HNP-18-004, License Amendment Request to Change Shearon Harris Nuclear Power Plant, Unit 1, Emergency Plan Emergency Action Level Scheme

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License Amendment Request to Change Shearon Harris Nuclear Power Plant, Unit 1, Emergency Plan Emergency Action Level Scheme
ML18226A022
Person / Time
Site: Harris 
Issue date: 08/13/2018
From: Hamilton T
Duke Energy Progress
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
HNP-18-004
Download: ML18226A022 (111)
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{{#Wiki_filter:Tanya M. Hamilton Vice President Harris Nuclear Plant 5413 Shearon Harris Road New Hill NC 27562-9300 10 CFR 50.90 August 13, 2018 Serial: HNP-18-004 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Shearon Harris Nuclear Power Plant (HNP), Unit 1 Docket No. 50-400 Renewed License No. NPF-63

Subject:

License Amendment Request to Change Shearon Harris Nuclear Power Plant, Unit 1, Emergency Plan Emergency Action Level Scheme Ladies and Gentlemen: In accordance with the provisions of 10 CFR 50.90, Duke Energy Progress, LLC, (Duke Energy) is requesting approval of proposed changes in the Emergency Plan Emergency Action Levels (EALs) used at the Shearon Harris Nuclear Power Plant, Unit 1 (HNP). The proposed changes to the EAL scheme will correct deficiencies identified and bring the site into alignment with the approved EAL methodology, Nuclear Energy Institute (NEI) 99-01 Revision 6, Development of Emergency Action Levels for Non-Passive Reactors. provides an evaluation of the proposed changes to the EAL scheme. Enclosure 2 provides the impacted pages of the EAL Technical Bases Document (clean version) for HNP. Enclosures 3 provides a markup version of the impacted pages of the EAL Technical Bases Document. Enclosure 4 provides the supporting calculation for the changes made to the containment radiation monitor EAL threshold values listed in the HNP EAL Table F-1, Fission Product Barrier Threshold Matrix. Enclosure 5 provides the supporting calculation for the change made to the radiation monitor reading for core uncovery during refueling that is used for cold shutdown EAL thresholds. Enclosure 6 provides a markup version of the HNP EAL Wallcharts for the proposed changes. Duke Energy requests Nuclear Regulatory Commission review and approval of this license amendment request within one year of acceptance. The amendment shall be implemented within 180 days following approval. In accordance with 10 CFR 50.91, a copy of this application, with enclosures, is being provided to the designated North Carolina State Official. This document contains no new regulatory commitments. Should you have any questions regarding this submittal, please contact Jeff Robertson, HNP Regulatory Affairs Manager, at (919)-362-3137.

U.S. Nuclear Regulatory Commission Serial HNP-18-004 Page 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on August 13, 2018. Sincerely, Tanya M. Hamilton

Enclosures:

1. Evaluation of the Proposed Changes to the Emergency Plan Emergency Action Level Scheme
2. Proposed Harris Nuclear Plant Emergency Action Level Technical Bases Document Changes, EP-EAL (Clean)
3. Proposed Harris Nuclear Plant Emergency Action Level Technical Bases Document Changes, EP-EAL (Markup)
4. Calculation for Containment Radiation EAL Threshold Values
5. Calculation for Radiation Monitor Readings for Core Uncovery during Refueling
6. Proposed Harris Nuclear Plant Emergency Action Level Wallchart Changes (Markup) cc:

J. Zeiler, NRC Sr. Resident Inspector, HNP W. L. Cox, Ill, Section Chief, N.C. DHSR M. Barillas, NRC Project Manager, HNP NRC Regional Administrator, Region II

U.S. Nuclear Regulatory Commission Serial HNP-18-004 Page 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on August, 2018. Sincerely, Tanya M. Hamilton

Enclosures:

1. Evaluation of the Proposed Changes to the Emergency Plan Emergency Action Level Scheme
2. Proposed Harris Nuclear Plant Emergency Action Level Technical Bases Document Changes, EP-EAL (Clean)
3. Proposed Harris Nuclear Plant Emergency Action Level Technical Bases Document Changes, EP-EAL (Markup)
4. Calculation for Containment Radiation EAL Threshold Values
5. Calculation for Radiation Monitor Readings for Core Uncovery during Refueling
6. Proposed Harris Nuclear Plant Emergency Action Level Wallchart Changes (Markup) cc:

J. Zeiler, NRC Sr. Resident Inspector, HNP W. L. Cox, III, Section Chief, N.C. DHSR M. Barillas, NRC Project Manager, HNP NRC Regional Administrator, Region II

U.S. Nuclear Regulatory Commission Serial HNP-18-004, Enclosure 1 SERIAL HNP-18-004 ENCLOSURE 1 EVALUATION OF THE PROPOSED CHANGES TO THE EMERGENCY PLAN EMERGENCY ACTION LEVEL SCHEME SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-63

U.S. Nuclear Regulatory Commission Page 1 of 15 Serial HNP-18-004, Enclosure 1 1.0

SUMMARY

DESCRIPTION In accordance with the provisions of 10 CFR 50.90, Duke Energy Progress, LLC (Duke Energy), is proposing three changes to the Shearon Harris Nuclear Power Plant, Unit 1 (HNP), Emergency Plan Emergency Action Level (EAL) scheme. The changes are associated with the Fission Product Barrier (FPB) Degradation EAL thresholds and the Cold Shutdown/Refueling System Malfunction EAL thresholds. Duke Energy proposes the following three changes.

1. Hot Operating Mode Loss of Reactor Coolant System Threshold: The HNP EAL Technical Bases Document, EP-EAL, includes declarations during hot operating modes (Modes 1 through 4) based on FPB integrity described in Table F-1, Fission Product Barrier Loss/Potential Loss Matrix and Bases. EP-EAL, Table F-1, includes a condition for determining Loss of Reactor Coolant System (RCS) based on a noble gas radiation monitor (REM-1LT-3502A-SA) used for containment leak detection. The EP-EAL basis description states that the listed threshold readings assume the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with normal operating concentrations (i.e., within Technical Specifications) into the containment atmosphere. The design of the noble gas monitor does not support this basis statement. Instantaneous release of all reactor coolant would trigger a safety injection (SI) and would isolate containment, which isolates flow to the radiation monitor rendering it unavailable. The current threshold does not adversely impact EAL declaration since a safety injection signal would alert decision makers at HNP of a RCS loss. Duke Energy proposes a change to the 'Loss of RCS loss' EAL that is more appropriate for an instantaneous release of all reactor coolant.
2. Hot Operating Mode Loss of Fuel Clad and Containment Thresholds: The EP-EAL, Table F-1, includes a Loss of Fuel Clad Barrier (FC) declaration and a Potential Loss of Containment (CNMT) Barrier declaration based on CNMT radiation levels as measured at the Containment High-Range Radiation Monitors (CHRRMs), RM-1CR-3589SA or RM-1CR-3590SB. The existing threshold values were developed in accordance with the methodology described in NUREG-0654/FEMA-REP-1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," November 1980 (Reference 6.5). The methodology approved for use at HNP is Nuclear Energy Institute (NEI) 99-01 Revision 6, Development of Emergency Action Levels for Non-Passive Reactors," as identified in Reference 6.2.

Duke Energy proposes a change to the CHRRM threshold values for 'Loss of FC' and 'Potential Loss of CNMT' in Table F-1 to values developed in accordance with NEI 99-01, Revision 6, methodology.

3. Cold Shutdown Loss of RCS Inventory Threshold: The EP-EAL includes declarations for core uncovery during cold shutdown and refueling modes under EALs CS1.3 and CG1.2 that are based upon CNMT radiation levels as measured at the CHRRMs. The current threshold value for EALs CS1.3 and CG1.2 is above radiation levels anticipated at the CHRRMs if core uncovery were to occur. In addition, due to the CHRRMs location and distance from the core, the CHRRMs are not ideal for monitoring loss of water shielding from the reactor cavity due to conditions leading to core uncovery. Alternate thresholds remain available to ensure core uncovery would result in the appropriate EAL declaration. However, establishing an appropriate method for monitoring radiation levels

U.S. Nuclear Regulatory Commission Page 2 of 15 Serial HNP-18-004, Enclosure 1 in CNMT would restore diversity for declarations associated with core uncovery. Thus, Duke Energy proposes a change in the method for measuring CMNT radiation levels from a method utilizing the CHRRMs to a method more capable of determining expected CNMT radiation levels in the event of core uncovery during cold shutdown or refueling. 2.0 DETAILED DESCRIPTION 2.1 Hot Operating Mode Loss of RCS Threshold 2.1.1 System Design and Operation As part of the Emergency Plan, Duke Energy maintains the ability to systematically declare EALs per the pre-planned scheme described in EP-EAL at HNP. The purpose of the EP-EAL document is to provide an explanation and rationale for each EAL. Decision makers use this document as a technical reference in support of EAL interpretation. The EAL scheme includes Category F for FPB degradation. This EAL category is applicable in hot operating modes (RCS temperature > 200°F, which is applicable to Modes 1 through 4) and represent threats to the defense-in-depth design concept that precludes the release of highly radioactive fission products to the environment. This EAL scheme evaluates threats to the FC, RCS, and CNMT. Assessment of each FPBs Loss or Potential Loss is performed per the Fission Product Barrier Threshold Matrix, Table F-1. Loss and Potential Loss signify the relative damage and threat of damage to a FPB. Loss means the FPB no longer assures containment of radioactive materials. Potential Loss means integrity of the FPB is threatened and could be lost if conditions continue to degrade. The number of FPBs that are lost or potentially lost determine the appropriate emergency classification level per the following criteria: Alert: Any loss or any potential loss of either FC or RCS Site Area Emergency (SAE): Loss or potential loss of any two barriers General Emergency (GE): Loss of any two barriers and loss or potential loss of third barrier Table F-1 includes thresholds for determining Loss of RCS based on the readings from the containment leak detection noble gas monitor, REM-1LT-3502A-SA. The purpose of this monitor is to identify a RCS release into CNMT with reactor coolant noble gas and iodine inventory at normal operating levels. Because of the very high fuel clad integrity, only small amounts of noble gases would be dissolved in the primary coolant. Thus, the REM-1LT-3502A-SA threshold value is lower than thresholds used to identify Loss of FC. 2.1.2 Current Emergency Plan Requirements The current Emergency Plan EAL scheme would declare a Loss of RCS whenever REM-1LT-3502A-SA detected noble gas concentrations greater than 8.3E-3 microcuries per milliliter (Ci/ml). This threshold value would lead to an Alert declaration, since at this point in an accident, the RCS would be considered lost, but FC and CNMT would not be impacted. This value is consistent with a significant leak of RCS with reactor coolant activity equal to Technical Specification limits.

U.S. Nuclear Regulatory Commission Page 3 of 15 Serial HNP-18-004, Enclosure 1 2.1.3 Reason for Proposed Change The current EAL threshold for determining Loss of RCS should be based on the expected radiation readings if there were an instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory into CNMT with reactor coolant activity equal to Technical Specification limits. This is the method specified per NEI 99-01, Revision 6 (Reference 6.3), which is the current EAL methodology approved by the NRC for use at HNP. However, the design of REM-1LT-3502A-SA does not support operation during an instantaneous loss of all RCS inventory. The inlet to this radiation monitor is equipped with a CNMT isolation valve designed to close on SI initiation. A SI would be expected with instantaneous loss of all RCS, rendering this monitor unavailable. Additionally, the current threshold values are a single number, meant to be representative of a wide array of events over a wide array of time periods. Duke Energy has identified that expected conditions in CNMT can vary greatly as an accident progresses, with dosage in CNMT changing relative to time from the accident. Thus, Duke Energy proposes using multiple thresholds for determining Loss of RCS based on the time from initiation of the accident. While not required, multiple thresholds will enhance the overall accuracy of the scheme and reduce the possibility of unnecessary declarations. 2.1.4 Description of Proposed Change Duke Energy proposes implementing a new method for declaring Loss of RCS in cases where no FC damage is present. Instead of using REM-1LT-3502A-SA, Duke Energy will use the Containment Ventilation Isolation (CVI) radiation monitors, as these monitors are located inside containment, at the operating deck level around the refueling cavity, and will continuously monitor containment radiation levels in the event of a safety injection. The thresholds used will be consistent with a loss of all RCS inventory assuming that reactor coolant activity equals Technical Specification allowable limits. This method is also consistent with the Reference 6.3 specified threshold of, Containment Radiation monitor reading greater than (site specific value). Thus, the current threshold in the Fission Product Barrier Matrix Table F-1 for Loss of RCS will be changed as shown below: Current threshold for declaring Loss of RCS: Containment Leak Detection Monitor Noble Gas (REM-1LT-3502A-SA) > 8.3E-3 Ci/ml. Proposed threshold for declaring Loss of RCS: A Containment Ventilation Isolation Radiation Monitor (RM-1CR-3561A-SA, RM-1CR-3561B-SB, RM-1CR-3561C-SA, or RM-1CR-3561D-SB): > 1.37E+03 mR/hr [millirem per hour] from T=0 hr to T=1 hr, > 1.12E+03 mR/hr from T>1 hr to T=2 hrs, > 6.35E+02 mR/hr from T>2 hrs to T=8 hrs > 1.37E+02 mR/hr from T> 8 hrs. The proposed changes to the EAL threshold description and bases are shown in Enclosure 2 (clean version of EP-EAL that incorporates changes) and Enclosure 3 (redline and strikeout version of EP-EAL changes). Enclosure 6 shows the proposed changes to the HNP EAL Wallcharts.

U.S. Nuclear Regulatory Commission Page 4 of 15 Serial HNP-18-004, Enclosure 1 The proposed EAL threshold is within the CVI radiation monitor's design operating range of 101 - 107 mR/hr (Reference 6.11). 2.2 Hot Operating Mode Loss of Fuel Clad and Containment Thresholds 2.2.1 System Design and Operation The HNP EAL scheme includes Category F for declaring FPB degradation. This EAL scheme includes the Fission Product Barrier Threshold Matrix, Table F-1, which is a matrix of thresholds for determining event classification based on conditions affecting FPBs. This matrix includes thresholds for determining FPB Loss and Potential Loss based on CNMT radiation levels. Two of these thresholds include Loss of FC declared when CHRRMs register radiation levels above 150 Rem per hour (R/hr), and Potential Loss of CNMT declared when CHRRMs register radiation levels above 600 R/hr. 2.2.2 Current Emergency Plan Requirements The current Emergency Plan EAL scheme requires a SAE declared when the CHRRMs identify radiation levels over 150 R/hr and a GE declared when radiation levels are over 600 R/hr. Note, a declaration of Loss of RCS is assumed, since without a loss of RCS it would not be possible to achieve the FC damage required to drive radiation levels up to 150 R/hr or 600 R/hr. Thus, 150 R/hr would result in a classification of a loss of two barriers, RCS and FC, and an SAE would be declared. The 600 R/hr radiation level may only occur after the loss of RCS and FC, and adds a Potential Loss of CNMT declaration, resulting in a GE. The current CHRRM threshold values were developed in accordance with the methodology described in NUREG-0654 (Reference 6.5). 2.2.3 Reason for Proposed Change The current licensing basis for the HNP EAL scheme is NEI 99-01, Revision 6 (Reference 6.3), as approved by the NRC in a letter dated April 13, 2016 (Reference 6.2). The EAL scheme transitioned from a NUREG-0654 (Reference 6.5) based scheme to an NEI 99-01, "Methodology for Development of Emergency Action Levels," Revision 5, based scheme following NRC approval in a letter dated April 25, 2010 (Reference 6.1). The CHRRM thresholds contained within the current EAL scheme are based upon calculation methodologies used for the NUREG-0654 (Reference 6.5) based scheme instead of NEI 99-01, Revision 6 (Reference 6.3), EAL scheme guidance. Duke Energy proposes changing the CHRRM threshold values in Table F-1 to values developed in accordance with the methodology described in NEI 99-01, Revision 6 (Reference 6.3), to correct this condition. Additionally, the current threshold values are a single number, meant to be representative of a wide array of events over a wide array of time periods. Duke Energy has identified that expected conditions in CNMT can vary greatly as an accident progresses. The dosage at the CHRRMs for a given amount of FC damage will change relative to time from the accident. Thus, Duke Energy proposes using multiple thresholds for determining a SAE or GE at the CHRRMs based on the time from initiation of the accident. While not required, multiple thresholds will enhance the overall accuracy of the scheme and reduce the possibility of unnecessary declarations.

U.S. Nuclear Regulatory Commission Page 5 of 15 Serial HNP-18-004, Enclosure 1 2.2.4 Description of the Proposed Change The current thresholds in the Fission Product Barrier Matrix Table F-1 for CNMT Radiation / RCS Activity will be revised as described below: Current threshold for declaring Loss of FC: > 150 R/hr at the CHRRMs Future threshold for declaring Loss of FC: > 130 R/hr at the CHRRMs from T=0 hr to T=1 hr, > 110 R/hr at the CHRRMs from T >1 hrs to T=2 hrs, > 70 R/hr at the CHRRMs from T >2 hrs to T=8 hrs, > 21 R/hr at the CHRRMs for T > 8 hrs. Current threshold for declaring Potential Loss of CNMT: > 600 R/hr at the CHRRMs Future threshold for declaring Potential Loss of CNMT: > 2360 R/hr at the CHRRMs from T=0 hr to T=1 hr, > 2000 R/hr at the CHRRMs from T >1 hr to T=2 hrs, > 1300 R/hr at the CHRRMs from T >2 hrs to T=8 hrs, > 390 R/hr at the CHRRMs for T > 8 hrs. The proposed EAL thresholds are within the CHRRM's design operating range of 100 - 108 R/hr (Reference 6.11). The proposed changes to the EAL threshold descriptions and bases are shown in Enclosure 2 (clean version of EP-EAL that incorporates changes) and Enclosure 3 (redline and strikeout version of EP-EAL changes). Enclosure 6 shows the proposed changes to the HNP EAL Wallcharts. 2.3 Cold Shutdown Loss of RCS Inventory Threshold 2.3.1 System Design and Operation The EAL scheme includes Category C for declarations based on the status of safety system functions associated with cold shutdown (Mode 5) or refueling/defueled. This EAL category represents the performance capabilities of malfunctioning systems with consideration given to RCS integrity, CNMT closure, and fuel clad integrity for the operating mode. Subcategory 1 pertains to RCS level, as reactor vessel or RCS water level is directly related to the status of adequate core cooling and fuel clad integrity in cold operating modes. CS1.3 and CG1.2 are used to declare a SAE and a GE based on site conditions that include indications of RCS level being degraded to the point of core uncovery. These EALs use multiple, redundant indications for declaring core uncovery, including the associated sump or tank level, erratic source range monitor indication, and CNMT radiation levels as measured at the CHRRMs. 2.3.2 Current Emergency Plan Requirements CS1.3 and CG1.2 use CNMT radiation levels as an indicator, with the threshold being CHRRM readings greater than 10,000 R/hr. From the EAL basis description, In the Refueling Mode, 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 indications on installed area radiation monitors (RM-1CR-3589SA or RM-1CR-3590SB). If these radiation monitors reach and exceed 10,000 R/hr, a loss of inventory with potential to uncover the core is likely to have occurred. The cited

U.S. Nuclear Regulatory Commission Page 6 of 15 Serial HNP-18-004, Enclosure 1 reference in EP-EAL for this threshold is AOP-31-BD, Loss of Refueling Cavity Integrity-Basis Document. 2.3.3 Reason for Proposed Change AOP-031-BD states, The dose rate to personnel on the edge of the Refueling Cavity is estimated to be as high as 10,000 R/hr with only six inches of one assembly extended above the water. The CHRRMs do not measure radiation levels at the edge of the refueling cavity, as their location is some distance away within CNMT. Thus, it was determined that the threshold listed in the associated EALs does not have an appropriate technical basis to serve as prompt indication of core uncovery. Also, due to the CHRRMs location and distance from the core, the CHRRMs are not ideal for monitoring loss of water shielding from the reactor cavity. The proposed change will implement a more appropriate method and threshold for monitoring CNMT radiation levels for detecting a loss of refueling cavity level. 2.3.4 Description of Proposed Change Duke Energy proposes changing the method for measuring CNMT radiation levels from a method utilizing the CHRRMs to a method more capable of monitoring CNMT radiation levels based on loss of refueling cavity level. Duke Energy proposes utilizing the CVI radiation monitors, as these monitors have improved sensitivity and the location of these monitors allows them to detect significant changes in the loss of water shielding from the reactor cavity, which can be used to detect conditions leading to core uncovery. The threshold proposed is a CVI radiation monitor value greater than 2.6E+04 mR/hr; however, single calculated radiation monitor reading does not represent core uncovery for all event scenarios due to multiple variables that can affect the source term (such as number of fuel assemblies in the core, fuel burn up in each assembly, and time after shutdown). The proposed threshold of 2.6E+04 mR/hr for the CVI radiation monitors will be low enough to be a valid indication of fuel uncovery, but high enough so that inadvertent classifications are not caused by maintenance events or other conditions that may increase radiation levels without a coincident loss of reactor vessel water level. Current threshold for declaring fuel uncovery per Category C: Containment radiation > 10,000 R/hr (RM-1CR-3589-SA or RM-1CR-3590-SB) Future threshold for declaring fuel uncovery per Category C: A Containment Ventilation Isolation Radiation Monitor > 2.6E+04 mR/hr (RM-1CR-3561A-SA, RM-1CR-3561B-SB, RM-1CR-3561C-SA, or RM-1CR-3561D-SB). The proposed EAL threshold is within the CVI radiation monitor's design operating range of 101 - 107 mR/hr (Reference 6.11). The proposed changes to the EAL threshold descriptions and bases are shown in Enclosure 2 (clean version of EP-EAL that incorporates changes) and (redline and strikeout version of EP-EAL changes). Enclosure 6 shows the proposed changes to the HNP EAL Wallcharts.

U.S. Nuclear Regulatory Commission Page 7 of 15 Serial HNP-18-004, Enclosure 1

3.0 TECHNICAL EVALUATION

3.1 Hot Operating Mode Loss of RCS Threshold The proposed method and threshold value for declaring a Loss of RCS (Alert) per HNP EP-EAL, Table F-1, Fission Product Barrier Matrix, Category C for CNMT Radiation / RCS Activity, was developed in accordance with Reference 6.3, Table 9-F-3, PWR EAL Fission Product Barrier Table, Thresholds for LOSS or POTENITIAL LOSS of Barriers, RCS Activity / CNMT Radiation, Loss 3.A. The Table 9-F-3 entry for PWR RCS Barrier Thresholds, RCS Activity / CNMT Radiation, Loss 3.A states the following: The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold 3.A since it indicates a loss of the RCS Barrier only. The CVI radiation monitors are already considered equipment important to Emergency Preparedness, as they are already used in the HNP EAL scheme for declaring initiating conditions related to abnormal radiation levels caused by an irradiated fuel event (Category R, Subcategory 2 per the HNP EAL scheme). The proposed thresholds will provide an indication of an instantaneous release of all reactor coolant mass with reactor coolant activity equal to Technical Specification allowable limits (with no FC damage). Duke Energy performed a calculation for the containment radiation EAL threshold values in accordance with the guidance specified in Reference 6.3, which is provided in Enclosure 4 of this submittal. Source Term for Thresholds Duke Energy used NUREG-1940, Revision 0, RASCAL 4: Description of Models and Methods, December 2012 (Reference 6.6), to develop the source terms used for the analysis. Guidance contained in Reference 6.6 is considered representative of the wide spectrum of possible events that establish the planning basis of emergency preparedness and provides radiological consequence assessment methods that are acceptable to the NRC. Additionally, the source term used to develop the HNP effluent EAL thresholds and in the Unified RASCAL Interface/Radiological Assessment System for Consequence Analysis (URI/RASCAL) dose assessment model is from Reference 6.6. Thus, Reference 6.6 is consistent with NRC guidance and with other source term bases used within the HNP Emergency Preparedness Program. Consistent with the graphs in Reference 6.6, the instantaneous release of the RCS to the containment is assumed to occur one hour after the damage event / reactor scram to account for damage progression, dispersion of activity and decay of the very short half-life isotopes. lists the HNP calculation that equates CHRRM readings to CVI radiation monitor readings for RM-1CR-3561A-SA using the ratio of 12.2 R/hr CVI to 17.5 R/hr CHRRM or approximately 70%. RM-1CR-3561A-SA is located on the inner wall of containment at the end of the refueling cavity, and is furthest away from the centerline of the reactor core. RM-1CR-3561B-SB and RM-1CR-3561C-SA are located on the inboard side of the 'A' Steam Generator Missile shield, which is adjacent to the Reactor Refueling Cavity. RM-1CR-3561D-SB is located on the inboard side of the 'C' Steam Generator Missile shield which is adjacent to the Reactor Refueling Cavity. NUREG/BR-0150, "Response Technical Manual (RTM-96)," Volume 1,

U.S. Nuclear Regulatory Commission Page 8 of 15 Serial HNP-18-004, Enclosure 1 Revision 4, Method A.4, "Evaluation of Containment Radiation," establishes a basis to use a radiation monitor for a core damage assessment as: "Confirm that the containment radiation monitor "sees" more than 50% of the shaded area shown in Fig. A-3 (PWR)." Based upon plant drawings, all four CVI radiation monitors satisfy the NUREG/BR-0150 Fig. A-3 (PWR) guidelines. All four CVI radiation monitors would be immersed within the same plume and will "see" a similar volume of the containment building; therefore, all four CVI radiation monitors are equivalent and can be used for EAL Classification. Conclusion Utilizing the CVI radiation monitors at the proposed thresholds for Loss of RCS declaration is consistent with the intent and methodology prescribed by NEI 99-01, Revision 6 (Reference 6.3), Table 9-F-3, PWR EAL Fission Product Barrier Table, Thresholds for LOSS or POTENTIAL LOSS of Barriers. The thresholds were derived using source terms per Reference 6.6. This method provides the level of redundancy and diversity in the HNP EAL scheme, Table F-1, Fission Product Barrier Matrix, that is prescribed from the guidance of Reference 6.3. 3.2 Hot Operating Mode Loss of Fuel Clad and Containment Thresholds The proposed values for declaration of Loss of FC (SAE) and Potential Loss of CNMT (GE) were developed in accordance with the methodology described in NEI 99-01, Revision 6 (Reference 6.3), Table 9-F-3, PWR EAL Fission Product Barrier Table, Thresholds for LOSS or POTENITIAL LOSS of Barriers. Loss of Fuel Clad (SAE) Barrier The new value for Loss of FC corresponds to the Reference 6.3, Table 9-F-3 entry for PWR Fuel Clad Barrier Thresholds, RCS Activity / CNMT Radiation, Loss 3.A. This table entry specifies the following: The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300Ci/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier. The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold 3.A since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the emergency classification level to a Site Area Emergency. Duke Energy performed a calculation for the containment radiation EAL threshold values in accordance with the guidance specified in Reference 6.3, which is provided in Enclosure 4 of this submittal. The FC FPB threshold value is based on an instantaneous release of reactor coolant into the CNMT at the percentage (%) of FC damage equivalent to 300 Ci/cubic centimeter (cc) dose equivalent I-131 RCS activity. Enclosure 4 identifies that 300Ci/cc is equivalent to 1.08% FC damage. This % FC damage value is multiplied as a ratio to the expected CNMT radiation reading for 100% fuel clad damage to establish the FC FPB threshold value in R/hr.

