ML031570202
| ML031570202 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 05/19/2003 |
| From: | Izyk C Entergy Nuclear Northeast |
| To: | NRC/Document Processing Center |
| References | |
| Download: ML031570202 (184) | |
Text
ENTERGY NUCLEAR NORTHEAST.,~
JAMESA FITZPATRICK NUCLEAR POWERPLANT
- 11.
O
,N 13 9 h.fDOCUMENT TANSMIALAND RECEIPT ACKNOWLEDGEMENT FORM DATE: May 19, 2003 CONTROLLED COPY NUMBER: 34 TO:
FROM:
U.S.N.R.C. Document Center/Washington, DC CATHY IZYK - EMERGENCY PLANNING DEPARTMENT
SUBJECT:
FMERGENCY PLAN AND IMPLEMENTING PROCEDURFS Enclosed are revisions to your assigned copy of the JAFNPP Emergency Plan and Implementing Procedures.
Please remove and DISCARD the old pages. Insert the attached, initial and date this routing sheet and return the completed routing sheet to Cathy Izyk in the Emergency Planning Department within 15 days.
If this transmittal is not returned within 15 days, your name will be removed from the controlled list.
__'>,, 'l_______________9i4
^
VOLUME I Upd t i Dated M A 30. 2003
_________5; DOCUMENT PAGES REV. #
INITIALS/DATE APPENDIX C REPLACE ALL 26 SECTION 5 REPLACE ALL 38
________________0;,
6,,
>tiS S1>i VOLUM E U date List Date dM A 30 003
__ ____5
_i _- L__
DOCUMENT PAGES REV. #
INITIALS/DATE EAP-4 REPLACE ALL AND PLACE 3 COLORED FLOW CHARTS IN THE 32 PROCEDURE BY PAGE NUMBER EAP-4.1 REPLACE ALL 16
°
- 'VOLUME 3 Update List Dated MAY 30. 2003 7_jE____i DOCUMENT PAGES REV.#
INITIALS/DATE EAP-44 REPLACE ALL 5
SAP-20 REPLACE ALL 22
EMERGENCY PLAN / VOLUME 1 UPDATE LIST I CONTROLLED COPY #A I
Date of Issue:
MAY 30, 2003 Procedure,-
Procedure
-.Revision DateofLast Number Title
'Number Review N/A TABLE OF CONTENTS REV. 23 05/03 SECTION 1 DEFINITIONS/ACRONYMS REV. 20 05/03 SECTION 2 SCOPE AND APPLICABILITY REV.
19 05/03 SECTION 3
SUMMARY
OF THE JAFNPP EMERGENCY PLAN REV.
10 05/03 SECTION 4 EMERGENCY CONDITIONS REV.
19 05/03 SECTION 5 ORGANIZATION REV. 38 05/03 SECTION 6 EMERGENCY MEASURES REV. 25 05/03 SECTION 7 EMERGENCY FACILITIES AND EQUIPMENT REV. 25 05/03 SECTION 8 MAINTAINING EMERGENCY PREPAREDNESS REV. 25 05/03 SECTION 9 RECOVERY REV.
17 05/03 APPENDIX A EMERGENCY PLAN IMPLEMENTING REV.
18 05/03 APPENIX A PROCEDURES REV._IS_05_0 APPENDIX B NYPA POLICY STATEMENT REV.
6 05/03 APPENDIX C LETTERS OF AGREEMENT REV. 26 05/03 APPENDIX D NEW YORK STATE PLAN AND PROCEDURES REV.
6 05/03 APPENDIX E OSWEGO COUNTY PLANS AND PROCEDURES REV.
6 05/03 APPENDIX F TYPICAL SUPPORT COMPANIES AND REV.
12 05/03 ORGANIZATIONS APPENDIX G DELETED (2/98) l APPENDIX H PUBLIC INFORMATION PROGRAM REV. 28 05/03 APPENDIX I EMERGENCY EQUIPMENT KITS REV.
10 05/03 APPENDIX J SUPPORTING DOCUMENTS REV.
9 05/03 APPENDIX K EVACUATION TRAVEL TIME ESTIMATES AND REV.
7 05/03 POPULATION DISTRIBUTION FOR THE JAF/NINE MILE POINT EMERGENCY PLANNING ZONE APPENDIX L NUREG-0654/FEMA-REP-1 CROSS REFERENCE REV. 12 05/03 APPENDIX M DELETED (5/84)
APPENDIX N TYPICAL FEDERAL SUPPORT RESOURCES REV. 13 05/03 Page 1 of 1
ENTERGY NUCLEAR OPERATIONS, INC.
JAMES A. FITZPATRICK NUCLEAR POWER PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE DOSE ASSESSMENT CALCULATIONS EAP-4 REVISION 32 APPROVED BY:
DATE:
5-17 :
RESPONSIBLE PROCEDURE OWNER EFFECTIVE DATE:
7
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--~C -
-I FIRST ISSUE FULL REVISION LIMITED REVISION E PERIODIC REVIEW DUE DATE:
INFORMATIONAL USE TSR I
ADMINISTRATIVE
~CONTROLLED COPY #
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MAY 2008
DOSE ASSESSMENT CALCULATIONS EAP-4 REVISION
SUMMARY
SHEET REV. NO.
32 Add "onsite and in-plant" to section 4.2 for clarity
- Added met data access method for simulator
- Incorporated several notes into procedure steps in section 5.2.3.Q and 5.2.3.A
- Add direction to section 5.2.3.Q regarding ERPAs for evacuation
- utilizing both the model and the flowchart, and the concern for lake/land breeze considerations.
- Change the title of section 5.4 to read USING MONITOR READINGS TO ESTIMATE WHEN A PAR WILL BE REACHED BASED ON PROJECTED DOSE
- Added direction in 5.4 for use of TEDE dose.
- Added note in section 5.4 to reference attachment 7 for calculations.
- Incorporated cautions into boxes and added wording regarding notifying offsite agencies within 15 minutes of PAR changes.
- Added Section 5.5.
- Modified Attachment 1 and 2 to add a caution in the GE section for 15 minute notification requirements.
- Add caution for lake/land breeze on Attachment 2.
- Updated Attachment 4 with current 2001 population estimates ERPA'S
- Change to Attachment 5 to add lines to the 3-inch section for data recording.
- Added attachment 6.
- Add attachment 7.
- Added attachment 8.
3 1
- In the table of section 5.2.1 and 5.2.3.F added "Direct Connect OR".
- Added note in section 5.2.3.N.3.
- Added section 5.4 -
Monitor reading estimate.
30 Added 5.5.2 D, to turn on modem.
Rev. No.
32 Page 2
of 26
DOSE ASSESSMENT CALCULATIONS EAP-4 TABLE OF CONTENTS SECTI 1.0 2.0 2.1 2.2 3.0 4.0 5.0 6.0
- ON PAGE PURPOSE................................................. a REFERENCES............................................. 4 Performance References................................. 4 Developmental References............................... 4 INITIATING EVENTS...................................... 4 RESPONSIBILITIES....................................... 4 PROCEDURE.............................................. 5 ATTACHMENTS........................................... 16
- 1. INITIAL PROTECTIVE ACTIONS............................ 17
- 2. AUGMENTED DOSE ASSESSMENT AND PROTECTIVE ACTIONS......
18
- 3. ANALYZED ACCIDENT TYPES............................... 19
- 4. 2001 POPULATION ESTIMATES EMERGENCY RESPONSE PLANNING AREAS................................................. 20
- 5. DOWNWIND SURVEY WORKSHEET............................. 21
- 6. DOSE ASSESSMENT ACCIDENT TYPE SELECTION FLOWCHART.....
22
- 7. CALCULATION FORM WHEN USING MONITOR READINGS TO ESTIMATE WHEN A PAR WILL BE REACHED BASED ON PROJECTED DOSE......................................... 23
- 8. SOURCE TERM ENTRY FOR FIRST 15 MINUTE TIME STEP OF A REFUEL ACCIDENT WHEN BUILDING ISOLATON HAS OCCURRED AND THE RELEASE IS THROUGH THE STACK......... 24
- 9. SOURCE TERM FOR A REFUEL ACCIDENT THAT RESULTS IN AN UNFILTERED RELEASE PATHWAY............................ 26 Page 3
of 26 Rev. No.- 32
DOSE ASSESSMENT CALCULATIONS EAP-4 1.0 PURPOSE To provide the methods for performing dose assessment and determining protective actions during accident conditions at James A. FitzPatrick Nuclear Power Plant.
2.0 REFERENCES
2.1 Performance References 2.1.1 EAP-4.1, RELEASE RATE DETERMINATION 2.1.2 EAP-5.3, ONSITE/OFFSITE DOWNWIND SURVEYS AND ENVIRONMENTAL MONITORING 2.1.3 EAP-42, OBTAINING METEOROLOGICAL DATA 2.2 Developmental References 2.2.1 EAP-4.1, RELEASE RATE DETERMINATION 2.2.2 EAP-5.3, ONSITE/OFFSITE DOWNWIND SURVEYS AND ENVIRONMENTAL MONITORING 2.2.3 EAP-42, OBTAINING METEOROLOGICAL DATA 3.0 INITIATING EVENTS 3.1 A General Emergency has been declared OR 3.2 A vented gaseous release exceeds alarm setpoints OR 3.3 An unmonitored gaseous release is suspected or underway.
4.0 RESPONSIBILITIES 4.1 Shift Manager/Emergency Director (SM/ED)
The SM/ED is responsible for ensuring that Protective Action Recommendations (PARs) are developed in accordance with this procedure.
Rev. No.
32 Page 4 of 26
DOSE ASSESSMENT CALCULATIONS EAP-4 4.2 TSC Rad Support Coordinator (TSC RSC)
The TSC RSC is responsible to the Emergency Director for managing the radiological monitoring and assessment aspects on-site and in-plant during an emergency and of those functions specified in Step 4.3 until relieved of those functions by the EOF.
4.3 EOF Rad Support Coordinator (EOF/RSC)
The EOF/RSC is responsible to the Emergency Director for managing the radiological monitoring and assessment aspects offsite during an emergency.
4.4 Dose Assessment Coordinator (DAC)
The DAC is responsible for managing the offsite dose aspects of an emergency, in order to assess the radiological consequences to the public.
4.5 Chemistry Technician/Rad Protection Technician The on-shift chemistry technician is responsible to the Emergency Director for conducting dose assessment from the control room and assisting the SM/ED with information related to offsite notification and protective action recommendations.
The on-shift radiation protection technician is responsible to the SM/ED for conducting surveys as directed.
5.0 PROCEDURE 5.1 Control Room Dose Assessment and Protective Action Recommendations Utilize Attachment 1, Initial Protective Actions, for control room dose assessment and protective action recommendations.
Rev. No.
32 Page 5 of 26
DOSE ASSESSMENT CALCULATIONS 5.2 Augmented Dose Assessment Dose projection shall be completed using the EDAMS computer located in the Control Room, Technical Support Center or Emergency Operations Facility as follows:
5.2.1 General Information A. Locations EDAMS software and follows:
hardware is located as LOCATION HARDWARE PRINTERS METE DATA CONNECTION Control Room Personal HP LaserJet Direct Connect Computer and Monitor Technical Personal Seiko D-Scan, Direct Connect Support Center Computer and HP LaserJet Monitor Emergency Operations Facility 1 Personal Computer Monitor and Seiko D-Scan, HP LaserJet Direct Connect OR Dial-up Modem Emergency Personal HP LaserJet Direct Connect Operations Computer and OR Dial-up Facility #2 Monitor Modem B. Computer problems IF at any time problems are experienced with the computer, MOVE to another location that has the EDAMS software and continue.
C. The dose assessment program is called RADDOSE V and is part of the EDAMS package.
D. Meteorological data is automatically sent (via direct connect or modem) to RADDOSE V and EDAMS by the Meteorological Monitoring System (MMS).
The user can use this data or manually input meteorological data.
E. Source term and release rate determination is discussed in procedure EAP-4.1.
F. Software documentation is available for the EDAMS code and is maintained by the Emergency Planning Coordinator.
Page 6
of 26 EAP-4 Rev. No.
3 2
DOSE ASSESSMENT CALCULATIONS EAP-4 s~.
5.2.2 EDAMS Dose Model Limitations A. The EDAMS menu from the EDAMS icon only allows the operation of one DOS application at a time.
B. Dose rates and deposition rates reported by the model are the maximum for the sector; not necessarily the dose rate or deposition rate at the center of the sector. This avoids the situation of a narrow (stable) plume slipping between receptor points and being missed.
C. Deposition data reported is not intended for an environmental evaluation; its intent is to indicate areas of potentially high ground level concentrations.
D. Forecast mode results may at times exceed real-time results; this is due to the forecast mode having a greater internal time step.
E. A calculation limitation of the dose assessment model occurs when an extreme wind (direction) shift takes place.
The model may not calculate doses in sectors that the plume skips over entirely within a single 15 minute interval advection step.
5.2.3 Dose Assessment Using EDAMS Computer NOTE:
The dose assessment model has many capabilities beyond those used in this procedure.
Utilize the "EDAMS Operators Manual" for further reference.
CAUTION:
Protective Action Recommendations (PARs) must be transmitted to the State and County within 15 minutes of declaring an Emergency, changing emergency classification, or changing Protective Action Recommendations.
A. IF during the course of dose assessment, the dose to the population is projected to exceed 1 Rem TEDE or 5 Rem CDE Thyroid, THEN immediately advise the ED that the General Emergency criteria has been met.
Rev. No.
32 Page 7 of 26
DOSE ASSESSMENT CALCULATIONS--
EAP-4 B. Use Attachment 2 Augmented Dose Assessment Protective Actions, for guidance when performing dose assessment activities at the TSC or EOF.
C. Ensure that the black switch on the CR and TSC meteorological panels is positioned to the Niagara Mohawk (B) position.
D. Energize the EDAMS computer power strip to provide power to the computer, monitor, and printer.
E. Ensure the modem is ON (powered on).
F. Select the "Login" icon from the EDAMS icons and select "Continue" at the plant picture screen.
G. Select the appropriate menu item based on your location as follows:
Location:
Menu Choice:
CR Direct connect to Met Data TSC Direct connect to Met Data EOF Direct connect OR Automatic Dial-in to Met Data Simulator Automatic Dial-in to Met Data H. When the login routine finishes, close the login window screen by selecting "OK".
I. From the EDAMS icons, select JAF Raddose-V.
J. Select "Continue" at the plant picture screen.
K. From the Raddose-V start up menu, select "Begin New Incident."
Rev. No.
32 Page 8 of 26
DOSE ASSESSMENT CALCULATIONS EAP-4
.~~~~~~
-1 Rev. No.
32 L. At the Raddose-V Accident Scenario Definition screen, enter the following information:
- 1. Reactor Trip Date -
date the reactor was scrammed or shutdown.
- 2.
Reactor Trip Time (24-hour format) -
time the reactor was scrammed or shutdown.
- 3.
Release Date -
date the release to atmosphere began, or is projected to begin.
- 4.
Release Time (24-hour format) -
time the release to atmosphere began, or is projected to begin.
- 5.
Lake Surface Temp (Degrees F) -
Enter the known lake surface temperature, or use the historical default value provided.
- 6.
Operator Initials -
Enter 2 or 3 initials, then press ENTER.
- 7.
Select "Accept" to accept and continue.
M. At the Raddose-V main menu, select "Enter/Edit Source Term Data."
N. At the Raddose-V Source Term Data Entry Screen, proceed as follows:
- 1. Utilize Attachment 6, DOSE ASSESSMENT ACCIDENT TYPE SELECTION FLOWCHART as a guide to determine the most appropriate accident type.
- 2.
Select "Accident Type" by pressing the "F2" key, or by using the mouse, then choosing the accident type which most closely matches current conditions.
Your selection determines which default isotopic mix is used for upcoming calculations (refer to, Analyzed Accident Types).
Page 9 of 26
DOSE ASSESSMENT CALCULATIONS EAP-4
- 3. When asked, "Is this release Elevated?"
select "Yes" for elevated releases or "No" for ground releases.
(A stack release is elevated; all other releases are ground releases.)
NOTE:
Back calculation cannot be used on first time step.
- 4. Select the "Method" used to determine the release rate by pressing the "F2" key, or by using the mouse, then choosing the appropriate method based on available information.
- 5. Select the Iodine release rate "Method" by hitting the "F2" key, or by using the mouse. Enter the "Monitor Reading" and "Release Rate" if required.
- 6. Up to three Accident Types (three release paths) can be entered by using the down arrow key (9) to select type 2 and 3.
- 7. When the source term data entry screen has been completed, select "Accept" to accept data and return to the Raddose-V main menu.
- 0. At the Raddose-V main menu, the menu bar will highlight the appropriate elevated and/or ground meteorological data choices based on your input in step M.2.
- 1. If direct met data input is being used, the appropriate ground and/or elevated met data will automatically be displayed for the current time step.
- 3. IF met data is not available via the MMS, THEN enter met data obtained from alternate sources, as outlined in EAP-42, OBTAINING METEOROLOGICAL DATA.
Rev. No.
32 Page 10 of 26 f
I
DOSE ASSESSMENT CALCULATIONS EAP-4 NOTE: To determine an estimated monitor reading to reach a PAG, refer to step 5.4 P. At the Raddose-V main menu, select "Perform Calculations."
- 1. The map of the 10 mile Emergency Planning Zone (EPZ) will appear with centerline dose rates after the model has calculated the actual model doses.
NOTE:
This data SHOULD NOT be used for PARs.
PARs should be based on forecast data, which will be the dose to be avoided by the protective action.
- 2. Select "Continue" to continue.
- 3. At the Raddose-V output menu, select "Continue Calculations".
- 4. At the Raddose-V main menu, select "Perform a Forecast".
NOTE:
A new time step must be added to perform a forecast.
- 5. Verify meteorology and source term data as required.
Select "Accept" to accept.
- 6. Enter "Forecast Period" (i.e. release duration).
Use 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> as a default value.
Select "OK".
- 7. When asked "Has a General Emergency been declared?" enter "Yes" or "No".
Select "Continue" to continue.
- 9. Select "Go to Report Menu".
- 10. Select "Print Complete Dose/Dose Rate Report".
Rev. No.
32 Page 11 of 26
DOSE ASSESSMENT CALCULATIONS EAP-4 NOTE:
County and State Protective Action Recommendations (PARs) take many factors into account that NMP/JAF procedures do not (i.e.,
road conditions, special population needs, Evacuation Travel Time Estimates, evacuation scenarios, and shelter vs. evacuation doses).
Therefore, differences in PARs may occur.
The Rad Support Coordinator must account for differences in Protective Action Recommendations when those exist.
PARs should not be modified to match County or State PARs without justification.
CAUTION:
Protective Action Recommendations (PARs) must be transmitted to the State and County within 15 minutes of declaring an Emergency, changing emergency classification, or changing Protective Action Recommendations.
Q.
Protective Action Recommendations (PARs)
- 1.
Since the Nine Mile Point/J.A. FitzPatrick (NMP/JAF) Site is contained in ERPA 1, any recommendation made for ERPA 1 must also apply to all NMP/JAF personnel not required to be onsite for the emergency.
- 2.
The RADDOSE model factors in meteorological conditions such as lake/land breeze that the Attachment 2 flowchart does not.
Therefore, the model needs to be considered even during situations where no release is occurring.
- 3.
The EDAMS RADDOSE V program will recommend PARs for each ERPA, based upon the dose assessment (in forecast mode).
- a. Ensure that both the RADDOSE ERPAs recommended for Evacuation, and the ERPAs recommended for evacuation are considered when developing PARs.
Rev. No.
32 Page 12 of 26 I
DOSE ASSESSMENT CALCULATIONS EAP-4
- b. For ERPAs recommended by RADDOSE for evacuation, ensure that both Dose PARs and Plant PARs are included.
- 4. Initiate or revise PARs based upon this recommendation (and previous recommendations, if made).
R. Notification
- 1. Record the revised PAR for each ERPA on the Part 1 Notification Form (EAP-1.1, ) and give to the ED for approval.
S. Update the RADDOSE-V Model at 15 minute intervals or as directed by the Dose Assessment Coordinator.
5.3 Downwind Survey Dose Estimates 5.3.1 Use Attachment 5, Downwind Survey Worksheet, to record field data transmitted to the dispatch center.
5.3.2 Projected Deep Dose Equivalent (DDE) is approximately equal to TEDE Projected Dose A. Use field information recorded on the Downwind Survey Worksheet, Attachment 5, to perform projected dose calculations.
B. Obtain the estimated duration of release from the Emergency Director and record it on. If it is unknown, use 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> as a first estimate.
C. Complete the calculations, as shown, on to determine DDE projected doses for each sampling location.
Rev. No.
32 Page 13 of 26
DOSE ASSESSMENT CALCULATIONS EAP-4 5.3.3 CDE Thyroid Projected Dose A. Use field information recorded on the Downwind Survey Worksheet, Attachment 5, to perform CDE Thyroid Projected Dose.
B. Obtain the estimated duration of release from the Emergency Director and record it on.
C. Calculate the I-131 concentration in accordance with Attachment 5.
D. Complete the calculations, as shown, on to determine thyroid rate and CDE thyroid projected doses for each location for which data has been recorded.
5.3.4 Results A. Provide the results of the DDE and CDE thyroid projected dose calculations from steps 5.3.2 and 5.3.3 to the individual tasked with calculating protective action recommendations.
5.4 Using Monitor Readings to Estimate When a Par Will be Reached Based on Projected Dose NOTE: may be utilized to perform the following calculations:
5.4.1 To determine the estimated monitor reading needed to reach a Child Thyroid dose of Srem CDE, perform the following:
A. Perform a forecast using EDAMS; B. Divide the Child Thyroid CDE dose (forecasted),
by the release point monitor reading(mr/hr,cpm,cps);
C. Divide the result from 5.4.1.B into 5 rem. This result becomes the estimated monitor reading corresponding to 5rem CDE for the forecast duration, based on the inputted flow rate, monitor reading and meteorology at the time the data was taken.
Rev. No.
32 Page 14 of 26
- -1
DOSE ASSESSMENT CALCULATIONS EAP-4 D. To determine the estimated monitor reading needed to reach a TEDE dose of 1 rem, perform the following:
- 1. Perform a forecast using EDAMS;
- 2. Divide the TEDE dose (forecasted), by the release point monitor reading (mr/hr,cpm,cps);
- 3. Divide the result from 5.4.2.B into 1 rem.
This result becomes the estimated monitori-reading corresponding to lrem CDE for the forecast duration, based on the inputted flow rate, monitor reading and meteorology at the time the data was taken.
5.5 Protective Action Recommendations (PARs) Beyond 10 Miles 5.5.1 If projected doses exceed the following values at 10 miles; EPA Protective Action Guidelines TEDE (Rem)
CDEt (Rem)
>1
>5 Then PARs need to be developed beyond the 10 mile EPZ.
PARs can be developed using the EDAMs routine that calculates Dose Rates at a point of interest.
5.5.2 Chose several points of interest that will encompass the postulated plume beyond the 10 miles (bound the plume).
Determine the dose rate.
Multiply the dose rate by the expected duration of the release to determine the TEDE or CDEt.
Make additional PARs based on this data and by using existing geo-political boundaries (i.e. towns, cities, etc.).
List those recommendations on the Part 1 form or provide additional detail on supplemental forms.
If PARs extend beyond the border of Oswego County request assistance from NY State staff to make proper notifications.
Rev. No.
32 Page 15 of 26
DOSE ASSESSMENT CALCULATIONS EAP-4 6.0 ATTACHMENTS
- 1.
INITIAL PROTECTIVE ACTIONS -
Pull out color flowchart
- 2.
AUGMENTED DOSE ASSESSMENT AND PROTECTIVE ACTIONS -
Pull out color flowchart
- 3.
ANALYZED ACCIDENT TYPES
- 4.
2001 POPULATION ESTIMATES EMERGENCY RESPONSE PLANNING AREAS
- 5.
DOWNWIND SURVEY WORKSHEET
- 6.
DOSE ASSESSMENT ACCIDENT TYPE SELECTION FLOWCHART
- 7.
CALCULATION FORM WHEN USING MONITOR READINGS TO ESTIMATE WHEN A PAR WILL BE REACHED BASED ON PROJECTED DOSE
- 8.
SOURCE TERM ENTRY FOR FIRST 15 MINUTE TIME STEP OF A REFUEL ACCIDENT WHEN BUILDING ISOLATON HAS OCCURRED AND THE RELEASE IS THROUGH THE STACK
- 9.
SOURCE TERM FOR A REFUEL ACCIDENT THAT RESULTS IN AN UNFILTERED RELEASE PATHWAY Page 16 of 26 Rev. No.
3 2
ANALYZED ACCIDENT TYPES Page 1 of 1 NewAccident Narnes/Analyze Accients Loss of Coolant Dop Refueling Steam Line Break Sam Lne Brak per Attactiment A of EAP-4 Accident Accident TwoPae e loca.jaf crd.jaf rfa.jaf sb2jaf s1b2.jaf esf.af Loss of Coolant Steam Line Break Two Containment Design Basis OLD EDAMS Accenl ame Used AcietRdAccident Phase Accident Analvzed Release Point Elevated Ground Elevated Ground Ground Elevated Nuclide LOCA CRD RFA SLB1 SLB2 CDBA Kr 83N 1.353E+00 1.577E-03 3.552E-04 1.517E-05 1.517E-05 1.154E-02 Kr 85M 2.906E+00 3.386E-03 1.657E-01 2.725E-05 2.725E-05 1.508E-04 Kr 85 1.301E-01 1.156E-04 9.144E-01 8.917E-08 8.917E-08 3.658E-09 Kr 87 5.572E+00 6.494E-03 2.695E-05 8.917E-05 8.917E-05 O.OOOE+00 Kr 88 7.894E+00 9.200E-03 5.252E-02 8.917E-05 8.917E-05 O.OOOE+00 Kr 89 9.817E+00 1.144E-02 O.OOOE+00 5.800E-04 5.800E-04 O.OOOE+00 lW Krsubtotal 2.767E+01 3.221E-02 1.133E+00 8.008E-04 8.008E-04 1.508E-04 co li LU 1Xel3lm 6.8252-02 7.953E-05 1.669E-01 6.692E6.69 2.692-08 7.994E-05
-i J
gJXel33m 9.942E-01 1.159E-03 1.991E+00 1.292E-06 1.292E-06 1.934E-03 0
Z1 Xe33 2.386E+01 2.781E-02 5.379E+01 3.658E-05 3.658E-05 2.769E-02 Xe135 3.081 E+00 3.589E-03 1.238E+01 9.833E-05 9.833E-05 1.952E-01 Xel35m 4.494E+00 5.239E-03 6.803E-01 1.158E-04 1.158E-04 5.686E-01 Xe37 2.094E+01 2.440E-02 O.OOOE+00 6.692E-04 6.692E-04 O.OOOE+00 Xel38 1.988E+01 2.316E-02 O.OOOE+00 3.975E-04 3.975E-04 O.OOOE+00 Xe subtotal 7.332E+01 8.544E-02 6.901E+01 1.319E-03 1.319E-03 7934E-01 Noble Gas (NG) subtotal 1.010E+02 1.176E-01 7.014E+01 2.120E-03 2.120E-03 7.936E-01 1131 3.406E-02 1.323E-04 2.439E-02 9.808E-04 9.808E-04 1.918E-03 1132 4.975E-02 1.933E-04 2.794E-05 7.628E-03 7.628E-03 2.803E-03 u
1133 7.119E-02 2.766E-04 2.498E-02 6.536E-03 6.536E-03 4.011E-03 1134 7.839E-02 3.044E-04 3.467E-10 1.380E-02 1.380E-02 4.417E-03 O
1135 6.725E-02 2.612E-04 4.233E-03 9.075E-03 9.075E-03 3.789E-03 fodi,,e subtotal 3.006E-01 1.168E-03 5.363E-02 3.802E-02 3.802E-02 1.694E-02 c
CS137 3.583E-03 1.671E-05 3.360E-03 1.198E-05 1.198E-05 2.019E-04 TE132 8.178E-03 O.OOOE+00 O.OOOE+00 6.900E-04 6.900E-04 4.606E-04 rn SR 89 2.132E-03 O.OOOE+00 O.OOOE+00 1.489E-04 1.489E-04 1.201E-04 SR 90 2.228E-04 O.OOOE+00 O.OOOE+00 1.126E-05 1.126E-05 1.255E-05 1
Ba1401 4.094E-03 O.OOOE+00 O.OOOE+00 4.358E-04 4.358E-04 2.306E-04 L140 4.336E-05 O.OOOE+00 O.OOOE+00 O.OOOE+00 0.000E+00 2.443E-06 Particulate subtotal 1.83E-02 1.67E-05 3.36E-03 1.30E-03 1.30E-03 1.03E-03 RELEASE RATE TOTALS (Cilsec) 1.01E+02 1.19E-01 7.02E+01 4.14E-02 4.14E-02 8.12E-01 Acident Duration Used for EDAMS 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 4 hours 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 2 hours 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 2 hours TOTAL Release Assumed (Ci) 2.92E+06 1.71E+03 5.05E+05 2.98E+02 2.98E+02 5.84E+03 Loss of Coolant RCl Rod Drop Rfueling Steam Line Break S
L LOCA C
Engineered Safety Accident l
Accident Two Phase Stam Lne Break Feature Componnt RATIOS Leakage 0
Iodine / Noble Gas Ratio 2.98E-03 9.93E-03 7.65E-04 1.79E+01 1.79E+01 2.13E-02 Noble gas / Iodine Ratio 3.36E+02 1.01E+02 1.31E+03 5.58E-02 5.58E-02 4.69E+01 Noble Gas / Particulate Ratio 5.53E+03 7.04E+03.
