ML20027C314
| ML20027C314 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 10/12/1982 |
| From: | Notaro J, Poltrino R LONG ISLAND LIGHTING CO., STONE & WEBSTER, INC. |
| To: | |
| Shared Package | |
| ML20027C225 | List: |
| References | |
| ISSUANCES-OL, NUDOCS 8210150281 | |
| Download: ML20027C314 (65) | |
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LILCO, October 12, 1982
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety a'nd Licensing Board In the Matter of ) -;
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LONG ISLAND LIGHTING COMPANY ) Docket No. 50-322 (OL)
) (Emergency Planning--
(Shoreham Nuclear Power Station, ) Phase I)
Unit 1) )
TESTIMONY OF JACK'A. NOTARO AND ROBERT L. POLTRINO FOR THE LONG ISLAND LIGHTING COMPANY ON PHASE I EMERGENCY PLANNING CONTENTION 13 --
INTERIM SAFETY PARAMETER DISPLAY SYSTEM PURPOSE
_ This testimony demonstrates that LILCO has designed a Safety Parameter Display System (SPDS) for Shoreham that will include the recommended functional capabilities of NUREG-0696 as modified by SECY 82-111. The interim SPDS is designed for use until the permanent SPDS is available.
The interim.SPDS provides for validation of data by pro-4 Viding redundant displays of most parameters, and has'a trend-ing capability for both radiological and non-radiological pa- .
rameters. All.SPDS displays and parameters (radiological and non-radiological) are available in the TSC. All radiological r displays and-radiological parameters are available in the EOF; ;
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EOF personnel may communicatie directly with the TSC for any additional technical parameters. The interim SPDS will provide graphic displays in the control room, to augment the informa-tion available to the operator for determining if 4. transient or accident has occurred, and to aid the operator in tracking the course of an accident.
ATTACHMENTS Attachment 13-1 Professional Qualifications of Jack A. Notaro Attachment 13-2 Professional Qualifications.of Robert L. Poltrino Attachment 13-3 SECY 82-111, "NRC Staff Recommendations on the Regairements for Emergency Response Capability" Attachment 13-4 Interim SPDS Parameters W-
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LILCO,-October 12, 1982 UNITED STATES OF AMEliICA l NUCLEAR REGULATORY COMMISSION l
! Before the Atomic Safety and Licensing Board In the Matter of )
) .
LONG ISLAND LIGHTING COMPANY ) Docket No. 50-322.(OL)
) (Emergency Planning---
(Shoreham Nuclear Power Station, ) Phase I)
Unit 1) )
TESTIMONY OF JACK A. NOTARO AND ROBERT L. POLTRINO FOR THE LONG ISLAND LIGHTING COMPANY ON PHASE I EMERGENCY PLANNING CONTENTION 13 --
INTERIM SAFETY PARAMETER DISPLAY SYSTEM Q1. Please state your names and business addresses.
A1. *My name is Jack A. Notaro; my business address is Shoreham Nuclear Power Station, P. O. Box 628, Wading River, New York, 11792.
My name is Robert L. Poltrino; my business address is Stone and Webster Engineering Corporation, 245 Summer '
Street, Boston, Massachusetts, 02110.
Q2 . . What are your present positions with your companies?
A2. [Notaro] I am the Operating Engineer at the Shoreham Nuclear Power Station for LILCO.
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[Poltrino] I am the principal electrical control engin-eer for Stone & Webster assigned.to the Shoreham pro-ject.
Q3. Please state your professional qualifications.
A3. {Notaro] My professional qualifications are attached to this testimony (Attachment 13-1). My familiarIity with the issues surrounding the interim SPDS stems from my daily responsibility for safe, reliable Plant Operations. This encompasses all integrated system op-erations, including use of the interim SPDS.
[Poltrino] My professional qualifications are attached to this testimony (Attachment 13-2). My familiarity with the issues surrounding the interim SPDS stems from my involvement with the design and design review of the Phase I Emergency Response Facilities (ERF), including the interim SPDS. Prior to my current assignment, I was an engineer in the electrical control group on Shoreham.
I have been involved with Shoreham for over eight years.
[Both witnesses sponsor the remaining answers in this testimony.]
Q4. Are you familiar with Suffolk County Contention EP 13?
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A4. Yes.
t What does that contention say?
QS.
AS. The contention reads as follows:
Suffolk County contends that the interim SPDS that LILCO proposes to utilize until the installation of a permanent SPDS is ,.
deficient because it does not meet mini-mum requirements for such a system.
Specifically, the interim SPDS does not:
A. provide all required parameters
[NUREG-0696 at 26];
B. provide for data verification
[NUREG-0696 at 24];
l C. provide trending capability
[NUREG-0696 at 25-26];
D. provide information to the TSC and EOF [NUREG-0696 at 25]; and E. provide the function of aiding the operator in the interpretation of transients and accidents, nor does it provide this function during and fol-lowing all events expected to occur during the life of the plant, includ-ing earthquakes [NUREG-0696 at 27].
Thus, the interim SPDS does not meet the requirements of 10 CFR 55 50.47(b)(4)(8),
and (9), 10 CFR Part 50, Appendix E, ,
Items IV.E.2 and 8, 10 CFR Part 50, Appendix A, CDC 13, and NUREGs-0696, 0737 and 0654, Item I.
I Q6. What is the thrust of the contention?
A6. Generally, Suffolk County contends that LILCO's interim
[ SPDS does not meet the guidelines of NUREG-0696 for such I
f a system.
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- f. Q7.- Are you familiar with the regulatory requirements and guidelines cited in EP 13? ,
A7. Yes.
Q8. Do 10 C.F.R. $$ 50.47(B)(4)/(8) and (9), 10 C.F.R. Part 50, Appendix E, Items IV.E.2 and 8, 10 C.F.R. Part 50,-
Appendix'A, GDC 13, and NUREG-0654, Item I establish re-quirements for SPDS?
A8. No.
Q9. What are the current regulatory requirements and guide-lines for the installation of a safety parameter display.
system at Shoreham, and how has LILCO implamented those
, requirements and guidelines?
A9. SECY 82-111, "NRC Staff Recommendations on the Requirements for Emergency Response Capability" (Attachment 13-3), outlines the current guidelines for SPDS. The original technical requirements and implemen-tation schedules for SPDS evolved through such NRC docu-ments as NUREG-0578, NUREG-0737 and NUREG-0696. -
Recognizing at the time these guidelines were issued that procurement and design of a new computer system fully meeting NUREG-0696 guidelines could not be achie-ved at Shoreham by fuel load, due to the late emergence of the definitive information provided in NUREG-0696,
LILCO chose to implement the SPDS for Shoreham in two
. phases. Phase I provides for an , interim SPDS, using the existing plant process computer; Phase II, a permanent SPDS, using a new computer base system. The interim l SPDS would be used only unti1 the permanent SPDS becomes available.
Prior to the issuance by the NRC of SECY 82-111, LILCO had obtained Staff concurrence to use an interim SPDS until at least April, 1983. With the revised criteria l and schedule guidelines set forth in SECY 82-111, two of the specific " requirements" listed by the County in its contention are no longer suggested by the NRC: (1) that SPDS be seismically qualified, and (2) that licensees meet certain schedule commitments that'had been sugges-ted. Implementation dates for SPDS are now to be deter-mined by discussions between the individual licensee and the NRC project manager.
Q10. Please describe the Phase I interim SPDS.
A10. The interim SPDS is based on (1) the existing plant pro-cess computer, (2) the Radiation Monitoring System (RMS) computers, and (3) a portion of the permanent data ac-quisition system. This additional diagnostic informa-tion will be available in a concentrated location to the station operators and technical and management level personnel.
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Qll. How does Phase II differ from Phase I?
L Phase II is based on a new multi-computer system consi-All.
l sting of four independent computers, two of which are seismically qualified. Any one of the four computers is capable of providing SPDS displays.
Q12. What " parameters" are suggested in NUREG-0696 Ior SPDS displays,;as referred to by the County in EP 13(A)?
A12. NUREG-0696 does not suggest any specific " parameters" to be incorporated in the SPDS. NUREG-0696 does, however, state:
The important plant functions related to the pri-mary display while the plant is generating power shall include, but not be limited to:
- a. Reactivity control
- b. Rea-tor core cooling and heat removal from primary system
- c. Reactor coolant system integrity
- d. Radioactivity control
- e. Containment integrity (Emphasis added.)
