ML18086B229

Emergency Response Facilities.

ML18086B229
Person / Time
Site:	Salem
Issue date:	12/31/1981
From:	Public Service Enterprise Group
To:
Shared Package
ML18086B163	List: ML18086B162 ML18086B228 ML18086B229
References
RTR-NUREG-0696, RTR-NUREG-0814, RTR-NUREG-696, RTR-NUREG-814 GL-81-10, NUDOCS 8201040130
Download: ML18086B229 (162)

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Ps l~G Public Service Electric and Gas Company

.Emergency Response Facilities

-* Salem Gen~rating Station Unit No. I Unit No. 2 Docket No. 50-272 Docket No. 50-311 License No. DPR-70 License No. DPR-75 DECEMBER 1981

• PSE&G RESPONSE TO NUREG 0696

1.0 INTRODUCTION

Public Service Electric & Gas Company (PSE&G) will use the facilities and systems described in this response to provide improved emergency response to accidents.

Facilities include the Control Room (CR), onsite Technical Support Center (TSC), onsite Operational Support Center (OSC), and nearsite Emergency Operations Facility ( EOF); systems include the safety parameter display system (SPDS) and nuclear data link (NDL). The TSC, EOF, and CR are required facilities specified by NRC regulations, 10 CFR SO, Appendix E, Article IV.E.8 and Appendix A, GDC 19. The SPDS and NDL systems and the OSC facility are included to provide the complete response to NUREG-0696. This document provides PSE&G's response on the functional criteria for Emergency Response Facilities (ERFs) and on the integrated support these facilities will provide to the* Control Room. ERF design and implementation have been integrated with PSE&G's implementation of NUREG-0654 (FEMA-REP-1), Revision 1, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants."

1.1 Background The accident at Three Mile Island led to studies and regulatory requirements which identified and mandated extensive improvements in the management of accidents at nuclear power plants. Identified improvements in PSE&G's Emergency Preparedness program include:

Establishing formal PSE&G, local, State, and Federal organizations to enhance management and effective coordination of emergency response support; Developing integrated emergency response facilities and data systems to aid in accident management; P81 95 01 1

Providing comprehensive and accurate informa-

• tion needed to assess conditions at Salem Generating Station and its environs prior to, during, and following an accident; Providing an improved capability of PSE&G to provide recommendations to State and local authorities for actions to protect the public; and Providing transmission of more accurate information to Federal, State, and local emergency response organizations, and general public.

To support PSE&G 's emergency preparedness activities and responsibilities, the emergency response facilities and systems to be provided are the TSC, OSC, EOF and SPDS. They will operate as an integrated system to support the Control Room in mitigation of the consequences of accide,nts and to enhance the capability to respond to abnormal plant conditions. These facilities and systems will help provide a graduated response capability dependent on the severity of an emergency. Severity conditions are classified into emergency classes (in order of increasing severity) by NUREG-0654, Revision 1, Appendix 1 and utilized in the Emergency Plan: *

(1) Notification of Unusual Event (2) Alert (3) Site Area Emergency (4) General Emergency 1.2 Control Room The Control Room is the onsite location from which the nuclear power plant is operated. It contains the instrumentation, controls and displays for:

Nuclear systems, Reactor coolant systems, Steam systems, Electrical systems, 1 Safety systems (including engineered safety features), and Accident monitoring systems *

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The Control Room is staffed by licensed reactor operators, senior reactor operators, and a senior reactor operator designated as the Senior Shift Supervisor as required by Technical Specifications.

Safe operation of the reactor and plant manipulations remain under the control of a licensed senior reactor operator, a reactor operator, or Senior Shift Supervisor at all times.

During abnormal operating conditions, these responsibilities increase significantly, and include actions to:

Diagnose abnormal conditions; Perform corrective actions; Mitigate abnormal conditions; Manage plant operations; Manage emergency reponse; Inform Federal, State, and local officials; Recommend public protective measures to State and local officials; Restore the plant to a safe condition; and Recover from abnormal conditions.

Initially, Control Room personnel assume all of these responsibilites.

1.3 Emergency Response Facilities In assisting Control Room personnel to mitigate accident consequences and respond to abnormal operating conditions, the emergency response facilities (ERFs) function during emergencies to provide the following services:

Assist reactor operators to determine the plant safety status; Relieve reactor operators of peripheral duties not directly related to reactor system manipulations; Prevent congestion in the Control Room; Provide assistance to operators by technical personnel having comprehensive plant data at their disposal; Provide a coordinated emergency response by technical and management personnel; P81 95 01 3

Provide reliable communications between onsite and offsite emergency response personnel; Provide a focal point for development of recommendations for offsite actions; and Provide relevant plant data to the NRC for analysis of abnormal plant operating conditions.

Many ERF functions will be performed through use of systems to gather, store, and process data for display in the Control Room, TSC, EOF, and NRC Operations Center. Emergency response facilities will be provided with a common Data Acquisition System (DAS), which will provide plant data to the emergency response facilities. This system will be fully integrated, with one system providing data to the SPDS, TSC, and EOF.

It will have provisions for future addition of an NDL.

The DAS will not rely on the P-250 plant process computer.

Personnel designated for ERF duty will be trained to follow procedures specified in the Emergency Plan to ensure timely emergency reponse. The Emergency Plan and Procedures provide for staffing levels, task assignments and procedures to be followed in the event of an emergency.

1. 3 .1 Technical Support Center The Technical Support Center (TSC) is an onsite facility located close to the Control Room. It will provide plant management and technical support to reactor operating personnel in the Control Room during emergency conditions. It has technical data displays and plant records available to assist in detailed analysis and diagnosis of abnormal plant conditions and any significant release of radioactivity to the environment. The TSC will be the primary communications center for the plant during an emergency. TSC resources will be managed by the Emergency Duty Officer who will assist Control Room operators by handling the administrative items, technical evaluations, and contact with offsi te activities, relieving the operators of these functions. The

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Technical Support Center is located onsite in a building adjacent to the Turbine Building and connected by an enclosed walkway. A detailed description of its location, construction, functions, and staffing appears in Section 2.

1.3.2 Operational Support Center The Operational Support Center (OSC} is an onsite assembly area separate from the Control Room and the TSC where operations support personnel report in an emergency. Direct communications are available between the OSC and the Control Room and between the OSC and the TSC so that personnel reporting to the OSC can be assigned duties to support emergency operations.

The Operational Support Center is a dual use area, comprised of the Senior Sh~ft Supervi-sor's office and a walled corrider between the Control Rooms. A detailed description of its location, construction, functions and staffing appears in Section 3.

1. 3. 3 Emergency Operations Facility The Emergency Operations Facility (EOF) is the nearsite support facility for management of overall PSE&G emergency response (including coordination with Federal, State, and local officials), coordination of radiological and environmental assessments, and determination of recommended public protective actions. The EOF will have appropriate technical data displays and plant records to assist in diagnosis of plant conditions to evaluate the potential or actual release of radioactive materials to the environment. A senior PSE&G official in the EOF, the Emergency Response Manager, has responsibility for organizing and managing PSE&G offsite resources to support the TSC and Control Room operators. An interim EOF is located in Quinton, New Jersey eight miles from the TSC. A new Emergency Operations Facility will be located in Salem, New Jersey, 7.5 miles P81 95 01 5

from the TSC. A detailed description of the new fa.cility, its location, construction, functions and staffing appears in Section 4.

1.3.4 Safety Parameter Display System The safety parameter display system (SPDS) will provide a display of plant parameters from which the safety status of plant operation may be assessed in the Control Room, TSC, and EOF.

The primary function of the SPDS is to aid Control Room operating personnel in making rapid assessments of plant safety status.

Duplication of SPDS displays in the TSC and EOF will improve the exchange of information between these facilites and the Control Room and assist management decision-making. The SPDS will be operable during normal plant operations modes and during all classes of emergencies. The Safety Parameter Display System will be a computer based system fully integrated with the data systems for the TSC and EOF. A detailed description of the SPDS appears in Section 5.

1.3.5 Nuclear Data Link The Nuclear Data Link (NDL) will not be provided as part of the emergency response facilities. Provision will be made for its future implementation, as described in Paragraph 6.

1.4 Activation and Use Activation and use of the ERFs will be determined by the emergency class and specific conditions surrounding an accident in accordance with the Emergency Plan. As a minimum, within the Operational Unavailabilty Criteria, the following conditions shall apply:

(1) The SPDS will be operable during all plant operating modes, including accidents.

(2) Activation of the onsite TSC and OSC is optional for a Notification of Unusual Event emergency cla.ss, and is required for Alert and higher classes *

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(3) Activation of the nearsite EOF is optional for Notification of Unusual Event and Alert emergency classes, and is required for Site Area Emergency and General Emergency classes.

Until the TSC, OSC, and EOF are activated, all of their important functions will be performed in the Control Room. When the TSC is functional, emergency response functions, except direct supervision of reactor operations and manipulation of reactor system controls, will shift to the TSC. Plant administration, technical support functions, and contact with offsite activities shall be performed in the TSC. The OSC will provide a place for operations support personnel to be in direct communication with the Control Room and other operations managers for assignment to duties in support of emergency operations. When the EOF is activated, the functions of providing overall emergency response management, monitoring and assessing radiological effluent and the environs, making offsite dose projections, providing recommendations to state and local officials, and coordinating with Federal officials will shift to the EOF. Table 1-1 outlines the transfer sequence of emergency response functions from the Control Room to the TSC and EOF under the various emergency classes in accordance with the

• Emergency Plan.

The level of ERF staffing will vary according to the severity of the emergency condition. Staffing criteria for each emergency class are fully detailed in the Emergency Plan.

1.5 Reliability ERF data systems, instrumentation, and facilities will be designed and constructed to provide a high degree of reliability. Reliability criteria for ERF systems, instrumentation, and facilities are described in terms of unavailability.

The operational unavailability of 0 .01 will be met on the basis of definitions outlined in this paragraph.

Operational unavailability = Downtime Operating Time P81 95 01 7.

where Operating Time is defined as the accumulated time the plant is above cold shutdown, beginning at the time the ERF is declared operational.

Downtime is defined as the accumulated time the ERF is unavailable when the reactor is above cold shutdown.

The following conditions will be used to determine availability of ERF Data Systems:

Loss of ERF data system hardware required for performance of required functions. Loss of peripheral devices will not be considered a loss of the ERF data system if sufficient peripherals remain to permit performance of function.

The ERF data system will not be declared unavailable due to loss of a redundant power source or input, unless an ERF function is lost. The purpose of redundancy is to assure continued operation of the ERF.

The ERF will not be declared unavailable due to loss of instrumentation not required by Technical Specifications for normal plant operation *

• The ERF will not be declared unavailable on the basis of unavailability of an input due to a scheduled outage. This is not a practical concern as the Technical Specifications preclude removal of key sensors from circuit during plant operation unless one or more redundant sensors are available. Such a requirement would seriously impact plant maintenance operations, and would discourage maintenance personnel from performing maintenance on equipment beyond the minimum requirements. The potential degradation of system performance caused by deferred maintenance greatly outweighs the need for availability of multiple inputs to a system which is essentially peripheral to plant operations.

If operating experience shows that ERF data systems reliability is significantly impacted by a device or class of devices, and corrective action is taken to improve the reliability of the device, the effect of the improvement will be factored into the unavailabili-ty calculation. The historical failure rate of the P81 95 01 8

device will be deleted from the calculation and the anticipated failure rate of the improved device will be substituted. This number will be revised again following the first year's operating experience to provide a base estimate of present system reliability.

The SPDS cold shutdown unavailability goal of 0.2 will be defined as:

SPDS cold shutdown unavailability = Downtime

-=c-0-=1-,d=--s"'""'h,....u-t-=d,....o_w_n_T-..,...im-e where Cold Shutdown Time is defined as the accumulated cold shutdown time beginning at the time the SPDS is declared operational.

Downtime is defined as the accumulated time the SPDS is unavailable when the reactor is at cold shutdown, including in the refueling mode.

The following conditions will be used to determine SPDS unavailability:

Loss of ERF data system hardware required for SPDS functions. The loss of peripheral devices will not be considered a loss of the SPDS if sufficient peripherals remain to *permit performance of its function.

The SPDS will not be declared unavailable due to unavailability of instrumentation not required by the SGS Technical Specification during cold shutdown or refueling.

The SPDS will not be declared unavailable due to loss of a redunaant power source or input, unless the SPDS function is lost. SPDS function will not be considered lost if instruments required by Technical Specification are available.

Scheduled outages for preventative maintenance will necessarily comply with Technical Specification requirements for instrument availability. The SPDS will not be declared unavailable due to removal from service of instrumentation in accordance with Technical Specification requirements.

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Specific justification for the above positions is:

1. A large number of instruments will become unavailable during plant maintenance programs.

For example, in-core instrumentation, in-core thermocouples, rod control, and pressurizer instrumentation are unavailable during most of a refueling cycle. It is obvious that a 0. 2 unavailability is unattainable if these devices are considered.

2. Imposition of a requirement to consider unavailable instrumentation in the cold shutdown unavailability goal, will impede plant maintenance and plant betterment outages, with the potential impact of reducing the quality and overall reliability of plant systems.

Power sources to the TSC and EOF will be redundant and HVAC systems will be provided with maintenance capability as required in Pargraph 4 of NUREG 0696.

2.0 TECHNICAL SUPPORT CENTER 2.1 Function

• The Technical Support Center (TSC) for Salem Generating Station, Units 1 and 2 will be controlled and operated by PSE&G. It will serve as the emergency operations work area for designated technical, engineering and station management personnel. Facilities will also be provided for a small staff of NRC personnel and other designated personnel required to provide technical support.

PSE&G commits to operating the TSC so in accordance with the functional requirements of Paragraph 2 .1 of NUREG-0696, with the exception of the minimum walking time requirement.

The TSC will provide facilities and equipment to support staff performance of four major functions:

( 1) Provide plant management and technical support to plant operations personnel during emergency conditions.

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(2) Relieve reactor operators of peripheral duties and

• communications not directly related to reactor system manipulations.

(3) Prevent congestion in the Control Room.

( 4) Perform EOF functions for the Alert, Site Area Emergency and General Emergency Classes until the EOF is functional.

The TSC will have facilities to support plant manage-ment and technical personnel assigned there. Use of the Technical Support Center during emergencies and recovery operations is described in the Emergency Plan. TSC facilities may be used by designated personnel for normal daily operation, training and emergency drills. The function of TSC facilities will, however, be restricted to functions compatible with those performed during an emergency. All plant data collected by the DAS can be accessed from the TSC.

2.2 Location The TSC will be located in the Clean Facilities Building adjacent to the Turbine Building and within the plant security boundary as shown in Figure 2 .1.

Average travel time between the TSC and the Control Room is 3. 5 minutes, including clearance of an automatic security checkpoint which controls access to the Control Room. Passage through the security checkpoint is acquired by using an magnetically encoded card incorporated into authorized employees' badges.

The security checkpoint can be disabled by security personnel if appropriate. The security checkpoints are designed to limit access to vital areas, while permitting personel to leave vital areas unhindered.

Therefore, there are no barriers to travel from the Control Room to the TSC. The route between the TSC and the control room is as follows:

1) Exit TSC through a door into the east third floor stairwell of the Clean Facilities Building.

2) Proceed down the stairwell to the first floor.

Exit the Clean Facilities Building into a covered walkway.

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3) Proceed through walkway to Turbine Building, and enter Service Building at the staircase located in the southeast corner.

4) Proceed up the stairwell to the third level.

Proceed through security checkpoint (card reader) into Service Building, and walk to Control Room Complex. Proceed through the second security checkpoint into the OSC, and through a door into the Control Room.

Radiation monitors having a sensitivity range of

• 0001-10 R/hr are located along the TSC-Control Room route as follows:

1) East third floor stairwell of the Clean Facilities Building.

2) On the wall of the covered walkway between the Clean Facilities Building and the Turbine Building.

3) Turbine Building at the foot of the stairway for use of personnel returning to the TSC.

4) Turbine Building stairway, first level (lOO'elevation).

5) Turbine Building stairway, third level.

The Turbine Building stairway is the route of travel between the Control Room TSC and Health Physics checkpoint which controls access to the Auxiliary Building and Containment.

PSE&G requests exemption from the TSC location requirement concerning walking time. No location which could meet the habi tibili ty and size requirements is available within existing buildings at Salem Generating Station. As described in this document, the TSC will have a significantly improved communication and instrumentation capability, when compared to the interim TSC upon which this requirement was based.

These enhanced capabilities should decrease the need for face to face interaction. In addition, there appears to be no technical basis for the two minute requirement.

