mr/hr R-69                    Guardhouse                                        0.1-107 mr/hr 2-R-72                  D6 Cable Spreading Room                          0.1-107 mr/hr 2-R-73                  D6 Bus 26 4KV SWGR Room                          0.1-107 mr/hr 2-R-74                  D6 Bus 221 & 222 480V SWGR Room                  0.1-107 mr/hr NOTE:          All Area Monitors on this page have Ion Chamber type detectors.
 
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E-PLAN EP                              EMERGENCY PLAN                              REV:      58 Page 121 of 164 Table 7      Radiation Monitors Process Monitors Radiation Channel No.              Detector/Location                          Instrument Range 1-R-11                  Scint/Containment and Shield Bldg Particulate Unit 1                          101-106 cpm 2-R-11                  Scint/Containment and Shield Bldg Particulate Unit 2                          101-106 cpm 1-R-12                  GM/Containment and Shield Bldg Gas Unit 1                                      101-106 cpm 2-R-12                  GM/Containment and Shield Bldg Gas          101-106 cpm Unit 2 1-R-15                  Scint/Condenser Air Ejector Unit 1          101-106 cpm 2-R-15                  Scint/Condenser Air Ejector Unit 2          101-106 cpm R-16                    Scint/Fan Coils Wtr Disch Unit 1 & Unit 2  101-106 cpm R-18                    Scint/Waste Disposal Liquid Effluent        101-106 cpm 1-R-19                  Scint/Steam Generator Blowdown Unit 1      101-106 cpm 2-R-19                  Scint/Steam Generator Blowdown Unit 2      101-106 cpm R-21                    Scint/Circulating Water Dsch                101-106 cpm 1-R-22                  GM/Shield Bldg Vent Gas Unit 1              101-106 cpm 2-R-22                  GM/Shield Bldg Vent Gas Unit 2              101-106 cpm R-23                    GM/Control Room Vent                        101-106 cpm R-24                    GM/Control Room Vent                        101-106 cpm
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                          EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                              REV:      58 Page 122 of 164 Table 7      Radiation Monitors Process Monitors Radiation Channel No.              Detector/Location                          Instrument Range R-25                    GM/Spent Fuel Pit Vent                    101-106 cpm R-26                    GM/11/21 RHR Cubicle Vent                  101-106 cpm R-27                    GM/12/22 RHR Cubicle Vent                  101-106 cpm 1-R-30                  GM/Aux Bldg Vent Gas Unit 1                101-106 cpm 2-R-30                  GM/Aux Bldg Vent Gas Unit 2                101-106 cpm R-31                    GM/Spent Fuel Pit Vent                    101-106 cpm R-35                    GM/Rad Waste Bldg Vent Gas                101-106 cpm 1-R-37                  GM/Aux Bldg Vent Gas Unit 1                101-106 cpm 2-R-37                  GM/Aux Bldg Vent Gas Unit 2                101-106 cpm R-38                    Scint/Fan Coils Wtr Dsch Unit 1 & Unit 2  101-106 cpm 1-R-39                  Scint/Component Cooling Liquid Unit 1      101-106 cpm 2-R-39                  Scint/Component Cooling Liquid Unit 2      101-106 cpm R-41                    GM/Waste Gas High Level Loop              101-106 cpm 1-R-50                  Ion Chamber/Shield Bldg Vent Gas Unit 1    0.1-107 mr/hr 2-R-50                  Ion Chamber/Shield Bldg Vent Gas Unit 2    0.1-107 mr/hr 1-R-51                  GM/Steam Line Unit 1, Loop A              0.1-105 mr/hr 1-R-52                  GM/Steam Line Unit 1, Loop B              0.1-105 mr/hr 2-R-51                  GM/Steam Line Unit 2, Loop A              0.1-105 mr/hr 2-R-52                  GM/Steam Line Unit 2, Loop B              0.1-105 mr/hr
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                    EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                                        REV:      58 Page 123 of 164 Table 8          Radiation Monitoring Instruments and Devices Portable Survey Instruments Types                                          Range(s)
GMs                                            0-70,000 cpm 0-1000 R/hr Ion Chambers                                    0-50 R/hr Scintillation                                  0-500,000 cpm Tissue Equivalent                              .001 mR/hr-999 R/hr Proportional Counter                            .001 mR-999 R/hr Portable Air Sampling Equipment Types                                          Range(s)
Continuous Air                                  50-50,000 cpm Monitors                                        10-106 cpm Air Samplers                                    2.5-20 CFM 0-80 LPM Analysis Equipment Types Tritium Liquid Scintillation Detection Gamma Spectroscopy Analysis Proportional Alpha/Beta Counters GM Counter Personnel Monitoring Equipment Types                                          Range(s)
Self-Reading Dosimeters (manual)                0-200 mR 0-1 R 0-5 R 0-100 R Self Reading                                    1 mR - 1000 R Dosimeters (electronic)
DLRs                                          All Ranges Finger Rings                                    All Ranges Portal Monitors                                0-30,000 cps Exact quantities and locations are described in the plants NOTE:          surveillance program procedures.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                            EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                              REV:      58 Page 124 of 164 Table 9          Instruments Available for Monitoring Major Systems MEASURED PARAMETER    TYPE OF READOUT    INDICATOR RANGE    INDICATOR LOCATION
: 1. Source Range Neutron Level      Log scale                0 to 106 cps  System cabinets indicator, Recorder                    Main Control Boards Computer output                        ERCS Computer Startup Rate      Linear scale            -0.5 to 5 DPM  Main Control Boards indicator,                              ERCS Computer Computer output Hot Shutdown Panel Neutron Flux      Log Scale              10-1 to 105 cps System cabinets Monitor (N51/N52)  Linear Scale            -1 to 7 DPM    ERCS Computer
: 2. Intermediate Range Neutron Level Log scale                10-11 to 10-3 System cabinets indicator,                    Amp      Main Control Boards Computer output                        ERCS Computer Neutron Level Recorder                10-11 to 10-3  Main Control Board Amp Startup Rate Linear scale            -0.5 to 5 DPM  Main Control Boards indicator                              ERCS Computer
: 3. Power Range Neutron Level      Linear scale          0-120% Full Power System cabinets indicator,                              Main Control Boards Computer output                        ERCS Computer Neutron Level      Recorder              0-120% Full Power Main Control Boards Neutron Flux      Log Scale                10-8 to 100%  System cabinets Monitor (N51/N52)  Linear Scale            -1 to 7 DPM    ERCS computer Linear Scale              10 to 200%
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                  REV:    58 Page 125 of 164 Table 9          Instruments Available for Monitoring Major Systems MEASURED PARAMETER    TYPE OF READOUT        INDICATOR RANGE    INDICATOR LOCATION
: 4. RC System Hot Leg            Linear scale recorder,      50-700&deg;F      Hot Shutdown Panel Temperature        Computer output                            Main Control Boards ERCS Computer Cold Leg          Linear scale recorder        50-700&deg;F      Hot Shutdown Panel Temperature        ERCS Computer                              Main Control Boards ERCS Computer Subcooling        Digital Scale                Variable    ERCS Computer Temperature                                                  Inadequate Core Cooling and pressure                                                  Monitoring Cabinet ERCS SAS Display Avg.              Linear scale                520-620&deg;F      Main Control Boards Temperature        indicator recorder,                        ERCS Computer Computer output                            ERCS SAS Display Temp.              Linear scale                  0-150%      Main Control Boards Difference        indicator, recorder                        ERCS Computer (Delta T)          Computer output Pressure          Linear scale                0-3000 psig    Main Control Boards indicator, recorder                        Hot Shutdown Panel Computer output                            ERCS Computer Low Range          Linear scale                0-750 psig    Main Control Boards Pressure          recorder Flow              Linear scale            0-110% rated flow Main Control Boards indicator,                                ERCS Computer Computer output
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                REV:      58 Page 126 of 164 Table 9          Instruments Available for Monitoring Major Systems MEASURED PARAMETER    TYPE OF READOUT        INDICATOR RANGE  INDICATOR LOCATION
: 5. Pressurizer Level (cold)        Linear scale                  0-100%    Main Control Boards indicator,                                Hot Shutdown Panel Computer output                          ERCS Computer Level              Linear scale                  0-100%    Main Control Boards indicator,                                ERCS Computer Recorder, Computer output, Annunciator Temperature        Linear scale                  0-700&deg;F    Main Control Boards (Vapor              indicator,                                ERCS Computer temperature and    Computer output liquid temperature)
: 6. RWST Level              Linear scale indicator,        0-100%    Main Control Boards Computer output,                          ERCS Computer Annunciator Valve Status        Indicator light                ---------- Main Control Boards
: 7. Steam Generator Narrow Range Level  Linear scale                  0-100%    Main Control Boards indicator, Recorder,                      ERCS Computer Annunciator Wide Range Level Linear scale                  0-100%    Main Control Boards indicator,                                Hot Shutdown Panel Computer output                          ERCS Computer Pressure Linear scale                0-1400 psig  Main Control Boards indicator,                                Hot Shutdown Panel Computer output                          ERCS Computer
: 8. Station Electric Distribution        Linear scale                0-5000 Volts Main Control Boards (Safeguards        indicator, Indicator        0-600 Volts  ERCS Computer AC and DC)          light, Computer output
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                    EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                      REV:      58 Page 127 of 164 Table 9        Instruments Available for Monitoring Major Systems MEASURED PARAMETER  TYPE OF READOUT        INDICATOR RANGE        INDICATOR LOCATION
: 9. Aux. FW Status Flow              Linear scale indicator,        0-250 gpm        Main Control Boards Computer output                                Hot Shutdown Panel ERCS Computer Pressure          Linear scale indicator,        0-2000 psig      Main Control Boards Computer output                                Hot Shutdown Panel ERCS Computer
: 10. Containment Vessel Pressure          Linear scale                    0-60 psia      Main Control Boards indicator,                      0-30 psia      ERCS Computer Computer output, Annunciator, Recorder Post-Accident      Log scale                      1-108R/hr      System cabinets Radiation (R-48)  indicator,                                      ERCS Computer (Containment)      Computer output Water Level        Linear scale                (Sump B) 0-100%    Main Control Boards indicator                (Containment) 0-12 ft Isolation Status  Indicator light,                -------------  Main Control Boards Computer output                                ERCS Computer Temperature        Computer output,                0-400&deg;F        ERCS Computer Air Recirc. Fan    Indicator light,                -------------  Main Control Boards Status            Computer output                                ERCS Computer Air Cooling        Indicator light                -------------  Main Control Boards System Status Air Cooling        Indicator light, Flow Status        Computer output                -------------  Main Control Boards ERCS Computer Spray Pump &      Indicator light, Valve Status      Computer output                -------------  Main Control Boards ERCS Computer
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                  REV:      58 Page 128 of 164 Table 9          Instruments Available for Monitoring Major Systems MEASURED PARAMETER    TYPE OF READOUT        INDICATOR RANGE    INDICATOR LOCATION
: 11. Safety Injection Flow              Linear scale                0-500 gpm    Main Control Boards indicator,                  0-1000 gpm    ERCS Computer Computer output Pump & Valve      Indicator light,            ------------- Main Control Boards Status            Computer output                            ERCS Computer
: 12. Resident Heat Removal Flow (RHR)        Linear scale                0-3000 gpm    Main Control Boards indicator,                  0-6000 gpm    ERCS Computer Computer output Pressure          Linear scale indicator      0-750 psig    Main Control Boards Pump & Valve      Indicator light,              ----------  Main Control Boards Status            Computer output                            ERCS Computer Emerg. Sump        Indicator light,            __________    Main Control Boards Valve Status      Computer output                            ERCS Computer Decay Heat        Linear scale                  50-400&deg;F    Main Control Boards Pump Suction      recorder,                                  ERCS Computer Temperature        Annunciator Decay Heat        Linear scale                  50-400&deg;F    Main Control Boards Cooler Outlet      recorder,                                  ERCS Computer Temperature        Annunciator
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                            EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                REV:    58 Page 129 of 164 Table 9          Instruments Available for Monitoring Major Systems MEASURED PARAMETER    TYPE OF READOUT    INDICATOR RANGE      INDICATOR LOCATION
: 13. Accumulator Accumulator        Linear scale            0-800 psig    Main Control Boards Pressure          indicator,                              ERCS Computer Computer output printout Annunciator Accumulator        Linear scale              0-100%        Main Control Boards Level              indicator,                              ERCS Computer Annunciator Valve Status      Indicator light,          ----------    Main Control Boards Annunciator
: 14. Emergency Ventilation System Fan & Damper      Indicator light,          -----------  Main Control Boards Status            Computer output                        ERCS Computer
: 15. Reactor Vessel    Digital Scale            Variable      Inadequate Core Cooling Level Instrument                                          Monitoring Cabinet System                                                    Computer
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                              EMERGENCY PLAN NUMBER:
E-PLAN EP                                  EMERGENCY PLAN                                REV:    58 Page 130 of 164 Table 9 Instruments Available For Monitoring Major Systems HOT SHUTDOWN PANEL AREA INDICATIONS
: 1. Steam Generator A Level (wide range)                            0-100%
: 2. Steam Generator B Level (wide range)                            0-100%
: 3. Pressurizer (cold) Level                                        0-100%
: 4. Steam Generator A Pressure                                      0-1400 psig
: 5. Steam Generator B Pressure                                      0-1400 psig
: 6. Letdown Valve Status and Control                                Indicating Lights
: 7. Auxiliary Feedwater Control and Status                          Indicating Lights
: 8. Charging Pump Control and Status                                Indicating Lights
: 9. Reactor Coolant System Hot Leg Temperature                      50-700&deg;F
: 10. Reactor Coolant System Cold Leg Temperature                    50-700&deg;F
: 11. Wide Range Reactor Coolant System Pressure                      0-3000 psig
: 12. Pressurizer Heater Control and Status                          Indicating Lights
: 13. Boric Acid Transfer Pump Control and Status                    Indicating Lights
: 14. Neutron Flux Level                                              0.1-105 cps
: 15. Steam Generator PORV                                            Controls
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                  EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                    REV:      58 Page 131 of 164 7.3.2    Facilities and Equipment for Offsite Monitoring A. Meteorological Several locations, exterior to the plant site, can be used to obtain offsite meteorological conditions. Locations and outputs are summarized in Table 10.
B. Assessment Equipment
: 1. The EOF Count Room contains a GEM detector system and Geiger-Mueller counter to analyze offsite samples.
: 2. The emergency lockers in the Assembly Points have the equipment necessary to collect and analyze air samples (particulate and iodine) and portable instruments for measuring radiation levels.
: 3. The hospital emergency kit at Mayo Clinic Health System has instruments for measuring radiation levels and contamination levels of radiation casualties arriving at the medical center for medical treatment.
: 4. All Monticello Nuclear Plant counting room and portable radiation detection equipment is available for analysis of samples from Prairie Island NGP.
: 5. There are TLD badges and airborne particulate and iodine sampling stations installed in areas surrounding the plant. The badges and air sampling stations are installed as part of the Radiation Environmental Monitoring Program. During an emergency, these badges and/or air sampling filters or cartridges may be used for dose assessment purposes.
: 6. All onsite portable equipment and count room equipment at Prairie Island NGP may be used for required offsite radiation surveys or analysis of offsite samples (liquid or airborne).
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                              EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                REV:      58 Page 132 of 164 Table 10          Offsite Meteorological Equipment
: 1. Lock and Dam #3 (a) Temperature (b) Wind Direction (c)  Wind Speed Meteorological information from Lock and Dam #3 is NOTE:          available on a twenty-four hour per day basis.
: 2. National Weather Service Twin Cities Local Area      (a)  Temperature (b)  Wind Direction (c)  Wind Speed (d)  Stability Class
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                    REV:      58 Page 133 of 164 7.4    Protective Facilities and Equipment 7.4.1    Assembly Points The primary protective facility for onsite personnel is the evacuation to an assembly point. Either the Distribution Center or the North Warehouse may be used for an assembly point depending on wind direction. The Emergency Director SHALL designate which one is to be used.
The assembly area emergency locker contains equipment that will be used for personnel contamination checks, personnel decontamination, radiation detection equipment to assess conditions at the assembly area and communication equipment for contact with the Emergency Director.
7.4.2    Operational Support Center The Operational Support Center locker contains all the equipment necessary for reentry into the plant. This includes protective clothing, respiratory protection, monitoring devices, and radiation meters. Air sampling and contamination survey equipment is available for onsite surveys. Decontamination and first aid equipment is available for treatment of onsite personnel.
7.4.3    Emergency Operations Facility The EOF can be designated as an alternate assembly area. Facilities are available for gathering personnel into a specific area. An emergency locker contains equipment necessary for determining personnel contamination and for decontamination of individuals. A decontamination shower and retention system is available for collection of contaminated waste. A spare Field Survey Team Equipment Kit is located at the EOF.
Communication equipment (radio and telephone) is available for contacting emergency personnel both onsite and offsite.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                  REV:      58 Page 134 of 164 7.4.4    Mayo Clinic Health System Mayo Clinic Health System has the equipment required to handle medical emergencies complicated by radioactive contamination. Monitoring equipment, decontamination materials and waste storage (solid and liquid) are available.
7.4.5    Red Wing Fire Station DLR's and Electronic dosimeters (SRD's) will be issued to Emergency responders eg. when they arrive on site before entering the Protected Area.
All dosimeters and DLRs are maintained by plant personnel.
7.4.6    Technical Support Center Emergency Locker The Technical Support Center emergency locker contains the necessary survey instruments, dosimetry and protective clothing to allow reentry or access into the plant during emergency conditions.
7.5    First Aid and Medical Facilities First Aid Kits are available at various emergency lockers in the plant. Any injury requiring medical treatment will be treated at the local medical center. All medical support is covered by Section F4 of the Operations Manual, Medical Support and Casualty Care.
7.6    Damage Control Equipment and Supplies The maintenance area has a completely supplied machine shop with equipment necessary to machine all but the largest pieces of equipment, (e.g., turbine rotors).
One shop area, located in the Auxiliary Building, is for contaminated items. The other shop, located in the Service Building, is for non-contaminated items.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                    REV:      58 Page 135 of 164 8.0    MAINTAINING EMERGENCY PREPAREDNESS 8.1    Organizational Preparedness 8.1.1    Emergency Response Training To achieve and maintain an acceptable level of emergency preparedness, training SHALL be conducted for members of the on-site Emergency Response Organization in accordance with the Prairie Island Nuclear Generating Plant Emergency Plan Training Program.
Training for all on-site Emergency Response Organization members consists of a review of the Emergency Plan in the form of a general overview. In addition to Emergency Plan overview training, personnel assigned key on-site emergency response positions SHALL receive training specific to their position.
Key Emergency Response Organization members SHALL receive Emergency Plan training on an annual basis.
Monticello & Prairie Island offsite support will make provisions for the training of those off-site organizations who may be called upon to provide assistance in the event of an emergency.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                  EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                    REV:      58 Page 136 of 164 8.1.2    Exercises, Drills, and Tests The conduct of periodic drills and exercises are conducted in accordance with the guidance provided in FP-EP-WI-14, Emergency Preparedness Drill and Exercise Manual and FP-EP-WI-24, Emergency Preparedness Drill and Exercise Objectives.
A. Exercises Exercises which test the integrated capability and a major portion of the basic elements existing within the Emergency Plan SHALL be conducted at least every 2 years. This exercise may be included in the full participation biennial exercise which tests the offsite emergency plans.
B. Drills Drills are supervised instructional periods aimed at testing, developing and maintaining skills in a particular operation and are a part of the continuous training program.
In order to ensure that adequate emergency response capabilities are maintained during the interval between biennial exercises, drills SHALL be conducted including at least one drill, during the off exercise year, involving a combination of some of the principal functional areas of the onsite emergency response capabilities. The principal functional areas of emergency response include activities such as management and coordination of emergency response, accident assessment, protective action decision making, and plant system repair and corrective actions. During these drills, activation of all of the Emergency Plans response facilities (TSC, OSC, and EOF) would not be necessary, opportunities to consider accident management strategies would be given, supervised instruction would be permitted, operating staff would have the opportunity to resolve problems (success paths) rather than have controllers intervene, and the drills could focus on onsite training objectives.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                  EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                    REV:      58 Page 137 of 164 Drills SHALL be conducted in the following areas at the designated minimal frequency. Additional drills may be scheduled by plant management if dictated by response of personnel to previous drills.
: 1. Fire Fire drills SHALL be conducted in accordance with Prairie Island Administrative Work Instructions (AWIs) and/or the NSPMs Quality Assurance Topical Report.
: 2. Medical Emergency Medical emergency drills involving the transport of a simulated contaminated individual causing the participation of local support agencies SHALL be conducted annually.
: 3. Radiological The periodic radiological and health physics drills described below may be conducted as part of the annual Radiation Protection Specialist continuing training program in the form of walkthroughs or job performance measured activities. These drills may also be conducted as part of an annual plant wide full scale drill or facility drill.
a      Health Physics Drills which involve response to, and analysis of, simulated elevated airborne and/or liquid samples and direct radiation measurements in the environment SHALL be conducted semi-annually.
b      Radiological monitoring drills which include the collection and analysis of environmental samples for the purpose of ground deposition assessment SHALL be conducted annually.
c      Post accident sampling drills which include the analysis of in-plant liquid samples (with simulated elevated radiation levels) including the use of the Post Accident Sampling System (PASS) SHALL be conducted annually.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                  EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                                      REV:      58 Page 138 of 164
: 4. Security Hostile Action Drills will be conducted to verify readiness to mitigate after a terrorist event. These drills will be conducted in accordance with FP-EP-WI-14 and FP-EP-WI-24.
: 5. Emergency Organization Augmentation Semi-annual Emergency Organization Augmentation Drills are conducted to provide an ongoing verification that the emergency organization can augment the shift organization in a timely fashion.
C. Tests A test is a functional test of equipment to verify that the equipment is operable.
: 1. Communications with state, local and tribal governments within the plume exposure pathway SHALL be tested monthly.
: 2. Communications with Federal response organizations and State governments within the ingestion pathway SHALL be tested quarterly.
: 3. Communications between Prairie Island, Minnesota and Wisconsin Emergency Operating Centers and all local Emergency Operations Centers, and radiation monitoring teams SHALL be checked annually.
: 4. Communication from the Control Room, TSC and EOF to the NRC Operations Center SHALL be tested monthly.
: 5. The Emergency Response Data System (ERDS) SHALL be tested on a quarterly basis.
: 6. The fixed siren portion of Public Alert and Notification System (PANS) SHALL be tested and verified operational on a weekly and monthly basis.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                  EMERGENCY PLAN NUMBER:
E-PLAN EP                                  EMERGENCY PLAN                                    REV:      58 Page 139 of 164 These communication tests SHALL be used not only to check the equipment operation but also that the various NOTE:          phone numbers and links are correct and 2-way communication can be established.
8.2    Review and Updating of the Plan and Procedures The Plant Manager has authority and responsibility for the Prairie Island Emergency Plan and the Emergency Plan Implementing Procedures.
The Plant Manager has the responsibility for the development and updating of the Emergency Plan, the Emergency Plan Implementing Procedures and coordination of the plan with offsite response organizations.
The Emergency Plan will be reviewed on an annual basis to ensure it is current according to the plants controlled procedure program. The update will take into account changes identified during drills and exercises.
Quarterly, all telephone numbers contained in the Emergency Plan Implementing Procedures SHALL be verified correct and updated as a result of the required communication tests.
8.2.1    Organization of Plan The organization of the Emergency Plan is reviewed and updated yearly by the Emergency Preparedness Manager. Reorganization may be necessary as the result of the following:
A. drills or exercises indicating need for changes B. changes in key personnel C. changes in the plants organization structure D. changes in the organization of offsite response agencies E. experience gained under actual emergency situations
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                  REV:      58 Page 140 of 164 8.2.2    Maintenance and Inventory of Emergency Equipment and Supplies Radiation protection equipment at each of the emergency facilities is checked monthly for operability according to surveillance and testing program.
Emergency plan portable radiation instruments SHALL receive a Channel Check and Channel Operational Test monthly and a Channel Calibration at least every 24 months. If any emergency plan portable radiation instrument is found inoperable, then immediate actions SHALL be initiated to restore operability or replacement.
All supplies are inventoried quarterly and dated equipment and material are periodically replaced according to surveillance and testing program.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                  EMERGENCY PLAN NUMBER:
E-PLAN EP                                  EMERGENCY PLAN                                  REV:      58 Page 141 of 164 9.0    RECOVERY In general, the plant will be responsible for the short term recovery, that is recovery from an emergency condition in which no core damage or serious release of radioactivity to the environments has occurred.
If it is clear that a high potential exists for core damage and/or a serious release of radioactivity to the environment, a Recovery Phase will be activated to provide for the long-term recovery actions and for establishing support arrangements.
In general, before re-occupying buildings after an emergency, certain recovery criteria must be satisfied: (1) There must be assurance that the problem encountered is solved and that this same incident cannot immediately recur; (2) The general occupancy areas must be free of significant contamination; (3) Radiation areas and High Radiation areas must be properly defined; and (4) Airborne radioactivity must be eliminated or controlled.
9.1      Investigation of Incidents All incidents SHALL be investigated in conjunction with corporate event response procedures.
9.2      Recovery Procedures All recovery operations SHALL be performed in accordance with written procedures. These procedures SHALL include the following activities:
A. Investigation of the cause of the incident B. Investigation of plant conditions following an accident C. Repair and restoration of facilities D. Testing and startup of restored facilities Methods for determining the extent of radioactive contamination and general protective measures to be taken for personnel performing recovery operations are established in Section F2, Radiation Safety, of the Operations Manual, and in the Radiation Protection Manual, RPIPs.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                  EMERGENCY PLAN NUMBER:
E-PLAN EP                                  EMERGENCY PLAN                                  REV:      58 Page 142 of 164 Written procedures for recovery of the facility from the specific post accident conditions will be prepared by qualified plant staff members and submitted to the Plant Operating Review Committee. Plant Operating Review Committee approval of all such procedures is required prior to their initiation.
9.3    Criteria for Resumption of Operations If the plant is shutdown as the result of an emergency, it will be restarted only when:
A. The conditions which caused the emergency are corrected.
B. The cause of the emergency is understood.
C. Restoration, repair and testing is completed as required.
D. No unreviewed safety questions exist.
E. All conditions of the license and technical specifications are satisfied.
9.4    Transition to Recovery If it is clear that extensive plant damage exists and contamination of plant systems have occurred, then a recovery phase may be necessary.
Transition to the recovery phase will take place in an incremental manner as the functions change from operational to engineering/construction. The decision to make the transition from the emergency phase to the recovery phase should be a joint decision by the ED and EM. The Recovery Manager should possess the qualifications of an Emergency Manager. This position should be occupied by personnel representing the company executive level.
Should transition to the recovery phase become necessary, the site engineering/construction staff would provide the nucleus of the organization.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                  REV:      58 Page 143 of 164 This plant staff would be augmented as required by specialists from the site organization and the offsite support groups. In addition, appropriate assistance would be secured from the Architect-Engineer and the NSSS vendor organizations.
This support could be broadened as required by consultant help from the several organizations familiar with Prairie Island NGPs organization. The overall organization envisioned for a substantial Recovery Phase would be a blend of site staff, and appropriate vendor and consultant personnel. On a prior basis it is counterproductive to define in detail the extensive organization that might be involved in a sizable Recovery Phase because of the unlimited variation of conditions that could result from plant emergencies. However, the nucleus organization has been identified together with guidelines on how the organization might be expanded to meet the requirements demanded at the time.
When the Emergency Manager and Emergency Director agree that the onsite emergency condition has been terminated, a complete transfer of the responsibilities for offsite support may be made to the Recovery Organization. The EOF would then become the Recovery Center and function as Command Center for the Recovery Organization and the recovery effort. Details of Recovery Organization activation and implementing criteria are contained in the Emergency Plan Implementing Procedures.
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                              EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                  REV:      58 Page 144 of 164 THIS PAGE IS INTENTIONALLY LEFT BLANK
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                              EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                  REV:      58 Page 145 of 164 Attachment A          Emergency Plan Implementing Procedures A.1    PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES The following is a listing of F3 procedures (Emergency Plan Implementing Procedures) which SHALL be used by plant emergency organization personnel to implement the emergency plan. This may not be a complete detailed procedure list but is meant to serve as a basis for procedure development.
Procedure Number                                Procedure Title                    Affected Plan Section F3-1            Onsite Emergency Organization                          5.3, 5.4 F3-2            Classifications of Emergencies                          4.0 F3-3            Responsibilities During a Notification of Unusual Event 6.1, 6.3, 6.7 F3-4            Responsibilities During an Alert, Site Area            6.1, 6.2, 6.3, 6.4, 6.5, 6.6, or General Emergency                                    6.7 F3-5            Emergency Notifications                                5.3.3(D), 6.1.2, 6.1.3, 6.1.4, 7.2 F3-5.1          Switchboard Operator Duties                            6.1.2, 6.1.3, 6.1.4 F3-5.2          Response to False Siren Activation                      6.1.2 F3-5.3          Deleted F3-6            Activation and Operation of Technical Support Center    5.3, 7.1.1 F3-7            Activation and Operation of Operational Support        5.3, 7.1.2 Center F3-8            Recommendations for Offsite Protective Actions          6.4, 6.6 F3-8.1          Deleted F3-9            Emergency Evacuation                                    6.6.1, 7.4.1
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                            EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                                REV:      58 Page 146 of 164 Attachment A Emergency Plan Implementing Procedures Procedure Number          Procedure Title                                      Affected Plan Section F3-10          Personnel Accountability                              6.6.1 F3-11          Search and Rescue                                    6.6.1, 6.7.1 F3-12          Emergency Exposure Control                            6.6.1, 6.7.1 F3-13          Offsite Dose Calculations                            6.4, 6.6 F3-13.0        Deleted F3-13.1        Rad & Met Data for Dose Projections                  6.4, 6.6, 7.3 F3-13.2        Deleted F3-13.3        Manual Dose Calculations                              6.4, 6.6 F3-13.4        Deleted F3-13.5        Alternate Meteorological Data                        6.4, 6.6, 7.3 F3-13.6        Weather Forecasting Information                      6.4, 6.6, 7.3 F3-14.1        Onsite Radiological Monitoring                        6.4, 7.3 F3-14.2        Deleted F3-15          Responsibilities of the Radiation Survey Teams During 6.4.2 a Radioactive Airborne Release F3-16          Responsibilities of the Radiation Survey Teams During 6.4.2 a Radioactive Liquid Release F3-17          Core Damage Assessment                                6.4.1 F3-17.1        Core Damage Determination F3-17.2        Long Term Cooling F3-18          Thyroid Iodine Blocking Agent (Potassium Iodide)      6.6.2
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                          EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                              REV:      58 Page 147 of 164 Attachment A Emergency Plan Implementing Procedures Procedure Number          Procedure Title                                    Affected Plan Section F3-19          Personnel and Equipment Monitoring and              6.6.3, 6.7.2 Decontamination F3-20          Determination of Radioactive Release Concentrations 6.4, 7.3 F3-20.1        Determination of Steam Line Dose Rates              6.4, 7.3 F3-20.2        Determination of Vent Stack Dose Rates              6.4, 7.3 F3-21          Establishment of a Secondary Access Control Point  6.6.3 F3-22          Deleted F3-23          Emergency Sampling                                  7.3 F3-23.1        Emergency Hotcell Procedure                        7.3 F3-23.2        Deleted F3-23.3        Deleted F3-23.4        Deleted F3-23.5        Deleted F3-23.6        Deleted F3-23.7        Deleted F3-23.8        Deleted F3-24          Record Keeping During an Emergency                  6.2 F3-25          Re-Entry                                            7.3, 9.0 F3-26          Deleted
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                    EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                        REV:      58 Page 148 of 164 Attachment A Emergency Plan Implementing Procedures Procedure Number          Procedure Title                              Affected Plan Section F3-26.1        Operation of the ERCS Display                7.1, 7.3 F3-26.2        Radiation Monitor Data on ERCS                7.1, 7.3 F3-26.3        ERDS - NRC Data Link                          6.4, 7.2.2.E F3-27          Response to Railroad Grade Crossing Blockage  5.6.4(D)
F3-28          Deleted F3-29          Emergency Security Procedures                5.3, 6.1 F3-30          Transition to Recovery                        9.0 F3-31          Response to Security Related Threats          5.3, 6.1
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                            EMERGENCY PLAN NUMBER:
E-PLAN EP                                EMERGENCY PLAN                                REV:      58 Page 149 of 164 Attachment A Emergency Plan Implementing Procedures A.2    EOF EMERGENCY PLAN IMPLEMENTING PROCEDURES The following is a listing of F8 procedures (EOF Emergency Plan Implementing Procedures) which SHALL be used by EOF emergency organization personnel to implement the emergency. This may not be a complete detailed procedure list but is meant to serve as a basis for procedure development.
Procedure Number                                Procedure Title                  Affected Plan Section F8-1            Emergency Operations Facility Organization            5.4, 5.5, 5.6 F8-2            Deleted F8-3            Activation and Operation of the EOF                    5.4, 7.1.3 F8-4            Emergency Support and Logistics                        5.6, 5.7 F8-5            Offsite Dose Assessment and Protective                6.4, 6.6 Action Recommendations F8-6            Radiological Monitoring and Control at                6.6, 6.7, 7.4 the EOF F8-8            Offsite Agency Liaison Activities                      5.6, 5.7 F8-9            Event Termination or Recovery                          5.5, 9.0 F8-10          Record Keeping in the EOF                              6.2 F8-11          Transfer to the Backup EOF                            7.1.3 F8-12          Emergency REMP                                        7.3.2
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                          EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                              REV:      58 Page 150 of 164 Attachment B      Summary of Emergency Supplies
: 1. Old Receiving Warehouse Locker and Emergency Vehicle Field Survey Kits A. Beta Gamma Survey Meters B. Offsite Sample Kits (2)
: 1. Airborne Sample Equipment (Particulate, Iodine, Gaseous)
: 2. Liquid Sample Equipment C. Personnel Dosimetry D. Portable Communication Radios E. Foul weather gear F. Protective clothing G. Potassium Iodide Potassium Iodide (KI)
: 2. North Warehouse and Distribution Center Assembly Points (each location)
A. Beta Gamma Survey Meters B. Portable Communications Radio C. 1 Copy of Emergency Plan Implementing Procedures (F-3)
D. Personnel Decontamination Kit E. Airborne Sample Equipment F. Small First Aid Kit G. Area Radiation Monitor H. Protective Clothing
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                      EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                          REV:      58 Page 151 of 164 Attachment B Summary of Emergency Supplies
: 3. Operational Support Center A. Beta Gamma Survey Meters B. Air Sampling Equipment (Battery and AC Powered)
C. Personnel Dosimetry D. Portable Communications Radios (located in Control Room)
E. Area Radiation Monitor F. Portable Lanterns and Batteries G. Copies of Emergency Plan Implementing Procedures (F-3)
H. Plant Floor Plans I. Protective Clothing (Including Waterproof)
J. Respiratory Protection (SCBAs) and spare bottles K. First Aid Kit L. Continuous Air Monitor (Control Room and OSC)
M. Drager Toxic Chemical Air Sampler N. Full Face Respirators and Iodine Canisters
: 4. Technical Support Center A. Beta Gamma Survey Meters B. Airborne Sampling Equipment C. Personnel dosimetry D. Portable Communications Radio E. Area Radiation Monitor
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                            EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                                  REV:    58 Page 152 of 164 Attachment B Summary of Emergency Supplies F. Continuous Air Monitor G. Copies of Emergency Plan Implementing Procedures (F-3)
H. Protective Clothing I. Respiratory Protection (SCBAs) and spare bottles J. Plant Floor Plans K. Potassium Iodide (KI) Distribution
: 5. Mayo Clinic Health System A. Beta Gamma Survey Meters B. Personnel dosimetry C. Copy of Operations Manual F-4 D. Supplies (Disposable clothing, solid waste containers, and liquid waste containers)
: 6. Hot Cell A. Beta Gamma Survey Meters B. Protective clothing C. Alpha Survey Meter D. Sample team communication gear E. Copy of Emergency Plan Implementing Procedure (F-3)
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                        EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                            REV:      58 Page 153 of 164 Attachment B Summary of Emergency Supplies
: 7. Fire Brigade Dress Out Area A. Self-Contained Breathing Apparatus
: 8. Emergency Operations Facility A. Beta Gamma Survey Meters B. Offsite Sample Kit (1)
: 1. Airborne Sampling Equipment (Particulate, Iodine, Gaseous)
: 2. Liquid Sampling Equipment C. Personnel Dosimetry D. Airborne Sampling Equipment (Local)
E. Portable Communication Radios F. Personnel Decontamination Kit G. Area Radiation Monitor H. Continuous Air Monitor I. GEM detector for Isotopic Analysis of Samples J. Decontamination Shower K. Potassium Iodide (KI) Distribution
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                            EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                                REV:      58 Page 154 of 164 Attachment C          NUREG-0654/PI E-Plan Cross Reference NUREG-0654 ELEMENT                        PI E-PLAN REFERENCE (by Section)
A. Assignment of Responsibility (Organization Control)
A.1.a                                5.6, 5.7 A.1.b                                5.6, 5.7 A.1.c                                Figure 3 A.1.d                                1.5, 1.6, 5.3.1, MT & PI Offsite Plan A.1.e                                5.2, 5.3, 5.3.1, 5.3.3, Table 1 A.2.a                                State/Local Plans A.2.b                                State/Local Plans A.3                                  MT & PI Offsite Plan A.4                                  5.3.1 B. Onsite Emergency Organization B.1                                  5.1, 5.2, 5.3, 5.4 B.2                                  5.3.1 B.3                                  5.3.1 B.4                                  5.3.1 B.5                                  5.3.2, 5.3.3, 5.3.4, Table 1, 5.4 B.6                                  5.6, 5.7, Figure 4 B.7                                  Table 1, 5.4, MT & PI Offsite Plan B.7.a                                5.4, MT & PI Offsite Plan B.7.b                                5.4, 5.5 B.7.c                                5.4, MT & PI Offsite Plan B.7.d                                MT & PI Offsite Plan B.8                                  5.6.3, MT & PI Offsite Plan B.9                                  5.6.4, 5.7, 6.7.3, MT & PI Offsite Plan
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                            EMERGENCY PLAN NUMBER:
E-PLAN EP                              EMERGENCY PLAN                                REV:      58 Page 155 of 164 Attachment C        NUREG-0654/PI E-Plan Cross Reference NUREG-0654 ELEMENT                        PI E-PLAN REFERENCE (by Section)
C. Emergency Response Support and Resources C.1.a                                5.3.1 C.1.b                                MT & PI Offsite Plan C.1.c                                MT & PI Offsite Plan C.2.a                                State/Local Plan C.2.b                                MT & PI Offsite Plan C.3                                  Attach A.2, (F8-4), MT & PI Offsite Plan C.4                                  Attach A.2 (F8-4), 5.6, MT & PI Offsite Plan D. Emergency Classification System D.1                                  4.0 D.2                                  4.0, Annex A D.3                                  State/Local Plan D.4                                  State/Local Plan E. Notification Methods and Procedures E.1                                  6.1.2, 6.1.4 E.2                                  6.1.1 E.3                                  5.3.1 E.4                                  5.3.1 E.4.a                                5.3.1 E.4.b                                5.3.1 E.4.c                                5.3.1 E.4.d                                5.3.1 E.4.e                                5.3.1
 
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E. Notification Methods and Procedures (contd)
E.4.f                                5.3.1 E.4.g                                5.3.1 E.4.h                                5.3.1 E.4.i                                5.3.1 E.4.j                                5.3.1 E.4.k                                5.3.1 E.4.l                                5.3.1 E.4.m                                5.3.1 E.4.n                                5.3.1 E.5                                  7.2.3, State/Local Plan E.6                                  5.3.1, 7.2.3, State/Local Plan E.7                                  MT & PI Offsite Plan F. Emergency Communications F.1.a                                6.1.2, 6.1.3, Table 6 F.1.b                                6.1.2, 6.1.3, 7.2.2, Table 6 F.1.c                                Table 6 F.1.d                                72.2., Table 6 F.1.e                                6.1.1, 7.2.2, Table 6 F.1.f                                7.2.2, Table 6 F.2                                  7.2.2 F.3                                  8.1.2
 
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G. Public Education and Information G.1                                  MT & PI Offsite Plan G.2                                  MT & PI Offsite Plan G.3.a                                MT & PI Offsite Plan G.3.b                                Attach A (F8-8)
MT & PI Offsite Plan G.4.a                                MT & PI Offsite Plan G.4.b                                MT & PI Offsite Plan G.4.c                                MT & PI Offsite Plan G.5                                  MT & PI Offsite Plan H. Emergency Facilities and Equipment H.1                                  7.1.1, 7.1.2 H.2                                  7.1.3 H.3                                  State/Local Plan H.4                                  6.1.1, MT & PI Offsite Plan H.5                                  7.3 H.5.a                                7.3.1, Table 7, Table 11 H.5.b                                7.3.1, Table 8, Table 9 H.5.c                                7.3.1, Table 10 H.5.d                                7.3.1 H.6.a                                7.3.2, Table 11 H.6.b                                7.3.2, MT & PI Offsite Plan H.6.c                                7.3.2, MT & PI Offsite Plan
 
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H. Emergency Facilities and Equipment (contd)
H.7                                  7.3.2, 7.4.3, 7.4.4, 7.4.5, 7.4.6, Attach B.
H.8                                  6.4, 7 .3.1, 7.3.2, Table 11 H.9                                  7.1.2, 7.4.2, Attach B H.10                                8.2.2 H.11                                Attach B H.12                                7.1.3 I. Accident Assessment I.1                                  Annex A, Table 7 Table 8, Table 10 I.2                                  6.4, 7.3.1 I.3.a                                6.4 I.3.b                                6.4.1 I.4                                  6.4.1 I.5                                  6. 4.1, 7.3.1, 7.3.2 Table 11 I.6                                  6.4.1 I.7                                  6.4.2 I.8                                  6.4.1, 6.4.2 I.9                                  6.4.2 I.10                                6.4.1 I.11                                State Plan
 
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J. Protective Response J.1.a                                6.6.1 (A)
J.1.b                                6.6.1 (A)
J.1.c                                6.6.1 (A)
J.1.d                                6.6.1 (A)
J.2                                  6.6.1 (A)
J.3                                  6.6.1 (A)
J.4                                  6.6.1 (A)
J.5                                  5.3.1, 6.6.1 (A)
J.6.a                                6.6.2 (A)
J.6.b                                6.6.2 (A)
J.6.c                                6.6.2 (B)
J.7                                  6.6.2 I Tables 3 and 4 J.8                                  6.6.1 (D), MT & PI Offsite Plan J.9                                  State/Local Plan J.10.a                              6.4.2, MT & PI Offsite Plan J.10.b                              6.6, MT & PI Offsite Plan J.10.c                              7.2.3 J.10.d                              State/Local Plan J.10.e                              State/Local Plan J.10.f                              State/Local Plan J.10.g                              State/Local Plan J.10.h                              State/Local Plan J.10.i                              State/Local Plan
 
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J. Protective Response [Contd]
J.10.j                              State/Local Plan J.10.k                              State/Local Plan J.10.l                              State/Local Plan J.10.m                              6 .6.1 I Tables 3 and 4 J.11                                State Plan J.12                                State/Local Plan K. Radiological Exposure Control K.1.a                                6.7.1, Table 4 K.1.b                                6.7.1, Table 4 K.1.c                                6.7.1, Table 4 K.1.d                                6.7.1, Table 4 K.1.e                                6.7.1, Table 4 K.1.f                                6.7.1, Table 4 K.1.g                                6.7.1, Table 4 K.2                                  6.7.1 K.3.a                                6.7.1 K.3.b                                6.7.1 K.4                                  State/Local Plan K.5.a                                6.6. 3, Table 5 K.5.b                                6.7.2 K.6.a                                6.6.3 (A)
K.6.b                                6.6.3 (A)
K.6.c                                6.6.3 (A)
K.7                                  6.6.3, 6.7.2
 
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L. Medical and Public Health Support L.1                                  5.6.4 (B), 6.7.3 L.2                                  6.7.2 L.3                                  State Plan L.4                                  5.6.4 I, 6.7.2, 6.7.3 M. Recovery and Re-entry Planning and Post Accident Operations M.1                                  5.5, 9.0 M.2                                  5.5, 9.4 M.3                                  5.5,9.4 M.4                                  6.4.1 N. Exercises and Drills N.1.a                                8.1.2 N.1.b                                8.1.2 N.2.a                                8.1.2 N.2.b                                8.1.2 N.2.c                                8.1.2 N.2.d                                8.1.2 N.2.e (1)                            8.1.2 N.2.e (2)                            8.1.2 N.3.a                                Site Drill/Exercise Manual N.3.b                                Site Drill/Exercise Manual N.3.c                                Site Drill/Exercise Manual N.3.d                                Site Drill/Exercise Manual N.3.e                                Site Drill/Exercise Manual
 
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N. Exercises and Drills (contd)
N.3.f                                Site Drill/Exercise Manual N.4                                  Site Drill/Exercise Manual N.5                                  Site Drill/Exercise Manual O. Radiological Emergency Response Training O.1                                  8.1.1 O.1.a                                8.1.1 O.1.b                                State/Local Plan O.2                                  8.1.1, 8.1.2 Site Drill/Exercise Manual O.3                                  8.1.1 O.4.a                                8.1.1 O.4.b                                8.1.1 O.4.c                                8.1.1 O.4.d                                8.1.1 O.4.e                                8.1.1 O.4.f                                8.1.1 O.4.g                                MT & PI Offsite Plan O.4.h                                8.1.1, MT & PI Offsite Plan O.4.i                                MT & PI Offsite Plan O.4.j                                8.1.1 O.5                                  8.1.1, MT & PI Offsite Plan
 
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P. Responsibility for the Planning Effort: Development, Periodic Review and Distribution of Emergency Plans P.1                                  MT & PI Offsite Plan P.2                                  8.2 P.3                                  8.2 P.4                                  8.2.1 P.5                                  MT & PI Offsite Plan P.6                                  2.0, MT & PI Offsite Plan P.7                                  Attach A P.8                                  Table of Contents Attach C P.9                                  MT & PI Offsite Plan P.10                                8.2
 
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EP                                  ANNEX A REV:
E-PLAN 58 EMERGENCY ACTION LEVEL MATRIX The site-specific Emergency Action Levels (EALs) are presented in the attached Emergency Action Level Matrix. These EALs are based on the NEI 99-01 EAL scheme.
Emergency Plan Implementing Procedure Classification of Emergency also contains the same Emergency Action Level Matrix.
Page 1 of 1
 
Prairie Island Nuclear Generating Plant                                                                                                                                                                          EMERGENCY ACTION LEVEL MATRIX GENERAL EMERGENCY                                                                          SITE AREA EMERGENCY                                                                                                    ALERT                                                                                                NUE                                                  HOT & COLD RG1      Release of gaseous radioactivity resulting in offsite dose                            RS1      Release of gaseous radioactivity resulting in offsite dose                            RA1      Release of gaseous or liquid radioactivity resulting in offsite                        RU1      Release of gaseous or liquid radioactivity greater than 2 times greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE.                                        greater than 100 mrem TEDE or 500 mrem thyroid CDE.                                            dose greater than 10 mrem TEDE or 50 mrem thyroid CDE.                                            the ODCM limits for 60 minutes or longer.
Notes:
* The Emergency Director should declare the General Emergency promptly                Notes:
* The Emergency Director should declare the Site Area Emergency promptly              Notes:
* The Emergency Director should declare the Alert promptly upon determining            Notes:
* The Emergency Director should declare the Unusual Event promptly upon upon determining that the applicable time has been exceeded, or will likely be                  upon determining that the applicable time has been exceeded, or will likely be                  that the applicable time has been exceeded, or will likely be exceeded.                          determining that 60 minutes has been exceeded, or will likely be exceeded.
exceeded.                                                                                      exceeded.
* If an ongoing release is detected and the release start time is unknown,
* If an ongoing release is detected and the release start time is unknown,
* If an ongoing release is detected and the release start time is unknown,
* If an ongoing release is detected and the release start time is unknown,                      assume that the release duration has exceeded 15 minutes.                                        assume that the release duration has exceeded 60 minutes.
assume that the release duration has exceeded 15 minutes.                                      assume that the release duration has exceeded 15 minutes.
* If the effluent flow past an effluent monitor is known to have stopped due to
* If the effluent flow past an effluent monitor is known to have stopped due to
* If the effluent flow past an effluent monitor is known to have stopped due to
* If the effluent flow past an effluent monitor is known to have stopped due to                actions to isolate the release path, then the effluent monitor reading is no longer              actions to isolate the release path, then the effluent monitor reading is no longer actions to isolate the release path, then the effluent monitor reading is no longer            actions to isolate the release path, then the effluent monitor reading is no longer            valid for classification purposes.                                                                valid for classification purposes.
valid for classification purposes.                                                              valid for classification purposes.
* The pre-calculated effluent monitor values presented in EAL RA1.1 should be
* The pre-calculated effluent monitor values presented in EAL RG1.1 should be
* The pre-calculated effluent monitor values presented in EAL RS1.1 should be                  used for emergency classification assessments until the results from a dose RU1.1        1          2        3        4          5        6        DEF used for emergency classification assessments until the results from a dose                    used for emergency classification assessments until the results from a dose                    assessment using actual meteorology are available.
assessment using actual meteorology are available.                                              assessment using actual meteorology are available.                                                                                                                                      Reading on ANY effluent radiation monitor in the table below greater than the listed values Effluents / Release Rates                                                                                                                                                                                                                                                                                                                                                                                                                                                      Effluents / Release Rates for 60 minutes or longer:
Gaseous Effluents 1R-22 Unit 1 Shield Building Vent Rad Monitor            1 x 104 cpm RA1.1        1        2        3        4        5          6        DEF                                  Unit 1  1R-30 Unit 1 Aux Building Vent Rad Monitor              1 x 103 cpm Reading on EITHER of the following radiation monitors greater than the reading shown for                              1R-37 Unit 1 Aux Building Vent Rad Monitor              1.5 x 103 cpm 15 minutes or longer.                                                                                                  2R-22 Unit 2 Shield Building Vent Rad Monitor            1 x 104 cpm RG1.1        1        2        3        4        5          6        DEF                    RS1.1        1        2        3        4        5          6        DEF                                          1R-50 High Range Stack Gas Monitor                  450 mR/hr                          Unit 2  2R-30 Unit 2 Aux Building Vent Rad Monitor              6 x 103 cpm Reading on EITHER of the following radiation monitors greater than the reading shown for        Reading on EITHER of the following radiation monitors greater than the reading shown for                              2R-50 High Range Stack Gas Monitor                  450 mR/hr                                  2R-37 Unit 2 Aux Building Vent Rad Monitor              2 x 103 cpm 15 minutes or longer:                                                                          15 minutes or longer:                                                                                                                                                                                                  R-25 Spent Fuel Pool Vent Rad Monitor                    1 x 104 cpm 1R-50 High Range Stack Gas Monitor                45,000 mR/hr                                1R-50 High Range Stack Gas Monitor                  4,500 mR/hr          RA1.2        1        2          3          4        5      6        DEF                                              R-31 Spent Fuel Pool Vent Rad Monitor                    1 x 104 cpm 2R-50 High Range Stack Gas Monitor                45,000 mR/hr                                2R-50 High Range Stack Gas Monitor                  4,500 mR/hr          Dose assessment using actual meteorology indicates doses greater than 10 mrem TEDE                                    R-35 Radwaste Building Vent Rad Monitor                  3 x 103 cpm or 50 mrem thyroid CDE at or beyond the site boundary.                                                                                        Liquid Effluents Abnormal Rad Release / Rad Effluent                                                                                                                                                                                                                                                                                                                                                                                                                                                      Abnormal Rad Release / Rad Effluent Unit 1  1R-19 SG Blowdown Radiation Monitor                      2 x 104 cpm RG1.2      1        2        3          4        5        6        DEF                      RS1.2        1        2          3        4        5        6        DEF                      RA1.3          1          2        3          4        5          6        DEF                              Unit 2  2R-19 SG Blowdown Radiation Monitor                      4 x 104 cpm Dose assessment using actual meteorology indicates doses greater than 1,000 mrem                Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE            Analysis of a liquid effluent sample indicates a concentration or release rate that would                              R-18 Waste Effluent Liquid Monitor                      5 x 104 cpm TEDE or 5,000 mrem thyroid CDE at or beyond the site boundary.                                  or 500 mrem thyroid CDE at or beyond the site boundary.                                        result in doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the site                                R-21 Circ Water Discharge Monitor                        2 x 104 cpm boundary for one hour of exposure.
RG1.3          1          2        3        4          5        6        DEF                RS1.3          1          2        3        4          5        6        DEF                RA1.4          1          2        3        4          5        6        DEF                  RU1.2        1          2          3        4          5        6        DEF Field survey results indicate EITHER of the following at or beyond the site boundary:          Field survey results indicate EITHER of the following at or beyond the site boundary:          Field survey results indicate EITHER of the following at or beyond the site boundary:            Reading on ANY effluent radiation monitor greater than 2 times the calculated limit
* Closed window dose rates greater than 1,000 mR/hr expected to continue for 60
* Closed window dose rates greater than 100 mR/hr expected to continue for 60
* Closed window dose rates greater than 10 mR/hr expected to continue for 60              established by a current radioactivity discharge permit for 60 minutes or longer.
minutes or longer.                                                                              minutes or longer.                                                                              minutes or longer.
* Analyses of field survey samples indicate thyroid CDE greater than 5,000 mrem
* Analyses of field survey samples indicate thyroid CDE greater than 500 mrem
* Analyses of field survey samples indicate thyroid CDE greater than 50 mrem for          RU1.3        1          2        3          4        5        6        DEF for one hour of inhalation.                                                                    for one hour of inhalation.                                                                    one hour of inhalation.                                                                Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times the ODCM limits for 60 minutes or longer.
RG2      Spent fuel pool level cannot be restored to at least 729.16                          RS2        Spent fuel pool level at 729.16 elevation.                                          RA2      Significant lowering of water level above, or damage to,                                RU2      UNPLANNED loss of water level above irradiated fuel.
elevation for 60 minutes or longer.                                                                                                                                                            irradiated fuel.
Note:    The Emergency Director should declare the General Emergency promptly upon            RS2.1        1          2          3          4        5          6        DEF                RA2.1        1          2          3        4          5          6      DEF                  RU2.1        1          2        3        4          5        6      DEF determining that 60 minutes has been exceeded, or will likely be exceeded.            Lowering of spent fuel pool level to 729.16 elevation.                                        Uncovery of irradiated fuel in the REFUELING PATHWAY.                                            a.        UNPLANNED water level drop in the REFUELING PATHWAY as indicated by RA2.2        1          2          3        4          5          6      DEF                            ANY of the following:
Spent Fuel Pool Level / In Plant Rad                                                                                                                                                                                                                                                                                                                                                                                                                                            Spent Fuel Pool Level / In Plant Rad RG2.1        1          2        3          4          5          6        DEF                                                                                                                Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by
* Level less than SFP low water level alarm (752.5 feet elevation)
Spent fuel pool level cannot be restored to at least 729.16 elevation for 60 minutes or                                                                                                        ANY of the following radiation monitors:
* Refueling Canal Level (365" on Control Board) longer.                                                                                                                                                                                                  1(2)R-2 Containment Vessel Area Monitor                              1 x 10-1 R/hr
* Visual Observation (752.5 feet elevation) 1(2)R-12 Containment/SBV Radio Gas Monitor                            9.0E+2 cpm        AND R-5 Fuel Handling Area Monitor                                        1 x 10 -1 R/hr    b.        UNPLANNED rise in area radiation levels as indicated by 5                    ANY of the following radiation monitors:
R-25 Spent Fuel Pool Vent Rad Monitor                                2 x 10 cpm R-29 Shipping and Receiving Area Monitor                            1 x 102 mR/hr
* R-5 Fuel Handling Area Monitor 5
R-31 Spent Fuel Pool Vent Rad Monitor                                2 x 10 cpm
* 1(2)R-2 Containment Vessel Area Monitor
* Other Portable Area Radiation Monitoring Instrumentation RA2.3        1        2          3        4        5 6 DEF Lowering of spent fuel pool level to 739.16 elevation.
RA3      Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.
Note:    If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted.
RA3.1        1        2        3        4          5        6      DEF Dose rate greater than 15 mR/hr in EITHER of the following areas:
* Control Room (R-1)
* Central Alarm Station (R-69)
RA3.2                            3        4          5 An UNPLANNED event results in radiation levels that prohibit or impede access to any of the Table H1 plant rooms or areas.
PINGP 1576, Rev. 11 Doc. Type/Sub Type: EP/EVT Retention: Lifetime +                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                  1 of 9
 
Prairie Island Nuclear Generating Plant                                                                                                                                                                    EMERGENCY ACTION LEVEL MATRIX GENERAL EMERGENCY                                                                  SITE AREA EMERGENCY                                                                                  ALERT                                                                                            NUE                                                      HOT & COLD HG1    Release ofACTION HOSTILE    gaseous  radioactivity resulting      resulting in loss        in offsite of physical controldose of the                HS1      Release ofACTION HOSTILE    gaseous  radioactivity within          resulting AREA.
the PROTECTED      in offsite dose        HA1      Release ofACTION HOSTILE    gaseous  or liquid within      radioactivity the OWNER        resulting inAREA CONTROLLED        offsite or                HU1      Release of SECURITY Confirmed  gaseous orCONDITION liquid radioactivity greater than 2 times or threat.
greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE.
facility.                                                                                greater than 100 mrem TEDE or 500 mrem thyroid CDE.                          dose greater airborne    than attack    10 mrem threat withinTEDE  or 50 mrem thyroid CDE.
30 minutes.                                                  the ODCM limits for 60 minutes or longer.
HG1.1      1        2        3          4        5      6      DEF                  HS1.1        1          2          3    4        5        6    DEF        HA1.1      1        2          3        4      5        6    DEF                        HU1.1          1        2  3          4        5      6      DEF
: a.      A HOSTILE ACTION is occurring or has occurred within the PROTECTED              A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as  A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED                    A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by AREA as reported by Security Shift Supervisor.                                  reported by Security Shift Supervisor.                                      AREA as reported by Security Shift Supervisor.                                              Security Shift Supervisor.
Security                                                                                                                                                                                                                                                                                                                                                                                                                                                    Security AND
: b.      EITHER of the following has occurred:                                                                                                                        HA1.2        1            2      3          4          5          6      DEF            HU1.2          1          2          3          4        5          6        DEF
: 1. ANY of the following safety functions cannot be controlled or maintained.                                                                                A validated notification from NRC of an aircraft attack threat within 30 minutes of PINGP. Notification of a credible security threat directed at PINGP.
* Reactivity control
* Core cooling                                                                                                                                                                                                                                    HU1.3        1            2        3        4        5          6          DEF
* RCS heat removal                                                                                                                                                                                                                                A validated notification from the NRC providing information of an aircraft threat.
OR
: 2. Damage to spent fuel has occurred or is IMMINENT.
Phenomenon Natural & DestructiveEffluents / Release Rates                                                                                                                                                                                                                                                                                                                                                                                                                                    Phenomenon Natural & DestructiveEffluents / Release Rates HU2      Seismic event greater than OBE levels.
NOTE:
Depending upon the plant HU2.1          1          2      3        4          5      6      DEF mode at the time of the              Seismic event greater than Operational Basis Earthquake (OBE) as indicated by an event escalation criteria              "OBE Exceedance" alarm on Seismic Monitoring Panel.
for HU2 & HU3 would be                HU3      Hazardous event.
CA6 or SA9.                      Note:    EAL HU3.4 does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.
HU3.1          1          2      3        4          5      6      DEF Table H1                                                                                                                                                                A tornado strike within the PROTECTED AREA.
Mode                          Building                                Area/Room Abnormal Rad Release / Rad Effluent Hazards                                                                                                                                                                                                                                                                                                                                                                                                                                                      Abnormal Rad Release / Rad Effluent Hazards HU3.2          1          2        3        4          5          6      DEF Mode 3                    Auxiliary Building                      735' General Area Internal room or area flooding of a magnitude sufficient to require manual or automatic Mode 4                    Auxiliary Building                      695' General Area                                                                                                                electrical isolation of a SAFETY SYSTEM component needed for the current operating Mode 5                    Auxiliary Building                      695' General Area                                                                                                                mode.
HU3.3          1        2          3          4          5          6        DEF Movement of personnel within the PROTECTED AREA is impeded due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release).
HU3.4          1          2        3        4          5          6        DEF A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles.
HU3.5          1          2        3        4          5          6        DEF High or low river water level occurrences affecting the PROTECTED AREA as indicated by EITHER of the following:
* River Intake level greater than 692 ft MSL
* River Intake level less than 669.5 ft MSL HU4      FIRE potentially degrading the level of safety of the plant.
NOTE:
Table H2                                                                                                Depending upon the plant              Note:    The Emergency Director should declare the Unusual Event promptly upon mode at the time of the                        determining that the applicable time has been exceeded, or will likely be Building                                                      Rooms                                                                                                    event escalation criteria                      exceeded.
Control Building                                              Control Room                                                                                                for HU4 would be                  HU4.1        1          2          3          4          5        6      DEF CA6 or SA9.                    a.        A FIRE is NOT extinguished within 15-minutes of ANY of the following FIRE Control Room Chiller Room                                                                                                                                detection indications:
* Report from the field (i.e., visual observation)
Turbine Building                                              Battery Rooms
* Receipt of multiple (more than 1) fire alarms or indications AFW Rooms
* Field verification of a single fire alarm AND D1/D2 Diesel Generator Rooms                                                                                                                  b.        The FIRE is located within ANY of the Table H2 plant rooms or areas.
Safeguards Switchgear Rooms Fire                                                                                                                                                                                                                                                                                                              HU4.2          1        2          3          4          5          6        DEF Fire Relay Room                                                                                                                                    a.        Receipt of a single fire alarm (i.e., no other indications of a FIRE).
AND 480 V Switchgear Rooms                                                                                                                        b.        The FIRE is located within ANY of the Table H2 plant rooms or areas.
AND Event Montoring Rooms                                                                                                                          c.        The existence of a FIRE is not verified within 30-minutes of alarm receipt.
D5/D6 Diesel Generator Building                              All HU4.3        1          2        3          4          5          6    DEF Auxiliary Building                                            All areas within AB Special Vent Zone                                                                                                          A FIRE within the plant PROTECTED AREA or ISFSI PROTECTED AREA not extinguished within 60-minutes of the initial report, alarm or indication.
Plant Screen House                                            Safeguards Cooling Water Pump Rooms Safeguards Traveling Screen Room                                                                                                              HU4.4          1        2            3        4        5        6  DEF A FIRE within the plant PROTECTED AREA or ISFSI PROTECTED AREA that requires Shield Building/Containment                                  All                                                                                                                                            firefighting support by an offsite fire response agency to extinguish.
PINGP 1576, Rev. 11 Doc. Type/Sub Type: EP/EVT Retention: Lifetime +                                                                                                                                                                                                                                                                                                                                                                                                                                                                                      2 of 9
 
Prairie Island Nuclear Generating Plant                                                                                                                                                                                                EMERGENCY ACTION LEVEL MATRIX GENERAL EMERGENCY                                                                                  SITE AREA EMERGENCY                                                                                                      ALERT                                                                                                NUE                                                      HOT & COLD Release of gaseous radioactivity resulting in offsite dose                                              Release of gaseous radioactivity resulting in offsite dose                              HA5      Release of Gaseous    gaseous release    or liquid impeding    radioactivity access          resulting to equipment      in offsite necessary                                    Release of gaseous or liquid radioactivity greater than 2 times Toxic, Corrosive,                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                            Toxic, Corrosive, greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE.                                                greater than 100 mrem TEDE or 500 mrem thyroid CDE.                                              dose for  greater normal    than plant  10 mrem TEDE operations,      or 50 or cooldown  mrem  thyroid CDE.
shutdown.                                            the ODCM limits for 60 minutes or longer.
Note:    If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is Asphyxiant or                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                Asphyxiant or warranted.
HA5.1                              3        4          5
: a.        Release of a toxic, corrosive, asphyxiant or flammable gas into any of the Table Flammable Gases                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                              Flammable Gases H1 plant rooms or areas.
AND
: b.        Entry into the room or area is prohibited or impeded.
HS6      Inability to control a key safety function from outside the                            HA6      Control Room evacuation resulting in transfer of plant control Control Room.                                                                                    to alternate locations.
Shift Manager /                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                              Shift Manager /
Note:    The Emergency Director should declare the Site Area Emergency promptly upon            HA6.1        1        2          3          4        5          6      DEF determining that 15 minutes has been exceeded, or will likely be exceeded.              An event has resulted in plant control being transferred from the Control Room to the ISFSI                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                        ISFSI remote Hot Shutdown Panels.
HS6.1          1        2        3          4          5        6 Control Room                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                  Control Room
: a.        An event has resulted in plant control being transferred from the Control Room to the remote Hot Shutdown Panels.
AND
: b.        Control of ANY of the following key safety functions is not reestablished within Abnormal Rad Release / Rad Effluent Hazards Continued                                                                                                                                                                                                                                                                                                                                                                                                                                                                        Abnormal Rad Release / Rad Effluent Hazards Continued 15 minutes.
Emergency DirectorEffluents / Release Rates                                                                                                                                                                                                                                                                                                                                                                                                                                                                                  Emergency DirectorEffluents / Release Rates
* Reactivity control (Modes 1, 2, and 3 only)
* Core cooling
* RCS heat removal HG7        Other conditions exist which in the judgment of the                                          HS7      Other conditions exist which in the judgment of the                                    HA7      Other conditions exist which in the judgment of the                                    HU7        Other conditions exist which in the judgment of the Evacuation                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                    Evacuation Emergency Director warrant declaration of a General                                                    Emergency Director warrant declaration of a Site Area                                            Emergency Director warrant declaration of an Alert.                                                Emergency Director warrant declaration of a NUE.
Emergency.                                                                                              Emergency.
HG7.1          1        2          3        4          5          6        DEF                        HS7.1          1          2          3        4          5        6      DEF                HA7.1          1          2        3          4        5          6        DEF                  HU7.1        1          2          3          4          5          6      DEF Other conditions exist which in the judgment of the Emergency Director indicate that                    Other conditions exist which in the judgment of the Emergency Director indicate that              Other conditions exist which, in the judgment of the Emergency Director, indicate that            Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial                    events are in progress or have occurred which involve actual or likely major failures of          events are in progress or have occurred which involve an actual or potential substantial          events are in progress or have occurred which indicate a potential degradation of the level Judgment                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                      Judgment core degradation or melting with potential for loss of containment integrity or HOSTILE                  plant functions needed for protection of the public or HOSTILE ACTION that results in            degradation of the level of safety of the plant or a security event that involves probable life  of safety of the plant or indicate a security threat to facility protection has been initiated. No ACTION that results in an actual loss of physical control of the facility. Releases can be              intentional damage or malicious acts, (1) toward site personnel or equipment that could          threatening risk to site personnel or damage to site equipment because of HOSTILE                releases of radioactive material requiring offsite response or monitoring are expected ICS/EALS                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                      ICS/EALS reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for                lead to the likely failure of or, (2) that prevent effective access to equipment needed for the  ACTION. Any releases are expected to be limited to small fractions of the EPA Protective          unless further degradation of safety systems occurs.
more than the immediate site area.                                                                      protection of the public. Any releases are not expected to result in exposure levels which        Action Guideline exposure levels.
exceed EPA Protective Action Guideline exposure levels beyond the site boundary.
Table E1                                      EU1 Damage to a loaded cask CONFINEMENT BOUNDARY.
Gamma      Neutron Cask                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                          Cask Location                                                EU1.1      1            2      3          4        5          6        DEF 1                      2                                3                                4                    5                                6                                  DEF mrem/hr    mrem/hr Center of top protective cover              90        20 Confinement                                  Between cask flange and side neutron 160        380 Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading on the surface of the spent fuel cask greater than ANY Table E1 value:              Modes:              Power                      Startup                      Hot Standby 350 F Hot Shutdown            Cold Shutdown Refueling                              Defueled Confinement shield 200 F Boundary                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                      Boundary 350 F > Tavg > 200 F Mid-height of side neutron shield            80        70 Between cask bottom and side neutron 170        1860 shield PINGP 1576, Rev. 11 Doc. Type/Sub Type: EP/EVT Retention: Lifetime +                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                              3 of 9
 
Prairie Island Nuclear Generating Plant                                                                                                                                                    EMERGENCY ACTION LEVEL MATRIX GENERAL EMERGENCY                                                                      SITE AREA EMERGENCY                                                                                              ALERT                                                                                                NUE                                                                  HOT SG1        Prolonged loss of all offsite and all onsite AC power to                    SS1        Loss of all offsite and all onsite AC power to emergency buses                SA1        Loss of all but one AC power source to emergency buses for                          SU1          Loss of all offsite AC power capability to emergency buses for emergency buses.                                                                        for 15 minutes or longer.                                                                15 minutes or longer.                                                                            15 minutes or longer.
Note:                                                                                  Note:                                                                                    Note:      The Emergency Director should declare the Alert promptly upon determining that      Note:        The Emergency Director should declare the Unusual Event promptly upon The Emergency Director should declare the General Emergency promptly upon              The Emergency Director should declare the Site Area Emergency promptly upon 15 minutes has been exceeded, or will likely be exceeded.                                        determining that 15 minutes has been exceeded, or will likely be exceeded.
determining that 4 hours has been exceeded, or will likely be exceeded.                determining that 15 minutes has been exceeded, or will likely be exceeded.
SG1.1        1        2      3      4                                                SS1.1          1      2        3      4                                                SA1.1          1      2      3      4                                                        SU1.1          1        2    3        4
: a. Loss of ALL offsite and ALL onsite AC power to both Safeguards                      Loss of ALL offsite and ALL onsite AC power to both Safeguards                          a. AC power capability to both Safeguards Buses  15 and 16                                      Loss of ALL offsite AC power capability (Table S2) to both Safeguards Buses 15 and 16 (25 Buses 15 and 16 (25 and 26).                                                        Buses 15 and 16 (25 and 26) for 15 minutes or longer.                                      (25 and 26) is reduced to a single power source (Table S1) for                              and 26) for 15 minutes or longer.
Loss of Power                                                                                                                                                                                                                                                                                                                                                                                                                                  Loss of Power AND                                                                                                                                                                                15 minutes or longer.
: b. EITHER of the following:                                                                                                                                                      AND                                                                                                                                          Table S2
* Restoration of at least one AC emergency bus in less than 4 hours is not likely.                                                                                            b. Any additional single power source failure will result in a loss of all AC power to SAFETY                            Unit 1                                    Unit 2
* Core Cooling CSF - RED                                                                                                                                                        SYSTEMS.                                                                                                          1R Transformer                            2RY Transformer SG8        Loss of all AC and Vital DC power sources for 15 minutes or                  SS8                Loss of all Vital DC power for 15 minutes or longer.                                                                                                                                      CT11 Transformer                          CT12 Transformer longer.                                                                                                                                                                                                    Table S1 Note:      The Emergency Director should declare the General Emergency promptly upon    Note:                                                                                                        Unit 1                                  Unit 2 The Emergency Director should declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.                                                                                                            1R Transformer                        2RY Transformer determining that 15 minutes has been exceeded, or will likely be exceeded.
CT11 Transformer                      CT12 Transformer SG8.1        1        2      3      4                                                  SS8.1          1      2      3      4                                                          D1 Diesel Generator                    D5 Diesel Generator
: a. Loss of ALL offsite and ALL onsite AC power  to both Safeguards                      Indicated voltage is less than 111 VDC on both 125 VDC                                                D2 Diesel Generator                    D6 Diesel Generator Buses 15 and 16 (25 and 26) for 15 minutes or longer.                                Panels 11 and 12 (21 and 22) for 15 minutes or longer.                                                                Aligned Available Cross-Ties AND
: b. Indicated voltage is less than 111 VDC on both 125 VDC Panels 11 and 12 (21 and 22) for 15 minutes or longer.
SA2        UNPLANNED loss of Control Room indications for 15 minutes                          SU2          UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.                                              or longer.
Note:      The Emergency Director should declare the Alert promptly upon determining that      Note:        The Emergency Director should declare the Unusual Event promptly upon System Malfunctions                                                                                                                                                                                                                                                                                                                                                                                                                            System Malfunctions Control Room Indications                                                                                                                                                                                                                                                                                                                                                                                                                        Control Room Indications 15 minutes has been exceeded, or will likely be exceeded.                                        determining that 15 minutes has been exceeded, or will likely be exceeded.
SA2.1          1        2      3        4
: a. An UNPLANNED event results in the inability to monitor one or more of the following              SU2.1          1        2        3      4 parameters from within the Control Room for 15 minutes or longer.
Reactor Power                                                                    a. An UNPLANNED event results in the inability to monitor one or more of the following Pressurizer Level                                                                  parameters from within the Control Room for 15 minutes or longer.
RCS Pressure Core Exit Temperature Level in at least one steam generator                                                        Reactor Power Auxiliary Feed Water Flow                                                                    Pressurizer Level AND                                                                                                          RCS Pressure
: b. ANY of the following transient events in progress.                                                        Core Exit Temperature
* Automatic or manual runback greater than 25% thermal reactor power                                  Level in at least one steam generator
* Electrical load rejection greater than 25% full electrical load                                    Auxiliary Feed Water Flow
* Reactor trip
* ECCS (SI) actuation SU3          Reactor coolant activity greater than Technical Specification allowable limits.
SU3.1          1        2        3      4 Reactor Coolant System                                                                                                                                                                                                                                                                                                                                                                                                                          Reactor Coolant System 1(2)R-9 reading greater than 1.0 R/hr.
SU3.2          1        2        3      4 Sample analysis indicates that a reactor coolant activity value is greater than an allowable limit specified in Technical Specification 3.4.17 Condition C (Dose Equivalent I-131 > 30 &#xb5;Ci/gm).
SU4        RCS leakage for 15 minutes or longer.
Note:      The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
SU4.1        1      2      3        4 RCS unidentified or pressure boundary leakage greater than 10 gpm for 15 minutes or longer.
SU4.2          1      2        3      4 RCS identified leakage greater than 25 gpm for 15 minutes or longer.
SU4.3          1      2        3      4 Leakage from the RCS to a location outside containment greater than 25 gpm for 15 minutes or longer.
PINGP 1576, Rev. 11 Doc. Type/Sub Type: EP/EVT Retention: Lifetime +                                                                                                                                                                                                                                                                                                                                                                                                                          4 of 9
 
Prairie Island Nuclear Generating Plant                                                                                                                                  EMERGENCY ACTION LEVEL MATRIX GENERAL EMERGENCY                                                      SITE AREA EMERGENCY                                                                                              ALERT                                                                                                NUE                                                                      HOT SG1  Prolonged loss of all offsite and all onsite AC power to  SS1 SS5        Loss of all Inability to offsite shutdownandthe all onsite reactorAC poweratochallenge causing    emergency  buses to core              SA1 SA5        Loss of all or Automatic  butmanual one ACtrip power failssource to emergency to shutdown          buses the reactor, andfor                      SU1 SU5        Loss of all or Automatic  offsite AC trip manual  power    capability fails          to emergency to shutdown            buses for the reactor.
emergency buses.                                                      for 15 minutes cooling  or RCSor  longer.
heat removal.                                                          15 minutes or subsequent    longer.actions taken at the main control boards manual                                                                              15 minutes or longer.
are not successful in shutting down the reactor.
Note:      Heat Sink CSF should not be considered RED if total AFW flow is less than 200  Note:      A manual action is any operator action, or set of actions, which causes the control  Note:      A manual action is any operator action, or set of actions, which causes the control gpm due to operator action.                                                                rods to be rapidly inserted into the core, and does not include manually driving in              rods to be rapidly inserted into the core, and does not include manually driving in SS5.1          1                                                                                  control rods or implementation of boron injection strategies.                                    control rods or implementation of boron injection strategies.
: a. An automatic or manual trip did not reduce reactor power to less than 5%.
Loss of Power ATWS                                                                                                                                                                                                                                                                                                                                                                                                                            Loss of Power ATWS AND                                                                                            SA5.1          1                                                                                  SU5.1          1
: b. All manual actions to reduce reactor power to less than 5% have been unsuccessful.      a. An automatic  or manual trip did not reduce reactor power to less than 5%.                    a. An initial automatic or manual trip did not reduce reactor power to less than 5%.
AND                                                                                              AND AND                                                                                        b. Manual actions taken at the main control boards are not successful in reducing reactor        b. EITHER of the following:
: c. EITHER of the following conditions exist:                                                  power to less than 5%.
* A subsequent manual action taken at the main control boards is successful in shutting
* Core Cooling CSF - RED                                                                                                                                                                        down the reactor.
* Heat Sink CSF - RED                                                                                                                                                                                  o Reactor Trip (Switch) o AMSAC/DSS (Switch) o Turbine Trip (Pushbutton)
* A  subsequent  automatic trip is successful in shutting down the reactor.
SA9        Hazardous event affecting a SAFETY SYSTEM needed for the                              SU6          Loss of all onsite or offsite communications capabilities.
current operating mode.
Note:
* If the affected SAFETY SYSTEM train was already inoperable or out of service System MalfunctionsContinued SystemMalfunctions                                                                                                                                                                                                                                                                                                                                                                                                                    System MalfunctionsContinued SystemMalfunctions before the hazardous event occurred, then this emergency classification is not            SU6.1          1        2      3      4 warranted.
Communications / Hazard Events                                                                                                                                                                                                                                                                                                                                                                                                          Communications / Hazard Events
* If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of        Loss of ALL of the following onsite communication methods:
degraded performance to at least one train of a SAFETY SYSTEM, then this
* Sound Powered Phones emergency classification is not warranted.
* Plant Paging System SA9.1          1      2        3        4
* Plant Telephone Network
: a. The  occurrence of ANY of the following hazardous events:
* Plant Radio System
* Seismic event (earthquake)
* Internal or external flooding event                                                          SU6.2          1        2      3      4
* High winds or tornado strike
* FIRE                                                                                    Loss of ALL of the following Offsite Response Organization (ORO) communications methods:
* EXPLOSION
* Plant Telephone Network
* Low River Water Level
* Plant Radio System (dedicated offsite channels)
* Other events with similar hazard characteristics as determined by the Shift Manager.                                                                                  SU6.3          1        2      3      4 AND
: b. 1. Event damage has caused indications of degraded performance on one train of a              Loss of ALL of the following NRC communications methods:
SAFETY SYSTEM needed for the current operating mode.
* Plant Telephone Network AND
* ENS Network
: 2. EITHER of the following:
* Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM needed for the current operating mode.
* The event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode.
SU7        Failure to isolate containment or loss of containment pressure control.
SU7.1          1      2        3      4 Containment                                                                                                                                                                                                                                                                                                                                                                                                                              Containment
: a. Failure of containment to isolate when required by an actuation signal.
AND
: b. ALL required penetrations are not closed within 15 minutes of the actuation signal.
SU7.2        1          2      3      4
: a. Containment  pressure  greater than 23 psig.
AND
: b. Less than one full train of containment spray and any two containment fan coils is operating per design for 15 minutes or longer.
PINGP 1576, Rev. 11 Doc. Type/Sub Type: EP/EVT Retention: Lifetime +                                                                                                                                                                                                                                                                                                                                                                                                                  5 of 9
 
