ML023330228

Part 2 of 2 - Carolina Power & Light Co., H. B. Robinson Steam Electric Plant, Unit No. 2 Plant Operating Manual Volume 2 Part 5 Emergency Procedure EPRAD-03 Dose Projections, Revision 12

ML023330228
Person / Time
Site:	Robinson
Issue date:	11/22/2002
From:	Carolina Power & Light Co
To:	Document Control Desk, Office of Nuclear Security and Incident Response
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Download: ML023330228 (143)

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CAROLINA POWER & LIGHT COMPANY H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 PLANT OPERATING MANUAL VOLUME 2 PART 5 EMERGENCY PROCEDURE EPRAD-03 DOSE PROJECTIONS REVISION 12 I EPRAD-03 I

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Page 1 of 8o0 CP&L R

Reference Use

SUMMARY

OF CHANGES IEPRAD-03 I

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Page 2 of 801 STEP #

REVISION COMMENTS.3.5.7 Added instructions to use detector efficiencies from Table 2 for completing Parts A and E of the attachment.

Added instructions to use detector efficiencies from Table 1 for completing Parts F and G of the attachment...3.5.9 Added Table 2 to the attachment representing detector efficiencies for each dose projection rad monitor vs time per EC 49849.

Entire Procedure Page numbering revised to reflect AP-007 format.

Corrected typographical errors throughout the procedure.

TABLE OF CONTENTS PAGE SECTI 8.3.1 8.3.2 8.3.3 8.3.4 8.3.5 Definitions and Abbreviations...................................................

45 General Information.................................................................

48 Quality Codes...........................................................................

49 Core Uncovery Time Determination..........................................

50 Accident Mitigation Systems.....................................................

51 Obtaining and Updating Meteorological Data...........................

52 Source Time Determination.....................................................

57 Flow Rates................................................................................

61 Detector Sensitivities...............................................................

66 Measuring Radiation Level on Main Steam Lines.....................

71 Typical RMS Values..................................................................

72 RMS Monitored Systems..........................................................

74 W eather Service Data...............................................................

75 Onsite Meteorological Data......................................................

76 Meteorological Forecast Form..................................................

77 HBRDOSE/RASCAL Comparison Matrix..................................

78 Manual Calculation of Curies Released....................................

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Page 3 of 80 ION PURPOSE.............................................................................................

4 RESPONSIBILITIES..............................................................................

4 INSTRUCTIONS...................................................................................

4 8.3.3.1 Use of the Dose Projection Program in the Control Room.......... 4 8.3.3.2 Use of the Dose Projection Program by the Dose Projection Team.........................................................................................

19 8.3.3.3 Interpretation of the Dose Projection Summary Table.............. 41 RECORDS...........................................................................................

43 ATTACHMENTS.................................................................................

43 8.3.5.1 8.3.5.2 8.3.5.3 8.3.5.4 8.3.5.5 8.3.5.6 8.3.5.7 8.3.5.8 8.3.5.9 8.3.5.10 8.3.5.11 8.3.5.12 8.3.5.13 8.3.5.14 8.3.5.15 8.3.5.16 8.3.5.17

8.3.1 PURPOSE This procedure provides instruction for performing dose projections in case of possible offsite emergencies from a release of airborne radioactivity.

8.3.2 RESPONSIBILITIES

1.

Operations personnel under the direction of the Site Emergency Coordinator are responsible for performing the Control Room portion of this procedure until the Dose Projection Team is activated.

2.

The Radiological Control Manager or the Dose Projection Team Leader is responsible for calculating the TEDE and the thyroid CDE, to be used by the Radiological Control Manager and the Emergency Response Manager in determining and evaluating possible off-site consequences from a release of airborne radioactivity.

NOTE:

Due to the complexity and branching nature of this procedure a slightly different numbering convention from other Emergency Procedures (EP) is used.

Additionally, this section contains several Attachments to assist the user that are not specifically referenced in the body of the section.

8.3.3 INSTRUCTIONS

1.

USE OF THE DOSE PROJECTION PROGRAM IN THE CONTROL ROOM 1.1 Accessing The Dose Projection Computer Program.

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8.3.3 (Continued)

NOTE:

This section represents a systematic approach to access the dose projection computer program. Steps must be followed in order and you must be logged into EDS with an event declared. Any problems in accessing the program must be promptly reported to Information Technology personnel for resolution.

1.1.1 IF the ERFIS terminal to be used is initially aligned to the site LAN, THEN align the terminal to the ERFIS system.

NOTE:

Inability to link with the ERFIS host is indicated by the following:

IF initially in ERFIS during the failure an "ERR1 1 COMMUNICATIONS TIMED OUT!!!" message on the top line of the man machine interface, and the EDS icon in the upper right corner will turn red.

IF initially in the site LAN during the failure Error text on the screen and a login prompt.

1.1.2 IF the ERFIS terminal was in the site LAN at the time of the failure and error text and a login prompt are displayed, THEN proceed to Step 1.1.5 for "local mode" operations.

1.1.3 IF the ERFIS terminal to be used is aligned and linked to ERFIS, THEN dose projection may be accessed by typing the turn on code "hbrdose" in the man-machine interface or alternately, from the main menu select the "EP" Menu, then choose "hbrdose."

1.

IF the dose projection program is accessed, THEN proceed to Section 1.2 "Control Room Dose Projection."

2.

IF the dose projection program is not accessed, THEN proceed to the next step.

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8.3.3 (Continued) 1.1.4 IF the ERFIS terminal to be used is aligned but not linked to ERFIS, THEN dose projection may still be used, however, radiological and meteorological data must be manually entered.

1.

Type the turn on code "hbrdose" in the man-machine interface.

2.

Notify Information Technology personnel of problems as soon as practical.

3.

IF the dose projection program is accessed, THEN proceed to Section 1.2 "Control Room Dose Projection."

4.

IF the dose projection is not accessed, THEN proceed to the next step.

1.1.5 IF the ERFIS terminal to be used can not be aligned to ERFIS, THEN continue in this section to configure the ERFIS terminal for "local mode."

NOTE:

As long as the ERFIS terminal has power the following sub-steps should align the terminal to perform dose projection in "local mode." This method will require manual entry of radiological and meteorological data.

1.

Reboot the ERFIS terminal by depressing CTRL, ALT, SHIFT, DEL (numeric keypad delete must be used) simultaneously. The computer may take up to 10 seconds to respond to this key sequence.

2.

Choose ERFIS/EDS from the System Commander.

3.

IF a grey QNX window appears, THEN press the right mouse button to get the menu, choose "exit," and confirm the exit choice. Otherwise skip this step and proceed to the next step.

4.

When the cursor is displayed, possibly after various system messages, press "ENTER" if the login prompt is not displayed. A "Login" prompt should appear.

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1.1.5 (Continued)

5.

At the login prompt type "hbrdose" and press "Enter."

6.

At the password prompt type "hbrdose" and press "Enter."

Hbrdose will automatically start after this step.

7.

Do not attempt to print or make electronic notifications in "local mode," as this will further degrade execution of the program.

8.

IF "local mode" worked, go to Section 1.2, "Control Room Dose Projection."

Contact Information Technology personnel for instructions to return the ERFIS terminal to normal when dose projections are complete.

1.1.6 IF "local mode" did not work on the initial ERFIS terminal, THEN repeat Step 1.1.5 on another ERFIS terminal, and request that Information Technology personnel immediately bring a computer to the Control Room with a current version of the dose projection program installed.

1.

Information Technology personnel will provide instructions on accessing the program.

2.

Proceed to Section 1.2, "Control Room Dose Projection."

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1.2 Control Room Dose Projection:

1.2.1 With the mouse, left click the Projection menu item.

1.2.2 Left click the Control RM menu item.

NOTE:

Within each release pathway, the left mouse button can be used to move the cursor to the desired field. Depressing the left mouse button will also select or deselect any of the monitors.

CAUTION DO NOT USE radiation monitors that are out-of-service for dose projections.

Verify that ERFIS data is correct by comparing it to the control room readouts if the RMS/ERFIS interface multiplexer is in alarm.

When manually entering data in hbrdose do not leave blank spaces between characters, (e.g., use 3,000,000 or 3E6 NOT 3 E6).

1.2.3 CONTAINMENT== ENVIRONMENT Group

a.

IF no monitor in the CONTAINMENT=*ENVIRONMENT group is in alarm, Make sure that none of the check boxes are selected. You can deselect any monitor by pressing the left mouse button on the desired monitor. Go to step 1.2.4.

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IF at any time the computer locks up while performing the following steps, THEN depress CTRL, ALT, DEL simultaneously, or if that is unsuccessful depress CTRL, ALT, SHIFT, DEL (numeric keypad delete must be used) simultaneously, AND GO TO step 1.1.

1.2.3 (Continued)

NOTE:

mR/hr while When performing the following step be aware that R-2 is in units of R-32 A&B are in R/hr.

b.

IF R-32A OR, R-32B OR R-2 are in alarm, THEN select the alarming monitor that has the highest radiation level AND GO TO step 1.2.3.e.

c.

IF R-12 is in alarm and is aligned to the CV, THEN select R-12 AND GO TO Step 1.2.3.e.

d.

IF R-12 is in alarm AND is aligned to the plant vent, THEN GO TO Step 1.2.4.

e.

IF a bad quality code OR no data is being displayed for the selected monitor, THEN manually enter the reading from the radiation monitor drawer.

NOTE:

The default CV release flow of 1.5 CFM is based on the CV design leak rate at design basis CV pressure.

f.

IF containment integrity is maintained, THEN go to step 1.2.4.

g.

IF containment integrity is not maintained, THEN enter the leakrate that is escaping through an unmonitored pathway in the FLOW field (next to R-32A).

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1.2.4 PLANT VENT STACK Group

a.

IF no monitors in the PLANT VENT STACK group are in alarm, THEN GO TO Step 1.2.5.

b.

IF R-14E is above 50 cpm, THEN select R-14E AND GO TO Step 1.2.4.f.

c.

IF R-14D is above 12 cpm, THEN select R-14D AND GO TO Step 1.2.4.f.

d.

IF R-14C is in alarm, THEN select R-14C AND GO TO Step 1.2.4.f.

e.

IF R-21 is in alarm, THEN select R-21.

f.

IF a bad quality code OR no data is being displayed for the selected monitor, THEN manually enter the reading from the radiation monitor drawer.

g.

IF R-21 was selected, THEN GO TO Step 1.2.5.

h.

IF a good quality code is provided for stack FLOW THEN GO TO Step 1.2.4.i.

i.

Select the ventilation units which are operating.

1.2.5 R-1 2

a.

IF R-12 is not in alarm THEN GO TO Step 1.2.6.

b.

IF R-12 is aligned to the CV THEN GO TO Step 1.2.6.

c.

IF R-14C OR R-14D OR R-14E were selected in Step 1.2.4 THEN GO TO Step 1.2.6.

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1.2.5 (Continued)

d.

IF R-21 was selected in Step 1.2.4 AND HVE-15 is the only release pathway THEN GO TO Step.1.2.6.

e.

Select R-12.

f.

IF a bad quality code OR no data is being displayed for R-1 2, THEN manually enter the reading from the radiation monitor drawer.

NOTE:

The groups of HVE units in the PLANT VENT STACK group or.3.5.8, Flow Rates, can be used to determine the following flowrate.

g.

Move the cursor to the plant vent stack flow field and enter the flow that is going up the plant vent stack.

1.2.6 LOWER FHB => ENVIRONMENT Group

a.

IF no monitors in the Lower FHB =t ENVIRONMENT group are in alarm, THEN GO TO Step 1.2.7.

b.

IF R-20 is in alarm AND has not failed high, THEN select R-20 AND GO TO Step 1.2.6.d.

c.

IF R-20 has failed high, THEN select R-30.

d.

IF a bad quality code OR no data is being displayed for the selected monitor, THEN manually enter the reading from the radiation monitor drawer.

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1.2.7 SECONDARY RELEASE

a.

IF NO R-31 monitors are above one mrem/hr, THEN GO TO Step 1.2.8.

b.

IF steam/water from the Main Steam Line that has a monitor above one mrem/hr is escaping through a faulted Main Steam System outside containment, THEN GO TO step 1.2.7.d.

c.

IF NO PORVs AND NO SRVs are open, THEN GO TO Step 1.2.8.

d.

Select PORV/SRV.

e.

IF a bad quality code OR no data is being displayed for the selected monitor, THEN manually enter the reading from the radiation monitor drawer.

f.

IF the release is due to a Main Steam System fault as described in Step 1.2.7.b., THEN using the Main Steam indications on the RTGB and Attachment 8.3.5.8 to compare flowrates, enter the PORV and SRV combination that would produce the same flow rate as the fault, in the fields below the monitor(s) in alarm, AND GO TO Step 1.2.7.1.i.

g.

IF the PORV on the Main Steam Line(s) that has the monitor(s) in alarm are open, THEN enter 1 in the PORV field below the monitor(s)in alarm.

h.

IF any SRV(s) on the Main Steam Line(s) that has(have) the monitor(s) in alarm are open, THEN enter the number open in the SRV field below the monitor(s) in alarm.

i.

IF a bad quality code OR no data is being displayed in the SG Press field(s) below the monitor(s) in alarm, THEN manually enter the correct pressure as obtained from control room readouts.

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1.2.8 STEAM = CONDENSER

a.

IF R-15 is not in alarm, THEN GO TO Step 1.2.9.

b.

IF R-14C OR R-14D OR R-14E were not selected in the Plant Vent Stack group OR R-1 2 was not selected in Step 1.2.5.f, THEN select Steam = Condenser.

c.

IF R-15 is in alarm and not failed high THEN select STEAM=,CONDENSER AND go to 1.2.8.d.

d.

IF a bad quality code OR no data is being displayed for R-1 5, THEN manually enter the reading from the radiation monitor drawer.

e.

IF only one vacuum pump is running, THEN select 310 CFM flow AND GO TO Step 1.2.9.

f.

IF two vacuum pumps are running, THEN select 610 CFM flow.

1.2.9 Left click the Done button.

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1.2.10 SPECTRUM DETERMINATION

a.

IF the incident does not involve the reactor (i.e. spent fuel, waste gas, old spent fuel), THEN GO TO step 1.2.1 0.f

b.

IF the incident involves the reactor AND the core has not been uncovered, THEN GO TO step 1.2.10.e.

c.

Select the time that the core has been uncovered:

< 30 minutes 0.5 - 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (30 minutes - 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 48 minutes)

> 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 48 minutes)

d.

GO TO step 1.2.11.

e.

IF the incident involves mechanical damage to fuel in the reactor, THEN select'< 30 minutes' AND GO TO step 1.2.11.

f.

IF the incident involves a Waste Gas Decay Tank, THEN select WASTE GAS AND GO TO step 1.2.11.

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The core uncovery time will be the time that a RED status occurs on the CORE COOLING critical safety function status tre6 until the tree conditions return to yellow status (core covered and core exit thermocouples

< 7000 F). This determination can be performed on the TV monitor or on the manual board.

1.2.10 (Continued)

NOTE:

Spent fuel that is being shipped or is in preparation for shipment should be classified as OLD SPENT FUEL. Assume thafthe spent fuel has been out of the reactor core for less than three years if the true time is unknown.

g.

IF the incident is a fuel handling accident AND involves spent fuel that has been out of the reactor core for less than three years THEN select SPENT FUEL AND GO TO step 1.2.11.

h.

IF the incident is a fuel handling accident AND involves spent fuel that has been out of the reactor core for more than three years THEN select OLD SPENT FUEL.

1.2.11 FILTRATION/CV SPRAYS/PARTITIONING NOTE:.3.5.5, Accident Mitigation Systems, of this procedure describes if Filtration/CV Sprays/Partitioning are "Effective" or "Not Effective".

a.

IF Filtration OR CV Sprays OR Partitioning are effective, THEN select Effective AND GO TO Step 1.2.12.

b.

IF Filtration OR CV Sprays OF Partitioning are not effective, THEN select Not Effective.

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1.2.12 METEOROLOGY CAUTION If direct access to the meteorological tower has failed, the data will appear colored red in the input fields. DO NOT USE THIS DATA.

a.

IF meteorological data with a good quality code is being displayed, THEN GO TO Step 1.2.12.m.

b.

IF a bad quality code OR no data is being displayed for meteorological data, THEN left click the REFRESH button to make a second attempt to acquire meteorological data from ERFIS.

c.

IF the computer makes the connection to the meteorological tower AND the meteorological data is properly updated (in accordance with Caution Statement above), THEN GO TO Step 1.2.12.m.

d.

IF meteorological data is not available from the control room computer, THEN manually update the meteorological data.

e.

Call the CP&L offsite meteorological contact (See the ERO Phone Book for number).

f.

IF meteorological data is available from the CP&L offsite meteorological contact, THEN manually update the meteorological data AND GO TO Step 1.2.12.k.

NOTE:

If the Florence Airport or the National Weather Service office is called, the only information that can be obtained is the wind direction, wind speed, and ambient temperature. Stability factor must be obtained from Step 1.2.12.k of this procedure. If wind speed and direction are only supplied for one point enter these values in both the elevated and ground fields. Do not enter wind gust as the wind speed, and if no Delta T is supplied do not enter one.

g.

Call the Florence Airport: (See the ERO Phone Book for numbers)

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1.2.12 (Continued)

h.

IF meteorological data is available from the Florence Airport, THEN manually update the meteorological data AND GO TO Step 1.2.13.

L.

Call the National Weather Service office in Columbia, South Carolina: (See the ERO Phone Book for numbers)

j.

IF meteorological data is available from the National Weather Service, THEN manually update the meteorological data.

k.

IF there is no stability class data available, THEN make an estimate of the current Atmospheric Stability Class by visual observation, using the following table:

light wind or calm

(< 11.5 mph) moderately strong wind(- 11.5 mph)

Rain, Day or NghtD D

D Sunny Day B

C Cloudy Day C

D Cloudy Night E

D Clear Night F

D I.

Enter the stability class in the appropriate field.

m.

Left click the Shutdown time field.

1.2.13 REACTOR SHUTDOWN TIME

a.

IF the reactor is not shutdown, THEN GO TO Step 1.2.14.

b.

IF the reactor is shutdown AND the shutdown time is not displayed OR is not correct, THEN manually enter the date and time of shutdown in the space provided.

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RELEASE DURATION NOTE:

The estimated release duration should be from the start of the release until the projected time that the release should stop. This can be determined by estimating the completion of a damage control mission, performance of a repair to stop the release, or the estimated time until the RCS, CV Sump, or Steam Generator is below 2000 F.

a.

IF the release duration is known, THEN manually enter the release duration in the field provided AND go to Step 1.2.15.

b.

IF the release duration is unknown AND an estimate is available, THEN enter the estimated time in the field provided AND go to Step 1.2.15.

c.

IF the release duration is unknown AND no estimate is available, THEN enter 1 in the field provided AND go to Step 1.2.15.

1.2.15 Left click the Done button.

1.2.16 The dose projection will be given on the screen.

1.2.17 Using the information supplied notify the government officials as per the requirements of EPNOT-01.

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2.

USE OF THE DOSE PROJECTION PROGRAM BY THE DOSE PROJECTION TEAM 2.1 Access the software using Section 1.1 of this procedure as a guideline, and return to this section instead of the Control Room Dose Projection section.

2.2 The main menu screen will appear. The items in this menu can be accessed by clicking the left button on any of these items.

2.3 Six menu topics are available for use. They are listed here along with the section in this procedure which explains their use.

Projection 2.4 Contingency 2.5 Int Phase 2.6 Graphics 2.7 Utilities 2.8 Exit 2.9 NOTE:

The Dose Projection Program should be used to calculate the "Total Dose" from the start of the release until the projected end. To do so conservatively, the Dose Projection Team may decide to use the estimated peak release rate throughout the release period. If no information is available, the current release rate should be considered constant throughout the release period.

2.4 PROJECTION This menu item should be used to perform early phase dose projections based on plant radiation monitors, plant samples, or environmental samples.

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2.4.1 Control Room This function should be used by control room personnel. Its use is described in Step 1.2 of this procedure.

2.4.2 RMS This function should be used by the dose projection team to perform dose projections when adequate data is available from the plant effluent monitors. The following steps are for guidance and are not required to be performed in entirety or in the order in which they are given.

2.4.2.1 The first screen that will appear when this menu option is selected is the ROBINSON EFFLUENT MONITORS screen.

The following items should be considered when using this screen:

CAUTION DO NOT perform an official dose projection using data obtained from a radiation monitor that is out-of-service. The control room may be contacted to determine any monitors that are out-of-service that may have good quality codes on ERFIS.

This could occur when the RMS/ERFIS interface multiplexer fails.

When manually entering data in hbrdose do not leave blank spaces between characters, (e.g., use 3,000,000 or 3E6 NOT 3 E6).

NOTE:.3.5.12, RMS Monitored Systems can be used to determine the relationship between radiation monitors and effluent pathways.

Data that is on this screen will have quality codes in.3.5.3, Quality Codes of this procedure.

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2.4.2.1 (Continued)

Dose projections for more than one release pathway can be performed using this program. The monitors are grouped on this screen according to the release pathway that they monitor. Therefore, only one monitor from each group can be selected each time the dose projection program is executed.

