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Enclosure 1 - Millstone, Units 2 and 3 and ISFSI; North Anna, Units 1 and 2 and ISFSI; and Surry, Units 1 and 2 and ISFSI - Response to EAL Scheme Change RAIs
	ML19249B767
Person / Time
Site:	Millstone, Surry, North Anna, 07200002, 07200055
Issue date:	08/29/2019
From:	; Dominion Energy Nuclear Connecticut, Virginia Electric & Power Co (VEPCO)
To:	; Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
References
	19-296
	Download: ML19249B767 (86)
	v • d • e

Serial No.19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 ENCLOSURE 1 RESPONSE TO EAL SCHEME CHANGE RAls Dominion Energy Nuclear Connecticut, Inc. (DENC)

Virginia Electric and Power Company (Dominion Energy Virginia)

Millstone Power Station Units 2 and 3 and ISFSI North Anna Power Station Units 1 and 2 and ISFSI Surry Power Station Units 1 and 2 and ISFSI

Serial No.19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 ENCLOSURE 1 Attachment 1 RAI RESPONSE MATRIX Dominion Energy Nuclear Connecticut, Inc. (DENC)

Virginia Electric and Power Company (Dominion Energy Virginia)

Millstone Power Station Units 2 and 3 and ISFSI North Anna Power Station Units 1 and 2 and ISFSI Surry Power Station Units 1 and 2 and ISFSI

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Station IC/EAL As described in the license amendment request (LAR) dated The proposed Table R 1, "Unit [1, 2, or 3 as applicable]

January 4, 2019 (Agencywide Documents Access and Gaseous Effluent Monitor Classification Thresholds,"

Management System (ADAMS) Accession Number that is used for EALs RU1, RA1, RS1, and RG1 ML19011A237}, certain release points within the Dominion Notification of Unusual Event (NOUE) thresholds, are Energy fleet generate incongruent Unusual Event (UE) EAL based on a dose that considers only a fraction of the thresholds compared to calculated ALERT EAL thresholds that U.S. Environmental Protection Agency Early Phase are based on offsite dose of 10 mrem TEDE. These UE EALs Protective Action Guides rather than a low level were calculated following the NEI 99-01, Revision 6 guidance of radiological release that exceeds regulatory "2 times the site-specific effluent release controlling document commitments for an extended period (e.g., an limits for 60 minutes or longer". For some release points, using uncontrolled release).

the Offsite Dose Calculation Manual (ODCM) (site-specific Please explain what features, that are unique to effluent release controlling document) methods and limits to Dominion facilities, require a deviation from the NRG- determine the UE EALs resulted in calculated UE values that endorsed EAL scheme or provide threshold values that were greater than ALERT EAL threshold values or did not MPS2 are consistent with NEI 99 01, Revision 6, such that an provide a factor of 10 separation from the ALERT EAL MPS3 NOUE would be declared for a low level radiological threshold. Therefore, an alternative approach was proposed Table R-1 NAPS release that exceeds regulatory commitments for an for U E EALs based on dose 10 times lower than the ALERT 10 SPS extended period (e.g., an uncontrolled release). mrem limit. Following discussion with NRC staff and understanding that this alternative approach will constitute a significant deviation from the NRC approved guidance in Revision 6 of NEI 99-01, Dominion Energy has revised the alternative approach for these gaseous effluent releases.

Specifically, the Initiating Condition (IC) for gaseous effluent pathways has been changed to be the same for both the liquid and gaseous effluent pathways, as follows:

"Release of gaseous or liquid radioactivity greater than 2 times the allocated ODCM limits for 60 minutes or longer."

This method will provide a justifiable basis for NOUE thresholds based on established methods and setpoints provided in the

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Station IC/EAL facility ODCM. The proposed NOUE values will classify events based on degradation in the level of safety of the plant and will maintain a near linear escalation between all four classification levels (i.e., NOUE, ALERT, Site Area Emergency (SAE) and General Emergency (GE)). The revised IC for gaseous effluent pathways is the same IC definition currently used for gaseous pathways in the North Anna and Surry NEI 99-01, Revision 4 EALs.

The Technical Bases Documents for MPS2/MPS3, NAPS and SPS have been updated to change the Initiating Condition (IC) for gaseous effluent pathways to be the same for both liquid and gaseous effluent pathways, as shown in Enclosures 3, 4 and 5, respectively.

A review of Calculation RP-18-08 shows that NAPS and Dominion Energy has verified the calculation of each effluent SPS each have one gaseous radiation monitor that has release UE EAL threshold following the NEI 99-01, Revision 6 Offsite Dose Calculation Manual (ODCM) based guidance of "2 times the site-specific effluent release controlling setpoints that are not consistent with other Dominion document limits for 60 minutes or longer". Specifically, NAPS gaseous radiation monitor setpoints or other Dominion radiation monitor GW-RM-178 and SPS radiation monitor GW-gaseous radiation monitor setpoints that were provided RM-130 were confirmed to result in NOUE threshold values that as threshold values. The outlying monitors have ODCM would exceed calculated ALERT threshold values.

NAPS based setpoints that would be greater than the 2 Table R-1 Following the alternative approach discussed in the response to SPS proposed Alert threshold values, which are RAI 1 above, gaseous NOUE thresholds have been determined approximately an order of magnitude higher that the using an 'allocated' fraction of the ODCM limits for each station.

current threshold values. These limits should roughly Use of this alternative approach allows NOUE threshold values correspond to 500 mrem/year (approximately .06 mrem per hour). As such, it appears that the alarm setpoints to utilize administrative controls established to prevent unintentional releases, and to control and monitor intentional for these effluent flow paths are above both the releases. Allocation factors were determined by maintaining a technical specification and Alert threshold values.

release rate at least an order of magnitude lower than Therefore, the NRG staff could not determine how the corresponding ALERT EAL release rates for each release

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Station IC/EAL setpoint that corresponds to greater than 10 mrem per pathway. Comparing that release rate against ODCM release hour was obtained. rate limits determines the fraction of ODCM (allocation) credited in each NOUE EAL value. Radiation monitor setpoints will be

a. Please verify that those gaseous radiation adjusted to control releases and correspond to this NOUE IC monitors (NAPS GW RM-178 and SPS GW RM radiological release criteria. This approach is the basis of the 130), which would result in NOUE setpoints that current NEI 99-01, Revision 4 EAL scheme at the stations, and are greater than the Alert threshold value, were was submitted and approved as an acceptable deviation on correctly calculated as directed by the site 1/28/11 by the NRG (see ADAMS ML103220114). This specific ODCM, and revise setpoints if justified.

approach has also been applied to the MPS2 and MPS3 EAL

b. If Dominion determines that gaseous radiation schemes.

monitor setpoints for NAPS GW RM-178 and Controls set within the station effluent setpoint determination SPS GW RM 130 were correctly calculated, process will assure radiation monitor alarm indications occur please provide threshold values that would alert prior to and/or at the UE EAL limit (i.e., any effluent control the operators of excessive effluent levels that alarm indications will apprise Operators of changing radiological are approximately an order of magnitude less conditions).

than the threshold values for an Alert classification level.

The proposed Table R 1, "Unit [1, 2, or 3 as applicable] Dominion Energy has eliminated the main steam radiation Gaseous Effluent Monitor Classification Thresholds," monitors used to classify safety/PORV steam releases and that is used for EALs RA1, RS1, and RG1 has auxiliary feedwater (Terry Turbine) exhausts based on threshold values based on main steam line radiation abnormal radiation levels from Table R-1. This elimination is MPS2 monitors. Dominion does not propose main steam line based on the fact that main steam radiation monitors are not a MPS3 radiation monitor threshold values for EAL RU1 normal effluent pathway and therefore are not included in the 3 Table R-1 because there are no ODCM limits on the main steam MPS2, MPS3, NAPS or SPS ODCM.

NAPS SPS and/or the auxiliary feedwater exhausts and the limited The Technical Bases Documents for MPS2/MPS3, NAPS and ability for the respective radiation monitors to detect low SPS EALs EAL RU1 have been updated to remove these level radioactivity. The steam line monitors are typically radiation monitors from Table R-1, as shown in Enclosures 3, 4 installed to provide information relative to steam and 5, respectively.

generator tube leakage while operating at power. This information includes which steam generator, if any, is

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Station IC/EAL the most affected and a relative quantity of steam generator tube leakage. In using threshold values for EALs RA1, RS1, and RG1 based on main steam line monitor radiation readings, the licensee must make assumptions about both the flow rate and the isotopic mix at the time of the accident. These assumptions could adversely impact the accuracy of the threshold value.

Please explain how the main steam radiation monitors can provide an accurate offsite dose indication by using an assumed steam flow, that could change by several orders of magnitude post-accident, concurrent with detectors that may not accurately assess the post accident isotopic mix or consider removing the steam generator discharge radiation monitors from the proposed Dominion EAL Schemes.

The proposed Table R 1, "Unit [1, 2, or 3 as applicable] The tables provided in Enclosure 1, Attachment 2 shows Gaseous Effluent Monitor Classification Thresholds," comparisons between current EAL values and new proposed that is used for EALs RA1, RS1, and RG1 have Revision 6 EALs for each station. Provided below each table is threshold values that have substantially changed from a brief description of what method, assumption or parameter MPS2 the current threshold values. Enclosure 7, "Summary of causes a difference between the proposed values from current MPS3 Calculations," of the letter dated January 4, 2019, values.

4 Table R-1 provides that the Meteorological Information and Dose NAPS SPS Assessment System (MIDAS) was used to determine the projected threshold values. However, it was not clear if the methodology used in determining the proposed threshold values is the same methodology currently being used for dose assessment. As such, the basis for the change in the threshold values for

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Station IC/EAL EALs RA1, RS1, and RG1 were not apparent to the NRC staff.

Please provide an explanation that supports the changes to the threshold values in the for EALs RA 1, RS1, and RG1 from their existing EAL values to the proposed EAL values. This explanation should address the fact that MIDAS was used to determine the proposed EAL threshold values and appears to be the same program that is currently used for dose assessment.

The proposed threshold values for EAL RU1 do not MPS2:

include the below effluent instruments that are identified Discharge from the Waste Gas Decay Tank, monitored by RM-in the respective unit's offsite dose calculation manuals.

9095, is aligned to the site stack ventilation pathway from MPS2

MPS Unit 2: Waste Gas Decay Tank Monitor which is monitored by the site stack downstream radiation (RM9095), Reactor Building Closed Cooling monitor RM-8169. Therefore, RM-9095 does not need to be Water (RM6038); included as an effluent pathway monitor.

MPS Unit 3: Engineering Safeguards Building RM-6038 is included in the REMODCM because MPS2 has no MPS2 Monitor (HVQ RE49); service water radiation monitor to act as a final effluent monitor MPS3 5 RU1 for service water. The Reactor Building Closed Cooling Water NAPS

NAPS: Liquid Radwaste Effluent Line (1-LW-RM-(RBCCW) radiation monitor, RM-6038, is a leak-detection SPS 111 ), Steam Generator High Capacity Slowdown monitor that monitors closed cooling water, not service water.

Line (1 (2) SW RM 130(230)), Condenser Air The purpose of RBCCW radiation monitor RM-6038 is to warn Ejector (1 (2) SV RM 121 (221 )); and of abnormal radioactivity in the RBCCW system to prevent

SPS: Service Water Effluent Line (1 SW RM 107 releases to the Service Water system which could result in A, B, C, and D), Radwaste Facility Effluent Line exceeding Technical Specification release limits. An alarm (RE-RRM-131 ), Condenser Air Ejector (1 (2) SV generated from RM-6038 will alert operations personnel to take RM 121 (221 )), Ventilation Vent No. 1 (1-VG-RM- action to prevent releases into Service Water from the RBCCW 104), Radwaste Facility Vent (RRM-101 ). system.

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Station IC/EAL Please revise the proposed EAL RU 1 threshold values MPS3:

to include the above effluent instruments or provide Ventilation releases from the Engineered Safeguards Facility justification that supports excluding those instruments (ESF) building are monitored by radiation monitor HVQ49.

from the proposed EAL scheme. For example, some Upon initiation of a Safety Injection signal, ESF building effluent flow path instrumentation may be retired in ventilation is isolated and re-directed to the site stack, place.

monitored by radiation monitor HVR19. Dominion Energy agrees that HVQ RE49 should be included as an effluent radiation monitor to declare an Unusual Event, as releases through this pathway could exist prior to isolation. ALERT, SAE, and GE thresholds are not required, as this pathway will be isolated upon an accident. Inclusion of HVQ RE49 is shown in the comparison table provided in RAI 4 response for Millstone

3. The NOUE EAL threshold for HVQ RE49 is calculated using NEI 99-01, Revision 6 guidance of 2 times the site-specific effluent release controlling document limits for 60 minutes or longer.

NAPS:

The Liquid Radwaste Effluent line, 1-LW-RM-111; Steam Generator High Capacity Slowdown lines, SS-RM-125(225);

and Service Water System Effluent line, SW-RM-108, discharge into the circulating water tunnel and are upstream of the Circulating Water Discharge Tunnel radiation monitors, SW-RM-130(230). As such, they are not included as EAL radiation monitors.

The Condenser Air Ejector radiation monitors, SV RM-121 (221 ),

are normally aligned to Ventilation Stack 'A' and are monitored by VG-RM-179. They are not accident radiation monitors.

Upon receipt of a Hi-Hi alarm on RM-121 (221 ), air ejector flow is diverted into Containment. The Hi-Hi alarm on RM-121 (221)

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Station IC/EAL is set appreciably below the ODCM setpoint limit. Therefore, the effluent will be diverted into Containment well before meeting the criteria to classify an Unusual Event where levels need to exceed 2 times the ODCM limit for 60 minutes. Based on these facts, the condenser air ejector radiation monitors are not included as EAL radiation monitors.

SPS:

The Service Water Effluent line radiation monitors, 1-SW-RM-107 A, B, C, D, and the Radwaste Facility Effluent line radiation monitor, RE-RRM-131, discharge into the Circulating Water Discharge Tunnel upstream of the respective Discharge Tunnel radiation monitors, SW-RM-120(220). As such, they are not included as EAL radiation monitors.

The Condenser Air Ejector radiation monitors, SV-RM-111 (211 ), divert air ejector flow into Containment upon receipt of an alarm. The alarm setpoint is set appreciably below the ODCM setpoint limit. Therefore, the effluent will be diverted into Containment well before meeting the criteria to classify an Unusual Event where levels need to exceed 2 times the ODCM limit for 60 minutes. Based on this fact, the Condenser Air Ejector radiation monitors are not included as EAL radiation monitors.

Inputs to Ventilation Vent No. 1 radiation monitor, VG-RM-104, are from the Service Building which houses the Chemistry Lab and RP Hoods and general area ventilation in the dosimetry and personnel decontamination area. Ventilation Vent No. 1 is a final release point and therefore, VG-RM-104 is required as part of the ODCM. However, the source of radioactive material to this pathway is limited with insufficient potential or quantity to

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Station IC/EAL create a release rate sufficient to meet the criteria to classify an Unusual Event where levels need to exceed 2 times the ODCM limit for 60 minutes. As such, radiation monitor VG-RM-104 is not included as an EAL radiation monitor.

Surry Power Statiqn processes various quantities and types of radwaste in the Radwaste Facility. Ventilation from the Radwaste Facility exhausts directly to the environment, monitored by radiation monitor RRM-101. The facility is designed with a fire protection sprinkler system to detect and arrest any potential fire. Due to the limited source of radioactive material capable of being released airborne from this facility and the protection features designed to extinguish potential fire sources, there is insufficient risk, potential or quantity of radioactive material to create a protracted release rate exceeding 2 times the ODCM limit for 60 minutes to meet Unusual Event criteria. As such, radiation monitor RRM-101 is not included as an EAL radiation monitor.

The cited note is applicable when the RCS is not intact or in The threshold value for EAL CA3.1 includes a note that reduced inventory provided containment closure is established indicates the EAL is not applicable if heat removal is in and the heat-up duration of 20 minutes has not been exceeded operation and reactor coolant system (RCS) (Table C-5).

temperature is being reduced within the applicable heat MPS2 The note is not applicable under conditions when containment up duration. This is not appropriate when the RCS is MPS3 closure is not established because the Table C-5 heat-up 6 CA3:1 not intact or in reduced inventory.

NAPS duration is O minutes and thus escalation to the Alert is SPS Please explain how the Emergency Director would not immediate.

potentially apply the note indicating that EAL CA3.1 is not applicable if a RCS heat removal system is in operation and reducing RCS temperature to high temperature conditions, when the RCS is not intact or in

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Station IC/EAL reduced inventory.

The threshold value for EAL CA3.1 does not include a The CA3.1 EAL thresholds submitted in the LAR dated January note indicating that the threshold value for an increase 4, 2019 for MPS3, NAPS and SPS include the parenthetical MPS2 in RCS pressure does not apply when the RCS is in statement excluding water solid plant conditions.

MPS3 water solid conditions.

7 CA3.1 As justified in the MPS2 Comparison Matrix in the same NAPS Please explain how EAL CA3.1 will be assessed in a submittal, the cited parenthetical was deleted because MPS2 SPS timely and accurate manner during solid water does not perform solid water plant operations.

conditions or consider adding a clarifying note to the threshold value for EAL CA3.1.

The proposed Fuel Clad and RCS Potential Loss B.3 MPS2 Fuel Clad Barrier Potential Loss B.3 and RCS Barrier threshold values state, "Applicable RCS and Core Heat Potential Loss B.3 thresholds have been revised as follows:

Removal (HR) Safety Function Status Check "Applicable RCS and Core Heat Removal, HR-1 or HR-2, acceptance criteria not met." Since the proposed Basis Safety Function Status Check acceptance criteria not met for Fuel Clad Potential Loss B.3 includes HR Once Through Core Cooling (OTCC), this threshold value OR would not be met until it was determined that OTCC is FC Loss B.3 OTC (HR-3) is required" not effective. This is not consistent with the NEI 99 01, 8 MPS2 Revision 6, guidance which states, "[t]his condition The MPS2 EAL Technical Bases Document has been updated, RCS Pot.

indicates an extreme challenge to the ability to re.move as shown in Enclosure 3, Attachment 2, to reflect the changes Loss B.3 RCS heat using the steam generators (i.e., loss of an to the Fuel Clad Barrier Potential Loss B.3 and RCS Barrier effective secondary side heat sink)." Potential Loss B.3 thresholds described above.

