ML030910406

Attachment I Comparison of NUREG-0654 to Ginna Nerp, and Proposed Enhancements, Attachment II Evaluattion of NUREG-0654 Basis for on Shift Staffing, Attachment III Evaluation of NUREG-0654 Basis for Timing Staff Augmentation, Attachment IV

ML030910406
Person / Time
Site:	Ginna
Issue date:	03/11/2003
From:	Clark R NRC/NRR/DLPM/LPD1
To:	Richard Laufer NRC/NRR/DLPM/LPD1
References
TAC MB7240
Download: ML030910406 (46)

Text

Rochester Gas and Electric Corporation (RG&E) has evaluated the outstanding on shift and emergency plan staff augmentation issue raised by NRC Inspection Report 50-244/02-09 and TIA 2002-02. This evaluation was performed from a technical standpoint. Specifically, the issues were analyzed with consideration of what actions were required (and when), and why such actions were a requirement. Available NRC guidance (e.g., NUREG-0654), recent NRC Safety Evaluation Reports, surveys of the nuclear power industry, and the Ginna Nuclear Emergency Response Plan (NERP) and procedures formed the basis for this review. The attachments which follow contain detailed assessments of the following topics:

I.

Comparison of NUREG-0654 to Ginna NERP, and Proposed Enhancements II.

Evaluation of NUREG-0654 Basis for On Shift Staffing III.

Evaluation of NUREG-0654 Basis for Timing Staff Augmentation IV.

Evaluation of RG&E Current On Shift Staffing V.

RG&E NERP Staff Augmentation Estimated Travel Times While the attachments contain specific details, the following is a brief summary of the overall conclusions obtained from these evaluations.

a.

RG&E proposes to add 17 one hour responder positions to the 13 current one hour responders. RG&E is also proposing to add one 30 minute responder to address the on shift difference with respect to the Rad/Chem Technician position. This 30 minute responder is actuated upon declaration of an unusual event or reactor trip, which ever occurs first. Attachment I contains additional details.

b.

Ginna is a small compact site/facility in comparison with most plants in the U.S. nuclear power industry. As such, it is much easier and quicker to gain access to the facility and to the necessary components. This feature allows for fewer individuals to be required to immediately respond to an event. Based on surveys with 12 other plants, the Ginna on shift organization does not appear to be significantly different from a number of other sites of similar size and type.

c.

Ginna is easily accessible by the response organization. There is a four lane highway only three miles from the site, with straight access roads, which provides for reasonable response times of the staff augmentation. The majority of the response organization lives within a relatively short distance from the site as shown in Attachment V. A large percentage of the proposed one hour responders (> 60%) have a 30 minute or less travel time to the plant, with > 80% having a travel time of less than 45 minutes. The relatively short estimated travel time is due to the fact that there are numerous suburbs of the city of Rochester, N.Y. near the plant.

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d.

All NERP facilities activate at the ALERT level (which by definition has a low potential for offsite impact). At the ALERT, the need for offsite assistance to immediately contain and mitigate the event is small. RG&E voluntarily changed the EOF activation requirement to the ALERT many years ago, recognizing the need for offsite support (we previously only required EOF activation within one hour of a SITE AREA EMERGENCY). A number of the current and proposed one hour response individuals are cross trained to provide further depth in positions. Also, three one hour responders are activated at an Unusual Event level. While these positions are not called out in NUREG-0654, they can help to identify additional resources which are needed for slow moving accidents.

e.

Based on a review of available analyses, the accidents which have the greatest potential for early releases or which create the most impact on shift personnel were evaluated to confirm that all expected actions could be completed. Attachments III and IV contain additional details.

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Attachment I Comparison of NUREG-0654 to Ginna NERP, and Proposed Enhancements Draft - For Information Only 3/11/03

Comparison of NUREG-0654 to Ginna NERP, and Proposed Enhancements NUREG-0654 Ginna Current Ginna Proposed Notes Im-~

Position Title or Expertise On 30 60 On 30 60 On 30 60 Shift Min.

Min.

Shift Min.

Min.

Shift Min.

Min.

(Note 1)

(Note 2)

Plant Operations and Assessment of Operational Aspects I

Shift Supervisor (SRO) 1 1

1 2

Shift Foreman (SRO) 1 1

3 Control Room Operators 2

2 2

4 Auxiliary Operators 2

2 2

Emergency Direction and Control (Emergency Coordinator)

Supervisor, or designated 1*

facility manager I

I I

I_

I Notification/Communication 6 l (There is no NUREG title for l1 l

1 l

2 l

1 l

}

l 1

l2 lNote 4 this position - Communicator) 1 1 2

lNote4 Radiological Accident Assessment and Support of Operational Accident Assessment 7

Senior Manager (EOF 1

1 1

Note 5 Director) l l

8 Senior Health Physics (HP) 1 2

2 Note 6 Expertise l

l l

l 1_I I

Pagel1 of 10 Draft - For Information Only 3/11/03

NUREG-0654 Ginna Current Ginna Proposed Notes Position Title or Expertise On 30 60 On 30 60 On 30 60 Shift Min.

Min.

Shift Min.

Min.

Shift Min.

Min.

(Note 1)

(Note 2) l l

l 9

(There is no NUREG title for 2

2 4

Note 7 this position - Offsite Surveys) 10 (There is no NUREG title for 1

1-2 Note 8 this position - Onsite Surveys) 11 HP Technicians (In-plant 1

1 1

1 1

1 2

Note 9 surveys) 12 Rad/Chem Technicians 1

1 Note 10 Plant System Engineering, Repair and Corrective Actions 13 Shift Technical Advisor 1

=

11

=

=

14 Core/Thermal Hydraulics 12 I

1 Note 1 1 15 Electrical 1-1 Note 11 16 Mechanical 1

1 Note I1 17 Mechanical Maintenance 1*

1 1*

1*

1 Note 12 18 Rad Waste Operator 1

Note 13 19 Electrical Maintenance 1*

1 I

1*

-Note 14 20 Instrument & Control 1

1 Note 14 Technician Protective Actions (In-plant) 21 l HP Technicians 2*

2 2

1*

1 1*

3 Note 15 Page2 of 1 0 Draft - For Information Only 3/11/03

NUREG-0654 Ginna Current Ginna Proposed Notes Position Title or Expertise On 30 60 On 30 60 On 30 60 Shift Min.

Min.

Shift Min.

Min.

Shift Min.

Min.

(Note )

(Note 2)

Fire Fighting 22 (There is no NUREG title for per Local support Local support l Local support Note 16 l

this position - Fire Brigade) t FPP p

N 16 Rescue Operations and First-Aid 23 (There is no NUREG title for 2*

Local Support 5

Local Support 5

Local Support Note 17 this position)

I I

I I

I 0!4 A

o1LU access Control and Personnel Accountability

')A I I

T dl.t I

Security Personnel Total I

Per NUREG-0654, this position may be provided by shift personnel assigned other functions.

The Ginna 60 minute totals do not include credit for the following current NERP required one hour responders, which have been found to be required to effectively implement the plan:

TSC Operations Assessment Manager Technical Assessment Manager Survey Center Manager EOF Nuclear Operations Manager Engineering Manager News Center Manager Page3 of IO Draft - For Information Only 3/11/03

Note 1 RG&E is not currently committed to any 30 minute responders in the Nuclear Emergency Response Plan (NERP). This was provided to the NRC in a letter dated 5/1/8 1. The NRC provided approval of the Ginna emergency plan on May 5, 1983 and considered NUREG-0737 Item III.A.2.1 as complete.

Note 2 The operators on shift are capable of performing minor electrical and mechanical maintenance. This could include replacing fuses, tightening valve packing, etc. This provides the capability for initial repair and corrective actions. The operators have received radiation protection training as part of their training and can provide that capability if required. The Ginna Shift RP Technicians have the training and expertise to provide radiation surveys and are cross-trained in radiochemistry/chemistry analysis. The current on shift personnel have the training and ability to place the plant in a safe condition as documented in Attachment IV. Though many of the responders live close to the site and have travel times less than 30 minutes per Attachment V, RG&E is only proposing that a HP or Rad/Chem qualified individual 30 minute responder be specifically committed to as part of the NERP.

Plant Operations and Assessment of Operational Aspects The current Ginna shift complement meets the guidance of Table B-I for this functional area.

Emergency Direction and Control (Emergency Coordinator)

Note 3 The Emergency Coordinator position is initially filled by the Shift Supervisor. He is relieved of this duty by the TSC Director (one hour responder) who becomes the TSC Emergency Coordinator when the TSC assumes command and control.

Notification/Communication Note 4 Ginna currently has a Communicator on shift (I of 3 Auxiliary Operators) and a TSC Communicator as a required NERP one hour responder. Additional Communicator qualified personnel are notified to respond as part of the automated emergency notification process as described in the RG&E letter dated November 6, 2002. The addition of the Emergency Response Data System (ERDS) has also provided the NRC with the ability to remotely monitor key Ginna parameters. As an enhancement, RG&E proposes to add an EOF Communicator as a NERP required one hour responder to provide for a Page 4 of 10 Draft - For Information Only 3/11/03

Communicator in each of the three command and control facilities. There are 4 qualified TSC Communicators and 7 EOF Communicators with less than a 30 minute travel time per Attachment V.

Radiological Accident Assessment and Support of Operational Accident Assessment Note 5 Ginna currently has an EOF Recovery Manager as a required NERP one hour responder.

Note 6 Ginna currently utilizes the 1 Ginna Shift RP Technician to perform this function initially, as a collateral duty. The STA is also trained to perform the dose assessment calculation per EPIP 2-18. This calculation is done by the plant process computer, and is backed up with a simple calculation form. Ginna also has a TSC Dose Assessment Manager and an EOF Dose Assessment Manager as required NERP one hour responders, with 5 of these individuals having less than a 30 minute travel time per Attachment V.

Note 7 Ginna currently does not have off-site survey personnel listed in the NERP as required one hour responders, though they are notified to respond as part of the automated emergency notification process as described in the RG&E letter dated November 6, 2002. Effluent monitor calculations and other plant RG 1.97 indications are the preferred method for rapid determination of EALs and PARs. That is the basis for the Ginna Emergency Action Level (EAL) classification and Protective Action Recommendation (PAR) procedures. Offsite radiological survey tasks such as soil, water, and vegetation sampling or environmental TLD retrieval can be performed when additional augmentation personnel arrive. These types of radiological survey tasks would be considered in the recovery phase following an offsite release of radioactive material and are not needed for the immediate protection of the public health and safety. As an enhancement, RG&E proposes to add 4 off-site survey personnel as NERP required one hour responders. There are a total of 15 qualified survey personnel (off-site and on-site) with less than a 30 minute travel time per Attachment V.

