ML20195C363
| ML20195C363 | |
| Person / Time | |
|---|---|
| Issue date: | 11/10/1998 |
| From: | Stewart Magruder NRC (Affiliation Not Assigned) |
| To: | Essig T NRC (Affiliation Not Assigned) |
| References | |
| PROJECT-689 NUDOCS 9811170093 | |
| Download: ML20195C363 (8) | |
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j-NUCLEAR REGULATORY COMMISSION 2
WASHINGTON, D.C. ennas annt
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November 10, 1998
' MEMORANDUM TO: Thomas H. Essig, Acting Chief Generic issues and Environmental Projects Branch Division of Reactor Program Management Office of Nuclear Reactor Regulation FROM:
Stewart L. Magruder, Project Manager M L Generic issues and Environmental Projects Branch L
Division of Reactor Program Management Office of Nuclear Reactor Regulation l
SUBJECT:
SUMMARY
OF WORKSHOP WITH THE NUCLEAR ENERGY l
lNSTITUTE (NEI) REGARDING PERFORMANCE INDICATORS FOR ASSESSING RADIATION PROTECTION PROGRAMS l
l On October 20,1998, representatives of the Nuclear Energy Institute (NEI) met with representatives of the Nuclear Regulatory Commission (NRC) at the NRC's offices in Rockville, l
Maryland. Attachment 1 provides a list of workshop attendees.
The purpose of the workshop was to continue discussion and development of performance indicators (PI) to be used by the NRC to help assess the licensee's radiation protection programs at power reactors. NRC manager.1ent provided an introduction to start the meeting, i
encouraged public participation in the meeting, and summarized the goals and time schedules for the project. Both NRC and NEl participants started the discussion by reviewing and reinforcing the points of agreement that were reached at the three-day NRC workshop in late September.
l The strategic performance area (radiation safety) and the comer stones (public and
(
occupational) were discussed and the objectives for both corner stones were generally agreed to - keep worker and public doses below regulatory limits and as low as reasonably achievable
-(ALARA).' The group agreed that the cross-cutting issues (human performance, safety-conscious work place, etc.) would not need separate Pls, but would be evident by overall licensee performance.
For performance assessment, several points were discussed and will ' id in further development a
of specific Pls (e.g., reporting frequency needs to be established, the acceptable number of events should be established as a threshold for reporting, with trending and corrective actions taken by the licensee before NRC intervention). Thresholds should be clearly defined -
regulatory and safety thresholds which separate broad levels of licensee performance must be
. established for the Pls. For occupational radiation safety, two possible Pls were discussed in some detail, it was agreed that one Pl should capture "significant unplanned individual dose" events. Based on a lack of consensus, more work will be necessary to reach closure on these Pls. For the public radiation area, it was agreed to examine the need for Pls in effluents, 9
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l'd T. Essig November 10, 1998
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transportation and the control of radioactive materials. Existing regulations (Part 50, Appendix i
I) may make the development of these Pls relatively straightforward. No final consensus agreement was reached relative to specific Pls. The NRC shared candidate Pls with NEl, which were discussed at the meeting (Attachment 2 contains copies of the NRC handouts provided to l
NEl and the Public present).
NEl shared plans to attempt to verify and vel date draft Pls against historical events, plant 3
operating experience and regulatory violations. This effort will help support the validation and 1
verification efforts for the Pls, prior to implementation.
)
l The meeting closed with plans to meet and continue Pl development at the NEl offices,' along
. with the industry working group, on October 29. This information was shared with the two I
members of public present.
l l
Project No. 689
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Attachments: As stated cc w/att: See next page i
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November 10, 1998 transportation and the control of radioactive materials. Existing regulations (Part 50, Appendix
- 1) may make the development of these Pls relatively straightforward. No final consensus agreement was reached relative to specific Pls. The NRC shared candidate Pls with NEl, which were discussed at the meeting (Attachment 2 contains copies of the NRC handouts provided to NEl and the Public present).
NEl shared plans to attempt to verify and validate draft Pts against historical events, plant operating experience and regulatory violations. This effort will help support the valida+ ion and verification efforts for the Pls, prior to implementation.
The meeting closed with plans to meet and continue P1 development at the NEl offices, along with the industry working group, on October 29. This information was shared with the two members of public present.
I Project No. 689 Attachments: As stated cc w/att: See next page i
1 l
DIETRIBUTION: See attached page l
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SC:Phb OFFICE PM:PGEB PERB BC:PERB NAME SMagrudeIN JWiggl dn CMiller p FMNwicz f
DATE 11/9 /98 11/k/98 11/ M/98 11/h/98 i
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l-Distribution Mtg.' Summary w/ NEl Dated November 10, 1998 Hard Conv
,NS DERB R/F
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- SMagruder JWigginton EMail TEssig CMiller i
REmch JWhite, Rl
.' KBarr, Ril GShear, Rlli BMurray, RIV' NShah-GKuzo RPedersen SKlementowicz AMadison I
PBaranowsky
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i Radiation Protection Performance Indicator Meeting 10/20/98 List of Attendees Name Oraanization
. Charles Miller USNRC.
Pat Baranowsky USNRC Alan Madison USNRC Steve Klementowicz USNRC George Kuzo USNRC Roger Pedersen USNRC Nirodh Shah USNRC' Jim Wigginton USNRC Ralph Anderson NEl Paul Genoa NEl Kim Green NUS Information Services
' David Stellfox McGraw-Hill l
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CAUSES OF OVEREXPOSURES AT NPPs l
1.
. Failure to perform ADEQUATE SURVEYS (work hazards analysis and evaluations)
L 2.
Failure to follow procedures
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3.
Poor HP technician response to changing radiological conditions / work scope l
l 4.
Inadequate Radiation Work Permit
' 5.
Lack of first line supervisor involvement Note:
Need to ensure that the above are somehow factored in and play a role in the l
development of the performance indicator program. Since the above causes should be viewed as precursors to loss of exposure control events, the licensee's i
self-assessment program should be searching for these problems.
CANDIDATE Pls FOR RADIOLOGICAL CONTROLS j
Occupational Collective station TEDE dose (person-rem; some time-rolling average, OR use licensee-develped goals) l CR - ratio of collective station dose delivered at >1.5 rem to individuals to site collective dose Number of overexposures and events with substantial potential for overexposures (would logically use weighting factors - DDE> skin).
High Radiation area access controls challenges (unauthorized entries into >100 mR/h, inadequate RWP, transient dose rates mishandled (e.g., HPT failed to stop work), failure to provide adequate survey,)
Number of hours outside the BWR or PWR Owners Group Guidelines for Chemistry (trending these indicators of "out-of-spec" chemistry parameters important to demonstrate management's commitment to program) l Loss of one or more barriers (or a percentage, say 25% of barriers) when working in Locked HRAs (dose rates > 10R/h) and Very HRAs Willful bypassing or violation of plant HP procedures while working in a
posted airborne, HRA, or VHRA.
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Significant unplanned intakes (>0.05 All )
Significant unplanned DDEs (>100 mrem)
Evidence of lack of HP technician qualification leading to incident i
Effluents / Environmental Number of unplanned effluent releases (both liquid and gaseous)
Some fraction of Appendix I dose design objectives for various dose receptors Number of hours (or events) plant process or effluent radiation monitoring systems are not capable of performing their intended safety function.
i Failure to perform aspects of environmental monitoring program Loss of control of radioactive materials - unauthorized release of materials outside the RCA.
j Radwaste/ Transportation Number of problems with improper shipments noted by LLW receiving authority (includes DOT and PCP discrepancies)
Number of violations of the package's Certificate of Compliance i
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Nuclear Energy Institute Project No. 689 cc: Mr. Ralph Beedle Senior Vice President and Chief Nuclear Officer
- Nuclear Energy Institute Suite 400 1776 l Street, NW Washington, DC 20006-3708 Mr. Alex Marion, Director Programs Nuclear Energy Institute Suite 400 1776 l Street, NW Washington, DC 20006-3708 Ms. Lynnette Hendricks, Director Plant Support Nuclear Energy institute Suita 400 1776 i Street, NW Washington, DC 20006-3708 Mr. Steven Driscol Radiation Protection Institute of Nuclear Power Operation 700 Galleria Parkway Atlanta, Georgia 30339-5957 b
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RIvi:lon 100998 NEl 97-03 l;
Draft Final Rev. 3 i
(NUMARC/NESP-007) i i
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- Methodology for Development i
of Emergency Action Levels 4
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.v Revision 100998 ACKNOWLEDGEMENTS Revision 3 of this report incorporates numerous suggestions provided by utilities that have implemented the NUMARC/NESP 007 EALs, and input provided by the staff of the NRC. NEl acknowledges the valuable input provided, and the extensive technical support provided by the members of the EAL Task Force.
i NOTICE j
This report was prepared as an account of work sponsored by the Nuclear Energy Institute
~(NEI). Neither NEl nor any of its employees, members, or consultants make any warranty,
' expressed or implied, or assume any legal liability or responsibility for the accuracy, completeness or usefulness of any information, apparatus, product or process disclosed in this report, or represent that its use would not infringe privately-owned rights.
The opinions, conclusions, and recommendations :,at fodh in this report are those of the authors and do not necessarily represent the views of NEl, its employees, members or consultants.
Because NEl is supported in part by Federal funds, NEl's activities are subject to Title VI of the
. Civil Rights Act of 1964, which prohibits discrimination based on race, color, or national origin,
- and other federal laws and regulations, rules, and orders issued thereunder prohibiting discrimination. ' Written complaints of exclusion, denial of benefits or other discrimination of l
those bases under this program may be filed with the United States Nuclear Regulatory l
. Commission, Washington, DC 20555 or any other appropriate federal regulatory agency or, among others, the Tennessee Valley Authority (TVA), Omce of Equal Employment Opportunity, 400 West Summit Hill Drive, Knoxville, TN 37902.
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FOREWORD Revision 3 to NUMARC/NESP-007 presents the methodology for development of emergency action levels as an alternative to NRC/ FEMA guidelines contained in Appendix 1 of NUREG-0654/ FEMA-REP-1, Rev. 2 " Criteria for Preparation and Evaluation of Radiological Emergency Response Pfans and Preparedness in Support of Nuclear Power Plants," October 1980 and 10 CFR 50.47 (a)(4). Revision 3 incorporates changes associated with lessons learned by the utilities that have implemented Revision 2 of this document.
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Revision 100998 EAL Revision Task Force 1996-1998 Rodney Brown Duke Power Company Gary Cerkas Arizona Public Service Company Dennis M. Emborsky Northeast Utilities James D. Jones Niagara Mohawk Power Corporation Ron Jorgensen Washington Public Power Supply System Stephen F. LaVie Duquesne Light Company Walter H. Lee Southem Nuclear Operating Company Mark Luksic Tennessee Valley Authority 1
Alan Nelson Nuclear Energy !nstitute i
Kevin Morris Detroit Edison Company David W. Stobaugh Commonwealth Edison Company Martin Vonk Commonwealth Edison Company l
i Liaison to the Task Force Frederick Hasselberg U.S. Nuclear Regulatory Commission James O'Brien U.S. Nuclear Regulatory Commission Randy Sullivan U.S. Nuclear Regulatory Commission Warren U.S. Nuclear Regulatory Commission l
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RIvision 100998 The following persons participated in the development of Revision 2 of the NUMARC/NESP-007 document.
EAL Revision Task Force Reginald C. Rodgers, Chairman Northeast Utilities Thomas B. Blount GPU Nuclear Corporation Roy K. Bros!
Duquesne Light Company Nick S. Catron Tennessee Valley Authority David L. Fugere Consumers Power Company Albert L. Garrou Carolina Power & Light Company i
George J. Giangi GPU Nuclear Corporation f'
Terry Gilman Commonwealth Edison Company l
- Ronald E. Harris Duke Power Company L
Roger Hoyt Wolf Creek Nuclear Operating Corporation i.
John Jenkins Duke Power Company Stephen F. LaVie Duquesne Light Company i
l Theresa M. Lechton Commonwealth Edison Company j
James M. Minneman Pennsylvania Power & Light Company Kevin Moles Wolf Creek Nuclear Operating Corporation Patrick Taylor Pennsylvania Power & Light Company Alan P. Nelson, Project Manager NUMARC Independent Industry Review Group o
l Craig Adams Public Service Electric & Gas Company Rod Kneger Indiana & Michigan Electric Company Milton Stiller Union Electric Company Liaison to the Task Force Frederick Hasselberg U.S. Nuclear Regulatory Commission Michael Jamgochian U.S. Nuclear Regulatory Commission Aby Mohseni U.S. Nuclear Regulatory Commission Edward Podolak U.S. Nuclear Regulatory Commission William Reckley U.S. Nuclear Regulatory Commission Craig Wingo -
Federal Emergency Management Agency Edison Carmack Institute of Nuclear Power Operations
. ERC Environmental and Energy Services Company (ERCE) Study Team John S. Fuoto, Study Team Leader l
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TABLE OF CONTENTS -
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EXEC UTIVE SUM MARY......................................................................
l ACRONYMS.............................................................................................................
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1.0 METHODOLOGY FOR DEVELOPMENT OF EMERGENCY A C TIO N L EVE LS.................................................................
1.1 Background....................................................................................................1.1 l
1.2 Task Force C ha rte r....................................................................
1.3 Structure of the Study................................................................................ 1.2 L
2.0 CHANGES INCORPORATED IN REVISION 3...................................................... 2.1 I
l 3.0 DEVELOPMENT OF BASIS FOR GENERIC APPROACH...................................... 3.1 3.1 Reg ulatory Context................................................................................. 3.1 3.2 Definitions Needed To Develop EAL Methodology................................ 3.3 3.3 Differences in Perspective.............................................................................. 3.4 l
3.4 Recog nition Categories................................................................................ 3.4 3.5 Design Differences..................................................................................... 3.6 3.6 Required Characteristics............................................................................... 3.6 3.7 Emergency Class Descriptions........................................................................ 3.7 3.8 Emergency Class Thresholds.....,.................................................................. 3.8 l
3.9 Emergency Action Levels............................................................................... 3. 9 l
3.10 Treatment Of Multiple Events and Emergency Class Upgrading.................. 3.11 3.11 Emergency Class Downgrading.................................................................... 3.12 3.12 Classifying Transient Events.......................................................................... 3.12 3.13 Interface Between Classification and Activation of Emergenc Shutdown IC/EALs................................................................y Facilities....... 3.13 i
3.14
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3.15 Operating M ode Applicability......................................................................... 3.14 L
4.0
' H U MAN FACTORS C ONSIDERATIONS............................................................... 4.1 4.1 Level of Integration of EALs with Plant Procedures......................................... 4.1 4.2 Method of Presentation.....................................................................
4.3 Symptom-Based, Event-Based or Barrier-Based EALs.................................. 4.2 i
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Revision 100998 TABLE OF CONTENTS (continued)
PAGE 5.0 -
G E N E RIC EA L G U l DA NC E....................................................................................
5.1 Ge neric Arrange me nt.................................................................................. 5.1 5.2 G e n e ric B a se s........................................................................................
5.3 Site Specific I mplementation.......................................................................... 5.3 5.4 Defi n ition s..............................................................................................
INITIATING CONDITIONS MATRICES Category A Abnormal Rad Levels / Radiological Effluent............................................... 5-A-1 Category F Fission Product Barrier Degradation............................................................ 5-F-1 Category H Hazards and Other Conditions Affecting Plant Safety................................ 5-H-1 Category S System M alfu nction.................................................................................. 5-S-1 l
APPENDIX A BASIS FOR RADIOLOGICAL EFFLUENT INITIATING CONDITIONS.......A.1 1
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Revision 100998 TABLE OF CONTENTS (continued) l i
I PAGE TABLES TABLE 5-A-1: Recognition Category A initiating Condition Matrix................................. 5-A-1 TABLE 5 F-1 Recognition Category F Initiating Condition Matrix.................................... 5-F-1 l
TABLE 5-F-2 BWR Fission Product Barrier Reference Table........................................ 5-F-2 TABLE 5-F-3 PWR Fission Product Barrier Reference Table........................................ 5 F-9
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TABLE 5-H-1: Recognition Category H Initiating Condition Matrix............................ 5-H-1 1
I TABLE 5-S-1 Recognition Category S Initiating Condition Matrix................................... 5-S-1 I
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1 EXECUTWE
SUMMARY
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3 Nuclear utilities must respond to a formal set of thresho!:1 conditions that require plant personnel to 4
take specific actions with regard to notifying state and loca govemments and the public when certain j
5 off-normal indicators or events are recognized. Emerger.cy classes are defined in 10 CFR 50.
I 6
Levels of response and the conditions leading to those responses are defined in a joint NRC/ FEMA j
7 guidelines contained in Appendix 1 of NUREG-0654/ FEMA-REP-1, Rev.1, " Criteria for Preparation j
8 and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear 9
Power Plants," October 1980.
10 i
11 In 1988, A NUMARC/NESP project was initiated to re-evaluate the emergency action levels (EALs) in 12 the context of utility operating experience. At that time, the nuclear utility industry had over ten years 13 of experience in adapting the NRC guidelines to specific plant configurations, using them both in 14 exercises and under actual emergency conditions. As a result, a number of improvements had been t
15 identified as NUREG-0654, Appendix 1 guidelines had been applied in the development of plant 16 EALs.
I 17 18 The NUMARC/NESP EAL Task Force developed a systematic approach and supporting basis for 19 EAL development. This methodology developed a set of generic EAL guidelines, together with the 20 basis for each, such that they could be used and adapted by each utility on a consistent basis. The 21 review uf the industry's experiences with EALs, in conjunction with regulatory considerations, was 22 applied directly to the development of this generic set of EAL guidelines. The generic guidelines were 23 intended to clearly define conditions that represent increasing risk to the public and can give 24 consistent classifications when applied at diffarent sites. The NUMARC/NESP-007 document resulted 9.5 from that effort. The draft NUMARC/NESP-O')7 methodology was reviewed by individuals from the
.6 industry, independent of the task force, was submitted to the entire industry for review, was exercised 27 in a table top exercise with the NRC, underwant a regulatory analysis by the NRC, was published for 28 public comment in the Federal Register, and was endorsed by the NRC as an acceptable altemative 29 to the guidance in NUREG-0654 in Revision 3 to Regulatory Guide 1.101, " Emergency Planning and i
30 Preparedness for Nuclear Power Reactors". The methodology was presented to the industry in a 31 workshop conducted in St Louis in September 1992.
32 33 Close to the end of the process described above, concems developed regarding the classification of 34 events which occur during periods of plant shutdowns and refueling. Industry experience had shown 35 that plants could be susceptible to a variety of events that could challenge safety during shutdown 36 operations. While these events had neither posed nor indicated an undue risk to public health and 37 safety, they did indicate the need to consider emergency action levels applicable during shutdown 38-modes. Since the issue was still under evaluation, shutdown EALs were not included in Revision 2, 39 but were deferred to a later revision of NUMARC/NESP-007. A special task force was formed to 40 address this issue and draft shutdown EALs were prepared in conjunction with efforts of the 41 NUMARC Shutdown Plant issues Working Group to coordinate industry activities relating to 42 shutdown safety.
43 44 As utilities implemented the NUMARC/NESP-007 areas of possible improvement were identified. In 45 addition, the staff of the NRC provided suggestions for improvement based on their review of utility 46 submittals. A task force was assembled to incorporate the implementation experiences. NEl 97-03, 47 Revision 3, is the successor to NUMARC/NESP-007.
48 v
Revision 100998 1
The guidance presented here is not intended to be applied to plants a=-is. It is intended to give the 2
user the logic for developing site-specific EALs (i.e., instrument readings, etc.) using site-specific EAL 3
presentation methods (formats). Basis information is provided to aid station personnel in preparation 4
of their own site-specific EALs, to provide necessary information for training, and for explanation to 5
state and local officials. In addition, state and local requirements have not been reflected in the 6
generic guidance and should be considered on a case-by-case basis with appropriate state and local 7
emergency response organizations. It is important that the NEl EALs be treated as an integrated 8
package. Selecting only portions of this guidance for use in developing site-specific EALs could lead 9
to inconsistent or incomplete EALs.
10 11 Although the basic concerns with barrier integrity and the major safety problems of nuclear power 12 plants are similar across plant types, design differences will have a substantial effect on EALs. The 13 major differences are found between a BWR and a PWR. In these cases, EAL guidelines unique to '
14 BWRs and PWRs must be specified Even among PWRs, however, there are substantial differences 15 in design and in types of containment used. There is enough commonality among plants that many 16 ICs will be the same or very similar. However, others will have to match plant features and safety.
17 system designs that are unique to the plant type or even to the specific plant. The EAL Task Force 18 believes that there is sufficient information provided in the basis of the EALs to allow the EALs to be 19 implemented at plants from all NSSS LWR vendors. However, this generic guidance is not 20 considered to be applicable to advanced LWR designs, or to decommissioned facilities, or to away 21 from site radioactive material storage facilities.
22 23 The original EAL Task Force identified eight characteristics that were to be incorporated into model t
24' EALs. Experience to date has shown these considerations to be VALID. These were:
25 26 (1)
Consistency (i.e., the EALs would lead to similar decisions under similar circumstances at 27 different plants);
i 28 29 (2)
Human engineering and user friendliness; 30 31 (3)
Potential for classification upgrade only when there is an increasing threat to public health 32 and safety; 33 34 (4)
Ease of upgrading and downgrading; 35 36 (5)
Thoroughness in addressing, and disposing of, the issues of completeness and accuracy 37 raised regarding NUREG-0654, Appendix 1; 38 39 (6)
Technical completeness and appropriateness for each classification level; 40 41 (7)
A logical progression in classification for combinations of multiple events; 42 43 (8)
Objective, observable values.
44 45 Based on the information gathered and reviewed, the Task Force has developed generic EAL
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46 guidance. Because of the wide variety of presentation methods (formats) used at different utilities,
- 47. the Task Force believes that specifying guidance as to what each IC and EAL should address, and 48 including sufficient basis information for each EAL will best assure uniformity of approach. The 49 information is presented by Recognition Category:
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A - Abnormal Rad Levels / Radiological Effluent 2
3 F - Fission Product Barrier Degradation 4
5-H - Hazards and Other Conditions Affecting Plant Safety
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S - System Malfunction 4
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Each of the EAL guides in Recognition Categories A, H, and S is structured in the following way:
10 11 Recognition Category - As described above.
12 1
13-Emergency Class - NOUE, Alert, Site Area Emergency or General Emergency.
14 15 i
Initiating Condition - Symptom-or Event-Based, Generic identification and Title.
16 17 Operating Mode Applicability - Power Operation, Hot Standby, Hot Shutdown, Cold Shutdown,
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18 Refueling, Defueled or All.
19 20 Example Emergency Action Level (s) corresponding to the IC.
4 21 22 Basis information for plant-specific readings and factors that may relate to changing the generic 23 IC or EAL to a different emergency class, such as for Loss of All AC Power.
24 l
'25 For Recognition Category F, the EAL information is presented in a matrix format. The presentation 26 method was chosen to clearly show the synergism among the EALs and to support more accurata 27 dynamic assessments. For category F, the EALs are arranged by safety function, or fission product 28 barrier. Classifications are based on various combinations of function or barrier challenges.
29 30 The EAL Guidance has the primary threshold for NOUE as operation outside the safety envelope for 31 the plant as defined by plant technical specifications, including LCOs and Action Statement Times. In 32 addition, certain precursors of more serious events such as loss of offsite AC power and earthquakes 33 are included in NOUE EALs. This provides a clear demarcation between the lowest emergency class 34 and "non-emergency" notifications specified by 10 CFR 50.72.
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Revision 100998 1-2 ACRONYMS 3
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AC Altemating Current 6
AEOD NRC Office for Analysis and Evaluation of Operational Data 7
ATWS Anticipated Transient Without Scram
- 8 B&W Babcock and Wilcox 9~
BWR Boiling Water Reactor 10 CCW Component Cooling Water -
11 CE Combustion Engineering
.12 CFR_
Code of Federal Regulations 13 CMT Containment 14:
CSF Critical Safety Function 15 CSFST
. Critical Safety Function Status Tree 16 DC Direct Current 17.-
-DHR Decay Heat Removal 18, DOT Department of Transportation 19 EAL-Emergency Action L6'rel 20 ECCS
, Emergency Core Cooling System 21
-ECL Emergency Classification Level 22 EOF Emergency Operations Facility L
23 ~
EOP Emergency Operating Procedure l
24.
EPA.
Environmental Protection Agency 25 EPG Emergency Procedure Guideline 26 EPIP Emergency Plan Implementing Procedure 27 EPRI Electric Power Research Institute 28-ERG Emergency Response Guideline -
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.ESF Engineered Safeguards Feature l-
'30 ESWL Emergency Service Water i
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FEMA Federal Emergency Management Agency 32 FSAR Final Safety Analysis Report 33' GE General Electric 34 HPCI High Pressure Coolant injection 35 HPS! ~
High Pressure Safety injection-36 IC Initiating Condition
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ACRONYMS (continued) 2 3_
4-LCO Limiting Condition of Operation 5
LER Licensee Event Report 6
-LOCA Loss of Coolant Accident 7
LPSI-Low Pressure Safety injection
'8-MSIV Main Steam isolation Valve 9
mR millirem 10-Mw Megawatt 11:
NEl Nuclear Energy Institute 12 NESP.
National Environmental Studies Project 13 NRC Nuclear Regulatory Commission 14 NSSS Nuclear Steam Supply System -
1 15 NOUE Notification Of Unusual Event 1
16 NUMARC Nuclear Management and Resources Council 17-
-OBE Operating Basis Earthquake 18 ODCM Offsite Dose Calculation Manual i
- 19-
.PRA Probabilistic Risk Assessment 20 PWR Pressurized Water Reactor -
21'
- PSIG
- Pounds per Square Inch Gauge 22-R
. Rem 23 RCIC Reactor Core Isolation Cooling 24 RCS-Reactor Coolant System 95-_
RPS Reactor Protection System 6
SBGTS-Stand-By Gas Treatment System 27 SG-Steam Generator 28 SI
. Safety injection 29 SPDS
. Safety Parameter Display System 30 SRO Senior Reactor Operator 31 SSE Safe Shutdown Earthquake i
-32 TSC Technical Support Center
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'34 WOG Westinghouse Owners Group ix
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1 1.0 METHODOLOGY FOR DEVELOPMENT OF EMERGENCY ACTION j
2 LEVELS l
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1.1 BACKGROUND
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7 Nuclear utilities must respond to a formal set of threshold conditions that require plant personnel to i
8 take specific actions with regard to notifying state and local govemments and the public when certain 9
off-normalindicators or events are recognized. Emergency classes are defined in 10CFR50. Levels 10 of response and the conditions leading to those responses are defined in a joint NRC/ FEMA 11 guidelines contained in Appendix 1 of NUREG-0654/ FEMA-REP-1, Rev.1, " Criteria for Preparation l-12 and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear l
13 Power Plants," October 1980.
i 14 15 in 1988, A NUMARC/NESP project was initiated to re-evaluate the emergency action levels (EALs) in 16 the context of utility operating experience. At that time, the nuclear utility industry had over ten years 17 of experience in adapting the NRC guidelines to specific plant configurations, using them both in 18 exercises and under actual emergency conditions. As a result, a number of improvements had been 19 identified as NUREG-0654, Appendix 1. Guidelines have been applied in the development of plant l
20 EALs.
i 21 l
22 The NUMARC/NESP EAL Task Force developed a systematic approach and supporting basis for l
23 EAL development. This methodology developed a set of generic EAL guidelines, together with the 24 basis for each, such that they could be used and adapted by each utility on a consistent basis. The 25 review of the industry's experiences with EALs, in conjunction with regulatory considerations, was l-36 applied directly to the development of this generic set of EAL guidelines. The generic guidelines 27 were intended to clearly define conditions that represent increasing risk to the public and can give 28 consistent classifications when applied at different sites. The NUMARC/NESP-007 document 29 resulted from that effort. The draft NUMARC/NESP-007 methodology was reviewed by individuals 30 from the industry, independent of the task force, was submitted to the entire industry for review, was 31 exercised in a table top exercise with the NRC, underwent a regulatory analysis by the NRC, was 32 published for public comment in the Federal Register, and was endorsed by the NRC as an l
33 acceptable attemative to the guidance in NUREG-0654 in Revision 3 to Regulatory Guide 1.101, 34
" Emergency Planning and Preparedness for Nuclear Power Reauors". The methodology was 35 presented to the industry in a workshop conducted in St Louis in September 1992.
36 37 Close to the end of the process described above, concems developed regarding the classification of 38 events which occur during periods of plant shutdowns and refueling. Industry experience had shown l
39 that plants could be susceptible to a variety of events that could challengo safety during shutdown 40 operations. While these events had neither posed nor indicated an undue risk to public health and 41 safety, they did indicate the need to consider emergency action levels applicable during shutdown 42 modes. Since the issue was still under evaluation, shutdown EALs were not included in Revision 2, 43 but were deferred. Guidance which addresses shutdown, defueled, and long term storage EALs will 44 be issued as part of_NEl 99-01. NEl 99-01.will address both NUMARC/NESP-007 and NUREG-0654 l
45 users.
l 46 47 As utilities implemented the NUMARC/NESP-007, areas of possible improvement were identified. In 48 addition, the staff of the NRC provided suggestions for improvement based on their review of utility i
1.1 L
s 1 - submittals. A task force was assembled to incorporate the implementation. NEl-97-03, Revision 3, is 2 -. the successor to NUMARC/NESP-007.-
t 1
i i
i 9
4 f
5 t
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2.0 CHANGES INCORPORATED WITH REVISION 3 2
3 This section summarizes the more significant changes made to the EAL methodology with Revson3.
4 This is not intended to be a complete tabulation. Numerous editorial changes were made in the 5
interest of clarity and/or consistent formatting. These changes are not tabulated herein.
G 7
2.1 Section 3.0, Development of Basis for Generic Approach 8
9 New sections were added to make recommendations regarding the (1) classification of transient 10 events, (2) the interface between classification and activation of emergency facilities, and (3) 11 operating mode applicability.
12 13 2.2 Section 4.0, Human Factors Considerations 14 15 No significant changes.
16 17 2.3 Section 5.0, Generic EAL Guidance 18 19 Additional information regarding site-specific implementation was added in response to numerous 20 questions received during utility implementation efforts.
21 22 A set of definitions was added to this section. These words and phrases are defined terms having 23 specific meanings as they relate to the EALs. This terms appear in capital letters in the IC / EALs, 24 and bases. These defined terms were added in Revision 3 to further enhance consistent 25 clsssifications.
26 27 The initiating condition matrices for each recognition category were re-arranged slightly to align event 28 progressions where possible. While the individual ICs are presented in sequence by IC designator, 29 the IC entries in the initiating condition matrices may not be in sequence.
30 31 References to the need to address multi-unit sites has been deleted from the individual ICs and 32 relocated to the generic text at start of Section 5.0.
33 34-2.4 Section 5.0, Recognition Category A 35 36 A significant change in the philosophy of classifying abnormal radiological effluent events was 37 incorporated in Revision 3. In Revision 2, indications on radiation monitors would trigger a dose 38 assessment using real-time meteorology, and the classification would be based on the results of the 39 dose assessment, which was required to be completed within 15 minutes. With Revision 3, this 40 requirement has been deleted as a prerequisite for classification. Appropriate dose assessments are 41 still required to be performed as accident assessments by NUREG-0654, but are no longer required 42 as part of the classification process. If results from dose assessments are available at the time the 43 classification is made, the revised ICs require the Site Area and General Emergency classifications to 44
.be basedsn these results...The largest number of questions received on the Revision 2 methodology 45 were associated wi'.h this recognition category. In order to address these questions and to better 46 explain the basis of the effluent ICs, Appendix A, " Basis for Radiological Effluent initiating Conditions" 47 was added I Revision 3.
48 2.1
.. -.-.~ - -. - -
j 1
Dose quantities implemented with the revisions to 10 CFR 20, such as TEDE, CDE, etc., have been i
2 incorporated in the ICs and EALs.
3 L
4 In AU2 and AA2, guidance as to how level decrease could be detected was added to the basis. The 5
AU1 reference to increase in alioome concentration was deleted. Text was also added to bases to 6
restrict dry storage applicability in AU2 to dry storage licensed under 10 CFR 50, not that licensed 7
under 10 CFR 72.
8 9
The field survey EALs were revised to specify " closed window" redings. This was done to eliminate 10 confusion regarding constructions such as TEDE-rate.
11 l
12 2.5 Section 5.0, Recognition Category F l
13 14 A significant change to the initiating condition matrix was to revise the definition of a Site-Area 15 Emergency. The new definition simply calls for the loss or potential loss of ANY two barriers. The 16 previous definition was found to be unnecessarily complex, and a review of the possible 17 combinations of EALs that could comprise a Site Area Emergency, indicated that the same endpoint l
18 would be reached with the revised definition. For PWRs, this change in definition required some re-l 19 arrangement in the steam generator related EALs.
20 21 For the BWR RCS Barrier EAL for RCS Leak Rate, the word "unisolable" was added to the main
.22 steam line break EAL. If the break was isolable, then the event should not be declared.
23 24 The BWR Containment Barrier EAL for Reactor Vessel Water Level was revised to remove reference 25 to the maximum core uncovery time curve, and replace it with the EAL ' Primary containment flooding 26 required".
1 27 28 The PWR SGTR-related EALs were revised in order to compensate for the revision to the definition i
29 of the combination of barrier challenges that comprise a Site Area Emergency. The changes ensure 30 that a Site Area Emergency will be declared for a event that results in a RUPTURED, FAULTED 31 steam generator. The previous Containment Barrier EAL for primary-to-secondary leakage with 32 secondary leakage greater than technical specification allowable was revised to be applicable for 33 leakrates greater than 10 gpm in the presence of non-isolable steam release from affected SG.
34 l
35 The bases discussions for several of the EALs were updated to address comments and to reflect the l
36 changes identified above.
l 37 1
38 2.6 Section 5.0, Recognition Category H 39 l
40 The ICs HU1 and HA1, Natural Phenomena, were revised to address several comments. In 41 particular, the EALs in HA1 were re-formatted to reflect the logic originally intended. The original 42 intent was that the occurrence of the phenomena of a magnitude sufficient to cause damage would 43 constitute a NOUE. If the event caused plant damage that was indicated by " VISIBLE DAMAGE" or 44.
by plant performance changes observable on control room indications, that escalation to an Alert 45 would be approonste. Because of a foiin.iin,yerror, th!: bgic was lost in HA1:--
46-
~~
~
i l
47 An EAL was added to HU1 and HA1 to address uncontrolled intemal flooding. This EAL was original 1
48. proposed as a new SID IC. However, these events were deemed applicable to power operating 49-modes as well.
- 2. 2 l
1 2
The basis for HU2, Fire, was upgraded to provide guidance as to the starting of the specified "15-3 minute clock" 4
5 Several changes were made to the Security Event EALs to resolve apparent overlap in the EALs for 6
the four classifications. The current EALs are believed to have clearer classification thresholds.
7 Significant in these improvements was the added definitions for several terms, e.g., Hostile Armed 8
Force.
9 10 2.7 Section 5.0, Recognition Category 8 11 12 A new IC, SU8, Inadvertent Criticality, has been added. This IC was originally proposed as a new 13 S/D IC. However, these events were deemed applicable to other operating modes as well.
14 15 Clarifications were made to the ICs and Bases for the loss of annunciator, fuel clad degradation, and 16 protection system failure events.
17 4
- 2. 3
s ;
1 3.0 DEVELOPMENT OF BASIS FOR GENERIC APPROACH 2
i 3
This section addresses several key considerations that were incorporated into the development of 4
the original NUMARC/NESP EALs. An understanding of these considerations will facilitate the 5
implementation of this generic guidance into site-specific programs.
In prior revisions to this 6
document, this section described the process by which the Task Force identified and resolved these 7
considerations. Since much of this was deemed to be historical in nature, it has been removed from 8
this revision.
9 10 Literature reviews, review of plant-specific EALs, and on-site utility interviews were performed as 11 preparation for the drafting of the generic guidance. The review led to the conclusion that the current 12 regulatory structure was not an impediment to the development of the appropriate EALs. Rather, the 13 detailad guidance currently in place could be enhanced.
14 15 The generic guidance provided in this document is intended to address radiological emergency l
16 preparedness. Non-radiological events are included in the classification scheme only to the extent 17 that these events represent challenges to the continued safety of the reactor plant and its operators.
18 There are existing reporting requirements (EPA, OSHA) under which utilities operate. There are also 19 requirements for emergency preparedness involving hazardous chemical releases. While the 20 proposed classification structure could be expanded to include these non-radiological hazards, these 21 events are beyond the scope of this document.
22 23 This classification scheme is based on the four classification levels promulgated by the NRC as the 24 standard for the United States. This scheme is different from the intemational severity scale, which is 25 not addressed in this generic guidance. The NRC has determined that US nuclear faciHties would 26 continue to classify events using the four classification levels and that the NRC would re-classify the 27 event in any intemational communication.
28 29 3.1 Regulatory Context 30 31 Title 10, Code of Federal Regulations, Part 50 provides the regulations that govem emergency 32 preparedness at nuclear power plants. Nuclear power reactor licensees are required to have 33 NRC-approved " emergency response plans" for dealing with " radiological emergencies." The 34 requirements call for both onsite and offsite emergency response plans, with the offsite plans being 35 those approved by FEMA and used by the State and local authorities. This document deels with the 36 utilities' approved onsite plans and procedures for response to radiological emergencies at nuclear 37 power plants, and the links they provide to the offsite plans.
38 39 Section 50.47 of Title 10 of the Code of Federal Regulations (10 CFR 50.47), entitled " Emergency 40 Plans," states the requirement for such plans. Part (a)(1) of this regulation states that "no operating 41 license will be issued unless a finding is made by NRC that there is reasonable assurance that 42 adequate protective measures can and will be taken in the event of a radiological emergency."
43 44 The major portion of 10 CFR 50.47 lists " standards" that emergency response plans must meet. The 45 standards constitute a detailed list of Rems to be addressed in the plans. Of padicular importance to 46 this project is the fourth standard, which addresses " emergency classification" and " action levels."
47 These terms, however, are not defined in the regulation.
48 49 10 CFR 50.54, " Conditions of licenses," emphasizes that power reactor licensees must " follow, and 50 maintain in effect, emergency plans which meet the standards in Part 50.47(b) and the requirements 3.1
i 1
1 in Appendix E to this pirt." Th3 rem ind:r of this part d::Is primarily with required implementation l
2 dates.
l 3
(
4 10 CFR 50.54(q) allows licensees to make changes to emergency plans without prior Commission 5
approval only if: (a) the changes do not decrease the effectiveness of the plans and (b) the plans, as l
6 changed, continue to meet 10 CFR 50.47(b) standards and 10 CFR 50 Appendix E requirements.
l 7
The licensee must keep a record of any such changes. Proposed changes that decrease the 8
effectiveness of the approved emergency plans may not be implemented without application to and i
9 approval by the Commission.
10 11 10 CFR 50.72 deals witn "Immediate notification requirements for operating nuclear power reactors."
12 The "immediate" notification section actually includes three types of reports: (1) immediately after 13 notification of State or local agencies (for emergency classification events); (2) one-hour reports; and, 14 (3) four-hour reports.
