ML17144A285

Summary Teleconference with Exelon Generation Company, LLC, on Planned Submittals of License Amendment Requests to Change Emergency Action Levels Associated with Lake Ontario Water Level

ML17144A285
Person / Time
Site:	Nine Mile Point, FitzPatrick
Issue date:	06/02/2017
From:	Marshall M Plant Licensing Branch 1
To:	Exelon Generation Co
Marshall M L/DORL/LPL1/415-2871
References
Download: ML17144A285 (47)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 June 2, 2017 LICENSEE: Exelon Generation Company, LLC FACILITY: Nine Mile Point Nuclear Station, Units 1 and 2 James A. FitzPatrick Nuclear Power Plant

SUMMARY

OF APRIL 25, 2017, MEETING WITH EXELON GENERATION COMPANY, LLC, ON PLANNED SUBMITTAL OF LICENSE AMENDMENT REQUEST TO CHANGE EMERGENCY ACTION LEVELS ASSOCIATED WITH LAKE ONTARIO WATER LEVEL On April 25, 2017, a Category 1 public meeting was held between the U.S. Nuclear Regulatory Commission (NRC) and representatives of Exelon Generation Company, LLC (Exelon, the licensee) by teleconference. The purpose of the meeting was for Exelon to discuss a license amendment request (LAR) that the licensee planned to submit to the NRC that would change emergency action levels (EALs) associated with Lake Ontario water level for the Nine Mile Point Nuclear Station, Units 1 and 2 (Nine Mile Point), and James A. FitzPatrick Nuclear Power Plant (FitzPatrick). The licensee planned to submit the LAR to the NRC by April 28, 2017. The meeting notice and agenda, dated April 24, 2017, are available in the Agencywide Documents Access and Management System (ADAMS) at Accession No. ML17114A409. A list of attendees is provided as Enclosure 1.

The licensee presented information using excerpt from the FitzPatrick Updated Final Safety Analysis Report (Enclosure 2), excerpt from an EAL bases document for FitzPatrick (Enclosure 3), and draft LAR for Nine Mile Point (Enclosure 4). Prior to presenting an overview of the technical basis for the planned LAR, the licensee provided background and context for the need to change the EAL in response to the increasing level of Lake Ontario. Both FitzPatrick and Nine Mile Point are located on the banks of Lake Ontario. The licensee presented two options for changing the EALs at FitzPatrick and Nine Mile Point.

According to the licensee, the International Joint Commission, which is the organization that licenses and regulates certain water resource projects along the United States and Canadian border, changed the water management strategy for Lake Ontario in December 2016. The change allows the lake level to be managed at a level above the unusual event condition for certain EALs at FitzPatrick and Nine Mile Point. In addition to the change in management strategy, the licensee noted that the water level in Lake Ontario has been impacted by an abnormally wet spring and heavy rains. The first option described by Exelon using FitzPatrick as an example would increase the high lake water level of the EAL involving Lake Ontario. This option would be permitted without prior NRC approval in accordance with Title 10 of the Code of Federal Regulations (10 CFR) Section 50.54(q)(3) if the change in water level of the EAL did not reduce the effectiveness of the site's emergency plan. The second option described by Exelon using Nine Mile Point as an example would seek NRC approval, on an exigent basis, of a license amendment to remove the high lake water level as an initiating condition for providing notification of an unusual event. Exelon wanted to change the EALs that contain high lake level

threshold to avoid entering and remaining in an unusual event for a prolonged period and to avoid contributing to existing public concerns attributed to the unexpected environmental conditions (i.e., record high lake levels).

During the meeting, the NRC staff acknowledged that Exelon could change the EALs for FitzPatrick and Nine Mile Point per 10 CFR 50.54(q)(3) if the conditions for making changes to the emergency plan without prior NRC approval are met. Regarding the second option (i.e.,

license amendment), the NRC staff informed the licensee that more substantive technical justification than the one provided during the meeting would need to be included in the LAR.

Specifically, the licensee would need to provide a more substantive description of how the change in the EAL would continue to demonstrate compliance with the requirements for emergency plans.

Members of the public were in attendance. At the end of the meeting, but prior to adjourning the meeting, a representative of the State of New York asked clarifying questions about the 10 CFR 50.54(q)(3) regulation. Public meeting feedback forms were not received.

No regulatory decisions were made during the meeting. At the time of the public teleconference, Exelon had not decided which of the two options it would pursue in changing the EALs. Subsequent to the meeting, Exelon verbally informed the NRC that it would implement the first option (i.e., increasing the high lake level threshold) for both FitzPatrick and Nine Mile Point by using the provisions of 10 CFR 50.54(q)(3).

Please direct any inquiries to me at 301-415-2871 or Michael.Marshall@nrc.gov.

~/~

Michael L. Marshall, Jr., Senior Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-220, 50-333, and 50-410

Enclosures:

1. List of Attendees

2. Excerpt from FitzPatrick Updated Final Safety Analysis Report

3. Excerpt from Emergency Action Level Bases Document for FitzPatrick

4. Draft License Amendment Request for Nine Mile Point cc w/encls: Distribution via Listserv

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• • • • • :.....:..*L Alvse Peterson State of New York Art Daniels Exelon Generation Comoanv, LLC Bill Drews Exelon Generation Comoanv, LLC Booma Venkataraman U.S. Nuclear Reaulatory Commission Bridaette Frymire State of New York Dave Burch Exelon Generation Comoanv, LLC Dave Gudaer Exelon Generation Comoanv, LLC Dennis Moore Exelon Generation Comoanv, LLC Don Johnson U.S. Nuclear Regulatorv Commission Doua Walker Exelon Generation Comoanv, LLC Jerry Collin State of New York Jim Jones Exelon Generation Comoanv, LLC John Metro State of New Yark Katv Yurkon Exelon Generation Comoanv, LLC Ken Kristensen Exelon Generation Comoanv, LLC Larry Baker Exelon Generation Comoanv, LLC Mark Hawes Exelon Generation Comoanv, LLC Michael Marshall U.S. Nuclear Reaulatory Commission Ray Hoffman U.S. Nuclear Reaulatory Commission Ron Reynolds Exelon Generation Comoanv, LLC Ryan Pritch Exelon Generation Comoanv, LLC Enclosure 1

Enclosure 2 Excerpt from FitzPatrick Updated Final Safety Analysis Report

JAF FSAR UPDATE In addition, a one-dimensional steady-state model was used to predict the maximum probable setup at the JAF site using the maximum probable wind storm as input.

Comparisons were made to Lake Erie, Lake Michigan, and Lake Ontario based on the results of the one-dimensional model.

Two-dimensional Time Dependent Model

a. Description of the Model During winter or during a large wind storm, Lake Ontario is in a barotropic or "homogeneous fluid" condition (Ref. 10). Therefore, the hydrodynamics of the lake can be represented by a two-dimensional barotropic mathematical model in a vertically integrated form. The complete momentum equations in two horizontal directions, including the pressure term vertically integrated, and the continuity equation, are solved for the water level fluctuation using an imposed time variable wind field as input on the surface of the lake. Effects of bottom friction, bottom topography and lateral boundary configuration are included in the model. The effect of the rotation of the earth is represented by a constant coriolis parameter.

The wind driven circulation patterns of the lake were first determined by the development of the vorticity equation from the governing equations (Refs. 11, 12).

The velocity field of the lake, the continuity equation, and the divergence of the vector form of the momentum equations were then used to obtain the water level fluctuation in the lake. A square grid of 5 km was used to form the basis for the circulation and setup model of Lake Ontario. Figure 2.4-1 shows the depth and rectangular step approximation of the outline of Lake Ontario. The important parameters which appear to have strong influence on the magnitude of the setup of the lake are wind stress, bottom stress, depth of water, and bottom configuration

b. Validation of the Model The wind storm used to validate the model occurred on January 25, 1972. The storm generated a considerable surge in the eastern end of Lake Ontario. The storm was selected on the basis of easy access to the wind data and water level records.

The lowest pressure of the storm was located about 300 miles north of Lake Ontario at 7 a.m., January 25, 1972, with a minimum central pressure of 982 mb. The center of the low moved toward the northeast with a speed of approximately 40 miles per hour. During the period of high wind speed, the direction of wind over the lake was persistently toward east, which coincides with the long axis of Lake Ontario. The wind field over the lake was compiled from ten weather stations located along the shorelines of the lake in the United States and Canada. At the eastern end of the lake, wind speeds as high as 52 miles per hour for about three hours were recorded at Kingston, Ontario. A complete spacial time dependent wind field over the lake for the period of the storm was used to compute the wind stress at each grid point on the lake surface. A typical wind speed and direction at the JAF site, grid index of the mathematical model - (55, 5), is shown in Figure 2.4-2 The water level fluctuation over the entire extent of Lake Ontario was generated by the mathematical model using the prescribed wind field as input data. Figure 2.4-3 depicts the predicted and observed water levels at Toronto, Rochester, and Oswego, which represent the western, middle, and eastern sections of Lake Ontario, respectively. It can be seen that at all three locations the predicted and observed water levels compare reasonably well.

Since the wind was blowing toward the east, the water level at the western end of the lake was depressed while the eastern end of the lake was higher and the 2.4-3 Rev. 3 4/15

JAF FSAR UPDATE mid-section of the lake experienced minimal fluctuations. Figure 2.4-4 shows the water level variation over the entire lake at the hour of maximum setup at the eastern end.

Prediction of the Maximum and Minimum Probable Setup - Two-dimensional Time Dependent Model The maximum probable wind storm which will occur in the lake area was based on a modification of the storm of January 25, 1972. The duration and wind directions of the January 25, 1972 storm were retained; however, the wind speeds were increased by a statistical technique developed by Gumbel (Ref. 13) to reflect a storm with a recurrence interval of 10,000 years. The resulting storm, Figure 2.4-5, has a maximum sustained wind speed of 88 mph, lasting for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. The mathematical model was used to simulate the circulation and water level fluctuations of Lake Ontario under the influence of the probable maximum wind storm. Figure 2.4-6 is the result of the numerical simulation showing the water level variations on the entire lake at the hour of maximum setup at the eastern end. A vertical profile across the axis of the lake is also shown in Figure 2.4-6.

Due to the combination of on-shore and along-shore currents resulting from the bottom and boundary configurations, the setup along the shore is relatively higher than that at some distance off-shore. At JAF site, the maximum setup was calculated to be 4.1 feet. With west wind the maximum water level of Lake Ontario occurs at the east shore and the minimum water level occurs at the west shore. The maximum instantaneous water variations occurred at both the west shore and the east shore with a magnitude of 4.6 ft.

The probable maximum decrease in lake level at JAF site when the wind blows from the opposite direction was assumed to be the same as the probable maximum setup as described above, namely 4.1 ft. Since the fetch and flow resistance in the lake would be the same for both east wind or west wind, the nodal point or the pivot point of the lake water surface would also remain essentially unchanged.

The hydrograph of water level versus time near JAF site during the storm period is shown in Figure 2.4-7. According to the prediction, the maximum setup of 4.1 ft above mean water level will last for about one hour.

Steady State - One-dimensional Model The steady state one-dimensional model was used to calculate the maximum setup for comparison.

The stress coefficient of the one-dimensional model was adjusted by using the observed wind and water level data for the January 25, 1972 storm. Water levels over Lake Ontario were then calculated using the maximum probable windstorm described in the above paragraph as input to the adjusted one-dimensional model. The water level profile along the long axis of the lake is shown in Figure 2.4-8. The maximum setup at JAF site was calculated to be 3.5 ft and that at the eastern end of the lake was calculated to be 4.5 ft. These values compare reasonably well with those calculated by the two-dimensional model.

Comparison of Setup in Lakes Ontario, Michigan, and Erie - One-dimensional Model Since the one-dimensional model can estimate setup of Lake Ontario reasonably well as compared to the observed setup and because it is a simple model, it was used to calculate the setup values on Lake Michigan and Erie for the purpose of comparing the setup values of these three lakes.

Setup values for each lake are calculated using the one-dimensional model with the same wind speed and stress coefficient. These values are shown in Table 2.4-1. The greater setup (and setdown) values calculated for Lake Erie and Michigan, reflect the major influences of the shallower depth of Lake Erie and longer fetch of Lake Michigan, as compared to Lake Ontario.

Since the magnitude of wind surges depends primarily on the wind speed, the fetch length over which the wind blows, and the depth of the lake, it is possible to make an estimate of the magnitude 2.4-4 Rev. 3 4/15

JAF FSAR UPDATE of wind setups on the Great Lakes for a given wind speed with the wind direction along the longest fetch. Such an estimate serves as an index for relative magnitude of wind driven surges in the lakes.

Table 2.4-2 shows the mean depth, fetch, and the relative setup index of the Great Lakes. The relative setup index is the relative magnitudes of wind surge among the five Great Lakes. Assuming maximum wind setup on Lake Erie as 100 percent, the percentages in the index column show the relative maximum setups on the other lakes for any given wind speed. For example, if for a given wind speed the maximum setup on Lake Erie were 10.0 ft, for the same wind speed, the maximum setup on Lake Ontario would be 18 percent of 10.0 ft, or 1.8 ft. The values shown on Table 2.4-2 agree well with the results of the one- dimensional model as indicated on Table 2.4-1.

2.4.3.4 Computing Maximum Wave Runup The maximum probable wind storm was used in conjunction with deep water wave forecasting curves of the Bretschneider-revised Sverdrup-Munk method (Ref. 15) to determine the "significant" deep water wave height and period. The significant deep water wave produced by the maximum probable storm had a height of 35 ft (trough to crest) and a period of 13.5 sec.

The portion of the storm with average hourly winds greater than 50 mph was taken to be the critical portion of the storm for wave formation. The duration of winds 50 mph and greater is 23 hrs. Using the significant wave period of 13.5 sec, approximately 6, 100 consecutive waves reached the shore during the critical portion of the storm. The probability that a single wave with the highest combination of height and period will occur during this period is 1/6100 or 0.16 per 1,000 consecutive waves. Using the data from Bretschneider (Ref. 15) as to the joint distribution of height and period per 1,000 consecutive waves, this probability corresponds to a deep water wave 50 ft in height with a period of 20 sec.

Standard orthogonal refraction techniques were used to determine the variation in wave height as the waves approached the shoreline. All waves were assumed to break when the ratio of wave height to water depth was 0.78. This criterion is derived from solitary wave theory and is consistent with data on the breaking of oscillatory waves (Ref. 16).

The magnitude of the wave runup was determined according to the composite slope method of Saville (Ref. 17). This technique, based on model test data, considers the following parameters:

a. The location and water depth at which the wave breaks

b. The wave period and breaking height

c. The lake bottom topography

d. The shore topography.

A maximum wave runup of 7.5 ft was estimated.

2.4.3.5 Monthly Mean High Water Level of Record The all-time monthly mean high water level (el. 249.3) occurred in June, 1952 when Lake Ontario was unregulated and subject to changes in levels produced by natural inflows and outflows.

Regulation of Lake Ontario was implemented in April, 1960 at the direction of the International Joint Commission (IJC). One of the primary objectives of the plan of regulation is that the monthly mean level of Lake Ontario shall not exceed el. 248. The agreement states that lake level is regulated within a range of elevations from "244 feet (navigation season) to 248 feet as nearly as may be ... "

The regulation of Lake Ontario is carried out under the general supervision of the International St.

Lawrence River Board of Control, composed of ten members representing Canada and the United States. The Board has appointed regulation representatives who oversee the day-to-day operations conducted under the regulation plan. In addition, the Board formed a five member operations advisory group whose members include NYPA, Ontario Hydro, Hydro Quebec, the St. Lawrence Seaway Development Corporation, and the Canadian Coast Guard. This group, together with the 2.4-5 Rev. 3 4/15

JAF FSAR UPDATE regulation representatives, meets at weekly intervals to consider Lake Ontario and St. Lawrence River conditions as they affect regulation. Regulation is physically implemented through the control of the flows through the St. Lawrence Project. To this end, Ontario Hydro and the Authority receive advice each Thursday as to the flow to be released during the following week which meets the IJC's objectives insofar as Lake Ontario regulation is concerned.

In view of the strict surveillance exercised by the IJC, through the St. Lawrence River Board of Control and the Board's operations advisory group, with respect to the levels of Lake Ontario, lake level will not be regulated in excess of el. 248 through operator error or unilateral action on the part of the Authority or Ontario Hydro. Furthermore, there is no likelihood that the International Joint Commission will allow the level of Lake Ontario to be regulated in excess of el. 248 due to the adverse effect of such action on development along the lake shorefront.

The plan of regulation is based upon historic norms for the period 1860 to 1954. Since regulation began, transient conditions in excess of historic norms have caused monthly mean lake level to exceed el. 248 on a few separate occasions, each of three to four months duration. In each case, levels were lower than would have been experienced under natural control. The highest monthly mean lake level experienced since regulation began was el. 249.1 in May 1973.

The original FSAR concluded that the all-time monthly mean high water level, or any other level above el. 248 would be experienced only if regulation were suspended and Lake Ontario were to revert to natural control; and that it was proper to assume that the maximum storm occurs when the lake stage is at el. 248, the maximum controlled still water level. Since it has been shown that input in excess of historic norms can occur and can cause lake level to exceed el. 248 on rare occasions, an additional evaluation was performed (Ref. 40). This evaluation concluded that no condition adverse to safety exists with a maximum lake level at or below el. 250.

2.4.3.6 Design Minimum Low Water The design minimum low water level of Lake Ontario for the FitzPatrick Plant is el. 236.5. This elevation is based on superposition of the following effects:

1. Minimum still water level of Lake Ontario el. 240.6

2. Instantaneous lowering of the still water level due 4.1 ft to the maximum probable seiche on Lake Ontario It should be noted that the actual minimum still water levels of Lake Ontario, observed during a period of record beginning in 1860 and continuing through the present time, were:

1. Lowest monthly mean water surface level recorded prior to construction of the St. Lawrence Power Project (November, 1934) el. 242.7

2. Lowest mean water surface level for a quarter-month recorded prior to construction of the St. Lawrence Power Project (third quarter of December, 1934) el. 242.6

3. Lowest monthly mean water surface level recorded subsequent to the commencement of regulation of Lake Ontario by the St. Lawrence Power Project (January, 1965) el. 243.0 The Effect of Failure of the St. Lawrence Power Project on Low Lake Levels The St. Lawrence Power Project constructed jointly by the Authority and Ontario Hydro consists of two dams and a hydroelectric power plant in the St. Lawrence River. Iroquois Dam is a gated gravity-type structure located 78 miles below the outlet of the lake. Long Sault Dam, a second gated 2.4-6 Rev. 3 4/15

JAF FSAR UPDATE gravity type structure, is located 102 miles below the outlet of the lake and serves as the spillway for the power dam. The central element of the St. Lawrence Power Project is the Robert Moses-Robert H. Saunders Power Dam, a massive gravity-type structure located 106 miles below the outlet of the lake. Iroquois Dam and the Moses-Saunders Power Dam ordinarily pass the full flow of the St.

Lawrence River, less relatively small diversions, including the quantity of water needed to operate the locks of the St. Lawrence Seaway which, in the aggregate, amount to much less than 1 percent of the river's total flow. Long Sault Dam is operated infrequently and customarily discharges no water at all.

In the course of normal operation, Lake Ontario is regulated at the Power Dam pursuant to a plan of regulation approved by the IJC. Actual operation has demonstrated that the lake also can be regulated by manipulating the gates at Iroquois Dam. The Power Project, therefore, provides a redundant safeguard against any loss of capability to regulate Lake Ontario.

The failure of any part of the St. Lawrence Power Project is beyond reasonable expectations. The three gravity dams are massive structures which were conservatively designed to withstand all static and dynamic forces by wide margins. For example, the design criteria applicable to these dams provided factors of safety of at least 2.0 against overturning and at least 3.0 against sliding.

Gravity structures were designed to resist seismic forces equivalent to 0.05 gin both the horizontal and vertical directions, while the seismic design of slender structures was based on dynamic analysis. All phases of construction which had any bearing on the ultimate safety of any structure were continuously inspected, during performance of the work, by qualified engineers and subjected to rigorous quality control requirements.

Since completion of the project, all structures have been under continuous surveillance by the Authority and Ontario Hydro. Both Ontario Hydro and the Authority conduct rigorous programs of inspection and preventive maintenance with respect to all elements of the project under their jurisdiction. Once every five years, or more often if necessary, a complete safety inspection of the project is performed by independent consultants who report to the Federal Power Commission with respect to the condition of dams and other structures. The integrity of the dams which comprise the St. Lawrence Power Project and regulate the levels of Lake Ontario is, therefore, ensured by the combination of conservative design, rigorous quality control during construction. and continuous surveillance following completion of the project.

In November, 1968, the St. Lawrence Study Office at the Canadian Department of Energy, Mines and Resources analyzed possible upstream and downstream effects resulting from failure of the St.

Lawrence Power Project structures. Under the adverse assumptions of this study, which postulated the sudden destruction of the above mentioned dams and the lowest supply sequence on record, it was determined that the lake level would decline gradually and, approximately one year following the assumed failure, be no more than 2.1 ft below the lowest level attained during regulation, i.e.,

the lake level would decline from el. 242.7 to el. 240.6. The study concluded that once the lake level had declined to about el. 240.6, natural controls, such as existed before the project. would be re-established and the lake levels would rise and fall thereafter in accordance with natural supplies delivered to Lake Ontario from the Great Lakes watershed.

Lake Ontario is the source of the St. Lawrence River and the last in the chain of five Great Lakes.

These lakes drain an area of approximately 300,000 sq miles in the United States and Canada.

Taken together, they makeup the largest body of fresh water in the world. Each lake acts as an enormous natural regulating reservoir which smooths out variations in inflow and tends to equalize outflows from season to season.

Lake Ontario obtains its principal supply of water from the Niagara River which drains the four upper lakes (Erie, Huron, Michigan and Superior) and discharges 200,000 cu ft per sec on an annual average basis. Other inflows are received from precipitation and springs within the Lake Ontario watershed.

2.4-7 Rev. 3 4/15

JAF FSAR UPDATE The outflow from Lake Ontario, which corresponds to the flow in the St. Lawrence River, is equal to the quantity of water supplied from the Niagara River and all other sources, less the change in the quantity of water stored in the lake. Although storage in Lake Ontario is now governed by artificial controls, the lake was self-regulating in its natural state.

Flow in the St. Lawrence River is characterized by its extreme regularity. The maximum flow (350,000 cu ft per sec) is about twice the minimum flow (170,000 cu ft per sec). This condition existed before construction of the St. Lawrence Power Project, and would continue today if the project did not exist. In the improbable event of the simultaneous failure of the Iroquois Dam and the Moses-Saunders Power Dam or Long Sault Dam, the actual level to which Lake Ontario could fall would be governed by supplies of water to the lake during the period following such failure and the natural resistance of the lakes to sudden changes in levels and flows. These effects would guarantee that the actual minimum still water level of the lake would be well above the design minimum water level elevation of the FitzPatrick Plant during any period necessary to re-establish control of the lake.

It is concluded that the simultaneous failure of the dams which regulate the levels of Lake Ontario is beyond any reasonable probability of occurrence. In the event of any such catastrophe, the levels of Lake Ontario would decline gradually. The full effect should be experienced about a year following the failure, by which time the still water level might fall to a minimum at el. 240.6. Superposition of the maximum probable seiche would produce a further lowering of 4.1 ft. to el. 236.5 over a short term.

2.4.3.7 Implication of the Maximum and Minimum Lake Levels on Power Station The cross-sectional profiles through the intake and discharge screen houses and tunnels are shown in Figure 2.4-9. The average ground elevation outside the screenhouse is 272.0 ft. Concerning the flooding of the exterior access of the power plant, the maximum wave run up of 7 .5 ft, the maximum wind setup of 4.1 ft, and the maximum rainfall of 0.35 ft were added to the maximum controlled still water level of 248 ft resulting in a maximum probable flood level of el. 260 at the JAF site. The grade elevation of the power plant, 272.0 ft, is well above the probable coincident maximum flood level of 260 ft at the power plant site with a freeboard of 12 ft. Consideration of a maximum lake level of el.

250 would still result in approximately 1Oft. of freeboard.

The original design basis maximum flood level in the screenhouse was determined as el. 252.5.

Since the intake screenhouse top deck ceiling is at el. 253.0, which is 0.5 ft above maximum probable flood level in the screenhouse, no damage from flooding would be expected. The revised design basis flood level of el. 255 coincides with the floor level in the screenwell. Any uplift forces on the floor slab resulting from this higher floor level are more than offset by the weight of the slab (Ref.

40). Again. no damage from flooding is expected. All seismic Class I equipment in the screenwell area is mounted at or above el. 255.

The effect of waves in the lake on the performance of the circulating water system is negligible and no loss of cooling water resulting from surging in the intake conduit could occur.

A mathematical model of the circulating water system was developed. The model uses the equation of continuity and momentum and solves for the flow system and the pumps characteristics simultaneously.

The maximum wave height that can exist without breaking at the location of the intake with a maximum probable lake elevation of el. 252.5 is 22 ft. Different wave periods of 8 sec, 13.9 sec, and 27.8 sec, representing a range of possible occurrences, were considered. The variations in the circulating water system due to the action of the above mentioned waves under the evaluated conditions is summarized below:

2.4-8 Rev. 3 4/15

JAF FSAR UPDATE Screenwell Pump Flow Intake Tunnel Water Level Wave Hgt Wave Period Fluctuation Flow Fluctuation Fluctuation ft sec cfs cfs inch 22 8 -1.0 +/-85.0 +0.1,-0.4 22 13.9 -2.0 +/-148.0 +0.5,-1.2 22 27.8 -6.0 +/-300.0 +2.3,-5.4 Figure 2.4-10 shows the effects of a 22 ft wave having a period of 13.9 sec., which represents the most probable occurrence of a 22 ft wave over the lake structure.

It can be seen from Figure 2.4-10 that although the circulating water system is subjected to a continuous train of 22 ft high waves over the intake structure, the screenwell water level experiences only about two inches of fluctuation and the variation in the circulating water pump flow is negligible. Because of the energy damping effects of the long intake and discharge tunnels, the resonance period of the circulating water system is much longer than the periods of waves acting over the intake and discharge structures.