Prairie Island Nuclear Generating Plant                                                                                                                                            EMERGENCY ACTION LEVEL MATRIX HOT SG1        Prolonged loss of all offsite and all onsite AC power to                          SS1        Loss of all offsite and all onsite AC power to emergency buses                SA1        Loss of all but one AC power source to emergency buses for                        SU1          Loss of all offsite AC power capability to emergency buses for emergency buses.                  GE                                                        for 15 minutes or longer.                                      SAE                      15 minutes or longer.                                                                              ALERT 15 minutes or longer.
NOTE:            To classify the event: Determine which combination of the three barriers meet the Loss or Potential Loss threshold and check each applicable box.
                                                                *Containment Radiation Monitors are sensitive to temperature changes which can cause thermally induced current errors. These monitors should not be used for classification in the first 5 minutes after a Steam Line Break or LOCA in Containment. Once 5 minutes has expired these monitors can be used for EAL determination.
FG1 - Loss of any two barriers and Loss or Potential Loss of the third barrier.                                                                  FS1 -Loss or Potential Loss of any two barriers.                                                        FA1 - Any Loss or any Potential Loss of either the Fuel Clad or RCS barrier.
1            2          3          4                                                                                        1        2          3        4                                                                                          1          2            3          4 Loss of Power                                                                                                                                                                                                                                                                                                                                                                                                                    Loss of Power Fuel Clad Barrier                                                                                                              RCS Barrier                                                                                                            Containment Barrier LOSS                                                    POTENTIAL LOSS                                                          LOSS                                                POTENTIAL LOSS                                                        LOSS                                                  POTENTIAL LOSS
: 1. RCS or SG Tube Leakage                                                                                                1. RCS or SG Tube Leakage                                                                                              1. RCS or SG Tube Leakage A. Core Cooling CSF - ORANGE entry                                  A. An automatic or manual                          A. GREATER THAN or EQUAL to 60 gpm leak rate                        A. A leaking or RUPTURED SG is FAULTED conditions met.                                                      ECCS (SI) actuation is required by                  excluding normal reductions in RCS inventory                          outside of containment.
EITHER of the following:                            (example, letdown system or RCP seal leak-off) caused by EITHER of the following:
: 1. UNISOLABLE RCS leakage OR                                                          1. UNISOLABLE RCS leakage
: 2. SG tube RUPTURE.                                          OR
: 2. SG tube leakage.
OR B. RCS Integrity CSF - RED entry conditions System Malfunctions FissionSystem ProductMalfunctions Barrier Degradation met.
: 2. Inadequate Heat Removal                                                                                                      2. Inadequate Heat Removal                                                                                                  2. Inadequate Heat Removal A. Core Cooling CSF - RED entry conditions        A. Core Cooling CSF - ORANGE entry                                                                                            Note: Heat Sink CSF should not be considered RED                                                                                  A. Core Cooling CSF - RED entry met.                                              conditions met.                                                                                                                if total AFW flow is less than 200 gpm due to                                                                                    conditions met for 15 minutes or longer.
OR                                                                                                                        operator action.
Note:  Heat Sink CSF should not be considered                                                                                A. Heat Sink CSF - RED entry conditons met.
RED if total AFW flow is less than 200 gpm due to operator action.
Fission Product Barrier Degradation B. Heat Sink CSF - RED entry conditions met.
: 3. RCS Activity/Containment Radiation                                                                                              3. RCS Activity/Containment Radiation                                                                                          3. RCS Activity/Containment Radiation
* A. Containment radiation monitor reading
* A. Containment radiation monitor reading
* A. Containment radiation monitor reading greater greater than 5,500 R/hr on 1(2)R-48 or 49.                                                                                        greater than 40 R/hr on 1(2)R-48 or 49.                                                                                                                                                than 20,000 R/hr on 1(2)R-48 or 49.
OR B.            Coolant activity greater than 300 &#xb5;Ci/gm dose equivalent Iodine-131.
: 4. Containment Integrity or Bypass                                                                                          4. Containment Integrity or Bypass                                                                                          4. Containment Integrity or Bypass A. Containment isolation is required.                  A. Containment CSF - RED entry conditions AND                                                      met.
EITHER of the following:                                      OR
: 1. Containment integrity has been lost                  B. Containment H2 concentration greater than based on Emergency Director                                or equal to 6%.
judgment.                                                  OR OR                                                    C. 1. Containment pressure greater than 23
: 2. UNISOLABLE pathway from the                                    psig.
containment to the environment exists.                        AND
: 2. Less than one full train of containment OR                                                              spray and any two containment fan coils B. Indications of RCS leakage outside of                            is operating per design for 15 minutes or containment.                                                    longer.
: 5. Emergency Director Judgment                                                                                                        5. Emergency Director Judgment                                                                                            5. Emergency Director Judgment A. ANY condition in the opinion of the Emergency        A. ANY condition in the opinion of the Emergency                              A. ANY condition in the opinion of the Emergency        A. ANY condition in the opinion of the Emergency                  A. ANY condition in the opinion of the Emergency        A. ANY condition in the opinion of the Emergency Director that indicates Loss of the Fuel Clad            Director that indicates Potential Loss of the                                Director that indicates Loss of the RCS Barrier.        Director that indicates Potential Loss of the                    Director that indicates Loss of the Containment          Director that indicates Potential Loss of the Barrier.                                                Fuel Clad Barrier.                                                                                                                    RCS Barrier.                                                      Barrier.                                                Containment Barrier.
HOT CONDITIONS PINGP 1576, Rev. 11 Doc. Type/Sub Type: EP/EVT Retention: Lifetime +                                                                                                                                                                                                                                                                                                                                                                                                          6 of 9
 
Prairie Island Nuclear Generating Plant                                                                                                                                                  EMERGENCY ACTION LEVEL MATRIX GENERAL EMERGENCY                                                                      SITE AREA EMERGENCY                                                                                              ALERT                                                                                      NUE                                                                                        COLD CG1      Loss of RPV inventory affecting fuel clad integrity with                        CS1    Loss of RPV inventory affecting core decay heat removal                          CA1        Loss of RPV inventory.                                                      CU1        UNPLANNED loss of RPV inventory for 15 minutes or longer.
containment challenged.                                                                  capability.
Note:    The Emergency Director should declare the General Emergency promptly upon                                                                                                                                                                                        Note:      The Emergency Director should declare the Unusual Eventt promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.      CS1.1                                              5                                    CA1.1                                                  5        6                              determining that 15 minutes has been exceeded, or will likely be exceeded.
: a.      CONTAINMENT CLOSURE not established.                                            Loss of RPV inventory as indicated by level less than ANY of the following:
CG1.1                                              5        6                            AND
* 10 Refueling Canal                                                            CU1.1                                              5          6
: a.        RPV level less than 56% RVLIS Full Range (Mode 5) for 30 minutes or longer. b.      RPV level less than 65% RVLIS Full Range (Mode 5).
* 10 ERCS DP                                                                  UNPLANNED loss of reactor coolant results in RPV level less than a procedurally required
* 69% RVLIS Full Range (Mode 5).                                              lower limit for 15 minutes or longer.
AND                                                                                        CS1.2                                              5
: a.      CONTAINMENT CLOSURE established.                                                                                                                                          CU1.2                                                  5          6
: b.        ANY indication from the Containment Challenge Table C1.                          AND                                                                                      Note:      The Emergency Director should declare the Alert promptly upon determining  a.        RPV level cannot be monitored.
Inventory Control                                                                                                                                                                                                                                                                                                                                                                                                                          Inventory Control
: b.      RPV level less than 56% RVLIS Full Range (Mode 5).                                          that 15 minutes has been exceeded, or will likely be exceeded.              AND CG1.2                                                5        6                                                                                                                  CA1.2                                                5        6                      b.        UNPLANNED increase in Containment Sumps A or C, or Waste Holdup Tank
: a.        RPV level cannot be monitored for 15 minutes or longer.                                levels.
: a.        RPV level cannot be monitored for 30 minutes or longer in Mode 5 or 6.                                                                                                    AND AND                                                                                                                                                                                b.        UNPLANNED increase in Containment Sumps A or C, or Waste Holdup Tank
: b.        Core uncovery is indicated by ANY of the following:                                                                                                                                levels due to a loss of RPV inventory.
* 1(2)R-2, Containment Vessel Area Monitor, reading greater than          Note:  The Emergency Director should declare the Site Area Emergency promptly upon Cold Shutdown / Refuel System Malfunction                                                                                                                                                                                                                                                                                                                                                                                                          Cold Shutdown / Refuel System Malfunction 0                                                                    determining that 30 minutes has been exceeded, or will likely be exceeded.
1 R/hr (10 R/hr).
* Erratic source range monitor indication.
* UNPLANNED increase in Containment Sumps A or C, or Waste                CS1.3                                              5        6 Holdup Tank levels of sufficient magnitude to indicate core uncovery. a.      RPV level cannot be monitored for 30 minutes or longer in AND                                                                                                Mode 5 or 6.
: c.        ANY indication from the Containment Challenge Table C1.                        AND Containment Challenge Table C1                                    b.      Core uncovery is indicated by ANY of the following:
* CONTAINMENT CLOSURE not established*
* 1(2)R-2, Containment Vessel Area Monitor, reading greater than 0
* H2 concentration greater than or equal to 6% inside containment                                          1 R/hr (10 R/hr).
* UNPLANNED increase in containment pressure
* Erratic source range monitor indication.
* If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit,
* UNPLANNED increase in Containment Sumps A or C, or Waste then declaration of a General Emergency is not required.                                                    Holdup Tank levels of sufficient magnitude to indicate core uncovery.
CA2        Loss of all Offsite and all Onsite AC power to emergency                    CU2        Loss of all but one AC power source to emergency buses for buses for 15 minutes or longer.                                                        15 minutes or longer.
Note:      The Emergency Director should declare the Alert promptly upon determining  Note:      The Emergency Director should declare the Unusual Event promptly upon that 15 minutes has been exceeded, or will likely be exceeded.                        determining that 15 minutes has been exceeded, or will likely be exceeded.
CA2.1                                                5        6      DEF              CU2.1                                                  5          6      DEF Loss of ALL Offsite and ALL Onsite AC Power to both Safeguards Buses 15 and 16        a.        AC power capability to both Safeguards Buses 15 and 16 (25 and 26) for 15 minutes or longer.                                                            (25 and 26) is reduced to a single power source (Table S1) for 15 minutes or longer.
AND Loss of Power                                                                                                                                                                                                                                                                                                                                                                                                                              Loss of Power
: b.        Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS.
Table S1 Unit 1                            Unit 2 1R Transformer                2RY Transformer CT11 Transformer                CT12 Transformer D1 Diesel Generator            D5 Diesel Generator D2 Diesel Generator            D6 Diesel Generator Aligned Available Cross-Ties CU4        Loss of Vital DC power for 15 minutes or longer.
Note:      The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
CU4.1                                                  5            6 Indicated voltage is less than 111 VDC on required 125 VDC Panels 11 and 12 (21 and 22) for 15 minutes or longer.
PINGP 1576, Rev. 11 Doc. Type/Sub Type: EP/EVT Retention: Lifetime +                                                                                                                                                                                                                                                                                                                                                                                                                                                    7 of 9
 
Prairie Island Nuclear Generating Plant                            EMERGENCY ACTION LEVEL MATRIX GENERAL EMERGENCY  SITE AREA EMERGENCY                                                ALERT                                                                                        NUE                                                                                                  COLD CA3        Inability to maintain the plant in cold shutdown.                                CU3      UNPLANNED increase in RCS temperature.
Note:      The Emergency Director should declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.          CU3.1                                              5        6 UNPLANNED increase in RCS temperature to greater than 200 F.
Temperature Control                                                                                                                                                                                                                                                                                                    Temperature Control CA3.1                                              5        6 UNPLANNED increase in RCS temperature to greater than 200 F for greater than the            Note:    The Emergency Director should declare the Unusual Event promptly upon duration specified in Table C2.                                                                        determining that 15 minutes has been exceeded, or will likely be exceeded.
Table C2: RCS Heat-up Duration Thresholds CONTAINMENT                    Heat-up              CU3.2                                                5        6 RCS CLOSURE                    Duration              Loss of ALL RCS temperature and RPV level indication for 15 minutes or longer.
Not intact or in RCS Reduced          Not Established              0 minutes Inventory                  Established                20 minutes*
Intact (capable of being N/A                  60 minutes*
pressurized)
* If RHR (or at least 1 S/G when RCS intact) is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.
CA3.2                                                  5          6 UNPLANNED RCS pressure increase greater than 25 psig.
(This Threshold does not apply during water-solid plant conditions.)
Cold Shutdown / Refuel System Malfunction Continued                                                                                                                                                                                                                                                                            Cold Shutdown / Refuel System Malfunction Continued CU5      Loss of all onsite or offsite communications capabilities.
CU5.1                                                5        6      DEF Loss of ALL of the following onsite communication methods:
* Sound Powered Phones Communications                                                                                                                                                                                                                                                                                                          Communications
* Plant Paging System
* Plant Telephone Network
* Plant Radio System CU5.2                                                  5        6    DEF Loss of ALL of the following Offsite Response Organizations (ORO) communications methods:
* Plant Telephone Network
* Plant Radio System (dedicated offsite channels)
CU5.3                                              5          6      DEF Loss of ALL of the following NRC communications methods:
* Plant Telephone Network
* ENS Network CA6        Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.
Note:
* If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then this emergency classification is not warranted.
* If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.
CA6.1                                                  5        6 Hazardous Events                                                                                                                                                                                                                                                                                                        Hazardous Events
: a.        The occurrence of ANY of the following hazardous events:
* Seismic event (earthquake)
* Internal or external flooding event
* High winds or tornado strike
* FIRE
* EXPLOSION
* Low River Water Level
* Other events with similar hazard characteristics as determined by the Shift Manager.
AND
: b.            1. Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode.
AND
: 2.      EITHER of the following:
* Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM needed for the current operating mode.
* Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode.
COLD CONDITIONS PINGP 1576, Rev. 11 Doc. Type/Sub Type: EP/EVT Retention: Lifetime +                                                                                                                                                                                                                                                                                                                                  8 of 9
 
Prairie Island Nuclear Generating Plant                                                                                                                                                EMERGENCY ACTION LEVEL MATRIX CONFINEMENT BOUNDARY: The barrier(s) between spent fuel and the environment once the spent fuel is DEFINITIONS HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment,    PROTECTED AREA: The area encompassing all controlled areas within the security protected area fence as shown processed for dry storage.                                                                                      take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using  in USAR Figure 1.1-3, Site Plan Prairie Island Security Fence. This area does not include the ISFSI.
guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy CONTAINMENT CLOSURE: No open containment penetrations exist as identified in C19.9, Containment                the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or  REFUELING PATHWAY: The reactor refueling cavity, spent fuel pool, or fuel transfer canal.
Boundary Control during Mode 5, Cold Shutdown and Mode 6, Refueling. The definition of an open containment felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to penetration is a penetration that provides direct access from the containment atmosphere to the outside        address such activities (i.e., this may include violent acts between individuals in the owner controlled area).      RUPTURED: The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to environment with no automatic closure available.                                                                                                                                                                                    require a safety injection.
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold EXPLOSION: A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical        IMMINENT: The trajectory of events or conditions is such that an EAL will be met within a relatively short period of shutdown condition, including the ECCS. These are typically systems classified as safety-related.
reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical      time regardless of mitigation or corrective actions.
component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an                                                                                                                      SECURITY CONDITION: Any Security Event as listed in the approved security contingency plan that constitutes a explosion. Such events may require a post-event inspection to determine if the attributes of an explosion are  INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): A complex that is designed and constructed for threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the present.                                                                                                        the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.        plant. A SECURITY CONDITION does not involve a HOSTILE ACTION.
UNISOLABLE: An open or breached system line that cannot be isolated, remotely or locally.
FAULTED: The term applied to a steam generator that has a steam leak on the secondary side of sufficient size ISFSI PROTECTED AREA: The area surrounding the Independent Spent Fuel Storage Installation encompassed by to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely          the double chain link fence surrounding the ISFSI as defined in the Security Plan; the ISFSI Protected Area is      UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected depressurized.                                                                                                  excluded from the Plant Protected Area.                                                                              plant response to a transient. The cause of the parameter change or event may be known or unknown.
FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large    OWNER CONTROLLED AREA: Land owned or leased by Prairie Island Nuclear Generating Plant. This area is                VISIBLE DAMAGE: Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing or quantities of smoke and heat are observed.                                                                      bounded by a wire mesh, owner controlled fence. Unauthroized personnel are not allowed within this area.            analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train.
NOTE: Refer to F3-2.1, Emergency Action Level Technical Bases, as well as this Wall Chart to Classify an Event PINGP 1576, Rev. 11 Doc. Type/Sub Type: EP/EVT Retention: Lifetime +                                                                                                                                                                                                                                                                                                                                                        9 of 9
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                PLANT SAFETY PROCEDURE NUMBER:
F3                  CORE DAMAGE DETERMINATION REV:
F3-17.1 6
REFERENCE USE
* Procedure should be at the work location.
* Procedure segments may be performed from memory.
* Use the procedure to verify segments have been completed.
* When required, sign or initial appropriate blocks to certify that all segments are complete.
APPROVAL:
PCR #: 602000015508 Page 1 of 21
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                    PLANT SAFETY PROCEDURE NUMBER:
F3                    CORE DAMAGE DETERMINATION REV:
F3-17.1 6
1.0    PURPOSE 1.1    The purpose of this procedure is to provide a means to determine if reactor core damage has occurred.
2.0    APPLICABILITY 2.1    In the absence of Nuclear Engineering Staff, this procedure should be used during the initial phases of an emergency to determine the likelihood of core damage.
3.0    PRECAUTIONS C      3.1    The values obtained by the use of this procedure are best estimates, but should be used for decisions until a more detailed evaluation can be completed. The Inadequate Core Cooling Monitor (lCCM) should be used during Emergency Operating Procedure (EOP) usage when checking Reactor Coolant System (RCS) subcooling. Emergency Response Computer System (ERCS) subcooling values should be used for information and comparison only. Any EOP decisions should be based on ICCM indications.
3.2    Iodine spiking may occur after a shutdown or significant power change, usually during the 2 to 6 hour period following the power change. Iodine spiking is a characteristic of the condition where an increase in the normal primary coolant activity is noted, but no damage to the cladding has occurred.
4.0    SPECIAL CONSIDERATION NONE 5.0    RESPONSIBILITIES 5.1    TSC technical staff are responsible to determine if reactor core damage may or has occurred according to the guidance provided in this procedure.
Page 2 of 21
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                  PLANT SAFETY PROCEDURE NUMBER:
F3                  CORE DAMAGE DETERMINATION REV:
F3-17.1 6
6.0    DISCUSSION 6.1    The approach utilized in this procedure to determine potential or actual core damage is based on the review of key indicators that measure core exit temperatures, reactor vessel water level, containment radiation monitors, and containment hydrogen concentration. These indicators are readily available at the onset of an event.
6.2    As time and conditions permit, these estimates may be compared to the calculated core damage assessment results that are based on measured fission product distribution in the RCS and containment or post accident sampling.
7.0    PREREQUISITES 7.1    An emergency of an Alert, Site Alert, or General Emergency has been declared.
Page 3 of 21
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                          PLANT SAFETY PROCEDURE NUMBER:
F3                  CORE DAMAGE DETERMINATION REV:
F3-17.1 6
8.0    PROCEDURE The plant indicators described in this procedure are available via an ERCS Group Tabular Display for each unit.
NOTE:          See ERCS Main Menu - Tabular Display - Load Group -
1_F3-17-1 or 2_F3-17-1.
8.1    Review the following indicators to estimate core damage.
8.1.1    Containment Hydrogen Concentration:
A. Obtain the containment hydrogen concentration (%).
Within the accuracy of this methodology, it is assumed that recombiners will have an insignificant effect on the NOTE:          hydrogen concentration when it is indicated that extensive zirconium-steam reaction could have occurred.
B. From Figure 1, determine the percentage (%) zirconium water reaction.
C. Table 1 can be used to estimate the extent of core damage estimate.
8.1.2    Core Exit Thermocouple Readings:
For core exit thermocouple temperatures, the maximum NOTE:          temperatures achieved during the event should be used in determining core damage.
A. Obtain as many core exit thermocouple readings as possible for evaluation of core temperature conditions.
If a thermocouple reads greater than 1650F or is reading NOTE:          considerably different than neighboring thermocouples, thermocouple failure should be considered.
B. Compare the thermocouple readings with those in Table 1 to estimate core damage.
Page 4 of 21
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                        PLANT SAFETY PROCEDURE NUMBER:
F3                    CORE DAMAGE DETERMINATION REV:
F3-17.1 6
Radiation Monitors in containment may experience errors NOTE:          during first 4 hours after a DBA LOCA due to thermally induced errors. See Attachment 1 for more information.
8.1.3    Containment Radiation Monitor:
A. Obtain the containment dome monitor readings, R/Hr, from R-48 and/or R-49.
B. From Figure 2 determine the core damage estimate. The exposure rate in Figure 2 is based on the release of only noble gases to the containment. Halogens and other fission products were not considered to be significant contributors to the containment monitor reading.
8.2    All indicators should confirm any core damage estimates. If some indicators do not agree on core damage estimates, then recheck of indications may be performed, or certain indicators may be discounted, based on engineering judgment.
8.3    As time and conditions permit, the TSC nuclear engineer will be conducting a core damage estimate based on post accident sampling.
 
==9.0    REFERENCES==
 
9.1    CAP 1301548, EP Drill-Source of RCS Subcooling Values, (C)
Page 5 of 21
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                PLANT SAFETY PROCEDURE NUMBER:
F3                  CORE DAMAGE DETERMINATION REV:
F3-17.1 6
10.0    ATTACHMENTS 10.1    Table 1 - Characteristics of Categories of Fuel Damage 10.2    Table 2 - Expected Fuel Damage Correlation With Fuel Rod Temperature For Information Only 10.3    Figure 1 - Containment Hydrogen Concentration Based on Zirconium Water Reaction 10.4    Figure 2 - Percent Noble Gases in Containment 10.5    Attachment 1 - Technical Evaluation High Range Containment Radiation Monitors Thermal Induced Current Page 6 of 21
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                                                                    PLANT SAFETY PROCEDURE NUMBER:
F3                                                        CORE DAMAGE DETERMINATION REV:
F3-17.1 6
Table 1        Characteristics of Categories of Fuel Damage Percent (%)
and Type                                        Containment      Core Exit Core                                                of Fission                      Fission        Rad Monitor      Thermocouples  Core              Hydrogen Damage                                              Products                        Product        R/hr 10 hrs      Reading        Uncovery          Monitor Category
* Released                        Ratio ***      after shutdown**  (Deg F)        Indication        (Vol % H2)
No clad damage                                      KR-87      < 1 X 10-3            Not Applicable          --            < 750    No uncovery      Negligible Xe-133    < 1 x 10-3 I-131      < 1 X 10-3 I-133      < 1 X 10-3 0-50% clad damage                                    Kr-87      10 0.01          Kr-87 = 0.022          0 - 50        750 - 1300 Core uncovery        0-6 Xe-133    10 0.1            I-133 = 0.71 I-131      10 0.3 I-133      10 0. 1 50 - 100% clad damage                                Kr-87      0.01 - 0.02          Kr-87 = 0.022        50 to 100      1300 - 1650 Core uncovery      6 - 13 Xe-133    0.1 - 0.2            I-133 = 0.71 I-131      0.3 - 0.5 I-133      0.1 - 0.2 0 - 50% fuel pellet                                  Xe-Kr, Cs, I 1 - 20              Kr-87 = 0.22      100 to 1.15E4        > 1650  Core uncovery      6 - 13 overtemperature                                      Sr-Ba      0 - 0.1              I-133 = 2.1 50-100% fuel pellet                                  Xe-Kr, Cs, I 20 - 40            Kr-87 = 0.22    1.15E4 to 2.3E4      > 1650  Core uncovery      6 - 13 overtemperature                                      Sr-Ba      0.1- 0.2              I-133 = 2.1 0 - 50% fuel melt                                    Xe, Kr, Cs, I 40-70              Kr-87 = 0.22      2.3E4 to 2.7E4      > 1650  Core uncovery      6 - 13 Sr-Ba 0.2 - 0.8                  I-133 = 2.1 Pr 0.1 - 0.8 50 - 100% fuel melt                                  Xe, Kr, Cs, I, Te > 70          Kr-87 = 0.22          > 2.7E4          > 1650  Core uncovery      6 - 13 Sr, Ba > 24                      I-133 = 2.1 Pr > 0.8 Characteristics of Categories of Fuel Damage
* This table is intended to indicate whether there is fuel damage.
**  These values are from Figure 2 and should be revised for times other than 10 hours.
Kr - 87 I - 133 Xe - 133 I - 131 Page 7 of 21
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                PLANT SAFETY PROCEDURE NUMBER:
F3                    CORE DAMAGE DETERMINATION REV:
F3-17.1 6
Table 2      Expected Fuel Damage Correlation With Fuel Rod Temperature For Information Only - See Note Below Page 8 of 21
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                          PLANT SAFETY PROCEDURE NUMBER:
F3                  CORE DAMAGE DETERMINATION REV:
F3-17.1 6
Figure 1      Containment Hydrogen Concentration Based on Zirconium Water Reaction Page 9 of 21
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                PLANT SAFETY PROCEDURE NUMBER:
F3                    CORE DAMAGE DETERMINATION REV:
F3-17.1 6
Figure 2      Percent Noble Gases in Containment Page 10 of 21
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                        PLANT SAFETY PROCEDURE NUMBER:
F3                    CORE DAMAGE DETERMINATION REV:
F3-17.1 6
Attachment 1        Technical Evaluation High Range Containment Radiation Monitors Thermal Induced Current ECR 608000000176 Sections IV and X discussed in this attachment refer to the NOTE:          sections within the ECR. References listed in brackets can be found in the ECR.
1.0    Background From recent NRC Environmental Qualification Component Design Bases Inspections (EQ CDBI) several stations were issued green non-cited violations (NCVs) for not adequately documenting operability of their CHRRMs or to restoring the capability to classify emergency action levels [12, 13, 14]. NRC Information Notice 97-45 Supplement 1 had identified a potential transient operational deficiency on the coaxial signal cables associated with CHRRMs from thermally induced currents (TIC). The potential exists for thermally induced currents on the signal cables associated with these monitors to cause erratic and inaccurate readings.
The TIC phenomena appears to be dependent on the temperature change magnitude, rate of change, the type of cabling used for the plant, the service temperature history, and the uniformity of the cable manufacturing run. TR-112582 [19] notes the sources of TIC include trapped space charge in the insulation of coaxial cables and polar properties of a polysulfone layer used in some cables. The thermal resistance of the cable jacket and the high thermal conductivity of the cable braid create a thermal stimulus at the cable dielectric. The thermal wave resulting from an external temperature change expands the dielectric. If free charges are available in the dielectric, a charge is induced on the braid and center conductor of the cable. The charge is dissipated as current into the impedance of the detector electronics. TIC therefore causes small currents to be induced due to temperature differential between the inner and outer conductors of the coax cable. The currents are too small to impact most instrumentation and control cables, are transient in nature, and are only present during significant temperature transients. The current/dose relationship for Sorrento/GA High Range Radiation Monitoring Systems is shown in TR-112582 [19]
as:
Rad Monitor Cable Current Indicated Dose (nA)
Rate (R/hr)
                                    -2 10                1
                                    -1 10                10 1                102 10                103 102              104 103              105 104              106 Page 11 of 21
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                      PLANT SAFETY PROCEDURE NUMBER:
F3                    CORE DAMAGE DETERMINATION REV:
F3-17.1 6            Technical Evaluation High Range Containment Radiation Monitors Thermal Induced Current ECR 608000000176 Testing performed for Southern California Edison confirmed the magnitude and direction of the spurious signal was a function of the temperature gradient across the cable insulation. The NRC lnformation Notice indicated that the duration of the spurious signal could be as long as 15 minutes, with the worst impact over in approximately 1 minute.
PINGP evaluated the potential impact of TIC on PINGP instrumentation and the only devices impacted at PINGP are the containment high range radiation monitors (CHRRM). From SAP FLOC information, cable lengths inside containment associated with these detectors are 1R-48 (1CMR-1): 80ft, 1R-49 (1CMW-1): 80ft, 2R-48 (2CMR-1): 90ft, and 2R-49 (2CMW-1): 65ft. The PINGP cable is type CBLTP 362 (Rockbestos type RSS-6-104) 2-1/C-22 COAX with a solid #22 AWG conductor. The TIC phenomenon could cause these monitors to go into high alarm during the rapid temperature increases associated with LOCAs and MSLBs inside containment. They could also alarm as failed low due to negative TIC effects during rapid cooling for the same events. CHRRMs are used to estimate post-accident fuel damage which factors into post-accident emergency classifications and in certain EOPs. PINGPs response to this issue was addressed in CHAMPS Condition Reports No. 19980371 [5] and No.
19980809 [6].
A recent review of condition reports no. 19980371 and no. 19980809 determined that no attempt was made to quantify the TIC effects from a PINGP LOCA or MSLB. Also, discussions about whether the CHRRMs meet RG 1.97 requirements with the TIC phenomenon were based on having alternate indications for fuel damage and not expecting to have fuel damage in the first 10-15 minutes of a LOCA. They did not describe how CHRRMs meet RG 1.97 requirements themselves. With recent NRC violations for other sites not adequately addressing the TIC phenomenon in their CHRRMs, a new Passport Action Request 01555551 [7], was written to provide the mechanism to assure a new review of the OE is performed.
Based on the lack of quantitative data of the TIC effect at PINGP for 1R-48/1R-49 &
2R-48/2R-49, Engineering Evaluation ECR 608000000013 was undertaken to determine the site-specific TIC effect of these CHRRMs during a worst-case LOCA and MSLB. Based on the results of this evaluation, CAP 501000001861 was written on 8/22/2017 and 1R-48/1R-49 & 2R-48/2R-49 were declared INOPERABLE by Operations.
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                    PLANT SAFETY PROCEDURE NUMBER:
F3                  CORE DAMAGE DETERMINATION REV:
F3-17.1 6            Technical Evaluation High Range Containment Radiation Monitors Thermal Induced Current ECR 608000000176 Since this time, the NRC has approved PINGPs LAR L-PI-17-006 [20] in NRC SER on March 6, 2018 [21] revising fission product barrier containment radiation EALs for R-48
        & R-49 based on approved calculation GEN-PI-092 [9]. Specifically, the EAL containment radiation levels for a RCS barrier loss increase from 7 to 40 R/hr, a fuel cladding barrier loss increase from 200 to 5,500 R/hr, and a potential containment barrier loss increase from 800 to 23,000 R/hr. This evaluation will consider whether the changes in these EALs along with proposed changes to guidance in associated emergency and operations procedures are adequate to demonstrate that 1R-48/1R-49
        & 2R-48/2R-49 can perform their intended functions when required.
2.0    Analysis - Design Basis Accidents and Quantifying Tic Effects From PINGP accident analyses, the design basis accidents that cause the largest temperature transients in containment that would translate to the largest TIC effects are the Large-Break Loss-Of-Coolant Accident (LOCA) and the Main Steam Line Break (MSLB). From PINGPs LOCA analyses of record [8], the LOCA scenario with the highest peak temperature ramps from 120&deg;F to around 266&deg;F within approximately 15 seconds (See Figure A-1 below). The black line represents a temperature that bounds all analyzed scenarios with margin for EQ Program purposes and is not an actual DBA temperature profile.
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                        PLANT SAFETY PROCEDURE NUMBER:
F3                    CORE DAMAGE DETERMINATION REV:
F3-17.1 6            Technical Evaluation High Range Containment Radiation Monitors Thermal Induced Current ECR 608000000176 Only four of the seven analyzed LOCA profiles extend to the highest LOCA temperature peak that occurs at around 27 minutes into the event. Since all the LOCA temperature profiles have a very similar shape, its not expected that TIC effects would be greatly different between the individual LOCA profiles. However, to account for the uncertainty of not obtaining the highest TIC effect from the shorter profiles, all the longer temperature profiles were run on the TIC software and the maximum difference between TIC effect results were added to the maximum TIC result from each of the longer profiles to obtain the maximum TIC effect result. Also, as previously mentioned, cable lengths have a direct effect on the TIC values and cable lengths differ between each radiation monitor. Therefore, TIC values for each radiation monitor were calculated from the TIC software results and cable lengths to show the variation in readings operators or observers would see during a LOCA.
For a MSLB, containment temperature ramps from 120&deg;F to 311&deg;F in 80 seconds followed by a relatively fast temperature reduction to 245 &deg;F after 800 seconds (13 min.) [10] (See Figure B-1 below).
Although the MSLB temperature profile extends to only 800 seconds post-accident, from steam line break descriptions in USAR Section - 14 [2], no additional temperature spikes are expected from the event. So, the trend should continue as a slow decrease in temperature.
Page 14 of 21
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                  PLANT SAFETY PROCEDURE NUMBER:
F3                  CORE DAMAGE DETERMINATION REV:
F3-17.1 6            Technical Evaluation High Range Containment Radiation Monitors Thermal Induced Current ECR 608000000176 3.0    RESULTS - CHRRM POST-ACCIDENT RESPONSE TO A LOCA TIC software results for the LOCA accident profiles [35] show that during the initial temperature spike that peaks at approximately 15 seconds, the CHRRMs will produce peak TIC errors greater than +2,550 R/hr at around 9 seconds post-LOCA in the radiation monitor most effected by the TIC phenomenon (2R-48) because it has the longest cable length. As containment cools, the TIC errors decrease quickly until they become negative at around 90 seconds. The TIC errors become more negative until about 120 seconds (2 minutes) when they bottom in the -27 R/hr. in 2R-48. TIC errors then trend upward again until they reach zero at around 300 seconds (5 minutes) post-LOCA. Then, as temperatures increase again, a secondary positive TIC error peak of
        +22 R/hr in 2R-48 occurs around 7 minutes post-LOCA. Then, the TIC errors slowly decrease until around 18 minutes when they quickly drop and turn negative until they bottom out around -11 R/hr. in 2R-48 at approximately 20 minutes post-LOCA.
Fluctuation in TIC errors continue in decreasing magnitudes as containment air temperatures change throughout the first 65 minutes of the event. At the end of the 6,000 seconds (100 minutes) post-LOCA simulation, TIC values are slightly negative ranging from -0.49 R/hr on 2R-48 to -0.35 R/hr on 2R-49. The full LOCA TIC response chart for all CHRRMs is shown in Figure D below. Note: The TIC error values used to construct Figure D were calculated from the TIC software output of the LOCA profile that produced the highest peak and lowest trough response in the 5-20 minute timeframe multiplied by 1.26 to account for the 26% maximum variation in the peak magnitudes seen in all four LOCA responses. The reasoning behind using the output with highest TIC error magnitudes from the 5-20 minute timeframe is that it would be the worst case for making an EAL classification decision during an event.
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                  PLANT SAFETY PROCEDURE NUMBER:
F3                    CORE DAMAGE DETERMINATION REV:
F3-17.1 6              Technical Evaluation High Range Containment Radiation Monitors Thermal Induced Current ECR 608000000176 Figure D: Post-LOCA TIC Error for 1R-48, 1R-49, 2R-48, and 2R-49.
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                  PLANT SAFETY PROCEDURE NUMBER:
F3                    CORE DAMAGE DETERMINATION REV:
F3-17.1 6            Technical Evaluation High Range Containment Radiation Monitors Thermal Induced Current ECR 608000000176 4.0    Results - CHRRM Post-Accident Response to a MSLB TIC software results for the MSLB accident profile [34] show that during the initial temperature spike that peaks at approximately 85 seconds, the CHRRMs will produce a peak TIC error of +1,922 R/hr at around 27 seconds post-MSLB in the radiation monitor most effected by the TIC phenomenon (2R-48). As containment cools, the TIC error decreases quickly until it becomes negative at around 117 seconds (1.95 minutes). The TIC error becomes more negative until around 158 seconds (2.6 minutes) when it bottoms out at -144 R/hr for 2R-48. TIC errors then trend less negative as temperature in containment slowly cool. However, at the end of the 800 seconds (13.3 minutes) post-MSLB simulation, TIC values are still negative and range from -7 R/hr on 2R-48 to -5 R/hr on 2R-49. See Figure E below.
Figure E: Post-MSLB TIC Error for 1R-48, 1R-49, 2R-48, and 2R-49.
Page 17 of 21
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                          PLANT SAFETY PROCEDURE NUMBER:
F3                    CORE DAMAGE DETERMINATION REV:
F3-17.1 6              Technical Evaluation High Range Containment Radiation Monitors Thermal Induced Current ECR 608000000176 5.0    Evaluation - TIC Impact on Design and License Bases Functions For Emergency Planning, the Containment High-Range Radiation Monitors are used to determine the condition of fission product barriers during an event. Specifically, they are used to identify a RCS Barrier Loss, a Fuel Clad Barrier Loss, and a Containment Barrier Potential Loss from the Emergency Action Level (EAL) Matrix in the Emergency Plan [24]. The recently approved License Amendment Request [20, 21] raises the EAL for an RCS Barrier Loss from 7 R/hr to 40 R/hr on both R-48 and R-49; a Fuel Clad Barrier Loss from 200 R/hr to 5,500 R/hr, and a potential Containment Barrier Loss from 800 R/hr to 20,000 R/hr. So, this evaluation will use and will only apply when these new EALs that are implemented.
In evaluating the ability of the CHRRMs to perform their functions of detection of significant releases and emergency plan actuation, the time constraint of significance is the requirement to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an EAL has been exceeded.[24] However, in practice, the individual making the event classification needs the information no later than 13 minutes post-accident to document and verify the correct classification in the EAL tables to meet this deadline. The likelihood of an emergency classification being declared in the first 5 minutes of an event is small due to the high demands on the Main Control Room staff at the beginning of an event. So, for this evaluation, the critical timeframe when information from the CHRRMs will be needed to perform their design and license bases functions is considered to be between 5 and 13 minutes post-accident.
For the CHRRMs to function properly during a LOCA or MSLB, their readings need to provide Operations information to correctly classify the event. Radiation readings must be accurate enough that they dont lead to under- or over-classification of the actual event. With this criteria in mind, CHHRM acceptance criteria for emergency plan actuation and release assessment is determined to be the following: Expected radiation level for fission product barrier loss + TIC error falls in the correct EAL range given for the fission product barrier loss during the 5-13 minute post-event timeframe.
Page 18 of 21
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                      PLANT SAFETY PROCEDURE NUMBER:
F3                    CORE DAMAGE DETERMINATION                                    F3-17.1 REV:      6            Technical Evaluation High Range Containment Radiation Monitors Thermal Induced Current ECR 608000000176 By definition, during a LOCA there is a loss of the RCS fission product barrier. Loss of the RCS fission product barrier alone would result in an alert event classification according to the EAL tables. From the current PINGP calculation of record [9], a loss of the RCS barrier without fuel failure would produce radiation levels in containment measured at 43 R/hr on the CHRRMs (not including the TIC error). The next-highest level of classification based on CHRRM readings is site area emergency due to fuel cladding failure. From the calculation of record [9], a 5% fuel cladding damage event (corresponding to a loss of fuel cladding barrier) would produce radiation levels in containment measured at 5,957 R/hr on the CHRRMs (not including the TIC error).
Finally, the calculation of record [9] determined that a 20% fuel cladding damage event (corresponding to a potential loss of containment) would produce radiation levels of 23,830 R/hr. These radiation values will be used as the expected radiation levels in containment during loss of fission product barriers when evaluating the ability of CHRRMs to perform their intended functions of emergency plan actuation and release assessment during the 5-13 minute post-event timeframe.
During a MSLB in containment, the breach should be in only the non-radioactive secondary coolant system. Containment radiation should show no significant increase.
Thus, based solely on CHRRM readings, acceptance criteria for the CHRRMs during an MSLB is indicated radiation levels should be below the RCS barrier loss EAL in the 5-13 minute post-accident timeframe resulting in no event being declared (The classification would be based on other plant indications). A review of MSLB data from the TIC software output [34] indicates that the range of expected TIC errors during the 5-13 minutes post-MSLB timeframe are -4 R/hr at 5 minutes to -52 R/hr at 7 minutes post-accident. A summary of expected radiation readings from the CHRRMS during MSLBs and LOCAs with the corresponding acceptance criteria is presented in Table 1 below.
One additional item of note is TR-112582 mentions the Sorrento/GA detector has a 10-11 amp normal output or keep alive signal current. If this signal is lost, the system issues a fail alarm to the operator. Therefore, if a negative TIC error signal of 10-11 amp or more occurs, the keep alive signal will be negated and a fail alarm will be generated. The following table compiles all the previously described TIC software results, information, acceptance criteria, and required time frames to summarize the ability of Operations to correctly classify the LOCA and MSLB events based on CHRRMs readings indications:
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                            PLANT SAFETY PROCEDURE NUMBER:
F3                      CORE DAMAGE DETERMINATION REV:
F3-17.1 6                Technical Evaluation High Range Containment Radiation Monitors Thermal Induced Current ECR 608000000176 Event              Expected      TIC Error 5-  CHRRMs            Acceptance      Acceptance Criteria Radiation    13 min. (R/hr) Readings 5 -13    Criteria (R/hr) Met?
Level (R/hr)                min. (R/hr)
(Expected + TIC)
LOCA              43            0 to 22        43 to 65          40 to 5,500    Yes.
LOCA              5,957        0 to 22        5,957 to 5,979    5,500 to 20,000 Yes w/Cladding Failure LOCA              23,830        0 to 22        23,830 to 23,852  > 20,000        Yes w/potential loss of containment MSLB              Normal area  -4 to -52      Failed low      < 40            Yes, with additional value <1                                                      guidance to address Failed low Table 1: Comparison of CHRRM Indication vs. Acceptance Criteria Table 1 shows that TIC errors during the 5 to 13 minutes period post-LOCA (without fuel cladding damage) are expected to produce readings on all four CHRRMs in the range of 43 to 65 R/hr. This range of radiation readings falls within the new EAL range of >40 and <5,500 R/hr corresponding to a RCS barrier loss. So, this meets the established acceptance criteria. For the case of a LOCA with fuel cladding failure, Table 1 shows CHRRMs would produce readings between 5,957 and 5,979 R/hr.
These values also fall in the corresponding EAL acceptance criteria range for fuel cladding failure of 5,501 to 20,000 R/hr. For the case of potential loss of containment barrier, Table 1 shows CHRRMs would produce readings between 23,830 and 23,852 R/hr. These readings are above the EAL level of 20,000 R/hr that would indicate a potential loss of containment barrier. So, CHRRMs readings during all potential loss of fission product barriers due to LOCAs meet the acceptance criteria for emergency plan actuation and release assessment during the 5-13 minute post-accident timeframe.
For the case of an MSLB, Table 1 shows that TIC errors during the 5 to 13 minutes are expected to be of sufficient magnitude to negate the detector signal from normal containment radiation and the keep alive signal, producing a fail low alarm according to the detector logic diagram [29]. If appropriate guidance were provided in procedures and documents identified in Section IV that failed low alarms were a normal response to a cooldown event in containment for these detectors and were not failed detectors, operators and/or the emergency director could correctly conclude the fission product barriers were intact and would not classify the event based on the CHRRM readings.
Therefore, CHRRMs acceptance criteria are also met for emergency plan actuation and release assessment during the 5-13 minute post-event timeframe post-MSLB.
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                      PLANT SAFETY PROCEDURE NUMBER:
F3                    CORE DAMAGE DETERMINATION REV:
F3-17.1 6            Technical Evaluation High Range Containment Radiation Monitors Thermal Induced Current ECR 608000000176 Assessing the ability of CHRRMs to perform its remaining intended function, long-term surveillance, it can be seen in both LOCA and MSLB TIC Error charts the TIC errors decrease as stored energy from the accident has fully released and temperatures stabilize in containment. TIC errors will trend to zero with no additional releases of energy and CHRRMs will provide accurate measurement of actual radiation conditions in containment as time passes. Therefore, it is concluded that CHRRMs are able to perform their design and license bases function of long-term surveillance of radiation levels in containment during accident conditions.
6.0    Conclusions and Recommendations Based on this evaluation with the new containment radiation EALs in place, enhanced procedural guidance on the TIC effect specified in Section X., and associated training of Operations, RP, and Emergency Response performed, the CHRRMs are able to perform their license and design bases functions. This evaluation shows that five minutes after either a LOCA or a MSLB Operations will be able to use CHRRMs measurements to detect significant radiation releases and actuate the emergency plan (Classify events) in the timeframe necessary (5-13 minutes). If personnel question alarms or data indicating high readings in the first five minutes after an event or failed low readings, there will be appropriate guidance in the alarm response or emergency procedure to assure them that this is expected after a thermal transient. Also, CHRRMs will be accurate enough after five minutes for personnel to monitor containment radiation levels in the long-term. The accuracy of CHRRMs five minutes after LOCAs and MSLBs meets the acceptance criteria for all its intended functions with enhanced procedural guidance and training in place.
One recommendation is to ensure all the conditions of this evaluation are met by creating a tracking action to verify all required procedure changes and training from Section X have been completed before returning 1R-48/1R-49 & 2R-48/2R-49 to operable.
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                PLANT SAFETY PROCEDURE NUMBER:
F3                    CORE DAMAGE ASSESSMENT REV:
F3-17 16 REFERENCE USE
* Procedure should be at the work location.
* Procedure segments may be performed from memory.
* Use the procedure to verify segments have been completed.
* When required, sign or initial appropriate blocks to certify that all segments are complete.
APPROVAL:
PCR #: 602000015507 Page 1 of 52
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                    PLANT SAFETY PROCEDURE NUMBER:
F3                      CORE DAMAGE ASSESSMENT REV:
F3-17 16 1.0    PURPOSE 1.1    The purpose of this procedure is to provide a means to best estimate the degree of reactor core damage from the measured fission product concentrations in water and gas samples taken for the primary system and containment under accident conditions.
2.0    APPLICABILITY 2.1    This procedure SHALL be used once sample activity data is available through step 8.1; otherwise, use F3-17.1 for core damage indication.
3.0    PRECAUTIONS 3.1    The numbers obtained using this procedure, are at best, estimates only.
3.2    When making core damage calculations as per this procedure, considerations should be given to other plant indicators, for example:
3.2.1      Incore Thermocouples.
3.2.2      Containment Radiation Monitors (R48/49).
3.2.3      Hydrogen Concentration in the Containment Atmosphere 3.3    Spiking may occur after a shutdown or significant power change, usually during the 2 to 6 hour period following the power change. Iodine spiking is a characteristic of the condition where an increase in the normal primary coolant activity is noted, but no damage to the cladding has occurred.
3.4    Keep all generated documents in accordance with F3-24, Record Keeping During an Emergency.
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                        PLANT SAFETY PROCEDURE NUMBER:
F3                      CORE DAMAGE ASSESSMENT REV:
F3-17 16 4.0    SPECIAL CONSIDERATIONS 4.1    The fuel pellet inventories listed in Table 2 are based on a 1650 MWth 2-loop Westinghouse plant. The Core Damage Assessment software, CDA, and Section 8.12 of this procedure for a manual calculation, scale these values to the current maximum licensed power level of 1677 MWth.
5.0    RESPONSIBILITIES 5.1    Persons qualified under PI-BEP-ERO-029 are responsible to estimate the degree of reactor core damage according to the guidance provided in this procedure. Persons not qualified SHALL assess damage per F3-17.1.
6.0    DISCUSSION 6.1    The approach utilized in this methodology of core damage assessment is measurement of fission product concentrations in the primary coolant system, and containment, when applicable, utilizing the post accident sampling system.
6.2    Certain nuclides have been selected to be associated with each particular core damage state, i.e., clad damage, fuel overheat and fuel melt. These nuclides reach equilibrium quickly within the fuel cycle. Once equilibrium conditions are reached, a fixed inventory of the nuclides is assumed to exist within the fuel pellet. For these nuclides which reach equilibrium, their relative ratios within the fuel pellet can also be considered to be constant. During operation, certain volatile fission products collect in the gap. The relative ratios in the gap can also be considered to be constant, however, the distribution of the nuclides in the gap is not in the same proportion as the fuel pellet inventory since the migration of each nuclide into the gap is dependent on its particular diffusion rate. The relative ratios of the nuclides analyzed during an accident may be compared to the predicted relative ratios existing in the gap and fuel pellet to determine the source of the fission product release, i.e., gap release or fuel pellet.
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                      PLANT SAFETY PROCEDURE NUMBER:
F3                        CORE DAMAGE ASSESSMENT REV:
F3-17 16 6.3    Clad damage is characterized by the release of these fission products, i.e.,
isotopes of the noble gases, iodine, and cesium which have accumulated in the gap and during the operation of the plant. When the cladding ruptures, it is assumed that the fission product gap inventory of the damaged fuel rods is instantaneously released to the primary system. For this methodology it is assumed that the noble gases will escape through the break of the primary system boundary to the containment atmosphere and the iodines will stay in solution and travel with primary system water during the accident.
6.4    Fission product release associated with overtemperature fuel conditions arises initially from the portion of the noble gas, cesium and iodine inventories that was previously accumulated in grain boundaries. In addition, small amounts of the more refractory elements, barium-lanthanum, and strontium are also released.
6.5    Fuel pellet melting leads to rapid release of many noble gases, halides, and cesiums remaining in the fuel after overheat conditions. Significant release of the strontium, barium-lanthanum chemical groups is perhaps the most distinguishing feature of melt release conditions.
6.6    Auxiliary indicators such as core exit thermocouples, reactor vessel water level, containment radiation monitors, and the containment hydrogen concentration are available for estimating core damage. These indications should confirm the core damage estimates which in turn are based on the radionuclide analysis.
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                        PLANT SAFETY PROCEDURE NUMBER:
F3                      CORE DAMAGE ASSESSMENT REV:
F3-17 16 7.0    PREREQUISITES 7.1    An emergency of an Alert, Site Alert, or General Emergency has been declared.
8.0    PROCEDURE The CDA (Core Damage Assessment) program may be used NOTE:          whenever core damage estimates are desired.
8.1    Request the Radiation Protection Group to obtain the applicable samples to enable an adequate assessment of core damage. See Table 1 for suggested sampling locations.
Monitor via LAN ERCS if available; otherwise, obtain printouts of ERCS screens every hour from ERCS terminal in NOTE:          TSC or control room. Monitoring frequency may be increased/decreased per evaluators discretion. Values for CDA should be taken at the time of the sample.
8.2    Obtain the following plant data at the approximate sample time:
8.2.1    Incore Thermocouple Map
* ERCS code TC and TC1
* IF ERCS is unavailable, THEN request the information from the control room ICCM panel via the three-way communicator.
* IF ICCM is unavailable, THEN contact I&C to request manual readings.
8.2.2    Containment Pressure
* ERCS point 1U5015A[2U5015A] available on SAS containment panel (code:XT14)
* IF ERCS is unavailable, THEN request the information from the control room via the three-way communicator.
8.2.3    Containment Temperature
* ERCS point 1U5013A[2U5013A]
* IF ERCS is unavailable, THEN request the information from the control room via the three-way communicator.
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                PLANT SAFETY PROCEDURE NUMBER:
F3                    CORE DAMAGE ASSESSMENT REV:
F3-17 16 8.2.4    Containment Hydrogen Concentration
* ERCS point 1U5021A[2U5021A] available on the SAS containment panel (code:XT14)
* IF ERCS is unavailable, THEN request the information from the control room via the three-way communicator.
8.2.5    Containment Radiation Level
* ERCS point 1U5022A[2U5022A] available on the SAS containment panel (code:XT14)
* IF ERCS is unavailable, THEN request the information from the control room via the three-way communicator.
8.2.6    Containment Sump Level
* ERCS point 1U5017A[2U5017A] available on the SAS normal operation panel (code: XS11)
* IF ERCS is unavailable, THEN request the information from the control room via the three-way communicator.
8.2.7    Containment Sump Temperature
* When considering the containment sump liquid mass, assume containment sump temperature is less than 200 F. In the event of a large break LOCA, USAR Appendix K, Figure K-15 (upper curve) can be used to estimate the sump temperature.
8.2.8    RVLIS Level
* ERCS point 1U5011A[2U5011A] available on the SAS Primary Core Cooling Panel (code: XT12)
* IF ERCS is unavailable, THEN request the information from the control room via the three-way communicator.
Page 6 of 52
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                          PLANT SAFETY PROCEDURE NUMBER:
F3                      CORE DAMAGE ASSESSMENT REV:
F3-17 16 8.2.9    RCS Temperature
* ERCS panel RCS temperature (code: XT21). Use the lowest available water temperature.
* IF ERCS is unavailable, THEN request the information from the control room via the three-way communicator.
8.3    Perform a core damage assessment according to the instructions in SWI NE-5 (23). Continue with Step 8.15 of this procedure when the CDA run is complete.
If the computer is not available, perform the following NOTE:          manual calculations to obtain core damage estimates.
8.4    Decay correct the specific activities determined by the sample analysis, back to the time of reactor shutdown, as follows:
The decay correction may have been accomplished by the NOTE:          computer during the spectrum analysis. Therefore, this step may not need to be completed.
A A0 =                - i t e
Where:
A    =        measured specific activity, Ci/gm or Ci/cc i  =        decay constant of isotope i, sec-1; where i = 0.693/(half-life)i t    =        time elapsed from reactor shutdown to time of sampling, sec.
A0 =          decay corrected specific activity Ci/gm or Ci/cc Page 7 of 52
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                                    PLANT SAFETY PROCEDURE NUMBER:
F3                    CORE DAMAGE ASSESSMENT REV:
F3-17 16 8.5    If a parent-daughter relationship exists for a specific isotope, the following steps should be followed to calculate the fraction of the measured activity due to the decay of the daughter that was released and then to calculate the activity of the daughter released at shutdown.
8.5.1    Calculate the hypothetical daughter concentration (QB) at the time of the sample analysis assuming 100 percent release of the parent and daughter source inventory:
QB (t)    =      Ki B
B  Ai
(                    )
Qo AI e  Ai t  e  B t + Q0B e-  B t Where:
Q0Ai      =      100% source inventory (Ci) of parent i, Table 2 or Table 4.
Q 0B      =      100% source inventory (Ci) of daughter, Table 2 or Table 4.
QB (t)    =      hypothetical daughter activity (Ci) at sample time.
Ki        =      if parent has 2 daughters, Ki is the branching factor, Table 3.
Ai      =      decay constant of parent i, sec -1; where Ai = 0.693/(half-life)Ai B        =      daughter decay constant, sec -1; where B = 0.693/(half-life)B t        =      time period from shutdown to time sample, sec.
8.5.2    Determine the contribution of only the decay of the initial inventory of the daughter to the hypothetical daughter activity at sample time:
Q 0B e - B t Fr    =
QB (t) 8.5.3    Calculate the amount of decay corrected sample specific activity associated with just the daughter that was released.
M 0B  =    Fr X A0 Where: A0 =      decay corrected specific activity (Ci/gm or Ci/cc) as determined by the analysis.
Page 8 of 52
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                        PLANT SAFETY PROCEDURE NUMBER:
F3                    CORE DAMAGE ASSESSMENT REV:
F3-17 16 8.6    Determine the total volume or mass of the medium which was sampled.
8.6.1    Containment Volume:
V  =      containment free volume (ccs)
                            =      3.74 X 1010 ccs When considering the containment sump liquid mass, assume containment sump temperature is less than 200F.
NOTE:          In the event of a large break LOCA, USAR Appendix K, Figure K-15 (upper curve) can be used to estimate the sump temperature.
8.6.2    Liquid Mass:
A. Liquid temperature < 200F 28 .3 X 10 3 cc Mass (gms)      =    volume (ft3) X STP X ft 3 Where: STP =          water density at STP = 1.0 gm/cc B. Liquid temperature > 200F 28 .3 X 10 3 cc Mass (gms)          =                  3 volume (ft ) X      (2) X STP X STP                      ft 3 Where:            (2)    =    water density ratio at medium STP temperature, from Figure 1 STP          =    water density at STP = 1.0 gm/cc Page 9 of 52
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                        PLANT SAFETY PROCEDURE NUMBER:
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F3-17 16 8.7    Determine the total activity of each isotope in each medium.
8.7.1    Containment Atmosphere:
Curie Total containment      =    A0 (Ci/cc) X V (ccs) X 1 X 106 Ci Activity (curies)
Where:      A0      =    Specific activity of containment atmosphere (Ci/cc), decay corrected to time of reactor shutdown and temperature/pressure corrected.
V        =    containment free volume (ccs)
                                              =    3.74 X 1010 ccs 8.7.2    Liquid Sample:
Curie Total Liquid      =      Liquid          x A0 (Ci/cc)    x 1 X 106 Ci Activity (Curies)        MASS (gms)
Where: A0      =    Specific activity of liquid sample (Ci/gm), decay corrected to time of reactor shutdown.
8.8    The approximate total activity of each isotope in the liquid samples can now be calculated.
Total Water Activity    =    RCS Activity + Sump Activity + Activity Leaked to Secondary System.
8.9    Now the total activity of each isotope released at the time of the accident can be determined:
Total Activity  =    Total Water +        Containment Released              Activity      Atmosphere Activity Page 10 of 52
 
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F3-17 16 8.10    Utilizing the total activity of each isotope released, calculate the activity ratios of the released fission products.
Noble Gas Activity 8.10.1    Noble Gas Ratio        =
Xe  133 Activity Iodine Activity 8.10.2    Iodine Ratio      =
I - 131 Activity Steady state power conditions may be assumed where NOTE:            power does not vary by more than  10% from the time averaged value.
8.11    Determine the power history prior to reactor shutdown.
8.12    Using the power history, determine a power correction factor for each isotope, in accordance with the following guidelines:
Steady state power condition is assumed where the power NOTE:            does not vary by more than  10% from the time averaged value.
8.12.1    Steady State power prior to shutdown.
A. Half-life of nuclide < 1 day Average Power Level (Mwt) for Prior 4 Days Power Correction Factor =
Power Level (Mwt) used in Table 2 B. Half-life of nuclide > 1 day Average Power Level (Mwt) for Prior 30 Days Power Correction Factor =
Power Level (Mwt) used in Table 2 C. Half-life of nuclide ~ 1 year Average Power Level (Mwt) for Prior 1 year Power Correction Factor =
Power Level (Mwt) used in Table 2 Page 11 of 52
 
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F3-17 16 8.12.2  Transient power history in which the power has not remained constant prior to reactor shutdown.
For the majority of the selected nuclides, the 30-day power NOTE:          history prior to shutdown is sufficient to calculate a power correction factor.
A. Power Correction Factor          =
(
j Pj 1 - e
                                                                                    - i t j
                                                                                              )e - i t 0 j RP Pj    =    average power level (Mwt) during operating period tj RP    =    power level (Mwt) used in Table 2 tj    =    operating period in days at power Pj where power does not vary more than 10 percent power from the time averaged value (Pj).
I    =    decay constant of nuclide I in inverse days.
where I = 0.693/(half-life)I T0j    =    time between end of period j and time of reactor shutdown in days.
B. For the few nuclides with half-lives around one year or longer, a power correction factor which ratios effective full power days to total calendar days of cycle operation is applied.
Actual Operating EFPD of equilibrium cycle Power Correction Factor = Total expected EFPD of equilibrium cycle operation Where: Equilibrium Cycle = three (3) cycles of core operation 8.12.3  For Cs-134, Figure 2 is used to determine the power correction factor.
To use Figure 2, the average power during the entire operating period is required.
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                        PLANT SAFETY PROCEDURE NUMBER:
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F3-17 16 8.13    The total inventory of fission products available for release at reactor shutdown are calculated by applying the power correction factors to the equilibrium, end-of-life core inventories.
Equilibriu m Inventory at          Power Corrected Inventory =          end  of  life (Ci)    X    Correction (Table 2)                  Factor 8.14    Determine the percentage of inventory released, for each isotope.
Total Activity Released (Ci)
Release            =                                    X 100 Corrected Inventory (Ci)
Percentage (%)
8.15    The results of radionuclide analysis may now be used to determine an estimate of the extent of core damage.
8.15.1    From Figure 3 thru 15, estimate the extent of core damage by categorizing the percentage of clad damage, fuel over-temperature, and fuel melt.
8.15.2    Compare the calculated activity ratios with those listed in Table 5.
Measured relative ratios greater than the gap activity ratios listed in Table 5 are indicative of more severe failures, e.g., fuel overheat.
Page 13 of 52
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                          PLANT SAFETY PROCEDURE NUMBER:
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F3-17 16 The plant indicators described in this procedure are available via an ERCS Group Tabular Display for each unit.
NOTE:          See ERCS Main Menu - Tabular Display - Load Group -
1_F3-17-1 or 2_F3-17-1.
8.16    To verify the conclusion of the radionuclide analysis, other indicators should now be used to provide verification of the estimate of core damage.
8.16.1  Containment Hydrogen Concentration:
A. Obtain the containment hydrogen concentration (%).
Within the accuracy of this methodology, it is assumed that recombiners will have an insignificant effect on the NOTE:          hydrogen concentration when it is indicated that extensive zirconium-steam reaction could have occurred.
B. From Figure 16, determine the percentage (%) zirconium water reaction.
C. Table 6 can be used to validate the extent of core damage estimate.
8.16.2  Core Exit thermocouple Readings:
For core exit thermocouple temperatures, the maximum NOTE:          temperatures achieved during the event should be used in determining core damage.
A. Obtain as many core exit thermocouple readings as possible for evaluation of core temperature conditions.
If a thermocouple reads greater than 1650F or is reading NOTE:          considerably different than neighboring thermocouples, thermocouple failure should be considered.
B. Compare the thermocouple readings with those in Table 6 to confirm the core damage estimate.
Page 14 of 52
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                        PLANT SAFETY PROCEDURE NUMBER:
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F3-17 16 Radiation Monitors in containment may experience errors NOTE:          during first 4 hours after a DBA LOCA due to thermally induced errors. See Attachment 1 for more information.
8.16.3    Containment Radiation Monitor:
A. Obtain the containment dome monitor readings, in R/Hr, from R-48 and/ or R-49.
B. From Figure 17, verify core damage estimate. The exposure rate in Figure 17 is based on the release of only noble gases to the containment. Halogens and other fission products were not considered to be significant contributors to the containment monitor reading.
8.17    All indicators should confirm any core damage estimates. If radio-nuclide analysis and auxiliary indicators do not agree on core damage estimates, then recheck of indications may be performed, or certain indicators may be discounted, based on engineering judgment.
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                  PLANT SAFETY PROCEDURE NUMBER:
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F3-17 16
 
==9.0    REFERENCES==
 
NONE 10.0 ATTACHMENTS 10.1    Table 1 - Suggested Sampling Locations 10.2    Table 2 - Fuel Pellet Inventory 10.3    Table 3 - Parent-Daughter Relationships 10.4    Table 4 - Source Inventory of Related Parent Nuclides 10.5    Table 5 - Isotopic Activity Ratios of Fuel Pellet and Gap 10.6    Table 6 - Characteristics of Categories of Fuel Damage 10.7    Table 7 - Expected Fuel Damage Correlation With Fuel Rod 10.8    Figure 1 - Water Density Ratio (Temperature vs. STP) 10.9    Figure 2 - Power Correction Factor For Cs-134 Based on Average Power During Operation 10.10 Figure 3 - Relationship of % Clad Damage With % Core Inventory Released of Xe-133 10.11 Figure 4 - Relationship of % Clad Damage With % Core Inventory Released of I-131 10.12 Figure 5 - Relationship of % Clad Damage With % Core Inventory Released of I-131 W/Spiking 10.13 Figure 6 - Relationship of % Clad Damage With % Core Inventory Released of Kr-87 10.14 Figure 7 - Relationship of % Clad Damage With % Core Inventory Released of Xe-131M 10.15 Figure 8 - Relationship of % Clad Damage With % Core Inventory Released of I-132 10.16 Figure 9 - Relationship of % Clad Damage With % Core Inventory Released of I-133 Page 16 of 52
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                              PLANT SAFETY PROCEDURE NUMBER:
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F3-17 16 10.17 Figure 10 - Relationship of % Clad Damage With % Core Inventory Released of I-135 10.18 Figure 11 - Relationship of % Fuel Over Temperature With % Core Inventory Released of Xe, Kr, I, or Cs 10.19 Figure 12 - Relationship of % Fuel Over Temperature With % Core Inventory Released of Ba or Sr 10.20 Figure 13 - Relationship of % Fuel Melt With % Core Inventory Released of Xe, Kr, I, Cs or Te 10.21 Figure 14 - Relationship of % Fuel Melt With % Core Inventory Released of Ba or Sr 10.22 Figure 15 - Relationship of % Fuel Melt With % Core Inventory Released of Pr 10.23 Figure 16 - Containment Hydrogen Concentration Based on Zirconium Water Reaction 10.24 Figure 17 - Percent Noble Gases in Containment 10.25 Attachment 1 - Thermally Induced Current Errors in Containment Radiation Monitors Page 17 of 52
 
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F3-17 16 Table 1    Suggested Sampling Locations Page 18 of 52
 
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F3-17 16 Table 2        Fuel Pellet Inventory Fuel Pellet Inventory*
Nuclide                  Half Life                Inventory Curies**
Kr 85m                      4.4 h                      1.0 x 107 Kr 87                      76 m                      1.85 x 107 Kr 88                      2.8 h                    2.69 x 107 Xe 131m                    11.8 d                    2.94 x 105 Xe 133                    5.27 d                    9.26 x 107 Xe 133m                    2.26 d                    1.35 x 107 Xe 135                    9.14 h                    1.77 x 107 I 131                      8.05 d                    4.54 x 107 I 132                      2.26 h                    6.65 x 107 I 133                      20.3 h                    9.26 x 107 I 135                      6.68 h                    8.33 x 107 Rb 88                      17.8 m                    2.69 x 107 Cs 134                      2 yr                    1.09 x 107 Cs 137                      30 yr                    4.96 x 106 Te 129                    68.7 m                    1.51 x 107 Te 132                    77.7 h                    6.65 x 107 Sr 89                      52.7 d                    3.70 x 107 Sr 90                      28 yr                    3.36 x 106 Ba 140                    12.8 d                    7.91 x 107 La 140                    40.22 h                    8.33 x 107 La 142                    92.5 m                    7.07 x 107 Pr 144                    17.27 m                    5.81 x 107
: 1. Inventory based on ORIGEN run for equilibrium, end-of-life core.
            ** Westingouse, 2-Loop 1650 Mwt Plant Page 19 of 52
 
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F3-17 16 Table 3    Parent-Daughter Relationships Page 20 of 52
 
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F3-17 16 Table 4        Source Inventory of Related Parent Nuclides Page 21 of 52
 
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F3-17 16 Table 5      Isotopic Activity Ratios of Fuel Pellet and Gap Page 22 of 52
 
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F3                                                            CORE DAMAGE ASSESSMENT REV:
F3-17 16 Table 6        Characteristics of Categories of Fuel Damage Containment Percent                                        Radiogas and Type                                        Monitor          Core Exit Core                                                of Fission                        Fission      R/hr 10 hrs      Thermocouples  Core              Hydrogen Damage                                              Products                          Product      after            Reading        Uncovery          Monitor Category                                            Released                          Ratio***      shutdown**        (Deg F)        Indication        (Vol % H2)
No clad damage                                      Kr-87      < 1 X 10-3            No Applicable          --            < 750    No uncovery      Negligible Xe-133    < 1 x 10-3 I-131      < 1 X 10-3 I-133      < 1 X 10-3 0-50% clad damage                                    Kr-87      10 0.01            Kr-87 = 0.022        0 - 50        750 - 1300 Core uncovery        0-6 Xe-133    10 0.1 I-131      10 0.3            I-133 = 0.71 I-133      10 0. 1 50 - 100% clad damage                                Kr-87      0.01 - 0.02            Kr-87 = 0.022      50 to 100      1300 - 1650 Core uncovery      6 - 13 Xe-133    0.1 - 0.2 I-131      0.3 - 0.5              I-133 = 0.71 I-133      0.1 - 0.2 0 - 50% fuel pellet                                  Xe-Kr, Cs, I                      Kr-87 = 0.22    100 to 1.15E4        > 1650  Core uncovery      6 - 13 overtemperature                                      1 - 20 Sr-Ba      0 - 0.1                I-133 = 2.1 50-100% fuel pellet                                  Xe-Kr, Cs, I                      Kr-87 = 0.22    1.15E4 to 2.3E4      > 1650  Core uncovery      6 - 13 overtemperature                                      20 - 40 Sr-Ba      0.1- 0.2              I-133 = 2.1 0 - 50% fuel melt                                    Xe, Kr, Cs, I 40-70              Kr-87 = 0.22    2.3E4 to 2.7E4      > 1650  Core uncovery      6 - 13 Sr-Ba 0.2 - 0.8 Pr 0.1 - 0.8                      I-133 = 2.1 50 - 100% fuel melt                                  Xe, Kr, Cs, I, Te                Kr-87 = 0.22        > 2.7E4          > 1650  Core uncovery      6 - 13
                                                        > 70 Sr, Ba > 24                      I-133 = 2.1 Pr > 0.8 Characteristics of Categories of Fuel Damage*
* This table is intended to indicate whether there is fuel damage.
**  These values are from Figure 17 and should be revised for times other than 10 hours.
Kr - 87 I - 133 xe - 133 I - 131 Page 23 of 52
 
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F3-17 16 Table 7      Expected Fuel Damage Correlation With Fuel Rod Temperature For Information Only - See Note Below Page 24 of 52
 
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F3-17 16 Figure 1 Water Density Ratio (Temperature vs. STP)
Page 25 of 52
 
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F3-17 16 Figure 2      Power Correction Factor For Cs-134 Based on Average Power During Operation Page 26 of 52
 
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F3-17 16 Figure 3      Relationship of % Clad Damage With % Core Inventory Released of Xe-133 Page 27 of 52
 
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F3-17 16 Figure 4 Relationship of % Clad Damage With % Core Inventory Released of I-131 Page 28 of 52
 
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F3-17 16 Figure 5 Relationship of % Clad Damage With % Core Inventory Released of I-131 W/Spiking Page 29 of 52
 
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F3-17 16 Figure 6      Relationship of % Clad Damage With % Core Inventory Released of Kr-87 Page 30 of 52
 
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F3-17 16 Figure 7 Relationship of % Clad Damage With % Core Inventory Released of Xe-131M Page 31 of 52
 
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F3-17 16 Figure 8 Relationship of % Clad Damage With % Core Inventory Released of I-132 Page 32 of 52
 
