Text

Revisions to the Exelon Nuclear Standardized Radiological Emergency Plan Implementing Procedure
	ML033210112
Person / Time
Site:	Dresden, Byron, Braidwood, Clinton, Quad Cities, LaSalle
Issue date:	11/05/2003
From:	Jury K; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RS-03-208
	Download: ML033210112 (68)
	v • d • e

ExeknaM Exelon Generation www.exeloncorp.com Nuclear 4300 Winfield Road Warrenville, IL 60555 10 CFR 50, Appendix E RS-03-208 November 5, 2003 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Braidwood Station, Units 1 and 2 Facility Operating License NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457 Byron Station, Units 1 and 2 Facility Operating License NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455 Clinton Power Station, Unit 1 Facility Operating License NPF-62 NRC Docket Nos. STN 50-461 Dresden Nuclear Power Station, Units 2 and 3 Facility Operating License DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249 LaSalle County Station, Units I and 2 Facility Operating License NPF-1 and NPF-18 NRC Docket Nos. 50-373 and 50-374 Quad Cities Nuclear Power Station, Units 1 and 2 Facility Operating License DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265

Subject:

Revisions to the Exelon Nuclear Standardized Radiological Emergency Plan Implementing Procedure In accordance with 10 CFR 50, Appendix E, Section V, "Implementing Procedures," Exelon Generation Company, LLC (EGC) and AmerGen Energy Company (AmerGen) are submitting changes to several Emergency Plan procedures.

November 5, 2003 U. S. Nuclear Regulatory Commission Page 2 Specifically, the following Exelon procedures have been revised:

> EP-AA-112-401, Nuclear Duty Officer"

> EP-AA-110-301 "Core Damage Assessment (BWR)"

> EP-MW-110-200 Dose Assessment."

Procedure EP-AA-112-401 has been revised to identify that the Nuclear Duty Officer (NDO) may be located in either the MidWest or MidAtlantic, but will respond to an event at any Exelon or AmerGen nuclear station. The activities of the NDO have been modified such an NDO in either region is capable of effectively responding to an event.

Procedure EP-MW-1 10-200 has been revised to reflect the adoption of the DAPAR dose assessment software. Corresponding changes to reflect the deletion of the previous dose assessment methodology have been incorporated. The adoption of this software allows a common dose assessment methodology and process between Exelon and the State of Illinois.

Procedure EP-AA-1 10-301 has been revised to reflect the implementation of the CDAM core damage assessment software at the Limerick and Peach Bottom stations. In addition, several administrative and editorial revisions were incorporated in response to user feedback.

The revised procedures are included in the attachments to this letter. These procedure changes were implemented between October 10 and October 22, 2003, and therefore must be submitted prior to November 9, 2003.

Should you have any questions concerning this letter, please contact Mr. T.W. Simpkin at 630-657-2821.

Repctfully, Keith R. Jury Director - Licensing and Regulatory Affairs Exelon Generation Company, LLC AmerGen Energy Company, LLC cc:

Regional Administrator - NRC Region IlIl (two copies)

NRC Senior Resident Inspector - Braidwood Station NRC Senior Resident Inspector - Byron Station NRC Senior Resident Inspector - Clinton Power Station NRC Senior Resident Inspector - Dresden Nuclear Power Station NRC Senior Resident Inspector - LaSalle County Station NRC Senior Resident Inspector - Quad Cities Nuclear Power Station Attachments:

Attachment A-Exelon Nuclear Procedure EP-AA-112-401, Nuclear Duty Officer" Attachment B - Exelon nuclear Procedure EP-MW-110-200, "Dose Assessment" Attachment C - Exelon Nuclear Procedure EP-AA-1 10-301, "core Damage Assessment (BWR)"

ATTACHMENT A EP-AA-112-401 NUCLEAR DUTY OFFICER

EP-AA-1 12-401 Exekon.

Revision 2 Pag 1 of 4 Nuclear Level 2 - Reference'Use NUCLEAR DUTY OFFICER (NDO)

1.

PURPOSE 1.1.

This procedure describes the responsibilities and actions of the Exelon Nuclear Duty Officer (NDO), which is a designated U24 17" duty position in either the Mid-West or Mid-Atlantic Region.

l When the Shift Manager decides that a situation warrants activation of the Emergency Response Organization (ERO) under the Emergency Plan, this procedure becomes applicable.

2.

TERMS AND DEFINITIONS None

3.

RESPONSIBILITIES 3.1.

- The Nuclear Duty Officer (NDO) is responsible for functioning as the initial Exelon Nuclear Corporate Management contact when an emergency event is classified at an Exelon Nuclear station. The NDO shall decide the appropriate response for events not classified under the emergency plan.

3.2.

The NDO is also responsible for interface with the State Duty Officers (or designated points of contact) regarding event information until the Corporate Emergency Director position is staffed.

4.

MAIN BODY 4.1.

INITIATE the appropriate Emergency Plan activities using the position checklist contained in Attachment 1.

5.

DOCUMENTATION None

6.

REFERENCES None

7.

ATTACHMENTS 7.1., Nuclear Duty Officer (NDO) Checklist

EP-AA-1 12-401 Revision 2 Page 2 of 4 ATTACHMENT I NUCLEAR DUTY OFFICER (NDO) CHECKLIST Page 1 of 3 Section 1, Event Classification Section 2, Notification of Transportation Accident Section 3, Hazardous Materials Emergency Section 4, Activation of Electric Operations Emergency Load Conservation Program l

1.

EVENT CLASSIFICATION NOTE:

The NDO will receive an alpha page message from the Emergency Response Organization (ERO) Callout System identifying the affected station, event classification,.and facilities. being activated.

1.1.

CONTACT the affected Station Duty Manager to VERIFY and OBTAIN

updated information concerning emergency response actions and event status.

1.2.

NOTIFY the Nuclear Duty Executive (NDE) and the Chief Nuclear Officer (CNO).

1.3.

NOTIFY the Exelon Nuclear Regional Communications Duty Officer of the event.

1.4..

REVIEW news releases created by Exelon Generation Communications & Public Affairs for accuracy prior to activation of the Emergency Public Information Organization.

1.5.

RESPOND to requests for information concerning the event from the State Duty Officer(s), if contacted.

1.6._

UPDATE the Exelon Nuclear Management, using the appropriate method listed below, until the EOF is in Command and Control or the event has been terminated:

1.6.1.

MAINTAIN the NDO Message voice mailbox up to date.

1.6.2.

If the event involves a Mid-Atlantic station, then UPDATE the Employee Emergency Notification Line for the affected station, using the instructions found in EP-MA-110-1 00.

1.7.

MAINTAIN a record of activities.

EP-AA-112-401 Revision 2 Page'3 of 4 ATTACHMENT I NUCLEAR DUTY OFFICER (NDO) CHECKLIST Page 2 of 3 1.8.

For an Alert or higher classification, PERFORM the following:

1.8.1.

NOTIFY industry support organizations per the Reportability Manual, using the contact numbers listed in the ERF Telephone Directory:

Institute of Nuclear Power Operations (INPO)

American Nuclear Insurers (ANI) 1.8.2.

TRANSFER responsibility for notification of INPO and ANI to the EOF Logistics Manager once the EOF is activated.

Mid-West Region lCantera NDO may relocate to the ECIF to continue to assist with notifications to ANI, lINPO and other support /governmental agencies.l

2.

NOTIFICATION OF TRANSPORTATION ACCIDENT NOTE:

A Transportation Accident is defined in 49 CFR 171.15 and 49 CFR 171.16.

2.1.

REVIEW OP-AA-106-102, "Accidents or Incidents Involving the Transportation of Rad Material."

2.2.

CONTACT the affected Station Duty Manager to get information on the event.

2.3.

CONTACT Exelon Nuclear Communication Services and coordinate press releases if necessary.

2.4.

CONTACT INPO to initiate the Voluntary Assistance Program, if necessary.

2.5.

CONTACT American Nuclear Insurers (ANI) to provide'them with status of the event.

2.6.

MAINTAIN a record of activities.

I EP-AA-1 12-401 Revision 2 Page 4 of 4 ATTACHMENT I NUCLEAR DUTY OFFICER (NDO) CHECKLIST Page 3 of 3 MID-WEST Region

3.

ACTIVATION OF ELECTRIC OPS EMERGENCY LOAD CONSERVATION PROGRAM 3.1 NOTIFY Distribution Dispatch Center (DDC)/Load Dispatcher that a station is in an emergency condition. DDC will maintain power to that station's Emergency Planning Zone (EPZ) for the Alert and Notification System (ANS) sirens.

3.2 NOTIFY the ANS siren maintenance vendor of possible rolling power outages, and INSTRUCT the vendor to monitor siren performance.

3.3 TRANSFER the ELCP Bridge phone call to the Corporate Operations Center.

3.4 UPDATE the NDO Message voice mailbox. MAINTAIN the NDO Message voice mailbox up to date until the EOF is activated or the event has been terminated.

3.5 If support from the Site Restoration Management Team at a nuclear station is necessary, then COORDINATE plant access for the Team.

3.6 If the ANS siren maintenance vendor was notified of rolling power outages and the outages have been terminated, then NOTIFY them that the outages have been terminated.

3.7 If DDC/Load Dispatch was notified of an emergency condition, and the emergency is terminated, then NOTIFY the DDC/Load Dispatch that there is no longer an emergency condition at the affected station.

ATTACHMENT B EP-MW-110-200 DOSE ASSESSMENT

EP-MW-1 10-200 Revision 2 Exe___________

Page !1 of 34 Nuclear Level 2 - Reference Use DOSE ASSESSMENT

1.

PURPOSE This procedure provides the methods and instructions for performing offsite dose assessment and projection by the Emergency Response Organization.

-~~~~~~~~~~

2.

TERMS AND DEFINITIONS 2.1 Centerline (plume): An imaginary line drawn in the middle of the plume along its downwind travel direction. The plume concentrations and deposition are assumed to be the highest along the centerline 2.2 Cloud Shine: Gamma radiation from radioactive materials in the air (plume) 2.3 Committed Dose Equivalent (CDE): The internal dose equivalent to parts of the body that will be received from an intake of radioactive material by an individual over a 50-year period of time.

2.4 Committed Effective Dose Equivalent (CEDE): The sum of the internal

. dose equivalent for 50 years following intake (inhalation or ingestion) of a rapionuclide to each organ multiplied by a weighting factor.

2.5 Core Damage: Damage to the components that comprise the reactor core.

Core damage typically refers to the failure of fuel cladding and/or fuel melting as a result of overheating.

2.6 Curie (Ci): A unit of radioactivity equal to 3.7E+10 disintegrations per second.

2.7 DAPAR

Exelon Dose Assessment and Protective Action Recommendations (DAPAR) software provides two major functions (Quick Assessment and Full Assessment) in order to perform dose assessment.

A. Quick Assessment is used by the Control Room to arrive at offsite dose projections and PARs, or to verify classifications in as'quick a time as' possible during fast breaking events without taking too much time away.

from their event mitigating actions. A monitored release is the only method used in the quick assessment. Some assumptions and standard numbers are used to limit the amount of data Control Room personnel must enter prior to calculating a PAR.

B. Full Assessment is used by the called-in ERO Staff in the TSC/EOF and allows for more detailed assessment of a release. The following methods may be used to project offsite doses:

EP-MW-1 10-200 Revision 2 Page 2 of 34 Monitored Release: Offsite radiological assessment related to a monitored value taken at one of several release locations (Plant Vent Stack, Waste Processing Vent Stacks and Turbine Building Vent Stack) within the plant.

Containment Leakage/Failure: Offsite radiological assessment related to a default, known, or predicted level of containment leakage or failure.

Field Team Survey and Sample Analysis: Offsite radiological assessment related to comparisons of field team radiological survey and isotopic sample concentrations with predicted plume dispersion.