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The aforementioned travel route is covered, but not shielded or ventilated. Personnel exposure will be limited by controlling the number of trips and through the use of protective apparatus, as appropriate.

2.3 Staffing and Training Upon activation of the TSC, designated personel will report and achieve full functional operation in accordance with the Emergency Plan. Activation of the TSC will ensure that only designated operating personnel are in the Control Room during the emergency and that needed technical support will be provided without obstructing actual plant manipulations or overcrowding the Control Room.

The designated TSC staff consists of sufficient technical, engineering, and station management personnel to provide the needed support to the Control Room during emergency conditions. The Emergency Duty Officer will coordinate activities in the TSC and direct personnel who will interface with the Control Room, the OSC, and the EOF.

The level of TSC staffing varies according to the severity of the emergency condition. Staffing for each emergency class is detailed in the Emergency Plan.

PSE&G has developed training programs to ensure the TSC functions effectively and TSC staff personnel are aware of their responsibilities during an accident. In addition, to maintain proficiency, the TSC staff will participate in periodic TSC activation drills in accordance with the Emergency Plan. Operating procedures and staff training

• in the use of data systems and instrumentation will contain guidance on the limitations of instrument readings including whether the information can be relied upon following such events as accidents resulting from earthquakes or the release of radiation.

2.4 Size The TSC consists of a complex of adjacent areas. The TSC complex size and arrangement fully comply with the requirements of Paragraph 2.4 of NUREG-0696. The TSC is divided into five functional areas.

PBl 95 01 1;3

( 1) The operational area and enclosed NRC room with work facilities for 25 people, approximately 2000 ft 2 *

( 2) An enclosed conference room, approximately 300 ft 2 *

( 3) A Bunk Room for 10 persons, approximately 250 ft 2 *

(4) Personnel Support Facilities.

( 5) The Technical Document Room, approximately 900 ft2.

The first four areas are located on the third floor of the Clean Facilities Building. Attachment 2.1 shows the assignment of - working space, and equipment.

Attachment 2. 2 shows the interaction patterns between personnel. Attachment 2. 3 shows the functional arrangement/relationships. Attachment 2.4. depicts the communications and functional interrelationships of the TSC organization. The Technical Document Room (TDR) is located on the second floor directly beneath the other areas and is readily accessible via an adjacent stairwell. Personnel Support Facilities, (lockers, toilets and showers) are located on the third floor.

2.5 Structure The TSC complex will be able to withstand the most adverse conditions reasonably expected during the design life of the plant, including earthquakes, high winds and floods. The TSC is not located in a Seismic Category I structure and will 'not be qualified as an Engineered Safety Feature (ESF).

The Clean Facilities Building is steel framed, supported on concrete filled pipe piles with 24" thick floor slab, block walls and/or prefabricated steel

,siding panels, and a steel roof deck covered with built-up roofing. The second and third floors are 6" thick concrete and the interior walls are concrete block.

The building is connected to the Turbine Building by an enclosed walkway. Access to the Control Room is gained by walking from the Clean Facilities Building through P81 95 01 14

the enclosed walkway to the Turbine Building, and then directly into the Service Building. The Central corridor of the Service Building provides the normal access to the Control Room.

2.6 Habitability The Technical Support Center Operational Area, NRC Room, TDR, and Conference Room will be provided with radiological habitability equivalent to that required for the Control Room. It will comply with General Design Criterion 19, Standard Review Plan 6.4, and NUREG-0737 "Clarification of TMI Action Plan Requirements" Item II.b.2. The ventilation sy~tem will be activated by TSC personnel. It will not be Seismic Category I qualified, redundant, instrumented in the Control Room, or automatically activated. A redundant source of power and repair parts sufficient to perrni t restoration in the event of failure will be provided. The TSC Ventilation System will include high efficiency particulate air and charcoal filters. The system will be divided into zones to permit the most vital areas (the operational area, NRC Room, enclosed conference room, and Technical Document Room) to be isolated from the balance of facility. The functional status of key components and dampers of the ventilation system will be indicated in the TSC operational area.

Shielding from direct radiation will be accomplished through use of cement block walls and a poured concrete roof over the most vital TSC areas.

The radiation monitoring system provided for the TSC consists of portable monitoring equipment dedicated to the TSC. The type and placement were determined by separating the various rooms and walkway areas into "contamination zones, and providing monitors at each end of the zones so that levels can be established prior to passage. This equipment will be used to continuously indicate radiation dose rates and airborne radioactivity concentrations inside and outside the TSC while it is in use during an emergency. This system includes local alarms set to provide early warning to TSC personnel of adverse conditions that may affect the habitability. Detectors will be able to distinguish the presence or absence of radio-iodines in

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concentrations as low as lo-7 microcuries/cc. Potas-sium iodide will be provided for use by TSC and Control Room personnel. TSC ventilation filter system capacity will be independent of the thyroid blocking provisions.

Personnel protection equipment will be provided in the TSC for the staff who must travel between the TSC and the Control Room or EOF under adverse radiological conditions. The protective equipment will allow TSC personnel to continue to function in the presence of low level airborne radioactivity or radioactive surface contamination. Anticontamination clothing and respiratory protection equipment will be provided.

Equipment provided in the TSC is listed in the Emergency Plan. This equipment will be inventoried and maintained to assure availability during an emergency.

The personnel assigned to the center during an emergency are trained in the use of the equipment.

Instructions are also provided for their use.

If the TSC becomes uninhabi tabl'e, vital TSC functions will be transferred in accordance with the Emergency Plan. Habitability will be determined by a trained Health Physics Technician.

2.7 Communciations The TSC will be the primary onsite communications center for the station during an emergency. It will have reliable voice communications to the Control Room, the osc, EOF, and NRC. The primary function of this system will be plant management communications and the immediate exchange of information on plant status and operations. Communications with State and local operations centers will be furnished in the TSC to provide early notification and recommendations to offsite authorities prior to activation of the EOF.

The TSC voice communications facilities will include means for reliable primary and backup communication.

The TSC voice communications will include private line telephones, commercial telephones, radio networks, and intercommunication systems as appropriate to accomplish TSC functions during emergency operating conditions.

PSE&G will provide a means for TSC telephone access to commercial telephone common-carrier service that PBl 95 01 16

bypasses any local telephone switching facilities that may be susceptible to loss of power during emergen-cies. Spare commercial telephone lines to the plant will be available for use by the TSC during emergencies.

The TSC voice communications equipment will include:

Hotline telephone (located in the NRC portion of the conference room) on the NRC emergency notification system (ENS) to the NRC Operations Center; Dedicated telephone (located in the NRC portion of the conference room) on the NRC health physics network (HPN);

Dedicated telephones for management communica-tion with direct access to the Control Room, OSC, and EOF; Dial telephones that provide access to onsi te and offsite locations; Dial telephones utilizing the PSE&G microwave system to the Bell System in Newark, N.J. This system bypasses the local Bell System network and provides Newark foreign exchanges to the TSC and other key locations on-site.

Dial telephone between work areas of the TSC (Operational area and Technical Document Room)

Communications to PSE&G mobile monitoring teams and to State and local operations centers prior to EOF activation.

PSE&G will provide two dial telephone lines for NRC use when the TSC is activated. PSE&G will furnish onsite access to the NRC to facilities and cables for the ENS and HPN telephones.

Facsimile transmission capability between the TSC, EOF, and NRC Operations Center will be provided *

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A station of the plant paging system will be located within the TSC. to provide additional communications flexibility.

A further description of how the TSC communication needs are met is presented in the Emergency Plan.

2.8 Instrumention, Data System Equipment, and Power Supplies Technical Support Center instrumentation data system equipment and power supplies will function independent-ly of actions in the Control Room and without degrading or interfering with Control Room and plant functions.

TSC instrumentation will derive its information from the Data Acquistion System described in Section 7. 0.

The TSC electrical equipment load will not degrade the capability or reliability of any safety related power source. Circuit transients or power supply failures and fluctuations will not cause a loss of any stored data vital to TSC functions. Sufficient alternate or backup sources will be provided to maintain continuity of TSC functions and to immediately resume data acquistion, storage and display of TSC data if loss of primary TSC power occurs. Total TSC data *system reliability will be designed to achieve an operational

• unavailability goal at 0.01 during all plant operation conditions above cold shutdown as described in Section 1.5 of this submittal. Plant data processors and related instrumentation and equipment will not be used in providing essential TSC functions. Provisions may be made in the TSC for utilizing data obtainable from these devices in peripheral tasks. SPDS display equipment used in the TSC will not be seismically qualified but will meet TSC data system equipment reliablity and performance criteria. The SPDS display in the Technical Support Center will be capable of retrieving the same displays available in the Control Room.

2.9 Technical Data and Data System The TSC Technical Data System will not retreive, store or process data acquired directly from the plant. It will receive its data from the common DAS, which will have this capability *

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The data system will provide access to accurate and reliable information sufficient to determine:

( 1) Plant steady-state operating conditions prior to the accident.

(2) Transient conditions producing the initiating event.

(3) Plant systems dynamic behavior throughout the course of the accident.

The data set available to the TSC will be the complete data set from the Data Acquistion System (DAS). This data set is described in detail in Attachment 8.1, "Salem Nuclear Generating Station Unit *2 Compliance with USNRC Regulatory Guide

1. 9 7", which was transmitted to USNRC Licensing Branch 3 on April 2, 1981.

Data storage and recall capability will be provided for the TSC data. Fourteen hours of data storage capacity will be provided. The time resolution will be one sample per second for the majority of variables; some variables may be recorded at a slower interval consistent with the nature of the data. This capability will permit a minimum of two hours of pre-event data to be recorded. Capacity wi 11 be provided to record at least two weeks of additional post event data with reduced time resolution. Archival data storage capability will be provided with the potential to transfer data between active memory and archival data storage without interrupting the TSC data acquisition real time operation.

Four 19", color CRT's will be provided, plus a hardcopy device to allow TSC personnel to perform their assigned task consistent with Emergency Plan requirements. Each CRT is portable and can be moved to any of the work*

stations shown in Attachment 2.1. TSC displays will include alphanumeric and limited graphical representation of plant systems, variables, in-plant radiological variables, and meteorological information. The system will have the ability to trend information using a time history display.

P81 95 01 19

2.10 Records Availability and Management

• The TSC will have a complete and up-to-date depository of plant records and procedures at the disposal of TSC personnel to aid in technical analysis and evaluation of emergency conditions. This will be accomplished by making the Technical Document Room, which contains all plant records, a dual use facility used by the Station as a technical library during normal operations and by the TSC during an emergency. This will insure that TSC personnel have ready access to up-to-date records, operational specifications, procedures, and drawings.

Drawings included are schematics, assembly and wiring diagrams, and other drawings which show the as-built condition of plant structures, systems down to the component level and the in-plant system arrangements.

Station administrative procedures provide for the continual update of documents stored in this facility.

3.0 OPERATIONAL SUPPORT CENTER 3.1 Functions The Operational Support Center (OSC) is an onsite area separate from the Control Room and the TSC where operations support personnel will assemble in an emergency. The OSC:

Provides a location where plant logistic support can be coordinated during an emergency.

Restricts Cont ro 1 Room access to support

,personnel specifically requested by shift supervision.

When the OSC is fully activated, it will be supervised by the OSC Coordinator, as designated in the Emergency Plan.

3.2 Habitability No commitment to OSC habitability is made herein, although in fact the OSC is located within the Control Room complex. Therefore, OSC habitability is comparable to that of the Control Room. osc P81 95 01 20

habitability is verified by the fixed and portable monitoring equipment which is used to fulfill control room monitoring requirements. Additionally, the Emergency Plan includes provisions for performance of OSC functions by essential support personnel from another designated onsite location. Although provisions for a backup OSC have been provided, there is no requirement for a backup OSC and no commitment is made herein.

3.3 Communications The OSC has direct communications with the Control Room and TSC so personnel reporting to the OSC can be assigned to duties in support of emergency operations.

The OSC communications system consists of communica-tions equipment available in the Senior Shift Supervisor's office, which will support OSC operation.

This includes a dedicated telephone extension to the TSC and one telephone capable of reaching onsi te and offsite locations. Direct voice intercommunications and reliable direct radio communications will be used to supplement telephone communication links. A list of equipment provided in the Control Room complex is contained in the Emergency Plan *

• 4.0 4.1 EMERGENCY OPERATIONS FACILITY Emergency Operations Facility Functions and Integration with Overall Emergency The Emergency Operations Facility (EOF) for Salem Planning Generating Station Uni ts 1 & 2 will be controlled and operated by Public Service Electric & Gas Company

( PSE&G) . It will serve as the nears i te support center for management of the aggregate response to a radiological emergency as defined by NUREG-0654, Revision 1, Appendix 1. PSE&G commits to operating the EOF so as to fulfill the functional requirements of paragraph 4.1 of NUREG-0696.

The EOF will provide facilities and equipment to support staff performance of four major functions:

(1) Management of overall emergency response activities P81 95 01 21

( 2) Coordination of radiological and environmental

• (3)

( 4) assessment Determination of recommendations for protective actions for the public Coordination of emergency response operations with Federal, state and local agencies in accordance with the Emergency Plan.

Activation and use of the Emergency Operations Facility will be at the option of PSE&G in the Notification of Unusual Event and Alert emergency classes. The option will be exercised depending upon management's evalua-tion of the potential consequences of the situation based upon the nature of initiating conditions, trends subsequently perceived, and results of actions taken to mitigate potential consequences. EOF activation will be mandatory in the event of declaration of a Site Area Emergency or General Emergency.

The activated EOF will be managed by a PSE&G Vice President, who is designated Emergency Response Manager. He will direct PSE&G's offsite response activites and coordinate actions with and provide appropriate support to the Technical Support Center Emergency Duty Officer. The EOF will be staffed by PSE&G, federal, state, and other emergency personnel

• designated by the SGS Emergency Plan.

A full array of equipment will be provided in the EOF for aquisition, recording, display and evaluation of plant operating parameters, and radiological and meteorological data. The data will be analyzed and evaluated to determine the nature and scope of any protective measures which may be recommended to state and local officials for protection of the public health and safety, if the magnitude and potential effects of a radioactive release dictate. The equipment array will include a duplicate of the display of information collected by the Safety Parameter Display System

( SPDS), which is described in detail in Section 5.

Radiological and meteorological data systems are described in paragraph 4.8. All equipment, displays, and instrumentation to be used to perform essential EOF functions will be located within the EOF.

P81 95 01 22 ,

Facilities will be provided in the EOF (as depicted in the EOF layout, Attachment 4.1) for NRC, FEMA, New Jersey, Delaware and local emergency response agency personnel responsible for implementing emergency response actions for protection of the general public.

This arrangement will enhance coordination of activities and exchange of information among participa-ting agencies and the PSE&G emergency response organization. These agencies will also operate from other offsite control centers located at their respective agency facilities.

Facilities will be provided for PSE&G, NRC, New Jersey, and Delaware public information offices. A Press Center will be provided for periodic dessemination of information to the public through the news media by spokespersons designated in the Emergency Plan (normally the Emergency Response Manager).

PSE&G will provide normal industrial security for the EOF complex to ensure EOF activation readines,s, including lock and key control, a personnel identifica-tion system, exterior lighting, and periodic patrols by police and/or private security guards. When the EOF is activated security protection will be upgraded. At least one security guard will be on duty at all times

• to control access. All access to the facility, with the exception of the Press Center, will be through a single controlled portal. All other access points will be secured. The Press Center will be isolated from the remainder of the facility and a separate press entrance will be provided (See Attachment 4.1).

In accordance with the Emergency Plan, the EOF will be activated periodically for personnel training and emergency preparedness exercises. Normal use of the facility, which is within PSE&G's Southern Training Center, will not degrade EOF activation readiness, operations or equipment reliability.

4.2 Location, Structure, and Habitability The Salem Emergency Operations Facility will be located in PSE&G's Southern Training Center on Chestnut Street in Salem, New Jersey. This site is located 7. 5 miles from the Technical Support Center. The site location P81 95 01 23

• is judged to provide operational and logistical benefits with regard to its relationship to the area's transportation system. Salem is at the intersection of the two state highways (Routes 45 & 49) in the area and to county and municipal emergency response off ices and resources. Salem is approximately 8 miles by road from the New Jersey Turnpike and Interstate 295, via Route

49. Three county highways, Routes 557, 540, and 581 connect to Routes 45 and 49.

The city of Salem is served by a freight-only railroad and an airfield capable of accommodating small commercial aircraft. In addition, the

• Southern Training Center has a helicopter landing pad. There is also a landing pad at the Salem Generating Station.