2.4-9 Rev. 3 4/15

Enclosure 3 Excerpt from Emergency Action Level Bases Document for FitzPatrick

CLASSIFICATION OF EMERGENCY CONDITIONS IAP-2 ATTACHMENT 3 - EAL BASES Category: H - Hazards Subcategory: 1 - Natural & Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the Protected Area EAL:

HUl.5 Unusual Event Lake water level > 248 ft OR ESW intake bay water level < 237 ft Mode Applicability:

ALL NEI 99-01 Basis:

This EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant operators.

This EAL addresses other site specific phenomena (such as hurricane, flood, or seiche) that can also be precursors of more serious events.

JAFNPP Basis:

The high lake level is based upon the maximum attainable controlled lake water level as specified in the FSAR (ref. 1). The low level is based on ESW intake bay water level and corresponds to the design minimum lake level (ref. 2).

JAFNPP Basis Reference(s):

1. FSAR Section 2. 4. 3

2. Safety Evaluation JAF-SE-93-034 "Evaluation of Maximum and Minimum Water Levels at Screenwell for Safe Operation of Class I Equipment" Rev. No. 34 Page 134 of 325

Enclosure 4 Draft License Amendment Request for Nine Mile Point

10 CFR 50.90 10 CFR 50.54(q)

NMP1L3153 April XX, 2017 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Nine Mile Point Nuclear Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-63 and NPF-69 NRC Docket Nos. 50-220. 50-410. and 72-1036

SUBJECT:

License Amendment Request- Change Emergency Action Level HU1 .5 to Remove High Lake Level Initiating Condition for Unusual Event Emergency Classification In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," and pursuant to 10 CFR 50.54(q), "Emergency Plans," Exelon Generation Company, LLC (Exelon) is requests amendments to the licenses for the facilities listed above.

Specifically, the proposed changes involve revising the Emergency Plans for the affected facilities to revised Emergency Action Level (EAL) HU1 .5 to remove the high lake level as an initiating condition for entering a Notice of Unusual Event (UE) emergency classification.

Exelon has concluded that the proposed changes present no significant hazards consideration under the standards set forth in 10 CFR 50.92, "Issuance of amendments."

The proposed changes have been reviewed by the Plant Operations Review Committee in accordance with the requirements of the Exelon Quality Assurance Program.

This LAR contains no regulatory commitments. provides the evaluation of the proposed changes. Attachment 2 provides a copy of the markup of the proposed Emergency Action Level Matrices pages. Attachment 3 provides a copy of the clean proposed Emergency Action Level Matrices pages provides a copy of the marked up EAL Technical Bases pages that reflect the proposed changes and Attachment 5 provides the clean EAL Technical Bases pages.

License Amendment Request Revise EAL HU1 .5 to Remove High Lake Level NRC Docket Nos. 50-220 and 50-41 O April XX, 2017 Page 2 Exelon requests approval of this LAR by May 24, 2017, to support the predicted rise in lake level as described in Attachment 1.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"

paragraph (b), Exelon is notifying the State of New York of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official.

If you have any questions or require additional information, please contact Ron Reynolds at (610) 765-5247.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the XXth day of April 2017.

Respectfully, Jim Barstow Director, Licensing & Regulatory Affairs Exelon Generation Company, LLC Attachments: 1. Evaluation of Proposed Changes

2. Markup of Proposed Emergency Action Level Matrices Pages

3. Clean Proposed Emergency Action Level Matrices Pages

4. Markup of Proposed Emergency Action Level Bases Technical Pages

5. Clean Emergency Action Level Bases Technical Pages cc: Regional Administrator - NRC Region I w/ attachments NRC Senior Resident Inspector - NMP NRC Project Manager, NRR - NMP A.L. Peterson, NYSERDA

License Amendment Request Revise EAL HU1 .5 to Remove High Lake Level NRC Docket Nos. 50-220 and 50-410 April XX, 2017 Page 3 bee: Senior Vice President - Mid-Atlantic Operations wlo attachments Site Vice President - NMP Plant Manager - NMP Director, Operations - NMP "

Director, Site Engineering - NMP Director, Site Training - NMP "

Manager, Regulatory Assurance - NMP wl attachments Manager, Licensing, KSA R. Reynolds, KSA Commitment Coordinator - KSA Correspondence Control Desk - KSA

ATTACHMENT 1 License Amendment Request EVALUATION OF PROPOSED CHANGES

Subject:

License Amendment Request - Change Emergency Action Level HU1 .5 to Remove High Lake Level Initiating Condition for Unusual Event Emergency Classification 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration 4.4 Conclusions

6.0 REFERENCES

License Amendment Request Attachment 1 Revise EAL HU1 .5 to Remove High Lake Level Page 1 of 11 NRG Docket Nos. 50-220 and 50-41 O Evaluation of Proposed Changes 1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," and pursuant to 10 CFR 50.54(q), "Emergency Plans,"

Exelon Generation Company, LLC (Exelon) is requesting approval for proposed changes to the Emergency Action Level (EAL) of Renewed Facility Operating License Nos. DPR-63 and NPF-69 for Nine Mile Point Nuclear Station, Unit 1 (NMP1) and Unit 2 (NMP2).

The proposed changes involve revising the Emergency Action Level (EAL) HU1 .5 for NMP1 and NMP2 by removing the Notification of Unusual Event (UE) emergency classification threshold for high lake water level. This change is necessary because of recent changes in water management strategy and of natural causes on Lake Ontario.

Exelon requests approval of this license amendment request by May 24, 2017, as a result of the change in water management strategy for the lake and the predicted rise in lake level as a result of natural causes.

2.0 DETAILED DESCRIPTION The Boundary Waters Treaty of 1909 (BWT) established the International Joint Commission (IJC) as a cornerstone of United States - Canada relations in the boundary region. Under the BWT, the IJC licenses and regulates certain water resource projects along the border that affect levels and flows on the other side. The IJC also alerts the governments to emerging issues that might have negative impacts on the quality or quantity of boundary waters.

The BWT was established in 1909 and later amended in 1987 and 2012. The 2012 amendment was implemented on December 8, 2016, via a Supplementary Order of Approval. The Order required the regulated monthly mean level of Lake Ontario not to exceed monthly values which have occurred between 1900 and 2005. Trigger levels were established for actions to prevent lake level from getting too high or too low. The high trigger levels for April (248.03 ft.), May (248.46 ft.), June (248.33 ft.), and July (248.13 ft.),

exceed 248.2 ft. The International Lake Ontario - St. Lawrence River Board ensures the provisions of the Order relating to water levels are adhered to.

Since there was a change in water management strategy of Lake Ontario instituted by a Supplemental Order which was implemented in December 2016 under the provisions of the BWT between the U.S. and Canada. The Order required the regulated monthly mean level of Lake Ontario not to exceed monthly values and action levels were established to prevent lake level from getting too high or too low. Under the current water management strategy mandated by the Order, certain action levels will allow the lake level to be managed above the threshold for the UE condition. As result, Exelon is requesting the elimination of high lake water level as an initiating condition for a UE emergency classification. As of April 18, 2017, the current level of Lake Ontario is approximately 247.3 ft. The U.S. Army Corp of Engineers projects a rise in lake level by another 6-inches by May 14, 2017, based on forecasted level management by the St. Lawrence dam controls. In addition, it is an abnormally wet spring season and recent heavy rain is also impacting water level in Lake Ontario.

License Amendment Request Attachment 1 Revise EAL HU 1.5 to Remove High Lake Level Page 2of11 NRC Docket Nos. 50-220 and 50-410 Evaluation of Proposed Changes Further projected increases may exceed the high water level defined in the NMP1 Updated Final Safety Analysis Report (UFSAR) and NMP2 Updated Safety Analysis Report (USAR). The UFSAR and USAR high lake level is an initiating condition for entering EAL HU1 .5. Once the threshold is exceeded, the plants are required to declare a UE emergency classification in accordance with HU1 .5. Therefore, the plants will enter a UE for a prolonged period of time which will compound existing public concerns caused by the unexpected environmental conditions being experienced in the area. This UE will remain in place until the lake water level recedes below the action level limits that would require entry into the UE classification. As noted in Section 3.0, the water levels are managed on a month-by-month bases from April through July. These levels, during this period, may exceed the high water level initiating condition for entering the UE. Therefore, if entered, the UE could last through June based on the trigger levels described in Section 3.0.

Proposed Emergency Action Level Changes:

The proposed change to the NMP1 and NMP2 EALs (EP-AA-1013, Addendums 3 and 4),

is to delete the Notice of UE threshold for high lake water level as provided in Recognition Category H - Hazards and Other Conditions Affecting Plant Safety, EAL HU 1.5. See proposed changes to the EAL and EAL Basis shown below.

NMP1 EP-AA-1013, Addendum 3 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 1 - Natural or Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the PROTECTED AREA EAL: HU1 .5 Unusual Event Lake water level> 248.2 ft QR Intake water level < 238. 8 ft Plant-Specific This threshold addresses higf:J aRd low bay water level conditions that could be a precursor of more serious events (ref 1, 2).

The htgh lake lev-el is based 1:1poR the maxtm1:1m attaiRable 1:1RooRtFOJ!ed lake 111ater ie'l-OI as speoified iR the N.'>AP 2 USAR. Dams OR the St. La11mmoe Rtv-er, 1:1Rder the a1:1thor.ity of the IRtematioRa! St. Lawr:eRoe Rt~'er Board of CoRtfOI, are R01111:1sed to rog1:1late the Jake

.'-0~'0!. The low limit is set for el 74.37 m (244 #) OR Apri.' 1 aRd ts matRtaiRed at or abo*,'()

that e!evatioR d1:1riRg the eRttro RavtgatioR seasoR (April 1 to f\'-e*.'ember 3Q). The 1:1pper limit of the Jake !eve! ts el 75. 59 m (248. 2 #) (ref 3).

License Amendment Request Attachment 1 Revise EAL HU1 .5 to Remove High Lake Level Page 3of11 NRC Docket Nos. 50-220 and 50-410 Evaluation of Proposed Changes The low level is based on intake forebay level and corresponds to the minimum intake water level for operability of Emergency Service Water, Emergency Diesel Generator cooling water, Containment Spray Raw Water and Diesel and Electric FIRE Pump (ref. 4-9).

During planned evolutions such as intake water gate manipulation for reverse flow operations in which continuous monitoring of the intake level is being accomplished, entry into this EAL would not be warranted unless UNPLANNED /unexpected conditions and/or indications occur.

Generic This EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant operators.

This EAL addresses other site specific phenomena that can also be precursors of more serious events.

NMP2, EP-AA-1013, Addendum 4 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 1 - Natural or Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the PROTECTED AREA EAL:

HU1.5 Unusual Event Lake v1ator !e*1-0! > 248. 2 ft OR Intake water level < 237 ft Mode Applicability:

All Basis:

Plant-Specific This threshold addresses high aRd low lake water level conditions that could be a precursor of more serious events.

Tho high lako !o~'el is based l:JPOR tho maximum attaiRablo URGORtro!lod lako water level as specified iR tho USAR. Dams OR the St. LawroRGO River, uRder tho authority of tho

!RtornatioRal St. Lav/FORGO Ri~'er Board of CoRtro!, aro Row usoEf to regulate tho !a.'<-e level. The !ew limit is set for el 74.37 m (244 ft) OR AprN 1 aREf is maiRtaiReEf at or abo'l-0 that ele*1-atioR EfuriRg the eRtiro RavigatioR seasoR (April 1 to November 30). The upper limit of tho lake Jovel is e! 7a.a9 m (248.2 ft) (r:ef. 1).

License Amendment Request Attachment 1 Revise EAL HU1 .5 to Remove High Lake Level Page 4of11 NRG Docket Nos. 50-220 and 50-41 O Evaluation of Proposed Changes The low level is based on intake water level and corresponds to the design minimum lake level. The probable minimum low water level of Lake Ontario at the site has been determined to be 72.0 m (236.3 ft.) resulting from a setdown caused by a Probable Maximum Wind Storm concurrent with the lowest probable lake level. (ref 2)

Generic This EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant operators.

This EAL addresses other site specific phenomena that can also be precursors of more serious events.

3.0 TECHNICAL EVALUATION

Background

Lake Ontario, the easternmost of the Great Lakes, is an international body of water forming part of the border between the United States and Canada. The lake is 193 mi long and 53 mi wide at its largest points, and has a surface area of 7,340 sq mi. It has a maximum depth of 802 ft., an average depth of approximately 283 ft., and a volume of 393 cu mi.

The International Lake Ontario - St. Lawrence River Board uses the Moses-Saunders dam on the St. Lawrence River to regulate the lake level. The low limit is set for elevation 244 ft., on April 1 and is maintained at or above that elevation during the entire navigation season (April 1 to November 30).

The BWT established the International Joint Commission (IJC) as a cornerstone of United States - Canada relations in the boundary region. Under the BWT, the IJC licenses and regulates certain water resource projects along the border that affect levels and flows on the other side. The IJC also alerts the governments to emerging issues that might have negative impacts on the quality or quantity of boundary waters.

The 2012 amendment which was implemented on December 8, 2016, via a Supplementary Order of Approval, required the regulated monthly mean level of Lake Ontario not to exceed monthly values which have occurred between 1900 and 2005.

Trigger levels were established for actions to prevent lake level from getting too high or too low. The high trigger levels for May (248.46 ft.) and June (248.33 ft.) exceed 248.2 ft.,

which is the existing threshold The International Lake Ontario - St. Lawrence River Board ensures the provisions of the Supplemental Order relating to water levels are adhered to.

Water surface setup and seiche are produced by winds and atmospheric pressure gradients. These short-term lake fluctuations are generally less than 2 ft., in amplitude.

Winds are directly related to the formation of surface waves, the magnitude of which varies between 0 and 15 ft., in height during a given year. Tide magnitudes amount to less than 1-in.

License Amendment Request Attachment 1 Revise EAL HU1 .5 to Remove High Lake Level Page 5of11 NRC Docket Nos. 50-220 and 50-410 Evaluation of Proposed Changes Lake Ontario provides a heat sink for processing and operating heat from safety related components during a Design Basis Accident (OBA) or transient, as well as during normal operation. Thus, two of the principal functions of Lake Ontario at NMP1 and NMP2 are the dissipation of residual heat after reactor shut down and dissipation of residual heat after an accident.

Flood Protection All safety-related facilities except the intake structure are protected from flooding by a revetment ditch system. The system is constructed along the lakeshore in front of NMP2.

The top of the revetment is at an elevation of 263 ft., and prevents possible plant flooding due to lake level wave action. All safety-related facilities, systems, and equipment are protected against flood damage resulting from the probable maximum storm and historical maximum lake level, historical maximum precipitation and probable maximum lake level and surge with wind-wave action from Probable Maximum Wind Surge (PMWS).

All personnel entrances to Category I structures are at elevation 261 ft., or higher. The revetment ditch system was approved by the NRC in a December 1977 letter.

Lake Level Determination Lake level determination at the NMP site is performed by accessing a website maintained by the National Oceanic and Atmospheric Administration (NOAA). Current lake level in Oswego, NY is calculated by adding the "Observed Height" to 243.3 ft., (International Great Lakes Datum of 1985 (IGLD85)). Note that IGLD85 is a reference elevation system used to define water levels within the Great Lakes-St. Lawrence River system.

Emergency Action Level Revision NEI 99-01, Revision 5 Currently, the EAL scheme used at NMP1 and NMP2 are based on the guidance provided in NEI 99-01, Revision 5. NEI 99-01, Revision 5, EAL HU1, Natural or destructive phenomena affecting the PROTECTED AREA includes guidance for establishing a threshold (#5) based on "(Site specific occurrences affecting the PROTECTED AREA)."

The basis for threshold #5 states:

This EAL addresses other site specific phenomena (such as hurricane, flood, or seiche) that can also be precursors of more serious events.

This is consistent with the definition of an UE found in NEI 99-01, Revision 5, which states:

Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring off-site response or monitoring are expected unless further degradation of safety systems occurs.

License Amendment Request Attachment 1 Revise EAL HU1 .5 to Remove High Lake Level Page 6of11 NRC Docket Nos. 50-220 and 50-410 Evaluation of Proposed Changes Discussion: Potential degradation of the level of safety of the plant is indicated primarily by exceeding plant technical specification Limiting Condition of Operation (LCO) allowable action statement time for achieving required mode change. Precursors of more serious events should also be included because precursors do represent a potential degradation in the level of safety of the plant. Minor releases of radioactive materials are included. In this emergency classification level, however, releases do not require monitoring or off-site response.

In response to the NEI 99-01, Revision 5 guidance, NMP1 and NMP2 EALs have established EAL HU1 .5 to provide the minimum and maximum lake levels to serve as a precursor to a more serious event. Specifically, the NMP EALs state the UE maximum lake level of 248.2 ft., in HU1 .5 is:

..based on upon the maximum attainable uncontrolled lake water level as specified in the USAR. Dams on the St. Lawrence River, under the authority of the International St. Lawrence River Board of Control, are now used to regulate the lake level.

However, the BWT currently controls lake water level in excess of the UE threshold at NMP1 and NMP2. An amendment to the Boundary Waters Treaty of 1909 was implemented on December 8, 2016, via a Supplementary Order of Approval. The Order required the regulated monthly mean level of Lake Ontario not to exceed monthly values which have occurred between 1900 and 2005. Trigger levels were established for actions to prevent lake level from getting too high or too low. The high trigger levels for May (248.46 ft.) and June (248.33 ft.) exceed the 248.2 ft., UE EAL thresholds at NMP1 and NMP2.

Note that the spring of 2017 has been characterized by high rain levels and low snow and lake freezing levels. As of April 20, 2017, the U.S. Army Corps of Engineers stated that, over the last month, water levels have risen on all of the Great Lakes. Lake Ontario has risen 15 inches. In the next month, Lake Ontario is currently expected to rise approximately 6 inches. The forecast levels exceed the long-term average by 10 inches for the months in question.

License Amendment Request Attachment 1 Revise EAL HU1 .5 to Remove High Lake Level Page 7of11 NRC Docket Nos. 50-220 and 50-410 Evaluation of Proposed Changes Lalie Ontario 51100 i *>00

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Detroit District ii 500 Cili!l1datum,l4J31 http://www.lre.usace.army.mll 4120/2017 As noted, the Lake Ontario lake levels are being controlled to a level which exceeds the UE threshold at NMP1 and NMP2. Because of the IJC actions to control the lake level above the EAL threshold, the UE EAL must be revised to prevent potentially frequent and unjustified entries into an emergency (i.e., UE) condition.

A review of the basis for the NMP maximum lake level threshold of 248.2 ft., indicates there is no operational impact that corresponds to the value. The HU1 .5 threshold was selected solely based on being a precursor to a more serious event. Note that the NMP1 and NMP2 EALs provide an "Alert" EAL HA1 .5 with a maximum lake level of> 254 ft.

This "Alert" level is approximately 6 ft., above the UE level and represents a large differential considering the size of the lake and the time required for the lake levels to change by this amount (i.e., approximately 6 ft.).

Exelon proposes to revise the HU1 .5 EAL threshold to remove the UE maximum lake level threshold. NMP site utilizes Operating Procedures to monitor the lake level twice daily.

Changes to lake level occur incrementally and large changes occur over extended periods of time. Observed changes/trends are communicated to Station Management for review and appropriate actions. The IJC also provides predictions for significant lake level changes. Station Management is alerted and has opportunity to evaluate trends as the "Alert" EAL threshold is approached.

Given the gradual rate at which significant lake level changes occur, the approximate 6 ft.,

difference between the UE and "Alert" threshold noted above, provides little value in being a precursor to the "Alert" classification. Considering there is no operational impact to the station until after the 254 ft., "Alert" level, the UE threshold is not necessary or useful considering the frequent monitoring performed by Operations at each unit. UE declarations would be declared unnecessarily with little association or correlation to potential to reach the "Alert" threshold.

License Amendment Request Attachment 1 Revise EAL HU1 .5 to Remove High Lake Level Page 8of11 NRC Docket Nos. 50-220 and 50-41 O Evaluation of Proposed Changes

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria The following regulatory requirements have been considered:

The proposed changes have been evaluated to determine whether applicable regulations and requirements continue to be met.

10 CFR 50.47, Emergency Plans: 10 CFR 50.47(b)(4) states, "A standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures." The revised EAL criteria continue to meet the requirements of this regulation.

10 CFR 50.54(q) states: " ... (2) A holder of a nuclear power reactor operating license under this part ... shall follow and maintain in effect emergency plans which meet the requirements in Appendix E of this part .... (4) The changes to a licensee's emergency plan that reduce the effectiveness of the plan ... may not be implemented without prior approval by the NRC. A licensee desiring to make such a change ... shall submit an application for an amendment to its license .... " This LAR for revising the EAL is being submitted in accordance with this regulation. 10 CFR Part 50, Appendix E, Emergency Planning and Preparedness for Production and Utilization Facilities,Section IV.B states, "emergency plans are to include EALs, which are to be used as criteria for determining the need for notification and participation of State and local agencies, the NRC and other Federal agencies .... " The revised actions continue to meet the requirements of this regulation.

10 CFR 50, Appendix E, Emergency Planning and Preparedness for Production and Utilization Facilities,Section IV.B states, "emergency plans are to include EALs, which are to be used as criteria for determining the need for notification and participation of State and local agencies, the NRC and other Federal agencies .... "The revised actions continue to meet the requirements of this regulation.

Regulatory Guide 1.219, Revision 1, "Guidance on Making Changes to Emergency Plans for Nuclear Power Reactors," dated July 2016. This Regulatory Guide describes a method that the NRC considers acceptable to implement the requirements of 10 CFR 50.54(q) related to emergency preparedness and specifically to making changes to emergency response plans.

4.2 Precedent There is no precedent for this proposed change.

License Amendment Request Attachment 1 Revise EAL HU1 .5 to Remove High Lake Level Page 9of11 NRC Docket Nos. 50-220 and 50-410 Evaluation of Proposed Changes 4.3 No Significant Hazards Consideration In accordance with 10 CFR 50.90, Application for amendment of/icense, construction permit, or early site permit," Exelon Generation Company, LLC (Exelon) requests license amendments for the facilities listed below in support of Emergency Plan changes to revise Emergency Action Level (EAL) HU1 .5 related to the declaration of a Notice of Unusual Event (UE) emergency classification on high lake level.

Nine Mile Point Nuclear Station, Unit 1 (NMP1)

Nine Mile Point Nuclear Station, Unit 2 (NMP2)

The proposed changes have been reviewed considering the applicable requirements of, 10 CFR 50.47, 10 CFR 50, Appendix E, 10 CFR 50.54(q}, and other applicable NRC guidance. Exelon has evaluated the proposed changes to the affected facilities Emergency Plans and determined that the changes do not involve a Significant Hazards Consideration. In support of this determination, an evaluation of each of the three (3) standards, set forth in 10 CFR 50.92, "Issuance of amendment," is provided below.

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes to EAL HU1 .5 do not reduce the capability to meet the emergency planning requirements established in 10 CFR 50.47 and 10 CFR 50, Appendix E. The proposed changes do not reduce the functionality, performance, or capability of the Emergency Response Organization (ERO) to respond in mitigating the consequences of accidents or transients. All required ERO functions at the facilities will continue to be performed as required.

The probability of a reactor accident requiring implementation of Emergency Plan EALs has no relevance'in determining whether the proposed changes to the EAL HU1 .5 reduce the effectiveness of the Emergency Plans for NMP1 and NMP2. As discussed in Section D, "Planning Basis," of NUREG-0654, Revision 1, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants":

" ... The overall objective of emergency response plans is to provide dose savings (and in some cases immediate life saving) for a spectrum of accidents that could produce offsite doses in excess of Protective Action Guides (PAGs). No single specific accident sequence should be isolated as the one for which to plan because each accident could have different consequences, both in nature and degree. Further, the range of possible selection for a planning basis is very large, starting with a zero point of requiring no planning at all because significant offsite radiological accident consequences are unlikely to occur, to planning for the worst possible accident, regardless of its extremely low likelihood .... "

License Amendment Request Attachment 1 Revise EAL HU1 .5 to Remove High Lake Level Page 10 of 11 NRC Docket Nos. 50-220 and 50-41 O Evaluation of Proposed Changes Therefore, Exelon did not consider the risk insights regarding any specific accident initiation or progression in evaluating the proposed changes.

The proposed changes do not involve any physical changes to plant equipment or systems, nor do they alter the assumptions of any accident analyses. The proposed changes do not adversely affect accident initiators or precursors nor do they alter the design assumptions, conditions, and configuration or the manner in which the plants are operated and maintained. The proposed changes do not adversely affect the ability of Structures, Systems, or Components (SSCs) to perform their intended safety functions in mitigating the consequences of an initiating event within the assumed acceptance limits.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed changes to EAL HU1 .5 do not involve any physical changes to plant systems or equipment. The proposed changes do not involve the addition of any new plant equipment. The proposed changes will not alter the design configuration, or method of operation of plant equipment beyond its normal functional capabilities. All Exelon ERO functions will continue to be performed as required. The proposed changes do not create any new credible failure mechanisms, malfunctions, or accident initiators.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from those that have been previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed changes to EAL HU1 .5 do not alter or exceed a design basis or safety limit. There is no change being made to safety analysis assumptions, safety limits, or limiting safety system settings that would adversely affect plant safety as a result of the proposed changes. There are no changes to setpoints or environmental conditions of any SSC or the manner in which any SSC is operated. Margins of safety are unaffected by the proposed changes. The applicable requirements of 10 CFR 50.47 and 10 CFR 50, Appendix E will continue to be met.

Therefore, the proposed changes do not involve any reduction in a margin of safety.

