View source for NG-04-0093, Revision 19 to EAL Table of Contents, Rev. 8 to EAL EBD-A, Rev. 5 to EAL EBD-F, Rev. 8 to EAL EBD-H, and Rev. 6 to EAL EBD-S

{{Adams
| number = ML040550092
| issue date = 02/12/2004
| title = Revision 19 to EAL Table of Contents, Rev. 8 to EAL EBD-A, Rev. 5 to EAL EBD-F, Rev. 8 to EAL EBD-H, and Rev. 6 to EAL EBD-S
| author name = Peifer M
| author affiliation = Nuclear Management Co, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000331
| license number = DPR-049
| contact person = 
| case reference number = NG-04-0093, TAM-116, TAM-117, TAM-118
| document report number = EAL EBD-A, Rev 8, EAL EBD-F, Rev 5, EAL EBD-H, Rev 8, EAL EBD-S, Rev 6
| document type = Emergency Preparedness-Emergency Plan Implementing Procedures, Letter type:NG
| page count = 282
}}

=Text=
{{#Wiki_filter:NMCA Committed to Nuclear Exellence                                         Duane Arnold Energy Center Operated by Nuclear Management Company, LLC February 12, 2004                                                                      NG-04-0093 10 CFR 50.54 Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station 0-P1-17 Washington, DC 20555-0001 DUANE ARNOLD ENERGY CENTER DOCKET 50-331 LICENSE No. DPR-49 DUANE ARNOLD ENERGY CENTER EMERGENCY PLANNING DEPARTMENT PROCEDURE TRANSMITTAL ACKNOWLEDGEMENT MEMOs TAM-1 16,117 and 118 Enclosed is the Duane Arnold Energy Center EMERGENCY PLANNING DEPARTMENT PROCEDURE TRANSMITTAL ACKNOWLEDGEMENT MEMOs TAM-116, 117 and 118 which contain revisions to various Emergency Planning Department documents. These revisions have been reviewed pursuant to 10 CFR 50.54 (q) and it has been determined they do not decrease the effectiveness of the approved emergency plan.
This letter makes no new commitments or changes to any existing commitments.
Sincerely, Mark A. Peifer Site Vice President, Duane Arnold Energy Center

==Enclosures:==
DAEC EMERGENCY PLANNING DEPARTMENT PROCEDURE TRANSMITTAL ACKNOWLEDGEMENT MEMOs TAM-1 16,117 and 118 CC        Regional Administrator, USNRC, Region III (2 copies)
Project Manager NRC Resident Inspector 3277 DAEC Road
* Palo, Iowa 52324-9785
                                                                     -                          AcWS Telephone: 319.851.7611

NI\MC Committed to Nuclear Excellenf I
DAEC EMERGENCY PLANNING DEPARTMENT PROCEDURE TRANSMITTAL ACKNOWLEDGEMENT MEMO (TAM-1 16)
To:             NRC-NRR Document Control Desk US NRC Washington DC 20555 Re:             Entire EAL Basis Document (Table of Contents Rev) (Copy 91)
PSM

==Title:==
n/a Distribution Date:              02 /04 /2004 Effective Date of Change:       02/05/2004 Return by:                      02 / 19 / 2004 Please perform the following to your assigned manual. If you have any questions regarding this TAM please contact Don A. Johnson at 319-851-7872.
REMOVE              INSERT EAL Table of Contents Revision                                   Rev. 18             Rev. 19 EAL EBD-A (PWR: 24090)                                           Rev. 7              Rev. 8 EAL EBD-F (PWR: 24096)                                           Rev. 4              Rev. 5 EAL EBD-H (PWR: 24092)                                           Rev. 7              Rev. 8 EAL EBD-S (PWR: 24094)                                           Rev. 5              Rev. 6 PERFORMED BY:
Print Name                                   Sign Name                          Date Please return to:     K. Dunlap PSC/Emergency Planning 3313 DAEC Rd.
Palo, IA 52324 To be completed by DAEC EP personnel only:
Date TAM returned:
EPTools updated:                                                                             Page 1 of 1

N Committed to Nuclear Exceh ene Duane Arnold Energy Center Operated by Nuclear Management Company, LLC Tuesday, February 3, 2004 NRC-NRR Document Control Desk US NRC Washington, DC 20555 To: NRC-NRR Document Control Desk From: DAEC Emergency Planning Department Re: Description of changes to the following documents EAL               EBD-A               Abnormal Rad Levels/Radiological Effluent Category For EAL AA2: Added the IC for ARM Hi Rad Alarm back to ensure consistency with intial NRR approved EAL and NEI 99-01 rev 4.
EAL               EBD-F               Fission Product Barrier Degradation Category For the RPV Level column: remove refernce to RPV Levels that "cannot be restored" or have "no injection source available". These statements could cause a delay in classification and were specifically asked by NRR to not be used.
EAL               EBD-H               Hazards and Other Conditions Affecting Plant Safety Category For EAL HUI: Returned the IC for earthquakes back to the initial NRR approved wording to remove any potential overclassfication issues.
EAL               EBD-S               System Malfunction Category For EAL SU1: Removed the word 'unplanned' as this could potentially result in an underclassification issue and is inconsistent with NEI 99-01 rev 4.
For EAL SU6: Removed 'sound powered phones' as a viable routine communications method as this system is unreliable at DAEC.
For EAL SA3: Returned this EAL to the original NRR approved wording to remove any potential underclassification issues.
Please contact Paul Sullivan, Manager of Emergency Preparedness at DAEC, (319)851-7191, if you require further information.
3313 DAEC Road - Palo, lowa 52324-9785 Telephone: 319.851.7191 - Fax: 319.851.7364

EAL BASES DOCUMENT                          Rev. 19 INDEX                              Page 1 of 1 PROCEDURE                   TITLE                   REV #  REV. DATE Introduction                                1    2/01/2000 Definitions                                 3    11/21/2003 Organization of Basis Information           3    8/5/2003 EBD-A   Abnormal Rad Levels/Radiological Effluent   8    2/5/2004 Category EBD-F  Fission Product Barrier Degradation         4    2/5/2004 Category EBD-H   Hazards and Other Conditions Affecting      8    2/5/2004 Plant Safety Category EBD-S  System Malfunction Category                 5    2/5/2004 EBD-E  ISFSI Abnormal Events Category              0    8/5/2003

   \
                ^l EAL BASES DOCUMENT.
                                                       .e..-               I
                                                                           .EBD-A Rev. 8 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT CATEGORY                            PAGE 1 of 27 Usage Level Reference Use                 j lEffective Date:   >2 /&sect; /og                   l TECHNICAL REVIEW Prepared by:      Qo-I      A.4LO                        Date:    1 1tz./o Reviewed by:                                             Date:   c1 ;fq' o Independent Ril     wer Reviewed by:                                              Date:       I lze ((q O~perations Reviewer PROCEDURE APPROVAL I am responsible for the technical content of this procedure and for obtaining the necessary approval from the State and County Emergency Management officials prior to implementation.
Documentation of State and County Emergency Management approval is via NEP-2004-0005.
Approved by:           y     :      _                     Date:      I Manager, Emergency Planning                        I  /

EBD-A EAL BASES DOCUMENT-                                                               E      -
                           .         -         -         i;-io                        -         -eva.          8 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT CATEGORY                                                    PAGE 2 of 27j-Table of Contents AU1 - Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment That Exceeds Two Times the Offsite Dose Assessment Manual (ODAM) and is Expected to Continue For 60 Minutes or Longer ..........................................................                                             A-3 AU2 - Unexpected Increase in Plant Radiation ................................................. A-8 AA1 - Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 200 Times the Offsite Dose Assessment Manual (ODAM) and is Expected to Continue for 15 Minutes or Longer .........................................................                                             A-1 1 AA2 - Major Damage to Irradiated Fuel or Loss of Water Level that Has or Will Result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel                                            ......................................................... A -16 AA3 - Release of Radioactive Material or Increases in Radiation Levels Within the Facility That Impedes Operation of Systems Required to Maintain Safe Operations or to Establish or to Maintain Cold Shutdow n                                          ......................................................... A -19 AS1 - Site Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 mrem TEDE or 500 mrem CDE Thyroid for the Actual or Projected Duration of the Release ................. A-21 AGI - Site Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity that Exceeds 1,000 mrem TEDE or 5,000 mrem CDE Thyroid for the Actual or Projected Duration of the Release .........................................................                                            A-24

EBD-A EAL BASES DOCUMENT Rev. 8 ABNORMAL RAD LEVELSIRADIOLOGICAL EFFLUENT CATEGORY                                PAGE 3 of 27 AU1 Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment That Exceeds Two Times the Offsite Dose Assessment Manual (ODAM) Limit and is Expected to Continue For 60 Minutes or Longer EVENT TYPE: Offsite Rad Conditions OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE: (I or 2 or 3 or 4)
: 1. Valid Reactor Building ventilation rad monitor (Kaman 3/4, 5/6, 7/8) or Turbine Building ventilation rad monitor (Kaman 1/2) reading above I E-3   [tCi/cc and is expected to continue for 60 minutes or longer.
OR Valid Offgas Stack rad monitor (Kaman 9/10) reading above 2.0 E-l pCi/cc and is expected to continue for 60 minutes or longer.
OR Valid LLRPSF rad monitor (Kaman 12) reading above 1.0 E-3 ptCi/cc and is expected to continue for 60 minutes or longer.
OR Valid GSW rad monitor (RIS-4767) reading above 3E+3 CPS and is expected to continue for 60 minutes or longer.
OR Valid RH-IRSW & ESW rad monitor (RM-1997)reading above 8E+2 CPS and is expected to continue for 60 minutes or longer.
OR Valid RHRSW & ESW Rupture Disc rad monitor (RM4268) reading above I E+3 CPS and is expected to continue for 60 minutes or longer.
OR
: 2. Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates in excess of 2 times ODAM limit and is expected to continue for 60 minutes or longer.
OR
: 3. Valid dose assessment indicating dose rates beyond the site boundary above 0.1 mremlhr TEDE and is expected to continue for 60 minutes or longer.
AU1

EBD-A, EAL BASES DOCUMENT                                                     .B.
Rev. 8 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT CATEGORY                                   PAGE 4 of 27 DAEC EAL INFORMATION:
Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results.
UNPLANNED, as used in this context, includes any release for which a radioactivity discharge permit was not prepared, or a release that exceeds the conditions (e.g.,
minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit. The EC/OSM should not wait until 60 minutes has elapsed, but should declare the event as soon as it is determined that the release duration has or will likely exceed 60 minutes. Also, if an ongoing release is detected and the starting time for that release is unknown, the EC/OSM should, in the absence of data to the contrary, assume that the release has exceeded 60 minutes.
The approach taken for calculation of gaseous radioactive effluent EAL setpoints includes use of the ODAM Table 3-2 source term computed by BWR-GALE for the DAEC Base Case. The release is assumed to be from a single release point. Multiple release points would be difficult to present as explicit EAL threshold values and in any case, are addressed by off-site dose assessment by MIDAS, which is the preferred method for determining this condition. The calculation methods for setpoint determination are from ODAM Section 3.4 and are based on Regulatory Guide 1.109 methodology. The table below lists the results of the gaseous effluent EAL calculations. The Kaman extended range capability is used because the General Electric Offgas Stack monitor has a limited range.
GASEOUS EFFLUENT EALS                                         l AU1

EBD-A EAL BASES DOCUMENT Rev. 8 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT CATEGORY                                      PAGE 5 of 27 Turbine Bldg (Kaman 1/2) and Offgas Stack Kaman 9110 Reactor Bldg (Kaman 3/4, 5/6, 7/8)
Maximum flow (CFM)          10,000                            72,000 Release Limits              Concentration     Release Rate    Concentration        Release Rate (ViCi/cc)         (uiCi/sec)      (pCi/cc)             (pCi/sec)
Tech Spec                   l .E-I            5.2E+5          6.2E-4               2.1 E+4 Unusual Event (2 x TS)      2.0E-1             1.OE+6         l .2E-3              4.2E+4 Alert (60 x TS)             6.OE+0            3.OE+7          3.7E-2               l .3E+6 LLRPSF Kaman      12 Maximum flow (CFM)          99,000 Release Limits              Concentration                                  Release Rate (tLCi/cc)                                      ([LCi/sec)
Tech Spec                   5.9E-4                                         2.8E+4 Unusual Event (2 x 1.OE-3                                                  5.6E+4 TS/ODAM)
Alert (200 x TS)             1.OE-l                                        5.6E+6 The off-gas stack is treated as an elevated release and the turbine building and reactor building vents are treated as mixed-mode releases. The ground level setpoints are taken from the default setpoint calculations from the quarterly surveillance tests performed by DAEC Chemistry teclnicians. Reactor Building, Turbine Building, LLRPSF (Low Level Radwaste Processing and Storage Facility) and Offgas Stack Noble Gas Monitor alarm setpoints are calculated based on achieving the Tech Spec/ODAM instantaneous release limit, assuming annual average meteorology as defined in the ODAM. The Tech Spec/ODAM Limit currently corresponds to a reactor building or turbine building ventilation alarm setpoint of 6.2 E-04 pCi/cc. The monitor alarm setpoint can be periodically adjusted but typically does not vary by much. The DAEC EAL therefore addresses valid radiation levels exceeding 2 times the alarm setpoint for greater than 60 minutes. Rounded off, this corresponds to I E-3 p.Ci/cc. The corresponding offigas stack monitor value is l.IE-l pICi/cc, rounded off to I E-l ltCi/cc. The Tech Spec Limit currently for the LLRPSF building ventilation alarm setpoint is 5.9 E-04 pCi/cc. The DAEC EAL therefore addresses valid radiation levels exceeding 2 times the alarm setpoint for greater thlan 60 minutes. This corresponds to I E-3 pCi/cc.
Technical specification setpoints for radioactive liquid radiation monitors are 10 times the 10 CFR 20 Appendix B, Table 2, Water Effluent Concentration (WEC) limits. It is the policy of DAEC to process all liquid radwaste so that no release of radioactive liquid to the environment is allowed. The radwaste effluent line which could be used as a batch release mechanism has a trip function that prevents exceeding the DAEC release limit, however, an EAL has been provided. The other pathways to the environment (RHRSW - to cooling tower, RHRSW - to discharge canal) have radiation monitors with readouts going to the AU1

EBD-A EAL BASES DOCUMENT Rev. 8 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT CATEGORY                                   PAGE 6 of 27 I--
Control Room. These systems could become contaminated if heat exchanger leaks develop; however, historically this has not occurred in the service water systems at DAEC. These monitors are displayed on panels I C02 and I C I0.
Reactor water is the likely source of contamination through the service water systems as opposed to floor drain, detergent drain, and chemical waste discharge. The floor drain and detergent drains go to Radwaste Processing and would be batch released to the Radwaste effluent discharge line (if such a release were to occur). The chemical discharge sump is normally a radioactivity clean system and is tested by Chemistry to ensure no contamination prior to discharging to the canal.
The setpoints for the three service water radiation effluent monitors vary because of differences in detector efficiencies and background. Setpoints based on the same reactor water sample are listed below to show the differences. The rounded off readings will be used for the EALs for ease of reading the monitor scales.
Monitor                       TS Limit       Reading          UE Level       Alert Level GSW                            1,555 CPS       1.5E+3 CPS      3E+3 CPS       33E+5 CPS     l RHRSW & ESW to cooling        413 CPS        413+2 CPS        813+2 CPS      8E+4 CPS RHRSW & ESW to                507 CPS        513+2 CPS        I E+3 CPS      I E+5 CPS Discharge Canal               _____
There are no significant deviations from the generic EALs. However, DAEC does not have a telemetered radiation monitoring system. As an alternative, use of field instruments was considered. It is not practical to establish an EAL based on field survey readings of 0.1 mr/hr for greater than 60 minutes because field instruments in use for emergency response do not have a threshold of detection to meet such criteria.
Hourly Whole Body Dose Corresponding to 2 x ODAMI Limit for Gaseous Release ODAM limit = 500 mrem/year Whole Body Dose (10CFR20, Appendix B limit = 50 mrem/year Whole Body Dose) 2 x ODAM limit = [2 x 500 mrem/year]/8760 hours/year = 0.114 mrem Whole Body, in one hour Rounded off to 0.1 mrem/hr Dose assessment using MIDAS is based on the EPA-400 methodology, e.g., use of Total Effective Dose Equivalent (TEDE). This is somewhat different from whole body dose from AUI

EBD-A EAL BASES DOCUMENT Rev. 8 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT CATEGORY                                PAGE 7 of 27 gaseous effluents determined by ODAM methodology which forms the basis for the radiation monitor readings calculated in accordance with the generic methodology. The gaseous effluent radiation monitors can only detect noble gases. The contribution of iodine's to TEDE could therefore only be determined either by: (I) utilizing MIDAS, or (2) gaseous effluent sampling. DAEC EAL 4 is written in terms of TEDE and the gaseous effluent radiation monitor readings are determined based on ODAM.

==REFERENCES:==
: 1. Offsite Dose Assessment Manual Section 6.0, 6.1.2 and 7.1.2 Bases
: 2. Emergency Plan Implementing Procedure (EPIP) 3.3, Dose Assessment and Protective Action
: 3. Radiation Protection Calculation No. 95-001-C, Emergency Actions Levels Based on Effluent Radiation Monitors, January 24, 1995
: 4. UFSAR Section 11.5, Process and Effluent Radiation Monitoring and Sampling Systems
: 5. ATE!Ifetlhodologyfor Development of Emergency Action Levels NUAMRCANESSP-00 7 Revision 4, May 1999 AUI

EBD-A EAL BASES DOCUMENT Rev. 8 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT CATEGORY                                   PAGE 8 of 27 AU2 Unexpected Increase in Plant Radiation EVENT TYPE: Onsite Rad Conditions OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:                (I or 2)
: 1. Uncontrolled loss of reactor cavity or fuel pool water level with all spent fuel assemblies remaining covered by water as indicated by ANY of the following:
* Report to the control room.
* Valid fuel pool level indication (LI-3413) below 36 feet and lowering.
* Valid WR GEMAC Floodup indication (LI-4541) coming on scale.
OR
: 2. Unexpected ARM reading offscale ligh or above 1000 times normnal* readings.
* Normal levels can be considered as the highest reading in the past twenty-four hours excluding the current peak value.
DAEC EAL INFORMATION:
There are no significant deviations from the generic EALs. DAEC does not have a spent fuel transfer canal.
Uncontrolledmeans that the condition is not the result of planned actions by the plant staff in accordance with procedures. Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results.
There are three methods to determine water level decreases of concern. The first method is by report to the control room. The other methods include use of the Floodup level indicator and the spent fuel pool level indicator. These are further described below.
During preparation for reactor cavity flood up prior to entry into refuel mode, reactor vessel level instrument LI-4541 (WR GEMAC, FLOODUP) on control room panel IC04 is placed AU2

EBD-A EAL BASES DOCUMENT Rev. 8 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT CATEGORY                                    PAGE 9 of 27 in service by I&C personnel connecting a compensating air signal after the reference leg is disconnected from the reactor head. Normal refuel water level is above the top of the span of this flood up level indicator. A valid indication (e.g., not due to loss of compensating air signal or other instrument channel failure) of reactor cavity level coming on span for this instrument is used at DAEC as an indicator of uncontrolled reactor cavity level decrease.
DAEC Technical Specifications require a minimum of 36 feet of 'water in the spent fuel pool. During refueling, the gates between the reactor cavity and the refueling cavity are removed and the spent fuel pool level indicator LI 3413 is used to monitor refueling water level. Procedures require that a normal refueling water level be maintained at 37 feet 5 inches. A lowv level alarm actuates when spent fuel pool level drops below 37 feet I inch.
Symptoms of inventory loss at DAEC include visual observation of decreasing water levels in reactor cavity or spent fuel storage pool, Reactor Building (RB) fuel storage pool radiation monitor or refueling area radiation monitor alarms, observation of a decreasing trend on the spent fuel pool water level indicator, and actuation of the spent fuel pool low water level alarm. To eliminate minor level perturbations from concern, DAEC uses LI 3413 indicated water level below 36 feet and lowering.
Increased radiation levels can be detected by the local refueling floor area radiation monitors, the refueling floor Continuous Air Monitor (CAM) alarm, refueling areas radiation monitors, fuel pool ventilation exhaust monitors, and by Standby Gas Treatment (SGBT) System automatic start. Applicable area radiation monitors include those that are displayed on Panel I C02 and alarmed on Panel I C04B. The DAEC EAL has also been written to reflect the case where an ARM may go offscale high prior to reaching 1,000 times the normal reading.
NOTE: On Annunciator Panel IC04B, the indicators listed below are expected alarms during pre-planned transfers of highly radioactive material through the affected area. If anl HP Technician is present, sending an Operator is not required. Radiation levels other than those expected should be promptly investigated. The indicators are high radiation alarms from the Hot Laboratory or Administrative Building, the new fuel storage area, and the radwaste building.
AU2

EBD-A EAL BASES DOCUMENT Rev. 8 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT CATEGORY                                PAGE 10 of 27p'

==REFERENCES:==
: 1. Alarm Response Procedure (ARP) IC04B, Reactor Water Cleanup and Isolation
: 2. Technical Specification 3.7.8, Spent Fuel Pool Water Level
: 3. Emergency Plan Implementing Procedure (EPIP) 3.1, Inplant Radiological Monitoring, Attachment 1, ARM Locations
: 4. Emergency Operating Procedures (EOP) Basis Document, Breakpoints for RC/L & L
: 5. Surveillance Test Procedure (STP) 3.0.0.0-OIPA, Daily and Shift Instrument Checks
: 6. Integrated Plant Operating Instruction (IPOI) 8 , Outage and Refueling Operations
: 7. Core Alterations, RFP403, Procedure for Moving Core Components Between Reactor Core and Spent Fuel Pool, Within the Reactor Core, or Within the Spent Fuel Pool
: 8. NEI MIethodologyfor Development of Emergency Action Levels NUAJARC'/NE SP-007 Revision 4, May 1999 AU2

EBD-A EAL BASES DOCUMENT Rev. 8 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT CATEGORY                                   PAGE 11 of 27 AA1 Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 200X the Offsite Dose Assessment Manual (ODAM) Limit and is Expected to Continue for 15 Minutes or Longer EVENT TYPE: Offsite Rad Conditions OPERATING MOD)E APPLICABILITY: All EAL THIRESIIOLD VALUE:               (I or 2 or 3 or 4)
: 1. Valid Reactor Building ventilation rad monitor (Kaman 3/4, 5/6, 7/8 ) or Turbine Building ventilation rad monitor (Kaman 1/2) reading above 3 1.-2 jiCi/cc and is expected to continue for 15 minutes or longer.
OR Valid Offgas Stack rad monitor (Kaman 9/10) reading above 6 E+-0 jCi/cc and is expected to continue for 15 minutes or longer.
OR Valid LLRPSF rad monitor (Kaman 12) reading above I E-l tCi/cc and is expected to continue for 15 minutes or longer.
OR Valid GSW rad monitor (RIS-4767) reading above 3E+5 CPS and expected to continue for 15 minutes or longer.
OR Valid RHIRSW & ESW rad monitor (RM-1997) reading above 8E+4 CI'S and expected to continue for 15 minutes or longer.
OR Valid RH-IRSW & ESW Rupture Disc rad monitor (RM-4268) reading above I E+5 CPS and expected to continue for 15 minutes or longer.
OR
: 2. Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates writh a release duration expected to continue for 15 minutes or longer in excess of 200 times ODAM limit.
OR AA1

EBD-A EAL BASES DOCUMENT Rev. 8 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT CATEGORY                                   PAGE 12 of 27 1-
: 3. Valid site boundary radiation reading of greater than 10 mremlhr above normal background and is expected to continue for 15 minutes or longer.
OR
: 4. Valid dose assessment indicating dose rates beyond the site boundary above 10 mrenl/hr TEDE and is expected to continue for 15 minutes or longer.
DAEC EAL INFORNM TION:
Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results. In a case where data from Kaman readings is being used to determine whether an EAL threshold value has been exceeded, Valid means that flow through the associated Kaman Monitor has been verified and does exist as indicated in
[tCi/sec on SPRAD.
UNPLANNED, as used in this context, includes any release for which a radioactivity discharge permit was not prepared, or a release that exceeds the conditions (e.g.,
minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit. The EC/OSM should not wait until 15 minutes has elapsed, but should declare the event as soon as it is determined that the release duration has or wvill likely exceed 15 minutes. Also, if an ongoing release is detected and the starting time for that release is unknown, the EC/OSM should, in the absence of data to the contrary, assume that the release has exceeded 15 minutes.
AA1

EBD-A EAL BASES DOCUMENT Rev. 8 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT CATEGORY                                     PAGE 13 of 27 Gaseous Effluent EALs Offgas Stack Kaman 9/1 0                Turbine Bldg (Kaman 1/2) and Reactor Bldg (Kaman 3/4, 5/6, 7/8)
MaLximum flow (CFM)          10,000                                 72,000 Release Limits              Concentration        Release Rate       Concentration      Release Rate (iCifcc)         -(Ci/sec)              (LCi/cc)           (piCi/sec)
Tech Spec                    1.113-1             5.213+5            6.213-4            2.1 13+4 Unusual Event (2 x TS)      2.013-1               1.OE+6             1.213-3           4.213+4 Alert (60 x TS)             6.0E+0               3.OE+7             3.7E-2              1.31E+6 LLRPSF Kaman 12 Maximum flow (CFM)          99,000 Release Limits              Concentration               Release Rate (pCi/cc)                    (ylCi/sec)
Tech Spec                   5.913-4                     2.8E1+4 Untustual Event (2 x TS)     1.013-3                    5.6E1+4 Alert (200 x TS)             1.013- I                   5.6E+6 The off-gas stack is treated as an elevated release and the turbine building and reactor building vents are treated as mixed-mode releases. The ground level setpoints are taken from the default setpoint calculations from the quarterly surveillance tests pcrformed by DAEC Chemistry technicians. Reactor Building, Turbine Building, LLRPSF (Low Level Radwaste Processing and Storage Facility) and Offgas Stack Noble Gas Monitor alann setpoints are calculated based on achieving the Tech Spec instantaneous release limit assuming annual average meteorology as defined in the ODAM. The Tech Spec Limit currently corresponds to a reactor building or turbine building ventilation alarm sctpoint of 6.2 E-4 ,uCi/cc. The monitor alarm setpoint can be periodically adjusted but typically does not vary by much. For the Offgas Stack, Reactor Building and Turbine building KAMAN monitor readings, DAEC chose to multiply the technical specification concentration by a factor of 60 (instead of 200) in order to allow for a logical step progression in monitor setpoints from the AU 1 through AA 1 to AS 1. The DAEC EAL therefore addresses valid radiation levels exceeding 60 times the alarm setpoint for greater than 15 minutes. Rounded down, this corresponds to 3 E-2 pCi/cc. The corresponding offgas stack monitor value is 6.6 pCi/cc, rounded down to 6 13+0 p[Ci/cc. The Tech Spec/ODAM Limit currently for thle LLRPSF building ventilation alarm setpoint is 5.9 E-04 pLCi/cc. Thie DAI C EAL therefore addresses valid radiation levels exceeding 200 times the alarm setpoint for greater than 15 minutes. This corresponds to I E-1 pCi/cc.
Technical specification setpoints for radioactive liquid radiation monitors are 10 times the 10 CFR 20 Appendix B, Table 2, Water Effluent Concentration (WEC) limits. It is the policy of DAEC to process all liquid radwaste so that no release of radioactive liquid to the AA1

EBD-A EAL BASES DOCUMENT Rev. 8 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT CATEGORY                                    PAGE 14 of 27     l-environment is allowed. The radwaste effluent line which could be used as a batch release mechanism has a trip function that prevents exceeding the DAEC release limit, and therefore no EAL limits are provided. The other pathways to the environment (RHRSW - to cooling tower, RHIRSW - to discharge canal) have radiation monitors with readouts going to the Control Room. These systems could become contaminated if heat exchanger leaks develop; however, historically this has not occurred in the service water systems at DAEC.
These monitors are displayed on panels 1C02 and I C 10.
Reactor water is the likely source of contamination through the service water systems as opposed to floor drain, detergent drain, and chemical waste discharge. The floor drain and detergent drains go to Radwaste Processing and would be batch released to the Radwaste effluent discharge line (if such a release were to occur). The chemical discharge sump is normally a radioactivity clean system and is tested by Chemistry to ensure no contamination prior to discharging to the canal.
The setpoints for the three service wvater radiation effluent monitors vary because of differences in detector efficiencies and background. Setpoints based on the same reactor water sample are listed below to show the differences. The rounded off readings will be used for the EALs for ease of reading the monitor scales.
Monitor                      TS/ODAM Limit        Reading             UE Level            Alert Level GSW                           1,555 CPS            1.5E+3 CPS          3E+3 CPS            3E+5 CPS RFIRSW & ESW to cooling      413 CPS              4E+2 CPS            8E+2 CPS            8E+4 CPS tower RHRSW & ESW to               507 CPS              5E+2 CPS            I E+3 CPS           IE+5 CPS Discharge Canal DAEC does not have a telemetered radiation monitoring system. As an alternative, DAEC uses valid field survey readings outside the site boundary greater than 10 mr/hr or greater than 50 mr/hr CDE Thyroid.
Hourly Whole Body Dose Corresponding to 200 x ODAMNI Limit for Gaseous Release ODAM limit = 500 mrem/year Whole Body Dose (IOCFR20, Appendix B limit = 50 mremn/year Whole Body Dose) 200 x ODAE limit = [200 x 500 mrem/year]/8760 hours/year = 11.4 mrem Whole Body in one hour Rounded off to 10 mremlhr AA1

EBD-A EAL BASES DOCUMENT Rev. 8 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT CATEGORY                                PAGE 15 of 27 Dose assessment using MIDAS is based on the EPA-400 methodology, e.g., use of Total Effective Dose Equivalent (TEDE). This is somewhat different from whole body dose from gaseous effluents determined by ODAM methodology which forms the basis for the radiation monitor readings calculated in AUI in accordance with the generic methodology.
The gaseous effluent radiation monitors can only detect noble gases. The contribution of iodine's to TEDE could therefore only be determined either by: (I) utilizing MIDAS, or (2) gaseous effluent sampling. DAEC EAL 4 is written in terms of TEDE and the gaseous effluent radiation monitor readings are determined based on ODAM.

==REFERENCES:==
: 1. Offisite Dose Assessment Manual Section 6.0, 6.1.2 and 7.1.2 Bases
: 2. Emergency Plan Implementing Procedure (EPIP) 3.3, Dose Assessment and Protective Action
: 3. Radiation Protection Calculation No. 95-001 -C, Emergency Actions Levels Based on Effluent Radiation Monitors, January 24, 1995
: 4. UFSAR Section 11.5, Process and Effluent Radiation Monitoring and Sampling Systems
: 5. EPA 400-R-92-00 1, Alanual of ProtectiveAction Guides and J'rolecaiheActions/for Aluclear Incidents
: 6. ANEI AMetlhodology for Development of Emergency, Action Levels A'UAIA RCC/AESP-007 Revision 4, May 1999 AA1

EBD-A EAL BASES DOCUMENT Rev. 8 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT CATEGORY                                   PAGE 16 of 27 AA2 Major Damage to Irradiated Fuel or Loss of Water Level that Has or Will Result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel EVENT TYPE: Onsite Rad Conditions OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:                 (I or 2 or 3 or 4)
I. Report of either of the following:
* Valid ARM Fli Rad alarn for the Refueling Floor North End (RNM 9163), Refueling Floor South End (RM 9164), New Fuel Storage (RM 9153), or Spcnt Fuel Storage Area (RM 9178).
* Valid Refueling Floor North End (RM-91 63), Refueling Floor South End (RM-9164), or New Fuel Storage Area (RM-9153) ARM Reading above I0 mr/hr
* Valid Spent Fuel Storage Area ARM (RM-9178) Reading above 100 mr/hr OR
: 2. Report of Visual observation of irradiated fuel uncovered.
OR
: 3. Valid water level reading below 450" as indicated on LI-4541 (floodup) for the Reactor Refueling Cavity that will result in Irradiated Fuel uncovering.
OR
: 4. 4. Valid Fuel Pool water level indication (11-3413) below 16 feet that will result in Irradiated Fuel uncovering.
DAEC EAL INFORINIATION:
Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results. Valid alarms are solely due to damage to irradiated fuel or loss of water level that has or will result in the uncovering of irradiated fuel.
There are no significant deviations from the generic EALs. Increased radiation levels can be detected by the local radiation monitors, in-plant radiological surveys, new fuel and spent fuel storage area radiation monitor alarms displayed on panel 1C04B, fuel pool ventilation exhaust monitors, and by Standby Gas Treatment (SBGT) System automatic start.
AA2

EBD-A EAL BASES DOCUMENT Rev. 8 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT CATEGORY                                       PAGE 17 of 27 Applicable area radiation monitors include RM-9163, RM-9164, RM-9153, and RM-9178.
These monitors are located in the north end of the refuel floor, the south end of the refuel floor, the new fuel vault area, and near the spent fuel pool, respectively.
Per ARP IC04B, the applicable area radiation monitor alarms actuate when radiation levels increase above 100 mr/hr in the spent fuel pool area or above 10 mr/hr in the other three areas of concern. If a valid actuation of these alarms Nvere to occur, the refileling floor would be immediately evacuated. 'Thus, a report ofa fuel handling accident with either valid actuation of the fuel area alarms on panel I C04B or with measured radiation levels in the spent fuel pool or north fuel area are used to address the generic concern consistent with DAEC design and procedures.
During preparation for reactor cavity flood up prior to entry into refuel mode, reactor vessel level instrument 1,1-4541 (W\VR GEMAC, FLOODUP) oln control room panel IC04 is placed in service by I&C personnel connecting a compensating air signal after the reference leg is disconnected from the reactor head. Normal refuel water level is above the top of the span of this flood up level indicator. A valid on-scale indication (e.g., not due to loss of compensating air signal or other instrument channel failure) from this instrument can be used to determine uncontrolled loss of water level in the reactor cavity.
During refuieling, the gates between the reactor cavity and the refueling cavity are removed and the spent fuel pool level indicator LI 3413 is used to monitor refueling water level. This measures the common water level in the reactor cavity and the fuel pool. The bottom of the fitiel transfer slot between the spent fuel pool and the reactor cavity is 16 feet above the bottom of the spent fuel pool. The top of the active fuel in the spent fuel storage racks is slightly less than 13 feet 9 inches above the bottom of the spent fuel pool. Therefore, postulated failures which drain the reactor cavity through the reactor vessel cannot uncover Fuel in the spent fuel storage racks. HIowever, valid indication ofspent fuel pool level less than 16 feet would indicate that spent fuel in the storage racks may potentially become uncovered.
RF1P403 requires that upon a loss of water level situation, that the refueling crew on the refueling floor shall discharge any fuel assembly on the fuel grapple as follows:
* If a fuel assembly is currently being withdrawn from a slot in the core or spent fiuel pool, immediately reinsert it into that slot.
  *Ifia fuel assembly is being transferred and is still over or near the core, insert it into the closest available slot in the core.
* If a fuel assembly is being transferred and is over or near the spent fuel pool, insert it into the closest available slot in the spent fuel racks.
AA2

EBD-A EAL BASES DOCUMENT Rev. 8 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT CATEGORY                                   PAGE 18 of 27 Following these actions, the refueling floor is to be evacuated of all personnel. The DAEC EAL is written to address the generic concern that a spent fuel assembly was not fuilly covered by water. This can either be by visual observation of an uncovered spent fuel assembly or by trending fuel pool level in the control room if a spent fuel assembly could not be placed in a safe storage location specified by RFP 403 as described above.

==REFERENCES:==
: 1. Alarn Response Procedure (ARP) I C04B, Reactor Water Cleanup and Isolation
: 2. Technical Specification 3.7.8, Spent Fuel Pool Water Level
: 3. Emergency Operating Procedures (EOP) Basis Document, Breakpoints for RC/L & L
: 4. Emergency Plan Implementing Procedure (EPIP) 3.1, Inplant Radiological Monitoring, Attachment 1, ARM Locations
: 5. Surveillance Test Procedure (STP) 3.0.0.0-01, Daily and Shift Instrument Checks
: 6. Integrated Plant Operating Instruction (IPOI) 8 , Outage and Refueling Operations
: 7. Core Alterations, RFP403, Procedure for Moving Core Components Between Reactor Core and Spent Fuel Pool, Within the Reactor Core, or Within the Spent Fuel Pool
: 8. Bechtel Drawing C-492, Reactor Building - Reactor Well, Spent Fuel & Dryer-Separator Pool General Arrangement, Rev. 6
: 9. Bechtel Drawing C-493, Reactor Building - Spent Fuel Liner Plan Elevations and Details, Sheet 1, Rev. 6
: 10. Holtec International Drawing No. 1045, Rack Construction - Spent Fuel Storage Racks, Rev. 3
: 11. NEI Mfethodologyfor Development of Emtergency Action Levels NUMfARC/NVESP-007 Revision 4, May 1999 AA2

EBD-A EAL BASES DOCUMENT Rev. 8 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT CATEGORY                                     PAGE 19 of 27 AA3 Release of Radioactive Material or Increases in Radiation Levels Within the Facility That Impedes Operation of Systems Required to Maintain Safe Operations or to Establish or to Maintain Cold Shutdown EVENT TYPE: Onsite Rad Conditions OPEIRATING MODE APPLICA1I1LITY: All EAL THIRESIIOLI) VALUE:               (1 or 2)
: 1. Valid Control Room Area Radiation Monitor (RM-9162) reading above 15 mr/hr.
OR
: 2. Valic North CRD Module Area Radiation Monitor (RM-9 168) reading above 500) mr/hr.
affecting the Remote Shutdown Panel, 1C388.
I)AEC EAL INFORMATION:
Vatlidl means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results.
There are no significant deviations from the generic EALs. Per the UFSAR, the control room is the only area that is required to be continuously occupied to achieve and maintain safe shUtdoxvn following design basis accidents. The capability exists ior plant shutdown from outside the main control room in the event that the control room becomes uninhabitable using remote shutdown panel IC388. The RB 757 CRD North AIRM-9168 is in the vicinity of the Remote Shutdown Panel and is used to monitor radiation levels to determine habitability for that area.
Expected increases in monitor readings due to controlled evolutions (such as lifting thie steam dryer during refueling) do not result in emergency declaration. Nor should momentary increases due to events such as resin transfers or controlled movement of radioactive sources result in emergency declaration. In-plant radiation level increases that would result in emergency declaration, are also tOnplanne1Cd, e.g., outside the limits established by an existing radioactive discharge permit.
AA3

EBD-A EAL BASES DOCUMENT Rev. 8 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT CATEGORY                                PAGE 20 of 27

==REFERENCES:==
: 1. Alarm Response Procedure (ARP) 1C04B, Reactor Water Cleanup and Isolation
: 2. Abnormnal Operating Procedure (AOP) 913, Fire
: 3. Abnornal Operating Procedure (AOP) 914, Security
: 4. Abnormal Operating Procedure (AOP) 915, Shutdown Outside Control Room
: 5. Surveillance Test Procedure (STP) 3.0.0.0-01, Daily and Shift Instrument Checks
: 6. Integrated Plant Operating Instruction (IPOI) 8 , Outage and Refueling Operations
: 7. Emergency Plan Implementing Procedure (EPIP) 3.1, Inplant Radiological Monitoring
: 8. UFSAR Section 6.4, Habitability Systems
: 9. Bechtel Calculation DA-4, Project Number 265-002, Control Room Habitability, 9/3/80
: 10. NE! I fetlhodologyfor Development of Em nergency Action Levels NUM1RC/NERSII-007 Revision 4, May 1999 AA3

EBD-A EAL BASES DOCUMENT Rev. 8 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT CATEGORY                                    PAGE 21 of 27 ASI          Site Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 mrem TEDE or 500 mrem CDE Thyroid for the Actual or 1Projected l)uration of the Release EVENT TYP'E: Orfsitc Rad Conditions OPERATING MOI)E AlPP'LICABILITY: All EAL TIHRESHIOLI) VALUE: (I or 2 or 3)
I. Valid Reactor Building ventilation rad monitor (Kaman 3/4, 5/6, 7/8) or Turbinc Building ventilation rad monitor (Kaman 1/2) reading above 6 E-2 FICi/cc and is expected to continue for 15 minutes or longer. (Dose assessment not available.)
O1R Valid Offgas Stack rad monitor (Kaman 9/10) reading above 4 E+l pCi/cc and is expected to continue for 15 minutes or longer. (Dose assessmcnt not available)
OR
: 2. Field sunrcy results indicate site boundary dose rates exceeding 100 mrerlh/r expected to continue for more than one hour; or analyses of field survey samples indicate CDE Thyroid of 500 mrem for one hour of inhalation.
OR
: 3. I)ose assessment determines integrated accident dose projection outside the site boundary above 100 mrem TEDE or above 500 mrem CDE Thyroid.
I)AEC EAL INFORMATION:
ailidl means that the reading is from instrumentation determined to be operable in accordance wvith the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survcy results. In a case where data from Kaman readings is being used to determine whether an EAL threshold value has been exceeded, 'ialidlmeans that flow through the associated Kaman Monitor has been verified and does exist as indicated in
[tCi/sec on SI'RAD.
The preferred method for declaration of AS I is by means of Dose Assessment using the MIDAS computer model. Ilowever, if Kaman monitor readings are sustained for longer than 15 minutes and the required MIDAS dose assessments cannot be completed wvithin this period, then the declaration can be made using Kaman readings PROVIDEID the readings ASI

EBD-A EAL BASES DOCUMENT Rev. 8 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT CATEGORY                                   PAGE 22 of 27 are not from an isolated flow path. If Kaman readings are not valid, field survey results may be utilized.
DAEC's Meteorological Information and Dose Assessment System (MIDAS) wvzas utilized to determine the Kaman monitor limits. Eight separate combinations of release point, source term, meteorological conditions and equipment status were analyzed. Pathways considered were the offgas stack, the turbine building exhaust vent and a single reactor building exhaust vent. Multiple release points were not considered. In this same vein, it was assumed that only one of the three reactor building vents is on during the release.
The source terms used have been pre-loaded into MIDAS and are the default mixes associated with a loss of coolant accident (LOCA) and a control rod drop (CRD). The LOCA mix. was used in conjunction with a release via the offgas stack while the CRD mllix was used for releases via the turbine or reactor building vents. The source term for a release via the offgas stack is further impacted by the status of the standby gas treatment system.
The status of that system was also taken into consideration.
Based on 1995 data (NG-96-0987), the atmospheric stability was classified as Pascal E 33%
of the time. Consequently, both classifications were evaluated. Based on the stune report, the most common wind speeds were:
Pascal Class                         Altitude                    Speed (mph)
D                                    156'                        8-12 D                                    33'                         8-12 E                                    156'                        8-12 E                                    33'                         4-7 Though the temperature setting has no impact on the MIDAS calculations, a value must be entered in order for the program to run. Consequently, the temperature was arbitrarily set at 50 F.
The rain estimate was set at zero, to eliminate any on site washout of radioactive material.
For the first MIDAS runs a I Ci/cc concentration was assumed. The results of these runs were then normalized to the limits, thus generating a theoretical Kaman limit. Additional MIDAS runs were made with these theoretical limits as input to verify the normalization process. In addition to the total integrated dose, MIDAS calculates a peak whole body DDE rate resulting from the plume and a peak thyroid CDE rate resulting from inhalation.
Because the AS I and AGI KAMAN limits are to be based on a one-hour exposure, establishing concentration limits so these peak values match the NUMARC limits is acceptable.
ASI

EBD-A EAL BASES DOCUMENT Rev. 8 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT CATEGORY                                     PAGE 23 of 27 Site Area          Geneml Initiating Condition                                    Emergency'         Eiergency ASI                AGI Valid Turbine or Reactor Building ventilation rad monitor (KAMAN) reading for more than 15 0.06 [tCi/cc                      0.6 fCi/cc millutes above:
I)AEC does not have a telemetered radiation monitoring systemi. As an alternative, )AElC uses valid field survey readings outside the site boundary to determine il'doses are greater than 100 mrhAr TEDE or greater than 500 mr/lr CDE Thyroid.
Dose assessment using MIDAS is based on the EPA-400 methodology, e.g., use of Iotal Eflective Dose Equivalent (T[IDE) and Committed I)ose Equivalent (C)lE) Thyroid.
TEDE is somewhat different from whole body dose from gaseous effluents determ-ined by ODlAM methodology which forms the basis for the radiation monitor readings calculated in AU 1. These factors can introduce differences that are at least as large as those introduced by using T'll)E versus whole body dose. The gaseous elffluent radiation monitors can only detect noble gases. The contribution of iodine's to TEDE and C)E Th1yroid could therefore only be determnined either by: (1) utilizing the source term mixture in MII)AS, or (2) gaseous effluent sampling. Therefore, DAEC EAL Threshold \Talue 3 is xvritten in terms of' TEDE and CDE Thyroid.
REFERrENCES:
I. Olffsite Dose Assessment Manual, Section 6.0, 6.1.2 and 7.1.2, Bases
: 2. EImergency Plan Implementing Procedure (EPIP) 3.3, Dose Assessment and lProtective Action
: 3. Radiation Protection Calculation No. 95-001-C, Emergency Actions Levels Based on Iffluent Radiation Monitors, January 24, 1995
: 4. Radiation E-ngineering Calculation No. 96-007-A, Determination of DAIC Radioactive Release Initiating Conditions for AS I & AG I Emergency Classifications, July 3, 1996
: 5. UFSAR Section 11.5, Process and Effluent Radiation Monitoring and Sampling Systems
: 6. EPA 400-R-92-00 1, iA'anucl of Protective Action Guides and Ih-otectii e Actions fir Nuclear Incidents
: 7. NEI AlkthotiologyJfi Dei elopnient of Eniergency Action Lcvels NUA IA RCYNESP-00 7 RIeision 4, May 1999 ASI

EBD-A EAL BASES DOCUMENT Rev. 8 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT CATEGORY                                         24 of 27 AGI Site Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity that Exceeds 1,000 mrem TEDE or 5,000 mrem CDE Thyroid for the Actual or Projected Duration of the Release EVENT TYPE: Offsite Rad Conditions OPERATING NIODE APPLICABILITY: All EAL THRESHOLD VALUE: (1 or 2 or 3)
: 1. Valid Reactor Building ventilation rad monitor (Kaman 3/4, 5/6, 7/8) or Turbille Building ventilation rad monitor (Kaman 1/2) reading above 6 E-l ttCi/cc and expected to continue for 15 minutes or longer. (Dose assessment not available)
OR Valid Offgas Stack rad monitor (Kaman 9/10) reading above 4 E+2 pCi/cc and expected to continue for 15 minutes or longer. (Dose assessment not available)
OR
: 2. Field survey results indicate site boundary dose rates exceeding 1,000 mrem/hr expected to continue for more than one hour; or analyses of field survey samples indicate CDE Thyroid of 5,000 mrem for one hour of inhalation.
OR
: 3. Dose assessment determines integrated accident dose projection outside the site boundary above 1,000 mrem TEDE or above 5,000 mrem CDE Thyroid.
DAEC EAL INFORMATION:
Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results. In a case where data from Kaman readings is being used to determine whether an EAL threshold value has been exceeded, Valid means that flow through the associated Kaman Monitor has been verified and does exist as indicated in pCi/sec on SPRAD.
AGI

EBD-A EAL BASES DOCUMENT Rev. 8 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT CATEGORY                                    PAGE 25 of 27 The preferred method for declaration of AGI is by means of Dose Assessment using the MIDAS computer model. H-lowever, if Kaman monitor readings are sustained for longer than 15 minutes and the required MIDAS dose assessments cannot be completed within this period, then the declaration can be made using Kaman readings PROVIDEI) the readings are not from an isolated flow path. If Kaman readings are not valid. field survey results may be utilized.
DAEC's Meteorological Information and Dose Assessment System (MIDAS) wvas utilized to determine the Kaman monitor limits. Eight separate combinations of release point, source term, meteorological conditions and equipment status were analyzed. Iathways considered wvere the offgas stack, the turbine building exhaust vent and a single reactor building exhaust vent. Multiple release points were not considered. In this same vein, it was assumed that only one of the three reactor building vents is on during the release.
The source terms used have been pre-loaded into MIDAS and are the deflault mixes associated with a loss of coolant accident (LOCA) and a control rod drop (CRD). 'l'he LOCA mix was used in conjunction With a release via the offgas stack while the CR) mlix was used for releases via the turbine or reactor building vents. The source term lor a release via the offgas stack is further impacted by the status of the standby gas treatment system.
The status of that system wvas also taken into consideration.
Based of 1995 data (NG-96-0987), the atmospheric stability was classified as Pascal B 33%
of the time. Consequently, both classifications were evaluated. Based oln the same report, the most common wind speeds were:
Pascal Class                 Altitude                    Speed (mrph)
: 1)                            156'                       8- 12
: 1)                           33'                         8- 12 E                             156'                       8- 12 1                            33'                         4-7 Though the temperature setting has no impact on the MIDAS calculations, a value must be entered in order for the program to rin. Consequently, the temperature auras arbitrarily set at 50 F.
Tile rain estimate was set at zero, to eliminate any on site wvashout of radioactive material.
I-or the first MIDAS runs a ICi/ec concentration was assumed. The results of these runs were then normalized to the limits, thus generating a theoretical Kaman limit. Additional MIDAS runs were made with these theoretical limits as input to verify the normalization process.
AG1

EBD-A EAL BASES DOCUMENT Rev. 8 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT CATEGORY                                     PAGE 26 of 27 1-'
In addition to the total integrated dose, MIDAS calculates a peak whole body DDE rate resulting from the plume and a peak thyroid CDE rate resulting from inhalation. Because the ASI and AGI Kaman limits are to be based on a one-hour exposure, establishing concentration limits so these peak values match the NUMARC limits is acceptable.
Site Area        General Initiating Condition                                      Emergency        Emergency ASI              AGI Valid Turbine or RB ventilation rad monitor (Kaman) reading for more than 15 minutes above:           0.06 pCi/cc      0.6 pCi/cc Valid Offgas Stack ventilation rad monitor (Kaman) reading for more than 15 minutes above:                   40 pCi/cc        400 pCi/cc DAEC does not have a telemetered radiation monitoring system. As an alternative, DAEC uses valid field survey readings outside the site boundary to determine if doses are greater than 1,000 mr/hr TEDE or greater than 5,000 mr/hr CDE to the Thyroid.
Dose assessment using MIDAS is based on the EPA-400 methodology, e.g., use of Total Effective Dose Equivalent (TEDE) and Committed Dose Equivalent (CDE) Thyroid.
TEDE is somewhat different from whole body dose from gaseous effluents determined by ODAM methodology which forms the basis for the radiation monitor readings calculated in AU 1. These factors can introduce differences that are at least as large as those introduced by using TEDE versus whole body dose. The gaseous effluent radiation monitors can only detect noble gases. The contribution of iodine's to TEDE and CDE Thyroid could therefore only be determined either by: (1) utilizing the source term mixture in MlIDAS, or (2) gaseous effluent sampling. Therefore, DAEC EAL Threshold Value 4 is xvritten in terms of TEDE and CDE Thyroid.