15 16 Although 10 CFR 50.72 contains significant detail, it does not define either " Emergency Class" or 17
" Emergency Action Level." But one-hour and four-hour reports are listed as "non-emergency events,"
18 namely, those which are "not reported as a declaration of an Emergency Class." Certain 19 10CFR 50.72 events can also meet the Notification of Unusual Event emergency classification if they 20 are precursors of more serious events. These situations also warrant anticipatory notification of state 21 and local officials. (See Section 3.7, " Emergency Class Descriptions".)
22 23 By footnote, the reader is directed from 10 CFR 50.72 to 10 CFR 50 Appendix E, for information 24 conceming " Emergency Classes."
25 26 10 CFR 50.73 describes the " Licensee event report system," which requires submittal of follow-up 27 written reports within thirty days of required notification of NRC.
28 29 10 CFR 50 Appendix E, Section B, " Assessment Actions," mandates that emergency plans must 30 contain " emergency action levels" EALs are to be described for: (1) determining the need for 31 notification and participation of various agencies, and (2) determining when and what type of 32 protective measures should be considered. Appendix E continues by stating that the EALs are to be 33 based on:(1)in-plant conditions;(2)in-plantinstrumentation;(3) onsite monitoring; and (4) offsite 34 monitoring.
35 36 10 CFR 50 Appendix E, Section C, " Activation of Emergency Organization," also addresses 37
" emergency classes" and " emergency action levels." This section states that FALs are to be based 38 on: (1) onsite radiation monitoring information; (2) offsite radiation monitoring information; and, 39 (3) readings from a number of plant sensors that indicate a potential emergency, such as 40 containment pressure and the response of the Emergency Core Cooling System. This section also 41 states that " emergency classes" shall include: (1) Notification of NOUEs, (2) Alert, (3) Site Area 42 Emergency, and (4) General Emergency.
43 44 These regulations are supplemented by various regulatory guidance documents. A significant 45 document that has dealt specifically.with EALs is NUREG-0654/ FEMA-REP-1, " Criteria for 46 Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in l
47 Support of Nuclear Power Plants," October 1980.
l 3.2
1 2
3.2 Definitions Used in Develop!ng EAL Methodology 3_
4 Based on the above review of regulations, review of common utility usage of terms, discussions 5
among Task Force members, and existing published information, the following definitions apply to the 6
generic EAL methodology:
7 8
EMERGENCY CLASS: One of a minimum set of names or titles, established by the Nuclear 9
Regulatory Commission (NRC), for grouping off-normal nuclear power plant conditions 10 according to (1) their relative radiological seriousness, and (2) the time-sensitive onsite and 11 off-site radiological emergency preparedness actions necessary to respond to such 12 conditions. The existing radiological emergency classes, in ascending order of seriousness, 13 are cal'ad:
14 15
. Notification of Unusual Event 16 17 Alert 18 19 Site Area Emergency 20 21 General Emergency 22 23
!NITIATING CONDITION (IC): One of a predetermined subset of nuclear power plant 24 conditions where either the potential exists for a radiological emergency, or such an 25 emergency has occurred.
26 27 Discussion:
28 29 In NUREG-0654, the NRC introduced, but does not ueline, the term " initiating condition."
30 Since the term is commonly used in nuclear power plant emergency planning, the definition 31 above has been developed and combines both regulatory intent and the greatest degree of 32 common usage among utilities.
33 34 Defined in this manner, an lC is an emergency condition which sets'it apart from the broad 35 class of conditions that may or may not have the potential to escalate into a radiological 36 emergency. It can be a continuous, measurable function that is outside technical 37 specifications, such as elevated RCS temperature or falling reactor coolant level (asymplom). It 38 also encompasses occurrences such as FIRE (an event) or reactor coolant pipe failure (an 39 event or a barrier breach).
40 41 EMERGENCY ACTION LEVEL (EAL): A pre-determined, site-specific, observable threshold 42 for a plant initiating Condition that places the plant in a given emergency class. An EAL can 43 be: an instrument reading; an equipment status indicator; a measurable parameter (onsite or 44 offsite); a discrete, observable event; results of analyses; entry into specific emergency 45 operating procedurespor another phenomenon which41taccurs, indicates entry into a l
46 particular emergency class'.
l 47 48 Discussion:
49-3.3 l
l
i 1
The ttrm "cmtrg:ncy cction level" has been defined by example in the regulations, as noted 2
in the above discussion conceming regulatory background. The term had not, however, been l
3 defined operationally in a manner to address all contingencies.
4 5
There are times when an EAL will be a threshold point on a measurable continuous function, 6
such as a primary system coolant leak that has exceeded technical specifications for a
[
7 specific plant.
4 8
9 At other times, the EAL and the IC will coincide, both identified by a discrete event that places i
10-the plant in a particular emergency class. For example, " Train Derailment Onsite" is an 4
11 example of an "NOUE"IC in NUREG-0654 that also can be an event-based EAL.
j 12
]
13' 3.3 Differences In Perspective j
14 15 The purpose of this effort is to define a methodology for EAL development that will better assure a 16 consistent emergency classification commensurate with the level of risk. The approach must be 4
17 easily understood and applied by the individuals responsible for onsite and offsite emergency 18 preparedness and response. In order to achieve consistent application, this recommended i
19 methodology must be accepted at all levels of application (e.g., licensed operators, health physics 20 personnel, facility managers, offsite emergency agencies, NRC and FEMA response organizations, 21 etc.).
22 23 Commercial nuclear facilities are faced.with a range of public service and public acceptance 24 pressures. It is of utmost importance that emergency regulations be based on as accurate an 25 assessment of the risk as possible. There are evident risks to health and safety in understating the 26 potential hazard from an event. However, there are both risks and costs to alerting the public to an 27 emergency that exceeds the true threat. This is true at all levels, but particularly if evacuation is 28 recommended.
29 30 3.4 Recognition Categories 31 32 ICs and EALs can be grouped in one of several schemes. This generic classification scheme 33 incorporates symptom-based, event-based, and barrier-based ICs and EALs.
34 35 The symptom-based category for ICs and EALs refers to those indicators that are measurable over 36 some continuous spectrum, such as core temperature, coolant levels, containment pressu.re, etc.
37 When one or more of these indicators begin to show off-normal readings, reactor operators are 38 trained to identify the probable causes and potential consequences of these " symptoms" and take 39 corrective action. The level of seriousness indicated by these symptoms depends on the degree to 40 which they have exceeded technical specifications, the other symptoms or events that are occurring 41 contemporaneously, and the capability of the licensed operators to gain control and bring the 42 indicator back to safe levels.
43 44 Event-based EALs and ICs refer to occurrences with potential safety significqnce, such as the failure 45 of a high-pressure safety injection pump, a safety valve failure, or a loss of electric power to some 46 part of the plant. The range of seriousness of these " events"is dependent on the location, number of 47 contemporaneous events, remaining plant safety margin, etc.
48 49 Barrier-based EALs and ICs refer to the level of challenge to principal barriers used to assure 50 containment of radioactive materials contained within a nuclear power plant. For radioactive materials 51' that are contained within the reactor core, these barriers are: fuel cladding, reactor coolant system 3.4
1 pressure boundary, and containment. The level of challenge to these barriers encompasses the 2
extent of damage (loss or potential loss) and the number of barriers concurrently under challenge. In 3
reality, barrier-based EALs are a subset of symptom-based EALs that deal with symptoms indicating 4
fission product barrier challenges. These barrier-based EALs are primarily derived from Emergency 5
Operating Procedure (EOP) Critical Safety Function (CSF) Status Tree Monitoring (or their 6
equivalent). Challenge to one or more barriers generally is initially identified through instrument 7
readings and periodic sampling. Under present barrier-based EALs, deterioration of the reactor 8
coolant system pressure boundary or the fuel clad barrier usually indicates an " Alert" condition, two 9
barriers under challenge a Site Area Emergency, and loss of two barriers with the third barrier under 10 challenge is a General Emergency. The fission product barrier matrix described in Section 5-F is a 11 hybrid approach that recognizes that some events may represent a challenge to more than one 12 barrier, and that the containment barrier is weighted less than the reactor coolant system pressure 13 boundary and the fuel clad barriers.
14 15 Symptom-based ICs and EALs are most easily identified when the plant is in a normal startup, 16 operating or hot shutdown mode of operation, with all of the barriers in place and the plant's 17 instrumentation and emergency safeguards features fully operational as required by technical 18 specifications. It is under these circumstances that the operations staff has the most direct 19 information of the plant's systems, displayed in the main control room. As the plant moves through 20 the decay heat removal process toward cold shutdown and refueling, barriers to fission products are 21' reduced (i.e... reactor coolant system pressure boundary may be open) and fewer of the safety 22 systems required for power operation are required to be fully operational. Under these plant 23 operating modes, the identification of an IC in the plant's operating and safety systems becomes 24 more event-based, as the instrumentation to detect symptoms of a developing problem may not be 25 fully effective; and engineered safeguards systems, such as the Emergency Core Cooling System 26 (ECCS), are partially disabled as permitted by the plant's Technical Specifications.
27 28 Barrier-based ICs and EALs also are heavily dependent on the ability to monitor instruments that 29 indicate the condition of plant operating and safety systems. Fuel cladding integrity and reactor 30 coolant levels can be monitored through several indicators when the plant is in a normal operating 31 mode, but this capability is much more limited when the plant is in a refueling mode, when many of 32 these indicators are disconnected or off-scale. The need for this instrumentation is lessened, 33 however, and attemate instrumentation is placed in service when the plant is shut down.
l 34 35 It is important to note that in some operating modes there may not be definitive and unambiguous 36 indicators of containment integrity available to control room personnel. For this reason, barrier-based 37 EALs should not place undue reliance on assessments of containment integrity in all operating i
38 modes. Generally, Technical Specifications relax maintaining containment integrity requirements in l
39 modes 5 and 6 in order t'o provide flexibility in performance of specific tasks during shutdown 40 conditions. Containment pressure and temperature indications may not increase if there is a pre-41 existing breach of containment integrity. At most plants, a large portion of the containment's exterior 42 cannot be monitored for leakage by radiation monitors.
43 44 Several categories of emergencies have no instrumentation to indicate a developing problem, or the A5 atyent may ha idantifiad.before any,other indications.nra racnnnizad A reactor _coolanLpipe.could 46 break; FIRE alarms could sound; radioactive materials could be released; and any number of other 47 events can occur that would place the plant in an emergency condition with little waming. For 48 emergencies related to the reactor system and safety systems, the ICs shift to an event based 49 scheme as the plant mode moves toward cold shutdown and refueling modes. For non-radiological 50 events, such as FIRE, extemal floods, wind loads, etc., as described in NUREG-0654 Appendix 1, 51 event-based ICs are the norm.
52 3.5
e 1
in miny casts, a combination of sympte,.i, event-and barrier-based ICs will be present as an
-- 2 emergency. develops. In a loss of coolant accident (LOCA), for example:
3 4
Coolant level is dropping; (symptom) 5-6 e.
There is a. leak of some magnitude in the system (pipe break, safety valve stuck open) that 7
exceeds plant capabilities to make up the loss; (barrier breach or event) 8 9
Core (coolant) temperature is rising; (symptom) and 10 11 At some level, fuel failure begins with indicators such as high off-gas, high coolant activity 12 samples, etc. (barrier breach or symptom) 13 14 3.5 Design Differences 15
'16 Although the same basic concems with barrier integrity and the major safety problems of nuclear 17-power plants are similar across plant types, design differences will have a substantial effect on EALs.
j 18 The major differences are found between a BWR and a PWR. In these cases, EAL guidelines unique 19 -to BWRs and PWRs must be specified. Even among PWRs, however, there are substantial 20 differences in design and in types of containment used.
21 22 There is enough commonality among plants that many ICs will be the same or very similar. However, 23 others will have to match plant features and safety system designs that are unique to the plant type 24 or even to the specific plant. The basis for each EAL guideline should supply sufficient information as 25 to what is required for a site-specific EAL.
26 27 3.6 Required Characteristics 28 29 Eight characteristics that should be incorporated into model EALs are identified below:
30
'31 (1)
Consistency (i.e., the EALs would lead to similar decisions under similar circumstances at 32 different plants);
33 34 (2)
Human engineering and user friendliness; 35 36 (3)
Potential for classification upgrade only when there is an increasing threat to public health 37 and safety; 38 39 (4)
Ease of upgrading and downgrading; 40 41 (5)
Thoroughness in addressing, and disposing of, the issues of completeness and accuracy 42 raised regarding NUREG-0654 Appendix 1; 43 44 (6)
Technical completeness for each classification level; 45 46 (7)
A logical progression in classification for multiple events; and 47 48 (8)
Objective, observable values.
49 50 The EAL development procedure pays careful attention to these eight characteristics to assure that 51 all are addressed in the proposed EAL methodology. The most pervasive and complex of the eight is 3.6
j e
1 the :firsWconsistency." The common denominator that is most appropriate for measuring 2
consistency among ICs and EALs is relative risk.' The approach taken in the development of these
.3.
EALs is based on risk assessment to set the boundaries of the emergency classes and assure that 4
all EALs that trigger that emergency class are in the same range of relative risk. Precursor conditions 5
of more serious emergencies also represent a potential risk to the.public and must be appropriately 6-classified.
l l
l f
3.7
l 1
2 3.7-Emergency Class Descriptions i
l.
3 i
l 4
There are three considerations related to emergency classes. These are:
l 5
6 (1)
The potential impact on radiological safety, either as now known or as can be reasonably 7
projected; 8
9 (2)
How far the plant is beyond its predefined design, safety, and operating envelopes; and l
10 l
11 (3)
Whether or not conditions that threaten health are expected to be confined to within the site L
12 boundary.
l 13 14 The ICs deal explicitly with radiological safety impact by escalating from levels corresponding to 15 releases within regulatory limits to releases beyond EPA Protective Action Guideline (PAG) plume 16 exposure levels. In addition, the " Discussion" sections below include offsite dose consequence 17 considerations which were not included in NUREG-0654 Appendix 1.
18 19 NOTIFICATION OF UNUSUAL EVENT: Events are in process or have occurred which 20 indicate a potential degradation of the level of safety of the plant. No releases of radioactive 21 material requiring offsite response or monitoring are expected unless further degradation of l
22 safety systems occurs.
~
23 24 Discussion:
25 26 Potential degradation of the level of safety of the plant is indicated primarily by exceeding 27 plant technical specification Limiting Condition of Operation (LCO) allowable action statement 28 time for achieving required mode change. Precursors of more serious events should also be 29 included because precursors do represent a potential degradation in the level of safety of the
.30 plant. Minor releases of radioactive materials are included. In this emergency class, however, 31 releases do not require monitoring or offsite response (e.g., dose consequences of less than 32 10 millirem).
33 34 ALERT: Events are in process or have occurred which involve an actual or potential 35 substantial degradation of the level of safety of the plant. Any releases are expected to be 36 limited to small fractions of the EPA Protective Action Guideline exposure levels.
37 38 Discussion:
39 40 Rather than discussing the distinguishing features of " potential degradation" and " potential 41 substantial degradation," a comparative approach would be to determine whether increased 42 monitoring of plant functions is warranted at the Alert level as a result of safety system 43 degradation. This addresses the operations staff's need for help, independent of whether an 44 actual decrease in plant safety is determined. This increased monitoring can then be used to 45 better determine the actual plant safety state, whether escalation to a higher emergency class 46 is warranted, er whether de-escalat!on or termination of the emergency class declaration is 47 warranted. Dose consequences from these events are small fractions of the EPA PAG plume l
48 exposure levels, i.e., about 10 millirem to 100 millirem.
l 49 i
50 SITE AREA EMERGENCY: Events are in process or have occurred which involve actual or 51 likely major failures of plant functions needed for protection of the public. Any releases are not j
3.8
a]
1 expected to result in exposure levels which exceed EPA Protective Action Guideline exposure 2
levels except near the site boundary.
3 4
Discussion:
5 6
The discriminator (threshold) between Site Area Emergency and General Emergency is 7
whether or not the EPA PAG plume exposure levels are expected to be exceeded outside the 8
site boundary. This threshold, in addition to dynamic dose assessment considerations 9
discussed in the EAL guidelines, clearly addresses NRC and offsite emergency response 10 agency concerns as to timely declaration of a General Emergency.
11 12 GENERAL EMERGENCY: Events are in process or have occurred which involve actual or 13 imminent substantial core degradation or melting with potential for loss of containment 14 integrity. Releases can be reasonably expected to exceed EPA Protective Action Guideline 15 exposure levels offsite for more than the immediate site area.
I 16 17 Discussion:
18 19 The bottom line for the General Emergency is whether evacuation or sheltering of the general 20 public is indicated based on EPA PAGs, and therefore should be interpreted to include 21 radionuclide release regardless of cause. In addition, it should address concems as to 22 uncertainties in systems or structures (e.g. containment) response, and also events such as 23 waste gas tank releases and severe spent fuel pool events postulated to occur at high 24 population density sites. To better assure timely notification, EALs in this category must 25 primarily be expressed in terms of plant function status, with secondary reliance on dose 26 projection. In terms of fission product barriers, loss of two barriers with potential loss of the n
27 third barrier constitutes a General Emergency.
28 29 3.8 Emergency Class Thresholds 30 31 The most common bases for establishing these boundaries are the technical specifications and 32' setpoints for each plant that have been developed in the design basis calculations and the Final 33 Safety Analysis Report (FSAR).
34 35 For those conditions that are easily measurable and instrumented, the boundary is likely to be the 36 EAL (observable by plant staff, instrument reading, alarm setpoint, etc.) that indicates entry into a 37 particular emergency class. For example, the main steam line radiation monitor may detect high
)
38 radiation that triggers an alarm. That radiation level also may be the setpoint that closes the main 39 steam isolation valves (MSIV) and initiates the reactor scram. This same radiation level threshold, 40 depending on plant-specific parameters, also may be the appropriate EAL for a direct entry into an 41 emergency class.
42 43 in addition to the continuously measurable indicators, such as coolant temperature, coolant levels, 44 leak rates, containment pressure, etc., the FSAR provides indications of the consequences 45 associated with design basis events. Examples would include stem pipe breaks, MSIV malfunctions, 46 and other anticipated events that, upon occurrence, place the plant immediately into an emergency 47 class.
48 49 Another approach for defining these boundaries is the use of a plant-specific probabilistic safety 50 assessment (PSA - also known as probabilistic risk assessment, PRA). PSAs have been completed 51 for several individual plants, but this is by no means comprehensive. There are, however, PSAs that 3.9
- ~.
b i-I have been completed for representative plant types such as is done in NUREG-1150, " Severe
- Accident Risks
- An Assessment for Five Nuclear Power Plants," as well as several other 3
utility-sponsored PSAs. Existing PSAs can be used as a good first approximation of the relevant ICs 4
and risk associated with emergency conditions for existing plants. Generic insights from PSAs and 5
related severe accident assessments which apply to EALs and emergency class determinations are:
6 7
- 1. Core damage frequency at many BWRs is dominated by seque'nces involving prolonged loss of
.8l all AC power. In addition, prolonged loss of all AC power events are extremely important at 9'
PWRs.' This.would indicate that should this occur, and AC power is not restored within 15 10 minutes, entry into the emergency class at no lower than a Site Area Emergency, when the plant 11.
was initially at power, would be appropriate. This implies that precursors to loss of all AC power 12' events should appropriately be included in the EAL structure.
13 14
- 2. For severe core damage events, uncertainties exist in phenomena important to accident 15 progressions leading to containment failure. Because of these uncertainties, predicting 16 containment integrity may be difficult in these conditions. This is why maintaining containment 17 integrity alone following sequences leading to severe core damage may be an insufficient basis 18.
for not escalating to a General Emergency.
19 20
- 3. A review of four full-scope PRAs (3 PWR,1 BWR) showed that leading contributors to latent 21 fatalities were containment bypass, large LOCA with early containment failure, station blackout 22 greater ttian 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (e.g., LOCA consequences of Station Blackout), and reactor coolant pump 23 seal failure. This indicates that generic EAL methodology must be sufficiently rigorous to cover 24 these sequences in a timely fashion.
25 26 Another critical element of the analysis to arrive at these threshold '(boundary) conditions is the time 27-that the plant might stay in that condition before moving to a higher emergency class. In particular, 28 station blackout coping analyses performed in response to 10 CFR 50.63 and Regulatory Guide 29 1.155, " Station Blackout," may be used to determine whether a specific plant enters a Site Area 30 Emergency or a General Emergency directly, and when escalation to General Emergency is 31 indicated. The time dimension is critical to the EAL since the purpose of the emergency class for 32 state and local officials is to notify them of the level of mobilization that may be necessary to handle 33 the emergency. This is particularly true when a " Site Area Emergency" or " General Emergency" is 34 imminent. Establishing EALs for such conditions must take estimated evacuation time into 35 consideration to minimize the potential for the plume to pass while evacuation is underway.
36 37 Regardless of whether or not containment integrity is challenged, it is possible for significant 38 radioactive inventory within containment to result in EPA PAG plume exposure levels being 39 exceeded even assuming containment is within technical specification allowable leakage rates. With 40 or without containment challenge, however, a major release of radioactivity requiring offsite 41 protection actions from core damage is not possible unless a major failure of fuel cladding allows 42 radioactive material to be released from the core into the reactor coolant. NUREG-1228, " Source 43 Estimations During Incident Response to Severe Nuclear Power Plant Accidents," indicates that such 44 conditions do not exist when the amount of clad damage is less than 20%.
45 46 3.9 Emergency Action Levels 47 48 With the emergency classes defined, the thresholds that must be met for each EAL that is to be 49 placed under the emergency class can be determined. There are two basic approaches to 50 determining these EALs. EALs and emergency class boundaries coincide for those continuously 51 measurable, instrumented ICs, such as radioactivity, core temperature, coolant levels, etc. For these 3.10
1 ICs, the EAL will be the threshold reading that most closely corresponds to the emergency class 2
description using the best available information.
3 4
For discrete (discontinuous) events, the approach will have to be somewhat different. Typically, in 5
this category are intemal and extemal hazards such as FIRE or earthquake. The purpose for 6
including hazards in EALs is to assure that station personnel and offsite emergency response 7
organizations are prepared to deal with consequential damage these hazards may cause. If, indeed, 8
hazards have caused damage to safety functions or fission product barriers, this should be confirmed 9
by symptoms or by observation of such failures. Therefore, it appropriate to enter an Alert status for 10 events approaching or exceeding design basis limits such as Operating Basis Earthquake, design 11 basis wind loads, FIRE within VITAL AREAS, etc. This would give the operating staff additional 12 support and improved ability to determine the extent of plant damage unless damage to barriers or 13 challenges to Critical Safety Functions (CSFs) have occurred or are identified, then the additional
-14 support can be used to escalate or terminate. The Emergency Class could be escalated or 15 terminated based on what is then found. Of course, security events must reflect potential for 16 increasing security threat levels.
17 18 Plant emergency operating procedures (EOPs) are designed to maintain and/or restore a set of 19 CSFs which are listed in the order of priority for restoration efforts during accident conditions. While 20 the actual nomenclature of the CSFs may vary among plants, generally the PWR CSF set includes:
21 Subcriticality 22 23 Core cooling 24 Heat sink Pressure-temperature-stress (RCS integrity) 25 Containment 26 27 RCS inventory 28 29 There are diverse and redundant plant systems to support each CSF. By monitoring the CSFs 30 instead of the individual system component status, the impact of multiple events is inherently 31 addressed, e.g., the number of operable components available to maintain the function.
32 33 The EOPs contain detailed instructions regarding the monitoring of these functions and provides a 34 scheme for classifying the significance of the challenge to the functions. In providing EALs based on 35 these schemes, the emergency classification can flow from the EOP assessment rather than being 36 based on a separate EAL assessment. This is desirable as it reduces ambiguity and reduces the 37 time necessary to classify the event.
38 39 As an example, consider that the Westinghouse Owner's Group (WOG) Emergency Response 40 Guidelines (ERGS) classify challenges as YELLOW, ORANGE, and RED paths. If the core exit 41 thermocouples exceed 1200 degrees F or 700 degrees F with low reactor vessel water level, a RED 42 path condition exists. The ERG considers a RED path as "... an extreme challenge to a plant function 43-necessary for the protection of the public..." This is almost identical to the present NRC NUREG-44 0654 description of a site area emergency "... actual or likely failures of plant functions needed for 45 -the protet;f N ~^%dt-reasonably follows,that4f-any-CSF enters-a-RED pathr+ site area 46 emergency exists. A general emergency could be considered to exist if core cooling CSF is in a RED 47 path and the EOP function restoration procedures have not been successful in restoring core cooling.
l 3.11 i
v 1
2 3.10 Treatment Of Multiple Events And Emergency Class Upgrading 3
4 The above discussion deals primarily with simpler emergencies and events that may not escalate 5
rapidly. However, usable EAL guidance must also consider rapidly evolving and complex events.
.6 Hence, emergency class upgrading and consideration of multiple events must be addressed.
7 8
There are three approaches presently in use for covering multiple events and emergency class 9
upgrading. These approaches are:
10 11 (U1)
Multiple contemporaneous events are counted and are the basis for escalating to a higher 12 emergency class. For example, two or more contemporaneous Alerts escalate to a Site Area 13 Emergency, l
14 15 (U2)
The emergency class is based on the highest EAL reached. For example, two Alerts remain 16 in the Alert category. Or, an Alert and a Site Area Emergency is a Site Area Emergency.
17 18 (U3)
Emergency Director judgment. Although all emergency classifications require judgment, some 19 utilities rely on Emergency Directorjudgment with little or no additional explicit guidance.
20 21 An additiona! approach for plants with PRAs is to make use of event tree analysis to define 22 combinations of events which lead to equivalent risks. Such event sequences should have an equal 23 emergency classification assigned. However, the chief drawback to this approach as well as (U1) 24 above, is that multiple events may be masked when they actually occur. Further, for plants using 25 symptom-based (and barrier-based) emergency procedures, direct perception of multiple events is 26 unnecessary.
27
'28 Emergency class upgrading for multi-unit stations with shared safety-related systems and functions 29 must also consider the effects of a loss of a common system on more than one unit (e.g. potential for 30 radioactive release from more than one core at the same site). For example, many two-unit stations 31 have their control panels for both units in close proximity within the same room. Thus, control room 32 evacuation most likely would affect both units. There are a number of other systems and functions 33 which may be shared at a given multi-unit station. This must be considered in the emergency class 34 declaration and in the development of appropriate site-specific ICs and EALs based on the generic 35 EAL guidance.
36 37 Although the majority of the EALs provide very specific thresholds, the Emergency Director must 38 remain alert to events or conditions that lead to the conclusion that exceeding the EAL threshold is 39 imminent. If, in the judgment of the Emergency Director, an imminent situation is at hand, the 40 classification should be made as if the thresholds has been exceeded. While this is particularly 41 prudent at the higher emergency classes (as the early classification may provide for more effective 42 implementation of protective measures), it is nonetheless applicable to all emergency classes.
43 44 RECOMMENDATION:
45 46 The best approacn is (U2) above with appropriate consideration for Emergency 47 Director judgment EALs. Properly structured EALs on a fission product barrier basis 48 and which include equivalent risk, will appropriately escalate multiple events to a 49 higher emergency class. For example, common cause failures such as loss of ultimate 50 heat sink or loss of all AC power, will result in multiple contemporaneous symptoms 51 indicating safety system functional failures and increasing challenge to fission product 3.12
a 1
barriers. It is the existence of these symptoms (barrier challenges) that escalate the 2
emergency class, whether there are one or multiple causes.
3 4
3.11 Emergency Class Downgrading 5
6 Another important aspect of usable EAL guidance is the consideration of what to do when the risk 7
posed by an emergency is clearly decreasing. There are several approaches presently in use for 8
. emergency class downgrading. These approaches are:
9 10 (D1)
Terminate the emergency class declaration.
11 l
12 (D2) ' Recovery from emergency class.
13 14 (D3)
Combination of downgrading approaches. Many utilities reviewed include the option to 15 downgrade to a lower emergency class. This is consistent with actions called for in 16
-NUREG-0654 Appendix 1. However, these utilities state that their experience more closely
.17 resembles (D1) and (D2) above as practical choices.
18 19 Another approach possible with risk-based EALs is a relatively simple approach for upgrading to a 20 higher emergency class when the risk increases and downgrading when risk decreases. The 21 boundaries for emergency categories are defined in terms of risk in this approach, and discrete 22 events fall into these categories based on risk. This means that within each emergency class, there 23 is uniformity to the relative levels of risk to human health and safety from radiological accidents, j
24 However, this option may not be practical when applied to actual emergencies, especially those 25 involving General Emergencies.
26 27 RECOMMENDATION:
4 28 29 A combination approach involving recovery from General Emergencies and some Site 30 Area Emergencies and termination from NOUEs, Alerts, and certain Site Area 31 Emergencies causing no long-term plant damage appears to be the best choice.
32 Downgrading to lower emergency classes adds notifications but may have merit under 33 certain circumstances.
34 35 3.12 Classifying Transient Events 36 37 For some events, the condition may be corrected before a declaration has been made. For example, 38 an emergency classification is warranted when automatic and manual actions taken within the control 39 room do not result in a required reactor scram. However, it is likely that actions taken outside of the 40 control room will be successful, probably before the Emergency Director classifies the event. The key 41 consideration in this situation is to determine whether or not further plant damage occurred while the 42 corrective actions were being taken. In some situations, this can be readily determined, in other 43 situations, further analyses (e.g., coolant radiochemistry sampling, may be necessary). There are 44 several approaches presently in use for handling transient events. These approaches are:
45 46 (T1)
Classify the event as indicated and terminate the emergency once assessment shows that 47 there were no consequences from the event and other termination criteria are met.
48 49 (T2)
No emergency declaration is made, but the event is reported and notifications are made.
3.13
2.-
1 2
RECOMMENDATION 3
4 Option (T1) is believed to be appropriate for events at higher emergency 1
5 classifications. Option (T2) may be appropriate for events that might have been 6
classified as NOUEs, but might not be sufficient for some events (e.g., ATWS). It is 7
recommended that the program incorporate aspects of both options with examples of i
8 when each would be appropriate. Many of the generic event-based IC's and EAL's have 9
discriminators based on time or magnitude. Generally, if the discriminator is exceeded, 10 the event should be classified. In implementing the generic guidance into site-specific 11 programs, care should be taken to ensure that the ICs and EALs minimize the need for 12 these ad hoc decisions on transient events.
13 14 There may be cases in which a plant condition that exceeded an EAL threshold was not recognized 15 at the time of occurrence, but is identified well after the condition has occurred (e.g., as a result of 16 routine log or record review) and the condition no longer exists. In these cases, an emergency should 17 not be declared. Normal reporting requirements (e.g.,10 CFR 50.72) are applicable in these cases.
18 19 3.13 Interface Between Classification and Activation of Emergency Facilities 20 21 Existing regulations call for the activation of various emergency facilities at different levels of 22 emergency efassification. The intent of activating these facilities is to provide needed support to the 23 on-shift complement. A question often arises, "If I utilize the TSC as a precautionary measure do 1 24 have to declare an Alert emergencyT There are two possible situations:
25 26 The Emergency Director is faced with an event or series of events which individually may not 27 constitute an Alert emergency, but in combination, is causing the Emergency Director with 28 concem over his ability to contend with the situation using his on-shift resources. This should be 29 clearly recognized as a case in which the Emergency Director judgment ICs apply, and the 30 emergency classification is probably warranted.
31 32 The site has received waming of severe weather. Site management deems it prudent to utilize 33 the onsite emergency facilities to ensure the availability of personnel should the weather cause 34 plant damage while personnel travel is hindered. This situation wouldn't warrant an Alert 35 classification unless the severe weather waming was such that damage comparable to an Alert 36 IC was expected.
37 38 RECOMMENDATION 39 40 The key consideration is not the fact that the facilities were utilized, but rather, the 41 reason for that use.
Facilities may be used for events that may not warrant 42 classification of an emergency.
43 44 3.14 Shutdown IC/EALs 45 46 Generic Letter 88-17, Loss of Decay Heat Removal, SECY-91-283, Evaluation of Shutdown and 47 Low Power Risk issues, SECY-93-190, Regulatory Approach to Shutdown and Low-power 48 Operation, NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power 49 Plants in the United States, and NUMARC 91-06, Guidelines for Industry Actions to Assess 50 Shutdown Management, all address nuclear power plant safety issues that are applicable to periods 51-when the plant is shutdown. These evaluations identify a number of variables which significantly 3.14
+,
I affect the probability and consequences of losing decay heat removal capability during shutdown 2
periods. In addition, NUREG--1449 discusses that the need to respond appropriately, including 3
emergency classification and notification, still exists during cold-shutdown and refueling conditions.
l
'4 Both SECY-93-190 and NUREG-1449 have been reviewed and issues concerning shutdown effects 5
on declaring emergencies have been addressed.
6 7
Given the variability of plant configurations (e.g., systems out-of-service for maintenance, 8
containment open, reduced AC power redundancy, time since shutdown) during these periods, the 9
consequences of any given initiating event can vary greatly. For example, a loss of decay heat i
10 removal capability that occurs at the end of an extended outage has less significance than a similar 11 loss occurring during the first week after shutdown. Compounding these events is the likelihood that 12 instrumentation necessary for assessment may also be inoperable.
- 13 14 Guidance which addresses shutdown, defueled, and long term storage EALs will be issued as part of 15 NEl 99-01. NEl 99-01 will address both NUMARC/NESP-007 and NUREG-0654 users.
16 17 3.15 Operating Mode Applicability 18 19 Emergency action levels have typically been written without regard to the operating mode to which 20 they apply. While the applicable operating modes are obvious for some initiating conditions (e.g.,
j 21 failure of the reactor protection system), the situation is not as clear for others.
1 22 23 The plant operating mode that existed at the time that the event occurred, prior to any protective 24 system or operator action initiated in response to the condition, is compared to the mode applicability 25 of the EALs. If an event occurs, and a lower or higher plant. operating mode is reached before the 26 emergency classification can be made, the declaration shall be based on the mode that existed at 27 the time the event occurred.
28 29-For events that occur in Cold Shutdown or Refueling, escalation is via EALs that have Cold 30 Shutdown or Refueling for mode applicability, even if Hot Shutdown (or a higher mode) is entered 31 during any subsequent heat-up. In particular, the Fission Product Barrier Matrix EALs are applicable 32 only to events that occur in Hot Shutdown or higher.
33 34 The modes identified in the EALs were based on the standard t'ech'nical specifications for BWRs and 35 Westinghouse PWRs. To aid in interpreting these modes for PWRsfrom other NSSSs and for plant 36 with non-standard technical specifications, the modes are described below.
i
~-
3.15
~_ -
l r 1
2 3.15.1 BWR Operating Modes 3
L 4
Power Operations (1):
Mode Switch in Run 5
6 Startup (2): Mode Switch in Startup/ Hot Standby or l
7 Refuel (with all vessel head bolts fully tensioned) 8 9
Hot Shutdown (3): Mode Switch in Shutdown, Average 10 Reactor Coolant Temperature >200 F 11
.12 Cold Shutdown (4): Mode Switch in Shutdown, Average 13 Reactor Coolant Temperature s 200 *F 14
-15 Refueling (5):
Mode Switch in Shutdown orRefuel, l
16 and one or more vessel head bolts less than fully tensioned.
' 17 -
18 Defueled (None)
All reactor fuel removed from reactor 19 pressure vessel.
20 l
21 3.15.2 PWR Operating Modes l
22 23 Power Operating (1):
Reactor Power > 5%, Keg 2 24 0.99 25 l
26 Startup (2): Reactor Power s 5%, Kerr 2 0.99 L
27
(
28 Hot Standby (3):
RCS 2 350 F, Keg < 0.99
!29 30 Hot Shutdown (4): 200 *F < RCS < 350 *F, Kerr< 0.99 31 l
32 Cold Shutdown (5): RCS < 200 F, Keg < 0.99 l
33 34 Refueling (6):
One or more vessel head closure bolts 35 less than fully tensioned 36' 37 Defueled (6): All reactor fuel removed from reactor pressure l
38 vessel.
l 39 t
a 3.16 l
c.
i i
1 4.0 HUMAN FACTORS CONSIDERATIONS
?
2 3
4 Some factors that should be considered in determining the method of presentation of EALs:
5 6
Who is the audience (user) for this information? A senior utility executive would likely want 7
information presented differently than a licensed operator. Offsite agencies and the NRC would 8
have entirely different information needs.
9.
The conditions under which the information must be read, understood, and acted upon. Since 10 11 the subject matter here is emergency actions, it is highly likely that the user of the EALs will be 12 under high stress during the conditions where they are required to be used, particularly under l
13 conditions corresponding to Site Area Emergency and General Emergency.
14 15 What is the user's perception as to the importance of the EALs compared to other actions and 16 decisions that may be needed at the same time? To allow a licensed operator to discharge his 17 responsibilities for dealing with the situation and also provide prompt notification to outside 18 agencies, the emergency classification and notification process must be rapid and concise.
19 20 Is the EAL consistent with the user's knowledge of what constitutes an emergency situation?
l 21 22 How much help does the user receive in deciding which EAL and emergency class is involved?
23 An offsite Emergency Director has many more resources immediately at his disposal than the 24 licensed operator (typically, the Shift Supervisor) who has to make'the initial decisions and take 25 first actions.
26 27 Based on review of a number of plants' EALs and associated information, interviews with utility 28 personnel, and a review of drill experience some recommendations can be made.
29 30 4.1 Level Of Integration Of EALs With Plant Procedures 31 32 A rigorous integration of EALs and emergency class determinations into the plant procedure set, 33 although having some benefits, is probably unnecessary. Such a rigorous integration could well 34 make it more difficult to keep documentation up-to-date. However, keeping EALs totally separated 35 from plant procedures and relying on licensed operator or other utility Emergency Director memory 36 during infrequent, high stress periods is insufficient.
37 38 RECOMMENDATION:
39 40 Visual cues in the plant procedures that it is appropriate to consult the EALs is a 41 method currently used by several utilities. This method can be effective when it is tied 42 to appropriate training. Notes in the appropriate plant procedures to consult the EALs 43 can also be used. It should be noted that this discussion is not restricted to only the 44 emergency procedures; alarm. recognition procedures, abnormal operating
~45 Cprocedu er; ent norms! operatitig procedureir that" apply"to~ cold shutdewn and I
46 refueling modes should also be inciuded. In addition, EALs can bo based on entry into 47 particular procedures or existence of perticular Criticai Safety Function conditions.
4.1 l
e 1
2 4.2 Method of Presentation 3
4 A variety of presentation methods are presently in use. Methods range from directly copying 5
NUREG-0654 Appendix 1 language, adding plant-specific indications to clarify NUREG-0654, use of 6
procedure language including specific tag numbers for instrument readings and alarms, deliberate 7
omission of instrument tag numbers, flow charts, critical safety function status trees, checklists, and 8
combinations of the above.
9 10 What is clear, however, is that the licensed operator (typically the Shift Supervisor) is the first user of 11 this information, has the least amount of help in interpreting the EALs, and also has other significant 12 responsibilities to fulfill while dealing with the EALs. Offsite agencies and emergency directors 13 outside the control room to whom responsibilities are tumed over have other resources and advisors 14 available to them that a licensed operator does not when he is first faced with an emergency 15 situation. In addition, as an emergency situation evolves, the operating staff and the health physics 16 staff are the personnel who must first deal with information that is germane to changing the 17 emergency classification (up, down, or out of the emergency class).
j 18
)
19 RECOMMENDATION:
20 21 The method of presentation should be one with which the operations and health 22 physics staff are comfortable. As is the case for emergency procedures, bases for 23 steps should be in a separate (or separable) document suitable for training and for 24 reference by emergency response personnel and offsite agencies. Each nuclear plant 25 should already have presentation and human factors standards as part of its 26 procedure writing guidance. EALs that are consistent with those procedure writing 27 standards (in particular, emergency operating procedures which most closely 28 correspond to the conditions under which EALs must be used) should be the norm for 29 each utility.