This will make possible rapid movement of personnel between the station and the

• EOF. This transportation network makes the EOF readily accessible by road and air, to designated personnel of all agencies and activities assigned an emergency response role by the Emergency Plan. An extensive network of local roads, depicted on the Eastern United States (Wilmington)

U.S. Geological Survey Map (Attachment 4.2) assures rapid coverage of the Emergency Planning Zone by radiological monitoring teams.

The physical structure of the facility has been well engineered for the design life of the plant. The building will be a 65, 0 00 square foot structure on reinforced concrete footings and floor slab, with supporting steel columns, beams and joists. The built-up roofing material will be supported on a steel deck.

The EOF will conform to all applicable building codes and has been designed to withstand winds and floods with 100 year recurrence frequency.

The State of New Jersey Department of Environmental Protection identifies the 10 year and 100 year high water levels at the EOF site as 7.1 feet and 8.9 feet above mean s'ea level, respectively. The floor elevation of the EOF is 9.0 feet. The elevation of the road to the EOF is slightly over 4 feet. Thus, record high water levels would flood the access road and preclude access to the EOF by vehicle and could P81 95 01 24

hamper activities of mobile monitoring teams in some

• areas.

helicopter.

The EOF would continue to be accessible by Internal EOF operations would continue without adverse impact.

The SGS Final Safety Analysis Report, Environmental Report, Operating License Stage, Appendix B-Report, Site Environmental Studies, identifies high winds with a 100 year recurrence frequency as having a maximum velocity of 100 miles per hour. It is not anticipated that such winds will significantly affect self-contained internal EOF operations. This is due to the strength of building construction and the availability of back-up power. However, activities of mobile monitoring teams would have to be suspended. Under such conditions, extremely high dispersion factors would provide high assurance that radiation exposures would be correspondingly low. Remote monitoring would continue to be available to the extent transmission lines survive. Similarly, data transmission could be adversely impacted by damage to microwave and radio antennae and transmission lines, particularly, if winds were accompanied by electrical storms, which are often associated with squall lines, tornadoes and hurri-canes. Under such circumstances, atmospheric condi-tions* could be expected to intermittently af feet data transmission and communcations.

PSE&G does not plan to establish a backup EOF as required by NUREG 0696, Paragraph 4.2. PSE&G's analysis (Attachment 4.3), substantiates the contention that the Salem EOF can, with regard to radiological considerations, be treated as if it were farther than 10 miles from the TSC and therefore requires no backup EOF.

This analysis employed assumptions contained in Regulatory Guide 1.4 with respect to dispersion factors utilized in accident analysis and dose assessment.

These values were used to determine radiation exposure at a "typical" site 10 miles from the TSC. These values were compared to factors calculated using the methodology specified in Regulatory Guide 1.45 to determine exposures at the EOF, 7. 5 miles from the TSC. The site specific dispersion factors are based upon the relatively favorable area meteorology as documented in the SGS Final Safety Analysis Report

• P81 95 01 2~

(Environmental Report Operating License Stage, Appendix A-1), (Attachment 4.4). Downwind radiation exposures at the EOF for a given release from Salem would range from 22 to 95 percent of that of an EOF located 10 miles from a "typical" reactor site for the first four days. Subsequent exposures extending through 30 days of the release would be only 19 percent higher than at the "typical" 10 mile EOF. However, analytical studies such as the WASH 1400 Reactor Safety Study indicate the source term diminishes rapidly with respect to time following an accident. In excess of four days dispersion factors are very low at either the 7.5 or 10 mile distance (i.e. X/Q approx. 2xl0 -7).

Therefore, the EOF will be accessible during periods of radiation release.

NUREG-0696 requires that EOF areas in which dose assessments, communications, and decision making will take place, be shielded from external radiation to a protection factor greater than or equal to 5. However, the foregoing analysis supports the position that the location of the EOF 7.5 miles from the TSC is at least equivalent in terms of exposure and habitability to a "typical" reactor site EOF situated ten miles from the TSC. PSE&G has, on this basis, requested relief from the requirement of NUREG-0696, paragraph 4.2., for the

• need for a backup EOF and installation of special habitability features. This request dated April 20, 1981 was directed to the USNRC Division of Licensing, which has not yet ruled on the matter.

The favorable meteorology and EOF location will ensure that performance of EOF functions would not be interrupted by radiation releases and thus, will provide continuity of dose assessment, prediction, communication and decision-making.

Twenty sets of protective clothing will be maintained at the EOF, in accordance with the Emergency Plan. In addition, mutual support agreements with other utilities in the region include providing emergency equipment, including radiation survey devices and protective clothing.

Sufficient potassium iodide for the staff will be stored in the EOF emergency equipment locker.

• P81 95 01 26

Additional supplies will be available from Radiation Management Corporation, Philadelphia, Pennsylvania.

Emergency procedure EPI-15 provides for issuance of potassium iodide when a calculated iodine does of 10 REM or greater to the thyroid gland is likely to be received. A description -of the dual channel analyzer used to determine airborne I-131 concentrations is presented in Section 10 of the Emergency Plan and in Section IV of the Emergency Procedures. Detection limits for I-131 are less than IE-7 micro ci/cc if not masked by noble gases. Masking is not expected to be a factor due to use of silver zeolite filter cartridges in sample collections.

Ten full-face respirators with charcoal filters are maintained in the EOF. However, as discussed previously, airborne contamination is not expected to present a major problem at the EOF due to its location, the favorable area meteorology and the upgraded ventilation system.

4.3 Staffing and Training The EOF will be staffed in accordance with the Emergen-cy Plan Section 3, to ensure proper and effective overall management of available material and human resources, including on-going evaluation and integration of response activites during and following an accident.

EOF functions will initially be performed in the Technical Support Center until the EOF becomes fully functional. Total EOF activation has been shown to require between four and six hours, due to times required for notification of EOF personnel and for many of them to travel from northern to southern New** Jersey.

These times are justified by initial performance of EOF functions in the TSC until EOF activation. EOF activation time and the interim management of initial overall response from the TSC are in accordance with the Emergency Plan.

Activation of the EOF and notification of the EOF staff are in accordance with Emergency Procedures EPII-1 through EPII 7. These procedures have been reviewed by

• P81 95 01 27

• the NRC, Region I. The procedures are structured as check lists of sequential action statements which clearly define each element of the EOF activation/personnel notification process. A listing of procedure titles indicates the scope of this process.

EPII-1 Emergency Response Manager Preparation to Assume Responsibilities.

EPII-2 Site Support Manager Preparation to Assume Responsibilities.

EPII-3 Radiological Emergency Manager Response.

EPII-4 Notification of Corporate Emergency Response Personnel EPII-5 Emergency Paging of Corporate Emergency Response Personnel EPII-6 Off-site Administrative Support EPII-7 Testing of Emergency Procedure EPII-4.

All EOF equipment can be operational within 60

• minutes. Emergency Procedures have been developed based upon detailed analysis of emergency drills. The procedures delineate in detail and in sequence, the tasks essential to EOF operation. Task performance is allocated to appropriately trained and qualified individuals. Long-term Augment, Emergency Response Organization (Attachment 4.5) depicts the relationship of the EOF staff to the overall emergency organization. The Proposed Equipment Layout (Attachment 4. 6) depicts positioning of equipment in functional areas.

Functional Arrangement Relationship Charts (Attachment 4.7) and Interaction Patterns Diagrams (Attachment 4.8) explain how the location of functional areas and equipment is cons is tent with pat terns of vital inter-personal interaction and machine utilization.

The essential pathways of information exchange within the EOF are depicted in the Communication/Function P81 95 01 28

Relationship Charts (Attachment 4.9). Operational

• Sequence Diagrams (Attachment 4 .10) depict the direc-tional flow of essential EOF tasks. Staff training has been and will continue to be provided to ensure all personnel are aware of and proficient in performing their assigned duties. These procedures and the associated training will include characteristics and use of all data systems and instrumentation described in this document (as equipment becomes available),

including limitations upon their effectiveness due to conditions which may be encountered following a serious accident.

To maintain operational proficiency, periodic EOF activation drills will be conducted in accordance with the Emergency Plan. Such drills will include partici-pation of the full EOF staff and operation of all faci-lities and equipment designated for performance of EOF functions.

The combined effect of technical and operational train-ing, realistic drills and assignment of experienced technical engineering and management personnel who nor-mally perform related or analogous duties to emergency response positions, is a highly qualified, trained and coordinated EOF organization. Attachment 4. 5 depicts the relationship of the EOF organization to the overall emergency response organization.

4.4 Size The EOF will meet or exceed the space requirements of paragraph 4.4 of NUREG-0696. Approximately 18,170 square feet of floor space in the Southern Training Center will be designated for use as the Emergency Operations Facility. This area includes a 2650 square foot Press Center. Of the remaining 15, 520 square feet, 12,509 square feet will provide 75 square feet of work space per person for a staff of up to 164 persons and 2432 square feet for conference rooms.

Additional space will be available in the building to accomodate another 100 persons in the unlikely event of a situation in which a greatly augmented staff would be required. Normal EOF occupancy by all concerned parties/agencies, is not expected to exceed 80 persons. The Emergency Plan designtes a minimum total of 71 emergency personnel assignments *

• P81 95 01 29

• The functional layout of the EOF (Attachment 4.1) depicts designated work spaces:

(1) Space for EOF data system equipment for data transmission and reception (Data Center, Communications Center)

( 2) Space to repair, maintain and service equipment displays and instrumentation (in Southern Training Center workshops & labs)

(3) Space to accommodate communications equipment and its use by EOF personnel to perform their assigned functions (Communications Center, dedicated telephone lines throughout the EOF)

(4) Space for ready access to functional displays of EOF data. (Data Center; provisions for installation of remote terminals in the Dose Assessment Area)

(5) Space for storage of plant records and historical data or space for the means to readily acquire and display the records (Southern Training Center Library)

(6) Space for PSE&G emergency response activities (7) Office space for state, local, and FEMA personnel (8) Separate office space to accommodate a minimum of ten NRC personnel during emergency activation of the EOF. (NRC offices)

Attachment 4 .1 depicts the EOF functional layout and illustrates that work areas are situated to facilitate the effective interaction necessary to accomplish oper-ational objectives. Pe.rsonnel are assigned to work areas in functional groups. Groups which perform rela-ted tasks and therefore would have the most need for face-to-face interaction are , in most cases, located adjacent to one another. Each work station will be equipped and staffed as appropriate to its function.

The following chart provides a listing *of work area floor space and personnel capacity given 75 square feet of work space per person:

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Floor Personnel Personnel

• Area Emergency Response Mgr.

Site Support Mgr.

Site Engg Support Mgr.

PSE&G Public Information Space (Ft2) 552 408 532 408 Capacity 7

5 7

5 Assigned 2

5 2

5 NRC Management 534 7 2*

NRC Technical 942 12 5*

NRC Public Information 408 5 2*

PSE&G Admin. Support 3400 45 10 FEMA 324 4 2*

NJ Bureau of Radiation Protection 324 4 4*

Delaware Emerg. Plan'g & 360 4 2*

Operations PSE&G Rad. Emerg. Mgr. 650 8 2 Data Center 228 3 2 Communications Center 399 5 4 State Rad. Mon & Dose Assessment 960 13 5*

PSE&G Rad. Mon & Dose Assessment 960 13 8 NRC Rad. Mon & Dose Assessment 960 13 5*

NJ State Police 360 4 4*

12,509 164 71 Conference Rooms 2,432 Thus, the EOF will provide 75 square feet of working space for up to 164 persons. Normal EOF manning during full scale activation will be approximately 71 persons.

• Anticipated assignments, agencies may vary assigned strength.

The following list identifies functional areas which will be provided with maps, diagrams, and/or drawings as appropriate for performance of assigned functions and provides the minimum wall space, in linear feet available for their display:

1. PSE&G Radiation Monitoring and Dose Assessment 56'

2. State and Federal Radiation Monitoring and Dose Assessment 45'

• P81 95 01 31

3. PSE&G Emergency Response Manager 32'

4. NRC Management 32'

5. FEMA 32'

6. NJ Bureau of Radiation Protection 32'

7. NJ State Police 32'

8. Delaware Emergency Planning & Operations 32'

9. PSE&G Public Information Office 32'

10. NRC Public Information Office 32'

11. Press Center 64' 4.5 Radiological Monitoring The EOF will comply with the provisions of paragraph 4.5 of NUREG-0696 by providing a radiological monitoring system in the EOF to ensure radiological protection of EOF personnel. The system will employ two state-of-the-art radiation detection monitors which will be dedicated to monitoring the EOF environment.

Radiation dose rates and airborne radioactivity concentrations (beta, gamma, iodine, particulate) inside the EOF will be continuously monitored during all periods when the facility is activated in an emergency. The system will warn EOF personnel of adverse radiological conditions, by automatic activation of local alarms set to trip at preset radiation levels. System sensitivity will permit detectors to distinguish the presence or absence of radio iodines at concentrations as low as lo-7 microcuries/cc. The monitors will be portable uni ts which will be moved to various points in the facility during an emergency to ensure comprehensive coverage.

Four survey meters will be provided which have sensitivity ranges up to 1000 REM/hr. additional EOF radiation-monitoring equipment will include 20 dosimeters (10 high range, 10 low range}, 100 TLD's, 2 air samples and a dual channel analyzer. All radiation monitoring equipment will be stored in the emergency equipment locker.

Instrumentation for continuous monitoring of beta air concentrations will be equipped with a strip chart recorder, an alarm light, and bell. The alarm setting is variable. The instrumentation is maintained and calibrated on a regular schedule by station personnel, in accordance with the schedule contained in the Salem Generating Technical Specifications. The Radiological

• P81 95 01 32

• Emergency Manager's Staff at the EOF will monitor the habitability when the EOF is activated. Personnel assigned to the Radiological Emergency Manager's office have a variety of radiological health physics experience, including nuclear reactor power plant experience.

The alarm for the beta air concentration monitor will be set slightly above background to give an early warning of adverse conditions which may affect EOF habitability. The instrument provides a visual and audible warning of radiation levels.

In addition to the dual channel analyzer for iodine air sample analysis, a beta counter is provided in the EOF for air particulate and swipe sample analysis. The counting room is accessible to EOF personnel and monitoring teams. The counting equipment is stored in the EOF. Backup counting and more definitive analysis is available at the station (emergency situation permitting), the PSE&G Research and Test Laboratory in Maplewood, NJ., and at Radiation Management Corporation in Philadelphia, Pa.

4 .,6 Communications EOF Communications Systems will be in compliance with the provisions of NUREG-0696, paragraph 4.5. Reliable voice commmunications will be available from the EOF to the TSC, Control Room, NRC, New Jersey, Delaware and county emergency operations centers. Normal communications between the Control Room and the EOF will be through the TSC.

The Emergency Plan provides for establishing communications with the following organizations which may be involved in emergency response:

1. State of New Jersey

2. State of Delaware

3. Salem County (NJ)

4. Cumberland County (NJ)

5. New Castle County (Del.)

6. Kent County (Del.)

7. PSE&G (Including radiological monitoring teams)

8. U.S. Nuclear Regulatory Commission P81 95 01 33

The primary functions of the EOF voice communications

• will be:

1.

2.

EOF management communciations with the Emergency Duty Officer in the TSC Communications resources to manage emergency response

3. Communications to coordinate radiological monitoring

4. Communications to coordinate offsite emergency response activities

5. Communications to disseminate information and recommended protective actions to responsible government agencies Reliable EOF voice communications will be provided by primary and backup means, including private commercial telephones, microwave telephones, and radio networks, to facilitate accomplishment of EOF functions under emergency conditions.

Selected EOF telephones have the capability to bypass local telephone switching facilities that may be susceptible to loss of power during emergencies. These systems and the microwave telephone system provide non-interruptable communcations between the EOF, TSC and the control room. At least 12 local switchboard independent commercial telephone lines will be available in the EOF.

Appropriate telephone lines will be provided in all EOF functional areas. A Communications Center will be established and equipped with radios and telephones to provide emergency communications with the agencies listed above.

The following means of primary and secondary communications will be provided to each agency:

4.6.l Emergency Communications with the State of New Jersey Primary communciations between Salem Generating station and the State of New Jersey will utilize private line from the Senior Shift Supervisor's Office, TSC and the EOF to the State Police

• P81 95 01 34

Communications Bureau. This private line is a pick up and ring hot line which will be used to notify and exchange information with the State for all the emergency action levels.

Secondary communications will be provided by unlisted number telephones. These telephones are located in the TSC, Control Room (Senior Shift Supervisor's Office) and the EOF.