==REFERENCES:==

I. Offsite Dose Assessment Manual, Section 6.1.2 and 7.1.2, Bases
: 2. Emergency Plan Implementing Procedure (E'PIP) 3.3, Dose Assessment and Protective Action
: 3. Radiation Protection Calculation No. 95-001-C, Emergency Actions Levels Based on Effluent Radiation Monitors, January 24, 1995
: 4. Radiation Engineering Calculation No. 96-007-A, Determination of DAEC Radioactive Release Initiating Conditions for AS I & AGI Emergency Classifications, July 3, 1996
: 5. UFSAR Section 11.5, Process and Effluent Radiation Monitoring and Sampling Systems AG1

EBD-A EAL BASES DOCUMENT Rev. 8 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT CATEGORY                                       PAGE 27 of 27
: 6. EPA 400-R-92-001, Alrnuatrl ofProtectiv ct iion Giciies (anc lIr olectiie Actionis.oI Ai'tclear Inciclenis
: 7. ATEI Ajfethodlology for Deilelopemenn of EmzerpygencjyAction Levels \UAIAR(/S'I/I\'S'vI-OO7 Revision 4, May 1999.
AG1

EBD-F EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                            PAGE 1 of 29 Usage Level Reference Use
[Effective  Date:   X                ]
TECHNICAL REVIEW Prepared by:       i         )     l                    Date:    IL?      /o Reviewed by:      H      a&0.
Independent Reflwer Date:    01i2q Reviewed by:                                             Date:    i (Zq /01 U J      Operations Staff PROCEDURE APPROVAL I am responsible for the technical content of this procedure and for obtaining the necessary approval from the State and County Emergency Management officials prior to implementation.
Documentation of State and County Emergency Management approval is via NEP-2004-0005 Approved by:    C2.e4D                                   Date:     I /-3/ov  '
Manager, Emergency Planning                    I      I

EBD-F EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                                   PAGE 2 of 29l Table of Contents FU1 - Any Loss or Any Potential Loss of Primary Containment Barrier ...................           3 FAI - Any Loss or Any Potential Loss of Either Fuel Clad Or RCS Barrier .............. 4 FS1 - Loss Or Potential Loss of Any Two Barriers ................................................ 5 FG1 - Loss of Any Two Barriers AND Potential Loss of the Third Barrier ............. 6 FUEL CLAD BARRIER INDICATORS ........................................................             7 RCS BARRIER INDICATORS .............            .......................................... 14 PRIMARY CONTAINMENT BARRIER INDICATORS ........................................ 22

EBD-F EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                            PAGE 3 of 29l FUI Any Loss or Any Potential Loss of Primary Containment Barrier EVENT TYPE: See Fission Barrier Table OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EAL Threshold Values:
See the Fission Barrier Table indicators discussed later in this section.
DAEC INFORMATION:
The entry conditions for this Initiating Condition are shown by the logic chart located to the right of the Fission Barrier Table. This logic is simplified from the generic NUMARC/NESP-007 logic based on the following considerations:
: 1. Human Factors - It is easier to understand and to remember the escalation from Alert to Site Area Emergency to General Emergency using the simpler logic.
: 2. Comprehensiveness - A comparison was made of the combinations of barrier losses and potential losses between the DAEC logic and the NUMARC/NESP-007 logic. All six generic barrier loss/potential loss combinations are addressed in the DAEC logic that addresses 12 combinations of barrier loss/potential loss. No sequences addressed by the NUMARC/NESP-007 logic are significantly affected by the simplified logic when applied to a BWR.

==REFERENCES:==

See the Fission Barrier Table indicators discussed later in this section.
FUI

EBD-F EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                             PAGE 4 of 29l FAI Any Loss or Any Potential Loss of Either Fuel Clad Or RCS Barrier EVENT TYPE: See Fission Barrier Table OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EAL Threshold Values:
See the Fission Barrier Table indicators discussed later in this section.
DAEC INFORMATION:
The entry conditions for this Initiating Condition are shown by the logic chart located to the right of the Fission Barrier Table. This logic is simplified from the generic logic based on the following considerations:
: 1. Human Factors - It is easier to understand and to remember the escalation from Alert to Site Area Emergency to General Emergency using the simpler logic.
: 2. Comprehensiveness - A comparison was made of the combinations of barrier losses and potential losses between the DAEC logic and the NUMARC/NESP-007 logic. All six generic barrier loss/potential loss combinations are addressed in the DAEC logic that addresses 12 combinations of barrier loss/potential loss. No sequences addressed by the NUMARC/NESP-007 logic are significantly affected by the simplified logic when applied to a BWR.

==REFERENCES:==

See the Fission Barrier Table indicators discussed later in this section.
FAt

EBD-F EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                             PAGE 5 of 29l FS1 Loss Or Potential Loss of Any Two Barriers EVENT TYPE: See Fission Barrier Table OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EAL Threshold Values:
See the Fission Barrier Table indicators discussed later in this section.
DAEC INFORMATION:
The entry conditions for this Initiating Condition are shown by the logic chart located to the right of the Fission Barrier Table. DAEC uses "Loss Or Potential Loss of Any Two Barriers." This logic is simplified from the generic logic based on the following considerations:
: 1. Human Factors - It is easier to understand and to remember the escalation from Alert to Site Area Emergency to General Emergency using the simpler logic.
: 2. Comprehensiveness - A comparison was made of the combinations of barrier losses and potential losses corresponding to Site Area Emergency between the DAEC logic and the NUMARC/NESP-007 logic. All six generic barrier loss/potential loss combinations are addressed in the DAEC logic that addresses 12 combinations of barrier loss/potential loss. No sequences addressed by the NUMARC/NESP-007 logic are significantly affected by the simplified logic when applied to a BWR.

==REFERENCES:==

See the Fission Barrier Table indicators discussed later in this section.
FS1

EBD-F EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                              PAGE 6 of 29l FG1 Loss of Any Two Barriers AND Potential Loss of the Third Barrier EVENT TYPE: See Fission Barrier Table OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EAL Threshold Values:
See the Fission Barrier Table indicators discussed later in this section.
DAEC INFORMATION:
The entry conditions for this Initiating Condition are shown by the logic chart located to the right of the Fission Barrier Table. This logic is simplified from the generic logic based on the following considerations:
: 1. Human Factors - It is easier to understand and to remember the escalation from Alert to Site Area Emergency to General Emergency using the simpler logic.
: 2. Comprehensiveness - A comparison was made of the combinations of barrier losses and potential losses between the DAEC logic and the NUMARC/NESP-007 logic. All six generic barrier loss/potential loss combinations are addressed in the DAEC logic that addresses 12 combinations of barrier loss/potential loss. No sequences addressed by the NUMARC/NESP-007 logic are significantly affected by the simplified logic when applied to a BWR.

==REFERENCES:==

See the Fission Barrier Table indicators discussed later in this section.
FGI

EBD-F EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                            PAGE 7 of 29l FISSION BARRIER: Fuel Clad DAEC INDICATOR: Radiation/Core Damage EAL THRESHOLD VALUE:
Clad Damage Determination LOSS - Fuel Damage assessment (PASAP 7.2) indicates at least 5% fuel clad damage.
POTENTIAL LOSS - None DAEC INFORMATION:
As a site-specific loss indicator, DAEC uses determination of at least 5% fuel clad damage, which is consistent with the containment rad monitor reading indicators described previously. This can be determined per FUEL DAMAGE ASSESMENT, PASAP 7.2.

==REFERENCES:==
: 1. Post Accident Sampling and Analysis Procedure (PASAP) 7.2, Fuel Damage Assessment Fuel Clad Barrier Radiation/Core Damage

EBD-F EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                              PAGE 8 of 29l FISSION BARRIER: Fuel Clad DAEC INDICATOR: Radiation/Core Damage EAL THRESHOLD VALUE:
Drywell/Torus Radiation Monitoring LOSS - Drywell Area Hi Range Rad Monitor RIM-9184A or B reading ABOVE 7E+2 R/hr OR LOSS - Torus Area Hi Range Rad Monitor RIM-9185A or B reading ABOVE 3E+1 R/hr POTENTIAL LOSS - None DAEC INFORMATION:
Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, coolant sampling or radiological survey results.
There is no significant deviation from the generic "loss" indicator. Per NUMARC/NESP-007, the (site-specific) reading (Drywell/Torus Rad - above) is a value that indicates release into the drywell of reactor coolant with elevated activity corresponding to about 2% to 5% fuel clad damage. This activity level is well above that expected from iodine spiking. It is intended that determination of barrier loss be made whenever the indicator threshold is reached until such time that core damage assessment is performed, at which time direct use of containment rad monitor readings is no longer required.
As documented by NG-88-0966, General Electric performed a study to predict dose rate readings from fuel damage calculations for emergency planning. The calculations were performed to obtain gamma ray dose rates at the locations of the containment atmospheric monitoring system radiation detectors in the drywell and torus locations for assumed releases of gap activity from the core. These calculations were based on "nominal" estimates of fuel rod gap fission product inventory fractions, which are considered to be more appropriate for determining a minimum threshold reading than inventory assumptions found in the NRC Regulatory Guides. The Regulatory Guide inventory assumptions applicable to dose assessments are larger and therefore non-conservative for determination of this EAL threshold. Two separate cases were evaluated.
In the first case, the released activity was assumed to be contained in the drywell atmosphere. This case is considered representative of conditions following a line break in Fuel Clad Barrier RadiationlCore Damage

EBD-F EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                              PAGE 9 of 29l which activity is released directly into the drywell. In the second case, the released activity was assumed to be contained in the torus. This could be applied for an event which results in vessel isolation and blowdown to the suppression chamber. The results for each case were provided for each case in the form of gamma ray dose rate versus time profiles for assumed releases of 100% and 20% of the gap activity from the core. The dose rate calculations were carried out independent of any specific information on details of construction or response characteristics of the detector systems. The figures show a drywell reading of about 2.9 x 10 3 R/hr or a torus reading of about 1.1 x 102 R/hr associated with 20% gap release at two hours after shutdown. Scaling this down to 5%
gap release:
Calculation of Drywell and Torus Monitor Readings Assuming 5% Gap Release NG-88-0966 value 20% Gap Release at 2 hours for drywell = 2.9 x 103 R/hr Drywell reading = 2.9 x 103 R/hr x [5 % / 20 %] = 7.25 x 102 R/hr, round off as 7 E+2 R/hr NG-88-0966 value 20% Gap Release at 2 hours for torus = 1.1 x 102 R/hr Torus reading = 1.1 x 102 R/hr x [5 % /20 %] = 2.75 x 101 R/hr, round off as 3 E+1 R/hr The results are rounded off for ease of reading the respective radiation monitors' scales.
The two-hour point was picked because it allows ample time for the Technical Support Center to be operational and core damage assessment to begin. These indicators correspond to about 2.5% gap release if they occur immediately after shutdown. Thus, the indicators address the 2%-5% fuel clad damage range of concern described by the generic guidance.

==REFERENCES:==
: 1. Office Memo NG-88-0966, G.E. Fuel Damage Documentation/Dose Rate Calculations, 03/18/88 Fuel Clad Barrier Radiation/Core Damage

EBD-F EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                            PAGE 10 of 29l FISSION BARRIER: Fuel Clad DAEC INDICATOR: Radiation/Core Damage EAL THRESHOLD VALUE:
Primary Coolant Activity Level LOSS - Coolant activity ABOVE 300 tLCi/gm dose equivalent 1-131.
POTENTIAL LOSS - None DAEC INFORMATION:
There is no significant deviation from the generic indicator. Consistent with the generic methodology, DAEC uses a coolant activity value of 300 jlCi/gm 1-131 equivalent. This value is well above that expected for iodine spikes and would indicate fuel clad damage has occurred.

==REFERENCES:==
: 1. Post Accident Sampling and Analysis Procedure (PASAP) 7.2, Fuel Damage Assessment Fuel Clad Barrier Radiation/Core Damage

EBD-F EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                             PAGE 11 of 29l FISSION BARRIER: Fuel Clad DAEC INDICATOR: RPV Level EAL THRESHOLD VALUE:
Reactor Vessel Water Level LOSS - RPV Level BELOW -25 Inches POTENTIAL LOSS - RPV Level BELOW 15 Inches DAEC INFORMATION:
The loss indicator is based on a value that corresponds to the minimum value to assure core cooling without further degradation of the fuel clad. DAEC uses the Minimum Steam Cooling RPV Water Level of -25 inches. This is defined to be the lowest RPV water level at which the covered portion of the reactor core will generate sufficient steam to preclude any clad temperature in the uncovered portion of the core from exceeding 1500 0F.
Consistent with the EOPs, an indicated RPV level below -25 inches is used.
The potential loss indicator corresponds to the water level at the top of the active fuel (TAF). Consistent with the EOPs, an indicated RPV level below 15 inches.

==REFERENCES:==
: 1. Emergency Operating Procedure (EOP)-1, RPV Control, Sheet 1 of 1
: 2. ATWS Emergency Operating Procedure (EOP)-RPV Control, Sheet 1 of 1
: 3. Emergency Operating Procedure (EOP) Basis, Curves and Limits, C5, Minimum Steam Cooling RPV Water Level Fuel Clad Barrier RPV Level

EBD-F EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                            PAGE 12 of 29l FISSION BARRIER: Fuel Clad DAEC INDICATOR: EC/OSM Judgement EAL THRESHOLD VALUE:
EC/OSM's Judgement Any condition which in the EC/OSM's judgement indicates loss or potential loss of the fuel clad barrier due to Imminent barrier degradation OR the barrier may be considered lost or potentially lost due to the inability to monitor the barrier.
DAEC INFORMATION:
There is no significant deviation from the generic indicator. Per EPIP 2.5, Control Room Emergency Response Operation, the Emergency Coordinator/Operations Shift Manager (EC/OSM) performs the emergency director function at DAEC.
EC/OSM considerations for determining whether any barrier "Loss" or "Potential Loss" include imminent barrier degradation, degraded barrier monitoring capability, and consideration of dominant accidentsequences.
Any condition which in the judgement of the EC/OSM indicates a LOSS or POTENTIAL LOSS of the FUEL CLAD barrier such as, but not limited to:
* Degraded barriermonitoringcapability from loss of/lack of reliable indicators.
* Consideration for instrumentation operability.
* Portable instrumentation readings.
* Offsite monitoring results.
* Complete loss of 125 VDC.
* Loss of decay heat removal.
* ATWS with failure of Standby Liquid Control.
* Prolonged station blackout.
* Loss of offsite power with early HPCI/RCIC failure Imminent means that no turnaround in safety system performance is expected and that General Emergency conditions can be expected to occur within two hours. Imminent fission barrier degradation must be considered by the EC/OSM to assure timely declaration of a General Emergency and to better assure that offsite protective actions can be effectively accomplished.
Fuel Clad Barrier EC/OSM Judgement

EBD-F EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                            PAGE 13 of 29l Degraded barriermonitoringcapability from loss of/lack of reliable indicators must also be considered by the EC/OSM when determining if a fission barrier loss or potential loss has occurred.
This assessment should also include consideration for instrumentation operability and portable instrumentation readings.
Offsite monitoring results may be an indication of Fission Product Barrier degradation causing an unmonitored release.
Dominant accident sequences can lead to loss of all Fission Barriers. Based on the IPE, the dominant accident sequences leading to core damage at DAEC include complete loss of 125 VDC, loss of decay heat removal, ATWS with failure of Standby Liquid Control, prolonged station blackout, and loss of offsite power with early HPCI/RCIC failure. The EC/OSM should also consult System Malfunction EALs, as appropriate, to assure timely emergency classification declaration.

==REFERENCES:==
: 1. Emergency Plan Implementing Procedure (EPIP) 2.5, Control Room Emergency Response Operation
: 2. Duane Arnold Energy Center Individual Plant Examination (IPE) November 1992 Fuel Clad Barrier ECIOSM Judgement

EBD-F EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                              PAGE 14 of 29l FISSION BARRIER: RCS DAEC INDICATOR: Radiation/Core Damage EAL THRESHOLD VALUE:
Drywell Radiation Monitoring LOSS - Drywell Area Hi Range Rad Monitor RIM-9184A or B reading ABOVE 5 R/hr after Reactor Shutdown POTENTIAL LOSS - None DAEC INFORMATION:
There is no significant deviation from the generic indicator. This loss indicator is based on conditions after reactor shutdown to assure that it is not misapplied, i.e., to exclude readings due to N-16 effects which are typically 5 to 8 R/hr at full power conditions.
The 5 R/hr value for this loss indicator corresponds to instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with normal operating concentrations (i.e., within Technical Specifications) into the drywell atmosphere. The reading will be less than that specified for the loss indicator for Radiation/Core Damage that applies to the Fuel Clad barrier. Thus, this indicator would be indicative of a RCS leak only. If the radiation monitor reading increased to that value specified by the Radiation/Core indicator applying to the Fuel Clad barrier, fuel damage would also be indicated.
As documented by NG-88-0966, General Electric performed a study to predict dose rate readings from fuel damage calculations for emergency planning. The calculations were performed to obtain gamma ray dose rates at the locations of the containment atmosphere monitoring system radiation detectors in the drywell and torus locations for assumed releases of gap activity from the core. These calculations were based on "nominal" estimates of fuel rod gap fission product inventory fractions, which are considered to be more appropriate for determining a minimum threshold reading than inventory assumptions found in the NRC Regulatory Guides. The Regulatory Guide inventory assumptions applicable to dose assessments are larger and therefore non-conservative for determination of this EAL threshold. Two separate cases were evaluated. In the first case, the released activity was assumed to be contained in the drywell atmosphere. This case is considered representative of conditions following a line break in which activity is released directly into the drywell. In the second case, the released activity was assumed to be contained in the torus. This could be applied for an event which results in vessel isolation and blowdown to the suppression chamber. The results for each case were RCS Barrier RadiationlCore Damage

EBD-F EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                               PAGE 15 of 29l provided for each case in the form of gamma ray dose rate versus time profiles for assumed releases of 100% and 20% of the gap activity from the core. The dose rate calculations were carried out independent of any specific information on details of construction or response characteristics of the detector systems. The figures show a drywell reading of about 2.1 x 10 4 R/hr associated with a 100% gap release immediately after shutdown. Assuming 99.99% fuel clad integrity (0.01% gap release) and uniform dispersal of radionuclides into the drywell immediately after shutdown, a drywell monitor reading is calculated:
Calculation of Drywell Monitor Reading Assuming 0.01% Gap Release NG-88-0966 value for 100% Gap Release at 0.01 minutes = 2.1 x         10 4 R/hr (2.1 x 10 4 ) R/hrx [(1 x 10-2) percent / 100 percent] = (2.1) x 1044 R/hr = 2.1 x 10&deg; R/hr = 2 R/hr To assure an indicator that is readily discernible on the drywell radiation monitor scale, DAEC uses a valid reading above 5 R/hr after reactor shutdown.

==REFERENCES:==
: 1. Office Memo NG-88-0966, G.E. Fuel Damage Documentation/Dose Rate Calculations, 03/18/88
: 2. Technical Specification 3.4.5, Drywell Leak Detection Instrumentation RCS Barrier Radiation/Core Damage

EBD-F EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                              PAGE 16 of 29l FISSION BARRIER: RCS DAEC INDICATOR: RPV Level EAL THRESHOLD VALUE:
Reactor Vessel Water Level LOSS - RPV Level BELOW 15 Inches POTENTIAL LOSS - None DAEC INFORMATION:
There is no significant deviation from the generic indicator. This loss indicator corresponds to the water level at the top of the active fuel (TAF). In order to provide normal means to cool the fuel, water level must be maintained above the top of active fuel otherwise extraordinary means must be taken to assure that adequate core cooling exists. In certain failure event sequences reactor vessel water level may be procedurally lowered to the top of active fuel and the reactor coolant system depressurized to allow for steam cooling of the core. Even though fuel clad damage is not predicted under these conditions several safety system failures need to have occurred to reach the condition where steam cooling would be procedurally required. Therefore this is indicative of a loss of the reactor coolant system boundary. Water levels below this value indicate a challenge to core cooling which is a precursor to more serious events.

==REFERENCES:==
: 1. Emergency Operating Procedures (EOP) Basis, Breakpoints RCS Barrier RPV Level

EBD-F EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                            PAGE 17 of 29l FISSION BARRIER: RCS DAEC INDICATOR: Leakage EAL THRESHOLD VALUE:
RCS Leak Rate LOSS - None POTENTIAL LOSS - RCS leakage ABOVE 50 GPM OR POTENTIAL LOSS - Unisolable primary system leakage outside drywell as indicated by area temperatures or ARMs exceeding the Max Normal Limits per EOP 3, Table 6.
DAEC INFORMATION:
There are no significant deviations from the generic potential loss indicators applying to RCS leakage and indications of unisolable primary system leakage.
If an SRV is stuck open or is cycling and no other emergency conditions exist, an emergency declaration may not be appropriate. RCS leakage inside the drywell excludes Safety-Relief Valve (SRV) discharge through the SRV discharge piping into the torus below the water line. However, if the fuel is damaged and the SRV is allowing fission products to escape into primary containment, a loss of RCS should be determined as having occurred. The EC/OSM should also consult SU5, RCS Leakage, to determine if RCS leakage exceeds the threshold required for declaration of an Unusual Event.
Unisolable primary system leakage is considered a Potential loss of RCS based on RCS leakage outside the drywell. Site-specific RCS leakage is determined from temperature or area radiation alarms (ARMs) exceeding the Max Normal limits listed in Table 6, EOP 3. Unisolable primary system leakage in the areas of the steam tunnel, main turbine generator, RCIC, HPCI, etc., indicates a direct path from the RCS to areas outside primary containment. It should be confirmed that the indicators are caused by RCS leakage. Area temperatures or area radiation alarms above Max Normal limits are the criteria for declaration of an Alert classification. An unisolable leak which is indicated by exceeding Max Safe limits escalates to a Site Area Emergency when combined with Primary Containment Barrier loss (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded.
DAEC does not use the generic "loss" indicator for main steam line break. NUMARC Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 2 Questions and Answers, June 1993, discloses that the main steam line break RCS Barrier Leakage

EBD-F EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                              PAGE 18 of 29l with isolation does not have to be included as a fission barrier table indicator. This event can be appropriately classified in the System Malfunction Recognition Category. This event was classified as a RCS barrier loss indicator in the generic guidance because this event typically results in a puff release with dose consequences greater than 10 millirem whole body, i.e., offsite dose consequences consistent with declaration of an Alert in accordance with AA1, Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 200 Times Radiological Technical Specifications for 15 Minutes or Longer. However, UFSAR Section 15.6.6, Table 15.6-1, Steam-Line Break -
Radiological Effects for Puff Release at 47 Meters, Total Dose, shows a maximum dose of 0.58 mrem (5.8E-04 rem) passing cloud whole body dose using conservative assumptions. Therefore, because this event at DAEC has dose consequences similar to those of AU1, Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 2 Times Radiological Technical Specifications for 60 Minutes or Longer, it has been added as an Unusual Event EAL in SU5, RCS Leakage.

==REFERENCES:==
: 1. Alarm Response Procedure (ARP) 1C04B, Reactor Water Cleanup and Recirculation
: 2. Alarm Response Procedure (ARP) 1C04C, Reactor Water Cleanup and Recirculation
: 3. Emergency Operating Procedure (EOP) 3, Secondary Containment Control
: 4. UFSAR Section 15.6.6, Loss-of-Coolant-Accident
: 5. NEI Methodology for Development of Emergency Action Levels Revision 4, May 1999
: 6. NUMARC Methodology for Development of Emergency Action Levels, NUMARC/NESP-007, Revision 2, Questions and Answers, June 1993 RCS Barrier Leakage

EBD-F EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                              PAGE 19 of 29l FISSION BARRIER: RCS DAEC INDICATOR: Primary Containment Atmosphere EAL THRESHOLD VALUE:
Drywell Pressure LOSS - Drywell Pressure ABOVE 2 psig and not caused by a loss of DW Cooling POTENTIAL LOSS - None DAEC INFORMATION:
There is no significant deviation from the generic indicator. The value for this loss indicator corresponds to the drywell high pressure ECCS initiation signal setpoint of 2.0 psig.
DAEC also specifies that drywell cooling is operating to assure that the indicator is not misapplied to conditions that do not indicate RCS leakage into the drywell, i.e., the drywell pressure increase is not due to loss of drywell cooling.
DAEC uses a GE Mark I Containment. During reactor operation, with drywell cooling in operation and the drywell inerted, the normal operating pressure in the drywell is between 0.5 and 1.0 psig. Analysis at the DAEC shows that a 50 gpm RCS leak would result in a 2 to 3 psig pressure rise over a six minute time period. Since a 2 psig rise would place DAEC above the ECCS initiation setpoint, (2 psig) it is necessary to select the DAEC ECCS initiation setpoint of 2 psig to indicate an actual loss of the RCS. Drywell cooling is not isolated at the 2 psig ECCS initiation setpoint, therefore further pressure rise would be indicative of a RCS leak.

==REFERENCES:==
: 1. Emergency Operating Procedures (EOP) Bases, Breakpoints
: 2. Emergency Operating Procedures (EOP) -1, RPV Control
: 3. Emergency Operating Procedures (EOP) -2, Primary Containment Control RCS Barrier Pri. Cont. Atmosphere

EBD-F EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                              PAGE 20 of 29l FISSION BARRIER: RCS DAEC INDICATOR: EC/OSM's Judgement EAL THRESHOLD VALUE:
* Any condition which in the EC/OSM's judgement indicates loss or potential loss of the RCS barrier due to Imminent barrier degradation OR the barrier may be considered lost or potentially lost due to the inability to monitor the barrier.
DAEC INFORMATION:
There is no significant deviation from the generic EAL. Per EPIP 2.5, Control Room Emergency Response Operation, the Emergency Coordinator/Operations Shift Manager (EC/OSM) performs the emergency director function at DAEC. EC/OSM considerations for determining whether any barrier "Loss" or "Potential Loss" include imminent barrier degradation, degraded barriermonitoring capability, and consideration of dominant accident sequences.
Any condition which in the judgement of the EC/OSM indicates a LOSS or POTENTIAL LOSS of the RCS barrier such as, but not limited to:
* Degraded barrier monitoring capability from loss of/lack of reliable indicators.
* Consideration for instrumentation operability.
* Portable instrumentation readings.
* Offsite monitoring results.
* Complete loss of 125 VDC.
* Loss of decay heat removal.
* ATWS with failure of Standby Liquid Control.
* Prolonged station blackout.
* Loss of offsite power with early HPCI/RCIC failure Imminent means that no turnaround in safety system performance is expected and that General Emergency conditions can be expected to occur within two hours. Imminent fission barrier degradation must be considered by the EC/OSM to assure timely declaration of a General Emergency and to better assure that offsite protective actions can be effectively accomplished.
RCS Barrier EC/OSM Judgement

EBD-F EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                             PAGE 21 of 291 Degraded barier monitoring capability from loss of/lack of reliable indicators must also be considered by the EC/OSM when determining if a fission barrier loss or potential loss has occurred.
This assessment should also include consideration for instrumentation operability and portable instrumentation readings.
Offsite monitoring results may be an indication of Fission Product Barrier degradation causing an unmonitored release.
Dominant accident sequences can lead to loss of all Fission Barriers. Based on the IPE, the dominant accident sequences leading to core damage at DAEC include complete loss of 125 VDC, loss of decay heat removal, ATWS with failure of Standby Liquid Control, prolonged station blackout, and loss of offsite power with early HPCI/RCIC failure. The EC/OSM should also consult System Malfunction EALs, as appropriate, to assure timely emergency classification For the RCS barrier, the EC/OSM should also consider safety-relief valves (SRVs) open or cycling. If an SRV is stuck open or is cycling and no other emergency conditions exist, an emergency declaration may not be appropriate. However, if the fuel is damaged and the SRV is allowing fission products to escape into primary containment, a loss of RCS should be determined as having occurred. The EC/OSM should also consult SU5, RCS Leakage, to determine if RCS leakage exceeds the threshold required for declaration of an Unusual Event.

==REFERENCES:==
: 1. Emergency Plan Implementing Procedure (EPIP) 2.5, Control Room Emergency Response Operation
: 2. Duane Arnold Energy Center Individual Plant Examination (IPE) November 1992
: 3. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, May 1999 RCS Barrier EC/OSM Judgement

EBD-F EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                             PAGE 22 of 29l FISSION BARRIER: Primary Containment DAEC INDICATOR: Radiation/Core Damage EAL THRESHOLD VALUE:
Significant Radioactive Inventory in Containment LOSS - None POTENTIAL LOSS - Drywell Area Hi Range Rad Monitor RIM-9184A or B reading ABOVE 3E+3 R/hr OR POTENTIAL LOSS - Torus Area Hi Range Rad Monitor RIM-9185A or B reading ABOVE 1E+2 R/hr DAEC INFORMATION:
There is no significant deviation from the generic indicators. The potential loss (site-specific) indicator value corresponds to at least 20% fuel clad damage with release into the primary containment. This indicator corresponds to loss of both the Fuel Clad and RCS barriers with Potential Loss of the Primary Containment barrier, and would result in declaration of a General Emergency. The basis for the 20% fuel clad damage threshold is described under the 20% core damage assessment indicator. It is intended that determination of banierpotential loss be made whenever the indicator threshold is reached until such time that core damage assessment is performed, at which time direct use of containment rad monitor readings is no longer required.
As documented by NG-88-0966, General Electric performed a study to predict dose rate readings from fuel damage calculations for emergency planning. The calculations were performed to obtain gamma ray dose rates at the locations of the containment atmospheric monitoring system radiation detectors in the drywell and torus locations for assumed releases of gap activity from the core. These calculations were based on "nominal" estimates of fuel rod gap fission product inventory fractions, which are considered to be more appropriate for determining a minimum threshold reading than inventory assumptions found in the NRC Regulatory Guides. The Regulatory Guide inventory assumptions applicable to dose assessments are larger and therefore non-conservative for determination of this EAL threshold. Two separate cases were evaluated.
In the first case, the released activity was assumed to be contained in the drywell atmosphere. This case is considered representative of conditions following a line break in which activity is released directly into the drywell. In the second case, the released activity was assumed to be contained in the torus. This could be applied for an event which results in vessel isolation and blowdown to the suppression chamber. The results for each case Primary Containment Barrier Radiation/Core Damage

EBD-F EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                            PAGE 23 of 29l were provided for each case in the form of gamma ray dose rate versus time profiles for assumed releases of 100% and 20% of the gap activity from the core. The dose rate calculations were carried out independent of any specific information on details of construction or response characteristics of the detector systems. The figures show a drywell reading of about 2.9 x 103 R/hr and a torus reading of about 1.1 x 102 R/hr associated with 20% gap release at two hours after shutdown. These values are rounded to 3 E+3 R/hr and 1 E+2 R/hr, respectively. The two hour point was picked because it allows ample time for the Technical Support Center to be operational and core damage assessment to begin.

==REFERENCES:==
: 1. Office Memo NG-88-0966, G.E. Fuel Damage Documentation/Dose Rate Calculations, 03/18/88 Primary Containment Barrier Radiation/Core Damage

EBD-F EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                             PAGE 24 of 29l FISSION BARRIER: Primary Containment DAEC INDICATOR: Radiation/Core Damage EAL THRESHOLD VALUE:
Clad Damage Determination LOSS - None POTENTIAL LOSS - Fuel Damage assessment procedures indicate at least 20% fuel clad damage.
DAEC INFORMATION:
As a site-specific "potential loss" indicator, DAEC uses determination of at least 20% fuel clad damage, which is consistent with the level of fuel damage indicated by the drywell and torus radiation monitor readings used earlier with this Indicator. This can be determined using appropriate fuel damage assessment procedures. Regardless of whether primary containment integrity is challenged, it is possible for significant radioactivity within the primary containment to result in EPA PAG plume exposure levels being exceeded even assuming that the primary containment is within technical specification allowable leakage rates. With or without primary containment challenge, however, a major release of radioactivity requiring off-site protective actions from core damage is not possible unless a major failure of the fuel clad barrier allows radioactive material to be released from core into the reactor coolant. NUREG-1228 indicates that such conditions do not exist when the amount of fuel clad damage is less than 20%.
Other indicators were also considered. No other reliable indicators for Primary Containment "loss" or "potential loss" could be determined.

==REFERENCES:==
: 1. Post Accident Sampling and Analysis Procedure (PASAP) 7.2, Fuel Damage Assessment
: 2. NUREG-1228, Source Term Estimations During Incident Response to Severe Nuclear Power Plant Accidents, October 1988 Primary Containment Barrier Radiation/Core Damage

EBD-F EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                             PAGE 25 of 29l FISSION BARRIER: Primary Containment DAEC INDICATOR: RPV Level EAL THRESHOLD VALUE:
Reactor Vessel Water Level LOSS - None POTENTIAL LOSS - RPV level BELOW -39 inches DAEC INFORMATION:
The underlying concern for this indicator is a threshold that represents significant uncovering of the core and imminent core damage. Imminent means that no turnaround in safety system performance would be expected and that General Emergency conditions would be expected within two hours.
Consistent with the underlying concern, the DAEC indicator addresses conditions where the water level is below the Minimum Zero-Injection RPV Water Level of -39. The Minimum Zero-injection RPV Water Level is defined to be the lowest RPV water level at which the covered portion of the reactor core will generate sufficient steam to preclude any fuel clad temperature in the uncovered portion of the core from exceeding 1800 'F. The Minimum Zero-Injection RPV Water Level is utilized to preclude significant fuel clad damage and hydrogen generation for as long as possible when no sources of RPV makeup water are available.
Thus, for RPV water level below -39 inches, water levels would continue to decrease and the fuel clad temperature would be expected to continue to rise. Due to large uncertainties in severe accident progression, it should be assumed that severe core melt is imminent if this condition were to occur. It would not be acceptable to delay the declaration of the General Emergency and issuance of Protective Action Recommendations beyond this point.

==REFERENCES:==
: 1. Emergency Operating Procedure (EOP) RPV/F - RPV Flooding
: 2. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, May 1999 FISSION BARRIER: Primary Containment Primary Containment Barrier Leakage

EBD-F EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                               PAGE 26 of 29.
DAEC INDICATOR: Leakage EAL THRESHOLD VALUE:
Containment Isolation Valve Status After Containment Isolation Signal LOSS - Failure of both valves in any one line to close AND a downstream pathway to the environment exists.
OR LOSS - Unisolable primary system leakage outside the drywell as indicated by area temps or ARMs exceeding the Max Safe Limits per EOP 3, Table 6, when Containment Isolation is required.
OR LOSS - Primary containment venting in progress per EOPs.
POTENTIAL LOSS - None DAEC INFORMATION:
The "loss" indicators used at DAEC directly correspond to the generic indicators. Venting of the primary containment can be performed in accordance with EOP 2 irrespective of the offsite radioactivity release rate that will occur and by defeating isolation interlocks as necessary. The consequences of not doing so may be the loss of primary containment integrity, core damage, and an uncontrolled radioactive release much greater than might otherwise occur. Primary containment venting is performed only as necessary to reduce and then maintain torus pressure below the Primary Containment Pressure Limit (PCPL) of 53 psig.
This EAL is intended to cover the inability to isolate the containment when containment isolation is required. In addition, the presence of area radiation or temperature alarms above the Max Safe limits listed in Table 6, EOP 3 after a containment isolation, indicate an unisolable primary system leakage outside the drywell. The indicators should be confirmed to be caused by RCS leakage. Also, an intentional venting of primary containment for pressure control per EOPs to the secondary containment and/or the environment is considered a loss of containment. Containment venting for temperature or pressure when not in an accident situation should not be considered.

==REFERENCES:==
: 1. Emergency Operating Procedure (EOP) 2, Primary Containment Control
: 2. Emergency Operating Procedure (EOP) 3, Secondary Containment Control
: 3. Emergency Operating Procedures (EOP) Bases, Breakpoints Primary Containment Barrier Leakage

EBD-F EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                                PAGE 27 of 29 FISSION BARRIER: Primary Containment DAEC INDICATOR: Primary Containment Atmosphere EAL THRESHOLD VALUE:
Drywell Pressure/Atmosphere LOSS - Rapid unexplained decrease following initial increase in pressure.
OR LOSS - Drywell pressure response not consistent with LOCA conditions.
OR POTENTIAL LOSS - Torus Pressure reaches 53 PSIG.
OR POTENTIAL LOSS - Drywell or Torus H2 CANNOT be determined to be below 6%
AND Drywell or torus 02 CANNOT be determined to be below 5%.
DAEC INFORMATION:
There are no significant deviations from the generic indicators. The "loss" indicators used at DAEC directly correspond to the generic indicators.
The first "potential loss" indicator is torus pressure of 53 psig, which is the Primary Containment Pressure Limit (PCPL) used in the EOPs. The second "potential loss" indicator is based on determination of explosive mixture in accordance with the SAGs.
DAEC SAGs require control of drywell and torus atmosphere gas concentrations to less than 6% H2 and less than 5% 02 to assure that an explosive mixture does not exist. This "potential loss" indicator is written to be consistent with the SAGs.

==REFERENCES:==
: 1. Emergency Operating Procedure (EOP) 2, Primary Containment Control
: 2. Severe Accident Guideline - 3 (SAG-3), Hydrogen Control Primary Containment Barrier Primary Containment Atmosphere

EBD-F          I EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                               PAGE 28 of 29l FISSION BARRIER: Primary Containment DAEC INDICATOR: ECIOSM Judgement EAL THRESHOLD VALUE:
Any condition which in the EC/OSM's judgement indicates loss or potential loss of the primary containment barrier due to Imminent barrier degradation OR the barrier may be considered lost or potentially lost due to the inability to monitor the barrier.
DAEC INFORMATION:
There is no significant deviation from the generic indicator. Per EPIP 2.5, Control Room Emergency Response Operation, the Emergency Coordinator/Operations Shift Manager (EC/OSM) performs the emergency director function at DAEC. EC/OSM considerations for determining whether any barrier "Loss" or "Potential Loss" include imminent barrier degradation, degraded barrier monitoring capability, and consideration of dominant accident sequences.
Any condition which in the judgement of the EC/OSM that indicates LOSS or POTENTIAL LOSS of the Primary Containment Barrier such as, but not limited to:
* Degraded barrier monitoring capability from loss of/lack of reliable indicators.
* Consideration for instrumentation operability.
* Portable instrumentation readings.
* Offsite monitoring results.
* Complete loss of 125 VDC.
* Loss of decay heat removal.
* ATWS with failure of Standby Liquid Control.
* Prolonged station blackout.
* Loss of offsite power with early HPCI/RCIC failure Imminent means that no turnaround in safety system performance is expected and that General Emergency conditions can be expected to occur within two hours. Imminent fission barrier degradation must be considered by the EC/OSM to assure timely declaration of a General Emergency and to better assure that offsite protective actions can be effectively accomplished.
Primary Containment Barrier EC/OSM Judgement

EBD-F EAL BASES DOCUMENT Rev. 5 FISSION PRODUCT BARRIER DEGRADATION CATEGORY                             PAGE 29 of 29 Degraded barriermonitoring capability from loss of/lack of reliable indicators must also be considered by the EC/OSM when determining if a fission barrier loss or potential loss has occurred.
This assessment should also include consideration for instrumentation operability and portable instrumentation readings.
Offsite monitoring results may be an indication of Fission Product Barrier degradation causing an unmonitored release.
Dominant accident sequences can lead to loss of all Fission Barriers. Based on the IPE, the dominant accident sequences leading to core damage at DAEC include complete loss of 125 VDC, loss of decay heat removal, ATWS with failure of Standby Liquid Control, prolonged station blackout, and loss of offsite power with early HPCI/RCIC failure. The EC/OSM should also consult System Malfunction EALs, as appropriate, to assure timely emergency classification

==REFERENCES:==
: 1. Emergency Plan Implementing Procedure (EPIP) 2.5, Control Room Emergency Response Operation
: 2. Duane Arnold Energy Center Individual Plant Examination (IPE) November 1992 Primary Containment Barrier EC/OSM Judgement

EAL BASES DOCUMENT                                    EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                       Rev. 8 CATEGORY                                      PAGE 1 of 31 I                                                                               ____________________________
Usage Level I.        REFERENCE USE Effective Date:    :2  -, -,
TECHNICAL REVIEW Prepared and Verified by:
                                              '-7                Date:     k 4-404 Reviewed by:                                             Date:   01/z9 /O4 Emergency Plan in Staff Reviewed by:                                             Date:    I /), IC)l I                          K I Operations Reviewer C)
PROCEDURE APPROVAL I am responsible for the technical content of this procedure and for obtaining the necessary approval from the State and County Emergency Management officials prior to implementation.
Documentation of State and County Emergency Management approval is via NEP-2004-0005.
Approved by:      A      x   <>add<>                      Date:     C/3c,/o c Manager, Emergency Planning

EAL BASES DOCUMENT                                                            EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                                                     Rev. 8 CATEGORY                                                           PAGE 2 of 31 Table of Contents HUI - Natural and Destructive Phenomena Affecting the Protected Area ...................H-3 HU2 - Fire Within Protected Area Not Extinguished Within 15 Minutes of Detection ...H-7 HU3 - Release of Toxic or Flammable Gases Deemed Detrimental to Safe Operation of the Plant ..................................................                                                           H-8 HU4 - Confirmed Security Event Which Indicates a Potential Degradation in the Level of Safety of the Plant ..................................................                                             H-9 HU5 - Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of an Unusual Event ........................................................                          H-i 1 HAI - Natural and Destructive Phenomena Affecting the Plant Vital Area ................. H-12 HA2 - Fire Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown .................................................. H-16 HA3 - Release of Toxic or Flammable Gases Within a Facility Structure Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown ........................................................                                   H-18 HA4 - Confirmed Security Event in a Plant Protected Area ....................................... H-20 HA5 - Control Room Evacuation Has Been Initiated .................................................. H-22 HA6 - Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of an Alert ....................................................... H-23 HSI - Confirmed Security Event in a Plant Vital Area .................................................                   H-24 HS2 - Control Room Evacuation Has Been Initiated and Plant Control Cannot Be Established
        .                                                             .................................................. H-26 HS3 - Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of a Site Area Emergency .                         .................................................. H-28 HG1 - Security Event Resulting in Loss Of Ability to Reach and Maintain Cold Shutdown.H-29 HG2 - Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of a General Emergency ..................................................H-30

EAL BASES DOCUMENT                                     EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                         Rev. 8 CATEGORY                                       PAGE 3 of 31 HU1 Natural and Destructive Phenomena Affecting the Protected Area EVENT TYPE: Natural Disasters and Destructive Phenomena OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:
: 1. Earthquake detected per AOP 901, Earthquake OR
: 2. Report of tornado touching down within Plant Protected Area or within switchyard.
OR
: 3. Assessment by the control room that a destructive event has occurred.
OR
: 4. Vehicle crash into plant structures or systems within Plant Protected Area.
OR
: 5. Report of an unanticipated explosion within the Plant Protected Area resulting in visible damage to permanent structures or equipment.
OR
: 6. Report of turbine failure resulting in casing penetration or damage to turbine or generator seals.
OR
: 7. River level above 757 feet.
OR
: 8. Any water level above Max Normal Operating Limit.
OR
: 9. River water level below 725 feet. 6 inches.
HUI

EAL BASES DOCUMENT                                     EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                         Rev. 8 CATEGORY                                      PAGE 4 of 31 DAEC EAL INFORMATION:
The Plant Protected Area is the area within the security fence. This includes ISFSI and the Intake Structure. Although the switchyard is included in this EAL, it is not part of the Plant Protected Area.
DAEC EAL Threshold Value 1 addresses earthquakes that are detected in accordance with AOP 901. For DAEC, a minimum detectable earthquake that is indicated on panel 1C35 is an acceleration greater than +/- 0.01 Gravity.
DAEC EAL Threshold Value 2 addresses report of a tornado striking within the Plant Protected Area or within the plant switchyard.
DAEC EAL Threshold Value 3 allows the control room to determine that an event has occurred and take appropriate action based on personal assessment as opposed to verification. No attempt is made to assess the actual magnitude of the damage. Such damage can be due to collision, tornadoes, missiles, or any other cause. Damage can be indicated by report to the control room, physical observation, or by Control Room/local control station instrumentation. Such items as scorching, cracks, dents, or discoloration of equipment or structures required for safe shutdown are addressed by this EAL.
DAEC EAL Threshold Value 4 addresses a vehicle (automobile, aircraft, forklift, truck or train) crash that may potentially damage plant structures containing functions and systems required for safe shutdown of the plant. This does not include vehicle crashes with each other or damage to office or warehouse structures. Escalation to Alert under HA1 would occur if damage was sufficient to affect the ability to achieve or maintain safe shutdown, e.g., damage made required equipment inoperable or structural damage was observed such as bent supports or pressure boundary leakage.
Safe ShutdownNital Areas Category                                             Area Electrical      Switchyard, 1G31 DG and Day Tank Rooms, 1G21 DG and Day Tank Power         Rooms, Battery Rooms, Essential Switchgear Rooms, Cable Spreading Power      Room Heat Sink/
Coolant        Torus Room, Intake Structure, Pumphouse Supply Containment       Drywell, Torus Emergency        NE, NW, SE Corner Rooms, HPCI Room, RCIC Room, RHR Valve Room, Systems        North CRD Area, South CRD Area, CSTs Other        Control Building, Remote Shutdown Panel 1C388 Area, Panel 1C55/56 Area, SBGT Room HU1

EAL BASES DOCUMENT                                    EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                        Rev. 8 CATEGORY                                      PAGE 5 of 31 DAEC EAL Threshold Value 5 addresses explosions within the Plant Protected Area. As used here, an explosion is a rapid, violent, unconfined combustion, or a catastrophic failure of pressurized equipment, that significantly imparts significant energy to near-by structures or equipment. Damage can be indicated by report to the control room, physical observation, or by Control Room/local control station instrumentation. Such items as scorching, cracks, dents, or discoloration of equipment or structures are addressed by this EAL. The Emergency Director needs to consider the security aspects of the explosion, if applicable.
DAEC EAL Threshold Value 6 addresses turbine failure causing observable damage to the turbine casing or damage to turbine or generator seals.
DAEC EAL Threshold Value 7 addresses the observed effects of flooding in accordance with AOP 902. Plant site finished grade is at elevation 757.0 ft. Personnel doors and railroad and truck openings at or near grade would require protection in the event of a flood above elevation 757.0 ft.
Therefore, EAL 6 uses a threshold of flood water levels above 757.0 ft.
 )AEC EAL Threshold Value 8 addresses internal flooding can be due to system malfunctions, component failures, or repair activity mishaps (such as failed freeze seal) that can threaten safe operation of the plant. Therefore, this EAL is based on a valid indication that the water level is higher than the maximum normal operating limits. The Maximum Normal Operating Limits are defined as the highest values of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly.
Exceeding these limits is an entry condition into EOP 3, Secondary Containment Control and may be an indication that water from a primary system is discharging into secondary containment.
Exceeding the maximum normal operating limit is interpreted as a potential degradation in the level of the safety of the plant and is appropriately treated as an Unusual Event emergency classification. The maximum normal operating water level limits are taken from AOP 902 and EOP 3 and are shown in the table below:
Maximum Operatinci Limits - Water Levels                                             l Affected Location                    Indicator       Maximum         Maximum Safe Normal OL       OL HPCI Room Area                      Li 3768          2 inches       6 inches RCIC Room Area                      LI 3769          3 inches       6 inches A RHR Corner Room SE Area            LI 3770          2 inches       10 inches B RHR Corner Room NW Area           Li 3771          2 inches       10 inches Torus Area                          LI 3772         2 inches        12 inches HU1

EAL BASES DOCUMENT                                    EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                        Rev. 8 CATEGORY                                       PAGE 6 of 31 EAL Threshold Value 9 addresses the effects of low river water level. The intake structure for the safety-related water supply systems (river water, RHR service water, and emergency service water) is located on the west bank of the Cedar River. An overflow-type barrier across the river was designed and constructed in accordance with Seismic Category I criteria to intercept the stream bed flow and divert it to the intake structure. This makes the entire flow of the river available to the safety-related water supply systems. A minimum flow of 13 cubic feet per second (cfs) from a minimum 1000-year river flow of 60 cfs must be diverted. The top of the barrier wall is at elevation 725 ft. 6 in. River water level below this level represents a potential degradation in the level of safety of the plant and is addressed by EAL Threshold Value 9.
In this EAL, "Vital Area" is defined as plant structures or areas containing equipment necessary for a safe shutdown, i.e., synonymous with Safe Shutdown Area.