Note 8 Ginna currently does not have on-site survey personnel listed in the NERP as required one hour responders, though they are notified to respond as part of the automated emergency notification process as described in the RG&E letter dated November 6, 2002. Effluent monitor calculations and other plant RG 1.97 indications are the preferred method for rapid determination of EALs and PARs. That is the basis for the Ginna EAL and PAR procedures. As an enhancement, RG&E proposes to add 2 on-site survey personnel as NERP required one hour responders. There are a total of 15 qualified survey personnel (off-site and on-site) with less than a 30 minute travel time per Attachment V.

Note 9 Ginna currently has 1 Ginna Shift RP Technician. Also, a RP/Chem Manager is a current NERP one hour responder that would assist in providing senior HP expertise. Additional HP qualified individuals (Ginna Shift RP Technicians and RP Page5 of 10 Draft - For Information Only 3/11/03

Communicator in each of the three command and control facilities. There are 4 qualified TSC Communicators and 7 EOF Communicators with less than a 30 minute travel time per Attachment V.

Radiological Accident Assessment and Support of Operational Accident Assessment Note 5 Ginna currently has an EOF Recovery Manager as a required NERP one hour responder.

Note 6 Ginna currently utilizes the I Ginna Shift RP Technician to perform this function initially, as a collateral duty. The STA is also trained to perform the dose assessment calculation per EPIP 2-18. This calculation is done by the plant process computer, and is backed up with a simple calculation form. Ginna also has a TSC Dose Assessment Manager and an EOF Dose Assessment Manager as required NERP one hour responders, with 5 of these individuals having less than a 30 minute travel time per Attachment V.

Note 7 Ginna currently does not have off-site survey personnel listed in the NERP as required one hour responders, though they are notified to respond as part of the automated emergency notification process as described in the RG&E letter dated November 6, 2002. Effluent monitor calculations and other plant RG 1.97 indications are the preferred method for rapid determination of EALs and PARs. That is the basis for the Ginna Emergency Action Level (EAL) classification and Protective Action Recommendation (PAR) procedures. Offsite radiological survey tasks such as soil, water, and vegetation sampling or environmental TLD retrieval can be performed when additional augmentation personnel arrive. These types of radiological survey tasks would be considered in the recovery phase following an offsite release of radioactive material and are not needed for the immediate protection of the public health and safety. As an enhancement, RG&E proposes to add 4 off-site survey personnel as NERP required one hour responders. There are a total of 15 qualified survey personnel (off-site and on-site) with less than a 30 minute travel time per Attachment V.

Note 8 Ginna currently does not have on-site survey personnel listed in the NERP as required one hour responders, though they are notified to respond as part of the automated emergency notification process as described in the RG&E letter dated November 6, 2002. Effluent monitor calculations and other plant RG 1.97 indications are the preferred method for rapid determination of EALs and PARs. That is the basis for the Gina EAL and PAR procedures. As an enhancement, RG&E proposes to add 2 on-site survey personnel as NERP required one hour responders. There are a total of 15 qualified survey personnel (off-site and on-site) with less than a 30 minute travel time per Attachment V.

Note 9 Ginna currently has 1 Ginna Shift RP Technician. Also, a RP/Chem Manager is a current NERP one hour responder that would assist in providing senior HP expertise. Additional HP qualified individuals (Ginna Shift RP Technicians and RPI Page5 of 10 Draft - For Information Only 3/11/03

Technicians) are notified to respond as part of the automated emergency notification process as described in the RG&E letter dated November 6, 2002. The on-shift Control Room operators, STA and Ginna Shift RP Technician have remote indication of in-plant area radiation monitors, process monitors, and effluent monitors in the Control Room. These initially would guide the assessment of in-plant radiological conditions, and deployment of Auxiliary Operators and Fire Brigade members.

As an enhancement, RG&E proposes to add 1 individual qualified in either HP functions or Rad/Chem functions as a 30 minute responder who would respond to off-normal events (Unusual Event or reactor trip). Also, RG&E proposes to add two HP qualified individuals as NERP required one hour responders to support in-plant surveys. There are a total of 16 HP qualified personnel and RP/Chemistry Managers with less than a 30 minute travel time per Attachment V.

Note 10 Ginna currently does not have a separate on shift Rad/Chem Technician. Additional Rad/Chem qualified individuals (Ginna Shift RP Technicians and Chem Technicians) are notified to respond as part of the automated emergency notification process as described in the RG&E letter dated November 6, 2002. Since the completion of the inspection on April 17, 2002, RG&E has conducted an analysis to determine the adequacy of the on-shift staffing during the following high manpower intensive events:

Grid Failure, Direct entry into Station Blackout Procedure, Security available Security event in Switchyard, Loss of circuits 767 and 751, Security not available Explosion in Screen House, Loss of Buses 17 & 18, Security available Security event in Screen House, Loss of Buses 17 & 18, Security not available Fire in the Aux. Bldg., Security available Security event and subsequent fire in the Aux. Bldg., Security not available LOCA outside containment LOCA outside containment with large fire Design Basis SGTR Design Basis SGTR with large fire The analysis (see Attachment IV) found that, although some activities would not be covered with the current on-shift Ginna Shift RP Technician staffing, those activities are not critical to the mitigation or recovery of the event. Specifically, there are no critical chemistry samples required by operations procedures to mitigate the events. The Shift Supervisor prioritization of non-critical activities would ensure that the activities were done as timely as possible.

Though the Rad/Chem Technician function itself may not be time critical, RG&E has determined that an augmentation of the Ginna Shift RP Technician on shift function with a new 30 minute responder would mean that these activities would be Page 6 of IO Draft - For Information Only 3/11/03

performed in a more timely manner. Since the Ginna Shift RP Technicians are qualified to perform chemistry analysis as well as radiological surveys and dose analysis, this augmentation would allow for the completion of the non-critical activities mentioned above as well as other activities. As discussed in notes 9 and 15, additional responders, including a RP/Chem Manager, would also have the ability to support chemistry analysis, and as such, no separate Rad/Chem Technician one hour responder would be added.

Plant System Engineering, Repair and Corrective Actions Technical support personnel are provided to support supplemental actions need to ensure the plant remains in a stable condition, restore capabilities needed for control of the plant, and assist in planning/preparing necessary corrective maintenance. As such, these functions are not needed during the initial stage of an emergency. The technical support personnel are needed for assessing the extent and impact of damage, practical long-term stabilization options, priority corrective maintenance, and other plant recovery work.

Due to the time needed to stabilize the plant and assess the event, the initial phase of an accident scenario is not expected to involve a large need for maintenance personnel for activities that could not be performed by the on shift complement. Only after the plant is in stable and understood status can attention be refocused to corrective maintenance that may be needed to restore plant conditions. Until the reactor plant is stabilized and the causal agents are discerned, actual repairs or realignment of plant equipment should not require large-scale maintenance support.

Note 11 The on-shift STA is able to provide the core/thermal hydraulics expertise until the arrival of a dedicated individual. Ginna currently has a TSC Technical Assessment Manager and an EOF Engineering Manager listed in the NERP as required one hour responders who can preliminarily fulfill these functions. The specific engineering discipline personnel are also currently notified to respond as part of the automated emergency notification process as described in the RG&E letter dated November 6, 2002. As an enhancement, RG&E proposes to supplement the current one hour engineering responders with 3 engineering discipline specific personnel as NERP required one hour responders. There are a total of 11 qualified engineering support personnel with less than a 30 minute travel time per Attachment V.

Note 12 Ginna currently has a Maintenance Assessment Manager listed in the NERP as a required one hour responder whose task is to determine and prioritize the repair activities. Mechanical Maintenance Technician expertise is not needed until after the plant has been placed in a safe condition since these tasks typically require significant planning and coordination. All equipment manipulations would be initially performed by auxiliary operators, who could also perform minor activities such as tightening valve packing. The Emergency Coordinator directs the call-in of technicians to troubleshoot and correct equipment malfunctions whenever equipment problems are identified. Since the Shift Supervisor assumes the duties of the Emergency Page 7 of 10 Draft - For Information Only 3/11/03

Coordinator at the classification of the event, the necessary technicians would be called in the perform the necessary troubleshooting and repair of equipment early during the event. As an enhancement, RG&E proposes to add a Mechanical Maintenance Technician as a NERP required one hour responder. There are a total of 21 Mechanical Maintenance I

personnel with less than a 30 minute travel time per Attachment V.

Note 13 Ginna currently does not have a Rad Waste Operator as a required NERP one hour responder. There is no need for a radiological waste operator until well after the event has been mitigated. Any radiological waste processing would be performed by an auxiliary operator as part of their normal duties during the recovery phase of the event. Therefore, Ginna does not propose the addition of a separate Rad Waste Operator as a NERP one hour responder.

Note 14 As stated in Note 12, the Maintenance Assessment Manager is a required one hour responder whose task is to determine and prioritize the repair activities. Electrical/Instrument & Control Technician expertise is not needed until after the plant has been placed in a safe condition. All equipment manipulations would be initially performed by auxiliary operators, who could also perform minor activities such as replacing fuses and closing breakers. The Emergency Coordinator directs the call-in of I

technicians to troubleshoot and correct equipment malfunctions whenever equipment problems are identified. Since the Shift l

Supervisor assumes the duties of the Emergency Coordinator at the classification of the event, the necessary technicians would l

be called in the perform the necessary troubleshooting and repair of equipment early on during the event. As an enhancement, RG&E proposes to add an Electrician and an Instrument and Control Technician as NERP required one hour responders. There are a total of 10 Electrical/Instrument & Control personnel with less than a 30 minute travel time per I

Attachment V.

Protective Actions (In-plant)

I Note 15 Ginna currently has a Ginna Shift RP Technician who could assist with protective actions as prioritized by the Shift Supervisor. The on shift Auxiliary Operators and Fire Brigade members receive basic radiation monitoring training. Radiation exposure monitoring has improved dramatically since NUREG-0654, Table B-1, was issued. The on shift Auxiliary Operators, I

Fire Brigade, and Security Officers all use alarming dosimeters with dose and dose rate alarms. The Auxiliary Operators are also trained to use some portable radiation instrumentation. On shift personnel can self-frisk (using PCMs). A RP/Chem l

Manager and Dose Assessment Manager are current NERP required one hour responders reporting to the TSC. These I

individuals are more experienced and can coordinate a more in-depth assessment of radiological conditions inside or outside the plant. Additional HP qualified individuals are notified to respond as part of the automated emergency notification process as described in the RG&E letter dated November 6, 2002. As an enhancement, RG&E proposes to add two additional HP qualified individuals as NERP required one hour responders to support in-plant protective actions. There are a total of

]

16 qualified HP qualified personnel and RP/Chemistry Managers with less than a 30 minute travel time per Attachment V.