Additional communications capability will be provided by a New Jersey State Police Radio (EMRAD) located in the EOF Communications Center, for direct contact with the New Jersey and Delaware State Police.

4.6.2 Emergency Communications with the State of Delaware Primary communications between Salem Generating Station and the State of Delaware will utilize the State of Delaware's National Warning System (NAWAS) link from the Senior Shift Supervisors Office, TSC and the EOF to the State Police Communications Bureau. This system will be a pick up and ring hot line used to notify the

• State for all emergency classes

• Secondary communications will be provided by unlisted number telephones located in the TSC, Control Room (Senior Shift Supervisor's Office) and the EOF.

Additional communications will be provided by the New Jersey State Police Radio (EMRAD) for direct contact with the New Jersey and Delaware State Police.

4.6.3 Emergency Communications with Salem and Cumberland County, New Jersey Primary communications between Salem Generating Station and Salem County, utilizes a dedicated direct line from the Senior Shift Supervisor's Office, TSC and the EOF to County police dispatchers. This system will be a pick up and ring hot line for notifying the County in the event of a General Emergency *

• P81 95 01 35

• Secondary communication is provided by unlisted number telephones located in the Operational Support Center, TSC, Control Room (Senior Staff Supervisor's Office and the EOF.

Additional communications capability will be provided by New Jersey State Police Radio (EMRAD) which will provide direct contact with the local (county) Police.

4.6.4 Emergency Communications with New Castle and Kent County, Delaware Primary communications between Salem Generating Station and New Castle and Kent Counties will utilize a private line from the Senior Shift Supervisor's Office, TSC and the EOF to the county police dispatcher. This link will be a pick-up andring hot line to notify the county in the event of a General Emergency.

Secondary communication is provided by unlisted number telephones located in the Control Room (Senior Shift Supervisor's Off ice), the OperationalSupport Center, TSC and the EOF.

Additional communication capability is provided by a New Jersey State Police Radio (EMRAD) with direct contact with the Delaware State Police, who have the capability to contact county policy by radio and telephone.

4.6.5 PSE&G (Internal Communications)

In addition to the links described above between the Senior Shift Supervisor's Off ice, TSC, and EOF, the normal PSE&G communications system will be used to alert site personnel of emergencies, initiate implementation of Emergency Plan Proce-dures, and coordinate activities with the EOF.

The system will be divided into two functional areas, internal station communications and station-to-offsite communications. The latter system will consist of microwave telephones, police and station security radios and radio telephones.

PBl 95 01 36

The station Security Radio System is a two

• channel system capable of communications within the Plume Exposure Pathway EPZ.

normally used by Station Security.

during an emergency.

This system is One channel will be assigned to the emergency organization Sufficient radios are available to support Security and Emergency Planning needs. It will be used to communicate between the station's onsite and offsite survey teams and personnel performing radiological assessment at the TSC and off-site radiological data collection at the EOF. The EOF telephone system will provide intercommunications between all EOF functional areas, including between the emergency response manager and the EOF super;...

visory staff. An additional independent intercom system will be provided.

Facsimile transmission capability between the EOF, TSC, and the NRC Operations Center will be provided. The facimile device will be tested periodically to ensure its capatibility and operability.

4.6.6 Emergency Communications with The NRC The system for communication with the NRC will consist of 2 direct lines*which will be used to transmit general accident information and radio-logical assessment information. Extensions of both lines will be installed in the EOF, in NRC off ice space. The lines will be on the NRC Emergency Notification System (ENS) and the NRC Health Physics Network (HPN). Two additional dedicated telephones will be provided for use by NRC personnel.

4.7 Instrumentation, Data System Equipment, and Power Supplies The Salem EOF will comply with the provisions of paragraph 4.7 of NUGEG-0696 by providing Safety Parameter Display System (SPDS) equipment, described in detail in Section 5. This equipment will be designed to achieve an operational unavailability goal .of 0. 01 during all plant operating conditions above cold shutdown.

P81 95 01 37

The EOF data system will perform its functions

• independent of personnel actions in the Control Room and will not degrade or interfere with Control Room and plant functions.

This system will gather, store, and display data needed for EOF and NRC personnel to analyze and exchange information on plant conditions with the Emergency Duty Officer. EOF instrumentation will derive its information from the Data Acquisition System described in paragraph 7. 0. The SPDS display used in the EOF will not be seismically qualified, *but will meet the EOF data system equipment reliability and performance criteria. The SPDS display in the EOF will be capable of retrieving the same data displays as the Control Room.

Backup power will be provided to ensure data system availability. Back-up power will be supplied by a natural gas driven engine generator in conjunction with an automatic transfer switch which will activate the generator upon loss of power. The generator will provide electrical output sufficient to supply all facility lighting, the telephone system and all EOF data and communications systems described in this document. Electrical equipment load in the EOF will not affect any safety-related power source. The data system will be designed to preclude loss of any stored data vital to EOF functions due to power supply failure or circuit transient (Reference SPDS, Section 5).-

4.8 Technical Data and Data System The EOF Technical Data System will not retrieve, store, or process data acquired directly from the plant. It will derive its data from the common plant DAS, which will have this capability. The data available will be sufficient to provide general information on the condition of the plant for utility resource management.

The EOF data set will include radiological, meteorolo-gical, and other environmental data as needed to:

(1) Assess environmental conditions (2) Coordinate radiological monitoring activities (3) Recommend implementation of offsite emergency plans.

P81 95 01 38

The data set available to the EOF will be the complete

• data set available from the Data Acquisition System (DAS). This data set is described in detail in Attachment 8.1, "Salem Nulcear Generating Station Uriit No. 2 Compliance with Regulatory Guide 1.97." Also included in the data set will be meteorological variables specified in proposed Revision 1 to Regulatory Guide 1. 23, "Meteorological Measurement Programs in Support of Nuclear Power Plants".

The accuracy of data in the EOF will be consistent with the data accuracy needed to perform EOF functions. EOF data display accuracy* will be equivalent to TSC data displays. The time resolution of data acquisition will be sufficient to provide data without loss of information during transient conditions, al though slower than that available in the TSC, because of data transmission limitations.

EOF data displays of Regulatory Guide 1. 97 variables will meet the criteria for EOF data, but will not be designed to the Design and Qualification Criteria of Regulatory guide 1.97 for display of those variables in the Control Room.

Data storage and recall capability will be provided for the EOF data set. This capacity will be integral with the Data Acquisition System located on site, and will not be located within the EOF. Fourteen hours of data storage capacity will be provided. The time resolution will be one sample per second for the majority of vari-ables, some variables may be recorded at a slower interval consistent with the nature of the data. This capability will permit a minimum of two hours of pre-event data to be recorded. Capacity will be provided to record at least two weeks of additional post-event data with reduced time resolution. Archival data stor-age and the capability to transfer data between active memory and archival data storage without interrupting EOF data acquisition and displays will be provided for all EOF data. A sufficient number of data display devices will be provided in the EOF to allow personnel to perform their assigned tasks with unhindered access to alphanumeric and/or graphical representations of:

P81 95 01 39

(1) Plant systems variables (2) In-plant radiological variables (3) Meteorological information The system will have the ability to trend information using a time history *display. EOF displays will be designed so that callup, manipulation, and presentation of data can be easily performed. Displays will be partitioned to facilitate retrieval of information by the different functional groups in the EOF. This will be accomplished using two display units, and by logically ordering information display pages , on a call up basis. EOF displays will be formated so as to be easily understood by EOF personnel. Display devices will not be provided for news media briefings.

Human-factor engineering will be incorporated in the design of the EOF Technical Data System.

The SPDS will be displayed in the EOF in order to provide PSE&G management personnel and NRC representa-tives with information on current reactor systems status and to facilitate communications among the Control Room, TSC, and EOF. The SPDS system will be capable of presenting multiple displays in the same manner as in the Control Room.

4.9 Records Availability and Management The EOF will comply with the records availability and management provisions of paragraph 4.9 of NUREG-0696 by maintaining in the EOF, hard copy records and other documents as follows:

1. Salem Unit 1 and Unit 2 Technical Specifications

2. Station Operating Procedures

3. Final Safety Analysis Report

4. Salem personnel radiation exposure histories

5. SGS Emergency Plan (includes off-site population distribution data and evacuation plan)

6. SGS Emergency Plan Procedures

7. New Jersey Radiological Emergency Plan

8. Delaware Radiological Emergency Plan

9. Emergency Planning Zone population distribution data

10. EPZ radiological monitoring records

11. Detailed area maps (U. s. Geological Survey)

P81 95 01 40

12. Drawings schematics, diagrams showing plant structures and systems including down to the component level and showing their locations.

13. Employee radiation exposure histories (available over computer from the station)

The above records will be stored in the Southern Training Center Library. They will be periodically updated to ensure currency, completeness and therefore operational utility in an emergency.

Historical/baseline meteorological data will be available in the FSAR. Current plant meteorological data will be available from primary and backup sources (the redundant on-site meteorological towers). This data will be transmitted in digital form to the control room and is relayed over telephone wires to the TSC and EOF. Dose assessment procedures are designed to use either data set.

5.0 SAFETY PARAMETER DISPLAY SYSTEM 5.1 Function The SPDS will provide a continuous display of plant information (parameters or derived variables) during normal and abnormal operating conditions to assist control room personnel in evaluating the safety status of the plant.

The SPDS displays will serve to concentrate a set of plant parameters to aid in assessing plant safety status without surveying the entire Control Room. SPDS displays will be designed with an information level concept. A primary display, having a concise presentation of information, with secondary displays, will present more detailed information about individual areas of the plant to aid in assessing causes of abnormalities and determining corrective actions.

Data for display will be validated. Validation will occur in real time using redundant sensors, sensor comparison with derived variables, etc. When an unsuccessful validation occurs, the SPDS will identify the impacted parameter and will have the capability to display the data from the sensors involved.

P81 95 01 41

A qualification program will be established to demonstrate SPDS conformance to the above functional criteria.

5.2 Location The SPDS will be located in the Control Room with duplicate SPDS displays in the TSC and EOF.

The SPDS display system in the Control Room will consist of CRT's with keyboards. To provide maximum accessibility and visibility, one CRT/keyboard will be placed in a central location in the control board with the other on a nearby vertical panel (RP-3).

Four CRTs with keyboards will be available in the TSC and EOF in order to allow personnel in these locations to access the SPDS displays. This capability will not cause any interference with Control Room operations.

5.3 Size The SPDS display will be of a size compatible with the Control Room design. The SPDS display will be visible to personnel standing in the Control Room. Two displays will be used to allow simultaneous viewing of the primary and a secondary display. It will not interfere normal movement or visual access of the reactor opera-tor to the control console.

SPDS displays will utilize existing space in the Control Room. Provisions will be made for one SPDS display to be visible to the Senior Reactor Operator licensed supervisor, who is supervising emergency operations.

5.4 Status The SPDS display will be designed so that only operat-ing personnel normally assigned to the Control Room will be required for its operation.

J 5.5 Display Considerations The SPDS CRT displays have undergone a preliminary human factors engineering principles review. Attention has been given to the qualitative and quantitative aspects of displayed information.

P81 95 01 42

Multi-color graphic CRT's will be used to facilitate pattern recognition so the operator can readily determine the status of systems displayed by the SPDS, identify problems and determine their severity.

Provisions will be made for the addition of predictive techniques (e.g., Disturbance Analysis System) when they become available.

The SPDS will *be responsive to transient and accident sequences and will indicate the status of the plant during normal and abnormal operating conditions. The SPDS display system will consist of a primary and several secondary displays. The primary SPDS display will automatically be displayed if the operator is not using the CRT to obtain other information (secondary displays).

The primary display will be a pressure/temperature dia-gram consisting of a plot of reactor saturation tempe-rature versus reactor pressure and a plot of allowable termperature {margin) versus reactor pressure. The hottest in core temperature and T hot and T cold for each reactor coolant loop will be contained in the pressure/temperature diagram. Also the primary display will include the location of the hottest in core temperature, reactor power level, reactor pressure and whether reactivity control is normal or abn~rmal.

When the operator observes an abnormality on the pressure/temperature diagram, secondary SPDS displays can be selected to provide more detailed information.

As a minimum set, displays of the following plant systems will be provided:

Reactor Coolant System Residual Heat Removal Safety Injection Chemical & Volume Control System Steam Systems Feedwater Systems Radiation Systems P81 95 01 43

When required, a third level of displays containing

• 6.0 information be available.

on NUCLEAR DATA LINK individual components (e.g.,

Pressurizer), down to the status of sensors (compared to redundant measurements or calculations), will also The Nuclear Data Link will be capable of transmitting all sensor and parameter information. The data link will conform to IBM 2780 Protocol. Details concerning specific transmissions will be based upon additional NRC guidance and will be determined at a later date.

7. 0 ACQUISITION AND CONTROL OF TECHNICAL DATA (DATA AQUISITION SYSTEM) 7.1 Sources of Technical Data The parameters provided to and by this system are listed in Section 8. Isolation devices are provided for all signal interfaces with safety systems to prevent interference, degradation, or damage to any element of the safety system as specified in General Design 24, and IEEE Standard 279-1971, Section 4. The signals will come directly from process or from the process computer interface cabinets, depending upon the reliability and qualification requirements for the data application. Except for some meteorological and radiological data, these inputs will not be processed by a software-programmable device before they enter the DAS.

7.2 Acquisition of Data The data acquisition and distribution system planned for Salem is shown on Figure 7 .1. All inputs that interface with a safety system signal will be isolated by an isolation device prior to connection to the DAS.

7.3 DAS Functional Limitations The DAS will not be subjected to external demands for processing or services that could degrade reliability under accident conditions and will not be interrupted, delayed,

• or in any way impeded or degraded in its P81 95 01 44

function by any such external demands or software installation or changes in any plant equipment. The only exceptions will be the system's internal calibration and self-diagnostic routines. Output data from the DAS will be consistent with readings available to Control Room operators. The DAS will achieve this objective by performing validation tests on A and B variables as defined in Regulatory Guide 1.93.

• 7.4 DAS Design, Verification, and Configuration Control Because the data acquisition system will be the basic source of data for all emergency facilities, its hardware and software configurations and changes will be verified for reliability. Procedures to demonstrate and evaluate the integrity of software and the integrated system will be similar to procedures contained in Nuclear Safety Analysis Center publication NSAC/39, December, 1981, "Verification and Validation for Safety Parameter Display Systems".

Upon acceptance of PSE&G's Implementation Plan for conformance to NUREG 0696, an ERF DAS will be purchased from a qualified supplier.

The questions in NUREG 0814 (pages 6-3 to 6-5, and 7-4 to 7-7) will be answered after vendor is selected.

7.5 DAS Reliability The data acquisition system will provide data access for the SPDS, TSC, and EOF, and the NDL. Therefore, DAS reliability will be consistent with the reliability goals of the SPDS, TSC, EOF, and NDL. *A moderate amount of excess capacit~ and capability will be included in the original design to permit modest increases in parameter monitoring without serious impact on the system.

8.0 EMERGENCY RESPONSE FACILITY INTEGRATION The Emergency Response Facility (ERF) system design insures that the following criteria are satisfied:

0 Operation of any system does not degrade the reliability or performance of any reactor safety or control system or of any safety related displays in the Control Room.

P81 95 01 45

0 Control Room action and system operation does not cause degradation of or interference with functional operation of the ERF.

0 Normal operation of any ERF system or subsystem does not cause degradation of or interference with the functional operations of other systems in those facilities.

0 The station security system will restrict access to DAS hardware and software which are thus protected against unauthorized manipulation of or interference with input signals, data processing, data storage, and data output.

The ERF is a fully integrated data processing system serving all emergency response facilities and systems.

The data set available for display and use in the TSC and EOF will be that described in Attachment 8.1, "Salem Nuclear Generating Station, Unit 2 Compliance with USNRC Regulatory Guide 1.97".

Meteorological variables, will be provided, including temperature, wind direction and speed at three plant levels, temperature differential, dew point temperature (or humidity), precipitation and degree of standard deviation of wind direction, as specified in Regulatory Guide 1.23.

Inputs to the SPDS will include, but not be limited to the following systems:

Neutron Flux RCS Cold Leg Water Temperature RCS Hot Leg Water Temperature RCS Pressure Core Exit Temperature Coolant Level in Reactor P81 95 01 46

• Containment Pressure Radioactivity Concentration or Radiation Level in Circulating Primary Coolant Containment Area Radiation Effluent Radioactivity-Noble Gas Effluent from Condenser Air Removal System Exhaust RHR System Flow Pressurizer Level Steam Generator Level Steam Generator Pressure Main Feedwater Flow Auxiliary Feedwater Flow Plant Vent Radiation Concentration and Flow ERF signals are derived from the DAS which will be interfaced with plant systems, as previously stated in response to paragraph 7.1 of NUREG-0696.

