RC-14-0032, Virgil C. Summer Unit 1, License Amendment Request LAR-14-02392, Request for NRC Approval of Proposed Changes to Emergency Action Levels. Attachments Iii, Iv, V & Vi: Technical Bases for Proposed EALs (Clean Copy), NEI 99-01 Revision 6 EA
Text
Document Control Desk Attachment III LAR-14-02392 RC-14-0032 Page 1 of 347 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) Unit I DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 ATTACHMENT III Technical Bases Document for the Proposed VCSNS EALs (Clean Copy)
EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]ENCLOSURE I Emergency Action Level Technical Bases DRAFT E CKW 3/20/14 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]TABLE OF CONTENTS SECTION TITLE PAGE 1.0 P U R P O S E .............................................................................................
..1 2.0 D IS C U S S IO N ..........................................................................................
1 2.1 B ackground
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1 2.2 Fission Product Barrier Thresholds
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2 2.3 Fission Product Barrier Classification Criteria ...............................
3 2.4 EAL Organization
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3 2.5 Technical Bases Information
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6 2.6 Operating Mode Applicability
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7 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS
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9 3.1 General Considerations
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9 3.1.1 Classification Timeliness
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9 3.1.2 Valid Indications
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9 3.1.3 Imminent Conditions
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9 3.1.4 Planned vs. Unplanned Events .......................................
10 3.1.5 Classification Based on Analysis ....................................
10 3.1.6 Emergency Director Judgment .........................................
10 3.2 Classification Methodology
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11 3.2.1 Classification of Multiple Evenets and Conditions
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11 3.2.2 Consideration of Mode Changes During Classification
........ 11 3.2.3 Classification of Imminent Conditions
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12 3.2.4 Emergency Classification Level Upgrading and Downgrading
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12 3.2.5 Classification of Short-Lived Events .................................
13 3.2.6 Classification of Transient Conditions
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13 3.2.7 After-the-Fact Discovery of an Emergency Event or C ondition ........................................................................
..14 3.2.8 Retraction of an Emergency Declaration
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14 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]TABLE OF CONTENTS SECTION TITLE PAGE
4.0 REFERENCES
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15 4.1 Developmental
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15 4.2 Implementing
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15 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS
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17 5.1 D efinitions
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..17 5.2 Acronyms & Abbreviations
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23 6.0 VCSNS-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE
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26 7.0 ATTACHMENTS
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30 7.1 Attachment 1 -Emergency Action Level Technical Bases ...........
31 Category R Abnormal Rad Levels / Rad Effluent ........................
32 Category C Cold Shutdown / Refueling System Malfunction
........ 83 Categqory H Hazards and Other Conditions Affecting Plant Safety...
143 Cateqory S System Malfunction
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197 Cateqory F Fission Product Barrier Degradation
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259 C ategory I IS FS I .............................................................................
266 7.2 Attachment 2 -Fission Product Barrier Matrix and Bases .................
269 7.3 Attachment 3 -Figures ......................................................................
328 7.4 Attachment 4 -Safe Operation
& Shutdown Areas Tables R-2 & H-3 Bases ....................................................................
339 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]1.0 PURPOSE This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the EAL Upgrade Project for Vigil C. Summer Nuclear Station (VCSNS).It should be used to facilitate review of the VCSNS EALs and provide historical documentation for future reference.
Decision-makers responsible for implementation of EPP-001, Activation and Implementation of Emergency Plan, may use this document as a technical reference in support of EAL interpretation.
This information may assist the Emergency Director in making classifications, particularly those involving judgment or multiple events. The basis information may also be useful in training and for explaining event classifications to off-site officials.
The expectation is that emergency classifications are to be made as soon as conditions are present and recognizable for the classification, but within 15 minutes in all cases of conditions present. Use of this document for assistance is not intended to delay the emergency classification.
2.0 DISCUSSION
2.1 Background
EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the VCSNS Emergency Plan.In 1992, the NRC endorsed NUMARC/NESP-007 "Methodology for Development of Emergency Action Levels" as an alternative to NUREG-0654 EAL guidance.NEI 99-01 (NUMARC/NESP-007)
Revisions 4 and 5 were subsequently issued for industry implementation.
Enhancements over earlier revisions included: " Consolidating the system malfunction initiating conditions and example emergency action levels which address conditions that may be postulated to occur during plant shutdown conditions.
- Initiating conditions and example emergency action levels that fully address conditions that may be postulated to occur at permanently Defueled Stations and Independent Spent Fuel Storage Installations (ISFSIs).Page 4 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Simplifying the fission product barrier EAL threshold for a Site Area Emergency.
Subsequently, Revision 6 of NEI 99-01 has been issued which incorporates resolutions to numerous implementation issues including the NRC EAL FAQs. Using NEI 99-01 Revision 6 Final, November 2012 (ADAMS Accession Number ML110240324) (ref. 4.1.1), VCSNS conducted an EAL implementation upgrade project that produced the EALs discussed herein.2.2 Fission Product Barrier Thresholds Fission product barrier thresholds represent threats to the defense in depth design concept that precludes the release of radioactive fission products to the environment.
This concept relies on multiple physical barriers, any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment.
Many of the EALs derived from the NEI methodology are fission product barrier threshold based. That is, the conditions that define the EALs are based upon thresholds that represent the loss or potential loss of one or more of the three fission product barriers."Loss" and "Potential Loss" signify the relative damage and threat of damage to the barrier.A "Loss" threshold means the barrier no longer assures containment of radioactive materials; A "Potential Loss" threshold implies an increased probability of barrier loss and decreased certainty of maintaining the barrier.The primary fission product barriers are: A. Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.B. Reactor Coolant System (RCS): The RCS Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves.C. Containment (CMT): The Containment Barrier includes the containment building and connections up to and including the outermost containment isolation valves.This barrier also includes the main steam, feedwater, and blowdown line extensions Page 5 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]outside the containment building up to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for escalation of the emergency classification level (ECL) from Alert to a Site Area Emergency or a General Emergency.
2.3 Fission
Product Barrier Classification Criteria The following criteria are the bases for event classification related to fission product barrier loss or potential loss: Alert: Any loss or any potential loss of either Fuel Clad or RCS barrier Site Area Emergency:
Loss or potential loss of any two barriers General Emergency:
Loss of any two barriers and loss or potential loss of the third barrier 2.4 EAL Organization The VCSNS EAL scheme includes the following features:* Division of the EAL set into three broad groups: o EALs applicable under all plant operating modes -This group would be reviewed by the EAL-user any time emergency classification is considered.
o EALs applicable only under hot operating modes -This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, Hot Standby, Startup, or Power Operation mode.o EALs applicable only under cold operating modes -This group would only be reviewed by the EAL-user when the plant is in Cold Shutdown, Refuel or Defueled mode.The purpose of the groups is to avoid review of hot condition EALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a Page 6 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]hot condition.
This approach significantly minimizes the total number of EALs that must be reviewed by the EAL-user for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identification of the EAL that applies to the emergency.
Within each group, assignment of EALs to categories/subcategories:
Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user.
The VCSNS EAL categories are aligned to and represent the NEI 99-01 "Recognition Categories." Subcategories are used in the VCSNS scheme as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds.
The VCSNS EAL categories/subcategories are listed below.Page 7 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]EAL Groups, Categories and Subcategories EAL Group/Category EAL Subcategory Any Operating Mode: R -Abnormal Rad Levels / Rad Effluent 1 -Radiological Effluent 2 -Irradiated Fuel Event 3 -Area Radiation Levels H -Hazards and Other Conditions 1 -Security Affecting Plant Safety 2 -Seismic Event 3 -Natural or Technological Hazard 4 -Fire or Explosion 5 -Hazardous Gas 6- Control Room Evacuation 7 -Judgment I -Independent Spent Fuel Storage 1 -Confinement Boundary Installation (ISFSI)Hot Conditions:
S -System Malfunction 1 -Loss of ESF AC Power 2 -Loss of Vital DC Power 3 -Loss of Control Room Indications 4 -RCS Activity 5 -RCS Leakage 6 -RTS Failure 7 -Loss of Communications 8 -Containment Isolation Failure F -Fission Product Barrier Degradation None Cold Conditions:
C -Cold Shutdown / Refueling System 1 -RCS Level Malfunction 2 -Loss of ESF AC Power 3 -RCS Temperature 4 -Loss of Vital DC Power 5 -Loss of Communications Page 8 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]The primary tool for determining the emergency classification level (ECL) is the EAL Classification Matrix. The user of the EAL Classification Matrix may (but is not required to)consult the EAL Technical Bases Document in order to obtain additional information concerning the EALs under classification consideration.
The user should consult Section 3.0, and Attachments 1 & 2 of this document for such information.
2.5 Technical
Bases Information EAL technical bases are provided in Attachment 1 for each EAL according to EAL group (Any, Hot, Cold), EAL category (R, C, H, S, F and E) and EAL subcategory.
A summary explanation of each category and subcategory is given at the beginning of the technical bases discussions of the EALs included in the category.
For each EAL, the following information is provided: Category Letter & Title Subcategory Number & Title Initiating Condition (IC)Site-specific description of the generic IC given in NEI 99-01 Rev. 6.EAL Identifier (enclosed in rectangle)
Each EAL is assigned a unique identifier to support accurate communication of the emergency classification to onsite and offsite personnel.
Four characters define each EAL identifier:
- 1. First character (letter):
Corresponds to the EAL category as described above (R, C, H, S, F or I)2. Second character (letter):
The emergency classification (G, S, A or U)G = General Emergency S = Site Area Emergency A = Alert U = Unusual Event 3. Third character (number):
Subcategory number within the given category.Subcategories are sequentially numbered beginning with the number one (1). If Page 9 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]a category does not have a subcategory, this character is assigned the number one (1).4. Fourth character (number):
The numerical sequence of the EAL within the EAL subcategory.
If the subcategory has only one EAL, it is given the number one (1).Classification (enclosed in rectangle):
Unusual Event (U), Alert (A), Site Area Emergency (S) or General Emergency (G)EAL (enclosed in rectangle)
Exact wording of the EAL as it appears in the EAL Classification Matrix Mode Applicability One or more of the following plant operating conditions comprise the mode to which each EAL is applicable:
1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown, 5 -Cold Shutdown, 6 -Refueling, D -Defueled, or All. (See Section 2.6 for operating mode definitions)
Definitions:
If the EAL wording contains a defined term, the definition of the term is included in this section. These definition can also be found in Section 5.1.Basis: A Plant-Specific basis section that provides VCSNS-relevant information concerning the EAL. This is followed by a Generic basis section that provides a description of the rationale for the EAL as provided in NEI 99-01 Rev. 6.VCSNS Basis Reference(s):
Site-specific source documentation from which the EAL is derived 2.6 Operating Mode Applicability (ref. 4.1.7)1 Power Operation Keff -0.99 and rated thermal power > 5%.Page 10 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]2 StartuD Ke ff- 0.99 and rated thermal power < 5%.3 Hot Standby Keff < 0.99 and average coolant temperature Tavg > 350 0 F.4 Hot Shutdown Keff < 0.99 and average coolant temperature 350°F > Tavg > 200OF 5 Cold Shutdown Keff < 0.99 and average coolant temperature Tavg -200 0 F.6 Refuel Keff < 0.95 and average coolant temperature Tavg -140°F Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.D Defueled All reactor fuel removed from reactor pressure vessel (full core off load during refueling or extended outage).The plant operating mode that exists at the time that the event occurs (prior to any protective system or operator action being initiated in response to the condition) should be compared to the mode applicability of the EALs. If a lower or higher plant operating mode is reached before the emergency classification is made, the declaration shall be based on the mode that existed at the time the event occurred.Page 11 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS
3.1 General
Considerations When making an emergency classification, the Emergency Director must consider all information having a bearing on the proper assessment of an Initiating Condition (IC). This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes and the informing basis information.
In the Recognition Category F matrices, EALs are based on loss or potential loss of Fission Product Barrier Thresholds.
3.1.1 Classification
Timeliness NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate ECL. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01, Interim Staff Guidance, Emergency Planning for Nuclear Power Plants (ref. 4.1.14).3.1.2 Valid Indications All emergency classification assessments shall be based upon VALID indications, reports or conditions.
An indication, report, or condition is considered to be VALID when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed.Implicit in this definition is the need for timely assessment.
3.1.3 Imminent
Conditions For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing radiological release is detected and the release Page 12 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]start time is unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.3.1.4 Planned vs. Unplanned Events A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that 1) the activity proceeds as planned and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component.
In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected.
Events or conditions of this type may be subject to the reporting requirements of 10 § CFR 50.72 (ref. 4.1.4).3.1.5 Classification Based on Analysis The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments, chemistry sampling, RCS leak rate calculation, etc.); the EAL and/or the associated basis discussion will identify the necessary analysis.
In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available).
The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift).
3.1.6 Emergency
Director Judgment While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary.
The NEI 99-01 scheme provides the Emergency Director with the ability to classify events and conditions based upon judgment using EALs that are consistent with the ECL definitions (refer to Category H). The Emergency Director will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL Page 13 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]definition.
A similar provision is incorporated into the Fission Product Barrier Tables;judgment may be used to determine the status of a fission product barrier.3.2 Classification Methodology To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded.
The evaluation of an EAL(s) must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, then the IC is considered met and the associated ECL is declared in accordance with plant procedures.
When assessing an EAL that specifies a time duration for the off-normal condition, the"clock" for the EAL time duration runs concurrently with the emergency classification process "clock". For a full discussion of this timing requirement, refer to NSIR/DPR-ISG-01 (ref. 4.1.14).3.2.1 Classification of Multiple Events and Conditions When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared.
For example:* If an Alert EAL and a Site Area Emergency EAL are met, whether at one unit or at two different units, a Site Area Emergency should be declared.There is no "additive" effect from multiple EALs meeting the same ECL. For example:* If two Alert EALs are met, whether at one unit or at two different units, an Alert should be declared.Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events (ref. 4.1.2).3.2.2 Consideration of Mode Changes During Classification The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable.
If an Page 14 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]event or condition occurs, and results in a mode change before the emergency is declared, the ECL is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared).
Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition.
For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response.
In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher.3.2.3 Classification of Imminent Conditions Although EALs provide specific thresholds, the Emergency Director must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT).
If, in the judgment of the Emergency Director, meeting an EAL is IMMINENT, the emergency classification should be made as if the EAL has been met. While applicable to all ECLs, this approach is particularly important at the higher ECLs since it provides additional time for implementation of protective measures.3.2.4 Emergency Classification Level Upgrading and Downgrading By generic industry classification guidance, an ECL may be downgraded when the event or condition that meets the highest IC and EAL no longer exists, and other site-specific downgrading requirements are met. However, per VCSNS Emergency Plan implementing procedure guidance, down-grading of the ECL is not performed, rather the ECL is simply terminated.
As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02 (ref. 4.1.2).Page 15 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]3.2.5 Classification of Short-Lived Events Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance.
By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed.
If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration.
Examples of such events include a failure of the reactor protection system to automatically scram/trip the reactor followed by a successful manual scram/trip or an earthquake.
3.2.6 Classification
of Transient Conditions Many of the ICs and/or EALs contained in this document employ time-based criteria.These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted.
In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes).
The following guidance should be applied to the classification of these conditions.
EAL momentarily met during expected plant response -In instances where an EAL is briefly met during an expected (normal) plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are performed in accordance with procedures.
EAL momentarily met but the condition is corrected prior to an emergency declaration
-If an operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the applicable EAL is not considered met and the associated emergency declaration is not required.
For illustrative purposes, consider the following example: An ATWS occurs and the Emergency Feedwater system fails to automatically start.Steam generator levels rapidly decrease and the plant enters an inadequate RCS heat removal condition (a potential loss of both the fuel clad and RCS barriers).
If an operator manually starts the Emergency Feedwater system in accordance with Page 16 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]an EOP step and clears the inadequate RCS heat removal condition prior to an emergency declaration, then the classification should be based on the ATWS only.It is important to stress that the 15-minute emergency classification assessment period is not a "grace period" during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event; emergency classification assessments must be deliberate and timely, with no undue delays. The provision discussed above addresses only those rapidly evolving situations where an operator is able to take a successful corrective action prior to the Emergency Director completing the review and steps necessary to make the emergency declaration.
This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions.
3.2.7 After-the-Fact Discovery of an Emergency Event or Condition In some cases, an EAL may be met but the emergency classification was not made at the time of the event or condition.
This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery.
This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process.In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1022 (ref. 4.1.3) is applicable.
Specifically, the event should be reported to the NRC in accordance with 10 CFR § 50.72 (ref. 4.1.4) within one hour of the discovery of the undeclared event or condition.
The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements.
3.2.8 Retraction
of an Emergency Declaration Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022 (ref. 4.1.3).Page 17 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]
4.0 REFERENCES
4.1 Developmental
4.1.1 NEI 99-01 Rev. 6 Final, Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML1 10240324.4.1.2 RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007.4.1.3 NUREG-1022 Event Reporting Guidelines:
10CFR50.72 and 50.73 4.1.4 10 § CFR 50.72 Immediate Notification Requirements for Operating Nuclear Power Reactors 4.1.5 10 § CFR 50.73 License Event Report System 4.1.6 VCSN Offsite Dose Calculation Manual (ODCM)4.1.7 Technical Specifications Table 1.1 4.1.8 EP-100 Radiation Emergency Plan 4.1.9 OAP-108.4 Operations Outage Control of Containment Penetrations 4.1.10 SSP-004 Outage Safety Review Guidelines 4.1.11 EPP-001 Activation and Implementation of Emergency Plan, Section 3.7 4.1.12 Drawing SS-024-019 Site Plan 4.1.13 OAP-103.5 EOP/AOP Writer's Guide 4.1.14 NSIR/DPR-ISG-01, Interim Staff Guidance, Emergency Planning for Nuclear Power Plants 4.1.15 Certificate of Compliance No. 1032 Appendix A Technical Specifications for the HI-STORM FW MPC Storage System 4.2 Implementing 4.2.1 EPP-001 Activation and Implementation of Emergency Plan 4.2.2 NEI 99-01 Rev. 6 to VCSNS EAL Comparison Matrix Page 18 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]4.2.3 VCSNS EAL Matrix Page 19 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS
5.1 Definitions
(ref. 4.1.1 except as noted)Selected terms used in Initiating Condition and Emergency Action Level statements are set in all capital letters (e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document.
The definitions of these terms are provided below.Confinement Boundary The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As applied to the VCS ISFSI, the confinement boundary is defined to be the HI-STORM Multi-Purpose Canister (MPC) (ref. 4.1.15).Containment Closure The procedurally defined conditions or actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.
OAP-1 08.4, Operations Outage Control of Containment Penetrations, and SSP-004, Outage Safety Review Guidelines, prescribe Containment Closure actions and associated conditions.
A containment condition in which all penetrations providing direct access from the containment atmosphere to the outside atmosphere are closed by at least one automatic isolation valve, blind flange, or manual valve (ref. 4.1.9). Containment Closure is applicable to Mode 5 reduced RCS inventory operation and during Mode 6 Core Alterations or movement of spent fuel in the Reactor Building.
The Reactor Building penetrations shall be closed or capable of being closed prior to the onset of core boiling upon loss of the decay heat removal capability (ref. 4.1.10): A. The equipment door held in place by a minimum of four bolts.B. A minimum of one door in each personnel airlock closed.C. Each penetration providing direct access from the Reactor Building atmosphere to the outside atmosphere shall be either: 1. Closed by an isolation valve, blind flange, or manual valve, or Page 20 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]2. Capable of being closed by an Operable automatic Reactor Building Purge and Exhaust isolation valve.D. All temporary penetrations are sealed.Emergency Action Level (EAL)A pre-determined, site-specific, observable threshold for and Initiating Condition (IC) that, when met or exceeded, places the plant in a given emergency classification level (ECL).Emergency Classification Level (ECL)One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions.Emergency Control Officer (ECOL)A senior VCSNS employee with overall responsibility for coordinating emergency response actions of the station, and the ERO with the affected state(s) and county agencies.EPA PAGs Environment Protection Agency Protective Action Guidelines.
The EPA PAGs are expressed in terms of dose commitment:
1 Rem TEDE or 5 Rem CDE Thyroid. Actual or projected offsite exposures in excess of the EPA PAGs requires VCSNS to recommend protective actions for the general public to offsite planning agencies.Explosion A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization.
A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion.
Such events require a post-event inspection to determine if the attributes of an explosion are present.Faulted Page 21 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.
Fire Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.Flooding A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.Hostage A person(s) held as leverage against the station to ensure that demands will be met by the station.Hostile Action An act toward VCSNS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on VCSNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA) (ref. 4.1.12).Hostile Force One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.
Page 22 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Imminent The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.Intrusion The act of entering without authorization.
Discovery of a bomb in a specified area is indication of intrusion into that area by a HOSTILE FORCE.Owner Controlled Area Area between the vehicle barrier system and the PROTECTED AREA barrier (ref. 4.1.12).Plant Operator Any member of the plant staff who, by virtue of training and experience, is qualified to assess the indications or reports for validity and to compare the same to the EAL Matrix in Attachment I. The plant operator has the authority to declare the appropriate EAL and activate the Radiation Emergency Plan. A Plant Operator does not encompass plant personnel such as chemists, radiation protection technicians, security personnel, and others whose position require they report rather than assess abnormal conditions to the Control Room or Technical Support Center. The Plant Operator is the Duty Shift Supervisor or the Emergency Director responsible for managing the Emergency Response (ref. 4.2.1 ).Projectile An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.Protected Area An area encompassed by physical barriers and to which access is controlled.
The Protected Area refers to the designated security area around the process buildings and is depicted in Drawing SS-024-019 Site Plan (ref. 4.1.12).Page 23 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]RCS Intact The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams).Reduced Inventory The procedurally defined condition when the Reactor Vessel level is greater than three (3)feet (36") below the Reactor Vessel flange with fuel in the vessel. This level corresponds to RCS level less than 434'-7.43" (ref. 4.1.10).Refueling Pathway The reactor refueling cavity, spent fuel pool and fuel transfer canal comprise the refueling pathway.Ruptured The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
Safety System A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary;(2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Page 24 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Security Condition Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A security condition does not involve a hostile action.Site Boundary All land and water areas inside the one mile radius of the Reactor Building, use of which must be authorized by SCE&G (ref. 4.1.12).Unisolable An open or breached system line that cannot be isolated, remotely or locally.Unplanned A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown.Valid An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.
Visible Damage Damage to a equipment or structure that is readily observable without measurements, testing, or analysis.
The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected equipment or structure.
Page 25 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]5.2 Acronyms & Abbreviations OF .......................................................................................................
Degrees Fahrenheit
..................................................................................................................
Feet or M inutes" ...........................................................................................................
Inches or Seconds Ir ...........................................................................................................................
Gam ma rq ...........................................................................................................................
Neutron AB ..........................................................................................................
Auxiliary Building AC ........................................................................................................
Alternating Current AO P .................................................................................
Abnorm al Operating Procedure ATW S ......................................................................
Anticipated Transient W ithout Scram CDE ......................................................................................
Com m itted Dose Equivalent CFR .....................................................................................
Code of Federal Regulations CMT .............................................................................................................
Containment CSF ...............................................................................................
Critical Safety Function CSFST ......................................................................
Critical Safety Function Status Tree DBA ...............................................................................................
Design Basis Accident DC ...............................................................................................................
Direct Current EAL .............................................................................................
Em ergency Action Level ECCS ............................................................................
Em ergency Core Cooling System ECL ..................................................................................
Em ergency Classification Level EO F ..................................................................................
Em ergency Operations Facility EO P ..............................................................................
Emergency Operating Procedure EPA ...............................................................................
Environm ental Protection Agency EPIP .................................................................
Em ergency Plan Im plem enting Procedure EPRI .............................................................................
Electric Power Research Institute ERG ...............................................................................
Em ergency Response Guideline ESF .................................................................................
Engineered Safeguards Feature FEMA ...............................................................
Federal Em ergency Managem ent Agency FSAR ....................................................................................
Final Safety Analysis Report HSI .............................................................................................
Hum an System Interface IC .........................................................................................................
Initiating Condition ID ..............................................................................................................
Inside Diam eter IPEEE .................
Individual Plant Examination of External Events (Generic Letter 88-20)ISFSI ............................................................
Independent Spent Fuel Storage Installation Keff ..........................................................................
Effective Neutron M ultiplication Factor LCO ..................................................................................
Lim iting Condition of Operation LOCA .........................................................................................
Loss of Coolant Accident Page 26 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]M SIV ......................................................................................
Main Steam Isolation Valve M SL ........................................................................................................
Main Steam Line m R, m Rem , m rem , m REM ...............................................
m illi-Roentgen Equivalent Man MW ....................................................................................................................
Megawatt NEI ...............................................................................................
Nuclear Energy Institute NPP ..................................................................................................
Nuclear Power Plant NRC ...............................................................................
Nuclear Regulatory Com m ission NSSS ................................................................................
Nuclear Steam Supply System NO RAD ...................................................
North Am erican Aerospace Defense Com mand (NO)UE ............................................................................. (Notification Of) Unusual Event NUMARC 1 .............................
..................... Nuclear Managem ent and Resources Council O BE ......................................................................................
Operating Basis Earthquake OCA ..............................................................................................
Owner Controlled Area O DCM/O DAM ..........................................
Offsite Dose Calculation (Assessm ent) Manual O RO .................................................................................
Off-site Response O rganization PA ..............................................................................................................
Protected Area PAG ........................................................................................
Protective Action Guideline PRA/PSA .....................
/ Probabilistic Safety Assessment PW R .......................................................................................
Pressurized W ater Reactor PS .........................................................................................................
Protection System PSIG ...............................................................................
Pounds per Square Inch Gauge R ........................................................................................................................
Roentgen RB ...........................................................................................................
Reactor Building RCC ............................................................................................
Reactor Control Console RCS ............................................................................................
Reactor Coolant System Rem , rem , REM .......................................................................
Roentgen Equivalent Man RETS .........................................................
Radiological Effluent Technical Specifications RPS ........................................................................................
Reactor Protection System RPV ...........................................................................................
Reactor Pressure Vessel RVLIS .......................................................
Reactor Vessel Level Instrumentation System SAR ...............................................................................................
Safety Analysis Report SAS ..........................................................................................
Safety Autom ation System SBO .........................................................................................................
Station Blackout SCBA ......................................................................
Self-Contained Breathing Apparatus SG ..........................................................................................................
Steam Generator 1 NUMARC was a predecessor organization of the Nuclear Energy Institute (NEI).Page 27 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]SI ..............................................................................................................
Safety Injection SPDS ...........................................................................
Safety Parameter Display System SRO ............................................................................................
Senior Reactor Operator SSC ..........................................................................
Structures, System s & Com ponents TEDE ...............................................................................
Total Effective Dose Equivalent TOAF ....................................................................................................
Top of Active Fuel TSC ...........................................................................................
Technical Support Center W OG ...................................................................................
W estinghouse Owners Group Page 28 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]6.0 VCNS-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of a VCNS EAL within the NEI 99-01 IC/EAL identification scheme. Further information regarding the development of the VCNS EALs based on the NEI guidance can be found in the EAL Comparison Matrix.VCNS NEI 99-01 Rev. 6 EAL IC Example EAL RU1.1 AU1 1,2 RU1.2 AU1 3 RU2.1 AU2 1 RA1.1 AA1 1 RA1.2 AA1 2 RA1.3 AA1 4 RA2.1 AA2 1 RA2.2 AA2 2 RA2.3 AA2 3 RA3.1 AA3 1 RA3.2 AA3 2 RS1.1 AS1 1 RS1.2 AS1 2 RS1.3 AS1 3 RS2.1 AS2 1 RG1.1 AG1 1 RG1.2 AG1 2 RG1.3 AG1 3 Page 29 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]VCNS NEI 99-01 Rev. 6 EAL IC Example EAL RG2.1 AG2 1 CUI.1 CUl 1 CU1.2 CUl 2 CU2.1 CU2 1 CU3.1 CU3 1 CU3.2 CU3 2 CU4.1 CU4 1 CU5.1 CU5 1,2,3 CA1.1 CA1 1 CA1.2 CA1 2 CA2.1 CA2 1 CA3.1 CA3 1 CA3.2 CA3 2 CS1.1 CS1 1 CS1.2 CS1 2 CS1.3 CS1 3 CG1.1 CG1 1 CG1.2 CG1 2 FA1.1 FA1 1 FS1.1 FS1 1 FG1.1 FG1 1 HU1.1 HU1 1,23 HU2.1 HU2 1 Page 30 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]VCNS NEI 99-01 Rev. 6 EAL IC Example EAL HU3.1 HU3 1 HU3.2 HU3 2 HU3.3 HU3 3 HU3.4 HU3 4 HU4.1 HU4 1 HU4.2 HU4 2 HU4.3 HU4 3 HU4.4 HU4 4 HU7.1 HU7 1 HA1.1 HA1 1,2 HA2.1 CA6 1 SA9 1 HA3.1 CA6 1 SA9 1 HA4.1 CA6 1 SA9 1 HA5.1 HA5 1 HA6.1 HA6 1 HA7.1 HA7 1 HS1.1 HS1 1 HS6.1 HS6 1 HS7.1 HS7 1 HG1.1 HG1 1 HG7.1 HG7 1 Page 31 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]VCNS NEI 99-01 Rev. 6 EAL IC Example EAL SU1.1 SUl 1 SU3.1 SU2 1 SU4.1 SU3 1 SU4.2 SU3 2 SU5.1 SU4 1 SU6.1 SU5 1 SU6.2 SU5 2 SU7.1 SU6 1,2,3 SU8.1 SU7 1 SU8.2 SU7 2 SA1.1 SAl 1 SA3.1 SA2 1 SA6.1 SA5 1 SS1.1 SS1 1 SS2.1 SS8 1 SS6.1 SS5 1 SG1.1 SG1 1 SG1.2 SG8 1 IUl E-HU1 1 Page 32 of 346 EPP-1 08 ENCLOSURE I REVISION 01 [DRAFT E]7.0 ATTACHMENTS
7.1 Attachment
1, Emergency Action Level Technical Bases 7.2 Attachment 2, Fission Product Barrier Matrix and Basis 7.3 Attachment 3, Figures Page 33 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]ATTACHMENT 1 EMERGENCY ACTION LEVEL TECHNICAL BASES Page 34 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category R -Abnormal Rad Levels / Radiological Effluent EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)Many EALs are based on actual or potential degradation of fission product barriers because of the elevated potential for offsite radioactivity release. Degradation of fission product barriers though is not always apparent via non-radiological symptoms.
Therefore, direct indication of elevated radiological effluents or area radiation levels are appropriate symptoms for emergency classification.
At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems or precursors to more significant releases.
At higher release rates, off site radiological conditions may result which require offsite protective actions. Elevated area radiation levels in plant may also be indicative of the failure of containment systems or preclude access to plant vital equipment necessary to ensure plant safety.Events of this category pertain to the following subcategories:
- 1. Radiological Effluent Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits. Projected offsite doses, actual offsite field measurements or measured release rates via sampling indicate doses or dose rates above classifiable limits.2. Irradiated Fuel Event Conditions indicative of a loss of adequate shielding or damage to irradiated fuel may preclude access to vital plant areas or result in radiological releases that warrant emergency classification.
- 3. Area Radiation Levels Sustained general area radiation levels which may preclude access to areas requiring continuous occupancy also warrant emergency classification.
Page 35 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
R -Abnormal Rad Levels / Radiological Effluent Subcategory:
1 -Radiological Effluent Initiating Condition:
Release of gaseous or liquid radioactivity
> 2 times the ODCM limits for 60 minutes or longer.EAL: RU1.1 Unusual Event Reading on any Table R-1 effluent radiation monitor > column "UP" for > 60 min.(Notes 1,2, 3)Note 1: Note 2: Note 3: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert I UE 2 X Hi-Rad Main Plant Vent RM-A3 (gas) N/A 280,000 cpm 28,000 cpm alarm Exhaust RM-A13 14 mR/hr N/A N/A N/A 0 2 X Hi-Rad 0 RM-A4 (gas) N/A N/A N/APurge exhaust alarm RM-A14 740 mR/hr 74 mR/hr 7.4 mR/hr N/A Main Steam Line RM-G19 535 mR/hr 53.5 mR/hr 5.4 mR/hr N/A (Note 4) A/B/C Liquid Waste and 2 X Hi-Rad Nuclear Blowdown RM-L-9 N/A N/A N/A-J Discharge alarm Mode Applicability:
All Page 36 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Definition(s):
None Basis: Plant-Specific Liquid Releases The column "UE" liquid release value in Table R-1 is two times the alarm setpoint of RM-L9. The setpoint is established to ensure the ODCM release limits are not exceeded (ref.1, 5, 6, 7). RM-L9, Liquid Waste and Nuclear Blowdown Discharge, monitors the radioactivity of the combined effluent flow path from the Liquid Waste Processing and Storage System and from the Nuclear Blowdown Processing System. This monitor provides effluent measurement for radioactivity before the discharge goes to the penstocks.
High alarm interlock closes the liquid waste discharge valve XVD-691 0-LW.The liquid waste effluent line is a batch type release point. The monitor setpoint is set considering the tank concentration and isotopic content. Also, the available dilution is considered such that the concentration in the uncontrolled effluent does not exceed the limits of the ODCM and 10CFR20. Once isolated, control of radioactivity release through this path is reestablished and the EAL threshold for this monitor no longer applies. Fluid monitored by RM-L5 and by RM-L7 is monitored by RM-L9 prior to discharge from the plant. (ref. 2, 3, 4)Gaseous Releases The column "UE" gaseous release values in Table R-1 represent two times the alarm setpoint of the specified monitors.
The setpoints are established to ensure the ODCM release limits are not exceeded. (ref. 1, 6, 7, 8)RM-A3/RM-A1 3 -Main Plant Vent Exhaust: The main plant vent exhaust has the potential for release of radioactive particulate, iodine and noble gases and RM-A3 is designed to monitor these releases.
The sensitivity of RM-A3 provides for adequate measurement of approach to the maximum release allowable by 1 OCFR20 limits for unrestricted area. The setpoint of the monitor is established in accordance with the Offsite Dose Calculation Page 37 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Manual. It must be called to the attention of plant operators that the response of RM-A3 to low concentrations of particulate and iodine is seriously impaired by a simultaneous large amount of noble gases which will tend to make the particulate and iodine channel read higher than normal. Under these conditions laboratory measurements of filter media are required to quantify the release. RM-A1 3 provides extended range backup to channel RM-A3 for indication of high level noble gas discharge up tol0 5 pCi/cc as referenced to XE-133. (ref. 2, 3, 4)RM-A4/RM-A14
-RB Purge Exhaust: The Reactor Building is a potential release path during building venting preliminary to refueling operations.
RM-A4 and RM-A14 are monitors similar to RM-A3 and RM-A1 3. The 36" purge is not used during Modes 1-4. The 6" purge is allowed in all modes for pressure control and to clean up the air in the RB.However leakage past either of these pathways due to a valve failure or high RB pressure would be indicated by RM-A4 and RM-A14 if it was aligned to that particular pathway. The HI-RAD interlock (Gas Channel) stops the 36 in. and 6 in. Reactor Building purge. Once the purge is isolated, control of radioactivity release through this path is reestablished and the EAL threshold no longer applies to this monitor. RM-A1 4 provides extended range backup to channel RM-A4 for indication of high level noble gas discharge up to10 5 pCi/cc as referenced to XE-133. (ref. 2, 3, 4, 5)Generic This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release).
It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment.
Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases.
The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.Page 38 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
Classification based on effluent monitor readings assumes that a release path to the environment is established.
If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.Releases should not be prorated or averaged.
For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.This EAL addresses normally occurring continuous as well as planned batch radioactivity releases from monitored gaseous or liquid effluent pathways.Escalation of the ECL would be via IC RAI.VCSNS Basis Reference(s):
- 1. Offsite Dose Calculation Manual 2. Design Bases Document -Radiation Monitoring System (RM)3. SOP-1 24 Process and Area Radiation Monitoring System 4. HPP-904 Use of the Radiation Monitoring System (RMS)5. EPP-3 Plant Radiological Surveys 6. TWR 11.0/6.2-07-010, "Effluent Monitor Calculations for New EAL's" 7. TWR 11.0-07-011, "Dose-Based Effluent Response Thresholds (EAL's) Using MIDAS for Alert, Site Area Emergency, and General Emergency Classifications
- 8. TWR 11.0/6.2-07-013, "RM-L1 Calculations for New EAL's" 9. NEI 99-01 AU1 Page 39 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
R -Abnormal Rad Levels / Radiological Effluent Subcategory:
1 -Radiological Effluent Initiating Condition:
Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.EAL: RU1.2 Unusual Event Sample analysis for a gaseous or liquid release indicates a concentration or release rate> 2 x ODCM limits for -> 60 min. (Notes 1, 2)Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Mode Applicability:
All Definition(s):
None Basis: Site Specific None Generic This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release).
It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment.
Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional Page 40 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]releases.
The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
Releases should not be prorated or averaged.
For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.).Escalation of the ECL would be via IC RAI.VCSNS Basis Reference(s):
- 1. VCSNS Off-Site Dose Calculation Manual (ODCM)2. NEI 99-01 AU1 Page 41 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category: Subcategory:
Initiating Condition:
R -Abnormal Rad Levels / Radiological Effluent 1 -Radiological Effluent Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL: RA1.1 Alert Reading on any Table R-1 effluent radiation monitor > column "ALERT" for > 15 min.(Notes 1,2, 3, 4, 5)Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4: During a tube rupture with reactor at power RM-G19A/B/C monitor readings are affected by 1 6 N therefore they are not reliable until reactor has tripped and the monitors stable.Note 5 The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor [ GE [ SAE [ Alert I UE 2 X Hi-Rad Main Plant Vent RM-A3 (gas) N/A 280,000 cpm 28,000 cpm alarm Exhaust RM-A13 14 mR/hr N/A N/A N/A 0 2 X Hi-Rad h RM-A4 (gas) N/A N/A N/A 8 RB Purge exhaust alarm RM-A14 740 mR/hr 74 mR/hr 7.4 mR/hr N/A Main Steam Line RM-G19 535 mR/hr 53.5 mR/hr 5.4 mR/hr N/A (Note 4) A/B/C Liquid Waste and 2 X Hi-Rad"- Nuclear Blowdown RM-L-9 N/A N/A N/A alaRm" Discharge alarm Mode Applicability:
All Page 42 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Definition(s):
None Basis: Plant-Specific The column "ALERT" gaseous release values in Table R-1 represent offsite dose resulting from an actual or imminent release of gaseous radioactivity that exceeds 10 mRem TEDE or 50 mRem thyroid CDE (1% of the EPA PAGs). The emergency dose assessment methodology/model was used to calculate monitor readings at or beyond the SITE BOUNDARY that would yield the limiting EPA PAG dose assuming (ref. 4, 6, 7):* Design basis RCS source term* Annual average meteorology (wind speed and stability)
- Default release duration* Most limiting wind direction (highest 0/0)RM-A3/RM-A1 3 -Main Plant Vent Exhaust: The main plant vent exhaust has the potential for release of radioactive particulate, iodine and noble gases and RM-A3 is designed to monitor these releases.
The sensitivity of RM-A3 provides for adequate measurement of approach to the maximum release allowable by 1 OCFR20 limits for unrestricted area. It must be called to the attention of plant operators that the response of RM-A3 to low concentrations of particulate and iodine is seriously impaired by a simultaneous large amount of noble gases which will tend to make the particulate and iodine channel read higher than normal. Under these conditions laboratory measurements of filter media are required to quantify the release. RM-A1 3 provides extended range backup to channel RM-A3 for indication of high level noble gas discharge up tol 05 pCi/cc as referenced to XE-133. (ref. 1, 2, 3)RM-A4/RM-A1 4 -RX BLDG Purge Exhaust: The Reactor Building is a potential release path during building venting preliminary to refueling operations.
RM-A4 and RM-A14 are monitors similar to RM-A3 and RM-A1 3. The 36" purge is not used during Modes 1-4. The Page 43 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]6" purge is allowed in all modes for pressure control and to clean up the air in the RB.However leakage past either of these pathways due to a valve failure or high RB pressure would be indicated by RM-A4 and RM-A14 if it was aligned to that particular pathway. The rationale for selection of range, sensitivity and setpoint is the same as for the plant vent exhaust. RM-A14 provides extended range backup to channel RM-A4 for indication of high level noble gas discharge up tol0 5 pCi/cc as referenced to XE-133. The HI-RAD interlock (Gas Channel) stops the 36 in. and 6 in. Reactor Building purge. Once the purge is isolated, control of radioactivity release through this path is reestablished and the EAL threshold no longer applies to this monitor. (ref. 1, 2, 3, 5)RM-G19A/B/C
-Main Steam Line: Each MSL header, upstream of the relief valves, is provided with a high range gamma sensitive monitor to provide indication of the steam activity.
These monitors measure the dose rate of the steam lines and are used to monitor a potential release of radioactivity through the main steam relief valves. They are also used to provide information to the control room operators of a primary-to-secondary leak via a tube rupture and to provide for the determination of which steam generator is affected. (ref. 1, 2, 3) During a tube rupture with reactor at power RM-G19A/B/C monitor readings are affected by N16 gamma therefore they are not reliable until reactor has tripped and the monitors have stabilized.
Liquid effluents are classified under EAL RA1.3.Generic This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1 % of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases.
Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully Page 44 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to the environment is established.
If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.Escalation of the ECL would be via IC RS1.VCSNS Basis Reference(s):
- 1. Design Bases Document -Radiation Monitoring System (RM)2. SOP-1 24 Process and Area Radiation Monitoring System 3. HPP-904 Use of the Radiation Monitoring System (RMS)4. TWR 11.0/6.2-07-010, "Effluent Monitor Calculations for New EAL's" 5. EPP-3 Plant Radiological Surveys 6. TWR 11.0-07-011, "Dose-Based Effluent Response Thresholds (EAL's) Using MIDAS for Alert, Site Area Emergency, and General Emergency Classifications
- 7. TWR 11.0/6.2-07-013, "RM-L1 Calculations for New EAL's" 8. NEI 99-01 AA1 Page 45 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
R -Abnormal Rad Levels / Radiological Effluent Subcategory:
1 -Radiological Effluent Initiating Condition:
Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL: RA1.2 Alert Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY (Notes 3, 4, 5)Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4: During a tube rupture with reactor at power RM-G19A/B/C monitor readings are affected by 1 6 N therefore they are not reliable until reactor has tripped and the monitors stable.Note 5: The pre-calculated effluent monitor values presented in EALs RAI.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Mode Applicability:
All Definition(s):
SITE BOUNDARY -All land and water areas inside the one mile radius of the Reactor Building, use of which must be authorized by SCE&G Basis: Plant-Specific Dose assessment may be performed by either manual computer based methods (ref. 1).Generic This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1 % of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases.
Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).Page 46 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at 1 % of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to the environment is established.
If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.Escalation of the ECL would be via IC RS1.VCSNS Reference(s):
- 1. EPP-005 Offsite Dose Calculations
- 2. NEI 99-01 AA1 Page 47 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
R -Abnormal Rad Levels / Radiological Effluent Subcategory:
1 -Radiological Effluent Initiating Condition:
Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL: RA1.3 Alert Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: " Closed window dose rates > 10 mR/hr expected to continue for _> 60 min.* Analyses of field survey samples indicate thyroid CDE > 50 mrem for 60 min. of inhalation.(Notes 1,2)Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Mode Applicability:
All Definition(s):
SITE BOUNDARY-All land and water areas inside the one mile radius of the Reactor Building, use of which must be authorized by SCE&G.Basis: Plant-Specific EPP-007, Environmental Monitoring, provides guidance for emergency or post-accident radiological environmental monitoring (ref. 1).Generic This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases.
Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of Page 48 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Escalation of the ECL would be via IC RS1.VCSNS Reference(s):
- 1. EPP-007 Environmental Monitoring
- 2. NEI 99-01 AA1 Page 49 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
R -Abnormal Rad Levels / Radiological Effluent Subcategory:
1 -Radiological Effluent Initiating Condition:
Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL: RS1.1 Site Area Emergency Reading on any Table R-1 effluent radiation monitor > column "SAE" for > 15 min.(Notes 1,2, 3, 4, 5)Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4: During a tube rupture with reactor at power RM-G19A/B/C monitor readings are affected by 16N therefore they are not reliable until reactor has tripped and the monitors stable.Note 5: The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Table R-1 Effluent Monitor Classification Thresholds Release Point [ Monitor [ GE ] SAE f Alert ] UE 2 X Hi-Rad Main Plant Vent RM-A3 (gas) N/A 280,000 cpm 28,000 cpm alarm Exhaust RM-A13 14 mR/hr N/A N/A N/A U)0 2 X Hi-Rad RM-A4 (gas) N/A N/A N/A RB Purge exhaust alarm RM-A14 740 mR/hr 74 mR/hr 7.4 mR/hr N/A Main Steam Line RM-G19 535 mR/hr 53.5 mR/hr 5.4 mR/hr N/A (Note 4) A/B/C 2 Liquid Waste and 2 X Hi-Rad Nuclear Blowdown RM-L-9 N/A N/A N/A" Discharge Mode Applicability:
All Definition(s):
None Page 50 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Basis: Plant-Specific The column "SAE" gaseous release values in Table R-1 represent offsite dose resulting from an actual or imminent release of gaseous radioactivity that exceeds 100 mRem TEDE or 500 mRem thyroid CDE (10% of the EPA PAGs). The emergency dose assessment methodology/model was used to calculate monitor readings at or beyond the SITE BOUNDARY that would yield the limiting EPA PAG dose assuming (ref. 5, 6, 7): " Design basis RCS source term" Annual average meteorology (wind speed and stability)" Default release duration" Most limiting wind direction (highest O/Q)RM-A3/RM-A1 3 -Main Plant Vent Exhaust: The main plant vent exhaust has the potential for release of radioactive particulate, iodine and noble gases and RM-A3 is designed to monitor these releases.
The sensitivity of RM-A3 provides for adequate measurement of approach to the maximum release allowable by 1 OCFR20 limits for unrestricted area. It must be called to the attention of plant operators that the response of RM-A3 to low concentrations of particulate and iodine is seriously impaired by a simultaneous large amount of noble gases which will tend to make the particulate and iodine channel read higher than normal. Under these conditions laboratory measurements of filter media are required to quantify the release. RM-A13 provides extended range backup to channel RM-A3 for indication of high level noble gas discharge up to10 5 pCi/cc as referenced to XE-133. (ref. 1,2, 3)RM-A4/RM-A14
-RX BLDG Purge Exhaust: The Reactor Building is a potential release path during building venting preliminary to refueling operations.
RM-A4 and RM-A14 are monitors similar to RM-A3 and RM-A1 3. The 36" purge is not used during Modes 1-4. The 6" purge is allowed in all modes for pressure control and to clean up the air in the RB.However leakage past either of these pathways due to a valve failure or high RB pressure would be indicated by RM-A4 and RM-A14 if it was aligned to that particular pathway. The Page 51 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]rationale for selection of range, sensitivity and setpoint is the same as for the plant vent exhaust. RM-A14 provides extended range backup to channel RM-A4 for indication of high level noble gas discharge up tol0 5 pCi/cc as referenced to XE-133. The HI-RAD interlock (Gas Channel) stops the 36 in. and 6 in. Reactor Building purge. Once the purge is isolated, control of radioactivity release through this path is reestablished and the EAL threshold no longer applies to this monitor. (ref. 1, 2, 3, 4)RM-G19A/B/C
-Main Steam Line: Each MSL header, upstream of the relief valves, is provided with a high range gamma sensitive monitor to provide indication of the steam activity.
These monitors measure the dose rate of the steam lines and are used to monitor a potential release of radioactivity through the main steam relief valves. They are also used to provide information to the control room operators of a primary-to-secondary leak via a tube rupture and to provide for the determination of which steam generator is affected. (ref. 1, 2, 3) During a tube rupture with reactor at power RM-G1 9A/B/C monitor readings are affected by N16 gamma therefore they are not reliable until reactor has tripped and the monitors have stabilized.
Generic This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases.
Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to the environment is established.
If the effluent flow past an effluent monitor is known to have Page 52 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.Escalation of the ECL would be via IC RG1.VCSNS Basis Reference(s):
- 1. Design Bases Document -Radiation Monitoring System (RM)2. SOP-124 Process and Area Radiation Monitoring System 3. HPP-904 Use of the Radiation Monitoring System (RMS)4. EPP-3 Plant Radiological Surveys 5. TWR 11.0/6.2-07-010, "Effluent Monitor Calculations for New EAL's" 6. TWR 11.0-07-011, "Dose-Based Effluent Response Thresholds (EAL's) Using MIDAS for Alert, Site Area Emergency, and General Emergency Classifications
- 7. TWR 11.0/6.2-07-013, "RM-L1 Calculations for New EAL's" 8. NEI 99-01 AS1 Page 53 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category: Subcategory:
Initiating Condition:
R -Abnormal Rad Levels / Radiological Effluent 1 -Radiological Effluent Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL: RS1.2 Site Area Emergency Dose assessment using actual meteorology indicates doses > 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the SITE BOUNDARY (Notes 3, 4, 5)Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4: During a tube rupture with reactor at power RM-G19A/B/C monitor readings are affected by 1 6 N therefore they are not reliable until reactor has tripped and the monitors stable.Note 5: The pre-calculated effluent monitor values presented in EALs RAI.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Mode Applicability:
All Definition(s):
SITE BOUNDARY-All land and water areas inside the one mile radius of the Reactor Building, use of which must be authorized by SCE&G.Basis: Plant-Specific Dose assessment may be performed by either manual or computer based methods (ref. 1).Generic This IC addresses a release of gaseous radioactivity that results in projected or actual off site doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases.
Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.Radiological effluent EALs are also included to provide a basis for classifying events and Page 54 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to the environment is established.