I Page 8 of 10 Draft - For Information Only 3/11/03

Fire Fighting Note 16 The Ginna Fire Protection Program requires a five person Fire Brigade. The on shift Fire Brigade consists of two Auxiliary Operators, who perform captain duties, and three additional separate dedicated Fire Brigade members. The local volunteer fire department is approximately 4 miles from the site and is able to respond very rapidly.

Rescue Operations and First-Aid Note 17 Rescue Operations and First-Aid is provided on shift by the Fire Brigade.

Site Access Control and Personnel Accountability The current Ginna security complement will not be discussed.

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Summary of Proposed Ginna NERP Staff Augmentation NERP 30 Minute Responders*

HP or Rad/Chem qualified individual Proposed NERP One Hour Responders Onsite TSC Emergency Coordinator Operations Assessment Manager Technical Assessment Manager Maintenance Assessment Manager RP/Chemistry Manager TSC Dose Assessment Manager TSC Communicator Survey Center Manager Administrative Manager Off-site Survey (4)

On-site Survey (2)

HP qualified individuals (4)

Nuclear Assessment I&C/Electrical Assessment Mechanical/Hydraulic Assessment Mechanical Maintenance Electrician I&C Technician Current Current Current Current Current Current Current Current Current Proposed Proposed Proposed Proposed Proposed Proposed Proposed Proposed Proposed Off site Recovery Manager Engineering Manager Nuclear Operations Manager EOF Dose Assessment Manager News Center Manager EOF Communicator Current Current Current Current Current Proposed I

Il

• The 30 minute responder is actuated upon declaration of an Unusual Event or reactor trip, which ever occurs first.

I Page 10 of 10 Draft - For Information Only 3/11/03

Attachment II Evaluation of NUREG-0654 Basis for On Shift Staffing Draft - For Information Only 3/11/03

Evaluation of NUREG-0654 Basis for On Shift Staffing With regards to the NUREG-0654, Table B-1, guidance for on shift staffing, RG&E currently only differs in the area of a Rad/Chem Technician. RG&E has one Ginna Shift RP Technician who is cross-trained to performed this function and is proposing to add a second individual as a 30 minute responder. Listed below are the types of activities that could be expected of the HP and Rad/Chem Technician taken from draft NUREG/CR-3903 and how they would be covered by the Ginna on shift personnel until augmented by the response staff.

Activity RG&E Coverage Radio-chemical The Ginna Shift RP Technician is qualified to obtain samples Sampling/Analysis and perform chemistry analysis. Based on a review performed of various scenarios (see Attachment IV), these activities are not expected to be performed within the first 30 minutes and are not critical to the mitigation or recovery of the event. Specifically, there are no critical chemistry samples required by Emergency procedures to mitigate an event. The Shift Supervisor prioritization of these activities would ensure that the activities were done as timely as possible, until the arrival of augmented staffing.

In-plant Radiological The on shift Control Room operators, STA and Ginna Shift RP Monitoring Technician have remote indication of in-plant area radiation monitors, process monitors, and effluent monitors in the Control Room. Many of these monitors are RG 1.97 instruments, providing added assurance of their reliability and availability post accident. These monitors would initially guide the assessment of in-plant radiological conditions, and deployment of Auxiliary Operators and Fire Brigade members. The in-plant radiation monitoring could be further supplemented by the Ginna Shift RP Technician as necessary.

Meteorological Assessment The on shift STA and Ginna Shift RP Technician have remote indication of meteorological data from the Control Room, and therefore require little time to observe and record. This data would initially guide the assessment for rapid determination of Emergency Action Level (EAL) classification and Protective Action Recommendations (PAR). These instruments are also RG 1.97 qualified.

Page 1 of 2 Draft - For Information Only 3/11/03

Activity RG&E Coverage Dose Ginna currently utilizes the 1 Ginna Shift RP Technician to Projection/Assessment perform this function initially, as a collateral duty. The STA is also trained to perform the dose assessment calculation per EPIP 2-18. This calculation is done by the plant process computer, and is backed up with a simple calculation form.

Protective Action The Ginna determination of PARs is based initially on EAL Recommendation classification and wind direction by the Emergency Coordinator. The wind direction is available within the control room using RG 1.97 instrumentation.

Radiological Exposure The on shift Auxiliary Operators and Fire Brigade members Control receive basic radiation monitoring training. Radiation exposure monitoring has improved dramatically since NUREG-0654, Table B-i was issued. The on shift Auxiliary Operators, Fire Brigade, and Security Officers all use alarming dosimeters with dose and dose rate alarms. The Auxiliary Operators are also trained to use some portable radiation instrumentation. On shift personnel can self-frisk (using PCMs). The Ginna Shift RP Technician could assist with this activity as necessary.

Search and Rescue Search and rescue activities are not always required in an emergency. The dedicated Fire Brigade members are trained to provide this function as well as provide first-aid.

Decontamination Decontamination activities would not be required during the initial portion of an emergency except for personnel decontamination. In this instance, and the Ginna Shift RP Technician could assist with this activity as necessary.

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Attachment III Evaluation of NUREG-0654 Basis for Timing Staff Augmentation Draft - For Information Only 3/11/03

Evaluation of NUREG-0654 Basis for Timing of Staff Augmentation

Background

NUREG-0654, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants,Section II, Planning Standards and Evaluation Criteria, contains Table B-I entitled "Minimum Staffing Requirements for NRC Licensees for Nuclear Power Plant Emergencies". This table identifies staffing expectations for on-shift and at 30 minute and 60 minute intervals post accident. This attachment provides an evaluation of the overall basis for these times to better understand the technical basis behind them. Specifically, it is intended to identify those events which have the potential to result in an early radiological release to ensure these are evaluated in Attachment IV.

Section I.D.3 of NUREG-0654 states the following:

The range of times between the onset of accident conditions and the start of a major release is of the order of one-half hour to several hours. The subsequent time period over which radioactive material may be expected is of the order of one-half hour (short-term release) to a few days (continuous release). Table 2 summarizes the guidance on the time of the release, which has been used in developing the criteria for notification capabilities in Part II.

Table 2 GUIDANCE ON INITIATION AND DURATION OF RELEASE Time from the initiating event to start of atmospheric release Time period over which radioactive material may be continuously released Time at which major portion of release may occur Travel time for release to exposure point (time after release) 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to one day 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to several days 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to 1 day after start of release 5 miles - 0.5 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 10 miles - 1 to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> NUREG-0654 states that NUREG-0396 provides a planning basis for many of its requirements.

As such, NUREG-0396 and more current information were reviewed further.

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Evaluation Section HLI.C of NUREG-0396, Planning Basis for the Development of State and Local Government Radiological Emergency Response Plans in Support of Light Water Nuclear Power Plants describes the basis for the time factors associated with releases. Specifically, this report states that:

The planning time frames are based on design basis accident considerations and the results of calculations reported in the Reactor Safety Study(5. The guidance cannot be very specific because of the wide range of time frames associated with the spectrum of accidents considered. Therefore, it will be necessary for planners to consider the possible different time periods between the initiating event and arrival of the plume and possible time periods of releases in relationship to time needed to implement protective actions.

The Reactor Safety Study indicates, for example, that major releases may begin in the range of one-half to as much as 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> after an initiating event and that the duration of the releases may range from one-half hour to several days with the major portion of the release occurring well within the first day. In addition, significant plume travel times are associated with the most adverse meteorological conditions that might result in large potential exposures far from site. For example, under poor dispersion conditions associated with low windspeeds, two hours or more might be required for the plume to travel a distance of five miles. Higher wind speeds would result in shorter travel times but would provide more dispersion, making high exposures at long distances much less likely. Therefore, in most cases, significant advance warning of high concentrations should be available since NRC regulations (4,5) require early notification of offsite authorities for major releases of radioactive material. The warning time could be somewhat different for reactors with different containment characteristics than those analyzed in the Reactor Safety Study. The range of times, however, is judged to be suitably representative for the purpose of developing emergency plans. Shorter release times are typically associated with design basis events of much smaller potential consequences or with more severe Reactor Safety Study accident sequences.

The planning basis for the time dependence of a release is expressed as a range of time values in which to implement protective action. This range of values prior to the start of a major release is of the order of one-half hour to several hours. The subsequent period over which radioactive material may be expected to be released is of the order of one-half hour (short-term release) to a few days (continuous release). Table 2 summarizes the Task Force guidance on the time of the release. [Table 2 is the same as that in NUREG-0654]

Since the Reactor Safety Study is the primary basis for shorter time frames for major releases, a review was performed of WASH-1400. Section 5 of WASH-1400 contains discussions of reactor accident risks, including the definition of radioactive release categories. Table 5-1 (attached) includes a summary of these release categories with the time of the releases between one-half hour and 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> as described above. The only events which lead to a release within approximately one hour (i.e., < 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) following the accident were PWR Release Categories 8 and 9. As described within Section 5.2.1 of WASH-1400:

Page 2 of 4 Draft - For Information Only 3/11/03

In categories 8 and 9 the core doesn't melt, and only some of the activity in the gaps of the fuel rods is released. Category 8 involves gap releases with failure of the containment to isolate properly. In category 9, containment isolates correctly.

It is also noted that Table 5-1 shows these releases lasting only 30 minutes with no warning times for evacuations.

Table 5-2 of WASH-1400 compares the dominant PWR accident sequences versus the release categories. With respect to Categories 8 and 9, only large and small (2" to 6") LOCAs were identified. Consequently, the early releases were attributed to LOCAs. It is noted that WASH-1400 also evaluated LOCAs outside containment separately; however, no timing studies are provided.

Since WASH-1400 was developed over 25 years ago, a review was also made of more recent studies. Specifically, NUREG-l 150 and the basis for the Westinghouse Severe Accident Management Guidelines (SAMGs) were reviewed.

NUREG-1150, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants, includes an evaluation of Surry (which was evaluated in WASH-1400), Sequoyah, and Zion which are all PWRs of varying containment design. While there is no table in NUREG-1 150 comparable to Table 5-1 in WASH-1400, a review was made of the results to determine which scenarios contributed the most to early radiological releases. This review determined the following:

a.