The acquisition of data is consistent with DAS and meets the requirements of Regulatory Guide 1. 97 for acquisition only for those parameters required for display by SPDS.

JS/JD:dmb 12-15-81 P81 95 01 47

Table 1-1. Transfer of Emergency Response Functions from the Control Room to the Technical Support Center and the Emergency Operations Facility Emergency Class Notification of Unusual Site Area General Emergency Response Functions Event Alert Bmergency Emergency Supervision of reactor CR CR CR CR operations and manipulation of controls Management of plant CR(TSC} TSC TSC TSC operations Technical support to CR(TSC} TSC TSC TSC reactor operations Management of corporate CR(TSC,EOF} TSC(EOF} EOF EOF emergency response resources Radiological effluent and CR(TSC,EOF} TSC(EOF} EOF EOF environs monitoring, assess-nt, and ~ose projections nform Federal, State, and CR(TSC,EOF} TSC(EOF} EOF EOF local emergency response organizations and make recommendations for public protective actions Event monitoring by NRC CR TSC(EOF) TSC&EOF TSC&EOF regional emergency *response team*

Management of recovery CR(TSC,EOF} TSC(EOF} EOF EOF operations Technical support of CR(TSC,EOF} TSC TSC TSC recovery operations Note: (CR}, (TSC}, (EOF}, or (TSC, EOF} indicates that activation of this facility (or the performance of this function} is optional for the indicated emergency class.

• One NRC individual also may be stationed in the Control Room.

1 95 01 48.*

• FIGURES AND ATTACHMENTS

Table l.*1 Transfer Sequence of Emengency Responsibilities Emergency Function On Shift & Short Term Long Term Initial Augment Augment Augment Control of Operations CR CR CR Emergency Direction & Control CR TSC EOF Notification of States (Event Class) CR TSC TSC Notification of NRC CR TSC EOF Prot. Act. Recommendations for States CR TSC EOF Prot. Act. Recommendations for Station CR TSC TSC Accident Assessment CR TSC TSC Radiological Exposure Control (Station) CR TSC TSC

-,

Repair and Corrective Actions (Station) CR osc osc Support of Operations CR osc osc Request Federal Assistance CR TSC EOF Public Information CR(HQ) CR(HQ) EOF CR-Control Room OSC-Operations Support Ctr.

TSC-Technical Support Ctr.

EOF-Emergency Operations Facility (HQ)-Corporate Offices Newark

UNIT 2 L_ . J UNIT l a D

• SR. SHll='T SUPERVISORS OFF'ICE CONTROL ROOM NO. 2 UNIT N0.1 UNIT J

AOMINIS. BLD'G.

TURBINE. GEN. BLO'G.

CLEAN SERVICE 8LO'G.

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I F'IG. 2.1 LOCATION OF TECHNICla.L SUPPORT CENTER IN RELATION TO THE CONTROL ROOM

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RELATIONSHIP CHART FUNCTIONS -

TSC - SHORT TERM AUGMENT AL CLOSENESS 0

Absolutely A

Necessary B/U EDD E Especially Important I Important RPE (or) SRPT TSS or Oper 0 Ordinary Close Assm't u Unimportant Eng. Assm't x Undesirable Ad.min Support NJSP Comm DSP (ENS)

Comm NRC (HFN)

NRC (DEPO)

Comm NJ (BRP)

Comm Co's Plume Plotting Q:d REASON Direct face to face Conference l. contact

2. Visual Contact TOR of PRINTS

3. Share equipment

4. Convienience CRT TERMINAL

5. Req. for job XEROX/TC AREA

6. Noise Disturbance TTACHMENT 2.3 FUNCTIONAL ARRANGEMENT 1 of 1 RELATIONSHIP CHART

RELATIONSHIP CHART COMMUNICATIONS -

EDO -

SHORT 'l'ERM AUGMENT VAL CLOSENESS Absolutely Neces.

A Direct,Radio or P B/U EDO E Espec. Important Dedicated Line I

Important SECURITY Limited Use Line 0 Ordinary rdinar Phone INFO SERV Unimportant u No Equip. Require x Undesirable NRC REGION ADMIN SUPP SSS OP/ENG ASSMT PT (RPE)

NJSP DEL SP NJ/DEL Counties REASON

l. Direct Supervision DATA BASE

2. Exchange Info OSC COORD.

3. Improper Pathway OTHER SUPPORT 4. By Face to Face

s. Req. Plans ect. print

.

HQ GROUPS

6. Computer Link

7. Met Only ATTACHMENT 2.4 COMM/FUNCT RELATIONSHIP l of 3 CHART 8. Interc (Emergency Management)

• RELATIONSHIP CHART COMMUNICATIONS -

OPERATIONAL/ENG ASSESSMENT SHORT TERM AUGMENT VAL CLOSENESS Absolutely Neces.

A Direct,Radio or P CONTROL ROOM Espec. Important E

Dedicated Line I Important DATA BASE Limited Use Line 0 Ordinary Ordinar Phone TDR Unimportant u No Equip. Require HQ ENG. x Undesirable HQ COMPUTER INFO ENG-CORE, THERM, HUD ENG -

OFF SITE REASON

1. Direct Supervision

2. Exchange Info

3. Improper Pathway

4. By Face to Face

s. Re~uest plans, print ,

ec *

6. Computer Link

7. Using Installed Inst .

ATTACHMENT 2.4 COMM/FUNCT RELATIONSHIP 2 of 3 e. Intercom CHART (Operational/Eng Assessment)

RELATIONSHIP CHART COMMUNICATIONS -

'l'SC RAD ASSESSMENT SHORT 'l'ERM AUGMENT VAL CLOSENESS RPE or (SPRT)

(TSC) Absolutely Neces.

A Direct,Radio or P CONTROL ROOM AAL

• E Espec. Important Dedicated Line CHEM TECHS I Important Limited Use Line 0 Ordinary ONSITE RAD rdinar Phone MONS Unimportant u No Equip. Require RAD PROT ACT x Undesirable OFF SITE RAD MON TEAMS NJ BRP DEPO NRC HPN DATA BASE PLUME PLOTTER (TSC)

REASON

1. Direct Supervision

2. Exchange Info

3. Improper Pathway

4. Have Face to Face

s. When State has Field Teams ue J.o ed

6. Computer Link

7. Using installed instruments ATTACHMENT 2.4 COMM/FUNCT RELATIONSHIP CHART 3 of 3 B* MET ONLY (Radiological Assessment)

• RAD ASSESSMENT SITE PAR RAD ASSESSMENT OFFSITE PAR OPERATIONAL ASSESSMENT OF PLANT COND CR REOUESTS FOR REPAIR

' CORRECTIVE ACT SUPPORT CR REOUEST TECHNICAL EVALUATION OF - - * -

• OPERATIONAL PROB LEH

• TECHNICAL ASSESSMENT INDICATES SYST/

EOUIPMENT PROB WHICH MAY AFFECT OPS OPERATIONAL OPERATIONAL ASSESSMENT ASSESSMENT PAR PAR DIRECT DETERMINATION OF

"'

UI SITE OFF SITE REPAIR &

CORRECITIVE ALTERNATIVES AND EVALUATE PAR PAR ACT NECESSITY

& PRIORITY OPERATIONAL ASSESSMENT RAD STATE ASSESSMENT INPUT ON ENGINEERING OF REPAIR ' ASSESSMEN'l' LOCAL GOV. CORRECTIVE NOTIFICATION ACTIONS RAD REPAIR ' ASSESSMENT CORRECTIVE J\CT DECISION SELECT BEST ALTERNATIVE DIRECT CR ' NOTIFY SECURITY STATE '

LOCAL GOV OF PAR DIRECT OSC COORDINATOR TO PERFORM INFORM CR OF REPAIR ' OPERATIONAL CORRECTIVE ACT RECOMMENDATION

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• ATTACHMENT 4.3 COMPARISON OF METEOROLOGICAL DISPERSION FACTOR EMERGENCY OPERATIONS FACILITY SALEM NUCLEAR GENERATING STATION X/Q AT 7.5 MILES RATIO SALEM @ 7.5 TIME INTERVAL X/Q AT 10 MILES REG. GUIDE 1.145 MILES VS. TYPICAL AFTER ACCIDENT REG. GUIDE 1.4 ASSUMPTIONS(!) METHODOLOGY (2) SITE @ 10 MILES *

(sec/m 3 (sec/m 3 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> l.2xlo- 5 6.3Xl0- 6 0.52 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> l.2Xl0-S 2.1x10- 6 0.22 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 2.35Xl0- 6 1. exio- 6 0.77 1 - 4 days 7.5x10- 7 1.1x10- 7 0.95 4 - 30 days 1. 6Xl0- 7 l.9Xl0- 7 1.19 (1) for typical site (2) using 5% probability technique and Salem site data (3) for a 624 consecutive hour period the dose would not be significantly affected by the slightly higher 4/Q valve

SMITH-SINGER, METEOROLC GISTS, INC.