The Containment to Environment release path is a valid release path in most situations even when no containment leakage has been identified. The 1.5 cfm flow is the design basis release rate when the CV is pressurized to design basis pressure. If containment leakage is into the Auxiliary Building and the release is monitored by a stack monitor a dose projection using the containment monitors is not necessary.

NOTE:

If a projection is made using R-1 2 aligned to the plant vent it will be based on a ground level release pathway instead of a mixed mode release.

R-12 is usually lined up to the containment atmosphere. If a dose projection is performed using R-12, ensure that it aligned the way that it is being used.

R-12 can be used to perform a dose projection when it is aligned to the plant vent. In order to accomplish this the R-12 plant vent flowrate must be manually entered into the containment monitors flowrate field.

If this is performed, then it can not be accomplished at the same time that a dose projection is being performed based on containment leakage. For this reason, if a projection is needed based on both release paths, they must be performed separately and manually added together.

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2.4.2.1 (Continued)

The flow rate for the R-1 4 monitors will be automatically updated by ERFIS to reflect the plant vent stack when ERFIS is available.

R-21 has a default flow rate which is equivalent to the flow of HVE-15. This default value should normally be used, because this is the effluent volume that this detector is monitoring.

R-20 and R-30 have a default flow rate which is equivalent to the flow of HVE-14.

R-31A, 31B, and 31C should not normally be used if they are below 1 mrem/hr. However, if they are, they must be background corrected and manually entered.

Obtain the latest valid normal reading from the weekly background/alarm setpoint check or other source (Attachment 8.3.5.11, Typical RMS Data, may be used) and subtract the normal reading from the control room readout and enter this value.

If a release is due to a faulted steam line, a dose projection can be performed by selecting the number of SRV's and PORV that would approximate the release (use Attachment 8.3.5.8, Flow Rates). The UNKNOWN MIX under the CONTINGENCY menu can be used to perform a dose projection under this condition using Attachment 8.3.5.7, Source Term Determination, Part E.

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2.4.2.1 (Continued)

NOTE:

If a dose projection is performed using R-15 and the release is due to a break in the line between the condenser vacuum pumps and the plant vent stack, the projection will be based on a mixed mode release, instead of a ground level release.

The program will allow you to perform dose projections using the R-1 5 monitor and the plant vent monitors at the same time. However, this should only be done under the following circumstances:

R-15 is above background and the line from the condenser vacuum pumps to the plant vent is allowing leakage, OR R-15 is above background and NEITHER R 14C, R-14D, R-14E, NOR R-12 when it is aligned to the plant vent, are being used for a dose projection.

2A.2.2 Once selections have been made on this screen select the DONE field.

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2.4.2.3 The SPECTRUM DETERMINATION screen is the next screen that will appear. Several characteristics of the incident must be entered on this screen in order to identify the source term.

If the incident involves the reactor, the time that the reactor core has been uncovered must be selected using the guidelines in Attachment 8.3.5.4, Core Uncovery Time Determination of this procedure.

If the incident involves Spent Fuel, regardless of the location, you must identify if the fuel is Spent Fuel or Old Spent Fuel. Old Spent Fuel is fuel that has not been in the reactor while critical for three years or more.

If the incident involves a Waste Gas Decay Tank select the Waste Gas option.

The RELEASE DURATION should be determined and entered in the appropriate field. If an estimate of the time is not known one hour can be used here until better information is available.

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Only one of the following conditions can exist for each execution of the Dose Projection Program. If more than one of the following conditions exist, execute the projection more than once using the appropriate effluent monitors to accurately quantify the effluent.

NOTE:

The estimated release duration should be from the start of the release until the projected time that the release should stop. This can be determined by estimating the completion of a damage control mission, performance of a repair to stop the release, or the estimated time until the RCS, CV Sump, or Steam Generator is below 2000 F.

2.4.2.3 (Continued)

NOTE:

A Plant Operations Advisor, SRO on the Accident Assessment Team, or the Shift Technical Advisor should be consulted to determine whether these mitigation systems are operable.

The effectiveness of FILTRATION/CV SPRAYS/PARTITIONING should be determined.

Use the guidelines in Attachment 8.3.5.5, Accident Mitigation Systems, of this procedure to make this determination.

If the quality codes for the meteorology data are good they may be used. If the quality codes are bad or there is other reason to question them, complete this section using Attachment 8.3.5.6, Obtaining and Updating Meteorological Data, of this procedure for guidance.

If the plant has SHUTDOWN enter the shutdown date and time in the appropriate fields. Otherwise, these fields can be left as they appear.

The DONE field should be selected when all of the information on this screen has been entered.

2.4.2.4 The Projection Screen is the final screen to appear. It is explained in Step 3 of this procedure.

2.4.3 PLANT SAMPLE This function is for use by the dose projection team to perform dose projections based on plant samples of the effluent stream. It should be used as needed by the dose projection team.

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2.4.3.1 The screen that will appear when this menu option is selected should be completed with the help of the following guidelines.

Enter the activity of each nuclide that is listed on the screen that is available from the plant sample analysis.

Identify the release height,of the effluent. Select mixed if the release is through the plant vent regardless of the wind speed. Select ground if the release is by any other pathway, or if the pathway is unknown.

The time from sample to release is provided to correct the sample activity for any radioactive decay that has occurred in the sample effluent between the time the sample was collected and the time of the release. DO NOT enter a value in this field unless you wish to decay correct the effluent stream.

Enter the flowrate in cfm of the effluent stream. Care should be taken to understand where the sample was obtained, and ensure that the FLOW field data corresponds with the flow of the sampled air with no further dilution. A flowrate can be manually entered using the flowrates in Attachment 8.3.5.8, Flow Rates, as a reference, or a flow can be selected by selecting the FLOWS field.

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2.4.3.1 (Continued)

NOTE:

The estimated release duration should be from the start of the release until the projected time that the release should stop. This can be determined by estimating the completion of a damage control mission, performance of a repair to stop the release, or the estimated time until the RCS, CV Sump, or Steam Generator is below 2000 F.

The release duration should be determined and entered in the appropriate field. If an estimate of the time is not known, one hour can be used until better information is available.

If the quality codes for the meteorology data are good they may be used. If the quality codes are bad or there is other reason to question them, complete this section using Attachment 8.3.5.6, Obtaining and Updating Meteorological Data of this procedure for guidance.

If the plant has shutdown, enter the shutdown date and time in the appropriate fields.

The DONE field should be selected when all of the information on this screen has been entered.

2.4.3.2 The Projection Screen is the final screen to appear. It is explained in Step 3 of this procedure.

2.4.4 ENVIRONMENTAL SAMPLE This function is for use by the dose projection team to perform dose projections based on environmental samples. It should be used as needed by the dose projection team.

EPRAD-03 Rev. 12 Page 27 of 80]

NOTE:

Protective action recommendations are required to be made within 15 minutes of obtaining environmental sample results. (AR #48774) 2.4.4.1 The screen that will appear when this menu option is selected is titled as the ENVIRONMENTAL MONITORING TEAM. It should be completed with the help of the following guidelines.

Enter the closed window dose rate (in mrem/hr) that is obtained at a height of approximately one meter above the ground. The value should be obtained from the Environmental Monitoring Team Leader and should reflect the most recent data that is available from near the centerline of the plume.

For the air sample dose rate select the CART field, and enter data in the appropriate fields using the following guidance:

Enter the sample volume in cubic feet.

Select whether count rate or dose rate will be entered.

Enter the count rate or dose rate on contact with the iodine cartridge. This data should be obtained from the Environmental Monitoring Team Leader and should reflect the most recent data that is available from the centerline of the plume.

Select the DONE field and the program will calculate the Thyroid Committed Dose Rate.

(This calculation is based on Attachments in EPRAD-01, Environmental Monitoring.)

Select the CANCEL field to exit this window or click the mouse outside of the window.

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2.4.4.1 (Continued)

Enter the downwind distance from the plant stack to the sample collection location..

Enter the direction from the plant that the sample was collected in degrees.

The release duration should be determined and entered in the appropriate field. If the time is not known one hour should be used here until better information is available.

Identify the release height of the effluent. Select mixed if the release is through the plant vent regardless of the wind speed. Select ground if the release is by any other pathway, or if the pathway is unknown.

If the quality codes for the meteorology data are good they may be used. If the quality codes are bad or there is other reason to question them, complete this section using step Attachment 8.3.5.6, Obtaining and Updating Meteorological Data, of this procedure for guidance.

If the plant has shutdown enter the shutdown date and time in the appropriate fields.

The DONE field should be selected when all of the information on this screen has been entered.

2.4.4.2 The Projection Screen is the final screen to appear. It is explained in Step 3 of this procedure.

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2.5 CONTINGENCY Contingency calculations are typically "what if" types of calculations that allow the Dose Projection Team to make predictions of off-site dose based on a projected event. However, they can be used to make actual dose projections.

2.5.1 KNOWN MIX This function allows the user to input the isotopic analysis of the release in order to perform a dose projection.

2.5.1.1 The screen that will appear when this menu option is selected should be completed with the help of the following guidelines.

Enter the activity of each nuclide that is listed on the screen which could be in a postulated release.

Identify the release height of the effluent. Select mixed if the release is through the plant vent regardless of the wind speed. Select ground if the release is by any other pathway, or if the pathway is unknown.

Enter the time from when the activities were determined until the release could begin. This is not required, it should only be entered when it is expected that the activity has decayed since the sample was pulled.

Enter the number of Curies that could be released..3.5.7, Source Term Determination, of this procedure can be used to help determine this value.

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2.5.1.1 (Continued)

If the quality codes for the meteorology data are good they may be used. If the quality codes are bad or there is other reason to question them, complete this section using Attachment 8.3.5.6, Obtaining and Updating Meteorological Data, of this procedure for guidance.

If the plant has shutdown enter the shutdown date and time in the appropriate fields.

The DONE field should be selected when all of the information on this screen has been entered.

2.5.1.2 The Projection Screen is the final screen to appear. It is explained in Step 3 of this procedure.

2.5.2 UNKNOWN MIX This function allows the user to project what the offsite dose to the public would be due to a release if the total activity of the release is known but the isotopic abundances are not known.

2.5.2.1 The screen that will appear when this menu option is selected should be completed with the help of the following guidelines.

NOTE:

Only one of the following conditions can exist for each execution of the Dose Projection Program. If more than one of the following conditions exist, execute the projection more than once using the appropriate effluent monitors to accurately quantify the effluent.

If the incident involves the reactor core the time that the reactor core has been uncovered or could be uncovered must be selected. Use Attachment 8.3.5.4, Core Uncovery Time Determination, to help make this determination.

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2.5.2.1 (Continued)

If the incident involves Spent Fuel, regardless of the location, you must identify if the fuel is Spent Fuel or Old Spent Fuel. Old Spent Fuel is fuel that has not been in the reactor while critical for three years or more.

If the incident involves a Waste Gas Decay Tank select the Waste Gas option.

Enter the number of Curies that could be released..3.5.7, Source Term Determination, of this procedure can be used to help determine this value.

Identify the release height of the effluent. Select mixed if the potential release is through the plant vent regardless of the wind speed. Select ground if the release is by any other pathway, or the pathway is unknown.

The effectiveness of Filtration/CV Sprays/Partitioning should be determined. The guidelines in Attachment 8.3.5.5, Accident Mitigation Systems, of this procedure should be used to make this determination.

If the quality codes for the meteorology data are good they may be used. If the quality codes are bad or there is other reason to question them, complete this section using.3.5.6, Obtaining and Updating Meteorological Data, of this procedure for guidance.

If the plant has shut down enter the shutdown date and time in the appropriate fields.

The DONE field should be selected when all of the information on this screen has been entered.

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2.5.2.2 The Projection Screen is the final screen to appear. It is explained in Step 3 of this procedure.

2.5.3 DEFAULTS This function allows the user to hypothesize what the offsite dose to the public would be due to a postulated release if plant conditions are unknown. A default should only be used when neither the total activity nor the isotopic analysis of the potential release are known.

2.5.3.1 The screen that will appear when this menu option is selected should be completed with the help of the following guidelines.

NOTE:

Only one of the following conditions can exist for each execution of the Dose Projection Program. If more than one of the following conditions exist, execute the projection more than once using the appropriate effluent monitors to accurately quantify the effluent.

If the incideni involves the reactor core the time that the reactor core has been uncovered or could be uncovered must be selected. Use Attachment 8.3.5.4, Core Uncovery Time Determination, to make this determination.

If the incident involves Spent Fuel, whether in the containment or in the Fuel Handling Building, you must identify if the fuel is Spent Fuel or Old Spent Fuel. Old Spent Fuel is fuel that has not been in the reactor while critical for three years or more.

EPRAD-03 Rev. 12 Page 33 of 80 CAUTION Calculated dose using defaults are EXTREMELY conservative and may assume all of the core activity is released, depending on the spectrum determination.

2.5.3.1 (Continued)

If the incident involves a Waste Gas Decay Tank select the Waste Gas option.

• Identify the release height of the effluent. Select mixed if the potential release is through the plant vent regardless of the wind speed. Select ground if the release is by any other pathway, or the pathway is unknown.

The effectiveness of Filtration/CV Sprays/Partitioning should be determined. The guidelines Attachment 8.3.5.5, Accident Mitigation Systems, of this procedure should be used for making this determination.

If the quality codes for the meteorology data are good they may be used. If the quality codes are bad or there is other reason to question them, complete this section using Attachment 8.3.5.6, Obtaining and Updating Meteorological Data, of this procedure for guidance.

If the plant has shutdown enter the shutdown date and time in the appropriate fields.

The DONE field should be selected when all of the information on this screen has been entered.

2.5.3.2 The Projection Screen is the final screen to appear. It is explained in Step 3 of this procedure.

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2.6 INTERMEDIATE PHASE Intermediate phase calculations are used during the intermediate phase of an emergency to project the one year, two year,. and fifty year committed dose to the public due to exposure from contamination deposited on the ground. The calculations are based on environmental data.

2.6.1 DOSE RATE This function is used to calculate the projected doses using dose rate data from environmental monitoring teams.

2.6.1.1 The screen that will appear when this menu option is selected should be completed with the help of the following guidelines.

It should be determined if weathering (radioactive decay is also included in this factor) will be considered when performing this function. To do this select the UTILITIES function from the main menu, and follow the guidelines in Step 2.8.1 of this procedure.

Enter the closed window dose rate in mrem/hr taken at approximately one meter from the ground in the 1 meter dose rate field.

Enter the straight line distance in miles or fractions of miles from the plant vent that the sample was taken.

Enter the bearing in degrees from the plant for the sample location.

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2.6.1.1 (Continued)

NOTE:

If no data has been entered under the "Sample" then the "Average Spectrum" choice will not appear, and only the Default Spectrum can be used.

Select if the Default Spectrum or the Average Spectrum should be used to perform the projection.

The Average Spectrum should be selected here when adequate data has been entered in the "Sample" screen (Step 2.6.2).

Select the DONE field when complete and the dose will be given.

2.6.2 Sample This function is used to calculate the projected dose using isotopic analysis of samples collected by environmental monitoring teams.

2.6.2.1 The screen that will appear when this menu option is selected should be completed with the help of the following guidelines.

It should be determined if weathering (radioactive decay is also included in this factor) will be considered when performning this function. To do this select the UTILITIES function from the main menu, and follow the guidelines in Step 2.8.1 of this procedure.

Enter the activity of each nuclide that is present in the sample that is listed on this screen. These activities should be entered in units of pCi/m 2. The nuclides that are listed on this screen are the only ones in RNP's anticipated source term that have a long enough half-life to contribute significant dose.

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2.6.2.1 (Continued)

Enter the sample identification number. This will normally be our radiochemistry form number.

Determine if the sample should be added to the sample data base from which the average deposition is calculated. Choosing this option will also include the sample in the average spectrum function of the DOSE RATE option.

Enter the straight line distance in miles or fractions of miles from the plant vent that the sample was taken.

Enter the direction in degrees from the plant for the sample location.

Select the DONE field when complete and the dose will be given.

2.7 GRAPHICS This menu item provides a graphic display of the 10 mile and 50 mile Emergency Planning Zones. It should be used as an aid by the Dose Projection Team to help with Protection Action Recommendations, and determining the adequacy of the environmental monitoring efforts.

NOTE:

The latest graphics of the 10 MILE ISOPHLETHS and the 10 MILE PARs are automatically saved to a disk file. They can be printed using the Microsoft Paintbrush program.

2.7.1 10 MILE ISOPLETHS This function provides a display of the isopleths within the 10 mile Emergency Planning Zone where the TEDE and Thyroid CDE limits are exceeded. If isopleths do not appear the EPA PAGs are not exceeded by the latest dose projection.

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2.7.2 10 MILE PARS This function displays the evacuation Protective Action Recommendations for the 10 Mile Emergency Planning Zone. This display consist of a five mile radius with a two mile keyhole superimposed on the map. The two mile keyhole applies to only the two mile sector (AO). Any of the five mile sectors (Al, B1, C1, D1, and El) that are intersected by the five mile radius keyhole should be evacuated. If the dose at the centerline of the plume exceeds the EPA PAGs at any distance five miles or beyond, then the radius of the keyhole is extended to ten miles. Any of the ten mile sectors (A2, B2, C2, D2, and E2) which are intersected by the ten mile keyhole should be evacuated.

2.7.3 10 MILE EMT POINTS This function provides a method to enter and display Environmental Monitoring Team Data in the 10 mile EPZ.

Click the mouse on the map location were the sample was taken.

Enter the closed window dose rate taken at approximately one meter above the ground. Use units of mrem/hr and depress the enter key when complete.

The computer will update the sample point with a color coded circle depending on the dose rate recorded at the location. These color codes are given in the upper right hand corner of the screen.

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2.7.4 50 MILE INT PHASE This function provides a method to enter Environmental Monitoring Team Data in the 50 mile EPZ, and calculate and display the 1, 2, and 50 year committed doses along with the skin dose.

Click the mouse on the map location were the sample was taken.

Enter the closed window dose rate taken at approximately one meter above the ground. Use units of mrem/hr and depress the enter key when complete.

The computer will update the sample point with a color code that represents if the program is above the EPA limits, above normal background, at background level. It will also calculate the 1, 2, and 50 year committed doses along with the skin dose.

2.8 UTILITIES This menu item is provided to assist the Dose Projection Team. These functions can be used at any time they are needed while using the program.

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2.8.1 WEATHERING This function is used when performing an Intermediate Phase Dose Projection to account for reductions in the source term due to weathering and radioactive decay.

2.8.2 PRINTING SETUP This function can be used to configure printing from this application. The user can configure custom printers and enable or disable the automatic printing of dose projection screens.

2.8.3 DISTANCES This function allows the user to adjust the distances from the plant that dose projections are calculated. This is done by identifying the maximum distance from the plant and the increment between each distance that is desired. This function is especially useful in determining distances close in to the site or beyond 10 miles.

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2.8.4 NOTIFICATION This function will automatically print out a State Notification Form in the proper format with the dose projection information completed when it is selected.

2.8.5 MANUAL SCREEN PRINT Use this menu item to print the currently displayed dialog or screen.

The printout will be directed to the default printer for the workstation unless a custom printer has been selected.

2.9 EXIT This menu item will exit the dose projection program when it is selected.

3.

INTERPRETATION OF THE DOSE PROJECTION

SUMMARY

TABLE This summary table appears on the screen after the dose projection calculation has been completed.

3.1 The first column at the top of this table is the distance from the plant.

These distances default to Site Boundary, 2 Miles, 5 Miles, and 10 Miles.

The distance along the centerline of the plume is identified in the Max row.

3.2 The second column at the top of this table is the TEDE in mrem. This will give the value that is entered on the Notification Form in the appropriate location.

3.3 The third column at the top of this table is the Thyroid CDE in mrem. This will give the value that is entered on the Notification Form in the appropriate location.

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3.4 The next three columns give the Effective Dose Equivalent due to Immersion in the plume, the Committed Effective Dose Equivalent due to inhalation, and the Effective Dose Equivalent due to ground deposition.

These columns are provided for information only.

3.5 The final column at the top of the table gives the X/Q value in s/m 3. This value should be provided to the State and Federal Emergency Response Officials when requested. For stability classes E, F, and G in MIXED MODE RELEASES, the X/Q is extremely small at the site boundary when compared with the other X/Q values.

3.6 The Dose Projection Summary Table also contains the Dose Projection Meteorology Data.

3.7 The reactor shutdown time is also found on this table.

3.8 The Projection Time which the Notification Form refers to is listed on the table as the Calculation Time.

NOTE:

The following two steps are very important for proper correlation between the dose projections performed by HBR's Dose Projection Team and the projections performed by State personnel. The Xe-1 33 Equivalent Release and the 1-131 Equivalent Release values are used by South Carolina Department of Health and Environmental Control for input into their dose assessment program.

3.9 The Xe-133 Equivalent Release is provided on this table and it is the value that should be entered as the Noble Gas Activity on the Notification Form. If any results are questionable, then Attachment 8.3.5.17 should be used to calculate this value.

3.10 The 1-131 Equivalent Release is provided on this table and it is the value that should be entered as the Iodine Activity on the Notification Form. If any results are questionable, then Attachment 8.3.5.17 should be used to calculate this value.

3.11 The Dosimeter Correction Factor that is provided on this table should only be used when the Radiological Control Manager has directed that a Dosimeter Correction Factor is necessary, and there is not adequate data to calculate one using environmental data.

