ML031550423

Response to NRC Memorandum Dated December 31, 2002, Regarding the Minimum On-Shift and Augmentation Staffing for Radiological Emergencies

ML031550423
Person / Time
Site:	Ginna
Issue date:	05/23/2003
From:	Mecredy R Rochester Gas & Electric Corp
To:	Clark R Document Control Desk, Office of Nuclear Reactor Regulation
References
IR-02-009, TIA 02-002
Download: ML031550423 (54)

Text

Robert C. Mecredy Rroff Vice President Always at Your Service Nuclear Operations May 23, 2003 Mr. Robert L. Clark Office of Nuclear Regulatory Regulation U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Response to NRC Memorandum dated December 31, 2002, Regarding the Minimum On-shift and Augmentation Staffing for Radiological Emergencies Rochester Gas and Electric Corporation R.E. Ginna Nuclear Power Plant Docket No. 50-244

Dear Mr. Clark,

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 and 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 VI.

Results of Unannounced Off-hours Call Out Drill VII.

Regulatory Commitments An equal opportunity employer 89 East Avenue I Rochester, NY 14649 tel (585) 546-2700 www.rge.com l-7+E An Energy East Company

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 18 one hour responder positions to the 13 current one hour responders. RG&E is 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 unplanned reactor trip, which ever occurs first. RG&E is also proposing to add a Dose Assessment Manger to the list of responders who are activated at the declaration of an Unusual Event. 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.

d.

The results of an unannounced off-hours call out drill performed on May 8, 2003 are provided as Attachment VI. This drill was specifically performed to evaluate the response of the proposed 30 minute responder (HP or Rad/Chem) and the proposed I&C/Electrical maintenance 60 minute responders. Due to the large number of other proposed responders, it was determined to be prudent to perform a complete 60 minute responder drill. The overall results of the drill were that the current Nuclear Emergency Response Plan (NERP) 60 minute response requirements were met and, with the exception (by six minutes) of two of the proposed Survey Team members, the new proposed response requirements were met.

The call out drill overall response times and the number of responders was negatively impacted by the automated activation system. Unanticipated issues with the call out process were identified due to the large number of proposed individuals who must now be contacted at an early stage following the event. There were also additional issues associated with personal pagers. These issues have been entered into the Ginna corrective action process (ACTION Report tracking number 2003-1009). RG&E is working with the automated activation system vendor to streamline the call out process and is handling the pager issues on an individual basis. Additional unannounced testing (without an actual response required) will be performed to validate the process prior to July 31, 2003. RG&E is also providing for additional qualified maintenance individuals to supplement the three individuals per position who were initially assigned during the call out drill.

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e.

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). Though the historical response of the RG&E emergency response organization (ERO) has been timely, there will be a reinforcement by senior management of the expectation that NERP responders will respond immediately upon being notified and not wait for additional time.

This expectation has also been added to the annual responder training, which currently includes a discussion of the automated call out process, and will be discussed within the NERP.

f.

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. The TSC Director, Operations Assessment Manager, and Technical Assessment Manager or their' alternates will normally report to the TSC to provide the following assistance:

Provide and coordinate activities to relieve the Control Room of communications, Emergency Assessment and manpower utilization.

Direct and coordinate operations personnel in accident confirmation, mitigation and recovery.

Assist Control Room with technical assessment of the event and other activities that are not essential Control Room functions.

As discussed above, in addition to the 3 current positions, a Dose Assessment Manager will also report to the TSC to provide assistance.

f.

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.

RG&E is currently planning on completing the implementation of the proposed enhancements, including training of the additional personnel and the completion of the corrective actions associated with the unannounced off-hours call out drill, and submitting the revised Nuclear Emergency Response Plan (NERP) to the NRC by July 31, 2003. If you should have any questions regarding this submittal, please contact Mr. Thomas Harding, 585-771-3384.

ly yours, Robert C. Mecredy Page 3 of 4

xc:

Mr. Robert Clark (Mail Stop 0-8-C2)

Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 11555 Rockville Pike Rockville, MD 20852 Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 U.S. NRC Ginna Senior Resident Inspector Peter R. Smith, Acting President New York State Energy, Research, and Development Authority 17 Columbia Circle Albany, NY 12203-6399 Mr. Charles Puglisi NYS Department of Public Service 3 Empire Plaza Albany, NY 12223-1350 Mr. Robert Reynolds Federal Emergency Management Agency, Region II Jacob K. Javits Federal Building 26 Federal Plaza, Room 1337 New York, New York 10278-0002 Ms. Thelma Wideman Director, Wayne County Emergency Management Office Wayne County Emergency Operations Center 7336 Route 31 Lyons, NY 14489 Ms. Mary Louise Meisenzahl Administrator, Monroe County Office of Emergency Preparedness 1190 Scottsville Road, Suite 200 Rochester, NY 14624-5159 Mr. Andrew Feeney State Emergency Management Office Public Security Building State Campus Albany, NY 12226-5000 Page 4 of 4

Attachment I Comparison of NUREG-0654 to Ginna NERP, and Proposed Enhancements

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

Plant Operations and Assessment of Operational Aspects 1

Shift Supervisor (SRO) 1 1

1 2

Shift Foreman (SRO) 1 1

1 3

Control Room Operators 2

=

2

=

=

2

==-

4 Auxiliary Operators 2

2 2

Emergency Direction and Control (Emergency Coordinator) s Shift Technical Advisor, Shift 1(a)

(a) l I

1(a)

Note 3 Supervisor, or designated I I l

l l

l 1

facility manager l

l I

l l

l l

Notification/Communication 6 l (ThereisnoNUREGtitlefor 1 1 1

2 1

1 1-2 Note4 this position - Communicator) l l

l l

l l

l l

1 l

Radiological Accident Assessment and Support of Operational Accident Assessment 7

Senior Manager (EOF 1

1 1

Note 5 Director) l l

l l

l l

l l

8 Senior Health Physics (HP) 1 2

2 Note 6 Expertise l

l l

l l

l l

l l

l Page 1 of 11 5/23/03

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.

l ___________________________________

_ l__l l_l (N ote I) l(N ote 2) l

.(N t

lo e

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 1

Note 9 surveys) 12 Rad/Chem Technicians 1

1 1

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

1 1

l 14 Core/Thermal Hydraulics 1

2 1

Note 11 15 Electrical 1

1 Note 11 16 Mechanical 1

1 Note 11 17 Mechanical Maintenance l(a) 1 (a) 1 (a) 1 Note 12 18 Rad Waste Operator

=-

1 T

l Note 13 19 Electrical Maintenance 1(a)

I 1

1(a) 1(a) 1 Note 14 20 Instrument & Control 1

l 1

Note 14 Technician Protective Actions (In-plant) 217 HP Technicians 2(a) 2 2

1(a) 1 1(a) l 4

Note 15 Page2of 11 5/23/03

Page3of 11 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 Fire Fighting 22 (There is no NUREG title for per Local support 5(b)

Local support 5(b)

Local support l Note 16 this position - Fire Brigade)

FPP l

l l_

l Rescue Operations and First-Aid 23 (There is no NUREG title for 2(a)

Local Support 3(C)

Local Support 3(C)

Local Support Note 17 this position )

Site Access Control and Personnel Accountability Security Personnel Total I

24 1 5/23/03

Table Notation (a)

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

(b)

Includes two Auxiliary Operators as collateral duties and three dedicated Fire Brigade members.

