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. Attachment Ii: Technical Bases Document for the Proposed EALs (Marked-Up Copy)
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Document Control Desk Attachment II LAR-14-02392 RC-14-0032 Page 1 of 360 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) Unit I DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 ATTACHMENT II Technical Bases Document for the Proposed VCSNS EALs (Marked-Up 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 1.0 2.0 3.0 TITLE PAGE P U R P O S E ..............................................................................................
..1 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 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 R E FE R E N C ES .......................................................................................
15 4.1 D evelopm ental ..............................................................................
15 4.2 Im plem enting .................................................................................
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 Category S System Malfunction
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197 Category F Fission Product Barrier Degradation
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259 C ateg o ry 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 1 of 359 EPP-1 08 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 2 of 359 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 Emerqency:
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 3 of 359 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 4 of 359 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 5 of 359 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, For 1)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 6 of 359 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 7 of 359 EPP-1 08 ENCLOSURE I REVISION 01 [DRAFT E]2 Startup Keff >- 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 > 200°F 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 8 of 359 EPP-1 08 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 9 of 359 EPP-1 08 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 10 of 359 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 11 of 359 EPP-1 08 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 12 of 359 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 13 of 359 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-1 022 (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 14 of 359 EPP-1 08 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 15 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]4.2.3 VCSNS EAL Matrix Page 16 of 359 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 17 of 359 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 18 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT El 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 19 of 359 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 20 of 359 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 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;(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 21 of 359 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 22 of 359 EPP-1 08 ENCLOSURE I REVISION 01 [DRAFT E]5.2 Acronyms & Abbreviations OF .......................................................................................................
Degrees Fahrenheit
..................................................................................................................
Feet or M inutes.. ..........................................................................................................
Inches or Seconds r ...........................................................................................................................
G a m m a r. ............................................................................................................................
N e u tro n AB ..........................................................................................................
Auxiliary Building AC ........................................................................................................
Alternating Current AOP .................................................................................
Abnormal 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 ............................................................................
Emergency Core Cooling System ECL ..................................................................................
Emergency Classification Level EOF ..................................................................................
Emergency Operations Facility EOP ..............................................................................
Emergency Operating Procedure EPA ...............................................................................
Environmental Protection Agency EPIP .................................................................
Emergency Plan Im plem enting Procedure EPRI .............................................................................
Electric Power Research Institute ERG ...............................................................................
Em ergency Response Guideline ESF .................................................................................
Engineered Safeguards Feature FEMA ...............................................................
Federal Em ergency Management Agency FSAR ....................................................................................
Final Safety Analysis Report HSI ............................................................................................
Human System Interface IC .........................................................................................................
Initiating Condition ID ..............................................................................................................
Inside Diameter 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 23 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]MSIV ......................................................................................
Main Steam Isolation Valve MSL ........................................................................................................
Main Steam Line m R, m Rem , m rem , m REM ...............................................
m illi-Roentgen Equivalent M an 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 m and (NO)UE ............................................................................. (Notification Of) Unusual Event NUMARC 1 ...........................
....................... Nuclear M anagem ent and Resources Council OBE ......................................................................................
Operating Basis Earthquake OCA ..............................................................................................
Owner Controlled Area O DCM/O DAM ..........................................
Offsite Dose Calculation (Assessm ent) M anual 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 Instrum entation 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 24 of 359 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 25 of 359 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 AUl 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 26 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]VCNS NEI 99-01 Rev. 6 EAL IC Example EAL RG2.1 AG2 1 CUI.1 Cui 1 CU1.2 Cui 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 27 of 359 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 28 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]VCNS NEI 99-01 Rev. 6 EAL IC Example EAL SUI.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 29 of 359 EPP-108 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 30 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]ATTACHMENT 1 EMERGENCY ACTION LEVEL TECHNICAL BASES Page 31 of 359 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, offsite 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 32 of 359 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 "UE" for > 60 min.(Notes 1,2, 3)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.Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor I GE S 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 U)0, 2 X Hi-Rad 4 RM-A4 (gas) N/A N/A N/A 10RB 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*2- alarm I Discharge Mode Applicability:
All Page 33 of 359 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-6910-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 1 OCFR20. 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 10CFR20 limits for unrestricted area. The setpoint of the monitor is established in accordance with the Offsite Dose Calculation Page 34 of 359 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 tol 05 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-A13. 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-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. (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 35 of 359 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.EAL--#1--This EAL addresses normally occurring continuous as well as planned batch radioactivity releases from monitored gaseous or liquid effluent pathways.EAL #2 -This EAL addres.es radioactiVity releaser that cau.e effluent radi-tion monitor readings to exceed 2 times thc limit established by a radioactivity discharge permit. This-EAL Will typically be associGated with plaRned batch releases froM noncontinuous, release pathways (e.g., radwaste, waste gas).EAL #3 -Thi6 EAL addresscs UncontFOlled gaseous or liquid releasos that are detected by sample analyses or enviEronmental sur~veys, pa~tiGUlarly on u1Rnmonitoed pathways (e.g., spills, of radioactive liquids into storm drains, heat eXchanger leakaeirir water Escalation of the emnergency class~ifiation leve!ECL would be via IC AA4-.RA1.VCSNS Basis Reference(s):
- 1. Offsite Dose Calculation Manual 2. Design Bases Document -Radiation Monitoring System (RM)3. SOP-124 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" Page 36 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]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 AUM Page 37 of 359 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 38 of 359 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.