==REFERENCES:==
: 1. Abnormal Operating Procedure (AOP) 901, Earthquake
: 2. Abnormal Operating Procedure (AOP) 902, Flood
: 3. Abnormal Operating Procedure (AOP) 903, Tornado
: 4. Emergency Operating Procedure (EOP)-3, Secondary Containment Control
: 5. EOP Basis Document, EOP-3, Secondary Containment Control
: 6. UFSAR Chapter 3, Design of Structures, Components, Equipment, and Systems
: 7. Bechtel Drawing BECH-M017, Equipment Location - Intake Structure Plans at Elevations, Rev. 6 HU1

EAL BASES DOCUMENT                                      EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                         Rev. 8 CATEGORY                                       PAGE 7 of 31 HU2 Fire Within Protected Area Not Extinguished Within 15 Minutes of Detection EVENT TYPE: Fire OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:
: 1. Fire in buildings or areas contiguous to any of the following areas not extinguished within 15 minutes of control room notification or verification of a control room alarm:
* Reactor, turbine, control, admin/security
* Intake structure
* Pump house DAEC EAL INFORMATION:
The purpose of this EAL is to address the magnitude and extent of fires that may be potentially significant precursors to damage to safety systems. This includes such items as fires within the administration building, and security building (buildings contiguous to the reactor building, turbine building and control building), yet, excludes fires in the warehouse or construction support center, waste-basket fires, and other small fires of no safety consequence. As used here, Detection is visual observation and report by plant personnel or sensor alarm indication. The 15 minute time period begins with a credible notification that a FIRE is occurring, or notification of a VALID fire detection system alarm. Verification of a fire detection system alarm includes actions that can be taken within the control room or other nearby location to ensure that the alarm is not spurious. A verified alarm is assumed to be an indication of a FIRE unless it is disproved within the 15-minute period by personnel dispatched to the scene. In other words, a personnel report from the scene may be used to disprove a sensor alarm if received within 15 minutes of the alarm, but shall not be required to verify the alarm.
Per AOP 913, the location of a fire can be determined by observing 1C40B alarm messages, Zone Indicating Unit (ZIU) alarms, or fire annunciators on panels 1C40 and 1C40A. The location of a fire can also be determined by verbal report of the person discovering the fire. Verification of the alarm in this context means those actions taken to determine that the control room alarm is not spurious.

==REFERENCES:==
: 1. Abnormal Operating Procedure (AOP) 913, Fire
: 2. Abnormal Operating Procedure (AOP) 914, Security HU2

EAL BASES DOCUMENT                                      EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                         Rev. 8 CATEGORY                                        PAGE 8 of 31 HU3         Release of Toxic or Flammable Gases Deemed Detrimental to Safe Operation of the Plant EVENT TYPE: Other Hazards and Failures OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:
: 1. Report or detection of toxic or flammable gases that could enter within the site area boundary in amounts that can affect normal operation of the plant.
OR
: 2. Report by Local, County or State Official for potential evacuation of site personnel based on offsite event.
DAEC EAL INFORMATION:
This Threshold Value is based on releases in concentrations within the site boundary that will affect the health of plant personnel or affecting the safe operation of the plant with the plant being within the evacuation area of an offsite event (i.e., tanker truck accident releasing toxic gases, etc.) The evacuation area is as determined from the DOT Evacuation Tables for Selected Hazardous Materials, in the DOT Emergency Response Guide for Hazardous Materials.
For the purposes of this EAL, CO2 (such as is discharged by the fire suppression system) is not toxic. CO2 can be lethal if it reduces oxygen to low concentrations that are immediately dangerous to life and health (IDLH). C0 2 discharge into an area is not basis for emergency classification under this IC unless: (1)Access to the affected area is required, and (2) C02 concentration results in conditions that make the area uninhabitable or inaccessible (i.e., IDLH).

==REFERENCES:==
: 1. UFSAR Section 2.2, Nearby Industrial, Transportation, and Military Facilities
: 2. UFSAR Section 6.4, Habitability Systems HU3

EAL BASES DOCUMENT                                        EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                           Rev. 8 CATEGORY                                         PAGE 9 of 31 HU4 Confirmed Security Event Which Indicates a Potential Degradation in the Level of Safety of the Plant EVENT TYPE: Security OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:
: 1. Suspected sabotage device discovered within the plant Protected Area.
OR
: 2. Suspected sabotage device discovered outside the Protected Area, in the plant switchyard or ISFSI.
OR
: 3. Confirmed tampering with safety related equipment.
OR
: 4. A hostage situation that disrupts normal plant or ISFSI operations.
OR
: 5. Civil disturbance OR strike which disrupts normal plant or ISFSI operations.
OR
: 6. Internal disturbance that is not short lived or that is not a harmless outburst involving one or more individuals within the Protected Area or ISFSI.
OR
: 7. Credible Security Threat of "LO" Severity.
DAEC EAL INFORMATION:
Security events which do not represent at least a potential degradation in the level of safety of the plant are reported under 10 CFR 73.71 or in some cases under 10 CFR 50.72. The term "suspected sabotage device" is used in place of "bomb device" for consistency with the DAEC Safeguards Contingency Plan.
Consultation with Security supervision is required to determine these Threshold Values.
EAL 1 describes a suspected sabotage device discovered within the Protected Area. It is a potential degradation of the level of safety of the plant and is an UNUSUAL EVENT.
HU4

EAL BASES DOCUMENT                                      EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                           Rev. 8 CATEGORY                                         PAGE 10 of 31 EAL 2 describes a suspected sabotage device discovered in the plant switchyard or ISFSI representing a potential degradation of the level of safety of the plant.
EAL 3 is for confirmed tampering and is adapted from the list of security plan contingencies.
EAL 4 identifies a hostage situation that disrupts normal plant or ISFSI operations. A hostage situation is considered to disrupt normal operations if it results in the inability to perform surveillance activities, alters unit operations, or as described in the security plan.
EAL 5 describes a civil disturbance or strike if considered to be a spontaneous activity that disrupts normal plant or ISFSI operations. A civil disturbance or strike is considered to disrupt normal plant operations if it initially disrupts normal ingress or egress to the owner controlled or protected area, or if it requires assistance from the Local Law Enforcement Agencies (LLEA) to control.
EAL 6 deals with suspicious internal disturbances that may have been planned by unauthorized personnel as a diversion to gain entry to the site property.
EAL 7 ensures that appropriate notifications for the security threat are made in a timely manner.
The determination of a Credible Security Threat of "LO" or "Hi" Severity is based on information found in NMC SE-0018, "Security Threat Assessment". The emergency response to a Credible Security Threat of "LO" Severity is initiated through AOP 914, "Security Events" and EPIP 2.8, "Security Threat". A Credible Security Threat of "Hi" Severity would escalate this classification to the ALERT status as an HA4. Only the plant to which the specific threat is made need declare the Notification of Unusual Event.
Suspected sabotage devices discovered within the plant Vital Area would result in escalation via other Security EALs.

==REFERENCES:==
: 1. Abnormal Operating Procedure (AOP) 914, Security Events
: 2. NMC SE-001 8, "Security Threat Assessment"
: 3. EPIP 2.8, "Security Threat"
: 4. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 NEI 99-01 Revision 4, January 2003
: 5. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 NEI 97-03 August 1997 HU4

EAL BASES DOCUMENT                                    EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                       Rev. 8 CATEGORY                                       PAGE 11 of 31 HU5 Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of an Unusual Event EVENT TYPE: Emergency Director Judgment OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:
: 1. Other conditions exist which in the judgment of the Emergency Director indicate that events are in process or have occurred which indicate a potential degradation of the level of safety of the plant. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.
DAEC EAL INFORMATION:
The EAL addresses conditions that fall under the Notification of Unusual Event emergency
 ;lassification description contained in NUMARC/NESP-007, NEI 99-01, Rev. 4 January 2003, that is retained under the generic methodology.

==REFERENCES:==
: 1. Emergency Plan Implementing Procedure (EPIP) 2.5, Control Room Emergency Response Operation
: 2. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007, NEI 99-01 Revision 4, January 2003 HU5

EAL BASES DOCUMENT                                EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                    Rev. 8 CATEGORY                                   PAGE 12 of 31 HAI Natural and Destructive Phenomena Affecting the Plant Vital Area EVENT TYPE: Natural Disasters and Destructive Phenomena OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:
: 1. Receipt of the Amber Operating Basis Earthquake Light and the wailing seismic alarm on 1C35
(+/- 0.06 gravity).
OR
: 2. Report of tornado striking Plant Vital Area.
OR
: 3. Assessment by the control room that damage has affected Safe Shutdown Areas.
OR
: 4. Vehicle crash affecting Plant Vital Areas.
OR
: 5. Sustained wind speed at or above 95 miles MPHI, affecting Plant Vital Areas.
OR
: 6. Turbine failure-generated missiles affecting Safe Shutdown Areas.
OR
: 7. River level above 767 feet.
OR
: 8. Water level above Max Safe Operating Limit in 2 or more areas AND Reactor shutdown is required.
OR
: 9. River level below 724 feet 6 inches.
HA1

EAL BASES DOCUMENT                                    EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                        Rev. 8 CATEGORY                                        PAGE 13 of 31 DAEC EAL INFORMATION:
There are no significant deviations from the generic EALs. For the events of concern here, the key issue is not the wind speed, earthquake intensity, etc., but whether there is resultant damage to equipment or structures required to achieve or maintain safe shutdown, regardless of the cause.
Determination of damage affecting the ability to achieve or maintain safe shutdown can be indicated by reports to the control room, physical observation or by Control Room/local control station instrumentation.
EAL Threshold Value 1 addresses OBE events that are detected in accordance with AOP 901.
For DAEC, the OBE is associated with a peak horizontal acceleration of +/- 0.06 Gravity.
DAEC EAL Threshold Value 2 addresses report of a tornado striking a plant vital area.
DAEC EAL Threshold Value 3 addresses a report to the control room of damage affecting safe shutdown areas. The reported damage can be from tornadoes, high winds, flooding, missiles, collisions, or any other cause. The missiles mentioned here can be from any cause, e.g., tornado-jenerated; turbine, pump or other rotating machinery catastrophic failure; or generated from an explosion.
DAEC EAL Threshold Value 4 addresses vehicle (automobile, aircraft, forklift, truck or train) confirmed crashes affecting plant vital areas. This does not include vehicle crashes with each other or damage to office or warehouse structures.
DAEC EAL Threshold Value 5 addresses sustained high wind speeds as measured by the 33-Foot or 156-Foot elevations on the Meteorological Tower. Sustained wind speed means the baseline wind speed measured by meteorological tower that does not include gusts. The design basis wind speed is 105 miles per hour. However, the meteorological instrumentation is only capable of measuring wind speeds up to 100 miles per hour. Thus the alert level for sustained high wind speed, 95 miles per hour, is selected to be on-scale for the meteorological instrumentation and to conservatively account for potential measurement errors.
DAEC EAL Threshold Value 6 addresses Turbine failure-generated missiles affecting safe shutdown areas. This threshold addresses the threat to safety related equipment from missiles generated by main turbine rotating component catastrophic failures.
HAI

EAL BASES DOCUMENT                                     EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                        Rev. 8 CATEGORY                                       PAGE 14 of 31 PerAOPs 913 and 914, the following areas are identified as safe shutdown areas and are shown on the EAL tables. This table is displayed as an aid to the Emergency Coordinator in deternmining appropriate areas of concem.
Safe ShutdownNital Areas                                I Category                                   Area Electrical       Switchyard, 1G31 DG and Day Tank Rooms, 1G21 DG and Day Power            Tank Rooms, Battery Rooms, Essential Switchgear Rooms, Cable Spreading Room Heat Sink/
Coolant          Torus Room, Intake Structure, Pumphouse Supply Containment       Drywell Torus Emergency         NE, NW, SE Corner Rooms, HPCI Room, RCIC Room, RHR Systems           Valve Room, North CRD Area, South CRD Area, CSTs 0theher1          Control C55156Building,  Remote Area, SBGT    RoomShutdown Panel 1C388 Area, Panel DAEC EAL Threshold Value 7 addresses river water levels exceeding design flood water levels.
All Seismic Category I structures and non-seismic structures housing Seismic Category I equipment are designed to withstand the hydraulic head resulting from the "maximum probable flood" to which the site could be subjected. The design flood water is at elevation 767.0 ft. Major equipment penetrations in the exterior walls are located above elevation 767.0 ft. Openings below the flood level are either watertight or are provided with means to control the inflow of water in order to ensure that a safe shutdown can be achieved and maintained. Consideration has also been given to providing temporary protection for openings in the exterior walls up to flood levels of 769.0 ft. All buildings were also checked for uplift (buoyancy) for a flood level at elevation 767.0 ft, and the minimum factor of safety used was 1.2. Therefore, DAEC EAL 7 uses as its threshold flood water levels above 767 feet.
DAEC EAL Threshold Value 8 addresses internal flooding consistent with the requirements of EOP 3, Secondary Containment Control. If RPV pressure reduction will decrease leakage into secondary containment then this is due to leakage from the primary system, which is addressed by the Fission Barrier Table indicators and System Malfunction EALs, and is not addressed here.
Therefore, EAL 8 addresses conditions in which water level in two or more areas is above Maximum Safe Operating Limits and reactor shutdown is required. Required means that the reactor shutdown was procedurally mandated by EOP 3 and is not merely performed as a precaution or inadvertently. Maximum Safe Operating Limits are defined as the highest parameter value at which neither (1) equipment necessary for safe shutdown of the plant will fail nor (2) personnel access necessary for the safe shutdown of the plant will be precluded. The internal flooding can be due to system malfunctions, component failures, or repair activity mishaps HA1

EAL BASES DOCUMENT                                        EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                            Rev. 8 CATEGORY                                          PAGE 15 of 31 (such as failed freeze seal) that can threaten safe operation of the plant. This includes water intrusion on equipment that is not designed to be submerged (e.g., motor control centers).
The maximum safe operating water level limits are taken from EOP 3 and are shown on the table below:
Maximum Operating Limits - Water Levels Affected Location                    Indicator       Maximum        Maximum Safe Normal OL       OL HPCI Room Area                       LI 3768          2 inches      6 inches RCIC Room Area                       LI 3769          3 inches      6 inches A RHR Corner Room SE Area            LI 3770          2 inches       10 inches B RHR Corner Room NW Area            LI 3771          2 inches       10 inches Torus Area                           Ll 3772         2 inches        12 inches DAEC EAL Threshold Value 9 addresses the effects of low river water level. The intake structure for the safety-related water supply systems (river water, RHR service water, and emergency service water) is located on the west bank of the Cedar River. The overflow weir is at elevation 724 feet 6 inches. River level at or below this elevation will result in all river flow being diverted to
'he safety related water supply systems. The top of the intake structure around the pump wells is at elevation 724 feet. If the river water level dropped to this level, the pump suction would have no continuous supply. Therefore, this EAL uses a threshold of water level below 724 feet 6 inches as a potential substantial degradation of the ultimate heat sink capability.
In this EAL, "Vital Area" is defined as plant structures or areas containing equipment necessary for a safe shutdown, i.e., synonymous with Safe Shutdown Area.

==REFERENCES:==
: 1. Abnormal Operating Procedure (AOP) 901, Earthquake
: 2. Abnormal Operating Procedure (AOP) 902, Flood
: 3. Abnormal Operating Procedure (AOP) 903, Tornado
: 4. Abnormal Operating Procedure (AOP) 913, Fire
: 5. Abnormal Operating Procedure (AOP) 914, Security Events
: 6. UFSAR Chapter 3, Design of Structures, Components, Equipment, and Systems
: 7. Bechtel Drawing BECH-M017, Equipment Location - Intake Structure Plans at Elevations, Rev. 6
: 8. EOP Basis Document, EOP 3 - Secondary Containment Control Emergency Operating Procedure (EOP) 3, Secondary Containment Control HA1

EAL BASES DOCUMENT                                    EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                         Rev. 8 CATEGORY                                       PAGE 16 of 31 HA2 Fire or Explosion Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown EVENT TYPE: Fire OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:
: 1. Fire or explosion affecting any of the following areas:
* Reactor, Turbine, Control, Admin/Security
* Intake Structure
* Pump house AND
: 2. Affected system parameter indications show degraded performance or plant personnel report VISIBLE DAMAGE to permanent structures or equipment within the specified area.
DAEC EAL INFORMATION:
There is no significant deviation from the generic EAL. Of particular concern for this EAL are fires that may be detected in the reactor building, control building, turbine building, pumphouse, and intake structure as shown in Tabs 1 and 3 of AOP 913. Damage from fire or explosion can be indicated by physical observation, or by Control Room/local control station instrumentation. No attempt is made in this EAL to assess the actual magnitude of the damage.
Per AOP 913, the location of a fire can be determined by observing 1C40B alarm messages, Zone Indicating Unit (ZIU) alarms, or fire annunciators on panels 1C40 and 1C40A.
NOTE Scope of Systems and Equipment of concern was established by review of Appendix R Safe Shutdown credited systems. Only those systems directly affecting safe shutdown or heat removal are listed for consideration, due to fire damage. Support Systems and equipment such as HVAC and specific instrumentation, while included in Appendix R analysis is not considered an immediate threat to the ability to shutdown the plant and remove decay heat.
HA2

EAL BASES DOCUMENT                                   EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                        Rev. 8 CATEGORY                                     PAGE 17 of 31 Systems & Equipment of Concern
                        . Reactivity Control
                        . Containment (DrywelllTorus)
                        . RHR/Core Spray/SRVs
                        . HPCI/RCIC
                        . RHRSW/River Water/ESW
                        . Onsite AC Power/EDGs
                        . Offsite AC Power
                        . Instrument AC
                        . DC Power
                        . Remote Shutdown Capability This EAL addresses a FIRE / EXPLOSION and not the degradation in performance of affected systems. System degradation is addressed in the System Malfunction EALs. The reference to damage of systems is used to identify the magnitude of the FIRE / EXPLOSION and to discriminate against minor FIREs I EXPLOSIONs. The reference to safety systems is included to discriminate against FIREs / EXPLOSIONs in areas having a low probability of affecting safe operation. The significance here is not that a safety system was degraded but the fact that the FIRE / EXPLOSION was large enough to cause damage to these systems. Thus, the designation of a single train was intentional and is appropriate when the FIRE I EXPLOSION is 3rge enough to affect more than one component. Lagging fires, fires in waste containers or any miscellaneous fires that may be in the vicinity of safety systems, but do not cause damage to these systems, should NOT be considered for this EAL.
With regard to EXPLOSIONS, only those EXPLOSIONS of sufficient force to damage permanent structures or identified equipment required for safe operation, should be considered. As used here, an EXPLOSION is a rapid, violent, unconfined combustion, or a catastrophic failure of pressurized equipment, that potentially imparts significant energy to near-by structures and materials. The occurrence of the EXPLOSION with reports of evidence of damage (e.g.,
deformation, scorching) is sufficient for the declaration. The Emergency Director also needs to consider any security aspects of the EXPLOSIONS, if applicable.

==REFERENCES:==
: 1. Abnormal Operating Procedure (AOP) 913, Fire
: 2. Abnormal Operating Procedure (AOP) 914, Security Events
: 3. Abnormal Operating Procedure (AOP) 915, Shutdown Outside Control Room
: 4. UFSAR Section 6.4, Habitability Systems HA2

EAL BASES DOCUMENT                                  EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                       Rev. 8 CATEGORY                                      PAGE 18 of 31 HA3 Release of Toxic or Flammable Gases Within a Facility Structure Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown EVENT TYPE: Other Hazards and Failures OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:
: 1. Report or detection of toxic gases within a Safe Shutdown Area in concentrations that will be life threatening to plant personnel.
OR
: 2. Report or detection of flammable gases within a Safe Shutdown Area in concentrations that will affect the safe operation of the plant.
DAEC EAL INFORMATION:
This EAL, in addition to EAL HAS, also addresses entry of toxic gases that may result in control room evacuation in accordance with AOP 915.
For the purposes of this EAL, CO2 (such as is discharged by the fire suppression system) is not toxic. CO2 can be lethal if it reduces oxygen to low concentrations that are immediately dangerous to life and health (IDLH). C02 discharge into an area is not basis for emergency classification under this IC unless: (1) Access to the affected area is required, and (2) C0 2 concentration results in conditions that make the area uninhabitable or inaccessible (i.e., IDLH).
TOXIC - Exposure to the worker in excess of the limits specified in 29 CFR 1910.1000. In practice, this should be considered for concentrations which are capable of producing incapacitation of the worker.
The source of the release is NOT of immediate concern for these threshold values. The concern is for the health and safety of plant personnel and their ability to maintain the plant in a safe operating condition.
This EAL is based on gases that have entered plant structures that will affect the safe operation of the plant. These structures include buildings and areas contiguous to plant vital areas and other significant buildings or areas. The intent of this EAL is NOT to include buildings or other areas that are NOT contiguous or immediately adjacent to plant vital areas.
HA3

EAL BASES DOCUMENT                                  EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                     Rev. 8 CATEGORY                                     PAGE 19 of 31 PerAOPs 913 and 914, the following areas are identified as safe shutdown areas and are shown on the EAL tables. This table is displayed as an aid to the Emergency Director in determining appropriate areas of concem.
Safe Shutdown/Vital Areas Category                                   Area Electrical      Switchyard, 1G31 DG and Day Tank Rooms, 1G21 DG and Day Power           Tank Rooms, Battery Rooms, Essential Switchgear Rooms, Cable Spreading Room Heat Sink/
Coolant         Torus Room, Intake Structure, Pumphouse Supply Containment     Drywell, Torus Emergency       NE, NW, SE Corner Rooms, HPCI Room, RCIC Room, RHR Systems         Valve Room, North CRD Area, South CRD Area, CSTs
 )ther          Control Building, Remote Shutdown Panel 1C388 Area, Panel 1C55/56 Area, SBGT Room

==REFERENCES:==
: 1. Abnormal Operating   Procedure (AOP) 913, Fire
: 2. Abnormal Operating   Procedure (AOP) 914, Security Events
: 3. Abnormal Operating   Procedure (AOP) 915, Shutdown Outside Control Room
: 4. UFSAR Section 6.4,   Habitability Systems HA3

EAL BASES DOCUMENT                                    EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                         Rev. 8 CATEGORY                                       PAGE 20 of 31 HA4 Confirmed Security Event in a Plant Protected Area EVENT TYPE: Security OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE
: 1. Intrusion into plant Protected Area by a hostile force.
OR
: 2. Sabotage device discovered in the plant Protected Area.
OR
: 3. Any security event of increasing severity that persists for > 30 minutes:
: a. Credible bomb threats
: b. Extortion
: c. Suspicious Fire or Explosion
: d. Significant Security System Hardware Failure
: e. Loss of Guard Post Contact OR
: 4. Credible Security Threat of "HI" Severity DAEC EAL INFORMATION:
Consultation with Security supervision is required to determine these Threshold Values.
EAL 1 is an intrusion of a hostile force into the Protected Area representing a potential for a substantial degradation of the level of safety of the plant. A civil disturbance, which penetrates the Protected Area, can be considered a hostile force.
EAL 2 is the discovery of a sabotage device in the Plant Protected area.
EAL 3 security events represent an escalated threat to plant safety above that contained in the Unusual Event. Under this EAL, adversaries within the Protected Area are not yet affecting nuclear safety systems, engineered safety features, or reactor shutdown capability that are located within HA4

EAL BASES DOCUMENT                                    EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                         Rev. 8 CATEGORY                                      PAGE 21 of 3 the vital area. A security event is considered to be "of increasing severity" if events are NOT under control of the security force within 30 minutes. Intrusion into a vital area by a hostile force will escalate this event to a Site Area Emergency.
EAL 4 is the determination of "Credible Security Threat of HI Severity" based on information found in NMC SE-0018, "Security Threat Assessment". The emergency response to a "Credible Security Threat of HI Severity" is initiated through AOP 914, "Security Events" and EPIP 2.8, "Security Threat".

==REFERENCES:==
: 1. NMC SE-001 8, "Security Threat Assessment"
: 2. Abnormal Operating Procedure (AOP) 914, Security Events
: 3. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 NEI 99-01 Revision 4, January 2003
: 4. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 NEI 97-03 August 1997 HA4

EAL BASES DOCUMENT                                EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                    Rev. 8 CATEGORY                                   PAGE 22 of 31 HA5 Control Room Evacuation Has Been Initiated EVENT TYPE: Control Room Evacuation OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:
: 1. Entry into AOP 915 and initiation of control room evacuation.
DAEC EAL INFORMATION:
The applicable procedure for control room evacuation at DAEC is AOP 915.
Evacuation of the Control Room represents a potential for substantial degradation of the level of safety of the plant and therefore requires an ALERT declaration. Additional support, monitoring and direction is required and accomplished by activation of the Technical Support Center at the ALERT classification level.

==REFERENCES:==
: 1. Abnormal Operating Procedure (AOP) 915, Shutdown Outside Control Room
: 2. UFSAR Section 6.4, Habitability Systems HA5

EAL BASES DOCUMENT                                   EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                         Rev. 8 CATEGORY                                      PAGE 23 of 31 HA6 Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of an Alert EVENT TYPE: Emergency Director Judgment OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:
: 1. Other conditions exist which in the judgment of the Emergency Director indicate that events are in process or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.
DAEC EAL INFORMATION:
The EAL addresses conditions that fall under the Alert emergency classification description contained in NUMARC/NESP-007, NEI 99-01, Rev. 4 January 2003.

==REFERENCES:==
: 1. Emergency Plan Implementing Procedure (EPIP) 2.5, Control Room Emergency Response Operations
: 2. NEI Methodology for Development of Emergency Action Levels, NUMARC/NESP-007, NEI 99-01 Revision 4, January 2003 HA6

EAL BASES DOCUMENT                                    EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                         Rev. 8 CATEGORY                                      PAGE 24 of 31 HSI Confirmed Security Event in a Plant Vital Area EVENT TYPE: Security OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:
: 1. Intrusion into plant Vital Area by a hostile force.
OR
: 2. A security event that results in the loss of control of any Vital Area (other than the Control Room).
OR
: 3. IMMINENT loss of physical control of the facility (remote shutdown capability) due to a security event.
OR
: 4. A confirmed sabotage device discovered in a Vital Area.
DAEC EAL INFORMATION:
Consultation with Security supervision is required to determine these Threshold Values.
IMMINENT - Mitigation actions have been ineffective and trended information indicates that the event or condition will occur within 2 hours.
This threshold value escalates from the ALERT Protected Area intrusion to a Vital Area intrusion of a hostile force.
A security event is as defined in the Safeguards Contingency Plan.
Loss of physical control of the Control Room OR loss of physical control of the remote shutdown capability due to a security event, is to be classified as a GENERAL EMERGENCY per Initiating Condition HG1.
A "confirmed sabotage device" is a determination made by the security force through the Security Plan, Contingency procedures and other guidance documentation.
HS1

EAL BASES DOCUMENT                                     EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                         Rev. 8 CATEGORY                                        PAGE 25 of 31 This class of security events represents an escalated threat to plant safety above that contained in HA4, Security Event in a Plant Protected Area, in that a hostile force has progressed from the Protected Area to the Vital Area. Under the condition of concern here, the adversaries are considered to be in a position to directly and negatively affect nuclear safety systems, engineered safety features, or reactor shutdown capability.

==REFERENCES:==
: 1. Abnormal Operating Procedure (AOP) 914, Security Events
: 2. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, January 2003 HS1

EAL BASES DOCUMENT                                      EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                         Rev. 8 CATEGORY                                       PAGE 26 of 31 HS2 Control Room Evacuation Has Been Initiated and Plant Control Cannot Be Established EVENT TYPE: Control Room Evacuation OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:
: 1. Control room evacuation has been initiated.
AND
: 2. Control of the plant cannot be established per AOP 915 within 20 minutes.
DAEC EAL INFORMATION:
There is no significant deviation from the generic EAL. The applicable procedure for control room evacuation at DAEC is AOP 915. Based on the results of the analysis described below, DAEC uses 20 minutes as the site-specific time limit for establishing control of the plant. DAEC has satellite panels associated with the remote shutdown panel at various locations through out the plant. Control of the plant from outside the control room is assumed when the controls are transferred to remote shutdown panel 1C388 in accordance with AOP 915.
The Emergency Director is expected to make a reasonable, informned judgment within the 20 minute time limit that control of the plant from the remote shutdown panel has been established.
The intent of the EAL is that control of important plant equipment and knowledge of important plant parameters has been achieved in a timely manner. Primary emphasis should be placed on those components and instruments that provide protection of and information about safety functions. At a minimum, consistent with the Appendix R safe shutdown analysis described above, these safety functions include reactivity control, maintaining reactor water level, and decay heat removal.
General Electric performed analyses to demonstrate compliance with the requirements of 10 CFR 50 Appendix R for DAEC. The evaluation of Reactor Coolant Inventory was performed using the GE evaluation model (SAFE). The SAFE code determines if the reactor coolant inventory is above the TAF during the safe shutdown operation. If core uncovery occurs, the fuel clad integrity evaluation is performed by determining the duration of the core uncovery and the resulting peak cladding temperature (PCT). The PCT calculations were performed by incorporating the SAFE output into the Core Heatup Analysis code (CHASTE). The details of these calculations are provided in Section 4 of the final report for DAEC Appendix R analyses ("Safe Shutdown Appendix R Analyses for Duane Arnold Energy Center", MDE-44-036).
HS2

EAL BASES DOCUMENT                                   EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                         Rev. 8 CATEGORY                                      PAGE 27 of 31 The required analyses include evaluation of the safe shutdown capability of the remote shutdown system for various control room fire events assuming: (1) no spurious operation of equipment, (2) spurious operation of a safety-relief valve (SRV) for 20 minutes, (3) spurious operation of a SRV for 10 minutes, and (4) spurious leakage from a one-inch line. The analyses show that the worst case spurious operation of SRV or isolation valves on a one-inch liquid line (high-low pressure interface) will not affect the safe shutdown ability of the remote shutdown system for DAEC in case of a fire requiring control room evacuation before the identified time limit for the necessary operator actions at the auxiliary shutdown panels. For the limiting cases of worst case spurious leakage from a one-inch line and spurious operation of a SRV, operator control within 20 minutes would not impact the integrity of the fuel clad, the reactor pressure vessel, and the primary containment.

==REFERENCES:==
: 1. Abnormal Operating Procedure (AOP) 915, Shutdown Outside Control Room
: 2. General Electric Report MDE-44-0386, Safe Shutdown Appendix R Analysis for DAEC, March 1986
: 3. UFSAR Section 6.4, Habitability Systems
: 1. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, January 2003 HS2

EAL BASES DOCUMENT                                     EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                         Rev. 8 CATEGORY                                        PAGE 28 of 31 HS3 Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of Site Area Emergency EVENT TYPE: Emergency Director Judgment OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:
: 1. Other conditions exist which in the Judgment of the Emergency Director indicate that events are in process or have occurred which involve actual or likely major failures of plant functions needed for protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.
DAEC EAL INFORMATION:
There is no significant deviation from the generic EAL.
The EAL addresses conditions that fall under the Site Area Emergency classification description contained in NUMARC/NESP-007, NEI 99-01, Rev. 4 January 2003.

==REFERENCES:==
: 1. Emergency Plan Implementing Procedure (EPIP) 2.5, Control Room Emergency Response Operation
: 2. NEI Methodology for Development of Emergency Action Levels, NUMARC/NESP-007, NEI 99-01 Revision 4, January 2003 HS3

EAL BASES DOCUMENT                                  EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                      Rev. 8 CATEGORY                                     PAGE 29 of 31 HG1 Security Event Resulting in Loss Of Ability to Reach and Maintain Cold Shutdown EVENT TYPE: Security OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:
: 1. Loss of physical control of the control room due to security event.
OR
: 2. Loss of physical control of remote shutdown capability due to security event.
DAEC EAL INFORMATION:
This EAL is an escalation of the SITE AREA EMERGENCY, HS1 declaration for a hostile force intrusion of a Vital Area taking physical control of either the Control Room OR taking over the emote shutdown capabilities which results in the loss of physical control of the facility. This also includes areas where any switches that transfer control of safe shutdown equipment to outside the control room are located.

==REFERENCES:==
: 1. Abnormal Operating Procedure (AOP) 914, Security Events
: 2. UFSAR Section 6.4, Habitability Systems HG1

EAL BASES DOCUMENT                                 EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                       Rev. 8 CATEGORY                                     PAGE 30 of 31 HG2 Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of General Emergency EVENT TYPE: Emergency Director Judgment OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:
Other conditions exist which in the Judgment of the Emergency Director indicate that events are in process or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.
DAEC EAL INFORMATION:
IMMINENT - Mitigation actions have been ineffective and trended information indicates that the event or condition will occur within 2 hours.
POTENTIAL - Mitigation actions are not effective and trended information indicates that the parameters are outside desirable bands and not stable or improving.
This Emergency Action Level allows for classification of events which in the judgment of the Emergency Director warrant the GENERAL EMERGENCY classification but do not fit into any other GENERAL EMERGENCY criteria. Emergency Director judgment is to be based on known conditions and the expected response to mitigating activities within a short time period arbitrarily set at 2 hours. Classification of a GENERAL EMERGENCY is not to be delayed pending an extended evaluation of possibilities and probabilities. If time allows and the offsite response organizations are active, consultation with the effected state and the NRC is prudent prior to classification.
HG2

EAL BASES DOCUMENT                              EBD-H HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY                Rev. 8 CATEGORY                                  PAGE 31 of 31

==REFERENCES:==
: 1. Emergency Plan Implementing Procedure (EPIP) 2.5, Control Room Emergency Response Operation
: 2. NUREG-0654/FEMA-REP-1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, Revision 1, October 1980, Appendix 1
: 3. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, January 2003 HG2

I .
                           -EAL BASES DOCUMENT-.                               EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 1 of 43 I
                                              -Usage Level
                                                 ,e e     Us l       ~Reference Usel I[   Effective Date:
k ll I
I TECHNICALUREVIEW Prepared by:                  c      Z      4             Date:    1 Reviewed by:     &eV              OX       J7leA   -      Date: 01    2q9 4 Independent RevieweU Reviewed by-                                              Date:    i /i'cY Operations Reviewer PROCEDURE-APPROVAL I am responsible for the technical content-of this procedure and for obtaining the necessary approval from the State and County Emergency Management officials prior to implementation.
Documentation of State and County Emergency Management approval is via NEP-2004-0005.
I .
Approved by:         '-C                                  Date:
I I Manager, Emergency Planning                        .

I d0
                 *L.EAL BASES DOC'UMENtW-                            ''"                           EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 2 of 43 Table of Contents SU1 Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes                                                      .4 SU2 Inability to Reach Required Shutdown Within Technical Specification Limits ...............                                                                                5 SU3 Unplanned Loss of Most or All Safety'System Annunciation or Indication in the Control Room for Greater Than 15 Minutes                                                  .       6 SU4 Fuel Clad Degradation                                                      .8 SU5 RCS Leakage.12 SU6 Unplanned Loss of All Onsite or Offsite Communications Capabilities .15 SU7 Unplanned Loss of Required DC Power During Cold Shutdown or Refuel Mode For Greater Than 15 Minutes .17 SA1 Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses During Cold Shutdown or Refuel Conditions .                                                       19 SA2 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was Successful........................................................................................................ 21 SA3 Inability to Maintain Plant in Cold Shutdown .23 SA4 Unplanned Loss of Most or All Safety System Annunciation or Indication in Control Room With Either (1) a Significant Transient in Progress, or (2) Compensatory Non-Alarming Indicators are Unavailable ..........                                                                                             25

:EAL BASES DOCUMENT.",'..:-.
                   .-                                          .'   EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 3 of 43 SA5 AC Power Capability to Essential Busses Reduced to a Single' Power Source for Greater Than 15 Minutes Such That Any Additional Single Failure Would Result in Station Blackout ................. 27.
SS1 Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses .........................                                        28 SS2 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was NOT Successful..........................   !                                      29 SS3 Loss of All Vital DC Power ..............................                          31 SS4 Complete Loss of Heat Removal Capacity Error! Bookmark not defined.-
SS5 Loss of Water Level in the Reactor Vessel That Has or Will Uncover Fuel in the Reactor Vessel .........................           ...        35 SS6 Inability to Monitor a Significant Transient in Progress ......................... 37 SG1 Prolonged Loss of All Offsite.Power and Prolonged Loss of All Onsite AC Power                        .............         .                    39 SG2 Failure of the Reactor Protection System to Complete an Automatic Scram and Manual Scraam was NOT Successful and There is' Indication of an Extreme Challenge to the Ability to Cool the Core.... 42

I
:-'     ,LBASE&ocuMENT               i:''.:'.                EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY' Page 4 of 43 SUI Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:.
The following conditions exists:
: 1. Loss of power to Startup (1X3) and Standby (1X4) transformers is expected to last for greater than 15 minutes.
AND
: 2. Emergency Busses 1A3 and 1A4 are powered by their respective Standby Diesel Generators.
DAEC EAL INFORMATION:
This event is a precursor of a more serious Station Blackout condition and is thus considered as a potential'degradation of the level of safety of the plant. It is possible to be operating within Technical Specification LCO Action Statement time limits and make a declaration of an Unusual Event in accordance with this EAL.
The intent of this EAL is to declare an UNUSUAL EVENT when offsite power has been lost and both of the emergency diesel generators have successfully started and energized their respective ECCS bus.

==REFERENCES:==
: 1. Abnormal Operating Procedure (AOP) 301, Loss of Essential Electrical Power
: 2. UFSAR Section 8.2, Offsite Power System
: 3. NEI Methodology for Development of Emergency Action Levels, NUMARC/NESP-007 Revision 4, January 2003 Sul

                        ', '  .EAL BASES DOCUMENT ,-                            EBD-S Rev. 6 SYSTEM MALFUNCTIONCATEGORY Page 5 of 43  :
SU2 Inability to Reach Required Shutdown Within Technical Specification Limits EVENT TYPE: Inability to Reach or Maintain Shutdown Conditions OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EAL THRESHOLD VALUE:
The following condition exists:
: 1. Plant NOT brought to required mode within the applicable LCO Action Statement Time.
DAEC EAL INFORMATION:
Limiting Conditions for Operations (LCO) require the plant to be brought to a specific condition when an LCO has been entered. Depending on the circumstances this may or
. may not be an emergency or a precursor to a more serious event. In any case when a plant initiates a shutdown due to having entered an LCO action statement a one-hour report must be made under 10CFR50.72(b) non-emergency events. The'plant'is within its safety envelope when being shutdown within the allowable action statement time of a Technical Specification. An immediate classification of UNUSUAL EVENT should be made when the plant is NOT brought to the required mode within the allowable action statement time of any Technical Specification LCO. Declaration is based on the time at which the LCO Action Statement specified time period elapses and is NOT related to how long a condition may have existed.