Surry - LOCAs outside of containment dominate early releases results (88% of the internal large early release frequency or LERF).

b.

Sequoyah - LOCAs, both inside and outside of containment, and Station Blackout (SBO) dominate early release results (93% of LERF).

c.

Zion - LOCAs, both inside and outside of containment, and SGTR dominate early release results (94% of LERF).

It is noted that of these plants, Zion is most closely associated with the Ginna Station design of a large, dry containment.

With respect to SAMGs, the Executive Volume for the Westinghouse Owners Group (WOG)

Program Report, Revision 0, states the following:

Based on a survey of WOG utility member E-plan requirements, the TSC is not required to be functional until approximately one hour after the declaration to activate the TSC is given. Since the TSC may not be fully functional at the time of core damage for a limited set of "fast-acting" accident sequences, it was necessary to develop a limited set of guidance for use by the control room staff. The resulting control room guideline (designated SACRG-1) is limited in two respects: it only deals with a limited set of accidents and it only considers actions which need to be taken in the first hour or two of these fast-acting events. Severe accidents initiated by a large Loss of Coolant Accident (LOCA) or an Anticipated Transient Without Scram (ATWS) are the only two events Page 3 of 4 Draft - For Information Only 3111/03

which progress to core damage before the TSC would be functional.

Conclusion The timing considerations within NUREG-0654 Table B-I are primarily based on the presumption that major releases could occur as soon as 30 minutes following an accident as described within NUREG-0396 and taken fromWASH-1400. However, WASH-1400 shows that only accidents in which there is no core damage, have releases which occur at approximately one hour, with no warning time available for evacuation. Therefore, NUREG-0654 appears to be quite conservative in estimating the contribution of these early releases with respect to defining shift augmentation requirements.

More recent studies indicate that LOCAs, both inside and outside containment, SBO, large LOCAs, SGTR, and ATWS events have the potential to result in core damage within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> with subsequent early releases. With respect to SBO, large LOCAs, and ATWS events, the potential for large early releases is primarily dependent upon the status of the containment isolation functions. At Ginna Station, there are 94 mechanical containment penetrations broken down as follows:

a.

Normally Closed at Power 45

b.

Normally Open, Requires Active Valve to Isolate 8

c.

Normally Open, Uses Closed System to Isolate 17

d.

Essential System (Required Post Accident) 24 94 As can be seen, there are very few penetrations (8 of 94) which have the potential to result in a rapid release to the outside environment. These penetrations have redundant valves and are verified closed by Step 12 of procedure E-0 which occurs within minutes of the initiation of an accident. Consequently, the focus for accidents which have the potential for early releases should be on LOCAs outside containment and SGTRs. This is evaluated further in Attachment IV.

Page 4 of 4 Draft - For Information Only 3/11/03

0 Chapter 5 Reactor Accident Risks

5.1 INTRODUCTION

AND

SUMMARY

This chapter presents the results of the nuclear plant accident risk assessments.

These assessments, made according to the methodology outlined in Chapter 4, are fully described in the appendices to this report.

Although the information presented in this chapter derives, to some extent, from all appendices, the majority of the study results reported herein come from Appendix V - Quantita-tive Results of Accident Sequences and Appendix VI Calculations of Reactor Accident Consequences.

Section 5.2 describes how the radioac-tive releases associated with nuclear plant accidents are categorized and notes the principal characteristics of the different release categories.

Section 5.3 provides the probabilities associated with each of the release categories and describes the dominant accident sequences, i.e.,

those that contribute significantly to the proba-bility associated with each release

. category.

Section 5.4 discusses the initiation of nuclear plant accidents by external causes, noting that deliberate human acts are not accounted for in the risk assessment. While the initiation of core melt sequences by earthquakes, tornadoes, floods, aircraft impacts, and tidal waves is possible, the probabil-ities are expected to be low and their contribution to risk is predicted to be small compared to that of the dominant accident sequences discussed in section 5.3.

A discussion of nuclear plant accident risks, in terms of fatalities, injuries, long-term health effects, and property damage is provided in section 5.5.

Sections 5.2 -

5.5 provide summaries of the information that serve as the basis for predictions of the accident risks associated with a total of 100 nuclear power plants in the U.S.

These predictions are discussed in section 5.6.

5.2 RADIOACTIVE RELEASE CATEGORIES As set forth in Chapter 4, the quanti-ties of various isotopes released from the containment following a

given O

accident are calculated using the CORRAL Code described in Appendices V and VII.

Rather than calculate each of the approximately 1000 core melt sequences with CORRAL, it seemed desirable to reduce the number to be calculated to those necessary to adequately determine the accident risk.

To achieve this objective, the core melt sequences involved in the large LOCA event tree were carefully reviewed to identify those involving distinctly different physical processes and different combinations of ESF system failures.

5.2.1 PWR RELEASE CATEGORIES In reviewing the PWR accident sequences it was found that the large majority of the sequences in all the event trees involved quite similar processes.

It was thus possible to group the sequences into the one of 38 cases involving dif-ferences in timing or physical processes taking place during the accident.

Each of these 38 cases was then analyzed using the CORRAL Code to obtain the magnitude of radioactivity released to the atmosphere.

From these results it was found that the spectrum of releases could be well represented by a set of nine different radioactive release categories. These categories are shown in Table 5-1.

This table includes additional items of information which will be discussed later.

One of the largest releases, category 1, is associated with a potential steam explosion in the reactor vessel.

Such accidents would involve a large volume of molten U02 falling into a pool of water in the bottom of the reactor ves-sel, and becoming finely dispersed in the water to mix efficiently enough with it to produce a steam explosion.

This could potentially release large enough amounts of energy to rupture the vessel and, in some cases, even the containment as a result of missiles generated by the vessel rupture.

Because of the heavy concrete shielding around the reactor

vessel, a missile having sufficient energy to rupture the containment would almost certainly go up through the containment dome.

The one half of the molten core that was finely dispersed in water is assumed to be ejected into the containment oxidizing atmosphere, thus producing a large release, energetically discharged, from the upper part of the containment.

Although such a release is predicted to be very unlikely, it cannot be ruled out completely on the basis of 'w,;A -tJ - / 0cc

A present evidence.

This involves failure of the removal systems that are containment.

category also radioactivity located in the The category 2 releases are also associ-ated with core melt and basically involve failure of radioactivity removal systems to operate, followed by rupture of the containment caused by hydrogen' burning and steam over-pressure.

Category 3 includes some of the cases that are similar to those in categories 1 and 2, but involve partial success of radioactivity removal systems.

Category 4 involves core melt cases in which the containment is not fully isolated and the containment radioactivity removal systems have failed.

Category 5 is similar to 4 except that radioactivity removal systems are operating.

Cate-gories 6 and 7 cover cases in which the molten core melts through the bottom on the containment, with and without radioactivity removal systems operating, but the above ground part of the con-tainment remains intact.

In categories 8

and 9 the core doesn't melt, and only some of the activity in the gaps of the fuel rods is released.

Category 8

involves gap releases with failure of the containment to isolate properly.

In category 9,

the containment isolates correctly.

Considerable effort was spent in trying to identify possible accidents in which a release larger than that of category 1 might be produced.

The possibility of processes that might physically eject the entire core outside the containment was examined.

No such process could be identified that appeared to be consist-ent with the energy available and the physical constraints of the containment.

Even if such an event were to occur and the core melted outside of containment, a release larger than that of category 1 would not be expected to occur.

This is so because these accidents already involve a

large energetic dispersal of the molten fuel in the form of small particles where the large surface to volume ratio enhances both fuel oxidation and the release of radioactivity from the fuel.

for removal of radioactivity is largely ineffective in a number of the core melt cases.

Thus the principal mechanism for removal of radioactivity is natural deposition on the surfaces inside the containment and the reactor building.

For these reasons, the BWR release cate-gories are different than those for the PWR.

As in the PWR, the release categories were determined from CORRAL Code runs of those accident sequences involving dif-ferent physical processes.

Twenty-three CORRAL runs were

made, and subsequent analyses identified the five release categories shown in Table 5-1.

As in the

PWR, category 1 involves a steam explosion in the reactor vessel in which about half the core is involved.

The steam explosion ejects this half of the core from the containment.

The resulting exposure of the finely dis-persed molten fuel to an oxidizing atmosphere results in a very large release of radioactive material to the atmosphere.

Category 2

involves a

core meltdown-after containment overpressure rupture caused by loss of decay heat removal systems.

In this category a limited amount of deposition of the radioactive materials occurs and the release is made directly to the atmosphere.

The magni-tude of release is roughly comparable to category 1 for a number of the isotopes.

Category 3 covers overpressure ruptures of containment similar to category 2 but in this category the radioactive materi-als released from the core escape through the reactor building to the atmosphere.

The radioactive release magnitude is smaller than category 2

releases since deposition and some scrubbing action by the torus water enhances retention of the radioactivity.

Category 4 covers the cases in which the containment fails to properly isolate and the leakage is enough to prevent containment overpressure rupture.

In this category, the magnitude of radio-activity release is significantly reduced by additional deposition in the containment due to the longer release times and by deposition in the reactor building.

In some

cases, processing through gas treatment systems achieves further reductions.

Category 5 covers the case where the core does not melt and a small amount of I

a 5.2.2 DWR RELEASE CATEGORIES The paths to release of radioactivity in a BWR are quite different than for the PWR.

Although the BWR has containment

sprays, they are not designed as ESFs and are not credited for removal of radioactivity.