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NOTE:

The Emergency Action Level provided by the dose projection program is for INFORMATION ONLY. All Emergency Classifications shall be made by using the EAL procedures.

3.12 Emergency Action Levels If an Emergency Action Level due to a dose projection is exceeded, the output screen will indicate the appropriate classification. Evaluate the EAL Flow Charts and compare the dose calculation against the EAL's to determine the Emergency Classification. Notify the RCM of all Emergency Action Levels that the progqram recommends.

3.13 HBRDOSE/RASCAL.3.6.16, HBRDOSE/RASCAL Comparison Matrix can be used to discuss differences in plant dose projections and those performed using the NRC's RASCAL program.

8.3.4 RECORDS N/A 8.3.5 ATTACHMENTS 8.3.5.1 8.3.5.2 8.3.5.3 8.3.5.4 8.3.5.5 8.3.5.6 Definitions and Abbreviations General Information Quality Codes Core Uncovery Time Determination Accident Mitigation Systems Obtaining and Updating Meteorological Data I EPRAD-03 Rev. 12 1

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8.3.5.7 8.3.5.8 8.3.5.9 8.3.5.10 8.3.5.11 8.3.5.12 8.3.5.13 8.3.5.14 8.3.5.15 8.3.6.16 I EPRAD-03 I

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Page 44 of 80 Source Term Determination Flow Rates Detector Sensitivities Measuring Radiation Level on Main Steam Lines Typical RMS Values RMS Monitored Systems Weather Service Data Onsite Meteorological Data Meteorological Forecast Form HBRDOSE/RASCAL Comparison Matrix

ATTACHMENT 8.3.5.1 Page 1 of 3 DEFINITIONS/ABBREVIATIONS Definitions:

Atmosphere Dispersion Factor (X/Q. - the fraction of activity released that will reach the point of interest (sec/m3).

Committed Dose Equivalent-The dose equivalent to organs or tissue of reference that will be received from an intake of radioactive material by an individual during the 50 year period following the intake.

Committed Effective Dose Equivalent -The sum of the products of the weighting factors applicable to each of the body organs or tissues that are irradiated and the committed dose equivalent to these organs or tissues.

Core Uncovery Time - The time that inadequate core cooing occurs until the time that adequate core cooling is restored. (See Attachment 8.3.5.4).

Early Phase - The period at the beginning of a nuclear incident when immediate decisions for effective use of protective actions are required, and must be based primarily on predictions of radiological conditions in the environment. This phase may last from hours to days. For the purpose of dose projection, it is assumed to last for four days.

Effective Dose Equivalent - The sum of the products of the dose equivalent to each "organ and a weighting factor, where the weighting factor is the ratio of the risk of mortality from delayed health effects arising from irradiation of a particular organ or tissue to the total risk of mortality from delayed health effects when the whole body is irradiated uniformly to the same dose. This unit is considered equivalent to be the Deep Dose Equivalent for the purposes of dose projections because the external exposures are considered to be uniform across the whole body.

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ATTACHMENT 8.3.5.1 Page 2 of 3 DEFINITIONS/ABBREVIATIONS Intermediate Phase - The period beginning after the incident source and releases have been brought under control and reliable environmental measurements are available for use as a basis for decisions on additional protective actions and extending until these protective actions are terminated. This phase may overlap the early and late phases and may last from weeks to many months. For the purpose of dose projection, it is assumed to last for one year.

Late Phase - The period beginning when recovery action designed to reduce radiation levels in the environment to permanently acceptable levels are commenced, and ending when all recovery actions have been completed. This period may extend from months to years (also referred to as the recovery phase).

Release Duration - The period of time from the beginning of the release until the end of the release or the projected end of the release. This can be determined by estimating the completion of a damage control mission, performance of a repair to stop the release, or the estimated time until the RCS, CV Sump, or Steam Generator temperature is below 2000 F.

Release Rate (Q) - The term in the dose projection which describes the amount of activity that is being released. This is recorded in Curies per second. The total curies released may be calculated from the release rate (Q) and the release duration in seconds.

Total Effective Dose Equivalent - The sum of the deep-dose equivalent (for external exposures) and the committed effective dose equivalent(for internal exposures).

Weatherinqi/Weathering Factor - The fraction of radioactivity remaining after being affected by average weather conditions for a specified period of time.

Abbreviations:

1.

ALARA - As Low As is Reasonably Achievable

2.

BSEP - Brunswick Steam Electric Plant

3.

CDE - Committed Dose Equivalent

4.

CFM - Cubic Feet per Minute

5.

CPM - Counts Per Minute

6.

CV - Containment Vessel

7.

EAL - Emergency Action Level

8.

EMT - Environmental Monitoring Team

9.

EOF - Emergency Operations Facility

10.

ERFIS - Emergency Response Facility Information System

11.

ERO - Emergency Response Organization

12.

GPM - Gallons Per Minute

13.

HNP - Harris Nuclear Project

14.

LAN - Local Area Network

15.

LOCA - Loss Of Coolant Accident EPRAD-03 Rev. 12 Page 46 of 80

ATTACHMENT 8.3.5.1 Page 3 of 3 DEFINITIONS/ABBREVIATIONS PORV - Power Operated Relief Valve RCS - Reactor Coolant System RMS - Radiation Monitoring System SDS - Satellite Display System SRO - Senior Reactor Operator SRV - Safety Relief Valve STA - Shift Technical Advisor TEDE - Total Effective Dose Equivalent 1EPRAD-03 Rev. 12 Page 47 of 80 1

16.

17.

18.

19.

20.

21.

22.

23.

ATTACHMENT 8.3.5.2 Page 1 of 1 GENERAL INFORMATION Backup Capability:

If ERFIS or a computer with the dose projection program are not available, contact computer support personnel and request that they provide a computer with the current revision of the dose projection software installed on it.

R-14 C, D, and E operate as follows:

R-14C is the Normal range Noble Gas monitor.

R-14D is the Intermediate range Noble Gas monitor.

R-14E is the High range Noble Gas monitor.

R-14D and R-14E normally read between 10 and 11 CPM.

R-14C when increasing will reach its predetermined alarm setpoint. Further increase will cause R-14C to reach its predetermined swap-over setpoint. When the swap-over setpoint is reached, R-14C will fail to 1 Meg (1M) which also will cause R-14D and R-14E to activate and start providing intermediate and high range noble gas readings.

If R-14C is reading 1 Meg, this SHOULD NOT be used as a valid reading and RMS data SHOULD BE obtained from R-14D and R-14E.

Special attention should be paid to the quality code of the data on the program. Quality code color schemes are given in Attachment 8.3.5.3.

In order to select an item when performing the dose projection press the space bar.

Pressing the space bar will also deselect the item if it had already been selected.

The help menu may be accessed at any time while using the dose projection program.

This can be accomplished by pressing the Flfunction key. The function can be exited by clicking the mouse on the EXIT field or by pressing the F1 key.

Messages are displayed at the bottom of each screen to describe the function that the cursor is on.

Attachments 8.3.5.14, Onsite Meteorological Data, and 8.3.5.15, Meteorological Forecast Form, can be used to record weather conditions and forecast.

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ATTACHMENT 8.3.5.3 Page 1 of 1 QUALITY CODES Color of Data Meaning T

Action Red Stars Computer Entered Bad Data Do Not Use This Data Green Computer Entered Good This "Data May Be Used Data Normal Level Yellow Computer Entered Good This Data May Be Used Data Alert Level Red Computer Entered Good This Data May Be Used Data Alarm Level White Manually Entered Data This Data May Be Used I EPRAD-03 Rev. 12 Page 49 of 80 1

ATTACHMENT 8.3.5.4 Page 1 of 1 CORE UNCOVERY TIME DETERMINATION NOTE:

The time determination below is based upon the core level and temperature that the core cooling is insufficient to prevent the cladding from overheating and failing. This basis is conservative for all fuel damage scenarios which result from core uncovery. This time can be determined by consulting a SRO or RO with access to plant data.

Core uncovery time is defined for dose projection purposes to be the point in time that inadequate core cooling occurs until the time that adequate core cooling is restored. For the purposes of dose projection core uncovery time will be the time that a RED status occurs on the CORE COOLING critical safety function status tree until the tree conditions return to YELLOW status (core covered and core exit thermocouples < 7000 F).

There are other possible accidents that may result in fuel damage. These events could be initiated by core flow blockage from debris or by localized melting from a rod ejection accident, pump failures, etc. as analyzed by the UFSAR. In this case, judgment may be applied using the bases information for CORE UNCOVERY TIME DETERMINATION above to most closely describe the fuel damage situation. In general choice of "uncovery < 30 min", corresponding to a release of 100% of the gap activity will conservatively account for most mechanical and miscellaneous fuel damage situations.

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ATTACHMENT 8.3.5.5 Page 1 of 1 ACCIDENT MITIGATION SYSTEMS The capability to take credit for accident release mitigation systems is built into the dose projection program. Credit is given one of three ways, charcoal filtration, containment sprays, and water partitioning in the steam generator.

NOTE: It is important to note that if the release is mitigated by ANY of the following: Charcoal Filtration, CV Sprays, or Partitioning, then assume mitigating effects are EFFECTIVE unless information is known to be otherwise.

Only if the release pathway is direct to the environment without mitigation, should NOT EFFECTIVE be selected.

Filtration Various fans can be aligned to cleanup effluent from leaking systems. When the release is passing through any one of the following fans, filtration can be considered effective. The general area(s) where the fan draws a suction is listed in parenthesis.

HVE-1A or HVE-1B (Containment Purge)

HVE-3 or HVE-4 (Containment Air, In pre-purge mode)

HVE-5A or HVE-5B (Auxiliary Building Exhaust)

HVE-15A (Spent Fuel Pit during refueling)

The CV Spray System The CV Spray System is designed to remove radioiodine from containment in the event a radioactive release (typically a LOCA) occurs inside containment. If such a release occurs and the CV Spray System (with NaOH added) is operating, then the CV Sprays are considered effective.

Water Partitioning Occurs during a release through the steam generators (e.g., a tube leak or tube rupture) and level in the affected steam generator is above the top of the tubes.

Partitioning is effective for removing iodines and some particulates when the steam generator level is greater than 10% on the Narrow Range Steam Generator Level Indicator.

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ATTACHMENT 8.3.5.6 Page 1 of 5 OBTAINING AND UPDATING METEOROLOGICAL DATA In the manual data entry mode, meteorological data may not be available from ERFIS.

Determine wind direction, wind speed, and atmospheric stability class using one of six methods listed in preferred order of use.

NOTE:

Meteorological data will normally display a green value and an "OK" quality code. If the values are displayed in white, or the quality code is "BAD",

DO NOT USE THIS DATA.

1.

If operable, use the data from control room readouts to obtain the atmospheric stability class, wind speed, and wind direction.

OR

2.

Call the CP&L offsite meteorological contact (See ERO Phone Book for number).

NOTE:

If The Florence Airport or the National Weather Service office is called, the only information that can be obtained is the wind direction, wind speed, and ambient temperature. Stability factor must be obtained from Step 5 of this Attachment.

If wind speed and direction are only supplied for one point enter these values in both the elevated and ground fields. Do not enter wind gust as the wind speed, and if no Delta T is supplied do not enter one.

3.

Call the Florence Airport for Weather Information (See ERO Phone Book for number).

OR EPRAD-03 Rev. 12 Page 52 of 80

ATTACHMENT 8.3.5.6 Page 2 of 5 OBTAINING AND UPDATING METEOROLOGICAL DATA

4.

Call the National Weather Service office in Columbia, South Carolina, for daily weather information or in Wilmington, North Carolina, for severe weather information. Use Attachment 8.3.5.13 to document this: (See ERO Phone Book for number)

OR

5.

If there is no stability class data readily available, a general estimate of the current Atmospheric Stability Class can be made by visual observation, using the following table:

light wind or calm

(< 11.5 mph) moderately strong wind(> 11.5 mph)

Rain, Day or NiDht D

D Sunny Day B

C Cloudy Day C

D Cloudy Night E

D Clear Night F

D OR I EPRAD-03 I

Rev. 12 1

Page 53 of 801

ATTACHMENT 8.3.5.6 Page 3 of 5 OBTAINING AND UPDATING METEOROLOGICAL DATA

6.

A manual method may be used to acquire data from the meteorological tower.

The following method may be used to manually obtain this data:

A.

Obtain the Meteorological Tower Building key from E&RC or Plant Security.

B.

Locate the Met Tower recorder inside the building.

C.

Locate the Upper Display key on the recorder.

D.

Depress the Upper Display key one or more times until the Upper Display is placed in manual control. "MAN" will be displayed in the Lower Display of the recorder.

E.

Locate the Channel Up (CH Up) and Channel Down (CH Down) keys on the recorder.

F.

Using the Channel Up (CH Up) and Channel Down (CH Down) keys, scroll through the recorder channels to obtain the necessary information required on the "Manual Meteorological Collection Data Sheet" included in this attachment.

G.

Using the Differential Temperature values obtained from the recorder, determine the Stability Class as per the table included in this attachment.

I EPRAD-03 Rev. 12 1

Page 54 of 80 1

ATTACHMENT 8.3.5.6 Page 4 of 5 OBTAINING AND UPDATING METEOROLOGICAL DATA EXAMPLE OF RECORDER CHANNEL SELECTIONS LT1 - Lower Temperature #1 ( ambient temperature)

DT1 - Differential Temperature #1 DT2 - Differential Temperature #2 LWS - Lower Wind Speed LWD - Lower Wind Direction UWS - Upper Wind Speed UWD - Upper Wind Direction DPT - Dew Point MANUAL METEOROLOGICAL COLLECTION DATA SHEET WIND SPEED UPPER WIND SPEED MPH LOWER WIND SPEED MPH WIND DIRECTION UPPER WIND DIRECTION DEGREES LOWER WIND DIRECTION DEGREES AMBIENT TEMPERATURE TEMPERATURE DEGREES F DIFFERENTIAL TEMPERATURE DT1 =

DT2 =

C/1 00M C/1 00M STABILITY CLASS DT1 + DT2=

2 IEPRAD-03 I

Rev. 12 Page 55 of 801 CH 01 CH 02 CH 03 CH 04 CH 05 CH 06 CH 07 CH 08 C/1 00M

ATTACHMENT 8.3.5.6 Page 5 of 5 OBTAINING AND UPDATING METEOROLOGICAL DATA STABILITY CLASS DIFFERENTIAL TEMP. C/1OOM (circle one)

A

<-1.9 B

-1.9 TO -1.7 C

-1.7 TO -1.5 D

-1.5 TO -0.5 E

-0.5 TO +1.5 F

+1.5 TO +4.0 G

I EPRAD-03 I

Rev. 12 1

Page 56 of 80

ATTACHMENT 8.3.5.7 Page 1 of 4 SOURCE TERM DETERMINATION Part A - Determination of Curies in Containment Atmosphere This calculation can be performed by obtaining the activity in the containment from the RCD or E&RC lead technician, or by calculating it using the radiation monitor data and their sensitivities. When calculations are performed utilizing radiation levels obtained from R-2, consideration should be given to background correcting the radiation level.

1)

Containment atmospheric activity:

As determined by sampling CV atmosphere:

IJCi/cc Calculation of atmospheric activity from a CV radiation monitor:

Sensitivity CV Activity R-12

___cpm R-2 mrem/hr R-32A/B rem/hr

/ ___cpm/(liCvcc)

=

_iCi/cc

/ _

(mrem/hr)/([lCi/cc) = _

_CVcc

/

_(rem/hr)/(i[CVcc) = _

IiCi/cc

2)

Equation for determining curies in containment:

Curies in containment1

= (CV activity ((lCi/cc]) (5.7 x 10 4 [Ci-cc/pCi])

= (

[liCVcc]) (5.7 x 104 [Ci-cc/[lCil)

=

Ci

• This value can be determined by referencing Attachment 8.3.5.9, Table 2. Ensure that the sensitivity that corresponds to the correct accident and shutdown time are used.

1 Containment volume as calculated per RNP-C/CONT-1 002, Determination of Containment Heat Sink, is 2.013 x 106 cubic feet (5.7 x 1010 cc). The value of 5.7 x 104 is used to account for 1iCi to Ci conversion.

I EPRAD-03 I

Rev. 12 Page 57 of 80 1 Monitor Reading

ATTACHMENT 8.3.5.7 Page 2 of 4 SOURCE TERM DETERMINATION Part B - Determination of Curies in the Reactor Coolant System (RCS)

Obtain the RCS activity from the RCD or the E&RC lead technician to perform this calculation.

1)

RCS activity:

__Ci/ml

2)

Equation for determining curies in the RCS:

Curies in RCS

= (RCS activity [l.Ci/ml]) (2.65 x 102 Ci-ml/iLCi)

= (.

[lCi/ml]) (2.65 x 102 Ci-ml/l.Ci)

=

Ci Part C - Determination of Sump Source Term Obtain the sump activity and the sump volume from the RCD or the E&RC lead technician in order to perform this calculation.

1)

Quantity of liquid in sump gal

2)

Sump activity ltCi/cc

3)

Equation for determining curies in sump:

Curies in the sump =

(volume of liquid in sump [gal]) (activity of sump ((lCi/ml]) (3.79 x 10,3)

=

[gal]) (

[.0Ci/ml]) (3.79 x 10-3 Ci-mVl/Ci-gal)

=_

Ci Part D - Determination of Primary to Secondary Leakage Source Term Obtain the primary to secondary leak rate and RCS activity in order to perform this calculation.

1)

Primary to Secondary Leakage gal/min

2)

Source Term (Ci)

= (Leakrate gal/min)(6.3E-5)(RCS Activity i.Ci/cc)

= (

gpm)(6.3E-5)(

I.Ci/cc)

= (

CVsec)(

hrs)(3600 sec/hr)

Ci I EPRAD-03 I

Rev. 12 I

Page 58 of 80 1

ATTACHMENT 8.3.5.7 Page 3 of 4 SOURCE TERM DETERMINATION Part E - Determination of Source Term Released Due To Secondary Leakage Determine the leakrate from the PORV or SRVs using Attachment 8.3.5.8. If the leak is due to a faulted Main Steam System obtain an estimate of the leakrate can be obtained from the Accident Assessment Team.

1)

Secondary Leakrate cc/sec

2)

Source Term (Ci) =

mrem/hr x hr x cc/sec x 3.6E-03 Ci-sec/lpCi-hr R-31 Rad Level Duration Leakrate (mrem/hr)/([iCi/cc)

R-31 Sensitivity Ci

• This value can be determined by referencing Attachment 8.3.5.9, Table 2. Ensure that the sensitivity that corresponds to the correct accident and shutdown time are used.

Part F - Determination of Source Term Released Through Main Steam Using Direct Survey Request that the RCD dispatch a member of the plant monitoring team with an extendable probe survey instrument to a location one level below the Main Steam lines as indicated by Attachment 8.3.5.10. The probe should be extended to a position adjacent to the low point of each steam line (or as directed by the Dose Projection Teamleader or RCD) to determine the contact dose rate. The status (open/closed) of the PORV and SRVs on the monitored lines should also be noted.

1)

Contact radiation level on Steam Line:

mrem/hr

2)

Flow Rate:

cc*/sec (Attachment 8.3.5.8 or Accident Assessment Team)

3)

Detector Sensitivity from Attachment 8.3.5.9, Table 1:

(mrem/hr)/(iCi/cc)

4)

Source Term (Ci) =

mrem/hr x hr x cc /sec x 3.6E-03 Ci-sec/[tCi-hr Rad Level Duration Flow Rate (mrem/hr)/(. Ci/cc)

Detector Sensitivity Ci

• Substitute ml for cc when calculations are performed for water filled main steam lines.

IEPRAD-03 I

Rev. 12 Page 59 of 801

ATTACHMENT 8.3.5.7 Page 4 of 4 SOURCE TERM DETERMINATION Part G - Determination of Source Term Released Plant Vent Stack Request that the RCD dispatch a member of the plant monitoring team with an extendable probe survey instrument to obtain a contact radiation level on the side of the plant stack (an instrument with a remote probe can also be tused). The measurement should be made inside the shielded orifice which is approximately 4 feet above the Auxiliary Building roof on the south side of the stack.

1)

Contact radiation level on Plant Vent Stack:

mrem/hr

2)

Release rate = (Use Attachment 8.3.5.8)

= (Stack Flow Rate [cfm]) (28320 [cc/ft3]) (60 [min/hr])

= (_cfm)

(28320 cc/ft3) (60 min/hr)

=_

cc/hr

3)

Detector Sensitivity from Attachment 8.3.5.9, Table 1:

(mrem/hr)/([,Ci/cc)

4)

Source Term (Ci) =

mrem/hr x hr x cc/hr x 1E-06 Ci/lICi Rad Level Duration Release Rate S

v (mrem/hr)/([lCi/cc)

"Sensitivity =

Ci I EPRAD-03 Rev. 12 1

Page 60 of 80

ATTACHMENT 8.3.5.8 Page 1 of 5 FLOW RATES R-11, R-12, R-14 HVE-2A/B....................................................................

4.4 x 104 cfm HVE-2A/B and HVE-15/15A.........................................

5.5 x 104 cfm HVE-2A/B and HVE-1A/B............................................