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F3-17 16 Figure 9 Relationship of % Clad Damage With % Core Inventory Released of I-133 Page 33 of 52
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                          PLANT SAFETY PROCEDURE NUMBER:
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F3-17 16 Figure 10 Relationship of % Clad Damage With % Core Inventory Released of I-135 Page 34 of 52
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                              PLANT SAFETY PROCEDURE NUMBER:
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F3-17 16 Figure 11 Relationship of % Fuel Over Temperature With % Core Inventory Released of Xe, Kr, I, or Cs Page 35 of 52
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                            PLANT SAFETY PROCEDURE NUMBER:
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F3-17 16 Figure 12 Relationship of % Fuel Over Temperature With % Core Inventory Released of Ba or Sr Page 36 of 52
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                              PLANT SAFETY PROCEDURE NUMBER:
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F3-17 16 Figure 13 Relationship of % Fuel Melt With % Core Inventory Released of Xe, Kr, I, Cs or Te Page 37 of 52
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                            PLANT SAFETY PROCEDURE NUMBER:
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F3-17 16 Figure 14 Relationship of % Fuel Melt With % Core Inventory Released of Ba or Sr Page 38 of 52
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                              PLANT SAFETY PROCEDURE NUMBER:
F3                    CORE DAMAGE ASSESSMENT REV:
F3-17 16 Figure 15 Relationship of % Fuel Melt With % Core Inventory Released of Pr Page 39 of 52
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                          PLANT SAFETY PROCEDURE NUMBER:
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F3-17 16 Figure 16 Containment Hydrogen Concentration Based on Zirconium Water Reaction Page 40 of 52
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                            PLANT SAFETY PROCEDURE NUMBER:
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F3-17 16 Figure 17 Percent Noble Gases in Containment Page 41 of 52
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                        PLANT SAFETY PROCEDURE NUMBER:
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F3-17 16 Attachment 1        Technical Evaluation High Range Containment Radiation Monitors Thermal Induced Current ECR 608000000176 Sections IV and X discussed in this attachment refer to the NOTE:          sections within the ECR. References listed in brackets can be found in the ECR.
1.0    Background From recent NRC Environmental Qualification Component Design Bases Inspections (EQ CDBI) several stations were issued green non-cited violations (NCVs) for not adequately documenting operability of their CHRRMs or to restoring the capability to classify emergency action levels [12, 13, 14]. NRC Information Notice 97-45 Supplement 1 had identified a potential transient operational deficiency on the coaxial signal cables associated with CHRRMs from thermally induced currents (TIC). The potential exists for thermally induced currents on the signal cables associated with these monitors to cause erratic and inaccurate readings.
The TIC phenomena appears to be dependent on the temperature change magnitude, rate of change, the type of cabling used for the plant, the service temperature history, and the uniformity of the cable manufacturing run. TR-112582
[19] notes the sources of TIC include trapped space charge in the insulation of coaxial cables and polar properties of a polysulfone layer used in some cables. The thermal resistance of the cable jacket and the high thermal conductivity of the cable braid create a thermal stimulus at the cable dielectric. The thermal wave resulting from an external temperature change expands the dielectric. If free charges are available in the dielectric, a charge is induced on the braid and center conductor of the cable. The charge is dissipated as current into the impedance of the detector electronics. TIC therefore causes small currents to be induced due to temperature differential between the inner and outer conductors of the coax cable. The currents are too small to impact most instrumentation and control cables, are transient in nature, and are only present during significant temperature transients. The current/dose relationship for Sorrento/GA High Range Radiation Monitoring Systems is shown in TR-112582 [19]
as:
Rad Monitor Cable Current Indicated Dose (nA)
Rate (R/hr)
                                      -2 10                1
                                      -1 10                10 1                102 10                103 102              104 103              105 104              106 Page 42 of 52
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                  PLANT SAFETY PROCEDURE NUMBER:
F3                      CORE DAMAGE ASSESSMENT REV:
F3-17 16            Technical Evaluation High Range Containment Radiation Monitors Thermal Induced Current ECR 608000000176 Testing performed for Southern California Edison confirmed the magnitude and direction of the spurious signal was a function of the temperature gradient across the cable insulation. The NRC lnformation Notice indicated that the duration of the spurious signal could be as long as 15 minutes, with the worst impact over in approximately 1 minute.
PINGP evaluated the potential impact of TIC on PINGP instrumentation and the only devices impacted at PINGP are the containment high range radiation monitors (CHRRM). From SAP FLOC information, cable lengths inside containment associated with these detectors are 1R-48 (1CMR-1): 80ft, 1R-49 (1CMW-1): 80ft, 2R-48 (2CMR-1): 90ft, and 2R-49 (2CMW-1): 65ft. The PINGP cable is type CBLTP 362 (Rockbestos type RSS-6-104) 2-1/C-22 COAX with a solid #22 AWG conductor. The TIC phenomenon could cause these monitors to go into high alarm during the rapid temperature increases associated with LOCAs and MSLBs inside containment. They could also alarm as failed low due to negative TIC effects during rapid cooling for the same events. CHRRMs are used to estimate post-accident fuel damage which factors into post-accident emergency classifications and in certain EOPs. PINGPs response to this issue was addressed in CHAMPS Condition Reports No. 19980371 [5] and No.
19980809 [6].
A recent review of condition reports no. 19980371 and no. 19980809 determined that no attempt was made to quantify the TIC effects from a PINGP LOCA or MSLB. Also, discussions about whether the CHRRMs meet RG 1.97 requirements with the TIC phenomenon were based on having alternate indications for fuel damage and not expecting to have fuel damage in the first 10-15 minutes of a LOCA. They did not describe how CHRRMs meet RG 1.97 requirements themselves. With recent NRC violations for other sites not adequately addressing the TIC phenomenon in their CHRRMs, a new Passport Action Request 01555551 [7], was written to provide the mechanism to assure a new review of the OE is performed.
Based on the lack of quantitative data of the TIC effect at PINGP for 1R-48/1R-49 &
2R-48/2R-49, Engineering Evaluation ECR 608000000013 was undertaken to determine the site-specific TIC effect of these CHRRMs during a worst-case LOCA and MSLB. Based on the results of this evaluation, CAP 501000001861 was written on 8/22/2017 and 1R-48/1R-49 & 2R-48/2R-49 were declared INOPERABLE by Operations.
Page 43 of 52
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                    PLANT SAFETY PROCEDURE NUMBER:
F3                    CORE DAMAGE ASSESSMENT REV:
F3-17 16            Technical Evaluation High Range Containment Radiation Monitors Thermal Induced Current ECR 608000000176 Since this time, the NRC has approved PINGPs LAR L-PI-17-006 [20] in NRC SER on March 6, 2018 [21] revising fission product barrier containment radiation EALs for R-48
        & R-49 based on approved calculation GEN-PI-092 [9]. Specifically, the EAL containment radiation levels for a RCS barrier loss increase from 7 to 40 R/hr, a fuel cladding barrier loss increase from 200 to 5,500 R/hr, and a potential containment barrier loss increase from 800 to 23,000 R/hr. This evaluation will consider whether the changes in these EALs along with proposed changes to guidance in associated emergency and operations procedures are adequate to demonstrate that 1R-48/1R-49
        & 2R-48/2R-49 can perform their intended functions when required.
2.0    Analysis - Design Basis Accidents and Quantifying Tic Effects From PINGP accident analyses, the design basis accidents that cause the largest temperature transients in containment that would translate to the largest TIC effects are the Large-Break Loss-Of-Coolant Accident (LOCA) and the Main Steam Line Break (MSLB). From PINGPs LOCA analyses of record [8], the LOCA scenario with the highest peak temperature ramps from 120&deg;F to around 266&deg;F within approximately 15 seconds (See Figure A-1 below). The black line represents a temperature that bounds all analyzed scenarios with margin for EQ Program purposes and is not an actual DBA temperature profile.
Page 44 of 52
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                    PLANT SAFETY PROCEDURE NUMBER:
F3                    CORE DAMAGE ASSESSMENT REV:
F3-17 16            Technical Evaluation High Range Containment Radiation Monitors Thermal Induced Current ECR 608000000176 Only four of the seven analyzed LOCA profiles extend to the highest LOCA temperature peak that occurs at around 27 minutes into the event. Since all the LOCA temperature profiles have a very similar shape, its not expected that TIC effects would be greatly different between the individual LOCA profiles. However, to account for the uncertainty of not obtaining the highest TIC effect from the shorter profiles, all the longer temperature profiles were run on the TIC software and the maximum difference between TIC effect results were added to the maximum TIC result from each of the longer profiles to obtain the maximum TIC effect result. Also, as previously mentioned, cable lengths have a direct effect on the TIC values and cable lengths differ between each radiation monitor. Therefore, TIC values for each radiation monitor were calculated from the TIC software results and cable lengths to show the variation in readings operators or observers would see during a LOCA.
For a MSLB, containment temperature ramps from 120&deg;F to 311&deg;F in 80 seconds followed by a relatively fast temperature reduction to 245 &deg;F after 800 seconds (13 min.) [10] (See Figure B-1 below).
Although the MSLB temperature profile extends to only 800 seconds post-accident, from steam line break descriptions in USAR Section - 14 [2], no additional temperature spikes are expected from the event. So, the trend should continue as a slow decrease in temperature.
Page 45 of 52
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                  PLANT SAFETY PROCEDURE NUMBER:
F3                    CORE DAMAGE ASSESSMENT REV:
F3-17 16            Technical Evaluation High Range Containment Radiation Monitors Thermal Induced Current ECR 608000000176 3.0    RESULTS - CHRRM POST-ACCIDENT RESPONSE TO A LOCA TIC software results for the LOCA accident profiles [35] show that during the initial temperature spike that peaks at approximately 15 seconds, the CHRRMs will produce peak TIC errors greater than +2,550 R/hr at around 9 seconds post-LOCA in the radiation monitor most effected by the TIC phenomenon (2R-48) because it has the longest cable length. As containment cools, the TIC errors decrease quickly until they become negative at around 90 seconds. The TIC errors become more negative until about 120 seconds (2 minutes) when they bottom in the -27 R/hr. in 2R-48. TIC errors then trend upward again until they reach zero at around 300 seconds (5 minutes) post-LOCA. Then, as temperatures increase again, a secondary positive TIC error peak of
        +22 R/hr in 2R-48 occurs around 7 minutes post-LOCA. Then, the TIC errors slowly decrease until around 18 minutes when they quickly drop and turn negative until they bottom out around -11 R/hr. in 2R-48 at approximately 20 minutes post-LOCA.
Fluctuation in TIC errors continue in decreasing magnitudes as containment air temperatures change throughout the first 65 minutes of the event. At the end of the 6,000 seconds (100 minutes) post-LOCA simulation, TIC values are slightly negative ranging from -0.49 R/hr on 2R-48 to -0.35 R/hr on 2R-49. The full LOCA TIC response chart for all CHRRMs is shown in Figure D below. Note: The TIC error values used to construct Figure D were calculated from the TIC software output of the LOCA profile that produced the highest peak and lowest trough response in the 5-20 minute timeframe multiplied by 1.26 to account for the 26% maximum variation in the peak magnitudes seen in all four LOCA responses. The reasoning behind using the output with highest TIC error magnitudes from the 5-20 minute timeframe is that it would be the worst case for making an EAL classification decision during an event.
Page 46 of 52
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                  PLANT SAFETY PROCEDURE NUMBER:
F3                      CORE DAMAGE ASSESSMENT REV:
F3-17 16            Technical Evaluation High Range Containment Radiation Monitors Thermal Induced Current ECR 608000000176 Figure D: Post-LOCA TIC Error for 1R-48, 1R-49, 2R-48, and 2R-49.
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                  PLANT SAFETY PROCEDURE NUMBER:
F3                      CORE DAMAGE ASSESSMENT REV:
F3-17 16            Technical Evaluation High Range Containment Radiation Monitors Thermal Induced Current ECR 608000000176 4.0    Results - CHRRM Post-Accident Response to a MSLB TIC software results for the MSLB accident profile [34] show that during the initial temperature spike that peaks at approximately 85 seconds, the CHRRMs will produce a peak TIC error of +1,922 R/hr at around 27 seconds post-MSLB in the radiation monitor most effected by the TIC phenomenon (2R-48). As containment cools, the TIC error decreases quickly until it becomes negative at around 117 seconds (1.95 minutes). The TIC error becomes more negative until around 158 seconds (2.6 minutes) when it bottoms out at -144 R/hr for 2R-48. TIC errors then trend less negative as temperature in containment slowly cool. However, at the end of the 800 seconds (13.3 minutes) post-MSLB simulation, TIC values are still negative and range from -7 R/hr on 2R-48 to -5 R/hr on 2R-49. See Figure E below.
Figure E: Post-MSLB TIC Error for 1R-48, 1R-49, 2R-48, and 2R-49.
Page 48 of 52
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                        PLANT SAFETY PROCEDURE NUMBER:
F3                      CORE DAMAGE ASSESSMENT REV:
F3-17 16            Technical Evaluation High Range Containment Radiation Monitors Thermal Induced Current ECR 608000000176 5.0    Evaluation - TIC Impact on Design and License Bases Functions For Emergency Planning, the Containment High-Range Radiation Monitors are used to determine the condition of fission product barriers during an event. Specifically, they are used to identify a RCS Barrier Loss, a Fuel Clad Barrier Loss, and a Containment Barrier Potential Loss from the Emergency Action Level (EAL) Matrix in the Emergency Plan [24]. The recently approved License Amendment Request [20, 21] raises the EAL for an RCS Barrier Loss from 7 R/hr to 40 R/hr on both R-48 and R-49; a Fuel Clad Barrier Loss from 200 R/hr to 5,500 R/hr, and a potential Containment Barrier Loss from 800 R/hr to 20,000 R/hr. So, this evaluation will use and will only apply when these new EALs that are implemented.
In evaluating the ability of the CHRRMs to perform their functions of detection of significant releases and emergency plan actuation, the time constraint of significance is the requirement to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an EAL has been exceeded.[24] However, in practice, the individual making the event classification needs the information no later than 13 minutes post-accident to document and verify the correct classification in the EAL tables to meet this deadline. The likelihood of an emergency classification being declared in the first 5 minutes of an event is small due to the high demands on the Main Control Room staff at the beginning of an event. So, for this evaluation, the critical timeframe when information from the CHRRMs will be needed to perform their design and license bases functions is considered to be between 5 and 13 minutes post-accident.
For the CHRRMs to function properly during a LOCA or MSLB, their readings need to provide Operations information to correctly classify the event. Radiation readings must be accurate enough that they dont lead to under- or over-classification of the actual event. With this criteria in mind, CHHRM acceptance criteria for emergency plan actuation and release assessment is determined to be the following: Expected radiation level for fission product barrier loss + TIC error falls in the correct EAL range given for the fission product barrier loss during the 5-13 minute post-event timeframe.
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                      PLANT SAFETY PROCEDURE NUMBER:
F3                      CORE DAMAGE ASSESSMENT REV:
F3-17 16            Technical Evaluation High Range Containment Radiation Monitors Thermal Induced Current ECR 608000000176 By definition, during a LOCA there is a loss of the RCS fission product barrier. Loss of the RCS fission product barrier alone would result in an alert event classification according to the EAL tables. From the current PINGP calculation of record [9], a loss of the RCS barrier without fuel failure would produce radiation levels in containment measured at 43 R/hr on the CHRRMs (not including the TIC error). The next-highest level of classification based on CHRRM readings is site area emergency due to fuel cladding failure. From the calculation of record [9], a 5% fuel cladding damage event (corresponding to a loss of fuel cladding barrier) would produce radiation levels in containment measured at 5,957 R/hr on the CHRRMs (not including the TIC error).
Finally, the calculation of record [9] determined that a 20% fuel cladding damage event (corresponding to a potential loss of containment) would produce radiation levels of 23,830 R/hr. These radiation values will be used as the expected radiation levels in containment during loss of fission product barriers when evaluating the ability of CHRRMs to perform their intended functions of emergency plan actuation and release assessment during the 5-13 minute post-event timeframe.
During a MSLB in containment, the breach should be in only the non-radioactive secondary coolant system. Containment radiation should show no significant increase.
Thus, based solely on CHRRM readings, acceptance criteria for the CHRRMs during an MSLB is indicated radiation levels should be below the RCS barrier loss EAL in the 5-13 minute post-accident timeframe resulting in no event being declared (The classification would be based on other plant indications). A review of MSLB data from the TIC software output [34] indicates that the range of expected TIC errors during the 5-13 minutes post-MSLB timeframe are -4 R/hr at 5 minutes to -52 R/hr at 7 minutes post-accident. A summary of expected radiation readings from the CHRRMS during MSLBs and LOCAs with the corresponding acceptance criteria is presented in Table 1 below.
One additional item of note is TR-112582 mentions the Sorrento/GA detector has a 10-11 amp normal output or keep alive signal current. If this signal is lost, the system issues a fail alarm to the operator. Therefore, if a negative TIC error signal of 10-11 amp or more occurs, the keep alive signal will be negated and a fail alarm will be generated. The following table compiles all the previously described TIC software results, information, acceptance criteria, and required time frames to summarize the ability of Operations to correctly classify the LOCA and MSLB events based on CHRRMs readings indications:
Page 50 of 52
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                          PLANT SAFETY PROCEDURE NUMBER:
F3                      CORE DAMAGE ASSESSMENT REV:
F3-17 16              Technical Evaluation High Range Containment Radiation Monitors Thermal Induced Current ECR 608000000176 Event              Expected      TIC Error 5-  CHRRMs            Acceptance        Acceptance Criteria Radiation    13 min. (R/hr) Readings 5 -13    Criteria (R/hr)  Met?
Level (R/hr)                min. (R/hr)
(Expected + TIC)
LOCA              43            0 to 22        43 to 65          40 to 5,500      Yes.
LOCA              5,957        0 to 22        5,957 to 5,979    5,500 to 20,000  Yes w/Cladding Failure LOCA              23,830        0 to 22        23,830 to 23,852  > 20,000          Yes w/potential loss of containment MSLB              Normal area  -4 to -52      Failed low      < 40              Yes, with additional value <1                                                        guidance to address Failed low Table 1: Comparison of CHRRM Indication vs. Acceptance Criteria Table 1 shows that TIC errors during the 5 to 13 minutes period post-LOCA (without fuel cladding damage) are expected to produce readings on all four CHRRMs in the range of 43 to 65 R/hr. This range of radiation readings falls within the new EAL range of >40 and <5,500 R/hr corresponding to a RCS barrier loss. So, this meets the established acceptance criteria. For the case of a LOCA with fuel cladding failure, Table 1 shows CHRRMs would produce readings between 5,957 and 5,979 R/hr.
These values also fall in the corresponding EAL acceptance criteria range for fuel cladding failure of 5,501 to 20,000 R/hr. For the case of potential loss of containment barrier, Table 1 shows CHRRMs would produce readings between 23,830 and 23,852 R/hr. These readings are above the EAL level of 20,000 R/hr that would indicate a potential loss of containment barrier. So, CHRRMs readings during all potential loss of fission product barriers due to LOCAs meet the acceptance criteria for emergency plan actuation and release assessment during the 5-13 minute post-accident timeframe.
For the case of an MSLB, Table 1 shows that TIC errors during the 5 to 13 minutes are expected to be of sufficient magnitude to negate the detector signal from normal containment radiation and the keep alive signal, producing a fail low alarm according to the detector logic diagram [29]. If appropriate guidance were provided in procedures and documents identified in Section IV that failed low alarms were a normal response to a cooldown event in containment for these detectors and were not failed detectors, operators and/or the emergency director could correctly conclude the fission product barriers were intact and would not classify the event based on the CHRRM readings.
Therefore, CHRRMs acceptance criteria are also met for emergency plan actuation and release assessment during the 5-13 minute post-event timeframe post-MSLB.
Page 51 of 52
 
PRAIRIE ISLAND NUCLEAR GENERATING PLANT                                      PLANT SAFETY PROCEDURE NUMBER:
F3                      CORE DAMAGE ASSESSMENT REV:
F3-17 16            Technical Evaluation High Range Containment Radiation Monitors Thermal Induced Current ECR 608000000176 Assessing the ability of CHRRMs to perform its remaining intended function, long-term surveillance, it can be seen in both LOCA and MSLB TIC Error charts the TIC errors decrease as stored energy from the accident has fully released and temperatures stabilize in containment. TIC errors will trend to zero with no additional releases of energy and CHRRMs will provide accurate measurement of actual radiation conditions in containment as time passes. Therefore, it is concluded that CHRRMs are able to perform their design and license bases function of long-term surveillance of radiation levels in containment during accident conditions.
6.0    Conclusions and Recommendations Based on this evaluation with the new containment radiation EALs in place, enhanced procedural guidance on the TIC effect specified in Section X., and associated training of Operations, RP, and Emergency Response performed, the CHRRMs are able to perform their license and design bases functions. This evaluation shows that five minutes after either a LOCA or a MSLB Operations will be able to use CHRRMs measurements to detect significant radiation releases and actuate the emergency plan (Classify events) in the timeframe necessary (5-13 minutes). If personnel question alarms or data indicating high readings in the first five minutes after an event or failed low readings, there will be appropriate guidance in the alarm response or emergency procedure to assure them that this is expected after a thermal transient. Also, CHRRMs will be accurate enough after five minutes for personnel to monitor containment radiation levels in the long-term. The accuracy of CHRRMs five minutes after LOCAs and MSLBs meets the acceptance criteria for all its intended functions with enhanced procedural guidance and training in place.
One recommendation is to ensure all the conditions of this evaluation are met by creating a tracking action to verify all required procedure changes and training from Section X have been completed before returning 1R-48/1R-49 & 2R-48/2R-49 to operable.
Page 52 of 52
 
REC/RPSS HANDBOOK INFORMATION USE
* Procedure should be available, but not necessarily at the work location.
* Procedure may be performed from memory.
* User remains responsible for procedure adherence.
REVISION 26 APPROVAL:
PCR #: 602000019055
 
REC/RPSS HANDBOOK REV. 26 PREFACE This booklet contains information that may be useful to the Radiological Emergency Coordinator and/or Radiation Protection Support Supervisor for use during a radiological emergency at the Prairie Island Nuclear Generation Plant. Some of the information is factual data that may be used in formulating response decisions. Other information presents suggested solutions to pre-identified problems or situations that may exist during an emergency situation Page 2 of 4
 
REC/RPSS HANDBOOK REV. 26 TABLE OF CONTENTS REV. PAGE DESCRIPTION                                                    NO.      NO.
Title Page 26        1 Preface 26        2 Table of Contents 26        3 TAB                                                            REV. NO. OF DESCRIPTION NO.                                                            NO.      PAGES 1        Representative Gamma Shielding Factors                  0        2 2        Guidelines for Contamination of Human Food and Animal Feed                                    0        1 3        R-50 OOS Grab Samples                                    0        1 4        R-51/R-52 OOS Monitoring Alternatives                    0        1 5        R-48/R-49 TIC Errors & OOS Alternatives                26        4 6        Condenser Contamination After SGTR                        (deleted) 7        10CFR20 Appendix B, Annual Limits on Intake (ALIs) and Derived Air Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent Concentrations:
Concentrations for Release to Sewerage                  2        89 8        Shield Bldg. Stack Effluent Release Flow Paths          16        1 9        Units & Conversion Factors                              1        17 10      Containment Volume                                      0        1 11      Cember Health Physics (Half Life I131, Dose due to total decay I131, Surface contamination limits)      0        4 12      Effective Half-lives                                    0        1 13      Xe133 Equivalent Activity Chart                          0        1 Page 3 of 4
 
REC/RPSS HANDBOOK REV. 26 TABLE OF CONTENTS (Cont.)
TAB                                                          REV. NO. OF DESCRIPTION NO.                                                          NO. PAGES Harmerski Xe-133 Eq. Curies in Core vs Time &
14                                                            0        6 I131 Eq. Curies in Core 15  Radioiodine concentration vs exposure                      (deleted) 16  Noble Gas concentration vs. exposure                        (deleted) 17  DELETED                                                  26        1 Regulatory Guide 8.29, Instruction Concerning Risks 18  From Occupational Radiation Exposure,                    1        17 Revision 1, February 1996.
19  DELETED                                                  26        1 20  DELETED                                                  20        1 21  DELETED                                                  26        1 22  Radiation Monitors EAL Matrix                            19        1 Power Supplies for Effluent Rad Monitors &
23                                                            0        4 Ventilation Systems Steam Generator Iodine Partitioning 24                                                            0        3 (Attachment E to 1ES-3.1)
Core Damage Assessment (discussion) 25                                                            17        1 F3-17 page 3 of 39 26  Radiation Monitor Control Functions                      26        6 Safety Injection Signal / Containment Isolation Signal /
27  Containment Ventilation Isolation Signal / Control Room  17        1 Ventilation Isolation Signal 28  Calculating Field Gas Activities                          19        1 Guidance for Selecting Geopolitical Subareas when 29                                                            21        1 Making a Protective Action Recommendation CAP 01522424 Guidance for spraying a radioactive 30                                                            25        3 plume with water.
Page 4 of 4
 
REPRESENTATIVE SHIELDING FACTORS FROM GAMMA CLOUD SOURCE*
Structure or Location                  Shielding Factor (a)                Representative Range Outside                                                1.0 Vehicles                                                1.0 Wood-frame House (b)                                  . 0.9 (no basement)
Basement of wood house                                  0.6                              0.1 to 0. 7 (c)
Masonry House (no basement)                            0.6                              0.4 to 0. 7 (c)
Basement of masonry house                              0.4                              0.1 to 0.5  (c)
Large office or industrial building                    0.2                            0.1 to 0.3 (c, d)
(a)      The ratio of the dose received inside the structure to the dose that would be received outside the structure.
(b)      A wood frame house with brick or stone veneer is approximately equivalent to a masonry house for shielding purposes.
(c)      This range is mainly due to different wall materials and different geometries.
(d)      The shielding factor depends on where the personnel are located within the building (e.g., the basement or an inside room)
SELECTED SHIELDING FACTORS FOR AIRBORNE RADIONUCLIDES Wood house, no basement                                                      0.9 Wood house, basement                                                        0.6 Brick house, no basement                                                    0.6 Brick house, basement                                                        0.4 Large office or industrial building                                          0.2 Outside                                                                      1.0
* Taken from SAND 77-1725 (Unlimited Release)
Page 1 of 2                                  TAB 1 Rev. 0
 
REPRESENTATIVE SHIELDING FACTORS FOR SURFACE DEPOSITED RADIONUCLIDES
* Structure or Location                            Shielding Factor (a)                Representative Range 1m above an infinite smooth surface                      1.00 1m above ordinary ground                                0.70                              0.47-0.85 1m above center of 50-ft roadways,                      0.55                              0.4-0.6 50% decontaminated Cars on 50-ft road:
Road fully contaminated                                  0.5                              0.4-0.7 Road 50% decontaminated                                  0.5                              0.4-0.6 Road fully decontaminated                                0.25                              0.2-0.5 Trains                                                  0.40                              0.3-0.5 One and two-story wood-frame                            0.4 (b)                            0.2-0.5 house (no basement)
One and two-story block and brick                      0.2 (b)                            0.04-0.40 house (no basement)
House basement, one or two walls ful-                  0.1 (b)                            0.03-0.15 ly exposed One story, less than 2 ft of basement,                0.05 (b)                            0.03-0.07 walls exposed Two stories, less than 2 ft of base-                  0.03 (b)                            0.02-0.05 ment, walls exposed Three or four story structures, 5000 to 1,000 ft2 per floor; First and second floors;                                0.05(b)                            0.01 -0.08 Basement                                                0.01(b)                          0.001 -0.07 Multistory structures> 10,000 ft2 per floor:
Upper floors:                                          0.01 (b)                          0.001 -0.02 Basement                                              0.005 (b)                          0.001 - 0.015 (a)        The ratio of the dose received inside the structure to the dose that would be received outside the structure.
(b)        Away from doors and windows.
* Taken from SAND 77-1725 (Unlimited Release)
Page 2 of 2                                    TAB 1 Rev. 0
 
GUIDELINES FOR CONTAMINATION OF HUMAN FOOD AND ANIMAL FEED*
PREVENTIVE PAG'S 1.5 rem projected dose commitment to thyroid 0.5 rem projected dose commitment to whole body, bone marrow, or any other organ RESPONSE LEVELS FOR                                    1-131    Cs-134      Cs-137      Sr-90        Sr-89 PREVENTIVE PAG Initial Activity Area                                    0.13        2          3          0.5          8 Deposition (&#xb5;Ci/m 2)
Forage Concentration (&#xb5;Ci/kg)                            0.05        0.8        1.3        0.18          3 Peak Milk Activity (&#xb5;Ci/liter)                          0.015      0.15        0.24        0.009        0.14 Total Intake (&#xb5;Ci)                                      0.09        4          7          0.2        2.6 EMERGENCY PAG'S 15 rem projected dose commitment to thyroid 5 rem projected dose commitment to whole body, bone marrow, or any other organ
 
===RESPONSE===
LEVELS FOR EMERGENCY PAG                  1-131        Cs-134        Cs-137            Sr-90            Sr-89 Infant    Adult Infant  Adult  Infant    Adult  Infant  Adult    Infant  Adult Initial Activity Area            1.3      18    20    40      30      50      5      20        80    1600 Deposition (&#xb5;Ci/m 2)
Forage Concentration            0.5      7      8    17      13      15      1.8      8        30      700
(&#xb5;Ci/kg)
Peak Milk Activity              0.15      2    1.5    3      2.4        4    0.09    0.04      1.4      30
(&#xb5;Ci/liter)
Total Intake (&#xb5;Ci)              0.9      10    40    70      70      80      2      7        26      400
 
==Reference:==
Accidental Radioactive Contamination of Human Food and Animal Feeds; Food and Drug Administration Recommendations for State and Local Agencies, Federal Register, October 22, 1982.
Page 1 of 1                                  TAB 2 Rev. 0
 
R-50 OOS GRAB SAMPLES I. ISSUE:
What are other means of monitoring high level rad releases via the Shield Building Stack when R-50 is Out Of Service?
II. SUGGESTIONS:
1.0    If the R-50 vacuum pump is out of service,
: 1. 1    Install and start a portable air pump.
1.2      Spare portable air pumps are located at Access Control.
NOTE:      If a portable air pump is used, the flow will usually bypass the R-50 monitor, but still allow sampling of the sample stream.
1.3    An R-50 grab sample may be taken as necessary per F3-20.2 (about once every hour).
2.0    If the monitor R-50 is out of service, refer to F3-20.2 for instructions on obtaining local samples and dose rates for releases from the Shield Building Vent Stacks.
Page 1 of 1 TAB 3 Rev. 0
 
R-51/R-52 OOS MONITORING ALTERNATIVES I. ISSUE:
What are other means of monitoring high level rad levels in the steam lines when R-51/52 is Out Of Service?
II. SUGGESTIONS:
If R-51/52 is out of service, consider following the instructions provided in F3-20.1 concerning obtaining readings from the AM-2 remote monitor.
Page 1 of 1 TAB 4 Rev. 0
 
R-48/R-49 Thermally Induced Errors & OOS Alternatives I. ISSUES:
: 1. R-48/R-49 Thermally Induced Errors R-48/R-49 Thermally Induced Current Errors Erroneous readings will occur on R-48/R-49 monitors during LOCA or MSLB accidents due to thermally induced current (TIC) in the detector cabling. These detectors can still be used to classify an emergency as follows:
* Ignore the readings for the first five minutes following a LOCA or MSLB
* During a LOCA without clad damage, the TIC readings on R-48/R-49 will be higher than the Alert EAL threshold of 40 R/h and this classification will be corroborated by a separate RCS leakage EAL
* During a LOCA without clad damage, the TIC readings on R-48/R-49 will not be high enough to exceed the Site Area EAL threshold of 5,500 R/h
* R-48/R-49 readings that exceed the Site Area EAL threshold of 5,500 R/h or the General Emergency threshold of 20,000 R/h should considered valid readings and emergency classification proceed as required
* It is expected that the MSLB will cause the monitors to experience downscale failure after five minutes Page 1 of 4 TAB 5, Rev. 26
 
R-48/R-49 Thermally Induced Errors & OOS Alternatives 50 LOCA TIC Error (R/hr) 40 30                                                                                - -  TIC error (R/hr) 1R-49 20
                                                                                                      - - TIC error (R/hr) 2R-49
                                                                                                      - -  TIC error (R/hr) 2R-48 TIC Error (R/hr) 10                              \
                                                                                                      - -  TIC error (R/hr) 1R-48
                                        '\
                                              ,Jo 0
0
                                                        'I.,,...                      19" 5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100105
                  -10
                  -20
                  -30
                  -40
                  -50 Time (Min) 200 MSLB TIC Error (R/hr) 150 100 l
50 R/hr                0
                                \                                                                          TIC error (R/hr) 1R-49 TIC error (R/hr) 2R-49
                    -50
                  -100 0
2 1
I    4              6 V  8          10    12  14        TIC error (R/hr) 2R-48 TIC error (R/hr) 1R-48
                                        ~
V
                  -150
                  -200 Time (min)
Page 2 of 4 TAB 5, Rev. 26
 
R-48/R-49 Thermally Induced Errors & OOS Alternatives
: 2. R-48/R-49 Out of Service If R-48/49 are Out Of Service, see the alternative methods to monitor high level radiation in containment described below.
II. SUGGESTIONS:
: 1. Investigate the usefulness of low range R-2 or R-7 radiation monitors. Use if they are functional.
1.1. If the loop monitors read low, then containment probably does not indicate failed fuel atmosphere.
1.2. If the loop monitors read high, then the RCS and/or containment is highly contaminated. An air sample of containment will determine if containment is contaminated.
: 2. Obtain radiation monitor readings exterior to the containment shield building.
Use the predicted results of F3-25, Reentry, as a guide for determining the percent of design basis accident fuel failure.
: 3. Obtain core exit temperatures.
Temperature                          Potential Damage
                          < 750 Deg. F            No Cladding Damage 750 - 1300 Deg. F            0 - 50% Clad Damage 1300 - 1650 Deg. F            50 - 100% Clad Damage
                          > 1650 Deg. F            Zr-HOH Reaction, 0 - 50% Fuel Melt
                          > 3450 Deg. F            50 - 100% Fuel Melt
 
==Reference:==
F3-17, Table 6 Characteristics of Categories of Fuel Damage.
Page 3 of 4 TAB 5, Rev. 26
 
R-48/R-49 Thermally Induced Errors & OOS Alternatives
: 4. Obtain RVLIS readings.
Reactor Vessel Full Range Indication displays the level from NOTE:      the bottom of the reactor vessel to top of reactor vessel when both RCPs are stopped.
Reactor Vessel Full Range reading of < 50% indicates a level below top of core.
(
 
==Reference:==
Ops Manual B4B, Rx Vessel Level Instrumentation System, Rev. 3.)
Page 4 of 4 TAB 5, Rev. 26
 
REC RPSS 06 DELETED Refer to C21.1.1 AOP , Processing Condensate Following a Steam Generator Tube Rupture Page 1 of 1                    TAB 6, Rev. 21
 
10CFR20 Appendix B Annual Limits on Intake (ALIs) and Derived Air Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent Concentrations; Concentrations for Release to Sewerage Attached are pages B-1 to B-88 of Appendix B to Part 20.
The new liquid effluent limits are ten (10) times NOTE:    the Water Effluent Concentrations of Table 2, Column 2 IAW T.S. 5.5.4.b.
Page 1 of 1              TAB 7, Rev. 2
 
APPENDIX B Statement of Requirement:
Note: The accompanying tables (3) and footnotes/notes that comprise the rest of Appendix B can be found on the web at: <http://www.nrc.gov/NRC/CFR/fABLES/ISOTOPES/PART020-APPB/radionuclides.htrnl>.
Appendix B to Part 20 - Annual Limits on Intake (ALis) and Derived Air Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent Concentrations; Concentrations for Release to Sewerage Introduction For each radionuclide, Table 1 indicates the chemical form which is to be used for selecting the appropriate ALl or DAC value. The ALis and DACs for inhalation are given for an aerosol with an activity median aerodynamic diameter (AMAD) of 1 m and for three classes (D,W,Y) of radioactive material, which refer to their retention (approximately days, weeks or years) in the pulmonary region of the lung. This classification applies to a range of clearance half-times of less than 10 days for D, for W from 10 to 100 days, and for Y greater than 100 days. The class (D, W, or Y) given in the column headed "Class" applies only to the inhalation ALis and DACs given in Table 1, columns 2 and 3. Table 2 provides concentration limits for airborne and liquid effluents released to the general environment. Table 3 provides concentration limits for discharges to sanitary sewer systems.
Notation The values in Tables 1, 2, and 3 are presented in the computer "E" notation. In this notation a value of 6E-02 represents a value of 6x I 0*2 or 0.06, 6E+2 represents 6x 102 or 600, and 6E+0 represents 6x 10&deg; or 6.
Note that the columns in Table 1 of this appendix captioned "Oral Ingestion ALI," "Inhalation ALI," and "DAC," are applicable to occupational exposure to radioactive material.
The ALis in this appendix are the annual intakes of a given radionuclide by "Reference Man" which would result in either: (1) a CEDE of 5 rems (stochastic ALI); or (2) a committed dose equivalent of 50 rems to an organ or tissue (non-stochastic ALI). The stochastic ALis were derived to result in a risk, due to irradiation of organs and tissues, comparable to the risk associated with deep-dose equivalent to the whole body of 5 rems. The derivation includes multiplying the committed dose equivalent to an organ or tissue by a weighting factor, wT. This weighting factor is the proportion of the risk of stochastic effects resulting from irradiation of the organ or tissue, T, to the total risk of stochastic effects when the whole body is irradiated uniformly. The values of wT are listed under the definition of weighting factor in 10 CPR 20.1003. The non-stochastic ALis were derived to avoid non-stochastic effects, such as prompt damage to tissue or reduction in organ function.
B-1                                    NUREG- 1736
 
APPENDIX B A value of wT=0.06 is applicable to each of the five organs or tissues in the "remainder" category receiving the highest dose equivalents, and the dose equivalents of all other remaining tissues may be disregarded. The following parts of the GI tract - stomach, small intestine, upper large intestine, and lower large intestine - are to be treated as four separate organs. Note that the dose equivalents for extremities (hands and forearms, feet and lower legs), skin, and lens of the eye are not considered in computing the CEDE, but are subject to limits that must be met separately.
When an ALI is defined by the stochastic dose limit, this value alone, is given. When an ALI is determined by the non-stochastic dose limit to an organ, the organ or tissue to which the limit applies is shown, and the ALI for the stochastic limit is shown in parentheses. (Abbreviated organ or tissue designations are used: LLI wall = lower large intestine wall; St. wall = stomach wall; Blad wall= bladder wall; and Bone surf= bone surface.)
The use of the ALis listed first, the more limiting of the stochastic and non-stochastic ALis, will ensure that non-stochastic effects are avoided and that the risk of stochastic effects is limited to an acceptably low value. If, in a particular situation involving a radionuclide for which the non-stochastic AL I is limiting, and use of that non-stochastic ALI is considered unduly conservative, the licensee may use the stochastic ALI to determine the CEDE. However, the licensee shall also ensure that the 50-rem dose equivalent limit for any organ or tissue is not exceeded by the sum of the external deep-dose equivalent plus the internal committed dose to that organ (not the effective dose). For the case where there is no external dose contribution, this would be demonstrated if the sum of the fractions of the nonstochastic ALis (ALl05 ) that contribute to the committed dose equivalent to the organ receiving the highest dose does not exceed unity (i.e., (intake (in &#xb5;Ci) of each radionuclide/ALl05 ) <1 .0). If there is an external deep-dose equivalent contribution of Hd then this sum must be less than l-(H/50) instead of being
<1.0.
The derived air concentration (DAC) values are derived limits intended to control chronic occupational exposures. The relationship between the DAC and the ALI is given by:
DAC = ALI (in &#xb5;Ci)/(2,000 hours per working year x 60 minutes/hour x 2 x 104 ml per minute) = [ALl/2.4 x 109) ,&i/ml, where 2 x 104 ml is the volume of air breathed per minute at work by "Reference Man" under working conditions of "light work."
The DAC values relate to one of two modes of exposure: either external submersion or the internal committed dose equivalents resulting from inhalation of radioactive materials. Derived air concentrations based upon submersion are for immersion in a semi-infinite cloud of uniform concentration and apply to each radionuclide separately.
The ALI and DAC values relate to exposure to the single radionuclide named, but also include contributions from the in-growth of any daughter radionuclide produced in the body by the decay of the parent. However, intakes that include both the parent and daughter radionuclides should be treated by the general method appropriate for mixtures.
NUREG- 1736                                      B-2
 
APPENDIX B The value of ALI and DAC -do not apply directly when the individual both ingests and inhales a radionuclide, when the individual is exposed to a mixture of radionuclides by either inhalation or ingestion or both, or when the individual is exposed to both internal and external radiation (see 10 CFR 20.1202). When an individual is exposed to radioactive materials which fall under several of the translocation classifications (i.e., Class D, Class W, or Class Y) of the same radionuclide, the exposure may be evaluated as if it were a mixture of different radionuclides.
It should be noted that the classification of a compound as Class D, W, or Y is based on the chemical form of the compound and does not take into account the radiological half-life of different radioisotopes. For this reason, values are given for Class D, W, and Y compounds, even for very short-lived radionuclides.
The columns in Table 2 of this appendix captioned "Effluents," "Air," and "Water," are applicable to the assessment and control of dose to the public, particularly in the implementation of the provisions of 10 CFR 20.1302. The concentration values given in Columns 1 and 2 of Table 2 are equivalent to the radionuclide concentrations which, if inhaled or ingested continuously over the course of a year, would produce a total effective dose equivalent of 0.05 rem (50 millirem or 0.5 millisieverts).
Consideration of non-stochastic limits has not been included in deriving the air and water effluent concentration limits because non-stochastic effects are presumed not to occur at the dose levels established for individual members of the public. For radionuclides, where the non-stochastic limit was governing in deriving the occupational DAC, the stochastic ALI was used in deriving the corresponding airborne effluent limit in Table 2. For this reason, the DAC and airborne effluent limits are not always proportional as was the case in Appendix B to 10 CFR 20.1- 20.601.
The air concentration values listed in Table 2, Cplumn 1, were derived by one of two methods.
For those radionuclides for which the stochastic limit is governing, the occupational stochastic inhalation ALI was divided by 2.4 x 109ml, relating the inhalation ALI to the DAC, as explained above, and then divided by a factor of 300. The factor of 300 includes the following components: a factor of 50 to relate the 5-rem annual occupational dose limit to the 0. I-rem limit for members of the public, a factor of 3 to adjust for the difference in exposure time and the inhalation rate for a worker and that for members of the public; and a factor of 2 to adjust the occupational values (derived for adults) so that they are applicable to other age groups.
For those radionuclides for which submersion (external dose) is limiting, the occupational DAC in Table 1, Column 3, was divided by 2 19. The factor of 2 19 is composed of a factor of 50, as described above, and a factor of 4.38 relating occupational exposure for 2,000 hours per year to full-time exposure (8,760 hours per year). Note that an additional factor of 2 for age considerations is not warranted in the submersion case.
B-3                                  NUREG- 1736
 
APPENDIX B The water concentrations were derived by taking the most restrictive occupational stochastic oral ingestion ALl and dividing by 7.3 x 107
* The factor of 7.3 x 107 (ml) includes the following components: the factors of 50 and 2 described above and a factor of 7.3 x 1()5 (ml) which is the annual water intake of "Reference Man."
Note 2 of this appendix provides groupings of radionuclides which are applicable to unknown mixtures of radionuclides. These groupings (including occupational inhalation ALis and DACs, air and water effluent concentrations and sewerage) require demonstrating that the most limiting radionuclides in successive classes are absent. The limit for the unknown mixture is defined when the presence of one of the listed radionuclides cannot be definitely excluded either from knowledge of the radionuclide composition of the source or from actual measurements.
The monthly average concentrations for release to sanitary sewers are applicable to the provisions in 10 CFR 20.2003. The concentration values were derived by taking the most restrictive occupational stochastic oral ingestion ALl and dividing by 7.3 x 106(ml). The factor of 7.3 x 106 (ml) is composed of a factor of 7 .3 x 105 (ml), the annual water intake by "Reference Man," and a factor of 10, such that the concentrations, if the sewage released by the licensee were the only source of water ingested by a reference man during a year, would result in a CEDE of0.5 rem.
List of Elements Name                                              Atomic Symbol                          No.
Actinium                                            Ac                              89 Aluminum                                            Al                              13 Americium                                          Am                              95 Antimony                                            Sb                              51 Argon                                              Ar                              18 Arsenic                                            As                              33 Astatine                                            At                              85 Barium                                              Ba                              56 Berkelium                                          Bk                              97 Beryllium                                          Be                                4 Bismuth                                            Bi                              83 Bromine                                              Br                              35 NUREG- 1736                                      B-4
 
APPENDIX B Name          Atomic Svmbol        No.
Cadmium              Cd          48 Calcium              Ca          20 Californium          Cf          98 Carbon                C            6 Cerium                Ce          58 Cesium                cs          55 Chlorine              Cl            17 Chromium              Cr          24 Cobalt                co          27 Copper                cu          29 Curium                Cm          96 Dysprosium            Dy          66 Einsteinium          Es          99 Erbium                Er          68 Europium              Eu          63 Fermium              Fm          100 Fluorine              F            9 Francium              Fr          87 Gadolinium            Gd          64 Gallium              Ga          31 Germanium            Ge          32 Gold                  Au          79 Hafnium              Hf          72 Holmium              Ho          67 Hydrogen              H              I B-5                  NUREG- 1736
 
APPENDIX B Name          Atomic Svmbol        No.
Indium                In          49 Iodine                  I          53 Iridium                Ir          77 Iron                  Fe          26 Krypton              Kr          36 Lanthanum            La          57 Lead                  Pb          82 Lutetium              Lu          71 Magnesium            Mg            12 Manganese            Mn          25 Mendelevium          Md          101 Mercury              Hg          80 Molybdenum            Mo          42 Neodymium            Nd          60 Neptunium            Np          93 Nickel                Ni          28 Niobium              Nb          41 Osmium                Os          76 Palladium            Pd          46 Phosphorus            p          15 Platinum              Pt          78 Plutonium            Pu          94 Polonium              PO          84 Potassium            K            19 Praseodymium          Pr          59 NUREG-1736        B-6
 
APPENDIX B Name          Atomic Svmbol        No.
Promethium            Pm          61 Protactinium          Pa          91 Radium                Ra          88 Radon                Rn          86 Rhenium              Re          75 Rhodium              Rh          45 Rubidium              Rb          37 Ruthenium            Ru          44 Samarium              Sm          62 Scandium              SC          21 Selenium              Se          34 Silicon                Si          14 Silver                Ag          47 Sodium                Na          11 Strontium              Sr          38 Sulfur                s          16 Tantalum              Ta          73 Technetium            Tc          43 Tellurium            Te          52 Terbium              Tb          65 Thallium              Tl          81 Thorium              Th          90 Thulium              Tm          69 Tin                  Sn          50 Titanium              Ti          22 B-7                NUREG- 1736
 
APPENDIXB Name          Atomic Symbol        No.
Tungsten            w            74 Uranium            u            92 Vanadium            V            23 Xenon              Xe          54 Ytterbium          Yb          70 Yttrium            y            39 Zinc                Zn          30 Zirconium          Zr          40 NUREG-1736      B-8
 
APPENDIX B Table 1                            Table 2            Table 3 Occupational Values                    Effluent          Releases to Concentrations          Sewers Col.1      Col.2        Col. 3      Col.1          Col.2 Oral                                                              Monthly Ingestion          Inhalation                                      Average Atomic                                                    ALl        ALl      DAC              Air        Water    Concentration No. Radionuclide                Class                (&#xb5;Ci)      (&#xb5;Ci)      (&#xb5;Ci/ml)      (&#xb5;Cf/ml)      (&#xb5;Ci/mJ)      (&#xb5;Ci/ml)
Hydrogen-3      Water, DAC includes skin        8E+4        8E+4          2E-5          IE-7          lE-3        lE-2 absorption Gas (HT or T2) Submersion 1: Use above values as HT and T, oxidize in air and in the body to HTO 4  Beryllium-7      W, all compounds except those given for Y Y, oxides, halides, and 4E+4 I  2E+4 2E+4 9E-6 SE-6 I      3E-8 3E-8 I    6E-4 I    6E-3 nitrates 4  Beryllium-10    W, see 'Be                      1E+3        2E+2          6E-8        2E-10            -            -
LLI wall (1E+3)          -                                      2E-S          2E-4 Y,see 'Be                          -          IE+!        6E-9        2E-l I            -
2 6  Carbon- 11'  '  Monoxide                          -          JE+6        5E-4          2E-6 Dioxide                            -        6E+S          3E-4        9E-7            -
Compounds                        4E+S        4E+S          2E-4        6E-7          6E-3          6E-2 6  Carbon-14        Monoxide                                    2E+6          7E-4      l      2E-6      I    -  I    -
Dioxide                            -        2E+S          9E-5          3E-7            -
Compounds                        2E+3        2E+3          I E-6        3E-9          3E-5          3E-4 9  Fluorine- 18' 21 D, fluorides ofH, Li. Na.        SE+4        7E+4          3E-5          I E-7                        -
K. Rb, Cs, and Fr              St wall                            I                    I-(5E+4)          -            -            -          7E-4          7E-3 W. fluorides of Be. Mg, Ca,        -        9E+4          4E-5          IE-7            -            -
Sr. Ba. Ra, Al. Ga. In. Tl, As, Sb, Bi. Fe, Ru. Os. Co, Ni. Pd. Pt, Cu. Ag, Au. Zn.
Cd, Hg, Sc. Y. Ti. Zr, V, Nb, Ta, Mn, Tc, and Re Y, lanthanum fluoride                    I  8E+4          3E-5          IE-7 II  Sodium-22        D, all compounds                4E+2        6E+2          3E-7        9E-10          6E-6          6E-5 II  Sodium-24        D, all compounds                4E+3        SE+3          2E-6        7E-9          SE-5          SE-4 B-9                                                        NUREG- 1736
 
APPENDIX B Table 1                          Table 2            Table 3 Occupational Values                  Effluent          Releases to Concentrations          Sewers Col.1  I  Col.2    I  Col. 3 I Col. 1          Col. 2  I Oral                                                          Monthly Ingestion I        Inhalation      I                            Average Atomic                                                ALI        ALI      DAC            Air        Water    Concentration No.      Radionuclide            Class            (&#xb5;Ci)      (&#xb5;Ci)      (&#xb5;Ci/ml)    (&#xb5;Ci/ml)      (&#xb5;Ci/ml)    (uCi/ml) 12  Magnesium-28    D, all compounds except      7E+2        2E+3          7E-7      2E-9          9E-6        9E-5 those given for W W, oxides, hydroxides,                    IE+3          SE-7      2E-9 carbides, halides, and nitrates 13    Aluminum-26    D, all compounds except      4E+2        6E+l          3E-8      9E-11            6E-6        6E-5 those given for W W, oxides, hydroxides,          -        9E+l          4E-8      IE-10            -            -
carbides, halides, and nitrates 14    Silicon-3 I    D, all compounds except      9E+3        3E+4          IE-5      4E-8            lE-4        IE-3 those given for W and Y W, oxides, hydroxides,                    3E+4          IE-5      SE-8              -
carbides, and nitrates Y. aluminosilicate glass                  3E+4          IE-5      4E-8 14  Silicon-32      D, see "Si                    2E+3 LLI wall  I  2E+2 I    IE-7 I JE-1    0            -
(3E+3)        -                                    4E-5        4E-4 1
W.see ; Si                                IE+2          SE-8      2E-1O            -
1 Y, see; Si                                5E+0          2E-9      2E-12            -
15  Phosphorus-32  D, all compounds except      6E+2        9E+2          4E-7        I E-9        9E-6        9E-5 phosphates given for W W, phosphates ofZn2*. s3*,                4E+2          2E-7      SE-10            -
Mg1*. Fe3*. Bil+, and lanthanides 15  Phosphorous-33  D, see)!p                    6E+3        8E+3          4E-6        IE-8          SE-5        8E-4 W. see*'!p                                3E+3          IE-6      4E-9 NUREG-1736                                                B-10
 
APPENDIX B Table 1                        Table 2          Table3 Occupational Values                Effluent        Releases to Concentrations        Sewers Col.1      CCol.2                3  Col.1        Col.2 Oral                                                        Monthly Ingestion          Inhalation                                Average Atomic                                                    ALl          ALI      DAC            Air        Water  Concentration No.      Radionuclide                Class            (&#xb5;Ci)        (&#xb5;Ci)    (&#xb5;Ci/ml)    (&#xb5;Ci/ml)      (&#xb5;Ci/ml)  (.uCi/mI) 16  Sulfur-35        Vapor                                        18+4        6E-6      2E-8 D, sulfides and sulfates        1E+4        2E+4          7E-6      2E-8          -
except those given for W    LLI wall (8E+3)                                  -          lE-4        lE-3 W, elemental sulfur,            6E+3        2E+3          9E-7      3E-9 sulfides of Sr, Ba, Ge, Sn, Pb, As, Sb, Bi, Cu, Ag, Au, Zn, Cd, Hg, W, and :Ml.
Sulfates of Ca, Sr, Ba, Ra, As, Sb, and Bi 17  Chlorine-36      D, chlorides of H, Li, Na,      2E+3        2E+3          IE-6      3E-9          2E-5        2E-4 K, Rb, Cs, and Fr W, chlorides of lanthanides.      -        2E+2          1 E-7    38-10 Be, Mg, Ca. Sr, Ba, Ra, Al, 3a, In. Tl, Ge, Sn, Pb, As, Sb, Bi, Fe, Ru, Os, Co, Rh, Ir. Ni, Pd, Pt, Cu, Ag, Au, Zn, Cd, Hg, Sc, Y, Ti, Zr, Hf, V, Nb, Ta, Cr, :Ml, W,
                        \-In, Tc, and Re 17  ::::hlorine-38'21 D,see 36CI                      2E+4        4E+4          2E-5      6E-8 St. wall (3E+4)          -            -          -          3E-4        3E-3 36 N, see CI                        .          5E+4        2E-5      6E-8 17  :hlorine-39'~1    ). see 3"CI                    2E+4        5E+4        2E-5      7E-8 St. wall (4E+4)          -            -          -          SE-4        SE-3 N, see 3"CI                      -        6E+4          2E-5      BE-8 18  \.rgon-37        ;ubmersion  111
                                                          -            -          IE+0      6E-3 18    \.rgon-39        ,ubmersion'  1
                                        >                  -            -          2E-4      8E-7 18  \.rgon-41        iubmersion< 1'                    -            -          3E-6      lE-8 19  'otassium-40      >. all compounds              3E+2        4E+2          2E-7      6E-I0        4E-6        4E-5 19    >otassium-42      ), all compounds              5E+3        5E+3          2E-6      7E-9          6E-5        6E-4 19    'otassium-43      ), all compounds              6E+3        9E+3          4E-6      IE-8        9E-5        9E-4 B-11                                                NUREG- 1736
 
                                                                                  ---------~
APPENDIX B Table 1                      Table 2          Table 3 Occupational Values              Effluent        Releases to Concentrations        Sewers Col. I      Col.2        Col.3  Col. I        Col. 2 Oral                                                      Monthly Ingestion          Inhalation                              Average Atomic                                              ALI          AL1      DAC        Air        Water  Concentration No. Radionuclide                Class        (&#xb5;Ci)        (&#xb5;Ci)      (&#xb5;Ci/ml) (&#xb5;Ci/ml)      (&#xb5;Ci/ml)  (&#xb5;Ci/ml) 19  Potassium-44<2>  D, all compounds        2E+4        7E+4          3E-5    9E-8            -
St. wall (4E+4)          -                                5E-4      5E-3 19  Potassium-45< 2>  D, all compounds        3E+4          IE+5        5E-5    2E-7 St. wall (5E+4)          -                                7E-4      7E-3 20    Calcium-41        W, all compounds        3E+3        4E+3          2E-6 Bone Surf    Bone Surf (4E+3)      (4E+3)            -    5E-9          6E-5      6E-4 20    Calcium-45      W. all compounds          2E+3        8E+2          4E-7    lE-9        2E-5      2E-4 20    Calcium-47      W. all compounds          8E+2        9E+2          4E-7    IE-9          IE-5      IE-4 21    Scandium-43      Y, all compounds          7E+3        2E+4          9E-6    3E-8          IE-4      IE-3 21    Scandium-44m    Y, all compounds          5E+2        7E+2          3E-7    IE-9        7E-6      7E-5 21    Scandium-44      Y, all compounds          4E+3          IE+4        5E-6    2E-8          5E-5      SE-4 21    Scandium-46      Y, all compounds          9E+2        2E+2          IE-7  3E-10          IE-5      IE-4 21    Scandium-47      Y. all compounds          2E+3        3E+3          IE-6    4E-9            -        -
LLI wall (3E+3)          -            -        -          4E-5      4E-4 21    Scandium-48      Y, all compounds          8E+2        1E+3          6E-7    2E-9          JE-5      IE-4 21    Scandium-49' 2l  Y, all compounds          2E+4        5E+4          2E-5    8E-8          3E-4      3E-3 22    Titanium-44      D, all compounds except  3E+2        IE+I          SE-9  2E-11          4E-6      4E-5 those given for W and Y W, oxides, hydroxides,                3E+l          IE-8  4E-ll            -
carbides, halides, and nitrates Y,SrTiO3                              6E+0          2E-9  SE-12            -
22    Titanium-45      D,see~i                  9E+3        3E+4          IE-5    3E-8          I E-4      IE-3 W, see-'1'i                            4E+4          I E-5  5E-8            -
Y, see "'1'i                          3E+4          IE-5    4E-8            -
NUREG- 1736                                            B-12
 
APPENDIX B Table 1                      Table2          Table3 Occupational Values              Effluent        Releases to Concentrations        Sewers Col.t        Col.2        Col.3  Col. t      Col.2 Oral                                                    Monthly Ingestion          Inhalation                            Average Atomic                                              ALI          AL1      DAC        Air        Water  Concentration No. Radionuclide                Class        (&#xb5;Ci)        (&#xb5;Ci)      (&#xb5;Ci/ml) (&#xb5;Ci/ml)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml) 23  Vanadium-4'12>  D, all compounds except    3E+4        8E+4          3E-5    IE-7            -
those given for W        St. wall (3E+4)          -                              4E-4      4E-3 W, oxides, hydroxides,                  1E+5          4E-5    lE-7            -
carbides, and halides 23  Vanadium-48      D, see 47 V                6E+2        IE+3          SE-7    2E-9          9E-6      9E-5 47 W, see V                                6E+2          3E-7  9E-10            -
47 23  Vanadium-49      D, see V                  7E+4        3E+4          lE-5 LLI wall    Bone Surf (9E+4)      (3E+4)            -    5E-8          JE-3      JE-2 W, see V 47 2E+4          8E-6    2E-8            -
24  Chromium-48      D, all compounds except    6E+3          IE+4        SE-6    2E-8          8E-5      8E-4 those given for W and Y W, halides and nitrates                7E+3          3E-6    IE-8 Y, oxides and hydroxides                7E+3          3E-6    IE-8 21          48 4E-4 24  Chromium-49'    D, see Cr                  3E+4        8E+4          4E-5    IE-7                    4E-3 48 W, see Cr                              IE+5          4E-5    lE-7            -
Y,see 48Cr                              9E+4          4E-5    IE-7 24  Chromium-5 1    *o. see  4 "Cr              4E+4        5E+4          2E-5    6E-8          5E-4      SE-3 W, see "'Cr                            2E+4          IE-5  3E-8 4
Y, see "Cr                              2E+4          8E-6    3E-8 25  Manganese-5 1m  D, all compounds except    2E+4        5E+4          2E-5    7E-8        3E-4      3E-3 those given for W W, oxides, hydroxides,                  6E+4          3E-5    SE-8            -
halides, and nitrates 25  Manganese-      D, see~ 1Mn                3E+4        9E+4          4E-5    IE-7 52m'?)                                    St. wall (4E+4)          -                                SE-4      SE-3 W, see~ 1Mn                              IE+S          4E-5    IE-7 51 25    Manganese-52    D, see Mn                  7E+2        IE+3          5E-7    2E-9          IE-5      IE-4 51 W, see Mn                              9E+2          4E-7    IE-9 B-13                                              NUREG- 1736
 
APPENDIX B Tablel                      Table2          Table3 Occupational Values              Effluent      Releases to Concentrations      Sewers Col. I      Col.2        Col,3    Col.1        Col. 2 Oral                                                    Monthly Ingestion        Inhalation                              Average Atomic                                              ALI        ALI      DAC          Air      Water  Concentration No. Radionuclide              Class          (&#xb5;Ci)      (&#xb5;Ci)    (&#xb5;Ci/ml}  (&#xb5;Ci/ml}    (&#xb5;Ci/ml)  (&#xb5;Cl/ml}
25    Manganese-53  D, see 51 Mn                SE+4        lE+4        SE-6        -        7E-4      7E-3 Bone Surf
                                                      -        (2E+4)          -      3E-8          -        -
W, see Mn51
                                                      -        1E+4        SE-6      2E-8          -          -
51 25    Manganese-54  D, see Mn                    26+3        9E+2        48-7      1E-9        3E-5      3E-4 W, see"Mn                      -        8E+2        38-7      18-9          -          -
51 25    Manganese-56  D, see Mn                    5E+3        2E+4        6E-6      2E-8        7E-5      7E-4 W,see Mn 51
                                                      -        2E+4        9E-6      3E-8          -          -
26    Iron-52        D, all compounds except      9E+2        3E+3          lE-6    4E-9        lE-5      lE-4 those given for W W, oxides, hydroxides, and    -        2E+3          IE-6    3E-9          -          -
halides 26    lron-5S        D, see 5iFe                  9E+3        2E+3        SE-7      3E-9        IE-4      IE-3 W. see s!Fe                    -        4E+3        2E-6      6E-9          -        -
52 26    lron-59        D, see Fe                    8E+2        3E+2          IE-7    SE-10        IE-5      lE-4 W, see 52Fe                    -        5E+2        2E-7    7E-10            -        -
26    lron-60        D, see siFe                  3E+I        6E+O        3E-9    9E-12        4E-7      4E-6 5
W. see ~Fe                    -        2E+I        SE-9    3E-1 I          -        -
27    Cobalt-SS      W, all compounds except      IE+3        3E+3          IE--6  4E-9        2E-S      2E-4 those given for Y Y, oxides, hydroxides,        -        3E+3          lE-6    4E-9          -        -
halides. and nitrates 27    Cobalt-56      W. see 51 Co                5E+2        3E+2          IE-7  4E-10        6E-6      6E-5 55 Y. see Co                    4E+2        2E+2        SE-8    3E-10            .        -
27    Cobalt-57      W, see 55Co                  8E+3        3E+3          IE-6    48-9        6E-5      6E-4 Y. see 55 Co                4E+3        7E+2        3E-7    9E-10            -        -
55 27    Cobalt-58m    W, see Co                    6E+4        9E+4        4E-5      IE-7        8E-4      8E-3 Y, sec 51
* Co                  -        6E+4        3E-5    9E-8          -          -
27    Cobalt-58      w. see  55 Co                2E+3        1E+3        SE-7    2E-9        2E-5      2E-4 Y, see ''Co                  IE+3        7E+2        3E-7      IE-9          -
NUREG- 1736                                              B-14
 
APPENDIX    13 Table I                      Table 2            Table 3 Occupational Values                Effluent        Releases to Concentrations        Sewers Col. I      Col.2        Col. 3  Col. I        Col.2 Oral                                                        Monthly Ingestion          Inhalation                                Average Atomic                                              ALI          ALI      DAC          Air        Water  Concentration No. Radionuclide                Class          (&#xb5;Ci)        (&#xb5;Ci)      (&i/ml)  (&#xb5;Ci/ml)      (&#xb5;Ci/ml)    (&#xb5;Ci/ml) 27  Cobalt-60m121  W, see 55Co                  IE+6        4E+6          2E-3    6E-6            -
St wall (IE+6)          -                                  2E-2        2E-l Y, see 55Co                              3E+6          IE-3    4E-6 55 27  Cobalt-60      W, see Co                    5E+2        2E+2          7E-8    2E-10          3E-6        3E-5 55 Y, see Co                    2E+2        3E+l          IE-8    SE-11            -
27  Cobalt-61 121  W, see 55Co                  2E+4        6E+4          3E-5    9E-8        3E-4        3E-3 55 Y, see Co                    2E+4        6E+4          2E-5    BE-8            -
27  Cobalt-62mr.i  w, see    55 Co              4E+4        2E+5          7E-5    2E-7            -
St. wall (5E+4)          -                                  7E-4        7E-3 55 Y, see Co                                  2E+5          6E-5    2E-7 28  Nickel-56      D, all compounds except        IE+3        2E+3          BE-7    3E-9          2E-5      2E-4 those given for W W, ox.ides, hydroxides, and      -        IE+3          SE-7    2E-9 carbides vapor                                      IE+3          SE-7    2E-9 28    Nickel-57      D, see 56Ni                  2E+3        5E+3          2E-6    7E-9          2E-5        2E-4 W, see Ni55 3E+3          IE-6    4E-9            -
Vapor                                      6E+3          3E-6    9E-9            -
28    Nickel-59      D. see 56 Ni                  2E+4        4E+3          2E-6    5E-9          3E-4        3E-3 W, see""Ni                                7E+3          3E-6    IE-8 vapor                                    2E+3          8E-7    3E-9 28    Nickel-63      D, see 56 Ni                9E+3        2E+3          7E-7    2E-9          IE-4        IE-3 56 W, see Ni                                3E+3          lE-6    4E-9            -
Vapor                                    8E+2          3E-7    IE-9 28    Nickel-65      D, see''"Ni                  8E+3        2E+4          IE-5    3E-8          I E-4      lE-3 1
W, see 'Ni                                3E+4          IE-5    4E-8 Vapor                                    2E+4          7E-6    2E-8 28    Nickel-66      D, see 56 Ni                4E+2        2E+3          7E-7      2E-9            -
LLI wall B-15                                                NUREG- 1736
 
APPENDIX B Table 1                      Table2          Table3 Occupational Values              Effluent        Releases to Concentrations      Sewers Col. I      Col.2        Col.3  Col. I      Col. 2 Oral                                                    Monthly Ingestion          Inhalation                            Average Atomic                                                ALI        ALI      DAC        Air        Water  Concentration No. Radionuclide              Class            (&#xb5;Ci)      (&#xb5;Ci)      (&#xb5;Ci/ml) (&#xb5;Ci/ml)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml)
(5E+2)          -            -      -        6E-6      6E-5 W,see 56Ni                      -        6E+2          3E-7  9E-10          -          -
Vapor                          -        3E+3          IE-6  4E-9          -          -
29  Copper-60'  21 D, all compounds except      3E+4        9E+4          4E-5    lE-7          -          -
those given forW and Y    St. wall (3E+4)          -            .      .        4E-4      4E-3 W, sulfides, halides, and      .        1E+5          SE-5    2E-7          -          -
nitrates Y*. oxides and hydroxides      -        IE+5        4E-5    lE-7          -          -
29  Copper-61      D, see 60Cu                    lE+4      3E+4          IE-5  4E-8        2E-4      2E-3 W, see Cu 60                    .        4E+4          2E-5    6E-8          .          -
00 Y, see cu                      -        4E+4          IE-5  5E-8          -          -
29  Copper-64      D, see "&deg;Cu                    IE+4      3E+4          IE-5  4E-8        2E-4      2E-3 W, see 60Cu                      -        2E+4          IE-5  3E-8          -          -
Y, see "&deg;Cu                      -        2E+4          9E-6    3E-8          -          -
60 29  Copper-67      D, see Cu                    5E+3        8E+3          3E-6    IE-8        6E-5      6E-4 W,see 60Cu                      -        SE+3          2E-6    7E-9          -          .
Y, see 00Cu                      -        5E+3          2E-6    6E-9          -          -
30  linc-62        Y. all compounds              IE+3        3E+3          IE-6  4E-9        2E-5      2E-4 121 30  linc-63        Y, all compounds              2E+4        7E+4          3E-5    9E-8          -          -
St. wall (3E+4)          -            .      .          3E-4      3E-3 30  linc-65        Y. all compounds            4E+2        3E+2          IE-7  4E-10        SE-6      SE-5 30  linc-69m        Y, all compounds            4E+3        7E+3          3E-6    IE-8        6E-5      6E-4 30  Zinc-69"'      Y, all compounds            6E+4        IE+5          6E-5    2E-7        8E-4      8E-3 30  Zinc-7 Im      Y, all compounds            6E+3        2E+4          7E-6    2E-8        8E-5      8E-4 30    Zinc-72        Y, call compounds            IE+3        IE+3          SE-7    2E-9        IE-5      IE-4 121 31    Jallium-65      D, all compounds except      5E+4        2E+5          7E-5    2E-7 those given for W          St. wall (6E+4)          -                                9E-4      9E-3 NUREG- 1736                                                B-16
 
APPENDIX B Table 1                      Table 2            Table 3 Occupational Values                Effluent        Releases to Concentrations        Sewers Col. I      Col. 2        Col. 3  Col. 1        Col. 2 Oral                                                      Monthly Ingestion          Inhalation                                Average Atomic                                            ALI        ALI      DAC          Air        Water  Concentration No. Radionuclide              Class        (&#xb5;Ci)      (&#xb5;Ci)      (&#xb5;Ci/ml)  (&#xb5;Ci/ml)      (&#xb5;Ci/ml)    (&#xb5;Ci/ml)
W, oxides, hydroxides,                2E+5          8E-5      3E-7          -
carbides, halides, and nitrates 31  Gallium-66    D, see 65 Ga              IE+3        4E+3          IE-6    SE-9        IE-5        IE-4 W, see 65 Ga                          3E+3          IE-6    4E-9 65 31  Gallium-67    D, see Ga                  7E+3        IE+4          6E-6      2E-8        IE-4        lE-3 65 W, see Ga                              IE+4          4E-6      ]E-8 31  Gallium-68"'  D, see 65 Ga              2E+4        4E+4          2E-5    6E-8          2E-4        2E-3 W, see 65Ga                            5E+4          2E-5    7E-8            -
2            65 31  Gallium-7fj >  D, see Ga                  5E+4        2E+5          7E-5      2E-7 St. wall (7E+4)          -                                lE-3        IE-2 W, see Ga 65 2E+5          8E-5      3E-7          -
31  Gallium-72    D, see 6 ~Ga                IE+3      4E+3          IE-6    SE-9        2E-5        2E-4 W, see Ga 65 3E+3          lE-6    4E-9            -
65 31  Gallium-73    D, see Ga                  5E+3        2E+4          6E-6      2E-8        7E-5        7E-4 W, see 65 Ga                          2E+4          6E-6      2E-8          -
32  Gennanium-66  D. all compounds except    2E+4        3E+4          IE-5    4E-8          3E-4        3E-3 those given for W W, oxides, sulfides, and              2E+4          8E-6      3E-8          -
halides 32  Germanium-6721 D, see "Ge                3E+4        9E+4          4E-5      lE-7 St. wall (4E+4)          -                                6E-4        6E-3 W, see 66Ge                            IE+S          4E-5      IE-7          -
32  Gennanium-68  D, see "Ge                5E+3        4E+3          2E-6    SE-9          6E-5        6E-4 W, seeMGe                              IE+2        4E-8    IE-IO          -
66 32  Gennanium-69  D, see Ge                  IE+4        2E+4          6E-6    2E-8          2E-4        2E-3 66 W, see Ge                              8E+3          3E-6      lE-8 32  Gennanium-7 I  D, seeMGe                  SE+S        4E+5          2E-4    6E-7          7E-3        7E-2 W, see "Ge                            4E+4          2E-5    6E-8 B-17                                                NUREG- 1736
 
APPENDIX B Table 1                      Table 2          Table 3 Occupational Values              Effluent        Releases to Concentrations        Sewers Col. I      Col.2        Col.3  Col. I        Col.2 Oral                                                    Monthly Ingestion          Inhalation                              Average Atomic                                      ALI          ALI      DAC          Air        Water  Concentration No. Radionuclide              Class  (&#xb5;Ci)        (&#xb5;Ci)    (&#xb5;Ci/ml) (&#xb5;Ci/ml)      (&#xb5;Ci/ml)  (&#xb5;Ci/ml) 32    Germanium-75'21 D,see 66Ge          4E+4        8E+4          3E-5    lE-7 St. wall
{7E+4)          -            -      -          9E-4      9E-3 W,see 66Ge            -        8E+4        4E-5    IE-7 32    Germanium-77    D, see 66Ge        9E+3        IE+4        4E-6    IE-8          IE-4      IE-3 W, see Ge66
                                                -        6E+3          2E-6  8E-9 32  Germanium-78121  D, see 66Ge        2E+4        2E+4          9E-6  3E-8 St. wall (2E+4)          -            -      -          3E-4      3E-3 W, see 66 Ge          -        2E+4          9E-6  3E-8 33    Arsenic-69"'    W, all compounds    3E+4        IE+5          SE-5  2E-7 St. wall (4E+4)          -            -      -          6E-4      6E-3 33  Arsenic-1rJ  2
                        > W. all compounds    IE+4        5E+4          2E-5  7E-8          2E-4      2E-3 33    Arsenic-7 I    W, all compounds    4E+3        5E+3          2E-6  6E-9          SE-5      SE-4 33    Arsenic-72      W, all compounds    9E+2        IE+3        6E-7    2E-9          IE-5      IE-4 33    Arsenic-73      W, all compounds    8E+3        2E+3        7E-7    2E-9          I E-4    IE-3 33  Arsenic-74      W, all compounds    IE+3        8E+2        3E-7    IE-9          2E-5      2E-4 33  Arsenic-76      W, all compounds    IE+3        IE+.i        6E-7    2E-9          IE-5      IE-4 33  A.rsenic-77      W. all compounds    4E+3        5E+3          2E-6    7E-9 LLI wall (5E+3)          -                                6E-5      6E-4 33    <\rsenic-78' 2'  vv. all compounds  8E+3        2E+4        9E-6    3E-8          IE-4      IE-3 NUREG- 1736                                        B-18
 
APPENDIX B Table 1                      Table 2          Table3 Occupational Values              Effluent        Releases to Concentrations        Sewers Col.1      Col.2    I    Col.3  Col.1          Col.2 Oral                                                    Monthly Ingestion          Inhalation                              Average Atomic                                                  ALI        ALI      DAC        Air        Water  Concentration No. Radionuclide                  Class            (&#xb5;Ci)      (&#xb5;Ci)      (&#xb5;Ci/ml) (&#xb5;Ci/ml)      (&#xb5;Ci/ml)  (&#xb5;Ci/ml) 34  Selenium-7rJ2>    D, all compounds except      2E+4        4E+4          2E-5    SE-8          ]E-4        IE-3 those given for W W, oxides, hydroxides,        1E+4      4E+4          2E-5    6E-8 carbides, and elemental Se 34  Seleniu*m-73m12l  D, see 70Se                  6E+4        2E+5          6E-5    2E-7        4E-4      4E-3 70 W, se Se                    3E+4        IE+S        6E-5    2E-7 70 34  Selenium-73        D, see Se                    3E+3        1E+4        SE-6    2E-8          4E-5      4E-4 W, see 70Se                    -        2E+4          7E-6    2E-8 70 34  Selenium-75        D. see Se                    5E+2        7E+2          3E-7    lE-9        7E-6      7E-5 W, see Se70
                                                          -        6E+2          3E-7  SE-10 34  Selenium-79        D, see 70Se                  6E+2        8E+2        3E-7    IE-9        8E-6      SE-5 W, see Se 70
                                                          -        6E+2        2E-7    SE-IO 34  Selenium-8 I m*~l  D, see Jose                  4E+4        7E+4          3E-5  9E-8          3E-4      3E-3 w, see Jose                  2E+4        7E+4          3E-5    IE-7 34  Selenium-81'2>    D. see Jo5e                  6E+4        2E+S          9E-5    3E-7 St. wall (8E+4)          -            -                  IE-3      IE-2 W, see Jose                      -        2E+5          lE-4  3E-7 34  Selenium-83"'    D, see "'Se                  4E+4        IE+5        SE-5    2E-7          4E-4      4E-3
: w. see Jose                  3E+4        IE+5        SE-5    2E-7 35  Bromine-74m'2l    D, bromides ofH, Li, Na,      IE+4        4E+4          2E-5    5E-8 K, Rb, Cs, and Fr            St. wall  I          I (2E+4)          -                                3E-4      3E-3 W, bromides oflanthanides,                4E+4          2E-5    6E-8 Be, Mg, Ca, Sr, Ba, Ra, Al.
Ga, In, Tl, Ge, Sn, Pb, As, Sb, Bi, Fe, Ru, Os, Co, Rh, Ir. Ni, Pd, Pt, Cu, Ag, Au, Zn, Cd, Hg, Sc, Y, Ti, Zr, Hf, V, Nb, Ta, Mn, Tc, and Re 35    Brornine-74' 2)  D. see 1*"'Br                2E+4        7E+4          3E-5    lE-7 St. Wall B-19                                              NUREG-1736
 
                                                                                              *---- ----------~-
APPENDIX B Table 1                      Table 2            Table3 Occupational Values              Effluent          Releases to Concentrations          Sewers Col. l      Col. 2        Col.3  Col. I      Col.2 Oral                                                        Monthly Ingestion          Inhalation                                Average Atomic                                      ALI          ALI      DAC        Air        Water    Concentration No.      Radionuclide            Class  (&#xb5;Ci)        (&#xb5;Ci)      (&#xb5;Ci/ml) (&#xb5;Ci/ml)      (&#xb5;Ci/ml)    (&#xb5;Ci/ml)
(4E+4)          -            -        -        SE-4        SE-3
: w. see  7
                                  *mer        -        8E+4          4E-5    lE-7            -          -
35    Bromine-75"'          74 D, see "!Br        3E+4        5E+4          2E-5    7E-8            -          -
St. wall (4E+4)          -            -      . -        SE-4        SE-3 74 W, see mBr            -        5E+4          2E-5    7E-8            -          -
35    Bromine-76    D, see nmBr        4E+3        5E+3          2E-6    7E-9          SE-5        SE-4 W, see ,.'"Br          -        4E+3          2E-6    6E-9            -          -
7 D, see *msr 3s    Bromine-77                        2E+4        2E+4          IE-5  3E-8          2E-4        2E-3 7
W, see *mer            -        2E+4          8E-6    3E-8            -          -
1 35    Bromine-80m    D, see *"'Br        2E+4        2E+4          7E-6    2E-8          3E-4        3E-3 W, see 7JmBr            -        IE+4        6E-6    2E-8            -          -
35    Bromine-8(1 >2        74 D, see mBr          5E+4        2E+5          8E-5    3E-7            -          -
St. wall
{9E+4)          -            -        -          IE-3        IE-2 74 W, see '"Br            -        2E+5          9E-5    3E-7            -          -
35    Bromine-82    D, see 7*mer        3E+3        4E+3          2E-6    6E-9          4E-5        4E-4 W, see 1*mer          -        4E+3          2E-6    SE-9            -          -
3s    Bromine-83            7 D, see *msr        5E+4        6E+4          3E-5    9E-8            -          -
St. wall (7E+4)          -            -        -        9E-4        9E-3 W, see 74 m8r          -        6E+4          3E-5    9E-8            -            -
1 35    Bromine-84"'  D, see *"'Br        2E+4        6E+4          2E-5    8E-8            -            -
St. wall (3E+4)          -            -        -        4E-4        4E-3 74 W, see "'Br            -        6E+4          3E-5    9E-8            -          -
36    Krypton-7412>  Submersion&deg;1          -          -          3E-6    lE-8            -            -
36    Krypton-76      Submersion111          -          -          9E-6    4E-8            -          -
36    Krypton-71  2
                    >  Submersion'!)          -          -          4E-6    2E-8            -            -
36    Krypton-79      Submersion' ii        -          -          2E-5    7E-8            -            -
NUREG-1736                                        B-20
 
APPENDIX B Table 1                    Table2          Table3 Occupational Values              Effluent      Releases to Concentrations      Sewers Col. I      Col.2        Col.3  Col. I      Col.2 Oral                                                    Monthly Ingestion          Inhalation                            Average Atomic                                                  ALI          ALI      DAC        Air      Water  Concentration No.      Radionuclide                Class          (&#xb5;Ci)        (&#xb5;Ci)      (&#xb5;Ci/ml) (&#xb5;Ci/ml)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml) 36  1 Krypton-8 I      SubmersionCll                    -          -          7E-4    3E-6          -        -
36    Krypton-83m<2>  Subrnersionm                    -          -          IE-2    5E-5          -        -
36    Krypton-85m      Submersion1n                    -          -          2E-5    lE-7          -        -
36    Krypton-85      Subrnersionm                    -          -          IE-4    7E-7          -        -
36    Krypton-81 >2 Submersion<n                    .          -          SE-6    2E-8 .        -        -
36    Krypton-88      Submersion&deg;>                    .          .          2E-6    9E-9          -        .
37    Rubidium-7g >  2 D, all compounds            4E+4          IE+S        SE-5    2E-7          -        -
St. wall (6E+4)          -            .      -        BE-4      8E-3 37    Rubidium-81 m'!'  D, all compounds            2E+5        3E+5          IE-4    5E-7          -        -
St. wall (3E+5)          -            -      .        4E-3      4E-2 37    Rubidium-81      D, all compounds            4E+4        5E+4          2E-5    7E-8        SE-4      SE-3 37    Rubidium-82m      D, all compounds              JE+4        2E+4          7E-6    2E-8        2E-4      2E-3 37    Rubidium-83      D, all compounds            6E+2        IE+3          4E-7    IE-9        9E-6      9E-5 37 I Rubidium-84        D, all compounds            5E+2        8E+2          3E-7    IE-9        7E-6      7E-5 37    Rubidium-86      D, all compounds            5E+2        8E+2          3E-7    IE-9        7E-6      7E-5 37    Rubidium-87      D, all compounds            IE+3        2E+3          6E-7    2E-9        IE-5      IE-4 37    Rubidium-8s<21    D, all compounds            2E+4        6E+4          3E-5    9E-8          -        .
St. wall (3E+4)          -            .      .        4E-4      4E-3 37    Rubidium-891! 1  D. all compounds            4E+4        IE+S          6E-5    2E-7          .        -
St. wall (6E+4)          -            .      .        9E-4      9E-3 21 38    Strontium-80'    D. all soluble compounds    4E+3        IE+4          SE-6    2E-8        6E-5      6E-4 except SrTiO3 Y, all insoluble compounds      .        IE+4          SE-6    2E-8          .          -
and SrTiO3 B-21                                            NUREG-1736
 
-\PPENDIX B Table 1                    Table2          Table3 Occupational Values              Effluent      Releases to Concentrations      Sewers Col. I      Col.2        Col.3  Col. I        Col.2 Oral                                                    Monthly Ingestion          Inhalation                            Average Atomic                                                ALI          ALI      DAC          Air      Water  Concentration No. Radionuclide                  Class        (&#xb5;Ci)        (&#xb5;Ci)    (&#xb5;Ci/ml) (&#xb5;Ci/ml)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml) 38  Strontium-81 12>    D, see 80Sr              3E+4        8E+4          3E-5    lE-7        3E-4      3E-3 80 Y, see Sr                  2E+4        8E+4          3E-5    IE-7          -          -
38  Strontium-82        D, see 80Sr                3E+2        4E+2        2E-7  6E-10          -          -
LLiwall (2E+2)          -            -      -        3E-6      3E-5 Y, see 80Sr                2E+2        9E+l        4E-8    lE-10          -          -
38  Strontium-83        D, seellOSr                3E+3        7E+3        3E-6    IE-8        3E-5      3E-4 Y, see "Sr                2E+3        4E+3          lE-6  5E-9          -          -
12        80 38  Strontium-8Sm >    D, see Sr                  2E+5        6E+5        3E-4    9E-7        3E-3      3E-2 80 Y, see Sr                    -          8E+5        4E-4    lE-6          -          -
110 38  Strontium-85        D, see Sr                  3E+3        3E+3          IE-6  4E-9        4E-5      4E-4 Y. see 110Sr                -          2E+3        6E-7    2E-9          -          -
80 38  Strontium-87m      D, see Sr                  5E+4        1E+5        SE-5    2E-7        6E-4      6E-3 80 Y, see Sr                  4E+4        2E+5        6E-5    2E-7          -          -
38  Strontium-89        D, see"&deg;Sr                6E+2        8E+2        4E-7    IE-9          -          -
LLiwall (6E+2)          -            -      -        SE-6      8E-5 Y, see "&deg;Sr                5E+2        IE+i        6E-8  2E-10          -          -
38  Strontium-90        D. seel!OSr                3E+l        2E+J        8E-9      -          -          -
Bone Surf    Bone Surf (4E+l)      (2E+I)          -  3E-l l        5E-7      5E-6 Y, see l!OSr                -          4E+0        2E-9  6E-12          -          -
38  Strontium-91        D, see"&deg;Sr                2E+3        6E+3        2E-6    8E-9        2E-5      2E-4 Y. see "&deg;Sr                  -          4E+3          IE-6  5E-9          -          -
38  Strontium-92        D, see 80Sr                3E+3        9E+3        4E-6    lE-8        4E-5      4E-4 Y, see 80Sr                  -          7E+3        3E-6    9E-9          -          -
2 39  Yttrium-86m'  '    W, all compounds except    2E+4        6E+4        2E-5    8E-8        3E-4      3E-3 those given for Y Y. oxides and hydroxides    -          5E+4        2E-5    SE-8          -          -
NUREG-I 736                                                  B-22
 
APPENDIX B Table 1                        Table 2            Table 3 Occupational Values                  Effluent        Releases to Concentrations        Sewers Col. I      Col.2        Col.3      Col. I      Col.2 Oral                                                          Monthly Ingestion            Inhalation                                  Average Atomic                                          ALI          ALI      I DAC            Air      Water    Concentration No. Radionuclide              Class    (&#xb5;Ci)    I (&#xb5;Ci)    I  (&#xb5;Ci/ml)  I (&#xb5;Ci/ml)    (&#xb5;Ci/ml)    (&#xb5;Ci/ml) 39    Yttrium-86    W, see 86mY            1E+3        3E+3            lE-6      SE-9        2E-5          2E-4 Y, see 86mY                          3E+3            lE-6    5E-9            -            -
39    Yttrium-87    W, see 86 mY            2E+3          3E+3          IE-6      5E-9          3E-5        3E-4 Y, see 86 mY      I    -    I  3E+3      I    IE-6  I    SE-9          -            -
39    Yttrium-88    w,  see 86mY            IE+3        38+2          IE-7      3E-10        IE-5        lE-4 Y, see 86 mY                        2E+2            IE-7    3E-10          -            -
39    Yttrium-90m    W,seeR6mY              8E+3        lE+4          5E-6      2E-8          IE-4        IE-3 Y, see R6my                -          IE+4          5E-6      2E-8          -            -
39    Yttrium-90    W, see somy            4E+2        7E+2          3E-7    9E-10            -            -
LLI wall (5E+2)          -              -        -          7E-6        7E-5 Y, see 86 mY              -        6E+2          3E-7      9E-10          -            -
39    Yttrium-91 mm  W, see Mmy              IE+S        2E+5            IE-4      3E-7        2E-3        2E-2 Y, see R6mY                -        2E+5          7E-5      2E-7            -            -
39    Yttrium-9 I    W, see 86mY            5E+2        2E+2          7E-8    2E-IO            -            -
LLlwall (6E+2)            -              -        -        BE-6        8E-5 Y, see 86mY                -        IE+2          5E-8    2E-10 39    Yttrium-92    W,see 86 '"Y            3E+3        9E+3          4E-6      IE-8        4E-5        4E-4 Y. see **my                -        8E+3          3E-6      IE-8 39    Yttrium-93    W, see 86my            IE+3        3E+3          IE-6      4E-9        2E-5        2E-4 86 Y, see "'Y                -        2E+3          lE-6      3E-9 39    Yttrium-94"'  W, see **my            2E+4        8E+4          3E-5      lE-7          -            -
St. wall (3E+4)            -            -          -        4E-4        4E-3 Y, see **my                -        8E+4          3E-5      IE-7 39    Yttrium-95 12
                      > W, see 86my            4E+4          2E+5          6E-5      2E-7          -            -
St. wall (5E+4)          -              -        -        7E-4    I    7E-3 Y, see 86my                -        IE+5          6E-5      2E-7          -
l      -
B-23                                                  NUREG-1736
 