U.S. Nuclear Regulatory Commission Page 9 of 15 Serial HNP-18-004, Enclosure 1 Potential Loss of CNMT Barrier The new value for Potential Loss of CNMT corresponds to the Reference 6.3, Table 9-F-3 entry for PWR Containment Barrier Thresholds, RCS Activity / Containment Radiation, Potential Loss 3.A. This table entry specifies the following: The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds. NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment, which would then escalate the emergency classification level to a General Emergency. NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, provides the basis for using the 20% fuel cladding failure value. Unless there is a site-specific analysis justifying a different value, the reading should be determined assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with 20% fuel clad failure into the containment atmosphere. Duke Energy performed a calculation for the containment radiation EAL threshold values in accordance with the guidance specified in Reference 6.3, which is provided in Enclosure 4 of this submittal. The CNMT FPB threshold value is based on an instantaneous release of reactor coolant into the CNMT at an equivalent of 20% FC damage. The 20% FC damage value is multiplied as a ratio to the expected CNMT radiation reading for 100% FC damage to establish the CNMT FBP threshold value in R/hr. Enclosure 4 identifies that 300Ci/cc is equivalent to approximately 1.08% FC damage. Therefore, the ratio between the radiation reading proposed for FC damage and the expected CNMT radiation reading for 20% FC damage is approximately 1.08/20. Source Term for Thresholds Duke Energy used Reference 6.6 to develop the source terms used for the analysis. Guidance contained in Reference 6.6 is considered representative of the wide spectrum of possible events that establish the planning basis of emergency preparedness and provides radiological consequence assessment methods that are acceptable to the NRC. Additionally, the source term used to develop the HNP effluent EAL thresholds and in the Unified RASCAL Interface/Radiological Assessment System for Consequence Analysis (URI/RASCAL) dose assessment model is from Reference 6.6. Thus, Reference 6.6 is consistent with NRC guidance and with other source term bases used within the HNP Emergency Preparedness Program. Consistent with the graphs in Reference 6.6, the instantaneous release of the RCS to the containment is assumed to occur one hour after the damage event / reactor scram to account for damage progression, dispersion of activity and decay of the very short half-life isotopes.

U.S. Nuclear Regulatory Commission Page 10 of 15 Serial HNP-18-004, Enclosure 1 Conclusion The proposed values for declaration of Loss of FC and Potential Loss of CNMT' developed in accordance with the methodology established in NEI 99-01, Revision 6 (Reference 6.3), Table 9-F-3, PWR EAL Fission Product Barrier Table, Thresholds for LOSS or POTENITIAL LOSS of Barriers, use source terms derived in accordance with Reference 6.6. This method is consistent with NRC-approved guidance for deriving the specified EAL threshold values. Thus, these threshold values are appropriate for use in classifying and declaring initiating conditions affecting the sites FPBs. 3.3 Cold Shutdown Loss of RCS Inventory Threshold The current basis for declaring core uncovery during cold shutdown with the reactor vessel head removed utilizes the CHRRMs with a threshold of 10,000 R/hr. Duke Energy proposes replacing this method with a method that utilizes the CVI radiation monitors, using a reading of greater than 2.6E+04 mR/hr as the threshold. The CVI radiation monitors are positioned to be able to detect a loss of refueling cavity level. The use of the CVI radiation monitor value of 2.6E+04 mR/hr as the threshold will allow decision makers at HNP to distinguish normal CNMT radiation level fluctuations from a significant issue with the potential to lead to core uncovery. The proposed threshold will provide an indication of core uncovery. Duke Energy performed a calculation for the proposed radiation monitor reading for core uncovery during refueling in accordance with the guidance specified in Reference 6.3, which is provided in Enclosure 5 of this submittal. Source Term for Thresholds Duke Energy used NUREG-1940, Revision 0, RASCAL 4: Description of Models and Methods, December 2012 (Reference 6.6), to develop the source terms used for the analysis. Guidance contained in Reference 6.6 is considered representative of the wide spectrum of possible events that establish the planning basis of emergency preparedness and provides radiological consequence assessment methods that are acceptable to the NRC. Additionally, the source term used to develop the HNP effluent EAL thresholds and in the Unified RASCAL Interface/Radiological Assessment System for Consequence Analysis (URI/RASCAL) dose assessment model is from Reference 6.6. Thus, Reference 6.6 is consistent with NRC guidance and with other source term bases used within the HNP Emergency Preparedness Program. As described in Enclosure 5, the CVI radiation monitors do not have a direct line of site to the core. Therefore, the threshold value is calculated by backscatter from the containment dome. Design inputs for containment geometry, radiation monitor detector location, source materials weight and geometry, source geometry and receptor geometry were used to calculate results for each of the CVI radiation monitors. Based upon monitor accuracy/readability, human factors, and the similarity of results between monitors, the CS1.3 and CG1.2 EAL thresholds are established at a value greater than 2.6E+04 mR/hr. A single calculated radiation monitor reading does not represent core uncovery for all event scenarios due to multiple variables that can affect the source term (such as number of fuel assemblies in the core, fuel burn up in each assembly, and time after shutdown). The proposed threshold of a reading greater than 2.6E+04 mR/hr on the CVI radiation monitors will be low enough to be a valid indication of fuel uncovery, but high enough so that inadvertent classifications are not caused by maintenance events or other conditions that may increase radiation levels without a coincident loss of reactor vessel water level.

U.S. Nuclear Regulatory Commission Page 11 of 15 Serial HNP-18-004, Enclosure 1 This proposed method for declaring core uncovery during cold shutdown is consistent with the guidance issued per Reference 6.3, Table C-1, Recognition Category C Initiating Condition Matrix, for declaring SAE and GE (CS1 and CG1). The EAL Developer Notes state, As water level in the reactor vessel lowers, the dose rate above the core will increase. Enter a site-specific radiation monitor that could be used to detect core uncovery and the associated site-specific value indicative of core uncovery. The Duke Energy proposed method will both utilize a site-specific radiation monitor and a site-specific value, and will allow the site to identify conditions indicative of core uncovery. This method of determining elevated risk of core uncovery is consistent with current EAL methodology for declaring CS1.3 and CG1.2. Currently, the 10,000 R/hr is credited as indicative of a loss of inventory with potential to uncover the core. An alternate method for declaring CS1.3 and CG1.2 is UNPLANNED increase in any C-1 sump or tank of sufficient magnitude to indicate core uncovery. Conclusion Thus, the proposed method and threshold value will properly classify and declare conditions that indicate core uncovery during a refueling shutdown with the reactor vessel head removed. This change is consistent with the current EAL scheme and will improve the means to detect core uncovery during a refueling shutdown.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria Requirements for the Emergency Plans EAL Scheme: The regulations in 10 CFR 50.54(q) provide direction to licensees seeking to revise their Emergency Plan. The requirements related to nuclear power plant Emergency Plans are provided in the standards in 10 CFR 50.47, "Emergency Plans," and the requirements of Appendix E, "Emergency Planning and Preparedness for Production and Utilization Facilities," to 10 CFR 50. Paragraph (a)(1) of 10 CFR 50.47 states in part that, no initial operating license for a nuclear power reactor will be issued unless a finding is made by the NRC that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency. Section 50.47 establishes standards that onsite and offsite emergency response plans must meet for the NRC staff to make such a positive finding. One of these standards, 10 CFR 50.47(b)(4), stipulates that Emergency Plans include a standard emergency classification and action level scheme. 10 CFR 50.47(b)(4) states, "A standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures." Table B17-1, Conformance with QA Regulatory Guides and Industry Standards of the Duke Energy Quality Assurance Program Document DUKE-QAPD-001 -A-, Amendment 43, confirms HNP will follow a format for emergency procedures in accordance with 10 CFR 50, Appendix E. 10 CFR 50 Appendix E, Section IV. Content of Emergency Plans, Item B, Assessment Actions states:

U.S. Nuclear Regulatory Commission Page 12 of 15 Serial HNP-18-004, Enclosure 1

1. "The means to be used for determining the magnitude of, and for continually assessing the impact of, the release of radioactive materials shall be described, including emergency action levels that are to be used as criteria for determining the need for notification and participation of local and State agencies, the Commission, and other Federal agencies, and the emergency action levels that are to be used for determining when and what type of protective measures should be considered within and outside the site boundary to protect health and safety. The emergency action levels shall be based on in-plant conditions and instrumentation in addition to onsite and offsite monitoring.

The emergency action levels shall be based on in-plant conditions and instrumentation in addition to onsite and offsite monitoring

2. "A licensee desiring to change its entire emergency action level scheme shall submit an application for an amendment to its license and receive NRC approval before implementing the change. Licensees shall follow the change process in 10 CFR 50.54(q) for all other emergency action level changes."

NRC Regulatory Guide 1.101, Revision 4, endorsed Nuclear Management and Resources Council, Inc./National Environmental Studies Project (NUMARC/NESP)-007, Revision 2, issued in January 1992, and NEI 99-01, Revision 4, EAL guidance as acceptable alternatives to the guidance provided in NUREG-0654 for development of EALs to comply with 10 CFR 50.47 and Appendix E to 10 CFR Part 50. NRC RIS 2005-02, Revision 1, also discusses that a change in an EAL scheme to incorporate the improvements provided in NUMARC/NESP-007 or NEI 99-01 would not decrease the overall effectiveness of the Emergency Plan. However, due to the potential safety significance of the change, the change needs prior NRC review and approval. In a letter dated March 28, 2013 (Reference 6.4), the NRC staff concluded that the guidance contained in NEI 99-01, Revision 6, provides an acceptable method to develop an EAL scheme in accordance with the requirements of Appendix E to 10 CFR Part 50. In a letter dated April 13, 2016 (Reference 6.2), the NRC staff approved the HNP EAL scheme change to implement Reference 6.3 guidance, which is the basis for the current HNP EAL scheme. As evaluated per Section 3.0 of this submittal, proposed changes to the HNP Emergency Plans EAL scheme will restore or enhance compliance to the applicable regulations and guidance. 4.2 Precedent Other operating plants have submitted LARs to address deficiencies in their EAL scheme. The following examples involve EAL changes that improved the licensees ability to declare a condition based upon expected plant conditions as described in the NEI 99-01 guidance: (1.) Vogtle Electric Generating Plant, Units 1 and 2: Southern Company determined that the range of the credited main steam radiation monitors in their Category R declaration scheme was insufficient to cover the entire range of the EAL thresholds and submitted a LAR that requested removing the affected monitors from their EAL scheme and relying on existing alternate means for making effluent declarations (Reference 6.7). The amendment was issued by the NRC in the Safety Evaluation (Reference 6.8) dated September 30, 2014. (2.) Prairie Island Nuclear Generating Plant, Units 1 and 2: Xcel Energy determined that the effluent threshold established for an Alert classification level in their Category R declaration scheme was not appropriate - the emergency classification level exceeded the radiation monitor's indicated range. They also determined that for the loss of fuel clad barrier, the RCS

U.S. Nuclear Regulatory Commission Page 13 of 15 Serial HNP-18-004, Enclosure 1 letdown line radiation monitor value used in their scheme was inconsistent with the intent of the loss criteria established in the NEI 99-01 guidance. A LAR was submitted to revise the radiation monitor value used for the effluent threshold and to remove the RCS letdown line monitor from the loss of fuel clad barrier EAL (Reference 6.9). The amendment was issued by the NRC in the Safety Evaluation (Reference 6.10) dated January 25, 2014. 4.3 Significant Hazards Consideration In accordance with the provisions of 10 CFR 50.90, Duke Energy Progress, LLC (Duke Energy), is proposing changes to the Shearon Harris Nuclear Power Plant, Unit 1 (HNP), Emergency Plan Emergency Action Level (EAL) scheme. Duke Energy has evaluated whether or not a significant hazards consideration (SHC) is warranted with the proposed amendment by addressing the three criterion set forth in 10 CFR 50.92(c) as discussed below. (1) Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? The proposed changes affect the HNP Emergency Plan EAL scheme and do not alter any of the requirements of the Operating License or the Technical Specifications. The proposed changes do not reduce the effectiveness of the HNP Emergency Plan or the HNP Emergency Response Organization. The proposed changes do not modify any plant equipment and do not impact any failure modes that could lead to an accident. Additionally, the proposed changes do not impact the consequence of any analyzed accident since the changes do not affect any equipment related to accident mitigation. Based on this discussion, the proposed amendment does not increase the probability or consequences of an accident previously evaluated. (2) Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed changes affect the HNP Emergency Plan EAL scheme and do not alter any of the requirements of the Operating License or the Technical Specifications. These changes do not modify any plant equipment and there is no impact on the capability of the existing equipment to perform their intended functions. No new failure modes are introduced by the proposed changes. The proposed amendment does not introduce any accident initiator or malfunctions that would cause a new or different kind of accident. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. (3) Does the proposed amendment involve a significant reduction in a margin of safety? These changes affect the HNP Emergency Plan EAL scheme and do not alter any of the requirements of the Operating License or the Technical Specifications. The proposed changes do not affect any of the assumptions used in the accident analysis, nor do they affect any operability requirements for equipment important to plant safety. Therefore, the proposed changes will not result in a significant reduction in the margin of safety.

U.S. Nuclear Regulatory Commission Page 14 of 15 Serial HNP-18-004, Enclosure 1 Based on the above, Duke Energy concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c). Accordingly, a finding of no significant hazards consideration is justified. 4.4 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed change in the HNP Emergency Plan, (2) operation of HNP will continue to be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

Duke Energy has determined that the proposed amendment would not change requirements with respect to use of a facility component located within the restricted area, as defined by 10 CFR 20, nor would it change inspection or surveillance requirements. Duke Energy has evaluated the proposed change and has determined that the change does not involve: I. A Significant Hazards Consideration, II. A significant change in the types or significant increase in the amounts of any effluent that may be released off site, or III. A significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. Letter from the NRC to Carolina Power & Light Company, Shearon Harris Nuclear Power Plant, Unit 1 - Changes to the Emergency Action Level Scheme (TAC No.

ME1227), dated April 25, 2010 (ADAMS Accession No. ML100610685)

2. Letter from the NRC to Duke Energy, Shearon Harris Nuclear Power Plant, Unit 1 -

Issuance of Amendment to Adopt Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors" (CAC No. MF6196), dated April 13, 2016 (ADAMS Accession Number ML16057A838)

3. NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors, dated November 2012 (ADAMS Accession Number ML12326A805)
4. Letter from NRC to NEI, U.S. Nuclear Regulatory Commission Review and Endorsement of NEI 99-01, Revision 6, Dated November, 2012, dated March 28, 2013 (ADAMS Accession Number ML12346A463)

U.S. Nuclear Regulatory Commission Page 15 of 15 Serial HNP-18-004, Enclosure 1

5. NRC and Federal Emergency Management Agency, NUREG-0654/FEMA-REP-1, Revision 1, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," dated November 30, 1980 (ADAMS Accession No. ML040420012)
6. NRC, NUREG-1940, Revision 0, RASCAL 4: Description of Models and Methods, dated December 2012 (ADAMS Accession No. ML13031A448)
7. Letter from Southern Nuclear Operating Company, Inc. to NRC, License Amendment Request to Revise the Vogtle Electric Generating Plant Emergency Plan, dated August 20, 2013 (ADAMS Accession No. ML13233A112)
8. Letter from NRC to Southern Nuclear Operating Company, Inc., Vogtle Electric Generating Plant, Units 1 and 2 - Issuance of Amendments to the Emergency Plan (TAC Nos. MF2594 and MF2595), dated September 30, 2014 (ADAMS Accession No. ML14170A911)
9. Letter from Xcel Energy to the NRC, License Amendment Request (LAR) to Revise Emergency Plan (EP) Emergency Action Levels (EALs): RA1.2 and Fuel Clad Barrier Loss Criteria," dated December 13, 2012 (ADAMS Accession No. ML14170A911)
10. Letter from the NRC to Xcel Energy, Prairie Island Nuclear Generating Plant, Units 1 and 2 - Issuance of Amendments Re: Emergency Plan Changes (TAC Nos. MF0379 and MF0380), dated January 25, 2014 (ADAMS Accession No. ML13270A279)
11. HNP Final Safety Analysis Report Update (FSAR), Table 12.3.4-1, Area Radiation Monitors, Amendment 61

U.S. Nuclear Regulatory Commission Serial HNP-18-004, Enclosure 2 SERIAL HNP-18-004 ENCLOSURE 2 PROPOSED HARRIS NUCLEAR PLANT EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT CHANGES, EP-EAL (CLEAN) SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-63

ATTACHMENT 1 EAL Bases EP-EAL Rev. XXX DRAFT Page 70 of 220 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting core decay heat removal capability EAL: CS1.3 Site Area Emergency RCS water level cannot be monitored for 30 min. (Note 1) AND Core uncovery is indicated by any of the following:

  • UNPLANNED increase in any Table C-1 sump or tank of sufficient magnitude to indicate core uncovery
  • A Containment Ventilation Isolation Radiation Monitor > 2.6E+04 mR/hr (RM-1CR-3561A-SA, RM-1CR-3561B-SB, RM-1CR-3561C-SA, or RM-1CR-3561D-SB)
  • Erratic source range monitor indication Note 1:

The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Table C-1 Sumps / Tanks

  • Containment sumps
  • PRT
  • RCDT
  • CCW surge tank
  • RAB sumps
  • RWST
  • RMWST
  • Recycle Holdup Tank Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Definition(s): 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.

ATTACHMENT 1 EAL Bases EP-EAL Rev. XXX DRAFT Page 71 of 220 Basis: In this EAL, all RCS water level indication would be unavailable for greater than 30 minutes, and the RCS inventory loss must be detected by indirect leakage indications. Sump level increases must be evaluated against other potential sources of leakage. If the make-up rate to the RCS unexplainably rises above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Sumps and tanks where RCS leakage may accumulate are listed in listed in Table C-1(ref. 1, 2). In the Refueling Mode, as water level in the reactor vessel lowers, the dose rate above the core will increase. The dose rate due to this loss of water shielding from the reactor cavity will result in indications on the Containment Ventilation Isolation (CVI) area radiation monitors. If these radiation monitors reach and exceed 2.6E+04 mR/hr, with RCS water level indication unavailable for greater than 30 minutes, then a loss of inventory with potential to uncover the core is likely to have occurred. Post-TMI accident studies indicated that the installed PWR nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations. This IC addresses a significant and prolonged loss of RCS inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS level cannot be restored, fuel damage is probable. The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. The inability to monitor RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS. EAL values should be determined with a margin sufficient to ensure that an accurate monitor reading is available. For example, an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor reading.

ATTACHMENT 1 EAL Bases EP-EAL Rev. XXX DRAFT Page 72 of 220 A single calculated radiation monitor reading does not represent core uncovery for all event scenarios due to multiple variables that can affect the source term (i.e., number of fuel assemblies in the core, fuel burn up in each assembly, and time after shutdown). Therefore, consideration has been given so that the threshold value of > 2.6E+04 mR/hr on the CVI radiation monitors will be low enough to be a valid indication of fuel uncovery, but high enough so that inadvertent classifications are not caused by maintenance events or other conditions that may increase radiation levels without a coincident loss of reactor vessel water level. HNP utilizes CVI area radiation monitors RM-1CR-3561A-SA, 1RM-1CR-3561B-SB, RM-1CR-3561C-SA, and RM-1CR-3561D-SB for evaluating conditions in containment. The sensitivity and location of these monitors allows them to detect significant changes in the loss of water shielding from the reactor cavity, which can be used to detect core uncovery. This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. Escalation of the emergency classification level would be via IC CG1 or RG1. HNP Basis Reference(s):

1. GP-001, Reactor Coolant System Fill and Vent Mode 5
2. GP-008, Draining the Reactor Coolant System
3. GP-009, Refueling Cavity Fill, Refueling and Drain of the Refueling Cavity Modes 5-6-5
4. CSD-EP-HNP-0101-06, Radiation Monitor Readings for Core Uncovery during Refueling
5. NEI 99-01 CS1

ATTACHMENT 1 EAL Bases EP-EAL Rev. XXX DRAFT Page 76 of 220 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting fuel clad integrity with containment challenged EAL: CG1.2 General Emergency RCS level cannot be monitored for 30 min. (Note 1) AND Core uncovery is indicated by any of the following:

  • UNPLANNED increase in any Table C-1 sump or tank of sufficient magnitude to indicate core uncovery
  • A Containment Ventilation Isolation Radiation Monitor > 2.6E+04 mR/hr (RM-1CR-3561A-SA, RM-1CR-3561B-SB, RM-1CR-3561C-SA, or RM-1CR-3561D-SB)
  • Erratic source range monitor indication AND Any Containment Challenge indication, Table C-2 Note 1:

The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 6: 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

  • Containment sumps
  • PRT
  • RCDT
  • CCW surge tank
  • RAB sumps
  • RWST
  • RMWST
  • Recycle Holdup Tank Table C-2 Containment Challenge Indications
  • CONTAINMENT CLOSURE not established (Note 6)
  • Containment hydrogen concentration 4%
  • UNPLANNED rise in Containment pressure

ATTACHMENT 1 EAL Bases EP-EAL Rev. XXX DRAFT Page 77 of 220 Mode Applicability: 5 - Cold Shutdown, 6 - Refueling Definition(s): CONTAINMENT CLOSURE - The procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. As applied to HNP, Containment Closure is established when containment penetration closure is established in accordance with Technical Specifications 3/4.9.4. 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. Basis: In this EAL, all RCS water level indication would be unavailable for greater than 30 minutes, and the RCS inventory loss must be detected by indirect leakage indications. Sump level increases must be evaluated against other potential sources of leakage. If the make-up rate to the RCS unexplainably rises above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Sumps and tanks where RCS leakage may accumulate are listed in listed in Table C-1 (ref. 1, 2). In the Refueling Mode, as water level in the reactor vessel lowers, the dose rate above the core will increase. The dose rate due to core uncover will result in indications on the Containment Ventilation Isolation (CVI) area radiation monitors. If these radiation monitors reach and exceed 2.6E+04 mR/hr, with RCS water level indication unavailable for greater than 30 minutes, then a loss of inventory with potential to uncover the core is likely to have occurred. Post-TMI accident studies indicated that the installed PWR nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.

ATTACHMENT 1 EAL Bases EP-EAL Rev. XXX DRAFT Page 78 of 220 Three conditions are associated with a challenge to containment integrity:

  • 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 the 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. An explosive mixture can be formed when hydrogen gas concentration in the containment atmosphere is greater than 4% by volume in the presence of oxygen.
  • Any unplanned increase in containment pressure in the Cold Shutdown or Refueling mode indicates a potential loss of containment closure capability. Unplanned containment pressure increases indicates containment closure cannot be assured and the containment cannot be relied upon as a barrier to fission product release.

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 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. 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.

ATTACHMENT 1 EAL Bases EP-EAL Rev. XXX DRAFT Page 79 of 220 The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. The inability to monitor RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS. EAL values should be determined with a margin sufficient to ensure that an accurate monitor reading is available. For example, an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. A single calculated radiation monitor reading does not represent core uncovery for all event scenarios due to multiple variables that can affect the source term (i.e., number of fuel assemblies in the core, fuel burn up in each assembly, and time after shutdown). Therefore, consideration has been given so that the threshold value of > 2.6E+04 mR/hr on the CVI radiation monitors will be low enough to be a valid indication of fuel uncovery, but high enough so that inadvertent classifications are not caused by maintenance events or other conditions that may increase radiation levels without a coincident loss of reactor vessel water level. HNP utilizes CVI area radiation monitors RM-1CR-3561A-SA, 1RM-1CR-3561B-SB, RM-1CR-3561C-SA, and RM-1CR-3561D-SB for evaluating conditions in containment. The sensitivity and location of these monitors allows them to detect significant changes in the loss of water shielding from the reactor cavity, which can be used to detect core uncovery. This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. HNP Basis Reference(s):

1. GP-001, Reactor Coolant System Fill and Vent Mode 5
2. GP-008, Draining the Reactor Coolant System
3. GP-009, Refueling Cavity Fill, Refueling and Drain of the Refueling Cavity Modes 5-6-5
4. CSD-EP-HNP-0101-06, Radiation Monitor Readings for Core Uncovery during Refueling
5. NEI 99-01 CG1

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases EP-EAL Rev. XXX DRAFT Page 177 of 220 Table F-1 Fission Product Barrier Threshold Matrix Fuel Clad (FC) Barrier Reactor Coolant System (RCS) Barrier Containment (CNMT) Barrier Category Loss Potential Loss Loss Potential Loss Loss Potential Loss A RCS or SG Tube Leakage None None

1. An automatic or manual ECCS (SI) actuation required by EITHER:
  • UNISOLABLE RCS leakage
  • SG tube RUPTURE
1. CSFST Integrity-RED Path entry conditions met
1. A leaking or RUPTURED SG is FAULTED outside of containment None B

Inadequate Heat Removal

1. CSFST Core Cooling-RED Path entry conditions met
1. CSFST Core Cooling-ORANGE PATH entry conditions met
2. CSFST Heat Sink-RED Path entry conditions met AND Heat sink is required None
1. CSFST Heat Sink-RED Path entry conditions met AND Heat sink is required None
1.