Please explain how the Emergency Director would make an accurate assessment of Fuel Clad and RCS Potential Loss B.3 based on a loss of the ability to remove heat using the steam generators rather than based on an inability to implement OTCC.

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Station IC/EAL Table 4, "MPS2 Comparison Matrix," provides that The Difference Justification column of the MPS2 Table 4, "MPS2 does not start a standby charging pump if "MPS2 Comparison Matrix," Category F: Fission Product Barrier inventory cannot be maintained with operating makeup, Degradation, RCS P-Loss 1, RCS Pot. Loss A.1, has been rather SI [safety injection] would be initiated." updated as shown in Enclosure 6.

Considering that Combustion Engineering plants The pump capacity of a standby MPS2 positive displacement typically have a pressurizer level control system that charging pump, started on decreasing pressurizer level, is not would automatically start a second charging pump on indicative of a potential loss of the RCS barrier. Control room lowering level, and abnormal operating procedures indications are available to provide the operator adequate (AOP) for a RCS leak would direct starting a second capability to maintain pressurizer level within specified limits and charging pump and would not typically initiate safety identify UNISOLABLE RCS or SG tube leakage> 50 gpm injection without tripping the reactor, the NRG staff excluding normal reductions in RCS inventory (e.g., letdown, could not understand why the alternate threshold RCP seal leakage). MPS2 has implemented the alternative wording was required. Additionally, it appears that threshold wording consistent with NEI 99-01, Rev. 6, RCS operators would have to subtract reactor coolant pump Potential Loss 1, Developer's Notes.

RCS Pot.

9 MPS2 leakage and letdown from 50 gallons per minute. This Loss A.1 MPS2 has a pressurizer level control system that automatically appears to require the operators to make a calculation, that could involve potentially changing values, to starts a second charging pump on lowering level. A deviation in assess this RCS Potential Loss A.1. pressurizer level would cause letdown flow to throttle back .

(down to a minimum flow of approximately 28 gpm letdown 1f If it is not reasonable that a second charging would necessary). If pressurizer level continued to lower, the first either start automatically on lowering pressurizer level backup charging pump would receive a start signal at or as an abnormal operating procedure response to a approximately -2.5% level deviation. Operations personnel RCS leak, then provide a readily available threshold would enter Abnormal Operating Procedure (AOP) 2568, value for RCS Potential Loss A.1 that does not involve "Reactor Coolant System Leak Abnormal Operating Procedure,"

a mass balance calculation. If a RCS leakage AOP and proceed to stabilize RCS inventory, reactor power, RCS directs that the reactor be tripped or otherwise rapidly temperature and quantify the RCS leakage.

shut down based on RCS leakage, it may be reasonable to use those conditions as a threshold The design flow of the MPS2 positive displacement charging value. pumps is 44 gpm. Decreasing pressurizer level with RCS power and temperature stable would be indicative of a> 50 gpm leak rate with two charging pumps running (88 gpm), a minimum

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Station IC/EAL design letdown flow of approximately 28 gpm and assumed RCP bleedoff flow of up to 10 gpm. Therefore, continued decreasing level with two charging pumps running would not require operators to make a mass balance calculation. By contrast, two charging pumps running and system level stabilized would likely indicate leakage < 50 gpm and not constitute an inability to maintain pressurizer level within specified limits.

If increasing charging flow does NOT stop the decreasing pressurizer level trend while operating in MODEs 1 or 2, then the operator is directed to trip the reactor. If the RCS leak occurs in MODE 3 and increasing charging flow does NOT stop the decreasing pressurizer level trend, then the operator is provided with guidance on how to address initiation of a safety injection actuation signal.

Table 4, "NAPS (SPS) Comparison Matrix," provides The evaluation to determine whether this threshold has been that "Starting of a standby charging pump is not met does not require a detailed mass balance calculation. The representative of RCS leak size relative to charging information to determine whether this threshold has been met pump capacity." Although the charging systems for a can be determined by using readily available information.

Westinghouse plant are not all identical, either system Specifically, the following RCS data from the control board is design or operator actions will typically maximize available:

NAPS RCS Pot. charging flow in response to lowering pressurizer level.

charging make-up flow 10 SPS Loss A.1 The threshold value for RCS Potential Loss A.1 is

seal injection flow based on the capability of one charging pump to

letdown flow and seal return flow maintain pressurizer level after either the automatic Operations personnel and ST As are trained to perform this system response or operator actions have maximized task in a timely manner, while simultaneously managing a leak charging flow and not on a specific RCS leak rate. The in the RCS.

developer notes for RCS Potential Loss A.1 allow for an appropriate site specific value. However, the threshold

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Station IC/EAL value provided is based on nominal charging pump capacity, and the basis discussion indicates that normal reductions in RCS inventory, such as letdown and RCP seal leakoff, should be excluded. This appears to require the operators to make a calculation, which could involve changing values, to assess this RCS Potential Loss A.1.

If it is not reasonable that a second charging would either start automatically on lowering pressurizer level or as operator response to a RCS leak, then provide a readily available threshold value for RCS Potential Loss A.1 that does not involve a mass balance calculation.

Note: if a RCS leakage AOP or alarm response manual directs that the reactor be tripped or otherwise rapidly shut down based on RCS leakage, it may be reasonable to use those conditions as a threshold value.

The proposed threshold value for MPS 2 EAL HU 2.1 MPS2 and MPS3 share a common Protected Area. The requires notification by MPS Unit 3 during a seismic magnitude of any seismic event will impact both units equally.

event. Depending on the potential impacts, MPS Unit 3 Therefore, for a seismic event greater than the OBE, MPS3 will may not be readily available to contact MPS Unit 2. classify the event because the instrumentation triggered by the seismic event (i.e., used for the event classification) is located The proposed threshold value for SPS Units 1 and 2 is MPS2 in the MPS3 Control Room. The MPS2 EAL bases states that 11 HU2.1 modified by a note, which indicates that EAL HU 2.1 SPS MPS3 will notify the MPS2 Control Room to initiate MPS2 plant should be declared once the event has not been operations response to the seismic event, not for emergency diagnosed within 15 minutes provided control room classification purposes. If the magnitude of the event results in personnel felt a seismic event. a loss of communications capability, classification would be per Please explain why the alternative guidance in the NEI EAL MU7 .1 /CU5.1.

99 1, Revision 6, developer notes was not included as The alternative generic guidance provided in the developer's

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Station IC/EAL an alternate threshold value since this would provide notes is for plants that have their seismic instrument indications clear and consistent guidance for declaring EAL HU2.1 external to the Control Room. SPS seismic indications are if the primary method was either not available or located inside the Control Room envelope but SPS does not delayed. have an OBE exceeded annunciator or indicator on the main control panels. An OBE determination is made in accordance with O-AP-37.00, "Seismic Event". This determination can be completed within 15 minutes of receipt of the seismic trigger annunciator (VSP-45 (E-7) on the main control panels). Note 13 has been included to ensure that the event is classified for a felt earthquake in the unlikely event that the OBE determination, in accordance with O-AP-37.00, is not completed in a timely manner.

The NAPS (SPS) Fire Areas for EALs HU 4.1 and HU For Appendix "R", NAPS and SPS conform to the guidance of 4.2 in the proposed Table H 1 includes the entire Appendix "A" to Branch Technical Position (BTP) 9.5-1. The turbine, auxiliary, fuel, and decontamination buildings. shutdown methodology was performed in accordance with The threshold values for EALs HU 4.1 and HU 4.2 Section 111.G.3 of Appendix "R". This compliance approach should represent fires that could potentially degrade the resulted in some large fire areas. In this case, the 1) Turbine level of safety of the plant. It appears that the proposed Building and 2) Auxiliary Building, which also includes fire Table H-1 could result in an NOUE declaration for fires zones for the Fuel and Decontamination Buildings. The NAPS that do not potentially degrade the level of safety of the Appendix 'R' analysis of record does not support the detailed 12 Table H-1 SPS plant. segregation of these buildings by elevations and/or rooms.

Please review the NAPS (SPS) Fire Areas for EALs HU 4.1 and HU 4.2 in the proposed Table H1 and verify that a fire that cannot be extinguished within 15 minutes anywhere in the turbine, auxiliary, fuel handling, or decontamination buildings would potentially degrade the level of safety of the plant.

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Station IC/EAL Dominion proposes to deviate from a standard EAL a. The justification for the proposed deviation to eliminate the scheme by eliminating the site specific restoration time SBO coping time criteria for AC power restoration from from the threshold value for EAL MG 1.1. The NRG MG 1.1 is not based on the existence or availability of FLEX staff does agree that, as stated in the basis discussion equipment or the procedures directing use of FLEX for EALs MS 1.1 and MG 1.1, which states "credit can equipment. The basis for the justification is the procedural be taken for any [alternating current] AC power source guidance put in place as part of FLEX implementation for an that has sufficient capability to operate equipment Extended Loss of AC Power (ELAP).

necessary maintain a safe shutdown condition, such as As stated in the deviation justifications submitted:

the FLEX generators." The NRG staff also does not agree that the existence of FLEX equipment and "In accordance with plant EOPs, operators will declare an appropriate procedures to use that equipment provides ELAP within 60 min. of the loss of all AC power to the justifies the removal of the "site specific" time to restore emergency buses and direct implementation of FLEX AC power. Additionally, the basis discussion that credit Support Guidelines, including the deployment of MPS2 can be taken for any AC power source is not reflected dedicated portable equipment and performance of DC MPS3 in the threshold values for EALs MS 1.1 and MG 1 .1, load shedding. Even if no AC emergency bus is 13 MG1.1 nor is it consistent with Emergency Preparedness energized, these actions will maintain or restore core NAPS SPS Frequently Asked Question (EPFAQ) 2015-015, cooling, containment, and spent fuel pool cooling "Consideration of listing site specific power sources capabilities indefinitely. Therefore, the underlying basis applicable for consideration for loss of power EALs." for the generic EAL SBO coping time statement, that power must be restored to an AC emergency bus within a

a. Please explain what features, which are unique to fixed amount of time to avoid a severe challenge to one or Dominion facilities, require a deviation from the more fission product barriers, is not valid for MPS2 NRG-endorsed EAL scheme or provide threshold

[MPS3, NAPS and SPS.]"

values that are consistent with NEI 99 01, Revision 6, such that a General Emergency classification The SBO analyses and derived coping times were level would be declared for an extended loss of AC determined in accordance with 10 CFR 50.63 and power with sufficient capacity to operate Regulatory Guide 1.155. These analyses do not take credit equipment necessary to maintain a safe shutdown for the current plant capabilities in place to mitigate the condition. effects of an extended loss of AC power (ELAP) and are not appropriate criteria for escalation to a General Emergency

b. To ensure timely and accurate assessment of MS when adequate core cooling is available. Escalation to the 1.1 and MG 1.1, please include either a condition

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RAI#

Station IC/EAL or a note for EALs MS 1 .1 and MG 1.1 threshold General Emergency should be based on actual indications values that clearly indicates that "credit can be of degraded core cooling during loss of AC power events.

taken for any AC power source that has sufficient The appropriateness of this deviation is not based on any capability to operate equipment necessary unique design feature of the Dominion Energy facilities. The maintain a safe shutdown condition, such as the majority of the industry NEI 99-01, Revision 6 license FLEX generators."

submittals were made prior to full industry implementation of BDBEE and ELAP guidance.

b. The suggested note would be applicable to the EALs associated with the loss of all AC power to emergency busses for greater than 15 min.; CA2.1 and MS1 .1 MPS2, MPS3, NAPS and SPS EALs CA2.1 and MS1 .1 thresholds have been revised to include the following note:

"For this EAL, credit can be taken for any AC power source that has sufficient capability to operate equipment necessary to maintain a safe shutdown condition, provided it can be aligned within the 15 minute classification criteria."

The EAL Technical Bases Documents for MPS2/MPS3, NAPS and SPS, EALs CA2.1 and MS1 .1 have been updated to include the note indicated above, as shown in Enclosures 3, 4 and 5, respectively.

The proposed threshold value for EALs NAPS MU 4.1 Dominion Energy agrees that the dose rate values selected for and SPS MU 4.1 appear to use radiation monitor the MU4.1 limits respectively, do not align with full power NAPS readings that are based on an accident mix after 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months threshold values applicable to Mode 1. The MU4.1 values 14 MU4.1 SPS of decay, which appears to reflect post shut down selected align with the one hour decayed responses at conditions. Although the associated Technical Technical Specification limiting iodine spikes of 10 uCi/gm DEi Specification is applicable in Mode 1, no full power for Surry and 60 uCi/gm DEi for North Anna.

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Station IC/EAL threshold value was provided. The NAPS and SPS threshold values for MU4.1 have been revised to be based on calculated full power values as follows:

Please explain how EALs NAPS MU 4.1 and SPS MU 4.1 will be accurately assessed in a timely manner while NAPS MU4.1 has been changed from the current operating at full power or revise accordingly. value of 1.5E+04 mrem/hr to 2.4E+04 mrem/hr.

SPS MU4.1 value has changed from the current value of 3.0E+05 cpm to 1.0E+06 cpm.

The NAPS and SPS Technical Bases Documents for EAL MU4.1 have been updated to include the change in threshold values indicated above, as shown in Enclosures 4 and 5, respectively.

The proposed threshold value for EALs MPS MU 4.1, A review of the calculation of Technical Specification coolant NAPS MU 4.2, and SPS MU 4.2 appears to use post activity dose rates performed assuming the Technical shutdown radiation levels that would be consistent with Specification limiting iodine spike of 10 uCi/gm DEi for Surry an iodine spike at full power that corresponds to the and 60 uCi/gm DEi for all other plants also concluded that the Technical Specification limit. Although the associated dose rate limit selected for the (.$. 2 hr) time period did not Technical Specification is applicable in Mode 1, no full consider the full power value applicable to Mode 1. The value power threshold value was provided for the applicable selected to represent this time period was based on a range MPS2 MU4.1 EALs. starting after 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of decay. The current values in each MPS3 15 respective EAL table represent a median limit selected NAPS MU4.2 Please explain how EALs MPS MU 4.1, NAPS MU 4.2, between dose rate values calculated between 1 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of SPS and SPS MU 4.2 will be accurately assessed in a timely decay using a method to minimize error from the highest to manner while operating at full power.

lowest dose rate within the range. When compared to a median limit selected between dose rate values calculated between O to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of decay, the limit for the (< 2 hr) value in the table will increase.

The values in Table M-4 for the (.$. 2 hr) time period have been revised to better represent the threshold including

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Station IC/EAL consideration of a value at full power, as follows:

Table M-4 Tech. Spec. Coolant Activity Dose Rates (SPS Only):

Shutdown (hrs) mR/hr/ml S2 0.16

>2 - s 8 0.10

>8 0.05 Table M-4 Tech. Spec. Coolant Activity Dose Rates (NAPS, MPS2, MPS3):

Shutdown (hrs) mR/hr/ml S2 0.80

>2- s 8 0.50

>8 0.30 The Technical Bases Documents for EALs MPS MU 4.1, NAPS MU 4.2, and SPS MU 4.2 have been updated to include the changes indicated above, as shown in Enclosures 3, 4 and 5, respectively.

The proposed threshold value for EAL SPS MU 4.2 The Intermediate Shutdown mode applicability for SPS EALs 16 SPS MU4.2 appears to be for conditions when the Technical MU4.1, MU4.2 and MU4.3 has been modified to be applicable Specifications would no longer apply. Specifically, it only when RCS temperature is greater than 500°F consistent appears EAL SPS MU 4.2 would no longer be with the applicability of SPS Technical Specification 3.1.D

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Station IC/EAL applicable when the reactor is shut down and RCS limits.

temperature is reduced to less than S00°F.

The Technical Bases Documents for SPS EALs MU 4.1, MU Please explain how EAL SPS MU 4.2 will accurately 4.2 and MU 4.3 have been updated to include the change assessed, such that an NOUE classification level would indicated above, as shown in Enclosure 5.

only be declared when the associated Technical Specification is applicable.

EALs MPS MU 4.2, NAPS MU 4.3, and SPS MU 4.3 have been revised as follows to consider that the plant has 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months to recover reactor coolant activity concentration per Technical The proposed threshold value for EALs MPS MU 4.2, Specifications prior to declaration of a NOUE classification NAPS MU 4.3, and SPS MU 4.3 appear to apply to level:

conditions where the Technical Specifications allow continued operation for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months while the licensee MPS2 MU4.2:

makes attempts to restore either I 131 or Xe 133 Sample analysis indicates that a reactor coolant activity value concentrations to within the Technical Specification is> any of the following Technical Specification 3.4.8 limits:

MPS2 limits. However, it may not be appropriate to declare an MU4.2 NOUE classification level when the Technical Dose equivalent 1-131 > 1.0 uCi/gm for> 48 hrs MPS3 17 Specification Limiting Condition for Operation (LCO) is Dose equivalent 1-131 > 60 uCi/gm NAPS MU4.3 met. This is especially true when LCO allows continued Dose equivalent Xe-133 > 1,100 uCi/gm tor > 48 hrs SPS operation at full power for an extended time. MPS3 MU4.2:

Please explain how EALs MPS MU 4.2, NAPS MU 4.3, Sample analysis indicates that a reactor coolant activity value and SPS MU 4.3 will be accurately declared for is> any of the following Technical Specification 3.4.8 limits:

conditions which indicate a potential degradation of the Dose equivalent 1-131 > 1.0 uCi/gm for> 48 hrs level of safety of the plant consistent with the Dose equivalent 1-131 > 60 uCi/gm declaration of an NOUE classification level.

Dose equivalent Xe-133 > 81.2 uCi/gm for> 48 hrs NAPS MU4.3:

Sample analysis indicates that a reactor coolant activity value

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Station IC/EAL is> any of the following Technical Specification 3.4.16 limits:

Dose equivalent 1-131 > 1.0 uCi/gm for> 48 hrs Dose equivalent 1-131 > 60 uCi/gm Dose equivalent Xe-133 > 197 uCi/gm for > 48 hrs SPS MU4.3:

Sample analysis indicates that a reactor coolant activity value is> any of the following Technical Specification 3.1.D limits:

Dose equivalent 1-131 > 1.0 uCi/gm for> 48 hrs Dose equivalent 1-131 > 10 uCi/gm Dose equivalent Xe-133 > 234 uCi/gm for > 48 hrs The Technical Bases Documents for EALs MPS2/MPS3 MU 4.2, NAPS MU 4.3, and SPS MU 4.3 have been updated to include the changes indicated above, as shown in Enclosures 3, 4 and 5, respectively.