6.2 x 104 cfm HVE-2A/B and HVE-1A/B & HVE-15/15A................... 7.2 x 104 cfm R-15, Air Eiector - Noble Gas Flow Rate = 3.10 x 102 cfm (for one vacuum pump running)

Flow Rate = 6.10 x 102 cfm (for two vacuum pumps running)

R-20, R-30, Fuel Building Basement Exhaust - Low and High Range Noble Gas Flow Rate = 1.0 x 104 cfm R-21, Fuel Building UPPER Level Exhaust Flow Rate = 1.34 x 104 cfm R-31 A, R-31 B, R-31 C - Steam-Line Monitors (at 800 psi)

PORV (100% lift)....................................

1.92E06 cc/sec (4.57E05 Ibm/hr)

PORV and 1 SRV.................................

4.00E06 cc/sec (9.51 E05 Ibm/hr)

PORV and 2 SRV..................................

6.11 E06 cc/sec (1.45E06 Ibm/hr)

PORV and 3 SRV.................................

9.19E06 cc/sec (2.19E06 Ibm/hr)

I EPRAD-03 I

Rev. 12 Page 61 of 80 1

ATTACHMENT 8.3.5.8 Page 2 of 5 FLOW RATES R-31A, R-31B, R-31C - Steam-Line Monitors (Filled with Water)

PORV (100% lift)...............................................

7.32E04 ml/sec PORV AND 1 SRV...........................................

1.56EO5 ml/sec" PORV AND 2 SRV...........................................

2.42EO5 ml/sec PORV AND 3 SRV...........................................

3.68EO5 ml/sec R-2, R-32A, R-32B - Containment Radiation Monitors Containment isolated with no discharge via plant vent.......... 1.5 CFM1 Containment vented via plant vent.........................................

2500 CFM 1 Design basis leakage for containment at 0.1% containment volume per day.

IEPRAD-03 Rev. 12 1

Page 62 of 801

ATTACHMENT 8.3.5.8 Page 3 of 5 FLOW RATES STEAM LINE FLOW RATE CALCULATION FOR A DRY STEAM GENERATOR 1.0

1.

RCS Leak Rate (RCSLR) gpm'

2.

RCS Temperature OF

3.

RCS Pressure psig

4.

S/G Pressure psig

5.

S/G Temp OF 2.0

1.

From the Steam Tables determine the specific volume of RCS Fluid (RCSsv) at conditions in 1.0 ft3/Ib

2.

From the Steam Tables determine the specific volume of S/G Fluid (SG sv) at condition in 1.0 ft3/lb 3.0 Determine the RCS Mass Release Rate (RCSMRR) into S/G by using the following formula:

RCSLR (gal/my)

= lb/min (7.48 gaVft3) (RCSsv ft3/lb)

L..RCSLR gal/min )

Slb/min (7.48 gaVft3) (

RCSsv ft3/lb) 4.0 Determine the steam flow rate using the following formula:

RCSMRR (lb/min) SGsv (ft3/ib) (472 cc/sec) = cc/sec ft3/min

)(

(472cc/sec)

RCSMRR (lb/min) SGsv (ft /Ib) ft/min cc/sec 5.0 Performed by:

Verified by:

I EPRAD-03 I

Rev. 12 Page 63 of 80 1

ATTACHMENT 8.3.5.8 Page 4 of 5 FLOW RATES CONVERSION OF STEAM MASS FLOW RATE TO VOLUMETRIC FLOW RATE (

1.

Obtain and record the steam mass flow rate in lbs/hr from the Accident Assessment Team.

lbs/hr [1]

2.

Obtain and record the main steam pressure in psig.

I psig

3.

Use the figure on the following page to determine the specific volume (cc/lb) for the pressure determined in step 2.

cc/lb [2]

4.

Determine the volumetric flow rate using the following formula:

(lbs/hr) (1 hr/3600 sec) (cc/lb) = cc/sec

(...

Ib/hr) (1 hr/3600 sec) (._

cc/lb) = _

cc/sec

[1]

[2]

Performed by:

/

Date Time Verified by:

/

Date Time (1) For use with R-31 readings under any conditions.

I EPRAD-03 Rev. 12 Page 64 of 80 1

ATTACL....CNT 8.3.5.8 Page 5 of 5 FLOW RATES STEAM PRESSURE VS SPECIFIC VOLUME PSIG VS CC PER POUND PSIG 1200 1100 1000 900 800 700 600 500 400 300 200 100 0

150 175 200 225 250 SPECIFIC VOLUME - CC/LB (THOUSANDS)

IEPRAD-03 I

Rev. 12 1

Page 65 of 80 1 C

CC, w

i-0 25 50 75 100 125

ATTACHMENT 8.3.5.9 Page 1 of 5 DETECTOR SENSITIVITIES Determine the appropriate accident scenario classification (1-10) utilizing the following table.

ACCIDENT PLANT CONDITIONS FILTRATION/PARTITIONING/SPRAYS SCENARIO 1

Core not uncovered Effective or Not Effective 2

Core uncovered <30 minutes Effective 3

Core uncovered <30 minutes Not Effective 4

Core uncovered 0.5 - 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Effective 5

Core uncovered 0.5 - 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Not Effective 6

Core uncovered > 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Effective 7

Core uncovered > 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Not Effective 8

New Spent Fuel Effective or Not Effective 9

Old Spent Fuel Effective or Not Effective 10 Waste Gas Decay Tank Effective or Not Effective Determine the sensitivity of the appropriate detector using the following tables and accident scenarios. (The sensitivities in Table 1 below are based on nuclide mixes at reactor shutdown). Table 2 provides detector efficiencies (sensitivities) for eight (8) designated accident categories and time steps for each accident sequence.

Table 1 ACCIDENT PLANT VENT STEAM FILLEDMAIN STEAM WATER FILLED MAIN SCENARIO STACK(mrem/hr)/([.Ci/cc)

LINE(mrem/hr)/(tiCi/cc)

STEAM LIN E(mrem/hr)/([lCi/ml) 1 3.28E+02 1.49E+01 4.51 E+00 2

1.42E+03 7.11E+01 2.15E+01 3

2.63E+03 9.66E+01 2.35E+01 4

1.40E+03 7.08E+01 2.16E+01 5

1.98E+03 7.94E+01 2.09E+01 6

1.40E+03 7.08E+01 2.15E+01 7

2.20E+03 8.24E+01 2.06E+01 8

2.80E+03 N/A N/A 9

5.44E+00 N/A N/A 10 2.48E-01 N/A N/A EPRAD-03 Rev. 12 Page 66 of 80]

C ATTAC(...,-tNT 8.3.5.9 Page 2 of 5 DETECTOR SENSITIVITIES Table 2: Summary of Detector Sensitivities(Efficiencies) for Designated Accident Scenarios C

Accident R-14C R-14D, R-14E,

o R-12

, R-20 R-21e, R-15 R-02 R-30 R-31A,B,C R-32A R-32B Time step cpm per-cpm per',,,i :cpm per-

-cpmper, cpm per,

,cpmp

cpmper, mR/hrper, mR/hrper mR/hrper R/hrppe R/hr per "u"Cc C

'cc uc c

. u" uCVcc-'*.

ub uCVcc UCi/cc uCVccr uCVcc Accident 1 Normal RCS t=0 1.50E+07 3.94E+03 2.58E+00 3.03E+07 3.17E+07 3.25E+07 4.42E+05 9.2E+03 6.1E+01 1.1E+01 9.2E+00 9.2E+00 t=0.5 1.41 E+07 3.71 E+03 2.43E+00 2.76E+07 2.90E+07 2.97E+07 3.54E+05 6.9E+03 3.6E+01 6.3E+00 6.9E+00 6.9E+00 t=1 1.38E+07 3.63E+03 2.38E+00 2.68E+07 2.81 E+07 2.87E+07 3.23E+05 6.2E+03 2.9E+01 4.9E+00 6.2E+00 6.2E+00 t=2 1.36E+07 3.57E+03 2.34E+00 2.61 E+07 2.74E+07 2.80E+07 2.97E+05 5.4E+03 2.OE+01 3.4E+00 5.4E+00 5.4E+00 t=4 1.33E+07 3.50E+03 2.29E+00 2.54E+07 2.66E+07 2.73E+07 2.64E+05 5.OE+03 1.5E+01 2.5E+00 5.OE+00 5.OE+00 t=8 1.31 E+07 3.43E+03 2.25E+00 2.47E+07 2.59E+07 2.65E+07 2.20E+05 4.3E+03 8.4E+00 1.3E+00 4.3E+00 4.3E+00 t=16 1.29E+07 3.38E+03 2.21E+00 2.41E+07 2.53E+07 2.59E+07 1.73E+05 4.OE+03 4.5E+00 6.6E-01 4.OE+00 4.OE+00 t=32 1.29E+07 3.40E+03 2.23E+00 2.41 E+07 2.53E+07 2.59E+07 1.36E+05 3.8E+03 3.3E+00 5.OE-01 3.8E+00 3.8E+00 Accident 2 Core uncovered < 30 minutes/ mechanical damage with sprays/filtration effective t=O 3.26E+07 8.59E+03 5.62E+00 6.42E+07 6.74E+07 6.90E+07 3.41 E+06 4.2E+04 4.0E+02 7.2E+01 4.2E+01 4.2E+01 t=0.5 3.10E+07 8.16E+03 5.34E+00 5.33E+07 5.60E+07 5.73E+07 3.70E+06 3.3E+04 3.OE+02 5.4E+01 3.3E+01 3.3E+01 t=W 3.04E+07 8.OOE+03 5.24E+00 4.84E+07 5.08E+07 5.20E+07 3.87E+06 2.8E+04 2.5E+02 4.4E+01 2.8E+01 2.8E+01 t=2 3.01E+07 7.91E+03 5.18E+00 4.44E+07 4.66E+07 4.77E+07 4.07E+06 2.4E+04 2.OE+02 3.6E+01 2.4E+01 2.4E+01 t=4 2.98E+07 7.85E+03 5.14E+00 4.02E+07 4.22E+07 4.32E+07 4.33E+06 1.9E+04 1.4E+02 2.5E+01 1.9E+01 1.9E+01 t=8 2.96E+07 7.78E+03 5.10E+00 3.53E+07 3.70E+07 3.79E+07 4.63E+06 1.3E+04 8.OE+01 1.3E+01 1.3E+01 1.3E+01 t=16 2.94E+07 7.73E+03 5.06E+00 3.05E+07 3.20E+07 3.27E+07 4.90E+06 9.OE+03 3.6E+01 4.7E+00 9.OE+00 9.OE+00 t=32 2.92E+07 7.69E+03 5.04E+00 2.67E+07 2.80E+07 2.87E+07 5.11 E+06 7.7E+03 2.OE+01 1.8E+00 7.7E+00 7.7E+00 Detector sensitivities derived from Calculation No. RNP-M/MECH 1746.

EPRAD-03 Rev. 12 Page 67 of 801

Q IEPRAD-03 Rev. 12 Page 68 of 801 ATTACo...I'T 8.3.5.9 Page 3 of 5 DETECTOR SENSITIVITIES Table 2: Summary of Detector Sensitivities(Efficiencies) for Designated Accident Scenarios

(

Accident R-14C R-14D R-14E

,R-12 R-20 R21 R02 -

R30 R31A,B,C R32A R32B Time step cpm per cpm per, cpm per,-

cpm per

-. cpm'per' cpm per; cpm pe

'mR/hr per mR/hr per mR/hr per R/hr pe r per uCV cc

?

-uccc 0c cc uOVcc 1:

IuCVcc'ý

t. ° uCi cc.,,

uCVcc' uCl/cc uCVcc uCccr uCVcc Accident 3 Core uncovered < 30 minutes/ mechanical damage with sprays/filtration NOT effective t=0 1.55E+07 4.08E+03 2.67E+00 7.03E+07 7.38E+07 7.55E+07 1.62E+06 7.3E+04 7.4E+02 1.3E+02 7.3E+01 7.3E+01 t=0.5 1.34E+07 3.53E+03 2.31 E+00 6.60E+07 6.93E+07 7.09E+07 1.60E+06 6.7E+04 6.8E+02 1.2E+02 6.7E+01 6.7E+01 t=1 1.30E+07 3.42E+03 2.24E+00 6.39E+07 6.71E+07 6.86E+07 1.65E+06 6.2E+04 6.2E+02 1.1E+02 6.2E+01 6.2E+01 t=2 1.32E+07 3.47E+03 2.27E+00 6.15E+07 6.45E+07 6.60E+07 1.79E+06 5.4E+04 5.3E+02 9.1E+01 5.4E+01 5.4E+01 t=4 1.37E+07 3.59E+03 2.35E+00 5.82E+07 6.11 E+07 6.25E+07 1.98E+06 4.4E+04 4.2E+02 7.2E+01 4.4E+01 4.4E+01 t=8 1.42E+07 3.72E+03 2.44E+00 5.38E+07 5.65E+07 5.78E+07 2.21 E+06 3.4E+04 3.2E+02 5.4E+01 3.4E+01 3.4E+01 t=16 1.50E+07 3.95E+03 2.59E+00 4.81E+07 5.05E+07 5.17E+07 2.50E+06 2.5E+04 2.3E+02 3.8E+01 2.5E+01 2.5E+01 t=32 1.63E+07 4.28E+03 2.81E+00 4.1OE+07 4.30E+07 4.40E+07 2.84E+06 1.9E+04 1.6E+02 2.5E+01 1.9E+01 1.9E+01 Accident 4 Core uncovered 0.5 < 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> with sprays/filtration effective t=0 3.32E+07 8.74E+03 5.72E+00 6.40E+07 6.71 E+07 6.87E+07 3.47E+06 4.2E+04 4.OE+02 7.2E+01 4.2E+01 4.2E+01 t=0.5 3.17E+07 8.34E+03 5.46E+00 5.28E+07 5.54E+07 5.67E+07 3.78E+06 3.3E+04 3.OE+02 5.4E+01 3.3E+01 3.3E+01 t=W 3.11E+07 8.18E+03 5.35E+00 4.78E+07 5.01E+07 5.13E+07 3.96E+06 2.8E+04 2.5E+02 4.4E+01 2.8E+01 2.8E+01 t=2 3.07E+07 8.07E+03 5.29E+00 4.37E+07 4.59E+07 4.70E+07 4.16E+06 2.4E+04 2.OE+02 3.6E+01 2.4E+01 2.4E+01 t=4 3.04E+07 7.99E+03 5.23E+00 3.95E+07 4.14E+07 4.24E+07 4.41E+06 1.9E+04 1.4E+02 2.5E+01 1.9E+01 1.9E+01 t=8 3.01E+07 7.92E+03 5.18E+00 3.46E+07 3.63E+07 3.72E+07 4.71E+06 1.3E+04 8.OE+01 1.3E+01 1.3E+01 1.3E+01 t=16 2.98E+07 7.84E+03 5.14E+00 2.99E+07 3.13E+07 3.21E+07 4.97E+06 9.OE+03 3.6E+01 4.7E+00 9.OE+00 9.OE+00 t=32 2.96E+07 7.79E+03 5.1OE+00 2.63E+07 2.76E+07 2.82E+07 5.17E+06 7.7E+03 2.OE+01 1.8E+00 7.7E+00 7.7E+00 Detector sensitivities derived from Calculation No. RNP-M/MECH 1746, (EC 49849, Set-Point, Declaration Evaluation for EP)

C ATTACi....JNT 8.3.5.9 Page 4 of 5 DETECTOR SENSITIVITIES Table 2: Summary of Detector Sensitivities(Efficiencies) for Designated Accident Scenarios C

Accident R-14C -

D,1,,1;R R-21:< *,R-15 K-'

R-02,'R-30 R-31A,B.C R-32A R-32B TimeACdentstep cpm per; cpmRper

, cm Rperp nAcpm R2per pe, cpmper pm pe'.*

cpm ler, nR/hrper.

mR/hr per' mR/hr per R/hr pe R/hr per uCcc

.,~u~c

'u~c C*.C uc

u~/C' c

,,:,uCVcc I*,

uOVcc-,

-* uOVcc uCVccr uCVcc uC~ucvc

ý.,u ccu~c V

c, c uuvcc uccc

,Ccc

• Cc::

uCc,"l/

Ccc**,'*'...

Accident 5 Core uncovered 0.5 < 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> with sprays/filtration NOT effective t=O 2.26E+07 5.94E+03 3.89E+00 6.67E+07 7.00E+07 7.16E+07 2.36E+06 5.9E+04 5.9E+02 1.OE+02 5.9E+01 5.9E+01 t=0.5 2.03E+07 5.33E+03 3.49E+00 5.97E+07 6.26E+07 6.41 E+07 2.42E+06 5.3E+04 5.2E+02 9.0E+01 5.3E+01 5.3E+01 t=W 1.97E+07 5.18E+03 3.39E+00 5.64E+07 5.92E+07 6.06E+07 2.51E+06 4.8E+04 4.7E+02 8.1E+01 4.8E+01 4.8E+01 t=2 1.97E+07 5.19E+03 3.40E+00 5.32E+07 5.58E+07 5.71E+07 2.67E+06 4.1E+04 4.0E+02 6.8E+01 4.1E+01 4.1E+01 t=4 2.OOE+07 5.26E+03 3.44E+00 4.93E+07 5.18E+07 5.30E+07 2.90E+06 3.3E+04 3.OE+02 5.2E+01 3.3E+01 3.32+01 t=8 2.03E+07 5.34E+03 3.50E+00 4.45E+07 4.67E+07 4.78E+07 3.18E+06 2.4E+04 2.1E+02 3.5E+01 2.4E+01 2.4E+01 t=16 2.09E+07 5.49E+03 3.60E+00 3.90E+07 4.09E+07 4.19E+07 3.48E+06 1.8E+04 1.4E+02 2.2E+01 1.8E+01 1.8E+01 t=32 2.18E+07 5.72E+03 3.75E+00 3.32E+07 3.49E+07 3.57E+07 3.80E+06 1.3E+04 8.9E+01 1.4E+01 1.3E+01 1.3E+01 Accident 6 Core uncovered > 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> with sprays/filtration effective t=0 3.29E+07 8.64E+03 5.66E+00 6.38E+07 6.70E+07 6.85E+07 3.43E+06 4.2E+04 4.0E+02 7.2E+01 4.2E+01 4.2E+01 t=0.5 3.13E+07 8.22E+03 5.38E+00 5.28E+07 5.54E+07 5.66E+07 3.73E+06 3.3E+04 3.02+02 5.4E+01 3.3E+01 3.3E+01 t=W 3.06E+07 8.06E+03 5.27E+00 4.78E+07 5.01E+07 5.13E+07 3.90E+06 2.82+04 2.5E+02 4.4E+01 2.8E+01 2.8E+01 t=2 3.02E+07 7.95E+03 5.21E+00 4.37E+07 4.59E+07 4.70E+07 4.10E+06 2.4E+04 2.0E+02 3.6E+01 2.4E+01 2.4E+01 t=4 2.99E+07 7.88E+03 5.16E+00 3.95E+07 4.15E+07 4.24E+07 4.35E+06 1.9E+04 1.4E+02 2.5E+01 1.9E+01 1.9E+01 t=8 2.97E+07 7.80E+03 5.11E+00 3.46E+07 3.64E+07 3.72E+07 4.64E+06 1.3E+04 8.OE+01 1.3E+01 1.3E+01 1.3E+01 t=16 2.94E+07 7.732+03 5.06E+00 2.99E+07 3.14E+07 3.21E+07 4.90E+06 9.OE+03 3.6E+01 4.7E+00 9.02+00 9.0E+00 t=32 2.92E+07 7.68E+03 5.03E+00 2.62E+07 2.75E+07 2.82E+07 5.1 OE+06 7.7E+03 2.OE+01 1.8E+00 7.7E+00 7.7E+00 Detector sensitivities derived from Calculation No. RNP-M/MECH 1746, (EC 49849, Set-Point, Declaration Evaluation for EP).

