ML20041F878
| ML20041F878 | |
| Person / Time | |
|---|---|
| Issue date: | 03/10/1982 |
| From: | NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | |
| Shared Package | |
| ML20041F870 | List: |
| References | |
| NUDOCS 8203170503 | |
| Download: ML20041F878 (31) | |
Text
4' NRC STAFF RECOMMENDATIONS ON THE REQUIREMENTS FOR EMERGENCY RESPONSE CAPABILITY 4
i March 10,1982 6
4 8203170503 820315' 1
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CONTENTS M
1.
INTRODUCTION..................................................
1 2.
USE OF EXISTING DOCUMENTATION.................................
3 3.
C0 ORDINATION AND INTEGRATION OF INITIATIVES...................
4 4.
SAFETY PARAMETER DISPLAY SYSTEM (SPDS)........................
7 Current Regulatory Requirements Functional Statement Recommended Requirements Integration Reference Documents 5.
DETAILED CONTROL ROOM DESIGN REVIEW...........................
10 Current Regulatory Requirements Functional Statement Recommended Requirements Documentation and NRC Review Integration Reference Documents 6.
REGULATORY GUIDE 1.97 - APPLICATION TO EMERGENCY RESPONSE FACILITIES....................................................
13
' Current Regulatory Requirements Functional Statement Recommended Requirements Documentation and NRC Review 7.
UPGRADE EMERGENCY OPERATING PROCEDURES (E0Ps).................
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Current Regulatory Requirements Functional Statement Recommended Requirements j
Documentation and NRC Review Reference Documents 8.
EMERGENCY RESPONSE FACILITIES.................................
17 i
Current Regulatory Requirements Technical Support Center....................................
19 Functional Statement Recommended Requirements l
L CONTENTS (Continued)
Operational Support 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 Emergencies.........................
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EMERGENCY RESPONSE CAPABILITY 4
s 1.
INTRODUCTION This report was prepared as a result of a review by the Committee to Review Generic Requirements (CRGR). The recommendations herein have I
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 Response Capability from the wide range of guidance documents that NRC has issued.
It is not intended that these guidance documents (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 emergency 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 g
permits for nuclear power plants.
For applicants for a construction permit (CP) or manufacturing license (ML), the requirements described in this document must be supplemented with the specific pr(visions in the rule specifying licensing requirements for pending CP and ML applications.
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In this regard, it is expected that the staff would review CP and ML 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
'he 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, promulgated to NRC staff, and incorporated as l
l regulatory requirements (e.g., in the Standard Review Plan or by regulation 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 the subject with the licensee and mutually agree on schedules and completion dates. The implementation dates would then be formalized into an enforceable l
document.
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j 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 staff believes to be essential to upgraded emergency response capabilities.
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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i 2.
USE OF EXISTING DOCUMENTATION The NRC staff recommeni
'; hat the following NUREG documents are intended to be used as sources of guidance and information, and the Regulatory Guides are to be considered as guidance or as an acceptable approach to meeting formal
[
requirements.
The items by virtue of their inclusion in these documents shall i
not be misconstrued as requirements to be levied on licensees or as inflexible 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 l
Draft Criteria for Preparation of Emergency Operating Procedures 0799 0801 - Evaluation Criteria for Control Room Design Reviews 0814 - Methodology for Evaluation of Emergency Response Facilities 0818 - Emergency Action Levels for Light Water Reactors 0835 Human Factors Acceptance Criteria for SPDS Regulatory Guides Meteorological Measurement Program for Nuclear Power Plants 1.23 (Rev. 1)
Instrumentation for Light-Water Cooled Nuclear Power Plants 1.97 (Rev. 2) to Assess Plant and Environs Conditions During and Following an Accident 1.101 (Rev. 2) - Emergency Planning for Nuclear Power Plants Bypassed and Inoperable Status Indication for Nuclear Power 1.47 Plant Safety Systems 1
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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 Regulatory Guide 1.97 guidance, control room design review, development of symptom oriented emergency operating proce-dures, and operating staff training should be integrated with respect to the overall enhancement of operator ability to comprehend 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.
However, 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 design 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 Emergency Operations Facility (E0F) 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 E0F facilities are not necessarily dependent on each other.
The Operational Support Center (OSC) is independent of TSC and E3F.
4.
The three groups of initiatives--SPDS, control room improvements, and emergency response facilities (TSC, E0F, OSC)--should have the following interrelationships:
a.
The SPDS 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 some 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 E0F functions.
c.
The SPDS and control room improvements are essential elements in operator training programs and the upgraded plant-specific emergency operating procedures.
d.