==REFERENCES:==
: 1. DAEC Technical Specifications
: 2. NEI Methodology forDevelopment of Emergency Action Levels NUMARC/NESP-007 Revision 4, January 2003 SU2

EArBASES&DOCUMENT-                                 EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 6 of 43 SU3 Unplanned Loss of Most or All Safety System Annunciation or Indication in the Control Room for Greater Than 15 Minutes EVENT TYPE: Instrumentation/Communication OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EAL THRESHOLD VALUE:
The following conditions exist:
: 1. Unplanned loss of most or all 1C03, 1C04 and 1C05 Annunciators or indicators associated with Critical Safety Functions for greater than 15 minutes.
AND
: 2. Compensatory non-alarming indications are available.
DAEC EAL INFORMATION:
Control room panels 1C03, 1C04, and 1C05 contain the annunciators associated with safety systems at DAEC. Therefore, the DAEC EAL addresses unplanned loss of most annunciators on these panels. Compensatory non-alarming indications include the plant process computer, SPDS, plant recorders, or plant instrument displays in the control room. Unplanned loss of annunciators or indicators excludes scheduled maintenance and testing activities.
Under the conditions of concem , entry into AOP 302.2, Loss of Alarm Panel Power, would be made. The procedure requires alerting operators on shift to the nature of the lost annunciation. It further requires that operators be attendant and responsive to abnormal indications that relate to those systems and components that have lost annunciation.
                                                                                         'VI SU3

                               .-,EAL         BASES DOCUIMENT.-                    EBD-S
                        , .. -  -A - bi.;,-..   .I>. :-- . -
Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 7 of 43 MOST - 75% of safety system annunciators or indicators are lost OR a significant risk that a degraded plant condition could go undetected exists. The use and definition of MOST is not intended to require a detailed count of lost annunciators or indicators but should be used as a guide to assess the ability to monitor the operation of the plant.
Unplanned loss of critical safety function indicators (i.e., EOP/EAL parameters) for greater than 15 minutes may preclude operators from taking actions to mitigate a transient.
Annunciators on 1C03, 1C04, and 1C05 share a common power supply from 125 VDC Division I that is fed through circuit breaker 1D13.
Indications of loss of annunciators associated with safety systems include:
125 VDC charger, battery, or system annunciators on control room panel 1C08 Loss of 'sealed in" annunciators at-affected panels Failure of affected annunciator panels shiftily testing by plant operators
* Expected alarms are not received V.>
* Computer point ID B350 indicates "NSS ANN DC LOSS TRBL." (Loss of DC power to panels 1C03, 1C04, and 1C05)

==REFERENCES:==
: 1. Operating Instruction (01) No. 317.2 Annunciator System
: 2. Abnormal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power
: 3. Abnormal Operating Procedure (AOP) 302.2, Loss of Alarm Panel Power
: 3. NEI Methodology for Development of EmergencyAction Levels NUMARC/NESP-007 Revision 4, January 2003 SU3

                - 'i--  EAL BASES, DOCUMENT&#xa2;-,:- '            -                EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 8 of 43  l SU4 Fuel Clad Degradation EVENT TYPE: Coolant Activity OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EAL THRESHOLD VALUE:
One of the following:.
: 1. Pretreat radiation monitor (RM-4104) reading ABOVE 4E+3 mr/hr.
OR
: 2. Reactor Coolant sample activity value indicating ABOVE 1.2 0Ci/ml dose equivalent 1-131.
DAEC EAL INFORMATION:
There are no significant deviations from the generic EALs. These EALs are precursorsof more serious fuel clad degradation and are thus considered as indicating a potential degradation of the level of safety of the plant. Thus, it is possible to be operating within Technical Specification LCO Action Statement time limits foriodine spikes and make a declaration of an Unusual Event. DAEC mode applicability for these EALs are consistent with the Tech Specs.
EAL 1 addresses pretreat rad monitor (RM-4104) exceeding 4E+3 mr/hr. The calculation supporting this value is described below. This indication would be validated by ensuring the pretreat rad monitor reading is determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or coolant sampling results. This reading would be displayed on Control Room panels 1C-02 and 1C-10 on recorder RR-4104.
As specified in the generic methodology, DAEC EAL 2 addresses coolant samples exceeding technical specification 3.4.6, coolant activity less than or equal to 1.2 gCi/ml dose equivalent 1-131.
SU4

                             '.EAL^BASESDOCUMENT'                                    EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 9 of 43    .
Radiological Engineering Calculation 94-014A and UFSAR Table 15.4-1 were reviewed to determine a suitable EAL threshold for the pretreat rad monitor reading corresponding to the Tech Spec 3.4.6 coolant activity limit of 1.2 pCVml of dose equivalent 1-131. Using the condenser noble gas source term for the control rod drop accident of 2.38 E +06 Curies shown on UFSAR Table 15.4-1 and the condenser free volume of 55,000 cubic'feet, an initial noble gas concentration in the condenser offgas line is determined. Because the offgas flow rate is very small (about 50 standard cubic feet per minute) compared to the total condenser free volume, dilution of the condenser noble gas concentration due to offgas'flow is not considered in the calculation shown below. Decrease in the noble gas source term due to decay of short-lived noble gas radioisotopes and offgas flow dilution effects are addressed by rounding down the value calculated as shown below.
Calculation 94-014A used an exposure rate method based on using a source term consisting of a defined mixture of noble gases and iodine from the control rod drop accident as described in the DAEC UFSAR, Section 15.4. The calculation assumed that the activity is released instantly and immediately reached in equilibrium with the reactor coolant inventory. Using this calculation, using dose correction factors (DCFs) for child KS thyroid dose from Reg. Guide ;1 09, and adjusting for the specific gravity (0.736) of saturated water at 1050 psia (fluid conditions assumed in the calculation) to adjust for standard conditions, the 1-131 dose equivalent (in units of pCi/ml assuming 1 cc equals 1 ml) is determined for this event. This 'result is then linearly scaled for rad monitor readings corresponding to the Tech'Spec'3.4:6 allowable primary coolant activity of 1.2 pCi/ml 1-131 dose equivalent,'i.e., the relative mixture of noble gases and iodine is assumed to remain constant. 1-129 is ignored because it has no effect on the calculation result.
Isotope          DCF          Concentration            Correction Factor     1-131 DEQ ([pCi/cc)
(mrem/pci)          (pCicc)
                                      -.          -.   [DCFISOTOPE / DCF- 131] i
                                              ._         .       0.736 1-131       4.39  E-03        1.6.E+01'              ... 1.4 E+00               2.2 E+01 1-132       5.23  E-05        2.2 E+01    .       .      1.6 E-02               3.6 E-01 1-133       1.04  E-03        3.1 E+01.        -         3.2 E-01               1.0 E+01 1-134       1.37  E-05        3.4 E+01                   4.2 E-03                1.4 E-01 1-135        2.14  E-04        2.9 E+01                   6.6 E-02               1.9 E+00 TOTAL             -                                                               3.4 E+01 Therefore, for this event, a coolant activity of 34 pCilcc 1-131 dose equivalent is calculated. Scaling the results for 1.2 pC/cc 1-131 dose equivalent, a suitable condenser SU4

EAL' BASES' DOCUMENT.S" '.;-                         EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 10 of 43 I
source term and corresponding initial concentration in the offgas flow is then determined.
This is then converted to a pretreat rad monitor reading by use of the monitor efficiency factor:               -      -
Pretreat Rad Monitor (RM-4104) Reading NG concentrationclad damage = NG concentrationROD DROP X [1.2 1Ci/cc /34 gCi/cc ]
= [2.38 E +6 Ci x 1 E+6 0Ci ICi]/ [5.5 E+4 ft3 x 2.83 E+4 cC/t 3 ] X [1.2 pCi/cc /34 pCi/cc]
= 1529 pCi x 0.0353 = 54.0 pCi/cc Pretreat rad monitor reading = NG concentration X Rad monitor efficiency Rad monitor efficiency = 89.2 mrlhr/ plCi/cc, therefore:
Pretreat rad monitor reading = 89.2 X 54.0 = 4800 mr/hr To account for isotopic decay and dilution effects of offgas flow, round down to 4E+03 mr/hr.-
The calculation results were also reviewed to determine if suitable values for the main steam line (MSL) radiation monitors could be developed. As shown above, the rod drop accident corresponds to coolant activity of 34 pCi/cc 1-131 dose equivalent. As determined by the reference calculation, this corresponds to' a MSL radiation monitor reading of about 5.7 R/hr. Scaling the results for 1.2 pCi/ml 1-131 dose equivalent:
lMSL Reading Corresponding to 1.2 pCi/ml 1-131 dose equivalent
((1.2 pCi/cc] / [34 gCicc)) X 5.7 R/hr = 0.2 R/hr = 200 mr/hr 200 mr/hr is at the lower end of the normal MSL monitor readings during full power.
Because this value is not distinguishable, and hydrogen water chemistry' system malfunctions that result in increased production of N-16 can also result in increased main steam line radiation levels, it is not appropriate at DAEC to use the main steam line monitor readings.
SU4

            ' ::            BASES'DOCUMENT.,
                    -i.'-.<jEAL                                            ' EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY                                 P Page 11 of 43

==REFERENCES:==
: 1. Abnormal Operating Procedure (AOP) 672.2, Offgas Radiation/Reactor Coolant High Activity
: 2. Technical Specification 3.4.6, Coolant Chemistry
: 3. Radiological Engineering Calculation No. 94-014A, Main Steam Line Radiation Monitor Setpoint Calculation, August 29, 1994
: 4. Surveillance Test Procedure (STP) No. 3.4.6-01, Reactor Coolant Gamma and Iodine Activity
: 5. Annunciator Response Procedure (ARP) 1C03A, Reactor and Containment Cooling and Isolation
: 6. Annunciator Response Procedure (ARP) 1C05B, Reactor'Control'-
: 7. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, January 2003 I I, . .,       ,-   ,,- .      *   -.    :
SU4

                    , .'.,' -EAL'BASES.DOCUMENTE^'r
                             ' '                                            EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 12 of 43 SU5 RCS Leakage EVENT TYPE: Coolant Leakage OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EAL THRESHOLD VALUE:
One of the following:
: 1. Unidentified or pressure boundary leakage ABOVE 10 gpm.
OR
: 2. Identified leakage ABOVE 25 gpm.
OR
: 3. Indication of Main Steamline Break.
DAEC EAL INFORMATION:
EAL Threshold Values I and 2 are precursorsof more serious RCS barrier challenges and are thus considered as a potential degradationof the level of safety of the plant.
Thus, it is possible to be operating within Technical Specification LCO Action Statement time limits and make a declarationof an Unusual Event in accordancewith these EALs.
Credit for the action statement time limit should only be given when leakage exceeds technical specification limits but has not yet exceeded the UnusualEvent EAL thresholds describedabove.
The DAEC Tech Spec Section 3.4.4 coolant system leakage LCO limits are: (1) < 5 gpm unidentified leakage, (2)
* 25 gpm total leakage averaged over the previous 24 hour period, and (3) < 2 gpm increase in unidentified leakage within the previous 24 hour period in Mode 1. Total leakage is defined as the sum of identified and unidentified leakage.
DAEC EAL Threshold Value 1 uses the generic value of 10 GPM for unidentified leakage or pressure boundary leakage. The 10 gpm value for the unidentified or pressure boundary leakage was selected as it is observable with normal control room indications.
DAEC EAL Threshold Value 2 uses identified leakage set at a higher value due to the SU5

:EAL BASES DOCUMENT- . ,.i:                           EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 13 of 43 lesser significance of identified leakage in comparison to unidentified or pressure boundary leakage.
DAEC EAL Threshold Value 3, indication of main steam line break, is discussed in NUMARC Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 2 Questions and Answers, June.1993, Fission Product Barrier-BWR section. This was in response to question 4 which states that the main steam line break with isolation can be classified under System Malfunctions.
The Main Steamline Break indication would be validated by ensuring the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results.
A stuck open or cycling SRV has been added here as discussed in NUMARC Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 2 Questions and Answers, June 1993, Fission Product Barrier-BWR section.
K If an SRV is stuck open or is cycling and no other emergency conditions exist, an emergency declaration may not be appropriate. RCS leakage inside the drywell excludes Safety-Relief Valve (SRV) discharge through the SRV discharge piping into the torus below the waterline. However, if the fuel is damaged and the SRV is allowing fission products to escape into primary containment, a loss of RCS should be determined as having occurred.
Abnormal operating procedures (AOPs) or technical specifications may require the plant to shut down when an SRV is stuck open or cycling. For cycling SRVs, technical specifications for suppression pool temperature and /or level, and SRV operability, may drive operators to shut down the plant. This is within the analyzed operating envelope of the plant and does not represent degradation in the level of safety. This also does not represent a significant precursor to further plant degradation.
SU5

:-EALI BASES DOCUMENTS               -            EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 14 of 43

==REFERENCES:==
: 1. Technical Specification 3.4.4, Coolant Leakage
: 2. Surveillance Test Procedure No. (STP) 3.0.0.0-01, Reactor Coolant System Leak Rate Calculation
: 3. Operating Instruction No. (01) 920, Drywell Sump System
: 4. Alarm Response Procedure (ARP) 1C04B, Reactor Water Cleanup and Recirculation
: 5. Alarm Response Procedure (ARP) IC04C, Reactor Water Cleanup and Recirculation
: 6. UFSAR Section 5.2.5, Detection of Leakage through Reactor Coolant Pressure Boundary
: 7. UFSAR Section 15.6.6, Loss-of-Coolant-Accident
: 8. NEI Methodology for Development of EmergencyAction Levels NUMARC/NESP-007 Revision 4, January 2003
: 9. NUMARC Methodology for Development of Emergency Action Levels NUMARCINESP-007 Revision 2, Questions and Answers, June 1993 Fission Product Barrier-BWR section SU5

   .               .it';,.:z1't,- AL'BASESAD CU   ENT ve    s,,              EBD-S; Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 15 of 43 SU6 Unplanned Loss of All Onsite or Offsite Communications Capabilities EVENT TYPE: InstrumentationlCommunication OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:
One of the following groups of communication losses:
: 1. Loss of ALL of the following onsite communication capabilities affecting the ability to perform routine operation:
* Plant Operations Radio System
* In-plant Telephones
* Plant Paging System OR
: 2. Loss of ALL of the following offsite communications capability:
* All telephone lines (commercial)
* Microwave Phone System
* FTS-2001 phone system (ENS & HPN)
* Cellular Phones
* Police Radio                            -
* Rad Survey Radio systems              '     I DAEC EAL INFORMATION:
There is no significant deviation from the generic EAL. The communications methods used at DAEC are described in the Emergency Plan. In-plant and external agency telephone communication methods include PABX lines, direct-ring lines, and NRC telephones which are extensions for the Emergency Notification System. There is also a microwave system to provide backup emergency telephone communications.
The availability of one method of ordinary offsite communication is sufficient to inform state and local authorities of plant problems. This EAL is intended to be used only when SU6

               -.: -I>-  Ei-LEA, BASES DOCUMENT".
a
                                                         ;:.                 EBD-S-.S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 16 of 43 extraordinary means (relaying of information from radio transmissions, individuals being sent to offsite locations, etc.) are being utilized to make communications possible.
The DAEC plant operations radio system is a UHF system with consoles located in the Control Room, Technical Support Center, Operational Support Center, and the Central Alarm Station. Hand-held transceivers are used in this system to provide simplex communications within the plant and onsite. The DAEC Radiological Survey Radio System is an 800 MHz trunked/conventional repeater system that provides base-to-portable communications throughout the DAEC EPZ. A secondary high-band system provides back-up capability for the 800 MHz radio. Consoles are located in the Technical Support Center and the Emergency Operations Facility at the Alliant Tower. The DAEC Security (backup radiological survey) Radio System provides base-to-portable security communication within the plant and with the Linn County Sheriffs Office using a mobile relay (repeater) type base station and two VHF frequencies. Control consoles are located in the Secondary Alarm Station, Central Alarm Station, Security Control Point, Technical Support Center, and Emergency Operations Facility. The DAEC also has a base station licensed for operation in the Police Radio Service on the law enforcement state-wide, point-to-point VHF frequency. The transmitter and one control console are located at the Secondary Alarm Station and in the Central Alarm Station. This station is for communications with Iowa Department of Public Safety radio station, Linn County Sheriffs office, and the Benton County Sheriffs office. This point-to-point channel is also used by the Linn County Emergency Management and other public-safety organizations throughout the state of Iowa.

==REFERENCES:==
: 1. Emergency Plan, Section F, Emergency Communications
: 2. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, January 2003 SU6

: .EAL.BASESDOCUMENT.                                EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 17 of 43 SU7 Unplanned Loss of Required DC Power During Cold Shutdown or Refuel Mode For Greater Than 15 Minutes EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Cold Shutdown, Refuel EAL THRESHOLD VALUE:
The following conditions exist:                  --
: 1. Unplanned Loss of Div 1 and Div 2 125 VDC busses based on bus voltage BELOW 105 VDC indicated.
AND.
: 2. Failure to restore power to at least one required 125 VDC bus within 15 minutes from time of loss.
DAEC EAL INFORMATION:
There is no significant deviation from the generic EAL. Unplanned loss of Div. I and Div.
: 11. 125 VDC busses exclude scheduled maintenance and testing activities. Under the conditions of concern, AOP 302.1, Loss of 125 VDC Power, would be entered. The DAEC EALs address the loss of both divisions of the 125 VDC systems consistent with AOP 302.1.
The 125 VDC system is divided into two independent divisions - Division I (1D1) and Division II (1D2) - each with separate AC and DC (battery) power supplies. Loss of both 125 VDC Divisions could compromise the ability to monitor and control the removal of decay heat during cold shutdown or refueling operations. These EALs are intended to be anticipatory in as much as the operating crew may not have necessary indication and control of equipment needed to respond to the loss. If this loss results inthe inability to maintain cold shutdown, the escalation to an Alert will be per SA3 "Reactor Coolant temperature to exceed Technical Specification limit of 212 F or UNCONTROLLED temperature rise approaching the Technical Specification limit of 212".
Bus voltage is based on the minimum bus voltage necessary for the operation of safety related equipment and the loss may be indicated by the illumination of annunciators "125 SU7

BASES DOCUMENT -EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 18 of 43 VDC System I Trouble" on IC08A, A-9 andlor "125 VDC System IITrouble" on 1C08B, A-4.

==REFERENCES:==
: 1. Abnormal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power
: 2. Abnormal Operating Procedure (AOP) 388, Loss of 250 VDC Power
: 3. Technical Specification 3.8, Electric Power Systems
: 4. UFSAR Section 8.3, Onsite Power Systems
: 5. UFSAR Table 8.3-6, Plant Battery System - DC Power, Instrumentation, and Control, Principle DC Loads (125V)
: 6. ARP 1C08A A-9
: 7. ARP 1C08B A-4
: 4. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, January 2003 SU7

                        .EA BASES&DOCUMENT                                  EBD-S Rev. 6 SYSTEM MALFUNCTION-CATEGORY                                .
Page 19 of 43 '
SAI Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses-During Cold Shutdown or Refuel Conditions EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Cold Shutdown, Refuel, Defueled EAL THRESHOLD VALUE:
The following conditions exist:
: 1. Loss of power to Startup (1X3) and Standby (1X4) transformers.
AND
: 2. Failure of A Diesel Generator (1G-31) and B Diesel Generator (1G-21) to supply power to emergency busses 1A3 and 1A4.
AND
: 3. Failure to restore power to at least one emergency bus, 1A3 or 1A4, within 15 minutes from the time of loss of both offsite and onsite AC power.
DAEC EAL INFORMATION:
Under the conditions of concern, entry into AOP 301.1, Station Blackout, would be made under Tab 1. Indications/alarms related to station blackout are displayed on control room panel IC08 and are listed in the procedure under "Probable Indications."
The loss of both offsite and onsite AC power to the emergency buses when in Cold Shutdown, Refuel or Defueled modes, compromises safety systems required for decay heat removal and is a substantial degradation of the level of safety of the plant. An ALERT is declared in Cold Shutdown and Refueling modes due to the less severe threat to the protection of the health and safety of the public because of the much longer time available to restore power and decay heat removal systems.
15 minutes was selected to exclude transient or momentary power losses.

==REFERENCES:==
: 1. Abnormal Operating Procedure (AOP) 301.1, Station Blackout
: 2. Abnormal Operating Procedure (AOP) 301, Loss of Essential Electrical Power SA1

EAL BASES DOC&MENT g;.                      EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 20 of 43
: 3. Technical Specifications Section 3.8, Electrical Power Systems
: 4. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, January 2003 SA1

                 '.E.
:i:'-ALBASES`DOCUMENT                 '                       EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY                                             ;
Page 21 of 43 .
SA2 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was Successful EVENT TYPE:. RPS Failure OPERATING MODE APPLICABILITY: Run, Startup EAL THRESHOLD VALUE:
The following conditions must exist to declare this EAL:
: 1. Auto Scram Failure AND
: 2. Operator actions to reduce power are SUCCESSFUL as indicated by either:
: a. ALL Rods Full-In, OR
: b. Reactor Shutdown Under All Conditions Without Boron, OR
: c. Reactor power below the APRM Downscale Alarm on ALL valid APRM instruments DAEC EAL INFORMATION:                -
The condition of concern is failure of the Reactor Protection System (RPS) to scram the reactor when a valid scram signal is present. This condition is more than a potential degradation of a safety system in that a front line automatic protection system did not function in response to a plant transient and thus plant safety has been compromised and design limits of the fuel may have been exceeded.
The EAL evaluation should occur after operators have taken actions'from the main control room to insert a manual scram and reduce reactor power. Permissible actions include all actions that can be performed quickly'from the main control room by on-shift operators (e.g., use of the Manual Scram pushbuttons, ARI, placing the Mode Switch in Shutdown, individual scram test switches, etc.). It is not appropriate to delay the EAL
                                -                                                        SA2

              -   f   .jEAL BASES DOCUMENT t                                    EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 22 of 43 evaluation until other time consuming actions are completed such as manual rod insertion or completion of in-plant EOP Support Procedures for rod insertion (e.g.,
venting the over-piston areas of individual CRDs).
Operator actions are considered successful if any of the following results are achieved:
* All control rods inserted to at least position 00 - this is defined in EOPs as the Maximum Subcritical Banked Withdrawal Position and is the lowest control rod position to which all control rods may be withdrawn in a bank and the reactor will none the less remain shutdown under all conditions, irrespective of reactor coolant temperature and any boron which may have been injected into the RPV.
* Determination that the Reactor is "Shutdown under ALL conditions without boron" -
this can be determined by relying on the Technical Specification demonstration of adequate shutdown margin:
   - One control rod is out beyond position 00 AND
   - All other control rods are at position 00 For other combinations of rod patterns and boron concentration, reactor engineering will need to perform a shutdown margin calculation.
* Reactor power is below the APRM Downscale Alarm Setpoint on ALL valid APRM instruments.
Note - If the mode switch is in Startup and the rods are fully inserted (i.e., the reactor is shutdown) prior to the automatic signal failure, then declaration of an Alert would not be required. In this case, the event would be reported under 10 CFR 50.72 (b) (2) (1)as a four hour report.

==REFERENCES:==
: 1. Integrated Plant Operating Instruction (IPOI) No. 5, Reactor Scram
: 2. ATWS Emergency Operating Procedure (EOP) - RPV Control
: 3. Emergency Operating Procedure (EOP) 1 - RPV Control: X
: 4. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, January 2003 SA2

EAL BASES DOCUMENT,.;;                                EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 23 of 43 SA3 Inability to Maintain Plant in Cold Shutdown EVENT TYPE: Inability to Reach or Maintain Shutdown Conditions OPERATING MODE APPLICABILITY: Cold Shutdown, Refuel EAL THRESHOLD VALUE:
: 1. Loss of Decay Heat Removal systems required to maintain Cold Shutdown.
AND
: 2. Temperature rise that exceeds 212 OF.
OR
: 3. Uncontrolled temperature rise approaching 212 IF.
DAEC EAL INFORMATION:
Under the conditions of concern for EAL Threshold Value 1, AOP 149, Loss of Decay Heat Removal, would be entered under Tab 1, Loss of Shutdown Cooling.
Indications/alarms related to loss of shutdown cooling are displayed on control room panels 1C03 and I C05 and are listed in the procedure under "Probable Indications." The procedure requires that shutdown cooling be re-established.
The procedure provides curves of maximum water heat up rates which provide an upper bound of the heatup until an estimated time to boil calculation can be completed by Engineering.
The DAEC EAL is written to imply an RCS temperature rise above 2120 F that is not allowed by plant procedures. This corresponds to the inability to maintain required temperature conditions for Cold Shutdown. "Uncontrolled" means that system temperature increase is not the result of planned actions by the plant staff. The wording is also intended to eliminate minor cooling interruptions occurring at the transition between Hot Shutdown and Cold Shutdown or temperature changes that are permitted to occur during establishment of alternate core cooling so that an unnecessary declaration of an Alert does not occur. The uncontrolled temperature rise is necessary to preserve SA3

EAL BASES DOCUMENT              E    I          EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 24 of 43 the anticipatory philosophy of NUREG-0654 for events starting from temperatures much lower than the cold shutdown temperature limit.

==REFERENCES:==
: 1. Abnormal Operating Procedure (AOP) 149, Loss of Decay Heat Removal
: 2. DAEC Technical Specifications
: 3. Surveillance Test Procedure (STP) 3.4.9-01, Heatup and Cooldown Rate Log
: 4. NUREG 1449, Shutdown and Low-Power Operation at CommercialNuclearPower Plants in the United States, September 1993
: 5. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, January 2003 SA3

                       ' ' 'EAL'BASESDOCUMENT              -'   .:               EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 25 of 43 SA4 Unplanned Loss of Most or All Safety System Annunciation or Indication in Control Room With Either (1) a Significant Transient in Progress, or (2) Compensatory Non-Alarming Indicators Unavailable EVENT TYPE: Instrumentation/Communication OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EAL THRESHOLD VALUE:
The following conditions exist:
: 1. Unplanned loss of most or all 1C03, 1C04 and 1CO5 annunciators or indicators associated with Critical Safety Functions for greater than 15 minutes.
AND i'~ 2. Either of the following conditions exist:         -
: a. A significant plant transient in progress.
OR
: b. Loss of compensatory non-alarming indications.
DAEC EAL INFORMATION:
Control room panels 1C03, 1C04, and 1C05 contain the annunciators associated with safety systems at DAEC. Therefore',the DAEC EAL addresses unplanned loss of annunciators on these panels. Compensatorynon-alarmiing indications includes the plant process computer, SPDS, plant recorders, or plant instrument displays in the control; .
room. Unplanned loss of annunciators or indicators excludes scheduled maintenance and testing activities. Significant transient includes response to automatic or manually initiated functions such as scrams, runbacks involving greater than 25% thermal power change, ECCS injections, or thermal power oscillations of 10% or greater.
Under the conditions of concem, entry into AOP 302.2, Loss of Alarm Panel Power, would be made. The procedure requires alerting operators on shift to the nature of the lost annunciation. It further requires that operators be attendant and responsive to abnormal indications that relate to those systems and components that have lost annunciation.
SA4

EAL BASES DOCUMENT.                  -       '      EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 26 of 43 Therefore, the generic criterion related to specific opinion of the Operations Shift Manager that additional operating personnel will be required to safely operate the unit is not included in the DAEC EAL because the concern is addressed by the AOP.
MOST - 75% of safety system annunciators or indicators are lost OR a significant risk that a degraded plant condition could go undetected exists. The use and definition of MOST is not intended to require a detailed count of lost annunciators or indicators but should be used as a guide to assess the ability to monitor the operation of the plant.
Unplanned loss of critical safety function indicators (i.e., EOP/EAL parameters) for greater than 15 minutes may preclude operators from taking actions to mitigate a transient.
Annunciators on 1C03, 1C04, and 1C05 share a common power supply from 125 VDC Division I that is fed through circuit breaker 1D1 3.
Indications of loss of annunciators associated with safety systems include:
* 125 VDC charger, battery, or system annunciators on control room panel 1C08
* Loss of usealed in" annunciators at affected panels
* Failure of affected annunciator panels shiftily testing by plant operators
* Expected alarms are not received
* Computer point ID B350 indicates "NSS ANN DC LOSS TRBL." (Loss of DC power to panels 1C03, 1C04, and IC05)

==REFERENCES:==
: 1. Operating Instruction (01) No. 317.2 Annunciator System
: 2. Abnormal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power
: 3. Abnormal Operating Procedure (AOP) 302.2, Loss of Alarm Panel Power
: 4. NEI Methodology for Development of EmergencyAction Levels NUMARC/NESP-007 Revision 4, January 2003 SA4

                           .-EAL-BASES DOCUMENT.'--                               EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 27 of 43 _
SA5 AC Power.Capability to Essential Busses Reduced to a Single Power Source for Greater Than 15 Minutes Such That Any Additional Single Failure Would Result in Station Blackout EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EAL THRESHOLD VALUE:
The following condition exists:
Only one AC power source remains available to supply Bus 1A3 or Bus 1A4 for greater than 15 minutes AND if it is lost, a Station Blackout will occur.
DAEC EAL INFORMATION:
K_> The DAEC EAL is written to address the underlying concern, i.e., only one AC power.
source remains and if it is lost, a Station Blackout will occur.. Under the conditions of concern, entry into AOP 301, Loss of Essential Electrical Power, would be made under Tab 1, Loss of One Essential 4160V Bus, and/or under Tab 3, Loss of Offsite Power.
Indications/alarms related to degraded AC power are displayed on control room panel 1C08 and are listed in AOP 301 under uProbable Indications."
At DAEC, the Essential Buses of concern are 4160V.Buses 1A3 and 1A4. Each of these buses feed their associated 480V and 120V AC busses through step down transformers.
Onsite power sources at DAEC include the A and B Diesel Generators, 1G-31 and 1G-21, respectively.

==REFERENCES:==
: 1. Abnormal Operating Procedure (AOP) 301, Loss of Essential Electrical Power
: 2. UFSAR Chapter 8 Electrical Power
: 3. Technical Specifications Section 3.8. Electrical Power Systems
: 4. NEI Methodology for Development of.EALs, NUMARC/NESP-007 Revisionr4, January 2003                                -
SA5

                        .KEALBASES DOCUMENT.:'-I                -          EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 28 of 43 SSI Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EAL THRESHOLD VALUE:
The following conditions exist
: 1. Loss of power to Startup (1X3) and Standby (1X4) transformers.
AND
: 2. Failure of A Diesel Generator (1G-31) and B Diesel Generator (IG-21) to supply power to emergency busses 1A3 and 1A4.
AND                                                                                  K>
: 3. Failure to restore power to at least one emergency bus within 15 minutes from the time of loss of both offsite and onsite AC power.
DAEC EAL INFORMATION:
There is no significant deviation from the generic EAL. In accordance with the generic guidance, DAEC is using a threshold of 15 minutes for Station Blackout to exclude transient or momentary power losses.
Under the conditions of concern, entry into AOP 301.1, Station Blackout, would be made under Tab 1. Indications/alarms related to station blackout are displayed on control room panel 1C08 and are listed in the procedure under "Probable Indications."

==REFERENCES:==
: 1. Abnormal Operating Procedure (AOP) 301.1, Station Blackout
: 2. Technical Specifications Section 3.8, Electrical Power Systems
: 3. UFSAR Chapter 8, Electric Power
: 4. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, January 2003 SS1

                   '''-'ELLBASES DOCUMENT               '..  :                 EBD-S Rev. 6
                 'SYSTEM MALFUNCTION CATEGORY Page 29 of 43 SS2 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was NOT Successful EVENT TYPE: RPS Failure OPERATING MODE APPLICABILITY: Run, Startup EAL THRESHOLD VALUE:
Failure of automatic scram and actions taken by operators in the Control Room to shut down the reactor OR reduce reactor power below the APRM downscales have been INEFFECTIVE.
The following conditions must exist to declare this EAL:
: 1. In ATWS EOP AND
: 2. Operator actions to reduce power are UNSUCCESSFUL as indicated by either:
: a. Reactor'power above the APRM Downscale Alarm on ANY valid APRM instrument, OR
: b. Boron Injection Initiation Temperature (BIIT) Curve,(EOP Graph 6) exceeded.
DAEC EAL INFORMATION:
This EAL addresses conditions where failure of an automatic scram has occurred and manual actions performed in the Control Room to reduce reactor power have been unsuccessful., -         '  -       '             -
Under the conditions of concern for this EAL, the reactor may be producing more heat than the maximum decay heat load for which safety systems are designed. A Site Area Emergency is warranted because conditions exist that may lead to the potential loss of the fuel cladding or primary containment. Aitholigh'this EAL may be viewed as redundant to the Fission Barrier Table, its inclusion is hecessary to better assure timely recognition and emergency response.'                 '          .
SS2

EALBASES;DOCUMENE-             .-t                   EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 30 of 43 The EAL evaluation should occur after operators have taken actions from the main control room to insert a manual scram and reduce reactor power. Permissible actions include all actions that can be performed quickly from the main control room by on-shift operators (e.g., use of the Manual Scram pushbuttons, ARI, placing the Mode Switch in Shutdown, individual scram test switches, etc.). It is not appropriate to delay the EAL evaluation until other time consuming actions are completed such as manual rod insertion or completion of in-plant EOP Support Procedures for rod insertion (e.g.,
venting the over-piston areas of individual CRDs).
The purpose of the ATWS EOP is to maintain adequate core cooling, shutdown the reactor and cooldown the RPV to cold shutdown conditions. The ATWS EOP is implemented when it cannot be determined that control rod insertion alone will assure that the reactor will remain shutdown under all conditions.
Reactor power above the APRM downscale setpoint is indicative of power generation above the decay heat levels which primary containment is designed to suppress.
Furthermore, if reactor power is above the APRM downscale setpoint, it is likely that the core bulk boiling boundary would be above that which provides suitable stability margin for operation at high powers and low flows.
Exceeding the Boron Injection Initiation Temperature (BIIT) limit (EOP Graph 6) is an indirect indication that the reactor is at power and that excessive decay heat is being added to the suppression pool.
The higher the reactor power level is, the more heat energy will be rejected to the torus thus requiring a lower torus temperature for initiation of boron injection if the Heat Capacity Limit is not to be exceeded before reactor shutdown is achieved.
As long as the core remains submerged (the preferred method of core cooling), fuel integrity and RPV integrity are not directly challenged even under failure-to-scram conditions. However, a scram failure coupled with an MSIV isolation results in rapid heatup of the torus due to the steam discharged from the RPV via SRVs. The challenge to the primary containment will thus become a limiting factor.

==REFERENCES:==
: 1. Integrated Plant Operating Instruction (IPOI) No. 5, Reactor Scram
: 2. ATWS Emergency Operating Procedure (EOP) - RPV Control
: 3. NEI Methodology for Development of EALs NUMARC/NESP-007 Rev 4, January 2003 SS2

EAL BASES DOCUMENT                   7-             EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 31 of 43 SS3 Loss of All Vital DC Power EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EAL THRESHOLD VALUE:             i         .      -
The following conditions exist:
: 1. Loss of Div 1 AND Div 2 125V DC busses based on bus voltage BELOW 105 VDC indicated.
AND
: 2. Failure to restore power to at least one required 125V DC bus within 15 minutes from time of loss.
DAEC EAL INFORMATION:
Under the conditions of concem, AOP 302.1, Loss of 125 VDC Power, would be entered under Tab 3, Complete Loss of 125 VDC. Consequently, the DAEC EAL addresses loss of both divisions of the 125V DC system consistent with AOP.
At DAEC, the 125V DC Systems ensure power is available for the reactor to be shutdown safely and maintained in a safe condition. The 125V System is divided into two independent divisions - Division I and Division II - with separate DC power supplies.
These power supplies consist of two separate 125V batteries and chargers serving systems such as RCIC, RHR, EDGs, and HPCI.
Complete loss of both 125V DC Divisions could compromise the ability to monitor and control the removal of decay heat during cold shutdown or refueling operations.
SS3

                      -. EAL BASES DOCUMENT":                         EBD-Rev. 6 .
SYSTEM MALFUNCTION CATEGORY                   l        Page

==REFERENCES:==
: 1. Abnormal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power
: 2. Abnormal Operating Procedure (AOP) 388, Loss of 250 VDC Power
: 3. Technical Specification 3.8, Electrical Power Systems
: 4. UFSAR Section 8.3, Onsite Power Systems
: 5. UFSAR Table 8.3-6, Plant Battery System - DC Power, Instrumentation, and Control, Principle DC Loads (125V)
: 6. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, January 2003 SS3

'*'''';        'W:.,          :-.;:> BASES   DOCUMENTED ......;-'::
i'EAL                                EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 33 of 43 SS4 Complete Loss of Heat Removal Capability EVENT TYPE: Inability to Reach or Maintain Shutdown Conditions OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown.
EAL THRESHOLD VALUE:
EOP Graph 4 Heat Capacity Limit is exceeded.
DAEC EAL INFORMATION:
This EAL addresses complete loss of functions, including ultimate heat sink and reactivity control, required for hot shutdown with the reactor at pressure and temperature. Under these conditions, there is an actual major failure of a system intended for protection of the public.
This EAL represents an escalation from the conditions of concern in SA3, Inability to Maintain Cold Shutdown, because the reactor is at operating pressure and temperature and decay heat levels are higher.
Per DAEC Technical Specifications, the following systems are necessary to achieve or maintain Hot Shutdown conditions:
* Reactor Protection System Instrumentation
* Core and Containment Cooling Systems Instrumentation
* Reactivity Control
* Standby Liquid Control System
* Core and Containment Cooling Systems
* Primary System Boundary
* Auxiliary Electrical Systems SS4
* s         EAL BASES, DOCUMENTED:                              EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 34 of 43 Loss of instrumentation is addressed by SS6,, Inability to Monitor a Significant Transient in Progress. SS1, Station Blackout, and SS3, Loss of all Vital DC Power address the Auxiliary Electrical System and the Core and Containment Cooling Systems; therefore they are not covered here. The Fission Barrier Table and SU5, RCS Leakage cover failure of the primary system boundary. Reactivity Control, including Standby Liquid Control System, is addressed in EALs SA2, SS2 and SG2.

==REFERENCES:==
: 1. Abnormal Operating Procedure (AOP) 149, Loss of Decay Heat Removal
: 2. Emergency Operating Procedure (EOP) 1 - RPV Control
: 3. ATWS Emergency Operating Procedure (EOP) - RPV Control
: 4. Emergency Operating Procedure ALC - Alternate Level Control
: 5. Emergency Operating Procedure (EOP) Basis, EOP Breakpoints
: 6. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, January 2003 SS4

                              '.EALBASES DOCUMENT!                           EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 35 of 43 SS5 Loss of Water Level in the Reactor Vessel That Has or Will Uncover Fuel in the Reactor Vessel EVENT TYPE: Inability to Reach or Maintain Shutdown Conditions OPERATING MODE APPLICABILITY: Cold Shutdown, Refuel EAL THRESHOLD VALUE:
The following conditions exist:
: 1. RPV level below 15 inches, indicating'that the core is or will be uncovered.
AND
: 2. Loss of all decay heat removal.
DAEC EAL INFORMATION:
The DAEC EAL is written in terms of the general concern that no cooling water source is lined up or available for injection into the RPV and water level is decreasing below the top of the active fuel (TAF). Under the conditions of concern for EAL Threshold Value 1,AOP 149, Loss of Decay'Heat Removal, would be entered under Tab 1, Loss of Shutdown Cooling. Indications/alarms related to loss of shutdown cooling are displayed on control room panels 1C03 and 1C05 and are listed in the procedure. Consistent with the value used in the EOPs, the EAL uses an indicated RPV level of 15 inches for the water level corresponding to TAF.
The conditions address concerns raised by the NRC AEOD Report AEOD/EGO9, "BWR Operating Experience Involving Inadvertent Draining of the Reactor Vessel", dated August 8,1986. This report states:
In broadest terms, the dominant cause of inadvertent reactor vessel draining are related to the operational and design problems associated with the residual heat removal system when it is entering into or exiting from the shutdown cooling mode. During this transitional period water is drawn from the reactor vessel, cooled by RHR heat exchangers (from the cooling provided by the service water system), and returned to the reactor vessel. First there are piping and valves in the residual heat removal SS5

EAL BASES DOCUMENT, -                               EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 36 of 43 system which are common to both the shutdown cooling mode and other modes of operation such as low pressure coolant injection and suppression pool cooling. These valves, when improperly positioned provide a drain path for the reactor coolant to flow from the reactor vessel to the suppression pool or the radwaste system. Second, establishing or exiting the shutdown cooling mode of operation is entirely manual making such evolutions vulnerable to personnel and procedural errors. Third, there is no comprehensive valve interlock arrangement for all the residual heat removal system valves that could be activated during shutdown cooling. Collectively, these factors have contributed to the repetitive occurrences of the operational events involving the inadvertent draining of the reactor vessel.

==REFERENCES:==
: 1. Abnormal Operating Procedure (AOP) 149, Loss of Decay Heat Removal
: 2. Emergency Operating Procedure (EOP)-1, RPV Control, Sheet 1 of 1
: 3. Emergency Operating Procedure (EOP) Basis, EOP Breakpoints
: 4. NRC AEOD Report AEOD/EGO9, UBWR Operating Experience Involving Inadvertent Draining of the Reactor Vessel", August 8, 1986
: 5. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, January 2003 SS5

                        --EAL BASES DOCUMENT                                   EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY -
Page 37 of 43 SS6 Inability to Monitor a Significant Transient in Progress EVENT TYPE: Instrumentation/Communication OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EAL THRESHOLD VALUE:
The following conditions exist:
: 1. Significant transient in progress and ALL of the following:
: a. Loss of most or all annunciators on Panels 1C03, 1C04 and 1C05 AND
: b. Loss of compensatory non-alarming indications.
AND
: c. Loss of indicators needed to monitor criticality, OR core heat removal, OR Fission Product Barrier status.
DAEC EAL INFORMATION:
The DAEC EAL is written in terms of a significant transient in progress with loss of both safety system annunciators and loss of compensatory non-alarming instrumentation. The DAEC EAL structure, which addresses all the key points in the generic EAL, better assures that the condition of concern for this EAL will be readily recognized.
Significant transient includes response to automatic or manually initiated functions such as scrams, runbacks involving greater than 25% thermal power change, ECCS injections, or undamped thermal power oscillations greater than 10%.
Compensatory non-alarming indications include the plant process computer, SPDS, plant recorders, or plant instrument displays in the control room. These indications are needed to monitor (site-specific) safety functions that are of concern in the generic EAL.
SS6

             -     -. EALBASES DOCUMENT.-:                                EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 38 of 43 Control room panels 1C03, 1C04, and 1C05 contain the annunciators associated with safety systems at DAEC. Annunciators on 1C03, 1C04, and 1C05 share a common power supply from 125 VDC Division I that is fed through circuit breaker 1D13.
MOST - 75% of safety system annunciators or indicators are lost OR a significant risk that a degraded plant condition could go undetected exists. The use and definition of MOST is not intended to required a detailed count of lost annunciators or indicators but should be used as a guide to assess the ability to monitor the operation of the plant.
Indications of loss of annunciators associated with safety systems include:
* 125 VDC charger, battery, or system annunciators on control room panel 1C08
* Loss of "sealed in" annunciators at affected panels
* Failure of affected annunciator panels shiftily testing by plant operators
* Expected alarms are not received
* Computer point ID B350 indicates "NSS ANN DC LOSS TRBL." (Loss of DC power to panels 1C03, 1C04, and 1C05)

==REFERENCES:==
: 1. Operating Instruction (01) No. 317.2, Annunciator System
: 2. Abnormal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power
: 3. Abnormal Operating Procedure (AOP) 302.2, Loss of Alarm Panel Power
: 4. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, January 2003 SS6

                ''':'.'; EALMBASES DOCUMENT!'..,                            EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY                         -
Page 39 of 43 SG1 Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EAL THRESHOLD VALUE:
The following conditions exist:
: 1. Loss of power to both Startup (1X3) and Standby (1X4) transformers.
AND
: 2. Failure of A Diesel Generator (1G-31) and B'Diesel Generator (1G-21) to supply power to emergency busses 1A3 and 1A4.
AND
: 3. ANYONE OF THE FOLLOWING:
: a. Restoration of power to either Bus'1A3 or 1A4 is not likely within 4 hours.
OR
: b. RPV level is indeterminate.
OR
: c. RPV level is below +15 inches.
SG1

              <-    2      AL.BASES DOCUMENi, ' .:>.-.          >      -iEBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY                                                l Page 40 of 43 DAEC EAL INFORMATION:
There is no significant deviation from the generic EAL. Under prolonged Station Blackout (SBO) conditions, fission product barrier monitoring capability may be degraded.
Although it may be difficult to predict when power can be restored, it is necessary to give the EC/OSM a reasonable idea of how quickly a General Emergency should be declared based on the following considerations:
* Are there any present indications that core cooling is already degraded to the point where a General Emergency is IMMINENT (i.e., loss of two barriers and a potential loss of the third barrier)?
* Ifthere are presently no indications of degraded core cooling, how likely is it that power can be restored prior to occurrence of a General Emergency?
The first part of this EAL corresponds to the threshold conditions for Initiating Condition SS1, Station Blackout - namely, entry into AOP 301.1, Station Blackout. The second part of the EAL addresses the conditions that will escalate the SBO to General Emergency.
Occurrence of any of the following is sufficient for escalation: (1) SBO coping capability exceeded, or (2) loss of drywell cooling that continues to make RPV water level measurements unreliable, or (3) indications of inadequate core cooling. Each of these conditions is discussed below:
: 1. SBO Coping Capability Exceeded DAEC has a SBO coping duration of four hours. The likelihood of restoring at least one emergency bus should be based on a realisticappraisalof the situation since a delay in an upgrade decision based on only a chance of mitigating the event could result in a loss of valuable time in preparingand implementing public protective actions.
: 2. RPV Water Level Measurements Remaining Unreliable Flashing of the reference leg water will result in erroneously high RPV water level readings giving a false indication of actual water inventory and potentially indicating adequate core cooling when it may not exist. EOP Graph 1, RPV Saturation Temperature, defines the conditions under which RPV level instrument leg boiling may occur.
SGI

                '--     EAL BASES DOCUMENT".               i :'* .          EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 41 of 43
: 3. Indications of Inadequate Core Cooling.
DAEC uses the RPV level that is used for the Fuel 'Clad "potential loss" condition in the Fission Product'Barrier Matrix. This is RPV level below +15 inches.

==REFERENCES:==
: 1. Abnormal Operating Procedure (AOP) 301.1, Station Blackout
: 2. Letter NG-92-0283, John F. Franz, Jr. to Dr. Thomas E. Murley, Response to Safety Evaluation by NRC-NRR "Station Blackout Evaluation Iowa Electric Light and Power Company Duane Arnold Energy Center," February 10, 1992
: 3. Emergency Operating Procedure (EOP)1 - RPV Control
: 4. Emergency Operating Procedure (EOP) ALC 'Alternate Level Control
: 5. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007, Revision 4, January 2003
                                      ,,  :1  - n    .
i, I   .1     .   . . . .     .
                                                                             . . I SG1

I
      .                 EAL BASES.DOCUMENT:S            -';                EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 42 of 43 SG2 Failure of the Reactor Protection System to Complete an Automatic Scram and Manual Scram was NOT Successful and There is Indication of an Extreme Challenge to the Ability to Cool the Core EVENT TYPE: RPS Failure OPERATING MODE APPLICABILITY: Run, Startup EAL THRESHOLD VALUE:
Failure of automatic and manual scrams AND conditions exist that no longer assure adequate core cooling or adequate decay heat removal.
The following conditions must exist to declare this EAL:
: 1. InATWS EOP AND
: 2. Loss of adequate core cooling or decay heat removal capability as indicated by either:
: a. RPV level cannot be maintained above -25 inches.
OR
: b. HCL Curve (EOP Graph 4) exceeded.
DAEC EAL INFORMATION:
This EAL addresses conditions where failure of an automatic scram has occurred and manual actions performed in the Control Room to reduce reactor power have been unsuccessful AND a subsequent loss of adequate core cooling or decay heat removal capability occurs. If either of these challenges exists during an ATWS, a core melt sequence exists. In this situation, core degradation can occur rapidly. For this reason, the General Emergency declaration is intended to be anticipatory of the fission product barrier matrix declaration to permit maximum offsite intervention time.
The purpose of the ATWS EOP is to maintain adequate core cooling, shutdown the reactor and cooldown the RPV to cold shutdown conditions. The ATWS EOP is SG2

                       - .:.EAL-BASES In'          DOCUMENT.'         'i                EBD-S Rev. 6 SYSTEM MALFUNCTION CATEGORY Page 43 of 43 implemented when it cannot be determined that control rod insertion alone will assure that the reactor will remain shutdown under all conditions.
If injection with all available Preferred and Alternate ATWS Injection Systems fails to provide sufficient injection to restore and maintain level above -25 inches (Minimum Steam Cooling RPV Water Level), adequate core cooling is threatened and submergence of the core is attempted by flooding the primary containment. This is accomplished by transfer to and implementation of the DAEC Severe Accident Guidelines (SAGs).
The Heat Capacity Limit (EOP Graph 4) is defined to be the highest torus temperature at which initiation of RPV depressurization will not result in exceeding the Primary Containment Pressure Limit (the PCPL is 53 psig at the DAEC) before the rate of energy transfer from the RPV to the primary containment is within the capacity of the containment vent.
Control of torus temperature relative to the Heat Capacity Limit is directed in the Primary Containment Control Guideline, EOP 2. If the actions being taken in EOP 2 to preserve torus heat capacity are inadequate or not effective, RPV pressure must be reduced in order to remain below the Heat Capacity Limit. Therefore, actions in the RPV pressure control section of the ATWS EOP must accommodate these requirements. Failure to do so may lead to failure of the containment or loss of equipment necessary for the safe shutdown of the plant.