Further, the vapor sup-pression system that has some capability a 4fP4 - 1A/Yo

a TABLE 5-1

SUMMARY

OF ACCIDENTS INVOLVING CORE DURATION WARNING ELEVATION CONTAINWENT POAIIYTIME OF or TIME EOR OF ENERGY REL£ASE per RELEASE RELEASE EVACUATION RELEASE REE FRACTION OF CORE INVENTORY RELEASED CATEGORY Reactor-Yr (Er)

(Hr)

(Hr)

(Meters) tlO6 Btu/Hr) Xe-Kr Org. I I

Cs-Rb Te-Sb Ba-Sr Ru b)

La' PWR 1 9xlO 7

2.5 0.5 1.0 25 5 2 0 (d) 0.9 Ex103 0.7 0.4 0.4 0.05 0.4 3X10 3 PWR 2 aXio_6 2.5 0.5 1-0 0

170 0.9 7X10-3 0.7 0.5 0.3 0.06 0.02 4x10-3 PiR 3 4x10 6 5.0 1.5 2.0 0

6 0.8 6X10 3 0.2 0.2 0.3 0.02 0.03 3xl03 PWR 4 5X10 7

2.0 3.0 2.0 0

1 0.6 2x10 3

0.09 0.04 0.03 5x10 3

3x10 3 4x1O0 4

PWR 5 7X10 7

2.0 4.0 1.0 0

0.3 0.3 2xlO 0.03 9xl0 3 5xlO 3

lxlO 6xlO 4 7xO1 PWR 6 6X10 12.0 10.0 1.0 0

N/A 0.3 2X10 8X10 4

8X10 4

x10 3

9XO 7x10 5

1X10 PNR 7 4xO 5

lO.O 10.0 1.O O

N/A 6X10-2x10-2xlO lx10 2xlO 1x0l ls10X 2x10 7

PWR 8 4x10 0.5 0.5 N/A 0

N/A 2X107 5X1O 6 1X10 4

5X10 4 lxlo06 Wo-8 o

PWR 9 4x10 4

0.5 0.5 N/A 0

N/A 3x106 7X1O 1X1 6x10 7

1X10 9

1X10 0

i BWR 1 lxlO 2.0 2.0 1.5 25 130 1.0 7x103 0.40 0.40 0.70 0.05 0.5 5X10 3

BSR 2 6X10 30.0 3.0 2.0 0

30 1.0 7x10 3

0.90 0.50 0.30 0.10 0.03 4x10 3

BWR 3 2x10 30.0 3.0 2.0 25 20 1.0 7xLO 3

0.10 0.10 0.30 0.01 0.02 3x10 DWR 4 2xlO 5.0 2.0 2.0 25 N/A 0.6 7X10 SX10o 5X10 3

4x10 6x10 4

6x10 4

1X10-DWR 5 1X1O 3.5 5.0 N/A 150 N/A 5x10 2x10 6X10 1 1 4X10 89X1012 OxlO 14 0

0 (a) A discussion of the isotopes used in the study is found in Appendix VI.

Background on the isotope groups and release mechanisms is found In Appendix Viz.

(b)

Includes Mo, Rh, Tc, Co.

(a)

Includes Nd, Y. Ce, Pr, La, Nb, An, CM, Pu, Np, Zr.

(d)

A lower energy release rate than this value applies to part of the period over which the radioactivity is being released.

The effect of lower energy release rates on consequences Is found in Appendix VI.

6 VASI -/qoc

X

,TABLE 52 -PWR-DOINANT-ACCIDENT.. SEUNCSv.

rLAE CATEGORIES SAECA2RE R~Cr

• 1 oe I

. 1

• 1 0

A,

• 1 0

' ~ i : A-T R G M O u

'lx A

'S llV

?-dA3Ft.

• '2la~

1*1 1

At-1 o..

9*10

'SI4LLL LA sd S

-uS~..

S-G 10

-Y-

• N*.

2 3y S

-a

! 20012 SOB D

S

-c 1*1 2H*1-I

• 0

11.

218 2

2T SN -l 3

e

~

SC.t S0I-527C Pr~ab11t~*

1*0 3*0 3*0 2*10 2*10 2a0 2*1 5~~+/-0

~ *f~

110 r0 ACy a

2 10

• 0 11

.ASAO8VEASI.57 AWV* *2*

1*10 a

a7

z.

211*-

7*L!

-7

-6 27

-B

-7 450%= VALUE)

,g W"6-S Ta1r U-910 8.D*

0w J*O 43

.377 A,

£1A*J

/(6ci AU-

6 KEY TO PWR ACCIDENT SEQUENCE SYMBOLS I

Q T

V a

T 6y

£5 A -

Intermediate to large LOCA.

B -

Failure of electric power to ESFs.

B' - Failure to recover either onsite or offsite electric power within about 1 to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> following an initiating transient which is a loss of offsite AC power.

C -

Failure of the containment spray injection system.

D

- Failure of the emergency core cooling injection system.

F Failure of the containment spray recirculation system.

- Failure of the containment heat removal system.

I

- Failure of the emergency core cooling recirculation system.

Failure of the reactor protection system.

L - Failure of the secondary system steam relief valves and the auxiliary feedwater system.

I-Failure of the secondary system steam relief valves and the power conversion system.

2 - Failure of the primary system safety relief valves to reclose after opening.

4 Massive rupture of the reactor vessel.

A small LOCA with an equivalent diameter of about 2 to 6 inches.

A small LOCA with an equivalent diameter of about 1/2 to 2 inches.

Transient event.

LPIS check valve failure.

- Containment rupture due to a reactor vessel steam explosion.

- Containment failure resulting from inadequate isolation of containment openings and penetrations.

- Containment failure due to hydrogen burning.

- Containment failure due to overpressure.

Containment vessel melt-through.

I

'EY TO TABLE 5-2 a VWAf S -N06

Attachment IV Evaluation of RG&E Current On Shift Staffing Draft - For Information Only 3/11/03

Evaluation of RG&E Current On Shift Staffing Attached are the results of the on shift staffing evaluation that was performed prior to RG&E providing a dedicated Fire Brigade to replace the use of Security personnel as brigade members in 2002. The scenarios evaluated were based on their high impact on the shift organization and include beyond design basis events. The evaluation was performed using a PC based version of the Ginna simulator and the Ginna procedures. The following are the parameters and assumptions used to complete the matrix.

1.

Minimum staffing level on a weekend.

2.

Worst case scenario requiring actions to be taken per the Response Not Obtained (RNO) column of the EOP's.

3.

Off-site fire department is conservatively assumed to not be available until the 45 minute point.

4.

Both fire brigade captains are involved with the fire until it is out (i.e. 45 minutes).

5.

Time for S/G sample is 90 minutes, with a 60 minute count.

6.

Time for a full RCS sample is 120 minutes.

7.

EPIP classification is not made until directed by procedure.

8.

The top priority of the security force is a security threat.

9.

The Ginna Shift RP Technician monitors the fire brigade when in a controlled area, if not involved in dose assessment activities in the control room.

10.

Auxiliary Operators cannot perform chemistry analysis per the EOP's.

Definitions:

SS Shift Supervisor (SRO)

CRF Control Room Foreman (SRO)

HCO Head Control Operator (RO)

CO Control Operator (RO)

STA Shift Technical Advisor PRIMARY Auxiliary Operator (primary plant)

SECONDARY Auxiliary Operator (secondary plant)

EXTRA Auxiliary Operator (extra)

SHIFT TECH Shift RP Technician Page 1 of 13Draft - For Information Only 3/11/03

Grid Failure, Direct entry Into ECA-O.O, Security available Ss CRF HCO I

CO STA PRIMARY SECONDARY EXTRA SECURITY SHIFT TECH NOT COVERED EPIP CR

• Procedure

Prmr, Secondary,lmm CR, Lead ECA0 0 Immediate Aate Actions AB, Tour TB, Tour TB, AVT 00 CA00 Actions edaeAtos; CRLead CR, Procedure

Primary, Secondary AB Tour ResetngTDAFWP CR,

• 05 ECA-0 0 ECA-0 0 ECA-0 0 governor valve Communicator CR Lead CR, Procedure Attempt to Attempt to CR, Locally start LocaCy start Diesels

10 C ECA-0 0 restore power restore power Monitoring Diesels Communicator CR, Lead CR, Procedure Pull stop Open Reactor CR, Isolate RCP Backup cooling to CR, Open vital area Site Area

15 ECA-0 0 equipment Protection doors Monitonng seals TDAFWP Communicator doors Emergency CR, Procedure

Primary, Secondary, CR, Isolate RCP Locally isolate CR, Open vital area Locally check Elec. Locally monitor

.20 Le ECA0 0 ECA- 0 ECA-0.0 Monitoring seals valves Communicator doors steam d

ine DC power supply CR, Procedure Primary.

Secondary, CR, Degass ATT.

CR Locally check 25 CR, Lead ECA-0 0 ECA-0 0 ECA-0 0 Monitoring Generator Faulted/Ruptured Communicator radiation CR, Procedure Primary, Secondary, CR, Degass A

li.

CR, AO: Locally operate CR, Lead ECA-0.0 ECA-0 0 ECA-0 0 Monitoring Generator FaultedlRuptured Communicator ARrbs

• 30

_S IG_

CR ed CR, Procedure Primary, Secondary, Energize Degass ATT.

CR, Sample RCS and A:E-F.

CR ed ECA-0 0 ECA-0 0 ECA-0.0 source ranges Generator Faulted/Ruptured Communicator PZR for boron A:E-F.

.35

_SIG_

CR, Lead CR, Procedure Primary.

Secondary, CR, Degass Locally Isolate CR, Sample RCS and AO: ER-AFW.1 40 ECA-0.0 ECA-0.0 ECA-0.0 Monitoring Generator Ci1CVI valves Communicator PZR for boron CR, Lead CR, Procedure Primary, Secondary, CR, Locally Isolate CR, Sample RCS and 45 ECA-0 0 ECA-0.0 ECA-0.0 Monitoring CIICVI valves Communicator PZR for boron At this point, we would be looping in ECA-O.O until power is restored

.55 56 0

zr5 id 13

Security event in Switchyard, Loss of circuits 767 and 751, Secunty not available SS CRF NCO Co STA PRIMARY SECONDARY EXTRA C PoedrePImmdarye Secondary~lmm CR, Lead ECA-,0 P

Immedeate udlat Actmons AB, Tour TB, Tour TB, AVT ECAO 0 ActioActon CR. Prcdur Primay Secondary

Reseting, CR, Lead 0er ay y

'IO AB, Tour TDAFWP 05 ECA-0.0 ECA-O 0 ECA-0 0 Communicator CR, Lead CR, Procedure Attempt to Attempt to CR, Monitorng Locally start Locally start CR,

10 ECA-0 0 restore power restore power Diesels Diesels Communicator CR, Lead CR, Procedure Pull stop Open Reactor Moi IsolateRCP Backuproobngto CR,

.15 ECA-O 0 equipment Protection doors seals TDAFWP Communicator CR, Lead CR, Procedure Primary, Secondary, CR Monitoring ocally Isolate CR,

20 ECA-O 0 ECA-0 0 ECAr0 0 valves Communicator CR, Procedure

Primary, Secondary, Degass ATT.