APPENDIX B Table 1                            Table2                Table3 Occupational Values                    Effluent            Releases to Concentrations            Sewers Col. I        Col.2        Col.3        Col. I      Col.2 Oral                                                                Monthly Ingestion            Inhalation                                        Average Atomic                                              ALI          ALI      DAC              Air        Water      Concentration No. Radionuclide              Class          (&#xb5;Ci)        (&#xb5;Ci)      (&#xb5;Ci/ml)      (&#xb5;Ci/ml)    (&#xb5;Ci/ml)        (&#xb5;Ci/ml) 40    Zirconium-86    D, all compounds except    IE+3          4E+3          2E-6        6E-9        2E-5            2E-4 those given for W and Y W, oxides, hydroxides,                  3E+3          IE-6        4E-9          -
halides, and nitrates Y, carbide                                2E+3          IE-6        3E-9 40    Zirconium-88    D, see86zr                4E+3          2E+2          9E-8        3E-IO        5E-5            5E-4 8
W, see 6zr                                5E+2          2E-7        7E-IO            -
8 Y, see 6zr                              3E+2          lE-7        4E-IO          -
40    Zirconium-89    D, see""Zr                2E+3          4E+3          IE-6        5E-9        2E-5            2E-4 W, see "'Zr                              2E+3          IE-6        3E-9          -
Y. see **zr                              2E+3          IE-6        3E-9 40    Zirconium-93          8 D. see 6zr                1E+3          6E+0          3E-9            -
Bone Surf Bone Surf (3E+3)      I    (2E+I)    I    -          2E-1l    I    4E-5    I    4E-4 I                I- I              -            -
I 8
W, see 6zr                                2E+I            lE-8 Bone Surf
                                                      -    I    (6E+l)      I            I  9E-l l  I        -    I      -
Y. see 86Zr                              6E+I          2E-8 Bone Surf (7E+I)            -        9E-l 1          -
40    Zirconium-95    D. see""Zr                IE+3            IE+2 Bone Surf    I    5E-8 I  -          2E-5            2E-4 (3E+2)            -        4E-10          -
W, see **zr                              4E+2          2E-7        SE-10          -
Y, see""Zr                                3E+2          IE-7        4E-I0          -
40    Zirconium-97  D, see 86zr                6E+2          2E+3          8E-7          3E-9        9E-6          9E-5 W, see 86zr                              1E+3          6E-7          2E-9 Y, see B6zr                              IE+3          5E-7          2E-9 41    Niobium-88'"  W, all compounds except    5E+4          2E+5          9E-5        3E-7          -
those given for Y        St. wall
                                                  <7E+4\      I      -          -      I    -      I    IE-3      I      IE-2 NUREG- 1736                                            B-24
 
APPENDIX B Table 1                      Table 2          Table 3 Occupational Values                Effluent        Releases to Concentrations        Sewers Col. 1      Col.2        Col.3    Col.1        Col. 2 Oral                                                      Monthly Ingestion          Inhalation                              Average Atomic                                            ALI        ALI      DAC          Air        Water  Concentration No. Radionuclide              Class          (&#xb5;Ci)      (&#xb5;Ci)      (&#xb5;Cl/ml)  (&#xb5;Ci/ml)      (&#xb5;Ci/ml)  (&#xb5;Ci/ml)
Y, oxides and hydroxides                2E+S          9E-S    3E-7            -
2        88 41  Niobium-89m    W, see Nb                  1E+4        4E+4          2E-S    6E-8          IE-4      IE-3 (66min)
Y, see 88Nb                            4E+4          2E-S    SE-8 41  Niobium-89    W, see 88Nb                SE+3        2E+4          XE-6    3E-8          7E-S      7E-4 (122min)              88 Y, see Nb                              2E+4          6E-6      2E-8 41  Niobium-90    W, see 88Nb                1E+3        3E+3          lE-6    4E-9          lE-5      IE-4 Y, see 88Nb                            2E+3          lE-6    3E-9            -
8 41  Niobium-93m    W, see RNb                  9E+3        2E+3          SE-7    3E-9 LLiwall (IE+4)          -                                2E-4      2E-3 88 Y, see Nb                              2E+2          7E-8    2E-1O            -
41  Niobium-94    W, see 88Nb                9E+2        2E+2          SE-8    3E-IO          IE-5      IE-4 Y, see 88 Nb                            2E+I          6E-9    2E-ll            -
8 41  Niobium-95m    W, see" Nb                  2E+3        3E+3          1 E-6    4E-9 LLiwall (2E+3)          -                                3E-5      3E-4 88 Y, see Nb                              2E+3          9E-7      3E-9            -
41  Niobium-95    W, see 88Nb                2E+3        1E+3          SE-7      2E-9        3E-S      3E-4 Y, see 88Nb                              IE+3        SE-7      2E-9            -
41  Niobium-96    W, see""Nb                  1E+3      3E+3          I E-6    4E-9          2E-5      2E-4 88 Y. see Nb                              2E+3          I E-6    JE-9            -
41  Niobium-91 2>  W, see 88 Nb                2E+4        8E+4          JE-5      IE-7        3E-4      3E-3 Y, see 88Nb                            7E+4          JE-5      IE-7          -
41  Niobium-98"'  W, see""Nb                  1E+4      5E+4          2E-5      SE-8        2E-4      2E-3 Y, see""Nb                              5E+4          2E-5    7E-8            -
42  Molybdenum-90  D, all compounds except    4E+3        7E+3          3E-6      IE-8        3E-5      3E-4 those given for Y Y,oxides,hydroxides,and    2E+3        SE+3          2E-6    6E-9            -
MoS 2 42  Molybdenum-    D, see 90 Mo                9E+3        2E+4          7E-6      2E-8        6E-5      6E-4 B-25                                                NUREG- 1736
 
APPENDIX B Table 1                      Table2            Table3 Occupational Values              Effluent        Releases to Concentrations        Sewers Col. I      Col.2        Col.3  Col. I        Col.2 Oral                                                    Monthly Ingestion          Inhalation                              Average Atomic                                              ALI        ALI      DAC          Air        Water  Concentration No.      Radionuclide              Class        (&#xb5;Ci)      (&#xb5;Ci)      (&#xb5;Ci/mJ) (&#xb5;Ci/mJ)      (&#xb5;Ci/mJ)  (&#xb5;Ci/mJ)
Y, see 90Mo                4E+3        IE+4          6E-6    2E-8            -          -
42    Molybdenum-93  D,see~o                    4E+3        5E+3          2E-6    8E-9        SE-5      5E-4 Y,see~o                    2E+4        2E+2          SE-8    2E-10          -          -
42    Molybdenum-99  D,see~o                    2E+3        3E+3          lE-6    4E-9          -          -
LLI wall (IE+3)          -            -        -          2E-5      2E-4 Y,see~o                    IE+3        IE+3        6E-7    2E-9          -          -
42    Molybdenum-    D.see~o                    4E+4        lE+S          6E-5    2E-7          -          -
101 121                                  St. wall (5E+4)          -            -        -          7E-4      7E-3 Y,see~o                      -        IE+S          6E-5    2E-7          -          -
43    Technetium-    D, all compounds except    7E+4        2E+5          6E-5    2E-7        IE-3      IE-2 93m12)          those given for W W, oxides, hydroxides,        -        3E+5          IE-4    4E-7            -          -
halides, and nitrates 43    Technetium-93  D, see 93 "'Tc            3E+4        7E+4          3E-5    !E-7        4E-4      4E-3 93 W, see "'Tc                  -        IE+5          4E-5    !E-7          -          -
43    Technetium-    D, see 93 "'Tc            2E+4        4E+4          2E-5    6E-8          3E-4      3E-3 94m< 2i W, see 93 "'Tc                -        6E+4          2E-5    8E-8          -          -
93 43    Technetium-94  D, see "'Tc                9E+3        2E+4          SE-6    3E-8          IE-4      IE-3 93 W, see "'Tc                            2E+4          lE-5    3E-8            -
93 43    Technetium-95m  D, see "'Tc                4E+3        5E+3          2E-6    8E-9          5E-5      5E-4 93 W, see "'Tc                            2E+3          SE-7    3E-9          -
93 43    Technetium-95  D, see "'Tc                IE+4        2E+4          9E-6    3E-8          IE-4      IE-3 W, see 93"'Tc                          2E+4          SE-6    3E-8            -
93 43    Technetium-    D, see "'Tc                2E+5        3E+5          IE-4    4E-7          2E-3      2E-2
::l6m12>                93 W, see "'Tc                            2E+5          IE-4    3E-7            -
43    Technetium-96  D, see 93 "'Tc            2E+3        3E+3          IE-6    SE-9          3E-5      3E-4 W, see 93 "'Tc                        2E+3          9E-7    3E-9            -
43    fechnetium-97m        93 D, see "'Tc                5E+3        7E+3          3E-6      -          6E-5      6E-4 St. wail NUREG- 1736                                              B-26
 
APPENDIX B Table 1                      Table2          Table3 Occupational Values                Effluent      Releases to Concentrations      Sewers Coll        Col.2        Col.3    Col. I      Col.2 Oral                                                    Monthly Ingestion          Inhalation                              Average Atomic                                              ALI          ALI      DAC          Air        Water  Concentration No. Radionuclide                CIB5S        (&#xb5;Ci)        (&#xb5;Ci)      (&#xb5;Ci/m1)  (&#xb5;Ci/m1)    (&#xb5;Ci/mi)  (&#xb5;Ci/ml)
                                                        -      (7E+3)            -      lE-8            -        -
93 W, see "'Tc                  -        1E+3        SE-7      2E-9            -        -
43  Technetium-97      D, see 93"'Tc            4E+4        5E+4          2E-5    ?E-8      . SE-4      5E-3 93 W, see "'Tc                    -        6E+3          2E-6    SE-9            -        -
43  Technetium-98      D, see 93"'Tc              IE+3        2E+3          7E-7      2E-9          lE-5      lE-4 93 W, see "'Tc                    -        3E+2          lE-7*  4E-10            -        -
93 43  Technetium-99m    D, see "'Tc                8E+4        2E+5          6E-5      2E-7          lE-3      lE-2 93 W, see "'Tc                    -        2E+5          IE-4    3E-7            -        -
43  Technetium-99            93 D, see "'Tc                4E+3        5E+3          2E-6        -          6E-5      6E-4 St. wall
                                                        -      (6E+3)            -      SE-9            -        -
93 W, see "'Tc                    -        7E+2          3E-7    9E-10            -        -
43    Technetium-101(2) 93 D, see "'Tc                9E+4        3E+5          lE-4    5E-7            -        -
St. wall (1E+5)          -            -        -          2E-3      2E-2 W, see 93 "'Tc                -        4E+5          2E-4    5E-7            -        -
43    Technetium-              93 D, see "'Tc                2E+4        7E+4          3E-5      lE-7          -        -
104121                                      St. wall (3E+4)          -            -        -          4E-4      4E-3 W. see 93 "'Tc                -        9E+4          4E-5    lE-7            -        -
121 44    Ruthenium-94      D. all compounds except    2E+4        4E+4          2E-5    6E-8          2E-4      2E-3 those given for W and Y W, halides                    -        6E+4          3E-5    9E-8            -        -
Y, oxides and hydroxides      -        6E+4          2E-5    8E-8            -        -
44    Ruthenium-97      D, see<NRu                8E+3        2E+4          8E-6    3E-8          IE-4      lE-3 W, see Ru94
                                                      -        IE+4        5E-6      2E-8            -        -
04 Y. see Ru                    -        IE+4        5E-6      2E-8            -        -
44    Ruthenium- I 03  D, see 94Ru                2E+3        2E+3          7E-7      2E-9        3E-5      3E-4 94 W, see Ru                    -        IE+3          4E-7      IE-9            -        -
94 Y, see Ru                    -        6E+2          3E-7    9E-10            -        -
B-27                                              NUREG- 1736
 
APPENDIX B Table 1                        Table2              Table3 Occupational Values                  Effluent          Releases lo Concentrations          Sewers Col.1      Col.2        Col.3      Col.1        Col.2 Oral                                                          Monthly Ingestion          Inhalation                                  Average Atomic                    (,t                        ALI        ALI      DAC            Air        Water    Concentration No. Radionuclide                Class          (&#xb5;Ci)      (&#xb5;Ci)      (&#xb5;Ci/ml)  (&#xb5;Ci/ml)      (&#xb5;Ci/ml)    (&#xb5;Ci/ml) 44  Ruthenium- I 05  D, see 94Ru                5E+3        IE+4        6E-6      2E-8          7E-5        7E-4 W, see 94 Ru                  -        IE+4        6E-6      2E-8            -            -
94 Y, see Ru                    -          IE+4        5E-6      2E-8            -            -
44    Ruthenium- 106          94 D, see Ru                  2E+2        9E+l          4E-8    . IE~lO          -            -
LLI wall (2E+2)          -            -          -          3E-6        3E-5 W,see 'Ru9
                                                          -        SE+l          2E-8      SE-11            -            -
94 Y. see Ru                      -        IE+l        SE-9*    2E-ll            -            -
45    Rhodium-99m      D, all compounds except    2E+4        6E+4          2E-5      8E-8          2E-4        2E-3 those given for W and Y W, halides                    -        8E+4          3E-5        IE-7          -            -
Y, oxides and hydroxides      -        7E+4          3E-5      9E-8            -            -
45  Rhodium-99        D,see~h                    2E+3        3E+3          lE-6      4E-9          3E-5        3E-4 W, see 99"'Rh                -        2E+3          9E-7      3E-9            -            -
99 Y, see "'Rh                  -        2E+3          8E-7      3E-9            -            -
45    ~hodium-100      D, see 9\lmRh              2E+3        5E+3          2E-6      7E-9          2E-5        2E-4 99 W, see "'Rh                  -        4E+3          2E-6      6E-9            -            -
Y, see 99"'Rh                -        4E+3          2E-6      SE-9            -            -
99 45    ~hodium-1O l m  D, see "'Rh                6E+3        IE+4          SE-6      2E-8          8E-5        SE-4 99 W, see "'Rh                    -        8E+3          4E-6      JE-8            -            -
Y, see 99"'Rh                  -        8E+3          3E-6      JE-8            -            -
911 45    ~hodium-1O1      D, see "'Rh                  2E+3        5E+2          2E-7      7E-10          3E-5        3E-4 911 W, see mRh                    -        8E+2          3E-7      lE-9            -            -
911 Y,see "'Rh                    -        2E+2          6E-8      2E-1O            -            -
45    ~hodium- I02m    D, see99mRh                  IE+3        5E+2          2E-7    7E-10            -            -
LLI wall (IE+3)          -            -          -          2E-5        2E-4 99 W, see "'Rh                    -        4E+2          2E-7    SE-1O            -            -
99 Y, see "'Rh                    -        1E+2          SE-8    2E-1O            -            -
45    ~hodium-102      D, see 99"'Rh              6E+2      I  9E+l    I    4E-8  I  IE-10    I    SE-6    I    8E-5 NUREG- 1736                                                B-28
 
APPENDIX B Table I                    Table2          Table3 Occupational Values              Effluent      Releases to Concentrations      Sewers Col. I      Col.2        Col.3  Col.I      Col.2 Oral                                                  Monthly Ingestion          Inhalation                            Average Atomic                                                ALI        ALI      DAC        Air      Water  Concentration No.      Radionuclide                  Class        (&#xb5;Ci)      (&#xb5;Ci)      (&#xb5;Ci/ml) (&#xb5;Ci/ml)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml)
W,see!l'JmRh                  -        2E+2          7E-8  2E-IO          -          -
99 Y, see mRh                    -        6E+l          2E-8  8E-11          .          -
45    Rhodium-I 03mm    D,see 99 ~h                4E+5        1E+6          5E-4    2E-6*        6E-3      6E-2 W,see 99 ~h                  .        1E+6          5E-4    2E-6          -          .
Y,see 99~h                    -        1E+6          5E-4    2E-6 .        .          -
45    Rhodium- I 05    D, see 99 mRh              4E+3        IE+4*        SE-6  2E-8          -          -
LLI wall (4E+3)          -            -      -          5E-5      SE-4 W, see !l'JmRh                -        6E+3          3E-6  9E-9          -          -
Y, see 'l'l"'Rh              -        6E+3          2E-6  8E-9          -          -
45    Rhodium- I 06m    D,see!l'J~h                8E+3        3E+4          IE-5  4E-8        IE-4      lE-3 W,see99mRh                    -        4E+4          2E-5  SE-8          -          -
99 Y, see mRh                    -        4E+4          IE-5  5E-8          -          -
2        99 45    Rhodium-101 >    D, see "'Rh                7E+4        2E+5          IE-4  3E-7          -          -
St. wall (9E+4)          -            -      -          IE-3      IE-2 W, see 'l'l"'Rh              -        3E+5          lE-4  4E-7          -          -
Y, see 'l'lmRh                -        3E+5          IE-4  3E-7          -          -
46    Palladium- I00  D, all compounds except    IE+3        IE+3          6E-7    2E-9        2E-5      2E-4 those given for W and Y W, nitrates                    -        IE+3          SE-7    2E-9          -          -
Y, oxides and hydroxides      -        IE+3          6E-7    2E-9          -        -
1 46    Palladium-IOI    D, se~ Ci0pd                IE+4        3E+4          IE-5  5E-8        2E-4      2E-3 W, see 1Ci0pd                  -        3E+4          IE-5    5E-8          -          -
1 Y, see Ci0pd                          3E+4          IE-5  4E-8 46    Palladium- I 03  D, see 100pd              6E+3        6E+3          3E-6    9E-9          -
LLI wall (7E+3)          -                              IE-4      IE-3 W, see 100pd                          4E+3          2E-6    6E-9          -
1 Y, see Ci0pd                          4E+3          I E-6  5E-9          -
B-29                                            NUREG-1736
 
APPENDIX B Table 1                    Table2          Table3 Occupational Values            Effluent        Releases to Concentrations        Sewers Col. I      Col.2        Col.3  Col. I        Col.2 Oral                                                    Monthly Ingestion          Inhalation                            Average Atomic                                              ALI          ALI      DAC        Air        Water  Concentration No. Radionuclide                Class        (&#xb5;Ci)        (&#xb5;Ci)    (&#xb5;Ci/ml) (&#xb5;Ci/ml)    (&#xb5;Ci/ml)  (&#xb5;_Ci/ml) 46  Palladium- 107  D, see 100pd                3E+4        2E+4        9E-6      -            .          -
LLlwall      Kidneys (4E+4)        (2E+4)          -    3E-8        5E-4      SE-3 W,see 00pd1
                                                      -          7E+3        3E-6    IE-8            -        -
1 Y, see 00pd                  -          4E+2        2E-7  6E-10            -        -
46  Palladium- 109  D, see 100pd                2E+3        6E+3        3E-6    9E-6        3E-5      3E-4 W. see 100pd                  -          5E+3        2E-6    8E-9            -          -
1 Y, see 00pd                  -          5E+3        2E-6    6E-9            -        -
47  Silver- I 0212
                      > D, all compounds except    5E+4        2E+5        8E-5    2E-7          -          -
those given for W and Y  St. wall (6E+4)          -            -      -          9E-4      9E-3 W, nitrates and sulfides      -          2E+5        9E-5    3E-7          -          -
Y. oxides and hydroxides      -          2E+5        8E-5    3E-7          -          -
102 47  Silver- 103(2>  D, see      Ag            4E+4        lE+S        4E-5    IE-7        SE-4      SE-3 W, see    102 Ag              -          lE+S        SE-5    2E-7          -          -
102 Y, see      Ag                -          lE+S        SE-5    2E-7          -          -
47  Silver-104m12>  D, see ,u 2Ag              3E+4        9E+4        4E-5    1E*7        4E-4      4E-3 W, see '02 Ag                -          IE+S        SE-5    2E-7          -          -
Y, see  102 Ag                -          IE+S        SE-5    2E-7          -          -
102 47  Silver- l 0412> D, see      Ag              2E+4        7E+4        3E-5    IE-7        3E-4      3E-3 W,see    102 Ag              -          lE+S        6E-5    2E-7          -          -
Y, see  102 Ag                -          lE+S        6E-5    2E-7          -          -
102 47  Silver-105      D. see      Ag              3E+3        IE+3        4E-7    IE-9        4E-5      4E-4 W, see rn2Ag                  -          2E+3        7E-7    2E-9          -          -
Y, see  102 Ag                -          2E+3        7E-7    2E-9          -          -
102 47    Silver- 106m    D, see      Ag              8E+2        7E+2        3E-7    IE-9        lE-5      lE-4 W, see  102 Ag              -          9E+2        4E-7    lE-9          -          -
                        '{, see 102 Ag                -          9E+2        4E-7    lE-9          -          -
47    Silver-106'"    D, see  102 Ag              6E+4        2E+5        8E-5    3E-7          -          -
St. wall NUREG- 1736                                              B-30
 
APPENDIX B Table 1                      Table 2          Table 3 Occupational Values              Effluent        Releases to Concentrations        Sewers Col. 1      Col.2        Col.3  Col. 1        Col.2 Oral                                                    Monthly Ingestion          Inhalation                              Average Atomic                                              AL1        ALI      DAC          Air        Water  Concentration No.      Radionuclide                Class      (&#xb5;Ci)      (&#xb5;Ci)      (&#xb5;Ci/ml) (&#xb5;Ci/ml)      (&#xb5;Ci/ml)  (&#xb5;Ci/ml)
(6E+4)          -                                9E-4      9E-3 W, see    102 Ag                      2E+5          9E-5    3E-7            -
02 Y, see ' Ag                            2E+5          8E-5  3E-7            -
47  Silver-108m    D, see 102Ag                6E+2        2E+2          8E-8  3E-10          9E-6      9E-5 W, see    102 Ag                      3E+2          lE-7  4E-10            -
102 Y, see      Ag                        2E+l          IE-8  3E-11            -
47  Silver- 11 Om    D, see ' 02Ag              5E+2        IE+2        SE-8  2E-10          6E-6      6E-5 102 W, see        Ag                      2E+2          8E-8  3E-10            -
102 Y, see        Ag                        9E+I          4E-8    IE-10            -
47  Silver-I 11    D, see '02Ag                9E+2        2E+3          6E-7 LLI wall      Liver (IE+3)      (2E+3)            -    2E-9          2E-5      2E-4 W. see 102Ag                            9E+2          4E-7    lE-9            -
02 Y, see ' Ag                            9E+2          4E-7    lE-9 02 47  Silver-I 12    D, see ' Ag                3E+3        8E+3          3E-6    IE-8        4E-5      4E-4 W, see 102 Ag                          IE+4          4E-6    IE-8 Y, see 102Ag                            9E+3          4E-6    IE-8 47  Silver- 11 5121 D, see Ag102 3E+4        9E+4          4E-5    IE-7            -
St. wall (3E+4)          -                                4E-4      4E-3 102 W, see        Ag                        9E+4          4E-5    IE-7 Y, see  102 Ag                        8E+4          3E-5    IE-7            -
48  Cadmium-104m    D, all compounds except    2E+4        7E+4          3E-5    9E-8          3E-4      3E-3 those given for W and Y W. sulfides, halides, and              IE+5          SE-5    2E-7            -
nitrates Y, oxides and hydroxides                1E+5          5E-5    2E-7 48    Cadmium- 107    D, see 104Cd                2E+4      5E+4          2E-5    BE-8          3E-4      3E-3 04                                                                -
W, see' Cd                            6E+4          2E-5    BE-8 104 Y, see Cd                              5E+4          2E-5  7E-8            -
B-31                                              NUREG-1736
 
APPENDIX B Table 1                            Table 2            Table 3 Occupational Values                    Effluent          Releases to Concentrations          Sewers Col. I      Col.2          Col.3        Col. I        Col.2 Oral                                                              Monthly Ingestion          Inhalation Atomic No. Radionuclide                Class ALI
(&#xb5;Ci)    I  ALI
(&#xb5;Ci)
I DAC
(&#xb5;Ci/ml)
Air
(&#xb5;Ci/ml)
Water
(&#xb5;Ci/ml)
Average Concentration
(&#xb5;Ci/ml)
I                I- I - I I I                            --
1 48    Cadmium- 109    D, see 0-ICd          3E+2            4E+l            IE-8                    -
Kidneys        Kidneys (4E+2)    I    (SE+!)    I          I    7E-l l  I    6E-6        6E-5 I            II - I II -- I I                                -
1().1Cd W, see                            IE+2          SE-8            -
Kidneys
                                                -    I  (1E+2)                        2E-10                        -
Y, see 104 Cd                      IE+2          5E-8        2E-10            -
104                                        IE-9 48    :::admium-l 13m D, see    Cd        2E+l        2E+0 Kidneys      Kidneys (4E+l)      (4E+O)            -          SE-12          5E-7        5E-6 W,see 10-ICd                      8E+0          4E-9            -
Kidneys
                                                -    I  (IE+!)      I    -          2E-11            -            -
Y, see 104 Cd                  I  IE+!    I    SE-9    I    2E-11      I    - I        -
NUREG- 1736                                        B-32
 
APPENDIX B 1ao1e 1                      *1able l          1ame.:1 Occupational Values              Effluent          Releases to Concentrations          Sewers Col. I      Col.2        Col.3  Col. I          Col. 2 Oral                                                      Monthly Ingestion          Inhalation                                Average Atomic                                            ALI        ALI      DAC        Air          Water    Concentration No. Radionuclide              Class        (&#xb5;Ci)      (&#xb5;Ci)      (&#xb5;Ci/ml) (&#xb5;Ci/ml)      (&#xb5;Ci/ml)    (&#xb5;Ci/ml) 48  Cadmium- 113  D, see 111.iCd            2E+t        2E+O          9E-to      -
Kidneys      Kidneys (3E+t)      (3E+O)          -    5E-12            4E-7      4E-6 W, see 1&deg;td                            8E+0          3E-9 Kidneys (lE+t)          -    2E-tl              -
Y, see  11).1Cd tE+l        6E-9    2E-tt              -
1                                                                              4E-5 48  Cadmium-I 15m  D, see &deg;'Cd                3E+2        5E+l          2E-8                    4E-6 Kidneys (8E+l)          -    IE-10              -
W, see 104Cd                            IE+2          SE-8  2E-10              -
1 Y, see &deg;'Cd                            IE+2          6E-8  2E-10              -
48  Cadmium- 115  D, see '&deg;'Cd              9E+2        IE+3          6E-7    2E-9 LLiwall (IE+3)          -                                  IE-5      1 E-4 W, see '&deg;'Cd                            tE+3        5E-7    2E-9              -
Y, see ""Cd                            tE+3        6E-7    2E-9 104 48  Cadmium-117m  D, see Cd                  5E+3        tE+4          5E-6    2E-8            6E-5      6E-4 04
                      'W, see ' Cd                          2E+4          7E-6    2E-8 Y, see 104Cd                          tE+4          6E-6    2E-8 48  Cadmium- 117  D. see ""Cd                5E+3        IE+4        5E-6    2E-8          6E-5      6E-4 1
W, see u.iCd                          2E+4          7E-6    2E-8 Y. see ""Cd                            IE+4          6E-6    2E-8 49  Indium-109    D, all compounds except    2E+4        4E+4          2E-5    6E-8            3E-4      3E-3 those given for W W, oxides, hydroxides,                6E+4          3E-5    9E-8              -
halides, and nitrates 49  Indium-I IO'z' D. see 10Yln              2E+4        4E+4          2E-5    6E-8            2E-4      2E-3 (69.lmin)
W, see ""In                            6E+4          2E-5    8E-8              -
49  Indium- I IO  D, see ,min                5E+3        2E+4          7E-6    2E-8            7E-5      7E-4 (4.9h)
W,see'"ln                              2E+4          8E-6    3E-8              -
B-33                                                  NUREG- 1736
 
APPENDIX B Table 1 Occupational Values Table 2 Eftluent Concentrations I    Table3 Releases to Sewers Col. I      Col. 2      Col. 3    Col. I      Col.2 Oral                                                        Monthly Ingestion          Inhalation                                Average Atomic                                                    ALI          ALI    IDAC            Air        Water  Concentration No.      Radionuclide                        Class      (.uCn        <.uCi)    l.uCi/ml) (&#xb5;Ci/ml)    (&#xb5;Ci/ml)    (&#xb5;Ci/ml) 49    Indium- 111          D, see 1~n            I  4E+3      I  6E+3    I    3E-6    9E-9        6E-5        6E-4 1
W, see ooin                            6E+3          3E-6    9E-9          -
2                  1 49    Indium-I 1z< i        D, see ooin                2E+5        6E+5          3E-4    9E-7        2E-3        2E-2 W, see 1ooin                  -        7E+5          3E-4    IE-6          -            .
49    Indium-l 13mc2i      D, see 1ooin              5E+4          IE+5          6E-5    2E-7        7E-4          7E-3 1                                                                    .
W, see ooin                            2E+5          BE-5    3E-7                        -
49    Indium-I 14m                  1 D, see @in                3E+2        6E+l          3E-8  9E-ll            -            -
LLiwall (4E+2)          -            .        .        SE-6        5E-5 W, see *win                              IE+2          4E-8    IE-10          .            -
49    Indium- I ISm        D, see 1&deg;"In                IE+4        4E+4          2E-5    6E-8        2E-4        2E-3 W, see IOY!n                            5E+4          2E-5    7E-8          -
1 49  Indium-I 1.5          D, see cwin                4E+I          IE+O        6E-10  2E-12        SE-7        SE-6 1
W, see cwin                            5E+O          2E-9  BE-12          .            -
21          1119 lndium- I 16m'        D. see      1n            2E+4        8E+4          3E-5      IE-7        3E-4        3E-3 W, see *win                              IE-5        SE-5    2E-7          .            -
12            1 49  Indium- I I 7m      i D, see ~n                  IE+4        3E+4          IE-5    SE-8        2E-4        2E-3 109                                                                  .
W, see        1n                        4E+4          2E-5    6E-8                        -
2 49  Indium- 111    '      D. see 11191n              6E+4        2E+S          7E-5    2E-7        BE-4        8E-3 1
W, see &deg;"1n                              2E+5          9E-5    3E-7          -
49  Indium-I 19m    12 i
1 D. see cwin                4E+4        IE+S          SE-5    2E-7          .
St. wall (5E+4)          .                    -          7E-4  I    7E-3 1
W, see cwin                              IE+S          6E-5    2E-7      I      -  I      -
50  Tin-110              D. all compounds except those given for W 4E+3        IE+4          SE-6    2E-8          SE-5 I  SE-4 W, sulfides, oxides,                    1E+4          SE-6    2E-8 hydroxides, halides, nitrates, and stannic phosphate NUREG- 1736                                                    B-34
 
APPENDIX B Table 1                          Table 2          Table 3 I                                      I Occupational Values                  Effluent        Releases to Concentrations        Sewers I                                            Col. 1      CoJ.2        Col. 3      Col. 1      Col. 2 Oral                                                        Monthly Ingestion          Inhalation                                Average Atomic                                              ALl        ALI      DAC            Air        Water  Concentration No.      Radionuclide                  Class      (&#xb5;Ci)      (&#xb5;Cl)      (&#xb5;Ci/ml)    (&#xb5;Ci/ml)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml) so    Tin-I 11'"      D, see 110sn              7E+4        2E+5          9E-5        3E-7          lE-3      IE-2 W, see 110Si:i                        3E+5          IE-4      4E-7 50    Tin-I 13        D, see 110sn              2E+3 LU wall    I    IE+3 I      SE-7 I    2E-9 I- I        -
(2E+3)          -                                  3E-5        3E-4
: w. see 110sn        -                SE+2          2E-7      BE-10          -
50    Tin-117m          D, see  110 Sn            2E+3        1E+3          SE-7        -
LU wall    Bone Surf (2E+3)      (2E+3)            -        3E-9        3E-5      3E-4 I W, see  110 sn                      IE+3          6E-7        2E-9            -
so    Tin-I 19m                110 D, see sn                3E+3        2E+3          IE-6      3E-9            -
LU wall (4E+3)          -                                  6E-5      6E-4 110 W, see      sn                        1E+3        4E-7        lE-9 so    Tin-12lm          D, see 110sn              3E+3        9E+2          4E-7        lE-9 LU wall (4E+3)        -                                    SE-S      SE-4 I W, see  110 Sn                        5E+2          2E-7      BE-10          -
so    Tin-121          D, see 110 Sn            6E+3 LLlwall    I    2E+4 I      6E-6 I    2E-8 I- I      -
(6E+3)          -                                    BE-5      BE-4 110 W, see      Sn                        IE+4          SE-6        2E-8 110 50    Tin-123m<21      D, see      Sn            SE+4        IE+S          SE-S        2E-7        7E-4      7E-3 W, see 110Sn                          IE+S          6E-5        2E-7          -
50    Tin- 123                1111 D, see Sn                5E+2      6E+2          3E-7      9E-IO          -          -
LLI wall (6E+2)          -            -          -          9E-6        9E-S W, see  110 Sn              -        2E+2*          7E-8      2E-10          -          -
B-35                                                  NUREG- 1736
 
APPENDIX B Table I                      Table2            Table3 Occupational Values              Effluent        Releases to Concentrations        Sewers Col.I      Col.2        Col.3  Col.I        Col.2 Oral                                                    Monthly Ingestion          Inhalation                              Average Atomic                                                      ALL        ALI      DAC          Air        Water  Concentration No.      Radionuclide                  Class            (&#xb5;Ci)      (&#xb5;Ci)      (&#xb5;Ci/ml) (&#xb5;Ci/ml)      (&#xb5;Ci/ml)  (&#xb5;Ci/ml) 50    Tin- 125          D, see  110 Sn                  4E+2        9E+2          4E-7    lE-9            -          -
LLI wall (5E+2)          -            -      -          6E-6      6E-5 W,see    110 Sn                    -        4E+2          lE-7  SE-10 *          -          -
110 50    Tin- 126          D, see      Sn                  3E+2        6E+I          2E-8  SE-I I        4E-6      4E-5 W, see    110 Sn                    -        7E+l          3E-8  9E-l 1            -          -
110 so    Tin- 127          D, see      sn                  7E+3        2E+4          SE-6  3E-8          9E-5      9E-4
: w. see 110sn                      -        2E+4          8E-6  3E-8            -          -
110 50    Tin-128(2)        D, see      Sn                  9E+3        3E+4          lE-5  4E-8          IE-4      IE-3 110 W, see      Sn                    -        4E+4          lE-5  5E-8            -          -
51    Antimony- I I s- >
2 D, all compounds except        8E+4        2E+5          JE-4  3E-7          lE-3      lE-2 those given for W W, oxides, hydroxides,            -        3E+5          lE-4  4E-7            -          -
halides, sulfides, sulfates, and nitrates 115 51    Antimony-          D, see      Sb                  2E+4        7E+4          3E-5    IE-7          3E-4      3E-3 ll6m' 2>
W,see    115 Sb                    -        IE+S          6E-5    2E-7            -          -
51    Antimony-I t6<  2
                          > D, see  115 Sb                  7E+4        3E+5          lE-4  4E-7            -          -
St. wall (9E+4)          -              -      -          IE-3      IE-2 W,see    115 Sb                  -        3E+5          IE-4    SE-7            -          -
          <\ntimony-117            115 51                        D. see      Sb                7E+4        2E+5          9E-5    3E-7          9E-4      9E-3 115 W.see        Sb                  -        3E+5          lE-4  4E-7            -          -
115 51    <\ntimony-118m    D, see    Sb                  6E+3        2E+4          8E-6  3E-8          7E-5      7E-4 W, see  115 Sb                5E+3        2E+4          9E-6  3E-8            -          -
115 51    <\ntimony-119      D. see    Sb                  2E+4        SE+4          2E-5    6E-8          2E-4      2E-3 W, see 1"Sb                    2E+4        3E+4          IE-5  4E-8            -          -
51    <\ntimony-120      D, see 115 Sb                  lE+S        4E+5          2E-4    6E-7            .          -
:J6min)                                        St. wall (2E+S)          -            -      -          2E-3      2E-2 115 W, see      Sb                    -        5E+5          2E-4    7E-7            -          -
NUREG-I 736                                                      B-36
 
APPENDIX B Table 1                        Table 2            Table 3 Occupational Values                Effluent          Releases to Concentrations          Sewers Col.1    I  Col. J        Col.3    Col. I        Col.2 Oral                                                        Monthly Ingestion I        Inhalation                                  Average Atomic                                          ALl        ALl      DAC            Air        Water    Concentration No. Radionuclide              Class      (&#xb5;Ci)      (&#xb5;Ci)      (&#xb5;Ci/ml)  (&#xb5;Ci/ml)      (&#xb5;Ci/ml)    (&#xb5;Ci/ml) 51  Antimony- 120  D, see 115Sb            1E+3        2E+3          9E-7      3E-9          lE-5        IE-4 (5.76 d)
W, see  115 Sb          9E+2        IE+3          SE-7      2E-9            -
51  Antimony-122  D, see 115Sb            8E+2        2E+3          lE-6      3E-9            -
LU wall (8E+2)                                              lE-5        IE-4 W, see 115Sb        I    7E+2        1E+3          4E-7      2E-9 115                                                    IE-6          3E-3        3E-2 51  Antimony-      D, see    Sb      I    3E+5        8E+5          4E-4 124m12l W, see  115 Sb      I    2E+5        6E+5    I    2E-4  I    8E-7      I      - I 51  Antimony- 124  D, see 115 Sb            6E+2        9E+2          4E-7      IE-9        7E-6        7E-5 W,see    115 Sb          5E+2        2E+2          IE-7      3E-10            -          -
115                                                    3E-9        3E-5        3E-4 51  Antimony- 125  D, see    Sb            2E+3        2E+3          IE-6 W,see msb                  -        5E+2          2E-7      7E-10            -          -
51  Antimony-      D, see 115 Sb            5E+4        2E+5          8E-5      3E-7            -          -
126m12>                              St. wall (7E+4)          -            -        -          9E-4        9E-3 W, see  115 Sb            -        2E+5          8E-5      3E-7            -          -
51  Antimony- 126  D, see  115 Sb          6E+2        IE+3        5E-7      2E-9        7E-6        7E-5 W, see "'Sb              5E+2        5E+2          2E-7      7E-10            -          -
51  Antimony-I 27  D,see  115 Sb          8E+2        2E+3          9E-7      3E-9            -          -
LLI wall (8E+2)          -            -        -          IE-5        IE-4 W,see    115 Sb          7E+2        9E+2          4E-7      lE-9            -          -
51  Antimony-I 28m D.see msb                8E+4        4E+5          2E-4      SE-7            -          -
(I0.4min)                            St. wall (IE+S)          -                                  IE-3        IE-2 W, seemsb                            4E+5          2E-4      6E-7            -
115                                                                2E-5        2E-4 51  Antimony-I 28  D, see    Sb            1E+3        4E+3          2E-6      6E-9 (9.0lh)
W, see 115Sb                        3E+3          IE-6      5E-9            -
115                                                                              4E-4 51  Antimony- 129  D, see    Sb            3E+3        9E+3          4E-6      IE-8        4E-5 115 W, see      Sb                      9E+3          4E-6      I E-8 B-37                                                    NUREG- 1736
 
APPENDIX B Table 1                      Table2          Table3 Occupational Values                Effluent      Releases to Concentrations      Sewers Col. I      Col.2      Col.3      Col. I      Col.2 Oral                                                      Monthly Ingestion          Inhalation                              Average Atomic                                                      ALI          ALI      DAC            Air      Water  Concentration No. CRadionuclide              a          s          s  (&#xb5;Ci)        (&#xb5;Ci)    (&#xb5;Ci/ml)    (&#xb5;Ci/ml)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml) 51  Antimony- I '3rJ2l    D,see 115Sb                    2E+4        6E+4        3E-5      9E-8        3E-4      3E-3 115 W, see      Sb                            8E+4        3E-5      lE-7 I
121          115 51  Antimony-131          D, see      Sb                JE+4 Thyroid 2E+4 Thyroid  I    IE-5      -
(2E+4)      (4E+4)          -      6E-8        2E-4      2E-3 W, see Sb 115
                                                                -          2E+4          lE-5        -
Thyroid (4E+4)          -      6E-8 52  rellurium-116        D, all compounds except        8E+3        2E+4        9E-6      3E-8        IE-4      IE-3 those given for W W, oxides, hydroxides, and                  3E+4          IE-5      4E-8 nitrates 52  felluri um-121 m      D, see ""Te                    SE+2        2E+2          BE-8 Bone Surf    Bone Surf (7E+2)      (4E+2)          -      SE-JO        IE-5      lE-4 W. see 11 "Te                    -          4E+2        2E-7      6E-IO 11 52  fellurium-121        D, see "Te                    3E+3        4E+3        2E-6      6E-9        4E-5      4E-4 I W, see  11 6Te                  -          3E+3        lE-6      4E-9 52  fellurium- I23m      D, see ""Te                    6E+2        2E+2        9E-8        -
Bone Surf    Bone Surf (IE+3)      (5E+2)          -      BE-10        IE-5      IE-4 W, see 11 "Te                              SE+2        2E-7      BE-10 52  fellurium-123        D, see 11 "Te                  5E+2        2E+2        8E-8        .
Bone Surf    Bone Surf (1E+3)      (SE+2)          -      7E-10        2E-5      2E-4 11 W, see "Te                      -          4E+2        2E-7        -
Bone Surf
                                                                -        (IE+3)          -        2E-9 NUREG- 1736                                                      B-38
 
APPENDIX B Table 1                            Table 2          Table3 Occupational Values                      Effluent        Releases to Concentrations        Sewers Col.1        Col.2        Col.3        Col.1        Col.2 Oral                                                            Monthly Ingestion          Inhalation                                    Average Atomic                                        ALI          ALI      DAC                Air        Water  Concentration No. Radionuclide                Class    (&#xb5;Ci)        (&#xb5;Ci)      (&#xb5;Ci/ml)      (&#xb5;Ci/ml)      (&#xb5;Ci/ml)  (&#xb5;Ci/ml) 52  Tellurium-125rn  D. see 11 "Te        1E+3 Bone Surf 4E+2 Bone Surf      I    2E-7 I
(1E+3)      (IE+3)  I        -      I    IE-9        2E-5      2E-4 W, see ""Te                        7E+2          3E-7          IE-9 52  Telluriwn-1 27m          11 D, see 6Te            6E+2        3E+2          IE-7            .          9E-6      9E-5 Bone Surf (4E+2)            -          6E-10 11 W, see 6Te              -          3E+2          lE-7        4E-IO 11 52  Tellurium-127    D, see 6Te            7E+3        2E+4          9E-6          3E-8          IE-4      IE-3 W. see ""Te                        2E+4    I    7E-6    I    2E-8 11 52  Tellurium-l 29rn D, see 6Te            5E+2        6E+2          3E-7        9E-10          7E-6      7E-5 11 W, see 6Te                        2E+2          lE-7        3E-10 52  Tellurium- 129"' D, see 11 "Te        3E+4        6E+4    I    3E-5      I  9E-8          4E-4      4E-3 W, see 11 "Te                      7E+4          3E-5          IE-7 52  Tellurium- 1 Im        11 D, see "Te            3E+2        4E+2          2E-7            -
Thyroid      Thyroid (6E+2)      (1E+3)            -          2E-9          SE-6      8E-5 11 W, see "Te                        4E+2          2E-7            -
Thyroid (9E+2)            -            IE-9 52  Tellurium-13 I"'        11 D, see "Te            3E+3        5E+3        2E-6              .
Thyroid      Thyroid (6E+3)      (IE+4)            -          2E-8          SE-5      SE-4 I    -
11 W. see "Te                        5E+3            2E-6 Thyroid    I (IE+4)      I    -      I  2E-8 I                I 11 52    Tellurium- 132  D. see "Te            2E+2        2E+2          9E-8            -
Thyroid      Thyroid (7E+2)      (8E+2)            -          IE-9          9E-6      9E-5 W, see 11 "Te                    2E+2          9E-8            .
Thyroid (6E+2)            -          9E-IO B-3                                                      NUREG- 1736
 
APPENDIX B Table 1                      Table2          Table3 Occupational Values                Effluent      Releases to Concentrations      Sewers Col. I        Col.2      Col.3    Col.1        Col.2 Oral                                                    Monthly Ingestion            InhaJation                            Average Atomic                                                    ALI          ALI      DAC          Air      Water  Concentration No.      Radionuclide                          Class    (&#xb5;Ci)        (&#xb5;Ci)    (&#xb5;Ci/ml) (&#xb5;Ci/ml)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml) 11 52    Tellurium-              D, see "Te                3E+3        5E+3        2E-6      -            -        -
133m'2>                                      Thyroid      Thyroid (6E<+-3)      (1E+4)          -    2E-8        9E-5      9E-4 W,see 6Te 11
                                                              -          5E+3        2E-6      -            -          -
Thyroid
                                                              -        (1E+4)          -    2E-8          -          -
52    Tellurium-l33m                  11 D,see 6Te                1E+4        2E+4        9E-6      -            -        -
Thyroid      Thyroid (3E+4)        (6E+4)          -    8E-8        4E-4      4E-3 11 W, see 6Te                  -          2E+4        9E-6      -            -        -
Thyroid
                                                              -        (6E+4)          -    8E-8            -        -
52    Tellurium- 134"'      :o. see 111
                                              'Te          2E+4        2E+4          JE-5      -            -        -
Thyroid      Thyroid (2E+4)      (5E+4)          -    7E-8        3E-4      3E-3 11 W. see "Te                  -          2E+4        IE-5      -          -          -
Thyroid
                                                            -        (5E+4)          -    7E-8          -          -
2 53    Iodine- l 20m'      >  D, all compounds        IE+4          2E+4        9E-6    3E-8          -          -
Thyroid (IE+4)            -            -      -          2E-4      2E-3 1
53    Iodine- I 20'  >      D, all compounds        4E+3          9E+3        4E-6      -            -          -
Thyroid      Thyroid (8E+3}      (IE+4)            -    2E-8        IE-4      IE-3 53    Iodine-121              D, all compounds        IE+4        2E+4        8E-6      -            -        -
Thyroid      Thyroid (3E+4)        (5E+4)          -    7E-8        4E-4      4E-3 53    Iodine- 123            D. all compounds        3E+3          6E+3        3E-6      -
Thyroid      Thyroid
()E+4)      (2E+4)            -    2E-8        I E-4      IE-3 I 63 d i n e - I 2 4            D, all compounds        5E+I          8E+I        3E-8      -
Thyroid      Thyroid (2E+2)      (3E+2)            -    4E-IO        2E-6      2E-5 NUREG-1736                                                      B-40
 
APPENDIX B Table 1                        Table 2        Table 3 Occupational Values                Effluent      Releases to Concentrations      Sewers Col. I      CoJ.2        Col.3    Col.1      Col. 2 Oral                                                    Monthly Ingestion            Inhalation Average Atomic                                      ALI          ALI      DAC          Air        Water  Concentration No.      Radionuclide            Class    (&#xb5;Ci)        (&#xb5;Ci)      (&#xb5;Ci/ml)  (&#xb5;Ci/ml)    &i/ml)    (&#xb5;Ci/ml) 53    Iodine- 125    D, all compounds    4E+I        6E+l          3E-8      -
Thyroid      Thyroid (1E+2)      (2E+2)            -  3E-10        2E-6      2E-5 53    Iodine- 126    D, all compounds    2E+l        4E+l          IE-8      -
Thyroid      Thyroid (7E+l)      (1E+2)            -    2E-10          lE-6      IE-S 53    Iodine- I zgm  D, all compounds    4E+4          IE+S        SE-5    2E-7 St. wall (6E+4)          -                                SE-4      SE-3 53    Iodine- 129    D. all compounds    SE+0        9E+0          4E-9 Thyroid      Thyroid (2E+l)      (3E+I)            -    4E-ll        2E-7      2E-6 53  lodine-130      D, all compounds    4E+2        7E+2          3E-7 Thyroid      Thyroid (IE+3)      (2E+3}            -    3E-9        2E-5      2E-4 53  Iodine- 13 I    D. all compounds    3E+l        SE+l          2E-8 Thyroid      Thyroid (9E+l)      (2E+2)            -    2E-10        IE-6      IE-5 53  Iodine-I 32m< 2> D, all compounds    4E+3        8E+3          4E-6 Thyroid      Thyroid (IE+4)      (2E+4)            -    3E-8        IE-4      IE-3 53  Iodine-132      D. all compounds  4E+3          8E+3          3E-6      -
Thyroid      Thyroid (9E+3)      OE+4)              -    2E-8        I E-4    IE-3 53  Iodine- 133      D, all compounds    IE+2        3E+2          I E-7 Thyroid      Thyroid (5E+2)      {9E+2}            -    IE-9        7E-6      7E-5 53  Iodine-134121    D, all compounds  2E+4        5E+4          2E-5    6E-8          -
Thyroid (3E+4)          -                                4E-4      4E-3 53    Iodine-I 35      D, all compounds  8E+2        2E+3          7E-7 Thyroid      Thyroid
{3E+3)      (4E+3)            -    6E-9        3E-S      3E-4 B-41                                              NUREG- 1736
 
APPENDIX B Table 1                      Table 2            Table 3 Occupational Values              Effluent          Releases to Concentrations          Sewers Co). I      Col.2        Col.3    Col. I        Col.2 Oral                                                        Monthly Ingestion          Inhalation                                Average Atomic                                            AL1          AL1      DAC          Air        _Water    Concentration No. Radionuclide                Class      (&#xb5;Ci)        (&#xb5;Ci)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml)      (&#xb5;Ci/ml)    (&#xb5;Ci/ml) 54    Xenon- 12(12>      Submersion"'                                    IE-5    4E-8            -
54    Xenon-121<21      Submersion"'                                    2E-6    lE-8 54    Xenon-122          Submersion"'                                    7E-5    3E-7            -
54  Xenon- 123        Submersion"'                                    6E-6    3E-8 54    Xenon- 125        Submersion"'                                    2E-5    ?E-8 54    Xenon-127          Submersion"'                                    I E-5  6E-8            -
54    Xenon-129m        Submersioncn            -          -          2E-4    9E-7            -          -
54  Xenon-131m        Submersionm              -          -          4E-4    2E-6            -          .
54    Xenon-133m        Submersion' 11          -          -          lE-4    6E-7            -          -
54  Xenon-133          Submersion 111
                                                    -          -          IE-4    SE-7            -            -
54    Xenon-l 35ma    1 Submersion111            -          -          9E-6    4E-8            -          -
54  Xenon-135          Submersionrn            -          -          JE-5    7E-8            -          -
54    Xenon- l 38Cl1    Submersionu>            -          -          4E-6    2E-8            -          -
55    Cesium- I 25m      D, all compounds      5E+4        IE+5        6E-5    2E-7            -          -
St. wall (9E+4>          -            -        -          IE-3        IE-2 55  Cesium-127          D, all compounds      6E+4        9E+4        4E-5    lE-7          9E-4        9E-3 55    Cesium-129        D, all compounds      2E+4        3E+4          IE-5    SE-8          3E-4        3E-3 55    Cesiurn-130' 2
                        >  D, all compounds      6E+4        2E+5        8E-S    3E-7            -          -
St. wall (1E+5)          -            -        -          IE-3        IE-2 55    Cesiurn-131        D, all compounds      2E+4        3E+4          IE-5    4E-8          3E-4        3E-3 55  Cesium-132          D, all compounds      3E+3        4E+3        2E-6    6E-9          4E-5        4E-4 55    Cesium- I 34m      D. all compounds      IE+S        IE+5        6E-5    2E-7            -            -
St. wall (IE+S)          -            -        -          2E-3        2E-2 55    Cesium- 134        D, all compounds      7E+l        IE+2        4E-8    2E-IO          9E-7        9E-6 55    Cesium- I35m'2 '  D. all compounds      IE+S        2E+5        8E-5    3E-7          IE-3        IE-2 55    Cesium- 13 5      D, all compounds      7E+2        1E+3        5E-7    2E-9          IE-5        lE-4 55  :::esium-136        D, all compounds      4E+2        7E+2        3E-7    9E-IO          6E-6        6E-5 NUREG-1736                                            B-42
 
APPENDIX B Table 1                      Table2          Table3 Occupational Values              Effluent      Releases to Concentrations      Sewers Col. I      Col.2    I    Col.3  Col.I        Col.2 Oral                                                    Monthly Ingestion          Inhalation                            Average Atomic                                                ALI          ALI      DAC          Air        Water  Concentration No.      Radionuclide                  Class      (&#xb5;Ci)        (&#xb5;Ci)      (&#xb5;Ci/ml) (&#xb5;Ci/ml)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml) ss    Cesium-137        D, all compounds          1E+2        2B+2          6E-8  2E-10          lE-6      IE-S S5    Cesium- I 3s< 21 D, all compounds          2E+4          6B+4          2E-5    SE-8          -          -
St. wall (3E+4)          -            -        .        4E-4      4E-3 2
S6    Barium-126< >    D, all compounds          6E+3        2E+4          6E-6    2E-8        8E-5      8E-4 56    Barium-128        D, all compounds          5E+2        2E+3          7E-7    2E-9        7E-6        7E-5 56    Barium-13tm<2>  D, all compounds          4E+5          1E+6          6E-4  2E-6            -          .
St. wall (5E+S)          -            -      .-        7E-3      7E-2 56    Barium-131        D, all compounds          3E+3        8E+3          3E-6    IE-8        4E-5      4E-4 56    Barium-133m      D, all compounds          2E+3        9E+3          4E-6    IE-8          -          -
LLiwall (3E+3)          -            .      .        4E-5      4E-4 56    Barium-133      D, all compounds          2E+3        7E+2          3E-7  9E-I0        2E-5      2E-4 56    Barium-135m      D, all compounds          3E+3        lE+4          SE-6  2E-8        4E-5      4E-4 56    Barium-139' 21  D, all compounds            IE+4        3E+4          IE-5  4E-8        2E-4      2E-3 56      Barium-140      D, all compounds          5E+2        IE+3          6E-7    2E-9          .          .
LLI wall (6E+2)          .            .        .        8E-6      8E-5 56      Barium-141 121  D, all compounds          2E+4        7E+4          JE-5    JE-7        3E-4      3E-3 121 56    Barium-142        D, all compounds          5E+4        IE+5          6E-5    2E-7        7E-4      7E-3 57    Lanthanum-        D, all compounds except    SE+4        IE+5          5E-5    2E-7        6E-4      6E-3
      . 131' 21          those given for W W, oxides and hydroxides      -        2E+5          7E-5    2E-7          .          -
131                                                                          4E-4 57    Lanthanum-132    D. see      La            3E+3        IE+4          4E-6    IE-8        4E-5 W.see    131 La              -          IE+4          5E-6  2E-8            -          .
57    Lanthanum-1 35    D, see 1J 1La            4E+4          IE+5          4E-S    IE-7        SE-4      SE-3 W. see  131 La                        9E+4          4E-5    IE-7          -
B-43                                              NUREG- 1736
 
APPENDIX B Table 1                      Table 2          Table 3 Occupational Values                Effluent      Releases to Concentrations      Sewers Col. I      Col.2        Col. 3  Col. I        Col. 2 Oral                                                    Monthly Ingestion          Inhalation                              Average Atomic                                                ALI          ALl      DAC          Air      Water  Concentration No.      Radionuclide                Class        (&#xb5;Ci)        (&#xb5;Ci)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml) 57  Lanthanum- 137  D, see 131La                1E+4        6E+l        3E-8        -        2E-4      2E-3 Liver
                                                          -      (7E+1)            -    lE-10          -          -
W,see      131 La                -        3E+2          lE-7      -            -          -
Liver
                                                          -      (3E+2)            -    4E-10            -          -
131 57  Lanthanum- 138  D, see      La              9E+2          4E+O          lE-9  SE-12          IE-5      IE-4 W, see    131 La              -          IE+l          6E-9  2E-11            -          -
131 57  Lanthanum- 140  D, see      La              6E+2          1E+3          6E-7    2E-9        9E-6      9E-5 W, see    131 La                -        IE+3        SE-7    2E-9            -          -
57  Lanthanum-14 I  D,see mLa                    4E+3        9E+3          4E-6      IE-8        SE-5      SE-4 W,see      131 La                -        IE+4          SE-6    2E-8          -          -
131 57    Lanthanum-      D, see      La              8E+3          2E+4        9E-6      3E-8        IE-4      lE-3 14212)
W,see    131 La                -        .3E+4          IE-5    SE-8          -          -
57    Lanthanum-      D,see mLa                    4E+4          IE+5        4E-5      lE-7          -          -
143121                                      St. wall (4E+4)          -            -        -        SE-4      SE-3 W,see mLa                      -        9E+4          4E-5      IE-7          -          -
58    Cerium-134      W, all compounds except      5E+2        7E+2          3E-7    IE-9          -          -
those given for Y          LLI wall (6E+2)          -            -        -        8E-6      SE-5 Y, oxides, hydroxides, and      -        7E+2          3E-7    9E-IO          -          -
fluorides 134 58    ::erium-135    W, see      Ce              2E+3        4E+3          2E-6    SE-9        2E-5      2E-4 Y. see  134 Ce                -        4E+3          IE-6    SE-9          -          -
134 58    ::erium-137m    W, see Ce                    2E+3        4E+3          2E-6    6E-9          -
LLI wall (2E+3)          -                                3E-5      3E-4 Y,see 134Ce                              4E+3          2E-6    5E-9          -
13 58    ::erium-137    W, see 'Ce                  SE+4        lE+S          6E-5    2E-7        7E-4      7E-3 13 Y, see 'Ce                                IE+5          SE-5    2E-7          -
NUREG-1736                                                  B-44
 
APPENDIX B Tablet                      Table2          Table3 Occupational Values              Effluent        Releases to Concentrations      Sewers Col. I      Col. 2        Col.3  Col. 1      Col.2 Oral                                                  Monthly Ingestion          Inhalation Average Atomic                                              ALI        .Airl    DAC                    Water  Concentration No. Radionuclide              Class        (&#xb5;Ci)      (&#xb5;Ci)      (&#xb5;Ci/ml) (&#xb5;Ci/ml)    (&#xb5;Cl/ml)  (uCi/ml) 58  Cerium-139    W, see 134Ce                5E+3        8E+2          3E-7    lE-9        7E-5      7E-4 1
Y.see 'te              I              7E+2          3E-7  9E-10          -
58  Cerium- 141    W, see  134 Ce              2E+3        7E+2          3E-7    IE-9          -
LLI wall (2E+3)          -            -        -        3E-5      3E-4 Y,see  134 Ce                  -        6E+2          2E-7    SE-10          -
58  Cerium-143    W,see  134 Ce              1E+3        2E+3          8E-7    3E-9          -
LLI wall (1E+3)          -            -      -        2E-5      2E-4 Y,see  134 Ce                  -        2E+3          7E-7    2E-9          -
58  Cerium-144    W, see i)*ce                2E+2        3E+I          IE-8  4E-1 l          -
LLI wall (3E+2)          -            -      -        3E-6      3E-5 13 Y, see 'Ce                    -        IE+l          6E-9  2E-1 l          -
59  Praseodymium-  W, all compounds except    5E+4        2E+5          lE-4    3E-7          -
136(2)        those given for Y          St. wan (7E+4)          -            -      -          lE-3      lE-2 Y, oxides, hydroxides,        -        2E+5          9E-5    3E-7          -
carbides, and fluorides 59  Praseodymium-  W, see  136 Pr              4E+4        2E+5          6E-5    2E-7        5E-4      5E-3 137121                1 Y, see )6Pr                    -        IE+S          6E-5    2E-7          -
59  Praseodymium-          13 W, see 6l>r                  IE+4      5E+4          2E-5    8E-8        lE-4      IE-3 138m Y, see 13 6Pr                  -        4E+4          2E-5    6E-8          -
136                4E+4        IE+5          5E-5    2E-7        6E-4      6E-3 59  Praseodymium-  W,see      Pr 139 Y, see 136 Pr                  -        IE+5          5E-5    2E-7          -
59  Praseodymium-  W, see 136Pr                8E+4        2E+5          7E-5    2E-7        lE-3      IE-2 142m121 Y, see 136Pr                  -        1E+5          6E-5    2E-7          -
59  Praseodymium-  W, see  13 6Pr              IE+3        2E+3          9E-7    3E-9        IE-5      IE-4 142 Y, see 136 Pr                  -        2E+3          8E-7    3E-9          -
B-45                                            NUREG- 1736
 
APPENDIX B Table 1                      Table2            Table3 Occupational Values                Effluent        Releases to Concentrations        Sewers Col. I      Col.2          Col.3  Col. I        Col. 2 Oral                                                      Monthly Ingestion            Inhalation                              Average Atomic                                            ALI          ALI        DAC          Air        Water  Concentration No. Radionuclide              Class      (&#xb5;Ci)        (&#xb5;Ci)      (&#xb5;Ci/ml) (&#xb5;Ci/ml)      (&#xb5;Ci/ml)  (&#xb5;Ci/rill) 59    Praseodymium-  W, see 136I>r            9E+2        8E+2          3E-7    lE-9            -
143                                  LLiwall (1E+3)          -                                2E-5      2E-4 Y, see 136I>r                          7E+2          3E-7  9E-10            -
59    Praseodymium-  W, see  136 Pr          3E+4          lE+S          SE-5    2E-7            -
144(2)                                  St. wall (4E+4)            -                                6E-4      6E-3 136 Y, see    Pr                          1E+5          SE-5    2E-7            -
59    Praseodymium-  W,see*~                  3E+3        9E+3          4E-6    IE-8        4E-5      4E-4 145 Y, see  13 6Pr                          8E+3          3E-6    lE-8          -
59    Praseodymium-  W, see  136 Pr          5E+4        2E+5          SE-5    3E-7            -
147'2)                                  St. wall (8E+4)          -                                lE-3        IE-2 Y, see 136 Pr                          2E+5          SE-5    JE-7            -
60    Neodymium-    W. all compounds except    IE+4        6E+4          2E-5    8E-8          2E-4      2E-3 136(2)        those given for Y Y, oxides, hydroxides,                SE+4          2E-5    SE-8            -
carbides, and fluorides 60    Neodymium- 138 W, see 136Nd              2E+3        6E+3          3E-6    9E-9          JE-5      3E-4 136 Y. see    Nd                          5E+3          2E-6    7E-9 60    Neodymium-    W, see 136Nd              SE+3        2E+4          7E-6    2E-8          7E-5      7E-4 139m Y, see 136Nd                          1E+4          6E-6    2E-8 16 60    Neodymium-    W, see  l Nd            9E+4        3E+5          I E-4  SE-7          IE-3      IE-2 139m                  136 Y, see    Nd                          3E+5          IE-4    4E-7 60    Neodymium-141  W, see 136Nd              2E+S        7E+S          3E-4    IE-6          2E-3      2E-2 Y, see 136Nd                          6E+5          3E-4    9E-7            -
16 60    Neodymium-147  W, see  ) Nd            IE+3        9E+2          4E-7    IE-9            -
LLI wall (IE+3)          -                                2E-5      2E-4 6
Y, see 1) Nd                          8E+2          4E-7    lE-9 NUREG- 1736                                          B-46
 
APPENDIX B Table I              I          Table 2      I    Table3 Occupational Values                  Eftluent          Releases to Concentrations          Sewers Col~I      Col. 2        Col.3    Col. I        Col. 2 Oral                                                          Monthly Ingestion            Inhalation                                  Average Atomic                                            ALI          ALI      DAC            Air        Water    Concentration No.      Radionuclide              Class        (&#xb5;Ci)        (&#xb5;Ci)      (&#xb5;Ci/ml)  (&#xb5;Ci/ml)      (&#xb5;Ci/ml)    (&#xb5;Ci/ml) 60  Neodymium-      W, see 13 6Nd              IE+4        3E+4          IE-5    4E-8          IE-4        IE-3 14gl2>                13 Y, see 6Nd                              2E+4          IE-5    3E-8            -
136 60  Neodymium-      W, see      Nd            7E+4        2E+5          SE-5    3E-7          9E-4        9E-3 151 121 Y, see  13 6Nd                          2E+5          SE-5    3E-7            -
61  Promethium-    W, all compounds except    5E+4        2E+5          BE-5    3E-7            -
141(2)        those given for Y        St. wall (6E+4)          -                                  SE-4          SE-3 Y, ox.ides, hydroxides,                2E+5          7E-5      2E-7            -
carbides, and fluorides 61  Promethium- 143 W,see'"Pm                  5E+3        6E+2          2E-7      SE-10        7E-5        7E-4 Y, see '"Pm                            7E+2          3E-7      lE-9 141 61  Promethium- 144 W, see      Pm            lE+3        lE+2          5E-8    2E-10          2E-5        2E-4 141 Y, see Pm                              IE+2          SE-8    2E-10            -
61  Promethium- 145 W, see 141 Pm              IE+4        2E+2          7E-8                    IE-4        IE-3 Bone Surf (2E+2)            -      3E-10            -
Y, see "'Pm                            2E+2          8E-8    3E-IO            -
61    Promethium- 146 W, see ""Pm                2E+3        5E+l          2E-8    7E-11          2E-5        2E-4 Y, see"'Pm                            4E+l          2E-8    6E-1 l          -
61    Promethium- 147 W, see "'Pm              4E+3          IE+2          SE-8        -
LLiwall      Bone Surf (5E+3)      (2E+2)            -      3E-10          7E-5        7E-4 Y, see '"Pm                            IE+2          6E-8    2E-1O            -
61    Promethium-    W, see "'Pm              7E+2        3E+2          IE-7    4E-10          IE-5        IE-4 148m Y, see"'Pm                            3E+2          IE-7    SE-1O            -
61    Promethium-148  W, see"'Pm                4E+2        SE+2          2E-7    SE-1O            -
LLI wall (5E+2)          -                                  7E-6        7E-5 Y, see "'Pm                            5E+2          2E-7    7E-10            -
B-47                                                    NUREG-I736
 
APPENDIX B Table 1                        Table 2          Table 3 Occupational Values                Effluent        Releases to Concentrations        Sewers Col. I      Col.2        Col.3  Col. 1        Col. 2 Oral                                                      Monthly Ingestion            Inhalation                              Average Atomic                                            ALl          ALl      DAC          Air        Water  Concentration No.      Radionuclide                  Class    (&#xb5;Ci)        (&#xb5;Ci)      (&#xb5;Cl/ml) (&#xb5;Ci/ml)      (&#xb5;Cl/ml)  (&#xb5;Cl/ml) 61    Promethium- 149  W, see  141 Pm        IE+3        28+3          SE-7  . 3E-9.            -          -
LLlwall (1E+3)          -            -        -          2E-5.      2E-4 Y,see  I4I Pm            -        2E+3          8E-7    2E-9            -          -
141 61  Promethium- 150  W,see        Pm        SE+3        2E+4          BE-6    3E-8          7E-S      7E-4 1 1 Y, see
* Pm              -        2E+4          7E-6    2E-8            -          -
14I 61  Promethium- 15 1  W,see        Pm        2E+3        4E+3          lE-6    SE-9          2E-5        2E-4 Y, see  141 Pm            -        3E+3          lE-6    4E-9            -          -
62  Samarium-        W, all compounds      3E+4          IE+S          4E-5    IE-7        4E-4      4E-3 14lm121 62  Samarium-141<2>  W, all compounds      5E+4        2E+5          8E-5    2E-7            -          -
St. wall (6E+4)            -            -      -          SE-4      SE-3 62    Samarium- 142'21  W, all compounds      8E+3        3E+4          lE-5    4E-8          IE-4      lE-3 62    Samarium- 145    W, all compounds      6E+3        5E+2          2E-7  7E-10          SE-5      SE-4 62    Samarium- 146    W, all compounds      IE+!        4E-2          IE-11      -              -          -
Bone Surf    Bone Surf (3E+l)        (6E-2)          -    9E-14          3E-7      3E-6 62    Samarium-I 47    W, all compounds      2E+I        4E-2        2E-11      -            -          -
Bone Surf    Bone Surf (3E+l)        (7E-2)          -    IE-13          4E-7      4E-6 62    Samarium-15 I    W, all compounds      IE+4        IE+2          4E-8      -            -          -
LU wall      Bone Surf (IE+4)      (2E+2)            -    2E-10          2E-4      2E-3 62    Samarium- 153    W, all compounds      2E+3        3E+3          lE-6    4E-9            -          -
LLI wall (2E+3)            -                                3E-5      3E-4 62    Samarium- I 55*!> W, all compounds      6E+4        2E+S          9E-S    3E-7            -
St. wall (8E+4)            -                                lE-3      IE-2 62    Samarium- 156    W, all compounds      SE+3        9E+3          4E-6    IE-8        7E-5        7E-4 63    Europium- 145    W, all compounds      2E+3        2E+3          BE-7    3E-9          2E-5      2E-4 NUREG-1736                                            B-48
 
APPENDIX B Table 1                      Table 2          Table3 Occupational Values                Effluent        Releases to Concentrations        Sewers I
Col. I        Col.2        Col.3  Col.I        Col.2 Oral                                                      Monthly Ingestion            Inhalation                              Average Atomic                                                ALI          ALI      DAC          Air        Water  Concentration No. Radionuclide              Class            (&#xb5;Ci)        (&#xb5;Ci)      (&#xb5;Ci/ml)  (&#xb5;Ci/ml)    (&#xb5;Ci/ml)  <uCl/mll 63  EuroJ>ium-146    W, all compounds            1E+3          IE+3        ~7      2&9*          l_E-5    lE-4 63  Europium-147    W, all compounds            3E+3          2E+3          7E-7    2E-9        4E-5      4E-4 63  Europium-148    W, all compounds            1E+3        . 4E+2          lE-7  SE-10.        .IE-5      IE-4 63  Europium-149    W, all compounds              IE+4        3E+3          IE-6    4E-9          2E-4      2E-3 63  Europium-I 50    W, an compounds              3E+3          8E+3          4E-6
* lE-8        4E-5      4E-4 (12.62h) 63  Europium- I 50  W, all compounds            8E+2          2E+I          8E-9    3E-1 I        IE-5      IE-4 (34.2y) 63  Europium- I S2m  W, all compounds            3E+3          6E+3          3E-6    9E-9        4E-S      4E-4 63  Europium-152    W. all compounds            8E+2          2E+l          IE-8  3E-ll          IE-5      IE-4 63  Europium-I S4    W, all compounds            5E+2          2E+l          8E-9    3E-II        7E-6      7E-S 63  Europium-155    W, all compounds            4E+3          9E+I          4E-8      -          5E-5      5E-4 Bone Surf
                                                        -        (IE+2)            -    2E-10            -
63  Europium-156    W, all compounds            6E+2          5E+2          2E-7    6E-10        8E-6      8E-S 63  Europium-157    W, all compounds            2E+3          5E+3          2E-6    7E-9        3E-5      3E-4 63  Europium-15&<21  W, all compounds            2E+4          6E+4          2E-5    8E-8        3E-4      3E-3 64  Gadolinium-      D. all compounds except      5E+4          2E+5          6E-5    2E-7            -
1451~>          those given for W          St. wall (5E+4)            -            -      -          6E-4      6E-3 W, oxides, hydroxides, and      -          2E+5          7E-5    2E-7            -
fluorides 64  Gadolinium-146  D,see 145Gd                  1E+3          IE+2          SE-8    2E-10        2E-5      2E-4 W,see " 5Gd                    -          3E+2          IE-7    4E-10            -
64  Gadolinium- I 47          5 D, see " Gd                  2E+3          4E+3          2E-6    6E-9        3E-5      3E-4 W, see  15
                                ~  Gd                  -          4E+3          IE-6    SE-9            -
B-49                                                NUREG- 1736
 
APPENDIX B Table 1                      Table 2            Table 3 Oc upational Values                Effluent        Releases to Concentrations        Sewers Col.I        CoL2        Col.3    Col. I        Col.2 Oral                                                      Monthly Ingestion          Inhalation                              Average Atomic                                          ALI          ALI      DAC          Air        Water  Concentration No. Radionuclide                Class  (&#xb5;Ci)        (&#xb5;Ci)    (&#xb5;Ci/ml) (&#xb5;Ci/ml)      (&#xb5;Ci/ml)    (&#xb5;Ci/ml) 145 64  Gadolinium- 148  D, see    Gd          lE+l        8E+3        3E-12      -            -          -
Bone Surf  Bone Surf (2E+l)      (2E+2)            -  2E-14          3E-7        3E-6 145 W, see      Gd          -          3E-2        IE-11      -            -          -
Bone Surf (6E-2)          -  SE-14            -
145 64  Gadolinium- 149  D, see    Gd        3E+3          2E+3        9E-7    3E-9        4E-5        4E-4 W, seeu 5Gd                        2E+3          IE-6    3E-9          -          -
145 64    Gadolinium- 15 1 D, see    Gd        6E+3          4E+2        2E-7      -          9E-5        9E-4 Bone Surf (6E+2)            -    9E-IO          -          -
145 W, see      Gd                    IE+3        5E-7    2E-9            -
64    Gadolinium- 152  D, see 145 Gd        2E+l          lE-2      48-12      -
Bone Surf  Bone Surf (3E+l)      * (2E-2)          -  3E-14          4E-7        4E-6 W. see  145 Gd                      4E-2        2E-11      -            -          -
Bone Surf (8E-2)          -    IE-13          -          -
64    Gadolinium-153  D. see 145Gd          5E+3          IE+2        6E-8                  6E-5        6E-4 Bone Surf (2E+2)          -    3E-IO            -
W, see  145 Gd                      6E+2        2E-7  8E-IO            -
145 64    Gadolinium- 159  D, see    Gd        3E+3        8E+3          3E-6    IE-8        4E-5        4E-4 W, see " 5Gd                      6E+3          2E-6    8E-9            -
65    Terbium-14&deg;12>  W, all compounds      9E+3        3E+4          IE-5    SE-8          IE-4        IE-3 65    Terbium- 149    W, all compounds      5E+3        7E+2          3E-7    IE-9          7E-5        7E-4 65    rerbium- 150    W, all compounds      5E+3        2E+4          9E-6    3E-8          7E-5        7E-4 65    Terbium- 15 1    W. all compounds      4E+3        9E+3          4E-6    IE-8          5E-5        5E-4 65    rerbium- 153    W, all compounds      5E+3        7E+3          3E-6    lE-8          7E-5        7E-4 65    rerbium- 154    W, all compounds      2E+3        4E+3          2E-6    6E-9          2E-5        2E-4 65    rerbium- 155    W, all compounds      6E+3        8E+3          3E-6    IE-8    I    8E-5 I    8E-4 NUREG-1736                                          B-S
 
APPENDIX B Table 1                        Table 2          Table3 Occupational Values                  Effluent      Releases to Concentrations      Sewers Col. I        Col.2        Col. 3    Col. I        Col.2 Oral                                                        Monthly Ingestion            Inhalation                                Average Atomic                                          ALl          ALI      DAC            Air      Water  Concentration No.      Radionuclide            Class      (&#xb5;Ci)        (&#xb5;Ci)      (&#xb5;Ci/ml)    (&#xb5;Ci/ml)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml) 65    Terbium- 156m  W, all compounds      2E+4          3E+4          IE-5      4E-8        2E-4      2E-3 (5.0h) 65    Terbium- 156m    W, all compounds      7E+3          8E+3          3E-6        lE-8        lE-4        lE-3 (24.4h) 65    rerbium- 156    W, all compounds      IE+3          IE+3        6E-7      2E-9          lE-5      lE-4 65    rerbium- 157    W, all compounds      5E+4          3E+2          IE-7 LLI wall      Bone Surf (5E+4)        (6E+2)            -      SE-10        7E-4      7E-3 65    rerbium- 158    W, all compounds      IE+3          2E+l          8E-9      3E-11        2E-5      2E-4 65    rerbium- 160    W. all compounds      8E+2          2E+2          9E-8      3E-10        lE-5      lE-4 65    rerbium- 161    W, all compounds      2E+3          2E+3          7E-7      2E-9 LLI wall (2E+3)            -            -          -        3E-5      3E-4 66    \Dysprosium- 155 W, all compounds      9E+3          3E+4          IE-5      4E-8        IE-4        IE-3 66  !Dysprosium- 157  W, all compounds      2E+4          6E+4          3E-5      9E-8        3E-4      3E-3 66  !Dysprosium-I 59  W, all compounds      1E+4          2E+3          IE-6      3E-9        2E-4      2E-3 66  !Dysprosium- 165  W, all compounds      IE+4          5E+4          2E-5      6E-8        2E-4      2E-3 66    !Dysprosium- 166  W, all compounds      6E+2          7E+2          3E-7      lE-9 LLI wall (8E+2)            -                                  lE-5      lE-4 67    !Holmium- 15512>  W, all compounds      4E+4          2E+5          6E-5      2E-7        6E-4      6E-3 2
67    Holmium-151 >    W, all compounds      3E+5          1E+6          6E-4      2E-6        4E-3      4E-2 2
67    IHolmium- I59'  > W. all compounds      2E+5          IE+6        4E-4      IE-6        3E-3      3E-2 67    IHolmium-161      W, all compounds      IE+5          4E+5          2E-4      6E-7        IE-3      lE-2 67    \Holmium-        W, all compounds      5E+4          3E+5          I E-4    4E-7        7E-4      7E-3 162m12 >                          I              I 67    IHolmium- J6i<2>  W, all compounds    5E+5          2E+6          lE-3      3E-6 St. wall (8E+5)            -                                  IE-2      IE-I 67    \Holmium-
      'l64m' 2>
W, all compounds I    IE+5 I    3E+5 I    IE-4 I
4E-7        IE-3      lE-2 B-5 I                                                  NUREG-1736
 