.4' I EPRAD-03 Rev. 12 Page 69 of 80i

C ATTACL,.._njT 8.3.5.9 Page 5 of 5 DETECTOR SENSITIVITIES Table 2: Summary of Detector Sensitivities(Efficiencies) for Designated Accident Scenarios

(

Accident R-14C R-14D;,

R-14E-R-12.-,

R-20 r R21

- R15

'R-02,,4 R-30 R-31ABC-R-32A R-32B Time step cpm per cpm perK,cpm per: 'cp'm'*Pper cpm per',

cpm pr, ccp pe R/hr per mR/hr per, mR/h'r per,,

R/hr pe R/hr per S"

uCVcc uCVcc ucc" uc "C' c

, ucuCVcc Cc

-uClcc,,

- uCVccCV e

uCi/cc'*,-

, uCVccr uCLiUc Accident 7 Core uncovered > 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> with sprays/filtration NOT effective t=0 1.75E+07 4.60E+03 3.01 E+00 6.32E+07 6.63E+07 6.78E+07 1.83E+06 6.6E+04 6.6E+02 1.1 E+02 6.6E+01 6.6E+01 t=0.5 1.51 E+07 3.98E+03 2.60E+00 5.70E+07 5.98E+07 6.12E+07 1.80E+06 6.OE+04 5.6E+02 1.OE+02 6.OE+01 6.OE+01 t=1 1.45E+07 3.82E+03 2.50E+00 5.40E+07 5.67E+07 5.80E+07 1.85E+06 5.5E+04 5.1E+02 9.4E+01 5.5E+01 5.5E+01 t=2 1.45E+07 3.81E+03 2.49E+00 5.09E+07 5.34E+07 5.46E+07 1.96E+06 4.8E+04 4.7E+02 8.OE+01 4.8E+01 4.8E+01 t=4 1.46E+07 3.84E+03 2.51E+00 4.70E+07 4.93E+07 5.04E+07 2.12E+06 3.8E+04 3.7E+02 6.3E+01 3.8E+01 3.8E+01 t=8 1.47E+07 3.88E+03 2.54E+00 4.22E+07 4.43E+07 4.53E+07 2.31 E+06 2.9E+04 2.6E+02 4.5E+01 2.9E+01 2.9E+01 t=16 1.51E+07 3.98E+03 2.61E+00 3.67E+07 3.85E+07 3.94E+07 2.53E+06 2.1E+04 1.7E+02 3.OE+01 2.1E+01 2.1E+01 t=32 1.59E+07 4.17E+03 2.73E+00 3.08E+07 3.23E+07 3.30E+07 2.77E+06 1.6E+04 1.2E+02 1.9E+01 1.6E+01 1.6E+01 Accident 8 Spent Fuel Assembly (Gap) t=0 3.26E+07 8.59E+03 5.62E+00 6.42E+07 6.74E+07 6.90E+07 N/A 4.2E+04 4.OE+02 N/A 4.2E+01 4.2E+01 t=0.5 3.1OE+07 8.16E+03 5.35E+00 5.33E+07 5.60E+07 5.73E+07 N/A 3.3E+04 3.OE+02 N/A 3.3E+01 3.3E+01 t=W 3.04E+07 8.OOE+03 5.24E+00 4.84E+07 5.08E+07 5.20E+07 N/A 2.8E+04 2.5E+02 N/A 2.8E+01 2.8E+01 t=2 3.01 E+07 7.91 E+03 5.18E+00 4.44E+07 4.66E+07 4.77E+07 N/A 2.4E+04 2.OE+02 N/A 2.4E+01 2.4E+01 t=4 2.98E+07 7.85E+03 5.14E+00 4.02E+07 4.22E+07 4.32E+07 N/A 1.9E+04 1.4E+02 N/A 1.9E+01 1.9E+01 t=8 2.96E+07 7.78E+03 5.1OE+00 3.53E+07 3.70E+07 3.79E+07 N/A 1.3E+04 8.OE+01 N/A 1.3E+01 1.3E+01 t=16 2.94E+07 7.73E+03 5.06E+00 3.05E+07 3.20E+07 3.27E+07 N/A 9.OE+03 3.6E+01 N/A 9.OE+00 9.OE+00 t=32 2.92E+07 7.69E+03 5.04E+00 2.67E+07 2.80E+07 2.87E+07 N/A 7.7E+03 2.OE+01 N/A 7.7E+00 7.7E+00 Detector sensitivities derived from Calculation No. RNP-M/MECH 1746, (EC 49849, Set-Point, Declaration Evaluation for EP)

EPRAD-03 Rev. 12 Page 70 of 80]

ATTACHMENT 8.3.5.10 Page 1 of 1 MEASURING RADIATION LEVEL ON MAIN STEAM LINES Steel Support Wall 11' 4"1 Reference IEPRAD-03 I

Rev. 12 1

Page 71 of 80 Platform Grating

ATTACHMENT 8.3.5.11 Page 1 of 2 TYPICAL RMS VALUES The background and alarm setpoint for radiation monitors should be obtained from the control room or other current sources if they are needed to perform dose projections.

The following two tables provide the typical values for the background and alarm setpoints for radiation monitors, however these values should not be used for performing dose projections unless no other data is available. Table 1 contains all of the radiation monitors that are used for dose projections, while Table 2 contains other monitors that may be of interest to the dose projection team.

TABLE 1:

RADIATION CHANNEL DESCRIPTION TYPICAL*

SCALE MONITOR I

BKG/SETPOINT R2 CV LOW RANGE AREA

- 10 / 100 mR/HR 0.1 - 10,000 mR/HR R12 CV AIR GAS

-1 K/2.2KCPM 10 -10,000,000 CPM

= 1.8 times BKG R14C PLANT VENT GAS LOW 30-60/-10,000 CPM 10 - 1,000,000 CPM Default 1 M in high range.

switch to high range or 10 M at monitor

- 700 k CPM all R-14 channels R14D PLANT VENT GAS MID 10-11 /130 CPM 10 - 1,000,000 CPM Default 10 in low range R14E PLANT VENT GAS HIGH 10-11 /BKG + 17 CPM 10 - 1,000,000 CPM Default 10 in low range R15 CONDENSER AIR EJECTOR 10-15 /100+BKG CPM 10 - 1,000,000 CPM R20 LOWER FUEL HANDLING 10-40 / 9,800 CPM 10 -10,000,000 CPM BUILDING low range R21 UPPER FUEL HANDLING 10-25 / 9,730 CPM 10 -10,000,000 CPM FUEL HANDLING R30 LOWER FUEL HANDLING

- 0.5 !18+BKG mR/HR 1 - 100,000 mR/HR BUILDING high range R31A MAIN STEAM LINE A

- 0.3 /12 mR/HR 1 - 100,000 mR/HR R31 B MAIN STEAM LINE B

- 0.4 /12 mR/HR 1 - 100,000 mR/HR R31C MAIN STEAM LINE C

- 0.6/12 mR/HR 1 - 100,000 mR/HR R32A CV HIGH RANGE

<1/ 10&1,000 R/HR 1-10,000,000 R/HR R32B CV HIGH RANGE

<1/10&1,000 RPHR 1-10,000,000 R/HR

• A printscreen can be performed on either the EDS terminal or ERFIS at the onset of an accident to provide more current backgrounds for monitors that are not yet being effected by the accident.

EPRAD-03 Rev. 12 Page 72 of 80

ATTACHMENT 8.3.5.11 Page 2 of 2 TYPICAL RMS VALUES TABLE 2:

- RADIATION CHANNEL DESCRIPTION TYPICAL BKG/SETPOINT SCALE MONITORIII R1 CONTROL ROOM AREA

<1 1 2.5 mR/HR 0.1 - 10,000 mR/HR R3 PASS PANEL AREA 0.1-0.3/ 20 mR/HR 0.1 -10,000 mR/HR R4 CHARGING PUMP AREA

-4 / 50 mR/HR 0.1 - 10,000 mR/HR R5 SPENT FUEL BLDG. AREA

<1 / 50 mR/HR 0.1 - 10,000 mR/HR R6 SAMPLING ROOM AREA

<1 / 50 mR/HR 0.1 - 10,000 mR/HR R7 IN-CORE INSTRUMENT AREA

-4 / 200 mR/HR 0.1 - 10,000 mR/HR R8 DRUMMING ROOM 1-2 / 50 mR/HR 0.1 - 10,000 mR/HR R9 LETDOWN LINE AREA 10-40/ 3000 mR/HR 1 - 100,000 mR/HR Ri1 CV AIR PARTICULATE

-20 K / 3.6E4 CPM 10 - 1,000,000 CPM

= 1.8 times BKG R14A PLANT VENT PARTICULATE

-500 / 2E6 CPM 10 - 1,000,000 CPM R14B PLANT VENT IODINE

-10 / 90,000 CPM 10 - 1,000,000 CPM R16 HVH COOLING WATER

-300 /1,900 CPM 10 - 1,000,000 CPM R17 COMPONENT COOUNG WATER

-300 / 830 CPM 10 - 1,000,000 CPM R18 LIQUID WASTE DISPOSAL

-18,500/ VARIES 10 - 1,000,000 CPM R19A SG "A" BLOWDOWN

<2,000/ -10 K CPM 10 -10,000,000 CPM R19B SG "B" BLOWDOWN

<1,000/ -8 K CPM 10 -10,000,000 CPM R19C SG "C" BLOWDOWN

<2,000/-10 K CPM 10 -10,000,000 CPM R22P E&RC BLDG. PARTICULATE

-300/ 10,000 CPM 1 - 1,000,000 CPM R221 E&RC BUILDING IODINE

-15 /300 CPM 1 -1,000,000 CPM R22NG E&RC BUILDING NG

-40 11,000 CPM 1 - 1,000,000 CPM R23P RADWASTE BLDG. PART.

-60/9,700 CPM 1 - 1,000,000 CPM R231 RADWASTE BLDG. IODINE

<10 / 1090 CPM 1 - 1,000,000 CPM R23NG RADWASTE BLDG. NG

-20 / 387 CPM 1 - 1,000,000 CPM R33 MONITOR BLDG. AREA

<1 / 10 mR/HR 1 - 100,000 mR/HR R37 COND. POLISHER

-100 /18,500 CPM 10 -10,000,000 CPM R38P EOF PARTICULATE

-900 / 32,000 CPM 10 - 1,000,000 CPM R381 EOF IODINE

-10/802 CPM 10 - 1,000,000 CPM R38NG EOF NOBLE GAS

-25/935 CPM 10 - 1,000,000 CPM I EPRAD-03 Rev. 12 Page 73 of 80

1 08 10 tll-,06ud I

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ATTACHMENT 8.3.5.13 Page 1 of 1 WEATHER SERVICE DATA

1.

Call the Weather Service at the Florence Airport, Columbia, South Carolina or Wilmington, North Carolina. Ask for the forecaster on duty and identify yourself by saying, "This is (your name) at the Carolina Power & Light Company (CP&L)

H. B. Robinson Nuclear Plant. This is an emergency (or emergency drill). May I have the last hour surface weather observation from Florence, South Carolina?"

If the last hour data is not available from Florence, then request the last hour surface weather observation from Columbia. The following data should be obtained:

1-Hour Forecast Station for which data is given Wind Speed (MPH)

Cloud Cover (in tenths of total)

Cloud Ceiling (feet above ground)

Wind Direction (from N,S,E,W,etc.)

Wind Direction Trends (steady, shifting, variable)

Precipitation Activity Probability of Precipitation

2.

Also, obtain a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> forecast for Florence from the meteorologist on duty.

3-Hour Forecast Station for which data is given Wind Speed (MPH)

Cloud Cover (in tenths of total)

Cloud Ceiling (feet above ground)

Wind Direction (from N,S,E,W,etc.)

Wind Direction Trends (steady, shifting, variable)

Precipitation Activity Probability of Precipitation

3.

Other Information:

Date:

Time:

Name:

I EPRAD-03 Rev. 12 1

Page 75 of 80

ATTACHMENT 8.3.5.14 Page 1 of 1 ONSITE METEOROLOGICAL DATA Date:

Time Ground Wind Speed (mph)

Elevated Wind Speed (mph)

Ground Wind Dir. (From)

Elevated Wind Dir. (From)

AMB Temp. (OF)

AT (0/1100m)

Stability Class Time Ground Wind Speed (mph)

Elevated Wind Speed (mph)

Ground Wind Dir. (From)

Elevated Wind Dir. (From)

AMB Temp. (OF)

AT (°0/100m)

Stability Class IEPRAD-03 I

Rev. 12 Page 76 of 80 1

ATTACHMENT 8.3.5.15 Page 1 of 1 METEOROLOGICAL FORECAST FORM Date:

Time Issued:

Issued By:

Received By:

Forecast Location:

A)

Next 1 Hour

1)

Wind Direction: Sector Deg.

2)

Winds Should Remain (Steady; Shifting; Variable) 2a) Variation Should Be Deg.

3)

Wind Velocity:

to (MPH)

4)

Stability Class

5)

Precipitation Activity Will Be (None, Scattered, Steady)

6)

Precipitation Type (Rain, Rainshowers, Thunderstorms, Ice, Snow) 7)'

Precipitation Intensity (Light, Moderate, Severe)

B)

Next 3 Hours:

C)

Next 3 Days:

D)

Remarks:

I EPRAD-03 Rev. 12 Page 77 of 80

(

ATTAC-,..,,i-T 8.3.5.16 Page 1 of 2 HBRDOSE/RASCAL COMPARISON MATRIX I HBRDOSE RASCAL[

DIFFERENCES/COMMENTS I

EFFECTS EPA 400 Dose Yes Partial See Note 1 HBRDOSE will give higher ground exposure factors doses. Rascal will give higher external doses.

TEDE may be affected in either direction.

Reg Guide Yes No RASCAL uses a single equation with RASCAL generally will use higher X-Q values,

1. 145 XJQs non site-specific wake factors. Also, especially for ground level releases at closer probably doesn't use plume meander distances. When using default cap on Sigma z, default. RASCAL cap on Sigma z is Rascal will calculate higher doses during 500 m. R.G. 1.145 shows 3000 m.

unstable met conditions.

Deposition No Yes RASCAL calculates deposition as a NONE separate dose quantity, which does not effect TEDE.

Depletion No No NONE Wet Deposition No Yes RASCAL uses a mass balance for RASCAL immersion and inhalation doses will wet deposition.

be lower than HBRDOSE. RASCAL ground doses will be higher.

Finite Model Yes Yes RASCAL converts to a semi-infinite Conversion point of finite model, semi-infinite model at Sigma y = 400 m. HBRDOSE model will cause negligible difference. RASCAL uses Sigma y = 500 m. RASCAL uses use of Sigma y instead of an average sigma as horizontal dispersion coefficient only in described in "Meteorology and Atomic Energy" determining plume size. HBRDOSE may cause a big difference in the calculated uses average Sigma.

gamma dose for non-isotropic plumes.

Decay for TAS Yes Yes NONE Downwind No No NONE Decay I

I I___I I EPRAD-03 Rev. 12 Page 78 of 80

C Dose factors for thyroid are identical between the two models. For external dose, RASCAL includes the contribution of short lived daughters in the external dose factors, which EPA-400 does not do. Similarly, RASCAL includes short lived daughter products in the inhalation and ground exposure dose factors. The most obvious result of this is that some of the noble gases (i.e. Kr-88), are included in inhalation and ground exposure dose in RASCAL. Ground exposure dose factors are calculated in RASCAL assuming a 0.3 cm/s deposition rate and further correction factor of 0.5 to account for rough ground. EPA-400 dose factors assume a deposition velocity of 1 cm/s for iodines and 0.1 cm/s for particulates with no correction factor.

I EPRAD-03 Rev. 12 1

Page 79 of 801 ATTAC(LI..C, 'JT 8.3.5.16 Page 2 of 2 HBRDOSE/RASCAL COMPARISON MATRIX HBRDOSE RASCAL DIFFERENCES/COMMENTS EFFECTS Daughter No Yes RASCAL calculates ingrowth of HBRDOSE may underestimate doses, Ingrowth daughter decay products.

particularly inhalation doses.

Source Term RTM-96 NUREG-Different isotopes, RASCAL can During LOCA sequences, Spent Fuel accidents, 1228 dynamically calculate spectrum or Waste Gas Decay Tank rupture, there should based on particular accident be little difference. Other sequences may cause sequence.

large differences.

Uses monitor Yes No NRC will probably be doing worst NRC predicted dose will be higher. CP&L should reading for case analysis based on specific consider a method of providing gross noble gas, source term accident and PRA instead of actual iodine, and particulate release rates to the NRC.

release.

DO NOT ALLOW THE NRC TO USE THE EQUIVALENT RELEASE RATES CALCULATED BY HBRDOSE.

Source term Yes Yes HBRDOSE uses gross (cpm or NRC results will not be available for several based on EMT mrem/hr) inputs. RASCAL requires hours, but may be more accurate.

samples isotopic analysis.

Intermediate Yes No NONE Phase Calculations Note 1:

ATTACHMENT 8.3.5.17 Page 1 of 1 MANUAL CALCULATION OF CURIES RELEASED (FOR DOSE PROJECTION TEAM USE)

Use this manual calculation for stability classes E, F, and G in MIXED MODE RELEASES when X/Q is extremely small at the site boundary when compared with the other X/Q values.

The Xenon Dose Equivalent and the Iodine Dose Equivalent are the source term values used by South Carolina Department of Health and Environmental Control for input into their dose assessment program.

Use the following formula, and the data from the most reasonable distance, to calculate Q in Curies.

Q= [3600(D)] / [(1000)(X/Q)(DCF)]

where:

Q = the calculated Dose Equivalent source term in Curies, 3600 converts the release from Ci/sec to Curies 1000 = millirem - rem conversion D = TEDE or thyroid CDE in mRem, both from the dose projection program printout, X/Q = the dispersion factor (from the same distance as the dose, DCF = the dose conversion factor (Rem per uCi - cm

- hr); 20 for Xenon dose equivalent or 1.3 E +06 for Iodine dose equivalent.

Q= [3600(D)]/ [(1 000)(X/Q)(DCF)]

Q = [3.6(D)] / [(XIQ)(DCF)]

Q(Xe) = [3.6(

Q(l) = [3.6(

)] [ (

)] / [(

)(20)] = Curies Xe-133 Dose Equivalent

) (1.3E+06)] = Curies Iodine Dose Equivalent I

Rev. 12 PPage 80 of 80 1 F-EPRAD-03

CP&L R

Reference Use CAROLINA POWER & LIGHT COMPANY H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 PLANT OPERATING MANUAL VOLUME 2 PART 5 EMERGENCY PROCEDURE EPTSC-04 RADIOLOGICAL CONTROL DIRECTOR REVISION 5

EPTSC-04 Rev. 5 Page 1 of o1

SUMMARY

OF CHANGES I EPTSC-04 Rev. 5 Page 2 of 10 1 Step #

REVISION COMMENTS 8.4.3.8 Revised step from "Turn on the Post Accident Sampling System" to "Direct sampling activities, as necessary, to assist in accident assessment." Also added reference to RNP RA/01 -0164 and NRC Amendment No. 192.

Entire Revised page numbering to reflect AP-007 format.

Procedure

TABLE OF CONTENTS SECTION PAGE Q UICK STA RT G UIDE...................................................................................................

4 8.4.1 PURPO SE.................................................................................................

5 8.4.2 RESPO NSIBILITIES.............................................................................

5 8.4.3 INSTRUCTIO NS....................................................................................

5 8.4.4 RECO RDS...........................................................................................

10 8.4.5 ATTACHM ENTS...................................................................................

10 IEPTSC-04 I

Rev. 5 Page 3 of 101

RADIOLOGICAL CONTROL DIRECTOR (RCD) QUICK START GUIDE NOTE:

Blanks are provided for place keeping q/s only, logs are the official record. This is a summary level guide and does not replace the procedure steps.

1.

Sign-in on facility sign-in board. Log on Electronic Display System (EDS).

2.

If dialogic was utilized for callout, upon arrival at the Technical Support Center (TSC), notify dialogic of your arrival at the facility (857-1777).

3.

Obtain a briefing on Plant Status.

4.

Determine if contaminated individuals have been released from the plant.

5.

Obtain wind direction (degrees blowing from). Request 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, and 3 day weather forecasts.

6.

Coordinate with the Radiological Control Manager (RCM) in the Emergency Operations Facility (EOF).

7.

Determine the E&RC staff available in the emergency facilities, request additional resources as needed.

8.

Request updates on the E&RC Team status every 30 minutes from an available E&RC Supervisor or assigned "lead" person.

9.

Determine status of habitability for the TSC (from the RCM) and for the OSC (from the OSC Leader).

10.

Notify the SEC of readiness to activate.

11.

Refer to procedure steps.

I EPTSC-04 Rev. 5 Page4of10

8.4 RADIOLOGICAL CONTROL DIRECTOR (RCD) 8.4.1 PURPOSE This procedure describes the functional responsibilities and procedure steps for the Radiological Control Director (RCD).

8.4.2 RESPONSIBILITIES

1.

Manage the radiological control activities in the Technical Support Center (TSC).

2.

Monitor meteorology, onsite radiological consequences and dose projections.

3.

Liaison with the Radiation Control Manager (RCM) in the Emergency Operations Facility (EOF).

8.4.3 INSTRUCTIONS NOTE:

The Radiation Control (RC) Technician on shift will report to the Superintendent-Shift Operations and support Operations during an emergency.

The technician will continue reporting to the SSO after activation of the Operations Support Center (OSC) unless higher priority actions are required as deemed necessary by the OSC Leader.

1.

Advise the E&RC Team Lead (either E&RC Supervisor or "lead" technician) of monitoring locations and sample collection points in the plant, collection of required data and assessment of radiological conditions at these points.

2.

Request in-plant samples to assess plant/fuel conditions.

3.

Report to the Site Emergency Coordinator (SEC) regarding:

a.

Radiological monitoring and assessment,

b.

Radiation exposure control,

c.

Team direction & supporting missions, EPTSC-04 Rev. 5 Page 5 of 10

8.4.3.3 (Continued)

d.

Emergency facility habitability, TSC/EOF Building status as decided by the RCM and ERM.

e.

Sampling and analysis, and

f.

Liaison with Offsite Radiation Control (RC) personnel and the RCM in the EOF

4.

Advise the Environmental & Radiation Control (E&RC) Team Supervisor or Lead person regarding:

a.

Prioritizing tasks,

b.

Determining protective gear and dosimetry,

c.

Development of precautions for the reentry team briefing,

d.

Deviations from a full set of anti-contamination clothing, and

e.

Changes to requirements for protective equipment.

5.

Determine the need for on-site protective sheltering or evacuation, along with routes (to and from the plant) based on plant data, dose projections and meteorology.

a.

Recommend site evacuation assembly location.