Release Point Sample Analysis: Offsite radiological assessment related to a measured isotopic concentration taken at the point of release to the environment.

GEMS Analysis: Allows for offsite dose assessment using isotopic release rates (uCi/sec of each isotope) provided by the State of Illinois (Division of Nuclear Safety) Gaseous Effluent Monitoring System..

2.8 Delta T: The difference in temperature from the lower temperature sensor and the upper temperature sensor on the Exelon meteorological tower. Delta-T is used to calculate stability class.

2.9 Deposition

The contamination found on the surface of the ground.

2.10 Dose Commitment: The dose that will be accumulated by a specific organ over a specified period following uptake.

2.11 Dose Conversion Factor (DCF): The dose equivalent per unit intake of a radionuclide (mrem/uCi) or the effects of exposure to a given concentration of an isotope in a plume. R/hr per uCi/cc.

2.12 Dose Proiection: The calculation of individual radiation exposure at a given location at some time in the future. Dose projections are performed in response to an actual or anticipated release of radioactive material to the environment.

2.13 Effective Dose Equivalent (EDE): The sum of the dose equivalent from external exposure to each organ multiplied by a weighting factor. EDE is used to estimate the risk of delayed health effects.

2.14 2.15 2.16 2.17 2.18 2.19 2.20 221 2.22 EP-MW-1 10-200 Revision 2 Page § of 34 Emergency Planning Zone EPZ): An area around a nuclear power plant in which plans are in place for an emergency at the plant.. Plans are in place to take immediate protective actions for individuals located within 10 miles of the Nuclear Plants. This area is called the Plume Exposure Emergency Planning Zone. In addition, longer-term plans are in place for the Ingestion Pathway Emergency Planning Zone which is within 50 miles of the plant.

Evacuation ExPosure Period: The period during which those being evacuated are exposed to the radioactive plume.

I Millirem (mR): One one-thousandth of a Rem. The Rem is a unit of measure that defines the extent of biological injury that results to the body when it is' exposed to radiation.

Pilot Operated Relief Valve PORV): A valve which serves to reduce pressure in the reactor coolant system or main steam system by allowing steam to escape from the Pressurizer or the steam generators. The PORVs can be operated remotely by Plant Operators or automatically by high pressure.

Plant Parameter Display System (PPDS): Electronic graphical display of plant, meteorological and radiological data needed for accident and dose assessment.

Protective Action Guidelines (PAGs): Radiation exposure guidelines established by the Environmental Protection Agency which are used to determine'the appropriate protective actions to be taken on the part of.

emergency workers and the general public. These actions include sheltering and evacuation.

Protective Action Recommendations (PARs): A recommendation made by Exelon personnel to the offsite authorities on the appropriate protective actions to be taken on the part of the'general public. The PARs are'based on plant conditions or dose projections using the PAGs for guidance.

Safety Relief Valve: A valve that serves to reduce pressure in a fluid system should the pressure become too high. Both the reactor coolant system (located on pressurizer) and the main steam system (located on'steam generators) have safety'and relief valves to protect them from being damaged by excessive pressure.

Site Boundarv:' Defined as a circle with a radius of or % mile (depending on the site specific ODCM) and the containment building as its center.

EP-MW-110-200 Revision 2 Page 4 of 34 2.23 Station Vent: That part of the plant's ventilation system through which the containment building and auxiliary building air may be processed to the outside atmosphere. The discharge of the station vent is continuously monitored for abnormal amounts of radiation and would be isolated long before radiation levels approach federal limits.

2.24 Subareas: Pre-designated areas offsite in which Protective Actions such as evacuation of sheltering will be performed.

2.25 Total Effective Dose Equivalent (TEDE): A method of converting exposure to radiation to the biological effects that it will cause to the human body. It combines the external and internal ionizing radiation exposure. The TEDE is the sum of Deep Dose Equivalent and Committed Effective Dose Equivalent.

3.

RESPONSIBILITIES 3.1.1 The Shift Manager, or designated on-shift individual, shall perform required dose assessments prior to responsibility being transferred to either the Technical Support Center (TSC) or Emergency Operations Facility (EOF).

3.1.2 The TSC Radiological Controls Coordinator shall relieve the Control Room and perform required assessments if the transfer of PAR / dose assessment responsibilities to the EOF is delayed.

3.1.3 The EOF Dose Assessor shall relieve the TSC Radiological Controls Coordinator when directed by the EOF Dose Assessment Coordinator, and perform required dose assessments. Responsibility for dose assessments can be assumed directly from the Control Room.

4.

MAIN BODY 4.1 Dose Assessment and Protective Action Recommendation (DAPAR)

REFER to Attachment 1 for user guidelines.

4.2 Obtaining IDNS Monitor Data from the EOF REFER to Attachment 2 for user guidelines.

5.

DOCUMENTATION None

EP-MW-1 10-200 Revision 2 Page 6of34

6.

REFERENCES 6.1 EP-MW-1 23-1002, "Dose Assessment and Protective Action Recommendation (DAPAR) Program Technical Basis".

7.. -

ATTACHMENTS 7.1, DAPAR Users Guide 7.2, Obtaining IDNS Monitor Data from the EOF

EP-MW-1 10-200 Revision 2 Page 6 of 34 ATTACHMENT I DAPAR USERS GUIDE Page 1 of 27

1.

OVERVIEW 1.1 As a Windows based application designed in Access; DAPAR, uses many standard user interfaces. Instructions are not provided in basic computer operations in the Windows environment. The user must be familiar with these to efficiently operate the program.

l 1.2 It is also assumed user is familiar with health physics fundamentals.

Emergency Response Organization training will provide an overview of dose assessment methodologies.

1.3 DAPAR Procram Use: The program is to be used to estimate the offsite consequences of a release or potential release of radioactive materials from an Exelon Station during an emergency. The primary purpose of these dose projections is to arrive at a Protective Action Recommendation given by Exelon management to offsite authorities. These PARs'will be used by those authorities in their decision making process to take actions to protect the general public.'

1.4 Limitations and Pre-Conditions for use DAPAR:

1.4.1 The program should not be used to calculate the actual dose received by populations exposed to radioactive materials from a release. Results may be used as part of the post accident investigations, but a much more in-depth analysis is needed to actually assign doses received by members of the public.

1.4.2 DAPAR should be used only when an emergency has been. The program makes many conservative assumptions.to ensure proper actions are taken offsite prior to exposing the general public to any release of radioactive materials.

1.4.3 DAPAR release paths are based on the generalized BWR/PWR gaseous effluent pathways described in NUREG-1228. The Process Reduction Factor (PRF) of any additional release path(s) should be approximated by using one of the existing DAPAR pathways.

NOTE:

Use of the DAPAR program to project doses based on routine plant readings would indicate offsite doses many magnitudes higher than actual offsite doses. Care should be taken in'making a Protective Action' Recommendation or notification based on DAPAR output if there are no indications of Core Damage (ODCM calculations should be used to calculate offsite doses when no core damage is expected).

EP-MW-1 10-200 Revision 2 Page of 34 ATTACHMENT I DAPAR USERS GUIDE Page 2 of 27 NOTE:

There is a different version of DAPAR for each Station. Each version takes into account diff6rent release points and gaseous effluent radiation monitor conversion factors. Always verify that the correct station version of the program is being used.

2.

START UP 2.1 Start the computer.

2.2 The application is accessed by one of the folowing:

2.2.1 Open the DAPAR folder desktop icon on applicable dose assessment computers.

1. Start the appropriate DAPAR program for the plant that has declared an emergency.

2. Programs are labeled <Station Name> - DAPAR v2.0.

2.2.2 Select RUN from the the 'Start Bar' and type in the file path and name as follows:

C:\\DAPAR\\Braidwood DAPAR v2_O.mdb C:\\DAPAR\\Byron DAPAR v2_O.mdb C:\\DAPAR\\Clinton DAPAR v2

.mdb C:\\DAPAR\\Dresden'DAPAR v2

.mdb C:\\DAPAR\\LaSalle DAPAR v2 O.mdb C:\\DAPAR\\Quad Cities DAPAR v2

.mdb 2.3 IF the assigned Dose Assessment Computer cannot access the application or the DAPAR program will not run, THEN Install DAPAR on any computer from CDs or Disks located in the Control Room, TSC or the EOF Library. DAPAR is installed by copying appropriate file to computer's hard drive.

EP-MW-1 10-200 Revision 2 Page 8 of 34 ATTACHMENT 1 DAPAR USERS GUIDE Page 3-of 27

3.

BASIC PROGRAM FLOW DIAGRAM Repo PAR-Report Dose Projection - Reports Receptor-Reports (full only)

I The above diagram shows basic tasks that can be performed by the DAPAR program and how a user would navigate between them.

NOTE:

Values for specific plant data is obtained from the Plant Parameter Display System (PPDS) or the affected unit's Control Room.

EP-MW-1 10-200 Revision 2 Page of 34 ATTACHMENT I DAPAR USERS GUIDE Page 4 of 27

4.

TITLE SCREEN The title screen showvs the application version and offers the user three opti6ns to direct program flow.

4.1 Quick Assessment - This option is designed to be used by the Control Room and TSC.

It performs assessments based on design basis default source terms.

4.2 Full Assessment - This option is designed to be used by the Dose Assessment Staff in the TSC and EOF. It allows user more options in performing calculations.

4.3 Quit - Exits the Program..

-1,

In I

ly~~~Of

~~ Dos Assc~esmet NOTE:

-Once the User selects "Quick Assessment" or "Full Assessment," returning to the title screen will reset all program values.

4.4 SELECT either "Full Assessment" or 'Quick Assessment" and then GO TO either:

4.4.1 Section 5 of this procedure for Quick Assessment.

4.4.2

Section 7 of this procedure for Full Assessment.

EP-MW-110-200 Revision 2 Page 10 of 34 ATTACHMENT 1 DAPAR USERS GUIDE Page 5.of 27

5.

QUICK ASSESSMENT The Quick Assessment operations and calculations are identical to the Full Assessment method for a monitored release, but utilizes default release path and core damage assumptions for the determination of offsite doses. This allows for a rapid assessment from the Main Control Room.

Selecting Main Steam Line' activates Relief Inputs' (Only appears on PWR Stations)

.d S

.e 4

ed MPH)ji

___OrTi i

te-.T~em1 Wind Detion (F-romi ILJ1(L~i/secV Ii rnRV~Oj~~n~flt tabsl Class (A-G),'

I al 4..

Jaadsk.I4*(a

])

Allows input from 2 release points. Will be Units for PWRs or Main Stack (Chimney.) or Rx Bldg Vent for BWRs NOTE:

User must choose the appropriate Channel, "Lo", "Mid",- or Hi". The same pCi/sec reading for the different channels can make a significant difference in projected doses offsite. Channel selection information is obtained from the Process Book Plant Data Displays or the Control Room.

5.1 Monitor Information - User chooses the appropriate monitor from the listed effluent monitors.

5.1.1 For PWRs the following choices are available:

1. Aux Bldg. Vent -'SELECT this option for releases from the Auxiliary Building Vent.

EP-MW-110-200 Revision 2 Page 1 of 34 ATTACHMENT 1 DAPAR USERS GUIDE Page 6 of 27

2. Main Steam Line - SELECT this option for releases from the Main Steam Line.

The Main Steam Line radiation monitors read out in mr/hr, therefore a flow has to be determined to calculate the uCi/sec release rate.

This is accomplished by entering the steam pressure and the number of relief valves that are open in the section labeled Relief Inputs.

5.1.2 For BWRs, normally there will be no choice available or needed here..

Dresden Station DAPAR allows choice for a Unit 1 release point.