Classsification based on effluent monitor readings ass-umes6 that a releasie path to the environmet i6 established.
if the effluent floW past an effluent mon9itor is knoWn to have stopped due to actions to isolate the release path, then the effuen moni49torAt reading is no!onger 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.EW-^ #1 -This EAL addresses normally occurring continuous...
radioactivity relea.es.
from.Kmotnitored gaseous, or liquid effluent pathways-.
EAL #2 -This EAL addresses releases that cause efflunt rladiation monitor readings to eXceed 2 times the limit established by a radieacti~ty discharge pri.Ti EAL Will typically be associated with planned batch releases froM nO-on coninuous1 rANelease pathways (e.g., radwaste, waste gas).EA-#.3--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 emergency classification 1evelEOL would be via IC AAI-RA1.VCSNS Basis Reference(s):
- 1. VCSNS Off-Site Dose Calculation Manual (ODOM)2. NEI 99-01 AWl Page 39 of 359 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-G1 9A/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 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 a 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 mF!/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"3 Discharge Mode Applicability:
All Page 40 of 359 EPP-1 08 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/c)RM-A3/RM-A13
-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-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 Page 41 of 359 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 42 of 359 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 om..rge..Y level CIL would be via IC AS1-RS1.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. 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 43 of 359 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-G1 9A/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.
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 44 of 359 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 emergency c!assif*cation 1evelECL would be via IC A-1-RS1.VCSNS Reference(s):
- 1. EPP-005 Offsite Dose Calculations
- 2. NEI 99-01 Wl Page 45 of 359 EPP-1 08 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: RA1.3 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 46 of 359 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.Class6ification based on effluent monitor readings assum~es that a rolcase path to the eniro-nment s established.
If the ef!fuent flow past an monitor is knOWn to have stopped due to actions to isolate the relearse path, then the effluent mon)itorF reading is no longer valid for classification purposes.Escalation of the em.ergencY classificatio.n evelECL would be via IC AS4-RS1.VCSNS Reference(s):
- 1. EPP-007 Environmental Monitoring
- 2. NEI 99-01 AA1 Page 47 of 359 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: RS1.1 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-G1 9A/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.
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 R RM-A4 (gas) N/A N/A N/A 10RB 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"5 Nuclear Blowdown RM-L-9 N/A N/A N/A alarm"3 Discharge Mode Applicability:
All Definition(s):
None Page 48 of 359 EPP-1 08 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-A13
-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-A13 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-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-A13. 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 49 of 359 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 to10 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 50 of 359 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 emergency cla6sificGation lcvc!ECL would be via IC AG-1-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 51 of 359 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: RS1.2 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-G1 9A/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 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 52 of 359 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 emergency classification levelECL would be via IC AG-1-RG1.VCSNS Reference(s):
- 1. EPP-005 Offsite Dose Calculations
- 2. NEI 99-01 AS1 Page 53 of 359 EPP-1 08 ENCLOSURE I REVISION 01 [DRAFT E]Category: Subcategory:
Initiating Condition:
R -Abnormal Rad Levels / Radiological Effluent 1 -Radiological Effluent 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: RS1.3 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 54 of 359 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.Classification based on cifluont moenitor readings assumes that a release path to the enviGronment is established.
if the effluent flo)W past an effluent moniRtor is knoWn to havedue to actionsr to isolate the release path, then the r-eading i6 nR longer valid for clas-ification purpo.se.Escalation of the emergency classification leve!ECL would be via IC AG4-1RG1.VCSNS Reference(s):
- 1. EPP-007 Environmental Monitoring
- 2. NEI 99-01 AS1 Page 55 of 359 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 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-G1 9A/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.
Table R-1 Effluent Monitor Classification Thresholds Release Point I Monitor J 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 0 2 X Hi-Rad RB Purge exhaust RM-A4 (gas) N/A N/A N/A 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"I Discharge Mode Applicability:
All Definition(s):
None Basis: Page 56 of 359 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 e/Q)RM-A3/RM-A13
-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-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-A13. 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 57 of 359 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 to10 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-G19A/B/C monitor readings are affected by N1 6 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 58 of 359 EPP-1 08 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 59 of 359 EPP-1 08 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 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-G1 9A/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 60 of 359 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 61 of 359 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 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 62 of 359 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.Clarssification based on; off luent monitor reading6 assumes that a releasc path to the envionmet is- established.
it the effluent flow past an effluent mo~nitor i6 known to have stopped due to ac:tionsr.
to isolIate the release path, then the effluent mnonitor reading is no longer valid for classification puprposes.
VCSNS Reference(s):
- 1. EPP-007 Environmental Monitoring
- 2. NEI 99-01 AG1 Page 63 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category: Subcategory:
Initiating Condition:
EAL: R -Abnormal Rad Levels / Rad Effluent 2 -Irradiated Fuel Event Unplanned loss of water level above irradiated fuel 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 64 of 359 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-G1 7A/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 65 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Page 66 of 359 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 emergency clasificGation lcv-lECL would be via IC AA2-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 67 of 359 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 68 of 359 EPP-1 08 ENCLOSURE I REVISION 01 [DRAFT El* 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-(e-ee Drv.lopcr 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.This IC applies to irradiated fuel that is, licensed for dry storage up to the point that the loaded storage cask is scaled. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC EIHUl.Page 69 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Escal-ation of the emergency WOUld be ba-sed en either AoeGnqition Category' A 9r G EAL#1 This EAL escalates from AU2-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.EAL#2 This EAL addresses a release of radioacte materia!l aused by mechanial darmnage to i rradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assemnbly.