1D3 CEDAR SHORE DRIVE MASSAPEQUA, N. Y. 11?SS TEL: 516 - 798*1445 516 - '724-9654

~~~0?-9l*OGTC~L ~~CTIO~

~AT.EM f'SAR I'..!1..Bl.Jr. ~~RyJr.r.~ F._IEC'I'Rir:: ANn GA~ r.o.

l. GFNF RAT, MF'T'l<'Q'ROl.OG Tr. ..\ L FF. A 'f'URF.~.

'T'h~ ~~lP-m  !';i te is rem~'l"'kAbly fl::tt nnrt uncom-pli_cat~rt in all r~sp1=1octs. ThH m~~n winci ~p~ede"'. arP.

relatively ~trong, And thP diRtrihution of tu~-h,dcnce condi. tirm~ typical of open land area~ dPspi t~ thP prrrd.mi ty of Delaware Bay.

Thi:> huT"ri~anl) j R the 111ost prob'i hlP. m01tPoroloei c~ 1 h~~~rd, with a rea$onablP. likeJih~od of 100 milP nPr hou,.. 'D""~k winds ::JR wP.ll ft!== t.h?. f'Rsnci~t.A.rl di-=:turb~.n~P of ttiP 1.ocaJ tide ann wavP. p~_ttE'rns *

• Tn the ~AR. thP P"r.:=d.bi1ity of frt;-ri11Pn~v TP~T'lt i "nf>d 1 becanc:;e of tl'i~ lon~ n~"' ~trP.t~h t>f w~ t"r 1?Y.te.11ni,,~ ~out.hPRSt n.f the site.

s:i'l'I unusuFi11y hi.p.:h of st~blP. condi ti one; wi.th onqhorP fJ ow w::t~

Ca.,.'?f11l ~tu:iy nf t,i,,-.

d::i ti:'! h~~ nnt rf'.'V~R led anything unusu~ l wi tli wi_nti-: fror:i th~ snutheast~ proh?bly be~~u~~ of thP n~rrownr~~ ~f thP np~n wRt~r ~e~tnr ~nd also bAcauqe th~ Ray t"MP"r-at11rP~ ~rP not uF.u~.lly VPry cold.

Tn th~ prP1hrdnary P.VJ:t.lu~ti.on it wa~ con~1ud£'n th~t thP meteornl~eicAl p~ttern~ Rt th~ ~itP woulrl hr f::wor~b1i:- for l!! nuclPar .1n~tallation, Rnd thPrr ~"Ptn$

to h~ nn ~~~d to modify that cnncluF.ion.

IT. SOURr.r~ OF DATA.

A. ~~in !!,P.t~~~nloe\~~l Tow~r.

DA.ta hpr.~mf' Available frnm th~ )00 ft m~t~o.,.ologic~l tow~r loeatPd on Artific,p1 T~l~nn in JunP. of 1969. Thn towPr iP po~Jti~n~d ju~t north of thP aetual plal'\.t ,:; i tr. in Fie;urP l.

A detAilP.d repreRPntation of t~~ mP.tPoro1ogical facility is not nP.cesF;~rv bPcAu~:" nf t.hP ~hnl'1i

~it.y nf the terl""sdn. 'l'h~ tow"r tfatR useri ;,,,* .

• ATTACHMENT 4.4 A.1-1 Amendment 2

this study iR primarily that from the JO and JOO-foot levels, although some data were obtained at the intermediate 150-foot ele-vation. The wind instrumP.ntation consisted of Aerovanes, and the temperature-difference measurements WP.re obtained from aRnirated resistance thermometers. The compiP.te array is shown in Figure 2. The usual precipitation, humidity and auxiliary data are on record also if they are needed for general environmental applications.

B. Delaware City Wind Data.

The Tidewater Oil Company (Delaware City) ha~ operated a wind direction and speed system at an elevation of approximately 250 feet above ground for a considerable period of time.

These data were analyzed some years*ago for a dispersion study and they served as a partial basis for the PSAR estimates. Now they provide an interesting comparison with the site data itself.

c. Standard NOAA Records.

There are a number of regular reporting stations nearby, and they have been drawn upon for additional information where a very long record js desirable.

III. DATA PROCESSING AND ANALYSIS.

The data from the Salem site have been initially processed and transferred to punch card form by the personnel of the Maplewood Laboratory of Public Service. ThP. card data and the original chart rolls have then been forwarded to Smith-Singer for checking and further processing to determine the turbulence classifications. The data have then been trans-ferred to magnetic tape anrt subjected to a series of computer analyses that are on file at Public Service .and Smith-Singer. In the following sections, only those portions of the studies that are considered directly pertinent to the FSAR deliberations are reviewed. Other portions of the meteorological studies have been adapted for specific safety anal-ysis evaluation, and these are reflected in approp-riate portions of the overall document.

A.1-2 Amendment 2

JV. S'T'ANDAJ3J>_MF.TEO'ROT,OGICAL DAT.A.

A. 'T'Qmner~ture.

The monthly ~nd annu;tl distribution~ nf thP. low-lev~l temp~rature are gi.w~n in 'l'ab1~ 1, 1

Rnd a typically mild seacoAst climate is indi-

~atPd. ThP record of this and all other data P.xtends from June, 1.969, through NovP.mbAr, 19?0.

Data are be ine obtained cnntinuouRly, howew~r, nr.ti h~vP hP.en "Partially procesFH~d ~t the ti.mP of this report, but thPrP is little likPlihond that thPy will in any way alter the r.:oncl.u~ion~

rP.~chen hP.re.

B. PrPr.initation *

.

• Simil~r r~cordR of pr~cipit~tion ~re pre-

~entAd in Table £, where the most notahle featurP is thP sucr.:Pc;Rion of very riry month~*during thQ warm half of 19?0.

C. lH P.'h Winrh:;.

ThP main metAoroJoPic?l h~~ard at the SalAm RitP is the hurr5ca~e, ;. ph~nomPnon which fortu-nRt-ely o~cur~ infr~quent]y. The Jocql wind re~ord!=: during thA 1-! year period ::trP ohvi.nnsly in~uffic:ient to providP. ins1ght into the pn~~iblP pef'.lr winti valuP.~, and the 25-yeRr rP.cord av::til-a.ble from thP. Philadelphia International Airport

~ite i~ prPSP.ntP.d as more indicative. of the prob-ahlP distribution (Table 3).

In adrlition. the annlicatinn of P.xtrPm~

v~l.ue thP.ory to existiTl~trecords indic0.tes that onP should anticipate R fa~teRt-milP. wind of

~pproximatPly 100 m/hr R.t S~lPm on R 100-yPar rer.urrP-nce intervRl. SpeC'i fie calculation~ for Atlantic City and Philadelphia give fA.st,:.Rt-mile V?..1ues of 98 and 92 m/hr rP.spectivP.Jy.

Tornadoes ~rP. quite rR.rP btlt nnt unknown ir. this ~rea. The very cnmnletP. ~ummRry prepar~d by Wolfor~l ~howR thRt the ~lo~Pst torn::trin tra~k wa8 cbi:;P.rvPd 011 August ?.~. 1941

• whP.n "* fairly

• A.1-3 Amendment 2

small tornado passed through Wilmington, Delaware and moved on to Swedesboro, N.,J. Using 'J'hom

• s technique 2 for estimating the probability of a torn~do aetually affecting the site, we obtain a prob~bility of 1 in 4JOO years.

D. Ice Storm~.

The site is so close to the open water of Delaware Bay that severe ice storms are very infr1=3quent.

v~ DIFFUSION METEOROLOGY.

A. General Philosophy.

The most difficult aspect of developing a diffusion climatology for a site is the assembly of data that will adequately define the turbulence of the wiYld flt:>w in the layer approprfatP for t~e problem. Numerous bi-directional wi11d vanes have appeared on th~ cornmercil!ll.l market, h1;t n.on~ of them hRVP, oper~ted satisfactorily over an extended period under normal field conditions. A bivane was installed on the 150-foot level at Salem, with the usual extensive loss of data, e.nd lack of con-fidence in thP. records that were obtained.

Alternative techniques of estimating the turbulence usually involve one of two methodRa approximating it from a combination of lapse rate and wind speed measurements, or from the fluctu-ations of a standard wind instrument such as an Aerovane. We believe the latter to be more repre-sentative of the typical problems, and accordingly this presentation is largely based on wind direction range and gustiness data. The lapse rate classi-fication has been used, however, and some of the data are summarized in the report. In this instance the two techniques are in good agreement.

The system used for defining the turbulence is that developed originally by Singer and Smith3 and widely applied in both nuclear and fossil power plant evaluati.ons. The classification is depicted A.1-4 Amendment 2

in Figur'?! 3. wherP. ClasR11>~ T and TT repre~ent.

unst~blP r.onlii tions, Class TI I is the ovP.rc~st stormy i=ituation, Qnrt Class TV is thP. !=tt::tblP.,

inversion flow pattern.

In the PSAR the distributjon of tnrbulen~P clf'ssifi.o::i.t.irms obt~ined from ttiP. nel~w~re City s]tP. tP.n milP.~ NNW of Salem wa~ nrPsentP.d R~

probahly typicRl of thP dii:;pPrRion rPe,imPi:;.

Tn Ta bl P. 4 the new Sa lP.m d::i ta ( )00-foot lev~ l) arP. comp~rP.rt with thP. ~Rrli~r summary from DelRwarP r.ity, And th~ RgrePmPnt i~ vP.ry goo~

oespitP thP. f~ct th~t the inform~tion wa~ obtainPd in diffP.rPnt yP.ars. ThP. n~ly notRhle diffP.rP.ncP i.R that Salem showed a more marked tendP.ncy toward thP. nP.utral, ~la!=;s ITT turbulPnce than din nelaw;::irP City. ThiR R.bberation may be real, but i t ii:; morP.

likPlv th~t the w~ter towP.r nn whi~h thP. Delaware r.ity lnRt.rument wa~ located prodtJCP.d s.omPwhat.

hroRrter and morP. turbulent directjon traces th~n the clP.an installation at Salem. Jn R.ny casP. ~

the niffP.rence has no great significanne ]n the di.spP.rsion P.val1J::i.tion.

At both R~tes, the di~tributions ~~em qujt~

normal for opPn, mi d-1ati tudP locA.tions. The Clai::.s II turb1..t1.P.nr.e dom]nate~ the distribut]on~,

accounting for approximately 60% of all hnurR, and thP stable cases are found in rouehlv 25% of the rem~d.nrier. We had antir.inatAd a not\cP.a.hle inr.rl!'~E;e in thP. frP.quency of f:1a~~ TV condition~

d11r:iT1g the 1 ==ttP spring ;:1.nd early f;UTnmPr at Sa] Pm ..

heCRllRe. i. t is di rP~tly P.XpORed to OVP.r-w~t.P.r f] ow which mi.eht bP. stRblP, but s:tpparently thr> ~omhin ation of infrP.quent wind~ from the lJ0-160° ~e~t.o'r and thP relqtivP.ly mild h~y tempP.rRturP~ d5n not

prod\Jr.P. tnP. ex-p~c+.ed incre~!=le.

C. La use Ra t.P.s.

Jn TablP 5, th~ di~trihution nf lap~~ rates ovP.r the vPar )s i:=hown. These d.~_ta R.irrP.P w~11 wit:h thP. 1nrti~ations of thP. tl\rhuJ.P.n~e cJ~~sif.i cation, in thRt :?.4% of thP. hour~ s::tppeRr to bP

~t~ble, 14% neutr.Ftl and the. rPma:indP.r unf;t:able ..

A.1-5 Amendment 2

Another indication that the water influence iR fairly small at this site is that the diurnal variation of the lapse rate in June (Figure 4) does not Rhow any tendency toward stability in the afternoon hours, and in fact is quite similar to the December (Figure 5) and the annual (Figure 6) patterns.

D. Relation Between Lapse Rates and Turbulence Classes.

As a final comparison between turbulence classes and the lapse rate data, Table 6 is pre-sented. In it,it is clear that the two methods of estimating turbulence are compatible at this site. The vast majority of Class I and Class II turbulence hours are associated with unstable lapse rates, the Class III hours are more nearly neutral, and the Class IV hours are primarily inverRion periods as they ought to be.

E. Wind Patterns.

Complete multivariate distributions of wind direction, speed, lapse rate and turbulence class are available in the computer analysis files, but in this report only the summary data are required and presented.

The distributions of wind speeds at the JO and 300-foot levels as a function of turbulence class are presented in Table ?, where the most notable feature is the very low frequency of calms.

Normally, with an Aerovane as a sensing instrument, calms at the JO-foot elevation are prominent, but the very flat terrain and the air-sea interaction at Salem obviously favor a yigorous wind flow.

Also, the percentage of hours having relatively high ~peeds, reflected in both Tables 7 and 8, is quite large, as one would anticipate in this locality.

The wind direction distributions for all hours and for stable hours only are shown in Figures 7 and 8, respectively. The overall data show the typical northwest peak commonly associated with st~tions in this portion of the country, but the corresponding southwest peak is a bit skewed toward south and southeast, reflecting the typical seabreeze flow in the late spring and summer.

• A.1-6 Amendment 2 L

The stable flow patternR of FigurP 8 arp Amone the most evenly di~tribut~d on rP.nord, with juRt a tiny peak from thP south-southeai:::t.

F. DiRnerston Param~tPrs~

Th~ di~tributions of lapsP. rateR, winn~ ~nd turbu1P.nce c-lA.SRPS alrP.~dy prP.C!P.nt.P.d are adP.qllatP.

to defjnP. thP. diffusion ~eteornlogy of thi~ sitP as quite normal And unC'omplicatPd, hut it i~

important to transl~ t~ the da tR. R~ a.ecura te ly RS posRible into the dispersion p~rametP.rs actually used in nlJmerica l evAl1Ja tions. Since. the exper-iPnC".' with the bi-directional wind vane waF: typi-cally unsuccessful, the m~asuremPnt of hou~ly wind n:i rec ti on ranee waR evaluated a.nd used for pi:;ti-matP.~ of e. These data, se~ar~tPd according to turbulencE'! class, are given for thP. entirP. PP.r.i on of ob~P.:rvat.i.on in Table 9, And it is apparent. that the wind fluctuations at thi~ si.te i:tre. very nP-i:irly idPnticfll to thnsP ~t BrookhavP.n N~t.io.nfl) Labnr~

tnry4 wh~rP the turhulPnc~ clas~ification w~~

originR.lly dPvelopert. 1t thPrPfnrP is rea.i:;on~b1.P to lltilizP. thP. d5f~1slon par~metPrs developPd at that. !;]tP.5 :in this sturly.

OnP f11rt:hP.r poi rit is imoort::ant. ~nrt th~ t is to bi:-. c;urP. that. ri iffu~ ion with sou th-~outheR~ t winrlq frn~ thP opP.n w~ters nf nel~wRrP. Bay ls not significa~t1y different from th~t occurring with Ot~Pr W)nn directions. ~~hlP 10 is ~ rerliCA Of

'T'~blP. 9. ~x~Ppt that only south-sou+.h~a~t w~nrl~

ArP. repri:>~PY"ltP.ti., Obviom~ly tliPrl'- is no di ffPrPnCP.

~~ ~11 othPr indic~tnrA in thP. ~tudy h~v~ sug~P~tPrt, VT. APPl.TCA'J'TON TO 'T'HF !=iAFPTY }.NAT.YSTS 1 Tn t.hP PSAR fl table of thP. leR~t fRvnrAhJ~ nis-pP.r~inn co~ditinns w~R presP.ntert 1 b~~Pn 1Rrg~ly on th~ turhulencP. cla~sifi~atjons And wjnd di~tribution~

at DPJ~w~r~ City. The sub~equPnt ~tudtP~ ~t thp i:;itP i~Relf in no w~y inv~lidatP. the oriRi.nal ~~timatP.

ey~ept. to indic~tP. very po~itivPly th~t thA rhanr.~

of h~vi~g ~ neArl.y fixPd wind dirP~tion ac~ompanied hv Rt~h~Q (~lR~~ TV) c~nrti~ionR Pnrt lipht win~~ i~

P~t.rP.ml'.' 1V rl=!mOtP * "fi'Or th!?.' fi.ri=:t. ho11r or twn fol 1 owi nr

i po~tulat.Pn ::tr.rirfont, ~ 1 m/sP.r. wind from virtuRl1y ~

j:rny rljrP.r.tinn h~R an ar.~P.ptRb]i:> 1"..,..nh~hili.t.y, but fnr A.1-7 Amendment 2

any period longer than two hours, a higher wind speed must be used. Examination of the Class IV turbulence data, for example, shows that only 5.6% of the 30-foot wind data fall in the combined "Calm" and "2-'.3 mi/hr" groups, regardless of direc-tion, so that it is unrealistic to postulatP. an exceptionally restrictive initial 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the accidental release.

VII. CONCLUSION.

Review of the meteorological data obtained at the site, together with similar information from the Delaware City tower on the other side of the river, indicates that the area has a completely orderly an~ cont:! if~tent pattern of di fful'!' 4 on co"l-

~.; ti (ln~, a~~mnrRni.ed by Ufl,:-na11y "1!!.~l:l"lrou~ wino speeds and a varied wind direction di~tribution.

No evidence of unusual or adverse patterns of any kind appearR in the record. .'

• A.1-8 Amendment 2

'

1. Wolfnrd, L. V,t Tornado Occurrenr.e.s i~ the Un:itP.ct StatP.s, u. s. nept. of Commer~P, TP.~h. Paper #20. 1960.

2. ~hom, H. C.s TornRdo Probabilities, Monthly WP.a. Rev., v. 91, #10-12, 196).

J. SingP.r. I. A. and M. R. ~miths RP.lat5on of GuRtiness to Other Meteorological ParamPter~, Jour.

1Vleteor.-;-ro7"T2J. 171-126, 1953.

4. Singer, I. A** John A. Friz~ola. and M. F., Smith!

A ~implifien MP.thod of Rstimati~~ Atmosph~rir.

Diffur;ion ParamP.ters, APCA .Tour., v. 16, 1}11.NOV: 1966.

5. Ri.ne;P.r, I. A. and M. R. Smith: ~j:mrH:::pheric D:i~per~i.on at. BroC'lkhRvPn N::tt.]onal Lahoratorv, Int .. ,Jour.

nf Air & Water Po11ut.ion, 10,-*f25-135, 1966 *

• A.1-9 Amendment 2

TABLE 1

• DISTRIBUI'ION OF HOURLY TEMPERATURES (PERCENT)

TEMPERATURE CLASSF.S (Op)

-20 -10 0 +10 +20 +30 +40 +50 +60 +70 +80 +90

< to to to to to to to to to to to to ill.... +40 MONTH

-

-20 -10 0

-- +10 +20 .+/-2.Q__ +60 +/-1.Q_ +80 .+/-22..... +100 JAN 6 19 44 25 6 <l FEB 6 31 42 17 4 MAR 9 52 35 4 <l APR 9 35 38 15 3 <l MAY 8 36 34 14 6 2

• JUN >9 48 J6 7 <l

• JUL 1 28 54 16 1

• Affi <l 18 54 24 SEP 2 15 30 43 8 2 OCT <l 6 19 33 34 8 <l

• NOV <l 5 20 42 29 4 DEC 1 25 59 14 1 ANNUAL <1 1 10 18 15 14 17 18 5

• 2 months of data A.1-10 Amendment 2

TABLE 2 PRECIPITATION (IN. WATER) 1970 Ra~e of Maximum Hourl~ Rate MONTH 1969 JAN 0.65 .01 to .10 FEB 1.70 .11 to .20 MAR 3.03 .21 to .30 APR 4.54 .51 to .60 MAY 1.39 .21 to .30 JUN 1.. 87 3.89 .51 to .60 JUL 7.18 2.82 1.00 Plus AUG 3.75 1.29 .71 to .80 SEP 2.02 1.47 .41 to .50 OCT 2.92 2.13 .61 to .70 NOV 1.64 5.46 .51 to .60 DEC 6.92 .51 to .60 A.1-11 Amendment 2