License Amendment Request Attachment 1 Revise EAL HU 1.5 to Remove High Lake Level Page 11 of 11 NRC Docket Nos. 50-220 and 50-41 O Evaluation of Proposed Changes 4.4 Conclusions In conclusion, and based on the considerations discussed above: 1) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed changes to EAL HU1 .5; 2) the changes will be in compliance with the NRC's regulations; and 3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

The proposed amendment would change requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 because the amendment approves an acceptable EAL change which is required for operation of the facility. Exelon has evaluated the proposed change and has determined that the change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. Regulation Plan 2014 for the Lake Ontario and the St. Lawrence River

2. Nuclear Energy Institute Guidance Document 99-01, Revision 5

3. Nine Mile Point Unit 1Final Safety Analysis Report (Updated), Revision 24

4. Updated Safety Analysis Report Nine Mile Point Unit 2

ATTACHMENT 2 License Amendment Request Change Emergency Action Level HU1 .5 to Remove High Lake Level Initiating Condition for Unusual Event Emergency Classification Markup of Proposed Emergency Action Level Matrices Pages NMP1 EAL Matrices Pages 1

2 NMP2 EAL Matrices Pages 1

2

Attachment 2 Page 1 of 4 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT RS1.1 1 2 3 4 D RA1.1 1 2 3 4 D RU1.1 1 2 3 4 5 D D SG1.1 1 2 SS1.1 1 2 SA1.1 1 2 SU1.1 1 2 Loss of all offsite and all onsite AC power, Table S-1, to 4.16 Loss of all offsite and all onsite AC power, Table S-1, to AC power capability to 4.16 kV emergency buses reduced to Loss of all offsite AC power, Table S-1, to 4.16 kV ANY monitor reading > Table R-1 SAE column ANY gaseous monitor reading > Table R-1 Alert column ANY gaseous monitor reading > Table R-1 UE column kV emergency buses 4.16 KV emergency buses for 15 min. (Note 4) a single power source, Table S-1, for 15 min. (Note 4) emergency buses for 15 min. (Note 4) for 15 min. (Note 1) for 15 min. (Note 2) for 60 min. (Note 2) AND EITHER: AND

• Do not delay declaration awaiting dose assessment results

• If dose assessment results are available, declaration 1 Restoration of at least one 4.16 kV emergency bus within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely ANY additional single power source failure will result in a loss of all 4.16 kV emergency bus power Table S-1 AC Power Sources RA1.2 1 2 3 4 D RU1.2 1 2 3 4 D OR should be based on dose assessment instead of Loss of RPV water level cannot be restored and maintained radiation monitor values (see EAL RS1.2) AC Power above -84 in. or RPV water level cannot be determined Onsite ANY liquid monitor reading > Table R-1 Alert column ANY liquid monitor reading > Table R-1 UE column

• DG 102 for 15 min. (Note 2) for 60 min. (Note 2)

• DG 103 None

• T-101N Offsite RA1.3 1 2 3 4 D RU1.3 1 2 3 4 D

• T-101S SS2.1 1 2 Confirmed sample analyses for gaseous or liquid releases Confirmed sample analyses for gaseous or liquid releases 2 Loss of None < 106 VDC on both Battery Board 11 and Battery Board 12 None

• T-10 backfed from offsite through T-1 or T-2 (only if already aligned) indicate concentrations or release rates > 200 x ODCM limits indicate concentrations or release rates > 2 x ODCM limits for 15 min. (Note 4)

DC Power 1 for 15 min. (Note 2) for 60 min. (Note 2)

SG3.1 1 SS3.1 1 SA3.1 1 SU3.1 2 Offsite Rad An automatic scram fails to shut down the reactor An automatic scram failed to shut down the reactor as An automatic scram failed to shut down the reactor An UNPLANNED sustained positive period observed on Conditions as indicated by reactor power > 6% indicated by reactor power > 6% AND nuclear instrumentation RG1.2 1 2 3 4 D Dose assessment using actual meteorology indicates doses RS1.2 1 2 3 4 D Dose assessment using actual meteorology indicates doses 3 AND All manual actions fail to shut down the reactor as indicated by reactor power > 6%

AND Manual actions taken at the reactor control console (mode Manual actions taken at the reactor control console (mode switch in shutdown, manual scram push buttons or switch in shutdown, manual scram push buttons and ARI) ARI) successfully shut down the reactor as indicated by

> 1,000 mRem TEDE or 5,000 mRem thyroid CDE at or > 100 mRem TEDE or 500 mRem thyroid CDE at or beyond None None Criticality AND EITHER of the following exist or have occurred: failed to shut down the reactor as indicated by reactor reactor power 6%

beyond the SITE BOUNDARY the SITE BOUNDARY & RPV water level cannot be restored and maintained power > 6%

RPS above -109 in. or RPV water level cannot be determined RG1.3 1 2 3 4 D RS1.3 1 2 3 4 D Failure OR Torus water temperature and RPV pressure cannot be Field survey results indicate closed window dose rates Field survey results indicate closed window dose rates maintained below the Heat Capacity Temperature Limit R > 1,000 mRem/hr expected to continue for 60 min. at or beyond the SITE BOUNDARY (Note 1)

> 100 mRem/hr expected to continue for 60 min. at or beyond the SITE BOUNDARY (Note 1)

None None (N1-EOP-4 Figure M)

Abnorm.

OR Analyses of field survey samples indicate thyroid CDE OR Analyses of field survey samples indicate thyroid CDE 4 SU4.1 1 2 Rad Release

/ Rad

> 5,000 mRem for 1 hr of inhalation at or beyond the SITE BOUNDARY (Note 1)

> 500 mRem for 1 hr of inhalation at or beyond the SITE BOUNDARY (Note 1) S Inability to Reach or Maintain None None None Plant is not brought to required operating mode within Technical Specifications LCO required action completion time Shutdown Effluent Conditions System RA2.1 1 2 3 4 D RU2.1 1 2 3 4 D Malfunct.

SS5.1 1 2 SA5.1 1 2 SU5.1 1 2 Alarm on ANY of the following radiation monitors due to UNPLANNED water level drop in a reactor refueling pathway damage to irradiated fuel or loss of water level: as indicated by inability to restore and maintain SFP level > low Loss of > approximately 75% of annunciation or indication UNPLANNED loss of > approximately 75% of annunciation or UNPLANNED loss of > approximately 75% of annunciation or Table R-1 Effluent Monitor Classification Thresholds

• ARM 18 (West end of shield wall) water level alarm (Note 3) on Control Room panels L, K, H, F and G for 15 min. indication on Control Room panels L, K, H, F and G for 15 indication on Control Room panels L, K, H, F and G for 15

• ARM 25 (Rx building - east wall) AND Table S-2 Significant Transients (Note 4) min. (Note 4) min. (Note 4)

Monitor GE SAE ALERT UE

• ARM 29 (Refuel bridge (LOW RANGE)) Area radiation monitor reading rise on ANY of the following: AND AND EITHER:

• Refuel Bridge (HIGH RANGE)

• ARM 18 (West end of shield wall)

• Turbine runback > 25% thermal reactor power A significant transient is in progress, Table S-2 A significant transient is in progress, Table S-2 2

• Reactor Building Vent Radiation Monitor 5

GASEOUS Stack (RN 10A/B) N/A N/A 3.0E4 cps 300 cps

• ARM 25 (Rx building - east wall)

• Electric load rejection > 25% full electrical load AND OR

• ARM 29 (Refuel bridge (LOW RANGE)) Compensatory indications are unavailable (Plant Computer, Compensatory indications are unavailable (Plant

• Refuel Bridge (HIGH RANGE)

• Reactor scram SPDS)

Inst. Process Computer, SPDS)

Onsite Rad EC Vent N/A 300 mRem/hr 30 mRem/hr 10 mRem/hr

• ECCS injection Conditions

&

• Thermal power oscillations > 10%

Spent Fuel SW Effluent N/A N/A 90,000 cpm 900 cpm LIQUID Events RA2.2 1 2 3 4 D RU2.2 1 2 3 4 D RW Discharge N/A N/A 200 x batch 2 x batch A water level drop in a reactor refueling pathway that will UNPLANNED area radiation readings rise by a factor of result in irradiated fuel becoming uncovered 1,000 over NORMAL LEVELS SU6.1 1 2 Table S-3 Communications Systems 6 None None System Onsite Offsite (internal) (external)

Loss of all Table S-3 onsite (internal) communication methods affecting the ability to perform routine operations OR Comm. Loss of all Table S-3 offsite (external) communication RA3.1 1 2 3 4 D PBX (normal dial telephones) X X methods affecting the ability to perform offsite notifications 3 None None Dose rates > 15 mRem/hr in EITHER of the following areas requiring continuous occupancy to maintain plant safety Gaitronics X SU7.1 1 2 Hand-Held Portable Radio (station radio) X CR/CAS Rad functions:

Control Room OR 7 Control Room installed satellite phones (non portable) X Reactor coolant activity > 4 µCi/gm I-131 Equivalent CAS Fuel Clad None ENS None X SU7.2 1 2 Degradation RECS X HA1.1 1 2 3 4 D HU1.1 1 2 3 4 D Offgas radiation monitor RN-12A or RN-12B hi-hi alarm for 15 min.

NMP-2 seismic instrumentation indicates > 0.075 g Seismic event identified by ANY two of the following:

Table H-1 Safe Shutdown Areas AND

• Annunciator H2-1-6 SEISMIC DETECTION SU8.1 1 2 Earthquake confirmed by ANY of the following: EQUIPMENT EVENT indicates seismic event detected

• Reactor Building (including Primary Containment)

• Control Room

• Earthquake felt in plant

• JAFNPP seismic instrumentation

• Control Room indication of degraded performance of

• Confirmation of earthquake received on NMP-2 or JAFNPP seismic instrumentation 8 None None None Unidentified drywell leakage > 10 gpm OR

• Earthquake felt in plant RCS Identified reactor coolant drywell leakage > 25 gpm

• Screenhouse systems required for the safe shutdown of the plant Leakage

• Turbine Building HA1.2 1 2 3 4 D HU1.2 1 2 3 4 D

• Battery Rooms Tornado striking Tornado striking within PROTECTED AREA boundary FG1.1 FS1.1 FA1.1 FU1.1 F

• Battery Board Rooms OR 1 2 1 2 1 2 1 2 OR Sustained high winds > 90 mph Sustained high winds > 90 mph

• Cable Spreading Room Loss of ANY two fission product barriers Loss or potential loss of ANY two fission product barriers ANY loss or ANY potential loss of EITHER Fuel Clad ANY loss or ANY potential loss of Containment barrier resulting in EITHER: Fission Product

• Main Steam Isolation Valve Room VISIBLE DAMAGE to ANY SAFETY-RELATED AND (Table F-1) barrier OR RCS barrier (Table F-1) (Table F-1)

STRUCTURE, SYSTEM or COMPONENT within ANY Table Barrier Loss or potential loss of third fission product barrier

• Diesel Generator Engine and Board Rooms Degradation H-1 area (Table F-1)

• Security OR

• Central Alarm Station Control Room indication of degraded performance of ANY SAFETY-RELATED STRUCTURE, SYSTEM or

• Secondary Alarm Station COMPONENT within ANY Table H-1 area

• Security Uninterruptible Power Supply Room HA1.3 1 2 3 4 D HU1.3 1 2 3 4 D Table F-1 Fission Product Barrier Matrix Internal flooding Internal flooding that has the potential to affect ANY resulting in EITHER: SAFETY-RELATED STRUCTURE, SYSTEM, OR Fuel Clad Barrier Reactor Coolant System Barrier Containment Barrier An electrical shock hazard that precludes access to operate COMPONENT required by Technical Specifications for the or monitor ANY SAFETY-RELATED STRUCTURE, current operating mode in ANY Table H-1 area Loss Potential Loss Loss Potential Loss Loss Potential Loss Notes SYSTEM, OR COMPONENT within ANY Table H-1 area 1 1. The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time OR Control Room indication of degraded performance of ANY

1. Primary Containment Flooding is required

1. RPV water level cannot be 1. RPV water level cannot be restored and maintained restored and maintained

1. Primary Containment Flooding is required SAFETY-RELATED STRUCTURE, SYSTEM or Natural or Destructive Phenomena

2. The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the release duration has exceeded, or will likely exceed, the applicable time.

In the absence of data to the contrary, assume that the release duration has exceeded the applicable COMPONENT within ANY Table H-1 area HA1.4 1 2 3 4 D HU1.4 1 2 3 4 D A above -84 in. following depressurization of the RPV or RPV water level cannot above -84 in. or RPV water level cannot be determined None None RPV Water time if an ongoing release is detected and the release start time is unknown. Level be determined Turbine failure-generated PROJECTILEs Turbine failure resulting in ANY of the following:

3. If loss of water level in the refueling pathway occurs while in Mode 3, 4 or D, consider classification

• Casing penetration resulting in EITHER:

under EALs CU3.1, CU3.2 or CU3.3

• Damage to turbine seals VISIBLE DAMAGE to or penetration of ANY SAFETY-

4. The ED should not wait until the applicable time has elapsed, but should declare the event as soon RELATED STRUCTURE, SYSTEM or COMPONENT

• Damage to generator seals 2. Torus pressure > 35 psig as it is determined that the condition has exceeded, or will likely exceed, the applicable time. within ANY Table H-1 area 2. Primary Containment 1. Primary Containment and rising OR pressure > 3.5 psig due to pressure rise followed by a

5. If the equipment in the stated area was already inoperable, or out of service, before the event RCS leakage rapid UNPLANNED drop in occurred, then EAL HA3.1 should not be declared as it will have no adverse impact on the ability of Control Room indication of degraded performance of 3. Explosive mixture exists the plant to safely operate or safely shutdown beyond that already allowed by Technical Specifications at the time of the event.

ANY SAFETY-RELATED STRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 area B Primary Containment pressure inside Primary Containment

( 6% H2 and 5% O2)

Primary None None None HA1.5 1 2 3 4 D HU1.5 1 2 3 4 D Containment

2. Primary Containment Pressure / 4. Torus water temperature Lake water level > 254 ft Lake water level > 248.2 ft pressure response not Temperature and RPV pressure cannot OR OR consistent with LOCA be maintained below the Intake water level < 236 ft Intake water level < 238.8 ft conditions Heat Capacity Temperature Limit (N1-EOP-4 Figure M)

HA1.6 1 2 3 4 D

3. Release pathway exists 1. UNISOLABLE primary 3. Failure of all Primary Vehicle crash outside Pirmary system leakage outside Containment isolation resulting in EITHER: Containment resulting from Primary Containment as valves in ANY one line to VISIBLE DAMAGE to ANY SAFETY-RELATED isolation failure in ANY of indicated by exceeding close following auto or H STRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 area OR Control Room indication of degraded performance of ANY the following systems (excluding normal process system flowpaths from an UNISOLABLE system):

EITHER:

ANY N1-EOP-5 Detail T area temperature alarm setpoint manual initiation AND Direct downstream pathway outside Primary SAFETY-RELATED STRUCTURE, SYSTEM or

• Main steam line OR Containment and to the Hazards COMPONENT within ANY Table H-1 area

• EC steam line ANY N1-EOP-5 Detail R environment exists

&

• RWCU area radiation alarm Other HA2.1 1 2 3 4 D HU2.1 1 2 3 4 D

• Feedwater setpoint 4. Intentional Primary Containment venting per Condi-tions FIRE or EXPLOSION resulting in EITHER:

FIRE not extinguished within 15 min. of Control Room notification or verification of a Control Room FIRE alarm in C None None

4. RPV blowdown is required EOPs None Affect-ing 2 None None VISIBLE DAMAGE to ANY SAFETY-RELATED STRUCTURE, SYSTEM or COMPONENT within ANY Table ANY Table H-1 area, RadWaste Solidification and Storage Bldg, or Security West Bldg (Note 4)

Isolation 5. UNISOLABLE primary system leakage outside Plant H-1 area Primary Containment as Safety Fire or indicated by exceeding Explosion OR HU2.2 1 2 3 4 D EITHER:

Control Room indication of degraded performance of ANY Maximum safe general SAFETY-RELATED STRUCTURE, SYSTEM or EXPLOSION of sufficient force to damage permanent area temperature of COMPONENT within ANY Table H-1 area structures or equipment within the PROTECTED AREA 135ºF OR HA3.1 1 2 3 4 D HU3.1 1 2 3 4 D Maximum safe area radiation of 8 R/hr 3 Access to ANY Table H-1 area is prohibited due to toxic, corrosive, asphyxiant or flammable gases which jeopardize operation of systems required to maintain safe operations or Toxic, corrosive, asphyxiant or flammable gases in amounts that have or could adversely affect NORMAL PLANT OPERATIONS 2. Drywell radiation 5. Drywell radiation 80 R/hr 5. Drywell radiation None None Hazardous safely shutdown the reactor (Note 5) 3,000 R/hr 4.0 E4 R/hr Gas HU3.2 1 2 3 4 D Recommendation by local, county or state officials to D 3. Reactor coolant activity None None None evacuate or shelter site personnel based on an offsite event Rad > 300 µCi/gm I-131 Equivalent HG4.1 1 2 3 4 D HS4.1 1 2 3 4 D HA4.1 1 2 3 4 D HU4.1 1 2 3 4 D 4. ANY condition in the 2. ANY condition in the 6. ANY condition in the 2. ANY condition in the opinion 6. ANY condition in the opinion 6. ANY condition in the opinion opinion of the Emergency opinion of the Emergency opinion of the Emergency of the Emergency Director of the Emergency Director of the Emergency Director A HOSTILE ACTION has occurred such that plant personnel are unable to operate equipment required to A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Site A HOSTILE ACTION is occurring or has occurred within the Owner Controlled Area as reported by the Security Site A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Security Site Supervisor E Director that indicates loss of the Fuel Clad barrier Director that indicates potential loss of the Fuel Director that indicates loss of the Reactor Coolant that indicates potential loss of the Reactor Coolant that indicates loss of the Containment barrier that indicates potential loss of the Containment barrier 4 maintain safety functions HG4.2 1 2 3 4 D Supervisor Supervisor OR OR A credible site-specific security threat notification Judgment Clad barrier System barrier System barrier A validated notification from NRC of an AIRLINER attack OR Security A HOSTILE ACTION has caused failure of Spent Fuel threat within 30 min. of the site A validated notification from NRC providing information of an Cooling systems aircraft threat AND IMMINENT fuel damage is likely HS5.1 1 2 3 4 D HA5.1 1 2 3 4 D 5 None Control Room evacuation has been initiated Control Room evacuation has been initiated None Control AND Room Control of the plant cannot be established within 15 min.

Evacuation HG6.1 1 2 3 4 D HS6.1 1 2 3 4 D HA6.1 1 2 3 4 D HU6.1 1 2 3 4 D Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial have occurred which involve actual or likely major failures of have occurred which involve an actual or potential have occurred which indicate a potential degradation of the 6 core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; (1) toward site personnel or equipment that substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring Judgment reasonably expected to exceed EPA Protective Action could lead to the likely failure of or; (2) that prevent effective HOSTILE ACTION. ANY releases are expected to be limited are expected unless further degradation of safety systems Prepared for Constellation by: Operations Support Services, Inc. - www.ossi-net.com (Rev. 23 1/18/13)

Guideline exposure levels (1,000 mRem TEDE or 5,000 access to equipment needed for the protection of the public. to small fractions of the EPA Protective Action Guideline occurs mRem thyroid CDE) offsite for more than the immediate site ANY releases are not expected to result in exposure levels exposure levels (1,000 mRem TEDE or 5,000 mRem thyroid area which exceed EPA Protective Action Guideline exposure CDE) levels (1,000 mRem TEDE or 5,000 mRem thyroid CDE) beyond the SITE BOUNDARY EU1.1 1 2 3 4 D EAL Identifier E None None None Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by measured dose rates > then ANY of the following:

• 400 mRem/hr at 3 feet from the HSM surface

• 100 mRem/hr outside HSM door on centerline XXX.X Category (R, H, E, S, F, C) Sequential number within subcategory/classification ISFSI

• 20 mRem/hr end shield wall exterior Emergency classification (G, S, A, U) Subcategory number (1 if no subcategory)

Modes:

EPIP-EPP-01-EAL 1

Power Operation 2

Hot Shutdown 3

Cold Shutdown 4

Refuel D

Defueled Nine Mile Point Nuclear Station Unit 1 Attachment 1 Revision 234 EAL Matrix Unit 1 Page 1 of 2 MODE 1 or 2

Attachment 2 Page 2 of 4 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT RS1.1 1 2 3 4 D RA1.1 1 2 3 4 D RU1.1 1 2 3 4 5 D D CA1.1 3 4 D CU1.1 3 4 D Table C-1 AC Power Sources ANY monitor reading > Table R-1 SAE column for 15 min. (Note 1)

ANY gaseous monitor reading > Table R-1 Alert column for 15 min. (Note 2)

ANY gaseous monitor reading > Table R-1 UE column for 60 min. (Note 2) 1 Loss of all offsite and all onsite AC power, Table C-1, to 4.16 kV emergency buses for 15 min. (Note 4)

AC power capability to 4.16 kV emergency buses reduced to a single power source, Table C-1, for 15 min. (Note 4)

Onsite None

• Do not delay declaration awaiting dose assessment

• DG102 AND Loss of results

• DG103 ANY additional single power source failure will result in a loss

• If dose assessment results are available, declaration AC Power RA1.2 1 2 3 4 D RU1.2 1 2 3 4 D of all 4.16 kV emergency bus power should be based on dose assessment instead of radiation monitor values (see EAL RS1.2)

• T-101N ANY liquid monitor reading > Table R-1 Alert column ANY liquid monitor reading > Table R-1 UE column Offsite for 15 min. (Note 2) for 60 min. (Note 2)

• T-101S CU2.1 3 4 2

• T-10 backfed from offsite through T-1 or T-2 (only if already aligned) None < 106 VDC on required 125 VDC buses (Battery board 11, Loss of Battery board 12) for 15 min. (Note 4)

RA1.3 1 2 3 4 D RU1.3 1 2 3 4 D DC Power CG3.1 3 4 CS3.1 3 4 CA3.1 3 4 CU3.1 3 Confirmed sample analyses for gaseous or liquid releases Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 200 x ODCM limits indicate concentrations or release rates > 2 x ODCM limits RPV water level < -84 in. for 30 min. (Note 4) With CONTAINMENT CLOSURE not established, RPV RPV water level < +5 in. RCS leakage results in the inability to maintain or restore RPV 1 for 15 min. (Note 2) for 60 min. (Note 2)

AND ANY Containment Challenge Indication, Table C-3 water level < -1 in. OR RPV water level cannot be monitored for 15 min. with ANY water level > +53 in. for 15 min. (Note 4)

Offsite Rad UNPLANNED RPV leakage indication, Table C-2 (Note 4)

Conditions RG1.2 1 2 3 4 D RS1.2 1 2 3 4 D CG3.2 3 4 CS3.2 3 4 CU3.2 4 Dose assessment using actual meteorology indicates doses Dose assessment using actual meteorology indicates doses RPV water level cannot be monitored with core uncovery With CONTAINMENT CLOSURE established, RPV water

> 1,000 mRem TEDE or 5,000 mRem thyroid CDE at or > 100 mRem TEDE or 500 mRem thyroid CDE at or beyond None None UNPLANNED RPV water level drop below EITHER of the indicated by ANY of the following for 30 min. (Note 4): level < -84 in.

beyond the SITE BOUNDARY the SITE BOUNDARY following for 15 min. (Note 4):

• ANY UNPLANNED RPV leakage indication, Table C-2

• Erratic Source Range Monitor indication

• 0 ft Flange Level (RPV flange)

RG1.3 1 2 3 4 D RS1.3 1 2 3 4 D CS3.3 3 4

• RPV water level band (when the RPV water level band AND is established below the RPV flange)

ANY Containment Challenge Indication, Table C-3 RPV water level cannot be monitored for 30 min. (Note 4)

Field survey results indicate closed window dose rates Field survey results indicate closed window dose rates with a loss of RPV inventory as indicated by ANY of the R > 1,000 mRem/hr expected to continue for 60 min. at or beyond the SITE BOUNDARY (Note 1)

OR

> 100 mRem/hr expected to continue for 60 min. at or beyond the SITE BOUNDARY (Note 1)

OR None None 3 following:

• ANY UNPLANNED RPV leakage indication, Table C-2

• Erratic Source Range Monitor indication CU3.3 4 Abnorm.

Rad Release Analyses of field survey samples indicate thyroid CDE

> 5,000 mRem for 1 hr of inhalation at or beyond the SITE BOUNDARY (Note 1)

Analyses of field survey samples indicate thyroid CDE

> 500 mRem for 1 hr of inhalation at or beyond the SITE BOUNDARY (Note 1)

C RPV Water Level RPV water level cannot be monitored with a loss of RPV inventory as indicated by ANY UNPLANNED RPV leakage indication, Table C-2

/ Rad Table C-2 RPV Leakage Indications Effluent Cold SD/

Refuel RA2.1 1 2 3 4 D RU2.1 1 2 3 4 D

• Drywell equipment drain tank level rise System Alarm on ANY of the following radiation monitors due to UNPLANNED water level drop in a reactor refueling pathway Malfunct.