CSFST Core Cooling-RED Path entry conditions met AND Restoration procedures not effective within 15 min. (Note 1) C CNMT Radiation / RCS Activity

1. (RM-1CR-3589SA or RM-1CR-3590SB) > Table F-2 Column FC Barrier Loss
2. Dose equivalent I-131 coolant activity > 300 µCi/gm None
1. (RM-1CR-3561A-SA, RM-1CR-3561B-SB, RM-1CR-3561C-SA, or RM-1CR-3561D-SB) > Table F-2 Column RCS Barrier Loss None None
1. (RM-1CR-3589SA or RM-1CR-3590SB) > Table F-2 Column CNMT Potential Loss D

CNMT Integrity or Bypass None None None None

1. Containment isolation is required AND EITHER:
  • Containment integrity has been lost based on Emergency Coordinator judgment
  • UNISOLABLE pathway from Containment to the environment exists
2. Indications of RCS leakage outside of containment
1.

CSFST Containment-RED Path entry conditions met

2. Containment hydrogen concentration

> 4%

3. Containment pressure > 10 psig with < one full train of depressurization equipment operating (one CNMT spray pump and two CNMT fan coolers) per design for 15 min.

(Note 1) E EC Judgment

1. Any condition in the opinion of the Emergency Coordinator that indicates loss of the fuel clad barrier
1.

Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the fuel clad barrier

1.

Any condition in the opinion of the Emergency Coordinator that indicates loss of the RCS barrier

1.

Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the RCS barrier

1.

Any condition in the opinion of the Emergency Coordinator that indicates loss of the containment barrier

1.

Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the containment barrier Table F-2 Containment Radiation Time After S/D (Hours) FC Barrier Loss R/hr RCS Barrier Loss mR/hr CNMT Potential Loss R/hr 0 - 1 130 1.37E+03 2360 1 - 2 110 1.12E+03 2000 2 - 8 70 6.35E+02 1300 > 8 21 1.37E+02 390

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases EP-EAL Rev. XXX DRAFT Page 183 of 220 Barrier: Fuel Clad Category: C. CNMT Radiation / RCS Activity Degradation Threat: Loss Threshold:

1. RM-1CR-3589SA or RM-1CR-3590SB > Table F-2, Column FC Barrier Loss Table F-2 Containment Radiation Time After S/D (Hours)

FC Barrier Loss R/hr RCS Barrier Loss mR/hr CNMT Potential Loss R/hr 0 - 1 130 1.37E+03 2360 1 - 2 110 1.12E+03 2000 2 - 8 70 6.35E+02 1300 > 8 21 1.37E+02 390 Definition(s): None Basis: Containment radiation monitor readings greater than Table F-2, FC Barrier Loss indicate the release of reactor coolant, with elevated activity indicative of fuel damage, into the Containment. The radiation monitor reading is derived assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 µCi/cc dose equivalent I-131 into the Containment atmosphere (ref. 1). Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within technical specifications and are therefore indicative of fuel damage. RM-1CR-3589SA and RM-1CR-3590SB are the Containment High Range Monitors that provide indication of radiation levels in Containment during and after postulated accidents (ref. 2). Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases EP-EAL Rev. XXX DRAFT Page 184 of 220 The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold C.1 since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the ECL to a Site Area Emergency. HNP Basis Reference(s):

1. EP-EALCALC-HNP-1701, Containment Radiation EAL Threshold Values
2. DBD-304, Radiation Monitoring System and Gross Failed Fuel Monitor
3. NEI 99-01 CNMT Radiation / RCS Activity Fuel Clad Loss 3.A

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases EP-EAL Rev. XXX DRAFT Page 195 of 220 Barrier: Reactor Coolant System Category: C. CNMT Radiation/ RCS Activity Degradation Threat: Loss Threshold:

1. A Containment Ventilation Isolation Radiation Monitor (RM-1CR-3561A-SA, RM-1CR-3561B-SB, RM-1CR-3561C-SA, or RM-1CR-3561D-SB) > Table F-2, Column RCS Barrier Loss Table F-2 Containment Radiation Time After S/D (Hours)

FC Barrier Loss R/hr RCS Barrier Loss mR/hr CNMT Potential Loss R/hr 0 - 1 130 1.37E+03 2360 1 - 2 110 1.12E+03 2000 2 - 8 70 6.35E+02 1300 > 8 21 1.37E+02 390 Definition(s): None Basis: A Containment Ventilation Isolation radiation monitor reading greater than Table F-2, RCS Barrier Loss (ref. 1) indicates the release of reactor coolant into containment. The readings assume the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with normal operating concentrations (i.e., within Technical Specifications) into the Containment atmosphere. Because of the very high fuel clad integrity, only small amounts of noble gases would be dissolved in the primary coolant. The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold C.1 since it indicates a loss of the RCS Barrier only. There is no Potential Loss threshold associated with RCS Activity / Containment Radiation. HNP Basis Reference(s):

1. EP-EALCALC-HNP-1701, Containment Radiation EAL Threshold Values
2. DBD-304, Radiation Monitoring System and Gross Failed Fuel Monitor
3. NEI 99-01 CNMT Radiation / RCS Activity RCS Loss 3.A

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases EP-EAL Rev. XXX DRAFT Page 208 of 220 Barrier: Containment Category: C. CNMT Radiation/RCS Activity Degradation Threat: Potential Loss Threshold:

1. RM-1CR-3589SA or RM-1CR-3590SB > Table F-2, Column CNMT Potential Loss Table F-2 Containment Radiation Time After S/D (Hours)

FC Barrier Loss R/hr RCS Barrier Loss mR/hr CNMT Potential Loss R/hr 0 - 1 130 1.37E+03 2360 1 - 2 110 1.12E+03 2000 2 - 8 70 6.35E+02 1300 > 8 21 1.37E+02 390 Definition(s): None Basis: Containment radiation monitor readings on RM-1CR-3589SA or RM-1CR-3590SB > Table F-2 column CNMT Potential Loss indicate significant fuel damage well in excess of that required for loss of the RCS barrier and the Fuel Clad barrier (ref. 1). The readings are higher than that specified for Fuel Clad Loss C.1 and RCS Loss C.1. Containment radiation readings at or above the Containment barrier Potential Loss threshold, therefore, signify a loss of two fission product barriers and Potential Loss of the third, indicating the need to upgrade the emergency classification to a General Emergency. RM-1CR-3589SA and RM-1CR-3590SB are the Containment High Range Monitors that provide indication of radiation levels in Containment during and after postulated accidents (ref. 2). The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds.

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases EP-EAL Rev. XXX DRAFT Page 209 of 220 NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the ECL to a General Emergency. HNP Basis Reference(s):

1. EP-EALCALC-HNP-1701, Containment Radiation EAL Threshold Values
2. DBD-304, Radiation Monitoring System and Gross Failed Fuel Monitor
3. NEI 99-01 CNMT Radiation / RCS Activity Containment Potential Loss 3.A

U.S. Nuclear Regulatory Commission Serial HNP-18-004, Enclosure 3 SERIAL HNP-18-004 ENCLOSURE 3 PROPOSED HARRIS NUCLEAR PLANT EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT CHANGES, EP-EAL (MARKUP) SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-63

ATTACHMENT 1 EAL Bases EP-EAL Rev. 17 Page 70 of 219 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting core decay heat removal capability EAL: CS1.3 Site Area Emergency RCS water level cannot be monitored for. 30 min. Note 1 AND Core uncovery is indicated by any of the following: x UNPLANNED increase in any Table C-1 sump or tank of sufficient magnitude to indicate core uncovery x Containment radiation > 10,000 R/hr (RM-1CR-3589-SA or RM-1CR-3590-SB) x Erratic source range monitor indication Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Table C-1 Sumps / Tanks x Containment sumps x PRT x RCDT x CCW surge tank x RAB sumps x RWST x RMWST x Recycle Holdup Tank Mode Applicability: 5 - Cold Shutdown, 6 - Refueling Definition(s): 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. Basis: In this EAL, all RCS water level indication would be unavailable for greater than 30 minutes, and the RCS inventory loss must be detected by indirect leakage indications. Sump level increases must be evaluated against other potential sources of leakage. If the make-up rate to the RCS unexplainably rises above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Sumps and tanks where RCS leakage may accumulate are listed in listed in Table C-1(ref. 1, 2). Replace with: "A Containment Ventilation Isolation Radiation Monitor > 2.6E+04 mR/hr (RM-1CR-3561A-SA, RM-1CR-3561B-SB, RM-1CR-3561C-SA, or RM-1CR-3561D-SB)"

ATTACHMENT 1 EAL Bases EP-EAL Rev. 17 Page 71 of 219 In the Refueling Mode, 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 indications on installed area radiation monitors (RM-1CR-3589-SA or RM-1CR-3590-SB). If these radiation monitors reach and exceed 10,000 R/hr, a loss of inventory with potential to uncover the core is likely to have occurred (ref. 4). Post-TMI accident studies indicated that the installed PWR nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations. This IC addresses a significant and prolonged loss of RCS inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS level cannot be restored, fuel damage is probable. The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. The inability to monitor RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS. This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-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 level would be via IC CG1 or RG1 HNP Basis Reference(s):

1. GP-001, Reactor Coolant System Fill and Vent Mode 5
2. GP-008, Draining the Reactor Coolant System
3. GP-009, Refueling Cavity Fill, Refueling and Drain of the Refueling Cavity Modes 5-6-5
4. AOP-031-BD, Loss of Refueling Cavity Integrity-Basis Document
5. NEI 99-01 CS1 Add INSERT B Replace with INSERT A Replace with: "CSD-EP-HNP-0101-06, Radiation Monitor Readings for Core Uncovery during Refueling"

INSERT A: INSERT B: EAL values should be determined with a margin sufficient to ensure that an accurate monitor reading is available. For example, an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. A single calculated radiation monitor reading does not represent core uncovery for all event scenarios due to multiple variables that can affect the source term (i.e., number of fuel assemblies in the core, fuel burn up in each assembly, and time after shutdown). Therefore, consideration has been given so that the threshold value of > 2.6E+04 mR/hr on the CVI radiation monitors will be low enough to be a valid indication of fuel uncovery, but high enough so that inadvertent classifications are not caused by maintenance events or other conditions that may increase radiation levels without a coincident loss of reactor vessel water level. HNP utilizes CVI area radiation monitors RM-1CR-3561A-SA, RM-1CR-3561B-SB, RM-1CR-3561C-SA, and RM-1CR-3561D-SB for evaluating conditions in containment. The sensitivity and location of these monitors allows them to detect significant changes in the loss of water shielding from the reactor cavity, which can be used to detect core uncovery. The dose rate due to this loss of water shielding from the reactor cavity will result in indications on the Containment Ventilation Isolation (CVI) area radiation monitors. If these radiation monitors reach and exceed 2.6E+04 mR/hr, with RCS water level indication unavailable for greater than 30 minutes, then a loss of inventory with potential to uncover the core is likely to have occurred.

ATTACHMENT 1 EAL Bases EP-EAL Rev. 17 Page 75 of 219 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting fuel clad integrity with containment challenged EAL: CG1.2 General Emergency RCS level cannot be monitored for 30 min. (Note 1) AND Core uncovery is indicated by any of the following: x UNPLANNED increase in any Table C-1 sump or tank of sufficient magnitude to indicate core uncovery x Containment radiation > 10,000 R/hr (RM-1CR-3589-SA or RM-1CR-3590-SB) x Erratic source range monitor indication AND Any Containment Challenge indication, Table C-2 Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 6: 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 x Containment sumps x PRT x RCDT x CCW surge tank x RAB sumps x RWST x RMWST x Recycle Holdup Tank Table C-2 Containment Challenge Indications x CONTAINMENT CLOSURE not established (Note 6) x Containment hydrogen concentration 4% x UNPLANNED rise in Containment pressure Replace with: "A Containment Ventilation Isolation Radiation Monitor > 2.6E+04 mR/hr (RM-1CR-3561A-SA, RM-1CR-3561B-SB, RM-1CR-3561C-SA, or RM-1CR-3561D-SB)"

ATTACHMENT 1 EAL Bases EP-EAL Rev. 17 Page 76 of 219 Mode Applicability: 5 - Cold Shutdown, 6 - Refueling Definition(s): CONTAINMENT CLOSURE - The procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. As applied to HNP, Containment Closure is established when containment penetration closure is established in accordance with Technical Specifications 3/4.9.4. 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. Basis: In this EAL, all RCS water level indication would be unavailable for greater than 30 minutes, and the RCS inventory loss must be detected by indirect leakage indications. Sump level increases must be evaluated against other potential sources of leakage. If the make-up rate to the RCS unexplainably rises above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Sumps and tanks where RCS leakage may accumulate are listed in listed in Table C-1 (ref. 1, 2). In the Refueling Mode, 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 indications on installed area radiation monitors (RM-1CR-3589-SA or RM-1CR-3590-SB). If these radiation monitors reach and exceed 10,000 R/hr, a loss of inventory with potential to uncover the core is likely to have occurred (ref. 4). Post-TMI accident studies indicated that the installed PWR nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations. Three conditions are associated with a challenge to containment integrity: x CONTAINMENT CLOSURE is not established. x 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 the 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. An explosive mixture can be formed when hydrogen gas concentration in the containment atmosphere is greater than 4% by volume in the presence of oxygen. x Any unplanned increase in containment pressure in the Cold Shutdown or Refueling mode indicates a potential loss of containment closure capability. Unplanned containment pressure increases indicates containment closure cannot be assured and the containment cannot be relied upon as a barrier to fission product release. Add INSERT C

INSERT C: The dose rate due to core uncover will result in indications on the Containment Ventilation Isolation (CVI) area radiation monitors. If these radiation monitors reach and exceed 2.6E+04 mR/hr, with RCS water level indication unavailable for greater than 30 minutes, then a loss of inventory with potential to uncover the core is likely to have occurred.

ATTACHMENT 1 EAL Bases EP-EAL Rev. 17 Page 77 of 219 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 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. In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged. The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. The inability to monitor RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS. This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-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. HNP Basis Reference(s):

1. GP-001, Reactor Coolant System Fill and Vent Mode 5
2. GP-008, Draining the Reactor Coolant System
3. GP-009, Refueling Cavity Fill, Refueling and Drain of the Refueling Cavity Modes 5-6-5
4. AOP-031-BD, Loss of Refueling Cavity Integrity-Basis Document
5. NEI 99-01 CG1 Add INSERT D Replace with: "CSD-EP-HNP-0101-06, Radiation Monitor Readings for Core Uncovery during Refueling"

INSERT D: EAL values should be determined with a margin sufficient to ensure that an accurate monitor reading is available. For example, an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. A single calculated radiation monitor reading does not represent core uncovery for all event scenarios due to multiple variables that can affect the source term (i.e., number of fuel assemblies in the core, fuel burn up in each assembly, and time after shutdown). Therefore, consideration has been given so that the threshold value of > 2.6E+04 mR/hr on the CVI radiation monitors will be low enough to be a valid indication of fuel uncovery, but high enough so that inadvertent classifications are not caused by maintenance events or other conditions that may increase radiation levels without a coincident loss of reactor vessel water level. HNP utilizes CVI area radiation monitors RM-1CR-3561A-SA, 1RM-1CR-3561B-SB, RM-1CR-3561C-SA, and RM-1CR-3561D-SB for evaluating conditions in containment. The sensitivity and location of these monitors allows them to detect significant changes in the loss of water shielding from the reactor cavity, which can be used to detect core uncovery.

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases EP-EAL Rev. 17 Page 175 of 219 Table F-1 Fission Product Barrier Threshold Matrix Fuel Clad (FC) Barrier Reactor Coolant System (RCS) Barrier Containment (CNMT) Barrier Category Loss Potential Loss Loss Potential Loss Loss Potential Loss A RCS or SG Tube Leakage None None 1. An automatic or manual ECCS (SI) actuation required by EITHER: x UNISOLABLE RCS leakage x SG tube RUPTURE 1. Operation of a standby charging pump is required by EITHER: x UNISOLABLE RCS leakage x SG tube leakage 2. CSFST Integrity-RED Path entry conditions met

1. A leaking or RUPTURED SG is FAULTED outside of containment None B

Inadequate Heat Removal 1. CSFST Core Cooling-RED Path entry conditions met 1. CSFST Core Cooling-ORANGE PATH entry conditions met 2. CSFST Heat Sink-RED Path entry conditions met AND Heat sink is required None 1. CSFST Heat Sink-RED Path entry conditions met AND Heat sink is required None 1. CSFST Core Cooling-RED Path entry conditions met AND Restoration procedures not effective within 15 min. (Note 1) C CNMT Radiation / RCS Activity

1. Containment radiation

>150 R/hr (RM-1CR-3589-SA or RM-1CR-3590-SB) 2. Dose equivalent I-131 coolant activity > 300 Ci/gm None 1. Containment Leak Detection Monitor Noble Gas (REM-1LT-3502A-SA) > 8.3E-3 Ci/ml None None 1. Containment radiation >600 R/hr (RM-1CR-3589-SA or RM-1CR-3590-SB) D CNMT Integrity or Bypass None None None None 1. Containment isolation is required AND EITHER: x Containment integrity has been lost based on Emergency Coordinator judgment x UNISOLABLE pathway from Containment to the environment exists 2. Indications of RCS leakage outside of containment 1. CSFST Containment-RED Path entry conditions met

2.

Containment hydrogen concentration > 4%

3.

Containment pressure > 10 psig with < one full train of depressurization equipment operating (one CNMT spray pump and two CNMT fan coolers) per design for 15 min. (Note 1) E EC Judgment 1. Any condition in the opinion of the Emergency Coordinator that indicates loss of the fuel clad barrier 1. Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the fuel clad barrier 1. Any condition in the opinion of the Emergency Coordinator that indicates loss of the RCS barrier 1. Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the RCS barrier 1. Any condition in the opinion of the Emergency Coordinator that indicates loss of the containment barrier 1. Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the containment barrier Revise to "1. (RM-1CR-3589SA or RM-1CR-3590SB) > Table F-2 Column FC Barrier Loss" Revise to

1. " (RM-1CR-3561A-SA, RM-1CR-3561B-SB, RM-1CR-3561C-SA, or RM-1CR-3561D-SB) > Table F-2 Column RCS Barrier Loss" Revise to "1. (RM-1CR-3589SA or RM-1CR-3590SB) > Table F-2 Column CNMT Potential Loss" Add INSERT H

INSERT H: Table F-2 Containment Radiation Time After S/D (Hours) FC Barrier Loss R/hr RCS Barrier Loss mR/hr CNMT Potential Loss R/hr 0 - 1 130 1.37E+03 2360 1 - 2 110 1.12E+03 2000 2 - 8 70 6.35E+02 1300 > 8 21 1.37E+02 390

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases EP-EAL Rev. 17 Page 181 of 219 Barrier: Fuel Clad Category: C. CNMT Radiation / RCS Activity Degradation Threat: Loss Threshold:

1. Containment radiation >150 R/hr (RM-1CR-3589-SA or RM-1CR-3590-SB)

Definition(s): None Basis: Containment radiation monitor readings greater than 150.3 R/hr, rounded to 150 R/hr for readability, indicate the release of reactor coolant, with elevated activity indicative of fuel damage, into the Containment. The reading is derived assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 Ci/cc dose equivalent I-131 into the Containment atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within technical specifications and are therefore indicative of fuel damage (approximately 5% clad failure depending on core inventory and RCS volume). (ref. 1) RM-1CR-3589-SA and RM-1CR-3590-SB are the Containment High Range Monitors that provide indication of radiation levels in Containment during and after postulated accidents. The Alert alarms are set at 6.5 R/hr and the High alarms are set at 17.5 R/hr. (ref. 2, 3). The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300 Ci/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier. The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold C.1 since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the ECL to a Site Area Emergency. HNP Basis Reference(s):

1. Calculation 3-B-12-022, DHRAM-Response to a Fuel and RCS Breach
2. DBD-304, Radiation Monitoring System and Gross Failed Fuel Monitor
3. HP-500, Radiation Monitoring System Data Base Manual
4. NEI 99-01 CNMT Radiation / RCS Activity Fuel Clad Loss 3.A Replace with 'INSERT E' from the following page:

Revise: "EP-EALCALC-HNP-1701, Containment Radiation EAL Threshold Values" RM-1CR-3589SA and RM-1CR-3590SB 3.

INSERT E:

1. RM-1CR-3589SA or RM-1CR-3590SB > Table F-2, Column FC Barrier Loss Table F-2 Containment Radiation Time After S/D (Hours)

FC Barrier Loss R/hr RCS Barrier Loss mR/hr CNMT Potential Loss R/hr 0 - 1 130 1.37E+03 2360 1 - 2 110 1.12E+03 2000 2 - 8 70 6.35E+02 1300 > 8 21 1.37E+02 390 Definition(s): None Basis: Containment radiation monitor readings greater than Table F-2, FC Barrier Loss indicate the release of reactor coolant, with elevated activity indicative of fuel damage, into the Containment. The radiation monitor reading is derived assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 Ci/cc dose equivalent I-131 into the Containment atmosphere (ref. 1).

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases EP-EAL Rev. 17 Page 193 of 219 Barrier: Reactor Coolant System Category: C. CNMT Radiation/ RCS Activity Degradation Threat: Loss Threshold:

1. Containment Leak Detection Monitor Noble Gas (REM-1LT-3502A-SA)

> 8.3E-3 Ci/ml Definition(s): None Basis: Containment radiation monitor readings on REM-1LT-3502A-SA noble gas channel greater than 8.3E-3 Ci/ml (ref. 1) indicate the release of reactor coolant to the Containment. The readings assume the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with normal operating concentrations (i.e., within Technical Specifications) into the Containment atmosphere. Because of the very high fuel clad integrity, only small amounts of noble gases would be dissolved in the primary coolant. The Containment High Range Monitors (RM-1CR-3589-SA or RM-1CR-3590-SB) are bugged to read at least 1 R/hr and are not capable of detecting this radiation level (ref. 2, 3). The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold C.1 since it indicates a loss of the RCS Barrier only. There is no Potential Loss threshold associated with RCS Activity / Containment Radiation. HNP Basis Reference(s):

1. Calculation HNP-M/MECH-1074, Alternate Source Term Effect on REM-3502A Response to RCS Breach with Non-Failed Fuel
2. DBD-304, Radiation Monitoring System and Gross Failed Fuel Monitor
3. HP-500, Radiation Monitoring System Data Base Manual
4. NEI 99-01 CNMT Radiation / RCS Activity RCS Loss 3.A Revise:

"EP-EALCALC-HNP-1701, Containment Radiation EAL Threshold Values" Replace with 'INSERT F' from the following page: 3.

INSERT F:

1. A Containment Ventilation Isolation Radiation Monitor (RM-1CR-3561A-SA, RM-1CR-3561B-SB, RM-1CR-3561C-SA, or RM-1CR-3561D-SB) > Table F-2, Column RCS Barrier Loss Table F-2 Containment Radiation Time After S/D (Hours)

FC Barrier Loss R/hr RCS Barrier Loss mR/hr CNMT Potential Loss R/hr 0 - 1 130 1.37E+03 2360 1 - 2 110 1.12E+03 2000 2 - 8 70 6.35E+02 1300 > 8 21 1.37E+02 390 Definition(s): None Basis: A Containment Ventilation Isolation radiation monitor reading greater than Table F-2, RCS Barrier Loss (ref. 1) indicates the release of reactor coolant into containment.

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases EP-EAL Rev. 17 Page 206 of 219 Barrier: Containment Category: C. CNMT Radiation/RCS Activity Degradation Threat: Potential Loss Threshold: 1. Containment radiation > 600 R/hr (RM-1CR-3589-SA or RM-1CR-3590-SB) Definition(s): None Basis: Containment radiation monitor readings greater than 601.2 R/hr, rounded to 600 R/hr for readability, (ref. 1) indicate significant fuel damage well in excess of that required for loss of the RCS barrier and the Fuel Clad barrier. The readings are higher than that specified for Fuel Clad Loss C.1 and RCS Loss C.1. Containment radiation readings at or above the Containment barrier Potential Loss threshold, therefore, signify a loss of two fission product barriers and Potential Loss of the third, indicating the need to upgrade the emergency classification to a General Emergency. RM-1CR-3589SA and RM-1CR-3590SB are the Containment High Range Monitors that provide indication of radiation levels in Containment during and after postulated accidents. The Alert alarms are set at 6.5 R/hr and the High alarms are set at 17.5 R/hr. (ref. 2, 3). The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds. NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the ECL to a General Emergency. HNP Basis Reference(s):

1. Calculation 3-B-12-022 DHRAM, Response to a Fuel and RCS Breach
2. DBD-304, Radiation Monitoring System and Gross Failed Fuel Monitor
3. HP-500, Radiation Monitoring System Data Base Manual
4. NEI 99-01 CNMT Radiation / RCS Activity Containment Potential Loss 3.A Replace with 'INSERT G' from the following page:

Revise: "EP-EALCALC-HNP-1701, Containment Radiation EAL Threshold Values" RM-1CR-3589SA and RM-1CR-3590SB 3.

INSERT G:

1. RM-1CR-3589SA or RM-1CR-3590SB > Table F-2, Column CNMT Potential Loss Table F-2 Containment Radiation Time After S/D (Hours)

FC Barrier Loss R/hr RCS Barrier Loss mR/hr CNMT Potential Loss R/hr 0 - 1 130 1.37E+03 2360 1 - 2 110 1.12E+03 2000 2 - 8 70 6.35E+02 1300 > 8 21 1.37E+02 390 Definition(s): None Basis: Containment radiation monitor readings on RM-1CR-3589SA or RM-1CR-3590SB > Table F-2 column CNMT Potential Loss indicate significant fuel damage well in excess of that required for loss of the RCS barrier and the Fuel Clad barrier (ref. 1).

U.S. Nuclear Regulatory Commission Serial HNP-18-004, Enclosure 4 SERIAL HNP-18-004 ENCLOSURE 4 CALCULATION FOR CONTAINMENT RADIATION EAL THRESHOLD VALUES SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-63

....-..-i Harris Nuclear Plant (HNP) Containment Radiation EAL Threshold Values EP-EALCALC-HNP-1701 Revision 2 Revision 2 Author: Caryl D. Ingram Technical Reviewer: William P. Cerame EPM Reviewer: David A. Thompson CHP Revision 2 Author: 7/2/2018 Technical Reviewer: __:~=--:~=-=---~_:::.::::::: =: *=::-=-----~::* '.'===_:_7~/2/~20~1:::_8 7/2/2018

HNP EAL Technical Bases Calculations - Containment Radiation EAL Threshold Values EP-EALCALC-HNP-1701 Page 2 of 25 Revision 2 List of Affected Pages Calculation Number: EP-EALCALC-HNP-1701 Revision Number: 2 Body of Calculation (including appendices) Supporting Documents Rev. # Pages Revised Pages Deleted Pages Added Rev. # Type Pages Revised Pages Deleted Pages Added 0 1 - 15 1 1 - 15 1 - 21 2 1 - 21 1 - 25

HNP EAL Technical Bases Calculations - Containment Radiation EAL Threshold Values EP-EALCALC-HNP-1701 Page 3 of 25 Revision 2 Revision Summary Form Calculation Number: EP-EALCALC-HNP-1701 Revision Number: 2 Revision Summary Revision Summary 0 Original Issue. 1 Provided interpolation of Containment High Range Radiation Monitor threshold values between 1 hr and 24 hrs. 2 Provided EAL Table F-2 RCS Barrier Loss containment radiation monitor threshold values using the Containment Ventilation Isolation monitors (CVIs) in lieu of the Containment High Range Radiation Monitor's minimum instrument range value of 5 R/hr.