CR CR, Lead ECA-OO ECA-OO ECAOO CR, Monitoring tenerator Faulted/Ruptured

25 ECA-0 a CA-0.0

_A-0 0

Gnerator SIG Communicator CR, Procedure

Primary, Secondary, Degass A1T.

CR CR. Lead CR, Monitoring Faulted/Ruptured

30 ECA-0 0 ECA-0 0 ECA-0 0 Generator SIG Communicator CR, Procedure

Primary, Secondary, Energize source Degass ATT.CR

35

, Lea ECA-0 0 ECA-0 0 ECA-0 0 ranges Generator Faulted/Ruptured Communicator CR, Lead CR, Procedure

Primary, Secondary, C Montorin Degass Locally Isolate CR, 40 ECA-0 0 ECA-0 0 ECA r0 CR, oniong Generator CIICVI valves Communicator CR, Lead CR, Procedure Primary, Secondary, CR, Monitoring Localy isolate Co c

.45ECA0 0

ECA0 0 ECA0 0CI/CVl valves ICommunicator

-r-

-I NUT I.UVERKU EI'lP' Security:Open vital area doors SEte Area Emergenr y

.ocally check Elec: Locally monitor DC power

team line supply Security.Open vital area adlation doors

.ocally check steam line radlation__

AO: Locally operate ARV's 3ample RCS and AO, ER-AFW I PZR for boron

'ample RCS and AO: ER-AFW I

~IZR for boron 50 At this point, we would be looping In ECA-O.O until power is restored

55 1

__I ___I6______I0___

Pr,5,- I C If 15

Explosion In Screen House, Loss of Buses 17 & 18, Security available.

I SS CRF HCO Co STA PRIMARY SECONDARY EXTRA SECURITY SHIFT TECH.

NOT COVERED EPIP CR, roceurePrimary, CR,,LPad mmediate econdary,lmm

-00CR, Lead ECR Proc edure ediate Actions AB, Tour TB, Tour TB, AVT

00 EAActions______

CR ed CR, Procedure Primary, Secondary, Lff ReAsTurtDAnP CR, 05 ECA-0.0 ECA-0 0 ECA-4 0 W

vauve Communcator CR, Lead CR, Procedure Attempt to Aempt CR, Monitoring Locally start Locally start CR, 10 ECA-0 0 restore power restore power Diesels Diesels Communicator CR, Lead CR, Procedure Pull stop pn o

CR, Motorng solate RCP Backup cooling to CR, Open Vital area Site Area CR, ead ECA-0 0 equipment dottons CR oioigseals TDAFWP Communicator doors Emergency CR, CR, Procedure Primary.

Secondary, Locally Isolate CR, Open vital area Locally check Elec. Locally monitor

20 Lead ECA-0.0 ECA-0.0 ECA-0 0

SesCR, Monitoring ves Communicator doors steamine DC power supply

30_

CR,_____

radCPocdriationA00 CR oiongGnrtr Fale/utue omncao OLcal prt CR, Procedure

Primary, Secondary, Degass ATT.

CR, Locally check

.25 ECA-0 0 ECA-0 0 ECA-0.

CR, Monitoring Generator Faulted/Ruptured Communicator line

__25___

___SIG

__ radiation CR, Procedure Pnimary.

Secondary, Degass ATT.

CR oclyoprt CR, Lead ECA-o.0 ECA-0 0 ECA-0 0 CR, Monitoring Generator Faulted/Ruptured Communicator AO: LR-Alypa

30

_SIGAR CR, Procedure

Primary, Secondary, Energize Degass ATT.

CR, Sample RCS CR, Lead EA00 EA0 EA00 sucragsGnatr Faulted/Ruptured Comnctrand PZR for AO: ER-AFW.1

.35 EC1 C-C-

suc agsGnrtr SIG Comnctrboron CR, Procedure Primary.

eodrDgs oal slt R

Sample RCS CR, Lead EC-C-

SEConary 0

oioigGenerator Ci/CVI valves Communicator and PZR for AO: ER-AFW.1

4 CR, Lead ECA-0 0 ECA-0 0 ECA-0.0 CR, Monitoring Locally isoate Comunator and PZR For

__45_________boron At this point, we would be looping in ECA-O.O until power is restored

55

60 reu e-q o

/5

Security event In the Screen House, Loss of Buses 17 & 18, Security not avaflable I

AAii-AA1

)L)

I

_M.P S

TN SS1 I

CRFt i

HCO G(;

I STA I

PRIMARY I SECO1N.AR.Y I

EX.TRA 1-CR CR, Procedure Immediate Sendary,lmm _

00 Lea WEA-a a Acmedons ediate Actions AB, Tour TO. Tour TOAV 00 TBtions CR ed CR, Procedure Primary.

Secondary, Reseting CR,

.05 CR, Lead ECA-0 0 ECA-0 0 ECA-0 0 A

AB. Tour TDAFWP Communicator

_ _ __ _ _ governor valve C m u i ao CR, Lead CR, Procedure Attempt to Attempt to CR, Monitonng Locally start Locally start CR,

10 ECA-0 0 restore power restore power Diesels Diesels Communicator CR, Lead CR, Procedure Pull stop Open Reactor OR Monitofin Isolate RCP Backup cooling to CR,

16 ECA-0 0 equipment Protection doors C M t

seals TDAFWP Communicator CR, Lead CR, Procedure

Primary, Secondary, CR, Monitoing valses Communicator

'20 ECA-0 0 ECA-a S

0 Monitoring L

latee Communicator CR, Procedure Primary, Secondary, s

Degass ATT.

CR, 5CR, Lead ECA-o 0 ECA-0 0 ECA-0 0 re n

g Generator Faulted/Ruptured Communicator CR, Procedure Primary, Secondary, R

RDegass A

OR, 30 ECA-0 0 ECA-0 0 ECA40 0 Generator CIICVI valves Communicator CR, Lead CR, Procedure Pnmary-Secondary, Energize source Degass Faute ed Communicator 35 Lad ECA-0 0 ECA-0 0 ECA-0 0 ranges Generator Sa/G fupue omuiao CR ed CR. Procedure Primary, Secondary.,

R ontrn Degass Locally Isolate CR,

-40 ECA-0 0 ECA-0 0 ECA-0 0 Generator CI/CVI valves Communicator CR, Lead CR, Procedure Primary, Secondary, CR, Monitoring Locally Isolate CR, 45ECA-0 0

ECA-0 0 ECA-n 0 CIfCVI valves Communicator lSHIFT TECH l

NOT COVERED I

EPIP Security:Open vital area doors Site Area Emergency steam che Elec: Locally monitor DC power supply radiabon Security Open vital area doors Locally check steam line radiation AO: Locally operate ARVs Sample RCS and AO: ER-AFW.1 PZR for boron Sample RCS and PZR for boron Sample RCS and PZR for boron

• 50 At this point, we would be looping in ECA-O.O until power Is restored 160 f'4l5e 6 ~

5&/

Fire in the Aux. Bldg, Security available (ER-FIRE.3)

_SSE CRF HCO CO I

STA PRIMARY SECONDARY EXTRA SECURITY [ SHIFT TECH.

NOT COVERED EPIP Control Room Control Room Control Room Control Room Actions Actions Actions ActionsABTorB ouTAr Control Room Control Room Control Room Control Room Fire Brigade Fire Bngade CR, Monitoring Fire

05 Actions Evacuation Evacuation Evacuation Captain Captain Communicator Fire Brigade Brigade Control Room Trip Both MG Locally verify Fire Brigade Fire Brigade CR.

Monitoring Fire Actions sets RX triplMSIV's Don SCBA WA DIG Room Captain Captain Communicator Fire Brigade Brigade

10 Shut Control Room Trip both RCP's Isolate l.A. to Close MOV-856 Start A Fire Brigade Fire Brigade CR, Fire Brigade Monitoring Fire

15 Actions nCNMT Captain Captain Communicator Brigade Control Room Trip Intake Start and contro a bus 14 MonitorA!DIG Fire Brigade Fire Brigade CR, Fire Brigade Monitoring Fire Alert

20 Actions heaters TDAFW pump Captain Captain Communicator Brigade Control Room Start Diesel air Start and contra Energize bus 14 Verify Natural Fire Brigade Fire Brigade CR, Monitoring Fire

25 Actions compressor TDAFW pump from the DIG Circulation Captain Captain Communicator Fire Brgade Brigade Start Diesel air Locally Line up Fire Brigade Fire Brigade CR, Monitoring Fire and start W irAB'gd

30 compressor Charging pump Captain Captain Communicator re Brigade Brigade Locally Line up Fire Brigade Fire Brigade FiretoBngader and start W Captain Captain nre Bigade Brigade

35 Charging pump Ca Unload bus 16 Fire Brigade Fire Brigade Monitoring Fire

40 Captain Captain Brigade Bria Fire Brigade Fire Brigade Fire Brigade Monitoring Fire

45 Captain Captain Brigade 55

==

==

===-

-=

.6 0 pe'g e a

/5

Security event and subsequent fire in the Aux. BLDG, Secudty not available.

SS CRF HCO Co STA PRIMARY l SECONDARY [

EXTRA Control Room Control Room Control Room Control Room X

Actions Actions Actions Actions AB, Tour TB, Tour TB, AVT Control Room Control Room Control Room Control Room Fire Brigade Fire Brigade CR,

05 Actions Evacuation Evacuation Evacuation Captain Captain Communicator Control Room Trip Both MG Locally venfyFieBgae irBiad R

RX Actionsr thlp/MSIVs Don SCBA A DIG Room Captain Captain Communicator Acton stsShut_______

Control Room Trip both RCPs Isolate I A. to Close MOV-856 Start 'A' DIG Fire Brigade Fire Bngade CR,

15 Actions CNMT Captain Captain Communicator Control Room Trip Intake Start and Fire Brgade Fire rlgade

20 Actions healers control TDAFW Unload bus 14 MontitorADlG Fir CiB CR i

0 Captan Captain Communicator Control Room Start Diesel air cotrol ad Energize bus 14 Verify Natural Fire Bnigade Fire Brigade CR,

25 Actions compressor control TDa from the D/G Circulation Captain Captain Communicator Control Room Start Diesel air Locally Unire Brgade Fre Bfigade CR,

40 Actions compreshutor n

c o

l d sChtartown m Captain Captain Communicator Control Room Cooldown to Cooldown to ocally Un u Verify Natural Fire Brigade Fire Brigade CR, Actions cold shutdown cold shutdown and start 'A Circulation Captain Captain communicator

35

_______Charging pump_______

Control Room Cooldown to Cooldown to Unload bus 1 Monitor AD/G Fire Bngade Fire Brigade CR.