5.2 ENTER Reading Information - Enter the appropriate monitor reading in pCi/sec or mr/hr. Choose the appropriate channel, "Lo", "Mid" or "Hi". Both pCi/sec readings should be entered, if one reading is 2 1000 times the other, 0 may be entered for smaller number.

IF using Main Steam Line monitor release path THEN enter Steam Generator Pressure, Number of Relief Valves Open and PORV open or shut.

5.3 ENTER Time After Shutdown Information - Enter the time since the reactor was shutdown in hours and minutes (hh:mm).

5.4 ENTER Meteorological Data - Enter the appropriate data from plant instruments as follows:

NOTE:

Met Data is available from the Plant Parameter Display System (PPDS) screens or the Control Room. If Met Tower data is unavailable from these locations, another source of meteorological data may be used such as the Meteorological Vendor, National Weather Service or a local TV or Radio broadcast stations. The stability class can be estimated from the Table 1-1 or Table 1-2 if Meteorological Vendor provides AT/Az or ao:

-5.4.1 Wind Speed (MPH) - Obtain and enter wind speed in Miles per Hour (MPH) 5.4.2 Wind Direction (From) - Obtain and enter the direction the wind is coming FROM in degrees. (0°-3600)

EP-MW-1 10-200 Revision 2

. Page 12 of 34 ATTACHMENT I DAPAR USERS GUIDE Page 7.of 27 5.4.3 Stability Class (A-G) - Obtain and enter the stability class.

A, Extremely unstable conditions B, Moderately unstable conditions C, Slightly unstable conditions D, Neutral conditions E, Slightly stable conditions F, Moderately stable conditions G, Extremely stable conditions If the stability class is not available use the following tables to choose appropriated value:

'Table

'Stability Class Determinato

'on Daytime Solar Radiation Surface (For moderate cloud cover move Nighttime Conditions Wind Speed one column to the right)

(mph)

Summer Spring/Fall Heavy Thin overcast

< 3/8 Heavy Sky Clear Sky Overcast Winter

(>112 cloud cloud Overcast

'Clear Sky Sky.

Rain cover) cover Rain

< 9.0 A

A-B 0

B 0

to 9.0 A-B B

D C

E F

D to 13.5 B

B-C D

C D

E D

> 13.5 C

C-D 1

D D

9,,.<..,Table 1-2.'< - '

'I S tability Class Determinati n-.

(deg rees)

A

<-1.9 222.5 B

> -1.9.to*< -1.7

<22.5 to217.5 C

> -1.7 to

-1.5

<17.5 to 12.5.

D

> -1.5 to

-0.5

<12.5 to 27.5 E

> -0.5 to

+1.5

<7.5 to 23.8 F

> +1.5 to* +4.0

.<3.8 to 2.1 G

> +4.0 0.0 to <2.1

EP-MW-1 10-200 Revisioh 2 Page 13 of 34 ATTACHMENT 1 DAPAR USERS GUIDE Page 8 of 27 NOTE:

The conditions Good or Adverse in the following step are in relation to the weather. Adverse is heavy rain or any other condition that would hinder the flow of traffic.

5.5 Set Evacuation Conditions - Click l

on the Conditions button to open dilib:~<jsTim of Da.'

the Weather Conditions Wlndow.

5.6 SELECT the appropriate conditions

D8Yt, 6..

for the program to calculate the,~p~EeigjriiSM Maximum Evacuation Time

/>Time of Week Estimate (ETE). Once the user sets B

the evacuation conditions, the.

g j

.2&

l program will place the Max ETE lI value in the Release Duration and Alle Max ETE text boxes.

Not all choices are available for every station or condition. The Program only allows appropriate choices.

5.7 IF the exact release duration is known, THEN change the displayed time to the known release duration. If a.good estimate of the release duration cannot be determined, use the default ETE value entered by the program.

5.8 SELECT the PARs button - The program will calculate the downwind doses based on user inputs and display Protective Action Recommendation Window.

EP-MW-1 10-200 Revision 2 Page 14 of 34 ATTACHMENT I DAPAR USERS GUIDE Page 9 of 27 PROTECTIVE ACTION RECOMMENDATIONS The Protective Action Recommendation (PAR) form displays a summary of the downwind dose projections with a map showing which Subareas (colored areas) where Protective Actions Recommendations should be made.

6.

6.1.

I 0-MM-M-Mr. R.;

.1] Monior eadig (QuIck)

IR I

> dEvacuation'ConditionsI.

17 1.2 A

12 1n 23 3

27 RI or:

- 400-;

!'.,I i;_Ii-_1 P!

1

'.I I

"I

! -1 NI 1.;

Downwind Subareas colored only if PARs should be

-made Displays

\\

Isotopic Group

.Release Rates t!,.

I--

14 20 2D 25 I

I R;~].

4 f

a,,,

~~ I:

_ _ _ _ _ _ _ __C E I hy in g l e s n o e i nK-6.2 Explanation of displayed data:

6.2.1 Assessment Method - Method used to calculate downwind doses.

6.2.2 Evacuation Conditions - Entered evacuation conditions and meteorological data, along with Release Duration (form displays the meteorological data use to determine PAR).

6.2.3 Subareas to be Evacuated - Form displays an EPZ map. Subareas where the population may receive doses exceeding a PAG are colored.

6.2.4 Affected Areas - This is the downwind Subareas that are affected by the release.

EP-MW-1 10-200 Revision 2 Page 16'of 34 ATTACHMENT I DAPAR USERS GUIDE Page 10 of 27 6.2.5 TEDE and CDE Thy - Shows the highest doses (no protection and sheltered) in Rem for each Ring.

6.3 6.4 7.

The RR button vill display the total release rates for isotopic groups in Ci/sec.

User can SELECT Print to print the PAR report or SELECT the Go Back button and modify inputs.

This will return user to either Quick Assessment Form or one of the Assessment Method forms available in the Full Assessment mode.

~ Noble Gases:>, 5.24E+/-02; Part&~1ltesY 1:'.

O.OE+OO, I

FULL ASSESSMENT The Full Assessment operations and calculations are identical to the Quick Assessment method for a monitored release, but it allows the user to make more choices in performing dose projection calculations.

Choosing the Full Assessment option directs the program to a baseline data entry window. The window is divided into four input areas.

_____________________________I__________________________________

I ~K;1111 Only available in I PWR DAPARs

][fneactof ffore A;;idn]

_Type Damage:

ayMel Amotz~~~~~~~~~~tot~~~~~~amage(~~~~~~~......

I )a 1 1 I:

I -I -

4 !:,.; 7- %-

-.-

-- , %-

_.., -: -.. 'I!, 5 1 I

r¢ Spent Fuel Accident.- CNewfm Old FuelStatus> l i.UndeiV'aterj -

Did iWaste Gas Decay Tank Accident

~~~~~

~~~

.Z-

%so.

-- Sp 1

'; HausAfeS/D fh'mr}

"'j-r....

-22 4 Wind Speed [MPHL I1.0 jIVeci tio!

N1one j IIT. r i onitored Release.

41 Not Entered 4,PRF.

-B___

7.1 71 Source Term - This allows user to choose the appropriate source term depending on plant conditions and the type of accident that has occurred.

EP-MW-1 10-200 Revision 2 Page 16 of 34 ATTACHMENT 1 DAPAR USERS GUIDE Page 11 of 27 7.1.1 SELECT Reactor Core Accident if the source of the release is from the reactor core.

SELECT Gap or Melt and ENTER the % Damage based on core damage estimates or known conditions in the plant.

1. SELECT Fuel Handling Accident if the release is caused by damage to the spent fuel.

CHOOSE between New Fuel or Old Fuel based on the type of fuel that has been damaged. New Fuel refers to fuel freshly transferred from the reactor to the fuel pool (i.e.,. program uses an estimate of the minimum time after shutdown allowed by the station before beginning transfer of fuel from the reactor). Old Fuel refers to fuel transferred to the fuel pool at an earlier outage (i.e., program uses a time after shutdown corresponding to the estimated minimum time between refuel outages).The program uses a gap release scenario and defaults to a reactor Time After Shutdown based on this choice.

NOTE:

The choice of Waste Gas Decay Tank Accident is only available for PWRs.

2. SELECT Waste Gas Decay Tank Accident if the release is caused by damage to a failure of a waste gas decay tank. The program sets source to one failed Waste Gas Decay Tank inventory.

7.2 Dominant Release Path - This allows user to choose the most appropriate release path:

NOTE:

DAPAR release paths arebased on the generalized BWR/PWR gaseous effluent pathways described in NUREG-1228. The Process Reduction Factor (PRF) of any additional release path(s) should be approximated by using one of the existing DAPAR pathways.

EP-MW-1 10-200 Revision 2 Page 1 of 34 ATTACHMENT 1 DAPAR USERS GUIDE Page 12 of 27 7.2.1 PWR Paths:

a; Sii i iill

.,~~~~~~~~~~~~

..e

[

PC 11

~ ~

~ ~ ~ ~ ~ ~ ~ ~

uu.0 7.2.2 BWR Paths:

NOTE:

Dresden Station DAPAR has one additional path (N) through the Isolation Condenser.

~.

T4-

+

t t f

, z Ls Di.L.....______________________

J-i',

EP-MW-1 10-200 Revision 2 Page 18 of 34 ATTACHMENT 1 DAPAR USERS GUIDE Page 13 of 27 7.2.3 Depending on the 'path chosen, the user will be presented with more choices to pick the appropriate Process Reductions for the release.

1. If a release through the Steam Generators is chosen:

Determine the status of the secondary side of the steam generator and select the appropriate condition.

2. If a release'through containment or drywell is chosen:

I

First determine if containment sprays have been used since release of fission product materials to Containment.

A l

.

-I U

I iSecondarsiBoiringJ..

I

';.

4() Se&ndaryDry 1

i -: -,-

i ! ,-

1 I

" I

,. - -

1 :

ontahmn SWy'if~

I.

onha~rrent Spray'O`J' 1114 -, J., '. "

.. eE, -,.

Representative Screens similar for PWR or BWR s

r,.

Then determine Containment / Drywell holdup time.

Q()2'24H6tw I(3h~24HoPM:

- 3. If a release through the Aux Bldg or Rx Bldg.

Determine the RAB holdup time.

U U

U

2A

1-

QSpecificFfodUpTuie 9

A'

EP-MW-110-200 Revision 2 Page 10 of 34 ATTACHMENT DAPAR USERS GUIDE Page 14of27

4. If a release through a filtered Vent or SBGT.

Choose if the filters are working or not. 'i Re Wok(e. rAg:'SlBd

.ilter; NoF WodrjNg (eag, LOCA o;rd contarnv.

If the release has been ongoing for a long time or contains a large amount of liquids filters may not be working.

NOTE:

BWR DAPARs also have a choice for Torus or Suppression Pool status, uSubcooledl,"Saturated"' or "Bypassed".

7.3 ENTER Meteorological Data - Enter the appropriate data from plant instruments.

NOTE:

'Met Data is available from the Plant Parameter Display System (PPDS) screens or the Control Room. If Met Tower data is unavailable from these locations, another source of meteorological data may be used such as the Meteorological Vendor, National Weather Service or a local TV or Radio broadcast stations. The stability class can be estimated from the Table.1-1 or Table 1-2 if Meteorological Vendor provides AT/Az or 0:

7.3.1 Wind Speed (MPH) - Obtain and enter wind speed in Miles per Hour (MPH).

7.3.2 Wind Direction (From) - Obtain and enter the direction the wind is coming FROM in degrees. (0-3600) 7.3.3 Stability Class (A-G) - Obtain and enter the stability class.

A, Extremely unstable conditions B, Moderately unstable conditions C, Slightly unstable conditions D, Neutral conditions E, Slightly stable conditions F, Moderately stable conditions G, Extremely stable conditions

EP-MW-1 10-200 Revision 2 Page 20 of 34 ATTACHMENT I DAPAR USERS GUIDE Page 15 of 27 If the stability classis not available use the following tables to choose appropriated value:

..* Table 1-1 Stability Class Determination Daytime Solar Radiation Surface (For moderate cloud cover move Nighttime Conditions Wind Speed one column to the ight).