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Station IC/EAL For the purposes of emergency classification, the MPS Unit 3, NAPS and SPS method used to determine the reactor is shutdown following a reactor trip is consistent with the plant EOPs. This method is based on reactor power indicating The proposed threshold value for EALs MPS Unit 3 greater than that requiring entry into Critical Safety Function MU/MA/MS 6.1, NAPS MU/MA/MS 6.1, and MU/MA/MS Status Trees (CSFST) Subcriticality Red Path (5%). This is 6.1, when combined with the proposed basis document also the power level that defines power operation in the discussion, could be interpreted to imply that as long as Technical Specifications.

reactor power is less than 5%, then an emergency As specified in the generic developer's guidance, "Developers declaration is not required, based on either the Mode 1 may include site-specific EOP criteria indicative of a successful definition of power operation or the capacity of the reactor shutdown in an EAL statement, the Basis or both (e.g.,

plant's decay heat removal systems rather than a a reactor power level)." Additionally, the generic NEI 99-01, failure of the reactor protection system to shut down the MU6.1 Revision 6 bases states that a successful reactor trip results in MPS3 reactor. This is not consistent with NEI 99 01, Revision a power level within the capability of decay heat removal 18 NAPS MA6.1 6, which requires the reactor to be shut down.

systems (the basis for the CSFST Subcriticality Red Path entry SPS All current Dominion reactor protection system failure condition).

MS6.1 EAL threshold values, as well as the proposed EAL Therefore, the specified power levels are the site-specific threshold value for MPS Unit 2, do not present a human indication of a successful reactor trip for emergency factor concern and are consistent with NEI 99 01, classification for those plants that implement WOG Emergency Revision 6. Please revise the proposed EALs MPS Response Guidelines (ERGs) and CSFSTs.

Unit 3 MU/MA/MS 6.1, NAPS MU/MA/MS 6.1, and MU/MA/MS 6.1, to reflect a reactor shutdown as MPS Unit 2 does not implement WOG ERGs, but instead defined in station emergency operating procedures implements Combustion Engineering reactivity control Safety (EOPs), which include reactor power lowering, or use Function Status Check acceptance criteria. Therefore, for the wording consistent with NEI 99 01, Revision 6. purpose of emergency classification, indications for a successful reactor trip have been specified consistent with those criteria. MPS2 has not identified an equivalent power level associated with the capability of decay heat removal systems.

Serial No.19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 ENCLOSURE 1 Attachment 2

/'

RAI 4-EAL COMPARISON TABLES Dominion Energy Nuclear Connecticut, Inc. (DENC)

Virginia Electric and Power Company (Dominion Energy Virginia)

Millstone Power Station Units 2 and 3 and ISFSI North Anna Power Station Units 1 and 2 and ISFSI Surry Power Station Units 1 and 2 and ISFSI

Serial No.19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 1 Attachment 2 Page 1 of 4 Comparison of Current and New Table R-1 Effluent Monitor Setpoints Millstone 2 Pathway Site Stack MP2 Vent MP2 Vent Rad Monitor RM-8169 RM-8169 RM-81328 RM-81328 RM-8168 RM-8168 EAL Revision Current New Rev.6 Current New Rev.6 Current NewRev.6

~2xREMODCM ~2xAlloc REMODCM ~2xREM0DCM ~2xAlloc REMODCM ~2xAlloc REMODCM (0.026 uCi/cc) (0.2 uCi/cc) (8.4E4 CPM) (4.4ES CPM) N/A (0.016 uCi/cc) 1 uCi/cc 3.6 uCi/cc N/A N/A 0.02uCi/cc 0.16uCi/cc lOuCi/cc 36 uCi/cc N/A N/A 0.2 uCi/cc 1.6 uCi/cc 30uCi/cc 360 uCi/cc N/A N/A 2uCi/cc 16 uCi/cc Dose Model ADAM code MIDAS 1.5.17 ADAM code MIDAS 1.5.17 C

0 (Alert & SAE) Met Avg (SAE & ALERT) Predominant MET data Avg (SAE & ALERT) Predominant MET data

c 95% (GE) Predominant MET data 95% (GE) Predominant MET data I- (GE) Met Data N/A N/A w TID-14844 Source Term NUREG-1465 TID-14844 NUREG-1465

~

Flow 12,000cfm 12,000cfm 64,000cfm 64,000cfm ODCM Release Point NOUE developed using ODCM Release Point NOUE developed using 83% ODCM Release Point NOUE developed using 83%

NOTES apportion (13%) applied to 100% Allocation. apportion (33%) applied to Allocation. apportion (33%) applied to Allocation.

the NOUE Limit. No Release Point apportion the NOUE Limit. No Release Point apportion the NOUE Limit. No Release Point apportion applied to REMODCM limit. applied to REMODCM limit. applied to REMODCM limit.

I-z Alert 0.00027 uCi/cc 0.00027 uCi/cc N/A N/A 0.02 uCi/cc 0.16 uCi/cc 5

~ Alarm 0.013 uCi/cc ~0.1 uCi/cc SOOCPM SOOCPM 0.2 uCi/cc 1.6 uCi/cc MPS2 NOUE values are different due to 'Current' values crediting station release point apportion factors defined in the REMODCM. Revision 6 NOUE values are based on Technical Specification limits (Instantaneous Release Rate Limits - IRRL) defined in the REMODCM without crediting release pathway apportion.

ALERT, SAE, and GE values are different between 'Current' and 'Rev. 6' values primarily because of different dose models used in their determination. The 'Current' values were generated using an in-house Accident Dose Assessment Model (ADAM) versus use of the MIDAS code.

Additionally, the 'Current' values used different meteorological and source term assumptions.

Serial No.19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 1 Attachment 2 Page 2 of 4 Comparison of Current and New Table R-1 Effluent Monitor Setpoints Millstone 3 Pathway Site Stack MP3 Vent MP3 ESF Vent Rad Monitor RE-19 RE-19 RE-10 RE-10 RE-49 RE-49 EAL Revision Current New Rev.6 Current New Rev.6 Current New Rev.6

.e'.2xREMODCM .e'.2xAlloc REMO DCM .e'.2xREMODCM .e'.2xAlloc REMODCM .e'.2xREMODCM .e'.2xAI Ioc REMO DCM (0.026 uCi/cc) (0.2 uCi/cc) (0.0017 uCi/cc) (0.0059 uCi/cc) (0.0012 uCi/cc) (0.12 uCi/cc) 1 uCi/cc 3.6uCi/cc 0.01 uCi/cc 0.059 uCi/cc N/A ~/A 10uCi/cc 36uCi/cc 0.1 uCi/cc 0.59 uCi/cc N/A N/A 30uCi/cc 360 uCi/cc 0.8 uCi/cc 5.9 uCi/cc N/A N/A Dose Model ADAM code MIDAS 1.5.17 ADAM code MIDAS 1.5.17 C

0 (Alert & SAE) Met Avg (SAE & ALERT) Predominant MET data Avg (SAE & ALERT) Predominant MET data

I:

I- (GE) Met Data 95% (GE) Predominant MET data 95% (GE) Predominant MET data N/A N/A w

Source Term TID-14844 NUREG-1465 TID-14844 NUREG-1465.

i Flow 12,000cfm 12,000cfm 210,000cfm 210,000cfm ODCM Release Point NOUE developed using ODCM Release Point NOUE developed using ODCM Release Point NOUE developed using NOTES apportion (13%) applied to 100% Allocation. apportion (33%) applied to 100% Allocation. apportion (1%) applied to 100% Allocation.

the NOUE Limit. No Release Point apportion the NOUE Limit. No Release Point apportion the NOUE Limit. No Release Point apportion a lied to REMODCM limit. a lied to REMODCM limit. a lied to REMO DCM limit.

I-z Alert 0.00006 uCi/cc 0.00006 uCi/cc 0.000013 uCi/cc 0.000013 uCi/cc 3.0E-04 uCi/cc 3.0E-04 uCi/cc 6

0..

I;:; Alarm 0.011 uCi/cc 0.011 uCi/cc 0.00084 uCi/cc 0.0029 uCi/cc 5.9E-04 uCi/cc 5.9E-04 uCi/cc MPS3 NOUE values are different due to 'Current' values crediting station release point apportion factors defined in the REMODCM. Revision 6 NOUE values are based on Technical Specification limits {Instantaneous Release Rate Limits - IRRL) defined in the REMODCM without crediting release pathway apportion.

ALERT, SAE, and GE values are different between 'Current' and 'Rev. 6' values primarily because of different dose models used in their determination. The 'Current' values were generated using an in-house Accident Dose Assessment Model (ADAM) versus use of the MIDAS code.

Additionally, the 'Current' values used different meteorological and source term assumptions.

Serial No.19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 1 Attachment 2 Page 3 of 4 Comparison of Current and New Table R-1 Effluent Monitor Setpoints North Anna Pathway Vent Stack A Vent Stack B Process Vent Rad Monitor VG-Rl-179 VG-Rl-179 VG-Rl-180 VG-Rl-180 GW-Rl-178 GW-Rl-178 EAL Revision Current New Rev.6 Current New Rev.6 Current New Rev.6

,::,:2xAlloc ODCM ,::,:2xAlloc ODCM ,::,:2xAlloc ODCM ,::,:2xAlloc ODCM _::,:2xAlloc ODCM _::,:2xAlloc ODCM

{3.6E+OS uCi/sec) (2.6E+OS uCi/sec) (3.6E+OS uCi/sec) (2.0E+OS uCi/sec) (2.8E+OS uCi/sec) (3.SE+OS uCi/sec) 4.56E+06 uCi/sec 2.6E+06 uCi/sec 4.07E+06 uCi/sec 2.0E+06 uCi/sec 4.22E+06 uCi/sec 3.SE+06 uCi/sec 4.0E+07 uCi/sec 2.6E+07 uCi/sec 3.57E+07 uCi/sec 2.0E+07 uCi/sec 3.7E+07 uCi/sec 3.SE+o7 uCi/sec 4.0E+08 uCi/sec 2.6E+08 uCi/sec 3.57E+08 uCi/sec 2.0E+08 uCi/sec 3.7E+08 uCi/sec 3.SE+08 uCi/sec 0 Dose Model MIDAS 1.5.11 MIDAS 1.5.17 MIDAS 1.5.11 MIDAS 1.5.17 MIDAS 1.5.11 MIDAS 1.5.17 0

i:: Met Data Predominant MET data Predominant MET data Predominant MET data Predominant MET data Predominant MET data Predominant MET data I-Source Term NUREG-1465 (NG only) NUREG-1465 (NG, I, Cs) NUREG-1465 (NG only) NUREG-1465 (NG, I, Cs) NUREG-1465 (NG only) NUREG-1465 (NG, I, Cs)

"':ii: Flow 142,300 cfm 40,000cfm 108,700cfm 12,000cfm 310cfm 300cfm NOUE developed using NOUE developed using NOUE developed using NOUE developed using NOUE developed using 10% NOUE developed using NOTES 100% Allocation of Tech ODCM limit and 72% 100% Allocation ofTech ODCM limit and 55% Allocation of Tech Spec ODCM limit and 12.5%

Spec dose limit. Allocation. Spec dose limit. Allocation. dose limit. Allocation.

1-z 0 Value s_3.6E+OS uCi/sec s_2.0E+OS uCi/sec s_3.6E+OS uCi/sec s_2.0E+05 uCi/sec s_2.8E+OS uCi/sec .5 3.SE+OS uCi/sec

~

Ill The difference in NAPS NOUE values from 'Current' to new proposed 'Rev. 6' EALs are from different 'allocation' factors determined for each to maintain sufficient separation from respective calculated ALERT threshold values. Revision 6 NOUE values are based on Technical Specification limits (Instantaneous Release Rate Limits - IRRL} defined in the ODCM without crediting release pathway apportion. Another factor that creates a difference is normal operational pathway flow credited in the ODCM versus expected flow under accident conditions.

ALERT, SAE, and GE values are different between 'Current' and 'Rev. 6' values primarily due to different source term and flow assumptions.

Differences caused by different versions of MIDAS are expected to be minimal.

Serial No.19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 1 Attachment 2 Page 4 of 4 Comparison of Current and New Table R-1 Effluent Monitor Setpoints Surry Pathway Vent #2 Process Vent Rad Monitor VG-Rl-131 VG-Rl-131 GW-Rl-130 GW-Rl-130 EAL Revision Current NewRev.6 Current New Rev.6

,:;:2xAlloc ODCM ,:;:2xAlloc ODCM ,:;:2xAlloc ODCM ,:;:2xAI I oc ODCM

{5.67E+04 uCi/sec) {7.2E+04 uCi/sec) {3.68E+05 uCi/sec) (2.8E+05 uCi/sec) 9.12E+05 uCi/sec 7.2E+05 uCi/sec 3.12E+06 uCi/sec 2.8E+06 uCi/sec 8.0E+06 uCi/sec 7.2E+06 uCi/sec 2. 74E+07 uCi/sec 2.8E+07 uCi/sec 8.0E+07 uCi/sec 7.2E+07 uCi/sec 2.74E+08 uCi/sec 2.SE+OS uCi/sec C Dose Model MIDAS 1.S.11 MIDAS 1.5.17 MIDAS 1.5.11 MIDAS 1.5.17 0

c Met Data Predominant MET data Predominant MET data Predominant MET data Predominant MET data I-Source Term NUREG-1465 {NG only) NUREG-1465 {NG, I, Cs) NUREG-1465 (NG only) NUREG-1465 (NG, I, Cs)

a: Flow 37,500cfm 34,000cfm 300cfm 300cfm NOUE developed using 30"/o NOUE developed using NOUE developed using 2.5% NOUE developed using NOTES Allocation of Tech Spec ODCM limit and 38% Allocation of Tech Spec ODCM limit and 2%

dose limit. Allocation. dose limit. Allocation.

1-z 6 Value 5.5.67E+04 uCi/sec 5.7.2E+04 uCi/sec 5.3.68E+05 uCi/sec ~ 2.8E+05 uCi/sec

~

The difference in SPS NOUE values from 'Current' to new proposed 'Rev. 6' EALs are from different 'allocation' factors determined for each to maintain sufficient separation from respective calculated ALERT threshold values. Revision 6 NOUE values are based on Technical Specification limits (Instantaneous Release Rate Limits - IRRL) defined in the ODCM without crediting release pathway apportion. Another factor that creates a difference is normal operational pathway flow credited in the ODCM versus expected flow under accident conditions.

ALERT, SAE, and GE values are slightly different between 'Current' and 'Rev. 6' values primarily due to different source term and flow assumptions. Differences caused by different versions of MIDAS are expected to be minimal

Serial No.: 19-296 Docket Nos.: 50-280/281 72-2/55 ENCLOSURE 2 SPS FUEL CLAD LOSS C.6 DEVIATION Virginia Electric and Power Company (Dominion Energy Virginia)

Surry Power Station Units 1 and 2 and ISFSI

Serial No.: 19-296 Docket Nos.: 50-280/281 72-2/55 Enclosure 2 Page 1 of 1 NEI SPS Example FPB FPB Threshold Threshold Description Fuel Clad 3.B C.6 During the review of calculations to answer RAI 14, it was identified Loss that the Fuel Clad Barrier (FCB) Loss C.6 Letdown radiation monitor threshold value for Surry deviates from the NEI 99-01, Revision 6 guidance, but failed to be identified in the EAL Comparison Matrix Document. The following has been added to correct the omission:

With letdown in service, Reactor Coolant Letdown Radiation Monitor CH-RI-() 18/19 > 5E+06 cpm Per Engineering Calculation PA-0236-0-A, the calculated letdown reading at 1 hr of decay corresponding to 300 uCi/gm DEl-131 , is above the upper limit of detection of 1E+07 cpm. A threshold value of 5E+06 cpm was chosen for CH-RI-( )18/19 based on the midpoint of the highest decade of readable scale of the monitors. While the threshold value is conservative compared to a value corresponding to 300 uCi/gm DEl-131, it represents a significant reactor coolant concentration caused by failure of fuel cladding that is at least an order of magnitude above the Technical Specification iodine spike limit and falls within an approximate range of 2% to 5 % fuel clad failure.

The Fuel Clad Loss threshold value is a deviation from the NEI 99-01, Revision 6, Fuel Clad Loss 3.8 generic wording and bases but is deemed acceptable consistent with the above justification.

Serial No.: 19-296 Docket No.: 50-336 ENCLOSURE 6 MPS2 COMPARISON MATRIX RCS POT. LOSS A.1 Dominion Energy Nuclear Connecticut, Inc. (DENC)

Millstone Power Station Unit 2

Serial No.: 19-296 Docket Nos.: 50-336 Enclosure 6 Page 1 of 1 Table 4 - MPS2 Comparison Matrix Category F: Fission Product Barrier Degradation RCS RCS or SG Tube Leakage RCS Pot. 1. UNISOLABLE RCS or The pump capacity of a standby MPS2 positive P-Loss 1 A. Operation of a standby Loss SG tube leakage > 50 displacement charging pump, started on decreasing A.1 gpm excluding normal charging (makeup) pressurizer level, is not indicative of a potential loss of pump is required by reductions in RCS the RSC barrier. Control room indications are inventory (e.g.

EITHER of the following: available to provide the operator adequate capability letdown, RCP seal

1. UNISOLABLE RCS to maintain pressurizer level within specified limits and leakage) leakage identify UNISOLABLE RCS or SG tube leakage

> 50 gpm excluding normal reductions in RCS OR inventory (e.g., letdown, RCP seal leakage).

2. SG tube leakage. The design flow of the MPS2 positive displacement OR charging pumps is 44 gpm. Decreasing pressurizer level would be indicative of a> 50 gpm leak rate with

8. RCS cooldown rate two charging pumps running (88 gpm), a minimum greater than (site- design letdown flow of approximately 28 gpm and specific pressurized assumed RCP bleedoff flow of up to 10 gpm.

thermal shock Therefore, continued decreasing level with two criteria/limits defined by charging pumps running would not require operators site-specific to make a mass balance calculation. By contrast, two indications). charging pumps running and system level stabilized would likely indicate leakage < 50 gpm and not constitute an inability to maintain pressurizer level within specified limits.