Q13. Do the interim SPDS displays for Shoreham monitor all the important plant functions suggested in NUREG-0696?
A13. Yes.
e Q14. What functions are available from the interim SPDS?
A14. The interim SPDS monitors the status of the following functions:
! a. Reactivity control l b. Reactor core coolant / heat removal + i j c. Reactor coolant system integrity l d. Containment integrity / heat removal *
- e. Radioactivity control ,.
Functions (a) through (d) do have parameters that are displayed on a color CRT located on panel lH11*PNL-603, directly in front of tb.eygperator's desk in the main control room.
The Radiation Monitoring System (RMS) providing the radioactivity control information listed in (e), above, is a computer-based radiological information system that provides the status of radiation levels throughout the plant. Data from individual radiation monitors is inputted to the computer system to provide continuous real time radiological status for the plant. The compu-ter system also provides trending capability and auto-matic report generation. Set points are provided for each radiation monitor; these set points will alarm to '
-alert plant personnel of unusual conditions. The RMS system provides the station operators with parameters that monitor the status of radioactivity control.
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Thus, the interim SPDS for Shoreham addresses each of the five categories referenced in NUREG-0696.
Q15. What specific parameters does.the interim SPDS monitor?
A15. For a complete list of the specific parameters monitored by Shoreham's interim SPDS, see Attachment 13-4, attached to this testimony.
Q16. In response to EP 13(B), does the interim SPOS provide for data verification?
A16. Yes. Data validation is the process by which the accu-racy of the data being displayed is determined. Many of the interim SPDS parameters displayed or available for display have at least two redundant channels for every parameter used for the interim SPDS. Therefore, the op-erator can readily validate data presented by.ths interim SPDS by comparing redundant parameter channels.
The operator would be aware of any discrepancy between redundant channels, and could refer to other SPDS param-eters or board-mounted instrumentation to resolve'any conflicting readings.
Q17. In response to EP 13(C),-doesthe interim SPDS provide a trending capability?
A17. Yes,.the interim SPDS has the capability.to provide C
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E analyses of trends. The RMS has historical logs and graphic trend displays, both of which can provide trend information for a radiological parameter or group of pa-rameters. These printouts and displays are available at the RMS printer and the CRT in the control room. For the remainder of the SPDS, the operator has available in the control room a two-pen trend recorder. Anhanalog point in the process computer data base can be assigned to this recorder to provide trending information. (This recorder is located on panel 1H11*PNL-602.) In addi-tion, there is available to the control room operator eight special logs to which the operator can assign any analog computer point in the process computer data base along with the desired scan period for historical log-
- ging and trend analysis.
Q18. In response to EP 13(D), does the interim SPDS provide information to the Technical Support Center (TSC) and the Emergency Operations Facility (EOF)?
A18. Yes. The SPDS parameters and, in fact, the entire pro- ~
cess computer data base, is available from printers lo-cated in the TSC. In addition, the CRT in the TSC may be utilized to display individual parameters from the process computer data base or to display any of the gra-phics available from the plant process computer. The
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[ entire contents of.the data base of the RMS -- both CRT graphics and hard copy printouts -- are available'at the TSC.
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Only radiological parameters are provided at the EOF.
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It is this radiological information which is of prime importance at the EOF. Any additional information of interest to the personnel in the EOF is available via direct communication to the TSC.
SECY 82-111 deleted the suggested guideline that SPDS displays be included in the TSC and EOF.
Q19. In response to EP 13(E), does the interim SPDS aid'the operator in the interpretation of transients and acci-dents?
A19. Yes. The SPDS displays provide the station operator.
with additional diagnostic information in a conce4trated location, a'iding the operator in tracking the course of-an accident.
Q20. Will the interim SPDS be available following all-events expected to occur during the life of the plant, includ-ing earthquakes?
A20. The plant process computer is powered from a seismically.
qualified' inverter, which, in turn, is fed from a
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s safety-grade diesel generator and battery. The process computer is located in a room with a redundant safety-grade ventilation system in a non-harsh environment.
With the exception of seismic events, the interim SPDS will function during any event expected to occur during the life of the plant. The plant process computer was not designed to be seismically qualified. SECN 82-111 has deleted all seismic provisions for the SPDS.
However, the permanent SPDS will have two seismically qualified CRT's in the control room, driven by two re-dundant seismically qualified computers.
Q21. Please summarize your testimony.
A21. LILCO has designed a Safety Parameter Display System
- (SPDS) for Shoreham that will include the recommended functional capabilities of NUREG-0696 as modified by SECY 82-111. The interim SPDS is designed for use until the permanent SPDS is available.
The interim SPDS provides for validation of data by pro-viding redundant displays of most parameters, and has a trending capability for both radiological and non-radiological parameters.
All SPDS displays and parameters (radiological and non-radiological) are available in the TSC. All
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I radiological displays and radiological parameters are available in the EOF; EOF personnel may communicate di-rectly with the TSC for any additional technical parame-ters. The interim SPDS will provide graphic displays in the control room, to augment the information available to the operator for determining if a transient or acci-
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dent has occurred, and to aid the operator in tracking the course of an accident.
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,. Attachmsnt 13 - 1 PROFESSIONAL QUALIFICATIONS JACK A. NOTARO Operating Engineer LONG ISLAND LIGHTING COMPANY My name is Jack A. Notaro. My business address is Long Island Lighting Company, Shoreham Nuclear Power Station, P.O.
Box 628, Wading River, New York, 11792. I am employed.by Long Island Lighting Company as an Operating Engineer and have held this position since July 1978. I am assigned as the Operating I Engineer at the Shoreham Nuclear Power Station. In this capac-ity, I am responsible for the development and implementation of the Station's operational activities including the direct. n of startup, day-to-day operation and shutdown of station equip-i ment. My responsibilities also include the implementation of I initial, requalification and replacement training programs for licensed and unlicensed operators, and the development, review and implementation of the operations section of the Station Operating Manual.
I received a Bachelor of Mechanical Engineering degree from City College of New York in 1970; and a Master of Business Administration from Adelphi University in 1974. In July 1976, I completed the General Electric Boiling Water Reactor Simulator Program and obtained certification as a Senior
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Reactor Operator. I have completed several industry seminars and training programs including: BWR Design Orientation, BWR Technology, BWR Observation Training, Nuclear Power Plant Technology, Radiation Protection, Basic Health Physics, l Vibration Analysis, Statics, Strength of Materials & Dynamics, Management of Maintenance Storekeeping & Inventories, QA for the Nuclear Industry, Inservice Inspection & QA During Operations, Basic Radiography, Magnetic Particle & Liquid Penetrant Testing, Basic Ultrasonics, Nuclear Power QA, Inservice Inspection Sy=posium, Operations Quality Assurance, Fire Fighting Training, Limerick Simulator Capability, and Simulator Refresher Training.
I joined Long Island Lighting Company in June 1970 and was assigned to the Maintenance Section at the Northport Power Station. In January 1972 I was transferred to the Electric Production Department and my assigned duties included maintenance scheduling, manpower allocation, equipment testing and station performance analysis.
In March 1973 I was assigned to the Quality Assurance Section at the Shoreham Nuclear Power Station. I was subse-quently promoted to Station Operating Quality Assurance Engineer in July 1974. My duties included initial development and implementation of the operational quality assurance pro-gram. This program included reviews, audits, surveillances, inspections, selection and training of personnel, development I
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of procedures and instructions, and the utilization of consultants and contractors. In addition, I was responsible for licensing and inspection activities associated with the U.S. Nuclear Regulatory Commission.
In August 1978 I was assigned to the Vermont Yankee Nuclear Power Station to observe startup of the unit following a refueling outage. I witnessed the completion of the ints-grated leak rate test, the primary systems hydrostatic pressure test, and the drywell inspection. I observed approach to crit-icality, criticality, plant heat-up and transfer to run. I also witnessed a half-scram recovery during plant heat-up.
From March 1981 to May 1981 I was assi'gned to the Operations Section of the Millstone Nuclear Power Station for the completion of the Unit 1 refueling outage. I participated in all significant pre- and post-refueling outage surveillance testing and inspections, and I actually took part in refuel bridge operations including control rod removal and replacement, channeled and dechanneled fuel movements, core inspections and verifications, dropped fuel bundle evaluetions, and recovery.