APPENDIX B Table 1                    Table 2          Table 3 Occupational Values              Effluent      Releases to Concentrations      Sewers Col. I      Col. 2        Col.3  Col. 1      Col.2 Oral                                                    Monthly Ingestion          Inhalation                            Average Atomic                                        ALI          ALI      DAC          Air      Water  Concentration.
No.      Radionuclide              Class    (&#xb5;Ci)        (&#xb5;Ci)    (&#xb5;Ci/ml) (&#xb5;Ci/ml)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml) 67    Holmium- I 6412> W, all compounds    2E+5        6E+5        3E-4  *9E-7            -          -
St. wall (2E+5)          -            -      -        3&3        3E-2 67    Holmium- 166m    W, all compounds    6E+2        7E+0        3E-9    9E-12        9E-6
* 9E-5 67    Holmium-166      W, all compounds    9E+2        2E+3        7E-7    2E-9          .,        -
ILi wall (9E+2)          -            -      -          IE-5      lE-4 67    Holmium- 167    W, all compounds    2E+4        6E+4        2E-5    8E-8        2E-4        2E-3 68    Erbium-161      W, all compounds    2E+4        6E+4        3E-5    9E-8        2E-4        2E-3 68    Erbium- 165      W. all compounds    6E+4        2E+5        8E-5    3E-7        9E-4        9E-3 68    Erbium- 169      W, all compounds    3E+3        3E+3          lE-6    4E-9 LU wall (4E+3)          -                              SE-5        SE-4 68  Erbium-171        W, all compounds    4E+3        1E+4        4E-6    lE-8        SE-5        SE-4 68  Erbium- 172      W, all compounds    IE+3        IE+3        6E-7    2E-9          -
LU wall (IE+3)          -                              2E-5        2E-4 21 69  Thulium- 162'    W, all compounds    7E+4        3E+5          lE-4  4E-7 St. wall (7E+4)          .                              IE-3        IE-2 69    Thulium- 166      W, all compounds    4E+3        IE+4        6E-6    2E-8        6E-5        6E-4 69  Thulium- 167      W, all compounds    2E+3        2E+3          8E-7    3E-9 LU wall (2E+3)          -                              3E-5      3E-4 69    Thulium-170      W, all compounds  8E+2        2E+2          9E-8  3E-10          -
LLiwall (IE+3)          -            -      -          lE-5      IE-4 69    Thulium-171      W, all compounds    IE+4        3E+2          IE-7      -
LLI wall    Bone Surf (IE+4)      (6E+2)          -    8E-10        2E-4      2E-3 NUREG- 1736                                        B-52
 
APPENDIX B Table 1                    Table 2          Table3 Occupational Values              Effluent      Releases to Concentrations      Sewers Col. I      Col. 2        Col.3  Col. I      Col.2 Oral                                                  Monthly Ingestion          Inhalation                            Average Atomic                                                    ALl        ALI      DAC        Air      Water  Concentration No. Radionuclide                      Class          (&#xb5;Ci)      (&#xb5;Ci)      (&#xb5;Ci/ml) (&#xb5;Ci/ml)    (uCi/mll  (&#xb5;Ci/m1) 69  Thulium- 172        W, all compounds              7E+2        IE+3        5E-7    2E-9          -
LLiwall (8E+2)          -            -      ~          IE-5      IE-4 69  Thulium- 173        W, all compounds              4E+3        IE+4        5E-6    2E-8        6E-5      6E-4 69  Thulium- I 7s<2>    W, all compounds              7E+4        3E+5          lE-4    4E-7 St wall (9E+4)          -            -      -          lE-3      I E-2 121 70  Ytterbium- 162      W, all compounds ex.cept      7E+4        3E+S          IE-4    4E-7        IE-3      IE-2 those given for Y Y, ox.ides, hydrox.ides, and      -        3E+5          IE-4    4E-7          -
fluorides 162                                                                2E-5      2E-4 70  Ytterbium-I 66      W,see        Yb                IE+3      2E+3          8E-7    3E-9 Y, see  162 Yb                  -        2E+3          8E-7    3E-9 70  Ytterbium- 16121    W,see    162 Yb              3E+S        8E+S          3E-4    IE-6        4E-3      4E-2 Y, see  162 Yb                  -        7E+S          3E-4    IE-6 162 70  Ytterbium- 169      W,see        Yb              2E+3        8E+2          4E-7    IE-9        2E-5      2E-4 Y, see 162vb                      -        7E+2          3E-7    IE-9 162 70  Ytterbium- 175      W, see        Yb              3E+3        4E+3          IE-6  -5E-9 LLlwall (3E+3) .        -            -      -        4E-5      4E-4 162 Y, see      Yb                    -        3E+3          IE-6    5E-9 162 70  Ytterbium- 17121    W, see      Yb                2E+4        5E+4          2E-5    7E-8        2E-4      2E-3 Y,see  16 2Yb                  -        5E+4          2E-5    6E-8 121          162 2E-4      2E-3 70  Ytterbium- 178      W,see        Yb                IE+4        4E+4          2E-5    6E-8 Y, see 162Yb                              4E+4          2E-5    SE-8 71  Lutetium- 169      W, all compounds except        3E+3        4E+3          2E-6    6E-9        3E-5      3E-4 those given for Y Y, oxides, hydroxides, and        -        4E+3          2E-6    6E-9 fluorides 71  Lutetium- 170      W, see 169Lu                  IE+3        2E+3          9E-7    3E-9        2E-5      2E-4 169 Y, see      Lu                            2E+3          8E-7    3E-9 B-53                                            NUREG- 1736
 
APPENDIX B Table 1                      Table2          Table3 Occupational Values                Effluent      Releases to Concentrations        Sewers Col. I        Col.2        Col.3    Col. I      Col. 2 Oral                                                      Monthly Ingestion            Inhalation                              Average Atomic                                          ALI          ALI      DAC          Air        Water  Concentration No. Radionuclid~                Class    (&#xb5;Ci)        (&#xb5;Ci)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml) 71  Lutetium-171    W; see 1691...u        2E+3          2E+3          86-7
* 3E-9.        3E-5      3E-4 16 Y, see 91...u**            -        *2E+3          8E-7    3E-9            -          -
169 71  Lutetium- 172  W, see      Lu          IE+3          1E+3        SE-7    2E-9          lE-5      IE-4 Y,see 1~u                  -          IE+3        5E-7    2E-9            -          -
16 71  Lutetium- 173  W, see 91..u            5E+3          3E+2          IE-7      -          7E-5      7E-4 Bone Surf
                                                    -        (5E+2)          -    6E-10            -          -
1 Y, see 691...u            -          3E+2          IE-7    4E-10            -        -
71  Lutetium-17 4m  W,see *~u              2E+3          2E+2          lE-7      -            -          -
LLlwall      Bone Surf (3E+3)        (3E+2)          -    5E-10        4E-5      4E-4 Y, see  1611 Lu            -          2E+2          9E-8    3E-10            -          -
71  Lutetium- 174  W; see 1691...u        5E+3          1E+2        SE-8        -          7E-S      7E-4 Bone Surf
                                                    -        (2E+2)            -    3E-10            -        -
16 Y, see 91.u                -          2E+2        6E-8    2E-10          -          -
16 71  Lutetium-l 76rn W. see 91.u            8E+3          3E+4          lE-5    3E-8        lE-4      lE-3 169 Y, see      Lu            -          2E+4        9E-6      3E-8          -          -
16 71  Lutetium-176    W, see 91.u            7E+2          SE+O        2E-9        -          lE-5      IE-4 Bone Surf
                                                  -        (IE+I)            -    2E-l l          -          -
16 Y, see 91..u              -          8E+O        3E-9    IE-11          -          -
71  Lutetium- l 77m W, see  169 Lu        7E+2          1E+2        5E-8        -          lE-5      IE-4 Bone Surf
                                                  -        (1E+2)          -      2E-10          -          -
Y, see 1611 Lu            -          8E+l        3E-8    IE-10          -          -
1 71  Lutetium- 177  W.see ~u              2E+3          2E+3          9E-7      3E-9          -          -
LLI wall (3E+3)            -            -        -          4E-5      4E-4 1
Y, see wi..u              -          2E+3          9E-7    3E-9            -          -
NUREG- 1736                                          B-54
 
APPENDIX B
* Table 1                    Table 2          Table 3 Occupational Values              Effluent      Releases to Concentrations        Sewers Col. t      Col.2        Col.3  Col.1        Col.2 Oral                                                    Monthly Ingestion          Inhalation                            Average Atomic                                              ALI          ALl      DAC        Air        Water  Concentration No. Radionuclide                    Class      (&#xb5;Ci)        (&#xb5;Ci)    (&#xb5;CVml)  (&#xb5;Ci/ml)    (&#xb5;CVml)    (&#xb5;Ci/ml) 71  Lutetium- I 78mC2> W, see 169Lu              5E+4        2E+5        8E-5    3E-7            .
St. wall (6E+4)          .                              8E-4      SE-3 Y, see 169Lu                          2E+S        7E-5    2E-7 71  Lutetium- l 78C7->          16 W, see 91.u              4E+4        IE+S        SE-5    2E-7            .
St. wall (4E+4)          .                              6E-4      6E-3 16 Y, see 91.u                            1E+5        SE-5    2E-7            .
71  Lutetium- 179      W, see 169Lu              6E+3        2E+4        8E-6    3E-8        9E-5      9E-4 Y, see  169 Lu                        2E+4        6E-6    3E-8            .
72  Hafnium- 170      D, all compounds except  3E+3        6E+3        2E-6    8E-9        4E-5      4E-4 those given for W W, oxides, hydroxides,                5E+3          2E-6    6E-9            .
carbides, and nitrates 72  Hafnium- 172      D, see 170Hf              IE+3      9E+O          4E-9                2E-5      2E-4 Bone Surf (2E+l)          . 3E-1 l          .
17 W, see &deg;Hf                            4E+l          2E-8 Bone Surf (6E+I)          . 8E-1 I          .
72  Hafnium- 173      D, see 17&deg;Hf              5E+3        1E+4          SE-6    2E-8        7E-5      7E-4 W,see 17&deg;Hf                          IE+4          SE-6    2E-8          .
72  Hafnium-175        D, see 17&deg;Hf              3E+3        9E+2          4E-7                4E-5      4E-4 Bone Surf (IE+3)          . IE-9 W, see  170 Hf                      IE+3          SE-7    2E-9          .
72  Hafnium-177m' 2>  D, see 17&deg;Hf              2E+4        6E+4          2E-5    8E-8        3E-4      3E-3 170                                                                .
W, see      Hf                      9E+4          4E-5    IE-7 B-55                                            NUREG-1736
 
APPENDIX B Table 1                          Table 2              Table3 Oral h;      1 upational Values Col. 2  I Col. 3 I  Col.
Effluent Concentrations 1  I  Col. 2 Releases to Sewers Monthly Ingestion                          Inhalation                                          Average Atomic                      *ALI                          ALI      DAC          A i r        W  a t e      Concentration No. Radionuclide      (&#xb5;Ci) Class                  (&#xb5;Ci)    (&#xb5;Ci/ml)      (&#xb5;Cl/ml)      (&#xb5;Ci/ml)      (&#xb5;Ci/ml) 72    Hafnium- 178m    D, see 170Hf        3E+2        lE+O        SE-10          -          3E-6          3E-5 Bone Surf
                                                -        (2E+O)          -        3E-12            -            -
17 W, see &deg;Hf                      5E+O          2E-9          -              -              -
Bone Surf (9E+O)          -        lE-11            -              -
17 72    Hafnium- 179m    D, see &deg;Hf          1E+3        3E+2          lE-7                        lE-5          IE-4 Bone Surf (6E+2)          -        8E-IO            -
17 W, see &deg;1-If    I    -        6E+2          3E-7        8E-10            -
72    Hafuium-l 80m    D, see  170 Hf  I    7E+3        2E+4          9E-6        3E-8          lE-4          IE-3 W, see  170 Hf I      -        3E+4    I      lE-5    I      4E-8      I    -  I      -
17 72    Hafuium-181      D, see &deg;1-If        1E+3        2E+2          7E-8          -          2E-5          2E-4 Bone Surf
                                                -        (4E+2)          -        6E-10              -
W, see  170 Hf  I    -        4E+2          2E-7      6E-10              -
72    Hafnium-I 82mm  D, see 170Hf    I    4E+4        9E+4        4E-5          I E-7        5E-4          5E-3 W, see  110 Hf  I    -        IE+5        6E-5          2E-7            -            -
72    Hafnium- 182            17 D, see &deg;1-If        2E+2        SE-I        3E-10            -              -            -
Bone Surf  Bone Surf (4E+2)      (2E+O)          -        2E-12          5E-6          5E-5 W. see 17&deg;1-If        -        3E+O          IE-9 Bone Surf
                                                -        (7E+O)          -        IE-11            -
21        170 72    Hafnium-183'    D. see    Hf        2E+4        5E+4        2E-5        6E-8          3E-4          3E-3 W. see  110 Hf        -        6E+4        2E-5        8E-8            -
72    Hafnium- I 84    D. see 17&deg;Hf    I    2E+3        8E+3          3E-6        lE-8      l    3E-5  I    3E-4 17 W, see &deg;1-If    I      -        6E+3    I    3E-6    I      9E-9      I    -  I      .
NUREG- 1736                                        B-56
 
APPENDIX B Table 1                      Table 2          Table3 Occupational Values                Effluent        Releases to Concentrations        Sewers Col.I      Col.2        Col.3    Col. I        Col.2 Oral                                                      Monthly Ingestion        Inhalation                                Average Atomic                                                ALI        ALI      DAC            Air        Water  Concentration No. Radionuclide                    Class        (&#xb5;Ci)      (&#xb5;Ci)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml)      (&#xb5;Ci/ml)  (&#xb5;Ci/ml) 73  Tantalum-17i<2>    W, all compounds except    4E+4        lE+S          5E-5    2E-7          5E-4      SE-3 those given for Y Y, elemental Ta, oxides,              lE+S          4E-5      IE-7 hydroxides, halides, carbides, nitrates, and nitrides 73  Tantalum-173    I W, see "'ra 1
7E+3        2E+4          8E-6      3E-8        9E-5      9E-4 Y, see 1"'ra                          2E+4          7E-6      2E-8 12            172 73  Tantalum- 174 >    W, see      Ta            3E+4        1E+5        4E-5      lE-7        4E-4      4E-3 17 Y, see 2Ta                            9E+4          4E-5      IE-7 73  Tantalum- 175      W, see 172Ta              6E+3        2E+4          7E-6      2E-8        8E-5      8E-4 17 Y, see 1Ta                              IE+4        6E-6      2E-8 73  Tantalum-176      W, see 17 2Ta              4E+3        1E+4        SE-6      2E-8        SE-5      5E-4 Y, see 172Ta                            1E+4        5E-6      2E-8 17 73  Tantalum- 177      W, see 1'a                  IE+4      2E+4          8E-6      3E-8        2E-4      2E-3 17 Y, see 1'a                            2E+4          7E-6      2E-8 73  Tantalum-I 78      W, see 172Ta              2E+4        9E+4          4E-5      IE-7        2E-4      2E-3 172 Y, see Ta                              7E+4          3E-5      IE-7 17 73  Tantalum- 179      W, see 2Ta                2E+4        5E+3          2E-6      8E-9        3E-4      3E-3 Y, see 172Ta                          9E+2          4E-7      lE-9 172 73  Tantalum- I 80m    W, see      Ta            2E+4        7E+4          3E-5      9E-8        3E-4      3E-3 172 Y, see      Ta                        6E+4          2E-5    8E-8 73  Tantalum- 180      W, see 172Ta              IE+3        4E+2          2E-7    6E-10        2E-5      2E-4 Y, see 172Ta                  -        2E+l          IE-8    3E-l 1 73  Tantalum-          W, see inTa                2E+5        5E+5          2E-4    8E-7 182m12>                                      St. wall (2E+5)        -            -        -          3E-3      3E-2 Y, see 172Ta                  -        4E+5        2E-4      6E-7 73  Tantalum- 182      W. see 178I'a              8E+2        3E+2          IE-7    SE-10          IE-5      IE-4 17 Y. see 2Ta                  -    I  IE+2    I    6E-8  I  2E-10 B-57                                                NUREG- 1736
 
APPENDIX B Table 1                        Table2            Table3 Occupational Values                Effluent        Releases to Concentrations        Sewers Col. I      Col.2        Col.3    Col. I      Col. 2 Oral                                                      Monthly Ingestion          Inhalation                                Average Atomic                                                    ALI        ALI      DAC          Air        Water  Concentration No. Radionuclide                  Class          (&#xb5;Ci)      (&#xb5;Ci)      (&#xb5;Ci/ml) _ (&#xb5;Ci/ml)      (&#xb5;Ci/ml)  (&#xb5;Ci/ml) 73    Tantalum- 183      W, see 172Ta                9E+2        JE+3        5E-7      2E-9            -          -
LLI wall (IE+3)          -            -        -          2E-5      2E-4 17 Y, see 2Ta                    -          IE+3        4E-7      IE-9            -          -
73    Tantalum- I 84    W,see 1~a                    2E+3        5E+3          2E-6      SE-9          3E-5      3E-4 17 Y, see 2Ta                    -        5E+3          2E-6      7E-9            -          -
121          17 73    Tantalum- 185      W, see 2Ta                  3E+4        7E+4          3E-5      IE-7        4E-4      4E-3 Y,seemTa                      -        6E+4          3E-5      9E-8            -          -
73    Tantalum-186121    W, see 172Ta                5E+4        2E+S          IE-4    3E-7            -          -
St. wall (7E+4)          -            -        -          IE-3      IE-2 Y, see 17 2Ta                  -        2E+5          9E-5      3E-7            -          -
74    Tungsten- 176      D, all compounds            IE+4        5E+4          2E-5      7E-8          IE-4      JE-3 74    Tungsten- 177      D, all compounds              2E+4        9E+4          4E-5      JE-7        3E-4      3E-3 74    Tungsten-I 78      D, all compounds              5E+3        2E+4          8E-6      3E-8          7E-5      7E-4 74    Tungsten-179'2>    D, all compounds              5E+5        2E+6          7E-4      2E-6          7E-3      7E-2 74    Tungsten- 18 I    D, all compounds              2E+4        3E+4          IE-S      5E-8          2E-4      2E-3 74    Tungsten- 185      D, all compounds              2E+3        7E+3          3E-6      9E-9            -
LLlwall (3E+3)          -                                  4E-5      4E-4 74    Tungsten- 187      D, all compounds              2E+3        9E+3          4E-6      IE-8          3E-5      3E-4 74    Tungsten-188      D, all compounds            4E+2        IE+3          5E-7      2E-9            -          -
LLiwall (5E+2)          -            -        -          7E-6      7E-5 75    Rhenium-171    2
                          ' D, all compounds except      9E+4        3E+5          IE-4      4E-7            -          -
those given for W          St. wall (IE+5)          -            -        -          2E-3      2E-2 W, oxides, hydroxides, and      -        4E+5          IE-4      5E-7            -          -
nitrates NUREG- 1736                                                  B-58
 
APPENDIX B Table 1                      Table 2          Table 3 Occupational Values                Effluent        Releases to Concentrations        Sewers Col. I        Col.2        Col. 3  Col. I      Col.2 Oral                                                        Monthly Ingestion            Inhalation                                Average Atomic                                          ALl            ALl      DAC          Air        Water  Concentration No.        Radionuclide                Class  (&#xb5;Ci)          (&#xb5;Ci)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml)      (&#xb5;Ci/ml)    (&#xb5;Ci/ml) 75  ; *Rhenium-17g!2> D, see 177Re          7E+4          3E+5
* lE-4
* 4E-7            -          -
St. wall (1E+5)            -            -        -          lE-3        lE-2 W, see 177Re              -          3E+5          lB-4  -4E-=7            -          -
177 75      Rheni_um-181  D, see      Re        5E+3          9E+3          4E-6    IE-8        7E-5        7E-4 W, see "Re17
                                                    -          9E+3          4E-6    lE-8            -          -
177 75      Rhenium-182    D,see        Re        7E+3          1E+4          58-6    2E-8        9E-5
* 9E-4 (12.7h)
W, see 177Re              -          2E+4          68-6    2E-8            -          -
177 75      Rhenium-182    D, see      Re        1E+3          2E+3          IE-6    3E-9        2E-5        2E-4 (64.0h)
W, see    177 Re          -          2E+3          9E-7    3E-9            -          -
17 75    Rhenium-184m    D, see "Re            2E+3          3E+3          )E-6    4E-9          3E-5        3E-4 W, see mRe                -          4E+2        28-7    6E-10            -          -
177 75    Rhenium-184    D, see      Re        2E+3          4E+3          IE-6    5E-9        3E-5        3E-4 W, see    177 Re          -          1E+3        6E-7      2E-9.          -          -
17 75    Rhenium-186m    D, see 7Re            1E+3          2E+3          7E-7        -            -          -
St. wall      St. wall (2E+3)        (2E+3)            -    3E-9          2E-5        2E-4 W, see  177 Re          -          2E+2          6E-8    28-10            -          -
177 75      Rhenium-186    D, see      Re        2E+3          3E+3          lE-6    4E-9          3E-5        3E-4 W, see  177 Re          -          2E+3          7E-7    2E-9            -          -
75      Rhenium~ 187    D, see  177 Re        6E+5          8E+5          4E-4      -          8E-3        SE-2 St. wall
                                                  -        (9E+5}            -    IE-6            -          -
W, see  177 Re          -          IE+5          4E-5    IE-7            -          -
177 75      Rhenium-I 88mrn D,see      Re        8E+4          IE+S          6E-5    2E-7          IE-3        IE-2 W, see  177 Re          -          IE+5          6E-5    2E-7            -          -
177 75    Rhenium- 188    D,see      Re        2E+3          3E+3          IE-6    4E-9        2E-5        2E-4 W. see  177 Re          -          3E+3          IE-6    4E-9            -          -
177 75      Rhenium-I 89    D, see      Re        3E+3          5E+3          2E-6    7E-9          4E-5        4E-4 177 W, see      Re                      4E+3          2E-6    6E-9            -
8-59                                                NUREG- 1736
 
APPENDIX B Table 1                          Table2            Table3 Occupational Valoes                  Effluent        Releases to Concentrations          Sewers Col. I      CoJ.2        Col.3    Col. I          CoJ.2 Oral                                                          Monthly Ingestion          Inhalation                                    Average Atomic                                              ALI          ALI      DAC            Air          Water    Concentration No. Radionuclide                Class        (&#xb5;Ci)        (&#xb5;Ci)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml)        (&#xb5;Ci/ml)    (&#xb5;Ci/ml) 76    Osmium- 1sci2> D, all compounds except      IE+5        4E+5          2E-4      SE-1            lE-3        IE-2 those given for W and Y W, halides and nitrates        -        SE+S          2E-4      7E-7          ...  -          -
Y, oxides and hydroxides.      -        SB+S.        28-4. _(,E.;._7          -          -
180 76    Osmium-181 121 D, see      0s              IE+4        4E+4          2E-5      6E-8            2E-4        2E-3 W, see    180 0s              -        SE+4          2E-5      6E-8              -          -
Y, see  180 0s                -        4E+4          2E-5      6E-8              -          -
1 76    Osmium-I 82    D, see S00s                2E+3        6E+3        2E-6      SE-9            3E-5        3E-4 W, see    180 0s                -        4E+3        2E-6      6E-9              -          -
1 Y, see S00s                    -        4E+3        2E-6      6E-9              -          -
180 76    Osmium- 185    D, see      0s              2E+3        5E+2          2E-7    7E-IO            3E-5        3E-4 W, see    180 0s                -        8E+2        3E-7      IE-9              -          -
Y, see  180 0s                -        8E+2        3E-7      IE-9              -          -
180 76    Osmium- 189m  D, see      0s              8E+4        2E+S          lE-4    3E-7            lE-3        IE-2 W, see    180 0s                -        2E+5        9E-5      3E-7              -          -
1 Y, see S00s                    -        2E+5        7E-5      2E-7              -          -
180 76    ::>smium-19lm  D, see      0s              IE+4        3E+4        IE-5      4E-8            2E-4        2E-3 W, see nioos                  -        2E+4        SE-6      3E-8              -          -
Y, see 180 0s                -        2E+4        7E-6      2E-8              -          -
76    Jsmium-191    D, see nioos                2E+3        2E+3          9E-7    . 3E-9              -          -
LLlwall (3E+3)          -            -          -            3E-5        3E-4 W,see    180 0s                -        2E+3        7E-7      2E-9              -          -
Y, see 180 0s                -        1E+3        6E-7      2E-9              -          -
10 76    )srniurn-193  D, see
* 0s              2E+3        5E+3        2E-6      6E-9              -          -
LLiwall (2E+3)          -            -          -            2E-5        2E-4 W, see  180 0s                -        3E+3          IE-6    4E-9                -          -
Y, see 1 0
                                "  0s                -        3E+3          IE-6    4E-9                -          -
NUREG- 1736                                              B-60
 
APPENDIX B Table 1                      Table 2          Table 3 Occupational Values              Effluent      Releases to Concentrations      Sewers Col. t      Col. 2        Col.3  Col.1        Col.2 Oral                                                    Monthly Ingestion          Inhalation                            Average Atomic                                              ALl        ALI      DAC          Air      Water  Concentration No.      Radionuclide                  Class      (&#xb5;Ci)      (&#xb5;Ci)      (&#xb5;Cl/ml) (&#xb5;Ci/ml)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml) 76  Osmium- 194    D, see 1800s                4E+2        4E+l          2E-8  6E-11          -
LLiwall (6E+2)          -                              8E-6      8E-5 W, see 1S00s                            6E+l          2E-8  8E-ll          -
I Y, see S00s                              8E+O          3E-9    lE-11          -
77  Iridium-182'2)  D, all compounds except      4E+4        IE+S        6E-5    2E-7          -
those given for W and Y    St. wall (4E+4)          -                              6E-4      6E-3 W, halides, nitrates, and      -        2E+5          6E-5    2E-7          -          -
metallic iridium Y, oxides and hydroxides      -          IE+S        5E-5    2E-7          -        -
182 77  Iridium-184    D, see      lr              8E+3        2E+4          lE-5    3E-8        IE-4      IE-3 W, see    I82 lr              -        3E+4          lE-5    5E-8          -        -
Y, see  182 lr                -        3E+4          lE-5    4E-8          -        -
1112 77  Iridium- 185    D, see        lr            5E+3        IE+4        5E-6    2E-8        7E-5      7E-4
: w. see    1112 Ir              -        IE+4          5E-6    2E-8          -        -
Y. see  I 2 K  1r                -        IE+4          4E-6    IE-8          -        -
I82 77  Iridium- 186    D. see      Ir              2E+3        8E+3          3E-6    IE-8        3E-5      3E-4 W, see  I82 Ir              -        6E+3          3E-6    9E-9          -          -
Y. see I112 lr                -        6E+3          2E-6    8E-9          -          -
I112 77  Iridium- 187    D. see      lr              1E+4        3E+4          IE-5    5E-8        IE-4      IE-3 W, see  I 2
                                  "  1r              -        3E+4          IE-5    4E-8          -        -
Y, see I112 1r                -        3E+4          IE-5    4E-8          -        -
I82 77  Iridium- 188    D, see      Ir              2E+3        5E+3          2E-6    6E-9        3E-5      3E-4 18 W, see 2Jr                    -        4E+3          IE-6    SE-9          -        -
Y. see 1112 lr                -        3E+3          IE-6    SE-9          -        -
B-6 1                                            NUREG- 1736
 
APPENDIXB Table 1                      Table2            Table3 Occupational Values              Effluent        Releases to Concentrations        Sewers CoJ. l      CoJ.2        Col.3  Col.l          CoJ.2 Ora)                                                      Monthly Ingestion            Inhalation                              Average Atomic                                              ALI          ALI      DAC          Air        Water  Concentration No.      Radionuclide                    Class    (uCi)        (uCi)      (uCi/ml) (uCi/ml)      (uCi/mJ)  (&#xb5;Ci/ml) 77    Iridiwn-1 89    D, see 182Jr              5E+3        5E+3          2E-6    7E-9            -
LLI wall (5E+3)          -                    -          7E-S      7E-4 W, see    182 lr                      4E+3          2E-6    SE-9      I -
Y, see lr182
                                                      -          4E+3          IE-6    5E-9            -
18 77    lridium-190m12J D, see 2Jr                2E+S        2E+S          8E-5    3E-7        2E-3      2E-2 W, see 182Jr                            2E+5          9E-5    3E-7            -
Y. see Ir182 2E+5          SE-5    3E-7            -        -
182 77    Iridium- 190    D, see        lr          1E+3        9E+2          4E-7    IE-9          IE-5      lE-4 182 W, see        Ir                        IE+3        4E-7    IE-9 Y, see 182lr                            9E+2          4E-7    IE-9            -
182 77    lridium- l 92m  D, see        lr          3E+3        9E+l          4E-8    IE-10        4E-5      4E-4 18 W, see 2Ir                              2E+2          9E-8    3E-10            -
18 Y. see 2Ir                              2E+I          6E-9  2E-ll            -
182 77    lridium-192    D, see        lr          9E+2        3E+2          IE-7  4E-10          lE-5      IE-4 W, see    11
* 1r            -          4E+2          2E-7  6E-10            -        -
Y, see  12
                                  " Ir              -          2E+2          9E-8  3E-10            -          -
182 77  Iridium- l 94m  D, see        Ir          6E+2        9E+l          4E-8    lE-10        9E-6      9E-5 182 W, see        Ir            -          2E+2          7E-8  2E-l0            -          -
Y, see* 182 lr            -          IE+2          4E-8    IE-10          -          -
11 77  lridium-194      D, see    " Ir            IE+3        3E+3          IE-6    4E-9          lE-5      IE-4 W, see '" 2Ir                -          2E+3          9E-7    3E-9            -          -
Y, see  1 2
                                  " Ir                -          2E+3          SE-7    3E-9            -          -
18 77    Iridium- l 95m  D, see 2:tr                8E+3        2E+4          IE-5    3E-8          IE-4      lE-3 W, see  182 lr            -          3E+4          lE-5    4E-8            -          -
Y, see  182 lr              -          2E+4          9E-6    3E-8            -          -
2 77    lridium-195      D, see '" 1r              IE+4        4E+4          2E-5    6E-8          2E-4      2E-3 W, see  12
                                    ~ lr                        5E+4          2E-5    7E-8            -
11 Y, see  " Ir                          4E+4          2E-5    6E-8            -
NUREG-1736                                              B-62
 
APPENDIX B Table 1                      Table 2          Table 3 Occupational Values                Effluent      Releases to Concentrations      Sewers Col. 1      Col. 2        Col. 3  Col. 1        Col.2 Oral                                                    Monthly Ingestion          Inhalation                              Average Atomic                                                ALl        ALl      DAC          Air        Water  Concentration No. Radionuclide                  Class        (&#xb5;Cl)      (&#xb5;Ci)      (&#xb5;Ci/ml)  (&#xb5;Ci/ml)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml) 78  Platinum- 186      D, all compounds          1E+4        4E+4          2E-5    5&8          2E-4      2E-3 78  Platinum-188        D, all compounds          2E+3        2E+3          7E-7    2E-9        2E-5      2E-4 78  Platinum- 189      D, all compounds          1E+4        3E+4          IE-5    4E-8        lE-4        IE-3 78  Platinum-191        D, all compounds          4E+3        8E+3          4E-6      IE-8        SE-5      SE-4 78  Platinum-193m      D, all compounds          3E+3        6E+3          3E-6    SE-9          -
LU wall (3E+4)          -                                4E-5      4E-4 78  Platinum-193        D, all compounds          4E+4        2E-t4          IE-5    3E-8 LU wall (5E+4)          -                                6E-4      6E-3 78  Platinum-195m      D, all compounds          2E+3        4E+3          2E-6    6E-9          -
LLI wall (2E+3)          -                                3E-5      3E-4 21 78  Platinum- l 97m'    D, all compounds          2E+4        4E+4          2E-5    6E-8        2E-4      2E-3 78  Platinum-197        D, all compounds          3E+3        IE+4          4E-6    IE-8.        4E-S.      4E-4 78  Platinum-199' 21    D, all compounds          5E+4        IE+S        6E-5    2E-7        7E-4      7E-3 78  Platinum-200        D, all compounds          1E+3        3E+3          lE-6    5E-9        2E-5      2E-4 79  Gold-193            D, all compounds except    9E+3        3E+4          JE-5    4E-8        JE-4      IE-3 those given for W and Y W, halides and nirrares                2E+4          9E-6    3E-8          -
Y, oxides and hydroxides              2E+4          SE-6    3E-8          -
79    Gold-194            D, see 193Au              3E+3        8E+3          3E-6    lE-8        4E-5      4E-4 W, see 193Au                          5E+3          2E-6    8E-9          -
Y, see  193 Au                          5E+3          2E-6    7E-9          -
3 79    Gold-195            D, see '" Au              5E+3        IE+4          SE-6    2E-8        7E-5      7E-4 W, see "~Au                            IE+3          6E-7    2E-9 Y, see 193Au                          4E+2          2E-7    6E-10          -
B-63                                              NUREG-1 736
 
APPENDIX B Table 1                      Table2          Table 3 Occupational Values              Effluent      Releases to Concentrations      Sewers Col.1        Col.2        Col.3  Col.1        Col.2 I                            Oral                                                    Monthly Ingestion            Inhalation                            Average Atomic                                              ALl          ALl      DAC        Air        Water  Concentration No.      Radionuclide                  Class      (&#xb5;Ci)        (&#xb5;Ci)      (&#xb5;Ci/ml) (&#xb5;Ci/ml)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml) 193 79    Gold-198m      ! D, see      Au            1E+3        3E+3          IE-6  4E-9          IE-5      IE-4 W,see 193 Au                -          IE+3        5E-7    2E-9          -          -
Y, see  193 Au                -        1E+3        5E-7    2E-9          -          -
193 79    Gold-198          D, see      Au            IE+3        4E+3          2E-6    SE-9        2E-5      2E-4 W, see  193 Au              -        2E+3          8E-7    JE-9            -          -
Y,see  193 Au                -        2E+3          7E-7    2E-9          -          -
79    Gold-199          D,see  193 Au            3E+3 LLlwall 9E+3          4E-6    IE-8          -        -
(3E+3)            -            -      -          4E-5      4E-4 W,see    193 Au              -        4E+3          2E-6    6E-9            -        -
* Y, see  193 Au              -        4E+3          2E-6    5E-9            -          -
193 79    Gold-200m        D. see    Au            lE+J        4E+3          IE-6  5E-9        2E-5      2E-4 W, see '"3Au                  -        3E+3          IE-6  4E-9            -          -
Y. see  193 Au              -          2E+4          IE-6  3E-9          -          -
21              193 79    Gold-200          D,see      Au            3E+4          6E+4          3E-5    9E-8        4E-4      4E-3 W,see    193 Au              -          8E+4          3E-5    IE-7          -*        -
Y. see  193 Au              -          7E+4          3E-5    IE-7          -          -
79    Gold-20 1121 D. see  193 Au            7E+4          2E+5          9E-5    JE-7          -          -
St. wall (9E+4)          -            -      -          IE-3      IE-2 W. see mAu                  -          2E+5          IE-4    3E-7          .          .
Y. see  13
                                    " Au                .          2E+5          9E-5    3E-7          -          -
80    Mercury-193m      Vapor                        -        8E+3          4E-6    IE-8          -          -
Organic D                4E+3          IE+4          SE-6    2E-8        6E-5      6E-4 D, sulfates              3E+3        9E+3          4E-6    lE-8        4E-5      4E-4 W, oxides, hydroxides,      -        8E+3          3E-6    IE-8          -          -
halides . nitrates, and sulfides NUREG- I 736                                                B-64
 
APPENDIX B Table 1                      Table 2          Table 3 Occupational Values                Effluent      Releases to Concentrations      Sewers Col.1      Col. 2        Col. 3  Col.1        Col. 2 Oral                                                    Monthly Ingestion          Inhalation                              Average Atomic                                    ALI        ALI      DAC          Air      Water  Concentration No. Radionuclide              Class  (&#xb5;Ci)      (&#xb5;Ci)      (&#xb5;Ci/ml)  (&#xb5;Ci/ml)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml) 80  Mercury- 193 Vapor                -        3E+4          IE-5    4E-8            -        -
Organic D          2E+4        6E+4          3E-5    9E-8        3E-4      3E-3 D, see 193mHg      2E+4        4E+4          2E-5    6E-8        2E-4      2E-3 W, see 193 mHg        -        4E+4          2E-5    6E-8            -        -
80  Mercury- 194 Vapor                -        3E+l          lE-8  4E-ll            -        -
OiganicD            2E+l        3E+l          lE-8  4E-ll        2E-7      2E-6 D, see 193 mHg      8E+2        4E+I          2E-8    6E-11        lE-5      lE-4 W, see 193 mHg        -        1E+2          SE-8    2E-10            -        -
80  Mercury-195m Vapor                -        4E+3          2E-6    6E-9            -        -
Organic D          3E+3        6E+3          3E-6      SE-9        4E-5      4E-4 D, see IY)mHg      2E+3        5E+3          2E-6      7E-9        3E-5      3E-4 W, see 193 mHg        -        4E+3          2E-6      5E-9          -        -
80  Mercury- 195 Vapor                -        3E+4          lE-5    4E-8            -        -
OrganicD            2E+4        5E+4          2E-5    6E-8        2E-4      2E-3 D, see 193mHg      JE+4        4E+4          lE-5    5E-8        2E-4      2E-3 W' see 193mHg        -        3E+4          lE-5    5E-8            -        -
80  Mercury-197m Vapor                -        5E+3          2E-6      7E-9          -        -
Organic D          4E+3        9E+3          4E-6      IE-8        SE-5      5E-4 D. see m"'Hg        3E+3        7E+3          3E-6      lE-8        4E-5      4E-4 W, see IYJmHg        -        5E+3          2E-6    7E-9            -        -
80  Mercury-197  Vapor                -        8E+3          4E-6      lE-8          -        -
OrganicD            7E+3        1E+4          6E-6    2E-8        9E-5      9E-4 D, see 193mHg      6E+3        IE+4          5E-6    2E-8        SE-5      SE-4 W, see 193mHg                  9E+3          4E-6      IE-8 B-65                                              NUREG-1736
 
APPENDIX B Table 1                      Table 2          Table 3 Occupational Values              Effluent        Releases to Concentrations        Sewers Col. I      Col.2        Col. 3  Col. I        Col. 2 Oral                                                      Monthly Ingestion          Inhalation                              Average Atomic                                              ALI          ALI      DAC          Air        Water  Concentration No.      Radionuclide                    Class  (&#xb5;Ci)        (&#xb5;Ci)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml)      (&#xb5;Ci/ml)  (&#xb5;Ci/ml) 80  Mercury-l99m<2>    Vapor                    -        8E+4          3E-5    IE-7            -          -
Organic D              6E+4        2E+5        7E-5      2E-7            -          -
St. wall (1E+5)          -            -        -          IE-3      IE-2 D, see 193"'Hg        6E+4        lE+S        6E-5    2E-7          8E-4        BE-3 W, see 193"'Hg            -        2E+5        7E-5      2E-7            -          -
80  Mercury-203        Vapor                    -        SE+2        4E-7      lE-9            -          -
Organic D              5E+2        8E+2        3E-7      IE-9        7E-6      7E-5 D, see 193rnHg        2E+3        1E+3        5E-7    2E-9          3E-5      3E-4 W, see  193
                                          '"Hg        -        1E+3        SE-7    2E-9            -          -
81  Thallium-l 94m'!1 D, all compounds          SE+4        2E+5        6E-S    2E-7            -          -
St. wall (7E+4)          -            -        -          IE-3      lE-2 81  Thallium-19412)      D, all compounds      3E+5        6E+S        2E-4    BE-7            -          -
St. wall (3E+5)          -            -        -          4E-3      4E-2 11 81  Thallium-195'        D, all compounds      6E+4        IE+S        5E-5    2E-7          9E-4      9E-3 81  Thallium-197        D, all compounds      7E+4        1E+5        5E-5    2E-7          IE-3      lE-2 21 81  fhallium-l 98m'      D, all compounds      3E+4        5E+4        2E-5    8E-8          4E-4      4E-3 81  fhallium-198        D, all compounds      2E+4        3E+4        IE-5    SE-8          3E-4      3E-3 81  fhallium-199        D, all compounds      6E+4        8E+4        4E-5    IE-7          9E-4      9E-3 81  Thallium-200        D. all compounds      8E+3        IE+4        SE-6    2E-8          IE-4      IE-3 81  fhallium-201        D. all compounds      2E+4        2E+4        9E-6    3E-8          2E-4      2E-3 81    fhallium-202        D, all compounds      4E+3        5E+3          2E-6    7E-9          SE-5      SE-4 81    fhallium-204        D, all compounds      2E+3        2E+3          9E-7    3E-9        2E-S        2E-4 82    :..ead-195m12l      D, all compounds      6E+4        2E+5          XE-5    3E-7        8E-4        BE-3 82    Jead-198            D, all compounds      3E+4        6E+4          3E-5    9E-8        4E-4      4E-3 82    Jead-199(2)          D, all compounds      2E+4        7E+4          3E-5    IE-7        3E-4        3E-3 82    Jead-200            D, all compounds      3E+3        6E+3          3E-6    9E-9        4E-5      4E-4 82    ~d-201              D, all compounds      7E+3        2E+4          BE-6    3E-8          IE-4      IE-3 NUREG- 1736                                              B-66
 
APPENDIX B Table 1                      Table 2          Table 3 Occupational Values              Effluent        Releases to Concentrations        Sewers Col.I        Col.2        Col.3  Col. I        Col. 2 Oral                                                    Monthly Ingestion          Inhalation                              Average Atomic                                            ALI        AL1      DAC          Air        Water  Concentration No. Radionuclide                  Class      (&#xb5;Ci)      (&#xb5;Ci)      (&#xb5;Ci/ml) (&#xb5;Ci/ml)      (&#xb5;Ci/ml)  (&#xb5;Ci/ml) 82  Lead-202m        D, all compounds          9E+3        3E+4          IE-5    4E-8          IE-4      IE-3 82  Lead-202        D, all compounds          1E+2        SE+I          2E-8  7E-ll          2E-6      2E-S 82  Lead-203        D, all compounds          SE+3        9E+3          4E-6    IE-8          7E-5      7E-4 82  Lead-205        D, all compounds        4E+3          1E+3          6E-7    2E-9          SE-5      SE-4 82  Lead-209        D, all compounds          2E+4        6E+4          2E-S    SE-8          3E-4      3E-3 82  Lead-210        D, all compounds          6E+l        2E+l          IE-10      -
Bone Surf    Bone Surf (lE+O)      (4E-l)            -  6E-13          IE-8      IE-7 2
82  Lead-21 1< >    D, all compounds          1E+4        6E+2          3E-7  9E-10        2E-4      2E-3 82  Lead-21 2        D, all compounds          8E+l        3E+l          IE-8  SE-II            -
Bone Surf
                                              *(1E+2)          -                                2E-6        2E-5 12 82  Lead-214    >    D, all compounds        9E+3        8E+2          3E-7    lE-9          IE-4      IE-3 83  Bismuth-200 2>  D, nitrates              3E+4        8E+4          4E-5    lE-7        4E-4      4E-3 W, all other compounds                IE+S          4E-S    lE-7          -
83  Bismuth-201"'    D,see~i                  IE+4        3E+4          IE-5  4E-8          2E-4      2E-3 0
W, seew Bi                            4E+4          2E-5    5E-8            -
2 83  Bismuth-202<  > D,see~i                  IE+4        4E+4          2E-5    6E-8          2E-4      2E-3 W, seeiooei                          8E+4          3E-5    IE-7          -
1 83  Bismuth-203      D. see ooi3i            2E+3        7E+3          3E-6    9E-9          3E-5      3E-4 2
W, see  ooi3i                      6E+3          3E-6    9E-9            -
83  Bismuth-205      D,see~i                  1E+3        3E+3          IE-6    3E-9          2E-5      2E-4 W,see~i                              IE+3          5E-7    2E-9            -
83    Bismuth-206      D,see~i                  6E+2        IE+3          6E-7    2E-9          9E-6      9E-5 W,see~i                              9E+2          4E-7    IE-9            -
83    Bismuth-207      D,see~i                  1E+3        2E+3          7E-7    2E-9          IE-5      IE-4 W,see~i                              4E+2          lE-7  5E-10            -
B-67                                                NUREG- 1736
 
APPENDIX B Table 1                        Table 2          Table 3 Occupational Values                Effluent        Releases to Concentrations        Sewers Col.1      Col.2        Col.J    Col. I      Col. 2 Oral                                                      Monthly Ingestion          Inhalation                                Average Atomic                                                    ALl        ALI      DAC          Air        Water    Concentration No. Radionuclide                    Class          (&#xb5;Ci)      (&#xb5;Ci)      (&#xb5;Ci/ml)  (&i/ml      (&#xb5;Ci/ml)    (&#xb5;Ci/ml) 83  Bismuth-21Om      D, see 2'1li                  4E+l        SE+0          2E-9        -
Kidneys    Kidneys (6E+l)      (6E+0)            -    9E-12        8E-7        8E-6 W, see2ooai                                7E-I        3E-10    9E-13            -
83  Bismuth-210        D, see2ooai                    8E+2        2E+2          IE-7                  lE-5        lE-4 Kidneys (4E+2)            -    SE-10            -
2 W, see '1li                      -        3E+l          IE-8    4E-ll            -          -
83  Bismuth-2 I zc 2 2
D, see '1li                    5E+3        2E+2          IE-7    3E-10        7E-5        7E-4 W, see 2'1li                      -        3E+2          lE-7    4E-10            -          -
121 83  Bismuth-213        D, see 2ooai                  7E+3        3E+2          IE-7    48-10          lE-4        IE-3 W, see 2ooai                      -        4E+2          IE~7    SE-10            -          -
21 83  Bismuth-214'      D, see 2ooai                  2E+4        8E+2          3E-7      lE-9            -          -
St. wall (2E+4)        -              .        -          3E-4        3E-3 W,see:?OOsi                                9E-2        4E-7      lE-9 84  Polonium-203m      D, all compounds except        3E+4        6E+4          3E-5      9E-8        3E-4        3E-3 those given for W W, oxides, hydroxides, and        -        9E+4          4E-5      IE-7 nitrates 84  Polonium-205"'    D, see'&deg;3Po                I  2E+4        4E+4          2E-5      SE-8        3E-4        3E-3 W, see 103Po              I      -        7E+4          3E-5      IE-7                      -
20 84  Polonium-207      D, see 31>o                I  8E+3        3E+4          IE-5      3E-8        lE-4        IE-3 W, see  203 Po            I      -        3E+4          lE-5      4E-8          -
84  Polonium-210      D, see:?03Po              I  3E+0        6E-1        3E-10    9E-13        4E-8        4E-7 W, see:?03Po              I      -        6E-1        3E-10    9E-l3          -
21 85    Astatine-207      D, halides                I  6E+3        3E+3          IE-6    4E-9          8E-5        8E-4 w                          I      -        2E+3          9E-7    3E-9            -
85    Astatine-21 I      D, halides                I  IE+2        8E+I          3E-8    IE-10        2E-6        2E-5 w                          I      -        SE+l    I    2E-8  I    SE-II      I    - I    -
NUREG- 1736                                                    B-68
 
APPENDIX B Table 1                        Table 2            Table3 Occupational Values                Effluent          Releases to Concentrations          Sewers Col.1      Col.2        Col.3    Col.1        Col.2 Oral                                                          Monthly Ingestion          Inhalation                                  Average Atomic                                                ALI        A.LI      DAC          Air        Water      Concentration No. Radionuclide            Class              (.uCi)      (&#xb5;Ci)      (&#xb5;Ci/ml)  (&#xb5;Ci/ml)      (&#xb5;Ci/ml)      (&#xb5;Ci/ml) 86  Radon-220      With    daughters  removed  I  -        2E+4          7E-6    2E-8            -            -
With daughters present                    2E+l          9E-9    3E-ll                          -
(or 12        (or 1.0 working      working level        level) months) 86  Radon-222      With daughters removed                    1E+4          4E-6      IE-8            -            -
With daughters present                    1E+2          3E-8    IE-10            -            -
(or 4      (or 0.33 working      working level        level) months) 87  Francium-22i21  D, all compounds              2E+3      5E+2            2E-7  6E-10          3E-5          3E-4 87  Francium-223121 D, all compounds          I  6E+2        8E+2            3E-7    I E-9        SE-6    I    8E-5 Radium-223      W, all compounds              5E+O        7E-l        3E-IO    9E-13            -            -
II 88 Bone Surf (9E+O)                                              IE-7        IE-6 88  Radium-224      W, all compounds              SE+O        2E+O          7E-10    2E-12            -            -
Bone Surf (2E+l)                                            2E-7          2E-6 88  Radium-225      W, all compounds              8E+O        7E-l        3E-10    9E-13            -            -
I Bone Surf I  (2E+l)                                            2E-7          2E-6 88  Radium-226      W, all compounds              2E+O        6E-I        3E-10    9E-13            -            -
Bone Surf (5E+O)                                            6E-8          6E-7 88  Radium-221 2>  W, all compounds              2E+4        IE+4          6E-6 Bone Surf  Bone Surf (2E+4)      (2E+4)            -      3E-8          3E-4          3E-3 88  Radium-228      W, all compounds              2E+O        IE+O          SE-10    2E-12            -            -
Bone Surf (4E+O)                                            6E-8          6E-7 B-69                                                  NUREG- 1736
 
APPENDIX B Table 1                        Table2          Table3 I          Oc upational Values                  Effluent Concentrations Releases to Sewers Col. I        Col. 2 I Col. 3          Col. I      Col.2 Oral                                                        Monthly Ingestion            Inhalation                                Average Atomic                                              ALI            ALI      DAC            Air        Water  Concentration No. Radionuclide            Class            (&#xb5;Ci)          (&#xb5;Ci)  I (&#xb5;Ci/ml)    (&#xb5;Ci/ml)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml) 89    Actinium-224  D, all compounds except      2E+3          3E+1          lE-8          -          -
those given for W and Y    LUwall      Bone Surf (2E+3)        (4E+l)          -        SE-1 1    I 3E-S      3E-4 W, halides and nitrates  I      -          SE+l    I    2E-8      7E-11 Y, oxides and hydroxides I      -          SE+l    I    2E-8      6E-ll                    -
89  Actinium-225  D, see'Ac                    SE+l          3E-1        lE-10          -          -
LLlwall      Bone Surf (SE+!)        (SE-I)          -      7E-13        7E-7      7E-6 W,see  224 Ac            I      -    I    6E-l    I    3E-I0  I    9E-13    I    .
Y,seemAc                        -          6E-l        3E-10      9E-13 89  Actinium-226  D,see 224Ac                  IE+2          3E+O          lE-9                                -
LLI wall    Bone Surf I  (IE+2)        (4E+O)          -        SE-12      2E-6      2E-5 W, see 224Ac                    .          SE+O          2E-9      7E-12          -
Y,see  224 Ac                                5E+O          2E-9      6E-12          -
89  Actinium-227  D, seemAc                    2E-l          4E-4        2E-13        -
Bone Surf    Bone Surf (4E-1)        (SE-4)          -        IE-15      SE-9      SE-8 W. seell'Ac                                2E-3        7E-13        -
Bone Surf (3E-3)          -        4E-15          -
Y, seell~Ac                  -    I  4E-3      I  2E-12    I    6E-l5    I    -
89  Actinium-228  D,seemAc                    2E+3          9E+0        4E-9                    3E-5      3E-4 Bone Surf (2E+I)          -        2E-11          .
W, seemAc                                  4E+I          2E-8          -
Bone Surf (6E+l)          -        SE-11          .
Y,see=Ac                                  4E+I          2E-8      6E-11          -
NUREG- 1736                                                B-70
 
APPENDIX B Table l                            Table 2                  Table 3 Occupational Values                      Effluent              Releases to Concentrations                Sewers I
Col.1        Col.2        Col.3        CoLl            Col.2 Oral                                                                    Monthly Ingestion            Inhalation                                            Average Atomic                                                ALl            ALI      DAC              Air          Water        Concentration No. Radionuclide              Class              (uCi)          (&#xb5;Ci)    (&#xb5;Ci/ml)      (&#xb5;Ci/ml)      (&#xb5;Ci/ml)          (&#xb5;Ci/ml) .
90  Thorium-2262 > W, all compounds except        5E+3          2E+2          6E-8        2E-10            -
those given for Y            St. wall (5E+3)                        -    l      -    l      7E-5              7E-4 Y, oxides and hydroxides  I      -          IE+2    I    6E-8      I      2E-10      I    -      I    -
90  Thorium-227            22 W, see 6Tb                I    1E+2      3E-l              IE-10      SE-13          2E-6              2E-5 Y, see 226Th                      -          3E-1        IE-10        SE-13            -
90  Thorium-228            22 W, see 6Tb                      6E+O      I    IE-2        4E-1.2          -
Bone Surf      Bone Surf (IE+I)        (2E-2)          -        3E-14          2E-7              2E-6 22 Y, see 6Th                I      -          2E-2        7E-12        2E-14            -                -
90  Thorium-229            22 W, see 6Th                      6E-1          9E-4        4E-13          -              -                -
Bone Surf : Bone Surf
{lE+O)          (2E-3)          -        3E-15          2E-8              2E-7 22 Y, see 6Th                                    2E-3          IE-12          -
Bone Surf (3E-3)          -        4E-15            -
90  Thorium-230            22 W, see "Th                      4E+O Bone Surf 6E-3 Bone Surf      1    3E-12 l l  -            -
I    -
(9E+O)        (2E-2)      I    -    I  2E-14      I    IE-7      I      IE-6
                                                ~
22 Y. see "rh                                    2E-2        6E-12            -
Bone Surf (2E-2)            -        3E-14            -
90  Thorium-23 1  W, see 226n,              I    4E+3          6E+3          3E-6        9E-9          5E-5              SE-4 Y, see 226Tb                      -          6E+3          3E-6        9E-9            -
I                I    -
I    -
I    -
22 90  Thorium-232    W, see "rh                      7E-I          IE-3              SE-13 Bone Surf      Bone Surf (2E+O)        (3E-3)    I      -    I    4E-15      I    3E-8    I      3E-7 22 Y, see "rh
                                              ,1
                                                        -          3E-3 Bone Surf I    IE-12 I
I I
                                                        -          (4E-3)            -    I      6E-15        I    -      I B-71                                                            NUREG- 1736
 
APPENDIX B Table 1                      Table2            Table3 Occupational Values              Effluent          Releases to Concentrations          Sewers Col. l      Col.2        Col.3    Col. I      Col.2 Oral                                                        Monthly Ingestion            Inhalation                                Average Atomic                                                ALI          ALI      DAC          Air      Water      Concentration No. Radionuclide                Class        (&#xb5;Ci)        (&#xb5;Ci)    (&#xb5;Ci/ml) (&#xb5;Ci/ml)    (&#xb5;Ci/ml)      (&#xb5;Ci/ml) 90  Thorium-234      W,see226Th                  3E+2          2E+2        SE-8  3E-10          -            -
LLI wall (4E+2)          .            -        -        SE-6          5E-5 Y,seen&ib                    -          2E+2        6E-8    2E-10          -            .
91  Protactinium-    W, all compounds except    4E+3          1E+2        5E-8    2E-10        SE-5          5E-4 22-r 2>          those given for Y Y, oxides and hydroxides      -          IE+2        4E-8    IE-10          -            -
91    Protactinium-228 W, see 227Pa                1E+3          IE+l        SE-9      -        2E-5          2E-4 Bone Surf
                                                        -        (2E+l)          -    3E-11          -            -
Y, see 227Pa                  -          IE+l        SE-9    2E-1 I        -            -
91  Protactinium-230 W, see mpa                  6E+2          5E+0        2E-9  7E-l2          -            -
Bone Surf (9E+2)            -            -        -        IE-5          IE-4 Y, see 227Pa                  -          4E+O        IE-9  SE-12          -            -
91    Protactinium-231 W,see  227 Pa              2E-l          2E-3        6E-13      -          -            -
Bone Surf    Bone Surf (SE-I)        (4E-3)          -    6E-15        6E-9          6E-8 Y, see 227Pa                              4E-3        2E-12 Bone Surf (6E-3)          -    SE-15          -
227 91    Protactinium-232 W, see      Pa              1E+3          2E+l        9E-9      -        2E-5          2E-4 Bone Surf (6E+l)          -    8E-1 I          -
227 Y, see    Pa                            6E+I          2E-8      -
Bone Surf (7E+I)          -    lE-10          -
91    Protactinium-233 W,see  227 Pa              IE+3        7E+2          3E-7    IE-9          -
LLlwall (2E+3)            -            -        -        2E-5          2E-4 Y, see 227 Pa                            6E+2          2E-7  8E-IO          -
227 91    Protactinium-234 W, see      Pa            2E+3          8E+3          3E-6    IE-8        3E-5          3E-4 NUREG-1736                                                B-72
 
APPENDIX B Table 1                      Table 2          Table 3 Occupational Values              Effluent        Releases to Concentrations.      Sewers Col. I      Col.2        Col.3  Col. I        Col.2 Oral                                                      Monthly Ingestion            Inhalation                              Average Atomic                                                  ALI          ALI      DAC          Air        Water  Concentration No.      Radionuclide                    Class        (&#xb5;Ci)        (&#xb5;Ci)      (&#xb5;Ci/ml) (&#xb5;Ci/ml)      (&#xb5;Cl/ml)  (&#xb5;Ci/ml)
Y, see 227Pa                    -        7E+3          3E-6    9E-9          -
92  Uranium-230        D, UF6, UO2F 2* UO2(NO3) 2    4E+O          4E-l        2E-10      -            -          -
Bone Surf    Bone Sutf
{6E+O)        {6E-l)          -    SE-13          SE-8      SE-7 W, UO,, UF4, UCl 4              -          4E-l        IE-10  SE-13            -          -
Y,UO2, UJOR                      -          3E-l        IE-10  4E-13            -          -
23 92  ll.Jraniwn-23 I    D, see    &deg;1.J              5E+3          8E+3        3E-6    lE-8          -          -
LLI wall (4E+3)          -            -        -          6E-5      6E-4 W,seel3&deg;1.J                      -        6E+3          2E-6    8E-9          -
Y, see  13
                                    &deg;1.J                  -          5E+3        2E-6    6E-9          -          -
92  IUranium-232      D, see  23
                                    &deg;1.J              2E+O          2E-l        9E-ll      -            -          -
Bone Surf    Bone Surf
{4E+0)        {4E-l)          -    6E-13          6E-8      6E-7 W, see 23&deg;1.J                  -          4E-l        2E-10  SE-13            -
Y, see 23&deg;U                    -          8E-3        3E-12    lE-14          -          -
92  llJraniwn-233      D. see 23
                                    &deg;1.J              IE+l          IE+0        SE-IO      -                      -
Bone Surf    Bone Surf (2E+I)        (2E+0)          -    3E-12          3E-7      3E-6 W, see iiou                    -          7E-l        3E-10    IE-12 Y, see i,ou                    .          4E-2        2E-l l  5E-14 92  Uranium-234    131 D, see Z3&deg;tJ                  IE+l        IE+O        SE-10      -
Bone Surf    Bone Surf (2E+l)        (2E+O)          -    3E-l2          3E-7      3E-6 2                      .
W. see 3&deg;1.J                              7E-l        3E-10  IE-12 Y, see 23&deg;1.J                  .          4E-2        2E-l l  SE-14 B-73                                              NUREG- 1736
 
APPENDIX B Table 1                  I        Table 2          Table3 Occupational Values                      Effluent        Releases to I    Concentrations      Sewers Col. I        Col.2        Col.3        Col. I        Col.2 Oral                                                            Monthly Ingestion            Inhalation                                    Average Atomic                                            ALI          ALI      DAC                Air        Water  Concentration No. Radionuclide                  Class    (&#xb5;Ci)        (&#xb5;Ci)      (&#xb5;Ci/ml)        (&#xb5;Ci/ml)    (&#xb5;Ci/ml)  (uCi/mll 92    Uranium-235"'    D, see 23 &deg;U            lE+l          lE+0          6E-10            .
Bone Surf      Bone Surf (2E+l)        (2E+O)            -          3E-12        3E-7      3E-6 W, see 23 &deg;U                          SE-I        3E-10          IE-12 Y,seell&deg;lJ                -          4E-2          2E-l l        6E-14 92    Uranium-236      D, seell&deg;lJ              IE+l        lE+O          5E-10            -
Bone Surf      Bone Surf (2E+l)        (2E+0)            -          3E-12        3E-7      3E-6 W, seell&deg;lJ                -          8E-1        3E-10          lE-12          -
23 Y. see &deg;U                  -          4E-2        2E-11        6E-14 23 92    Uranium-237    D, see &deg;U              2E+3          3E+3            IE-6          4E-9 LLI wall (2E+3)            -              -            -          3E-5      3E-4 w, see mu                  -    I    2E+3    I    7E-7    I      2E-9 Y, see 23 &deg;U                      I  2E+3    I    6E-7    I    2E-9 I              I            I-131 92    Uranium-238      D,see2l&deg;tJ              lE+l              lE+0          6E-10 Bone Surf        Bone Surf (2E+l)        (2E+O)            -          JE-12        3E-7      3E-6 W,see    23
                                      &deg;U                        8E-1          3E-10          lE-12 23 Y, see    &deg;U                        4E-2          2E-l l        6E-14 92    Uranium-239"'    D, see 23&deg;U            7E+4          2E+5          BE-5          3E-7        9E-4      9E-3 2
W, see    3&deg;U                        2E+5          7E-5          2E-7 23 Y, see &deg;U                            2E+5          6E-5          2E-7 23 92    Uranium-240      D, see    &deg;U            IE+3          4E+3          2E-6          5E-9        2E-5      2E-4 W,see 23
                                    &deg;U                    I  3E+3    I    lE-6      I  4E-9 Y, see 23
                                    &deg;U                          2E+3          IE-6          3E-9 NUREG- 1736                                            B-74
 
APPENDIX B Table I                      Table 2          Table 3 Occupational Values                Effluent      Releases to Concentrations      Sewers Col. I      Col.2        Col.3    Col. I      Col.2 Oral                                                    Monthly Ingestion          Inhalation                              Average Atomic                                                ALI          ALI      DAC          Air      Water  Conceniratfon.
No. i Radionuclide                      Class      (&#xb5;Ci)        (&#xb5;Ci)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml) 92    Uranium-          D, see  23
                                      &deg;U              18+1        lE+O        SE-10 naturitl'3>
* Bone Surf    Bone Surf (2E+l)      (2E+O)                  3E-12        3E-7      3E-6 W, see  2 3<\i                        SE-I        3E-10    9E-13 Y,see 23&deg;U                              5E-2        2E-l l  9E-14 93    Neptunium-23:t2>  W, all compounds          lE+S        2E+3          7E-7                  2E-3      2E-2 Bone Surf (5E+2)                  6E-9 121 93    Neptunium-233      W, all compounds          8E+5        3E+6          lE-3    4E-6        lE-2      lE-1 93    Neptunium-234      W, all compounds          2E+3        3E+3          lE-6    4E-9        3E-5      3E-4 93  Neptunium-235        W, all compounds          2E+4        8E+2          3E-7 LLI wall    Bone Surf (2E+4)      (1E+3)                  2E-9        JE-4      3E-3 93    Neptunium-236      W, all compounds          3E+o        2E-2        9E-12 (1.15E+5y)                                Bone Surf    Bone Surf (6E+o)      (SE-5)              . SE-14        9E-8      9E-7.
93    Neptunium-          W, all compounds        *3E+3
* 3E+I          IE-8 236m (22.531)                              Bone Surf    Bone Surf (4E+3)      (7E+I)                  IE-10        5E-5      SE-4 93    Neptunium-237      W, all compou~ds .        SE-1        4E-3        2E-12 Bone Surf . Bone Surf (IE+O)      (lE-2)                  IE-14        2E-8      2E-7 93    Neptunium-238      W, all compounds          IE+3        6E+l          3E-8                  2E-5      2E-4 Bone Surf (2E+2)                  2E-10 93    Neptunium-239      W, all compounds          2E+3        2E+3          9E-7    3E-9 LLI wall (2E+3)                                            2E-5      2E-4 93    Neotunium-24(>'2'  W, all compounds          2E+4        8E+4          3E-5      IE-7        3E-4      3E-3 B-75                                              NUREG-1736
 
APPENDIX B Table 1                        Table2          Table3 Occupational Values                  Effluent      Releases to Concentrations      Sewers Col. I      Col.2        Col.3      Col. I      Col.2 Oral                                                        Monthly Ingestion          Inhalation                                Average Atomic                                              ALI        ALI      DAC            Air      Water  Concentration No. Radionuclide                Class      (&#xb5;Ci)        (&#xb5;Ci)      (&#xb5;Ci/ml)    (&#xb5;Ci/ml)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml) 94  Plutonium-234    W, all compounds except    8E+3        2E+2          9E-8      3E-10        lE-4      1B-3 Pu02 Y,Pu02                      -.        2E+2          SE-8      3E-10          -          -
12          234 94  Plutonium-235 >  W, see      Pu          9E+5        3E+6          IE-3      4E-6        IE-2      lE-1 Y, see 234 Pu                -        3E+6          IE-3
* 3E-6          -          .
94  Plutonium-236    W, see 231>u              2E+O        2E-2          SE-12...      .            -          -
Bone Surf    Bone Surf (4E+-O)      (4E-2)            -      5E-14        6E-8      6E-7 Y, see 234Pu                  -        4E-2          2E-11      6E-14          .          -
234 94  Plutonium-237    -W,see      Pu            IE+4        3E+3          IE-6      5E-9        2E-4      2E-3 Y,seemPu                      -        3E+3          IE-6      4E-9          -          -
94  Plutonium-238    W, see 234Pu              9E-l        7E-3        3E-12          -          -          -
Bone Surf    Bone Surf (2E+-O)      (lE-2)          .        2E-14        2E-8      2E-7 Y,see 23 "Pu                .        2E-2        SE-12      2E-14          -          .
94  Plutonium-239    W. see 234Pu              SE-I        6E-3        3E-12          -          -          .
Bone Surf    Bone Surf (IE+O)      (I E-2)          -        2E-14        2E-8      2E-7 Y. see ~"Pu                  .        2E-2        7E-12          -          -          -
Bone Surf
                                                      -        (2E-2)          -        2E-14          -          -
94  ?lutonium-240    W, see  234 Pu            8E-1        6E-3        3E-12          -          .          -
Bone Surf    Bone Surf (lE+-0)      (IE-2)          -        2E-14        2E-8      2E-7 Y, see 234 Pu                .        2E-2        78-12          -          -          -
Bone Surf
                                                      -        (2E-2)          -        2E-14          -          -
NUREG- 1736                                              B-76
 
APPENDIX B Table 1                        Table 2            Table 3 Occupational Values                  Effluent          Releases to Concentrations          Sewers Col.1        CoJ.2        CoJ.3      CoJ.1      CoJ.2 Oral                                                          Monthly Ingestion            Inhalation                                  Average Atomic                                            ALl          ALI      DAC            Air        Water    Concentration No.      Radionuclide                Class    (&#xb5;Ci)        (&#xb5;Ci)      (&#xb5;Ci/ml)    (&#xb5;Ci/mJ)    (&#xb5;Ci/mJ)      (&#xb5;Ci/ml) 94    Plutonium-241    W, seemPu              4E+l          3E-I        lE-10        -            -
Bone Surf    Bone Surf (7E+1)        (6E-l)          -      SE-13        IE-6          IE-5 Y,see 234Pu                            SE-I        3E-l0        -            -
Bone Surf (lE+O)            -      IE-12          -
94    Plutonium-242    W, see23-4Pu            SE-I          7E-3        3E-12        -
Bone Surf    Bone Surf (lE+o)        (lE-2)          -      2E-14      2E-8          2E-7 Y, see  234 Pu            -          2E-2        7E-12        -            -
Bone Surf
[      -          (2E-2)    I    -  I    2E-14    I    -
94  Plutonium-243              13 W, see 'Ptt            2E+4          4E+4    I    2E-S I    SE-8    I  2E-4          2E-3 Y,see mPu          I      -          4E+4          2E-5      SE-8          -
94  Plutonium-244    W, see 234Pu            SE-1          7E-3        3E-12        -            -
Bone Surf    Bone Surf (2E+o)        (IE-2)            -      2E-14        2E-8          2E-7 Y. see n,Pu                            2E-2        7E-12        -          -
Bone Surf (2E-'.?)          -      2E-14          -
94  Plutonium-245    W, seei 3'Pu            2E+3          5E+3          2E-6      6E-9        3E-5          3E-4 13 Y, see 'Pu                            4E+3          2E-6      6E-9          -
94  Plutonium-246            23 W, see 'Pu              4E+2      3E+2            IE-7        4E-IO              -
LU wall (4E+2)      -                -            -            6E-6    6E-5 2
Y, see J.LPu              -          3E+2          IE-7    4E-10          -
95 95 Americium-237"'
Americium-W, all compounds W, all compounds I  8E+4 4E+4 3E+5 3E+3 IE-4 IE-6 4E-7 IE-3 SE-4 IE-2 SE-3 23812)                                              Bone Surf (6E+3)            -      9E-9          -
95    Americium-239  I W. all comoounds        SE+3          IE+4    I    SE-6  I    2E-8    I  7E-5          7E-4 B-77                                                    NUREG- 1736
 
APPENDIX B Table 1                        Table2          Table3 Occupational Values                  Effluent      Releases to Concentrations        Sewers Col, I      Col.2        Col.3    Col. I        Col.2 Oral                                                        Monthly Ingestion          Inhalation                                Average Atomic                                      ALI          ALI      DAC            Air        Water  Concentration No.      Radionuclide            Class    (&#xb5;Ci)        (&#xb5;Ci)      (&#xb5;Ci/ml) * (&#xb5;Ci/ml)      (&#xb5;Ci/ml)  (&#xb5;Ci/ml) 95  Americium-240  W,  al.I compounds  2E+3          38+3          IE-6      4:S,9        3E-S      38-4 95  Americium-241  W, all compounds      SE-I        6E-3        3E-l'!        -            -        -
Bone Surf    Bone Surf (IE+O)        (IE-2)          -    28-14          2E-8      2E-7 95    Americium-    W, all compounds      88-1        6E-3        38-12        -            -        -
242m                              Bone Surf    Bone Surf (lE+O}        (lE-2}          -    2E-l4          2E-8      28-7 95  Americium-242  W, all compounds    4E+3          SE+l        4E-8        -          SE-S      SE-4 Bone Surf
                                                -        (9E+l)            -      lE-10            -        -
95  Americium-243  W, all compounds      88-1        68-3        38-12        -            -        -
Bone Surf    Bone Surf (IE+O}        (IE-2)          -    28-14          2E-8      28-7 95  Americium-      W, all compounds    6E+4          4E+3        28-6        -            -        -
244m12>                            St. wall    Bone Surf (8E+4)      (7E+3)          -        lE-8          IE-3      lE-2 95    Americium-244  W, all compounds    3E+3        2E+2          SE-8        -          4E-5      48-4 Bone Surf
                                                -        (3E+2)          -      4E-10            -        -
95  l\mericium-245  N, all compounds    3E+4        8E+4          38-5      IE-7        4E-4      4E-3 95  l\mericium-    N, all compounds    5E+4        2E+5          SE-5      3E-7            -        -
Z46m'21                            St. wall (6E+4)          -            -          -          88-4      88-3 95  i\.mericium-    W. all compounds    3E+4          IE+S        4E-5      IE-7          4E-4      4E-3 246' 2 '
96  ::urium-238    W, all compounds    2E+4        1E+3          SE-7      2E-9          2E-4      2E-3 96  Zurium-240      W, all compounds    6E+I          6E-I        2E-IO        -
Bone Surf    Bone Surf (8E+l)      (6E-1)          -      98-13          IE-6      lE-5 96    hrium-241      &#xa5;, all compounds    IE+3        3E+l          lE-8                    2E-5      2E-4 Bone Surf (4E+l)          -      5E-1 l          -
NUREG- 1736                                        B-78
 
APPENDIX B Table 1                        Table 2        I    Table3 Occupational Values                Effluent            Releases to Concentrations      I    Sewers Col.1        Col.2        CoL3    Col. I      Col.2 Oral                                                          Monthly Ingestion            Inhalation                                  Average Atomic                                          ALl        ALl      DAC        Air        Water      Concentration No.        Radionuclide            Class    (&#xb5;Ci)        (&#xb5;Ci)      (&#xb5;Ci/ml) (&#xb5;Ci/ml)      (&#xb5;Ci/ml)      (&#xb5;Ci/ml) 96      Curium-242    W, all compounds      3E+l          3E-1        lE-10                    -              .
Bone Surf    Bone Surf (SE+l)      (3E-1)            . 4E-13          7E-7          7E-6 96    Curium-243      W, all compounds      lE+O        9E-3        4E-12    -              -            -
Bone Surf    Bone Surf (2E+O)      (2E-2)            -  2E-14          3E-8          3E-7 96      Curium-244    W, all compounds      IE+O        IE-2        SE-12                                  -
Bone Surf    Bone Surf (3E+O)      (2E-2)            -  3E-14          3E-8          3E-7 96    Curium-245      W, all compounds      7E-I        6E-3        3E-12    -
Bone Surf    Bone Surf (IE+O)      (IE-2)            -  2E-14          2E-8          2E-7 96    Curium-246      W. all compounds      7E-l        6E-3        3E-12    -              -            -
Bone Surf    Bone Surf (IE+O)      (IE-2)          -    2E-14          2E-8          2E-7 96    Curium-247      W, all compounds      8E-l        6E-3          3E-12                  -              -
Bone Surf    Bone Surf (IE+O)      (lE-2)          -    2E-14          2E-8          2E-7 96    ::::urium-248  W, all compounds      2E-l        2E-3          7E-13
                                          *Bone Surf    Bone Surf (4E-l)      (3E-3)            -    4E-15          SE-9          SE-8 96      Zufium-249"'    W, all compounds      5E+4        2E+4          7E-6                  7E-4          7E-3 Bone Surf
                                                  -      (3E+4)            -    4E-8            -              -
96      :urium-250      -N, all compounds    4E-2        3E-4          IE-13                  -              -
Bone Surf    Bone Surf
{6E-2)      (SE-4)            -    8E-16        9E-IO          9E-9 97      3erkelium-245  N, all compounds      2E+3        IE+3          SE-7  2E-9          3E-5          3E-4 97      :lerkelium-246  N, all compounds      3E+3        3E+3          lE-6  4E-9          4E-5          4E-4 97    I :lerkelium-247  N, all compounds      SE-I        4E-3          2E-12                    -            -
Bone Surf    Bone Surf (IE+O)      (9E-3)            -    IE-14        2E-8    I      2E-7 B-79                                                  NUREG- 1736
 
APPENDIX B Table I                      Table2          Table3 Occupational Values              Effluent      Releases to Concentrations      Sewers Col. I      Col.2        Col.3  Col. I      Col.2 Oral                                                    Monthly Ingestion          Inhalation                            Average Atomic                                                ALI        ALI      DAC          Air      Water  Concentration No.      Radionuclide                Class        (&#xb5;Ci)      (&#xb5;Cl)      (&#xb5;Ci/ml) (&#xb5;Ci/ml)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml) 97    Berkelium-249    W, all compounds          2E+2        2E+O          7E-10      -          -          -
Bone Surf  Bone Surf
{5E+2)      (4E+O)            -    5E-12        6E-6      6E-5 97    Berkelium-250    W, all compounds          9E+3        3E+2          lE-7      -        lE-4      lE-3 Bone Surf
                                                        -      (7E+2)            -    lE-9          -          -
98    Californium-24412)
W, all compounds except    3E+4        6E+2          2E-7  SE-10          -          -
those given for Y        St. wall (3E+4)          -            -      -        4E-4      4E-3 Y, oxides and hydroxides      -        6E+2          2E-7  8E-10          -          -
98    Califomium-246    W, see'Cf                  4E+2        9E+O          4E-9  lE-11        SE-6      SE-5 Y,see'Cf                      -        9E+O          4E-9  lE-11          -          -
98  CalifomilDllll-ffl W,seemCf                  8E+O        6E-2*        3E-11      -          -          -
Bone Surf  Bone Surf (2E+I)      (IE-I)            -  2E-13        2E-7      2E-6 Y, seemCf                    -        IE-I        4E-11    lE-13          -          -
2 98    Califomium-249    W, see 1Cf              5E-1        4E-3        2E-12      -          -          -
Bone Surf  Bone Surf (lE+O)      (9E-3)            -    lE-14        2E-8      2E-7 Y, see ?4-JCf                -        lE-2        4E-12      -          -          -
Bone Surf
                                                        -      (lE-2)          -    2E-14          -        -
98    Califomium-250    W, see 1..,Cf              IE+O        9E-3        4E-12      -          -        -
Bone Surf  Bone Surf (2E+O)      (2E-2)          -    3E-l4        3E-8      3E-7 Y, see 1..,Cf                -        JE-2          IE-II  4E-14          -        -
98  Califomium-25 1    W, see 244Cf                5E-l        4E-3        2E-12      -
Bone Surf  Bone Surf (IE+O)      (9E-3)            -    IE-14        2E-8      2E-7 2
Y, see 4-JCf                            IE-2        4E-12      -
Bone Surf (lE-2)            -  2E-14          -
NUREG-1736                                                B-80
 
APPENDIX B Table 1                      Table 2          Table3 Occupational Values              Effluent        Releases to Concentrations      Sewers Col. 1      Col. 2        Col. 3  Col. 1      Col.. 2 Oral                                                    Monthly Ingestion          Inhalation                              Average Atomic                                            ALl        ALl      DAC          Air        Water  Concentration No.        Radionuclide                Class  (&#xb5;Ci)      (&#xb5;Ci)      (&#xb5;Ci/ml)  (&#xb5;Ci/ml)    (&#xb5;Ci/ml)  (&#xb5;Ci/ml) 98      Califomium-252  w. see 244Cf        2E+O        2E-2          8E-12      -            -          -
Bone Surf  Bone Surf (5E+0)      (4E-2)            -    5E-14        7E-8      7E-7 Y, see 244 Cf          -          3E-2        lE-1 l  SE-14            -          -
98      Califomium-253  W, see  244 Cf        2E+2        2E+O
* SE-10    3E-12          -          -
Bone Surf (4E+2)          -            -        -        5E-6      SE-5
                          *Y,see 2~f              -        2E+0          7E-10  2E-12            -          -
98      Califomium-254  W,see 2.>>Cf          2E+0        2E-2        9E-12    3E-14        3E-8      3E-7 2
Y, see .>>Cf            -          2E-2        7E-12    2E-14          -          -
99      Einsteinium-250  W, all compounds    4E+4        5E+2          2E-7      -        6E-4      6E-3 Bone Surf
                                                  -        (IE+3)            -      2E-9          -          -
99      Einsteinium-251  W, all compounds    7E+3        9E+2          4E-7      -          lE-4      IE-3 Bone Surf
                                                  -        (IE+3)            -    2E-9            -          -
99      Einsteinium-253  W, all compounds    2E+2        lE+0          6E-10  2E-12        2E-6      2E-5 99      Einsteinium-    W. all compounds    3E+2        IE+I          4E-9    IE-I I          -          -
254m                                LU wall (3E+2)        -              -      -          4E-6      4E-5 99      Einsteinium-254  W, all compounds    SE+0        7E-2          3E-1 I    -            -          -
Bone Surf  Bone Surf (2E+l)      (IE-I)            -    2E-13        2E-7      2E-6 100  I Fennium-252        W, all compounds    5E+2        IE+I          SE-9  2E-I 1        6E-6      6E-5 100    I Fermium-253      W, all compounds    IE+3        IE+l          4E-9    lE-11        lE-5      IE-4 100  I Fermium-254        W. all compounds    3E+3        9E+l          4E-8    IE-10        4E-5      4E-4 100      Fennium-255      W, all compounds    5E+2        2E+l          9E-9  3E-l l        7E-6      7E-5 JOO      Fermium-257      W, all compounds    2E+l        2E-l          7E-l l    -
Bone Surf  Bone Surf (4E+l)      (2E- l)          -    3E-13        SE-7      5E-6 8-81                                              NUREG- 1736
 