MPS2 has implemented the alternative threshold wording consistent with NEI 99-01, Rev. 6, RCS Potential Loss 1, Developer's Notes.

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 ENCLOSURE 7

SUMMARY

OF CALCULATIONS (UPDATED)

Dominion Energy Nuclear Connecticut, Inc. (DENC)

Virginia Electric and Power Company (Dominion Energy Virginia)

Millstone Power Station Units 2 and 3 and ISFSI North Anna Power Station Units 1 and 2 and ISFSI Surry Power Station Units 1 and 2 and ISFSI

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 1 of 54

SUMMARY

OF CALCULATIONS (UPDATED)

Calculation Summaries (Updated)

General:

One of the goals of the EAL reanalysis effort was to apply consistent practices and methods between the Dominion fleet plants and to define common response criteria across the fleet for certain EAL initiating conditions (i.e. fuel clad degradation and fuel barrier failure criteria, RCS sample dose rates and sample line dose rates).

Uncertainties in the proposed methodologies are similar to anticipated and acceptable uncertainties known to exist in most radiological assessment methods and techniques in support of emergency response. The primary variables used in the reanalysis of dose rate EAL thresholds response are, (1) source term, (2) shielding geometry, and (3) source volume. New calculated EAL values are based on expected plant conditions.

For instance, core isotopic release fractions are based on realistic recommendations from NUREG 1228,"Source Term Estimation during Incident Response to Severe Nuclear Power Plant Accidents", October 1988" rather than conservative and bounding design basis guidance of NUREG 1465,"Accident Source Terms for Light-Water Nuclear Power Plants", dated February 1995. Equilibrium coolant concentrations are taken from calculations of Technical Specification RCS Coolant Activity applicable to the fuel clad degradation initiating condition criteria. RCS volumes are updated to assume hot full-power conditions. Single dose rate response thresholds are used wherever applicable across all Fleet plants. Common fleet EAL values for similar category thresholds are deemed important to enhance familiarity between facilities/units to reduce the potential for human error. Calculation summaries have been provided in this Enclosure which contain additional information for critical calculations used in our EAL reanalysis.

The basis for the gaseous Unusual Event IC and associated thresholds has beeh revised to correspond to any unplanned release of gaseous effluent radioactivity to the environment that will result in release 2 times the allocated site-specific effluent release controlling document limits for 60 minutes or longer. This Unusual Event gaseous release criterion is being used consistently across all operating nuclear units at Dominion Energy nuclear stations at Millstone, North Anna and Surry. The word

'allocated' is required because for some release points, using ODCM methods and limits to determine the UE EALs, the UE values calculated were greater than ALERT EAL threshold values or did not provide a factor of 10 separation from the ALERT EAL threshold. To maintain consistency across the Fleet and reduce confusion and human error potential, a single Initiating Condition (IC) definition for gaseous and liquid releases at the NOUE level is being used. The Initiating Condition (IC) will be worded:

"Release of gaseous or liquid radioactivity greater than 2 times the allocated ODCM

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 2 of 54 limits for 60 minutes or longer." This method provides a justifiable basis for NOUE thresholds based on established methods and setpoints provided in the facility ODCM.

The proposed NOUE values will classify events based on degradation in the level of safety of the plant and will maintain a near linear escalation between all four classification levels (i.e., NOUE, ALERT, Site Area Emergency (SAE) and General Emergency (GE)). The IC being used is the same IC definition currently used for gaseous pathways in the North Anna and Surry NEI 99-01, Revision 4 EALs.

Classification thresholds within Table R-1 were generated using the MIDAS dose assessment code. Inputs to MIDAS use most prevalent meteorological data and expected release point parameters. Most prevalent meteorology represents conditions that wou Id most likely exist (based on the most prevalent stability class and average wind speed within that stability class). Dispersion based on most prevalent meteorology differs from that assumed , in the ODCM which uses annual average meteorology.

Dispersion based on actual meteorological conditions at the time of the emergency (most prevalent) can be 10 - 20 times higher than the annual average dispersion prescribed for use in an ODCM. Assumptions of one-hour decay since shutdown and a one-hour release duration are applied. Mitigating reduction mechanisms (e.g., decay, sprays, filters) input into MIDAS for each accident type determined the radiological release source term consistent with the guidance provided in NUREG-1228.

Dose rate values specified in Tables F-2, F-3 and F-4 were developed using a method to minimize error(+/-) for the threshold value within defined time (decay) periods. Time periods were chosen to fit monitor response (fast changes in response early following reactor shutdown are broken up into smaller time periods to better approximate expected change). Values were chosen within each time period to minimize error

(<50%) from the highest to lowest response within each time range. Fuel clad barrier loss thresholds are each calculated based on 5% fuel clad damage which represents a significant amount of fuel clad damage comparable to 300 uCi/gm DEl-131.

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 3 of 54 Calculation Summary Gaseous Effluent Radiation Monitor Thresholds The following pertinent information has been extracted from Millstone Calculation RP-18-08, "MPS1 Abnormal Rad Release Gaseous EAL Thresholds Based on NEI 99-01, Revision 6". It is provided to assist technical reviewers that will be evaluating this license amendment request.

Purpose:

To determine new Emergency Action Level release threshold using updated guidance from NEI 99-01 Rev 6 for Millstone Unit 1 continuous monitored pathway from the Spent Fuel Pool Island Vent.

References:

1. Nuclear Energy Institute NEI 99-01, Rev. 6, "Methodology for Development of Emergency Action Levels," November 2012.

2. MP-22-REC-BAP01, Rev.29, "Millstone Radiological Effluent and Off-site Dose Calculation Manual (REMODCM)."

3. MP-22-REC-REF03, Rev. 6, "REMODCM Technical Information Document", Oct 18, 2016.

4. Calculation RA-0016, Revision O Addendum A, "Radiological Consequences of Release of all Gap Activity in the Spent Fuel Pool at MP1", Nov. 29, 2010.

5. Software-Meteorological Information and Dose Assessment System, MIDAS, Version 1.5.17.022218.

6. MIDAS Software QA Documentation, SQA-MIDAS-DOM-20180614.

7. EPA-400/R-17/001, "PAG Manual: Protective Action Guides and Planning Guidance for Radiological Incidents", January 2017.

8. MP-26-EPI-FAP10, Rev. 11, "Dose Assessment."

9. M2EAL-03053R2, Rev. 2, "MP2 EAL Offsite Dose Parameters".

Computer Codes Used:

MIDAS software (Ref 5) was utilized to determine the projected EDE, TEDE and Thyroid COE for a one (1) hour release duration. Integrated TEDE for a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months release duration is used for the purpose of calculating emergency action levels for the ALERT classification. MIDAS is classified per the Software Quality Assurance program as class 3 software (Ref .6).

Methodology:

The meteorology and source terms used to develop the threshold values were chosen to best represent the conditions that would be expected at the time of the emergency.

The calculated threshold value considers a source of 100% Kr-85 release from damage of irradiated assemblies in the Unit 1 SFP and meteorology in accordance with NEI 99-

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 4 of 54

01. The resulting values are adequately conservative and represent the best estimate of the release rates that would result in exceeding the dose criteria of NEI 99-01.

To determine the ALERT radiological threshold for the MP1 SFPI vent, MIDAS was used to predict expected doses based on best estimate meteorological and plant conditions. Inputs to MIDAS use most prevalent met data and expected release point parameters and normalized source terms of 1 Ci/sec. No mitigating reduction mechanisms (decay, sprays, filters, etc.) were used as input into MIDAS for this particular calculation as iodine and particulate removal mechanisms have no effect to the release of Kr-85. The MIDAS outputs generated represent a radiological prediction normalized to the source entered (e.g., 1 Ci/sec of Kr-85).

The maximum projected EDE, TEDE at or beyond the site boundary distance were obtained from the MIDAS outputs. The TEDE dose was divided into the applicable EAL criteria to determine the radioactivity concentration (uCi/cc) seen by the radiation monitor which would yield the referenced dose criteria for the ALERT classification.

This concentration is the rad monitor action level for the ALERT classification. Thyroid CDE limits were not included in this evaluation per new guidance in EPA-400 (Ref. 7) and agreement with the State of Connecticut to remove Thyroid PAGs in the EALs.

From the predicted release to obtain levels that achieve the ALERT EAL threshold limit of > 10 mrem TEDE, the corresponding number of irradiated fuel assemblies that would need to be damaged to achieve that source term will be determined.

==

Conclusions:==

The condition where the NOUE threshold is exceeded for 60 minutes is indicative of the inability to terminate radioactive release within prescribed regulatory and license limits and therefore represents a loss of plant control and degraded safety. For the Unusual Event (NOUE) threshold value based on station release limits, the methodology and assumptions established with the Millstone REMODCM were followed.

For the ALERT threshold value determined, the release conditions required to produce 10 mrem TEDE from a pure Kr-85 release using predominant meteorological conditions was calculated using the MIDAS accident dose software code. The ALERT threshold value determined, while equivalent to release conditions that would produce 10 mrem TEDE, requires that 50% of the spent fuel pool irradiated assemblies will need to fail and release within 15 minutes to the environment to produce radiological conditions that produce such dose.

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 5 of 54 MP1 SFPI Vent RM-SFPl-02 EAL Thresholds

, I MPl Sf Pl Vent EAL Escalation

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Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 6 of 54 Calculation Summary Gaseous Effluent Radiation Monitor Thresholds The following pertinent information has been extracted from Millstone Calculation RP-18-02, "MPS2 Abnormal Rad Release Gaseous EAL Thresholds Based on NEI 99-01, Revision 6". It is provided to assist technical reviewers that will be evaluating this license amendment request.

Purpose:

Calculation of new Emergency Action Levels were determined for radioactive releases from the MP2 Ventilation Vent and Millstone Stack based on updated guidance from NEI 99-01, Rev 6 and revision to EPA-400 for removal of thyroid COE PAG limits.

References:

1. Nuclear Energy Institute NEI 99-01, Rev. 6, "Methodology for Development of Emergency Action Levels," November 2012.

2. Software-Meteorological Information and Dose Assessment System, MIDAS, Version 1.5.17.022218.

3. MIDAS Software QA Documentation, SQA-MIDAS-DOM-20180614.

4. NUREG-1228, "Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents",

1. McKenna, T. J. and Gutter, U.S. Nuclear Regulatory Commission, Washington, D.C, 1988.

5. MP-22-REC-BAP01, Rev.27-01, "Millstone Radiological Effluent and Off-site Dose Calculation Manual (REMODCM)."

6. MG-EV-99-0004, Rev 0, "Units 1, 2, 3 Radiological Boundaries", July 20, 1999.

7. MP-26-EPI-FAP10, Rev. 11, "Dose Assessment."

8. Nuclear Energy Institute NEI 99-01, Rev. 4, "Methodology for Development of Emergency Action Levels," January 2003.

9. RERM-02906-R2, Rev. 1, "Millstone Unit 2 Vent Radiation Monitor (RM- 81328)

High Range Setpoint", Jan 16, 2003.

10. MP-22-REC-REF03, Rev. 6, "REMODCM Technical Information Document", Oct 18, 2016.

11. DWG 25203-20098, Rev. 05, "Main Steam Piping Plan - Containment & Aux.

Bldg", dated 10/10/2003.

12. Millstone Unit 2 Radiation Monitor Manual

13. EPA-400/R-17/001, "PAG Manual: Protective Action Guides and Planning Guidance for Radiological Incidents", January 2017.

14. M2EAL-03053R2, Rev. 2, "MP2 EAL Offsite Dose Parameters" MIDAS Dose Software:

MIDAS software (Ref 3.2) was utilized to determine the projected EDE, TEDE and Thyroid COE for a one (1) hour release duration. Integrated TEDE for a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months release duration are used for the purpose of calculating emergency action levels for ALERT,

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 7 of 54 Site Area and General Emergency classifications. MIDAS is classified per the Software Quality Assurance program as class 3 software (Ref. 3).

MIDAS Assumptions and Inputs:

For all MIDAS runs, the stability class=D, ambient temperature=50°, and the direction=252° (from) to result in MIDAS shortest distance to the site boundary= 496 m (-0.31 mO and 620 m (-0.39 mi) in the ENE direction. (5° added to centerline to align over calculation point within MID.AS tabular report at 0.37 mites).

"" Discharge flow per release point was taken from Ref. 7.

,,,. Release from a LOCA is considered leak of contaiflment GAS.

Based on the information above, one run is required for each of the release points as all iflput assumptions are identical and can be rnti~~lo the applicable EAL threshold.

Method of Calculation:

The meteorology and source terms used to develop the threshold values were chosen to best represent the conditions that would be expected at the time of the emergency for each respective action level.

The calculated threshold values consider appropriate source term and meteorology in accordance with NEI 99-01. The resulting values are adequately conservative and represent the best estimate of the release rates that would result in exceeding the dose criteria of NEI 99-01. The values determined show consistent classification escalation from RU1 through RG1.

The RU1 thresholds based on the REMODCM Instantaneous Release Rate Limits that utilize annual average meteorology are shown to be essentially equal to 1 mrem TEDE using most prevalent met conditions. This shows the same principles of dose and maintains consistency with the Technical Specifications. Sufficient margin exists between plant setpoint alarms and the EAL thresholds to provide sufficient awareness to the Operators prior to reaching the emergency condition. The Unusual Event (UE)

EALs are calculated for release points controlled in the REMODCM, Ref. 3.

To determine the EAL radiological thresholds for MP2 Ventilation Vent and Millstone Stack, and release points, MIDAS was used to predict expected doses based on best estimate meteorological and plant conditions. Inputs to MJDAS use most prevalent met data and expected release point parameters together with event tree, core condition, mitigating reduction factors, and normalized source terms of 1 uCi/cc for vent and stack

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 8 of 54 rad monitors. . An assumed one-hour decay time since shutdown and a one-hour duration of release are applied in each computer run. The mitigating reduction mechanisms (decay, sprays, filters, etc.) input into MIDAS for a given accident event determine the final radiological release source term mix. The MIDAS outputs generated for each release option represent a radiological prediction normalized to the source entered (e.g., 1 uCi/cc). For MP2 Vent and Millstone Stack releases, a LOCA accident type is selected for the event tree. A fuel handling accident was not run in MIDAS since an additional mitigation reduction factor of 100 for the pool water would logically result in lower site boundary doses which would then lead to higher emergency action level thresholds for the MP2 Vent and Millstone Stack.

The maximum projected EDE, TEDE and Thyroid COE dose at or beyond the site boundary distance were obtained from the MIDAS outputs .. The TEDE dose was divided into the applicable EAL criteria to determine the radioactivity concentration (uCi/cc) seen by the radiation monitor, which would yield the referenced dose criteria for a given emergency classification. These concentrations are the rad monitor action levels for the various emergency classifications. Thyroid COE limits were not included in this evaluation per new guidance in EPA-400 (Ref .13) and agreement with the State of Connecticut to remove Thyroid PAGs in the EALs.

==

Conclusions:==

Following the guidance of NEI 99-01 Revision 6, recommended values for Millstone 2 release point EAL thresholds were calculated.

For the Unusual Event (NOUE) threshold values determined, the NOUE values are set at 2 times the 'allocated' site-specific effluent release controlling document limits for 60 minutes or longer. To maintain Dominion Fleet commonality, this difference from the prescribed guidance in NEI 99-01 Revision 6 is needed by adding the word 'allocated'.

The difference is necessitated because some release pathways at the Surry and North Anna stations following the ODCM guidance would result in NOUE threshold values greater than corresponding ALERT threshold values. The NOUE thresholds when exceeded for 60 minutes are indicative of the inability to terminate a radioactive release within prescribed regulatory and license limits and therefore represent a loss of plant control and degraded safety.

The ALERT, SAE and GE threshold values determined, represent a radioactive release that results in 1%, 10%, and 100% of the revised EPA Protective Action Guideline TEDE limits. These threshold limits were calculated using expected meteorological conditions based on 5 years of meteorological data collected from the plant MET tower.

Dose analyses were performed using the most prevalent stability class and wind speed conditions at each respective level on the MET tower. The selection and use of predominant meteorological dispersion is appropriate and in accordance with the intent of NEI 99-01.

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 9 of 54 Figure 3 graphically displays the relationship between monitor effluent control setpoint values, the Technical Specification limit, and the four EAL threshold values for the two normal operational discharge release pathways from Millstone 2. This figure demonstrates that the four EALs are sufficiently separated and show escalation from the NOUE level up through the GE level. Sufficient margin exists between plant setpoint alarms and the EAL thresholds to provide sufficient awareness to the Operators prior to reaching the lowest EAL threshold .

Figure 3 - MP2 Vent RM8168 and Stack RM8169 EAL Escalation MP2 Vent and Stack!EAl Escalation

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 10 of 54 Calculation Summary Gaseous Effluent Radiation Monitor Thresholds The following pertinent information has been extracted from Millstone Calculation RP-18-03, "MPS3 Abnormal Rad Release Gaseous EAL Thresholds Based on NEI 99-01, Revision 6". It is provided to assist technical reviewers that will be evaluating this license amendment request.

Purpose:

Calculation of new Emergency Action Levels were determined for radioactive releases from the MP3 Ventilation Vent, Millstone Stack, and MP3 ESF Vent based on updated guidance from NEI 99-01, Rev 6 and revision to EPA-400 for removal of thyroid COE PAG limits.

References:

1. Nuclear Energy Institute NEI 99-01, Rev. 6, "Methodology for Development of Emergency Action Levels,"

1. November 2012

2. Software-Mete9rological Information and Dose Assessment System, MIDAS, Version 1.5.17.022218

3. MIDAS Software QA Documentation, SQA-MIDAS-DOM-20180614

4. NUREG-1228, "Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents",

5. MP-22-REC-BAP01, Rev.27-01, "Millstone Radiological Effluent and Off-site Dose Calculation Manual (REMODCM)"

6. MG-EV-99-0004, Rev 0, "Units 1, 2, 3 Radiological Boundaries", July 20, 1999

7. MP-26-EPI-FAP10, Rev. 11, "Dose Assessment"

8. Nuclear Energy Institute NEI 99-01, Rev. 4, "Methodology for Development of Emergency Action Levels," January 2003.

9. Blank

10. MP-22-REC-REF03, Rev. 6, "REMODCM Technical Information Document"

11. EPA-400/R-17/001, "PAG Manual: Protective Action Guides and Planning Guidance for Radiological Incidents", January 2017

12. Millstone Unit 3 Radiation Monitor Manual

13. EPA-400/R-17/001 , "PAG Manual: Protective Action Guides and Planning Guidance for Radiological Incidents", January 2017.