(c)

Performed by the three dedicated Fire Brigade members, as collateral duties.

(d)

The current and proposed Ginna on shift totals include the three dedicated Fire Brigade members.

(e)

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

TSC Operations Assessment Manager News Center Manager Survey Center Manager EOF Nuclear Operations Manager TSC Maintenance Manager (f)

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

TSC Operations Assessment Manager News Center Manager Technical Assessment Manager Survey Center Manager EOF Nuclear Operations Manager TSC Maintenance Manager EOF Engineering Manager TSC Technical Assessment Manager Page4of 11 5/23/03

Notes 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 5/1/81. 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 Auxiliary 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 majority of the Ginna Auxiliary Operators have a Nuclear Navy background which provides them with sound basis for working with equipment in the field. The Auxiliary Operators have received radiation protection training as part of their training and can provide that capability if required. They are also qualified as communicators. The Ginna Shift RP Technicians have the training and expertise to provide radiation surveys and are cross-trained in radiochemistry/chemistry (Rad/Chem) 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-1 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 designated Communicator on shift (1 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 Page5of 11 5/23/03

Response Data System (ERDS) has also provided the NRC with the ability to remotely monitor key Ginna parameters. The STA could also provide assistance with technical support of the NRC emergency phone line if required. As an enhancement, RG&E proposes to add an EOF Communicator as a NERP required one hour responder to provide for a 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. For a fast breaking General Emergency, the Protective Action Recommendation (PAR) procedure is implemented from the control room and recommends evacuation in a two mile radius and five miles down wind, with sheltering the remainder of the populace. The PAR recommendation is based solely on the event classification and the wind direction. Ginna also has a TSC Dose Assessment Manager and an EOF Dose Assessment Manager as required NERP one hour responders who would provide enhanced dose assessment support, with 5 of these individuals having less than a 30 minute travel time per Attachment V. As an enhancement, RG&E proposes to add the TSC Dose Assessment Manager position to the list of individuals who are notifiled to respond at the declaration of an Unusual Event.

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.

Page 6 of 11 5/23/03

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 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 one 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 unplanned reactor trip). Also, RG&E proposes to add one additional HP qualified individual as a NERP required one hour responder 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 Auxiliary Building, Security Available Security Event and Subsequent Fire in the Auxiliary Building, Security Not Available LOCA Outside Containment LOCA Outside Containment with Large Fire Page7of 11 5/23/03

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 (Note 9) would mean that these activities would be performed in a more timely manner. Since the Ginna Shift RP Technicians are qualified to perforn 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. Also, RG&E proposes to add one additional Rad/Chem qualified individual as a NERP required one hour responder to support chemistry analysis. There are a total of 10 HP qualified personnel and RP/Chemistry Managers with less than a 30 minute travel time.

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 Page 8 of 11 5/23/03

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 perfonn 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 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 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 technicians to troubleshoot and correct equipment malfunctions whenever equipment problems are identified. Since the Shift Supervisor assumes the duties of the Emergency Coordinator at the classification of the event, the necessary technicians would 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 Attachment V.

Protective Actions (In-plant)

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 and Page9of 11 5/23/03

training on the effects of radiation received during emergency events. Radiation exposure monitoring has improved dramatically since NUREG-0654, Table B-1, 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 for steam line monitoring. On shift personnel can self frisk when leaving a restricted area where PCMs are not available. A RP/Chem Manager and Dose Assessment Manager are current NERP required one hour responders reporting to the TSC. These 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 three 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.

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 three dedicated Fire Brigade members.

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

Page lOof 11 5/23/03

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

HP or Rad/Chem qualified individual Proposed NERP 60 Minute Resnonders 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 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 Rad/Chem Technician Current Current Current Current Current Current Current Current Proposed 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

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

Page 11 of 1 55/23/03

Attachment II Evaluation of NUREG-0654 Basis for On Shift Staffing

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 Radio-chemical Sampling/Analysis In-plant Radiological Monitoring

+

Meteorological Assessment RG&E Coverage The Ginna Shift RP Technician is qualified to obtain samples 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.

• 1-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. 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.

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 5/23/03

Page 2 of 2 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-1 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.

5/23/03

Attachment III Evaluation of NUREG-0654 Basis for Timing Staff Augmentation

Evaluation of NUREG-0654 Basis for Timing of Staff Augmentation Backg-round 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-1 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-termn 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 -

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.

Page I of 4 5/23/03

Evaluation Section III.C of NUREG-0396, Planning Basis for the Development of State and Local Government Radiological Emergency Response Plans in Support ofLight 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 Page 2 of 4 5/23/03

and 9. As described within Section 5.2.1 of WASH-1400:

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-1 150 and the basis for the Westinghouse Severe Accident Management Guidelines (SAMGs) were reviewed.

NUREG-1 150, Severe Accident Risks: An Assessmentfor 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 NJREG-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 Page 3 of 4 5/23/03

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

Conclusion The timing considerations within NUREG-0654 Table B-1 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 from WASH-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 5J23/03

Chapter 5 Reactor Accident Risks 5.1 INTIIODUCTION AND SUAMIARY 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 IADIOACTIVE RELEASE CATEGORIES As set forth in Chapter 4, the quanti-ties of various isotopes released from the containment following a given acc.i(cnt ar calculaLedl usiiig Hi.e COIIRAL 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 conLtailmeniL.

AlLliouli ucli.1 it:

predicted to be very unlikely, it cannot be ruled out completely on the basis of VAS qoo

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.