30 31 4.3 Symptom-based, Event-bmsd, Or Barrier-based EALs 32 33 A review of the emergency class descriptions provided elsewhere in this document shows that 34 NOUEs and Alerts deal primarily with sequences that are precursors to more serious emergencies or 35 that may have taken a plant outside of its intended operating envelope, but currently pose no danger 36 to the public. Observable indications in these classes can be events (e.g. natural phenomena),
37 symptoms (e.g., high temperature, low water level), or barrier-related (e.g., challenge to fission 38 product barrier). As one escalates to Site Area Emergency and General Emergency, potential 39 radiological impact to people (both onsite and offsite) increases. However, at this point whatever the 40 root cause event (s) leading to the emergency class escalation matter far less than the i1 creased 41 (potential for) radiological releases. Thus, EALs for these emergency classes should be primarily 42 symptom-and barrier-based. It should be noted again, as stated in Section 3.4, that barrier 43 monitoring is a subset of symptom monitoring, i.e., what readings (symptoms) indicate a challenge to
'44 a fission product barrier.
45 46 RECOMMENDATION:
47 48 A combination approach that ranges from primarily event-based for NOUEs to 49 primarily symptom-or barrier-based for General Emergencies is recommended. This 50 is to better assure that timely recognition and notification occurs, that events 4.2
1 occurring during refueling and cold shutdown are appropriately covered, and that 2
multiple events can be effectively treated in the EALs.
4 l
t 4.3
.~
1 1 --
5.0 GENERIC EAL GUIDANCE 2;
i 3 - This section provides generic EAL guidance based on the information gathered and reviewed by the 4
Task Force. Because of the wide rariety of presentation methods'used at different utilities, this 5
document specifies guidance as to what each IC and EAL should address, and including sufficient 6
basis information for each will best assure uniformity of approach. This approach is analogous to 7
reactor vendors' owners groups developing generic emergency procedure guidelines which are 8
converted by each _ utility into plant-specific emergency operating procedures.
Each utility is i
.9
. reminded. however, to review the " Human Factors Considerations" section of this document as part 10 of implementation of the attached Generic EAL Guidance.
11 12 5A
. Generic Arrangement 4
'13
'.14 - ' The information is presented by Recognition Categories:
2 15 i
16 A - Abnormal Rad Levels / Radiological Effluent 17 18
+
+
F - Fission Product Barrier Degradation 19 20-H - Hazards and Other Conditions Affecting Plant Safety 21
]
22 S - System Malfunction 23 l
24: The initiating Conditions for each of the above Recognition Categories A, H, and S are in the order of 25 NOUE, Alert, Site Area Emergency, and General Emergency. For Recognition Category F, the 26 barrier-based EALs are presented in Tables F-1 and F-2 for BWRs and PWRs respectively. For all i
27 Recognition Categories, an initiating Condition matrix versus Emergency Class is first shown.
1 28
. Separate BWR and PWR initiating Condition matrices are not required. The purpose of the IC i-29 matrices is to provide the reader with an overview of how the ICs are logically related under each 30 Emergency Class.
i 31
[
32 Each of the EAL guides in Recognition Categories A, H, and S is structured in the following way:
i 33 2
34 e.
Recognition Category - As described above.
35 -.
36-Emergency Class - NOUE, Alert, Site Area Emergency or General Emergency.
37 38 e
initiating Condition - Symptom-or Event-Based, Generic Identification and Title.
39 40 Operating Mode Applicability - refers to the operating mode (PWRs) or operating condition j
41 (BWRs) during which the IC / EAL is applicable - Power Operation (includes Startup Mode in 14 2 PWRs), Hot Standby (includes Hot Standby / Startup Condition in BWRs), Hot Shutdown, Cold
- 43.
Shutdown, Refueling, Defueled or Ali. These modes are defined in each licensee's technical 44 specifications. The mode classifications and terminology appropriate to the specific facility should 45 be used. See also Section 3.15.
[
46 47 If an IC or EAL includes an explicit reference to a technical specification, and the technical
.48 specification is not applicable because of operating mode, then that particular IC or EAL is also
- not applicable.
- ~~
50 4
- 5.1
~
1 6
'd i
Example Emergency Action Level (s)- these EALs are examples of conditions and indications 1
2 that were considered to meet the criteria of the IC. These examples were not intended to be all 3
encompassing, and some may not apply to a particular facility. Utilities should generally address 4
each example EAL that applies to their site. If an example EAL does not apply because of its 5
wording, e.g., specifies instrumentation not available at the site, the utility should identify other 6
available means for entry into the IC. Ideally, the example EALs used will be unambiguous, 7
expressed in site-specific nomenclature, and be readily discemible from control room 8
instrumentation.
9 Basis - provides information that explains the IC and example EALs. The bases are written to 10 11 assist the personnel implementing the generic guidance into site-specific procedures. Site-12 specific deviations from the IC / EALs should be compared to the Basis for that IC to ensure that
'13 the fundamental intent of the IC and EALs is met. Some bases provide information intended to 14 assist with establishing site-specific instrumentation values. Appendix A provides detailed 15 guidance on implementing the Radiological Effluent ICs.
16 17 For Recognition Category F, basis information is presented in a format consistent with Tables 3 and 18
- 4. The presentation method shown for Fission Product Barrier Function Matrix was chosen to clearly 19 show the synergism among the EALs and to support more accurate dynamic assessments. Other 20 acceptable methods of achieving these goals which are currently in use include flow charts, block 21 diagrams, and checklist tables. Utilities selecting these attemative need to ensure that all possible 22 EAL combinations in the Fission Product Barrier Function Matrix are addressed in their presentation 23 method.
24 25 5.2 Generic Bases 26
~27 - The generic guidance has the primary threshold for NOUEs as operation outside the safety envelope 28 for the plant as defined by plant technical specifications, including LCOs and Action Statement 29 Times. In addition, certain precursors of more serious events such as loss of offsite AC power and 30 earthquakes are included in NOUE IC / EALs. This provides a clear demarcation between the lowest 31 emergency class and "non-emergency" notifications specified by 10 CFR 50.72.
32
^
33 For a number of Alerts, IC / EALs are chosen based on hazards which may cause damage to plant 34 safety functions (i.e., tomadoes, hurricanes, FIRE in plant VITAL AREAS) c'r require additional help 35-directly (control room evacuation) and thus increased monitoring of the plant is warranted. The 36 symptom-basec' and barrier-based IC / EALs are sufficiently anticipatory to address the results of 37 multiple failures, regardless of whether there is a common cause or not. Declaration of the Alert will 38 already result in the manning of the TSC for assistance and additional monitoring. Thus, direct 39 escalation to the Site Area Emergency is unnecessary. Other Alerts that have been specified 40 correspond to conditions which are consistent with the emergency class description.
41 42 The basis for Site Area Emergencies and General Emergencies is primarily the extent and severity of 43 fission product ba rier challenges, based on plant conditions as presently known or as can be 44 reasonably projected.
-45
.m....
.g m.
46 With ~ regard to thi.s,... Hazards Recogrdtion Category'^ the existence of a hazara that represents a 47 potential degradation in the level of cafety of the plant is the basis of NOUE classification. If the 48 hazard results in VISIBLE DAMAGE to plant structures or equipment associated with safety systems 49 or if system performance is affected, the event may be escalated to an alert. The reference to 50 duration or to damage to safety systems is intended only to size the event. Consequential damage 5.2
e I
from such hazzrds, if observed, would be the basis for escalation to Site Area Emergency or General
~
2 Emergency, by entry to System Degradation or Fission Product Barrier IC / EALs.
3 4
-5 6
5.3 Site Specific Implementatiu, 7
8 The guidance presented here is not intended to be applied to plants as-is. The generic guidance is I
9 intended to give the logic for developing site-specific IC / EALs using site-specific IC / EAL 10 presentation methods. Each utility will need to revise the IC / EALs to meet site-specific needs with 11 regard to instrumentation, nomenclature, plant arrangement, and method of presentation, etc. Such 12 revision is expected and encouraged provided that the intent of the generic guidance is retained.
13-Deviations from the intent may be acceptable, but will need to be justified during regulatory review.
14 Items associated with presentation, e.g., format, sequencing of IC / EALs, IC numbering, recognition 15 categories are at the option of the utility.
16 17 The generic guidance includes both ICs and example EALs. It is the intent of this guidance that both 18 be included in the site-specific implementation. Each serves a specific purpose. The IC is intended i
19 to be the fundamental criteria for the declaration, whereas, the EALs are intended to represent 20 unambiguous examples of conditions that may meet the IC. There may be unforeseen events, or
]
21 combinations of events, for which the EALs may not be exceeded, but in the judgment of the 22 Emergency birector, the intent of the IC may be met. While the generic guidance does include 23 Emergency Director judgment ICs, the additional detail in the individual ICs will facilitate 24 classifications over the broad guidance of the ED judgment ICs.
25 26 For sites involving more than one reactor unit, consideration needs to be given to how events 27 involving shared safety functions may affect more than one unit, and whether or not this may be a 28 factor in escalating the event.
29 30 State and local requirements have not been reflected in the generic guidance and should be 31 considered on a case-by-case basis with appropriate state and local emergency response 32 organizations.
j 33 34 Although not a requirement, utilities should consider preparing a basis document, or including basis 35 information with the IC / EALs. The bases provided for each IC will provide a starting point for 36 developing these site-specific bases. This information may assist the Emergency Director in making 37 classifications, particularly those involving judgment or multiple events. The basis information may be 38 useful in training, for explaining event classifications to offsite officials, and would facilitate regulatory 39 review and approval of the classification scheme.
40 1
41 42 5.4 Definitions 43 44 in the IC / EALs, selected words have been set in all capital letters. These words are defined terms 45 having specific meanings as they relate to this procedure. Definitions of these terms are provided 46 below.
47 48 AFFECTING SAFE SHUTDOWN: Event in progress has adversely affected functions that are 49 necessary to bring the plant to and maintain it in the applicable HOT or COLD SHUTDOWN 50 condition. Plant condition applicability is determined by Technical Specification LCOs in effect.
51 5.3
r 1
Example 1: Event causes damage that results in entry into an LCO that requires the plant to 2
- be placed in HOT SHUTDOWN. HOT SHUTDOWN is achievable, but COLD SHUTDOWN is 3
not. This event is not "AFFECTING SAFE SHUTDOWN."
)
-4 Example 2: Event causes damage that results in entry into an LCO that requires the plant to 5.
be placed in COLD SHUTDOWN. HOT SHUTDOWN is achievable, but COLD SHUTDOWN 6
' is not. This event is "AFFECTING SAFE SHUTDOWN."
7 8
BOMB: refers to an explosive device suspected of having sufficient force to damage plant systems or j
9 structures.
10 11 CIVIL DISTURBANCE: is a group of (site-specific #) or more persons violently protesting station 12 ' operations or activities at the site.
13 14 - EXPLOSION: is a rapid, violent, unconfined combustion, or catastrophic failure of pressurized 15 equipment that imparts energy of sufficient force to potentially damage permanent structures,
'16 systems, or components.
17 e
18 EXTORTION: is an attempt to cause an action at the station by threat of force.
I 19 20 FAULTED: (PWRs) in a steam generator, the existence of secondary side leakage that results in an 21 uncontrolled decrease in steam generator pressure or the steam generator being completely 22' depressurized.
23 24 FIRE: is combustion characterized by heat and light. Sources of smoke such as slipping drive belts 25 or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is 26-NOT required if large quantities of smoke and heat are observed.
27 28 HOSTAGE: is a person (s) held as leverage against the station to ensure that demands will be met by 29 the station.
!~
30 1
31' HOSTILE FORCE: one or more individuals who are engaged in a determined assault, overtly or by
- 32 stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing 33 destruction.
34 35 ' INTRUSION / INTRUDER: is a person (s) present in a PROTECTED AREA without authorization.
36 37 NORMAL PLANT OPD APONS: activities at the plant site associated with routine testing, 38 maintenance, or equipnd egoerations, in accordance with normal operating or administrative 39 procedures. Entry into abnormal or emergency operating procedures, or deviation from normal 40 security or radiological controls posture, is a departure from NORMAL PLANT OPERATIONS.
41 42 PROTECTED AREA: is an area which normally encompasses all controlled areas within the security 43 protected area fence (site-specdic).
44 l
45 RUPTURED: - (PWRs) in a steam generator,. existence of primary-to-secondary leakage of a l
46 Enagnitude'dufficient to require or cause a're' actor trip sind safety injection?
~~
~
l 47 48 SABOTAGE: is deliberate damage, mis-alignment, or mis-operation of plant equipment with the 49 intent to render the equipment inoperable.. Equipment found tampered with or damaged due to 50 malicious mischief may NOT meet the definition of SABOTAGE until this determinah is made by
-51 security supervision.
52 5.4 l
l
.e 1 ' SIGNIFICANT TRANSIENT: is an UNPLANNED event involving one or more of the foliowing: (i) 2 automatic turbine runback >25% thermal reactor power, (2) electrical load rejection >25% full 3
electrical load, (3) Reactor Trip, (4) Safety injection Activation, or (5) thermal power oscillations >10%
4 5
STRIKE ACTION: is a work stoppage within the PROTECTED AREA by a Nedy of workers to -
6 enforce compliance with demands made on (site-specific). The STRIKE ACTION must threaten to 7
interrupt NORMAL PLANT OPERATIONS.
8 9
UNPLANNED: a parameter change or an event that is not the result of an intended evolution and 10 requires corrective or mitigative actions.
11 12 VALID: an indication, report, or condition, is considered to be VALID when it is conclusively verified 13 by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by 14 direct observation by plant personnel, such that doubt related to the indicator's operability, the 15 condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for 16 timely assessment.
17 18 VISIBLE DAMAGE:
is damage to equipment or structure that is readily observable without 19 measurements, testing, or analysis. Damage is sufficient to cause concem regarding the continued 20 operability or reliability of affected safety structure, system, or component. Example damage 21 includes: deformation due to heat or impact, denting, penetration, rupture, cracking, paint blistering.
22 Surface blemishes (e.g., paint chipping, scratches) should not be included.
23 24 VITAL AREA: is any area within the PROTECTED AREA which contains equipment, systems, l
25 components, or material, the failure, destruction, or release of which could directly or indirectly i
26 endanger the public health and safety by exposure to radiation (site-specific).
27 I
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v Table 5-A-1 Recognition Category A Abnormal Rad Levels / Radiological Effluent INITIATING CONDITION MATRIX NOUE ALERT SITS AREA EMERGENCY -
GENERAL EMERGENCY AU1 Any UNPLANNED Release of AA1 Any UNPLANNED Release of AS1 Offsite Dose Resulting from an
.AG1 Offsde Dose Resulting from an -
Gaseous or Liquid Radio-Gaseous or Liquid Radioactnrity Actual or imminent Release of Actual orImminent Release of activity to the Environment that to the Environment that Exceeds Gaseous Radioactivity Exceeds Gaseous Radioactnnty Exceeds Exceeds Two Times the Radio-200 Times the Radiological 100 mR TEDE or 500 mR 1000 mR TEDE or 5000 mR logical Effluent Technical EffluentTechnicalSw: :+.as Thyroid CDE for the Actual or Thyrcnd CDE for the Actual or Spoci,cetime for60 Minutes or for 15 Minutes or Longer.
Projected Duration of the Projected Duration of the Longer.
Op. Modes:M Release.
Release Using Actual Op. Modes:M Op. Modes:M M ;2uk,gy.
Op. Modes:M AU2 Unexpected increase in Plant AA3 Release of Radnosctive Material Radiation.
or increases in Radiation Levels Op. Modes:M Within the Facility That Impedes t
Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown Op. Modes M AA2 Damage to irradiated Fuel or Loss of Water Level that Has or WiR Result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel.
Op. Modes:M t
5-A-1
_. _. _. _ _ -. ~.. _ _ _ _. _.. _. _ _ _ _ _.
s
. ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT AU1 2
3-Initiating Condition - NOTIFICATION OF UNUSUAL EVENT 4
{
5 Any UNPLANNED Release of Gaseous or Liquid Radioactivity to the Enviromnent that 6-Exceeds Two Times the Radiological Effluent Technical Specifications for 60 Minutes or 7
Longer.
8 9_
Operating Mode Applicability:
All
.10 11 Example Emergency Action Levels: (1 or 2 or 3 or 4 or 5) 12 13 1.
VALID reading on any effluent monitor that exceeds two times the alarm setpoint 14 established by a current radioactivity discharge permit for 60 minutes or longer.
115 16 2.
VALID reading on one or more of the following radiation monitors that exceeds the 17 reading shown for 60 minutes or longer:
18 19 (site-specific list) 20 21 3.
Confirmed sample analyses for gaseous or liquid releases indicates concentrations or 22 release rates, with a release duration of 60 minutes or longer, in excess of two times (site-23 specific technical specifications).
24 25 4.
VALID reading on perimeter radiation monitoring system greater than 0.10 mR/hr above 26 normal background sustained for 60 minutes or longer [for sites having telemetered perimeter 27 monitors).
28 29 5.
VALID indication on automatic real-time dose assessment capability greater than (site-30 specific value) for 60 minutes or longer (for sites having such capability).
31 32 Basis:
33 34 Refer to Appendix A for a detailed basis of the radiological effluent IC / EALs.
35 36 This IC addresses a potential or actual decrease in the level of safety of the plant as indicated by a 37 - radiological release that exceeds regulatory commitments for an extended period of time. Nuclear 38 power plants incorporate features intended to control the release of radioactive effluents to the 39 environment. Further, there are administrative controls established to prevent unintentional releases, 40 or control and monitor intentional releases. These controls are located in the Offsite Dose Calculation 41 Manual (ODCM), and for plants that have not implemented Generic Letter 89-01, in the Radiological 42 Effluent Technical Specifications (RETS). The occurrence of extended, uncontrolled radioactive
- 45 releasee to the er.vimament is indicative of a degradation la there features andhr cent"ols. Some 44
' sit 6s may find it advantageous to address gaseous and liquid releases wim t.eparate initiating 45 conditions and EALs.
46 47-The RETS multiples are specified in ICs AU1 and AA1 only to distinguish between non-emergency 48_ conditions, and from each other. While these multiples obviously correspond to an offsite dose or 49 dose rate, the emphasis in classifying these events is the degradation in the level of safety of the A-2
I plant, NOT the mrgnituds of ths tssociated dose or dose rate. Releases should not be prorated or 2
averaged. For example, a release exceeding 4x RETS for 30 minutes does not meet the threshold 3
for this IC.
4 5
UNPLANNED, as used in this context, includes any release for which a radioactivity discharge permit 6
was not prepared, or a release that exceeds the conditions (e.g., minimum dilution flow, maximum 7
discharge flow, alarm setpoints, etc.) on the applicable permit. VAL /D means that a radiation monitor 8
reading has been confirmed by the operators to be correct by channel check or comparison to 9
redundant monitors, etc. The Emergency Director should not wait until 60 minutes has elapsed, but 10 should declare the event as soon as it is determined that the release duration has or will likely l-11 exceed 60 minutes. Also, if an ongoing release is detected and the starting time for that release is 12 unknown, the Emergency Director should, in the absence of data to the contrary, assume that the
{
13 release has exceeded 60 minutes.
i 14 15 EAL #1 addresses radioactivity releases that for whatever reason cause effluent radiation monitor l.
16 readings that exceed two times the alarm setpoint established by the radioactivity discharge permit.
17 This alarm setpoint may be associated with a planned batch release, or a continuous release path. In 18 either case, the setpoint is established by the ODCM to wam of a release that is not in compliance t
i 19 with the RETS. Indexing the EAL threshold to the ODCM setpoints in this manner insures that the
(
'20 EAL threshold will never be less than the setpoint established by a specific discharge permit.
l 21 22 EAL #2 is similar to EAL #1, but is intended to address effluent or accident radiation monitors on non-23 routine release pathways (i.e., for which a discharge permit would not normally be prepared). The 24 ODCM establishes a methodology for determining effluent radiation monitor setpoints. The ODCM l
25 specifies default source terms and, for gaseous releases, prescribes the use of pre-determined j
26 annual average meteorology in the most limiting downwind sector for showing compliance with the j
27 regulatory commitments These monitor reading EALs should be determined using this methodology.
28 l
29 EAL #3 addresses uncontrolled releases that are detected by sample analyses, particularly on i
30 unmonitored pathways, e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in j
31 river water systems, etc.
32 33 The 0.10 mR/hr value in EAL #4 is based on a release rate not exceeding 500 mrem per year, as 34 provided in the ODCM / RETS, prorated over 8766 hours0.101 days <br />2.435 hours <br />0.0145 weeks <br />0.00334 months <br />, multiplied by two, and rounded. (500 +
35 8766 x 2 = 0.114). This is also the basis of the site specific value in EAL #5.
36 l
37 EALs #1 and #2 directly correlate with the IC since annual average meteorology is required to be 38 used in showing compliance with the RETS and is used in calculating the alarm setpoints. EALs #4 39 and #5 are a function of actual meteorology, which will likely be different from the limiting annual 40 average value. Thus, there will likely be a numerical inconsistency. However, the fundamental basis 41 of this IC is NOT a dose or dose rate, but rather the degradation in the level of safety of the plant 42 implied by the uncontrolled release. Exceeding EAL #4 or EAL #5 is an indication of an uncontrolled 43 release meeting the fundamental basis for this IC.
44 a
I-4 1
5-A-3 l
1 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT 2
AU2 3
Initiating Condition - NOTIFICATION OF UNUSUAL EVENT 4
5 Unexpected increase in Plant Radiation.
6 7.
Operating Mode Applicability:
All 8
9 Example Emergency Action Levels: (1 or 2 or 3) 10 11 1.
VALID (site-specific) indication of uncontrolled water level decrease in the reactor 12 refueling cavity, spent fuel pool, or fuel transfer canal with all irradiated fuel assemblies 13
' remaining covered by water.
14 15 2.
VALID (site-specific) radiation reading for irradiated spent fuel in dry storage.
16 17-3.
VAllD Direct Area Radiation Monitor readings increases by a factor of 1000 over normal
- 18 levels.
19 20
- Normal levels can be considered as the highest reading in the past twenty-four hours 21 excluding the current peak value.
22 23 Basis:
24 25 This IC addresses events that have resulted, or may result, in unexpected increases in radiation 26 dose rates within plant buildings. Such increases represent a loss of control over radioactive material 27 and may represent a potential degradation in the level of safety of the plant.
28
'29 VAllD means that a radiation monitor reading has been confirmed by the operators to be correct by 30 channel check or comparison to redundant monitors, etc.
31 32 In light of Reactor Cavity Seal failure incidents at two different PWRs and loss of water in the Spent 33 Fuel Pit / Fuel Transfer Canal at a BWR, explicit coverage of these types of events via EAL #1 is 34 appropriate given their potential for increased doses to plant' staff. Classification as a NOUE is 35 warranted as a precursor to a more serious event. Site-specific indications may include 36-instrumentation such as water level and local area radiation monitors, and personnel (e.g., refueling 37 crew) reports. If available, security video cameras may allow remote observation. Depending on 38 available level instrumentation, the declaration threshold may need to be based on indications of
- 39. water makeup rate or decrease in refueling water storage tank level. While a radiation monitor could 40 detect an increase in dose rate due to a drop in the water level, it might not be a reliable indication of 41 - ' whether or not the fuel is covered. For example, the reading on an area radiation monitor located on 42 the refueling bridge may increase due to planned evolutions such as head lift, or even a fuel 45 noembly bsing raised in the mudpulator inast. Geners!!y, increased radiation moWr 'Ad! cation: wir 44 need to wmbined with enother indicator (or personnel report) of water loss. This event e calates to 45 an Alert per IC AA2 if irradiated fuel outside the reactor vessei is uncovered. For events involving 46 irradiated fuel in the reactor vessel, escalation would be via the Fission Product Barrier Matrix for 47 events in operating modes 1-4.
48 t
5-A-4
l l
i 1
EAL #2 addresses the degradation of irradiated spent fuel stored onsite in dry storage modules or l-2. casks. These modules are designed to standards identified in 10 CFR Part 72. The dry storage 3
modules are routinely monitored by site Radiation Protection / Health Physics personnel, such that any 4.
degradation would be detected. Readings of (site specific dose rate) are indicative of degradation of 5
the irradiated spent fuel or storage cask / module. The value of (site specific dose rate) should be 6
based on not exceeding a dose rate of 0.10 mR/hr at the closest point for public access, which 7
should be considered the site's Restricted Area Boundary. This value is consistent with 10 CFR Part 8
20 limits for members of the public..
9 10 EAL #3 ' addresses UNPLANNED increases in in-plant radiation levels that represent a degradation in l
11 the control of radioactive material, and represent a potential degradation in the level of safety of the l
12 plant. This event escalates to an Alert per IC AA3 if the increase in dose rates impedes personnel l
13 access necessary for safe operation.
(
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5-A-5 l.
1 2
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT 3
AA1 4
initiating Condition - ALERT 5
6 Any UNPLANNED Release of Gaseous or Liquid Radioactivity to the Environment that 7
Exceeds 200 Times the Radiological Effluent Technical Specifications for 15 Minutes or 8
Longer.
9 10 Operating Mode Applicability:
All 11 12 Example Emergency Action Levels: (1 or 2 or 3 or 4 or 5) 13 14 1.
VAllD reading on any effluent monitor that exceeds 200 times the alarm setpoint 15 established by a current radioactivity discharge permit for 15 minutes or longer.
16 17 2.
VALID reading on one or more of the following radiation monitors that exceeds the 18 reading shown for 15 minutes or longer-19 20 (site-specific list) 21 22 3.
Confirmed sample analyses for gaseous or liquid releases indicates concentrations or 23 release rates, with a release duration of 15 minutes or longer, in excess of 200 times (site-24 specific technical specifications).
25 26 4.
VALID reading on perimeter radiation monitoring system greater than 10.0 mR/hr above 27 normal background sustained for 15 minutes or longer [for sites having telemetered perimeter 28 monitors).
29 30 5.
VALID indication on automatic real-time dose assessment capability greater than (site-31 specific value) for 15 minutes or longer [for sites having such capability).
32 33 Basis:
34 35 Refer to Appendix A for a detailed basis of the radiological effluent IC /EALs.
36 37 This IC addresses a potential or actual decrease in the level of safety of the plant as indicated by a 38 radiological release that exceeds regulatory commitments for an extended period of time. Nuclear 39 power plants incorporate features intended to control the release of radioactive effluents to the 40 environment. Further, there are administrative controls established to prevent unintentional releases, 41 or control and monitor intentional releases. These controls are located in the Offsite Dose Calculation 42 Manual (ODCM), and for plants that have not implemented Generic Letter 89-01, in the Radiological 43 Driuent TechrE Specificatic:
.. ETS). ?e occurrence of c.xtended s.c.ontre..ed rsdioactive 44 releases to the environment is it'dicative of a cagradation in these features and/or controls. Some 45 sites may find it advantageous to address gaseous and liquid releases with separate initiating l
46 conditions and EALs.
47 48 The RETS multiples are specified in ICs AU1 and AA1 only to distinguish between non-emergency 49 conditions, and from each other. While these multiples obviously correspond to an offsite dose or 5-A-6
- - ~..- -...--.--._._.-
1 dose rate, the emphasis in classifying these events is the degradation in the level of safety of the 2
plant, NOT the magnitude of the associated dose or dose rate. Releases should not be prorated or 3
averaged. For example, a release exceeding 400x RETS for 30 minutes does not meet the threshold 4
for this IC.
5 6
UNPLANNED, as used in this context, includes any release for which a radioactivity discharge permit 7
was not prepared, or a release that exceeds the conditions (e.g., minimum dilution flow, maximum 8
discharge flow, alarm setpoints, etc.) on the applicable permit. VAL /D means that a radiation monitor l
9 reading has been confirmed by the operators to be correct by channel check or comparison to 10 redundant monitors, etc. The Emergency Director should not wait until 15 minutes has elapsed, but 11 should declare the event as soon as it is determined that the release duration has or will likely 12 exceed 15 minutes. Also, if an ongoing release is detected and the starting time for that release is 13 unknown, the Emergency Director should, in the absence of data to the contrary, assume that the
- 14. release has exceeded 15 minutes.
15 16 EAL #1 addresses radioactivity releases that for whatever reason cause effluent radiation monitor l
17 readings that exceed two hundred times the alarm setpoint established by the radioactivity discharge l
18 permit. This alarm setpoint may be associated with a planned batch release, or a continuous release 19 path.'In either case, the setpoint is established by the ODCM to wam of a release that is not in 20 compliance with the RETS. Indexing the EAL threshold to the ODCM setpoints in this manner insures l
21 that the EAL threshold will never be less than the setpoint established by a specific discharge permit.
22 23 EAL #2 is similar to EAL #1, but is intended to address effluent or accident radiation monitors on non-24 routine release pathways (i.e., for which a discharge permit would not normally be prepared). The 25 ODCM establishes a methodology for determining effluent radiation monitor setpoints. The ODCM 26 specifies default source terms and, for gaseous releases, prescribes the use of pre-determined 27 annual average meteorology in the most limiting downwind sector for showing compliance with the l
28 regulatory commitments. These monitor reading EALs should be determined using this methodology.
29 30 EAL #3 addresses uncontrolled releases that are detected by sample analyses, particularly on 31 unmonitored pathways, e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in l
32 river water systems, etc.
33 34 The 10.0 mR/hr value in EAL #4 is based on a release rate not exceeding 500 mrem per year, as j
35 provided in the ODCM / RETS, prorated over 8766 hours0.101 days <br />2.435 hours <br />0.0145 weeks <br />0.00334 months <br />, multiplied by 200, and rounded. (500 +
36 8766 x 200 = 11.4). This is also the basis of the site specific value in EAL #5.
37 38 EALs #1 and #2 directly correlate with the IC since annual average meteorology is required to be 39 used in showing compliance with the RETS and is used in calculating the alarm setpoints. EALs #4 l
40 and #5 are a function of actual meteorology, which will likely be different from the limiting annual l
41 average value. Thus, there will likely be a numerical inconsistency. However, the fundamental basis 42 of this IC is NOT a dose or dose rate, but rather the degradation in the level of safety of the plant 43 implied by the uncontrolled release. Exceeding EAL #4 or EAL #5 is an indication of an uncontrolled 44 release meeting the fundamental basis for this IC.
45 46 Due to the uncertairwy associated with meteorology, emergency implementing procedures should call 47 for the timely performance of dose assessments using actual (real-time) meteorology in the event of l
48 a gaseous radioactivity release of this magnitude. The results of these assessments should be 49 compared to the ICs AS1 and AG1 to determine if the event classification should be escalated.
.50 Contrary to the practices specified in revision 2 of this document, classification should not be delayed 51
'pending the results of these dose assessments.
5-A-7
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A"NERMAL RAD LEVELS / RADIOLOGICAL EFFLUENT 2
AA2 3.
Initiating Condition - ALERT 4
5 Damage to Irradiated Fuel or Loss of Water Level that Has or Will Result in the i
6 Uncovering of irradiated Fuel Outside the Reactor Vessel.
7 l
8 Operating Mode Applicability:
All l
9 l
10 Example Emergency Action Levels: (1 or 2)
L 11 12 1.
A VALID (site-specific) alarm or reading on one or more of the following radiation l
13 monitors: (site-specific monitors)
'14 i
15 Refuel Floor Area Radiation Monitor I
16 Fuel Handling Building Ventilation Monitor 17 Refueling Bridge Area Radiation Monitor 18 19 2.
Water level less than (site-specific) feet for the reactor refueling cavity, spent fuel pool and 20 fuel trarisfer canal that will result in irradiated fuel uncovering.
21 22 Basis:
23 24 This IC addresses specific events that have resulted, or may result, in unexpected increases in 25 radiation dose rates within plant buildings, and may be a precursor to a radioactivity release to the i
26 environment. These events represent a loss of control over radioactive material and represent a 27 degradation in the level of safety of the plant. These events escalate from IC AU2 in that fuel activity 28 has been released, or is anticipated due to fuel heatup. In this classification, the radiation protection l
29 concem shifts from in-plant personnel issue to an issue involving persons offsite as well. This IC 30 applies to spent fuel requiring water coverage and is not intendedTo address spent' fuel which is 31 licensed for dry storage, which is discussed in IC AU2.
l 32 33 VAllO means that a radiation monitor reading has been confirmed by the operators to be correct by 34 channel check or comparison to redundant monitors, etc.
35 4
L 36 EAL #1 addresses radiation monitor indications of fuel uncovery and/or fuel damage. Increased
!^
37 readings on ventilation monitors may be indication of a radioactivity release from the fuel, confirming 38 that damage has occurred. Increased background at the monitor due to water level decrease may 39 mask increased ventilation exhaust airbome activity and needs to be considered. While a radiation 40- monitor could detect an increase in dose rate due to a drop in the water level, it might not be a 41 reliable indication of whether or not the fuel is covered. For example, the monitor could in fact be 42-properly responding to a known event involving transfer or relocation of a source, stored in or near 43-the fuel poc! or responding to a planned evolution such as removal of the reactor head. Application 44 of these initiating Conditions requires understanding of the actual radiological conditions present in 45 the vicinity of the monitor. Information Notice No. 90-08, "KR-85 Hazards from Decayed Fuel"should i.
46 be considered in establishing radiation monitor EAL thresholds.
l
-47 48 in EAL #2, site-specific indications may include instrumentation such as water level and local area l
49 radiation monitors, and personnel (e.g., refueling crew) reports. If available, security video cameras l
5-A-9
s f
11 may allow remote observation. Depending on available leve! indication, the declaration threshold may i
- 21 need to be based on indications of water makeup rate or decrease in refueling water storage tank l
3 level.
i 4
5 Escalation, if appropriate, would occur via IC AS1 or AG1 or Emergency Director judgment.
m f
0 1
b-5-A-10
e l
1 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT l
2 AA3 l
3 Initiating Condition - ALERT 4
5 Release of Radioactive Material or increases in Radiation Levels Within the Facility That 6
Impedes Operation of Systems Required to Maintain Safe Operations or to Establish or 7
Maintain Cold Shutdown 8
l 9
Operating Mode Applicability:
All 10 11 Example Emergency Action Levels: (1 or 2) 12 13 1.
VAllD (site-specific) radiation monitor readings GREATER THAN 15 mR/hr in areas 14 requiring continuous occupancy to maintain plant safety functions:
15 16 (Site-specific) list 17 18 2.
VALID (site-specific) radiation monitor readings GREATER THAN < site specific > values in 19 areas requiring infrequent access to maintain plant safety functions.
20 21 (Site-specific) list 22 23 Basis:
24 25 This IC addresses increased radiation levels that impede necessary access to operating stations, or 26 other areas containing equipment that must be operated manually or that requires local monitoring, in 27 order to maintain safe operation or perform a safe shutdown. It is this impaired ability to operate the 28 plant that results in the actual or potential substantial degradation of the level of safety of the plant.
29 The cause and/or magnitude of the increase in radiation levels is not a concern of this IC. The 30 Emergency Director must consider the source or cause of the increased radiation levels and 31 determine if any other IC may be involved. For example, a dose rate of 15 mR/hr in the control room 32 may be a problem in itself. However, the increase may also be indicative of high dose rates in the 33 containment due to a LOCA. In this latter case, an SAE or GE may be indicated by the fission 34 product barrier matrix ICs.
35 36 VALID means that a radiation monitor reading has been confirmed by the operators to be correct by 37 channel check or comparison to redundant monitors, etc. At multiple-unit sites, the example EALs 38 could result in declaration of an Alert at one unit due to a radioactivity release or radiation shine 39 resulting from a major accident at the other unit. This is appropriate if the increase impairs operations 40 at the operating unit.
41 42 This IC is not meant to apply to increases in the containment dome radiation monitors as these are 43 events which are addressed in tha fission product barrier matrix ICs. Nor is it intended to apply to 44 anticipated temporary increases due to planned events (e.g., incore detector movement, radwaste 45 container movement, depleted resin transfers, etc.)
46 47 Areas requiring continuous occupancy includes the control room and, as appropriate to the site, any 48 other control stations that are manned continuously, such as a radwaste control room or a central 49 security alarm station. The value of 15mR/hr is derived from the GDC 19 value of 5 rem in 30 days 5-A-11 i
l
I with adjustment for expected occupancy times. Although Section til D.3 of NUREG-0737, 2
" Clarification of TMI Action Plan Requirements", provides that the 15 mRlbr value can be averaged 3
over the 30 days, the value is used here without averaging, as a 30 day duration implies an event 4
potentially more significant than an Alert.
5 6
For areas requiring infrequent access, the site-specific value(s) should be based on radiation levels 7
which result in exposure control measures intended to maintain doses within normal occupational 8
exposure guidelines and limits (i.e.,10 CFR 20), and in doing so, will impede necessary access. As 9
used here, impede, includes hindering or interfering provided that the interference or delay is 10 sufficient to significantly threaten the safe operation of the plant.
11 12 Emergency planners developing the site-specific lists may refer to the site's abnormal operating 13 procedures, emergency operating procedures, the 10 CFR 50 Appendix R analysis, and/or, the 14 analyses performed in response to Section 2.1.6b of NUREG-0578, "TMI-2 Lessons teamed Task 15 Force Status Report and Short-term Recommendations", when identifying areas containing safe 16 shutdown equipment. Do not use the dose rates postulated in the NUREG-0578 analyses as a basis 17 for the radiation monitor readings for this IC, as the design envelope for the NUREG-0578 analyses 18 correspond to general emergency conditions.
19 5-A-12
1 1
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT 2
AS1 3
Initiating Condition - SITE AREA EMERGENCY l
4 l
5 Offsite Dose Resulting from an Actual or imminent Release of Gaseous Radioactivity 6
Exceeds 100 mR TEDE or 500 mR Thyroid CDE for the Actual or Projected Duration of l
7 the Release.
l 8
l 9
Operating Mode Applicability:
All l
10 l
11 Examp!e Emergency Action Levels: (1 or 2 or 3 or 4) l 12 13 Note: If dose assessment results are available at the time of declaration, the classification 14 should be based on EAL #2 instead of EAL #1.While necessary declarations should not be 15 delayed awaiting results, the dose assessment should be initiated / completed in order to 16 determine if the classification should be subsequently escalated.
17 18 1.
VAllD reading on one or more of the following radiation monitors that exceeds or is 19 expected to exceed the reading shown for 15 minutes or longer i
20 21 (site-specific list) 22 23 2.
Dose assessment using actual meteorology indicates doses greater than 100 mR TEDE 24 or 500 mR thyroid CDE at or beyond the site boundary.
25 26 3.
A VAllD reading sustained for 15 minutes or longer on perimeter radiation monitoring 27 system greater than 100 mR/hr. (for sites having telemetered perimeter monitors) l 28 29 4.
Field survey results indicate closed window dose rates exceeding 100 mR/hr expected to 30 continue for more than one hour; or analyses of field survey samples indicate thyroid CDE of 31 500 mR for one hour of inhalation, at or beyond the site boundary.