2.09E+04 1.632+00 1.63E+00 7.72E+02 Iodine / Particulate Ratio 1.65E+01 6.99E+01 1.60E+01 2.93E+01 2.93E+01l 1.65E+01 NG Particulate + Iodine Ratio 3.17E+02 9.93E+01 1.23E+031 5.39E-021 5.39E-021 4.42E+01 EAP-4 DOSE ASSESSMENT ATTACHMENT 3 Rev. No.
32 CALCULATIONS Page 19 of 26
2001 POPULATION ESTIMATES EMERGENCY RESPONSE PLANNING AREAS Page 1 of I 1
29 28 LEGEND 1 ERPA Number
=
ERPA Population 2001 Population Estimates J.A. FitzPatrick/Nine Mile Point Emergency Response Radiological Emergency Response Planning Areas (ERPAs)
Plan and Procedures EAP-4 DOSE ASSESSMENT ATTACHMENT 4 Rev. No.
32 CALCULATIONS Page 20 of 26 I
I I
I I
I I
I I
I I
I
DOWNWIND SURVEY WORKSHEET
(
Page 1 of 1 Team Number Sample Date Sample Time Map Location Description Miles Downwind Degrees Sector Dose Survey Results Calculation 3 inches 3 feet Closed Window Release Conversion Projected DDE Dose Rate Dose Rate Duration mrem to rem (DDE. TEDE)
Open (mrem/hour) 4')
(hours) 1 x lo" Rem Window Dose Rate Closed l
Window I
X X
Sample Air Sample Results Net K
Volume cpm Factor 2)
Child Thyroid (DCF)
Release Projected (gross - bkg)
Duration CDE Thyroid Iodine Gross cpm' (hrs)
(Rem)
X 6 x lo-1 X
2.6 x 16 (3)
X Partictilate Bkg cpm Particulate Net X
4 x 10-l'
=
Particulate (gross -
bkg)
Concentration Particulate Gross cpm (lCi/cc) or (Ci/m3 )
(1)If iodine net cpm is >8500, iodine cartridge should be returned for isotopic analysis on a priority basis.
(2)K factors assume 25 ft3 sample volume.
- 3) Child Thyroid DCF reflects correction (1.3 x 106 x 2) based on EPA-400-R-92-001, May 1992.
(4)Closed window dose rate equals DDE rate.
Completed by
/
/
/
Print Initial Date Time EAP-4 Rev. No.
32 DOSE ASSESSMENT CALCULATIONS ATTACHME11T Page 21 of 26
(
CALCULATION FORM WHEN USING MONITOR READINGS TO ESTIMATE WHEN A PAR WILL BE REACHED BASED ON PROJECTED DOSE Page 1 of THYROID
- 1. Record EDAMS forecasted thyroid dose
- 2. Record the corresponding plant effluent monitor reading that the thyroid dose in step 1 is based on
- 3. Divide the value in step 1 by the value in step 2.
forecasted thyroid dose example:
effluent monitor reading workspace:
5 rem workspace:
=
_rem ulnits
- 5. The result in step 4 corresponds to the monitor reading that will equal the 5rem PAG for CDE Thyroid based on the given flow, meteorology and monitor value at the time the data was taken.
- 2. Record the corresponding plant effluent monitor step 1 is based on reading that the TEDE dose in
- 3. Divide the value in step 1 by the value in step 2.
pie:
forecasted TEDE dose example effluenlt m onitor reading work-space:
1 rem work space:
I=ren_
units
- 5. The result in step 4 corresponds to the monitor reading that will equal the lrem PAG for TEDE based on the given flow, meteorology and monitor value at the time the data was taken.
EAP-4 DOSE ASSESSMENT ATTACHMENT 7 Rev. No.
32 CALCULATIONS Page 23 of 26 rem units rem ulnits
SOURCE TERM ENTRY FOR FIRST 15 MINUTE TIME STEP OF A REFUEL ACCIDENT WHEN BUILDING ISOLATON HAS OCCURRED AND THE RELEASE IS THROUGH THE STACK Page 1 of 2 The following information is extracted from JEP-00-034 for entry of Refuel Accident source term for the first 15 minute time step when building ventilation occurs as expected.
BACKGROUND:
During the Design Basis Refuel Accident, there will be two release pathways.
The first release pathway is through the Reactor Building Vent for eight (8) seconds.
This is a ground level, unfiltered release.
The reason for this release is that isolation of the reactor building takes place in 18 seconds following the accident, and the reactor building ducting was sized to provide a ten (10) second delay. In addition, during the transition from the Normal Reactor Building Ventilation mode to the Isolated mode, a potential exfiltration activity release occurs.
Exfiltration is the pressurization of the building. Radioactive gases, which escape the fuel pool, are released to the atmosphere 10 seconds after the accident for 8 seconds with a small additional release via exfiltration.
The remainder of the activity which escapes the pool is released to the atmosphere via the Standby Gas Treatment System (SBGT) and the stack (elevated release) beginning 18 seconds after the accident.
ENTRY STEPS FOR FIRST 15 MINUTE TIME STEP:
The following steps should be performed to estimate the offsite dose following a refuel accident using EDAMS.
In Raddose "Enter/Edit Source Term Data" Option, perform the following:
The first "time step" in the Source Term Data Entry Screen should contain two release pathways as follows:
- 2. On the first release path line, in the Accident type column, select "RFA -
Refuel Accident".
This will insert the filtered portion of the release into the model.
- a. Select "Y" for Elevated Release.
- b.
In the METHOD column, select "FSAR -
Default Release Rate".
- c. Select the appropriate Iodine method.
- 3. On the second release path line of time step 1, in the Accident type column, select "USER-User Defined Accident".
- a. Select "N" for Elevated Release.
- b.
Next, enter the following isotopic data into the isotopic entry screen:
(Nuclide PCi/sec I NuclidelpCi/secl I Nuclide pCi/sec lNuclide lpCi/sec1 Kr 83m I2.32E+01 Kr 85m 1.08E+04 Kr 85 5.94E+04 Kr 87 1.76E+00 Kr 88 3.42E+03 Kr 89 0.OOE+00 Xel31m 1.1OE+04 Xel33m 1.30E+05 Xel33 3.49E+06 Xel35 8.05E+05 Xel35m 4.42E+04 Xel37 0.OOE+00 Xe138 O.OOE+00 I131 l.90E+04 1132 1.82E+01 1133 1.63E+04 1134 2.28E+04 1135 2.76E+03 Cs137 2.19E+03 Te 132 O.OOE+00 Sr 90 O.OOE+00 Ba 140 O.OOE+00 Sr 89 O.OOE+00 La 140 O.OOE+00 EAP-4 DOSE ASSESSMENT ATTACHMENT 8 Rev. No.
32 CALCULATIONS Page 24 of 26
SOURCE TERM ENTRY FOR FIRST 15 MINUTE TIME STEP OF A REFUEL ACCIDENT WHEN BUILDING ISOLATON HAS OCCURRED AND THE RELEASE IS THROUGH THE STACK Page 2 of 2 This isotopic data is the ground level, unfiltered, 8-second release time averaged over the 15 minute RADDOSE V time step.
- 3. Press the "Accept" button. RADDOSE V will now calculate the first time step wth both the ground level and elevated portions of the release accounted for.
When this data is entered the first time step will appear similar to the following:
ADV STEP PATH ACCIDENT FLOWRATE METHOD MONITOR TOTAL REL I
IODINE STP TIME TYPE READING RATE MEHTOD MONITOR REL RATE (Ci/sec)
(Ci/sec' I
12:51 IE RFA FSAR 3.66E.01 FSAR 4.55E-03 20 USER USER 4.62E.06 4.62E.0o USER 6.09E.04 6.09E-02 3
NONE STEP Total 4.13E.01 Total Iodine.
6.54E-02
- 4. During the second time step, and all others, ONLY USE the default Refuel Accident.
- 5. When this data is entered the second time step will appear similar to the following:
ADV STEP PATH ACCIDENT FLOhRATE METHOD MONITOR TOTAL REL I
IODINE I
STP TIME TYPE READING RATE METHOD MONITOR REL RATE (Ci/sec)
(Ci/aec) 2 13:06 IE RFA PSAR 3.66E501 FSAR 4.55E-03 2
NONE I
3 NONE STEP Total =
3.66E01 Total Iodine =
4.55E-03 EAP-4 DOSE ASSESSMENT ATTACHMENT 8 Rev. No.
32 CALCULATIONS Page 25 of 26
SOURCE TERM FOR A REFUEL ACCIDENT THAT RESULTS IN AN UNFILTERED RELEASE PATHWAY Page 1 of 1 BACKGROUND NOTE: should be utilized if building ventilation isolation has occurred, and the release is being filtered through Stand-by Gas Treatment.
The following accident source term will be utilized prior to the availability of chemistry data for a refuel accident that results in an un-filtered release through either the stack, a building vent or unmonitored pathway.
The information is taken from the EAP-4 Attachment 3 tab for decay time 0.0.
- 1. Select USER as the accident type.
- 2. You will be queried as to whether the release is elevated.
a.Respond as appropriate for the conditions.
- 3. You will then be prompted to enter each isotope in uCi/sec.
a.Utilize the following information for the release source-term:
ISOTOPE ] uCi/sec ISOTOPE uCi/sec
- 4. Obtain and utilize isotopic data atmosphere as soon a possible.
1131 2.09E+ 06 1132 2.40E+03 1133 2.14E+06 1134 3.OOE-02 1135 3.63E+05 Subtotal 4.62E+06 Cs137 2.88E+05 Tel32
- 0. OOE+OO Sr 89 0.OOE+00 Sr 90
- 0. OOE+OO Ba140 O.OOE+O0 Lal40 O.OOE+O0 Subtotal 2.88E+07 Total l6.06E+O8 from the refuel floor vent or refuel floor Kr 83m 3.06E+09 Kr 85m 1.42E+06 Kr 85 7.83E+06 Kr 87 2.32E+02 Kr 88 4.51E+05 Kr 89
- 0. OOE+OO Xel3lm 1.43E+06 Xel33m
- 1. 71E+06 Xel33 4.GlE+08 Xel35 1.06E+08 Xel35m 5.82E+06 Xel37
- 0. OOE+O0 XeI38
- 0. OOE+00 Subtotal 6.01E+08 EAP-4 DOSE ASSESSMENT ATTACHMENT 9l Rev. No.
32 CALCULATIONS Page 26 of 26
-1 ENTERGY NUCLEAR OPERATIONS, INC.
JAMES A. FITZPATRICK NUCLEAR POWER PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE RELEASE RATE DETERMINATION EAP-4.1 REVISION 16 APPROVED BY:
DATE: D t
r RESPONSIBLE PROCEDURE OWNER EFFECTIVE DATE: ja-ul 3G, RO63 r
FIRST ISSUE 0 FULL REVISION 3 LIMITED REVISION E INFORMATIONAL USE INFORMATIONAL USE TSR ADMINISTRATIVE CONTROLLED #:
r PERIODIC REVIEW DATE:
6/- 7
.6200'v~~~~~.
. I I
I
- TSR al4tt
RELEASE RATE DETERMINATION EAP-4.1 REVISION
SUMMARY
SHEET REV. NO.
16 Added attachment 13.
- Replaced Attachment 12 to make it consistent with EAP-4, Attachment 3 source term information.
- Re-ordered stack and vent calculations on Attachment 11.
- In section 4.1 - added 2 nd note in reference to new attachment 13.
- In section 4.1.1.A -
added note for EPIC RRC display info.
- In section 4.1.1.D table changed K Factor rate from negative 1 to negative 7, and changed to Ci from uCi.
- On Attachment 1 updated the K factor information.
- On Attachment 3 NOTE: added or 2.75 CFM to complete the information for the primary containment and reactor building leak rate.
- On Attachment 11 deleted &p" for the iodine percent of tech spec calculation.
- Added a note to clarify section 4.1.4.H on use of flowrate for CHRM calcs.
Added information to calc box for flowrate to "Utilize rated or calculated primary containment leakage".
15 On Attachment 3 in the Note section changed the followlng items:
added (1298 cc/sec) after 1.5W per day and changed RO-13 to RO-15 and changed the containment leak rate from - 1,437 to -1,569.
- Also, on attachment three deleted reference to Tech Spec's per section 3.2.a, volume 1B because of implementation of the ITS Program and a referenced the document Offsite Dose Calculation Manual (ODCM).
- Also in attachment three in the box Attachment 4 Location on Graph -
containment volume was changed to 7.4.8E-9cc.
Rev. No.
16 Page 2 of 27
I RELEASE RATE DETERMINATION EAP-4.1 r;
i i
TABLE OF CONTENTS PAGE PURPOSE...........................
REFERENCES........................
..................... 4
...................... 4 INITIATING EVENTS.......................
4 PROCEDURE.......................
5 Release Rate Determination................
Default Accident Source Terms....
Unmonitored Release..............
........... 5
......................12
......................13 ATTACHMENTS.......................
13
- 1.
FLOW CHART TO DETERMINE RELEASE RATE FROM LOW RANGE EFFLUENT MONITORS..........................
- 2.
FLOW CHART TO DETERMINE RELEASE RATE FROM HIGH RANGE EFFLUENT MONITORS (HREM)...................
- 3.
WORK SHEET TO DETERMINE RELEASE RATE FROM CONTAINMENT RAD MONITORS.........................
14 15 16 FITZPATRICK HRCRM READINGS........
1.5% LEAKAGE SOURCE TERM ESTIMATE..
10% LEAKAGE SOURCE TERM ESTIMATE..
25% LEAKAGE SOURCE TERM ESTIMATE..
19 Rev. No.
- 8.
50% LEAKAGE SOURCE TERM ESTIMATE................. 21
- 9.
100% LEAKAGE SOURCE TERM ESTIMATE................. 22
- 10.
CATASTROPHIC LEAKAGE SOURCE TERM ESTIMATE........ 23
- 11. CALCULATION METHOD FOR DETERMINING PERCENT OF TECHNICAL SPECIFICATION FOR NRC EVENT NOTIFICATION WORKSHEET............................ 24
- 12. ANALYZED ACCIDENT TYPES.......................... 26
- 13. CALCULATION TO DETERMINE EFFLUENT MONITOR READING TO REACH 100% OF TECH SPECS FOR NOBLE GAS AND IODINE................................... 27 16 Page 3 of 27 SECTION 1.0 2.0 3.0 4.0 4.1 4.2 4.3 5.0 4.
5.
6.
7.
RELEASE RATE DETERMINATION EAP-4'.1 1.0 PURPOSE This procedure provides instructions for manually estimating release rates in the event of an accidental release of radioactivity to the environment.
2.0 REFERENCES
2.1 Performance References None 2.2 Developmental References 2.2.1 EAP-5.3, ONSITE/OFFSITE DOWNWIND SURVEYS AND ENVIRONMENTAL MONITORING 2.2.2 EAP-42, OBTAINING METEOROLOGICAL DATA 2.2.3 NUREG-0654, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants 2.2.4 JAF FSAR Chapter 14 2.2.5 EAP-4, DOSE ASSESSMENT CALCULATIONS 2.2.6 High Range Containment Monitor Response to Post Accident Fission Product Releases -
James A.
FitzPatrick Nuclear Power Plant, SL-4370, Sergeant Lundy, May 1985 3.0 INITIATING EVENTS, 3.1 An emergency classification has been declared as-defined in IAP-2, and 3.2 A release-of radioactivity exceeding technical specifica-tions is-suspected or underway.
Rev. No.
16 Page 4 of 27
RELEASE RATE DETERMINATION EAP-4.1 4.0. PROCEDURE 4.1 Release Rate Determination NOTE:
Use Attachment 11 to calculate the percent of Tech.
Spec. in order to determine if the Tech. Spec.
release rate has been exceeded and for completion of the NRC Event Notification Worksheet, EAP-1.1,.
NOTE: 3 may be utilized to calculate effluent monitor readings that correspond to 100% Tech Spec release rate values.
4.1.1 Low Range Effluent Monitor calculation A. Record date, time and name of individual performing calculations in upper right-hand corner of Attachment 1.
NOTE: EPIC RRC display calculated release rates for low range monitors are in uCi/sec, these values must be-converted to Ci/sec.
B. Record observed gross count rate with appropriate units for the Reactor Building (RxB)
(lower floors), Refuel Floor (RF), Radwaste (RW), Turbine Building (TB) and/or Stack.
This data may be obtained from the EPIC computer.
IF computer points are-unavailable, Control Room AND/OR local monitors can be used for this data.
NOTE:
For stack releases, it is important to determine whether any dilution fan is operating.
C. For Building Vent Releases, multiply the gross count rate (cpm) by the default K factor listed in table on following page, until updated K factors are available based on recent chemistry sample data.
Rev. No.
16 Page 5 of 27
RELEASE RATE DETERMINATION EAP-4. 1 D. For Stack Releases, multiply the gross count rate (cps) by the default K factor listed below, unless an updated K factor is available based on recent Chemistry sample data.
Page 6 of 27
.i I Monitor R Factor Normal Flow RatesBased on (cfm)
Reactor Bldg.
3.2E-7 Ci/sec/cpm 61,000 Below Refuel Floor(Pt. 3337)
Refuel 3.7E-7 Ci/sec/cpm 70,000 Floor(Pt. 3338)
Radwaste 1.7E-7 Ci/sec/cpm 32,500 Bldg.(Pt. 3340)
Turbine 5.6E-7 Ci/sec/cpm 107,000 Bldg.(Pt. 3339)
Stack(Pt. 3336) 6.OE-7 Ci/sec/cps 6,600 IF flow rates differ from the Normal Flow Rates listed above, THEN a correction to the K factor is necessary as follows:
K(corrected)
=
New Flow Rate lX K Factor(listed) ]
-w Normal Flow RateJ Rev. No.
16
RELEASE RATE DETERMINATION EAP-4.1 NOTE:
The accuracy of ventilation flow rate indications at the low end of an instrument range should be confirmed with appropriate instrument calibration procedures.
E. An estimate of the iodine release rate can be obtained by multiplying the I/NG ratio from a chemistry sample by the NG release rate.
IF a chemistry sample is not available, THEN the iodine release rate can be estimated by multiplying a default I/NG ratio by the NG release rate.
For default release rates and I/NG ratio, refer to Attachment 12.
F. IF the low range effluent monitors are inopera-tive or off-scale, THEN the appropriate high range effluent monitor must be used.
4.1.2 High Range Effluent Monitor (HREM) Calculation A. Record date, time-and individual performing calculation in upper right-hand corner of.-
B. Record observed dose rate for the Stack, Turbine Building -(TB) and/or Radwaste (RW).
This data may be obtained from the EPIC computer.
IF computer points are unavailable, THEN Control Room monitors can be used for this data.
NOTE:
For stack releases, it is important to determine whether any dilution fan is operating.
C. Multiply the dose rate by the K factor listed below.
Rev. No.
16 Page 7 of 27
- 1
RELEASE RATE DETERMINATION EAP-4.1 NOTE: These conversion constants are based on normal flow-rates listed below.
A conversion factor of 0.45 (Ci/cc)/(mR/hr) was applied to the normal flow rate.
This value is given by General Electric and is based on the monitor response to Xe-133.
- IF flow rates differ from the Normal Flow Rates listed above, THEN a correction to the.K factor is necessary as follows:
K(corrected) r New Flow Rate lX
[ Factor(listed)1 LNormal Flow Rate J L
D. An estimate of the iodine release rate can be obtained by multiplying the I/NG ratio from a 6hemistry sample by the NG release rate.
IF a chemistry sample is not available, TEDN the iodine release rate can be estimated by multiplying a default I/NG ratio by the NG release rate.
For default release rates and I/NG ratios, refer to Attachment 12.
E. A back calculated release rate may be estimated from field survey data in lieu of or in addition to the estimate from low and high range effluent' monitors.
Page 8 of 27 HREM R FACTOR NORMAL FLOW RATES (cfm)*
STACK One SGT train 1.40 (Ci/sec)/(mR/hr) based on 6,600 operating One SGT train and one stack 2.54 (Ci/sec)/(mR/hr) based on 12,000 dilution fan operating; TURBINE BLDG, 22.6 (Ci/sec)/(mR/hr) based on 107,000 RADWASTE BLDG 6.85 (Ci/sec)/(mR/hr) based on 32,500 Rev. No.
16
RELE.ASE RATE DETERMINATION EAP-4.1 4.1.3 Back Calculations from Downwind Survey Dose Rate Data using EDAMS A. Start the EDAMS program and from the EDAMS icons, select "EDAMS".
NOTE:
The mouse does NOT work in this DOS Sub-routine.
B. Select "Release Rate Calculations".
C. Select "James A. FitzPatrick".
D. Select "Back calculate".
E. Enter the time survey data was obtained (24-hour format).
F. Enter a number representing one of the accident types listed.
G. Enter the wind speed (MPH).-
H. Enter "E" for elevated/stack or "G" for ground/vent release.
I. Enter the stability class (A -
G).
J. Enter the three (3) foot closed window reading from the ion chamber (mR/hr).
K. Enter the downwind distance that the above reading was obtained.
(Use 0.87 miles if the reading is taken at the site boundary.)
L. Hit the F9 key to calculate.
Record or print the results.
Page 9 -of 27
AI Rev. No.
16
RELEASE RATE DETERMINATION EAP-4.1 4.1.4 Release Rate Estimation Using Containment High Range Radiation Monitor Data A. Record date, time and individual performing calculations in upper right-hand corner of.
B. Record containment rad monitor I.D. (i.e.,
either 27-RE-104 A or B) in Column 1 or an average of the two.
C. Record the containment rad monitors average reading (dose rate) or the individual monitor reading (dose rate) in Column 2. Obtain readings from EPIC.
D. Record the time the containment rad monitor dose rate was observed in Column 3.
E. Record the time of shutdown in Column 4.
F. Determine the time in hours after shutdown that the containment radiation monitor reading was taken (Column 4 -
Column 3) and record in Column 5..
Page 10 of 27 Rev. No.
16
RELEASE RATE DETERMINATION EAP-4.1 NOTE:
Ensure that credit is taken for any dilution provided to the value calculated in step 4.1.4.G prior to it entering the effluent pathway to the environment (i.e. dilution by Reactor Building volume, etc.).
G. Determine and record in Column 6 the calculated concentration in containment for the time after shutdown reading using the curves in Attachment 4 and the following core damage estimates:
- Concentrations derived using EAP-44 estimates of core inventory and a containment volume of 7.42E+9cc (i.e. drywell and torus gas space volume).
NOTE:
The expected flow rate to the environment is the flow rate from the primary containment outward -
it is NOT stack flow.
For example: if there is no apparent abnormal leakage then the rated primary containment leakage rate should be used as provided on, until a leakage rate can be calculated by TSC engineering.
-H. Determine the expected flow rate (cc/sec) to the environment and record in Column 7. Assistance from TSC engineering staff may be necessary in determining flow rates.
I. Determine the estimated release rate by multiplying Column 6 by Column 7. Record in Column 8.
Rev. No.
16 Page 11 of 27 Calculated Concentration
- Location on Graph (Ci/cc)
Area above Case #1 5.20E-2 Area between Case # and Case #2 3.45E-2 Area between Case #2 and Case #3 1.09E-2 Area between Case #3 and Case #4 3.30E-4 Area between Case #4 and Case #5 l.91E-5 Area between Case #5 and Case #6 l.91E-6 Area below Case #6 Normal
RELEASE RATE DETERMINATION NOTE:
EPIC provides release rates based on default K factors and normal flow rates.
4.1.5 Obtaining Release Rate Using EPIC A. Call up the Radioactivity Release Control (RRC) display on EPIC.
B. Obtain and record release rate data from RRC display for release pathway of concern.
4.2 Default Accident Source Terms 4.2.1 Various types of design basis accidents have been analyzed and source terms estimated.
Refer to 2 for estimated values.
4.2.2 In addition, source term estimates have been developed based on differing amounts of core damage for accidents resulting in leakage of activity through the drywell boundary.
A. Attachments 5 through 10-provide correlation between stack source term estimates for given containment leak rates and containment high range radiation monitor readings.
B. These attachments can be used to project what a release rate may be given a break in containment and containment failure imminent.
C. These source terms are only estimates and should be input with the understanding of the assumptions used in their development.
D. The source terms correspond to test cases in the Sergeant Lundy study "High Range Containment Monitor Response to Post Accident Fission Product Release" and are plotted on Attachments 5 through 10.
These graphs are based on calculation [[::JAF-89-003|JAF-89-003]] filed in. the original procedure EAP-4 master file.
Rev. No.
16 Page 12 of 27 EAP-4.1
RELEASE RATE DETERMINATION EAP-4.1 4.3 Unmonitored Release All likely release pathways are monitored.
If there is a release through an unmonitored pathway, the release should be evaluated based on a source term (area monitors, process monitors, and/or local grab samples, as appropriate) or back calculations from downwind readings as described in Section 4.1.3 of this procedure.
5.0 ATTACHMENTS
- 1.
FLOW CHART TO DETERMINE RELEASE RATE FROM LOW RANGE EFFLUENT MONITORS
- 2.