5.2.2 BWR 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 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 dopodiLion ort the tjurfacoo inBid te 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 W4S

-vqoo

TABLE 5-1

SUMMARY

OF ACCIDENTS INVOLVING CORE DURATION WARNING ELEVATION CONTAINMENT TIME OF O

IEFO F

ENERGY(a PROBABILITY OF O

TIEOR F

RELEASE FRACTION OF CORE INVENTORY RELEASED(a RELEASE per RELEASE RELEASE EVACUATION RELEASE 6

(b) c)

CATEGORY Reactor-Yr (Hr)

(1r)

(fir)

(Meters)

(10 Btu/Hr)

Xe-Kr Org. I I

Cs-Rb Te-Sb Ba-Sr Ru La PWR 1 9x1O 2.5 0.5 1.0 25 5 2 0 (d) 0.9 6xO 0.7 0.4 0.4 0.05 0.4 3x1O PWR 2 8x10 2.5 0.5 1.0 0

170 0.9 7xlO 0.7 0.5 0.3 0.06 0.02 4x10 3

PWR 3 4x10 5.0 1.5 2.0 0

6 0.8 6xlO 0.2 0.2 0.3 0.02 0.03 3x10 3

PWR 4 5x10 2.0 3.0 2.0 0

1 0.6 2xI0 0.09 0.04 0.03 SxlO 3xlO 4x1O PWR 5 7xO7 2.0 4.0 1.0 0

0.3 0.3 2x10 0.03 9x10 5x10 lx10 6xlO 7x10 PWR 6 6xO 12.0 10.0 1.0 0

N/A 0.3 2x10-3 8x-4 8X-4

_x 3

9x_5 7x 5 x1O5 PWR 7 4x10 10.0 10.0 1.0 0

N/A 6xlO 2x10 2x10 lxlO 2x10 lxlO lxlO 2xlO PWR 8 4x10 0.5 0.5 N/A 0

N/A 2xlO 5x10 lxlO 5xlO lxlO ixlO 0

0 PWR 9 4xlO 0.5 0.5 N/A 0

N/A 3x10 7x10 1xlO 6xlO lxlO lxlO1 0

0 BWR 1 lxlO 2.0 2.0 1.5 25 130 1.0 7xlO 0.40 0.40 0.70 0.05 0.5 5x10 3

BWR 2 6xlO 30.0 3.0 2.0 0

30 1.0 7xlO 0.90 0.50 0.30 0.10 0.03 4x10 3 BWR 3 2xO 30.0 3.0 2.0 25 20 1.0 7xO 0.10 0.10 0.30 0.01 0.02 3xlO 3

BWR 4 2x10 5.0 2.0 2.0 25 N/A 0.6 7xl0 Bx10 5x10 4xl0 3 6x10 6xI0 lxlO BWR 5 lxlO 3.5 5.0 N/A 150 N/A 5x10 2x10 6xlO1 4x10 8xlO 8xO1 0

0 (a) A discussion of the isotopes used in the mechanisms is found in Appendix VII.

study is found in Appendix VI.

Background on the isotope groups and release (b)

Includes Mo, Rh, Tc, Co.

Cc) Includes Nd, Y, Ce, Pr, La, Nb, Am, 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. wAS14 - toc

TABLE 5-2 PWR DOMINANT ACCIDENT SEQUENCES vs. RELEASE CATEGORIES RELEASE CATEGORIES Core 4elt No Core 2

3 4

S 6

7 9W.

9-AD -

11 8~A1 -10 AD-..

ACo-B-1 AD-B A9I-AD-.

6 A-R A I 4

LARGErSCA AR-a 10 Ae;6 11 Ix1 9H-

-6 AIF-A x10 I10 2x10 1.10 4.10 x10 2.10 2.10 1.10 Ar-AR-AXAr A-I O

58y 31x00 3

0 Al-

_6 LARGE LCA 7x10.I 410- i 0

5 a10 3*0x I*10 A

ACD-10 AIIF-y 11 ADF-4 0

-11 2.1

-Ro2 A("-.

AG-1 A Probabilities 2x10 19 110 xio 1.10 410 3107 3x10 1.10-5 I

1-4 SR'-

I SR l

iDa scoO SN0-IFE SID-S-

S1 x

° s

S 9D-a B

S1Cn-B 1

S1 -B 8

5 3.-

1 6-0 3x

'li0-9 24.10 6x10 x1 0

3.

3NIoD6 1

3I-0 SKMALL LCA SICD__a s

-B S N-a S0 4-SB-C S -

SMALLOA jC2x f

10

-10 1

-N 9

1 9

1 7.10 3x10 10 30 3x1 6 10-9 2.10 31O '

I I-4-1 S1HF-f-11 s,r6 IN-4*

10 3_10

_E1dO 3.10 R

--10

-8T-7 I-

-7

-6

-5 4

S Probabilities 3x10 210 210 x10 3

0 I10 6x10 6x10 310 310 S

2"-aO S

o D-a s nG-B s2 n-B S B-C 2_-

V*10 21*109 9xl0 2110-1 210109 9

s2-O S NF-Y1 S

2N-52 HN-B S CD-C N S

2H-c -6 1109 2TML-10 6x10 110 260xI 6*L10-SKANLL INA T

S TXQS_

N-C EVENT-T4210 62x10 21 S2 221ODI to-1 2

11

22.

-1 102It

-61 9.1*10 2*106 S2C-n SG8 2 ~

~

tiQa-

-T 2Q; -N 2*10 9*10 S

Probabilities 1x10 3x10 3x10 310 3x10 2x10 6 2*10 12 EA1y

-S O

A T7R 2.10

_310

• .10 lxi -7 REACTOR VESSEL AT-6 RUPTURE -

S.10 11 AC-1 2 (5t VXLUE) 9x1 10

-9i

-59x 0

-91

-x1

-7x0 410 41-R Probbilities 2x10 1x1l 110 2x10 110 2x10 1x10 4x INTERFACING SYSTEMS LInCA 40_

(CNECK VLVE

- '4*0 V Proabilities 4Te50 4.10o6 4t 1Q7 4x eoq ThLR~-a TNLRy3 TKL-a TAL-B0 TMLR-C_7 TML.-C 3*108 f7*107 6*10en 3*1 IC) b 6*10 6*o0 TRANSIENT TMLSR-6l TKaE EVENT - T 20 3.10 3*10 3*10 T-0b TIOQ-EC6 T IO0 1.10

• Proabilties 3xI-7 3xO-6 4xO-7

-x_8 2l-7 xl_6 l

o-5 (1

SM9ATION Or ALL ACCIDENT SEQUENCES PER R(ELEASE CATEGORY MEDJAN-7--6-

-5S 150% VALUE) 9.107 8.10-

4. 106 5,107 7x107 6.10-4*105 4*10 5 4*104 LOnK BOUND

~~-7

-7 7

-6

-5

-65 (5% VLtE) 9xl0-e 8*10 6*10 9.10 2*10 2*10 1*10 4*107 4.105 UPPER OUND

-_6

-5 5x

.1-6

-x06 21-5

-x04

-4 4I0 (95% VALUE) 9*50 8.0 4

5.10 4,0 2*0 2*0 4*104 103 Nates The probabilities for each release category for each event tree and te for all accident equences are te median.

val..s of the dominant accident sequences sned by Mnte Carla simulation plus a 10% contribution from te adjacent FraE,.6.,asf.g.y 1.a.ifity KEY TO TABLE 5-2 ON FOLLOWING PAGE Ws -

o O

KEY TO PWR ACCIDENT SEQUENCE SYMBOLS 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.

G - Failure of the containment heat removal system.

H - Failure of the emergency core cooling recirculation system.

K - Failure of the reactor protection system.

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

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

Q -

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

R -

Massive rupture of the reactor vessel.

s1 - A small LOCA with an equivalent diameter of about 2 to 6 inches.