TABLE 3 DISTRIBtJrION OF PEAK WINDS PHILADELPHIA INI'ERNATIONAL AIRPOHT (25-year record}

Fastest Mile Month SJ2eed(mJ2h} Direction JAN 61 NE FEB 59 NW MAR 56 NW APR 59 SW MAY 56 SW JUNE 73 w JULY 67 E SEP 49 NE OCT 66 SW NOV 60 SW DEC 47 NW Fastest Mile Observed in Area: 88 mph, North, July, 1931 Estimated Peak Hourly Value: 70 mph A.1-12 Amendment 2

TABLE 4 PERCENTAGE FREQUENCY OF

'I' UR BULENCE CLASSES Salem & Delaware City CU.SS MONI'H I II III IV JAN 6 (2) 62 (65) 13 ( 2) 19 (31)

FEB 4 (3) 57 (64) 16 (5) 23 (28)

MAR 7 (3) 59 ( 66) 12 (6) 22 ( 25)

APR 6 (2) 60 ( 72) 15 (9) 19 (17)

MAY 12 (11) 59 (63) 6 (1) 23 (25)

• JUN 13 (12) 57 (58) *10 (1) 20 (29)

• JUL 12 ( 4) 58 ( 64) 10 (0) 20 (32)

• Affi 12 (3) 53 (65) 10 ( 0) 25 ( 32)

• SEP 14 ( 4) 50 (62) 12 (7) 24 ( 27)

• OCT 8 (6) 52 (62) 14 (5) 26 (27)

• NOV 6 (7) 56 (64) 13 (15) 25 (14)

DEC 4 (8) 72 (51) ' 12 ( 12) 12 (29)

ANNUAL 8 (6) 58 (62) 12 (5) 22 (27)

• 2 mos. of data

( ) data for Delaware City A.1-13 Amendment 2

TABLE 5 PERCENI'AGE FREQUENCY

• OF LAPSE RATES LAPSE RATE GROUP ~ TJOO - T30 1 °F)

-1.6 -0.4 +0.6 -1.6 +2.6 +J.6

< to to to to to to >

MONTH -1.7 -2.!.2 +2.!.2 +b.2. +~ +hl ~ +Ii.6

-

JAN 18 46 11 8 5 5 2 5 FEB 18 37 14 10 6 6 3 6 MAR 20 47 14 6 4 3 2 4 APR 19 45 12 7 5 6 0 6 MAY 30 27 10 8 6 7 5 7

• JUN 32 40 12 6 4 3 1 2

• JUL 25 45 13 7 5 3 1 1

• AUG 30 32 14 8 9 4 2 1

• SEP 24 32 18 9 7 5 3 2

• OCT 19 33 20 10 7 4 2 5

• NOV 13 43 20 8 6 3 3 4 DEC 18 57 15 5 3 1 <l 1 ANNUAL 22 40 14 8 6 4 2 4

• 2 mos. of data A.1-14 Amendment 2

TABLE 6 RELATION BETWEEN LAPSE RATES AND TURBULENCE CLASSES

(-:.;\

TENIP'ft:RATURE OIFFERENCE 1 T~OO-T~O 'PT ~OF)

-1.6 -o.4 o.6 1,6 2.6 3,6

• ru RBUl.ENC"F: to to to to to to

-*- GI~ASS- *--** < -)

• 7

-

-0.5 0,5 1.5 .2. 5 3.5 4.5

-> 4.6 I 5.6 3.2 o.; 0.1 0.1 0.1 0.1 0.1 II 15.4 26 .. 4 7. '.3 3.1 1.6 0.9 o.4 o.6 III 0.7 5.9 2.8 1,0 o.6 o.4 0.1 0.2 IV 1.0 3.7 4,5 3.8 3,6 2.7 1.5 2.4 A.1-15 Amendment 2

TABLE 7 PERCENTAGE FREQUENCY OF WIND SPEED CLASSES

• TURB.

30 FT WIND SPEED(MPH) 4-7 8-12 13-18 19+ ALL

~

CLASS

-

CALM 0.3 o.o 9.5 I o.6 2.5 4.4 1.7 0.7 4.1 20.9 20.0 8.6 1.8 56.1 II o.o 0.3 2.6 5.3 2.6 0.7 11.4 III 1 .. 4 4.2 11.J 5.0 0.9 0.1 22.9 IV 2.8 11.l 39.2 32.0 12.3 2.6 100.0 ALL

~OOFT WIND SPEED(MPH) 0.7 1.9 4.1 2.1 o.6 0.2 9.6 I

II 0.2 1.1 7.2 18.o 18.6 11.4 56.5 III o.o o.o 0.1 0.9 4.8 6.o 11.8 o.4 1.0 3.8 7.1 6.8 3.1 22.2 IV 1.3 4.0 15.2 28.1 30.B 20.B 100.0 ALL l

A.1-16 Amendment 2

TABLE 8 MEAN ANNllA.L WIND ~PEF.DR AT VARIOUS LRVEIB (MPH)

'I'URBULF.NCF.

CLAS~ ':\O FT J00 F'T' t 5.0 6.o II B.O 13.0 III 10.0 lO.O

. ,

IV 5.0 12.0

• AT,L HOURS 7. () ] J. f)

A.1-17 Amendment 2

TABLE 9

• AVERAGE HORIZONTAL RANGE (0)

TURBULENCE CLASS MONTH I II III IV ALL JAN 60 30 20 <10 25 FEB 60 30 20 <10 30 MAR 70 30 20 <10 25 APR 60 JO 20 <10 JO MAY 70 25 20 <10 25

• JUN 55 25 20 10 25

• JUL 65 25 15 10 20

• AUG 65 20 20 10 20

• SEP 60 25 20 10 25

• OCT 60 JO 20 <10 25 NOV 55 30 20 <10 30

• DEC 50 30 20 <10 30 ANNUAL 60 30 20 <10 sigma e 12 6 J-4 <2

• 2 mos. of data A.1-18 Amendment 2

TABT.E 10 AVERAGF. liORIZONT~L RANG'R ( O) FOR WTND DIRF.CTJONS BETWEEN 130 AND 1.60 DEGREES MO~TH I IT ITJ TV ALI.

,JAN 90 40 ?O <10 10 80 30 20 <l.O 10 FEB MAR 60 30 JO <10 10 40  ?.O <10 40 APR 50 MA.Y 70 30 20 <10 30

• JUN 70 30 ?O 10 30

• ,JlJTJ 60 ,o 20 10 20

• AlJf'; ?r> 30 JO <10 30

• SEP ?O JO '.30 <10 30 60 30 20 <10 20 OCT NOV 60 30 30 <10 30 DEC 60 30 JO JO ANNUAL ?O 30 ?0-30 10

• 2 months nf datA A.1-19 Amendment 2

Fig 1 SOURCES OF DATA

' ...\.-1 ..._/,,,,

• Philadeph1a

'...

"

• '

Airport

'

ILMINGTM...

.

Millville

\

A.1-20 Amendment 2

Figure 2

• Salem Nuclear Generating Station Meteorological Tower Schematic 300' 300' f

150' 150' Key

• Ground Level Temperature Wind Speed, 85' Wind Direction r IT 30' 30' A.1-21 Amendment 2

DEFINITION OF FIG. J TURBULENCE ClASSES CLASS I LARGE, LAZY CONVECTIVE EDDIES CAUSED BY HEATING AIR CLOSE TO THE GROUND.

MOST FREQUENT ON SUMMER MORNINGS WHEN WIND SPEEDS ARE LIGHT AND LAKE BREEZES ARE NOT f'RESENT.

CLASS II TYPICAL DAYTIME TRACE HAVING A MIXTURE OF CON-VECTl VE AND MECHANICAL TURBULENCE. FLUCTUATIONS ARE MORE SUBDUED WITH ON-SHORE WINDS THAN OFFSHORE.

CLASS Ill TYPICAL TURBULENCE ASSOCIATED WITH OVERCAST, STORMY, OR NOCTURNAL SITUATIONS HAVING RELA-TIVEL*Y STRONG WINDS .

.MECHANICAL TURBULENCE PREDOMINATES.

CLASS IV CLASSIC TEMPERATURE INVERSION CASE WITH ALMOST NO TURBULENCE EITHER NOCTURNAL OR OR ASSOCIATED WITH DAYTIME LAKE BREEZES, ESPECIALLY IN THE SPRING.

L A.l-22 Amendment 2

FTGURF: 4 DIURNAL VARIATTON OF I.APSE RATE JUJ\fR lQ?O

+3

+2

,.

z 0

H er.

o:;

~

z 1-1

...--.

PL.. +l c........

c

(""\

~

0 0

c

(""\

E-ex::

v,

-l P-c(

H

-2 0 2 4 6 8 10 12 14 16 18 20 22 24 HOURS

• A.l-23 Amendment 2

FIGURE 5 DIURNAL YARIATION OF LAPSE RATE DEC 1970

+3

,

+2 z0 H

Cl)

~

-

r:..

~

z

-

0 +l H 0

rri E-<

0 0 0

rri E-<

2 4 6 8 10 12 14 16 18 20 22 24 HOURS A.1-24 Amendment 2

FIGURE 6

• DTURNlL VART~TTON OF T.APSE R~ 'l'F.

ANNUAL

+3

+2 z

• • --

0 H

u:

0:::

~

zH Ii-0 +l 0

(""\

e-..

0 0 0

C""'I E-<

Ct-i c.r, p...

-1 <

t-1

-2 0 2 4 6 8 10 12 14 16 18 20 22 24 HOURS A.1-25 Amendment 2

STATION SALEM HEIGHT 300 FEET

• PERIOD JUN 1969-NOV 1970 ALL HOURS INCLUDED 350 360 10 290 280 80 260 100 250 110 240 120 230 130 220 140 210 150 200 170 160 190 180

• PERCENT OF ALL HOURS FIGURE 7 A.1-26 Amendment 2

STATION SALEM HEIGHT 300 FEET PERIOD JUN 1969 - NOV 1970 STABLE HOURS ONLY 350 360 10 290 280 80

21 Ffl::.r--t-+-+--l_J 90 260 100 250 110 240 120 230 130 220 140 210 150 200 170 160 190 180 PERCENT OF AI,L HOURS FIGURE 8 A.1-27 Amendment 2

• ADDENDUM 1 TO METEOROLOGICAL SECTION The following is a compilation of atmospheric dispersion factors determined from the data analyzed in the previous section.

.~

'

A.1-28 Amendment 2

• I. LEAST FAVORABLE 30 DAYS.

Estimating the sequence of the meteorological conditions that may result in the least favorable diffusion over a JO-day period following a postu-lated accidental release is not quite the simple problem usually encountered where the site is sur-rounded entirely by populated land. From the purely meteorological standpoint, the least favorable com-bination of conditions would occur with winds that would carry the effluent over water for three or four miles before reaching populated areas, and therefore is not a valid estimate of the problem.

The Salem site 1s substantially identical to many others in that one may anticipate a very stable dispersion situation existing with a 1 m/sec wind speed from almost any wind direction at a time of an accident. Such a situation might maintain for an hour or two. but since the site is subjected to brisk winds under any stability condition, one would anti-cipate that a 2 m/sec wind would be a minimum. estimate for the succeeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, even if stable conditions continued. For the remainder of the JO days following the postulated accident, one must review the accumu-

• lated meteorological data for reasonable probabilities, and essentially three different combinations appear.

The month of June, 1969, is clearly the least favorable of the entire set of observed data with the following combination of conditions:

Wind Direction:

Class I Frequency:

Class II Frequency:

Class III Frequency:

Class IV. Frequency:

This wind direction. however, carries the effluent completely over water for several miles, and cannot be considered limiting except in terms of producing the highest potential value at relatively great dis-tances. Discarding this case as unrealistic, two other candidates appear in the record, both of which

• A.1-29 Amendment 2

have onshore winds. The first is December, 1969, in which over 15% of the Class II cases came from 300° accompanied by 1% of the Class IV cases. The other situation was found in AUgust, 1969, where the Class IV condition had a high frequency of occurrence (3.1%),

but Class II occurred in only 1.5% of the hours from this same direction.

These oases have been analyzed, and 1t 1s appar-ent that the dominance or the Class IV situation is such that one must postulate the least favorable case as an extension of the latter situation. Accordingly we estimate that the following extreme set of conditions represents the least favorable.

Wind Direction: 240° 0-2 hours: Class IV, 1 m/sec wind 2-24 hours: Class IV, 2 m/sec wind

• '

1-5 days: Class II, 4 m/sec, 5%

Class III, 5 m/sec, 2%

Class IV, 2.5 m/sec, 6%

6-JO days: Class II, 4 m/sec, J%

Class III, 5 m/sec, 1%

Class IV, 2.5 m/sec, 4%

This computation is in our judgment a conceivable sequenoe of conditions, and it has been translated into X/Q values in the attached table.

L A.1-30 Amendment 2

LEAST FAVORABLE DIFFUSION SEQUENCE 0-30 DAYS (X/Q)

D1stance(km} 0-2 Hrs J-24 Hrs 1-,:2 D&s 6-~o Dazs 1.5 4.oxio-4 2.ox10-4 3.2x10-6 i.9x10- 6 2.5 l.9Xl*o- 4 9.3x10- 5 i.2x10- 6 7.2x10-7 5.0 7.ox10- 5 3.5x10-5 3.7x10-7 2.1x10-7 10.0 2.9x10-5 i.5x10-5 i.2x10-7 6.7x10-B

A.1-31 Amendment 2

• .II. ANNUAL CONCENTRATION PATTERNS.

The data summarized 1n the meteorological sec-t ion of the FSAR are for the period extending from June, 1969, through November 1970, but for the purpose of estimating the annual distribution of concentrations, it is preferable to select a single year to prevent undue weighting of the data. Consequently, the patterns have been developed from the records as they were sum-marized from June, 1969, to May, 1970.

The computer summarization includes detailed es-timates of the frequency of the three major turbulence classes (Class I is included with Class II for these purposes) in each io 0 sector around the site, X/Q values have been developed from them. Presentation of the material is made in tabular form, listing the concentra-tions at four distances, and in graphical form as well *

• A.1-32 Amendment 2

ANNUAL &Q VALUES cx10- 8 )

(Based on 6/69-5/70)

Distance(km)

Bear1ng~ 0 l 1.,2 2.5 s.o 10.0 10 86 34 11 4 20 64 25 8 3 JO 91 36 11 4 40 55 21 7 2 50 44 17 5 2 60 64 26 8 3 70 73 29 9 3 80 69 27 8 3 90 81 32 10 J 100 66 25 B J 110 62 25  ? 2 120 95 34 10 3 lJO 66 25 8 2 1.l+o 70 27 8 .- J 150 49 19 6 2 160 38 15 4 l 170 26 10 J 1 180 113 45 14 5 190 74 29 9 3 200 92 36 11 4 210 74 29 9 3 220 ?0 27 9 J 230 30 12 4 1 240 35 15 4 1 250 25 10 3 1 260 13 5 2 1 270 67 26 8 3 280 57 22 7 2 290 61 2.5 8 2 300 112 44 14 5 310 131 51 15 5 320 76 JO 9 3 330 82 32 10 3 340 72 28 9 3 350 67 26 8 J 360 107 42 13 4

• A.1-33 Amendment 2

ANNUAL X/Q

(% io-8) 350 360

~~\H*-*~LJ~c oJ

• A.1-34 Amendment 2

ADDENDUM 2 TO METEOROLOGICAL SECTION The following additio~al information has been obtained since the meteorological section was compiled. A summary of pertinent data collected over the two year period from June, 1969 to May, 1971 is included (the meteorological section is based on data collected from June, 1969 to November, 1970) *

• A.1-35 Amendment 2

PERCENTAGE FREQUENCY OF OCCURRENCE OF WIND SPEED CLASSES BY LARGE RATE GROUPS JUNE 2 1969 - MAY 2 1971 Wind SEeed {MPH}

Type Rate 30 Ft

{oc/lOOM} Calm 1-3 4-7 8-12 13-18 19+

-1.9 .4 1. 7 1.4 .6 .2

-1.89 to -1. 7 .2 .8 .8 .5 .1

-1.69 to -1.5 .4 1.0 1.3 .6 .3

-1.49 to -0.5 4.8 15.9 16.5 7.1 1.9

-0.40 to +1. 5 5.5 15.0 9.0 1.8 .4 .!

+1.5 to +4.0 1.9 4.6 1.9 .2 .o

+4.0 to +6.0 .5 1.4 .4 .1 .o

+6.0 .1 .2 *. 1 .o .o

.>

'""'

I 300 Ft w

°'

-1.9 .2 .5 1.4 1.3 1.5

-1.89 to -1. 7 .1 .4 .7 .8 .1

-1.69 tb -1. 5 .3 .7 1.1 1.0 1.0

-1.49 to -0.5 2.7 7.8 13.0 15.1 13.2

-0.49 to +1. 5 1.6 4.7 10.0 11.9 7.6

+1.5 to +4.0 .o .1 ' .2 .1 .1 5"ro +4.0 to +6.0 .o .1 .1 .1 .o tj 0.

El

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TURBULENCE CIASS 30 FT 150 P'T 300 FT I s.o 6.0 6.0 II s.o 13.0 14.0 III 10.0 16.0 19.0 IV s.o 9.0 12.0 ALL HOURS 1.0 11.0 14.0 A.1-37 Amendment 2

PERCENTAGE FREQUENCY OF WIND SPEED CLASSES

-30 FT WIND SPEED (MPH)

TURB.

CLASS CALM ~ 7  !::ll 13-18 lli I 2.e 4.0 1.7 .2 .o II 4.9 22.0 19.6 7.8 2.1 III IV

.3 s.e 3.4 11.2 S.6 4.3 2.4

.7

• '.1 ALL .3 13.8 40.6 31.3 11.1 2.9 300 PT WIND SPEED (MPH)

I 2.2 3.8 2.0 *' .6 .2 II 1.2 6.5 16.9 18.3 14.0 III o.o .o .e 4.9 6.8 IV 1.3 3.7 1.0 6.5 3.1 ALL .l 4.7 14.1 26. 7 30.3 24.2 l

A.1-38 Amendment 2

STATION SALEM HEIGHT 300 FEET PERIOD JUN 1969 Ma?: 1971

• 350 360 10 ALL HOURS INCLUDED 290 280 80 90 "ti 100 260 250 II 0 240 120 230 130 220 140 210 150 200 160 190 180 170 PERCENT OF ALL HOURS

• A.1-39 Amendment 2

STATION . SALEM HEIGHT 300 FEET PERIOD JUN 1969 - May 1971

• STABLE HOURS ONLY 350 360 10 330 320 290 280 80 100 260 250 110 240 120 230 130 140 220 210 150 200 170 160 190 180 PERCENT OF ALL HOURS L

A.1-40 Amendment 2

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RELATIONSHIP CHART COMMUNICATIONS -

EOF - OPERATIONAL -

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RELATIONSHIP CHAP.T COMMUNIATION -

EOF - RAD ASSESSMENT -

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CHART COMMUNICATIONS -

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• ACTIVATED I* PER UNIT BASIS

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..

• o PS~G Public Service Electric and Gas Company 80 Park Plaza, T16D Newark, N.J. 07101 201/430.8217 Robert L. Mittl General Manager* Licensing and Environment April 2, 1981 Director of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, D. c. 20555 Attention: Mr. F. J. Miraglia, Chief Licensing Branch 3 Division of Licensing Gentlemen:

COMPLIANCE WITH REGULATORY GUIDE 1.97 NO. 2 UNIT SALEM NUCLEAR GENERATING STATION DOCKET NO. 50-311 Public Service Electric and Gas hereby submits, in the enclosure to this letter, its initial evaluation of Salem 2 to determine the degree of compliance with Regulatory Guide 1.97, along with a schedule for bringing the unit into compliance.

Should you have any questions, do not hesitate to contact us.

Very truly yours, Enclosure ATTACHMENT 8.1 The .Energy People

SALEM NUCLEAR GENERATING STATION UNIT #2 COMPLIANCE WITH USNRC REGULATORY GUIDE 1.97 "INSTRUMENTATION FOR LIGHT WATER-COOLED NUCLEAR POWER PLANTS TO ASSESS PLANT AND ENVIRONS CONDITIONS DURING AND FOLLOWING AN ACCIDENT" INTRODUCTION PSE&G has completed an initial evaluation of the Salem Unit #2 instrumentation systems to determine its degree of compliance with Reg. Guide 1.97. Since the Salem design bases for instrument systems were developed and approved by the NRC significantly prior to the issuance of this guide, the evaluation was based on compliance with the overall intent of the guide.

Demonstration of compliance with the intent of the Reg. Guide required that any specific differences between the Salem design bases and those of the Reg. Guide be identified. The identification of these differences are

• specified in Section I

• The key elements of the overall evaluation can be summarized as follows:

o Compliance of existing systems and instrumentation is based upon meeting the intent of the Reg.

Guide.

o Compliance of new equipment is based upon appli-cation of the Reg. Guide to the extent that exist-ing design can accommodate the change without com-promising the existing system.

o Previous commitments to modify existing equipment or to add new equipment were considered (e.g.

NUREG-0588 and NUREG-0737) in the context of those commitments which pre-date Reg. Guide 1.97.

The results of this evaluation have been classified into five basic types of "compliance levels". These com-pliance levels have been selected to illustrate the resolu-tion actions planned for the equipment to demonstrate the overall plant compliance with Re~. Guide 1.97. This infor-mation is included in Sections II and III.

MP Bl 101 03/1

• I*

Identification of Design Basis Differences From Reg.

Guide 1.97.