(mph)

Summer SpringFall vHeavy.

Thin overcast

< 3/8 Heavy (mph SClerk Clengarl Overcast Winter

(>112 cloud cloud Overcast Rain cover) cover Rain

< 9.0 A

A-B

.D B

D to 9.0 A-B B

D C

E F

D to 13.5 B

B-C D

C D

E

.D

>13.5 C

C-D D

.D D

D

,Table 1-2' Stability Class Determination Clas ATI,&z a

9 '-

CasS(

0C lO.,m)

(degrees)t'-

A

< -1.9 222.5 B

> -1.9 to

-1.7

<22.5 to 17.5 C

>-1.7 to

-1.5

<17.5 to 12.5 D

> -1.5 to

-0.5

<12.5 to 7.5 E

> -0.5 to

+1.5

<7.5 to 3.8 F

>+1.5' to*+4.0

<3.8toŽ2.1.

G

>+4.0 0.Oto<2.1 7.4 Set Precipitation Conditions - Click on the Change Precipitation button to open Precipitation Window.

None - No rain or snow Light Rain - Light Drizzle < 0.1 inchs per hour

~'66ht Rai,6 Moderate Rain -

Heavy Drizzle 0.1 to 0.3 inches per hour

.}ModezatefRain Heavy Rain - Greater than 0.3 inches per hour eFHeavv Rain Light Snow - Visibility 0.63 miles or greater

iht Snow Moderate Snow - Visibility 0.31 to 0.63 miles

.Moderaie Snow Heavy Snow - Visibility < 0.31 miles

EP-MW-1 10-200 Revision 2 Page 21 of 34 ATTACHMENT I DAPAR USERS GUIDE Page 16 of 27 NOTE:

The conditions Good or Adverse in the following step are in relation to the weather. Advers6 is heavy rain or any other condition that would hinder the flow of traffic.

7.5 Set Evacuation Conditions - Click on the Conditions button to open the Weather Conditions Window.

Condition..

ThneofDar.

SEthe appropriate conditions Go -'

for the program to calculate the, Maximum Evacuation Time Estimate Time of We.k (ETE). Once the user sets the Weka evacuation conditions, the program will place the Max ETE value in the l

i Release Duration'and Max ETE text

' boxes.

Not all choices are available for'every station or condition. The Program only allows appropriate choices.

7.7 IF a good estimate of the release duration (i.e., how long release is expected to last) cannot be determined, use the default ETE value entered by the program. The release duration is then assumed to be the time it takesfor the population to be evacuated from the affected area. IF the exact release duration is known, THEN change the displayed time to the known release duration.

[

NOTE:

User. may switch back and forth between assessment methods as more information becomes available or conditions change. With the exception of the Time After Shutdown, which updates each time user returns to main form, the data on the Full Assessment form will not change unless user changes it.

7.8 Assessment Methods - CHOOSE the appropriate assessment method based on available inputs. The Assessment methods are:

7.8.1 Monitored Release - SELECT this method for a release through a plant vent or through the Main Steam Relief Valves. GO TO Section 8.

7.8.2 Containment Leakage - SELECT this method for containment failure scenarios. GO TO Section 9.

7.8.3 Field Team Data - SELECT this method if field team survey or sample data is available. GO TO Section 10.

EP-MW-1 10-200 Revision 2 Page 22 of 34 ATTACHMENT I DAPAR USERS GUIDE Page 17 of 27 7.8.4 Release Path - SELECT this method for a sample of a release has been obtained and a release flow rate can be estimated. GO TO Section 11.

7.8.5 GEMS Analysis-SELECT this method for when entering release rate.from the GEMS monitors. GO TO Section 12.

8.

MONITORED RELEASE Sample of Monitor Release Monito Rei In u PAG i';ance mile:

Screen.

v

~

-Ei 0

Each J

.E Station's M-eine J

I CD i

DAPAR has

_-_i_.*_____________Jr R~~~~~~~~~~~~~~el Our (hbbJ slight 1

(IS)~

O0E.21" veqrinqtinnr.

EP-MW-110-200 Revision 2 Page23 of 34 ATTACHMENT I DAPAR USERS GUIDE Page 18 of 27 8.1 CHOOSE the appropriate monitor:

8.1.1 For PWRs

1. Aux Bldg. Vent - SELECT this option for releases from the Auxiliary Building Vent.

2. Main Steam Line - SELECT this option for releases from the Main Steam Line.

The Main Steam Line radiation monitors read out in mr/hr, therefore a flow has to be determined to calculate the uCi/sec release rate.

This is accomplished by entering the steam pressure and the number of relief valves that are open in the section labeled Relief Inputs.

8.1.2 For BWRs, normally there will be no choice available or needed here.

Dresden Station DAPAR allows choice for Unit 1 release point.

NOTE:

User must choose the appropriate Channel, "Lo", "Mid", or uHi". The same.

pCi/sec reading for the different channels can make a significant difference in projected doses offsite. Channel selection information is obtained from the Process Book Plant Data Displays or the Control Room.

8.2 Input Reading Information - User enters the appropriate monitor reading in uCi/sec or mr/hr. IF two (2) units or vent entries are shown a value must be entered in each before program calculates offsite dose (enter 0 if no release).

8.3 After User enters data the program calculates offsite doses. The user can now perform one of the following items:

8.3.1 SELECT the Print Receptor Report button to print a report of projected dose rates, TEDE and CED Doses and Ground Deposition at pre-designated receptor points.

8.3.2 SELECT the Print button to print a report of offsite dose projections based on the monitored release.

8.3.3 SELECT the Back button to change input data on the Full Assessment Form.

8.3.4 SELECT the PARs button to view PAR form - GO TO section 3.6 8.3.5 SELECT a different monitor and/or change readings to recalculate doses and update PAR.

EP-MW-1 10-200 Revision 2 Page 24 of 34 ATTACHMENT I DAPAR USERS GUIDE Page 19 of 27

9.

CONTAINMENT LEAKAGE/FAILURE Wimr-MT171:~~~~~~

Ina.. I--.

I Method R:- lee iMode PAG Diitan -mi-iles 77-~~___-

10%

C;Gr Rap g

) Failwe to Iolate (1 00%. per Day) l0 QI

-Catai-ophic Fah(i' 1

I~kII~

~

hr PA stI (ef 7

777;

., I 0-_-_._.__._-

1--__

. 1,_, -f ; -

I',; - -.

, --

.,: -,

. s

Release Dujation hh mm) l

.1.0".I a,324E+00 -. 7.28E.03~,j I.90E-01-t I 9.86E-02 I 2.9:Z5E-01;,

4.51E+00k..

1. 21 5E+00

.83E43

~:1.26E.

1..6.55E-02, j-.196E-01.-

2.

2.0,

- 1.55E+00, ~-3.48E*03 "905E02 1,4.1E-02,-'

-1.41E01 as J1.65E+00,- :-3.72E J 93E*02:

'3.60E-02'~i.J9*0 3.01i-l 1;32E+00.-1 I2.98E.03 I 5.55E-a I a~,289E-02.:

-8.-73E.

3.5 1.1OE+00 [ 2.47E-03 4.59E*02 ]1; 2.39E 02

- 7.

.1 4.0' l-9.29E-01. l 2.09E-03 J: 3.89E.02 J

.a3E-02 i

-1.32E

1.09E:+OO

. a

" 7 :

9.1 SELECT the appropriate containment methods:

9.1.1

% Damage calculates doses based on release of total available fission products based on type and amount of core damage.

9.1.2 Drywell Reading calculates doses based on the drywell radiation monitor readings.

9.2 9.2.1 SELECT the appropriate containment release mode:

Leakage - Program defaults to 0.1% per day which is the Design Leakage rate per the UFSAR. If a different percentage of leak rate has been calculated by TSC engineers enter that value in the % per day text box.

9.2.2 Failure to Isolate - Assumes 100% of the isotopes available for release are released in a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months time period.

9.2.3 Catastrophic Failure -- Assumes 100% of the isotopes available for release are released in a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months time period.

EP-MW-1 10-200 Revision 2 Page 2 of 34 ATTACHMENT I DAPAR USERS GUIDE Page 20 of 27 9.2.4 Leak Rate (cfm) - calculates release rate base on estimated leak rate from the containment.'

9.3 After User enters data the program calculates offsite doses. The user can now have the program perform the following items:

9.3.1 SELECT the Print Receptor Report button to print a report of projected dose rates, TEDE and CED Doses and Ground Deposition at pre-designated receptor points.

9.3.2 SELECT-the Print button to print offsite dose projections based on containment failure.

9.3.3 SELECT the Back button to change input data on the Full Assessment Form.

9.3.4 SELECT the PARs button to view PAR form - GO TO section 6.

I

EP-MW-1 10-200 Revision 2 Page 26 of 34 ATTACHMENT I DAPAR USERS GUIDE Page 21 of 27

10.

FIELD TEAM ANALYSIS NOTE:

The program calculates the plume Travel Time and Release Time to allow Dose Assessment personnel to compare previous dose assessment reports with data measured in the field.

10.1 Dose Rate Survey-SELECT this method if Field Team Survey Data is available.

EP-MW-1 10-200 Revision 2 Page 2t of 34 ATTACHMENT I DAPAR USERS GUIDE Page 22 of 27 10.2 Air Sample Results - SELECT this method if Field Team Air Sample Data is available.

-Downiwind (les): f cris~ rz:oe W2 4 rj Suvy~ Tire 1

J-(Mr-.

+/-

IIUUIL1 LU

[

I~f~+U3 1.. I+UU, 4.to~

t+UU 21 \\

jt+

U,

1.5_ -...

443 19EO 1.171-49EO01'

'3OE.O--.

2.0 [-9.96E+02 I

.54E+00 ]

1E1I64EO 231E.0O.

25.5 18.33E.02'-

1.29E00

-8.75E.O2~

48E0

..0E.OO

.0o I65E0*

1OEO I-6.8XE02,, :-:3.80E-01 1.4E 0 V,

10.3 NOTE:

ENTER the Field Team information as follows:

The program will not allow mr/hr readings for sample data or isotopic results for survey data.

10.3.1 ENTER Downwind (miles) - straight-line distance from release point to sample location.

10.3.2 ENTER Crosswind (miles)- the distance the team was away from the centerline when the sample was taken. The program will warn user if reported sample location is wider than expected plume width. The maximum width of any plume for the most unstable stability class is 2.96 miles 10 miles downwind.

10.3.3 IF the analysis basis is Dose Rate Survey-ENTER the Field Team Survey reading in the box labeled Level.

EP-MW-1 10-200 Revision 2 Page 28 of 34 ATTACHMENT I DAPAR USERS GUIDE Page 23 of 27 10.3.4 IF Analysis basis is Air Sample Results - ENTER the uCi/cc values for each known isotope in the table at the upper right section of the form.

10.3.5 ENTER Survey Time - Enter the time the survey or sample was taken.

10.4 After User enters data the program calculates offsite doses. The user can now have the program perform the following items:

10.4.1 SELECT the Print button to print offsite dose projection reports based on Field Team Analysis.

10.4.2 SELECT the Receptor Point Distance button to view/print a report on the downwind and crosswind distance to pre-designated receptor points. This report can help in selection of appropriate sampling locations by categorizing each receptor point as being within 1, 2, or 3 Sigma Y of centerline. Sampling should be performed as close to centerline as possible.

10.4.3 SELECT the Back button to change input data on the Full Assessment Form.

10.4.4 IF Field Team samples were the selected basis, SELECT the PARs button to view PAR form - GO TO section 6.

EP-MW-1 10-200 Revisioh 2 Page 2§ of 34 ATTACHMENT I DAPAR USERS GUIDE Page 24 of 27

11.