FLOW CHART TO DETERMINE RELEASE RATE FROM HIGH RANGE EFFLUENT MONITORS (HREM)
- 3.
WORK SHEET TO DETERMINE RELEASE RATE FROM CONTAINMENT RAD MONITORS
- 4.
FITZPATRICK HRCRM READINESS
- 5. -1.5% LEAKAGE SOURCE TERM ESTIMATE
- 6.
10% LEAKAGE SOURCE TERM ESTIMATE
- 7.
25% LEAKAGE SOURCE TERM ESTIMATE
- 8.
50% LEAKAGE SOURCE TERM ESTIMATE
- 9.
100% LEAKAGE SOURCE TERM ESTIMATE
- 10. CASTASTROPHIC LEAKEAGE SOURCE TERM ESTIMATE
- 11. CALCULATION METHOD FOR DETERMINING PERCENT OF TECHNICAL SPECIFICATION FOR NRC EVENT NOTIFICATION WORKSHEET
- 12. ANALYZED ACCIDENT TYPES
- 13. -CALCULATION TO DETERMINE EFFLUENT MONITOR READING TO
-REACH 100% OF TECH SPECS FOR NOBLE GAS AND IODINE Rev. No.
16 Page 13 of 27
FLOW CHART TO DETERMINE RELEASE RATE FROM LOW RANGE EFFLUENT MONITORS p'age 1 of 1 DATA:
Rx B (cpm)
RF _
(cpm)
Stack l Rx B EPIC (cpm)
Pt I.D.
3337 Refuel EPIC Floor Pt. I.D.
(cpm) 3338 Radwasto EPIC (cpm)
Pt. I.D.
3340 Turbine EPIC Building Pt. I.D.
(cpm) 3339 Stack EPIC (cps)
Pt. I.D.
3336 (cps)
RW _
_ (cpm)
TB (cpm)
(cpm) x (
) = RR (CiUs)
I x (
)
](epm) x (
) = RR (CUs)
I x (
)
](cpm) x (
) = RR (CVs)
J I
(
,=
I (cpm) x (;
) = RR (CUs)
I x(
)=
(cps) x (
) = RR (CVs) x (.
)
SAMPLE INL
.4
~~~~~~~~.....
Refer to Section 4.1.1. C and D for additional guidance i
Normal Flow Rates Based on Monitor K Factor (dm)
Reactor Bldg. (Pt 3337) 3.2E-7 Ci/sec/cpm 61,000 Refuel Floor (Pt. 3338) 3.7E-7 Ci/sec/cpm 70,000 Radwaste Bldg. (Pt 3340) 1.7E-7 Ci/sec/cpm 32,500 Turbine Bldg. (Pt 3339) 5.6E-7 Ci/seccpm 107,000 Stack (Pt 3336) 6.OE-7 Ci/sec/cps 6,600 IF flow rates differ from the Norm al Flow Rates listed above, THEN a correction to the K factor is necessary as folows:
New Flow Rate K(orrected)
=
New Fw Rate K Factor (isted NOTE: The accuracy of ventilation flow rate indications at the low end of an I instrument range should be confirmed with appropriate instrument calibration Lprocedures.
DATE:
TIME:
LNAME:
Actual Iodine Release Rate (RR)
(Actual l/NG Ratio) x RR(NG) = RR(lodine) Ci/sec
______________ x
=
C /sec Estimated Iodine Release Rate (RR)
( **
)
x RR(NG)
= RR (lodine) C/sec x
=
C/sec Noble Gas Release Rate RR(NG)
_ CUsec Iodine / Noble Gas Ratio RATIOS Loss of Coolant Accident 2.98E-03 Control Rod Drop 9.93E-03 Refueling Accident 1.24E-Q4 Steam Line Break 1.79E+01 Containment Design Basis 2.13E-02 Accident EAP-4.1 RELEASE RATE DETERMINATION ATTACHMENT 1 Rev. N 6-16 Page 14 27
(
FLOW CHART TO DETERMINE RELEASE RATE FROM HIGH RANGE EFFLUENT MONITORS (HREM)
DATA: Stack (mR/hr)
TB (mR/hr)
RW I_(mR/hr)
(mR/hr)
(1.40)
=
RR (Ci/s)
______I ___
(1.40)
=
l(mRJhr).
.. (2.54)
=
RR (Ciis)
(2.54)
=
Based on G.-E. Data for monitor response under normal flow rates listed on page 7.
DATE:
TIME:
NAME:
Noble Gas Release Rate RR (NG)
(Ci/sec)
C Page 1 of 1 I Artual loduine RAIAA RatA RR}
(Actual VNG Ratio) x RR(NG)
= RR(lodine) CVsec:
x
=_
. Estimated Iodine Release (RR) Rate
- (
) x RR(NG)
= RR(lodine) Ci/sec
(
)x
=
Stack (HREM)
(mRthr) 1191 Stack (HREM)
1 Fan 1191 (mR/hr)
Pt. I.D.
(mR/hr) 1194
.(mR/hr)
(22.6)
=
RR (Ci/s)
(22.6)
=
Pt. I.D..
(mR/hr) 1195:]
- ~~~~~~~~~~~~~~~~~~~~~~~~~
(mR/hr)
(6.85)
=
RR (Cils)
(6.85) =
Iodine / Noble Gas Ratio RATIOS **
Loss of Coolant Accident 2.98E-03 Control Rod Drop 9.93E-03 Refueling Accident 1.24E-04 Steam Line Break Single Phase 1.79E+01 Steam Line Break Two Phase 1.79E+01 Containment Design Basis Accident 2.1 3E-02 EAP-4.1 I
RELEASE RATE DETERMINATION".
ATTACHMENT 2 Rev. No.
16 Page 15 of 27
(
WORK SHEET TO DETERMINE RELEASE RATE FROM CONTAINMENT RAD MONITORS TIME:
NAME:
Column 1 Column 2 Column 3 Column 4 Column 5 Column 6 Column 7 Column 8 Containment Containment Time of Time of Time of Calculated Expected Flow Rate Estimated Rad Monitor Rad Monitor Reading Shutdown Reading Concentration to Environment Release Rate in containment
- (cc/sec)
(Ci/sec)
I. D.
Dose Rate After (ci/cc)
(R/hr)
Shutdown Utilize rated or (hr) AT calculated primary
.hr)
AT containment leakage x
x
- To convert cfm to cc/sec. multiply cfm by 472 from CRC handbook of Chemistry and Physics, 64th Edition, pg. F-308.
NOTE:
The Primary Containment and Reactor Building leak rate default value is 1.5% per day (1298 cc/sec or 2.75 CFM). The as-left Primary Containment leak rate calculated after RO-15 was approximately 1,569 scf/day (1.1 CFM or 514.4 cc/sec).
The dose rate at the site boundary is 500 mr/yr whole body from noble gas, 1,500 mr/yr for any organ from iodines and particulates with half lives greater than 8 days, per Offsite Dose Calculation Manual (ODCM).
The current total as-left Primary Containment leakage can be found in the last run of ST-39B attachment 7. The last run ST-39B is in a binder in the bottom drawer of the middle file cabinet in the Operations file area adjacent to the Control Room. Attachment 7 is used to update the leakage totals subsequent to a complete run of ST-39B (i.e., forced outage LLRTs). The most recently dated forms of attachment 7 will contain the current as-left Minimum Pathway leakage and Maximum Pathway leakage. These numbers will provide the least and most amount of leakage projected for all Primary Containment leakage pathways. These numbers are reported in Standard Liters per Minute (SLM). To convert to cc/sec, as required in EAP-4. 1, divide SLM by 28.31 to get CFM, them multiply CFM by 472 to get cc/sec.
I -
Rev. No.
16
.~~
RELEASE RATE DETERMINATION ATTACHMENT 3 Page 16 of 27 DATE:
I, I
Calculated Concentration* Location on Graph (Cl/cc)
Area above Case #1 5.20E-2 Area between Case #1 and Case #2 3.45E-2 Area bet'ween Case #2 and Case #3 1.09E-2 Area between Case #3 and Case #4 3.30E-4 Area between Case #4 and Case #5 1.9 1E-5 Area between Case #5 and Case #6 1.91E-6 Area below Case #6 Normal
- Concentrations derived using EAP-44 estimates of core inventory and a containment volume of 7.48E+9cc (i.e.
drywell and torus gas space volume).
I i
Page 1 of 1
RELEASE RATE DETERMINATION EAP-4 1
- st 4A, -
A 1
V
^
ll nlmcnFA n,-
_~
4 k L. L%LLkL=lA L.
FITZPATRICK HRCRM READINGS
.rage o
L FITZPATRICK -
HRCR!t REROINGS Page 17 of 27 7
ins 104 to 0:
=
cc C-z i
Mc a:
let Ioa Rev. No.
16
1.5% PRIMARY CONTAINENI BOUNDRY I FAKACE ACCIDENI SOURCE TERM ESIIMAILS lME IN HOURS AFTER PIPEBRAKE IN DRYWELL (D
H 50 40 r3 20 Q
U 0
(t 10 t-4 Lxj ti txi H
H H
0 to
.D to H
IOD 0
I-h H
I 1 I
II a >
C lH t
30 H 0
HA 11
(
10% PRIMARY CONTAINME N BuNDRY LEAKAGE ACCIDENT SOURCE TERM ESrMA[LS
- I
~~~~~~~~~~~~~~I 1
3 12 18 24 IITINE IN HOURS AFTER PIPEBRAKE IN DRYWELL
(
400
- a
.z 0
H-a'N U) ti tTlH
'-4 0z U
u 0u u
Bv U.
I to I%D 0
I-h t11yfi 0 PiIFr m
H t4i H
30
° H
.II
800'_
700-600 0c 500 3 400
. 8 i~
A300 1'
200 10 0
I 25% PRIMARY CONTAINMENT BOUNDRY LEAKACE ACCIDENT SOURCE [ERM ESTIMATES
./
a 7
2
-91" A-1 C-~s, _.3 ------
e, ify~~~
I 3
12 I-Y 18 TIME IN HOURS AFTER PIPEBRAKE IN DRYWELL 24 Po at w
tO 0
LrI H
HH 0z toI U1 Lxi dP 0
En t
q o
- a 0 (X K H y Lxi
-3 H
LI3 tz i
9' w
_.A 30 I-A
:;
l I
I Ii
/. _......I 4~~~~~~~
C 50% PRIMARY CONTAINMENT BOUNDRY LEAKAGE ACCIDENT SOURCE TERM ESTIMATES
(
(D la.
0 H-M.~
U IW U
U 0
o CluH 1d to M
10 to En 0
dP t-4 M
a M
In 0
0 Lt M
H M
r3 H
H MX LXI M
En ri ri 0
1~~~~~~~~~~~~~~~~
I a,
rtpS 0
0, 1
I3 I 12 18 24 TINE I HOURS AFTER PIPEBRAKE IN DRYWELL
-4 30 f
0
4000. Ip 3000 U:
^ 2000 U
E 0
10 1 00 CQ 0'
7 t
1 3
-- I I I 1007 PRIMARY CONTAINMENT BOUNDRY LEAKAGE ACCIDENT SOURCE TERM ESTIMATES
~~1 NC e,se Cse #3 Ca se3 12 18 TIME IN HOURS AFTER PIPEBRAKE IN DRYWELL la (D
0
'-0 0op tI Li Ca 0
1-43 ct tm 0
wm 0rr w
trx tTi ri W
W H
z0 H
0h 41 24 3
to 0
g I
== -
- --
fz
(
C~~~~~~~~~~
8S) 0 r ~
CATASTROPHIIC FAILURE BOUNDRY IEAKAGE SZ:.
ACCIDE;NT SOURCE TERM ESTIMATES t
.~~~~~ 1L0, 1a,=.
7q000I H
5 °P.l
- //
1 6
40,°°°00 _7f
_ _____ c]
W 5 1.1 v
4/
HY~~~~~~~~~~~~~~~~~~~~~~~~
I z~~~~~~~~~~~~~~~~~~~~~~~
2Om 0-A 20OOO__
,1 IOpO L-4.PeJ La D
I I
]C ase iY
_I
_1_W 1
T-MEIO3RS 12 18 24 30 1-'
ME I IOURS AFTER PIPEBRAKE IN DRYWELL FH
RELEASE RATE DETERMINATION EAP-4.1 1 Page 1 of 2 CALCULATION METHOD FOR DETERMINING PERCENT OF TECHNICAL SPECIFICATION FOR NRC EVENT NOTIFICATION WORKSHEET
- 1.
The formula for calculating the percent of Technical Specification of Airborne releases was derived from the JAF Offsite Dose Calculation Manual, Revision 7.
The following assumptions apply:
The release Technical Specification limit of 500 mrem/year ([CTS] Technical Specifications, Appendix B, Section 3.2.a. 1 [ITS] 5.5.4) was used as the basis for the noble gas instantaneous release limit.
The Technical Specification of 1,500 mrem/year to any organ ([CTS] Technical Specifications, Appendix B, Section 3.2.a.2 [ITS15.5.4) was used as the basis for the radioiodine, tritium and eight day particulate instantaneous release limit.
The most conservative X/q (4.83E-7 sec/m3 ) for ground based receptors at the site boundary was used for all cases and is the FSAR defined accident X/q.
AR assumptions and conservatism of the ODCM were applied.
- 2.
As a result of these assumptions and conservatism, these formulae should be used only to estimate the initial percent of Technical Specifications as required by the NRC Event Notification Worksheet (EAP-1.1, ). As more detailed source term and meteorological data become available, a more accurate determination of percent Technical Specification should be performed.
- 3.
Technical Specification release rates for stack and vent noble gas releases are obtained from the setpoint release rates for these points described in the ODCM.
- 4.
Calculation method for determining the initial percent of Technical Specifications for NRC Event Worksheet (use calculation worksheet on next page).
For Noble Gas Vent Release:
% T.S. = RRNO (Ci/s) x 1432 For Noble Gas Stack Release:
% T.S. = RRNO (Ci/s) x 333 Where RRNG = noble gas release rate in curies per second 1432 = 100% divided by vent setpoint release rate of 0.0698 Ci/sec 333 = 100% divided by stack setpoint release rate of 0.3 Ci/sec These equations assume an instantaneous release rate, ODCM dose conversion factors, and historical meteorological data.
For Gross Liquid Release excluding Tritium:
% T.S.
FL (gal/r) x CL (LCi/ml) x 2120 Where FL = flow rate in gallons per minute CL = concentration of liquid effluent in iCi/ml 2120 = unit and dose conversion factor For iodine, tritium and particulates with half-lives greater than 8 days:
% T.S.i meanRparR,t*l,
= ERine (Ci/sec) x 40.48
% TS.tnjtj
= RRfti, (Ci/sec) x 0.32 Total % T.S. = E % T.S. for all release points Rev. No.
16 Page 24 of 27
-r 1 Page 2 of 2 CALCULATION METHOD FR DETERMINING PERCENT OF TECHNICAL SPECIFICATION FOR NRC EVENT NOTIFICATION WORKSHEET Noble Gas STACK Release
% T.S.,
RRNG (Ci/s) x 333
% T. S.
=
RRNG __
I Ci/s x 333
% T.S. =
Noble Gas VENT Release
% T.S. = RRNG x 1432
% T.S.=
RRNG I Ci/s x 1432 9a T.S. =
NOTE:
This is indication of a release in progress.
1009 NG Stack Release = 0.3 ci/sec 100% NG Vent Release= 0.0698 ci/sec NOTE:
If
'. TS 100 - Fill out Part II Form.
Gross Liquid Release excluding Tritium
% T.S. = FL x CL (Ci/ml) x2120
% T.S.
=
[FL I
% T.S. =
gal/m x
[CL Iodine and Particulate with half lives greater than 8 days
% T.S. ip E RR(Ci/s)i (all vents) x 40.48
% T.S.
i&p
=
(RRi __)
Ci/s + (RRi
) Ci/
+
stack T.B. vent RRi
) Ci/ s + (RRi
) Ci/s +
Rx Bldg vent Refuel Floor (RRi
) Ci /s] x 40.48 Radwaste Bldg
% T. S. ip
= (RRi
% T.S. ip
=
_) Ci/s x 40.48 total vents Tritium
%9 9%
T.S.
Tritium T.-S.
Tritium E RR(Ci/)Tritium (all vents) x 0.32
= [(RRTritiu. _
Ci/
+ (RRTritium
) Cil/s +
stack T.B. vent (RRTritium
)_ )Ci/s
+ (RRTii
_)
Ci/s
+
Rx Bldg vent, Refuel Floor (RRTritiuM
) Ci/s] x 0.32 Radwaste Bldg
% T. S.
= (RRTritium
% T.S.
ritium -
total
) Ci/s x 0.32 I vents Where:
RR= release rate (Ci/sec)
FL - Flow Rate (gal/rm)
CL Concentration of liquid effluent (pCi/ml)
EAP-4.1 RELEASE RATE DETERMINATION ATTACHMENT 11 Rev. No.
16 Page 25 of 27
]1 lCi/ml x 212 0]
I RELEASE RATE DETERMINATION EAP-4.1 2 Page 1 of.l ANALYZED ACCIDENT TYPES NwAccident Names/AnalyzedAccidents Los ofCoolant Control Rod Drop Rcing Stam Lin Brak Steam Line Break L
p..
per Attachment A of EAP-4 Accident C________
loca.jaf crd.jaf rfa.jaf sIb2.jaf sib2.jaf esf.Jaf OLD EDAMS Accident Name Used Loss o Coolant Control Rod Drop Refueling Accident SteamLinBe BekTwo Accident lAnalzed Release Point Eevated Ground Elevated Ground Ground Elevated
's e
- Nucltde k
LOCA *f z
RFA s
SLBt'.
SLBZ;t;
.COA-t Kr 83M 1.353E+00 1.577E-03 3.552E-04 1.517E-05 1.517E-05 1.154E-02 Kr 85M 2.906E+00 3.386E-03 1.657E-01 2.725E-05 2.725E-05 1.508E-04 Kr85 1.301E-01 1.156E-04 9.144E-01 8.917E-08 8.917E-08 3.658E-09 Kr 87 5.572E+00 6.494E-03 2.695E-05 8.917E-05 8.917E-05 O.OOOE+00 Kr 88 7.894E+00 9.200E-03 5.252E-02 8.917E-05 8.917E-05 0.000E+00 Kr 89 9.817E+00 1.144E-02 0.OOOE+00 5.800E-04 5.800E-04 0.000E+OO co u
Kr subtotal 2.767E+01 3.221E-02
- 1. 133E+00 8.008E-04
&008E-04 1.508E-04 c
lo ui <>1Xel31m 6.825E-02 7.953E-05 1.669E-01 6.692E-08 6.692E-08 7.994E cO 21XeI33m 9.942E-01 1.159E-03 1.991 E+00 1.292E-06 1.292E-06 1.934E-03 0z Xel33 2.386E+01 2.781 E-02 5.379E+01 3.658E-05 3.658E-05 2.769E-02 Xcl35 3.081 E+00 3.589E-03 1.238E+01 9.833E-05 9.833E-05 1.952E-01 XeI35m 4.494E+00 5.239E-03 6.803E-01 1.158E-04 1.158E-04 5.686E-01 Xe137 2.094E+01 2.440E-02 O.OOOE+00 6.692E-04 6.692E-04 O.OOOE+00 Xe138 1.988E+01 2.316E-02 O.OOOE+00 3.975E-04 3.975E-04 O.OOOE+00 Xe subtotal 7.3322+01 8.544E402 6.901E+01 1.319E-03 1.319E-03 7.934E-01 Noble Gas (NC) subtotal 1.010E+02 1.176E-01 7.014E+01 2.120E-03 2.120E-03 7.936E-01 1131 3.406E-02 1.323E-04 2.439E-02 9.808E-04 9.808E-04 1.918E-03 1132 4.975E-02 1.933E-04 2.794E-05 7.628E-03 7.628E-03 2.803E-03 1133 7.119E-02 2.766E-04 2.498E-02 6.536E-03 6.536E-03 4.011E-03 o.<
1134 7.839E-02 3.044E-04 3.467E-10 1.380E-02 1.380E-02 4.417E-03 11351 6.725E-02 2.612E-04 4.233E-03 9.075E-03 9.075E-03 3.789E-03 Iodine subtotal 3.0061-01 1.168E-03 5.363E-02 3.802E-02 3.802E-02 1.694E-02 CS137 3.583E-03 1.671E-05 3.360E-03 1.198E-05 1.198E-05 2.019E-04 TE132 8.178E-03 O.OOOE+00 O.OOOE+00 6.900E-04 6.900E-04 4.606E-04
_n l
-SR89 2.132E-03 O.OOOE+00 O.OOOE+00 1.489E-04 1.489E-04 1.201E-04 LU lSR 90 2.228E-04 O.OOOE+00 0.000E+00 1.126E-05 1.126E-05 1.255E-05 Bal40 4.094E-03 O.OOOE+00 0.00012+00 4.358E-04 4.358E-04 2.306E-04 40La4 4.336E-05 O.OOOE+00 0.0002+00 O.OOOE+00 O.OOOE+00 2.443E-06 P ariculate subtotal 1.83E-02 1.67E-05 3.36E-03 1.30E-03 1.30E-03 1.03E-03 RELEASE RATE TOTALS (CVsec) 1.01E+02 1.19E-01 7.02E+01 4.14E-02 4.14E-02 8.12E-01 Accident Duration Used for EDAMS 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 4 hours 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 2 hours 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 2 hours TOTAL Release Assumed (Ci) 2.92E+0B 1.71E+03 5.05E+05 2.98E+02 2.98E+02 5.84E+03 LOCA -Enginee S1 Loss of Coolant ntrol Rod Drop Refuellng Steam Lhe Stem Lne Brak Feature Component RATIOS Acciden Cnto Accldnt Two Phase Leakage o
Iodine I Noble Gas Ratio 2.98E-03 9.93E-03 7.65E-04 1.79E+01 1.79E+01 2.13E-02 Noblegas/lIodlneRatio 3.36E+02 1.012+02 1.31E+03 5.58E-02 5.58E-02 4.69E+01 Noble Gas / ParUculate RatIo 5.53E+03 7.04E+03 2.09E+04 1.63E+00 1.63E+00 7.72E+02 Iodine / Particulate Ratio 1.65E+01 6.99E+01 1.602E+01 2.93E+01 2.93E+01 1.652+01
___NG
/ Particulate + Iodine Ratlo 3.17E+021 9.932+01 1.23E+031 5.3912-021 5.39E-021 4.42E+01 Page 26 of 27 I
I Rev.
No.
16
CALCULATION TO DETERMINE EFFLUENT MONITOR READING TO REACH 100% OF TECH SPECS FOR NOBLE GAS AND IODINE Page 1 of 1 NOTE:
100% Noble Gas STACK release rate equals 0.3 Ci/sec 100% Noble Gas BUILDINb VENT release rate equals 0.0698ti ic 100% IODINE release rate from vent or stack equals 2.47 Ci/sec
- 1. 100% Noble gas - STACK reading:
0.3 Ci/sec Equation: flowcorrected k factor Workspace:
0.30C flow ci
=monitor value in CPS Isec
,rrected k factor nonitor value in CPS
- 2. 100 % Noble gas - BUILDING VENT reading:
0.0698Ci/sec flow corrected k factor =monitor value in CPM Workspace:
=
monitor value in CPM flow corrected k factor
- 3. 100% Iodine release rate - any source:
noblegas releaseratein Cilsec flowcorrected k factor 2.47Ci/sec Eruation Iodine/ Noble gasi monitorvaluein CPS or CPM
= noblegasreleaseratein Ci/sec ratio 2.47 C/sec I_INGratio noble gas release rate in Ci / sec flow corrected k factor MONITOR k - Factor (cfm)
Rx Bldg below 3.2E-7 Ci/sec/cpm 61,000 refuiel floor Refuel floor 3.7E-7 Ci/sec/cpm 70,000 Radwaste Bldg 1.7E-7 Ci/sec/cpm 32,500 Turbine Bldg 5.6E-7 Ci/sec/cpm 107,000 Stack 6.OE-7 Ci/sec/cps 6,600 noblegas releaserate in Ci/sec monitor value in CPS or CPM Accident Type Iodine/Noble LOCA 2.98E-3 Control Rod Drop 9.93E-3 Refuel Accident 1.24E-4 Steam Line Break 1.79E+1 Containment Design Basis 2.13E-2 Accident Equation: K(flowcorrected) = newflow x (K listed) normalflow new flow qn/Mwo x
normal flow K listed =
K corrected NOTE: For building vents units are Counts Per Minute (CPM)
For the stack units are Counts Per Second (CPS)
Equation:
Workspace:
Work.
Rev. No.
16 RELEASE RATE DETERMINATION Page 27 of 27 I
I 00698Cse
EMERGENCY PLAN IMPLEMENTING PROCEDURESIVOLUME 3 UPDATE LIST I Fnf%TjTQT T T-%
I I I\\'JL,Ld LJ '
rL X rt,z I Date of Issue:
MAY 30, 2003 Procedure Procedr RiSioMaeof Ueo Last-Number T itle
.Nuiher e
"jProcedure N/A TABLE OF CONTENTS REV. 23 12/98 N/A EAP-26 PLANT DATA ACQUISITION SYSTEM REV. 12 11/02 Informational ACCESS__
ESTIMATION OF POPULATION DOSE EAP-27 WITHIN 10 MILE EMERGENCY PLANNING REV. 10 06/02 Informational ZONE EAP-28 EMERGENCY RESPONSE DATA SYSTEM REV. 6 0700 Reference EAP-28 (ERDS) AICTIVATION REV.__6
_07_00_Reference EAP-29 EOF VENTILATION ISOLATION DURING RV 50 nomtoa EAP-29 AN EMERGENCY REV. 6 05/03 Informational EAP 30 TASTOTORCVY*REV.
1I 7
05/03 Inforrnational TRANSITION TO RECOVERY*
EAP-31 RECOVERY MANAGER*
REV. 2 05/03 Inforrational EAP-32 RECOVERY SUPPORT GROUP*
REV. 9 05/03 Infornational DEVELOPMENT OF A RECOVERY ACTION 0
r EAP-33 PLAN*
-REV.
105103 Inforrnational ACCEPTANCE OF ENVIRONMENTAL REV 4 EAP-34 SAMPLES AT THE EOF/EL DURING AN 05/03 Informational EMERGENCY EAP-35 EOF TLD ISSUANCE DURING AN REV. 7 EMERGENCY
-05/03 Inorniational EAP-36 ENVIRONMENTAL LABORATORY USE REV.5 05/03 Informational DURING AN EMERGENCY R._53 Ioan SECURITY OF THE EOF AND EL DURING REV 0
EAP-37 DRILLS, EXERCISES AND ACTUAL 02/03 Infomiational EVENTS EAP-39 DELETED (02/95)
EAP-40 DELETED (02/98)
EAP41 DELETED (12/85)
EAP-42 OBTAINING METEOROLOGICAL DATA REV. 19 05/03 Informational EAP-43 EMERGENCY FACILITIES LONG TERM REV. 60 05/03 Infomiational STAFFINGRE.6050 Inomtna EAP-44 CORE DAMAGE ESTIMATION REV. 5 05/03 Infornational EMERGENCY RESPONSE DATA SYSTEM EAP-45 (ERDS CONFIGURATION CONTROL REV. 6 07/00 Informational PROGRAM)
SAP-I MAINTAINING EMERGENCY REV. 17 02/03 Informational PREPAREDNESS 02/03 Inforrnabonal SAP-2 EMERGENCY EQUIPMENT INVENTORY REV. 35 01/03 Reference SAP-3 EMERGENCY COMMUNICATIONS REV. 73 02/03 Reference SAP-3____
TESTING 02/03 Reference Page 1 of2 i
t 4n)1 44 (d I
EMERGENCY PLAN IMPLEMENTING PROCEDURESNOLUME 3 UPDATE LIST Date of Issue:
MAY 30. 2003 Rev.-.-
-Risi Date of Use o Proced'ure
_Proced'ure Last1"Rviio,U~
Numbert
-;. :; ;; :: -. -: -^ Title.