If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.Escalation of the ECL would be via IC RG1.VCSNS Reference(s):
- 1. EPP-005 Offsite Dose Calculations
- 2. NEI 99-01 AS1 Page 55 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
R -Abnormal Rad Levels / Radiological Effluent Subcategory:
1 -Radiological Effluent Initiating Condition:
Offsite dose resulting from an actual or imminent release of gaseous radioactivity exceeds 100 mRem TEDE or 500 mRem thyroid CDE for the actual or projected duration of the release using actual meteorology EAL: RS11.3 Site Area Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: " Closed window dose rates > 100 mR/hr expected to continue for > 60 min." Analyses of field survey samples indicate thyroid CDE > 500 mrem for 60 min. of inhalation.(Notes 1,2)Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Mode Applicability:
All Definition(s):
SITE BOUNDARY-All land and water areas inside the one mile radius of the Reactor Building, use of which must be authorized by SCE&G.Basis: Plant-Specific EPP-007, Environmental Monitoring, provides guidance for emergency or post-accident radiological environmental monitoring (ref. 1).Generic This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases.
Releases of this magnitude are Page 56 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]associated with the failure of plant systems needed for the protection of the public.Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Escalation of the ECL would be via IC RG1.VCSNS Reference(s):
- 1. EPP-007 Environmental Monitoring
- 2. NEI 99-01 AS1 Page 57 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
R -Abnormal Rad Levels / Radiological Effluent Subcategory:
1 -Radiological Effluent Initiating Condition:
Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL: RG1.1 General Emergency Reading on any Table R-1 effluent radiation monitor > column "GE" for -> 15 min.(Notes 1,2, 3, 4, 5)Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4: During a tube rupture with reactor at power RM-G19A/B/C monitor readings are affected by 16N therefore they are not reliable until reactor has tripped and the monitors stable.Note 5: The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Table R-1 Effluent Monitor Classification Thresholds Release Point ]Monitor I GE I SAE I Alert I UE 2 X Hi-Rad Main Plant Vent RM-A3 (gas) N/A 280,000 cpm 28,000 cpm alarm Exhaust RM-A13 14 mR/hr N/A N/A N/A 0 2 X Hi-Rad A RM-A4 (gas) N/A N/A N/A RB Purge exhaust alarm RM-A14 740 mR/hr 74 mR/hr 7.4 mR/hr N/A Main Steam Line RM-G19 535 mR/hr 53.5 mR/hr 5.4 mR/hr N/A (Note 4) NB/C 2 Liquid Waste and 2 X Hi-Rad"- Nuclear Blowdown RM-L-9 N/A N/A N/A alarm"3 Discharge alarm Mode Applicability:
All Definition(s):
None Basis: Page 58 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Plant-Specific The column "GE" gaseous release values in Table R-1 represent offsite dose resulting from an actual or imminent release of gaseous radioactivity that exceeds 1000 mRem TEDE or 5000 mRem thyroid CDE (100% of the EPA PAGs). The emergency dose assessment methodology/model was used to calculate monitor readings at or beyond the SITE BOUNDARY that would yield the limiting EPA PAG dose assuming (ref. 5, 6, 7):* Design basis RCS source term* Annual average meteorology (wind speed and stability)
- Default release duration" Most limiting wind direction (highest 0/0)RM-A3/RM-A1 3 -Main Plant Vent Exhaust: The main plant vent exhaust has the potential for release of radioactive particulate, iodine and noble gases and RM-A3 is designed to monitor these releases.
The sensitivity of RM-A3 provides for adequate measurement of approach to the maximum release allowable by 10CFR20 limits for unrestricted area. It must be called to the attention of plant operators that the response of RM-A3 to low concentrations of particulate and iodine is seriously impaired by a simultaneous large amount of noble gases which will tend to make the particulate and iodine channel read higher than normal. Under these conditions laboratory measurements of filter media are required to quantify the release. RM-A1 3 provides extended range backup to channel RM-A3 for indication of high level noble gas discharge up to10 5 pCi/cc as referenced to XE-133. (ref. 1, 2, 3)RM-A4/RM-A14
-RX BLDG Purge Exhaust: The Reactor Building is a potential release path during building venting preliminary to refueling operations.
RM-A4 and RM-A14 are monitors similar to RM-A3 and RM-A1 3. The 36" purge is not used during Modes 1-4. The 6" purge is allowed in all modes for pressure control and to clean up the air in the RB.However leakage past either of these pathways due to a valve failure or high RB pressure would be indicated by RM-A4 and RM-A14 if it was aligned to that particular pathway. The rationale for selection of range, sensitivity and setpoint is the same as for the plant vent Page 59 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]exhaust. Release through this path of gaseous radioactivity that exceeds the EPA PAGs would cause an off-scale high reading on RM-A4. The Table R-1 entry for this monitor is therefore "N/A" at the General Emergency classification level. RM-A14 provides extended range backup to channel RM-A4 for indication of high level noble gas discharge up tol 05 pCi/cc as referenced to XE-133. The HI-RAD interlock (Gas Channel) stops the 36 in. and 6 in. Reactor Building purge. Once the purge is isolated, control of radioactivity release through this path is reestablished and the EAL threshold no longer applies to this monitor.(ref. 1,2, 3, 4)RM-G19A/B/C
-Main Steam Line: Each MSL header, upstream of the relief valves, is provided with a high range gamma sensitive monitor to provide indication of the steam activity.
These monitors measure the dose rate of the steam lines and are used to monitor a potential release of radioactivity through the main steam relief valves. They are also used to provide information to the control room operators of a primary-to-secondary leak via a tube rupture and to provide for the determination of which steam generator is affected. (ref. 1,2, 3) During a tube rupture with reactor at power RM-G19A/B/C monitor readings are affected by N16 gamma therefore they are not reliable until reactor has tripped and the monitors have stabilized.
Generic This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases.
Releases of this magnitude will require implementation of protective actions for the public.Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Page 60 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Classification based on effluent monitor readings assumes that a release path to the environment is established.
If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.VCSNS Basis Reference(s):
- 1. Design Bases Document -Radiation Monitoring System (RM)2. SOP-124 Process and Area Radiation Monitoring System 3. HPP-904 Use of the Radiation Monitoring System (RMS)4. EPP-3 Plant Radiological Surveys 5. TWR 11.0/6.2-07-010, "Effluent Monitor Calculations for New EAL's" 6. TWR 11.0-07-011, "Dose-Based Effluent Response Thresholds (EAL's) Using MIDAS for Alert, Site Area Emergency, and General Emergency Classifications
- 7. TWR 11.0/6.2-07-013, "RM-L1 Calculations for New EAL's" 8. NEI 99-01 AG1 Page 61 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
R -Abnormal Rad Levels / Radiological Effluent Subcategory:
1 -Radiological Effluent Initiating Condition:
Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL: RG1.2 General Emergency Dose assessment using actual meteorology indicates doses > 1000 mrem TEDE or 5000 mrem thyroid CDE at or beyond the SITE BOUNDARY (Notes 3, 4, 5)Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4: During a tube rupture with reactor at power RM-G19A/B/C monitor readings are affected by 16N therefore they are not reliable until reactor has tripped and the monitors stable.Note 5: The pre-calculated effluent monitor values presented in EALs RAI.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Mode Applicability:
All Definition(s):
SITE BOUNDARY -All land and water areas inside the one mile radius of the Reactor Building, use of which must be authorized by SCE&G.Basis: Plant-Specific Dose assessment may be performed by either manual or computer based methods (ref. 1).Generic This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases.
Releases of this magnitude will require implementation of protective actions for the public.Radiological effluent EALs are also included to provide a basis for classifying events and Page 62 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to the environment is established.
If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.VCSNS Reference(s):
- 1. EPP-005 Offsite Dose Calculations
- 2. NEI 99-01 AG1 Page 63 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT El Category: Subcategory:
Initiating Condition:
R -Abnormal Rad Levels / Radiological Effluent 1 -Radiological Effluent Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL: RG1.3 General Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: " Closed window dose rates > 1000 mR/hr expected to continue for -- 60 min." Analyses of field survey samples indicate thyroid CDE > 5000 mrem for 60 min. of inhalation.(Notes 1,2)Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Mode Applicability:
All Definition(s):
SITE BOUNDARY-All land and water areas inside the one mile radius of the Reactor Building, use of which must be authorized by SCE&G.Basis: Plant-Specific EPP-007, Environmental Monitoring, provides guidance for emergency or post-accident radiological environmental monitoring (ref. 1).Generic This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases.
Releases of this magnitude will require implementation of protective actions for the public.Page 64 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.VCSNS Reference(s):
- 1. EPP-007 Environmental Monitoring
- 2. NEI 99-01 AG1 Page 65 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
R -Abnormal Rad Levels / Rad Effluent Subcategory:
2 -Irradiated Fuel Event Initiating Condition:
Unplanned loss of water level above irradiated fuel EAL: RU2.1 Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by any of the following:
- Refueling Cavity: LI-7403 MCB annunciator XCP-609 2-6 (REFUEL CAV LVL HI/LO)" Spent Fuel Pool: LI-7431 and LI-7433 MCB annunciators XCP 608(609) 1-2 (SFP LVL HI/LO)" Fuel Transfer Canal: LI-7405 MCB annunciator XCP-612 1-6 (FUEL XFER CANAL LVL HI/LO)AND UNPLANNED rise in area radiation levels as indicated by any of the following radiation monitors: " RM-G6 Rx Bldg Refueling Bridge* RM-G17A/B Rx Bldg Manipulator Crane" RM-G8 FHB Refueling Bridge Area Gamma Mode Applicability:
All Definition(s):
UNPLANNED-.
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown.REFUELING PATHWAY-.
The reactor refueling cavity, spent fuel pool and fuel transfer canal comprise the refueling pathway.Basis: Plant-Specific Indications of decreasing level include (ref. 1, 2):* Refueling Cavity: Page 66 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]LI-7403 is equipped with alarms at 461'-7" and 461'-3.5", MCB annunciator XCP-609 2-6 (REFUEL CAV LVL HI/LO)" Spent Fuel Pool: LI-7431 and LI-7433 have a range of 16 ft and cover a level elevation from 445'-4" to 461'-4". Thus providing adequate coverage of the normal operating level which is kept above the elevation of the skimmer piping at elevation 460'-3". Alarms are provided at 461'-0" and 460'-4", MCB annunciators SFP LVL HI/LO (XCP 608(609) 1-2)." Fuel Transfer Canal: LI-7405 has a range of 27 ft corresponding to an elevation of 434'-9" to 461'-9".Alarms are provided at 461'-3" and 460'-3", MCB annunciator XCP-612 1-6 (FUEL XFER CANAL LVL HI/LO)Allowing level to decrease could result in spent fuel being uncovered, reducing spent fuel decay heat removal and creating an extremely hazardous radiation environment.
Technical Specification LCO 3/4.7.10 requires at least 23 ft of water above the Spent Fuel Pool storage racks. Technical Specification LCO 3/4.9.9 requires at least 23 ft of water above the reactor vessel flange in the refueling cavity during refueling operations.
This maintains sufficient water level in the fuel transfer canal, refueling cavity, and spent fuel pool to retain iodine fission product activity in the water in the event of a fuel handling accident. (ref. 3, 4)Radiation monitors that may indicate a loss of shielding above irradiated fuel include (ref.2, 5):* RM-G6 Rx Bldg Refueling Bridge RM-G17A/B
-Rx Bldg Manipulator Crane These monitors provide monitoring of radiation due to a dropped fuel assembly during refueling operations.
This results in a fuel cladding rupture with the release of the gap activity.
The noble gases are expected to bubble up to the surface of the pool where the monitors will provide detection and alarm of the condition.
Page 67 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Page 68 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]RM-G17A/B provide purge isolation in the event of a fuel drop cladding rupture and are only installed in Mode 6.RM-G8 -FHB Refueling Bridge Area Gamma: This monitor provides a similar function as the monitors located on the Reactor Building bridge.Generic This IC addresses a decrease in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant.A water level decrease will be primarily determined by indications from available level instrumentation.
Other sources of level indications may include reports from plant personnel (e.g., from a refueling crew) or video camera observations (if available).
A significant drop in the water level may also cause an increase in the radiation levels of adjacent areas that can be detected by monitors in those locations.
The effects of planned evolutions should be considered.
For example, a refueling bridge area radiation monitor reading may increase due to planned evolutions such as lifting of the reactor vessel head or movement of a fuel assembly.
Note that this EAL is applicable only in cases where the elevated reading is due to an UNPLANNED loss of water level.A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes.Escalation of the ECL would be via IC RA2.VCSNS Reference(s):
- 1. Design Bases Document -Spent Fuel Cooling System (SF)2. AOP-123.1 Decreasing Level in the Spent Fuel Pool or Refueling Cavity During Refueling 3. Technical Specifications LCO 3/4.7.10 4. Technical Specifications LCO 3/4.9.9 5. Design Bases Document -Radiation Monitoring System (RM)6. NEI 99-01 AU2 Page 69 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
R -Abnormal Rad Levels / Rad Effluent Subcategory:
2 -Irradiated Fuel Event Initiating Condition:
Significant lowering of water level above, or damage to, irradiated fuel EAL: RA2.1 Alert Uncovery of irradiated fuel in the REFUELING PATHWAY Mode Applicability:
All Definition(s):
REFUELING PATHWAY-.
The reactor refueling cavity, spent fuel pool and fuel transfer canal comprise the refueling pathway.Basis: Plant-Specific Indications of decreasing water level with the potential to uncover irradiated fuel include (ref. 1,2):* Refueling Cavity: LI-7403 is equipped with alarms at 461'-7" and 461'-3.5", MCB annunciator XCP-609 2-6 (REFUEL CAV LVL HI/LO)" Spent Fuel Pool: LI-7431 and LI-7433 have a range of 16 ft and cover a level elevation from 445'-4" to 461'-4" thus providing adequate coverage of the normal operating level which is kept above the elevation of the skimmer piping at elevation 460'-3". Alarms are provided at 461'-0" and 460'-4", MCB annunciators SFP LVL HI/LO (XCP 608(609) 1-2).Page 70 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]* Fuel Transfer Canal: LI-7405 has a range of 27 ft corresponding to an elevation of 434'-9" to 461'-9".Alarms are provided at 461'-3" and 460'-3", MCB annunciator XCP-612 1-6 (FUEL XFER CANAL LVL HI/LO)Technical Specification LCO 3/4.7.10 requires at least 23 ft of water above the Spent Fuel Pool storage racks. Technical Specification LCO 3/4.9.9 requires at least 23 ft of water above the reactor vessel flange in the refueling cavity during refueling operations.
This maintains sufficient water level in the fuel transfer canal, refueling cavity, and spent fuel pool to retain iodine fission product activity in the water in the event of a fuel handling accident. (ref. 3, 4) Allowing level to decrease could result in spent fuel being uncovered, reducing spent fuel decay heat removal and creating an extremely hazardous radiation environment.
Plant procedures require termination of fuel and core component movements and evacuation of the Reactor Building and Fuel Handling Building if elevated radiation levels are detected.
All core alternations are stopped and transient fuel assemblies and core components are placed in a safe position in the reactor vessel, Spent Fuel Pool or fuel transfer cart to the extent practicable (ref. 2). Figure 1 illustrates the elevations (rounded)at which fuel assemblies could become uncovered when seated in the reactor vessel, Spent Fuel Pool, Reactor Building and Fuel handing Building upenders, and fuel transfer tube (ref. 5, 6).Generic This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment.
As such, they represent an actual or potential substantial degradation of the level of safety of the plant.EAL#1 Page 71 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]This EAL escalates from RU2.1 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters.
Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations.
While an area radiation monitor could detect an increase in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered.
To the degree possible, readings should be considered in combination with other available indications of inventory loss.A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes.Escalation of the ECL would be via IC RS1.VCSNS Reference(s):
- 1. Design Bases Document -Spent Fuel Cooling System (SF)2. AOP-123.1 Decreasing Level in the Spent Fuel Pool or Refueling Cavity During Refueling 3. Technical Specifications LCO 3/4.7.10 4. Technical Specifications LCO 3/4.9.9 5. Drawing SS-024-021
- 6. Drawing E-002-001 7. NEI 99-01 AA2 Page 72 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
R -Abnormal Rad Levels / Rad Effluent Subcategory:
2 -Irradiated Fuel Event Initiating Condition:
Significant lowering of water level above, or damage to, irradiated fuel EAL: RA2.2 Alert Damage to irradiated fuel resulting in a release of radioactivity as indicated by a Hi-Rad alarm on any of the following radiation monitors: " RM-G8 FHB Refueling Bridge Area Gamma" RM-A6 Fuel Handling Bldg Exhaust" RM-G6 Rx Bldg Refueling Bridge* RM-G17A/B Rx Bldg Manipulator Crane Mode Applicability:
All Definition(s):
None Basis: Plant-Specific When considering escalation, information may come from: " Radiation monitor readings" Sampling and surveys* Dose projections/calculations
- Reports from the scene regarding the extent of damage (e.g., refueling crew, radiation protection technicians)
Radiation monitors listed in this EAL are (ref. 1, 2):* RM-G8 -FHB Refueling Bridge Area Gamma: This monitor provides a similar function as the monitors located on the Reactor Building bridge.Page 73 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]" RM-A6 -Fuel Handling Bldg Exhaust: This monitor measures the particulate, iodine and noble gas activity in the exhaust duct leading to the stack and allows to determine the source of potential releases should the stack monitor indicate an unexpected plant release." RM-G6 Rx Bldg Refueling Bridge RM-G17A/B
-Rx Bldg Manipulator Crane These monitors provide monitoring of radiation due to a dropped fuel assembly during refueling operations.
This results in a fuel cladding rupture with the release of the gap activity.
The noble gases are expected to bubble up to the surface of the pool where the monitors will provide detection and alarm of the condition.
RM-G17A/B provide purge isolation in the event of a fuel drop cladding rupture. Both of these monitors are only installed in Mode 6.Plant procedures require termination of fuel and core component movements and evacuation of the Reactor Building and Fuel Handling Building if elevated radiation levels are detected.
All core alternations are stopped and transient fuel assemblies and core components are placed in a safe position in the reactor vessel, Spent Fuel Pool or fuel transfer cart to the extent practicable (ref. 1). Figure 1 illustrates the elevations (rounded)at which fuel assemblies could become uncovered when seated in the reactor vessel, Spent Fuel Pool, Reactor Building and Fuel handing Building upenders, and fuel transfer tube (ref. 3, 4).Generic This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment.
As such, they represent an actual or potential substantial degradation of the level of safety of the plant.Escalation of the emergency would be based on either Recognition Category R or C ICs.Page 74 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly.
A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident).
Escalation of the ECL would be via IC RS1.VCSNS Reference(s):
- 1. AOP-123.1 Decreasing Level in the Spent Fuel Pool or Refueling Cavity During Refueling 2 Design Bases Document -Radiation Monitoring System (RM)3. Drawing SS-024-021
- 4. Drawing E-002-001 5. NEI 99-01 AA2 Page 75 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
R -Abnormal Rad Levels / Rad Effluent Subcategory:
2- Irradiated Fuel Event Initiating Condition:
Significant lowering of water level above, or damage to, irradiated fuel EAL: RA2.3 Alert Lowering of spent fuel pool level to Level 2 (ele. 455' 6")Mode Applicability:
All Definition(s):
None Basis: Plant-Specific The VCS1 SFP is located in the western end of the fuel handling building.
The surface of the water is normally maintained at plant elevation 461.5 ft. by scuppers that act as skimmers.
This results in a minimum water depth of 24 ft. of water shielding over the stored spent fuel assemblies (ref. 1).Indications of SFP decreasing water level include LI-7431 and LI-7433. LI-7431 and LI-7433 and have a range of 16 ft. and cover a level elevation from 445'-4" to 461'-4" thus providing adequate coverage of the normal operating level which is kept above the elevation of the skimmer piping at elevation 460'-3". A low SFP level alarm is provided at 461'-0" (Level 1), MCB annunciators (XCP 608(609) 1-2 (SFP LVL HI/LO) (ref. 2, 3).Additionally, Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level that provides adequate shielding above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3). For VCS1, the SFP Level 2 setpoint is plant elevation 455' 6" or approximately 19' above the stored fuel assemblies (ref. 5).Generic Page 76 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment.
As such, they represent an actual or potential substantial degradation of the level of safety of the plant.Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool.Escalation of the emergency classification level would be via IC RS1.VCSNS Reference(s):
- 1. USFSAR Section 9.1.2.2 Facilities Description
- 2. Design Bases Document -Spent Fuel Cooling System (SF)3. ARP-001-XCP-608 1-2 (SFP LVL HI/LO)4. AOP-123.1 Decreasing Level in the Spent Fuel Pool or Refueling Cavity During Refueling 5. Letter RC-13-0119 from T. D. Gatlin to NRC 8/28/2013 Attachment 1 Virgil C. Summer Nuclear Station Unit 1 -Response to Request for Additional Information
-Overall Integrated Plan in Response to Commission Order Modifying License Requirements for Reliable Spent Fuel Pool Instrumentation (Order EA-12-051)
- 6. NEI 99-01 AA2 Page 77 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
R -Abnormal Rad Release/ Rad Effluent Subcategory:
2 -Irradiated Fuel Event Initiating Condition:
Spent fuel pool level at the top of the fuel racks EAL: RS2.1 SiteArea Emergency Lowering of spent fuel pool level to Level 3 (ele. 437' 0")Mode Applicability:
All Definition(s):
None Basis: Plant-Specific The VCS1 SFP is located in the western end of the fuel handling building.
The surface of the water is normally maintained at plant elevation 461.5 ft. by scuppers that act as skimmers.
This results in a minimum water depth of 24 ft. of water shielding over the stored spent fuel assemblies (ref. 1).Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level that provides adequate shielding above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3). For VCS1, the SFP Level 3 setpoint is plant elevation 437' 0" (ref. 2).Generic This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.
It is recognized that this IC would likely not be met until well after another Site Area Emergency IC was met; however, it is included to provide classification diversity.
Page 78 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Escalation of the emergency classification level would be via IC RG1 or RG2.VCSNS Reference(s):
- 1. USFSAR Section 9.1.2.2 Facilities Description
- 2. Letter RC-13-0119 from T. D. Gatlin to NRC 8/28/2013 Attachment 1 Virgil C. Summer Nuclear Station Unit 1 -Response to Request for Additional Information
-Overall Integrated Plan in Response to Commission Order Modifying License Requirements for Reliable Spent Fuel Pool Instrumentation (Order EA-12-051)
- 3. NEI 99-01 AS2 Page 79 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
R -Abnormal Rad Release / Rad Effluent Subcategory:
2 -Irradiated Fuel Event Initiating Condition:
Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer EAL: RG2.1 General Emergency Spent fuel pool level cannot be restored to at least Level 3 (ele. 437' 0") for > 60 min.(Note 1)Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability:
All Definition(s):
None Basis: Plant-Specific The VCS1 SFP is located in the western end of the fuel handling building.
The surface of the water is normally maintained at plant elevation 461.5 ft. by scuppers that act as skimmers.
This results in a minimum water depth of 24 ft. of water shielding over the stored spent fuel assemblies (ref. 1).Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level that provides adequate shielding above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3). For VCS1, the SFP Level 3 setpoint is plant elevation 437' 0" (ref. 2).Generic This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment.
Page 80 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]It is recognized that this IC would likely not be met until well after another General Emergency IC was met; however, it is included to provide classification diversity.
VCSNS Reference(s):
- 1. USFSAR Section 9.1.2.2 Facilities Description
- 2. Letter RC-13-0119 from T. D. Gatlin to NRRC 8/28/2013 Attachment 1 Virgil C. Summer Nuclear Station Unit 1 -Response to Request for Additional Information
-Overall Integrated Plan in Response to Commission Order Modifying License Requirements for Reliable Spent Fuel Pool Instrumentation (Order EA-12-051)
- 3. NEI 99-01 AG2 Page 81 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
R -Abnormal Rad Levels / Rad Effluent Subcategory:
3- Area Radiation Levels Initiating Condition:
Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.EAL: RA3.1 Alert Dose rate > 15 mR/hr in EITHER of the following areas: " Control Room (RM-G1)" Central Alarm Station (by survey)Mode Applicability:
All Definition(s):
None Basis: Plant-Specific RM-G1 is the permanently installed Control Room area radiation monitor and, along with local radiation surveys, may be used to assess this EAL threshold.
Permanently installed area radiation monitoring is not installed in the CAS and, therefore, radiation levels in this area must be assessed with local radiation survey techniques (ref. 1, 2, 3).Generic This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown.
As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should consider the cause of the increased radiation levels and determine if another IC may be applicable.
Escalation of the ECL would be via Recognition Category R, C or F ICs.VCSNS Basis Reference(s):
Page 82 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]1. Design Bases Document -Radiation Monitoring System (RM)2. SOP-124 Process and Area Radiation Monitoring System 3. HPP-904 Use of the Radiation Monitoring System (RMS)4. NEI 99-01 AA3 Page 83 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
R -Abnormal Rad Levels / Rad Effluent Subcategory:
3 -Area Radiation Levels Initiating Condition:
Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.EAL: RA3.2 Alert An UNPLANNED event results in radiation levels that prohibit or impede access to any Table R-2 area (Note 6)Note 6: If the equipment in the listed area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted.
Table R-2 Safe Operation
& Shutdown Areas Area Mode Applicability Auxiliary Building 374' 3 Auxiliary Building 388' 3, 4, 5 Auxiliary Building 400' 4, 5 Auxiliary Building 412 3, 4, 5 Auxiliary Building 436' 1,2, 3, 4, 5 Auxiliary Building 463' 3, 4, 5 Intermediate Building 412' 3 Intermediate Building 436' 4, 5 Intermediate Building 463' 3, 4, 5 Control Building 412' 2, 3 Control Building 436' 3, 4, 5 Turbine Building (All levels) 1, 2 Mode Applicability:
All Definition(s):
UNPLANNED-.
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown.Page 84 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Basis: Plant-Specific The Table R-2 safe operation and shutdown areas (with entry-related mode applicability) are those plant areas that contain equipment which require a manual/local action as specified in general operating procedures (and procedures referenced by them) used for normal plant operation, cooldown and shutdown.
The list specifies the plant operating modes during which entry would be required for each area and thus specifying when a loss of access or impeded access is applicable to this EAL (ref. 1).Plant areas where actions of a contingent or emergency nature might be needed to be performed. (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) were not considered for inclusion.
Additionally, areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections) were not considered for inclusion.
Refer to Attachment 4 "Safe Operation
& Shutdown Areas Tables R-2 & H-3 Bases." If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.Generic This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown.
As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should consider the cause of the increased radiation levels and determine if another IC may be applicable.
For RA3.2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the Page 85 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]elevated radiation levels. The emergency classification is not contingent upon whether entry is actually necessary at the time of the increased radiation levels. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., installing temporary shielding, requiring use of non-routine protective equipment, requesting an extension in dose limits beyond normal administrative limits).An emergency declaration is not warranted if any of the following conditions apply." The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example, the plant is in Mode 1 when the radiation increase occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4." The increased radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., radiography, spent filter or resin transfer, etc.)." The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections)." The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action.* If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.Escalation of the ECL would be via Recognition Category R, C or F ICs.VCSNS Basis Reference(s):
- 1. EPP-108 Emergency Action Level Technical Bases Attachment 4 "Safe Operation
&Shutdown Areas Tables R-2 & H-3 Bases." 2. NEI 99-01 AA3 Page 86 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Categorv C -Cold Shutdown / Refuelinq System Malfunction EAL Group: Cold Conditions (RCS temperature
< 200'F);EALs in this category are applicable only in one or more cold operating modes.Category C EALs are directly associated with cold shutdown or refueling system safety functions.
Given the variability of plant configurations (e.g., systems out-of-service for maintenance, containment open, reduced AC power redundancy, time since shutdown)during these periods, the consequences of any given initiating event can vary greatly. For example, a loss of decay heat removal capability that occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown.
Compounding these events is the likelihood that instrumentation necessary for assessment may also be inoperable.
The cold shutdown and refueling system malfunction EALs are based on performance capability to the extent possible with consideration given to RCS integrity, containment closure, and fuel clad integrity for the applicable operating modes (5 -Cold Shutdown, 6 -Refueling, D -Defueled).
The events of this category pertain to the following subcategories:
- 1. RCS Level reactor vessel or RCS water level is directly related to the status of adequate core cooling and, therefore, fuel clad integrity.
- 2. Loss of Engineered Safeguards Features (ESF) AC Power Loss of ESF plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity.
This category includes loss of onsite and offsite power sources for 7.2 KV safeguards buses 1 DA and 1 DB.3. RCS Temperature Uncontrolled or inadvertent temperature or pressure increases are indicative of a potential loss of safety functions.
- 4. Loss of Vital DC Power Page 87 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Loss of emergency plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity.
This category includes loss of power to or degraded voltage on the 125VDC safeguards buses.5. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.
- 6. Hazardous Event Effecting Safety Systems Certain hazardous natural and technological events may result in VISIBLE DAMAGE to or degraded performance of SAFETY SYSTEMS warranting classification.
Page 88 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT El Category:
C -Cold Shutdown / Refueling System Malfunction Subcategory:
1 -RCS Level Initiating Condition:
UNPLANNED loss of reactor vessel/RCS inventory for 15 minutes or longer.EAL: CU1.1 Unusual Event UNPLANNED loss of reactor coolant results in reactor vessel/RCS level less than a required lower limit for -- 15 min. (Note 1)Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability:
5 -Cold Shutdown, 6 -Refueling Definition(s):
UNPLANNED-.
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown.Basis: Plant-Specific With the plant in Cold Shutdown, RCS water level is normally maintained above the pressurizer low level setpoint of 17%. When pressurizer level drops to 17%, letdown isolates and pressurizer heaters are deenergized.
This condition is signaled by MCB annunciator XCP-616 1-3 (BLCK HTRS ISOL LTDN PZR LCS LO) (ref. 1, 2, 3).Pressurizer level is indicated on LI-459A, LI-460, LI-461 and LR-459 on MCB XCP-6109 (ref. 2). However, if pressurizer level is being controlled below 17%, or if level is being maintained in a designated band in the reactor vessel it is the inability to maintain level above the low end of the designated control band due to a loss of inventory resulting from a leak in the RCS that is the concern.With the plant in Refueling mode, RCS water level is normally maintained above the reactor vessel flange. The reactor vessel flange mating surface is at 437'-7.43" elevation Page 89 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]rounded to 437'-7" (ref. 5). RCS elevations are illustrated in Figures 2 and 3 (ref. 5, 9).RCS level can be monitored by one or more of the following (ref. 6, 7): " LI-462, COLD CAL LEVEL % (ref. 8)* Control Room tygon hose TV monitor and RB camera* Mid Loop Monitoring System* Mansell Level Monitoring System (MLMS)" RVLIS Upper Plenum reading of 84.3% corresponds to the reactor vessel flange mating surface (ref. 5, 6, 9)Regardless of where RCS level is intentionally being controlled, either above or below the reactor vessel flange, as in Cold Shutdown, it is the inability to maintain level above the low end of the designated control band due to a loss of inventory resulting from a leak in the RCS that is the concern.Generic This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor reactor vessel/RCS level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.Refueling evolutions that decrease RCS water inventory are carefully planned and controlled.
An UNPLANNED event that results in water level decreasing below a procedurally required limit concurrent with indications of coolant leakage warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered.This EAL recognizes that the minimum required reactor vessel/RCS level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented.
This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum Page 90 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]level is typically specified in the applicable operating procedure but may be specified in another controlling document.The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level.Continued loss of RCS inventory may result in escalation to the Alert ECL via either IC CA1 or CA3.VCSNS Basis Reference(s):
- 1. Setpoint Bases (SB)2. OAP-103.2 Emergency Operating Procedure Setpoint Document 3. ARP-001-XCP-616 Panel XCP-616 4. 201-325 Control Panel XCP-6109 5. GOP-9 Mid-Loop Operation 6. SOP-1 01 Reactor Coolant System 7. SOP-1 15 Residual Heat Removal 8. 201-325 Main Control Board Instrumentation Control Panel XCP-6109 9. GOP-7 Core Refueling (Mode 5 to Mode 6, Defuel, and Refuel to Mode 6)10. NEI 99-01 CU1 Page 91 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
C -Cold Shutdown / Refueling System Malfunction Subcategory:
1 -RCS Level Initiating Condition:
UNPLANNED loss of reactor vessel/RCS inventory for 15 minutes or longer EAL: CU1.2 Unusual Event Reactor vessel/RCS level cannot be monitored AND UNPLANNED increase in any Table C-1 sump or tank levels due to a loss of reactor vessel/RCS inventory Table C-1 Sumps & Tanks" RB Sump" CCW surge tank" PRT* RCDT Mode Applicability:
5 -Cold Shutdown, 6 -Refueling Definition(s):
UNPLANNED-.
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown.Basis: Plant-Specific RCS elevations are illustrated in Figures 2 and 3 (ref. 1, 5). RCS level can be monitored by one or more of the following (ref. 2, 3): " LI-462, COLD CAL LEVEL % (ref. 4)* Control Room tygon hose TV monitor and RB camera Page 92 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]" Mid Loop Monitoring System* Mansell Level Monitoring System (MLMS)" RVLIS (ref. 1,2, 5)The following disagreements between the tygon hose and Mansell Level Monitoring System, or the Mid-Loop Monitoring System require RCS draindown termination and Operations Management resolution of the cause of the level discrepancy (ref. 5): " When RCS level is above the reactor vessel Flange mating surface and disagreement of greater than one foot exists." When RCS level is below the reactor vessel Flange mating surface and disagreement of greater than six inches exists.In this EAL, all water level indication is unavailable, and the reactor vessel inventory loss must be detected by sump or tank level changes (Table C-1). Plant design and procedures provide the capability to detect and assess primary system leakage (ref. 6, 7, 8, 9).Generic This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor reactor vessel/RCS level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.Refueling evolutions that decrease RCS water inventory are carefully planned and controlled.
An UNPLANNED event that results in water level decreasing below a procedurally required limit concurrent with indications of coolant leakage warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered.The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level.Page 93 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]This EAL addresses a condition where all means to determine reactor vessel/RCS level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump/tank levels. Sump/tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the reactor vessel/RCS.
Continued loss of RCS inventory may result in escalation to the Alert ECL via either IC CA1 or CA3.VCSNS Basis Reference(s):
- 1. GOP-9 Mid-Loop Operation 2. SOP-1 01 Reactor Coolant System 3. SOP-1 15 Residual Heat Removal 4. 201-325 Main Control Board Instrumentation Control Panel XCP-6109 5. GOP-7 Core Refueling (Mode 5 to Mode 6, Defuel, and Refuel to Mode 6)6. ARP-001-XCP-615
- 7. FSAR Section 5.2.7.1.3 8. AOP-101.1 Loss of Reactor Coolant not Requiring SI 9. FSAR Section 5.2.7.1.3.8
- 10. NEI 99-01 CUL1 Page 94 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
C -Cold Shutdown / Refueling System Malfunction Subcategory:
1 -RCS Level Initiating Condition:
Loss of reactor vessel/RCS inventory EAL: CA1.1 Alert Loss of reactor vessel/RCS inventory as indicated by level < 429'-6" elevation, < 64.5%RVLIS Narrow Range (bottom of hot leg penetration)
Mode Applicability:
5 -Cold Shutdown, 6 -Refueling Definition(s):
None Basis: Plant-Specific When reactor vessel water level drops to 429'-6.5" elevation (ref. 1), the inside diameter of the bottom of the RCS hot leg penetration is uncovered.
Hot leg centerline:
430'-9" elevation Hot leg inside diameter:
-29" Bottom of hot leg: 430'-9" -29"/2 = 429'-6.5" elevation (rounded to 429'-6")RCS elevations are illustrated in Figures 2 and 3 (ref. 1, 5). RCS level can be monitored by one or more of the following (ref. 2, 3): " LI-462, COLD CAL LEVEL % (ref. 4)" Control Room tygon hose TV monitor and RB camera* Mid Loop Monitoring System" Mansell Level Monitoring System (MLMS)" RVLIS Narrow Range reading of 64.2% (rounded to 64.5%) is the bottom of the hot leg penetration (ref. 1, 2, 5)Page 95 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]When RCS level is below the reactor vessel Flange mating surface and disagreement of greater than six inches exists between the tygon hose and either the Mansell Level Monitoring System or the Mid-Loop Monitoring System, any RCS draindown must be terminated and Operations Management must resolve the cause of the level discrepancy (ref. 5).Generic This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier).
This condition represents a potential substantial reduction in the level of plant safety.For this EAL, a lowering of water level below 429'-6" indicates that operator actions have not been successful in restoring and maintaining reactor vessel/RCS water level. The heat-up rate of the coolant will increase as the available water inventory is reduced. A continuing decrease in water level will lead to core uncovery.Although related, this EAL is concerned with the loss of RCS inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Decay Heat Removal suction point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS1 If the reactor vessel/RCS inventory level continues to lower, then escalation to Site Area Emergency would be via IC CS1.VCSNS Basis Reference(s):
- 1. GOP-9 Mid-Loop Operation 2. SOP-1 01 Reactor Coolant System 3. SOP-1 15 Residual Heat Removal 4. 201-325 Main Control Board Instrumentation Control Panel XCP-6109 5. GOP-7 Core Refueling (Mode 5 to Mode 6, Defuel, and Refuel to Mode 6)6. NEI 99-01 CA1 Page 96 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
C -Cold Shutdown / Refueling System Malfunction Subcategory:
1 -RCS Level Initiating Condition:
Loss of reactor vessel/RCS inventory EAL: CA1.2 Alert Reactor vessel/RCS level cannot be monitored for -15 min. (Note 1)AND UNPLANNED increase in any Table C-1 sump or tank levels due to a loss of reactor vessel/RCS inventory Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table C-1 Sumps & Tanks 0 RB Sump 0 CCW surge tank 0 PRT e RCDT Mode Applicability:
5 -Cold Shutdown, 6 -Refueling Definition(s):
UNPLANNED-.
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown.Basis: Plant-Specific In this condition, all water level indication would be unavailable, and the reactor vessel inventory loss must be detected by sump or tank level changes (Table C-1). Plant design and procedures provide the capability to detect and assess primary system leakage (ref. 1, 2, 3, 4).Page 97 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Generic This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier).
This condition represents a potential substantial reduction in the level of plant safety.For this EAL, the inability to monitor reactor vessel/RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation.
If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump/tank levels. Sump/tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the reactor vessel/RCS.
The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CSI.If the reactor vessel/RCS inventory level continues to lower, then escalation to Site Area Emergency would be via IC CSI.VCSNS Basis Reference(s):
- 1. ARP-001-XCP-615
- 2. FSAR Section 5.2.7.1.3 3. AOP-1 01.1 Loss of Reactor Coolant not Requiring SI 4. FSAR Section 5.2.7.1.3.8
- 5. NEI 99-01 CA1 Page 98 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
C -Cold Shutdown / Refueling System Malfunction Subcategory:
1 -RCS Level Initiating Condition:
Loss of reactor vessel/RCS inventory affecting core decay heat removal capability EAL: CS1.1 Site Area Emergency CONTAINMENT CLOSURE not established AND Reactor vessel level < 429' elevation, < 63% RVLIS Narrow Range (6" below the bottom of the hot leg penetration)
Mode Applicability:
5 -Cold Shutdown, 6 -Refueling Definition(s):
CONTAINMENT CLOSURE -The procedurally defined conditions or actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.
OAP-1 08.4, Operations Outage Control of Containment Penetrations, and SSP-004, Outage Safety Review Guidelines, prescribe Containment Closure actions and associated conditions.
A containment condition in which all penetrations providing direct access from the containment atmosphere to the outside atmosphere are closed by at least one automatic isolation valve, blind flange, or manual valve (ref. 9). Containment Closure is applicable to Mode 5 reduced RCS inventory operation and during Mode 6 Core Alterations or movement of spent fuel in the Reactor Building.
The Reactor Building penetrations shall be closed or capable of being closed prior to the onset of core boiling upon loss of the decay heat removal capability (ref. 10): A. The equipment door held in place by a minimum of four bolts.B. A minimum of one door in each personnel airlock closed.Page 99 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]C. Each penetration providing direct access from the Reactor Building atmosphere to the outside atmosphere shall be either: 1. Closed by an isolation valve, blind flange, or manual valve, or 2. Capable of being closed by an Operable automatic Reactor Building Purge and Exhaust isolation valve.D. All temporary penetrations are sealed.Basis: Plant-Specific When reactor vessel water level decreases to 429'-0.5" elevation, water level is six inches below the elevation of the bottom of the RCS hot leg penetration (ref. 1).Hot leg centerline:
430'-9" elevation Hot leg inside diameter:
-29" Bottom of hot leg less six inches: 430'-9" -29"/2 -6" = 429'-0.5" elevation rounded to 429'When reactor vessel water level drops significantly below the elevation of the bottom of the RCS hot leg penetration (six inches or more), all sources of RCS injection have failed or are incapable of making up for the inventory loss. RCS elevations are illustrated in Figures 2 and 3 (ref. 1, 5). RCS level can be monitored by one or more of the following (ref. 2, 3):* LI-462, COLD CAL LEVEL % (ref. 4)0 Control Room tygon hose TV monitor and RB camera" Mid Loop Monitoring System" Mansell Level Monitoring System (MLMS)" RVLIS Narrow Range reading of 62.9% (rounded to 63%) is six inches below the bottom of the hot leg penetration (ref. 1, 2, 5)Page 100 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RCS or reactor vessel water level decrease and potential core uncovery.
The inability to restore and maintain level after reaching this setpoint infers a failure of the RCS barrier and potential loss of the Fuel Clad barrier.Generic This IC addresses a significant and prolonged loss of reactor vessel/RCS inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.
Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions.
The difference in the specified RCS/reactor vessel levels of CS1.1 and CS1.2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.
This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.
Escalation of the ECL would be via IC CG1 or RG1.VCSNS Basis Reference(s):
- 1. GOP-9 Mid-Loop Operation 2. SOP-1 01 Reactor Coolant System 3. SOP-1 15 Residual Heat Removal 4. 201-325 Main Control Board Instrumentation Control Panel XCP-6109 Page 101 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT El 5. GOP-7 Core Refueling (Mode 5 to Mode 6, Defuel, and Refuel to Mode 6)6. OAP-1 08.4 Operations Outage Control of Containment Penetrations
- 7. SSP-004 Outage Safety Review Guidelines
- 8. NEI 99-01 CS1 Page 102 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
C -Cold Shutdown / Refueling System Malfunction Subcategory:
1 -RCS Level Initiating Condition:
Loss of reactor vessel/RCS inventory affecting core decay heat removal capability EAL: CS1.2 Site Area Emergency CONTAINMENT CLOSURE established AND Reactor vessel level < 427' elevation, < 58% RVLIS Narrow Range (top of active fuel)Mode Applicability:
5 -Cold Shutdown, 6 -Refueling Definition(s):
CONTAINMENT CLOSURE- The procedurally defined conditions or actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.
OAP-108.4, Operations Outage Control of Containment Penetrations, and SSP-004, Outage Safety Review Guidelines, prescribe CONTAINMENT CLOSURE actions and associated conditions.
A containment condition in which all penetrations providing direct access from the containment atmosphere to the outside atmosphere are closed by at least one automatic isolation valve, blind flange, or manual valve (ref. 6). CONTAINMENT CLOSURE is applicable to Mode 5 reduced RCS inventory operation and during Mode 6 Core Alterations or movement of spent fuel in the Reactor Building.
The Reactor Building penetrations shall be closed or capable of being closed prior to the onset of core boiling upon loss of the decay heat removal capability (ref. 7): A. The equipment door held in place by a minimum of four bolts.B. A minimum of one door in each personnel airlock closed.Page 103 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]C. Each penetration providing direct access from the Reactor Building atmosphere to the outside atmosphere shall be either: 1. Closed by an isolation valve, blind flange, or manual valve, or 2. Capable of being closed by an Operable automatic Reactor Building Purge and Exhaust isolation valve.D. All temporary penetrations are sealed.Basis: Plant-Specific When reactor vessel water level drops below 427'-0.27" elevation rounded to 427' (ref. 1), core uncovery is about to occur. RCS elevations are illustrated in Figures 2 and 3 (ref. 1, 5). RCS level can be monitored by one or more of the following (ref. 2, 3):* LI-462, COLD CAL LEVEL % (ref. 4)* Control Room tygon hose TV monitor and RB camera* Mid Loop Monitoring System* Mansell Level Monitoring System (MLMS)* RVLIS Narrow Range reading of 57.9% (rounded to 58%) is top of active fuel (ref. 1, 2,5)Generic This IC addresses a significant and prolonged loss of reactor vessel/RCS inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.
Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.Page 104 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions.
The difference in the specified RCS/reactor vessel levels of CS1.1 and CS1.2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.
This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.
Escalation of the ECL would be via IC CG1 or RG1.VCSNS Basis Reference(s):
- 1. GOP-9 Mid-Loop Operation 2. SOP-101 Reactor Coolant System 3. SOP-1 15 Residual Heat Removal 4. 201-325 Main Control Board Instrumentation Control Panel XCP-6109 5. GOP-7 Core Refueling (Mode 5 to Mode 6, Defuel, and Refuel to Mode 6)6. OAP-108.4 Operations Outage Control of Containment Penetrations
- 7. SSP-004 Outage Safety Review Guidelines
- 8. NEI 99-01 CS1 Page 105 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
C -Cold Shutdown / Refueling System Malfunction Subcategory:
1 -RCS Level Initiating Condition:
Loss of reactor vessel/RCS inventory affecting core decay heat removal capability EAL: CS1.3 Site Area Emergency Reactor vessel/RCS level cannot be monitored for _> 30 min. (Note 1)AND Core uncovery is indicated by any of the following:
- RM-G6 Rx Bldg Refueling Bridge or RM-G17A/B Rx Bldg Manipulator Crane offscale-high" Erratic source range monitor indication" UNPLANNED increase in any Table C-1 sump / tank level of sufficient magnitude to indicate core uncovery Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table C-1 Sumps & Tanks* RB Sump* CCW surge tank* PRT* RCDT Mode Applicability:
5 -Cold Shutdown, 6 -Refueling Definition(s):
UNPLANNED-.
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown.Basis: Plant-Specific Page 106 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]If all means of level monitoring are not available, the reactor vessel inventory loss may be detected by the Containment area radiation monitors, erratic Source Range Monitors, or indication or sump/tank level increases: " As water level in the reactor vessel lowers, the dose rate above the core will increase.
The dose rate due to this core shine should result in off-scale indication on the listed monitors.
RM-G6 (Rx Bldg Refueling Bridge) and RM-G17A/B (Rx Bldg Manipulator Crane) are located on the Refueling Bridge in the Containment and provide monitoring of radiation due to a dropped fuel assembly during refueling operations.
RM-G17A/B are only installed in Mode 6. This results in a fuel cladding rupture with the release of the gap activity.
The noble gases are expected to bubble up to the surface of the pool where the monitors will provide detection and alarm at___1 R/hr. RM-G17A and RMG-17B provide purge isolation in the event of a fuel drop cladding rupture. RM-G6 and RM-G17A/B have an indication range of 1 -10 5 mR/hr. If any of these radiation monitors reach and exceed 10 5 mR/hr (offscale-high), a loss of inventory with potential to uncover the core is likely to have occurred.
RM-G7 and RM-G18 are the Containment High Range Radiation Monitors but are not located in the Containment with sufficient line-of-sight to the irradiated fuel in the reactor vessel to be of use in detecting loss of inventory above the core.(ref. 1)" Post-TMI studies indicate that the installed nuclear instrumentation will operate erratically when the core is uncovered and source range monitors can be used as a tool for making such determinations.
Figure 4 shows the response of the source range monitor during the first few hours of the TMI-2 accident.
The instrument reported an increasing signal about 30 minutes into the accident.
At this time, the reactor coolant pumps were running and the core was adequately cooled as indicated by the core outlet thermocouples.
Hence, the increasing signal was the result of an increasing two-phase void fraction in the reactor core and vessel downcomer and the reduced shielding that the two-phase mixture provide to the source range monitor (ref. 2, 3). Source range count rate is indicated in the Control Page 107 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Room on MCB Panel XCP-6110 Source Range Monitors NI-31B and NI-32B, and NIS Recorder NR-45 (ref. 4): If water level monitoring capability is unavailable, the reactor vessel inventory loss may be detected by sump or tank level changes (Table C-1). Plant design and procedures provide the capability to detect and assess primary system leakage (ref. 5, 6, 7, 8).Generic This IC addresses a significant and prolonged loss of reactor vessel/RCS inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.
Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties).
It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.
The inability to monitor reactor vessel/RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation.
If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump/tank levels. Sump/tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the reactor vessel/RCS.
This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-Page 108 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.
Escalation of the ECL would be via IC CG1 or RG1.VCSNS Basis Reference(s):
- 1. OAP-1 08.4 Operations Outage Control of Containment Penetrations
- 2. Severe Accident Management Guidance Technical Basis Report, Volume 1: Candidate High-Level Actions and Their Effects, pgs 2-18, 2-19 3. Nuclear Safety Analysis Center (NSAC), 1980, "Analysis of Three Mile Island -Unit 2 Accident," NSAC-1 4. 201-326 Main Control Board Instrumentation Control Panel XCP-61 10 5. ARP-001-XCP-615
- 6. FSAR Section 5.2.7.1.3 7. AOP-101.1 Loss of Reactor Coolant not Requiring SI 8. FSAR Section 5.2.7.1.3.8
- 9. NEI 99-01 CS1 Page 109 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT El Category: Subcategory:
Initiating Condition:
C -Cold Shutdown / Refueling System Malfunction 1 -RCS Level Loss of reactor vessel/RCS inventory affecting fuel clad integrity with containment challenged EAL: CG1.1 General Emergency Reactor vessel level < 427' elevation, < 58% RVLIS Narrow Range (top of active fuel) for>30 min.AND Any of the following indications of containment challenge: " CONTAINMENT CLOSURE not established (Note 7)" Containment hydrogen concentration
> 4%* UNPLANNED increase in Containment pressure Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 7: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required.Mode Applicability:
5 -Cold Shutdown, 6 -Refueling Definition(s):
CONTAINMENT CLOSURE -The procedurally defined conditions or actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.
OAP-1 08.4, Operations Outage Control of Containment Penetrations, and SSP-004, Outage Safety Review Guidelines, prescribe CONTAINMENT CLOSURE actions and associated conditions.