S2 - A small LOCA with an equivalent diameter of about 1/2 to 2 inches.

T -

Transient event.

V - LPIS check valve failure.

- Containment rupture due to a reactor vessel steam explosion.

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

y - Containment failure due to hydrogen burning.

6 Containment failure due to overpressure.

e Containment vessel melt-through.

KEY TO TABLE 5-2 hw4s o

Attachment IV Evaluation of RG&E Current On Shift Staffing

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 13 5/23/03

Grid Failure, Direct Entry into ECA-O.O, Security Available SS I

CRF I

HCO I

co STA I

PRIMARY SECONDARY EXTRA SECURITY SHIFT TECH.

I NOT CUVERL-I EPIP CR, roceurePrmary, S

CR, Lead CR Procedure ma Secondary,lmm AB

00 ECA-0.0 Act~~~Imsediate

ActosB, Tour TB, Tour TB, AVT CR, Procedure

Primary, Secondary, AB, Tour Reseting TDAFWP CR,

05 ECA-0.0 ECA-0.0 ECA-0.0 Z

govemor valve Communicator CR, Lead CR, Procedure Attempt to Attempt to CR, Locally start Locatly start Diesels CR,

10 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 Monitoring seals TDAFWP Communicator doors Emergency CR, Procedure

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

20 CR, Lead ECA-0.0 ECA-0.0 ECA-0.0 Monitoring seals valves Communicator doors sdiation DC power supply

20 CR ed C,P eueEA00 EA00 Mnon Dgs ale/utrd Cmuiao radiation CR, Procedure Primary, Secondary, CR, Degass ATT.

CR, Locally check CR, Lead ECA-0.0 ECA-0.0 ECA-0.0 Monitoring Generator Su/GVI lved Communicator Pa on CR Lad CR, Procedure

Primary, Secondary, CR, Degass ATT.

CR AO: Locally operate CR, Lead

~~~~~~~~~~~~~~~~~~~Faulted/Ruptured

30 ECA-0.0 ECA-0.0 ECA-0.0 Monitoring Generator S/G Communicator ARVs CR, Procedure Primary, Secondary, Energize Degass A.CR, Sample RCS and CR, Lead Faulted/Ruptured AO: ER-AFW.1

35 ECA-0.0 ECA-0.0 ECA-0.0 source ranges Generator SIG Communicator PZR for boron 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 Cl/C VI 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 ECA4.0 Monitoring CI/CVI valves Communicator PZR for boron At this point, we would be looping in ECA-O.O until power is restored

50

50~~~

~ ~

c o__

55

_SECURI1_Y_SHIFT__TECH.___NOT__COVERED__

60_

==_

=_ _

==_

_ 1=

=__

==_

p) qD 0'-

Security Event in Switchyard, Loss of Circuits 767 and 751, Security Not Available SS CRF HCO CO I

STA PRIMARY SECONDARY EXTRA SECURITY SHIFT TECH.

NOT COVERED EPIP CR, Lead CR, Procedure Immediate Secondary,mm AB, Tour TB, Tour TB, AVT

00 ECA(~I:,

ACAion0 ediate Actions Reseting CR, Lead CR, Predure CA4.

0 ECA0 0 AB, Tour TDAFWP Cr, ECA-0.0 ECA-0.0 ECA-0.0 Cmuiao

05

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

~~~~~~~~~~~CR, MonitoringLoalstr Lclysat CR C, Lead ECA-0.0 restore power restore power Diesels Diesels Communicator

10 CR, Procedure Pull Stop Open Reactor Isolate RCP Backup cooling to CR, SitOe Area

20 CR, Lead EA4 ECA-0.0 equipment Protection doors ing seals TDAFWP Communicator s

t en viEl area doorsEmergency

30 CR, Lead CT, Pmc edum Fnmaly, ECA4 0

CF, MonRDnw Dass FalRed/lluP umd cpmLOcallycheck EleAOLocallylmonitoraDC powe CR, Procedure Primary, Secondary, Locally Isolate CR, Lall e

RCS a

n d

CR, Lead ECA4,0 ECA.0 ECA.0 CR, Monitoring Csupply Securlty:Open vital area

20

__ __ __r adiation doors Degass ATT.

CRLocally check CR, Procedure Primary, Secondary, CRDoiorn eeator CaultedRupture CR, La this point,ECA-0.0 ECA-0.0 ECA-0.0 Sbe lopn Communicatorsa

25 SG

=

=

radiation CR, Lead CR, Procedure

Primary, Secondary, CR oioigDegass ATTh CR, A:LclyoeaeAV

30 ECA-0.0 ECA-0.0 ECA-0.0 Generator S/G Communicator ATT.

CR, Procedure

Primary, Secondary, Energize source Degass CR, Sample RCS and CR,3ead ECA-0.0 ECA-0.0 ECA-0.0 ranges Generator Fae/RutuedCommunicator PZR for boron A:E-F CR, Procedure

Primary, Secondary, SamplesRCocandyAIsolRteFWR1 CR, Lead

~~~~~~~~~~CR, Monitoring DegassleLocallyAIsolateFCR, CR,4ead ECA-0.0 ECA-0.0 ECA-0.0 Generator CI/CVI valves Communicator PZR for boron CR, Lead CR, Procedure Primary, Secondary, Locally Iolate CR,

• A~b*Sample RCS and

45

~~~ECA-0.0 ECA-0.0 ECA-0.0 CR,Mnioin lCVI valves Communicator PZR for boron At this point, we would be looping In ECA-0.0 until power Is restored

6 0 1 _ _

)

0 w

Explosion in Screen House, Loss of Buses 17 & 18, Security Available.

SS I

CRF I

HCO I

CO I

STA I

PRIMARY SECONDARY EXTRA SECURITY I SHIFT TECH.

NOT COVERED EPIP

Primalry, ScnaylmA,Tu CR. Lead CR, Procedure Immediate ea Acions AB, Tour TB, Tour TB, AVT