Acquisition, processing, and. management of data for SPDS, control room improvements, and emergency response facilities should be coordinated but need not be centralized.
5 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 should develop procedures identifying the following:
a.
The respective roles of NRR, IE, and Regional Offices in managing implementation, checking licensee rate of progress, and verifying compliance, including the extent to which NRC review and inspection is necessary during implementation.
b.
Procedural methods and enforcement measures that could be used to ensure NRC staff and licdnsee 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 done by licensees in a good-faith effort to meet requirements as they understand them.
For each case in which a licensee would have to remove or rip out emergency response facilities or equipment that was installed in good faith 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 Enforcement 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 facilitids 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.
l 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 j
principles to select parameters, data display, and functions to j
be incorporated in the SPDS.
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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 design.
Apply the results of this review to:
(1) Verify SPDS paramete'r selection, data display, and functions.
(2) Develop plant-specific E0Ps.
(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 implement Regulatory Guide 1.97.
(4) Train and qualify plant operating staff regarding E0Ps and modifications.
d.
Verify, prior to finalization 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 room operators and provide adequate emergency response capabilities).
Implement E0Ps and install control room modifications coincident with e.
scheduled outages as necessary, and train operators in advance of these changes as they are phased into operation.
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4.
SAFETY PARAMETER DISPLAY SYSTEM (SPDS)
Current Reculatory Requirements j
r 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 avoid a degraded core.
This can be particularly important during anticipated transients and the initial phase of an accident.
Recommended Requirements 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 are. 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 l
to control room instrumentation are not needed for this augmentation (e.g., GDC 2, 3, 4 in Appendix A; 10 CFR Part 100; single-failure require-ments).
The 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 l
SPDS.
After the SPDS has been installed, operating procedures should be l
available that will allow timely and correct safety status assessment when l
the SPDS is not available.
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3.
There is a wide range of useful information that can be provided by l
vari ~ous systems.
This information is reflected in such staff documents as NUREG-0696, NUREG-0835, and Regulatory Guide 1.97.
Prompt implementation of an SPDS can provide an important contrib' tion to u
plant safety.
The selection of specific information that should be provided for h particular plant shall be based on engineering judgment of individual plant licensees, taking into account the importance of prompt
' implementation.
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4 The SPDS display shall be designed to incorporate accepted human factors z,
principles so that the displayed information can be readily perceived and comprehended by SPDS users.
aq 5.
Minimum information to be provided shall be sufficient to provide informa-1 tion to plant operators about:
a.
Reactivity control b.
Reactor core cooling and heat removal from the primary system Reactor coolant system integrity c.
d.
Radioactivity control e.
Containment conditions' The specific parameters to be displayed shall be determined by the licensee.
1 6.
The licensee shall prepare a written safety analysis describing the basis l
on which the selected parameters are sufficient to assess the safety status of each identified function for a wide range of events, which include symptoms of severe accidents.
Such analysis, along with the specific implementation plan for SPDS shall be reviewed as described below.
7.
The licensee's proposed implementation of an SPDS system shall be reviewed in accordance with the licensee's technical specifications to determine whether the changes involve an unreviewed safety question or change of technical specifications.
normal fashion with prior NRC review.If they do, they shall be processed in the If the changes do not involve an unreviewed safety question or a change in the technical specifications, the licensee may implement such changes without prior approval by NRC.
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 licensee's proposed system, or analysis is seriously inadequate.
Integration Prompt implementation of an SPDS is a design 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 Documents NUREG-0660
-- Need for SPDS identified
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NUREG-0737 Specified SPDS Functional criteria for SPDS NUREG-0696 Specific acceptance criteria keyed to 0696 y
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 g
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DETAILED CONTROL ROOM DESIGN REVIEW Current Regulatory Requirements As specified in Item I.D.1 in NUREG-0737, the implementation schedule is still to 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 accidents if they occur by improving the information provided to them" (from 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 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 temporary decline in safety after modifications resulting from the need to relearn maintenance and operating procedures.
This should be carefully reviewed by persons competent in human factors engineering and risk analysis.
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Recommended Requirements 1.
Conduct a control room design review to identify human engineering dis-crepancies.
The review shall consist of:
a.
The establishment of a qualified multidisciplinary review team and a review program incorporating accepted human engineering principles.
b.
The use of function and task analysis (that had been used as the basis for developing emergency operating procedure Technical Guide-lines) to identify control room operator tasks and information and i
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control requirements during emergency operations.
This analysis has multiple purposes and should also serve as the basis for developing I
training and staffing needs and verifying SPDS parameters.