I was assigned to the Operations Section of the Millstone Nuclear Power Station in June 1981. I participated in routine BOP and NSSS system surveillance testing, and high risk I&C and operations equipment and system surveillance testing. I conducted heat balances, core flow calculations and
subsequent nuclear instrumentation calibrations with, and without, the main computer available. In addition, I witnessed implementation of emergency notification procedures, as well as manipulated controls for power downs, return to power, Tech Spec LCO's, control rod repositioning, and stuck control rod surveillance testing. I also participated in half scram and full scram recoveries and the subsequent investigations, eval-uations and notifications.
I am a member of the American Society for Quality Control and past member of the Edison Electric Institute -
Quality Arsurance Task Force (EEI-QATF) and the EEI-QATF Operations Subcommittee.
Attachment 13 - 2 PROFESSIONAL QUALIFICATIONS ROBERT L. POLTRINO Control Engineer, Control Systems Division STONE & WEBSTER ENGINEERING CORPORATION
(. My name is Robert Poltrino. My business address is 245 Summer Street, Boston, Massachusetts 02110. I am employed by Stone & Webster Engineering Corporation as the Principal Electrical Control Engineer on the Shoreham Project and have held this position since September 1980. In this capacity, I have overall responsbility for the design and layout of auxil-iary control and relay panels, preparation of the purchase spe-cifications for those panels, review and approval of nuclear steam supply system (NSSS) control panel layout drawings, and the preparation of elementary control diagrams.
I was awarded a Bachelor of Science degree in electri-cal engineering from Northeastern Univeristy. Since degree conferral, I have successfully completed all but ten credit hours towards the Master of Science degree in electrical engi-neering at Northeastern.
Prior to joining Stone & Webster, I served in the United States Army from July 1969 - July 1973. As a Project Officer in the U.S. Army Training Device Agency, I was
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responsible for preparation of conceptual system design and descriptions for computer-aided training systems. Prior to this, I was involved in the installation and management of high capacity _ microwave and tropospheric scatter communications sys-tems and associated telecommunications switching centers.
These systems were installed at some 20 different sites '
throughout Southeast Asia. I also monitored and evaluated con-tractor operations and performance overseas.
Upon joining Stone & Webster Engineering Corporation (November *1973) as an Engineer, I was assigned to the electri-cal control group en the Long Island Lighting Company (LILCO)
Shoreham Nuclear Power Station (SNPS) Unit 1. My activities included review of GE NSSS control board layout, review of ven-dor elementary drawings, preparation of purchase specifications for the auxiliary control and relay panels, design and layout of auxiliary control panels and benchboards, design and layout of start-up transient monitoring system interface, and the preparation of elementary controls diagrams.
Since September 1980 to the present, I have functioned as Principal Electrical Control Engineer on the Shorehe.
Project. Since then I have been scheduling and assigning >
tanks, and establishing completion dates within the electrical control group. I supervise assigned personnel, and review and
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I am a Registered Professional Engineer in the Commonwealth'of Massachusetts and the State of California.
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Attachment 13 - 3
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NRC STAFF RECOMMENDATIONS ON THE REQUIREMENTS FOR EMERGENCY RESPONSE CAPABILITY Parch 10,1982 .
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. . CONTENTS Pace .
.' 1. INTRODUCTION.................................................. ,
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- 2. USE OF EXISTING DOCUMENTATION................................. 3
- 3. COORDINATION AND INTEGRATION OF INITIATIVES................... 4
- 4. SAFETY PARAMETER DISPLAY SYSTEM (5P05)........................ 7
- Current Regulatory Requirements Functional Statement
- Recommended Requirements
, Integration Reference Documents
- 5. DETAILED CONTROL ROOM DESIGN REVIEW........................... 10 Current Regulatory Requirements Functional Statement
. 'f Recommended Requirements .
\ - Documentation and NRC Review
-- Integration
- Reference Documents .
- 6. REGULATORY GUIDE 1.97 - APPLICATION TO EMERGthCY RESPONSE -
FACILITIES.................................................... 13
' Current Regulatory Requirements Functional Statement Recommended Requirements .
Documentation and NRC Review
- 7. UPGRADE EMERGENCY OPERATING PROCEDURES (E0Ps)................. 15 Current Regulatory Requirements .
Functional Statement Recommended Requirements Documentation and NRC Review -
Reference Documents
- 8. EMERGENCY RESPONSE FACILITIES................................. 17 Current Regulatory Requirements . .
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Technical Support Center.................................... 19 Functional Statement Recommended Requirements -
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CONTENTS (Continued)
Operational Suppcrt Center.................................. . 21 Functional Statement Recommended Requirements
- Emergency Operations Facility.............................. 22 Functional Statement Recommended Requirements Documentation and NRC Review
- Reference Documents Table 1 - Emergency Operations Facility Location Options.......... 25 Table 2 - Minimum Staffing Requirements for NRC Licensees for Nuclear Power Plant Emer.gencies......................... 26
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- EMERGENCY RESPONSE CAPABILITY .
- 1. INTRODUCTION .
This report was prepared as a result of a review by the Committee to Review Generic Requirements (CRGR). The recommendations herein have been developed by the program offices and are supported by CRGR. The report represents- the staff's attempt to distill the fundamental requirements for nuclear plant Emergency Respolise Capability from the wide range of guidance documents that NRC has issued. It is not intended that these guidance docuinents (NUREG reports and Regulatory Guides) be ignored; they are still useful sources of guidance for licensees and NRC staff regarding acceptable means for meeting the fundamental requirements contained in this document.
These fundamental requirements are further specification of the general guidance specified previously by the Commission in its regulations, orders and policy statements on emergen:y planning and TMI issues. It is intended that these fundamental requirements would be applicable to
. licensees of operating nuclear power plants and holders of construction
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permits for nuclear power plants. For applicants for a construction
- permit ICP) or manufacturing license (ML), the requirements described in this ~ document must be supplemented with the specific provisior.s in the rule specifying licensing requirements for pending CP and ML applications.
In this regard, it is expected that the staff would review CP and ML 4 applications against the guidance in the current Standard Review Plan, -
and this might lead to more detailed requirements than prescribed in this document.
Based on discussions with licensees, the staff has learned that many of
- the Commission approved schedules for emergency response facilities
'; probably will not be met. In recognition of this fact and the difficulty of implementing generic deadlines, the staff proposes that plant-specific schedules be established which take into account the unique status of each plant. The following sequence for developing implementation schedules
- is proposed.
When the basic requirements for emergency response capabilities and
.-. facilities are finalized, they should be transmitted to licensees by a generic letter from NRR, promul regulatory requirements (e.g., in gated to NRC the Standard staff, Plan Review and or incorporated by regulationas .
> or Order, as appropriate). The letter to licensees should request that licensees submit a proposed schedule for completing actions to comply with the basic requirements. Each licensee's proposed schedules would then be reviewed by the assigned NRC Project Manager, who would discuss I
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( . the subject ~with the licensee and mutually agree on schedules and completion j i
dates. The implementation dates would then be formalized into an enforceable
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- The basic requirements in this document do not alter previously issued guidance, which remains in effect. This document does attempt to place that guidance in perspective by identifying the elements that the NRC j staff believes to be essential to upgraded emergency response capabilities.
l The proposal to formalize implementation dates in an enforceable document reflects the level of importance which the NRC staff attributes to these basic requirements. The NRC staff does not recommend that existing guidance be imposed in this manner, but rather that it be used as guidance to be considered in upgrading emergency response capabilities. This indicates the distinction which the staff believes should be made between the basic requirements and guidance.
The following sections describe NRC staff recommendations on basic re-quirements, their interrelationships, and NRC actions to improve manage-ment of emergency response regulation. Reference documents are cited with a description of content as it relates to specific initiatives.
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- 2. USE OF EXISTING DOCUMENTATION The NRC staff recommends that the follosing NUREG documents are intended to be .
. used as sources cf guidance and information, and the Regulatory Guides are to I be considered as guidance or as an acceptable approach to meeting formal requirements. The items by ' virtue of their inclusion in these documents shall not be misconstrued as requirements to be levied on licensees or as inflexible f criteria to be used by NRC staff reviewers.
NUREG Report Titles 0696 -
Functional Criteria for Emergency Response Facilities 0700 -
Guidelines for Control Room Design Reviews 0799 -
Draft.Criter'a for Preparation of Emergency Operating Procedures 0801 -
Evaluation criteria for Control Room Design Reviews 0814 -
Methodology for Evaluation of Emergency Response Facilities J l
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Emergency Action Levels for Light Water Reactors .