==REFERENCES:==
: 1. Emergency Operating Procedure ATWS EOP - RPV Control
: 2. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, January 2003 SG2

Comrititted to Nuclear Excefiene DAEC EMERGENCY PLANNING DEPARTMENT PROCEDURE TRANSMITTAL ACKNOWLEDGEMENT MEMO (TAM-1 17)
To:             NRC-NRR Document Control Desk US NRC Washington DC 20555 Re:             Entire EPIP Document (Copy 28)
PSM

==Title:==
n/a Distribution Date:              02 / 04 / 2004 Effective Date of Change:       02 /05 /2004 Return by:                      02 /19 /2004 Please perform the following to your assigned manual. If you have any questions regarding this TAM please contact Don A. Johnson at 319-851-7872.
REMOVE              INSERT EPIP Table of Contents Revision                                  Rev. 149            Rev. 150 EPIP 3.1 (PWR: 23538)                                            Rev. 13             Rev. 14 EPIP 4.3 (PWR: 24115)                                            Rev. 11             Rev. 12 EPIP EAL-01 (PWR: 24091)                                         Rev. 5              Rev. 6 EPIP EAL-02 (PWR: 24097)                                         Rev. 4              Rev. 5 EPIP EAL-03 (PWR: 24093)                                         Rev. 5              Rev. 6 PERFORMED BY:
Print Name                                   Sign Name                          Date Please return to:     K. Dunlap PSC/Emergency Planning 3313 DAEC Rd.
Palo, IA 52324 To be completed by DAEC EP personnel only:
Date TAM returned:
EPTools updated:                                                                             Page 1 of 2

Comnmitted to Nuclear Excellenf              _
DAEC EMERGENCY PLANNING DEPARTMENT PROCEDURE TRANSMITTAL ACKNOWLEDGEMENT MEMO (TAM-117)
To:             NRC-NRR Document Control Desk US NRC Washington DC 20555 Re:             Entire EPIP Document (Copy 28)
PSM

==Title:==
n/a Distribution Date:              02 /04 /2004 Effective Date of Change:       02/05/2004 Return by:                      02 /19 / 2004 Please perform the following to your assigned manual. If you have any questions regarding this TAM.
please contact Don A. Johnson at 319-851-7872.
REMOVE              INSERT EPIP EAL-04 (PWR: 24095)                                         Rev. 4              Rev. 5 -
EPIP OSC-09 (PWR: 24027)                                         Rev. 0              Rev. 1 EPIP OSC-1 1 (PWR: 24089)                                        Rev. 0              Rev. 1 EPIP TSC-03 (PWR: 24021)                                         Rev. 2              Rev. 3 EPIP TSC-41 (PWR: 24099)                                         Rev. 0              Rev. 0 Remove this form.
PERFORMED BY:
Print Name                                   Sign Name                         Date Please return to:     K. Dunlap PSC/Emergency Planning 3313 DAEC Rd.
Palo, IA 52324 To be completed by DAEC EP personnel only:
Date TAM returned:
EPTools updated:                                                                             Page 2 of 2

i             NMC
  ~Committedto Nuclearfxellen DuaneManagement Operated by Nuclear   Arnold Energy      Center Company,  LLC Friday, February 6, 2004 NRC-NRR Document Control Desk US NRC Washington, DC 20555 To: NRC-NRR Document Control Desk From: DAEC Emergencv Planninq Department Re: (TAM-1 17) Description of changes to the following documents EPIP        3.1          Inplant Radiological Monitoring Eliminate reference to the Post Accident Sampling system and any procedures related to it.
Page 16, Add: Attachment 2 - LOCATIONS FOR POST ACCIDENT SAMPLING Per TS Amendment #252, PASS Elimination, this is one of the procedures to be changed to reflect the change in requirement to perform a PASS sample or have and maintain the system. PASS will be 'Abandoned In Place'.
EPIP        4.3          Rescue and Emergency Repair Work Change page 9 reference from PASS team to post accident sampling team.
EPIP        EAL-01       Abnormal Rad Levels/Radioactive Effluent EAL Table For EAL AA2: Added the ICfor ARM Hi Rad Alarm back to this EAL to be consistent with the initial NRR approved EAL and NEI 99-01 rev 4.
EPIP        EAL-02       Fission Barrier EAL Table For the RPV Level row: Removed the statements "that cannot be restored" and "no injection source available" from this row due to potential underclassification issues, and due to the NRR asking that these statements not be used.
EPIP        EAL-03       Hazards and Other Conditions Affecting Plant Safety EAL Table For EAL HU1: Retumed the ICfor earthquakes back to the initial NRR approved wording to remove any potential overclassification issues.
EPIP        EAL-04       System Malfunction EAL Table For EAL SU1: Removed the word "Unplanned" as it was not consistent with the initial NRR approved wording or NEI 99-01 rev 4.
For EAL SU6: Removed reference to 'Sound power/ed telephones" as this system is not a viable option for routine communications at DAEC.                      /
For EAL SA3: Retumed the wording to that of the initial NRR approved wording to ensure consistency and remove any potential underclassification issues.
EPIP        OSC-09 HP Supervisor Checklist Page 2 of 2, Item 14: Delete 'personnel to perform PASS sampling, and the dispatch of. No one will be dispatched to perform PASS sampling.
Per TS Amendment #252 - PASS Elimination, this is one of the procedures to be changed to reflect the elimination of the requirement to perform a PASS sample or have and maintain the system.
PASS will be 'Abandoned In Place'.
EPIP        OSC-11       Emergency Assignment Staffing Board 3313 DAEC Road - Palo, Iowa 52324-9785 Telephone: 319.851.7191 - Fax: 319.851.7364

Duane Arnold Energy Center Committed to Nuclear        eltene                             Operated by Nuclear Management Company, LLC EPIP        TSC-03       Site Radiation Protection Coordinator Checklist PASS SAMPLE section:
CHANGE - 'PASS SAMPLE' to 'POST ACCIDENT SAMPLING' DELETE - Large Volume and Small Volume ADD - Reactor Coolant and Suppression Pool This revision changes the reference from PASS sample system specifics to post accident sampling points without using PASS. These sample points are described in Chemistry procedures, PASAP 2.6, 2.7 and 2.8.
Per TS Amendment #252 - PASS Elimination, this is one of the procedures to be changed to reflect the elimination of the requirement to perform a PASS sample or have and maintain the system. PASS will be
     'Abandoned In Place'.
EPIP        TSC-41       PASS Capabilities Eliminate form. The information on this form is valuable, but EPIP 3.1 no longer directs anyone to this procedure.
Add to EPIP 3.1, IN-PLANT RADIOLOGICAL MONITORING as an Attachment.
Please contact Paul Sullivan, Manager of Emergency Preparedness at DAEC, (319)851-7191, if you require further information.
Sincerely,
/"-          W lis->                nDate:  X-/C, 4&deg;o 3313 DAEC Road - Palo, Iowa 52324-9785 Telephone: 319.851.7191 - Fax: 319.851.7364

,'EMERGEN, CYPLAN IMPLEMENTING PROCEDURES'             Rev.150 YEEGEC-                                             Rev 14 INDEX                         PAGE 1 of 7 Procedure                         Title                      Revision    Date Number 1.1   Determination of Emergency Action Levels            22     11/21/03 1.2   Notification                                        29     10/13/03 1.3   Plant Assembly and Site Evacuation                    9    09/12/01 1.4   Release of Emergency-Related Information              5    10/13/03 1.5   Activation and Operation of the EOF                   5    10/13/03 2.1    Activation and Operation of the OSC                  14    10/13/03 2.2    Activation and Operation of the TSC                 24     10/13/03 2.3    Operation of the FTS-2001 Phone Network               7    10/13/03 2.4    Activation and Operation of the ORAA                  8    09/12/01 2.5    Control Room Emergency Response Operation            15    10/13/03 2.6    Activation and Operation of the ORAL                 10    02105/04 2.7    Activation and Operation of the ODEF                  6    10/15/01 2.8    Security Threat                                      2      7/30/03 3.1    In-Plant Radiological Monitoring                     14      2/5/04 I 3.2    Field Radiological Monitoring                        14    10/13/03 3.3    Dose Assessment and Protective Action               21      1/21/04 4.2    First Aid, Decontamination and Medical Support       7      8/23/02 4.3    Rescue and Emergency Repair Work                     12      2/5/04 4.5    Administration of Potassium Iodide (KI)              7      2110/03 5.2    Recovery and Re-entry                                10    10/13/03

-4EMERGENCY PLAN1PLEMENTING             ROC DU E INDEX                              PAGE 2 of 7 Form                       Title                           Revision  Referencing Number                                                      Number     Procedure CR-01    OSM/OSS Checklist                                    Rev. 4     EPIP 2.5 CR-03    Dose Projection & ARM Data Sheet                     Rev. 0     EPIP 2.5 CR-04    Control Room to TSC Transfer Checklist               Rev. 2     EPIP 2.5 EAL-01   Abnormal Rad Levels/Radioactive Effluent Table       Rev. 6     EPIP 1.1 EAL-02   Fission Barrier Table                                Rev. 5     EPIP 1.1 EAL-03   Hazards & Other Conditions Affecting Plant Safety    Rev. 6     EPIP 1.1 EAL-04   System Malfunction Table                             Rev. 5     EPIP 1.1 EAL-O5   ISFSI Table                                          Rev. 0     EPIP 1.1 EOF - 02 NRC - HPN Communicator Checklist                     Rev. 3     EPIP 1.5 EOF - 03 Technical Recorder Checklist                         Rev. 2     EPIP 1.5 EOF - 04 Summary of Computer Data Backup Collection           Rev. 2     EPIP 1.5 Activities EOF - 05 EOF Information Services Representative Checklist    Rev. 3     EPIP 1.5 EOF - 06 DAEC Key Parameter Log                               Rev. 0     EPIP 1.5 EOF - 07 Emergency Response and Recovery Director             Rev. 5     EPIP 1.5 Checklist EOF - 08 Rad & EOF Manager Checklist                          Rev. 7   EPIP 1.5,3.3 EOF-09   EOF OPS Liaison Checklist                            Rev. 1     EPIP 1.5 EOF -10  EOF-Communicator Checklist                           Rev. 3     EPIP 1.5 EOF -11  Support Services Coordinator Checklist               Rev. 3     EPIP 1.5 EOF - 12 Field Team Director Checklist                        Rev. 1   EPIP 1.5, 3.3

      . MREC>>lMPLEMENTINGPROCEDURs                     Rev. 150 INDEX                          PAGE 3 of 7 EOF - 14  EOF MIDAS Operator Checklist                       Rev. I EPIP 1.5, 3.3 EOF - 15  Radiological Data Plotter Checklist                Rev. 0 EPIP 1.5,3.3 EOF - 16  Radiological Assessment Coordinator Checklist      Rev. 2 EPIP 1.5,3.3 EOF - 17  EOF Security Access Clerk Checklist                Rev. 3   EPIP 1.5 EOF - 18  EOF Staffing Accountability Roster                 Rev. 4   EPIP 1.5 EOF - 19  Drill Announcement Message                         Rev. 0 EPIP 1.4, 1.5 EOF - 20  Emergency Announcement Message                     Rev. I EPIP 1.4, 1.5 EOF -21   Personnel Access Log                               Rev. 1 EPIP 1.4, 1.5 EOF -22   Registration Form                                  Rev. 0 EPIP 14, 1.5 EOF -23   Security Post Log                                  Rev. 2 EPIP 1.4, 1.5 EOF -24   First Floor Security Post Description              Rev. 2 EPIP 1.4, 1.5 EOF - 25  Fourteenth Floor Security Post Description        Rev. 11   EPIP 1.5 EOF - 27  Status Update Message - EOF Communicator           Rev. 0  EPIP 1.5 EOF -28   Verbal Closeout Summary                            Rev. 0  EPIP 1.5 EOP -29   Written Closeout Summary                           Rev. 0  EPIP 1.5 EOF - 30  Status Board                                       Rev. 1  EPIP 1.5 EOF - 31  Access B adge Example                             Rev. 0   EPIP 1.5 EOF -32   EOF Staff Response                                Rev. 4   EPIP.15 EOF - 33  Recovery Issues                                   Rev. 0   EPIP 5.2 EOF - 34  EOF Activities                                    Rev. 0   EPIP 5.2 EOF - 35  Recovery Phase Plan Outline Guidance              Rev. 0   EPIP .52 EOF - 36  RE-Entry Briefing Guide                           Rev. 0   EPIP 5.2 EOF - 37  RE-Entry Debriefing Guide                         Rev. 0   EPIP 5.2 EOF-38    EOF Messenger Checklist                           Rev. 1   EPIP 1.5 ERO - 01  ERO Position Equivalency Table                    Rev. I   EPIP 1.5 JPIC - 01 JPIC Manager Checklist                             Rev. 4   EPIP 1.4 JPIC - 03 Alliant Spokesperson Checklist                     Rev. 3  EPIP 1.4

kEMEiRGENCY PLAN IMPLEMENTING'PROCEDURES                    Rv15 INDEX                              PAGE 4 of 7 JPIC - 04  Technical Liaison Checklist                          Rev. 4   EPIP 1.4 JPIC - 05  Sequence of Events                                   Rev. 0   EPIP 1.4 JPIC - 06  Public Information Officer Support Checklist         Rev. 5   EPIP 1.4 JPIC - 07  Logistics Coordinator Checklist                      Rev. 4   EPIP 1.4 JPIC - 08  Logistics Support Checklist                          Rev. 4   EPIP 1.4 JPIC -09   Audiovisual Support Checklist                        Rev. 4   EPIP 1.4 JPIC - 11  Rumor Control Coordinator I Checklist                Rev. 3   EPIP 1.4 JPIC - 12  Rumor Control Event Summary Log                      Rev. I   EPIP 1.4 JPIC - 13  Rumor Control Coordinator II Checklist               Rev. 2   EPIP 1.4 JPIC -14   Public Rumor Control Checklist                       Rev. 2   EPIP 1.4 JPIC -  15 News Media Rumor Control Checklist                   Rev. 4   EPIP 1.4 JPIC -16   Assistant JPIC Manager Checklist                     Rev. 3   EPIP 1.4 JPIC -17   JPIC Security Access Control Checklist               Rev. 3   EPIP 1.4 JPIC -18   Sixth Floor Security Post Description                Rev. 2   EPIP 1.4 JPIC - 19  JPIC Distribution List                               Rev. 2   EPIP 1.4 NOTE-01    ERO Notification-Phone System Callout                Rev. 5   EPIP 1.2 NOTE-02    ERO Notification - Alphanumeric Paging System        Rev. 3   EPIP 1.2 Callout NOTE-03    Event Notification Worksheet                         Rev. 1   EPIP 1.2 NOTE-04    Plant Assembly Notification                          Rev. 2   EPIP 1.2 NOTE-05    Emergency Action Level Notification                  Rev. 7   EPIP 1.2 NOTE-06    Plant Page for Emergency Classification Changes      Rev. I   EPIP 1.2 NOTE-07    Basic Notification Flowpath                          Rev. 2   EPIP 1.2 ODEF-01    ODEF Decontamination Waiting Area                    Rev. 0   EPIP 2.7 ODEF-02    Floor Plan for ORAL/ODEF                             Rev. 0   EPIP 2.7 ODEF-03    Travel Route to ORAIJODEF                            Rev. 0   EPIP 2.7 ODEF-04    12'h Avenue Entrance to ORAL/ODEF                    Rev. 0   EPIP 2.7 ORAA-01    Offsite Relocation and Assembly Area Supervisor's    Rev. 2   EPIP 2.4

     .EEGNYP      NIMPALEMENTINGPROCEDURES'.Rev. 150 INDEX                              PAGE 5 of 7 Checklist ORAA-02  Health Physics Support for the Offsite Relocation    Rev. 1   EPIP 2.4 and Assembly Area ORAA-03  Security Support for the Offsite Relocation and      Rev. I   EPIP 2.4 Assembly Area ORAA-04  Offsite Relocation and Assembly Area                 Rev. 0   EPIP 2.4 ORAA-05  Offsite Relocation and Assembly Area Parking and     Rev. 0   EPIP 2.4 Vehicle Monitoring OSC-01   OSC Layout                                           Rev. 0   EPIP 2.1 OSC-03   Minimum Staffing Level                               Rev. 2   EPIP 2.1 OSC-04   Recommended Log Entry Topics                         Rev. 0   EPIP 2.1 OSC-05   Emergency Event Log Sheet                            Rev. 0   EPIP 2.1 OSC-06   Personal Statement Concerning Incident               Rev. 0   EPIP 2.1 OSC-07   Emergency Exposure Tracking Log                      Rev. 0   EPIP 2.1 OSC-08   OSC Supervisor Checklist                             Rev. 0   EPIP 2.1 OSC-09   Health Physics Supervisor Checklist                  Rev. I   EPIP 2.1 I OSC-10   IC/EM & Mechanical Supervisor Checklist              Rev. 1   EPIP 2.1 OSC-11   Emergency Assignment Staffing Board Duties           Rev. 1   EPIP 2.1 I OSC-12   External Exposure Limits                             Rev. 0   EPIP 4.3 OSC-13    Guidance on Dose Limits for Workers Performing      Rev. 0   EPIP 4.3 Emergency Services OSC-14    Guidelines Regarding Selection of Volunteers        Rev. 0   EPIP 4.3 OSC-15    OSC Repair Team Work Order                          Rev. 0   EPIP 4.3 OSC-16    Repair Team Datasheet Flowpath                      Rev. 0   EPIP 4.3 OSC-17    Field Monitoring Team Checklist                     Rev. 0   EPIP 2.1 PAR -01   PAR Decision Making - Recommendations               Rev. 0   EPIP 3.3 PAR- 02   PAR Decision Making - Flowchart                      Rev. 0  EPIP 3.3 PASE-02   Onsite Assembly Locations                            Rev. 2  EPIP 1.3 PASE-05   Site Evacuation Routes                               Rev. I  EPIP 1.3

     ' 'EMER, GENCYSPLAN IMPLLEMENTINoG PROCEDURS ;, Rev. 150 K>                              INDEX                            PAGE 6 of 7 SAM-01      EOP-SAG Transition Checklist                       Rev. 0   EPIP 2.2 TSC-01      Emergency Coordinator Checklist                    Rev. 4   EPIP 2.2 TSC-02      TSC-EOF Transfer Checklist                         Rev. 3   EPIP 2.2 TSC-03      Site Radiation Protection Coordinator Checklist    Rev. 3   EPIP 2.2 I
TSC-04      Technical & Engineering Supervisor Checklist       Rev. 4   EPIP 2.2 TSC-05      Quality Assurance Checklist                        Rev. I   EPIP 2.2 TSC-06      Security & Support Supervisor Checklist            Rev. 2   EPIP 2.2 TSC-07      Administrative Supervisor Checklist                Rev. 2   EPIP 2.2 TSC-08      Material Management Supervisor Checklist           Rev. 2   EPIP 2.2 TSC-09      TSC Communicator Checklist                         Rev. 3   EPIP 2.2 TSC-10      CR-Communicator Checklist                          Rev. 4   EPIP 2.2 TSC-12      ENS Communicator Checklist                         Rev. 2   EPIP 2.2 TSC-13      HPN Communicator Checklist                         Rev. 1   EPIP 2.2 TSC-14      TSC/OSC Operations Liaison Checklist               Rev. 2   EPIP 2.2 TSC-15      Radiological Support Staff Checklist               Rev. 2   EPIP 2.2 TSC-18      TSC MIDAS Operator Checklist                       Rev. 0   EPIP 2.2 TSC-19      Technical & Analysis Engineer Checklist            Rev. 2   EPIP 2.2 TSC-20      TSC Operations Supervisor                          Rev. 3   EPIP 2.2 TSC-21      Electrical Engineer Checklist                      Rev. 1   EPIP 2.2 TSC-23      Mechanical Engineer Checklist                      Rev. 0   EPIP 2.2 TSC-24      Reactor Engineer Checklist                         Rev. 2   EPIP 2.2 TSC-26      Information Services Representative Checklist      Rev. 2   EPIP 2.2 TSC-27      Fire Marshall Checklist                            Rev. I   EPIP 2.2 TSC-28      NRC Roles During A Nuclear Power Plant             Rev. 0   EPIP 2.2 Emergency Checklist TSC-29      TSC Minimum Staffing Level                         Rev. 3   EPIP 2.2 TSC-30      Emergency Action Request Log                       Rev. 0   EPIP 2.2 K>,
TSC-3 1     Radio Operator Log                                 Rev. I   EPIP 2.2

~-.EMERGENCY8P3LAN:IMPL:EMENTIN>G.PROCEDURES.        e  5 INDEX                       PAGE 7 of 7 TSC-33    Typical Organization of the NRC Site Team    Rev. 0    EPIP 2.2 TSC-34    TSC Organization Chart                       Rev. 4    EPIP 2.2 TSC-35    Assignment Form                              Rev. 0    EPIP 5.2 TSC-36    Deactivation Report                          Rev. 0    EPIP 5.2 TSC-37    Plant Operations Status                      Rev. 0    EPIP 5.2 TSC-38    TSC/Control Room/OSC Activities              Rev. 0    EPIP 5.2 TSC-39    TSC Clerical Checklist                       Rev. 0    EPIP 2.2 TSC-40    ARM Locations                                Rev. 0   EPIP 3.1/4.3 TSC-41    Deleted I
TSC-42    On-Site Map                                  Rev. 0    EPIP 3.2 TSC-43    ESB Logon Instructions (TSC/CR/EOF)          Rev. 0    EPIP 2.2

S
 >.'.-....'EMERGENCY PLAN IMPLEMENTING1PROCEDURE                 '.      EPIP 3.1 Rev. 14 IN-PLANT RADIOLOGICAL-MONITORING Page 1 of 17 Effective Date:     /1d
                  -e ,    .    ,        #      L REVIEW-                     ..-     ..
Prepared by:              ,1.Dae:'                                  /Z-      0g Reviewed by:                                                Date:        o I dependent Reviewer PROCEDURE APPROVAL I am responsible for the technical content of this procedure.
Approved by:                       c                        Date:
Manager, Emergency Planning

   - EMERGENCY PLAN IMPLEMENTING PROCEDURES i"                                                                  EPIP 3.1 Rev. 14 IN-PLANT RADIOLOGICAL MONITORING Page 2 of 17 Table of Contents.
Page 1.0   PURPOSE ................................................                                                                       3 2.0   DEFINITIONS ................................................                                                                   3 3.0   INSTRUCTIONS .................................................                                                                 3 3.1    RESPONSIBLITIES ................................................ 3 3.2    DETERMINATION OF ASSEMBLY AREA HABITABILITY ..............                                                              5 3.3    IN-PLANT RADIOLOGICAL MONITORING ...................................... 10 3.4    POST ACCIDENT SAMPLING ..........................                                                ...................... 11 4.0   RECORDS .................................................                                                                      13

==5.0   REFERENCES==
 ................................................                                                                    13 6.0   ATTACHMENTS                                              ................................................                      14 ATTACHMENT 1 HP: E-Plan Habitability Data Sheet ...................... 15 ATTACHMENT 2: LOCATIONS FOR POST ACCIDENT SAMPLING 16

EMERGENCY PLAN IMPLEMENTINGPROCEDURE               no,.
Rev. 14 IN-PLANT RADIOLOGICAL'MONITORING Page 3 of 17 1.0   PURPOSE (1)   This procedure provides instructions for performing radiological monitoring activities during an emergency including determination of habitable areas, In-Plant radiological monitoring,'and post accident sampling. This procedure supports the implementation of requirements in 10CFR50, Appendix A, NUREG 0654 and the DAEC Plan.
(2)   This procedure is applicable to the Site Radiation Protection Coordinator, Chemistry Technicians, the OSC Health Physics Supervisor, Health Physics Technicians and other In-Plant Monitoring Team personnel involved in in-Plant radiological monitoring activities.: --.
(3)   This procedure should be implemented for events classified as an ALERT or' greater, and may be implemented upon activation of the Operational Support Center (OSC) for an event classified as a NOTIFICATION OF UNUSUAL EVENT if necessary.
2.0   DEFINITIONS (1)   None
                                   -3.0 ''INSTRUCTIONS 3.1     RESPONSIBILITIES                                                  -
3.1.1 SITE RADIATION PROTECTION COORDINATOR (1) Ensure that DAEC personnel are dispatched to monitor the environs in and around the plant for radiological consequences associated with the event.
(2)   Conduct an initial evaluation and assessment of the results of radiological monitoring activities. Upon activation of the Emergency Operations Facility (EOF), this function will be assumed by the Rad & EOF Manager for all Off-Site monitoring activities.

      -EMERGENCY PLAN IMPLEMENJTING PROCEDURE                         EPIP 3.1 Rev. 14 IN-PLANT RADIOLOGICAL MONITORING Page 4 of 17 (3)   Assess the On-Site radiological consequences and directing protective measures, including evacuation of the plant.
(4)   During the initial stages of the event, appraise local and State authorities, through the Emergency Coordinator, of the results of radiological monitoring activities and providing protective action recommendations based upon the projected radiological consequences to the population-at-risk. Upon activation of the EOF, this function will be assumed by the Rad & EOF Manager.
3.1.2 OSC HEALTH PHYSICS SUPERVISOR (1)    Ensure that personnel dispatched from the OSC are properly outfitted with protective clothing and equipment, are briefed regarding ALARA and are apprised of existing and potential radiological hazards.
(2)  Coordinate with the Site Radiation Protection Coordinator to obtain information regarding plant status, problems, response options, significant radiological releases in progress, offsite dose rates, plume location and meteorological conditions as necessary.
(3)  Ensure the determination of habitability of assembly areas and ERO facilities.
(4)   Coordinate the dispatch of monitoring teams, post accident sampling teams and Offsite Relocation Assembly Area (ORAA) and Offsite Radiological Analysis Laboratory/Offsite Decontamination Facility (ORAUODEF) personnel when necessary. Ensure the continued briefing of the ORAA in terms of plant and radiological conditions.
(5)   Supervise efforts to prepare injured/contaminated personnel for transport to offsite medical facilities.
(6)   Brief rescue and emergency repair team personnel regarding radiological hazards which exist, or which potentially may be encountered, and provide guidance regarding precautions to be taken and limits that shall not be exceeded.

 -,.:,,:EMERGENCYrPLAN IMPLEMENTING PROCEDURE..,,..,                           EPIP 3.1 Rev. 14 IN-PLANT RADIOLOGICALWMONITORING Page 5 of 17 3.2      DETERMINATION OF ASSEMBLY AREA HABITABILITY (1)   The assembly areas of concem'rat the DAEC include the following:
(a)    Control Room (b)    TSC (c)    OSC, which includes the OSC Break Room, Security Building, the Health Physics Access Control Area, and interconnecting passageways.
(d)    Hot Lab (e)    Warehouse (f)    Badging Center
          '(g)    Training Center          '
(h)    Plant Support Center (PSC).
(2)    The OSC HP Supervisor should'determine the habitability of the PSC,'Badging Center and Training Center through the use of an On-Site Radiological monitoring team or, if unavailable, use additional H.P. personnel to perform this function.
(3)    Initial determinations of habitability' for the Control Room, TSC and Hot Lab should be accomplished by ensuring that the radiation levels being detected by the permanently installed radiati6n monitoring (Area Radiation Monitoring) system are normal.
:' ,.7, -   : ',..- B , - , . . ' . , - ,. Za

EMERGENCYPLANWiMPLEMENTING PROCEDURE                             EPIP 3.1 Rev. 14 IN-PLANT RADIOLOGICAL MONITORING Page 6 of 17 (a)    When the Plant Process Computer is operational, this data is available to the TSC and EOF. The CR Communicator need only to forward the Main Steamline Rad Monitor data to the Site Radiation Protection Coordinator as these do not have computer points associated with them NOTE EPIP Form TSC-40, "ARM Locations",
provides a listing of the installed Area Radiation Monitors throughout the plant, including their locations and ranges.
(b)    As specified in EPIP 2.2, "Activation and Operation of the TSC", the three TSC radiation monitors (two monitoring the ventilation intake duct work and one in the TSC general area) shall be monitored upon initially activating the TSC, by the Radiation Support Staff. This information should be forwarded to the Site Radiation Protection Coordinator.
(4)   If radiation levels being monitored in the above areas are greater than normal, the Site Radiation Protection Coordinator should be advised and protective measures instituted.
NOTE If the monitors are alarming, the Site Radiation Protection Coordinator may recommend evacuation of such areas to the Emergency Coordinator.
If radiation levels are higher than normal, priority should be given to radiation survey and sampling activities in those areas to determine their acceptability for continued access.

EMERGENCY PLAN IMPLEMENTING PROCEDURE/
* EPIP 3.1
                .                     ...       .            ... Rev. 14 IN-PLANT RADIOLOGICAL MONITORING                         R Page 7 of 17 (5) The OSC HP Supervisor shall be responsible to initiate radiation survey and airborne sampling of the OSC, TSC,: PSC, 'Warehouse, Badging Center, Training Center, Control Room, or other areas, based upon the priority assigned by the Site Radiation Protection Coordinator. The following activities will be conducted by Health Physics Technicians assigned by the Emergency Assignment Tag Board.:'
(a) Area radiation surveys shall be conducted and documented as prescribed in HP Procedures and Attachment 1, HP: E-Plan Habitability Data Sheet.
(b) *Airborneradioactivity surveys shall be conducted and documented as prescribed in HP Procedures.
(6) The results of such survey and sampling activities shall be reviewed by the OSC HP Supervisor, and their degree of conformance' to normal DAEC Administrative Guidelines determined., The OSC Supervisor and Site
* Radiation Protection Coordinator (SRPC) shall be informed of any abnormal conditions.
                                        -NOTE Normal personnel external and internal exposure limits in effect are prescribed in ACP 1411.18 "Personnel Dosimetry".
Standard    area contamination limits to be observed are prescribed "in ACP- 1411.22 "Personnel Access and Egress in Radiological Areas".
(7) Survey and sampling' activities prescribed above should be conducted in the' Control Room, TSC, and Hot Lab if increasing trends are detected by the installed radiation monitors or as directed by the Site Radiation Protection Coordinator.     .                      -                .
:-;:, lEMERGENCY PLAN IMPLEMENTING PROCEDURE*                           EPIP 3.1 Rev. 14 IN-PLANT RADIOLOGICAL MONITORING Page 8 of 17 (8)   If an assembly area is determined to be uninhabitable or not suitable for long term habitability, the following should occur.
(a)   The Site Radiation Protection Coordinator shall recommend evacuation of selected assembly areas or of the site, and, as appropriate, relocation of essential personnel to alternate assembly areas on site. Evacuation of non-essential personnel is required for a SITE or GENERAL EMERGENCY.
(b)   Non-essential personnel should be evacuated to the ORAA, as prescribed in EPIP 2.4, "Activation and Operation of the ORAA".
(c)   Essential personnel shall be relocated to another assembly area determined to be safe, or occupancy times established for the affected areas and relief shift assignments made so that essential functions can be conducted on a continuing basis.
(9)   If evacuation of one or more assembly areas is required, essential personnel should be relocated to an alternate area, as follows:
Primary Area                   1 5t Alternate               -2   Alternate Control Room             As required to achieve a stable, safe, shutdown condition TSC/OSC                   Control Room                     Security Bldg or ORAA Hot Lab                  ORAL                             N/A Warehouse                 ORAA                             N/A Badging Center           ORAA                             N/A Training Center          ORAA                             N/A PSC                      ORAA                             N/A ORAA                     ODEF                             N/A

EMERGENCY PLAN IMPLEMENTING PROCEDURE                              EPIP 3.1 Rev. 14 IN-PLANT RADIOLOGICAL MONITORING                  -
Page 9 of 17 NOTE When         selecting; "essential"     personnel, consideration should be given to the minimum shift staffing requirements defined in the DAEC Plan, Table B-I. The Emergency Telephone Book may be used as a reference for obtaining shift staffing designations.
Guidance regarding TSC personnel considered to be essential is contained in EPIP 1.3, "Plant and Site Evacuation".'-'
(10) The OSC HP Supervisor shall assure that follow-up monitoring of Assembly Areas is conducted, and shall obtain/seek the concurrence of the OSC Supervisor.
(a) Survey and sampling activities to be conducted and their frequency should be based upon the potential radiological hazards as well as the results of prior monitoring activities and the capability of installed instrumentation to detect changes' in radiological conditions.
(b) In defining the type, frequency and location of monitoring activities to be conducted, consideration should be given to increases in radiation levels throughout the plant.''
(11)''The results 'of such monit6rinig activities shall be reviewed by the OSC HP Supervisor. The OSC Supervisor and Site Radiation Protection' Coordinator shall be advised of any problems or adverse trends detected.
(a) The Site Radiation Protection o'C6rdinator should provide guidance to the OSC Supervisor regardirig'further monitoring activities to be conducted and protective measures to be instituted for personnel in those areas; for example, use of protective clothing and respirators, implementation (or modification) of occupancy times, administration of KI, etc.

  -k,-:-''EMERGENCY PLAN IMPLEMENTING PROCEDURE                             EPIP 3.1            i "
I Rev. 14 IN-PLANT RADIOLOGICAL MONITORING Page 10 of 17 NOTE Use of accepted Health Physics practices (i.e.,
clothing and limiting the stay time) should only be considered for Control Room and TSC personnel.
3.3        IN-PLANT RADIOLOGICAL MONITORING (1)      Radiological monitoring activities inside the plant shall be conducted to support required response actions.
(2)    A Health Physics Technician should accompany each rescue, repair and reentry team upon initial entry into an area where an actual or potential radiological hazard exists.
(a)    As discussed in EPIP 4.3, "Rescue and Emergency Repair Work", repair activities may be conducted during the course of several staged entries; therefore, entries subsequent to the initial entry may be conducted without the presence of a Health Physics Technician, providing the hazards are well known and the possibility of a transient occurring is non-existent.
(b)    The determination as to whether or not a Health Physics Technician shall accompany a repair team or an operator into the plant and remain in their presence shall be made by the Site Radiation Protection Coordinator or OSC HP Supervisor.
(3)    The OSC HP Supervisor shall provide guidance to the Health Physics Technician on the radiological conditions to expect and monitoring activities to be conducted prior to dispatch of personnel into an actual or potential radiological hazard area.
(a)   The OSC HP Supervisor shall ensure that required protective clothing is donned properly.

~.,' ' ,'''EMERGENCY PLANa IM\PLEMENTING 'PROCEDURE                         EPIP 3.1 Rev. 14 IN-PLANT RADIOLOGICALIMONITORING Page 11 of 17 (b) Respiratory protection equipment, if required, will be checked and determined to be operable, and the user aware of equipment operating limitations.
(c) Dosimetry of sufficient range will be worn to adequately monitor whole body and extremity doses, where applicable.
(d) Equipment and supplies anticipated to be required for the assignment are available and checked to ensure operability.
(e) Briefings will be conducted, to the extent required, to ensure that personnel are aware of the jobs to be performed.
(4) Upon being dispatched, the Health Physics Technician should, in addition to monitoring dose rates and total exposures:
(a) Provide guidance to team members regarding precautions to be taken to minimize their total exposure.          .             .    , --
(b) Advise the OSC HP Supervisor of unexpected radiological conditions encountered.
(5)' The OSC HP Supervisor or design6e should, as time permits, monitor the activities of in-plant personnel utilizing the network of cameras available on-site.
(6) Upon egress from the plant, the Health Physics Technician should provide guidance, as required, in discarding protective clothing and in the performance of personnel monitoring and decontamination, if required, prior to passing through the' Health Physics Access Control Area.
                                            . *;    A D *- - *M 3.4         POST ACCIDENT SAMPLING >-;.,                  .                -..
NOTE .'
Any individual involved in obtaining a post accident sample shall comply with the radiological controls as specified by the Emergency Response Organization radiological staff. Any sampling entry shall be preceded by a radiological evaluation and a briefing of the sampling team covering expected radiological conditions, protective measures, monitoring requirements and exposure'reduction 'techniques.

EMERGENCY PLAN IMPLEMENTING PROCEDURED                           EPIP 3.1 Rev. 14 IN-PLANT RADIOLOGICAL MONITORING Page 12 of 17 NOTE Due to the probability of high concentration of noble gases, radiological protective measures specified by the ERO radiological staff will likely include full protective clothing and a self-contained breathing apparatus.
(1) The OSC Supervisor and OSC HP Supervisor shall be advised by the Site Radiation Protection Coordinator of post accident sampling and analysis activities to be conducted.
(2)   If a particulate filter and charcoal cartridge sample from the KAMAN Effluent Monitoring System is to be obtained for analyses, (a)    The Site Radiation Protection Coordinator should determine which cartridge(s) are to be obtained and the radiation level of the cartridge, as monitored by the KAMAN System.
The OSC HP Supervisor should be advised of the projected radiation levels and review with the Site Radiation Protection Coordinator appropriate radiological protective measures to be taken.
(3)   Based upon the expected dose rates, an estimate should be made of the total exposure to be received while obtaining and analyzing a sample and exposure limit increases authorized, if required, as specified in EPIP 4.3, "Rescue and Emergency Repair Work".
(4)   The OSC HP Supervisor shall designate the Chemistry Technician(s) accompanied by a Health Physics Technician to obtain the samples. The OSC HP Supervisor will conduct a briefing consisting of, at a minimum, the expected radiological conditions and associated precautions, increased exposure limits, stay times, access routes and that adequate protective clothing, dosimetry and respiratory protection equipment is donned before each entry.
NOTE Prior to dispatching the Sampling Team personnel into the Power Block Structure, the OSM/OSS should be advised and Emergency Coordinator authorization received.

K>
 ......."-_-"-EMERGENCY PLAN IMPLEMENTING PROCEDURE                   ,      EPIP 3.1 Rev. 14 IN-PLANT RADIOLOGICAL MONITORING Page 13 of 17 (a)   Sample retrieval from the Off-Gas Stack should be accomplished in accordance with PASAP's.
(b)   Sample retrieval from the Reactor Building Stacks and Turbine Building Vent should be accomplished in accordance with PASAP's.
(c)   Grab samples from the Reactor Building should be obtained as prescribed in PASAP's.
(5)  Personnel dispatched to remote locations to obtain samples should maintain communications with the' OSC, advising the OSC HP Supervisor of any unusual phenomena observed.
(6)  Analyses of samples and documentation of results should be conducted using the PASAP's.
4.0   RECORDS (1) All records generated 'as a result of this procedure shall be forwarded to the Site Radiation Protection Coordinator for retention and is considered a QA Record (exception is for records bgenerated'during'drills/exercises) in accordance with EPDM 1007.

==5.0   REFERENCES==

(1)   DAEC Emergency Plan (2)  NUREG 0654, Rev. 1, Planning Standard H, I, J, K (3)  EPIP 1.3, "Plant and Site Evacuation" (4)  EPIP 2.5, "Control Room Emergency Response Organization"

EMERGENCY PLAN IMPLEMENTING PROCEDURESs                  E.:i-.PIP 3.1 Rev. 14 IN-PLANT RADIOLOGICAL MONITORING Page 14 of 17 (5)  EPIP 2.2, "Activation and Operation of the TSC" (6)  HP Procedures (7)  ACP 1411.22 "Personnel Access and Egress in Radiological Areas" (8)  ACP 1411.18 "Personnel Dosimetry" (9)  EPIP 4.3, "Rescue and Emergency Repair Work" (10) EPIP 2.4, "Activation and Operation of the ORAX' (11) ENVIRONMENTAL SAMPLING PROCEDURES (12) PASAP's (13) OFFSITE DOSE ASSESSMENT MANUAL (ODAM) 6.0  ATTACHMENTS (1)  Attachment 1, HP: E-Plan Habitability Data Sheet (2)  Attachment 2, LOCATIONS FOR POST ACCIDENT SAMPLING

I   .
EMERGENCY          PLAN IMPLEMENT I)r 'ROCEDURE                           EPIP 3.1                                 (
C(                                       ~~~~~~~~~~~~~~~~~~.:.....e. . .. :Sli/... .             lS Rev. 14 IN-PLANT RADIOLOGICAL MONITORING                                                                                 I1 Page 15 of 17 ATTACHMENT I HP: E-Plan HabitabilityData Sheet DATE:
TIME                          TIME             4-           TIME -                           TIME        -
DOSE       CONTAM.            DOSE       CONTAM.           DOSE        CONTAM.        I;,     DOSE      CONTAM.
I-'
AREA                  mr/hr          dpm/100cm2   I  mr/hr        dpm/100cm2      mr/hr          dpm/100cm2         mr/hr         dpm/100cm2 4-         7        7        1tI         7       7                               l                               l III        I                     I11I                      + X: -I              I1 I~.-
I I
j I              .
I I
                                                                                                                         ..                  l AIR SAMLES:
COMMENTS:
Health Physics Supervisor Reviewed:                                                   Time:                                  Date:
Surveying HP Tech:                                                                                                             Page:      of-EPIP 3.1

EMERGENCY PLAN IMPLEMENTING PROCEDURE                         EPIP 3.1 Rev. 14 IN-PLANT RADIOLOGICAL MONITORING Page 16 of 17 ATTACHMENT 2 LOCATIONS FOR POST ACCIDENT SAMPLING NOTE Conditions may vary widely and change quickly during preparation for sampling and actual sampling during the period following an accident. The Site Radiation Protection Coordinator (SRPC)/Rad & EOF Manager must be conferred with prior to attempting to retrieve a sample. Assembly of temporary shielding may be necessary.
(1)    Locations from which samples can be drawn include the following points listed below.
Containment isolation logic circuitry modifications have been provided, where required, to permit obtaining samples under isolated conditions.
(a) A depressurized Reactor Coolant System sample from the RHR Heat Exchanger discharge line when RHR is in the shutdown cooling mode of operation.
UI (b) A torus sample from the RHR/Core Spray Fill Pump 1P-70 casing drain, if pump is running and RHR is in the LCPI, torus cooling, or test mode of operation.
(c) A torus sample from the RHR Heat Exchanger discharge line when RHR is in the LCPI, torus cooling, or test mode of operation.
(d) A drywell atmospheric sample from the Containment Atmosphere Monitoring System analyzer sample lines.
(e) A torus atmospheric sample from the Containment Atmospheric Monitoring system analyzer sample lines.
EPIP 3.1

EMERGENCY PLAN IMPLEMENTING' PRCEDUREPP3.
Rev. 14 IN-PLANT RADIOLOGICAL MONITORING Page 17 of 17 (2) High Range Effluent Monitoring (KAMAN) System particulate and iodine filter cartridges can be drawn from:
(a)    Off-Gas Stack (b)    Reactor Building Stacks(s)
(c)    Turbine Building Vent System (3) Reactor Building grab samples can be taken from the backup sampler located on the Turbine Building roof.
EPIP 3.1

   . EMERGENCY PLAN IMPLEMENTING 'PROCEDURE                              EP; 4.,3 EPIP       4.3
.0
                       .      .Rev.                                                  12 RESCUE AND EMERGENCY REPAIR WORK
            .-_        _        __                                         w          1 of 19
                                                                                      .Page Effective Date:                   I 1,,,,
              ;,,,,      ,,,            :CCAL;        EVIEW.-
Prepared by:         _    4            J7..>Date:                           /4 Reviewed by:                           d                       Date:  ________
                                      'Inpendent Reviewer
     ,           4-                              .                                         i,_
I am responsible for the technical content of this procedure.
Approved by:          _                                        Date:    / c2 a Manager, Emergency Planning

3;-;; EMERGENCY PLAN IMPLEMENTiNG PROCEDURE -           EPIP 4.3 Rev. 12 RESCUE AND EMERGENCY REPAIR WORK Page 2 of 19 Table of Contents Pace 1.0     PURPOSE ............                                          3 2.0     DEFINITIONS ............                                      3 3.0     INSTRUCTIONS ...........                                      3 3.1  RESPONSIBILITIES .....................                   3 3.2  RADIOLOGICAL CONCERNS .....................              10 3.3  REQUIREMENTS .....................                       13 3.4  RESCUE OPERATIONS .....................                  14 3.5  EMERGENCY REPAIR .....................                   16 4.0     RECORDS .........                                             19

==5.0     REFERENCES==
 .........                                          19

EMERGENCY, PLAN IMPLEMENTING PROCEDURE                   -ii.1EPIP 4.3       -
Rev. 12 RESCUE AND EMERGENCY REPAIR WORK Page 3 of 19 1.0    PURPOSE (1) This procedure provides guidance for rescue and emergency repair team preparation and activities.
(2) This procedure is applicable to key Emergency Response Organization personnel located in the Control Room, TSC and OSC and to personnel involved in rescue and emergency repair team activities.
2.0      DEFINITIONS (1) None 3.0     INSTRUCTIONS 3.1    RESPONSIBILITIES (1) Security &Support Supervisor (a) Conducting an accountability check for all personnel within the protected area.
(b) -Conducting a security sweep of the,Owner Controlled Area to ensure that everyone, (NMC employees, contractors, visitors, etc) are informed of the event and are proceeding to their assigned Assembly Area.
(c) Directing notification of the Emergency Response Organization, as required.        -     I   ;      :   *
(d) Limiting access into the facility to only those personnel who are required for emergency response efforts.      -:       :
(e) Establishing measures that will enable continuous accountability for all personnel within the protected area once the initial accountability check has been completed.

EMERGENCY PLAN IMPLEMENTING PROCEDURE                           EPIP 4.3 Rev. 12 RESCUE AND EMERGENCY REPAIR WORK Page 4 of 19 (f) Ensuring that no unauthorized personnel gain access to the site.
(g) Assigning personnel for first aid duties, as required.
(h) Ensuring that an accountability check for all personnel within the protected area is conducted in a timely fashion and that requisite security posts are filled.
(i) Ensuring that all individuals in the Owner Controlled Area are accounted for and are in their designated Assembly Area.
(j) Ensuring that the Emergency Response Organization notification process as described in the Emergency Plan Implementing Procedures have been initiated and is successfully completed.
(k) Determining existing and potential administrative support needs and providing direction to the Administrative Supervisor, as required.
(2)  TSC Operations Supervisor (a) Serve as an interface between the Operations Shift Supervisor and all other non-operating personnel and groups.
(b) Provide direction and assistance as necessary to achieve and maintain stable plant conditions.
(c) Assist the Operations Shift Supervisor in coordinating operational activities.
(d)  Monitor operational activities to assure that the plant is operated and maintained in as safe a condition as possible.
(e)  If no abnormal radiological conditions are present, dispatch of operations personnel from the control room to plant locations where local operational response actions may be required to initially mitigate the emergency condition is permitted providing:
: 1. The Emergency Coordinator is informed during his Control Room briefing soon after their dispatch of the location and tasks of any operators initially dispatched into the plant prior to the activation of the OSC.

 ' EMERGENCY PLAN IMPLEMENTING PROCEDURE                    -   EPIP 4.3 Rev. 12 RESCUE AND EMERGENCY REPAIR WORK Page 5 of 19
: 2. After OSC activation, dispatch of any operator into the plant by the Control Room is permitted providing the TSC Operations Supervisor and the OSC Supervisor is informed. The OSC may elect to dispatch a Health Physics technician to accompany these operators. The Health Physics technician shall keep the HP Supervisor or the Site Radiation Protection Coordinator informed of any abnormal radiation conditions that may exist.
                                -;    NOTE.
If additional individuals/resources (other than operations and/or HP personnel) are required to accompany the operator, such as Mechanics, Electricians,'and I & C Technicians then they shall be classified as a team and shall be- coordinated and -controlled by the TSC/OSC.
(f) Coordinate with the OSM/OSS and the Site Rad Protection Coordinator, the dispatch of operators when radiological conditions make entry into the plant hazardous. Inform the OSC and HP Supervisor as soon as possible.
(g) Coordinate, as necessary, with the OSC Supervisor to maintain an awareness of the status of corrective actions in progress by repair teams which have been dispatched and rescue activities which may be in progress. Ensure to'inform'the'OSM/OSS of the repair team progress and results.
(3) Emergency Coordinator (a) Ensure the activation of the DAEC Emergency Response Organization as appropriate for the classification and circumstances of the emergency condition.
(b) Coordinate efforts to return the plant to, and maintain it in a safe, stable condition.
(c) Coordinate accident assessment and analyses efforts to determine the full scope and impact of the emergency.