A containment condition in which all penetrations providing direct access from the containment atmosphere to the outside atmosphere are closed by at least one automatic isolation valve, blind flange, or manual valve (ref. 9). CONTAINMENT CLOSURE is applicable to Mode 5 reduced RCS inventory operation and during Mode 6 Core Alterations or movement of spent fuel in the Reactor Building.
The Reactor Building Page 110 of 346 EPP-1 08 ENCLOSURE I REVISION 01 [DRAFT E]penetrations shall be closed or capable of being closed prior to the onset of core boiling upon loss of the decay heat removal capability (ref. 10): A. The equipment door held in place by a minimum of four bolts.B. A minimum of one door in each personnel airlock closed.C. Each penetration providing direct access from the Reactor Building atmosphere to the outside atmosphere shall be either: 1. Closed by an isolation valve, blind flange, or manual valve, or 2. Capable of being closed by an Operable automatic Reactor Building Purge and Exhaust isolation valve.D. All temporary penetrations are sealed.UNPLANNED-.
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown.Basis: Plant-Specific When reactor vessel water level drops below 427'-0.27" elevation rounded to 427' (ref. 1), core uncovery is about to occur. RCS elevations are illustrated in Figure 2 and 3 (ref. 1, 5).RCS level can be monitored by one or more of the following (ref. 2, 3): " LI-462, COLD CAL LEVEL % (ref. 4)* Control Room tygon hose TV monitor and RB camera" Mid Loop Monitoring System" Mansell Level Monitoring System (MLMS)* RVLIS Narrow Range reading of 57.9% (rounded to 58%) is top of active fuel (ref.1, 2,4)Three indications are associated with a challenge to Containment:
Page 111 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]CONTAINMENT CLOSURE is not established.
In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive mixture of dissolved gases in Containment.
However, Containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that an explosive mixture exists. A combustible mixture can be formed when hydrogen gas concentration in the Containment atmosphere is greater than 4% by volume (ref. 8, 9). All hydrogen measurements are referenced to concentrations in dry air even though the actual Containment environment may contain significant steam concentrations.
The plant has two hydrogen monitoring systems. Sample points are located near each recombiner and near the RBCUs on the 530' Level. Manual action is required to start the redundant hydrogen analyzers.
The analyzers have a range of 0-10% and 0-20% of H 2 in air (by volume) and an accuracy of +/- 2% of range. Hydrogen concentration in the Reactor Building is indicated in the control room (ref. 7).An UNPLANNED pressurization that can breach the containment barrier signifies a challenge to the Containment pressure retaining capability which is dependent on the status of the containment.
If containment integrity is established for full power operation, a breach could occur if the design containment pressure is exceeded (57 psig). For this condition, a small UNPLANNED pressure rise above atmospheric pressure does not challenge containment.
If in refueling operations, however, a breach could occur if the UNPLANNED pressure rise exceeded the capability of a temporary containment seal. This would occur at a much lower pressure than the containment design pressure.
Use of the verb "...can breach...:
instead of"breaches" provides the Emergency Director with the latitude to assess the magnitude and rate of the containment pressure rise with respect to the barrier status (for the existing operating mode) and determine that the containment challenge exists due to elevated pressure either before or at the time that the actual breach of the barrier occurs.Page 112 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Generic This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged.
This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity.
Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment.
If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity.
It therefore represents a challenge to Containment integrity.
The plant has two hydrogen monitoring systems. Sample points are located near each recombiner and near the RBCUs on the 530' Level. Manual action is required to start the redundant hydrogen analyzers.
The analyzers
[CI-8257 (8258)] have a range of 0-10%and 0-20% of H 2 in air (by volume) and an accuracy of +/- 2% of range. Hydrogen concentration in the Reactor Building is indicated in the control room (ref. 3, 12).In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment.
If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use Page 113 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT El the other listed indications to assess whether or not containment is challenged.
The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties).
It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.
This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.
VCSNS Basis Reference(s):
- 1. GOP-9 Mid-Loop Operation 2. SOP-1 01 Reactor Coolant System 3. SOP-1 15 Residual Heat Removal 4. 201-325 Main Control Board Instrumentation Control Panel XCP-6109 5. GOP-7 Core Refueling (Mode 5 to Mode 6, Defuel, and Refuel to Mode 6)6. SOP-1 22 Post Accident Hydrogen Removal System 7. FSAR Section 6.2.5.5.3 8. FSAR Section 6.2.3.5.1 9. OAP-1 08.4 Operations Outage Control of Containment Penetrations
- 10. SSP-004 Outage Safety Review Guidelines
- 11. SOP-1 22 Post Accident Hydrogen Removal System 12. NEI 99-01 CG1 Page 114 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
C -Cold Shutdown / Refueling System Malfunction Subcategory:
1 -RCS Level Initiating Condition:
Loss of reactor vessel/RCS inventory affecting fuel clad integrity with containment challenged EAL: CG1.2 General Emergency Reactor vessel/RCS level cannot be monitored for > 30 min. (Note 1)AND Core uncovery is indicated by any of the following:
- RM-G6 Rx Bldg Refueling Bridge or RM-G17A/B Rx Bldg Manipulator Crane offscale-high
- Erratic source range monitor indication
- UNPLANNED increase in any Table C-1 sump / tank level of sufficient magnitude to indicate core uncovery AND Any of the following indications of containment challenge: " CONTAINMENT CLOSURE not established (Note 7)* Containment hydrogen concentration
> 4%* UNPLANNED increase in Containment pressure Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 7: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required.Table C-1 Sumps & Tanks* RB Sump* CCW surge tank* PRT e RCDT Mode Applicability:
5 -Cold Shutdown, 6 -Refueling Definition(s):
Page 115 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]CONTAINMENT CLOSURE -The procedurally defined conditions or actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.
OAP-1 08.4, Operations Outage Control of Containment Penetrations, and SSP-004, Outage Safety Review Guidelines, prescribe CONTAINMENT CLOSURE actions and associated conditions.
A containment condition in which all penetrations providing direct access from the containment atmosphere to the outside atmosphere are closed by at least one automatic isolation valve, blind flange, or manual valve (ref. 15). CONTAINMENT CLOSURE is applicable to Mode 5 reduced RCS inventory operation and during Mode 6 Core Alterations or movement of spent fuel in the Reactor Building.
The Reactor Building penetrations shall be closed or capable of being closed prior to the onset of core boiling upon loss of the decay heat removal capability (ref. 16): A. The equipment door held in place by a minimum of four bolts.B. A minimum of one door in each personnel airlock closed.C. Each penetration providing direct access from the Reactor Building atmosphere to the outside atmosphere shall be either: 1. Closed by an isolation valve, blind flange, or manual valve, or 2. Capable of being closed by an Operable automatic Reactor Building Purge and Exhaust isolation valve.D. All temporary penetrations are sealed.UNPLANNED-.
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown.Basis: Plant-Specific Page 116 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]When reactor vessel water level drops below 427'-0.27" elevation rounded to 427' (ref. 1), core uncovery is about to occur. RCS elevations are illustrated in Figures 2 and 3 (ref. 1, 5). RCS level can be monitored by one or more of the following (ref. 2, 3): " LI-462, COLD CAL LEVEL % (ref. 4)* Control Room tygon hose TV monitor and RB camera* Mid Loop Monitoring System" Mansell Level Monitoring System (MLMS)" RVLIS Narrow Range reading of 57.9% is top of active fuel (ref. 1. 2, 5)If all means of level monitoring are not available, the reactor vessel inventory loss may be detected by the Containment area radiation monitors, erratic Source Range Monitors, or indication or sump/tank level increases:
As water level in the reactor vessel lowers, the dose rate above the core will increase.
The dose rate due to this core shine should result in off-scale indication on the listed monitors.
RM-G6 (Rx Bldg Refueling Bridge) and RM-G17A/B (Rx Bldg Manipulator Crane) are located on the Refueling Bridge in the Containment and provide monitoring of radiation due to a dropped fuel assembly during refueling operations.
RM-G17A/B are only installed in Mode 6. This results in a fuel cladding rupture with the release of the gap activity.
The noble gases are expected to bubble up to the surface of the pool where the monitors will provide detection and alarm at1 R/hr. RM-G17A and RMG-17B provide purge isolation in the event of a fuel drop cladding rupture. RM-G6 and RM-G17A/B have an indication range of 1 -10 5 mR/hr. If any of these radiation monitors reach and exceed 105 mR/hr (offscale-high), a loss of inventory with potential to uncover the core is likely to have occurred.
RM-G7 and RM-G18 are the Containment High Range Radiation Monitors but are not located in the Containment with sufficient line-of-sight to the irradiated fuel in the reactor vessel to be of use in detecting loss of inventory above the core.(ref. 6)Page 117 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]* Post-TMI studies indicate that the installed nuclear instrumentation will operate erratically when the core is uncovered and source range monitors can be used as a tool for making such determinations.
Figure 4 shows the response of the source range monitor during the first few hours of the TMI-2 accident.
The instrument reported an increasing signal about 30 minutes into the accident.
At this time, the reactor coolant pumps were running and the core was adequately cooled as indicated by the core outlet thermocouples.
Hence, the increasing signal was the result of an increasing two-phase void fraction in the reactor core and vessel downcomer and the reduced shielding that the two-phase mixture provide to the source range monitor (ref. 7, 8). Source range count rate is indicated in the Control Room on MCB Panel XCP-61 10 Source Range Monitors NI-31 B and NI-32B, and NIS Recorder NR-45 (ref. 9): " If water level monitoring capability is unavailable, the reactor vessel inventory loss may be detected by sump or tank level changes (Table C-1). Plant design and procedures provide the capability to detect and assess primary system leakage (ref.10, 11, 12, 13).Three indications are associated with a challenge to Containment: " CONTAINMENT CLOSURE is not established.
- In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive mixture of dissolved gases in Containment.
However, Containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that an explosive mixture exists. A combustible mixture can be formed when hydrogen gas concentration in the Containment atmosphere is greater than 4% by volume (ref. 2, 3). All hydrogen measurements are referenced to concentrations in dry air even though the actual Containment environment may contain significant steam concentrations.
The plant has two hydrogen monitoring systems. Sample points are located near each recombiner and near the RBCUs on the 530' Level. Manual action is required to start the redundant hydrogen analyzers.
Page 118 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]The analyzers have a range of 0-10% and 0-20% of H 2 in air (by volume) and an accuracy of +/- 2% of range. Hydrogen concentration in the Reactor Building is indicated in the control room (ref. 1).An UNPLANNED pressurization that can breach the containment barrier signifies a challenge to the Containment pressure retaining capability which is dependent on the status of the containment.
If containment integrity is established for full power operation, a breach could occur if the design containment pressure is exceeded (57 psig). For this condition, a small UNPLANNED pressure rise above atmospheric pressure does not challenge containment.
If in refueling operations, however, a breach could occur if the UNPLANNED pressure rise exceeded the capability of a temporary containment seal. This would occur at a much lower pressure than the containment design pressure.
Use of the verb "...can breach...:
instead of"breaches" provides the Emergency Director with the latitude to assess the magnitude and rate of the containment pressure rise with respect to the barrier status (for the existing operating mode) and determine that the containment challenge exists due to elevated pressure either before or at the time that the actual breach of the barrier occurs.Generic This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged.
This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity.
Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.Page 119 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment.
If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity.
It therefore represents a challenge to Containment integrity.
In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment.
If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged.
The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties).
It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.
The inability to monitor reactor vessel/RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation.
If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump/tank levels. Sump/tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the reactor vessel/RCS.
This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Page 120 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.
VCSNS Basis Reference(s):
- 1. GOP-9 Mid-Loop Operation 2. SOP-1 01 Reactor Coolant System 3. SOP-1 15 Residual Heat Removal 4. 201-325 Main Control Board Instrumentation Control Panel XCP-6109 5. GOP-7 Core Refueling (Mode 5 to Mode 6, Defuel, and Refuel to Mode 6)6. Design Bases Document -Radiation Monitoring System (RM)7. Severe Accident Management Guidance Technical Basis Report, Volume 1: Candidate High-Level Actions and Their Effects, pgs 2-18, 2-19 8. Nuclear Safety Analysis Center (NSAC), 1980, "Analysis of Three Mile Island -Unit 2 Accident," NSAC-1 9. 201-326 Main Control Board Instrumentation Control Panel XCP-6110 10. ARP-001-XCP-615
- 11. FSAR Section 5.2.7.1.3 12. AOP-1 01.1 Loss of Reactor Coolant not Requiring SI 13. FSAR Section 5.2.7.1.3.8
- 14. OAP-1 03.2 Emergency Operating Procedure Setpoint Document 15. OAP-1 08.4 Operations Outage Control of Containment Penetrations
- 16. SSP-004 Outage Safety Review Guidelines
- 17. FSAR Section 6.2.3.5.1 18. SOP-1 22 Post Accident Hydrogen Removal System 19. FSAR Section 6.2.5.5.3 20. NEI 99-01 CG1 Page 121 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
C -Cold Shutdown / Refueling System Malfunction Subcategory:
2 -Loss of ESF AC Power Initiating Condition:
Loss of all but one AC power source to ESF buses for 15 minutes or longer.EAL: CU2.1 Unusual Event AC power capability to 7.2 KV ESF buses 1 DA and 1 DB reduced to a single power source (Table C-2) for > 15 min. (Note 1)AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table C-2 AC Power Supplies Offsite:* 115 KV power to XTF-4 and XTF-5* 230 KV power to XTF-31* Parr Hydro Plant 13.8 KV power to ESF bus 1DA or 1DB Onsite:* Diesel Generator A* Diesel Generator B Mode Applicability:
5 -Cold Shutdown, 6 -Refueling, D -Defueled Definition(s):
SAFETY SYSTEM- A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):
Page 122 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary;(2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis: Plant-Specific As used in this EAL the term "capability" means an AC power source is either currently powering essential loads on one or more 7.2 KV ESF buses or is capable of energizing and powering essential loads on at least one 7.2 KV ESF bus within 15 min.Table C-2 lists AC sources capable of powering ESF buses. Safeguards power originates offsite from two independent sources (ref. 1):* The Parr Generating Complex supplies 115 KV power to the two Engineered Safety Feature (ESF) transformers (XTF-4 and XTF-5). The transformer outputs are combined at 7.2 KV bus 1 DX and then supplied to 7.2 KV ESF bus 1 DA (Train A).This is the preferred or normal power source to Train A and the alternate power source for Train B.* 7.2 KV ESF bus 1 DB (Train B) is supplied from the emergency auxiliary transformer (XTF-31).
The emergency auxiliary transformer receives 230 KV power from the Virgil C. Summer substation (switchyard) bus 3. This transformer is the preferred power source for Train B and the alternate power source for Train A.The Parr Hydro Plant provides a 13.8 KV AC line to the 7.2 KV ESF buses. This Alternate AC (AAC) Power Supply has the capacity to supply only one fully loaded ESF bus (ref. 4).The AAC is designed to provide back-up power to either ESF bus whenever one of the Diesel Generators is out of service. The AAC is verified available and an operational readiness status check is performed when it is anticipated that one of the Diesel Page 123 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Generators will be inoperable for longer than the allowed outage time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The design of the AAC is capable of providing the required safety and non-safety related loads in the event of a total loss of offsite power and if both Diesel Generators fail to start and load. During these events it is assumed that there is no seismic event or an event that requires safeguards actuation (e.g., safety injection, containment spray, etc.) The AAC is capable of supplying sufficient power to mitigate the effects of an accident.
The AAC is not credited in the safety analysis.
The AAC is, however, capable of mitigating the dominant core damage sequences and provides a significant overall risk reduction for station operation. (ref. 7).The two trains of 7.2 KV safeguards power are also provided with an onsite standby source of power for supplying power when the ESF and emergency auxiliary transformers are not available.
The Diesel Generators A and B are capable of supplying all loads on the distribution network of their respective train (ref. 1, 2, 3, 4, 5).This cold condition EAL is equivalent to the hot condition loss of all offsite AC power EAL SA1.1.Generic This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment.
When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an Alert because of the increased time available to restore another power source to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant.An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to a safeguards bus. Some examples of this condition are presented below.Page 124 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]" A loss of all offsite power with a concurrent failure of all but one safeguards power source (e.g., an onsite diesel generator)." A loss of all offsite power and loss of all safeguards power sources (e.g., onsite diesel generators) with a single train of safeguards buses being back-fed from the unit main generator.
- A loss of safeguards power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from an offsite power source.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2.VCSNS Basis Reference(s):
- 1. FSAR Section 8 2. EOP-6.0 Loss of All ESF AC Power 3. EOP-1.0 Reactor Trip/Safety Injection Actuation 4. SOP-304 115KV/7.2KV Operations
- 5. SOP-306 Emergency Diesel Generator 6. AOP-304.4 Loss of all ESF AC Power While in Shutdown (Modes 5 and 6)7. Technical Specifications Bases 3/4.8 8. NEI 99-01 CU2 Page 125 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
C -Cold Shutdown / Refueling System Malfunction Subcategory:
2 -Loss of ESF AC Power Initiating Condition:
Loss of all offsite and all onsite AC power to ESF buses for 15 minutes or longer.EAL: CA2.1 Alert Loss of all offsite and all onsite AC power (Table C-2) capability to 7.2 KV ESF buses 1DA and 1DB for -a 15 min. (Note 1)Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table C-2 AC Power Supplies Offsite: 0 115 KV power to XTF-4 and XTF-5* 230 KV power to XTF-31 0 Parr Hydro Plant 13.8 KV power to ESF bus 1DA or 1DB Onsite:* Diesel Generator A* Diesel Generator B Mode Applicability:
5 -Cold Shutdown, 6 -Refueling, D -Defueled Definition(s):
SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary;Page 126 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E](2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis: Plant-Specific This EAL is indicated by the loss of all offsite and onsite AC power to 7.2 KV ESF buses 1 DA and 1 DB. Table C-2 lists AC sources capable of powering ESF buses. As used in this EAL the term "capability" means an AC power source is either currently powering essential loads on one or more 7.2 KV ESF buses or is capable of energizing and powering essential loads on at least one 7.2 KV ESF bus within 15 min.Safeguards power originates offsite from two independent sources (ref. 1):* The Parr Generating Complex supplies 115 KV power to the two Engineered Safety Feature (ESF) transformers (XTF-4 and XTF-5). The transformer outputs are combined at 7.2 KV bus 1 DX and then supplied to 7.2 KV ESF bus 1 DA (Train A).This is the preferred or normal power source to Train A and the alternate power source for Train B.* 7.2 KV ESF bus 1 DB (Train B) is supplied from the emergency auxiliary transformer (XTF-31).
The emergency auxiliary transformer receives 230 KV power from the Virgil C. Summer substation (switchyard) bus 3. This transformer is the preferred power source for Train B and the alternate power source for Train A.The Parr Hydro Plant provides a 13.8 KV AC line to the 7.2 KV ESF buses. This Alternate AC Power Supply has the capacity to supply only one fully loaded ESF bus (ref. 4).The AAC is designed to provide back-up power to either ESF bus whenever one of the Diesel Generators is out of service, particularly in Modes 1 through 4. The AAC is verified available and an operational readiness status check is performed when it is anticipated that one of the Diesel Generators will be inoperable for longer than the allowed outage time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The design of the AAC is capable of providing the required safety and Page 127 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]non-safety related loads in the event of a total loss of offsite power and if both Diesel Generators fail to start and load. During these events it is assumed that there is no seismic event or an event that requires safeguards actuation (e.g., safety injection, containment spray, etc.) Although the AAC is not designed for DBA loads, it is capable of supplying sufficient power to mitigate the effects of an accident.
The AAC is not credited in the safety analysis.
The AAC is, however, capable of mitigating the dominant core damage sequences and provides a significant overall risk reduction for station operation.
The AAC alone is adequate to supply electrical power to affect a safe shutdown of the plant (ref. 7).The two trains of 7.2 KV safeguards power are also provided with an onsite standby source of power for supplying power when the ESF and emergency auxiliary transformers are not available.
The Diesel Generators A and B are capable of supplying all loads on the distribution network of their respective train (ref. 1, 2, 3, 4, 5).This EAL is the cold condition equivalent of the hot condition loss of all AC power EAL SS1.1.Generic This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the increased time available to restore a safeguards bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Escalation of the ECL would be via IC CS1 or RS1.Page 128 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]VCSNS Basis Reference(s):
- 1. FSAR Section 8 2. EOP-6.0 Loss of All ESF AC Power 3. EOP-1.0 Reactor Trip/Safety Injection Actuation 4. SOP-304 115KV/7.2KV Operations
- 5. SOP-306 Emergency Diesel Generator 6. AOP-304.4 Loss of all ESF AC Power While in Shutdown (Modes 5 and 6)7. Technical Specifications Bases 3/4.8 8. NEI 99-01 CA2 Page 129 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
C -Cold Shutdown / Refueling System Malfunction Subcategory:
3 -RCS Temperature Initiating Condition:
UNPLANNED increase in RCS temperature.
EAL: CU3.1 Unusual Event UNPLANNED increase in RCS temperature to > 200OF Mode Applicability:
5 -Cold Shutdown, 6 -Refueling Definition(s):
UNPLANNED-.
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown.Basis: Plant-Specific Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (200 0 F, ref. 1). The wide range RTDs indicate RCS temperature over a range of 0 0 F to 700'F. This range is necessary for transients and for heatup or cooldown operations.
The temperature is displayed on two separate pen recorders (TR-410, TR-413) located on MCB Panel XCP-6109.
Recorder TR-410 displays the cold leg wide range temperatures, and recorder TR-413 displays the hot leg wide range temperatures.
The wide range hot and cold leg temperatures are also displayed on meters. TI-41 0 and TI-420 display loop A and loop B cold leg temperatures respectively, and TI-413 and TI-423 display loop A and loop B hot leg temperatures (ref.2). RCS temperature is also monitored by Integrated Plant Computer System (IPCS)computer points. The IPCS Heatup/Cooldown program provides continuous updates of calculated Reactor Coolant System (RCS) and Pressurizer (PZR) rates based upon measured conditions of the plant. The Heatup/Cooldown program is executed through the turn-on-code (TOC) 'HUMMI' or the dedicated function key 'HUMMI'. Displays are Page 130 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]available at the SDS displays throughout the facility (ref. 3, 4) Heatup and Cooldown rate limitations are provided in Technical Specifications (ref. 5, 6).Generic This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limitand represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to IC CA3.A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.
This EAL involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications.
During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.
During an outage, the level in the reactor vessel will normally be maintained above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled.
A loss of forced decay heat removal at reduced inventory may result in a rapid increase in reactor coolant temperature depending on the time after shutdown.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.
Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.VCSNS Basis Reference(s):
- 1. Technical Specifications Table 1.1 2. 201-325 Main Control Board Instrumentation Control Panel XCP-6109 Page 131 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]3. STP-1 03.001 Reactor Coolant System and Pressurizer Heatup/Cooldown Surveillance
- 4. OAP-107.1 Control of IPCS Functions 5. Technical Specifications 3.4.9.1 6. Technical Specifications 3.4.9.2 7. NEI 99-01 CU3 Page 132 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
C -Cold Shutdown / Refueling System Malfunction Subcategory:
3 -RCS Temperature Initiating Condition:
UNPLANNED increase in RCS temperature.
EAL: CU3.2 Unusual Event Loss of all RCS temperature and reactor vessel/RCS level indication for __ 15 min.(Note 1)Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability:
5 -Cold Shutdown, 6 -Refueling Definition(s):
None Basis: Plant-Specific Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (200 0 F, ref. 1). The wide range RTDs indicate RCS temperature over a range of 0°F to 700 0 F. This range is necessary for transients and for heatup or cooldown operations.
The temperature is displayed on two separate pen recorders (TR-410, TR-413) located on MCB Panel XCP-6109.
Recorder TR-41 0 displays the cold leg wide range temperatures, and recorder TR-413 displays the hot leg wide range temperatures.
The wide range hot and cold leg temperatures are also displayed on meters. TI-410 and TI-420 display loop A and loop B cold leg temperatures respectively, and TI-413 and TI-423 display loop A and loop B hot leg temperatures (ref.2). RCS temperature is also monitored by Integrated Plant Computer System (IPCS)computer points. The IPCS Heatup/Cooldown program provides continuous updates of calculated Reactor Coolant System (RCS) and Pressurizer (PZR) rates based upon measured conditions of the plant. The Heatup/Cooldown program is executed through the turn-on-code (TOC) 'HUMMI' or the dedicated function key 'HUMMI'. Displays are Page 133 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]available at the SDS displays throughout the facility (ref. 3, 4) Heatup and Cooldown rate limitations are provided in Technical Specifications (ref. 5, 6).RCS elevations are illustrated in Figure 2 (ref. 7). RCS level can be monitored by one or more of the following (ref. 8, 9): " LI-462, COLD CAL LEVEL % (ref. 2)" Control Room tygon hose TV monitor and RB camera* Mid Loop Monitoring System" Mansell Level Monitoring System (MLMS)" RVLIS (ref. 2, 7, 10)The following disagreements between the tygon hose and Mansell Level Monitoring System, or the Mid-Loop Monitoring System require RCS draindown termination and Operations Management resolution of the cause of the level discrepancy prior to continued draining (ref. 10): " When RCS level is above the reactor vessel Flange mating surface and disagreement of greater than one foot exists.* When RCS level is below the reactor vessel Flange mating surface and disagreement of greater than six inches exists.Generic This EAL addresses the inability to determine RCS temperature and level, and represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to IC CA3.This EAL reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.
Page 134 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.
Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.VCSNS Basis Reference(s):
- 1. Technical Specifications Table 1.1 2. 201-325 Main Control Board Instrumentation Control Panel XCP-6109 3. STP-1 03.001 Reactor Coolant System and Pressurizer Heatup/Cooldown Surveillance
- 4. OAP-107.1 Control of IPCS Functions 5. Technical Specifications 3.4.9.1 6. Technical Specifications 3.4.9.2 7. GOP-9 Mid-Loop Operation 8. SOP-1 01 Reactor Coolant System 9. SOP-1 15 Residual Heat Removal 10. GOP-7 Core Refueling (Mode 5 to Mode 6, Defuel, and Refuel to Mode 6)11. NEI 99-01 CU3 Page 135 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
C -Cold Shutdown / Refueling System Malfunction Subcategory:
3 -RCS Temperature Initiating Condition:
Inability to maintain the plant in cold shutdown.EAL: CA3.1 Alert UNPLANNED increase in RCS temperature to > 200°F for > Table C-3 duration (Note 1)Note 1: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.Table C-3: RCS Heat-up Duration Thresholds RCS Status Containment Closure Heat-up Duration Status Intact AND not at N/A 60 min.*REDUCED INVENTORY Not intact OR at established 20 min.*REDUCED INVENTORY not established 0 min.* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.
Mode Applicability:
5 -Cold Shutdown, 6 -Refueling Definition(s):
CONTAINMENT CLOSURE -The procedurally defined conditions or actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.
OAP-1 08.4, Operations Outage Control of Containment Penetrations, and SSP-004, Outage Safety Review Guidelines, prescribe CONTAINMENT CLOSURE actions and associated conditions.
A containment condition in which all penetrations providing direct access from the containment atmosphere to the outside atmosphere are closed by at least one automatic isolation valve, blind flange, or manual valve (ref. 7). CONTAINMENT CLOSURE is applicable to Mode 5 reduced RCS inventory operation and during Mode 6 Core Alterations or movement of spent fuel in the Reactor Building.
The Reactor Building Page 136 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]penetrations shall be closed or capable of being closed prior to the onset of core boiling upon loss of the decay heat removal capability (ref. 8): A. The equipment door held in place by a minimum of four bolts.B. A minimum of one door in each personnel airlock closed.C. Each penetration providing direct access from the Reactor Building atmosphere to the outside atmosphere shall be either: 1. Closed by an isolation valve, blind flange, or manual valve, or 2. Capable of being closed by an Operable automatic Reactor Building Purge and Exhaust isolation valve.D. All temporary penetrations are sealed.UNPLANNED-.
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown.Basis: Plant-Specific A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.
The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams).Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (200 0 F, ref. 1). The wide range RTDs indicate RCS temperature over a range of 0°F to 700 0 F. This range is necessary for transients and for heatup or cooldown operations.
The temperature is displayed on two separate pen recorders (TR-410, TR-413) located on MCB Panel XCP-6109.
Recorder TR-410 displays the cold leg wide range temperatures, and recorder TR-413 displays the hot leg wide range temperatures.
The wide range hot and cold leg temperatures are also Page 137 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]displayed on meters. TI-410 and TI-420 display loop A and loop B cold leg temperatures respectively, and TI-413 and TI-423 display loop A and loop B hot leg temperatures (ref.2). RCS temperature is also monitored by Integrated Plant Computer System (IPCS)computer points. The IPCS Heatup/Cooldown program provides continuous updates of calculated Reactor Coolant System (RCS) and Pressurizer (PZR) rates based upon measured conditions of the plant. The Heatup/Cooldown program is executed through the turn-on-code (TOC) 'HUMMI' or the dedicated function key 'HUMMI'. Displays are available at the SDS displays throughout the facility (ref. 3, 4) Heatup and Cooldown rate limitations are provided in Technical Specifications (ref. 5, 6).Generic This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant.A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.
The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact, or RCS inventory is reduced (e.g., mid-loop operation in PWRs). The 20-minute criterion was included to allow time for operator action to address the temperature increase.The RCS Heat-up Duration Thresholds table also addresses an increase in RCS temperature with the RCS INTACT. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature increase without a substantial degradation in plant safety.Page 138 of 346 EPP-1 08 ENCLOSURE I REVISION 01 [DRAFT E]Finally, in the case where there is an increase in RCS temperature, the RCS is not intact or is at REDUCED INVENTORY, and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes).
This is because 1) the evaporated reactor coolant may be released directly into the Containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel.Escalation of the ECL would be via IC CS1 or RS1.VCSNS Basis Reference(s):
- 1. Technical Specifications Table 1.1 2. 201-325 Main Control Board Instrumentation Control Panel XCP-6109 3. STP-1 03.001 Reactor Coolant System and Pressurizer Heatup/Cooldown Surveillance
- 4. OAP-107.1 Control of IPCS Functions 5. Technical Specifications 3.4.9.1 6. Technical Specifications 3.4.9.2 7. OAP-1 08.4 Operations Outage Control of Containment Penetrations
- 8. SSP-004 Outage Safety Review Guidelines
- 9. NEI 99-01 CA3 Page 139 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
C -Cold Shutdown / Refueling System Malfunction Subcategory:
3 -RCS Temperature Initiating Condition:
Inability to maintain the plant in cold shutdown.EAL: CA3.2 Alert UNPLANNED RCS pressure increase > 10 psig (This EAL does not apply during water-solid plant conditions)
Mode Applicability:
5 -Cold Shutdown, 6 -Refueling Definition(s):
UNPLANNED-.
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown.Basis: Plant-Specific A 10 psig RCS pressure increase can be monitored on PI-402A1 (0-600 psig) on MCB Panel XCP-6108 (ref. 1), PI-402A (0-600 psig) on MCB Panel XCP-6109 (ref. 2), or computer point U6019 (AVG RCS PRESSURE -NR OR WR) (ref. 3).Generic This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed.This EAL provides a pressure-based indication of RCS heat-up.Escalation of the ECL would be via IC CS1 or AS1.VCSNS Basis Reference(s):
- 1. 201-324 Main Control Board Instrumentation Control Panel XCP-6108 2. 201-325 Main Control Board Instrumentation Control Panel XCP-6109 3. STP-1 03.001 Reactor Coolant System and Pressurizer Heatup/Cooldown Surveillance Page 140 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]4. NEI 99-01 CA3 Page 141 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
C -Cold Shutdown / Refueling System Malfunction Subcategory:
4 -Loss of Vital DC Power Initiating Condition:
Loss of Vital DC power for 15 minutes or longer.EAL: CU4.1 Unusual Event< 108 VDC on required DC buses (Train A or Train B vital 125 VDC system) for -15 min.(Note 1)Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability:
5 -Cold Shutdown, 6 -Refueling Definition(s):
SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary;(2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis: Plant-Specific The fifteen minute interval is intended to exclude transient or momentary power losses.Class 1 E 125 VDC power consists of two separate vital main distribution panels. These panels are DPN-1 HA and DPN-1 HB for the Train A and Train B vital 125 VDC systems Page 142 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E](ref. 1). They are both located on the 412' level of the Intermediate Building.
Each main panel is supplied DC power through a battery charger (XBC-1 A and XBC-1 B) and is backed up by a 60 cell, lead-acid storage battery (ref. 2).Minimum DC bus voltage is 108 VDC (ref. 3, 4). MCB annunciators XCP-636 4-6 and XCP-637 4-6 (DC SYS OVRVOLT/UNDRVOLT) signal low Train A and Train B voltage at 126 VDC (ref. 5, 6). Train A and Train B voltage may be monitored on MCB Panel XCP-6116 voltmeters (ref. 7).This EAL is the cold condition equivalent of the hot condition loss of DC power EAL SS2.1.Generic This IC addresses a loss of Vital DC power which compromises the ability to monitor and control operable SAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode. In these modes, the core decay heat load has been significantly reduced, and coolant system temperatures and pressures are lower; these conditions increase the time available to restore a vital DC bus to service. Thus, this condition is considered to be a potential degradation of the level of safety of the plant.As used in this EAL, "required" means the Vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment.
For example, if Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency classification.
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Depending upon the event, escalation of the ECL would be via IC CAl or CA3, or an IC in Recognition Category R.VCSNS Basis Reference(s):
- 1. FSAR Figure 8.3-2aa 2. FSAR Section 8.3.2.1 3. EOP-6.0 Loss of All ESF AC Power Page 143 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]4. FSAR Section 8.3.2.1.3 5. ARP-001-XCP-636 Annunciator Point 4-6 6. ARP-001-XCP-637 Annunciator Point 4-6 7. 201-332 Main Control Board Instrumentation Control Panel XCP-6116 8. NEI 99-01 CU4 Page 144 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
C -Cold Shutdown / Refueling System Malfunction Subcategory:
5 -Loss of Communications Initiating Condition:
Loss of all onsite or offsite communications capabilities.
EAL: CU5.1 Unusual Event Loss of all Table C-4 onsite communication methods OR Loss of all Table C-4 ORO/NRC communication methods Table C-4 Communication Methods System Onsite ORO/NRC Gai-Tronics system X Radio system X Internal Telephone system X Telephone land lines X X Fiberoptic links X Satellite phone system X Federal Telephone System (ENS) X ESSX X Mode Applicability:
5 -Cold Shutdown, 6 -Refueling, D -Defueled Definition(s):
None Basis: Plant-Specific Page 145 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]The Table C-4 list for onsite communications loss encompasses the loss of all means of routine communications (e.g., commercial and internal telephones, page party system (Gai-Tronics) and radios) (ref. 1, 2, 3).The Table C-4 list for offsite (ORO/NRC) communications loss encompasses the loss of all means of communications with offsite authorities.
This includes the FTS (ENS), commercial telephone lines and dedicated phone systems (fiberoptic and satellite) (ref. 1, 2,3).This EAL is the cold condition equivalent of the hot condition EAL SU7.1.Generic This IC addresses a significant loss of on-site or offsite communications capabilities.
While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC.This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations.
The second EAL condition addresses a total loss of the communications methods used to notify all OROs of an emergency declaration.
The OROs referred to here are the State, Fairfield, Newberry, Lexington and Richland County EOCs as well as the NRC.VCSNS Basis Reference(s):
- 1. FSAR 9.5.2 2. EP-100 Radiation Emergency Plan, Section 7.5 3. EP-100 Radiation Emergency Plan, Figure 7-2 4. NEI 99-01 CU5 Page 146 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Cate-gory H -Hazards and Other Conditions Affectinq Plant Safety EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)Hazards are non-plant system-related events that can directly or indirectly affect plant operation, reactor plant safety or personnel safety.The events of this category pertain to the following subcategories:
- 1. Security Unauthorized entry attempts into the PROTECTED AREA, bomb threats, sabotage attempts, and actual security compromises threatening loss of physical control of the plant.2. Seismic Event Natural events such as earthquakes have potential to cause plant structure or equipment damage of sufficient magnitude to threaten personnel or plant safety.3. Natural or Technological Hazard Other natural and non-naturally occurring events that can cause damage to plant facilities include tornados, FLOODING, hazardous material releases and events restricting site access warranting classification.
- 4. Fire or Explosion FIRES and EXPLOSIONS can pose significant hazards to personnel and reactor safety. Appropriate for classification are FIRES or EXPLOSIONS within the site PROTECTED AREA or which may affect operability of equipment needed for safe shutdown 5. Hazardous Gas Toxic, corrosive, asphyxiant or flammable gas leaks can affect normal plant operations or preclude access to plant areas required to safely shutdown the plant.Page 147 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]6. Control Room Evacuation Events that are indicative of loss of Control Room habitability.
If the Control Room must be evacuated, additional support for monitoring and controlling plant functions is necessary through the emergency response facilities.
- 7. ED Judgment The EALs defined in other categories specify the predetermined symptoms or events that are indicative of emergency or potential emergency conditions and thus warrant classification.
While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and subsequent implementation of the Emergency Plan, a provision for classification of emergencies based on operator/management experience and judgment is still necessary.
The EALs of this category provide the Emergency Director the latitude to classify emergency conditions consistent with the established classification criteria based upon Emergency Director judgment.Page 148 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
H -Hazards and Other Conditions Affecting Plant Safety Subcategory:
1 -Security Initiating Condition:
Confirmed SECURITY CONDITION or threat.EAL: HU1.1 Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by Security Team Leader OR Notification of a credible security threat directed at the site OR A validated notification from the NRC providing information of an aircraft threat Mode Applicability:
All Definition(s):
SECURITY CONDITION
-Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A security condition does not involve a hostile action.HOSTILE ACTION -An act toward VCSNS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on VCSNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA).Basis: Plant-Specific If the Security Team Leader determines that a threat notification is credible, the Security Team Leader will notify the Shift Supervisor that a "Credible Threat" condition exists for Page 149 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]VCSNS. Generally, VCSNS procedures address standard practices for determining credibility.
The three main criteria for determining credibility are: technical feasibility, operational feasibility, and resolve. For VCSNS, a validated notification delivered by the FBI, NRC or similar agency is treated as credible (ref. 1, 2).Generic This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1, HS1 and HG1.Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations.
Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].The first threshold references the Security Team Leader because these are the individuals trained to confirm that a security event is occurring or has occurred.
Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR § 2.390 information.
The second threshold addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with the VCSNS Security Plan (ref. 1).The third threshold addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft.
The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with the VCSNS Security Plan (ref. 1).Page 150 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information.
This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location.
Security-sensitive information should be contained in non-public documents such as the Security Plan.Escalation of the ECL would be via IC HAl.VCSNS Basis Reference(s):
- 1. Virgil C. Summer Nuclear Station Security Plan 2. SPP-1 18 Security Notification
- 3. NEI 99-01 HU1 Page 151 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
H -Hazards and Other Conditions Affecting Plant Safety Subcategory:
1 -Security Initiating Condition:
HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes.EAL: HA1.1 Alert A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the Security Team Leader OR A validated notification from NRC of an aircraft attack threat within 30 min. of the site Mode Applicability:
All Definition(s):
HOSTILE ACTION -An act toward VCSNS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on VCSNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA).OWNER CONTROLLED AREA -Area between the vehicle barrier system and the PROTECTED AREA barrier.Basis: Plant-Specific The OWNER CONTROLLED AREA is depicted in Drawing SS-024-019, Site Plan (ref. 1).Generic This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require Page 152 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact.Timely and accurate communications between the Security Team Leader and the Control Room is essential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering).
The Alert declaration will also heighten the awareness of Offsite Response Organizations (OROs), allowing them to be better prepared should it be necessary to consider further actions.This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.The first threshold is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. This includes any action directed against an ISFSI that is located outside the plant PROTECTED AREA.The second threshold addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state of readiness.
This EAL is met when the threat-related information has been validated in accordance with SPP-1 18 Security Notification (ref. 3).Page 153 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft.
The status and size of the plane may be provided by NORAD through the NRC.In some cases, it may not be readily apparent if an aircraft impact within the OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one based on other ICs/EALs, should not be unduly delayed while awaiting notification by a Federal agency.Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information.
This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location.
Security-sensitive information should be contained in non-public documents such as the Security Plan.Escalation of the ECL would be via IC HS1.VCSNS Basis Reference(s):
- 1. Drawing SS-024-019 Site Plan 2. Virgil C. Summer Nuclear Station Security Plan 3. SPP-1 18 Security Notification
- 4. NEI 99-01 HA1 Page 154 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
H -Hazards and Other Conditions Affecting Plant Safety Subcategory:
1 -Security Initiating Condition:
HOSTILE ACTION within the PROTECTED AREA EAL: HS1.1 Site Area Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Team Leader Mode Applicability:
All Definition(s):
HOSTILE ACTION -An act toward VCSNS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on VCSNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA).PROTECTED AREA -An area encompassed by physical barriers and to which access is controlled.
The PROTECTED AREA refers to the designated security area around the process buildings and is depicted in Drawing SS-024-019 Site Plan (ref. 1).Basis: Plant-Specific None Generic This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA.This event will require rapid response and assistance due to the possibility for damage to plant equipment.
Page 155 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Timely and accurate communications between Security Team Leader and the Control Room is essential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering).
The Site Area Emergency declaration will mobilize Offsite Response Organization (ORO) resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions.
This IC does not apply to a HOSTILE ACTION directed at an ISFSI PROTECTED AREA located outside the plant PROTECTED AREA; such an attack should be assessed using IC HAl. It also does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information.
This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location.
Security-sensitive information should be contained in non-public documents such as the VCSNS Security Plan (ref. 2).Escalation of the ECL would be via IC HG1.VCSNS Basis Reference(s):
- 1. Drawing SS-024-019 Site Plan 2. Virgil C. Summer Nuclear Station Security Plan 3. NEI 99-01 HS1 Page 156 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category: Subcategory:
Initiating Condition:
EAL: H -Hazards and Other Conditions Affecting Plant Safety 1 -Security HOSTILE ACTION resulting in loss of physical control of the facility HG1.1 General Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Team Leader AND EITHER of the following has occurred: Any of the following safety functions cannot be controlled or maintained" Reactivity control" Core cooling" RCS heat removal OR Damage to spent fuel has occurred or is IMMINENT Mode Applicability:
All Definition(s):
HOSTILE ACTION -An act toward VCSNS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on VCSNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA).PROTECTED AREA -An area encompassed by physical barriers and to which access is controlled.
The Protected Area refers to the designated security area around the process buildings and is depicted in Drawing SS-024-019 Site Plan (ref. 1).Page 157 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]IMMINENT-The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.Basis: Plant-Specific None Generic This IC addresses an event in which a HOSTILE FORCE has taken physical control of the facility to the extent that the plant staff can no longer operate equipment necessary to maintain key safety functions.
It also addresses a HOSTILE ACTION leading to a loss of physical control that results in actual or IMMINENT damage to spent fuel due to 1) damage to a spent fuel pool cooling system (e.g., pumps, heat exchangers, controls, etc.) or, 2)loss of spent fuel pool integrity such that sufficient water level cannot be maintained.
Timely and accurate communications between Security Team Leader and the Control Room is essential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information.
This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location.
Security-sensitive information should be contained in non-public documents such as the VCSNS Security Plan (ref. 2).VCSNS Basis Reference(s):
- 1. Drawing SS-024-019 Site Plan 2. Virgil C. Summer Nuclear Station Security Plan 3. NEI 99-01 HG1 Page 158 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category: Subcategory:
Initiating Condition:
EAL: H -Hazards and Other Conditions Affecting Plant Safety 2 -Seismic Event Seismic event greater than OBE levels HU2.1 Unusual Event Seismic event > OBE as indicated by EITHER:* Triaxial Seismic Switch MCB annunciator XCP-638 3-5 (RB FOUND SEIS SWITCH OBE EXCEED)" Any red OBE light on the Triaxial Response Spectrum Recorder Mode Applicability:
All Definition(s):
None Basis: Plant-Specific The instrumentation used to indicate a seismic event greater than the Operational Basis Earthquake (OBE) includes the Triaxial Seismic Switch and the Triaxial Response Spectrum Recorder.
The specified annunciator, XCP-638 3-5 (RB FOUND SEIS SWITCH OBE EXCEED), is sounded in the Control Room whenever the Triaxial Seismic Switch senses the OBE plant design level of 0.1 Og for the horizontal directions or 0.067g for the vertical direction.
The Response Spectrum Recorder located on the RB foundation mat is connected to the Response Spectrum Annunciator located on the far right side of the Main Control Board in the Control Room. This Annunciator has a set of yellow and red lights which relate to each of the 12 frequencies in each orthogonal direction, with the yellow lights connected to switches set at 2/3 of the OBE level and the red lights connected to switches set at the OBE level. (ref. 1, 2, 3)Confirmation of the seismic event is not included in the EAL, however, to avoid inappropriate emergency classification resulting from spurious actuation of the seismic Page 159 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT El instrumentation, an offsite agency (USGS, National Earthquake Information Center) can confirm that an earthquake has occurred in the area of the plant. Such confirmation should not, however, preclude a timely emergency declaration.
The NEIC can be contacted by calling (303) 273-8500.
Select option #1 and inform the analyst you wish to confirm recent seismic activity in the vicinity of VCSNS. Provide the analyst with the following VCSNS coordinates:
34 deg. 17 min. 54.1 sec. north latitude, 81 deg. 18 min.54.6 sec. west longitude (ref. 4, 5). Alternatively, near real-time seismic activity can be accessed via the NEIC website: http://earthquake.
usgs.gov/eqcenterI Generic This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE). An earthquake greater than an OBE but less than a Safe Shutdown Earthquake (SSE) should have no significant impact on safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk-downs and post-event inspections).
Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant.Event verification with external sources should not be necessary during or following an OBE. Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event (e.g., lateral accelerations in excess of 0.1Og). The Shift Supervisor or Emergency Director may seek external verification if deemed appropriate (e.g., a call to the USGS, check internet news sources, etc.); however, the verification action must not preclude a timely emergency declaration.
Depending upon the plant mode at the time of the event, escalation of the ECL would be via IC CA6 or SA9.VCSNS Basis Reference(s):
- 1. ES-426 Earthquake Response Procedure 2. FSAR Section 3.7 Page 160 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT El 3. EPP-015 Natural Emergency 4. FSAR Section 2.1.1 5. ARP-001-XCP-638
- 6. NEI 99-01 HU2 Page 161 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
H -Hazards and Other Conditions Affecting Plant Safety Subcategory:
2 -Seismic Event Initiating Condition:
Seismic event affecting a SAFETY SYSTEM needed for the current operating mode EAL: HA2.1 Alert Seismic event resulting in EITHER of the following: " Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode" The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode Mode Applicability:
All Definition(s):
SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary;(2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
VISIBLE DAMAGE -Damage to a component or structure that is readily observable without measurements, testing, or analysis.
The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.
Page 162 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT El Basis: Plant-Specific The significance of seismic events are discussed under EAL HU2.1 (ref. 1).Generic This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant.The first conditional addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available.
The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.The second conditional addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components.
Operators will make this determination based on the totality of available event and damage report information.
This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage.Escalation of the ECL would be via IC CS1 or RS1.VCSNS Basis Reference(s):
- 1. ES-426 Earthquake Response Procedure 2. NEI 99-01 CA6 Page 163 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
H -Hazards and Other Conditions Affecting Plant Safety Subcategory:
3 -Natural or Technological Hazard Initiating Condition:
Hazardous event EAL: HU3.1 Unusual Event A tornado strike within the PROTECTED AREA Mode Applicability:
All Definition(s):
PROTECTED AREA -An area encompassed by physical barriers and to which access is controlled.
The Protected Area refers to the designated security area around the process buildings and is depicted in Drawing SS-024-019 Site Plan (ref. 1).Basis: Plant-Specific Weather information may be received from Impact Weather, a contract weather reporting service for SCE&G. This information can be received by telephone or log in on the website at www.impactweather.com (ref. 2, 3).Generic This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.EAL HU3.1 addresses a tornado striking (touching down) within the PROTECTED AREA.Escalation of the ECL would be based on ICs in Recognition Categories R, F, S or C.VCSNS Basis Reference(s):
- 1. Drawing SS-024-019 Site Plan 2. OAP-109.1 Guidelines for Severe Weather 3. EPP-015 Natural Emergency 4. NEI 99-01 HU3 Page 164 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
H -Hazards and Other Conditions Affecting Plant Safety Subcategory:
3 -Natural or Technological Hazard Initiating Condition:
Hazardous event EAL: HU3.2 Unusual Event Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode Mode Applicability:
All Definition(s):
FLOODING -A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary;(2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis: Plant-Specific Refer to VCSNS IPE Internal FLOODING Analysis Workbook to identify susceptible internal Flooding Areas (ref. 1).Generic Page 165 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.This EAL addresses FLOODING of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns.
Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode.Escalation of the ECL would be based on ICs in Recognition Categories R, F, S or C.VCSNS Basis Reference(s):
- 1. VCSNS IPE Internal Flooding Analysis Workbook 2. NEI 99-01 HU3 Page 166 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
H -Hazards and Other Conditions Affecting Plant Safety Subcategory:
3 -Natural or Technological Hazard Initiating Condition:
Hazardous event EAL: HU3.3 Unusual Event Movement of personnel within the PROTECTED AREA is impeded due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release)Mode Applicability:
All Definition(s):
PROTECTED AREA -An area encompassed by physical barriers and to which access is controlled.
The Protected Area refers to the designated security area around the process buildings and is depicted in Drawing SS-024-019 Site Plan (ref. 1).Basis: Plant-Specific As used here the term "offsite" is meant to be areas external to the VCSNS PROTECTED AREA.EPP-014 Toxic Release (ref. 2) provides additional information on hazardous substances and spills.The HazMat Risk Assessment is a document that provides guidance on the hazards that are in areas and facilities of the plant site, individual chemical information, and Emergency Response guidelines.