• Drywell floor drain tank level rise damage to irradiated fuel or loss of water level: as indicated by inability to restore and maintain SFP level > low

• Reactor building equipment sump level rise Table R-1 Effluent Monitor Classification Thresholds

• ARM 18 (West end of shield wall) water level alarm (Note 3)

• ARM 25 (Rx building - east wall)

• Reactor Building floor drain sump level rise AND Monitor GE SAE ALERT UE

• ARM 29 (Refuel bridge (LOW RANGE)) Area radiation monitor reading rise on ANY of the following:

• Torus water level rise

• Refuel Bridge (HIGH RANGE)

• ARM 18 (West end of shield wall)

• UNPLANNED rise in RPV make-up rate 2

• Reactor Building Vent Radiation Monitor GASEOUS Stack (RN 10A/B) N/A N/A 3.0E4 cps 300 cps

• ARM 25 (Rx building - east wall)

• ARM 29 (Refuel bridge (LOW RANGE))

• Observation of UNISOLABLE RCS leakage

• Refuel Bridge (HIGH RANGE)

Onsite Rad EC Vent N/A 300 mRem/hr 30 mRem/hr 10 mRem/hr Conditions

& Table C-3 Containment Challenge Spent Fuel SW Effluent N/A N/A 90,000 cpm 900 cpm LIQUID Events Indications CA4.1 3 4 CU4.1 3 4 RA2.2 1 2 3 4 D RU2.2 1 2 3 4 D RW Discharge N/A N/A 200 x batch 2 x batch A water level drop in a reactor refueling pathway that will UNPLANNED area radiation readings rise by a factor of 4

• CONTAINMENT CLOSURE not established

• Explosive mixture exists inside Primary Containment An UNPLANNED event results in EITHER:

RCS temperature > 212°F for > Table C-4 duration OR Unplanned event results in RCS temperature > 212°F result in irradiated fuel becoming uncovered 1,000 over NORMAL LEVELS (H2 6% and O2 5%) CU4.2 3 4 RCS RPV pressure increase > 10 psi due to an UNPLANNED Temp. loss of decay heat removal capability Loss of all RCS temperature and RPV water level

• UNPLANNED rise in Primary Containment pressure indication for 15 min. (Note 4)

• RB area radiation > 8 R/hr RA3.1 1 2 3 4 D 3 None None Dose rates > 15 mRem/hr in EITHER of the following areas requiring continuous occupancy to maintain plant safety 5 CU5.1 3 4 An UNPLANNED sustained positive period observed on CR/CAS functions: Inadvertent Table C-4 RCS Reheat Duration nuclear instrumentation Rad Control Room Criticality Thresholds OR CAS

• If an RCS heat removal system is in operation within this time CU6.1 3 4 D HA1.1 1 2 3 4 D HU1.1 1 2 3 4 D frame and RCS temperature is being reduced, the EAL is not Table C-5 Communications Systems applicable Table H-1 Safe Shutdown Areas NMP-2 seismic instrumentation indicates > 0.075 g Seismic event identified by ANY two of the following: 6 RCS Status CONTAINMENT CLOSURE Status Duration System Onsite Offsite (internal) (external)

Loss of all Table C-5 onsite (internal) communication methods affecting the ability to perform routine operations AND

• Annunciator H2-1-6 SEISMIC DETECTION Comm.

OR Earthquake confirmed by ANY of the following: EQUIPMENT EVENT indicates seismic event detected Loss of all Table C-5 offsite (external) communication

• Reactor Building (including Primary Containment)

• Earthquake felt in plant INTACT N/A 60 min.* PBX (normal dial telephones) X X

• Confirmation of earthquake received on NMP-2 or methods affecting the ability to perform offsite notifications

• Control Room

• JAFNPP seismic instrumentation JAFNPP seismic instrumentation Gaitronics X

• Control Room indication of degraded performance of Established 20 min.*

• Screenhouse

• Earthquake felt in plant Hand-Held Portable Radio (station radio) X systems required for the safe shutdown of the plant Not INTACT

• Turbine Building Not established 0 min. Control Room installed satellite phones (non portable) X HA1.2 1 2 3 4 D HU1.2 1 2 3 4 D

• Battery Rooms ENS X Tornado striking Tornado striking within PROTECTED AREA boundary RECS X

• Battery Board Rooms OR OR

• Cable Spreading Room Sustained high winds > 90 mph Sustained high winds > 90 mph resulting in EITHER:

• Main Steam Isolation Valve Room VISIBLE DAMAGE to ANY SAFETY-RELATED

• Diesel Generator Engine and Board Rooms STRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 area

• Security OR

• Central Alarm Station Control Room indication of degraded performance of ANY SAFETY-RELATED STRUCTURE, SYSTEM or

• Secondary Alarm Station COMPONENT within ANY Table H-1 area

• Security Uninterruptible Power Supply Room HA1.3 1 2 3 4 D HU1.3 1 2 3 4 D Internal flooding Internal flooding that has the potential to affect ANY resulting in EITHER: SAFETY-RELATED STRUCTURE, SYSTEM, OR An electrical shock hazard that precludes access to operate COMPONENT required by Technical Specifications for the or monitor ANY SAFETY-RELATED STRUCTURE, current operating mode in ANY Table H-1 area Notes SYSTEM or COMPONENT within ANY Table H-1 area 1 1. The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time OR Control Room indication of degraded performance of ANY

2. The ED should not wait until the applicable time has elapsed, but should declare the event as soon SAFETY-RELATED STRUCTURE, SYSTEM or Natural or as it is determined that the release duration has exceeded, or will likely exceed, the applicable time. COMPONENT within ANY Table H-1 area Destructive Phenomena In the absence of data to the contrary, assume that the release duration has exceeded the applicable HA1.4 1 2 3 4 D HU1.4 1 2 3 4 D time if an ongoing release is detected and the release start time is unknown.

3. If loss of water level in the refueling pathway occurs while in Mode 3, 4 or D, consider classification Turbine failure-generated PROJECTILEs Turbine failure resulting in ANY of the following:

under EALs CU3.1, CU3.2 or CU3.3 resulting in EITHER:

• Casing penetration VISIBLE DAMAGE to or penetration of ANY SAFETY-

• Damage to turbine seals

4. The ED should not wait until the applicable time has elapsed, but should declare the event as soon RELATED STRUCTURE, SYSTEM or COMPONENT

• Damage to generator seals as it is determined that the condition has exceeded, or will likely exceed, the applicable time. within ANY Table H-1 area

5. If the equipment in the stated area was already inoperable, or out of service, before the event OR occurred, then EAL HA3.1 should not be declared as it will have no adverse impact on the ability of Control Room indication of degraded performance of the plant to safely operate or safely shutdown beyond that already allowed by Technical ANY SAFETY-RELATED STRUCTURE, SYSTEM or Specifications at the time of the event. COMPONENT within ANY Table H-1 area HA1.5 1 2 3 4 D HU1.5 1 2 3 4 D Lake water level > 254 ft Lake water level > 248.2 ft OR OR Intake water level < 236 ft Intake water level < 238.8 ft HA1.6 1 2 3 4 D Vehicle crash resulting in EITHER:

VISIBLE DAMAGE to ANY SAFETY-RELATED H STRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 area OR Control Room indication of degraded performance of ANY Hazards SAFETY-RELATED STRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 area Other HA2.1 1 2 3 4 D HU2.1 1 2 3 4 D Condi-tions FIRE or EXPLOSION FIRE not extinguished within 15 min. of Control Room resulting in EITHER: notification or verification of a Control Room FIRE alarm in Affect-ing 2 None None VISIBLE DAMAGE to ANY SAFETY-RELATED STRUCTURE, SYSTEM or COMPONENT within ANY Table ANY Table H-1 area, RadWaste Solidification and Storage Bldg, or Security West Bldg (Note 4)

Plant H-1 area Safety Fire or Explosion OR HU2.2 1 2 3 4 D Control Room indication of degraded performance of ANY SAFETY-RELATED STRUCTURE, SYSTEM or EXPLOSION of sufficient force to damage permanent COMPONENT within ANY Table H-1 area structures or equipment within the PROTECTED AREA HA3.1 1 2 3 4 D HU3.1 1 2 3 4 D 3 None None Access to ANY Table H-1 area is prohibited due to toxic, corrosive, asphyxiant or flammable gases which jeopardize operation of systems required to maintain safe operations or Toxic, corrosive, asphyxiant or flammable gases in amounts that have or could adversely affect NORMAL PLANT OPERATIONS Hazardous safely shutdown the reactor (Note 5)

Gas HU3.2 1 2 3 4 D Recommendation by local, county or state officials to evacuate or shelter site personnel based on an offsite event HG4.1 1 2 3 4 D HS4.1 1 2 3 4 D HA4.1 1 2 3 4 D HU4.1 1 2 3 4 D A HOSTILE ACTION has occurred such that plant A HOSTILE ACTION is occurring or has occurred within the A HOSTILE ACTION is occurring or has occurred within the A SECURITY CONDITION that does not involve a HOSTILE personnel are unable to operate equipment required to PROTECTED AREA as reported by the Security Site Owner Controlled Area as reported by the Security Site ACTION as reported by the Security Site Supervisor 4 maintain safety functions HG4.2 1 2 3 4 D Supervisor Supervisor OR OR A credible site-specific security threat notification A validated notification from NRC of an AIRLINER attack OR Security A HOSTILE ACTION has caused failure of Spent Fuel threat within 30 min. of the site A validated notification from NRC providing information of an Cooling systems aircraft threat AND IMMINENT fuel damage is likely HS5.1 1 2 3 4 D HA5.1 1 2 3 4 D 5 Control Room evacuation has been initiated Control Room evacuation has been initiated Control None AND None Room Control of the plant cannot be established within 15 min.

Evacuation HG6.1 1 2 3 4 D HS6.1 1 2 3 4 D HA6.1 1 2 3 4 D HU6.1 1 2 3 4 D Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial have occurred which involve actual or likely major failures of have occurred which involve an actual or potential have occurred which indicate a potential degradation of the core degradation or melting with potential for loss of plant functions needed for protection of the public or substantial degradation of the level of safety of the plant or a level of safety of the plant or indicate a security threat to containment integrity or HOSTILE ACTION that results in an HOSTILE ACTION that results in intentional damage or security event that involves probable life threatening risk to facility protection has been initiated. No releases of actual loss of physical control of the facility. Releases can be malicious acts; (1) toward site personnel or equipment that site personnel or damage to site equipment because of radioactive material requiring offsite response or monitoring 6 reasonably expected to exceed EPA Protective Action could lead to the likely failure of or; (2) that prevent effective HOSTILE ACTION. ANY releases are expected to be limited are expected unless further degradation of safety systems Prepared for Constellation by: Operations Support Services, Inc. - www.ossi-net.com (Rev. 23 1/18/13)

Guideline exposure levels (1,000 mRem TEDE or 5,000 access to equipment needed for the protection of the public. to small fractions of the EPA Protective Action Guideline occurs Judgment mRem thyroid CDE) offsite for more than the immediate site ANY releases are not expected to result in exposure levels exposure levels (1,000 mRem TEDE or 5,000 mRem thyroid area which exceed EPA Protective Action Guideline exposure CDE) levels (1,000 mRem TEDE or 5,000 mRem thyroid CDE) beyond the SITE BOUNDARY EU1.1 1 2 3 4 D EAL Identifier XXX.X E None None None Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by measured dose rates > then ANY of the following:

• 400 mRem/hr at 3 feet from the HSM surface Category (R, H, E, S, F, C) Sequential number within subcategory/classification ISFSI

• 100 mRem/hr outside HSM door on centerline Emergency classification (G, S, A, U) Subcategory number (1 if no subcategory)

• 20 mRem/hr end shield wall exterior Modes:

EPIP-EPP-01-EAL 1 2 3 4 D Power Operation Hot Shutdown Cold Shutdown Refuel Defueled Nine Mile Point Nuclear Station Unit 1 Attachment 1 Revision 234 EAL Matrix Unit 1 Page 2 of 2 MODE 3, 4 or D

Attachment 2 Page 3 of 4 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT RG1.1 1 2 3 4 5 D RS1.1 1 2 3 4 5 D RA1.1 1 2 3 4 5 D RU1.1 1 2 3 4 5 D SG1.1 1 2 3 SS1.1 1 2 3 SA1.1 1 2 3 SU1.1 1 2 3 Loss of all offsite and all onsite AC power, Table S-1, to 4.16 Loss of all offsite and all onsite AC power, Table S-1, to AC power capability to 4.16 KV emergency buses Loss of all offsite AC power, Table S-1, to 4.16 KV ANY monitor reading > Table R-1 GE column ANY monitor reading > Table R-1 SAE column ANY gaseous monitor reading > Table R-1 Alert column ANY gaseous monitors > Table R-1 UE column KV emergency buses 2ENS*SWG101 and 2ENS*SWG103 4.16 KV emergency buses 2ENS*SWG101 and 2ENS*SWG101 and 2ENS*SWG103 reduced to a single emergency buses 2ENS*SWG101 and 2ENS*SWG103 for 15 min. (Note 1) for 15 min. (Note 1) for 15 min. (Note 2) for 60 min. (Note 2) AND EITHER: 2ENS*SWG103 for 15 min. (Note 4) power source, Table S-1, for 15 min. (Note 4)

• Do not delay declaration awaiting dose assessment results

• If dose assessment results are available, declaration

• Do not delay declaration awaiting dose assessment results

• If dose assessment results are available, declaration 1 Restoration of 4.16 KV emergency bus 2ENS*SWG101 or 2ENS*SWG103 within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely OR AND ANY additional single power source failure will result in a loss of all power to 4.16 KV emergency buses 2ENS*SWG101 Table S-1 AC Power Sources should be based on dose assessment instead of should be based on dose assessment instead of RA1.2 1 2 3 4 5 D RU1.2 1 2 3 4 5 D Loss of RPV water level cannot be restored and maintained and 2ENS*SWG103 radiation monitor values (see EAL RG1.2) radiation monitor values (see EAL RS1.2)

• 2EGS*EG1 AC above -14 in. or RPV water level cannot be determined ANY liquid monitor > Table R-1 Alert column ANY liquid monitor reading > Table R-1 UE column Onsite Power

• 2EGS*EG3 for 15 min. (Note 2) for 60 min. (Note 2)

• 2EGS*EG2 (with 2ENS*SWG102 crosstied to 2ENS*SWG101 or 2ENS*SWG103)

• Reserve Transformer A 2 Offsite RA1.3 1 2 3 4 5 D RU1.3 1 2 3 4 5 D SS2.1 1 2 3

• Reserve Transformer B 1 Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 200 x ODCM limits Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 2 x ODCM limits Loss of DC Power None < 105 VDC on both 2BYS*SWG002A and 2BYS*SWG002B for 15 min. (Note 4)

None

• Aux Boiler Transformer for 15 min. (Note 2) for 60 min. (Note 2)

Offsite Rad SG3.1 1 2 SS3.1 1 2 SA3.1 1 2 SU3.1 3 Conditions RG1.2 1 2 3 4 5 D RS1.2 1 2 3 4 5 D An automatic scram fails to shut down the reactor An automatic scram failed to shut down the reactor as An automatic scram failed to shut down the reactor An UNPLANNED sustained positive period observed on as indicated by reactor power > 4% indicated by reactor power > 4% AND nuclear instrumentation Dose assessment using actual meteorology indicates doses

> 1,000 mRem TEDE or 5,000 mRem thyroid CDE at or Dose assessment using actual meteorology indicates doses

> 100 mRem TEDE or 500 mRem thyroid CDE at or beyond None None 3 AND All manual actions fail to shut down the reactor as indicated by reactor power > 4%

AND Manual actions taken at the reactor control console (mode Manual actions taken at the reactor control console (mode switch in shutdown, manual scram push buttons or beyond the SITE BOUNDARY the SITE BOUNDARY switch in shutdown, manual scram push buttons and ARI) ARI) successfully shut down the reactor as indicated by Criticality AND EITHER of the following exist or have occurred: failed to shut down the reactor as indicated by reactor reactor power 4%

& RPV water level cannot be restored and maintained power > 4%

RPS above -39 in. or RPV water level cannot be determined RG1.3 1 2 3 4 5 D RS1.3 1 2 3 4 5 D OR Failure Suppression pool temperature and RPV pressure R Field survey results indicate closed window dose rates

> 1,000 mRem/hr expected to continue for 60 min. at or beyond the SITE BOUNDARY (Note 1)

Field survey results indicate closed window dose rates

> 100 mRem/hr expected to continue for 60 min. at or beyond the SITE BOUNDARY (Note 1)

None None cannot be maintained below the Heat Capacity Temperature Limit (N2-EOP-PC Figure M)

OR OR 4

Abnorm. SU4.1 1 2 3 Analyses of field survey samples indicate thyroid CDE Analyses of field survey samples indicate thyroid CDE Rad Release

/ Rad

> 5,000 mRem for 1 hr of inhalation at or beyond the SITE BOUNDARY (Note 1)

> 500 mRem for 1 hr of inhalation at or beyond the SITE BOUNDARY (Note 1)

S Inability to Reach or Maintain None None None Plant is not brought to required operating mode within Technical Specifications LCO required action completion time Effluent RA2.1 1 2 3 4 5 D RU2.1 1 2 3 4 5 D Shutdown Table R-1 Effluent Monitor Classification Thresholds Conditions Alarm on ANY of the following radiation monitors due to UNPLANNED water level drop in a reactor refueling pathway System Monitor GE SAE ALERT UE damage to irradiated fuel or loss of water level: as indicated by inability to restore and maintain SFP level > Malfunct.

• 2RMS-RE111 SS5.1 1 2 3 SA5.1 1 2 3 SU5.1 1 2 3 low water level alarm (Note 3)

• 2RMS-RE112 AND GASEOUS Radwaste/RB Vent Effluent 5.5E+7 µCi/s 5.5E+6 µCi/s 200 x Alarm 2 x Alarm Loss of > approximately 75% of annunciation or indication UNPLANNED loss of > approximately 75% of annunciation or UNPLANNED loss of > approximately 75% of annunciation or

• 2RMS-RE113 Area radiation monitor reading rise on ANY of the following:

2 Main Stack Effluent 1.0E+10 µCi/s 1.0E+9 µCi/s 200 x Alarm 2 x Alarm

• 2RMS-RE114

• 2RMS-RE140

• 2HVR*RE14A

• 2RMS-RE111

• 2RMS-RE112

• 2RMS-RE113 Table S-2 Significant Transients on all of the following Control Room panels for 15 min.

(Note 4):

• 2CEC*PNL601 indication on all of the following Control Room panels for 15 min. (Note 4):

• 2CEC*PNL601 indication on all of the following Control Room panels for 15 min. (Note 4):

• 2CEC*PNL601 Onsite Rad

• 2CEC*PNL602

• 2CEC*PNL602

• 2CEC*PNL602 5

• 2HVR*RE14B

• 2RMS-RE114 Conditions

• 2RMS-RE140

• Automatic turbine runback > 25% thermal reactor power

• 2CEC*PNL603

• 2CEC*PNL603

• 2CEC*PNL603 Service Water Effluent N/A N/A 200 x DRMS High (red) 2 x DRMS High (red)

• 2CEC*PNL851

&

• Electric load rejection > 25% full electrical load

• 2CEC*PNL851

• 2CEC*PNL851 Spent Fuel

• 2CEC*PNL852

• 2CEC*PNL852

• 2CEC*PNL852 Inst.

LIQUID Events

• Reactor scram AND AND EITHER:

Liquid RadWaste Effluent N/A N/A N/A 2 x DRMS High (red)

• ECCS injection A significant transient is in progress, Table S-2 A significant transient is in progress, Table S-2 RA2.2 1 2 3 4 5 D RU2.2 1 2 3 4 5 D

• Thermal power oscillations > 10% AND OR Compensatory indications are unavailable (Plant Computer, Compensatory indications are unavailable (Plant Process Cooling Tower Blowdown N/A N/A 200 x DRMS High (red) 2 x DRMS High (red)

A water level drop in a reactor refueling pathway that will UNPLANNED area radiation readings rise by a factor of SPDS) Computer, SPDS) result in irradiated fuel becoming uncovered 1,000 over NORMAL LEVELS Table S-3 Communications Systems SU6.1 1 2 3 RA3.1 1 2 3 4 5 D 3 Dose rates > 15 mRem/hr in EITHER of the following areas requiring continuous occupancy to maintain plant safety 6 None System Onsite Offsite (internal) (external)

Loss of all Table S-3 onsite (internal) communication methods affecting the ability to perform routine operations OR None None functions: None Comm. Loss of all Table S-3 offsite (external) communication CR/CAS Control Room PBX (normal dial telephones) X X methods affecting the ability to perform offsite notifications Rad OR CAS Gaitronics X SU7.1 1 2 3 HA1.1 1 2 3 4 5 D HU1.1 1 2 3 4 5 D 7 Station radio (portable)

Control Room installed satellite phones (non portable)

X X Reactor coolant activity > 4 µCi/gm I-131 Equivalent Fuel Clad None ENS X Seismic event > OBE (0.075g) Seismic event identified by ANY two of the following:

Degradation X SU7.2 1 2 3 as indicated by EITHER:

• Annunciator 842121 SEISMIC ACCELERATION RECS Computer Point ERSNC02, OBE Detected EXCEEDED indicates seismic event detected OR Offgas radiation DRMS high (red) alarm for 15 min.

• Confirmation of earthquake received on NMP-1 or Table H-1 Safe Shutdown Areas ANY amber LED light lit at the Seismic Monitor Panel, JAFNPP seismic instrumentation Response Spectrum Annunciator SU8.1 1 2 3

• Earthquake felt in plant 8

• Reactor Building (including Primary Containment) AND Earthquake confirmed by ANY of the following: Unidentified or reactor coolant pressure boundary

• Control Room

• Earthquake felt in plant leakage > 10 gpm None None None

• Diesel Generator Engine and Board Rooms

• JAFNPP seismic instrumentation RCS OR

• Control Room indication of degraded performance of Leakage Identified reactor coolant leakage > 25 gpm

• Standby Switchgear and Battery Rooms systems required for the safe shutdown of the plant

• HPCS Switchgear and Battery Rooms HA1.2 1 2 3 4 5 D HU1.2 1 2 3 4 5 D F

• Remote Shutdown Rooms FG1.1 1 2 3 FS1.1 1 2 3 FA1.1 1 2 3 FU1.1 1 2 3 Tornado striking Tornado striking within PROTECTED AREA boundary

• Control Building HVAC Rooms OR OR Sustained high winds > 90 mph Loss of ANY two fission product barriers Loss or potential loss of ANY two fission product barriers ANY loss or ANY potential loss of EITHER Fuel Clad ANY loss or ANY potential loss of Containment barrier

• Service Water Pump Rooms Sustained high winds > 90 mph Fission Product resulting in EITHER: AND (Table F-1) barrier OR RCS barrier (Table F-1) (Table F-1)

• Electrical Protection Assembly Room Barrier Loss or potential loss of third fission product barrier VISIBLE DAMAGE to ANY SAFETY-RELATED STRUCTURE, SYSTEM, OR COMPONENT within ANY Degradation (Table F-1)

• PGCC Relay Room Table H-1 area OR Control Room indication of degraded performance of ANY SAFETY-RELATED STRUCTURE, SYSTEM, OR COMPONENT within ANY Table H-1 area Table F-1 Fission Product Barrier Matrix HA1.3 1 2 3 4 5 D HU1.3 1 2 3 4 5 D Notes Internal flooding Internal flooding that has the potential to affect ANY Fuel Clad Barrier Reactor Coolant System Barrier Containment Barrier

1. The ED should not wait until the applicable time has elapsed, but should declare the event as soon resulting in EITHER: SAFETY-RELATED STRUCTURE, SYSTEM, OR as it is determined that the condition will likely exceed the applicable time An electrical shock hazard that precludes access to operate COMPONENT required by Technical Specifications for the Loss Potential Loss Loss Potential Loss Loss Potential Loss

2. The ED should not wait until the applicable time has elapsed, but should declare the event as soon or monitor ANY SAFETY-RELATED STRUCTURE, current operating mode in ANY Table H-1 area SYSTEM or COMPONENT within ANY Table H-1 area 1. Primary Containment 1. RPV water level cannot be 1. RPV water level cannot be 1. Primary Containment 1

as it is determined that the release duration has exceeded, or will likely exceed, the applicable time.