0 :.~; 0 X 0.
f.5 Number:
Procedure SAP-4 NYS/OSWEGO COUNTY EMERGENCY Informational PREPAREDNESS PHOTO IDENTIFICATION REV. 10 05/03 CARDS SAP-5 DELETED (3/98)
SAP-6 DRILL/EXERCISE CONDUCT REV. 19 03/03 Infonnational SAP-7 MONTHLY SURVEILLANCE PROCEDURE Informational FOR ON-CALL EMPLOYEES REV. 36 08/02 SAP-8 PROMPT NOTIFICATION SYSTEM Informational FAILURE/SIREN SYSTEM FALSE REV. 13 12/02 ACTIVATION SAP-9 DELETED (02/94)
SAP-10 METEOROLOGICAL MONITORING REV. 1 03/02 Informational SYSTEM SURVEILLANCE SAP-1l EOF DOCUMENT CONTROL REV. 11 06/02 Informational SAP-13 EOF SECURITY AND FIRE ALARM Informational SYSTEMS DURING NORMAL OPERATIONS REV. 4 06/02 SAP-14 DELETED (02/95)
SAP-15 DELETED (11/92)
SAP-16 UTILIZING EPIC IDT TERMINALS FROM Infonnational DESTINY SYSTEM REV. 4 06/02 SAP-17 EMERGENCY RESPONSE DATA SYSTEM Continuous (ERDS) QUARTERLY TESTING REV. 7 07/00 SAP-19 SEVERE WEATHER REV. 4 01/01 Infonnational SAP-20 EMERGENCY PLAN ASSIGNMENTS REV. 22 05/03 Informational SAP-21 DELETED (04/01)
SAP-22 EMERGENCY PLANNING PROGRAM SELF REV. 2 05/03 Informational ASSESSMENT REV.
05/03 Page 2 of 2 I
ENTERGY NUCLEAR OPERATIONS, INC.
JAMES A. FITZPATRICK NUCLEAR POWER PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE CORE DAMAGE ESTIMATION EAP-44 REVISION 5 APPROVED BY:
4Ž DATE:
/?
RESPONSIBL-'fCEDURE -OWNT EFFECTIVE DATE:
FIRST ISSUE a
3
,Cb3 tULL REVISION LIMITED REVISION PERIODIC REVIEW DUE DATE:
dgo INFORMATIONAL USE TSR
- i ADMINISTRATIVE lCONTROLLED COPY#
X Si. AL~01
CORE DAMAGE ESTIMATION REVISION
SUMMARY
SHEET REV. NO.
5 4
- Updated pertinent plant parameters in section 4.5.1.
- Updated performance references in section 4.8.2
- Reformat per AP-02.01, Rev.5
- Change total Zr to 9.71E04 lbs in step 4.12.5 -
changes are as a result of power uprate.
- Change core thermal power from 2436 Mwt -
changes are as a result of power uprate.
- Change total daily gross thermal generation in Mwt-hr to 60,864 -
changes are as a result of power uprate.
- Increase core inventories on Attachment 6 by 4.1 changes are as a result of power uprate.
- Change level of use to "informational" per AP-02.04.
Page 2
of 72 EAP-44 Rev. No.
CORE DAMAGE ESTIMATION TABLE OF CONTENTS SECTION 1.0 2.0 2.1 2.2 3.0 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 4.10 4.11 4.12 4.13 4.14 Rev. No.
PAGE PURPOSE..............................
5 REFERENCES..............................
5 Performance References.............................
5 Developmental References.............................
5 REQUIREMENTS...............
6 PROCEDURE
.6 Methods for Estimating Core Damage.
6 Core Damage Mechanisms and Fission Product Sources 6
General.
7 Characteristic Nuclides 7
Plant Parameters
.8 Calculation of Core Inventories
.9 Isotope Decay Correction.13 Estimation Procedure Preliminary Guide for PASS Samplesl4 Core Damage Estimation Using Gaseous PASS Samples.
15 Core Damage Estimation Using Liquid PASS Samples.22 Core Damage Estimation Using High Range Containment Monitor Readings (HRCM).27 Core Damage Estimation Using Containment Atmosphere Hydrogen Concentration....................
....... 31 Integration of Other Parameters Into The Estimate.
36 Development of a Final Estimate.39 5
Page 3 of 72 EAP-44
CORE DAMAGE ESTIMATION EAP-44 5.0 ATTACHMENTS 41
- 1.
SUMMARY
OF CORE DAMAGE ESTIMATION TECHNIQUES..
. 42
- 2.
DESCRIPTION OF CORE RELEASE COMPONENTS.
43
- 3.
BEST ESTIMATE FISSION PRODUCT RELEASE FRACTIONS.. 44
- 4.
ISOTOPE ATTRIBUTES 45
- 5.
CHARACTERISTIC ISOTOPES 47
- 6.
CORE INVENTORIES (CI) 48
- 7.
FUEL CYCLE DATA 50
- 8.
SAMPLES MOST REPRESENTATIVE OF CORE CONDITIONS DURING AN ACCIDENT FOR THE ESTIMATION OF CORE DAMAGE.
.. 51
- 9.
CATEGORIES OF CORE DAMAGE EVENTS.
.. 52
- 10. RATIOS OF ISOTOPES IN CORE INVENTORY AND FUEL GAP.
.. 53
- 11. FISSION PRODUCT CONCENTRATIONS IN REACTOR WATER AND DRYWELL GAS SPACE DURING REACTOR SHUTDOWN UNDER NORMAL CONDITIONS.
.. 54
- 12. TECHNICAL BASIS AND REFERENCE GUIDE FOR CORE DAMAGE ESTIMATION.
.. 55
- 13.
TYPICAL HYDROGEN PRODUCTION RATE FROM ALUMINUM AND ZINC VS. TEMPERATURE..................................
.. 65
- 14.
SPECIFIC RADIOLYTIC HYDROGEN PRODUCTION VS. TIME.......
.. 66
- 15.
PERCENT OF FUEL RODS WITH RUPTURED CLAD VS. CORE CLAD OXIDATION.......................................
.. 67
- 16.
PERCENT OF THE FUEL RODS WITH OXIDATION EMBRITTLEMENT VS TOTAL CORE OXIDATION..............................
.. 68
- 17.
100% NOBLE GAS AIRBORNE IN DRYWELL..................
.. 69
- 18.
25% HALOGENS AIRBORNE IN DRYWELL...................... 70
- 19. 50% HALOGENS DILUTED IN TORUS WATER.................
.. 71
- 20. 1% REMAINING FISSION PRODUCTS DILUTED IN TORUS WATER
..... 72 Page 4
of 72 1,
Rev. No.
CORE DAMAGE ESTIMATION EAP-44 1.0 PURPOSE The purpose of this procedure is to describe three methods for estimating core damage post accident conditions.
The first method involves using the measured fission product concentrations in either water or gas samples taken from the primary system.
The second method involves use of the High Range Drywell Monitor readings.
The third method involves containment atmosphere hydrogen concentration.
This procedure also describes the technical basis of each method, including a discussion of its limitations.
Each method provides the capability to assess the degree of core damage and place it in one of the ten predefined NRC categories of core damage (i.e., defined categories of No Fuel Damage, Cladding Failure, Fuel Overheat or Fuel Melt).
2.0 REFERENCES
2.1 Performance References 2.1.1 PSP-17, PASS OPERATING PROCEDURE 2.1.2 AM-03.02, POST ACCIDENT SAMPLE ANALYSIS 2.2 Developmental References 2.2.1 Compliance with NUREG-0737, Item II.B.3 (Ref. 5.2) 2.2.2 NEDO-22215, Procedures for the Determination of the Extent of Core Damage Under Accident Conditions, August, 1982 2.2.3 Memo RES-83-0279, NUREG-0737 Item II.B.3, BWRDG-8324 (June 17, 1983) Attachment 2, Integration of Other Plant Parameters into Core Damage Estimate 2.2.4 PSP-17, PASS OPERATING PROCEDURE 2.2.5 AM-03.02, POST ACCIDENT SAMPLE ANALYSIS 2.2.6 Richard Bradshaw, "Core Damage Assessment-and PASS," Presentation Handout, Combustion Engineering, Inc.
2.2.7 "Reactor Safety Study; An Assessment of Accident Risk in U.S. Commercial Nuclear Power Plants",
WASH-1400 (NUREG-75/014), U.S. NRC, October 1975 Rev. No.
5 Page 5 of 72
CORE DAMAGE ESTIMATION EAP-44 2.2.8 D.C. Kocher, "Radioactive Decay Data Tables",
Health and Safety Research Division Oak Ridge National Laboratory, 1981 2.2.9 J. Stoer and R. Bulirsch, "Introduction to Numerical Analysis", Springer Verlag, 1980 2.2.10 J.M. Rich, B.J. Andrews, G.P. Lahti, and D. J.
Pichurski, "High Range Containment Monitor Response to Post Accident Fission Product Releases", SL-4370, Sargent and Lundy, May 24, 1985 REFERENCES 3.0 REQUIREMENTS None 4.0 PROCEDURE 4.1 Methods for Estimating Core Damage 4.1.1 There are three basic techniques which can be used to estimate the degree of core damage:
A. Primary system sample isotopic concentration, B. Drywell dose rate, C. Containment atmosphere hydrogen concentration 4.1.2 These techniques may be used independently or in parallel to estimate core damage.
4.1.3 This procedure addresses all three techniques.
Each of the three techniques has different capabilities.
A brief summary of the capabilities of each technique is contained in Attachment 1.
4.2 Core Damage Mechanisms and Fission Product Sources 4.2.1 The mechanism of fission product release from a degraded fuel rod is identified through the presence of characteristic fission products in the sample medium.
4.2.2 The identification of the source of fission product release is used to verify the release mechanism determined by isotope identity.
Page 6 of 72 Rev. No.
5
CORE DAMAGE ESTIMATION EAP-44 4.2.3 Reference 2.2.7 describes four conditions or times during core melt at which major driving forces for release exist.
The four major release components are Gap Release, Meltdown Release, Vaporization Release and Oxidation Release.
A brief description of the physical processes occurring during each of these release components is shown in Attachment 2. The fraction of total fission product inventory released during each of these four major release components is shown in for each of eight fission product elemental groupings.
4.3 General 4.3.1 Core damage will not take place uniformly among
-all the fuel rods; thus, a combination of fuel damage categories may exist simultaneously.
4.3.2 Calculation of isotopic inventories in the primary system at post scram times is not always straight forward.
In some cases, isotopes have a strong neutron removal cross section while the reactor is operating.
4.3.3 Consideration must be given to including the contribution of remnant activity from preceding fuel cycles for longer lived isotopes.
The current core load and the two preceding core loads will contribute to the total activity.
4.4 Characteristic Nuclides 4.4.1 All nuclides listed in each of the element groups of Attachment 3 were reviewed to determine which are best suited to represent that element group.
Several attributes of each isotope were evaluated during-this review process.
The isotopes reviewed, their attributes-and results of the evaluation are shown in Attachment 4. The attributes reviewed are discussed below.
Rev. No.
5 Page 7 of 72
CORE DAMAGE ESTIMATION 4.4.2 4.5 Plant Pa3 4.5.1 4.5.2 4.5.3 Rev. No.
5 Based on this review, the best characteristic isotopes were chosen for each elemental group and are shown in Attachment 5 in order of preference.
It was not always possible to choose a characteristic isotope for a particular element group which met all of the criteria because of a limited selection.
- ameters The pertinent plant parameters for the FitzPatrick plant are given below:
The inventories for some major fission products in the full new core of the FitzPatrick Plant at the time of the reactor shutdown are shown in for those various durations of full power operation.
For the purposes of this procedure, power levels over long periods of time will be measured in units of Effective Full Power Days (EFPD).
It is assumed that EFPD electric is equivalent to EFPD thermal.
Any error inherent in this assumption is negligible in comparison to the accuracy of the estimation methods described in this procedure.
Page 8 of 72 FitzPatrick Plant Rated Reactor Thermal Power Level 2536 MWt Number of Fuel Bundles 560 Total Primary Coolant Mass 3.21E9 g (Reactor Water Plus Suppression Pool Water)
Reactor Water Mass 2.14E8 g Suppression Pool Water Mass 3.00E9 g Primary Containment Free Volume 7.48E9 cc (Torus Plus Drywell Free Volumes)
Drywell Free Volume 4.25E9 cc Torus Free Volume 3.23E9 cc EAP-44
CORE DAMAGE ESTIMATION EAP-44 4.5.4 Daily thermal output will be measured using gross thermal generation in MWt-hr.
This information is readily obtained from the daily report.
The maximum gross thermal generation output is assumed to be 2536 MWt or 60,864 MWt-hr in one day.
4.6 Calculation of Core Inventories 4.6.1 Since a reactor does not run continuously at a single power level, it is necessary to correct the core inventories in Attachment 6 for power level and operating history before using them.
4.6.2 only the current fuel cycle need be considered for isotopes with half-lives less than fifteen days.
4.6.3 The residual radioactivity from the two previous fuel cycles must also be considered for isotopes with half-lives greater than fifteen days.
A. In order to perform this correction, it is necessary to maintain a file of operating history for the current and two previous fuel cycles.
B. The required fuel cycle data is shown in.
4.6.4 Correction for current fuel cycle operating history is performed as follows:
A. Obtain the half-life of the isotope of interest from Attachment 4.
B. Review the current fuel cycle operating history from the current date backwards to determine the continuous power irradiation period.
- 1. The first extended outage encountered during this period ends the continuous
.power irradiation period.
- 2. An extended outage is defined as an outage lasting longer than four half-lives of the isotope of interest.
Rev. No.
5 Page 9 of 72
CORE DAMAGE ESTIMATION
- 3. The continuous power irradiation period should be rounded to the nearest day and can be no shorter than one day and no longer than the length of the current fuel cycle.
- 4. The continuous power irradiation history is the span of time from the current date backwards to the first extended outage.
C. If the continuous power irradiation period matches one of the operating times shown in then the isotope core inventory can be obtained directly from the table.
D. If not, then an interpolated value must be calculated. Newton's interpolation formula of divided differences is used to perform the interpolation as follows (from Ref. 2.2.9):
where:
to
=
Closest operating time period on Attachment 6 less than the continuous power irradiation period, (days) tl, t2 The next two operating time periods from Attachment 6 after to, (days)
CO, C, C2
=
The core inventories of the isotope of interest from associated with the operating time periods to, t1 and t2, (Ci) t The operating time period of interest, (days)
C
=
The core inventory of the isotope of interest associated with operating time period t, (Ci)
Fol = C1 Co, F12
= C2 C1 F
2
= F1 2 Fol t-to t 2
-tl t2
-to C = C
+F01 (t-to) + F012 (t-to)
(t-tl)
Page 10 of 72 EAP-44 Rev. No.
CORE DAMAGE ESTIMATION EAP-44 4.6.5 Correction for power level history is performed as follows:
A. Multiply the half-life of the isotope of
-interest by six and round off to the nearest day.
- 1. This is the power level history period of interest.
- 2. It is a minimum of one day and a maximum equal to the continuous power irradiation period previously calculated, beginning with the current day.
B. Determine the total thermal output of the plant during the power level history period of interest by summing the daily thermal outputs from Attachment 7.
C. The power history corrected core activity is calculated as follows:
CF (C)(P)
(60,864)(tp) where:
CF Total corrected core activity of the isotope of interest, (Ci)
C Core activity of the isotope of interest calculated in 4.6.4.D, (Ci)
P
=
Total gross thermal output calculated in 4.6.5.B, (MWt-hr) tp
=
Operating days of the power level history period calculated in 4.6.5.A, (days) 60864
=
Maximum daily gross thermal power output of plant (MWt-hr)/day Rev. No.
5 Page 11 of 72
CORE DAMAGE ESTIMATION 4.6.6 An additional correction for residual activity from prior fuel cycles is required when using isotopes with half-lives greater than fifteen days.
This correction factor is calculated using the information in Attachment 7 as follows:
A. Determine the core activity levels C2 and C for the two prior fuel cycles using t2 and t1 respectively, as the continuous power irradiation periods (Attachment 7).
B. Calculate the core inventory for the isotope of interest for each prior fuel cycle using the interpolation method of 4.6.4.D.
C. Correct the core inventories calculated in 4.6.6.B above for power history as follows:
CR1 =
(EFPDl)(Cl) ti CR2 =
(EFPD2) (C2) t2 where:
C1
=
Core inventory for fuel cycle current-1, calculated in 4.6.6.B (Ci)
C2
=
Core inventory for fuel cycle current-2, calculated in 4.6.6.B (Ci) ti
=
Duration of power level history, duration of fuel cycle current-1, from Attachment 7, days) t2
=
Duration of power level history, duration of fuel cycle current-2, from Attachment 7, (days)
EFPD1
=
Total power of fuel cycle current-1, from Attachment 7, (effective full power days)
EFPD2
=
Total power of fuel cycle current-i, from Attachment 7, (effective full power days)
CR2
=
Core inventory for fuel cycle current-2 corrected for power level history, (Ci)
CR1
=
Core inventory for fuel cycle current-1 corrected for power level history, (Ci)
Page 12 of 72 EAP-44 Rev. No.
5
CORE DAMAGE ESTIMATION EAP-44 D. Calculate the residual activity from the two preceding fuel cycles.
CRF2 = CR2__e i(R2,1 + tl
+ R1,c + tc) 3 CRF1 =
2 CR1 ei(R3,c
+ tc) 3 where:
R2,1= Refueling outage duration between fuel cycle current-2 and fuel cycle current-1, (days)
Rl,c =
Refueling outage duration between fuel cycle current-1 and the current fuel cycle, (days) ti
=
The total duration of fuel cycle current-1, (days) tc
=
The total duration of the current fuel cycle up to the present day, (days)
CRF2 =
The current residual activity from fuel cycle current-2, (Ci)
CRF1 =
The current residual activity from fuel cycle current-1, (Ci)
Xi
=
The decay constant for isotope "i",
(days)
E. The total current core inventory of isotope "i" is obtained by adding CRF2 and CRF1 to CF calculated in 4.6.5.C.
4.7 Isotope Decay Correction 4.7.1 The concentration of any isotope detected in a PASS sample must be decay corrected to the time of reactor shutdown before the Core Damage Estimation procedure can be applied.
4.7.2 Supply the counting room technician with the time of sample collection and time of reactor shutdown for automatic isotope decay correction or else decay correct as shown in the following step.
Rev. No.
5 Page 13 of 72
CORE DAMAGE ESTIMATION EAP-44 4.7.3 Correct the measured liquid or gaseous concentration (C, or C) for decay to the time of reactor shutdown, using the following equations:
Cd
= C, et or Cdg = Cg eXit where:
Cdw
=
The liquid concentration of isotope "i" corrected for decay, (pCi/g)
Cdg
=
The gaseous concentration of isotope "i" corrected for decay, (pCi/cc)
A\\i
=
The decay constant of isotope "i" from, (hr-1) t
=
The time between the reactor shutdown and the midpoint of sample counting, (hrs) 4.8 Estimation Procedure Preliminary Guide for PASS Samples 4.8.1 Obtain samples from the Post Accident Sampling System (PASS), following the procedure outlined in PSP-17 (Ref. 2.2.4).
A. It is recommended that both the water and gas phase samples be taken and analyzed in order to reduce the uncertainty in core damage estimations.
B. Samples acquired for the estimation of core damage should be taken from locations that are consistent with the line break case and system conditions, as outlined in Attachment 8. This will ensure the viability of results reported and provide the best estimation of core damage.
4.8.2 Perform gamma ray spectrometry on the samples and determine the concentration, in pCi/g or pCi/cc of a fission product i, as outlined in AM-03.02 (Ref.
2.1.1).
A. In water, the concentration is represented as C,.
B.
In gas, the concentration is represented as Cd.
Rev. No.
5 Page 14 of 72
CORE DAMAGE ESTIMATION EAP-44 4.9 Core Damage Estimation Using Gaseous PASS Samples 4.9.1 Ensure that any isotopic concentrations used in this section have been decay corrected to the time of reactor shutdown as described in 4.7.
4.9.2 Gaseous activity concentrations in the drywell and torus may have to be corrected for pressure and temperature.
A. This correction must be made if significant differences exist between the sample vial temperature and pressure, and drywell or torus temperature and pressure.
B. If both drywell and torus samples are available, this correction must be applied to each separately as follows:
Cdgt = (Cdg) (P2) (T1)
(P1)(T2) where:
Cdgt =
Gaseous isotope concentration corrected for decay and temperature and pressure, (pci/cc)
Cdg =
Gaseous isotope concentration corrected for decay (4.7), (pCi/cc)
P1
=
Sample vial pressure (psia = psig +
14.7)
P2
=
Drywell or torus atmospheric pressure (psia = psig + 14.7)
Tl
=
Sample vial temperature (ER = EF + 460)
T2
=
Drywell or torus atmospheric temperature (ER = EF + 460)
Rev. No.
5 Page 15 of 72
CORE DAMAGE ESTIMATION 4.9.3 If either a drywell or torus atmosphere sample is available, it is assumed to be applicable to both atmospheric volumes.
A. If there is reason to assume that little or no mixing has occurred between the two air volumes then the drywell sample should be used.
B. If both drywell and torus atmospheric samples are available, then the total gaseous activity of isotope "i" should be determined using both samples.
C. If there is a big difference in sampling time or a significant reactor change has occurred between samples, then only the most recent sample should be used.
The activity is calculated as follows:
A. one sample (good mixing):
Ag =
(Cdgt)(Drywell + Torus free volume) (106)
B. one sample (poor mixing):
Ag =
(Cdgt) (Free volume) (10 6)
C. two samples:
Ag =
[(Cdgt) (Drywell free volume) +
(Cdgt) (Torus free volume)] (106) where:
Ag =
total primary containment gaseous activity of isotope "i", (Ci) 10-6
=
conversion factor microcuries to curies, (Ci/uci)
Rev. No.
5 Page 16 of 72 I
CORE DAMAGE ESTIMATION EAP-44 Choose one or more Noble gas isotope(s) from that was detected in the gaseous PASS sample.
A. Ensure that steps 4.9.1 through 4.9.3 have been performed on the measured concentration of this isotope.
B. Calculate the core inventory of this isotope using the methods described in Section 4.6.
Perform the following calculations:
NG1
=
Ag (O. 03) (CF) x 100, NG2
=
Ag x 100, NG3
=
Ag (O.01) (CF)
(0. 12) (CF )
= Total corrected core activity of isotope of interest calculated in 4.6.5.C or 4.6.6.E (Ci)
= Best estimate of % cladding failure
= Upper estimate of cladding failure
= Lower estimate of % cladding failure Page 17 of 72 4.9.4 x 100 where:
CF NG, NG2 NG3 Rev. No.
5
CORE DAMAGE ESTIMATION For each gaseous sample obtained using PASS, determine the concentrations of the following isotopes listed below:
A. Decay correct the concentrations to the time of reactor shutdown as described in 4.7.
B. It is not necessary to perform the pressure and temperature correction.
Calculate the following concentration ratios:
Ratio Kr-87 concentration Xe-133 concentration Kr-88 concentration Xe-133 concentration Kr-85m concentration Xe-133 concentration N/A N/A I-132 concentration I-131 concentration I-133 concentration I-131 concentration I-134 concentration I-131 concentration I-135 concentration I-131 concentration Page 18 of 72 4.9.5 Isotope Kr-87 Kr-88 Kr-85m Xe-133 I-131 I-132 I-133 I-134 I-135 EAP-44 Rev. No.
5
CORE DAMAGE ESTIMATION EAP-44 Compare the ratios calculated in 4.9.5 above to those shown in Attachment 10 to determine the release source.
A. Fuel cladding failure is assumed if the source is the fuel gap.
B. At least some fuel pellet overheating or melting is assumed if the source is the core.
Were any isotopes from the alkali metals and tellurium group detected in the sample?
If yes, then perform steps 4.9.1 through 4.9.4 on one or more of these isotopes.
Perform the following calculations:
H1
=
Ag (N) (CF) x 100 H2 Ag(LL) (CF) x 100, H3
=
where:
H1, H2, H3
=
Best, upper and lower estimate of % fuel pellet overheating N, LL, UL
=
Nominal, lower and upper limits of fuel pellet melt release from Attachment 3.
4.9.8 Were any isotopes from the noble metals, alkaline earths, rare earths or refractories detected in the sample?
If yes, then perform steps 4.9.1 through 4.9.4 on one or more of these isotopes.
Perform the following--calculations:
Ag x 100, M2 =
Ag (N) (CF)
=
where:
MI, M2, M3 N, LL, UL
= Best, upper and lower estimate of fuel pellet melting
= Nominal, lower and upper limits of fuel pellet melt release from Attachment 3 Page 19 of 72 4.9.6 4.9.7 Ag (UL) (CF) x 100 x 100 Ag (UL) (CF)
Rev. No.
5
CORE DAMAGE ESTIMATION 4.9.9 Based on the results of the analysis performed in 4.9.1 through 4.9.8, place the core damage into one of the categories described in Attachment 9.
Use the following guidance to assist in this categorization process:
A. After completing the analysis of 4.9.4.
- 1. If the upper estimate of % cladding failure is greater than 100% then some fuel pellet overheat has definitely occurred, proceed to 4.9.9.C below.
- 2. If it is less than 100%, then fuel pellet overheating may have occurred, proceed to 4.9.9.B below.
B. Complete the analysis described in 4.9.5 and 4.9.6.
- 1. If the activity ratio calculated in these steps is more than double that of the fuel gap activity ratio, it is assumed that some fuel pellet overheating has definitely occurred, proceed to 4.9.9.C below.
- 2. If the activity ratio is less than double the gap ratio, it is assumed that fuel pellet overheating has not occurred and only clad failure damage has occurred.
- 3. Use the best estimate of percent cladding failure calculated in 4.9.4 to assign a core damage category from Attachment 9.
- 4. If the best estimate result is greater than 100%, assign core damage category 4.
Rev. No.
5 Page 20 of 72 t -
CORE DAMAGE ESTIMATION EAP-44 C. If the upper estimate of % clad failure is greater than 100% or the activity ratio is more than double that of the fuel gap activity, then it is assumed that some fuel pellet overheating has occurred.
- 1. The best estimate of fuel pellet overheating may be used to determine the percent of fuel pellets which have undergone initial overheating.
- 2. Since all noble gases and iodine are released during initial fuel overheat, these isotopes cannot be used to predict core damage past initial fuel overheat (category 5).
- 3. Thus, all positive determinations of fuel pellet overheat using iodine and noble gases are classified as core damage category 5, regardless of the percent of fuel pellets which have undergone initial overheat.