Monitor personnel at access points as required.

6.

Consult the Dose Projection Team Leader (DPTL) in the EOF to determine affected zones in the 10 mile Emergency Planning Zone (EPZ). Assign priorities as necessary.

7.

Contaminated, injured personnel should be treated on site if possible.

8.

Direct sampling activities, as necessary, to assist in accident assessment.{RNP RAI01-0164; NRC Amendment No. 192}

9.

Notify the RCM regarding Phase "A" Isolation.

IEPTSC-04 I

Rev. 5 Page 6 of 10 1

8.4.3 (Continued)

10.

Provide guidance to the E&RC Team Supervisor or Lead Person for establishing personnel and vehicle decontamination areas when required.

a.

Determine if an alternate means of transporting personnel from the plant is needed.

b.

Based on wind direction and magnitude of release, determine an appropriate area to set up for vehicle decon.

c.

Determine the proper method of decon and area setup (i.e.,

masslin wipe down, wash down with soap and water, water supply, water containment, decon supplies, etc.)

d.

Determine release limits.

e.

Consider personnel transport in CP&L vehicle(s), and deferring vehicle decon until part of the recovery effort.

f.

If radiation levels on site prohibit adequate decontamination or monitoring these functions may be performed at county operated locations.

Inform county emergency management officials if this contingency must be used.

11.

Coordinate with the State and the Nuclear Regulatory Commission (NRC) as required.

12.

Ensure exposure control and that Special Radiation Work Permits (RWPs) are issued as necessary. Approve exposure extensions.

13.

Ensure that necessary information is posted on displays and status boards. Including:

a.

Onsite radiological status

b.

Protective Action Recommendations (PARs)

c.

10 mile emergency planning zone (EPZ) map

d.

TSC Habitability Status.

I EPTSC-04 I

Rev. 5 Page7 of 10 1

8.4.3 (Continued)

14.

Recommend the administration of potassium iodide (KI) to CP&L personnel and contract employees when the Committed Dose Equivalent (CDE) to the thyroid is > 25 Rem.

a. Determine if KI is required for personnel in buildings designed to maintain habitability such as the Control Room and TSC/EOF building.

15.

Regulatory limits shall be observed for planned radiation exposures to emergency workers unless the Plant General Manager (PGM),

the Radiological Control Director (RCD) or the Site Emergency Coordinator (SEC) authorizes the individual to exceed 5 Rem TEDE in a year.

16.

Follow these Emergency Worker Dose Guidelines:

I EPTSC-04 I

Rev. 5 Page 8 of 10 1 NOTE:

In all cases, it is the responsibility of each individual, to maintain the total effective dose equivalent ALARA.

Declared pregnant women shall not participate in these actions.

Internal exposures shall be minimized by respiratory protection and contamination controlled by the use of protective clothing.

Entry into High Radiation Areas shall not be permitted unless instrumentation capable of measuring the anticipated radiation levels is provided.

Entry into a High Radiation Area shall require wearing a self-reading dosimeter capable of measuring the expected exposure to be received.

Entry into Radiation Fields of > 100 Rem/hr. shall not be permitted unless specifically authorized by the PGM or RCD. In their absence the SEC shall authorize.

8.4.3.16 (Continued)

a.

Repair/Reentry efforts may require individuals to enter a hazardous area to protect valuable installations, or to make the facility more secure against events which could lead to radioactivity releases (i.e., assessment actions or entry of damage repair parties who are to repair valve leaks or add iodine-fixing chemicals to spilled liquids).

In such instances, planned dose to emergency workers shall not exceed 10 Rem TEDE to the whole body, 30 Rem to the lens of the eye, or 100 Rem to any other organ including skin and extremities.

b.

Lifesaving Actions or Protection of Large Population efforts may require personnel to search for and remove injured persons or entry to prevent conditions that would probably injure numbers of people, a planned dose shall not exceed 25 Rem TEDE to the whole body, 75 Rem to the lens of the eye, or 250 Rem to any other organ including skin and body extremities. This applies to:

The removal of injured persons if the saving of life is possible.

Entry to prevent conditions that, if left uncorrected, could lead to damage or releases that would probably injure numbers of people on or offsite.

Justifiable dose limits for situations in which the collective dose avoided by the emergency operation is significantly larger than that incurred by the workers involved.

c.

Actions requiring a dose > 25 Rem shall consider the following in addition:

Rescue personnel shall be volunteers and shall be instructed about the risks involved. Refer to EPOSC-04, Emergency Work Control.

Volunteers above the age of 45 shall be selected when possible for the purpose of avoiding unnecessary genetic effects.

[EPTSC-04 Rev.

I Page9of 10

8.4.3 (Continued)

17.

Review PLP-021, "Chemical Storage, Inventory, Spill and Hazard Communication Program", for items to consider in the event of a chemical spill or accident.

NOTE:

Contact numbers for the Environmental Compliance Unit are listed in the Emergency Response Organization Phone Book.

a. Contact the Environmental Compliance Unit to determine reportability.

b. Ensure the settling pond is isolated from the discharge canal for spills directed toward storm drains.

18.

Develop recovery strategy.

8.4.4 RECORDS N/A 8.4.5 ATTACHMENTS N/A I EPTSC-04 Rev. 5 Page 10 of 10

CP&L R

Reference Use CAROLINA POWER & LIGHT COMPANY H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 PLANT OPERATING MANUAL VOLUME 2 PART 5 EMERGENCY PROCEDURE EPTSC-07 DAMAGE ASSESSMENT REVISION 4

EPTSC-07 Rev. 4 Page

of ý53

SUMMARY

OF CHANGES STEP #

REVISION COMMENTS Step 8.7.1.1 Deleted sentence requiring analysis of system radionuclide concentration Paragraph 2 for confirming direct interpretation results.

Step 8.7.3 Re-word note to identify the methods to be used for core damage Note assessment: direct interpretation of plant instrumentation and radioanalysis.

Step 8.7.3.2 Re-worded step to describe sampling and radioanalysis methods used in work packages 2, 3, and 4.

Step 8.7.3.3 Re-worded step to indicate that work packages 5, 6, and 7 should be Sentence 2 used initially to determine core damage.

Attachment Revised radioisotope concentrations to reflect per RNP-M/MECH-1742, 8.7.5.2, page Design Inputs for E-Plan Dose Assessment, Source Term Rad Monitor 3 of 3 Information, Table 1 Attachment Revised radioisotope concentrations to reflect per RNP-M/MECH-1 742, 8.7.5.3, page Design Inputs for E-Plan Dose Assessment, Source Term Rad Monitor 2 of 5 Information, Table 1 Attachment Revised radioisotope concentrations to reflect per RNP-M/MECH-1742, 8.7.5.4, page Design Inputs for E-Plan Dose Assessment, Source Term Rad Monitor 2 of 7 Information, Table 1 Attachment Revised work package 5 to reflect core damage assessment process per 8.7.5.5 EC 47067, Revised Core Damage Assessment Process.

Attachment Revised work package 6 to reflect core damage assessment process per 8.7.5.6 EC 47067, Revised Core Damage Assessment Process.

Attachment Revised work package 7 to reflect core damage assessment process per 8.7.5.7 EC 47067, Revised Core Damage Assessment Process.

Attachment Deleted work package flowchart.

8.5.7.12 Entire Re-formatted page numbering to reflect AP-007 format.

Procedure IEPTSC-07 I

Rev. 4 L

Page2 of53 I

TABLE OF CONTENTS SECTION 8.7.1 8.7.2 8.7.3 8.7.4 8.7.5 8.7.5.1 8.7.5.,

8.7.5.*

8.7.5.z 8.7.5.,

8.7.5.'

8.7.5.

8.7.5.

8.7.5.

8.7.5.

8.7.5.

8.7.5.

PAGE P U R PO S E..............................................................

4 RESPONSIBILITIES............................................................................

4 INSTRUCTIONS...................................................................................

4 R ECO R D S..............................................................................................

5 ATTACHMENTS...................................................................................

6 Work Package 1 - General Information Needed for All Assessment Methods.................................................................

7 Work Package 2 - Assessment Using Radionuclide Analysis - Reactor Power Constant for 30 Days or More.......... 11 3

Work Package 3 - Assessment Using Radionuclide Analysis - Reactor Power Constant for 4-30 Days................... 14 4

Work Package 4 - Assessment Using Radionuclide Analysis - Reactor Power Constant for Less Than 4 Days............ 19 5

Work Package 5 - High Level Core Damage Assessment.....

26 6

Work Package 6 - Fuel Rod Clad Damage...............................

31 7

Work Package 7-Fuel Overtemperature Damage.................... 33 8

Work Package 8 - Summary of Assessments.......................... 37 9

Characteristics of Categories of Fuel Damage.........................

48 10 Core Protection.......................................................................

49 11 Function Restoration Procedures............................................

51 12 Definitions................................................................................

53 I EPTSC-07 Rev. 4 I

Page3of 53

8.7 DAMAGE ASSESSMENT 8.7.1 PURPOSE

1.

The purpose of this procedure is to provide guidance and direction to the Technical Analysis Director and the Accident Assessment Team in the evaluation of core damage and implementation of accident assessment actions of PLP-007, Robinson Emergency Plan.

This procedure provides methods used to identify the four major fuel conditions using containment radiation monitor readings, core exit thermocouple readings, hydrogen concentration, and subcooling: 1) no damage; 2) clad damage; 3) fuel overtemperature; and, 4) fuel melt.

8.7.2 RESPONSIBILITIES

1.

The Technical Analysis Director is responsible for ensuring appropriate Accident Assessment Team activation and utilization of this procedure.

2.

The Accident Assessment Team is responsible for implementation of this procedure.

8.7.3 INSTRUCTIONS NOTE:

This procedure uses two methods of core damage assessment: direct interpretation of plant instrumentation and radionuclide analysis.

1.

This procedure is arranged into 8 Work Packages.

Work Package 1 may be used for either core damage assessment method; however, it is not required when assessing core damage with Work Packages 5, 6, and 7.

IEPTSC-07 I

Rev. 4 1

Page4of 531

8.7.3 (Continued)

2.

Work Packages 2, 3, and 4 are based on sampling and subsequent analysis and are dependent on how long the reactor has been at a constant power level. This method may be used to confirm the assessment in Work Packages 5, 6, and 7.

3.

Work Packages 5, 6, and 7 are plant instrument specific. These work packages should be used to initially determine core damage.

4.

Work Package 8 is a summary of assessment activities, incorporating available information from the other work packages.

5.

The following Attachments are provided as informational materials or as a summary of information contained within the work packages.

a..7.5.9, Characteristics of Categories of Fuel

Damage,

b..7.5.10, Core Protection,

c..7.5.11, Functional Restoration Procedures,

d..7.5.12, Definitions.

8.7.4 RECORDS N/A I EPTSC-07 I

Rev. 4 L

Page 5of53

8.7.5 ATTACHMENTS 8.7.5.1 Work Package 1 - General Information Needed for All Assessment Methods 8.7.5.2 Work Package 2 - Assessment Using Radionuclide Analysis

- Reactor Power Constant 30 Days or More 8.7.5.3 Work Package 3 - Assessment Using Radionuclide Analysis

- Reactor Power Constant 4-30 Days 8.7.5.4 Work Package 4 - Assessment Using Radionuclide Analysis

- Reactor Power Constant Less Than 4 Days 8.7.5.5 Work Package 5 - High Level Core Damage Assessment 8.7.5.6 Work Package 6 - Fuel Rod Clad Damage 8.7.5.7 Work Package 7 - Fuel Overtemperature Damage 8.7.5.8 Work Package 8 - Summary of Assessments 8.7.5.9 Characteristics of Categories of Fuel Damage 8.7.5.10 Core Protection 8.7.5.11 Function Restoration Procedures 8.7.5.12 Definitions IEPTSC-07 I

Rev. 4 Page 6 of 531

ATTACHMENT 8.7.5.1 Page 1 of 4 WORK PACKAGE 1-GENERAL INFORMATION NEEDED FOR ALL ASSESSMENTS Work Package 1 - General Information Needed for Assessment Methods.

1.

Use this work package first when using Work Packages 2, 3, and 4 to determine core damage. Work Packages 5, 6, and 7 may be completed without completing this work package. The information required for Work Packages 5, 6, and 7 is recorded within each package.

2.

Obtain plant data and radiochemistry sample data as necessary to complete this package.

3.

Use an ERFIS/EDS Terminal to obtain Plant data, as follows:

a.

Access the group library function Located under turn on core "Real," for real time data display on EDS.

b.

Select "COREDAMG"

c.

A copy of the group may be printed for convenience.

4.

Record sample times, dates, temperatures, pressures and corrected system volumes on the worksheet.

5.

Use this package to correct reactor coolant and RHR system density where requested.

6.

Use this package to convert sump level to total gallons of RHR system volume in containment.

EPTSC-07 Rev. 4 Page 7of 53

ATTACHMENT 8.7.5.1 Page 2 of 4 WORK PACKAGE 1-GENERAL INFORMATION NEEDED FOR ALL ASSESSMENTS CORE DAMAGE ASSESSMENT SAMPLE DATA WORKSHEET 1.0 CONTAINMENT ATMOSPHERE A. Date and time sample drawn:

/.1L hours B.

CV temperature @ time of sample: _F

+ 460 =

°R C. CV pressure @ time of sample:

psig + 14.7 = __

psia D. CV volume (corrected): (1.612 E09) x (Step B - Step C) = __

cc @ STP 2.0 REACTOR COOLANT SYSTEM A. Date and time sample drawn:

/

I, hours B. RCS temperature @ time of sample:

OF C. Pressurizer level @ time of sample: __

x.01 =

D. Water density ratio:

= p(t)/p(@STP)

E.

RCS volume (corrected):

[(2.29 E08) + (Step C)(3.34 E07)] x (Step D) =

cc 3.0 RHR SYSTEM A. Date and time sample drawn:

.__/___/__/,

_hours B.

CV sump level @ time of sample:

(LI-801 or LI-802 )

(gallons in sump) x 3785.6 cc/gal cc C. If the RHR System is used during a LOCA event, then the RHR volume is determined as:

Step B + 3.7856 E07 cc =

cc If the RHR System is used in a cooldown mode with the RCS intact, then the RHR volume is determined as:

[(2.29 E08) + (Step C in Section 2.0) (3.34 E07)] cc +

3.7856 E07 cc =

cc D. Water density ratio:

= p(t)Ip(@STP)

E. RHR volume (corrected): (Step C) x (Step D) =

cc 4.0 TIME OF REACTOR SHUTDOWN A. Time C. Percent power prior to shutdown B. Date D. Core Burnup EFPD IEPTSC-07 Rev. 4 Page 8 of 53

ATTACHMENT 8.7.5.1 Page 3 of 4 WORK PACKAGE 1-GENERAL INFORMATION NEEDED FOR ALL ASSESSMENTS WATER DENSITY RATIO VS. TEMPERATURE I

I I

I

.1 0.

6.

50 100 I50 200 250 300 350 400 Temoerature, degrees F 450 500 550 600 This graph assumes 2250 small (< 1 %) error.

psia, however it can b used with lower pressures with a very I EPTSC-07 I

Rev. 4 Page 9 of 53 1 FC,,

4.'

0 4.'

'U m

V L.

V

'U 650

ATTACHMENT 8.7.5.1 Page 4 of 4 WORK PACKAGE 1-GENERAL INFORMATION NEEDED FOR ALL ASSESSMENTS CONTAINMENT WATER VOLUME VS. SUMP LEVEL P_ o r

i.

z

/'

inches

-ap P4)

W CM M

TO

('1)

VIZ WO 0 O

W 0-POT Q 0 Mq M

(INCHES) ABOVE BoTIom OF CONTADIMENT SUMP EPTSC-07 Rev. 4 Page 10 of 53]

ATTACHMENT 8.7.5.2 Page 1 of 3 WORK PACKAGE 2 - ASSESSMENT USING RADIONUCLIDE ANALYSIS REACTOR POWER CONSTANT 30 DAYS OR MORE NOTE:

Constant reactor power is defined as the operating condition where there

ý is less than 10% rated thermal power variation during the period.

1.

Obtain the results of the radionuclide analysis from the Radiological Control Director and complete this package.

2.

Enter the uCilcc sample activity in the appropriate space (i.e., RCS, RHR, or CV) for each nuclide.

NOTE:

Because of the long counting time associated with accurate strontium analysis, Ba-1 40 will serve as the initial fuel melt indicator until the strontium results are obtained and confirmed.

3.

Using the decay constant provided in Column (2) for each isotope, divide the Column (1) sample activity by the Column (2) value which is the product of the decay constant and the time difference between sample time and reactor shutdown, to determine the corrected specific activity in Column (3).

4.

Record the specific system corrected volume from Work Package 1 in Column (4). Multiply the corrected specific activity from Column (3) by the volume in Column (4) to determine the corrected system total uCi content.

Record the results in Column (5).

I EPTSC-07 Rev. 4 Page 11 of 53 NOTE:

On this package, the sample activity in Column (1) is corrected during laboratory analysis back to the original activity at the time of sampling. To accurately assess core damage, this activity must be corrected to the specific activity at shutdown on Column (3) and the total activity in Column (5) which would have yielded that sample activity if the release had occurred at shutdown.

This activity is compared to the adjusted power source term in Column (6) to estimate % of nuclide release. Columns (3) and (5) are representative activities if the release occurs at the instant of reactor shutdown. All total activities are corrected to time of shutdown to make calculations easier.

RCS pressure, temperature, or power transients may result in increased RCS iodine concentrations without clad damage (iodine spiking). Do not use iodine concentrations alone as evidence of fuel clad damage.

ATTACHMENT 8.7.5.2 Page 2 of 3 WORK PACKAGE 2 - ASSESSMENT USING RADIONUCLIDE ANALYSIS REACTOR POWER CONSTANT 30 DAYS OR MORE

5.

Obtain the percent constant reactor power prior to the incident and the Effective Full Power Days (EFPD) of fuel used in the cycle from the Plant Operations Director. Use this data where applicable to determine the correction terms identified in Column (6).

6.

Divide Column (5) by Columns (6) and (7) and multiply by 100 to obtain the percent released per system in Column (8).

7.

To obtain the total percent released, sum all three sample system results from Column (8) and record in Column (9).

8.

Proceed tb Work Package 8 - Summary of Assessments.

I EPTSC-07 Rev. 4 Page 12 of 53

C C

ATTACHMENT 8.7.5.2 Page 3 of 3 WORK PACKAGE 2 - ASSESSMENT USING RADIONCLIDE ANALYSIS - REACTOR POWER CONSTANT 30 DAYS OR MORE 230 DAY NUCLIDE RELEASE WORKSHEET CORRECTED

=

VOLUME cc CORRECTED SYSTEM TOTAL ACTIVITY uCI (1)

(2)

(3)

(4)

(5)

RCS

+

e-.545t

=

x

+

Kr-87 RHR

+

e-545t

=

x CV

+

a-_

=

x

+

RCS

+

e-244t

=

x

+

Kr-88 RHR

+

e-244t

=

x 4

CV

+

e-244t

=

x

=

RCS

+

e- 004t

=

x

+

1-131 RHR

+

e- 004t

=

x

=

CV

+

e- 004t

=

x

=

RCS

+

1

=

x

=

+

Cs-134 RHR

+

1

=

x CV

+

1

=

x RCS

+

1

=

x

=

Cs-137 RHR

+

1

=

x

=

CV

+

1

=

x RCS

+

e- 009t

=

x

+

Te-132 RHR

+

e- 009t

=

x CV

+

e- 009t

=

x

+

RCS

+

1

=

x

+

Sr-89 RHR

+

1

=

x

+

CV

+

1

=

x 4

RCS

+

1

=

x Sr-90 RHR

+

1

=

x

+

CV

+

1

=

RCS 4

e- 002t

=

x

+

Ba-140 RHR e.002t

=

x CV

+

e- 002t

=

x

=

+

REACTOR POWER LEVEL CONSTANT FOR 30 DAYS (LESS THAN 10% CHANGE)

NOTE: Due

= (hours) time difference between sample time and reactor shutdown analysis the EFPD = effective full power days rcdedp X = (hours') decay constant (1) RNP.M/MECH-1742, Design Inputs for E-Plan Dose Assessment, Source Term Rad Monitor Information, Table 1

+ CORRECTION TERM +

SOURCEP)

TERM uCI (6)

(7)

"% power/100+ 3 03E13

"% power/100 + 3 03E13

% power/1 00+ 303E13

% power/100+ 4.20E13

% power/100+ 4.20E13

% power/100+ 4.20E13

% power/100+ 6.20E13

% power/1 00 + 6.20E1 3

% vower/1 00+-6 20E13

.9

+ 1.25E13

.9

+ 1.25013 F

9

+ 1.28E12 EFPD/1140 + 8.87212 EFPD/1140 + 8,87E12

% power/100 + 8.91 213

% power/100+ 8.91E13

% power/100+ 8 91E13 EFPD/1140 + 5.90E13 EFPD/1140 + 5.90E13 EFPD/1140 + 5 90E13 EFPD/1140 + 6.16E12 EFPD/1140 + 6.16E12 E FPD/1140 + 6.16212

% power/100 + 1.13E14

% power/100 + 1.13E14 o/ nnwer/100 + 1.13E14 x 100 PERCENT RELEASE (q) x 100 =

x 100 =

x 100 =

x 100 =

x 100 =

x 100 =

x 100 =

x 100=

x 100 =

x 100=

x 100 =

x 100 =

x 100=

x 100=

x 100=

x 100=

x 100 =

x 100 =

x 100=

x 100 =

x 100 =

x 100 =

x 100 =

x 100 =

x 100 =

x 100 =

x 100 =

x 100 =

to the long analysis time associated with Strontium se columns will not be completed initially, but can be on receiving sample results.