In the absence of data to the contrary, assume that the release duration has exceeded the applicable OR Flooding is required restored and maintained restored and maintained Flooding is required Control Room indication of degraded performance of ANY A

time if an ongoing release is detected and the release start time is unknown. above -14 in. following above -14 in. or RPV water SAFETY-RELATED STRUCTURE, SYSTEM or depressurization of the RPV level cannot be determined Natural or 3. If loss of water level in the refueling pathway occurs while in Mode 4, 5 or D, consider classification COMPONENT within ANY Table H-1 area or RPV water level cannot None None Destructive under EALs CU3.1, CU3.2 or CU3.3 RPV Water 1 2 3 4 5 D HU1.4 1 2 3 4 5 D be determined Phenomena HA1.4

4. The ED should not wait until the applicable time has elapsed, but should declare the event as soon Level as it is determined that the condition has exceeded, or will likely exceed, the applicable time. Turbine failure-generated PROJECTILEs Turbine failure resulting in ANY of the following:

5. If the equipment in the stated area was already inoperable, or out of service, before the event resulting in EITHER:

• Casing penetration

2. Primary Containment 1. Primary Containment 2. Primary Containment occurred, then EAL HA3.1 should not be declared as it will have no adverse impact on the ability of VISIBLE DAMAGE to or penetration of ANY SAFETY-

• Damage to turbine seals pressure > 1.68 psig due to pressure rise followed by a pressure > 45 psig and the plant to safely operate or safely shutdown beyond that already allowed by Technical RELATED STRUCTURE, SYSTEM or COMPONENT

• Damage to generator seals RCS leakage rapid UNPLANNED drop in rising Specifications at the time of the event. within ANY Table H-1 area Primary Containment 3. Explosive mixture exists OR Control Room indication of degraded performance of ANY SAFETY-RELATED STRUCTURE, SYSTEM or B pressure

2. Primary Containment inside Primary Containment

( 6% H2 and 5% O2)

Primary None None None COMPONENT within ANY Table H-1 area pressure response not 4. Suppression pool Containment HA1.5 1 2 3 4 5 D HU1.5 1 2 3 4 5 D consistent with LOCA temperature and RPV Pressure /

conditions pressure cannot be Temperature Lake water level > 254 ft Lake water level > 248.2 ft maintained below the Heat OR OR Capacity Temperature Limit Intake water level < 233 ft Intake water level < 237 ft (N2-EOP-PC Figure M)

HA1.6 1 2 3 4 5 D 3. Release pathway exists 1. UNISOLABLE primary 3. Failure of all Primary outside Primary system leakage outside Containment isolation Vehicle crash H

Containment resulting from Primary Containment as valves in ANY one line to resulting in EITHER: isolation failure in ANY of indicated by exceeding close following auto or VISIBLE DAMAGE to ANY SAFETY-RELATED the following systems EITHER: manual initiation STRUCTURE, SYSTEM or COMPONENT within ANY Table (excluding normal process RB area temperature AND H-1 area system flowpaths from an above an isolation Direct downstream pathway Hazards OR UNISOLABLE system): setpoint outside Primary Control Room indication of degraded performance of ANY

• Main steam line OR Containment and to the

& SAFETY-RELATED STRUCTURE, SYSTEM or Other

• RCIC steam line RB area radiation above environment exists COMPONENT within ANY Table H-1 area

• RWCU an alarm setpoint Condi-

• Feedwater 4. Intentional Primary tions Affect-HA2.1 1 FIRE or EXPLOSION 2 3 4 5 D HU2.1 1 2 3 4 5 D FIRE not extinguished within 15 min. of Control Room C None None

4. RPV blowdown is required Containment venting per EOPs None 2

ing resulting in EITHER: notification or verification of a Control Room FIRE alarm in Isolation Plant VISIBLE DAMAGE to ANY SAFETY-RELATED ANY Table H-1 area or Turbine Building (Note 4) 5. UNISOLABLE primary None None STRUCTURE, SYSTEM or COMPONENT within ANY Table system leakage outside Safety Primary Containment as H-1 area HU2.2 1 2 3 4 5 D Fire or indicated by exceeding OR Explosion EITHER:

Control Room indication of degraded performance of ANY EXPLOSION of sufficient force to damage permanent RB area maximum safe SAFETY-RELATED STRUCTURE, SYSTEM or structures or equipment within the PROTECTED AREA temperature value COMPONENT within ANY Table H-1 area (N2-EOP-SC Detail S)

OR HA3.1 1 2 3 4 5 D HU3.1 1 2 3 4 5 D RB area radiation >

3 8.00E+3 mR/hr Access to ANY Table H-1 area is prohibited due to toxic, Toxic, corrosive, asphyxiant or flammable gases in amounts corrosive, asphyxiant or flammable gases which jeopardize that have or could adversely affect NORMAL PLANT None None operation of systems required to maintain safe operations or OPERATIONS Hazardous 2. Drywell area radiation 5. Drywell area radiation 5. Drywell area radiation safely shutdown the reactor (Note 5)

Gas HU3.2 1 2 3 4 5 D 3100 R/hr (3.1 E6 41 R/hr 6.0 E4 R/hr D

mRem/hr) (4.1 E4 mRem/hr) (6.0 E7 mRem/hr)

Recommendation by local, county or state officials to None None None evacuate or shelter site personnel based on an offsite event 3. Reactor coolant activity Rad > 300 µCi/gm I-131 HG4.1 1 2 3 4 5 D HS4.1 1 2 3 4 5 D HA4.1 1 2 3 4 5 D HU4.1 1 2 3 4 5 D Equivalent A HOSTILE ACTION has occurred such that plant A HOSTILE ACTION is occurring or has occurred within the A HOSTILE ACTION is occurring or has occurred within the A SECURITY CONDITION that does not involve a HOSTILE personnel are unable to operate equipment required to PROTECTED AREA as reported by the Security Site Owner Controlled Area as reported by the Security Site ACTION as reported by the Security Site Supervisor 4. ANY condition in the 2. ANY condition in the 6. ANY condition in the 2. ANY condition in the opinion 6. ANY condition in the opinion 6. ANY condition in the opinion 4 maintain safety functions HG4.2 1 2 3 4 5 D Supervisor Supervisor OR OR A credible site-specific security threat notification E opinion of the Emergency Director that indicates loss of the Fuel Clad barrier opinion of the Emergency Director that indicates potential loss of the Fuel opinion of the Emergency Director that indicates loss of the Reactor Coolant of the Emergency Director that indicates potential loss of the Reactor Coolant of the Emergency Director that indicates loss of the Containment barrier of the Emergency Director that indicates potential loss of the Containment barrier A validated notification from NRC of an AIRLINER attack OR Judgment Clad barrier System barrier System barrier Security A HOSTILE ACTION has caused failure of Spent Fuel threat within 30 min. of the site A validated notification from NRC providing information of an Cooling systems aircraft threat AND IMMINENT fuel damage is likely 5

HS5.1 1 2 3 4 5 D HA5.1 1 2 3 4 5 D Control Room evacuation has been initiated Control Room evacuation has been initiated Control None AND None Room Control of the plant cannot be established within 15 min.

Evacuation HG6.1 1 2 3 4 5 D HS6.1 1 2 3 4 5 D HA6.1 1 2 3 4 5 D HU6.1 1 2 3 4 5 D Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions 1 exist 2 which3 in the 4 judgment 5 ofDthe Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial have occurred which involve actual or likely major failures of have occurred which involve an actual or potential have occurred which indicate a potential degradation of the core degradation or melting with potential for loss of plant functions needed for protection of the public or substantial degradation of the level of safety of the plant or a level of safety of the plant or indicate a security threat to containment integrity or HOSTILE ACTION that results in an HOSTILE ACTION that results in intentional damage or security event that involves probable life threatening risk to facility protection has been initiated. No releases of actual loss of physical control of the facility. Releases can be malicious acts; (1) toward site personnel or equipment that site personnel or damage to site equipment because of 6

radioactive material requiring offsite response or monitoring reasonably expected to exceed EPA Protective Action could lead to the likely failure of or; (2) that prevent effective HOSTILE ACTION. ANY releases are expected to be limited are expected unless further degradation of safety systems Prepared for Constellation by: Operations Support Services, Inc. - www.ossi-net.com (Rev. 22 1/18/13)

Guideline exposure levels (1,000 mRem TEDE or 5,000 access to equipment needed for the protection of the public. to small fractions of the EPA Protective Action Guideline occurs mRem thyroid CDE) offsite for more than the immediate site ANY releases are not expected to result in exposure levels exposure levels (1,000 mRem TEDE or 5,000 mRem thyroid Judgment area which exceed EPA Protective Action Guideline exposure CDE) levels (1,000 mRem TEDE or 5,000 mRem thyroid CDE) beyond the SITE BOUNDARY EU1.1 1 2 3 4 5 D EAL Identifier E None None None Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by measured dose rates > then ANY of the following:

• 400 mRem/hr at 3 feet from the HSM surface

• 100 mRem/hr outside HSM door on centerline XXX.X Category (R, H, E, S, F, C) Sequential number within subcategory/classification ISFSI

• 20 mRem/hr end shield wall exterior Emergency classification (G, S, A, U) Subcategory number (1 if no subcategory)

MODE 1, 2 or 3 EPIP-EP-02-EAL Modes: 1 Power Operation 2

Startup 3

Hot Shutdown 4

Cold Shutdown 5

Refuel D

Defueled Nine Mile Point Nuclear Station Unit 2 Attachment 1 Revision 223 EAL Matrix Unit 2 Page 1 of 2

Attachment 2 Page 4 of 4 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT RG1.1 1 2 3 4 5 D RS1.1 1 2 3 4 5 D RA1.1 1 2 3 4 5 D RU1.1 1 2 3 4 5 D CA1.1 4 5 D CU1.1 4 5 D ANY monitor reading > Table R-1 GE column for 15 min. (Note 1)

ANY monitor reading > Table R-1 SAE column for 15 min. (Note 1)

ANY gaseous monitor reading > Table R-1 Alert column for 15 min. (Note 2)

ANY gaseous monitors > Table R-1 UE column for 60 min. (Note 2) 1 Table C-1 AC Power Sources Loss of all offsite and all onsite AC power, Table C-1, to 4.16 KV emergency buses 2ENS*SWG101 and 2ENS*SWG103 AC power capability to 4.16 KV emergency buses 2ENS*SWG101 and 2ENS*SWG103 reduced to a single power source, Table C-1, for 15 min. (Note 4)

Onsite

• Do not delay declaration awaiting dose assessment

• Do not delay declaration awaiting dose assessment Loss of

• 2EGS*EG1 for 15 min. (Note 4)

AND results results AC

• 2EGS*EG3 ANY additional single power source failure will result in a loss

• If dose assessment results are available, declaration

• If dose assessment results are available, declaration of all power to 4.16 KV emergency buses 2ENS*SWG101 Power should be based on dose assessment instead of should be based on dose assessment instead of RA1.2 1 2 3 4 5 D RU1.2 1 2 3 4 5 D and 2ENS*SWG103 radiation monitor values (see EAL RG1.2) radiation monitor values (see EAL RS1.2)

• Reserve Transformer A Offsite ANY liquid monitor > Table R-1 Alert column ANY liquid monitor reading > Table R-1 UE column

• Reserve Transformer B 2

CU2.1 4 5 for 15 min. (Note 2) for 60 min. (Note 2)

• Aux Boiler Transformer None < 105 VDC on required 125 VDC emergency buses Loss of for 15 min. (Note 4)

DC Power RA1.3 1 2 3 4 5 D RU1.3 1 2 3 4 5 D CG3.1 4 5 CS3.1 4 5 CA3.1 4 5 CU3.1 4 1 Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 200 x ODCM limits Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 2 x ODCM limits RPV level < -14 in. for 30 min. (Note 4)

AND With CONTAINMENT CLOSURE not established, RPV water level < 11.8 in.

RPV water level < 17.8 in.

OR RCS leakage results in the inability to maintain or restore RPV water level > 159.3 in. for 15 min. (Note 4) for 15 min. (Note 2) for 60 min. (Note 2)

Offsite Rad ANY Containment Challenge Indication, Table C-3 RPV water level cannot be monitored for 15 min. with ANY Conditions UNPLANNED RPV leakage indication, Table C-2 (Note 4)

RG1.2 1 2 3 4 5 D RS1.2 1 2 3 4 5 D CG3.2 4 5 CS3.2 4 5 Dose assessment using actual meteorology indicates doses Dose assessment using actual meteorology indicates doses CU3.2 5

> 1,000 mRem TEDE or 5,000 mRem thyroid CDE at or > 100 mRem TEDE or 500 mRem thyroid CDE at or beyond None None RPV water level cannot be monitored with core uncovery With CONTAINMENT CLOSURE established, RPV water beyond the SITE BOUNDARY the SITE BOUNDARY UNPLANNED RPV water level drop below EITHER of the indicated by ANY of the following for 30 min. (Note 4): level

• ANY UNPLANNED RPV leakage indication, Table C-2 < -14 in. following for 15 min. (Note 4):

• Erratic Source Range Monitor indication

• 364 in. (RPV flange)

RG1.3 1 2 3 4 5 D RS1.3 1 2 3 4 5 D CS3.3 4 5

• RPV water level band (when the RPV water level band AND is established below the RPV flange)

R ANY Containment Challenge Indication, Table C-3 RPV water level cannot be monitored for 30 min. (Note 4)

Field survey results indicate closed window dose rates Field survey results indicate closed window dose rates with a loss of RPV inventory as indicated by ANY of the

> 1,000 mRem/hr expected to continue for 60 min. at or beyond the SITE BOUNDARY (Note 1)

OR

> 100 mRem/hr expected to continue for 60 min. at or beyond the SITE BOUNDARY (Note 1)

OR None None 3 following:

• ANY UNPLANNED RPV leakage indication, Table C-2 CU3.3 5 Abnorm.

C Analyses of field survey samples indicate thyroid CDE Analyses of field survey samples indicate thyroid CDE

• Erratic Source Range Monitor indication RPV Rad > 5,000 mRem for 1 hr of inhalation at or beyond the SITE > 500 mRem for 1 hr of inhalation at or beyond the SITE Water RPV water level cannot be monitored with a loss of Release BOUNDARY (Note 1) BOUNDARY (Note 1) RPV inventory as indicated by ANY UNPLANNED RPV Level leakage indication, Table C-2

/ Rad Effluent RA2.1 1 2 3 4 5 D RU2.1 1 2 3 4 5 D Cold SD/ Table C-2 RPV Leakage Indications Table R-1 Effluent Monitor Classification Thresholds Refuel Alarm on ANY of the following radiation monitors due to UNPLANNED water level drop in a reactor refueling pathway System

• Drywell equipment drain sump level rise Monitor GE SAE ALERT UE damage to irradiated fuel or loss of water level: as indicated by inability to restore and maintain SFP level > Malfunct.

• Drywell floor drain sump level rise

• 2RMS-RE111 low water level alarm (Note 3)

• 2RMS-RE112 AND

• Reactor building equipment sump level rise GASEOUS Radwaste/RB Vent Effluent 5.5E+7 µCi/s 5.5E+6 µCi/s 200 x Alarm 2 x Alarm

• 2RMS-RE113 Area radiation monitor reading rise on ANY of the following:

• Reactor Building floor drain sump level rise 2 Main Stack Effluent 1.0E+10 µCi/s 1.0E+9 µCi/s 200 x Alarm 2 x Alarm

• 2RMS-RE114

• 2RMS-RE140

• 2HVR*RE14A

• 2RMS-RE111

• 2RMS-RE112

• 2RMS-RE113

• Suppression Pool level rise

• UNPLANNED rise in RPV make-up rate Onsite Rad

• 2HVR*RE14B

• 2RMS-RE114 Conditions Service Water Effluent N/A N/A 200 x DRMS High (red) 2 x DRMS High (red)

• 2RMS-RE140

• Observation of UNISOLABLE RCS leakage Spent Fuel LIQUID Events Liquid RadWaste Effluent N/A N/A N/A 2 x DRMS High (red) 1 2 3 4 5 D 1 2 3 4 5 D Table C-3 Containment Challenge RA2.2 RU2.2 CA4.1 4 5 CU4.1 4 5 Cooling Tower Blowdown N/A N/A 200 x DRMS High (red) 2 x DRMS High (red) Indications A water level drop in a reactor refueling pathway that will result in irradiated fuel becoming uncovered UNPLANNED area radiation readings rise by a factor of 1,000 over NORMAL LEVELS 4

• CONTAINMENT CLOSURE not established

• Explosive mixture exists inside Primary Containment An UNPLANNED event results in EITHER:

RCS temperature > 200°F for > Table C-4 duration OR UNPLANNED event results in RCS temperature > 200°F RCS CU4.2 4 5 RA3.1 1 2 3 4 (H2 6% and O2 5%) RPV pressure increase > 10 psi due to an UNPLANNED 5 D Temp. loss of decay heat removal capability Loss of all RCS temperature and RPV water level 3 None None Dose rates > 15 mRem/hr in EITHER of the following areas requiring continuous occupancy to maintain plant safety functions:

None

• UNPLANNED rise in Primary Containment pressure

• RB area radiation > 8.00E+3 mR/hr indication for 15 min. (Note 4)

CR/CAS Control Room CU5.1 4 5 Rad OR CAS 5 Table C-4 RCS Reheat Duration Table C-5 Communications Systems An UNPLANNED sustained positive period observed on nuclear instrumentation Inadvertent HA1.1 1 2 3 4 5 D HU1.1 1 2 3 4 5 D Criticality Thresholds Onsite Offsite System Seismic event > OBE (0.075g) Seismic event identified by ANY two of the following:

• If an RCS heat removal system is in operation within this time (internal) (external)

CU6.1 4 5 D as indicated by EITHER:

• Annunciator 842121 SEISMIC ACCELERATION frame and RCS temperature is being reduced, the EAL is not 6

Computer Point ERSNC02, OBE Detected EXCEEDED indicates seismic event detected applicable PBX (normal dial telephones) X X OR Loss of all Table C-5 onsite (internal) communication methods

• Confirmation of earthquake received on NMP-1 or CONTAINMENT Gaitronics X affecting the ability to perform routine operations ANY amber LED light lit at the Seismic Monitor Panel, RCS Status Duration Table H-1 Safe Shutdown Areas Response Spectrum Annunciator JAFNPP seismic instrumentation Comm.

CLOSURE Status Station radio (portable) X OR

• Earthquake felt in plant Loss of all Table C-5 offsite (external) communication AND Control Room installed satellite phones (non portable) X INTACT N/A 60 min.* methods affecting the ability to perform offsite notifications

• Reactor Building (including Primary Containment) Earthquake confirmed by ANY of the following:

• Earthquake felt in plant ENS X

• Control Room

• JAFNPP seismic instrumentation Established 20 min.* RECS X

• Diesel Generator Engine and Board Rooms

• Control Room indication of degraded performance of Not INTACT systems required for the safe shutdown of the plant Not established 0 min.

• Standby Switchgear and Battery Rooms HA1.2 1 2 3 4 5 D HU1.2 1 2 3 4 5 D

• HPCS Switchgear and Battery Rooms

• Remote Shutdown Rooms Tornado striking Tornado striking within PROTECTED AREA boundary OR OR

• Control Building HVAC Rooms Sustained high winds > 90 mph Sustained high winds > 90 mph

• Service Water Pump Rooms resulting in EITHER:

VISIBLE DAMAGE to ANY SAFETY-RELATED

• Electrical Protection Assembly Room STRUCTURE, SYSTEM or COMPONENT within ANY Table

• PGCC Relay Room H-1 area OR Control Room indication of degraded performance of ANY SAFETY-RELATED STRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 area HA1.3 1 2 3 4 5 D HU1.3 1 2 3 4 5 D Notes Internal flooding Internal flooding that has the potential to affect ANY resulting in EITHER: SAFETY-RELATED STRUCTURE, SYSTEM, OR

1. The ED should not wait until the applicable time has elapsed, but should declare the event as soon An electrical shock hazard that precludes access to operate COMPONENT required by Technical Specifications for the as it is determined that the condition will likely exceed the applicable time or monitor ANY SAFETY-RELATED STRUCTURE, current operating mode in ANY Table H-1 area

2. The ED should not wait until the applicable time has elapsed, but should declare the event as soon SYSTEM or COMPONENT within ANY Table H-1 area 1

as it is determined that the release duration has exceeded, or will likely exceed, the applicable time. OR In the absence of data to the contrary, assume that the release duration has exceeded the applicable Control Room indication of degraded performance of ANY time if an ongoing release is detected and the release start time is unknown. SAFETY-RELATED STRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 area Natural or 3. If loss of water level in the refueling pathway occurs while in Mode 4, 5 or D, consider classification Destructive under EALs CU3.1, CU3.2 or CU3.3 HA1.4 1 2 3 4 5 D HU1.4 1 2 3 4 5 D Phenomena 4. The ED should not wait until the applicable time has elapsed, but should declare the event as soon Turbine failure-generated PROJECTILEs Turbine failure resulting in ANY of the following:

as it is determined that the condition has exceeded, or will likely exceed, the applicable time. resulting in EITHER:

• Casing penetration

5. If the equipment in the stated area was already inoperable, or out of service, before the event VISIBLE DAMAGE to or penetration of ANY SAFETY-

• Damage to turbine seals occurred, then EAL HA3.1 should not be declared as it will have no adverse impact on the ability of RELATED STRUCTURE, SYSTEM or COMPONENT

• Damage to generator seals the plant to safely operate or safely shutdown beyond that already allowed by Technical within ANY Table H-1 area Specifications at the time of the event. OR Control Room indication of degraded performance of ANY SAFETY-RELATED STRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 area HA1.5 1 2 3 4 5 D HU1.5 1 2 3 4 5 D Lake water level > 254 ft Lake water level > 248.2 ft OR OR Intake water level < 233 ft Intake water level < 237 ft HA1.6 1 2 3 4 5 D Vehicle crash H

resulting in EITHER:

VISIBLE DAMAGE to ANY SAFETY-RELATED STRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 area OR Hazards Control Room indication of degraded performance of ANY

& SAFETY-RELATED STRUCTURE, SYSTEM or Other COMPONENT within ANY Table H-1 area Condi-HA2.1 1 2 3 4 5 D HU2.1 1 2 3 4 5 D tions Affect- FIRE or EXPLOSION FIRE not extinguished within 15 min. of Control Room ing Plant Safety 2 None None resulting in EITHER:

VISIBLE DAMAGE to ANY SAFETY-RELATED STRUCTURE, SYSTEM or COMPONENT within ANY Table notification or verification of a Control Room FIRE alarm in ANY Table H-1 area or Turbine Building (Note 4)

H-1 area HU2.2 1 2 3 4 5 D Fire or Explosion OR Control Room indication of degraded performance of ANY EXPLOSION of sufficient force to damage permanent SAFETY-RELATED STRUCTURE, SYSTEM or structures or equipment within the PROTECTED AREA COMPONENT within ANY Table H-1 area HA3.1 1 2 3 4 5 D HU3.1 1 2 3 4 5 D 3 None None Access to ANY Table H-1 area is prohibited due to toxic, corrosive, asphyxiant or flammable gases which jeopardize operation of systems required to maintain safe operations or Toxic, corrosive, asphyxiant or flammable gases in amounts that have or could adversely affect NORMAL PLANT OPERATIONS Hazardous safely shutdown the reactor (Note 5)

Gas HU3.2 1 2 3 4 5 D Recommendation by local, county or state officials to evacuate or shelter site personnel based on an offsite event HG4.1 1 2 3 4 5 D HS4.1 1 2 3 4 5 D HA4.1 1 2 3 4 5 D HU4.1 1 2 3 4 5 D A HOSTILE ACTION has occurred such that plant A HOSTILE ACTION is occurring or has occurred within the A HOSTILE ACTION is occurring or has occurred within the A SECURITY CONDITION that does not involve a HOSTILE personnel are unable to operate equipment required to PROTECTED AREA as reported by the Security Site Owner Controlled Area as reported by the Security Site ACTION as reported by the Security Site Supervisor 4 maintain safety functions HG4.2 1 2 3 4 5 D Supervisor Supervisor OR OR A credible site-specific security threat notification A validated notification from NRC of an AIRLINER attack OR Security A HOSTILE ACTION has caused failure of Spent Fuel threat within 30 min. of the site A validated notification from NRC providing information of an Cooling systems aircraft threat AND IMMINENT fuel damage is likely 5

HS5.1 1 2 3 4 5 D HA5.1 1 2 3 4 5 D Control Room evacuation has been initiated Control Room evacuation has been initiated None None Control AND Room Control of the plant cannot be established within 15 min.

Evacuation HG6.1 1 2 3 4 5 D HS6.1 1 2 3 4 5 D HA6.1 1 2 3 4 5 D HU6.1 1 2 3 4 5 D Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial have occurred which involve actual or likely major failures of have occurred which involve an actual or potential have occurred which indicate a potential degradation of the 6 core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; (1) toward site personnel or equipment that substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring Judgment reasonably expected to exceed EPA Protective Action could lead to the likely failure of or; (2) that prevent effective HOSTILE ACTION. ANY releases are expected to be limited are expected unless further degradation of safety systems Prepared for Constellation by: Operations Support Services, Inc. - www.ossi-net.com (Rev. 22 1/18/13)

Guideline exposure levels (1,000 mRem TEDE or 5,000 access to equipment needed for the protection of the public. to small fractions of the EPA Protective Action Guideline occurs mRem thyroid CDE) offsite for more than the immediate site ANY releases are not expected to result in exposure levels exposure levels (1,000 mRem TEDE or 5,000 mRem thyroid area which exceed EPA Protective Action Guideline exposure CDE) levels (1,000 mRem TEDE or 5,000 mRem thyroid CDE) beyond the SITE BOUNDARY EU1.1 1 2 3 4 5 D EAL Identifier E None None None Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by measured dose rates > then ANY of the following:

• 400 mRem/hr at 3 feet from the HSM surface

• 100 mRem/hr outside HSM door on centerline XXX.X Category (R, H, E, S, F, C) Sequential number within subcategory/classification ISFSI Emergency classification (G, S, A, U) Subcategory number (1 if no subcategory)

• 20 mRem/hr end shield wall exterior MODE 4, 5 or D EPIP-EPP-02-EAL Modes: 1 Power Operation 2

Startup 3

Hot Shutdown 4

Cold Shutdown 5

Refuel D

Defueled Nine Mile Point Nuclear Station Unit 2 Attachment 1 Revision 223 EAL Matrix Unit 2 Page 2 of 2

ATTACHMENT 3 License Amendment Request Change Emergency Action Level HU1.5 to Remove High Lake Level Initiating Condition for Unusual Event Emergency Classification Clean Proposed Emergency Action Level Matrices Pages NMP1 EAL Matrices Pages 1

2 NMP2 EAL Matrices Pages 1

2

Attachment 3 Page 1 of 4 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT RS1.1 1 2 3 4 D RA1.1 1 2 3 4 D RU1.1 1 2 3 4 5 D D SG1.1 1 2 SS1.1 1 2 SA1.1 1 2 SU1.1 1 2 Loss of all offsite and all onsite AC power, Table S-1, to 4.16 Loss of all offsite and all onsite AC power, Table S-1, to AC power capability to 4.16 kV emergency buses reduced to Loss of all offsite AC power, Table S-1, to 4.16 kV ANY monitor reading > Table R-1 SAE column ANY gaseous monitor reading > Table R-1 Alert column ANY gaseous monitor reading > Table R-1 UE column kV emergency buses 4.16 KV emergency buses for 15 min. (Note 4) a single power source, Table S-1, for 15 min. (Note 4) emergency buses for 15 min. (Note 4) for 15 min. (Note 1) for 15 min. (Note 2) for 60 min. (Note 2) 1 AND EITHER: AND

• Do not delay declaration awaiting dose assessment Restoration of at least one 4.16 kV emergency bus ANY additional single power source failure will result in a loss results within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely of all 4.16 kV emergency bus power

• If dose assessment results are available, declaration RA1.2 1 2 3 4 D RU1.2 1 2 3 4 D OR Table S-1 AC Power Sources should be based on dose assessment instead of Loss of RPV water level cannot be restored and maintained radiation monitor values (see EAL RS1.2) AC Power above -84 in. or RPV water level cannot be determined Onsite ANY liquid monitor reading > Table R-1 Alert column ANY liquid monitor reading > Table R-1 UE column

• DG 102 for 15 min. (Note 2) for 60 min. (Note 2)

• DG 103 None

• T-101N Offsite RA1.3 1 2 3 4 D RU1.3 1 2 3 4 D

• T-101S 2

SS2.1 1 2

• T-10 backfed from offsite through T-1 or T-2 Confirmed sample analyses for gaseous or liquid releases Confirmed sample analyses for gaseous or liquid releases None < 106 VDC on both Battery Board 11 and Battery Board 12 None (only if already aligned) indicate concentrations or release rates > 200 x ODCM limits indicate concentrations or release rates > 2 x ODCM limits Loss of for 15 min. (Note 4) 1 for 15 min. (Note 2) for 60 min. (Note 2) DC Power SG3.1 1 SS3.1 1 SA3.1 1 SU3.1 2 Offsite Rad An automatic scram fails to shut down the reactor An automatic scram failed to shut down the reactor as An automatic scram failed to shut down the reactor An UNPLANNED sustained positive period observed on Conditions as indicated by reactor power > 6% indicated by reactor power > 6% AND nuclear instrumentation RG1.2 1 2 3 4 D Dose assessment using actual meteorology indicates doses RS1.2 1 2 3 4 D Dose assessment using actual meteorology indicates doses 3 AND All manual actions fail to shut down the reactor as indicated by reactor power > 6%

AND Manual actions taken at the reactor control console (mode Manual actions taken at the reactor control console (mode switch in shutdown, manual scram push buttons or switch in shutdown, manual scram push buttons and ARI) ARI) successfully shut down the reactor as indicated by

> 1,000 mRem TEDE or 5,000 mRem thyroid CDE at or > 100 mRem TEDE or 500 mRem thyroid CDE at or beyond None None Criticality AND EITHER of the following exist or have occurred: failed to shut down the reactor as indicated by reactor reactor power 6%

beyond the SITE BOUNDARY the SITE BOUNDARY & RPV water level cannot be restored and maintained power > 6%

RPS above -109 in. or RPV water level cannot be determined RG1.3 1 2 3 4 D RS1.3 1 2 3 4 D Failure OR Torus water temperature and RPV pressure cannot be R

Field survey results indicate closed window dose rates Field survey results indicate closed window dose rates maintained below the Heat Capacity Temperature Limit

> 1,000 mRem/hr expected to continue for 60 min. at or > 100 mRem/hr expected to continue for 60 min. at or (N1-EOP-4 Figure M) beyond the SITE BOUNDARY (Note 1) beyond the SITE BOUNDARY (Note 1) 4 OR OR None None SU4.1 1 2 S

Abnorm. Analyses of field survey samples indicate thyroid CDE Analyses of field survey samples indicate thyroid CDE Rad > 5,000 mRem for 1 hr of inhalation at or beyond the SITE > 500 mRem for 1 hr of inhalation at or beyond the SITE Inability to Plant is not brought to required operating mode within BOUNDARY (Note 1) BOUNDARY (Note 1) Reach or Technical Specifications LCO required action completion Release None None None

/ Rad Maintain time Shutdown Effluent Conditions System RA2.1 1 2 3 4 D RU2.1 1 2 3 4 D Malfunct.