D. Use the best estimate of percent of fuel pellet overheating from 4.9.7 in conjunction with to determine the category of core damage due to fuel overheat.
E. Use the best estimate of percent of fuel pellet melting from 4.9.8 in conjunction with to determine the category of core damage due to fuel pellet melting.
4.9.10 There are several areas of concern, which may affect the accuracy of core damage estimation using gaseous PASS samples.
These issues must be considered when determining the validity of the core damage estimation.
The major areas of concern are found in Section 5.0 of Attachment 12 to this procedure.
Fission product concentrations under normal conditions are presented in 1 for additional information and guidance.
Rev. No.
5 Page 21 of 72
CORE DAMAGE ESTIMATION 4.10 Core Damage Estimation Using Liquid PASS Samples 4.10.1 Ensure that any isotopic concentrations used in this sample have been decay corrected to the time of reactor shutdown as described in 4.7.
4.10.2 If either a reactor vessel or torus water sample is available, it is assumed to be applicable to both water volumes.
A. If there is reason to assume that little or no mixing has occurred between the two water volumes then the vessel sample should be used.
B. If both reactor vessel and torus water samples are available, then the total water activity of isotope "i" is determined using both samples.
C. If there is a big difference in sampling time or a significant reactor change has occurred between samples, then only the most recent sample should be used.
The activity is calculated as follows:
- 1. one sample (good mixing):
Ag =
(Cdw) (Reactor Vessel + Torus Water mass) (106)
- 2. one sample (poor mixing):
(NOTE: Only applicable when sample is from the vessel)
Ag =
(Cdw) (Reactor Vessel Water mass) (10-6)
- 3. two samples:
Ag
=
[(Cdw)(Reactor Vessel Water mass) +
(Cdw)(Torus Water mass)](10-6) where:
Ag
= Total primary containment liquid activity of isotope "i",
(Ci) 10-6 = conversion factor microcuries to curies, (Ci/uci)
Page 22 of 72 EAP-44 Rev. No.
5
CORE DAMAGE ESTIMATION IEAP-44 4.10.3 Choose one or more iodine isotopes from Attachment 5 that were detected in the liquid PASS sample.
A. Ensure that steps 4.10.1 and 4.10.2 have been performed on the measured concentration of this isotope.
B. Calculate the core inventory of this isotope using the methods of 4.6.
Perform the following calculations:
I1=
Ag x 100, I2
=
(0. 017) (CF)
- Ag (0.001) (CF) x 100, I 3 Ag x 100 (0.20) (CF) where:
Ill I21 3
Best, upper and lower estimate of % clad failure CF
Total corrected core activity of isotope of interest calculated in 4.6.5.C or 4.6.6.E (Ci).
4.10.4 For each liquid sample obtained from PASS, obtain the concentrations of the following isotopes listed below:
A. Decay correct them to the time of reactor shutdown as described in 4.7.
Calculate the following concentration ratios.
Isotope I-131 I-132 I-133 I-134 I-135 Ratio N/A I-132 concentration I-131 concentration I-133 concentration I-131 concentration I-134 concentration I-131 concentration I-135 concentration I-131 concentration Page 23 of 72 Rev. No.
5
CORE DAMAGE ESTIMATION EAP-44 4.10.5 Compare the ratios calculated in 4.10.4 above to those supplied in Attachment 10 to determine the release source.
A. Fuel cladding failure is assumed if the source is the fuel gap.
B. At least some fuel pellet overheating or melting is assumed if the source is the core.
4.10.6 Were any isotopes from the Alkali Metals and Tellurium Group detected in the sample?
If yes, then perform steps 4.10.1 and 4.10.2 on one or more of these isotopes.
Perform the following calculations:
H (1
=
Ag (N) (CF) x 100, H2 =
Ag x 100, H3 (N) (CF)
Ag x 100 (UL) (CF) where:
H H2 H3 N, LL, UL
= Best, upper and lower estimate of % fuel pellet overheating
= Nominal, lower and upper limits of fuel pellet melt release from Attachment 3 4.10.7 Were any isotopes from the Noble Metals, Alkaline Earths, Rare Earths or Refractories detected in the sample?
If yes, then perform steps 4.10.1 and 4.10.2 on one or more of these isotopes.
Perform the following calculations.
M1 =
Ag x 100, M2 =
Ag x 100, M3 Ag x 100 (N) (CF) where:
(UL) (CF)
M1' 2' M3 N, LL, UL
= Best, upper and lower estimate of fuel pellet melting
= Nominal, lower and upper limits of fuel pellet melt release from Attachment 3 Page 24 of 72 I
Rev. No.
5
CORE DAMAGE ESTIMATION EAP-44 4.10.8 Based on the results of the analysis performed in 4.10.1 through 4.10.7, place the core damage into one of the categories described in Attachment 9.
Use the following guidance to assist in the categorization process:
A. After completing the analysis of 4.10.3.
- 1. If the upper estimate of percent cladding failure is greater than 100% then some fuel pellet overheat has definitely occurred, proceed to 4.10.8.C below.
- 2. If it is less than 100%, then fuel pellet heating may have occurred, proceed to 4.10.8.B below.
B. Complete the analysis described in 4.10.4 and 4.10.5.
- 1. If the activity ratios calculated in these steps is more than double that of the fuel gap activity ratio, it is assumed that some fuel pellet overheating has definitely occurred, proceed to 4.10.8.C below.
- 2. If the activity ratio is less than double the gap ratio, it is assumed that fuel pellet overheating has not occurred and only clad failure damage has occurred.
- 3. Use the best estimate of percent clad failure calculated in 4.10.3 to assign a core damage category from Attachment 9.
- 4. If the best estimate result if greater than 100%, assign core damage category 4.
Rev. No.
5 Page 25 of 72
CORE DAMAGE ESTIMATION C. If the upper estimate of percent clad failure is greater than 100% or the activity ratio is more than double that of the fuel gap activity, then it is assumed that some fuel pellet overheating has occurred.
- 1. The best estimate of percent fuel overheating may be used to determine the percent of fuel pellets which have undergone initial overheating.
- 2. Since all iodines are released during initial fuel overheat, these isotopes cannot be used to predict core damage past initial fuel overheat (category 5).
- 3. Thus, all positive determinations of fuel pellet overheat using iodines are classified as core damage category 5, regardless of the calculated percent of fuel pellets which have undergone initial overheat.
D. Use the best estimate of percent of fuel pellet overheating from 4.10.6 in conjunction with to determine the category of core damage due to fuel overheat.
E. Use the best estimate of percent of fuel pellet melting from 4.10.7 in conjunction with to determine the category of core damage due to fuel pellet melting.
4.10.9 There are several areas of concern, which may effect the accuracy of core damage estimation using liquid PASS samples.
These issues must be considered when determining the validity of the core damage estimation.
The major areas of concern are found in Section 6.0 of Attachment 12 to this procedure.
Fission product concentrations under normal conditions are presented in 1 for additional information and guidance.
Rev. No.
5 Page 26 of 72 EAP-44
CORE DAMAGE ESTIMATION, EAP-44 4.11 Core Damage,Estimation Using High Range Containment Monitor Readings (HRCM) 4.11.1 The HRCM response is due largely to,airborne noble gas and iodine isotopes.
Secondary contributors include halogens and other solids in the primary coolant piping and those accumulating on the drywell floor.
4.11.2 Most of these isotopes have half-lifes on the order of several hours to several days.
The core inventory of these isotopes is affected by the recent core power history as previously described in section 4.6.4.
4.11.3 Because the HRCM response is due to a spectrum of isotopes all with different half-lifes, an isotope specific correction factor cannot be used as described in section 4.6.4.
A. Instead, a power history correction factor must be based on the predominant isotopes affecting the HRCM reading-at a given time after core shutdown.
B. A power correction factor for each of three periods after core shutdown adequately accounts for core power history and variations in isotopic distributions.
C. The following Power,History Durations are specified:
Time After Power History Shutdown Duration 0-24 hrs.
2 days 1 day -
7 days 7 days
>7 days 30 days Rev. No.
5 Page 27 of 72
CORE DAMAGE ESTIMATION EAP-44 4.11.4 The power history correction factor is calculated using the following equation:
PH =
p (60,864)(tp) where:
PH p
60864 tp
= Power history correction factor
= Total thermal power output over last tp days from Attachment 7, (MWt)-hr
= Maximum daily gross thermal power output of the plant, (MWt-hr)/day
= Power history duration specified in 4.11.3 above, (days).
NOTE:
If the selected duration is less than latest full power operating period, set tp equal to the latter.
4.11.5 Calculation of core damage using HRCM reading requires use of the HRCM response curves from Reference 2.2.10.
A. Core clad failure is estimated based on assumptions described in Attachment 12 to this procedure.
Based on the time of reactor shutdown, obtain the following HRCM responses from Attachments 17-20 (as extracted from Ref.
2.2.10):
- 1. Due to noble gases airborne in the drywell from Attachment 17.
- 2. Due to halogens airborne in the drywell from Attachment 18.
- 3. Due to halogens in primary piping and liquids accumulating on the drywell floor 9.
Page 28 of 72 Rev. No.
5
CORE DAMAGE ESTIMATION EAP-44 The following equation is used to calculate the percent clad failure:
%CF =
(HRCM)(100)
[(NGR)(0.03) + (IR)(0.017)(SF+0.3) + (IRT)(0.017)] (PH) where:
%CF NGR IR IRT 0.03 0.017 0.3 SF PH HRCM
=
Percent of cladding failure in the core HRCM noble gas response from Attachment 17, (rad/hr)
HRCM iodine response from Attachment 18, (rad/hr)
HRCM response due to halogens in primary piping and liquids accumulating on the drywell floor, from Attachment 19 (rad/hr)
Nominal noble gas gap release fraction from Nominal iodine gap release fraction from Adjustment factor to account for contribution of halogens depositing on drywell internal surfaces Containment spray reduction factor, SF=1 if the containment sprays are not used, else the user may input any value less than 1 Power'history correction factor Actual HRCM reading at time specified after reactor'shutdown, (rad/hr)
Rev. No.
5 Page 29 of 72
CORE DAMAGE ESTIMATION B. Fuel pellet overheat core damage is based on assumptions described in Attachment 12 to this procedure.
Based on the time of reactor shutdown, obtain the following HRCM responses from Attachments 17-20 (as extracted from Ref. 2.2.10):
- 1. Due to noble gases airborne in drywell from 7.
- 2. Due to halogens airborne in the drywell from Attachment 18.
- 3. Due to halogens in primary piping and liquids accumulating on the drywell floor from Attachment 19.
- 4. Due to alkali metals and Te group nuclides in primary piping and liquids accumulating on the drywell floor from Attachment 20, with adjustment.
The following equation is used to calculate the percent fuel pellet overheat.
%PO =
(HRCM)(100)
[(NGR)(0.873)+(IR)(0.885)(SF+0.3)+(IRT)(0.885)+(RFPT)(7)]
(PH) where:
%PO
=
Percent of fuel pellet overheating in the core RFPT
=
HRCM response due to alkali metals and Te group nuclides in primary piping and liquids accumulating on the drywell floor, from 0 (rad/hr) 0.873
=
Nominal noble gas melt release fraction from 0.885
=
Nominal iodine melt release fraction from 7
=
Adjustment factor to reflect the HRCM response from the alkali metals and tellurium group released during core overheat.
Page 30 of 72 EAP-44 Rev. No.
5
CORE DAMAGE ESTIMATIONi EAP-44 4.11.6 Calculation of CF greater than 100 does not necessarily mean that 100% of the fuel cladding is damaged.
A. A similar analysis would result if less than 100% of the fuel cladding is damaged and a fraction of the fuel pellets has undergone initial overheating.
B. However, this method of core damage estimate cannot differentiate between the two scenarios.
4.11.7 Based on the calculations in paragraph 4.11.5.A and 4.11.5.B, place the core damage into one of the categories described in Attachment 9.
A. Because of the limitations of this method, the worst category which can be assigned is 5, initial fuel overheat.
B. Hence, any indication of fuel pellet overheat from 10% to 100% would place the damage in this category only and not any worse category.
4.11.8 There are several areas of concern, which may affect the accuracy of core damage estimation using HRCM readings.
These issues must be considered when determining the validity of the core damage estimation.
The major areas of concern are found in Section 8.0 of Attachment 12 to this procedure.
4.12 Core Damage Estimation Using Containment Atmosphere Hydrogen Concentration 4.12.1 There are multiple sources of hydrogen released during severe accidents.
Each of these sources must be evaluated to yield a determination of that amount of hydrogen which has been generated by core material oxidation.
These sources include:
A. Hydrogen present in-the reactor coolant for normal chemistry control.
B. Oxidation of various metals within the containment building.
C. The radiolytic decomposition of water.
D. Oxidation of the zirconium metal in the core.
Rev. No.
5 Page 31 of 72
CORE DAMAGE ESTIMATION 4.12.2 Oxidation of zirconium in the fuel clad occurs at all temperatures but is significant only above 1800EF.
The following two scenarios will yield lower than actual damage assessment.
A. Slower uncovery which yields the same total amount of hydrogen and causes greater oxidation along a shorter length of clad.
B. Rapid uncovery following a large LOCA causes fuel heat up to a higher temperature but oxidation is limited by the consumption of available steam.
4.12.3 Evaluation of core damage using atmosphere hydrogen concentrations must consider all sources of hydrogen addition to the containment as listed in paragraph 4.12.1.
A. Hydrogen which is normally present in solution in primary coolant is released to the containment atmosphere following a break in the reactor coolant system.
- 1. This amount is negligibly small compared to the total quantity produced in a degraded core accident and is near the limit of sensitivity of measurement for typical PASS containment measurement capability.
- 2. Hence, hydrogen from this source is not quantified or considered in the analysis of core damage from containment hydrogen concentration.
B. A variety of metals in the containment are oxidized by steam to produce hydrogen.
These are principally aluminum and zinc.
- 1. Based on the sample time, duration of steam atmosphere in containment and containment temperature, use Attachment 13 to estimate the total volume of hydrogen (Vm) produced by this mechanism.
- 2. This may be done by summing the production during successive periods at the corresponding containment temperatures.
Rev. No.
5 Page 32 of 72 EAP-44
CORE DAMAGE ESTIMATION EAP-44 C. Production of hydrogen by radiolytic decomposition of water becomes significant when fission products are dispersed in the coolant.
- 1. Hence, a prerequisite knowledge of the category of fuel pellet overheat (initial, intermediate, major) must be available in order to determine the hydrogen volume produced by this mechanism.
- 2. This information may be obtained from one of the other damage assessment methods described in this procedure or assumed.
- 3. Based on sample time and duration of fuel pellet overheat category, use Attachment 14 to estimate the total volume of hydrogen (Vr) produced by this mechanism.
- 4. An appropriate thermal power level (MWt) for the prior 30 day operating period is used to obtain the total radiolytic hydrogen production using Attachment 14.
The appropriate power level is determined as follows:
PL=
(P)
(24)(tp) where:
PL =
Power level for use with Attachment 14, (MWt)
P
=
Total gross thermal power output over last tp days from Attachment 7 (MWt-hr)
Tp =
Duration of power level history, (30 days or latest full power operating period, whichever is less) 24 =
Number of hours per day Rev. No.
S Page 33 of 72
CORE DAMAGE ESTIMATION 4.12.4 Production of hydrogen by oxidation of zirconium metal in the core is readily determined by subtracting the hydrogen not produced by clad oxidation from the total measured hydrogen.
The net hydrogen produced by core zirconium oxidation is calculated using the following equations:
Vgross
=
(H 2 ) (C) (Pl) (To) (3.531E-5) 100 T1 Po Vnet
=
Vgross Vm Vr where:
H2
=
Is the measured containment hydrogen concentration as a percent of total C
=
Drywell free volume, (4.37E9 cc)
P1
=
Pressure of drywell when sample was taken, (PSIa)
T,
=
Temperature of drywell when sample was taken, (R
=
F + 0460)
PO
=
Standard pressure, (14.7 PSIa)
To
=
Standard temperature, (460 R)
Vgross
=
Volume of H2 gas measured in the drywell at STP, (SCF)
Vm
=
Volume of H2 gas in the drywell at STP due to oxidation of containment metals, (SCF)
Vr
=
Volume of H2 gas in the drywell at STP due to radiolytic decomposition of water, (SCF)
Vnet
=
Volume of H2 gas in the drywell at STP due to core zirconium oxidation 3.531E-5
=
Conversion factor cc to cubic feet, (ft3/cc)
Rev. No.
5 Page 34 of 72 L-EAP-44
CORE DAMAGE ESTIMATION EAP-44 4.12.5 The percent oxidation of core zirconium volume is calculated using the following equation:
%OX (Vnet) (100)
(ZM) (Convi) where:
%OX ZM Convl
=
Percent oxidation of total core zirconium volume
=
Total core zirconium mass, (9.71E4 lb)
=
Conversion factor, 7.92 SCF of H2 per pound of zirconium reacted 4.12.6 The percent of rods with clad rupture is determined using Attachment 15.
A. This attachment shows the percentage of rods with clad rupture as a percent of the core clad volume oxidized.
B. Three curves are shown for three temperatures which typically span the rupture temperatures during boil off.
C. These temperatures correspond to the values of system pressure in new prepressurized fuel.
D. Older fuel with accumulated fission product gas will rupture at lower temperatures.
E. Hence, the damage estimate obtained by utilizing the hydrogen measurement is a lower limit of damage.
F. The ordinate of Attachment 15 is used to place the core damage into one of the categories of fuel failure described in Attachment 9.
Rev. No.
5 Page 35 of 72
CORE DAMAGE ESTIMATION EAP-44 4.12.7 The percent of rods with local oxidation embrittlement is determined using Attachment 16.
A. This attachment shows the percent of fuel rods with local embrittlement as a function of the percent of the core clad volume oxidized during boil off scenarios.
B. At a given percent core oxidation, there is a range of possible embrittlement.
C. The oxidation limit during boil off occurs at about the time of clad melt, which is the upper limit of the fuel pellet overheat category.
D. The ordinate of Attachment 16 is used to place the core damage into one of the categories of fuel overheating described in Attachment 9.
4.12.8 There are several areas of concern, which may affect the accuracy of core damage estimation using containment hydrogen concentration.
These issues must be considered when determining the validity of the core damage estimation.
The major areas of concern are found in Section 10.0 of 2 to this procedure.
4.13 Integration of Other Parameters Into The Estimate The useful parameters are described briefly here, and no exact procedure for working them into the core damage estimation is provided.
Instead, these parameters are meant to be used by the individual performing the core damage estimation as an aid in determining the best method of core damage estimation and as a check on the reasonableness of determined core damage estimate. These parameters by themselves cannot be used to perform a core damage estimate.
Rev. No.
5 Page 36 of 72 I
CORE DAMAGE ESTIMATION EAP-44 4.13.1 Reactor Vessel Water Level -
This parameter is used to determine the fraction of the core which is uncovered during a LOCA and the duration of the uncovery.
A. If the core has never been uncovered, it is unlikely that any significant damage will have occurred, although it is possible to have some clad failure without uncovery of the core.
B. The fraction of core uncovered must remain uncovered for at least 5 to 10 minutes before fuel overheat and worse damage occurs.
C. This parameter is most useful with core damage estimation using reactor coolant core exit temperature.
4.13.2 Reactor Vessel Pressure -
High reactor vessel pressure may indicate a core damage event has occurred.
A. This indication is ambiguous, because there are many non-degraded core events which could also produce a high reactor vessel pressure.
B. This parameter is most useful with core damage estimation using reactor core exit temperature.
4.13.3 Primary Containment Integrity - Any breach in the Drywell or Torus free volume will result in the loss of activity and hydrogen from the primary containment system.
A. This will effect core damage estimates using gaseous PASS samples, hydrogen concentration and HRCM readings.
B. Any breach in the Torus liquid volume will also result in the loss of activity from the primary containment.
This should not affect core damage estimates using liquid PASS samples since uniform mixing will have taken place within the torus water and any leakage would not change the isotopic concentrations.
Rev. No.
5 Page 37 of 72
CORE DAMAGE ESTIMATION EAP-44 4.13.4 Pressure Vessel Integrity -
If the pressure vessel has a liquid or steam leak which bypasses primary containment it will result in the loss of activity and hydrogen from the primary containment system.
A. This will affect core damage estimate using PASS samples, hydrogen concentrations and HRCM readings.
4.13.5 Torus Free Volume Mixing -
Use Drywell and Torus pressure and other reactor system parameters to determine if the Drywell has blown down to the Torus.
A. Review recent events and system parameters to determine if the pressure vessel has been pressure relieved to the Torus via the automatic depressurization system, safety relief valves or any other means.
B. This information will help determine if both Torus and Drywell gaseous PASS samples are needed.
C. This information will help determine if both Torus and Reactor Vessel liquid PASS samples are needed.
4.13.6 Torus Liquid Recirculated -
Determine if Torus water is being recirculated through the pressure vessel and if so how long has this been ongoing.
A. This information will help determine if both Torus and pressure vessel liquid PASS samples are obtained.
4.13.7 Drywell Spray -
Determine if the Drywell sprays have been utilized.
A. This information is needed for the evaluation of core damage using HRCM readings.
Rev. No.
5 Page 38 of 72
CORE DAMAGE ESTIMATION EAP-44 4.13.8 Reactor Building Indications A. Reactor Building area radiation monitors, sump levels and samples, and Standby Gas Treatment System effluent monitors can be used to determine if there is a Primary Containment breach into the Reactor Building.
4.14 Development of.a Final Estimate All of the data obtained using sections 4.9 through 4.13 of this procedure should be used in the development of a final estimate.
It is not possible to provide a single method of converting all this data into the best estimate of core damage because of the wide diversity of physical phenomena and interactions which can occur during a core damage event.
The following provides some guidance on combining the estimates obtained using different methods:
4.14.1 The best application of each method of core damage estimation is:
A. gaseous PASS sample -
fuel clad failure and initial fuel overheat failure (category 5)
B. liquid PASS sample -
fuel overheat failure and fuel melting C. containment atmosphere hydrogen concentration -
fuel clad failure and fuel overheat failure D. HRCM readings -
fuel clad failure and initial fuel overheat failure (category 5) 4.14.2 Review the core damage estimation performed.
A. Review the technical basis and the areas of concern described in the Attachment 12 to this procedure.
Determine if any conditions exist that effect the validity of the estimate.
B. Review the other parameters described in section 4.13 to determine if they support the validity of the estimation or indicate that it is invalid.
Do this for each estimation of core damage.
Rev. No.
5 Page 39 of 72
CORE DAMAGE ESTIMATION EAP-44 4.14.3 Compare different valid core damage estimations obtained by using one method repeatedly or several different methods.
A. All core damage estimation data should be chronologically consistent.
B. Core damage estimates performed close to the same time should indicate similar degrees of core damage.
C. If a core damage estimate does not meet these criteria, use the review described in 4.14.2 to choose the most valid estimate for a particular time or to eliminate an estimate that is chronologically inconsistent.
4.14.4 Use any or all of the core damage evaluation methods and information contained in this procedure to assign the final core damage estimate.
More than one category may be indicated (for example, in order for fuel melt to occur, some fuel overheat and cladding failure must have occurred).
Assign the highest category of core damage indicated as the final estimate of core damage.
Page 40 of 72 Rev. No.
5
CORE DAMAGE ESTIMATION EAP-44 5.0 ATTACHMENTS
- 1.
SUMMARY
OF CORE DAMAGE ESTIMATION TECHNIQUES.
- 2. DESCRIPTION OF CORE RELEASE COMPONENTS.
- 3. BEST-ESTIMATE FISSION PRODUCT RELEASE FRACTIONS (FROM REF. 2.2.2).-
- 4. ISOTOPE ATTRIBUTES.
- 5. CHARACTERISTIC ISOTOPES.
- 6. CORE INVENTORIES.
- 7. FUEL CYCLE DATA.
- 8. SAMPLES MOST REPRESENTATIVE OF CORE CONDITIONS DURING AN ACCIDENT FOR ESTIMATION OF CORE DAMAGE.
- 9. CATEGORIES OF CORE DAMAGE EVENTS (FROM REF. 2.2.2).
- 10. RATIOS OF ISOTOPES IN CORE INVENTORY AND FUEL GAP (FROM REF. 2.2.2).
- 11. FISSION PRODUCT CONCENTRATIONS-IN REACTOR WATER AND DRYWELL GAS SPACE DURING REACTOR SHUTDOWN UNDER NORMAL CONDITIONS (FROM REF. 2.2.2).
- 12. TECHNICAL BASIS AND REFERENCE GUIDE FOR CORE DAMAGE ESTIMATION.
- 14. SPECIFIC RADIOLYTIC HYDROGEN PRODUCTION VS TIME.
- 15. PERCENT OF RODS WITH RUPTURE CLAD VS CORE CLAD OXIDATION.
- 16. PERCENT OF THE FUEL RODS WITH OXIDATION EMBRITTLEMENT VS TOTAL CORE OXIDATION.
- 17. 100% NOBLE GAS AIRBORNE IN DRYWELL (FROM REF. 2.2.10).
- 18. 25% HALOGENS AIRBORNE IN DRYWELL (FROM REF. 2.2.10).
- 19.
50% HALOGENS DILUTED IN TORUS WATER (FROM REF.
2.2.10).
- 20. 1% REMAINING FISSION PRODUCTS DILUTED IN TORUS WATER (FROM REF. 2.2.10).
Page 41 of 72 Rev. No.
5
CORE DAMAGE ESTIMATION ATTACHMENT 1 Page 1 of 1
SUMMARY
OF CORE DAMAGE ESTIMATION TECHNIQUES
- 1. Primary System Sample Isotope Concentrations The information obtained from the radiological concentrations of the sampled media is applicable to all categories of core damage.
The type of core damage is known when both the mechanism and the source of the fission product release has been determined.
The following information is obtained from sample radiological concentrations.
- a. Mechanism determination by fission product identity.
- b. Source determination by isotope ratios.
- c. Extent determination by fission product inventory.
- 2. Drywell Dose Rate The information obtained from dose rate measurements as an indication of fission product release is most applicable within the cladding failure and fuel overheat categories of core damage because of the dependence upon transport of the fission products.
The dose rate is most representative of the highly volatile fission products released through the clad burst and grain diffusion mechanisms.
The measurement of dose rate is used to yield a prediction of the percent of the fission product release from the fuel.
It cannot be used to distinguish between the identity of fission products.
Therefore, the ability to distinguish between major cladding failure and initial fuel overheat is limited.
- 3. Containment Atmosphere Hydrogen Concentration The information obtained from hydrogen measurements as an indication of fuel cladding oxidation is most applicable within the fuel overheat category of core damage.
Within this category, the clad surface temperatures are sufficiently high to result in the production by oxidation of measurable quantities of hydrogen but are below that which results in fuel clad material melting.
The measurement of hydrogen gas is used to yield:
- a. a prediction of fuel clad embrittlement (overheat).
- b. a prediction of fuel clad rupture.
Page 42 of 72 EAP-44 Rev. No.
5
CORE DAMAGE ESTIMATION:
EAP-44 ATTACHMENT 2 Page 1 of 1
DESCRIPTION OF CORE RELEASE COMPONENTS
- 1. Gap Release Fission product release which occurs when the fuel claddings experience initial rupture.
It consists mostly of activity that was released to void spaces within the fuel rods during normal reactor operation and rapid depressurization of contained gases provide the driving force for escape.