,EMERGENCY PLAN IMPLEMENTING PROCEDURE                         EPIP 4.3 Rev. 12 RESCUE AND EMERGENCY REPAIR WORK Page 6 of 19 (d) Ensure appropriate initial notification of NMC, local, State, and Federal officials and agencies. This function will be assumed by the Corporate Emergency Response Organization upon activation of the Emergency Operations Facility.
(e) Provide initial protective action recommendations, as appropriate, to local and State authorities who are responsible for offsite protective measures.
This function will be assumed by the Corporate Emergency Response Organizations upon activation of the Emergency Operations Facility.
(f) Apprise NMC, local, State, and Federal officials and agencies of updated information pertaining to the emergency condition.
(g) Classify/Reclassify the event as necessary.
(h) Approve extensions on occupational dose limits, as necessary.
(i) Prepare the DAEC Emergency Response Organization for an orderly transfer of responsibilities to the recovery organization.
(j) Authorize issuance of KI.
(k) Authorize the dispatch of Rescue and Emergency Repair Teams.
(I) Authorize the formation of an OSC Quick Response Team.
NOTE While the administrative aspects of most of these responsibilities may be delegated by the Emergency Coordinator, the responsibilities for items (e), (g), and (h) may not be delegated except as indicated herein.

--'' 'I';EMERGENCY1PLAN       IMPLEMENTING PROCEDURE                     EPIP 4.3 Rev. 12 RESCUE AND EMERGENCY REPAIR WORK Page 7 of 19 (3)    Technical and Engineering Supervisor (a) Verifying that the TSC is fully activated and staffed as described in the Emergency Plan Implemeenting Procedures.
(b) Evaluating plant status and providing support to the operations staff as requested.
(c) Providing overall management and direction to the technical staff assembled in the TSC.
(d) Evaluating recommendations for corrective action provided by the technical staff and operating crew and recommending to the Emergency Coordinator a course of action to be taken to combat the existing situation.
(e) Recommending changes in the Emergency Classification based upon:
: 1. Plant status changes, with or without radiological releases in progress.
: 2. Actual or potential radiological release parameters.
: 3. The progress of activities undertaken to mitigate the situation and their probability for success.
(f) Providing direction to the technical staff comprised of support personnel such as SPDS Operat6r, React6r Engineer, and consultant/contractor representatives to analyze plant conditions and define courses of action to mitigate the emergency situation.
(g) Providing direction to the engffiee'ring staff in the' Technical Support Center to aid in analysis of plant conditions and defines courses of action to mitigate the emergency situation.
(h) Coordinating engineering activities undertaken by the Corporate Emergency Response Organization with efforts being taken at the DAEC to mitigate the event and establish stable plant conditions.
(i) Providing support to the OSC Supervisor as necessary for coordinating all repair/corrective action efforts conducted at the DAEC.

EMERGENCY PLAN IMPLEMENTING PROCEDURE:;                        EPIP 4.3 Rev. 12 RESCUE AND EMERGENCY REPAIR WORK Page 8 of 19 (4)  Site Radiation Protection Coordinator (a)  Verifying with the Security & Support Supervisor that everyone in the Owner Controlled Area is accounted for and in their designated Assembly Area or have been evacuated as applicable.
(b)  Evaluate projected exposure to Rescue and Emergency Repair Teams.
(c)  Make recommendations to the Emergency Coordinator on dose extensions, as necessary.
(d)  Evaluating projected exposure to Rescue and Emergency Repair Teams.
(e)  Ensuring that DAEC personnel are dispatched to monitor the environs in and around the plant for radiological consequences associated with the event.
(f)  Conducting an initial evaluation and assessment of the results of radiological monitoring activities. Upon activation of the Emergency Operations Facility, this function will be assumed by the Radiological Assessment Coordinator for all offsite monitoring activities.
(g)  Assessing the onsite radiological consequences and directing protective measures, including evacuation of the plant in part or in whole.
(h)  During the initial stages of the event, apprising local and State authorities, through the Emergency Coordinator, of the results of radiological monitoring activities and providing protective action recommendations based upon the projected radiological consequences to the population-at-risk. Upon activation of the Emergency Operations Facility, this function will be assumed by the Radiological Assessment Coordinator.
(5)  OSC Supervisor (a)  Supervising the implementation of the tasks and staffing of the OSC.
(b)  Providing general supervision and direction to personnel who report to the Operational Support Center.

" -EMERGENCY-PLAN IMPLEMENTING, PROCEDURE--                     : EPIP 4.3'        -
Rev. 12 RESCUE AND EMERGENCY REPAIR WORK Page 9 of 19 (c)  Coordinating evacuation from the site and all unnecessary personnel in the event that the Operational Support Center is not habitable and during events classified as a SITE AREA or GENERAL EMERGENCY, once
:such an evacuation has b6en' authorized by the Emergency Coordinator.
(d)  Coordinating all repair/corrective action efforts conducted at the DAEC to achieve stable plant conditiori and to'terminate any uncontrolled or excessive radiological release. 7 (e)   Ensuring that personnel dispatched from the Operational Support Center are properly briefed and equipped for their assignment in regards to technical contents as well as ALARA including existing and potential radiological hazards.
(6)  Health Physics Supervisor (a)   Ensuring that personnel dispatched from the Operational Support Center are properly outfitted with protective -clothing and equipment, are briefed regarding ALARA and are apprised of existing and potential radiological hazards.
(b) -Coordinating with the Site Radiation Protective Coordinator in'the TSC to obtain information regarding plant status, problems, response options, significant radiological releases in progress, offsite dose rates, plume location and meteorological conditions as necessary.
(c)   Ensure that determination of habitability of assembly areas and ERO facilities.
(d)  Coordinate the dispatch of monitoring teams, post accident sampling
         -team and ORAA and ORAL/ODEF personnel. Ensure the continued briefing of the ORAA and ORAL/ODEF in terms of plant and radiological conditions.           ;
(e)  Supervise efforts to prepare injured/contaminated personnel for transport to offsite medical facilities.
(f)  Brief rescue and emergency repair team personnel regarding radiological hazards which exist or which potentially may be encountered and provide guidance regarding precautions to be taken and limits that shall not be exceed.

EMERGENCY PLAN IMPLEMENTING PROCEDURE> > -                       EPIP 4.3 Rev. 12 RESCUE AND EMERGENCY REPAIR WORK Page 10 of 19 3.2     RADIOLOGICAL CONCERNS (1) An evaluation of projected exposures to team members shall be made prior to dispatch.
(a)   DAEC limits defined by 10 CFR 20.1201 and EPA-400-R-92-001 are prescribed in ACP 1411.17, "Exposure Limits and Upgrades," and are partially reproduced on OSC-12, "Extemal Exposure Limits".
(b)   If exposures in excess of 5 Rem are projected for team members, the Emergency Coordinator may authorize exposures up to the limits specified on OSC-1 3, "Guidance on Dose Limits for Workers Performing Emergency Services", in accordance with 10 CFR 50.47(b)(1 1), providing that team members are volunteers.
(2)  Individuals assigned to rescue and emergency repair teams shall take reasonable precautions to maintain their exposures as low as possible, consistent with ALARA considerations and job assignment.
(a)   Respiratory equipment, if required, shall be fitted and determined to be functioning properly, as prescribed in ACP 1411.20 "Radiological Respiratory Protection".
(b)   Area radiation surveys shall be conducted as prescribed in HP Procedures and team members shall position themselves to minimize their own exposures due to localized hot spots.
(c)   Personal dosimetry shall be checked periodically, as specified in ACP 1411.18, "Personnel Dosimetry", so as to monitor the total dose received and to ensure that the exposure limits established are not exceeded.
(3)  Based upon the most probable locations where individuals determined to be missing may be located, an assessment of potential radiological exposures should be conducted.
(a)   If no significantly increased radiation levels are indicated by the Area Radiation Monitors (ARMs), projected exposures can be estimated by reviewing the latest survey forms, RWPs and ARM readings coupled with estimated time to conduct rescue activities.

iEMERGENCY PLAN IMPLEMENTINGhPROCEDURE- -'                        EPIP 4.3 Rev. 12 RESCUE AND EMERGENCY REPAIR WORK Page 11 of 19 (b) If significant radiation levels are indicated, i.e., Hi Hi or upscale ARM indications and/or elevated Containment High Range Radiation monitor readings, an estimate of projected exposures can be made by reviewing the Shielding Study or reviewing the results of accident analyses contained in Chapter 15 of the UFSAR coupled with-estimated time to conduct rescue activities.
(4)  In estimating projected exposures consideration' needs to be given to ingress/egress routes and expected dose rates which may be encountered enroute.
                                        .NOTE Operations personnel .should be consulted when evaluating ingress/egress routes due to their familiarity with           physical    layout considerations and awareness of how such routes may be impacted due to changing plant conditions.
(5) The Emergency Coordinator shall approve changes to the established DAEC dose levels, if warranted, provided they are in accordance with 10 CFR 20.1201 and/or EPA 400-R-92-001.
(a) The Site Radiation Protection Coordinator shall be notified of the increased exposure limits.-
(b) The increased levels, the individuals to whom such levels apply and the name of the individual who authorized such levels shall be documented in the Site Radiation Protection Coordinator's Log.
NOTE During'an emergency, the'normal process of requesting and approving increased exposure' limits, prescribed in ACP 1411.17, "Exposure Limits and Upgrades",' need not be followed.'
                                    . ~  .. .. *I

EMERGENCY PLAN IMPLEMENTING PROCEDURE                                         EPIP 4.3 k.&#xa2;4*        *i
* rt-)-  V- +1< I;'P'*,-''5> .;:z;;-
                                                    'V t.' *-.*~t'5hltS.-ik'a';8 Rev. 12 RESCUE AND EMERGENCY REPAIR WORK Page 12 of 19 (6)     The exposure records of personnel assigned to the rescue team should be maintained to ensure that the projected exposure to be received during rescue activities will not result in any member of the team exceeding the limits defined by EPA-400-R-92-001.
(a)   Team membership should be modified if personnel assigned are approaching such limits.
(b)   Team membership shall be comprised of volunteers if projected exposures for the proposed rescue activities will cause personnel exposures to exceed limits specified in 10 CFR 20.1201 and/or EPA-400-R-92-001.
: 1. Documentation that individuals so assigned are volunteers and that they understand the risks and limits established for the assignment shall be recorded in the OSC Supervisor's log along with the names of the volunteers.
(c)    If significant airborne iodine activity is anticipated, consideration should be given to thyroid blocking, as prescribed in EPIP 4.5, "Administration of Potassium Iodide (KI)".
: 1. Guidance with respect to implementation of emergency exposure limits and use of volunteers is contained in OSC-14, "Guidelines Regarding Selection of Volunteers".
(7)     For those response options which will require personnel access into radiation areas in the plant, an assessment of the potential radiological exposures should be conducted, as prescribed in this procedure.
(8)     The benefit of the proposed repair/damage control activity should be evaluated considering:
(a) The magnitude and/or duration of the potential or actual radiological release if the repair/damage control activity is not performed.
(b)    The associated delay or inability to establish control of the plant and/or restore stable conditions.
(c)    The magnitude of the projected exposure to the public calculated in accordance with PASAP's and EPIP 3.3, "Dose Assessment and Protective Action ".

      'EMERGENCY PLAN IMPLEMENTING PROCEDURE-                     rP;  EPIP 4.3 Rev. 12 RESCUE AND EMERGENCY REPAIR WORK Page 13 of 19 (d) Exposures received by skilled personnel during performance of the activity which may make themiriunavailable for further assignments.
(9) If the course of action selected will result in emergency repair team personnel exceeding 5 Rem, see guidance contained in this procedure.
(10) Monitoring shall be performed by the Health Physics Technician, as specified in this procedure.
3.3     REQUIREMENTS (1) Once the TSC has been activated, rescue and emergency repair teams shall not be dispatched without the prior approval of the Emergency Coordinator or designated representative when-radiological conditions make entry into the plant hazardous.
(2) The TSC Operations Supervisor'shall be informed, in advance whenever possible, of the dispatch of personnel to locations where local operational response actions are required.,
(3) Prior to dispatch of rescue and emergency repair teams, briefings shall be conducted by the OSC Supervisor and appropriate discipline supervisor which cover at a minimum:
(a) The activity to be performed.
(b) The specific area where the activity is to be performed.
NOTE For rescue activities, the area designated may only be a best estimate..
(c)    Routes to be used to' gain acce'ssto the area.
(d)    Estimated radiation levels in 'the specific area and radiation levels estimated to be encountered along the route.,
(e) Authorized exposure limits for team members and dose rate limits which should not be exceeded.. -

EMERGENCY PLAN IMPLEMENTING PROCEDURES ---.:JEPIP 4.3 i; a:..   .:-o :,a.i.
                        . 4 *>k>
s  ;.o d'c.- 4e , .~enw- i-,@ .: ,:. ;,'! t'-*;, r.- %- ;
Rev. 12 RESCUE AND EMERGENCY REPAIR WORK Page 14 of 19 (f)     Communications requirements and methods.
(g)     Specific technical and administrative requirements associated with performance of the activity.
(h)     Precautions to be taken during performance of the activity.
3.4      RESCUE OPERATIONS (1)       If accountability for all personnel within the Protected Area, as prescribed in EPIP 1.3, "Plant and Site Evacuation", can not initially be established, one or all of the following actions should be taken in an effort to locate the missing individual(s) and determine if rescue operations are required.
(a)     Contact emergency response facility supervisory personnel to determine if the individual(s) is/are, in fact, present.
(b)     Contact the individual's direct supervisor and personnel with whom the individual was working to determine the job assignment and ascertain last-known whereabouts.
NOTE The Personnel Statement Concerning Incident form may provide information regarding last known whereabouts.
(c)     Check the security computer system for location and time of the employee's last entry or exit.
(d)     Attempt contact of the individual(s) using the plant paging system.
(2)        Rescue teams should be comprised of the following personnel, as necessary:
(a)      One (1) Health Physics Technician.
(b)     Sufficient number of first aid trained individuals to support the applicable emergency.
(c)    Additional personnel as specified by the Security and Support Supervisor to support search and rescue activities.

'EMERGENCYY          PLAN IMPLEMENTING PROCEDURE.                      EPIP 4.3 E:-,
Rev. 12 RESCUE AND EMERGENCY REPAIR WORK
                                          -                            Page 15 of 19 (3)   Prior to dispatch of a rescue team, the OSC Supervisor shall ensure that a briefing is conducted.
(a) The HP Supervisor or his designee should brief team personnel regarding expected radiological hazards, limits which are applicable, considerations associated with injured personnel who may be contaminated with open wounds and/or in high radiation fields, and communications requirements.
(b) The Security & Support Supervisor should check to ensure that first aid personnel have equipment and supplies expected to be necessary and are familiar with techniques which may be required.
(4)  Communications between the rescue team and the OSC Supervisor should be conducted using the Plant Page system or telephones.
(5) The OSC Supervisor will then dispatch and coordinate direction of the rescue teams upon being authorized by the Emergency Coordinator or designated representative.
(6) While enroute, the Health Physics Technician shall:
(a) Monitor ahead of the team to detect areas of high radiation.
(b) Monitor total dose for each team member.
(7) Upon locating the missing individual(s):
(a) The extent of injuries, if any, should be determined.
(b) The OSC Supervisor shall be'advised.
(8) If no injuries were sustained, the individual shall be escorted back to the'OSC.
(9) If injuries are evident, treatment priorities shall be based on the radiological and the medical condition of the injured.:
(a) Such determination should be made jointly between the Health Physics Technician and First Aid personnel.
(b) If both the injury and level of radiation is life threatening, the victim should first be relocated to a "safe" radiation area, then first aid treatment administered.                     S..

EMERGENCY PLAN IMPLEMENTING PROCEDURE'                               EPIP 4.3 Rev. 12 RESCUE AND EMERGENCY REPAIR WORK Page 16 of 19 NOTE As a general rule, consideration should be given to immediately relocating a victim if the radiation level is >25 R/hr.
(c)   If the injury is life threatening but the level of radiation is not, immediate first aid treatment should be provided.
(d)   If the injury is not life threatening, priority should be given to relocating the victim to a safe area, preferably the First Aid Room, before administering first aid.
(10) Unless directed otherwise during the briefing or during subsequent communications with the HP Supervisor, rescue teams should return with the injured to the Health Physics Access Control area.
(11) Additional medical treatment and documentation of injuries, radiological problems and the like shall be performed, as described in EPIP 4.2, "First Aid, Decontamination, and Medical Support".
3.5     EMERGENCY REPAIR (1)   Equipment malfunctions, operational trends and proposed corrective actions should be transmitted to the OSC Supervisor and Technical and Engineering Supervisor by the TSC Operations Supervisor.
(2)  Corrective action options which may include emergency repair and damage control activities should be evaluated and developed by the Technical and Engineering Supervisor, OSC Supervisor, Emergency Coordinator, and TSC Operations Supervisor, priorities should be assigned to each work task. OSC-15 "OSC Repair Team Work Order" should be utilized when forming an emergency repair team. The flow path of the OSC Repair Team Work Order is depicted on OSC-16.

EMERGENCY PL-AN IMPLEMENTING PROCEDURE.                      a-; EPIP 4.3 Rev. 12 RESCUE AND EMERGENCY REPAIR WORK Page 17 of 19 (3)  If it is determined that an quick OSC Repair Team response is necessary, the Emergency Coordinator shall make a PA announcement informing the TSC/OSC of this need. An individual shall be named as the owner, typically this would be the applicable craft OSC Supervisor. The owner shall take the lead in ensuring that all the necessary paperwork is expedited and, coordinating all the necessary briefings. All of the applicable personnel in the TSC and OSC shall give this OSC Quick Response Team their full attention.
All applicable procedures regarding the formation, dispatch, and debriefing of OSC Repair Teams shall be adhered to.
(4)  Where several courses of action can be taken, the Emergency Coordinator shall determine which option is to be pursued.
(a)     Approval of the Emergency Response and Recovery Director is obtained.
(b)     Deviation from such standard practices are documented as Action Requests (AR) under the direction of the Manager, Quality Assurance.
(5)  Personnel initially assigned to emergency repair work by the Emergency Assignment Tag Board, defined in EPIP 2.1, "Activation and Operation of the OSC", include:
(a)     One Mechanic, one Instrument and Controls Technician and two Electricians and one HP Technician (uniless waived by the SRPC).
(b)     Additionally, one Health Physics Technician should be assigned for emergency repair work if work must be accomplished in a radiation area.
(6)  The OSC Supervisor should verify that individuals to be assigned to emergency repair work by the Emergency Assignment Tag Board are appropriately qualified to execute the'desired activities.
(7)  The composition of emergency repair teams should be modified based upon the considerations defiried in this procedure.
(a)     The OSC Supervisor shall redesignate, as necessary, the individuals to be assigned.^e (b)     The OSC Supervisor shall ensure that personnel so identified are available and ready for such assignment.

  ' EMERGENCY PLAN IMPLEMENTING PROCEDURE ' ';                     EPIP 4.3 Rev. 12 RESCUE AND EMERGENCY REPAIR WORK                                                 K)j Page 18 of 19 (8)   Prior to dispatch of an emergency repair team, the OSC Supervisor shall ensure that a briefing is conducted.
(a)   The HP Supervisor or his designee should brief team personnel regarding expected radiological hazards and limits which are'applicable.
(b)   The Technical and Engineering Supervisor or his designee should ensure that team personnel understand the repair or damage control functions to be performed, actions to take if other than expected conditions are discovered, and communications requirements.
(9) Communications between the emergency repair team and the Technical and Engineering Supervisor should be conducted using the Plant Page system or telephones.
(10): The OSC Supervisor shall dispatch the emergency repair team upon receiving authorization from the Emergency Coordinator or designated representative.
(11) Emergency repair or damage control activities shall be conducted in accordance with the instructions provided during the briefing.
(a)    Problems encountered or unexpected conditions discovered should be reported to the Technical and Engineering Supervisor.
(b) The Technical and Engineering Supervisor shall coordinate emergency repair team activities, as required, with the TSC'Operations Supervisor.
(12) Upon completion of emergency repair/damage control activities or as otherwise directed, emergency' repair team members 'shall return to the OSC.
(a)    Normal radiation protection practices shall be observed at the access area unless otherwise'directed by the HP Supervisor.
(b)    Decontamination, if required, shall be performed in accordance with the requirements defined in EPIP 4.2, "First Aid, Decontamination, and Medical Support".
(c) - Debriefing of all teamrs should be performed with the appropriate Craft Supervisor, HP Supervisor, and Engineering as necessary..
(d) All deviations from procedures' or hardware nonconformances will be documented via Action Request (AR) initiation.'
KJ-

EMERGENCY PLAN IMPLEMENTING PROCEDURE                    --;; EPIP
                                                            ;-   -     4.3 Rev. 12 RESCUE AND EMERGENCY REPAIR WORK Page 19 of 19 (13) Emergency repair and damage control activities conducted shall be documented in the Technical and Engineering Supervisor's and OSC Supervisor's log.
4.0   RECORDS (1)  All records generated by this procedure shall be retained IAW QA Record Retention Requirements (except records generated during drills and exercises) in accordance with EPDM 1007.

==5.0    REFERENCES==

(1)  Duane Arnold Energy Center Emergency Plan (2)  Manual of Protective Action Guides and Protective Actions for Nuclear Incidents (EPA 400-R-92-001 dated October 1991 with 2nd printing May 1992)
(3)  10 CFR 20.1201 (4)  Updated Final Safety Analysis Report (UFSAR), Chapter 15 (5)  ACP 1411.20 "Radiological Respiratory Protection" (6)  ACP 1411.17, "Exposure Limits and Upgrades" (7)  ACP 1411.18, "Personnel Dosimetry" (8)  HP Procedures (9)  EPIP 4.5, "Administration of Potassium Iodide (KI)"
(10) EPIP 4.2, "First Aid, Decontamination, and Medical Support" (11) EPIP 3.3, "Dose Assessment and Protective Action" (12) EPIP 2.1, "Activation and Operation of the OSC" (13) PASAP's

HEALTH PHYSICS SUPERVISOR CHECKLIST The following checklist is available for use by the Health Physics Supervisor or another individual, so designated, as an aid in ensuring that emergency response actions are completed.
* ITEM
: 1.      Relocate nameplate on the Emergency Assignment Staffing Board.
2;      Report to the OSC Supervisor.
: 3.      Assign experienced Health Physics personnel to supervise and/or assist others performing selected functions.
: 4.      Ensure inventory instrument checks and zeroing of dosimeters are accomplished.
: 5. Determine if an onsite radiological problem exists and provide direction to personnel who need to exit the building.
: 6. Ensure action is initiated to determine the habitability of assembly areas including:
* OSC
>
* Assembly Building -
* TSC Hot Lab
* Control Room
* Badging Center
* Training Center
* Warehouse     -
* Data Acquisition Center
: 7. Review "Personal Statement Concerning Incident" forms, evaluate the results of the review and communicate areas of concern to the OSC Supervisor, Site Radiation Protection Coordinator and/or Emergency Coordinator, as appropriate.
: 8. Ensure that team status board is updated.
OSC-AR Rev.I                                         Page 1 of 2

HEALTH PHYSICS SUPERVISOR CHECKLIST ITEM
: 9. Coordinate with the Site Radiation Protection Coordinator (SRPC) to obtain information regarding plant status, problems, response options, significant radiological releases in progress, offsite dose rates, plume location and meteorological conditions.
: 10. Conduct briefings for Field Radiological Monitoring Teams (onsite and offsite) to verify that Teams are ready to be dispatched and advise the OSC Supervisor.
: 11. Supervise efforts required to prepare injured/contaminated personnel for transport to offsite medical facilities and assign Health Physics personnel, as required, to accompany personnel to offsite facilities.
: 12. Provide assistance, as requested, by the TSC Supervisor/OSC Supervisor to provide an HP to accompany repair teams, assure that team personnel are properly clothed, have the required dosimetry and the required respiratory protection equipment, and are briefed regarding ALARA.
: 13. Review the results of radiological monitoring activities and advise the SRPC of problems discovered, unexpected results, etc.
: 14. Coordinate with the OSC Supervisor/SRPC the dispatch of ORAIJODEF personnel.
: 15. Coordinate with ORAA Supervisor and assign two Health Physics Techs to activate the ORAA and assist in evacuating personnel.
: 16. Identify services required to support ongoing radiological monitoring activities and advise the SRPC of such needs.
: 17. Schedule Health Physics Technicians to provide continuous radiological coverage, as necessary.
: 18. Ensure personnel check personal dosimeters as necessary.
: 19. Provide periodic briefings to ORAA staff on plant/radiological conditions.
(SC-09                                                                                           K-Rev. I                                   Page 2 of 2

EMERGENCY ASSIGNMENT STAFFING BOARD DUTIES ASSIGNMENT:          RESCUE, REPAIR AND MONITORING
* JOB CLASSIFICATR )N: HEALTH PHYSICS TECHNICIAN INSTRUCTIONS:    A. Locker #1, Shelf #1 -;Verify operability of the equipment identified on the appropriate inventory list posted on the inside of the locker door.
B. Dress in one full set of protective clothing.
C. Inform the HP Supervisor that you have completed equipment verification checks and are standing by for a briefing.
D. When directed, attend briefing with HP Supervisor.
:tI E. Obtain additional protective clothing, respiratory equipment, etc., as required by the HP Supervisor.
F. Advise HP Supervisor that you have obtained all required equipment and are awaiting further instructions.,
Page 1 of 21 i~R    oscv Rev. I

                                                                                                        /
i.
EMERGENCY ASSIGNMENT STAFFING BOARD DUTIES ASSIGNMENT:         RESCUE, REPAIR AND MONITORING
* JOB CLASSIFICATION: HEALTH PHYSICS TECHNICIAN INSTRUCTIONS:    A. Locker #1, Shelf #5 - Obtain and verify operability of the equipment identified on the appropriate inventory list posted on the inside of the locker door.
B. Dress in one full set of protective clothing.
C. Inform the HP Supervisor that we have completed equipment verification checks and are standing by for a briefing.
D. When directed, attend briefing with OSC Supervisor.
E. Obtain additional protective clothing, respiratory equipment, etc., as required by the HP Supervisor.
F. Advise the HP Supervisor that you have obtained all required equipment and are awaiting further instructions.
                                                                                                     <21-Page 2 of 21 0osc-Rev. I

EMERGENCY ASSIGNMENT STAFFING BOARD DUTIES ASSIGNMENT:         HABITABILITY MONITORING
* JOB CLASSIFICATION: HEALTH PHYSICS TECHNICIAN INSTRUCTIONS:    A. Locker #1, Shelf #3 - Verify operability of the equipment identified on the appropriate inventory list posted on the inside of the locker'door.
B. Inform the BP Supervisor that you are standing by at the Emergency Locker Area and are ready to verify habitability.
C. Perform a radiation smear and airborne survey to verify habitability of assembly areas as directed by the'HP Supervisor.
NOTE If normal counting equipment is available, such equipment should be used to analyze smears'and airborne filter cartridges.
D. Setup -and' Operate OSC and TSC continuous air monitoring systems as necessary to ensure habitability.
E. Record all results (sample time, count time, location and results), using appropriate HPP forms, etc..
F. Report results to the HP Supervisor as they are obtained (include samiple time, location, and results).
G. Collect and review Personal Statement Concerning Incident forms.
H. Repeat Steps, D, F and G on a periodic basis as directed by the HP Supervisor.
Page'3 df21 oso.)'
Rev. I

EMERGENCY ASSIGNMENT STAFFING BOARD DUTIES ASSIGNMENT:         ONSITE FIELD RADIOLOGICAL MONITORING (TEAM LEADER)
* JOB CLASSIFICATION: HEALTH PHYSICS TECHNICIAN INSTRUCTIONS:    A. Locker #2, Shelf #1 - Verify operability of the equipment identified on the appropriate inventory list posted on the inside of the locker door.
B. Notify the HP Supervisor that the Field Onsite Radiological Monitoring Team is ready for their briefing, and attend the briefing, when scheduled.
D. Obtain additional protective clothing, respiratory           equipment, hi-range dosimetry, etc., as required by the HP Supervisor.
NOTE If a severe radiological problem exists onsite, don protective clothing and respiratory protection equipment, as directed by the HP Supervisor.
E. Transport all necessary equipment to the vehicle and load the materials.
F. Inform the HP Supervisor that the Field Onsite Radiological Monitoring TeanxmJ is prepared to be dispatched and proceed, as directed.
Page 4 of 21 osMi Rev. I

EMERGENCY ASSIGNMENT STAFFING BOARD DUTIES ASSIGNMENT:          OFFSITE FIELD RADIOLOGICAL MONITORING (A-TEAM LEADER)
* JOB CLASSIFICATION:  HEALTH PHYSICS TECHNICIAN INSTRUCTIONS:    A. Locker #2, Shelf #2 - Obtainr and verify operability of the equipment identified on the appropriate inventory list posted on the inside of the locker door.
B. Notify the HP Supervisor that the Field Radiological Monitoring Team A is ready for their briefing, and attend the briefing, when scheduled.
C. Obtain additional protective clothing, respiratory equipment,             hi-range dosimetry, etc., as required by the HP Supervisor.
NOTE If a severe radiological problem exists onsite, don protective clothing and respiratory protection equipment, as directed by the HP Supervisor.
1                  D. Transport all necessary equipment to the vehicle and load the materials.
E. Inform the HP Supervisor that the Field Radiological Monitoring Team A is prepared to be dispatched and proceed, as directed.
Page 5 of 21 osc-II Rev. I

EMERGENCY ASSIGNMENT STAFFING BOARD DUTIES ASSIGNMENT:         RESCUE, REPAIR AND MONITORING                  *
* JOB CLASSIFICATION: HEALTH PHYSICS TECHNICIAN INSTRUCTIONS:    A. Locker #1, Shelf #2 - Verify operability of the equipment identified on the appropriate inventory list posted on the inside of the locker door.
B. Dress in one full set of protective clothing.
C. Inforrm the HP Supervisor that you have completed equipment verification checks and are standing by for a briefing.
D. When directed, attend briefing with HP Supervisor.
E. Obtain additional protective clothing, respiratory equipment, etc., as required by the HP Supervisor.
F. Advise the HP Supervisor that you have obtained all required equipment and are awaiting further instructions.
Page 6 of 21 osc4I Rev. I

EMERGENCY ASSIGNMENT STAFFING BOARD DUTIES ASSIGNMENT:          PERSONNEL MONITORING *
* JOB CLASSIFICATION:  HEALTH PHYSICS TECHNICIAN INSTRUCTIONIS: A. Locker #1, Shelf #4'-Obtain the equipment identified on the appropriate inventory list posted on the inside of the locker door.
                  'B. Inform' the HP Supervisor that you are reporting to Access Control to assist with evacuated personnel monitoring.
C. Pass out Personal Statement Concerning Incident forms to evacuated personnel located in the OSC Tag Board Area, OSC Break Room, and Access Control Area. Instruct evacuated personnel to fill out the form and return to you when completed.
D. Provide assistance in surveying and decontaminating personnel evacuating the Power Block Structure.'
E. Inform HP Supervisor that personnel monitoring activities have been completed at Access Control.
F. 'Issue Self-Alarming Dosimeters (Alnors) as needed.
Page 7 of 2"',
I  osc-I ReV. I

EMERGENCY ASSIGNMENT STAFFING BOARD DUTIES ASSIGNMENT:         OFFSITE FIELD RADIOLOGICAL MONITORING (B-TEAM LEADER) * -*
JOB CLASSIFICATION: HEALTH PHYSICS TECHNICIAN INSTRUCTIONS:    A. Locker #2, Shelf #3 - Obtain and verify operability of the equipment identified on the appropriate inventory list posted on the inside of the locker door.
B. Locate the Offsite Field Radiological Monitoring B-Team Member and notify the HP Supervisor that the Field Radiological Monitoring Team B is ready for their briefing, and attend the briefing, when scheduled.
C. Obtain additional protective clothing, respiratory equipment, hi-range dosimetry, etc., as required by the HP Supervisor.
NOTE If a severe radiological problem exists onsite, don protective clothing and respiratory protection equipment, as directed by the IlP Supervisor.
D. Transport all necessary equipment to the vehicle and load the materials.
E. Inform the HP Supervisor that the Field Radiological Monitoring Team B is prepared to be dispatched and proceed, as directed.
Page 8 of21 osx-I Rev. I

EMERGENCY ASSIGNMENT STAFFING BOARD DUTIES ASSIGNMENT:          RESCUE AND REPAIR '*
JOB CLASSIFICATION:  INSTRUMENT TECHNICIAN INSTRUCTIONS:    A. Locker #3, Shelf #2-- Verify operability'of the equipment identified on the appropriate inventory list posted on the inside of the locker door.
NOTE
                   - Inventory of the Instrumentation Tool Kit'is not necessary if the seal is intact. If the seal is broken, take an inventory of tools using the inventory list located inside the tool kit.'
B. Dress in one full set of protective clothing.
C. Inform the I&C Maintenanc'e Supervisor that'you have completed equipment verification checks and are standing by for a briefing.
D. When directed, attend briefing with I&C Maintenance and HP Supervisor.
BE. Obtain additional protective clothing, respiratory equipment, etc., as required by the HP Supervisor.
F. Upon receiving an assignment, break the seal on the Instrumentation Tool Kit, check to see that you have the necessary tools for the assigned task, and obtain additional tools and equipment, if necessary.
G. Advise I&C Supervisor that you have obtained all required equipment and are awaiting further instructions: '         -
Page 9 of 21 osc-II Rev. I

EMERGENCY ASSIGNMENT STAFFING BOARD DUTIES ASSIGNMENT:         RESCUE AND REPAIR
* JOB CLASSIFICATION: ELECTRICIAN INSTRUCTIONS:    A. Locker #3, Shelf #3 - Verify operability of the equipment identified on the appropriate inventory list posted on the inside of the locker door.
NOTE Inventory of the Electrician's Tool Kit is not necessary if the seal is intact.
If the seal is broken, take an inventory of tools using the inventory list located inside the tool kit.
B. Dress in one fill set of protective clothing.
C. Inform the Electrical Maintenance Supervisor that you have completed equipment verification checks and are standing by for a briefing.
D. When directed, attend briefing with Electrical Maintenance and HP Supervisor.
E. Obtain additional protective clothing, respiratory equipment, etc., as require,_>
by the HP Supervisor.
F. Upon receiving an assignment, break the seal on the Electrician's Tool Kit, check to see that you have the necessary tools for the assigned task, and obtain additional tools and equipment, if necessary.
G. Advise Electrical Supervisor that you have obtained all required equipment and are awaiting further instructions.
Page 10 of 21                                                   V>
osc-u Rev. I

EMERGENCY ASSIGNMENT STAFFING BOARD DUTIES ASSIGNMENT:          RESCUE AND REPAIR             *
* JOB CLASSIFICATION:  ELECTRICIAN INSTRUCTIONS:    A. Locker #3, Shelf #4 -'Verify operability of the equipment identified on the appropriate inventory 'list from EPIP 6.1 posted on the inside of the locker door.
NOTE Inventory of the Electrician's Tool Kit is not necessary if the seal is intact.
If the seal is broken, take an inventory of tools using the inventory list located inside the tool kit.
B. Dress in one full set of protective clothing.
C. Inform the Electrical !'Maintenance Supervisor that you have completed equipment verification checks and are standing by for a briefing.
D. When directed, attend briefing with Electrical Maintenance and HP Supervisor.
E. Obtain additional protective'clothing, respiratory equipment, etc., as required by the HP Supervisor.
F. Upon receiving an assignment, break the seal on the Electrician's Tool Kit, check to see that you have the necessary tools for the assigned task, and obtain additional tools and equipment, if necessary.
G. Advise Electrical Supervisor that you have obtained all required equipment and are awaiting further instructions.
Page II of 21 osc-u Rev. I

EMERGENCY ASSIGNMENT STAFFING BOARD DUTIES ASSIGNMENT:         RESCUE AND REPAIR JOB CLASSIFICATION: MECHANIC *
* INSTRUCTIONS:    A. Locker #3, Shelf #5 - Verify operability of the equipment identified on the appropriate inventory list posted on the inside of the locker door.
NOTE Inventory of the Mechanical Tool Kit is not necessary if the seal is intact.
If the seal is broken, take an inventory of tools using the inventory list located inside the tool kit.
B. Dress in one full set of protective clothing.
C. Inform the Mechanical Maintenance Supervisor that you have completed equipment verification checks and are standing by for a briefing.
D. When directed, attend briefing with Mechanical Maintenance and HP Supervisor.
E. Obtain additional protective clothing, respiratory equipment, etc., as required by the HP Supervisor.
F. Upon receiving an assignment, break the seal on the Mechanical Tool Kit, check to see that you have the necessary tools for the assigned task, and obtain additional tools and equipment, if necessary.
H. Advise Mechanical Supervisor that you have obtained all required equipment and are awaiting further instructions.
Page 12of21 oxsC-Rev. I

EMERGENCY ASSIGNMENT STAFFING BOARD DUTIES ASSIGNMENT:          OSC COMMUNICATOR'AND DATA RECORDER
* JOB CLASSIFICATION: RADWASTE OPERATOR INSTRUCTIONS:    A. Locker #1, Shelf #4 - Verify operability of the equipment identified on the appropriate inventory list posted on the inside of the locker door. Retrieve notebook from locker.
B. Proceed to the OSC.
C. Verify operability of communication equipment in the OSC.
D. Review Attachment 8, "Recommended Log Entry Topics".
E. Initiate maintenance of the OSC Log.
F. Assist the OSC Supervisor with communications by functioning as the OSC Supervisor's communicator.
G. When all 30 rmin. positions have been staffed, relocate the Emergency Assignment Staffing Board to the OSC office area.
H. Inform the OSC Supervisor of all pertinent information received.
P 1 Page 13 'of 21 "
,\J ~Osc-a Rev. I

EMERGENCY ASSIGNMENT STAFFING BOARD DUTIES ASSIGNMENT:           OFFSITE FIELD RADIOLOGICAL MONITORING (A-TEAM MEMBER)*
JOB CLASSIFICATI, ON: RADWASTE OPERATOR INSTRUCTIONS:     A. Locker #2, Shelf #5 - Obtain and verify operability of the equipment identified on the appropriate inventory list posted on the inside of the locker door.
B. Go to the Security Control Point (SCP) and:
: 1. Obtain keys for a vehicle, unless already checked out.
C. Obtain 800 MHz hand held radio from the OSC.
D. Check with the HP Supervisor to determine if severe radiological problems exist onsite.
NOTE If a severe radiological problem exists onsite, don protective clothing an&)
mask, as directed by the HP Supervisor.
E. Check to see if the vehicle is operable and fill the gas tank, as necessary. (Do not leave vehicle unattended inside the protected area).
F. Mount the radio antenna on the vehicle.
G. Park the vehicle outside the fence near the main entrance.
Page 14 of21l oSC4I Rev. I

                     -EMERGENCY ASSIGNMENT STAFFING BOARD DUTIES ASSIGNM] ENT:         OFFSITE FIELD RADIOLOGICAL MONITORING (A-TEAM MEMBER)
* JOB CLASE;IFICATION: RADWASTE OPERATOR INSTRUCI IONS: H. Obtain a 12-volt battery (automotive type) from the Battery Storage Locker in the OSC hallway.
I. Locate the Offsite Field Radiological Monitoring A-Team Leader and advise him that you are ready for the briefing.
J. Attend the briefing conducted by the HP Supervisor.
K. Obtain additional protective clothing, respiratory         equipment, hi-range dosimetry, etc., as required by the OSC Supervisor.
L. Assist in loading equipment and supplies into the vehicle.
M. Perform radio check with TSC.
Page 15 of 21 K.'  osC-Re. I

EMERGENCY ASSIGNMENT STAFFING BOARD DUTIES ASSIGNMENT:         ONSITE FIELD RADIOLOGICAL MONITORING (TEAM MEMBER) *
* JOB CLASSIFICATION: RADWASTE OPERATOR INSTRUCTIONS:    A. Locker #2, Shelf #4 - Verify operability of the equipment identified on the appropriate inventory list posted on the inside of the locker door.
B. Obtain the Operations (onsite) radio from OSC and perform radio check with CAS.
C. Obtain keys for a vehicle unless already checked out.
D. Check with the HP Supervisor to determine if severe radiological problems exist on site.
NOTE If a severe radiological problem exists onsite, don protective clothing and mask, as directed by the HP Supervisor.
E. Take the Operations (onsite) radio and vehicle key with you to the vehicle.
(Do not leave the vehicle unattended inside the protected area).
F. Check to see if the vehicle is operable and fill the gas tank, as necessary.
G. Park the vehicle outside the fence near the main entrance.
Page 16 of 21 osc-1 Rev. I

EMERGENCY ASSIGNMENT STAFFING BOARD DUTIES ASSIGNMENT:          ONSITE FIELD RADIOLOGICAL MONITORING (TEAM MEMBER) *
* JOB CLASSIFICATI ON: RADWASTE OPERATOR INSTRUCTIONS:    H. Obtain a 12-volt battery (automotive type) from the Battery Storage Locker in the
           - . '        OSC hallway.
I. Attend with the Onsite Field Monitoring Team Leader, the briefing conducted by the HP Supervisor.
J. Obtain additiohal protective clothing, respiratory equipment, etc., as required by the HP Supervisor.
K. Assist in loading equipment and supplies into the vehicle.
K)
Pag&'17 of 21 osc-l1 Revp I

I EMERGENCY ASSIGNMENT STAFFING BOARD DUTIES ASSIGNMENT:          OFFSITE FIELD RADIOLOGICAL MONITORING (B-TEAM MEMBER) *
* JOB CLASSIFICATI1ON: RADWASTE OPERATOR INSTRUCTIONS:    A. Locker #3, Shelf #1 - Obtain and verify operability of the equipment identified on the appropriate inventory list posted on the inside of the locker door.
B. Go to the Security Control Point (SCP) and:
: 1. Obtain keys for a vehicle, unless already checked out.
B. Obtain 800 MHz hand held radio from the OSC.
C. Check with the HP Supervisor to determine if severe radiological problems exist onsite.
NOTE If a severe radiological problem exists onsite, don protective clothing         V..)
mask, as directed by the HP Supervisor.
D. Check to see if the vehicle is operable and fill the gas tank, as necessary.
E. Mount the radio antenna on the vehicle.
F. Park the vehicle outside the fence near the main entrance.
Page 18 of 21                                                    J osc41 Rev. I

                           ,~- t !.
EMERGENCY ASSIGNMENT STAFFING BOARD DUTIES ASSIGNMENT:             OFFSITE FIELD RADIOLOGICAL MONITORING (B-TEAM MEMBER) *
* JOB CLASSIFICATI lON:   RADWASTE OPERATOR INSTRUCTIONS:     G. Obtain a 12-volt battery (automotive type) from the Battery Storage Locker in the OSC hallway.
H. Locate the Offsite Monitoring B-Team Leader and advise him that you are ready for the briefing conducted by the HP Supervisor.
I. Obtain additional protective clothing, respiratory equipment, hi-range dosimetry, etc., as required by the HP Supervisor.
J. Assist in loading equipment and supplies into the vehicle.
K. Perform radio check with TSC.
Page 19 of 21 osc41 Rev. I

EMERGENCY ASSIGNMENT STAFFING BOARD DUTIES ASSIGNMENT:         CHEMISTRY SUPPORT
* JOB CLASSIFICATION: CHEMISTRY TECHNICIAN INSTRUCTIONS:    A. Locker #4 -Shelf #1- Obtain a 5R and 10R Dosimeter. Obtain additional protective clothing, respiratory protection equipment, special dosimetry etc., as required by HP Supervisor.
B. Inform the HP Supervisor that you have assumed responsibility for Chemistry support. Inquire as to the need for post accident sampling     I activities.
: 1. If post accident sampling activities are not currently necessary,     I advise the OSC Supervisor that you are proceeding to the Hot Lab to provide Count Room support.
NOTE Verify that the Hot Lab ARM is Not Upscale or Alarming.
: 2. If post accident sampling is necessary, obtain briefing from HP Supervisor.
C. Advise HP Supervisor when you have obtained all required equipment and are awaiting fuirther instruction.
D. When directed by HP Supervisor, commence sampling as required.
Page 20 of 21 osMal Rep. I

I EMERGENCY ASSIGNMENT STAFFING BOARD DUTIES ASSIGNMENT:         CHEMISTRY SUPPORT *
* JOB CLASSIFICATION: CHEMISTRY TECHNICIAN INSTRUCTIONS:    A. Locker #4, Shelf #1 - Obtain a 5R and IOR Dosimeter. Obtain additional protective clothing, respiratory protection equipment, special dosimetry etc., as required by HP Supervisor.
B. Inform the HP Supervisor that you have assumed responsibility for Chemistry support. Inquire as to the need for post accident sampling activities.
: 1. If post accident sampling activities are not currently        I necessary, advise the OSC Supervisor that you are proceeding to the Hot Lab to provide Chemistry Support.
NOTE Verify that the Hot Lab ARM is Not Upscale or Alarming.
: 2. If post accident sampling is necessary, obtain briefing from  I HP Supervisor.
C. Advise HP Supervisor when you have obtained all required equipment and are awaiting further instruction.
D. When directed by HP Supervisor, commence sampling as required.
Page 21 of 21 oscI Rev.)I

*    [d7-7,                 - DAEC-EMERGENCY RESP.ONSE RGANIZATION.-..
POSITION          PECIFICZHECKLIST;                 .`
FACILITY:         TSC         ERO POSITION: SITE RAD PRO COORDINATOR EPIP FORM TSC-03              REVISION #: 3 NAME:                                                                         DATE:_
NOTE This checklist.is intended to be an aid in your response to the ERO.                      Reference the applicable Emergency Procedures often, as time permits, to ensure compliance.
REFERENCES DAEC EMERGENCY PLAN                                                 DAEC EPIP's Section B, 'Emergency Response Organization'              1.3, 'Plant Assembly and Accountability' Section J, 'Protective Response'                          2.1, 'Activation & Operation of the OSC' Section K, 'Radiological Exposure Control'                2.2, 'Activation & Operation of the TSC' Section L, 'Medical and Public Health Support'            3.3, 'Dose Assessment & Protective Action' Section M, 'Recovery and Reentry Planning and             4.5, 'Administration of KI' Post-Accident Operation' 5.2, 'Recovery & Reentry' BASIC PURPOSE OF THIS ERO POSITION:
This position is responsible for all on-site radiological activities and protection. Also responsible for the On-Site Field Team and the Off-Site Field Teams until the EOF relieves the SRPC of Off-Site Field Team responsibility. Also, assist the EC in making PARs.
TSC-03                                          Page 1 of 5 Rev. 03

lT AECEMERGENCYtRESPONSEORGANIZATION^8Ntt, .-!- ;--.
QPOSITION SPECIFIC CHECKLIST                        K; ACTIVATION Report to the TSC.
Swipe in the emergency accountability card reader, sign in on the Emergency Assignment Staffing Board, acquire the Site Radiation Protection Coordinator badge.
Locate and utilize the Site Radiation Protection Coordinator Position Specific Manual and checklist.
Verify the following positions have been staffed.
MIDAS Terminal Operator (Chem Tech) (shared resource with the EOF)
Radiological Support Staff Member Field Team Director - Onsite Field Team Director - Offsite NRC HPN Communicator Report staffing results to the Tech & Eng Supervisor OPERATION Verify MIDAS is operable.
Verify TSC & OSC is habitable.
Verify EC or Tech & Eng Supervisor initiates a PA announcement with regard to facility habitability.
Verify operational status of SBGT Trains, inform MIDAS Operator of status.
Verify reactor trip. Time                 (All rods full in)
Verify Radiological Support Staff person is acquiring ARM data.
Advise EC of dose extensions. (Only EC may authorize dose extensions)
Review MIDAS projections every 15-30 minutes and issue Protective Action Recommendations (PAR's) to the EC, as necessary.
TSC-03                                   Page 2 of 5 Rev. 03