SS5.1 1 2 SA5.1 1 2 SU5.1 1 2 Alarm on ANY of the following radiation monitors due to UNPLANNED water level drop in a reactor refueling pathway damage to irradiated fuel or loss of water level: as indicated by inability to restore and maintain SFP level > low Loss of > approximately 75% of annunciation or indication UNPLANNED loss of > approximately 75% of annunciation or UNPLANNED loss of > approximately 75% of annunciation or Table R-1 Effluent Monitor Classification Thresholds

• ARM 18 (West end of shield wall) water level alarm (Note 3) on Control Room panels L, K, H, F and G for 15 min. indication on Control Room panels L, K, H, F and G for 15 indication on Control Room panels L, K, H, F and G for 15

• ARM 25 (Rx building - east wall) AND Table S-2 Significant Transients (Note 4) min. (Note 4) min. (Note 4)

Monitor GE SAE ALERT UE

• ARM 29 (Refuel bridge (LOW RANGE)) Area radiation monitor reading rise on ANY of the following: AND AND EITHER:

• Refuel Bridge (HIGH RANGE)

• ARM 18 (West end of shield wall)

• Turbine runback > 25% thermal reactor power A significant transient is in progress, Table S-2 A significant transient is in progress, Table S-2 2

• Reactor Building Vent Radiation Monitor 5

GASEOUS Stack (RN 10A/B) N/A N/A 3.0E4 cps 300 cps

• ARM 25 (Rx building - east wall)

• Electric load rejection > 25% full electrical load AND OR

• ARM 29 (Refuel bridge (LOW RANGE)) Compensatory indications are unavailable (Plant Computer, Compensatory indications are unavailable (Plant

• Refuel Bridge (HIGH RANGE)

• Reactor scram SPDS)

Inst. Process Computer, SPDS)

Onsite Rad EC Vent N/A 300 mRem/hr 30 mRem/hr 10 mRem/hr

• ECCS injection Conditions

&

• Thermal power oscillations > 10%

Spent Fuel SW Effluent N/A N/A 90,000 cpm 900 cpm LIQUID Events RA2.2 1 2 3 4 D RU2.2 1 2 3 4 D RW Discharge N/A N/A 200 x batch 2 x batch A water level drop in a reactor refueling pathway that will UNPLANNED area radiation readings rise by a factor of SU6.1 1 2 result in irradiated fuel becoming uncovered 1,000 over NORMAL LEVELS Table S-3 Communications Systems 6 None None System Onsite Offsite (internal) (external)

Loss of all Table S-3 onsite (internal) communication methods affecting the ability to perform routine operations OR Comm. Loss of all Table S-3 offsite (external) communication RA3.1 1 2 3 4 D 3

PBX (normal dial telephones) X X methods affecting the ability to perform offsite notifications Dose rates > 15 mRem/hr in EITHER of the following areas Gaitronics X requiring continuous occupancy to maintain plant safety SU7.1 1 2 None None CR/CAS Rad functions:

Control Room OR 7 Hand-Held Portable Radio (station radio)

Control Room installed satellite phones (non portable)

X X

Reactor coolant activity > 4 µCi/gm I-131 Equivalent CAS Fuel Clad None ENS None X SU7.2 1 2 Degradation RECS X HA1.1 1 2 3 4 D HU1.1 1 2 3 4 D Offgas radiation monitor RN-12A or RN-12B hi-hi alarm for 15 min.

NMP-2 seismic instrumentation indicates > 0.075 g Seismic event identified by ANY two of the following:

Table H-1 Safe Shutdown Areas AND

• Annunciator H2-1-6 SEISMIC DETECTION SU8.1 1 2 Earthquake confirmed by ANY of the following: EQUIPMENT EVENT indicates seismic event detected

• Reactor Building (including Primary Containment)

• Control Room

• Earthquake felt in plant

• JAFNPP seismic instrumentation

• Control Room indication of degraded performance of

• Confirmation of earthquake received on NMP-2 or JAFNPP seismic instrumentation 8 None None None Unidentified drywell leakage > 10 gpm OR

• Earthquake felt in plant RCS Identified reactor coolant drywell leakage > 25 gpm

• Screenhouse systems required for the safe shutdown of the plant Leakage

• Turbine Building HA1.2 1 2 3 4 D HU1.2 1 2 3 4 D

• Battery Rooms Tornado striking Tornado striking within PROTECTED AREA boundary F

• Battery Board Rooms FG1.1 1 2 FS1.1 1 2 FA1.1 1 2 FU1.1 1 2 OR OR Sustained high winds > 90 mph Sustained high winds > 90 mph

• Cable Spreading Room Loss of ANY two fission product barriers Loss or potential loss of ANY two fission product barriers ANY loss or ANY potential loss of EITHER Fuel Clad ANY loss or ANY potential loss of Containment barrier resulting in EITHER: Fission Product

• Main Steam Isolation Valve Room VISIBLE DAMAGE to ANY SAFETY-RELATED AND (Table F-1) barrier OR RCS barrier (Table F-1) (Table F-1)

STRUCTURE, SYSTEM or COMPONENT within ANY Table Barrier Loss or potential loss of third fission product barrier

• Diesel Generator Engine and Board Rooms Degradation H-1 area (Table F-1)

• Security OR

• Central Alarm Station Control Room indication of degraded performance of ANY SAFETY-RELATED STRUCTURE, SYSTEM or

• Secondary Alarm Station COMPONENT within ANY Table H-1 area

• Security Uninterruptible Power Supply Room HA1.3 1 2 3 4 D HU1.3 1 2 3 4 D Table F-1 Fission Product Barrier Matrix Internal flooding Internal flooding that has the potential to affect ANY resulting in EITHER: SAFETY-RELATED STRUCTURE, SYSTEM, OR Fuel Clad Barrier Reactor Coolant System Barrier Containment Barrier An electrical shock hazard that precludes access to operate COMPONENT required by Technical Specifications for the Notes or monitor ANY SAFETY-RELATED STRUCTURE, current operating mode in ANY Table H-1 area Loss Potential Loss Loss Potential Loss Loss Potential Loss SYSTEM, OR COMPONENT within ANY Table H-1 area 1 1. The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time OR Control Room indication of degraded performance of ANY

1. Primary Containment Flooding is required

1. RPV water level cannot be 1. RPV water level cannot be restored and maintained restored and maintained

1. Primary Containment Flooding is required A

SAFETY-RELATED STRUCTURE, SYSTEM or Natural or 2. The ED should not wait until the applicable time has elapsed, but should declare the event as soon above -84 in. following above -84 in. or RPV water COMPONENT within ANY Table H-1 area Destructive as it is determined that the release duration has exceeded, or will likely exceed, the applicable time. depressurization of the RPV level cannot be determined Phenomena In the absence of data to the contrary, assume that the release duration has exceeded the applicable HA1.4 1 2 3 4 D HU1.4 1 2 3 4 D RPV Water or RPV water level cannot None None time if an ongoing release is detected and the release start time is unknown. Level be determined Turbine failure-generated PROJECTILEs Turbine failure resulting in ANY of the following:

3. If loss of water level in the refueling pathway occurs while in Mode 3, 4 or D, consider classification

• Casing penetration resulting in EITHER:

under EALs CU3.1, CU3.2 or CU3.3

• Damage to turbine seals VISIBLE DAMAGE to or penetration of ANY SAFETY-

4. The ED should not wait until the applicable time has elapsed, but should declare the event as soon RELATED STRUCTURE, SYSTEM or COMPONENT

• Damage to generator seals 2. Torus pressure > 35 psig as it is determined that the condition has exceeded, or will likely exceed, the applicable time. within ANY Table H-1 area 2. Primary Containment 1. Primary Containment and rising OR pressure > 3.5 psig due to pressure rise followed by a

5. If the equipment in the stated area was already inoperable, or out of service, before the event RCS leakage rapid UNPLANNED drop in B

occurred, then EAL HA3.1 should not be declared as it will have no adverse impact on the ability of Control Room indication of degraded performance of 3. Explosive mixture exists Primary Containment the plant to safely operate or safely shutdown beyond that already allowed by Technical ANY SAFETY-RELATED STRUCTURE, SYSTEM or inside Primary Containment pressure Specifications at the time of the event. COMPONENT within ANY Table H-1 area ( 6% H2 and 5% O2)

Primary None None None HA1.5 1 2 3 4 D HU1.5 1 2 3 4 D Containment Pressure / 2. Primary Containment

4. Torus water temperature Lake water level > 254 ft Intake water level < 238.8 ft Temperature pressure response not and RPV pressure cannot OR consistent with LOCA be maintained below the Intake water level < 236 ft conditions Heat Capacity Temperature Limit (N1-EOP-4 Figure M)

HA1.6 1 2 3 4 D

3. Release pathway exists 1. UNISOLABLE primary 3. Failure of all Primary Vehicle crash outside Pirmary system leakage outside Containment isolation resulting in EITHER: Containment resulting from Primary Containment as valves in ANY one line to VISIBLE DAMAGE to ANY SAFETY-RELATED isolation failure in ANY of indicated by exceeding close following auto or H STRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 area OR Control Room indication of degraded performance of ANY the following systems (excluding normal process system flowpaths from an UNISOLABLE system):

EITHER:

ANY N1-EOP-5 Detail T area temperature alarm setpoint manual initiation AND Direct downstream pathway outside Primary SAFETY-RELATED STRUCTURE, SYSTEM or

• Main steam line OR Containment and to the Hazards COMPONENT within ANY Table H-1 area

• EC steam line ANY N1-EOP-5 Detail R environment exists

&

• RWCU area radiation alarm Other HA2.1 1 2 3 4 D HU2.1 1 2 3 4 D

• Feedwater setpoint 4. Intentional Primary Condi-tions FIRE or EXPLOSION FIRE not extinguished within 15 min. of Control Room C None None Containment venting per EOPs None 2

Affect- resulting in EITHER: notification or verification of a Control Room FIRE alarm in 4. RPV blowdown is required Isolation 5. UNISOLABLE primary VISIBLE DAMAGE to ANY SAFETY-RELATED ANY Table H-1 area, RadWaste Solidification and Storage ing Bldg, or Security West Bldg (Note 4) system leakage outside None None STRUCTURE, SYSTEM or COMPONENT within ANY Table Plant H-1 area Primary Containment as Safety Fire or indicated by exceeding Explosion OR HU2.2 1 2 3 4 D EITHER:

Control Room indication of degraded performance of ANY Maximum safe general SAFETY-RELATED STRUCTURE, SYSTEM or EXPLOSION of sufficient force to damage permanent area temperature of COMPONENT within ANY Table H-1 area structures or equipment within the PROTECTED AREA 135ºF OR HA3.1 1 2 3 4 D HU3.1 1 2 3 4 D Maximum safe area radiation of 8 R/hr 3 Access to ANY Table H-1 area is prohibited due to toxic, corrosive, asphyxiant or flammable gases which jeopardize operation of systems required to maintain safe operations or Toxic, corrosive, asphyxiant or flammable gases in amounts that have or could adversely affect NORMAL PLANT OPERATIONS 2. Drywell radiation 5. Drywell radiation 80 R/hr 5. Drywell radiation Hazardous None None safely shutdown the reactor (Note 5) 3,000 R/hr 4.0 E4 R/hr Gas HU3.2 1 2 3 4 D Recommendation by local, county or state officials to D 3. Reactor coolant activity None None None evacuate or shelter site personnel based on an offsite event Rad > 300 µCi/gm I-131 Equivalent HG4.1 1 2 3 4 D HS4.1 1 2 3 4 D HA4.1 1 2 3 4 D HU4.1 1 2 3 4 D 4. ANY condition in the 2. ANY condition in the 6. ANY condition in the 2. ANY condition in the opinion 6. ANY condition in the opinion 6. ANY condition in the opinion A HOSTILE ACTION has occurred such that plant personnel are unable to operate equipment required to A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Site A HOSTILE ACTION is occurring or has occurred within the Owner Controlled Area as reported by the Security Site A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Security Site Supervisor E opinion of the Emergency Director that indicates loss opinion of the Emergency Director that indicates opinion of the Emergency Director that indicates loss of the Emergency Director that indicates potential loss of the Emergency Director that indicates loss of the of the Emergency Director that indicates potential loss 4

of the Fuel Clad barrier potential loss of the Fuel of the Reactor Coolant of the Reactor Coolant Containment barrier of the Containment barrier maintain safety functions Supervisor Supervisor OR Judgment Clad barrier System barrier System barrier HG4.2 1 2 3 4 D OR A credible site-specific security threat notification A validated notification from NRC of an AIRLINER attack OR Security A HOSTILE ACTION has caused failure of Spent Fuel threat within 30 min. of the site A validated notification from NRC providing information of an Cooling systems aircraft threat AND IMMINENT fuel damage is likely HS5.1 1 2 3 4 D HA5.1 1 2 3 4 D 5 Control Room evacuation has been initiated Control Room evacuation has been initiated Control None AND None Room Control of the plant cannot be established within 15 min.

Evacuation HG6.1 1 2 3 4 D HS6.1 1 2 3 4 D HA6.1 1 2 3 4 D HU6.1 1 2 3 4 D Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial have occurred which involve actual or likely major failures of have occurred which involve an actual or potential have occurred which indicate a potential degradation of the 6 core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; (1) toward site personnel or equipment that substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring Judgment reasonably expected to exceed EPA Protective Action could lead to the likely failure of or; (2) that prevent effective HOSTILE ACTION. ANY releases are expected to be limited are expected unless further degradation of safety systems Prepared for Constellation by: Operations Support Services, Inc. - www.ossi-net.com (Rev. 23 1/18/13)

Guideline exposure levels (1,000 mRem TEDE or 5,000 access to equipment needed for the protection of the public. to small fractions of the EPA Protective Action Guideline occurs mRem thyroid CDE) offsite for more than the immediate site ANY releases are not expected to result in exposure levels exposure levels (1,000 mRem TEDE or 5,000 mRem thyroid area which exceed EPA Protective Action Guideline exposure CDE) levels (1,000 mRem TEDE or 5,000 mRem thyroid CDE) beyond the SITE BOUNDARY EU1.1 1 2 3 4 D EAL Identifier E None None None Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by measured dose rates > then ANY of the following:

• 400 mRem/hr at 3 feet from the HSM surface

• 100 mRem/hr outside HSM door on centerline XXX.X Category (R, H, E, S, F, C) Sequential number within subcategory/classification ISFSI

• 20 mRem/hr end shield wall exterior Emergency classification (G, S, A, U) Subcategory number (1 if no subcategory)

MODE 1 or 2 EPIP-EPP-01-EAL Modes: 1 Power Operation 2

Hot Shutdown 3

Cold Shutdown 4

Refuel D

Defueled Nine Mile Point Nuclear Station Unit 1 Revision 24 EAL Matrix Unit 1 Page 1 of 2

Attachment 3 Page 2 of 4 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT RS1.1 1 2 3 4 D RA1.1 1 2 3 4 D RU1.1 1 2 3 4 5 D D CA1.1 3 4 D CU1.1 3 4 D ANY monitor reading > Table R-1 SAE column for 15 min. (Note 1)

ANY gaseous monitor reading > Table R-1 Alert column for 15 min. (Note 2)

ANY gaseous monitor reading > Table R-1 UE column for 60 min. (Note 2) 1 Table C-1 AC Power Sources Loss of all offsite and all onsite AC power, Table C-1, to 4.16 kV emergency buses for 15 min. (Note 4)

AC power capability to 4.16 kV emergency buses reduced to a single power source, Table C-1, for 15 min. (Note 4)

Onsite None

• Do not delay declaration awaiting dose assessment

• DG102 AND Loss of results

• DG103 ANY additional single power source failure will result in a loss

• If dose assessment results are available, declaration AC Power RA1.2 1 2 3 4 D RU1.2 1 2 3 4 D of all 4.16 kV emergency bus power should be based on dose assessment instead of radiation monitor values (see EAL RS1.2)

• T-101N ANY liquid monitor reading > Table R-1 Alert column ANY liquid monitor reading > Table R-1 UE column Offsite for 15 min. (Note 2) for 60 min. (Note 2)

• T-101S 2

CU2.1 3 4

• T-10 backfed from offsite through T-1 or T-2 (only if already aligned) None < 106 VDC on required 125 VDC buses (Battery board 11, Loss of Battery board 12) for 15 min. (Note 4)

RA1.3 1 2 3 4 D RU1.3 1 2 3 4 D DC Power CG3.1 3 4 CS3.1 3 4 CA3.1 3 4 CU3.1 3 Confirmed sample analyses for gaseous or liquid releases Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 200 x ODCM limits indicate concentrations or release rates > 2 x ODCM limits 1

RPV water level < -84 in. for 30 min. (Note 4) With CONTAINMENT CLOSURE not established, RPV RPV water level < +5 in. RCS leakage results in the inability to maintain or restore RPV for 15 min. (Note 2) for 60 min. (Note 2)

AND water level < -1 in. OR water level > +53 in. for 15 min. (Note 4)

ANY Containment Challenge Indication, Table C-3 RPV water level cannot be monitored for 15 min. with ANY Offsite Rad UNPLANNED RPV leakage indication, Table C-2 (Note 4)

Conditions RG1.2 1 2 3 4 D RS1.2 1 2 3 4 D CG3.2 3 4 CS3.2 3 4 CU3.2 4 Dose assessment using actual meteorology indicates doses Dose assessment using actual meteorology indicates doses RPV water level cannot be monitored with core uncovery With CONTAINMENT CLOSURE established, RPV water

> 1,000 mRem TEDE or 5,000 mRem thyroid CDE at or > 100 mRem TEDE or 500 mRem thyroid CDE at or beyond None None UNPLANNED RPV water level drop below EITHER of the indicated by ANY of the following for 30 min. (Note 4): level < -84 in.

beyond the SITE BOUNDARY the SITE BOUNDARY following for 15 min. (Note 4):

• ANY UNPLANNED RPV leakage indication, Table C-2

• Erratic Source Range Monitor indication

• 0 ft Flange Level (RPV flange)

RG1.3 1 2 3 4 D RS1.3 1 2 3 4 D CS3.3 3 4

• RPV water level band (when the RPV water level band AND is established below the RPV flange)

ANY Containment Challenge Indication, Table C-3 RPV water level cannot be monitored for 30 min. (Note 4)

R Field survey results indicate closed window dose rates Field survey results indicate closed window dose rates with a loss of RPV inventory as indicated by ANY of the

> 1,000 mRem/hr expected to continue for 60 min. at or beyond the SITE BOUNDARY (Note 1)

> 100 mRem/hr expected to continue for 60 min. at or beyond the SITE BOUNDARY (Note 1)

None None 3 following:

• ANY UNPLANNED RPV leakage indication, Table C-2 CU3.3 4 C

OR OR Analyses of field survey samples indicate thyroid CDE Analyses of field survey samples indicate thyroid CDE

• Erratic Source Range Monitor indication Abnorm. RPV

> 5,000 mRem for 1 hr of inhalation at or beyond the SITE > 500 mRem for 1 hr of inhalation at or beyond the SITE RPV water level cannot be monitored with a loss of Rad Water BOUNDARY (Note 1) BOUNDARY (Note 1) RPV inventory as indicated by ANY UNPLANNED RPV Release Level leakage indication, Table C-2

/ Rad Table C-2 RPV Leakage Indications Effluent Cold SD/

Refuel RA2.1 1 2 3 4 D RU2.1 1 2 3 4 D

• Drywell equipment drain tank level rise System Alarm on ANY of the following radiation monitors due to UNPLANNED water level drop in a reactor refueling pathway Malfunct.

• Drywell floor drain tank level rise damage to irradiated fuel or loss of water level: as indicated by inability to restore and maintain SFP level > low

• Reactor building equipment sump level rise Table R-1 Effluent Monitor Classification Thresholds

• ARM 18 (West end of shield wall) water level alarm (Note 3)

• ARM 25 (Rx building - east wall)

• Reactor Building floor drain sump level rise AND Monitor GE SAE ALERT UE

• ARM 29 (Refuel bridge (LOW RANGE)) Area radiation monitor reading rise on ANY of the following:

• Torus water level rise

• Refuel Bridge (HIGH RANGE)

• ARM 18 (West end of shield wall) 2

• Reactor Building Vent Radiation Monitor

• UNPLANNED rise in RPV make-up rate GASEOUS Stack (RN 10A/B) N/A N/A 3.0E4 cps 300 cps

• ARM 25 (Rx building - east wall)

• ARM 29 (Refuel bridge (LOW RANGE))

• Observation of UNISOLABLE RCS leakage

• Refuel Bridge (HIGH RANGE)

Onsite Rad EC Vent N/A 300 mRem/hr 30 mRem/hr 10 mRem/hr Conditions

& Table C-3 Containment Challenge Spent Fuel SW Effluent N/A N/A 90,000 cpm 900 cpm Indications CA4.1 4 CU4.1 4 LIQUID Events 3 3 RA2.2 1 2 3 4 D 1 2 3 4 D 4

RU2.2

• CONTAINMENT CLOSURE not established An UNPLANNED event results in EITHER: Unplanned event results in RCS temperature > 212°F RW Discharge N/A N/A 200 x batch 2 x batch RCS temperature > 212°F for > Table C-4 duration A water level drop in a reactor refueling pathway that will UNPLANNED area radiation readings rise by a factor of

• Explosive mixture exists inside Primary Containment OR result in irradiated fuel becoming uncovered 1,000 over NORMAL LEVELS RCS (H2 6% and O2 5%) CU4.2 3 4 RPV pressure increase > 10 psi due to an UNPLANNED Temp. loss of decay heat removal capability Loss of all RCS temperature and RPV water level

• UNPLANNED rise in Primary Containment pressure indication for 15 min. (Note 4)

• RB area radiation > 8 R/hr RA3.1 1 2 3 4 D 3 None None Dose rates > 15 mRem/hr in EITHER of the following areas requiring continuous occupancy to maintain plant safety 5 CU5.1 3 4 An UNPLANNED sustained positive period observed on CR/CAS functions: Inadvertent Table C-4 RCS Reheat Duration nuclear instrumentation Rad Control Room OR Criticality Thresholds CAS

• If an RCS heat removal system is in operation within this time CU6.1 3 4 D HA1.1 1 2 3 4 D HU1.1 1 2 3 4 D frame and RCS temperature is being reduced, the EAL is not Table C-5 Communications Systems 6

applicable Loss of all Table C-5 onsite (internal) communication methods NMP-2 seismic instrumentation indicates > 0.075 g CONTAINMENT System Onsite Offsite affecting the ability to perform routine operations Table H-1 Safe Shutdown Areas Seismic event identified by ANY two of the following: RCS Status Duration (internal) (external)

AND CLOSURE Status OR

• Annunciator H2-1-6 SEISMIC DETECTION Comm.

Earthquake confirmed by ANY of the following: EQUIPMENT EVENT indicates seismic event detected Loss of all Table C-5 offsite (external) communication

• Reactor Building (including Primary Containment)

• Earthquake felt in plant INTACT N/A 60 min.* PBX (normal dial telephones) X X

• Confirmation of earthquake received on NMP-2 or methods affecting the ability to perform offsite notifications

• Control Room

• JAFNPP seismic instrumentation JAFNPP seismic instrumentation Gaitronics X

• Control Room indication of degraded performance of Established 20 min.*

• Screenhouse

• Earthquake felt in plant Hand-Held Portable Radio (station radio) X systems required for the safe shutdown of the plant Not INTACT

• Turbine Building Not established 0 min. Control Room installed satellite phones (non portable) X HA1.2 1 2 3 4 D HU1.2 1 2 3 4 D

• Battery Rooms ENS X Tornado striking Tornado striking within PROTECTED AREA boundary RECS X

• Battery Board Rooms OR OR

• Cable Spreading Room Sustained high winds > 90 mph Sustained high winds > 90 mph resulting in EITHER:

• Main Steam Isolation Valve Room VISIBLE DAMAGE to ANY SAFETY-RELATED

• Diesel Generator Engine and Board Rooms STRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 area

• Security OR

• Central Alarm Station Control Room indication of degraded performance of ANY SAFETY-RELATED STRUCTURE, SYSTEM or

• Secondary Alarm Station COMPONENT within ANY Table H-1 area

• Security Uninterruptible Power Supply Room HA1.3 1 2 3 4 D HU1.3 1 2 3 4 D Internal flooding Internal flooding that has the potential to affect ANY resulting in EITHER: SAFETY-RELATED STRUCTURE, SYSTEM, OR An electrical shock hazard that precludes access to operate COMPONENT required by Technical Specifications for the or monitor ANY SAFETY-RELATED STRUCTURE, current operating mode in ANY Table H-1 area Notes 1

SYSTEM or COMPONENT within ANY Table H-1 area

1. The ED should not wait until the applicable time has elapsed, but should declare the event as soon OR as it is determined that the condition will likely exceed the applicable time Control Room indication of degraded performance of ANY

2. The ED should not wait until the applicable time has elapsed, but should declare the event as soon SAFETY-RELATED STRUCTURE, SYSTEM or Natural or as it is determined that the release duration has exceeded, or will likely exceed, the applicable time. COMPONENT within ANY Table H-1 area Destructive Phenomena In the absence of data to the contrary, assume that the release duration has exceeded the applicable HA1.4 1 2 3 4 D HU1.4 1 2 3 4 D time if an ongoing release is detected and the release start time is unknown.