- 2. Meltdown Release Fission product release which occurs from the fuel while it first heats to melting and becomes molten.
High gas flows in the core during this period sweep the activity out of the core region.
- 3. Vaporization Release Fission product release which occurs after large amounts of molten core material fall into the reactor cavity from the pressure vessel.
Turbulence caused by internal convection and melt sparing by gaseous decomposition products of concrete produce the driving forces of the escape.
- 4. Oxidation Release Fission product release which occurs just after and is a result of a steam explosion event.
Finely divided fuel material is scattered into an oxygen atmosphere and undergoes extensive oxidation which liberates specific fission products.
Rev. No.
5 Page 43 of 72
CORE DAMAGE ESTIMATION EAP-44 ATTACHMENT 3 BEST ESTIMATE FISSION PRODUCT RELEASE FRACTIONS Page 1 of 1 Release GAP RELEASE MELTDOWN RELEASE OXIDATION RELEASE VAPORIZATION RELEASE Upper Upper Lower Upper Lower Upper Lower Upper Lower Nominal Limit Limit Nominal Limit Limit Nominal Limit Limit Nominal Limit Limit Noble GasesO.030 0.030 0.010 0.12 0.873 0.485 0.970 0.087 0.078 0.097 0.010 0.010 0.010 (Xe,Kr)
Halogens 0.017 0.001 0.20 0.885 0.492 0.983 0.088 0.078 0.098 0.010 0.010 0.010 (I,Br)
Alkali Metals 0.050 0.004 0.30 0.760 0.380 0.855 0.190 0.190 0.190 (Cs,Rb)
Tellurium Group 0.0001 3x10-7 0.04 0.150 0.05 0.250 0.510 0.340 0.680 0.340 0.340 0.340 (Te,Se,Sb)
Noble Metals 0.030 0.01 0.10 0.873 0.776 0.970 0.005 0.001 0.024 (RU,Rh,Pd,Mo,Tc)
Alkaline Earths 1X10-6 3x10-9 0.0004 0.100 0.02 0.20 0.009 0.002 0.045 (Sr,Ba)
Rare Earths 0.003 0.001 0.01
-0.010 0.002 0.050 (Y,La,Ce,Nd,Pr, Eu,Pm,Sm,Np,Pu)
Refractories 0.003 0.0 0.01 (Zr,Nb)
Rev. NP 5
Page 44 of
CORE DAMAGE ESTIMATION ATTACHMENT 4 Page 1 of 2 ISOTOPE ATTRIBUTES GROUP ISOTOPE T1/2 (hrs)
Ah-1 Ancestor Amount Gamma T1/2 between A(hrs
)
Test Delectability 4 & 360 hrs Noble Kr-85 9.39E4 7.38E-6 X
(a)
Poor No Gases Kr-85m 4.48 1.55E-1 X
1.5 Good Yes Kr-87 1.27 5.45E-1 X
2.7 Good No Kr-88 2.84 2.44E-1 X
3.7 Good No Xe-133 1.26E2 5.51E-3 X
6.5 Poor Yes Halogens I-133 1.93E2 3.59E-3 X
2.9 Good Yes I-133 20.8 3.33E-2 X
6.5 Good Yes I-135 6.61 1.05E-1 X
5.9 Good Yes Alkali Cs-134 1.81E4 3.84E-5 X
(a)
Good No Metals Cs-137 2.64E5 2.62E-6 X
(a)
Good No Tellurium Sb-127 9.24E1 7.50E-3 X
0.25(b)l Good Yes Group Sb-129 4.4 1.58E-1 X
1.0 Good Yes Te-129m 8.06E2 8.59E-4 X
0.34(b)l Poor No Te-132 78.2 8.86E-3 X
4.4 Good Yes Te-134 6.97E-1 9.95E-1 X6.7 Good No Moble MO-99 66.02 1.05E-2 X
6.1 Good Yes Metals RU-103 9.44E2 7.34E-4 X
2.9 Good No RU-105 4.44 1.56E-1 X
0.9(b)
Good Yes Alkaline Sr-89 1.21E3 5.71E-4 X
4.8 Poor No Earths Sr-90 2.51E5 2.77E-6 X
(a)
None No Sr-91 9.5 7.29E-2 X
5.9 Good Yes Sr-92 2.71 2.56E-1 X
6.1 Good Yes Page 45 of 72 K.
EAP-44 Rev. No.
5
CORE DAMAGE ESTIMATION ATTACHMENT 4 Page 2 of 2 ISOTOPE ATTRIBUTES GROUP ISTOPE T1/(hrs
-1 Ancestor Aon Gamma T1/2 between GROUP l ISOTOPE l T1/2(hrs Tes Amount Detectability 4 & 360 hrs Rare Y-91 1.40E3 4.94E-4 x
5.9 poor No Earths Y-93 10.1 6.86E-2 x
6.5 moderate Yes La-141 3.94 1.76E-1 x
6.0 moderate No Ce-141 7.80E2 8.88E-4 x
6.0 good No Ce-143 33.0 2.10E-2 x
6.2 good Yes Ce-144 6.82E3 1.02E-4 x
6.1 moderate No Pr-142 19.13 3.62E-2 x
moderate Yes Pr-143 3.25E2 2.13E-3 x
6.2 none Yes Pr-145 5.98 1.16E-1 x
4.2 good Yes Nd-147 2.64E2 2.63E-3 x
2.6 good Yes Nd-149 1.73 4.01E-1 x
1.3 good No Pm-147 2.30E4 3.02E-5 x
2.6 none No Pm-148 1.29E2 5.38E-3 x
good Yes Pm-149 53.08 1.31E-2 x
1.3 moderate Yes Pm-151 28.4 2.44E-2 x
0.5(b) good Yes Sm-153 46.7 1.48E-2 x
0.15(b) moderate Yes Ev-154 7.71E4 8.99E-6 x
(a) good No Refrac-Zr-95 1.54E3 4.51E-4 x
6.4 good No tories Zr-97 16.9 4.10E-2 x
6.2 moderate Yes Long half-life causes low nuclides decay away.
activity levels and won't be detectable until short lived Low yield may make it difficult to detect in the presence of other nuclides.
Denotes passed ancestor test, as described in Attachment 12, Section 3.0.
5 Page 46 of (a)
(b)
(x)
Rev. N EAP-44
CORE DAMAGE ESTIMATION' EAP-44 ATTACHMENT 5 CHARACTERISTIC ISOTOPES Page 1 of 1
- Does not meet all criteria Page 47 of 72 Group Preferences Isotope Noble 1
Kr-85m Gases 2
Kr-88*
3 Xe-133*
4 Kr-85*
Halogens 1
I-131 2
I-133 3
I-135 Alkali 1
Cs-134*
Metals 2
Cs-137*
Te-132 Group 2
Sb-129 3
Sb-127 Noble 1
Mo-99 Metals 2
Ru-105 3
Ru-103*
Alkaline 1
Sr-91 Earths 2
Sr-92 Rare 1
Ce-143 Earths 2
ND-147 3
PM-148 4
Pm-151 Refactories 1
Zr-97 2
Zr-95*
Rev. No.
5
CORE DAMAGE ESTIMATION ATTACHMENT 6 CORE INVENTORIES (CI)
Page 1 of 2 Group Noble Gases Halogens Alkali Metals Tellurium Group Noble Metals Alkaline Earths Rare Earths Refractories Isotope Kr-85 Kr-85m Kr-88 Xe-133 I-131 1-133 1-135 Cs-134 Cs-137 Sb-127 Sb-129 Te-129m Te-132 Te-134 Mo-99 Ru-103 Ru-105 Sr-89 Sr-90 Sr-91 Sr-92 Y-91 Y-93 La-141 Ce-141 Ce-143 Ce-144 Pr-142 Pr-143 Pr-145 Nd-147 Nd-149 Pm-147 Pm-148 Pm-149 Pm-151 Sm-153 Zr-95 Zr-97
- 1. ODAY
- 9. 896E+02
- 2. 554E+07
- 7. 465E+07 6.115E+06 4.447E+06 7.588E+07 1.220E+08 7.647E-01
- 8. 110E+03
- 6. 838E+05 1.572E+07
- 3. 689E+04
- 1. 756E+07 1.420E+08
- 2. 826E+07 1.545E+06
- 2. 577E+07 1.319E+06
- 7. 615E+03
- 9. 957E+07 1.235E+08 7.421E+05
- 1. 076E+08
- 1. 194E+08
- 1. 964E+06
- 4. 794E+07 2.747E+05
- 7. 892E+00 1.304E+06
- 7. 695E+07
- 2. 911E+06 2.387E+07
- 1. 063E+03
- 9. 609E+O1
- 5. 735E+06
- 4. 230E+06
- 1. 170E+06 1.429E+06
- 7. 731E+07
- 3. ODAY
- 2. 968E+03
- 2. 624E+07 7.480E+07
- 2. 932E+07 1.319E+07 1.278E+08 1.326E+08
- 9. 289E+O0 2.439E+04
- 1. 846E+06
- 1. 617E+07 1.299E+05 4.331E+07 1.420E+08 6.731E+07
- 3. 642E+06
- 2. 684E+07
- 3. 950E+06 2.288E+04
- 1. 198E+08
- 1. 237E+08 3.442E+06 1.328E+08 1.235E+08
- 7. 005E+06
- 9. 524E+07 8.223E+05 2.218E+02 8.583E+06 8.223E+07 8.282E+06 2.389E+07
- 9. 245E+03 2. 089E+03 1.399E+07
- 7. 953E+06
- 2. 582E+06 4.271E+06 1.170E+08
- 7. ODAY
- 6. 924E+03 2.615E+07 7.454E+07 7.361E+07
- 2. 738E+07 1.409E+08 1.329E+08 1.345E+02
- 5. 696E+04
- 3. 238E+06 1.628E+07 3.060E+05
- 7. 117E+07 1.417E+08 1.051lE+08 8.233E+06 2.748E+07
- 8. 982E+06
- 5. 331E+04
- 1. 201E+08 1.234E+08
- 8. 830E+06 1.337E+08 1.235E+08 1.649E+07
- 1. 187E+08
- 1. 908E+06
- 1. 822E+03 2.756E+07 8.225E+07 1.718E+07 2. 394E+07
- 4. 638E+04
- 2. 138E+04 2.074E+07 9.499E+06
- 3. 654E+06
- 9. 767E+06 1.233E+08
- 14. ODAY 1.384E+04 2.597E+07
- 7. 394E+07 1.145E+08
- 4. 332E+07 1.419E+08
- 1. 332E+08
- 1. 086E+03
- 1. 139E+05 4.241E+06
- 1. 653E+07
- 5. 847E+05
- 8. 742E+07 1.411E+08 1.232E+08 1.563E+07 2.886E+07 1.708E+07
- 1. 062E+05
- 1. 192E+08 1.226E+08
- 1. 760E+07 1.334E+08 1.236E+08
- 3. 124E+07 1.220E+08 3.775E+06 8.274E+03 5.570E+07 8.212E+07
- 2. 822E+07 2.404E+07
- 1. 615E+05
- 1. 219E+05
- 2. 323E+07
- 9. 773E+06
- 4. 087E+06
- 1. 880E+07 1.235E+08
- 30. ODAY
- 2. 961E+04
- 2. 557E+07 7.257E+07 1.324E+08 5.783E+07 1.432E+08 1.336E+08 8.708E+03 2.441 E+05
- 4. 815E+06 1.708E+07
- 1. 105E+06
- 9. 267E+07 1.395E+08 1.272E+08
- 2. 991E+07 3.191E+07 3.260E+07
- 2. 249E+05
- 1. 173E+08 1.210E+08 3.481E+07 1.321E+08 1.236E+08
- 5. 760E+07
- 1. 219E+08
- 7. 890E+06
- 3. 986E+04
- 8. 993E+07 B. 180E+07
- 4. 079E+07
- 2. 424E+07
- 5. 523E+05
- 6. 180E+05 2.405E+07
- l. 011E+07 4.493E+06 3.691E+07 1.236E+08 Page 48 of 72 EAP-44 Rev. No.
CORE DAMAGE ESTIMATION EAP-44 I
ATTACHMENT 6 CORE INVENTORIES (CI)
Page 2 of 2 GROUP Noble Gases Halogens Alkali Metals Tellurium Group Noble Metals Alkaline Earths Rare Earths Refractories ISOTOPE Kr-85 Kr-85m Kr-88 Xe-133 I-131 1-133 1-135 Cs-134 Cs-137 Sb-127 Sb-129 Te-129m Te-132 Te-134 Mo-99 Ru-103 Ru-105 Sr-89 Sr-90 Sr-91 Sr-92 Y-91 Y-93 La-141 Ce-141 Ce-143 Ce-144 Pr-142 Pr-143 Pr-145 Nd-147 Nd-149 Pm-147 Pm-148 Pm-149 Pm-151 Sm-153 Zr-95 Zr-97 180.ODAY
- 1. 753E+05 2.308E+07 6.416E+07 1.474E+08
- 6. 916E+07 1.476E+08 1.372E+08
- 4. 534E+05 1.478E+06 6.465E+06
- 2. 062E+07
- 2. 813E+06
- 9. 949E+07 1.312E+08 1.317E+08 8.424E+07
- 5. 185E+07
- 7. 973E+07 1.240E+06
- 1. 055E+08
- 1. 109E+08
- 9. 541E+07 1.241E+08 1.244E+08 1.202E+08
- 1. 190E+08
- 3. 841E+07 6.480E+05 1.185E+08 8.022E+07
- 4. 859E+07 2.571E+07 4.546E+06 6.890E+06
- 3. 014E+07
- 1. 196E+07
- 8. 848E+06
- 1. 111E+08 1.247E+08
- 1. OYR 3.498E+05
- 2. 006E+07 5.482E+07 1.429E+08
- 6. 914E+07 1.429E+08 1.329E+08 1.692E+06 2.984E+06 7.123E+06
- 2. 182E+07
- 3. 154E+06 9.844E+07 1.198E+08 1.279E+08
- 9. 833E+07
- 6. 361E+07 7.466E+07 2.312E+06
- 9. 136E+07 9.772E+07 9.422E+07 1.118E+08
- 1. 183E+08
- 1. 182E+08 1.109E+08
- 6. 100E+07 1.446E+06 1.104E+08 7.505E+07 4.645E+07 2.559E+07 7.733E+06
- 1. 130E+07 3.282E+07 1.273E+07
- 1. 362E+07
- 1. 199E+08
- 1. 189E+08 1.5YR 5.1642+05 1.821E+07 4.914E+07 1.400E+08 6.891E+07 1.399E+08 1.303E+08 3.463E+06 4.450E+06 7.464E+06 2.238E+07 3.282E+06 9.759E+07 1.137E+08 1.258E+08 1.045E+08 7.071E+07 6.716E+07 3.247E+06
- 8. 277E+07
- 8. 966E+07 8.644E+07
- 1. 041E+08
- 1. 144E+08 1.149E+08
- 1. 105E+08 7.365E+07 2.216E+06 1.054E+08 7.192E+07 4.520E+07
- 2. 558E+07
- 9. 627E+06 1.395E+07 3.452E+07 1.323E+07 1.800E+07 1.166E+08 1.150E+08 3.OYR 9.851E+05 1.450E+07 3.783E+07
- 1. 350E+08 6.865E+07 1.347E+08 1.264E+08 1.101E+07 8.749E+06
- 8. 108E+06 2.337E+07 3.451E+06 9.669E+07 1.037E+08 1.227E+08 1.155E+08
- 8. 618E+07
- 4. 968E+07 5.543E+06 6.569E+07 7.361E+07 6.833E+07
- 8. 892E+07
- 1. 070E+08 1.068E+08
- 9. 639E+07 8.531E+07 4.732E+06
- 9. 593E+07 6.618E+07 4.318E+07 2.602E+07
- 1. 153E+07
- 1. 717E+07 3.726E+07 1.446E+07 3.058E+07
- 1. 048E+07
- 1. 079E+08 Page 49 of 72 Rev. No.
5
CORE DAMAGE ESTIMATION EAP-44 ATTACHMENT 7 Page 1 of 1 FUEL CYCLE DATA Previous Fuel Cycles Fuel Cycle Total Days Refueling Fuel Cycle Total Power Fuel Cycle Fuel Cycle in Fuel Outage EFPD Start (Day)
Stop (Day)
Cycle Duration Days Current - 2 P
t2 R
Current - 1 P
t R
Current Fuel Cycles Daily Power Output Cumulative Power Date Fuel Cycle Day Gross MWt-hr Output MWt-hr 10 2
3 4
5 7
8 9
10 11 12 13 14 EFPD
=
Effective Full Power Days Rev. No.
5 Page
_0 of 72
SAMPLES MOST REPRESENTATIVE OF CORE CONDITIONS DURING AN ACCIDENT FOR THE ESTIMATION OF CORE DAMAGE Pag SAMPLE LOCATION Supp.
Supp.
Jet Pool Pool Pump Liquid Atmos.
RHR Drywell Other Instructions Small Liquid Line Break, Reactor Yes Yes' Y_
Yes2 Power 1%
Small Liquid Line Break, Reactor Yes' Yes Yes2
- a. RHR must be in shutdown Power <1%
- b. Reactor water level must and flow from moisture Small Steam Line Break, Reactor Yes Yes' Y_
Yes2 Power 21%
Small Steam Line Break, Reactor Yes' Yes Yes2
- a. RHR must be in shutdown Power 1%
cooling mode
- b. Reactor water level must be raised and flow from moisture separators Large Liquid Line Break, Reactor Yes3 Yes4 Yes' Yes2
- a. Suppression pool must be Power >1%
suppression cooling mode Large Liquid Line Break, Reactor Yes4 Yes' Yes3 Yes2
- a. RHR must be in shutdown Power <1%
- b. Suppression pool must be suppression cooling mode
- c. Reactor water level must and flow from moisture Superscripts on the Sample Location indicate system sample order of preference e 1 of 1 EAP-44 C
ATTACHMENT 8 Rev. No.
5 CORE DAMAGE ESTIMATION Page 51 of 72
CORE DAMAGE ESTIMATION EAP-44 ATTACHMENT 9 Page 1 of 1 CATEGORIES OF CORE DAMAGE EVENTS Degree of Minor Intermediate Major Degradation
(<10%)
(10% -
50%)
(>50%)
No Fuel Damage Cladding Failure 2
3 4
Fuel Overheat 5
6 7
Fuel Melt 8
9 10 Page 52 of 72 Rev. No.
5
CORE DAMAGE ESTIMATION EAP-44 ATTACHMENT 10 Page 1 of 1 RATIOS OF ISOTOPES IN CORE INVENTORY AND FUEL GAP Activity Ratio* in Activity Ratio* in Isotope Half-Life Core Inventory Fuel Gap Kr-87 76.3 m 0.233 0.0234 Kr-88 2.84h 0.33 0.0495 Kr-85m 4.48h 0.122 0.023 Xe-133 5.25d 1.0*
1.0*
I-134 52.6 m 2.3 0.155 I-132 2.3 h 1.46 0.127 I-135 6.61h 1.97 0.364 I-133 20.8 h 2.09 0.685 1-131 8.04d 1.0*
1.0*
- Ratio = Noble gas isotope concentration for noble gases Xe-133 concentration
= Iodine isotope concentration for iodines I-131 concentration Page 53 of 72 Rev. No.
5
CORE DAMAGE ESTIMATION ATTACHMENT 11 Page 1 of 1 FISSION PRODUCT CONCENTRATIONS IN REACTOR WATER AND DRYWELL GAS SPACE DURING REACTOR SHUTDOWN UNDER NORMAL CONDITIONS Reactor Water, pCi/q Drywell Gas (pci/cc)
Isotope Upper Limit Nominal Upper Limit Nominal I-131 29 0.7 CS-137c 0.3a 0.03 b Xe-133 10-4a io-5b Kr-85 4xlO-5a 4x10-'b aObserved experimentally, in an operating BWR-3 with MK I containment, data obtained from GE unpublished document, DRF 268-DEV-0009.
bAssuming 10% of the upper limit values.
cRelease of CS-137 activity would strongly depend on the core inventory which is a function of fuel burnup.
Page 54 of 72
- -l EAP-44 Rev. No.
5
CORE DAMAGE ESTIMATION EAP-44 ATTACHMENT 12 PAGE 1 OF 10 TECHNICAL BASIS AND REFERENCE GUIDE FOR CORE DAMAGE ESTIMATION 1.0 Core Damage Mechanisms and Fission Product Sources 1.1 The mechanism of fission product release from a degraded fuel rod is identified through the presence of characteristic fission products in the sample medium.
1.1.1 Fission products are selected to differentiate between the three major mechanisms of release: clad burst which releases the highly volatile fission products found in the gas gap; grain diffusion which releases the moderately volatile fission products found in the fuel pellet; and melting which releases the non-volatile fission products.
1.1.2 Multiple criteria are applied to the selection of the specific isotopes used in the procedure:
A.
The samples may not be analyzed for as long as eight hours following an accident.
Therefore, the half-life should be sufficiently long and the core inventory sufficiently large to ensure that a measurable quantity is present in the sample media for the release mechanism in question.
B.
The rate at which the isotope reaches equilibrium in the core inventory should be fast with respect to power plant transients.
This rate is dependent upon a number of parameters; however, use of a short half-life is implied.
This criteria is imposed to minimize the effect of short plant power transients upon the equilibrium core inventory.
The best compromise between this criteria and criterion (a) includes those isotopes with half-lives between four hours and fifteen days, when such isotopes are available.
C.
The isotopes present in a sample must represent a specific mechanism of core damage.
This selection is dependent upon the available information on degraded core transport mechanisms.
D.
The isotope should be detectable within a postulated fission product mixture using standard semiconductor and multichannel analyzer techniques.
E.
Employ those isotopes for which the chemical behavior is most-well known.
Rev. No.
5 Page 55 of 72
CORE DAMAGE ESTIMATION ATTACHMENT 12 PAGE 2 OF 10 F.
Employ those isotopes for which primary system concentration is based on simple radiological decay with little or no ingrowth from parent and grandparent nuclides.
1.1.3 The categories of core damage identified as cladding failures are characterized by the release of fission products through the mechanisms of clad burst and gas gap diffusion.
The characteristic fission products are the noble gases and halogens.
These fission products are released in highly volatile chemical species.
1.1.4 The categories of core damage identified as fuel overheat are characterized by the release of fission products through grain boundary diffusion and diffusion within the U02 grains.
The characteristic fission products are cesium, rubidium and tellurium.
The fission products released in this category include the less volatile species driven off by the high temperatures in addition to increased quantities of the highly volatile species discussed with regard to cladding failure in 1.1.3 above.
1.1.5 The categories of core damage identified as fuel melt are characterized by the release of fission products through escape from molten fuel.
The characteristic fission products are barium, lanthanum and promethium.
The fission products released in this category include the nonvolatile species driven off by melting of U02 in addition to increased quantities of the less volatile species discussed with regard to fuel overheat in 1.1.4 above.
1.2 The identification of the source of fission product release is used to verify the release mechanism determined by isotope identity.
1.2.1 For a particular accident, the radial variation in peak fuel cladding temperature can be significant.
Therefore, accident scenarios can be postulated in which a limited number of fuel rods may experience fuel pellet overheating while the majority of the fuel may not reach the temperature required for a cladding burst.
During such an accident, the identity and quantity of individual fission products detected in reactor coolant samples is insufficient information to determine the type of damage, which has occurred.
1.2.2 The added information needed to evaluate the accident is the source of the detected fission products.
Specifically, it is necessary to determine whether the fission products have been released from the fuel rod gas gap or from the fuel pellet.
This determination can be performed using the relative ratio's of the gap and fuel pellet for the isotope of a given fission product.
Rev. No.
5 Page 56 of 72 EAP-44
CORE DAMAGE ESTIMATION EAP-44 ATTACHMENT 12 PAGE 3 OF 10 1.2.3 The relative ratio of the isotopes of a given fission product is, in effect, a constant at the time of production.
The value of the ratio is dependent upon the material being fissioned and the energy of the neutron which induces the fission.
Each isotope has its own characteristic half-life; therefore, the ratios of the isotopes will vary as a function of time following their production.
1.2.4 Those fission products which migrate along the fuel pellet temperature gradient and reach the gas gap will.
consist of material which has existed in the pellet for sufficient time for this migration to take place and therefore, may be considered to consist of the older collection of material.
1.2.5 The relative ratios of the isotopes of fission products found in the gas gap, is, therefore, different from that found in the fuel pellet. Thus, calculation may be employed to determine typical ratios for isotopes of fission products in a given region of the core.
Comparison of the ratios obtained from sample data with these calculated values determines the source of the fission product release.
1.3 Reference 5.6 describes four conditions or times during core melt at which major-driving forces for release exist.
The four major release components are Gap Release, Meltdown Release, Vaporization Release and Oxidation Release. A brief description of the physical processes occurring during each of these release components is shown in Attachment 2. The fraction of total fission product inventory released during each of these four major release components is shown in Attachment 3 for each of eight fission product elemental groupings.
1.3.1 This procedure is concerned with estimating core damage; hence, it will only address the Gap Release and Meltdown Release components, since the Vaporization Release and Oxidation Release components only occur after the core is 100% melted.
1.3.2 For the purpose of this procedure, the Meltdown Release will be divided into two categories of fuel damage; fuel pellet overheating and-fuel pellet melting.
The Meltdown Release fraction assigned to each category will be based on the element groups as follows: Noble Gases, Halogens, Alkali Metals and Tellurium Group are assigned to the fuel pellet overheat category; and Noble Metals, Alkaline Earths, Rare Earths and Refractories are assigned to the fuel pellet melt category.
Page 57 of 72 Rev. No.
5
CORE DAMAGE ESTIMATION ATTACHMENT 12 PAGE 4 OF 10 2.0 General Information for Core Damage Estimation 2.1 When estimating the mechanism, source and extent of core damage, the results are not necessarily clear cut and may exhibit some degree of overlap between various categories.
This is a result of the fact that core damage will not take place uniformly among all the fuel rods.
A combination of categories may exist simultaneously.
2.2 Calculation of isotopic inventories in the primary system at post scram times is not always straight forward.
In some cases, a particular isotope may have a strong neutron removal cross section while the reactor is operating.
Hence, its equilibrium concentration with its parents and grandparents under operating conditions will be much less than that which will occur after reactor scram.
For example, Xe-135 is quite prone to this effect since it has a strong neutron absorption cross section and is the daughter of Xe-135m and granddaughter of I-135.
2.3 Core inventories are readily calculated for all Iodine and Noble Gases of concern based on the current fuel cycle history, because of their relatively short half-lives. This is not the case for isotopes with long half-lives such as Cs-134 and Cs-137.
Consideration must be given to including the contribution of remnant activity from preceding fuel cycles for longer lived isotopes.
Since approximately 1/3 of the core is changed out in each refueling it is necessary to maintain information on core history for the current core load and the two preceding core loads.
3.0 Characteristic Nuclides All nuclides listed in each of the element groups of Attachment 3 were reviewed to determine which are best suited to represent that element group.
Several attributes of each isotope were evaluated during this review process.
The isotopes reviewed, their attributes and results of the evaluation are shown in Attachment 4. The attributes reviewed are discussed below.