RELEASE POINT ANALYSIS

-~~~~

~

.I.

.Ve46Flo'teSCFM).

00k Eee

IKr-88......-._1r

' s..

Xe-31m RehaW6 Ouwtionthkrn):

3:00w Ia;hG rond..

Xe-133 2.OOE+05 d F Xe-i 33m

'6ackDi e

i' 11Xe 3

2.20E+ 01 1

PtRpceptor s

3.40E+05

~'

i 01 ie l: I

.40E+05

-01E+03

4.16E+02 1 39E+02

' 157E+03.

1i.39E+04

0.5

, 3.40E+05 --..1.01E+03,-

416E+02 1;39E02 57E+03-

-;39E+04 '

A 1.0 2.40E05

7.1 BE+02

,2 57E+02i;

.8._5E+01 1

.6E+03 8.58E+03 1.5

.6E05 4.75E+02-1 71E+02 569E01 7.03E+02

. 5.9E+03,..

.2.0 1.15E05 3.42E+02

. 1.23E+02

141OE+01.5.06E+02

.4.1OE+03;

-2.5.; -, 1.23E+05-,. ;a66E+02 9.41E+01

'3.14E+01-4.91E+02: -. ::3.14E+03. `

3.0.: -9184E+04 2.93E+02

7.54E+01' 2.51 E+01 3.93E+02/- :2.51E+03'r i3.5, :-:8.14E+04i,. 242E+02 !;

6.24E+01.

-208E+01 326E+02

-2.08E+03 4.0

-6.90E+04;

. 2..

-2.05E+02

.5.29E+01 1.76E+01 2.76E+02

.1.7E+03 4.......

9....E.....4 I+0 1

.E R

eo C

c t

i f

e it

(

- u 11.1 ENTER the known Isotopic Concentration for each isotope (if unknown leave blank).

11.2 ENTER Vent Flow Rate (or estimate flow rate for other releases) in SCFM.

EP-MW-1 10-200 Revision 2 Page 30 of 34 ATTACHMENT 1 DAPAR USERS GUIDE Page 25 of 27 SELECT the proper release point (Elevated or Ground) to provide the proper atmospheric dispersion calculations..The table below provides guidance:

11.3 STATION PATH Rel Pt:

Chimney Elevated '.

Dresden/Quad Reactor Bldg. Vent.

Ground

.ea

,Isolation Condenser (Dresden only)

Ground Hole in Wall / Other Ground Ventilation Stack Ground Braidwood/Byron Steam Generator Safety Valves/PORV Ground Hole in Wall / Other Ground Station Vent Stack-Elevated LaSalle Standby Gas Treatment Stack' Elevated Hole in Wall [Other Ground HVAC Stack Ground Clinton' Standby. Gas Treatment Stack Ground Hole in Wall / Other Ground After User enters data the program calculates offsite doses. The user can now have the program perform the following items:

11.4 11.4.1 SELECT the Print Doses button to print offsite dose projections based on Release Point Analysis.

I 11.4.2 SELECT the Print Receptor Report button to print a report of projected dose rates, TEDE and CED Doses and Ground Deposition at pre-designated receptor points.

11.4.3 SELECT the Back button to change input data on the Full Assessment Form.

1.1.4.4 SELECT the PARs button to view PAR form - GO TO section 6.

I

EP-MW-1 10-200 Revisioh 2 Page 3i of 34 ATTACHMENT 1 DAPAR USERS GUIDE Page 26 of 27 12.

GEMS ANALYSIS

.I 12.1 SELECT Total, Noble Gas, lodine and/or Particulate.

Release Rates if Total values are available from

GEMS, 11 j*Totel Nob Gas, L wor P&a e ReeaeR 5 lndv I-ileseRates

,K-88

. Xe-1 31 m 4.00E+07

Xe-1 33 290EO08

.Xe-1 733m_

Xe-135.

3.20E.08 PAG Dan miles i '

JTEDEt~~

-2

'i;i-tE 'tli.DEtiD-45--

3.tj *;

- Z "1

it >,}......

..... ~ -

t-

-W v,--c.v

'.; RatcaxoDaon Ikil 30015 K'

4y#

... ss.r.-." "'a.

as

.a n#s..

wS>.n.

4) s F-

P,

'Pett Rt FIt Dosci :

e-38 1:

1-131 Z10E.05 or.

1.-.

-S.B.'. I234E403:l GL36EE00:

'-7.0E0 Z53E.00'l 1.65E+M I 21502 i I 5

.1.58E+03 430E +00

3.51E.00

-,1.1?E+W -

8.98E.00 A17E.02 1.0 92.-

.,&6E+(2

-Z52E.00.

1.31E+0; -

4.37E1-

-4.26E+00 437E+01

-1.5.-

6.63E+02 1 80E.

.7.13E01

-Z38E01 a-Z75E400.

Z38E+01.

7 20

.&511E+02

.1.39E+O.

,.66E01

.1.55E41'

.2.O1E.OJ00. 55E+01,

2--5 196E+02-I B00!W

'-3 40E.01 1.1 E0o 71.5SW7 -1.13E,.01 ao 9E 8.39E-01

-2.M501

-8.2E2

-1.19E+W~

I L5 j Z52E+02

6. ME-l

-21SE

,1. I-7.20E02 973E01 7.20EO*

-4.0 213E02 5.78E2E-01 1.-.6.07E

4.

.a2lE4 1.

607E004 8 12.2 SELECT Individual' Isotopic Release Rates

'I

'lX;,cidNctle Ga, lnealr Pte Relese es JhdlseaseRate I

ot~b*IeGases.

[

D.gOE+06 iePstance llea

,,+..

5

-TEDE.

...2.lI I

cD~E(hwi~o i25 IP~1 Patkl 6Ees 4

Wsec-I_*J 1Q,j ~-P~-bt.-.'y-4~

-<','.K t

-.. li; ',llasiT;Pon S..

1.42E+02 lC: :224E01- -- 7.15E+00 -

211E+00 :S4£.00 -lJ6801

0.5;

.07E+01 -.

1.27E1 4.06E+.0 0-1.20E+00

-39E+.

aBSE&MA 1.0

'7.0E401

.1.12E01.

1.4E40 5.44E41 Z5OE00

. 1.01 1.5 5.01E+01 1.92E02 A.10E+'Wt0,0 - 24E-01--1 1E400 t04E 0.1 1-4.02E01-l 36E.02 7.53E01

-a 7

2E212-.1.04E+00. 7-? 16E00 2.5 a3E.01..

63242: -5.61E1i 7i.6SE4-l 7.90E01 33E00?

.ao

-- 264E+01 4.18E02.

441E.01 l..1t

£13E'01 l.4.19EO0.

-a-, 12.37E+01 74E42 -

aEOEol-1.0.1

'03E01

- a42E4W'0i:

II I

k t

I W

k t

V I 4'.0 Z14+Ol 1:

a37E-U1 '3.U02E1 I`UUMQZ.-.4.25E41

_ Z__L+__ _I-12.3 ENTER the known sotopic Concentration for each group or each isotope (if unknown leave blank).

EP-MW-1 10-200 Revision 2 Page 32 of 34 ATTACHMENT DAPAR USERS GUIDE Page 27 of 27 12.4 After User enters data the program calculates offsite doses. The user can now have the program perform the following items:

12.4.1 SELECT the Print Doses button to print offsite dose projections based on Release Point Analysis.

12.4.2 SELECT the Print Receptor Report button to print a report of projected dose rates, TEDE and CED Doses and Ground Deposition at pre-designated receptor points.

12.4.3 SELECT the Back button to change input data on the Full Assessment Form.

12.4.4 SELECT the PARs button to view PAR form - GO TO section 6.

EP-MW-1 10-200 Revision 2 Page 33 of 34 ATTACHMENT 2 OBTAINING IDNS MONITOR DATA FROM THE EOF Page 1 of 2 This procedure provides instructions for obtaining Illinois Department of Nuclear Safety (IDNS) effluent radioactivity and offsite radiation data via a dedicated computer in the EOF.

For each Exelon nuclear power plant in Illinois, the following data is available from IDNS:

GEMS (Gaseous Effluent Monitoring System) data on radioactivity in airborne effluents; Reuter-Stokes detector offsite radiation data

1.

PRECAUTIONS None

2.

LIMITATIONS AND ACTIONS

. None

3.

PROCEDURE 3.1 TURN ON the Reverse State Link computer. This is a non-networked computer located in the Environmental Area of the EOF.

3.2 CLICK OK on the messages and dialog boxes that appear (including error messages) until you get to the Windows desktop. No password is required to log onto this computer.

3.3 DOUBLE-CLICK the icon labeled pw" to obtain the connection password.

This will be needed for Step 3.7, substep 6.

3.4 CLOSE the file which provides the connection password.

3.5 DOUBLE-CLICK the icon labeled IDNS.h 3.6 WAIT for the modem to establish a connection with the IDNS computer network.

3.7 EP-MW-1 10-200 Revision 2 Page 34 of 34 ATTACHMENT 2 OBTAINING IDNS MONITOR DATA FROM THE EOF Page2of2 INTERACT with the IDNS network per the following dialog:

Prompt User Response I

Blinking Yellow Rectangle

<enter>

2 reac <enter>

.3 Enter username>

exelon <enter>

4 Local>

c eagle <enter>

5 Username:

exelon <enter>

6 Password:

[Use password determined in Step 3.2]

<enter>

7 Eagle$

.onfs <enter>.

8 Enter Option (1-24),.

env <enter> displays Environmental Menu

9.

Enter Option (1-24),

One of the following:

metra <enter>

[for meteorology, Reuter-Stokes data, and GEMS gross noble gas, iodine and particulate release rates]

gems <enter>

[for GEMS isotopic release rates]

.ex <enter>

(to begin process of exiting from the IDNS network]

When using the metra or gems option, follow the on-screen instructions for display choices or to exit to the Environmental Menu. If you are using the

. - -metra ALL display option; use Control-c to exit to the Environmental Menu.

After entering ex from the Environmental Menu, continueas follows:

10 Eagle$

lo <enter>

(letters ell and "oh" for "logout")

11 Local>

Alt (cd) (While holding Alt key, type cd) 12 Blinking Yellow Rectangle.

Alt (fx) (While holding Alt key, type fx)

FThis returns the computer to the Windows desktop.]

3.8 When access to the IDNS network is no longer needed, SHUTDOWN the computer-per standard Windows procedures.

ATTACHMENT C EP-AA-110-301 CORE DAMAGE ASSESSMENT (BWR)

EP-AA-1 10-301 Exelon.

Revision 2 Page 1 of 25 Nuclear Level 2-Reference Use CORE DAMAGE ASSESSMENT (BWR)

1.

PURPOSE 1.1.

This procedure provides emergency response personnel with the methodology to estimate the degree of possible core damage at Exelon Nuclear's Boiling Water Reactor (BWR) stations, with the exception of Oyster Creek Generating Station (OCGS). Refer to EP-AA-1 10-302 for methodology to estimate potential core damage for a Pressurized Water Reactor (PWR).

1.2.

This Core Damage Assessment process is designed to assist in estimating core damage after an accident with potential clad or core damage conditions, and is intended to provide an acceptable alternative to existing station core damage assessment models and methods utilized by Reactor Engineering to assist in the following:

Determining if the fuel barriers are breached to evaluate the appropriate Emergency Action Level (EAL) classification.

Providing input on core configuration (coolable or uncoolable) for prioritization of mitigating activities.

Determining the potential quantity and isotopic mix of a radiological release to project offsite doses.

Predicting the radiation protection actions that should be considered for long term recovery activities.