00 ECA-0.0 Actions eit cin

0 CR, Lead CR, Procedure

Prmary, Secondary, Reseting CR ECA-0.0 ECA-0.0 ECA-0.0

~~~~~~~~~~Communicator

05 CAO0 EA.0 EA0 governor valve Lead CR, Procedure Attempt to Attempt to C

t Locally start Locally start CR,

10CR, ECA-0.0 restore power restore power CR, Monionng Diesels Diesels Communicator CR, Lead CR, Procedure Pull stop Open Reactor I

isolate RCP Backup cooling to CR, Open vital area Site Area

15 ECA-0.0 equipment Protection doors CR, Monitorng seals TDAFWP Communicator doors Emergency CR, Procedure

Primary, Secondary, Locally isofate CR, Open vital area Locally check Elec. Locally monitor

20 CR, Lead ECA-0.0 ECA-0.0 ECA-0.0 CR, Monitorng valves Communicator doors asteam line DC power supply.

20 vR ed EA00 Pm Smd C,Mmon eeaD ale/utu Communicatr dor slaiatine CR, Lead CR, Procedure

Primary, Secondary, Cnitorn Degass ATT.

CR, StamlineCk ECA-0.0 ECA-0.0 ECA-0.0 r

n g

Generator Faulted/Ruptured Communicator stea fr no AO: ER-AFW.1

25 Comuncaorradiation CR, Lead CR, Procedure

Primary, Secondary, CR,Montorng AlTTdlpte CR, AO: Locally operate

40CR Lead ECA-0.o ECA-0.0 ECA-0.0 CR, Monitolng Generator Faled/Rupte Communicator ARV-s

30

_ _ S G CR, Lead CR, Procedure

Primary, Secondary, Energize source Degass ATT.

CR, Sample RCS and A:E-F.

CR35 ead ECA-0.0 ECA-0.0 ECA-0.0 ranges Generator Faulted/Ruptured Communicator PZR for boron A:E-F.

35 ~

~

~

_ SIG CR, Lead CR, Procedure

Primary, Secondary, C,MntngDegass Locally Isolate CR, Sample RCS and AO: ER-AFW.1

40 ECA-0.0 ECA-0.0 ECA-0.0 CR oioigGenerator CI/CVI valves Communicator PZR for boron CR, Lead CR, Procedure

Primary, Secondary, C M Locally isolate CR, Sample RCS and

45

, a ECA-0.0 ECA-0.0 ECA-0.0 CR, Monitonng CI/CVI valves Communicator PZR for boron

50 At this point, we would be looping in ECA-O.0 until power is restored

55

=

60

==

____=

=

0 I-w

Security Event In the Screen House, Loss of Buses 17 & 18, Security Not Available SS I

CRF I

HCO co STA I

PRIMARY I SECONDARY I

EXTRA I

SECURITY I SHIFT TECH.

NOT COVERED E-PIP

0CR, Lead C

Immedate Secondary,lmm AB, Tour TB, Tour TB, AVT ECA-0.0 ediate Actions

00 Actions CR, Procedure

Primary, Secondary, Reseting CR

0 CR, Lead ECA-0.0 ECA0.0 ECA-0.0

, Tour TAFWP Communicator

05 govemor valve CR, Led CR, Procedure Atempt to Attempt to CR, MOnitng Locally start Locally start CR,

10 CR, Lead ECA-0.0 restore power restore power CR, Monitoring Isl Diesels Communicator S

rea CR, Lead CR, Procedure Pull stop Open Reactor C,Mnong Isolate RCP Backup cooling to CR, Scrt:pnvtlae or ieAe

0CR, Lead EC4o EA.

C40 CR, Monitoring vavsCmuiao ta ie

.SrcurltY:Ope n via area doorSUpl

15 C

, LeECA-0.o equipment Protection doors seals TDAFWP Communicator Emergency CR, Procedure

Primary, Secondary, Locally Isolate CR, LElec:

Locally monitor DC power supply C40 wR, Lead ECA4.0 ECA-0.0 ECA40 CR, Monitoring vlV C m cre l

Securlty:Open vital area doors

20 radiation ATT.

~~~~~~~~~~~Locally check CR, Lead CR, Procedure

Pnmary, Secondary, CR,Mnog Degass atedRutuedCR,

steam line

25

~ECA-0.0 ECA-0.0 ECA-0.0 Mnon Generator Faltd/upurdCommunicatorraito 45 CR, Lead CR, Procedure

Primary, Secondary Degass CR o

o a

30 ECA-0.0 ECA-0.0 ECA-0.0 Generator S.until pu re Communicator ATT.CR CR, Lead CR, Procedure Primary.

Secondary, Energize source Degass CaltdRutue Sample RCS and A:E-F.

35

~~~ECA-0.0 ECA-0.0 ECA-0.0 ranges Generator Communicator PRor boron CR, Lead CR, Procedure

Primary, Secondary, CR. Monitoring Degass Locally isolate CR, Sample RCS and

40

~~~ECA-0.0 ECA-0.0 ECA-0.0 Generator CI/CVI valves Communicator PZR for boron CR, Lead CR, Procedure

Primary, Secondary.

C,MnongLocally isolate CR, Sample RCS and CR,4ead ECA-0.0 ECA-0.0 ECA-0.0 R ontrn Cl/CVI valves Communicator PZR for boron At this point, we would be looping In ECA.

until power is restored

55 I

CD 0

tk

Fire in the Auxiliary Building, Security Available (ER-FIRE.3)

SS I

CRF I

HCO I

CO STA I

PRIMARY SECONDARY EXTRA SECURITY I SHIFT TECH. I NOT COVEREDI EPIP Control Room Control Room Control Room Control Room

00 Actions Actions Actions Actions AB, T T,ouTBAV Control Room Control Room Control Room Control Room Fire Brgade Fire Brigade CR, Monitoring Fire

05 Actions Evacuation Evacuation Evacuation Captain Captain Communicator Fire Brigade Brgade Contrl Rom Trp Boh MG Locally verify ControltRoom Trp Both MG RX tp/MSIVs Don SCBA

'A" D/G Room Fire Brigade Fire Brgade CR.

Fire Brigade Monitoring Fire

10 Actions sets Shut Captain Captain Communicator Brigade Control Room Isolate.A. to Fire Brigade Fire Brigade CR, Monitoring Fire

2 oTrp both RCP's C T

Close MOV-856 Start"A D/G Captain Captan Communicator Fire Bgade Brigade

15 AtosCM ControlRoom Trp Intake Start and Fire Brigade Fire Brigade CR, Monitoring Fire

2 Actions heaters control TDAFW Unload bus 14 Monitor D/G Captain Captain Communicator Fire Brigade Brgade Alert

20 Chagin pump Control Room Start Diesel air Stro A Energize bus Verfy Natural Fire Brgade Fire Brigade CR, Fire Brigade Monitodng Fire 25Actions compressor conumpTAF 14 from the D/G Circulation Captain Captain Communicator FieBiae Brigade Start Diesel air Locally Line up Fire Brigade Fire Brigade CR, Monitoring Fire