3.1 A review was performed of each isotopes parent and grandparent radionuclides in a decay chain.
Only those isotopes whose activity is controlled by its own simple radioactive decay were included in Attachment 4. Isotopes with complicated in growth and decay were not chosen because of the difficulty in calculating the activity of such isotopes at times after reactor shutdown.
3.2 A review was performed of the relative amount of each isotope produced in the reactor core.
If the amount of a nuclide produced is significantly less than that of others, it will be difficult to detect and measure in a sample containing a mixture of fission products.
Rev. No.
5 Page 58 of 72 EAP-44
CORE DAMAGE ESTIMATION EAP-44 ATTACHMENT 12 PAGE 5 OF 10 3.3 Each isotope was reviewed to determine if it emitted gamma rays of a suitable energy and yield to be easily detected by Ge(Li) analysis.
3.4 The half-life of each isotope was reviewed to determine if it met the criteria established in 3.1 above, of between four hours and fifteen days.
4.0 Estimation Procedure Preliminary Guide for PASS Sample 4.1 The basis for core damage estimation is the quantity of characteristic fission products observed to be present in the sampled media and, therefore, is available for immediate release to the environment.
This quantity is expressed as a percent of the source inventory calculated to be present at the time of the accident.
The value of this inventory is dependent upon the source of the fission product release.
4.2 The quantity of released fission product is defined as that which is observed to be present in the sample media.
No inferences should be made regarding amounts of fission products which are not observed to be present, because of the limits on the present capabilities to predict fission product transport (see next paragraph for more detail).
4.3 This distinction is best explained by example.
Consider the case in which measured samples of the containment building atmosphere and reactor coolant indicate that 20% of the I-131 isotope calculated to be in the gas gap is now found in the sampled fluids.
This does not indicate that 20% of the fuel rods have been ruptured.
A greater number may be anticipated to have failed.
This number cannot be determined because the effects of oxidation within the core and plate out are not analytically known.
Therefore, it can only be stated that 20% of the gas gap source inventory is available for release to the environment.
Using the NRC core damage characteristics defined in Attachment 9, this would indicate Intermediate Cladding Failure.
5.0 Areas of Concern When Using Gaseous PASS Samples There are several areas of concern, which may affect the accuracy of core damage estimation using gaseous PASS samples.
These issues must be considered when determining the validity of the core damage estimation.
The major areas of concern are as follows:
5.1 Using only one sample to calculate the total activity in the free volumes of both^the Drywell and Torus introduces an inaccuracy of unknown magnitude.
If possible, it is always better to use samples from both volumes.
Page 59 of 72 Rev. No.
5
CORE DAMAGE ESTIMATION ATTACHMENT 12 PAGE 6 OF 10 5.2 If there is an atmospheric breach in the Drywell or Torus, or if there is a leak from the pressure vessel which bypasses primary containment, the estimation of total airborne activity will be incorrect and the validity of the core damage estimation will be in doubt.
If possible, an estimate of the airborne activity released through this pathway should be made and added to that calculated to be in the Drywell and Torus free volumes.
This will minimize the error due to this situation.
5.3 Very small amounts of the isotopes from the groups used for fuel pellet overheating and fuel pellet melting are likely to become airborne.
The majority of the activity will remain in solution and that which does become airborne will rapidly plate out and fall out.
Thus, it is unlikely that isotopes from these groups will be detectable in the presence of the large noble gas and halogen activities.
Even if they are detectable, they do not represent a good estimate of the total activity released, since the majority of activity will remain in solution.
Based on these limitations, it is not accurate to estimate core damage greater than category 5 using gaseous PASS samples.
Liquid PASS samples should be used to estimate core damage greater than category 5.
5.4 When estimating core damage using iodine and noble gas isotopes, it is preferable to use noble gases. A significant portion of iodine will remain in solution, a significant portion of the airborne iodine will plate out and fall out, and use of containment sprays will wash out an unknown fraction of airborne iodine.
Since airborne noble gases are little affected by these three mechanisms, they are a better estimate of total activity released.
6.0 Areas of Concern When Using Liquid Pass Samples There are several areas of concern, which may effect the accuracy of core damage estimation using liquid PASS samples.
these issues must be considered when determining the validity of the core damage estimation.
The major areas of concern are as follows:
6.1 Using only one sample to calculate the total activity in the liquid volumes of the Drywell and Torus introduces an inaccuracy of unknown magnitude.
If possible, it is always better to use samples from both volumes.
6.2 If there is a liquid leak form the pressure vessel, Drywell or Torus with an exit path outside of primary containment, the estimation of total liquid activity will be incorrect and the validity of the core damage estimation will be in doubt.
If possible, an estimate of the airborne activity released through this pathway should be made and added to that calculated to be in the pressure vessel and torus liquid volumes. This will minimize the error due to this situation.
Rev. No.
5 Page 60 of 72 EAP-44
CORE DAMAGE ESTIMATION EAP-44 ATTACHMENT 12 PAGE 7 OF 10 6.3 The fraction of iodine released from fuel pellets, due to clad failure and initial overheat, which remains in the liquid volume is unknown.
It is usually assumed that 50% of the release fraction will remain in the liquid volume.
Hence, gaseous PASS samples using noble gas isotopes are a more accurate means of estimating core damage due to clad failure and initial fuel overheat up through category 4.
6.4 The majority of the Alkali Metals, Tellurium Group, Nobel Metals, Rare Earths, and Refractory isotopes released from the core will remain in the liquid volume.
Hence, liquid PASS samples are best suited for calculating core damage greater than category 5.
7.0 Core Damage Estimation Using High Range Containment Monitor Readings 7.1 The same radiological parameters may be evaluated using measurement of the dose rate inside the containment building as were used for PASS sample concentration analysis.
The use of two different measurements for evaluation of the same physical parameters may be employed as a means to reduce uncertainly.
Or one may be used as a substitute for the other.
7.2 This method for core damage assessment is based upon the comparison between dose rates measured following an accident and analytically determined values of the realistic or best estimates of dose rates that would correspond to specific categories of core damage.
7.3 The radiation dose rates inside the containment building following an accident are dependent on many variables which include reactor power, fuel burnup, containment building geometry, the identity and quantity of the fission products released from the core, and the location within the building at which they are measured.
7.4 Cladding failure core damage is characterized by the release of fission products through the mechanism of clad burst and gas gap diffusion.
The following assumptions are used in evaluation of cladding failure:
- a. The characteristic fission products are the noble gases and
- halogens,
- b.
These fission products are released in highly volatile chemical species.
- c.
It is assumed that the source inventory is the equilibrium gas gap activity of Attachment 3.
- d.
It is assumed that 100% of the Noble gas and 25% of the halogen activity released from the gap remains airborne.
- e.
If containment sprays are operating, the user may enter a decontamination factor (DF).
- f. It is assumed that 50% of the Halogen activity released from the GAP is diluted in the Reactor Vessel Liquid Volume Page 61 of 72 Rev. No.
5
CORE DAMAGE ESTIMATION EAP-44 ATTACHMENT 12 PAGE 8 OF 10 7.5 Fuel pellet overheat core damage is characterized by the release of fission products through grain boundary diffusion and diffusion from within the U02 grains. The following assumptions are used in the evaluation of fuel pellet overheat:
- a. The characteristic nuclides are the Noble Gas, Halogen, Alkali Metals and Tellurium Group of Attachment 3.
- b.
It is assumed that the source inventory is the equilibrium Melt Release of Attachment 3 for the Element groups identified in (a) above.
- c.
It is assumed that 100% of the Noble gas and 50% of the Halogen activity released from the fuel becomes airborne and that 25% of the Halogen activity immediately plates out.
- d.
It is assumed that 1% of the Alkali Metals and Tellurium Group release becomes airborne and immediately plates out.
- e.
If Containment Sprays are operating, the user may enter a decontamination factor (DF).
- f.
It is assumed that 50% of the Halogen activity released and 99% of Alkali Metals and Tellurium activity released is diluted in the Primary Coolant.
7.6 The HRCM response is due largely to airborne noble gas and iodine isotopes.
7.6.1 Most of these isotopes have half-lifes on the order of several hours to several days.
The core inventory of these isotopes is affected by the recent core power history.
7.6.2 Because the HRCM response is due to a spectrum of isotopes all with different half-lifes, an isotope specific correction factor cannot be used.
Instead, a power history correction factor must be based on the predominant isotopes affecting the HRCM reading at a given time after core shutdown.
7.6.3 A review of the dose conversion factors and other isotope specific information indicates that a power correction factor for each of three periods after core shutdown adequately accounts for core power history and variations in isotopic distributions with time after reactor shutdown.
7.7 Calculation of core damage using HRCM reading requires use of the HRCM response curves from Reference 5.9.
7.7.1 Core clad failure is estimated using the isotopic release fractions and primary system activity distribution delineated in Paragraph 7.4 above.
Based on the time of reactor shutdown, obtain an HRCM response due to noble gases and iodines from Attachment 17 and Attachment 18 of Reference 5.9 respectively.
Rev. No.
5 Page 62 of 72 I
CORE DAMAGE ESTIMATION
_ ~~~~~;
A c~~~ - :
ATTACHMENT 12 PAGE 9 OF 10 7.7.2 Fuel pellet overheat core damage is estimated using the isotope release fractions and primary system activity distributions delineated in Paragraph 7.5 above.
Based on the time of reactor shutdown, obtain an HRCM response due to noble gas airborne in drywell, halogens airborne in drywell, halogens diluted in torus water, and remaining fission products diluted in torus water from Attachments 17, 18 and 19 of Reference 5.9 respectively.
8.0 Areas of Concern When Using HRCM Monitor Readings There are several areas of concern, which may affect the accuracy of core damage estimation using HRCM readings. These issues must be considered when determining the validity of the core damage estimation.
The major areas of concern are:
8.1 The attachments in Reference 5.9 are based on dilution of the airborne source term uniformly throughout the air space of the Drywell and Torus.
If the Drywell has not blown down to the Torus then-the core damage estimation must be modified to account for this.
8.2 If there is a breach in the Drywell such that airborne noble gases and iodines can escape, then the validity of this core damage estimation is doubtful.
8.3 If there is a steam leak directly from primary containment to some place other than the Drywell then the validity of the core damage estimation is doubtful.
8.4 Both HRCM's should have similar readings.
If one monitor reading is significantly different than the other, an attempt should be made to determine if either HRCM is malfunctioning.
If both HRCM's appear to be in proper working condition, then use the higher reading to perform the core damage estimation.
8.5 It is not possible to use HRCM readings to estimate core damage past the initial fuel overheat category because any additional airborne activity released after this category of fuel damage will have negligible impact on the HRCM reading.
9.0 Core Damage Estimation Using Containment Atmosphere Hydrogen Concentration 9.1 There are multiple sources of hydrogen released during severe accidents.
Each of these sources must be evaluated to yield a determination of that amount of hydrogen which has been generated by core material oxidation.
These sources include:
o Hydrogen present in the reactor coolant for normal chemistry control.
o Oxidation of various metals within the containment building.
O The radiolytic decomposition of water.
o Oxidation of the zirconium metal in the core.
Page 63 of 72 Rev. No.
5
CORE DAMAGE ESTIMATION EAP-44 ATTACHMENT 12 PAGE 10 OF 10 9.2 Oxidation of zirconium in the fuel clad occurs at all temperatures but is significant only above 18000 F. Such temperatures occur only if the coolant level falls below the top of the core.
The damage assessment procedure is based on the scenario of a boil-off of reactor coolant below the top of the core without inlet flow followed by a relatively rapid and complete recovery of reactor coolant level.
In actuality, other core damage scenarios may occur, as described below.
Use of this core damage procedure for either of these two alternate scenarios will yield a lower than actual damage assessment.
9.2.1 Slower uncovery which yields the same total amount of hydrogen and causes greater oxidation along a shorter length of clad.
9.2.2 Rapid uncovery following a large LOCA causes fuel heat up to a higher temperature but oxidation is limited by the consumption of available steam.
10.0 Areas of Concern When Using Containment Atmosphere Hydrogen Concentration There are several areas of concern, which may affect the accuracy of core damage estimation using containment hydrogen concentration.
These issues must be considered when determining the validity of the core damage estimation.
The major areas of concern are:
10.1 If the drywell has blown down to the Torus then the free olume and hydrogen measurements must be modified to account for the volume of hydrogen in the Torus.
10.2 If there is a breach in the primary containment or if hydrogen is removed by some other mechanism, then some of the hydrogen will escape and this method will underestimate core damage based on the primary containment hydrogen concentration.
10.3 Clad failure temperature depends on the age of the fuel; hence, the damage estimate obtained is a lower limit.
10.4 The core damage estimation was based on a particular boil off scenario.
Deviations from this scenario will affect the accuracy of the analysis.
10.5 The determination of radiolytic hydrogen production is dependent on the category of fuel overheat used to perform the analysis.
Use of the wrong category may cause a significant error in determining radiolytic hydrogen concentration, which would adversely affect the estimation of core damage.
10.6 The ability to detect clad rupture by measuring hydrogen depends on the sensitivity of the measurement.
Typically, the minimum measurable concentration in the containment atmosphere is 0.1% by volume.
This is equivalent to oxidation during boil off of 0.5% of the core clad volume. 6 shows that by the time a reliable concentration is measurable in the containment atmosphere, at least 40% to 100% of the rods are ruptured, depending on system pressure and fuel burnup.
Rev. No.
5 Page 64 of 72
CORE DAMAGE ESTIMATION EAP-44
-1
.1.
ATTACHMENT 13 Page 1 of 1 TYPICAL HYDROGEN PRODUCTION RATE FROM ALUMINUM AND ZINC VS. TEMPERATURE 8000 7600 7200 6800 6400 a: 6000 v
5600 U) 7 5200 C)
- I 4800
- 0. 4400 z
(D 4000 0
3600 o 3200 Ljl 2800 2400 2000 1600 1200 800 400 100 120 140 160 180 200 220 TEMPERATURE 'F 240 260 280 300 Page 65 of 72 Rev. No.
5
CORE DAMAGE ESTIMATION ATTACHMENT 14 Page 1 of 1 SPECIFIC RADIOLYTIC HYDROGEN PRODUCTION VS. TIME MAJOR FUEL OVERHEAT INTERMEDIATE FUEL OVERHEAT INMAL FUEL OVERHEAT 0
100 200 300 400 500 600 700 800 TIME. HOURS Page 66 of 72
'11 s
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Rev. No.
5
CORE DAMAGE ESTIMATION EAP-44 ATTACHMENT 15 Page 1 of 1 PERCENT OF FUEL RODS WITH RUPTURED CLAD VS. CORE CLAD OXIDATION 100 RUPTURE TEMPERATURE
< 1 00 PSIA
< 1200 PSIA
< 1650 PSIA FOR CURVE LABELED IS WITH TEMPERATURE 1200'F 1600'F 1800TF 0
0.5 1.0 1.5 2.0
% OXIDATION OF CORE CLAD VOLUME Page 67 of 72 600'F 80 60 40 20 0
0 CL 0:
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0 Rev. No.
5
CORE DAMAGE ESTIMATION ATTACHMENT 16 Page 1 of 1 PERCENT OF THE FUEL RODS WITH OXIDATION EMBRITTLEMENT VS TOTAL CORE OXIDATION FOR X TO X DECAY HEAT AND 300 PSIA TO 2600 PSIA WIlEN COOLANT LEIEL DROPS BY BOILOFF WITH NO INLET FLOV UNTIL CORE IS RAPIDLY QUENCHED f-E El:m z0 c
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0 100 80 C0 40 20 0
20 40 60 80 100
/e OXIDATION OF CORE CLAD VOLUME Page 68 of 72 EAP-44 Rev. No.
5
CORE DAMAGE ESTIMATION EAP-44 ATTACHMENT 17 Page 1 of 1 100% NOBLE GAS AIRBORNE IN DRYWELL I1 Sl,, II I
' 1 I
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HRCRM READINGS Page 69 of 72 1Oe 5
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Rev. No.
CORE DAMAGE ESTIMATION ATTACHMENT 18 25% HALOGENS AIRBORNE IN DRYWELL
-nME AFTER SHUTDOWN (HOURS)
FITZPATRICK -
HRCRM READINGS Page 1 of 1 Page 70 of 72 I
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I-I04 EAP-44 Rev. No.
5
CORE DAMAGE ESTIMATION EAP-44 ATTACHMENT 19 Page 1 of 1 50% HALOGENS DILUTED IN TORUS WATER Page 71 of 72 108 106 0:
I CY in('3 0
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(I 102 101 10o Rev. No.
5
CORE DAMAGE ESTIMATION ATTACHMENT 20 Page 1 of 1 1% REMAINING FISSION PRODUCTS DILUTED IN TORUS WATER lT-I u TIME AFTER SHUTDOWN (HOURS)
FITZPATRICK -
HRCRM READINGS Page 72 of 72 EAP-44 107 106 Of 0*
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.1, Rev. No.
5
ENTERGY NUCLEAR OPERATIONS, INC.
JAMES A. FITZPATRICK NUCLEAR POWER PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE EMERGENCY PLAN ASSIGNMENTS SAP-20 REVISION 22 APPROVED BY:
4 RESPONSIBLE PROCEDURE OWNER EFFECTIVE DATE:
FIRST ISSUE 0
<4 D,
xL REVIS 0
LL REVISION LIMITED REVISION INFORMATIONAL USE ADMINISTRATIVE PERIODIC REVIEW DUE DATE: M rch DATE:
!/2710E I*
TSR 207
EMERGENCY PLAN ASSIGNMENTS SAP-20 REVISION
SUMMARY
SHEET REV. NO.
22
- In Position 17 added number 5 that deals with EOF Security Coordinator.
Also on same legend, deleted boxes for Oswego County Liaison and NYS Liaison.
- On attachment 2 added reference to note 6 in the B&G Trades row and the RESP column.
- Attached section 4.4.
- Deleted SCBA quals for nuclear security guards.
21 Updated security's title changes from SECURITY COORDINATOR/SERGEANT to SHIFT SECURITY SUPERVISORS
- In position 17 deleted reference to EOF Security guard.
- For position 22 -
title change from Security Coord/Sergeant to Security Shift Supervisor
- In position 59 -
deleted reference to Security Guard at front desk at EOF.
- Deleted position 77 Security at the JNC.
- Updated 1 to reflect changes of security.
Page 2 of 55 I
Rev. No.
2 2
EMERGENCY PLAN ASSIGNMENTS TABLE OF CONTENTS SECTION 1.0 2.0 2.1 2.2 2.3 3.0 4.0 5.0 PAGE PURPOSE.............................................
4 REFERENCES...........................................
4 Performance References................................. 4 Developmental References............................... 4 Management Expectations................................ 4 INITIATING EVENTS......................................
4 PROCEDURE............................................
4 ATTACHMENTS............................................
5
- 1.
ORGANIZATION CHARTS AND LEGEND................... 41
- 2.
ERO TRAINING APPLICABILITIES..................... 48 Page 3
of 55 SAP-20 Rev. No.
2 2
EMERGENCY PLAN ASSIGNMENTS 1.0 PURPOSE This SAP provides job specific guidance for Emergency Plan assignments.
Positions that are defined in the normal plant organization chart are not defined within.
Each position includes an arbitrarily assigned reference number for that position. Adherence to specific instructions is very desirable as portions of this guidance may have been developed in response to drill comments or events, but procedural adherence is not required.
Individual sections may be copied and used by staff for reference, as needed.
2.0 REFERENCES
2.1 Performance References None 2.2 Developmental References 2.2.1 JAF Emergency Plan Section 5, ORGANIZATION 2.2.2 EAP-17, EMERGENCY ORGANIZATION STAFFING 2.3 Management Expectations 2.3.1 ACT-99-40398 (DER-99-00118) Ensure the Emergency Response Organization immediately reviews any procedural deviations or departures taken from approved plant procedures during emergencies.
3.0 INITIATING EVENTS None 4.0 PROCEDURE 4.1 Each individual called on to fill an emergency position in the Control Room, Technical Support Center, Operational Support Center or another facility should use as reference the appropriate enclosure for that emergency position found in this procedure. to this procedure includes the Emergency Organization charts for each facility.
The charts include the position title, the facility activation requirements and the designated alternate.
Page 4 of 55 SAP-20 Rev. No.
2 2
EMERGENCY PLAN ASSIGNMENTS SAP-20 4.2 All documentation generated through the implementation of this procedure should be forwarded to:
Emergency Planning Coordinator James A. FitzPatrick Nuclear Power Plant 4.3 includes a list of ERO Training Applicability.
4.4 The Medical Department (Occupational Health Nurse), The Training Department, and Radiation Protection Department are responsible to notify the Emergency Preparedness Manager and the individuals supervisor when an individual has a lapses or failure to maintain qualification to perform his/her ERO position.
5.0 ATTACHMENTS
- 1.
ORGANIZATION CHARTS AND LEGEND
- 2.
ERO TRAINING APPLICABILITY Page 5 of 55 Rev. No.
22
EMERGENCY PLAN ASSIGNMENTS EMERGENCY AUGMENTED FACILITY LEADS POSITION 1 EMERGENCY DIRECTOR/TSC MANAGER ALT.
"ED" Emergency Director -
TSC Responsibilities
Ensure that of EAP-14.1 (TSC Activation Checklist) is completed.
Ensure announcement is made when TSC becomes operational.
- 2. Use IAP-2 to classify emergency as either UNUSUAL EVENT, ALERT, SITE AREA EMERGENCY or GENERAL EMERGENCY.
- 3. After classifying the emergency, complete IAP-1 checklist to assure appropriate procedures are initiated.
- 4. Review and approve New York State/Oswego County Part I, II and III forms every half hour or upon significant event change (forms found in EAP-1.1.).
Descriptive information should not be of a highly technical nature.
- 5. Announce over Gai-Tronics an update on plant status at approximately half-hour intervals.
- 6. Approve protective action recommendations prior to approving Part I and II forms.
- 7. Assure NRC notification over ENS has been done by Control Room Communicator and continued by TSC Communicator.
- 8. Authorize on Attachment 1 of EAP-15 all emergency exposure limits.
- 9. Review all press releases from the Joint News Center.
- 10. Gather TSC coordinators into conference room to plan corrective actions and have TSC coordinators brief each other on status of activities.
- 11. Appoint TSC Manager as Acting Emergency Director when you are in transit to EOF, or at other times as necessary.
12.Approve Part I, II and III forms just prior to leaving the TSC for the EOF.
- 13. Assure status boards are updated.
- 14. Refer to Section 5.3.1 of the Emergency Plan for a listing of general responsibilities.
- 15. Ensure offsite agencies are notified prior to a site evacuation.
Page 6 of 55 SAP-20 Rev. No.
2 2
EMERGENCY PLAN ASSIGNMENTS SAP-20 EMERGENCY AUGMENTED FACILITY LEADS POSITION 1 EMERGENCY DIRECTOR/TSC MANAGER ALT.
"ED" (continued) 16.Notify NMPC if remote assembly area on Howard Road is to be used.
(Have RSC dispatch rad technicians and equipment to Howard Road for personnel and vehicle monitoring.)
- 17. Notify Environmental Lab of emergency classification, if during normal working hours, and have them initiate activation of the EOF.
18.Ensure Plant Computer Operator activates ERDS upon declaration of an Alert.
- 19. Include the status of repair team actions during periodic plant briefings.
21.Declare EOF operational upon arrival after discussion with TSC.
..EXP2.3.1 22.Review any deviations or departures from procedures during emergencies.
Initiate required notifications.
(reference AP-02.06, Section 7.0)
POSITION 2 EMERGENCY DIRECTOR AIDE "ED Aide"
- 1. Review EOF activities and ensure their compliance with emergency plan procedures.
- 2. Act as a contact point for offsite agencies.
POSITION 3 OPERATIONS COORDINATOR "OPS COORDINATOR"
- 1. Direct plant operational activities.
- 2. Advise the Emergency Director on matters concerning plant operations.
- 3. Direct Accident Management Team; act as decision maker regarding Severe Accident Management.
Rev. No.
22 Page 7 of 55
EMERGENCY PLAN ASSIGNMENTS POSITION 4 TECHNICAL SUPPORT CENTER MANAGER "TSC Manager"
- 1.
Activate TSC in accordance with EAP-14.1.
Complete Attachment 2 of EAP-14.1.
- 2.
Use IAP-2 to assist in classifying the emergency as UNUSUAL EVENT, ALERT, SITE AREA EMERGENCY or GENERAL EMERGENCY.
- 3. After classifying the emergency, complete IAP-1 checklist to assure appropriate procedures are initiated.
- 4. Assure Communications and Records Coordinator transmits Part I, II and III forms every 30 minutes at a minimum (until EOF is declared operational).
- 5. Assign Licensed SRO to staff the Control Room/OSC/TSC/EOF hot line.
- 6. Fulfill Emergency Director's responsibilities while he is in transit to the EOF, and at other times as necessary.
- 7. Assure the following coordinators fulfill their responsibilities:
- a.
Security Shift Security Suprvisors
- b.
Technical Coordinator
- c.
Communications and Records Coordinator
- d.
Emergency Maintenance Coordinator
- e.
Rad Support Coordinator
- f.
Emergency Director Aide
- 8. Refer to Section 5.3.2 of the Emergency Plan for a listing of general responsibilities.
- 9. Conduct formal conferences as required.
- 10. Emphasize TSC formality to all personnel.
POSITION 5 OPERATIONAL SUPPORT CENTER MANAGER "OSC Manager"
- 2. Determine requirements for facility operability based upon the guidance provided in EAP-14.5.
- 3. Assign communicator (preferably SRO) to staff the 4-way hot line.
Page 8 of 55 SAP-20 Rev. No.
2 2
EMERGENCY PLAN ASSIGNMENTS SAP-20 POSITION 5 OPERATIONAL SUPPORT CENTER MANAGER "OSC Manager" (continued)
- 4. Check communications equipment for operability (e.g. OSC page system, hot lines, telephones, plant page).
- 5. Conduct frequent OSC briefings using the guidance provided by of EAP-14.5, OSC Briefing Checklist.
- 6. Perform duties as specified in EAP-14.5, Section 4.3.3.
- 8. Ensure that team members have proper safety equipment (eg.
hard hats, flashlights, etc.).
- 9. Ensure workers radiological exposure limits and qualifications are identified as soon as the OSC is manned.
- 10. Repair teams should be briefed to call back to OSC when responding to PA announcements while working on emergency tasks in the plant (e.g. during a protected area evacuation).
11.Ensure all team members (including rad tech) are present at the briefing prior to dispatching them.
- 12. Repair teams should be reminded to leave work area as soon as work is completed.
- 13. Assure in-plant teams have been thoroughly briefed prior to being dispatched.
Formal briefings and debriefings need to be conducted.
- 14. Assure OSC work activity center maintains up-to-date status board for tracking the dispatching of in-plant teams.
Page 9 of 55 Rev. No.
22
EMERGENCY PLAN ASSIGNMENTS SAP-20 POSITION 6 EMERGENCY OPERATIONS FACILITY MANAGER "EOF Manager"
- 2. Assign communicators or other personnel to perform the following functions:
relay Part I data over RECS update status boards as needed telecopy Parts I, II, and III data as needed copy and distribute Parts I, II and III data within EOF
- 3. Ensure individuals and equipment are available for performing the following functions:
relaying of technical data from plant relaying required information to offsite agencies dose assessment activities logging EOF activities tracking emergency facilities long term staffing procurement of supplies, materials and services
- 4. Upon declaring the EOF operational, ensure Parts I, II and III forms are completed and disseminated as required.