                                                                                     .=1 f                 '        '   'EC EMERGENCY REPONSEORGANIZATION...
                     '  t-      J?    TSION SPECIFIC-CHECKLIST Verify Linn Co., Benton Co., and the State of Iowa EMA's are notified of any PAR changes within 15 minutes'of EC approval. The usual path for those notifications once the TSC is activated,-'is thr6ugh the Admin Supervisor.
Verify MIDAS information is consistent with data obtained from the Offsite Field Monitoring Teams.                                -
Authorize the administration'of Ki'as 66cessary. Inform the EC.
MEDICAL Name of Injured: ' -
Badge:                               -                    '-
Location and Contamination Levels'         -'
          -Verify hospital notified to receive coi'ntamrinrated/injured person Hospital name:                                                   Time:
Verify Medical Consultant notified of contaminated/injured person (telephone numbers in the ETB).
Personal information and condition of contaminated/injured person provided to the EC.
Verify follow-up of patient's condition.
Verify follow-up information provided to the EC.
TSC-03                                    Page 3 of 5 Rev. 03

DAEC-EMERGENCY-RESPONSEa ORGANIZATIOTNrus it                   arm ta V-ax-POSITIO'N SPECIFIC; CHECKLMIS RADIOLOGICAL MONITORING TEAMS Provide information regarding the projected or ongoing release to the HP Supervisor for dissemination to the Onsite and /or Offsite Radiological Monitoring Teams prior to dispatch.
Verify dispatch of the onsite field team. Time of dispatch:
Assume control of the onsite team upon being advised of their readiness for dispatch by the HP Supervisor.
Verify dispatch of the Offsite Field Team A. Time of dispatch:
Verify dispatch of the Offsite Field Team B. Time of dispatch:
EC informed of dispatch of Onsite and Offsite Field Teams.
Assume control of the offsite teams upon being advised of their readiness for dispatch by the HP Supervisor.
Transfer control of the offsite teams to the Radiological Assessment Coordinator (RAC) in the EOF. Time:
ORAA ACTIVATION Review current MIDAS projection for direction of plume travel prior to activation of ORAA.
Determine travel route for staff and evacuees.
Verify EC informed of impending ORAA activation and travel route.
Authorize OSC Supervisor to activate the ORAA. Time:
Verify HP Supervisor informed of impending ORAA activation.
Verify ORAA operational. Time:
Verify plume direction periodically; relocate ORAA personnel to the ORAL/ODEF as necessary.
TSC-03                                   Page 4 of 5 Rev. 03

EMERGENC,                NXSE ORGANIZATIONl POSITION SPECIFICqCHECKLIST ORAL1ODEF ACTIVATION Authorize OSC Supervisor to activate the ORAL/ODEF. Time:
Verify HP Supervisor informed of impending ORAUODEF activation.
Verify ORAUODEF operational. Time:
Verify EC informed the ORAUODEF is operational.
POST ACCIDENT SAMPLING (as necessary or requested)
Verify that post accident sampling is initiated.
Reactor Coolant
                   -     Suppression Pool Water
                   -     Offgas Stack Containment Atmosphere K>
Reactor Building Vents Turbine Building Vents RECOVERYIREENTRY Assist the EC is recovery/reentry activities as requested.
TSC-03                                 Page 5 of 5 Rev. 03

NMfi)
Committed to Nuclear Excellence DAEC EMERGENCY PLANNING DEPARTMENT PROCEDURE TRANSMITTAL ACKNOWLEDGEMENT MEMO (TAM-1 18)
To:             NRC-NRR Document Control Desk US NRC Washington DC 20555 Re:             Entire DAEC Emergency Plan (Table of Contents Rev) (Copy 91)
PSM

==Title:==
n/a Distribution Date:              02 / 04 /2004 Effective Date of Change:       02/05/2004 Return by:                      02 / 19 / 2004 Please perform the following to your assigned manual. If you have any questions regarding this TAM please contact Don A. Johnson at 319-851-7872.
REMOVE              INSERT DAEC E-PLAN App-2 (PWR: 24060)                                   Rev. 20             Rev. 21   -
DAEC E-PLAN Table of Contents Revision                           Rev. 26             Rev. 27 DAEC E-PLAN Plan 'B' (PWR: 23996)                                Rev. 25             Rev. 26 DAEC E-PLAN Plan 'C' (PWR: 23997)                                Rev. 20             Rev. 21   -
DAEC E-PLAN Plan 'I' (PWR: 23998)                                Rev. 20             Rev. 21 DAEC E-PLAN Plan 'N' (PWR: 23999)                                Rev. 20             Rev. 21 PERFORMED BY:
Print Name                                   Sign Name                          Date Please return to:     K. Dunlap PSC/Emergency Planning 3313 DAEC Rd.
Palo, IA 52324 To be completed by DAEC EP personnel only:
Date TAM returned:
EPTools updated:                                                                             Page 1 of 1

NiiiC Nu Duane Arnold Energy Center Operated by Nuclear Management Company, LLC Committed to Nuclear Ekce ncen Wednesday, February 4, 2004 NRC-NRR Document Control Desk US NRC Washington, DC 20555 To: NRC-NRR Document Control Desk From: DAEC Emergency Planning Department Re: Description of changes to the following documents DAEC E-PLA App-2                   Letters of Agreement Removed reference to a Letter of Agreement with Framatome.
For TS Amendment #252 - PASS Elimination, this is one of the procedures to be changed to reflect the elimination of the requirement to perform a PASS sample or have and maintain the system.
PASS will be 'Abandoned In Place'.
DAEC E-PLA Plan 'B'                Emergency Response Organization For TS Amendment #252 - PASS Elimination, this is one of the procedures to be changed to reflect the elimination of the requirement to perform a PASS sample or have and maintain the system.
PASS will be 'Abandoned In Place'.
DAEC E-PLA Plan 'C'                Emergency Response Support and Resources For TS Amendment #252 - PASS Elimination, this is one of the procedures to be changed to reflect the elimination of the requirement to perform a PASS sample or have and maintain the system.
PASS will be 'Abandoned In Place'.
DAEC E-PLA Plan 'I'                Accident Assessment For TS Amendment #252 - PASS Elimination, this is one of the procedures to be changed to reflect the elimination of the requirement to perform a PASS sample or have and maintain the system.
PASS will be 'Abandoned In Place'.
DAEC E-PLA Plan 'N'                Exercises and Drills For TS Amendment #252 - PASS Elimination, this is one of the procedures to be changed to reflect the elimination of the requirement to perform a PASS sample or have and maintain the system.
PASS will be 'Abandoned In Place'.
Please contact Paul Sullivan, Manager of Emergency Preparedness at DAEC, (319)851-7191, if you require further information.
3313 DAEC Road - Palo, Iowa 52324-9785 Telephone: 319.851.7191 - Fax: 319.851.7364

:'    DAEC EMERGENCY PLAN                                 APPENDIX2 APPENDIX 2                                 Rev. 21 LETTERS OF AGREEMENT                                 Page 1 of 1 IES/NMC has agreements established with the agencies and firms listed. The agreements are on file in the Corporate Office of IES Utilities Inc. in the Alliant Tower in downtown Cedar Rapids and with the Emergency Planning department onsite at the DAEC.
I Bechtel Power Corporation Benton County Sheriffs Department City of Palo Palo Fire Department General Electric INPO Linn County Sheriffs Department Mercy Medical Center Disaster Services Division, Iowa Department of Public Defense (now Emergency Management Division)
University of Iowa Hospital and Clinics Benton-Linn Ambulance Service USCEA

            >;D.-   iAEC  EMERGENCY PLAN--                               -.: ;      TOC Rev. 27 TABLE OF CONTENTS' Page 1 of 2 Procedure #    Title                                                                       Rev#
Introduction ........................................                         20 Section A      Assignment of Responsibilities (Organizational Control) ........ I... 21 Section B       Emergency Response Organization ........................................ 26 Section C       Emergency Response Support and Resources ........................ 21 Section D       Emergency Classification System .........................                     22 Section E       Notification Methods and Procedures .........................                 20 Section F       Emergency Communications.....................................................22 Section G       Public Education and Information ........................................ 20 Section H.. Emergency Facilities Staffing, Activation and Equipment ......... 23 Section I       Accident Assessment ........................................                  21 Section J       Protective Response ........................................                  21 Section K       Radiological Exposure Control ........................................        20 Section L       Medical and Public Health Support ........................................ 20 Section M       Recovery and Re-entry Planning and Post-Accident Operation ........................................                            20 Section N        Exercises and Drills ........................................                21 Section 0        Radiological Emergency Response Training ............................. 20 Section P        Planning Responsibilities ........................................           20

                ;.':  DAEC EMERGENCY PLAN 2<"->i'                                    TOC Rev. 27 TABLE OF CONTENTS Page 2 of 2 Procedure #          Title                                                                  Rev#
Appendix 1           Cross-Reference Matrix NUREG-0654, Rev. I vs. DAEC Emergency Plan .......................................         20 Appendix 2           Letters of Agreement .......................................            21 Appendix 3           Linn and Benton Counties Evacuation Time Estimates Within the Emergency Planning Zone for the DAEC ................. 21 Appendix 4           Emergency Plan Implementing Procedures .............................. 20 Appendix 5           Matrix of Implementing Procedures vs.
DAEC Emergency Plan .......................................             20 Appendix 6           Definitions .......................................                     22 NOTE: Appendix 3, "Linn & Benton County Evacuation Time Estimates within the EPZ for the DAEC", has been removed from the main body of the DAEC Emergency Plan and added as an addendum to the DAEC Emergency Plan.

    .-         DAEC EMERGENCY PLAN-                                       SECTION 'B' Rev. 26 EMERGENCY RESPONSE ORGANIZATION Page 1 of 34 I Effective Date:      faz              I TECHNICAL REVIEW Prepared by-.                                           Date:        I 73/q  -
Reviewed b:                                             Date:        1) 6bc I dependent Reviewer.
P            .P .O.
PROCEDURE APPROVAL I    I D  .a.t
                                                                ;           -          1 Approved by                                             Date: 1/Z, :2/,o i<
Manager, Emergency Planning
                                                 ..   .   .       ., i   ,
Reviewed by:                                            Date:       ,,          L l
Approved by:                                  I . IDate:                      b4-Authorized by:                                           Date:.         2      q

DAEC EMERGENCY PLAN,                                                   SECTION 'B' Rev. 26 EMERGENCY RESPONSE ORGANIZATION Page 2 of 34 Table of Contents Pace 1.0 PURPOSE ......................................................                                    3 2.0 REQUIREMENTS .......................................................                              3 2.1  RESPONSE POSITIONS .........................................                                 3 2.2  ONSITE RESPONSE ASSIGNMENTS .........................................                        3 2.3  OFFSITE (EOF & JPIC) RESPONSE ASSIGNMENTS .................... 17 2.4  LONG-TERM ORGANIZATION ............                             ............................ 23 2.5  INTERFACES ........................................                                          25 2.6  LOCAL SUPPORT SERVICES ........................................                              26 3.0 ATTACHMENTS ......................................................                                27 TABLE B-1, 'ON-SHIFT STAFFING & STAFF AUGMENTATION ASSIGNMENTS"............................................................................... 28 FIGURE B-1, "ONSITE AND CORPORATE EMERGENCY RESPONSE ORGANIZATION" ...................................................... 31 FIGURE B-2, "IMMEDIATE RESPONSE INTERFACE" ..................... 33 FIGURE B-3, "LONG-TERM RESPONSE INTERFACE" ................... 34

17                   . DAEC EMERGENCY PLAN                   -           SECTION 'B' Rev. 26 EMERGENCY RESPONSE'ORGANIZATION-Page 3 of 34

==1.0 PURPOSE==
(1)   This section describes the structure of the Emergency Response Organization and the specific responsibilities and authorities of key response personnel.
Support services available with the Asset Owner and the Nuclear Management Company (NMC) to augment the Emergency Response Organization are identified as well as those services that may be provided by the Asset Owner, the NMC; contractors and local organizations.
(2)   The NMC is assigned operational responsibility for the DAEC. However, the Asset Owner maintains corporate accountability for,activities at the DAEC and will participate when necessary in activities at the DAEC. The reference "Asset Owner/NMC" will be used throughout this procedure to signify this relationship.
Further details regarding this relationship can be found in the "Nuclear Power Plant Operating Services Agreement" (NPPOSA) between the Asset Owner and the NMC.
                                   ;2.0     REQUIREMENTS 2.1       RESPONSE POSITIONS, (1)  The Emergency Response Organization is as illustrated in Figure B-I, and in the text of Section 'B' and Section 'H' of the DAEC Emergency Plan.
Personnel qualified to fulfill the emergency response positions are identified in the Emergency Telephone Book and in the DAEC Training Records with the DAEC Training Records being the most up to date. The Emergency Response Organization'structure can be modified as required by the Emergency Coordinator or Emergency Response & Recovery Director.
2.2       ONSITE RESPONSE ASSIGNMENTS,.
(1)   Immediate response organizational assignments, lines of succession, and responsibilities are as described below, for the Onsite organization.
(A)   Emergency Coordinator (i)     Assignment -       -      ;

DAEC EMERGENCY PLAN'                                    SECTION 'B' Rev. 26 EMERGENCY RESPONSE ORGANIZATION Page 4 of 34 (a)   The Emergency Coordinator functions onsite, coordinates the total site response effort, and normally operates from the Technical Support Center. The Emergency Coordinator reports to the ER&RD and has full authority and responsibility to initiate emergency actions and to recommend appropriate offsite protective measures to local and state authorities during the initial stages of the event as discussed in Section A.
(ii)    Lines of Succession (a)   The Operations Shift Manager/Supervisor (OSM/OSS) functions as the Emergency Coordinator until relieved. A qualified person will assume the responsibility of. the Emergency Coordinator and receives turnover from the OSM/OSS. If necessary, the ER&RD will appoint the position of Emergency Coordinator and will inform the Operations Shift Manager/Supervisor of the appointment.
(iii)   Responsibilities (a)   The Emergency Coordinator exercises full responsibility and authority for all activities at the site. This position is a 30-minute ERO response reporting position. He/she is assigned the following functional responsibilities:
: a. Ensure the activation of the onsite Emergency Response Organization as appropriate for the classification and circumstances of the emergency condition.
: b. Coordinate efforts to return the plant to and maintain it in a safe, stable condition.
: c. Coordinate accident assessment and analyses efforts to determine the full scope and impact of the emergency.
: d. Ensure appropriate initial notification of Asset Owner/NMC, local, state, and federal officials and agencies. This function will be assumed by the Emergency Operations Facility when activated.
: e. Provide initial Protective Action Recommendations, as appropriate, to local and state authorities who are responsible for offsite protective measures. This function will be assumed by the Emergency Operations Facility upon activation of that facility.

. I i                DAEC'EMERGENCY. PLAN                                  SECTION'B' EMERGENCY RESPONSE ORGANIZATION                                Rev. 26 Page 5 of 34
: f. Apprise Asset Owner/NMC, local, state, and federal officials
                      -and agencies of updated information pertaining to the
                    - emergency condition.'
: g. Classify/reclassify the event as necessary.
: h. Approve extensions on exposure limits for emergency workers, if necessary.
: i. Select alternate location for the Offsite Radiological and Analytical Laboratory/Offsite Decontamination Facility if radiological release and meteorological conditions warrant a change.
: j. Prepare the Emergency Response Organization for an orderly transfer of responsibilities to the recovery organization.
(b)    While the administrative aspects of most of these responsibilities may be delegated by the Emergency Coordinator, the responsibilities for items e, g, and h may not be delegated except as indicated herein. Upon operation of the EOF, the ER&RD assumes responsibility for these non-delegable duties.
(B) Operations Shift Manager and Supervisor (OSM/OSS)
(i)    Assignment (a)    The OSM/OSS, located in the Control Room, shall provide direction as required to return the plant to or assure that it is maintained in a safe, stable configuration.
(ii)   Lines of Succession (a)    Generally, the'Operations Shift Manager, the senior individual, assumes the role of Emergency Coordinator. However, the Operations Shift Supervisor will assume the role of Emergency Coordinator if the Operations Shift Manager is incapacitated.
In the event that both the Operations Shift Manager and Supervisor are incapacitated, their responsibilities will immediately be assumed by a Licensed Senior Reactor Operator, if available,,or by the Nuclear Station Operating Engineer. Should this situation occur during normal work hours, the Operations Manager or Supervisor, upon his arrival in the Control Room, will assume the responsibilities and

DAEC EMERGENCY PLAN                                   SECTION'B' EMERGENCY RESPONSE ORGANIZATION                                   Rev. 26 Page 6 of 34 authorities normally assigned to the Operations Shift Manager/Supervisor. If this situation should occur during other than normal work hours, the first licensed Senior Reactor Operator who reports to the site will assume the functional responsibilities assigned to the Operations Shift Manager or Supervisor. Subsequent relief of this individual will be as directed by the Emergency Coordinator.
(iii)   Responsibilities (a) The Operations Shift Manager/Supervisor evaluates the abnormal condition and implement emergency response actions as specified in the Emergency Plan Implementing Procedures (EPIPs) including:
* Classifying the event
* Recommending Protective Actions, if appropriate
* Notifying county, state, and federal officials and offsite support agencies as necessary
* Notifying the Emergency Response Organization (C) TSC Operations Supervisor (i)      Assignment (a) The TSC Operations Supervisor will proceed to the TSC after receiving a plant status briefing, as needed, from the OSM/OSS and will assist the TSC in all matters pertaining to the Control Room and Operations Department activities. In the event that the TSC Operations Supervisor is needed to stay in the Control Room for an extended period of time, the Emergency Coordinator shall assume these duties, assign collateral responsibilities to another ERO position, or appoint an alternate.
(ii)     Lines of Succession (a)    The OSM/OSS functions as the TSC Operations Supervisor until relieved by a qualified individual. If necessary, the Emergency Coordinator will appoint the TSC Operations Supervisor.
(iii)   Responsibilities

i I 1,              DAECEEMERGENCY PAN                     I'.'' .           SECTION 'B' Rev. 26 EMERGENCY RESPONSE ORGANIZATION                                   Pe 7f Page 7 of 34 (a) This position is a 60-minute ERO response reporting time position.
(b) The TSC Operations Supervisor is responsible for the following activities:
* Assist the TSC in all matters pertaining to the Control Room and to Operations Department activities.
Providing direction and assistance, as necessary, to the OSMI/OSS to achieve and maintain stable plant conditions.--
                     *      . Assisting the OSM/OSS in coordinating operational activities.
* Monitoring operational activities to-assure that the plant is operated and maintained in as safe a condition as possible.-
* Evaluating recommendations for corrective action
                           ' provided by the technical staff and operating crew and recommending to the Emergency Coordinator a course of action to be taken to mitigate the situation.
                   - - -Recommending              changes to the Emergency Classification based upon:
(i)      Plant status changes, with or without radiological releases in progress.
(ii) - -Actual or potential radiological release
                                    -      parameters.
(iii)    The progress of those activities undertaken to
                                       -   mitigate the situation and their probability for
                                         - success.-;
(D) Site Radiation Protection Coordinator, (i)    Assignment (a) The Site Radiation Protection Coordinator will operate from the TSC and initiate those activities related to radiological assessment of the environs surrounding the plant during the initial stages of the'event. Offsite monitoring will be assumed

             'DAEC EMERGENCY PLAN-                                 SECTION 'B' Rev. 26 EMERGENCY RESPONSE ORGANIZATION Page 8 of 34 by the Radiological Assessment Coordinator upon activation of the EOF.
(ii)   Lines of Succession (a)   The Operations Shift Manager/Supervisor functions as the Site Radiation Protection Coordinator until officially relieved by the Emergency Coordinator. A qualified Site Radiation Protection Coordinator informs the Emergency Coordinator that he/she is ready to assume that position's responsibilities. If necessary, the Emergency Coordinator will appoint the Site Radiation Protection Coordinator.
(iii)   Responsibilities (a) This position is a 30-minute ERO response reporting time position.
(b)   The Site Radiation Protection Coordinator is responsible for the following activities:
* Ensuring that DAEC personnel are dispatched to monitor the environs in and around the plant for radiological consequences associated with the event.
* Conducting an initial evaluation and assessment of the results of radiological monitoring activities. Upon activation of the EOF, evaluation and assessment of all offsite monitoring activities will be assumed by the Radiological Assessment Coordinator.
* Assessing the onsite radiological consequences and directing protective measures, including the need for partial or complete evacuation of the plant.
During the initial stages of the event, apprising local and state authorities, through the Emergency Coordinator, of the results of radiological monitoring activities and providing protective action recommendations based upon the projected radiological consequences to the population at risk. Upon activation of the EOF, this function will be assumed by the Radiological Assessment Coordinator.
(E) Security and Support Supervisor

I
                 -.DAEC   EMERGENCY-PLAN i". .                         *SECTION   'B''
Rev. 26 EMERGENCY RESPONSE ORGANIZATION' K':                                                                     Page 9 of 34 (i)     Assignment (a)     The Security and Support Supervisor will exercise supervision and direction of the security staff and direction over the personnel assigned to the TSC support staff.
(ii)  ,Lines of Succession (a)     If necessary, the Emergency Coordinator will appoint the Security and Support Supervisor.
(iii)    Responsibilities (a)    This position is a 30-minute ERO response reporting time position.
(b)    Upon activation of the TSC, the Security and Support Supervisor is responsible for:
Assuring that an accountability check for all personnel within the protected area is conducted in a timely fashion and that requisite security posts are filled.
Ensuring that the Emergency Response Organization notification process as described in the Emergency Plan Implementing Procedures has been initiated and is successfully completed.
* Assuring the TSC closed ventilation system is operational and activated.
                    .      Limiting access into the facility to only those personnel who are members of the Emergency Response Organization, or otherwise are authorized.
* Establishing measures that will enable continuous accountability for all personnel within the protected area
* Eonce the initial accountability check has been completed.
                     . -Ensuring that no unauthorized personnel gain access to the site.
                     . Assigning personnel for first aid duties, as required.
                     .      Providing overall management and direction to the support staff assembled in the TSC.

      -       DAEC EMERGENCY PLAN                                  SECTION 'B' Rev. 26 EMERGENCY RESPONSE ORGANIZATION Page 10 of 34 (F) Technical and Engineering Supervisor (i)    Assignment (a) The Technical and Engineering Supervisor will exercise supervision and direction over the personnel assigned to the Technical Support Center while the Emergency Coordinator is in the Control Room receiving a turnover. The Technical and Engineering Supervisor will exercise overall management and supervision of engineering, analysis and corrective action efforts undertaken by engineering and maintenance personnel at the DAEC from the TSC. In addition, he/she will coordinate with the Emergency Response Organization for engineering support efforts undertaken at the request of the Emergency Coordinator or TSC Operations Supervisor.
(ii)    Lines of Succession (a)   If necessary, the Emergency Coordinator will appoint the Technical and Engineering Supervisor.
(iii)   Responsibilities (a) This position is a 30-minute ERO response reporting time position.
(b)    Upon activation of the site Emergency Response Organization and the Technical Support Center, the Technical and Engineering Supervisor is responsible for:
* Verifying that the TSC is fully activated and staffed as described in the Emergency Plan Implementing Procedures.
* Evaluating plant status and providing support to the operations staff as requested.
* Assisting the Emergency Coordinator and the TSC Operations Supervisor in establishing the priority for repair activities to be undertaken.
* Providing direction to the technical staff comprised of support personnel such as the Fire Marshal, Safety Supervisor, and consultant/contractor representatives to analyze plant conditions and define courses of action to mitigate the emergency situation.

                             .. ':DAEC EMERGENCY PLAN"                                SECTION 'B' K>i                                                                                   Rev. 26 EMERGENCY RESPONSE ORGANIZATION',
Page 11 of 34 Providing direction to' the engineering 'staff in TSC to aid in analysis of plant conditions and define courses of action to mitigate the emergency situation.
                                  -'     Coordinating'corporate engineering activities with efforts
                                        'being taken at the DAEC to mitigate the event and establish'stable plant conditions.
Providing support to the OSC Supervisor as necessary for coordinating all repair/corrective action efforts conducted at the DAEC.
(G)     Reactor Engineer (i)          Assignment (a)      The" Reactor Engineer will provide support to the operations crew located in the Control Room, and to the Emergency Coordinator from the TSC. He/she will provide
                                -'recommendations'for returning the reactor core to a safe and stable condition.
(ii)         Lines'of Succession '             !
(a)       If necessary, the Emergency Coordinator will appoint the Reactor Engineer.
:. I     - ...      -Z                             . ....        . .
(iii)       Responsibilities
( a )            o . t .Ts - a   ;-   mRsr
                                                                  .    * -, I   r (a)      This position is' a .30-minute ERO response reporting position.
I(b)'      Upon activation -of the Emergency Plan for conditions
: -; I     I     ;-     classified as an ALERT or greater, the Reactor Engineer is responsible for:' '.'            - --      -
Supporting the operating crew in bringing the reactor core to desired condition and maintaining it there.
                                   .      Determining and reporting the amount of failed fuel to the TSC Operations Supervisor
                                   '      Obtaining vendor feedback on the amount of failed fuel.
* Recommending fuels-related priorities in recovery/re-entry op6erations.

DAEC EMERGENCY PLAN;                                SECTION 'B' Rev. 26 EMERGENCY RESPONSE ORGANIZATION Page 12 of 34 (H) Administrative Supervisor (i)     Assignment (a) The Administrative Supervisor will provide administrative and logistics support, as required, in the event that activation of the site Emergency Response Organization is required from the Technical Support Center.
(ii)    Lines of Succession (a)   If necessary, the Security and Support Supervisor will appoint someone to fill this position.
(iii)   Responsibilities (a)   This position is a 60-minute ERO response reporting time position.
(b)   Services to be provided under the direction of the Administrative Supervisor include, but are not limited to:
Clerical, typing and copying services.
* Document retrieval.
* Food services, clothing and overnight accommodations.
* Coordination of transportation services and any facilities or office space needs.
* Determining existing and potential administrative support needs and providing recommendations to the Security &
Support Supervisor, as required.
(I) (I)     Operational Support Center (OSC) Supervisor (i)     Assignment (a) The OSC Supervisor will exercise supervision and direction over the personnel who report to the OSC. He/she will report to the Emergency Coordinator in the TSC and will coordinate repair/corrective action efforts conducted at DAEC.

II I
                   -DAEC EMERGENCY PLAN                                 SECTION'B' Rev. 26 EMERGENCY RESPONSE ORGANIZATION                           '
Page 13 of 34 (ii)     Lines of Succession (a)   If necessary, the Emergency Coordinator will appoint the OSC Supervisor.
(iii)    Responsibilities (a) This position is a 30-minute ERO response reporting time position.
(b)   Upon activation of the Emergency Response Organization the OSC Supervisor is responsible for:
* Supervising the implementation of the tasks and staffing delineated by the Emergency Assignment Staffing Board.
* Providing general supervision and direction to personnel who report to the OSC.
* Coordinating evacuation from the site of all unnecessary personnel during events classified as a SITE AREA or GENERAL EMERGENCY, once such an evacuation has been authorized by the Emergency Coordinator.
* Coordinating all repair/corrective action efforts conducted at the DAEC to achieve stable plant conditions and to terminate any uncontrolled or excessive radiological release.
                       '*    Ensuring that personnel dispatched from the OSC are properly briefed and equipped for their assignment in regards to technical content, as well as ALARA, including existing and potential radiological hazards.
    '(J) Health Physics Supervisor        .
(i)  -  Assignments             I.-
(a) The Health Physics'Supervisor will provide overall direction and supervision in regards to ALARA and radiological practices to personnel in the OSC. In addition, he/she will provide direction to personnel for habitability of assembly areas.
(ii)     Lines of Successionr          ,

DAEC EMERGENC'(PLAN                                 SECTION 'B' Rev. 26 EMERGENCY RESPONSE ORGANIZATION Page 14 of 34 (a)   If necessary, the OSC Supervisor will appoint the Health Physics Supervisor.
(iii)   Responsibilities (a)   This position is a 30-minute ERO response reporting time position.
(b)   Upon activation of the Emergency Response Organization the Health Physics Supervisor is responsible for:
* Ensuring that personnel d ispatched from the OSC are properly outfitted with protective clothing and equipnilent, briefed regarding ALARA, and apprised of existing and potential radiological hazards.
* Coordinating with the Site Radiation Protection Coordinator to obtain information regarding plant status, problems, response options, significant radiological releases in progress, offsite dose rates, plume location and meteorological conditions as necessary.
* Ensuring the determination of habitability of assembly areas and ERO facilities.
* Coordinating the dispatch of monitoring teams and Offsite Relocation and Assembly Area (ORAA) and Offsite Radiological and Analytical Laboratory/Offsite Decontamination Facility (ORAIJODEF) personnel.
* Ensuring the ORAA is briefed periodically in terms of plant and radiological conditions.
* Supervising efforts to prepare injured/contaminated personnel for transport to offsite medical facilities.
(K) Instrumentation and Control/Electrical Maintenance (IC/EM) Supervisor, and Mechanical Maintenance Supervisor (i)     Assignment (a)   The ICEM Supervisor and Mechanical Maintenance Supervisor will aid in the coordination of repair/corrective actions conducted at the DAEC to achieve stable plant conditions from the Operations Support Center.

                 '      '.;DAEC     EMERGENCY PLAN                    .SECTION              'B' Rev. 26 EMERGENCY RESPONSE ORGANIZATION
<9                                                                              Page 15 of 34 (ii)         Lines of Succession
                   -  (a) , If necessary, the OSC Supervisor will appoint the IC/EM Supervisor and Mechanical Maintenance Supervisor (iii)        Responsibilities (a) -These positions are 60-minute ERO response time reporting
                    . .positions.
(b)     Upon activation of the Emergency Response Organization, the IC/EM Supervisor and Mechanical Maintenance Supervisor are responsible for:.
* Planning work packages for repair activities with assistance of engineering personnel in the TSC.
                              .*     Selecting personnel for repair teams appropriate to the work being done.
                              ' .Conducting      briefing and debriefings to repair team personnel;.
(2) Minimum Stafffing (A)  On-shift staffing and staff augmentation assignments are identified in Table B-1. The staffing plan is consistent with the guidance contained in NUREG-0654..Details regarding the assignments and associated responsibilities are addressed in the EPIPs.
(3) Other DAEC Organizational Assignments (A)   In addition to'the key response personnel described in the preceding sections, other DAEC personnel will assume roles as necessary in supporting the overall emergency response. Assignments and responsibilities of these support groups follow:
(i)          Security            .   ..     --.
                     '(a)      Upon activation of the plan, for events classified as an ALERT or greater,'the DAEC Security Force is responsible for performing an accountability check for all personnel within the protected area and controlling access to the site property. If evacuation of the site is required, the Security Force will assist in the evacuation and conduct an accountability check of all personnel dispatched to the ORAA. In addition, during other

DAEC EMERGENCY PLAN-..                               SECTION 'B' EMERGENCY RESPONSE ORGANIZATION                               Rev. 26 Page 16 of 34 than normal working hours, security force members may assist with initial notification of offsite agencies and will conduct notification of the Emergency Response Organization.
(ii)   Administrative (a) The DAEC Administrative Support Staff will provide general logistical support functions to aid response organization activities. This includes long range planning for providing personnel, material, facilities and office and clerical services.
Additionally, the administrative staff will coordinate warehouse and procurement activities and obtain life support services such as food, clothing, and overnight accommodations.
(iii)  Technical/Engineering (a) The technical and maintenance support staffs will provide plant engineering, maintenance assistance and coordination of corrective actions taken to mitigate the emergency condition, or terminate a release. This includes analytical and engineering efforts of site and corporate engineers and supervision and technical direction of activities performed by engineering, maintenance or construction crafts.
(iv)    Quality Assurance (a) Quality Assurance can provide assistance to design, procurement, and construction activities that are required to establish cold shutdown conditions. Quality Assurance can define and track activities that are not conducted in accordance with normal established practices, and can ensure that post-accident evaluations are conducted to verify the acceptability of those activities for both short and long term service.
(v)     Materials Management (Warehouse)
(a) The Warehouse Supervisor will provide warehouse and procurement support in the event that procurement of specialized parts/equipment not currently stored on-site, or available locally, is required.
: i. The site staff has the ability to procure site stores, and locally available stores, (equipment, parts, etc.) without involving the Warehouse Supervisor.

             ; -i.'          DAEC EMERG'ENCY PLAN              --               SECTION 'B' Rev. 26 EMERGENCY RESPONSE ORGANIZATION K> I            .                                                               Page 17 of 34 (vi)    Contracts and Agreements            -
(a) Assistance to the emergency response effort will'be available from the Nuclear Steam Supply System supplier (General Electric), Architect-Engineer (Bechtel) and other consultants.
Agreements and contracts are currently in place with each of these organizations which retain their services and define the commercial conditions of those services on a routine basis. In case of an emergency condition at DAEC, these private contractors could be contacted to augment the services they are currently providing. Agreemernts have been negotiated for emergency services as necessary. Section A of the DAEC Emergency Plan provides further details on the availability of contractors and consultants.
(vii)   Safety (a)   The Safety Specialist can advise' Corporate Management in the area of safety. When a personal injury accident occurs at DAEC, the Safety Specialist can provide investigative reports reflecting the events that led up to the accident. He/she can indicate whether safety rules and procedures were followed and recommend follow-up corrective/disciplinary actions.
Additionally, the SafetySpecialist is responsible for reporting all serious accidents to the Iowa Occupational Safety and Health Administration, a branch of the Iowa Bureau of Labor.
2.3     OFFSITE (EOF & JPIC) RESPONSE ASSIGNMENTS (1)  The Emergency Operations Facility and Joint Public Information Center provide the following principal functions-in the overall response to an emergency at the DAEC once control is transferred from the Technical Support Center:
(A)    Establishes a 'singlefo6cai point for performing radiological dose assessment and Protective Actio'n Recommendation decision-making, including coordination and interface with local, state, and federal support groups (B)    Establishes a coordinated means to disseminate information related to the accident to public officials, the news media, and industry public relations forums. ''.
(2)  Response positions are as follows:

           '   ' DAEC EMERGENCY PLAN '                      -         SECTION B' Rev. 26 EMERGENCY RESPONSE ORGANIZATION Page 18 of 34 (A)  Emergency Response and Recovery Director (ER&RD)
(i)       Assignment (a) The ER&RD is responsible for the overall direction and control of Asset Owner/NMC's integrated emergency response and recovery effort and providing the financial resources and contractual capabilities to ensure requisite actions can be taken to protect the health and safety of the public.
(ii)      Responsibilities (a)   Ensuring that the Emergency Response Organization is staffed by qualified personnel and coordinating with these personnel to ensure that sufficient support for various functions is available, either from within Asset OwnerINMC or from outside organizations (i.e., other utilities, Architect Engineers, Nuclear Steam Supply System suppliers, INPO, consultants, etc.).
(b)   Authorizing the procurement of equipment, materials, and resources, as necessary, to effectively respond, control, and recover from an accident condition at DAEC.
(c)    Implementing the Emergency Plan Implementing Procedures (EPIPs).
(d)    Reviewing and concurring with Protective Action Recommendations prior to their issuance once the Emergency Operations Facility (EOF) is operational.
(B)  Radiological and EOF Manager (i)      Assignment (a) The Radiological and EOF Manager is responsible for coordinating and directing all offsite radiological monitoring and dose assessment programs and supervising activities within the EOF. He/she will be accountable to the ER&RD.
(ii)      Responsibilities (a)   Establishing communications with the TSC as necessary.
Obtaining information on the diagnosis and prognosis of the accident condition, the quantities of radioactive material' releases and the prevailing meteorological conditions.

DAEC EMERGENCY PLAN.                                 SECTION 'B' Rev. 26 EMERGENCY RESPONSE ORGANIZATION Page 19 of 34 (b) Coordinating the onsite and offsite radiological monitoring activities to provide anticipated release rates and projected dose rates.
(c) Assisting and interfacing with county, state, federal and support agencies to relate accident information necessary for the offsite authorities'to implement their county and state emergency plans and procedures. '
(d) Interpreting radiological data for updating the ER&RD, county, state, federal and support agencies in terms of projected radiological exposures and actual dose measurements. This includes providing estimates of total population exposure when necessary. Providing radiation protection for those assembled attheEOF.         ;
(e) Analyzing all information for'significant trends while developing Protective Action Recommendations (PARs). Review the recommendations with the Emergency Response and Recovery' Director and, when authorized, provide them to the state and county officials.
(f)     Providing assistance to county, state, and federal officials for recovery activities.     -
(g) Advising the ER&RD to ensure compliance with legal and regulatory requirements.
(C) Radiological Assessment Coordinator (i)    The Radiological Assessment Coordinator is responsible for the following:
(a)' Directing and coordinating offsite monitoring teams.
(b) Performing dose projection calculations.
(c) Providing Protective Action Recommendations, as required, to the Radiological and EOF Manager.
         '(d)
* Coodrdinating the necessary support to other Agencies and support groups in field assessment, data analysis and
             -    environmental sample analysis.
(e) Coordinating with the State of Iowa in tracking the plume offsite.

DAEC EMERGENCY.PLAN                                   SECTION 'B' Rev. 26 EMERGENCY RESPONSE ORGANIZATION Page 20 of 34 (D) Support Services Coordinator (i)   The Support Services Coordinator will assure that necessary resources and activities are provided by staff personnel. Specific areas are available to support the overall emergency response and recovery effort conducted both at the Alliant Tower and at the site. This includes security, communications, personnel, transportation, purchasing, industrial relations, and safety. He/she will also coordinate, as necessary, the application of resources and equipment available within Asset Owner/NMC departments.
(a)    Security The Support Services Coordinator in cooperation with the Corporate Security Manager is responsible for providing security for the Alliant Tower and controlling access to the EOF, JPIC, and backup facility to the JPIC as well as appropriate Asset Owner/NMC working areas in the building. Staff assistance for building security will be provided by the security force under contract, and, to the extent possible, the Cedar Rapids Police Department and the Linn County Sheriffs office. Further details and instructions for establishing and maintaining security are provided in the EPIPs.
(b)    Communications The Palo Cooperative Telephone Association and US West can provide engineering support for commercial telephone installations that are permanent or temporary.
He/she can respond to identified communication needs and problems and coordinate necessary engineering and maintenance support for restoration or rearrangement of Asset Owner/NMC communications systems.
(c)    Corporate Services The Corporate Services Department can provide required services and equipment such as record and document retrieval and reproduction, office supplies, office furniture, photography, facility and area maps, audio visual aids, graphics, printing, distribution services, and general housekeeping services.
(d)    Personnel

DAEC-EMERGENCYPLAN                                  SECTION 'B' Rev. 26 EMERGENCY RESPONSE ORGANIZATION Page 21 of 34 Administrative personnel can be contacted to provide personnel to augment the administrative and clerical support functions associated with initial activation and continued operation of the EOF and JPIC.
(e)  Logistics and Transportation Transportation personnel can respond'to identified transportation needs and emergency air and land transportation necessary for materials and personnel.
He/she can arrange for rental cars and hotel accommodations for those personnel temporarily assigned to the Cedar Rapids area in support of the response and recovery effort. They can also provide for repairing and maintaining a transportation fleet and implementing contracts with commercial carriers to obtain priority transportation.
(f)  Purchasing Purchasing and Materials personnel can respond to identified needs related to procurement of materials and services and coordinate onsite and offsite procurement activities to assure rapid delivery of materials. They can augment Asset Owner/NMC resources by activating contracts with outside agencies and requesting, through use of prepared lists, emergency equipment available in Cedar Rapids.,--,
(g)  Industrial Relations The Manager, Industrial Relations can assess and respond to contractual problems that may arise during the course of the event and apprise bargaining unit officials and trade counsels of existing or projected labor problems.- He/she can also assist in the response to manpower,needs through the use of established manning lists,- a computerized skill inventory of Asset Owner/NMC employees, and the activation of
            ;--.:.7established'contracts/agreements with outside organizations.    -       -
(h) Supplemental Resources

DAEC EMERGENCY PLAN-                                SECTION 'B' Rev. 26 EMERGENCY RESPONSE ORGANIZATION Page 22 of 34
* The Operations and Production Departments of Asset Owner/NMC can augment the DAEC staff during an emergency. Personnel and equipment are available to provide maintenance and construction services at the DAEC. Materials, equipment, and machine shop services are also available.
(E) EOF Ops Liaison (i)     The EOF Ops Liaison is responsible for the following:
(a)   Advising the ER&RD on Emergency Action Levels (b)    Providing an operational insight and tracking plant status (c)   Assisting in the recovery phase of the emergency (d)   Advising the ER&RD on matters that pertain to the plant (F) Joint Public Information Center (JPIC) Manager (i)     The JPIC Manager is responsible for-ensuring that accurate and timely information is provided to the public and the news media; coordinating press statements and news media briefings with local, state and federal public relations officials; and coordinating information with the Corporate Communications Department.
His/her specific responsibilities include:
(a)    Initiating notification of the Asset Owner/NMC Emergency Public Information Organization and determining the extent to which the Joint Public Information Center (JPIC) will be activated.
(b)    Directing activities at news conferences.
(c)    Coordinating the release of all information prepared by the JPIC and ensuring that it receives concurrence from the DAEC Spokesperson or hislher designee as being technically accurate prior to its release.
(d)   Providing overall direction and coordination of all emergency response activities conducted by the Asset Owner/NMC personnel in the JPIC.
(e)   Providing Asset Owner/NMC departments and employees with information consistent with that released to the media.

                       .'-DAEC 'EMERGENCYPLAN                     '.i         SECTION 'B'
              .    ~~~-           -                                - ..... :.           .....{ ......-
                                                                                               .-A, Rev. 26 EMERGENCY RESPONSE ORGANIZATION Page 23 of 34 2.4     LONG-TERM ORGANIZATION                -.
(1) Activation of the onsite Emerge-ncyResponse Organization will, directly and indirectly, result in a response by essentially all personnel normally associated with the DAEC, particularly by management and key support personnel.
Therefore, within several hours after the initiating event, decisions will be made to provide and prepare for a long term augmented emergency organization.
The Emergency Response and Recovery Director will determine when the Recovery Organization is to be implemented. Prior to implementation of the Recovery Organization, the situation may require that the onsite Emergency Response Organization remain in place-for a protracted period of time. As conditions allow, shift schedules will be developed by the supervisors in charge of the Control Room and several supp6rt centers. Prior to implementation, these schedules shall be'reviewed and concurred with by the Emergency Coordinator. The Emergency Response Organization is set up on a team concept for 24-hour coverage.
(2)  The Asset Owner and the NMC jointly maintain the resources and capabilities to support response and recovery activities in the event of an emergency or accident condition at the DAEC. These include, but are not limited to, the following:
                            . Management direction and control
* Corporate and government affairs
* Public information and public relations
* Communications systems
* Security'dand edministration
* Medical and first aid
* Transportation and accommodations
* Commissary and catering
* Purchasing and stores
* Construction, maintenance, and mobile equipment and materials
* Appropriate staff and work force augmentation

DAEC EMERGENCY PLAN                             SECTION 'B' Rev. 26 EMERGENCY RESPONSE ORGANIZATION Page 24 of 34 Engineering activities
: i. Nuclear fuel, core physics, and thermal hydraulics ii. Design and construction iii. Electrical, instrumentation, mechanical iv. Chemistry and metallurgy
* Planning and scheduling
* Radiological analysis and protection
* Accident analysis
* Meteorological monitoring and analysis Health physics and decontamination
* Fire Protection
* Procedure development
* Operations and maintenance
* Quality assurance and control
* Contracts and agreements
* Company records and files
* Safety

                             ;" EMERGENCY PLAN :
                          ' DAEC                                                  SECTION 'B' EMERGENCY RESPONSE'ORGANIZATION                         -'       Rev. 26 Page 25 of 34 2.5     INTERFACES (1)   Figures B-2 and B-3 illustrate the immediate and long-term response interfaces, respectively, between the Emergency Response Organization and organizations of affected local, state and federal offices and agencies. The TSC, OSC, EOF, and JPIC are described and discussed in Section H.
(A)    Public Information and Governmental Relations (i)     During an emergency situation, the Joint Public Information Center (JPIC), located on the sixth floor of the Alliant Tower, can provide timely and accurate information to the news media and to public officials. The sixth floor contains an auditorium, conference rooms, kitchenette, and restrooms. The rear auditorium will be a work area for the media. This area will be equipped with tables and telephones., News briefings and conferences will be held in the front auditorium Conference rooms will be used as working areas for news center personnel. If the Cedar Rapids/Marion metropolitan area is evacuated, JPIC spokespersons and appropriate support staff can be relocated to the Alliant Energy Hanger at the Eastern Iowa Airport to continue media briefings and news conferences.
(ii)    Press Briefings and Public Relations (a). The Joint Public Information Center will function as the principal focal point for distribution of information to the public regarding the emergency condition at the DAEC. Press briefinigs will be coordinated by the JPIC Manager, who will ensure that appropriate emergency response and corporate individuals are available to provide technical information and respond to inquiries from the assembled media personnel.
Information related to the plant as well as generic information related tothe nuclear industry will be available to the media.
                             'Further details and instructions related to press briefings and public relations are provided in the EPIPs.
(iii)   Apprising Public'Officials'and Agencies             -
(a)    The facilities in the JPIC'will be used, as appropriate, following the issuance of Protective Action Recommendations (PARs),
for follow-up discussions and briefings of government officials and industry spokespersons on the status of the event, actions being taken, and evaluations assessing the impact upon the public.

                          ..DAEC EMERGENCY PLANW-                              SECTION 'B' Rev. 26 EMERGENCY RESPONSE ORGANIZATION Page 26 of 34 (B)     Corporate Assistance (i)     Legal Counsel (a)   Legal Counsel can provide the ER&RD with advice to prevent Asset Owner/NMC from taking actions that could increase corporate liability or jeopardize indemnification agreements when handling claims and litigation.
(ii)    Insurance and Claims (a)   Insurance and Risk personnel can advise the ER&RD in the area of insurance and claims, and provide them with regular status reports on the injured or contaminated individuals treated at nearby medical facilities. They interface with American Nuclear Insurers and can apprise them of the details, the sequence of events, the impact of the emergency, and the actions being taken to mitigate its consequences.
They also interface with Nuclear Electric Insurance Limited and Nuclear Mutual Limited (NEILINML), a utility-owned captive insurance group, which covers the loss of generation and coordinates claims filed on behalf of Asset Owner/NMC.
(b)   Insurance and Risk personnel will coordinate with nearby medical facilities and backup medical facilities at the University of Iowa Hospitals and Clinics, as required, in the treatment of radiological and non-radiological injuries. Insurance and Risk personnel can also complete all insurance forms and document all events affecting insurance and claims during the emergency.
2.6     LOCAL SUPPORT SERVICES (1)  Agreements have been reached with local agencies and private support facilities with regard to the type of support that will be furnished to the DAEC in the event of an emergency. These agreements in the form of letters of agreement (refer to Appendix 2 for list of letters) and agreements of responsibility as described in the Linn and Benton County Radiological Emergency Response Plans have been developed to ensure that there is a clear understanding of assigned responsibilities and that there will be proper coordination of activities in the event of an emergency. The Letters of Agreement will be updated as necessary and confirmed as acceptable at least every two years.