Satisfying inquiries from local and federal government agencies and provide evidence that the utility knows the plant conditions.

1.3.

Core damage may be assessed by:

Evaluating the drywell radiation levels (and confirmed by evaluating the extent of time the core was uncovered),

Concentration of certain isotopes in a reactor coolant analysis, or Concentration of hydrogen in the primary containment.

History of Core Cooling

2.

TERMS AND DEFINITIONS 2.1.

BWR - Boiling Water Reactor 2.2.

Cladding The outer coating (usually zirconium alloy), which covers the nuclear fuel elements to prevent corrosion of the fuel and the release of fission products into the coolant.

EP-AA-11 0-301 Revision 2 Page 2 of 25 2.3.

Containment Type -

Clinton (Mark 111)

Dresden (Mark 1)

LaSalle (Mark 11)

Limerick (Mark 11): 764 assemblies Cont. Volume (384,570 3 ) = Suppression Pool (149,380 fW

3) + Drywell (235, 190 f3 )

Peach Bottom (Mark 1): 764 assemblies Cont. Volume (303,600 ft3) = Suppression Pool (127,800 3 ) + Drywell (175, 800 f3 )

Quad Cities (Mark I) 2.4.

Core Release Fraction - The fraction of each isotope in the core inventory that is assumed to be released from the core under given core conditions.

2.5.

Core Uncoverv Time - For BWRs this is the period of time when reactor water level is less than that'required for minimum steam cooling, or about >

20% of the core active fuel is uncovered.

2.6.

Cladding Failure

1.

Also referred to as "Cladding Oxidation", "Gap Release" or "Clad Rupture" in other documents.

2.

100% clad failure refers to the rupture of 100% of the fuel rods in the core. This would result in all fission products contained in the gap space being released to the reactor coolant system.

2.7.

Equilibrium - Conditions associated with evaluation of different volumes of liquid or gas that contain concentrations of radioactive materials or hydrogen, when these concentrations are approximately the same. This is normally an extended period of time following accident initiation.

2.8.

Fission Products - The nuclei (fission fragments) formed by the fission of.

heavy elements or by subsequent radioactive decay of the fission fragments.

2.9.

Fuel Melt

1.

Referred to as uCore Melt," "In-Vessel Melt" or "Over-temperature" damage in reference documents.

2.

100% fuel melt refers to high temperatures in the fuel pellets in 100%

of the fuel rods in the core. This would result in all the fission products contained in the fuel pellet matrix being released to the reactor coolant system.

2.10.

Gap -The space inside a reactor fuel rod that exists between the fuel pellet and the fuel rod cladding.

EP-AA-110-301 Revision 2 Page 3 of 25 2.11.

Gap Release - The release into containment of fission products in the fuel pin gap.

2.12.

In-Vessel Core Melt -A condition during a reactor accident in which some of the cladding or reactor fuel melts as a result of overheating the fuel and remains inside the reactor vessel.

2.13.

In-Vessel Core Melt-Release - A release into containment from the reactor vessel, which assumes the entire core has melted, releasing a representative mixture of radioisotopes.

2.14.

Minimum Steam Cooling RPV Water Level (MSCRWL

- The lowest RP water level at which the covered portion of the reactor core will generate sufficient steam to maintain the hottest clad temperature below 15OoF.

2.15.

Minimum Zero-Iniection RPV Water Level (MZIRWL) - The lowest RPV water level at which the covered portion of the reactor core will generate sufficient steam to maintain the hottest-clad temperature below 1800OF, assuming no injection into the RPV.

2.16.

Shutdown -

As defined by station emergency operating procedures.

2.17.

Slump - Relocation of molten reactor core during an accident.

2.18.

Source Term - The amount and isotopic composition of material released or the release rate, used in modeling releases of material to the environment.

2.19.

Spiked Coolant-Reactor coolant containing increased concentrations of non-noble isotopes, sometimes seen with rapid shutdown or depressurization of primary system.

2.20.

Spiked Coolant Release - The release into containment of 100 times the non-noble gas fission products found in the coolant.

2.21.

Subcritical - The reactor condition when the number of neutrons released by the fission is not sufficient to achieve a self-sustaining nuclear chain reaction.

Defined under station emergency operating procedures.

2.22.

Suppression Chamber - May also be referred to as Wetwell or Torus. The Large steel pressure vessel containing a large volume of water that acts as a heat sink for the Drywell.

2.23.

TID - Total Isotopic Distribution

EP-AA-110-301 Revision 2 Page of 25 2.24.

Vessel Melt-Through

1.

Referred to as "Ex-Vessel Melt" or "Melt Release" in reference documents.

2.

Core debris is relocated to the primary containment building after the reactor pressure vessel has failed.

3.

RESPONSIBILITIES 3.1.

The TSC Core/Thermal Hydraulic Engineer shall serve as the Core Damage Assessment Methodology (CDAM) Evaluator.

3.2.

The TSC Radiation Controls Engineer shall coordinate radiological and chemistry information with the Core/Thermal Hydraulic Engineer in support of core damage assessment.

3.3.

The TSC Technical Manager shall coordinate core damage assessment activities.

4.

, MAIN BODY 4.1.

REFER to Attachment 1, BWR CDAM User Guide for instructions on use of the Core Damage Assessment Methodology (CDAM) Software Program.

5.

DOCUMENTATION 5.1.

A Summary Form and method specific reports are generated by the BWR CDAM Software for use in documenting the results of the assessment.

6.

REFERENCES 6.1.

NEDO-22214, Procedures for the Determination of the Extent of Core Damage Under Accident Conditions 6.2.

NEDC-33045P, Rev 0 (July 2001), Methods of Estimating Core Damage in BWRs 6.3.

WCAP-14696 (July 1996) Westinghouse Owners Group Core Damage Assessment Guidance.

6.4.

WCAP-14696-A (November 1999), Westinghouse Owners Group Core Damage Assessment Guidance.

6.5.

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

EP-AA-110-301 Revision 2 Page 5 of 25 6.6.

Station Commitments 6.6.1.

Peach Bottom CM-1 T04511 (Attachment 1, 5.6) 6.6.2.

Limerick Bottom CM-2 T04512 (Attachment 1, 5.6)

7.

ATTACHMENTS 7.1., BWR CDAM User Guide

EP-AA-110-301 Revision 2 Page 6 of 25 Attachment I BWR CDAM User Guide Page 1 of 20 OVERVIEW 1.1.

As a windows based application designed in Microsoft Access, BWR CDAM, uses many standard user interfaces. Instructions are not provided in basic computer operations in the windows environment. The user must be familiar with these to efficiently operate the program.

1.2.

It is also assumed user is familiar with basic reactor physics and core damage fundamentals. Emergency Response Organization training will provide an overview of core damage assessment methodologies.

1.3.

The program should be used by qualified personnel as a tool to estimate the type and amount of core damage..

2.

DETERMINE APPROPRIATE AND AVAILABLE ASSESSMENT METHODS Mid-West Region Stations

lREFER to EP-MW-1 10-1 001 ifor a listing of appropriate plant parameter lpoints to be used following a LOCA.

2.1.

The magnitude and type of event, transport mechanism and time after shutdown will be influencing factors on the method(s) utilized to determine the extent of core damage. Damage estimates can be developed using one or more methods as they become available or applicable.

2.1.1.

Indications Of Core Damage

1.

The primary indicators of core damage that are available during the early phases of an event:

Drywell/Containment Radiation Monitor Readings Drywell/Containment Hydrogen Readings

2.

Auxiliary indicators that are used to confirm and better define the possible type of damage are:

Reactor Pressure Vessel Level Indication System readings Estimation of maximum temperature reached within the core Estimated core uncovery time Abnormal Source Range Monitor readings

EP-AA-110-301 Revision 2 Page 7 of 25 BWR CDAM User Guide Page 2 of 20

3.

Long Term Indicators (once liquid or gaseous samples can be safely obtained) are:

Isotopic Ratios Presence of high levels of rare isotopes Quantity of isotopes present in samples 2.1.2.

SELECT the assessment method(s) most appropriate for the existing conditions. Methods available for assisting in the determination of the extent of core damage include the following:

Method Use Comment Containment Early Indication of Core Uncertainties due to variables in release Radiation Monitor Damage of fission products from RCS and effects of containment sprays.

Core Conditions Indication of onset of May not be reliable during later phases of Core Damage, core overheating due to changes in core geometry.

RPV Level Indication of Core.

Indicates possible damage not useful in Uncovery estimating the quantity of damage.

Source Range Indication of Core Loss of water level leads to increase in Monitor Uncovery gamma detection.

Containment Early Indication of Core Significant uncertainties due to variable Hydrogen Monitor Damage Hydrogen generation in core and in release of Hydrogen from RCS and effects of containment sprays.

RCS Samples and Late Indication of Core Very large uncertainties until all systems Containment Sump Damage -Suppression have reached equilibrium. Useful in and Atmosphere Pool Samples provide planning long term recovery.

Samples indication of Rx Vessel

. Failure 3.

.3.1.

3.1.1.

START UP THE CDAM PROGRAM ACCESS the application by one of the following:

OPEN the BWR CDAM desktop icon on applicable computers.

1.

START the BWR CDAM program for the plant that has declared an emergency. Programs are labeled BWR CDAM.

2.

SELECT the appropriate icon or run from the 'start bar' and type in the file path and name as follows C:ICDAMIBWR CDAM.MDB

EP-AA-1 10-301 Revision 2 Page of 25 BWR CDAM User Guide Page 3 of 20 3.1.2.

If the assigned Core Damage Assessment Computer cannot access the application or the CDAM program will not run, then install BWR CDAM on any computer from CDs or Disks located in the TSC or the EOF Library.

1.

INSTALL CDAM by copying appropriate file to computer's hard drive.

2.

UPDATE the properties" of the file by deselecting write protection.

4.

SELECTION AND PERFORMANCE OF ASSESSMENT 4.1.

SELECT the assessment method(s) most appropriate for the existing conditions. Methods available for assisting in the determination of the extent of core damage include the following:

Containment Radiation Analysis - (Section 5.2)

Core Conditions Analysis (Cooling History) - (Section 5.3)

Containment Hydrogen Analysis - (Section 5.4)

Nuclide Analyses (Ratios.and Abnormal Isotopes) - (Section 5.5)

Liquid Samples Analysis - (Section 5.6)

Gaseous Samples Analysis - (Section 5.7)

Basic Program Flow Diagram

EP-AA-1 10-301 Revision 2 Page 9 of 25 Attachment I BWR CDAM User Guide Page 4 of 20

5.

PROGRAM SCREENS AND INPUTS 5.1.

When the program is started the following screen appears:

NOTE:

The value boxes. are empty when the program is originally launched. The examples below may deviate from the CDAM displays during use due to different software versions in use in the Mid-Atlantic and Midwest regions. The display differences do not Mid-Atlantic version lists impact the functionality of the program. Where Limerick, Peach Bottom station title differences exist, the titles applicable to (and Oyster Creek which the Mid-Atlantic stations are contained in "( )."

is currently not applicable).

==

zeunrin Methd:.-

Me.

Ml t L~Rad Mntr Seel 4%

.z I_

s-..-;*-'_--_?

r-1i>--N~e:;

;

I

_.: I I 11 I "

-__-___

._4V:I CoreCondlions,4:

-,-.6Co 6*Cofn-7

~~-r~--

See 5.5 1Cr~m

~See 5.7 K

See 6.1 F

i T

-7 T'

7-

~flpri~

___77e I

EP-AA-1 10-301 Revision 2 Page 1 of 25 BWR CDAM User Guide Page 5 of 20 CAUTION Selecting an "Affected Station" resets all inputs to default values.

SELECT the Affected Station before other "Assessment Methods."

CAUTION Pressing the "Quit" button exits the program. When the program is closed all data is reset. Program saves no information to disk; printed reports serve as record of core damage assessments.