3. If loss of water level in the refueling pathway occurs while in Mode 3, 4 or D, consider classification Turbine failure-generated PROJECTILEs Turbine failure resulting in ANY of the following:

under EALs CU3.1, CU3.2 or CU3.3 resulting in EITHER:

• Casing penetration VISIBLE DAMAGE to or penetration of ANY SAFETY-

• Damage to turbine seals

4. The ED should not wait until the applicable time has elapsed, but should declare the event as soon RELATED STRUCTURE, SYSTEM or COMPONENT

• Damage to generator seals as it is determined that the condition has exceeded, or will likely exceed, the applicable time. within ANY Table H-1 area

5. If the equipment in the stated area was already inoperable, or out of service, before the event OR occurred, then EAL HA3.1 should not be declared as it will have no adverse impact on the ability of Control Room indication of degraded performance of the plant to safely operate or safely shutdown beyond that already allowed by Technical ANY SAFETY-RELATED STRUCTURE, SYSTEM or Specifications at the time of the event. COMPONENT within ANY Table H-1 area HA1.5 1 2 3 4 D HU1.5 1 2 3 4 D Lake water level > 254 ft Intake water level < 238.8 ft OR Intake water level < 236 ft HA1.6 1 2 3 4 D Vehicle crash resulting in EITHER:

VISIBLE DAMAGE to ANY SAFETY-RELATED H STRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 area OR Control Room indication of degraded performance of ANY Hazards SAFETY-RELATED STRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 area Other HA2.1 1 2 3 4 D HU2.1 1 2 3 4 D Condi-tions FIRE or EXPLOSION FIRE not extinguished within 15 min. of Control Room Affect-ing 2 None None resulting in EITHER:

VISIBLE DAMAGE to ANY SAFETY-RELATED STRUCTURE, SYSTEM or COMPONENT within ANY Table notification or verification of a Control Room FIRE alarm in ANY Table H-1 area, RadWaste Solidification and Storage Bldg, or Security West Bldg (Note 4)

Plant H-1 area Safety Fire or Explosion OR HU2.2 1 2 3 4 D Control Room indication of degraded performance of ANY SAFETY-RELATED STRUCTURE, SYSTEM or EXPLOSION of sufficient force to damage permanent COMPONENT within ANY Table H-1 area structures or equipment within the PROTECTED AREA HA3.1 1 2 3 4 D HU3.1 1 2 3 4 D 3 None None Access to ANY Table H-1 area is prohibited due to toxic, corrosive, asphyxiant or flammable gases which jeopardize operation of systems required to maintain safe operations or Toxic, corrosive, asphyxiant or flammable gases in amounts that have or could adversely affect NORMAL PLANT OPERATIONS Hazardous safely shutdown the reactor (Note 5)

Gas HU3.2 1 2 3 4 D Recommendation by local, county or state officials to evacuate or shelter site personnel based on an offsite event HG4.1 1 2 3 4 D HS4.1 1 2 3 4 D HA4.1 1 2 3 4 D HU4.1 1 2 3 4 D A HOSTILE ACTION has occurred such that plant A HOSTILE ACTION is occurring or has occurred within the A HOSTILE ACTION is occurring or has occurred within the A SECURITY CONDITION that does not involve a HOSTILE personnel are unable to operate equipment required to PROTECTED AREA as reported by the Security Site Owner Controlled Area as reported by the Security Site ACTION as reported by the Security Site Supervisor 4 maintain safety functions HG4.2 1 2 3 4 D Supervisor Supervisor OR OR A credible site-specific security threat notification A validated notification from NRC of an AIRLINER attack OR Security A HOSTILE ACTION has caused failure of Spent Fuel threat within 30 min. of the site A validated notification from NRC providing information of an Cooling systems aircraft threat AND IMMINENT fuel damage is likely 5

HS5.1 1 2 3 4 D HA5.1 1 2 3 4 D Control Room evacuation has been initiated Control Room evacuation has been initiated Control None AND None Room Control of the plant cannot be established within 15 min.

Evacuation HG6.1 1 2 3 4 D HS6.1 1 2 3 4 D HA6.1 1 2 3 4 D HU6.1 1 2 3 4 D Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial have occurred which involve actual or likely major failures of have occurred which involve an actual or potential have occurred which indicate a potential degradation of the core degradation or melting with potential for loss of plant functions needed for protection of the public or substantial degradation of the level of safety of the plant or a level of safety of the plant or indicate a security threat to containment integrity or HOSTILE ACTION that results in an HOSTILE ACTION that results in intentional damage or security event that involves probable life threatening risk to facility protection has been initiated. No releases of 6 actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action malicious acts; (1) toward site personnel or equipment that could lead to the likely failure of or; (2) that prevent effective site personnel or damage to site equipment because of HOSTILE ACTION. ANY releases are expected to be limited radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems Prepared for Constellation by: Operations Support Services, Inc. - www.ossi-net.com (Rev. 23 1/18/13)

Guideline exposure levels (1,000 mRem TEDE or 5,000 access to equipment needed for the protection of the public. to small fractions of the EPA Protective Action Guideline occurs Judgment mRem thyroid CDE) offsite for more than the immediate site ANY releases are not expected to result in exposure levels exposure levels (1,000 mRem TEDE or 5,000 mRem thyroid area which exceed EPA Protective Action Guideline exposure CDE) levels (1,000 mRem TEDE or 5,000 mRem thyroid CDE) beyond the SITE BOUNDARY EU1.1 1 2 3 4 D EAL Identifier XXX.X E None None None Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by measured dose rates > then ANY of the following:

• 400 mRem/hr at 3 feet from the HSM surface Category (R, H, E, S, F, C) Sequential number within subcategory/classification ISFSI

• 100 mRem/hr outside HSM door on centerline Emergency classification (G, S, A, U) Subcategory number (1 if no subcategory)

• 20 mRem/hr end shield wall exterior Modes:

EPIP-EPP-01-EAL 1

Power Operation 2

Hot Shutdown 3

Cold Shutdown 4

Refuel D

Defueled Nine Mile Point Nuclear Station Unit 1 Revision 24 EAL Matrix Unit 1 Page 2 of 2 MODE 3, 4 or D

Attachment 3 Page 3 of 4 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT RG1.1 1 2 3 4 5 D RS1.1 1 2 3 4 5 D RA1.1 1 2 3 4 5 D RU1.1 1 2 3 4 5 D SG1.1 1 2 3 SS1.1 1 2 3 SA1.1 1 2 3 SU1.1 1 2 3 Loss of all offsite and all onsite AC power, Table S-1, to 4.16 Loss of all offsite and all onsite AC power, Table S-1, to AC power capability to 4.16 KV emergency buses Loss of all offsite AC power, Table S-1, to 4.16 KV ANY monitor reading > Table R-1 GE column ANY monitor reading > Table R-1 SAE column ANY gaseous monitor reading > Table R-1 Alert column ANY gaseous monitors > Table R-1 UE column KV emergency buses 2ENS*SWG101 and 2ENS*SWG103 4.16 KV emergency buses 2ENS*SWG101 and 2ENS*SWG101 and 2ENS*SWG103 reduced to a single emergency buses 2ENS*SWG101 and 2ENS*SWG103 for 15 min. (Note 1) for 15 min. (Note 1) for 15 min. (Note 2) for 60 min. (Note 2) AND EITHER: 2ENS*SWG103 for 15 min. (Note 4) power source, Table S-1, for 15 min. (Note 4)

• Do not delay declaration awaiting dose assessment results

• If dose assessment results are available, declaration

• Do not delay declaration awaiting dose assessment results

• If dose assessment results are available, declaration 1 Restoration of 4.16 KV emergency bus 2ENS*SWG101 or 2ENS*SWG103 within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely OR AND ANY additional single power source failure will result in a loss of all power to 4.16 KV emergency buses 2ENS*SWG101 Table S-1 AC Power Sources should be based on dose assessment instead of should be based on dose assessment instead of RA1.2 1 2 3 4 5 D RU1.2 1 2 3 4 5 D Loss of RPV water level cannot be restored and maintained and 2ENS*SWG103 radiation monitor values (see EAL RG1.2) radiation monitor values (see EAL RS1.2)

• 2EGS*EG1 AC above -14 in. or RPV water level cannot be determined ANY liquid monitor > Table R-1 Alert column ANY liquid monitor reading > Table R-1 UE column Onsite Power

• 2EGS*EG3 for 15 min. (Note 2) for 60 min. (Note 2)

• 2EGS*EG2 (with 2ENS*SWG102 crosstied to 2ENS*SWG101 or 2ENS*SWG103)

• Reserve Transformer A 2 Offsite RA1.3 1 2 3 4 5 D RU1.3 1 2 3 4 5 D SS2.1 1 2 3

• Reserve Transformer B 1 Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 200 x ODCM limits Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 2 x ODCM limits Loss of DC Power None < 105 VDC on both 2BYS*SWG002A and 2BYS*SWG002B for 15 min. (Note 4)

None

• Aux Boiler Transformer for 15 min. (Note 2) for 60 min. (Note 2)

Offsite Rad SG3.1 1 2 SS3.1 1 2 SA3.1 1 2 SU3.1 3 Conditions RG1.2 1 2 3 4 5 D RS1.2 1 2 3 4 5 D An automatic scram fails to shut down the reactor An automatic scram failed to shut down the reactor as An automatic scram failed to shut down the reactor An UNPLANNED sustained positive period observed on as indicated by reactor power > 4% indicated by reactor power > 4% AND nuclear instrumentation Dose assessment using actual meteorology indicates doses

> 1,000 mRem TEDE or 5,000 mRem thyroid CDE at or Dose assessment using actual meteorology indicates doses

> 100 mRem TEDE or 500 mRem thyroid CDE at or beyond None None 3 AND All manual actions fail to shut down the reactor as indicated by reactor power > 4%

AND Manual actions taken at the reactor control console (mode Manual actions taken at the reactor control console (mode switch in shutdown, manual scram push buttons or beyond the SITE BOUNDARY the SITE BOUNDARY switch in shutdown, manual scram push buttons and ARI) ARI) successfully shut down the reactor as indicated by Criticality AND EITHER of the following exist or have occurred: failed to shut down the reactor as indicated by reactor reactor power 4%

& RPV water level cannot be restored and maintained power > 4%

RPS above -39 in. or RPV water level cannot be determined RG1.3 1 2 3 4 5 D RS1.3 1 2 3 4 5 D OR Failure Suppression pool temperature and RPV pressure R Field survey results indicate closed window dose rates

> 1,000 mRem/hr expected to continue for 60 min. at or beyond the SITE BOUNDARY (Note 1)

Field survey results indicate closed window dose rates

> 100 mRem/hr expected to continue for 60 min. at or beyond the SITE BOUNDARY (Note 1)

None None cannot be maintained below the Heat Capacity Temperature Limit (N2-EOP-PC Figure M)

OR OR 4

Abnorm. SU4.1 1 2 3 Analyses of field survey samples indicate thyroid CDE Analyses of field survey samples indicate thyroid CDE Rad Release

/ Rad

> 5,000 mRem for 1 hr of inhalation at or beyond the SITE BOUNDARY (Note 1)

> 500 mRem for 1 hr of inhalation at or beyond the SITE BOUNDARY (Note 1)

S Inability to Reach or Maintain None None None Plant is not brought to required operating mode within Technical Specifications LCO required action completion time Effluent RA2.1 1 2 3 4 5 D RU2.1 1 2 3 4 5 D Shutdown Table R-1 Effluent Monitor Classification Thresholds Conditions Alarm on ANY of the following radiation monitors due to UNPLANNED water level drop in a reactor refueling pathway System Monitor GE SAE ALERT UE damage to irradiated fuel or loss of water level: as indicated by inability to restore and maintain SFP level > Malfunct.

• 2RMS-RE111 SS5.1 1 2 3 SA5.1 1 2 3 SU5.1 1 2 3 low water level alarm (Note 3)

• 2RMS-RE112 AND GASEOUS Radwaste/RB Vent Effluent 5.5E+7 µCi/s 5.5E+6 µCi/s 200 x Alarm 2 x Alarm Loss of > approximately 75% of annunciation or indication UNPLANNED loss of > approximately 75% of annunciation or UNPLANNED loss of > approximately 75% of annunciation or

• 2RMS-RE113 Area radiation monitor reading rise on ANY of the following:

2 Main Stack Effluent 1.0E+10 µCi/s 1.0E+9 µCi/s 200 x Alarm 2 x Alarm

• 2RMS-RE114

• 2RMS-RE140

• 2HVR*RE14A

• 2RMS-RE111

• 2RMS-RE112

• 2RMS-RE113 Table S-2 Significant Transients on all of the following Control Room panels for 15 min.

(Note 4):

• 2CEC*PNL601 indication on all of the following Control Room panels for 15 min. (Note 4):

• 2CEC*PNL601 indication on all of the following Control Room panels for 15 min. (Note 4):

• 2CEC*PNL601 Onsite Rad

• 2CEC*PNL602

• 2CEC*PNL602

• 2CEC*PNL602 5

• 2HVR*RE14B

• 2RMS-RE114 Conditions

• 2RMS-RE140

• Automatic turbine runback > 25% thermal reactor power

• 2CEC*PNL603

• 2CEC*PNL603

• 2CEC*PNL603 Service Water Effluent N/A N/A 200 x DRMS High (red) 2 x DRMS High (red)

• 2CEC*PNL851

&

• Electric load rejection > 25% full electrical load

• 2CEC*PNL851

• 2CEC*PNL851 Spent Fuel

• 2CEC*PNL852

• 2CEC*PNL852

• 2CEC*PNL852 Inst.

LIQUID Events

• Reactor scram AND AND EITHER:

Liquid RadWaste Effluent N/A N/A N/A 2 x DRMS High (red)

• ECCS injection A significant transient is in progress, Table S-2 A significant transient is in progress, Table S-2 RA2.2 1 2 3 4 5 D RU2.2 1 2 3 4 5 D

• Thermal power oscillations > 10% AND OR Compensatory indications are unavailable (Plant Computer, Compensatory indications are unavailable (Plant Process Cooling Tower Blowdown N/A N/A 200 x DRMS High (red) 2 x DRMS High (red)

A water level drop in a reactor refueling pathway that will UNPLANNED area radiation readings rise by a factor of SPDS) Computer, SPDS) result in irradiated fuel becoming uncovered 1,000 over NORMAL LEVELS Table S-3 Communications Systems SU6.1 1 2 3 RA3.1 1 2 3 4 5 D 3 Dose rates > 15 mRem/hr in EITHER of the following areas requiring continuous occupancy to maintain plant safety 6 None System Onsite Offsite (internal) (external)

Loss of all Table S-3 onsite (internal) communication methods affecting the ability to perform routine operations OR None None functions: None Comm. Loss of all Table S-3 offsite (external) communication CR/CAS Control Room PBX (normal dial telephones) X X methods affecting the ability to perform offsite notifications Rad OR CAS Gaitronics X SU7.1 1 2 3 HA1.1 1 2 3 4 5 D HU1.1 1 2 3 4 5 D 7 Station radio (portable)

Control Room installed satellite phones (non portable)

X X Reactor coolant activity > 4 µCi/gm I-131 Equivalent Fuel Clad None ENS X Seismic event > OBE (0.075g) Seismic event identified by ANY two of the following:

Degradation X SU7.2 1 2 3 as indicated by EITHER:

• Annunciator 842121 SEISMIC ACCELERATION RECS Computer Point ERSNC02, OBE Detected EXCEEDED indicates seismic event detected OR Offgas radiation DRMS high (red) alarm for 15 min.

• Confirmation of earthquake received on NMP-1 or Table H-1 Safe Shutdown Areas ANY amber LED light lit at the Seismic Monitor Panel, JAFNPP seismic instrumentation Response Spectrum Annunciator SU8.1 1 2 3

• Earthquake felt in plant 8

• Reactor Building (including Primary Containment) AND Earthquake confirmed by ANY of the following: Unidentified or reactor coolant pressure boundary

• Control Room

• Earthquake felt in plant leakage > 10 gpm None None None

• Diesel Generator Engine and Board Rooms

• JAFNPP seismic instrumentation RCS OR

• Control Room indication of degraded performance of Leakage Identified reactor coolant leakage > 25 gpm

• Standby Switchgear and Battery Rooms systems required for the safe shutdown of the plant

• HPCS Switchgear and Battery Rooms HA1.2 1 2 3 4 5 D HU1.2 1 2 3 4 5 D F

• Remote Shutdown Rooms FG1.1 1 2 3 FS1.1 1 2 3 FA1.1 1 2 3 FU1.1 1 2 3 Tornado striking Tornado striking within PROTECTED AREA boundary

• Control Building HVAC Rooms OR OR Sustained high winds > 90 mph Loss of ANY two fission product barriers Loss or potential loss of ANY two fission product barriers ANY loss or ANY potential loss of EITHER Fuel Clad ANY loss or ANY potential loss of Containment barrier

• Service Water Pump Rooms Sustained high winds > 90 mph Fission Product resulting in EITHER: AND (Table F-1) barrier OR RCS barrier (Table F-1) (Table F-1)

• Electrical Protection Assembly Room Barrier Loss or potential loss of third fission product barrier VISIBLE DAMAGE to ANY SAFETY-RELATED STRUCTURE, SYSTEM, OR COMPONENT within ANY Degradation (Table F-1)

• PGCC Relay Room Table H-1 area OR Control Room indication of degraded performance of ANY SAFETY-RELATED STRUCTURE, SYSTEM, OR COMPONENT within ANY Table H-1 area Table F-1 Fission Product Barrier Matrix HA1.3 1 2 3 4 5 D HU1.3 1 2 3 4 5 D Notes Internal flooding Internal flooding that has the potential to affect ANY Fuel Clad Barrier Reactor Coolant System Barrier Containment Barrier

1. The ED should not wait until the applicable time has elapsed, but should declare the event as soon resulting in EITHER: SAFETY-RELATED STRUCTURE, SYSTEM, OR as it is determined that the condition will likely exceed the applicable time An electrical shock hazard that precludes access to operate COMPONENT required by Technical Specifications for the Loss Potential Loss Loss Potential Loss Loss Potential Loss

2. The ED should not wait until the applicable time has elapsed, but should declare the event as soon or monitor ANY SAFETY-RELATED STRUCTURE, current operating mode in ANY Table H-1 area SYSTEM or COMPONENT within ANY Table H-1 area 1. Primary Containment 1. RPV water level cannot be 1. RPV water level cannot be 1. Primary Containment 1

as it is determined that the release duration has exceeded, or will likely exceed, the applicable time.

In the absence of data to the contrary, assume that the release duration has exceeded the applicable OR Flooding is required restored and maintained restored and maintained Flooding is required Control Room indication of degraded performance of ANY A

time if an ongoing release is detected and the release start time is unknown. above -14 in. following above -14 in. or RPV water SAFETY-RELATED STRUCTURE, SYSTEM or depressurization of the RPV level cannot be determined Natural or 3. If loss of water level in the refueling pathway occurs while in Mode 4, 5 or D, consider classification COMPONENT within ANY Table H-1 area or RPV water level cannot None None Destructive under EALs CU3.1, CU3.2 or CU3.3 RPV Water 1 2 3 4 5 D HU1.4 1 2 3 4 5 D be determined Phenomena HA1.4

4. The ED should not wait until the applicable time has elapsed, but should declare the event as soon Level as it is determined that the condition has exceeded, or will likely exceed, the applicable time. Turbine failure-generated PROJECTILEs Turbine failure resulting in ANY of the following:

5. If the equipment in the stated area was already inoperable, or out of service, before the event resulting in EITHER:

• Casing penetration

2. Primary Containment 1. Primary Containment 2. Primary Containment occurred, then EAL HA3.1 should not be declared as it will have no adverse impact on the ability of VISIBLE DAMAGE to or penetration of ANY SAFETY-

• Damage to turbine seals pressure > 1.68 psig due to pressure rise followed by a pressure > 45 psig and the plant to safely operate or safely shutdown beyond that already allowed by Technical RELATED STRUCTURE, SYSTEM or COMPONENT

• Damage to generator seals RCS leakage rapid UNPLANNED drop in rising Specifications at the time of the event. within ANY Table H-1 area Primary Containment 3. Explosive mixture exists OR Control Room indication of degraded performance of ANY SAFETY-RELATED STRUCTURE, SYSTEM or B pressure

2. Primary Containment inside Primary Containment

( 6% H2 and 5% O2)

Primary None None None COMPONENT within ANY Table H-1 area pressure response not 4. Suppression pool Containment HA1.5 1 2 3 4 5 D HU1.5 1 2 3 4 5 D consistent with LOCA temperature and RPV Pressure /

conditions pressure cannot be Temperature Lake water level > 254 ft Intake water level < 237 ft maintained below the Heat OR Capacity Temperature Limit Intake water level < 233 ft (N2-EOP-PC Figure M)

HA1.6 1 2 3 4 5 D 3. Release pathway exists 1. UNISOLABLE primary 3. Failure of all Primary outside Primary system leakage outside Containment isolation Vehicle crash H

Containment resulting from Primary Containment as valves in ANY one line to resulting in EITHER: isolation failure in ANY of indicated by exceeding close following auto or VISIBLE DAMAGE to ANY SAFETY-RELATED the following systems EITHER: manual initiation STRUCTURE, SYSTEM or COMPONENT within ANY Table (excluding normal process RB area temperature AND H-1 area system flowpaths from an above an isolation Direct downstream pathway Hazards OR UNISOLABLE system): setpoint outside Primary Control Room indication of degraded performance of ANY

• Main steam line OR Containment and to the

& SAFETY-RELATED STRUCTURE, SYSTEM or Other

• RCIC steam line RB area radiation above environment exists COMPONENT within ANY Table H-1 area

• RWCU an alarm setpoint Condi-

• Feedwater 4. Intentional Primary tions Affect-HA2.1 1 FIRE or EXPLOSION 2 3 4 5 D HU2.1 1 2 3 4 5 D FIRE not extinguished within 15 min. of Control Room C None None

4. RPV blowdown is required Containment venting per EOPs None 2

ing resulting in EITHER: notification or verification of a Control Room FIRE alarm in Isolation Plant VISIBLE DAMAGE to ANY SAFETY-RELATED ANY Table H-1 area or Turbine Building (Note 4) 5. UNISOLABLE primary None None STRUCTURE, SYSTEM or COMPONENT within ANY Table system leakage outside Safety Primary Containment as H-1 area HU2.2 1 2 3 4 5 D Fire or indicated by exceeding OR Explosion EITHER:

Control Room indication of degraded performance of ANY EXPLOSION of sufficient force to damage permanent RB area maximum safe SAFETY-RELATED STRUCTURE, SYSTEM or structures or equipment within the PROTECTED AREA temperature value COMPONENT within ANY Table H-1 area (N2-EOP-SC Detail S)

OR HA3.1 1 2 3 4 5 D HU3.1 1 2 3 4 5 D RB area radiation >

3 8.00E+3 mR/hr Access to ANY Table H-1 area is prohibited due to toxic, Toxic, corrosive, asphyxiant or flammable gases in amounts corrosive, asphyxiant or flammable gases which jeopardize that have or could adversely affect NORMAL PLANT None None operation of systems required to maintain safe operations or OPERATIONS Hazardous 2. Drywell area radiation 5. Drywell area radiation 5. Drywell area radiation safely shutdown the reactor (Note 5)

Gas HU3.2 1 2 3 4 5 D 3100 R/hr (3.1 E6 41 R/hr 6.0 E4 R/hr D

mRem/hr) (4.1 E4 mRem/hr) (6.0 E7 mRem/hr)

Recommendation by local, county or state officials to None None None evacuate or shelter site personnel based on an offsite event 3. Reactor coolant activity Rad > 300 µCi/gm I-131 HG4.1 1 2 3 4 5 D HS4.1 1 2 3 4 5 D HA4.1 1 2 3 4 5 D HU4.1 1 2 3 4 5 D Equivalent A HOSTILE ACTION has occurred such that plant A HOSTILE ACTION is occurring or has occurred within the A HOSTILE ACTION is occurring or has occurred within the A SECURITY CONDITION that does not involve a HOSTILE personnel are unable to operate equipment required to PROTECTED AREA as reported by the Security Site Owner Controlled Area as reported by the Security Site ACTION as reported by the Security Site Supervisor 4. ANY condition in the 2. ANY condition in the 6. ANY condition in the 2. ANY condition in the opinion 6. ANY condition in the opinion 6. ANY condition in the opinion 4 maintain safety functions HG4.2 1 2 3 4 5 D Supervisor Supervisor OR OR A credible site-specific security threat notification E opinion of the Emergency Director that indicates loss of the Fuel Clad barrier opinion of the Emergency Director that indicates potential loss of the Fuel opinion of the Emergency Director that indicates loss of the Reactor Coolant of the Emergency Director that indicates potential loss of the Reactor Coolant of the Emergency Director that indicates loss of the Containment barrier of the Emergency Director that indicates potential loss of the Containment barrier A validated notification from NRC of an AIRLINER attack OR Judgment Clad barrier System barrier System barrier Security A HOSTILE ACTION has caused failure of Spent Fuel threat within 30 min. of the site A validated notification from NRC providing information of an Cooling systems aircraft threat AND IMMINENT fuel damage is likely 5

HS5.1 1 2 3 4 5 D HA5.1 1 2 3 4 5 D Control Room evacuation has been initiated Control Room evacuation has been initiated Control None AND None Room Control of the plant cannot be established within 15 min.