A comparison of the display and control requirements with a control c.
l room inventory to identify missing and surplus (distracting) displays and controls, d.
A control room survey to identify deviations from accepted human factors principles.
This survey will include, among other things, assessment of contral room layout, the usefulness of audible and visual alarm systems, information recording and recall capability, and control room environment.
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Assess which human engineering dis'crepancies are significant and should be corrected.
Select design improvements that will correct those discrep-ancies.
Improvements that can be accomplished with an enhancement program
- (paint-tape-label) should be done promptly.
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Verify that ea'c'h selected design improvement will provide the necessary correction, and can be introduced in the control room without creating any unacceptable human engineering discrepancies because of significant contribution to increased risk, unreviewed safety questions, or situations in which a temporary reduction in safety could occur.
Improvements that are introduced should be 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.
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 be accomplished.
NRC approval ~is not required before licensees conduct their reviews.
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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.
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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 unco.rrected or partially corrected.
4.
Within.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 decisi,cn 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-implementation audit may or may not include licensees selected for in progress audits under paragraph 2.
'5 For control rooms selected'for pre-implementation onsite audit, within one i
month after receiptsof the summary report, the NRC will conduct:
l A' pre-implement 5 tion audit of proposed modifications (e.g., equb ment a.
auditions, deletions and relocations, and proposed modificatio.ns).
b.
An audit of the justification for those human engineering discrep-ancies of safety significance to be left uncorrected or only partially corrected.
l The audit will consist of a review of licensee's record of the control room reviews, discussions with the licensee review team, and usually a
, control room visit. Within a month-efter this onsite audit, NRC will
' issue its safety evaluation report (SER).'
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'for control rooms for which NRC does not perform a pre-implementation onsite audit. NRC will copduct a review and issue its SER within two e
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12 months after receipt of the licensee's summary report.
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 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 for 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 wil1 be required prior to licensing.
Integration Prompt implementation of an SPDS is a design 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 Documents i
NUREG-0585 Stater 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.
-- States that requirement was issued June, 1980, final'.
guidance not yet issued.
NUREG-0801 October 1981 draft for comment; staff evaluation criteria.
s 13 REGULATORY GUIDE 1.97 6.
APPLICATION TO EMERGENCY RESPONSE FACILITIES Current Regulatory Requirements l
No licensee action is required.
Functional Statement Regulatory Guide 1.97 provides data to assist control room operators in pre-venting and mitigating the conseque'nces of reactor accidents.
Recommended Requirements 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 development and consideration as requirements.
It 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 presently 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 cf 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.
2.
Technical Support Cencer (TSC)
The Type A, B, C, D, E variables that are essential for performance of TSC functions shall be indicated in the TSC.
BWR incore thermocouples and continuous offsite dose monitors are not a.
required pending their further development and consideration as requirements.
b.
The indicators and associated circuitry shall be of reliable design but need not meet Class 1E, single-failure or seismic qualification requi rem'ents.
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_ Emergency Operations 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 'ue 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 shall 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 1.97 (Rev. 2):
(a) instrument range (b) environmental qualification (as stipulated in~ guide or state criteria)
(c) seismic 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 1E, non-Class 1E, 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 l
explicitly shown, and supporting justification or alternatives should be presented.
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UPGRADE EMERGENCY OPERATING PROCEDURES (EOPs)
Current Regulatory Requirements NUREG-0737, Item I.C.1, which has been approved by the Commission for imple-mentation.
Functional Statement J
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 Requirements 1.
In accordance with NUREG-0737, Item I.C.1, reanalyze transients and 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.
2.
Upgrade E0Ps to be consistent with Technical Guidelines cnd an appropriate procedure Writer's Guide.
3.
Provide appropriate training of operating personnel on the use of upgraded E0Ps prior to implementation of the E0Ps.
4.
Implement upgraded E0Ps.
Documentation and NRC Review 1.
Submit Technical Guidelines to NRC for review.
hRC 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 guidelines.
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.
b.
A Writer's Guide that details the specific methods to be used by the licensee in preparing E0Ps based on the Technical Guidelines.
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A description of the program for validation of the E0Ps.
d.
A brief description of the training program for the upgraded E0Ps.
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All procedures generation packages will be reviewed.
On an audit basis
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for selected facilities, upgraded E0Ps will be reviewed. The details and i
extent of this review will be based on the quality of the procedures t
generation packages submitted to NRC.