30

~~~compressor Ch arn pump Captain Captain Communicator Fire Brigade Bgd Lally Linet up' Fire Bride Fire Bdg ir rgade Monitoring Fire

35 u p _

_ Captain Captain

__ Bgade_

Unload bus 16 Fire Brigade Fire Brid ir rgade Monitoring Fire

40UCaptain Captain Brigade

45 Fire Brigade Fire Brigade Monitoring Fire

45 Captain Captain FieBiae Brigade

50;:

60=

Security Event and Subsequent Fire in the Auxiliary Building, Security Not Available.

SS I

CRF I

HCO I

co STA I

PRIMARY SECONDARY EXTRA I

SECURITY I SHIT IECH.

NOT CuVELU I Control Room Control Room Control Room Control Room AB, Tour TB, Tour TB, AVT

00 Actions Actions Actions Actions Control Room Control Room Control Room Control Room Fire Brigade Fire Bfigade CR, Monitoring Fire Brd Actions Evacuation Evacuation Evacuation Captain Captain Communicator Bngade

0 C______

nb.

Com TFSohMaptaivny YDnSBA

/ Ro ngd Captaingd Communicator BMriade R-Fm dgd Control Room Trip Both MG Locally veriy MD S

a

'AD/ G Fire Brigade Fire Brigade CR, Monitoring Fire Fire Brigade

.10 Actions sets tripTMSlIV's Shut Dn bA A D/G Captain Captain Communicator Brgade Control Room t

t es Isolate l cAo to Close MOV-856 Start "A' D/G Fire Brigade Fire Brigade CR, Monitoring Fire Fire Brigade

15 Actions CNMT Captain Captain Communicator Brigade Control Room Tdp Intake Start and control Unload bus 14 Monitor AD/G Fire Brgade Fire Bdgade CR, Monitodng Fire Fre Bigade

20 Actions heaters TDAFW pump Captain Captain Communicator Brigade Fr rgd eea mrec Control Room Start Diesel air Start and control Energize bus 14 Verify Natural Fire Brigade Fire Brigade CR, Monitoring Fire Fre Brigade

5 Actions compressor TDAFW pump from the D/G Circulation Captain Captain Communicator fg.a. L Brigade

4Control Rs Fhow Unlomd bsdMonitrt*A Diesel aiR B

Locally Line up Control Room Startoiesel air Cooldown to

.ad t

MoitorNaAuDaG Fire Brigade Fire Brigade CR, Monitoring Fire Fre Brigade

0 Actions compressor Captain Captain Communicator Brigade Control Room Cooldown to Cooldown to ol n

Vefy Natura Fire Brigade Fire Brigade CR, Monitoring Fire and starto "A

Fire Brigade

5 Actions cold shutdown cold shutdown Circulation Captain Captain Communicator Bigade Control Room Cooldown to Cooldown to CMonitorin Firenio A"/

ir rgaeC

40 Ado Unload bus 16 MonitorAD/G Fire Brgade Fire Brigade CR, Comniae Fire Brigade 40 Actions cold shutdown cold shutdown Captain Captain Communicator B!d Control Room Cooldown to Cooldown to

-Cooldown to Verify Natural Fire Brigade Fire Brigade CR, Monitoring Fire FieBgd

45 Actions cold shutdown cold shutdown cold shutdown Circulation Captain Captain Communicator Brigade Fr rgd Control Room Cooldown to Cooldown to Cooldown to Moio ADG Fire Brigade CR,

50 Actions cold shutdown cold shutdown cold shutdown Captain Communicator Control Room Cooldown to Cooldown to Cooldown to Verity Natural Fire Brigade CR, 55 Atons cold shutdown cold shutdown cold shutdown Circulation Captain communicator Control Room Cooldown to Cooldown to Cooldown to -

oio ADG Fire Brigade CR,

60 Actions cold shutdown Icold shutdown cold shutdown Moio ADG Captain Communicator CD 0 qi EPIPl

LOCA Outside Contalnment I

SS CRF H

CO STA l

RIMARY SECONDARY EXTRA SECURITY SHIFT TECH.

NOT COVERED EPIP CR, Lead CR. Procedure Pnrmay.

Serndary,imm Locally trip the E-o PdatE ediate Actions reactor

00 I

_ __ _ Actfons__

CR. Pmcedure Pimary, E-Secondary CR,nmio Sampb S r

U CR. Lead CNMT Isorton Colato_

25 CK LeuE-0 0

E-:

RPa_C0

_CMoWb 5>

mwcatorW

05 v;0 0

ve-CRt,)£ Prcedure

Prma, E-Secondary CR, ose CR, CR. Lead EC-12 e E

Ary2 E-1 CR, Monitorig CNMT isolationCommunator C

0E.0 0

E-0 CRM uI valvesSO-Communicator CR1 Lad CR. Procedur Primary, E-Seuoftdary,Lcladsle CR.

apeSO uoySemUe

40C, rd ECA1 E-0'-Pn edr EC-12'etr CR, Monioringof 30AB AT. S Comunttalor SamPl Sis Swe,-emtm CR, Lead Transiton to Transitn t Transitton to CR, MonCondngtro Access t The lert Dcd

20 ECA-12 ECA12 ECA-12 Communicator Sae/s AUX.LDG Local Rdtare CRosLead CP, Proedue CR, Proedure CR, Proedum R Men Dm

_r CR, SLte SIGCN

25 CR ed ECA-1.2 ECA-1 2 ECA412 CMotrogATTh SO-I Communicatr rt CR, Lead CR, Proedure CR.,

E CP, Procedure CR, MontorngC CR.

Sample S/Gor CR,Lead ECA-12 ECA-12 ECA-1.2 CRDMonttog Communikator

20.

CR, Lead CR. Procedure CR, Poedure CR, Pcedt CR, MonitDrig CR, Samp Si's

35

~ECA.1.2 ECA-1.2 ECA.12 CommunicatrSape/s CR, Lead CRA Prcedure CR, Procedure CR, Procedure CR. Mooiing CR, Sample S/G's

40 ECA-12 ECA-2 eC. P Moniba Communicator CR. Lead Transfion to Ttrson oto Transtio t CR, Monitog

45 t-I1 E-1 T-I

50 LOCA OLtside Containmreot with a are Fire Ss I

CRF HCO Co STA PRIMARY SECONDARY EXTRA SECURITY SHIFT TECH.

NOT COVERED_.

EPiP CP CLead cedue Seoondary,imm ABocTour tBipTou E-0 edmediat ActAoTourreactor

00 Actions

_____Acton_

re__

C.

CR, Procedure Primsary.

ri

Scedary, F' ire Seorigade Fire Btigade CR.

Fire Brigade Localy cdose CNMT CR tea i-C 0

Ei-C Captain Captain Communicator Isotation values CR Lead CR, Procedure Primary, B-Seconidary, CR. MonitrinFe Brigade Fire Brigade CR.

Lucatfy close CNMT

110 E-C 0

ti-C 0m Captain Captain Communicator FmBiaeIsolation vnies CR, tLeadi CR Pocedure Primaryr. E-Se Ddf.

CR, Mon*iNg Fire Brigade Fire Brigade CR, FieBigd apl

/s 0AB adsunt to ti-C 0

ti-C Captain Captain Communicator Fr rgd apeSGs 3O&.Sre ta

15

_ _ _ _ __ _ _r__

_ _ _ _ _ _ _ Oes. A lT S -1 CR. L.ead Tramnso Tronson to Transiton to CR. Monltldg FieBiae Fire Bade CR.

ieBiae apeSG Corol~

A~n~

lert RDlated ECA.l12 ECA-I.2 ECA.12 Captain Captais Communicator AirUX.gde SaLleWG..es At SO-I ca adto

20 AUX. BLDG., ATT: SD-1___ tEserseny CR ed CR, Pocedure CR, Pocedure CR, Procedure CR ahf ieBiae Fire Brigade CR. u Fre Brigade Sample SOs ATT: S-I

25 CR ed ECA.12 ECA-2 EA1.2 C.Mn Captain Captain Communicator CR, Lead CR Pocedtar CR, ProcedureCR C R

CR,Proctedum Fire Brigade Fire Brigade CR.

ieBiae apeSO

30 ECA-12 ECA-1 2 ECA.l12 CR oioigCaptain Captain Communicatr FmBiae Sml /

CR, Lead CR Pocedure CR, Pocedure CR, Pro'edur CR, Montring FmVa&

Fire Brigade CR, Fr rgd apeSO

35 ECA-1.2 ECA-2 ECA-I.2 Captain Captain Communicatr CR, Lead CR, Pocedure CR, Procedure CR. Pocedure CR. MonodgFire Brigade FIre Brigade CR, Fire Brigade Sample SOs

40 ECA-12 ECA-1.2 ECA-1.2 5d 5 Captain Captain Communicatr CR, Lead Transtion to Transition to Transhion to CR, Montinrg Fire Brigade Fire Brigade CR, u

45 Ei-1 ti-1 ti-I Captain Captain Communiator rire Brigde

0

0

_ 6 00 0

-F

Design Basis SGTR Ss I

CRF HCO co STA PRIMARY SECONDARY EXTRA SECURITY SHIFT TECH.

NOT COVERED EPIP CR Lead

lCR, Procedum re marya SecondayIrm Locafly trip tie CR, ead I nm6edeate E-0 ediate ActionSraco

00 ActIns CR, Lead CR, Procedure Pdmary, E-Secondary, Locally clon CR,

115