The HazMat Risk Assessment lists quantities and response guidelines for release of these chemicals (ref. 3).Generic This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.Page 167 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]This EAL addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA.Escalation of the ECL would be based on ICs in Recognition Categories R, F, S or C.VCSNS Basis Reference(s):
- 1. Drawing SS-024-019 Site Plan 2. EPP-014 Toxic Release 3. HazMat Risk Assessment
- 4. NEI 99-01 HU3 Page 168 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
H -Hazards and Other Conditions Affecting Plant Safety Subcategory:
3 -Natural or Technological Hazard Initiating Condition:
Hazardous event EAL: HU3.4 Unusual Event A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles (Note 10)Note 10: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.
Mode Applicability:
All Definition(s):
None Basis: Plant-Specific None Generic This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.This EAL addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles.
Examples of such an event include site FLOODING caused by a hurricane, heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road.This EAL is not intended apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011.Page 169 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Escalation of the ECL would be based on ICs in Recognition Categories R, F, S or C.VCSNS Basis Reference(s):
- 1. NEI 99-01 HU3 Page 170 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
H -Hazards and Other Conditions Affecting Plant Safety Subcategory:
3 -Natural or Technological Hazard Initiating Condition:
Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode EAL: HA3.1 Alert The occurrence of any Table H-1 hazardous event resulting in EITHER of the following: " Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode" The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode Table H-1 Hazardous Events* Internal or external FLOODING event* High winds or tornado strike* Other events with similar hazard characteristics as determined by the Shift Supervisor Mode Applicability:
All Definition(s):
FLOODING -A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary;Page 171 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E](2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
VISIBLE DAMAGE -Damage to a component or structure that is readily observable without measurements, testing, or analysis.
The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.
Basis: Plant-Specific
- Internal FLOODING may be caused by events such as component failures, equipment misalignment, or outage activity mishaps (ref. 1).* Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 100 mph (sustained). (ref. 2).Generic This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant.The first conditional addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available.
The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.The second conditional addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components.
Operators will make this determination based Page 172 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]on the totality of available event and damage report information.
This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage.Escalation of the ECL would be via IC CS1 or RS1.VCSNS Basis Reference(s):
- 1. VCSNS IPE Internal Flooding Analysis Workbook 2. FSAR Section 3.3.1 3. NEI 99-01 CA6 Page 173 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
H -Hazards and Other Conditions Affecting Plant Safety Subcategory:
4- Fire Initiating Condition:
FIRE potentially degrading the level of safety of the plant EAL: HU4.1 Unusual Event A FIRE is NOT extinguished within 15 min. of any of the following FIRE detection indications (Note 1):* Report from the field (i.e., visual observation)" Receipt of multiple (more than 1) fire alarms or indications" Field verification of a single fire alarm AND The FIRE is located within any Table H-2 area Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table H-2 Fire Areas* Reactor Building* Auxiliary Building* Control Building* Fuel Handling Building* Intermediate Building* Diesel Generator Building" Turbine Building" Service water Pumphouse* Safe Shutdown Yard Areas:* RWST C OST* DG Fuel Oil Storage Mode Applicability:
All Definition(s):
Page 174 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.Basis: Plant-Specific VCSNS Fire Protection Evaluation Report, Section 4.0 "Hazards Analysis" was used to identify areas (Table H-2) containing functions and systems required for safe shutdown of the plant (ref. 1).Generic This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.For EAL HU4.1 the intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket).In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc.Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed.
Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report.Depending upon the plant mode at the time of the event, escalation of the ECL would be via IC CA6 or SA9.VCSNS Basis Reference(s):
- 1. VCSNS Fire Protection Evaluation Report 2. NEI 99-01 HU4 Page 175 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
H -Hazards and Other Conditions Affecting Plant Safety Subcategory:
4- Fire Initiating Condition:
FIRE potentially degrading the level of safety of the plant EAL: HU4.2 Unusual Event Receipt of a single fire alarm (i.e., no other indications of a FIRE)AND The fire alarm is indicating a FIRE within any Table H-2 area AND The existence of a FIRE is not verified within 30 min. of alarm receipt (Note 1)Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table H-2 Fire Areas" Reactor Building" Auxiliary Building" Control Building" Fuel Handling Building* Intermediate Building* Diesel Generator Building" Turbine Building" Service water Pumphouse* Safe Shutdown Yard Areas:* RWST* CST* DG Fuel Oil Storage Mode Applicability:
All Definition(s):
Page 176 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.Basis: Plant-Specific VCSNS Fire Protection Evaluation Report, Section 4.0 "Hazards Analysis" was used to identify areas (Table H-2) containing functions and systems required for safe shutdown of the plant (ref. 1).Generic This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed.
A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress.If an actual FIRE is verified by a report from the field, then HU4.1 is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted.
Depending upon the plant mode at the time of the event, escalation of the ECL would be via IC CA6 or SA9.Page 177 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]VCSNS Basis Reference(s):
- 1. VCSNS Fire Protection Evaluation Report 2. NEI 99-01 HU4 Page 178 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
H -Hazards and Other Conditions Affecting Plant Safety Subcategory:
4- Fire Initiating Condition:
FIRE potentially degrading the level of safety of the plant EAL: HU4.3 Unusual Event A FIRE within the plant PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1)Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability:
All Definition(s):
FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.PROTECTED AREA -An area encompassed by physical barriers and to which access is controlled.
The Protected Area refers to the designated security area around the process buildings and is depicted in Drawing SS-024-019 Site Plan (ref. 1).Basis: Plant-Specific None Generic This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.In addition to a FIRE addressed by EAL HU4.1 or HU4.2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety.Page 179 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Depending upon the plant mode at the time of the event, escalation of the ECL would be via IC CA6 or SA9.VCSNS Basis Reference(s):
- 1. Drawing SS-024-019 Site Plan 2. NEI 99-01 HU4 Page 180 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
H -Hazards and Other Conditions Affecting Plant Safety Subcategory:
4 -Fire Initiating Condition:
FIRE potentially degrading the level of safety of the plant EAL: HU4.4 Unusual Event A FIRE within the plant PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish Mode Applicability:
All Definition(s):
FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.PROTECTED AREA -An area encompassed by physical barriers and to which access is controlled.
The Protected Area refers to the designated security area around the process buildings and is depicted in Drawing SS-024-019 Site Plan (ref. 1).Basis: Plant-Specific None Generic This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.If a FIRE within the plant PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded.
The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the FIRE is beyond the capability of the Fire Brigade to extinguish.
Declaration is Page 181 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions.Depending upon the plant mode at the time of the event, escalation of the ECL would be via IC CA6 or SA9.VCSNS Basis Reference(s):
- 1. Drawing SS-024-019 Site Plan 2. NEI 99-01 HU4 Page 182 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
H -Hazards and Other Conditions Affecting Plant Safety Subcategory:
4 -FIRE Initiating Condition:
FIRE or EXPLOSION event affecting a SAFETY SYSTEM needed for the current operating mode EAL: HA4.1 Alert FIRE or EXPLOSION resulting in EITHER of the following:
- Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode* The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode Mode Applicability:
All Definition(s):
EXPLOSION
-A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization.
A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an EXPLOSION.
Such events require a post-event inspection to determine if the attributes of an EXPLOSION are present.FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: Page 183 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E](1) The integrity of the reactor coolant pressure boundary;(2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
VISIBLE DAMAGE -Damage to a component or structure that is readily observable without measurements, testing, or analysis.
The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.
Basis: Plant-Specific" Refer to VCSNS Fire Protection Evaluation Report, Section 4.0 "Hazards Analysis" to identify areas containing functions and systems required for safe shutdown of the plant (ref. 4)" An EXPLOSION (including a steam line explosion) that degrades the performance of a SAFETY SYSTEM train or visibly damages a SAFETY SYSTEM component or structure would be classified under this EAL. The need to classify a steam line break not considered an EXPLOSION itself is considered in fission product barrier degradation monitoring (EAL Category F).Generic This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant.The first conditional addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available.
The indications of Page 184 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.The second conditional addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components.
Operators will make this determination based on the totality of available event and damage report information.
This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage.Escalation of the ECL would be via IC CS1 or RS1.VCSNS Basis Reference(s):
- 1. VCSNS Fire Protection Evaluation Report 2. NEI 99-01 CA6 Page 185 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
H -Hazards and Other Conditions Affecting Plant Safety Subcategory:
5 -Hazardous Gases Initiating Condition:
Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown EAL: HA5.1 Alert Release of a toxic, corrosive, asphyxiant or flammable gas into any Table H-3 area AND Entry into the area is prohibited or impeded (Note 6)Note 6: If the equipment in the listed area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted.
Table H-3 Safe Operation
& Shutdown Areas Area Mode Applicability Auxiliary Building 374' 3 Auxiliary Building 388' 3, 4, 5 Auxiliary Building 400' 4, 5 Auxiliary Building 412 3, 4, 5 Auxiliary Building 436' 1,2, 3, 4, 5 Auxiliary Building 463' 3, 4, 5 Intermediate Building 412' 3 Intermediate Building 436' 4, 5 Intermediate Building 463' 3, 4, 5 Control Building 412' 2, 3 Control Building 436' 3, 4, 5 Turbine Building (All levels) 1, 2 Mode Applicability:
All Definition(s):
None Page 186 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Basis: Plant-Specific The Table H-3 safe operation and shutdown areas (with entry-related mode applicability) are those plant areas that contain equipment which require a manual/local action as specified in general operating procedures (and procedures referenced by them) used for normal plant operation, cooldown and shutdown.
The list specifies the plant operating modes during which entry would be required for each area and thus specifying when a loss of access or impeded access is applicable to this EAL (ref. 1).Plant areas where actions of a contingent or emergency nature might be needed to be performed. (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) were not considered for inclusion.
Additionally, areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections) were not considered for inclusion.
Refer to Attachment 4 "Safe Operation
& Shutdown Areas Tables R-2 & H-3 Bases.".If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.Generic This IC addresses an event involving a release of a hazardous gas that precludes or impedes access to equipment necessary to maintain normal plant operation, or required for a normal plant cooldown and shutdown.
This condition represents an actual or potential substantial degradation of the level of safety of the plant.An Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the gaseous release. The emergency classification is not contingent upon whether entry is actually necessary at the time of the release.Page 187 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Evaluation of the IC and EAL do not require atmospheric sampling; it only requires the Emergency Director's judgment that the gas concentration in the affected room/area is sufficient to preclude or significantly impede procedurally required access. This judgment may be based on a variety of factors including an existing job hazard analysis, report of ill effects on personnel, advice from a subject matter expert or operating experience with the same or similar hazards. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).
An emergency declaration is not warranted if any of the following conditions apply.* The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the gaseous release).
For example, the plant is in Mode 1 when the gaseous release occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4." The gas release is a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., fire suppression system testing)." The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections)." The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action.* If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed Page 188 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]environment.
This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death.This EAL does not apply to firefighting activities that automatically or manually activate a fire suppression system in an area.Escalation of the ECL would be via Recognition Category R, C or F ICs.VCSNS Basis Reference(s):
- 1. EPP-108 Emergency Action Level Technical Bases Attachment 4 "Safe Operation
&Shutdown Areas Tables R-2 & H-3 Bases." 2. NEI 99-01 HA5 Page 189 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category: Subcategory:
Initiating Condition:
H -Hazards and Other Conditions Affecting Plant Safety 6- Control Room Evacuation Control Room evacuation resulting in transfer of plant control to alternate locations EAL: HA6.1 Alert An event has resulted in plant control being transferred from the Control Room to the Control Room Evacuation Panels (CREP)Mode Applicability:
All Definition(s):
None Basis: Plant-Specific Per AOP-600.1 Control Room Evacuation (ref. 1) plant control is established at the CREP when: " Emergency boration capability exists, if required* Charging and letdown flow can be controled to maintain Pressurizer level." EFW flow can be controlled to maintain SG levels.* RCS natural circulation can be established.
Generic This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. The loss of the ability to control the plant from the Control Room is considered to be a potential substantial degradation in the level of plant safety.Page 190 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown locations.
The necessity to control a plant shutdown from outside the Control Room, in addition to responding to the event that required the evacuation of the Control Room, will present challenges to plant operators and other on-shift personnel.
Activation of the ERO and emergency response facilities will assist in responding to these challenges.
Escalation of the ECL would be via IC HS6.VCSNS Basis Reference(s):
- 1. AOP-600.1 Control Room Evacuation
- 2. FEP-4.0 Control Room Evacuation Due To Fire.3. FSAR Section 7.4.1.3 4. NEI 99-01 HA6 Page 191 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT El Category:
H -Hazards and Other Conditions Affecting Plant Safety Subcategory:
6- Control Room Evacuation Initiating Condition:
Inability to control a key safety function from outside the Control Room EAL: HS6.1 Site Area Emergency An event has resulted in plant control being transferred from the Control Room to the Control Room Evacuation Panels (CREP)AND Control of any of the following key safety functions is not reestablished within 15 min.(Note 1): " Reactivity control" Core cooling" RCS heat removal Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability:
All Definition(s):
None Basis: Plant-Specific Per AOP-600.1 Control Room Evacuation (ref. 1) plant control is established at the CREP when: , Emergency boration capability exists, if required* Charging and letdown flow can be controled to maintain Pressurizer level.* EFW flow can be controlled to maintain SG levels." RCS natural circulation can be established.
Page 192 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Generic This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time.The determination of whether or not "control" is established at the remote safe shutdown location(s) is based on Emergency Director judgment.
The Emergency Director is expected to make a reasonable, informed judgment within 15 minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s).
Escalation of the ECL would be via IC FG1 or CG1 VCSNS Basis Reference(s):
- 1. AOP-600.1 Control Room Evacuation
- 2. FEP-4.0 Control Room Evacuation Due To Fire.3. FSAR Section 7.4.1.3 4. NEI 99-01 HS6 Page 193 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category: H -Hazards and Other Conditions Affecting Plant Safety Subcategory:
7 -Judgment Initiating Condition:
Other conditions existing that in the judgment of the Emergency Director warrant declaration of a UE EAL: HU7.1 Unusual Event Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.
No releases of radioactive material requiring off site response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs.Mode Applicability:
All Definition(s):
None Basis: Plant-Specific None Generic This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the ECL description for a Unusual Event.VCSNS Basis Reference(s):
- 1. NEI 99-01 HU7 Page 194 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
H -Hazards and Other Conditions Affecting Plant Safety Subcategory:
7 -Judgment Initiating Condition:
Other conditions exist that in the judgment of the Emergency Director warrant declaration of an Alert EAL: HA7.1 Alert Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.Mode Applicability:
All Definition(s):
HOSTILE ACTION -An act toward VCSNS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on VCSNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA).Basis: Plant-Specific None Page 195 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Generic This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the ECL description for an Alert.VCSNS Basis Reference(s):
- 1. NEI 99-01 HA7 Page 196 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category: Subcategory:
H -Hazards and Other Conditions Affecting Plant Safety 7 -Judgment Initiating Condition:
Other conditions existing that in the judgment of the Emergency Director warrant declaration of a Site Area Emergency EAL: HS7.1 Site Area Emergency Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY.Mode Applicability:
All Definition(s):
HOSTILE ACTION -An act toward VCSNS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on VCSNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA)Basis: Plant-Specific None Page 197 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Generic This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the ECL description for Site Area Emergency.
VCSNS Basis Reference(s):
- 1. NEI 99-01 HS7 Page 198 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
H -Hazards and Other Conditions Affecting Plant Safety Subcategory:
7 -Judgment Initiating Condition:
Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency EAL: HG7.1 General Emergency Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.
Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.Mode Applicability:
All Definition(s):
HOSTILE ACTION -An act toward VCSNS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on VCSNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA).IMMINENT-The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.Basis: Plant-Specific None Page 199 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Generic This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the ECL description for a General Emergency.
VCSNS Basis Reference(s):
- 1. NEI 99-01 HG7 Page 200 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category S -System Malfunction EAL Group: Hot Conditions (RCS temperature
> 200 0 F);EALs in this category are applicable only in one or more hot operating modes.Numerous system-related equipment failure events that warrant emergency classification have been identified in this category.
They may pose actual or potential threats to plant safety.The events of this category pertain to the following subcategories:
- 1. Loss of Engineered Safeauards Features (ESF) AC Power Loss of ESF plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity.
This category includes loss of onsite and offsite power sources for 7.2 KV safeguards buses 1 DA and 1 DB.2. Loss of Vital DC Power Loss of emergency plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity.
This category includes loss of power to or degraded voltage on the 125VDC safeguards buses.3. Loss of Control Room Indications Certain events that degrade plant operator ability to effectively assess plant conditions within the plant warrant emergency classification.
Losses of indicators are in this subcategory.
- 4. RCS Activity During normal operation, reactor coolant fission product activity is very low. Small concentrations of fission products in the coolant are primarily from the fission of tramp uranium in the fuel clad or minor perforations in the clad itself. Any significant increase from these base-line levels (2% -5% clad failures) is indicative of fuel failures and is Page 201 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]covered under Category F, Fission Product Barrier Degradation.
However, lesser amounts of clad damage may result in coolant activity exceeding Technical Specification limits. These fission products will be circulated with the reactor coolant and can be detected by coolant sampling.5. RCS Leakaqe The reactor vessel provides a volume for the coolant that covers the reactor core. The reactor vessel and associated pressure piping (reactor coolant system) together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail. Excessive RCS leakage greater than Technical Specification limits indicates potential pipe cracks that may propagate to an extent threatening fuel clad, RCS and Containment integrity.
- 6. RTS Failure This subcategory includes events related to failure of the Reactor Trip System (RTS) to initiate and complete reactor trips. In the plant licensing basis, postulated failures of the RTS to complete a reactor trip comprise a specific set of analyzed events referred to as Anticipated Transient Without Trip (ATWS) events. For EAL classification however, ATWS is intended to mean any trip failure event that does not achieve reactor shutdown.
If RTS actuation fails to assure reactor shutdown, positive control of reactivity is at risk and could cause a threat to fuel clad, RCS and Containment integrity.
- 7. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.
- 8. Containment Isolation Failure Failure of containment isolation capability (under conditions in which the containment is not currently challenged) warrants emergency classification.
- 9. Hazardous Event Affecting Safety Systems Page 202 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Various natural and technological events that result in degraded plant SAFETY SYSTEM performance or significant VISIBLE DAMAGE warrant emergency classification under the sub-category.
Page 203 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category: Subcategory:
Initiating Condition:
S -System Malfunction 1 -Loss of ESF AC Power Loss of all offsite AC power capability to ESF buses for 15 minutes or longer.EAL: SUl.1 Unusual Event Loss of all offsite AC power (Table S-1) capability to 7.2 KV ESF buses 1 DA and 1 DB for >15 min. (Note 1)Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table S-1 AC Power Supplies Offsite:* 115 KV power to XTF-4 and XTF-5* 230 KV power to XTF-31* Parr Hydro Plant 13.8 KV power to ESF bus 1DA or 1DB Onsite: " Diesel Generator A" Diesel Generator B Mode Applicability:
1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown, Definition(s):
None Basis: Plant-Specific As used in this EAL the term "capability" means an AC power source is either currently powering essential loads on one or more 7.2 KV ESF buses or is capable of energizing and powering essential loads on at least one 7.2 KV ESF bus within 15 min.Page 204 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Table S-1 lists AC sources capable of powering ESF buses. Safeguards power originates offsite from two independent sources (ref. 1):* The Parr Generating Complex supplies 115 KV power to the two Engineered Safety Feature (ESF) transformers (XTF-4 and XTF-5). The transformer outputs are combined at 7.2 KV bus 1 DX and then supplied to 7.2 KV ESF bus 1 DA (Train A).This is the preferred or normal power source to Train A and the alternate power source for Train B.* 7.2 KV ESF bus 1 DB (Train B) is supplied from the emergency auxiliary transformer (XTF-31).
The emergency auxiliary transformer receives 230 KV power from the Virgil C. Summer substation (switchyard) bus 3. This transformer is the preferred power source for Train B and the alternate power source for Train A.The Parr Hydro Plant provides a 13.8 KV AC line to the 7.2 KV ESF buses. This Alternate AC Power Supply has the capacity to supply only one fully loaded ESF bus (ref. 4).The AAC is designed to provide back-up power to either ESF bus whenever one of the Diesel Generators is out of service, particularly in Modes 1 through 4. The AAC is verified available and an operational readiness status check is performed when it is anticipated that one of the Diesel Generators will be inoperable for longer than the allowed outage time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The design of the AAC is capable of providing the required safety and non-safety related loads in the event of a total loss of offsite power and if both Diesel Generators fail to start and load. During these events it is assumed that there is no seismic event or an event that requires safeguards actuation (e.g., safety injection, containment spray, etc.) Although the AAC is not designed for DBA loads, it is capable of supplying sufficient power to mitigate the effects of an accident.
The AAC is not credited in the safety analysis.
The AAC is, however, capable of mitigating the dominant core damage sequences and provides a significant overall risk reduction for station operation.
The AAC alone is adequate to supply electrical power to affect a safe shutdown of the plant (ref. 7).The two trains of 7.2 KV safeguards power are also provided with an onsite standby source of power for supplying power when the ESF and emergency auxiliary transformers Page 205 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]are not available.
The Diesel Generators A and B are capable of supplying all loads on the distribution network of their respective train (ref. 1, 2, 3, 4, 5).Generic This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC engineered safeguard features (ESF) buses. This condition represents a potential reduction in the level of safety of the plant.For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the ESF buses, whether or not the buses are powered from it.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power.Escalation of the ECL would be via IC SAI.VCSNS Basis Reference(s):
- 1. FSAR Section 8 2. EOP-6.0 Loss of All ESF AC Power 3. EOP-1.0 Reactor Trip/Safety Injection Actuation 4. SOP-304 115KV/7.2KV Operations
- 5. SOP-306 Emergency Diesel Generator 6. AOP-304.1 Loss of Bus 1 DA(1 DB) with the Diesel not Available 7. Technical Specifications Bases 3/4.8 8. NEI 99-01 SUW Page 206 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
S -System Malfunction Subcategory:
1 -Loss of ESF AC Power Initiating Condition:
Loss of all but one AC power source to ESF buses for 15 minutes or longer.EAL: SA1.1 Alert AC power capability to 7.2 KV ESF buses 1 DA and 1 DB reduced to a single power source (Table S-1) for -- 15 min. (Note 1)AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table S-1 AC Power Supplies Offsite:* 115 KV power to XTF-4 and XTF-5* 230 KV power to XTF-31* Parr Hydro Plant 13.8 KV power to ESF bus 1DAor 1DB Onsite: 0 Diesel Generator A 9 Diesel Generator B Mode Applicability:
1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown, Definition(s):
SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: Page 207 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E](1) The integrity of the reactor coolant pressure boundary;(2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis: Plant-Specific As used in this EAL the term "capability" means an AC power source is either currently powering essential loads on one or more 7.2 KV ESF buses or is capable of energizing and powering essential loads on at least one 7.2 KV ESF bus within 15 min.Table S-1 lists AC sources capable of powering ESF buses. Safeguards power originates off site from two independent sources (ref. 1):* The Parr Generating Complex supplies 115 KV power to the two Engineered Safety Feature (ESF) transformers (XTF-4 and XTF-5). The transformer outputs are combined at 7.2 KV bus 1 DX and then supplied to 7.2 KV ESF bus 1 DA (Train A).This is the preferred or normal power source to Train A and the alternate power source for Train B.0 7.2 KV ESF bus 1 DB (Train B) is supplied from the emergency auxiliary transformer (XTF-31).
The emergency auxiliary transformer receives 230 KV power from the Virgil C. Summer substation (switchyard) bus 3. This transformer is the preferred power source for Train B and the alternate power source for Train A.The Parr Hydro Plant provides a 13.8 KV AC line to the 7.2 KV ESF buses. This Alternate AC Power Supply has the capacity to supply only one fully loaded ESF bus (ref. 4).The AAC is designed to provide back-up power to either ESF bus whenever one of the Diesel Generators is out of service, particularly in Modes 1 through 4. The AAC is verified available and an operational readiness status check is performed when it is anticipated that one of the Diesel Generators will be inoperable for longer than the allowed outage Page 208 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The design of the AAC is capable of providing the required safety and non-safety related loads in the event of a total loss of offsite power and if both Diesel Generators fail to start and load. During these events it is assumed that there is no seismic event or an event that requires safeguards actuation (e.g., safety injection, containment spray, etc.) Although the AAC is not designed for DBA loads, it is capable of supplying sufficient power to mitigate the effects of an accident.
The AAC is not credited in the safety analysis.
The AAC is, however, capable of mitigating the dominant core damage sequences and provides a significant overall risk reduction for station operation.
The AAC alone is adequate to supply electrical power to affect a safe shutdown of the plant (ref. 7).The two trains of 7.2 KV safeguards power are also provided with an onsite standby source of power for supplying power when the ESF and emergency auxiliary transformers are not available.
The Diesel Generators A and B are capable of supplying all loads on the distribution network of their respective train (ref. 1, 2, 3, 4, 5).The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. If the capability of a second source of ESF bus power is not restored within 15 minutes, an Alert is declared under this EAL.Generic This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment.
This IC provides an escalation path from IC Sul.An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an Engineered Safeguard Features (ESF) bus. Some examples of this condition are presented below.e A loss of all offsite power with a concurrent failure of all but one ESF power source (e.g., an onsite diesel generator).
Page 209 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]* A loss of all offsite power and loss of all ESF power sources (e.g., onsite diesel generators) with a single train of ESF buses being back-fed from the unit main generator.
- A loss of ESF power sources (e.g., onsite diesel generators) with a single train of ESF buses being back-fed from an offsite power source.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.Escalation of the ECL would be via IC SS1.VCSNS Basis Reference(s):
- 1. FSAR Section 8 2. EOP-6.0 Loss of All ESF AC Power 3. EOP-1.0 Reactor Trip/Safety Injection Actuation 4. SOP-304 115KV/7.2KV Operations
- 5. SOP-306 Emergency Diesel Generator 6. AOP-304.1 Loss of Bus 1 DA(1 DB) with the Diesel not Available 7. Technical Specifications Bases 3/4.8 8. NEI 99-01 SA1 Page 210 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
S -System Malfunction Subcategory:
1 -Loss of ESF AC Power Initiating Condition:
Loss of all offsite and all onsite AC power to ESF buses for 15 minutes or longer.EAL: SS1.1 Site Area Emergency Loss of all offsite and all onsite AC power (Table S-1) capability to 7.2 KV ESF buses 1 DA and 1 DB for >- 15 min. (Note 1)Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table S-1 AC Power Supplies Offsite:* 115 KV power to XTF-4 and XTF-5* 230 KV power to XTF-31* Parr Hydro Plant 13.8 KV power to ESF bus 1DA or 1DDB Onsite: , Diesel Generator A 0 Diesel Generator B Mode Applicability:
1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):
None Basis: Plant-Specific As used in this EAL the term "capability" means an AC power source is either currently powering essential loads on one or more 7.2 KV ESF buses or is capable of energizing and powering essential loads on at least one 7.2 KV ESF bus within 15 min.Page 211 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Table S-1 lists AC sources capable of powering ESF buses. Safeguards power originates off site from two independent sources (ref. 1):* The Parr Generating Complex supplies 115 KV power to the two Engineered Safety Feature (ESF) transformers (XTF-4 and XTF-5). The transformer outputs are combined at 7.2 KV bus 1 DX and then supplied to 7.2 KV ESF bus 1 DA (Train A).This is the preferred or normal power source to Train A and the alternate power source for Train B.* 7.2 KV ESF bus 1 DB (Train B) is supplied from the emergency auxiliary transformer (XTF-31).
The emergency auxiliary transformer receives 230 KV power from the Virgil C. Summer substation (switchyard) bus 3. This transformer is the preferred power source for Train B and the alternate power source for Train A.The Parr Hydro Plant provides a 13.8 KV AC line to the 7.2 KV ESF buses. This Alternate AC Power Supply has the capacity to supply only one fully loaded ESF bus (ref. 4).The AAC is designed to provide back-up power to either ESF bus whenever one of the Diesel Generators is out of service, particularly in Modes 1 through 4. The AAC is verified available and an operational readiness status check is performed when it is anticipated that one of the Diesel Generators will be inoperable for longer than the allowed outage time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The design of the AAC is capable of providing the required safety and non-safety related loads in the event of a total loss of offsite power and if both Diesel Generators fail to start and load. During these events it is assumed that there is no seismic event or an event that requires safeguards actuation (e.g., safety injection, containment spray, etc.) Although the AAC is not designed for DBA loads, it is capable of supplying sufficient power to mitigate the effects of an accident.
The AAC is not credited in the safety analysis.
The AAC is, however, capable of mitigating the dominant core damage sequences and provides a significant overall risk reduction for station operation.
The AAC alone is adequate to supply electrical power to affect a safe shutdown of the plant (ref. 7).The two trains of 7.2 KV safeguards power are also provided with an onsite standby source of power for supplying power when the ESF and emergency auxiliary transformers Page 212 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]are not available.
The Diesel Generators A and B are capable of supplying all loads on the distribution network of their respective train (ref. 1, 2, 3, 4, 5).This EAL is the hot condition equivalent of the cold condition loss of all AC power EAL CAl.1. When in Cold Shutdown, Refueling, or Defueled mode, the event can be classified as an Alert because of the significantly reduced decay heat, lower temperature and pressure, increasing the time to restore one of the ESF buses, relative to that existing when in hot conditions.
Generic This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. In addition, fission product barrier monitoring capabilities may be degraded under these conditions.
This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Escalation of the ECL would be via ICs RG1, FG1 or SG1.VCSNS Basis Reference(s):
- 1. FSAR Section 8 2. EOP-6.0 Loss of All ESF AC Power 3. EOP-1.0 Reactor Trip/Safety Injection Actuation 4. SOP-304 115KV/7.2KV Operations
- 5. SOP-306 Emergency Diesel Generator 6. AOP-304.1 Loss of Bus 1 DA(1 DB) with the Diesel not Available 7. Technical Specifications Bases 3/4.8 8. NEI 99-01 SS1 Page 213 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
S -System Malfunction Subcategory:
1 -Loss of ESF AC Power Initiating Condition:
Prolonged loss of all offsite and all onsite AC power to ESF buses or loss of all AC and vital DC power sources for 15 minutes or longer.EAL: SG1.1 General Emergency Loss of all offsite and all onsite AC power capability to 7.2 KV ESF buses 1 DA and 1 DB (Table S-1)AND EITHER of the following:
- Restoration of at least one ESF bus within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely (Note 1)* CSFST Core Cooling-RED path conditions met Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table S-1 AC Power Supplies Offsite:* 115 KV power to XTF-4 and XTF-5* 230 KV power to XTF-31 0 Parr Hydro Plant 13.8 KV power to ESF bus 1DA or 1DDB Onsite: " Diesel Generator A" Diesel Generator B Mode Applicability:
1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):
None Basis: Page 214 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Plant-Specific As used in this EAL the term "capability" means an AC power source is either currently powering essential loads on one or more 7.2 KV ESF buses or is capable of energizing and powering essential loads on at least one 7.2 KV ESF bus within 15 min.Table S-1 lists AC sources capable of powering ESF buses. Safeguards power originates offsite from two independent sources (ref. 3):* The Parr Generating Complex supplies 115 KV power to the two Engineered Safety Feature (ESF) transformers (XTF-4 and XTF-5). The transformer outputs are combined at 7.2 KV bus 1 DX and then supplied to 7.2 KV ESF bus 1 DA (Train A).This is the preferred or normal power source to Train A and the alternate power source for Train B.* 7.2 KV ESF bus 1 DB (Train B) is supplied from the emergency auxiliary transformer (XTF-31).
The emergency auxiliary transformer receives 230 KV power from the Virgil C. Summer substation (switchyard) bus 3. This transformer is the preferred power source for Train B and the alternate power source for Train A.The Parr Hydro Plant provides a 13.8 KV AC line to the 7.2 KV ESF buses. This Alternate AC Power Supply has the capacity to supply only one fully loaded ESF bus (ref. 4).The AAC is designed to provide back-up power to either ESF bus whenever one of the Diesel Generators is out of service, particularly in Modes 1 through 4. The AAC is verified available and an operational readiness status check is performed when it is anticipated that one of the Diesel Generators will be inoperable for longer than the allowed outage time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The design of the AAC is capable of providing the required safety and non-safety related loads in the event of a total loss of offsite power and if both Diesel Generators fail to start and load. During these events it is assumed that there is no seismic event or an event that requires safeguards actuation (e.g., safety injection, containment spray, etc.) Although the AAC is not designed for DBA loads, it is capable of supplying sufficient power to mitigate the effects of an accident.
The AAC is not credited in the safety analysis.
The AAC is, however, capable of mitigating the dominant core damage Page 215 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]sequences and provides a significant overall risk reduction for station operation.
The AAC alone is adequate to supply electrical power to affect a safe shutdown of the plant (ref. 7).The two trains of 7.2 KV safeguards power are also provided with an onsite standby source of power for supplying power when the ESF and emergency auxiliary transformers are not available.
The Diesel Generators A and B are capable of supplying all loads on the distribution network of their respective train (ref. 3, 4, 5, 6, 7).Indication of continuing core cooling degradation is manifested by entry to Critical Safety Function Status Tree (CSFST) Core Cooling-RED or ORANGE path (ref. 8).Critical Safety Function Status Tree (CSFST) Core Cooling-RED or ORANGE path is given in Figure 5 and indicates significant core exit superheating and core uncovery.Generic This IC addresses a prolonged loss of all power sources to AC engineered safeguard (ES)buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buses will lead to a loss of one or more fission product barriers.
In addition, fission product barrier monitoring capabilities may be degraded under these conditions.
The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG1. This will allow additional time for implementation of offsite protective actions.Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one 7.2KV ES bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers.The estimate for restoring at least one 7.2KV ES bus should be based on a realistic appraisal of the situation.
Mitigation actions with a low probability of success should not be Page 216 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public.The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.VCSNS Basis Reference(s):
- 1. FSAR Section 8.3.2.1.2 2. FSAR Section 8.4.1 3. FSAR Section 8 4. EOP-6.0 Loss of All ESF AC Power 5. EOP-1.0 Reactor Trip/Safety Injection Actuation 6. SOP-304 115KV/7.2KV Operations
- 7. SOP-306 Emergency Diesel Generator 8. EOP-1 2.0 Monitoring of Critical Safety Functions 9. AOP-304.1 Loss of Bus 1 DA(1 DB) with the Diesel not Available 10.NEI 99-01 SG1 Page 217 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
S -System Malfunction Subcategory:
1 -Loss of ESF AC Power Initiating Condition:
Prolonged loss of all offsite and all onsite AC power to ESF buses or loss of all AC and vital DC power sources for 15 minutes or longer.EAL: SG1.2 General Emergency Loss of all offsite and all onsite AC power (Table S-1) capability to 7.2 KV ESF buses 1 DA and 1 DB for >- 15 min.AND< 108 VDC on both Train A and Train B vital 125 VDC systems for _> 15 min.(Note 1)Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table S-1 AC Power Supplies Offsite:* 115 KV power to XTF-4 and XTF-5 o 230 KV power to XTF-31 0 Parr Hydro Plant 13.8 KV power to ESF bus 1DA or 1DB Onsite: " Diesel Generator A" Diesel Generator B Mode Applicability:
1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):
None Basis: Plant-Specific Page 218 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]ESF AC Power As used in this EAL the term "capability" means an AC power source is either currently powering essential loads on one or more 7.2 KV ESF buses or is capable of energizing and powering essential loads on at least one 7.2 KV ESF bus within 15 min.Table S-1 lists AC sources capable of powering ESF buses. Safeguards power originates offsite from two independent sources (ref. 5):* The Parr Generating Complex supplies 115 KV power to the two Engineered Safety Feature (ESF) transformers (XTF-4 and XTF-5). The transformer outputs are combined at 7.2 KV bus 1 DX and then supplied to 7.2 KV ESF bus 1 DA (Train A).This is the preferred or normal power source to Train A and the alternate power source for Train B.* 7.2 KV ESF bus 1 DB (Train B) is supplied from the emergency auxiliary transformer (XTF-31).
The emergency auxiliary transformer receives 230 KV power from the Virgil C. Summer substation (switchyard) bus 3. This transformer is the preferred power source for Train B and the alternate power source for Train A.The Parr Hydro Plant provides a 13.8 KV AC line to the 7.2 KV ESF buses. This Alternate AC Power Supply has the capacity to supply only one fully loaded ESF bus (ref. 4).The AAC is designed to provide back-up power to either ESF bus whenever one of the Diesel Generators is out of service, particularly in Modes 1 through 4. The AAC is verified available and an operational readiness status check is performed when it is anticipated that one of the Diesel Generators will be inoperable for longer than the allowed outage time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The design of the AAC is capable of providing the required safety and non-safety related loads in the event of a total loss of offsite power and if both Diesel Generators fail to start and load. During these events it is assumed that there is no seismic event or an event that requires safeguards actuation (e.g., safety injection, containment spray, etc.) Although the AAC is not designed for DBA loads, it is capable of supplying sufficient power to mitigate the effects of an accident.
The AAC is not credited in the safety analysis.
The AAC is, however, capable of mitigating the dominant core damage Page 219 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]sequences and provides a significant overall risk reduction for station operation.
The AAC alone is adequate to supply electrical power to affect a safe shutdown of the plant (ref. 7).The two trains of 7.2 KV safeguards power are also provided with an onsite standby source of power for supplying power when the ESF and emergency auxiliary transformers are not available.
The Diesel Generators A and B are capable of supplying all loads on the distribution network of their respective train (ref. 1, 2, 3, 4, 5).DC Vital Power Class 1 E 125 VDC power consists of two separate main distribution panels. These panels are DPN-1 HA and DPN-1 HB for the Train A and Train B vital 125 VDC systems (ref. 8).They are both located on the 412' level of the Intermediate Building.
Each main panel is supplied DC power through a battery charger (XBC-1A and XBC-1 B) and is backed up by a 60 cell, lead-acid storage battery (ref. 9).Minimum DC bus voltage is 108 VDC (ref. 10, 11). MCB annunciators XCP-636 4-6 and XCP-637 4-6 (DC SYS OVRVOLT/UNDRVOLT) signal low Train A and Train B voltage at 126 VDC (ref. 12,13). Train A and Train B voltage may be monitored on MCB Panel XCP-6116 voltmeters (ref. 14).Generic This IC addresses a concurrent and prolonged loss of both AC and Vital DC power. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A loss of Vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both AC and DC power will lead to multiple challenges to fission product barriers.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met.VCSNS Basis Reference(s):
Page 220 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]1. FSAR Section 8 2. EOP-6.0 Loss of All ESF AC Power 3. EOP-1.0 Reactor Trip/Safety Injection Actuation 4. SOP-304 115KV/7.2KV Operations
- 5. SOP-306 Emergency Diesel Generator 6. AOP-304.1 Loss of Bus 1 DA(1 DB) with the Diesel not Available 7. Technical Specifications Bases 3/4.8 8. FSAR Figure 8.3-2aa 9. FSAR Section 8.3.2.1 10. EOP-6.0 Loss of All ESF AC Power 11. FSAR Section 8.3.2.1.3 12. ARP-001 -XCP-636 Annunciator Point 4-6 13. ARP-001 -XCP-637 Annunciator Point 4-6 14.201-332 Main Control Board Instrumentation Control Panel XCP-6116 15. NEI 99-01 SG8 Page 221 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
S -System Malfunction Subcategory:
2 -Loss of Vital DC Power Initiating Condition:
Loss of all vital DC power for 15 minutes or longer.EAL: SS2.1 Site Area Emergency< 108 VDC on both Train A and Train B vital 125 VDC systems for - 15 min. (Note 1)Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability:
1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):
None Basis: Plant-Specific Class 1 E 125 VDC power consists of two separate main distribution panels. These panels are DPN-1 HA and DPN-1 HB for the Train A and Train B vital 125 VDC systems (ref. 1).They are both located on the 412' level of the Intermediate Building.
Each main panel is supplied DC power through a battery charger (XBC-1 A and XBC-1 B) and is backed up by a 60 cell, lead-acid storage battery (ref. 2).Minimum DC bus voltage is 108 VDC (ref. 3, 4). MCB annunciators XCP-636 4-6 and XCP-637 4-6 (DC SYS OVRVOLT/UNDRVOLT) signal low Train A and Train B voltage at 126 VDC (ref. 5, 6). Train A and Train B voltage may be monitored on MCB Panel XCP-6116 voltmeters (ref. 7).Generic This IC addresses a loss of Vital DC power which compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure of plant functions needed for the protection of the public.Page 222 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Escalation of the ECL would be via ICs RG1, FG1 or SG1.VCSNS Basis Reference(s):
- 1. FSAR Figure 8.3-2aa 2. FSAR Section 8.3.2.1 3. EOP-6.0 Loss of All ESF AC Power 4. FSAR Section 8.3.2.1.3 5. ARP-001-XCP-636 Annunciator Point 4-6 6. ARP-001-XCP-637 Annunciator Point 4-6 7. 201-332 Main Control Board Instrumentation Control Panel XCP-6116 8. NEI 99-01 SS8 Page 223 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
S -System Malfunction Subcategory:
3 -Loss of Control Room Indications Initiating Condition:
UNPLANNED loss of Control Room indications for 15 minutes or longer.EAL: SU3.1 Unusual Event An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for >- 15 min. (Note 1)Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table S-2 Safety System Parameters" Reactor power* Reactor vessel/pressurizer level* RCS pressure* Core Exit TCs* Level in at least one SG* EFW/AFW flow Mode Applicability:
1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):
UNPLANNED
-A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown.Basis: Plant-Specific Display information important in evaluating the performance of a safeguards system during periodic test, continuous normal operation, or post-accident operation is provided on the Main Control Board (MCB) panels XCP-61 01 through XCP-6117.
Sufficient process indicators, alarms, and recorders are provided to enable the operator to determine whether Page 224 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]a system is performing normally or if there is some unanticipated failure within a system (ref. 1, 2). The Integrated Plant Computer System (IPCS) monitors selected instrument channels to supplement the display information (ref. 3).CSFST paramters are normally monitored using the SPDS display on the Integrated Plant Computer System (IPCS) (ref. 4).Generic This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant.As used in this EAL, an "inability to monitor' means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s).
For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required.
The event would be reported if it significantly impaired the capability to perform emergency assessments.
In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.
This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition.
In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well.For example, if the value for reactor vessel level [cannot be determined from the Page 225 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT El indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.
Escalation of the ECL would be via IC SA3.VCSNS Basis Reference(s):
- 1. FSAR Section 7.5 2. FSAR Section 7.6 3. OAP-107.1 Control of IPCS Functions 4. EOP-12.0 Monitoring of Critical Safety Functions 5. NEI 99-01 SU2 Page 226 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
S -System Malfunction Subcategory:
3 -Loss of Control Room Indications Initiating Condition:
UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.EAL: SA3.1 Alert An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for -> 15 min. (Note 1)AND Any of the following transient events in progress:* Automatic or manual runback greater than 25% thermal reactor power" Electrical load rejection greater than 25% full electrical load" Reactor trip* ECCS actuation Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table S-2 Safety System Parameters" Reactor power" Reactor vessel/pressurizer level" RCS pressure" Core Exit TCs" Level in at least one SG* EFW/AFW flow Mode Applicability:
1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):
UNPLANNED
-A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown.Basis: Page 227 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Plant-Specific Display information important in evaluating the performance of a safeguards system during periodic test, continuous normal operation, or post-accident operation is provided on the Main Control Board (MCB) panels XCP-6101 through XCP-6117.
Sufficient process indicators, alarms, and recorders are provided to enable the operator to determine whether a system is performing normally or if there is some unanticipated failure within a system (ref. 1, 2). The Integrated Plant Computer System (IPCS) monitors selected instrument channels to supplement the display information (ref. 3).CSFST paramters are normally monitored using the SPDS display on the Integrated Plant Computer System (IPCS) (ref. 4).Generic This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant.As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s).
For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required.
The event would be reported if it significantly impaired the capability to perform emergency assessments.
In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.
Page 228 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition.
In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well.For example, if the value for steam generator level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.
Escalation of the ECL would be via ICs FS1 or IC RS1.VCSNS Basis Reference(s):
- 1. FSAR Section 7.5 2. FSAR Section 7.6 3. OAP-107.1 Control of IPCS Functions 4. EOP-12.0 Monitoring of Critical Safety Functions 5. NEI 99-01 SA2 Page 229 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
S -System Malfunction Subcategory:
4 -RCS Activity Initiating Condition:
Reactor coolant activity greater than Technical Specification allowable limits.EAL: SU4.1 Unusual Event With letdown in service, RM-L1 high range monitor > 39,000 cpm Mode Applicability:
1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):
None Basis: Plant-Specific The status of coolant activity and radiation levels is routinely monitored to detect the onset of fuel failure (ref. 1). This EAL addresses reactor coolant letdown line radiation levels sensed by RM-L1 in excess of Technical Specification allowable limits. Primary coolant letdown line radiation monitor RM-L1 provides a means of detecting the presence of failed fuel by indication of an increase in letdown activity which is then verified by analysis of samples. Two detectors with overlapping range are provided.
The low range is designed for the monitor to be on range with the radioactivity resulting from tramp uranium and the corrosion products.
The range of overlap between the low and high range detectors is such that two detectors would be operational in the range of concentrations relating to plant operation with failed fuel (ref. 2). Alarms are received on the MCB panel from the low and high range detectors (ref. 3). The high range alarm is _< 5 X EQUIL, where EQUIL is the normal or expected reading of the monitor when radioactivity is normally present in the sample stream. The low range alarm is <_ 2 X EQUIL (ref. 4).Page 230 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]The specified EAL threshold setpoint was calculated using RCS activities given in Table 11.1-2 of the FSAR and included all activities in the table scaled to 1.0 pCi/gm dose equivalent iodine (ref. 5, 6).Generic This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications.
This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.Escalation of the ECL would be via lCs FA1 or the Recognition Category R ICs.VCSNS Basis Reference(s):
- 1. SAP-154 Failed Fuel Action Plan 2. VCSNS Design Bases Document -Radiation Monitoring System (RM)3. ARP-01 9-XCP-642 4. HPP-904 Use of the Radiation Monitoring System (RMS)5. TWR 11.0/6.2-07-013 RM-L1 Calculations for New EAL's 6. FSAR Table 11.1-2 7. NEI 99-01 SU3 Page 231 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
S -System Malfunction Subcategory:
4 -RCS Activity Initiating Condition:
Reactor coolant activity greater than Technical Specification allowable limits.EAL: SU4,2 Unusual Event Sample analysis indicates that a primary coolant activity value is > an allowable limit specified in Technical Specifications 3/4.4.8 Mode Applicability:
1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):
None Basis: Plant-Specific This EAL addresses primary coolant samples exceeding Technical Specification LCOs 3/4.4.8, which are applicable in Modes 1, 2 and 3 with RCS average temperature (Tavg)> 500°F (ref. 1). The Technical Specification limits accommodate an iodine spike phenomenon that may occur following changes in thermal power. The Technical Specification LCO limits are established to minimize the offsite radioactivity dose consequences in the event of a steam generator tube rupture (SGTR) accident (ref. 2).Generic This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications.
This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.Escalation of the ECL would be via ICs FA1 or the Recognition Category R ICs.VCSNS Basis Reference(s):
- 1. Technical Specifications 3/4.4.8 2. Technical Specifications Bases 3/4.4.8 Page 232 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]3. NEI 99-01 SU3 Page 233 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
S -System Malfunction Subcategory:
5 -RCS Leakage Initiating Condition:
RCS leakage for 15 minutes or longer.EAL: SU5.1 Unusual Event RCS unidentified or pressure boundary leakage > 10 gpm for > 15 min.OR RCS identified leakage > 25 gpm for > 15 min.OR Leakage from the RCS to a location outside containment
> 25 gpm for > 15 min.(Note 1)Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability:
1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):
None Basis: Plant-Specific Unidentified leakage and identified leakage are determined by performance of the RCS water inventory balance (IPCS CHGNET, LRATE). Pressure boundary leakage would first appear as unidentified leakage and can only be positively identified by inspection (ref. 1).STP-1 14.002 is used to ensure RCS leakage is within Technical Specification limits (ref.2). MCB annunciator XCP-615 3-6 (RCS LEAK DET >1 GPM) signals RCS leakage into the Reactor Building sump that challenges Technical Specifications LCO limits (ref. 1, 4).The rate of primary-to-secondary leakage is determined by comparing the ratio of the activity of a given isotope measured in the secondary plant (i.e., steam generators, condensate or condenser off-gas) to that same isotope in the Reactor Coolant System (ref.5).Page 234 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Technical Specifications (ref. 6) defines RCS leakage as follows: " Controlled Leakage: Seal water flow supplied to the reactor coolant pump seals.* Identified Leakage: o Leakage (except Controlled Leakage) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or o Leakage into the containment atmosphere from sources that are both specifically located and unknown either not to interfere with the operation of leakage detection systems or not to be Pressure Boundary Leakage, or o Reactor coolant system leakage through a steam generator to the secondary system." Unidentified Leakage: All leakage (except Controlled Leakage) that is not identified leakage.* Pressure Boundary Leakage: Leakage (except steam generator tube leakage)through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.RCS leakage outside of the containment that is not considered identified or unidentified leakage per Technical Specifications includes leakage via CVCS/Letdown and interfacing system leakage such as RCS to the Component Cooling Water (CCW system) and RCS sampling system (ref. 7).General symptoms of RCS leakage include the following (ref. 7): " Decreasing Pressurizer level with increased charging flow and normal letdown flow" Increasing radiation level in Containment or the Auxiliary Building indicated by any of the following:
o RM-G5, RB PERSONNEL ACCESS AREA GAMMA 0 RM-G6, RB REFUEL BRIDGE AREA GAMMA Page 235 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]o RM-A2, RB SAMPLE LINE PARTICULATE (IODINE)(GAS)
ATMOS MONITOR o RM-A3, MAIN PLANT VENT EXH PARTICULATE(IODINE)(GAS)
ATMOS MONITOR o RM-A1 1, AB VENT GAS ATMOS MONITOR" Increasing sump level in Containment or the Auxiliary Building* Increased VCT makeup frequency* Increasing radiation level in the CCW System as indicated on RM-L2A(B), COMPONENT COOLING LIQUID MONITOR" Any of the following Main Control Board annunciators in alarm: o RBCU 1A/2A DRN FLO HI (XCP-606 2-2)o RBCU 1 B/2B DRN FLO HI (XCP-607 2-2)o RCS LEAK DET >1GPM (XCP-615 3-6)Generic This IC addresses RCS leakage which may be a precursor to a more significant event. In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant.The first and second EAL conditions are focused on a loss of mass from the RCS due to"unidentified leakage", "pressure boundary leakage" or "identified leakage" (as these leakage types are defined in the plant Technical Specifications).