14. M2EAL-03053R2, Rev. 2, "MP2 EAL Offsite Dose Parameters" MIDAS Dose Software:

MIDAS software (Ref 2) was utilized to determine the projected EDE, TEDE and Thyroid COE for a one (1) hour release duration. Integrated TEDE for a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months release duration is used for the purpose of calculating emergency action levels for ALERT, Site Area and General Emergency classifications. MIDAS is classified per the Software Quality Assurance program as class 3 software (Ref. 3).

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 11 of 54 MIDAS Assumptions and Inputs:

For all MIDAS runs, the stability dass=D, ambient temperature=50\ and the direction=252° (from)to result in MIDAS shortest distance to the sile boundary= 496 m {-0.31 mi) and620 m (-0..39 mi) inlhe ENE direction. (5" added lo centerline to align over calculation point within MID.AS tabular report at 0 ..37 miles) .

.,. Discharge flow per release point was taken from Ref. 7.

-~.J3rliilW from a LOCA is considered leak of containment GAS.

Based on !he information above, one run is required for each of !he release points as all input assumptions are identical and can be 4~ lo !he applicable EAL threshold.

Method of Calculation:

The meteorology and source terms used to develop the threshold values were chosen to best represent the conditions that would be expected at the time of the emergency for each respective action level.

The calculated threshold values consider appropriate source term and meteorology in accordance with NEI 99-01. The resulting values are adequately conservative and represent the best estimate of the release rates that would result in exceeding the dose criteria of NEI 99-01. The values determined show consistent classification escalation from RU1 through RG1.

RU1 thresholds based on the ODCM Instantaneous Release Rate Limits that utilize annual average meteorology are compared against dose criteria to maintain a logical and consistent escalation between the UE and ALERT thresholds. Both are based on the same principles of dose and maintain consistency with the Technical Specifications.

Sufficient margin exists between plant setpoint alarms and the EAL thresholds to provide sufficient awareness to the Operators prior to reaching the NOUE emergency condition.

To determine the EAL radiological thresholds for MP3 Ventilation Vent and Millstone Stack MIDAS was used to predict expected doses based on best estimate meteorological and plant conditions. Inputs to MIDAS use most prevalent met data and expected release point parameters together with event tree, core condition, mitigating reduction factors, and normalized source terms of 1 uCi/cc for vent and stack rad monitors. An assumed one-hour decay time since shutdown and a one-hour duration of release are applied in each computer run. The mitigating reduction mechanisms (decay, sprays, filters, etc.) input into MIDAS for a given accident event determine the final radiological release source term mix. The MIDAS outputs generated for each

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 12 of 54 release option represent a radiological prediction normalized to the source entered (e.g., 1 uCi/cc). For MP3 Ventilation Vent and Millstone Stack releases, a LOCA accident type is selected for the event tree. A fuel handling accident was not run in MIDAS since an additional mitigation reduction factor of 100 for the pool water would logically result in lower site boundary doses which would then lead to higher emergency action levels thresholds for the MP3 Ventilation Vent and Millstone Stack.

The maximum projected EDE, TEDE and Thyroid COE dose at or beyond the site boundary distance were obtained from the MIDAS outputs .. The TEDE dose was divided into the applicable EAL criteria to determine the radioactivity concentration (uCi/cc) seen by the radiation monitor, which would yield the referenced TEDE dose criteria for a given emergency classification. These concentrations are the rad monitor action levels for the various emergency classifications. Thyroid COE limits were not included in this evaluation per new guidance in EPA-400 (Ref. 13) and agreement with the State of Connecticut to remove Thyroid PAGs in the EALs.

==

Conclusions:==

Following the guidance of NEI 99-01 Revision 6, recommended values for Millstone 3 release point EAL thresholds based on the results of this calculation are summarized in Table 1 of the EAL Matrices.

For the Unusual Event (NOUE) threshold values determined, the NOUE values are set at 2 times the 'allocated' site-specific effluent release controlling document limits for 60 minutes or longer. To maintain Dominion Fleet commonality, this difference from the prescribed guidance in NEI 99-01 Revision 6 is needed by adding the word 'allocated'.

The difference is necessitated because some release pathways at the Surry and North Anna stations following the ODCM guidance would result in NOUE threshold values greater than corresponding ALERT threshold values. The NOUE thresholds when exceeded for 60 minutes are indicative of the inability to terminate a radioactive release within prescribed regulatory and license limits and therefore represent a loss of plant control and degraded safety.

The ALERT, SAE and GE threshold values determined, represent a radioactive release that results in 1%, 10%, and 100% of the revised EPA Protective Action Guideline TEDE limits. These threshold limits were calculated using expected meteorological conditions based on 5 years of meteorological data collected from the plant MET tower.

Dose analyses were performed using the most prevalent stability class and wind speed conditions at each respective level on the MET tower. The selection and use of predominant meteorological dispersion is appropriate and in accordance with the intent of NEI 99-01.

Figure 3 graphically displays the relationship between monitor effluent control setpoint values, the Technical Specification limit, and the four EAL threshold values for the two normal operational discharge release pathways from Millstone 3. This figure demonstrates that the four EALs are sufficiently separated and show escalation from

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 13 of 54 the NOUE level up through the GE level. Sufficient margin exists between plant setpoint alarms and the EAL thresholds to provide sufficient awareness to the Operators prior to reaching the lowest EAL threshold.

The Millstone 3 ESF Vent is isolated upon Safety Injection signal. Therefore, there is no ALERT, SAE, or GE threshold for this pathway since this pathway would be isolated prior to reaching levels sufficient to warrant higher classification.

Figure 3 - MP3 Vent {RE10) and Stack {RE1 9) EAL Escalation

___, ______ MP3 .;,;;;;;;;nd Stack EAL Escalat~ ---1

[______ ----- --- ---------- I!

r l

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Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 14 of 54 Calculation Summary Gaseous Effluent Radiation Monitor Thresholds The following pertinent information has been extracted from North Anna Calculation RP-08-22, "North Anna Abnormal Rad Release Gaseous EAL Thresholds Based on NEI 99-01, Revision 6". It is provided to assist technical reviewers that will be evaluating this license amendment request.

Purpose:

Calculation of new Emergency Action Levels were determined for radioactive releases from the NAPS Ventilation Vent and Process Vent based on updated guidance from NEI 99-01, Rev 6 and revision to EPA-400.

References:

1. Nuclear Energy Institute NEI 99-01, Rev. 6, "Methodology for Development of Emergency Action Levels," November 2012.

2. Software-Meteorological Information and Dose Assessment System, MIDAS, Version 1.5.17.022218.

3. MIDAS Software QA Documentation, SQA-MIDAS-DOM-20180614 and all previous files.

4. NUREG-1228, "Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents", McKenna, T. J. and Gutter, U.S. Nuclear Regulatory Commission, Washington, D.C, 1988.

5. VPAP-2103N, Revision 28, "Off-site Dose Calculation Manual (North Anna)."

6. NA-ENGT-000-CME 97-0010, Rev. 0, "Evaluation of the Required Tech Spec Flow Rate Value for Process Vent Blowers ... " Feb. 10, 1997.

7. EPIP-4.03, Revision 22, "North Anna Power Station Dose Assessment Team Controlling Procedure".

8. Nuclear Energy Institute NEI 99-01, Rev. 4, "Methodology for Development of Emergency Action Levels," January 2003.

9. Calculation PA-0225, Revision 0, Addendum (00, OOA, OOB, OOC), "North Anna Radiation Monitor Conversion Factors and EAL Readings".

10. HP-3010.040, Revision 27, "North Anna Power Station - Radiation Monitoring System Setpoint Determination".

11. EPA-400/R-17/001, "PAG Manual: Protective Action Guides and Planning Guidance for Radiological Incidents", January 2017.

12. RP-AA-151, Revision 0, "Radiation Protection Technical Bases Analyses or Calculations".

13. NAPS UFSAR, Revision 53.03.

14. NE-GL-0035N, "PC-MIDAS Guideline", Revision 11.

15. 1-E-O, Rev. 50, North Anna Emergency Procedure - "Reactor Trip or Safety Injection".

16. DC NA-11-01082, Rev. 01, "Main Steam Radiation Monitor Replacement".

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 15 of 54 Computer Codes Used:

MIDAS Dose Software MIDAS software (Ref. 2) was utilized to determine the projected EDE, TEDE and Thyroid COE for a one (1) hour release duration. Integrated TEDE for a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months release duration is used for the purpose of calculating emergency action levels for ALERT, Site Area and General Emergency classifications. MIDAS is classified per the Software Quality Assurance program as class 3 software (Ref .3).

MIDAS Assumptions and Inputs:

For all l\.*11DAS mns, the stability class=D, ambienttemperature=50',andthe direction=252" (froni)to resultin MIDAS shortest distance to the site boundary= 5000!!(-0.94 mi). {5" added to align calculation pointwithin MIDAS tabular report at 1 mile)..

""AccidentflowsfromRef.15,AttachrnentA.

- '"Two plantaccidentconditions applyto effluentsdischargedfromtheProcess Vent, (1) LOCA!GAP/Spray/Filter/RGS Leak-that represents activity in the RCS that is leaking iokl.J!:!M~~r.fil.l.,QQ&MeJNQ:.S.R(fil'l,8Jm.Q'.~~fmm.C.NMI- that represents dischargefromCNMT during early onset of a LOCA before isolation. lnMIDAS, the isotopic mix for both of these scenarios is the same with a total credited OF of 3000.

Based on the information above, one run isrequiredforeach of the release points as all input assumptions are identical and can be urti.11.~Jtto the applicable EALthreshold..

Methodology:

The meteorology and source terms used to develop the threshold values were chosen to best represent the conditions that would be expected at the time of the emergency for each respective action level.

The calculated threshold values consider appropriate source term and meteorology in accordance to NEI 99-01. The resulting values are adequately conservative and represent the best estimate of the release rates that would result in exceeding the dose criteria of NEI 99-01. The values determined show consistent classification escalation from RU1 through RG1.

RU1 thresholds, based on the ODCM Instantaneous Release Rate Limits that utilize annual average meteorology, are compared against dose criteria to maintain a logical and consistent escalation between the UE and ALERT thresholds. Both are based on the same principles of dose and maintain consistency with the Technical Specifications.

Sufficient margin exists between plant setpoint alarms and the EAL thresholds to provide sufficient awareness to the Operators prior to reaching the NOUE emergency condition. The Unusual Event (UE) EALs are calculated for release points controlled in the ODCM, Ref. 5.

Serial No.: 19-296 Docket Nos.: 50-336/ 423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 16 of 54 To determine the EAL radiological thresholds for Ventilation Vent and Process Vent, MIDAS was used to predict expected doses based on best estimate meteorological and plant conditions. Inputs to MIDAS use most prevalent met data and expected release point parameters together with event tree, core condition, mitigating reduction factors, and normalized source terms of 1 uCi/cc for vent and process vent rad monitors. An assumed one-hour decay time since shutdown and a one-hour duration of release are applied in each computer run. The mitigating reduction mechanisms (decay, sprays, filters, etc.) input into MIDAS for a given accident event determine the final radiological release source term mix. The MIDAS outputs generated for each release option represent a radiological prediction normalized to the source entered (e.g., 1 uCi/cc).

For Ventilation Vent and Process Vent releases, a LOCA accident type is selected for the event tree. A fuel handling accident was not run in MIDAS since an additional mitigation reduction factor of 100 for the pool water would logically result in lower site boundary doses which would then lead to higher emergency action levels thresholds for the Ventilation Vent and Process Vent.

The maximum projected EDE, TEDE and Thyroid COE dose at or beyond the site boundary distance were obtained from the MIDAS outputs. These doses were divided into the applicable EAL criteria to determine the radioactivity concentration (uCi/cc) seen by the radiation monitor, which would yield the referenced dose criteria for a given emergency classification. These concentrations are the rad monitor action levels for the various emergency classifications. The lowest predicted concentration between each dose analyzed is selected as the applicable EAL limit.

==

Conclusions:==

Following the guidance of NEI 99-01 Revision 6, recommended values for North Anna release point EAL thresholds based on the results of this calculation are summarized in Table R-1 of the EAL Matrices.

For the Unusual Event (NOUE) threshold values determined, the NOUE values are set at 2 times the 'allocated' site-specific effluent release controlling document limits for 60 minutes or longer. A difference from the prescribed guidance in NEI 99-01 Revision 6 by adding the word 'allocated' is necessitated because the Process Vent instantaneous release rate limits following the ODCM guidance would result in NOUE threshold values greater than corresponding ALERT threshold values. The NOUE thresholds when exceeded for 60 minutes are indicative of the inability to terminate a radioactive release within prescribed regulatory and license limits and therefore represent a loss of plant control and degraded safety.

The ALERT, SAE and GE threshold values determined, represent a radioactive release that results in 1%, 10%, and 100% of the revised EPA Protective Action Guideline TEDE limits. These threshold limits were calculated using expected meteorological

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/ 16/47/55/56 Enclosure 7 Page 17 of 54 conditions based on 5 years of meteorological data collected from the plant MET tower.

Dose analyses were performed using the most prevalent stability class and wind speed conditions at each respective level on the MET tower. The selection and use of predominant meteorological dispersion is appropriate and in accordance with the intent of NEI 99-01.

Figure 6 graphically displays the relationship between monitor effluent control setpoint values, the Technical Specification limit, and the four EAL threshold values for the two normal operational discharge release pathways from North Anna. This figure demonstrates that the four EALs are sufficiently separated and show escalation from the NOUE level up through the GE level. Additionally, it can be seen that the NOUE threshold is set at or above the setpoint limit, thus assuring the operator will be alerted due to radiation monitors going into alarm prior to or when the NOUE level is exceeded.

Figure 6 - NAPS Vent and Proces s Vent EAL Esca lation

.- - * ' ' ' " " " " " " ' & : : * * , ,,.,,r - ,_ - << - , - " ° 'r NAPS Vents and Process Vent EAL Escalation

urn;;es 2 .6E-OS 2.0£<,();$

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J

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 18 of 54 Calculation Summary Gaseous Effluent Radiation Monitor Thresholds The following pertinent information has been extracted from Surry Calculation RP 01, "Surry Abnormal Rad Release Gaseous EAL Thresholds Based on NEI 99-01, Revision 6". It is provided to assist technical reviewers that will be evaluating this license amendment request.

Purpose:

Calculation of new Emergency Action Levels were determined for radioactive releases from the SPS Ventilation Vent and Process Vent based on updated guidance from NEI 99-01, Rev 6 and revision to EPA-400.

References:

1. Nuclear Energy Institute NEI 99-01, Rev. 6, "Methodology for Development of Emergency Action Levels," November 2012.

2. Software-Meteorological Information and Dose Assessment System, MIDAS, Version 1.5.17.022218.

3. MIDAS Software QA Documentation, SQA-MIDAS-DOM-20161219 and all previous files.

4. NUREG-1228, "Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents", McKenna, T. J. and Gutter, U.S. Nuclear Regulatory Commission, Washington, D.C, 1988.

5. VPAP-2103S, Revision 20, "Off-site Dose Calculation Manual (Surry)."

6. EPIP-4.03, Revision 20, "Surry Power Station - Dose Assessment Team Controlling Procedure".

7. Nuclear Energy Institute NEI 99-01, Rev. 4, "Methodology for Development of Emergency Action Levels," January 2003.

8. Calculation PA-0224, Revision 0, Addendum (00, OOA, 008, OOC,D, and E), Surry Power Station Radiation Monitor Emergency Action Levels (EALs) for the Process Vent and Ventilation Vent #2, Steam Line, and Auxiliary Feed water Exhaust".

9. HP-3010.040, Revision 36, "Surry Power Station - Radiation Monitoring System Setpoint Determination".

10. EPA-400/R-17/001, "PAG Manual: Protective Action Guides and Planning Guidance for Radiological Incidents", January 2017.

11. RP-AA-151, Revision 0, "Radiation Protection Technical Bases Analyses or Calculations".

12. DC SU-10-01083, Rev. 01, "NRC Radiation Montiors Replacement Project".

13. NE-GL-0035S, "PC-MIDAS Guideline", Revision 11.

MIDAS Dose Software:

MIDAS software (Ref 2) was utilized to determine the projected EDE, TEDE and Thyroid COE for a one (1) hour release duration. Integrated TEDE for a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months release

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 19 of 54 duration is used for the purpose of calculating emergency action levels for ALERT, Site Area and General Emergency classifications.

MIDAS is classified per the Software Quality Assurance program as class 3 software (Ref. 3).

MIDAS Assumptions and Inputs:

Met"> .

LOCA - RCS Leak ProcYnt GE LOCA - RCS leak

For all MIDAS runs, the stability class=E, ambient temperature=50°, and the direction=252° (from} to result in MIDAS shortest distance to the site boundary= 503 m (-0.31 mi). (5° added to centerline to align over calculation point within MIDAS tabular report at 0.37 miles).

- The Process Vent system takes input from the Aux ..6l.@ atmosphere, Aerated Waste system. Containment Vacuum Pump discharge, Waste Gas Decay tanks, and from the Gas Stripper Surge drum. Two likely plant accident conditions could apply to effluents discharged from the Process Vent, (1) LOCNGAP/Spray/Filter/RCS leak - that represents activity in the RCS that is leakingkJ.to,Joo)JM~J:Y...m:12.LLO,.CA,:G8P.1N.Q.}.~R.WY.lEiJ1erl~mU.~aKftwn__GN.MI-that represents discharge from CN!\IT during early onset of a LOCA before isolation. In MIDAS, the isotopic mix for both of these scenarios is the same with a total credited OF of 3000.

Based on the information above, one run is required for each of the release points as all input assumptions are identical and can be ralio~d to the applicable EAL threshold.