32 33 Basis:
34 35 Refer to Appendix A for a detailed basis of the radiological effluent IC /EALs.
36 37 This IC addresses radioactivity releases that result in doses at or beyond the site boundary that 38 exceed a small fraction of the EPA Protective Action Guides (PAGs). Releases of this magnitude are l
39 associated with the failure of plant systems needed for the protection of the public. While these 40 failures are addressed by other ICs, this IC provides appropriate diversity and addresses events 41 which may not be able to be classified on the basis of plant status alone, e.g., fuel handling accident 42 in spent fuel building.
43 44 The TEDE dose is set at 10% of the EPA PAG, while the 500 mR thyroid CDE was established in 45 consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.
46 47 VALIO means that a radiation monitor reading has been confirmed by the operators to be correct by 48 channel check or comparison to redundant monitors, etc. The Emergency Director should not wait 5-A-13
.1 until 15 minutes has elapsed, but should declare the event as soon as it is determined that the 2
release duration has or will likely exceed 15 minutes.
3 4
The (site specific) monitor list in EAL #1 should include monitors on all potential release pathways.
5 The EPA PAGs are expressed in terms of the sum of the effective dose equivalent (EDE) and the 6
committed effective dose equivalent (CEDE), or as the thyroid committed dose equivalent (CDE). For 7
the purpose of these IC / EALs, the dose quantity total effective dose equivalent (TEDE), as defined 8
in 10 CFR 20, is used in lieu of "... sum of EDE and CEDE...." The EPA PAG guidance provides for 9
the use adult thyroid dose conversion factors. However, some states have decided to calculate child 10 thyroid CDE. Utility IC / EALs need to be consistent with those of the states involved in the facilities 11 emergency planning zone.
12 IS The monitor reading EALs should be determined using a dose assessment method that back 14 calculates from the dose values specified in the IC. The meteorology and source term (noble gases, 15 particulates, and halogens) used should be the same as those used for determining the monitor 16 reading EALs in ICs AU1 and AA1. This protocol will maintdn intervals between the EALs for the four 17 classifications. Since doses are generally not monitored in real-time, it is suggested that a release 18 duration of one hour be assumed, and that the EALs be based on a site boundary (or beyond) dose 19 of 100 mR/ hour whole body or 500 mR/ hour thyrold, whichever is more limiting (as was done for 20 EALs #3 and #4). If individual site analyses indicate a longer or shorter duration for the period in i
21 which the substantial portion of the activity is released, the longer duration should be used.
22 23 Since dose assessment is based on actual meteorology, whereas the monitor reading EALs are not, 24 the results from these assessments may indicate that the classification is not warranted, or may 25 indicate that a higher classification is warranted. For this reason, emergency implementing 26 procedures should call for the timely performance of dose assessments using actual meteorology 27 and release information. If the results of these dose assessments are available when the 28 classification is made (e.g., initiated at a lower classification level), the dose assessment results 29 override the monitor reading EALs. Contrary to the practices specified in revision 2 of this document, 30 classification should not be delayed pending the results of these dose assessments.
31 e
1 l
I l
5 5-A-14 L -
l 1
ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT 2
AG1 l
3 Initiating Condition - GENERAL EMERGENCY 4
i 5
Offsite Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity I
6 Exceeds 1000 mR TEDE or 5000 mR Thyroid CDE for the Actual or Projected Duration of l
7 the Release Using Actual Meteorology.
8 9
Operating Mode Applicability:
All 10 11 Example Emergency Action Levels: (1 or 2 or 3 or 4) 12 13 Note: If dose assessment results am available at the time of declaration, the classification 14 should be based on EAL #2 instead of EAL #1.While necessary declarations should not be 15 delayed awaiting results, the dose assessment should be initiated / completed in order to 16 determine if the classification should be subsequently escalated.
17 18 1.
VALID reading on one or more of the following radiation monitors that exceeds or 19 expected to exceed the reading shown for 15 minutes or longer 20 21 (site-specific list) 22 23 2.
Dose assessment using actual meteorology indicates doses greater than 1000 mR TEDE
{
24 or 5000 mR thyroid CDE at or beyond the site boundary.
25 26 3.
A VAllD reading sustained for 15 minutes or longer on perimeter radiation monitoring 27 system greater than 1000 mR/hr. [for sites having telemetered perimeter monitors]
28 29 4.
Field survey results indicate closed window dose rates exceeding 1000 mRlbr expected to 30 continue for more than one hour; or analyses of field survey samples indicate thyroid CDE of 31 5000 mR for one hour of inhalation, at or beyond site boundary.
32 i
33 Basis:
34 35 Refer to Appendix A for a detailed basis of the radiological effluent IC/EALs.
36 37 This IC addresses radioactivity releases that result in doses at or beyond the site boundary that 38 exceed the EPA Protective Action Guides (PAGs). Public protective actions will be necessary.
39 Releases of this magnitude are associated with the failure of plant systems needed for the protection 40 of the public and likely involve fuel damage. While these failures are addressed by other ICs, this IC 41 provides appropriate diversity and addresses events which may not be able to be classified on the 42 basis of plant status alone. It is important to note that, for the more severe accidents, the release l
43 ~ may be'unmonitored or there may be large uncertainties associated with the source term and/or l.
44 meteorology. For this reason, this IC should not be used to override a plant status IC (e.g., fission 45 product barrier matrix).
46 47 VAL /D means that a radiation monitor reading has been confirmed by the operators to be correct by 48 channel check or comparison to redundant monitors, etc. The Emergency Director should not wait 5-A-15
. _ _ _ _ _ _ _l s
1 until 15 minutes has elapsed, but should declare the event as soon as it is determined that the 2
release duration has or will likely exceed 15 minutes.
3 4
The (site specific) monitor list in EAL #1 should include monitors on all potential release pathways.
5 The EPA PAGs are expressed in terms of the sum of the effective dose equivalent (EDE) and the 6
committed effective dose equivalent (CEDE), or as the thyroid committed dose equivalent (CDE). For 7
the purpose of these IC / EALs, the dose quantity total effective dose equivalent (TEDE), as defined 8
in 10 CFR 20, is used in lieu of "... sum of EDE and CEDE...." The EPA PAG guidance provides for 9
the use adult thyroid dose conversion factors. However, some states have decided to calculate child 10 thyroid CDE. Utility IC / EALs need to be consistent with those of the states involved in the facilities 11 emergency planning zone.
12 13 The monitor reading EALs should be determined using a dose assessment method that 14 backcalculates from the dose values specified in the IC. The meteorology and source term (noble 15 gases, particulates, and halogens) used should be the same as those used for determining the 16 monitor reading EALs in ICs AU1 and AA1. This protocol will maintain intervals between the EALs for 17 the four classifications. Since doses are generally not monitored in real-time, it is suggested that a 18 release duration of one hour be assumed, and that the EALs be based on a site boundary (or 19 beyond) dose of 1000 mR/ hour whole body or 5000 mR/ hour thyroid, whichever is more limiting (as 20 was done for EALs #3 and #4). If individual site analyses indicate a longer or shorter duration for the 21 period in which the substantial portion of the activity is released, the longer duration should be used.
22 23 Since dose assessment.s based on actual meteorology, whereas the monitor reading EALs are not, 24 the results from these assessments may indicate that the classification is not warranted, or may 25 indicate that a higher classification is warranted. For this. reason, emergency implementing 26 procedures should call for the timely performance of dose assessments using actual meteorology 27 and release information. If the results of these dose assessments are available when the 28. classification is made (e.g., initiated at a lower classification level), the dose assessment results 29 override the monitor reading EALs. Contrary to the practices specified in revision 2 of this document, 30 classification should not be delayed pending the results of these dose assessments.
31 l
l
.~.
i 5-A-16 i
Table 5-F-1 Recognition Category F Fission Product Barrier Degradation INITIATING CONDITION MATRIX See Table 3 for BWR Example EALs See Table 4 for PWR Example EALs NOUE ALERT SITE AREA EMERGENCY GENERAL EMERGENCY FU1 ANY Loss or ANY Potential Loss FA1 ANY Loss or ANY Potential Loss FS1 Loss or Potential Loss of ANY FG1 Loss of ANY Two Barriers AND
=
of Containment of EITHER Fuel Clad OR RCS Two Barriers Potential Loss of Third Barrier Op. Modes: Power Operation, Op. Modes: Power Operation, Op. Modes: Power Operation, Op. Modes: Pdwer Operation, Hot Standby, Startup, Hot Hot Standby, Startup, Hot Hot Standby, Startup, Hot Hot Standby, Startup Hot Shutdown Shutdown Shutdown Shutdown NOTES 1.The logic used for these initiating conditions reflects the following considerations:
The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than the Containment Barrier (See Sections 3.4 and 3.8). NOUE ICs associated with RCS and Fuel Clad B viers are addressed under System Malfunction ICs.
At the Site Area Emergency level, there...ust be some ability to dynamically assess how far present conditions are from the threshold for a General Emergency. For example, if Fuel Clad and RCS Barrier " Loss" EALs existed, that, in addition to offsite dose assessments, would require continual assessments of radioactive inventory and containment integrity. Altematively, if both Fuel Clad and RCS Barrier " Potential Loss" EALs existed, the Emergency Director would have more assurance that there was no immediate need to escalate to a General Emergency.
The ability to escalate to higher emergency classes as an event deteriorates must be maintained. For example, RCS leakage steadily increasing i
would represent an increasing risk to public health and safety.
2.
Fission Product Barrier ICs must be capable of addressing event dynamics. Thus, the EAL Reference Table 3 and 4 state that imminent (i.e., within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) Loss or Potential Loss should result in a classification as if the affected threshold (s) are already exceeded, particularly for the higher emergency classes.
5-F-1
TABLE 5-F-2 BWR Esmergency Actlen Level Fission Product Barrier Reference Table Thresholds For LOSS er POTENTIAL LOSS of Barriers
- i
- Determine which combinahon of the three barriers are lost of have a potentialloss and use the foHowing key to classify the event Also, multiple events could occur which resut in the condusion that exceeding the loss or Potentialloss threshokfs is imminent (i.e., within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />). In this imminent loss situation use judgment and classify as if the thresholds are exceeded b
UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY ANYloss orANY Potenti$1 Loss of ANY loss or ANY Potential Loss of EITHER Loss or Potential Loss of ANY two Barriers Loss of ANY two Barriers AND Containment Fuel Clad or RCS Potential Loss of Third Barrier Fuel Clad B'arrier Exaniele EALS RCS Barrier Exasnele EALS Containment Barrier Exannele EALS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS
- 1. Primary Coolant ditivity Level
- 1. Drywell Pressure
- 1. Drywell Pressure Coolant Activity GREATER Not Applicable Pressure GREATER THAN Not Applicable Rapid unexplained (Site-specife) PSIG and i
THAN (site-specific) Value (site-specific) PSIG decrease following initial increasing increase OR t
OR Explosive mixture exists Drywe5 pressure response not consistent with LOCA l
- 2. ReactorVesselWaterLevel
- 2. ReactorVesselWater Level
- 2. ReactorVesselWaterLevel Level LESS THAN (site-Level LESS THAN (site-Level LESS THAN (site-Not Applicable Not Applicable Primary containment specific value) specific value) specific value) flooding required OR OR i
- 3. RCS Leak Rate
- 3. CNMT Isolation Failure or Bypass t
(Site-specific) Indication of RCS leakage GREATER Failure of both valves in Not applicable an unisolable Main THAN 50 gpm inside the any one line to close AND Steamline Break dryweB downstream pathway to the OR environment exists Unisolable primary system OR leakage outside dryweR as Intentional venting per i
indicated by ares EOPs temperature orarea OR j
radiation alarm Unisolable primary system leakage outside drywett as indicated by ares temperature or area radiation alarm OR OR OR
+
8* ~.2 4
TABLE 5-F-2 BWWR Emnergency Action Level Fission Prochset Barrier Reference Table Thresheids For LOSS or POTENTIAL LOSS of Barriers *
- Determine which combination of the three barriers are lost or have a potential loss and use the fonowing key to classify the event. Also, multiple events could occur which result in the conclusion that exceeding the loss or Potereal loss thresholds is imminent (i.e., within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />). In this imminent loss situation use judgment and classify as if the thresholds are exceeded UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY ANYloss or ANY PotentielLoss of ANY loss or ANY Potential Loss of EITHER Loss or Potential Loss of ANY two Barriers Loss of ANY two Barriers AND Containment FuelClad or RCS Potential Loss of Third Barrier Fuel Clad Barrier Exaniele EALS RCS Barrier Exaniele EALS Containneent Barrier Exaneele EALS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS
- 3. Drywell Radiation Monitoring
- 4. Drywell Radiation Monitorino
- 4. Slanificant Radioactive Inventory In Containment DryweB Radiation monitor Not Applicable DryweB Radiation monitor Not Applicable Not applicable Drywell Radiation monitor reading GREATER THAN reading GREATER THAN reading GREATER THAN (site-specific) R/hr (site-specific) R/hr (site-specific) R/hr OR OR OR
- 4. Other(Site-Specific) Indications
- 5. Other(Site-Specific) Indications
- 5. Other(site-specific) Indications (Site specific) as (Site specific) as applicable (Site-specific) as applicable (Site-specific) as applicable (Site specific) as applicable (Site specific) as applicable applicable OR OR OR
- 5. Emeroency DirectorJudomont
- 6. Emeroency DirectorJudgment
- 6. Emerrency Director Judrment Any condition in the opinion of the Emergency Director that Any condition in the opinion of the Emergency Director that Any condition in the opinion of the Emergency Director indicates Loss or Potential Loss of the Fuel Clad Barrier indicates Loss or Potentia! Loss of the RCS Barrier that indicates Loss or Potential Loss of the Containment barrier i
P 5-F-3
s 1
1 Dccis Infcrm:tien Fcr Tgbis 5-F-2 2
BWR Emergency Action Level 3
Fission Product Barrier Reference Table 4
5 FUEL CLAD BARRIER EXAMPLE EALs:(1 or 2 or 3 or 4 or 5) 6 7
The Fuel Clad barrier is the zircalloy or stainless steel tubes that contain the fuel pellets.
8 9
1.
Primary Coolant Activity Level 10 11 This (site-specific) value corresponds to 300 pCl/gm 1131 equivalent. Assessment by the NUMARC 12 EAL Task Force Indicates that this amount of coolant activity is well above that expected for iodine 13 spikes and corresponds to less than 5% fuel clad damage. This amount of radioactivity indicates 14 significant clad damage and thus the Fuel Clad Barrier is considered lost. The value expressed can 15 be either in mR/hr observed on the sample or as uCi/gm results from analysis.
16 17 There is no equivalent " Potential Loss" EAL for this item.
18 19 2.
Reactor Vessel Water Level 20 21 The " Loss" EAL (site-specific) value corresponds to the level which is used in EOPs to indicate 22 challenge of core cooling. Depending on the plant this may be top of active fuel or 2/3 coverage of 23 active fuel. This is the minimum value to assure core cooling without further degradation of the clad.
24 The " Potential Loss" EAL is the same as the RCS barrier " Loss" EAL #2 below and corresponds to 25 the (site-specific) water level at the top of the active fuel. Thus, this EAL indicates a " Loss" of RCS 26 barrier and a " Potential Loss" of the Fuel Clad Barrier. This EAL appropriately escalates the 27 emergency class to a Site Area Emergency. If the " Loss" value is also the Top of Active Fuel, the 28
" Potential Loss" value must be a value indicating a higher level also corresponding to a higher level 29 indicated in the RCS barrier " Loss" EAL #2.
30 31 3.
Drywell Radiation Monitoring 32 33 The (site-specific) reading is a value which indicates the release.of reactor coolant, with elevated 34 activity indicative of fuel damage, into the drywell. The reading should be calculated assuming the 35 instantaneous release and dispersal of the reactor coolant noble gas'and iodine inventory associated 36 with a concentration of 300 Cilgm dose equivalent 1-131 or the calculated concentration equivalent 37 to the clad damage used in EAL #1 into the drywell atmosphere. Reactor coolant concentrations of 38 this magnitude are several times larger than the maximum concentrations (including iodine spiking) 39 allowed within technical specifications and are therefore indicative of fuel damage (approximately 2 -
40 5% clad failure depending on core inventory and RCS volume). This value is higher than that 41 specified for RCS barrier Loss EAL #4. Thus, this EAL indicates a loss of both Fuel Clad barrier and 42 RCS barrier.
43 44 Caution:it is important to recognize that in the event the radiation monitor is sensitive to shine from 45 the reactor vessel or piping, spatious readings will be present and another indicator of fuel c'ad 46 damage is necessary or compensated for in the threshold value.
47 48 There is no " Potential Loss" EAL associated with this item.
5-F-4
p 1 ~ 4.
Oth:r (Site-Specific) Indicat'i:ns 2
3 : This EAL is to cover other (site-specific) indications that may indicate loss or potential loss of the 4
Fuel Clad barrier, including indications from containment air monitors or any other (site-specific) 5 instrumentation.
6 7
5.
Emergency Director Judgment i
8 9
This EAL addresses any other factors that are to be used by the Emergency Director in determining l
10 whether the Fuel Clad barrier is lost or potentially lost, in addition, the inability to monitor the barrier l.
~11 should also be incorporated in this EAL as a factor in Emergency Director judgment that the barrier -
t 12. may be considered lost or potentially lost. (See also IC SG1, " Prolonged Loss of All Offsite Power 13 and Prolonged Loss of All Onsite AC Power", for additionalinformation.)
14 15 RCS BARRIER EXAMPLE EALs: (1 or 2 or 3 or 4 or 5 or 6) 16 j
l 17 The RCS Barrier is the reactor coolant system pressure boundary and includes the reactor vessel 18 and all reactor coolant system piping up to the isolation valves.
l 19 20 1.
Drywell Pressure -
21 22 The (site-specific) drywell pressure is based on the drywell high pressure set point which indicates a 23 LOCA by automatically initiating the ECCS or equivalent makeup system.
i.
24
(
25 There is no " Potential Loss" EAL corresponding to this item.
26 27 2.
Reactor Vessel Water Level 28 29 This " Loss" EAL is the same as " Potential Loss" Fuel Clad Barrier EAL #2. The (site-specific) water 30 level corresponds to the level which is used in EOPs to indicate challenge of core cooling. Depending 31 on the plant this may be top of active fuel or 2/3 coverage of active fuel. This EAL appropriately 32
. escalates the ememency class to a Site Area Emergency. Thus, this EAL indicates a loss of the RCS l
33 barrier and a Pote stial Loss of the Fuel Clad Barrier.
l
.34 L
35 There is no " Potential Loss" EAL corresponding to this item.
i 36 l
37-3.
RCS Leak Rate 38 39 An unisc'able MSL break is a breach of the RCS barrier. Thus, this EAL is included for consistency
[
40 with the Aid emergency classification. The potential loss of RCS based on leakage is set at a level 41 Indicative of a small breach of the RCS but which is well within the makeup capability of normal and 42 emergency high pressure systems. Core uncovery is not a significant concem for a 50 gpm leak, 43 however, break propagation leading to significantly larger loss of inventory is possible. Many BWRs 44 may be unable to measure an RCS leak of this size because the leak would likely increase drywell 45 pressure above the drywell isolation set point. The system normally used to monitor leakage is 46-typically isolated as part of the drywell isolation and is therefore unavailable. If primary system leak
-47 rate information is unavailable, other indicators of RCS leakage should be used.
48 49-Potential loss of RCS based on primary system leakage outside the drywell is determined from site-I 50 specific temperature or area radiation alarms low setpoint in the areas of the main steam line tunnel, 51' main turbine generator, RCIC, HPCI, etc., which indicate a direct path from the RCS to areas outside
~- 52 primary containment. The indicators should be confirmed to be caused by RCS leakage. The area 5-F-5
s 1
trmperrtura or radiation low clarm s:tpoints are indicut:d for this exampl3 to cn;bla en Alert 2
classification. An unisolable leak which is indicated by a high alarm setpoint escalates to a Site Area 3
Emergency when combined with Containment Barrier EAL 3 (after a containment isolation) and a 4
General Emergency when the Fuel Clad Barrier criteria is also exceeded.
5 6
7 4.
Drywell Radiation Monitoring 8
9 The (site-specific) reading is a value which indicates the release of reactor coolant to the drywell. The 10 reading should be calculated assuming the instantaneous release and dispersal of the reactor 11 coolant noble gas and iodine inventory associated with normal operating concentrations (i.e., within 12 T/S) into the drywell atmosphere. This reading will be less than that specified for Fuel Clad Barrier 13 EAL #3. Thus, this EAL would be indicative of a RCS leak only. If the radiation monitor reading 14 increased to that value specified by Fuel Clad Barrier EAL #3, fuel damage would also be indicated.
15 16 However, if the site specific physical location of the drywell radiation monitor is such that radiation 17 from a cloud of released RCS gases could not be distinguished from radiation from adjacent piping 18 and components containing elevated reactor coolant activity, this EAL should be omitted and other 19 site specific indications of RCS leakage substituted.
20 21 There is no " Potential Loss" EAL associated with this item.
22 23 5.
Other (Site-Specific) Indications 24 25 This EAL is to cover other (site-specific) indications that may indicate loss or potential loss of the 26 RCS barrier.
27 28 6.
Emergency Director Judgment 29 30 This EAL addresses any other factors that are to be used by the Emergency Director in determining 31 whether the RCS barrier is lost or potentially lost. In addition, the inability to monitor the barrier 32 should also be incorporated in this EAL as a factor in Emergency Director Judgment that the barrier 33 may be considered lost or potentially lost. (See also IC SG1, " Prolonged Loss of Offsite Power and 34 Prolonged Loss of All Onsite AC Power", for additional information.)
35 36 PRIMARY CONTAINMENT BARRIER EXAMPLE EALs: (1 or 2 or 3 or 4 or 5 or 6) 37 38 The Primary Containment Barrier includes the drywell, the wetwell, their respective interconnecting 39 paths, and other connections up to and including the outermost containment isolation valves.
40 Containment Barrier EALs are used primarily as discriminators for escalation from an Alert to a Site 41 Area Emergency or a General Emergency.
42 43 1.
Drywell Pressure 44 45 Rapid unexplained loss of pressure (i.e., not attributable to drywell spray or condensation effects) 46 following an-initial pressure increase indicates a loss of containment integrity. Drywell pressure 47 should increase as a reault of mass and energy release into containment from a LOCA. Thus., drywell 48 pressure not increasing under these conditions indicates a loss of containment i::tegrity. This 49 indicator relies on the operators recognition of an unexpected response for the condition and 50 therefore does not have a specific value associated. The unexpected response is important because 51 it is the indicator for a containment bypass condition. The (site-specific) PSIG for potential loss of 52 containment is based on the containment drywell design pressure. Existence of an explosive mixture 1
5-F-6
... ~.
. ~.
e 1
me:ns c hydrogen tnd oxygin concentritiori ci tat I: cst ths lower dsfirgrction limit curva cxists. This 2
applies to BWRs with Merk 111 containments, as well as Mark I and 11 containment designs when the 3
are de-inerted.
i Y
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2 i
a E
4
?
5-F !
1 2.
Rs ctsr Vrst Wct2r Laval 2
0 The entry into.the Primary Containment Flooding emergency procedure indicates reactor vessel 4
water level can not be restored and that a core melt sequence is in progress. EOPs direct the 5
operators to enter Containment Flooding when Reactor Vessel Level cannot be restored to greater 6
than a Site Specific value (generally 2/3 core height) or is unknown. Entry into Containment Flooding 7
procedures is a logical escalation in response to the inability to maintain reactor vessel level.
8 9
The conditions in this potential loss EAL represent imminent core melt sequences which, if not 10 corrected, could lead to vessel failure and increased potential for containment failure, in conjunction 11 with and an escalation of the level EALs in the Fuel and RCS barrier columns, this EAL will result in 12 the declaration of a General Emergency -loss of two barriers and the potential loss of a third. If the 13 emergency operating procedures have been ineffective in restoring reactor vessel level above the 14 RCS and Fuel Clad Barrier Threshold Values, there is not a " success" path and core melt is 15 imminent. Entry into Containment flooding procedures is a logical escalation in response to the 16-inability to maintain reactor vessel level.
17 18 Severe accident analysis (e.g., NUREG-1150) have concluded that function restoration procedures 19 can arrest core degradation with the reactor vessel in a significant fraction of the core damage 20 scenarios, and the likelihood of containment failure is.very small in these events. Given this, it is 21 appropriate to provide a reasonable period to allow emergency operating procedures to arrest the 22 core melt sequence. Whether or not the procedures will be effective should be apparent within the l
23 time provided. The Emergency Director should make the declaration as soon as it is determined that I
24 the procedures have been, or will be, ineffective. There is no " loss" EAL associated with this item.
25 26 3.
Containment isolation Failure or Bypass 27 l
28 This EAL is intended to cover the inability to isolate the containment when containment isclation is 29 required. In addition, the presence of area radiation or temperature alarms high setpoint indicating 30 unisolable primary system leakage outside the drywell are covered after a containment isolation. The l
31 indicators should be confirmed to be caused by RCS leakage. Also, an intentional venting of primary 32 containment for pressure control per EOPs to the secondary containment and/or the environment is 33 considered a loss of containment. Containment venting for temperature or pressure when not in an 34 accident situation should not be considered.
35-36 There is no " Potential Loss" EAL associated with this item.
37 38 4.
Significant Radioactive inventory in Containment 39 l
40 The (site-specific) reading is a value which indicates significant fuel damage well in excess of that 41 required for loss of RCS and Fuel Clad. As stated in Section 3.8, a major release of radioactivity 42 requiring offsite protective actions from core damage is not possible unless a major failure of fuel 43 cladding allows radioactive material to be released from the core into the reactor coolant. Regardless 44-of whether containment is challenged, this amount of activity in containment, if released, could have
.45 such severe consequences that it is prudent to treat this as a potential loss of containment, such that 46-a General Emergency decle**' ion is warranted. NUREG-1228. " Source Estimatione DurM Incident 47 Response to Severe Nuclear Pows: Plant Accidents," indicates that such conditions do not exist 48 when the amount of clad damage is less than 20%. Unless there is a (site-specific) analysis justifying 49 a higher value, it is recommended that a radiation monitor reading corresponding to 20% fuel clad 50 damage be specified here.
51 52 There is no " Loss" EAL associated with this item.
5-F-8 1
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T 1
1
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5-F-9 1
(
'1 5.
Other(Sit 2-Specific)Indic ti:n3 2
3 This EAL is to cover other (site-specific) indications that may indicate loss or potential loss of the 4
containment barrier.
5 6
6.
Emergency Director Judgment q.
8 This EAL addresses any other factors that are to be used by the Emergency Director in determining 9
whether the Containment barrier is lost or potentially lost. In addition, the inability to monitor the 10 barrier should also be incorporated in this EAL as a factor in Emergency Director judgment that the 11
. barrier may be considered lost or potentially lost. (See also IC SG1, " Prolonged Loss of All Offsite 12 Power and Prolonged Loss of All Onsite AC Power", for additional information.)
13 l
3 4
e i
h44 M Pu*'
- *%(
meen 49 e e MM f J 's
... e u y a e
5-F-10
TABLE 5-F-4 PWfR EmerBency Action Level Fission Product Barrier Reference Table Thresholds For LOSS er POTENTIAL LOSS of Barriers *
- Detennine which wnun,a of the three barriers are lost or have a potentia? loss and use the foHowing key to classify the event. Also an event for multiple events could occur which result in the conclusion that exceeding the loss or poter'tialloss thresholds is inwninent (i.e within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />). In this imminent loss situation use judgment and classify as if the thresholds are exceeded UNUSUAL EVENT ALERT SITE AltEA EMERGENCY GENERAL EMERGENCY ANY loss or ANY Potential Loss of ANY loss or ANY Potential Loss of EITHER Loss or Potential Loss of ANY two Baniers Loss of ANY two Barriers AND Containment Fuel Ctad or RCS Potential Loss of Third Barrier Fuel Clad B'arrier Example EALS RCS Barrier Example EALS Centainment Barrier Exanssie EALS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS L Critical Safety Function Status L Critical Safety Function Status
- 1. Critical Safety Function Status Core-Cooling Red Core Cooling-Orange OR Not Applicable RCS Integrity-Red OR Heat Not AppFcable Containrnent-Red Heat Sink-Red Sink-Red OR OR OR
- 2. Primary Coolant Activity Level
- 2. RCS Leak Rate
- 2. Containment Pressure Coolant Activity GREATER Not Applicable GREATER THAN available Unisofable leak exceeding Rapid or unexplained (Site-specific)PStG and THAN (site-specinc)Value makeup capacity as the capacity of one decrease foHowing initial increasing indicated by a loss of RCS charging pump in the increase OR subcoonng normal charging mode OR Explosive mixture exists Containment pressure or OR sump level response not Pressure greater than consistent with LOCA containment depressurizat-conditions ion actuation setpotnt with less than one full train of depressurization equipment operating OR OR
- 3. Core Exit Thermocouple Readings
- 3. Core Exit Thomocauole Reading GREATER THAN (site-GREATERTHAN (site-Not applicable Core extt thermocouples in specific) degree F specific) degree F excess of 1200 degrees and restoratm procedures not effectrve within 15 minutes; or, core thermocouples in excess of 700 degrees with reactor vessellevel below top of active fuel and restoration procedures not e8fectrve
[
within 15 minutes 5-F-11
TABLE 5-F-4 PWR Emergency Actlen Level Fission Product Barrier Reference Table Thresholds For LOSS er POTENTIAL LOSS of Barriers
- l
- Determine which combination of the three barriers are lost or have a potentialloss and use the fo8owing key to dassify the event. Also an event for multiple events could occur which resut in the conclusion that exceeding the loss or potentialloss thresholds is immment (i.e., within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />). In this imminent loss sRustion use judgment and dessify as if the thresholds are exceeded UNUSUAL EVENT ALERT SITE AREA EMEROENCY GENERAL ENIERGENCY ANY loss or ANY Potential Loss of ANY loss or ANY Potential Loss of E!THER Loss or Potentia! Loss of ANY two Barriers Loss of ANY two Barriers AND Containment Fuel Ctad or RCS Potential Loss of Third Barrier Fuel Clad Barrier Example EALS RCS Barrier Example EALS Containment Barrier Example EALS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOFS OR OR OR
- 4. ReactorVesselWaterLevel
- 3. SG Tube Rupture
- 4. SG Secondary Side Release with P-to-S Leakaoe Not Applicab!e Level LESS than (site-SGT9 that resuRs in an Not Applicable RUPTURED S/G is also Not app 5 cab'e specific) value ECCS (SI) Actuation FAULTED outside of containment OR Primary-tMecondary leakrate greater than 10 gpr~ with nonisolable steam release from affected S!G OR
- 5. CNMT isolation Valves Status After CNMT Isolation Valve (s) not closed AND Not Applicable downstream pathway to the environment exists OR OR OR
- 5. Containment Radiation Ironitorina
- 4. Containment Radiation Monitorina
- 6. Slanificant Radioactive Inventory in Containment Containment rad monitor Not Appucable Containment rad monitor Not Applicable Not Applicable Containment rad monitor reading GREATER THAN reading GREATER THAN reading GREATER THAN (site-specific) R/hr (sRe-specific) R/hr (site-specific) R/hr OR OR OR
- 6. Other(Site-SpecificiIndications
- 5. Other(Site-Specific) Indications
- 7. Other(site-specificilndications (Site specific ) as (Site specific) as applicable (Site-specific) as applicab!e (Site-specific) as applicable (Site specific) as applicable (Site specific) as appncable applicable 5-F-12 s
i s
TABLE 5-F-4 PWR EmerWency Action Level
^
t Fission Predect Barrier Reference Table l
Thresholds For LOSS er POTENTIAL LCSS of Barriers *
- Determine which cornbination of tho three barriers are lost or have a potential loss and use the fotowing key to classify the event. Also an event for multiple events could occur which result in the conclusion that exceeding the loss or potentialloss thresholds is imminent (i.e., within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />). In this imminent loss situation use judgrnent and classify as if the thresholds are exceeded UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY ANY loss or ANY Potential Loss of ANY loss or ANY Potential Loss of EITHER Loss or Potential Loss of ANY two Barriers Loss of ANY two Barriers AND Containment l
FuelClad or RCS Potentialloss of Third Barrier i
Fuel Clad Barrier Example EALS RCS Barrier Example EALS Containment Barrier Example EALS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS OR OR OR t
t 5-F-13
TABLE 5-F-4 PWR Eneergency Action Level Fission Product Barrier Reference Table Thresholds For LOSS er POTENTIAL LOSS of Barriers *
- Determine whk:h combination of the three barriers are lost or have a potentialloss and use the foRowing key to classify the event. Also an event for multiple events could occur wNch result in the conclusion that exceeding the loss or potential loss thresholds is imminent (i.e., within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />). In this imminent loss situation use judgment and classify as if the thresholds are exceeded r
UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY ANY loss or ANY Potential Loss of ANY loss or ANY Potential Loss of EITHER Loss or Potential Loss of ANY two Barriers Loss of ANY two Barriers AND Containment Fuel Clad or RCS Potential Loss of Third Barrier Fuel Clad Barrier Example EALS RCS Barrier Example EALS Containment Barrier Example EALS LOSS t
POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS PCTENTIAL LOSS
- 7. Emeroency Direc irJudament
- 6. Emeroency Director Judament
- 8. Emeroency Director Judoment e
Any condition in the opinion of the Emergency Director that Any condition in the opinion of the Emergency Director that Any condition in the opinion of the Emergency Director indicates Loss or PNe ntial Loss of the Fuel C!ad Danier indicate Loss or Potential Loss of the RCS Barrier that indicates Loss or Potential Loss of the Containment barrier i
5-F-14 e
i 1
Basis Information For Table 5-F-4 2
PWR Emergency Action Level 3
Fission Product Barrier Reference Table 4
i 5
FUEL CLAD BARRIER EXAMPLE EALs: (1 or 2 or 3 or 4 or 5 or 6) 6 7
The Fuel Clad Barrier is the zircalloy or stainless steel tubes that contain the fuel pellets.
8 9
1.
Critical Safety Function Status 10 11 This EAL is for PWRs using Critical Safety Function Status Tree (CSFST) monitoring and functional 12 recovery procedures. For more information, please refer to Section 3.9 of this report. RED path 13 indicates an extreme challenge to the safety function. ORANGE path indicates a severe challenge to 14 the safety function.
l 15 16 Core Cooling - ORANGE indicates subcooling has been lost and that some clad damage may occur.
17 Heat Sink - RED indicates the ultimate heat sink function is under extreme challenge and thus these 18 two items indicate potentialloss of the Fuel Clad Barrier.
19 20 Core Cooling - RED indicates significant superheating and core uncovery and is considered to 21 indicate loss of the Fuel Clad Barrier.
22 23 2.
Primary Coolant Activity Level 24 25 This (site-specific) value corresponds to 300 Ci/gm 1131 equivalent. Assessment by the NUMARC 26 EAL Task Force indicates that this amount of coolant activity is well above that expected for iodine 27 spikes and corresponds to less than 5% fuel clad damage. This amount of radioactivity indicates 28 significant clad damage and thus the Fuel Clad Barrier is considered lost. The value expressed can 29 be either in mR/hr observed on the sample or as uCi/gm results from analysis.
30 31 There is no equivalent " Potential Loss" EAL for this item.
32 33 3.
Core Exit Thermocouple Readings 34 35 Core Exit Thermocouple Readings are included in addition to the Critical Safety Functions to include 36 conditions when the CSFs may not be in use (initiation after Si is blocked) or plants which do not 37 have a CSF scheme.
38 39 The " Loss" EAL (site-specific) reading should correspond to significant superheating of the coolant.
40 This value typically corresponds to the temperature reading that indicates core cooling - RED in Fuel 41 Clad Barrier EAL #1 which is usually about 1200 degrees F.
42 43 The " Potential Loss" EAL (site-specific) reading should correspond to loss of subcooling. This value 44 typically corresponds to the temperature reading that indicates core cooling - ORANGE in Fuel Clad 45 Barrier EAL #1 which is usually about 700 to 900 degrees F.
46 47 4.
Reactor Vessel Water Level 48 49 There is no " Loss" EAL corresponding to this item because it is better covered by the other Fuel Clad 50 Barrier " Loss" EALs.
5-F-15
s 1
2 The (site-specific) value for the " Potential Loss" EAL corresponds to the top of the active fuel. For 3
sites using CSFSTs, the " Potential Loss" EAL is defined by the Core Cooling - ORANGE path. The 4
(site-specific) value in this EAL should be consistent with the CSFST value.
5 6
5.
Containment Radiation Monitoring 7
8 The (site-specific) reading is a value which indicates the release of reactor coolant, with elevated 9
activity indicative of fuel damage, into the containment. The reading should be calculated assuming 10 the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory 11 associated with a concentration of 300 pCl/gm dose equivalent 1-131 into the containment 12 atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the 13 maximum concentrations (including iodine spiking) allowed within technical specifications and are 14
,therefore indicative of fuel damage (approximately 2 - 5% clad fa;iure depending on core inventory 15 and RCS volume). This value is higher than that specified for RCS barrier Loss EAL #4. Thus, this 16 EAL indicates a loss of both the fuel clad barrier and a loss of RCS barrier.
17 18 There is no " Potential Loss" EAL associated with this item.
19 20 6.
Other (Site-Specific) Indications 21 t
22 This EAL is to cover other (site-specific) indications that may indicate loss or potential loss of the 23 Fuel Clad barrier, including indications from containment air monitors or any other (site-specific) 24-instrumentation.
25 26 7.
Emergency Director Judgment 27 28 This EAL addresses any other factors that are to be used by the Emergency Director in determining 29 whether the Fuel Clad barrier is lost or potentially lost. In addition, the inability to monitor the barrier j
30 should also be incorporated in this EAL as a factor in Emergency Director judgment that the barrier 31 may be considered lost or potentially lost. (See also IC SG1, " Prolonged Loss or All Offsite Power 32 and Prolonged Loss of All Onsite AC Power", for additionalinformation.)
33 34 RCS BARRIER EXAMPLE EALs: (1 or 2 or 3 or 4 or 5 or 6) 35 36 The RCS Barrier includes the RCS primary side and its connections up to and including the j
37 pressurizer safety and relief valves, and other connections up to and including the primary isolation 38 valves.
39 40 1.-
Critical Safety Function Status 41 42 This EAL is for PWRc using Critical Safety Function Status Tree (CSFST) monitoring and functional 43 recovery procedures. For more information, please refer to Section 3.9 of this report. RED path 44 indicates an extreme challenge to the safety function derived from appropriate instrument readings, 45 and these CSFs indicate a potential! css of RCS barrier.
46 47 There is no " Loss" EAL associated with this item.
48 49 2.
RCS Leak Ra e 50 1
5-F-16
t t
1 The " Loss" EAL addresses conditions where leakage from the RCS is greater than available 2
inventory control capacity such that a loss of subcooling has occurred. The loss of subcooling is the
'3 fundamental indication that the inventory control systems are inadequate in maintaining RCS
'4 pressure and inventory against the mass loss through the leak.