A sampling of ugpraded E0Ps will be l
reviewed for technical adequacy in conjunction with the NRC Reactor Inspection Program.
l Reference Documents l
NUREG-0660,-Item I.C.1,.I.C.8, I.C.9 NUREG-0799 I
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' EMERGENCY RESPONSE FACILITIES Current Regulatory Requirements' f
i 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 CFR 50, Appendix E.
10 CFR 50, Appendix E, Paragraph IV.E Requirement for:
"1.
Equipment at the site for personnel monitoring; "2.
Equipment for determining the magnitude of and for continuously 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 -ite emergency operations facility from which effective direction can be given and effective control can be exercised during an emergency; "9.
At least one onsite and one offsite communications system; each system shall have a backup power source.
18 All communication plans shall have arrangements for emergencies, including titles and alternates for those in charge at both ends of the communication links and the primary and backup means of communication. Where consistent with the function of the governmental 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.
7 Provision for c0mmunications with Federal emergency "b.
response organizations.
Such communications systems shall be tested annually.
"c.
Provision for communications amcng 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-gency 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 from 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 l
Facility (E0F) 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 section on emergency response facilities.
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Technical Support Center (TSC)
Functional Statement The TSC is the onsite technical support center for emergency response. When activated, the TSC is staffed by predesignated technical, engineering, senior management, and other licensee personnel, and five predesignated NRC personnel.
During periods of activation, the TSC will operate uninterrupted to provide 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 Requirements The TSC will be:
1.
Located within the site protected area so as to facilitate necessary interaction with control room, OSC, EOF and other personnel involved with the emergency.
2.
Sufficient to accommodate and support NRC and licensee predesignated personnel, equipment and documentation in the center.
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 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 E0F and reliable voice communciations with the OSC, NRC Operations Centers and state and local operations centers.
Capable of reliable data collection, storage, analysis, display and l
7.
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 (b) the meteorological variables in Regulatory Guide 1.97 (Rev. 2) for site vicinity and National Weather Service data available by voice communication for the region in which the plant is located.
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1 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 (drawings, schematic diagrams, etc.) essential for evaluation of the plant under accident conditions.
7 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 /> after activation.
10.
Designed taking into account good human factors engineering principles.
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21 Operational Support Center (OSC)
Functional Statement b
When activated, the OSC will'be the onsite area separate from the control room where predesignated operations support personnel will assemble.
A predesignated licensee official shall be responsible for coordinating and assigning the personnel to tasks designated by control room, TSC or E0F personnel.
Recommended Requirements The OSC will be:
1.
Located onsite to serve as an assembly point for support personnel and to facilitate performance of support functions and tasks.
2.
Capable of reliable voice communications with the control room, TSC and E0F.
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22 Emergency Operations Facility (EOF)
Functional Statement The EOF is a licensee controlled and operated facility.
The E0F provides for management of overall licensee emergency response, coordination of radiological.
I and environmental assessment, determination of recommended public protective actions, and coordination of emergency response activities with Federal, State, and local agencies.
When the E0F 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 E0F.
Facilities shall be provided in the EOF for the acquisition, display, and evaluation of radiological and meteorological data and containment coaditions necessary to determine protective measures.
These facilities will be used to evaluate the magnitude and effects of actual or potential radioactive releases from the plant and to determine dose projections.
Recommended Requirements The EOF will be:
1.
Located and provided with radiation protection features as described in Table 1 (previous guidance approved by the Commission) and with appropriate radiological monitoring systems.
2.
Sufficient to accommodate and support Federal, State, local and licensee predesignated personnel, equipment and documentation in the E0F.
3.
Structurally built in accordance with the National Uniform Building Code.
4.
Environmentally controlled to provide room air temperature, humidity and l
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 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.
l Variables from the following categories that are essential to EOF l
functions shall be available in the EOF:
(a) variables from the appropriate Table 1 or 2 Regulatory Guide 1.97 (Rev. 2), and 1
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(b) the meteorological variables in Regulatory Guide 1.97 (Rev. 2) for site vicinity and regional data available via communication from the
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National Weather Service.
7.
Provided with up to date plant records (drawings, schematic diagrams, f
etc.), procedures, emergency plans and environmental information (such as
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geophysical data) needed to perform EOF functions.
j 8.
Staffed in accordance with Table 2 (previous guiosnce approved by the j
Commission).
Rcasonable exceptions to the 30-minute and 1-hour time limits for staffing should be justified and will be considered by NRC staff.
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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 principles.
Documentation and NRC Review The conceptual design for emergency response facilities (TSC, OSC, and E0F) 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 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 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.
I Licensees should assure that the design of ERFs satisfies these basic.
l 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 i
requirements have been satisfied and that ERFs are capable of performing their l
intended functions.