Method of Calculation:

The meteorology and source terms used to develop the threshold values were chosen to best represent the conditions that would be expected at the time of the emergency for each respective action level.

The calculated threshold values consider appropriate source term and meteorology in accordance to NEI 99-01. The resulting values are adequately conservative and represent the best estimate of the release rates that would result in exceeding the dose criteria of NEI 99-01. The values determined show consistent classification escalation from RU1 through RG1. The RU1 thresholds are based on the REMODCM Instantaneous Release Rate Limits that utilize annual average meteorology. They are based on the same principles of dose and maintain consistency with the Technical Specifications. Sufficient margin exists between plant setpoint alarms and the EAL thresholds to provide sufficient awareness to the Operators prior to reaching the NOLIE emergency condition.

The Unusual Event (UE) EALs are calculated for release points controlled in the ODCM, Ref. 5.

To determine the ALERT, SAE and GE EAL radiological thresholds for Ventilation Vent and Process Vent, MIDAS was used to predict expected doses based on best estimate meteorological and plant conditions. Inputs to MIDAS use most prevalent met data and

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 20 of 54 expected release point parameters together with event tree, core condition, mitigating reduction factors, and normalized source terms of 1 uCi/cc for vent and process vent rad monitors. An assumed one-hour decay time since shutdown and a one-hour duration of release are applied in each computer run. The mitigating reduction mechanisms (decay, sprays, filters, etc.) input into MIDAS for a given accident event determine the final radiological release source term mix. The MIDAS outputs generated for each release option represent a radiological prediction normalized to the source entered (e.g., 1 uCi/cc).

For Ventilation Vent and Process Vent releases, a LOCA accident type is selected for the event tree. A fuel handling accident was not run in MIDAS since an additional mitigation reduction factor of 100 for the pool water would logically result in lower site boundary doses which would then lead to higher emergency action levels thresholds for the Ventilation Vent and Process Vent.

The maximum projected EDE, TEDE and Thyroid COE dose at or beyond the site boundary distance were obtained from the MIDAS outputs. These doses were divided into the applicable EAL criteria to determine the radioactivity concentration (uCi/cc) seen by the radiation monitor, which would yield the referenced dose criteria for a given emergency classification. These concentrations are the rad monitor action levels for the ALERT, SAE and GE emergency classifications. The lowest predicted concentration between each dose analyzed is selected as the applicable EAL limit.

==

Conclusion:==

Following the guidance of NEI 99-01 Revision 6, recommended values for Surry release point EAL thresholds based on the results of this calculation are summarized in Table 1 of the EAL Matrices.

For the Unusual Event (NOUE) threshold values determined, the NOUE values are set at 2 times the 'allocated' site-specific effluent release controlling document limits for 60 minutes or longer. A difference from the prescribed guidance in NEI 99-01 Revision 6 by adding the word 'allocated' is necessitated because the Process Vent instantaneous release rate limits following the ODCM guidance would result in NOUE threshold values greater than corresponding ALERT threshold values. The NOUE thresholds when exceeded for 60 minutes are indicative of the inability to terminate a radioactive release within prescribed regulatory and license limits and therefore represent a loss of plant control and degraded safety.

For the Unusual Event (NOUE) threshold values determined, the NOUE values are set at 2 times the 'allocated' site-specific effluent release controlling document limits for 60 minutes or longer. A difference from the prescribed guidance in NEI 99-01 Revision 6 by adding the word 'allocated' is necessitated because the Process Vent instantaneous release rate limits following the ODCM guidance would result in NOUE threshold values greater than corresponding ALERT threshold values. The NOUE thresholds when exceeded for 60 minutes are indicative of the inability to terminate a radioactive release

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 21 of 54 within prescribed regulatory and license limits and therefore represent a loss of plant control and degraded safety.

The ALERT, SAE and GE threshold values determined, represent a radioactive release that results in 1%, 10%, and 100% of the revised EPA Protective Action Guideline TEDE limits. These threshold limits were calculated using expected meteorological conditions based on 5 years of meteorological data collected from the plant MET tower.

Dose analyses were performed using the most prevalent stability class and wind speed conditions at each respective level on the MET tower. The selection and use of predominant meteorological dispersion is appropriate and in accordance with the intent of NEI 99-01.

Figure 3 graphically displays the relationship between monitor effluent control setpoint values, the Technical Specification limit, and the four EAL threshold values for the two normal operational discharge release pathways from Surry. This figure demonstrates that the four EALs are sufficiently separated and show escalation from the NOUE level up through the GE level. Additionally, it can be seen that the NOUE threshold is set at or above the setpoint limit, thus assuring the operator will be alerted due to radiation monitors going into alarm prior to or when the NOUE level is exceeded.

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 22 of 54 Figure 3 - SP S Vent and Process Vent EAL Escalation SPS Vent and Process Vent EAL Escalation 1:0E+CS - , , - - - - - - - - - - - - - - - - - - - - - - - ~

7.2.E+07 rx:.>=+/_ <<- . ~+o//} JW<<c:_**>,:./7 :* *-} ~W I

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, Release /j Teth U RU1 i

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Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 23 of 54 Calculation Summary Containment High Range Radiation Monitor Responses to a LOCA The following pertinent information has been extracted from Surry Calculation RA-0063, Expected Containment High Range Radiation Monitor Response to a LOCA Based on Fuel Rod Gap Fractions Defined in NUREG 1228. It is provided to assist technical reviewers that will be evaluating the Fission Product Barrier matrix portion of this license amendment request.

Purpose:

The purpose of this calculation is to define the expected containment high range radiation monitor response to a large break LOCA with containment and recirculation spray based on fuel gap fractions defined in NUREG 1228 for Emergency Action Level (EAL) values developed in accordance with NEI 99-01, Rev. 6. These detector responses will be used for event classification based upon Fuel Clad Degradation EALs and as a radiation indicator for Fuel Clad Barrier Loss.

References:

1. NEI 99-01, Rev. 6, "Development of Emergency Action Levels for Non-Passive Reactors," November 2012.

1. NUREG-1228, "Source Term Estimation during Incident Response to Severe Nuclear Power Plant Accidents," October 1988.

2. Drawing 11448-FM-1 E, Rev. 13, Sheet 1, "Mach. Loe. - Reactor Cont. Sections "A-A", "E-E" & "Z-Z" Surry Power Station - Unit 1.

3. Drawing 11448-FE-46C, Rev. 16, "Conduit Plan Reactor Containment El. 47' - 4" Surry Power Station - Unit 1."

4. PA-0163, Rev, 0, Add. D, "Calculation of the Surry AST LOCA Dose Consequences to Support the Gothic Containment Reanalysis for GSl-191."

5. RA-0008, Rev. 0 thru Add. B, Core Isotopic Inventories for Surry Dose Consequence Analyses Based on the Alternate Source Term, May, 2010.

6. Drawing 11448-FE-46D, Rev. 12, "Conduit Plan Reactor Containment El. 47' -4" Surry Power Station - Unit 1."

7. SEALTB Rev. 4, "Emergency Action Level Technical Bases Document", December 2013.

8. Drawing 11448-FP-13D, Rev. 14, Sheet 4, "Containment & Recirc Spray System Sh 4."

9. Radiological Health Handbook, January 1970.

10.Drawing 11448-FM-1G, Rev. 14, Sheet 1, "Mach. Loe. - Reactor Cont. Sections "C-C" & "D-D" Surry Power 11 . Station - Unit 1."

12.SEAL MATRICES Rev. 4, "Surry Power Station Emergency Action Level Matrix".

Computer Codes Used:

Microshield version 7.02

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 24 of 54 Key Inputs and Assumptions Used:

1. 5% fuel clad damage representing 300 uCi/gm

2. Instantaneous release and dispersal of reactor coolant noble gas and iodine inventory into containment

3. Removal of iodine from the containment atmosphere due to containment spray operating

4. Response from noble gas concentration in containment above the operating floor

5. Release fraction from the fuel gap: 3% noble gasses (Ref. 2)

Methodology:

Microshield is used in this analysis to calculate the expected response from the Containment High Range Monitors.

Results and/or

Conclusions:

Dose Rates vs Decay Time for the 5% Clad Damage Decay time (hrs) Dose Rate (R/hr) 0 4.27E+02 1 1.24E+02 2 8.97E+01 4 6.05E+01 8 3.52E+01 16 2.12E+01 24 1.76E+01 36 1.53E+01 48 1.40E+01 72 1.20E+01 Figure 1: Dore Rate vs Decay Time for ll'.e 5% Clad Damage r------**--* *---1

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DoseRate(R/hr)vs DecayTime(hr$) ---

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Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 25 of 54 e Rates vs Decay Time for the 20% Clad Failure Decay time (hrs) Dose Rate (R/hr) 0 1.71 E+03 1 4.97E+02 2 3.59E+02 4 2.42E+02 8 1.41 E+02 16 8.45E+01 24 7.02E+01 36 6.11 E+01 48 5.57E+01 72 4.79E+01 Figure 2: Dose Rate vs Decay lime for tire 20% Clad Failure Dose Rate (R/hr)vs Decayllme (hrs)

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,;MJI.,<a *. - - - - * * - - - - - - - - * - * - * * * * - * * * * - - - - - - - * - - * * - * * - - - - - - - - - - - - - - - - * - * - *

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Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 26 of 54 Calculation Summary Containment High Range Radiation Monitor Responses to a LOCA The following pertinent information has been extracted from North Anna Calculation RA-0064, Expected Containment High Range Radiation Monitor Response to a LOCA Based on Fuel Rod Gap Fractions Defined in NUREG 1228. It is provided to assist technical reviewers that will be evaluating the Fission Product Barrier matrix portion of this license amendment request.

Purpose:

The purpose of this calculation is to define the expected containment high range radiation monitor response to a large break LOCA with quench (containment) and recirculation spray based on fuel gap fractions defined in NUREG 1228 for Emergency Action Level (EAL) values developed in accordance with NEI 99-01, Rev. 6. This calculation supports a revision to the North Anna EALs. These detector responses will be used for event classification based upon Fuel Clad Degradation EALs and as a radiation indicator for Fuel Clad Barrier Loss.

References:

1. NEI 99-01, Rev. 6, "Development of Emergency Action Levels for Non-Passive Reactors," November 2012.

2. NUREG-1228, "Source Term Estimation during Incident Response to Severe Nuclear Power Plant Accidents," October 1988.

3. Drawing 12050-FM-1A, Rev. 19, Mach. Loe. - Reactor Cont. Sh. 1 Plan EL 291'-10" North Anna Power Station - Unit 2."

4. NEAL MATRICES Rev. 7, "North Anna Power Station Emergency Action Level Matrix".

5. PA-0186, Rev. 0, "Containment High Range Radiation Monitor Accident Response Curves for North Anna and Surry," March 4, 2002.

6. PA-0186, Rev. 0, Add. A, "Containment High Range Radiation Monitor Accident Response Curves for North Anna and Surry," Sept. 7, 2006.

7. Calculation 11715-ES-017, Rev. 0, "North Anna 1 & 2 Containment Free Volume,"

Aug. 31, 1971.

8. Drawing 12050-FM-1 C, Rev. 18, "Mach. Loe. - Reactor Cont. Sh. 3, Plan - El. 241' -

O" North Anna Power Station - Unit 2."

9. NEALTBD Rev. 7, "Emergency Action Level Technical Bases Document", March 2015.

10. Radiological Health Handbook, January 1970.

11. NA-W0-000-00426795-01, Work Order Task for Reactor Containment Elevation 291 Area High Rad Monitor, March 18, 2000.

12. Drawing 11715-FM-1 G, Rev. 18, Sheet 1, "Mach. Loe. - Reactor Cont. SH7 Sections 3-3 & 4-4 North Anna Power Station."

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 27 of 54 Computer Codes Used:

Microshield version 7.02 Key Inputs and Assumptions Used:

1. 5% fuel clad damage representing 300 uCi/gm

2. Instantaneous release and dispersal of reactor coolant noble gas and iodine inventory into containment

3. Removal of iodine from the containment atmosphere due to containment spray operating

4. Response from noble gas concentration in containment above the operating floor

5. Release fraction from the fuel gap: 3% noble gasses (Ref. 2)

Methodology:

Microshield is used in this analysis to calculate the expected response from the Containment High Range Monitors.

Results and/or

Conclusions:

Dose Rates vs Decay Time for the 5% Clad Damage Decay time (hrs) Dose Rate (R/hr) 0 5.81E+02 1 1.69E+02 2 1.22E+02 4 8.29E+01 8 4.91E+01 16 3.04E+01 24 2.56E+01 36 2.26E+01 48 2.07E+01 72 1.78E+01 Figure 1: Dose Rate vs Decay Time Tor the 5% Clad Damage Dose Rate (R/hr) vs Decay Time (hrs) 7.00[-f02 ...............................................................................................................................................................................................................

6.00[-f02 ............................................................................................................................................................................................................

5.00E->02 -<------------------

Dose Rate 4.00E+o2 --¥--------

(R/hr) 3.00E-f02 * ...........................................................................................................................................................................................................

2.00[-f02 *+,............................................................................................................................................................................................................

0 10 20 30 40 50 60 70 110 Hours After Shutdown

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 28 of 54 Dose Rates vs Decay Time for the 20% Clad Failure Decay time (hrs) Dose Rate (R/hr)

O 2.32E+03 1 6.75E+02 2 4.88E+02 4 3.31E+02 8 1.96E+02 16 1.21E+02 24 1.03E+02 36 9.01 E+01 48 8.29E+01 72 7.16E+01 Figure 2: Dose Rate vs Decay Time forlhe 20% Clad Failure Dose Rate (R/hr) vs Decay Time (hrs) 2.00E+03 Dose Rate 1.50E-f03 (R/hr) 1.00E+03 5.00E-f02 0 10 20 30 40 50 60 70 80 Hours After Shutdown

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 29 of 54 Calculation Summary Containment High Range Radiation Monitor Responses to a LOCA The following pertinent information has been extracted from Calculation RA-0074, Millstone Unit 2 Expected Containment High Range Radiation Monitor Response to a LOCA Based on Fuel Rod Gap Fractions Defined in NUREG 1228. It is provided to assist technical reviewers that will be evaluating the Fission Product Barrier matrix portion of this license amendment request.

Purpose:

The purpose of this calculation is to define the expected containment high range radiation monitor system response for Millstone Unit 2 to a large break LOCA with containment and recirculation spray based on fuel gap fractions defined in NUREG 1228 for Emergency Action Level (EAL) values developed in accordance with NEI 99-01, Rev. 6. This calculation supports a revision to the Millstone Unit 2 EALs. These detector responses will be used for event classification based upon Fuel Clad Degradation EALs and as a radiation indicator for Fuel Clad Barrier Loss.

References:

1. Vendor Calculation 3D00-005, Rev. 3, "Millstone Unit 2 Containment Heat Sinks,"

December 2006.

2. Nuclear Energy Institute Document NEI 99-01, Rev. 6, "Development of Emergency Actions Levels for Non Passive Reactors," November 2012.

3. Reference Manual MP-26-EPA-REF02, Rev. 24, "Millstone Unit 2 Emergency Action Level (EAL) Technical Basis Document," March 2016.

4. NUREG-1228, "Source Term Estimation during Incident Response to Severe Nuclear Power Plant Accidents," October 1988.

5. Drawing 25203-28014, Rev. 15, "Millstone Unit 2 Instrument Location Containment Plan El. 14'-6" & 38'-6".

6. Drawing 25203-27021, Rev. 3, "Millstone Nuclear Power Station Unit No. 2 General Arrgt - Containment & Aux. Bldg. Section A-A".

7. 7. Engineering Calculation M2AST-03105R2, Rev. 0, "Millstone 2 Alternate Source Term," January 2002.

8. 8. Radiological Health Handbook, January 1970.

9. Computer Code MicroShield, Version 7.02, Grove Software, Inc.

10. Millstone Unit 2 EALs MP-26-EPI-FAP06-002, Rev. 7.

11. Drawing 25203-20104, Rev. 3, "Millstone Nuclear Power Station Unit No. 2 Area 5 Piping Containment Spray% H2 Purge".

12. Drawing 25203-27022, Rev. 9, "General Arrangement Containment & Aux. Bldg.

Section 8-8".

13. Engineering Calculation NUC-181, Rev. 1, "MP-2 Design-Basis Loss of Coolant Accident - Radiation Source Terms," June 1998.

14. Vendor Technical Manual VTM-303-007A, "Energy Response Test & Dose Rate Calibration of Model RD-23 Det." June 1986.

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 30 of 54 Computer Codes Used:

MicroShield 7 .02 Key Inputs and Assumptions Used:

1 . 5% fuel clad damage representing 300 uCi/gm

2. Instantaneous release and dispersal of reactor coolant noble gas and iodine inventory into containment

3. Removal of iodine from the containment atmosphere due to containment spray operating

4. Response from noble gas concentration in containment

5. Release fraction from the fuel gap: 3% noble gasses (Ref. 2)

Methodology:

The region of containment measured by each of the radiation monitors is modeled using a simplified rectangular box configuration. This geometry is used in MicroShield Version 7.02 [Reference 9] with the receptor location being the radiation monitor location.

MicroShield is used in this analysis to calculate the expected response from the Containment High Range Radiation Monitors.

Results and Conclusions The following results depict the expected CHRRMS response in terms of the dose rates at the various times for 5% clad damage.

5% Clad Damage Table Decay Time(hrs) Dose Rate (R/hr) 0 266 1 78 2 55 4 34 8 16 16 6.0 24 3.7 36 2.6 48 2.1 72 1.7

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 31 of 54 Figure 13.2 5% Fuel Clad Damage Dose Rate (R/hr} vs. Decay nme (hrs)

to 0 10 ao ~o so 70 so Hours After Shutdm.,m The following results depict the expected CHRRMS response in terms of the dose rates at the various times for 5% clad damage. These results were generated by multiplying the 5% fuel clad damage by a factor of 4, since the concentrations of dispersed nuclides are 4 times greater between the 20% and 5% calculations.