I Page l3of EPTSC-07 Rev. 4 NUCLIDE SAMPLE SAMPLE

+

ACTIVITY uCicc x

CORRECTED SPECIFIC ACTIVITY uCt/cc TOTAL PERCENT RELEASED %

18%

I Page 13of5 e-11

ATTACHMENT 8.7.5.3 Page 1 of 5 WORK PACKAGE 3 - ASSESSMENT USING RADIONUCLIDE ANALYSIS REACTOR POWER CONSTANT 4 -30 DAYS NOTE:

Constant reactor power is defined as the operating condition where there is less than 10% rated thermal power variation during the'period.

1.

Obtain the results of the radionuclide analyses from the Radiological Control Director (RCD) and complete this package.

2.

Complete Columns (1) through (5) in accordance with the instructions of Work Package 2.

3.

Obtain the reactor power history for the 30 days prior to the incident and the Effective Full Power Days (EFPD) of fuel used in the cycle from the Plant Operations Director. As applicable, use the power history data, Pages 3 through 5 or the EFPD data to determine the correction terms identified in Column (6).

4.

Complete Columns (8) and (9) of this package.

5.

Proceed to Work Package 8.

EPTSC-07 Rev. 4 Page 14 of 53 NOTE:

If the thermal power has not been constant over the last 30 days prior to shutdown, then some isotopes have not reached equilibrium concentrations. If the power level has not been constant, adjustments for the effects of power changes must be made. Where core power levels have changed by more than 10% of rated thermal power, a separate line must be completed for 1-131, Te 132, and Ba-140 at each power level.

RCS pressure, temperature, or power transients may result in increased RCS iodine concentrations without clad damage (iodine spiking). Do not use iodine concentrations alone as evidence of fuel clad damage.

(

ATTACHMENT 8.7.5.3 Page 2 of 5 WORK PACKAGE 3 - ASSESSMENT USING RADIONUCLIDE ANALYSIS - REACTOR POWER CONSTANT 4 - 30 DAYS z4 DAY, <30 DAY NUCLIDE RELEASE WORKSHEET NUCLIDE SAMPLE SAMPLE +

ACTIVITY uCVcc e.lt CORRECTED SPECIFIC ACTIVITY uCVcc x

CORRECTED VOLUME cc

=

CORRECTED SYSTEM TOTAL ACTIVITY uCI

+-

CORRECTION TERM +

SOURCE',1 TERM uCI xl00 =

PERCENT RELEASE (t1 (2)

(31 (4)

(5)

(6)

(7)

(8)

(9(

RCS e-.545t

=

x

+ % power/100 + 3.03E13 x 100 =

Kr-87 RHR

+

e-545t

=

x

+ % power/100 + 3.03213 x 100 =

CV e-545t

=

x

+ % power/100 + 3 03E13 x 100 =

RCS

+

e-244t

=

x

+ + % power/100 + 4 20E13 x 100 =

Kr-88 RHR

+

e-244t

=

x

+ ÷ % power/100 + 4.20E13 x 100 =

CV

+

e-244t

=

x

+

+ % power/100 + 4 20E13 x100=

RCS

+

e-004t

=

x

+

• Complete 6 20E13 x100=

1-131 RHR

+

e- 004t

=

x

+,

Att. 8 7.5.3 6.20E13 x100=

CV e-,004t

=

x

+

6 20E13 x 100 =

RCS

÷ 1

=

x

+

9

+ 1.25E13 x100=

Cs-134 RHR 1

=

x

+

9

+ 1.25E13 x 100 =

CV

+

1

=

x

+

9

+ 1.25E13 x 100 =

RCS

+

1

=

x

+ EFPD/1140 + 8.87E12 x100=

Cs-137 RHR

+

1

=

x

+ EFPD/1140

• 8.87E12 x100=

CV

+

1

=

x

+ EFPD/1140 + 887E12 x100=

RCS

+

e- 009t

=

x

+ + Complete 8.91E13 x 100 =

Te-132 RHR e- 009t

=

x

+ Att. 8.7.5.3 8.91E13 x100=

CV

+

e- 009t

=

x

+

8 91E13 x 100 =

RCS

+

1

=

x

+ + EFPD/1140

+ 5.90E13 x100=

Sr-89 RHR

+

1

=

x

+ EFPD/1140

+ 5.90E13 x100=

Cv 1

=

x

+ EFPD/1140

+ 590E13 x100=

RCS 4

1

=

x

+ EFPD/1140

+ 6.16E12 xl00=

Sr-90 RHR

+

1

=

EFPD/1140

+ 6.16212 x100=

CV

+

1

=

x

+ EFPD/1140

+ 616E12 x100=

RCS

+

e- 002t

=

x

+ Complete 1.13E14 x 100 =

Ba-140 RHR

+

e- 002t

=

x

+ ÷ Att. 8.7.5.3 1.13E14 x 100=

CV

+

e- 002t

=

x

+ ÷ 1.13E14 x 100 REACTOR POWER LEVEL CONSTANT FORŽ4 DAYS and <30 DAYS (LESS THAN 10% CHANGE)

N01 t = (hours) time difference between sample time and reactor shutdown anal EFPD = effective full power days re X = (hours 1) decay constant (1) RNP-M/MECH-1742, Design Inputs for E-Plan Dose Assessment, Source Term Rad Monitor Information, Table I rE: Due to the long analysis time associated with Strontium lysis these columns will not be completed Initially, but can be irded upon receiving sample results.

rEPTSC-07 Rev. 4 Page 15 of 53 TOTAL PERCENT RELEASED %

ATTACHMENT 8.7.5.3 Page 3 of 5 WORK PACKAGE 3 - ASSESSMENT USING RADIONUCLIDE ANALYSIS REACTOR POWER CONSTANT 4-30 DAYS CORRECTION TERM WORKSHEET FOR 1-131 Sum (Cl)

(02)

(C3)

(C4)

(C5)

(C6)

Sum (C2) = Must be equal to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />.

Correction Term = Sum (C6)/100 P, = % Average reactor power for period j tj= (hours) Duration of operation at P, tj= (hours) Duration from end of interval t, to reactor shutdown 1-131

. = 0.004 HOURS-'

I EPTSC-07 Rev. 4 I

Page 16 of 53]

Pi1 t

tji 1 -exp (4Xt) exp(-AXtji)

(01 )*(04)*(05)

ATTACHMENT 8.7.5.3 Page 4 of 5 WORK PACKAGE 3 - ASSESSMENT USING RADIONUCLIDE ANALYSIS REACTOR POWER CONSTANT 4-30 DAYS CORRECTION TERM WORKSHEET FOR Te-132 (C1)

(C2)

(C3)

(C4)

(C5)

(C6)

Pj tj ti, 1 -exp (-Xt,)

exp(-Xt,,)

(C1)*(C4)*(C5)

Sum (C2) = Must be equal to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />.

Correction Term = Sum (C6)/100 Pj = % Average reactor power for period j tj= (hours) Duration of operation at P, tji= (hours) Duration from end of interval tj to reactor shutdown Te-132

), = 0.009 HOURS"1 EPTSC-07 Rev. 4 Page 17 of 53 Sum

ATTACHMENT 8.7.5.3 Page 5 of 5 WORK PACKAGE 3 - ASSESSMENT USING RADIONUCLIDE ANALYSIS REACTOR POWER CONSTANT 4-30 DAYS CORRECTION TERM WORKSHEET FOR Ba-140 (C4)

(C5)

(C6)

Pi tj ti, 1-exp (-Xtj) exp(-X.tj,)

(C1)*(C4)*(C5)

Sum (C2) = Must be equal to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />.

Correction Term = Sum (C6)/100 Pj = % Average reactor power for period j tj= (hours) Duration of operation at P1 t= (hours) Duration from end of interval t, to reactor shutdown Ba-1 40 X = 0.002 HOURS-1 IEPTSC-07 I

Rev. 4 Page 18 of 53 Sum (C 1 )

('C2)

(C3)

ATTACHMENT 8.7.5.4 Page 1 of 7 WORK PACKAGE 4 - ASSESSMENT USING RADIONUCLIDE ANALYSIS REACTOR POWER CONSTANT LESS THAN 4 DAYS

1.

Obtain the results of the radionuclide analyses from the Radiological Control Director (RCD) and complete this package.

2.

Complete Columns (1) through (5) in accordance with the instructions of Work Package 2.

3.

Obtain the reactor power history for the 4 days prior to the incident and the Effective Full Power Days (EFPD) of fuel used in the cycle from the Plant Operations Director. As applicable, use the power history data and the EFPD data to determine the correction terms identified in Column (6).

NOTE:

If the power level has not been constant, adjustments for the effects of power changes must be made. Where core power levels have changed by more than 10% of rated thermal power, a separate line must be completed for Kr-87, Kr-88, 1-131, Te-1 32, and Ba-1 40 for each power level.

RCS pressure, temperature, or power transients may result in increased RCS iodine concentrations without clad damage (Iodine spiking). Do not use iodine concentrations alone as evidence of fuel clad damage.

4.

Complete Columns (8) and (9) of this package.

5.

Proceed to Work Package 8.

I EPTSC-07 Rev. 4 I

Page 19 of 53

(

(

ATTACHMENT 8.7.5.4 Page 2 of 7 WORK PACKAGE 4 - ASSESSMENT USING RADIONUCLIDE ANALYSIS - REACTOR POWER CONSTANT LESS THAN 4 DAYS

<4 DAY NUCLIDE RELEASE WORKSHEET NUCLIDE SAMPLE SAMPLE +

ACTIVITY uCvcc CORRECTED SPECIFIC ACTIVITY uCicc x

CORRECTED VOLUME cc

=

CORRECTED SYSTEM TOTAL ACTIVITY uCi

+-

CORRECTION TERM +

SOURCEi'1 TERM uCi xl00 =

PERCENT RELEASE (1)

(2)

(3)

(4)

(5)

(6)

(7)

(8)

(9)

RCS

+

e-545t

=

x

+ Complete

+ 3.03E13 x 100 =

Kr-87 RHR

+

e-545t

=

x

+ + Att. 8.7.5.4

+ 3.03E13 x 100 =

CV

+

e-545t

=

x

+ +

• 3 03E13 x 100 RCS

+

e-244t

=

x

=

+ Complete

+4 20E13 x 100=

Kr-88 RHR

+

e-.244t

=

x

+ + Att. 8.7.5.4

+ 4.20E13 x 100 =

CV

+

e-244t

=

x

+ +

4 20E13 x 100 RCS

+

e- 004t

=

x

+ Complete 6 20E13 x100=

1-131 RHR

+

e-004t

=

x

+ Att. 8.7 5 4 6.20E13 x100=

Cv

+

e-.004t

=

x

+

6 20E13 x 100 =

RCS

+

1

=

x

=

+

.9

+ 1 25E13 x 100 =

Cs-134.

RHR

+

1

=

x

+

.9

+ 1.25E13 x 100 =

CV

+

1

=

x

+

+

.9

+ 1 25E13 x 100 =

RCS I

=

x

+ EFPD/1140 +887E12 x100=

Cs-137 RHR

+

1

=

x

+ EFPD/1140 + 8.87E12 x 100 =

CV

+

1 x

+ EFPD/1140 *887E12 x100=

RCS

+

e. 009t

=

x

+ Complete 8.91E13 x100=

Te-132 RHR

+

e- 009t

=

x

+, Att. 8.7.5.4 8.91E13 x100=

CV e- 009t

=

x

+,

8.91E13 x 100 =

RCS

+

1

=

x

=

+ EFPD/1140

+5.90E13 x100=

Sr-89 RHR

+

1

=

x

+ EFPD/1140

+ 5.90E13 x 100=

CV

+

1

=

x

+ EFPD/1140

+5.90E13 x100=

RCS 1

=

x

+ EFPD/1140

+6.16E12 x100=

Sr-90 RHR 1

=

x

+, EFPD/1140

+ 6.16E12 x 100=

CV

+

1

=x

+ EFPD/1140

+6.16E12 xl00=

RCS

+

e- 002t

=

x

+ Complete 1.13E14 x 100 =

Ba-140 RHR e-002t

=

x

+

• Att. 8.7.5.4 1.13E14 x 100 =

CV

+

e-.002t

=

x

+

1.13E14 x100 =

REACTOR POWER LEVEL CONSTANT FOR <4 DAYS (LESS THAN 10% CHANGE)

NO t = (hours) time difference between sample time and reactor shutdown anal EFPD = effective full power days re 1 = (hours") decay constant (1) RNP-M/MECH-1742, Design Inputs for E-Plan Dose Assessment, Source Term Rad Monitor Information, Table 1 rE: Due to the long analysis time associated with Strontium ysis these columns will not be completed initially, but can be irded upon receiving sample results.

I Page 20 of 53 tD-T J.A-Rev. 4 I

TOTAL PERCENT RELEASED %

a-it Rev. 4 Page 20 of 53 1 I

ATTACHMENT 8.7.5.4 Page 3 of 7 WORK PACKAGE 4 - ASSESSMENT USING RADIONUCLIDE ANALYSIS REACTOR POWER CONSTANT LESS THAN 4 DAYS CORRECTION TERM WORKSHEET FOR Kr-87 Sum (Cl)

(C2)

(C3)

(C4)

(C5)

(C6)

Pj It tjI 1-exp (-Xt,)

exp(-Xti) I (Cl)*(C4)*(C5)

L

.3.

.4-4-

4 4

• 1-I I

I

+

-i I

L 4

4 I

i J.

I I

I I

I I

I J _____

£ __________

A Sum (C2) = Must be equal to 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />.

Correction Term = Sum (C6)/l 00 Pi = % Average reactor power for period j tj = (hours) Duration of operation at Pj tji = (hours) Duration from end of interval tj to reactor shutdown Kr-87 I = 0.545 hour0.00631 days <br />0.151 hours <br />9.011243e-4 weeks <br />2.073725e-4 months <br />s-'

EPTSC-07 Rev.4 Page 21 o 53]

ATTACHMENT 8.7.5.4 Page 4 of 7 WORK PACKAGE 4 - ASSESSMENT USING RADIONUCLIDE ANALYSIS REACTOR POWER CONSTANT LESS THAN 4 DAYS CORRECTION TERM WORKSHEET FOR Kr-88 (Cl)

(C2)

(C3)

(C4)

(C5)

(C6)

P!

tj ti, 1 -exp (-.Xt,)

exp(-Xtj,)

(Cl)*(C4)*(C5)

Sum (C2) = Must be equal to 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />.

Correction Term = Sum (C6)/1 00 P, = % Average reactor power for period j tj = (hours) Duration of operation at Pj tji = (hours) Duration from end of interval t, to reactor shutdown Kr-88 X = 0.244 hours0.00282 days <br />0.0678 hours <br />4.034392e-4 weeks <br />9.2842e-5 months <br />"'

I EPTSC-07 I

Rev. 4 1

Page 22 of 53:]

Sum

ATTACHMENT 8.7.5.4 Page 5 of 7 WORK PACKAGE 4 - ASSESSMENT USING RADIONUCLIDE ANALYSIS REACTOR POWER CONSTANT LESS THAN 4 DAYS CORRECTION TERM WORKSHEET FOR 1-131 (C1l

[C(2)

(0C3)

(C4)

(C5)

(C6)

Pj tj t,

1-exp (-Xt,)

exp(-Xtt,)

(Cl)*(C4)*(C5)

Sum (C2) = Must be equal to 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />.

Correction Term = Sum (C6)/i 00 Pj = % Average reactor power for period j tj= (hours) Duration of operation at Pi t= (hours) Duration from end of interval t, to reactor shutdown 1-131 X = 0.004 HOURS-'

EPTSC-07 Rev. 4 Page 23 of 53 Sum

ATTACHMENT 8.7.5.4 Page 6 of 7 WORK PACKAGE 4 - ASSESSMENT USING RADIONUCLIDE ANALYSIS REACTOR POWER CONSTANT LESS THAN 4 DAYS CORRECTION TERM WORKSHEET FOR Te-132

('Cl)

('C2)

('C3)

(C4)

(C5)

(C6)

Pj t,

tj, 1-exp (-.t,)

exp(-Xtf)

(Cl)*(C4)*(C5)

Sum (C2) = Must be equal to 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />.

Correction Term = Sum (C6)/100 Pj= % Average reactor power for period j tj= (hours) Duration of operation at Pi tji= (hours) Duration from end of interval tj to reactor shutdown Te-132 X = 0.009 HOURS-'

IEPTSC-07 I

Rev. 4 1

Page 24 of 53 Sum

ATTACHMENT 8.7.5.4 Page 7 of 7 WORK PACKAGE 4 - ASSESSMENT USING RADIONUCLIDE ANALYSIS REACTOR POWER CONSTANT LESS THAN 4 DAYS CORRECTION TERM WORKSHEET FOR Ba-140 (C1)

(C2)

(C3)

(C4)

(C5)

(C6)

P, tj tj, 1 -exp (-Mt) exp(-.Xtj,)

(Cl)*(C4)*(C5)

Sum (C2) = Must be equal to 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />.

Correction Term = Sum (C6)/100 Pj = % Average reactor power for period j tj= (hours) Duration of operation at Pj tj= (hours) Duration from end of interval tj to reactor shutdown Ba-1 40 X = 0.002 HOURS"1 IEPTSC-07 I

Rev. 4 1

Page 25 of 53 Sum

ATTACHMENT 8.7.5.5 Page 1 of 5 WORK PACKAGE 5 - HIGH LEVEL CORE DAMAGE ASSESSMENT

1.

Identify Current Plant Status.

a. Using the table and figures below, determine the possible status of the reactor core.

b. Go to the appropriate Work Package as indicated from the table.

High Level Core Damage Assessment Plant Status Fuel Rod Fission Product Status Core Exit Thermocouples LESS THAN 700°F No Core Damage; AND Continue to Monitor Containment Radiation Levels Plant Parameters LESS THAN Figure 1 mRad/hr Core Exit Thermocouples LESS THAN 2000°F Possible Fuel Rod AND Clad Damage Containment Radiation Go To Work Package 6 LESS THAN Figure 2 or Figure 3 mRad/hr Core Exit Thermocouples GREATER THAN 20000F Possible Fuel OR Overtemperature Damage Containment Radiation Go To Work Package 7 GREATER THAN Figure 2 or Figure 3 mRad/hr I EPTSC-07 Rev. 4 I

Page 26 of 53

ATTACHMENT 8.7.5.5 Page 2 of 5 WORK PACKAGE 5 - HIGH LEVEL CORE DAMAGE ASSESSMENT Figure 1 Dose Rate vs. Time -(No Fuel Damage)

I 000E+021 000 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 Time Since Shutdown (Hours)

Figure 2 - Detector R-32A Dose Rate vs.Time (1% Failed Fuel) 1 OOOE+07

-*1 000E+06 1 000E+05 1.000E+04 4 0 00 IEPTSC-07 I

Rev. 4 Page 27 of 53 E

500 1000 1500 2000 2500 3000 3500 4000 4500 5000 Time Since Shutdown (Hours)

ATTACHMENT 8.7.5.5 Page 3 of 5 WORK PACKAGE 5 - HIGH LEVEL CORE DAMAGE ASSESSMENT Figure 3 - Detector R-32B Dose Rate vs. Time (1% Failed Fuel)

I OOOE+04 4

000 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 Time Since Shutdown (Hours)

Figure 4 - Detector R-32A (100% Failed Fuel, NUREG-1 465 Short Term Release)

I (XOE+07 0

E 0 0 OE0 5000 IEPTSC-07 Rev.4 Page 28 of 53 a

S 0 a 000 500 10.00 1500 2000 2500 3000 3500 4000 4500 Time Since Shutdown (Hours)

I

ATTACHMENT 8.7.5.5 Page 4 of 5 Figure 5 - Detector R-32B (100% Failed Fuel, NUREG-1465 Short Term Release) 000 500 1000 1500 2000 2500 300D 3500 4000 4500 5000 Time Since Shutdown (Hours)

Figure 6 -Detector R-32A (100% Failed Fuel, NUREG-1465 Release)

I EPTSC-07 Rev. 4 Page 29 of 53 0 0 1 U

E vrI

!1 00M 500 1000D 1500 2000 2500 3000 3500 4000 4500 5000 Time Since Shutdown (Hours)

ATTACHMENT 8.7.5.5 Page 5 of 5 Figure 7 - Detector R-32B (100% Failed Fuel, NUREG-1465 Release) 0c e

0 C

000 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 Time Since Shutdown (Hours)

I EPTSC-07 Rev. 4 Page 30 of 53

ATTACHMENT 8.7.5.6 Page 1 of 2 WORK PACKAGE 6 - FUEL ROD CLAD DAMAGE 1.

Estimate Fuel Rod Clad Damage Based on Containment Radiation Levels.

a.

Find containment radiation level for 100% clad damage from Figure 4 or Figure 5 for the respective detector.

b.