HNP EAL Technical Bases Calculations - Containment Radiation EAL Threshold Values EP-EALCALC-HNP-1701 Page 4 of 25 Revision 2 Table of Contents

1. PURPOSE............................................................................................................................ 5
2. DEVELOPMENT METHODS AND BASES........................................................................... 5 2.1. Fuel Clad Loss.............................................................................................................. 5 2.2. Reactor Coolant System Loss...................................................................................... 6 2.3. Containment Potential Loss.......................................................................................... 7 2.4. Source Term................................................................................................................. 8 2.5. Decay Considerations................................................................................................... 9
3. DESIGN INPUTS................................................................................................................ 11 3.1. Constants and Conversion Factors............................................................................. 11 3.2. Plant Inputs................................................................................................................ 11 3.3. Source Term............................................................................................................... 11
4. CALCULATIONS................................................................................................................. 12 4.1. Fuel Clad Damage Estimate Based on 300 Ci/g DEI-131......................................... 12 4.2. Fission Product Barrier Thresholds............................................................................. 13 4.3. Extension and Interpolation of Data Points................................................................. 14
5. CONCLUSIONS.................................................................................................................. 15
6. REFERENCES.................................................................................................................... 15 Attachments, 300 Ci/g DEI-131 Equivalent Fuel Clad Damage.............................................. 17, CHRRM Fission Product Barrier Threshold Values............................................ 18, NUREG-1940 Figure 1-1 PWR Containment Monitor Response........................ 19, CHRRM Decay Time Interpolation..................................................................... 20, CVI Radiation Monitor RCS Fission Product Barrier Threshold Values................24

HNP EAL Technical Bases Calculations - Containment Radiation EAL Threshold Values EP-EALCALC-HNP-1701 Page 5 of 25 Revision 2

1.

PURPOSE The Shearon Harris Nuclear Plant (HNP) Emergency Action Level (EAL) Technical Bases Manual contains background information, event declaration thresholds, bases and references for the EAL and Fission Product Barrier (FPB) values used to implement the Nuclear Energy Institute (NEI) 99-01 Revision 6 EAL guidance. This calculation document provides additional technical detail specific to the derivation of the FPB Containment High Range Radiation Monitor (CHRRM) readings developed in accordance with the guidance in NEI 99-01 Revision 6. Documentation of the assumptions, calculations and results are provided for the values associated with the NEI 99-01 Revision 6 Table 9-F-3, PWR EAL Fission Product Barrier Table, thresholds listed below. NEI Fuel Clad Loss 3.A NEI Reactor Coolant Loss 3.A NEI Containment Potential Loss 3.A

2.

DEVELOPMENT METHODS AND BASES 2.1. Fuel Clad Loss Guidance Criteria Per NEI 99-01 Revision 6, this radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals 300 Ci/gm dose equivalent I-131. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the fuel clad barrier. The radiation monitor reading in this threshold is higher than that specified for RCS barrier loss threshold since it indicates a loss of both the fuel clad barrier and the RCS barrier. The reading should be determined assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory, with RCS radioactivity concentration equal to 300 Ci/gm dose equivalent I-131, into the primary containment atmosphere. HNP Basis The fuel clad FPB threshold value is based on an instantaneous release of reactor coolant into the containment at that percent fuel clad damage equivalent to 300 Ci/cc DEI-131 RCS activity. That percent fuel clad damage value is ratioed to a containment radiation reading for 100% fuel clad damage to establish the fuel clad FBP threshold value in R/hr.

HNP EAL Technical Bases Calculations - Containment Radiation EAL Threshold Values EP-EALCALC-HNP-1701 Page 6 of 25 Revision 2 2.2. Reactor Coolant System Loss Guidance Criteria Per NEI 99-01 Revision 6, this radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for fuel clad barrier loss threshold since it indicates a loss of the RCS barrier only. The reading should be determined assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory, with RCS activity at Technical Specification allowable limits, into the primary containment atmosphere. RCS activity at this level will typically result in primary containment radiation levels that can be more readily detected by primary containment radiation monitors, and more readily differentiated from those caused by piping or component shine sources. If desired, a plant may use a lesser value of RCS activity for determining this value. In some cases, the site-specific physical location and sensitivity of the containment radiation monitor(s) may be such that radiation from a cloud of released RCS gases cannot be distinguished from radiation emanating from piping and components containing elevated reactor coolant activity. If so, determine if an alternate indication is available. HNP Basis The HNP higher technical specification stabilized value for DEI-131 is 60 Ci/g. This activity would yield a containment radiation monitor reading approximately 5x lower than the fuel clad loss fission product barrier containment radiation reading equivalent to 300 Ci/g. NUREG-1940 Figure 1-1 provides estimates for standard plant containment radiation based on spiked RCS activity, which is slightly less than half the value obtained by the 300 Ci/g to 60 Ci/g DEI-131 ratio. NUREG-1940 Figure 1-1 models a spiked RCS activity that is lower than the RCS activity equivalent to 60 Ci/g DEI-131 described above (the NUREG-1940 graph is based on a release into containment of 100 times the non-noble gas fission products normally found in the coolant). This is the preferred value for the RCS loss threshold as it provides for a containment monitor escalation of approximately one decade between fission product barrier thresholds at the 1 hour point. NEI 99-01 guidance criteria allows the use of a lesser value for RCS activity (see guidance criteria section above). The HNP RCS FPB threshold value is based on NUREG-1940 standard plant containment radiation readings for an instantaneous release of spiked reactor coolant, which is lower than 60 Ci/g DEI-131, and is adjusted for the site specific power rating.

HNP EAL Technical Bases Calculations - Containment Radiation EAL Threshold Values EP-EALCALC-HNP-1701 Page 7 of 25 Revision 2 2.3. Containment Potential Loss Guidance Criteria Per NEI 99-01 Revision 6, this radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous fuel clad and RCS barrier loss thresholds. NUREG-1228 indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the RCS and fuel clad barriers. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the emergency classification level to a General Emergency. NUREG-1228 provides the basis for using the 20% fuel cladding failure value. Unless there is a site-specific analysis justifying a different value, the reading should be determined assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with 20% fuel clad failure into the primary containment atmosphere. HNP Basis The containment FPB threshold value is based on an instantaneous release of reactor coolant into the containment at an equivalent of 20% fuel clad damage. The 20% fuel clad damage value is ratioed to a containment radiation reading for 100% fuel clad damage to establish the containment FBP threshold value in R/hr.

HNP EAL Technical Bases Calculations - Containment Radiation EAL Threshold Values EP-EALCALC-HNP-1701 Page 8 of 25 Revision 2 2.4. Source Term Guidance Criteria NEI 99-01 does not specify a basis for source term activity or reduction factors. RG 1.183 provides assumptions for a LOCA used as a reference for FSAR design basis event analysis. Per RG 1.183 Section 1.1.4, Emergency Preparedness Applications: Requirements for emergency preparedness at nuclear power plants are set forth in 10 CFR 50.47, Emergency Plans. Additional requirements are set forth in Appendix E, Emergency Planning and Preparedness for Production and Utilization Facilities, to 10 CFR Part 50. The planning basis for many of these requirements was published in NUREG-0396, Planning Basis for the Development of State and Local Government Radiological Emergency Response Plans in Support of Light Water Nuclear Power Plants. This joint effort by the Environmental Protection Agency (EPA) and the NRC considered the principal characteristics (such as nuclides released and distances) likely to be involved for a spectrum of design basis and severe (core melt) accidents. No single accident scenario is the basis of the required preparedness. The objective of the planning is to provide public protection that would encompass a wide spectrum of possible events with a sufficient basis for extension of response efforts for unanticipated events. These requirements were issued after a long period of involvement by numerous stakeholders, including the Federal Emergency Management Agency, other Federal agencies, local and State governments (and in some cases, foreign governments), private citizens, utilities, and industry groups. Although the AST provided in this guide was based on a limited spectrum of severe accidents, the particular characteristics have been tailored specifically for DBA analysis use. The AST is not representative of the wide spectrum of possible events that make up the planning basis of emergency preparedness. Therefore, the AST is insufficient by itself as a basis for requesting relief from the emergency preparedness requirements of 10 CFR 50.47 and Appendix E to 10 CFR Part 50. Thus, RG 1.183 is not used as a source basis for the containment radiation monitor thresholds. Guidance contained in NUREG-1940 is considered representative of the wide spectrum of possible events that make up the planning basis of emergency preparedness and provides radiological consequence assessment methods which are acceptable to the NRC. Additionally, the source term used to develop the HNP effluent EAL thresholds and in the Unified RASCAL Interface/Radiological Assessment System for Consequence Analysis (URI/RASCAL) dose assessment model is from NUREG-1940. Thus, NUREG-1940 has been selected as a source term basis for the fission product barrier containment radiation thresholds for conformance to NRC guidance and consistency with other source term bases used within the HNP emergency preparedness program.

HNP EAL Technical Bases Calculations - Containment Radiation EAL Threshold Values EP-EALCALC-HNP-1701 Page 9 of 25 Revision 2 HNP Basis 2.4.1. The NUREG-1940 source term inputs used for the fission product barrier containment radiation thresholds are as follows: Fuel Clad Damage Equivalent to 300 Ci/g DEI-131 - NUREG-1940 Table 1-1 equilibrium core activity, in conjunction with the NUREG-1940 Table 1-5 non-noble gas release fraction, is used to develop the site specific iodine source term. Fuel Clad and Containment Barrier Thresholds - NUREG-1940 Figure 1-1 for cladding failure is used as a basis to establish this threshold. RCS Barrier Threshold - NUREG-1940 Figure 1-1 for spiked coolant is used as a basis to establish this threshold. 2.4.2. NUREG-1940 source term is based on a generic plant with a power rating of 3000 MWt. Per FSAR Section 1.1.5, Core Thermal Power, and Renewed License No. NPF-63 Amendment No. 152, Section 2.C.(1) Maximum Power Level, the HNP site specific source term is taken from the Unit 1 licensed core thermal power output of 2948 megawatts (consistent with the posting on the NRC website). The EAL thresholds are not based on the FSAR Chapter 15 rated power assumption of 2958, which includes 0.34% uncertainty. 2.4.3. Dose equivalent iodine 131 (DEI-131) dose conversion factors are derived from values provided in EPA-400-R-92-001. EPA-400 is used as the source for emergency preparedness related dose conversion factors and is the basis for the protective action guidelines. The DEI-131 dose conversion factors are not based on the FSAR Chapter 15 or other 10 CFR 20 reference sources as those are not reflective of the assumptions used within the EPA guidance for emergency preparedness use. 2.5. Decay Considerations Guidance Criteria Fission product barrier thresholds and their associated EALs are applicable only when the plant is in Power Operation, Hot Standby, Startup, or Hot Shutdown modes (known as the hot operating modes, or modes 1-4). Per NEI 99-01, the events for these thresholds correspond to an instantaneous release of all reactor coolant mass into the primary containment.

HNP EAL Technical Bases Calculations - Containment Radiation EAL Threshold Values EP-EALCALC-HNP-1701 Page 10 of 25 Revision 2 HNP Basis Consistent with the NUREG-1940 graphs, the instantaneous release of the RCS to the containment is assumed to occur one hour after the damage event / reactor scram to account for damage progression, dispersion of activity and decay of the very short half-life isotopes. NUREG-1940 provides additional containment radiation results for 24 hours after the damage event / reactor scram. Although NEI 99-01 does not require the use of multiple thresholds for the EAL, HNP will include multiple thresholds for consistency with fleet EALs.

HNP EAL Technical Bases Calculations - Containment Radiation EAL Threshold Values EP-EALCALC-HNP-1701 Page 11 of 25 Revision 2

3.

DESIGN INPUTS 3.1. Constants and Conversion Factors 3.1.1. 453.592 g per lbm water conversion factor 3.2. Plant Inputs 3.2.1. Rated Power Standard Plant (NUREG-1940 Section 1.2.4)............................................. 3,000 MWt HNP (FSAR 1.1.5)...................................................................................... 2,948 MWt 3.2.2. Reactor Coolant Mass (HNP-M/MECH-1020 Table 3) RCS mass...................................................................................... 4.41619 E+05 lbm 3.2.3. Standard Plant Containment Radiation Reading (NUREG-1940 Figure 1-1) Spray Off 100% fuel clad damage...................................... 60,000 R/hr (@ 1 hr after shutdown) 100% fuel clad damage.................................... 30,000 R/hr (@ 24 hr after shutdown) 100% spiked coolant.................................................. 50 R/hr (@ 1 hr after shutdown) 100% spiked coolant.................................................. 6 R/hr (@ 24 hr after shutdown) Spray On 100% fuel clad damage...................................... 12,000 R/hr (@ 1 hr after shutdown) 100% fuel clad damage...................................... 2,000 R/hr (@ 24 hr after shutdown) 100% spiked coolant.................................................... 2 R/hr (@ 1 hr after shutdown) 100% spiked coolant............................................... 0.2 R/hr (@ 24 hr after shutdown) 3.3. Source Term 3.3.1. Source Term Activity (NUREG-1940 Table 1-1) Core Activity (Ci/MWt) I-131 2.67E+04 I-132 3.88E+04 I-133 5.42E+04 I-134 5.98E+04 I-135 5.18E+04

HNP EAL Technical Bases Calculations - Containment Radiation EAL Threshold Values EP-EALCALC-HNP-1701 Page 12 of 25 Revision 2 3.3.2. Release Fractions - RFCore (NUREG-1940 Table 1-5) Non-Noble Gasses (I, Cs, Rb) - Fuel Clad Damage.................................... 0.05 (5%) 3.3.3. Iodine Dose Conversion Factors - DCF (EPA-400 Table 5-2) Rem/hr per Ci/cc I-131 1.3E+06 I-132 7.7E+03 I-133 2.2E+05 I-134 1.3E+03 I-135 3.8E+04

4.

CALCULATIONS 4.1. Fuel Clad Damage Estimate Based on 300 Ci/g DEI-131 4.1.1. 100% Core Activity Equivalent Reactor Coolant Iodine Concentrations 100% ( 

) = (

) x () x 106 () Coolant Activity (Ci/g) I-131 3.93E+05 I-132 5.71E+05 I-133 7.98E+05 I-134 8.80E+05 I-135 7.62E+05 Total 3.40E+06 4.1.2. 100% Fuel Clad Activity Equivalent Reactor Coolant Iodine Concentrations 100% ( 

) = 100% (
) x Coolant Activity (Ci/g)

I-131 1.96E+04 I-132 2.86E+04 I-133 3.99E+04 I-134 4.40E+04 I-135 3.81E+04 Total 1.70E+05

HNP EAL Technical Bases Calculations - Containment Radiation EAL Threshold Values EP-EALCALC-HNP-1701 Page 13 of 25 Revision 2 4.1.3. 100% Fuel Clad Activity Equivalent Reactor Coolant DEI-131 Concentrations 100% ( 

) = 100% (
) x The DEI-131 value for each iodine isotope is determined as follows: 
=

400 5 2 (/ /) 400 5 2 131(/ /) Coolant Activity (Ci/g) I-131 1.96E+04 I-132 1.69E+02 I-133 6.75E+03 I-134 4.40E+01 I-135 1.11E+03 Total 2.77E+04 4.1.4. % Fuel Clad Activity Equivalent Reactor Coolant at 300 Ci/g DEI-131 % = 300

100% ( 

)

300 Ci/ml DEI-131 = 1.08% Fuel Clad Damage See Attachment 1 for the spreadsheet calculations that develop the fuel clad source term activity and the % fuel clad damage estimates. 4.2. Fission Product Barrier Thresholds 4.2.1. Containment Potential Loss (20% Fuel Clad Damage) 20% (

) = 100% (

) x

x 20% Time After Shutdown Spray Off Containment Potential Loss Spray On Containment Potential Loss 1 hour 1.18E+04 R/hr 2.36E+03 R/hr 24 hours 5.90E+03 R/hr 3.93E+02 R/hr See Attachment 2 for the spreadsheet calculations that develop the 20% fuel clad damage containment radiation monitor reading.

HNP EAL Technical Bases Calculations - Containment Radiation EAL Threshold Values EP-EALCALC-HNP-1701 Page 14 of 25 Revision 2 4.2.2. Fuel Clad Loss (300 Ci/g DEI-131 Equivalent Clad Damage) (

) = 100% (

) x

x % Time After Shutdown Spray Off Fuel Clad Loss Spray On Fuel Clad Loss 1 hour 6.38E+02 R/hr 1.28E+02 R/hr 24 hours 3.19E+02 R/hr 2.13E+01 R/hr See Attachment 2 for the spreadsheet calculations that develop the 300 Ci/g DEI-131 equivalent fuel clad damage containment radiation monitor reading. 4.2.3. RCS Loss (Spiked Coolant) (

) = (

) x

Time After Shutdown Spray Off RCS Loss Spray On RCS Loss 1 hour 4.91E+01 R/hr 1.97+00 R/hr 24 hours 5.90E+00 R/hr 1.97E-01R/hr See Attachment 2 for the spreadsheet calculations that develop the spiked RCS containment radiation monitor reading. 4.3. Extension and Interpolation of Data Points The CHRM response to activity disbursed instantaneously and homogeneously throughout containment during a postulated LOCA event is influenced by multiple factors. Photon energy and abundance of each isotope are the major contributors with decay, daughter ingrowth, washout, settling and plate-out being additional factors when considering the effects of time upon the monitor reading from the onset. While development of decay curves with complex consideration of the above influences is possible, thresholds for the EAL fission product barriers consistent with the values provided at 1 and 24 hours in NUREG-1940 is desired. Thus, the derivation of a relative decay factor based on the 1 hour and 24 hour data points is the method chosen to interpolate containment monitor reading at hourly intervals out to 72 hours (the time at which the modes applicable to the fission product barrier matrix are no longer likely). See Attachment 4 for the equations and results for the derived containment high range radiation monitor hourly readings from 1 to 72 hours.

HNP EAL Technical Bases Calculations - Containment Radiation EAL Threshold Values EP-EALCALC-HNP-1701 Page 15 of 25 Revision 2

5.

CONCLUSIONS 5.1. 300 Ci/g DEI-131 is equivalent to 1.08% Fuel Clad Damage 5.2. Time dependent containment high range radiation monitor fission product barrier values both with and without sprays are provided in Attachment 4. 5.3. Table F-2 presents selected values at 1, 2, 8 and 24 hours after shutdown (rounded for readability) as they will appear on the EAL Wallchart (see below). Values for sprays on were chosen based on operational strategies at Harris Nuclear Plant that actuate sprays during adverse containment conditions. Table F-2 Containment Radiation (RM-1CR-3589SA, RM-1CR-3590SB) Time After S/D (Hours) FC Barrier Loss R/hr RCS Barrier Loss R/hr* CNMT Potential Loss R/hr 0 - 1 130 Refer to for CVI Radiation Monitor FPB Threshold 2360 1 - 2 110 2000 2 - 8 70 1300 > 8 21 390

  • Attachment 5 provides EAL Table F-2 RCS Barrier Loss containment radiation monitor threshold values using the Containment Ventilation Isolation radiation monitors (CVIs) in lieu of the Containment High Range Radiation Monitors.

HNP EAL Technical Bases Calculations - Containment Radiation EAL Threshold Values EP-EALCALC-HNP-1701 Page 16 of 25 Revision 2

6.

REFERENCES 6.1. NEI 99-01 Revision 6, Development of Emergency Action Levels for Non-Passive Reactors, September 2012 6.2. EPA-400-R-92-001, Manual of Protective action Guides and Protective Actions for Nuclear Incidents, May 1992 6.3. WCAP-14696-A, Westinghouse Owners Group Core Damage Assessment Guide, Revision 1 6.4. NUREG-1940, RASCAL 4: Description of Models and Methods, December 2012 6.5. NUREG-1228, Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents, October 1988 6.6. Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000 6.7. Harris Nuclear Plant FSAR 1.1.5 Core Thermal Power 6.8. Renewed License No. NPF-63 Amendment No. 152, Section 2.C.(1) Maximum Power Level 6.9. Technical Specifications, Harris Nuclear Plant, Revision 170 Section 3.4.8, RCS Specific Activity 6.10. HNP-M/MECH-1020, Reactor Vessel and Reactor Coolant System Water Volumes, Revision 3 6.11. EPM-601, Core Damage Assessment Technical Basis, Revision 2 300 Ci/g DEI-131 Equivalent Fuel Clad Damage EP-EALCALC-HNP-1701 Page 17 of 25 Revision 2 NUREG-1940 Table 1-1 Core Activity (Ci/MWt) HNP Core Activity (Ci) HNP RCS Activity (Ci/g 100% Core) HNP RCS Activity (Ci/g 100% Clad) EPA-400 Table 5-2 Iodine DCFs (rem per Ci/cc) DEI-131 ICF RCS Activity (Ci/g 100% Clad DEI-131) I-131 2.67E+04 7.87E+07 3.93E+05 1.96E+04 1.3E+06 1.00E+00 1.96E+04 I-132 3.88E+04 1.14E+08 5.71E+05 2.86E+04 7.7E+03 5.92E-03 1.69E+02 I-133 5.42E+04 1.60E+08 7.98E+05 3.99E+04 2.2E+05 1.69E-01 6.75E+03 I-134 5.98E+04 1.76E+08 8.80E+05 4.40E+04 1.3E+03 1.00E-03 4.40E+01 I-135 5.18E+04 1.53E+08 7.62E+05 3.81E+04 3.8E+04 2.92E-02 1.11E+03 Total 2.31E+05 6.82E+08 3.40E+06 1.70E+05 2.77E+04 Volume Conversion (g/lbm): 453.592 Rated Power (MWt): 2948 Non-Noble Gas RF (%): 5.0% RC Liquid Mass (lbm): 4.42E+05 RC Liquid Mass (g): 2.00E+08 Target DEI-131 (Ci/g): 3.00E+02 % Clad Damage: 1.08% CHRRM Fission Product Barrier Threshold Values EP-EALCALC-HNP-1701 Page 18 of 25 Revision 2 Reading for 300 Ci/cc RCS (Fuel Clad FPB R/hr) Reading for Spiked RCS (RCS FPB R/hr) Reading for 20% Clad Failure (Containment FPB in R/hr) 1 hour - Spray Off 6.38E+02 4.91E+01 1.18E+04 24 hours - Spray Off 3.19E+02 5.90E+00 5.90E+03 1 hour - Spray On 1.28E+02 1.97E+00 2.36E+03 24 hours - Spray On 2.13E+01 1.97E-01 3.93E+02 % Damage for 300 Ci/g DEI-131 RCS Activity: 1.08% Standard Plant (MWt): 3000 HNP Rated Power (MWt): 2948 NUREG-1940 Figure 1-1 Containment Rad Readings 1 hr spray off 100% Clad Failure (R/hr): 6.00E+04 24 hr spray off 100% Clad Failure (R/hr): 3.00E+04 1 hr spray on 100% Clad Failure (R/hr): 1.20E+04 24 hr spray on 100% Clad Failure (R/hr): 2.00E+03 1 hr spray off Spiked Coolant (R/hr): 5.00E+01 24 hr spray off Spiked Coolant (R/hr): 6.00E+00 1 hr spray on Spiked Coolant (R/hr): 2.00E+00 24 hr spray on Spiked Coolant (R/hr): 2.00E-01 1e+6 1e+5 .c: 1e+4 a:: - "O (l) "O <i5 1e+3

c ti)

C O> 1e+2 C ~ co (l) a:: ~ 1e+1 0 .:t:: C 0

E C
  • Je+O (l)

E -~ co C 0 1e-1 0 1e-2 1e-3 EP-EALCALC-HNP-1701 NUREG-1940 Figure 1-1 PWR Containment Monitor Response ~ § -- i-- - i-- H - ~ ~ ~ ~ ~ ~ ~ I- - § ~ = r-- Key to Spray Status and r-- µ --- Damage Amount = ~ Sprays Sprays ~ Off On

  • -* 100%

. 100% - = ~ f--< - ...