~~~E-O 0

E-0 NMT solatlon ConrurricOn 15CR Lead 3CR Procedure P E-mary, CR Locally dose CR.

20 CR Lead 6RPndn C,rd-0 0

se*

CR, MontoRrig CNMT 5o Commnirartc

10 E-0 "naen Communicato CR raad nsItion to Transion to TransItIon to Locan adirlutne" Survey Steam CR.

Sml I'

CR. ead E-3 E-3 E-3 ICR, Monitorirg of 300A&B Linies CommuicaorSmle/s 2

CR, Lead CR. Procedure CR, Pmcedue CR Procedure Isolate flow from CR, Samp SG's CR,0Lead E-3 E-3 CR, Monilring, Rutue SG ComSapl

/G atorD.i.

3 CR, Lead CR, Pmcedure CR, Prcedure CR, Pmocedun C, Moniodn ATT. Ruptured Isolate flow from CR, Samp SG's C,ea E-3 E-3 5

E3 m

CR Moni RptrAd S/G CommunIcato Samp S/

4CR lead CR, Procedure CR, Procedure CR. Prcedure CR M tring ATT: Rurd CR C.0 a

E-3 E-3 E-3 H

CATR R

SDAR EXTR SHIG' DSCR, Lead CR, Pmedure Pm, oE R Pecdum mofb A= r ph ATTre Bdad CRmuiao Sampl Bdad ociy os Nt

30 CR.

E' EF3 Communicabr SampbS/G' 25 CR Lead CR, Procedure CR. Procedure CR. Procedure CR, MoniCR.ng ATTi Captai CRF Sample S/G's 30 R Led CR, Puw Cll, 1e

t P

CR. Monoirg AlT: Rurire AlT: Srbde CommunicatorFieBiae Smb/G ATD-C5 R, lead R. Prceur CR, Procedure CR. Proedure CR, Monitorng Captah SD-lal CR.uitO ltBDae SmhSGs lbl 1

lRpw do E-3 e-3 E-3 Communicatr Samp S/Go

5CR. Lead

50

6

=-

=-

Design Basis SGTR Coincedent with a Large File I

Ss CRF HCO ]

co STA PRIMARY ISECONDARY EXTRA SECURITy. SHIFT TECH.

NOT COVERED EPIP CR. Lead CR. Procedure Pmmeary, Seodry1mA Tour TB. Tour raco

00 E~~-0 AI1ons edaActionsreco CR, Lead CR, Procedure Primary.

B-Seonrdary.

Fire Brigade Fire Brigade CR.

Fire Brigade LO cl ose CNMT

05 E-0 0

E-0 Caytata Captain Communicator lsotation naices CR. Lead C-0 P ged Prmr.

E ecnay o6or Fire Brigade Fffe Brigade CR.

Fire Brigade Locarty diose CNMT

10 E-0CR Prcdr Prmr E

-0Scnay R

otrn Captain Captain Communicator Isolation "Ilves CR, Led Trasitio to Tamlfio to Tnsftntoroe Brigade FrBrgde C,Locat adisutmnt of CR,3LeE 3

E -3 CR, Aonitorig Fireain Captai CR.mn rFire Brigade Sample S/Gas 3D10A&S, Survey Steam

15 6-6-

ati ati omnctrfines CR. Lead CR, Procedur CR. Procedure CR. Procedum Re oitrn Fire Brigade Fire Brigade CR.

File Brigade Sample SIGs~ Isolate flow fom Raptured Alert Declared

20 E-~63 E-3 E-3 CR otngCaptain Captain Communicator S/G CR, Lead R Pod CR, Pro cedure CR Procedure CR. Monitoring FL-a Brigde Fire Brigade CR Fr rgd apeSG Isolate flow from Ruptured

25 6-3 6-3 E-3 Captain Coptain Communicator FrBigd SapeIGs S/G, ATT: Ruptured S/G CR. Lead CR Pmeu CR Pmedr CR CR. uedrn Fire Brigade Fire Briade CR, FirsBrgaetSmplfSGo o/G AT:RupuduS/G

30 E-3 E-3 6.3 CR MntrrgCaptain Captain Commrunicatur F Biae SmlSI' St.ATT D-Ruptured_SIG, Isolate flow from Ruptured CR, Lead CR,

'Proedure CR, Procedure CR meu R Mntering Fl Brgd Fire Briade CR.

Fire Brigade Sample S/Gas S/G, ATT: Ruptured S/G,

35 6-3 E-3 E-3 C,Mn Captain Captain Commrunikatr ATT: SD-I CR ed CR. Proedure CR. Procedure CR. Procedure CR. Monrorig Fire Briade Fire Brigade CR.

ieBiae SmpeSG T:S-CR,4ead E-3 E-3 E-3 Captain Captain CommunicaturFieBgae SmlSI' AT:D1 45CR. Lead ri

60

0:60 D

0W it-

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 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 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 Item Not Covered Discussion Security: Open vital area doors 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.

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.

Page 10 of 13 5/23/03

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. 1 Only required if Condensate Storage Tank Level is less than 5 feet. Should not be a concem 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 concem 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 Auxiliary Building. Security Available Item Not Covered Discussion no issues Security Event and Subsequent Fire in the Auxiliarv Building, 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 5/23/03

LOCA Outside Containment 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 RGI.97 instruments provide the initial basis for concluding that there is no SGTR.