A rise in readings on radiation monitors sh ould be considered
- n conjunction with in-plant reports or obseriations of a potential fuel damnaging cvcnt (e.g., a fuel handling aGGident).-EAL #3 Page 70 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Spent fuel pool water level at this value is Within the lower end of the level range necessary to prcvent significant dose conseq-uences from direct gamma radiation to personnel performing Operations in the vicinity of the spent fuel pool. T-his condition reflec-tsR a; sign~ificant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cOO! the irradiated fuel assem~bles stored *n the pool.Escalation of the emergenRcy classification lcve'ECL would be via ICs A&4-RS1e9-AS2-(see AS2 geve~oer Aietee).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 71 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category: Subcategory:
Initiating Condition:
EAL: R -Abnormal Rad Levels / Rad Effluent 2 -Irradiated Fuel Event Significant lowering of water level above, or damage to, irradiated fuel 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 72 of 359 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-4see
.ote ). 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.This IC applics to irradiated fuel that is lizensed for dry, storage up to the point that the leaded storage cask is sealed. Once 6eaed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC E-HUl.Page 73 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Escalation of the emergency would be based on either Recognition Category A-Ror C ICs.EAL #TThis EAL cecalate-froM AU2 in that the loss Of level, in the affoeted potioRn Of the REFUELING PATHW.AY, is of sufficient magntude to have resulted in uncover of irrFadiated fuel. Indicat*ions of irradiated fuel unRove,', may inRcdh direct or indirect vi6sial observation (e.g., repEort from poI sonnel or cmr iges), a well as signifi change. in water and radiation levels, oF other plaRt parametersV.
ComputatiRoal aids m.ay also be used (e.g., a boilloff curve). Classification of an eveRt using this EAL should be based on the totality of available indications, repot and obseration.
--Thiole an area radiation mniEto oruld de ma icaused n a dose rate due to a ioweringeof wate level in somevntsion of thheREFUELING PATHiAY, the reading mfay ot be a remlable nrdication of whether or not the fuel is actually uncovered.
To the degree ponsible, readings s hould be cons idered in combinatihn with other available indications of A drop iRwtrlevel above irradiaed fuel WJthin the roactE)F vess~el ma" be Gl6 ini accordance Recognition Categor,'
C during the Cold Shutdown and Refueling Modes.-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).EA--#a Spernt fuel pool water level at this value is within the lower end of the level range ,-,o,,ssoa,., to pre,,e.+ 61gnifi;an dos G..O o,-,, ,n,.o6 f..-, garn radiat+e to perSOnnel perfo)Frming EoperatiE)s in the vircinity of the spent fuel pool. This conditiona 6ignif..a.t 1los Of spent fuel pool wate r invento.'
and thus it is also a precursor Escalation of the emergency classification levelECL would be via ICs AS--RS1orAS2- (see AS2 DRvelfper Ntces).VCSNS Reference(s):
Page 74 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]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 359 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-1 2-051 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 359 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-(&ee D.ve.per ..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.Ltw n~W ^1 na T rrna n In~ 1]^ 1r n'au 7~ ata I.Ar a t1. j .! V r% .rt r% ti*-n I! A~i~ e4 #~~ I/~ 0Li C aw n Ln A n'r-uan inn~InA A 4 mraI ~I~~r #'cL ..I/ n"ruM 1^n,- Mt +I- In CONFINEMENT BOUNDARY is ,lassificd in acordance with C E HUI.Escalation of the emnergency would be based on either Recognition Categor~y A or C This E:AL escalates, froem AU2 in that the los6 Of level, in the affected portion of the REFUELING PATHWAY, is Of cuff icent magnitude to have resulted in uncover,'
of irradiated fuel. Indications of irradiated fuel uncover,'
may inelude direct or indirect visual aaaa. V ..ALi~. I .~., I a~.au La VI~l,. ~.%' I t.VaVmV.t..I am a..... I *'..* t4 mlm It.t~Jt.aJ, tAJ v.a.. t.ta aI~ IImISJt.ti m t:,i;l, u:-;Ii ~.;IAIt- cInr i rA c ti iti ^n .,-i- ^%Ms 1 il n il n ~ :i zs I 11-i cii
- ui-MMAArg-
^.Ar" iriici I :l *A+n- fMA A. A.---..-.-.-.
..,...-I..--
---MA------.
algz o Ie mwd (en C a hoil off eir-, MuQ Cý14IZU'ifiAtin Qf :an eount I uzinq thic FAl ghniiH ho k~~~~ +,a 4 a h~II. .. 4 ii I., k!,. ; A; +.4 ,.-.a e aa v+ A aket +;a, ni I A I i .4; A lewcring of water level in some portion of the REFUELING P.ATH\A.IAY, the reading may not bhe a rcliabl~e indic-ationA of wthether or no the fueli ctal uncoevered.
To the degree possible, readings should be considered iR combination with otherF available idctoso A drop in water level above irradiated f ue! witi the reactor vessel m1Fay be ciassified in aG*da~ .,,,,it*G Category. th.e.. Co-ld Shudow ;::'- ..... moes-EAL #!2 This EAL addcreeoC'p a mol ozi f r~adioaet.i ma nra' w*iad h" monrh~anu'nIq cnr.Q~qrw trn irrad*=rtarI futal flnma nn~ a,,ar*c. ma. *macuim Irhtt drin Jrtfnn hum Immlf r hineAnfn ^f -n ..... ,..VV 119 11 W -11 assemoby, or dropping a neavy lead onto an assembly.