The third condition addresses an RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These conditions thus apply to leakage into the containment, a secondary-side system (e.g., steam generator tube leakage) or a location outside of containment.
The leak rate values for each condition were selected because they are usually observable with normal Control Room indications.
Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation).
The first condition uses a Page 236 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]lower value that reflects the greater significance of unidentified or pressure boundary leakage.The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification.
An emergency classification would be required if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flow cannot be isolated).
The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.Escalation of the ECL would be via ICs of Recognition Category R or F.VCSNS Basis Reference(s):
- 1. ARP-001 -XCP-615 Annunciator Point 3-6 2. STP-1 14.002 Operational Leakage Calculation
- 3. STP-1 14.003 RCS Leak Detection Setpoint Determination
- 4. Technical Specification 3.4.6.2 5. CP-307 Primary-to-Secondary Leakage Rate Determination
- 6. Technical Specifications, Definitions
- 7. AOP-1 01.1 Loss of Reactor Coolant Not Requiring SI 8. ES-1 61 RCS Leakage Management Program 9. FSAR Section 5.2.7 10. FSAR Section 7.6.5 11. NEI 99-01 SU4 Page 237 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
S -System Malfunction Subcategory:
6 -RTS Failure Initiating Condition:
Automatic or manual trip fails to shut down the reactor EAL: SU6.1 Unusual Event An automatic trip did not shut down the reactor after any RTS setpoint is exceeded AND A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor as indicated by reactor power < 5% (Note 8).Note 8: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.
Mode Applicability:
1 -Power Operation Definition(s):
None Basis: Plant-Specific A reactor trip is automatically initiated by the Reactor Trip System (RTS) when certain continuously monitored parameters exceed predetermined setpoints (ref. 1): Following a successful reactor trip, rapid insertion of the control rods occurs. Nuclear power promptly drops to a few percent of the original power level and then decays to a level some 8 decades less at a startup rate of about -1/3 DPM. The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor power starts to be observable.
A predictable post-trip response from an automatic reactor trip signal should therefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentation and a negative startup rate as nuclear power drops into the source range (ref. 1).The operator recognizes that the reactor has tripped by observing (ref. 1): Page 238 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]" Any red first-out Reactor Trip annunciator lit* Rapid decrease in neutron flux level as indicated by the NI System* Shutdown and Control Rods fully inserted* Rod Bottom Lights lit If these responses cannot be verified, operators perform contingency actions that manually insert control rods, open the reactor trip and bypass breakers in the Reactor Trip Switchgear (IB-463), and tripping the Rod Drive MG sets in the Rod Drive MG Control Cabinet (IB-463).
Local opening of these breakers requires actions outside of the Control Room; rapid control rod insertion by these methods is therefore not considered a"successful" manual reactor trip. For purposes of emergency classification, a "successful" manual reactor trip, therefore, includes only those immediate actions taken by the reactor operator in the Control Room on the control consoles.
Manual reactor trip switches CS-CR01 and CS-CR01A are located on MCB panels XCP-61 10 and XCP-6114, respectively (ref. 3, 4). These switches and controls can be rapidly manipulated from the specified MCB panels.In the event that the operator identifies a reactor trip is imminent and successfully initiates a manual reactor trip before the automatic trip setpoint is reached, no declaration is required.
The successful manual trip of the reactor before it reaches its automatic trip setpoint or reactor trip signals caused by instrumentation channel failures do not lead to a potential fission product barrier loss. If manual reactor trip actions in the Control Room fail to reduce reactor power below the power associated with the SAFETY SYSTEM design (<5%) (ref. 2), the event escalates to an Alert under EAL SA6.1.Generic This IC addresses a failure of the RTS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.Page 239 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Following the failure on an automatic reactor trip, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip using a different switch). Depending upon several factors, the initial or subsequent effort to manually trip the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip. This action does not include manually driving in control rods or implementation of boron injection strategies.
Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".
The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the ECL will escalate to an Alert via IC SA6. Depending upon the plant response, escalation is also possible via IC FAI. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event.A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.Page 240 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Should a reactor trip signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied." If the signal causes a plant transient that should have included an automatic reactor trip and the RTS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.
- If the signal does not cause a plant transient and the trip failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted.
VCSNS Basis Reference(s):
- 1. EOP-1.0 Reactor Trip/Safety Injection Actuation 2. EOP-13.0 Response to Abnormal Nuclear Power Generation
- 3. 201-326 Main Control Board Instrumentation Control Panel XCP-61 10 4. 201-330 Main Control Board Instrumentation Control Panel XCP-6114 5. NEI 99-01 SU5 Page 241 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
S -System Malfunction Subcategory:
6 -RTS Failure Initiating Condition:
Automatic or manual trip fails to shut down the reactor EAL: SU6.2 Unusual Event A manual trip did not shut down the reactor after any manual trip action was initiated AND A subsequent automatic trip or manual trip action taken at the reactor control consoles is successful in shutting down the reactor as indicated by reactor power < 5% (Note 8).Note 8: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.
Mode Applicability:
1 -Power Operation Definition(s):
None Basis: Plant-Specific This EAL addresses a failure of a manually initiated trip in the absence of having exceeded an automatic RTS trip setpoint and a subsequent automatic or manual trip is successful in shutting down the reactor (reactor power < 5%).A reactor trip is automatically initiated by the Reactor Trip System (RTS) when certain continuously monitored parameters exceed predetermined setpoints (ref. 1): Following a successful reactor trip, rapid insertion of the control rods occurs. Nuclear power promptly drops to a few percent of the original power level and then decays to a level some 8 decades less at a startup rate of about -1/3 DPM. The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor power starts to be observable.
A predictable post-trip response from an automatic reactor trip signal should therefore consist of a prompt drop in reactor power as Page 242 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]sensed by the nuclear instrumentation and a negative startup rate as nuclear power drops into the source range (ref. 1).The operator recognizes that the reactor has tripped by observing (ref. 1): " Any red first-out Reactor Trip annunciator lit* Rapid decrease in neutron flux level as indicated by the NI System" Shutdown and Control Rods fully inserted" Rod Bottom Lights lit If these responses cannot be verified, operators perform contingency actions that manually insert control rods, open the reactor trip and bypass breakers in the Reactor Trip Switchgear (IB-463), and tripping the Rod Drive MG sets in the Rod Drive MG Control Cabinet (IB-463).
Local opening of these breakers requires actions outside of the Control Room; rapid control rod insertion by these methods is therefore not considered a"successful" manual reactor trip. For purposes of emergency classification, a "successful" manual reactor trip, therefore, includes only those immediate actions taken by the reactor operator in the Control Room on the control consoles.
Manual reactor trip switches CS-CR01 and CS-CR01A are located on MCB panels XCP-61 10 and XCP-6114, respectively (ref. 3, 4). These switches and controls can be rapidly manipulated from the specified MCB panels.Generic This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor (trip [PWR] / scram [BWR]) that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.Following the failure on an automatic reactor trip, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip). If these manual actions are successful in shutting down the reactor, core heat Page 243 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip using a different switch). Depending upon several factors, the initial or subsequent effort to manually trip the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip. This action does not include manually driving in control rods or implementation of boron injection strategies.
Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".
The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the ECL will escalate to an Alert via IC SA6. Depending upon the plant response, escalation is also possible via IC FAl. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event.A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.Should a reactor trip signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.Page 244 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]* If the signal causes a plant transient that should have included an automatic reactor trip and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated." If the signal does not cause a plant transient and the trip failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted.
VCSNS Basis Reference(s):
- 1. EOP-1.0 Reactor Trip/Safety Injection Actuation 2. EOP-13.0 Response to Abnormal Nuclear Power Generation
- 3. 201-326 Main Control Board Instrumentation Control Panel XCP-6110 4. 201-330 Main Control Board Instrumentation Control Panel XCP-6114 5. NEI 99-01 SU5 Page 245 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
S -System Malfunction Subcategory:
6- RTS Failure Initiating Condition:
Automatic or manual trip fails to shut down the reactor and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor.EAL: SA6.1 Alert An automatic or manual trip fails to shut down the reactor AND Manual actions taken at the reactor control console are not successful in shutting down the reactor as indicated by reactor power -> 5% (Note 8)Note 8: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.
Mode Applicability:
1 -Power Operation Definition(s):
None Basis: Plant-Specific This EAL addresses any automatic or manual reactor trip signal that fails to shut down the reactor followed by a subsequent manual trip that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the SAFETY SYSTEMS were designed (> 5%) (ref. 1). For purposes of emergency classification, a"successful" manual reactor trip, therefore, includes only those immediate actions taken by the reactor operator in the Control Room to actuate manual reactor trip switches CS-CR01 and CS-CR01A (located on MCB panels XCP-61 10 and XCP-6114, respectively) (ref. 2, 3). Although the reactor can be manually tripped using controls on MCB panel XCP-6115 (e.g., depressing MASTER TRIP/EMERGENCY TRIP pushbuttons) (ref. 4), the Page 246 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]turbine/generator trip is not considered a "successful" manual reactor trip when evaluating this EAL.Automatic and manual trips are not considered successful if action away from the Control Room is required to trip the reactor. Local operator actions to open the reactor trip and bypass breakers in the Reactor Trip Switchgear (IB-463), and tripping the Rod Drive MG sets in the Rod Drive MG Control Cabinet (IB-463) are not considered "successful" manual reactor trips. If any of the alternate recovery actions for emergency boration of the RCS listed in EOPs are required to reduce reactor power below the power associated with the SAFETY SYSTEM design (< 5%), the reactor trips have been unsuccessful.
Negative intermediate range startup rate (SUR) is used as an indicator of decreasing power and should be observed following any reactor trip from power (ref. 1).Generic This IC addresses a failure of the RTS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful.
This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RTS.A manual action at the reactor control console is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip). This action does not include manually driving in control rods or implementation of boron injection strategies.
If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control console (e.g., locally opening breakers).
Actions taken at backpanels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control console".The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the Page 247 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shut down the reactor is prolonged enough to cause a challenge to the core cooling or RCS heat removal safety functions, the ECL will escalate to a Site Area Emergency via IC SS6. Depending upon plant responses and symptoms, escalation is also possible via IC FS1. Absent the plant conditions needed to meet either IC SS6 or FS1, an Alert declaration is appropriate for this event.It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.VCSNS Basis Reference(s):
- 1. EOP-13.0 Response to Abnormal Nuclear Power Generation
- 2. 201-326 Main Control Board Instrumentation Control Panel XCP-6110 3. 201-330 Main Control Board Instrumentation Control Panel XCP-6114 4. 201-331 Main Control Board Instrumentation Control Panel XCP-6115 5. NEI 99-01 SA5 Page 248 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
S -System Malfunction Subcategory:
6 -RTS Failure Initiating Condition:
Inability to shut down the reactor causing a challenge to core cooling or RCS heat removal.EAL: SS6.1 Site Area Emergency An automatic or manual trip fails to shutdown the reactor AND All manual actions to shut down the reactor are not successful in shutting down the reactor as indicated by reactor power >- 5%AND EITHER of the following conditions exist: " CSFST Core Cooling-RED path conditions met" CSFST Heat Sink-RED path conditions met Mode Applicability:
1 -Power Operation Definition(s):
None Basis: Plant-Specific This EAL addresses the following:
- Any automatic or manual reactor trip signal followed by a manual trip that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the SAFETY SYSTEMS were designed (> 5%, ref. 1) (EAL SA6.1), and* Indications that either core cooling is extremely challenged (CSFST Core Cooling-RED path) or heat removal is extremely challenged (CSFST Heat Sink-Red path)(ref. 2.)Page 249 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT El Manual reactor trip switches CS-CR01 and CS-CR01A are located on MCB panels XCP-6110 and XCP-6114, respectively (ref. 3, 4). These controls can be rapidly manipulated from the specified MCB panels and constitute the normal methods of initiating a manual trip. These are the only manual trip methods applicable to evaluation of EAL SA6.1.At the Site Area Emergency classification level, however, additional capabilities away from the Control Room may be considered such as opening the reactor trip and bypass breakers in the Reactor Trip Switchgear (IB-463) and tripping the Rod Drive MG sets in the Rod Drive MG Control Cabinet (IB-463).Indication that core cooling is extremely challenged is manifested by entry to Critical Safety Function Status Tree (CSFST) Core Cooling-RED path (Figure 5) (ref. 2). Indication that heat removal is extremely challenged is manifested by entry to CSFST Heat Sink-RED path (Figure 6) (ref. 2).The setpoint values provided in brackets following the normal setpoint values in the CSFSTs are used under adverse containment conditions.
Adverse containment conditions are defined as either (ref. 6): " Hi-1 RB pressure (> 3.6 psig), or" Containment Hi-radiation
(__ 100,000 Rads integrated dose)The IPCS monitors these parameters and indicates when adverse containment values should be used. The instruments available to monitor these containment parameters are Containment pressure on PI-950, PI-951, PI-952 or PI-953, and Containment radiation on RM-G7 or RM-G18. When no adverse values are provided, the given setpoint is used for any containment condition, as the parameter measurement is independent of containment atmosphere.
If containment pressure decreases below the adverse pressure setpoint after it has been exceeded, the normal values are again used. Once the adverse radiation level setpoint is exceeded, the adverse containment values must be utilized through event recovery and establishment of normal operating procedures.
Engineering should then be requested to evaluate instrumentation inaccuracies.
Generic Page 250 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.
In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs.
This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shut down the reactor.The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor.A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.Escalation of the ECL would be via IC RG1 or FGI.VCSNS Basis Reference(s):
- 1. EOP-1 3.0 Response to Abnormal Nuclear Power Generation
- 2. EOP-12.0 Monitoring of Critical Safety Functions 3. 201-326 Main Control Board Instrumentation Control Panel XCP-61 10 4. 201-330 Main Control Board Instrumentation Control Panel XCP-6114 5. 201-331 Main Control Board Instrumentation Control Panel XCP-6115 6. OAP-103.4 EOP/AOP User's Guide 7. NEI 99-01 SS5 Page 251 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
S -System Malfunction Subcategory:
7 -Loss of Communications Initiating Condition:
Loss of all onsite or offsite communications capabilities.
EAL: SU7.1 Unusual Event Loss of all Table S-3 onsite communication methods OR Loss of all Table S-3 ORO/NRC communication methods Table S-3 Communication Methods System Onsite ORO/NRC Gai-Tronics system X Radio system X Internal Telephone system X Telephone land lines X X Fiberoptic links X Satellite phone system X Federal Telephone System (ENS) X ESSX X Mode Applicability:
1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):
None Basis: Plant-Specific Page 252 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]The Table S-3 list for onsite communications loss encompasses the loss of all means of routine communications (e.g., commercial and internal telephones, page party system (Gai-Tronics) and radios) (ref. 1, 2, 3).The Table S-3 list for offsite (ORO/NRC) communications loss encompasses the loss of all means of communications with offsite authorities.
This includes the FTS (ENS), commercial telephone lines and dedicated phone systems (fiberoptic and satellite) (ref. 1, 2,3).This EAL is the hot condition equivalent of the cold condition EAL CU5.1.Generic This IC addresses a significant loss of on-site or offsite communications capabilities.
While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC.This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations.
The second EAL condition addresses a total loss of the communications methods used to notify all OROs or NRC of an emergency declaration.
The OROs referred to here are the State, Fairfield, Newberry, Lexington and Richland County EOCs as well as the NRC.VCSNS Basis Reference(s):
- 1. FSAR 9.5.2 2. EP-100 Radiation Emergency Plan, Section 7.5 3. EP-100 Radiation Emergency Plan, Figure 7-2 4. NEI 99-01 SU6 Page 253 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category: Subcategory:
Initiating Condition:
S -System Malfunction 8 -Containment Isolation Failure Failure to isolate containment or loss of containment pressure control EAL: SU8.1 Unusual Event Containment isolation actuated AND At least one isolation valve in each penetration is not closed within 15 min. of the actuation (Note 1)Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Mode Applicability:
1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):
None Basis: Plant-Specific None Generic This IC addresses a failure of containment penetrations to automatically isolate (close)when required by an actuation signal. Absent challenges to another fission product barrier, this condition represents potential degradation of the level of safety of the plant.The containment isolation signal must be generated as the result on an off-normal/accident condition (e.g., a safety injection or high containment pressure);
a failure resulting from testing or maintenance does not warrant classification.
The determination of containment and penetration status -isolated or not isolated -should be made in accordance with the Page 254 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]appropriate criteria contained in the plant AOPs and EOPs. The 15-minute criterion is included to allow operators time to manually isolate the required penetrations, if possible.This event would escalate to a Site Area Emergency in accordance with IC FS1 if there were a concurrent loss or potential loss of either the Fuel Clad or RCS fission product barriers and a direct release pathway to the environment as a result of the failed isolation.
VCSNS Basis Reference(s):
- 2. NEI 99-01 SU7 Page 255 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
S -System Malfunction Subcategory:
8 -Containment Isolation Failure Initiating Condition:
Failure to isolate containment or loss of containment pressure control.EAL: SU8.2 Unusual Event Containment pressure > 12 psig AND< one full train of depressurization equipment (Table S-4) is operating per designfor>_ 15 min.(Note 1)Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table S-4 Full Train Depressurization Equipment RBCU Groups Containment Sprays Operating Operating 2 0 1 1 0 2 Mode Applicability:
1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):
None Basis: Plant-Specific The Containment pressure setpoint (12 psig, ref. 2, 3) is the pressure at which the Containment Spray System should actuate and begin performing its function.
The design Page 256 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]basis accident analyses and evaluations assume the loss of one Containment Spray System train (ref. 2, 3).Each spray subsystem is started by separate ESF containment isolation Phase A and spray actuation signals. Normally, both subsystems operate; however, they are independent and can operate individually.
Although the design basis for Reactor Building heat removal is one spray subsystem operating in conjunction with one Reactor Building Cooling Unit (with one RHR pump and one charging pump providing emergency core cooling water), two Reactor Building Cooling Units (RBCUs) with no spray pumps or two spray pumps with no RBCUs can handle all required heat loads Technical Specifications defined equipment that comprises one full train of depressurization equipment is given in the note (ref. 1,2, 4, 5).RBCU operation verification is performed in accordance with EOP-1.0 Attachment 3 SI Equipment Verification.
In order to take credit for a RBCU operating per design, each RBCU must meet minimum Service Water flow requirements (ref. 6).Generic This IC addresses an event that results in high containment pressure with a concurrent failure of containment pressure control systems. Absent challenges to another fission product barrier, this condition represents potential degradation of the level of safety of the plant.This EAL addresses a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible.
The inability to start the required equipment indicates that containment heat removal/depressurization systems (e.g., containment sprays ) are either lost or performing in a degraded manner.This event would escalate to a Site Area Emergency in accordance with IC FS1 if there were a concurrent loss or potential loss of either the Fuel Clad or RCS fission product barriers.Page 257 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]VCSNS Basis Reference(s):
- 1. FSAR Section 6.2.2.2.1.2
- 2. OAP-103.2 Emergency Operating Procedure Setpoint Document 3. EOP-12.0 Monitoring of Critical Safety Functions 4. Technical Specifications 3/4-6.2.1 5. Technical Specifications 3/4-6.2.3 6. EOP-1.0 Reactor Trip/Safety Injection Actuation Attachmen 3 SI Equipment Verification
- 7. NEI 99-01 SU7 Page 258 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
S -System Malfunction Subcategory:
9 -Hazardous Event Affecting Safety System Initiating Condition:
Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.EAL: SA9.1 Alert The occurrence of any Table S-5 hazardous event AND EITHER of the following: " Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode" The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode Table S-5 Hazardous Events* Seismic event (earthquake)
- Internal or external FLOODING event* High winds or tornado strike* FIRE* EXPLOSION* Other events with similar hazard characteristics as determined by the Shift Supervisor Mode Applicability:
1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):
EXPLOSION
-A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization.
A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an EXPLOSION.
Such Page 259 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT El events require a post-event inspection to determine if the attributes of an EXPLOSION are present.FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.FLOODING -A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary;(2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
VISIBLE DAMAGE -Damage to a component or structure that is readily observable without measurements, testing, or analysis.
The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.
Basis: Plant-Specific
- The significance of seismic events are discussed under EAL HU2.1 (ref. 1)." Internal FLOODING may be caused by events such as component failures, equipment misalignment, or outage activity mishaps (ref. 2).Page 260 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]" Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 100 mph (sustained). (ref. 3)." Refer to VCSNS Fire Protection Evaluation Report, Section 4.0 "Hazards Analysis" to identify areas containing functions and systems required for safe shutdown of the plant (ref. 4)* An EXPLOSION (including a steam line explosion) that degrades the performance of a SAFETY SYSTEM train or visibly damages a SAFETY SYSTEM component or structure would be classified under this EAL. The need to classify a steam line break not considered an explosion itself is considered in fission product barrier degradation monitoring (EAL Category F).Generic This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant.The first condition addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available.
The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.The second condition addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components.
Operators will make this determination based on the totality of available event and damage report information.
This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage.Escalation of the ECL would be via IC FS1 or RS1.VCSNS Basis Reference(s):
- 1. ES-426 Earthquake Response Procedure Page 261 of 346 EPP-108 ENCLOSUREI REVISION 01 [DRAFT E]2. VCSNS IPE Internal Flooding Analysis Workbook 3. FSAR Section 3.3.1 4. VCSNS Fire Protection Evaluation Report 5. NEI 99-01 SA9 Page 262 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category F -Fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature
> 200'F);EALs in this category are applicable only in one or more hot operating modes.EALs in this category represent threats to the defense in depth design concept that precludes the release of highly radioactive fission products to the environment.
This concept relies on multiple physical barriers any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment.
The primary fission product barriers are: A. Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.B. Reactor Coolant System (RCS): The RCS Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves.C. Containment (CMT): The Containment Barrier includes the containment building and connections up to and including the outermost containment isolation valves.This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency.
The EALs in this category require evaluation of the loss and potential loss thresholds listed in the fission product barrier matrix of Table F-1 (Attachment 2). "Loss" and "Potential Loss" signify the relative damage and threat of damage to the barrier. "Loss" means the barrier no longer assures containment of radioactive materials. "Potential Loss" means integrity of the barrier is threatened and could be lost if conditions continue to degrade.The number of barriers that are lost or potentially lost and the following criteria determine the appropriate ECL: Alert: Any loss or any potential loss of either Fuel Clad or RCS Page 263 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Site Area Emergency:
Loss or potential loss of any two barriers General Emergency:
Loss of any two barriers and loss or potential loss of the third barrier The logic used for Category F EALs reflects the following considerations: " The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than the Containment Barrier.* Unusual Event ICs associated with fission product barriers are addressed in Recognition Category S.For accident conditions involving a radiological release, evaluation of the fission product barrier thresholds will need to be performed in conjunction with dose assessments to ensure correct and timely escalation of the emergency classification.
For example, an evaluation of the fission product barrier thresholds may result in a Site Area Emergency classification while a dose assessment may indicate that an EAL for General Emergency IC RG1 has been exceeded.The fission product barrier thresholds specified within a scheme reflect plant-specific VCSNS Unit 1 design and operating characteristics.
As used in this category, the term RCS leakage encompasses not just those types defined in Technical Specifications but also includes the loss of RCS mass to any location-inside the containment, a secondary-side system (i.e., steam generator tube leakage), an interfacing system, or outside of the containment building.
The release of liquid or steam mass from the RCS due to the as-designed/expected operation of a relief valve is not considered to be RCS leakage.At the Site Area Emergency level, classification decision-makers should maintain cognizance of how far present conditions are from meeting a threshold that would require a General Emergency declaration.
For example, if the Fuel Clad and RCS fission product barriers were both lost, then there should be frequent assessments of containment Page 264 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]radioactive inventory and integrity.
Alternatively, if both the Fuel Clad and RCS fission product barriers were potentially lost, the Emergency Director would have more assurance that there was no immediate need to escalate to a General Emergency.
Page 265 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
Fission Product Barrier Degradation Subcategory:
N/A Initiating Condition:
Any loss or any potential loss of either Fuel Clad or RCS EAL: FA1 .1 Aert Any loss or any potential loss of either Fuel Clad or RCS (Table F-i)Mode Applicability:
1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):
None Basis: Plant-Specific Fuel Clad, RCS and Containment comprise the fission product barriers.
Table F-1 (Attachment
- 2) lists the fission product barrier thresholds, bases and references.
At the Alert classification level, Fuel Clad and RCS barriers are weighted more heavily than the Containment barrier. Unlike the Containment barrier, loss or potential loss of either the Fuel Clad or RCS barrier may result in the relocation of radioactive materials or degradation of core cooling capability.
Note that the loss or potential loss of Containment barrier in combination with loss or potential loss of either Fuel Clad or RCS barrier results in declaration of a Site Area Emergency under EAL FS1.1.Generic None VCSNS Basis Reference(s):
- 1. NEI 99-01 FA1 Page 266 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
Fission Product Barrier Degradation Subcategory:
N/A Initiating Condition:
Loss or potential loss of any two barriers EAL: FS1.1 Site Area Emergency Loss or potential loss of any two barriers (Table F-i)Mode Applicability:
1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):
None Basis: Plant-Specific Fuel Clad, RCS and Containment comprise the fission product barriers.
Table F-1 (Attachment
- 2) lists the fission product barrier thresholds, bases and references.
At the Site Area Emergency classification level, each barrier is weighted equally. A Site Area Emergency is therefore appropriate for any combination of the following conditions: " One barrier loss and a second barrier loss (i.e., loss -loss)" One barrier loss and a second barrier potential loss (i.e., loss -potential loss)* One barrier potential loss and a second barrier potential loss (i.e., potential loss -potential loss)At the Site Area Emergency classification level, the ability to dynamically assess the proximity of present conditions with respect to the threshold for a General Emergency is important.
For example, the existence of Fuel Clad and RCS Barrier loss thresholds in addition to offsite dose assessments would require continual assessments of radioactive inventory and Containment integrity in anticipation of reaching a General Emergency classification.
Alternatively, if both Fuel Clad and RCS potential loss thresholds existed, Page 267 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]the Emergency Director would have greater assurance that escalation to a General Emergency is less imminent.Generic None VCSNS Basis Reference(s):
- 1. NEI 99-01 FS1 Page 268 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
Fission Product Barrier Degradation Subcategory:
N/A Initiating Condition:
Loss of any two barriers and loss or potential loss of the third barrier EAL: FG1.1 General Emergency Loss of any two barriers AND Loss or potential loss of the third barrier (Table F-i)Mode Applicability:
1 -Power Operation, 2 -Startup, 3 -Hot Shutdown, 4 -Hot Standby Definition(s):
None Basis: Plant-Specific Fuel Clad, RCS and Containment comprise the fission product barriers.
Table F-1 (Attachment
- 2) lists the fission product barrier thresholds, bases and references.
At the General Emergency classification level each barrier is weighted equally. A General Emergency is therefore appropriate for any combination of the following conditions:
- Loss of Fuel Clad, RCS and Containment barriers* Loss of Fuel Clad and RCS barriers with potential loss of Containment barrier" Loss of RCS and Containment barriers with potential loss of Fuel Clad barrier" Loss of Fuel Clad and Containment barriers with potential loss of RCS barrier Generic None VCSNS Basis Reference(s):
- 1. NEI 99-01 FG1 Page 269 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category I -Independent Spent Fuel Storage Installation (ISFSI)EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)An independent spent fuel storage installation (ISFSI) is a complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. A significant amount of the radioactive material contained within a canister must escape its packaging and enter the biosphere for there to be a significant environmental effect resulting from an accident involving the dry storage of spent nuclear fuel. Formal offsite planning is not required because the postulated worst-case accident involving an ISFSI has insignificant consequences to the public health and safety.A Notification of Unusual Event is declared on the basis of the occurrence of an event of sufficient magnitude that a loaded cask confinement boundary is damaged or violated.Page 270 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category:
ISFSI Subcategory:
Confinement Boundary Initiating Condition:
Damage to a loaded cask CONFINEMENT BOUNDARY EAL: IU1.1 Unusual Event Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than the following on the surface of the spent fuel cask (overpack):
- 60 mrem/hr (r + r0) on the top of the overpack* 600 mrem/hr (" + rj on the side of the overpack Mode Applicability:
All Definition(s):
CONFINEMENT BOUNDARY-.
The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As applied to the VCS ISFSI, the CONFINEMENT BOUNDARY is defined to be the HI-STORM Multi-Purpose Canister (MPC).Basis: Plant-Specific Overpacks are the casks which receive and contain the sealed MPCs for interim storage on the ISFSI. They provide gamma and neutron shielding, and provide for ventilated air flow to promote heat transfer from the MPC to the environs.
The term overpack does not include the transfer cask (ref. 1).The values shown represent 2 times the limits specified in the ISFSI Certificate of Compliance Technical Specification section 5.3.4 for radiation external to a loaded MPC overpack (ref. 1).Generic Page 271 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a MPC containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the sealed MPC is loaded into the storage cask (overpack).
The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage.The existence of "damage" is determined by radiological survey. The technical specification multiple of "2 times", which is also used in Recognition Category R IC RU1, is used here to distinguish between non-emergency and emergency conditions.
The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the "on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask.Security-related events for ISFSls are covered under ICs HU1 and HAl.VCSNS Basis Reference(s):
- 1. Certificate of Compliance No. 1032 Appendix A Technical Specifications for the HI-STORM FW MPC Storage System 2. NEI 99-01 E-HU1 Page 272 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]ATTACHMENT 2 FISSION PRODUCT BARRIER MATRIX AND BASES Page 273 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Introduction Table F-1 lists the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment).
The table is structured so that the three barriers occupy adjacent columns. Each fission product barrier column is further divided into two columns; one for Loss thresholds and one for Potential Loss thresholds.
The first column of the table (to the left of the Fuel Clad Barrier Loss column) lists the categories (types) of fission product barrier thresholds.
The fission product barrier categories are: 1. RCS or SG Tube Leakage 2. Inadequate Heat Removal 3. CMT Radiation
/ RCS Activity 4. CMT Integrity or Bypass 5. Emergency Director Judgment Each category occupies a row in Table F-1 thus forming a matrix defined by the category rows and the Loss/Potential Loss columns. The intersection of each category row with each Loss/Potential Loss column forms a cell in which one or more fission product barrier thresholds appear. If NEI 99-01 does not define a threshold for a barrier Loss/Potential Loss, the word "None" is entered in the cell.Thresholds are assigned letters within each Loss and Potential Loss column beginning with "A." In this manner, a threshold can be identified by its category number and threshold letter. For example, the first Fuel Clad barrier Loss in Category 2 is "FC Loss 2.A," the third Containment barrier Potential Loss in Category 4 is "CMT P-Loss 4.C," etc.If a cell in Table F-1 contains more than one threshold, each of the thresholds, if exceeded, signifies a Loss or Potential Loss of the barrier. It is not necessary to exceed all of the thresholds in a category before declaring a barrier Loss/Potential Loss.Page 274 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Subdivision of Table F-1 by category facilitates association of plant conditions to the applicable fission product barrier Loss and Potential Loss thresholds.
This structure promotes a systematic approach to assessing the classification status of the fission product barriers.When equipped with knowledge of plant conditions related to the fission product barriers, the EAL-user first scans down the category column of Table F-i, locates the likely category and then reads across the row of fission product barrier Loss and Potential Loss thresholds in that category to determine if any threshold has been exceeded.
If a threshold has not been exceeded in that category row, the EAL-user proceeds to the next likely category and continues review of the row of thresholds in the new category The EAL-user must examine each of the three fission product barriers to determine if other barrier thresholds in the category are lost or potentially lost. For example, if Reactor Building radiation is sufficiently high (i.e., > 20,000 R/hr), a Loss of the Fuel Clad and RCS barriers and a Potential Loss of the Containment barrier exist. Barrier Losses and Potential Losses are then applied to the algorithms given in EALs FG1.1, FS1.1 and FA1.1 to determine the appropriate emergency classification.
In the remainder of this Attachment, the Fuel Clad barrier threshold bases appear first, followed by the RCS barrier and finally the Containment barrier threshold bases. In each barrier, the bases are given according to category Loss followed by category Potential Loss beginning with Category 1, then 2.. .5.Page 275 of 346 EPP-1 08 ENCLOSURE I REVISION 01 [DRAFT E]Table F-1 Fission Product Barrier Threshold Matrix Fuel Clad Barrier Reactor Coolant System Barrier Containment Barrier Category Loss Potential Loss Loss Potential Loss Loss Potential Loss A. An automatic or manual ECCS A. Operation of a standby charging (SI) actuation required by pump is required by EITHER: EITHER:
- UNISOLABLE RCS leakage A. A leaking or RUPTURED SG is RCS or None None Nn T UNISOLABLE RCS -SG tube RUPTURE FAULTED outside of containment Leakage leakage B. CSFST Integrity-RED path L SG tube RUPTURE conditions met A. CSFST Core Cooling-ORANGE path conditions met A. CSFST Heat Sink-RED path A. CSFST Core Cooling-RED path Inadequat A. CSFST Core Cooling-RED B. CSFST Heat Sink-RED path conditions met conditions met deat path conditions met conditions met None AND None AND Heat AND Heat sink is required Restoration procedures not Removal Heat sink is required effective within 15 min. (Note 1)3 A. RM-G7 or RM-G18 CNTMT HI CMT RNG Gamma > 2,000 R/hr A. RM-G7 or RM-G18 CNTMT HI A. RM-G7 or RM-G18 CNTMT HI RNG Radiation B. Dose equivalent 1-131 coolant None RNG Gamma > 100 R/hr None None Gamma > 20,000 R/hr/ RCS activity > 300 pCi/gm Activity A. Containment isolation is required A. CSFST Containment-RED path AND EITHER: conditions met 4 .Containment integrity has B. Containment hydrogen concentratior been lost based on ED > 4%CMT None None None None judgment C. Containment pressure > 12 psig Integrity
- UNISOLABLE pathway from AND or Bypass containment to the environment
< one full train of depressurization exists equipment (Table F-2) is operating B. Indications of RCS leakage per design for 2 15 min. (Note 1)outside of containment 5 A. Any condition in the opinion of A. Any condition in the opinion of A. Any condition in the opinion of A. Any condition in the opinion of the A. Any condition in the opinion of A. Any condition in the opinion of the the ED that indicates loss of the the ED that indicates potential the ED that indicates loss of the ED that indicates potential loss of the ED that indicates loss of the ED that indicates potential loss of ED fuel clad barrier loss of the fuel clad barrier RCS barrier the RCS barrier containment barrier the containment barrier Judgment Table F-2 Full Train Depressurization Equipment RBCU Groups Containment Sprays Operating Operating 2 0 1 1 0 2 Page 276 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Fuel Clad Category:
None Definition(s):
N/A Basis: Plant-Specific N/A Generic N/A VCSNS Basis Reference(s):
N/A Page 277 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Fuel Clad Category:
None Definition(s):
N/A Basis: Plant-Specific N/A Generic N/A VCSNS Basis Reference(s):
N/A Page 278 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Fuel Clad Category:
- 2. Inadequate Heat Removal Degradation Threat: Loss Threshold:
A. CSFST Core Cooling-RED path conditions met Definition(s):
None Basis: Plant-Specific Critical Safety Function Status Tree (CSFST) Core Cooling-RED path is given in Figure 5 and indicates significant core exit superheating and core uncovery.
The CSFSTs are normally monitored using the SPDS display on the Integrated Plant Computer System (IPCS). (ref. 1)The setpoint values provided in brackets following the normal setpoint values in the CSFSTs are used under adverse containment conditions.
Adverse containment conditions are defined as either (ref. 4): " Hi-1 RB pressure (_> 3.6 psig), or* Containment Hi-radiation
(_ 100,000 Rads integrated dose)The IPCS monitors these parameters and indicates when adverse containment values should be used. The instruments available to monitor these containment parameters are Containment pressure on PI-950, PI-951, PI-952 or PI-953, and Containment radiation on RM-G7 or RM-G18. When no adverse values are provided, the given setpoint is used for any containment condition, as the parameter measurement is independent of containment atmosphere.
If containment pressure decreases below the adverse pressure setpoint after it has been exceeded, the normal values are again used. Once the adverse radiation level setpoint is exceeded, the adverse containment values must be utilized through event Page 279 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]recovery and establishment of normal operating procedures.
Engineering should then be requested to evaluate instrumentation inaccuracies (ref. 3).Generic This reading indicates temperatures within the core are sufficient to cause significant superheating of reactor coolant.VCSNS Basis Reference(s):
- 1. EOP-1 2.0 Monitoring of Critical Safety Functions 2. EOP-14.0 Response to Inadequate Core Cooling 3. EOP-14.1 Response to Degraded Core Cooling 4. OAP-103.4 EOP/AOP User's Guide 5. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.A Page 280 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Fuel Clad Category:
- 2. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:
A. CSFST Core Cooling-ORANGE path conditions met Definition(s):
None Basis: Plant-Specific Critical Safety Function Status Tree (CSFST) Core Cooling-ORANGE path is given in Figure 5 and indicates subcooling has been lost and that some fuel clad damage may potentially occur (ref. 1).The CSFSTs are normally monitored using the SPDS display on the Integrated Plant Computer System (IPCS) (ref. 1).The setpoint values provided in brackets following the normal setpoint values in the CSFSTs are used under adverse containment conditions.
Adverse containment conditions are defined as either (ref. 4):* Hi-1 RB pressure (> 3.6 psig), or* Containment Hi-radiation
(__ 100,000 Rads integrated dose)The IPCS monitors these parameters and indicates when adverse containment values should be used. The instruments available to monitor these containment parameters are Containment pressure on PI-950, PI-951, PI-952 or PI-953, and Containment radiation on RM-G7 or RM-G18. When no adverse values are provided, the given setpoint is used for any containment condition, as the parameter measurement is independent of containment atmosphere.
If containment pressure decreases below the adverse pressure setpoint after it has been exceeded, the normal values are again used. Once the adverse radiation level setpoint is exceeded, the adverse containment values must be utilized through event Page 281 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]recovery and establishment of normal operating procedures.
Engineering should then be requested to evaluate instrumentation inaccuracies (ref. 3).Generic This reading indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.VCSNS Basis Reference(s):
- 1. EOP-1 2.0 Monitoring of Critical Safety Functions 2. EOP-14.0 Response to Inadequate Core Cooling 3. EOP-14.1 Response to Degraded Core Cooling 4. OAP-103.4 EOP/AOP User's Guide 5. NEI 99-01 RCS or SG Tube Leakage Potential Loss 1.A Page 282 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Fuel Clad Category:
- 2. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:
B. CSFST Heat Sink-RED path conditions met AND Heat sink is required Definition(s):
None Basis: Plant-Specific In combination with RCS Potential Loss 2.A, meeting this threshold results in a Site Area Emergency.
Critical Safety Function Status Tree (CSFST) Heat Sink-RED path is given in Figure 6 and indicates the ultimate heat sink function is under extreme challenge and that some fuel clad damage may potentially occur (ref. 1).The CSFSTs are normally monitored using the SPDS display on the Integrated Plant Computer System (IPCS) (ref. 1).The phrase "and heat sink required" precludes the need for classification for conditions in which RCS pressure is less than SG pressure or Heat Sink-RED path entry was created through operator action directed by an EOP. For example, EOP-1 5.0 is entered from CSFST Heat Sink-Red.
Step 1 tells the operator to determine if heat sink is required by checking that RCS pressure is greater than any non-faulted SG pressure and RCS Thot is greater than 350 0 F. If these conditions exist, Heat Sink is required.
Otherwise, the operator is to either return to the procedure and step in effect or place RHR in service for heat removal. For large LOCA events inside the Containment, the SGs are moot because heat removal through the containment heat removal systems takes place. Therefore, Heat Sink Page 283 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Red should not be required and, should not be assessed for EAL classification because a LOCA event alone should not require higher than an Alert classification. (ref. 2)The setpoint values provided in brackets following the normal setpoint values in the CSFSTs are used under adverse containment conditions.
Adverse containment conditions are defined as either (ref. 3): " Hi-1 RB pressure (__ 3.6 psig), or* Containment Hi-radiation
(> 100,000 Rads integrated dose)The IPCS monitors these parameters and indicates when adverse containment values should be used. The instruments available to monitor these containment parameters are Containment pressure on PI-950, PI-951, PI-952 or PI-953, and Containment radiation on RM-G7 or RM-G18. When no adverse values are provided, the given setpoint is used for any containment condition, as the parameter measurement is independent of containment atmosphere.
If containment pressure decreases below the adverse pressure setpoint after it has been exceeded, the normal values are again used. Once the adverse radiation level setpoint is exceeded, the adverse containment values must be utilized through event recovery and establishment of normal operating procedures.
Engineering should then be requested to evaluate instrumentation inaccuracies.
Generic This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the Fuel Clad Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.
VCSNS Basis Reference(s):
- 1. EOP-12.0 Monitoring of Critical Safety Functions 2. EOP-15.0 Response to Loss of Secondary Heat Sink 3. OAP-103.4 EOP/AOP User's Guide 4. NEI 99-01 Inadequate Heat Removal Fuel Clad Potential Loss 2.B Page 284 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Fuel Clad Category:
- 3. CMT Radiation
/ RCS Activity Degradation Threat: Loss Threshold:
A. RM-G7 or RM-G18 CNTMT HI RNG Gamma > 2,000 R/hr Definition(s):
None Basis: Plant-Specific The gamma dose rate resulting from a postulated loss of coolant accident (LOCA) is monitored by the high range Reactor Building monitors, RM-G7 and -G18. RM-G18 and RM-G7 are located inside containment outside the A and B Steam Generator cubicles, respectively, on the 469' elevation.
The detector range is approximately 1 to 1 E7 R/hr (logarithmic scale). Radiation Monitors RM-G18 and RM-G7 provide a diverse means of measuring the containment for high level gamma radiation.
The detectors for RM-G18 and RM-G7 are a stainless steel gamma sensitive ion-chambers (ref. 1).The value specified represents, based on Microshield calculations, the expected containment high range radiation monitor (RM-G7 and RM-G18) response (2.1E3 R/hr rounded down to nearest whole number for readability) based on a LOCA (Reg. Guide 1.4 Case LOCA with fuel failure), one hour after shutdown with -2% fuel failure (ref. 2).Generic The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300 pCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.Page 285 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold 3.A since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the ECL to a Site Area Emergency.
VCSNS Basis Reference(s):
- 1. VCSNS Design Bases Document -Radiation Monitoring System (RM)2. TWR 11.5-91-027 Accident Severity Estimates from RM-G7 and G18 Response 3. NEI 99-01 CMT Radiation
/ RCS Activity Fuel Clad Loss 3.A Page 286 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Fuel Clad Category:
- 3. CMT Radiation
/ RCS Activity Degradation Threat: Loss Threshold:
B. Dose equivalent 1-131 coolant activity > 300 pCi/gm Definition(s):
None Basis: Plant-Specific Elevated reactor coolant activity represents a potential degradation in the level of safety of the plant and a potential precursor of more serious problems.
The threshold dose equivalent 1-131 concentration is well above that expected for iodine spikes and corresponds to about 5% fuel clad damage. When reactor coolant activity reaches this level the Fuel Clad barrier is considered lost. (ref. 1)Generic This threshold indicates that RCS radioactivity concentration is greater than 300 pCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.There is no Potential Loss threshold associated with RCS Activity / Containment Radiation.
VCSNS Basis Reference(s):
- 1. NEI 99-01 CMT Radiation
/ RCS Activity Fuel Clad Loss 3.B Page 287 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Category: Degradation Threat: Threshold:
Fuel Clad 3. CMT Radiation
/ RCS Activity Potential Loss None Definition(s):
N/A Basis: Plant-Specific N/A Generic N/A VCSNS Basis Reference(s):
N/A Page 288 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Fuel Clad Category:
- 4. CMT Integrity or Bypass Degradation Threat: Loss Threshold:
None Definition(s):
N/A Basis: Plant-Specific N/A Generic N/A VCSNS Basis Reference(s):
N/A Page 289 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Fuel Clad Category:
- 4. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
None Definition(s):
N/A Basis: Plant-Specific N/A Generic N/A VCSNS Basis Reference(s):
N/A Page 290 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Fuel Clad Category:
- 5. ED Judgment Degradation Threat: Loss Threshold:
A. Any condition in the opinion of the ED that indicates loss of the fuel clad barrier Definition(s):
None Basis: Plant-Specific None Generic This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Fuel Clad Barrier is lost.VCSNS Basis Reference(s):
- 1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A Page 291 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Fuel Clad Category:
- 5. ED Judgment Degradation Threat: Potential Loss Threshold:
A. Any condition in the opinion of the ED that indicates potential loss of the fuel clad barrier Definition(s):
None Basis: Plant-Specific None Generic This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Fuel Clad Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
VCSNS Basis Reference(s):
- 1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A Page 292 of 346 EPP-1 08 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Reactor Coolant System Category:
A. An automatic or manual ECCS (SI) actuation required by EITHER:* UNISOLABLE RCS leakage* SGtube RUPTURE Definition(s):
UNISOLABLE
-An open or breached system line that cannot be isolated, remotely or locally.RUPTURE -The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
Basis: Plant-Specific ECCS (SI) actuation is caused by (ref. 1):* Pressurizer pressure < 1850 psig* Steamline pressure < 675 psig" Steamline differential pressure > 97 psid* Reactor Building (Containment) pressure _> 3.6 psig Generic This threshold is based on an UNISOLABLE RCS leak of sufficient size to require an automatic or manual actuation of the Emergency Core Cooling System (ECCS). This condition clearly represents a loss of the RCS Barrier.This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an Page 293 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]interfacing system. The mass loss may be into any location -inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.
A steam generator with primary-to-secondary leakage of sufficient magnitude to require an safety injection is considered to be RUPTURED.
If a RUPTURED steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold 1 .A will also be met.VCSNS Basis Reference(s):
- 1. EOP-1.0 Reactor Trip/Safety Injection Actuation 2. EOP-4.0 Steam Generator Tube Rupture 3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Loss 1.A Page 294 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Reactor Coolant System Category:
A. Operation of a standby charging pump is required by EITHER: " UNISOLABLE RCS leakage" SG tube RUPTURE Definition(s):
UNISOLABLE
-An open or breached system line that cannot be isolated, remotely or locally.RUPTURE -The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
Basis: Plant-Specific The Chemical and Volume Control System (CVCS) includes three charging pumps which take suction from the Volume Control Tank and return cooled, purified reactor coolant to the RCS. Normal charging flow is handled by one of the three charging pumps. Each charging pump is designed for a flow rate of 150 gpm at 2520 psid and a maximum flow rate of 650 gpm at 1040 psid. A second charging pump being required is indicative of a substantial RCS leak. (ref. 1, 2, 3, 4)Generic This threshold is based on an UNISOLABLE RCS leak that results in the inability to maintain pressurizer level within specified limits by operation of a normally used charging (makeup) pump, but an ECCS (SI) actuation has not occurred.
The threshold is met when an operating procedure, or operating crew supervision, directs that a standby charging (makeup) pump be placed in service to restore and maintain pressurizer level.Page 295 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location -inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.
If a leaking steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold 1 .A will also be met.VCSNS Basis Reference(s):
- 1. SOP-1 02 Chemical and Volume Control System 2. FSAR Section 9.3.4.1.6 3. FSAR Section 9.3.4.2.1 4. FSAR Table 9.3-4 5. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.A Page 296 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Reactor Coolant System Category:
B. CSFST Integrity-RED path conditions met Definition(s):
None Basis: Plant-Specific The "Potential Loss" threshold is defined by the CSFST RCS Integrity
-RED path (Figure 7). The values in this EAL are consistent with the CSFST value (ref. 1). CSFST RCS Integrity
-Red Path plant and associated Operational Curve Limit A is given in Figures 7 and 8 and indicates an extreme challenge to the safety function when plant parameters are to the right of the limit curve following excessive RCS cooldown under pressure (ref. 1).The CSFSTs are normally monitored using the SPDS display on the Integrated Plant Computer System (IPCS) (ref. 1).Generic This condition indicates an extreme challenge to the integrity of the RCS pressure boundary due to pressurized thermal shock -a transient that causes rapid RCS cooldown while the RCS is in Mode 3 or higher (i.e., hot and pressurized).
VCSNS Basis Reference(s):
- 1. EOP-12.0 Monitoring of Critical Safety Functions 2. OAP-103.4 EOP/AOP User's Guide 3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1..B Page 297 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Reactor Coolant System Category:
- 2. Inadequate Heat Removal Degradation Threat: Loss Threshold:
None Definition(s):
N/A Basis: Plant-Specific N/A Generic N/A VCSNS Basis Reference(s):
N/A Page 298 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Reactor Coolant System Category:
- 2. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:
A. CSFST Heat Sink-RED path conditions met AND Heat sink is required Definition(s):
None Basis: Plant-Specific In combination with Fuel Clad Potential Loss 2.B, meeting this threshold results in a Site Area Emergency.
Critical Safety Function Status Tree (CSFST) Heat Sink-RED path is given in Figure 6 and indicates the ultimate heat sink function is under extreme challenge and that some fuel clad damage may potentially occur (ref. 1).The CSFSTs are normally monitored using the SPDS display on the Integrated Plant Computer System (IPCS) (ref. 1).The phrase "and heat sink required" precludes the need for classification for conditions in which RCS pressure is less than SG pressure or Heat Sink-RED path entry was created through operator action directed by an EOP. For example, EOP-1 5.0 is entered from CSFST Heat Sink-Red.
Step 1 tells the operator to determine if heat sink is required by checking that RCS pressure is greater than any non-faulted SG pressure and RCS Thot is greater than 350 0 F. If these conditions exist, Heat Sink is required.