5
- 6 The " Potential Loss". EAL is based on the inability to maintain normal liquid inventory within the 7
Reactor Coolant System (RCS) by normal operation of the Chemical and Volume Control System 8
which is considered as one centrifugal charging pump discharging to the charging header. A second 9
charging pump being required is indicative of a substantial RCS leak. For plants with low capacity 10 charging pumps, a 50 gpm leak rate value may be used to indicate the Potential Loss. In conjunction 11 with the SG Tube Rupture " Potential Loss" EAL this assures that any event that results in significant 12 RCS inventory shrinkage or loss (e.g., events leading to reactor scram and ECCS actuation) will 13-result in no lower than an " Alert" emergency classification.
14 l
15 3.
SG Tube Rupture 16 l
17 This EAL is intended to address the full spectrum of Steam Generator (SG) tube rupture events in 18 conjunction with Containment Barrier " Loss" EAL #4 and Fuel Clad Barrier EALs The " Loss" EAL l
19 addresses RUPTURED SG(s) for which the leakage is large enough to cause actuation of ECCS
-20 (SI). This is consistent to the RCS Barrier " Potential Loss" EAL #2. For plants that have implemented 21 W.O.G. emergency response guides, this condition is described by ' entry into E-3 required by 22 EOPs". By itself, this EAL will result in the declaration of an Alert. However, if the SG is also l
23 FAULTED (i.e., two barriers failed), the declaration escalates to a Site Area Emergency per 24 ' Containment Barrier " Loss" EAL #4.
25 26 There is no " Potential Loss" EAL.
'27 28 4.
Containment Radiation Monitoring 29.
30-The (site-specific) reading is a value which indicates the release of reactor coolant to the
-31 containment. The reading should be calculated assuming the instantaneous release and dispersal of 32 the reactor coolant noble gas and iodine inventory associated with normal operating concentrations 33 (i.e., within T/S) into the containment atmosphere. This reading will be less than that specified for 34 Fuel Clad Barrier EAL #5. Thus, this EAL would be indicative of a RCS leak only. If the radiation 35 monitor reading increased to that specified by Fuel Clad Barrier EAL #5, fuel damage would also be 36 indicated.
37 38 However, if the site specific physical location of the containment radiation monitor is such that 39_ radiation from a cloud of released RCS gases could not be distinguished from radiation from nearby 40 piping and components containing elevated reactor coolant activity, this EAL should be omitted and 4
41 other site specific indications of RCS leakage substituted.
42 43 There is no " Potential Loss" EAL associate d with this item.
44 45 5..
Other(Site-Specific) Indications J
}_
47 This EAL is to cover other (site-specific) indications that may indicate loss or potential loss of the 48 RCS barrier, including indications from containment air monitors or any other (site-specific) 49 instrumentation.
50 51 6.
Emergency Director Judgment 52-1 5-F-17 F
1 This EAL addresses any other factors that are to be used by the Emergency Director in determining 2
whether the RCS barrier is lost or potentially lost. In addition, the inability to monitor the barrier i
3 should also be incorporated in this EAL as a factor in Emergency Director judgment that the barrier 4'
may be considered lost or potentially lost. (See also IC SG1, " Prolonged Loss of All Offsite Power 5
and Prolonged Loss of All Onsite AC Power", for additional information.)
6 7
CONTAINMENT BARRIER EXAMPLE EALs: (1 or 2 or 3 or 4 or 5 or 6 or 7 or 8) 8.
9 The Containment Barrier includes the containment building, its connections up to and including the 10 outermost containment isolation valves. This barrier also includes the main steam, feedwater, and 11 blowdown line extensions outside the containment bui! ding up to and including the outermost 4
12 secondary side isolation valve.
13 14 1.
Critical Safety Function Status 15-
~16' This EAL is for PWRs using Critical Safety Function Status Tree (CSFST) monitoring and functional 17 recovery procedures. For more information, please refer to Section 3.9 of this report. RED path 18 indicates an extreme challenge to the safety function derived from appropriate instrument readings 19 and/or sampling results, and thus represents a potential loss of containment. Conditions leading to a 20 containment RED path result from RCS barrier and/or Fuel Clad Barrier Loss. Thus, this EAL is 21 primarily a discriminator between Site Area Emergency and General Emergency representing a 22 potentialloss of the third barrier.
23 24 There is no " Loss" EAL associated with this item.
25 26 2.
Containment Pressure 27 28 Rapid unexplained loss of pressure (i.e., not attributable to containment spray or condensation 29 effects) following an initial pressure increase indicates a loss of containment integrity. Containment 30 pressure and sump levels should increase as a result of the mass and energy release into
'31. containment from a LOCA. Thus, sump level or pressure not increasing indicates containment 32 bypass and a loss of containment integrity. The (site-specific) PSIG for potential loss of containment 33 is based on the containment design pressure. Existence of an explosive mixture means a hydrogen
'34 and oxygen concentration of at least the lower deflagration limit curve exists. The indications of 35-potential loss under this EAL corresponds to some of those leading to the RED path in EAL #1 above 36 and may be declared by those sites using CSFSTs. As described above, this EAL is primarily a 37 discriminator between Site Area Emergency and General Emergency representing a potential loss of 38 the third barrier.
39 40 The second potential loss EAL represents a potential loss of containment in that the containment 41 heat removal /depressurization system (e.g., containment sprays, ice condenser fans, etc., but not
'42 - including containment venting strategies) are either lost or performing in a degraded manner, as L
43 indicated by containment pressure greater than the setpoint at which the equipment was supposed to 44 have actuated.
45 46 3.
Core Exit Thermocouples 47-48. In this EAL, the function restoration procedures are those emergency operating procedures that 49 address the recovery of the core cooling critical safety functions. The procedure is considered 50 effective if the temperature is decreasing or if the vessel water level is increasing. For units using the 5-F-18
e d
1-CSF status trees a direct correlation to those status trees can be made if the effectiveness of th 1
2 restoration procedures is also evaluated as stated below.
3 i
4 The conditions in this potential loss EAL represent an imminent core melt sequence which, if not 5
corrected, could lead to vessel failure and an increased potential for containment failure. In 6
conjunction with the Core Cooling and Heat Sink criteria in the Fuel and RCS barrier columns, this 7
EAL would result in the declaration of a General Emergency - loss of two barriers and the potential 8
loss of a third. If the function restoration procedures are ineffective, there is no " success" path.
9 10 Severe accident analyses (e.g., NUREG-1150) have concluded that function restoration procedures 11 can arrest core degradation within the reactor vessel in a significant fraction of the core damage i
12 scenarios, and that the likelihood of containment failure is very small in these events. Given this, it is 13 appropriate to provide a reasonable period to allow function restoration procedures to arrest the core 14 melt sequence. Whether or not the procedures will be effective should be apparent within 15 15 minutes. The Emergency Director should make the declaration as soon as it is determined that the j
16 procedures have been, or will be ineffective. The reactor vessel level chosen should be consistent 17 with the emergency response guides applicable to the facility.
{
18 19 There is no " Loss" EAL associated with this item.
i 20 21 4.
SG Secondary Side Release With Primary To Secondary Leakage 22 i
1
]
23 This " loss" EAL recognizes that SG tube leakage can represent a bypass of the containment barrier i
24 as well as a loss of the RCS barrier. The first " loss' EAL addresses the condition in which a 25 RUPTURED steam generator is also FAULTED. This condition represents a bypass of the RCS and j.
26 containment barriers. In conjunction with RCS Barrier " loss" EAL #3, this would always result in the i
27_
declaration of a Site Area Emergency.
28 29 The second " loss' EAL addresses SG tube leaks than exceed 10 gpm in conjunction with a i
30 nonisolable release path to the environment from the affected steam generator. The threshold for
{
31 establishing the nonisolable secondary side release is intended to be a prolonged release of 32 radioactivity from the RUPTURED steam generator directly to the environment. This could be j
33 expected to occur when the main condenser is unavailable to accept the contaminated steam (i.e.,
i 34 SGTR with concurrent loss of offsite power and the RUPTURED steam generator is required for 35 plant cooldown or a stuck open relief valve. If the main condenser is available, there may be releases 36 via air ejectors, glar.d seal exhausters, and other sirnitar controlled, and often monitored, pathways.
37 These pathways do not meet the intent of a nonisolable release path to the environment. These i
38 minor releases are assessed using Abnormal Rad Levels / Radiological Effluent ICs.
I 39 l
.41 threshold was leakage greater than T/S allowable. Since the prior revision, many plants have 40 The leakage threshold for this EAL has been increased with Revision 3. In the earlier revision, the 42 implemented reduced steam generator T/S limits (e.g.,150 gpd) as a defense in depth associated 43 with alternate steam generator plugging criteria. The 150 gpd threshold is deemed too low for use as j
44 an emergency threshold. A pressure boundary leakage of 10 gpm was used as the threshold in IC 45 SUS, RCS Leakage, and is deemed appropriate for this EAL For smaller breaks, mot exceeding thel 4
46 normal charging capacity threshold in RCS Barrier " Potential Loss" EAL #2 (RCS Leak Rate) or-rct j
47 resulting in ECCS actuation in EAL #3 (SG. Tube Rupture), this EAL results in a NOUE. For largey 48 breaks, RCS barrier EALs #2 and #3 would result in an Alert. For SG tube ruptures which may j
49 involve multiple steam generators or unisolable secondary line breaks, this EAL would exist in 50 conjunction with RCS barrier " Loss" EAL #3 and would result in a Site Area Emergency. Escalation to 51 General Emergency would be based on " Potential Loss" of the Fuel Clad Barrier.
?
5-F-19 i
i
1 2
5.
Containment isolation Valve Status After Containment isolation 3
4 This EAL is intended to address incomplete containment isolation that allows direct release to the 5
environment. It represents a loss of the containment barrier.
6 7-The use of the modifier direct in defining the release path discriminates against release paths 8
through interfacing liquid systems. The existence of an in-line charcoal filter does not make a 9
release path indirect since the filter is not effective at removing fission noble gases. Typical filters
'10-have an efficiency of 95-99% removal of iodine. Given the magnitude of the core inventory of iodine, 11 significant releases could still occur. In addition, since the fission product release would be driven by 12 boiling in the reactor vessel, the high humidity in the release stream can be expected to render the
.13 filters ineffective in a short period.
14 15 There is no " Potential Loss" EAL associated with this item.
16.
17 6.
Significant Radioactive inventory in Containment 18 19 The (site-specific) reading is a value which indicates significant fuel damage well in excess of the 20 EALs associated with both loss of Fuel Clad and loss of RCS Barriers. As stated in Section 3.8, a 21 major release of radioactivity requiring offsite protective actions from core damage is not possible 22 unless a major failure of fuel cladding allows radioactive material to be released from the core into 23 the reactor coolant.
24 25 Regardless of whether containment is challenged, this amount of activity in containment, if released, 26 could have such severe consequences that it is prudent to treat this as a potential loss of 27 containment, such that a. General Emergency declaration is warranted. NUREG-1228, " Source 28 Estimations During incident Response to Severe Nuclear Power Plant Accidents," indicates that such 29 conditions do not exist when the amount of clad damage is less than 20%. Unless there is a (site-30. specific) analysis justifying a higher value, it is recommended that a radiation rnonitor reading 31 corresponding to 20% fuel clad damage be specified here.
32 33' There is no " Loss" EAL associated with this item.
34 35 7.
Other(Site-Specific) Indications 36' 37 This EAL should cover other (site-specific) indications that may unambiguously indicate loss or 38-potential loss of the containment barrier, including indications from area or ventilation monitors in 39 containment annulus or other contiguous buildings. If site emergency operating procedures provide 40 for venting of the containment during an emergency as a means of preventing catastrophic failure, a 41 Loss EAL should be included for the containment barrier. This EAL should be declared as soon as
'42 such venting is imminent. Containment venting as part of recovery actions is classified in accordance 43 with the radiological effluent ICs.
44 45 8.
Emergency Director Judgment 46 47 This EAL addresses any other factors that are to be used by the Emergency Director in determining 48 whether the Containment barrier is lost or potentially lost. In addition, the inability to monitor the 49 barrier should also be incorporated in this EAL as a factor in Emergency Director judgment that the 50~ barrier may be considered lost or potentially lost. (See also IC SG1, " Prolonged Loss of All Offsite 51 ' Power and Prolonged Loss of All Onsite AC Power", for additional information.)
5-F-20
~
TABLE 5-H-1 Recognition Category H Hazards and Other Conditions Affecting Plant Safety INITIATING CONDITION MATRIX NOUE ALERT SITE AREA EMERGENCY GENERAL EMERGENCY HU1 Natural and Destructive HA1 Natural and Destruchve Phenomena Affecting the Phenomena Affecting the Plant PROTECTED AREA.
VITAL AREA.
Op. Modes: AH Op. Modes: AU HU2 FIRE Within PROTECTED AREA HA2 FIRE or EXPLOSION Affecting Boundary Not Extinguished the Operability of Plant Safety Within 15 Minutes of Detection.
Systems Required to Establish Op. Modes: A#
or Maintain Safe Shutdown.
Op. Modes: AR HU3 Release of Toxic or Flammable HA3 Release of Toxic or Flammable Gases Deemed Detrimentalto Gases Within a Facility Structure Safe Operation of the Plant.
Which Jeopardizes Operation of Op. Modes:A#
Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown.
Op. Modes: AR HU4 Confirmed Security Event Which HA4 Confirmed Security Event in a HS1 Confirmed Security Event in a HG1 Security Event Resulting in Loss Indicates a Potential Plant PROTECTED AREA.
Plant VITAL AREA.
Of PhysicalControlof the Degradation in the Level of Op. Modes A#
Op. Modes: A#
Facility.
Safety of the Plant.
Op. Modes: AH Op. Modes: AH HUS Other Conditions Existing Which HA6 Other Conditions Existing Which HS3 Other Conditions Existing Which HG2 Other Conditions Existing Which in the Judgment of the in the Judgment of the in the Judgment of the in the Judgment of the Emergency Director Warrant Emergency Director Warrant Emergency Director Warrant Emergency Director Warrant Declaration of a NOUE.
Declaration of an Alert.
Declaration of Site Area Declaration of General Op. Modes: AH Op. Modes: An Emergency.
Emergency.
Op. Modes: AH Op. Modes: AH HA5 Control Room Evacuation Has HS2 Control Room Evacuation Has Been Initiated.
Been initiated and Plant Control Op. Modes: A#
Cannot Be Established.
Op. Modes: AR 5-H-1
6 1
e
_F
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i
i 1
HAZARDS AND OTHER CONDITIONS 2
AFFECTING PLANT SAFETY 3
HU1 4
Initiating Condition - NOTIFICATION OF UNUSUAL EVENT 5
6 Natural and Destructive Phenomena Affecting the PROTECTED AREA.
7 8
Operating Mode Applicability:
All 9
10 Example Emergency Action Level:
(1 or 2 or 3 or 4 or 5 or 6 or 7) 11 12 1.
(Site-Specific) method indicates felt earthquake.
13 14 2.
Report by plant personnel of tornado striking within PROTECTED AREA boundary, 15 16 3.
Vehicle crash into plant structures or systems within PROTECTED AREA boundary.
17 18 4.
Report by plant personnel of an unanticipated EXPLOSION within PROTECTED AREA 10 boundary resulting in VISIBLE DAMAGE to permanent structure or equipment.
20 21 5.
Report of turbine failure resulting in casing penetration or damage to turbine or generator seals.
22 23 6.
Uncontrolled flooding in (site-specific) areas of the plant that has the potential to affect safety 24 related equipment needed for the current operating mode.
25 26 7.
(Site-Specific) occurrences affecting the PROTECTED AREA.
27 28 BASIS:
29 30 NOUE in this IC are categorized on the basis of the occurrence of an event of sufficient magnitude to 31 be of concem to plant operators. Areas identified in the EALs define the location of the event based 32 on the potential for damage to equipment contained therein. Escalation of the event to an Alert 33 occurs when the magnitude of the event is sufficient to result in damage to equipment contained in 34 the specified location.
35 36 EAL #1 should be developed on site-specific basis. Damage may be caused to some portions of the 37 site, but should not affect ability of safety functions to operate. Method of detection can be based on 38 instrumentation, validated by a reliable source, or operator assessment. As defined in the EPRI-39 sponsored " Guidelines for Nuclear Plant Response to an Earthquake", dated October 1989, a "/elt 40 earfhquake"is:
41
- _42
_..An. earthquake of sufficient intensity such that: (a) the vibratory ground motion is felt at the 43 nuclear plant site and recognized as an earthquake based on a consensus of controi room 44 operators on duty at the time, and (b) for plants with operable seismic instrumentation, the 45 seismic switches of the plant are activated. For most plants with seismic instrumentation, the 46 seismic switches are set at an acceleration of about 0.01g.
47 5-H-2
s 1
EAL #2 is based on the assumption that a tornado striking (touching down) within the PROTECTED l
2-AREA may have potentially damaged plant structures containing functions or systems required for i
3 safe shutdown of the plant. If such damage is confirmed visually or by other in-plant indications, the
)
4 event may be escalated to Alert.
5 6
EAL #3 is intended to address crashes of vehicle types large enough to cause significant damage to i
7 plant structures containing functions and systems required for safe shutdown of the plant. If the crash 8
is confirmed to affect a plant VITAL AREA, the event may be escalated to Alert.
9<
10 For EAL #4 only those EXPLOSIONS of sufficient force to damage permanent structures or 11-equipment within the PROTECTED AREA should be considered. No attempt is made in this EAL to 12 assess the actual magnitude of the damage. The occurrence of the EXPLOSION with reports of 13 evidence of damage is sufficient for declaration. The Emergency director also needs to consider any 14 security aspects of the EXPLOSION, if applicable.
{
15 16
'EAL #5 is intended to address main turbine rotating component failures of sufficient magnitude to 17 cause observable damage to the turbine casing or to the seals of the turbine generator. Of major 18 concem is the potential for leakage of combustible fluids (lubricating oils) and gases (hydrogen 19 cooling) to the plant environs. Actual FIRES and flammable gas build up are appropriately classified 20 via HU2 and HU3. Generator seal damage observed after generator purge does not meet the intent 21 of this EAL be.cause it did not impact normal operation of the plant. This EAL is consistent with the 22' definition of a NOUE while maintaining the anticipatory nature desired and recognizing the risk to 23 non-safety related equipment. Escalation of the. emergency classification is based on potential 24 damage done by missiles generated by the failure or by the radiological releases for a BWR, or in 25' conjunction with a steam generator tube rupture, for a PWR. These latter events would be classified 26 by the radiological ICs or Fission Product Barrier ICs.
27 28-EAL #6 addresses the effect of flooding caused by intemal events such as component failures, 29 equipment misalignment, or outage activity mishaps. The site-specific areas includes those areas 30 that contain systems required for safe shutdown of the plant, that are not designed to be wetted or 31 submerged. Escalation of the emergency classification is based on the damage caused or by access 32 restdctions that prevent necessary plant operations or systems monitoring.
33 34' EAL #7 covers other site-specific phenomena such as hurricane, flood, or seiche. These EALs can 35 also be precursors of more serious events. In particular, sites subject to severe weather as defined in 36 the NUMARC station blackout initiatives, should include an EAL based on activation of the severe 37 weather mitigation procedures (e.g., precautionary shutdowns, diesel testing, staff call-outs, etc.).
38 5-H-3 m
m 4
r 1
HAZARDS AND OTHER CONDITIONS 2
AFFECTING Pl. ANT SAFETY a
HU2 4
Initiating Condition - NOTIFICATION OF UNUSUAL EVENT 5
6 FIRE Within PROTECTED AREA Boundary Not Extinguished Within 15 Minutes of i
7 Detection.
8 9
Operating Mode Applicability:
All 10 11-Example Emergency Action Level:
12 13 1.
FIRE in buildings or areas contiguous to any 'of the following (site-specific) areas not 14 extinguished within 15 minutes of control room notification or verification of a control room 15 alarm:
16 17 (Site-specific) list 18 19 Basis:
20 21 The purpose of this IC is to address the magnitude and extent of FIRES that may be potentially 22 significant precursors to damage to safety systems. As used here, Detection is visual observation 23 and report by plant personnel or sensor alarm indication. The 15 minute time period begins with a 24. credible notification that a FIRE is occurring, or indication of a VALID fire detection system alarm.
25 Verification of a fire detection system alarm includes actions that can be taken with the control room 26 or other nearby site-specific location to ensure that the alarm is not spurious. A verified alarm is 27_
assumed to be an indication of a FIRE unless it is disproved within the 15 minute period by personno!
28 dispatched to the scene. In other words, a personnel report from the scene may be used to disprove 29 a sensor alarm if received within 15 minutes of the alarm, but shall not be required to verify the 30 alarm.
31 32 The intent of this 15 minute duration is to size the FIRE and to discriminate against small FIRES that 33 are readily extinguished (e.g., smoldering waste paper basket). The site-specific list should be limited 34 and applies to buildings and areas contiguous (in actual contact with or immediately adjacent) to 35 plant VITAL AREAS or other significant buildings or areas. The intent of this IC is not to include 36 buildings (i.e., warehouses) or areas that are not contiguous (in actual contact with or immediately 37 adjacent) to plant VITAL AREAS. This excludes FIRES within administration bijildings, waste-basket 38 FIRES, and other small FIRES of no safety consequence.
39 40 Escalation to a higher emergency class is by IC HA4, " FIRE Affecting the Operability of Plant Safety 41 Systems Required for the Current Operating Mode".
42
~ ' ~ -
- ' - ~ ~ ' ' ' ' ~ ~
~ ~ ~P -
5-H-4
s 1
HAZARDS AND OTHER CONDITIONS 2
AFFECTING PLANT SAFETY 3
HU3 4
Initiating Condition - NOTIFICATION OF UNUSUAL EVENT 5
6 Release of Toxic or Flammable Gases Deemed Detrimental to Safe Operation of the 7
Plant.
8 9
Operating Mode Applicability:
All 10 11 Example Emergency Action Levels: (1 or 2) 12 13 1.
Report or detection of toxic or flammable gases that has or could enter the site area boundary 14 in amounts that can affect NORMAL PLANT OPERATIONS.
15 16 2.
Report by Local, County or State Officials for evacuation or sheltering of site personnel based 17 on an offsite event.
18 19 Basis:
20 21 This IC is based on releases in concentrations within the site boundary that will affect the health of 22 plant personnel or affecting the safe operation of the plant with the plant being within the evacuation 23 area of an offsite event (e.g., tanker truck accident releasing toxic gases, etc.).
24 25 Gases within the site boundary that are below life-threatening or flammable concentrations am not 26 applicable to this IC. Concentrations at these levels would not affect plant personnel or the safe 27 operation of the plant. Gases within the site boundary that are above life-threatening or flammeple 28 concentrations, yet have not exceeded those concentrations within a facility structure, would satisfy 29 the first EAL and would require declaration of a NOUE.
30 31 Escalation to an Alert is by IC HA3, " Release of Toxic or Flammable Gases Within a Facility Structure 32 Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or to Establish or 33 Maintain Cold Shutdown'.
34 i
l i
5-H-5
1 HAZARDS AND OTHER CONDITIONS 2-AFFECTING PLANT SAFETY 3
HU4 4
initiating Condition - NOTIFICATION OF UNUSUAL EVENT 5
6 Confirmed Security Event Which Indicates a Potential Degradation in the Level of Safety 7
of the Plant.
8 9
Operating Mode Applicability:
All 10 11 Example Emergency Action Levels:
12 13 1.
Security events as determined from (site-specific) Safeguards Contingency Plan 14 and reported by the (site-specific) security shift supervision 15 16 BASIS:
17 18 This EAL is based on (site-specific) Site Security Plans. Security events which do not represent a 19 potential degradation in the level of safety of the plant, are reported under 10 CFR 73.71 or in some 20 cases under 10 CFR 50.72. Examples of security events that indicate Potential Degradation in the 21 Level of Safety of the Plant are provided below for consideration.
22 23 Consideration should be given to the following events:
24 25 SABOTAGE has or is occurring affecting Safety Related Equipment l
26 HOSTAGE / EXTORTION situation that threatens to interrupt NORMAL PLANT OPERATIONS l
27 CIVIL DISTURBANCE ongoing between the site perimeter (or other site specific nomenclature) e 28 and PROTECTED AREA 29 Hostile STRIKE ACTION at the facility which threatens to interrupt NORMAL PLANT 30 OPERATIONS (judgment based on behavior of Strikers and/or intelligence received) 31 32 BOMB devices discovered within the plant PROTECTED AREA affecting Safety Related Equipment 33 or hostile INTRUSION into the plant PROTECTED AREA would result in EAL escalation to an 54 ALERT.
35 36 l
l i
4 4
4 5-H-6 I
?
1
. HAZARDS AND OTHER CONDITIONS 2
AFFECTING PLANT SAFETY a
HU5 4
initiating Condition - NOTIFICATION OF UNUSUAL EVENT 5
6 Other Conditions Existing Which in the Judgment of the Emergency Director Warrant 7
Declaration of a NOUE.
8 9
Operating Mode Applicability:
All 10 11 Example Emergency Action Level:
12 13 1.
Other conditions exist which in the judgment of the Emergency Director indicate that 14 events are in process or have occurred which indicate a potential degradation of the level of 15 safety of the plant. No releases of radioactive material requiring offsite response or monitoring 16 are expected unless further degradation of safety systems occurs.
17 l
18 Basis:
19 20 This EAL is intended to address unanticipated conditions not addressed explicitly elsewhere but that 21 warrant declaration of an emergency because conditions exist which are believed by the Emergency 22 Director to fall under the NOUE emergency class.
23 24 From a broad perspective, one area that may warrant Emergency Director judgment is related to 25 likely or actual breakdown of site-specific event mitigating actions. Examples to consider include l
26 inadequate emergency response procedures, transient response either unexpected or not l
27 understood, failure or unavailability of emergency systems during an accident in excess of that l
28 assumed in accident analysis, or insufficient availability of equipment and/or support personnel.
29 e
5-H-7
i 1
HAZARDS AND OTHER CONDITIONS
~
2 AFFECTING PLANT SAFETY 3
HA1 4
initiating Condition - ALERT 5
6 Natural and Destructive Phenomena Affecting the Plant VITAL AREA.
7 8
Operating Mode Applicability:
All 9
10 Example Emergency Action Levels: (1 or 2 or 3 or 4 or 5 or 6) 11 12 1.
(Site-Specific) method indicates Seismic Event greater than Operating Basis Earthquake 13 (OBE).
14 15 2.
Tornado or high winds greater than (site-specific) mph within PROTECTED AREA 16 boundary and resulting in VISIBLE DAMAGE to any of the following plant structures or 17 equipment therein or controlindication of degraded performance of those systems.
18 19 Reactor Building 20 Intake Building 21 Ultimate Heat Sink 22 Refueling Water Storage Tank 23 Diesel Generator Building 24 Turbine Building 25 Condensate Storage Tank 26 Control Room 27 Other (Site-Specific) Structures.
28 29 3.
Vehicle crash within PROTECTED AREA boundary and resulting in VISIBLE DAMAGE to 30 any of the following plant structures or equipment therein or control indication of degraded 31 performance of those systems:
32 33 Reactor Building 34 Intake Building 35 Ultimate Heat Sink 36 Refueling Water Storage Tank 37 Diesel Generator Building 38 Turbine Building 39 Condensate Storage Tank 40 Control Room 41 Other (S?.0-S,secific) Structures.
42 43
-t Turbins fa!.we 3.maated missiles result in any VISIBLE DAMAGE to or penetration of e.cf J.
44 the following plant areas: (site-specific) list.
45 46 5.
Uncontrolled flooding in (site-specific) areas of the plant that results in degraded safety system 47 performance as indicated in the control room or that creates industrial safety hazards (e.g.,
48 electric shock) that precludes access necessary to operate or monitor safety equipment.
49 5-H-8
. - -.. -. - - -.. - - _... - - ~. -
_ ~.
r 1
6.
(Site-Specific) occurrences within PROTECTED AREA boundary and resulting in VISIBLE 2
DAMAGE to plant structures containing equipment necessary for safe shutdown, or has caused 3
damage as evidenced by control room indication of degraded performance of those systems.
4' 5
BASIS:
6 7
The EALs in this IC escalate from the NOUE EALs in HU1 in that the occurrence of the event has 8
resulted in VISIBLE DAMAGE to plant structures or areas containing equipment necessary for a safe
-9 shutdown, or has caused damage to the safety systems in those structures evidenced by control 10 indications of degraded system response or performance. The occurrence of VISlBLE DAMAGE 11 and/or degraded system response is intended to discriminate against lesser events. The inkial 12
" report" should not be interpreted as mandating a lengthy damage assessment prior to classification.
13 No attempt is made in this EAL to assess the actual magnitude of the damage. The significance here 14 is not that a particular system or structure was damaged, but rather, that the event was of sufficient 15 magnitude to cause this degradation. Escalation to higher classifications occur on the basis of other 16 ICs (e.g., System Malfunction).
17 18 EAL #1 should be based on site-specific FSAR design basis. Seismic events of this magnitude can 19-result in a plant VITAL AREA being subjected to forces beyond design limits, and thus damage may 20 be assumed to have occurred to plant safety systems. See EPRl-sponsored " Guidelines for Nuclear 21 Plant Response to an Earthquake", dated October 1989, for information on seismic event categories.
22 23 EAL #2 should be based on site-specific FSAR design basis. Wind loads of this magnitude can cause 24 damage to safety functions.
25 26 EAL #s 2,3,4,5 should specify site-specific structures or areas containing systems and functions 27 required for safe shutdown of the plant.
28 29 EAL #3 is intended to address crashes of vehicle types large enough to cause significant damage to 30
. plant structures containing functions and systems required for safe shutdown of the plant.
31 32 EAL #4 is intended to address the threat to safety related equipment imposed by missiles generated 33 by main turbine rotating component failures. This site-specific list of areas should include all areas 34 containing safety-related equipment, their controls, and their power supplies. This EAL is, therefore, 35 consistent with the definition of an ALERT in that if missiles have damaged or penetrated areas 36 containing safety-related equipment the potential exists for substantial degradation of the level of 37 safety of the plant.
38 39 EAL #5 addresses the effect of intemal flooding that has resulted in degraded performance of 40 systems affected by the flooding, or has created industrial safety hazards (e.g., electrical shock) that 41 preclude necessary access to operate or monitor safety equipment. The inability to operate or l
42 monitor safety equipment represents a potential for substantial degradation of the level of safety of l
- 43. the plant. This flooding may have been caused by intemal events such as component failures, 44 equipment misalignment, or outage activity mishaps. The site-specific areas includes those areas or 45 stterged 46 47 EAL #6 covers other site-specific phenomena such as hurricane, flood, or seiche. These EALs can i
48 also be precursors of more serious events.
49 5-H-9
I i
j 1
HAZARDS AND OTHER CONDITIONS 2
AFFECTING PLANT SAFETY I
i 3
HA2 4
Initiating Condition - ALERT 5
6 FIRE or EXPLOSION Affecting the Operability of Plant Safety Systems Required +o 7
Establish or Maintain Safe Shutdown.
8 9
Operating Mode Applicability:
All 10 11 Example Emergency Action Level:
12 13 1.
FIRE or EXPLOSION in any of the following (site-specific) areas:
14 15 (Site-specific) list 16 17 AND 18 19 Affected system parameter indications show degraded performance or plant personnel report 20 VISIBLE DAMAGE to permanent structures or equipment within the specified area.
21 22 Basis:
23 24 Site-specific areas containing functions and systems required for the safe shutdown of the plant 25 should be specified. Site-Specific Safe Shutdown Analysis should be consulted for equipment and 26 plant areas required for the applicable mode. This will make it easier to determine if the FIRE or 27 EXPLOSION is potentially affecting one or more redundant trains of safety systems. Escalation to a 28 higher emergency class, if appropriate, will be based on System Malfunction, Fission Product Barrier 29 Degradation, Abnormal Rad Levels / Radiological Effluent, or Emergency Director Judgment ICs.
30 31 This EAL addresses a FIRE / EXPLOSION and not the degradation in performance of affected 32 systems. System degradation is addressed in the System Malfunction EALs. The reference to 33 damage of systems is used to identify the magnitude of the FIRE / EXPLOSION and to discriminate 34 against minor FIRES / EXPLOSIONS. The reference to safety systems is included to discriminate 35 against FIRES / EXPLOSIONS in areas having a low probability of affecting safe operation. The 36 significance here is not that a safety system was degraded but the fact that the FIRE / EXPLOSION 37 was large enough to cause damage to these systems. Thus, the designation of a, single train was 38 intentional and is appropriate when the FIRE / EXPLOSION is large enough to affect more than one 39-component.
40 41 This situation is not the same as removing equipment for maintenance tnf. a enm :q a piam c 42 Technical Specifications. Removal of equipment for maintenance is a plarmd tdiv?
mr/ S 43 accordance with procedures and, as such, does not constitute a substantial degradation la the ievei 44 of safety of the plant. A FIRE / EXPLOSION !s an UNPLANNED activity and, as such, does 45 constitute a substantial degradation in the level of safety of the plant. In this situation, an Alert 46 classification is warranted.
47 48 The inclusion of a " report of VISIBLE DAMAGE" should not be interpreted as mandating a lengthy 49 damage assessment prior to classification. No attempt is made in this EAL to assess the actual 5-H-10
3 I
i 1
magnitude of the damage. The occurrence of the EXPLOSION with reports of evidence of damage is
'2 sufficient for declaration. The declaration of an Alert and the activation of tne Technical Support 3 - - Center will provide the Emergency Director with the' resources needed to perform these damage 4-assessments. -The Emergency Director also needs to consider.any security aspects of the 5
EXPLOSIONS,if applicable.
6 4
V
^
r l
r l1 l
L l-l l-5-H-11
t e
i 1
HAZARDS AND OTHER CONDITIONS 2
AFFECTING PLANT SAFETY 3
HA3 4~
Initiating Condition - ALERT J
5 6
Release of Toxic or Flammable Gases Within a Facility Structure Which Jeopardizes 7
Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain 8
Cold Shutdown.
9 10 Operating Mode Applicability:
All 11 12 Example Emergency Action Levels: (1 or 2) 13 14 1.
Report or detection of toxic gases within a facility structure in concantrations that will i.e life 15 threatening to plant personnel.
16 17 2.
Report or detection of flammable gases within a facility structure in concentrations that will 18 affect the safe operation of the plant.
19 20 Basis:
21 22 This IC is based on gases that have entered a plant structure affecting the safe operation of the 23 plant. This IC applies to buildings and areas contiguous to plant VITAL AREAS or other significant 24 buildings or areas (i.e., service water pump house). The intent of this IC is not to include buildings 25 (e.g., warehouses) or other areas that are not contiguous or immediately adjacent to plant VITAL 26 AREAS. It is appropriate that increased monitoring be done to ascertain whether consequential 27 damage has occurred. Escalation to a higher emergency class, if appropriate, will be based on 28 System Malfunction, Fission Product Barrier Degradation, Abnormal Rad Levels / Radioactive 29 Effluent, or Emergency Director Judgment ICs.
30 31 Flammable gasses, such as hydrogen and acetylene. are routinely used to maintain plant systems 32 (hydrogen) or to repair equipment / components (acetylene - used in welding). Flammable gasses 33 have a concentration range at which they can ignite / support ccmbustion. An uncontrolled ralease of 34 flammable gasses within a facility structure has the potential to affect safe operation of the plant by 35 limiting either operator or equipment operations due to the potential for ignition and resulting 36 equipment damage / personnel injury. Once it has been determined that an uncontrolled release is 37 occurring, then sampling must be donc to determine if the concentration of the released gas is within 38 this range.
39 40 5-H-12
1 HAZARDS AND OTHER CONDITIONS 2
AFFECTING PLANT SAFETY 3
HA4 4
Initiating Condition - ALERT 5
6 Confirmed Security Event in a Plant PROTECTED AREA.
7 8
Operating Mode Applicability:
All 9
10 Exemple Emergency Action Levets: (1 or 2 or 3) 11 12 1.
BOMB discovered within the PROTECTED AREA potentially affecting (site specific) Safety l
13 Related Equipment.
14 15 2.
INTRUSION into plant PROTECTED AREA by a HOSTILE FORCE.
16 17-3.
Other security events as determined from (site-specific) Safeguards Contingency Plan and 18 reported by the (site-specific) security shift supervision 19 20 BASIS:
21 22 This class of security events represents an escalated threat to plant safety above that contained in 23 the NOUE. A confirmed INTRUSION report is satisfied if physical evidence indicates the presents of 24 a HOSTILE FORCE within the PROTECTED AREA. A BOM3 discovered which has the potential to 25 prevent the plant from achieving and maintaining Cold Shutt own represents a Potential Substantial f
26 Degradation of the Level of Safety of the Plant.
27 28 INTRUSION into a VITAL AREA by a HOSTILE FORCE will escalate this event to a Site Area 29 Emergency.
30 l
5-H-13
t 1
HAZARDS AND OTHER CONDITIONS 2
AFFECTING PLANT SAFETY a
HA5 4
Initiating Condition - ALERT
.5 6
Control Room Evacuation Has Been initiated.
7 8
Operating Mode Applicability:
All 9
10 Example Emergency Action Level:
11 12 1.
Entry into (site-specific) procedure for control room evacuation.
13 14 BASIS:
15 16 With the control room evacuated, additional support, monitoring and direction through the Technical 17 Support Center and/or other emergency response facility is necessary. Inability to establish plant 18 control from outside the control room will escalate this event to a Site Area Emergency.
19 i
l 5-H-14
s 1
HAZARDS AND OTHER CONDITIONS 2
AFFECTING PLANT SAFETY a
HA6 4
Initiating Condition - ALERT 5
6 Other Conditions Existing Which in the Judgment of the Emergency Director Warrant 7
Declaration of an Alert.
8 9
Operating Mode Applicability:
All L
10 l
11 Example Emergency Action Level:
l 12.
13 1.
Other conditions exist which in the judgment of the Emergency Director indicate that events are 14 in process or have occurred which involve actual or likely potential substantial degradation of 15 the level of safety of the plant. Any releases are expected to be limited to small fractions of the 16.
EPA Protective Action Guideline exposure levels.
17 18 Basis:
19 20 This EAL is intended to address unanticipated conditions not addressed explicitly elsewhere but that 21 warrant declaration of an emergency because conditions exist which are believed by the Emergency 22 Director to fall under the Alert emergency class.
l 23 l
l 5-H-15
r
~
~
HAZARDS AND OTHER CONDITIONS 1
2 AFFECTING PLANT SAFETY 3
HS1 4
Initiating Condition - SITE AREA EMERGENCY 5
4 6
Confirmed Security Event in a Plant VITAL AREA.
7 8
Operating Mode Applicability:
All 9
10 Example Emergency Action Levels: (1 or 2) 11 12 1.
INTRUSION into plant VITAL AREA by a HOSTILE FORCE.
13 14 2.
Other security events as determined from (site-specific) Safeguards Contingency Plan and 15 reported by the (site-specific) security shift supervision 16 17 BASIS:
18 19 This class of security events represents an escalated threat to plant safety above that contained in 20 the Alert IC in that a HOSTILE FORCE has progressed from the PROTECTED AREA to a VITAL 21 AREA.
22 23 Loss of Plant Control would escalate this event to a GENERAL EMERGENCY.
5-H-16
,s I'
HAZARDS AND OTHER CONDITIONS 2
AFFECTING PLANT SAFETY 3
HS2 4
. Initiating Condition - SITE AREA EMERGENCY 5
6 Control Room Evacuation Has Been initiated and Plant Control Cannot Be Established.
7 8
Operating Mode Applicability:
All 9
10 Example Emergency Action Level:
11 12 1.
Control room evacuation has been initiated.