A rise inta reangs on ranaiaIG MGnitor56 shou~ld be GOR60dered in Gonilinetio
.. lh in ht renork er oh5~ervation'6 of a notetia!fue damninoeveto Afilial Mie h n"In AeerlpentV
.........
v ..... v ......I-Page 77 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]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 ICs A-4-RS1F -AS2-( ,ee S2Devlop-er.
Noter4.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 78 of 359 EPP-1 08 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 79 of 359 EPP-1 08 ENCLOSURE I REVISION 01 [DRAFT E]Escalation of the emergency classification level would be via IC AG4-RG1 or AG2RG2.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 80 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category: Subcategory:
Initiating Condition:
R -Abnormal Rad Release / Rad Effluent 2 -Irradiated Fuel Event 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 81 of 359 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 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 AG2 Page 82 of 359 EPP-1 08 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.For EAL #2, an Alert declaration i .warranted if entry' into the affected room/,roi y be, POdually required during the plant operating mode ineffect at the time of the elevated radiation levels. The emergency Page 83 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]plassification is not contingent upon wihether entry is actually necessary at the time of the icrGeased radiation levels. Access rshould be considered ar, impeded if extraordinar~y r~kalng-ý^+n.
f.ar.Ii4atim aintni, p~ areaninnaI
- anr th a ~ffor-tod rflflm/j~p'a (e aaling tem pr-ary -rni 1lU rg, rF ure in g u Of rT nR nnr utnlne p rnTe lc e nqu nm en T,,
- :9 )- -1.,,- ---., , v ,!.,. J. JX. , ..,, requesting an exten.ion in ,doe limits beyond normal administrative ,imits).An emergency declaration is not warranted if any of the folloGWing conditions apply.9 The plant is in an operating mode different than the moede specified for the affoctod room/area (i.e., entry i6 not required during the operating moede in effect at the time radiation increase occu~rs, and the procedures used for norm~al operation, cooldoWn nnA tin A r rI r , lir^ n+ , *nfn
- hr. ffa ,,t 4 r, ,i a, in+*! RA ar4rn A-~ ~~ ~ T h n r r r a l r -A -t a allr ' r .' a , a t a yn n n r .n i ~ t g * .t m i ...K v ccompensatorn measue, swhic haddre-,, sth e temporar yinaccessibilit F -rarea (e.g. ,radiographyh r-,pent filter 9r rFrei ntransferr ,etc.)" +Re aGtiE)nT()F;AfR1A-M ree-FWaFea eRt ' G 16 of aR admiRlWative OF FeGOFE]kee., i _ (e.g., nermal n-r r4eu-iine im6peGtieRs)." The aGGe66 GGAUG! measures aFe of a GGR69Fvative oF preGautionaFy natuFe, aRd WC)"'d ACAaGtually nreveRt eF impede a reeuiFed aGtiE)R.................
] Escalation of the AR, C or F ICs.eFReFqeRGy 1J1Ctbft1t1HCttK4H 1eYelECL would be via Recognition Category 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. NEI 99-01 AA3 Page 84 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category: Subcategory:
Initiating Condition:
R -Abnormal Rad Levels / Rad Effluent 3 -Area Radiation Levels 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 85 of 359 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 EAL-#2RA3.2, 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 Page 86 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]the 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 emerqency 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 em.rg.n.y classification le.e..OL would be via Recognition Category AR, 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." Page 87 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]2. NEI 99-01 AA3 Page 88 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category C -Cold Shutdown / Refuelinq System Malfunction EAL Group: Cold Conditions (RCS temperature
-200 0 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 89 of 359 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 90 of 359 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: CU1.1 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 91 of 359 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
[PW,'] or RPV' [,W'J) 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 leakaqge warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered.This EAL-#4 recognizes that the minimum required (reactor vessel/RCS
/t-fwJ orvRPV[B4W4'J 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 Page 92 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]longer. The minimum 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.EAL= #2 addresses a condition where all mneans to determine (reac~tor Yesce!RCS[P"4 or RPV [814q) level have been lo't. Rn th;i conditieo, may determine that an inventorY lo66 isocurn by obseP.,ng changes- in 66ump and/or tank levels. SUMP and/or tank le c.nhanges must be evaluated against other potential sources of water flow to en6sue they are indicatiVe of leakage from the Ovessol/RC
[vWvv or RPV Continued loss of RCS inventory may result in escalation to the Alert emergeRGY classification lev'IECL 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-101 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 93 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT El Category: Subcategory:
Initiating Condition:
EAL: C -Cold Shutdown / Refueling System Malfunction 1 -RCS Level UNPLANNED loss of reactor vessel/RCS inventory for 15 minutes or longer 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 0 RB Sump 0 CCW surge tank 0 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 94 of 359 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
[P' AJ]-or ,,RP\. [B--R]) 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 I procedurally required limit concurrent with indications of coolant leakaqe warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered.EAL #1 recognizes that thc minimumn requircd (reactor Yecsscl/RCS
[PWFJ or RP2V[BWFVjJ) level can changc several times; during the course of a refueling outagea ifrn plant configurations and system lncs armplcmcnted.