20% Clad Damage Table Decay Time(hrs) Dose Rat (R/hr) 0 1065 1 314 2 223 4 138 8 64 16 23 24 14 36 10 48

8.5 72 6.9

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 32 of 54 Flgurel3A 20% Fuel Clad Damage Dose Rate {R/hr) vs. Decay Time (hrs)

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 33 of 54 Calculation Summary Containment High Range Radiation Monitor Responses to a LOCA The following pertinent information has been extracted from Calculation RA-0075, Millstone Unit 3 Expected Containment High Range Radiation Monitor Response to a LOCA Based on Fuel Rod Gap Fractions Defined in NUREG 1228. It is provided to assist technical reviewers that will be evaluating the Fission Product Barrier matrix portion of this license amendment request.

Purpose:

The purpose of this calculation is to define the expected containment high range radiation monitor system response for Millstone Unit 3 to a large break LOCA with containment and recirculation spray based on fuel gap fractions defined in NUREG 1228 for Emergency Action Level (EAL) values developed in accordance with NEI 99-01, Rev. 6. This calculation supports a revision to the Millstone Unit 3 EALs. These detector responses will be used for event classification based upon Fuel Clad Degradation EALs and as a radiation indicator for Fuel Clad Barrier Loss.

References:

1. Vendor Calculation ES-227, Rev. 0, "Containment Structure Free Volume,"

November 1979.

2. Millstone Unit 3 EALs MP-26-EPI-FAP06-003, Rev. 11.

3. Computer Code MicroShield, Version 7.02, Grove Software, Inc.

4. Engineering Calculation RERM-04345R3, Rev. 0, "Millstone Unit 3 Containment High Range Radiation Monitors' Accident Responses to a LOCA," April 2008.

5. Nuclear Energy Institute Document NEI 99-01, Rev. 6, "Development of Emergency Actions Levels for Non-Passive Reactors," November 2012.

6. Reference Manual MP-26-EPA-REF03, Rev. 21, "Millstone Unit 3 Emergency Action Level (EAL) Technical Basis Document," May 2016.

7. NUREG-1228, "Source Term Estimation during Incident Response to Severe Nuclear Power Plant Accidents," October 1988.

8. Drawing 25212-27012, Rev. 16, "Millstone Nuclear Power Station Unit No. 3 Machine Location - Containment Structure- Plan El 51'-4"."

9. Engineering Calculation NUC-181, Rev. 1, 'MP-2 Design-Basis Loss of Coolant Accident - Radiation Source Terms," June 1998.

10. Radiological Health Handbook, January 1970.

11. Drawing 25212-27015, Rev. 12, "Millstone Nuclear Power Station Unit No. 3 Machine Location - Containment Structure - Section 3-3."

12. Engineering Calculation M3AST-01942R3, Rev. 1, "Millstone 3 Alternate Source Term," May 2006.

13. Vendor Technical Manual VTM-303-007A, "Energy Response Test & Dose Rate Calibration of Model RD-23 Det." June 1986.

Computer Codes Used:

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 34 of 54 MicroShield 7.02 Key Inputs and Assumptions Used:

1. 5% fuel clad damage representing 300 uCi/gm

2. Instantaneous release and dispersal of reactor coolant noble gas and iodine inventory into containment

3. Removal of iodine from the containment atmosphere due to containment spray operating

4. Response from noble gas concentration in containment above the operating floor

5. Release fraction from the fuel gap: 3% noble gasses (Ref. 2)

Methodology:

The source volume bounded by the annular crane wall is modeled using two right cylinder volumes, with a rectangular volume representing the pressurizer cubicle black body. This geometry is used in MicroShield Version 7.02 [Reference 3] with the receptor locations being the radiation monitor locations. MicroShield is used in this analysis to calculate the expected response from the Containment High Range Monitors.

Results and

Conclusions:

The following results depict the expected CHARMS response in terms of the dose rates at the various times for 5% clad damage.

5% Clad Damage Table Decay Time Dose Rate (R/hr) Dose Rate (R/hr)

(hrs) (3RMS*RE05A) (3RMS*RE04A) 0 703 550 1 191 149 2 132 103 4 83 65 8 40 31 16 16 13 24 8.5 6.8 48 7.3 5.8 72 6.0 4.8

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 35 of 54

- - - ---- -- Fie,,ireH.2 5% Fuel Clad Damage Dose Rate {R/hr) vs. Decay Time (hrs)

-.-)1<M5'"'R.EC5A 31\MS*itrotA

- - - - y*~-* ..* ,, , - _,; :

"----,.--- ....-'---~ - - . . - - ---- -s 1C 10 30 4r)

Houri .\tt¥:!S' Shutdown SO

"° The following results depict the expected CHARMS response in terms of the dose rates at the various times for 20% clad damage. These results were generated by multiplying the 5% fuel clad damage by a factor of 4, since the concentrations of dispersed nuclides are 4 times greater between the 20% and 5% calculations.

20% Clad Damage Table Decay Time Dose Rate (R/hr) Dose Rate (R/hr)

(hrs) (3RMS*RE05A) (3RMS*RE04A) 0 2814 2202 1 764 599 2 530 415 4 332 260 8 160 126 16 66 53 24 44 35 36 34 27 48 29 23 72 24 19

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/ 16/47/55/56 Enclosure 7 Page 36 of 54 Figure 13.4 20% Fuel Clad Damage Dose Rate (R/hr) vs. Decay Time (hrs) f' 'J~n.r !_ *.r ').' :. r1

(.*. .. ;.

,,'J ; *i: ., *..1 i . ! ! / .:-:. *., ..

O'"'eRatt:

(11/h*)

-\--"'~-'-~-~-*- -~--~--'-t------.---.-------.------~

40 10 20 30 50 60 70 80 Hours Afref Shutdown

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 37 of 54 Calculation Summary Detector Response to an RCS Sample The following pertinent information has been extracted from Fleet Calculation RA-0059, Detector Response to an RCS Sample for EAL Classification of Fuel Clad Degradation and Barrier Loss. It is provided to assist technical reviewers that will be evaluating the Fission Product Barrier matrix portion of this license amendment request.

Purpose:

The purpose of this calculation is to determine the detector response to various depressurized RCS samples measured with a gamma detectorat a distance of 1 ft. This calculation supports the Emergency Action Levels (EALs) for NAPS, SPS, MPS2 and MPS3. These detector responses will be used for event classification based upon Fuel Clad Degradation EALs and as a radiation indicator for Fuel Clad Barrier Loss.

References:

1. SEAL MATRICES Rev. 4, "Surry Power Station Emergency Action Level Matrix".

2. NEAL MATRICES Rev. 5, "North Anna Power Station Emergency Action Level Matrix".

3. MP-26-EPI-FAP06-002 Rev. 9, "Millstone Unit 2 Emergency Action Levels".

4. MP-26-EPI-FAP06-003 Rev. 8, "Millstone Unit 3 Emergency Action Levels".

5. NUREG-1228, "Source Term Estimation during Incident Response to Severe Nuclear Power Plant Accidents"

6. Federal Guidance Report 12, EPA-402-R-93-081, "External Exposure to Radionuclides in Air, Water and Soil"

7. 1304952001-UR-0001 Rev. 0 Add. A, "Primary Coolant Design/ Technical Specification Activity Concentrations"

8. RA-0008 Rev. 0 Add. 0, "Core Isotopic Inventories for Surry Dose Consequence Analyses Based on the Alternate Source Term", May 2010.

9. PA-0089 Rev. 0 through Add. B, "Surry Steam Generator Tube Rupture [SGTR]

Dose Calculations at the EAB,. the LPZ and in Control Room", August 2000.

10. Robert C. Weast, ed., "CRC Handbook of Chemistry and Physics 60th edition", pg.

F-324, CRC Press, Inc.

11. PA-0194 Rev. 0 Add. 0, "Radiological Consequences of a Steam Generator Tube Rupture at North Anna Based on the Alternate Source Term", April 2003.

12. PA-0186 Rev. 0 Add. 0, "Containment High Range Radiation Monitor Accident Response Curves for North Anna and Surry"

13. PA-0081 Rev. 0 Add. 0, "North Anna SGTR Doses at the EAB, LPZ, and in Control Room", February 1991 .

14.06-ENG-04217R3 Rev. 0 CCN 1, "MP3 SPU Primary Coolant Design and Technical Specification Activity Concentrations etc."

15. M3AST-01942R3 Rev. 1 CCN 1, "Millstone 3 Alternate Source Term", May 2006.

16. M3ASTSGTR-04072R3 Rev. 1, "MP3 Uprated AST Steam Generator Tube Rupture Dose Consequences Analysis", April 2007.

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 38 of 54

17. M2AST-04080R2 Rev. 0 Add. 1, "MP2 Coolant Activity for Accident Analyses",

March 2005.

18. M2AST-03105R2 Rev 0, "Millstone 2 Alternate Source Term", December 2002.

19. SEALTB Rev. 4, "Emergency Action Level Technical Bases Document", December 2013.

20. NEALTBD Rev. 5, "Emergency Action Level Technical Bases Document", December 2013.

21. MP-26-EPA-REF02 Rev. 022, "Millstone Unit 2 Emergency Action Level (EAL)

Technical Basis Document"

22. MP-26-EPA-REF03 Rev. 018, "Millstone Unit 3 Emergency Action Level (EAL)

Technical Basis Document"

23. NEI 99-01 Rev. 6, "Development of Emergency Action Levels for Non-Passive Reactors", November 2012.

Computer Codes Used:

MICROSHIELD Version 7.02 WATPROP Version 4 Key Inputs and Assumptions Used:

1. 5% fuel clad damage representing 300 uCi/gm

2. Tech Spec RCS coolant activities and limits

3. Response from iodine concentration in degassed RCS sample

4. Release fraction from the fuel gap: 2% iodine (Ref. 5)

5. Detector placed 12 inches from outer edge of sample volume

6. Relative insensitivity of detector response demonstrated in the calculation to minor differences in distance to the sample volume

7. Relative insensitivity demonstrated in the calculation to sample container geometry

8. Relative proportional dose rate response demonstrated in the calculation to sample volume Method of Analysis:

The Technical Specification DE 1-131 spike iodine concentrations and core inventory iodine activity along with the RCS mass will be used to determine the sources for the various sample volumes. This source is then modeled in the code Microshield 7.02 with a dose point one foot from the source.

Results and

Conclusions:

In summary, dose rate responses have been determined for SPS, NAPS, MPS3 and MPS2 for TS coolant activity spikes and 5% cladding failure, which are used to develop EALs for Fuel Clad Degradation and Fuel Clad Barrier Loss, respectively, in various sample sizes and at several times after shutdown. A summary of conservative values representative of the expected detector response for Fuel Clad Degradation and Fuel Clad Barrier Loss in terms of mR/hr/ml vs. decay post-shutdown are presented Tables 1 and 2 below.

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 39 of 54 Table 1: Summary of Unpressurized RCS Sample Dose Rates taken at 1 foot_

for Fuel Clad Degradation vs. Decay Post-Shutdown Station/Unit 1 hr 2 hr 4 hr 8 hr 12 hr 24 hr (mR/hr/ml) (mR/hr/ml) (mR/hr/ml) (mR/hr/ml) (mR/hr/ml) (mR/hr/ml)

SPS 0.15 0.13 0.10 0.07 0.06 0.04 NAPS 0.76 0.66 0.54 0.40 0.33 0.21 MPS2 0.76 0.66 0.54 0.40 0.33 0.21 MPS3 0.76 0.66 0.54 0.40 0.33 0.21 Table 2: Summary of Unpressurized RCS Sample Dose Rates taken at 1 foot for Fuel Barrier Loss vs. Decay Post-Shutdown Station/Unit 1 hr 2 hr 4 hr 8 hr 12 hr 24 hr (mR/hr/ml) (mR/hr/ml) (mR/hr/ml) (mR/hr/ml) (mR/hr/ml) (mR/hr/ml)

SPS 17 12.7 8.5 5.4 4.0 2.2 NAPS 17 12.7 8.5 5.4 4.0 2.2 MPS2 17 12.7 8.5 5.4 4.0 2.2 MPS3 17 12.7 8.5 5.4 4.0 2.2 The scale of the response factors in the tables above are normalized to 1 ml. Samples when obtained in the plant will likely consist of greater volume (e.g., 120 or 250 ml).

Plant sampling procedures will direct to take a reading from 1 foot and divide by the actual collection volume and report the reading as (mR/hr per ml).

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 40 of 54 Calculation Summary Post-Accident Radiation Response for Primary Sample Line The following pertinent information has been extracted from Fleet Calculation RA-0079, Post- Accident Radiation Response Curves for Primary Hot Leg Sample Lines. It is provided to assist technical reviewers that will be evaluating the Fission Product Barrier matrix portion of this license amendment request.

Purpose:

The purpose of this calculation is to document the generation of radiation response curves for primary hot leg sample lines assuming an intact RCS. The accident scenario analyzed modeled 5% failed fuel (gap release).

References:

1. ORNUTM-2005/39, Version 6.2.2, "SCALE Code System".

2. PA-0219, Rev. 0, "Post-Accident Radiation Response Curves for North Anna Primary Hot Leg Sample Lines."

3. ET-NAF-05-0013, Rev 0, "Post-Accident Radiation Measurement and Accident Classification Based on PA-0219."

4. Memorandum MP-CHEM-13-01 dated March 19, 2013, "Bases for Proposed M2 and M3 EALs."

5. MGP Instruments Document# 15-00031, Revision 5, "Area Monitor Probes AMP 50-100-200 Operations and Maintenance Manual."

Computer Codes Used:

SCALE 6.2 (Reference 1)

Microshield 7.02 Key Inputs and Assumptions Used:

1. 5% fuel clad damage representing 300 uCi/gm

2. Response from gap release of noble gas, halogens, and cesiums

3. Release fraction from the fuel gap: 3% (Ref. 2)

4. Detector placed 2 inches from outside of various diameter sample line tubing

5. Sample line tubing length of 240 cm Methodology:

Source Term The initial source term for 5% failed fuel (gap release) was determined by multiplying the 1% failed fuel source term by a factor of 5. The results were then scaled down by the ratio of the water volume in the tubing to the RCS liquid inventory. The resulting source term in the tubing was decayed and converted to a photon spectrum using the ORIGEN code from the SCALE 6.2.2 code package.

Shielding Model - Scenario 1:

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 41 of 54 The sample line tubing was modeled as straight length of tubing running vertically up a concrete wall. The shielding model included the wall, tubing, and water within the tubing. The SCALE module MAVRIC was used to perform the photon transport. This is an improvement over the QADS model in that scattering/reflection from the concrete surfaces will be included and a dose rate measurement volume can be used in lieu of a point location. A tally volume comprised of air will be used to calculate the dose rate based on SCALE-supplied ANSl-77 flux-to-dose conversion factors.

Tubing I Detector Shielding Model - Scenario 2:

The sample line tubing was modeled as straight length of tubing. The shielding model included the tubing and water within the tubing. Microshield was used to determine the dose rate at a point 2 inches from the outside of the tubing. This scenario is designed to support a broad spectrum of possible sample line locations and geometries.

Results &

Conclusions:

Dose rates from a primary hot leg sample line assuming an intact RCS and 5% failed fuel were calculated. Tabular results and an associated radiation response curve are provided below. An adjustment factor to correct for a range of tubing sizes is provided.

These results are valid for North Anna, Surry and Millstone Power Stations.

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 42 of 54 20 ------..----.----*----- --*-**---**---**- **-***-**------**-*

18 + - - ~ - -I. . - - - - - + - - - + - - - - - t - - - - - + - - - - - l

.c 14 12 16 +---~\--+-----+------1-------,1------t--------!

+--~~--+-----+---+-----t------i------,----1

\_ !

e10 !

~

-**-**---***********-*****-***t-~ ~ *-----*-* ---***--****-********-** -** *-*-* *-* * * *-*-* * * *

8 '

6 4

r-t*--t---==r~=-,;;;;:;;:;;;;;;;~:::::::i 2 -+-------\.-----+---+-----+-----!----!

0 +--~--..--.--,--,--+--,---,--,--+--,....-,.---,--+--.--,--,--+-.....--,---r--l 0 4 8 12 16 20 24 Decay (hrs)

In some instances the tubing diameter may be larger than 3/8" tubing. In those instances the results can be scaled up by the ratio of tubing cross sectional area. The following scaling factors should be used:

Tubing size lijuid Area Scaling (Nominal OD) (in:) Factor 3/8" 0.110 1.0 1/2" 0.196 1.8 3/4" 0.442 4.0 Note that the tubing size scaling factors are first-order approximations. Tubing wall thickness can vary somewhat, and the changing solid-angle between source and detector will also vary. However, the table is considered a reasonable indicator of the dose rate changing with tubing size.

Shielding - Scenario 2 The source term determined was used as a source term in MicroShield to determine the dose rate at a location 2" from the outside edge of a sample line. The sample line was modelled as a cylinder of water. The MicroShield output shows following results.

Decay Step Dose Rate (hrs) (R/hr) 1 4.5 2 3.45 4 2.35 8 1.5 16 0.95 24 0.7

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 43 of 54 5 .000 * ****-**--*---**-****-***** * - * - - - -**--*--*-**--- -*-**-***------ ---*-*-*-***-* *-***-**--****--**-**-****

4.000 + - - l l - - - + - - - + - - - - - + - - - - - + - - - . + - - - - - - !

.c

~

0 4 8 12 16 20 24 Decay (hrs)

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 44 of 54 Calculation Summary Post-Accident Radiation Monitor Response for Core Uncovery The following pertinent information has been extracted from Calculation RA-0078, Verification of Rad Monitor Response to Core Uncovery. It is provided to assist technical reviewers that will be evaluating this license amendment request for MPS3, NAPS & SPS. (Note: MPS2 is analyzed in M2EP-04164R2).

Purpose:

The purpose of this calculation is to determine a single, common value for radiation monitor responses to core uncover for Millstone Unit 3, North Anna Units 1 & 2, and Surry Units 1 & 2. This analysis is only applicable when the reactor head is off of the vessel.