APPENDIX B Table 1                            Table 2              Table 3 Occupational Values                    Effluent            Releases to Concentrations            Sewers Col. I      Col. 2    I Col. 3      Col. I        Col.2 Oral                                                                Monthly Ingestion          Inhalation                                        Average Atomic                                                            ALI          ALI      DAC              Air        Water      Concentration No.        Radionuclide                    Class                (&#xb5;Ci)        (&#xb5;Ci)      (&#xb5;Ci/ml)      (&#xb5;Ci/ml)      (&#xb5;Ci/ml)      (&#xb5;Ci/ml)
IOI      Mendelevium-          W, all compounds                7E+3          8E+l          4E-8                        IE-4          IE-3 257                                                              Bone Surf (9B+l)            -          IE-10 IOI      Mendelevium-          W, all compounds                3E+l          2E-l          IE-10 258                                                  Bone Surf    Bone Surf (SE+))        (3E-1)                      SE-13          6E-7          6E-6 Any single radionuclide not listed above with decay mode other than alpha emission or spontaneous fission and with radioactive half-life less than 2 hours                            Submersion"'                    2E+2          IE-7        lE-9                          -
Any single radionuclide not listed above with decay mode other than alpha emission or spontaneous fission and with radioactive half-life greater than 2 hours                                                              2E-l        IE-10        IE-12          IE-8          I E-7 Any single radionuclide not listed above that decays by alpha emission or spontaneous fission, or any mixture for which either the identity or the concentration of any radionuclide in the mixture is not known                                                        4E-4        2E-13        lE-15        2E-9          2E-8 FOOTNOTES:
<I>  "Submersion" means that values given are for submission in a hemispherical semi-infinite cloud of airborne material.
21
'    These radionuclides have radiological half-lives ofless than 2 hours. The total effective dose equivalent received during operations with these radionuclides might include a significant contribution from external exposure. The DAC values for all radionuclides, other than those designated Class "Submersion," are based upon the committed effective dose equivalent due to the intake of the radionuclide into the body and do NOT include potentially significant contributions to dose equivalent from external exposures. The licensee may substitute IE-7&#xb5;Ci/ml for the listed DAC to account for the submersion dose prospectively, but should use individual monitoring devices or other radiation measuring instruments that measure external exposure to demonstrate compliance with the limits. (See section 20.1203.)
13
  ' For soluble mixtures ofU-238, U-234. and U-235 in air, chemical toxicity may be the limiting factor (see section 20.120l(e)). If the percent by weight (enrichment) of U-235 is not greater than 5, the concentration value for a 40-hour workweek is 0.2 milligrams uranium per cubic meter of air average. For any enrichment, the product of the average concentration and time of exposure during a 40-hour workweek shall not exceed 8E-3 (SA)
&#xb5;Ci-hr/ml, where SA is the specific activity of the uranium inhaled. The specific activity for natural uranium is 6. 77E-7 curies per gram U. The specific activity for other mixtures of U-236, U-235, and U-234, if not known, shall be:
NUREG- 1736                                                                B-82
 
APPENDIX B SA  = 3.6E-7 curies/gram U          _ U-depleted SA =      [0.4 + 0.38 (enrichment) + 0.0034 (enrichment)'I] E-6, enrichment ::i: 0. 72 where enrichment is the percentage by weight of U-235, expressed in percent.
NOTE:
: 1. If the identity of each radionuclide in a mixture is known but the concentration of one or more of the radionuclides in the mixture is not known, the DAC for the mixture shall be the most restrictive DAC of any radionuclide in the mixture.
: 2. If the identity of each radionuclide in the mixture is not known, but it is known that certain radionuclides specified in the appendix are not present in the mixture, the inhalation ALI, DAC, and effluent and sewage concentration for the mixture are the lowest values specified in this*
appendix for any radionuclide that is not known to be absent from the mixture; or Table 1                          Table 2                Table3 Occupational Values                    Effluent              Releases to Concentrations              Sewers Col.1        Col.2          Col.3      Col.I        Col.2 Oral Ingestion          Inhalation                                        Monthly Average AU        AL1                DAC        Air        Water      Concentration Radionuclide                              (&#xb5;Ci)      (&#xb5;Ci)            (&#xb5;Ci/ml)  (&#xb5;Ci/ml)      (&#xb5;Ci/ml)        (&#xb5;Ci/ml)
If it is known that Ac-227-D and Cm-250-W are not present                -          7E-4          3E-13        -              -
if, in addition, it is known that Ac-227-W, Y, Th-229-W, Y, Th-230-W, Th-232-W, Y, Pa-231-W, Y, Np-237-W, Pu-239-W, Pu-240-W, Pu-242-W. Am-241-W, Am-242m-W, Am-243-W. Cm-245-W, Cm-246-W, Cm-247-W, Cm-248-W, Bk-247-W. Cf-249-W, and Cf-25 1-W are not present                                                -            7E-3          3E-12        -            -
If, in addition, it is known that Sm-146-W, Sm-147-W, Cd-148-D, W, Gd-152-D, W, Th-228-W, Y, Th-230-Y.
U-232-Y. U-233-Y, U-234-Y, U-235-Y, U-236-Y. U-238-Y, Np-236-W, Pu-236-W, Y, Pu-238-W, Y, Pu-239-Y, Pu-240-Y, Pu-242-Y, Pu-244-W, Y, Cm-243-W, Cm-244-W, Cf-248-W, Cf-249-Y, Cf-250-W, Y, Cf-251-Y.
Cf-252-W, Y, and Cf-254-W, Y are not present                            -            7E-2          3E-l 1      -            -              -
If, in addition, it is known that Pb-210-D, Bi-2l0m-W, Po-210-D, W, Ra-223-W, Ra-225-W, Ra-226-W, Ac-225-D. W, Y, Th-227-W, Y, U-230-D. W, Y, U-232-D, W, Pu-241 -W. Cm-240-W, Cm-242-W, Cf-248-Y, Es-254-W, Fm-257-W, and Md-258-W are not present                                                                  -          7E-1            3E-10        -            -
If, in addition, it is known that Si-32-Y. Ti-44-Y, Fe-60-D, Sr-90-Y. Zr-93-D. Cd-l 13m-D, Cd-113-D, In-I IS-D, W, La-138-D, Lu-176-W, Hf-178m-D, W, Hf-182-D, W, Bi-210m-D, Ra-224-W, Ra-228-W, Ac-226-D. W, Y, Pa-230-W, Y, U-233-D, W; U-234-D, W, U-235-D. W, U-236-D, W, U-238-D, W, Pu-241-Y, Bk-249-W, Cf-253-W, Y, and Es-253-W are not present                                -          7E+O            3E-9 *      -              -
B-83                                                        NUREG-1736
 
APPENDIX B Table 1                          Table 2              TableJ Occupational Values                      Effluent            Releases to Concentrations            Sewers CoLl          Col.2        Col.3        Col.1        Col.2 Oral Ingestion            Inhalation                                        Monthly Average ALl        ALI            DAC            Air        Water      Concentration Radionuclide                              (&#xb5;Ci)        (&#xb5;Ci)        (&#xb5;Ci/ml)      (&#xb5;Ci/ml)    (&#xb5;Ci/ml)      (&#xb5;Ci/ml)
If it is known that Ac-227-D, W, Y, Th-229-W, Y, Th-232-W, Y, Pa-231-W, Y, Cm-248-W, and Cm-250-W are not present                                                                                                lE-14 If, in addition, it is known that Sm-146-W, Gd-148-D, W, Gd-152-D, Th-228-W, Y, Th-'230,W; Y,~ U-232-Y, U-233-Y, U-234-Y, U~~3,~&#xa5;, U-236-Y, U-:~8'iY,;
U-Nat-Y, Np-236-W, Np-237-W, fu~2364W! Y, Pu-238-W, Y, Pu-239-W, Y, J?,q"2Ml-iW; Y, Pu-242-W, Y, Pu-244-W, Y, Arn-241-W, Am~i42nil\V~ Arn-243-W, Cm-243-W, Cm-244-W, Cm-245-W, Cm-246-W, Cm-247-W, Bk-247-W, Cf-249-W, Y, Cf-25o+W; Y, Cf-251-W, Y, Cf-252-W, Y, and Cf-254-W, Y are not present                                                                                                        lE-13 If, in addition, it is known that Sm-147-W, Gd-152-W, Pb-210-D, Bi-2l0m-W, Po-210-D, W, Ra-223-W, Ra-225-W, Ra-226-W, Ac-225-D, W, Y, Th-227-W, Y, U-230-D, W, Y, U-232-D, W, U-Nat-W, Pu-241-W, Cm-240-W, Cm-242-W, Cf-248-W, Y, Es-254-W, Fm-257-W, and Md-258-W are not present                                                                        lE-12 If, in addition, it is known that Fe-60, Sr-90, Cd-l 13m, Cd-113, Iri415,l-129, Cs-134, Sm-145, Sm-147, Gd-148, Gd-152, Hg-194 (organic), Qi-21()m, Ra-223, Ra-224, Ra-225, Ac-225, Th-228, Th-230, U-233, U-234, U-235, U-236, U-238, U-Nat, Cm-242, Cf-248, Es-254, Fm-257, and Md-258 are not present                                                                                                    lE-6          IE-5
: 3. If a mixture of radionuclides consists of uranium and its daughters in ore dust (lO&#xb5;m AMAD particle distribution assumed) prior to chemical separation of the uranium from the ore, the following values may be used for the DAC of the mixture; 6E-l l&#xb5;Ci of gross alpha activity from uranium-238, uranium-234, thorium-230, and radium-226 per milliliter of air; 3E- 11 &#xb5;Ci of natural uranium per milliliter of air; or 45 micrograms of natural uranium per cubic meter of air.
: 4. If the identity and concentration of each radionuclide in a mixture are known, the limiting values should be derived as follows: determine, for each radionuclide in the mixture, the ratio between the concentration present in the mixture and the concentration otherwise established in Appendix B for the specific radionuclide when not in a mixture. The sum of such ratios for all of the radionuclides in the mixture may not exceed "I" (i.e., ''unity").
Example: If radionuclides "A," "B," and "C" are present in concentrations CA, C8 , and Cc, and if the applicable DACs are DAC,, DAC,, and DA Cc, respectively, then the concentrations shall be limited so that the following relationship exists:
                                  ~            +        &#xa3;a              +
DAC,                  DAC,                                          <1 NUREG-1736                                                                  B-84
 
APPENDIX B Discussion:
The data tabulated in this appendix is intended to be used to show compliance with a number of sections of Part 20 that refer to one or more of the tables in the appendix. For example, monitoring of workers for intakes is required when annual intakes may exceed specified fractions of Columns (2) and (3) of Table (1 ). Licensees may show compliance, in part, with public dose limits for doses resulting from effluents by referring to the concentrations listed in Columns (1) and (2) of Table (2). Compliance with sewer release limits is shown, in part, by using the values in Table (3). In addition, certain requirements such as area posting, respiratory protection, and incident reporting use Appendix B values as triggers for these actions. Table (1) is based on occupational dose limits, Table (2) on dose limits to members of the public, and Table (3) is a special case of exposure to members of the public.
The values tabulated in Appendix B are all secondary limits or derived quantities, and each column in the appendix is based on a primary limit, which in this case is a dose. A secondary limit, such as the DAC, is derived from a primary limit. The difference is that the primary limit is absolute in that it is not to be exceeded in any routine situation. The secondary limit is not absolute in the sense that it is applicable only if certain conditions are met. It is not valid as a limit if these conditions are not met. For example, the ALI is a limit if there are no external exposures during the monitoring year and the only source of exposure is internal. If there are external exposures, limiting annual intakes to an ALI will lead to exceeding the primary dose limit, and hence a violation of NRC requirements. The ALI is also not a limit if the ingested or inhaled radioactive material contains more than one type of radionuclide. In such cases, the ALis of each of the components must be adjusted to take account of the presence of the other components.
A derived quantity, such as the DAC, is not a limit at all, and may be exceeded at any time provided certain restrictions apply. The DAC is tabulated for convenience and because it is an easily measured quantity. It is easily calculated from the ALI by assuming a suitable breathing rate and exposure time. Part 20 does not limit airborne concentrations at any given time to values below the DAC, and requirements in Part 20 that are specified in terms of airborne concentrations generally use only time-averaged concentrations and not instantaneous values.
In this appendix, the daughter products of the radionuclides listed were not included in the intake when calculating the tabulated values of ALI and DAC for the parent. However, the effects of the daughters that are produced in the body after intake of the parent are included in the calculations. For example, uranium decays in a long decay chain that includes many radioactive daughter products. When considering the inhalation of uranium, the calculations of ALis and DACs for the tables assume that only the parent uranium isotope is inhaled, and no daughters are considered with the uranium inhalation. The daughters produced in the body after the uranium is taken into the body are included in estimating the dose resulting from the intake. To properly account for the dose from a parent that produces one or more daughters, the parent and each of the daughters must be treated as separately inhaled or ingested radionuclides, and the dose from each added to produce a total dose. The parent and its daughters that may be in the inhaled air or B-85                                      NUREG- 1736
 
APPENDIX B in the ingested material are considered as a mixture of radionuclides and not as members of one chain.
One exception to this rule is the case in which the half-life of the daughter is very short (usually less than about 20 minutes) and much shorter than that of the parent. In this case, the tabulated values assume that the daughter is in secular equilibrium with the parent and that both are always inhaled or ingested together. The daughter in this case need not be considered separately in determining the ALI from the table, because it has been already included in calculating the ALI for the parent. In such situations, you will not fmd data in the table pertaining to short-lived daughter~ of tabulated radionuclides. In such cases, assume that the data for the missing daughter has been included in the data for the parent.
Column (1) in the appendix is the ALI for ingestion by occupationally exposed workers.
Ingestion means taking in the material by mouth via food or drink or as solid or liquid contamination in the workplace. The values are based on an annual dose limit of 5 rem effective dose equivalent (called the stochastic dose limit) or 50 rem organ dose equivalent (called the non-stochastic, or deterministic, dose limit), whichever is more limiting. Internal dose models described in ICRP Publication 30 were used to calculate the effective and organ doses that would result from intake of unit activity of each of the radionuclides listed in the table. The intakes that would lead to an effective dose of 5 rem or an organ dose of 50 rem are then calculated. The highest intake that does not result in exceeding any organ limit or the effective dose equivalent limit is then selected as the ALI and tabulated in Column (1). If the ALI is based on an organ dose, the organ is specified under the tabulated ALI, and the ALI that would result in an effective dose equivalent of 5 rem is listed in parentheses under that organ name. The stochastic ALI is specified in parentheses because it is sometimes needed to show compliance when several radionuclides are present in the ingested material. If the ALI is based on the effective dose equivalent, then only that value is listed, with no other information.
Column (2) in Table (1) is calculated in the same manner as that used for Column (1), except that the intake is by inhalation of airborne material rather than ingestion. The methods of calculation are the same, but the dosimetric models are those for inhalation rather than for ingestion. In addition, inhaled material is classified into one of three classes, called D, W, or Y, depending on how rapidly the material is cleared from the lungs after it is inhaled. Class D is cleared most rapidly, within a matter of days, and Class Y is cleared most slowly, within months or years.
Class W is intermediate. The same radionuclide may exist in one or more classes depending on its chemical and physical characteristics. For example, uranium as a fluoride is a Class D material, but some of its oxide forms are Class W, and other oxide forms are Class Y. Licensees should make a concerted effort to accurately classify the airborne material present at their sites because such classification will determine the ALI and the dose received by a worker following an intake.
The first step in classification is to know the chemical form of the airborne radioactive material.
With that knowledge, the licensee may refer to the listing in Appendix B, which specifies the classification of the most frequently encountered compounds of each radionuclide. If the specific NUREG-1736                                        B-86
 
APPENDIX B compound is not listed, then other references may be used, such as the tabulations by the ICRP.
See Reference { 1} listed below.
Note also that the values of the ALI are based on the assumption that the airborne radioactive material is in the form of particles with an activity median aerodynamic diameter of 1 micrometer, or micron. If the median particle size on site is known and is substantially different from 1 micron, for example 5 microns, then the ALis may be adjusted accordingly, but only after obtaining approval from NRC.' The method of adjustment is described in References
{ 1} and { 2}, listed below. If the particle size is not known, then 1 micron is assumed.
The values in Column (3), Table (1), the derived air concentrations (DACs), are calculated directly from Column (2) by dividing the respective ALis by the breathing rate of a standard person (1.2 m 3/hr) and the number of working hours per year, taken to be 2,000 hr/yr.
Exceptions to this method are those airborne radionuclides that pose an external rather than an internal hazard, and for which ALis are not given, such as the isotopes of xenon and krypton. In such cases, the DACs are calculated directly from the external doses, assuming immersion in a semi-infinite cloud of the gas.
The values in Table (2) differ from those in Table (1) in two major respects: they do not include values that are based on non-stochastic radiation effects, because the public dose limits are so low that such effects are no longer of concern; and they are based on a stochastic dose limit of 100 mrem/yr effective dose equivalent rather than 5 rem/yr. The concentrations in Column (1) of Table (2) were obtained by dividing the stochastic ALis in Table (1) by the breathing rate of 2,400 m 3/yr, then dividing by 3 to take into account the fact that members of the public breathe the air 24 hours per day all year, rather than 8 hr/day during work days, as is assumed for occupational exposure, and also to adjust for differences in inhalation rates between persons at work and members of the public. The result is then divided by 50 to adjust the values from a dose limit of 5,000 mrem/yr to 100 mrem/yr. Because the occupational ALis were calculated for healthy adult workers, but members of the public include groups that may be of varying health conditions as well as children, the results are again divided by a safety factor of 2 to allow for this effect.
In the case of radionuclides that pose an external hazard, the concentrations in Column (I) of Table (2) were obtained by adjusting the occupational DA Cs in Table (1) for the difference in dose limits, that is, by dividing by a factor of 50, and then adjusting for differences in exposure duration from 8 hours per day during work days to 24 hours per day every day.
The concentrations for liquid effluents in Column (2) of Table (2) were obtained by using the most restrictive value in Column (1) of Table (1) and then adjusting it in the same manner as that used to adjust the air values.
The monthly average concentrations in Table (3) were obtained by assuming that a person obtains all of his water from the licensee's sewer outfall, and then calculating the average concentration that would result in an annual ingestion dose of 500 mrem. Averaging the B-87                                  NUREG- 1736
 
APPENDIX B concentrations over a month rather than a year avoids excessive short-term peak discharges by seasonal discharges.
 
==References:==
 
I. International Commission on Radiological Protection, Publication 30 and addenda, Pergamon Press, Fairview Park, Elmsford, NY 10523.
: 2. NRC Regulatory Guide 8.9, Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program.
NUREG-1736                                    B-88
 
SHIELD BUILDING STACK EFFLUENT RELEASE FLOW PATHS UNIT 1 CONTAINMENT                                                                                                      UNIT 2 CONTAINMENT
                                                                              ~
                                                                                                            ~
s FROM 21 AUX BLOG                                                                                              ~              FROM 21 AUX BLOG                                                                            ~
SUPPLY FAN SUPPLY FAN CV
                                                                                                                                          ~                                                                                                          ~
HY_
_cv      HY_
                                                            ~ POST LOCA
                                                                                                                                                                                      ~POSTLOCA 1 - R
* 22 J                                                                                                ~
PURGE MAKELP
                                                                                                                                                                ~-------t,...,.,,t---1--          PURGE MAKEUP r-------,+c'-,-+---1--                  CONTAINMENT ~GE                                                  5000 CFH
                                                                                                                                                                ~------1-r~:t---t--              CONTAINMENT PURGE
                                          ~-----l,-'-''-+---+-            INSERVICE PURGE 3311100 CFM
                                                                                                                                                                      ,------t.-,"-'t----+--      INSERVICE Pl.AGE                              5000 CFM 572'! CFM 572"1 CFH 21!10 CFH 200 CFM
                                                      '----->-----<          R~~~RCf------,
                                                                                                                                                                                  ~--------<        R~~~RCf------,
11 OR 12 SHIELD BLOG VENT 21 OR 22 SHIELD BLOG VENT EXH )-..--......J
                                                                                            >----          FAN                                                                                        A            f-e-e----, EXH)-..--......J FAN 121 SFP SPECIAL AJIIJ INSERVlCE PURGE 122 SFP SPECIAL AND INSERVICE PURGE A            I    ~                                      I                                                                          L_~                              I
                                                                                            ~                                                                                                          A
                                                                                                                                                                                                                    ~
FROMSFP~                                                121 CONTAINMENT PI.RGE r----r-                                                          FROHSFP~
                  ~              Fffl>>,! Ul AIR EJECTOR, 121 AUX  BLDG SPECIAL A
                                                                                            ~
* F...
A I ~ I I
                                                                                                                          ~
                                                                                                                                      ~
I
* 122 AUX A
BLDG SPECIAL I~                                    I
                                                                                                                                                                                                                    ~
FROM U2 AIR EJECTOR, CD            UI l  U2 AUX, BLOG UI I U2 AUX, BLOG cc
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... \blcj97\STACK AEL\STACK REL.DGN 1/12/2007 2:30:58 PM
 
Pub. 73 UNITS AND
                                      };{l?''
                                      ..*., ::-;c_":~ ..-.
                                                  .~-=::* . *:; --
FOXBORO REGISTER.ED TRADEMARK TAB 9, Rev. 1
 
INTERNATIONAL SYSTEM OF UNITS CONVERSION FACTORS (ACCURATE TO+/- 0.1 %)
The Activity of a radionuclide is expressed in becguerels. (Sq), equal to one disintegration per second.
ACTIVITY                                                ACTIVITY 1 terabecquerel          = 1 TBq        = 27 curies            1 kilocurie      =1 kCi          = 37 terabecquerels 1 gigabecuerel          = 1 GBq        = 27 millicuries        1 curie          =1Ci            = 37 gigabecquerels 1 megabecquerel          = 1 MBq        = 27 microcuries        1 millicurie    =1mCi            = 37 megabecquerels 1 kilobecquerel          = 1 kBq        = 27 nanocuries        1 microcurie    =1 &#xb5;Ci          = 37 kilobecquerels 1 becquerel            = 1 Bq          = 27 picocuries        1 nanocurie    =1 nCi          = 37 becquerels The Dose Equivalent is expressed in the sievert (Sv) equal to 1 biologically weighted joule per kg.
DOSE EQUIVALENT                                  DOSE EQUIVALENT 1 sievert                = 1 Sv          = 100 rem              1 kilorem                = 1 krem        = 10 sieverts 1 millisievert          = 1 mSv          = 100 mi Iii rem        1 rem                    = 1 rem          = 10 millisieverts 1 microsievert          = 1 &#xb5;Sv          = 100 microrem          1 millirem              = 1 mrem        = 10 microsieverts 1 nanosievert          = 1 nSv          = 100 nanorem          1 microrem              = 1 &#xb5;rem        = 10 nanosieverts The Absorbed Dose is measured by the 9Ifil'... (Gy), equal to 1 joule of energy absorbed per kilogram.
ABSORBED DOSE                                            ABSORBED DOSE 1 kilogray              = 1 kGy          = 100 krad              1 kilorad                = 1 krad        = 10 grays 1 gray                  = 1 Gy          = 100 rad              1 rad                    = 1 rad          = 10 milligrays 1 milligray            = 1 mGy          = 100 mi Iii rad        1 millirad              = 1 mrad        = 10 micrograys 1 microgray            = 1 &#xb5;Gy          = 100 nanorad          1 mic::irad              = 1 &#xb5;rad        = 10 nanograys The Effective Dose Equivalent for internal or partial body exposure, measured in sieverts. equals the sum of organ dose equivalents times weighting factors, w1: = 0.25 (gonads), 0.15 (breast), 0.12 (marrow), 0.12 (lung) 0.03 (thyroid),
0.03 (bone surf), and 0.06 (each of next 5 remainder organs.
The Committed Dose Equivalent is the dose equivalent, in sieverts, received over 50 years after an internal uptake.
TAB 9, Rev. 1
 
UNITS AND CONVERSION FACTORS These conversion factors are among those found most                International d' Cnites) has been adopted by France as useful to instrument engineers. They are accurate to the              the only legal system of units. and by the National Bureau four or five significant figures given and higher accuracy            of Standards in this country. The kilogram is the unit is almost never warranted in industrial instrumentation.              of mass. The newton is the unit of force, the force re-However, high accuracy is sometimes required. and there                quired to give a mass of one kilogram an acceleration of are times, for example, when the variation in gravity at                one meter per second per second. (In contrast to the different locations and the distinction between weight                  newton is the weight of a kilogram mass under standard and mass should not be ignored. The standard intensity                  gravity, a unit now called the kilopond, and formerly of gravity at mean sea level, latitude 45&deg;, is 980.7 gals (for          referred to as a *'kilogram (weight)." In the cgs*system, Galileo). The actual intensity at the Bureau of Standards              the unit of force is the dyne, so that the standard gravity is 980.0. while it is 978.2 at the Canal Zone, 982.2 in                weight force of a one-gram mass is 980. 7 dynes.) The northern Alaska. about 167.0 on the surface of the moon,                unit of pressure is the pascal, a pressure of one newton and zero in the *'weightless" condition of free fall or in              per square meter, or the bar, equal to 100,000 pascals.
artificial satellite orbit. The range between Alaska and                  Below is a set of prefixes, some of which are in common the Canal Zone, about +/-0.2 percent, is twice the accuracy              use, as in megatons, kilowatts. millivolts and microinches, limit of +/-0.1 percent for the Type CXX Strain-Gauge-                  to indicate decimal multiples and submultiples of units.
Actuated Load Cells. The conversion factor that grams                  These prefixes have been internationally accepted:
x 980.7 = dynes must be used with care, as it applies only under standard conditions, such as exist near Portland,                      10 12 tera          IO    deka        10-*    micro Oregon. At other locations the correct conversion factor                          I 0"  giga          10- 1  deci        10-*    nano is equal to the local intensity of gravity.                                      10"  mega          10- 2  centi        10- 12  pico The International System of Cnits (designated MKSA                          10 3  kilo          I 0- 3 mi Iii      I 0- 10 femto for meter, kilogram, second, and ampere, or SI, for System                        10 2  hecto                              10-1*  atto ATMOSPHEREs-acm (Standard sea-level pressure)                          CENTIMETERS ---cm x 1.01325              Bars                                            X 0.393';"              = Inches x 76.0                  Centimeters of mercury, at O C x 29.92                Inches of mercury, at 0 C                  CENTIMETERS      OF  '.\IERCURY, at O C x 33.96                Feet of water, at 68 F                        x 0.013332                  Bars x 1.0332                Kilograms (kiloponds) per square              x 0.4468                    Feec of water, at 1,8 F centimeter                                  x 5.362                    Inches of water. at 68 F X 101325                Pascals                                        x 0.013595                  Kilograms (kiloponds) per square X 14.696                Pounds per square inch                                                        centimeter X 1.058]                Tons per square foot                          X  27.85                    Pounds per square foot X 760                  Torr                                          X  0.1933-:-                Pounds per square inch X  0.013158                Standard atmospheres BARRELS, LIQUID-bbl                                                        X  10                      Torr x 31.5              = Gallons (liquid)
CENTIMETERS PER SECOND-cm per sec BARRELS, O1L-bbl                                                            x 1.9685                = Feet per minute X 42                = Gallons (oil)                                    x 0.03281                = Feet per second x 0.03600                = Kilometers per hour BARS                                                                        x 0.6000                = Meters per minute X 33.52                Feet of water, at 68 F                        x 0.02237                = Miles per hour X 29.53                Inches of mercury, at O C                      x 3.728 x 10-*          = Miles per minute X 1.0191                Kilograms (kiloponds) per square centimeter                                CUBIC CENTIMETERs-cu cm or cm 3 X 100000              Pascals                                        x 3.5315 x 10-*          = Cubic feet X 14.50.\              Pounds per square inch                        x 6.1024 x 10- 2        = Cubic inches X 0.98692              Standard atmospheres                          x 1.3080 x 10-*          = Cubic yards X 1.0443            =  Tons per square foot                            x 2.642 x 10-*          = Gallons X 750.06            =  Torr                                          x 2.200 x 10-*            = Gallons (Imperial) x 1.0000 x 10-3          = Liters BRITISH THERMAL UNITS-Btu or B x 778                = Foot-pounds                                  CUBIC FEET-CU ft x 0.252              = Kilogram-calories                              X 2.832 X 10 4            = Cubic centimeters x 107.6              = Kilogram-meters                                X 1728                    = Cubic inches x 3.929 x 10-*      = Horsepower-hours                                X 0.02832                = Cubic meters x 2.930 x 10-*      = Kilowatt-hours                                  X 0.03704                = Cubic yards X 7.481                  = Gallons BRITISH THERMAL UNITS PER MINUTE-Btu per min                              X 6.229                  = Gallons  (Imperial) x 12.96              = Foot-pounds per second                          X 28.32                  = Liters x 0.0235';"          = Horsepower x 0.01758            = Kilowatts                                    CuBic FEET PER M1NUTE-cfm x 17.58              = Watts                                          x 472.0                  = Cubic centimeters per second x 0.1247                  = Gallons per second CENTARES                                                                  x 0.4720                  = Liters per second X 1                  = Square meters                                  x 62.32                  = Pounds of water per minute, at 68 F Revised May 5, 1965 TAB 9, Rev. 1                    .FOXBORO
 
CUBIC FEET PER SECOND-cfs                                  GALLONS-gal x 0.6463          = Million gallons per day            X 8.322            = Pounds of water. at 68 F in air x 448.8          = Gallons per minute                  X 8,330            = Pounds of water, at 68 F in \'acuo X 3785            = Cubic centimeters J
CUBIC INCHES-CU in X 16.387 X 5.787 X 10- 4 X 1.6387 X 10-5 X 2.143 X lQ-5
                            = Cubic centimeters
                            = Cubic feet
                            = Cubic meters
                            = Cubic yards X 0.13368 X 231 X 3.785 X lQ- 3 X 4.951 X 10- 3 X 0.832 7
                                                                                      = Cubic feet
                                                                                      = Cubic inches
                                                                                      = Cubic meters
                                                                                      = Cubic yards
                                                                                      = Gallons (Imperial)
X 4.329 X 10-3    = Gallons                              X 8                = Pints (liquid)
X 3.605 X 10-3    = Gallons (Imperial)                  X 4                = Quarts  (liquid)
X 1.6387 X 10-2  = Liters                              X 3.785            = Liters CUBIC METERs-cu m or m 3                                GALLONS, IMPERIAL-gal x 35.31          = Cubic feet                          x 10.000          = Pounds of water. at 62 F in air x 61.024          = Cubic inches                        x 4546              = Cubic centimctc-rs x 1.308          = Cubic yards                        x 0.16054          = Cubic feet x 264.2          = Gallons                            x 4.546 x 10- 3    = Cubic meters x 220.0          = Gallons (Imperial)                  x 5.946 x 10- 3    = Cubic yards x 1000.0          = Liters                              x 1.20094          = Gallons CUBIC YARDS-CU yd x 4.546            = Liters X 7.646 X 10 5    = Cubic centimeters                GALLONS PER MrNUTE-gpm X 27              = Cubic feet                          x 2.228 x 10- 3    = Cubic feet per second X 46,656          = Cubic inches                        x 8.021            = Cubic feet per hour X 0.7646          = Cubic meters                        x 0.2271 S        = Cubic meters per hour X 201.97          = Gallons                            x 0.06308          = Liters per second X 168.17          = Gallons (Imperial)
X 764.6            = Liters                          GRAINS, Avoirdupois or Troy Grains avoirdupois = Grains troy DEGREES, Angular---deg or 0                                x 0.0648          = Grams x 60              = Minutes x 0.017-153        = Radians                          GRAINS PER GALL.ON. at 68 T" x 3600            = Seconds                            x 17. 17          = Parts per million by weight in water x 142. 9          = Pounds per million _!:(a lions DEGREES PER SECOND, ANGULAR-deg per sec x 0.017453        = Radians per second                GRAINS PER IMPERIAi. GALLON at 62 F x 0.16667          = Revolutions per minute              x 14.29            = Parts per million by wei!:(ht in "*acer x 2.7778 x 10- 3  = Revolutions per second GRAMS--g DRAMS-dr                                                    X 980.7            =  Dvnes X  27.344        =  Grains                              X 15.432          =  G~ains X  0.0625        =  Ounces                              X 0.03527-1        =  Ounces
  )      X  1.7718        =  Grams                              X 0.032151 X 2.2046 X 10- 3
                                                                                      =
                                                                                      =
Ounces (troy)
Pounds FATHOMS x6                = Feet                              GRAMS PER CENTIMETER-g per cm X 1.8288          = l\,feters                            x 5.600 x 10- 3    = Pounds per inch FEET-ft                                                  GRAMS PER CURIC CENTIMETER-g per cu cm X 30.48il        = Centimeters                          x 62.43            = Pounds per cubic foot X 12              = Inches                              x 0.03613          = Pounds per cubic inch X 0.3048          = 1\-leters X 0.3333          = Yards                            GRAMS PER LITER --g per 1 x 58.42            = Grains per gallon FEET OF \YATER, at 68 F                                    x 8.345            = Pounds per 1000 gallons x 0.02984            Bars                              x 0.062-11          = Pounds per cubic foot x 0.8811              Inches of mercury, at 0 C          x 1003              = Parts per million by weight in water.
x 0.030-12            Kilograms (kiloponds) per square                            at 68 F          .
centimeter X 62.32              Pounds per square foot          HECTARES-ha X 0.4328              Pounds per square inch            X 1.076-J X 10'    = Square feet X 0.02945            Standard atmospheres HORSEPOWER-hp FEET PER MINUTE -fpm                                        X 33,000            = Foot-pounds per minutc X 0.5080          = Centimeters per second              X 550              = Foot-pounds per second X 0.016667        = Feet per second                    X 42.42            = British Thermal Units per minute X 0.01829          = Kilometers per hour                X 10.69            = Kilogram-calories per minute X 0.3048          = Meters per minute                  X 1.0139            = Horsepower (metric)
X 0.01136          = Miles per hour                      X 0.7457            = Kilowatts X 745.7            = Watts FEET PER SECOND PER SECOND-fpsps x 30.48            = Centimeters per second per second HORSEPOWER, Boiler-hp (gals)                          x 33,479            = British Thermal Units per hour x 0.3048          = Meters per second per second        x 9.803            = Kilowatts FooT-PouNDs-ft-lb                                        HoRSEPOWER-HouRs-hp-hr X 1.285 X l0- 3    = British Thermal  Units              x 1.98 x 10'        = Foot-pounds X 3.239 X lQ-<    = Kilogram-calories                  x 2545              = British Thermal Units X 0.13825          = Kilogram-meters                    x 641.3            = Kilogram-calori<.>s X 5.050 X 10- 7    = Horsepower-hours                    x 2.737 x 10 5      = Kilogram-meters X 3.766 X 10-1    = Kilowatt-hours                      x 0.7457            = Kilowatt-hours TAB 9, Rev_ 1
 
INCHES-in.                                                  METERs-m X 2.540            = Centimeters                            X 3.281          = Feet X 39.37          = Inches INCHES OF l\,fERCURY, at O C                                  X 1.0936          = Yards x 0.03386          = Bars x 1.135            = Feet of water, at 68 F            METERS PER M1NUTE-m per min x 13.62                Inches of water, at 68 F            x 1.6667          = Centimeters per second x 0.03453              Kilograms (kiloponds) per square    x 3.281          = Feet per minute centimeter                        x 0.05468        = Feet per second X 70.73                Pounds per square foot              x 0.0600          = Kilometers per hour X 0.4912              Pounds per square inch              x 0.03728        = Miles per hour X 0.03342              Standard atmospheres INCHES OF WATER, at 68 F ,                                  METERS PER SECOND-m per sec x 2.486 x 10- 3    = Bars                                  x 196.8          = Feet per minute x 0.07342            Inches of mercury, at 0 C            x 3.281          = Feet per second x 2.535 x 10-  3 Kilograms (kiloponds) per square      x 3.600          = Kilometers per hour centimeter                        x 0.0600          = Kilometers per minute X 0.5770              Ounces per square inch                x 2.237          = Miles per hour X 5.193              Pounds per square foot                x 0.03728        = Miles per minute X 0.03606              Pounds per square inch X 2.454 X 10-3        Standard atmospheres              MICRONS-mu or &#xb5;,
X 10- 6          = Meters KILOGRAMS-kg X 2.2046            = Pounds                            MILES X 1.102 X 10- 3    = Tons (short)                          X 1.6094 X 10 6  = Centimeters X 5280            = Feet KILOGRAMS PER METER-kg per m                                    X 1.6094          = Kilometers x 0.6720            = Pounds per foot                      X 1760            = Yards KILOGRAMS (KrLOPONDS) PER SQUARE CENTIMETER-kg (kp)          MILES PER HouR-mph per sq. cm.                                              x 44. 70          = Centimeters per second X 0.9807              Bars X 32.87 x 88              = Feet per minute Feet of water, at 68 F              x 1.4667          = Feet per second X 28.96                Inches of mercury, at O C X 2048 x 1.6094          = Kilometers per hour Pounds per square foot              x 0.8684          = Knots X 14.223              Pounds per square inch              x 26.82            = Meters per minute X 0.9678              Standard atmospheres KILOGRAMS PER SQUARE MILLIMETER-kg per sq mm                MILES PER MINUTE x 10'              = Kilograms per square meter            X 2682            = Centimeters per second X 88              = Feet per second KILOMETERS PER HouR-kmph                                      X 1.6094          = Kilometers per minute              '  l x 27. 78            = Centimeters per second                X 60              = Miles per hour                      ,J-x 0.9113            = Feet per second x 54.68            = Feet per minute                    MILLIERS (Metric Ton, Tonne) x 16.667            = Meters per minute                    x 10 3            = Kilograms x 0.5396            = Knots x 0.6214            = Miles per hour                    MINUTES, Angular X 2.909 X 10-*    = Radians KILOMETERS PER HouR PER SECOND-kmphps x 27.78              = Centimeters per second per second  OUNCES--OZ x 0.9113            = Feet per second per second            X 16              = Drams x 0.2778            = Meters per second per second        X 437.5            = Grains KILOWATTS-kw                                                    X 0.06250        =  Pounds X 4.425 X 10 4      = Foot-pounds per minute                X 28.35          =  Grams X 737.6              = Foot-pounds per second              X 0.9115          =  Ounces (troy)
X 56.9              = British Thermal Units per minute    X 2.790 X 10- 5    = Tons (long)
X 14.33              = Kilogram-calories per minute        X 2.835 X 10- 5    = Tons (metric)
X 1.3410            = Horsepower OUNCES, Troy--oz KILOWATTHOURS-kwhr                                            X 480              =  Grains x 2.655 x 10 6      = Foot-pounds                          X 20              =  Pennyweights (troy) x 3413              = British Thermal Units                X 0.08333          = Pounds (troy) x 860                = Kilogram-calories                    X 31.103          = Grams x 3.671 x 10 5      = Kilogram-meters                      X 1.0971          = Ounces (avoirdupois) x 1.3410            = Horsepower-hours LITERS-I                                                    OUNCES, Fluid--oz X 1000.0            = Cubic centimeters                    X 1.8046          = Cubic inches X 0.035315          = Cubic feet                          X 0.02957          = Liters X 61.024            = Cubic inches X 1.308 X 10- 3    = Cubic yards                        OuNcEs PER SQUARE INcH--oz per sq in.
X 0.2642            = Gallons                              x 0.06250          = Pounds per square inch X 0.2200            = Gallons (Imperial)                    x 4.395            = Grams per square centimeter LITERS PER    MINUTE-I per min                              PARTS PER MILLION-ppm by weight in water x 5.885 x  to-*    = Cubic feet per second                x 0.0582          = Grains per gallon, at 68 F x 4.403 x  to- 3  = Gallons per second                    x 0.0700          = Grains per Imperial gallon, at 62 F x 3.666 x  10- 3  = Gallons per second (Imperial)        x 8.322            = Pounds per million gallons, at 68 F TAB 9, Rev. 1
 
PENNYWEIGHTS, Troy-dwt                                        QUARTS, Liquid-qt x 24                = Grains                                  x 57.75              = Cubic inches x 1.5552          = Grams
  )  POISE QUINTAL (Metric)
X 100                =  Kilograms X 100              = Centipoise                                X 220.46            =  Pounds X 2.0886 X  10- 3  = Pound (weight) second per square foot    X 101.28                Pounds (Argentina)
X .06721          = Pound (mass) per foot-second              X 129.54                Pounds (Brazil)
X 101.41                Pounds (Chile)
PouNDs-Jb                                                          X 101.47                Pounds (Mexico)
X 16              = Ounces                                    X 101.43                Pounds (Peru)
X 256              =*Drams X 7000            = Grains                                STOKES X 5 X 10- 4        = Tons (short)                              X 1.076 X  10- 3    = Square feet per second X 453.6            = Grams X 1.2153            = Pounds (troy)                        TEMPERATURE-temp Degrees Celsius-C PouNDS, Troy-lb                                                      C  +  273.15      = K Kelvin X 5760 X 240
                            =
                            =
Grains Pennyweights (troy)
(C x 9 /5) +  32 = F Fahrenheit C x 4/5          = R Reaumur X 12                =  Ounces (troy)
X 373.2            = Grams                                    Degrees Fahrenheit-F X 0.8229            = Pounds (avoirdupois)                        F  +  459.67      = Rankine X 13.166            = Ounces (avoirdupois)                        (F - 32) x 5 /9  = C Celsius
)        X 3.6735 X 10- 4 X 4.1143 X 10-<
                            =
                            =
Tons (long)
Tons (short)
(F - 32) x 4 /9 Degrees Reaumur-R
                                                                                            = R Reaumur X 3.7324 X 10-<    =  Tons (metric)                              Rx 5/4            = C Celsius (Rx 9 /4) +32    = F Fahrenheit POUNDS OF WATER, at 68 F x 0.01604          = Cubic feet                            TONS, Long x 27.72            = Cubic inches                              X 1016              = Kilograms x 0.1200            = Gallons                                  X 2240              = Pounds X 1.1200            = Tons (short)
POUNDS OF WATER PER MINUTE, at 68 F x 2.673 x 10-*      = Cubic feet per second                  ToNs, Metric (Tonne, Millier) x 1000              = Kilograms PouNDs PER CUBIC FooT-lb per cu ft                                x 2204.6            = Pounds x 0.016018          = Grams per cubic centimeter x 16.018            = Kilograms per cubic meter              ToNs, Short x 5.787 x 10-*      = Pounds per cubic inch                    X 2000              =  Pounds
  )  PouNDS PER CUBIC INcH-lb per cu in.
X 32000 X 907.2
                                                                                            =
                                                                                            =
Ounces Kilograms x 27.68            = Grams per cubic centimeter                X 2430.6            =  Pounds (troy) x 2.768 x 10 4      = Kilograms per cubic meter                X 0.8929            =  Tons (long) x 1728              = Pounds per cubic foot                    X 0.9072            =  Tons (metric)
PouNDS PER FooT-lb per ft                                      ToNs OF WATER PER 24 HouRs. at 68 F x 1.488            = Kilograms per meter                      x 83.33              = Pounds water per hour x 14.88            = Grams per centimeter                      x 0.1667            = Gallons per minute x 1.337              = Cubic feet per hour PouNDs PER SQUARE FooT-psf x 0.01605          = Feet of water, at 68 F                WATTS-W x 4.882 x 10- 4    = Kilograms per square centimeter          X 0.0569            = British Thermal Units per minute x 6.944 x 10- 3    = Pounds per square inch                  X 44.25              = Foot-pounds per minute X 0.7376            = Foot-pounds per second PouNDs PER SQUARE INcH-psi                                        X 1.341 X ]0- 3      = Horsepower x 0.06805          = Atmospheres                              X 0.01433            = Kilogram-calories  per minute x 2.311            = Feet of water, at 68 F x 27.73            = Inches of water, at 68 F              w ATTHOURS-whr x 2.036            = Inches of mercury, at O C                X  3.413              = British Thermal  Units x 0.07031          = Kilograms per square centimeter          X  2655              = Foot-pounds X  1.341 X  10- 3    = Horsepower-hours QUARTS,    Dry-qt                                                X  0.860              = Kilogram-calories X  67.20            = Cubic inches                            X  367.1              = Kilogram-meters
  ./
TAB 9, Rev. 1
 
TEMPERATURE CONVERSION TABLES Fahrenheit and Celsius (Centigrade)
C
* F        C
* F      C
* F      C
* F    C
* F
      -273.15 -459.67          -17.2      I    33.8  10.6    51  123.8    43    110    230  266    510    950
    -268      -450            -16.7      2    35.6  I I.I  52  125.6    49    120. 248  271    520    968
    -262      -440            -16.1    3    37.4  11.7    53  127.4    54    130    266  277    530    986
    -257      -430            -15.6    4    39.2  12.2    54  129.2    60    140    284  282    540  1004
    -251      -420            -15.0    5    41.0  12.8    55  131.0    66    150    302  288    550  1022
    -246      -410            -14.4    6    42.8    13.3    56  132.8    71      160    320  293    560  1040
    -240      -400            -13.9    7    44.6    13.9    57  134.6    77      170    338  299    570  1058
    -234      -390            -13.3    8    46.4    14.4    58  136.4    82      180    356  304    580  1076
    -229      -380            -12.8    9    48.2    15.0    59  138.2    88      190  374    310    590  1094
    -223      -370            -12.2    10    50.0    15.6    60  140.0    93      200  392    316    600  1112
    -218      -360            -11.7    11    51.8    16.1    61  141.8    99      210  410  321    610  1130
    -212      -350            -11.1    12    53.6    16.7    62  143.6                          327    620  1148
    -207      -340            -10.6    13    55.4    17.2    63  145.4                        332    630  1166
    -201      -330            -10.0    14    57.2    17.8    64  147.2                        338    640  1184
    -196      -320              -9.4    15    59.0    18.3    65  149.0                          343    650  1202
    -190      -310              -8.9    16    60.8    18.9    66  150.8    100      212  413  349    660  1220
    -184      -300              -8.3    17    62.6    19.4    67  152.6                        354    670  1238
    -179      -290              -7.8    18    64.4  20.0    68  154.4                        360    680  1256
    -173      -280              -7.2    19    66.2  20.6    69  156.2                        366    690  1274
    -169      -273    -459.4    -6.7  20    68.0  21.1    70  158.0                        371    700  1292
    -168      -270    -454      -6.1  21    69.8  21.7    71  159.8                        377    710  1310
    -162      -260    -436      -5.6  22    71.6  22.2    72  161.6    104    220    428  382    720  1328
    -157      -250    -418      -5.0  23    73.4  22.8    73  163.4    110    230    446  388    730  1346
    -151      -240    -400      -4.4  24    75.2  23.3    74  165.2    116    240    464  393    740  1364
    -146      -230    -382      -3.9  25    77.0    23.9    75  167.0    121    250    482  399    750  1382
    -140      -220    -364      -3.3  26    78.8    24.4    76  168.8    127    260    500  404    760    1400
    -134      -210    -346      -2.8  27    80.6  25.0    77  170.6    132    270    518  410    770    1418
    -129      -200    -328      -2.2  28    82.4  25.6    78  172.4    138    280    536  416    780    1436
    -123      -190    -310      -1.7  29    84.2    26.1    79  174.2    143    290    554  421    790  1454
    -118      -180    -292      -1.1  30    86.0    26.7    80  176.0  149      300    572  427    800  1472    ",
                                                                                                                    *. .)
    -112      -170    -274      -0.6  JI    87.8    27.2    81  177.8  154      310    590  432    810  1490
    -107      -*160  -256        0    32    89.6    27.8    82  179.6  160      320    608  438    820  1508
    -101      -150    -238        0.6  33    91.4    28.3    83  181.4  166      330    626  443    830  1526
      -95.6  -140    -220        1.1  34    93.2    28.9    84  183.2  171      340    644  449    840  1544
      -90.0  -130    -202        1.7  35    95.0    29.4    85  185.0    177      350    662  454    850  1562
      -84.4  -120    -184        2.2  36    96.8  30.0    86  186.8    182      360  680    46J    860  1580
      -78.9  -110    -166        2.8  37    98.6  30.6    87  188.6    188      370  698  466    870  1598
      -73.3  -100    -148        3.3  38  100.4  31.1    88  190.4    193      380  716  471    880  1616
    -67.8      -90  -130        3.9  39  102.2  31.7    89  192.2    199      390  734  477    890  1634
    -62.2      -80  -112        4.4 40    104.0  32.2    90  194.0    204      400  752  482    900  1652 I    -56.7
    -51.1
                -70
                -60
                        -94
                        -76 5.0 5.6 41 42 105.8 107.6 32.8 33.3 91 92 195.8 197.6 210 216 410 420 770 788 488 493 910
                                                                                                      . 920 1670 1689
    -45.6      -50    -58        6.1 43    109.4  33.9    93  199.4    221      430  806  499    930  17()!,
    -40.0      -40    -40        6.7 44    111.2  34.4    94  201.2    227      440  824  504    940  1724
    -34.4      -30    -22        7.2 45    113.0  35.0    95  203.0    232      450  842  510    950  1742
    -28.9    -20      -4        7.8 46    114.8  35.6    96  204.8  238      460  86:)  516    960  1760
    -23.3    -10        14      8.3 47    116.6  36.1    97  206.6  243      470    878  521    970  1778
    -17.8        0      32      8.9 48    118.4  36.7    98  208.4  249      480    896  527    980  1796 9.4 49    120.2  37.2    99  210.2  254      490    914  532    990  1814 10.0 50    122.0  37.8    100  212.0  260      S00    932  538  1000    1832 INTERPOLATION VALUES C
* F    C
* F
    *In the center (colored) column, find the temper-ature to be converted. The equivalent temperature                        0.56        1  1.8  3.33      6  10.8 is in the left column, if converting to Celsius, and m                    1.11      2    3.6  3.89      7  12.6 the right column, if converting to Fahrenheit.                            1.67      3    5.4  4.44      8  14.4 2.22      4    7.2  5.00      9  16.2 2.78      s    9.0  5.56    10    18.0 TAB 9, Rev. 1
 
TEMPERATURE CONVERSION TABLES Fahrenheit and Celsius (Centigrade) (continued)
C        "'        F        C
* F        C        .      F        C
* F 543    1010      1850      821    1510      2750    1099    2010  3650      1377      2510    4550 549    1020      1868      827    1520      2768    1104    2020  3668      1382      2520    4568 554    1030      1886      832    1530      2786    1110    2030  3686      1388      2530    4586 560    1040      1904      838    1540      2804    1116    2040  3704      1393      2540    4604 566    1050      1922      843    1550      2822    1121    2050  3722      1399      2550    4622 571    1060      1940      849    1560      2840    1127    2060  3740      1404      2560    4640 577    1070      1958      854    1570      2858    1132    2070  3758      1410      2570    4658 582    1080      1976      860    1580      2876    1138    2080  3776      1416      2580    4676 588    1090      1994      866    1590      2894    1143    2090  3794      1421      2590    4694 593    1100      2012      871    1600      2912    1149    2100  3812      1427      2600    4712 599    1110      2030      877    1610      2930    1154    2110  3830      1432      2610    4730 604    1120      2048      882    1620      2948    1160    2120  3848      1438      2620    4748 610    1130      2066      888    1630      2966    1166    2130  3866      1443      2630    4766 616    1140      2084      893    1640      2984      1171    2140  3884      1449      2640    4784 621    1150      2102      899    1650      3002    1177      2150  3902      1454      2650    4802 627    1160      2120      904    1660      3020      1182    2160  3920      1460      2660    4820 632    1170      2138      910    1670      3038      1188    2170  3938      1466      2670    4838 638    1180      2156      916    1680      3056      1193    2180  3956      1471      2680    4856 643    1190      2174      921    1690      3074      1199    2190  3974      1477      2690    4874 649    1200      2192    927    1700      3092      1204    2200  3992      1482      2700    4892 654    1210      2210      932    1710      3110      1210    2210  4010      1488      2710    4910 660    1220    2228      938    1720      3128      1216    2220  4028      1493      2720    4928 666    1230      2246      943    1730      3146    1221      2230  4046      1499      2730    4946 671    1240      2264      949    1740      3164    1227      2240  4064      1504      2740    4964 677    1250      2282      954    1750      3182    1232      2250  4082        1510      2750    4982
.      682 688 693 1260 1270 1280 2300 2318 2336 960 966 971 1760 1770 1780 3200 3218 3236 1238 1243 1249 2260 2270 2280 4100 4118 4136 1516 1521 1527 2760 2770 2780 5000 5018 5036 699    1290      2354      977    1790      3254    1254    2290  4154      1532      2790    5054
  )    704    1300      2372      982    1800      3272    1260    2300  4172      1538      2800    5072 710    1310      2390      988    1810      3290    1266    2310  4190      1543      2810    5090 716    1320      2408      993    1820      3308    1271    2320  4208      1549      2820    5108 721    1330      2426      999    1830      3326    1277    2330  4226      1554      2830    5126 727    1340      2444    1004    1840      3344    1282    2340  4244      1560      2840    5144 732    1350      2462    1010    1850      3362    1288    2350  4262      1566      2850    5162 738    1360      2480    1016    1860      3380    1293    2360  4280      1571      2860    5180 743    1370      2498    1021    1870      3398    1299    2370  4298      1577      2870    5198 749    1380      2516    1027    1880      3416    1304    2380  4316      1582      2880    5216 754    1390      2534    1032    1890      3434    1310    2390  4334      1588      2890    5234 760    1400      2552    1038    1900      3452    1316    2400  4352      1593      2900    5252 766    1410      2570    1043    1910      3470    1321    2410  4370      1599      2910    5270 771    1420      2588    1049    1920      3488    1327    2420  4388      1604      2920    5288 777    1430      2606    1054    1930      3506    1332    2430  4406      1610      2930    5306 782    1440      2624    1060    1940      3524    1338    2440  4424      1616      2940    5324 788    1450      2642    1066    1950      3542    1343    2450  4442      1621      2950    5342 793    1460      2660    1071    1960      3560    1349    2460  4460      1627      2960    5360 799    1470      2678    1077    1970      3578      1354    2470  4478      1632      2970    5378 804    1480      2696    1082    1980      3596      1360    2480  4496      1638      2980    5396 810    1490      2714    1088    1990      3614      1366    2490  4514      1643      2990    5414 816    1500      2732    1093    2000      3632      1371    2500  4532      1649      3000    5432 Temperature Conversion Formulas Degrees Celsius (formerly Centigrade) C                      Degrees Fahrenheit-F                  Degrees Reaumur-R C + 273.15      = K Kelvin            F  + 459.67    = Rankine                Rx 5/4          = C Celsius (C x 9 /5) + 32 = F Fahrenheit        rr, - 32) x 5 /9 = C Celsius            (Rx 9 /4)  + 32  = F Fahrenheit C x 4/5          = R Reaumur          (F - 32) x 4 /9 = R Reaumur TAB 9, Rev. 1
 
PROPERTIES OF SATURATED STEAM DRY SATURATED STEAM: PRESSURE TABLE*
Specific Volume                Enthalpy                      Entropy                    Internal Enerrr Abs Press.,
Abs Press.,                                                                                                              Sal.
lb          Temp        Sat.              Sat        Sat              Sal. Sal.                    Sal.                      Sal.      lb Sq In.          F        U1uld              Vapor      Liquid  Evap    Vapor    Liquid      Evap        Vapor      li1uid          Vapor    Sq In.
p            I          ,,                '1          h      h11      hi      s,        511          Sg          u,            ul        p 333.6          69.70  1036.3  1106.0  0.1326    1.8456      1.9782      69.70        1044.3    1.0 1.0        101.74    0.01614 173.73        93.99  1022.2  1116.2  0.1749    1.7451      1.9200      93.98        1051.9    2.0 2.0        126.08    0.01623 0,01630          118.71      109.37  1013.2  1122.6  0.2008    1.6855      1.8863    109.36          1056.7    3.0 3.0        141.48 0.01636            90.63      120.86  1006,4  1127.3  0.2198    1.6427      1.8625    120.85          1060.2    4.0 4.0        152.97                                                                                                                            5,0 0.01640            73.52      130.13  1001.0  1131.1  0.2347    1.6094      1.8441    130.12          1063.1 5.0        162.24 0.01645            61.98      137.96    996.2  1134.2  0.2472    1.5820      1.8292    137.94          1065.4    6.0 6.0        170.06 0.01649            53.64    _144.76    992.1  1136.9  0.2581    1.5586      1.8167    144.74          1067.4    7.0 7.0        176.85                                                                                                                              8,0 0.01653            47.34      150.79    988.5  I I 39.3 0.2674    1.5383      1.8057    150.77          1069.2 8.0        I 82.86 0.01656            42.40      156.22    985.2  1 I 41.4 0.2759    1.5203      1.7962    156.19          1070.B    9.0 9.0        188.28 0.01659            38.42      I 61.17  982.J  1143.3  0.2835    1.5041      1.7876    161.14          1072.2      JO 10        193.21 26.80      I 80.07  970.3  I 150,4  0.3120    1.4446      1.7566    180.02          1077.5  14.696 14.696        212.00    0.01672 26.29      181.11    969.7  1150.B  0.3135    1.4415      1.7549    181.06          1077.B      15 15        213.03    0.01672 20.089    196.16    960.1  1156.3  0.3356    1.3962      1.7319    196.10          1081.9      20 20        227.96    0.01683 16.303    208.42    952.1  1160.6  0.3533    1.3606      1.7139    208.34          1085.1      25 25        240.07    0.01692 0.01701            13.746    218.82    945.3  1164.1  0.3680    1.331 3      1.6993    218.73          1087.B      30 30        250.33 11.898    227.91    939.2  1167.1  0.3807    1.3063      1.6870    227.80          1090.I      35 35        259.28    0.01708 0.01715              10.498    236.03    933.7  1169.7  0.3919    1.2844      1.6763    235.90          1092.0      40 40        267.25 9.401    243.36    928.6  1172.0  0,4019    1.2650      1.6669    243.22          1093.7      45 45        274,44    0.01721 so        281.01    0.01727              8.515    250.09    924.0  1174.I  0.4110    1.2474      1.6585    249.93          1095.3      so 7.787    256.30    919.6  1175.9  0.4193    1.2316      1.6509    256.12          1096.7      55 55        287.07    0.01732 0.01738              7.175    262.09    915.5  1177.6  0.4270    1.2168      1.6438    261.90          1097.9      60 60        292.71 0.01743              6,655    267.50    911.6  1179.I  0.4342    1.2032      1.6374    267.29          1099.1      65 65        297.97 0.01748              6.206    272.61    907.9  1180.6  0.4409    1.1906      1.631 5    272,38          1100.2      70 70        302.92 5,816    277.43    904.S  11 81.9  0.4472    1.1787      1.6259    277.19          1101.2      75 75        307.60    0.01753 0.01757              5.472    282.02    901.1  1183.1  0.4531    1.1676      1.6207    281.76          I 102.1      BO 80        312.03 BS        316.25    0.01761        II 5.168    286.39 290.56 897.B 894.7 1184.2 1185.3 0,4587 0.4641 I. I 571 1.1471 1.6 I 58 1.6112 286.11 290.27 1102.9 1103.7 BS 90 90        320.27    0.01766              4.896 95        324.12    0.01770        I    4,652    294.56    891.7  1186.2  0.4692    1.1376      1.6068    294.25          1104.5      95 I                        888.B  1187.2  0.4740    I. I 2 86    1.6026    298.08          1105.2    100 100          327.81    0.01774              4.432    298.40 0.01782              4.049    305.66    883.2  1188.9  0.4832    1.1117      1.5948    305.30          1106,5    110 110          334.77 0.01789              3.728    312.44    877.9  1190.4  0.4916    1.0962      1.5878    312.05          1107.6    120 120          341.25 3.455    318.81    872.9  1191.7  0.4995    1.0817      1.5812    318.38          1108.6    130 130          347,32    0.01796 3.220    324.82    868.2  1193.0  0.5069    1.0682      1.5751    324.35          I 109.6  140 140          353.02    0.01802 3.015    330.51    863,6  1194.1  0.5138    1.0556      1.5694    330.01          1110.5    150 I 50        358.42    0.01809 2.834    335.93    859.2  1195.1  0.5204    1.0436      1.5640    335.39          1111.2    160 160          363.53    0.01815 2.675    341.09    854.9  1196,0  0.5266    1.0324      1.5590    340.52          1111.9    170 170          368.41    0.01822 2.532    346.03    850.B  1196.9  0,5325    1.0217      1.5542    345.42          1112.5    1 BO 1 BO        373.06    0.01827 0.01833              2.404    350.79    846.B  1197.6  0,5381    1.0116      1.5497    350.15          1113.I    190 190          377.51 2.288    355.36    843.0  1198.4  0,5435    1.0018      1.5453    354.68          1113.7    200 200          381.79    0.01839 1.8438    376.00    825.1  1201.1  0,5675    0.9588      1.5263    375.14          1115.B    250 250          400.95    0.01865 1.5433    393.84    809,0  1202.B  0.5879    0.9225      1.5104    392.79          1117.1    300 300          417.33    0.01890 350 350 400 II 431.72 444,59 0.01913 0.0193 1.3260 1.1613 409.69 424.0 794,2 780.5 1203.9 1204.5 0,6056 0.6214 0,8910 0.8630 1.4966 1.4844 408.45 422.6 1118.0 1118.5    400 1.0320    437.2    767,4 ! 1204.6  0,6356    0.8378      1.4734    435.5          1118.7    450 450          456.28    0.0195 0.9278    449.4    755.0  1204.4  0,6487    0.8147      1.4634    447.6          1118.6    500 500          467.01    0.0197 0.8424    460.B    743.1  1203.9  0.6608    0.7934      1.4542    458.B          11 lB.2  550 550          476.94    0.0199 0.7698    471.6    731.6  1203.2  0.6720    0.7734      1.4454    469.4          1117.7    600 600          486.21    0.0201 0.7083    481.B    720.5  1202.3  0.6826    0.7548      1.4374    479.4          1117.1    650 650          494.90    0.0203 0.6554    491.5    709.7  1201.2  0.6925    0.7371      1.4296    488.B          1116.3    700 700          503.10    0.0205 0.6092    500.B    699.2  1200.0  0.7019    0.7204      1.4223    598.0          1115.4    750 750          510.86    0.0207 0.5687    509.7    688.9  1198.6  0.7108    0,7045      1.4153    506.6          1114.4    BOO 800          518.23    0.0209 0.0210                0.5327    518,3    678.B  1197.1  0.7194    0.6891      1.4085    515.0          1113.3    850 850          525.26 0.5006    526.6    668.8  1195.4  0.7275    0.6744      1.4020    523.I          1112.1    900 900          531.98    0.0212 0.4717    534.6    659.1  1193.7  0.7355    0.6602      1.3957    530.9          1110.8    950 950          538.43    0.0214 0.0216                0.4456    542.4    649.4  1191.B  0.7430    0.6467      1.3897    538.4          1109.4  1000 1000          544.61 0,0220              0,4001    557.4    630.4  1187.B  0.7575    0.6205      1.3780    552.9          1106.4  1100 1100          556.31                                                                                                                          1200 567.22    0,0223                0.3619    571.7    611.7  1183.4  0.7711    0.5956      1.3667    566.7          1103.0 1200                                                                                                                                          1300 577.46    0.0227                0.3293    585.4    593.2  1178.6  0.7840    0.5719      1.3559    580.0          1099.4 1300                                                                                                                                          1400 0.0231              0,3012    598.7    574.7    1173.4  0.7963    0.5491      1.3454    592.7          1095.4 1400          587.10                                                                                                                          1500 0.0235              0.2765    611.6    556.3    1167.9  0.8082    0.5269      1.3351    605.I          1091.2 1500          596.23 671.7    463.4  1135.1  0.8619    0.4230      1.2849    662.2          1065.6  2000 2000          635.82    0.0257              0.1878 730.6    360.5  I 091.1  0.9126    0.3197      1.2322    717.3          1030.6  2500 2500          668.13    0.0287              0.1307 802.5    217.B  1020.3  0.9731    0.1885      1.1615    783.4            972.7  3000 3000          695.36    0.0346              0.0858 902.7        0      902.7  1.0580          0      1.0580    872.9            872.9  3206.2 3206.2        705,40    0.0503              0.0503
    *Abridged from "Thermodynamic Properties of Steam" by Joseph H. Keenan and Frederick G. Keyes. Copyright, 1936, by Joseph H. Keenan and Frederick G.
Keyes.
Published by John Wiley & Sons, Inc., New York.
FOXBORO TAB 9, Rev. 1
 
      'C*                                                    0                                                                                                                                          0                                                                Q
      }                                                                                                                                      ..J .                                                                                                                                  ~
VISCOSITY                                              COMPARISON CHART 0          IOO              200              300              400              500          600              700          800            900            IOOO          1100            1200          1300          1400            1500 Centi stokes [>'
12 5                      25                      32 5                    50                675                      75                875          100                                    125 Mobilometer I>
100g 20cm 25                                  50                            7~                        100                  125                                150                          175 Engler        [>
2~              50                  75            IQO              12_5            1~0          175            2\)0        2?5          2~0            275          300            325            350        375          400 Ford 4        [>
25          50      75          100        125        150        175      200        225      250      275        300    325      350        375      400              450                5&#xa3;0                550              690
[>                I      I            I          I          I          I        I          I        I      I            I      I        I            I        I                I                                      I Ford 3              I 50              100                  150              200              250            300              350          400            450                500                550          600              650            700 Sayboll Furol  C>
500              1000                1500              2000            2500          3000              3500          4000          4500            5000                55_00          6000            65_00          7000 Sayboll        [>
Universal 250    500        750      1000      1250      1500              2000              2500              3000            3500              4000                4500                5000            5500          6000 Redwood        [>
200                300              400              500            600              700          800            900          1000 Ubbelohde 3    C>
MA2ABCD&#xa3;FG                    H    IJKLMNOP                            0    R S          T          u                                      V                          w Gardner-Holdt [> ~
SCALES ABOVE READ IN CENTISTOKES (TO CONVERT INTO CENTIPOISES, MULTIPLY BY LIQUID SPECIFIC GRAVITY)
                                                                                                                                                                                                                                                                                .,I SCALES BELOW READ DIRECTLY IN CENTIPOISES 12              15                      20                    25                      30                      35                      40                      4 Demmler 10 [>
15              20                                    30                              40                              50                            60                                  7, Zahn 5        [>
20              30                40                50              60 Zahn 3        [>
45              55                            65                                                    75                                          85                                              95 Kreb Stormer[>
Stormer Cyl. [>        15      25              50                                                    115                                                                          225 150g
-I                                              250              500                  750                1000              12~            1500              1750        2000            2250            2500            2750          3000              3250          3500
)>
0::,                  MacMichael [>
CD                                                                                                                                                                                                                    1100            1200          1300          1400          1500
~
Brookfield    [>.0            100              200              300              400              500            600              700          800            900            1000
::II                                    0            I()()            200                300                              500            600              700          800            900            IOOO          1100            1200          1300          1400          1500 CD                    Centi poises (> I      I        I      I    I    I        I        I      I  I  4&#xa3;0'              I              I                I            I      I  I  I              I          I  I          I  I    I    I    I
                                                                                                                                                                                                                                                          '  I    I  I
                                                                                                                                                                                                                                                                          '      '  I I
                                                                            '        I                                  I
                                                                                                                                '      I
                                                                                                                                                '      I        I  I
                                                                                                                                                                              '              '      I
.... NOTE This chorl is intended to be on aid in determining opproximote comparative                                                Rl.'rrintcJ wirh permission of BrookficlJ Eng. Lah:-. Im:. Swughton, M;1~:-.
viscosities of Newlonion fluids                                                                                                                                                                                                                                                AR-15-1
 
LIQUID GRAVITY TABLES AND WEIGHT FACTORS Liquid Heavier than Water Liquid Llthter than Water Sp Gr 60 F/60 F 0
19    0 API Lb per g
* I at 60 F In Lb per cu ft at 60 F Sp Gr 60 F/60 I
                                                                      ....  'Tw Lb per gal at 60 f In vacuo Lb per cu If at 60 F In vacuo Sp Gr 60 F/60 F    "B9      0 Tw Lb per gal at 60 Fin vacuo Lb per cu It at 60 F In vacuo vacuo      In vacuo 1.00S      0.72        1  8.3791      62.6817        I.SOS    '8.65      101  12.s,1a      93.8667 0.600    103.33    1003      5.0025      37.,219                                8.-4208    62.9936        1.510    48.97        102  12.5895      94.1785 102.38 1  S.OU2        37.7338    1.010      1.44      2
    .605    101.41                                      1.015      2.U        3    8.4625      63.305'        I.SIS    49.29        103  12.6312      9-4.4904
    .610
    .615
    .620 99.51 97.6' 95.80 100.*7 98.58 96.73 5.0858 5.1275 5.1692 3a.o*56 38.3575 38.6693 1.020 1.02S 2.8' 3.5'      *s  a.so,2 8.5459 63.6173 63.9291 1.520 1.52S
                                                                                                                          *9.61
                                                                                                                          *9.92 10, IOS 12.6729 12.7146 9 ... 8022 95.IUI 1.030      ,.22      6    8.5876      6'.2'10        1.530    50.23        100  12.7563      95 .* 259
    .625    9'.00      9 *.90  5.2109      38.9812                                            6'.5528        1.535    50.5'        107  12.7980      95.7378 5.2526      39.2930    1.035      *. 90      7    8.6293
    .630    92.22      93.10                            1.040      5.58        8  8.6709      6 ... 86 .. 7  1.s,o    S0.84        108  12.8397      96.0,96
    .635      90.*7    91.33  5.29'3      39.60*9                                8.7126      65.1765        1.S'5    SI.I 5      109  12.8813      96.361 S 5.3360      39.9167    I.OH        6.24        9
    .6'0      88.75      89.591                          1.050      6.91      10    8.75'3      6s.,aa,        1.550    s,.,s        110  12.9230      96.6733
    .6'5      87.05      87.881 5.3776      *0.2286 I.OSS      7.56      II    8.7960      65.8002        1.555    51.75        111  12.96 .. 7    96.9852
    .650      85.38      86,191 5.,193      ,o.5*o*                          12    8.8377      66,1121        1.560    s2.os      112  13.006'      97.2970 84.53! 5.4610      ,o.&523    1.060      8.21
    .655      83.7*                                      1.065      8.85      13    8.879*      66 .* 239      1.565    52.3S        113  13_0,a,      97.6089
    .660      82.12    82.89,  5.5027      "1.16" I    1.070                u    8.9211      66.7358        1.570    52.6'        IU  13.0898      97.9207 80.53    81.281  S,SA,U,      .. ,.,160              9.*9
    .665                                                1.075    10.12      15    8.9627      67.o,76        1.575    52.9.        115  13.1315      98.2326
    .670    78.95      79.691  5.5861      ,1.7B7e 1.080    10.1*      16    9.004'      67.3595        1.580    SJ.23        116  13.1732      98.5 .. 44
    .675    77.*1      78.13  5.6278      ,2.0997                                                            1.585    53.52        117  13.2148      98.8563 5.6695      42.4115    I.OBS    11.36      17    9.0 .. 61  67.6713
    .680    75.88      76.59                            1.090    11.97      18    9.0878      67.9832        1.590    53.81        118  13.2565      99.1681
    .685    7*.38      75.07  5.7111      42.723 ..                                                          1.595    S,.09        119  13.2982      99.* 800 5.7528      *3.0352    1.095    12.58      19    9.1295      68.2950
    .690    72.90      73.57                            1.100    13.18      20    9.1712      68.6069        1.600    54,38        120  13.3399      99.7918
    .695    1, .... 72.10  5.79'5      *3.3*71 I.IDS    13.78      21    9.2129      68.9187        1.605    5'.66      121  13.3816    100.1037 70.00      70.6'  5.8362      *3.6589
    .700
    .705
  .710
  .715 68.58 67.18 65.80
            ,, ..u 69.21 67.80 66.,o um 5.9613 1
                                            *3.9708 AA,2826
                                            .U.5945 1.110 1.115 1.120 1.125 14.37 1 ... 96 15.5' 16.11 22 23 2,
25 9.2546 9.2962 9.3380 9.3796 69.2306 69.5'2' 69.85'3 70.1661 1.610 1.615 1.620 l.625 5 ... 9 ..
SS.22 55.*9 55.77 122 123 12' 125 13.* 233 13.4650 13.5067 13.5483 I 00.'1 SS I 00.727*
101.0392 101.3511
  .720                65.03  6.0030      ... 9063 1.130    16.68      26    9** 213    7o.,7ao        1.630    56.0*        126  13.5900    101.6629
  .725      63.10      63.67  6.04'6      *5.2182                          27    9.4630      70.7898        1.635    56.32        127  13.6317    101.97*8 61.78      62.3'  6.0863      '5.5300    1.135      17.25
  .730                                                1.140      17.81      28    9,S0*7      71.1017        1.6 .. 0 56.SS        128  13.673*    102.2866
  .735      60.,a      61.02  6.1280      45.8419 I.US      18.36      29    9.546*      71.4135        1.645    56.85        129  13.7151    102.5985
  .7*0      59.19      59.72  6.1697      46.1537                                                            l.650    57.12        130  13.7568    102.9103 46,46S6    1.150      18.91      30    9.5881      71.7254
  .745      57.92      58.*3  6.21U 1.155    19.46      31    9.6297      72.0372        1.655    57.39        131  13.7985    103.2222
  .750      56.67      57.17  6.2531      l.6.777A                                                            1.660    57.65        132  13.a,02    103.5340 55.92  6.29'8      A7.0893    1.160      20.00      32    9.671'      72.3.91
  .755      55.*3                                      1.165    20.5 .. 33    9.7131      72.6609        l.665    57.91        133  13.8818    103.8-i59
  .760      s,.21      54.68  6.3365      *7.-4011                                                                    58.17      13'  13.9235    10,.1s11 I                                ,47,7130    1.170      21.07      3*    9.75 .. 8  72.9728        1.670
  .765      53.01      53.*7  6.3781                  1.175      21.60      35    9.7965      73.2846        1.675    sa.*3      135  13.9652    I o, ... 696
  .770      51.81      52.27  6.* 198    ,a.02,a 1.180    22.12      36    9.8382      73.5965        1.680    58.69        136  U.0069      10,.781"'
  .775
  .780
  .785 I  50.65 A9,49 A8.J,4 51;08
                        ,49.91
                        '8.75 6.4615 6.5032 6.54'9
                                            *B.3367 AS.6-485 48.9604 1.185 1.190 1.195 22.6' 23.15 23.66 37 38 39 9.8799 9.9216 9.9632 73.9083 1,.2202 7*.5320 1.685 1.690 l.695 58.95 59.20 59.45 137 138 139 1,.0.. 86 1,.0903 U.1320 105.0933 105 ... 051 105.7170
  .790      ,1.22      *7.61  6.5866      -49.2722                                                            1.700    59.71        140  14.1736    106.0287 6.6283      .9.58'1    1.200      2 ... 17  *o  10.00,9      1,.a,39
  .795      46.10      46.49 1.205      2,.67      *1  10.0466      75.1557        1.705    59.96        141  14.2153    106.3 .. 07
  .800      '5.00      45.38  6.6700      .9.8959                                10.0883      75 ... 676      1.710    60.20        142  I 4.2570    I 06.6525
                        .. ,.21 6.7116      50.2077    1.210      25.16      *2
  .805      43.91                                      1.215    25.66      '3  10.1300      75.779*        1.715    60 ... 5    143  1,.2987    I 06.96 .. 4
  .810      ,2.u      *3.19  6.7533      50.5196                                            76.0913        1.720    60.70        144  1,_3 .. 0,  I 07.2762
            ,1.78      .. 2.12 6.7950      50.8315      1.220    26.IS          10.1717
  .815                                                1.225      26.63      ,s  10.213*      76.4031        1.725    60.94        145  14.3821    I 07.5881
  .820      *o.73      *1.06  6.11367    51.1*33 1.230      27.11      46  10.2551      76.71 *9        1.730    61.1 8      146  14.4238    I 07.8999
  .825
  .830
  .835
  .uo 39.70 38.67 37.66 36.67
                        ,0.02 38.98 37.96 36.95 6.878*
6.9201 6.9618 7.003' 51.4552 51.7670 52.0789 52.3907 1.235 1.2,0 I.HS 27.59 28.06 28.53
                                                                              ,1
                                                                              *a
                                                                              *9 10.2967 I 0.338' 10.3801 77.0268 77.3387 77.6505 1.735 1.1,0 1.7'5 61.43 61.67 61.91 U7 148 149 14.,65S 14.5071 u.5 .. 88 108.2118 I 08.5236 I 08.8355 109.1473
                                                                                                                                                                    ,)
1.250      29.00      so  10.,218      77.962'        1.750    62.1 S      ISO  14.590S
  ,845      35.68      35.96  7.0'51      52.7026
  .850      3*.71      3*.97  7.0868      53.014'                          SI  10.4635      78.27*2        1.755    62.38        I SI 14.6322    109.4592 1.255    29.46 1.260      29:92      52  I0.505~      78.5861        1.760    62.61        I 52 14.6739    109.7710
  .855      33.7*      3'.00  7.1285      53.3263                                10.5469      78.8979        l.765    62.85        153  U.7156      110.0829 32.79      33.03  7.1702      53.6381    1.265    30.38      53
  .860                                                1.270      30.83      s,  I 0.5885      79.2098        1.770    63.08        15'  , .. _7573  110.39 .. 7
  .865      31.85      32.08  7.2119      53.9500                                I 0.6302      79.5216        1.775    63.31        155  U.7990      I 10.7066 31.U    7.2536      5'.2618    1.275      31.27      55
  .870      30.92
  .875      30.00      30.21  7.2953      5,_5737                                                              1.780    63.5A      I 56  , ... 8406  111.0184 1.280      31.72      56  10.6719      79.8335 1.285      32.16      57  10.7136        80.1'53        l.785    63.77        157  I A.8823    111.3303
  .880      29.09      29.30  7.3369      5'.8855                                                              1.790    6 ... oo    158  U.9240      111.6421 28.39  7.3786      5S.l97l    1.290      32.60      58  I 0.7553      80.4572
  .885      28.19                                      1.295      33.03      59  10.7970      80.7690        1.795    6'.22        159  14.9657    111.95'0
  ,890      27.30      27.*9  7.'"203    55.5092                                              81.0809        1.800    6 ....... 160  I S.007*    112.2658 26.60  7.4620      55.8211    1.300    33.46      60  I 0.8387
  .895      26.*2
  .900      25.56      25.72  7.5037      56.1329                                10.aao*      81.3927        I.BOS    6 ...67      161  1S.0*91    I 12.5777 1.305    33.89      61 1.310    34.31      62  10.9220      81.7046        1.810    6'.89        162  1 S.0908    112.8895
  .905      2'.70      2'.85  7.SAS-4    56.A,'48                                                            1.815    65.11      163  15.1325    113.201' 7.5871      56.7566    1.315    3'.73      63  I 0.9637      82.016'
  .910      23.85      23.99                          1.320    35.1 S      6'  11.005'      82.3283        1.820    65.33      16'  15.1741    113.5132
  .915      23.01      23.U    7.6288      57.0685                                11.0,71      82.6'01        1.825    65.SS        165  15.2158    113.8251 22.17      22.30  7.6704      57.3803      l.32S    35.57      65
  .920
  .925      21.35      21.47  7.7121      57.6922                                11.0888      82.9520        1.830    65.77        166  I S.2575    11,.1369 1.330    35.98      66 1.335    36.39      67  11.1305      83.2638        1.835    65.98        167  I S.2992    1u.,,aa
  .930      20.5'      20.65  7.7538      5a.oo*o                                11.1722'      83.5757        1.8'0    66.20        168  IS.3'09    11,.7606 19.8*  7.7955      58.3159    1.3,0      36.79      68
  .935      19.73                                      1.34S      37.19      69  11.2139      83.8875        I.US    66 ... I    169  I S.3826    115.0725
  .9,o      18.93      19.03  7.8372      58.6277                                11.2555      8'.199*        1.850    66.62        170  15.42'3    11 S.38'3
  .9,S      18.15      18.24  7.8789      58.9396    1.350    37.59      70
  .950      17.37      17.'5  7.9206      59.251
* 1.355    37.99      71  11.2972      8'.5112        l.855    66.83        171  15.4660    I I S.6962 1.360    38.38      12  11.3389      a,.a231        1.860    67.o*        172  I S.5076    116.0080
  .955      16.60    -16.67  7.9623      59.5633                                11.3806      85.13'9        1.865    67.25      173  I S.5'93    116.3199 15.83      15.90  8.0039      59.8751    l.365    38.77      73
  .960                                                1.370    39.16      7*  11.4223      BS.4468        1.870    67.*6      17*  15.5910    I 16.6317
  .965      IS.OB      15.13  8.0456      60.1870                          75  11.46'0      85.7586        1.875    67.67        175  I S.6327    I I 6.9,36 U.33        U.38  8.0873      60.* 988    1.375    39.SS
  .970
  .975      13.59      13.63  8.1290      60.8107                                11.5057      86.0705        1.880    67.87      176  15.67 .. 4  117.255' 1.380    39.93      76 1.385    ,o.31      77  I 1.5*7*      86.3823        l.885    68.08      177  15.7161    I I 7.5673
  .980      12.86      12.89  8.1707      61.1225                                                              1.890    68.28        178  15.7578    117.8791
  .985
  .990
  .995 12.13 10.70 12.15 11.43 10.71 8.212' 8.25'1 8.2958 61,4344 61.7462 62.0581 1.390 l.395 1.,00
                                                                  ,o.68
                                                                  '1.06
                                                                  ,u3 78 79 80 11.5890 11.6307 11.672*
86.69*2 87.0060 87.3179 1.895 1.900 68 ... 8 68.68 179 180 I S.799*
15.8411 118.1910 II 8.5028 1.000      10.00      10.00  8.337*      62.3699                                                            1.905    68.88      181  I S.8828    118.81'7 uos      ,1.eo      81  11.7141      87.6297 1.410    ,2.17      82  11,7558      87.9'16        1.910    69.08        182  15.92'5    119.126S 1.,1s    ,2.53      83  11.7975      88.253*        1.915    69.28      183  I S.9662    I I 9.4384 U20      *2.89      a,  11.8392      88.5653        1.920    69.*B      , a,  16.0079    119.7502 1.425    *3.25      85  11.8809      88.8771        1.925    69.68      I BS  16.0.96    120.0621 1.430    *3.60      86  11.9225      89.1890        1.930    69.87      186  16.0913    120.3739 1.435    *3.95      87  11.96'2      89.5008        1.935    70.06      187  16.1329    120.6858
                                                        , ..... 0 ..... 31    88  12.0059      89.8127        1.9A0    70.26      188  16.1746    120.9976
                                                        , ..... 5 '4.65      89  12.0'76      90.12*5        1.9'5    70.45      189  16.2163    121.3095 1.450    '5.00      90  12.0893      90.,36'        1.950    70.6'      190  16.2580    121.6213 I.HS      '5.3*      91  12.1310      90.7'82        1.955    70.83      191  16.2997    121.9332 1.460    ,S.68      92  12.1727      91.0600        1.960    71.02      192  16.341"'    122.2 .. so 1.465    .. 6.02    93  12.2143      91.3719        1.965    71.21      193  16.3831    122.5569 U70      46.36            12.2560      91.6838        1.970    11.,0      19*  16.,2.a    122.8687 9*                                                            16.466'    123.1806
                                                        , ... 75  46.69      95  12.2977      91.9956        1.975    71.58 "195 1.480    *7.03      96  12.339*      92.3075        1.980    71.77      196  16.5081    123.*92*
                                                        ,.,as                      12.3811      92.6193        1.985    71.95      197  16.5498    123.80 .. 3
                                                                  *7.36      97                                                            16.591 S    12,.1161 I.A90    *7.68      98  12.,228      92.9312        1.990    72.IA      198 1.,95    0.01        99  12.464.S      93.2'30        1.995    72.32      199  16.6332    *12U280 l.500    0.33      100    12.5062      93.5549        2.000    72.SO      200    16.67"'9    12*.7398 TAB 9, Rev. 1
 