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. - 0835 , . Human Factors Acceptance Criteria for SPDS Regulatory Guides ,
1.23 (Rev. 1) -
Meteorological Measurement Program for Nuclear Power Plants 1.97 (Rev. 2) -
Instrumentation for Light-Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following
- an Accident 1.101 (Rev. 2) -
Emergency Planning for Nuclear Power Plants 1.47 -
Bypassed and Inoperable Status Indication for Nuclear Power.
Plant Safety Systems e
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- 3. COORDINATION AND INTEGRATION OF INITIATIVES
- 1. The design of the Safety Parameter Display System (SPDS), design of instrument displays based on Regulatog Guide 1.97 guidance, control room design review, developnent of symptom oriented emergency operating proce-dures, and operating staff training should be integrated with respect to the overall enhancement of operator ability to ;omprehend plant conditions and cope with emergencies. Assessment of information needs and display formats and locations should be performed by individual licensees. The SPDS could affect other control room improvements that licensees may consider. In some cases, a good SPDS may obviate the need for large- scale control room modifications. 14cwever, installation of the SPDS should not be delayed by slower progress on other initiatives. The SPDS should not be contingent on completion of the control room cesign review. NRC does not plan to impose additional requirements on licensees regarding SPDS.
- 2. Implementation of part or all of Regulatory Guide 1.97 (Rev. 2) represents a control room improvement. The implementation of control room improve-ments is not contingent on implementing Technical Support Center (TSC) and l
Emergency Operations Facility (EOF) requirements.
- 3. The Technical Support Center (TSC) and Emergency Operations Facility (EOF)
\ are dependent on control room improvements in, terms of communication and
( instrumentation needs among the TSC, EOF, and control room. TSC and EOF
. facilities are not necessarily dependent on each other. The Operational Sbpport Center (OSC) is independent of TSC and EOF.
- 4. The three groups of initiatives--SPDS, control room improvements, and ,
emergency response facilities (TSC, EOF, OSC)--should have the following interrelationships:
. a. The SFDS is an improvement in the control room because it enhances operator ability to comprehend plant conditions and interact in situations that require human intervention. The SPDS could affect -
other control room improvements that licensees may consider. In seme cases, a good SPDS could obviate the need for extensive modifications
. to control rooms.
- b. New instrumentation that may be added to the control room should be considered a requirement for inclusion in the design of the TSC and EOF only to the extent that such instrumentation is essential to the' performance of TSC and EOF functions.
- c. The SPDS and control room improvements are essential elements in operator training programs and the upgraded plant-specific emergency operating procedures. ,
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] d. Act,uisition, processing, and. management of data for SPDS, control room improvements, and emergency response facilities should be coordinated but need not be centralized. -
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- 5. Specific implementation plans and reasonable, achievable schedules should be established by* agreement between the NRC Project Manager and each individual licensee. The NRC office responsible for implementing each .
requirement snould develop procecures icentifying the following:
. implementation, checking licensee rate of progress, and verifying compliance, including the extent to which HRC review and inspection is necessary during implementation.
- b. Procedural methods and enforcement measures that could be used to
. ensure NRC staff and licensee attention to meeting mutually agreed upon schedules without significant delays and extensions.
- 6. The NRC Project Manager for each nuclear power plant is assigned program management responsibility for NRC staff actions associated with imple-menting emergency response initiatives. The NRC Project Manager is the principal contact for the licensee regarding these initiatives.
- 7. NRC will make allowances for work already cone by licensees in a good-faitn effort te meet requirements as they uncerst.ane them. For each case ,
in which a licensee would have to remove or rip out emergency response a facilities or equipment that was installed in good fai,th to meet previous
{ guidance in order to meet the basic requirements described in this docu-ment, the Director of the Office of Nuclear Reactor Regulation or Inspec-tion and Enforce =ent will review the circumstances and determine whether removal is necessary or existing facilities or equipment represent an acceptable alternative. Any regulatory position that would require the removal or major modification of existing emergency response facilities or equipment requires the specific approval of the Office Director.
- 8. NRC recognizes that acceptable alternative methods of phasing and inte-grating emergency response activities may be developed. Each licensee .
needs flexibility in integrating these activities, taking into account the varying degree to which the licensee has implemented past requirements and guidance. An example of a way in which these activities could be inte- .
grated is discussed below. Other methods of integration proposed by licensees would be reviewed considering licensees' progress on each .
initiative. .
- a. SPDS (1) Review the functions of the nuclear power plant operating staff that are necessary to recognize and cope with rare events that
? (a) pose significant contributions to risk, (b) could cause operators to make cognitive errors.in diagnosing them, and (c) are not included in routine operator training programs. .
( - . .
(2) Combine the results of this review with accepted human factors i
prihciples to select parameters, data display, and functions to be incorporated in the SPDS. t i
6
(
(3) . Design, build, and install the SPDS in the control room and train its users.
- b. To be done parallel without delaying SPDS, complete emergency opera-ting procedure technical guidelines that will be used to develop plant-specific emergency operating procedures.
- c. Using these E0P technical guidelines, the SPDS design, and accepted human factors principles, conduct a review of the control room Apply the results of this review to:
! design.
(1) Verify SPDS paramete'r selection, data display, and functions.
(2) Develop plant-specific E0Ps. I I (3) Design control room modifications that correct conditions adverse to safety (reduce significant contributions to risk),
and add additional instrumentation that may be necessary to , i implement Regulatory Guide 1.97.
(4) Train and qualify plant operating staff regarding E0Ps and modifications.
f'
- d. Verify, prior te finali ation of designs for modifications and of procedures and training, that the functions of control room operators in emergencies can be accomplished (i.e. , that the individual initia-
-( tives have been integrated sufficiently to meet the needs of control I
.' room operators and provide adequate emergency response capabilities),
- e. Implement E0Ps and install control room modifications coincident with scheduled cutages as necessary, and train operators in advance of ,
these changes as they are phased into operation.
S e
e -
e 4 -
e e
e e e
i . .
l 7 .
l
(_
~
- 4. SAFETY PARAMETER DISP!.AY SYSTEM (SPDS)
Current Reculatory Recuirements .
. No licensee action is required. ,
Functional Statement The SPDS should provide a concise display of critical plant variables to the control room operators to aid them in rapidly and reliably determining the safety status of the plant. Although the SPDS will be operated during normal
- operations as well as during abnornial conditions, the principal purpose and function of the SPDS is to aid the control room personnel during abnormal and emergency conditions in determining the safety status of the plant and in assessing whether abnormal conditions warrant corrective action by operators to
- t. void a degraded core. This can be particularly important during , anticipated transients and the initial phase of an accident. -
Recommended Recuirements
(
,' 1. Each operating reactor shall be provided with a Safety Parameter Display System that is located convenient to the control room operators. This
- system will continuously display information from which the plant safety
( status can be readily and reliably assessed by control room personnel who !
- /
a,re. responsible for the avoidance of degraded and damaged core events.
- 2. The control room instrumentation required (see General Design Criteria 13 and 19 of Appendix A to 10 CFR 50) forms the basic safety components '
required for safe reactor operation under normal, transient, and accident conditions. The SPDS is used in addition to the basic components and serves to aid and augment these components. Thus, requirements applicable to control room instrumentation are not needed for this augmentation (e.g. , GDC 2, 3, 4 in Appendix A; 10 CFR Part 100; single-f ailure require-ments). Tae SPDS need not meet requirements of the single-failure criteria and it need not be qualified to meet Class 1E requirements. The SPDS shall be suitably isolated from electrical or electronic interference with equipment and sensors that are in use for safety systems. The SPDS need not be seismically qualified, and additional seismically qualified .
indication is not required for the sole purpose of being a backup for SPDS. After the SPDS has been installed, operating procedures should be available that will allow timely and correct safety status assessment when 1 the SPDS is not available. -
i
- 3. There is a wide range of useful information that can be provided by i vari ~ous systems. This information is reflected in such staff documents as NUREG-0696, NUREG-0835, and Regulatory Guide,1.97.
i Prompt. implementation of an SPDS can provide an important contribution to plant safety. The selection of specific information that should be provided for k particular plant shall be based on engineering judgment of individual plant licensees, taking into account the importance of prompt I
- implementation. ,
hi-- -
. . . . , , . . , m___. _ _ _ _ _ _ _ _ _. _ _ ----a._ -
. 8
( .