This EALismtfthmnmu level, specified for the current plant. coýnditioNns, cannot be maintained for 15 minutes 0o Page 95 of 359 EPP-1 08 ENCLOSURE I REVISION 01 [DRAFT E]longer. The mninimum level is typically specified in the applicable operating proccdurebu may be Specified in another contro.lling doument.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.This EAL-#2- addresses a condition where all means to determine (reactor vessel/RCSOr RPV [.\,.,WR) level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump apA/Gr/Itank levels.Sump apd/Gr-!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,,PWR. ,v RPV [BWR-.Continued loss of RCS inventory may result in escalation to the Alert ene-geRy cla6sification Icp-fpECL 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-1 01.1 Loss of Reactor Coolant not Requiring SI 9. FSAR Section 5.2.7.1.3.8 10.NEI 99-01 CU1 Page 96 of 359 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"72 = 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 97 of 359 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-#-I, a lowering of water level below (,ite Specific level) 429'-6" indicates that operator actions have not been successful in restoring and maintaining (reactor vessel/RCS
[PWRI or RP,.' [BWR]) 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-#4- 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 I Reidal Decay Heat Removal suction point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.For EAL #.2, the inability to (reacGto vessel/RCS
[P\R] or RP' [BWR]) leve!, may be cau sed by instrumentation and/or power. failures, or water level dropping below the range of available intue~to.if water level cannot be monitored, operators mnay levels. Sum an/o t ank level changes mnust be evaluated against other potential source or IPVI BWRD1.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
[PWR] or RPV D inventory level continues to lower, then escalation to Site Area Emergency would be via IC CS1.VCSNS Basis Reference(s):
Page 98 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]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 99 of 359 EPP-1 08 ENCLOSURE I REVISION 01 [DRAFT E]Category: Subcategory:
Initiating Condition:
EAL: C -Cold Shutdown / Refueling System Malfunction 1 -RCS Level Loss of reactor vessel/RCS inventory 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* 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 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 100 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT El 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 EgAL #1, a lowering of watef level below (bite-pecific lovel) indicates that operata actirmn have not been toruessful in retorring and maintaering (reateor veispORGS
[ptak oer PV [el. 1) ater level. The heat up Fat the coolant wil aincrst ao the available wate r is eod. A cortinuingdeicate n water level will lead to r URGeyeVy Althoughinu FeIted d Ai #1 io concered with the loss of le inVenbtoe
' and tof the potential concurrent effects On systems neede for decay heat removal (e.g., lo6s Of a Residual Heat Remova6l1sucion point). An incr-ease in RCS temperature caused by a loss of decay heat renmval capability is evaluated under ICn G C For this EAL-#2, the inability to monitor (reactor vessel/RCS[,]orR[,nRtr level 9Fnti o 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-andIo-Itank levels. Sump-~ani4Lorztank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the (reactor vessel/ROS
[PWAR] or RPV [B\AR]).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
[PWAIR or RPV [BWARI} inventory level continues to lower, then escalation to Site Area Emergency would be via IC CS1.VCSNS Basis Reference(s):
- 1. ARP-001-XCP-615
- 2. FSAR Section 5.2.7.1.3 3. AOP-101.1 Loss of Reactor Coolant not Requiring SI 4. FSAR Section 5.2.7.1.3.8
- 5. NEI 99-01 CA1 Page 101 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Page 102 of 359 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 103 of 359 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)* 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 104 of 359 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
[PW or ....RIV ,BW-R])-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 EALs-1-.bCS1.
1 and 2-.bCS1.2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.
!n EAL 3.a, the 304ninut criterion is ticd to a rcadily recognizable eve'nt start timne (i.c.,th total loss of ability to moenitor level), and allows suff icient time to moneitor, assess and correlate reactor and plant conditions to dotermine if corc uncover; has actually occurred (ieto acco~unt for vaou acient progression and "RstrumentatieR uncertainties).
it as alloWs sUfficient time for nortorancne of actions to terminate leakane reoerivntor'---.-..I-.----.-----
flf~rl rf' ~n n n flnx+ Afl, Ir+ rconaI,, Im ntaa The-inabilityt monitor (reactorF vessel/RCS
[PWR] or RP BR)level mnay be caused by instrumentatio~n and/or power failueres, or water level dropping below thc range of availabl instrumentation.
If water level cannot be monitoed, operators may determ~ine thata inventor; los isouring by obseR'ing change6 in sump and/or tank levels. Sump and/or ensure they are indicative of leakage from the (reactor vessel/RCS
[PWR] or RP'I [BWR])These-This EALs addresses concemns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-Page 105 of 359 EPP-1 08 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 emergency cGasification 1eve!ECL would be via IC CG1 or AG-I-RGI.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. OAP-108.4 Operations Outage Control of Containment Penetrations
- 7. SSP-004 Outage Safety Review Guidelines
- 8. NEI 99-01 CS1 Page 106 of 359 EPP-1 08 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 107 of 359 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 r rRPVinventory 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.
Page 108 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]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 EA1-44CSI.
1 and 2-bCS1.2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.
In EAL 3.a, the 30-minute criterion i6 tied to a readily recognizable ovont start time (i.e.,thloss of ability to monitor level), and allows sufficient time to monitor, assess and corre;Ato and plant conditionsrto doetoe .. Awe if c ,reR-Guncer, ha Ucurred allows SUfficient time for perform.ance_
Of acGtions to ter~inate leakage, recover ivno coentrol/,makeup equipment and/pr restore level monRitoring.