Obtain current containment radiation level readings for R-32A or R-32B as applicable.

c.

Estimate clad damage using:

Current Containment Radiation Level

% Clad DamageCRM = Predicted Containment Radiation Level at 100% Clad Damage

2.

Estimate Fuel Rod Clad Damage Based on Core Exit Thermocouple Readings.

a. With RCS Pressure GREATER THAN 1600 psig:

% Clad Damage

':- Number of CETs > 1400YF Total Number of Operable CETs

b. With RCS Pressure LESS THAN 1600 psig:

Number of CETs > 1200'F Ca a

= Total Number of Operable CETs

3.

Confirm Reasonableness of Clad Damage Estimates.

a. Compare to expected response Containment Hydrogen Concentration LESS THAN 0.4% volume percent RVLIS LESS THAN 54% AND GREATER THAN 39%

Hot Leg RTD GREATER Tsar AND THAN LESS THAN 650'F Source Range Monitor GREATER THAN 1000 cps Difference in clad damage estimates from containment radiation and core exit thermocouples LESS THAN 50%, using ABSOLUTE VALUE

% Clad DamagecRM - % Clad DamagecE 1 AV

=%

Clad DamagecRM I EPTSC-07 Rev. 4 Page 31 of 53

ATTACHMENT 8.7.5.6 Page 2 of 2 WORK PACKAGE 6 - FUEL ROD CLAD DAMAGE

4.

If the expected response is not obtained, determine if the deviation can be explained from the accident progression:

Injection of water to the RCS Bleedpaths form the RCS Direct radiation to the containment monitors or from conservatisms in the prediction model:

fuel burnup fission product retention in the RCS fission product removal from containment.

5.

IF clad damage estimates have increased by more than 1% in the past 30 minutes OR estimates exceed 2% clad damage, THEN report the potential for an upgrade in classification to the Technical Analysis Director and the Plant Operations Director.

6.

Complete appropriate sections of Work Package 8.

IEPTSC-07 I

Rev.4 1

Page32of 3

ATTACHMENT 8.7.5.7 Page 1 of 4 WORK PACKAGE 7 - FUEL OVERTEMPERATURE DAMAGE CAUTION As many TC's as possible should be used for evaluation of core.temperature conditions. Caution must be used if a TC reads offscale - low or is reading considerably different from neighboring TC's as it may have failed.

1.

Estimate Fuel Overtemperature Damage Based on Containment Radiation Levels.

a.

Find containment radiation level for 100% core overtemperature damage from Figure 6 or Figure 7.

b.

Obtain current containment radiation level readings for R-32A or R-32B as applicable.

c.

Estimate overtemperature damage using:

Current Ctmt Radiation Level

% Core DamagecRM = Predicted Ctmt Radiation Level at 100% OvertempDamage

2.

Estimate Fuel Overtemperature Damage Based on Core Exit Thermocouple Readings.

a.

Obtain current core exit thermocouple temperature readings.

b.

Estimate overtemperature damage using:

% Core Damage czr Number of CETs > 20000 F Total Number of Operable CETs

3.

Confirm Reasonableness of Fuel Overtemperature Damage Estimates.

a. Compare to expected response RVLIS LESS THAN 39%

Hot Leg RTD GREATER THAN 650°F I EPTSC-07 Rev. 4 Page 33 of 53

ATTACHMENT 8.7.5.7 Page 2 of 4 WORK PACKAGE 7 - FUEL OVERTEMPERATURE DAMAGE Source Range Monitor GREATER THAN 1000 cps Difference fuel overtemperature estimates from containment radiation and core exit thermocouples LESS THAN 5 0%, using:

ABSOLUTE VALUE =

% Core Damage CRM - % Core Damage LE 1

% Core Damage CRM I

0 Containment hydrogen concentration NOTE:

Containment radiation monitor and core exit thermocouple estimates are not expected to deviate from hydrogen estimate by more than 25% damage.

4.

Obtain containment hydrogen concentration at 100% core overtemperature from table below:

Core Overtemperature Estimate Based on Containment Hydrogen Concentration RCS Pressure Water injection Predicted Containment to the RCS Hydrogen Concentration Less Than Yes Obtain H2 volume percent for 50% Zirc reaction line from CA-3, 1050 psig Figure 3-1 No Obtain H2 volume percent for 25% Zirc reaction line from CA-3, Figure 3-1 Greater Than Yes Obtain H2 volumes percent for 50% and 75%

Zirc reaction line from CA-3, Figure 3-1 and obtain average No Obtain H2 volume percent for 25% Zirc reaction line from CA-3, Figure 3-1 EPTSC-07 Rev. 4 Page 34 of 53

ATTACHMENT 8.7.5.7 Page 3 of 4 WORK PACKAGE 7 - FUEL OVERTEMPERATURE DAMAGE

5.

Estimate overtemperature damage using:

Current H2 Concentration.

% Core DamageHyd = Predicted H2 Concentration at 100% Overtemp Damage

6.

If expected response is not obtained, determine if the deviation can explained from the accident progression:

Injection of water to the RCS Bleed paths from the RCS Direct radiation to the containment radiation monitors Hydrogen burn in containment or operation of hydrogen igniters or from conservatisms in the predictive model:

fuel burnup fission product retention in the RCS fission product removal from containment.

7.

Report fuel overtemperature estimate to Technical Analysis Director and Plant Operations Director.

8.

Complete appropriate sections of Work Package 8.

I EPTSC-07 Rev. 4 Page 35 of 53]

ATTACHMENT 8.7.5.7 Page 4 of 4 R

P N

Quadrant 4 K

TCH

'U TOO TC 01 02 03 04 05 06 07 08 09 10 11 12 13 14 is TOSC TO

-I 4-

+ ---

I -

-f --

1-

-t I'-

K J

H G

F Cel E

D C

B Quadrant 1 II ci t

LRVUS SIB CBACA B

i-R TC TC TC TC TC SBA CIA

_-C-A

-A TC TC TC TC TC TC TC TC TC TC B

CBA C

-A SBA TC CBS TC TC TC TC BED CIA ODA Si N

M L

app

• ra Quadrant 2 E

D C

Tc B

A M

I A

i-ni-TC TO Quadrant 3 A

P EPTSC-07 Rev. 4 Page36of 3 T C CIS 01 02 03 04 05 06 07 08 09 10 11 12 13 14 15

-C ic Tc c

TC Ic TC TC TC TC TC TC CQ&

TC

ATTACHMENT 8.7.5.8 Page 1 of 11 WORK PACKAGE 8 -

SUMMARY

OF ASSESSMENTS

1.

Use this work package to prepare the Core Damage Assessment Summary Report (Page 2 and 3).

a.

Complete Sections 1, a, b, and c of this package.

b.

If available, convert results from Work Packages 2, 3, or 4, using this package.

2.

Complete all sections for which data is available on previous work packages.

3.

Transmit full or partial report to Technical Analysis Director and Plant Operations Director.

4.

Repeat this procedure as directed.

5.

Records of actions taken (work sheets and major communications) will be given to the Technical Analysis Director.

IEPTSC-07 I

Rev.4 1

Page37of 3

ATTACHMENT 8.7.5.8 Page 2 of 11 WORK PACKAGE 8 -

SUMMARY

OF ASSESSMENTS CORE DAMAGE ASSESSMENT

SUMMARY

REPORT

1.

The implementation of EPTSC-07 has been completed. Based on an analysis of specific indicators the core damage status is as follows:

Analytical Source A.

B.

C.

D.

E.

F.

NO CLAD OVER TEMP MELT CV CHRM R32 A or B Core Exit Thermocouples/RVLIS Hydrogen Levels Containment Atmosphere Reactor Coolant System RHR System

2.

Core Damage Assessment Definition - complete the appropriate section based on the above source and status.

Since it is probable that more than one type of damage may have occurred in the core, it is best to predict a range of estimated core damage.

I EPTSC-07 I

Rev. 4 I

Page 38 of 53:]

ATTACHMENT 8.7.5.8 Page 3 of 11 WORK PACKAGE 8 -

SUMMARY

OF ASSESSMENTS

a.

Core degradation indicated by CV High Range Monitors R-32ANB No damage

> RCS + Pre-existing Iodine Spike

> 60% Gap Release

> 5% Core Release

b.

Core degradation identified by core exit thermocouples (CET/RCS Pressure CET <700 OF CET <700°F - <1200-F CET >1200°F/RCS Pressure <1050 psig CET >1600°F/RCS Pressure >1050 psig CET >2300'F

c.

Percent zirconium-water reaction based on hydrogen Total Containment H2 cc Total RCS H 2 cc

% clad reacted (from CA-3, Figure 3-1)

d.

Percent core damage indicated by isotope release fractions.

CLAD FAILURE %

FUEL OVERTEMP. %

FUEL MELT %

I EPTSC-07 Rev. 4 Page 39 of 53 ISOTOPE Kr-87 Kr-88 1-131 Cs-1 34 Cs-1 37 Te-132 Sr-89 Ba-1 40

ATTACHMENT 8.7.5.8 Page 4 of 11 WORK PACKAGE 8 -

SUMMARY

OF ASSESSMENTS Comments:

Initiated By:

Reviewed By:

Approved By:

Technical Analysis Director Plant Operations Director Date:

Time:

Date:

Date:

I EPTSC-07 Rev. 4 Page 40 of 53

ATTACHMENT 8.7.5.8 Page 5 of 11 WORK PACKAGE 8 -

SUMMARY

OF ASSESSMENTS RELATIONSHIP OF % CLAD DAMAGE WITH

% CORE INVENTORY RELEASED OF Kr-87, Kr-88 Core Inventory Released (%)

10" 10-2 LOL I0-3 ID-4 Io-5 10-1 101 102 Clad Damage (r/)

I EPTSC-07 I

Rev.4 Page41 of 3

ATTACHMENT 8.7.5.8 Page 6 of 11 WORK PACKAGE 8 -

SUMMARY

OF ASSESSMENTS RELATIONSHIP OF % CLAD DAMAGE WITH

% CORE INVENTORY RELEASE OF 1-131 Reeased 10-1 10-2 I0-3 10-,

10-5 10" 100 10D 102 Clad: Damage (%)

IEPTSC-07 I

Rev.4 Page42of 53

ATTACHMENT 8.7.5.8 Page 7 of 11 WORK PACKAGE 8 -

SUMMARY

OF ASSESSMENTS RELATIONSHIP OF % CLAD DAMAGE WITH

% CORE INVENTORY RELEASE OF 1-131 WITH SPIKING Core Inventory Released (%)

100 L-100 101 Clad Damage (%)

IEPTSC-07 I

Rev. 4 Page43of53 10-1 10-2 10-3 10-4 10"5 10-"

102

ATTACHMENT 8.7.5.8 Page 8 of 11 WORK PACKAGE 8 -

SUMMARY

OF ASSESSMENTS RELATIONSHIP OF % FUEL OVERTEMPERATURE WITH % CORE INVENTORY RELEASED OF Kr, I, Cs, OR Te Core Inventory Released (%)

102 10' 100 1

,1I I..

100 101 102 Fuel Over Temperahbwe (%)

1EPTSC-07 Rev. 4 Page 44 of 53

ATTACHMENT 8.7.5.8 Page 9 of 11 WORK PACKAGE 8 -

SUMMARY

OF ASSESSMENTS RELATIONSHIP OF % FUEL OVERTEMPERATURE WITH % CORE INVENTORY RELEASED OF Ba OR Sr 10 Fuel Over Temperature (%)

IEPTSC-07 Rev. 4 Page 45 of 53 1.E+O0 1.E-01 1.E-02 U) cc 0

1.E-03 1.E-04 100

ATTACHMENT 8.7.5.8 Page 10 of 11 WORK PACKAGE 8 -

SUMMARY

OF ASSESSMENTS RELATIONSHIP OF % FUEL MELT WITH

% CORE INVENTORY RELEASED OF Kr, I, Cs, OR Te Core Inventory Released (%)

102 101 100 I C).

101 102 Fuel Malt (%)

IEPTSC-07 I

Rev. 4 P Pageo46of 3 100

ATTACHMENT 8.7.5.8 Page 11 of 11 WORK PACKAGE 8 -

SUMMARY

OF ASSESSMENTS RELATIONSHIP OF % FUEL MELT WITH

% CORE INVENTORY RELEASED OF Ba OR Sr Core Inventory Released (%)

102 101 100 10-1 10-2 10' 102 Fuel Melt (%)

I EPTSC-07 Rev. 4 Page 47 of 53

(

ATTACHMENT 8.7.5.9 Page 1 of 1 CHARACTERISTICS OF CATEGORIES OF FUEL DAMAGE Fission Product Concentrations:

Kr-87 (Kr)

Xe-131 (Xe)

Xe-133 1-131 (I) 1-132 1-133 1-135 (Cs)

(Sr)

(Ba)

(Te)

(Pr)

Activity Ratios:

Kr-87/Xe-1 33 1-132/1-131 1-133/1-131 1-135/1-131 Core Exit Thermocouples*

RVLIS Indication:

Core Uncovery Begins Core Completely Uncovered Containment Hydrogen Concentration:

Containment Radiation Monitors*

No Fuel Dama-ge

<4E-5%

<4E-3%

<2E-3%

<3E-4%

<5E-5%

<1 E-4%

<8E-5%

<750%

>58%

Negligible

Background

0-50%

Cladding Failure

-0.005%

-0.08%

-005%

-0.01%

-0.02%

-0.05%

-0.03%

0.022 0.17 0.71 0.39 1300°F

<58%

<10%

<1500 R/hr 50-100%

Cladding Failure

-0.01%

-0.1%

-0.1%

-0.3%

-0.03%

-0.1%

-0.05%

0.022 0.17 0.71 0.39 1650°F

<58%

20%

3000 Page 48 of 53 EPTSC-07 Rev. 4 C

0-50%

Fuel OverTemp 1-25%

1-25%

1-25%

1-25%

.001-.08%

.001-08%

0.22 1.5 2.1 1.9

>1650'F

<58%

20%

2.7E5 50-100%

Fuel OverTemp

-50%

-50%

-50%

-50%

-0.15%

-0.15%

0.22 1.5 2.1 1.9

>1650' F

<58%

20%

5.3E5 0-50%

Core Melt

-50%

-50%

-50%

-50%

-15%

-15%

1-50%

0.1-0.8%

0.22 1.5 2.1 1.9

>1650TF

<58%

<41%

20%

7.8E5 50-100%

Core Melt

>50%

>50%

>50%

>50%

>15%

>15%

>50%

>0.8%

0.22 1.5 2.1 1.9

>1650°F

<58%

<25%

20%

>7.8E5 I

Page 48 of 53 I

ATTACHMENT 8.7.5.10 Page 1 of 2 CORE PROTECTION 1.

Critical Safety Function Status Tree (CSFST's)

a.

Subcriticality (FRP-S)

Verify Automatic Actions or Perform Manual Actions to Reduce Core Power Emergency Borate Check for Possible Sources of Positive Reactivity and Eliminate them Verify Subcriticality

b.

Core Cooling (FRP-C)

Establish Safety Injection Flow to the RCS Rapidly Depressurize SGs to Depressurize RCS Start RCPs and Open All RCS Vent Paths to Containment

c.

Heat Sink (FRP-H)

Attempt Restoration of Feed Flow to Steam Generators Initiation of RCS Bleed and Feed Heat Removal Restore and Verify Secondary Heat Sink Termination of RCS Bleed and Feed heat Removal

d.

Integrity (FRP-P)

Stop RCS Cooldown Terminate SI if Criteria Satisfied Depressurize RCS to Minimize Pressure Stress Establish Normal Operating Conditions and Stable RCS Conditions Soak if Necessary Prior to Further Restricted Cooldown

e.

Containment (FRP-J)

Verify Containment Isolation and Heat Removal Check for and Isolate a Faulted Steam Generator Check for Excessive Containment Hydrogen and Determine Appropriate Action

f.

Inventory (FRP-I)

Establish Charging and Letdown Reduce PZR Pressure Energize PZR Heaters and Control Charging and Letdown to Draw a Bubble I EPTSC-07 I

Rev. 4 Page 49 of 53

ATTACHMENT 8.7.5.10 Page 2 of 2 CORE PROTECTION

2.

DBA Large Break LOCA (1 ft2 total area up to double-ended break)

a.

Blowdown Reactor Trip Signal and SI Signal in about 1 second 2235 psig to atmosphere in about 24 secon'ds Break flow 70,000 Ibm/sec to zero by end of blowdown Sl accumulator flow initiates at 600 psig in about 16 seconds

b.

Refill 2000 ppm water from RWST injected into RCS cold legs

c.

Reflood Bottom of Core (BOC) recovery time about 45 seconds Accumulator empties at about 58 seconds

d.

Long Term Recirculation RHR pumps transferred when RWST level reach switchover setpoint Cooling water backflushed from Containment Sump to the hot legs Core maintained in shutdown state by borated water

3.

Possible Consequences of the DBA

a.

Cladding Failure

b.

Fuel Overtemperature

c.

Core Melt

4.

Instrument Errors/Malfunctions

a.

G-M tube saturation

b.

Steam voids

c.

Flooding of RTD connection blocks IEPTSC-07 I

Rev.4 I

Page 0of 3

ATTACHMENT 8.7.5.11 Page 1 of 2 FUNCTION RESTORATION PROCEDURES

1.

Subcriticality (FRP-S)

a.

S.1 Response to Nuclear Power Generation/ATWS

b.

S.2 Response to Loss of Core Shutdown'

2.

Core Cooling (FRP-C)

C.1 C.2 C.3 Response to Inadequate Core Cooling Response to Degraded Core Cooling Response to Saturated Core Cooling

3.

Heat Sink (FRP-H)

H.1 H.2 H.3 H.4 H.5 Response to Loss of Secondary Heat Sink Response to Steam Generator Overpressure Response to Steam Generator High Level Response to Loss of Normal Steam Release Capability Response to Steam Generator Low Level

4.

Integrity (FRP-P)

a.

P.1 Response to Imminent Pressurized Thermal Shock

b.

P.2 Response to Anticipated Pressurized Thermal Shock

5.

Containment (FRP-J)

J.1 J.2 J.3 Response to High Containment Pressure Response to Containment Flooding Response to High Containment Radiation Level

6.

Inventory (FRP-1) 1.1 1.2 1.3 Response to High Pressurizer Level Response to Low Pressurizer Level Response to Voids in Reactor Vessel I EPTSC-07 I

Rev.4 Page 51 of 53

a.

b.

C.

a.

b.

C.

d.

e.

a.

b.

C.

a.

b.

C.

AOP-OC AOP-OC AOP-OC AOP-O0 AOP-01 AOP-01 AOP-01 AOP-01 AOP-02 AOP-02 AOP-02 AOP-02 AOP-O2 AOP-03 AOP-03 APP-00 APP-0C APP-DC APP-0C APP-0C ATTACHMENT 8.7.5.11 Page 2 of 2 FUNCTION RESTORATION PROCEDURES ABNORMAL OPERATING PROCEDURES

)1 Malfunction of Reactor Control System

)3 Malfunction of Reactor Make-up Control

)4 Control Room Inaccessibility

)5 Radiation Monitoring System 0

Main Feedwater/Condensate Malfunction 3

Fuel Handling Accident 8

Reactor Coolant Pump Abnormal Conditions 9

Malfunction of RCS Pressure Control 20 Loss of Residual heat Removal (Shutdown Cooling)

_1 Seismic Disturbances 23 Loss of Containment Integrity 24 Loss of Instrument Bus 28 ISFSI Abnormal Events 33 Shutdown LOCA 35 S/G Tube Leak ANNUNCIATOR PANEL PROCEDURES

)1 Miscellaneous NSSS

)2 Engineering Safeguards

)3 RCS & Makeup Systems

)4 First Out Reactor Trips

)5 NIS & Reactor Control I EPTSC-07 Rev. 4 Page 52 of 53

ATTACHMENT 8.7.5.12 Page 1 of 1 DEFINITIONS

1.

Gap Activity-volatile fission products (noble gases, halogens, cesiums) produced during operation which migrate into the gap region of the fuel pin.

2.

Clad Damage-structural deformation of the zirconium clad housing the UO2 fuel allowing the escape of fission products to the reactor coolant. Usually predominate with core temperatures > 1300'F - 2000'F.

3.

Fuel Overtemperature-refers to the release of fission products from the grain boundary during fuel overtemperature conditions >2000 0 F - 34500 F.

4.

Fuel Melt-refers to fission product release from the fuel associated with melting temperatures >3450 OF.

5.

Spiking Phenomena Spiking is an increase in the normal primary coolant iodine activity due to Reactor Coolant System pressure, temperature or power transients, where in fact no clad damage has occurred.

6 Oxyvqen Concentration in the Containment A decrease in oxygen concentration may indicate a hydrogen burn has occurred. This should be considered during the evaluation of percent containment hydrogen.

7.

Steam Generator Tube Rupture or Outside Containment Loss of Coolant If core activity has been released to other systems (e.g. secondary system, component cooling water), this procedure will not accurately reflect actual core damage.

This will be identified by auxiliary methods which estimate more severe damage than the isotopic analysis. If accurate samples of these systems are available as well as reasonable estimates of the sample space volume or mass, the methods in this procedure may be applied to improve the accuracy of the nuclide release estimate of core damage.

I EPTSC-07 I

Rev.4 I

Page of 53