  • 10%

I I 10% - I- - I-I- 1% . 1% = 1h 24 h 1h 24 h 1h 24h 1h 24 h Normal Coolant Spiked Coolant Cladding Failure Core Melt Damage State and Time After Reactor Shutdown Figure 1-1 PWR containment monitor response 1-14 Page 19 of 25 Revision 2 CHRRM Decay Time Interpolation EP-EALCALC-HNP-1701 Page 20 of 25 Revision 2

1) The relative decay constants related to the containment radiation monitor results from NUREG-1940 Figure 1-1 are derived as follows:

= ln

Where: Rt Monitor Reading at time interval (t) Ri Initial Monitor Reading t time interval k decay constant Spray Off Ri (1 hr) Rt (24 hr) k Containment Potential Loss Threshold (R/hr) 1.18E+04 5.90E+03 2.89E-02 Fuel Clad Loss Threshold (R/hr) 6.38E+02 3.19E+02 2.89E-02 RCS Loss Threshold (R/hr) 4.91E+01 5.90E+00 8.83E-02 Spray On Ri (1 hr) Rt (24 hr) k Containment Potential Loss Threshold (R/hr) 2.36E+03 3.93E+02 7.47E-02 Fuel Clad Loss Threshold (R/hr) 1.28E+02 2.13E+01 7.47E-02 RCS Loss Threshold (R/hr) 1.97E+00 1.97E-01 9.59E-02 t (hours): 24 CHRRM Decay Time Interpolation EP-EALCALC-HNP-1701 Page 21 of 25 Revision 2

2) The containment radiation monitor readings for various times after shutdown are then determined as follows:

= x 1E-2 1E-1 1E+0 1E+1 1E+2 1E+3 1E+4 1E+5 1 6 12 18 24 30 36 42 48 54 60 66 72 Containment High Range Monitor (R/hr) Time After Shutdown (hours) Spray Off - CHRM Fission Product Barrier Thresholds Containment Potential Loss Fuel Clad Loss RCS Loss 1E-3 1E-2 1E-1 1E+0 1E+1 1E+2 1E+3 1E+4 1 6 12 18 24 30 36 42 48 54 60 66 72 Containment High Range Monitor (R/hr) Time After Shutdown (hours) Spray On - CHRM Fission Product Barrier Thresholds Containment Potential Loss Fuel Clad Loss RCS Loss CHRRM Decay Time Interpolation EP-EALCALC-HNP-1701 Page 22 of 25 Revision 2 Note: Values provided below are the calculation results given to illustrate trends. It is recognized that the low scale of the high range containment monitor is above the later values for the RCS loss fission product barrier and are not appropriate for use in event declaration. Spray Off Time After Shutdown (hours) Containment Potential Loss Threshold (R/hr) Fuel Clad Loss Threshold (R/hr) RCS Loss Threshold (R/hr) Time After Shutdown (hours) Containment Potential Loss Threshold (R/hr) Fuel Clad Loss Threshold (R/hr) RCS Loss Threshold (R/hr) 1 1.18E+04 6.38E+02 4.91E+01 37 4.05E+03 2.19E+02 1.87E+00 38 3.94E+03 2.13E+02 1.71E+00 2 1.11E+04 6.02E+02 4.12E+01 39 3.82E+03 2.07E+02 1.57E+00 3 1.08E+04 5.85E+02 3.77E+01 40 3.71E+03 2.01E+02 1.43E+00 4 1.05E+04 5.68E+02 3.45E+01 41 3.61E+03 1.95E+02 1.31E+00 5 1.02E+04 5.52E+02 3.16E+01 42 3.51E+03 1.90E+02 1.20E+00 6 9.92E+03 5.37E+02 2.89E+01 43 3.41E+03 1.84E+02 1.10E+00 7 9.63E+03 5.21E+02 2.65E+01 44 3.31E+03 1.79E+02 1.01E+00 8 9.36E+03 5.06E+02 2.42E+01 45 3.21E+03 1.74E+02 9.22E-01 9 9.09E+03 4.92E+02 2.22E+01 46 3.12E+03 1.69E+02 8.44E-01 10 8.83E+03 4.78E+02 2.03E+01 47 3.03E+03 1.64E+02 7.73E-01 11 8.58E+03 4.64E+02 1.86E+01 48 2.95E+03 1.60E+02 7.08E-01 12 8.34E+03 4.51E+02 1.70E+01 49 2.86E+03 1.55E+02 6.48E-01 13 8.10E+03 4.38E+02 1.56E+01 50 2.78E+03 1.51E+02 5.93E-01 14 7.87E+03 4.26E+02 1.43E+01 51 2.70E+03 1.46E+02 5.43E-01 15 7.65E+03 4.14E+02 1.31E+01 52 2.63E+03 1.42E+02 4.97E-01 16 7.43E+03 4.02E+02 1.20E+01 53 2.55E+03 1.38E+02 4.55E-01 17 7.22E+03 3.90E+02 1.09E+01 54 2.48E+03 1.34E+02 4.16E-01 18 7.01E+03 3.79E+02 1.00E+01 55 2.41E+03 1.30E+02 3.81E-01 19 6.81E+03 3.69E+02 9.17E+00 56 2.34E+03 1.27E+02 3.49E-01 20 6.62E+03 3.58E+02 8.40E+00 57 2.27E+03 1.23E+02 3.19E-01 21 6.43E+03 3.48E+02 7.69E+00 58 2.21E+03 1.19E+02 2.92E-01 22 6.25E+03 3.38E+02 7.04E+00 59 2.15E+03 1.16E+02 2.68E-01 23 6.07E+03 3.28E+02 6.44E+00 60 2.08E+03 1.13E+02 2.45E-01 24 5.90E+03 3.19E+02 5.90E+00 61 2.03E+03 1.10E+02 2.24E-01 25 5.73E+03 3.10E+02 5.40E+00 62 1.97E+03 1.06E+02 2.05E-01 26 5.57E+03 3.01E+02 4.94E+00 63 1.91E+03 1.03E+02 1.88E-01 27 5.41E+03 2.93E+02 4.52E+00 64 1.86E+03 1.00E+02 1.72E-01 28 5.25E+03 2.84E+02 4.14E+00 65 1.80E+03 9.76E+01 1.58E-01 29 5.10E+03 2.76E+02 3.79E+00 66 1.75E+03 9.48E+01 1.44E-01 30 4.96E+03 2.68E+02 3.47E+00 67 1.70E+03 9.21E+01 1.32E-01 31 4.82E+03 2.61E+02 3.18E+00 68 1.65E+03 8.95E+01 1.21E-01 32 4.68E+03 2.53E+02 2.91E+00 69 1.61E+03 8.70E+01 1.11E-01 33 4.55E+03 2.46E+02 2.66E+00 70 1.56E+03 8.45E+01 1.01E-01 34 4.42E+03 2.39E+02 2.44E+00 71 1.52E+03 8.21E+01 9.27E-02 35 4.29E+03 2.32E+02 2.23E+00 72 1.47E+03 7.98E+01 8.49E-02 36 4.17E+03 2.26E+02 2.04E+00 CHRRM Decay Time Interpolation EP-EALCALC-HNP-1701 Page 23 of 25 Revision 2 Spray On Time After Shutdown (hours) Containment Potential Loss Threshold (R/hr) Fuel Clad Loss Threshold (R/hr) RCS Loss Threshold (R/hr) Time After Shutdown (hours) Containment Potential Loss Threshold (R/hr) Fuel Clad Loss Threshold (R/hr) RCS Loss Threshold (R/hr) 1 2.36E+03 1.28E+02 1.97E+00 37 1.49E+02 8.06E+00 5.65E-02 38 1.38E+02 7.48E+00 5.13E-02 2 2.03E+03 1.10E+02 1.62E+00 39 1.28E+02 6.94E+00 4.66E-02 3 1.89E+03 1.02E+02 1.47E+00 40 1.19E+02 6.44E+00 4.23E-02 4 1.75E+03 9.47E+01 1.34E+00 41 1.10E+02 5.98E+00 3.85E-02 5 1.62E+03 8.79E+01 1.22E+00 42 1.03E+02 5.55E+00 3.49E-02 6 1.51E+03 8.15E+01 1.11E+00 43 9.52E+01 5.15E+00 3.18E-02 7 1.40E+03 7.57E+01 1.00E+00 44 8.83E+01 4.78E+00 2.88E-02 8 1.30E+03 7.02E+01 9.12E-01 45 8.20E+01 4.43E+00 2.62E-02 9 1.20E+03 6.52E+01 8.29E-01 46 7.61E+01 4.12E+00 2.38E-02 10 1.12E+03 6.05E+01 7.53E-01 47 7.06E+01 3.82E+00 2.16E-02 11 1.04E+03 5.61E+01 6.84E-01 48 6.55E+01 3.54E+00 1.97E-02 12 9.63E+02 5.21E+01 6.21E-01 49 6.08E+01 3.29E+00 1.79E-02 13 8.94E+02 4.83E+01 5.65E-01 50 5.64E+01 3.05E+00 1.62E-02 14 8.29E+02 4.49E+01 5.13E-01 51 5.24E+01 2.83E+00 1.47E-02 15 7.70E+02 4.16E+01 4.66E-01 52 4.86E+01 2.63E+00 1.34E-02 16 7.14E+02 3.86E+01 4.23E-01 53 4.51E+01 2.44E+00 1.22E-02 17 6.63E+02 3.59E+01 3.85E-01 54 4.19E+01 2.26E+00 1.11E-02 18 6.15E+02 3.33E+01 3.49E-01 55 3.88E+01 2.10E+00 1.00E-02 19 5.71E+02 3.09E+01 3.18E-01 56 3.61E+01 1.95E+00 9.12E-03 20 5.30E+02 2.87E+01 2.88E-01 57 3.35E+01 1.81E+00 8.29E-03 21 4.92E+02 2.66E+01 2.62E-01 58 3.11E+01 1.68E+00 7.53E-03 22 4.56E+02 2.47E+01 2.38E-01 59 2.88E+01 1.56E+00 6.84E-03 23 4.24E+02 2.29E+01 2.16E-01 60 2.67E+01 1.45E+00 6.21E-03 24 3.93E+02 2.13E+01 1.97E-01 61 2.48E+01 1.34E+00 5.65E-03 25 3.65E+02 1.97E+01 1.79E-01 62 2.30E+01 1.25E+00 5.13E-03 26 3.39E+02 1.83E+01 1.62E-01 63 2.14E+01 1.16E+00 4.66E-03 27 3.14E+02 1.70E+01 1.47E-01 64 1.98E+01 1.07E+00 4.23E-03 28 2.92E+02 1.58E+01 1.34E-01 65 1.84E+01 9.96E-01 3.85E-03 29 2.71E+02 1.46E+01 1.22E-01 66 1.71E+01 9.25E-01 3.49E-03 30 2.51E+02 1.36E+01 1.11E-01 67 1.59E+01 8.58E-01 3.18E-03 31 2.33E+02 1.26E+01 1.00E-01 68 1.47E+01 7.96E-01 2.88E-03 32 2.16E+02 1.17E+01 9.12E-02 69 1.37E+01 7.39E-01 2.62E-03 33 2.01E+02 1.09E+01 8.29E-02 70 1.27E+01 6.86E-01 2.38E-03 34 1.86E+02 1.01E+01 7.53E-02 71 1.18E+01 6.37E-01 2.16E-03 35 1.73E+02 9.36E+00 6.84E-02 72 1.09E+01 5.91E-01 1.97E-03 36 1.60E+02 8.68E+00 6.21E-02 CVI Radiation Monitor RCS Fission Product Barrier Threshold Values EP-EALCALC-HNP-1701 Page 24 of 25 Revision 2

1. Purpose The purpose of Attachment 5 is to provide EAL Table F-2 RCS Barrier Loss containment radiation monitor threshold values using the Containment Ventilation Isolation radiation monitors (CVIs) in lieu of the Containment High Range Radiation Monitors (CHRRMs) minimum instrument range value of 5 R/hr.

There are 4 CVIs (RM-1CR-3561A-SA, RM-1CR-3561B-SB, RM-1CR-3561C-SA, RM-1CR-3561D-SB) that are located at the operating deck level around the refueling cavity. The CVIs' primary advantage over the CHRRMs for application to the EAL Table F-2 RCS Barrier Loss threshold is that the CVIs have better low range sensitivity than the CHRRMs (101 - 107 mR/hr (CVIs) versus 100 - 108 R/hr (CHRRMs)) {3.1}.

2. Calculation Harris calculation TDR-RC-013 {3.2} provides a method to equate CHRRM readings to CVI radiation monitor readings for RM-1CR-3561A-SA using the ratio (12.2 R/hr CVI/17.5 R/hr CHRRM) (~70%). RM-1CR-3561A-SA is located on the inner wall of containment at the end of the refueling cavity, and is furthest away from the centerline of the reactor core. RM-1CR-3561B-SB and RM-1CR-3561C-SA are located on the inboard side of the 'A' Steam Generator Missile shield which is adjacent to the Reactor Refueling Cavity. RM-1CR-3561D-SB is located on the inboard side of the 'C' Steam Generator Missile shield which is adjacent to the Reactor Refueling Cavity. NUREG/BR-0150 {3.3}, Method A.4, "Evaluation of Containment Radiation" establishes a basis to use a radiation monitor for a core damage assessment as: "Confirm that the containment radiation monitor "sees" more than 50% of the shaded area shown in either Fig. A-3 (PWR) or Fig. A-4 (BWR)." Based upon plant drawings CAR-2166-G-453 {3.4} and CAR-2165-G-013 {3.5}, all four CVI radiation monitors satisfy the NUREG/BR-0150 Fig. A-3 (PWR) guidelines. All four CVI radiation monitors would be immersed within the same plume and will "see" a similar volume of the containment building, therefore all four CVI radiation monitors are equivalent and can be used for EAL Classification.

Applying the 12.2/17.5 ratio to the "Spray On" CHRRM time dependent readings (R/hr) shown on Attachment 4, page 23 of this calculation and multiplying by 1000 to convert R/hr to mR/hr results in the following CVI threshold values: Time After S/D (Hours) CHRRMs RCS Barrier Loss R/hr CVIs RCS Barrier Loss mR/hr 0 - 1 1.97 1.373E+03 1 - 2 1.62 1.129E+03 2 - 8 0.912 6.358E+02 > 8 0.197 1.373E+02 For example, 1.97 R/hr x 12.2/17.5 x 1000 mR/R = 1.373E+03 mR/hr CVI equivalent reading CVI Radiation Monitor RCS Fission Product Barrier Threshold Values EP-EALCALC-HNP-1701 Page 25 of 25 Revision 2 Truncating for use in EAL Table F-2 results in the following: Table F-2 Containment Radiation For FC Barrier Loss and CNTM Potential Loss (RM-1CR-3589SA, RM-1CR-3590SB) For RCS Barrier Loss (RM-1CR-3561A-SA, RM-1CR-3561B-SB, RM-1CR-3561C-SA, RM-1CR-3561D-SB) Time After S/D (Hours) FC Barrier Loss R/hr RCS Barrier Loss mR/hr CNMT Potential Loss R/hr 0 - 1 130 1.37E+03 2360 1 - 2 110 1.12E+03 2000 2 - 8 70 6.35E+02 1300 > 8 21 1.37E+02 390

3. References 3.1 Harris UFSAR, Table 12.3.4-1, "Area Radiation Monitors," Amendment 61.

3.2 Harris Calculation TDR-RC-013, "Determination of EAL Dose Rate for Monitor RM-1CR-3561A-SA," Rev. 0. 3.3 NUREG/BR-0150, Response Technical Manual (RTM-96), US Nuclear Regulatory Commission, Vol. 1, Rev. 4. 3.4 Drawing CAR-2166-G-453 (Fusion Drawing 6-G-0453), Instrument Location Arrangement, Unit 1, Plan El. 286', Revision 16. 3.5 Drawing CAR-2165-G-013 (Fusion Drawing 5-G-013), General Arrangement Containment Building Sections Sheet 1, Revision 17.

U.S. Nuclear Regulatory Commission Serial HNP-18-004, Enclosure 5 SERIAL HNP-18-004 ENCLOSURE 5 CALCULATION FOR RADIATION MONITOR READINGS FOR CORE UNCOVERY DURING REFUELING SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-63

NUCLEAR DEVELOPMENT / OPERATING FLEET / DECOMMISSIONED CSD-EP-HNP-0101-06 RADIATION MONITOR READINGS FOR CORE UNCOVERY DURING REFUELING REVISION 0 Effective Dates: N/A N/A N/A 06/28/2018 N/A Brunswick Catawba Crystal River Harris Lee N/A N/A N/A N/A N/A Levy McGuire Oconee Robinson NGO

Harris Nuclear Plant (HNP) Radiation Monitor Readings for Core Uncovery during Refueling CSD-EP-HNP-0101-06 Revision 0 Document Author: Scott McCain Technical Reviewer: William Cerame EP Reviewer: David A. Thompson CHP Document Author: 06/28/18 w~/J/] -* - Technical Reviewer: L,AA,a.mA, 06/28/18

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HNP EAL Technical Bases Calculation - CG1.2 and CS1.3 CSD-EP-HNP-0101-06 Page 3 of 30 Revision 0 Table of Contents 1 PURPOSE............................................................................................................................ 4 2 DEVELOPMENT METHOD AND BASES.............................................................................. 4 3 DESIGN INPUTS.................................................................................................................. 6 3.1 Constants and Conversion Factors............................................................................... 6 3.2 Plant Inputs.................................................................................................................. 6 3.3 Radiation Monitors........................................................................................................ 6 3.4 Source Term................................................................................................................. 7 4 CALCULATIONS................................................................................................................. 10 4.1 Initial Source Term (Reactor Core) Activity................................................................. 10 4.2 Source (Reactor Core) Volume................................................................................... 10 4.3 Source Materials Mixed Shielding Density.................................................................. 11 4.4 MicroShield Results.................................................................................................. 11 4.5 3561 Radiation Monitor Dose Rates for RCS Level at TOAF...................................... 12 5 CONCLUSION.................................................................................................................... 14 6 REFERENCES.................................................................................................................... 15 ATTACHMENTS, Source Volume and Effectiveness Density Determination.................................. 16, MicroShield Instructions for Source Density Determination............................... 17, MicroShield Calculation Report......................................................................... 19, Radiation Monitor Back-Scatter Determination................................................... 24, Radiation Monitor Back-Scatter Basis Reference............................................... 25

HNP EAL Technical Bases Calculation - CG1.2 and CS1.3 CSD-EP-HNP-0101-06 Page 4 of 30 Revision 0 1 PURPOSE The Shearon Harris Nuclear Plant (HNP) Emergency Action Level (EAL) Technical Bases Manual contains background information, event declaration thresholds, bases and references for the EAL and Fission Product Barrier (FPB) values used to implement the Nuclear Energy Institute (NEI) 99-01 Revision 6 EAL guidance. This calculation document provides additional technical detail specific to the derivation of cold shutdown radiation monitor readings in containment developed in accordance with the guidance in NEI 99-01 Revision 6. Documentation of the assumptions, calculations and results are provided for the HNP Cx1 series EAL values associated the NEI 99-01 Revision 6 EALs listed below. NEI EAL CS1.3 NEI EAL CG1.2 2 DEVELOPMENT METHOD AND BASES 2.1 CS1.3 Guidance Criteria The initiating condition for CS1 is Loss of RCS inventory affecting core decay heat removal capability. To be met, EAL CS1.3 requires two threshold conditions (inability to monitor reactor vessel RCS level and core uncovery indication). Per NEI 99-01 Revision 6, for EAL #3.b bullet one - As water level in the reactor vessel lowers, the dose rate above the core will increase. Enter a site-specific radiation monitor that could be used to detect core uncovery and the associated site-specific value indicative of core uncovery. It is recognized that the condition described by this IC may result in a radiation value beyond the operating or display range of the installed radiation monitor. In those cases, EAL values should be determined with a margin sufficient to ensure that an accurate monitor reading is available. For example, an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This provision notwithstanding, if the estimated/calculated monitor reading is greater than approximately 110% of the highest accurate monitor reading, then developers may choose not to include the monitor as an indication and identify an alternate EAL threshold. To further promote accurate classification, developers should consider if some combination of monitors could be specified in the EAL to build-in an appropriate level of corroboration between monitor readings into the classification assessment. HNP Bases The CS1.3 radiation monitor threshold is predicated on the loss of water above the core during refueling shutdown with the reactor vessel head removed. The monitors used do not have direct line of sight. Therefore, the threshold value will be calculated by backscatter from the containment dome.

HNP EAL Technical Bases Calculation - CG1.2 and CS1.3 CSD-EP-HNP-0101-06 Page 5 of 30 Revision 0 2.2 CG1.2 Guidance Criteria The initiating condition for CG1 is Loss of RCS inventory affecting fuel clad integrity with containment challenged. To be met, EAL CG1.2 requires three threshold conditions (inability to monitor reactor vessel RCS level, core uncovery indication and containment challenge indication). Per NEI 99-01 Rev. 6, for EAL #2.b bullet one - As water level in the reactor vessel lowers, the dose rate above the core will increase. Enter a site-specific radiation monitor that could be used to detect core uncovery and the associated site-specific value indicative of core uncovery. It is recognized that the condition described by this IC may result in a radiation value beyond the operating or display range of the installed radiation monitor. In those cases, EAL values should be determined with a margin sufficient to ensure that an accurate monitor reading is available. For example, an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This provision notwithstanding, if the estimated/calculated monitor reading is greater than approximately 110% of the highest accurate monitor reading, then developers may choose not to include the monitor as an indication and identify an alternate EAL threshold. To further promote accurate classification, developers should consider if some combination of monitors could be specified in the EAL to build-in an appropriate level of corroboration between monitor readings into the classification assessment. HNP Bases The CG1.2 radiation monitor threshold is predicated on the loss of water above the core during refueling shutdown with the reactor vessel head removed. The monitors used do not have direct line of sight. Therefore, the threshold value will be calculated by backscatter from the containment dome. 2.3 Event Conditions Guidance Criteria Per NEI 99-01 Revision 6, these ICs address a significant and prolonged loss of RCS inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. 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. HNP Bases A single calculated radiation monitor reading cannot represent the point of core uncovery for all event scenarios due to multiple variables that can affect the source term (such as number of fuel assemblies in the core, fuel burn up in each assembly and time after shutdown). Therefore, reasonable assumptions and inputs have been selected so that the calculated value will be low enough to be a valid indication of fuel uncovery, but high enough so that inadvertent classifications are not made due to maintenance events or other conditions which may increase radiation levels without a coincident loss of reactor vessel water level.

HNP EAL Technical Bases Calculation - CG1.2 and CS1.3 CSD-EP-HNP-0101-06 Page 6 of 30 Revision 0 3 DESIGN INPUTS 3.1 Constants and Conversion Factors 3.1.1 16.387 cc per in3 unit conversion factor 3.1.2 453.592 g per lbm unit conversion factor 3.2 Plant Inputs 3.2.1 Rated Power Standard Plant (NUREG-1940 Section 1.2.4)............................................. 3,000 MWt HNP (FSAR 1.1.5)...................................................................................... 2,948 MWt 3.2.2 Containment Geometry Reactor vessel head is removed. Top of Active Fuel (EAL technical basis manual).................................................. 249 Containment Spring Line height (Drawing CAR-2165-G-0013)............................. 376 Containment Dome radius (Drawing CAR-2165-G-0013)........................................ 65 Containment Dome height (sum of spring line and dome radius).......................... 441 3.3 Radiation Monitors 3.3.1 Monitor Ranges (FSAR TABLE 12.3.4-1) RM-1CR-3561A-SA.................................................................... 1E+1 - 1E+7 mR/hr RM-1CR-3561B-SB.................................................................... 1E+1 - 1E+7 mR/hr RM-1CR-3561C-SA.................................................................... 1E+1 - 1E+7 mR/hr RM-1CR-3561D-SB.................................................................... 1E+1 - 1E+7 mR/hr 3.3.2 Detector Location - Elevation (Drawing CAR-2166-G-0453) RE-1CR-3561-ASA.......................................................................................... 289 8 RE-1CR-3561-BSB.......................................................................................... 289 6 RE-1CR-3561-CSA.......................................................................................... 289 6 RE-1CR-3561-DSB.......................................................................................... 289 6

HNP EAL Technical Bases Calculation - CG1.2 and CS1.3 CSD-EP-HNP-0101-06 Page 7 of 30 Revision 0 3.3.3 Detector Location - Radial Distance from Core Center (Equipment Database System) RE-1CR-3561-ASA................................................................................................ 62 RE-1CR-3561-BSB................................................................................................ 26 RE-1CR-3561-CSA................................................................................................ 20 RE-1CR-3561-DSB................................................................................................ 16 3.4 Source Term 3.4.1 Source Term Activity (NUREG-1940 Table 1-1) Core Activity (Ci/MWt) Core Activity (Ci/MWt) Core Activity (Ci/MWt) Ba-139 4.74E+04 La-141 4.33E+04 Te-127 2.36E+03 Ba-140 4.76E+04 La-142 4.21E+04 Te-127m 3.97E+02 Ce-141 4.39E+04 Mo-99 5.30E+04 Te-129 8.26E+03 Ce-143 4.00E+04 Nb-95 4.50E+04 Te-129m 1.68E+03 Ce-144* 3.54E+04 Nd-147 1.75E+04 Te-131m 5.41E+03 Cm-242 1.12E+03 Np-239 5.69E+05 Te-132 3.81E+04 Cs-134 4.70E+03 Pr-143 3.96E+04 Xe-131m 3.65E+02 Cs-136 1.49E+03 Pu-241 4.26E+03 Xe-133 5.43E+04 Cs-137* 3.25E+03 Rb-86 5.29E+01 Xe-133m 1.72E+03 I-131 2.67E+04 Rh-105 2.81E+04 Xe-135 1.42E+04 I-132 3.88E+04 Ru-103 4.34E+04 Xe-135m 1.15E+04 I-133 5.42E+04 Ru-105 3.06E+04 Xe-138 4.56E+04 I-134 5.98E+04 Ru-106* 1.55E+04 Y-90 2.45E+03 I-135 5.18E+04 Sb-127 2.39E+03 Y-91 3.17E+04 Kr-83m 3.05E+03 Sb-129 8.68E+03 Y-92 3.26E+04 Kr-85 2.78E+02 Sr-89 2.41E+04 Y-93 2.52E+04 Kr-85m 6.17E+03 Sr-90 2.39E+03 Zr-95 4.44E+04 Kr-87 1.23E+04 Sr-91 3.01E+04 Zr-97* 4.23E+04 Kr-88 1.70E+04 Sr-92 3.24E+04 La-140 4.91E+04 Tc-99m 4.37E+04 Notes: NUREG-1940 radionuclides with an

  • are assumed to be present in secular equilibrium with short-lived daughters.

Additional source term from activation materials in the reactor vessel, such as vessel walls and internals, is not included. 3.4.2 Decay Time Note - Source term nuclide half-life and energies for the decay corrected core inventory, including daughter nuclides, are internal to the MicroShield application. HNP URI New Fuel Default Time After Shutdown.......................................... 72 hours

HNP EAL Technical Bases Calculation - CG1.2 and CS1.3 CSD-EP-HNP-0101-06 Page 8 of 30 Revision 0 3.4.3 Source Materials Weight Fuel Assembly dry weight (FS1-0030999 Table 2-2).................................... 1498.9 lb

  1. Fuel Assemblies (FSAR Table 4.1.1-1)............................................................... 157 Fuel (UO2) weight (FSAR Table 4.1.1-1)..................................................... 181,190 lb Zirconium alloy weight (FSAR Table 4.1.1-1)................................................ 41,415 lb Other Metals1 weight (see Section 4.2.2 and Attachment 1)....................... 1.27E+4 lb 3.4.4 Source Materials Density Uranium Dioxide - UO2 (NIST)................................................................... 10.96 g/cc Other Metals - iron (Attachment 1)............................................................... 7.86 g/cc Zirconium (Attachment 1)................................................................................ 6.5 g/cc Water (Attachment 1)...................................................................................... 1.0 g/cc Air (Attachment 1)................................................................................... 0.00122 g/cc 3.4.5 Source Geometry The source is considered a cylinder with core activity evenly distributed throughout the equivalent volume. For the outer liner, core barrel thickness is used, although ID/OD are adjusted for modeling simplification. The use of this geometry is not considered significant to the results as the dose receptor points are located over the source.