50 Actions cold shutdown cold shutdown Captain Captain Communicator Control Room Cooldown to Cooldown to Cooldown to Venfy Natural Fire Brigade Fire Brigade CR,

45 Actions cold shutdown cold shutdown cold shutdown Circulation Captain Captain Communicator Control Room Cooldown to Cooldown to Cooldown to Monito rA"DG Fire Brigade CR,

50 Actions cold shutdown cold shutdown cold shutdown Captain Communicator Control Room Cooldown to Cooldown to Cooldown to Verify Natural Fire Brigade CR, Acins cold shutdown cold shutdown cold shutdown Circulation Captain Communicator Control Room Cooldown to Cooldown to Cooldown to Monitor 'ADIG FrBigde CR,

50 Actions cold shutdown cold shutdown cold shutdown Captain Communicator SHIFT TECH I NOT COVERED EPIP Aonitoring Fire Fire Brigade irigade Aonitoring Fire Fire Brigade 3rgade Fire Brigade Monitoring Fire Fire Brigade 3rigade Aoriltoring Fire Fire Brigade General Emergency 3ngade Monitoring Fire Fire Brigade 3rigade ionitoring Fire Fire Brigade 3rIgade

__ontonng Fire Fire Brigade 3igade vionitoring Fire Fire Brigade 3ngade

/lonitoring Fire Fire Brigade 3rigade 104~p 7ef /I

0

-11 45 So6 LOCA Ouhtlde CNMT SS CRF HICO CR. Lead CR. Procedure orrary.

o E-Irenwdiare

~

Adbn s CR L CR. LrCeduOP Primary, E-0-0 0

CR. Lead R.

PrImaPr y,

E 0-0 0

CR, Lead CR. Procedure Prmary.

E E-0 0

CR, Lead Transition to Tran t

ECA-12 ECA-12 CR, Lead ECA 12 CR. Procedure CA12 ECA.12 CR, Lead CR, P12dur CR Procedure L

ECA.1.2 ECA.12 CR. Lead CR. Procedure CR. Procedure ECA.12 ECA-1.2 CR. Load CR. Procedure, CR, Procedure ECA-1.2 ECA-12 CR, Lead Transition to Trnstion to EC 1 Elr LOCA OUtISWe CNMT wMi a large tir I

r.

vv f-RMR Secondaryln1m edlate Adiorns Se nudary, E-0 Secondary,

_E-Transiton to ECA-12 CR, Procedure ECA t12 CR. Procedure ECA 1 2 R Procedure ECA.12 CR, Procedure ECA.12 rnshlon b Enuir I

0-I L ocally clae r CNtT isolaton LoGally rclse CR, Monitorihg CNMT solation atoes CR. Modntoring Local sdisutbrom ar00MUS CR. Monitorng CR, Monitorong CR.

Monotoring CR, Monitoring CR, Mon"erle 0

I SECONDARY ATP $014 mATr, SD-l

_Air SDJ.1 ATT S0.1 3_I 1 j EXTRA SECURITY Locally trip fth reactur lCR l

_I CoImmunirat1r CR, ICommunicator CR.

Comnarndeaor CR, lCnromrobrnior11 CR.

lCosnorricatlorl

lCR, Communicator CR.

Communnerc l

I' ICR.

Commuontalor l1 MM-F TF2.S

!_orCVRD Sample SIG'S SalPte SWS Sample S/G's Sample SIGS Sample clGlI Survey Steam Lines Control Acress to The AUX BLDG EPIP I

Alert DeCLard Local Radiation rEPIP

-I I

r l

55 iS CRF l CR Lead CR. Prowdun l 101lCRLead Ir4 lCR Lead CR, Procedure l CR, Load CR. Procedure 0.0 l CR, Load lCR. Pro~cedure

20 ECA4I2 CLad C.Procedure 35 lECA-1 2

lCR, Lead CR. Procedure Io ECA-1.Z CR ed CProcedure lCR.,Lead ITrarrubtto l

l55 1

~

1 50 610

_L HCO l

C Primary.

Scedri trmmediates Scnari Actions rr di leAciorn

Prlrraoy, E Secondary.

O E-0 Primary.

E-Secondary, P

Erray

- Seorirrrary.

Trnuiton to Transition to CA12 ECA.1.2 CProcedr CR. Procedurre EC-2 ECA-12 CProcedure CR, Procedure 1

ECA-1 2 CR, Procedure CR, Procedure ECA-1.2 ECA-12

,CR, Procedure CR, Procedure ECA-12 lECA-1.2 Trarr n to lTransltan to E-1 lE-1 7_

CPRIMARY SDECODAR EXTRA

,AS ATour TEhTOW La11)

• thye reactor Phe Brogade Fire Brigade CR.

Captain Captain Cormrnstrarcar CR, MntorftkV Fire Brigade Fire Brigade CR.

Captair Caplain Communicator Captan Capairs Cionmmuedain CR.

FrBrgd FieBgade CR.

Catn Captain Commenicatar Mor, ta FeBrigade FieBrae CR, CpanCaptain Communicator CMniirr reBrigad Fir Bge CR.

Captain Cpan Cmrrlau Mon CataeCapt ain Communrcator lCRMotoin lrnedoad alh do d CR lCR. Monitoring Fire Brigade lFireBrigade lCR, Captai Capban Communicator CR. Montaoring Flre Brigade Fire Brigade CR, Captain Captain Commurncatro l

-1 SECU lteBrigade l

LoycbeCNUAT l

lItsolation vaes lFre riadt l

Loai~etseCNMT l

g lIsolatin valves Fhr Brigad-e Samplett OOSB rrv~ ra

[Fndado lSample/'s lTTS-lieBriga~de lSampie StG r

EPI I- -

j Fir. Brigade L-I P5,.. g.

/5

Design Basis SGTR I

s

-*c I

.CO I

-C STA l

I--

CR. Lead ER Procedure Iiate 10CRk GC Inr~meredlae E

15 C, k

Tr~si-An b Ansiionsb CR, Lead CR Pocedure Primary, 3

E.GE 0

CR, irad Procedure Prim: y.

3 10 0-0 CR, Lead Transition to sntOWOo t

15 R

-3 E-3 CR, Lead CR. Procedure CR. procedure 20 E-3 E-3 CR. Lead CR. Procedure CR, Procedure 25 C-3 0.3 CR, Lead CR Procedure CR, Pocdure E-3 E-3 CR. Lead CR, Procedure CR, Procerntu

0. E3 E-3 CR, Lead CR. Procedure CR. Procedure 00.E3 0-3 CR, Lead Secondlary imrnm P~rV-edlate Acllortr

• Secondary.

LECadcose E.G CN1.IT Isolation E44 rubei Secondary, CR, Mon/urog o

E.0CG oilrn CNMT iolaloft Trarrsition to Lclaiute 0.3 CR, Monitodrng Locat r lsulme CR. Procedore CR

=

3P d

CR, Mon""

ATT, Ruptured CR, ProcedureCRMoioi Arupre i-3 CR, Mo SIGornrT R

rd C Procedure ArT. Raptured CIR.

Monitodrqe CR, Prorcedure 0.3 CR. Mortrn" -

l 4__.L
IKA Survey Steam Isolate Bow from Ruptured SIG Isolate Bow from Ruptured S/G ATT SD-1 ATT SD-l AT. SD-I I

7 SECURITY-lLOr taWYipi Locally trip em reacror CR Communrcaorl CR.

Conmrunicalor Communicator CR, IConrunorrcalorl ICR.

lCommuslcalorl CR, COrrmunoicatorI lCR.I

[Commronuotoc 1

I SHIFT TECH NOT COVERED lSample S/G's lSarople S/G'e Sample S/Gs Sample SIG's Sample S/Gas Sample l/Ga I

lSampie S/0'sl

-Mi lAhrt Deceared

_j Desin Basis SGTR coincedeon wI a large B I

I I

STSsur~o S

lCR, Lead 05 10

.25 35 CR, Lead CR. L.a, CR, Lead CR Lead CR. Lead ICR. Lead I

ICR. Lead I

lCR.

Lead CR, Lead E.G Imediate Seceedary,tmm AlIoE.

edilpe Actionr CR, Procedure Primary, Secondary.

EG 0

E.G CR, Procedure Primary, E Secrrndary, E.G 0

E.G lTransetu on b

Trarrotlor no b

Tramrrolon bt I

E-3 0-3 0E.3 CR.Procdure lCR, Procedure CR. Procedure E.3 lE4 0E3 CR, Procedure CR, Procedure lCR. Procedure E-3 0-3 E-3 CPrOcedure CR. ProceueC.Prcde E-3 0.3 0-CR, ProedureC CR P

rocedure E-3 E-3 03 lCR, Proceduret CR, Proceldure jCR, Procedure 0-3I 0_3 0.

l AB Tori lCR, Mo retriq ades

- ~~captain CR.

oolurlg Fre rigde Captain CR. Moniturino Fire Brigade Captain CR, Monitodno fIre Brigade Captain CR. Montrorlng Captai ade CR, Moulorlng Fi I

Brigade CaptaIn

'R, Monitoring Fime Brrgade Capltain I

- - ------ I l

ITB, Tour Fire Brigade Captain Fire Brigade ICpapirt FmBrigade Fire Brigade Captain Fire Brigade Captain FIre Brigade Captai r~imBfriade

.CtainI Loca-y trip toe

ICR,

{Corororrlcabr CR.

Commurdck I

CR,.

ConmunicatorI CR.

Commnknorl

ICR, lCoemericarorl CR.

ICommunicatoI Communicalor Fire Brigade Ftre Brigade ftre Brigade Fke Brigade Fire Brigade Fire Brigade Fire Brigade Sample StGa Sample S/G' ISaupl S/GoS Sample SIG's I

Sample S/Go; Sample S/Gos Loally close CNUT lsolarron valves Localiy cio# CNMT Isolaou rvralves iLoca adisutnrent of glae to from Raptured l30&, SD ve Sc Isolatetewfrom Ruptured S/G, ATT. Ruptured BIGl Isolaie tret rrmm RapturerrdI SIG. ATT: Ruptured SIG, ATT SD-I luar

_lo 1 upue

,r HA j

ECURI iC Alert Delared lR. Monuoring I

IFirs Brigade lCaptain nFre Brigade Captain CR, Communicator FAm B.gade Lt 5e I

I 50 I

~

-I i

i -

i Prx
e q o

1

Evaluation of On Shift Staffing Scenario Items Not covered Grid Failure, Direct entry into ECA-0.0, Security available Item Not Covered Discussion Electrician: Locally monitor DC Not a critical activity initially, DC voltage indication is availabl power supply control room. Batteries are good for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

e in the AO: Locally operate ARV's ARV control is available in the control room. Local control is used if necessary to cool down the primary system to delay the potential for RCP seal LOCA. This step does not occur until 30 minutes at which time the 30 minute responder becomes available, enabling transfer of duties.