Otherwise, the operator is to either return to the procedure and step in effect or place RHR in service for heat removal. For large LOCA events inside the Containment, the SGs are moot because heat removal through the containment heat removal systems takes place. Therefore, Heat Sink Page 299 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Red should not be required and, should not be assessed for EAL classification because a LOCA event alone should not require higher than an Alert classification. (ref. 2)The setpoint values provided in brackets following the normal setpoint values in the CSFSTs are used under adverse containment conditions.
Adverse containment conditions are defined as either (ref. 3):* Hi-1 RB pressure (> 3.6 psig), or" Containment Hi-radiation
(> 100,000 Rads integrated dose)The IPCS monitors these parameters and indicates when adverse containment values should be used. The instruments available to monitor these containment parameters are Containment pressure on PI-950, PI-951, PI-952 or PI-953, and Containment radiation on RM-G7 or RM-G18. When no adverse values are provided, the given setpoint is used for any containment condition, as the parameter measurement is independent of containment atmosphere.
If containment pressure decreases below the adverse pressure setpoint after it has been exceeded, the normal values are again used. Once the adverse radiation level setpoint is exceeded, the adverse containment values must be utilized through event recovery and establishment of normal operating procedures.
Engineering should then be requested to evaluate instrumentation inaccuracies.
Generic This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the RCS Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.
Meeting this threshold results in a Site Area Emergency because this threshold is identical to Fuel Clad Barrier Potential Loss threshold 2.B; both will be met. This condition warrants a Site Area Emergency declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and increase RCS pressure to the point where mass will be lost from the system.Page 300 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]VCSNS Basis Reference(s):
- 1. EOP-1 2.0 Monitoring of Critical Safety Functions 2. EOP-15.0 Response to Loss of Secondary Heat Sink 3. OAP-103.4 EOP/AOP User's Guide 4. NEI 99-01 Inadequate Heat Removal Potential Reactor Coolant System Loss 2.A.Page 301 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Reactor Coolant System Category:
- 3. CMT Radiation
/ RCS Activity Degradation Threat: Loss Threshold:
A. RM-G7 or RM-G18 CNTMT HI RNG Gamma> 100 R/hr Definition(s):
None Basis: Plant-Specific The gamma dose rate resulting from a postulated loss of coolant accident (LOCA) is monitored by the high range Reactor Building monitors, RM-G7 and -G18.:RM-G18 and RM-G7 are located inside containment outside the A and B Steam Generator cubicles, respectively, on the 469' elevation.
The detector range is approximately 1 to 1 E7 R/hr (logarithmic scale). Radiation Monitors RM-G18 and RM-G7 provide a diverse means of measuring the containment for high level gamma radiation.
The detectors for RM-G18 and RM-G7 are a stainless steel gamma sensitive ion-chambers (ref. 1).The value specified represents, based on Microshield calculations, the expected containment high range radiation monitor (RM-G7 and RM-G18) response (Reg. Guide 1.4 Case LOCA with fuel failure) (1.05E2 R/hr rounded down to nearest whole number for readability) based on a LOCA, one hour after shutdown with -0.1% fuel failure (ref. 2).Generic The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold 3.A since it indicates a loss of the RCS Barrier only.There is no Potential Loss threshold associated with RCS Activity / Containment Radiation.
Page 302 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]VCSNS Basis Reference(s):
- 1. VCSNS Design Bases Document -Radiation Monitoring System (RM)2. TWR 11.5-91-027 Accident Severity Estimates from RM-G7 and G18 Response 3. NEI 99-01 CMT Radiation
/ RCS Activity Reactor Coolant System Loss 2.A Page 303 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Category: Degradation Threat: Threshold:
Reactor Coolant System 3. CMT Radiation
/ RCS Activity Potential Loss None Definition(s):
N/A Basis: Plant-Specific N/A Generic N/A VCSNS Basis Reference(s):
N/A Page 304 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Reactor Coolant System Category:
- 4. CMT Integrity or Bypass Degradation Threat: Loss Threshold:
None Definition(s):
N/A Basis: Plant-Specific N/A Generic N/A VCSNS Basis Reference(s):
N/A Page 305 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Reactor Coolant System Category:
- 4. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
None Definition(s):
N/A Basis: Plant-Specific N/A Generic N/A VCSNS Basis Reference(s):
N/A Page 306 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT El Barrier: Reactor Coolant System Category:
- 5. ED Judgment Degradation Threat: Loss Threshold:
A. Any condition in the opinion of the ED that indicates loss of the RCS barrier Definition(s):
None Basis: Plant-Specific None Generic This threshold addresses any other factors that may be used by the Emergency Director in-determining whether the RCS Barrier is lost.VCSNS Basis Reference(s):
- 1. NEI 99-01 Emergency Director Judgment Reactor Coolant System Loss 6.A Page 307 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Reactor Coolant System Category:
- 5. ED Judgment Degradation Threat: Potential Loss Threshold:
A. Any condition in the opinion of the ED that indicates potential loss of the RCS barrier Definition(s):
None Basis: Plant-Specific None Generic This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
VCSNS Basis Reference(s):
- 1. NEI 99-01 Emergency Director Judgment Reactor Coolant System Potential Loss 6.A Page 308 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Containment Category:
A. A leaking or RUPTURED SG is FAULTED outside of containment Definition(s):
FAULTED -The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.
RUPTURED -The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
Basis: Plant-Specific None Generic This threshold addresses a leaking or RUPTURED Steam Generator (SG) that is also FAULTED outside of containment.
The condition of the SG, whether leaking or RUPTURED, is determined in accordance with the thresholds for RCS Barrier Potential Loss 1.A and Loss 1.A, respectively.
This condition represents a bypass of the containment barrier.FAULTED is a defined term within the NEI 99-01 methodology; this determination is not necessarily dependent upon entry into, or diagnostic steps within, an EOP. For example, if the pressure in a steam generator is decreasing uncontrollably (part of the FAULTED definition) and the FAULTED steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher Page 309 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]priority condition, the steam generator is still considered FAULTED for emergency classification purposes.The FAULTED criterion establishes an appropriate lower bound on the size of a steam release that may require an emergency classification.
Steam releases of this size are readily observable with normal Control Room indications.
The lower bound for this aspect of the containment barrier is analogous to the lower bound criteria specified in IC SU4 for the fuel clad barrier (i.e., RCS activity values) and IC SU5 for the RCS barrier (i.e., RCS leak rate values).This threshold also applies to prolonged steam releases necessitated by operational considerations such as the forced steaming of a leaking or RUPTURED steam generator directly to atmosphere to cooldown the plant, or to drive an auxiliary (emergency) feed water pump. These types of conditions will result in a significant and sustained release of radioactive steam to the environment (and are thus similar to a FAULTED condition).
The inability to isolate the steam flow without an adverse effect on plant cooldown meets the intent of a loss of containment.
Steam releases associated with the expected operation of a SG power operated relief valve or safety relief valve do not meet the intent of this threshold.
Such releases may occur intermittently for a short period of time following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown.
Steam releases associated with the unexpected operation of a valve (e.g., a stuck-open safety valve) do meet this threshold.
Following an SG tube leak or rupture, there may be minor radiological releases through a secondary-side system component (e.g., air ejectors, glad seal exhausters, valve packing, etc.). These types of releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.The ECLs resulting from primary-to-secondary leakage, with or without a steam release from the FAULTED SG, are summarized below.Page 310 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]P-to-S Leak Rate Less than or equal to 25 gpm Greater than 25 gpm Requires operation of a standby charging (makeup) pump (RCS Barrier Potential Loss)Requires an automatic or manual ECCS (SI) actuation (RCS Barrier Loss)Affected SG is FAULTED Outside of Containment?
Yes No No classification No classification Unusual Event per SU5 Unusual Event per SU5 Site Area Emergency per FS1 Site Area Emergency per FS1 Alert per FA1 Alert per FA1 There is no Potential Loss threshold associated with RCS or SG Tube Leakage.VCSNS Basis Reference(s):
- 1. EOP-3.0 Faulted Steam Generator Isolation 2. EOP-4.0 Steam Generator Tube Rupture 3. NEI 99-01 RCS or SG Tube Leakage Containment Loss 1.A Page 311 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Containment Category:
None Definition(s):
N/A Basis: Plant-Specific N/A Generic N/A VCSNS Basis Reference(s):
N/A Page 312 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Containment Category:
- 2. Inadequate Heat Removal Degradation Threat: Loss Threshold:
None Definition(s):
N/A Basis: Plant-Specific N/A Generic N/A VCSNS Basis Reference(s):
N/A Page 313 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Containment Category:
- 2. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:
A. CSFST Core Cooling-RED path conditions met AND Restoration procedures not effective within 15 min. (Note 1)Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Definition(s):
None Basis: Plant-Specific Critical Safety Function Status Tree (CSFST) Core Cooling-RED path is given in Figure 5 and indicates significant core exit superheating and core uncovery.
The CSFSTs are normally monitored using the SPDS display on the Integrated Plant Computer System (IPCS). (ref. 1)The function restoration procedures are those emergency operating procedures that address the recovery of the core cooling critical safety functions.
The procedure is considered effective if the temperature is decreasing or if the vessel water level is increasing (ref. 1, 2, 3).A direct correlation to status trees can be made if the effectiveness of the restoration procedures is also evaluated.
The setpoint values provided in brackets following the normal setpoint values in the CSFSTs are used under adverse containment conditions.
Adverse containment conditions are defined as either (ref. 4): 0 Hi-1 RB pressure (> 3.6 psig), or Page 314 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]* Containment Hi-radiation
(__ 100,000 Rads integrated dose)The IPCS monitors these parameters and indicates when adverse containment values should be used. The instruments available to monitor these containment parameters are Containment pressure on PI-950, PI-951, PI-952 or PI-953, and Containment radiation on RM-G7 or RM-G 18. When no adverse values are provided, the given setpoint is used for any containment condition, as the parameter measurement is independent of containment atmosphere.
If containment pressure decreases below the adverse pressure setpoint after it has been exceeded, the normal values are again used. Once the adverse radiation level setpoint is exceeded, the adverse containment values must be utilized through event recovery and establishment of normal operating procedures.
Engineering should then be requested to evaluate instrumentation inaccuracies (ref. 3, 4).This threshold indicates significant core exit superheating and core uncovery.
If core exit thermocouple (TC) readings are greater than 1,200 0 F (ref. 1), Fuel Clad barrier is also lost.Generic This condition represents an IMMINENT core melt sequence which, if not corrected, could lead to vessel failure and an increased potential for containment failure. For this condition to occur, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. If implementation of a procedure(s) to restore adequate core cooling is not effective (successful) within 15 minutes, it is assumed that the event trajectory will likely lead to core melting and a subsequent challenge of the Containment Barrier.The restoration procedure is considered "effective" if core exit thermocouple readings are decreasing and/or if reactor vessel level is increasing.
Whether or not the procedure(s) will be effective should be apparent within 15 minutes. The Emergency Director should escalate the ECL as soon as it is determined that the procedure(s) will not be effective.
Severe accident analyses (e.g., NUREG-1 150) have concluded that function restoration procedures can arrest core degradation in a significant fraction of core damage scenarios, and that the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide 15 minutes beyond the required entry point to determine if procedural actions can reverse the core melt sequence.Page 315 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]VCSNS Basis Reference(s):
- 1. EOP-12.0 Monitoring of Critical Safety Functions 2. EOP-14.0 Response to Inadequate Core Cooling 3. EOP-14.1 Response to Degraded Core Cooling 4. OAP-103.4 EOP/AOP User's Guide 54. NEI 99-01 Inadequate Heat Removal Containment Potential Loss 2.A Page 316 of 346 EPP-1 08 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Category: Degradation Threat: Threshold:
Containment
- 3. CMT Radiation
/ RCS Activity Loss None Definition(s):
N/A Basis: Plant-Specific N/A Generic N/A VCSNS Basis Reference(s):
N/A Page 317 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Containment Category:
- 3. CMT Radiation
/ RCS Activity Degradation Threat: Potential Loss Threshold:
A. RM-G7 or RM-G18 CNTMT HI RNG Gamma > 20,000 R/hr Definition(s):
None Basis: Plant-Specific The gamma dose rate resulting from a postulated loss of coolant accident (LOCA) is monitored by the high range Reactor Building monitors, RM-G7 and -G 18. RM-G18 and RM-G7 are located inside containment outside the A and B Steam Generator cubicles, respectively, on the 469' elevation.
The detector range is approximately 1 to 1 E7 R/hr (logarithmic scale). Radiation Monitors RM-G1 8 and RM-G7 provide a diverse means of measuring the containment for high level gamma radiation.
The detectors for RM-G18 and RM-G7 are a stainless steel gamma sensitive ion-chambers (ref. 1).The value specified represents, based on Microshield calculations, the expected containment high range radiation monitor (RM-G7 and RM-G18) response (Reg. Guide 1.4 Case LOCA with fuel failure) based on a LOCA, one hour after shutdown with -20% fuel failure (ref. 2).2.1 x 10 4 R/hr (rounded down the the nearest whole number) is a value which indicates significant fuel damage well in excess of the thresholds associated with both loss of Fuel Clad and loss of RCS barriers.
A major release of radioactivity requiring off-site protective actions from core damage is not possible unless a major failure of fuel cladding allows radioactive material to be released from the core into the reactor coolant (ref. 1, 2).Generic Page 318 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds.
NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the ECL to a General Emergency.
VCSNS Basis Reference(s):
- 1. VCSNS Design Bases Document -Radiation Monitoring System (RM)2. TWR 11.5-91-027 Accident Severity Estimates from RM-G7 and G18 Response 3. NEI 99-01 CMT Radiation
/ RCS Activity Containment Potential Loss 3.A Page 319 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Containment Category:
- 4. CMT Integrity or Bypass Degradation Threat: Loss Threshold:
A. Containment isolation is required AND EITHER: " Containment integrity has been lost based on ED judgment" UNISOLABLE pathway from containment to the environment exists Definition(s):
UNISOLABLE
-An open or breached system line that cannot be isolated, remotely or locally.Basis: Plant-Specific None Generic These thresholds address a situation where containment isolation is required and one of two conditions exists as discussed below. Users are reminded that there may be accident and release conditions that simultaneously meet both bulleted thresholds.
First Threshold
-Containment integrity has been lost, i.e., the actual containment atmospheric leak rate likely exceeds that associated with allowable leakage (or sometimes referred to as design leakage).
Following the release of RCS mass into containment, containment pressure will fluctuate based on a variety of factors; a loss of containment integrity condition may (or may not) be accompanied by a noticeable drop in containment pressure.
Recognizing the inherent difficulties in determining a containment leak rate during accident conditions, it is expected that the Emergency Director will assess this threshold using judgment, and with due consideration given to current plant conditions, and available operational and radiological data (e.g., containment pressure, readings on Page 320 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]radiation monitors outside containment, operating status of containment pressure control equipment, etc.).Refer to the middle piping run of Figure 10. Two simplified examples are provided.
One is leakage from a penetration and the other is leakage from an in-service system valve.Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure.Another example would be a loss or potential loss of the RCS barrier, and the simultaneous occurrence of two FAULTED locations on a steam generator where one fault is located inside containment (e.g., on a steam or feedwater line) and the other outside of containment.
In this case, the associated steam line provides a pathway for the containment atmosphere to escape to an area outside the containment.
Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components.
These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.Second Threshold
-Conditions are such that there is an UNISOLABLE pathway for the migration of radioactive material from the containment atmosphere to the environment.
As used here, the term "environment" includes the atmosphere of a room or area, outside the containment, that may, in turn, communicate with the outside-the-plant atmosphere (e.g., through discharge of a ventilation system or atmospheric leakage).
Depending upon a variety of factors, this condition may or may not be accompanied by a noticeable drop in containment pressure.Refer to the top piping run of Figure 10. In this simplified example, the inboard and outboard isolation valves remained open after a containment isolation was required (i.e., containment isolation was not successful).
There is now an UNISOLABLE pathway from the containment to the environment.
The existence of a filter is not considered in the threshold assessment.
Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to Page 321 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream.Leakage between two interfacing liquid systems, by itself, does not meet this threshold.
Refer to the bottom piping run of Figure 10. In this simplified example, leakage in an RCP seal cooler is allowing radioactive material to enter the Auxiliary Building.
The radioactivity would be detected by the Process Monitor. If there is no leakage from the closed water cooling system to the Auxiliary Building, then no threshold has been met. If the pump developed a leak that allowed steam/water to enter the Auxiliary Building, then second threshold would be met. Depending upon radiation monitor locations and sensitivities, this leakage could be detected by any of the four monitors depicted in the figure and cause the first threshold to be met as well.Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable containment leakage through various penetrations or system components.
Minor releases may also occur if a containment isolation valve(s) fails to close but the containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.The status of the containment barrier during an event involving steam generator tube leakage is assessed using Loss Threshold 1.A.VCSNS Basis Reference(s):
- 2. NEI 99-01 CMT Integrity or Bypass Containment Loss 4.A Page 322 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Containment Category:
- 4. CMT Integrity or Bypass Degradation Threat: Loss Threshold:
B. Indications of RCS leakage outside of containment Definition(s):
None Basis: Plant-Specific EOP-2.5 LOCA Outside Containment (ref. 1) provides instructions to identify and isolate a LOCA outside of the containment.
Potential RCS leak pathways outside containment include: " RHR* CVCS/Letdown
- RCP seals" PZR/RCS Loop sample lines Generic Containment sump, temperature, pressure and/or radiation levels will increase if reactor coolant mass is leaking into the containment.
If these parameters have not increased, then the reactor coolant mass may be leaking outside of containment (i.e., a containment bypass sequence).
Increases in sump, temperature, pressure, flow and/or radiation level readings outside of the containment may indicate that the RCS mass is being lost outside of containment.
Unexpected elevated readings and alarms on radiation monitors with detectors outside containment should be corroborated with other available indications to confirm that the source is a loss of RCS mass outside of containment.
If the fuel clad barrier has not been Page 323 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]lost, radiation monitor readings outside of containment may not increase significantly; however, other unexpected changes in sump levels, area temperatures or pressures, flow rates, etc. should be sufficient to determine if RCS mass is being lost outside of the containment.
Refer to the middle piping run of Figure 10. In this simplified example, a leak has occurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building.Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure and cause threshold 4.A to be met as well.To ensure proper escalation of the emergency classification, the RCS leakage outside of containment must be related to the mass loss that is causing the RCS Loss and/or Potential Loss threshold 1 .A to be met.VCSNS Basis Reference(s):
- 2. NEI 99-01 CMT Integrity or Bypass Containment Loss 4.B Page 324 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Containment Category:
- 4. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
A. CSFST Containment-RED path conditions met Definition(s):
None Basis: Plant-Specific Critical Safety Function Status Tree (CSFST) Containment-RED path (Figure 9) is entered if Containment pressure is greater than or equal to 55 psig and represents an extreme challenge to safety function (ref. 1).55 psig is based on the containment design pressure (ref. 2).Generic If containment pressure exceeds the design pressure, there exists a potential to lose the Containment Barrier. To reach this level, there must be an inadequate core cooling condition for an extended period of time; therefore, the RCS and Fuel Clad barriers would already be lost. Thus, this threshold is a discriminator between a Site Area Emergency and General Emergency since there is now a potential to lose the third barrier.VCSNS Basis Reference(s):
- 1. EOP-12.0 Monitoring of Critical Safety Functions 2. OAP-1 03.2 Emergency Operating Procedure Setpoint Document 3. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.A Page 325 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Containment Category:
- 4. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
B. Containment hydrogen concentration
> 4%Definition(s):
None Basis: Plant-Specific The lower limit of flammability of hydrogen in air is approximately 4% (ref. 1).In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive mixture of dissolved gases in Containment.
However, Containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that an explosive mixture exists. A combustible mixture can be formed when hydrogen gas concentration in the Containment atmosphere is greater than 4% by volume (ref. 1, 2). All hydrogen measurements are referenced to concentrations in dry air even though the actual Containment environment may contain significant steam concentrations.
The plant has two hydrogen monitoring systems. Sample points are located near each recombiner and near the RBCUs on the 530' Level. Manual action is required to start the redundant hydrogen analyzers.
The analyzers
[CI-8257 (8258)] have a range of 0-10% and 0-20% of H 2 in air (by volume) and an accuracy of +/- 2% of range. Hydrogen concentration in the Reactor Building is indicated in the control room (ref. 3, 4).To generate such levels of combustible gas, loss of the Fuel Clad and RCS barriers must have occurred.
With the Potential Loss of the Containment barrier, the threshold hydrogen concentration, therefore, will likely warrant declaration of a General Emergency.
Generic Page 326 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity.
It therefore represents a potential loss of the Containment Barrier.VCSNS Basis Reference(s):
- 1. FSAR Section 6.2.3.5.1 2. SOP-1 22 Post Accident Hydrogen Removal System 3. FSAR Section 6.2.5.5.3 4. SOP-1 22 Post Accident Hydrogen Removal System 5. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.B Page 327 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT El Barrier: Containment Category:
- 4. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
C. Containment pressure > 12 psig AND< one full train of depressurization equipment (Table F-2) operating per design for > 15 min. (Note 1 Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Table F-2 Full Train Depressurization Equipment RBCU Groups Containment Sprays Operating Operating 2 0 1 1 0 2 Definition(s):
None Basis: Plant-Specific The Containment pressure setpoint (12 psig, ref. 2, 3) is the pressure at which the Containment Spray System should actuate and begin performing its function.
The design basis accident analyses and evaluations assume the loss of one Containment Spray System train (ref. 2, 3).Each spray subsystem is started by separate ESF containment isolation Phase A and spray actuation signals. Normally, both subsystems operate; however, they are independent and can operate individually.
Although the design basis for Reactor Building Page 328 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]heat removal is one spray subsystem operating in conjunction with one Reactor Building Cooling Unit (with one RHR pump and one charging pump providing emergency core cooling water), two Reactor Building Cooling Units (RBCUs) with no spray pumps or two spray pumps with no RBCUs can handle all required heat loads Technical Specifications defined equipment that comprises one full train of depressurization equipment is given in Table F-3 (ref. 1, 2, 4, 5).RBCU operation verification is performed in accordance with EOP-1.0 Attachment 3 SI Equipment Verification.
In order to take credit for a RBCU operating per design, each RBCU must meet minimum Service Water flow requirements (ref. 6).Generic This threshold describes a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible.
This threshold represents a potential loss of containment in that containment heat removal/depressurization systems (e.g., containment sprays, etc., but not including containment venting strategies) are either lost or performing in a degraded manner.VCSNS Basis Reference(s):
- 1. FSAR Section 6.2.2.2.1.2
- 2. OAP-103.2 Emergency Operating Procedure Setpoint Document 3. EOP-12.0 Monitoring of Critical Safety Functions 4. Technical Specifications 3/4-6.2.1 5. Technical Specifications 3/4-6.2.3 6. EOP-1.0 Reactor Trip/Safety Injection Actuation Attachmen 3 SI Equipment Verification
- 7. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.C Page 329 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Containment Category:
- 5. ED Judgment Degradation Threat: Loss Threshold:
A. Any condition in the opinion of the ED that indicates loss of the containment barrier Definition(s):
None Basis: Plant-Specific None Generic This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is lost.VCSNS Basis Reference(s):
- 1. NEI 99-01 Emergency Director Judgment Containment Loss 6.A Page 330 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Barrier: Containment Category:
- 5. ED Judgment Degradation Threat: Potential Loss Threshold:
A. Any condition in the opinion of the ED that indicates potential loss of the containment barrier Definition(s):
None Basis: Plant-Specific None Generic This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
VCSNS Basis Reference(s):
- 1. NEI 99-01 Emergency Director Judgment Containment Potential Loss 6.A Page 331 of 346 EPP-1 08 ENCLOSURE I REVISION 01 [DRAFT E]ATTACHMENT 3 Figures Page 332 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Figure 1: Fuel Assembly Uncovery Elevations Transfer slot,-i" 461'-6w Normal WL pity SFP 439'top offuelin SFP UL L- and upenders-431' Transfer tube--, -----427' TOAF Page 333 of 346 (0 CA)CA)CA)In CD 0 C,, m DO m Cf, 0 z m z 0 m cn -m I m 0O EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Figure 3: RCS Elevations MANSELL/MIDLOOP MONITOR RVLIS NARROW ELEVATION RANGE (FT) (%)Top of Active Fuel 427' 57.9 Bottom of Hot Leg 429' 6" 64.2 RV FLANGE 437' 7.4" 84.3 Page 335 of 346 EPP-1 08 ENCLOSURE I REVISION 01 [DRAFT E]Figure 4: Response of the TMI-2 Source Range Measurement During the First Six Hours of the Accident Cq y- )C 4): cc, EE<0 0 z C (0 'C0 (sapeoap 6ol) puooaS ied slunoo Page 336 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Figure 5: CSFST Core Cooling GOTO EOP-14.0 GOTO EOP-14.0 NOftfWW N CORE WTn TO&(co=i EGO T U.MB THAN 120O YESEOP-14.1 GOTO-EOP-14.1 000E EOP-14 SGO TO r EOP-14.1 0 No GOTO ooo FAEOP-14-2 CSF SAT Page 337 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Figure 6: CSFST Heat Sink GO TO Mo EOP-1 5.0 tow NO 000000000000*
080000000000 GO- 5.1 0 00OO° 00000000000 EOP-15.2 o o 0 0 NO YES 80000000GO TO So o EOP-15.4 o 0 0 o UNIM N "r 0 11 ml r ... EOP-15.4 CSF SAT Page 338 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Figure 7: CSFST RCS Integrity GO TO EOP-16.0 STO THE.r MOWTO OF umWr A 7Y11 GO TO 0 EOP-16.0 LE ___.3c8 Go___ 8 N&WO=M N GIMIIM TH" ZiYE NO 00 GOTO F" , o .....W 0 EOP-16-1 NO CSF SAT CSF SAT N 01.m D 1.1ou LM Tm l" IN' GO TO I M L 6T0 I=' = .J M Y E S E O P -1 6 .0 m$PmoP, .
NO= E P1.GO TOC SF SAT Page 339 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Figure 8: Plant Operational Limits Curve PLANT OPERATIONAL LIMITS CURVE RCS PRESSURE (PSIG)3000 2500 2000 15S00 1000 500 0 100 150 200 250 300 3SO RCS Tcold (-F)400 Page 340 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Figure 9: CSFST Containment Page 341 of 346 EPP-1 08 ENCLOSURE I REVISION 01 [DRAFT El Figure 10: Containment Integrity or Bypass Examples------------
Effluent Auxiliary Building Monitor Seal Coolinq Page 342 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]ATTACHMENT 4 Safe Operation
& Shutdown Areas Tables R-2 & H-3 Bases Page 343 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Background NEI 99-01 Revision 6 ICs AA3 and HA5 prescribe declaration of an Alert based on impeded access to rooms or areas (due to either area radiation levels or hazardous gas concentrations) where equipment necessary for normal plant operations, cooldown or shutdown is located. These areas are intended to be plant operating mode dependent.
Specifically the Developers Notes For AA3 and HA5 states: The "site-specific list of plant rooms or areas with entry-related mode applicability identified" should specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown.
Do not include rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations).
In addition, the list should specify the plant mode(s) during which entry would be required for each room or area.The list should not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections).
Further, as specified in IC HA5: The list need not include the Control Room if adequate engineered safety/design features are in place to preclude a Control Room evacuation due to the release of a hazardous gas. Such features may include, but are not limited to, capability to draw air from multiple air intakes at different and separate locations, inner and outer atmospheric boundaries, or the capability to acquire and maintain positive pressure within the Control Room envelope.VCS1 Table R-2 and H-3 Bases A review of station operating procedures identified the following mode dependent in-plant actions and associated areas that are required for normal plant operation, cooldown or shutdown: MODE 1 (Power Operation)" FWP and FWBP's per SOP-210 (TB all levels)* PTP-1 02.001 Extraction System check valves (TB-436)* XVG02074 &2075 (TB-412)* XVG02210 (TB-463)* XVT02072A/B (TB-412)* 3062's Blow down temperature control (AB-436)MODE 2 (Startup)" FWP and FWBP's per SOP-210 (TB all levels)" XVTO1663 (TB-412)" Primary Chemistry Lab (CB-412)Page 344 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]MODE 3 (Hot Standby)* Primary Chemistry Lab (CB-412)* PZR heater disconnects (AB-436)* RCP seal injection adjustments (AB-412 west pen/IB-412 east pen)* SI Accumulator isolation valves (IB-463/AB-463)
- P-1 2 interlocks (CB-436 relay room)* RHR samples (AB-374)* CCW pump start (IB-412)" CHG/Sl pump cycle (AB-388/IB-436/IB-463)
MODE 4 (Hot Shutdown)/Mode 5 (Cold Shutdown* CHG/Sl Bkr alignment (IB-436/1B463)" RHR bkr alignment (AB-412/AB-463/IB-463)" ASI disable (AB-388/AB-400/AB-436)" P-12 interlocks (CB-436 relay room)" Steam Generator Shell Temp monitoring for MODE 5 (RB-412/RB-436)
Table R-2 & H-3 Results Table R-2 & H-3Safe Operation
& Shutdown Areas Area Mode Applicability Auxiliary Building 374' 3 Auxiliary Building 388' 3, 4, 5 Auxiliary Building 400' 4, 5 Auxiliary Building 412 3, 4, 5 Auxiliary Building 436' 1,2, 3, 4, 5 Auxiliary Building 463' 3, 4, 5 Intermediate Building 412' 3 Intermediate Building 436' 4, 5 Intermediate Building 463' 3, 4, 5 Control Building 412' 2, 3 Control Building 436' 3, 4, 5 Turbine Building (All levels) 1,2 Plant ODeratina Procedures Reviewed 1. GOP-4B Power Operation Mode 1 Descending
- 2. GOP-5 Reactor Shutdown From Startup to Hot Standby Mode 2 to Mode 3 Page 345 of 346 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]3. GOP-6 Plant Shutdown From Hot Standby to Cold Shutdown Mode 3 to Mode 5 4. SOP-21 0 Feedwater System 5. PTP-102.001 Main Turbine Tests 6. "Rooms Needed for Normal Plant Shutdown from Mode 1 to Mode 5" An Assessment performed by Doug Edwards 5/24/13 Page 346 of 346 Document Control Desk Attachment IV LAR-14-02392 RC-14-0032 Page 1 of 117 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) Unit I DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 ATTACHMENT IV V.C. Summer Unit I NEI 99-01 Revision 6 EAL Comparison Matrix A SCANA COAPANW V. C. Summer Unit 1 NEI 99-01 Revision 6 EAL Comparison Matrix Revision 0 [Draft E 3/20/14]
EAL Comparison Matrix OSSI Project #12-0202 VCS1 Table of Contents Section Page Introduction
1 Comparison Matrix Format ------------------------------------------------------------------------
1 EAL Wording ---------------------------------------------------------------------------------
1 EAL Emphasis Techniques
1 Global Differences
2 Differences and Deviations
3 Category A -Abnormal Rad Levels / Rad Effluents..............................................------------------------------------------------------
12 Category C -Cold Shutdown / Refueling System Malfunction
31 Category D -Permanently Defueled Station Malfunction
51 Category I -Events Related to Independent Spent Fuel Storage Installations
53 Category F -Fission Product Barrier Degradation
55 Category H -Hazards and Other Conditions Affecting Plant Safety ------------------------------------------
68 Category S -System Malfunction
86 Table 1 -VCS1 EAL Categories/Subcategories.................................................---------------------------------------------------------
5 Table 2 -NEI / VCS1 EAL Identification Cross-Reference
6 Table 3 -Summary of Deviations
11 iofi EAL Comparison Matrix OSSI Project #12-0202 VCS1 Introduction This document provides a line-by-line comparison of the Initiating Conditions (ICs), Mode Applicability and Emergency Action Levels (EALs) in NEI 99-01 Rev. 6 Final, Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML110240324, and the V. C. Summer Unit 1 (VCS1) ICs, Mode Applicability and EALs. This document provides a means of assessing VCS1 P differences and deviations from the NRC endorsed guidance given in NEI 99-01. Discussion of VCS1 EAL bases and lists of source document references are given in the EAL Technical Bases Document.
It is, therefore, advisable to reference the EAL Technical Bases Document for background information while using this document.Comparison Matrix Format The ICs and EALs discussed in this document are grouped according to NEI 99-01 Recognition Categories.
Within each Recognition Category, the ICs and EALs are listed in tabular format according to the order in which they are given in NEI 99-01. Generally, each row of the comparison matrix provides the following information:
- NEI EAL/IC identifier
- NEI EAL/IC wording* VCS1 EAL/IC identifier
- VCS1 EAL/IC wording* Description of any differences or deviations EAL Wording In Section 4.1, NEI recommends the following: "The guidance in NEI 99-01 is not intended to be applied to plants "as-is"; however, developers should attempt to keep their site-specific schemes as close to the generic guidance as possible.
The goal is to meet the intent of the generic Initiating Conditions (ICs) and Emergency Action Levels (EALs) within the context of site-specific characteristics
-locale, plant design, operating features, terminology, etc.Meeting this goal will result in a shorter and less cumbersome NRC review and approval process, closer alignment with the schemes of other nuclear power plant sites and better positioning to adopt future industry-wide scheme enhancements" To assist the Emergency Director (ED), the VCS1 EALs have been written in a clear and concise style (to the extent that the differences from the NEI EAL wording could be reasonably documented and justified).
As a result, any unnecessary words have been removed from the VCS1 EALs to reduce EAL-user reading burden to the extent practicable.
The wording reduction gained from elimination of a few characters in a given EAL may not appear to be advantageous within the context of one EAL.When applied to the composite set of EALs, however, significant gains are realized and reading efficiency is improved.
This supports timely and accurate classification in the tense atmosphere of an emergency event. The EAL differences introduced to reduce reading burden comprise almost all of the differences justified in this document.EAL Emphasis Techniques Due to the width of the table columns and table formatting constraints in this document, line breaks and indentation may differ slightly from the appearance of comparable wording in the source documents.
NEI 99-01 is the source document for the NEI EALs; the VCS1 EAL Technical Bases Document for the VCS1 EALs.Development of the VCS1 IC/EAL wording has attempted to minimize inconsistencies and apply sound human factors principles.
As a result, differences occur between NEI and VCS1 ICs/EALs for these reasons alone.When such difference may infer a technical difference in the associated NEI IC/EAL, the difference is identified and a justification provided.The print and paragraph formatting conventions summarized below guide presentation of the VCS1 EALs in accordance with the EAL writing criteria.Space restrictions in the EAL table of this document sometimes override this criteria in cases when following the criteria would introduce undesirable complications in the EAL layout.* Upper case-underline print is used for the logic terms AND, OR and EITHER.* Bold font is used for certain logic terms, negative terms (not, cannot, etc.), any, all." Upper case print is reserved for defined terms, acronyms, system abbreviations, logic terms (and, or, etc. when not used as a conjunction), annunciator window engravings.
1 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1* Three or more items in a list are normally introduced with "Any of the following..." or "All of the following..." Items of the list begin with bullets when a priority or sequence is not inferred.* The use of AND/OR logic within the same EAL has been avoided when possible.
When such logic cannot be avoided, indentation and separation of subordinate contingent phrases is employed.Global Differences The differences listed below generally apply throughout the set of EALs and are not repeated in the Justification sections of this document.
The global differences do not decrease the effectiveness of the intent of NEI 99-01.1. The NEI phrase "Notification of Unusual Event" has been changed to"Unusual Event" or abbreviated "UE" to reduce EAL-user reading burden.2. NEI 99-01 IC Example EALs are implemented in separate plant EALs to improve clarity and readability.
For example, NEI lists all IC HU3 Example EALs under one IC. The corresponding VCS1 EALs appear as unique EALs (e.g., HU3.1 through HU3.4).3. Mode applicability identifiers (numbers/letter) modify the NEI 99-01 mode applicability names as follows: 1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown, 5 -Cold Shutdown, 6 -Refueling, D -Defueled.
NEI 99-01defines Defueled as follows: "Reactor Vessel contains no irradiated fuel (full core off-load during refueling or extended outage)." 4. NEI 99-01 uses the terms greater than, less than, greater than or equal to, etc. in the wording of some example EALs. For consistency and reduce EAL-user reading burden, VCS1 has adopted use of boolean symbols in place of the NEI 99-01 text modifiers within the EAL wording.5. "min." is the standard abbreviation for "minutes" and is used to reduce EAL user reading burden.6. IC/EAL identification:
- NEI Recognition Category A "Abnormal Radiation Levels/Radiological Effluents" has been changed to Category R"Abnormal Rad Levels / Rad Effluents." The designator "R" is more intuitively associated with radiation (rad) or radiological events. NEI IC designators beginning with "A" have likewise been changed to "R." NEI 99-01 defines the thresholds requiring emergency classification (example EALs) and assigns them to ICs which, in turn, are grouped in "Recognition Categories." The Recognition Categories, however, are so broad and the IC descriptions are so varied that an EAL is difficult to locate in a timely manner when the EAL-user must refer to a set of EALs with the NEI organization and identification scheme. The NEI document clearly states that the EAL/IC/Recognition Category scheme is not intended to be the plant-specific EAL scheme for any plant, and appropriate human factors principles should be applied to development of an EAL scheme that helps the EAL-user make timely and accurate classifications.
VCS1 endeavors to optimize the NEI EAL organization and identification scheme to enhance usability of the plant-specific EAL set. To this end, the VCS1 IC/EAL scheme includes the following features: a. Division of the NEI EAL set into three groups: o EALs applicable under all plant operating modes -This group would be reviewed by the EAL-user any time emergency classification is considered.
o EALs applicable only under hot operating modes -This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, Hot Standby, Startup or Power Operation mode.o EALs applicable only under cold operating modes -This group would only be reviewed by the EAL-user when the plant is in Cold Shutdown, Refueling or Defueled mode.The purpose of the groups is to avoid review of hot condition EALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a hot condition.
This approach significantly minimizes the total number of EALs that must be reviewed by the EAL-user for a given plant condition, reduces EAL-user 2 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 reading burden and, thereby, speeds identification of the EAL that applies to the emergency.
- b. Within each of the above three groups, assignment of EALs to categories/subcategories
-Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user.Subcategories are used as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds.
The VCS1 EAL categories/subcategories and their relationship to NEI Recognition Categories are listed in Table 1.c. Unique identification of each EAL -Four characters comprise the EAL identifier as illustrated in Figure 1.Figure 1 -EAL Identifier EAL Identifier XXX.X Category (R, H, E, S, F, C) ' L Sequential number within subcategory/classitication Emergency classification (E, S, A, U) SuI L L Sbctegory number (Ilit no subcategory)
The first character is a letter associated with the category in which the EAL is located. The second character is a letter associated with the emergency classification level (G for General Emergency, S for Site Area Emergency, A for Alert, and U for Notification of Unusual Event). The third character is a number associated with one or more subcategories within a given category.
Subcategories are sequentially numbered beginning with the number "1". If a category does not have a subcategory, this character is assigned the number "1". The fourth character is a number preceded by a period for each EAL within a subcategory.
EALs are sequentially numbered within the emergency classification level of a subcategory beginning with the number "1".The EAL identifier is designed to fulfill the following objectives:
o Uniqueness
-The EAL identifier ensures that there can be no confusion over which EAL is driving the need for emergency classification.
o Speed in locating the EAL of concern -When the EALs are displayed in a matrix format, knowledge of the EAL identifier alone can lead the EAL-user to the location of the EAL within the classification matrix. The identifier conveys the category, subcategory and classification level. This assists ERO responders (who may not be in the same facility as the ED) to find the EAL of concern in a timely manner without the need for a word description of the classification threshold.
o Possible classification upgrade -The category/subcategory/identifier scheme helps the EAL-user find higher emergency classification EALs that may become active if plant conditions worsen.Table 2 lists the VCS1 ICs and EALs that correspond to the NEI ICs/Example EALs when the above EALJIC organization and identification scheme is implemented.
Differences and Deviations In accordance NRC Regulatory Issue Summary (RIS) 2003-18 "Use of Nuclear Energy Institute (NEI) 99-01, Methodology for Development of Emergency Action Levels" Supplements 1 and 2, a difference is an EAL change in which the basis scheme guidance differs in wording but agrees in meaning and intent, such that classification of an event would be the same, whether using the basis scheme guidance or the VCS1 EAL. A deviation is an EAL change in which the basis scheme guidance differs in wording and is altered in meaning or intent, such that classification of the event could be different between the basis scheme guidance and the VCS1 proposed EAL.Administrative changes that do not actually change the textual content are neither differences nor deviations.
Likewise, any format change that does not alter the wording of the IC or EAL is considered neither a difference nor a deviation.
The following are examples of differences:
3 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1" Choosing the applicable EAL based upon plant type (i.e., BWR vs.PWR)." Using a numbering scheme other than that provided in NEI 99-01 that does not change the intent of the overall scheme." Where the NEI 99-01 guidance specifically provides an option to not include an EAL if equipment for the EAL does not exist at VCS1 (e.g., automatic real-time dose assessment capability).
- Pulling information from the bases section up to the actual EAL that does not change the intent of the EAL." Choosing to state ALL Operating Modes are applicable instead of stating N/A, or listing each mode individually under the Abnormal Rad Level/Radiological Effluent and Hazard and Other Conditions Affecting Plant Safety sections.* Using synonymous wording (e.g., greater than or equal to vs. at or above, less than or equal vs. at or below, greater than or less than vs. above or below, etc.)* Adding VCS1 equipment/instrument identification and/or noun names to EALs.* Combining like ICs that are exactly the same but have different operating modes as long as the intent of each IC is maintained and the overall progression of the EAL scheme is not affected.* Any change to the IC and/or EAL, and/or basis wording, as stated in NEI 99-01, that does not alter the intent of the IC and/or EAL, i.e., the IC and/or EAL continues to: o Classify at the correct classification level.o Logically integrate with other EALs in the EAL scheme.o Ensure that the resulting EAL scheme is complete (i.e., classifies all potential emergency conditions).
The following are examples of deviations:
- Use of altered mode applicability.
- Altering key words or time limits.* Changing words of physical reference (protected area, safety-related equipment, etc.).* Eliminating an IC. This includes the removal of an IC from the Fission Product Barrier Degradation category as this impacts the logic of Fission Product Barrier ICs." Changing a Fission Product Barrier from a Loss to a Potential Loss or vice-versa.
- Not using NEI 99-01 definitions as the intent is for all NEI 99-01 users to have a standard set of defined terms as defined in NEI 99-01.Differences due to plant types are permissible (BWR or PWR).Verbatim compliance to the wording in NEI 99-01 is not necessary as long as the intent of the defined word is maintained.
Use of the wording provided in NEI 99-01 is encouraged since the intent is for all users to have a standard set of defined terms as defined in NEI 99-01." Any change to the IC and/or EAL, and/or basis wording as stated in NEI 99-01 that does alter the intent of the IC and/or EAL, i.e., the IC and/or EAL: o Does not classify at the classification level consistent with NEI 99-01.o Is not logically integrated with other EALs in the EAL scheme.o Results in an incomplete EAL scheme (i.e., does not classify all potential emergency conditions).
The "Difference/Deviation Justification" columns in the remaining sections of this document identify each difference between the NEI 99-01 IC/EAL wording and the VCS1 IC/EAL wording. An explanation that justifies the reason for each difference is then provided.
If the difference is determined to be a deviation, a statement is made to that affect and explanation is given that states why classification may be different from the NEI 99-01 IC/EAL and the reason for its acceptability.