4.
. The SPDS display shall be designed to incorporate accepted human factors principles sobythat comprehended SPDS the displayed information can be readily perceived and users.
l S.
), ,
Minimum information tion to plant operators about:to be provided shall be sufficient to provide informa--
- a. Reactivity control b.
- c. Reactor core cooling and heat removal from the primary system j d.
Reactor coolant system integrity >
Radicactivity control -
I e.
Containment conditions' The specific parameters to be displayed shall be determined by the
-licensee.
6.
The licensee shall prepare a written safety analysis describing the basis status of each identified function for a wide range of events include symotoms of severe accidents.
,- Suen analysis, along with the -
iq specific below. implementation plan for SPDS shall be reviewed as described k .
- 7. .
- in accordance with the licensee's technical specification whether technical the changes involve an unreviewed safety question or change of !!
specifications.
normal fashion with prior NRC review.If they do, they shall be processed in the i If the changes do not involve an
. I the licensee may implement such changes without prior a However, the licensee's analysis shall be submitted to NRC promptly on completion of ' review by the licensee's offsite committee.
Based on the -
results of NRC review, the Director of IE or the Director of NRR may question is posed.by the licensen's proposed system, or if th anniysis is seriously inadequate.
l I
)
Inteoration .
Prompt implementation of an SPDS is asdesign goal and of primary importance.
- The schedule for implementing SPDS should not be impacted by schedules for the control room design review and development of symptom-oriented -
emergency operating procedures. For this reason, licensees should develop and propose an integrated schedule for implementation in which the SPDS design is an input to the other initiatives. If reasonable, this schedule should be accepted by NRC. -
( Reference Dorunents .
-- Need for SPDS identified ,
e *
. ,j
u- -......a_.m- -w'-- -
a c
March 11,1962 ) SECY-82-111
%,......i -
POLICY ISSUE For: (Notation Vote) LICENSING DIVISIOR
.The Commissioners bIUMRY From: William J. Dircks Executive Director for Operations .
Subject:
REQUIREMENTS FOR EMERGENCY RESPONSE CAPASILITY Purcose: To request Commission approval of a set of basic re-quirements for emergency response capability and approval for the staff to work with licensees to develop plant-specific implementation schedules.
Discussion: One of the first issues reviewed by the Comittee to Revisw Gent-ic Recuirements (CRGR) was the broad arst cf-emergency responst facilities anc capamilities at nuclear plants. The Cons:itcas found tna- imple-(
mentation schedules were not being coordinated within the NRC. In addition, existing NRC dccuments published as guidance to licensees were sometimes being used as ^fi:T recuirements. Discussions with industry representatives and the staff indicated that some licensees had slowed down on work it *.his area pending NRC clarification of its requirmeents.
. Some utilities have virtually stopped work on some of the items, while others have proceeded and, in some cases, completed some of the items. The Committee recommended that steps be taken by the Office Directors involved to clarify the requirments and i j
implementation schedules for the Safety Parameter Display System (SPDS), Control Room Design Review, upgraded Emergency. Operating Procedures, Regulatory Guide 1.97, Technical Support Center (TSC). Operational Support Cen*.ar (OSC), and Emergency Operations Facility (EDF). In my meno to the Coeurission dated December 31, 1981, I noted that the DEDROGR staff would work with the program offices to clarify the basic requirements in this area and establish a revised implementation plan.
Enclosed are the staff's recommendations for the requireme.1ts in the broad aret of emergency response facilities and caoabilities outlined above. The requirements were developed by the program offices ,
Contact:
V. Stallo, Jr., DEDROGR -
49-29704
_ _ _____n-_-.--. - - - - - - - - - - =
The Comissioners _
3-and are supportec ey CRGR. Tne enclosure represents a distillation of fun::amental requirements from the .
broad range of guidanca documents that NRC has issued (principally NUREE reports and Regulatory Guides) . The staff intends that the guidance documents 1 refer ed to in the enclosure not be used to impose requirements on licensees, but rather that they be used as sources of guidance for NRC reviewers and licensees regarding acceptable means for meeting the fundamental requirements proposed.
In discussions w!th owners' groups and individual 2
licensees, the staff has learned that the Comission approved schedule of October 1,1982, for implementation r of the T5C and EOF probably cannot be met. In recognition of this fact and the difficulty of
- implementing generic deadlines, the staff is proposing 7 that plant-speci*ic schedules be established which
" u ti ince account One unique status of sten pian..
t Eacn licensee would be requested t: sutait a pmposed
( schedule for empleting the actions to comply with the fundamental recuirements. The NRC h cje::
Manage- for each plant should be knowledgeable of the ove all work effort going on at a plant and, based on guidance received from NRC management, could reach agreement with licensees on schedules which optimize use of utility and NRC resources. The
- agreed gen completion dates would be formalized in an order. By this approach, future staff coordination .
problems regarding implementation schedules will be avoided.
Resource iistimates: The costs to licensees to implement the requirements
' proposed in the enclosure were included in the estimates set out in NUREG-0660
_ Recommendation: That the Consission:
- 1. Acerove the fundamental requirements described in tne enclosure.
=
- 2. Accreve the issuance of the requirements in the enclosure by 50.54f letters as a revision to NUREG 0737 3
Amoreve the method for establishing plant.
( specific implementation schedules described in the enclosure.
G
_, _ - . . , - - _ ~ - - - - ~ - - " ' '
(
The Commissioners 4 Accreve One implementatior. of these requirements tnrougn plant-specific orders.
- 5. Note that the sta** intends to use the previously issued NUREG reports and Regulatory Guides as guidance documents only.
Scheduline: Licensees are currently required to establish a TSC and EOF by October 1. Prompt action on this paper is required in order to provide guidance to licensees.
. William J. Dircks Exec::tive Director for Operations
Enclosure:
l NR 3.zff Reconnenosticr.
on the Requiremen.: fop Eme gency Response Capability Cosm:issioners' c:mmen.s -snould be providec directly te he 0"fice of the Secretary by c.c.b. Monday March 29, 1982.
Concission Staff Office connents, if any, should be submitted to the-Comrissioners NLT Monday, March 22, 1982, with an infor1 nation copy to the Office of the Secretary. If the paper is of such a nature that it requires additional time for analytical review and coement, the Concissioners and the Secretariat should be apprised of when coaments may be expected.
DISTRIBUTTON counnssioners Comnission Staff Offices Exec Dir for Operations Is,ec Legal Director ACRS
Secretariat
4
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8ACKGR0Uji0 CRGR REVIEW OF OVEP.All MRC ACTIVITIES fil EMERGENCY RESP 0ffSE (12/3/81)
NRC.0FFICE ACTIVITIES NEEDEU RETIEll COORDINATION
- $~
ACTmt REQUIREMENis NOT CLEAlt .
-- NoiiEs5 AND Rt's sninS 56MElids uSEu ni sTarr AS riRM REquiREncNTS 4
- . 'i. . : . i O e h. . 9 4 I INITIAL DRAFJ OF STAFF RECOMENIMil0NS (12/29/81) 4 . e. . ..
ACRS BRIEFING (1/8/82)
FINAL STAFl* RECOMENDATIONS CONSIDERING All COMENTS (3/10/82)
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Specified SPDS NUREG-0737 ,
-NUREG-0696 --
Functional criteria for SPDS .
Specific a::eptan:e criteria keyed te 0596 NUREG-0835 ,
Reg. Guide 1.97 (Rev. 2) -- Instrumentation for Light-Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident I -
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- 5. DETAILED CONTROL ROOM DESIGN REVIEW j Current Reculatorv Reouirements As specified in Item I.D.1 in NUREG-0737, the implementation schedule is still !
t.o be developed.
Functional Statement The objective of the control room design review is to " improve the ability of !
nuclear power plant control room operators to prevent accidents or cope with l accidents if they occur by improvirig the information provided to them" (from i NUREG-0660, Item I.D.1). As a complement to improvements of plant operating !
_. staff capabilities in response to transients and other abnormal conditions that l will result from implementation of the SPDS and from upgraded emergency opera-ting procedures, this design review will identify any modifications of control room configurations that would contribute to a significant reduction of risk and enhancement in the safety of operation. Decisions to modify the control ,
room would include consideration of long-term risk reduction and any potential !
temocrary decline in safety after modifications resulting from the need to !