PHYSICAL CONSTANTS OF PARAFFIN HYDROCARBONS COMPOUND                    METHANI        ETHANE    PROPANE      ISO-BUTANE N-BUTANE ISO-PENTANE N-PENTANE N-HEXANE N-HEPTANE N-OCTANI                        N-NONANE N-DECANE MOLECULAR FORMULA **** ,,,, ** ,, ,              CH,        C.H,            C,H,          C,H,o                                                                        C,H11          C,H,o  i MOLECULAR WEIGHT ...... , ..... ,,              16.042      30.068        44.094        5l120      5l120      72.146          72.146      16.172      100.191    114.224        12l250    !  142.27&
MELTING POINT at 14.696 psla F .. .. ..        -296.5'1    -297.9        -305.1        -255.3      -217.0      -255.1          -201.5      -139.&        -131.1        -70.2        -64.4 I      -21.5
            "        "                C .. .. ..  -182.5'1    -113.3          -187.7        -159.5      -13l3      -159.9          -129.7        -95.3        -911.6      -56.1        -53.&        -29.7 IOILING POINT at 14.696 psla F .. .. .. *      -25l7      -127.5          --43.7            10.9        31.1        12.1          96.9        155.7        209.2      25l2          303.4        345.2
            "        "                C .. .. .. *  -161.5        --ll6          ---42.1        -11.7        ---0.5      27.9            36.1          61.7          91.4      125.7        150.1        174.0 DENSITY OF LIQUID at 60 F and 14.696 psla Specific Gmlty at 60/60 F* **********              0.3<    0.374        0.507Jh        0.5631*    0.51441>    0.6241          0.6312      0.6640        0.6112      o.7D61  !    0.7217      0.7341 0
API * *************************                340<        247          147.2          119.1      110.6        95.0            92.7          11.6          74.1        61.7          GU I          61.3 Lbper1alat60F* ............... .                  2.5<      3.11        4.224'          4,685h      4.163h      5.200          5.253        5.527        5.721        5.113        &.DOI i      6.114 Gal per lb maleat 60 F *************              u,        9.67          10.44*        12.40h      11.95h      13.U            13.74        15.59          17.49      19.41:        21.35 i      23.21 DENSITY OF VAPOR at 60 F and 14.696 psla                                                                                                                              i                        I Specific iravlty air =1.~eal 1as ****          0.554*      1.038*        1.522*        2.DD&e      2.006'      2.491          2.491        2.975        3.459  i    3.943 ,      4.421        4.912
                    " -actual (corrected) *****        0.555      1.046          1.546          2.066      2.010 Lb per M cu It-Ideal 1as ************          42.21*      79.23*        116.19'        153.15'    153.15'    190.11          190.11      227.01        264.03  i  300.99 !      337.95      314.91
)                    " -actual (corrected) *****
Cu fl vapor per 1al llq-ldeal 1as ******
42.35 59' 19.86 39.25' 11l0 36.35e 157.1 30.59' 158.1 31.75'      27.40          27.68        24.31        21.73      19.51        11.10      16.32
          "          "      -actual (corrected>                                35.11          29.10      30.77 RATIO, GAS VOL PER LIQ VOL-ldeal 1as                          293.&e        271.Se                    237.5'    205.0          207.1        112.4        162.6      146.5        133.2        122.1
            "          "      -actual <corrected)                                267.6          222.1      230.1                                                                            -      j CRITICAL CONDITIONS Temperature-F *******************            -116.5          90.1          206.3          275.0      305.6      370.0          315.9        454.5        512.6      565.2 I1    613.0  !    655.0
              "        C ................... .      --82.5        32.3          96.8          135.0      152.0      117.1          196.6        234.1          267.0      296.2  i      322.1        346.1 Pressure-atmosphere **************              45.8        41.2          42.0          36.0        31.5        32.9            33.3          29.9          27.0        24.6          23.5 I      21.1 Pressure-psla .................. .                613          708            617            529        551        413            490          440          391        362 I          345          320 GROSS HEAT OF COMBUSTION at 60 F Btu per lb rapar-ldeal 1as ***********        23,891    22,329          21,610        21,265      21,315      21,046          21,09-4      20,9-49        20,142      20,164  i    20,701      20,653 Btu per cu fl-Ideal 1as ............ .        1,010*      1,769'          2,511*        3,256'    3,264*      4,001          4,009        4,756        5,503                    6,996        7,743
            "      " -actual ccorrected) *******      1,012      1,713          2,558          3,354      3,361 Btu per 1al liq at 60 F ............ ..                  69,433(1)      91,044        99,097    103,047    108,120        110,125      115,069      111,651    1 121,420    I  123,625  I  121,4ss
  ) FLAMMABILITY LIMITS Lawer petcentln air ............... .            5.0        3.22          2.37          1.80        1.86        1.32            1.40          1.25          1.0  I    0.14 I
                                                                                                                                                                                        !      0.14  !      0.61 Upper percentln air .............. ..          15.0      12.45            9.50          1.44        1.41                        7.ID        6.90            6.0  I      3.Z  !,        2.9          2.1 CU FT OF AIR ta burn 1 cu fl 1as .......          9.53      16.67          23.12          30.97      30.97      3ll1            3ll1          45.26        52.41      59.55  I    66.70  I    73.15 HEAT OF VAPORIZATION at 14.696 psla Btu per lb at ballln1Palnt .......... ..      219.1      210.1          183.5          157.8      165.9      145.9          153.1        144.2  !    136.2      131.9    j    126.9  !    120.2 I
I SPECIFIC HEAT at 60 F and 14.696 psla C, vapor Btu per lb .............. ..        0.5271      0.4097        0.3185          0.3172      0.3970      0.3880          0.3972      0.3914        0.3992      0.3991  I    0.4003  i  0.4006 C, vapor Btu per lb ............... .        0.402      0.343          0.342          0.352      0.363      0.361          0.370        0.375    I    0.379        0.312        D.315        0.317 N C,/C, Btu per lb **************.*            1.308      1.193          1.133          1.097      1.094      1.076          1.074        1.062        1.052      1.046 :      1.040        1.034 C, liquid Blu per lb ............... .                              0.534 al--43 F 0.537 al 14 F 0.541 at OF    0.533          D.536        o.552        0.521      0.523        0.522        0.520 VAPOR PRESSURE at 100 F, psla ...... .                          7BOJ            190          72.2        51.6        20.4            15.6        4.96          1.62        0.54 '        0.11          0.01 ANILINE POINT-F ............... .                                            -              225.7      111.2      161.3          159.3        155,51        159.1      161.1        166.1        171.5 20 REFRACTIVE INDEX* N ; at 68 F....                                                                                  1.3537          1.3575      1.3749        1.3176      1.3974        1.4054      1.4120 DIELECTRIC CONSTANT* at 20 C .... ,                                      1.61 atO C                                1.143          1.144        1.190        1.924      1.941        1.972        1.991 J      &Air saturated hydrocarbons bAbsolute values from weights in vacuum
                                                                                    *Based on "perfect gas" t(alculated lCritical solution temperatures JExtropolated value cApparent values from weights in air                                        rApparent value for dissolved Methane at 60 F                  *Dielectric Constants from NBS Circular 514; other dAt saturation pressure (triple point)                                      h5aturation pressure                                          dala from NGAA Publicalion 2145-Revised 1957.
CONVERSION DATA FOR HYDROCARBON CALCULATIONS Atomic weights: Carbon-12.01; Hydrogen-1.008.                                                                                              141.5 Degrees API = - - - - - - -131.5 Molecular weight of air-28.966.                                                                                                  Sp Gr at 60/60 F Perfect gas at 32 F and 14.696 psia = 22.414 liter per                                                  Degrees Fahrenheit = 459.67 Rankine.
gram mole.
Perfect gas at 60 F and 14.696 psia = 379.498 cu ft per                                                760 mm Hg = 14.696 psia.
mole.                                                                                                  1 pound = 453.6 grams.
Specific gravity at 60/60 F x 0.999015 = density at 60 F                                                1 cu ft        28.32 liter.
in grams per cc.
Density of water at 60 F = 8.337 lb per gal = 0.999015                                                  1 cu ft = 7.481 gal.
grams per cc.                                                                                          1 gal = 3,785 ml.
TAB 9, Rev. 1
 
STAINLESS STEEL COMPOSITION (Percent)
AISI*        Cr          Ni        C        Mn          Si        p        s          Other                      REMARKS Type 201          16-18      3.5-5.5  .IS max    5.5-7.5    I max  .06 max  .03 max    N  .25 max      Low-nickel Equivalent al Type 301 202          17-19          4-6  .IS max    7.5-10      I max  .06 max  .03 max    N  .25 max      Low-nickel Equivalent of Type 302 301          16-18          6-8  .)5 max      2 max    I max  .045 max  .03 mall                    High Work-hardening 2 max          .045 max  .03 max                    General-Purpose "18-8 11 302          17-19          8-10  .IS max                I max 2 max    2-3    .045 max  .03 max                    More Scaling Resistance than Type 302 302B        17-19          8-10 .15 max
{ .6 Mo or 2 max              .2 max .IS min                    Free Machining  11 1 8-8"-Heavy Cuts 303          17-19          8-10 .IS mall              I max                        .6 Zr optional 2 max              .2 max .06 max    Se .IS min      Free Machining "1 8-8 11-Light Cuts 303Se        17-19          8-10 .IS max                I max
                                                                .045 max  .03 max                    Low-Carbon-For Welding 304          18-20          8-12  .08 max      2 max    I max
                                                                .045 max  .03 max                    lower-Carbon-For Welding 304L        18-20          8-12  .03 max      2 mall  I max
                                                                .045 max  .03 max                    lower Work-hardening Rate 305          17-19        10-13  .12 max      2 max    I max
                                                                .045 max  .03 max                    Welding Rod-for Ductility 308          19-21        10-12  .08 max      2 max    I max
                                                                .045 max  .03 max                    High-temp Strength and Scaling Resistance 309          22-24        12-15    .2 max      2 max    I max
                                                                .045 max  .03 max                    Low-Carbon Type 309-For Welding 309S        22-24        12-15  .08 max      2 max    I max
                                                                .045 max  .03 max                    Better High-temp Strength and Scaling Res.
310          24-26        19-22  .25 max      2 max  1.5 max
                                                                          .03 max                    Low-Carbon Type 310-For Welding 310S        24-26        19-22  .08 max      2 max  I.Smax  .045 max
                                                                .045 mall  .03                        Most Scaling Resistant 314          23-26        19-22  .25 max      2 max  1.5-3                  max
                                                                .045 max  .03        Mo  2-3        Increased Corrosion Resistance 316          16-18        10-14  .08 max      2 max    I max                max
                                                                          .03        Mo 2-3          Low-carbon Type 316-For Welding 316L        16-18        10-14  .03 max      2 max    I max  .045 max      max
                                                                .045 max  .03        Mo 3-4          More CorrosiOn Resistance than Type 316 317          18-20        11-15  .08 max      2 max    I max                max
                                                                .045 max  .03        Ti  SxC min    Stabilized Against Carbide Precipitation 321          17-19          9-12  .08 max      2 max    I max                max
                                                                .045 max  .03 max    Cb+Ta IOXCmin    Stabilized Against Carbide Precipitation
                                                                                                                                                  - )1 347          17-19          9-13  .08 max      2 max    l max
( Cb +Ta I OXC min
                                                                .045 max  .03 max                    Stabilized Against Carbide Precipitation 348          17-19          9-13  .08 max      2 max    I mcx
                                                                                    \ but    Ta .I max 403        11.5-13          -    .I 5 max      l max  .5  max  .04 max  .03 max                    Like Type 41 0-"Turbine Quality" 405        11.5-14.5        -    .08 max      l max    I max    .04 max  .03 max    Al .l-.3        Reduced Heat-treat-hardenability 11
                                                                  .04 max  .03 max                    General-Purpose        12 Cr" 410        11.5-13.5              .15 max      1 max    I max
                                                                  .04 max  .03 max                    Better Strength than Type 410 414        11.5-13.5    1.25-2.5  .IS max      1 max    I max
{ .6 Mc or 416          12-14          -    .I 5 max  1.25 max    I max    .06 max  .IS min    .6 Zr optional Free Machining "12 Cr"-Heavy Cuts 416Se        12-14          -    .15 max    1.25 max    I max    .06 max  .06 max    Se .IS min      free Machining ., 12 Cr"-:-light Cuts 420          12-14          -    ever  .15    I max    I max    .04 max  .03 max                    Like Type 41 0-For Higher Hardness 11
                                                                  .04 max  .03 max                    General-Purpose        17 Cr" 430          14-18                .12 max      I max    I max
[ .6 Mo er "t 7 Cr"-Heavy Cuts 430F        14-18          -    .12 max    1.25 max    I max    .06 max  .IS min l .6 Zr optional  Free Machining 430F Se      14-18          -    .12 max    1.25 max    I max    .06 max  .06 max    Se  .15 min    Free Machining "17 Cr"-light Cuts
                                                                  .04 max  .03                        Hardenable-High Impact Strength 431          15-17      1.25-2.5    .2 max      I max    I max                max 440A        16-18          -        .6-.75    I max    I max    .04 max  .03 max    Mo  .75 max    Higher Hardness than Type 420 440B        16-18          -      .75-.95      I max    I max    .04 max  .03 max    Ma .75 max      Higher Hardness than Type 440A 440C        16-18          -    .95-1.2      I max    I max    .04 max  .03 max    Mo  .75 max    Higher Hardness than Type 440B 446          23-27          -      .2 max    1.5 max    I max    .04 max  .03 max    N  .25 max      Scaling Resistance at Elev Temp Heat Resistant, Good Mechanical Proper-501          4-6            -    ever. I 0    I max    I max    .04 max  .03 max    Mo  0.40-0.65  ties at Elev Temp When Annealed, Greater Ductility, Lower 502          4-6            -    .10 max      I max    I max    .04 max  .03 max    Mo  0.40-0.65  Tensile Strength than Type 501
*American Iron & Steel Institute.
TAB 9, Rev. 1                FOXBORO
 
STANDARD DIMENSIONS FOR WELDED OR SEAMLESS                                                WIRE TABLE, STANDARD ANNEALED COPPER*
STEEL PIPE-Schedule 10, 40, 80                                                        American Wire Gauge (B. & S.)
(English and Metric Units)                                                              Metric and English Units Cross S~tlon      Diameter      Gau1e      Diameter        Resistance lntern31 Diameter                                                                                    Inches        Ohms/1000 fl Threa~s    Nominal            mm'              mm          No.
Nominal Pipe        00                                                  Per    Wel1ht Size      Inches      Inches        Inches      Inches      Inch      lb/fl mm          mm                                                  11.68        0000      ,4600              0,04901 Inches mm          mm Sch 10 I    Sch 40 I    Sch 80 107.2 85.03 67.43 10.40 9.266 ODO 00
                                                                                                                                            .4096
                                                                                                                                            .3648
                                                                                                                                                                  .06180
                                                                                                                                                                  .07793 1/8      .405        .307          .269          .215  27          0.244                            8.252              0    .3249                .09827 53.48 10.29        7.80          6.83        5.46                                                7.348              1    .2893                .1239 42.41 33.63              6.544              2    .2576                .1563 1/4      .540        .410          .364        .302    18          0.424                                                                          .1970 26.67              5.827              3    .2294 13.72        I 0.41          9.25        7.67                                                                    4    .2043                .2485 21.15              5.189 3/8      .675        .545          .493        .423    18          0.567 16.77              4.621              5    .1819                .3133 17.15        13.84        12.52        10.74 13.30              4.115              6    .1620      I        .3951
                                                              .546    14          o.850          10.55              3.665              7    .1443                .4982 1/2      .840        .674          .622 15.80        13.87                                8.366            3.264              8    .1285                .6282 21.34        17.12 6.634            2.906              9    .1144                .7921
                                    .884          .824        .742    14          1.130            5.261            2.588            10    .1019                .9989 3/4      1.050 26.67        22.45        20.93        18.85                                4.172            2.305            11    .09074              1.260 3.309            2.053            12    .08081              1.588 1.315        1.097        1.049          .957  I 1-1/2    1.678            2.624            1.828            13    .07196              2.003 I
33.40        27.86        26.64        24.31                                2.081            1.628            14    .06408              2.525 1-1/4      1.660        1.442        1.380        1.278    I 1-1/2    2.272            1.650            1.450            15    .05707              3.184 42.16        36.63        35.05        32.46                                1.309            1.291            16    .05082              4.016 1.038            1.150            17    .04526              5.064 1-1/2      1.900        1.682        1.610        1.500    11-1/2      2.717            0.8231          1.024            18    .04030              6.385 48.26        42.72        40.89        38.10                                  .6527          0.9116          19    .03589              8.051
                                                                                                    .5176            .8118          20    .03196            10.15 2          2.375        2.157        2.067        1.939    11-1/2      3.652                                              21    .02846            12.80
                                                                                                    .4105            .7230 60.33        54.79        52.50        49.25                                                                  22    .02535            16.14
                                                                                                    .3255            .6438
                                                                                                    .2582            .5733          23    .02257            20.36 2-1/2      2.875        2.635        2.469        2.323      8        5.793
                                                                                                    .2047            .5106          24    .02010            25.67 73.03        66.93        62.71        59.00 2.900      8        7.575            .1624            .4547          25    .01790            32.37 3          3.500        3.260        3.068                                                                                                        40.81 82.80        77.93      73.66                                  .1288            .4049          26    .01594 88.90                                                                                                                                    51.47
                                                                                                    .1021            .3606          27    .01419 4.260        4.026        3.826      8        I0.790            .08098          .3211          28    .01264            64.90 4          4.500 114.3        108.2        102.3        97.18                                  .06422          .2859          29    .01126            81.83
                                                                                                    .05093          .2546          30    .01003          103.2 6          6.625        6.357        6.065        5.761      8      18.974              .04039          .2268          31    .008928        130.1 I 68.3      161.5        154.1        146.3                                  .03203          .2019          32    .007950        164.I
                                                                                                    .02540          .1798          33    .007080        206.9 8          8.625        8.329        7.981        7.625      8      28.554              .02014          .1601          34    .006304        260.9 219.1        211.6        202.7        193.7
                                                                                                    .01597          .1426          35    .005614        329.0 10          10.750      10.420        10.020        9.564      8      40.483              .01267          .1270          36    .005000        414.8 273.1        264.7        254.5        242.9                                  .01005          .1131          37    .OO,U53        523.1
                                                                                                    .007967          .1007          38    .003965        659.6 12          12.750      12.390        11.938      11.376      8      43.773                                                      .003531        831.8
                                                                                                    .006318          .08969          39 323.9        314.7        303.2        289.0
                                                                                                    .005010          .07987          40    .003145        1049
                                                                                                                                  *Dimensions and resistance at 20 C (68 F).
DECIMAL EQUIVALENTS 1/64 .015625                                  17/64 .265625                      33/64 .515625                          49/64 .765625 9/32          .28125            17/32            .53125              25/32              .7812S 1/32              .03125 19/64 .296875                      35/64 .546875                          51/64 .796875 3/64 .046875 5/16                      .3125    9/16                      .5625        13/16                      .8125 1/16                        .0625 5/64 .078125                                  21/64 .32812S                      37/64 .578125                          53/64 .82812S 11/32          .34375            19/32            .59375              27/32              .84375 3/32              .09375 23/64 .359375                      39/64 .609375                          55/64 .859375 7/64 .109375 3/8                      .375      5/8                      .625          7/8                      .875 1/8                          .125 9/64 .140625                                  25/64 .39062S                      41/64 .64062S                          57/64 .89062S 13/32          .40625            21/32            .65625              29/32              .90625 5/32                .1562S 11/64 .171875                                  27/64 .421875                      43/64 .671875                          59/64 .921875
                                                                                  .4375    11/16                      .6875        15/16                      .9375 3/16                          .1875            7/16 13/64 .203125                                  29/64 .453125                      45/64 .703125                          61/64 .953125
                                                                                  .46875            23/32            .71875              31/32              .96875 7/32                .21875                      15/32 15/64 .234375                                  31/64 .484375                      47/64 .734375                          63/64 .984375
*,..,,.                                                                            .5        3/4                      .75          l                        1.
1/4                          .25              1/2 TAB 9, Rev. 1
 
BAROMETRIC PRESSURE 29.92 in. Hg
                                                                                                                                                                          .10
                                                                                                                                                                          .09 a
0
                                                            ~
                                                                                                                                                                          .08 LEGEND D Dew Point Temperature EJ Grains of Moisture El Wet Bulb Temperature a Pounds of Water El Cubic Feet Per Pound a Dry Bulb Temperature fl Relative Humidity HOW TO USE CHART Locate point on chart at intersection of any two values. Obtain other values                                                                                                                                    .06 by following chart lines.
Example: at a dry bulb temperature a  of 172 F and a wet bulb tempera-ture El of 102 F, the following values may be obtained from the chart: dew point temperature D 87.5 F; cubic feet per pound of dry air El 16. 7; rela-tive humidity fl 10.3 percent; grains                                                                                                                                          )
of moisture per pound of dry air El 201; 0 pounds of water per pound of dry air ~ ?,H~f...+.[..l...,.j..-l-i,+,-:~..j../A~+:;,!,+~'..+.jc:j,-..1+.;::,...;,.._,~P,.~+;:~-JZ:,..,;.;..:...~.w., ;,+-+r!s!.--4++l"'IH,+
a  .0281.
TAB 9, Rev. 1 FOXBORO
 
                                                                                                                  .I z
    -~                                                                                                      V
                            **                                                                            L    ~
uo
  ,:")                                                                                                  ,                                                      I
* 100 F
                                                                                                                                                            ~
OIO
                                                                                                                                                  ~
                                                                                                                                                        ~
i '                                                                                                                                              '
w                                                              ,
                                                                                .P I/
                                                                                                                                                                        '10 :=
ill                                                  ,:;
                                                                  ,,i...__ .....,...
OEWC(L w
00 : ,
                      ~
                                                                                                                                                                      ....  ~
                                                                                                                                                  &deg;"-IIATINI
                      ,_                                      i;.~~
C;i
                                                                  'Ii:  I....-~
                                                                        '--t-;"
                                                                                  ..*~
                                                                                            ~J..I ill
                      ~                                                                                                                                                    ~
  )
iii
:l:
                      <      I
                                                          ~
I----~,_~    . ~.,.:-,:,      ..
                                                              ' - .,, ....... ..._ ',.,(~
                                                            ~'::::.'::::~"                  17 IO
                                                                                                                                                                            ~
I
                                                                                ,                                                                                          iii
:l:
17
                                                                                                                                                                      ~    <
* OJ I
                                                      ,,                                                                                                            0 z
                                    ,                                                                                                                                -10
                                  ~ +-bl                                                                                                                      L I,,
                                  -17'                                                                                                                                *o bl                                                                                                        I,.
00
                              -60    -40    -to        0      ~            ~          ~      ~      ~    ~      ,.,o  ""    100    10      to      40    ZO  -*
DEW POINT TEMPERATURE 'F.                                                    RELATIVE HUMIDITY %
The Foxboro Dew Point Nleasuring System measures dew point temperatures between -50 F and + 142 F within the limits of ambient temperature shown.
The operating range is shown in the left chart in terms                                                    Dew points at higher operating temperatures are measured b_y of dew point temperature. For example, at an ambient tem-                                                cooling a sample to a temperature within the operating range shown.
perature of 60 F, dew point temperatures from 60 F down                                                      For example, with a furnace atmosphere at 1600 F tem-to 10 F may be measured.                                                                                perature and 2 F dew point, the sample is cooled in a For convenience, the same operating range is shown in                                                refrigeration unit to 35 F. Its dew point is then readily the right chart in terms of relative humidity. At the same                                              measured by the Dewcel element.
ambient temperature of 60 F, used in the example above,                                                      Blast furnace air at 300 F temperature and 70 F dew it can be seen that the corresponding relative humidity                                                  point, may be measured by passing. it through an unin-range is 100 percent down to 13 percent.                                                                sulated pipe leading to a sampling chamber. Here, it is Another example illustrates the use of the chart in deter-                                            cooled to 120-130 F, enabling accurate dew point meas-mining a working range of ambient temperatures. To meas-                                                urement.
ure a dew point temperature of 60 F, it will be seen that                                                    Both the Sampling Chamber and Refrigeration Unit the range of ambient temperatures must be between 60 F                                                  can be furnished by The Foxboro Company for most and 133 F.                                                                                              applications .
* Readings are easily converted tu:
ABSOLUTE HUMIDITY. The most accurate method of determining the moisture content of air is to measure its dew point. Dew point values are readily converted to absolute humidity units or vapor. pressure units by reference to standard tables. Instruments reading absolute humidity directly are also available.
RELATIVE HUMIDITY. A two-pen Recorder can be furnished if a relative humidity measurement is desired.
One pen records dew point; the second pen records dry-bulb temperature. A table or chart is used to convert these readings to relative humidity.
Foxboro \Vet- and Dry-Bulb instruments, and direct-reading Relative Humidity instrum('nts of the hair type are also available.
TAB 9, Rev. 1
 
FOXBORO MANUFACTURING FACILITIES ARE
* UNITED STATES The Foxboro Company WORLD-WIDE I  Neponset Avenue Foxboro, Massachusetts
* CANADA Foxboro Company, Ltd.
I Foxborq instruments, whether 707 Dollard Avenue LaSalle, Quebec shippe*d from the United States, Canada, England, Holland, or Japan,
* ENGLAND                                are of identical design - built to Foxboro-Yoxall, Ltd.
I the same standards of quality for Redhill, Surrey sustained high performance.
* NETHERLANDS                              Whenever you purchase instru-Foxboro (Nederland) N.V.
I  Koninginne Laan, 169                  ments through one of our world-Soest                                  wide offices, you are assured of
                                                                                                  /
complete Foxboro Service.
* JAPAN Yokogawa Electric Works, Ltd.            The engineering background and I  2-9 Nakacho Musashino-shi, Tokyo
* AUSTRALIA Foxboro Pty., Ltd.
factory training of our representa-tives, their work with all types of measurement and control problems, and their constant, direct contact I  Lillydale, Melbourne
* FRANCE with Foxboro Application and Sales Engineering Groups, ma:ke them an integrated part of the Foxboro world-Foxboro France, S. A.
I  Arras, Franee
* MEXICO wide organization. They are ready, no matter how large or small your processing problem, to provide you Foxboro, S. .A.
Durango 262                          with the benefits of "Instrumenta-Mexico City 7, D.F., Mexico          tion by Foxboro."
FOXBORO  REGISTERED TRADEMARK Specialists in Process and Energy Control Engineering Offices in Principal Cities throughout the World TAB 9, Rev_ 1 Printed in U.S.A.
 
CONTAINMENT VOLUME REV.O CONTAINMENT VOLUME= 3.74 E10 cm 3
                                  = 1.32 E6 ft3
 
==Reference:==
USAR 5.2.1.1 Page 1 of 1 TAB 10, Rev. 0
 
                                                                . )
                                                                    '\
Introduction to Health Physics BY HERMAN CEMBER Northwestern Unh*ersity
                                ,,,,..k,,
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:<    OXFORD    LONDON
                          ~Ml auu .. *s ._.,,..:,
TO IPIDVSI._T I I I I PERGAMON PRESS EDINBURGH
* NEW YORK a      TORONTO
* SYDNEY              PARIS      BRAUNSCHWEIG
 
RADIATION DOSIMETRY                                  167 166                INTRODUCTION TO HEAL TH PHYSICS iJ    Dose Due to Total Decay Effective Half-life                                                                  :t
                                                                                            . "      The dose, dD, during an infinitesimally small time period, dt, at a time The total dose absorbed during any given time interval after the uptake                  interval t after an initial dose rate D 0 is of the iodine in the thyroid may be calculated by integrating the dose rate over the required time interval. In making this calculation, two factors must            *-                        dD  = instantaneous dose rate x          dt be considered, viz.                                                                                                    = Do e-.,1E I dt,                                    (6.20)
I. In situ radioactive decay of the isotope.                        7'~
where D0 is the dose rate at time t = 0, the instant when the isotope was taken J
: 2. Biological elimination of the isotope.                                up by the tissue; and tis the elapsed time after uptake. The total dose during
                                                                                                ~
In most instances, biological elimination follows first-order kinetics. In this          f;  the time interval T after uptake of the isotope is J                                                  f T
case, the equation for the quantity of radioisotope within an organ at any time t after uptake of a quantity Q0 is given by                                      ~ --
D    =  D0      e-AEt dt,                    (6.21)
Q  =  (Qo e--,\Rt) (e-ABI),                (6.13)                                                        D which, when integrated, yields where >.R is the radioactive decay constant, and >.a is the biological elimination constant. The two exponentials in equation (6.13) may be combined D    = Do (1      - e-Aet).                    (6.22)
                                                                                                                                            ,\E Q  =    Qo e-(AR+Asl r,                  (6.14)
For an infinitely long time, that is, when the isotope is completely gone, and, if >.E  = >.R + >.s, we have                                            (6.15)          equation (6.22) reduces to Q  =    Qo  e-.,\Er,                      (6. I 6)                                        D =Do                                          (6.23)
                                                                                                                                            ,\E
* where ,\E is called the effective elimination constant. The effective half-life It should be noted that the dose due to total decay is merely equal to the pro-then is duct of the initial dose rate, D 0 , and the average life of the radioisotope within TE= 0.693_                          (6.17)          the organ, 1/>.E. For the case in Example 6.5, the total absorbed dose during
                                                          ,\E                                        the first 5 days after deposition of the radioiodine in the thyroid is, according to equation (6.22), and after substituting ,\E = 0.693/TE, From the relationship among        ,\E, ,\R, and As, we have 1
                                            - = -
I
                                                          +-,  1                    (6.18)
D  =  31.9 rad/hr  x  24 hr/day 0.693 x    7.7 day (i _  exp _ 0.693 7.7 x  5)
TE      TR      Ts or
                                                                                                              = 3090 rads, TE= TR X Ts_                            (6. 19)        and the dose due to complete decay is, from equation (6.22),
TR+ Ts D = 31.9 rad/hr x 24 hr/day            X  7.7 day  = 8520 rads.
For 1311, TR= 8 days and Ts, the biological half-life in the thyroid, is 180 days.                                        0.693 The effective half-life, therefore, is
-I                                                                                                  Gamma Emitters
*c:i                                TE 8 X 180
                                        = --- =              7.7 days,                                For a uniformly distributed gamma emitting isotope, the dose rate at any
.....                                      8  + 180                                                point P due to-the isotope in the infinitesimal volume d V at any other point and the effective elimination constant is                                                  at a distance r from point p, as shown in Fig. 6.7, is
:Il                                                                                                                                            e-llr CD                                                                                                                            dD = C I ' - dVrad/hr,                            (6.24)
  <                                  ,\E =  0 693
                                              *      =  0.09 day- 1 *                                                                            ,2 0                                            7.7
 
320                INTRODUCTION TO HEALTH PHYSICS                                                      INTERNAL RADIATION PROTECTION                            321 using the half-face respirator, the wearer must try to eliminate possible  degree of hazard from surface contamination is strongly dependent on the leaks around the face piece. Half-face respirators are considered suit-    degree to which the contaminant is fixed to the surface.
able for air-borne dust concentrations up to ten (10) times the recom-        Dealing with loose surface contamination limits is not as straightforward mended maximum atmospheric concentration; respirators with full            as dealing with contamination of air and water. In the case of air and water face masks are considered suitable up to fifty (50) times the recom-      contamination, safety standards can be easily set-at least in theory-on the mended maximum atmospheric concentration.                                  basis of recommended maximum absorbed doses (Dose = 5 (N-18), etc.).
: 2. Supplied air masks that may be used either against dusts or gases or        Using these criteria, we can calculate maximum permissible body burdens for both. In this category we have two subdivisions: (a) air line hoods,      each of the radioisotopes. From the calculated body burden we go one step which utilize uncontaminated air, under positive (with respect to the      further from the basic radiation safety criteria, and compute maximum con-atmosphere) pressure supplied from a remote source, and (b) self-          centrations in air and water which, if continuously inhaled or ingested, would contained breathing apparatus, in which breathing air is supplied either  result in a body burden less than the calculated maximum. For the case of from a bottle carried by the man, or from a cannister containing oxygen    surface contamination, we go one more step away from the basic criteria; we generating chemicals. The advantage of the supplied air device is that    try to estimate the surface contamination which, if it should be dispersed into the pressure in the breathing zone is higher than atmospheric pressure. the environment, would result in concentrations that may lead to an excessive As a consequence, all leakage is from the inside out. When using a        body burden. Thus, specification of limits for loose surface contamination is supplied air device, it is imperative to know the time limitation on the  three steps removed from the basic safety requirements.
supply of air.                                                                From the foregoing discussion, it is clear that maximum limits for surface A third type of respiratory protective device is the gas-mask. In this device* contamination cannot be fixed in the same sense as limits for the concen-contaminated air is cleared by chemicals in a cannister through which the air      tration of radionuclides in air and water. Nevertheless, it is useful to compute passes. *Because of the specific action of the chemical agents on the contam-      a number that may serve as a guide in the evaluation of the hazard to workers inant, different cannisters must be used for different gases. For this reason, as  from surface contamination, and to assist the health physicist in deciding well as for the fact that air may leak into the face-piece of a gas-mask, gas-    whether or not to require the use of special protective measures for workers masks are not recommended for use against radioactive gas.                        in contaminated areas.
On the basis of per-unit-quantity of radioactivity, inhalation is considered Surface Contamination Limits                            the most serious route of exposure. Surface contamination, therefore, is usually limited by the inhalation hazard that may arise from resuspension of Contamination of personnel and/or equipment may occur either from nor-        the contaminant. The quantitative relationship between the concentration mal operations or as a result of the breakdown of protective measures. An          of loose surface contamination and consequent atmospheric concentration exact quantitative definition of contamination that would be applicable in all    above the contaminated surface due to stirring up the surface is called the situations cannot be given. Generally, contamination means the presence of        resuspension factor, f,., and is defined by undesirable radioactivity-undesirable either in the context of health or for technical reasons, such as increased background, interference with tracer                            f,. =  atmospheric concentration, _&#xb5;.Ci/cm 3 _
(I 1.2) studies, etc. In this discussion, only the health aspects of contamination are                                surface concentration &#xb5;.Ci/cm 2 considered.                                                                      Experimental investigation of the resuspension of loose surface contamina-Surface contamination falls into two categories, fixed and loose. In the case  tion shows that the resuspension factor varies from about I0- 4 to 10-s, of fixed contamination, the radioactivity cannot be transmitted to p,:~sonnel,    depending on the conditions under which the studies were conducted. A value
~
c:o and the hazard, consequently, is that of external radiation. For fixed contam-    of 10-s is reasonable for the purpose of estimating the hazard from surface
........ ination, therefore, the degree of acceptable contamination is directly related to the external radiation dose rate. Setting a maximum limit for fixed surface contamination.
contamination thus becomes a relatively simple matter. The hazard from          Example I I .2
:::c      loose surface contamination arises mainly from the possibility of tactile (D                                                                                            Estimate the maximum surface contamination of 90 Sr dust that may be
<          transmission of the radioactive contaminant to the mouth or to the skin, or      allowed before taking special safety measures to protect personnel against 0          of resuspending the contaminant and then inhaling it. It follows that the        a contamination hazard.
 
322                    INTRODUCTION TO HEALTH PHYSICS                                                                      INTERNAL RADIATION PROTECTION                                323 The maximum atmospheric concentration of 00 Sr recommended by the                                  Various laboratories and nuclear installations have set their own limits ICRP is 2 x 10-10 &#xb5;Ci/cm 3
* Using a value of 1o-s cm- 1 for the resuspension                      for contamination of personnel, equipment, and protective clothing. Tables factor in equation (11.2), we have                                                                  11.2 and 11.3 are given to illustrate some of the contamination standards maintained by several large users of radioisotopes.
                                        = x                                                        10        3 10-s cm- 1                - -&#xb5;Ci/cm surface concentration TAULE 11.3. U.S.S.R. SURFACE CONTAMINATION LIMITS surface concentration      =  2  x 10-4 &#xb5;Ci/cm 2*                    I:
I It should be emphasized that a figure for loose surface contamination                                                                    Contamination from 150 cm 2 in 1 min calculated by the method of Example 11.2 is intended only as a guide. In any Alpha particles              Beta particles particular case, the health physicist may, at his discretion, and depending on                          Object of contamination the nature of the operation, the degree of ventilation, and other relevant                                                              Before        After      Before        After factors, insist on more or less stringent requirements for surface contamina-                                                          cleaning      cleaning    cleaning      cleaning tion before requiring the use of protective devices for the worker.                                Hands                                  75          bg          5000            bg Special linens and towels              75          bg          5000            bg TABLE 11.2. UNITED KINGDOM ATOMIC ENERGY AUTHORITY MAXIMUM PERMISSIBLE                        Cotton special work clothes            500            100      25,000          5000 LEVELS OF SURFACE CONTAMINATION, &#xb5;Ci/cm 2<*>                                "Pellicular" clothing                  500          200      25,000        10,000 Gloves, outside I  Special* shoes, outside 500 500 100 200 25,000 25,000 5000 5000 I
Type of surface            Principal alpha        Low toxicity    Beta emitters        Work surfaces and equipment            500          200      25,000          5000 emitters th1      alpha emitters<*>
10-5 I
Inactive and low activity areas                                  10-*              10-*                Note. No contamination of the body is permitted.
Active areas                                10-*                10- 3            10-3 10-5                10-*              10-*                (From Sanilary Regulations/or Work wilh Radioactive Substances and Sources of Ionizing Personal clothing                                                                                  Radiation, Ministry of Health, U.S.S.R., Moscow, 1960.)
Authority clothing not normally worn in active areas Skin 10-*
10-5 10- 3 10-5 10-3 10-*
f Waste Disposal ti    Proper collection and disposal of radioactive waste is an inherent part of The contamination of surfaces by radioactive materials may give rise to external radia-I tion or to intake of radioactive materials by persons. The control of surface contamination        contamination control and internal radiation protection. In one sense, we is, therefore, an essential part of the safe handling of radioactive materials. Surface con-        cannot dispose of radioactive waste. All other types (non-radioactive) of toxic tamination should be controlled to the lowest practicable levels, and in any case, within the maximum permissible levels specified above, unless relaxations (e.g. for firmly fixed          wastes can be treated either chemically, physically, or biologically in such a contamination or low toxicity contaminants) have been permitted on the advice of the                manner as to reduce their toxicity. In the case of radioactive wastes, on the health physicist. The requirements of the maximum permissible doses from external                  other hand, nothing can be done to decrease the radioactivity, and hence, the radiation must be observed.
inherent toxicity of the waste. The only means of ultimate disposal is time-
  <*> Averaging is permitted over inanimate areas of up to 300 cm2 or, for floors, walls, and ceilings, 1000 cm2
* Averaging is permitted over 100 cm 2 for skin; for the hands,      to allow for the decay of the radioactivity.
over the whole area of one hand, nominally 300 cm2
* Radioactive wastes originate from any operation in which radioisotopes are
  <b> All alpha emitters other than those listed in note (c).                                      used or produced. For purposes of management and treatment, wastes may be
  <*> Uranium isotopes.
Enriched and depleted uranium.                                                              classified as high, intermediate, and low level. Low-level wastes are defined as Natural uranium.                                                                            those that must be diluted by a factor of no more than 10 3 before discharge Natural thorium.                                                                            into the environment, for intermediate levels, 10 3 < DF ::;; 105, and for Short-lived radionuclides, such as radon daughters.
Thorium-232.                                                                                high-level waste, t.he required dilution factor would be greater than 10 5
* Thorium-228 and thorium-230 when diluted to a specific activity of the same order          High-level wastes are associated with the inventory of fission products in the as that of natural uranium and natural thorium.                                            burned-up fuel of nuclear reactors and with the chemical and metallurgical (From the U.K.A.E.A. Health and Safety Code, Maximum Permissihle Doses from                    processes involved in the separation of the fission products from the unspent Inhaled and Ingested Radioactive Materials, Authority Code No. E.1.2, Issue No. l, London, June 1961.)                                                                                uranium or plutonium in the burned-up fuel.
TAB 11, Rev. 0
 
APPENDIX        C      EFFECTIVE HALF-LIVES Radionuclide        To              T,          T.
aH                19 d          12.26 y      19 d
            'Be              400d 0          53 d        47 d uc                  35 d      5770y            35 d
          '"F                140 d'            1.85 h      1.8 h 24 Na              19 d          !5h          15 h a2p              1200d            14.3 d      14 d
          **K                  37 d      1.3 X 109 y      37 d "Ca                  soy          165 d        165 d 51 Cr 110 d          27.8 d      22 d
          ,eyv                50 d 0        16.1 d      12 d
          **zn    (23 d')  933 d          245 d        111 d 72 Ga                8.3 y*      14.1 h      14 h
          **sr                11 y'        28 y            7.9 y
          **y                500 d'          64.2 h        64 h
          **zr              180 d'          65 d          48 d
          **Nb                50 d 0        35 d          21 d
          **Mo              150 d 0          2.75 d        2.7 d
        '""Ru                20d            I.Oy        19 d nasn                149 d 0      118 d          66d 131J                138 d'            8.05 d        7.6 d ia1cs                17 d          30 y          17 d u*Ba                200 d          12.8 d        12 d 140La                  35 d 0        40.2 h          1.6 d 1uce                500d 0        285 d        180 d uapr                  50 d 0        13.7 d        11 d 147 Nd 35 d 0        11.1 d        8.4 d u1pm                1-00 d 0          2.5 y      90d
        '"'Sm                11oy*          90y          soy 154 Eu                4 y*        16 y          3.2 y
        ""Ho                  37 d''        27 h            1.1 d 170 Tm              110 d 0      127 d          59 d 177 Lu                6d            6.8 d        3.2 d IBIW 5 d'      130d            4.8 d 21opb                531 d          21y            1.4 y 21*po                  57 d        138.4 d        40 d
      ""Ra + 55 % da        55 y      1620 y          53 y 22 'Ac                33 y*        21.2 y        13y
      ..,Th-*a< Pa        110 y*          24.1 d      24 d 2330      (30 d)    300 d      1.62 X 105 y    300d ea*pu Sol./bone        120 y      2.44 x JO' y    119 y lnsol./lungs      1.0 y    2.44 X 10' y      1.0 y
      "'Am                890 d'        458 y        890 d 242 Cm 600 d"        163 d        130 d
        &deg; From M. Errera and A. Forssberg, eds., Mechanisms in Radiobiology, Vol. 2, Academic Press, New York, 1960; and K. Z. Morgan and J. E. Turner, Principles of Rad-iation Protection, John Wiley & Sons, Inc., New York, 1967.
* Critical organ.
TAB 12, Rev. O
 
i~3
[_QQ1l)IL-A0!
    .-'3~
1e-'      8 v\v'  6 lz_a    C
  &#xb5;P
      \    3
          ,p H-+-+--+-L;+l-h-1-.;.;#m--lf-LL--l 9 l->-+-'-~~-i-;.-+-4--4--1..4-1-;-<+;...4-;..;..J..j../....C 5
2    3      5 6  7 8 9 10 c.c.?(Y).
(_) S / I\) (-,  A            1u r)'h_f C,AS  l)I A Lo
                                                                                              ,./I' ,;l1'-'
Jhi:
LOGARITHMIC 3 CYCLE;:S BY 3 CYCLES                                                                              TAB 13, Rev. 0 CONTROL COMPANY
 
DAVID E. HAMERSKI, PH.0.
(            PHYSICIST TELEPHONE (!507) AIS2-32!54 1666 EDGEWOOD ROAD WINONA. MINNESOTA  5!5987 To:  Don Schuelke                                        11/11/81 NSP Prairie Island Nuclear Plant RR #2 Welch, MN          55089 From:  D, Hamerski
 
==Subject:==
CONTAINMENT HIGH-RANGE RADIATION MONITOR DOSE ASSESSMENT FACTORS BASED UPON Xe-133 AND*I-131 EQUIVALENTS
 
==Background:==
The calculated data presented here are based upon the following assumptions:
(1)    That the monitor views a large, uniform fraction of the containment volume.
(2)    That the monitor is not placed in an area which is protected by massive shielding.
(3)    The source terms consist of the OWn~s Group source terms-- 100\ noble gases, 50\ iodines.
(4)    That the monitor is basically a 3 inch diameter ion chamber.
(5)    That the net free containment volume is 3.73xlo 10 cc.
This report supercedes all prior reports.
TAB 14, Rev. 0
 
E. HAMERSKI, PH.0.
PHYSICIST Results:  I. Xe-133 Equivalent Data The dose-factor for Xe-133 equivalent was found to he:
DF = l.2xlo 4 (mr/hr)/(pCi/cc)
The curve of Xe-133 equivalent versus time is given as Figure 1 and Figure 2.
II. I-131 Equivalent Data (a) The dose-factor for I-131 equivalent versus time is given as Figure 3.
(b) The curve of I-131 equivalent versus time is given as Figure 4.
TAB 14, Rev. 0
 
(Houa.s)
                                        */-~_._~-'-'---:-'-48==
        ---'-'-_,:!.:..;.....;~~~-:_,:__
                                                              =- ,_-1_                _;_,_01~_=:,_;_:~'~~-:
: ;:1 I lfc'f-*~!!~- .'f'>t:-f( ,:'I==~:~:k  I            :* ! :* :                                              l 1--I---'*-'-*..;.*-+-.;_:_-'-__.-_ _ _ -----                                  -->--'-.-_-                __-_--.--+----..                          i        : :
                                                                                                                                                    ---; __;_ __ ;_ __ i                                    r--:-~
t--l-------------------~-+-1:...~-_;_-=-_--;_.L..;.1_;_.-1-'_;___;_..;._-+----,------t-1                                                                                                                : . _-_ _;..__ _
        --+------"----------- ------+-.-+--r------+--                                            __ 4 ___ ______ ;      !                              -----+-------i-------r--                            I 1
    ---*-------- - ----*- ----- - -**                                                                                                      *------------------!--: - I                                                    L_____ ._
                                                                                                                                                                                                                            ''      .      ''                      - . l I                                  ;                                                                                                                                                                  '
I : :**
I                                                                                                                                                                                      i
    -'--r--
I
                                                      ---l-----------*-------+-----'------+--                                                                                                                    - --r----
1 i                                          ------*- --------- --*-*7                                                                : * :- : l : : :.'. I : : :
* j
                                                                                                                                        -------- --** -- *-*- *-1--:*-_-:-"!~~--:-*--,
1
                                                                                                                                                                                                                      . : i*:_ '. -- .
:I
* j
                    !                                                                                                                                    . ' .            . ., ,_            -, * .        t                                            i-:
                                                                                                                                                -~ ~-~~~~-L * .                              i              i-- -- __ _;_ ______ .                        1 ***
I                                                                  I                I              1                -
i;                                                                    I
                                                                                                                                                                                            'i i
I
                                                                                                                                                                                            'rI                                                    .. 1 I
                                                                                                                                                                                                                                                      . Ir I
I
                                                                                                                                                                                                                                                    *-c+----J
        ,. --,      I'    ---------**
I i                                                                                                                                                                      '
        . .l----- ------ .-
("
j I
i-*
TT--
      "'"11 ____,_                                                            --
i
                                                                                            'i
______ j_ _________ ..
                                                                                                                                  ---i..  ----
i
                                                                                                                                                                    -*-**--**-**-*--+--------
                                                                                                                                                                      .                  i
                                                                                                                                      '                                            : i :                                                  I iI
* j ;                                                  !
i
                                                                                      ... i -- --- .                                                                            ---,-i-' . '**                            t-**--*----*i I
                                                                                                                                                                                          !                                l              I i
                                                                                                                                                                                                                          -              I
            .., __ *-          *-* ----- -----.                                                                                                                      L._. _ _ _L__ ._
                                                                                                                                                                                                                        -;- -: - -f --*--
            ~-+------ --- : : : : --------i-                              ----JI.
                                                                                                                                                                      \
                                                                                                                                                                              ----      t
                                                                                                                                                                                                                          '              -i_- ,------+---*-----<
                                                                                                                                                                                                                                                            - i-.
                                                                                                                                                                                                                                        .i              ' *t ;
_T___
                              --,----*** *-+--- __ _;_ ;___
                                                                                                *::ti, :i-iME- _, -: : : -__
_;_-1-_-_""_-_~-+-;_-*_*-_s'-'-1-~-~~_____: _ __:__
r,,_----+---                                      -~----;
: 1- i
                                                                                                                                                                                                                                                          --+- **
I-,-*-
                                                                                                  !~ ;.:~l~:i-i *; . : . ----- - -. -__I__:* :-*7T-:.; -                                                                        -* -                  *~=:-~
                                                                                              - , - i---,- *t.
r            I, ~                                  i    i: 1' I
I                .. ' -. . ,-
                                                                                    ,-+-'--i:. .!--~--:~-,-!_-+-. -.:-;-*- --*------.                ,                            , ;'. !                ;1' -          r-~:-::r:::-:*1*, --- : : :T~:::":
                                                              --------+--'-:.,..-~_-':-'_ -:...,_,--_;__;__;__;_-+_-j--~---*--+-r,                                              ,_;~ ~:~:::            --i-~-- . I:.~: ~-:-i~_: . ~-~ 1-:::**---
::.t:.                                                                : 1=-l=- =r-:-:*=-;            i ;
* I :.
i                                                                                    r:. .
                                                                  ... ---*-----**-+--~--....._r-._:;--1_ . :                        -;        T:  "1!7T i::;:                          tr I  1
                                                                                                                                                                                                                        ~
l TAB 14, Re*v: 0
                                                                                                                                                                                                                                                      ;_n_; ::
    ,QJ    '/ ; ___ __ :___ * *1 -                                                                -H*            1      j"                              i!i          ,*1*1*--i-:-1 I                          .    ,,11;
                                                                  -----'--'---l.-'---_._i-~1--'1--....-..:..i-_:_-'-'-'----"-1-'--'1_*-_!-"""";~----'-'_:_1_-.....-....-...1_-i~*-1_L_-__-__;___J1_--_:_1-..l.1...J.-i_-r:_*....--'-1~'--'---'--'--1-...........
\
111.,    :L,      i:,,:I
                                                      '                                .:l.                                    J                                'f                                  s-                            '                                ?
(u,,..,.,.,.\
 
:_~1-r: l '. *(~_;--=[~~~---
I  .          ,    - --
              "7--*--*-:*-:* 1*-*-'-*- **; .                                                                                                                                                                                                            r a _ __,______ ; _________ _                                                                                                                                                    __. -*+--- -. ---
              ,_J___ _                                                              -*
                                                                                          ---- *--~-------*--*-  . . ' .
I    *    *          '
I
__ J ___ ---                                        '
I                                                                      . . .          '                    .
( __ ! ...                                                            . *... _:_ *_; . _I_. __ . -' -*******                                                                                                        --~------- ---l---          *-*----*-----t
                                                                                          .              I , ...          i                      !
* I
              . *---y- ------------ -----.                                                  --------*--*-*. --**                                                                                                                ----*** -*--* -'*-----...-- *--,
                                                                                                                                                                                                                                                                                                                .I
______ !_________ :__ _--*-----                          ~
I I
                                                                                                                                                                                                                                    - -------*-"-            -----~---~
I I
l:
                                                                                                                                                                                                                                                                        .... .... j 1
                                                                                  . --- ____ __. ---~---- -- ~ ***- -                                                                    * - --- --------      - L      - -*- ------ ----*' -----*-**
                                                                                                                                                                                                                                                                        .. -. 'i ..' -.
* Xe-133 EQUIVALEN'I'.                                                                        .              - . . .            .
CURIES IN CORE-
* I vs.
TIME*
:;  ___ _                            -- ---r-----                              ... ii . --:
                                                                                                -T:*.-.*-***-*                    -
I i      *
                    --        ,-i  .                                                        -- l-----***-***---****
J __
j
                                                                                                                                                                                                                                                                          -----i I
                                                                                                                                                                                                                                                                  -**--*- *. _-                ~    -____ t
                                                                                                                                                                                                                                                                                                -*: .. 1
______: _-* J i
i-* --              :-      I
                                                                                                -+-*
I I
                                                                                                                                                                                                                                                                      .  --- *-          ------*        :i I
l..                                                                                                                                                                                                          I!
i                                                                                                                                                                                                              I s:                  i 10~                    r. I          -*                    *-- : i__                        : ~~----
                        ;-l **- : : : _I : *-                                    ~ - - : :--+***-* . *-                                    I
                      -~
                      --1----+---      !;                                                          i ' : 1 ; i : _:_ __ L                                                                                                                                                                                  ---1
                                                                                                  .!                                        i 7          --    . _)_, __                                      --<--~--.. .--..-----* -                            I
                                                                                                                                            *-t-*-    - - - - - - - - - - - --- --
6            --------*- *-***-****--**    I 1      1
                                                                                        --'1~.----"'":
                                                                                                                                          . r ..
I
                    ----~------              '                                  ---+--------+-----
I I                                                                                                              '
i
                                                ---------        -------  -                                                                                                          --***----.                        -~*-***--***                :*  *--- . . . .                        - *-.**:
I                                I I                                '                                                                                                                                                                      .            . ,                                      .              . l I                                                                                                                                                                                                        ; ~--* -~*---~- *i      ~ ~    :--~ !* ~- .: -~ -:--1 I          . ,                                                                                                                                                                                          !                I
                                                                                                                                                                                                                                                            *      *
* I                *    *
                    ?    7*--:1**--*                                                                                              ;
                                                                                                                                  ' -  ;                                                    : '. :. i
                                                                                                                                                                                            .. , I '
I II t-.
t . ** . -*
                                                                                                                                                                                                                                                                '**i.ifi j- '
i                      ,                                                                                                                                                                t I
1* - . - __ ,. ;-*
I
                                                                                                                    **1*------                                                            -- ,--------  ..                                                                                            -+-
                                                                              ---*---r--:          ---.
1                                                                                            I    :    I TAB 14, Rev. 0 i                                                                                                                                                                        I
                                                                                                          " .: !: I                                                        .
I    t
                                          ! I ..
__,.___J_:_ _
                                                          *-'13            I i , : I j ; : iJ : ' : :                                                                    I        I ,* :    , :.        .    :  1' ;    *:i* I*;
                                                                                                                                                                                                                            ;_**                                  !-J -                          -f
--., rni-J.n~.111t lnni,*
I 'J.
I'{
IS-L.!__l_          .
If#
t"    ;    :
                                                                                                                          - ~ - - - - ------- -- . --* - .
I **
17 II
                                                                                                                                                                                    ~
2./          , ,.
                                                                                                                                                                                        -~~_Ll__: ~.- ---- . . . : . . ~ - - -......'--'-'-'_._'............~'--*......
 
                                        -                        +t - i t
                                                                                              -s-
                                                                                                                                              --r t-I
                                            -----------o-r                                                                                                    I
                                                          ,- -i--t-1 r*r-
* i*-- ,--*t* -      j___
t - r
                                                          !' -*--ij-I tT                                  !  :                -t- 1-r i-;; - t-r-+-t-
                                                                                                                                            . t --~-~-
                                                                -i-r,-r**i
                                    '1                    !I_  --* -+i /_.!          j*                                                      ! ' '          1--        - i--t --r-
                    .- - ' ;        i                          ---    1, j
f,'    i -1                                                      !        lI  !:. 1i - --- -rr  Tr-
    *            --t I      I i
                          -1*---*    ,
i,,_
                                                                -r-T r**r*-1 1--~~----                                                    ; i I        ;
i.
                                                                ---- r-- , j" -1 t--,        t  j l    J                                                        '  i
                                                                      .u                                                                                i.
      )(1o'f                                                          l-H-r r-** t .
i-1                    j r **      1                          I 1Gi  UR      E    3 r*n :*n g.o
                  - ... t
      ?.O      +--------+-*--* - ----
t    !
l-+------+--,--------1--"--------+------------*                                        - - - - - - - - - - -- ---- - - - - - - - +
r s-.o                                        ---+--"""T""""----+-------- ----- -- ----- --------------                --------~--------!
~ \ ~-0 3,0 i
r    1
                                                                                                                                                              ' .. *t . :
                -* i-i !                                                                                                              '  .
r-r I
:z.o                                                    - -+-
I          ;  '
                                                                                  ' : I I *t
                                                                                                                                                                        -+--..-
                                                                - Ii-;---+
                                                              *ri*                                                                                                , -l-
                                                                          +.
I
:--J-~ 1                                                                                    i    l
                                                                                                                                                                                ' I I  i
                                                                    -f                    i
                                                              -r-r1 +
                                                              *rl**r
                                                                -*- i    --tI  -f--
                                                                                      .--1 1*-; :
I i                                                    !
i '
I l .o                                                          I          I  '  .
                                                            --TI j_i_                                                                        i1**:
i          I i--
__ j .  '
T            1 '                                                                                                  I
                                                                  !                                                                    TAB 14,              Rev.O j
                                                                            ;                                                                                                    I-
                                                                                                                                                              . : ; _!_J i I          I
                ~~~~-~--                - - - - - - - - - - L ~ - - - - - - * - - - - - - - ------"-------- ------- . ____.,:__.:..__~I-----'--'-...............
                              ~  3 'f                                                            I ft, I*, llh  l11d1
 
                                                                                                      -----~----****-*****--~--~---~--~--            l ... J__ !- ' :
                                                      . f-** 1 -Li-r1
                                                        ;+-l-1*1                        i
                                                        -t7-rt-*-*;*
:--1 -r-i---t --~
i*-*1--/* -t*----l ---l--
                                                        .. - ,-- r***    t        -                                                              1 f j                      !          I                            \
          ' - - - ~ - - -... -*- ---*- ---- --*    --1-..-i--'-4-...--'--4---------*.                                  *---~---*                          1  it          I  ! !
                                                                                                                                                                                                    *r1I j_j _j
: ;-l
                                                                                    ---i **
                                                                                          '  I i
i
                                                                                                                                                          ! '        1 I
1*
I
:-H-t-I i-
                                                                                                                                                                                            *1 ***
                                                                                                                                                                                                      -,-r
                                                                                                                                                                                                    - *. )
                                                                                                                                                                                                      \
a                                                            **r                                                                                      !
                                                                                                                                                                                  ,  'i 7 *7 -,**
i t '  t  *I X IO                                                                    t I
                                                                                                            . F I G U R 'E                *4 I    ~                '        l  *l--l ,
i '                                                              I    t                **  '**,
I : ; I I :
r*h
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I    '
* 1*-- +---, -**t  -
i r 1 !
3.0    a - - - - - - - - - - - * * ..        ---<  ---+-~-~f--'-----I-1 1---EQUIVALEN                                                                                                              !    !
t*.I 1i        ;                                                                                                        i
:  ;    t                                                                    !
lJ:;.-: .
I    '
                                                          , . I    ~-  -              ---+
                                                        ,_J_;            I    !      L
                                                                                                                                                                                      .r-J._l *f--1-! ~
:    :            i    :        J                              . vs .*.                              [    I
                                                        ... ; ** lr:rr 1
t l .
                                                                                                .                                                              I    I I,+f i*
t TIME' '
i ..
                                            --**--****--4-----'--'---+--*---**-****1-** --
ii        r~
I
                                                                                                                                                                                                .l I
                                                              !    !    ?      f
                                                        --- +-    f            "i'
                                                            *t:-j      _j      t*                                                                                                              '!
!I                              --***--*---------1----'-.;__;--+----
i i
I I
i
                                                                                                                  **-, ..*. ***--*--*. -*- t .* -----
                                                                                                                    !                            I I
  ,.o                                      *----!---.-----+----- --****,-----------                                !
1 i
I                                                                              i
                                                            -:~j : .                                                                              I                                    . i
                                                            . : *r*-~
                                                              '            I      j
                                                                                            .-                                                  I                                          ;**  f I
                                                --~*--1---;              l
                                                                ~-1_1_:~-......,-l,...*- ~ - - - - - - - - ________          - - - - - - - -
* ___ ... _;___ _-1-_ _ _                      t*..,--r_~
I    ;
: -i-- -+-Lt' . l.
                                                      *r -** * -Ft~r
* j-            ~
I    I . '
1
                                                                                            *! ... 1
    /. S"
            ;!1I
          -~J.J.;
i  I '
t-7=1f i                                                                                                  j ___ :. i
* j' l! '              I
                                                                                                                                                                                          */-+-l -l-I !        r t
1.,.I
                                                                                                                                                          ~
                                                                                                                                                              ;I-:- :
                                                                                                                                                          ;.! !*L 1*  -      l.
                                                                                                                                                                                  .. 1-r *r
* t * /--
1 1  i-lI II      -1 I I
i I  :
1    I
                                                                                                                                                                  -    - II - :' ' l -f-i_i-~::.
l - I 1*' - - ; ' :-*-i-*                    -+*
                                                                                                                                                          *. 1- _j __ .* ... -t--                        ,--
I        I1                    J          I    !    I I                                TAB 14, Rev. 0 i
j
                                                                                                                                .; *I Ij I
      /.0    *__
              *
* 1_*.:_~ -~-- --~1--~-~.L----'-~--........__ __. _                                                            I    I I
__;_...L..-L._ _I        1. . .
: 2.        '1            '        g              IO                        I l.    ''1      I~              18          '-0
 
DELETED TAB 15
 
DELETED TAB16
 
This Tab 17 has been Deleted Rev. 26 Page 1 of 1    TAB 17, Rev. 26
 
U.S. NUCLEAR REGULATORY COMMISSION                                                                    Revision 1 February 1996 REGULATORY GUIDE OFFICE OF NUCLEAR REGULATORY RESEARCH REGULATORY GUIDE 8.29 (Draft was issued as DG-8012)
INSTRUCTION CONCERNING RISKS FROM OCCUPATIONAL RADIATION EXPOSURE A. INTRODUCTION                                          dose limit for the embryo/fetus of an occupationally exposed declared pregnant woman, and explicitly Section 19.12 of 10 CFR Pan 19, "Notices, In-                              states that Part 20 is not to be construed as limiting structions and Reports to Workers: Inspection and In-                            action that may be necessary to protect health and vestigations," requires that all individuals who in the                          safety during emergencies.
course of their employment are likely to receive in a Any information collection activities mentioned in year an occupational dose in excess of 100 mrem (1 this regulatory guide are contained as requirements in mSv) be instructed in the health protection issues asso-10 CFR Part 19 or 10 CFR Part 20. These regulations ciated with exposure to radioactive materials or radi-provide the regulatory bases for this guide. The infor-ation. Section 20.1206 of 10 CFR Part 20, "Standards mation collection requirements in 10 CFR Parts 19 and for Protection Against Radiation," requires that before 20 have been cleared under 0MB Clearance Nos.
a planned special exposure occurs the individuals in-3150-0044 and 3150-0014, respectively.
volved are, among other things, to be informed of the estimated doses and associated risks.
B. DISCUSSION This regulatory guide describes the information                                  It is important to qualify the material presented in that should be provided to workers by licensees about                            this guide with the following considerations.
health risks from occupational exposure. This revision conforms to the revision of 10 CFR Part 20 that be-                                    The coefficient used in this guide for occupational came effective on June 20, 1991, to be implemented                                radiation risk estimates, 4 x 1o- 4 health effects per by licensees no later than January 1, 1994. The revi-                              rem, is based on data obtained at much higher doses sion of 10 CFR Part 20 establishes new dose limits                                and dose rates than those encountered by workers.
based on the effective dose equivalent (EDE), requires                            The risk coefficient obtained at high doses and dose the summing of internal and external dose, establishes                            rates was reduced to account for the reduced effective-a requirement that licensees use procedures and engi-                            ness of lower doses and dose rates in producing the neering controls to the extent practicable to achieve                            stochastic effects observed in studies of exposed occupational doses and doses to members of the public                            humans.
that are as low as is reasonably achievable (ALARA),                                    The assumption of a linear extrapolation from the provides for planned special exposures, establishes a                            lowest doses at which effects are observable down to USNRC REGUIATORY GUIDES                                  Written comments may be submitted to the Rules Review and Directives Branch, DFIPS, ADM, U.S. Nuclear Regulatory Commission, Washing-Regulatory Guides are issued to describe and make available to the public        ton, DC 20555-0001.
such Information as methods acceptable to the NRC staff for implement-ing specific parts of the Commission's regulations, techniques used by          The guides are Issued in the following ten broad divisions:
the staff In evaluating specific problems or postulated accidents, and          t, Power Reactors                    6. Products data needed by the NRC staff in its review of applications for permits and      2. Research and Test Reactors        7, Transportation licenses. Regulatory guides are not substitutes for regulations, and com-      3. Fuels and Materials Facilities    8, Occupational Health pliance with them is not required. Methods and solutions different from        4. Environmental and Siting          9. Antitrust and Financial Review those set out In the guides will be acceptable If they provide a basis for the  5. Materials and Plant Protection  10. General findings requlsile to the issuance or continuance of a permit or license by the Commission.                                                                Single copies of regulatory guides may be obtained free of charge by writ-Ing the Office of Administration, Attention: Distribution and Services Section, U. s. Nuclear Regulatory Commission, Washington, DC This guide was Issued after consideration of comments received from the        20555-0001; or by fax at (301)415-2260.
public. Comments and suggestions for Improvements In these guides are          Issued guides may also be purchased from the National Technical Infor-encouraged at all times, and guides will be revised, as appropriate, to        mation Service on a standing order basis. Details on this service may be accommodate comments and to reflect new Information or experience.              obtained by writing NTIS, 5285 Port Royal Road, Springfield, VA22161.
TAB 18, Rev. 1
 
the occupational range has considerable uncertainty.                        high, i.e., above 20 rems (0.2 Sv), acute ex-The report of the Committee on the Biological Effects                      posures).
of Ionizing Radiation (Ref. 1) states that                                  The normal incidence of effects from natural and
          "... departure from linearity cannot be ex-                    manmade causes is significant. For example, approxi-cluded at low doses below the range of obser-                    mately 20% of people die from various forms of cancer vation. Such departures could be in the direc-                  whether or not they ever receive occupational expo-tion of either an increased or decreased risk.                  sure to radiation. To avoid increasing the incidence of Moreover, epidemiologic data cannot rigor-                      such biological effects, regulatory controls are imposed ously exclude the existence of a threshold in                  on occupational doses to adults and minors and on the 100 mrem dose range. Thus, the possibil-                    doses to the embryo/fetus from occupational expo-ity that there may be no risk from exposures                    sures of declared pregnant women.
comparable to external natural background                            Radiation protection training for workers who are radiation cannot be ruled out. At such low                      occupationally exposed to ionizing radiation is an es-doses and dose rates, it must be acknowl-                      sential component of any program designed to ensure edged that the lower limit of the range of un-                  compliance with NRC regulations. A clear understand-certainty in the risk estimates extends to                      ing of what is presently known about the biological zero."                                                          risks associated with exposure to radiation will result in The issue of beneficial effects from low doses, or              more effective radiation protection training and should hormesis, in cellular systems is addressed by the                      generate more interest on the part of the workers in United Nations Scientific Committee on the Effects of                  complying with radiation protection standards. In ad-Atomic Radiation (Ref. 2). UNSCEAR states that" ...                    dition, pregnant women and other occupationally ex-it would be premature to conclude that cellular adap-                  posed workers should have available to them relevant tive responses could convey possible beneficial effects                information on radiation risks to enable them to make to the organism that would outweigh the detrimental                    informed decisions regarding the acceptance of these effects of exposures to low doses of low-LET                          risks. It is intended that workers who receive this in-radiation."                                                            struction will develop respect for the risks involved, rather than excessive fear or indifference.
In the absence of scientific certainty regarding the relationship between low doses and health effects, and                              C. REGULATORY POSITION as a conservative assumption for radiation protection Instruction to workers performed in compliance purposes, the scientific community generally assumes with 10 CFR 19.12 should be given prior to occupa-that any exposure to ionizing radiation can cause bio-tional exposure and periodically thereafter. The fre-logical effects that may be harmful to the exposed per-son and that the magnitude or probability of these ef-                quency of retraining might range from annually for li-censees with complex operations such as nuclear fects is directly proportional to the dose. These effects may be classified into three categories:                              power plants, to every three years for licensees who possess, for example, only low-activity sealed sources.
Somatic Effects: Physical effects occurring in                  If a worker is to participate in a planned special expo-the exposed person. These effects may be ob-                    sure, the worker should be informed of the associated servable after a large or acute dose (e.g., 100                risks in compliance with 10 CFR 20.1206.
rems 1 (1 Sv) or more to the whole body in a In providing instruction concerning health protec-few hours); or they may be effects such as tion problems associated with exposure to radiation, all cancer that may occur years after exposure to occupationally exposed workers and their supervisors radiation.
should be given specific instruction on the risk of bio-Genetic Effects: Abnormalities that may oc-                    logical effects resulting from exposure to radiation.
cur in the future children of exposed individu-                The extent of these instructions should be commensu-als and in subsequent generations (genetic ef-                  rate with the radiological risks present in the work-fects exceeding normal incidence have not                      place.
been observed in any of the studies of human                          The instruction should be presented orally, in populations).                                                  printed form, or in any other effective communication Teratogenic Effects: Effects such as cancer or                  media to workers and supervisors. The appendix to congenital malformation that may be ob-                        this guide provides useful information for demonstrat-served in children who were exposed during                      ing compliance with the training requirements in 10 the fetal and embryonic stages of develop-                      CPR Parts 19 and 20. Individuals should be given an ment (these effects have been observed from                    opportunity to discuss the information and to ask ques-tions. Testing is recommended, and each trainee 1 In the International System of Units (SI), the rem is replaced by    should be asked to acknowledge in writing that the in-the sievert; 100 rems is equal to 1 sievert (Sv).                    struction has been received and understood.
8.29-2 TAB 18, Rev. 1
 
D. IMPLEMENTATION                            complying with specified portions of the Commission's The purpose of this section is to provide informa-      regulations, the guidance and instructional materials in tion to applicants and licensees regarding the NRC          this guide will be used in the evaluation of applications staff's plans for using this regulatory guide.              for new licenses, license renewals, and license amend-Except in those cases in which an applicant or li-      ments and for evaluating compliance with 10 CFR censee proposes acceptable alternative methods for          19.12 and 10 CFR Part 20.
REFERENCES
: 1. National Research Council, Health Effects of Ex-      2. United Nations Scientific Committee on the Ef-posure to Low Levels of Ionizing Radiation, Re-              fects of Atomic Radiation (UNSCEAR), Sources port of the Committee on the Biological Effects of          and Effects of Ionizing Radiation, United Na-Ionizing Radiation (BEIR V), National Academy                tions, New York, 1993.
Press, Washington, DC, 1990.
8.29-3 TAB 18, Rev. 1
 