I AO: ER-AFW. 1 Only required if Condensate Storage Tank Level is less than 5 feet. Should not be a concern for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Note:

Primary AO is degassing the generator which is not initially time critical.

Security event in Switchyard. Loss of circuits 767 and 751, Security not available I

I.

I Item Not Covered I

TDiRC11

.iien Security: Open vital area doors Electrician: Locally monitor DC power supply AO: Locally operate ARV's Current commitment is to open the doors within 30 minutes for SBO. The actual activity could be performed by the AOs, but this would be a Security Plan breach being performed during a security event. This could be performed by Security following the security event with little effect on room temperature.

Nt a critical activity initially, DC voltage indication is available in the control room. Batteries are good for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

AVcontrol is available in the control room. Local control is used if necessary to cool down the primary system to delay the potential for RCP seal LOCA. Thic ss not occur until 30 minutes at which time the 30 minute responder becomes available, enabling transfer of duties.

Onyrequired if Condensate Storage Tank Level is less than 5 feet. Should not be a concern for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

the generator which is not initially time critical.

I AU: kPK-AtW.I lNote: Primary AO is degassing 1 Page 10 of 13 Draft - For Information Only 3/11/03

Explosion in Screen House. Loss of Buses 17 & 18. Security available Item Not Covered Discussion Electrician: Locally monitor DC Not a critical activity initially, DC voltage indication is available in the power supply control room. Batteries are good for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

AO: Locally operate ARV's ARV control is available in the control room. Local control is used if necessary to cool down the primary system to delay the potential for RCP seal LOCA. This step does not occur until 30 minutes at which time the 30 minute responder becomes available, enabling transfer of duties.

AO: ER-AFW.I Only required if Condensate Storage Tank Level is less than S feet. Should not be a concern for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Note:

Primary AO is degassing the generator which is not initially time critical.

Security event in Screen House, Loss of Buses 17 & 18, Security not available Item Not Covered Discussion Security: Open vital are doors The current commitment is to open the doors within 30 minutes for SBO.

This commitment assumes that the ventilation to vital areas will be lost. No ventilation is unavailable with a loss of only buses 17 and 18.

Electrician: Locally monitor DC Not a critical activity initially, DC voltage indication is available in the power supply control room. Batteries are good for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

AO: Locally operate ARV's ARV control is available in the control room. Local control is used if necessary to cool down the primary system to delay the potential for RCP seal LOCA. This step does not occur until 30 minutes at which time the 30 minute responder becomes available, enabling transfer of duties.

AO: ER-AFW. 1 Only required if Condensate Storage Tank Level is less than 5 feet. Should not be a concern for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Note:

Primary AO is degassing the generator which is not initially time critical.

Fire in the Aux. Bldg., Security available Item Not Covered Discussion no issues Security event and subsequent fire in the Aux. Bldg., Security not available Item Not Covered Discussion Fire Brigade To be addressed by 3 additional fire brigade members added to shift in response to 2/25/02 Security Order.

Page 11 of 13 Draft - For Information Only 3/11/03

LOCA outside CNMT Item Not Covered Discussion Survey steam lines This task can be delayed with no adverse impact on EOP implementation.

Steam line surveys are only used to verify that no concurrent SGTR has occurred (highly unlikely). This survey can eventually be performed by the 30 minute RP Technician. Installed RG1.97 instruments provide the initial basis for concluding that there is no SGTR.

Control access to Aux. Bldg.

The SS will re-prioritize the Shift RP Technician to control access as required. A SG sample is only used to verify that no concurrent SGTR has occurred (highly unlikely). This can eventually be performed by the 30 minute RP Technician. Installed RG1.97 instruments provide the initial basis for concluding that there is no SGTR.

LOCA outside CNMT with large fire Item Not Covered Discussion Locally close CNMT isolation Only required if the CIVs fail to close. Should only be an issue if this valves would assist in isolating the leak. Attachment III provides additional discussion.

Local adjustment of 300A&B This is only a consideration step for a large imbalance in RCP seal injection flow.

Survey steam lines This task can be delayed with no adverse impact on EOP implementation.

Steam line surveys are only used to verify that no concurrent SGTR has occurred (highly unlikely). This survey can eventually be performed by the 30 minute RP Technician. Installed RG1.97 instruments provide the initial basis for concluding that there is no SGTR.

ATT:SD-1 Low priority of performing normal secondary system alignments for shutdown.

Control access to Aux. Bldg.

The SS will re-prioritize the Shift RP Technician to control access as required. A SG sample is only used to verify that no concurrent SGTR has occurred (highly unlikely). This can eventually be performed by the 30 minute RP Technician. Installed RG1.97 instruments provide the initial basis for concluding that there is no SGTR.

Design Basis SGTR Item Not Covered Discussion no issuesl Page 12 of 13 Draft - For Information Only 3/11/03

Design Basis SGTR with large fire Item Not Covered Discussion Locally close CNMT isolation Only required if the CIV's fail to close. Should not be an issue for SGTR.

valves Attachment III provides additional discussion.

Local adjustment of 300A&B This is only a consideration step for a large imbalance in RCP seal injection flow.

Isolate flow from ruptured S/G Only required if valves can not be closed from the MCB.

ATT:Ruptured SIG Needs to be performed to limit release paths when the 30 minute responder becomes available, enabling transfer of duties.

ATT:SD-1 Low priority of performing normal secondary system alignments for shutdown.

Note:

The probability of a SGTR coincident with a fire is extremely low.

Page 13 of 13 Draft - For Information Only 3/11/03

Attachment V RG&E NERP Staff Augmentation Estimated Travel Times Draft - For Information Only 3/11/03

RG&E NERP Staff Augmentation Estimated Travel Times (Note 1)

NUREG-0654

< 15

>15 Min.

> 30 Min.

> 45 Min.

> 60 Min.

Ginna NERP Title Position Title or Expertise Min. I 30 Min.

s 45 Min.

s 60 Min.

Plant Operations and Assessment of Operational Aspects Shift Supervisor (SRO)

N/A N/A N/A N/A N/A 2

Shift Foreman (SRO)

N/A N/A N/A N/A N/A 3

Control Room Operators N/A N/A N/A N/A N/A 4

Auxiliary Operators N/A N/A N/A N/A N/A Emergency Direction and Control (Emergency Coordinator) 5 Shift Technical Advisor, 3

Emergency coordinator Shift Supervisor, or designated facility manager Notification/Communication 6

(There is no NUREG title 1

3 1

l l TSC Communicator for this position- -

I Communicator) 1 3

4 1

1 EOF Communicator Radiological Accident Assessment and Support of Operational Accident Assessment 7

Senior Manager (EOF 2

1 Recovery Manager Director)

___l_ll_

8 Senior Health Physics 1

4 1

1 TSC Dose Assessment (HP) Expertise Manager EOF Dose Assessment l_

l Manager Page 1 of 4 Draft - For Information Only 3/11 IflA

RG&E NERP Staff Augmentation Estimated Travel Times (Note 1)

NUREG-0654

• 15

>15 Min.

> 30 Min.

> 45 Min.

> 60 Min.

Ginna NERP Title Position Title or Expertise Min.

• 30 Min. l 45 Min.

• -s 60 Min.

9 (There is no NUREG title 6

9 4

4 1

Survey Center Manager for this position - Offsite Survey Team Member Surveys) 10 (There is no NUREG title Included (9) above.

Survey Team Member for this position - Onsite Surveys) l1 HP Technicians (In-plant 7

9 3

2 2

Shift RP Technicians (Note 2) surveys)

RP Technicians RP/Chem Manager 12 Rad/Chem Technicians (Note2) 2 (Note2)

(Note2)

(Note2)

Chem Technicians l

Shift RP Technicians (Note2)

Plant System Engineering, Repair and Corrective Actions 13 Shift Technical Advisor N/A N/A N/A NIA NIA 14 Core/Thermal Hydraulics 2

1 1

Nuclear Assessment 15 Electrical 4

1 1

Electrical/I&C Assessment 16 Mechanical 1

3-Mechanical/Hydraulic Assessment 17 Mechanical Maintenance 10 11 2

1 2

Maintenance Personnel (Note 3) 18 Rad Waste Operator Not required.

19 Electrical Maintenance 1

1 9

1 Maintenance Personnel (Note 3) 20 Instrument & Control 4

4 8

1 Maintenance Personnel Technician (Note 3)

Page2 of 4 Draft - For Information Only 3/11/03

RG&E NERP Staff Augrment-tinn 1wm.utims1t rr:---

T r AT. 1)

NUREG-0654 15

>15 Mi.

>30 Mi

> 45 Min.

> 60 Mi1 Ginna NERP Title Position Title or Expertise Min.

30 Min. l 45 Mi. l

• 60 Min. l Protective Actions (In-plant) 21 l HP Technicians lIncluded (I 1) above.

lShift RP Technicians (Note 2 1

HRP Technicians R/Chem. Manager Fire Fighting 22 (There is no NUREG title Local support N/A for this position - Fire Brigade)

Rescue Operations and First-Aid 23 l (There is no NUREG title Local support N/A for this position )

Site Access Control and Personnel Accountability 24 Security Personnel N/A Total 42 55 31 9

8 (29%)

(38%)

(21%)

(6%)

(6%)

Page 3 of 4 Draft - For Information Only 3/11/03

NOTES Note 1 The estimated staff augmentation travel times are based on the projections of Yahoo! Maps, using current employee home addresses and the Ginna address. These times are not intended to be a commitment, but are provided to furnish an overall understanding of the staff's current ability to respond if required.

Note 2 The Ginna Shift RP Technicians are qualified in both HP and radiochemistry analysis. The number of qualified Ginna Shift RP Technicians and estimated travel times are not repeated for the listed NUREG-0654 function of Rad/Chem Technicians, to provide a more accurate total number at the bottom of the table.

Note 3 The number of Ginna maintenance personnel listed are those currently available for call-in by the Shift Supervisor.

They may not all be considered as one-hour responders when this is actually implemented.

Page 4 of 4 Draft - For Information Only 3/11/03