13 14 AND 15 16 Control of the plant cannot be established per (site-specific) procedure within (site-specific) 17 minutes.
l 18 19 BASIS:
20 21 Expeditious transfer of safety systems has not occurred but fission product barrier damage may not l~
22 yet be indicated. The intent of this IC is to capture those events where control of the plant cannot be l
23 reestablished in a timely manner. Site-specific time for transfer based on analysis or assessments as l
24
.to how quickly control must be reestablished without core uncovering and/or core damage. This time l
25 should not exceed 15 minutes without additional justification. The determination of whether or not 26 control is established at the remote shutdown panel is based on Emergency Director (ED) judgment.
_27 The ED is expected to make a reasonable, informed judgment within the site-specific time for transfer l
28 that the licensee has control of the plant from the remote shutdown panel.
29 30 The intent of the EAL is to establish control of important plant equipment and knowledge of important 31-plant parameters in a timely manner. Primary emphasis should be placed on those components and 32 instruments that supply protection for and information about safety functions. Typically, these safety 33' functions are reactivity control (ability te shutdown the reactor and maintain it shutdown), reactor
'34 water level (ability to cool the core), and decay heat removal (ability to maintain a heat sink) for a
.35 BWR. The equivalent functions for a PWR are reactivity control, RCS inventory, and secondary heat 36 removal.
37 38 Escalation of this event, if appropriate, would be by Fission Product Barrier Degradation, Abnormal 39 Rad Levels / Radiological Effluent, or Emergency Director Judgment ICs.
40 41 5-H-17
r 1~
HAZARDS AND OTHER CONDITIONS 2
AFFECTING PLANT SAFETY a
HS3 4
Initiating Condition - SITE AREA EMERGENCY 5
6 Other Conditions Existing Which in the Judgment of the Emergency Director Warrant 7
Declaration of Site Area Emergency.
8 9
Operating Mode Applicability:'
All 10 11 Example Emergency Action Level:
12 13 1.
Other conditions exist which in the judgment of the Emergency Director indicate that events are 14 in process or have occurred which involve actual or likely major failures of plant functions 15 needed for protection of the public. Any releases are not expected to result in exposure levels 16 which exceed EPA Protective Action Guideline exposure levels except near the site boundary.
17 18 Basis:
19 20 This EAL is intended to address unanticipated conditions not addressed explicitly elsewhere but that 21 warrant declaration of an emergency because conditions exist which are believed by the Emergency 22 Director to fall under the emergency class description for Site Area Emergency.
23 i
l I
i 5-H-18
1 I
HAZARDS AND OTHER CONDITIONS 2
AFFECTING PLANT SAFETY 3
HG1 4
Initiating Condition - GENERAL EMERGENCY 5
6 Security Event Resulting in Loss Of Physical Control of the Facility.
7 8
Operating Mode Applicability:
All 9
10 Example Emergency Action Level:
11 12 1.
A HOSTILE FORCE has taken control of plant equipment such that plant personnel are unable 13 to operate equipment required to maintain safety functions.
l 14 l
15 BASIS:
l 16 l.
17 This IC encompasses conditions under which a HOSTILE FORCE has taken physical control of 18-VITAL AREAS (containing vital equipment or ccntrols of vital equipment) required to maintain safety 19 functions and control of that equipment cannot tse transferred to and operated from another location.
20 Typically, these safety functions are reactivity control (ability to shut down the reactor and keep it 21 shutdown) reactor water level (ability to cool the core), and decay heat removal (ability to maintain a l
22 heat sink) for a BWR. The equivalent functions for a PWR are reactivity control, RCS inventory, and i
23 secondary heat removal. If control of the plant equipment necessary to maintain safety functions can 24 be transferred to another location, then the above initiating condition is not met.
25 26 This EAL should also address loss of physical control of spent fuel pool cooling systems if imminent i
27 fuel damage is likely (e.g., freshly off-loaded reactor core in pool).
l 28 1
29 Loss of physical control of the control room or remote shutdown capability alone may not prevent the 30 ability to maintain safety functions per se. Site-specific design of the remote shutdown capability and 31 the location of the transfer switches should be taken into account. ' If the safety functions cannot be 32 maintained from the remote shutdown facility without initial control. room action (e.g., reactor trip,
-33 transfer of control, etc.), then both the control room and remote shutdown areas shouid be included 34 in the site-specific areas.
i 35 36 l
...m m -
m
.cn m
5-H-19
r 1
HAZARDS AND OTHER CONDITIONS 2
AFFECTING PLANT SAFETY a
HG2 4
initiating Condition - GENERAL EMERGENCY 5
6 Other Conditions Existing Which in the Judgment of the Emergency Director Warrant 7
Declaration of General Emergency.
8 9
Operating Mode Applicability:
All 10 11 Example Emergency Action Level:
12 13 1.
Other conditions exist which in the judgment of the Emergency Director indicate that events are 14 in process or have occurred which involve actual or imminent substantial core degradation or 15 melting with potential for loss of containment integrity. Releases can be reasonably expected to 16 exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate 17 site area.
18 19 Basis:
20 21 This EAL is intended to address unanticipated conditions not addressed explicitly elsewhere but that 22 warrant declaration of an emergency because conditions exist which are believed by the Emergency 23 Director to fall under the General Emergency class.
24 a
l 5-H-20
Recognition Category S System Maifunction INITIATING CONDITION MATRIX NOUE ALERT SITE AREA EMERGENCY _
GENERAL EMERGENCY SU1 Loss of All Offsite Power to SAS AC power capability to essential SS1 Loss of All Offsite Power and SG1 Prolonged Loss of All Offsite Essential Busces for Greater busses reduced to a single Loss of All Onsite AC Power to Power and Prolonged Loss of All Than 15 Minutes.
power source for greater than 15 Essential Busses.
Onsite AC Power.
Op. Modes:M minutes such that any additional Op. Modes: Power Operation, Op. Modes: Power Operation, single failure would result in Startup, Hot Standby, Hot Startup, Hot Standby, Hot station blackout.
Shutdown Shutdown Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SA2 Failure of Reactor Protection SS2 Failure of Reactor Protection SG2 Failure of the Reactor Protection System Instrumentation to Com-System Instrumentation to Com-System to Complete an Auto-plete or Initiate an Automatic plete or initiate an Automatic matic Scram and Manual Scram Reactor Scram Once a Reactor Reactor Scram Once a Reactor was NOT Successful and There Protection System Setpoint Has Protection System Setpoint Has is Indication of an Extreme Been Exceeded and Manual Been Exceeded and Manual Challenge to the Ability to Cool Scram Was Successful Scram Was NOT Successful.
the Core.
Op. Modes: Power Operation, Op. Modes: Power Operation.
Op. Modes:PowerOperaCon, Startup, Hot Standby Startup Startup SU2 Inability to Reach Required SA3 inability to Maintain Plant in Cold SS4 Complete Loss of Function Shutdown Within Technical Shutdown.
Needed to Achieve or Maintain Specification Limits.
Op. Modes: Cold Shutdown.
Hot Shutdown.
Op. Mode x FMwer Operation, Refueling Op. Modes: Power Operation, Startup, Hot Utandby, Hot Startup, Hot Standby, Hot Shutdown Shutdown SU3 UNPLANNED Loss of Most or All SA4 UNPLANNED Loss of Most or All SS6 Inability to Monitor a Safety Systein Annunciation or Safety System Annunciation or SIGNIFICANT TRANSIENT in Indication in The Control Room Indication in Control Room With Progress.
for Greater Than 15 Minutes Either(1) a SIGNIFICANT Op. Modes: Power Operation, Op. Modes:;bwer Operation, TRANSIENT in Progress, or (2)
Startup Hot Standby, Hot Startup, Hot Standby, Hot Compensatory Non-Alarming Shutdown Shutdown Indicators are Unavailable.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown 5-S-1
Recognition Category S System Malfunction INITIATING CONDITION MATRIX SU7 UNPLANNED Loss of Required SA1 Loss of All Offsite Power and SS3 Lo'ss of All Vital DC Power.
DC Power During Cold Loss of A!! Onsite AC Power to Op. Modes: Power Operation, Shutdown or Refueling Mode for Essential Busses During Cold Startup, Hot Standby, Hot Greater than 15 Minutes.
Shutdown Or Refueling Mode.
Shutdown Op. Modes: Cold Shutdown, Op. Modes: Gold Shutdown, Refueling Refueling Defueled SU4 Fuel Clad Degradation.
Op. Modes: P6wer Operation, Startup, Hot Standby, Hot Shutdown, Cold Shutdown SUS RCS Leakage.
SS5 Loss of Water Levelin the Op. Modes: Power Operation, Reactor Vessel That Has or Ws!!
Startup, Hot.etandby, Hot Uncover Fuelin the Reactor Shutdown Vessel.
Op. Modes: Cold Shutdown, Refueling SU6 UNPLANNED Loss of All Onsite or Offsite Communications Capabilities.
Op. Modes: Ali SU8 Inadvertent Criticality.
Op Modes: Startup, Hot Standby, Hot Shutdown, Cold Shutdown, Refueling 5-S-2 t
~
i 1
SYSTEM MALFUNCTION 2
SU1 3
Initiating Condition - NOTIFICATION OF UNUSUAL EVENT 4
5 Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes.
6 7
Operating Mode Applicability:
All 8
9 Example Emergency Action Level:
10 11 1.
Loss of power to (site-specific) transformers for greater than 15 minutes.
12 13 AND 14-15 At least (site-specific) emergency generators are supplying power to emergency busses.
l
'16 17 Basis:
18 19 Prolonged loss of AC power reduces required redundancy and potentially degrades the level of l
20 safety of the plant by rendering the plant more vulnerable to a complete Loss of AC Power (e.g.,
21 Station Blackout). Fifteen minutes was selected as a threshold to exclude transient or momentary 22 powerlosses.
t 23 l
24 Plants that have the capability to cross-tie AC power from a companion unit may take credit for the 25 redundant power r.ource in the associated EAL for this IC. Inability to effect the cross-tie within 15 26 minutes warrants for declaring a NOUE.
27 l
5-S-3 l
t i
-1 SYSTEM MALFUNCTION l
2 SU2 3
-Initiating Condition - NOTIFICATION OF UNUSUAL EVENT
'4 5'
Inability to Reach Required Shutdown Within Technical Specification Limits.
6 i
7 Operating Mode Applicability:
Power Operation 8
Startup 9
Hot Standby 10-Hot Shutdown 11 12 Example Emergency Action Leveh 13 14 1.
Plant is not brought to required operating mode within (site-specific) Technical Specifications 15 LCO Action Statement Time.
16 17 Basis:
18 19 Limiting Conditions of Operation (LCOs) require the plant to be brought to a required shutdown mode 20 when the Technical Specification required configuration cannot be restored. Depending on the 21 circumstances, this may or may not be an emergency or precursor to a more severe condition. In any 22 case, the initiation of plant shutdown required by the site Technical Specifications requires a one 23 hour2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br /> report under 10 CFR 50.72 (b) Non-emergency events. The plant is within its safety envelope 24 when being shut down within the allowable action statement time in the Technical Specifications. An 25 immediate NOUE is required when the plant is not brought to the required operating mode within the 26 allowable action statement time in the Technical Specifications. Declaration of a NOUE is based on 27 the time at which the LCO-specified action statement time period elapses under the site Technical 1
28 Specifications and is not related to how long a condition may have existed. Other required Technical 29 Specification shutdowns that involve precursors to more serious events are addressed by other 30_
System Malfunction, Hazards, or Fission Product Barrier Degradation ICs.
l 31 i
I 5-S-4
1-SYSTEM MALFUNCTION 2
SUJ 3
Initiating Condition - NOTIFICATION OF UNUSUAL EVENT 4
5-UNPLANNED Loss of Most or All Safety System Annunciation or Indication in The 6
Control Room for Greater Than 15 Minutes 7
8 Operating Mode Applicability:
Power Operation 9-Startup 10 Hot Standby 11 Hot Shutdown 12 13-Example Emergency Action Level:
14 15 1.
UNPLANNED loss of most or all (site-specific) annunciators associated with safety systems for 16 greater than 15 minutes.
17 18 BASIS:
19 20 This IC and its associated EAL are intended to recognize the difficulty associated with monitoring 21' changing plant conditions without the use of a major portion of the annunciation or indication 22 equipment.
23 24 Recognition of the availability of computer based indication equipment is considered (e.g., SPDS, 25 plant computer, etc.).
26 27-Quantification of "Most" is arbitrary, however, it is estimated that if approximately 75% of the safety l.
28 system annunciators or indicators are lost, there is an increased risk that a degraded plant condition 29 could go undetected. It is not intended that plant perso.nnel perform a detailed count of the 30 instrumentation lost but use the value as a judgment threshold for determining the severity of the 31 plant conditions. It is also not intended that the Shift Supervisor be tasked with making a judgment 32 ' decision as to whether additional personnel are required to provide-increased monitoring of system 33 operation.
34 35 - -It is further recognized that most plant designs provide redundant safety system indication powered 36 from separate uninterruptable power supp::,. ;. Wnile failure of a large portion of annunciators is more 37 likely than a failure of a large portion of indications, the concern is included in this EAL due to 38 difficulty associated with assessment of plant conditions. The loss of specific, or several, safety 39 system indicators should remain a function of that specific system or component operability status.
.40.
This will be addressed by the specific Technical Specification. The initiation of a Technical 41 Specification imposed plant shutdown related to the instrument loss will be reported via 10 CFR 50.72.
42 if the shu*down is not in compliance with the Technical Specification action, the NOUE is based on 43-SU2 '1nability to Reach Required Shutdown Within Technical Specification LimitsJ
'44 45 (Site-specific) annunciators or indicators for this EAL must include those identified in the Abnormal 46 Operating Procedures, in the Emergency Operating Procedures, and in other EALs (e.g., area, 47 process, and/or effluent rad monitors, etc.).
48 5-S-5
i t:
1; ~ Fifteen minut:s w:s sel cted as a threshold to exclude transierd or momentary power losses.
2 -Due to the limited number of safety systems in operation during cold shutdown, refueling, and 3
defueled modes, no IC is indicated during these modes of operation.
4 This NOUE wid be escalated to an Alert if a transient is in progress during the loss 'of annunciation or 5
l 6
indication.
7-1
\\
t i
l-5-S-6
1 SYSTEM MALFUNCTION 2
SU4 3
initiating Condition - NOTIFICATION OF UNUSUAL EVENT 4
5 Fuel Clad Degradation.
6 7
Operating Mode Applicability:
Power Operation 8
Startup 9
Hot Standby 10 Hot Shutdown 11 Cold Shutdown 12 13 Example Emergency Action Levels: (1 or 2) 14 15 1.
(Site-specific) radiation monitor readings indicating fuel clad degradation greater than Technical 16 Specification allowable limits.
17 18 2.
(Site-specific) coolant sample activity value indicating fuel clad degradation greater than 19 Technichl Specification allowable limits.
20 21 Basis:
22 23 This IC is included as a NOUE because it is considered to be a potential degradation in the level of 24 safety of the plant and a potential r -Oursor of more serious problems. EAL #1 addresses site-25 specific radiation monitor readings sucr. as BWR air ejector monitors, PWR failed fuel monitors, etc.,
26 that provide indication of fuel clad integrity. EAL #2 addresses coolant samples exceeding coolant 27 technical specifications for lodine spike. Escalation of this IC to the Alert level is via the Fission 28 Product Barrier Degradation Monitoring ICs. Though the referenced Technical Specification limits 29 are mode dependent, it is appropriate that the EAL's be applicable in all modes, as they indicate a 30 potential degradation in the level of safety of the plant.
31 32 33 5-S-7
t 1
SYSTEM MALFUNCTION 2
SUS 1
3 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT 4
5 RCS Leakage.
6 7
Operating Mode Applicability:
Power Operation 8
Startup 9
Hot Standby 10 Hot Shutdown 11 12 Example Emergency Action Levels: (1 or 2) 13 14 1.
Unidentified or pressure boundary leakage greater than 10 gpm.
15 16 2.
Identified leakage greater than 25 gpm.
17 18 BASIS:
19-20 This 10 is included as a NOUE because it may be a precursor of more serious conditions and, as 21 result, is considered to be a potential degradation of the level of safety of the plant. The 10 gpm 22 value for the unidentified and pressure boundary leakage was selected as it is observable with 23 normal control room indications. Lesser values must generally be determined through time-24 consuming surveillance tests (e.g., mass balances). The EAL for identified leakage is set at a higher 25 value due to the lesser significance of identified leakage in comparison to unidentified or pressure 26 boundary leakage. In either case, escalation of this IC to the Alert levelis via Fission Product Barrier 27 Degradation ICs.
28 I
w 5-S-8
s
~
~
1 SYSTEM MALFUNCTION 2
SU6 3
Initiating Condition - NOTIFICATION OF UNUSUAL EVENT 4
5 UNPLANNED Loss of All Onsite or Offsite Communications Capabilities.
6 7
Operating Mode Applicability:
All 8
9 Example Emergency Action Levels: (1 or 2) 10 11 1.
Loss of all (site-specific list) onsite communications capability affecting the ability to perform 12 routine operations.
13
-14 2.
Loss of all (site-specific list) offsite communications capability.
15 16' Basis:
17 18 The purpose of this IC and its associated EALs is to recognize a loss of communications capability 19 that either defeats the plant operations staff ability to perform routine tasks necessary for plant 20 operations or the ability to communicate problems with offsite authorities. The loss of offsite 21 communications ability is expected to be significantly more comprehensive than the condition 22 addressed by 10 CFR 50.72.
'23 24 The availability of one method of ordinary offsite communications is sufficient to inform state and 25
. local authorities of plant problems. This EAL is intended to be used only when extraordinary means l
26 (e.g., relaying of information from radio transmissions, individuals being sent to offsite locations, etc.)
L 27 are being utilized to make communications possible.
l 28 l
'29 Site-specific list for onsite communications loss must encompass the loss of all means of routine 30 communications (e.g., phones, sound powered phone systems, page party system and radios /
3132 _ walkie talkies).
33 Site-specific list for offsite communications loss must encompass the loss of all means of 34 communications with offsite authorities. This should include the ENS, Bell lines, FAX transmissions, 35 and dedicated phone systems.
36 P
b 5-S-9 L- -
e 1
SYSTEM MALFUNCTION 2
SU7 3
Initiating Condition - NOTIFICATION OF UNUSUAL EVENT i
4 5-UNPLANNED Loss of Required DC Power During Cold Shutdown or Refueling Mode for 6
Greater than 15 Minutes, 7
L 8
OPERATING MODE APPLICABILITY:
Cold Shutdown 9
Refueling 10 l
11 EXAMPLE EMERGENCY ACTION LEVEL:
l 12 i
13
- 1. Either of the following conditions exist:
14
'15
16 voltage indications, i
17 18 AND 19 i
20
- b. Failure to restore power to at least one required DC bus within 15 minutes from the time of 21 loss.
22 23 BASIS:
24 25 The purpose of this IC and its associated EALs is to recognize a loss of DC power compromising the 26 ability to monitor and control the removal of decay heat during Cold Shutdown or Refueling 27 operations. This EAL is intended to be anticipatory in as much as the operating crew may not have 28 necessary indication and control of equipment needed to respond to the loss.
29 i
30 UNPLANNED is included in this IC and EAL to preclude the declaration of an emergency as a result 31 of planned maintenance activities. Routinely plants will perform maintenance on a Train related 32 basis during shutdown periods. It is intended that the loss of the operating (operable) train is to be 33 considered. If this loss results in the inability to maintain cold shutdown, the escalation to an Alert will 34 be per SA3 " inability to Maintain Plant in Cold Shutdown."
35 36
. (Site-specific) bus voltage should be based on the minimum bus voltage necessary for the operation 37 of safety related equipment. This voltage value should incorporate a margin of at least 15 minutes of 38 operation before the onset of inability to operate those loads. This voltage is usually near the L
39 minimum voltage selected when battery sizing is performed. Typically the value for the entire battery 40 set is approximately 105 VDC. For a 60 cell string of batteries the cell voltage 1.75 Volts per cell.
41 For a 58 string battery set the minimum voltage is typically 1.81 Volts per cell.
a o
5-S-10
s 1
SYSTEM MALFUNCTION 2
SU8 3
Initiating Condition - NOTIFICATION OF UNUSUAL EVENT 4
5 Inadvertent Criticality.
6 7
OPERATING MODE APPLICABILITY Startup 8
Hot Standby 9
Hot Shutdown 10 Cold Shutdown 11 Refueling 12 13 EXAMPLE EMERGENCY ACTION LEVEL 14-15'
- 1. An extended and UNPLANNED positive period or sustained positive startup rate 16 observed on nuclear instrumentation 17 18 BASIS 19 20 This IC addresses inadvertent criticality events. While the primary concem of this IC is criticality 21. events that occur in Cold Shutdown or Refueling modes, the IC is applicable in other modes in which
~22 inadvertent criticalities are possible. This IC indicates a potential degradation of the level of safety of 23 the plant, warranting a NOUE classification. This IC excludes inadvertent criticalities that occur during 24 planned reactivity changes associated with reactor startups (e.g., criticallity earlier than estimated).
25 26 This condition can be identified using period monitors / startup rate monitor. The term " extended" is 27 used in order to allow exclusion of expected short term positive periods / startup rates from planned 28 fuel bundle or control rod movements during core alteration for PWRs and BWRs. These short term 29 positive periods / startup rates are the result of the increase in neutron population due to suberitical
'30 multiplication.
31 32 Escalation would be by the Fission Product Barrier Matrix, as apprdpriate to the operating mode at 33 the time of the event, or by Emergency Director Judgment.
5-S-11
lt I
4 L
i.~*
1 SYSTEM MALFUNCTION 2
SA1 3
Initiating Condition - ALERT 4
5 Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses During 6
Cold Shutdown Or Refueling Mode.
7 8
Operating Mode Applicability:
Cold Shutdown 9
Refueling 10 Defueled 11 12 Example Emergency Action Level:
13 14 1.
Loss of power to (site-specific) transformers.
15-16 AND 17 18 Failure of (site-specific) emergency generators to supply power to emergency busses.
19 20 AND 21 22 Failure to restore power to at least one emergency bus within 15 minutes from the time of loss 23 of both offsite and onsite AC power.
24 25 Basis:
26 27' Loss of all AC power compromises all plant safety systems requiring electric power including RHR, 28 ECCS, Containment Heat Removal, Spent Fuel Heat Removal and the Ultimate Heat Sink. When in 29 cold shutdown, refueling, or defueled mode the event can be classified as an Alert, because of the 30 significantly reduced decay heat, lower temperature and pressure, increasing the time to restore one 31 of the emergency busses, relative to that specified for the Site Area Emergency EAL. Escalating to 32 Site Area Emergency, if appropriate, is by Abnormal Rad Levels / Radiological Effluent, or 33 Emergency Director Judgment ICs. Fifteen minutes was selected as a threshold to exclude transient 34 or momentary power losses.
35 36 Consideration should be given to operable loads necessary to remove decay heat or provide Reactor 37 Vessel makeup capability when evaluating loss of AC power to essential busses. Even though an 38 essential bus may be energized, if necessary loads (i.e., loads that if lost would inhibit decay heat 39 removal capability or Reactor Vessel makeup capability) are not operable on the energized bus then 40 the bus should not be considered operable.
41 l
l l
I 5-S-12
s 1
SYSTEM MALFUNCTION 2
SA2 3
Initiating Condition - ALERT 4
5 Failure of Reactor Protection System Instrumentation to Complete or Initiate an 6-Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been 7
Exceeded and Manual Scram Was Successful.
8 9
Operating Mode Applicability:
Power Operation 10 Startup 11 Hot Standby 12 13 Example Emergency Action Level:
14 15 1.
Indication (s) exist that indicate that reactor protection system setpoint was exceeded and 16 automatic scram did not occur, and a successfui manual scram occurred.
17 18 Basis:
19 l
20 This condition indicates failure of the automatic protection system to scram the reactor. This 21 condition is more than a potential degradation of a safety system in that a front line automatic 22 protection system did not function in response to a plant transient and thus the plant safety has been 23 compromised, and design limits of the fuel may have been exceeded. An Alert is indicated because 24 conditions exist that lead to potential loss of fuel clad or RCS. Reactor protection system setpoint 25 being exceeded, rather than limiting safety system setpoint being exceeded, is specified here 26 because failure of the automatic protection system is the issue. A manual scram is any set of actions 27 by the reactor operator (s) at the reactor control console which causes control rods to be rapidly 28 inserted into the core and brings the reactor suberitical (e.g., reactor trip button, Altemate Rod 29 Insdrtion): Failure of manual scram would escalate the event to a Site Area Emergency. -
30 5-S-13
.~.
1 SYSTEM MALFUNCTION 2
SA3 3
Initiating Condition - ALERT 4
5 Inability to Maintain Plant in Cold Shutdown.
6 7
OPERATING MODE APPLICABILITY:
Cold Shutdown 8
Refueling 9
10 EXAMPLE EMERGENCY ACTION LEVEL:
11 12-
- 1. The following conditions exist:
13 14
- a. Loss of (site-specific) Technical Specification required functions to maintain cold shutdown.
15 16 AND 17 18
- b. Temperature increate that either:
19 20 Exceeds Technical Specification cold shutdown temperature limit
=
21 22 OR 23 24 Results in uncontrolled temperature rise approaching cold shutdown technical
=
25 specification limit.
26 27-BASIS:
28 29 This EAL addresses complete loss of functions required for core cooling during refueling and cold 30 shutdown modes. Escalation to Site Area Emergency or General Emergency would be via Abnormal 31' Rad Levels / Radiological Effluent or Emergency Director Judgement ICs.
32 33 For PWRs, this IC and its associated EAL are based on concems raised by Generic Letter 88-17, 34
" Loss of Decay Heat Removal." A number of phenomena such as pressurization, vortexing, steam 35 generator U-tube draining, RCS level differences when operating at a mid-loop condition, decay heat 36 removal system design, and level instrumentation problems can lead to conditions where decay heat 37 removal is lost and core uncovery can occur. NRC analyses show that sequences that can cause 38 core uncovery in 15 to 20 minutes and severe core damage within an hour after decay heat removal 39 is lost. Under these conditions, RCS integrity is lost and fuel clad integrity is lost or potentially lost, 40 which is consistent with a Site Area Emergency. (Site-specific) indicators for these EALs are those
.41
- methods used by the plant in response to Generic Letter 88-17 which include core exit temperature 42 monitoring and RCS water level monitoring. In addition, radiation monitor readings may also be 43 appropriate as an indicator of this condition.
44-45
" Uncontrolled" means that system temperature increase is not the result of planned actions by the
.46 plant staff.
47 5-S-14
s
'1 The EAL guidance related to uncontrolled temperature rise'is necessary to preserve the anticipatory 2
philosophy of NUREG-0654 for events starting from temperatures much lower than the cold 3
shutdown temperature limit.
4 5
A loss of Technical Specification components alone is not intended to constitute an Alert. The same 6
is true of a momentary UNPLANNED excursion above 200'F when the heat removal function is 7
available. Separate statements (1a and ib) are included to recognize additional plant capability to 8
maintain cooling of the reactor.
9 10' Escalation to the Site Area Emergency is by IC SS5, " Loss of Water Level in the Reactor Vessel That 11 Has or Will Uncover Fuel in the Reactor Vessel," or by Abnormal Rad Levels / Radiological Effluent 12 ICs.
13 14 Multi-unit stations with shared safety functions should further consider how this IC may affect more 15 than one unit and how this may be a factor in escalating the emergency class.
I l '
I i
t 5-S-15 L
l 't 1
SYSTEM MALFUNCTION 2
SA4 3
Initiating Condition - ALERT i
4 5
UNPLANNED Loss of Most or All Safety System Annunciation or Indication in Control 6
Room With Either (1) a SIGNIFICANT TRANSIENT in Progress, or (2) Compensatory Non-7 Alarming Indicators are Unavailable.
t 8
9 Operating Mode Applicability:
Power Operation l
-10 Startup 11 Hot Standby 12 Hot Shutdown 13 14 Example Emergency Action Level:
15 16 1.
UNPLANNED loss of most or all (site-specific) annunciators associated with cafety systems for 17 greater than 15 minutes.
18 19 AND 20 21 Either of the following: (a or b) 22 23 a.
A SIGNIFICANT TRANSIENT is in progress.
24 25 OR 26 27 b.
Compensatory non-alarming indications are unavailable.
28 29 BASIS:
30 31 This IC and its associated EAL are intended to recognize the difficulty associated with monitoring 32 changing plant conditions without the use of a major portion of the annunciation or indication 33 equipment during a transient. Recognition of the availability of computer based indication equipment 34 is considered (e.g., SPDS, plant computer, etc.).
35 36
" Planned" loss of annunciators or indicators includes scheduled maintenance and testing activities.
37 38-Quantification of "Most" is arbitrary, however, it is estimated that if approximately 75% of the safety 39 cystem annunciators or indicators are lost, there is an increased risk that a degraded plant condition 40 could go undetected. It is not intended that plant personnel perform a detailed count of the 41 instrumentation lost but use the value as a judgment threshold for determining the severity of the 42 plant conditions. It is also not intended that the Shift Supervisor be tasked with making a judgment 43 decision as to whether additional personnel are required to provide increased monitoring of system 44 operation.
45 46 It is further recognized that most plant designs provide redundant safety system indication powered 47 from separate uninterruptable power supplies. While failure of a large portion of annunciators is more l
48 likely than a failure of a large portion of indications, the concem is included in this EAL due to 5-S-16
i I
difficulty associated with assessment of plant conditions. The loss of specific, or several, safety 2
system indicetors should remain a function of that specific system or component operability status.
3 This will be addressed by the specific Technical Specification. The initiation of a Technical 4
Specification imposed plant shutdown related to the instrument loss will be reported via 10 CFR
-5 50.72. If the shutdown is not in compliance with the Technical Specification action, the NOUE is 6
based on SU2 " inability to Reach Required Shutdown Within Technical Specification Limits."
7 8
Site-specific annunciators or indicators for this EAL must include those identified in the Abnormal 9
Operating Procedures, in the Emergency Operating Procedures, and in other EALs (e.g., area, 10 process, and/or effluent rad monitors, etc.).
11 12
" Compensatory non-alarming indications" in this context includes computer based information such 13 as SPDS. This should include all computer systems available for this use depending on specific plant 14 design and subsequent retrofits. If both a major portion of the annunciation system and all computer 15 monitoring are unavailable, the Alert is required.
16 17 Due to the limited number of safety systems in operation during cold shutdown, refueling and 18 defueled modes, no IC is indicated during these modes of operation.
19 20 This Alert will be escalated to a Site Area Emergency if the operating crew cannot monitor the 21 transient in progress.
22 i
l l
i
.v s.w 4.d*
e.
swr-5-S-17
! e e
1 SYSTEM MALFUNCTION l
SAS 2
3
. initiating Condition - ALERT l
4 5
AC power capability to essential busses reduced to a single power source for greater G
than 15 minutes such that any additional single failure would result in station blackout.
7 8
Operating Mode Applicability:
Power Operation
~9 Startup 10 Hot Standby 11 Hot Shutdown l
12 13 Example Emergency Action Level:
14 15 1.
AC power capability to site-specific essential' busses reduced to a single power source for 16 greater than 15 minutes
'l 17 18 AND 19 20 Any additional single failure will result in station blackout.
21 22 Basis:
23 l
24 This IC and the associated EALs are intended to provide an escalation from IC SU1, " Loss of All l
25 Offsite Power To Essential Busses for Greater Than 15 Minutes." The condition indicated by this IC l
26 is the degradation of the offsite and onsite power systems such that any additional single failure 27' would result in a station blackout. This condition could occur due to a loss of offsite power with a 28 concurrent failure of one emergency generator to supply power to its emergency busses. Another 29 related condition could be the loss of all offsite power and loss of onsite emergency diesels with only i
30 one train of emergency busses being backfed from the unit main generator, or the loss of onsite 31 emergency diesels with only one train of emergency busses being backfed from offsite power. The 32 subsequent loss of this single power source would escalate the event to a Site Area Emergency in 33 accordance with IC SS1, " Loss of All Offsite and Loss of All Onsite AC Power to Essential Busses."
34 35 At multi-unit stations, the EALs should allow credit for operation of installed design features, such as 36 cross-ties or swing diesels, provided that abnormal or emergency operating procedures address their 37 use. However, these stations must also consider the impact of this condition on other shared safety 38 functions in developing the site specific EAL.
-39 l
5-S-18
i :
a
'1-SYSTEM MALFUNCTION SS1 2
3 Initiating Condition - SITE AREA EMERGENCY L4 5;
Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses.
6' 7-Operating Mode Applicability:
Power Operation 8
Startup 9
Hot Standby 10
- Hot Shutdown 11-12_
Example Emergency Action Level:
13=
14 1.
Loss of power to (site-specific) transformers.
~15-16' AND 17 18-Failure of (site-specific) emergency generators to supply power to emergency busses.
gg 20-AND l 22 Failure to restore power to at least one emergency bus within (site-specific) minutes from the 23 time of loss of both offsite and onsite AC power.
24 25-Basis:
26-27--. Loss of all AC power compromises all plant safety systems requiring electric power including RHR, 28 ECCS, Containment Heat Removal and the Ultimate Heat Sink. Prolonged loss of all AC power will 29-cause. core uncovering and loss of containment integrity, thus this event can escalate to a General 30~ Emergency. The (site-specific) time duration should be selected to ' exclude transient or momentary
~31 power losses, but should not exceed 15 minutes.
32 33 - Escalation to Ger.eral Emergency is via Fission Product Barrier Degradation or IC SG1, " Prolonged 34 Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power."
35 L
36 L - Consideration should be given to operable loads necessary to remove decay heat or provide Reactor 37 Vessel makeup capability when evaluating loss of AC power to essential busses..Even though an
-38 essential bus may be energized, if necessary loads (i.e., loads that if lost would inhibit decay heat
.39 removal capability or Reactor Vessel makeup capability) are not operable on the energized bus then
.40
.the bus should not be considered operable. If this bus was the only energized bus then a Site Area 41:. Emergency per SS1 should be declared.-
42 L
5-S-19
,. ~
4 I
i 1
SYSTEM MALFUNCTION 2
SS2 3
Initiating Condition - SITE AREA EMERGENCY I
4 1
5 Failure of Reactor Protection System Instrumentation to Complete or Initiate an 1
6 Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been 7
Exceeded and Manual Scram Was NOT Successful.
8 9
Operating Mode Applicability:
Power Operation 10 Startup j
11
]
12 Example Emergency Action Level:
i 13 14 1.
Indication (s) exist that automatic and manual scram were not successful.
15 16 Basis:
17 l
18 Automatic and manual scram are not considered successful if action away from the reactor control 19 console was required to scram the reactor.
20
)
1 21 Under these conditions, the reactor is producing more heat than the maximum decay heat load for
{
22 which the safety systems are designed. A Site Area Emergency is indicated because conditions erist 23 that lead to imminent loss or potential loss of both fuel clad and RCS. Although this IC may be 24 viewed as redundant to the Fission Product Barrier Degradation IC, its inclusion is necessary to l
25 better assure timely recognition and emergency response. Escalation of this event to a General i
26 Emergency would be via Fission Product Barrier Degradation or Emergency Director Judgment ICs.
]
i I
i
- i 1
I l
5-S-20 l
, - - - il l
v 1
'1 SYSTEM MALFUNCTION 2
SS3 3
Initiating Condition - SITE AREA EMERGENCY 4
5 Loss of All Vital DC Power.
6 7
' Operating Mode Applicability:
Power Operation 8
Startup 9
Hot Standby 10 Hot Shutdown 11 12 Example Emergency Action Level:
13 14 1.
Loss of All Vital DC Power based on (site-specific) bus voltage indications for greater than 15 15 minutes.
16 17 Basis:
18 19 Loss of all DC power compromises ability to monitor and control plant safety functions. Prolonged 20 loss of all DC power will cause core uncovering and loss of containment integrity when there is 21 significant decay heat and sensible heat in the reactor system. Escalation to a General Emergency 22 would occur by Abnormal Rad Levels / Radiological Effluent, Fission Product Barrier Degradation, or 23 Emergency Director Judgment ICs. Fifteen minutes was sele'cted as a threshold to exclude transient 24 or momentary power losses.
25 l
l l.
L 5-S-21 1
l L
~. -
I e
1 SYSTEM MALFUNCTION 2
SS4 3
initiating Condition -- SITE AREA EMERGENCY 4
5 Complete Loss of Heat Removal Capability.
6 7
Operating Mode Applicability:
Power Operation 8
Startup 9
Hot Standby 10 Hot Shutdown 11 l
12 Example Emergency Action Leveh 13 14 1.
Loss of core cooling and heat sink (PWR),
15 16 1.
Heat Capacity Temperature Limit Curve exceeded (BWR).
17 18 Basis:
19 l
20 This EAL addresses complete loss of functions, including ultimate heat sink, required for hot 21 shutdown with the reactor at pressure and temperature. Reactivity control is addressed in other 22 EALs. For BWRs the loss of heat removal function is indicated by the Heat Removal Capability l
23 Temperature Limit Curve being exceeded.
24 25 Under these conditions, there is an actual major failure of a system intended for protection of the 26 public. Thus, declaration of a Site Area Emergency is warranted. Escalation to General Emergency 27 would be via Abnormal Rad Levels / Radiological Effluent, Emergency Director Judgment, or Fission 28 Product Barrier Degradation ICs.
29 I
i 4
s 4
5-S-22 4
'l
)
~
'1 SYSTEM MALFUNCTION 2
SS5 3-Initiating Condition - SITE AREA EMERGENCY 4
5 Loss of Water Level in the Reactor Vessel That Has or Will Uncover Fuel in the Reactor 6
Vessel.
7
- 8. LOPERATING MODE APPLICABILITY:
Cold Shutdown 9-
. Refueling
'10
- 11' EXAMPLE EMERGENCY ACTION LEVEL:-
{12 13
- 1. Loss of Reactor Vessel Water Level as indicated by:
.14 Loss of all' decay heat removal cooling as determined by (site-specific) procedure.
15 a.
16
.17 AND 18 19
. b.~ (Site-specific) indicators that the core is or will be uncovered.
20 l
~21. BASIS:
Udder the conditions specified by this IC, severe core damage can occur and reactor coolant system 24-pressure boundary integrity may not be assured. For BWRs, it is intended to address concems
'25 raised by NRC Office for Analysis and Evaluation of Operational Data (AEOD) Report AEOD/EG09, l
26
. "BWR Operating Experience involving inadvertent Draining of the Reactor Vessel," dated August 8, 27 1986. This report states:
28 29-In broadest terms, the dominant causes of inadvertent reactor vessel draining are related to 30 the operational and design problems associated with the residual heat removal system when 31 -.
It is entering into or exiting from the shutdown cooling mode'. During this transitional period 32
- water is drawn from the reactor vessel, cooled by the residual heat removal system heat 33, exchangers (from the cooling provided by the service water system), and retumed to the 134
. reactor vessel. First, there are piping and valves in the residual heat removal system which 35 are common to both the shutdown cooling mode and other modes of operation such as low 36 pressure coolant injection and suppression pool cooling. These valves, when improperly 37-positioned, provide a drain path for reactor coolant to flow from the reactor vessel to the L
38 suppression pool or the radwaste system. Second, establishing or exiting the shutdown L
39 cooling mode of operation is entirely manual, makirig such evolutions vulnerable to personnel
_40_
and procedura; errors. Third, there is no comprehensive valve interlock arrangement for all 41 the residual heat removal system valves that could be activated during shutdown cooling.