APPENDIX INSTRUCTION CONCERNING RISKS FROM OCCUPATIONAL RADIATION EXPOSURE This instructional material is intended to provide          The basic unit for measuring absorbed radiation is the user with the best available information about the      the rad. One rad (0.01 gray in the International Sys-health risks from occupational exposure to ionizing ra-      tem of units) equals the absorption of 100 ergs (a small diation. Ionizing radiation consists of energy or small      but measurable amount of energy) in a gram of materi-particles, such as gamma rays and beta and alpha par-        al such as tissue exposed to radiation. To reflect bio-ticles, emitted from radioactive materials, which can        logical risk, rads must be converted to rems. The new cause chemical or physical damage when they deposit          international unit is the sievert (100 rems= 1 Sv). This energy in living tissue. A question and answer format is    conversion accounts for the differences in the effec-used. Many of the questions or subjects were devel-          tiveness of different types of radiation in causing dam-oped by the NRC staff in consultation with workers,          age. The rem is used to estimate biological risk. For union representatives, and licensee representatives ex-      beta and gamma radiation, a rem is considered equal perienced in radiation protection training.                to a rad.
This Revision 1 to Regulatory Guide 8.29 updates        2. What are the possible health effects or expo-the material in the original guide on biological effects          sure to radiation?
and risks and on typical occupational exposure. Addi-Health effects from exposure to radiation range tionally, it conforms to the revised 10 CFR Part 20, from no effect at all to death, including diseases such "Standards for Protection Against Radiation," which as leukemia or bone, breast, and lung cancer. Very was required to be implemented by licensees no later high (100s of rads), short-term doses of radiation have than January 1, 1994. The information in this appen-been known to cause prompt (or early) effects, such as dix. is intended to help develop respect by workers for vomiting and diarrhea, 1 skin burns, cataracts, and the risks associated with radiation, rather than unjusti-even death. It is suspected that radiation exposure may fied fear or lack of concern. Additional guidance con-be linked to the potential for genetic effects in the chil-cerning other topics in radiation protection training is dren of exposed parents. Also, children who were ex-provided in other NRC regulatory guides.
posed to high doses (20 or more rads) of radiation
: 1. What is meant by health risk?                            prior to birth (as an embryo/fetus) have shown an in-creased risk of mental retardation and other congenital A health risk is generally thought of as something    malformations. These effects (with the exception of that may endanger health. Scientists consider health        genetic effects) have been observed in various studies risk to be the statistical probability or mathematical      of medical radiologists, uranium miners, radium work-chance that personal injury, illness, or death may re-      ers, radiotherapy patients, and the people exposed to sult from some action. Most people do not think about        radiation from atomic bombs dropped on Japan. In health risks in terms of mathematics. Instead, most of      addition, radiation effects studies with laboratory ani-us consider the health risk of a particular action in        mals, in which the animals were given relatively high terms of whether we believe that particular action will,      doses, have provided extensive data on radiation-in-or will not, cause us some harm. The intent of this ap-      duced health effects, including genetic effects.
pendix is to provide estimates of, and explain the bases for, the risk of injury, illness, or death from occupa-            It is important to note that these kinds of health tional radiation exposure. Risk can be quantified in        effects result from high doses, compared to occupa-terms of the probability of a health effect per unit of      tional levels, delivered over a relatively short period of dose received.                                              time.
When x-rays, gamma rays, and ionizing particles              Although studies have not shown a consistent interact with living materials such as our bodies, they      cause-and-effect relationship between current levels of may deposit enough energy to cause biological dam-          occupational radiation exposure and biological effects, age. Radiation can cause several different types of          it is prudent from a worker protection perspective to events such as the very small physical displacement of      assume that some effects may occur.
molecules, changing a molecule to a different form, or ionization, which is the removal of electrons from atoms and molecules. When the quantity of radiation 1These  symptoms are early indicators of what is referred to as energy deposited in living tissue is high enough, biolog-      the acute radiation syndrome, caused by high doses delivered ical damage can occur as a result of chemical bonds          over a short time penod, which includes damage to the blood-being broken and cells being damaged or killed. These        forming organs such as bone marrow, damaJe to the gastroin-testinal system, and, at very high doses, can mclude damage to effects can result in observable clinical symptoms.          the central nervous system.
8.29-4 TAB 18, Rev. 1
: 3. What is meant by early effects and delayed                normal healthy cells turn into cancer cells. The poten-or late effects?                                          tial for these delayed health effects is one of the main concerns addressed when setting limits on occupation-EARLY EFFECTS                                                al doses.
Early effects, which are also called immediate or              A delayed effect of special interest is genetic ef*
prompt effects, are those that occur shortly after a          fects. Genetic effects may occur if there is radiation large exposure that is delivered within hours to a few        damage to the cells of the gonads (sperm or eggs).
days. They are observable after receiving a very large        These effects may show up as genetic defects in the dose in a short period of time, for example, 300 rads          children of the exposed individual and succeeding gen-(3 Gy) received within a few minutes to a few days.          erations. However, if any genetic effects (i.e., effects Early effects are not caused at the levels of radiation      in addition to the normal expected number) have been exposure allowed under the NRC's occupational limits.        caused by radiation, the numbers are too small to have Early effects occur when the radiation dose is large    been observed in human populations exposed to radi*
enough to cause extensive biological damage to cells so      ation. For example, the atomic bomb survivors (from that large numbers of cells are killed. For early effects    Hiroshima and Nagasaki) have not shown any signifi-to occur, this radiation dose must be received within a      cant radiation-related increases in genetic defects shon time period. This type of dose is called an acute        (Ref. 3). Effects have been observed in animal studies dose or acute exposure. The same dose received over a        conducted at very high levels of exposure and it is long time period would not cause the same effect. Our          known that radiation can cause changes in the genes in body's natural biological processes are constantly re-        cells of the human body. However, it is believed that pairing damaged cells and replacing dead cells; if the        by maintaining worker exposures below the NRC limits cell damage is spread over time, our body is capable of        and consistent with ALARA, a margin of safety is pro-repairing or replacing some of the damaged cells, re-          vided such that the risk of genetic effects is almost ducing the observable adverse conditions.                      eliminated.
For example, a dose to the whole body of about            4. What is the difference between acute and 300-500 rads (3-5 Gy), more than 60 times the annu-                chronic radiation dose?
al occupational dose limit, if received within a short              Acute radiation dose usually refers to a large dose time period (e.g., a few hours) will cause vomiting and        of radiation received in a short period of time. Chronic diarrhea within a few hours; loss of hair, fever, and          dose refers to the sum of small doses received repeat*
weight loss within a few weeks; and about a 50 percent        edly over long time periods, for example, 20 mrem (or chance of death if medical treatment is not provided.        millirem, which is 1-th.ousandth of a rem) (0.2 mSv)
These effects would not occur if the same dose were          per week every week for several years. It is assumed accumulated gradually over many weeks or months              for radiation protection purposes that any radiation (Refs. 1 and 2). Thus, one of the justifications for es-      dose, either acute or chronic, may cause delayed ef-tablishing annual dose limits is to ensure that occupa-        fects. However, only large acute doses cause early ef-tional dose is spread out in time.                            fects; chronic doses within the occupational dose limits It is imponant to distinguish between whole body        do not cause early effects. Since the NRC limits do not and partial body exposure. A localized dose to a small        permit large acute doses, concern with occupational volume of the body would not produce the same effect          radiation risk is primarily focused on controlling as a whole body dose of the same magnitude. For ex-            chronic exposure for which possible delayed effects, ample, if only the hand were exposed, the effect would        such as cancer, are of concern.
mainly be limited to the skin and underlying tissue of              The difference between acute and chronic radi*
the hand. An acute dose of 400 to 600 rads (4-6 Gy)            ation exposure can be shown by using exposure to the to the hand would cause skin reddening; recovery              sun's rays as an example. An intense exposure to the would occur over the following months and no long-            sun can result in painful burning, peeling, and growing term damage would be expected. An acute dose of this          of new skin. However, repeated short exposures pro-magnitude to the whole body could cause death within          vide time for the skin to be repaired between expo-a short time without medical treatment. Medical treat-        sures. Whether exposure to the sun's rays is long term ment would lessen the magnitude of the effects and the        or spread over shon periods, some of the injury may chance of death; however, it would not totally elimi-          not be repaired and may eventually result in skin nate the effects or the chance of death.                      cancer.
Cataracts are an interesting case because they can DELAYED EFFECTS                                              be caused by both acute and chronic radiation. A cer-Delayed effects may occur years after exposure.        tain threshold level of dose to the lens of the eye is These effects are caused indirectly when the radiation        required before there is any observable visual impair-changes pans of the cells in the body, which causes the      ment, and the impairment remains after the exposure normal function of the cell to change, for example,          is stopped. The threshold for cataract development 8.29-5 TAB 18, Rev. 1
 
from acute exposure is an acute dose on the order of          the total amounts allowed if no external radiation is 100 rads (1 Gy). Further, a cumulative dose of 800            received. The resulting dose from the internal radi-rads (8 Gy) from protracted exposures over many              ation sources (from breathing air at 1 DAC) is the years to the lens of the eye has been linked to some          maximum allowed to an organ or to the worker's whole level of visual impairment (Refs. 1 and 4). These doses      body.
exceed the amount that may be accumulated by the lens from normal occupational exposure under the              6. How does radiation cause cancer?
current regulations.                                              The mechanisms of radiation-induced cancer are not completely understood. When radiation interacts
: 5. What is meant by external and internal ex-                with the cells of our bodies, a number of events can posure?
occur. The damaged cells can repair themselves and A worker's occupational dose may be caused by          permanent damage is not caused. The cells can die, exposure to radiation that originates outside the body,      much like the large numbers of cells that die every day called "external exposure," or by exposure to radi-          in our bodies, and be replaced through the normal bio-ation from radioactive material that has been taken          logical processes. Or a change can occur in the cell's into the body, called "internal exposure." Most NRC-          reproductive structure, the cells can mutate and subse-licensed activities involve little, if any, internal expo-    quently be repaired without effect, or they can form sure. It is the current scientific consensus that a rem of    precancerous cells, which may become cancerous. Ra-radiation dose has the same biological risk regardless        diation is only one of many agents with the potential of whether it is from an external or an internal source.      for causing cancer, and cancer caused by radiation The NRC requires that dose from external exposure              cannot be distinguished from cancer attributable to and dose from internal exposure be added together, if        any other cause.
each exceeds 10% of the annual limit, and that the total be within occupational limits. The sum of external          Radiobiologists have studied the relationship be-and internal dose is called the total effective dose          tween large doses of radiation and cancer (Refs. 5 and equivalent (TEDE) and is expressed in units of rems          6). These studies indicate that damage or change to (Sv).                                                        genes in the cell nucleus is the main cause of radiation-induced cancer. This damage may occur directly Although unlikely, radioactive materials may en-        through the interaction of the ionizing radiation in the ter the body through breathing, eating, drinking, or          cell or indirectly through the-actions of chemical prod-open wounds, or they may be absorbed through the              ucts produced by radiation interactions within cells.
skin. The intake of radioactive materials by workers is      Cells are able to repair most damage within hours; generally due to breathing contaminated air. Radioac-        however, some cells may not be repaired properly.
tive materials may be present as fine dust or gases in        Such misrepaired damage is thought to be the origin of the workplace atmosphere. The surfaces of equipment          cancer, but misrepair does not always cause cancer.
and workbenches may be contaminated, and these                Some cell changes are benign or the cell may die; these materials can be resuspended in air during work              changes do not lead to cancer.
activities.
Many factors such as age, general health, inher-If any radioactive material enters the body, the        ited traits, sex, as well as exposure to other cancer-material goes to various organs or is excreted, depend-      causing agents such as cigarette smoke can affect sus-ing on the biochemistry of the material. Most radioiso-      ceptibility to the cancer-causing effects of radiation.
topes are excreted from the body in a few days. For          Many diseases are caused by the interaction of several example, a fraction of any uranium taken into the            factors, and these interactions appear to increase the body will deposit in the bones, where it remains for a        susceptibility to cancer.
longer time. Uranium is slowly eliminated from the body, mostly by way of the kidneys. Most workers are          7. Who developed radiation risk estimates?
not exposed to uranium. Radioactive iodine is prefer-Radiation risk estimates were developed by several entially deposited in the thyroid gland, which is located national and international scientific organizations over in the neck.
the last 40 years. These organizations include the Na-To limit risk to specific organs and the total body,    tional Academy of Sciences (which has issued several an annual limit on intake (ALI) has been established          reports from the Committee on the Biological Effects for each radionuclide. When more than one radionu-            of Ionizing Radiations, BEIR), the National Council on clide is involved, the intake amount of each radionu-        Radiation Protection and Measurements (NCRP), the clide is reduced proportionally. NRC regulations speci-      International Commission on Radiological Protection fy the concentrations o*f radioactive material in the air    (ICRP), and the United Nations Scientific Committee to which a worker may be exposed for 2,000 working            on the Effects of Atomic Radiation (UNSCEAR).
hours in a year. These concentrations are termed the          Each of these organizations continues to review new derived air concentrations (DACs). These limits are          research findings on radiation health risks.
8.29-6 TAB 18, Rev. 1
 
Several reports from these organizations present        delayed cancer because of that 1-rem dose (although new findings on radiation risks based upon revised esti-      the actual number could be more or less than 4) in mates of radiation dose to survivors of the atomic            addition to the 2,000 normal cancer fatalities expected bombing at Hiroshima and Nagasaki. For example,              to occur in that group from all other causes. This UNSCEAR published risk estimates in 1988 and 1993            means that a 1-rem (0.01 Sv) dose may increase an (Refs. 5 and 6). The NCRP also published a report in          individual worker's chances of dying from cancer from 1988, "New Dosimetry at Hiroshima and Nagasaki                20 percent to 20.04 percent. If one's lifetime occupa-and Its Implications for Risk Estimatesn (Ref. 7). In        tional dose is 10 rems, we could raise the estimate to January 1990, the National Academy of Sciences re-            20.4 percent. A lifetime dose of 100 rems may in-leased the fifth report of the BEIR Committee,                crease chances of dying from cancer from 20 to 24 "Health Effects of Exposure to Low Levels of Ionizing        percent. The average measurable dose for radiation Radiation" (Ref. 4). Each of these publications also          workers reported to the NRC was 0.31 rem (0.0031 provides extensive bibliographies on other published          Sv) for 1993 (Ref. 9). Today, very few workers ever studies concerning radiation health effects for those        accumulate 100 rems (1 Sv) in a working lifetime, and who may wish to read further on this subject.                the average career dose of workers at NRC-licensed facilities is 1.5 rems (0.015 Sv), which represents an
: 8. What are the estimates of the risk of fataJ                estimated increase from 20 to about 20.06 percent in cance~ from radiation exposure?                          the risk of dying from cancer.
We don't know exactly what the chances are of It is important to understand the probability fac-getting cancer from a low-level radiation dose, primari-tors here. A similar question would be, "If you select ly because the few effects that may occur cannot be one card from a full deck of cards, will you get the ace distinguished from normally occurring cancers. How-of spades?" This question cannot be answered with a ever, we can make estimates based on extrapolation simple yes or no. The best answer is that your chance is from extensive knowledge from scientific research on 1 in 52. However, if 1000 people each select one card high dose effects. The estimates of radiation effects at from full decks, we can predict that about 20 of them high doses are better known than are those of most will get an ace of spades. Each person will have 1 chemical carcinogens (Ref. 8).
chance in 52 of drawing the ace of spades, but there is From currently available data, the NRC has              no way we can predict which persons will get that card.
adopted a risk value for an occupational dose of 1 rem        The issue is further complicated by the fact that in a (0.01 Sv) Total Effective Dose Equivalent (TEDE) of          drawing by 1000 people, we might get only 15 suc-4 in 10,000 of developing a fatal cancer, or approxi-        cesses, and in another, perhaps 25 correct cards in mately 1 chance in 2,500 of fatal cancer per rem of          1000 draws. We can say that if you receive a radiation TEDE received. The uncertainty associated with this            dose, you will have increased your chances of eventu-risk estimate does not rule out the possibility of higher    ally developing cancer. It is assumed that the more ra-risk, or the possibility that the risk may even be zero at    diation exposure you get, the more you increase your low occupational doses and dose rates.                        chances of cancer.
The radiation risk incurred by a worker depends              The normal chance of dying from cancer is about on the amount of dose received. Under the linear              one in five for persons who have not received any oc-model explained above, a worker who receives 5 rems            cupational radiation dose. The additional chance of (0.05 Sv) in a year incurs 10 times as much risk as          developing fatal cancer from an occupational exposure another worker who receives only 0.5 rem (0.005 Sv).          of 1 rem (0.01 Sv) is about the same as the chance of Only a very few workers receive doses near 5 rems              drawing any ace from a full deck of cards three times in (0.05 Sv) per year (Ref. 9).                                a row. The additional chance of dying from cancer According to the BEIR V report (Ref. 4), approxi-      from an occupational exposure of 10 rem (0.1 Sv) is mately one in five adults normally will die from cancer      about equal to your chance of drawing two aces succes-from all possible causes such as smoking, food, alco-        sively on the first two draws from a full deck of cards.
hol, drugs, air pollutants, natural background radi-It is important to realize that these risk numbers ation, and inherited traits. Thus, in any group of are only estimates based on data for people and re-10,000 workers, we can estimate that about 2,000 search animals exposed to high levels of radiation in (20%) will die from cancer without any occupational short periods of time. There is still uncertainty with re-radiation exposure.
gard to estimates of radiation risk from low levels of To explain the significance of these estimates, we      exposure. Many difficulties are involved in designing will use as an example a group of 10,000 people, each        research studies that can accurately measure the proj-exposed to 1 rem (0.01 Sv) of ionizing radiation. Using      ected small increases in cancer cases that might be the risk factor of 4 effects per 10,000 rem of dose, we      caused by low exposures to radiation as compared to estimate that 4 of the 10,000 people might die from          the normal rate of cancer.
8.29-7 TAB 18, Rev. 1
 
These estimates are considered by the NRC staff            ly because below the limits the effect is small compared to be the best available for the worker to use to make          to differences in the normal cancer incidence from an informed decision concerning acceptance of the              year to year and place to place. The ICRP, NCRP, and risks associated with exposure to radiation. A worker          other standards-setting organizations assume for radi-who decides to accept this risk should try to keep expo-        ation protection purposes that there is some risk, no sure to radiation as low as is reasonably achievable            matter how small the dose (Curves 1 and 2). Some (ALARA) to avoid unnecessary risk.                            scientists believe that the risk drops off to zero at some low dose (Curve 3), the threshold effect. The ICRP
: 9. If I receive a radiation dose that is within                and NCRP endorse the linear quadratic model as a occupational limits, will it cause me to get              conservative means of assuring safety (Curve 2).
cancer?
For regulatory purposes, the NRC uses the straight Probably not. Based on the risk estimates pre-            line portion of Curve 2, which shows the number of viously discussed, the risk of cancer from doses below          effects decreasing linearly as the dose decreases. Be-the occupational limits is believed to be small. Assess-        cause the scientific evidence does not conclusively ment of the cancer risks that may be associated with            demonstrate whether there is or is not an effect at low low doses of radiation are projected from data avail-          doses, the NRC assumes for radiation protection pur-able at doses larger than 10 rems (0.1 Sv) (Ref. 3). For        poses, that even small doses have some chance of caus-radiation protection purposes, these estimates are              ing cancer. Thus, a principle of radiation protection is made using the straight line portion of the linear qua-        to do more than merely meet the allowed regulatory dratic model (Curve 2 in Figure 1). We have data on            limits; doses should be kept as low as is reasonably cancer probabilities only for high doses, as shown by          achievable (ALARA). This is as true for natural car-the solid line in Figure 1. Only in studies involving radi-    cinogens such as sunlight and natural radiation as it is ation doses above occupational limits are there de-            for those that are manmade, such as cigarette smoke, pendable determinations of the risk of cancer, primari-        smog, and x-rays.
DOSE (REMS)                            50REMS Figure 1. Some Proposed Models for How the Effects of Radiation Vary With Doses at Low Levels 8.29-8                                          TAB 18, Rev. 1
: 10. How can we compare the risk of cancer from                working in several types of industries. Table 2 shows radiation to other kinds of health risks?                average days of life expectancy lost as a result of fatal One way to make these comparisons is to compare          work-related accidents. Table 2 does not include non-the average number of days of life expectancy lost            accident types of occupational risks such as occupa-because of the effects associated with each particular        tional disease and stress because the data are not health risk. Estimates are calculated by looking at a        available.
large number of persons, recording the age when death              These comparisons are not ideal because we are occurs from specific causes, and estimating the average      comparing the possible effects of chronic exposure to number of days of life lost as a result of these early        radiation to different kinds of risk such as accidental deaths. The total number of days of life lost is then        death, in which death is inevitable if the event occurs.
averaged over the total observed group.                      This is the best we can do because good data are not Several studies have compared the average days of      available on chronic exposure to other workplace car-life lost from exposure to radiation with the number of      cinogens. Also, the estimates of loss of life expectancy days lost as a result of being exposed to other health        for workers from radiation-induced cancer do not take risks. The word "average" is important because an in-        into consideration the competing effect on the life ex-dividual who gets cancer loses about 15 years of life        pectancy of the workers from industrial accidents.
expectancy, while his or her coworkers do not suffer any loss.                                                    11. What are the health risks from radiation exposure to the embryo/fetus?
Some representative numbers are presented in During certain stages of development, the embryo/
Table 1. For categories of NRC-regulated industries fetus is believed to be more sensitive to radiation dam-with larger doses, the average measurable occupational        age than adults. Studies of atomic bomb survivors ex-dose in 1993 was 0.31 rem (0.0031 Sv). A simple cal-          posed to acute radiation doses exceeding 20 rads (0.2 culation based on the article by Cohen and Lee (Ref.
Gy) during pregnancy show that children born after
: 10) shows that 0.3 rem (0.003 Sv) per year from age receiving these doses have a higher risk of mental re*
18 to 65 results in an average loss of 15 days. These tardation. Other studies suggest that an association ex-estimates indicate that the health risks from occupa*
ists between exposure to diagnostic x-rays before birth tional radiation exposure are smaller than the risks as-and carcinogenic effects in childhood and in adult life.
sociated with many other events or activities we en*          Scientists are uncertain about the magnitude of the counter and accept in normal day-to-day activities.          risk. Some studies show the embryo/fetus to be more It is also useful to compare the estimated average      sensitive to radiation-induced cancer than adults, but number of days of life lost from occupational exposure        other studies do not. In recognition of the possibility of to radiation with the number of days lost as a result of      increased radiation sensitivity, and because dose to the Table 1 Estimated Loss of Life Expectancy from Health Risks 8 Estimate of Ufe Expectancy Lost Health Risk                                                            (average)
Smoking 20 cigarettes a day                                                6 years Overweight (by 15%)                                                        2 years Alcohol consumption (U.S. average)                                          1 year All accidents combined                                                      1 year Motor vehicle accidents                              207 days Home accidents                                        74 days Drowning                                              24 days All natural hazards (earthquake, lightning, flood, etc.)                    7 days Medical radiation                                                          6 days Occupational Exposure 0.3 rem/y from age 18 to 65                          1S days 1 rem/y from age 18 to 65                            51 days aAdapted from Reference 10.
8.29-9                                        TAB18, Rev.1
 
for women (Refs. 1 and 4). These doses are far greater Table 2 Estimated Loss of Life Expectancy                  than the NRC s occupational dose limits for workers.
from Industrial Accidents 8 Although acute doses can affect fertility by reduc-ing sperm count or suppressing ovulation, they do not Estimated Days of Life have any direct effect on one's ability to function sexu-industry Type              Expectancy Lost (Average) ally. No evidence exists to suggest that exposures with-in the NRC's occupational limits have any effect on the All industries                            60 ability to function sexually.
Agriculture                              320 Construction                            227                  13. What are the NRC occupational dose limits?
Mining and Quarrying                    16 7                      For adults, an annual limit that does not exceed:
Transportation and
* S rems (O.OS Sv) for the total effective dose equiv-Public Utilities                      160 alent (TEDE), which is the sum of the deep dose Government                                60                        equivalent (DOE) from external exposure to the Manufacturing                            40                        whole body and the committed effective dose Trade                                    27                        equivalent (CEDE) from intakes of radioactive Services                                  27                        material.
* SO rems (0.5 Sv) for the total organ dose equiva-
  &Adapted from Reference 10.
lent (TODE), which is the sum of the ODE from external exposure to the whole body and the com-embryo/fetus is involuntary on the part of the embryo/              mitted dose equivalent (CDE) from intakes of ra-fetus, a more restrictive dose limit has been established          dioactive material to any individual organ or tis-for the embryo/fetus of a declared pregnant radiation              sue, other than the lens of the eye.
worker. See Regulatory Guide 8.13, "Instruction Con-cerning Prenatal Radiation Exposure."
* 15 rems (0.1S Sv) for the lens dose equivalent (LDE), which is the external dose to the lens of If an occupationally exposed woman decJares her              the eye.
pregnancy in writing, she is subject to the more restric-tive dose limits for the embryo/fetus during the remain-
* SO rems (O.S Sv) for the shallow dose equivalent 1er of the pregnancy. The dose limit of 500 mrems (S              (SDE), which is the external dose to the skin or to mSv) for the total gestation period applies to the em-              any extremity.
bryo/fetus and is controlled by restricting the exposure            For minor workers, the annual occupational dose to the declared pregnant woman. Restricting the wom-          limits are 10 percent of the dose limits for adult work-an's occupational exposure, if she declares her preg-          ers.
nancy, raises questions about individual privacy rights, For protection of the embryo/fetus of a declared equal employment opportunities, and the possible loss pregnant woman, the dose limit is O.S rem (S mSv) of income. Because of these concerns, the declaration during the entire pregnancy.
of pregnancy by a female radiation worker is volun-tary. Also, the declaration of pregnancy can be with-              The occupational dose limit for adult workers of S drawn for any reason, for example, if the woman be-          rems (O.OS Sv) TEDE is based on consideration of the lieves that her benefits from receiving the occupational      potential for delayed biological effects. The S-rem exposure would outweigh the risk to her embryo/fetus          (O.OS Sv) limit, together with application of the con-from the radiation exposure.                                  cept of keeping occupational doses ALARA, provides a level of risk of delayed effects considered acceptable
: 12. Can a worker become sterile or impotent                  by the NRC. The limits for individual organs are below from normal occupational radiation                      the dose levels at which early biological effects are ob-exposure?                                              served in the individual organs.
No. Temporary or permanent sterility cannot be                The dose limit for the embryo/fetus of a declared caused by radiation at the levels allowed under NRC's          pregnant woman is based on a consideration of the occupational limits. There is a threshold below which          possibility of greater sensitivity to radiation of the em-these effects do not occur. Acute doses on the order of        bryo/fetus and the involuntary nature of the exposure.
10 rems (0.1 Sv) to the testes can result in a measur-able but temporary reduction in sperm count. Tempo-rary sterility (suppression of ovulation) has been ob-        14. What is meant by ALARA?
'-erved in women who have received acute doses of 1S0                ALARA means "as low as is reasonably achiev-ads (1.S Gy). The estimated threshold (acute) radi-          able." In addition to providing an upper limit on an ation dose for induction of permanent sterility is about      individual's permissible radiation dose, the NRC re-200 rads (2 Gy) for men and about 3S0 rads (3.S Gy)            quires that its licensees establish radiation protection 8.29-10 TAB 18, Rev. 1
 
programs and use procedures and engineering controls        al radiation dose of about 0.36 rem (3.6 mSv). By age to achieve occupational doses, and doses to the public,    20, the average person will accumulate over 7 rems {70 as far below the limits as is reasonably achievable.        mSv) of dose. By age 50, the total dose is up to 18 rems "Reasonably achievable" also means "to the extent          (180 mSv). After 70 years of exposure this dose is up practicable." What is practicable depends on the pur-      to 25 rems (250 mSv).
pose of the job, the state of technology, the costs for averting doses, and the benefits. Although implemen-tation of the ALARA principle is a required integral Table 3 Average Annual Effective Dose Equiva-part of each licensee's radiation protection program, it                lent to Individuals in the u.s.a does not mean that each radiation exposure must be kept to an absolute minimum, but rather that "reason-                                          Effective Dose able" efforts must be made to avert dose. In practice,      Source                              Equivalent (mrems)
ALARA includes planning tasks involving radiation exposure so as to reduce dose to individual workers        Natural and the work group.
Radon                        200 There are several ways to control radiation doses,            Other than Radon            .1QQ e.g., limiting the time in radiation areas, maintaining            Total                                  300 distance from sources of radiation, and providing          Nuclear Fuel Cycle                              0.05 shielding of radiation sources to reduce dose. The use      Consumer Productsb                              9 of engineering controls, from the design of facilities      Medical and equipment to the actual set-up and conduct of Diagnostic X-rays            39 work activities, is also an important element of the ALARA concept.                                                    Nuclear Medicine              li Total                                    53 An ALARA analysis should be used in determin-          Total                                about 360 ing whether the use of respiratory protection is advis-                                                  mrems/year able. In evaluating whether or not to use respirators, the goal should be to achieve the optimal sum of exter-      aAdapted from Table 8.1, NCRP 93 (Ref. 11).
nal and internal doses. For example, the use of respi-bincludes building material, television receivers, lumi-rators can lead to increased work time within radiation nous watches, smoke detectors, etc. (from Table 5.1, areas, which increases external dose. The advantage of NCRP 93, Ref. 11).
using respirators to reduce internal exposure must be evaluated against the increased external exposure and related stresses caused by the use of respirators. Heat      16. What are the typical radiation doses received stress, reduced visibility, and reduced communication            by workers?
associated with the use of respirators could expose a worker to far greater risks than are associated with the          For 1993, the NRC received reports on about a internal dose avoided by use of the respirator. To the      quarter of a million people who were monitored for extent practical, engineering controls, such as contain-    o_ccupational exposure to radiation. Almost half of ments and ventilation systems, should be used to re-        those monitored had no measurable doses. The other duce workplace airborne radioactive materials.              half had an average dose of about 310 mrem (3.1 mSv) for the year. Of these, 93 percent received an annual dose of less than 1 rem ( 10 mSv) ~ 98. 7 percent
: 15. What are background radiation exposures?                received less than 2 rems (20 mSv); and the highest reported dose was for two individuals who each re-The average person is constantly exposed to ioniz-    ceived between 5 and 6 rems (50 and 60 mSv).
ing radiation from several sources. Our environment and even the human body contain naturally occurring                Table 4 lists average occupational doses for work-radioactive materials (e.g., potassium-40) that contrib-    ers (persons who had measurable doses) in various oc-ute to the radiation dose that we receive. The largest      cupations based on 1993 data. It is important to note source of natural background radiation exposure is ter-      that beginning in 1994, licensees have been required to restrial radon, a colorless, odorless, chemically inert      sum external and internal doses and certain licensees gas, which causes about 55 percent of our average,          are required to submit annual reports. Certain types of nonoccupational exposure. Cosmic radiation originat-        licensees such as nuclear fuel fabricators may report a ing in space contributes additional exposure. The use        significant increase in worker doses because of the of x-rays and radioactive materials in medicine and          exposure to long-lived airborne radionuclides and the dentistry adds to our population exposure. As shown          requirement to add the resultant internal dose to the below in Table 3, the average person receives an annu-      calculation of occupational doses.
8.29-11 TAB 18, Rev. 1
: 18. What happens if a worker exceeds the annual dose limit?
Table 4 Reported Occupational Doses for 1993 8                    If a worker receives a dose in excess of any of the annual dose limits, the regulations prohibit any occu-Average Measurable        pational exposure during the remainder of the year in Occupational                          Dose per Worker        which the limit is exceeded. The licensee is also re-Subgroup                                  (millirems)        quired to file an overexposure report with the NRC and provide a copy to the individual who received the dose.
The licensee may be subject to NRC enforcement ac-Industrial Radiography                      540              tion such as a fine (civil penalty), just as individuals are Commercial Nuclear Power Reactors          310              subject to a traffic fine for exceeding a speed limit. The Manufacturing and Distribution                                fines and, in some serious or repetitive cases, suspen-of Radioactive Materials                  300              sion of a license are intended to encourage licensees to Low-Level Radioactive Waste                                  comply with the regulations.
Disposal                                  270                    Radiation protection limits do not define safe or Independent Spent Nuclear Fuel                                unsafe levels of radiation exposure. Exceeding a limit Storage                                  260              does not mean that you will get cancer. For radiation Nuclear Fuel Fabrication                    130              protection purposes, it is assumed that risks are related to the size of the radiation dose. Therefore, when your dose is higher your risk is also considered to be higher.
  &From Table 3.1 in NUREG-0713 (Ref. 9).                      These limits are similar to highway speed limits. If you drive at 70 mph, your risk is higher than at 55 mph, even though you may not actually have an accident.
: 17. How do I know how much my occupational                    Those who set speed limits have determined that the dose {exposure) is?
risks of driving in* excess of the speed limit are not ac-If you are likely to receive more than 10 percent of    ceptable. In the same way, the revised 10 CFR Part 20 the annual dose limits, the NRC requires your employ-          establishes a limit for normal occupational exposure of er, the NRC licensee, to monitor your dose, to main-          5 rems (0.05 Sv) a year. Although you will not neces-tain records of your dose, and, at least on an annual          sarily get cancer or some other radiation effect at doses
  ,asis for the types of licensees listed in 10 CFR            above the limit, it does mean that the licensee's safety 20.2206, "Reports of Individual Monitoring," to in-          program has failed in some way. Investigation is war-form both you and the NRC of your dose. The purpose            ranted to determine the cause and correct the condi-of this monitoring and reporting is so that the NRC can      tions leading to the dose in excess of the limit.
be sure that licensees are complying with the occupa-          19. What is meant by a "planned special tional dose limits and the ALARA principle.                        exposure"?
External exposures are monitored by using indi-                A "planned special exposure" (PSE) is an infre-vidual monitoring devices. These devices are required        quent exposure to radiation, separate from and in ad-to be used if it appears likely that external exposure        dition to the radiation received under the annual occu-will exceed 10 percent of the allowed annual dose, i.e.,      pational limits. The licensee can authorize additional 0.5 rem (5 mSv). The most commonly used monitor-              dose in any one year that is equal to the annual occu-ing devices are film badges, thermoluminescence do-          pational dose limit as long as the individual's total dose simeters (TLDs), electronic dosimeters, and direct            from PSEs does not exceed five times the annual dose reading pocket dosimeters.                                    limit during the individual's lifetime. For example, li-censees may authorize PSEs for an adult radiation With respect to internal exposure, your employer        worker to receive doses up to an additional S rems is required to monitor your occupational intake of ra-        (0.05 Sv) in a year above the 5-rem (0.05-Sv) annual dioactive material and assess the resulting dose if it ap-    TEDE occupational dose limit. Each worker is limited pears likely that you will receive greater than 10 per-        to no more than 25 rems (0.25 Sv) from planned spe-cent of the annual limit on intake (ALI) from intakes          cial exposures in his or her lifetime. Such exposures in 1 year. Internal exposure can be estimated by mea-          are only allowed in exceptional situations when alter-suring the radiation emitted from the body (for exam-          natives for avoiding the additional exposure are not ple, with a "whole body counter") or by measuring the          available or are impractical.
radioactive materials contained in biological samples                Before the licensee authorizes a PSE, the licensee i;uch as urine or feces. Dose estimates can also be          must ensure that the worker is informed of the purpose nade if one knows how much radioactive material was        and circumstances of the planned operation, the esti-in the air and the length of time during which the air        mated doses expected, and the procedures to keep the was breathed.
doses ALARA while considering other risks that may 8.29-12 TAB 18, Rev. 1
 
be present. (See Regulatory Guide 8.35, "Planned                Part 20 "shall be construed as limiting actions that may Special Exposures.")                                            be necessary to protect health and safety."
Rare situations may occur in which a dose in ex-
: 20. Why do some facilities establish administra-                cess of occupational limits would be unavoidable in or-tive control levels that are below the NRC                  der to carry out a lifesaving operation or to avoid a limits?                                                    large dose to large populations. However, persons There are two reasons. First, the NRC regulations          called upon to undertake any emergency operation state that licensees must take steps to keep exposures          should do so only on a voluntary basis and with full to radiation ALARA. Specific approval from the li-              awareness of the risks involved.
censee for workers to receive doses in excess of admin-              For perspective, the Environmental Protection istrative limits usually results in more critical risk-bene-    Agency (EPA) has published emergency dose guide-fit analyses as each additional increment of dose is            lines (Ref. 2). These guidelines state that doses to all approved for a worker. Secondly, an administrative              workers during emergencies should, to the extent prac-control level that is set lower than the NRC limit pro-          ticable, be limited to 5 rems (0.05 Sv). The EPA fur-vides a safety margin designed to help the licensee              ther states that there are some emergency situations for avoid doses to workers in excess of the limit.                    which higher limits may be justified. The dose resulting from such emergency exposures should be limited to
: 21. Why aren't medical exposures considered as                  10 rems (0.1 Sv) for protecting valuable property, and part of a worker's allowed dose?                          to 25 rems (0.25 Sv) for lifesaving activities and the protection of large populations. In the context of this NRC rules exempt medical exposure, but equal guidance, the dose to workers that is incurred for the doses of medical and occupational radiation have protection of large populations might be considered equal risks. Medical exposure to radiation is justified justified for situations in which the collective dose to for reasons that are quite different from the reasons for others that is avoided as a result of the emergency op-occupational exposure. A physician prescribing an x-eration is significantly larger than that incurred by the ray, for example, makes a medical judgment that the workers involved.
benefit to the patient from the resulting medical infor-mation justifies the risk associated with the radiation.              Table 5 presents the estimates of the fatal cancer This judgment may or may not be accepted by the pa-              risk for a group of 1,000 workers of various ages, as-tient. Similarly, each worker must decide on the bene-            suming that each worker received an acute dose of 25 fits and acceptability of occupational radiation risk,            rems (0.25 Sv) in the course of assisting in an emer-just as each worker must decide on the acceptability of          gency. The estimates show that a 25-rem emergency any other occupational hazard.                                    dose might increase an individual's chances of devel-oping fatal cancer from about 20% to about 21%.
Consider a worker who receives a dose of 3 rems (0.03 Sv) from a series of x-rays in connection with an injury or illness. This dose and any associated risk must be justified on medical grounds. If the worker had also                                    Table 5 received 2 rems (0.02 Sv) on the job, the combined                      Risk of Premature Death from Exposure dose of 5 rems (0.05 Sv) would in no way incapacitate                      to 25-Rems (0.25-Sv) Acute Dose the worker. Restricting the worker from additional job exposure during the remainder of the year would not                                                  Estimated Risk have any effect on the risk from the 3 rems (0.03 Sv)                    Age at                      of Premature Death already received from the medical exposure. If the in-                    Exposure                    (Deaths per 1,000 dividual worker accepts the risks associated with the                    (years)                    Persons Exposed) x-rays on the basis of the medical benefits and accepts the risks associated with job-related exposure on the                    20-30                                9.1 basis of employment benefits, it would be unreason-                      30-40                                7.2 able to restrict the worker from employment involving                    40-50                                5.3 exposure to radiation for the remainder of the year.                      50-60                                3.5
: 22. How should radiation risks be considered in                  Source: EPA-400-R-92-001 (Ref. 2).
an emergency?
Emergencies are "unplanned" events in which ac-            23. How were radiation dose limits established?
tions to save lives or property may warrant additional                  The NRC radiation dose limits in 10 CFR Part 20 doses for which no particular limit applies. The revised          were established by the NRC based on the recommen-10 CFR Part 20 does not set any dose limits for emer-            dations of the ICRP and NCRP as endorsed in Federal gency or lifesaving activities and states that nothing in        radiation protection guidance developed by the EPA 8.29-13 TAB 18, Rev. 1
 
(Ref. 12). The limits were recommended by the ICRP            tries and are considered acceptable by the scientific and NCRP with the objective of ensuring that working          groups that have studied them. An employer is not ob-in a radiation-related industry was as safe as working in      ligated to guarantee a transfer if a worker decides not other comparable industries. The dose limits and the          to accept an assignment that requires exposure to radi-inciple of ALARA should ensure that risks to work-          ation .
    *-s are maintained indistinguishable from risks from              Any worker has the option of seeking other em-background radiation.
ployment in a nonradiation occupation. However, the studies that have compared occupational risks in the
: 24. Several scientific reports have recommended nuclear industry to those in other job areas indicate that the NRC establish lower dose limits.
that nuclear work is relatively safe. Thus, a worker may Does the NRC plan to reduce the regulatory limits?                                                find different kinds of risk but will not necessarily find significantly lower risks in another job.
Since publication of the NRC's proposed rule in 1986, the ICRP in 1990 revised its recommendations            26. Where can one get additional information on for radiation protection based on newer studies of radi-          radiation risk?
ation risks (Ref. 13), and the NCRP followed with a                The following list suggests sources of useful infor-revision to its recommendations in 1993. The ICRP              mation on radiation risk:
recommended a limit of 10 rems (0.1 Sv) effective dose equivalent (from internal and external sources),
* The employer-the radiation protection or health over a 5-year period with no more than S rems (0.05                physics office where a worker is employed.
Sv) in 1 year (Ref. 13). The NCRP recommended a
* Nuclear Regulatory Commission Regional Offices:
cumulative limit in rems, not to exceed the individual's King of Prussia, Pennsylvania    (610) 337-5000 age in years, with no more than 5 rems (0.05 Sv) in any Atlanta, Georgia                (404)  331-4503 year (Ref. 14).                                                    Lisle, Illinois                  (708) 829-9500 The NRC does not believe that additional reduc-              Arlington, Texas                  (817) 860-8100 tions in the dose limits are required at this time. Be-
* U.S. Nuclear Regulatory Commission cause of the practice of maintaining radiation expo-                  Headquarters sures ALARA (as low as is reasonably achievable), the              Radiation Protection & Health Effects Branch werage radiation dose to occupationally exposed per-                Office of Nuclear Regulatory Research
    .1s is well below the limits in the current Pan 20 that          Washington, DC 20555
..,ccame mandatory January 1, 1994, and the average                Telephone: (301) 415-6187 doses to radiation workers are below the new limits recommended by the ICRP and the NCRP.
* Depanment of Health and Human Services Center for Devices and Radiological Health
: 25. What are the options if a worker decides that                  1390 Piccard Drive, MS HFZ-1 the risks associated with occupational radi-                  Rockville, MD 20850 ation exposure are too high?                                  Telephone: (301) 443-4690 If the risks from exposure to occupational radi-
* U.S. Environmental Protection Agency ation are unacceptable to a worker, he or she can re-                Office of Radiation and Indoor Air quest a transfer to a job that does not involve exposure            Criteria and Standards Division to radiation. However, the risks associated with the ex-            401 M Street NW.
posure to radiation that workers, on the average, ac-                Washington, DC 20460 tually receive are comparable to risks in other indus-              Telephone: (202) 233-9290 8.29-14                                            TAB 18, Rev. 1
 
REFERENCES
: 1. B.R. Scott et al., "Health Effects Model for Nu-                    Protection and Measurements Held on April 8-clear Power Plant Accident Consequence Analy-                        9, 1987 (1988).
sis," Part I: Introduction, Integration, and Sum-mary, U.S. Nuclear Regulatory Commission,                      8. National Council on Radiation Protection.and NUREG/CR-4214, Revision 2, Part I, October                          Measurements, Comparative Carcinogenicity of 1993.
* Ionizing Radiation and Chemicals, NCRP Report No. 96, March 1989.
: 2. U.S. Environmental Protection Agency, Manual of Protective Action Guides and Protective Ac-                9. C.T. Raddatz and D. Hagemeyer, "Occupational tions for Nuclear Incidents, EPA-400-R                            Radiation Exposure at Commercial Nuclear Pow-001, May 1992.                                                      er Reactors and Other Facilities, 1993," U.S.
Nuclear Regulatory Commission, NUREG-0713,
: 3. International Commission on Radiological Pro-                        Volume 15, January 1995.
* tection, Annals of the ICRP, Risks Associated with Ionising Radiation, Volume 22, No.1, Per*                10. B.L. Cohen and I.S. Lee, "Catalog of Risks Ex-gamon Press, Oxford, UK, 1991.                                      tended and Updated," Health Physics, Vol. 61, September 19 91.
: 4. National Research Council, Health Effects of Ex-posure to Low Levels of Ionizing Radiation, Re-                11. National Council on Radiation Protection and port of the Committee on the Biological Effects of                    Measurements, Ionizing Radiation Exposure of Ionizing Radiation (BEIR V), National Academy                        the Population of the United States, NCRP Re-Press, Washington, DC, 1990.                                        port No. 93, September 1987.
S. United Nations Scientific Committee on the Ef-                12. U.S. Environmental Protection Agency, "Radi-fects of Atomic Radiation (UNSCEAR); Sources,                        ation Protection Guidance to Federal Agencies Effects and Risks of Ionizing Radiation, Report                      for Occupational Exposure," Federal Register, E.88.IX.7, United Nations, New York, 1988.                            Vol. 52, No. 17, January 27, 1987.
: 6. United Nations Scientific Committee on the Ef-                  13. International Commission on Radiological Pro-fects of Atomic Radiation (UNSCEAR), Sources                        tection, 1990 Recommendations of the Interna-and Effects of Ionizing Radiation, United Na-                        tional Commission on Radiological Protection, tions, New York, 1993.                                              ICRP Publication 60, Pergamon Press, Oxford, UK, 1991.
: 7. National Council on Radiation Protection and Measurements, New Dosimetry at Hiroshima                        14. National Council on Radiation Protection and and Nagasaki and Its Implications for Risk Esti-                    Measurements, Limitation of Exposure to Ioniz-mates, Proceedings of the Twenty-third Annual                        ing Radiation, NCRP Report No. 116, March Meeting of the National Council on Radiation                        1993.
*Copies are available for inspection or copying for a fee from the NRC Public Document Room at 2120 L Street NW., Washington, DC; the PDR's mailing address is Mail Stop LL-6, Washington, DC 20555; telephone (202) 634-3273; fax (202) 634-3343. Copies maybe purchased at current rates from the U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20402-9328 (tele-phone (202) S12-2249); or from the National Technical ln!ormaUon Service by writing NTIS at S285 Port Royal Road, Springfield, VA 22161.
8.29-15 TAB 18, Rev. 1
 
BIBLIOGRAPHY Abrahamson, S., et al., "Health Effects Models for                  States and Canada from Natural Background Radi-Nuclear Power Plant Accident Consequence Analy-                      ation, NCRP Report No. 94, December 1987.
,is," Part II: Scientific Bases for Health Effects Mod-
.ds, U.S. Nuclear Regulatory Commission, NUREG/                      National Council on Radiation Protection and Mea-CR-4214, Rev. 1, Part II, May 1989. 1                                surements, Exposure of the U.S. Population From Oc-cupational Radiation, NCRP Report No. 101, June Abrahamson, S., et al., "Health Effects Models for                    1989.
Nuclear Power Plant Accident Consequence Analysis, Modifications of Models Resulting From Recent Re-                    National Council on Radiation Protection and Mea-ports on Health Effects of Ionizing Radiation, Low                  surements, Risk Estimates for Radiation Protection, LET Radiation," Part II: Scientific Basis for Health                NCRP Report No. 11S, December 1993.
Effects Models, U.S. Nuclear Regulatory Commission, NUREG/CR-4214, Rev.1, Part II, Addendum 1, Au-                        National Council on Radiation Protection and Mea-gust 1991.1                                                          surements, Limitation of Exposure to Ionizing Radi-ation, NCRP Report No. 116, March 1993.
Abrahamson, S., et al., "Health Effects Models for Nuclear Power Plant Accident Consequence Analysis,                    National Safety Council, Accident Facts, 1993 Edi-Modifications of Models Resulting From Addition of                    tion, Itasca, Illinois, 1993.
Effects of Exposure to Alpha-Emitting Radionu-U.S. Environmental Protection Agency, "Radiation clides," Part II: Scientific Bases for Health Effects Protection Guidance to Federal Agencies for Occupa-Models, U.S. Nuclear Regulatory Commission, tional Exposure," Federal Register, Vol. 52, No. 17, NUREG/CR-4214, Rev.1, Part II, Addendum 2, May January 27, 1987.
1993. 1 U.S. Nuclear Regulatory Commission, "Instruction International Commission on Radiological Protection, Concerning Prenatal Radiation Exposure," Regulatory Radiation Protection, Recommendations of the Inter-Guide 8.13, Revision 2, December 1987. 2 national Commission on Radiological Protection, ICRP Publication 26, Pergamon Press, Oxford, UK, January                  U.S. Nuclear Regulatory Commission, "Monitoring 1 977.
Criteria and Methods To Calculate Occupational Radi-ation Doses," Regulatory Guide 8.34, July 1992.2 National Council on Radiation Protection and Mea-surements, Public Radiation Exposure From Nuclear                    U.S. Nuclear Regulatory Commission, "Planned Spe-Power Generation in the United States, NCRP Report                  cial Exposures," Regulatory Guide 8.35, June 1992. 2 No. 92, December 1987.
U.S. Nuclear Regulatory Commission, "Radiation National Council on Radiation Protection and Mea-                    Dose to the Embryo/Fetus," Regulatory Guide 8.36, surements, Exposure of the Population in the United                  July 1992. 2 1Copies  are available for inspection or copying for a fee from the NRC Public Document Room at 2120 L Street NW.,                  2 Single copies of regulatory guides may be obtained free of Washington, DC; the PDR's mailing address is Mail Stop                charge by writing the Office of Administration, Attn: Distri-LL-6, Washington, DC 20555-0001; telephone (202)                    bution and Services Section, USNRC, Washington, DC 634-3273; fax (202) 634-3343. Copies may be purchased at            20555, or by fax at (301) 415-2260. Copies are available for current rates from the U.S. Government Printing Office, P.O.        inspection or copying for a fee from the NRC Public Document Box 37082, Washington, DC 20402-9328 (telephone (202)                Room at 2120 L Street NW., Washington, DC; the PDR's 512-2249); or from the National Technical lnformation Ser-          mailing address is Mail Stop LL-6, Washington, DC vice by writing NTIS at 5285 Port Royal Road, Springfield,          20555-0001; telephone (202) 634-3273; fax (202)
VA 22161.                                                            634-3343.
8.29-16 TAB 18, Rev. 1
 
REGULATORY ANALYSIS A separate regulatory analysis was not prepared        May 21, 1991 (56 FR 23360). The regulatory analysis for this Revision 1 to Regulatory Guide 8.29. A value/      prepared for 10 CFR Part 20 provides the regulatory impact statement, which evaluated essentially the same      basis for this Revision 1 of Regulatory Guide 8.29, and subjects as are discussed in a regulatory analysis, ac-    it examines the costs and benefits of the rule as i,m.:.
companied Regulatory Guide 8.29 when it was issued          plemented by the guide. A copy of the "Regulatory in July 1981.                                              Analysis for the Revision of 10 CFR Part 20" (PNL-6712, November 1988), is available for inspec-This Revision 1 to Regulatory Guide 8.29 is need-    tion and copying for a fee in the NRC's Public Docu-ed to conform with the Revised 10 CFR Part 20, "Stan-      ment Room at 2120 L Street NW., Washington, DC dards for Protection Against Radiation," as published      20555-0001.
8.29-17 TAB 18, Rev. 1
 
This Tab 19 has been Deleted Rev. 26 Page 1 of 1    TAB 19, Rev. 26
 
This Tab 20 has been Deleted Rev. 20 Page 1 of 1    TAB 20, Rev. 20
 
This Tab 21 has been Deleted Rev 26 Page 1 of 1    TAB 21, Rev. 26
 
Radiation Monitors EAL Matrix Emergency Action Level Matrix:
The EAL Matrix (PINGP 1576) provides the emergency action level thresholds for various plant radiation monitor readings. All plant radiation effluent and area radiation monitors are used in the EAL Matrix. Review PINGP 1576 EAL Matrix for actual EAL thresholds.
Viewing Rad Monitor Alarm Setpoints:
ERCS and Radiation Alarm setpoints may NOT be the same. Use NOTE:
the Calibration Card / Set Point Program for actual values.
VIA Calibration Card/Set Point File:
Alarm set points are maintained in the Calibration Card/Set Point File. All radiation monitor High Alarm setpoint may be printed by performing the following:
: 1. Go to Start Menu, Applications, Calibration Card & Set Point File.
: 2. Enter your own Username and password.
: 3. Select Module, Set Point.
: 4. For viewing the entire radiation monitor listing, choose System RD and enter Search button.
: 5. After about 1 minute of searching, the results should be shown on the screen.
: 6. To print the results, choose File, Print. The setpoint file will be printed to your selected printer.
VIA ERCS:
ERCS may be used to display radiation monitor alarms:
: 1. Go to either ERCS Tabular RADMON Group or ERCS Display RADMON.
: 2. Click value of the rad monitor of interest. The selected rad monitors Analog Point Attributes Definition display will be shown.
: 3. Click the Alarm Limits button in lower right corner of display. The rad monitors alarm setpoints will be displayed.
ODCM setpoints and limits:
ODCM setpoints are managed by the Rad Protection Group. RPIP 4523, Monthly Effluent Monitor Setpoint Determination is a responsibility of the Radiation Protection Group. Forms generated by this RPIP are:
Monthly Effluent Monitor Setpoint Determination - Computer Generated Monthly Gaseous Effluent Monitor Setpoint Determination - PINGP 627 Monthly Liquid Effluent Monitor Setpoint Determination - PINGP 628 Contact Rad Protection supervision to learn of the most current specific ODCM effluent monitor setpoints.
Trending Rad Monitors on ERCS (F3-26.2):
: 1. Choose CUSTOM TRENDS on ERCS Main Menu.
: 2. Choose the desired Trend window. Up to 12 rad monitors may be plotted.
: 3. Select the first data point name field. A list of data points available will be presented.
: 4. Choose the list of desired rad monitors.
: 5. Choose start and end times.
REC RPSS 22.docx                      Page 1 of 1                  TAB 22, Rev. 19
 
Power Supplies for Effluent Rad Monitors &
Ventilation Systems Rev. 0 Developed By Jim Payton, February, 2003 Page 1 of 4      TAB 23, Rev. 0
 
Effluent Radiation Monitors Sample Pump &
Loop Monitors (R-70/71) Power Supplies Effluent Monitor Panel              MCC            Bus                Bus 1R-11/12                            1T2            122 or 222        16 or 26 1R-22                                1T1            112 or 212        15 or 25 1R-30                                1T2            122 or 222        16 or 26 1R-37                                1T1            112 or 212        15 or 25 1R-50            1RPB8              1T2            122 or 222        16 or 26 1/2R-70          1RPB8              1T2            122 or 222        16 or 26 2R-11/12                            1T1            212 or 112        25 or 15 2R-22                                1T2            122 or 222        16 or 26 2R-30                                1T2            122 or 222        16 or 26 2R-37                                1T1            112 or 212        15 or 25 2R-50            1RPA8              1T1            112 or 212        15 or 25 1/2R-71          1RPA8              1T1            112 or 212        15 or 25 R-26                                1K1            111                15 R-27                                1K2            121                16 R-35                                1RW1 Bus1      290                Non Safeguards Ventilation System Exhaust Fan Power Supplies Ventilation Sys            Exhaust Fan      MCC            Bus              Bus SHIELD BUILDING VENTILATION
                          # 11              1M1            112              15
                          # 12              1M2            122              16
                          # 21              2M1            212              25
                          # 22              2M2            222              26 AUXILIARY BUILDING SPECIAL VENTILATION
                          # 121            1MA1          112 or 212      15 or 25
                          # 122            1MA2          122 or 222      16 or 26 SPENT FUEL POOL SPECIAL VENTILATION
                          # 121            1MA1          112 or 212      15 or 25
                          # 122            1MA2          122 or 222      16 or 26 Page 2 of 4                  TAB 23, Rev. 0
 
SHIELD BUILDING VENTILATION SYSTEM Dampers isolate the filters during standby conditions. The ventilation system for each Shield Building also contains a vent stack, which penetrates the Shield Building dome and discharges to the atmosphere. Air can be directed to the stack from the Shield Building vent filters, the Auxiliary Building special vent filters, the Containment Purge filters or the Containment In-service Purge filters.
The Shield Building Ventilation System has two modes of operation, exhaust and recirculation. When the Shield Building recirculation fan starts from either SI or the control switch, the system is in the exhaust mode. The exhaust fan discharge damper and the recirc fan discharge damper are both open and the recirc damper is closed. All airflow is being exhausted through the stack.
When the annulus delta-P reaches -2 inches of H20, the recirc damper opens and the system is in the recirculation mode. All three dampers will stay open until the Shield Building recirc fan is stopped. During recirc mode, air is recirculated to the annulus.
The system maintains negative pressure by removing in-leakage to the annulus through the exhaust path.
Design Limits Recirculation Fans Number                      2 per unit Type                        Axial Vane Capacity                    5,000 cfm Motor                        20 hp Exhaust Fan Number                      2 per unit Type                        Axial Vane Capacity                    200 cfm Motor                        5 hp Filter Assembly Number                      2 per unit Type                        Particulate, absolute, charcoal with demister Heater Capacity              16 kw Filter Efficiency            99.9% for particulate > O. 3 microns 99.9% for elemental iodine 95% for organic iodine Page 3 of 4                        TAB 23, Rev. 0
 
AUXILIARY BUILDING SPECIAL VENTILATION SYSTEM The Auxiliary Building Special Ventilation System collects air leakage into the Auxiliary Building from the Containment Vessel following an accident. The system filters the leakage and directs it to the Shield Building vent stack for discharge to the atmosphere.
The system is designed to create a partial vacuum in the area of highest possible contamination, producing airflow from areas of low contamination to areas of high contamination.
A single system serves both units. The system consists of two redundant trains with each train containing a fan, a filter and the associated ducting. Special ventilation exhaust fans
#121 and #122 are powered from safeguards MCC 1M Bus 1 and Bus 2 respectively.
Radiation detection equipment monitors the air flowing through the system.
Page 4 of 4                      TAB 23, Rev. 0
 
Steam Generator Iodine Partitioning Rev. 0 I. ISSUE:
The best estimate for a I/NG ratio should be used when determining the Iodine source term during a Steam Generator Tube Rupture accident. The appropriate I/NG ratio is dependant upon the RCS I/NG ratio and the amount of iodine partitioning during the liquid to steam transition.
Three documents provide guidance for iodine scrubbing ratios for steam generator tube ruptures or leaks; NUREG-0800, Standard Review Plan; NUREG-0844, NRC Integrated Program for the Resolution of Unresolved Safety Issues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity; and NUREG/CR-2659 PNL 3794, February 1983, Iodine Transport Predicted for a Postulated Steam Line Break with Concurrent Rupture of Steam Generator Tubes.
Section 15.6.3, Radiological Consequences of Steam Generator Tube Failure (PWR), of NUREG-0800 page 15.6.3-4 provides guidance for iodine transport from the steam generator. The section states that an iodine-partitioning coefficient of 100 may be conservatively assumed.
NUREG-844 states the effective partitioning factor will depend on whether the Steam generator water level is above or below the rupture location(s). If the rupture is above the water level then no partitioning factor is assumed. It is conservative to assume if any tubes are uncovered, then the rupture is uncovered.
NUREG/CR-2659 states with the SG water level below the dryer separators, the dryers and separators would be effective in capturing most of the droplets carrying iodine. A value of 0.001 is taken in this situation. The assumption of no partitioning for the cases with the dryers and separators covered is assumed.
In an Evaluation of Iodine/Noble Gas Ratios in SG Mr. M. Agen to Mr. D. A.
Schuelke it is concluded that an iodine to noble gas ratio of 0.1 should be used for the ratio from the RCS activity leaving the core.
From the above information it is concluded that a I/NG ratio of 0.001 should be used is the SG tubes are covered and the dryer and separator are not covered.
If the SG tubes are uncovered or the dryer and separator are covered then it is conservative to use the iodine to noble gas ratio of 0.1 due to the ratio of activities leaving the core.
II. RESOLUTION:
: 1. Use F3-20 to determine the correct I/NG ratio during a Steam Generator Tube Rupture accident.
Page 1 of 3                              TAB 24
 
Steam Generator Iodine Partitioning Rev. 0
: 2. Use the attached Attachment E to 1ES-3.1 during adverse containment conditions (Containment Pressure > 5 psig or Containment Radiation >1E4 r/hr) to determine if the SG water level is in the Acceptable Region.
III.
 
==REFERENCES:==
: 1. F3-20, Determination of Radioactive Release Concentrations
: 2. 1ES-3.1 (2ES-3.1)
: 3. NUREG-0800, Standard Review Plan
: 4. NUREG-0844, NRC Integrated Program for the Resolution of Unresolved Safety Issues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity
: 5. NUREG/CR-2659 PNL 3794, February 1983, Iodine Transport Predicted for a Postulated Steam Line Break with Concurrent Rupture of Steam Generator Tubes Page 2 of 3                              TAB 24
 
Steam Generator Iodine Partitioning Rev. 0 l ES-3 . 1 REV . 11 Page 1 of 1 SG vlide Range leve l  (%)
90                                                          I I I I I I I I I I I I  I I I I I I I I I ATT AC HM ENT E        .........
87 .5 85                                                0verfi 11 Region 82 .5 80 77 . 5                    Acceptable        ~~
Region 75 72 .5
                                    ~
                                          "~
70                    Tube Uncovery Region                  ~
67 .5 65              Thi s Figure To Be Used For Adverse Con ta i nment Conditi ons 62 .5 60 0    100    200    300  400 500 600 700            800      900      1000 1100 SG Pre ssure (ps ig )
ATTACHMENT E Wide Range SG Level For Controlling Inventory Page 3 of 3                                          TAB 24
 
REC RPSS 25                            CORE DAMAGE ASSESSMENT Taken from F3-17, Core Damage Assessment 5.0    DISCUSSION The approach utilized in this methodology of core damage assessment is measurement of fission product concentrations in the primary coolant system, and containment, when applicable, utilizing the post accident sampling system.
Certain nuclides have been selected to be associated with each particular core damage state, i.e., clad damage, fuel overheat and fuel melt. These nuclides reach equilibrium quickly within the fuel cycle. Once equilibrium condition are reached, a fixed inventory of the nuclides is assumed to exist within the fuel pellet. For these nuclides which reach equilibrium, their relative ratios within the fuel pellet can also be considered to be constant. During operation, certain volatile fission products collect in the gap. The relative ratios in the gap can also be considered to be constant, however, the distribution of the nuclides in the gap is not in the same proportion as the fuel pellet inventory since the migration of each nuclide into the gap is dependent on its particular diffusion rate. The relative ratios of the nuclides analyzed during an accident may be compared to the predicted relative ratios existing in the gap and fuel pellet to determine the source of the fission product release, i.e., gap release or fuel pellet.
Clad damage is characterized by the release of these fission products, i.e., isotopes of the noble gases, iodine, and cesium which have accumulated in the gap and during the operation of the plant. When the cladding ruptures, it is assumed that the fission product gap inventory of the damaged fuel rods is instantaneously released to the primary system. For this methodology it is assumed that the noble gases will escape through the break of the primary system boundary to the containment atmosphere and the iodines will stay in solution and travel with primary system water during the accident.
Fission product release associated with overtemperature fuel conditions arises initially from the portion of the noble gas, cesium and iodine inventories that was previously accumulated in grain boundaries. In addition, small amounts of the more refractory elements, barium-lanthanum, and strontium are also released.
Fuel pellet melting leads to rapid release of many noble gases, halides, and cesiums remaining in the fuel after overheat conditions. Significant release of the strontium, barium-lanthanum chemical groups is perhaps the most distinguishing feature of melt release conditions.
Auxiliary indicators such as core exit thermocouples, reactor vessel water level, reactor coolant loop radiation monitors, containment radiation monitors, and the containment hydrogen concentration are available for estimating core damage. These indications should confirm the core damage estimates which in turn are based on the radionuclide analysis.
Page 1 of 1                                  TAB 25, REV 17
 
REC RPSS 26          RADIATION MONITOR CONTROL FUNCTIONS RADIATION MONITOR CONTROL FUNCTIONS Channel No. Train          Title                      Control Function R-1  B    Control Room Area Monitor                None 1R-2 B      Unit 1 Containment Area Monitor          None 2R-2 A      Unit 2 Containment Area Monitor          None R-3 B      Radiochem Lab Area Monitor                None R-4 B      Charging Pump Area Monitor                None (11,12,13)
R-5  B    Spent Fuel Pool Area Monitor              None R-6  B    Sampling Room Area Monitor                None 1R-7  B    Unit 1 lncore Seal Table Area            None 2R-7  A    Unit 2 lncore Seal Table Area            None Monitor R-8 B
* Waste Gas Valve Gallery Area              None 1R-9 B      Unit 1 Letdown Line Area Monitor          None 2R-9 A      Unit 2 Letdown Line Area Monitor          None 1R-11B      Unit 1 Containment Particulate    Isolates Containment Monitor                            Purge and In-Service Purge 2R-11A      Unit 2 Containment Particulate    Isolates Containment Monitor                            Purge and In-Service Purge 1R-12B      Unit 1 Containment Gas            Isolates Containment Monitor                            Purge and In-Service Purge 2R-12 A    Unit 2 Containment Gas            Isolates Containment Monitor                            Purge and In-Service Purge 1R- 15B    Unit 1 Condenser Air Ejector Gas          None Monitor 2R- 15A    Unit 2 Condenser Air Ejector Gas          None Monitor R-16 B      Containment Fan Coil (12,14,22,24)        None Cooling Water Discharge Monitor R-18 B      Waste Disposal System Liquid      Closes Waste Liquid Effluent Monitor                  Common Discharge Header Valve 1R-19 B    Unit 1 Steam Generator            Closes SGB to Flash Tank Slowdown Monitor                  Valves and Flash Tank to River valve Page 1 of 6                TAB 26, REV 26
 
REC RPSS 26          RADIATION MONITOR CONTROL FUNCTIONS RADIATION MONITOR CONTROL FUNCTIONS Channel No. Train        ~                          Control Function 2R-19 A    Unit 2 Steam Generator            Closes SGB to Flash Tank Slowdown Monitor                  Valves and Flash Tank to River valve R-21 A      Circulating Water Discharge              None Monitor 1R-22 A    Unit 1 Shield Bldg Vent Gas      Isolates Containment Monitor                          Purge and In-Service Purge 2R-22 B    Unit 2 Shield Bldg Vent Gas      Isolates Containment Monitor                          Purge and In-Service Purge R-23 A      Control Room Air Supply          Closes 121 and 122 CR Monitor A                        outside air supply dampers; closes CR exhaust steam exclusion damper, 121 CR Cleanup Fan R-24 B      Control Room Air Supply          Closes 121 and 122 CR Monitor B                        outside air supply dampers, closes CR exhaust steam exclusion damper, starts 122 CR Cleanup Fan R-25 A      Spent Fuel Pool Air Monitor      Stops 121 SFP normal supply and exhaust fans, isolates Units 1 *and 2 In-Service Purge, starts 121 Spent Fuel and In-Service Purge exhaust fan R-26 A      RHR Cubicle Air Monitor                  None (11,21)
R-27 B      RHR Cubicle Air Monitor                  None (12,22)
R-28 A      New Fuel Pit Criticality                None Area Monitor Page 2 of 6              TAB 26, REV 26
 
REC RPSS 26        RADIATION MONITOR CONTROL FUNCTIONS RADIATION MONITOR CONTROL FUNCTIONS Channel No. Train                ~                  Control Function R-29. B  Shipping and Receiving                    None Area Monitor 1R-30 B  Unit 1 Aux Bldg Vent Gas          Starts 122 Aux Bldg Monitors                          Special Exhaust Fan 2R-30 B  Unit 2 Aux Bldg Vent Gas          Starts 122 Aux Bldg Monitor B                          Special Exhaust Fan, Closes Gas Decay Tank Release Isolation Valve R-31 B    Spent Fuel Pool Area Monitor      Stops 121 SFP normal supply and exhaust fans, isolates units 1 and 2 In-Service Purge, starts 122 Spent Fuel and In-Service Purge Exhaust Fan R-32 A    Rad Waste Bldg Control                    None Station Area Monitor R-33 A    Rad Waste Bldg Second Floor              None Area Monitor R-34 B    Rad Waste Bldg Cement Dump                None Area Monitor (Abandoned)
R-35 B    Rad Waste Bldg Vent Gas                  None Monitor R-36 B    Charging Pump Area Monitor                None (21,22,23) 1R-37 A  Unit 1 Aux Bldg Vent Gas          Starts 121 Aux Bldg Monitor A                          Special Exhaust Fan 2R-37 A  Unit 2 Aux Bldg Vent Gas          Starts 121 Aux Bldg Monitor A                          Special Exhaust Fan, Closes Gas Decay Tank Release Isolation Valve R-38 A    Containment Fan Coil                      None (II, 13,21,23) Cooling Water Discharge Monitor Page 3 of 6              TAB 26, REV 26
 
REC RPSS 26        RADIATION MONITOR CONTROL FUNCTIONS RADIATION MONITOR CONTROL FUNCTIONS Channel No. Train                TiUe              Control Function 1R-39 A    Unit 1 Component Cooling        Closes CC Surge Tank System Monitor                  Vent Valve 2R-39 B    Unit 2 Component Cooling        Closes CC Surge Tank System Monitor                  Vent Valve R-41 B    Waste Gas High Activity Loop          None Monitor R-42 NA    Heating Boiler Dearator              None (Abandoned)
Rad Monitor R-43 NA    11-12 Filter Demin Rad Monitor        None (Abandoned)
R-44 NA    12-13 Filter Demin Rad Monitor        None(Abandoned)
R-45 NA    21-22 Filter Demin Rad Monitor        None (Abandoned)
R-46 NA    22-23 Filter Demin Rad Monitor        None (Abandoned)
R-47 NA    121 Spent Resin Transfer Tank        None (Abandoned) 1R-48 B    Unit 1 Containment High Range        None Area Monitor 2R-48 B    Unit 2 Containment High Range        None Area Monitor 1R-49 A    Unit 1 Containment High Range        None Area Monitor 2R-49A    Unit 2 Containment High Range        None Area Monitor 1R-50 B    Unit 1 High Range Shield Bldg        None Vent Gas Monitor 2R-50A    Unit 2 High Range Shield Bldg        None Vent Gas Monitor 1R-51 NA  11 Main Steam Loop Rad Monitor        None 2R-51 NA  21 Main Steam Loop Rad Monitor        None 1R-52 NA  12 Main Steam Loop Rad Monitor        None 2R-52 NA  22 Main Steam Loop Rad Monitor        None RE-18305 NADrum Conveyor 36 in. Height          None Monitor RE-18306 NADrum Conveyor 2 in. Height            None Monitor RE-29003 NA121 Waste Monitoring Tank              None Area Monitor Page 4 of 6            TAB 26, REV 26
 
REC RPSS 26      RADIATION MONITOR CONTROL FUNCTIONS RADIATION MONITOR CONTROL FUNCTIONS Channel No. Train              nue                      Control Function 1R-53 NA  Unit 1 SI Pump Area, Monitor                None 2R-53 NA  Unit 2 SI Pump Area Monitor                  None 1R-54 NA  Unit 1 CS Pump Area Monitor                  None 2R-54 NA  Unit 2 CS Pump Area Monitor                  None 1R-55 NA  Unit 1 Aux Bldg 695 East Area Monitor        None 2R-55 NA  Unit 2 Aux Bldg 695 West Area Monitor        None 1R-56 NA  Unit 1 Aux Bldg 695 West Area Monitor        None 2R-56 NA  Unit 2 Aux Bldg 695 East Area Monitor        None 1R-57 NA  Unit 1 Aux Bldg 715 East Area Monitor        None 2R-57 NA  Unit 2 Aux Bldg 715 West Area Monitor        None 1R-58 NA  Unit 1 Aux Bldg 715 West Area Monitor        None 2R-58 NA  Unit 2 Aux Bldg 71 & East Area Monitor      None 1R-59 NA  Unit 1 Aux Bldg 715                          None Penet./1.tdn Area Monitor 2R-59 NA  Unit 2 Aux Bldg 715                          None Penet./Ltdn Area Monitor 1R-60 NA  Unit 1 Aux Bldg 735 North                    None Area Monitor 2R-60 NA  Unit 2 Aux Bldg 735 North                    None Area Monitor 1R-61 NA  Unit 1 A Sim Line Area Monitor              None 2R-61 NA  Unit 2 A Sim Line Area Monitor              None 1R-62 NA  Unit 1 Aux Bldg 755 East Area Monitor        None 2R-62 NA  Unit 2 Aux Bldg 755 West Area Monitor        None 1R-63 NA  Unit 1 Aux Bldg 755 West Area Monitor        None 2R-63NA  Unit 2 Aux Bldg 755 East Area Monitor        None 1R-64 NA  Unit 1 Turb Bldg 735 North Area Monitor      None 2R-64 NA  Unit 2 Turb Bldg 735 North Area MonitorNone 1R-65 NA  Operations Support Center Area Monitor        None 1R-66 NA  0 1 Diesel Gen. Room Area Monitor            None 2R-67 NA  Instruments & Control Shop Area Monitor      None 2R-68 NA  Technical Support Center Area Monitor        None 2R-69 NA  Guardhouse Area Monitor                      None 2R-72 NA  06 Bldg 707' Cable Spreading Room            None Area Monitor 2R-73 NA  06 Bldg 716' Bus 26 Room Area Monitor        None Page 5 of 6                  TAB 26, REV 26
 
REC RPSS 26    RADIATION MONITOR CONTROL FUNCTIONS RADIATION MONITOR CONTROL FUNCTIONS Channel No. Train                                  control Function 2R-74 NA  06 Bldg 735' Bus 221/222 Room Area Monitor None Page 6 of 6              TAB 26, REV 26
 
REC RPSS 27          SAFETY INJECTION SIGNAL / CONTAINMENT ISOLATION SIGNAL /
CONTAINMENT VENTILATION ISOLATION SIGNAL /
CONTROL ROOM VENTILATION ISOLATION SIGNAL SAFETY INJECTION SIGNAL The safety injection signal initiates 18 actions. The full explanation of safety injection actions can be found in reference B18C; Engineered Safeguards System. Those actions important to the REC and RPSS positions are listed below:
* A reactor trip is initiated
* A Containment Isolation signal is generated (see below)
* A Containment Ventilation Isolation signal is generated (see below)
* A Control Room Ventilation Isolation signal is generated (see below)
* The Shield Building Special Ventilation System starts
* The Auxiliary Building Special Ventilation System starts and the Auxiliary Building Normal Ventilation System stops CONTAINMENT ISOLATION SIGNAL The containment isolation signal initiates 11 actions. The full explanation of containment isolation actions can be found in reference B18C. Those actions important to the REC and RPSS positions are listed below:
* Letdown isolates (R-9 reading no longer reflects the current RCS activity)
* The steam generator blowdown isolation valves close (R-19 reading no longer reflects the current steam generator blowdown activity)
* The containment vacuum breaker closes (unless a negative pressure exists in containment)
* The pressurizer sample line isolates
* R-11 and R-12 isolate (R-11 and R-12 readings no longer reflect the current containment atmosphere activity)
CONTAINMENT VENTILATION ISOLATION SIGNAL The containment ventilation isolation signal initiates 2 actions. The full explanation of containment ventilation isolation actions can be found in reference B18C. Those actions important to the REC and RPSS positions are listed below:
* The containment purge supply and exhaust dampers close
* The in-service purge supply and exhaust dampers close CONTROL ROOM VENTILATION ISOLATION SIGNAL The control room ventilation isolation signal initiates 2 actions. The full explanation of control room ventilation isolation actions can be found in reference B18C. Those actions important to the REC and RPSS positions are listed below:
* The Control Room intake duct damper closes
* The 121 and 122 Control Room cleanup fans and chillers start Page 1 of 1                        TAB 27, REV 17
 
REC RPSS Calculating Field Gas Activities Please reference the tables in Prairie Island EPIP F3-15 and Monticello EPIP A.2-410 for information on calculation of field gas activities.
Page 1 of 1                            TAB 28, REV 19
 
REC RPSS 29 Guidance for Selecting Geopolitical Subareas when Making a Protective Action Recommendation Attachment A of PINGP 577 Emergency Notification Report Form contains a table for selecting Geopolitical Subareas when making a Protective Action Recommendation (PAR). The table has, for wind speeds > 5 MPH, line items for a range of wind from in degrees. Each range of upwind directions lists the five affected downwind sectors and affected geopolitical subareas. Twelve of the 16 wind direction ranges bisect a small portion of a geopolitical sub-area on the outermost sector and that geopolitical subarea is not listed as one of the affected areas at either 5 miles and/or 10 miles. The table in PINGP 577 is consistent with EP implementing procedure F3-8 Recommendations for Offsite Protective Actions.
PINGP utilizes 5 downwind sectors for evacuation. In the 1980s the Geopolitical Sub-Areas and which areas were affected by which downwind sectors were determined. The geopolitical areas were determined in conjunction with the downwind sectors and were agreed to by the states of Minnesota and Wisconsin. The site and the states realized that certain slivers of geopolitical subareas were bisecting a downwind sector at the outer edge of the five downwind sectors. All entities agreed that the PARs would not automatically include the geopolitical subareas which had a small sliver if the area in one of the outermost downwind sector. The Geopolitical Sub-Areas are drawn to existing political jurisdictions and cannot conform exactly to downwind sectors. The parties involved (site and the states) review the boundaries periodically as part of the review of their emergency plans. These Geopolitical Sub-Areas boundaries have not changed since they were established in the 1980s.
Page 1 of 1                      TAB 29, Rev. 21
 
REC RPSS 30 Guidance for spraying a radioactive plume with water Condition Evaluation CAP Number: 01522424 Description of problem:
This CAP documents a question resulting from the 5/17/2016 drill critique. During some drills one of the actions the EOF staff is tasked to do is bring in a fire truck to spray down a possible radioactive plume emitting from the shield building stack. It was questioned if spraying down the stack is an appropriate action for a drill or actual event.
It is not entirely clear that this is a desirable thing to do. Would spraying the plume down onto the aux building roof result in increased control room dose? Would spraying the plume right at the exit of the vent affect the flow characteristics of the vent system?
Would spraying the plume defeat the design basis for vent system as described in the USAR? It was also questioned if the RW fire dept. would even have the equipment to reach the exhaust point for the shield building vent system. If the station determines that this is a beneficial action to take, we should perform a modification to have a permanently installed spray system so the fire department can respond to other calls during an event.
Evaluation Details:
The PINGP Emergency Plan Section 5.6.4 states that one of the RW fire dept. tasks includes spraying to contain radiological release.
5.6.4 Local Support Services A. Fire Fighting The Red Wing Fire Department will provide assistance in the event of a fire occurring at the plant. The duties and responsibilities of the Plant Fire Brigade, insuring complete coordination with the Fire Department, are covered in the Operations Manual, Section F5, Fire Fighting.
The Red Wing Fire Department has various firefighting apparatus and water pumping equipment available for use. All Red Wing Fire Department apparatus can perform both fire fighting tasks, including rescue, and non-fire fighting tasks, including spraying to contain radiological releases Page 1 of 3 TAB 30, Rev. 25
 
REC RPSS 30 Guidance for spraying a radioactive plume with water Condition Evaluation From the letter of agreement with the City of Red Wing, LETTER OF AGREEMENT FOR EMERGENCY RESPONSE SERVICES BETWEEN THE CITY RED WING, CITY OF RED WING OFFICE OF EMERGENCY MANAGEMENT, CITY OF RED WING POLICE DEPARTMENT, CITY OF RED WING FIRE DEPARTMENT, AND NORTHERN STATES POWER COMPANY, A MINNESOTA CORPORATION
: 11. The Red Wing Fire Department has the capability to and will provide fire, rescue and other non-fire fighting services within the Red Wing Fire Department's Minnesota and Wisconsin service areas in PINGP's emergency planning zone. The Red Wing Fire Department has various firefighting apparatus, including pumpers and an aerial platform. All Red Wing Fire Department apparatus can perform both fire fighting tasks, including rescue, and non-fire fighting tasks, including spraying to contain radiological releases and pumping water into the plant for refilling and cooling purposes. In all cases, such operations can begin once the radiological and security threats are mitigated to insure the safety of both plant personnel and fire fighters. The Red Wing Fire Department facility shall be the location for the Emergency Worker Decontamination.
From these two documents its clear that there is intent to spray water on a radioactive release if its safe to do so in order to help contain the release. No PI generated calculation or study or other documentation could be located that quantified the effect of spraying water on a plume.
If you use URI (dose assessment software) and a postulated release from the shield building vent with a release rate of 2.8E9 Ci/sec the release is of sufficient size to exceed the protective action guidelines (PAGs) at the site boundary. This meets the definition of a General Emergency.
Using an input for heavy rain to simulate spraying the release you find that the TEDE value at the site boundary is higher than not spraying.
Precipitation          TEDE (mRem) SB None                  1.02E3 Heavy Rain            1.51E3 There is a difference of almost 500 mRem TEDE. (32%)
Page 2 of 3 TAB 30, Rev. 25
 
REC RPSS 30 Guidance for spraying a radioactive plume with water Condition Evaluation The ground deposition is also higher with heavy rain due to the water stripping radionuclides out of the plume. Because more radionuclides are removed close to the discharge site there are fewer contributing to accumulated dose downwind. In the example used the TEDE at 10 miles with no precipitation is higher than the TEDE with heavy rain.
Precipitation        TEDE (mRem) 10m None                36 Heavy Rain          27 The difference is 9 mRem TEDE. (25%)
The actual difference in TEDE is not high and the accumulated dose at the discharge site is not enough to impact emergency response actions. There is a measurable benefit to the public and the method of spraying water referenced in the Emergency Plan and practiced by the ERO should continue. The probability of needing to perform this action is very low and is not required by any regulation or licensing bases. No plant modification is recommended at this time.
Actions:
OTHA 01522424-02 Communicate the CE results to ERO duty teams.
Preparers Name: Brian J Carberry                                  Date: 6/22/2016 Page 3 of 3 TAB 30, Rev. 25}}

Revision as of 13:51, 17 January 2022

Independent Spent Fuel Storage Installation (ISFSI) - Revised Emergency Plan (E-Plan), Emergency Plan Implementing Procedures (EPIP) and Rec/Rpss Handbook
ML21139A311
Person / Time
Site: Prairie Island  Xcel Energy icon.png
Issue date: 05/19/2021
From:
Northern States Power Company, Minnesota, Xcel Energy
To:
Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
Shared Package
ML21139A309 List:
References
Download: ML21139A311 (428)
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