Active fuel height (FSAR Table 4.1.1-1)................................................................ 144 Core equivalent diameter (FSAR Table 4.1.1-1)................................................ 119.7 Core Barrel thickness (FSAR Table 4.1.1-1)............................................................ 4 D = 119.7 H = 144 Activated Fuel Assemblies Inside Core Barrel L = 4 1 Assembly metal weight (Inconel and other metals) density input as equivalent to iron.

HNP EAL Technical Bases Calculation - CG1.2 and CS1.3 CSD-EP-HNP-0101-06 Page 9 of 30 Revision 0 3.4.6 Receptor Geometry The reflected dose rate at the area radiation monitors is calculated using the methods of Davissons Gamma Ray Dose Albedos (see Attachment 5). The calculation is based on a concrete reflector at the top of containment, with a diameter equal to the containment dome. The distance from the reflector at o to the radiation monitor is equal to the hypotenuse of the triangle formed by the difference in elevations of the reflector and the monitor and the distance from core center to the monitor. Concrete is selected as the predominant containment dome reflector material. HNP containment has a carbon steel liner prior to the concrete so both materials contribute to reflection. References for albedo values representing layered materials were not available, thus the greater albedo of the two reflectors is used by itself. The reflected dose rate is related to the area of the reflector. Assuming a non-collimated beam originating from the core, the entire containment dome radius would be within line of sight. RPV Flange Elevation (ft): 260 RPV Inside Diameter (in): 155.5 TOAF Elevation (ft): 249 RPV Inside Diameter (ft): 13.0 Source to Flange (ft): 11 RPV Radius (ft): 6.5 Angle Between Core Center Line & RPV Flange (degrees): 30.5 Distance Where Beam Equals Containment Radius (ft): 122 Elevation Where Beam Equals Containment Radius: 314 a c b H = 192 Spring Line = 376 ele. TOAF = 249 ele. Dome Peak = 441 ele. Dome Radius = 65

HNP EAL Technical Bases Calculation - CG1.2 and CS1.3 CSD-EP-HNP-0101-06 Page 10 of 30 Revision 0 4 CALCULATIONS 4.1 Initial Source Term (Reactor Core) Activity The source term input to MicroShield is based on NUREG 1940 Table 1-1 isotopes and core activity, adjusted to HNP specific power rating. () = (

) x () 4.2 Source (Reactor Core) Volume See Attachment 1 for the spreadsheet results for the equations provided below. 4.2.1 Source Cylinder Total Volume () = 2() x () x 16.387

4.2.2 Weight of Other Metals () = () x #[2() + ()] NUREG-1940 Table 1-1 Core Activity (Ci/MWt) HNP Core Activity (Ci) NUREG-1940 Table 1-1 Core Activity (Ci/MWt) HNP Core Activity (Ci) NUREG-1940 Table 1-1 Core Activity (Ci/MWt) HNP Core Activity (Ci) Ba-139 4.74E+04 1.40E+08 La-141 4.33E+04 1.28E+08 Te-127 2.36E+03 6.96E+06 Ba-140 4.76E+04 1.40E+08 La-142 4.21E+04 1.24E+08 Te-127m 3.97E+02 1.17E+06 Ce-141 4.39E+04 1.29E+08 Mo-99 5.30E+04 1.56E+08 Te-129 8.26E+03 2.44E+07 Ce-143 4.00E+04 1.18E+08 Nb-95 4.50E+04 1.33E+08 Te-129m 1.68E+03 4.95E+06 Ce-144* 3.54E+04 1.04E+08 Nd-147 1.75E+04 5.16E+07 Te-131m 5.41E+03 1.59E+07 Cm-242 1.12E+03 3.30E+06 Np-239 5.69E+05 1.68E+09 Te-132 3.81E+04 1.12E+08 Cs-134 4.70E+03 1.39E+07 Pr-143 3.96E+04 1.17E+08 Xe-131m 3.65E+02 1.08E+06 Cs-136 1.49E+03 4.39E+06 Pu-241 4.26E+03 1.26E+07 Xe-133 5.43E+04 1.60E+08 Cs-137* 3.25E+03 9.58E+06 Rb-86 5.29E+01 1.56E+05 Xe-133m 1.72E+03 5.07E+06 I-131 2.67E+04 7.87E+07 Rh-105 2.81E+04 8.28E+07 Xe-135 1.42E+04 4.19E+07 I-132 3.88E+04 1.14E+08 Ru-103 4.34E+04 1.28E+08 Xe-135m 1.15E+04 3.39E+07 I-133 5.42E+04 1.60E+08 Ru-105 3.06E+04 9.02E+07 Xe-138 4.56E+04 1.34E+08 I-134 5.98E+04 1.76E+08 Ru-106* 1.55E+04 4.57E+07 Y-90 2.45E+03 7.22E+06 I-135 5.18E+04 1.53E+08 Sb-127 2.39E+03 7.05E+06 Y-91 3.17E+04 9.35E+07 Kr-83m 3.05E+03 8.99E+06 Sb-129 8.68E+03 2.56E+07 Y-92 3.26E+04 9.61E+07 Kr-85 2.78E+02 8.20E+05 Sr-89 2.41E+04 7.10E+07 Y-93 2.52E+04 7.43E+07 Kr-85m 6.17E+03 1.82E+07 Sr-90 2.39E+03 7.05E+06 Zr-95 4.44E+04 1.31E+08 Kr-87 1.23E+04 3.63E+07 Sr-91 3.01E+04 8.87E+07 Zr-97* 4.23E+04 1.25E+08 Kr-88 1.70E+04 5.01E+07 Sr-92 3.24E+04 9.55E+07 La-140 4.91E+04 1.45E+08 Tc-99m 4.37E+04 1.29E+08 Rated Power (MWt): 2948

HNP EAL Technical Bases Calculation - CG1.2 and CS1.3 CSD-EP-HNP-0101-06 Page 11 of 30 Revision 0 4.2.3 Source Materials (Fuel Assemblies) Volume () = () x 453.592

4.2.4 Source Water Volume () = () () 4.3 Source Materials Mixed Shielding Density The source materials mixed shielding density is calculated in accordance with the guidance provided in MicroShield user's manual as shown in Attachment 2. See Attachment 1 for the spreadsheet results for the equations provided below. 4.3.1 Source Material Volume Fractions (MVF) (%) = () () 4.3.2 Individual Source Material Density Component (IDC)

= (%) x

4.3.3 Effective Source Density (ESD) for MicroShield

=

4.4 MicroShield Results The relative dose rate at top of active fuel is developed using MicroShield with the inputs and interim calculated results described in Sections 3 and 4 above. Gamma exposure from buildup is selected as the representative MicroShield result of interest. Core Surface (water level TOAF)....................................................... 3.232E+09 mR/hr Containment Dome Peak................................................................... 7.628E+05 mR/hr Refer to Attachment 3 for the comprehensive results of the MicroShield relative dose rates at top of active fuel and containment dome peak.

HNP EAL Technical Bases Calculation - CG1.2 and CS1.3 CSD-EP-HNP-0101-06 Page 12 of 30 Revision 0 4.5 3561 Radiation Monitor Dose Rates for RCS Level at TOAF 3561 (/) = (/) x cos x 2 x (,,, ) Where: 3561s Dose rate at the 3561 radiation monitors (R/hr) RSDRo MicroShield dose rate incident on reflector(dome) surface at o (R/hr) o Incident angle with respect to the normal (0°) A Reflecting area (ft2) c Distance from center of reflecting area to monitor (ft) Dose albedo - (Eo, o,, ) Note - Since the incident angle (o ) is 0 determination and use of the azimuth angle () is not necessary. Refer to Attachment 4 for the spreadsheet results for the 3561 monitor back-scatter dose rates due to water level at TOAF. Refer to Attachment 5 for details and inputs from Davisson, Gamma Ray Dose Albedos. 4.5.1 Reflecting Area (A) (2) = 2 Where: r Containment Dome radius (ft) h Containment Dome height from springline (ft) 4.5.2 Distance from Reflector Surface at o to Detector (c) () = 2 + 2 4.5.3 Emerging Polar Angle of the Reflector to the Detector () = =

HNP EAL Technical Bases Calculation - CG1.2 and CS1.3 CSD-EP-HNP-0101-06 Page 13 of 30 Revision 0 4.5.4 Dose Albedo Energy Groupings (Eo) The reference albedos are given for incident gamma energies of 0.2, 0.662, 1.0, 2.5 and 6.13 MeV. The MicroShield result tables provided the photon energies and activities that are used to develop predominant groupings as follows: Energy (MeV) Activity (/sec) Activity (%) Group Abundance (%) 0.015 1.77E+19 17.01% 56.5% 0.02 6.68E+17 0.64% 0.03 4.79E+18 4.59% 0.04 2.52E+18 2.42% 0.05 3.30E+17 0.32% 0.06 4.23E+17 0.41% 0.08 1.86E+18 1.78% 0.1 1.91E+19 18.34% 0.15 5.54E+18 5.31% 0.2 5.97E+18 5.73% 0.3 7.02E+18 6.73% 22.5% 0.4 2.46E+18 2.36% 0.5 9.04E+18 8.67% 0.6 4.96E+18 4.76% 0.8 1.49E+19 14.33% 15.7% 1 1.47E+18 1.41% 1.5 5.09E+18 4.88% 5.2% 2 1.55E+17 0.15% 3 1.74E+17 0.17% 4 2.81E+03 0.00% 5 6.93E+07 0.00% Totals 1.04E+20 100% The first grouping is applied to the 0.2 MeV albedo table column. The second grouping is applied to the 0.662 MeV albedo table column. The third grouping is applied to the 1.0 MeV albedo table column. The fourth grouping is applied to the 2.5 MeV albedo table column.

HNP EAL Technical Bases Calculation - CG1.2 and CS1.3 CSD-EP-HNP-0101-06 Page 14 of 30 Revision 0 5 CONCLUSION 5.1 Individual Monitor Results Monitor mR/hr RM-1CR-3561A-SA 1.97E+4 RM-1CR-3561B-SB 2.60E+4 RM-1CR-3561C-SA 2.63E+4 RM-1CR-3561D-SB 2.65E+4 5.2 EAL Threshold Values Based on monitor accuracy/readability, human factors and the similarity of results between radiation monitors, the CS1.3 and CG1.2 EAL threshold bullets are established as follows: Any RM-1CR-3561 (A, B, C, or D) containment radiation monitor > 2.6E+4 mR/hr

HNP EAL Technical Bases Calculation - CG1.2 and CS1.3 CSD-EP-HNP-0101-06 Page 15 of 30 Revision 0 6 REFERENCES 6.1. NEI 99-01 Rev 6, Methodology for Development of Emergency Action Levels, November 2012 6.2. NUREG-1940, RASCAL 4: Description of Models and Methods, December 2012 6.3. MicroShield program, Version 7.02 (08-MSD-7.02-1532) 6.4. National Institute of Standards and Technology (NIST) - http://www.physics.nist.gov/cgi-bin/Star/compos.pl?matno=272 - density of Uranium Dioxide (UO2) 6.5. ANS/SD-75-14, A Handbook of Radiation Shielding Data, July 1976 6.6. HNP Final Safety Analysis Report Update (FSAR) Table 4.1.1-1, Original Core Reactor Design Comparison Table, Amendment 61 Table 12.3.4-1, Area Radiation Monitors, Amendment 61 6.7. FS1-0030999, Adv. W17 HTP' Mass & Volume Calculation, Revision 2 6.8. Drawing CAR-2165-G-0013, General Arrangement Containment Building, Sheet 1, Revision 17 6.9. Drawing CAR-2166-G-0453, Containment Building Instrument Location Arrangement, Revision 16 6.10. Drawing CAR 1364-001275, Reactor Vessel General Arrangement, Revision 2 Source Volume and Effective Density Determination CSD-EP-HNP-0101-06 Page 16 of 30 Revision 0 Source Height (in): 144 Fuel Assembly dry weight (lbs): 1.50E+03 Source Diameter (in): 119.7

  1. Fuel Assemblies:

157 Source Radius (in): 59.85 Weight of all fuel assemblies: 2.35E+05 Source Volume (cm3): 2.66E+07 Weight of UO2 and Zirc: 2.23E+05 Weight of other metals: 1.27E+04 Conversion (gm per lb): 4.54E+02 Density (g/cc) Mass (lb) Mass (g) Volume (cc) Material Volume Fraction - MVF Individual Density Complnent - IDC (g/cc) UO2 10.96 1.81E+05 8.22E+07 7.50E+06 28.2% 3.09 Zirc 6.5 4.14E+04 1.88E+07 2.89E+06 10.9% 0.71 Metal 7.86 1.27E+04 5.77E+06 7.34E+05 2.8% 0.22 Water 1 1.54E+07 1.54E+07 58.1% 0.58 Effective 4.60 1.22E+08 2.66E+07 MicroShield Instructions for Source Density Determination ~rove software RFIDIATIONSCFrWFIRE.COm MicroShield User's Manual 5.2 Case Material Screen Attenuation and buildup (scatter) of radiation between a source and a dose point are affected by all intervening materials. Shield materials determine the radiation attenuation and buildup characteristics used to calculate the dose rate. A library of material characteristics can be used to formulate custom materials. Published attenuation coefficient data from ANSI/ANS-6.4.3-1991, Gamma-Ray Attenuation Coefficients and Buildup Factors for Engineering Materials, from the American Nuclear Society is provided for 100 atomic elements as well as air, water, and concrete. MicroShield provides 12 built-in materials and allows adding as many as eight user-defined materials for each shield region. User-defined materials are called "custom" materials. They must first be created with the custom material selection under the tools menu before they are available for shield input. The 12 built-in materials are chosen because they are used often. These materials and their default densities are: Material Density fg/cm3) Material Density (g/cm3) Air 0.00122 Tin 7.3 Aluminum 2.702 Titanium 4.5 Concrete 2.35 Tuncisten 19.3 Iron 7.86 Uranium 18.75 Lead 11.3 Water 1.0 Nickel 8.9 Zirconium 6.5 The default standard density of air is 0.00122 g/cm3 to be consistent with ANSI/ANS-6.6.1-1979. American National Standard for Calculation and Measurement of Direct and Scattered Gamma Radiation from LWR Nuclear Power Plants, from the American Nuclear Society. The Case Material Screen in Figure 5-3 allows the user to enter the materials present in the source and all shields. The column headers show the name of the source or shield and its dimension. Each row represents the type of material present. The user can enter any density for any material by clicking on the appropriate cell and entering the density with the keyboard. As well, hitting the space bar or double-clicking will enter the default density of the material. © 2014 Grove Software - All rights reserved 5-8 CSD-EP-H N P-0101-06 Page 17 of 30 Revision 0 MicroShield Instructions for Source Density Determination ~rove software RFIDIFITIDNSCFTWFlfifE.COm MicroShield User's Manual When more than one material is specified for a shield, the user has two choices to decide the appropriate density. One choice is to pre-calculate the volume fraction or the effective density for each material in the region and use that as an entry. As an example of pre-calculated volume fractions, a source consists of a cylindrical tank of water with steel internal structures. The source region may be represented as a tank with a homogeneous mixture of steel and water. The volume of the tank is 10,000 liters and it contains 8,000 kg of water and 1560 kg of iron. The material densities to enter are then: 8x106g Water:---- = 0.8gicm 3 10xl06cm3 6 . I.56x10 g _ 3 Iron. 6 3 - 0.L6 g/cm IOxlO cm ~ Another way to estimate densities is to multiply each individual material volume fraction by its pure density. For example, if a volume contains 95% (by volume) water at a density of 1.0 g/cm3 and 5% (by volume) steel at a density of 7.85 g/cm3, then the individual density entries are: Water: 0.95 x 1.0 g/cm3 = 0.95 g/cm3 Iron: 0.05 x 7.85 g/cm3 = 0.3925 g/cm3 Reinforced concrete can similarly be represented with a bulk material iron density representative of the reinforcing rod and the balance being concrete. A note of caution is appropriate while using custom and mixed materials. The fundamental coefficient data for attenuation and buildup are based on specific materials for which measurements have been made or which have been derived by more precise computing methods than are used in MicroShield. The method in MicroShield for combining materials is an approximation and will vary somewhat from the "true" situation. Therefore, results of the use of custom materials and mixed materials should be viewed with caution until a user has confidence in their validity. If field measurements are available, it is suggested that densities be adjusted to "normalize* the results of calculations to measurements. The energy-dependent linear attenuation coefficients for any region in a specific case may be read from the case file display option on the home menu. Quite often this display is useful for assessing the cause of differences among various dose rate calculations. © 2014 Grove Software - All rights reserved 5-11 CSD-EP-HNP-0101-06 Page 18 of 30 Revision 0 MicroShield Calculation Report Case Summaiy ofTOAF Page 1 of 5 MicroShield 7.02 Duke Energy Corporation (08-1\\ISD-7.02-1532) I Date I By Checked 06/05/18 I William Cerame Scott McCain Filename Run Date I Rnn Time I Duration EALCALC 1801 HNP TOAF T72.ms7 I June 5, 201s I 9:30:25 AM I 00:00:00 Case Title Description Geometry Hei2ht I Radius I A X

Ill.. #1 0.0 cm (0.0 in)

~

  1. 2 0.0 cm (0.0 in)

Shield N Source Air Gap Wall Clad Nuclide Ac-225 Ac-227 Am-241 At-217 Ba-137m Ba-139 Ba-140 Bi-210 Bt-211 Bi-213 Bi-214 Ce-141 Ce-143 Ce-144 Project Info TOAF 72 Hour Decay, TOAF & Top of Dome 8 - Cylinder Volume - End Shields Source Dimensions 365.76 cm (12 ft) 152.019 cm (4 ft 11.9 in) Dose Points y z 368.3 cm (12 ft 1.0 in) 6.2e+3 cm (204 ft) 0.0 cm (0.0 in) 0.0 cm (0.0 in) Shields Dimension Material Densitv 2.66e+o7 cm' Uranmm 4.6 Air 0.00122 10.16 cm Iron 7.86 Source Input: Grouping Method - Standard Indices Number of Groups: 25 Lower Energy Cutoff: 0.015 Photom < 0.015: Included Library: Grove Ci Bq 11Ci/cm* 2.5673e-015 9.4989e-005 9.6679e-Ol7 l.6594e+o02 6.1397c+0l2 6.2489e+o00 9.0610e+006 3.3526e+017 3.4122e+005 3.1399c-008 1.1618e+oo3 1.1824e-009 l.l 899e+o08 4.4027e+ol8 4.4810e+o06 3.444le-015 l.2743c-004 l.2970e-016 2.4535e-018 9.0779c-008 9.2393e-020 l.8986e-015 7.0250e-005 7.1499e-017 l.216le+008 4.4997c+018 4.5798e+o06 2.6007e+007 9.6226e+0l 7

9. 7937e+o05 l.0324e+o08 3.8200e+o18 3.8879e+o06

'2 Bq/cm* 3.577le-012 2.3121c+o05 l.2625e+010 4.3750e-005 l.6580e+0l l

4. 7989c-012 3.4185e-015 2.6455e-012 l.6945e+oll 3.6237e+0l0 1.4385e+0l 1 6/5/2018 Note - Microshield measures distance to receptor from the bottom of the source, therefore an additional 12' is added to the "Y" axis to account for core height in the TOAF and containment dome dose rate calculations.

CSD-EP-HNP-0101-06 Page 19 of 30 Revision 0 I ~ MicroShield Calculation Report Case Summary ofTOAF Page 2 of 5 Cm-242 3.2582e+o06 l.2055e+0l 7 l.2270e+o05 4.5398e+009 Cs-134 l.3862e+007 5.1288e+017 5.2200e+o05 l.9314e+0l0 Cs-135 6.8746e-002 2.5436e+o09 2.5888e-003 9.5787e+001 0,-136 3.7484e+o06 l.3869e+ol 7 l.4116e+005 5.2228e+009 Cs-137 9.5782e+o06 3.5439e+0l 7 3.6070e+o05 l.3346e+0 I 0 Cs-138 4.3019e-033 1.5917e-022 l.6200e-034 5.9940e-030 Fr-221 Fr-223 1-129 2.6114e-003 9.6623e+o07 9.8342e-005 3.6386e+000 1-131 6.2469e+o07 2.3114e+0I8 2.3525e+o06 8.7042e+010 I-132 6.0956e+o07 2.2554e+ol 8 2.2955e+o06 8.4934e+0l 0 I-133 l.4524e+007 5.3740e+ol 7 5.4695e+o05 2.0237e+010 I-134 3.3276e-017 1.2312.:-006 l.2531e-018 4.6366e-014 1-135 8.0481e+o04 2.9778e+ol5 3.0308e+o03 l.1214e+008 Kr-83m l.2881e-005 4.7660e+005 4.8508e-007

1. 7948.:-002 Kr-85 8.1975e+005 3.033 le+Ol 6 3.0870e+o04 1.1422e+009 Kr-85m 2.6430e+002 9.7789e+0l2 9.9529e+o00 3.6826e+005 Kr-87 3.2810e-010 l.2140e+001 1.2356e-Oll 4.57 l 6e-007 Kr-88 l.1697e+o00 4.3281e+0l0 4.4050e-002 I.6299e+003 La-140 l.3228e+o08 4.8942e+ol8 4.9813e'+-006 l.843le+0ll La-141 4.0379e+o02 l.4940e+013 l.5206e+o01 5.6262e+005 La-142 2.8965e-006 1.0717e+o05
1. 0908e-007 4.0358e-003 Mo-99 7.3253e+007
2. 7104e+o18 2.7586e+o06 l.0207e+0l 1 1'."'b-93m l.7431e-005 6.4496e+o05 6.5643e-007 2.4288e-002 Nb-95 l.3272e+o08 4.9105e+0l8 4.9979e+o06 l.8492e+0ll Nb-95m 4.5097e+o05 l.6686e+ol6 l.6983e+o04 6.2836e+o08 Nb-97 7.0284e+o06 2.6005e+ol 7 2.6467e+005 9.7930e+009 Nb-97m 6.1829e+006 2.2877e+ol 7 2.3284e+o05 8.6150e+009 Nd-147 4.2697e+o07 l.5798e+018 l.6079e+o06 5.9492e+010 Np-237 3.3518e-007 l.2402e+004 l.2622e-008 4.6702e-004 Np-239 6.9475e+o08 2.5706e+0I9 2.6163e+o07 9.6804e+0l 1 Pa-231 l.8443e-017 6.8240e-007 6.9453e-019 2.5698e-014 Pa-233 8.5135e-009
3. I 500e+002 3.2060e-OIO l.1862e-005 Pb-209 Pb-210 3.3174e-015 l.2274e-004 l.2493e-016 4.6223e-012 Pb-211 Pb-214 1.0151e-015 3.7560e-005 3.8228e-017 I.4144e-012 Pm-147 l.0190e+005 3.7704e+0IS 3.8374e+o03 l.4198e+008 Po-210 2.7135.:-015 l.0040e-004 l.0219e-016 3.7809e-012 Po-211 l.2761e-020 4.7216e-010 4.8056e-022 l.7781e-017 Po-213 Po-214 2.3399e-015 8.6577e-005 8.8117e-017 3.2603e-012 Po-215 2.2776e-020 8.4272e-010 8.5770e-022
3. l 735e-0l 7 Po-218 4.9319e-016 l.8248e-005 l.8573e-017 6.8719e-013 Pr-143 l.0885e+008 4.0276e+0l8 4.0992e+o06 l.5167e+0ll 615/2018 CSD-EP-HNP-0101-06 Page 20 of 30 Revision 0 MicroShield Calculation Report Case Summary ofTOAF Page 3 of5 Pr-144 l.0325e+-008 3.8201e+0l8 3.8880e+o06 l.4386e+0ll Pr-144m l.4764e+006 5.4626e+016 5.5598e+004 2.057le+009 Pu-238 2.1274e+-002 7.8713e+o12 8.0113e+o00 2.9642e+005 Pu-239 2.6325e+o02 9.7403e+0l2 9.9135e+o00 3.6680e+005 Pu-241 l.2595e+-007 4.6602e+ol7 4.7430e+o05 l.7549e+0l0 Ra-223 Ra-225 Ra-226 1.0042e-015 3.7156c-005 3.7817e-017 l.3992e-012 Rb-86 l.3955e+-005 5.1633e+015 5.2552e+003 l.9444e+008 Rb-87 l.1133e-007 4.l 193e+003 4.1925e-009 l.5512e-004 Rb-88 l.3062e+-O00 4.8329e+0l0 4.9189e-002 1.8200e+o03 Rh-103m l.2121e+-008 4.4849e+018 4.5646e+o06 l.6889e+0ll Rh-105 2.3345e+007 8.6377e+0l7 8.7913e+005 3.2528e+0l0 Rh-105m 2.9109e+o02 1.0770e+0l3 l.0962e+-001 4.0559c+005 Rh-106 4.5443e+-007 1.6814e+Q18
1. 7113e+o06 6.3318e+0l0 Rn-219 Rn-222 4.3294e-016 l.6019e-005 1.6304e-017 6.0324e-013 Ru-103 l.2141e+o08 4.4922c+018 4.5721e+006 l.6917e+0ll Ru-105 l.1848e+-003 4.3837e+013 4.4617e+o01 l.6508e+006 Ru-106 4.5443e+o07 l.6814e+018 l.7113c+006 6.3317e+0l0 Sb-127 4.1079c+o06 l.5199e+ol 7 1.54 70e+o05 5.7237e+009 Sb-129 3.0359e+002 l.1233e+013 l.1433e+001 4.230Ie+005 Sm-147 2.8239e-009 1.0449e+o02 1.0634e-010 3.9347e-006 Sr-89 6.8139e+007 2.521 le+018 2.5660c+o06 9.494le+010 Sr-90 7.0486c+006 2.6080e+ol 7 2.6544e+005 9.82Ue+009 Sr-91 4.639lc+o05 l.7165e+ol6 1.7 4 ?0e+-004 6.4639e+008 Sr-92 9.5975e-001 3.551 le+ol0 3.6142e-002 l.3373e+003 Tc-99 3.ll 15e+00O l.1512c+0ll l.1717e-001 4.3354e+003 Tc-99m 7.1409e+007 2.6421c+0I 8 2.6891e+o06 9.9497e+010 Te-127 4.9399e+-006 l.8278e+0l 7 1.8603e+-005 6.8830e+009 Te-127m l.1653c+o06 4.3115e+016 4.3882e+o04 1.6236e+009 Te-129 2.945le+-006 1.0897e+0I 7 l.1091c+005 4.1035e+009 Tc-129m 4.6749e+006 l.7297e+017
1. 7 605e+005 6.5137e+009 Te-131
6. 7819e+-005 2.5093e+ol6 2.5539e+o04
  • 9.4496e+008 Te-13lm 3.0125e+006 l.1146c+017 1.1344e+o05 4.1974e+009 Te-132 5.9164e+-007 2.189le+018 2.2280e+006 8.2436e+0l0 Th-227 Th-229 Th-230 6.l699e-014 2.2829e-003 2.3235e-Ol 5 8.5969e-0l l Th-231 5.4638e-0l0 2.0216e+001 2.0576e-Ol l 7.6130e-007 Tl-207 6.522le-019 2.4132e-008 2.456le-020 9.0875e-0l6 Tl-209 U-233 2.6642e-0l 7 9.8575e-007 l.0033e-018 3.7122e-014 U-234 2.4821e-006 9.1838e+o04 9.347le-008 3.4584e-003 U-235 l.2194c-009 4.5119e+001 4.5922e-O 11 l.6991e-006 6/5/2018 CSD-EP-HNP-0101-06 Page 21 of 30 Revision 0 MicroShield Calculation Report Case Smnmary of TOAF Page4 of5 U-237 8.1829e+001 3.0277e+ol2 3.0815e+O00 l.1402e+005 Xe-13lm l.0289e+006 3.8071e+0l6 3.8748e+004 l.4337e+009 Xe-133 l.2684e+008 4.6931e+ol8 4.7765e+006 l.7673e+Oll Xe-133m 2.8555e+006 l.0565e+o17 l.0753e+005 3.9787e+009 Xe-135 l.6626e+006 6.1515e+0l6 6.2609e+004 2.3165e+009 Xe-135m l.3814c+004 5.lll4e+o14 5.2023e+002 1.9 248e+007 Xe-138 l.2372e-084 4.5778e-074 4.6592c-086
1. 7239e-081 Y-90 7.l276c+006 2.6372e+0l 7 2.684le+005 9.9313c+009 Y-91 9.0815e+007 3.3602e+018 3.4199e+006 l.2654c+0l 1 Y-9lm 2.9173e+005 l.0794e+o16 l.0986e+004 4.0648e+008 Y-92 3.0442e+002 1.1264e+013 1.1464e+o01 4.2417e+005 Y-93 5.3092e+o05 l.9644c+016 l.9994e+004 7.3976e+008 Zr-93 5.5552e-002 2.0554e+009 2.0920e-003
7. 7 404c+00 1 Zr-95 l.268le+008 4.692lc+018 4.7755e+o06 l.7670e+0ll Zr-97 6.5225e+006 2.4133e+ol 7 2.4563e+005 9.088le+009 Buildup: The material reference is Source Intel!ration Parameters Radial 20 Crrcumterenti.al 10 Y Direction (axial)