. Collectively, these factors have contributed to the repetitive occurrences of the operational 42 43 evente involving the inadvertant draining of the reactor vessel.
44 45; For PWRs, this IC covers sequences such as prolonged boiling following loss of decay heat removal.
.47 Thus, declaration of a Site Area Emergency is warranted under the conditions specified by the IC.
48L Escalation to a general emergency is via radiological effluent IC AG1.
5-S-23 W
t' et
=
~_
t
~
- 1 SYSTEM MALFUNCTION 2
SS6 3
Initiating Condition - SITE AREA EMERGENCY
-4 5
Inability to Monitor a SIGNIFICANT TRANSIENT in Progress.
4 6
7 Operating Mode Applicability:
Power Operation 8-Startup 9
Hot Standby -
10 Hot Shutdown 11 12 Example Emergericy Action Level:
13 14 1.
Loss of most or all (site-specific) annunciators associated with safety systems.
15 16 AND 17 18 Compensatory non-alarming indications are unavailable.
19 20 AND i
21 22 Indications needed to monitor (site-specific) safety functions are unavailable.
23 24 AND 4
25 26 SIGNIFICANT TRANSIENT in progress.
27 28 BASIS:
1 29 30 This IC and its associated EAL are intended to recognize the inability of the control room staff to 31 monitor the plant response to a transient. A Site Area Emergency is considered to exist if the control 32 room staff cannot monitor safety functions needed for protection of the public.
33 34 (Site-specific) annunciators for this EAL should be limited to include those identified in the Abnormal 4
^
35 Operating Procedures, in the Emergency Operating Procedures, and in other EALs (e.g., rad 36 monitors, etc.)
37 38
" Compensatory non-alarming indications" in this context includes computer based information such 39 as SPDS. This should include all computer systems available for this use depending on specific plant 40 design and subsequent retrofits.
41
~42 (Site-specific) indications needed to monitor safety functions necessary for protection of the public 43 must include control room indications, computer generated indications and dedicated annunciation 44 capability. The specific indications should be those used to determine such functions as the ability to 45 shut down the reactor, maintain the core cooled, to maintain the reactor coolant system intact, and to 46 maintain containment intact.
47~
5-S-24
t I
1" Planned" and " UNPLANNED" actions are not differentiated since the loss of instrumentation of this 2 '. magnitude is of such significance during a transient that the cause of the loss is not an ameliorating
' 3 factor.
l 9
-- w t..
a 5-S-25
e i
1 SYSTEM MALFUNCTION 2
SG1 3
4 initiating Condition - GENERAL EMERGENCY 5
6 Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power.
~
7 8
Operating Mode Applicability:
Power Operation 9
Startup 10 Hot Standby 11 Het Shutdown 12 4
13 Example Emergency Action Level:
14 15 1.
Loss of power to (site-specific) transformers.
16 17 AND l
18 19 Failure of (site-specific) emergency diesel generators to supply power to emergency busses.
20 21 AND 22 23 Either of the following: (a or b) 24 25 a.
Restoration of at least one emergency bus within (site-specific) hours is ptot likely q
26 27 OR 28 29 b.
(Site-Specific) Indication of continuing degradation of core cooling based on Fission 30 Product Barrier monitoring.
31 32 Basis:
33-34 Loss of all AC power compromises all plant safety systems requiring electric power including RHR, 35 ECCS, Containment Heat Removal and the Ultimate Heat Sink. Prolonged loss of all AC power will 36 lead to loss of fuel clad, RCS, and containment. The (site-specific) hours to restore AC power can be 37 based on a site blackout coping analysis performed in conformance with 10 CFR 50.63 and 38 Regulatory Guide 1.155, " Station Blackout," as available, with appropriate allowance for offsite 39 emergency response. Although this IC may be viewed as redundant to the Fission Product Barrier 40 Degradation IC, its inclusion is necessary to better assure timely recognition and emergency 41 response.
42 43 This IC is specified to assure that in the untiloly event of a prolonged station blackout, timely 44 recognition of the seriousness of the event occurs and that declaration of a General Emergency 45 occurs as early as is appropriate, based on a reasonable assessment of the event trajectory.
~46 5-S-26
i 1
The likelihood of restoring at least one emcrgency bus should be based on a realistic appraisal of the 2
situation since a delay in an upgrade decision based on only a chance of mitigating the event could 3
' result in a loss of valuable time in preparing and implementing public protective actions.
'4 In addition, under these conditions, fission product barrier monitoring capability may be degraded.
5 Although it may be difficult to predict when power can be restored, it is necessary to give the 6
Emergency Director a reasonable idea of how quickly (s)he may need to declare a General 7-Emergency based on two major considerations:
8 9
1.
Are there any present indications that core cooling is already degraded to the point that Loss or 10.
Potential Loss of Fission Product Barriers is imminent? (Refer to Tables 3 and 4 for more 11 information.)
12 l
13 2.
If there are no present indications of such core cooling degradation, how likely is it that power 14-can be restored in time to assure that a loss of two barriers with a potential loss of the third 15 barrier can be prevented?
16 l
17 Thus, ir.fc ton of continuing core cooiing degradation must be based on Fission Product Barrier 18 monitoring with particular emphasis on Emergency Director judgment as it relates to imminent Loss 19 or Potential Loss of fission product barriers and degraded ability to monitor fission product barriers.
20 l
l l
l
+ MOT 'Te e
. WO,.
e emu
- e.,F g
5-S-27
1 SYSTEM MALFUNCTION 2
SG2 3
Initiating Condition - GENERAL EMERGENCY 4
5 Failure of the Reactor Protection System to Complete an Automatic Scram and Manual 6
Scram was NOT Successful and There is indication of an Extreme Challenge to the 7
Ability to Cool the Core.
8 9
Operating Mode Applicability:
Power Operation 10 Startup 11 12 Example Emergency Action Level:
13 14 1.
Indications exist that automatic and manual scram were not successful.
15 16 AND 17-18 Either of the following: (a or b) 19 20
- a. Indication (s) exists that the core cooling is extremely challenged.
21 22 OR 23 24
- b. Indication (s) exists that heat removal is extremely challenged.
25 26 Basis:
27 28 Automatic and manual scram are not considered successful if action away from the reactor control 29 console is required to scram the reactor.
30 31 Under the conditions of this IC and its associated EALs, the efforts to bring the reactor suberitical 32 have been unsuccessful and, as a result, the reactor is producing more heat than the maximum 33 decay heat load for which the safety systems were designed. Although there are capab!ities away 34 from the reactor control console, such as emergency boration in PWRs, or standby liquid control in 35 BWRs, the continuing temperature rise indicates that these capabilities are not effective. This 36 situation could be a precursor for a core melt sequence.
37 1
38 For PWRs, the extreme challenge to the ability to cool the core is intended to mean that the core exit 39 temperatures are at or approaching 1200 degrees F or that the reactor vessel water level is below the 40 top of active fuel. For plants using CSFSTs, this EAL equates to a Core Cooling RED condition and 41 an entry into function restoration procedure FR-S.1. For BWRs, the extreme challenge to the ability 42 to cool.the core is intended to mean that the reactor vessel water level cannot be restored and 43 maintained above Minimum Steam Cooling RPV Water Level.
44 45 Another consideration is the inability to initially remove heat during the early stages of this sequence.
46 For PWRs, if emergency feedwater flow is insufficient to remove the amount of heat required by 47 c'asign from at least one steam generator, an extreme challenge should be considered to exist. For 48 plants using CSFSTs, this EAL equates to a Heat Sink RED condition. For BWRs, considerations 5-S-28
i 1
. include inability to remove heat via the main condenser, or via the suppression pool or torus (e.g.,
2-due to high pool water temperature).
3-4-
In the event either of these challenges exist at a time tbst the reactor has not been brought below the 5
power associated with the safety system design (typically 3 to 5% power) a core melt sequence 6
exists, in this situation, core degradation can occur rapidly. For this reason, the General Emergency 7'
declaration is intended to be anticipatory of the fission product barrier matrix declaration to permit 8
maximum offsite intervention time.
e P
e n-p.'*g.
.
- g I
5-S-29
t 1
Appendix A 2
Basis for Radiological Effluent initiating Conditions 3
4 Introduction 5
This appendix supplements the basis information provided in Section 5 for initiating conditions AU1, 6
AA1, AS1, and AG1. Since the publication of revision 2 of this methodology, there have been 7
numerous questions raised as utilities worked to implement the IC and EALs. Additional feedback 8
was provided by the staff of the Nuclear Regulatory Commission. It became apparent that the brief 9
basis provided for each IC was not sufficient. When revision 3 of this document was in preparation, it 10 was decided to incorporate this appendix to provide the needed additional guidance and clarification.
11 The NUMARC/NESP-007 effluent IC/EALs represent a departure from previous EAL practice and 12 understanding these differences and their technical bases will facilitate site specific implementation of 13 the NUMARC/NESP-007 classification methodology.
14 This appendix will be structured into seven major sections. They are:
15 1.
Purpose of the effluent ICs/EALs and their relationship to other ICs/EALs 16 2.
Explanation of the ICs 17 3.
Explanation of the example EALs and their relationship to the ICs 18 4.
Interface between the ICs/EALs and the Offsite Dose Calculation Manual (ODCM) 19 5.
Monitor setpoints versus EAL thresholds.
20 6.
The impact of meteorology 21 7.
The impact of source term 22 A.1 Purpose of the Effluent ICs/EALs 23 ICs AU1, AA1, AS1, and AG1 provide classification thresholds for UNPLANNED and/or uncontrolled 24 releases of radioactivity to the environment. In as much as the purpose of emergency planning at 25 nuclear power plants is to minimize the consequences of radioactivity releases to the environment, 26 these ICs would appear to be controlling. However, classification of emergencies on the basis of 27 radioactivity releases is not optimum, particularly those classifications based on radiation monitor 28 indications. Such classifications can be deficient for several reasons, including:
29 In significant emergency events, a radioactivity release is seldom the initiating event, but 30 rather, is the consequence of some other condition. Relying on an indication of a release 31 may not be sufficiently anticipatory.
32 The relationship between an effluent monitor indication caused by a release and the 33 offsite conditions that result is a function of several parameters (e.g., meteorology, source 34 term) which can change in value by orders of magnitude between normal and emergency 35 conditions and from event to event. The appropriateness of these classifications is 36 dependent on how well the parameter values assumed in pre-establishing the 37 classification thresholds match those that are present at the time of the incident.
38 Section 3.3 of NUMARC/NESP-007 emphasizes the need for accurate assessment and classification 39 of events, recognizing that over-classification, as well as under-classification, is to be avoided.
40 Primary emphasis is intended to be placed on plant conditions in classifying emergency events.
41 Effluent ICs were included, however, to provide a basis for classifying events that cannot be readily 42 classified on the basis of plant condition alone. Plant condition ICs are included to address the 43 precursors to radioactivity release in order to ensure anticipatory action. The effluent ICs do not 44 standalone, nor do the plant condition ICs. The inclusion of both categories more fully addresses the A.1
t 1
potential event spectrum and compensates for potential deficiencies in either. This is a case in which 2
the whole is greater than the sum of the parts.
3 From the discussion that follows, it should become clear how the various aspects of the 4
NUMARC/NESP-007 effluent ICs/EALs work together to provide for reasonably accurate and timely 5
emergency classifications. While some aspects of the radiological effluent EALs may appear to be 6
potentially unconservative, one also needs to consider IC / EALs in other recognition categories that
.7 compensate for this condition. During site specific implementation of these ICs/EALs changes to 8 - some of these aspects might appear advantageous. While site specific changes are anticipated, 9
caution must be used to ensure that these changes do not impact the overall effectiveness of the ICs 10
/ EALs.
11 A.2. Initiating Conditions 12 There are four radiological effluent ICs provided in NUMARC/NESP-007. The IC and the fundamental 13. basis for the ultimate classification for the four classifications are:
14 General (AG1)
Offsite Dose Resulting from an Actual or Imminent Release of Gaseous 15 Radioactivity Exceeds 1000 mR TEDE or 5000 mR Thyroid CDE for the 16 Actual or Projected Duration of the Release Using Actual Meteorology.
17 Site Area (AS1)
Offsite Dose Resulting from an Actual or imminent Release of Gaseous 18 Radioactivity Exceeds 100 mR TEDE or 500 mR Thyroid CDE for the Actual 19 or Projected Duration of the Release.
20 Alert (AA1)
Any UNPLANNED Release of Gaseous or Liquid Radioactivity to the 21 Environment that Exceeds 200 Times Radiological Technical Specifications 22 for 15 Minutes or Longer.
23 NOUE (AU1)
Any UNPLANNED Release of Gaseous or Liquid Radioactivity to the 24 Environment that Exceeds Two Times Radiological Technical Specifications 25 for 60 Minutes or Longer.
i 26 The fundamental basis of AU1 and AA1 ICs differs from that for AS1 and AG1 ICs. It is important to l
27 understand the differences, 28-The Radiological Effluent Technical Specifications (RETS) (similar controls are included in
.29 the ODCMs of those facilities that implemented Generic Letter 89-01) are associated with 30 particular offsite doses and dose rate limits. For showing compliance with these limits, 31 facility Offsite Dose Calculation Manuals (ODCM) establish methodologies for establishing 32' effluent monitor alarm setpoints, based on defined source term and meteorology 33 assumptions.
34 AU1 and AA1 are NOT based on these particular values of offsite dose or dose rate but,
-35 rather, on the loss of plant control implied by a radiological release that exceeds a 36 specified multiple of the RETS release limits for a specified period of time.
l 37 The RETS multiples are specified only to distinguish AU1 and AA1 from non-emergency 38 conditions and from each other. While these multiples obviously correspond to an offsite l
39 dose, the classification emphasis is on a release that does not comply with a license 40 commitment for an extended period of time.
41 While some of the example EALs for AU1 and AA1 use indications of offsite dose rates as 42 symptoms that the RETS may be exceeded, the IC, and the classification, are NOT t
43 concerned with the particular value of offsite dose. While there may be quantitative 44.
inconsistencies involved with this protocol, the qualitative basis of the EAL, i.e., loss of 45 plant control, is not affected.
i A.2
I
- t l
l l
1 The basis of the AS1 and AG1 ICs IS a particular value of ofisite dose for the event i
2 duration. AG1 is set to the value of the EPA PAG. AS1 is a fraction (10%) of the EPA t
3 PAG. As such, these ICs are consistent with the fundamental definitions of a Site Area 4
and General Emergency.
5 A.3 Example Emergency Action Levels l
6 For each of the classifications, NUMARC/NESP-007 provides some example emergency action 7
levists and bases. Ideally, the example EALs would correspond numerically with the thresholds 8
expressed in the respective IC. Two cases are applicable to the effluent EALs:
9 1.
The EAL corresponds numerically to the threshold in the respective IC. For example, a 10 field survey result of 1000 mrem /hr for a projected release duration of one hour 11 corresponds directly to AG1.
12 2.
The EAL corresponds numerically to the threshold in the respective IC under certain 13 assumed conditions. For example, an effluent monitor reading that equates to 100 14 mrem for the projected duration of the release corresponds numerically to AS1 if the 15 actual meteorology, source term, and release duration matches that used in establishing 16 the monitor thresholds.
17 There are four typical example EALs:
18 Effluent Monitor Readinos: These EALs are pre-calculated values that correspond to the 19 condition identified in the IC for a given set of assumptions.
20 Field Survey Results: These example Eats are included to provide a means to address 21 classifications based on results from field surveys.
22 Perimeter Monitor indications: For sites having them, perimeter monitors can provide a 23 direct indication of the offsite consequences of a release.
24 Dose Assessment Results: These example EALs are included to provide a means to 25 address classifications based on dose assessments.
26 A.3.1 Effluent Monitor Readings 27 As noted above, these EALs are pre-calculated values that correspond to the condition identified in 28 the IC for a given set of assumptions. The degree of correlation is dependent on how well the 29 assumed parameters (e.g., meteorology, source term, etc.) represent the actual parameters at the 30 time of the emergency.
31 AS1 and AG1 32 Classifications should be made under these EALs if VAllD (e.g., channel check, comparison to 33 redundant / diverse indication, etc.) effluent radiation monitor readings exceed the pre-calculated 34 thresholds. In a change from previous versions of this methodology, confirmation by dose 35 assessments is no longer required as a prerequisite to the classification. Nonetheless, dose 36 assessments are important components of the overall accident assessment activities when 37 significant radioactivity releases have occurred or are projected. Dose assessment results, when 38 they become available, may serve to confirm the validity of the effluent radiation monitor EAL, may 39 indicate that an escalation to a higher classification is necessary, or may indicate that the 40 classification wasn't warranted. AS1 and AG1 both provide that, if dose assessment results are 41 available, the classification should be based on the basis of the dose assessment result rather than 42 the effluent radiation monitor EAL.
l 43 AU1 and AA1 44 ODCMs provide a methodology for determining default and batch-specific effluent monitor alarm l
45 setpoints pursuant to Standard Technical Specification (STS) 3.3.3.9. These setpoints are intended A.3
e t
1 to show that releases are within STS 3.11.2.1. The applicable limits are 500 mrem / year whole body 2
or 3000 mrem / year skin from noble gases. (Inhalation dose rate limits are not addressed here since 3
the specified surveillance involves collection and analysis of composite samples. This after-the-fact 4
asse.sment could not be an made in a timely manner conducive to accident classification.) These 5
setpoints are calculated using default source terms or batch-specific sample isotopic results and 6
annual average x/Q. Since the meteorology data is pre-defined, there is a direct correlation between 7
the monitor setpoints and the RETS limits. Although the actual x/Q may be different, NUREG-1022, 8
Event Reportino Guidelines 10 CFR 50.72 and 50.73. provided "... Annual average meteorological 9
data should be used for determining offsite airbome concentrations of radioactivity to maintain 10 consistency with the technical specifications (TS) for reporfability threshcids." The ODCM 11 methodology is based on long term continuous releases. However, its use here in a short term 12 release situation is appropriate. Remember that the AU1 and AA1 ICs are based on a loss of plant 13 control indicated by the failure to comply with a multiple of the RETS release limits for an extended 14 period and that the ODCM provides the methodology for showing compliance with the RETS.
15 To obtain the EAL thresholds, multiply the ODCM setpoint for each monitor by 2 (AU1) or 200 (AA1).
16 It would be preferable to reference "2 x ODCM Setpoint" or "200 x ODCM Setpoint" as the EAL 17 threshold. In this manner, the EAL would always change in step with changes in the ODCM setpoint 18 (e.g., for a batch or special release. In actual practice, there may be an "waming" and a "high" alarm 19 setpoint. The setpoint that is closest in value to the RETS limit should be used. Facility ODCMs may 20' lower the actual setpoint to provide an administrative " safety margin". Also, if there is more than one 21 unit or release stack on the site, the RETS limits may be apportioned. Two possible approaches to 22 obtain the EAL thresholds are:
23 The "2x" and "200x" multiples could be increased to address the reduced setpoints. For 24 example, if the stack monitor were set to 50% of the RETS limit, the EAL threshold could 25 be set to "4x"and "400x"the setpoint on that monitor.
26 The reduced setpoints could be ignored and the "2x" and "200x" multiples used as 27 specified. While numerically conservative, using a single set of multipliers would probably 28 be desirable from a human engineering standpoint.
29 in a change from previous versions of this methodology, confirmation by dose assessments is no i
30 longer required as a prerequisite to the classification. While assessments with real meteorology may 31 have provided a basis for escalating to AS1 (or AG1), the assessments could not confirm the AU1 or 32 AA1 classifications since compliance with the RETS is demonstrated using annual average 33 meteorology - not - actual meteorology.
34 Nonetheless, dose assessments are important components of the overall accident assessment 35 activities when significant radioactivity releases have occurred or are projected. Dose assessment 36 results, when they become available, may indicate that an escalation to a higher classification is 37 necessary. AS1 and AG1 both provide that, if dose assessment results are available, the 38 classification should be based on the basis of the dose assessment result rather than the effluent 39 radiation monitor EAL.
40 in typical practice, the radiological effluent monitor alarms would have been set, on the basis of 41 ODCM requirements, to indicate a release that could exceed the RETS limits. Alarm response 42 procedures calLfor_an assessment of the_. alarm to determine whether or not RETS have been 43 exceeded. Utilities typical 9 have methoos for racidly assedng an abnomud relea.e in ordec to l
44 determine whether or not the situation is reportable under 10 CFR 50.72. Since a radioactivity I
45 release of a magnitude comparable to the RETS limits will not create a need for offsite protective 46 measures, it would be reasonable to use these abnormal release assessment methods to initiate 47 dose assessment techniques using actual meteorology and projected source term and release 48 duration.
l A.4
- -. - - - - - - - - _ - -.. ~ ~
.t 1
A.3.2 Pirim;ttr Monitor, Field Survey Results, Dose Projection Results 2
AS1 and AG1 3
The perimeter monitor and field survey results are included to provide a means for classification 4
based on actual measurements. There is a 1:1 correlation (with consideration of release d t
5 between these EALs and the IC since all are dependent on actual meteorology.
\\
l 6
Dose projection result EALs are included to provide a basis for classification based on results from l
7 assessments triggered at lower emergency classifications, if the dose assessment results are 8
available at the time that the classification is made, the results should be used in conjunction with this 9
EAL for classifying the event rather than the effluent radiation monitor EAL.
10 Although the IC references TEDE and thyroid CDE as criteria, field survey results and perimeter 11 monitor indications will generally not be reported in these dose quantities, but rather in terms of a 12 dose rate. For this reason, the field survey EALs are based on a p y dose rate and a thyroid CDE 13 value, both assuming one hour of exposure (or inhalation). If individual site analyses indicate a longer j-14 or shorter duration for the period in which the substantial portion of the activity is released, the longer
'15 duration should be used for the field survey and/or perimeter monitor EALs.
16 AU1 and AA1 L
17 As discussed previously, the threshold in these ICs is based on exceeding a multiple of the RETS for L
18 an extended period. The applicable RETS limit is the instantaneous dose rate provided in Standard t
19 Technical Specification (STS) 3.11.2.1 While these three EALs are also expressed in dose rate, they 20 are dependent on actual meteorology. However, compliance with the RETS is demonstrated using 21 annual average meteorology. Due to this, the only time that there would be a 1:1 correlation between 22-the IC and these EALs is when the value of the actual meteorology matched the annual average -
23 an unlikely situation. For this reason, these EALs can only be indirect indicators that the RETS may 24 be exceeded. The three example EALs are consistent with the fundamental basis of AU1 and AA1, l
25 that of a uncontrolled radioactivity release that indicates a loss of plant control. A dose rate, at or 26 beyond the site boundary, greater than 0.1 mR/hr for 60 minutes or 10.0 mR/hr for 15 minutes is 27 consistent with this fundamental basis, regaidless of the lack of numerical correlation to the RETS..
28 The time periods chosen for the NOUE AU1 (60 minutes) and Alert AA1 (15 minutes are indicative of l
29 the relative risks based on the loss of ability to terminate a release l
30 The numeric values shown in AU1 and AA1 are based on a release rate not exceeding 500 mrem per 31 year, converted to a rate of 500 + 8766 = 0.057 mR/hr. If we take a multiple of 2, as specified in the i
32 NOUE threshold, this equates to a dose rate of about 0.11 mR/hr, which rounds to the 0.1 mR/hr 33 specified in AU1. Similarly for the AA1 EALs, we obtain 10 mR/hr.
34 In AU1 and AA1, reference is made to automatic real-time dose assessment capability. In AS1 and 35 AG1, the reference is to dose assessment. This distinction was made since it is unlikely that a dose 36 assessment using manual methods would be initiated without some prior indication, e.g., a effluent 37 monitor EAL.
38 A.4 Interface Between ODCM and ICs/EALs 39 For AU1 and AA1, a strong link was established with the facility's ODCM. It was the intent of the 40 NUMARC/NESP EAL Task Force to have the AU1 and AA1 EALs indexed to the ODCM alarm 41 setpoints. This was done for several reasons:
4.
42 To allow the EALs to use the monitor setpoints already in place in the facility ODCM, thus 43 eliminating the need for a second set of values as the EALs. The EAL could reference "2x j
44-ODCM Setpoint" or "200x ODCM Setpoint" for the monitors addressed in the ODCM.
45 Extensive calculations would only be necessary for monitors not addressed in the ODCM.
I f
A.5
I To ensure that the operators had an alarm to indicate the abnormal condition. If the 1
2 monitor EAL threshold was less than the default ODCM setpoint, the operators could be in 3
the position of having exceeded an EAL and not knowing it.
To take advantage of the alarm setpoint calculational methodology already documented in 4-5 the facility ODCM.
To simplify the IC / EAL by eliminating the need to address planned and UNPLANNED 6
a 7
releases, continuous or batch releases, monitored or unmonitored releases. Any release 8
that complies with the radiological effluent technical specifications (RETS) (or ODCM 9
controls for utilities that have implemented GL 89-01) would not exceed a monitor EAL 10 threshold.
11
- To eliminate the possibility of a planned release (e.g., containment / drywell purge) 12 resulting in effluent radiation monitor readings that exceed an classification threshold that 13 was based on a different calculation method. ODCMs typically require specific alarm 14 setpoints for such releases. If the release can be authorized under the provisions of the 15 ODCM/RETS, an emergency classification is not warranted. If the monitor EAL threshold 16 is indexed to the ODCM setpoint (e.g., ".. 2 x ODCM setpoint...") the monitor EAL will 17 always change in step with the ODCM setpoint.
Although the ODCM is intended to address long term routine releases, its use here for 18 19 short term releases is appropriate. The IC is specified in terms of a release that exceeds 20 RETS for an extended period of time. Compliance to the RETS is shown using the ODCM 21 methodology.
22 A.5 Setpoints versus Monitor EALs 23 Effluent monitors typically have provision for two separate alarm setpoints associated with the level 24 of measured radioactivity. (There may be other alarms for parameters such as low sample flow.)
25 These setpoints are typically established by the facility ODCM. As such, at most sites the values of l
26 the monitor EAL thresholds will not be implemented as actual alarm setpoints, but would be tabulated 1
27 in the classification procedure. If the monitor EAL thresholds are calculated as suggested herein they i
28 will be higher than the ODCM alarm setpoints by at least a factor of two (i.e., AU1). This alarm alerts 29 the operator to compare the monitor indication to the EAL thresholds. The NUMARC/NESP-007 l
30 effluent EALs do NOT require alarm setpoints based on the monitor EALs., However, if spare alarm 31 channels are available (e.g., high range channels), the monitor EAL threshold could be used as the 32 alarm setpoint.
33 A.6 The Impact of Meteorology 34 The existence of uncertainty between actual event meteorology and the meteorology assumed in 35 establishing the EALs was identified above. It is important to note that uncertainty is present 36 regardless of the meteorology data set assumed. The magnitude of the potential difference and, 37-hence, the degree of conservatism will depend on the data set selected. Data sets that are intended 38 to ensure low probability of under-conservative assessments have a high probability of being over-39 conservative. For nuclear power plants, there are different sets of meteorological data used for L
40 different purposes. The two primary sets are:
41 For ace dd m;cn ;-.tposes, sector xiQ salues are set at inat value ina! is exceedad e
[
42 oniy 0.5% of the nours wind blows into the sector. The highest of the 16 sector values is 43 the maximum sector x/Q value. The site x/Q value is set at that value that is exceeded l
44 only 5% of the hours for all sectors. The higher of the sector or site x/Q values is used in l
45 accident analyses.
A.6
- i l
l 1-For routins raisess situations, annual average x/Q values are calculated for specified l
2 receptor locations and at standard distancec in each of the 16 radial sectors. In setting 3
ODCM alarm set poiYs, the annual average x/O value for the most restrictive receptor at 4
or beyond the site bc andary is used. The sector annual average x/Q value is normalized 5
for the percentage of time that the wind blows into that sector in an actual event, the wind l
6
. direction may be into the affected sector for the entire release duration. Many sites 7
experience typical sector x/Qs that are 10-20 times higher than the calculated annual 8
average for the sector.
i 1
9 in developing the effluent EALs, the NUMARC EAL Task Force elected to use annual average l
10- meteorology for establishing effluent monitor EAL thresholds. This decision was based on the l
11 following considerations.
12 Use of the accident x/Qs, may be too conservative. For some sites, the difference 13 between the accident x/O and the annual average x/Q can be a factor of 100-1000. With i
=14 this difference in magnitude, the calculated monitor EALs for AS1 or AG1 might actually l
15 be less than the ODCM alarm setpoints, resulting in unwarranted classifications for 16 releases that might be in compliance with ODCM limits.
17 The ODCM and the RETS are based in part on annual average x/Q (non-normalized).
l 18 ODCMs already provide alarm setpoints based on annual average x/Q that could be used 19 for AU1 and AA1.
~
20 Use of a x/Q more restrictive than the x/Q used to establish ODCM alarm setpoints could 21 create a situation in which the EAL value would be less than the ODCM setpoint. In this l
22 case, the operators would have no alarm indication to alert them of the emergency 23 condition.
24 Use of one x/Q value for AU1 and AA1 and another for AS1 and AG1 might result in 25 monitor EALs that would not progress from low to high classifications. Instead, the AS1 26 and AA1 EALs might overlap.
27 28 Plant specific consideration must be made to determine if annual average meteorology is adequately 29 conservative for site specific use. If not one of the two more conservative techniques described 30 above should be selected. It is incumbent upon the licensee to ensure that the selection is properly 31 implemented to provide consistent classification escalation.
32 33 The impact of the differences between the assumed annual average meteorology and the actual 34 meteorology depends on the particular EAL.
35 For the AU1 and AA1 effluent monitor EALs, there is no impact since the IC and the EALs 36 are based on annual average meteorology by definition.
37 For the field survey, perimeter monitor, and dose assessment results EALs in AS1 and 38 AG1, there is no impact since the IC and these EALs are based on actual meteorology.
39 For the AS1 and AG1 effluent monitor EALs, there may be differences since the IC is i
j 40 based on actual meteorology and the monitor EALs are calculated on the basis of annua; 41.
average meteorology or, on a site specific basis, one of the more conservative derivatives 42.
of annual average meteorology. This is considered as acceptable in that dose 43 assessments using actual meteorology will be initiated for significant radioactivity 44 releases. Needed escalations can be based on the results of these assessments. As 45 discussed previously, this delay was deemed to be acceptable since in significant release i
A.7 4
I 1
situations, the plant condition EALs should provide the anticipatory classifications 2
necessary for the implementation of offsite protective measures.
For the field survey, perimeter monitor, and dose assessment results EALs in AU1 and 3
4 AA1, there is an impact. These three EALs are dependent on actual meteorology.
5 However, the threshold values for all of the AU1 and AA1 EALs are based on the 6
assumption of annual average meteorology. If the actual and annual average meteorology 7
were equal, the IC and all of the EALs would correlate. Since it is likely that the actual 8
meteorology will exceed the annual average meteorology, there will be numerical 9
inconsistencies between these EALs and the IC. The three example EALs are consistent 10 with the fundamental basis of AU1 and AA1, that of a uncontrolled radioactivity release 11 that indicates a loss of plant control. A dose rate, at or beyond the site boundary, greater 12 than 0.1 mR/hr for 60 minutes or 10.0 mR/hr for 15 minutes is consistent with this 13 fundamental basis, regardless of the lack of numerical correlation to the RETS.
14 A.7 The Impact of Source Term 15 The ODCM methodology should be used for establishing the monitor EAL thresholds for these ICs.
16 The ODCM provides a default source term based on expected releases. In many cases, the ODCM 17 source term is derived from expected and/or design releases tabulated in the FSAR.
18 For AS1 and AG1, the bases suggests the use of the same source terms used for establishing 19 monitor EAL thrennc!ds for AU1 and AA1. This guidance is provided to avoid potential overlaps 20 between effluent n'onitor EALs for AA1 and AS1. Other source terms may be appropriate. In any 21 case, efforts should be made to obtain and use best estimate (For Example: NUREG 1465), as 22 opposed to conservative, source terms for all four ICs.
3 23 Even if the same source term is used for all four ICs, the analyst must consider the impact of overly 24 conservative iodine to rioble gas ratios. The AU1 and AA1 IC thresholds are based on extemal noble 25 gas exposure. The AS1 and AG1 ICs are based on either TEDE or thyroid CDE. TEDE includes a 26 contribution " om inhalation exposure (i.e., CEDE) while the thyroid CDE is due solely to inhalation 27 exposure. Trio inhalation exposure is sensitive to the lodine concentration in the source term. Since 28 AU1 and AA1 are based on noble gases, and AS1 and AG1 are dependent on noble gases gnld 29 lodine, an over conservative iodine to noble gas ratio could result in AS1 and AG1 monitor EAL 30 thresholds that either overlap or are too close to the AA1 monitor EAL thresholds.
31 As with meteorology, assessment of source terms has uncertainty. This uncertainty is compensated 32 for by the anticipatory classifications provided by ICs in other recognition categories.
l l
l A.8
NEI 97-03 Revision 2 to 3 Change Summary (Sorted by NEI 97-03 Page Number) s Notes:
The following cross reference should be used to determine the SOURCE of the change described:
I. NESP-007 Question and Answers (June 1993)
- 2. Response To NRC Staff Comments On NESP-007 EAL Lessons Learned (7/96 updated 1/24/97)
- 3. Letter from Zaleman to Nelson " Review of NEI EAL Guidance Document NEI 97-03, Draft Final Rev.3, August 1997" (3/13/98)
- 4. NEI Emergency Action Level (EAL) Issue Task Force recommendation (1994-1998) t
NEI 97-03 Rev 2 to 3 Change Summary See Cat Pg Source Cmat Pg Change Description Reason for Change No No No-1.
0 4
N/A Title page revised to NEI 97-03 dated October To standardize the document 1998 and NUMARC changed to NEI numbering and reflect correct titles 2.
0 4
N/A Acknowledgements, Notice and Foreword Updated to indicate correct titles and changed to add clarity to the purpose of the revision 3.
0 4
N/A EAL Revision Task Force members revised To correctly list contributors to the revision process 4.
i 4
N/A Table of Contents revised To update content and page numbers 5.
v 4
N/A Executive Summary revised Updated to clarify history of development and reflect changes made within the body of the document based on incorporation of the Q&A document and NRC Comments " Lessons Learned" 6.
viii 4
N/A Acronym list changed Revised to update needed acronyms based on current industry usage and usage within this document 7.
I 1.1 1.1 4
N/A Section 1.0 was significantly changed by This historical information was n 1.2 rewording section 1.1 and deleting theNESP-007 longer needed for discussion in this 1.3 discussion associated with the Scenario document and is forevercaptured Applications, Task Force Charter, and Structure historically in NESP-007 Rev. 2.
of the Study.
ChgSumR1. doc 2
10/18/98
NEI 97-03 Rev 2 to 3 Change Summary Sec Cat Pg Source Cmnt Pg Change Description Reason for Change No No No 8.
2 2.0 2.1 4
N/A Section 2.0 was completely rewritten to discuss Applicable portions of the old section 2.1 the changes incorporated into revision 3. The 2.0 that described current EAL usage 2j section had previously described the current was incorporated into section 1.0.
2.4 EAL usage.
The remaining information still exists 2.5 2.2 historically in revision 2. High level 2.6 descriptions ofchanges made in each 2.7 23 section of revision 3 are incorporated in sections 2.1 through 2.7.
t 9.
3 3.0 3.1 4
N/A Section 3.0 revised and expanded.
To add clarity to the discussion on development of the basis for the generic approach to the new EAL scheme.
10.
3 3.1 3.1 4
N/A Section 3.1 revised.
To provide a clearer introduction into 3.2 the regulatory context discussion and to eliminate repetitive descriptions and examples of the content of certain regulatory guidance.
I 1.
3 3.2 33 4
N/A Section 3.2 revised.
Changes made to decrease repetition 3.4 of historical perspective on EALs.
12.
3 33 3.4 4
N/A No changes made to Section 3.3 13.
3 3.4 3.4 4
N/A Deleted reference to figure I and made minor Figure I no longer needed because it 3.5 revisions within the text.
is replaced by section 3.15. Text 3.6 revisions made for clariiy.
14.
3 3.5 3.6 4
N/A No changes made to Section 3.5 15.
3 3.6 3.6 4
N/A Deleted last two paragraphs and made minor Clarification and simplification.
editorial changes to the remainder of the section 16, 3
3.7 3.7 4
N/A No changes made to Section 3.7 3.8 i
ChgSumRI. doc 3
10/18/98
t
=
NEI 97-03 Rev 2 to 3 Change Summary See Cat Pg Source Cmnt Pg Change Description Reason for Change No No No 17.
3 3.8 3.8 4
N/A Deleted reference to figure 2 and made minor Figure 2 no longer needed. Text 39 revisions within the text.
revisions made for clarity -
18.
3 3.9 3.9 4
N/A Minor text changes made in all sections.
To improve clarity.
3.10 3.10 3.11 3.11 3.12 19.
3 3.12 3.12 1
7 2
Section 3.12 is a new section titled Classifying Added to address Q&A item.
3.13 1
9 3
Transient Events.
20.
3 3.13 3.13 4
N/A Section 3.13 is a new section titled Interface Added to include the guidance Between Classification and Activation of provided in Reg Guide 1.101.
Emergency Facilities 21.
3 3.14 3.13 4
N/A Section 3.14 is a new section titled Shutdown Added to introduce Shutdown 3.14 IC/EALs IC/EALs and to reference NEI 99-01.
22.
3 3.15 3.14 1
2 1
Section 3.15 is a new section titled Operating Added to replace deleted text and 3.15 Mode Applicability figures that described mode applicability.
23.
4 4.0 4.1 4
N/A No changes made to section 4.x.
ITF decision 4.1 4.2 4.2 4.3 24.
5 5.0 5.1 4
N/A Section 5.0 was expanded and subdivided into ITF decision i
four subsections titled as follows: 5.1 Generic Arrangement,5.2 Generic Bases,5.3 Site Specific Implementation, and 5.4 Definitions 25.
5 5.1 5.1 1
14 4
Minor text changes and additions made.
Clarification.
5.2 ChgSumRI. doc 4-10/18/98
NEI 97-03 Rev 2 to 3 Change Summary Sec Cat Pg Source Cmnt Pg
(.ange Description Reason for Change No No No 26.
5 53 53 1
3 1
Section 5.3 added to address multiple NRC Clarification.
1 4
2 comments and Q&A responses.
1 5
2 1
6 2
2 I
I 2
20 10 2
40 16 27.
5 5.4 53 4
N/A Section 5.4 added definitions. If a word is used Section added to provide standard 5.4 in the document that is found in the definitions definitions section for use through out 5.5 section then that word has been placed in CAPS.
the document and to provide the user a quick way to identify that the word had a defined meaning when used in the document.
28.
5 A
All I
l-13 5-12 AUI, AU2, AAl, AA2, AA3, ASI, AGI -
Revisions made based on NRC 2
2-19 2-10 Section A has been extensively revised and Comments and Q&A. See Appendix 2
23 11 Appendix A has been added to help the user A discussion in this Change understand usage of this section.
Summary.
29.
5 F
1 1
1 16 No change made to NEI 97-03 based on this See Q&A answer.