Evacuation HG6.1 1 2 3 4 5 D HS6.1 1 2 3 4 5 D HA6.1 1 2 3 4 5 D HU6.1 1 2 3 4 5 D Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial have occurred which involve actual or likely major failures of have occurred which involve an actual or potential have occurred which indicate a potential degradation of the core degradation or melting with potential for loss of plant functions needed for protection of the public or substantial degradation of the level of safety of the plant or a level of safety of the plant or indicate a security threat to containment integrity or HOSTILE ACTION that results in an HOSTILE ACTION that results in intentional damage or security event that involves probable life threatening risk to facility protection has been initiated. No releases of actual loss of physical control of the facility. Releases can be malicious acts; (1) toward site personnel or equipment that site personnel or damage to site equipment because of 6

radioactive material requiring offsite response or monitoring reasonably expected to exceed EPA Protective Action could lead to the likely failure of or; (2) that prevent effective HOSTILE ACTION. ANY releases are expected to be limited are expected unless further degradation of safety systems Prepared for Constellation by: Operations Support Services, Inc. - www.ossi-net.com (Rev. 22 1/18/13)

Guideline exposure levels (1,000 mRem TEDE or 5,000 access to equipment needed for the protection of the public. to small fractions of the EPA Protective Action Guideline occurs mRem thyroid CDE) offsite for more than the immediate site ANY releases are not expected to result in exposure levels exposure levels (1,000 mRem TEDE or 5,000 mRem thyroid Judgment area which exceed EPA Protective Action Guideline exposure CDE) levels (1,000 mRem TEDE or 5,000 mRem thyroid CDE) beyond the SITE BOUNDARY EU1.1 1 2 3 4 5 D EAL Identifier E None None None Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by measured dose rates > then ANY of the following:

• 400 mRem/hr at 3 feet from the HSM surface

• 100 mRem/hr outside HSM door on centerline XXX.X Category (R, H, E, S, F, C) Sequential number within subcategory/classification ISFSI

• 20 mRem/hr end shield wall exterior Emergency classification (G, S, A, U) Subcategory number (1 if no subcategory)

MODE 1, 2 or 3 EPIP-EP-02-EAL Modes: 1 Power Operation 2

Startup 3

Hot Shutdown 4

Cold Shutdown 5

Refuel D

Defueled Nine Mile Point Nuclear Station Unit 2 Revision 23 EAL Matrix Unit 2 Page 1 of 2

Attachment 3 Page 4 of 4 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT RG1.1 1 2 3 4 5 D RS1.1 1 2 3 4 5 D RA1.1 1 2 3 4 5 D RU1.1 1 2 3 4 5 D CA1.1 4 5 D CU1.1 4 5 D ANY monitor reading > Table R-1 GE column for 15 min. (Note 1)

ANY monitor reading > Table R-1 SAE column for 15 min. (Note 1)

ANY gaseous monitor reading > Table R-1 Alert column for 15 min. (Note 2)

ANY gaseous monitors > Table R-1 UE column for 60 min. (Note 2) 1 Table C-1 AC Power Sources Loss of all offsite and all onsite AC power, Table C-1, to 4.16 KV emergency buses 2ENS*SWG101 and 2ENS*SWG103 AC power capability to 4.16 KV emergency buses 2ENS*SWG101 and 2ENS*SWG103 reduced to a single power source, Table C-1, for 15 min. (Note 4)

Onsite

• Do not delay declaration awaiting dose assessment

• Do not delay declaration awaiting dose assessment Loss of

• 2EGS*EG1 for 15 min. (Note 4)

AND results results AC

• 2EGS*EG3 ANY additional single power source failure will result in a loss

• If dose assessment results are available, declaration

• If dose assessment results are available, declaration of all power to 4.16 KV emergency buses 2ENS*SWG101 Power should be based on dose assessment instead of should be based on dose assessment instead of RA1.2 1 2 3 4 5 D RU1.2 1 2 3 4 5 D and 2ENS*SWG103 radiation monitor values (see EAL RG1.2) radiation monitor values (see EAL RS1.2)

• Reserve Transformer A Offsite ANY liquid monitor > Table R-1 Alert column ANY liquid monitor reading > Table R-1 UE column

• Reserve Transformer B 2

CU2.1 4 5 for 15 min. (Note 2) for 60 min. (Note 2)

• Aux Boiler Transformer None < 105 VDC on required 125 VDC emergency buses Loss of for 15 min. (Note 4)

DC Power RA1.3 1 2 3 4 5 D RU1.3 1 2 3 4 5 D CG3.1 4 5 CS3.1 4 5 CA3.1 4 5 CU3.1 4 1 Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 200 x ODCM limits Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 2 x ODCM limits RPV level < -14 in. for 30 min. (Note 4)

AND With CONTAINMENT CLOSURE not established, RPV water level < 11.8 in.

RPV water level < 17.8 in.

OR RCS leakage results in the inability to maintain or restore RPV water level > 159.3 in. for 15 min. (Note 4) for 15 min. (Note 2) for 60 min. (Note 2)

Offsite Rad ANY Containment Challenge Indication, Table C-3 RPV water level cannot be monitored for 15 min. with ANY Conditions UNPLANNED RPV leakage indication, Table C-2 (Note 4)

RG1.2 1 2 3 4 5 D RS1.2 1 2 3 4 5 D CG3.2 4 5 CS3.2 4 5 Dose assessment using actual meteorology indicates doses Dose assessment using actual meteorology indicates doses CU3.2 5

> 1,000 mRem TEDE or 5,000 mRem thyroid CDE at or > 100 mRem TEDE or 500 mRem thyroid CDE at or beyond None None RPV water level cannot be monitored with core uncovery With CONTAINMENT CLOSURE established, RPV water beyond the SITE BOUNDARY the SITE BOUNDARY UNPLANNED RPV water level drop below EITHER of the indicated by ANY of the following for 30 min. (Note 4): level

• ANY UNPLANNED RPV leakage indication, Table C-2 < -14 in. following for 15 min. (Note 4):

• Erratic Source Range Monitor indication

• 364 in. (RPV flange)

RG1.3 1 2 3 4 5 D RS1.3 1 2 3 4 5 D CS3.3 4 5

• RPV water level band (when the RPV water level band AND is established below the RPV flange)

R ANY Containment Challenge Indication, Table C-3 RPV water level cannot be monitored for 30 min. (Note 4)

Field survey results indicate closed window dose rates Field survey results indicate closed window dose rates with a loss of RPV inventory as indicated by ANY of the

> 1,000 mRem/hr expected to continue for 60 min. at or beyond the SITE BOUNDARY (Note 1)

OR

> 100 mRem/hr expected to continue for 60 min. at or beyond the SITE BOUNDARY (Note 1)

OR None None 3 following:

• ANY UNPLANNED RPV leakage indication, Table C-2 CU3.3 5 Abnorm.

C Analyses of field survey samples indicate thyroid CDE Analyses of field survey samples indicate thyroid CDE

• Erratic Source Range Monitor indication RPV Rad > 5,000 mRem for 1 hr of inhalation at or beyond the SITE > 500 mRem for 1 hr of inhalation at or beyond the SITE Water RPV water level cannot be monitored with a loss of Release BOUNDARY (Note 1) BOUNDARY (Note 1) RPV inventory as indicated by ANY UNPLANNED RPV Level leakage indication, Table C-2

/ Rad Effluent RA2.1 1 2 3 4 5 D RU2.1 1 2 3 4 5 D Cold SD/ Table C-2 RPV Leakage Indications Table R-1 Effluent Monitor Classification Thresholds Refuel Alarm on ANY of the following radiation monitors due to UNPLANNED water level drop in a reactor refueling pathway System

• Drywell equipment drain sump level rise Monitor GE SAE ALERT UE damage to irradiated fuel or loss of water level: as indicated by inability to restore and maintain SFP level > Malfunct.

• Drywell floor drain sump level rise

• 2RMS-RE111 low water level alarm (Note 3)

• 2RMS-RE112 AND

• Reactor building equipment sump level rise GASEOUS Radwaste/RB Vent Effluent 5.5E+7 µCi/s 5.5E+6 µCi/s 200 x Alarm 2 x Alarm

• 2RMS-RE113 Area radiation monitor reading rise on ANY of the following:

• Reactor Building floor drain sump level rise 2 Main Stack Effluent 1.0E+10 µCi/s 1.0E+9 µCi/s 200 x Alarm 2 x Alarm

• 2RMS-RE114

• 2RMS-RE140

• 2HVR*RE14A

• 2RMS-RE111

• 2RMS-RE112

• 2RMS-RE113

• Suppression Pool level rise

• UNPLANNED rise in RPV make-up rate Onsite Rad

• 2HVR*RE14B

• 2RMS-RE114 Conditions Service Water Effluent N/A N/A 200 x DRMS High (red) 2 x DRMS High (red)

• 2RMS-RE140

• Observation of UNISOLABLE RCS leakage Spent Fuel LIQUID Events Liquid RadWaste Effluent N/A N/A N/A 2 x DRMS High (red) 1 2 3 4 5 D 1 2 3 4 5 D Table C-3 Containment Challenge RA2.2 RU2.2 CA4.1 4 5 CU4.1 4 5 Cooling Tower Blowdown N/A N/A 200 x DRMS High (red) 2 x DRMS High (red) Indications A water level drop in a reactor refueling pathway that will result in irradiated fuel becoming uncovered UNPLANNED area radiation readings rise by a factor of 1,000 over NORMAL LEVELS 4

• CONTAINMENT CLOSURE not established

• Explosive mixture exists inside Primary Containment An UNPLANNED event results in EITHER:

RCS temperature > 200°F for > Table C-4 duration OR UNPLANNED event results in RCS temperature > 200°F RCS CU4.2 4 5 RA3.1 1 2 3 4 (H2 6% and O2 5%) RPV pressure increase > 10 psi due to an UNPLANNED 5 D Temp. loss of decay heat removal capability Loss of all RCS temperature and RPV water level 3 None None Dose rates > 15 mRem/hr in EITHER of the following areas requiring continuous occupancy to maintain plant safety functions:

None

• UNPLANNED rise in Primary Containment pressure

• RB area radiation > 8.00E+3 mR/hr indication for 15 min. (Note 4)

CR/CAS Control Room CU5.1 4 5 Rad OR CAS 5 Table C-4 RCS Reheat Duration Table C-5 Communications Systems An UNPLANNED sustained positive period observed on nuclear instrumentation Inadvertent HA1.1 1 2 3 4 5 D HU1.1 1 2 3 4 5 D Criticality Thresholds Onsite Offsite System Seismic event > OBE (0.075g) Seismic event identified by ANY two of the following:

• If an RCS heat removal system is in operation within this time (internal) (external)

CU6.1 4 5 D as indicated by EITHER:

• Annunciator 842121 SEISMIC ACCELERATION frame and RCS temperature is being reduced, the EAL is not 6

Computer Point ERSNC02, OBE Detected EXCEEDED indicates seismic event detected applicable PBX (normal dial telephones) X X OR Loss of all Table C-5 onsite (internal) communication methods

• Confirmation of earthquake received on NMP-1 or CONTAINMENT Gaitronics X affecting the ability to perform routine operations ANY amber LED light lit at the Seismic Monitor Panel, RCS Status Duration Table H-1 Safe Shutdown Areas Response Spectrum Annunciator JAFNPP seismic instrumentation Comm.

CLOSURE Status Station radio (portable) X OR

• Earthquake felt in plant Loss of all Table C-5 offsite (external) communication AND Control Room installed satellite phones (non portable) X INTACT N/A 60 min.* methods affecting the ability to perform offsite notifications

• Reactor Building (including Primary Containment) Earthquake confirmed by ANY of the following:

• Earthquake felt in plant ENS X

• Control Room

• JAFNPP seismic instrumentation Established 20 min.* RECS X

• Diesel Generator Engine and Board Rooms

• Control Room indication of degraded performance of Not INTACT systems required for the safe shutdown of the plant Not established 0 min.

• Standby Switchgear and Battery Rooms HA1.2 1 2 3 4 5 D HU1.2 1 2 3 4 5 D

• HPCS Switchgear and Battery Rooms

• Remote Shutdown Rooms Tornado striking Tornado striking within PROTECTED AREA boundary OR OR

• Control Building HVAC Rooms Sustained high winds > 90 mph Sustained high winds > 90 mph

• Service Water Pump Rooms resulting in EITHER:

VISIBLE DAMAGE to ANY SAFETY-RELATED

• Electrical Protection Assembly Room STRUCTURE, SYSTEM or COMPONENT within ANY Table

• PGCC Relay Room H-1 area OR Control Room indication of degraded performance of ANY SAFETY-RELATED STRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 area HA1.3 1 2 3 4 5 D HU1.3 1 2 3 4 5 D Notes Internal flooding Internal flooding that has the potential to affect ANY resulting in EITHER: SAFETY-RELATED STRUCTURE, SYSTEM, OR

1. The ED should not wait until the applicable time has elapsed, but should declare the event as soon An electrical shock hazard that precludes access to operate COMPONENT required by Technical Specifications for the as it is determined that the condition will likely exceed the applicable time or monitor ANY SAFETY-RELATED STRUCTURE, current operating mode in ANY Table H-1 area

2. The ED should not wait until the applicable time has elapsed, but should declare the event as soon SYSTEM or COMPONENT within ANY Table H-1 area 1

as it is determined that the release duration has exceeded, or will likely exceed, the applicable time. OR In the absence of data to the contrary, assume that the release duration has exceeded the applicable Control Room indication of degraded performance of ANY time if an ongoing release is detected and the release start time is unknown. SAFETY-RELATED STRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 area Natural or 3. If loss of water level in the refueling pathway occurs while in Mode 4, 5 or D, consider classification Destructive under EALs CU3.1, CU3.2 or CU3.3 HA1.4 1 2 3 4 5 D HU1.4 1 2 3 4 5 D Phenomena 4. The ED should not wait until the applicable time has elapsed, but should declare the event as soon Turbine failure-generated PROJECTILEs Turbine failure resulting in ANY of the following:

as it is determined that the condition has exceeded, or will likely exceed, the applicable time. resulting in EITHER:

• Casing penetration

5. If the equipment in the stated area was already inoperable, or out of service, before the event VISIBLE DAMAGE to or penetration of ANY SAFETY-

• Damage to turbine seals occurred, then EAL HA3.1 should not be declared as it will have no adverse impact on the ability of RELATED STRUCTURE, SYSTEM or COMPONENT

• Damage to generator seals the plant to safely operate or safely shutdown beyond that already allowed by Technical within ANY Table H-1 area Specifications at the time of the event. OR Control Room indication of degraded performance of ANY SAFETY-RELATED STRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 area HA1.5 1 2 3 4 5 D HU1.5 1 2 3 4 5 D Lake water level > 254 ft Intake water level < 237 ft OR Intake water level < 233 ft HA1.6 1 2 3 4 5 D Vehicle crash H

resulting in EITHER:

VISIBLE DAMAGE to ANY SAFETY-RELATED STRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 area OR Hazards Control Room indication of degraded performance of ANY

& SAFETY-RELATED STRUCTURE, SYSTEM or Other COMPONENT within ANY Table H-1 area Condi-HA2.1 1 2 3 4 5 D HU2.1 1 2 3 4 5 D tions Affect- FIRE or EXPLOSION FIRE not extinguished within 15 min. of Control Room ing Plant Safety 2 None None resulting in EITHER:

VISIBLE DAMAGE to ANY SAFETY-RELATED STRUCTURE, SYSTEM or COMPONENT within ANY Table notification or verification of a Control Room FIRE alarm in ANY Table H-1 area or Turbine Building (Note 4)

H-1 area HU2.2 1 2 3 4 5 D Fire or Explosion OR Control Room indication of degraded performance of ANY EXPLOSION of sufficient force to damage permanent SAFETY-RELATED STRUCTURE, SYSTEM or structures or equipment within the PROTECTED AREA COMPONENT within ANY Table H-1 area HA3.1 1 2 3 4 5 D HU3.1 1 2 3 4 5 D 3 None None Access to ANY Table H-1 area is prohibited due to toxic, corrosive, asphyxiant or flammable gases which jeopardize operation of systems required to maintain safe operations or Toxic, corrosive, asphyxiant or flammable gases in amounts that have or could adversely affect NORMAL PLANT OPERATIONS Hazardous safely shutdown the reactor (Note 5)

Gas HU3.2 1 2 3 4 5 D Recommendation by local, county or state officials to evacuate or shelter site personnel based on an offsite event HG4.1 1 2 3 4 5 D HS4.1 1 2 3 4 5 D HA4.1 1 2 3 4 5 D HU4.1 1 2 3 4 5 D A HOSTILE ACTION has occurred such that plant A HOSTILE ACTION is occurring or has occurred within the A HOSTILE ACTION is occurring or has occurred within the A SECURITY CONDITION that does not involve a HOSTILE personnel are unable to operate equipment required to PROTECTED AREA as reported by the Security Site Owner Controlled Area as reported by the Security Site ACTION as reported by the Security Site Supervisor 4 maintain safety functions HG4.2 1 2 3 4 5 D Supervisor Supervisor OR OR A credible site-specific security threat notification A validated notification from NRC of an AIRLINER attack OR Security A HOSTILE ACTION has caused failure of Spent Fuel threat within 30 min. of the site A validated notification from NRC providing information of an Cooling systems aircraft threat AND IMMINENT fuel damage is likely 5

HS5.1 1 2 3 4 5 D HA5.1 1 2 3 4 5 D Control Room evacuation has been initiated Control Room evacuation has been initiated None None Control AND Room Control of the plant cannot be established within 15 min.

Evacuation HG6.1 1 2 3 4 5 D HS6.1 1 2 3 4 5 D HA6.1 1 2 3 4 5 D HU6.1 1 2 3 4 5 D Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial have occurred which involve actual or likely major failures of have occurred which involve an actual or potential have occurred which indicate a potential degradation of the 6 core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; (1) toward site personnel or equipment that substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring Judgment reasonably expected to exceed EPA Protective Action could lead to the likely failure of or; (2) that prevent effective HOSTILE ACTION. ANY releases are expected to be limited are expected unless further degradation of safety systems Prepared for Constellation by: Operations Support Services, Inc. - www.ossi-net.com (Rev. 22 1/18/13)

Guideline exposure levels (1,000 mRem TEDE or 5,000 access to equipment needed for the protection of the public. to small fractions of the EPA Protective Action Guideline occurs mRem thyroid CDE) offsite for more than the immediate site ANY releases are not expected to result in exposure levels exposure levels (1,000 mRem TEDE or 5,000 mRem thyroid area which exceed EPA Protective Action Guideline exposure CDE) levels (1,000 mRem TEDE or 5,000 mRem thyroid CDE) beyond the SITE BOUNDARY EU1.1 1 2 3 4 5 D EAL Identifier E None None None Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by measured dose rates > then ANY of the following:

• 400 mRem/hr at 3 feet from the HSM surface

• 100 mRem/hr outside HSM door on centerline XXX.X Category (R, H, E, S, F, C) Sequential number within subcategory/classification ISFSI Emergency classification (G, S, A, U) Subcategory number (1 if no subcategory)

• 20 mRem/hr end shield wall exterior MODE 4, 5 or D EPIP-EPP-02-EAL Modes: 1 Power Operation 2

Startup 3

Hot Shutdown 4

Cold Shutdown 5

Refuel D

Defueled Nine Mile Point Nuclear Station Unit 2 Revision 23 EAL Matrix Unit 2 Page 2 of 2

ATTACHMENT 4 License Amendment Request Change Emergency Action Level HU1 .5 to Remove High Lake Level Initiating Condition for Unusual Event Emergency Classification Markup of Proposed Emergency Action Level Technical Bases Pages NMP1 EAL Page 74 NMP2 EAL Page 81

jAttachment 4 Page 1 of 2 Exelon Confidential/Proprietary UNIT 1 EMERGENCY CLASSIFICATION TECHNICAL BASES EP-AA-1013 Addendum 3 Revision 1 Page 74 of 246 Attachment 1, Emergency Action Level Technical Bases (Continued)

Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 1 - Natural or Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the PROTECTED AREA EAL:

HU1.5 Unusual Event Lake water level> 248.2 ft OR Intake water level< 238.8 ft Mode Applicability:

All Basis:

Plant-Specific This threshold addresses high and low bay water level conditions that could be a precursor of more serious events (ref. 1, 2).

The high lake level is based upon the maximum attainable uncontrolled lake *.vater level as specified in the ~JMP 2 US,A,R. Dams on the St. Lawrence River, under the authority of the International St.

Lawrenee River Board of Control, are no'A' used to regulate the lake level. The low limit is set for el 74 .37 m (24 4 ft) on April 1 and is maintained at or above that elevation during the entire navigation season (April 1 to November 30). The upper limit of the lake level is el 75.59 m (248.2 ft) (ref. 3).

The low level is based on intake forebay level and corresponds to the minimum intake water level for operability of Emergency Service Water, Emergency Diesel Generator cooling water, Containment Spray Raw Water and Diesel and Electric FIRE Pump (ref. 4-9).

During planned evolutions such as intake water gate manipulation for reverse flow operations in which continuous monitoring of the intake level is being accomplished, entry into this EAL would not be warranted unless UNPLANNED /unexpected conditions and/or indications occur.

!Attachment 4 Page 2 of 2 Exelon Confidential/Proprietary UNIT 2 EMERGENCY CLASSIFICATION TECHNICAL BASES EP-AA-1013 Addendum 4 Revision O Page 81 of 264 Attachment 1, Emergency Action Level Technical Bases (Continued)

Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 1 - Natural or Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the PROTECTED AREA EAL:

HU1.5 Unusual Event Lake 1Nater level> 248.2 ft OR Intake water level < 237 ft Mode Applicability:

All Basis:

Plant-Specific This threshold addresses high and low lake water level conditions that could be a precursor of more serious events.

The high lake level is based upon the mm<imum attainable uncontrolled lake water le'4'el as specified in the USAR Dams on the St. Lawrence River, under the authority of the International St. Lawrence River Board of Control, are now used to regulate the lake level. The low limit is set for el 74.37 m (244 ft) on April 1 and is maintained at or above that elevation during the entire navigation season (April 1 to November 30). The upper limit of the lal<e le\*el is el 76.69 m (248.2 ft) (ref. 1).

The low level is based on intake water level and corresponds to the design minimum lake level. The probable minimum low water level of Lake Ontario at the site has been determined to be 72.0 m (236.3 ft) resulting from a setdown caused by a Probable Maximum Wind Storm concurrent with the lowest probable lake level. (ref. 2)

Generic This EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant operators.

This EAL addresses other site specific phenomena that can also be precursors of more serious events.

NMP2 Basis Reference(s):

1. USAR Section 2.4.1.2

2. USAR Section 2.4.11.2

3. N2-0SP-LOG-W001, Weekly Checks

4. NEI 99-01 IC HU1

ATTACHMENT 5 License Amendment Request Change Emergency Action Level HU1 .5 to Remove High Lake Level Initiating Condition for Unusual Event Emergency Classification Clean Emergency Action Level Bases Technical Pages NMP1 EAL Page 74 75 NMP2 EAL Page 81

!Attachment 5 Page 1 of 3 UNIT 1 EMERGENCY CLASSIFICATION TECHNICAL BASES EP-AA-1013 Addendum 3 Revision 2d Page 74 of 246 Attachment 1, Emergency Action Level Technical Bases (Continued)

Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 1 - Natural or Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the PROTECTED AREA EAL:

HU1 .5 Unusual Event Intake water level< 238.8 ft Mode Applicability:

All Basis:

Plant-Specific This threshold addresses low bay water level conditions that could be a precursor of more serious events (ref. 1, 2).

The low level is based on intake forebay level and corresponds to the minimum intake water level for operability of Emergency Service Water, Emergency Diesel Generator cooling water, Containment Spray Raw Water and Diesel and Electric FIRE Pump (ref.

4-9).

During planned evolutions such as intake water gate manipulation for reverse flow operations in which continuous monitoring of the intake level is being accomplished, entry into this EAL would not be warranted unless UNPLANNED /unexpected conditions and/or indications occur.

!Attachment 5 Page 2 of 3 UNIT 1 EMERGENCY CLASSIFICATION TECHNICAL BASES EP-AA-1013 Addendum 3 Revision 2d Page 75 of 246 Attachment 1, Emergency Action Level Technical Bases (Continued)

HU1 .5 Unusual Event (Continued)

Generic This EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant operators.

This EAL addresses other site specific phenomena that can also be precursors of more serious events.

NMP1 Basis Reference(s):

1. USAR Section 111-F Screenhouse, Intake and Discharge Tunnels

2. USAR Section X-F Service Water System

3. NMP 2 USAR Section 2.4.11.2

4. N1-ARP-H2 Annunciator H2-1-3

5. N1-SOP-18.1 Service Water Failure/Low Intake Level

6. S13.1-100F003

7. S14-93F003

8. S16.9NPSHAM002

9. Cale No. S14-93-F007

10. NEI 99-01 IC HU1

!Attachment 5 Page 3 of 3 UNIT 2 EMERGENCY CLASSIFICATION TECHNICAL BASES EP-AA-1013 Addendum 4 Revision 1d Page 81 of 264 Attachment 1, Emergency Action Level Technical Bases (Continued)

Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 1 - Natural or Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the PROTECTED AREA EAL:

HU1.5 Unusual Event Intake water level< 237 ft Mode Applicability:

All Basis:

Plant-Specific This threshold addresses low lake water level conditions that could be a precursor of more serious events.

The low level is based on intake water level and corresponds to the design minimum lake level. The probable minimum low water level of Lake Ontario at the site has been determined to be 72.0 m (236.3 ft) resulting from a setdown caused by a Probable Maximum Wind Storm concurrent with the lowest probable lake level. (ref. 2)

Generic This EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant operators.

This EAL addresses other site specific phenomena that can also be precursors of more serious events.

NMP2 Basis Reference(s):

1. USAR Section 2.4.1.2

2. USAR Section 2.4.11.2

3. N2-0SP-LOG-W001, Weekly Checks

4. NEI 99-01 IC HU1