5.2.

5.3.

5.3.1.

I Drywell/Containment Radiation Monitor Method PRESS the "Cont Rad Monitors" button on the Summary Screen to open the following form:

I

___i.8.so--swl I

See 5.3.3

-.Key Parameters 5.3.4

,-Cor*S rys Off

'-'*..~Cont Sprys On

- i nesinceSD s) t

t.

.i 4-,

M 6itr(R/hr)

Assessment Results:

T " 7 ~

~

'U" I

Drywe.-

Met Dad'-,

.Nte

.::...--.CM 059nl 2 0OE.03t

,Durnage Estimate re4--l-7OZ lj

.,;CMM-O E..

100% eeng(R.

.7 0

.1 1 The highest monitor read i used ti the.damage

?i

.Reud (RA-rF

1. 78E03l-r8.11Er-l I~e

..21 Preliminary

esults

(affect of

. input data) are shown here.:

, iX5esment alculations. :-,-

. :?

,~'S' rXn

-;_§

.;sm r,-,iC-- -i d- -I---fl-I- -

-l 1rs-fi (MA: Suppression Chamber)

Containment:.

Melt Clad..

..¢->

-;R /H r le

,l,.

-,D amage Eti at c1 l L t Z i

.e'stimated reading'.within.- ;.,'.'

l1'00% Reading (R/Hrj 2.21E 5rE05 8 Er-03 Containment is used for the.

. -r

.damage assesment calculations. -

Reang'(F/Hrj 2.21 Er03 -l8.l1E.O1lj S e e 5.3.8 '

,i,. .Dryell Grph j r

-Contanmnt Graph-I+ -

heVs ';

adc;'..a

EP-AA-110-301 Revision 2 Page 11 of 25 BWR CDAM User Guide Page 6 of 20 NOTE:

Program allows entry from 2 high range monitors for Drywell location and 1 for Torus or Containment I Suppression Chamber, however a reading may be entered from any monitor or measurement taken external to suppression chamber, which accurately indicated containment radiation levels. If two entries are made only the highest is used.

5.3.2.

ENTER the highest Drywell radiation monitor reading that occurred in these boxes

1.

If Drywell radiation monitor readings are not available, then enter the containment / Suppression Chamber radiation monitor reading.

5.3.3.

SELECT Drywell/Containment Spray status:

1.

If the Drywell/Containment Spray system was operated for the majority of the time since the estimated time of the onset of core damage then choose Drywell Spray On."

2.

If the Drywell/Containment Spray system was not operated or only operated briefly (e.g., <10% of time since the estimated time of the onset of core damage) then choose "Drywell Spray Off."

5.3.4.

ENTER the time after reactor shutdown, which corresponds the time the containment radiation reading was taken. Value must be between 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after shutdown, which corresponds to the time period in which this method is considered effective.

NOTE:

Pressing "Reset" button resets values on this form only.

5.3.5.

PRESS "Containment Graph" or "Supp Chamber Graph" button to display.a screen similar to the following:

(See example display on next page.)

EP-AA-1 10-301 Revision 2 Page 12 of 25 Attachment I BWR CDAM User Guide Pacee7of20 AE+6

--~~~~~~

.2E0

~ E+5 u e l S D

~~

~_

,4 E 4 o i 8

I _ _ _ _

F~~~.cit C j~a n I E+3 -

_I R u t e

t

< f I E + 2 _ _ _ _ _ _ -~~~~~~~~~~~~~~~~~~~

_ _ _ _ _ _ I ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~

O a d

< i~ ~ ~ ~~~~~t

- I

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~~~~~~~~~~~~.

0~1

====;

6=2===.____

-E. ~

~

_ f _

______ it Cld4_

_I_____

A3.6

J3 3.7 NOTE:

Graph shows high and low containment radiation levels which correspond to 100% Melt or Clad or 1%

Melt or Clad damage. A dot shows the last containment radiation level entered into the program for assessment.

5.3.6.

PRESS the "Print" button to print a report of containment radiation method inputs and best estimate of damage.

5.3.7.

PRESS the "Done" button to return to the Containment Radiation Monitor Evaluation Screen.

5.3.8.

PRESS the "BACK" button to return to the Summary Screen.

5.4.

Core Conditions Methods NOTE:

Each of. these. four methods is an independent assessment method.

5.4.1.

PRESS the "Core Conditions" button on the Summary Screen to open the following form:

I (See example form on next page.)

EP-AA-110-301 Revision 2

- Page 13 of 25 Attachment I BWR CDAM User Guide Page 8 of 20 I ee4.2 I

,The core is paftialiiP uncovered'-'

but i cooled by steam: Clad.;

temperatte: bre erxpected to' renain below'1500F.' o core, damage is Fxpcted.

oke

. Fs,ee 5.4.3

' Souice Range Mon 6Ci RateV 60..

Ox. s S rma- [No ftYei,

+

~nTime 0 50 Core Telperatire 1800'

, a. --;.....,,.,,,;

Assessment Hesuhs Assessment Result:

Assessment Results.

,0 to2 $ hour nima! un ry Between i8oditand24b0oF-F.`"

'The' coe has'remained covered.

ie ocredmg eyr:r jcWae ecin Local damade may have, expected.

I frgen s " leased and the fuel'

!obctuted do to other events. No cone damage is expecled See 5.4.5

/

fails.

See 5.4.41.

I 5.4.2.

Under Reactor Pressure Vessel (RPV) Water Level ENTER the lowest recorded (or estimated) RPV level (range 0 to -350 inches) and core spray flow at time of lowest reading NOTE:

Steps 5.4.3 through 5.4.6 are based on inputs from Reactor Operators, TSC Staff and other engineering personnel (including outside sources such as General Electric personnel).

5.4.3.

Under Source Range Monitor REVIEW plant parameter history and if the SRM (Wide Range Monitor at Peach Bottom) had indications of a reading 10 times those expected check "Yes."

5.4.4.

PRESS the "Core Levels" button to view information regarding water levels associated with the Station reactor and vessel level indications.

(See example form on next page.)

EP-AA-110-301 Revision 2 Page 14 of 25 Attachment I BWR CDAM User Guide Page 9 of 20

~~.

Lmrick Sition Top of A tive Fe

.1617 Ah':

Minimu Steam CtingFR'Water Level:.

-186.,

H

. MinimumZero Inecton RisWater Lee

-201 I*

fit,;.njet PurrpSn Ciatio l -204. l i

..-. ^;.. -.:~Bottom of Active Fuet:l - -30 4 Id.

..i.

.. -- l ::Back; aI.

I 5.4.5.

ENTER the estimated time the reactor core (20% of top of active' core) was uncovered without steam (level below the Minimum Steam Cooling Rx Water Level) or spray cooling reactor core.

5.4.6.

ENTER the estimated highest temperature reached in the reactor core.

5.4.7.-

PRESS the "Print" button to print a report of inputs and results of core temperature methods of core damage assessment.

5.4.8.

-PRESS the "Back" button to return to the Summary Screen.

5.5.

Containment Hydrogen Evaluations CAUTION This CDAM assumes no ignitor operation.

Ignitor use limits containment hydrogen concentration affecting the reliability of this method.

EP-AA-110-301 Revision 2 Page 15 of 25 Attachment I BWR CDAM User Guide Page 10 of 20 5.5.1.

PRESS the "Cont Hydrogen" button on the Summary Screen to open the following form:

2,

,Utth'i'"

ra~pro 'riote i

'ff_

4l

.~~Ca csxed:tw'2'

->,.s.1"F f

3 0 k k

. $ t 4 A k.a

.. ' ~ Y X4 !, H Z:

. it t 10 0 0 Brz X m

-X10 I

Air

-gi NM g.

-2~~~~~~

%%MwX4<./<

5

~

C~

'.A.,,<~/~'

~.

A 4Z O idized 5

MW 'A: '

~Q2~~ >4>.e measured NOTE

Suppression Chamber reading can only be entered if user selects "no" under Equilibrium in step 5.5.3 below.

5.5.3.

SELECT the applicable System Equilibrium status based on the following:

1. -

If Containment and Suppression Chamber monitors read the same or only atmospheres are assumed equalized, then SELECT Yes" for equilibrium.

2.

If containment and suppression chamber atmospheres are not in equilibrium or only containment H2 reading is available, then SELECT "No" for equilibrium.

5.5.4.

PRESS the "Print" button to print a report of inputs and results of core level methods of core damage assessment.

I

EP-AA-110-301 Revision 2 Page 16 of 25 BWR CDAM User Guide Pagel of 20 5.5.5.

PRESS the "Back" button to return to the Summary Screen.

5.6.

Nuclide Analysis(CM-1, CM-2) 5.6.1.

PRESS the,"Nuclide Analysis",button on the Summary Screen to open the l

following form:

S I

Ratio Comp on See-62 Vibtetotpe-X Ratiolto'm'p~r' e-,s,,:

r See 5.6.3'T-ine Since Shudown (hoW ur :J'jj i: 1 J

-o I

g See 5.6.* l Noble Ga §Activh.

Melt.-

o

< Clad

.AkalineEaths l

ISee 5.6.3.1 Xe33j.1.

L w S r

.i.

O ~~~~_____

i.

.j8

OE0,

.:~l.................li00Zl Rfractouie.........

t 7E Kr-87:

l.2-OE. -0 )-

> 022-l 0.022 Kr-Ot l 330E.IJl )

l 0.29.

1 0-045 ta-.:

Xe 131mj 1

2-2 004 0.004 F

Pd

, W.

0.14 l -< 0.096..1 l -I',$Jl l;.

Mo: i ', Tc-!

Xe-35:

220E.Ol

!li 0.19

.051 l 1

ae E rths l

. 'Haloaen:

A".ctivii'.'< Melt SamDle Clad,_

'; l ui4'

-l~ fli lSee 5.63:.2 t>\\!11 3

1 I--'E!

i

. 1J 0

'10-d '

_1~I132 l

20E1l.-1.46 l< 0.127 1.lk) l N

F4P.~/...............

.;.-;--i

. 1 -tI133 O 3t 2.09 t< 0.685.

[H-.f'-

!-W; 11-34 -l220E.O1t- --<< 3

>.30 -

0155:

.- si l

135. j AcOE.i 1.87

-76 f Y-j

,A 0*

.zI>,

I~~ee 5.6.3.2 j

Nil.

D-i 'E~~~~__

l I.

I ee6.7 N6.8 5.6.2.

ENTER the time since reactor shutdown (time between shutdown and sample being drawn).

5.6.3.

ENTER isotopic sample results in uCi/cc. Sample results are to be decay corrected back to time after shutdown that the sample was drawn.

1.

Noble Gases are ratioed to Xe-133

2.

Halogens are ratioed to 1-131 5.6.4.

If the ratios evaluated above are greater than predicted melt ratio, then melt damage is predicted

.5.6.5.

If the ratios evaluated above are less than clad ratio, then clad damage is predicted.

EP-AA-110-301 Revision 2 Page 17 of 25 Attachment I BWR CDAM User Guide Page 12 of 20 5.6.6.

If abnormal levels of rare isotopes are presen then check "Yes" and check which i

sotopes are present.

5.6.7.

PRESS the "Print" button to print a report of inputs and results of core level methods of core damage assessment.

5.6.8.

PRESS the "Back" button to return to the Summary Screen.

5.7.

Liquid Samples 5.7.1.

PRESS the "Liquid Samples" button on the Summary Screen to open the following form:

Sample pe/rLocaton Power Histol-I v Powe'p,()

1 IiShort Lved]s ICgfLved Days in Petiod Ag Power 11

.. ' 'r-.r-

,i

_ i!,

v, t....................

n j nRec t

i S s ee 5.7.6 l

.7 CBo'RCS and Supp eson' Poli Em~~~~~~~~~~-

R ecr:ri =Ta~

iiF 11I f

! ¢ amo - <.-- t -;-

S ee 5.7

.41r,.

v --*-

r "5 ----

ISe5771 Sample inlowsation mates

__333_

0:_

r-_.4

,j

~

-- ~.~nAftt rS i

ese 0::i:, 1-7111 i

Grph

'SyteminEquGaryur

! mC., ~

Lowest:.