The inability to mAonitor (reactor Vessel!RCS
[PWR] or RPV [B3WR]) level may be caused b inStrumentation and/or power failures, or water level droppingbelow the range of ava~iabl insturumentation.
if water level cannot be monitored, operators may determine that an inventor' loss is occurring by ()bser.'ng change in up and/ortank levels. Sump and/e tank level changesmust be evaluated against other potential sourGes of waterF flow to ensure they arc indicative of leakage from the (reactorvesl!C
[PWR] or RPV [B3WR]).Th4eae-ThisEALs 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 emnergency Glassifiration Ieve!ECL would be via IC CG1 or AG-1-RG1.VCSNS Basis Reference(s):
1 .GOP-9 Mid-Loop Operation 2. SOP-i 01 Reactor Coolant System 3. SOP-i 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-1 08.4 Operations Outage Control of Containment Penetrations
- 7. SSP-004 Outage Safety Review Guidelines Page 109 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]8. NEI 99-01 CS1 Page 110 of 359 EPP-1 08 ENCLOSURE I REVISION 01 [DRAFT E]Category: Subcategory:
Initiating Condition:
C -Cold Shutdown / Refueling System Malfunction 1 -RCS Level 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 111 of 359 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 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 -105 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. 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 112 of 359 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 LPWRI 9 -RPV-\, f'WR]) 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.Outagc/shutdown contingcncy plans typic~ally provide for Fe establirshing or verifying CONTAINMENT CLOSURE folloWing a loss of heat rmoval oFr RoRtc Eontrol functions.
The difference in the rspecified RCS/reactor
'.,essel levels of EAAI=s 1 .b and 2.b refbe~t the fac-t thaRt with CO-NIT.AINMENT CLOSURE ostablishod, there is a loer probability of a fissionR product release to the A-irnmn.In-EAL 3.a7-tThe 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 vesseVRCS
[PWR] or RPV [BW.--R])
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 and/er/_tank levels. Sump Page 113 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]ampdi4-/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.rPWR]
9F.,PV (f3WRI).These-This EALs 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 emergency classification leve!ECL would be via IC CG1 or AG-1-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 114 of 359 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: CG1.1 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 115 of 359 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. 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 116 of 359 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 117 of 359 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 118 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]the other listed indications to assess whether or not containment is challenged.
inTEAI=2.b--tjhe 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 mRonitor (FeoGtor voss ll l/RC1IS [PKJ 9or RPV [BWVI) leVel May be caused by instru~mentation and/o pwe faiues, or water level dropping below the range of available instrumentation.
if water level cannot be moniRtored, operators may dcteFRmine that an inventoY los isocuring by Gbser~ing changes in sump and/or tank levels, Sump and/or tank level changes must be evaluated against other potentia!
sources of water flow to ensure they are ~nic4atiye of leakage from the (reactor Yvcscol_/RCS
[PVWRI or RPV [BWqj)-.Thisese EALs 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-122 Post Accident Hydrogen Removal System 7. FSAR Section 6.2.5.5.3 8. FSAR Section 6.2.3.5.1 9. OAP-108.4 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 119 of 359 EPP-1 08 ENCLOSURE I REVISION 01 [DRAFT E]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: CG1.2 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 0 PRT o RCDT Mode Applicability:
5 -Cold Shutdown, 6 -Refueling Definition(s):
Page 120 of 359 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-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. 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 121 of 359 EPP-1 08 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 -105 mR/hr. If any of these radiation monitors reach and exceed 105 mR/hr (off scale-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 122 of 359 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 123 of 359 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 124 of 359 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.
inFAL 2.b4The 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
[PWH. or RPV [LV'" 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-and/r-Itank levels. Sump and/or-ttank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the (reactor vessel/RCS'r
[P JorRPV 18WRI).Page 125 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT El Thisese EALs 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. 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-61 10 10. ARP-001-XCP-615
- 11. FSAR Section 5.2.7.1.3 12.AOP-101.1 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 126 of 359 EPP-1 08 ENCLOSURE I REVISION 01 [DRAFT E]Category: Subcategory:
Initiating Condition:
C -Cold Shutdown / Refueling System Malfunction 2 -Loss of ESF AC Power 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 0 230 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):
Page 127 of 359 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): 9 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 128 of 359 EPP-1 08 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 an erniegenGy-safeguards bus. Some examples of this condition are presented below.Page 129 of 359 EPP-1 08 ENCLOSURE I REVISION 01 [DRAFT E]* A loss of all offsite power with a concurrent failure of all but one eme-geRoy safecguards power source (e.g., an onsite diesel generator).
- A loss of all offsite power and loss of all 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 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 130 of 359 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 -.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 0 230 KV power to XTF-31 0 Parr Hydro Plant 13.8 KV power to ESF bus 1DA or 1DB Onsite: 0 Diesel Generator A 0 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 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 131 of 359 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): 0 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 DR (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 132 of 359 EPP-1 08 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 an emergency safequards 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 emcrgcncy classification levelECL would be via IC CS1 or ASI-RS1.Page 133 of 359 EPP-1 08 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 134 of 359 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 > 200°F 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°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 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 135 of 359 EPP-1 08 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 limit, or the inability to determine R.S temperature and leveland 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.
EAL-#!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.EAL #2 a condition where there has been a significant 1l6 Ofi capability necessar,'
to- monitor RCS conditions and operators would be unable to monto key parameters necessar,'
to assure core decay heat removal. During this condition, thr is noR immediate threat of fuel damage because the co~re decay heat load has been Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.
Page 136 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT El 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. NEI 99-01 CU3 Page 137 of 359 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-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 138 of 359 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 W,-EAL addresses an UNPLA^NNED increase in R"S temperature above the TechnRial Specification cold shutdo, W temperature limit, or 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.A momentar' UNPLANNED excursion above the TechniGal coldi'fiiGR
..Id shutdown temperature limit when the heat removal function is available does not warrant-a GaiG~a~Stien-Page 139 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT El E-AL 41 involve6 a lo66 of decay heat remo'.al capability, orF an -addition of heat to the RCS in excess of that which can curr~ently be removed, such that reactor coolan temperature cannot be maintained below the col'd s.hutdown temperaturc limit specifiwed in Technical Specifications.