In all cases, however, the differences and deviations do not decrease the effectiveness of the intent of NEI 99-01. A summary list of VCS1 EAL deviations from NEI 99-01 is given in Table 3.4 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 Table 1 -VCS1 EAL Categories/Subcategories VCS1 EALs NEI Category Subcategory Recognition Category Group: Any Operating Mode: 1 -Radiological Effluent Abnormal Rad Levels/Radiological Effluent R -Abnormal Rad Levelse/Rad Effluent 2 -Irradiated Fuel Event ICs/EALs 3 -Area Radiation Levels H -Hazards and Other Conditions Affecting 1 -Security Hazards and Other Conditions Affecting Plant Safety 2 -Seismic Event Plant Safety ICs/EALs 3 -Natural or Technological Hazard 4 -Fire 5 -Hazardous Gas 6 -Control Room Evacuation 7 -ED Judgment I- ISFSI None ISFSI ICs/EALs Group: Hot Conditions:
1 -Loss of ESF AC Power System Malfunction ICs/EALs 2 -Loss of Vital DC Power 3 -Loss of Control Room Indications S -System Malfunction 4- RCS Activity 5 -RCS Leakage 6 -RTS Failure 7 -Loss of Communications 8 -Containment Isolation Failure F -Fission Product Barrier None Fission Product Barrier ICs/EALs Group: Cold Conditions:
1 -RCS Level Cold Shutdown./
Refueling System 2 -Loss of ESF AC Power Malfunction ICs/EALs C -Cold Shutdown/Refueling System 3 -RCS Temperature I Malfunction 4 -Loss of Vital DC Power 5 -Loss of Communications 5 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 Table 2 -NEI / VCS1 EAL Identification Cross-Reference NEI VCS1 Example Category and Subcategory EAL EAL AU1 1 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RU1.1 AU1 2 R -Abnormal Rad Levels/ Rad Effluent, 1 -Radiological Effluent RU1.1 AU1 3 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RU1.2 AU2 1 R -Abnormal Rad Levels / Rad Effluent, 2 -Irradiated Fuel Event RU2.1 AA1 1 R -Abnormal Rad Levels/ Rad Effluent, 1 -Radiological Effluent RA1.1 AA1 2 R -Abnormal Rad Levels/ Rad Effluent, 1 -Radiological Effluent RA1.2 AA1 3 N/A N/A AA1 4 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RA1.3 AA2 1 R -Abnormal Rad Levels / Rad Effluent, 2 -Irradiated Fuel Event RA2.1 AA2 2 R -Abnormal Rad Levels / Rad Effluent, 2 -Irradiated Fuel Event RA2.2 AA2 3 R -Abnormal Rad Levels / Rad Effluent, 2 -Irradiated Fuel Event RA2.3 AA3 1 R -Abnormal Rad Levels / Rad Effluent, 3 -Area Radiation Levels RA3.1 AA3 2 R -Abnormal Rad Levels / Rad Effluent, 3 -Area Radiation Levels RA3.2 AS1 1 R -Abnormal Rad Levels/ Rad Effluent, 1 -Radiological Effluent RS1.1 AS1 2 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RS1.2 AS1 3 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RS1.3 6 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI VCS1 IC Example Category and Subcategory EAL EAL AS2 1 R -Abnormal Rad Levels / Rad Effluent, 2 -Irradiated Fuel Event RS2.1 AG1 1 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RG1.1 AG1 2 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RG1.2 AG1 3 R -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent RG1.3 AG2 1 R -Abnormal Rad Levels / Rad Effluent, 2 -Irradiated Fuel Event RG2.1 Cui 1 C -Cold SD/ Refueling System Malfunction, 1 -RCS Level CU1.1 CU1 2 C -Cold SD/ Refueling System Malfunction, 1 -RCS Level CU1.2 CU2 1 C -Cold SD/ Refueling System Malfunction, 2 -Loss of ESF AC Power CU2.1 CU3 1 C -Cold SD/ Refueling System Malfunction, 3 -RCS Temperature CU3.1 CU3 2 C -Cold SD/ Refueling System Malfunction, 3 -RCS Temperature CU3.2 CU4 1 C -Cold SD/ Refueling System Malfunction, 4 -Loss of Vital DC Power CU4.1 CU5 1,2, 3 C -Cold SD/ Refueling System Malfunction, 5 -Loss of Communications CU5.1 CA1 1 C -Cold SD/ Refueling System Malfunction, 1 -RCS Level CA1.1 CA1 2 C -Cold SD/ Refueling System Malfunction, 1 -RCS Level CA1.2 CA2 1 C -Cold SD/ Refueling System Malfunction, 1 -Loss of ESF AC Power CA2.1 CA3 1 C -Cold SD/ Refueling System Malfunction, 3 -RCS Temperature CA3.1 CA3 2 C -Cold SD/ Refueling System Malfunction, 3 -RCS Temperature CA3.2 CA6 1 H -Hazards and Other Conditions Affecting Plant Safety 7 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI VCS1 IC Example Category and Subcategory EAL EAL 2 -Seismic Event HA2.1 3 -Natural or Technological Hazard HA3.1 4 -Fire or Explosion HA4.1 0S1 1 C -Cold SD/ Refueling System Malfunction, 1 -RCS Level CS1.1 CS1 2 C -Cold SD/ Refueling System Malfunction, 1 -RCS Level CS1.2 CS1 3 C -Cold SD/ Refueling System Malfunction, 1 -RCS Level CS1.3 CG1 1 C -Cold SD/ Refueling System Malfunction, 1 -RCS Level CS1.1 CG1 2 C -Cold SD/ Refueling System Malfunction, 1 -RCS Level CG1.2 E-HU1 1 I -ISFSI IU1.1 FA1 1 F -Fission Product Barrier Degradation FA1.1 FS1 1 F -Fission Product Barrier Degradation FS1.1 FG1 1 F -Fission Product Barrier Degradation FG1.1 HU1 1, 2, 3 H -Hazards and Other Conditions Affecting Plant Safety, 1 -Security HUl.1 HU2 1 H -Hazards and Other Conditions Affecting Plant Safety, 2 -Seismic Event HU2.1 HU3 1 H -Hazards and Other Conditions Affecting Plant Safety, 3 -Natural or Technological Hazard HU3.1 HU3 2 H -Hazards and Other Conditions Affecting Plant Safety, 3 -Natural or Technological Hazard HU3.2 HU3 3 H -Hazards and Other Conditions Affecting Plant Safety, 3 -Natural or Technological Hazard HU3.3 HU3 4 H -Hazards and Other Conditions Affecting Plant Safety, 3 -Natural or Technological Hazard HU3.4 HU3 5 N/A N/A 8 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI VCS1 Example Category and Subcategory EAL HU4 1 H -Hazards and Other Conditions Affecting Plant Safety, 4 -Fire or Explosion HU4.1 HU4 2 H -Hazards and Other Conditions Affecting Plant Safety, 4 -Fire or Explosion HU4.2 HU4 3 H -Hazards and Other Conditions Affecting Plant Safety, 4 -Fire or Explosion HU4.3 HU4 4 H -Hazards and Other Conditions Affecting Plant Safety, 4 -Fire or Explosion HU4.4 HU7 1 H -Hazards and Other Conditions Affecting Plant Safety, 7 -Judgment HU7.1 HAl 1, 2 H -Hazards and Other Conditions Affecting Plant Safety, 1 -Security HAl .1 HA5 1 H -Hazards and Other Conditions Affecting Plant Safety, 5 -Hazardous Gases HA5.1 HA6 1 H -Hazards and Other Conditions Affecting Plant Safety, 6 -Control Room Evacuation HA6.1 HA7 1 H -Hazards and Other Conditions Affecting Plant Safety, 7 -Judgment HA7.1 HS1 1 H -Hazards and Other Conditions Affecting Plant Safety, 1 -Security HS1.1 HS6 1 H -Hazards and Other Conditions Affecting Plant Safety, 6 -Control Room Evacuation HS6.1 HS7 1 H -Hazards and Other Conditions Affecting Plant Safety, 7 -Judgment HS7.1 HG1 1 H -Hazards and Other Conditions Affecting Plant Safety, 1 -Security HG1.1 HG7 2 H -Hazards and Other Conditions Affecting Plant Safety, 7 -Judgment HG7.1 SUl 1 S -System.Malfunction, 1 -Loss of ESF AC Power SU1.1 SU2 1 S -System Malfunction, 3 -Loss of Control Room Indications SU3.1 SU3 1 S -System Malfunction, 4 -RCS Activity SU4.1 SU3 2 S -System Malfunction, 4 -RCS Activity SU4.2 9 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI VCS1 IC Example Category and Subcategory EAL EAL SU4 1,2, 3 S -System Malfunction, 5 -RCS Leakage SU5.1 SU5 1 S- System Malfunction, 6 -RTS Failure SU6.1 SU5 2 S -System Malfunction, 6 -RTS Failure SU6.2 SU6 1,2, 3 S -System Malfunction, 7 -Loss of Communications SU7.1 SU7 1 S -System Malfunction, 8-Containment Isolation Failure SU8.1 SU7 2 S -System Malfunction, 8 -Containment Isolation Failure SU8.2 SA1 1 S -System Malfunction, 1 -Loss of ESF AC Power SA1.1 SA2 1 S -System Malfunction, 3 -Loss of Control Room Indications SA3.1 SA5 1 S -System Malfunction, 6 -RTS Failure SA6.1 SA9 1 H -Hazards and Other Conditions Affecting Plant Safety 2 -Seismic Event HA2.1 3 -Natural or Technological Hazard HA3.1 4 -Fire or Explosion HA4.1 SS1 1 S -System Malfunction, 1 -Loss of ESF AC Power SS1.1 SS5 1 S -System Malfunction, 6 -RTS Failure SS6.1 SS8 1 S -System Malfunction, 2 -Loss of Vital DC Power SS2.1 SG1 1 S -System Malfunction, 1 -Loss of ESF AC Power SG1.1 SG8 2 S -System Malfunction, 1 -Loss of ESF AC Power SG1.2 10 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 Table 3 -Summary of Deviations NEI vCS1 IC Example EAL EAL Description AA1 3 N/A This example EAL is not implemented in the VCS1 EAL scheme. VCS1 does not have the capability to perform real-time offsite dose assessment for liquid releases.11 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 Category R Abnormal Rad Levels/ Radiological Effluent 12 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI I NEI IC Wording and Mode VCS1 VCS1 IC Wording and Mode Difference/Deviation Justification Applicability IC#(s) Applicability AU1 Release of gaseous or liquid RU1 Release of gaseous or liquid The VCS1 0DCM is the site-specific effluent release radioactivity greater than 2 times radioactivity
> 2 times the ODCM limits controlling document.the (site-specific effluent release for 60 minutes or longer.controlling document) limits for MODE: All 60 minutes or longer.MODE: All NEI Ex. NEI Example EAL Wording EAL VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #Reading on ANY effluent Reading on any Table R-1 effluent Example EALs #1 and #2 have been combined into a single radiation monitor greater than 2 radiation monitor > column "UE" for a EAL to simplify presentation.
times the (site-specific effluent 60 min. The NEI phrase "...effluent radiation monitor greater than 2 release controlling document) (Notes 1, 2, 3) times the (site-specific effluent release controlling limits for 60 minutes or longer: document)" and "effluent radiation monitor greater than 2 (site-specific monitor list and times the alarm setpoint established by a current radioactivity threshold values corresponding discharge permit " have been replaced with "...any Table R-1 to 2 times the controlling RU1.1 effluent radiation monitor> column "UE".document limits) UE thresholds for all VCS1 continuously monitored gaseous 2 Reading on ANY effluent and liquid release pathways are listed in Table R-1 to radiation monitor greater than 2 consolidate the information in a single location and, thereby, times the alarm setpoint simplify identification of the thresholds by the EAL user. The established by a current values shown in Table R-1 column "UE", consistent with the radioactivity discharge permit for NEI bases, represent two times the ODCM release limits for 60 minutes or longer, both liquid and gaseous release.3 Sample analysis for a gaseous or RU1.2 Sample analysis for a gaseous or The VCS1 ODCM is the site-specific effluent release liquid release indicates a liquid release indicates a concentration controlling document.concentration or release rate or release rate > 2 x ODCM limits for greater than 2 times the (site- a 60 min.specific effluent release (Notes 1, 2)controlling document) limits for 13 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI Ex. NEI Example EAL Wording EAL VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #60 minutes or longer.Notes
- The Emergency Director N/A Note 1: The Emergency Director The classification timeliness note has been standardized should declare the Unusual should declare the event across the VCS1 EAL scheme by referencing the "time limit" Event promptly upon promptly upon determining specified within the EAL wording.determining that 60 minutes that time limit has been has been exceeded, or will exceeded, or will likely be likely be exceeded.
exceeded." If an ongoing release is Note 2: If an ongoing release is The classification timeliness note has been standardized detected and the release detected and the release across the VCSl EAL scheme by referencing the "time limit" start time is unknown, start time is unknown, specified within the EAL wording.assume that the release assume that the release duration has exceeded 60 duration has exceeded the minutes. specified time limit." If the effluent flow past an Note 3: If the effluent flow past an effluent monitor is known to effluent monitor is known to None have stopped due to actions have stopped, indicating that to isolate the release path, the release path is isolated, then the effluent monitor the effluent monitor reading reading is no longer valid for is no longer VALID for classification purposes.
classification purposes.14 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 Table R-1 Effluent Monitor Classification Thresholds Release Point J Monitor GE SAE Alert UE 2 X Hi-Rad Main Plant Vent RM-A3 (gas) N/A 280,000 cpm 28,000 cpm alarm Exhaust RM-A13 14 mR/hr N/A N/A N/A o 2 X Hi-Rad RM-A4 (gas) N/A N/A N/APurge exhaust alarm RM-A14 740 mR/hr 74 mR/hr 7.4 mR/hr N/A Main Steam Line RM-G19 535 mR/hr 53.5 mR/hr 5.4 mR/hr N/A (Note 4) A/B/C Liquid Waste and"- Nuclear Blowdown RM-L-9 N/A N/A N/A alarm" Discharge 15 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI I NEI IC Wording and Mode VCS1 VCS1 IC Wording and Mode Difference/Deviation Justification Applicability IC#(s) Applicability AU2 UNPLANNED loss of water level RU2 Unplanned loss of water level above None above irradiated fuel. irradiated fuel MODE: All MODE: All NEI Ex. NEI Example EAL Wording VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #a. UNPLANNED water level RU2.1 UNPLANNED water level drop in the Site-specific level indications and area radiation monitors drop in the REFUELING REFUELING PATHWAY as indicated bulletized.
PATHWAY as indicated by by any of the following:
ANY of the following:
- Refueling Cavity: LI-7403 (site-specific level MCB annunciator XCP-609 2-indications).
6 (REFUEL CAV LVL HI/LO)AND AND* Spent Fuel Pool: LI-7431 and b. UNPLANNED rise in area Se lP:3a radiation levels as indicated LI-7433 by ANY of the following MCB annunciators XCP radiation monitors.
608(609) 1-2 (SFP LVL HI/LO)(site-specific list of area
- Fuel Transfer Canal: LI-7405 radiation monitors)
MCB annunciator XCP-612 1-6 (FUEL XFER CANAL LVL HI/LO)AND UNPLANNED rise in area radiation levels as indicated by any of the following radiation monitors: " RM-G6 Rx Bldg Refueling Bridge* RM-G17A/B Rx Bldg Manipulator Crane* RM-G8 FHB Refueling Bridge Area Gamma 16 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCSi 17 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification AA1 Release of gaseous or liquid RA1 Release of gaseous or liquid None radioactivity resulting in offsite radioactivity resulting in offsite dose dose greater than 10 mrem TEDE greater than 10 mrem TEDE or 50 or 50 mrem thyroid CDE. mrem thyroid CDE MODE: All MODE: All NEAL # NEI Example EAL Wording EAL VCS1 EAL Wording Difference/Deviation Justification 1 Reading on ANY of the following RA1.1 Reading on any Table R-1 effluent The VCS1 radiation monitors that detect radioactivity effluent radiation monitors greater than radiation monitor > column "ALERT" release to the environment are listed in Table R-1. UE, Alert, the reading shown for 15 minutes for > 15 min. (Notes 1, 2, 3, 4, 5) SAE and GE thresholds for all VCS1 continuously monitored or longer: gaseous release pathways are listed in Table R-1 to (site-specific monitor list and consolidate the information in a single location and, thereby, threshold values) simplify identification of the thresholds by the EAL-user.2 Dose assessment using actual RA1.2 Dose assessment using actual The site boundary is the site-specific receptor point.meteorology indicates doses meteorology indicates doses > 10 greater than 10 mrem TEDE or mrem TEDE or 50 mrem thyroid CDE 50 mrem thyroid CDE at or at or beyond the SITE BOUNDARY beyond (site-specific dose (Notes 3, 4, 5)receptor point).3 Analysis of a liquid effluent N/A N/A VCS1 does not have the capability to perform real-time offsite sample indicates a concentration dose assessment for liquid releases.
This is a deviation or release rate that would result from NEI 9-01 Revision 6.in doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond (site-specific dose receptor point) for one hour of exposure.18 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 4 Field survey results indicate EITHER of the following at or beyond (site-specific dose receptor point): " Closed window dose rates greater than 10 mR/hr expected to continue for 60 minutes or longer.* Analyses of field survey samples indicate thyroid CDE greater than 50 mrem for one hour of inhalation.
RA1.3 Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: " Closed window dose rates > 10 mR/hr expected to continue for> 60 min.* Analyses of field survey samples indicate thyroid CDE >50 mrem for 60 min. of inhalation.(Notes 1, 2)The site boundary is the site-specific receptor point.Notes" The Emergency Director should declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.N/A N/A Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.The classification timeliness note has been standardized across the VCS1 EAL scheme by referencing the "time limit" specified within the EAL wording.The classification timeliness note has been standardized across the VCS1 EAL scheme by referencing the "time limit" specified within the EAL wording.None Note 4 provides site-specific guidance on usability of MSL radiation monitors.Note 4: During a tube rupture with reactor at power RM-G19NB/C monitor readings are affected by 16N therefore they are not reliable until 19 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 reactor has tripped and the* The pre-calculated effluent monitors stable.monitor values presented in Note 5 The pre-calculated effluent Incorporated site-specific EAL numbers associated with EAL #1 should be used for monitor values presented in generic EAL#1.emergency classification EALs RA1.1, RS1.1 and assessments until the results RG1.1 should be used for from a dose assessment emergency classification using actual meteorology are assessments until the available, results from a dose assessment using actual meteorology are available.
20 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification AA2 Significant lowering of water RA2 Significant lowering of water level None level above, or damage to, above, or damage to, irradiated fuel irradiated fuel. MODE: All MODE: All NEI Ex. NEI Example EAL Wording EAL VCS1 EAL Wording Difference/Deviation Justification 1 Uncovery of irradiated fuel in the RA2.1 Uncovery of irradiated fuel in the None REFUELING PATHWAY. REFUELING PATHWAY 2 Damage to irradiated fuel RA2.2 Damage to irradiated fuel resulting in a Site-specific list of radiation monitors bulletized.
resulting in a release of release of radioactivity as indicated by Radiation monitor Hi-Rad alarms specified.
radioactivity from the fuel as a Hi-Rad alarm on any of the following indicated by ANY of the radiation monitors: following radiation monitors:* RM-G8 FHB Refueling (site-specific listing of radiation Bridge Area Gamma monitors, and the associated readings, setpoints and/or
- RM-A6 Fuel Handling Bldg alarms) Exhaust* RM-G6 Rx Bldg Refueling Bridge* RM-G17A/B Rx Bldg Manipulator Crane 3 Lowering of spent fuel pool level RA2.3 Lowering of spent fuel pool level to V. C. Summer Unit 1 expects to comply with the NRC Order to (site-specific Level 2 value). Level 2 (ele. 455' 6") implementation date for Order EA-1 2-051, Modifying Licenses[See Developer Notes] with Regard to Requirements for Reliable Spent Fuel Pool Instrumentation.
Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1 -ele. 461' 6"), SFP level -19 ft. above the top of the fuel racks (Level 2 -ele. 455' 6") and SFP level at the top of 21 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 I I I I Ithe fuel racks (Level 3 -ele. 437' 0").22 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI IC# NEI IC Wording ICs) VCS1 IC Wording Difference/Deviation Justification iC#(s)AA3 Radiation levels that impede RA3 Radiation levels that impede access to None access to equipment necessary equipment necessary for normal plant for normal plant operations, operations, cooldown or shutdown.cooldown or shutdown MODE: All MODE: All NEI Ex. VCS1 EAL # NEI Example EAL Wording EAL # VCS1 EAL Wording Difference/Deviation Justification 1 Dose rate greater than 15 mR/hr RA3.1 Dose rate > 15 mR/hr in EITHER of No other site-specific areas requiring continuous occupancy in ANY of the following areas: the following areas: exist at VCS1." Control Room
- Control Room (RM-G1) RM-G1 is the installed CR ARM." Central Alarm Station
- Central Alarm Station (by The CAS does not have installed area radiation monitoring
- (other site-specific survey) and thus must be determined by survey.areas/rooms) 2 An UNPLANNED event results RA3.2 An UNPLANNED event results in The site-specific list of plant rooms or areas with entry-related in radiation levels that prohibit or radiation levels that prohibit or impede mode applicability are tabularized in Table R-2.impede access to any of the access to any Table R-2 area (Note 6) Table R-2 lists plant areas, not specific rooms. Therefore the following plant rooms or areas: word "rooms" has been deleted.(site-specific list of plant rooms or areas with entry-related mode applicability identified)
Note If the equipment in the listed N/A Note 6 If the equipment in the listed Table R-2 lists plant areas, not specific rooms. Therefore the room or area was already area was already inoperable word "rooms" has been deleted.inoperable or out-of-service or out-of-service before the before the event occurred, then event occurred, then no no emergency classification is emergency classification is warranted.
warranted.
23 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 Table H-2 Safe Operation
& Shutdown Areas Area Mode Applicability Auxiliary Building 374' 3 Auxiliary Building 388' 3, 4, 5 Auxiliary Building 400' 4, 5 Auxiliary Building 412 3, 4, 5 Auxiliary Building 436' 1, 2, 3, 4, 5 Auxiliary Building 463' 3, 4, 5 Intermediate Building 412' 3 Intermediate Building 436' 4, 5 Intermediate Building 463' 3, 4, 5 Control Building 412' 2, 3 Control Building 436' 3, 4, 5 Turbine Building (All levels) 1, 2 24 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI IC# NEI IC Wording ICs) VCS1 IC Wording Difference/Deviation Justification IC#(s)AS1 Release of gaseous radioactivity RS1 Release of gaseous radioactivity None resulting in offsite dose greater resulting in offsite dose greater than than 100 mrem TEDE or 500 100 mrem TEDE or 500 mrem thyroid mrem thyroid CDE CDE MODE: All MODE: All MODE: All NEI Ex. NEI Example EAL Wording VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #Reading on ANY of the following RS1.1 Reading on any Table R-1 effluent The VCS1 radiation monitors that detect radioactivity effluent radiation monitors greater than radiation monitor > column "SAE" for > release to the environment are listed in Table R-1. UE, Alert, the reading shown for 15 15 min. SAE and GE thresholds for all VCS1 continuously monitored minutes or longer: (Notes 1, 2, 3, 4, 5) gaseous release pathways are listed in Table R-1 to (site-specific monitor list and consolidate the information in a single location and, thereby, threshold values) simplify identification of the thresholds by the EAL-user.2 Dose assessment using actual RS1.2 Dose assessment using actual The site boundary is the site-specific receptor point.meteorology indicates doses meteorology indicates doses > 100 greater than 100 mrem TEDE or mrem TEDE or 500 mrem thyroid CDE 500 mrem thyroid CDE at or at or beyond the SITE BOUNDARY beyond (site-specific dose (Notes 3, 4, 5)receptor point)3 Field survey results indicate RS1.3 Field survey results indicate EITHER The site boundary is the site-specific receptor point.EITHER of the following at or of the following at or beyond the SITE beyond (site-specific dose BOUNDARY: receptor point):
- Closed window dose rates >" Closed window dose rates 100 m R/hr expected to continue greater than 100 mR/hr for > 60 min.expected to continue for 60 minutes or longer. e Analyses of field survey* Analyses of field survey samples indicate thyroid CDE >samples indicate thyroid 500 mrem for 60 min. of CDE greater than 500 inhalation.
25 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 mrem for one hour of inhalation.(Notes 1,2)Notes" The Emergency Director should declare the Site Area Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes." N/A* The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.The classification timeliness note has been standardized across the VCS1 EAL scheme by referencing the "time limit" specified within the EAL wording.The classification timeliness note has been standardized across the VCS1 EAL scheme by referencing the "time limit" specified within the EAL wording.None Note 4 provides site-specific guidance on usability of MSL radiation monitors.Incorporated site-specific EAL numbers associated with generic EAL#1.Note 4: During a tube rupture with reactor at power RM-G19A/B/C monitor readings are affected by 16N therefore they are not reliable until reactor has tripped and the monitors stable.Note 5 The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
26 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCSi 27of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification AS2 Spent fuel pool level at (site- RS2 Spent fuel pool level at the top of the Top of the fuel racks is the site-specific Level 3 description.
specific Level 3 description) fuel racks MODE: All NEI Ex. NEI Example EAL Wording EAL VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #1 Lowering of spent fuel pool level RS2.1 Lowering of spent fuel pool level to V. C. Summer Unit 1 expects to comply with the NRC Order to (site-specific Level 3 value) Level 3 (ele. 437' 0") implementation date for Order EA-1 2-051, Modifying Licenses with Regard to Requirements for Reliable Spent Fuel Pool Instrumentation.
Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1 -ele. 461' 6"), SFP level -19 ft. above the top of the fuel racks (Level 2 -ele. 455' 6") and SFP level at the top of the fuel racks (Level 3 -ele. 437' 0").28 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification AGI Release of gaseous radioactivity RG1 Release of gaseous radioactivity None resulting in offsite dose greater resulting in offsite dose greater than 1,000 mrem TEDE or than 1,000 mrem TEDE or 5,000 5,000 mrem thyroid CDE. mrem thyroid CDE MODE: All MODE: All NEI Ex. NIEapeELWrig VCS1 EAL # NEI Example EAL # VCS1 EAL Wording Difference/Deviation Justification 1 Reading on ANY of the following RG1.1 Reading on any Table R-1 effluent The VCS1 radiation monitors that detect radioactivity effluent radiation monitors greater than radiation monitor > column "GE" release to the environment are listed in Table R-1. UE, Alert, SAE the reading shown for 15 for a 15 min. and GE thresholds for all VCS1 continuously monitored gaseous minutes or longer: (Notes 1, 2, 3, 4, 5) release pathways are listed in Table R-1 to consolidate the (site-specific monitor list and information in a single location and, thereby, simplify identification of threshold values) the thresholds by the EAL-user.2 Dose assessment using actual RG1.2 Dose assessment using actual The site boundary is the site-specific receptor point.meteorology indicates doses meteorology indicates doses >greater than 1,000 mrem TEDE 1000 mrem TEDE or or 5,000 mrem thyroid CDE at 5000 mrem thyroid CDE at or or beyond (site-specific dose beyond the SITE BOUNDARY receptor point). (Notes 3, 4, 5)3 Field survey results indicate RG1.3 Field survey results indicate The site boundary is the site-specific receptor point.EITHER of the following at or EITHER of the following at or beyond (site-specific dose beyond the SITE BOUNDARY: receptor point):
- Closed window dose rates >* Closed window dose rates 1000 mR/hr expected to greater than 1,000 mR/hr continue for a 60 min.expected to continue for 60 minutes or longer.
- Analyses of field survey* Analyses of field survey samples indicate thyroid CDE samples indicate thyroid CDE > 5000 mrem for 60 min. of greater than 5,000 mrem for inhalation.
29 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 I one hour of inhalation.
I___ [ (Notes 1, 2) 1 Notes" The Emergency Director should declare the Site Area Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded." If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes." N/A" The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until the results from a dose assessment Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.Note 4: During a tube rupture with reactor at power RM-G19A/B/C monitor readings are affected by 16N therefore they are not reliable until reactor has tripped and the monitors stable.Note 5 The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency The classification timeliness note has been standardized across the VCS1 EAL scheme by referencing the "time limit" specified within the EAL wording.The classification timeliness note has been standardized across the VCS1 EAL scheme by referencing the "time limit" specified within the EAL wording.None Note 4 provides site-specific guidance on usability of MSL radiation monitors.Incorporated site-specific EAL numbers associated with generic EAL#1.30 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 using actual meteorology are available.
classification assessments until the results from a dose assessment using actual meteorology are available.
31 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification AG2 Spent fuel pool level cannot be RG2 Spent fuel pool level cannot be Top of the fuel racks is the site-specific Level 3 description.
restored to at least (site-specific restored to at least the top of the fuel Level 3 description) for 60 racks for 60 minutes or longer minutes or longe MODE: All NEI Ex. NEI Example EAL Wording VCS1 VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #1 Spent fuel pool level cannot be RG2.1 Spent fuel pool level cannot be V. C. Summer Unit 1 expects to comply with the NRC Order restored to at least (site-specific restored to at least Level 3 (ele. 437' implementation date for Order EA-1 2-051, Modifying Licenses Level 3 value) for 60 minutes or 0") for > 60 min. (Note 1) with Regard to Requirements for Reliable Spent Fuel Pool longer Instrumentation.
Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1 -ele. 461' 6"), SFP level -19 ft. above the top of the fuel racks (Level 2 -ele. 455' 6") and SFP level at the top of the fuel racks (Level 3 -ele. 437' 0").Note The Emergency Director should N/A The Emergency Director should None declare the General Emergency declare the General Emergency promptly upon determining that promptly upon determining that 60 60 minutes has been exceeded, minutes has been exceeded, or will or will likely be exceeded.
likely be exceeded.32 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 Category C Cold Shutdown / Refueling System Malfunction 33 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification CUl UNPLANNED loss of (reactor CUi UNPLANNED loss of reactor None vessel/RCS
[PWR] or RPV vessel/RCS inventory for 15[BWR]) inventory for 15 minutes minutes or longer or longer. MODE: 5 -Cold Shutdown, MODE: Cold Shutdown, Refueling Refueling NEI Ex. VCS1 EAL # NEI Example EAL Wording EAL # VCS1 EAL Wording Difference/Deviation Justification 1 UNPLANNED loss of reactor CU1.1 UNPLANNED loss of reactor None coolant results in (reactor coolant results in reactor vessel/RCS
[PWRF or RPV vessel/RCS level less than a[BWR]) level less than a required lower limit for > 15 required lower limit for 15 min. (Note 1)minutes or longer.2 a. (Reactor vessel/RCS
[PWRI CU1.2 Reactor vessel/RCS level Table C-1 provides a tabularized list of site-specific applicable sumps cannot be monitored and tanks.or RPV [BWb]) level cannot AND be monitored.
UNPLANNED increase in any AND Table C-1 sump or tank levels due to a loss of reactor b. UNPLANNED increase in vessel/RCS inventory (site-specific sump and/or tank) levels.Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across the declare the Unusual Event Director should VCS1 EAL scheme by referencing the "time limit" specified within the promptly upon determining that declare the event EAL wording.15 minutes has been exceeded, promptly upon or will likely be exceeded.
determining that time limit has been exceeded, or will likely 34 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 I I I be exceeded.Table C-1 Sumps & Tanks* RB Sump* CCW surge tank* PRT* RCDT 35 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification CU2 Loss of all but one AC power CU2 Loss of all but one AC power ESF is the site-specific designation for the VCS1 emergency AC source to emergency buses for source to ESF buses for 15 buses.15 minutes or longer. minutes or longer.MODE: Cold Shutdown, MODE: 6 -Cold Shutdown, Refueling, Defueled Refueling, Defueled NEI Ex. VCS1 EA E NEI Example EAL Wording EAL VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #1 a. AC power capability to (site- CU2.1 AC power capability to 7.2 KV 7.2 KV ESF buses 1 DA and 1 DB are the site-specific emergency specific emergency buses) is ESF buses 1 DA and 1DB buses.reduced to a single power reduced to a single power Site-specific AC power sources are tabularized in Table C-2.source for 15 minutes or source (Table C-2) for > 15 min.longer. (Note 1)AND AND b. Any additional single power Any additional single power source failure will result in source failure will result in loss of loss of all AC power to all AC power to SAFETY SAFETY SYSTEMS. SYSTEMS Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across the declare the Unusual Event Director should declare VCS1 EAL scheme by referencing the "time limit" specified within promptly upon determining that the event promptly the EAL wording.15 minutes has been exceeded, upon determining that or will likely be exceeded.
time limit has been exceeded, or will likely be exceeded.36 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 Table C-2 AC Power Supplies Offsite:* 115 KV power to XTF-4 and XTF-5 0 230 KV power to XTF-31 0 Parr Hydro Plant 13.8 KV power to ESF bus 1 DA or 1DB Onsite:* Diesel Generator A* Diesel Generator B 37 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification CU3 UNPLANNED increase in RCS CU3 UNPLANNED increase in RCS None temperature temperature MODE: Cold Shutdown, MODE: Cold Shutdown, Refueling Refueling NEI Ex. VCS1 EA NEI Example EAL Wording EAL VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #1 UNPLANNED increase in RCS CU3.1 UNPLANNED increase in RCS 200°F is the site-specific Tech. Spec. cold shutdown temperature temperature to greater than (site- temperature to > 200°F limit.specific Technical Specification cold shutdown temperature limit)2 Loss of ALL RCS temperature CU3.2 Loss of all RCS temperature and None and (reactor vessel/RCS
[PWR] reactor vessel/RCS level or RPV [BWR]) level indication indication for > 15 min.for 15 minutes or longer. (Note 1)Note The Emergency Director should N/A Note 1: The Emergency Director The classification timeliness note has been standardized across the declare the Unusual Event should declare the event VCS1 EAL scheme by referencing the "time limit" specified within the promptly upon determining that promptly upon EAL wording.15 minutes has been exceeded, determining that time or will likely be exceeded limit has been exceeded, or will likely be exceeded.38 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification CU4 Loss of Vital DC power for 15 CU4 Loss of Vital DC power for 15 None minutes or longer. minutes or longer.MODE: Cold Shutdown, MODE: 5 -Cold Shutdown, 6 -Refueling Refueling NEI Ex. NEI Example EAL Wording VCS1 VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #1 Indicated voltage is less than CU4.1 < 108 VDC on required DC 108 VDC is the site-specific minimum vital DC bus voltage.(site-specific bus voltage value) buses (Train A or Train B vital 125 Train A or Train B vital 125 VDC system are the site-specific vital on required Vital DC buses for 15 VDC system) for > 15 min. (Note DC buses.minutes or longer. 1)Note The Emergency Director should N/A Note 1: The Emergency Director The classification timeliness note has been standardized across the declare the Unusual Event should declare the event promptly VCS1 EAL scheme by referencing the "time limit" specified within promptly upon determining that upon determining that time limit the EAL wording.15 minutes has been exceeded, has been exceeded, or will likely or will likely be exceeded.
be exceeded.39 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI IC# NEI IC Wording ICs) VCS1 IC Wording Difference/Deviation Justification IC#(s)CU5 Loss of all onsite or offsite CU5 Loss of all onsite or offsite None communications capabilities, communications capabilities.
MODE: Cold Shutdown, MODE: Cold Shutdown, Refueling, Defueled Refueling, Defueled NEI Ex. VCS1 EA NEI Example EAL Wording EAL VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #Loss of ALL of the following CU5.1 Loss of all Table C-4 onsite Example EALs #1, 2 and 3 have been combined into a single EAL onsite communication methods: communication methods for simplification of presentation.(site specific list of OR Example EAL condition
- 2 and #3 have been combined into a communications methods) Loss of all Table C-4 ORO/NRC single statement because the list of ORO and NRC 2 Loss of ALL of the following ORO communication methods communications methods are the same.communications methods: Table b-4 provides a site-specific list of onsite and ORO/NRC (site specific list of communications methods.communications methods)3 Loss of ALL of the following NRC communications methods: (site specific list of communications methods)Table C-4 Communication Methods System Onsite ORO/NRC Gai-Tronics system X Radio system X Internal Telephone system X Telephone land lines X X Fiberoptic links X Satellite phone system X Federal Telephone System (ENS) X ESSX X 40 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI IC# NEI IC Wording ICs) VCS1 IC Wording Difference/Deviation Justification IC#(s)CA1 Loss of (reactor vessel/RCS CA1 Loss of reactor vessel/RCS None[PWR] or RPV [BWR]) inventory inventory MODE: Cold Shutdown, MODE: Cold Shutdown, Refueling Refueling NEI Ex. VCS1 EAL # NEI Example EAL Wording EAL # VCS1 EAL Wording Difference/Deviation Justification Loss of (reactor vessel/RCS CAl.1 Loss of reactor vessel/RCS 429'-6" elevation, 64.2% (rounded to 64.5%) RVLIS Narrow Range is[PWR] or RPV [BWRJ) inventory inventory as indicated by level < the site-specific reactor vessel level corresponding to the bottom of as indicated by level less than 429'-6" elevation, < 64.5% the RCS hot leg penetration.(site-specific level). RVLIS Narrow Range (bottom of hot leg penetration) 2 a. (Reactor vessel/RCS
[PWR] CA1.2 Reactor vessel/RCS level Table C-1 provides a tabularized list of site-specific applicable sumps or RPV [BWR]) level cannot cannot be monitored for > 15 and tanks.be monitored for 15 minutes min. (Note 1)or longer AND AND UNPLANNED increase in any b. UNPLANNED increase in Table C-1 sump or tank levels (site-specific sump and/or due to a loss of reactor tank) levels due to a loss of vessel/RCS inventory (reactor vessel/RCS
[PWR]or RPV [BWR]) inventory.
Note The Emergency Director should N/A Note 1: The Emergency Director The classification timeliness note has been standardized across the declare the Alert promptly upon should declare the event VCS1 EAL scheme by referencing the "time limit" specified within the determining that 15 minutes has promptly upon determining that EAL wording.been exceeded, or will likely be time limit has been exceeded, or exceeded will likely be exceeded.41 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI IC# NEI IC Wording ICs) VCS1 IC Wording Difference/Deviation Justification IC#(s)CA2 Loss of all offsite and all onsite CA2 Loss of all offsite and all onsite ESF is the site-specific designation for the VCS1 emergency AC AC power to emergency buses AC power to ESF buses for 15 buses.for 15 minutes or longer minutes or longer.MODE: Cold Shutdown, MODE: Cold Shutdown, Refueling, Defueled Refueling, Defueled NEI Ex. NEVxapeCASorig1A EAL # NEI Example EAL Wording VCS1 VCS1 EAL Wording Difference/Deviation Justification 1 Loss of ALL offsite and ALL CA2.1 Loss of all offsite and all onsite 7.2 KV ESF buses 1 DA and 1DB are the site-specific emergency onsite AC Power to (site-specific AC power (Table C-2) capability buses.emergency buses) for 15 to 7.2 KV ESF buses Site-specific AC power sources are tabularized in Table C-2.minutes or longer. 1 DA and 1DB for > 15 min.(Note 1)Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across the declare the Unusual Event Director should declare VCS1 EAL scheme by referencing the "time limit" specified within the promptly upon determining that the event promptly EAL wording.15 minutes has been exceeded, upon determining that or will likely be exceeded.
time limit has been exceeded, or will likely be exceeded.42 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI IC# NEI IC Wording ICs) VCS1 IC Wording Difference/Deviation Justification IC#(S)CA3 Inability to maintain the plant in CA3 Inability to maintain the plant in None cold shutdown.
cold shutdown.MODE: Cold Shutdown, MODE: Cold Shutdown, Refueling Refueling NEI Ex. VCS1 EAL # NEI Example EAL Wording EAL # VCS1 EAL Wording Difference/Deviation Justification 1 UNPLANNED increase in RCS CA3.1 UNPLANNED increase in RCS 200°F is the site-specific Tech. Spec. cold shutdown temperature temperature to greater than temperature to > 200OF for > limit.(site-specific Technical Table C-3 duration (Note 1) Table C-3 is the site-specific implementation of the generic RCS Specification cold shutdown Reheat Duration Threshold table.temperature limit) for greater than the duration specified in the following table.2 UNPLANNED RCS pressure CA3.2 UNPLANNED RCS pressure 10 psig is the site-specific pressure increase readable by Control increase greater than (site- increase > 10 psig (This EAL Room indications.
specific pressure reading). (This does not apply during water-EAL does not apply during solid plant conditions) water-solid plant conditions.
[PWR])Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across the declare the Unusual Event Director should declare VCS1 EAL scheme by referencing the "time limit" specified within the promptly upon determining that the event promptly EAL wording.15 minutes has been exceeded, upon determining that or will likely be exceeded.
time limit has been exceeded, or will likely be exceeded.43 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 Table: RCS Heat-up Duration Thresholds RCS Status Containment Closure Status Heat-up Duration Intact (but not at reduced Not applicable 60 minutes*inventory
[PWR])Not intact (or at reduced Established 20 minutes*inventory
[PWR]) Not Established 0 minutes* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.
Table C-3: RCS Heat-up Duration Thresholds RCS Status Containment Closure Status Heat-up Duration Intact AND not at REDUCED N/A 60 min.*INVENTORY Not intact OR at REDUCED established 20 min.*INVENTORY not established 0 min.* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.
44 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification CA6 Hazardous event affecting a HA2 Seismic event affecting a Generic ICs CA6, SA9 and associated example EALs have been re-SAFETY SYSTEM needed for SAFETY SYSTEM needed for assigned to the associated event hazards sub-categories for human the current operating mode. the current operating mode. factors considerations.
Although ICs CA6 and SA9 are generically MODE: Cold Shutdown, MODE: All assigned symptomatically to the system malfunctions related Refueling categories, it is the initiating events such as seismic activity, natural HA3 Hazardous event affecting a or technological events or fires/explosions that trigger assessment of SAFETY SYSTEM needed for safety system degradation associated with these EALs. This re-the current operating mode. assignment within the VCS1 EAL scheme results in no change of intent.MODE: All Mode applicability has been expanded to ALL modes based on the HA4 FIRE or EXPLOSION event consolidation of ICs CA6 and SA9 and to address Spent Fuel Pool affecting a SAFETY SYSTEM system degradation during full core off load.needed for the current operating mode.MODE: All NEI Ex. VCS1 EAL # NEI Example EAL Wording EAL # VCS1 EAL Wording Difference/Deviation Justification 1 a. The occurrence of ANY of the HA2.1 Seismic event resulting in This EAL represents the seismic event portion of the generic CA6 following hazardous events: EITHER of the following:
and SA9 example EAL #1* Seismic event 0 Event damage has caused (earthquake) indications of degraded" Internal or external performance in at least one flooding event train of a SAFETY" High winds or tornado SYSTEM needed for the strike current operating mode" FIRE 0 The event has caused" EXPLOSION VISIBLE DAMAGE to a" (site-specific hazards) SAFETY SYSTEM* Other events with similar component or structure hazard characteristics as needed for the current determined by the Shift operating mode Manager HA3.1 The occurrence of any Table C- This EAL represents the non-seismic naturally occurring and 'other'45 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 AND b. EITHER of the following:
- 1. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.OR 2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.5 hazardous event AND EITHER of the following:
- Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode* The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode event portion of the generic CA6 and SA9 example EAL #1 The list of hazardous events has been tabularized in Table H-3 No other VCS1 -specific unique hazards have been identified for inclusion of the EAL.HA4.1 FIRE or EXPLOSION resulting in EITHER of the following:
- Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode* The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode This EAL represents the fire and explosion portion of the generic CA6 and SA9 example EAL #1 Table H-3 Hazardous Events* Internal or external FLOODING event" High winds or tornado strike 46 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1* Other events with similar hazard characteristics as determined by the Shift Supervisor 47 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI IC# NEI IC Wording ICs) VCS1 IC Wording Difference/Deviation Justification IC#(S)CS1 Loss of (reactor vessel/RCS CS1 Loss of reactor vessel/RCS None[PWR] or RPV [BWR]) inventory inventory affecting core decay affecting core decay heat heat removal capability removal capability.
MODE: Cold Shutdown, MODE: Cold Shutdown, Refueling Refueling NEI Ex. NIEapeELWrig VCS1 EAL # NEI Example EAL # VCS1 EAL Wording Difference/Deviation Justification
- a. CONTAINMENT CLOSURE CS1.1 CONTAINMENT CLOSURE not 429' elevation, 62.9% (rounded to 63%) RVLIS Narrow Range not established, established corresponds to the site-specific indication of 6" below the RCS hot AND AND leg penetration.
- b. (Reactor vessel/RCS
[PWR] Reactor vessel level < 429'or RPV [BWR]) level less than elevation, < 63% RVLIS Narrow (site-specific level). Range (6" below the bottom of the hot leg penetration) 2 a. CONTAINMENT CLOSURE CS1.2 CONTAINMENT CLOSURE 427' elevation, 57.9% (rounded to 58%) RVLIS Narrow Range established, established corresponds to the site-specific indication of when core uncovery is AND AND about to occur.b.(Reactor vessel/RCS
[PWR] Reactor vessel level < 427'or RPV [BWR]) level less than elevation, < 58% RVLIS Narrow (site-specific level). Range (top of active fuel)3 a. (Reactor vessel/RCS
[PWR] CS1.3 Reactor vessel/RCS level RM-G6 (Rx Refueling Bridge) and RM-G1 7A/B (Rx Bldg Manipulator or RPV [BWR]) level cannot cannot be monitored for > 30 Crane) are the site-specific range radiation monitors that would be be monitored for 30 minutes min. (Note 1) indicative of possible core uncovery in the Refueling mode.or longer. AND The dose rate due to core shine when the top of the core becomes AND Core uncovery is indicated by uncovered should result in off-scale indication on the listed monitors.b. Core uncovery is indicated by any of the following:
Table C-1 provides a tabularized list of site-specific applicable sumps 48 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 ANY of the following:
- (Site-specific radiation monitor) reading greater than (site-specific value)* Erratic source range monitor indication
[PWR]* UNPLANNED increase in (site-specific sump and/or tank) levels of sufficient magnitude to indicate core uncovery* (Other site-specific indications)
- RM-G6 Rx Bldg Refueling Bridge or RM-G17A/B Rx Bldg Manipulator Crane offscale-high
- Erratic source range monitor indication UNPLANNED increase in any Table C-1 sump / tank level of sufficient magnitude to indicate core uncovery and tanks.Note The Emergency Director should N/A Note 1: The Emergency Director The classification timeliness note has been standardized across the declare the Site Area should declare the event VCS1 EAL scheme by referencing the "time limit" specified within the Emergency promptly upon promptly upon EAL wording.determining that 30 minutes has determining that time been exceeded, or will likely be limit has been exceeded exceeded, or will likely be exceeded.49 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI IC# NEI IC Wording ICs) VCS1 IC Wording Difference/Deviation Justification IC#(s)CG1 Loss of (reactor vessel/RCS CG1 Loss of reactor vessel/RCS None[PWR] or RPV [BWR]) inventory inventory affecting fuel clad affecting fuel clad integrity with integrity with containment containment challenged challenged MODE: Cold Shutdown, MODE: Cold Shutdown, Refueling Refueling NEI Ex. VCS1 EA NEI Example EAL Wording EAL VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #a. (Reactor vessel/RCS
[PWR] or CG1.1 Reactor vessel level < 427' 427' elevation, 57.9% (rounded to 58%)RVLIS Narrow Rang RPV [BWR]) level less than elevation, < 58% RVLIS Narrow corresponds to the site-specific indication of when core uncovery is (site-specific level) for 30 Range (top of active fuel) for > 30 about to occur.minutes or longer. min. 4% hydrogen concentration in the presence of oxygen represents AND AND an explosive mixture in containment.
Any of the following indications The generic Containment Challenge table has been implemented Containment Challenge Table of containment challenge:
as a bulletized list within the EAL wording for usability.(see below).
- CONTAINMENT CLOSURE not established (Note 7)* Containment hydrogen concentration
> 4%" UNPLANNED increase in Containment pressure 2 a. (Reactor vessel/RCS
[PWR] or CG2.1 Reactor vessel/RCS level cannot RM-G6 (Rx Refueling Bridge) and RM-G17A/B (Rx Bldg RPV [BWR]) level cannot be be monitored for > 30 min. (Note Manipulator Crane) are the site-specific range radiation monitors monitored for 30 minutes or 1) that would be indicative of possible core uncovery in the Refueling longer. AND mode.AND CThe dose rate due to core shine when the top of the core becomes b. Core uncovery is indicated by uncovered should result in off-scale indication on the listed b. Core uncovery is indicated by any of the following:
monitors.50 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 ANY of the following:
- (Site-specific radiation monitor) reading greater than (site-specific value)* Erratic source range monitor indication
[PWR]* UNPLANNED increase in (site-specific sump and/or tank) levels of sufficient magnitude to indicate core uncovery* (Other site-specific indications)
AND c. ANY indication from the Containment Challenge Table (see below).* RM-G6 Rx Bldg Refueling Bridge or RM-G17A/B Rx Bldg Manipulator Crane offscale-high
- Erratic source range monitor indication" UNPLANNED increase in any Table C-1 sump / tank level of sufficient magnitude to indicate core uncovery AND Any of the following indications of containment challenge:
- CONTAINMENT CLOSURE not established (Note 7)* Containment hydrogen concentration
> 4%* UNPLANNED increase in Containment pressure Table C-1 provides a tabularized list of site-specific applicable sumps and tanks.4% hydrogen concentration in the presence of oxygen represents an explosive mixture in containment.
The generic Containment Challenge table has been implemented as a bulletized list within the EAL wording for usability.
Note The Emergency Director should declare the General Emergency promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.N/A N/A Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.Note 7: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required.The classification timeliness note has been standardized across the VCS1 EAL scheme by referencing the "time limit" specified within the EAL wording.Note 7 implements the asterisked note associated with the generic Containment Challenge table.L __________________________________
I. I +/-51 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 Containment Challenge Table* CONTAINMENT CLOSURE not established*
- (Explosive mixture) exists inside containment
- UNPLANNED increase in containment pressure* Secondary containment radiation monitor reading above (site-specific value) [BWR]* If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.52 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 Category D Permanently Defueled Station Malfunction 53 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI IC# NEI IC Wording ICs) VCS1 IC Wording Difference/Deviation Justification IC#(s)PD-AU1 Recognition Category D N/A N/A NEI Recognition Category PD ICs and EALs are applicable only to PD-AU2 Permanently Defueled Station permanently defueled stations.
VCS1 is not a defueled station.PD-SU1 PD-HU1 PD-HU2 PD-HU3 PD-AA1 PD-AA2 PD-HA1 PD-HA3 54 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 Category E Independent Spent Fuel Storage Installation 55 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI IC# NEI IC Wording ICs) VCS1 IC Wording Difference/Deviation Justification IC#(S)E-HU1 Damage to a loaded cask IU1 Damage to a loaded cask Changed IC identifies from E to I. Based on operations feedback CONFINEMENT BOUNDARY CONFINEMENT BOUNDARY the use of I makes better sense as in "ISFSI".MODE: All MODE: All NEI Ex. VCS1 EAL # NEI Example EAL Wording EAL # VCS1 EAL Wording Difference/Deviation Justification 1 Damage to a loaded cask IU1.1 Damage to a loaded cask 60 mrem/hr (T + n) on the top of the overpack and 600 mrem/hr (T CONFINEMENT BOUNDARY as CONFINEMENT BOUNDARY + r0 on the side of the overpack represent 2 times the site-specific indicated by an on-contact as indicated by an on-contact cask technical specification allowable levels per the ISFSI Technical radiation reading greater than (2 radiation reading greater than Specifications (CoC).times the site-specific cask the following on the surface of specific technical specification the spent fuel cask (overpack):
allowable radiation level) on the surface of the spent fuel cask. e 60 mrem/hr (r + r) on the top of the overpack* 600 mrem/hr (T + r0 on the side of the overpack 56 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 Category F Fission Product Barrier Degradation 57 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI IC# NEI IC Wording ICs) VCS1 IC Wording Difference/Deviation Justification IC#(s)FA1 Any Loss or any Potential Loss of FA1 Any loss or any potential loss of None either the Fuel Clad or RCS either Fuel Clad or RCS barrier. MODE: 1 -Power Operation, 2 -MODE: Power Operation, Hot Startup, 3 -Hot Standby, 4 -Hot Standby, Startup, Hot Shutdown Shutdown NEI Ex. VCS1 EAL # NEI Example EAL Wording EAL # VCS1 EAL Wording Difference/Deviation Justification 1 Any Loss or any Potential Loss of FA1.1 Any loss or any potential loss of Table F-1 provides the fission product barrier loss and potential loss either the Fuel Clad or RCS either Fuel Clad or RCS (Table thresholds.
barrier. F-1)58 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification FS1 Loss or Potential Loss of any two FS1 Loss or potential loss of any two None barriers barriers MODE: Power Operation, Hot MODE: 1 -Power Operation, 2 -Standby, Startup, Hot Shutdown Startup, 3 -Hot Standby, 4 -Hot Shutdown NEI Ex. VCS11 EAL# NEI Example EAL Wording EAL # VCS1 EAL Wording Difference/Deviation Justification 1 Loss or Potential Loss of any two FS1.1 Loss or potential loss of any two Table F-1 provides the fission product barrier loss and potential loss barriers barriers thresholds.
59 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI IC# NEI IC Wording ICs) VCS1 IC Wording Difference/Deviation Justification IC#(s)FG1 Loss of any two barriers and FG1 Loss of any two barriers and loss None Loss or Potential Loss of third or potential loss of the third barrier barrier MODE: Power Operation, Hot MODE: 1 -Power Operation, 2 -Standby, Startup, Hot Shutdown Startup, 3 -Hot Standby, 4 -Hot Shutdown NEI Ex. VCS1 EAL # NEI Example EAL Wording EAL # VCS1 EAL Wording Difference/Deviation Justification 1 Loss of any two barriers and FG1.1 Loss of any two barriers Table F-1 provides the fission product barrier loss and potential loss Loss or Potential Loss of third AND thresholds.
barrier Loss or potential loss of the third barrier (Table F-i)60 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 PWR Fuel Clad Fission Product Barrier Degradation Thresholds NEI VCS1 FPB NEI Threshold Wording FPB VCS1 FPB Wording Difference/Deviation Justification FPB# FPB #(s)FC Loss RCS or SG Tube Leakage N/A N/A N/A 1 Not Applicable FC Loss Inadequate Heat Removal FC Loss CSFST Core Cooling-RED Consistent with the generic guidance provided in PWR FPB 2 A. Core exit thermocouple 2.A path conditions met Developer Notes, VCS1 has opted to use CSFST conditions in lieu readings greater than (site- of specifying the parameters and values associated with this specific temperature value). threshold.
FC Loss RCS Activity/CMNT Rad FC Loss RM-G7 or RM-G18 CNTMT HI RM-G7 or RM-G18 are the site-specific containment high range 3 A. Containment radiation 3.A RNG Gamma > 2,000 R/hr radiation monitors.
The specified monitors and values are monitor reading greater than containment radiation monitor readings (rounded down to nearest (site-specific value) whole number for readability) corresponding to -2% fuel failure.OR FC Loss Dose equivalent 1-131 coolant None B. (Site-specific indications activity > 300 ICi/gm that reactor coolant activity is 3.B greater than 300 gCi/gm dose equivalent 1-131)FC Loss CNMT Integrity or Bypass N/A N/A N/A 4 Not Applicable FC Loss Other Indications N/A N/A No other site-specific Fuel Clad Loss indication has been identified 5 for VCS1.A. (site-specific as applicable)
FC Loss ED Judgment FC Loss Any condition in the opinion of None the ED that indicates loss of the 6 A. ANY condition in the opinion 5.A fue cladibarer of the Emergency Director that indicates Loss of the Fuel Clad Barrier.61 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI VCS1 FPB NEI Threshold Wording FPB VCS1 FPB Wording Difference/Deviation Justification FPB# FPB #(s)FC RCS or SG Tube Leakage N/A N/A N/A P-Loss A. RCS/reactor vessel level 1 less than (site-specific level)FC Inadequate Heat Removal FC CSFST Core Cooling-ORANGE Consistent with the generic guidance provided in PWR FPB P-Loss A. Core exit thermocouple P-Loss path conditions met Developer Notes, VCS1 has opted to use CSFST conditions in lieu 2 readings greater than (site- 2.A of specifying the parameters and values associated with this 2 ednsgetrta site- threshold.
specific temperature value)OR FC CSFST Heat Sink-RED path Consistent with the generic guidance provided in PWR FPB B. Inadequate RCS heat P-Loss conditions met Developer Notes, VCS1 has opted to use CSFST conditions in lieu removal capability via steam 2.B of specifying the parameters and values associated with this generators as indicated by (site- AND threshold.
specific indications).