, relearn ma'intenance and opera .ing pro:edures. This should be carefully i
[ reviewed by persons competent in human factors engineering and risk analysis. ;
(
Recommended Recuirements l
.. . i
- 1. donduct a control room design review to identify human engineering dis- [
crepancies. The review shall consist of: i t
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- a. The establishment of a qualified multidisciplinary review team and a i review program incorporating accepted human engineering principles. [
i
- b. The use of function and task analysis (that had been used as the i basis for developing emergency operating procedure Technical Guide- .- j lines) to identify control room operator tasks and information and i control requirements during emergency operations. This analysis has i multiple purposes and should also serve as the basis for developing i training and staffing needs and verifying SPDS parameters. j
- c. A comparison of the display and control requirements with a control i room inventory to identify missing and surplus (distracting) displays !
. and controls. j i
- d. A control room survey to identify deviations from accepted human i factors principles. This survey will include, among other things, '
assessment of control room layout, the usefulness of audible and visual alarm systems, information recording and recall capability, i and control room environment, t 1 . .
- 2. Assess which human engineering dis'erepancies are significant and should be.
. corrected. Select design improvements that will correct those discrep- ,
1 ancies. Improvements that can be accomplished with an enhancement program- i
'. (paint-tape-label) should be done promptly. l
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. 3. Verify that each selected design improvement will provide the necessary l correction, and can be introduced in the control roort without creating any l
unacceptable human engineering discrepancies because of significant ,
contribution to increased risk, unre.iewed safety questions, or situations '
in which a temporary reduction in safety could occur. Improvements that are introduced should b'e coordinated with changes resulting from other improvement programs such as SPDS, operator training, new instrumentation ;
(Reg. Guide 1.97, Rev. 2), and upgraded emergency operating procedures. ,
1 Documentation and NRC Review
- . 1. All licensees shall submit a program plan within two months of the start of the control room review that describes how items 1, 2 and 3 above will l be accomplished. NRC approval is not required before licensees conduct their reviews.
- 2. Selected licensees will undergo an in progress audit by the NRR human factors staff based on the program plans and advice from resident inspectors.
- 3. All licensees shall submit a summary report outlining proposed control room changes. The report will also provide a summary justification for human engineering discrepancies with safety significance to be left
..( uncorrected or partially corrected.
- 4. , Vithin two weeks after receipt of the licensee's summary report, the NRC ;
will inform the licensee whether it will conduct a ' pre-implementation '
onsite audit. The decision will be based on the content of the program plan, summary report, and results of NRR in progress audits, if any. 'The licensee selection for pre-implenentation audit may or may not include :
licensees selected for in-progress audits under paragraph 2. l S. For control rooms selected for pre-implementation onsite audit, within one month after receipt of the summary report, the NRC will conduct: -
4
- a. A pre-implementation audit of proposed modifications (e.g. , equipment I additions, deletions and relocations, and proposed modifications). 1
- b. An audit of the justification for those human engineering discrep-ancies of safety significance to be left uncorrected or only partially corrected. .
The audit will consist of a review of licensee's record of the control room reviews, discussions with the licensee review team, and usually &
control room visit. Within a month after this onsite audit, NRC will
- issue its safety evaluation report (SER). ,
( 6.. For control rooms for which NRC does not perform a pre-implementation onsite audit ,HRC will conduct a review and issue its SER within two
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' months after receipt of the lic'ensee's summary i eport. The review shall be similar to that conducted for pre-implementation plants under para-graph 5 above, except that it may or may not include a specific audit.
The SER shall indicate whether, based on the review carried out, changes in the licensee's modification plan are needed to assure operational
, safety. Flexibility is considered in the control room review, because !
i certain control board discrepancies can be overcome by techniques not involving control board changes. These techniques could include' improved procedures, improved training, or the SPDS. -
- 7. The following approach will be used f6r OL review. For OL applications with SSER dates prior to June 1983, licensing may be based on either a Preliminary Design Assessment or a Control Room Design Review (CRDR) at the applicant's option. However, applicants who choose the Preliminary Design Assessment option are required to perform a CRDR after licensing.
For applications with SSER dates after June 1983, Control Room Design -
Review will be required prior to licensing.
Inteoration Prompt im'lementatier.
p of an SPDS is a desigr. goal and of prima y imoortance.
The schedule for. implementing SPDS should not be impacted by schecules
, for the control room design review and development of symptom-oriented *
( emeroency operating procedures. For this reason, licensees should ,
- develop and propose an integrated schedule for implementation in which the SPDS design is an input to the other initiatives. . If reasonable, this schedule should be accepted by NRC.
Reference Documents ,
NUREG-0585 --
States, that licensees should conduct review.
NUREG-0660, Rev. 1 --
States that NRR will require reviews for operating ~
reactors and operating licensee applicants.
NUREG-0700 --
Final guidelines for CRDR.
NUREG-0737 --
States that requirement was issued June, 1980, final".
guidance not ye'. issued.
l NUREG-0801 --
October 1981 t' raft for comment; staff evaluation criteria. _
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1 REGULATORY. GUIDE 1.97
- 6. APPLICATION TO EMERGENCY RESPONSE FACILITIES
." Current Reculatory Reouirements No licensee action is required.
Functional Statement Regulatory Guide 1.97 provides data to assist control room operators in pre-
. venting and mitigating the consequences of reactor accidents.
Recommended Reovirements
- 1. Control Room ,
Provide measurements and indication of Type A, B, C, D, E variables listed -
in Regulatory Guide 1.97 (Rev. 2). Individual licensees may take excep-tions based on plant-specific design features. BWR incore thermocouples
- and continuous offsite dose monitors are not required pending their .
further deveicpment and consideration as requirements. 1; is acceptable to rely on currently installed equipment if it will measure over the range ~
.( indicated in Regulatory Guide 1.97 (Rev. 2), even if the equipment is p,resently not environmentally qualified. Eventually, all the equipment required to monitor the course of an accident would be environmentally qualified in accordance with the pending Commission rule on environmental qualification. .
Provide reliable indication of the meteorological variables (wind direc- ,
tion, wind speed, and atmospheric stability) specified in Regulatory Guide 1.97 (Rev. 2) for site meteorology. No changes in existing meteoro-logical monitoring systems are necessary if they have historically provided reliable indication of these variables that are representative of .
meteorological conditions in the vicinity of the plant site. Information on meteorological conditions for the region in which the site is located shall be available via communication with the National Weather Service.
The Type A, B, C, D, E variables that are essential for performance of TSC functions shall be indicated in the TSC.
- a. BWR incere thermocouples and continuous offsite dose monitors are not
-' required pending their further development and consideration as requirements.
(4 b. The indicators and associated circuitry shall be of reliable design but need not meet Class IE, single-failure or seismic qualification requiremtnts.
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- 3. Emergency Ocerations Facility (EOF)
- a. Those primary indicators needed to monitor containment conditions and releases of radioactivity from the plant shall be provided in the EOF.
- b. The EOF data indications and associated circuitry shall be of ,
reliable design but need not meet Class 1E, single-failure or seismic qualification requirements.
Documentation and NRC Review ,
NRC review is not a prerequisite for implementation. Staff review will be in ,
- the form of an audit that will include a review of the licensee's method of implementing Regulatory Guide 1.97 (Rev. 2) guidance and the licensee's sup- '
porting technical justification of any proposed alternatives.
The licensee sh'all submit a report describing how it meets these requirements.
The submittal should include documentation which may be in the form of a table ,
that includes the following information for each Type A, B, C, D, E variable shown in Regulatory Guide '. 97 (Rev. 2):
.( (a) instrument range -
(b) environmental qualification (as stipulated in, guide or state criteria)
(c) s,eismic qualification (as stipulated in guide or state criteria)
(d) quality assurance (as stipulated in guide or state, criteria)
(e) redundancy and sensor (s) location (s)
(f) power supply (e.g. , Class IE, non-Class IE, battery backed) ,
(g) location of display (e.g. , control room board, SPDS, chemical laboratory) '
(h) schedule (for installation or upgrade)
Deviations from the guidance in Regulatory Guide 1.97 (Rev. 2) should be :
explicitly shown, and supporting justification or alternatives should be ~
i presented.