During this conditioR, there is no immediate thret of fue l damage because the acoe decay heat load has been reduced since the cessation of power 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 conrol1led.
A loss of forced decay heat remova a reduc~ed inventor' ma" result in a rapid increase in reactor coolant temperature depending OR the time after shu1tdown.
EAL#42This 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.
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 140 of 359 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 REDUCED INVENTORY N 6 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-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. 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 141 of 359 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 142 of 359 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 143 of 359 EPP-108 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--P-WR, 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.EAL #2 prvidos a pressure-baed indication of RGSheRat-up.
Escalation of the .mergec ..y cla.ification leeIECL would be via IC CS1 or AS-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 144 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT El 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.-Either A momcnntan' UNPLANNED cX-ursion above the Technical Spocification cold shutdoln tempcraturc limit when the heat romoval function is available does not warrant a The RGS Heat-up Duration Thresholds table addressesa inceaein RGS temperature w.Ahen CODNTAINIMENT CLOSURE is es~tabiishod but the RCS is, not intact, or RCS Page 145 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT El fl~AfRA rL~.. nn m+/-. ~.:.a a& _Z_ a ma"'W"11XV4 1 A .4I .. .ip I ii. S-, --. ~ -The RCS Heatu, Durat.on Th.eshold. , table also addessc, an incrase, nCS tommnnratura wpth thin RC'Q On;rt+ii Tho attu.u, rif CON~ITANINAIMET 1 02 CIRE OR1 aRin" nrUimni BI 1 I %4 product relcase. The 60 mninute timne frame should allow SU#;cient time to address the temperature:ces without -A su-b~tan~tial degradation in plant safety.Finally, in the case where there is ainrse in RCS temperature, the RCS is not itc nr Or. ai rrfimdca nf( ,nrstnn, EJD1AZO 2n ~r-Oh CCdITAINtAIT LQ!C'QI 1a1= *a' rnt icratakihrvaiod ma 1ý A t',-,",Olnt may, be re ,,r,".,, into 'I the, C) ontinmuent ah n
,,--m--,"--
-,-, the ..i.r.nment.
and 2) there is-, reduced reactor colant.inventor above tho too of irradiated fuel p This EAL-#2 provides a pressure-based indication of RCS heat-up.Escalation of the e.mergoeny4 G ass
'-4eeECL 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
- 4. NEI 99-01 CA3 Page 146 of 359 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 147 of 359 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 e~mergency c!Gassification leveECL would be via IC CA1 or CA3, or an IC in Recognition Category AR.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 148 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT El 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 149 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category: Subcategory:
Initiating Condition:
EAL: C -Cold Shutdown / Refueling System Malfunction 5 -Loss of Communications Loss of all onsite or offsite communications capabilities.
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 150 of 359 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 EAL-#4-addresses a total loss of the communications methods used in support of routine plant operations.
The second EAL conditionEAL-#2 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 (see Developer Notes)as well as the NRC.of an emnergenc~y decalaration.
VCSNS Basis Reference(s):
- 1. FSAR 9.5.2 2. EP-100 Radiation Emergency Plan, Section 7.5 3. EP-1 00 Radiation Emergency Plan, Figure 7-2 4. NEI 99-01 CU5 Page 151 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Category H -Hazards and Other Conditions Affecting 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 152 of 359 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 153 of 359 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 154 of 359 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 Orcianizations.
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].E-AL-#--The first threshold references (6ite-speGifi-the seGUtY Security shift-Team su .qeLeader) 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.
EAL--#2The second threshold addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with (sit -specific prOcedure.the VCSNS Security Plan (ref. 1).EAL-#-3The 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 Page 155 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]NORAD through the NRC. Validation of the threat is performed in accordance with the VCSNS Security Plan (ref. ,,specifi, ,p ,,.dure,.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 emergency classification le,-eIECL 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 156 of 359 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 157 of 359 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 Shift Supe-.,iTeam 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.EAL-#!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.EAL-#2The 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.~J ..........T--.d. --r ........ I" Page 158 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT El 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 emergency classificatiGn
!evc!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 159 of 359 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: HSI.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 160 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]Timely and accurate communications between Security Shif, -ipepii,-nTeam 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 Or-qanization (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 Safeguards-Plan (ref. 2).Escalation of the ,me.geR.Y classification 1cv..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 161 of 359 EPP-1 08 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 162 of 359 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 Shift ...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 conceming 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 163 of 359 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 greater than OBE levels EAL: 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 164 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]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/eqcenter/
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., typioal-lateral accelerations are-in excess of 0.08gl__A).
The Shift Manage-rShift Supervisor or Emergency Director may seek external verification if deemed appropriate (e.g., a call to the USGS, check intemet 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 emergeRG classificGaton levelECL would be via IC CA6 or SA9.VCSNS Basis Reference(s):
- 1. ES-426 Earthquake Response Procedure 2. FSAR Section 3.7 Page 165 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]3. EPP-015 Natural Emergency 4. FSAR Section 2.1.1 5. ARP-001-XCP-638
- 6. NEI 99-01 HU2 Page 166 of 359 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 167 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]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.RI I h1bThe 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.E-A-L--1-2The 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 emcrgcncy classification
!evc!ECL would be via IC CS1 or AS-1-RS1.VCSNS Basis Reference(s):
- 1. ES-426 Earthquake Response Procedure 2. NEI 99-01 CA6 Page 168 of 359 EPP-1 08 ENCLOSURE I REVISION 01 [DRAFT El 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.EA--#1-EAL HU3.1 addresses a tornado striking (touching down) within the PROTECTED AREA.EAL #2 addresses flooding of a building room or area that results inoperas io power to a SAFETY SYSTEM component due to water lcvel or other wetting concerns.Classification is also required if the watcr level or related wetting causes an autom~atic isolation of a SAFETY SYSTEM component from its power sourcGe (e.g., a breaker or relay trip). To warrant Glassifiation, operability of the affected component must be required by Technical Specifications for, the current operating mode, Page 169 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]1~I-~L ir:i annre~u~se~~
a nazarnotJf~
mareriais CVOflT orininaTino aT an onsi~e IOC3T!Ofl ann OT .--.-.----.