V
           ..-.--. ,:,:. ,.:T.
                            -. DAEC.EMERGENCY.PLAN;..-.               SECTION 'B'
                                                            .  .Rev.      26 EMERGENCY RESPONSE ORGANIZATION                      Pe 27 0Page   A                            27 of 34
                                   -3.0   -'ATTACHMENTS (1) Table B-1, uOn'Shift Staffing & Staff Augmentation Assignments" (2) Figure B-1, "Onsite Emergency Response Organization" (3) Figure B-2, "Immediate Response Interface" (4) Figure B-3, "Long-Term Response Interface" K>,
7 I,
i II i
                                            -  I
                                        . I   .

DAEC EMERGENCY PLAN EMERGENCY RESPONSE ORGANIZATION IRev.
SECTION 'B' Rev. 2 26 Page 28 of 34 Table B-i On-Shift Staffing & Staff Augmentation Assignments Major Functional Areas                Major Tasks                       Position Title or Expertise             On-Shift     Capability for Rsponse                      Comments (All positions are 24-hour staffing)               see      additions    Location Comments 30           60 mRa         min Plant Operations and      Respond to condition and         Operations Shift Manager (SRO)                                                     CR     Provides early direction and control until assessment or operational mitigate operational event       Operations Shift Supervisor (SRO)                                                         relieved by the Emergency Coordinator (Plant aspects                   consequences                     Control Room Operators                                  2                                 Manager - DAEC or designee).
Auxiliary Operators                                     2 Shift Technical Advisor Emergency Direction and   Site utility Emergency           Plant Manager - DAEC
* I                TSC    Assumed by the Operations Shift Manager Control                   Management                       Manager Outage and Support, or selected                                                   /Supervisor until relieved by the Plant Senior Plant Supervisory Personnel                                                        Manager, DAEC or designee.
Notificatsoa/Communi-     Notify licensee, state, local &                                                           I        I            2 CRJTSC   Performed by on-shift Operations, and cations                   federal personnel and                                                                                                      Security Force Personnel, or designated maintain conmmunications                                                                                                   Emergency Pemonnd.
Radiological Accident     Overall utility Emergency        Vice President, Nuclear or Selected                                           1*   EOF    One-hour staffing commitment may be filled Assessment and Support of Management and offshit           Management Personnel                                                                      by the Emergency Coordinator in the TSC.
Operational Accident      agency interface Assessment Oftsite Dose Assessment and      Radiation Protection Manager or selected                          I               TSC     Staffed by Site Radiation Protection Protective Action                Radiation Protection Supervisory Personnel.                                               Coordinator Recommendations Ofrsite Surveys                  I IP Technician                                                  1*           1*  OSC     May be staffed by Plant Personnel trained in DAEC Staff Member                                                 I            I          the IIP role for Field Teams.
Onsite Surveys                   HP Technician                                                    1*               OSC     May be staffed by Plant Personnel trained in (Out of plant)                  DAEC Staff Member                                                                          the IHP role for Field Teams.
Page I Subtotal                    l        8      l 6            6C
(                                                                                           (                                                                         (

DAEC EMERGENCY P                                                                                                  SECTION 'B'                                                                                                                                                                     9                            r Rev. 26
                                          .EMERGENCY RESPONSE ORGANIZATION Page 29 of 34 Table B-1 On-Shift Staffing & Staff Augmentation Assignments Major Functional Areas              MajorTasks                       PositionTitle or Expertise              On-Shift                 Capability for          Response                        Comments (All positions are 24-hour staffing)
* see                     additions             Location Comments 30             60 min           min Inplant Surveys                HP Technician                                             i                   I              I          oSC ChemistrylRadio-               Chern Technician                                                                                        OSC       On-Shift staffing provided by Radwvaste until chemistry                                                                                                                                        relieved by Chem Technician        -
Plant System Engineering Technical Support              Core/Thermal Hydraulics                                                                                 TSC       Position filled by a Reactor Engineer Repair and Corrective                                   Electrical Engineer                                                           I     ..      I    .... TSC Actions                                                 Mechanical Engineer                                                                                     TSC Repair and corrective actions  Mechanical Maintenance
* oSc       On-shift staffing may be provided by shift personnel assigned other functions.
:Electrical Maintenance                                        .
:*   .    . _       I         I  .        OSC      On-shift staffing may be provided by shift personnel assigned other functions I .
Instrument and Control (i&C) Technician                                                                 OSC Protective Actions       Radiation Protection:          HP Technicians A) Access Control                                                                                                          I           OSC B) HP Coverage for Repair,                                                                                                             oSC Corrective Actions, Search &
Rescue, First aid, & Fire fighting C) Habitability                                                                                              I                        OSC D) Dosimetry                                                                                                            .I                      Health Physics Access wvindov assigned this function.
I -    . . . __ __ ____I Page2Subtotal              --    - -             . I- . 22           1.      66     1      88       1 Pae S boa
:              1,

DAEC EMERGENCY PLAN                                                                        SECTION 'B' EMERGENCY RESPONSE ORGANIZATION IRev. 26 Iev.       2 Page 30 of 34 Table B-I On-Shift Staffing & Staff Augmentation Assignments Major Functional Areas              MajorTasks                  Position Title or Expertise             On-Shift  Capability for   Response                      Comments (All positions are 24-hour staffing)
* see       additions     Location Comments 30            60 min          min Fire Fighting
* Local Support    Fire Brigade per Technical Specifications.
Rescue Operations and
* Local Support    May be provided by shift personnd assigned First Aid                                                                                                                                     other functions.
Site Access Control and  Security, fire fighting,                                                                                             All per Security Plan.
Personnel Accountability Communications, personnel accountability Page I     Subtotal                                    8      6             6 Page 2     Subtotal                                    2      6            8 Page 3      Subtotal                                   0      0             0 GRAND TOTAL                                           10      12           14 C                                                                        (

(.                                                             c                                                    c DAEC EMERGENCY PLAN.-                                     SECTION 'B' EMERGENCY RESPONSE ORGANIZATION                              Rev. 26 Page 31 of 34 Figure B- I ONSITE EMERGENCY RESPONSE ORGANIZATION (Page I of 2)
[
7r7I
 .. ISupeviorj                 Manager
     -a.r- - .-.- ;.-. :1 - -'  .-- .- _
                                                                                                                        .I
* Joint PositMon with EOF

DAEC EMERGENCY PLAN,                                SECTION 'B' EMERGENCY RESPONSE ORGANIZATION                        Rev. 26 Page 32 of 34 Figure B-1 Corporate Emergency Response Organization (pg. 2 of 2)
Emergency       I Response and I Recovery Director]
                        -     ;t. g Rad& EOF         EOF/OPS Liaison Manager Logistics Coordinator
* Joint Position with TSC                            County/State Technical Liaisons Support Services Coordinator
(                                (

   -DAEC EMERGENCY PLAN     ::-. .  - SECTION 'B' EMERGENCY RESPONSE ORGANIZATION       Rev. 26 Pagee33 of 34.
FIGURE B-2 IMMEDIATE RESPONSE INTERFACE

                                      ,,,        a.
                                          -. DAECEMERGENCYPLN^:
SECTION 'B' EMERGENCY RESPONSE ORGANIZATION              Rev. 26 Page 34 of 34 FIGURE B-3 LONG-TERM RESPONSE INTERFACE EMERGENCY RESPONSE                                       EMERGENCY RESPONSE ORGANIZATION (OFFSITEI                                   ORGANIZATION IONSITEl

DAEC EMERGENCY PLAN                                         SECTION 'C' Rev. 21 EMERGENCY RESPONSE SUPPORT AND RESOURCES Page 1 of 7 Effective Date:   21sjOq TECHNICAL REVIEW Prepared by:                                                        Date: 1 ZZ3(o Reviewed by.                                                        Date: //likq pendent Reviewer             -
PROCEDURE APPROVAL Approved by:        ____                                           -Date:
Manager, Emergency Planning
                    ..       ..      .~~~~.      . .. . .. ......
Reviewed by:                 \ I__                                  Date: /IZ)   /"'
:-Operations Co         e Chairman Approved by:                                                        Date:
iAanager,t             Nuclear Authorized by:                                                       Date:  7141pl l     W6e President, Nuclear

DAECEMERGENCY                                          SECTION 'C' Rev. 21 EMERGENCY RESPONSE SUPPORT AND RESOURCES Page 2 of 7 Table of Contents Paae 1.0   PURPOSE .................................................                   3 2.0   REQUIREMENTS ..................................................            3 2.1  FEDERAL SUPPORT ................................                       3 2.2  NON-IES/NMC STAFFING OF THE EMERGENCY OPERATIONS FACILITY .................................                  5 2.3  RADIOLOGICAL LABORATORY SUPPORT ................................. 5 2.4  IES/NMC STAFFING OF OFFSITE GOVERMENTAL EMERGENCY OPERATION CENTERS (EOC'S) .................................            5 3.0   ATTACHMENTS ................................................                6 FIGURE C-1, "TYPICAL ORGANIZATION OF THE NRC SITE TEAM"................................................             7
: s I  'DAEC EMERGENCY PLN,                -" i>&#xa2;4-       SECTION 'C' Rev. 21 EMERGENCY RESPONSE SUPPORT AND RESOURCES                                           -
Page 3of7
                                          -1.0O;; PURPOSE' (1)  This section describes the support and resources which IES and the Nuclear Management Company (NMC) can expect from local, state, and federal governments and local support'agencies to augment their capabilities. Figure C-1 outlines the typical organization of the NRC Site Team.
(2)  The NMC is assigned operational responsibility for the DAEC. However, LES maintains corporate accountability for-activities at the DAEC and will participate when necessary in activities at the DAEC. The reference "IES/NMC" will be used throughout this procedure to signify this relationship.
Further details regarding this relationship can be found in the "Nuclear Power Plant Operating Services'Agreement" (NPPOSA) between IES and the NMC.
2.0     REQUIREMENTS 2.1     FEDERAL SUPPORT                    " ;
(1)  The federal government maintains in-depth capability to assist IES/NMC, the State of Iowa, and local governments through the Federal Radiological Emergency Response Plan and the'Federal Response Plan. Requests for federal assistance will be madelfor the State of Iowa through the Federal Emergency Management Agency or directly through the Iowa Department of Public Health to the Nuclear Regulatory Commission, Environmental Protection Agency, Department of Energy or the Food and Drug Administration as the situation may warrant their'response. These agencies can provide support for emergencies beyond the capabilities of IES/NMC and the local and state governments, primarily in the areas of radiological monitoring and sampling. This:support would be'of importance during a protracted event or one that involved a release of largeamtounts of radioactivity. When deemed
             -appropriate, asdescrib&d above, lES/NMC will recommend that theWState of Iowa request and coordinate the'support of these federal'agencies.
(2)  NRC response modes are described in"RCM-96 Section Q, "Concept of Operations" and are generalized as:
(a)   Standby: The situation'is 'sufficientlycomplex or dncertain to require additional monitoring a6nd preparations to' increase the NRC response quickly should it prove to be necessary. The appropriate NRC regional office directs response activities.'

DAEC EMERGEN                                         SECTION 'C' Rev. 21 EMERGENCY RESPONSE SUPPORT AND RESOURCES Page 4 of 7 (i)      The above mode involves only a few NRC specialists because the significance of the safety concerns immediately evident is not yet clear. Other federal organizations are notified but are not directly involved. State and local authorities, who will have been notified by the licensee, may call the NRC for an interpretation of the event and response.
(b)  The following modes, which indicate that an event with clear safety concerns is underway, usually involve direct participation by other federal organizations.
(i)      Initial Activation: The event calls for the NRC to dispatch a management and analysis team to the site. During this mode, the response to an event is directed by the NRC Executive Team from the NRC Operations Center at headquarters in Rockville, Maryland, while an NRC team is on the way to the site.
(ii)     Expanded Activation: The NRC management and analysis team is operational at the site and is now called the Site Team. During this mode, the entire NRC response is directed from the site by the Director of Site Operations with operational authority delegated by the Director of the Executive Team. Other NRC teams act in support of the Director of Site Operations.
(iii)    Deactivation: A plan for follow-up activities, including reentry and recovery, is in effect and the NRC is reducing its role to that outlined in the plan with the concurrence of the Federal Coordinating Officer and the State Coordinating Officer.
(3) Provisions of the Federal Radiological Emergency Response Plan come into play when response activities of more than one federal organization need to be coordinated; in addition, the broader Federal Response Plan will take effect.
These overlaid federal plans neither delay nor change the thrust of NRC response activities when they take effect, but they do entail additional coordination to provide the best possible federal support to state and local authorities.
(4) The NRC maintains the flexibility needed to meet both the changing requirements of an emergency and the varying needs for working with other organizations by first, defining the functions that may need to be performed and second, training teams of functional specialists who can be deployed as the situation requires.

                     -DAEC     EMERGENGY.P-N                               SECTION !C' Rev. 21 EMERGENCY RESPONSE SUPPORT AND RESOURCES, Page 5 of 7 2.2     NON-IES/NMC STAFFING OF THE EMERGENCY OPERATIONS FACILITY (1)   Provisions exist to accommodate representatives from both Linn and Benton Counties and the State of Iowa at the Emergency Operations Facility.
Additionally, facilities have been made available for use by NRC and FEMA representatives.
2.3     RADIOLOGICAL LABORATORY SUPPORT (1)   In addition to laboratory facilities available at the DAEC, the Offsite Radiological and Analytical Laboratory/Offsite Decontamination Facility (ORALJODEF) is located at 1017 12th Avenue SW in Cedar Rapids, approximately ten miles from the DAEC. Equipment installed in the ORAL enables analyses of environmental, atmospheric, soil, vegetation, and water samples to evaluate the presence of significant radionuclides.                    I (2)   In addition to these two laboratories, several other sources of laboratory support exist. These include the following:
(a)   Laboratory facilities of IES/NMC's environmental consultant for the analysis of environmental sample media (b)   The laboratory facilities from neighboring utilities as invoked through the INPO Fixed Facility Agreement (c)   The University of Iowa Hygienic Laboratory as coordinated through the Emergency Management Division, Iowa Department of Public Defense (d)   The NRC Mobile Van as coordinated with the NRC Region IlIl Site Team Leader.
2.4     IES/NMC STAFFING OF OFFSITE GOVERMENTAL EOC'S (1)   IES/NMC will prepare to send a utility representative to selected offsite governmental Emergency Operations Centers (EOC) as outlined in the Emergency Telephone Book (ETB), Section A-1.

                .;MIAEC;EMEROENCYPLAN%                      SECTION 'C' Rev. 21 EMERGENCY RESPONSE SUPPORT AND RESOURCES Page 6 of 7 3.0    ATTACHMENTS (1) Figure C-1, "Typical Organization of the NRC Site Team"

:;                             AEMERGENCYPAN                                                             SECTION'C' Rev. 21 EMERGENCY RESPONSE SUPPORT AND RESOURCES Page 7 of 7 FIGURE C-1 TYPICAL ORGANIZATION OF THE NRC SITE TEAM (LEAD DURING EXPANDED ACTIVATION)
                                                             ;l                  DIRECTOR OF SITE OPERATIONS DSO)
___ ___    I Dashed bordered boxes indicate                ll If-       _    Critical Positions. Critical positions                                    EOF                         I
    '     i      constitute the mini mum staff necessary L  -_   _   to perform the critJcaI NRC response             -      -     -    -   -     -   -     -     -    -
    *        ~   functions preparat tor to A..Ln-Ibn of (OSO) a Director of Site                  'prtin 4Snecialit LIH2E]
F1 Jflicalo EornSCsc I-HsPN Note 1:      Either the PMCL Communicatoror the HPN Communicator win be a critcal position. not both.
Note 2:      Either the ENS Communicator or the RSCL Communicator wig be a critical position, not both.
Note 3:      SafeguardstSecurity Assistant Coordinator Is part of lhe Inilal Ste Team only In safeguardshecurfty type events.

,..':;:.       .    .DAEC EMERGENCY.PLAN.                           ;SECTION'I'
                  .      ,-Rev. ,                                               21 ACCIDENT ASSESSMENT.                                       21 Page 1 of 17 Effective Date:
TECHN ALREVIEW Prepared by   -                                          Date Reviewed                                                   Date: ________
Independent Reviewer, PROCEDURE APPROVAL                                  -
App rved by:    PwcJ                                       Date:              a Manager, Emergency Planning'       -7' Reviewed by: /(JmV\                        -              Date:     2-:3o0'-
Operations C itee          Chairman Approved by:                                              Date:
a       hManager, Nuclear Authorized by:                                               Date:
V     President, Nuclear

DAEC' EMERGENCY PLAN -:-                                                 SECTION 'I' Rev. 21 ACCIDENT ASSESSMENT Page 2 of 17 Table of Contents Page 1.0     PURPOSE .............................                                                                      4 2.0     REQUIREMENTS ..............................                                                                4 2.1   CHARACTERISTIC PLANT SYSTEM AND EFFLUENT PARAMETER VALUES............................................................................................4 2.2   ACCIDENT ASSESSMENT CAPABILITIES AND RESOURCES ..... 4 2.3   RADIOLOGICAL SOURCE TERM AND MAGNITUDE DETERMINATIONS..........................................................................9 2.4   EFFLUENT MONITOR READINGS VS. EXPOSURE AND CONTAMINATION LEVELS .                        ................................................ 9 2.5   METEORLOGICAL INFORMATION ACQUISITION/EVALUATION. 10 2.6   RELEASE RATE/PROJECTED DOSE METHODOLOGY FOR OFF-SCALE OR INOPERABLE INSTRUMENTS ..................................... 10 2.7   FIELD MONITORING ................................................                                     10 2.8   RADIOLOGICAL HAZARD ASSESSMENT ...................................... 11 2.9   DETECTION AND MEASUREMENT OF RADIOIODINE CONCENTRATIONS .........                            ....................................... 11 2.10  RELATIONSHIP OF MEASURED PARAMETERS TO DOSE RATES.............................................................................................. 12 3.0     ATTACHMENTS                          ..                             ....................................... 12 (1)    TABLE 1-1, "AREA RADIATION MONITORS". ......................................... 13 (2)    TABLE 1-2, "PROCESS RADIATION MONITORS"................................... 15

            'DAEC     EMERGENCY PLAN <;                                                                                     K,.      SECTION I Rev. 21 ACCIDENT ASSESSMENT-Page 3 of 17 (3) TABLE 1-3, "PROCESS RADIATION MONITORS FOR HIGH RANGE EFFLUENT MONITORING SYSTEM".........................................                                                                               16 (4) FIGURE 1-1, "DAEC EMERGENCY PLANNING ZONE"................................. 17
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     -            ;.        DAECGEMERGENCY PLAN .                                      id>'  SECTION' Rev. 21 ACCIDENT ASSESSMENT Page 4 of 17 1.0          PURPOSE (1)   This section describes the methods, systems and equipment currently available for assessing and monitoring actual or potential offsite consequences of a radiological emergency condition.
(2)   The Nuclear Management Company (NMC) is assigned operational responsibility for the DAEC. However, IES maintains corporate accountability for activities at the DAEC and will participate when necessary in activities at the DAEC. The reference "IES/NMC" will be used throughout this procedure to signify this relationship. Further details regarding this relationship can be found in the "Nuclear Power Plant Operating Services Agreement" (NPPOSA) between IES and the NMC.
2.0           REQUIREMENTS 2.1         CHARACTERISTIC PLANT SYSTEM AND EFFLUENT PARAMETER VALUES (1)   Table D-1 identifies plant conditions, parameters, and potentially hazardous occurrences in the environment which enable definition of the emergency classification. Instrumentation, equipment status and parameter values associated with each condition are included in the EAL Tables located in the EPIP's.
2.2         ACCIDENT ASSESSMENT CAPABILITIES AND RESOURCES (1)   Systems or equipment which will be available during the course of an event to monitor and assess the magnitude of an actual or potential radiological release at the DAEC include the following, each of these is further described in the following paragraphs:
* High Range Effluent Monitoring System
* Containment High Range Radiation Monitoring System
* Area Radiation Monitoring System
* Low Level Iodine Sampling and Analysis Equipment
* Dose Projection Program (A) The Kaman Effluent Monitoring System consists of 11 monitor units installed in the Turbine Building vent stack, three Reactor Building vent

:-,.-:.... ,DAEC EMERGENCY PLAN'                      '           SECTION 'I'
                                                                    *Rev. 21 ACCIDENT ASSESSMENT '
Page 5 of 17 stacks, Off-gas stack, and LLRSF vent. The Normal Range Monitor is capable of detecting gaseous'activity ranging from 5 x 107 'p.Cilcc to
:1 x 10-2 jiCilcc using a P scintillation detector.
        -(i)     Each of the monitors has particulate and iodine collectors for laboratory sample analysis as well as a means for obtaining a gas grab sample.
(ii)    The five Accident Range Monitors, consist of two Geiger-Mueller detectors within a shielded sample chamber and is capable of detecting gaseous activity ranging from I x 10.2 piCi/cc to 1 x 105 ptCi/cc. Each monitor, as well, includes three particulate/iodine shielded collection assemblies and associated Geiger-Mueller detectors.' Sample collecticin is automatically initiated within one assembly, shifting to the next upon reaching a pre-set radiation level and minimum set amount of time until all assemblies reach the maximum radiation levels at which point the last assembly will
-  .             continue to collect although assembly has met saturation
                -conditions.' Technicians can collect filter media and reset collection.                  -
(iii)   One microcomputer is provided for each radiation monitoring unit and provides for complete control over the monitor. Pulse inputs from the detectors are converted into counts per minute and based upon the sample flow rate is displayed in Ci/cc. Release rate calculations can be made using existing vent stack flow rate'
                 -monitoring instrumntation.' The microcomputer also calc'ulates average radiation levels over a 1-minute, 10-minute, hour, and hour period.Theseaverages can be displayed for the last 30 periods-calculated; e.g., the last thirty 1-minute periods, 10-minute periods, etc.    '
I

          ;-DAEC. EMERGENCY PLAN                              SECTION 'I' Rev. 21 ACCIDENT ASSESSMENT Page 6 of 17 (iv)     Control and readout of units can be exercised from the Control Room and the Chemistry Laboratory through a minicomputer and associated peripheral equipment. Color CRT displays are available in both locations while a logger and CRT display printer are available in the Chemistry Laboratory.
(v)      Alarm functions provided include alert and high level alarms, rate of change alarms, and equipment failure alarms. Automatic control functions provided include check source activities, purging, and sample flow control.
(B)  The High Range Containment Monitoring System consists of four y sensitive ion chambers, two in the torus area and the other two inside the drywell. The detectors are capable of measuring radiation levels up to I to 107 R/hr, and can be monitored in the Control Room.
(i)      Further information regarding how these monitor readings can be used to calculate offsite doses based on the potential for release are discussed in Section D and the Emergency Plan Implementing Procedures.
(ii)     In addition to direct readout meter indications in the Control Room, a recorder is provided as wvell as several high level and inoperable alarms.
(C)   In addition to the above mentioned high range radiation monitoring systems, additional process and area radiation monitoring capabilities are available which enable assessment of inplant radiological conditions, fuel clad deterioration and effluent releases. The area radiation monitors, their range and location are provided in Table 1-1. Information regarding the process monitors is provided in Tables 1-2 and 1-3.
(D)  Under accident conditions, the normal sampling stations for obtaining representative Reactor Coolant System samples or primary containment atmospheric samples may be inaccessible or, if accessible, obtaining such samples may result in an individual receiving exposures in excess of 10 CFR 20 limits (10 CFR 50).

-                  o DAEC:EMERGENCY`PLAN' 9                                                    SECTION 'I' Rev. 21 ACCIDENT ASSESSMENT:                                 R     21 Page 7 of 17 NOTE Conditions may vary widely and change quickly during preparation for sampling and actual sampling during the period following an accident. The Site Radiation Protection Coordinator (SRPC)/Ra'd & EOF Manager must be conferred with prior to attempting to retrieve a sample. Assembly of temporary shielding may be necessary.
(i)       Locations from which samples can be drawn include the following points listed below. Containment isolation logic circuitry modifications have been provided, where required, to permit obtaining sam'ples under isolated conditions.
A depressurized Reactor Coolant System sample from the RHR Heat Exchanger discharge line when RHR is in the shutdown cooling mode of operation.
                           '.      A torus sample from the RHR/Core Spray Fill Pump 1P-70 casing drain, if pump is running and RHR is in the LCPtorus cooling, or test mode of operation.
                            -      A torus sampple from the RHR Heat Exchanger discharge line 'when RHR is in the LCPI,' torus cooling, or test mode of operation.
                             -     A drywelI atmospheric sample from the Containment
                                   'Atmosphere Monitoring System analyzer sample lines.
                             -.-   A tortis atmospheric sample from the Containment Atmospheric Monitoring system analyzer sample lines.
(ii)      The lab is equipped with standard chemistry as well as special equipment used to handle high level samples, etc. Sample preparation and routine chemical analyses activities will be accomplished in a shielded hood provided in the lab.
(E) Radiological data from the Kaman Effluent Monitoring system and meteorological data from the met tower are sent to the SPDS computer and assembled into one-minute raw data averages. The raw data is then

DAEC EMERGENCY PLAN                :-              SECTION 'I' Rev. 21 ACCIDENT ASSESSMENT Page 8 of 17 transferred to the MIDAS program via DecNet software. The MIDAS program then assembles these raw one-minute averages into 15-minute average raw data files. An internal quality control program either validates the data in the 15-minute average raw data files, or identifies the data as "questionable" and rejects it. The 15-minute raw data files are then used with preset, site-specific information and the 24 accident parameters to complete dose projections.
(i)      An interactive computer code (MIDAS) has been developed to perform dose projection calculations. The code can calculate, print, and plot the plume dispersion results of a Class B model for a single release. The calculations produce results for each of 31 spatial intervals (or distances) using a time-dependent plume segment model. Plume trajectory is normally determined by changes in wind direction with time. The model is run assuming flat terrain.
(ii)    The MIDAS program uses the plume segment model (CLASS B) repeatedly for each 15-minute release period to compute cumulative doses. Doses from each plume track are overlaid successively on a finely spaced radial grid. Contours of equal dose can be drawn through the doses calculated in the fine grid to produce isopleths over the integration time period. Doses are calculated for up to four projection periods. Results can be plotted on the graphics CRT and can utilize up to four release points (each one treated with a separate plume).
(iii)   To support all types of dose calculations within the B model, both time-integrated and "snapshot" (for dose rates) processing will be used. Snapshot processing enables estimation of the plume location at the current time as well as dose rate estimates from deposited particulates (ground shine) after the plume has left.

I
:.:DAECEMERGENCYPLAN               .       -. '-SECTION       'I' Rev. 21 ACCIDENT ASSESSMENT:'              s   -
Page 9 of 17 2.3      RADIOLOGICAL SOURCE TERM AND MAGNITUDE DETERMINATION (1)   The ratio of iodine to total gaseous activity has been established in the program and companion'procedures assuming a TID 14844 source term.
Modifications to this source term can be input to the program based upon the results of isotopic analyses conducted on containment atmospheric sampling and effluent stream filter cartridges. The means available to obtain and analyze these samples is discussed in paragraph 2.2 preceding.
(2)   As discussed in paragraph 2.2,-the effluent monitoring system is capable of detecting and measuring a wide-range of effluent activity concentrations up to those that could be present-presuming the TID 14844 source term. In addition, the containment radiation monitors discussed in paragraph 2.2 will provide an indication of the quantity of radioactive material available for release using the relationship specified. A procedure has been developed to relate containment radiation monitor readings to offsite doses.
(A)     Process monitors are available to provide an indication of radioactivity released in effluent water streams. An estimate of the magnitude of activity released can be made using installed plant instrumentation; e.g., tank levels and flow rates, and isotopic analyses of the sourcdeof activity. Further refinements can be made by sampling and analyses of effluent streams, aquatic biota, etc. The MIDAS source code can'also be
                 'used to make dose projections for liquid releases.
2.4      EFFLUENT MONITOR READINGS VS. EXPOSURE AND CONTAMINATION LEVELS                     . i (1)   The MIDAS dose projection program discussed in paragraph 2.2 provides the mechanism to relate effluent monitor readings to onsite and offsite exposures.
(A)v Due to the inherent inaccuracies in attempting to predict plume shape, downwind meteorological conditions, elevated atmospheric conditions and the like, field monitoring and analyses of airborne, waterborne, and environmental media provide the only real means of assessing the impact of radiological releases that may occur.
(2)   Ifthe MIDAS dose projection program is not operable, a PC-based program (laptop computer) is available in the TSC and EOF as a backup.

DAEC EMERGENCY PLAN                              SECTION 'I' Rev. 21 ACCIDENT ASSESSMENT Page 10 of 17 2.5     METEOROLOGICAL INFORMATION ACQUISITIONIEVALUATION (1)   The DAEC onsite meteorological program was initiated January 10, 1971.
New redundant instrumentation was added in November, 1984. In accordance with the regulatory position on Regulatory Guide 1.97, Revision 2, the meteorological system was designed in accordance with proposed Revision 1 to Regulatory Guide 1.23. Instrumentation is provided that is capable of measuring wind direction, wind speed, and ambient air temperature at two levels on the DAEC meteorological tower. Instrumentation is also provided for measuring the dewpoint at one level. For a discussion of the instrumentation, refer to Chapter 2 of the Updated Final Safety Analysis Report. Meteorological parameters monitored are also identified in Section H of this plan.
2.6     RELEASE RATEIPROJECTED DOSE METHODOLOGY FOR OFF-SCALE OR INOPERABLE INSTRUMENTS (1)    Emergency Plan Implementing Procedures exist for estimating release rate based on drywell and torus containment radiation monitor readings.
2.7     FIELD MONITORING (1)    Field monitoring is performed by DAEC personnel entailing, at a minimum, dose rate measurements and airborne sampling in the Plume Exposure Emergency Planning Zone. Results are reported to the Radiological Assessment Coordinator in the Emergency Operations Facility (or to the Site Radiation Protection Coordinator in the Technical Support Center if the EOF is not yet activated) where direction and control of the teams is exercised.
Whereas DAEC personnel will most likely be the first radiation survey teams dispatched to monitor the environs surrounding the site, the State of Iowa is also responsible for offsite monitoring. State teams will be dispatched to conduct similar monitoring activities and DAEC personnel will continue to supplement the State efforts. IES/NMC will coordinate offsite monitoring efforts conducted by DAEC personnel with those conducted by the State of Iowa. Upon termination of the release, IES/NMC will coordinate as required with the State of Iowa in establishing a long term environmental monitoring program.

        -DAEC                EMERGENCYP                                  SECTION 'I' Rev. 21 ACCIDENT ASSESSMENT;:              -     '
Page 11 of 17 2.8     RADIOLOGICAL HAZARD ASSESSMENT (1)   Radiological hazard assessment offsite commences with activation of the emergency plan, for those events with actual or potential releases. Field monitoring teams are dispatched from the Operational Support Center along the probable plume path, to ascertain the magnitude and location of contamination and radiation areas.
(2)   Teams will be dispatched and report locations by using reference locations or grid coordinates as shown on the DAEC Emergency Planning Zone map provided as Figure 1-1.
(3)   Teams will normally be dispatched in'IES/NMC vehicles. The monitoring teams are equipped with portable radios (described in Section F), survey and dose rate instruments, airborne sampling equipment, protective clothing and respiratory protection equipment. A further delineation of emergency equipment carried by the Field Teams is specifically identified in the Emergency Plan Implementing Procedures.
2.9     DETECTION AND MEASUREMENT OF RADIOIODINE CONCENTRATIONS (1)    Field teams dispatched are capable of measuring radioiodine concentration in air in the Plume Exposure'EPZ as low as 10-7 jCi/cc. Estimates of airborne' concentrations made using a survey meter with pancake probe on contact with a Silver Zeolite cartridge are provided to the Radiological Assessment Coordinator in the EOF. Filters ahnd'cartridges will be retained and a more accurate estimate of airborne concentrations obtained using laboratory counting equipment available at the DAEC or offsite laboratory facilities. A further discussion of additional laboratory facilities is contained in Section C.

DAEC EMERGENCY PLAN-                              SECTION 'I' Rev. 21 ACCIDENT ASSESSMENT Page 12 of 17 2.10    RELATIONSHIP OF MEASURED PARAMETERS TO DOSE RATES (1)  Various radiological parameters (contamination levels, water activity concentrations, air activity concentrations, etc.) measured in the field following an incident may be related to dose rates through the identification of key isotopes and gross radioactivity measurements. As discussed in paragraph 2.2, the MIDAS dose projection computer program incorporates the use of real time meteorological information, effluent release data and appropriate plant status inputs to calculate Deep Dose Equivalent (whole body dose) rates and estimate both Deep Dose Equivalent (whole body) and Committed Dose Equivalent (thyroid dose) commitments.
(2)  The results of analysis of environmental media and calculations related to total population exposure through the inhalation and ingestion pathways will be accomplished in accordance with the existing Appendix I Program.
3.0     ATTACHMENTS (1)  TABLE 1-1, "AREA RADIATION MONITORS" (2)  TABLE 1-2, "PROCESS RADIATION MONITORS" (3)  TABLE 1-3, "PROCESS RADIATION MONITORS FOR HIGH RANGE EFFLUENT MONITORING SYSTEM" (1)   FIGURE 1-1, UDAEC EMERGENCY PLANNING ZONE"

4-
                *l',;,.'_'DAEC      ',EME      NCY PLA                  .. SECTION      'I' Rev. 21 ACCIDENT ASSESSMENT.,
Page 13 of 17 TABLE 1-1 AREA RADIATION MONITORS DMonitorM                                   -; Locabonw'                       Range Desiiriator                            j RE-9151              RW Control Room (786')        -                      0.1-10 3 mr/hr RE-9152              RW Centrifuge Hallway                  -             0- .1-103mr/hr RE-9153              RB New Fuel Storage Area (855')                      0.1-103 mr/hr RE-9154              RW Drumming Area (757'-6")                           0.1 -103 mr/hr RE-9155              RB Jungle Room (812')                                 1-10 6 mr/hr RE-9156              RB Water Clean-up Recirc Pump Area(786')               1-106 mr/hr RE-9157              RB Water Clean-up Heat Exch. (786')                   1_106 mr/hr RE-9158              TB Condensate Pumps Area (734')                      0.1-103 mr/hr RE-9159             TB Reactor Feed Pump Area (734)                       0.1-103 mr/hr RE-9160             TB Turbine Lube Oil Area (734')                       0.1-10 3 mr/hr RE-9161             TB Machine Shop Area (757'-6")                       0.01-1 02 mr/hr RE-9162             AB Control Room Area (786')                           0.1-10 3 mr/hr RE-9163              RB Refueling Floor- North End (855')                0.01-102 mr/hr RE-9164              RB Refueling Floor - South End (855')               0.01-102 mr/hr RE-9165             AB OSC Hallway (757'-6")                             0.01-102mr/hr RE-9166              RB SW Comer Room (716'-9")                           0.1-10 3 mr/hr RE-9167              RB   Rx Bldg Railroad Access (757'-6")               0.1-103 mr/hr RE-9168              RB   North CRD Module Area (757'-6")                 0.1-103 mr/hr RE-9169              RB   South CRD Module Area (757'-6")                 0.1-103 mr/hr RE-9170              RB   CRD Repair Room (757'-6")                       0.1-103 mr/hr RE-9171              RB   Exhaust Fan Room (812')                         0.1-103mr/hr RE-9172             AB    Rad. Chem. Lab (786')                           0.1-103 mr/hr RE-9173              RB   Spent Resin Tank Room (786')                    0.1-103 mr/hr RE-9174             TB    Sump Area Entrance (734')                       0.1-103mr/hr RE-9175              RB Cond. Phase Tank & Corridor Area (833'-6")        0.1-103 mr/hr RE-9176              RB TIP Drive Room (757'-6")                            1-104 mr/hr RE-9177              RB Outside Cleanup Phase Separators Tank Room          1-104mr/hr I I  RE-9178              R SnFlta(786')
RE-9178              RB Spent Fuel Storage Area (855')                     1_10 4 mr/hr

DAEC EMERGENCY PLANWSECTION                               'I' Rev. 21 ACCIDENT ASSESSMENT Page 14 of 17
 -'''Monitor:i^...-'--................-
                    '''Rn
                        '. ;n rto -~=-~;
DeiignaL                _,.-;.                     cratio'w -
_________________________,                   Range-RE-9179              TB Turbine Standard Area (780')                      1-104 mr/hr RE-9180              RB 1T-70, 73 Waste Collector/Floor Drain Tank        1-10 4mr/hr Room                                          _  _ _  _  _   _   _ _ _ _
RE-9184A               Drywell                                               1-10 7 R/hr RE-9184B               Drywell                                               1-10 7 R/hr RE-9185A               Torus                                                 1-107 R/hr RE-9185B               Torus                                                 1-10 7 R/hr CODE RW       Radwaste Bldg.
RB       Reactor Bldg.
TB       Turbine Bldg.
AB       Administrative Bldg.

O<^~~~~V, -     -         ,  ;S;9z A.-PvNJ.
:,,,AEC EMERG,,ENC               PLN.             S ~     .-; ;SECTION 'I'
                                                                           . Rev. 21 ACCIDENT ASSESSMENT Page 15 of 17 TABLE 1-2
                               - PROCESS RADIATION MONITORS
-.:Monator -onitor                                Locatio RE-4448A    Main Steamline                                                            1-106mr/hr RE-4448B    Main Steamline                                                            1-106 mr/hr RE-4448C     Main Steamline                                                            1-1 06 mr/hr RE-4448D    Main Steamline                                                            1-106mr/hr RE-3972    Radwaste Effluent                                                         0.l-106 cps RE-1 997   RHR and Emergency Service Water                                        . 0.1-10o cps RE-4820    Reactor Building Closed Cooling Water                                     0.1-106 cps RE-4767    Service Water Effluent                      -                             0.1-106 cps RE-4104    Off-Gas Pre-Treatment                                     --              1 06 mr/hr RE-4105    Off-Gas Linear FluxTilt -.                                           -    0-125 units:
  -RE4101A    Off-Gas Post-Treatment                 ---                                         cps RE-41 01B   Off-Gas Post-Treatment                                                    0.1-106  cps RE-411      ReOff-GasctorsBuilTrefmeling                                             0.01-1 0, m RE-4131A    Reactor Building Refueling Ventilation Exhaust                          0.01-100 mr/hr RE-4131B    Reactor Building Refueling Ventilation Exhaust                          0.01-100 mr/hr RE-7606A    Reactor Building (Mezzanine)                                             0.05-50 mr/hr RE-7606B    Reactor Building (Mezanine)                                              0.05-50 mr/hr RE-6101A    Control Building Intake                                                  0.05-50 mr/hr RE-61 01B   Control u              nteBuilding Intake                                0.05-50 mr/hr RE-7722A    Technical Support Center Building Intake (Left)                          0.1-104 mr/hr RE-7722B    Technical Support Center Building Intake (Right)                         0.1-104 mr/hr RE-7722C    Technical Support Center Working Area                                    0.1-_104 mr/hr RE-4138    Carbon Bed Vault (Off-Gas Building)                                       1-106 mr/hr RE-4268    RHRSW/ESW Effluent                                                        0.1-106 cps RE-4116A    Offgas                                                                    0.1-106 cps RE-4116B    Offgas                                                                    0.1-106 cps

DAEC EMERGENCY PLAN               -      SECTION 'I' Rev. 21 ACCIDENT ASSESSMENT Page 16 of 17 TABLE 1-3 PROCESS RADIATION MONITORS FOR HIGH RANGE EFFLUENT MONITORING SYSTEM Monitor~Io
:.-Monio;.                  nitorio~-;t-Lcatio%-   .;ange                :.
j:Designator                                                        Rng RE-5945      Turbine Building Ventilation Exhaust         5 x 10 1E' pCi/cc RE-5946      Turbine Building Ventilation Exhaust           0.01 - 10 5 pCi/cc RE-7645      Reactor Building Ventilation Exhaust         5 x 1 -_0- 1El' pCi/cc RE-7644      Reactor Building Ventilation Exhaust           0.01 - 10, pCi/cc RE-7647      Reactor Building Ventilation Exhaust         5 x 10 1E1 pCi/cc RE-7646      Reactor Building Ventilation Exhaust           0.01 - 105 pCi/cc RE-7649      Reactor Building Ventilation Exhaust         5 x 10 1E' pCi/cc RE-7648      Reactor Building Ventilation Exhaust           0.01 - 10 5 pCiCC RE-4176      Off-Gas Stack Discharge                      5 x 10 1E' pCi/cc RE-4175      Off-Gas Stack Discharge                        0.01 - 105 pCi/cc

-DAEC EMERGENCY PLAN                    - SECTION 'I' Rev. 21 ACCIDENT ASSESSMENT Page 17 of 17 FIGURE 1-1 DAEC EMERGENCY PLANNING ZONE TO L -
tI MT VERNON.
DAVENPORT.
AND LISBON TO IOWA CITY

.- I- -.
                      ;--:-::-:DAECEMERGENCY PLAN -: --     .      SECTION 'N' Rev. 21 EXERCISES AND DRILLS            -
Page 1 of 6 Effective Date:  7 TECHNICAL REVIEW Prepared by:           AV                           Date:  lZ$ "4 Reviewed by:K                                       Date:.
                     -nfa~ -Reviewer
[.            .- - ..                   ... .....
PROCEDURE APPROVAL Approved by:                                        Date: &#xa3; >- 6L/
Manager, Emergency Planning Reviewed by:                                        Date:   /z7 1/Oti Approved by:                                        Date:   Zl /04-
                                                                         - i - -
Authorized by:                                       Date:    7-14 loy

DAEC'EMERGENCY PLAN                  ;       SECTION'N' Rev. 21 EXERCISES AND DRILLS Page 2 of 6 Table of Contents Page 1.0 PURPOSE ............................                               3 2.0 REQUIREMENTS ...........................                           3 2.1  EXERCISES ........................                            3 2.2  DRILLS .......................                                4 2.3  SCENARIOS ........................                            5 2.4  OBSERVATION AND CRITIQUE .......................              6 2.5  MANAGEMENT CONTROL ........................                   6

Z'   '
 <b..   :2$si it     DAEC EMERGENCY PLAN.1 '             '; '             SECTION 'N' Rev. 21 EXERCISES AND DRILLS Page 3 of 6 1.0   PURPOSE (1) This section describes the periodic exercises and drills which are used to enhance and evaluate the capabilities of the Emergency Response
         - Organization..'
(2) The Nuclear Management Company (NMC) is assigned operational responsibility for the DAEC. However, IES maintains corporate accountability for activities at the DAEC and will participate when necessary in 'activities at the DAEC. The reference "IES/NMC" will be used'throughout this-procedure to signify this relationship. Further details regarding this relationship can be found in the "Nuclear Power Plant Operating Services Agreement" (NPPOSA) between IES and the NMC.
2.0     REQUIREMENTS The Manager, Emergency Planning shall be responsible for planning and scheduling drills and exercises, and the coordination of such with all affected groups within IES/NMC.
2.1     EXERCISES (1) Exercises will be conducted annually to evaluate the overall response and emergency capabilities of the ERO, to ascertain the level of familiarity with the emergency plan and procedures, and to demonstrate the effectiveness of the plan. These exercises will be coordinated with state and local response organizations to ensure their capability in mobilizing to the necessary level at least once every two years. The operational and readiness response to any exercise will be evaluated by observers and principal participants from agencies involved in the exercise. Exercises will be conducted in accordance with NRC/FEMA regulation. State and county participation is required on a biennial basis. -
(2) The scenarios for the exercises'shall be sufficiently varied such that all major portions of NUREG 0654, the plans and organizations are tested at least once every six years. An off-hours exercise will be scheduled to start between 6:00 p.m. and 6:00 a.m. once every six years. Where possible, the off-hours.
exercise should be unannounced.

.                    DAEC;EMERGENCY PLN,.-i,!                             SECTION'N' Rev. 21 EXERCISES AND DRILLS Page 4 of 6 2.2     DRILLS (1) A drill is a supervised instruction period that refers to an event involving organizational responses to a simulated accident to develop, test, and monitor specialized emergency skills that constitute one or more components of an emergency plan and/or procedures.
(2): Drills may be conducted as part of an exercise and will be evaluated as described in paragraph 2.4 of this section. Drills may be used as requalification training for ERO personnel. Specific types of drills and their frequencies are delineated below:
(A)   Fire Drills (i)     Fire drills will be conducted quarterly in accordance with the DAEC Fire Plan.                                                          KU (B)   Medical Emergency Drills (i)     Evaluated medical emergency drills will be conducted annually.
These drills will involve local support services and will use simulated contaminated and/or injured individuals.
(C)   Radiological Monitoring Drills (i)     Radiological monitoring drills will be conducted annually. These drills will involve both onsite and offsite monitoring teams and include the collection and analysis of sample media; e.g., water, air, vegetation, and soil.
        - (D)   Health Physics Drills, (i)     Health Physics drills will be conducted semi-annually. These drills will involve response to, and analysis of, simulated elevated airborne and liquid samples and direct radiation measurements at the DAEC.
(E) ERO Activation Drills

or
                    'DAEC EMERGENCY PAN                                     SECTION 'N' Rev. 21 EXERCISES AND DRILLS-                     .'             Re 21 Page 5 of 6 (i)    A minimum of two ERO activation drills will be conducted on an annual basis. These drills will assess the ability of the ERO to activate and staff the facility.
2.3      SCEO IARIOS     -
(1)   The conduct of each exercise and drill will be governed by staged scenarios.
Scenarios will specify and include at least the following:
Basic objectives of the drill or exercise. A measurable and observable objective must be specified for each major task/function.
                            -' *.Dates,      time period, places, personnel, and participating organizations.
                            -    'Simulated
                                 .-                events. .                ;
* Time schedule'of real and simulated initiating events.
* Narrative summary describing the conduct of the exercise or drill, including simulated casualties, offsite fire department assistance, rescue of personnel, use of protective clothing and associated equipment, deployment of personnel and radiological teams, and public information activities.
* A description of arrangements for and advance materials to be provided to official observers.

DAEC EMERGENCYrPRN-                               SECTION 'N' Rev. 21 EXERCISES AND DRILLS Page 6 of 6 2.4    OBSERVATION AND CRITIQUE (1) Qualified personnel will observe and perform a critique of exercises and drills.
Provisions will be made for federal, state, and local observers, as well as IES/NMC personnel, to observe and critique required exercises. A review and presentation of exercise critiques will be held as soon as practical after the   K)1-exercise and will be the subject of a formal report, as addressed in Paragraph 2.5. Drills will be evaluated, but a formal critique need not be generated.
However, on-the-spot corrections of erroneous performance will be made and overall results will be documented.
2.5    MANAGEMENT CONTROL (1) The results of exercise critiques, particularly comments on identified areas which require improvement or reevaluation, will be submitted to the Manager, Emergency Planning, or designee, for review. The Manager, Emergency Planning, or designee, will consult with responsible department heads and assign corrective action activities, as appropriate. Necessary plan and procedure changes will be handled as discussed in Section P; additional training will be included in the training program as discussed in Section 0.
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