Licensees need not document their actions on each specific
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item contained in NUREG-0696 or 0814.
Reference Documents (Emergency Response Facilities) 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.
24 NUREG-0660 -- Description of and implementation schedule for TSC, OSC and E0F.
Eisenhut letter to power reactor licensees 9/13/79 -- Request for commitment to meal requirements.
Denton letter to power reactor licensees 10/30/79 -- Clarification of requirements and implementation schedule.
Eisenhut letter to power reactor licensees 4/25/80 -- Clarification of requirements.
NUREG-0654 -- Radiological Emergency Response Plans NUREG-0696 -- Functional criteria for emergency response facilities.
NUREG-0737 -- Guidance on meteorological monitoring and dose assessment.
Eisenhut letter to power reactor license 2/18/81 -- Commission approved guidance on location, habitability and staff for emergency facilities.
Request and deadline for submittal of conceptual design of facilities.
NUREG-0814 (Draft Report for Comment) -- Methodology for evaluation of emergency 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 581-19, February 19, 1981 -- Commission approval of NUREG-0696 as general guidance only.
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s TABLE 1 EMERGENCY OPERATIONS FACILITY Option 1 Option 2 Two Facilities One Facility A.
Close-in Primary:
Reduce Habitability
- o At or Beyond 10 miles.
o within 10 miles o No special protection factor.
o protection factor = 5 o If beyond 20 miles, specific o ventilation isolation approval required by the with HEPA (no charcoal)
Commission, and some provi-sion for NRC site team closer to site.
o Strongly recommended location be coordinated with offsite authorities.
B.
Backup EOF o between 10-20 miles n3 o no separate, dedicated facility o arrangements for portable backup equipment o strongly recommended location be coordinated with offsite authorities o continuity of dose projection and decision making capability For both Options:
- located outside security boundary
- space for about 10 NRC employees
- none designated for severe phenomena, e.g., earthquakes
- Habitability requirements are only for the part of the EOF in which dose assessments communications and decision making take place.
If a utility has begun construction of a new building for an EOF that is located with 5 miles, that new facility is acceptable (with less than protection factor of 5 and ventilation isolation and HEPA) provided
" that a backup EOF similar to "B" in Option 1 is provided.
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TABLE 2 MINIMUM STAFFING REQUIREMENTS FOR NRC LICENSEES FOR NUCLEAR POWER PLANT EMERGENCIES Capability for Additions Position Title On Major Functional Area Major Tasks or Expertise Shift
- 30 min.
60 min.
Plant Operations and Shift supervisor (SRO) 1 Assessment of Shift foreman (SRO) 1 Operational Aspects Control-room operators 2
Auxiliary operators 2
Emergency Direction and Shift technical advisor, 1**
Control (Emergency shift supervisor, or Coordinator)***
designated facility manager Notification /
Nofity licensee, state 1
1 2
m Communication ****
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 (HP) expertise
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Offsite surveys 2
2 Onsite (out-of plant) 1 1
Inplant surveys HP technicians 1
1 1
Chemistry / radio-Rad / chem technicians 1
1 chemistry i
NOTE:
Source of this table is NUREG-0654, " Functional Criteria for Emergency Response Facilities."
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TABLE 2 (Continued)
Capability for Additions Position Title On Major Functional Area Major Tasks or Expertise Shift
- 30 min.
60 min.
Plant System Technical support Shift technical advisory 1
Engineering, Repair Core / thermal hydraulics 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
1 (I&C) technician Protective Actions Radiation protection:
HP technicians 2**
2 2
(In-Plant) a.
Access control Q3 b.
HP Coverage for repair, correc-tive actions, search and rescue first-aid, &
firefighting c.
Personnel monitor-ing d.
Dosimetry Firefighting Fire Local brigade support per techni-cal specifi-cation 2**
Local Rescue Operations and First-Aid support
s TABLE 2 (Continued)
Capability for Additions Position Title On Major Functional Area Major Tasks or Expertise Shift
- 30 min.
60 min.
Site Access Control Security, firefighting Security personnel All per and Personnel communications, per-security Accountability sonnel accountability plan Total 10 11 15 "For each unaffected nuclear unit in operation, maintain at least one shift foreman, one control-room operator, and one auxiliary operator except that units sharing a control room may share a shift foreman if all functions are covered.
Of
- May be provided by shift personnel assigned other functions.
- 0verall direction of facility response to be assumed by EOF director when all centers are fully manned.
Director of minute-to-minute facility operations remains with senior manager in technical support center or control room.
^^**May be performed by engineering aide to shift supervisor.
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