IO Results - Dose Point # 1 - (0,368.3,0) cm Fluence Rate Fluence Rate Exposure Rah Exposure Rah Energy (l\\ieY) Acthity (Photons/sec) l\\Ie V /cm'/sec MeY/cm'/~ec mR/hr mR/hr No Buildup With Buildup No Buildup With Bnildup 0.015 l.773e+l9 8.708e-28 8.572e-15 7.469e-29 7.352e-16 0.02 6.684e+17 5.747e-33 4.308e-16 l.991e-34 l.492e-17 0.03 4.787c+18 l.711e-15 6.393e-15 l.696e-17 6.336e-17 0.04 2.519e+18 l.200e-03 l.236c-03 5.306e-06 5.465c-06 0.05 3.295e+l7 l.160e+ol l.205e+0l 3.089e-02 3.21 lc-02 0.06 4.232e+l7 3.564c+03 3.744c+03 7.079e+00 7.436c+00 0.08 1.858e+18 3.333e+06 3.597e+06 5.275e+03 5.693e+o3 0.1 1.9lle+l9 5.965e+08 6.597e+08 9.126e+05 l.009e+o6 0.15 5.538e+18 5.448e+07 3.407c+o8 8.971e+04 5.610e+05 0.2 5.969e+l8 l.154e+09 l.750e+09 2.037e+06 3.088e+06 0.3 7.018e+l8 l.442e+I0 l.707e+IO 2.736e+07 3.239e+07 0.4 2.455e+18 1.399e+l0 l.625e+10 2.727e+07 3.166e+o7 0.5 9.04le+l8 9.747e+l0 l.145e+ll l.913e+08 2.247e+o8 0.6 4.957e+18 8.513e+l0 l.013e+ll 1.662e+08 l.977c-ru8 0.8 l.493e+19 4.972e+ll 6.165e+ll 9.458e+08 l.173e+09 1.0 l.474e+l8 7.832e+l0 9.997e+l0 1.444e+08 l.843e+08 1.5 5.086e+18 5.645e+ll 7.392e+ll 9.497e+08 l.244e+09 2.0 l.547e+17 2.599e+l0 3.472e+10 4.0l8e+07 5.369e+o7 3.0 l.743e+17 4.777e+10 6.396e+10 6.481c+07 8.677e+o7 4.0 2.813e+o3 l.037e-03 1.367e-03 l.283e-06 l.69lc-06 CSD-EP-HNP-0101-06 Page 22 of 30 Revision 0 MicroShield Calculation Report Case Smnmary ofTOAF Page 5 of5 I 5.0 I 6.929e+07 I 3.136e+0l I 4.238e+Ol I 3.595e-02 I 4.859e-02 I Totals 1.042e+20 I 1.427e+12 1.806e+12 2.560e+09 I 3.232e+09 I Results - Dose Point # 2 - (0,6217.92,0) cm Fluenre Rate Fluence Rate Exposure Rate Exposure Rah Energy (Me,') Acthity (Photons/ser) ~IeY/cm2/sec MeY/cm2/sec mR/hr mR/h1* No Buildup With Buildup ~o Buildup With Buildup 0.Dl5 l.773e+19 2.54le-33 5.129e-18 2.l 79e-34 4.399e-19 0.02 6.684e+l7 9.344e-36 2.578e-19 3.237e-37 8.930e-21 0.03 4.787e+18 2.994e-18 5.823e-18 2.967e-20 5.77le-20 0.04 2.519e+l8 5.567e-07 5.734e-07 2.462e-09 2.536e-09 0.05 3.295e+17 3.l 14e-03 3.237e-03 8.296e-06 8.624e-06 0.06 4.232e+17 7.l 71e-01 7.540e-0l l.424e-03 l.498e-03 0.08 l.858e+18 3.174e+02 3.417e+02 5.023e-0l 5.407e-0l 0.1 l.9lle+19 2.313e+04 2.548e+04 3.539e+0l 3.897e+Ol 0.15 5.538e+18 4.388e+03 l.947e+04 7.226e+00 3.206e+Ol 0.2 5.969e+IS 3.936e+o4 5.995e+04 6.947e+0I l.058e+o2 0.3 7.018e+18 1.309e+06 l.634e+06 2.483e+03 3.100e+03 0.4 2.455e+18 2.126e+06 2.663e+06 4.143e+03 5.189e+03 0.5 9.04le+18 l.748e+07 2.228e+07 3.430e+04 4.374e+04 0.6 4.957e+18 l.636e+07 2.120e+07 3.193e+04 4.139e+04 0.8 l.493e+19 l.030e+08 1.398e+08 l.960e+05 2.658e+05 1.0 1.474e+18 1.708e+07 2.388e+07 3.149e+04 4.40le+04 1.5 5.086e+18 l.339e+08 l.904e+08 2.253e+05 3.203e+05 2.0 l.547e+17 6.494e+06 9.364e+06 l.004e+04 l.448e+04 3.0 l.743e+l7 1.270e+07 l.813e+07 1.723c+04 2.459e+04 4.0 2.813e+o3 2.858e-07 3.970e-07 3.535e-10 4.912e-10 5.0 6.929e+o7 8.844e-03 1.258e-02 1.014e-05 l.442e-05 Totals 1.042e+20 3.106e+08 4.294e+08 5.530e+05 7.628e+05 6i5i2018 CSD-EP-HNP-0101-06 Page 23 of 30 Revision 0 Radiation Monitor Back-Scatter Determination CSD-EP-HNP-0101-06 Page 24 of 30 Revision 0 Radiation Monitor Elevation Detector to Core CL (a) Detector to Dome (b) Dome to Detector (c) Angle (degrees) Angle (radians) Energy Category (MeV) Energy Group (fraction) Dose Albedo () Fractional Dose (mR/hr) Total Dose (mR/hr) 0.2 0.57 0.0371 15901.68 0.6 0.23 0.0145 2475.11 1.00 0.16 0.0100 1187.58 2.50 0.05 0.0045 176.85 0.2 0.57 0.0432 20949.96 0.6 0.23 0.0172 3326.36 1.00 0.16 0.0114 1544.21 2.50 0.05 0.0045 198.68 0.2 0.57 0.0432 21197.56 0.6 0.23 0.0172 3365.67 1.00 0.16 0.0114 1562.46 2.50 0.05 0.0045 201.03 0.2 0.57 0.0432 21329.09 0.6 0.23 0.0172 3386.55 1.00 0.16 0.0114 1572.15 2.50 0.05 0.0045 202.28 Top of Active Fuel (ele ft): 249 Dome Hemisphere Reflecting Area (ft2): 2.65E+04 Dome Peak Dose Rate (mR/hr): 7.63E+05 Spring Line (ele ft): 376 Incident Angle - o (radians): 0 Dome Radius (ft): 65 Cos o 1 Dome Peak (ele ft): 441 TOAF to Dome (ft): 192 RM-01CR-3561DSB 289.5 16 151.5 152.3 6.03 0.11 2.65E+04 2.60E+04 RM-01CR-3561CSA 289.5 20 151.5 152.8 7.52 0.13 2.63E+04 0.39 1.97E+04 RM-01CR-3561BSB 289.5 26 151.5 153.7 9.74 0.17 RM-01CR-3561ASA 289.7 62 151.3 163.5 22.28 Radiation Monitor Back-Scatter Basis Reference A HANDBOOK OF RADIATION SHIELDING DATA CSD-EP-HNP-0101-06 J. C, COURTNEY, EDITOR Sponsored by: Nuclear Science Center Louisiana State University Baton Rouge l!lJnd Shielding and Dosimetry Division American Nuclear Society JULY, 19'76 Page 25 of 30 ANS/SD-76/14 Revision 0 Radiation Monitor Back-Scatter Basis Reference 5-27 Gamma Ray Dose Albedos C. M. Davisson U.S. Naval Research Laboratory The dose rate reflected from a surface as deduced from Reference 1 through 4 may be represented as: where D.R. D.R. Q Reflected dose rate D.R. 0 = Dose rate incident on surface at angJt 80 A Reflecting area r m Distance from center of reflecting area to receptor (A and r 2 must be in the same units) a(! 8. e. ~) = Dose albedo 0 0 The albedos, a(E, 8, 8, ~), for gammas incident on water, concrete, iron and leari have been calculited0by C. M. Davisson and L. A. Beach5 using Monte Carlo techniques in an extension of the original work by Theus and Beach6

  • The albedos are given for incident gamma energies of 0.2. 0,662, 1.0, 2,5 and 6.13 MeV and for 0

0 0 0 0 incident angles W1th respect to the normal of O

  • 22, 44, 66 and 88, as well as for point sources on the surface of the materials. The emerging polar angles, e1, as well as the emerging sectors or directions into which the emerging gammas were divided are shown in Fig. 5.13.

The values of the polar angles, e1, and of the azimuthal angles~. defining the emerging directions,~* are given on each page of Table 5,8. Note: The dose albedo values have statistical errors that range from 40% or 50! at very small albedo values to 5% or 10% at large albedo values. References 1 Reactor Shielding Design Manual, T, Rockwell III, editor. TID-7004 (March 1956)

p. 334, 2 D. J. Raso, "Monte Carlo Calculations on the Reflection and Transmission of Scattered Gamma Rays, 11 ~* ~* and Eng, 17, 411 (1963).

This report has a good discussion of the meaning of various terms and derived quantities. The dose albedos given here are those which he described in quotes, as "dose" albedos, 3 W. E. Selph, "Neutrons and Gamma-lay Albedos," DASA-1892-2 (May 1967), 0BNL-RSIC-21 (February 1968), or Chapter 4 of Weapons Radiation Shielding Handbook (NTIS No. AD-816 092). The dose albedos given here are those defined as a 1n this report. It R, L. French. and M, B, Wells, "An-Angle-Dependent Albedo for Fase~eutron Reflection Calculations, 11 Nucl. Sci. and Eng. 19, 441 (1964). 5 C, M. Davisson and L, A, Beach, 11Gamma-Ray Albedos of Iron," NRL Quarterly on Nucl. Sci, and Tech, (January 1 1 1960), p. 43; and private communication. 6 a. B. Theus and L. A. Beach, "Gamma-Ray Albedo," NRL Quarterly on Nucl. Sci. and Tech. (July-September 1955). CSD-EP-HNP-0101-06 Page 26 of 30 Revision 0 Radiation Monitor Back-Scatter Basis Reference S-28 Figure 5.13 NORMAL MATERIAL SURFACE ---- Geometry and Solid Angle Divisions CSD-EP-HNP-0101-06 Page 27 of 30 Revision 0 00 D ~~ DO Dd !~.. i] ..... (! 000000 ai2Ei-2 51 .-t:r+r-"

i~~i~

7i flo.... i CSD-EP-HNP-0101-06 Radiation Monitor Back-Scatter Basis Reference 5-31 Table 5.8 continued Page 28 of 30 r,H;:; ~8 Revision 0 A... i, CSD-EP-HNP-0101-06 Radiation Monitor Back-Scatter Basis Reference S-32 Table 5.8 continued Page 29 of 30 '. d "'~ 0 Revision 0 Radiation Monitor Back-Scatter Basis Reference S-33 Table 5.8 continued m~ g~ .t;~ ~-- ... ~ ~fi t~. f, t,!1': ';Ii! ~, 1.:: r:~ ~~ I -~ r: "{ -: 3 ai.,.., iit~l"'"':.; ir~i:;'~ ~ if 4 q'I LI", J.. ~ :i, s-* -:r.~r~*~~;r.: :r*. v,~*,.-1J-.,* r1ro1'1'1C'1 ~?:,.. ~ ~ ~~ 1,....~ _~..... 1u~ .~ ~ ~4~ ~ -~~~~{ ~,:~ ll ,. f,11,...... '.t~ ~*--

t. ~t.~ lit
~

~.. ~t'ti\\n~ =. ~":' ~~.~t:. ~ - ~ ~"":"! -~ _.. - ..,._"t\\e ........ [\\J p'J. ~.,.;..: ~.;.,~..;;

      • *...:._..;.:,"It':

_,,.,,.l.r-=s:...

  • 1

~ti.,._ ~~:!'1,t,;:~

  • "l,t,-m,.lt-tl'J

~.. *~i~=~ t; ;:~*~t: ~ it,J~~ a~~, f~iLlt~J i -~ ~ ~; (',_..,. ~.. -,, j !k ~,-! ~~~~~~ ~-.,. *<t*' * *.:i,. '°'. v:11,i..J!: !~ ~~ -'! '! ~ ~~~ j .~!*~':.;: ~ '! '!' ': \\~ ~ ~.,,7. ** j~ ~ -~ ~~.-. i e" s..,., ~'f,_ '£1!1 t;~S:::;'g: ':.lf'!l!~ 1r4..= thlrlt-,11\\.-r-,. Ela~;~~;*§ !;fl i: i;~ ~~t~c*, i :.n~ ~~ ~ ::,~.:t ~;~; 1r_J,, ~ ';;;--.-..... fl*~. ~._j!i '!~..... ~~'!~~~ ~~~~~#!~~~ ~~'!;!.:~~~f!q!i~~~ I ~ ~~

  • ! *.Tf~.:i:"}

-:--:~~-r~.-:~-=.. {; ~ **:. -;. ~ : --: ~ : -: ~ f ~~ ~"' i--.. -~ ~li!fS:;;ill *;l & i;f;;8J.1r.\\ ~~!Hi~~~~ !fl-i,;. tD,W ~ C'r,... -t1 (o,1,.::, fw V ~ -lh(J)atif.. !:fi~:! J ":~ ~ ~~ ~ f!.- r., '1.,: ~,.,.:i:.t.i* ~ 1 *,.. *. ~ ~* ;:( ~ ft.

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,i,l 0 t i i JR s:: !I CSD-EP-HNP-0101-06 Page 30 of 30 Revision 0

U.S. Nuclear Regulatory Commission Serial HNP-18-004, Enclosure 6 SERIAL HNP-18-004 ENCLOSURE 6 PROPOSED HARRIS NUCLEAR PLANT EMERGENCY ACTION LEVEL WALLCHART CHANGES (MARKUP) SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-63

CA1.1 Loss of RCS inventory as indicated by LI-403 or RCS standpipe level < - 82 in. CU1.1 UNPLANNED loss of reactor coolant results in RCS water level less than a required lower limit for ³ 15 min. (Note 1) RCS water level cannot be monitored AND EITHER UNPLANNED increase in any Table C-1 sump or tank due to a loss of RCS inventory Visual observation of UNISOLABLE RCS leakage CU1.2 RCS water level cannot be monitored for ³ 30 min. (Note 1) AND Core uncovery is indicated by any of the following: UNPLANNED increase in any Table C-1 sump or tank of sufficient magnitude to indicate core uncovery Containment radiation > 10,000 R/hr (RM-1CR-3589-SA or RM-1CR-3590-SB) Erratic source range monitor indication RCS water level cannot be monitored for ³ 30 min. (Note 1) AND Core uncovery is indicated by any of the following: UNPLANNED increase in any Table C-1 sump or tank of sufficient magnitude to indicate core uncovery Containment radiation > 10,000 R/hr (RM-1CR-3589-SA or RM-1CR-3590-SB) Erratic source range monitor indication AND Any Containment Challenge indication, Table C-2 RCS water level cannot be monitored for ³ 15 min. (Note 1) AND EITHER UNPLANNED increase in any Table C-1 sump or tank due to a loss of RCS inventory Visual observation of UNISOLABLE RCS leakage CA1.2 None Cold SD/ Refuel System Malfunct. Loss of Emergency AC Power Loss of all but one AC power source to emergency buses for 15 minutes or longer CU2.1 AC power capability, Table C-6, to emergency 6.9 KV buses 1A-SA and 1B-SB reduced to a single power source for ³ 15 min. (Note 1) AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS < 105 VDC bus voltage indications on Technical Specification required 125 VDC buses (DP-1A-SA, DP-1B-SB) for ³ 15 min. (Note 1) CU4.1 CG1.2 Loss of RCS inventory Loss of RCS inventory affecting core decay heat removal capability Loss of RCS inventory affecting fuel clad integrity with containment challenged RCS Level Loss of Comm. Loss of all onsite or offsite communications capabilities CU5.1 UNPLANNED loss of RCS inventory for 15 minutes or longer Loss of Vital DC power for 15 minutes or longer CA2.1 Loss of all offsite and all onsite AC power capability to 6.9 KV emergency buses 1A-SA and 1B-SB for ³ 15 min. (Note 1) Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer CS1.3 RCS Temp. UNPLANNED increase in RCS temperature to > 200°F CU3.1 UNPLANNED increase in RCS temperature Loss of all RCS temperature and RCS level indication for ³ 15 min. (Note 1) CU3.2 CA3.1 UNPLANNED increase in RCS temperature to > 200°F for > Table C-3 duration (Note 1) OR UNPLANNED RCS pressure increase > 10 psig (this does not apply during water-solid plant conditions) Inability to maintain plant in cold shutdown Hazardous Event Affecting Safety Systems C 1 3 5 6 The occurrence of any Table C-5 hazardous event AND EITHER: 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 2 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Loss of all Table C-4 onsite communication methods OR Loss of all Table C-4 offsite communication methods OR Loss of all Table C-4 NRC communication methods Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode CA6.1 EAL - COLD MODES 5, 6 & Defueled None None None None None None Loss of Vital DC Power 4 None None None 5 6 5 6 DEF 5 6 5 6 5 6 5 6 DEF 5 6 5 6 5 6 5 6 DEF 5 6 Table C-5 Hazardous Events Seismic event (earthquake) Internal or external FLOODING event High winds or tornado strike FIRE EXPLOSION Other events with similar hazard characteristics as determined by the Shift Manager

  • Containment sumps
  • PRT
  • RCDT
  • CCW surge tank
  • RAB sumps
  • RWST
  • RMWST
  • Recycle Holdup Tank Table C-1 Sumps / Tanks CS1.1 With CONTAINMENT CLOSURE not established, RCS level

< 70% RVLIS Full Range CS1.2 With CONTAINMENT CLOSURE established, RCS level < 63% RVLIS Full Range CG1.1 RCS level < 63% RVLIS Full Range for ³ 30 min. (Note 1) AND Any Containment Challenge indication, Table C-2

  • Containment Closure not established (Note 6)
  • Containment hydrogen concentration ³ 4%
  • Unplanned rise in Containment pressure Table C-2 Containment Challenge Indications 60 min.*