Heat sink is required The phrase "and heat sink required" has been added to preclude the need for classification for conditions in which RCS pressure is less than SG pressure or Heat Sink-RED path entry was created through operator action directed by an EOP.FC RCS Activity/CM NT Rad N/A N/A N/A P-Loss Not Applicable 3 FC CNMT Integrity or Bypass N/A N/A N/A P-Loss Not Applicable 4 FC Other Indications N/A N/A No other site-specific Fuel Clad Potential Loss indication has been P-Loss identified for VCS1.A. (site-specific as applicable) 5 62 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI VCS1 FPB NEI Threshold Wording FPB VCS1 FPB Wording Difference/Deviation Justification FPB# FPB #(s)FC Emergency Director FC Any condition in the opinion of None P-Loss Judgment P-Loss the ED that indicates potential 6 A. Any condition in the opinion 5.B loss of the fuel clad barrier of the Emergency Director that indicates Potential Loss of the Fuel Clad Barrier.63 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 PWR RCS Fission Product Barrier Degradation Thresholds NEI VCS1 FPB FPB NEI IC Wording #(s) VCS1 FPB Wording Difference/Deviation Justification FPB# #(s)RCS RCS or SG Tube Leakage RCS Loss An automatic or manual None Loss A. An automatic or manual 1.A ECCS (SI) actuation required 1 ECCS (SI) actuation is by EITHER: required by EITHER of the
- UNISOLABLE RCS following:
leakage 1. UNISOLABLE RCS 9 SG tube RUPTURE leakage OR 2. SG tube RUPTURE.RCS Inadequate Heat Removal N/A N/A N/A Loss Not Applicable 2 RCS RCS Activity/CMNT Rad RCS Loss RM-G7 or RM-G18 CNTMT HI The value specified represents, based on Microshield calculations, Loss A. Containment radiation 3.A RNG Gamma > 100 R/hr the expected containment high range radiation monitor (RM-G7 3 monitor reading greater than and RM-G18) response based on a LOCA, one hour after monite-speading gealue). tshutdown with -0.1% fuel failure (rounded down to the nearest (site-specific value). whole number).RCS CNMT Integrity or Bypass N/A N/A N/A Loss Not Applicable 4 RCS Other Indications N/A N/A No other site-specific RCS Loss indication has been identified for Loss VCS1, A. (site-specific as applicable) 5 RCS Emergency Director Judgment RCS Loss Any condition in the opinion None Loss A. ANY condition in the opinion of the ED that indicates loss 64 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI VCS1 FPB FPB NEI IC Wording #(S) VCS1 FPB Wording Difference/Deviation Justification FPB# #(s)6 of the Emergency Director that 5.A of the RCS barrier indicates Loss of the RCS Barrier.RCS RCS or SG Tube Leakage RCS Operation of a standby None P-Loss 1 A. Operation of a standby P-Loss 1.A charging pump is required by charging (makeup) pump is EITHER: required by EITHER of the 9 UNISOLABLE RCS following:
leakage 1. UNISOLABLE RCS
- SG tube RUPTURE leakage OR RCS CSFST Integrity-RED path Consistent with the generic guidance provided in PWR FPB 2. SG tube leakage. P-Loss 11.B conditions met Developer Notes, VCS1 has opted to use CSFST conditions in lieu OR of specifying the parameters and values associated with this threshold.
B. RCS cooldown rate greater than (site-specific pressurized thermal shock criteria/limits defined by site-specific indications).
RCS Inadequate Heat Removal RCS CSFST Heat Sink-RED path Consistent with the generic guidance provided in PWR FPB P-Loss 2 A. Inadequate RCS heat P-Loss 2.A conditions met Developer Notes, VCS1 has opted to use CSFST conditions in lieu of specifying the parameters and values associated with this removal capability via steam AND threshold.
generators as indicated by Ha iki eurdtrsod (site-specific indications).
Heat sink is required The phrase "and heat sink required" has been added to preclude the need for classification for conditions in which RCS pressure is less than SG pressure or Heat Sink-RED path entry was created through operator action directed by an EOP.RCS CS Activity/CM NT Rad N/A N/A N/A P-Loss 3 Not Applicable 65 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI VCS1 FPB FPB NEI IC Wording #(s) VCS1 FPB Wording Difference/Deviation Justification FPB# #(s)RCS CNMT Integrity or Bypass N/A N/A N/A P-Loss 4 Not Applicable RCS Other Indications N/A N/A No other site-specific RCS Potential Loss indication has been P-Loss 5 A. (site-specific as applicable) identified for VCS1.RCS Emergency Director Judgment RCS Any condition in the opinion of None P-Loss 6 A. ANY condition in the opinion P-Loss 5.A the ED that indicates potential P-Loss 6loss of the RCS barrier of the Emergency Director that indicates Potential Loss of the RCS Barrier.66 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 PWR Containment Fission Product Barrier Degradation Thresholds NEI VCS1 NEI IC Wording FPB VCS1 FPB Wording Difference/Deviation Justification FPB# FPB #(s)CNMT RCS or SG Tube Leakage CNMT A leaking or RUPTURED SG is None Loss A. A leaking or RUPTURED SG is Loss FAULTED outside of containment 1 FAULTED outside of containment.
1 .A CNMT Inadequate Heat Removal N/A N/A N/A Loss Not Applicable 2 CNMT RCS Activity/CMNT Rad N/A N/A N/A Loss Not applicable 3 CNMT CNMT Integrity or Bypass CNMT Containment isolation is required None Loss Loss A. Containment isolation is required AND EITHER: 4ND4.A AND 4* Containment integrity has been EITHER of the following:
lost based on ED judgment 1. Containment integrity has been
- UNISOLABLE pathway from lost based on Emergency containment to the environment Director judgment.
exists OR CNMT Indications of RCS leakage outside of None 2. UNISOLABLE pathway from Loss containment the containment to the 4.B environment exists.OR B. Indications of RCS leakage outside of containment.
67 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI NEI IC Wording VCS1VCS1 FPB Wording Difference/Deviation Justification FPB# FPB #(s)CNMT Other Indications N/A N/A No other site-specific Containment Loss indication has Loss been identified for VCSI.A. (site-specific as applicable) 5 CNMT Emergency Director Judgment CNMT Any condition in the opinion of the ED None Loss Loss that indicates loss of the containment 6 ANY condition in the opinion of the barrier Emergency Director that indicates Loss 5.A of the Containment Barrier.CNMT RCS or SG Tube Leakage N/A N/A N/A P-Loss Not Applicable 1 CNMT Inadequate Heat Removal CNMT CSFST Core Cooling-RED path Consistent with the generic guidance provided in PWR P-Loss A. 1. (Site-specific criteria for entry P-Loss conditions met FPB Developer Notes, VCS1 has opted to use CSFST 2 intos cr cooling restoratiaforentry 2conditions in lieu of specifying the parameters and 2 into core cooling restoration 2.A AND values associated with this threshold.
procedure)
Restoration procedures not effective addedaoted c ite th therhrhst AND ithi 15min.(Not 1)Added Note 1 consistent with other thresholds with a AND within 15 min. (Note 1) timing component.
- 2. Restoration procedure not effective within 15 minutes.CNMT RCS Activity/CMNT Rad CNMT RM-G7 or RM-G18 CNTMT HI RNG The value specified represents, based on Microshield P-Loss A. Containment radiation monitor P-Loss Gamma > 20,000 R/hr calculations, the expected containment high range 3 reading greater than (site-specific 3.A radiation monitor (RM-G7 and RM-G18) response value), based on a LOCA, one hour after shutdown with -20%fuel failure (rounded down to nearest whole number for readability).
CNMT CNMT Integrity or Bypass CNMT CSFST Containment-RED path Consistent with the generic guidance provided in PWR P-Loss A. Containment pressure greater than P-Loss conditions met FPB Developer Notes, VCS1 has opted to use CSFST 4 (site-specific value) 4.A conditions in lieu of specifying the parameters and values associated with this threshold.
68 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI NEI IC Wording VOSi VCS1 FPB Wording Difference/Deviation Justification FPB# FPB #(s)OR CNMT Containment hydrogen concentration
> 4% hydrogen concentration in the presence of oxygen B. Explosive mixture exists inside P-Loss 4% represents an explosive mixture in containment.
containment 4.B OR C. 1. Containment pressure greater than (site-specific pressure setpoint)
CNMT Containment pressure > 12 psig The Containment pressure setpoint (12 psig) is the AND P-Loss AND pressure at which the Containment Spray System AND Lshould actuate and begin performing its function.2. Less than one full train of (site- 4.C specific system or equipment)
<one full train of depressurization Table F-2 provides the site-specific combinations of is operating per design for 15 equipment (Table F-2) operating per operating RBCU and Sprays that constitute a full train minutes or longer, design for .15 mi. (Note 1) containment cooling systems.Added Note 1 consistent with other thresholds with a timing component.
CNMT Other Indications N/A N/A No other site-specific Containment Potential Loss P-Loss A. (site-specific as applicable) indication has been identified for VCS1.5 CNMT Emergency Director Judgment CNMT Any condition in the opinion of the ED None P-Loss A. ANY condition in the opinion of the P-Loss that indicates potential loss of the 6 Emergency Director that indicates 5.A containment barrier Potential Loss of the Containment Barrier.69 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 Category H Hazards and Other Conditions Affecting Plant Safety 70 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification HU1 Confirmed SECURITY HU1 Confirmed SECURITY None CONDITION or threat CONDITION or threat.MODE: All MODE: All NEI Ex. NEI Example EAL Wording VCS1 VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #1 A SECURITY CONDITION that HU1.1 A SECURITY CONDITION that Example EALs #1,2 and 3 have been combined into a single EAL does not involve a HOSTILE does not involve a HOSTILE for ease of presentation and use.ACTION as reported by the (site- ACTION as reported by Security specific security shift supervision).
Team Leader 2 Notification of a credible security OR threat directed at the site. Notification of a credible security threat directed at the site 3 A validated notification from the OR NRC providing information of an aircraft threat. A validated notification from the NRC providing information of an aircraft threat 71 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification HU2 Seismic event greater than OBE HU2 Seismic event greater than OBE Made the term 'level' plural as the VCS1 OBE is defined by different level levels ground accelerations between the horizontal and vertical axis.MODE: All MODE: All NEI Ex. VCS1 EI NEI Example EAL Wording EAL VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #1 Seismic event greater than HU2.1 Seismic event > OBE as The bulleted conditions represent the site-specific indications of Operating Basis Earthquake indicated by EITHER: exceeding an OBE.(OBE) as indicated by:* Triaxial Seismic Switch MCB (site-specific indication that a annunciator XCP-638 3-5 seismic event met or exceeded (RB FOUND SEIS SWITCH OBE limits) OBE EXCEED)* Any red OBE light on the Triaxial Response Spectrum Recorder 72 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI IC# NEI IC Wording ICs) VCS1 IC Wording Difference/Deviation Justification IC#(s)HU3 Hazardous event. HU3 Hazardous event None MODE: All MODE: All NEI Ex. NEI Example EAL Wording EAL VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #1 A tornado strike within the HU3.1 A tornado strike within the None PROTECTED AREA. PROTECTED AREA 2 Internal room or area flooding of a HU3.2 Internal room or area FLOODING None magnitude sufficient to require of a magnitude sufficient to manual or automatic electrical require manual or automatic isolation of a SAFETY SYSTEM electrical isolation of a SAFETY component needed for the current SYSTEM component needed for operating mode. the current operating mode 3 Movement of personnel within the HU3.3 Movement of personnel within the None PROTECTED AREA is impeded PROTECTED AREA is impeded due to an offsite event involving due to an offsite event involving hazardous materials (e.g., an hazardous materials (e.g., an offsite chemical spill or toxic gas offsite chemical spill or toxic gas release).
release)4 A hazardous event that results in HU4.1 A hazardous event that results in Added reference to Note 10.on-site conditions sufficient to on-site conditions sufficient to prohibit the plant staff from prohibit the plant staff from accessing the site via personal accessing the site via personal vehicles.
vehicles (Note 10)5 (Site-specific list of natural or N/A N/A No other site-specific hazard has been identified for VCS1.technological hazard events)Note EAL #3 does not apply to routine N/A Note 10: This EAL does not This note, designated Note #10, is intended to apply to generic traffic impediments such as fog, apply to routine traffic example EAL #4, not #3 as specified in the generic guidance.73 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 impediments such as fog, snow, ice, or vehicle breakdowns or accidents.
74 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification HU4 FIRE potentially degrading the HU4 FIRE potentially degrading the None level of safety of the plant. level of safety of the plant MODE: All MODE: All NEI Ex. VCS1 EAL # NEI Example EAL Wording EAL # VCS1 EAL Wording Difference/Deviation Justification
- a. A FIRE is NOT extinguished HU4.1 A FIRE is NOT extinguished Table H-1 provides a tabularized list of site-specific plant rooms and within 15-minutes of ANY of the within 15 min. of any of the areas.following FIRE detection following FIRE detection indications:
indications (Note 1): " Report from the field (i.e., 9 Report from the field (i.e., visual observation) visual observation)" Receipt of multiple (more
- Receipt of multiple (more than 1) fire alarms or than 1) fire alarms or indications indications" Field verification of a single
- Field verification of a single fire alarm fire alarm AND AND b. The FIRE is located within The FIRE is located within any ANY of the following plant rooms Table H-1 area or areas: (site-specific list of plant rooms or areas)2 a. Receipt of a single fire alarm HU4.2 Receipt of a single fire alarm Table H-1 provides a tabularized list of site-specific plant rooms and (i.e., no other indications of a (i.e., no other indications of a areas.FIRE). FIRE)AND AND b. The FIRE is located within The fire alarm is indicating a 75 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 ANY of the following plant rooms FIRE within any Table H-1 area or areas: AND (site-specific list of plant rooms or The existence of a FIRE is not areas) verified within 30 min. of alarm AND receipt (Note 1)c. The existence of a FIRE is not verified within 30-minutes of alarm receipt.3 A FIRE within the plant or ISFSl HU4.3 A FIRE within the plant VCS1 does not have an ISFSI located outside the plant Protected[for plants with an ISFSl outside PROTECTED AREA not Area.the plant Protected Area] extinguished within 60 min. of the PROTECTED AREA not initial report, alarm or indication extinguished within 60-minutes of (Note 1)the initial report, alarm or indication.
4 A FIRE within the plant or ISFSI HU4.4 A FIRE within the plant None[for plants with an ISFSl outside PROTECTED AREA that the plant Protected Area] requires firefighting support by PROTECTED AREA that requires an offsite fire response agency to firefighting support by an offsite extinguish fire response agency to extinguish.
Note Note: The Emergency Director N/A Note 1: The Emergency Director The classification timeliness note has been standardized across the should declare the Unusual Event should declare the event VCS1 EAL scheme by referencing the "time limit" specified within promptly upon determining that promptly upon the EAL wording.the applicable time has been determining that time exceeded, or will likely be limit has been exceeded, exceeded.
or will likely be exceeded.76 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification HU7 Other conditions exist which in the HU7 Other conditions existing that in None judgment of the Emergency the judgment of the Emergency Director warrant declaration of a Director warrant declaration of a (NO)UE UE MODE: All MODE: All NEI Ex. VCS1 EA E NEI Example EAL Wording EAL VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #Other conditions exist which in the HU7.1 Other conditions exist which in None judgment of the Emergency the judgment of the Emergency Director indicate that events are in Director indicate that events are progress or have occurred which in progress or have occurred indicate a potential degradation of which indicate a potential the level of safety of the plant or degradation of the level of safety indicate a security threat to facility of the plant or indicate a security protection has been initiated.
No threat to facility protection has releases of radioactive material been initiated.
No releases of requiring offsite response or radioactive material requiring monitoring are expected unless offsite response or monitoring further degradation of safety are expected unless further systems occurs. degradation of SAFETY SYSTEMS occurs.77 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification HA1 HOSTILE ACTION within the HAl HOSTILE ACTION within the None OWNER CONTROLLED AREA or OWNER CONTROLLED AREA airborne attack threat within 30 or airborne attack threat within 30 minutes. minutes MODE: All MODE: All NEI Ex. VCS1 EA NEI Example EAL Wording EAL VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #1 A HOSTILE ACTION is occurring or HA1.1 A HOSTILE ACTION is Example EALs #1 and #2 have been combined into a single EAL has occurred within the OWNER occurring or has occurred within for ease of use.CONTROLLED AREA as reported the OWNER CONTROLLED The Security Team Leader is the site-specific security shift by the (site-specific security shift AREA as reported by the supervision.
supervision).
Security Team Leader 2 A validated notification from NRC of OR an aircraft attack threat within 30 A validated notification from minutes of the site. NRC of an aircraft attack threat within 30 min. of the site 78 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification HA5 Gaseous release impeding HA5 Gaseous release impeding None access to equipment necessary access to equipment necessary for normal plant operations, for normal plant operations, cooldown or shutdown.
cooldown or shutdown MODE: All MODE: All NEI Ex. NEI Example EAL Wording VCS1 VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #a. Release of a toxic, HA5.1 Release of a toxic, corrosive, The site-specific list of plant rooms or areas with entry-related mode corrosive, asphyxiant or asphyxiant or flammable gas into applicability are tabularized in Table H-2.flammable gas into any of the any Table H-2 area Table H-2 lists plant areas, not specific rooms. Therefore the word following plant rooms or areas: AND "rooms" has been deleted.(site-specific list of plant rooms Entry into the area is prohibited or or areas with entry-related mode impeded (Note 6)applicability identified)
AND b. Entry into the room or area is prohibited or impeded.Note Note: If the equipment in the N/A Note 6: If the equipment in the listed Table H-2 lists plant areas, not specific rooms. Therefore the word listed room or area was already area was already inoperable "rooms" has been deleted.inoperable or out-of-service or out-of-service before the before the event occurred, then event occurred, then no no emergency classification is emergency classification is warranted.
warranted.
79 of 114 EAL Comparison -Matrix OSSI Project #12-0202 VCS1 Table H-2 Safe Operation
& Shutdown Areas Area Mode Applicability Auxiliary Building 374' 3 Auxiliary Building 388' 3, 4, 5 Auxiliary Building 400' 4, 5 Auxiliary Building 412 3, 4, 5 Auxiliary Building 436' 1, 2, 3, 4, 5 Auxiliary Building 463' 3, 4, 5 Intermediate Building 412' 3 Intermediate Building 436' 4, 5 Intermediate Building 463' 3, 4, 5 Control Building 412' 2, 3 Control Building 436' 3, 4, 5 Turbine Building (All levels) 1,2 80 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS 1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification HA6 Control Room evacuation HA6 Control Room evacuation None resulting in transfer of plant resulting in transfer of plant control to alternate locations, control to alternate locations MODE: All MODE: All NEI Ex. VCS1 NEI Example EAL Wording EAL VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #1 An event has resulted in plant HA6.1 An event has resulted in plant CREP are the site-specific remote shutdown panels/local control control being transferred from the control being transferred from the stations.Control Room to (site-specific Control Room to the Control remote shutdown panels and Room Evacuation Panels (CREP)local control stations).
81 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification HA7 Other conditions exist which in the HA7 Other conditions exist that in the None judgment of the Emergency Director judgment of the Emergency Director warrant declaration of an Alert. warrant declaration of an Alert MODE: All MODE: All NEI Ex. VCS1 EA E NEI Example EAL Wording EAL VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #Other conditions exist which, in the HA7.1 Other conditions exist which, in the None judgment of the Emergency Director, judgment of the Emergency Director, indicate that events are in progress or indicate that events are in progress or have occurred which involve an actual or have occurred which involve an actual or potential substantial degradation of the potential substantial degradation of the level of safety of the plant or a security level of safety of the plant or a security event that involves probable life event that involves probable life threatening risk to site personnel or threatening risk to site personnel or damage to site equipment because of damage to site equipment because of HOSTILE ACTION. Any releases are HOSTILE ACTION. Any releases are expected to be limited to small fractions expected to be limited to small fractions of the EPA Protective Action Guideline of the EPA Protective Action Guideline exposure levels. exposure levels.82 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification HS1 HOSTILE ACTION within the HS1 HOSTILE ACTION within the None PROTECTED AREA PROTECTED AREA MODE: All MODE: All NEI Ex. VCS1 EAL # NEI Example EAL Wording EAL # VCS1 EAL Wording Difference/Deviation Justification 1 A HOSTILE ACTION is occurring HS1.1 A HOSTILE ACTION is occurring or The Security Team Leader is the site-specific security shift or has occurred within the has occurred within the supervision.
PROTECTED AREA as reported PROTECTED AREA as reported by by the (site-specific security shift the Security Team Leader supervision).
83 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VoS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification HS6 Inability to control a key safety HS6 Inability to control a key safety function None function from outside the Control from outside the Control Room Room.MODE: All MODE: All NEI Ex. VCS1 EAL # NEI Example EAL Wording EAL # VCS1 EAL Wording Difference/Deviation Justification
- a. An event has resulted in plant HS6.1 An event has resulted in plant control CREP are the site-specific remote shutdown panels/local control being transferred from the being transferred from the Control Room control stations.Control Room to (site-specific to the Control Room Evacuation Panels remote shutdown panels and local (CREP)control stations).
AND.AND Control of any of the following key safety b. Control of ANY of the functions is not reestablished within 15 following key safety functions is min. (Note 1): not reestablished within (site-
- Reactivity control specific number of minutes)." Reactivity control
- Core cooling* Core cooling [PWR] / RPV 0 RCS heat removal water level [BWR]* RCS heat removal 84 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification HS7 Other conditions exist which in HS7 Other conditions existing that in the None the judgment of the Emergency judgment of the Emergency Director Director warrant declaration of a warrant declaration of a Site Area Site Area Emergency.
Emergency MODE: All MODE: All NEI Ex. NEI Example EAL Wording VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #Other conditions exist which in HS7.1 Other conditions exist which in the None the judgment of the Emergency judgment of the Emergency Director Director indicate that events are indicate that events are in progress or in progress or have occurred have occurred which involve actual or which involve actual or likely likely major failures of plant functions major failures of plant functions needed for protection of the public or needed for protection of the HOSTILE ACTION that results in public or HOSTILE ACTION that intentional damage or malicious acts, (1)results in intentional damage or toward site personnel or equipment that malicious acts, (1) toward site could lead to the likely failure of or, (2) that personnel or equipment that prevent effective access to equipment could lead to the likely failure of needed for the protection of the public.or, (2) that prevent effective Any releases are not expected to result in access to equipment needed for exposure levels which exceed EPA the protection of the public. Any Protective Action Guideline exposure releases are not expected to levels beyond the SITE BOUNDARY.result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.85 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI IC# NEI IC Wording ICs) VCS1 IC Wording Difference/Deviation Justification IC#(s)HG1 HOSTILE ACTION resulting in HG1 HOSTILE ACTION resulting in loss of None loss of physical control of the physical control of the facility facility.
MODE: All MODE: All NEI Ex. NEI Example EAL Wording VCS1 VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #1 a. A HOSTILE ACTION is HG1.1 A HOSTILE ACTION is occurring or has The Security Team Leader is the site-specific security shift occurring or has occurred within occurred within the PROTECTED AREA supervision.
the PROTECTED AREA as as reported by the Security Team Leader reported by the (site-specific AND EITHER of the following has security shift supervision).
occurred: AND Any of the following safety functions b. EITHER of the following has cannot be controlled or maintained occurred:
- Reactivity control 1. ANY of the following safety
- Core cooling functions cannot be controlled or maintained.
- RCS heat removal* Reactivity control OR* Core cooling Damage to spent fuel has occurred[PWR]/RPV water or is IMMINENT level [BWR]* RCS heat removal OR 2. Damage to spent fuel has occurred or is IMMINENT.86 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification HG7 Other conditions exist which in HG7 Other conditions exist which in the None the judgment of the Emergency judgment of the Emergency Director Director warrant declaration of a warrant declaration of a General General Emergency Emergency MODE: All MODE: All NEI Ex. NEI Example EAL Wording VCS1 VCS1 EAL Wording Difference/Deviation Justification EAL# EAL C EAL #Other conditions exist which in HG7.1 Other conditions exist which in the None the judgment of the Emergency judgment of the Emergency Director Director indicate that events are indicate that events are in progress or in progress or have occurred have occurred which involve actual or which involve actual or IMMINENT substantial core degradation IMMINENT substantial core or melting with potential for loss of degradation or melting with containment integrity or HOSTILE potential for loss of containment ACTION that results in an actual loss of integrity or HOSTILE ACTION physical control of the facility.
Releases that results in an actual loss of can be reasonably expected to exceed physical control of the facility.
EPA Protective Action Guideline Releases can be reasonably exposure levels offsite for more than the expected to exceed EPA immediate site area.Protective Action Guideline exposure levels offsite for more than the immediate site area.87 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 Category S System Malfunction 88 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI IC# NEI IC Wording ICs) VCS1 IC Wording Difference/Deviation Justification IC#(S)SUl Loss of all offsite AC power SUl Loss of all offsite AC power ESF is the site-specific designation for the VCS1 emergency AC capability to emergency buses for capability to ESF buses for 15 buses.15 minutes or longer. minutes or longer MODE: Power Operation, Startup, MODE: 1 -Power Operation, 2 -Hot Standby, Hot Shutdown Startup, 3 -Hot Standby, 4 -Hot Shutdown NEI Ex. NIEapeELWrig VCS1 EAL # NEI Example EAL # VCS1 EAL Wording Difference/Deviation Justification 1 Loss of ALL offsite AC power SU1.1 Loss of all offsite AC power 7.2 KV ESF buses 1DA and 1DB are the site-specific emergency capability to (site-specific (Table S-1) capability to 7.2 KV buses.emergency buses) for 15 minutes ESF buses 1 DA and 1 DB for > 15 Site-specific AC power sources are tabularized in Table S-1.or longer. min. (Note 1)Note The Emergency Director should N/A Note 1: The Emergency Director The classification timeliness note has been standardized across the declare the Unusual Event should declare the event VCS1 EAL scheme by referencing the "time limit" specified within the promptly upon determining that 15 promptly upon EAL wording.minutes has been exceeded, or determining that time will likely be exceeded.
limit has been exceeded, or will likely be exceeded.89 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI IC# NEI IC Wording ICs) VCS1 IC Wording Difference/Deviation Justification IC#(s)SU2 UNPLANNED loss of Control SU3 UNPLANNED loss of Control None Room indications for 15 minutes Room indications for 15 minutes or longer, or longer.MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Standby, 4 -Hot Shutdown Shutdown NEI Ex. NIEapeELWrig VCS1 ENEI Example EAL Wording VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #1 An UNPLANNED event results in SU3.1 An UNPLANNED event results in The site-specific Safety System Parameter list to tabulated in Table the inability to monitor one or the inability to monitor one or S-2.more of the following parameters more Table S-2 parameters from RCS level revised to read "Reactor vessel/pressurizer level". No from within the Control Room for within the Control Room for > 15 change in intent.15 minutes or longer. min. (Note 1)Note The Emergency Director should N/A Note 1: The Emergency Director The classification timeliness note has been standardized across the declare the Unusual Event should declare the event VCS1 EAL scheme by referencing the "time limit" specified within the promptly upon determining that promptly upon EAL wording.15 minutes has been exceeded, determining that time or will likely be exceeded.
limit has been exceeded, or will likely be exceeded.90 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1[BWR parameter list] [PWR parameter list]Reactor Power Reactor Power RPV Water Level RCS Level RPV Pressure RCS Pressure Primary Containment Pressure In-Core/Core Exit Temperature Suppression Pool Level Levels in at least (site-specific number)steam generators Suppression Pool Temperature Steam Generator Auxiliary or Emergency Feed Water Flow Table S-2 Safety System Parameters
- Reactor power* Reactor vessel/pressurizer level* RCS pressure* Core Exit TCs* Level in at least one SG* EFW/AFW flow 91 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification SU3 Reactor coolant activity greater SU4 Reactor coolant activity greater None than Technical Specification than Technical Specification allowable limits, allowable limits MODE: Power Operation, Startup, MODE: 1 -Power Operation, 2 -Hot Standby, Hot Shutdown Startup, 3 -Hot Standby, 4 -Hot Shutdown NEI Ex. VCS1 EAL # NEI Example EAL Wording EAL # VCS1 EAL Wording Difference/Deviation Justification 1 (Site-specific radiation monitor) SU4.1 With letdown in service, RM-L1 The specified EAL threshold setpoint was calculated using RCS reading greater than (site-specific high range monitor > 39,000 cpm activities given in Table 11.1-2 of the FSAR and included all activities value), in the table scaled to 1.0 pCi/gm dose equivalent iodine.2 Sample analysis indicates that a SU4.2 Sample analysis indicates that a Changed 'reactor coolant activity" to "primary coolant activity" to reactor coolant activity value is primary coolant activity value is > conform to site specific terminology.
greater than an allowable limit an allowable limit specified in specified in Technical Technical Specifications 3/4.4.8 Specifications.
92 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification SU4 RCS leakage for 15 minutes or SU5 RCS leakage for 15 minutes or None longer. longer MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Standby, 4 -Hot Shutdown Shutdown NEI Ex. VCS1 EI # NEI Example EAL Wording EAL VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #1 RCS unidentified or pressure SU5.1 RCS unidentified or pressure Example EALs #1, 2 and 3 have been combined into a single EAL boundary leakage greater than boundary leakage> 10 gpm for > for uasbility.(site-specific value) for 15 15 min.minutes or longer. OR 2 RCS identified leakage greater RCS identified leakage > 25 gpm than (site-specific value) for 15 for > 15 min.minutes or longer. OR 3 Leakage from the RCS to a Leakage from the RCS to a location outside containment location outside containment
> 25 greater than 25 gpm for 15 gpm for > 15 min.minutes or longer. (Note 1)Note The Emergency Director should N/A Note 1: The Emergency Director The classification timeliness note has been standardized across the declare the Unusual Event should declare the event VCS1 EAL scheme by referencing the "time limit" specified within promptly upon determining that promptly upon the EAL wording.15 minutes has been exceeded, determining that time or will likely be exceeded.
limit has been exceeded, or will likely be exceeded.93 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification SU5 Automatic or manual (trip SU6 Automatic or manual trip fails to None[PWR] / scram [BWR]) fails to shut down the reactor shutdown the reactor. MODE: 1 -Power Operation MODE: Power Operation NEI Ex. VCS1 EAL # NEI Example EAL Wording EAL # VCS1 EAL Wording Difference/Deviation Justification 1 a. An automatic (trip [PWR] / SU6.1 An automatic trip did not shut Added the word "...after any RTS setpoint is exceeded" to clarify that scram [BWR]) did not shutdown down the reactor after any RTS it is a failure of the automatic trip when a valid trip signal has been the reactor. setpoint is exceeded exceed.AND AND Reactor power < 5% is the site-specific indication of a successful
- b. A subsequent manual action A subsequent manual action reactor trip.taken at the reactor control taken at the reactor control consoles is successful in consoles is successful in shutting shutting down the reactor. down the reactor as indicated by reactor power < 5% (Note 8).2 a. A manual trip ([PWR] / SU6.2 A manual trip did not shut down Added the word "... after any manual trip action was initiated" to scram [BWR]) did not shutdown the reactor after any manual trip clarify that it is a failure of any manual trip when an actual manual the reactor. action was initiated trip signal has been inserted.AND AND Combined conditions b.1 and b.2 into a single statement to simplify b. EITHER of the following:
A subsequent automatic trip or the presentation.
manual trip action taken at the Reactor power < 5% is the site-specific indication of a successful 1.tio tasu enat mheeanl reactor control consoles is reactor trip.acotion takens hes r r successful in shutting down the sucntroul cnsoletisg reactor as indicated by reactor successful in shutting power < 5% (Note 8).down the reactor.OR 2 A subsequent automatic (trip [PWR] / scram 94 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1[BWR]) is successful in shutting down the reactor.Notes Note: A manual action is any N/A Note 8: A manual action is any None operator action, or set of actions, operator action, or set of which causes the control rods to actions, which causes be rapidly inserted into the core, the control rods to be and does not include manually rapidly inserted into the driving in control rods or implementation of boron include manually driving injection strategies.
in control rods or implementation of boron injection strategies.
95 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI IC# NEI IC Wording ICs) VCS1 IC Wording Difference/Deviation Justification IC#(S)SU6 Loss of all onsite or offsite SU7 Loss of all onsite or offsite None communications capabilities, communications capabilities.
MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Standby, 4 -Hot Shutdown Shutdown NEI Ex. NIEapeELWrig VCS1 EAL # NEI Example# EAL # VCS1 EAL Wording Difference/Deviation Justification 1 Loss of ALL of the following SU7.1 Loss of all Table S-3 onsite Example EALs #1, 2 and 3 have been combined into a single EAL onsite communication methods: communication methods for simplification of presentation.(site-specific list of OR Example EAL condition
- 2 and #3 have been combined into a single communications methods) Loss of all Table S-30RO/NRC statement because the list of ORO and NRC communications
_______Los of ll abl S-3ORONRC methods are the same.2 Loss of ALL of the following communication methods ORO communications methods: Table S-3 provides a site-specific list of onsite and ORO/NRC communications methods.(site-specific list of communications methods)3 Loss of ALL of the following NRC communications methods: (site-specific list of communications methods)96 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 Table S-3 Communication Methods System Onsite ORO/NRC Gai-Tronics system X Radio system X Internal Telephone system X Maintenance jack system X Telephone land lines X X Fiberoptic links X Satellite phone system X Federal Telephone System (ENS) X ESSX X 97 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI IC# NEI IC Wording ICs) VCS1 IC Wording Difference/Deviation Justification IC#(s)SU7 Failure to isolate containment or SU8 Failure to isolate containment or None loss of containment pressure loss of containment pressure control. [PWR] control MODE: Hot Standby, Hot MODE: 3 -Hot Standby, 4 -Hot Shutdown Shutdown NEI Ex. NEI Example EAL Wording VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #1 a. Failure of containment to SU8.1 Containment isolation actuated Reworded EAL to better describe the intent. Penetrations cannot isolate when required by an close, but they can be isolated by closure of one or more isolation actuation signal. AND valves associated with that penetration.
The revised wording AND At least one isolation valve in maintains the generic example EAL intent while more clearly each penetration is not closed describing failure to isolate threshold.
- b. ALL required penetrations within 15 min. of the actuation are not closed within 15 minutes (Note 1)of the actuation signal.2 a. Containment pressure SU8.2 Containment pressure > 12 psig The Containment pressure setpoint (12 psig) is the pressure at greater than (site-specific AND which the Containment Spray System should actuate and begin pressure).
performing its function.AND < one full train of Table S-4 provides the site-specific combinations of operating depressurization equipment RBCU and Sprays that constitute a full train containment cooling b. Less than one full train of (Table S-4) is operating per systems.(site-specific system or designfor
> 15 min.equipment) is operating per (Note 1)design for 15 minutes or longer.N/A N/A N/A Note 1: The Emergency Director Added Note 1 to be consistent in its use for EAL thresholds with a should declare the event timing component.
promptly upon determining that time limit 98 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 has been exceeded, or will likely be exceeded.Table S-4 Full Train Depressurization Equipment RBCU Groups Containment Sprays Operating Operating 2 0 1 1 0 2 99 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI IC# NEI IC Wording ICs) VCS1 IC Wording Difference/Deviation Justification IC#(S)SA1 Loss of all but one AC power SA1 Loss of all but one AC power ESF is the site-specific designation for the VCS1 emergency AC source to emergency buses for source to ESF buses for 15 buses.15 minutes or longer. minutes or longer.MODE: Power Operation, Startup MODE: 1 -Power Operation NEI Ex. VCS1 EA NEI Example EAL Wording EAL VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #1 a. AC power capability to (site- SAI.1 AC power capability to 7.2 KV 7.2 KV ESF buses 1 DA and 1 DB are the site-specific emergency specific emergency buses) is ESF buses 1 DA and 1 DB buses.reduced to a single power source reduced to a single power source Site-specific AC power sources are tabularized in Table S-1.for 15 minutes or longer. (Table S-1) for a 15 min. (Note 1)AND AND b. Any additional single power Any additional single power source failure will result in a loss source failure will result in loss of of all AC power to SAFETY all AC power to SAFETY SYSTEMS. SYSTEMS Note The Emergency Director should N/A Note 1: The Emergency Director The classification timeliness note has been standardized across the declare the Alert promptly upon should declare the VCS1 EAL scheme by referencing the "time limit" specified within the determining that 15 minutes has event promptly upon EAL wording.been exceeded, or will likely be determining that time exceeded.
limit has been exceeded, or will likely be exceeded.100 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 Table S-1 AC Power Supplies Offsite:* 115 KV power to XTF-4 and XTF-5* 230 KV power to XTF-31 0 Parr Hydro Plant 13.8 KV power to ESF bus 1DA or 1 DB Onsite: 0 Diesel Generator A e Diesel Generator B 101 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording lC#(s) VCS1 IC Wording Difference/Deviation Justification SA2 UNPLANNED loss of Control SA3 UNPLANNED loss of Control None Room indications for 15 minutes Room indications for 15 minutes or longer with a significant or longer.transient in progress.
MODE: 1 -Power Operation, 2 -MODE: Power Operation, Startup, 3 -Hot Standby, 4 -Hot Startup, Hot Standby, Hot Shutdown Shutdown NEI Ex. NVCxapeSA1orig A ENEI Example EAL Wording VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #An UNPLANNED event results in SA3.1 An UNPLANNED event results in The site-specific Safety System Parameter list to tabulated in Table the inability to monitor one or the inability to monitor one or S-2.more of the following parameters more Table S-2 parameters from from within the Control Room for within the Control Room for > 15 15 minutes or longer. min. (Note 1)AND AND ANY of the following transient Any of the following transient events in progress.
events in progress: " Automatic or manual
- Automatic or manual runback greater than 25% runback greater than 25%thermal reactor power thermal reactor power* Electrical load rejection 9 Electrical load rejection greater than 25% full greater than 25% full electrical load electrical load" Reactor scram [BWR] / trip
- Reactor trip[PWR]
- ECCS actuation* ECCS (SI) actuation* Thermal power oscillations greater than 10% [BWF]102 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 Note The Emergency Director should N/A Note 1: The Emergency Director The classification timeliness note has been standardized across the declare the Unusual Event should declare the event VCS1 EAL scheme by referencing the "time limit" specified within promptly upon determining that promptly upon the EAL wording.15 minutes has been exceeded, determining that time or will likely be exceeded.
limit has been exceeded, or will likely be exceeded.[BWR parameter list] [PWR parameter list]Reactor Power Reactor Power RPV Water Level RCS Level RPV Pressure RCS Pressure Primary Containment Pressure In-Core/Core Exit Temperature Suppression Pool Level Levels in at least (site-specific number)steam generators Suppression Pool Temperature Steam Generator Auxiliary or Emergency Feed Water Flow Table S-2 Safety System Parameters
- Reactor power* Reactor vessel level* RCS pressure* Core Exit TCs* Level in at least one SG* EFW/AFW flow 103 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification SA5 Automatic or manual (trip [PWR] SA6 Automatic or manual trip fails to None/ scram [BWR]) fails to shutdown shut down the reactor and the reactor, and subsequent subsequent manual actions manual actions taken at the taken at the reactor control reactor control consoles are not consoles are not successful in successful in shutting down the shutting down the reactor reactor. MODE: 1 -Power Operation MODE: Power Operation NEI Ex. VCS1 EA NEI Example EAL Wording EAL VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #a. An automatic or manual (trip SA6.1 An automatic or manual trip fails Reactor power < 5% is the site-specific indication of a successful
[PWR] / scram [BWR]) did not to shut down the reactor reactor trip.shutdown the reactor. AND AND Manual actions taken at the b. Manual actions taken at the reactor control console are not reactor control consoles are not successful in shutting down the successful in shutting down the reactor as indicated by reactor reactor. power > 5% (Note 8)Notes Note: A manual action is any N/A Note 8: A manual action is any None operator action, or set of actions, operator action, or set of which causes the control rods to actions, which causes be rapidly inserted into the core, the control rods to be and does not include manually rapidly inserted into the driving in control rods or implementation of boron injection include manually driving strategies.
nld aual rvn in control rods or implementation of boron injection strategies.
104 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 105 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification SA9 Hazardous event affecting a HA2 Seismic event affecting a Generic ICs CA6, SA9 and associated example EALs have been re-SAFETY SYSTEM needed for SAFETY SYSTEM needed for assigned to the associated event hazards sub-categories for human the current operating mode. the current operating mode. factors considerations.
Although ICs CA6 and SA9 are generically MODE: Power Operation, MODE: All assigned symptomatically to the system malfunctions related Startup, Hot Standby, Hot categories, it is the initiating events such as seismic activity, natural Shutdown HA3 Hazardous event affecting a or technological events or fires/explosions that trigger assessment of SAFETY SYSTEM needed for safety system degradation associated with these EALs. This re-the current operating mode. assignment within the VCS1 EAL scheme results in no change of intent.MODE: All Mode applicability has been expanded to ALL modes based on the HA4 FIRE or EXPLOSION event consolidation of ICs CA6 and SA9 and to address Spent Fuel Pool affecting a SAFETY SYSTEM system degradation during full core offload.needed for the current operating mode.MODE: All 106 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI Ex. VCS1 EA E NEI Example EAL Wording EAL VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #1 a. The occurrence of ANY of the following hazardous events: " Seismic event (earthquake)
- Internal or external flooding event" High winds or tornado strike" FIRE" EXPLOSION" (site-specific hazards)* Other events with similar hazard characteristics as determined by the Shift Manager AND b. EITHER of the following:
- 1. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.OR 2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.HA2.1 Seismic event resulting in EITHER of the following:
- Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode* The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode This EAL represents the seismic event portion of the generic CA6 and SA9 example EAL #1 HA3.1 The occurrence of any Table C-5 hazardous event AND EITHER of the following:
- Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode* The event has caused VISIBLE DAMAGE to a This EAL represents the non-seismic naturally occurring and 'other'event portion of the generic CA6 and SA9 example EAL #1 The list of hazardous events has been tabularized in Table H-3 No other VCS1 -specific unique hazards have been identified for inclusion of the EAL.I 107 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 SAFETY SYSTEM component or structure needed for the current operating mode HA4.1 FIRE or EXPLOSION resulting in EITHER of the following:
- Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode* The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operatina mode This EAL represents the fire and explosion portion of the generic CA6 and SA9 example EAL #1 Table H-3 Hazardous Events* Internal or external FLOODING event* High winds or tornado strike* Other events with similar hazard characteristics as determined by the Shift Supervisor 108 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification SS1 Loss of all offsite and all onsite SS1 Loss of all offsite and all onsite ESF is the site-specific designation for the VCS1 emergency AC AC power to emergency buses AC power to ESF buses for 15 buses.for 15 minutes or longer. minutes or longer MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Standby, 4 -Hot Shutdown Shutdown NEI Ex. VCS1 EAL # NEI Example EAL Wording EAL # VCS1 EAL Wording Difference/Deviation Justification 1 Loss of ALL offsite and ALL SS1.1 Loss of all offsite and all onsite 7.2 KV ESF buses 1 DA and 1DB are the site-specific emergency onsite AC power to (site-specific AC power (Table S-1) capability buses.emergency buses) for 15 minutes to 7.2 KV ESF buses 1 DA and Site-specific AC power sources are tabularized in Table S-1.or longer. 1DB for ? 15 min. (Note 1)Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across the declare the Unusual Event Director should declare VCS1 EAL scheme by referencing the "time limit" specified within the promptly upon determining that the event promptly EAL wording.15 minutes has been exceeded, upon determining that or will likely be exceeded.
time limit has been exceeded, or will likely be exceeded.109 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 NEI IC# NEI IC Wording ICs) VCS1 IC Wording Difference/Deviation Justification IC#(S)SS5 Inability to shutdown the reactor SS6 Inability to shut down the None causing a challenge to (core reactor causing a challenge to cooling [PWR] / RPV water level core cooling or RCS heat[BWR]) or RCS heat removal, removal MODE: Power Operation MODE: 1 -Power Operation NEI Ex. NEI Example EAL Wording EAL VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #1 a. An automatic or manual (trip SS6.1 An automatic or manual trip fails Reactor power < 5% is the site-specific indication of a successful
[PWR] / scram [BWR]) did not to shutdown the reactor reactor trip.shutdown the reactor. AND Indication that core cooling is extremely challenged is manifested by AND All manual actions to shut down entry to Critical Safety Function Status Tree (CSFST) Core Cooling-b. All manual actions to the reactor are not successful in RED path.shutdown the reactor have been shutting down the reactor as Indication that heat removal is extremely challenged is manifested unsuccessful.
indicated by reactor power > 5% by entry to CSFST Heat Sink-RED path.AND AND c. EITHER of the following EITHER of the following conditions exist: conditions exist:* (Site-specific indication of
- CSFST Core Cooling-RED an inability to adequately path conditions met remove heat from the core)
- CSFST Heat Sink-RED path conditions met* (Site-specific indication of an inability to adequately remove heat from the RCS)110 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification SS8 Loss of all Vital DC power for 15 SS2 Loss of all vital DC power for 15 None minutes or longer. minutes or longer.MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Standby, 4 -Hot Shutdown Shutdown NEI Ex. VCS1 EAL # NEI Example EAL Wording EAL # VCS1 EAL Wording Difference/Deviation Justification 1 Indicated voltage is less than SS2.1 < 108 VDC on both Train A and 108 VDC is the site-specific minimum vital DC bus voltage.(site-specific bus voltage value) Train B vital 125 VDC systems Train A or Train B vital 125 VDC system are the site-specific vital DC on ALL (site-specific Vital DC for > 15 min. (Note 1) buses.busses) for 15 minutes or longer.Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across the declare the Unusual Event Director should declare the VCS1 EAL scheme by referencing the "time limit" specified within the promptly upon determining that event promptly upon EAL wording.15 minutes has been exceeded, determining that time limit has or will likely be exceeded.
been exceeded, or will likely be exceeded.111 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCSl NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification SG1 Prolonged loss of all offsite and SG1 Prolonged loss of all offsite and ESF is the site-specific designation for the VCS1 emergency AC all onsite AC power to all onsite AC power to ESF buses.emergency buses. buses or loss of all AC and vital Combined ICs SG1 and SG8 under the loss of power category for MODE: Power Operation, DC power sources for 15 usability.
Startup, Hot Standby, Hot minutes or longer Shutdown MODE: 1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown NEI Ex. VCS1 EAL # NEI Example EAL Wording EAL # VCS1 EAL Wording Difference/Deviation Justification 1 a. Loss of ALL offsite and ALL SG1.1 Loss of all offsite and all onsite 7.2 KV ESF buses 1 DA and 1DB are the site-specific emergency onsite AC power to (site-specific AC power capability to 7.2 KV buses.emergency buses). ESF buses 1 DA and 11DB (Table Site-specific AC power sources are tabularized in Table S-1.AND 5-1) 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is the site-specific SBO coping analysis time.b. EITHER of the following:
AND CSFST Core Cooling-RED path conditions being met indicates* Restoration of at least EITHER of the following:
significant core exit superheating and core uncovery.one AC emergency bus in
- Restoration of at least one less than (site-specific ESF bus within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is hours) is not likely, not likely (Note 1)* (Site-specific indication of
- CSFST Core Cooling-RED an inability to adequately path conditions met remove heat from the core)Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across the declare the General Emergency Director should declare VCS1 EAL scheme by referencing the "time limit" specified within promptly upon determining that the event promptly the EAL wording.(site-specific hours) has been upon determining that exceeded, or will likely be time limit has been exceeded.
exceeded, or will likely be exceeded.112 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 113 of 114 EAL Comparison Matrix OSSI Project #12-0202 VCS1 VCS1 NEI IC# NEI IC Wording IC#(s) VCS1 IC Wording Difference/Deviation Justification SG8 Loss of all AC and Vital DC SG1 Prolonged loss of all offsite and ESF is the site-specific designation for the VCS1 emergency AC power sources for 15 minutes or all onsite AC power to ESF buses.longer. buses or loss of all AC and vital Combined ICs SG1 and SG8 under the loss of power category for MODE: Power Operation, DC power sources for 15 usability.
Startup, Hot Standby, Hot minutes or longer Shutdown MODE: 1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown NEI Ex. VCS1 EA NEI Example EAL Wording EAL VCS1 EAL Wording Difference/Deviation Justification EAL # EAL #1 a. Loss of ALL offsite and ALL SG1.2 Loss of all offsite and all onsite 7.2 KV ESF buses 1 DA and 1DB are the site-specific emergency onsite AC power to (site-specific AC power (Table S-1) capability buses.emergency buses) for 15 minutes to 7.2 KV ESF buses 1 DA and Site-specific AC power sources are tabularized in Table S-1.or longer. 1DB for > 15 min.108 VDC is the site-specific minimum vital DC bus voltage.AND AND Train A or Train B vital 125 VDC system are the site-specific vital DC b. Indicated voltage is less than < 108 VDC on both Train A and buses.(site-specific bus voltage value) Train B vital 125 VDC systems on ALL (site-specific Vital DC for > 15 min.busses) for 15 minutes or longer. (Note 1)Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across the declare the Unusual Event Director should declare the VCS1 EAL scheme by referencing the "time limit" specified within the promptly upon determining that 15 event promptly upon EAL wording.minutes has been exceeded, or determining that time limit has will likely be exceeded.
been exceeded, or will likely be exceeded.114 of 114 Document Control Desk Attachment V LAR-14-02392 RC-14-0032 Page 1 of 4 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) Unit I DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 ATTACHMENT V V.C. Summer Unit I NEI 99-01 Revision 6 EAL Wallcharts Information Only
Document Control Desk Attachment VI LAR-14-02392 RC-14-0032 Page 1 of 1 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT I DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 ATTACHMENT VI LIST OF REGULATORY COMMITMENTS The following table identifies those actions committed to by SCE&G, Virgil C. Summer Nuclear Station in this document and References.
Please direct questions regarding these commitments to Mr. Bruce L. Thompson, Manager, Nuclear Licensing, (803) 931-5042.COMMITMENT Due Date/Event V. C. Summer Unit 1 expects to comply with the NRC Order November 2015/ Prior to implementation date for Order EA-12-051, Modifying start up from Refueling Licenses with Regard to Requirements for Reliable Spent Outage 22 Fuel Pool Instrumentation.