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- 7. UPGPADE EME,RGENCY OPERATING PROCEDURES (EOPs)
Current Reculatory Recuirements ,
, NUREG-0737 Item I.C.1, which has been approved by the Commission for imple-mentation.
Functional Statement Symptom-based emergency operating procedures will improve human reliability and the ability to mitigate the consequences of a broad range of initiating events
. and subsequent multiple failures or operator errors.
Recommended Recuirements
- 1. In accordance with NUREG-0737, Item I.C.1, reanalyze transients ard accidents and prepare Technical Guidelines. These analyses will identify operator tasks, and information and control needs. The analyses also serve as the basis for integrating upgraded emergency operating procedures and the control room design review and verifying the SPDS design.
I 2. Upgrade E0Ps to be consistent with Tecnnical Guidelines and an approp-iate procedure Writer's Guide. . .
4
- 3. Provide appropriate training of operating personnel on the use of upgraded
. ESPs prior to implementation of the E0Ps.
- 4. Implement upgraded E0Ps. .
Documentation and NRC Review
- 1. Submit Technical Guidelines to NRC for review. NRC will perform a pre-implementation review of the Technical Guidelines and the Writer's Guide.
Within two months of receipt of the Technical Guidelines and Writer's .
Guide, NRC will advise the licensees of their acceptability.
- 2. Each licensee shall submit to NRC a procedures generation package at least three months prior to the date it plans to begin formal operator training ,
on the upgraded procedures. NRC approval of the submittal is not necessary prior to upgrading and implementing the E0Ps. The procedures generation package shall include:
- a. Plant-Specific Technical Guidelines - plant specific guidelines for plants not using generic technical guidelinen. For plants using generic technical guidelines, a description of the planned method for
- developing plant specific E0Ps from the generic guidelines, including
- plant specific information. .
q' .
A Writer's Guide that details the specific methods to be used by the b.
,1icensee*in preparing E0Ps based on the Technical Guidelines.
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16
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- c. A description of the program for validation of the E0Ps.
- d. A brief description of the training program for the upgraded E0Ps.
- 3. All procedures generation packages will be reviewed. On an audit basis for selected facilities, upgraded E0Ps will be reviewed. The details and extent of this review will be based on the quality of the procedures generation packages submitted te NRC. A sampling of ugpraded E0Ps will be reviewed for technical adequacy in conjunction with the NRC Reactor Inspection Program.
Reference Documents NUREG-0660, Item I.C.1,.I.C.8, I.C.9 NUREG-0799 e
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- 8. ' EMERGENCY RESPONSE FACILITIES Current Reculatory Reouirements- ,
.' 10 CFR 50.47(b)(6) (for Operating License applicants) -- Requirement for prompt communications among principal response organizations and to emergency personnel and to the public.
10 CFR 50.47(b)(8) -- Requirement for emergency facilities and equipment to support emergency response.
. 10 CFR 50.47(b)(9) -- Requirement that adequate methods, systems and equipment for assessing and monitoring actual or potential offsite consequences of a radiological emergency condition are in use.
10 CFR 50.54(q-) (for Operating Reactors) -- Same requirement as 10 CFR 50.47(b) plus 10 CF.R 50, Appendix E. .
10 CFR 56, Appendix E, Paragraph IV.E Requirement for:
"1. Equipment at the site for personnel monitoring;
.( "2. Equipment for determining the magnitude of and for continuous 1'y
. assessing the impact of the release of radioactive materials to the environment; "3. Facilities and supplies at the site for decontamination of onsite individuals; *
"4. Facilities and medical supplies at the site for appropriate emergency first aid treatment; "5. Arrangements for the services of physicians and other medical personnel qualified to handle radiation emergencies on site; "6. Arrangements for transportation of contaminated injured individ-uals from the site to specifically identified treatment facili .
ties outside the site boundary; "7. Arrangements for treatment of individuals injured in support.of licensed activities on the site at treatment facilities outside the site boundary; -
"8. A licensee onsite technical support center and a licensee
- near-site emergency operations fac.ility from which effective direction can be given and effective control can be exercised
'( - . during an emergency; .
.. "9. At ienst one onsite and one offsite communications system; each system shall have a backup power source.
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. All communication plans shall have arrangements for emergencies, including titles and alternates for these in charge at both ends l
of the communication ' inks and the primary and backup means of communication. Where consistent with the function of the goverr. mental agency, these arrangements will include:
"a. Provision for communications with contiguous State / local governments within the plume exposure pathway (emergency planning zone) EPZ. Such communications shall be tested monthly.
"b. Provision for c'ommunications with Federal emergency response organizations. Such communications systems shall be tested annually.
" c. Provision for communications among the nuclear power ,
reactor control room, the onsite technical support center, and the near-site emergency operations facility; and among the nuclear facility, the principal State and local emer-genqy operations centers, and the field assessment teams.
Such communications systems shall be tested annually.
" d. Provision for communications by the licensee with NRC .
( Headquarters and the appropriate NRC Regional Office
. Operations Center froc the nuclear power reactor control -
room, the onsite technical support center, and the near- ,
site emergency operations facility.' Such communications shall be tested monthly." ,
Within this section on emergency response facilities, the Technical Support Center (TSC), Operational Support Center (OSC) and Emergency Operations
. Facility (EOF) are addressed separately in terms of their functional statements and recommended requirements. The subsections on Documentation and NRC Review -
and Reference Documents that follow the EOF discussion apply to this entire . I section on ' emergency responra facilities. j e
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l Technical Support Center (TSC)
Functional Statement
. The TSC is the onsite technical support center for emergency response. When activated, the TSC is staffed by p' redesignated technical, engineering, senior management, and other licensee personnel, and five predesignated NRC personnel.
During periods of activation, the TSC will operate uninterrupted to provida plant management and technical support to plant operations personnel, and to relieve the reactor operators of peripheral duties and communications not directly related to reactor system manipulations. The TSC will perform EOF
- functions for the Alert Emergency class and for the Site Area Emergency class and General Emergency class until the EOF is functional.
Recommended Recuirements
' .The TSC will be: - -
- 1. l.ocated within the site protected area so as to facilitate necessary interaction with control room, OSC, EOF and other personnel involved with the emergen:y. .
- 2. Sufficient to accommodate and support NRC and licensee predesignated
-( personnel, equipment and documentation in the center.
- 3. ftructurally built in accordance with the National Uniform Building Code.
- 4. Environmentally controlled to provide room air temperature, humidity and ~
cleanliness appropriate for personnel and equipment.
- 5. Piovided with radiological protection and monitoring equipment necessary to assure that radiation exposure to any person working in the TSC would not exceed 5 rem whole body, or its equivalent to any part of the body, -
for the duration of the accident. .
- 6. Provided with reliable voice and data communications with the control room and EOF and reliable voice communciations with the OSC, NRC Operations Centers and state and local operations centers. .
- 7. Capable of reliable data collection, storage, analysis, display and communication sufficient to determine site and regional status, determine changes in status, forecast status and take appropriate actions. The '
following variables shall be available in the TSC: -
(a)
- the variables in the appropriate Table 1 or 2 of Regulatory
- Guide 1.97 (Rev. 2) that are essential for performance of TSC
!! functions; and -
M (b) the meteorological variables in Regulatory Guide 1.97 (Rev. 2) for
. site vic'inity and National Weather Service data available by voice communication for the region in which the plant is located.
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Principally those data must be available that would enable evaluating incident sequence, determining mitigating actions, evaluating damages and determining plant status during recovery operations.
- 8. Provided with accurate, complete and current plant records (orawings, schematic diagrams, etc.) essential for evaluation of the plant under accident conditions.
- 9. Staffed by sufficient technical, engineering, and senior designated licensee officials to provide needed support, and be fully operational within approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> af,ter activation.
- 10. Designed taking into account good human factors engineering principles.
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Operational Support Canter (OSC)
Functional Statement When activated, the OSC will' be the onsite area separate from the control room where predesignated operations support personnel will essemble. A
. predesignated licensee official shall be responsible for coordinating and assigning the personnel to tasks designated by control room, TSC or EOF
- personnel.
. Recommended Recuirements The OSC will be: .
- 1. teocated onsite to serve as an assembly point for support personnel and to facilitate . performance of support functions and tasks. .
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Emergency Operations facility (EOF)
Functional Statement .
The EOF is a licensee controlled and operated facility. The EOF provides for management of overall licensee emergency response, coordination of radiological, and environmental assessment, determination of recommended public protective actions, and coordination of emergency response activities with Federal, State, and local agencies.