~,.~----...-..---------
cff6ri~~ 6 + 4 iArk +r imre, A h~ +k +f rrr1 6+k,ihr +k, D~CnTrLC Tt~r AIA EAL #4 addre6se, a hazardous event that causes an on. itc impedimen+nt to -ehicle movement and sigRifi cant enough to prohibit the plant staff fm acces s.ng V thite using personal vehicles.
Examples Of suc~h an event inc~lUde site flooding caused by a hurricane, heavv rains UDrve ater roloasos.
damn failure, etc., or an on-site train derailMont blocking the access Froad.Th;s EAL ir- not intended apply to routine impediments sUch as fno,, i;e, or ,ehiclI breakdo'Wns or but rather to more significan.
conditionssuchas the Hurricane Andrew strike On Turkey Point iR 1092, the flodn Arond the Cooper Station during~ the MidwesA-1t floods1 of 1993. or the flooding around Ft. Cahon ta;tion in 2011 .EAL #5 addresses description).
Escalation of the eme..e.Categories AR, F, S or C.IUy classification ileveECL would be based on ICs in Recognition 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 170 of 359 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 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;(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 171 of 359 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.EAL #1 addresses a torado striking (tuching down) within the PRO T ECTIED AREA.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.uw icient magrnitude to impede the me'emont of po RRonnel Within the PROTECTED AREA.EAL #4 addresses.
a hazardous event that causes an On site imeietto veh*Gle movoene~t and signifficant enough to prohibi the plant staff from accessing the site using presOnal vehicles.
ExamplesE of Es;uh an event include site flooding caUsed by a hurrF;ane heavy rains, up. rvewater releases, dam failure, etc F oran on -site train derailment blocking the access road-.This EAL is not intended apply to routine impediments suGh a6 f, io or vehile breakdowns or accidents, but rather to more s ig nif icant- conditionsmi sý6 uch-1 as- the* Hurricane AnrGIew strike 9R TurkeyPiti 1992, the flooding around the Cooper Station during the MAd......t floodsof .1993, or. the......
f.. arou Ft. Calhoun Station in 20 1.EAL #5 addresses (site-specific description).
Escalation of the emergency classification leve;ECL would be based on ICs in Recognition Categories AR, F, S or C.VCSNS Basis Reference(s):
- 1. VCSNS IPE Internal Flooding Analysis Workbook 2. NEI 99-01 HU3 Page 172 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT El 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.E=A' #1 Addrpqrcs ai tornad sriing (ouching doWn) W4ti th PRO.-TECTD AREA.Page 173 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]T-h~-, AL a*96es feed~g o a uild~g orm9 area that results :n operators isolating power to a SAFETY S-YS-TEM.
cOMPOncnt due to water lovel or other wetting concerns.Cla.sification is al.o required if the water level or related wetting cause6 an automatic isol.atio of a SAFETY SYSTEM o n .fro 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.EAL-#3This 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.EAL #4 addresses a hazardo~us event that causcs an on site impediment to vehicle mEovemet and signifcant ereough to pfohibit the plant staff om accesinIg the site using p99rsonal vehicles.
Emxamplesof such ane evxent include site flooding c-aUsod by a hurricane, hCavy rains, e watr rlaCses, dam failure, etc., or an on site train derailment bloDkring the access road.This EAL is not intended apply to routine imapediments such as fo e, or vehicle breakdowns or accidents, but rather to more significaR"nt cndCiwtions suc--'-hI -as tihe Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the MidwA.est flo)ods of 1993, or the fleeding around Ft. Calhoun Station in 2011.EAL #5 add~ereses (site- specif ic desrGiption).
Escalation of the emnergency cl31ass-ifica3-tion levelECL would be based on ICs in Recognition Categories AR, F, S or C.VCSNS Basis Reference(s):
- 1. Drawing SS-024-019 Site Plan 2. EPP-01 4 Toxic Release 3. HazMat Risk Assessment
- 4. NEI 99-01 HU3 Page 174 of 359 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.EAL #1 addresses a tornado striking (touching down) within the PROTEGTED AREA.This EAL addresses flooding of a building room or area that results in operatrse iselat;gn power to a SAFETY SYSTEM component due to water level or other wetting conGernS.Classifiation is also required if the water lee or related wetting causes an automatic 6olation of a SAFETY SYSTEM compOnent frM its power source (e.g., a breaker or relay trip). To warrant Glassification, operability of the affected com~ponent must be required by Technical Specifications for the current operating mode.-EAL #3 addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED ARA EAL-#4This 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 Page 175 of 359 EPP-108 ENCLOSURE I REVISION 01 [DRAFT E]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.EAL #5 addresse6 (6ite-specific de6cription)-.
Escalation of the emergencY classification evelECL would be based on ICs in Recognition Categories AR, F, S or C.VCSNS Basis Reference(s):
- 1. NEI 99-01 HU3 Page 176 of 359