I 2
16 Q&A.
3 16 5
17 7
I8 9
19 II 20 30.
5 F
1 1
8 18 Discussion that exists within section 3.9 No additional discussion needed to addresses this question address Q&A ChgSumRI. doc 5
10/18/98 i
t
a.-
.t NEI 97-03 Rev 2 to 3 Change Summary Sec Cat Pg Source Cmnt Pg Change Description Reason for Change No No No 31.
5 F
1 2
21 11 Revised the Site Area Emergency criteria to any Clarification and logic simplification.
5 F
9 combination of two barriers. Revised PWR S/G l
Tube rupture to include also a Faulted S/G to make the logic compatible.
32.
5 F
2 1
4 17 Revised to indicate an unisolable MSL break.
Clarification i
33.
5 F
2 1
10 19 Revised to entry into containment flooding Clarification and simplific.dbn.
i gp procedure as more indicative of the desired classification.
L 34.
5 F
2 2
26 12 Added unisolable criteria and clarified basis.
Clarification 5
F 5
35.
5 F
2 3
2.3 9
Inadvertently placed in the loss column and has Fixed error.
been properly replaced in the potential loss column.
36.
5 F
4 2
22 11 Added compensation statement into the basis.
Clarification.
37.
5 F
4 2
25 12 Comment was no implemented Not valid and not implemented.
38.
5 F
5 1
6 17 Revised basis to clarify intent and eliminate Clarification confusing verbiage about leakage.
39.
5 F
5 2
27 12 Added unisolable criteria and clarified basis.
Clarification 40.
5 F
5 2
28 13 Clarified that the low setpoint is used and Clarification confirmed to be indicating a leak.
f 41.
5 F
5 2
29 13 Added unisolable criteria and clarified basis.
Clarification f
42.
5 F
5 2
30 13 Revised to indicate the actuation setpoint.
Clarification
[
43.
5 F
5 3
2.4 9
Clarified statement to indicate the low setpoint is Clarification.
used before a containment isolation and the high j
setpoint after containment isolation to get a reasonable escalation.
44.
5 F
6 2
31 13 Revised basis to indicate operator recognition of Clarification conditions in an accident situation.
t
?
ChgSumRI. doc 6
10/18/98 l
l
NEI 97-03 Rev 2 to 3 Change Summary See Cat Pg Source Cmat Pg Change Description Reason for Change No No No 45.
5 F
6 2
32 14 Revised basis to indicate operator recognition of Clarification conditions in an accident situation.
46.
5 F
7 2
33 14 Revised to eliminate max core uncovery time Simplified clarification i
34 and replaced with entry into the Containment i
Flooding procedure.4 i
47.
5 F
9 1
4 15 No change made to NEl 97-03 based on this See Q&A answer.
j I
5 15 Q&A.
48.
5 F
10 1
3 14 Revised all S/G tube rupture concerns to address Clarification and simplification.
l 6E(L this and the change in logic associated with the sciu4 3 4 L.
t change in the SAE definition.
49.
5 F
10 2
39 16 Revised PWR S/G Tube rupture to include also a Clarification and logic simplification.
Faulted S/G to make the logic compatible.
i 50.
5 F
12 2
35 14 Added text to PWR Fuel Clad barrier CETC.
To incorporate NRC comment for basis clarity.
51.
5 F
13 2
24 12 No change made Considered part ofOther(Site Specific) Indications and could be I
added for implementation based on this reference.
52.
5 F
15 1
1 13 Revised basis to include why the second Clarification i
2 charging pump is an indicator.
53.
5 F
15 2
36 15 Revised basis to include why the second Clarification i
charging pump is an indicator.
t 54.
5 F
15 2
37 15 Revised and clarified basis.
Clarification 55.
5 F
15 2
38 16 Revised PWR Containment barrier CETC basis To incorporate NRC comment. For f
clarity 56.
5 F
16 3
3.5 14 Clarified wording in basis.
Clarification i
ChgSumRI. doc 7
10/18/98
a NEI 97-03 Rev 2 to 3 Change Summary Sec Cat Pg Source Cmnt Pg Change Description Reason for Change No No No 57.
5 11 1
4 NA HG1 Changed IC from Loss of" Ability to Consistent with NUREG-0654.
t 5
11 19 Reach and Maintain Cold Shutdown" to Loss of Original IC is too specific and
" Physical Control of the Facility" suggests the core is loaded 58.
5 11 2
1 2
21 H Matrix - Initiating condition matrix for HUI Answer to Q&A. Wording in matrix 2
43 17 changed " occurring within" to "affecting" and IC made consistent with individual ICs I
wording changed per NRC comment 59.
5 11 2
1 4
21 HUI - EAL #3 for Control Room assessment Basis for change in answer to Q&A 1.
5 21, was deleted as an EAL and restated in paragraph
- 4 and #5.
2 41 16 1
2 53 19 60.
5 11 2
2 42 17 HUI - EAL was not revised to include affects on ITF decision.
a train ofcquipment due to ice buildup.
61.
5 11 2
1 6
22 HUI -EAL #3 basis revised To more clearly indicate what vehicles are being considered.
62.
5 II 2
1 3
21 HUl -EAL #4 was not revised to include ITF decision answer to this question.
63.
5 11 2
2 54 19 HUI - Added " Uncontrolled flooding in (site Added per comment 5
11 3
specific) areas of the plant that has the potential recommendation. Originally stated it to affect safety related equipment needed for the would be added as part of S/D EALs current operating mode" and applicable basis for however ITF concluded it has the example.
applicability to all operating modes.
Basis provides additional details as recommended.
64.
5 11 2
4 N/A HUl - Protected area in the IC is changed to ITF recommendation.
upper case and now defined in the Definitions section.
ChgSumRI. doc 8
10/18/98
y NEI 97-03 Rev 2 to 3 Change Summary See Cat Pg Source Cmat Pg Change Description Reason for Change No No No 65.
5 H
2 2
51 18 HUI - High winds were added to EAL #2 This is a site specific value indicative t
2 52 19 of the FSAR design basis threshold.
This changeis also included in the basis. The basis adequately addresses escalation to Alert via HAl.
66.
5 11 4
1 7
22 HU2 - Included statement that identifies when Basis for change is Q&A #7
- [
the 15 minute time frame begins j
67.
5 H
4 2
44 17 HU2-No changes made ITF believes that Q&A is already -
1 7
22 addressed adequately in the basis.
68.
5 11 4
2 45 17 HU2 - Provided further guidance on application ITF response to Comment 45 of the term ' areas contiguous to' 69.
5 11 4
2 46 17 HU2 - Added applicable information from ITF response to Comment 46.
l discussion in response to Q&A #13 I
70.
5 11 5
1 8
22 HU3 - Added information that flammable or Basis for change is Q&A #8 toxic gases that are above flammable or life-threatening concentrations are applicable to this f
EAL 71.
5 11 5
2 47 17 HU3 - Added infonnation that flammable or ITF response to Comment 47 I
toxic gases that are above flammable or life-i threatening concentrations are applicable to this EAL -
1 i
72.
5 H
5 3
3.8 14 HU3 - Added discussion on escalation of this IC Provided escalation criteria to Alert
+
to HA3 73.
5 H
6 2
48 18 HU4 - Added Bomb device to HA4 and removed Added per comment recommendation
{
5 H
13 reference from HU4 I
I t
i ChgSumRI.dc 9
10/18/98
. =
c NEI 97-03 Rev 2 to 3 Change Summary See Cat Pg Source - Cmnt Pg Change Description Reason for Change No No No 74 5
11 6
2 48 18 HU4 - EAL #1 deleted; Added Bomb device to Add per comment recommendation, 3
3.9 14 HA4 and removed reference from HU4 Bomb device in the Protected Area would indicate penetration by a Hostile Force 75.
5 11 6
4 NA HU4 - Expanded the Basis to include specific Definitions section added this defined events which should be considered for revision and events which generally 2
59 20 classification appear in Contingency Plans are included. Although Source #2 was asked specifically for HA4, it was considered more appropriate for HU4.
76.
5 II 6
4 NA HU4 - Protected Area definition removed from Definition Section added to 97-03 the Basis and when applicable, the defined term will be in Upper Case 77.
5 11 7
2 49 18 HU5 - Delete examples of actual event E;;amples were closely related to j
2 50 18 EAL's provided under other IC's 1
10 23 78.
5 11 7
2 60 21 HUS - Revise example EAL's to be consistent Judgement EAL's should use same with definition ofcorresponding Emergency language as ECL definition Classification Level (ECL) 79.
5 11 8
2 53 19 HAl -Wording added to EAL basis To provide more guidance on intent of site specific indications per the NRC comment.
ChgSumkt. doc 10 10/18/98
a o
NEI 97-03 Rev 2 to 3 Change Summary Sec Cat Pg Source Cmnt Pg Change Description Reason for Change No No No 80.
5 II 8
2 54 19 HAl - Added " Uncontrolled flooding in (site Added per comment specific) areas of the plant that results in recommendation. Basis provides decraded safety system performance as indicated additional details as recommended to in the control room or that creates industrial show escalation to the Alert, and safety hazards that precludes access necessary to provides details conceming flooding operate or monitor safety equipment.
events from intemal or extemal sources.
81.
5 11 8
1 11 23 HAl -No changes made to EALs based on these ITF believes that adequate 1
12 23 Q&A items information with in'the IC and 2
56 20 associated EALs exist for proper classification.
82.
5 11 to 1
13 24 HA2 - Added additional guidance to basis to Basis for change is Q&A #13 2
58 20 further clarify classification of this IC as an Alert 83.
5 II 10 2
57 20 HA2 - Added definition of Visible Damage to ITF response to Comment 57 Glossary of terms (page 5.22) 84.
5 II 10 3
3.6 14 HA2 - No additional changes made to NEI 97-03 No additional changes required - ITF based on the comment from the source document believes current EAL is appropriate.
- ITF believes curren't EAL is appropriate.
85.
5 11 10 3
3.10.1 15 HA2 - No change made based on this NRC No change required because the basis comment for this IC is related to specific areas of the site containing functions and systems required for safe shutdown of the plant.
86.
5 II 10 3
3.10.2 15 HA2 - No change made based on this NRC ITF reviewed this comment and comment believes that Explosions should still be referenced in this IC.
ChgSumRI. doc 11 10/18/98
n NEI 97-03 Rev 2 to 3 Change Summary Sec Cat Pg Source Cmnt Pg Change Description Reason for Change No No No 87.
5 11 12 4
N/A HA3 - Deleted text discussion relative to multi ITF believed that these changes add unit escalation and added text related to clarity to the EALs.
flammable gasses and potential for these gasses to affect safe operation.
88.
5 II 12 3
3.11 15 HA3 -Included further discussion relative to ITF response to NRC Comment 3.1I what is an appropriate concentration to classify based on this IC.
89.
5 11 13 1
9 22 HA4 - EAL #1 changed; Added Bomb device to Add per comment recommendation, 2
48 18 HA4 and removed reference from HU4 Bomb device in the Protected Area 3
3.9 14 would indicate penetration by a Hostile Force 90.
5 11 13 4
NA HA4 - EAL #1 & #2 became #2 & #3 Protected EAL #1 added for Bomb. Definitions Area, Hostile Force, and Vital Area were section added this revision changed to Upper Case 91.
5 11 13 4
NA HA4 - Deleted " Multi-unit stations" statement Included in section 5.3 " Site Specific following the basis Implementation" 92.
5 11 13 4
NA HA4 - Basis expanded for the new EAL #1 Bomb was moved from HU4 to HA4 93.
5 11 14 4
N/A HA5 - Deleted " Multi-unit stations" statement included in section 5.3 " Site Specific following the basis Implementation" 94.
5 11 14 1
14 24 HA5 -Changed Emergency Operations Center To include suggestion from Q&A.
to Emergency Response Facility 95.
5 11 15 2
60 21 HA6-Revised EAL wording To make the EAL contain wording which emulates the definition of an Alert.
ChgSumR1. doc 12 10/18/98 s
i
o i
NEI 97-03 Rev 2 to 3 Change Summary Sec Cat Pg Source Cmnt Pg Change Description Reason for Change No No No 96.
5 11 16 4
NA HS1 - Deleted " Multi-unit stations" statement Included in section 5.3 " Site Specific following the basis; Implementation";
Changed defined terms to upper case; Definition Section added to 97-03 and when applicable, the defined term will be in Upper Case 97.
5 11 16 4
NA IISI - Added " Confirmed" to the IC and Originally in the NOUE, now in all
" reported by the (site-specific) security shift four classification levels for supervision" to EAL #2 consistency 98.
5 II 17 1
15 25 HS2-Revised per the NRC comment.
To include information as described 2
61 21 in the Q&A.
99.
5 11 18 2
60 21 HS3 - Revised EAL wording and removed text To make the EAL contain wording on multi unit stations.
which emulates the definition of an SAE and to remove discussion on multi unit stations which is now discussed in the front matter.
100.
5 II 19 4
NA HG1 - Two original EALs combined into one The original EALs only addressed the and restmetured to reficct a function loss loss of specific locations and made no provisions for a core off-load scenario or maintaining the function by another means 101.
5 11 19 4
NA HG1 - Basis expanded to explain the new EAL EAL was changed from a location based to a function based. The basis now includes specific function examples for both types of Reactors 102.
5 II 19 4
NA HG1 - Deleted " Multi-unit stations" statement Included in section 5.3 " Site Specific following the basis Implementation" ChgSumRI. doc 13 10/18/98
o O
NEI 97-03 Rev 2 to 3 Change Summary Sec Cat Pg Source Cmnt Pg Change Description Reason for Change No No No 103.
5 11 20 2
60 21 HG2 - Revise example EAL's to be consistent Judgement EAL's should use same with definition ofcorresponding Emergency language as ECL definition Classification Level (ECL) 104.
5 S
I 4
N/A S Matrix - SS4 IC " Complete Loss of Function Consistency with SS4's event Needed to Achieve or Maintain Hot Shutdown" description (see page 5-S-22).
changed to " Complete Loss of Heat Removal Capability" 105.
5 S
3 1
1 26 SUI -Removed wording on multi unit affects Multi unit discussion covered in front and added text to allow taking credit for ability matter and therefore no longer to cross tie AC power from a companion unit needed. Cross tie added based Q&A..
106.
S 4
4 N/A SU2-minor test changes Clarity and consistency.
107.
5 S
5 2
62 21 SU3 - The IC was revised to equate indication The IC equates indication and and annunciators.
annunciators. However, the example EALs in NESP-007 only appear to address annunciatorloss.
108.
5 S
5 2
63 21 SU3 - The IC was revised to equate indication The IC equates indication and and annunciators.
annunciators. However, the example EALs in NESP-007 only appear to address annunciator loss.
109.
5 S
5 4
N/A SU3 -Initiating condition matrix for SU3 Formatting of defined terms as changed " unplanned loss" to " UNPLANNED specified in sections 2.3 and 5.
loss".
ChgSumRI. doc 14 10/18/98
=. _. _. _ - _. _ _ _.. _ _ - - - -
o-NEI 97-03 Rev 2 to 3 Change Summary -
t Sec Cat Pg Source Cmnt PR Change Description Reason for Change No No No 110.
5 S
5 4
N/A SU3 - Operating mode applicability was changed The modes identified in the EALs i
to include Startup for all designs were based on the standard technical i
specifications for BWRs and Westinghouse PWRs. To aid in
' I interpreting these modes for PWRs i
from other NSSSs and for plant with non-standard technical specifications,
'l the modes are now described in i
subsection 3.15.
111.
5 S
5 4
N/A SU3 - The EAL was modified to reflect an UNPLANNED is defined m r
unplanned loss as a classifiable event as opposed subsection 5.4, definition.
f to a " loss" being classifiable.
I12.
5 S
5 4
N/A SU3 - The basis for the EAL was modified to Verbiage related to the basis for j
reflect the EAL for SU-3.
UNPLANNED is contained within L
subsection 5.4, Definitions.
113.
5 S
5 4
N/A SU3 " Notification of Unusual Event" changed To use acronyms where appropriate.
[
to "NOUE" t
114.
5 S
5 4
N/A SU3 - Removed BOLD TEXT fonnatting from Added emphasis no longer f
sixth sentence in basis considered essential by ITF.
j 115.
5 S
7 2
64 22 SU4 - Revise basis to explain T/S mode Eliminate confusion regarding
}
3 3.13 13 applicability applicability j
116.
5 S
8 4
N/A SUS-Minor test changes.
To clarify escalation of this IC and to
[
1 provide consistency.
117.
5 S
8 1
12 4
SUS - No change made to NEI 97-03 based on See comment answer.
]
I 2
26 thisQ&A.
i l18.
5 S
9 2
65 22 SU6 - Revise basis to clarify applicability of Task Force suggestion i
1 4
27 EAL t
i t
[
ChgSumRI. doc 15 10/18/98 1
7 y
NEI 97-03 Rev 2 to 3 Change Summary:
Sec Cat Pg Source Cmat Pg Change Description Reason for Change No No No=
t i19. 5-s to
.4 -
N/A SU7 - Capitalized " UNPLANNED" Formatting ofdefined terms as-specified in sections 2.3 and 5.4 120.
5 s
10 4
N/A SU7 - Replaced single example EAL (with two To streamline presentation and be 1
discrete events listed as sub-clements, l.a and consistent with otherICs having 1.b) with two' sample EALs (I and 2, matching multiple example EALs.
the two events previously listed as sub-elements) 121.
5 s
10 4
N/A SU7 - Reference to escalation to the Alert level The applicable operating modes for by IC SA3 " Inability to Maintain Plant in Cold SU5 and SA3 are different. Section Shutdown" deleted.
3.15 says declaration is based on the mode that existed at the time the event occurred.
122.
5 s
to 4
N/A SU7 - Last sentence of basis, which described To eliminate redundancy; operating why operating modes were specified, deleted.
mode applicability listed separately.
l 123.
5 s
10-1 5
27 SU7-Basis was not revised to incorporate Q&A ITF agreement with Q&A answer.
.j guidance 124.
5 s
iI 4
N/A SU8 - Added new IC for Inadvertent Criticality The addition of this IC is based on the Shutdown EAL development process.
125.
5 s
12 1
9 29 SAI - Added wording to basis conceming loads To add Q&A intent to basis for i
which should be considered operable.
clarity.
f I
r i
i ChgSumRI. doc 16-10/18/98
a
+
NEI 97-03 Rev 2 to 3 Change Summary Sec Cat Pg Source Cmat Pg Change Description Reason for Change No No No 126.
5 S
13 2
75 26 SA2 - With regard to the ICs associated with Timing ofa reactor scram can be a protection system failure events (SA2, SS2, significant parameter in the event of SG2) several comments were received that took reactivity excursion events (e.g.,
exception to declaring an Alert in the event an PWR main steam line break), and automatic scram did not occur but the manual that it was not possible to generically trip from the control room was successful. The identify those transients for which task force considered a proposal that would link scram timing was significant and the failure to a concurrent plant transient for those events for which it was not.
which a scram was significant in preventing a reactivity excursion. After extensive review, the task force concluded that it was not possible to revise the IC at this time.
127.
5 S
13 2
66 22 SA2-Deleted site specific indications.
Deletion based on NRC comment.
2 67 22 ITF determined that no additional 2
68 23 changes were necessary to address 1
7 28 other NRC comments or Q&A.
128.
5 S
14 1
6 27 SA3 - Added new basis paragraph 5 To explain why both elements of the 2
69 23 incoq) orating information from Q&A explaining example EAL are needed why both elements of the example EAL are (incorporates answer to Q&A needed. [ NOTE: It was assumed that the questions 6.a and 6.b (also NRC reference to " June 1993 Q&A #4 under comment 69).
" SYSTEM MALFUNCTIONS"in this comment should have been to Q&A #6.]
129.
5 S
14 2
70 23 SA3 - No change made to NEI 97-03 based on Deferred for resolution under this NRC comment.
provisions of S/D EAL effort (NEl 99-01).
ChgSumRI. doc 17 10/18/98
s NEI 97-03 Rev 2 to 3 Change Summary See Cat Pg Source Cmnt Pg Change Description Reason for Change No No No 130.
5 s
16 2
62 21 SA4 - Initiating condition matrix for SA4 Formatting of defined terms as changed " unplanned loss" to " UNPLANNED specified in sections 2.3 and 5.4.
loss".
Changed significant transient to SIGNIFICANT TRANSIENT 131.
5 S
16 2
62 21 SA4 - The IC was revised to equate indication The IC equates indication and 2
63 21 and annunciators.
annunciators. However, the example EALs in NESP-007 only appear to address annunciator loss.
132.
5 s
16 4
N/A SA4 - Operating mode applicability was changed The modes identified in the EALs to include Startup for all designs were based on the standard technical specifications for BWRs and Westinghouse PWRs. To aid in interpreting these modes for PWRs from other NSSSs and for plant with non-standard technical specifications, the modes are now described in subsection 3.15.
133.
5 s
18 2
71 23 SAS -Revised EALs and minor revisions to text To address NRC comment # 71.
ofbasis.
134.
5 s
19 1
9 29 SSI - Added wording to basis concerning loads To incorporate intent of Q&A; ITF 2
72 24 which should be considered operable.
took no action on NRC comment 72 135.
.5 s
20 4
N/A SS2-Added Startup to mode applicability ITF decision. Also see writeup on SA2 concerning decision not to revise EAL 136.
5 S
20 1
7 28 SS2 - Changes were not made to the IC or EAL ITF decision based on this Question.
ChgSumRI. doc 18 10/18/98
a e
NEI 97-03 Rev 2 to 3 Change Summary Sec Cat Pg Source Cmat Pg Change Description Reason for Change No No No 137.
5 s
21 1
3 1
SS3 - Dele!-d reference to n alti, unit stations Multi unit concerns formerly from the basis addressed in the Q and A document have been subsumed into the generic text preceding section 5.0.
138.
5 s
22 2
55 19 SS4-Test was not changed to reflect concerns ITF believes that EAL already over loss of heat removal capability due to intake provides latitude for addressing this structure icing concern and that it does not warrant being a standalone EAL under the hazard ICs HUI or HA1.
139.
5 s
22 2
73 24 SS4 - Replaced generic example EAL with To address differences in generic separate examples for PWRs and BWRs.
plant design between PWRs and BWRs.
140.
5 s
22 2
74 25 SS4 - Revised basis to clarify that this IC does To clarify that suberiticality is not address reactivity control.
addressed in SS2.
141.
5 3
22 1
8 29 SS4 - No change made to NEI 97-03 based on See comment answer.
thisQ&A.
142.
5 s
22 4
N/A SS4 - Changed IC from " Complete Loss of To avoid confusion between ' loss of Function Needed to Achieve or Maintain Hot function needed to achieve hot Shutdown" to " Complete Loss of Heat Removal shutdown' and ' failure of RPS Capability" instmmentation to complete or initiate an automatic reactor scram once an RPS setpoint has been exceeded and manual scram was not successful (SS2)'.
ChgSumRI. doc 19 10/18/98 i
NEI 97-03 Rev 2 to 3 Change Summary Sec Cat Pg Source Cmnt Pg Change Description Reason for Change No No No 143.
5 s
22 4
N/A SS4 - Deleted statement addressed to multi-unit Generic guidance for multi-unit stations with shared safety functions suggesting stations is addressed in Section 3.10, consideration of the affect of the loss upon the i.e., " multi-unit stations with shared other unit and for escalation.
safety-related systems and functions must also consider the effects of a loss of a common system on more than one unit."
144.
5 s
23 4
N/A SSS - Changed effluence to effluent in basis.
To correct typographical error.
145.
5 s
24 4
N/A SS6 - Operating mode applicability was changed The modes identified in the EALs to include Startup for all designs were based on the standard technical specifications for BWRs and Westinghouse PWRs. To aid in interpreting these modes for PWRs from other NSSSs and for plant with non-standard technical specifications, the modes are now described in subsection 3.15.
146.
5 s
24 4
N/A SS6 - Initiating condition matrix for SS6 Formatting ofdefined terms as changed "significant transient" to specified in sections 2.3 and 5.4 "SIGNIFICANT TRANSIENT".
147.
5 S
24 4
N/A SS6 - Discussion of significant transient SIGNIFICANT TRANSIENTis removed from the basis defined in section 5.4, Definitions ChgSumR1. doc 20 10/18/98
Ty-NEI 97-03 Rev 2 to 3 Change Summary Sec Cat Pg Source Cmnt Pg Change Description Reason for Change No No No 148.
5 s
25 4
N/A SSti - Basis revised to include pl,anned and
" Planned" and
" UNPLANNED" unplanned actions are not difTerentiated since the loss of instnamentation of this magnitude is of such significance during a transient that the cause of the loss is not an ameliorating factor.
149.
5 s
26 4
N/A SGI -Minor text changes Consistency.
150.
5 s
28 4
N/A SG2 - Added Startup to mode applicability ITF decision. Also see writeup on.
SA2 concerning decision not to revise EAL l
I r
ChgSumRI. doc 21 10/18/98
NEI 97-03 Rev 2 to 3 Change Summary Sec Cat Pg Source Cmat Pg Change D scription Reason for Change No No No 151.
^PP-A 1
1-13 5-12 AUI, AU2, AAl, AA2, AA3, ASI, AGI - A significant change Clarification.
in the philosophy of classifying abnormal radiological effluent r
events was incorporated in Revision 3. In Revision 2, indications on radiation monitors would trigger a dose assessment usine real-time meteorology, and the classification would be based on the results of the dose assessment, which was required to be cornpleted within 15 minutes. With Revision 3, this requirement has been deleted as a prerequisite for classification. Appropriate dose assessments are still required to be performed as accident assessments by NUREG-0654, but are no longer recuired as part of the classification process. If results from dose assessments are available at the time the classification is made, the revised ICs require the Site Area ar4d General Emergency classifications to be based on these results. The largest number of questions received on the Revision 2 methodology were associated with this recognition category. In order to address these questions and to better explain the basis of the effluent ICs, Appendix A." Basis for Radiological Efiluent initiating Conditions was added in Revision 3.
Dose quantities implemented with the revisions to 10 CFR 20, such as TEDE, CDE, etc., have beer, incorporated in the ICs and EALs.
In AU2 and AA2, guidance as to how level decrease could be detected was added to the basis. The AUI reference to increase in airbome concentration was deleted. Text was also added to bases to restrict dry storage applicability in AU2 to dry storage licensed under 10 CFR 50, not that licensed under 10 CFR 72.
6 He field survey EALs were revised to specify " closed window" readings. His was done to eliminate confusion regarding constructions such as TEDE-rate.
ChgSumRI. doc
'22 10/18/98 i
kh' ISSUES FOR DISCUSSION DURING 10/21 TELECON WITH NEl EAL TASK FORCE AU1 Radioactive Effluent Would like to discuss the relationship between EALs 1 and 2 and the anticipated site-specific implementation of this guidance. In particular will some effluent monitors only be specified in EAL #2 and others in EAL #1. Will EAL #1 be applicable when a release occurs but one is not
, planned?
AA2 Loss of Water Level Would like to discuss how the alarms will be set and what the site-specific EALs may look like.
Fission Product Barrier:
Site Area Emergency Combinations Need additional information on appropriateness of declaring an Alert on the potential loss of containment.in combination with the loss or potential loss of other barriers.
Fission Product Barrier:
Steam Generator Tube Rupture We would like to discuss this new fission product EAL,
HU1 Natural and Destructive Phenomenon Affecting Plant Safety The change summary states that the effects of"high winds greater than (site specific) mph" was added to HU1, but the copy of NESP-007, Rev 3 reviewed did not add this item. Is it the intent to address high winds in HU17 HU3 Release of Toxic of Flammable Gases Deemed Detrimental to Safe Operation of the Plant HA3 contains wording regarding the sampling of gases. Should that wording be repeated here and should there be a time limit for sampling activities as suggested under questions to HA3?
The criteria for the EAL is confusing because its relationship to the criteria for the Alert EAL is not clear.
are both set at the flammable limit and it is the impact on plant systems that differs?
how does the life-threatening limit apply to HA3?
l A
l i
i'4
+
HA1#5 has been added to address intemal flooding.
Basis addresses inability to access systems and indication in control room that systems are degraded. Appears adequate to address intemal flooding hazards. Question of whether j
systems involved are assumed to be affected until otherwise proved? How does this relate to fire EAL, e.g., why'is this different than " fire in a affecting the operability of safety system"?
l Shouldn't the EALs be similar ?
HA3 Release of Toxic of Flammable Gases within a facility structure which jeopardizes operation of systems required to maintain safe operations or to establish or maintain cold shutdown The question in the 3/16/98 letter regarding the use of dissimilar language in HA2 and HA3 was not answered. Should the language describing safety systems in these two EALs be the same?
The request to define the concentration at which the EAL is met is discussed but not clear. Is the flammable limit the criteria for the EAL7 Should a time limit for sampling be established, beyond which the EAL would be considered as met?
HA4 Confirmed Security Event in a Plant Protected Area This revision adds an EAL for discovery of a bomb in the Protected Area that potentially affects safety related equipment. See discussion for HS1. If the bomb is in a vital area what wou!d the appropriate emergency class be?
HS1 Confirmed Security Event in a Plant Vital Area The changes clarify intent and are an improvement. However, the issue remains that a bomb found in a vital area may not be an SAE under this scheme. Rather it appears that a bomb found in the protected area that potentially affecting Safety Related Equipment is an Alert.
Does the discovery of a bomb in a vital area show that the safeguard protective measures have
- been seriously compromised and does this warrant an SAE declaration? Should this be stated in the EAL?
SU3 It is not clear how loss of indicators is supposed to be classified. Reference April 1996 comment on this issue.
SU4 Fuel Clad Degradation l
l The mode applicability was changed fam all modes to all modes except the defueled mode.
Additional information is needed to justify the change from all modes to all modes except the defueled mode. Notc that the basis states that the EAL is applicable in all modes.
I
y l
l SA1 Loss of AC Power i
The following statament was added to the basis for this EAL:
Co;asideration should be given to operable loads necessary to remove decay heat orprovide Reactor Vessel makeup capability when evaluating loss of AC l
po' ver to essential busses. Even though an essential bus may be energized, if
- ne :escary loads (i.e., loads that iflost would inhibit decay heat removal wpability or Reactor Vessel makeup capability) are not operable on the l
en stgized bus then the bus should not be considered operable.
l Additional information is need to understand how this information is to be used, i.e., is this to be i
- used in developing an EAL under this IC or to be referred to when classifying events so that i
' emergency directorjudgement may be applied.
SA2 Failure of RPS (Question on Revision 2)
Why does this EAL include the Hot Standby Mode, while the Site Area Emergency EAL does
- not? -
Is it warranted to describe that the scram does not have a power level associated with it (in i
contrast to the SAE level?
Also some confusion has existed over whether the RPS setpoints need to be listed (may want to put into basis). For the SAE level, is failure of a manualinitiated scram a SAE if the event was not initiated by a transient which exceeded a RPS setpdint.
p SA3 Loss of Cold Shutdown Function Need to evaluate basis information to clarify intent of EAL is to classify on both the loss of functions and exceeding temperature conditions.
SAS Single AC Power Source Although the intent of the change is acceptable, it seems that including the condition that "Any additional single failure will result in station blackout," should be modified to state that "Any additional single failure will result in loss of AC power to all the site-specific essential busses."
it is not clear if loss of all DGs with busses being feed by both trains of offsite power would be classified.
SS2 Failure of RPS (also added to SG2) l
. Added startup to the mode applicability. Need additional information justifying this change.
SS4 Complete Loss of Heat Removal Capability I-i
^
er
,,e
~
v n-w g
-w e,-r-n--
-, - = -,
s n
+
Q The Revision 2 EAL is:
Complete loss of any (site-specific) function required for hot shutdown.
This EAL was changed to provide address design differences between PWRs and BWRs as follows:
1.
Loss of core cooling and heat sink (PWR).
1.
Heat Capacity Temperature Limit Curve exceeded (BWR).
Further information is needed to ensure these are appropriate indications.
SG1 (issues on Revision 2)
Consider adding "to essential busses" to the IC to be consistent with other loss of power ICs.
Consider providing guidance regarding how to apply to basis statement "with appropriate allowance for offsite emergency response" Consider providing additional guidance in the basis related to EAL condition "b": site-specific indication of continuing degradation of core cooling. How is this to be applied in PWRs....
BWRs (ED judgement or specific setpoints)
SG2 Fal.ure of RPS Modified the basis to change the " indication for heat removal is extremely challenged" from 2/3 core height to Minimum Steam Cooling RPV Water Level.
Please provide additional information justifying this change.
i 1
4
e REVISION TO NESP-007 MILESTONES DATE (T/C) 1.
NEl to provide (1) revised NEl-97-03 with shutdown EAL guidance 10/98 removed and revision marked by revision bars, (2) analysis of modifications of EALs in NESP-007, and (3 ) response to issues identified in March 13,1998 comments to NEl on NEl-97-03 2.
Meet with NEl to discuss NEl-97-03' 10/19/98 3.
Comments resolved and final draft of NEl 97-03 issued to NRC for 12/98T cndorsement 4.
Revision to Regulatory Guide 1.101 (and companion Regulatory 2/99T
{
Analysis) endorsing NEl-97-03 developed in form of a draft guide for CRGR review 5.
CRGR/ACRS meeting on draft guide 2/99T 6.
Draft Guide issued for public comment 3/99T 7.
Public comments addressed (any needed revision to NEl-97-03 6/99T completed) 8.
CRGR/ACRS meeting on final guide 8/99T 9.
Regulatory Guide issued 10/99T
- s
+
i:
SHUTDOWN EAL GUIDANCE l
MILESTONES DATE (T/C) 1.
Meet with NEl to initiate dialog on industry effort to develop new 1/99 i
guidance for EALs applicable in the shutdown mode of operation i
2.
NEl to provide shutdown EAL guidance (in NEl-99-01) 3/99T 3.
NRC p ovides comments to NEl on NEl-99-01 5/99T 4.
Meet with NEl to discuss comments 6/99T 5.
Comments resolved and final draft of NEl-97-03 issued to NRC for 8.99T endorsement 6.
Revision to Regulatory Guide 1.101 (and companion Regulatory 10/99T Analysis) endorsing NEl-99-01 developed in form of a draft guide for CRGR/ACRS review. Draft Generic ! '.tter on shutdown EALs developed 7.
CRGR/ACRS meeting on draft guide and generic letter 12/99T 8.
Draft Guide issued for public comment 1/00T 9.
Public comments addressed (any needed revision to NEl-97-03 4/00T completed) 10.
CRGR/ACRS meeting on final guide and generic letter 6/00T 11.
Regulatory Guide and generic letter issued 8/00T l
i i
l i
w i
\\
EAL ITF j
TIMELINE Task #
NEl 97-03 Task Description Duration Due Date l
i18 ilTF migKNEltop w J Discuss shutdown EALNRC Letter received 8/3/98
' ' ' ' ~
y.,. gy..
12 daysj 8/3 4/98 l
M DetonnerEITF'sArectishMbundimMiden timeline Q n__.;;.
^
a l
.' Revise NEl9703 draft 163 Muddown EALs
~
s,*
Replace!SA3[SS5/SU7$
N M
.' sund word red liiw' comparison i
124 IMoonference; call to. update all on,lTE direction changej
- 60. mins; i8/7/98h pacopeM.timelinep _.
.. y
~L~
11.2 (EST)")
s s
.M 63(M6371pmeetinoID# is,'8480I 5 s
- 3@
- lTF;mtg.at Stof,
, < L. g Adays; 91 54-15/98j M (Duvoloplisl response lettei46 Shutdown.EAlfNRC lett,e(
'~ L,. ;i.~
received 8/3/98 sequestl6dplRCl response;"ts;lTFT co.m...ments by1.1/1/98i
~
cm x..
. m 1
.=
4m
%! EBulld a...m,_samplesummarydocumentwlthex _
.-es amples?
=?3AjssignNZ?$$$flaummaryE
'L[tol
^~
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in Build ' eferense docu,msn.ts 56. d di.s. thbule.._
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4.
ITF members:
5 days Bulld summary of changes to include cross-reference per e
of Q&A and NRC Industry lessons learned and any assignee scenarios deemed necessary for validation of changes 10/2 Send changes and scenarios to Walt Lee 5.
Compile summary document and scenarios - Lee 3 days 10/9 6.
Distribute compiled summary document for ITF review 10/12 prior to final proof - Nelson 7.
ITF mtg at NEl to:
Meet w!NRC to resolve 97-03 changes (Mtg-White 1 day 10/19 Flint) 1.5 days 10/20-21 Meet at NEl to build NEl 99-01 e
8.
NEl submit 97-03 for Industry review to include change 30 days 11/2 12-1 summary NEl compile and distribute 97-03 Industry comment.s 11/2 12/1 e
to ITF electronically for review (on-going)
NEl distribute final set of Industry comments to ITF 5 days 12/3 e
9.
ITF mtg at NEl to:
1.5 days 12/10-11 ITF review / propose resolution to industry comments Revise 97-03 to incorporate Industry comments e
10.
Submit 97-03 to NRC for Reg Guide 1.101 endorsement 12/18 l
I Page 1 8/4/98
s i '
11.
NRC lasue revised Reg. Guide 1.101 endorsing 97-03 10/99 Revise Reg. Guide 1.101 to endorse 97-03 Place 97-03 in Federal Register for Public comment l
Compile 97-03 comments l
Resolve 97-03 Public comments
=
Revise 97-03 to incorporate Public comments i
i i
r i
i 6
4 4
I Page 2 8/4/98
., A s.
}
}.1, i
l NEl 99-01 l
EAL ITF TIMELINE Task #
NEl 99-01 Task Description Duration Due Date i
1.
Develop defueled EAL strawman - Stobaugh 10/18/98 2.
Develop IFSIS/ Dry cask strawman - Costello 10/20/98 3.
ITF meet at NEl to develop draft NEl 99-01:
1.5 days 10/20-21 l
Shutdown ICs/EALs and bases section 1 day 1/19 Develop V&V scenarios based on NUREG 1449 e
]
Defueled ICs/EALs and bases section Dry Storage ICs/EALs and bases section 2.0 ITF to meet w/NRC at White Filnt to 2 days 1/20-E1 i.
roundtable/V&V Shutdown EALs and to present b7 2f
]
Defueled/ Dry Storage EALs J
ITF to meet at NEl to incorporate Shutdown EAL comments from NRC meeting ITF to meet w/NRC at White Flint to roundtable i
Shutdown EALs
-l ITF to meet at NEl to finalize Shutdown EALs 3.0 ITF meet to finalize NEl 99-01 2 days 3/3-4 4
4.0 NEl submit 9941 for Industry review 45 days 3/15-4/30 1
NEl compile 99-01 Industry comments and distribute to ITF for review
{
5.0 Meet at NEl to:
2 days 5/12-13 j
Resolve 99-01 Industry comments Revise 99-01 to incorporate industry comments 6.0 NEl 99-01 Panel (Part of NEl EP Forum - St.
0.5 day 6/7-8/99 Petersburg, Fl.)
7.0 ITF meet to:
?
?
l Incorporate any issues identified in Workshop Meet w/NRC to present and resolve any open or new changes NEl 99-01 final proof 8.0 Submit 99-01 to NRC for Reg Guide 1.101 7/15/99 endorsement 9.0 NRC lasue revised Reg. Guide 1.101 endorsing 99-01 Revise Reg. Guide 1.101 to endorse 99-01 Place 99-01 in Federal Register for Public comment Compile 99-01 comments Resolve 99-01 Public comments Revise 99-01 to incorporate Public comments Page 3 8/4/98
-. _.