When the EOF is activated, it will be staffed by predesignated emergency personnel identified in the emergen'cy plan. A designated senior licensee official will manage licensee activities in the EOF.
Facilities .; hall be provided in the EOF for the acquisition, display, and
. evaluation of radiological and meteorological data and containment conditions necessary to determine protective measures. These facilities will be used to l
evaluate the magnitude and effects of actual or potential radioactive releases from the plant and to determine dose projections.
Recommende'd Recuirements f
{ The EOF will be:
- 1. Located and provided with radiation protection features as described in -
Table 1 (previous guidance approved by the Commitsion) and with '
appropriate radiological monitoring systems. ;
- 2. Sufficient to accommodate and support Federal, State, local and licensee predesignated personnel, equipment and documentation in the EOF.
. 3. Structurally built in accordance with the National Uniform Building Code. .
- 4. Environmentally controlled to provide room air temperature, humidity and cleanliness appropriate for personnel and equipment.
- 5. Provided with reliable voice and data communications facilities to the TSC .
and control room, and reliable voice communication facilities to OSC and . I to NRC, State and local emergency operations centers.
- 6. Capable of reliable collection, storage, analysis, displays and communica-tion of information on containment conditions, radiological releases and meteorology sufficient to determine site and regional status, determine changes in status, forecast status and take appropriate actions.
Variables from the following categories that are essential to EOF r
- functions shall be available,in the EOF: .
fk (a) variables from the appropriate Table 1 or 2 Regulatory Guide 1.97 (Rev. 2), and 4
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(b) the meteorological variable's in Regulatory Guide 1.97 (Rev. 2) for site vicinity and regional data available via communication from the National Weather Service.
. 7. Provided with up to date plant records (drawings, schematic diagrams, etc.), procedures, emergency plans and environmental information (such as geophysical data) needed to perform EOF functions.
- 8. Staffed in accordance with Table 2 (previous guidance approved by the Commission). Reasonable exceptions to the 30-minute and 1-hour time limits for staffing should be justified and will be considered by NRC -
- staff.
- 9. Provided with industrial security when it is activated to exclude unauthorized personnel and when it is idle to maintain its readiness.
- 10. Designed taking into account good human factors engineering prindples. -
Documentation and NRC Review
- The conceptual design for emergency response f acilities (TSC, OSC, and EOF) have been submitted to NRC for review. In many cases, the lack of detail in these submittals has precluded an NRC decision of acceptability. Some designs
.ke have been disapproved because they clearly did not meet the intent of the applicable regulations. NRC does not intend to approve each design prior to implementation, but rather has provided in this document those " recommended requirements" which should be satisfied. These recommended requirements provided a degree of flexibility within which licensees can exercise management i prerogatives in designing and building emergency response facilities (ERF)'that satisfy specific needs of each licensee. The foremost consideration regarding ERFs is that they provide adequate capabilities of licensees to respond to emergencies. NUREG guidance on ERFs has been intended to address specific issues which the Commission believes should be considered in achieving improved capabilities. .
Licensees should assure that the design of ERFs satisfies these basic,
- requirements. Exemptions from or alternative methods of implementing these requirements should be discussed with NRC staff and in some cases could require, Commission approval. Licensees should continue work on ERFs to complete them according to schedules that will be negotiated on a plant-specific basis. NRC will conduct appraisals of completed facilities to verify that these requirements have been satisfied and that ERFs are capable of performing their intended functions. Licensees need not document their actions on each specific ites contained in NUREG-0696 or 0814.
Reference Documents (Emergency Response Facilities)
T. 10 CFR 50.47,(b) -- Requirements for emergency facilities and equipment for OLs.
10_CFR 50.54(q) and Appendix E, Paragraph IV.E -- Requirements for emergency facilities and equipment for ors.
9
. . ]
g 24
(
NUREG'-0550 -- Description of and imp 1'ementation schedule for TSC, OSC and EOF.
Eisenhut letter to powe reactor licensees S/13/79 -- Request for commitment to meet requirements.
Denton letter to power reactor licensees 10/30/79 -- Clarification of )
requiremer.ts )
and implementation schedule. l I
Eisenhut letter to power reactor licensees 4/25/80 -- Clarification of requirements.
NUREG-0554 -- Radiological Emergency Response Plans NUREG-0595--Functionaicriteriaforemergencyresponsefacilities.
NUREG-0737 -- Guidance on meteorological monitoring and dose assessment.
Eisenhut letter to power reactor license 2/18/81 -- Commission approved guidance en location, habitability and staff for emergency facilities.
Request and deadline for submittal of conceptual design of f acilities.
(, NUREG-0814 (Draft Report for Comment) -- Methodology for evaluation of ~
emerge,ncy response facilities. -
NUREG-0818 (Draft Report for Comment) -- Emergency Action Levels ,
Reg. Guide 1.97 (Rev. 2) -- Guidance for variables to be used in selected emergency response facilities.
COMJA-80-37, January 21, 1981 -- Commission approval guidance on EOF location and habitability. ,
Secretary memorandum 58l-19, February 19, 1981 --- Commission approval of NUREG-0595 as general guidance only.
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TABLE-MINIMUM STAFFING REQUIREMENTS FOR NRC LICENSEES *
. . . FOR NUCLEAR POWER PLANT EMERGENCIES CapabilityfohAdditions
- Position Title On Major Fur"*io'nal Area Major Tasks or Expertise Shift" 30 min. 60 min.
Plant Ope,silans and Shif t supervisor (SRO) 1 -- --
Sliirt foreman (SRO) 1 Assesshent of
. Operational Aspects Control-room operators 2 -- -- -
. Auxiliary operators . 2 Emergency Direction and Shift technical advisor, 1** -- --
shif t supervisor, or Control (Emergency designated facility Coordinator)*
- manager n>
Noti fication/ Hofity licensee, state 1 1 '2 "'
Communi ca tion **** local, and federal personnel & maintain communication ,
- Radiological Accident Emergency operations Senior manager -- --
1 Assessment and Support facility (EOF) director of Operational Accident Offsite dose Senior health physics --
1 --
Assessment assessment (llP) expertise Offsite surveys --
2 2 .
Onsite (out-of plant) --
1 1 Inplant surveys llP technicians 1 1 1 Chemistry / radio- Rad /ches technicians 1 --
1 chemistry N0lE: Source of this table is NUREG-0654, " Functional Criteria for Emergency Response Facil.ities."
- . .- i 0 .
- i TA8tE 2 (Chr. .) -
- CapabiliLy for Additions. .
- . Position Title On Major Functional Area Major Tasks or Expertise Shifta 30 min. 60 min.
Plant System Technical support Shift technical advisory 1 --
Core / thermal hydraulics Engineeting, Repair 1
. and Corrective Actions Electrical -- --
1 .-
Mechanical -- --
1 Repair and corrective Mechanical maintenance / 1** --
1 actions Radwaste operator 1
- Electrical maintenance / 1** 1 1 - .
Instrument and control 1
~
(I&C) technician 1 --
2**
Radiation protection: llP technicians 2 2 Protective Actions .
(In-Plant)' y
- a. Access control
- b. IIP Coverage for repair, correc-
- tive actions.
search and rescue first-aid, & ,
- firefighting
- c. Personnel monitor-ing
- d. Dosimetry ,
Firefighting -- --
Fire tocal brigade support per
- techni-cal specifi-cation
~
Rescue Operations .
2** Local . f cnd First-Aid support , i
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u . TABLE 2 ('Ce.. d) .
- *
- Capability for Additions _ '
- Position Title On-Major Tasks or Expertise Shifta 30 min. 60 min.
Major Functional Area Site Access Control Security, firefighting Security persannel All per communications, per- security tnd Personnel- plan Accountability sonnel accountability ,
- Total 10 11 15 i
' "For each unaf fected nuclear unit in operation, maintain at least one shif t foreman, one control-room i operator, and one auxiliary operator except that units sharing a control room may share a shift foreman -
l 1f all functions are covered. -
Q
- May be provided by shif t personnel assigned other functions. ,
- 0verall direction of facility response to be assumed by EOF director when all centers are fully manned. Director 1 I of minute-to-minute facility operations remains with senior manager in technical support center or control room.
Ca**May be performed by engineering side to shif t supervisor. ,
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- _ - _ _ _ _ _ _ _ _ _