References:

1. ORNUTM-2005/39, Version 6.2.2, "SCALE Code System".

2. LA-CP-14-00745, Rev. 0, "MCNP6 User's Manual, Code Version 6.1.1beta."

3. Calculation M3EP-04140R3 Rev. 0, "MP3 Rad Monitor Response to Core Uncovery."

4. MP-DWG-000-25212-27012 SH-00000000 Rev. 16, "Machine Location Cntmt Structure Plan El 51 Feet 4 Inches."

5. MP-DWG-000-25212-27013 SH-00000000 Rev. 13, "Machine Location Cntmt Structure Section 1-1 and 4-4."

6. MP-DWG-000-25212-27014 SH-00000000 Rev. 14, "Machine Location Containment Structure Section 2-2."

7. MP-DWG-000-25212-27015 SH-00000000 Rev. 12, "Machine Location Containment Structure Sect 3-3."

8. MP-DWG-000-25212-11060 SH-00000000 Rev. 9, "Plan Elevation 51 Feet 4 Inch Outline Containment Structure."

9. ETE-NAF-2014-0098 Rev. 0, "Millstone Unit 3 cycle 17 Nuclear Design Report."

10. M3AST-01942R3 Rev. 1, "Millstone 3 Alternate Source Term."

11. PNNL-15870 Rev. 1, "Compendium of Material Composition Data for Radiation Transport Modeling."

12. ET-NAF-06-0114 Rev. 0, "Dose Rate at the Containment Manipulator Crane Radiation Monitor Due to a Draindown Event Including Scatter from the air and Containment Dome."

13. PA-0227 Rev. 0, "Dose Rate at the Containment Manipulator Crane Radiation Monitor Due to a Draindown Event at North Anna or Surry."

14. M P-DWG-000-25212-11 075 SH-00000000 Rev. 3, "Containment Structure Section 1-1."

Computer Codes Used:

SCALE 6.2 (Reference 1)

MCNP 6.1 (Reference 2)

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 45 of 54 Key Inputs and Assumptions Used:

1. Water level at the top of the active fuel

2. Reactor head removed

3. Millstone 3 Core Inventory Methodology:

Source Term The LOCA core inventory source term will be decayed and converted to a photon spectrum using the ORIGEN code from the SCALE 6.2.2 code package.

Shielding Model The core, refueling cavity, and containment will be modeled using MCNP 6.1. The fuel region of the core will be modeled as a single homogeneous zone containing fuel and water. The water level is modeled as being level with the top of the active fuel. The gamma source in the fuel is distributed axially to model a nominal axial fuel burnup.

This distribution is intended to capture the self-shielding in the fuel zone. The gamma source has a uniform radial distribution in the fuel region. Structures inside the containment dome are limited to the crane wall and concrete structures close to the radiation monitors. These structures should reasonably model the photon backscatter to the radiation monitors. The radiation monitors will be modeled as finite air volumes used exclusively to tally the dose rate.

Results and/or

Conclusions:

Results from the MCNP output file provides an estimate of the average photon Mean Free Path (MFP) in each cell. An examination of these results indicates that the MFP in air varied between approximately 60 and 90 meters (197 and 295 feet). Given the distance between the operating floor and the containment dome, few photons interactions in air would occur during a photon's transit from the core to the containment dome and back to a radiation monitor.

Reference 3 also assumed that, other than air scatter, the primary contributor to radiation monitor dose rates was from photons that traveled vertically from the fuel through the air and scattered on the containment dome. This assumption was evaluated by modifying the MCNP input files to terminate any photon track that enters the refueling deck and refueling cavity concrete surfaces. The change was implemented by setting the importance of cell 10 to O (i.e. imp:p=O). The dose rates at the radiation monitors were reduced by -69% at RE-05A and -42% at RE-04A, bringing the results into reasonable agreement with the containment dome contributions calculated in Reference 3. analysis shows much of the dose rates at the radiation monitors in this calculation is due to photons scattering on or transiting through the refueling cavity concrete. Thus, while photon scattering on the containment dome is a contributor to the dose rates, the primary contributor is attributed to other concrete surfaces.

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 46 of 54 It was found that a drain-down event earlier in a refueling outage could increase the count rates by a factor of three. This is consistent with the change in photon intensity.

In summary, containment radiation monitor dose rates at Millstone, North Anna, and Surry are expected to be between 3 R/hr and 40 R/hr, depending on the unit, time after refueling, and radiation monitor location in containment. A value of 3 R/hr would be a conservative dose rate for use in identifying a potential drain-down event.

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 47 of 54 Calculation Summary Post-Accident Radiation Monitor Response for Core Uncovery The following pertinent information has been extracted from Calculation M2EP-04164R2, Verification of Rad Monitor Response to Core Uncovery. It is provided to assist technical reviewers that will be evaluating this license amendment request for MPS2.

Purpose:

The purpose of this calculation is calculate the radiological response of containment high range radiation monitors to core uncovery during refueling.

References:

1. M3EP-0414R3, Rev. 0, MP3 Rad Monitoring Response To Core Uncovery

2. ANS/SD-76/14, "Handbook of Radiation Shielding Data", July 1976

3. Cale. SFPGAMMA-04011 R2, Rev. 0, "MP2 Gamma Heating Analysis", dated 4/4/2003

4. MP2 Technical Specifications through Change# 318-01

5. Not Used

6. Not Used

7. Dwg 25203-29531, Rev.1, Millstone 2 Fuel Assembly

8. Dwg 25203-28014, Rev. 8, Instrument Location - Containment Plan El 14'-6" &38'-

6"

9. Dwg 25203-27022, Rev. 6, General Arrgt- Containment & Aux Bldg Section 8-8

10. Dwg 25203-29141, Sh. 100, Rev. 1, Unit 2 - Reactor Arrangement Sectional Elevation Layout 11 . SCALE 4.4a Code File Computer Code Used:

SCALE Package / QADS Key Inputs and Assumptions Used:

1. Water level at the top of the active fuel

2. Reactor head removed

3. Millstone 2 Core Inventory Method of Analysis:

EALs require the ability to measure gamma radiation from the reactor core while under conditions where the water level in the reactor vessel is at the top of the active core.

Considering the location of the core relative to RM-8240 and 8241, there does not appear to be any direct, line-of-sight communication of radiation. This requires crediting of air and concrete scattered radiations from the core to the rad monitors. Based upon a review of detector location (14'6" elevation on SG shield walls), and the need to credit multiple scattering there does not appear to sufficient radiation to register on these

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 48 of 54 detectors which have a low range limit of 1 R/hr. The insufficiency of radiation will be validated using RM-7891 which only requires a single scatter determination. RM-7891, the refuel floor rad monitor, is located adjacent to the refuel pool. No credit is taken for any radiation beyond single scattering.

The following scatter methods were validated in Reference 1. The methods are described in Reference 2 and applicable excerpts are retrievable as attachments to Reference 1.

Scattering by air is addressed by a method using a line beam response factor (LBRF) to evaluate the single scatter from direct radiation to a receptor located out of the direct beam. In the case of a reactor core, the gamma flux directly over the core at the elevation of the radiation monitors is determined using the QADS shielding code. This flux is converted to an equivalent line beam source by multiplying the flux by the horizontal surface area of the core. The line beam source is multiplied by the LBRF to get the dose rate at the rad monitor from air scattered radiation.

Scattering by the containment dome is addressed using a method that determines an appropriate albedo factor and appropriate dose rate response that reflects a single scatter from the containment structure above the core.

In this assessment, a lesser source term is more conservative to use (because it results in lower dose rates) and the twice burn, 5% enrichment provides a lesser source term than lower enrichments or single burns. The lesser source term results in a lower dose rate response ensuring that control room or SERO response to rad monitors used would be timely. In addition, since the core may be in a reduced inventory condition with containment closure not established, an assumption of 25 days decay will be assumed (this is roughly when draindown has occurred in order to put the reactor head back on the vessel).

This source term represents 1 fuel assembly decayed for 25 days using the ORIGENS code from the SCALE package (SQA Level 2). ORIGENS is a neutron depletion and decay code that will generate a source term that can be used in the QADS code with a multiplier to reach the equivalent of 217 assemblies. QADS is also from the SCALE package and is a point-kernel gamma shielding code. All of the fuel assemblies in a core will be modeled as a large cylindrical source. Receptor points will be located at the elevation of the rad monitors (for air scatter determination) and at the top of the containment dome (so that concrete/ steel scatter can be assessed).

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 49 of 54 Summary:

Dose rate at R7891 from a reactor vessel draindown condition to top of active core, with no reactor head in place, crediting scattered radiation and a 25 day decay time for fuel as determined by this calculation are:

R7891, R/hr Air Scatter 1.4 Dome Scatter 3.0 Total 4.4

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 50 of 54 Calculation Summary Post-Accident Letdown Radiation Monitor Response The following pertinent information has been extracted from North Anna Calculation PA-0234, Rev. 1, Post-Accident Letdown Radiation Monitor Response. It is provided to assist technical reviewers that will be evaluating this license amendment request.

Purpose:

This calculation provides the setpoint as a function of decay of the letdown line radiation monitor during accident condition for a 1% and 5% failed fuel (gap release) and iodine spike cases of 60 µCi/gm and 300 µCi/gm Dose Equivalent 1-131 released coolant activity in the RCS. During accident conditions, letdown radiation monitors support in determining the fuel failures that correspond to specific radiological criteria in the Emergency Action Levels (EALs).

References:

1. RF- DCP-000, 59-DCP-07-006, 11 Letdown Radiation Monitor Replacement/ North Anna I Unit 1 & 2 11 11

2. *Nuclear Energy Institute NEI 99-01, Rev. 4, Methodology for Development of Emergency Action Levels, 11 January 2003.

3. PA-DWG-000, 11715-FM-095A, Rev. 28, 11 FlowNalve Operating Numbers Diagram Chemical and Volume Control System", North Anna Power Station -

Unit 1 .

4. RF-DCP-000, 59-DCP-94-013, "Letdown Radiation Monitor Replacement - North Anna Unit 111 , September 1995

5. RF- DCP-000, 59-DCP-94-014, 11 Letdown Radiation Monitor Replacement - North Anna Unit 2 11 , March 1995.

11

6. RF - CALC-NFL, PA-0195, Rev. 0, Radiological Consequences of Fuel Handling Accident at North Anna Based on the Alternative Source Term", June 2003.

11

7. RF - CALC-NFL, PA-0219, Rev. 0, Post Accident Response Curves for North 11 Anna Primary Hot Leg Sample Lines , January 2005

8. RF - CALC-MEC, ME-0438, Rev. 2 "Reactor Coolant Letdown Radiation Monitor Setpoints", June 1995.

11

9. *PA-REF=OOO, NUREG-1228, Source Term Estimation during Incident 11 Response to Severe Nuclear Power Plant Accidents , McKenna, T. J. and Gutter, U .S. Nuclear Regulatory Commission, Washington, D .C, 1988

10. *PA-REF-0, REG. Gu-1 .109, 11 Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR PART 50, Appendix l, 11 Rev .. 1, U.S. Nuclear Regulatory Commission, Washington, D .C, 1977.

11. RF-CODE-000, MicroShield, Grove Software, Inc, Verification 7 .02.

12. RF - CALC-RAD, PA-0246, Rev. 1, "Letdown Radiation Monitor Setpoint for North Anna", Apri1 2008.

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 51 of 54 Computer Codes Used:

MicroShield Version 7.02 (Ref. 11)

Key Inputs and Assumptions Used:

1. Fuel rod gap fractions: 3% noble gas, 2% iodine, 5% cesium (Ref. 9)

2. Detector placed -0.5 inches from 2-inch pipe shielded with lead

3. North Anna Core Inventory Method of Analysis:

The 1% failed fuel is modeled in this calculation by calculating the dose rates resulting from 1% of the failed fuel gap inventory being released into the primary coolant that resulted following an accident. The 5% failed fuel was modeled by scaling up by a factor of 5 the results for the 1% failed fuel.

Dose rates to the monitor are calculated using MicroShield computer code. (Detector assembly is placed approximately 2 .5 inch (including O .5 inch of insulation of the pipe) from the center of a 2-inch diameter pipe source. The use of the MicroShield code is reasonable since the source is surrounded by lead shield and scattering is considered to be minimal. The MicroShield modeling assumes liquid source in a 1-inch radius pipe and 10 inches long with a thickness of O .154 inches. The dose point is assumed to be located at 2 .5 inch from the center (radial direction) of the source. The pipe and the detector are placed inside lead shield wall. It is assumed that the detector is located in the middle of the length of the pipe.

==

Conclusion:==

Tables 4 below provides summary of results of the letdown radiation monitor dose rates that can be used for the EAL radiological criteria.

The dose rate are calculated for the letdown line.

Table4 No11h Anna uldown Radiation Monitor Dose Dose Dose Dose Dose Timcihr.) iRlhr.) (R.lhr.) (RJhr.) CR/hr,)

!%FF 5%FF 60 nCi /r,m 30011Ci/,un 0.0 46.85 234.25 24.85

124.26 1.0 28.65 143.25 15.20 75.~

2.0 20.94 104.70 It.II S5.S4 4.0 14.20 71.00 7.S3 37.66 8.0 16.0 24.0 36.0 9.30 6.03 4.76 3.86 46.SO 30.15

"~fl 19.30 4.93 3.20 2.()5 24.67 15.99 1"62 10.24 48.0 3.41 *~ 9.04 72.0 2.97 14.85 1.58 7.SS

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 52 of 54 Calculation Summary Post-Accident Letdown Radiation Monitor Response The following pertinent information has been extracted from Surry Calculation PA-0236, Rev. 0., Add. A, Post-Accident Letdown Radiation Monitor Response for Surry. It is provided to assist technical reviewers that will be evaluating this license amendment request.

Purpose:

The purpose of this addendum is to determine the letdown radiation monitor (RM) response to source terms representative of gross reactor coolant activity and fuel failure that correspond to specific radiological criteria in the Emergency Action Levels (EALs) based on guidance from the Nuclear Energy Institute (NEI) for primary coolant activity level.

References:

1. NEI 99-01, Rev. 4, "Methodology for Development of Emergency Action Levels,"

January 2003.

2. NEI 99-01, Rev. 6, "Development of Emergency Action Levels for Non-Passive Reactors," November 2012.

3. Computer Code MicroShield, "MicroShield User's Manual", Grove Software, Inc.

4. MICROSHIELD-20170323-0-0, "MicroShield V. 7.02 Periodic Effectiveness Review 2017, Code Manager change and Code Owner change."

5. 958.398ABS Rev. A, "Calibration of a 903664 Letdown Monitor and 943-36 Detector, Victoreen, Inc., June 1996.

6. 11448/11548-7.57-148 Sheet 1, "Hi & Lo Range Letdown Monitor, June 1970 (Detector Shield Arrangement- Vendor Drawing 903664).

7. 11448/11548-7.57-26A Sheet 1, "Connections of Letdown Monitor, June 1970 (Sample line tubing details - Vendor Drawing 903742).

8. 11448/11548-7.57-268 Sheet 1, "Connections of Letdown Monitor, June 1970 (Sample line tubing details - Vendor Drawing 903742).

9. 903787, "[Aluminum] Spacer Hi-Lo", August 1970.

10.903751 Rev. X1, "[Lead] Plug", August 1970.

11. RA-0008 Rev. 0 through Addendum C, "Core Isotropic Inventories for Surry Dose Consequences Analyses Based on the Alternate Source Term", July 2010.

12. ETE-NAF-2017-0052 Rev. 1, "Input to a Proposed Surry License Amendment Request Adopting TSTF-490-A Rev. 0 (Dose Equivalent Xe-133) and Updated Alternative Source Term Analyses", January 2018.

13. NUREG-1228, "Source Term Estimation during Incident Response to Severe Nuclear Power Plant Accidents", McKenna, T. J. and Gutter, U.S. Nuclear Regulatory Commission, Washington, D.C, October 1988.

14. RA-0070 Rev. 0 through Addendum A, "Radiological Consequences of a Steam Generator Tube Rupture (SGTR) at Surry Power Station Based on the Alternative Source Term (AST)", May 2017.

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 53 of 54

15. Federal Guidance Report No. 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors For Inhalation, Submersion, and Ingestion", EPA 520/1-88-020, September 1988.

16. SQA-WATPROP-D-20170608, "SQA Documents for the Initial Release of WATPROP-D and Initial Periodic Effectiveness Review", June 2017.

Computer Codes Used:

MicroShield Version 7.02 [References 3 and 4]

Key Inputs and Assumptions Used:

1. Fuel rod gap fractions : 3% noble gas, 2% iodine, 5% cesium (Ref. 9)

2. Detector response from 0.5 inch sample tubing through lead shield

3. Surry Core Inventory Methodology:

Letdown radiation monitor response in counts per minute (cpm) was determined for accident source terms of 1% failed fuel, 10 µCi/cc DE 1-131, and 300 µCi/cc DE 1-131 at 0, 1, 2, 4, 8, 16, 24, 36, 48, and 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after shutdown. This was accomplished by using MicroShield to determine dose rate at the detector for the 1% failed source term at O hours. MicroShield was used to decay the 1% failed fuel source and determine the dose rates at the detector for each of the subsequent time steps. MicroShield dose rates were also determined for calibration standards of Co-60 and Mn-54 for which the vendor had determined cpm. Conversion factors for cpm to Dose rate for Co-60 and Mn-54 were derived and applied to the dose rate results for the 1% failed fuel source term at each time step.

==

Conclusion:==

The letdown radiation monitor response in counts per minute (cpm) to accident source terms of 1% failed fuel, 10 µCi/gm DE 1-131, and 300 µCi/gm DE 1-131 at 0, 0.5, 1, 2, 4,

Serial No.: 19-296 Docket Nos.: 50-336/423/338/339/280/281 72-2/16/47/55/56 Enclosure 7 Page 54 of 54 8, 16, 24, 36, 48, and 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after shutdown are documented in Table 4.

Table 4: Surry Letdown Radiation Monitor Response (cpm)

Time (hr.) 1%FF 10 µCi/gm 300 JtCi/gm 0.0 1.30E+07 1.18E+06 3.53E+07 0.5 5.98E+06 5.40E+05 1.62E+07 1.0 4.02E+06 3.62E+05 1.09E+07 2.0 2.33E+06 2.11E+05 6.32E+06 4.0 1.32E+06 1.19E+05 3.57E+06 8.0 7.05E+05 6.37E+o4 1.91E+06 16.0 3.08E+05 2.78E+04 8.35E+05 24.0 1.81E+05 1.64E+o4 4.91E+05 36.0 1.16E+05 1.04E+04 3.13E+05 48.0 9.34E+04 8.43E+o3 2.53E+05 72.0 7.94E+04 7.17E+o3 2.15E+051

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