20 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 0 min. Table C-3 RCS Heat-up Duration Thresholds Not intact OR At REDUCED INVENTORY Intact (but not REDUCED INVENTORY) RCS Status Containment Closure Status Heat-up Duration N/A established not established Prepared for Duke Energy by: Operations Support Services, Inc. - www.ossi-net.com (1/21/16) Modes: 1 Power Operations Defueled DEF 2 Startup 5 Cold Shutdown 3 Hot Standby 4 Hot Shutdown Harris Nuclear Plant Classification of Emergency EP-EAL Matrix Revision 1 6 Refuel Table C-4 Communication Methods Onsite Offsite System PABX telephone (desk phones) HE&EC PABX telephone Site paging system Satellite phone DEMNET Radio communications networks NRC ETS phone NRC HPN phone X X X X NRC X X X X X X X X RVLIS Full Range Plant El. 260.62' 70% 63% 252.04' Standpipe 0" 89% 249.01' Reactor Vessel Flange 6 in. < Bottom of Hotleg Top of Active Fuel RCS Levels Note 1: The Emergency Coordinator 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: 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. If dose assessment results are available, declaration should be based on dose assessment instead of radiation monitor values. Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required. Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents. Note 8: A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. NOTES RHR Pump Operations - 82" 253.75' Table C-6 AC Power Sources Offsite - SUT 1A - SUT 1B - UAT 1A/1B backfed via Main Transformer (only if already aligned) Onsite - EDG 1A-SA - EDG 1B-SB Irradiated Fuel Event RU1.1 Sample analysis for a gaseous or liquid release indicates a concentration or release rate > 2 x ODCM limits for ³ 60 min. (Notes 1, 2) RU1.2 Reading on any Table R-1 effluent radiation monitor > column UE for ³ 60 min. (Notes 1, 2, 3) RA1.1 Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4) RS1.1 Reading on any Table R-1 effluent radiation monitor > column SAE for ³ 15 min. (Notes 1, 2, 3, 4) RS1.2 Dose assessment using actual meteorology indicates doses > 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4) RA1.2 Reading on any Table R-1 effluent radiation monitor > column ALERT for ³ 15 min. (Notes 1, 2, 3, 4) RA1.3 Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2) RU2.1 UNPLANNED water level drop in the REFUELING PATHWAY as indicated by low water level alarm or indication (LI-01SF-5101A/LI-01SF-5102A/LI-01SF-5103A, LI-403 or RCS standpipe) AND UNPLANNED rise in corresponding area radiation levels as indicated by any Table R-2 area radiation monitors RA2.2 Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by high alarm on any of the following: - Table R-2 refueling pathway area radiation monitors - 1REM-*1FL-3508A-SA, FHB Emergency Exhaust - 1REM-*1FL-3508B-SB, FHB Emergency Exhaust RA3.1 Dose rates > 15 mR/hr in EITHER of the following areas: Control Room (RM-21RR-3560-SA) OR Central Alarm Station (by survey) Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE Unplanned loss of water level above irradiated fuel Significant lowering of water level above, or damage to, irradiated fuel RA2.1 Uncovery of irradiated fuel in the REFUELING PATHWAY Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown RS1.3 Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: - Closed window dose rates > 100 mR/hr expected to con-tinue for ³ 60 min. - Analyses of field survey samples indicate thyroid CDE > 500 mrem for 60 min. of inhalation. (Notes 1, 2) Abnorm. Rad Levels / Rad Effluent R GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT 2 Rad Effluent 1 None None None None HS1.1 A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Supervision HG1.1 A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Supervision AND EITHER of the following has occurred: Any of the following safety functions cannot be controlled or maintained - Reactivity - Core Cooling - RCS heat removal OR Damage to spent fuel has occurred or is IMMINENT HU2.1 HU4.1 A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications (Note 1): - Report from the field (i.e., visual observation) - Receipt of multiple (more than 1) fire alarms or indications - Field verification of a single fire alarm AND The FIRE is located within any Table H-1 area HA5.1 Release of a toxic, corrosive, asphyxiant or flammable gas into any Table R-3/H-2 rooms or areas AND Entry into the room or area is prohibited or IMPEDED (Note 5) Seismic event greater than OBE levels HU4.2 Receipt of a single fire alarm (i.e., no other indications of a FIRE) AND The fire alarm is indicating a FIRE within any Table H-1 area AND The existence of a FIRE is not verified within 30 min. of alarm receipt (Note 1) HU3.2 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 FIRE potentially degrading the level of safety of the plant Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown Confirmed SECURITY CONDITION or threat HU7.1 Other conditions existing that in the judgment of the Emergency Coordinator warrant declaration of a UE Other conditions exist which in the judgment of the Emergency Coordinator indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs. An event has resulted in plant control being transferred from the Control Room to the Auxiliary Control Panel HA6.1 Control Room evacuation resulting in transfer of plant control to alternate locations HS6.1 An event has resulted in plant control being transferred from the Control Room to the Auxiliary Control Panel AND Control of any of the following key safety functions is not reestablished within 15 min. (Note 1): - Reactivity - Core Cooling - RCS heat removal Inability to control a key safety function from outside the Control Room Other conditions exist which, in the judgment of the Emergency Coordinator, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels. HA7.1 Other conditions exist that in the judgment of the Emergency Coordinator warrant declaration of an Alert HS7.1 Other conditions exist which in the judgment of the Emergency Coordinator indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY. Other conditions existing that in the judgment of the Emergency Coordinator warrant declaration of a Site Area Emergency HG7.1 Other conditions exist which in the judgment of the Emergency Coordinator indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area Other conditions exist which in the judgment of the Emergency Coordinator warrant declaration of a General Emergency None Hazards H Seismic Event Fire Control Room Evacuation Hazardous Gases Security EC Judgment 2 4 5 1 6 7 None HU3.3 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) Area Rad Levels 3 RA1.4 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 ³ 60 min. - Analyses of field survey samples indicate thyroid CDE > 50 mrem for 60 min. of inhalation. (Notes 1, 2) RG1.1 Reading on any Table R-1 effluent radiation monitor > column GE for ³ 15 min. (Notes 1, 2, 3, 4) RG1.2 Dose assessment using actual meteorology indicates doses > 1000 mrem TEDE or 5000 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4) RG1.3 Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: - Closed window dose rates > 1000 mR/hr expected to con-tinue for ³ 60 min. - Analyses of field survey samples indicate thyroid CDE > 5000 mrem for 60 min. of inhalation. (Notes 1, 2) RA3.2 An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table R-3/H-2 rooms or areas (Note 5) HU3.1 A tornado strike within the PROTECTED AREA Hazardous event HU3.4 A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles (Note 7) Natural or Tech. Hazard 3 HU4.3 A FIRE within the plant PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1) HU4.4 A FIRE within the plant PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish None None None None None None None RA2.3 Lowering of spent fuel pool level < 270.7 ft. (Level 2) RS2.1 Lowering of spent fuel pool level < 260.7 ft. (Level 3) RG2.1 Spent fuel pool level at the top of the fuel racks Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer Spent fuel pool level cannot be restored to at least 260.7 ft. (Level 3) for > 60 min. (Note 1) 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF Table R-2 Refueling Pathway AreaRadiation Monitors Containment RM-1CR-3561A-SA Containment Ventilation Isolation RM-1CR-3561B-SB Containment Ventilation Isolation RM-1CR-3561C-SA Containment Ventilation Isolation RM-1CR-3561D-SB Containment Ventilation Isolation Fuel Handling Building RM-1FR-3564A-SA Spent Fuel Pool SW, SE, SW RM-1FR-3564B-SB Spent Fuel Pool SW, SE, SE RM-1FR-3565A-SA Spent Fuel Pool SW, SE, SW RM-1FR-3565B-SB Spent Fuel Pool SW, SE, SE RM-1FR-3566A-SA Spent Fuel Pool NE, NW, NE RM-1FR-3566B-SB Spent Fuel Pool NW, NE, NW RM-1FR-3567A-SA Spent Fuel Pool NW, NE, NW RM-1FR-3567B-SB Spent Fuel Pool NE, NW, NE Seismic event > OBE as indicated by any of the following: ALB-10/4-4, SEISMIC MON SYS OBE EXCEEDED is ALARMED ALARM light on Seismic Switch Power Supply is LIT Any red alarm light is LIT on the Response Spectrum Annunciator [Refer to EAL CA6.1 OR SA9.1 for escalation due to seismic event] [Refer to EAL CA6.1 OR SA9.1 for escalation due to natural or technological hazard] [Refer to EAL CA6.1 OR SA9.1 for escalation due to FIRE] None Note 1: The Emergency Coordinator 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: 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. If dose assessment results are available, declaration should be based on dose assessment instead of radiation monitor values. Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required. Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents. Note 8: A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. NOTES Gaseous Liquid Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE Plant Vent 1.05E+8 mCi/sec Secondary Waste Sample Tank Discharge Turbine Building Treated Laundry & Hot Shower Tank Discharge Waste Monitor/Waste Evaporator Condensate Tank Discharge RM-21AV-3509-1SA RM-1TV-3536-1 REM-1WL-3540 REM-21WL-3541 REM-21WS-3542 4.60E+8 mCi/sec 1.05E+7 mCi/sec 4.60E+7 mCi/sec 1.05E+6 mCi/sec 4.60E+6 mCi/sec 1.14E+4 mCi/sec 1.38E+4 mCi/sec 7.02E-04 mCi/ml 1.97E-03 mCi/ml 7.02E-04 mCi/ml Waste Process Building Vent 5 RM-1WV-3546-1 2.49E+5 mCi/sec Waste Process Building Vent 5A RM-1WV-3547-1 1.45E+4 mCi/sec 7.74E+9 mCi/sec 7.76E+9 mCi/sec 7.74E+8 mCi/sec 7.76E+8 mCi/sec 7.75E+7 mCi/sec 7.76E+7 mCi/sec None Table R-3/H-2 Safe Operation & Shutdown Rooms/Areas Room / Area Mode(s) - RAB 190 (RHR pumps) - RAB 216 (BIT) - RAB 236 (CSIP, Primary Sample Sink, AFW pumps, CCW pumps and HX, Boric Acid Transfer Pumps, Mezzanine Area) - RAB 261 (RHR Heat Exchangers, Demin. Valve Gallery, VCT Valve Gallery) - RAB 286 (Switchgear) - Steam Tunnel - ESW intakes 4 1, 2, 3, 4, 5 1, 2, 3, 4, 5 1, 2, 3, 4, 5 1, 2, 3, 4, 5 1, 2, 3, 4 1, 2, 3, 4, 5

  • Containment
  • Reactor Auxiliary Building
  • Emergency Diesel Generator Building
  • Diesel Fuel Oil Storage Building (DFOST)
  • ESW Intake Structure and Auxiliary Reservoir Intake Structure Table H-1 Fire Areas HU1.1 A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Security Shift Supervision HA1.1 A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the Security Shift Supervision Notification of a credible security threat directed at the site A validated notification from the NRC providing information of an aircraft threat to the site HU1.2 HU1.3 A validated notification from NRC of an aircraft attack threat within 30 min. of the site HA1.2 HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes HOSTILE ACTION within the PROTECTED AREA HOSTILE ACTION resulting in loss of physical control of the facility Confirmed SECURITY CONDITION or threat Replace with: "A Containment Ventilation Isolation Radiation Monitor

> 2.6E+04 mR/hr (RM-1CR-3561A-SA, RM-1CR-3561B-SB, RM-1CR-3561C-SA, or RM-1CR-3561D-SB)"

Prepared for Duke Energy by: Operations Support Services, Inc. - www.ossi-net.com (1/21/16) Modes: 1 Power Operations Defueled DEF 2 Startup 5 Cold Shutdown 3 Hot Standby 4 Hot Shutdown Harris Nuclear Plant Classification of Emergency EP-EAL Matrix Revision 1 Irradiated Fuel Event RU1.1 Sample analysis for a gaseous or liquid release indicates a concentration or release rate > 2 x ODCM limits for ³ 60 min. (Notes 1, 2) RU1.2 Reading on any Table R-1 effluent radiation monitor > column UE for ³ 60 min. (Notes 1, 2, 3) RA1.1 Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4) RS1.1 Reading on any Table R-1 effluent radiation monitor > column SAE for ³ 15 min. (Notes 1, 2, 3, 4) RS1.2 Dose assessment using actual meteorology indicates doses > 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4) RA1.2 Reading on any Table R-1 effluent radiation monitor > column ALERT for ³ 15 min. (Notes 1, 2, 3, 4) RA1.3 Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2) RU2.1 UNPLANNED water level drop in the REFUELING PATHWAY as indicated by low water level alarm or indication (LI-01SF-5101A/LI-01SF-5102A/LI-01SF-5103A, LI-403 or RCS standpipe) AND UNPLANNED rise in corresponding area radiation levels as indicated by any Table R-2 area radiation monitors RA2.2 Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by high alarm on any of the following: - Table R-2 refueling pathway area radiation monitors - 1REM-*1FL-3508A-SA, FHB Emergency Exhaust - 1REM-*1FL-3508B-SB, FHB Emergency Exhaust RA3.1 Dose rates > 15 mR/hr in EITHER of the following areas: Control Room (RM-21RR-3560-SA) OR Central Alarm Station (by survey) Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE Unplanned loss of water level above irradiated fuel Significant lowering of water level above, or damage to, irradiated fuel RA2.1 Uncovery of irradiated fuel in the REFUELING PATHWAY Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown RS1.3 Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: - Closed window dose rates > 100 mR/hr expected to con-tinue for ³ 60 min. - Analyses of field survey samples indicate thyroid CDE > 500 mrem for 60 min. of inhalation. (Notes 1, 2) Abnorm. Rad Levels / Rad Effluent R GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT 2 Rad Effluent 1 None None None None HS1.1 A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Supervision HG1.1 A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Supervision AND EITHER of the following has occurred: Any of the following safety functions cannot be controlled or maintained - Reactivity - Core Cooling - RCS heat removal OR Damage to spent fuel has occurred or is IMMINENT HU1.1 A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Security Shift Supervision HA1.1 HU2.1 HU4.1 A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications (Note 1): - Report from the field (i.e., visual observation) - Receipt of multiple (more than 1) fire alarms or indications - Field verification of a single fire alarm AND The FIRE is located within any Table H-1 area HA5.1 Release of a toxic, corrosive, asphyxiant or flammable gas into any Table R-3/H-2 rooms or areas AND Entry into the room or area is prohibited or IMPEDED (Note 5) Seismic event greater than OBE levels HU4.2 Receipt of a single fire alarm (i.e., no other indications of a FIRE) AND The fire alarm is indicating a FIRE within any Table H-1 area AND The existence of a FIRE is not verified within 30 min. of alarm receipt (Note 1) HU3.2 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 FIRE potentially degrading the level of safety of the plant Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown Confirmed SECURITY CONDITION or threat HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes HOSTILE ACTION within the PROTECTED AREA HOSTILE ACTION resulting in loss of physical control of the facility HU7.1 Other conditions existing that in the judgment of the Emergency Coordinator warrant declaration of a UE Other conditions exist which in the judgment of the Emergency Coordinator indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs. An event has resulted in plant control being transferred from the Control Room to the Auxiliary Control Panel HA6.1 Control Room evacuation resulting in transfer of plant control to alternate locations HS6.1 An event has resulted in plant control being transferred from the Control Room to the Auxiliary Control Panel AND Control of any of the following key safety functions is not reestablished within 15 min. (Note 1): - Reactivity - Core Cooling - RCS heat removal Inability to control a key safety function from outside the Control Room Other conditions exist which, in the judgment of the Emergency Coordinator, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels. HA7.1 Other conditions exist that in the judgment of the Emergency Coordinator warrant declaration of an Alert HS7.1 Other conditions exist which in the judgment of the Emergency Coordinator indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY. Other conditions existing that in the judgment of the Emergency Coordinator warrant declaration of a Site Area Emergency HG7.1 Other conditions exist which in the judgment of the Emergency Coordinator indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area Other conditions exist which in the judgment of the Emergency Coordinator warrant declaration of a General Emergency None Hazards H Seismic Event Fire Control Room Evacuation Hazardous Gases Security EC Judgment 2 4 5 1 6 7 None HU3.3 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) A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the Security Shift Supervision GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Area Rad Levels 3 RA1.4 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 ³ 60 min. - Analyses of field survey samples indicate thyroid CDE > 50 mrem for 60 min. of inhalation. (Notes 1, 2) RG1.1 Reading on any Table R-1 effluent radiation monitor > column GE for ³ 15 min. (Notes 1, 2, 3, 4) RG1.2 Dose assessment using actual meteorology indicates doses > 1000 mrem TEDE or 5000 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4) RG1.3 Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: - Closed window dose rates > 1000 mR/hr expected to con-tinue for ³ 60 min. - Analyses of field survey samples indicate thyroid CDE > 5000 mrem for 60 min. of inhalation. (Notes 1, 2) RA3.2 An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table R-3/H-2 rooms or areas (Note 5) HU3.1 A tornado strike within the PROTECTED AREA Hazardous event HU3.4 A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles (Note 7) Natural or Tech. Hazard 3 HU4.3 A FIRE within the plant PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1) HU4.4 A FIRE within the plant PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish None None None None None None None Table F-1 Fission Product Barrier Threshold Matrix Containment (CNMT) Barrier Fuel Clad (FC) Barrier Reactor Coolant System (RCS) Barrier Loss Potential Loss Loss Loss Potential Loss Potential Loss A. RCS or SG Tube Leakage B. Inadequate Heat Removal C. CNMT Radiation / RCS Activity D. CNMT Integrity or Bypass None None None None None

1. Any condition in the opinion of the Emergency Coordinator that indicates loss of the Fuel Clad barrier
1. Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the Fuel Clad barrier
1. An automatic or manual ECCS (SI) actuation required by EITHER:
  • UNISOLABLE RCS leakage
  • SG tube RUPTURE
1. Any condition in the opinion of the Emergency Coordinator that indicates loss of the RCS barrier
1. Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the RCS barrier
1. Any condition in the opinion of the Emergency Coordinator that indicates loss of the Containment barrier
1. Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the Containment barrier None None None E. EC Judgment
1. Operation of a standby charging pump is required due to EITHER:
  • UNISOLABLE RCS leakage
  • SG tube leakage
2. CSFST Integrity-RED Path entry conditions met
1. A leaking or RUPTURED SG is FAULTED outside of containment
1. CSFST Core Cooling-RED Path entry conditions met
1. CSFST Core Cooling-ORANGE Path entry conditions met
2. CSFST Heat Sink-RED Path entry conditions met AND Heat Sink is required None
1. CSFST Heat Sink-RED Path entry conditions met AND Heat Sink is required
1. CSFST Core Cooling-RED Path entry conditions met AND Restoration procedures not effective within 15 min. (Note 1)
1. Containment radiation >150 R/hr (RM-1CR-3589-SA or RM-1CR-3590-SB)
2. Dose equivalent I-131 coolant activity > 300 Ci/gm
1. Containment Leak Detection Monitor Noble Gas (REM-1LT-3502A-SA) > 8.3E-3 Ci/ml
1. Containment radiation > 600 R/hr (RM-1CR-3589-SA or RM-1CR-3590-SB)

None None None

1. Containment isolation is required AND EITHER Containment integrity has been lost based on Emergency Coordinator judgment UNISOLABLE pathway from Containment to the environment exists
2. Indications of RCS leakage outside of containment
1. CSFST Containment-RED Path entry conditions met
2. Containment hydrogen concentration > 4%
3. Containment pressure > 10 psig with

< one full train of depressurization equipment operating (one CNMT spray pump and two CNMT fan coolers) per design for > 15 min.(Note 1) System Malfunct. SA1.1 AC power capability, Table S-5, to 6.9 KV emergency buses 1A-SA and 1B-SB reduced to a single power source for ³ 15 min. (Note 1) AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS SG1.1 None None Fission Product Barriers FS1.1 Loss of any two barriers AND Loss or potential loss of third barrier (Table F-1) Loss or potential loss of any two barriers (Table F-1) FA1.1 Any loss or any potential loss of either Fuel Clad or RCS (Table F-1) FG1.1 SS1.1 Loss of Emergency AC Power Loss of all offsite AC power capability to emergency buses for 15 minutes or longer SU1.1 Loss of all offsite AC power capability, Table S-5, to 6.9 KV emergency buses 1A-SA and 1B-SB for ³ 15 min. (Note 1) Loss of all but one AC power source to emergency buses for 15 minutes or longer Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer Prolonged loss of all offsite and all onsite AC power to emergency buses Loss of all 125 VDC power based on battery bus voltage indications < 105 VDC on both emergency DC buses (DP-1A-SA, DP-1B-SB) for ³ 15 min. (Note 1) SS2.1 Loss of all vital DC power for 15 minutes or longer SA6.1 An automatic or manual trip fails to shut down the reactor as indicated by reactor power ³ 5% AND Manual trip actions taken at the reactor control console (actuation of MCB Reactor Trip Switch #1, #2 or MCB Turbine Trip switch) are not successful in shutting down the reactor as indicated by reactor power ³ 5% (Note 8) An automatic or manual trip fails to shut down the reactor as indicated by reactor power ³ 5% AND All actions to shut down the reactor are not successful as indicated by reactor power ³ 5% AND EITHER: Core Cooling RED Path entry conditions met Heat Sink RED Path entry conditions met SS6.1 Automatic or manual trip fails to shut down the reactor and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor Inability to shut down the reactor causing a challenge to core cooling or RCS heat removal Loss of all onsite or offsite communications capabilities SU7.1 Loss of CR Indications UNPLANNED loss of Control Room indications for 15 minutes or longer SU3.1 An UNPLANNED event results in the inability to monitor one or more Table S-1 parameters from within the Control Room for ³ 15 min. (Note 1) SA3.1 An UNPLANNED event results in the inability to monitor one or more Table S-1 parameters from within the Control Room for ³ 15 min. (Note 1) AND Any significant transient is in progress, Table S-2 UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode SA9.1 The occurrence of any Table S-4 hazardous event AND EITHER: 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 RCS activity greater than Technical Specification allowable limits SU4.1 RCS activity > Technical Specification Section 3.4.8 limits RCS leakage for 15 minutes or longer SU5.1 RCS unidentified or pressure boundary leakage > 10 gpm for ³ 15 min. OR RCS identified leakage > 25 gpm for ³ 15 min. OR Leakage from the RCS to a location outside containment > 25 gpm for ³ 15 min. (Note 1) Automatic or manual trip fails to shut down the reactor SU6.1 None Loss of all offsite and all onsite AC power capability to 6.9 KV emergency buses 1A-SA and 1B-SB for ³ 15 min. (Note 1) Loss of all offsite and all onsite AC power capability to 6.9 KV emergency buses 1A-SA and 1B-SB AND EITHER: - Restoration of at least one emergency bus in < 4 hours is not likely - Core Cooling RED Path entry conditions met F S 1 3 9 Loss of Comm. 7 An automatic trip did not shut down the reactor as indicated by reactor power ³ 5% after any RPS setpoint is exceeded AND A subsequent automatic trip or manual trip action taken at the reactor control console (actuation of MCB Reactor Trip Switch #1, #2 or MCB Turbine Trip switch) is successful in shutting down the reactor as indicated by reactor power < 5% (Note 8) None None None None None Loss of Vital DC Power 2 EAL-HOT MODES 1, 2, 3 & 4 SG1.2 Loss of all offsite and all onsite AC power capability to 6.9 KV emergency buses 1A-SA and 1B-SB for ³ 15 min. AND Loss of all 125 VDC power based on battery bus voltage indications < 105 VDC on both emergency DC buses (DP-1A-SA, DP-1B-SB) for ³ 15 min. (Note 1) None RCS Activity 4 RPS Failure 6 RCS Leakage 5 None None None SU6.2 A manual trip did not shut down the reactor as indicated by reactor power ³ 5% after any manual trip action was initiated AND A subsequent automatic trip or manual trip action taken at the reactor control console (actuation of MCB Reactor Trip Switch

  1. 1, #2 or MCB Turbine Trip switch) is successful in shutting down the reactor as indicated by reactor power < 5% (Note 8)

Loss of all Table S-3 onsite communication methods OR Loss of all Table S-3 offsite communication methods OR Loss of all Table S-3 NRC communication methods Hazardous Event Affecting Safety Systems None Table S-1 Safety System Parameters - Reactor power - RCS level - RCS pressure - Core exit T/C temperature - Level in at least one S/G - Auxiliary feed flow in at least one S/G RA2.3 Lowering of spent fuel pool level < 270.7 ft. (Level 2) RS2.1 Lowering of spent fuel pool level < 260.7 ft. (Level 3) RG2.1 Spent fuel pool level at the top of the fuel racks Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer Spent fuel pool level cannot be restored to at least 260.7 ft. (Level 3) for > 60 min. (Note 1) 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 5 6 1 2 3 4 DEF 6 Refuel 1 2 3 4 1 1 2 3 4 1 2 3 4 1 2 3 4 1 2 3 4 1 2 3 4 1 2 3 4 1 2 3 4 1 2 3 4 1 1 1 2 3 4 Failure to isolate containment or loss of containment pressure control SU8.1 EITHER: Any penetration is not isolated within 15 min. of a VALID containment isolation signal OR Containment pressure > 10 psig with < one full train of depressurization equipment operating (one CNMT spray pump and two CNMT fan coolers) per design for > 15 min. (Note 1) 1 2 3 4 1 2 3 4 Table S-4 Hazardous Events Seismic event (earthquake) Internal or external FLOODING event High winds or tornado strike FIRE EXPLOSION Other events with similar hazard characteristics as determined by the Shift Manager CNMT Failure 8 None None Loss of all emergency AC and vital DC power sources for 15 minutes or longer 1 2 3 4 1 2 3 4 1 2 3 4 1 2 3 4 None Table R-2 Refueling Pathway AreaRadiation Monitors Containment RM-1CR-3561A-SA Containment Ventilation Isolation RM-1CR-3561B-SB Containment Ventilation Isolation RM-1CR-3561C-SA Containment Ventilation Isolation RM-1CR-3561D-SB Containment Ventilation Isolation Fuel Handling Building RM-1FR-3564A-SA Spent Fuel Pool SW, SE, SW RM-1FR-3564B-SB Spent Fuel Pool SW, SE, SE RM-1FR-3565A-SA Spent Fuel Pool SW, SE, SW RM-1FR-3565B-SB Spent Fuel Pool SW, SE, SE RM-1FR-3566A-SA Spent Fuel Pool NE, NW, NE RM-1FR-3566B-SB Spent Fuel Pool NW, NE, NW RM-1FR-3567A-SA Spent Fuel Pool NW, NE, NW RM-1FR-3567B-SB Spent Fuel Pool NE, NW, NE Seismic event > OBE as indicated by any of the following: ALB-10/4-4, SEISMIC MON SYS OBE EXCEEDED is ALARMED ALARM light on Seismic Switch Power Supply is LIT Any red alarm light is LIT on the Response Spectrum Annunciator Table S-3 Communication Methods Onsite Offsite System PABX telephone (desk phones) HE&EC PABX telephone Site paging system Satellite phone DEMNET Radio communications networks NRC ETS phone NRC HPN phone X X X X NRC X X X X X X X X [Refer to EAL CA6.1 OR SA9.1 for escalation due to seismic event] [Refer to EAL CA6.1 OR SA9.1 for escalation due to natural or technological hazard] [Refer to EAL CA6.1 OR SA9.1 for escalation due to FIRE] SU4.2 Valid Gross Failed Fuel Detector (RS-7411A) high alarm (> 1E+04 cpm) None Note 1: The Emergency Coordinator 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: 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. If dose assessment results are available, declaration should be based on dose assessment instead of radiation monitor values. Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required. Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents. Note 8: A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. NOTES Note 1: The Emergency Coordinator 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: 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. If dose assessment results are available, declaration should be based on dose assessment instead of radiation monitor values. Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required. Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents. Note 8: A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. NOTES None Table S-2 Significant Transients - Reactor trip - Runback > 25% thermal power - Electrical load rejection > 25% electrical load - Safety injection actuation [Refer to fission product barrier EALs for escalation due to fuel clad failures] [Refer to fission product barrier EALs for escalation due to RCS leakage] Table R-3/H-2 Safe Operation & Shutdown Rooms/Areas Room / Area Mode(s) - RAB 190 (RHR pumps) - RAB 216 (BIT) - RAB 236 (CSIP, Primary Sample Sink, AFW pumps, CCW pumps and HX, Boric Acid Transfer Pumps, Mezzanine Area) - RAB 261 (RHR Heat Exchangers, Demin. Valve Gallery, VCT Valve Gallery) - RAB 286 (Switchgear) - Steam Tunnel - ESW intakes 4 1, 2, 3, 4, 5 1, 2, 3, 4, 5 1, 2, 3, 4, 5 1, 2, 3, 4, 5 1, 2, 3, 4 1, 2, 3, 4, 5 None Table S-5 AC Power Sources Offsite - SUT 1A - SUT 1B - UAT 1A/1B backfed via Main Transformer (only if already aligned) Onsite - UAT 1A/1B via Main Generator - EDG 1A-SA - EDG 1B-SB Gaseous Liquid Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE Plant Vent 1.05E+8 mCi/sec Secondary Waste Sample Tank Discharge Turbine Building Treated Laundry & Hot Shower Tank Discharge Waste Monitor/Waste Evaporator Condensate Tank Discharge RM-21AV-3509-1SA RM-1TV-3536-1 REM-1WL-3540 REM-21WL-3541 REM-21WS-3542 4.60E+8 mCi/sec 1.05E+7 mCi/sec 4.60E+7 mCi/sec 1.05E+6 mCi/sec 4.60E+6 mCi/sec 1.14E+4 mCi/sec 1.38E+4 mCi/sec 7.02E-04 mCi/ml 1.97E-03 mCi/ml 7.02E-04 mCi/ml Waste Process Building Vent 5 RM-1WV-3546-1 2.49E+5 mCi/sec Waste Process Building Vent 5A RM-1WV-3547-1 1.45E+4 mCi/sec 7.74E+9 mCi/sec 7.76E+9 mCi/sec 7.74E+8 mCi/sec 7.76E+8 mCi/sec 7.75E+7 mCi/sec 7.76E+7 mCi/sec Notification of a credible security threat directed at the site A validated notification from the NRC providing information of an aircraft threat to the site HU1.2 HU1.3 A validated notification from NRC of an aircraft attack threat within 30 min. of the site HA1.2 Confirmed SECURITY CONDITION or threat

  • Containment
  • Reactor Auxiliary Building
  • Emergency Diesel Generator Building
  • Diesel Fuel Oil Storage Building (DFOST)
  • ESW Intake Structure and Auxiliary Reservoir Intake Structure Table H-1 Fire Areas Revise to "1. (RM-1CR-3589SA or RM-1CR-3590SB) > Table F-2 Column FC Barrier Loss" Revise to 1. " (RM-1CR-3561A-SA, RM-1CR-3561B-SB, RM-1CR-3561C-SA, or RM-1CR-3561D-SB) > Table F-2 Column RCS Barrier Loss" Revise to "1. (RM-1CR-3589SA or RM-1CR-3590SB) > Table F-2 Column CNMT Potential Loss" Add INSERT A

INSERT A: Table F-2 Containment Radiation Time After S/D (Hours) FC Barrier Loss R/hr RCS Barrier Loss mR/hr CNMT Potential Loss R/hr 0 - 1 130 1.37E+03 2360 1 - 2 110 1.12E+03 2000 2 - 8 70 6.35E+02 1300 > 8 21 1.37E+02 390}}

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