To establish a baseline set of criteria for this eval-uation, the Reg. Guide 1.97 recommendations have been reviewed for similarity to the Salem plant design bases. In those instances where the Salem bases agree with the Reg. Guide 1.97, no differences are listed below. For those cases which involve differences, a comparison is provided below to demonstrate that the intent of the guide is adequately achieved.

1. The Salem plant design bases were effectively established prior to issuance of the Regulatory Guides referenced in Reg. Guide 1.97. Although the Salem plant conforms to the intent of the Reg-ulatory Guides, as stated in the Salem FSAR, strict compliance has not been required. In many cases, the Guides have been revised to incorporate subsequent revisions of referenced standards: and in some cases, the Guides are not applicable to the previously approved design (e.g., Reg. Guide

1. 75).

For the purpose of compliance with Reg. Guide 1.97, the Salem design will conform with the in-tent of the referenced Guides and Standards to the same extent as specified in previous responses to the NRC on the subject documents.

2. Reg. Guide 1.89 - "Qualification of Class IE Equipment for Nuclear Power Plant". The Salem plant review basis is NUREG-0588, Category II for existing instrumentation and NUREG-0588, Category I (i.e., IEEE 323-74) for new equipment. Evalua-tions for equipmen't in harsh environments have been completed. Evaluations for noncontrolled be-nign environments will be completed per NRC estab-lished schedules.

Recorders, indicators, and other instrumentation located in controlled benign environments such a~

the control room, have been considered as meeting M P81 101 03/2

the intent of Reg. Guide 1.97, pending the comple-tion of the NUREG 0588 benign environment review.

3. Reg. Guide 1.100 - "Seismic Qualification of Elec-tric Equipment for Nuclear Power Plants". The Salem plant review basis is IEEE 344-71, for exis-ting equipment and IEEE 344-74, for new equipment.

4. Reg. Guide 1.75- "Physical Independence of Elec-tric Systems". The Salem plant electric systems do not conform to the recommendations in Reg.

Guide 1.75, since this was not an original design criterion. New equipment will be integrated into our existing separation provisions. The Salem separation criteria has been approved by the NRC staff as described in Safety Evaluation Report, Supplement No. 4, Section 8.4.5.

5. Reg. Guide 1.32- "Criteria for Safety-Related Electric Power Systems for Nuclear Power Plants.

The Salem plant review basis is IEEE 308-71, "Class IE Electric Systems for Nuclear Power Gene-rating Stations".

6. Quality Assurance: Regulatory Guides

a. Reg. Guide 1.28, "Quality Assurance Program Requirements (Design and Construction)" Revi-sion 2, dated February, 1979. The Salem plant review basis is Safety Guide 28, which en-dorses ANSI 45.2.1 of 1971.

b. Reg. Guide 1.38, Revision 2, dated May, 1977, "Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage, and Handling of Items for Water-Cooled Nuclear Power Plants".

The Salem plant review basis is Reg. Guide 1.38, Revision 1, dated October, 1976.

c. Reg. Guide 1.64, dated June, 1976,. "Quality Assurance Requirements for the Design of Nu-clear Power Plants". The Salem plant review basis is Reg. Guide 1.64, Revision O, dated October, 1973 *

• M P81 101 03/3

d. Reg. Guide 1.123, dated July 7, 1977, "Quality Assurance Requirements for Control of Procure-ment of Items and Services for Nuclear Power Plants". The Salem plant review basis is ANSI 45.2.13, of 1976.

e. Reg. Guide 1.144, Revision 1, dated September, 1980 "Auditing of Quality Assurance Programs for Nuclear Power Plants". The Salem plant review basis is ANSI 45.2.12, Draft 4, Revision 2.

f. Reg. Guide 1.146, dated August 1980, "Qualifi-cation of Quality Assurance Program Audit Per-sonnel for Nuclear Power Plants". The Salem plant review basis does not include a commit-ment to Reg. Guide 1.146.

7. Unique Identification: The instruments are not specifically identified on the control panels as those intended for use under accident conditions.

The instrumentation on the control panels in the Salem Control Room is presently grouped on a functional basis. Additional markings could add confusion to a control panel layout that was favorably reviewed during the NRC "Human Factors Review of the Salem No. 2 Unit Control Room" in March of 1980.

8. Reg. Guide 1.118- "Periodic Testing of Electric Power and Protection Systems". The Regulatory Guide invokes the requirements of IEEE 338-1975, which is applicable to protection systems. The display information provided for Reg. Guide 1.97, is not considered to be part of the protection system and does not require all of the testing specified in IEEE-338. The plant equipment being used for compliance with Reg. Guide 1.97 has been designed to incorporate testing capabilities as discussed in the Salem FSAR, Chapter 7~2. Testing frequencies will be in accordance with the appli-cable Technical Specifications.

M P81 101 03/4

-s-

• 9. TYPE "A" VARIABLES (PLANT SPECIFIC)

The definition of Type *A" Variables given by Reg. Guide 1.97, Paragraph 1.1 is:

"Those variables to be monitored that provide the primary information required to permit the control room operators to take the specified manually con-trolled actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for design basis events."

The Salem Emergency Operating Procedures were re-viewed to determine which Reg. Guide 1.97, Type "A" parameters are required using the following baseline interpretation of Reg. Guide 1.97, Paragraph 1.1

• The operating procedures specify certain operator verification of automatic actions, and if the automatic actions have not been performed (presum-ably due to system failure), the operator is re-quired to manually perform those actions. The parameter selection does not include either the verification step or the manual backup action.

The parameter selection includes those required for operator actions needed for system functioning where no automatic signal/system exists.

Where important, the manual Type "A" operator ac-tions must be monitored to assure that the action has been performed. This monitoring of Type "A" operation is performed by "Type B" Variables as defined by Reg. Guide 1.97.

In reviewing the Emergency Operating Procedures, the event "end point" for parameter selection is a stable hot condition for all events except LOCA's (large or small) that cannot be isolat~d. The end point for LOCA's that cannot be isolated is a cold depressurized condition.

See Table I.l for an index of Class "A" Variables *

• M PBl 101 03/5

II. REG. GUIDE 1.97 COMPLIANCE LEVELS The evaluation revealed varying degrees of compliance with Reg. Guide 1.97 which were classified into five "com-pliance levels". These classifications evolved from con-sideration of the design bases, existing NRC commitments and specific new changes, where possible, to meet the Reg.

Guide. The overall results of this effort are summarized in Table II.l.

The description of each "compliance level" is provided below:

1. Items in Compliance The items categorized under this heading meet the PSE&G bases as outlined in Section I.

2. I~ms Where Design Precludes Compliance The items categorized under this heading are pres-ently installed, but* by nature of the present de-sign, may not meet the recommendations in the Guide such as environmental and seismic qualifica-tions. These items are generic to Westinghouse plants.

3. Items Which Are Being Replaced/Added The iteMs categorized under this heading deviate from one or more recommendations in the Guide.

These items will be replaced with devices modified to meet the appropriate recommendations. Re-placement/Installation/Modification for each item will be made to:

a. Meet requirements imposed by other regulations such as NUREG-0588 and NUREG-0737.

b. Meet the recommendations of Regula~ory Guide 1.97.

M P81 101 03/6

4. Items Which Are Not Being Replaced The items categorized under this heading are pres-ently installed, but deviate from one or more rec-ommendations of the Guide.

This is based on:

a. Devices located in non-harsh environment, that require qualification review in accordance with NUREG-0588 for benign environments which will be completed by June 30, 1982.

b. Devices located in a harsh environment that are not utilized in accident emergency in-structions for operators to maintain plant safety.

c. Devices currently meeting Tech. Spec. require-ments but the specified ranges do not meet the recommendations in Reg. Guide 1.97.

5. Items Not Part of Salem Design The items categorized under this heading are not part of the Salem design and are not being instal-led. Alternate capabilities are available which meet or will meet our requirements and provide ad-equate information for maintenance of plant safety.

III. PLANNED ACTIONS Compliance Level 1 No action planned. The instrumentation in this com-pliance level meets the intent of Reg. Guide 1.97 in accordance with the criteria specified in Section I.

Compliance Level 2 No equipment replacement planned at this time pending resolution of generic problems *

• M P81 101 03/7

,

Compliance Level 3a Instruments are being replaced or upgraded as a result of prior commitments related to NUREG-0588 and NUREG 0737. These devices will comply with Reg. Guide 1.97 in accordance with the criteria specified in Section I by the dates specified in previous correspondence to the NRC staff.

Compliance Level 3b Instruments will be upgraded to meet Reg. Guide 1.97 in accordance with the criteria specified in Section I by 6/1/83.

Compliance Level 4a The equipment will be evaluated in accordance with the requirements of the NUREG-0588 benign environment review, and appropriate actions will be taken where required.

Compliance Level 4b No action planned. The importance of the device is deemed to be relatively low or insignificant.

Compliance Level 4c The existing devices comply with Tech. Spec. require-ments and should not be modified.

Compliance Level 5 No action planned. Other provisions exist which ne-gate the need for the instrumentation.

All items currently planned to remain unchanged have been evaluated for potential effects on plant safety.

This evaluation concludes that plant safety is not af-fected by the lack of compliance to Reg. Guide 1.97 *

• M P81 101 03/8

TABLE I.1 INDEX TYPE "A" VARIABLES Variable Description Variable Reference No.

Reactor Coolant System Hot Leg 5 Water Temperature Reactor Coolant Pressure 6 Degrees of Subcooling 9 Containment Pressure 11 Effluent Radioactivity 16 Noble Gas Effluent from Condenser Air Removal System Exhaust Refueling Water Storage Tank Level 27 Pressurizer Level 30 Stearn Generator Pressure 36 Auxiliary Feedwater Flow 39 Auxiliary Feedwater Storage Tank Level 40 (Condensate Storage Tank)

Stearn Generator Radiation 73

• M P81 101 03/9

Table II.l Summary of Instrumentation Compliance with Reg. Guide 1.97.

Variable Variable Compliance Ref. No. Description Level 1 Neutron Flux 2 (Source Range, Intermediate range, Power range) - Monitors 2 Control Rod Position 1 3 RCS Soluble Boron Concentration 1 4 RCS Cold Leg Water Temperature

- RTD's 3a

- Indication 3b

• 5 RCS Hot Leg Water Temperature

- RTD's 3a

- Indication 3b 6 RCS Pressure - Transmitters 3a 7 Core Exit Temperature - Thermocouples 2 8 Coolant Level in Reactor 1 9 Degrees of Subcooling - Display 1 (inputs - See variable Ref. No. 6&7) 10 Containment Sump Water Level - 3a Transmitters 11 Containment Pressure (narrow and wide range) - Transmitters 3a 12 Containment Isolation Valve Position (excluding check valves) - Limit Switches 3a 13 Radioactivity Concentration or Radiation 5 Level in Circulating Primary Coolant M P81 101 03/10

Table II.1 (Cont'd)

Variable Variable Compliance Ref. No. Description Level 14 Analysis of Primary Coolant (Gamma Spectrum) 5 15 Containment Area Radiation - Monitors 3a 16 Effluent Radioactivity - Noble Gas Effluent from Condenser Air Removal System Exhaust - Monitors 3a 17 Containment Hydrogen Concentration - 3a Analyzers 18 Containment Effluent Radioactivity Nobles Gases from Identified Release Points - Monitors 3b 19 Radiation Exposure Rate (Electrical Penetration Area) - Monitor 3b 19A Radiation Exposure Rate (Fuel Handling Building & Penetration Area) - Monitors 4a 20 RHR System Flow - Transmitters 3a 21 RHR Heat Exchanger Outlet Temperature -

Thermocouples 4b 22 Accumulator Tank Level and Pressure -

- Transmitters 3b

- Transmitter Range 4c 23 Accumulator Isolation Valve Position 1 24 Boric Acid Charging Flow - Transmitters 3a 25 Flow in HPI System - Transmitters 3a 26 Flow in LPI System - Transmitters 3a M PB! 101 03/11

Table II.l (Cont'd)

Variable Variable Compliance Ref. No. Description Level 27 Refueling Water Storage Tank Level and Low Level Alarm - Transmitters 3a

- Transmitter Range 4c 28 Reactor Coolant Pump Status l 29 Primary System Safety Relief Valve l Positions (including PORV and code valves) or Flow through or Pressure in Relief Valve Lines 30 Pressurizer Level - Transmitters 3a Transmitter Range 4c 31 Pressurizer Heater Status (Current) - Heaters 2 32 Quench Tank Level (Pressurizer Relief Tank) - Transrniter 4c Range 33 Quench Tank Temperature (Pressurizer Relief Tank) - Transmitter 3b Range 34 Quench Tank Pressure (Pressurizer Relief Tank) - l 35 Stearn Generator Level - Transmitters 3a 36 Stearn Generator Pressure - Transmitters 3a 37 Safety/Relief Valve Positions or Main Stearn Flow - Transmitters 3a 38 Main Feedwater Flow 1 39 Auxiliary Feedwater Flow - Transmitters 3a M P81 101 03/12

• Table II.l (Cont'd)

Variable Variable Compliance Ref. No. Description Level 40 Condensate Storage Tank Water Level (Auxiliary Feedwater Storage Tank)

- Transmitters 3a 41 Containment Spray Flow 5 41A Containment Spray Flow Additive Rate

- Transmitters 3a 42 Heat Removal by the Containment Fan Heat Removal System - Transmitters 3a 43 Containment Atmosphere Temperature 4b 44 Containment Sump Water Temperature 5 45 CVCS Makeup Flow-in - Transmitters 3a 46 Letdown Flow - Transmitters 3b 47 Volume Control Tank Level - Transmitters 4a 48 Component Cooling Water Temperature to ESF System - Transmitters 4a 49 Component Cooling Water Flow to ESF System - Transmitters 4a 50 High-Level Radioactive Liquid Tank Level - Indication 3b 51 Radioactive Gas Holdup Tank Pressure-Indication 3b 52 Emergency Ventilation Damper Position -

Control Room Damper Limit Switches 4a 52A Emergency Ventilation Damper Position -

Ausiliary Bldg. Damper Limit Switches 4a

• M PBl 101 03/13

• Table II.l (Cont'd)

Variable Variable Compliance Ref. No. Description ~L~e~v~e~l~~-

52B Emergency Ventilation Damper Position -

Fuel Handling Bldg. 4a 53 Status of Standby Power 1 53A Status of Control Air 5 54 Containment of Purge Effluent N/A 55 Reactor Shield Building Annulus Effluent N/A 56 Auxiliary Building Effluent N/A 57 Condenser Air Removal System Exhaust N/A 58 Common Plant Vent or Multi-Purpose Vent Discharging any of above releases

- Monitor 3a 59 Vent from Steam Generator Saftey Relief Valves or Atmospheric Dump Valves -

Monitors 3a 60A All other identified Release Points (Decontamination Bldg.) - Monitor 3b 60B All other identified Release Points (Auxiliary Feed Pump Turbine Exhaust) -

Monitor 3a 61 All Identified Plant Release Points -

Monitor (Particulates and Alogens) 3a 62 Radiation Exposure Meters 5 63 Airborne Radioalogens and particulates (portable sampling with onsite analysis capability) - 1

• M P81 101 03/14

• Table II.1 (Cont'd)

.

Variable Variable Compliance Ref. No. Description Level 64 Plant and Environs Raidation -

Instrument Range 4c 65 Plant and Environs Radioactivity 1 66 Wind Direction 1 67 Wind Speed 1 68 Estimation of Atmospheric Stability 1 69 Primary Coolant (Grab sample) 1 70 Containment Air (Grab sample) 1 71 Containment Sump (Grab sample) 1 72 Effluent Radioactivity - Nobel Gases -

Monitor 3a 73 Stearn Generator Blow-down Radiation -

Monitor 3a

• M P81 101 03/15