.J~

-!;;. $See 5.7.10 ZZZZZZZA See 5.7.5.1. I'aZ 5.7.2.

SELECT appropriate isotope.

5.7.3.

SELECT sample location.

1.

If samples are available from both locations, then select both.

5.7.4.

ENTER Sample Information:

1.

Activity is isotopic sample results in uCi/cc (uCi/ml). Sample results are to be decay corrected back to time after shutdown that the sample was drawn

2.

Time After S/D (reactor shutdown) is the time between shutdown and sample being drawn.

5.7.5.

SELECT the appropriate System Equilibrium status:

EP-AA-1 10-301 Revisioh 2 Page 1 of 25 Attachment I BWR CDAM User Guide Page 13 of 20

1.

If sample was taken from only one location and systems are in equilibrium, then check yes" for "Systems in Equilibrium," otherwise check "no,'.

5.7.6.

ENTER power history (past to present, i.e. oldest steady state history as record number) of core since last refueling. Shutdown times are entered as the number of days with Ave Power (%) set at 0.

1.

For short-lived isotopes, EXTEND Power History at least 30 days.

2.

For long-lived isotopes, EXTEND power history at least 100 days, however the power history for the extent of the cycle is preferred.

3.

LIMIT variations in steady state power to +/- 20% within each operational period entered.

5.7.7.

Once all data has been entered, PRESS the "Calculate" button to display the

% Damage Estimates.

5.7.8.

' PRESS the "Volumes" button to display the follow screen:

~*~--~-. ~,.A2A$-

.See

.7.8.1 j

't - '"".

'. -Reactor CoolAnt System-RCS mW

<:. -- ;.. ppes 6rlo n hne;

.iqtjd (iip

___3__+O 1t

_26E091 See 5.7.8.2

'.Co.nin.t-t;,spher c4

-}

Suppression C erAtmsphee (cc):,l 3.32E.O9I A~~~~~~~l nd "Oliv e r Dresden staion____J kSe5784 See 5.7.8.4 Note7.8.

1.

Program enters default RCS volume, which the user may change based on RPV Level Readings at time of sample.

2.

Program enters default Suppression Chamber volume, which the user may change based on readings at time of sample.

3.

-Program enters default Containment free air volume which user may change based on conditions at time of sample. Unless there has been significant flooding of drywell this value will not change.

EP-AA-110-301 Revision 2 Page 19 of 25 BWR CDAM User Guide Page 14 of 20

4.

Program enters default Suppression Chamber free air volume which user may change based on conditions at time of sample. If there has been a significant increase or decrease in the water level in the Suppression Pool or Torus then the free air volume will change.

NOTE:

Pressing the "Reset" button will reset all volumes to default values.

5.

PRESS the "Back" button to-return to the Liquid or Gaseous screen, which user used to call volume form.

5.7.9.

PRESS the "Graph" button to display the following screen:

%FueI.eI,:::,

~

ampln raiion.

~iin~

7.27E-.Ol raimple Locatioun

,JS e

_E o O.

,,.-1E+4 3

-5 Sse nEulbim

'{l

-'.;N' t

s d

, Zj ElYe '.-EO No'. Ot NiA-i

'mates C..

F b u " > -...

S Dam! ~~~~~~~~~~~~~~~~~~Esmli.1:-r:;

@.1 E+2; ________

IE+ 1 See 5.7.9.1

/

See 5.7.9.2 (See Note on next page.)

EP-AA-1 10-301 Revision 2 Page 20 of 25 BWR CDAM User Guide Page 15 of 20 NOTE:

Graph on previous page shows High, Low, and Best melt curves; High, Low, and Best clad damage curves, and a red line across graph indicating entered corrected sample activity.

1.

PRESS the "Print" button to print a graph and summary of inputs.

2.

PRESS the "Back" button to go back to liquid or gaseous form which*

called this form.

5.7.10.

PRESS the "Back" button to return to the Summary Screen.

5.8.

Gaseous Samples 5.8.1.

PRESS the "Gas Samples" button on the Summary Screen to open the following form:

I.

1-1011-r

I~~ ~~~~~~~~~~~~

CA a £A' I I KapeTveLcto<.

~)ConAtm~ C~jsupp Cmber Atnos C= ot Sampe Inormaion JSee 5.8.4 1<

Sy~~~~ern~~~re~~(ps~~~gJ S~ee 5.8.6 Sy-e. ei1m23iiis 1

i,.j amage Estimates -

enemp(F):

3+0,V 2.OOE.O~j J*~~jf Highest:

1i.

Z3 erfrii Et im,6 ii

-e 1

I i... 1, -4.< `cIe I

7~---

-a7---,.

-'Do

=

Ad 5.8.2.

SELECT appropriate isotope.

5.8.3.

SELECT and sample location.

1.

If samples are available form both locations, then SELECT "Both" option.

EP-AA-110-301 Revision 2 Page 21 of 25 Attachment I BWR CDAM User Guide Page 16 of 20 5.8.4.

ENTER Sample Information:

1.

ENTER sample activity for selected isotope in uCi/cc (uCi/ml). Sample results are to be decay corrected back to time after shutdown that the sample was drawn

2.

ENTER Time After S/D that sample was taken.

3.

ENTER the pressure and temperature of the system sampled

4.

ENTER the end pressure and temperature of sample.

5.8.5.

ENTER power history (past to present, i.e. oldest steady state history as record number 1) of core since last refueling. Shutdown times are entered as the number of days with Avg Power (%) set at 0.

1.

-For short-lived isotopes, EXTEND Power History at least 30 days.

2.

For long-lived isotopes, EXTEND power history at least 100 days, however the Power history for the extent of the cycle is preferred.

3.

LIMIT variations in steady state power to + 20% within each operational period entered.

5.8.6.

Once all data has been entered PRESS the "Calculate" button to display the

% Damage Estimates.

5.8.7.

PRESS the Volumes" button to display the following screen (same as 5.7.8):

See 5.8.7.1l Reactor Codant System RCS(miJ.:

t Su~~ppression ChmbrL-qi (mU I3.26E+09 ee58..

Conamt Atmos ph*ere (cc)S 5.8.7.3 i '

<4 - :.' -':.'Sppression Chmbr Atisphere (c)b '.i.2+9,

~~r -i-7 7-,;

77"77

_F R

See 5.8.7.5 Note.87.

1.

Program enters default RCS volume, which the user may change based on RPV Level Readings at time of sample.

EP-AA-1 10-301 Revision 2 Page 22 of 25 BWR CDAM User Guide Page 17 of 20

2.

Program enters default Suppression Chamber volume, which the user may change based on readings at time of sample.

3.

Program enters default Containment free air volume which user may change based on conditions at time of sample. Unless there has been significant flooding of drywell this value will not change.

4.

Program enters default Suppression Chamber free air volume which user may change based on conditions at time of sample. If there has been a significant increase or decrease in the water level in the Suppression Pool or Torus then the free air volume will change.

NOTE:

Pressing the "Reset" button will reset all volumes to default values.

5.

PRESS the "Back" button to return to the Liquid or Gaseous screen, which user used to call volume form.

5.8.8.

PRESS the "Graph" button to display the following screen:.

.IR

.9M-I ILCcc F.60E-01 i

~Saimle Location.

~. -

ContirnentAtmosphee

~Syxtem in Eqiidurnm N/k

_ _ _ _ _ _ _ _ _ _ _ ~~;:, O ~ ei a OioU V

Na A. 3 DeL~

4_1 i

'IE -2

_L w Z Z

_10 Xi~%Cada~~

I See 5.8.8.1 r NOTE:

Graph shows High, Low, and Best melt curves; High, Low, and Best clad damage curves, and a red line across graph indicating entered.

EP-AA-1 10-301 Revision 2 Page 23 of 25 BWR CDAM User Guide Page 18 of 20

1.

PRESS the "Print" button to print a graph and summary of inputs.

2.

PRESS the "Back" button to go back to liquid or gaseous form which called this form.

5.8.9.

PRESS the "Back" button to return to the Summary Screen.

6.

CORE DAMAGE

SUMMARY

REPORT

.6.1.

Once the program user enters data for all available assessment methods and the program calculates damage based on inputs, SELECT the "Print" button to print a summary of all methods used.

6.2.

The values presented in the Assessment Methods section of the summary report show that they are in percent (%). Containment Hydrogen values are also in percent (but do not show the % symbol)..

(Sample report on next page.)

EP-AA-1 10-301 Revision 2 Page 24 of 25 Attachment I BWR CDAM User Guide Page 19 of 20 CDAM Method:

Core Damage Summary.

Station: 03 Clinton O Dresden O LaSalle, z Quad Cities

'Assessment Methods:

Melt Clad ContainmeitRadiation Mon'lors Cortainment:

29%

SupperssionChamber. l,C1%

l 2- -l Core Conditions -

-Core Cooling:...

Clad Damage Core.Uncovery.Time:

No Core Danage SRM 'Count Rate:

No Core Damage Core Temp:

Clad Falure.

Containment Hydrogen*.

1 l

2Q8 Sample'AnalVsis`

Ratios:

Fuo Melt

'Abnormal lsotopes:

.6 of 19 Present RCS: Liquid Samples: l 0% l%

Chamber:

GasSamples -'

100%.

. These methods should NOT be used for quartatbe or quantitative assessment except in the caseof a LOCA.

I Analyst's Estimate:

0 No Core Damage 0 Clad ding Failure 0 Fuel Melt Ariount.

NRC Core Condition Category:

Degree of Minor Intermediate Major Degraclation'

.(<10 )

'111A.50%6) 1(>50% )

No Core Da ae 1

1 1

Cladding Failure

.'2 3

'4 Fuel Overheat 5

6 7

F uel Melt

.-8.

9 10 Generaed BY:

Name: ____________________________

Date:,

12f05f02

'Time:'.

8:29 AM Core Darn~~~~~~~~~~~Je Summaiy Exelon BWRCDA~~~~~~~~~~~~~~~~~~~

Core Damage Summary Exelon BWR CDAM v1.0

EP-AA-110-301 Revision 2 Page 25 of 25 Attachment I BWR CDAM User Guide Page 20 of 20 6.3.

The Individual tasked with assessing core damage shall then ANALYZE the report to determine best estimate of type and amount of damage.

NOTE:

The CDAM program does not use the Fuel Overheat Condition Category 6.4.

Based on estimated type and amount of damage and following table (table also printed on summary report) ASSIGN NRC Core Condition Category (1-4 or 8 -10).

NRC Core Condition Categories Degree of Minor Intermediate Major Degradation

(<10%)

(10% to 50%)

(>50)

No Core Damage 1

I I

Cladding Failure 2

3 4

Fuel Overheat 5

6 7

Fuel Melt 9

10

7.

QUITING, OR EXITING, THE PROGRAM NOTE:

When the program is closed all data is reset.

CAUTION Program saves no information to disk; printed reports serve as record of core.damage assessments.

.7.1.

PRESS the "Quit" button on the Summary Screen exits the program.

RS-03-208, Revisions to the Exelon Nuclear Standardized Radiological Emergency Plan Implementing Procedure

Contents

Text

Subject:

2.7 DAPAR

2.9 Deposition

SUMMARY

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RS-03-208, Revisions to the Exelon Nuclear Standardized Radiological Emergency Plan Implementing Procedure

Text

Subject:

2.7 DAPAR

2.9 Deposition

SUMMARY

Navigation menu

Search