LOCA Outside Containment 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 RGI.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

[noissues l

Page 12 of 13 5/23/03

Design Basis SGTR with Large Fire Item Not Covered Discussion Locally close CNMT isolation Only required if the CIVs 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 S/G Needs to be performed to limit release paths when the 30 minute responder becomes available, enabling transfer of duties.

ATT:SD-l 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 5/23/03

Attachment V RG&E NERP Staff Augmentation Estimated Travel Times

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

NUREG-0654 l < 15

>15 Min.

>30 Min.

>45 Min.

> 60 Min.

Ginna NERP Title Position Title or Expertise

[ Min. 1 30 Min.

• 45 Min.

60 Min.

Plant Operations and Assessment of Operational Aspects 1

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 -

Communicator) 3 4

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

Senior Manager (EOF 2

1 l-Recovery Manager Director) l l

l l

8 Senior Health Physics 1

4 1

l 1

TSC Dose Assessment (HP) Expertise Manager EOF Dose Assessment Manager Page 1 of 4 5/23/03

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.

30 Min.

45 Min.

60 Min. X 9

(There is no NUREG title 6

9 5

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)

II HP Technicians (In-plant 7

9 3

2 2

Shift RP Technicians (Note2) surveys)

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

(Note2)

(Note2)

Chem Technicians Shift RP Technicians (Note2)

Plant System Engineering, Repair and Corrective Actions 13 Shift Technical Advisor N/A N/A N/A N/A N/A 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)

Page 2 of 4 5/23/03

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

Page 3 of 4 NUREG-0654

< 15

>15 Min.

>30 Min.

>45 Min.

> 60 Min.

Ginna NERP Title Position Title or Expertise l Min.

• 30 Min.

45 Min.

60 Min.__l Protective Actions (In-plant) 21 HP Technicians Included (11) above.

Shift RP Technicians (Note 2)

RP Technicians RP/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 (There is no NUREG title l

Local support N/A for this position)

Site Access Control and Personnel Accountability 24 SecurityPersonnel N/A I_

Total J42 2 55

+/-

32 T_ 9

_8 l

~~~~~~~(29%)

(38%)

l(21%)

l (6%)

l (6%)

5/23/03

NOTES Note 1 The estimated staff augmentation travel times are based on the projections of Yahoo! Maps, using 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 Ginna support organizations 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 the proposed enhancements are actually implemented.

Page 4 of 4 5/23/03

Attachment VI Results of Unannounced Off-hours Call Out Drill

Results of Unannounced Off-hours Call Out Drill (Notes 1, 4)

NUREG-0654

• 15 Min.

>15 Min.

> 30 Min.

> 45 Min.

> 60 Min.

Not Ginna NERP Title Position Title or Expertise 30 Min.

• 45 Min.

5 60 Min.

Available Emergency Direction and Control (Emergency Coordinator)

I Designated facility 2

1 TSC Emergency Coordinator manager Notification/Communication 2

(There is no NUREG 2

1 2

TSC Communicator title for this position -

7 Communicator) 1 E

C Radiological Accident Assessment and Support of Operational Accident Assessment 3

Senior Manager (EOF 1

1 1

EOF Recovery Manager Director) l 4

Senior Health Physics 2

1 4

TSC Dose Assessment (HP) Expertise Manager EOF Dose Assessment Manager 5

(There is no NUREG 1

4 5

15 Survey Center Manager title for this position -

(Note 4)

Survey Team Member Offsite Surveys) l 6

(There is no NUREG Included (5) above.

Survey Team Member title for this position -

Onsite Surveys) l 7

HP Technicians (In-4 7

3 2

7 Shift RP Technicians (Note 2) plant surveys)

RP Technicians RP/Chem Manager 8

Rad/Chem l

l l

l 2

Chem Technicians Technicians (Note2)

(Note2)

(Note2)

(Note2)

Shift RP Technicians (Note2)

Page 1 of 3 5/23/03

Results of Unannounced Off-hours Call Out Drill (Notes 1, 4)

NUREG-0654 I 15 Min.

>15 Min.

> 30 Min.

> 45 Min.

> 60 Min.

Not Ginna NERP Title Position Title or Expertise I 1*30 Min.

45 Min.

• 60 Min.

Available Plant System Engineering, Repair and Corrective Actions 9

Core/Thermal 2

Nuclear Assessment Hydraulics 10 Electrical 1

1 1

3 Electrical/I&C Assessment l l Mechanical 2

2 Mechanical/Hydraulic Assessment 12 Mechanical 1

1 1

Maintenance Personnel Maintenance (Note 3) 13 Rad Waste Operator Not required.

14 Electrical 1

1 1

Maintenance Personnel Maintenance (Note 3) 15 Instrument & Control 1

2 Maintenance Personnel Technician (Note 3)

Protective Actions (In-plant) 16 HP Included (7) above.

Shift RP Technicians (Note 2)

Technicians RP Technicians RP/Chem Manager TOTAL l

8 20 16 9

47 (Note 4)

Page 2 of 3 5/23/03

NOTES Note 1 The response times are from the declaration of the event until the time of arrival of the individuals at the designated emergency response facility.

Note 2 The Ginna Shift RP Technicians are qualified in both HP and radiochemistry analysis. The number of qualified Ginna Shift RP Technicians who responded and their response times are not repeated for the listed NTREG-0654 function of Rad/Chem Technician.

Note 3 The number of Ginna maintenance personnel listed are those that were designated at the time of the unannounced off-hours call out drill. Additional responders from each of the maintenance groups are being added.

Note 4 The overall response times and the number of responders was negatively impacted by issues associated with the automated activation system, including:

Outgoing phone calls by the automated system were still being made 60 minutes into the event There was an inability of a number of responders to call in to the 800 number to determine the basis for the pager activation.

Numerous issues with individuals pagers were reported Page 3 of 3 5/23/03

Attachment VII Regulatory Commitments

List of Regulatory Commitments The following table identifies those actions committed to by Rochester Gas & Electric in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments. Please direct questions regarding these commitments to Mr. Thomas Harding, 585-771-3384.

5/23/03 REGULATORY COMMITMENT l

DUE DATE Addition, including training, of 18 one hour July 31, 2003 responder positions to the 13 current one hour responders.

Addition, including training, of one 30 minute July 31, 2003 responder to address the on shift difference with respect to the Rad/Chem Technician position.

Addition of a Dose Assessment Manger to the July 31, 2003 list of responders who are activated at the declaration of an Unusual Event.

A reinforcement by senior management of the July 31, 2003 expectation that NERP responders will respond immediately upon being notified and not wait for additional time. This expectation will also be discussed within the NERP.

Completion of the corrective actions associated July 31, 2003 with the unannounced off-hours call out drill.

Submittal of the revised Nuclear Emergency July 31, 2003 Response Plan (NERP).