Use the following for guidance in distributing forms:
Part I forms Part II forms Part III forms Prepared by:
EOF Manager Rad Support Coord.
Technical Liaison Approved by:
Emergency Director Emergency Director Emergency Director Distribution:
Emergency Director Emergency Director Emergency Director EOF Manager EOF Manager EOF Manager Status Boards Keeper Dose Assessment Status Boards Keeper Rad Support Coord.
Boards Keeper RECS Communicator Rad Support Coord.
Technical Liaison Telecopiers Telecopiers Telecopiers
- 5. Ensure conferences between EOF Manager and Emergency Director are conducted as needed.
- 6. If an emergency generator is required for EOF power, call Auburn Armature at 1-800-333-0519.
- 7. Avoid conducting facility briefings while RECS line is in use.
Page 10 of 55 Rev. No.
2 2
EMERGENCY PLAN ASSIGNMENTS SAP-20 POSITION 7 JAF SPOKESPERSON "JAF SPOKESPERSON/JNC DIRECTOR" The position of Entergy's Spokesperson will be filled by the Director, Public Information, or designee.
The Spokesperson will coordinate all outgoing information.
The responsibilities of the Spokesperson will include:
- 1. Conducting routine interviews.
- 2. Serving as the source of statements.
- 3. Presiding at formal news conferences.
- 4. Coordinating the technical briefer and senior management available to the news media for information.
- 5. Coordinating the activities with the JNC Administrative Manager.
- 6. Maintaining contact with the Headquarters Office and securing any needed approvals.
- 7. Coordinating information with public information spokesperson for local, state, and federal agencies.
TECHNICAL SUPPORT CENTER EMERGENCY AUGMENTED STAFF POSITION 15 COMMUNICATIONS AND RECORDS COORDINATOR
- 2. Obtain copies of all Control Room communication forms for historical purposes.
- 3. Complete Parts I, II and III forms located in EAP-1.1 with appropriate input from Rad Support Coordinator and Technical Coordinator.
Descriptive information should not be of a highly technical nature.
- 4. Transmit Parts I, II and III forms located in EAP-1.1 every 30 minutes or upon significant event changes to Oswego County, New York State, EOF and JNC via telecopiers located at switchboard.
- 5. Designate a RECS communicator to transmit information over RECS phone when Part I of EAP-1.1 is completed.
Rev. No.
22 Page 11 of 55
EMERGENCY PLAN ASSIGNMENTS POSITION 15 COMMUNICATIONS AND RECORDS COORDINATOR (continued)
- 6. Designate a communicator or clerk to make copies of Parts I, II and III forms located in EAP-1.1 to distribute to:
- 1. Rad Support Coordinator
- 2. Plant Engineers
- 3. Technical Coordinator
- 4. OSC Manager
- 5. Public Information Officer
- 6. Emergency Director
- 7. Emergency Maintenance Coordinator
- 8. NRC Communicator
- 9. Communications and Records Coordinator
- 10. Telecopier
use any communicator if SRO is unavailable).
- 8. Assign an individual to contact all agencies on Attachment 8 of EAP-1.1 not already notified by Control Room Communicator, if Emergency Director so desires.
- 9. Maintain log of events in record book.
- 10. Once EOF is operational, they will transmit Parts I, II and III forms located in EAP-1.1 to Oswego County, New York State and JNC.
- 11. Terminate sending all telecopies to Oswego County and New York State once the EOF is operational.
- 12. Request to receive Parts I, II and III forms located in EAP-1.1 via telecopy from EOF after the EOF is operational and distribute them.
- 13. Instruct the TSC RECS Communicator to copy information from of EAP-1.1 after EOF is activated and distribute it.
NOTE:
Distribute either the Part I copied by the RECS communicator or the one telecopied from the EOF.
It is not necessary to distribute both.
Page 12 of 55 SAP-20 Rev. No.
2 2
EMERGENCY PLAN ASSIGNMENTS SAP-20
-POSITION 16 EMERGENCY MAINTENANCE COORDINATOR "EMC"
- 3. Coordinate with Operations the dispatching of damage repair teams after informing the TSC Manager of the intent to dispatch a team.
- 4. Update TSC with findings of damage repair teams after they have returned to OSC work activity center and have been debriefed.
- 5. Maintain log of events in record book.-
- 6. Emphasize the importance of prioritizing tasks to be worked on by OSC repair teams.
As priorities change during the event, the priorities of individual tasks may also change.
These priorities must be communicated to respective personnel.
- 7. The Emergency Maintenance Coordinator should set the priorities and discuss them with the OSC Manager.
The status of repair teams should be forwarded to the ED/TSC Manager to be included in the plant briefings.
- 8. Supervisors should maintain logs of their activities.
- 9. Repair teams should be briefed to call back to OSC when responding to PA announcements while working on emergency tasks in the plant (e.g. during a protected area evacuation).
- 10. Repair teams should be reminded to leave work area as soon as work is completed.
Page 13 of 55 Rev. No. 2 2
EMERGENCY PLAN ASSIGNMENTS POSITION 17 EMERGENCY SECURITY COORDINATOR
- 2. Coordinate assistance from Oswego County Sheriff's Department if they are needed for site access control.
- 3. Coordinate assistance from offsite fire agencies if they are needed.
- 4. Coordinate personnel accountability activities in accordance with EAP-8.
- 5. Assure;EOF Secu rity Coordinator as.i-been dispatched to theEOF upon -activation.
- 6. Establish emergency access control points to the site in accordance with EAP-23.
- 7. Inform Main Security to establish and update an emergency classification sign.
- 8. Maintain log of events in record book.
Page 14 of 55 SAP-20 Rev. No.
2 2
EMERGENCY PLAN ASSIGNMENTS SAP-20 POSITION 18 TECHNICAL COORDINATOR "TSC Technical Coordinator"
(Group)
- 2. Coordinate use of SPDS information for monitoring plant status.
(Engineers)
(Engineers)
- 4. Provide completed Part III form of EAP-l.1 every 30 minutes or upon significant change to Communications and Records Coordinator.
NOTE:
The EOF will perform this function upon being declared operational.
(Engineers)
- 5. Provide technical support to the Control Room regarding
-appropriate corrective measures.
Use available TSC drawings.
(Engineers)
- 6. Assist in emergency classifications in accordance with IAP-2.
(Coordinator)
- 7. Coordinate engineering decisions with G.E. liaison.
(Engineers)
- 8. Maintain documentation on plant forms, etc. which clearly describe any work activities or modifications not found in plant procedures.
(Engineers)
- 9. Coordinate in-plant repair activities with Emergency Maintenance Coordinator and OSC Manager.
(Coordinator)
- 10. Maintain log of events in record book.
(Group)
- 11. Provide input to NRC communicator on operations data for NRC review.
(Engineers)
Rev. No.
22 Page 15-of 55
EMERGENCY PLAN ASSIGNMENTS POSITION 19 RAD SUPPORT COORDINATOR "TSC Rad Support Coordinator"
- 2. Assure Rad Engineers verify equipment is operational.
- 3. Verify that equipment listed in EAP-14.6, Habitability of the Emergency Facilities, Section 3.0 (Initiating Events) is operational so that indicators of abnormal radiological conditions can be monitored.
- 4. Obtain copies of completed Part I and II forms of EAP-1.1 from Control Room along with protective action recommendations (as appropriate).
- 5. Assure Out-of-Plant Dispatcher verifies cellular phone and radio equipment is operational.
- 6. Assure Rad Protection Supervisor establishes CAM, IM-1A and ARM by switchboard is operational and has technician available for habitability surveys in emergency facilities and assembly areas.
- 7. Establish Rad Protection and Chemistry Supervisors in OSC for coordination of in-plant teams.
- 8. Review and provide completed Part I and II forms of EAP-1.1 to Communications and Records Coordinator for Emergency Director approval.
- 9. Approve completed protective action recommendations from data obtained through the use of EAP-4 and EAP-42.
10.Designate Rad Engineers to monitor plant parameters and determine source term.
- 11. Assure Rad Engineers estimate fuel damage as described in EAP-44, if appropriate.
- 12. Assure Plant Chemistry Supervisor makes appropriate provisions for PASS sampling in accordance with PSP-17, AM-03.01, etc.
- 13. Perform dose assessment on EDAMS, if necessary.
(EOF may take over this function as soon as dose assessment personnel arrive at the EOF.)
Ensure that a complete turnover of dose assessment functions is completed prior to transfer of dose assessment to the EOF.
- 14. Coordinate in-plant entries using EAP-6.
- 15. Coordinate dispatching of field teams (2 offsite and 1 onsite) using EAP-5.3.
- 16. Maintain log of events in record book.
- 17. Ensure equipment is operational and meteorological/survey team information is posted.
Contact a plant computer operator if equipment is inoperable.
- 18. Update of meteorological information and plant status to field L
teams must be done every 15 minutes and/or upon changing.
Page 16 of 55 SAP-20 Rev. No.
EMERGENCY PLAN ASSIGNMENTS SAP-20 POSITION 19 RAD SUPPORT COORDINATOR "TSC Rad Support Coordinator" (continued) 19.Consider the need for increased habitability monitoring and area surveys throughout the plant during loss of power scenarios (i.e. area rad monitors and/or process rad monitors not available).
20.OSC repair team members radiological exposure limits and qualifications need to be identified upon OSC activation.
- 21. Upon OSC activation recommend to OSC Manager,and Rad Protection Supervisor that repair teams dress-out in PCs.
- 22. Assist with Accident Management Team.
23.Evaluate radiological conditions that could impact a Protected Area evacuation (EAP-10) and/or Site Evacuation (EAP-11).
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EMERGENCY PLAN ASSIGNMENTS POSITION 21 NRC COMMUNICATOR
(Group)
- 2. Obtain copies of all forms completed in the Control Room prior to TSC activation.
- 3. Maintain continuous communication with NRC Operations Center via the ENS line.
Provide information needed to update their status boards.
Ensure Attachment 6 of EAP-1.1 is completed and transmitted as required.
Also, as part of the Records and Communications group, assist in the following, if necessary.
- 4. Complete Part I, II and III forms of EAP-1.1 with appropriate input from Radiation Support Coordinator, Technical Coordinator and/or Security Coordinator until EOF is operational.
(Communications/Records Coordinator)
- 5. Transmit information on Part I form of EAP-1.1 every 30 minutes or upon significant event changes to Oswego County and New York State via RECS until EOF is operational.
(RECS Communicator)
- 6. Telecopy Part I, II and III forms of EAP-1.1 every 30 minutes or upon significant event changes to Oswego County and New York State until EOF is operational.
(Telecopy/Switchboard Op)
(Telecopy/Switchboard Op)
- 8. Complete and maintain notifications to all agencies on of EAP-1.1 if directed to do so by the E.D.
(Communicator)
- 9. Copy and distribute Part I, II and III forms of EAP-1.1 to:
(Communicator)
Emergency Director Radiation Support Coordinator Technical Coordinator Emergency Maintenance Coordinator Communications/Records Coordinator Public Information Officer Security Coordinator ENS Communicator Rev. No.
22 Page 18 of 55 -
EMERGENCY PLAN ASSIGNMENTS SAP-20 POSITION 21 NRC COMMUNICATOR (continued)
- 10. Maintain log of events in record book.
(Group)
(Telecopy/Switchboard Operator)
- 12. Record all briefings by Emergency Director and information discussed in Coordinator's conferences.
(Emerg. Log Keeper)
POSITION 22 SECURITY SHIFT SUPERVISOR
- 1. Ensure accountability is conducted in accordance with EAP-8.
POSITION 23 PLANT ENGINEER
" TSC Plant Engineers"
- 1.
Assist in TSC set-up in accordance with EAP-14.1.
- 2. Use computer terminals and EPIC to obtain computer information.
- 3. Monitor EPIC computer emergency logs.
- 4. Update vessel level and pressure status boards as information changes.
- 5. Complete Part III forms of EAP-1.1 on a half hour basis or upon significant changes.
Route to Communications and Records Coordinator through Technical-Coordinator.
(When EOF is operational, Part III forms will be filled out and telecopied by EOF personnel.)
- 6. Provide technical support as directed by Technical Coordinator.
- 7. Develop corrective actions to solve problems utilizing all available resources (drawings, technical manuals, etc).
- 8. Verify plant status information with Licensed SRO communicator on the Control Room hotline.
- 9. Coordinate repair efforts with OSC personnel as required.
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EMERGENCY PLAN ASSIGNMENTS POSITION 24 RAD ENGINEER "TSC Rad Engineers"
- 1. Position reports to the Rad Support Coordinator.
- 3. Obtain meteorological data in accordance with EAP-42 and/or posted operator aid.
Mete data should be posted every 15 minutes on status board and updated to field teams at the same time.
The radio dispatcher can perform these tasks.
- 4. Obtain plant process data via EPIC and/or Plant Parameter terminals.
Use EAP-4 for calculating release rates and projecting doses.
This should be done every 30 minutes as a minimum.
Release rates and projected doses should be used to complete Part II forms of EAP-1.1 on same frequency.
- 6. Verify mete data information that was used for Part I Of EAP-1.1 in the Control Room.
- 7. Obtain protective action recommendations via EAP-4.
- 8. Interface with Plant Chemistry Supervisor to obtain a more representative isotopic breakdown of source term.
- 9. Estimate fuel damage via EAP-44.
Radiological Assessment Group
(Group)
- 2. Obtain copies of Part I forms of EAP-1.1 from the Control Room to determine meteorological/dose assessment information already sent to offsite agencies.
(Rad Engineers)
- 3. Verify that meteorological data from the computer and strip charts is accurate and current.
(Rad Engineers)
- 4. Assure technicians, Rad Protection and Chemistry Supervisors are available for OSC staffing and functions.
(Rad Support Coordinator)
- 5. Assure cellular phone and radio equipment is operable for survey teams.
(Out-of-Plant Dispatcher)
(Radiation Support Coordinator)
- 7. Establish habitability surveys in accordance with EAP-14.6 as conditions warrant.
(Radiation Protection Supervisor)
- 8. Complete information for Part I, II and III forms of EAP-1.1 forms and provide to Communications/Records Coordinator for Emergency Director approval.
(Radiological Support Coord.)
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EMERGENCY PLAN ASSIGNMENTS SAP-20 POSITION 24 RAD ENGINEER "TSC Rad Engineers" (continued)
- 9. Complete protective action recommendations per EAP-4 for supporting documentation on Part I and II forms of EAP-1.1.
(Rad Engineers)
- 10. Monitor effluent pathways and determine source terms using forms in EAP-4. (Rad Engineers)
- 11. Estimate fuel damage.
(Rad-Engineers)
- 12. Coordinate PASS activities.
(Plant Chemistry Supervisor)
- 13. Monitor in-plant work activities and record radiological data in accordance with EAP-6.
(Radiation Protection Supervisor)
- 14. Coordinate dose projections with EOF prior to transferring activities.
(Rad Engineers)
- 15. Maintain log of events.
(Group)
POSITION 25 COMMUNICATOR "TSC Communicator"
(Group)
- 2. Obtain copies of all forms completed in the Control Room prior to TSC being declared operational.
- 3. Complete Parts I, II and III forms of EAP-1.1 with.appropriate input from Radiation Support Coordinator, Technical Coordinator and/or Security Coordinator until EOF is operational.
(Communications/Records Coordinator)
- 4. Transmit information on Part I forms of EAP-1.1 every 30 minutes or upon significant event changes to Oswego County and New York State via RECS until EOF is operational.
- 5. Telecopy Parts I, II and III forms of EAP-1.1 every 30 minutes or upon significant event changes to Oswego County and New York State until EOF is operational.
(Telecopy/Switchboard Op)
(Telecopy/Switchboard Op)
- 7. Complete and maintain notifications to all agencies on of EAP-1.1 if directed to do so by the E.D.
(Communicator)
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EMERGENCY PLAN ASSIGNMENTS POSITION 25 COMMUNICATOR "TSC Communicator" (continued)
- 8. Copy and distribute Parts I, II and III forms of EAP-1.1 to:
(Communicator)
Emergency Director Radiation Support Coordinator Technical Coordinator Emergency Maintenance Coordinator Communications/Records Coordinator Public Information Officer Security Coordinator NRC Communicator
- 9. Maintain continuous communication with NRC Operations Center via the ENS line.
Provide information needed to update their status boards.
(NRC Communicator)
- 10. Maintain log of events in record book.
(Group)
(Telecopy/Switchboard Operator)
- 12. Record all briefings by Emergency Director and information discussed in Coordinator's conferences.
(Emerg. Log Keeper)
- 13. Complete, as necessary, all call-outs of additional plant personnel needed for support in accordance with EAP-1.1.
- 14. Maintain copies of all forms.
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EMERGENCY PLAN ASSIGNMENTS SAP-20 POSITION 26 TELEPHONE/TELECOPY/ACCOUNTABILITY "TSC Telephone/Telecopy Operators"
- 1. Perform switchboard activities for screening incoming calls.
- 2. Set-up telecopiers according to the following:
- a.
Transmit only telecopier should be telephone number 342-4268.
This should be used to transmit telecopies to Oswego County and New York State (Part I, II and III forms) until the EOF is operational.
- b.
Receive only telecopier should be telephone number 349-6053.
This should be used to receive Part II forms from the EOF after it is operational.
- 3. Maintain a copy of all telecopies with the attached transmission reports.
- 4. Provide the EOF with copies of all Part I, II and III forms transmitted prior to the EOF being operational.
- 5. Take accountability when directed, and report to one of the dedicated card readers.
NOTE:
Telecopies of Part I, II and III forms to NYS and Oswego County should take priority over routine telecopies (eg.
accountability forms to Staffing Coordinator).
POSITION 28 REACTOR ENGINEERING
- 1. Assist in the estimation of fuel damage via EAP-44.
- 2. Perform Reactor Engineering duties as required.
- 3. Provide support for Severe Accident Management Team.
- 4. Confirm Reactor shutdown.
- 5. Identify RPV breach.
- 6. Determine status of Torus spray, Drywell Spray, and Boron Injection.
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EMERGENCY PLAN ASSIGNMENTS POSITION 29 EMERGENCY LOG KEEPER
- 1. Maintain an historical log of TSC activities which include as a minimum:
timeline of activities (e.g. time facility declared operational, E.D. directives, etc.)
summarize discussions between E.D. and other TSC staff summarize coordinator briefings summarize TSC Manager discussions and briefings POSITION 30 PLANT COMPUTER OPERATOR "TSC Computer Operator"
- 3. Monitor process run on computer systems to assure emergency priorities are established.
- 4. Maintain log of events in record book.
- 5. Activate ERDS at the Alert or higher classification.
- 6. Assess and maintain computer operability in the Control Room, TSC, OSC and other areas as requested.
POSITION 32 RADIO DISPATCHER
- 1. Ensure all equipment is operational.
- 2. Monitor and log locations of teams.
- 3. Interface with Rad Support Coordinator/Rad Engineers and keep them informed regarding survey results.
- 4. Periodically brief team regarding plant conditions and significant events.
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EMERGENCY PLAN ASSIGNMENTS SAP-20 POSITION 51 STAFFING COORDINATOR
- 1. Upon arriving at the EOF, the Staffing Coordinator should consult emergency implementing procedure EAP-43, Emergency Facilities Long Term Staffing.
- 2. As personnel arrive at the EOF, the Staffing Coordinator shall update the EOF Organization Status Board.
Inform the EOF Manager when all JAF positions have been filled.
-3. The Staffing Coordinator shall complete step-4.2 of EAP-43 by obtaining copies of forms in file-cabinet.
- 4. Once above forms have been completed, assure copies are distributed.
To do this, the Staffing Coordinator shall direct an individual to send copies of completed forms to the appropriate facilities.
NOTE:
Control Room and OSC forms should be sent to the TSC with instructions for forwarding to the Ops Coordinator and OSC Manager, respectively.
POSITION 52 EMERGENCY LOG KEEPER EOF
- 1. Maintain a historical log of EOF activities which include as a minimum:
timeline of activities (i.e., when facility is operational, when offsite (NRC, State) representatives arrive, when Emergency Director arrives, etc.)
summarize discussions between Emergency Director and offsite agencies (NRC, State, County, etc.)
summarize discussions between Emergency Director and other EOF staff (i.e., EOF Manager, Rad Support Coordinator, Emergency Director Aide, Technical Liaison and Public Information Officer).
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EMERGENCY PLAN ASSIGNMENTS POSITION 53 CLERK "EOF Clerk"
- 1. Upon arriving at the EOF, collect all Parts I, II and/or III of EAP-1.1 telecopies received from the TSC on telecopier "A".
Make a sufficient number of copies and distribute throughout the EOF.
Log time and form numbers in "Incoming Logbook."
Perform other duties as assigned.
- 2. Upon EOF becoming operational, the EOF Manager will provide completed copies of Parts I, II and III forms of EAP-1.1 on a minimum half hour basis.
These forms should be telecopied to the State and County via telecopiers "B".
- 3. In addition, copies of completed Parts I, II and III forms of EAP-1.1 should be telecopied to the JNC and TSC via telecopier "Cit.
- 4. An individual will be assigned responsibility for making additional copies of completed Parts I, II and III forms of EAP-l.1 and distribute them throughout the EOF.
- 5. Telecopier "A" shall also be used to receive press releases from the JNC.
The press releases should be forwarded to the public information liaison.
- 6. Upon completion of transmitting telecopies, a transmission report will be produced and should be attached to the form and filed for a log of outgoing telecopies.
POSITION 54 COMPUTER OPERATORS EOF
- 1. Troubleshoot all inoperable computer equipment as needed.
- 2. Contact TSC computer operator for any appraisal of systems status.
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EMERGENCY PLAN ASSIGNMENTS SAP-20 POSITION 55 COMMUNICATOR (EOF)
"Status Board Comunicators"
- 1. Obtain completed sequence of events status sheets from the Technical Liaison and post.
- 2. Update the following status boards whenever a new Part I and/or III form of EAP-1.1 is generated:
Vessel Level/Pressure Graph (from Part III)
Plant Parameter Trends (from Part III)
Effluent Monitor Trends (from Part III)
FitzPatrick Protective Action Recommendations (from Part I)
- 2. Review past Part I forms generated from Control Room and/or TSC.
- 3. Relay information from completed Part I forms as directed in procedure EAP-1.1.
NOTE:
It is necessary to ensure the first EOF Part I transmission occurs within 30 minutes of the last TSC Part I transmission.
POSITION 56 OSWEGO COUNTY LIAISON
- 1. Report to the EOF and request initial briefing regarding plant conditions.
- 2. Report to Oswego County EOC and:
- a.
assist Oswego County personnel in the interpretation of plant data that has been transmitted to the County.
- b.
assist Oswego County personnel in reconciling any apparent discrepancies in data.
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EMERGENCY PLAN ASSIGNMENTS POSITION 57 PURCHASING/ACCOUNTING
- 1. Access computer systems as necessary.
- 2. Obtain necessary authorizations and provide for procurement of supplies, materials and/or services as needed.
POSITION 58 TECHNICAL LIAISON
- 1. Upon arrival at the EOF, establish continuous communications over CR-TSC-OSC-EOF hotline.
- 2. To establish an historical sequence of events it will be necessary to log all significant plant events as obtained over the dedicated hotline on tear-off sheets for posting.
NOTE:
It is this method of logging sequence of events that ensures consistency of displayed information throughout the emergency facilities.
- 3. Access plant computer information on the EPIC terminal, or by logging on to the WYSE terminal 708 system.
Use procedure EAP-26, Plant Data Acquisition System Access, for reference.
- 4. Complete, when directed by the EOF Manager, a New York State Plant Parameter Part III form using EPIC, the dedicated hotline and 708 system for data input.
NOTE:
At a minimum, these forms shall be completed on a half hour basis and/or significant plant event.
These forms may be computer generated on EPIC or on the 708 system.
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EMERGENCY PLAN ASSIGNMENTS SAP-20 POSITION 59 EOF SECURITY COORDINATOR "Offsite Security Coordinator"
- 1. Check to ensure all outside entrances are locked with the exception of the main entrance on the west side and north exit door leading to the JNC.
EOF Manager has the master key to these outside entrances.
- 2. Activate Security Alarm Control Panel as required to ensure security of the facility.
- 3. Fulfills responsibilities outlined in Section 4.3 of EAP-37.
- 4. Assure radio at front desk is turned up to monitor JAF Security communications.
Make sure time on radio is correct.
Supplies and Equipment Available Item Location EOF Master Key Registration Packet Badging Supplies Sign Lettering Phone ext. 5715 EOF Manager Security Office Cabinet.
Security Office Cabinet Security Office Cabinet Office Desk
- 6. Ensure emergency classification sign and barrier stanchions are established at main entrance to instruct arriving personnel.
Page 29 of 55 NOTE:
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EMERGENCY PLAN ASSIGNMENTS POSITION 60 RAD SUPPORT COORDINATOR "EOF Rad Support Coordinator"
- 1. Ensure personnel, equipment and communications are available for performance of dose assessment activities.
- 2. If any problems are encountered with dose assessment computer equipment, contact the EOF computer operator or the TSC computer operator.
- 3. Provide EOF Manager with dose assessment information needed to complete Part I and Part II forms of EAP-l.1 on a minimum half hour basis and/or significant plant changes.
- 4. If there has been a release, ensure personnel arriving at EOF are monitored for contamination.
- 5. Ensure personnel departing the EOF and entering the 10 mile Emergency Planning Zone are assigned dosimetry.
- 6. Ensure field teams are briefed and are continually updated regarding plant information after dispatching.
- 7. Act as liaison with the Emergency Director for providing offsite agencies with an understanding of dose assessment calculations and protective action recommendations.
- 8. Ensure procedures are properly utilized and forms used.
- 9. Ensure set-up and operability checks are made to equipment and that a proper turnover is conducted.
- 10. Ensure status boards are updated.
- 11. Remind personnel to resolve discrepancies between measured and projected doses, if necessary.
- 12. Ensure that personnel obtain meteorological information/
forecasts using EAP-42.
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EMERGENCY PLAN ASSIGNMENTS SAP-20 POSITION 62 DOSE ASSESSMENT COORDINATOR "EOF Dose Assessment Coordinator"
- 1. Ensure all dose assessment equipment is operational upon arriving at EOF.
- 2. Establish communications with TSC Rad Engineers to discuss eventual transfer of dose assessment function.
- 3. Upon EOF being declared operational, the dose assessment function shall be transferred to the EOF.
- 5. Using EDAMS, provide data to Rad Support Coordinator for completion of Part I and II forms of EAP-1.l-on a minimum half hour basis and/or significant change.
- 6. Modify model input as actual data becomes available, such as:
effluent monitor readings effluent stream/PASS sample results from TSC field team results
- 7. Compare field data with model results and inform Rad Support Coordinator of differences.
- 8. Operate EDAMS as required.
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EMERGENCY PLAN ASSIGNMENTS POSITION 63 RAD DATA COORDINATOR "EOF Rad Data Coordinator"
- 1. Ensure all equipment is operational (i.e. radios, phones, 708 data terminal, etc.).
- 2. Review past and present locations and data of any teams dispatched from the plant.
- 3. Provide routing for dispatching EOF field teams and assume control for routing of field teams depending upon meteorological conditions.
- 4. Interface with radio operator to continually update field teams regarding plant information and meteorological conditions.
- 5. Collect field data.
- 6. Interface with Rad Engineer to obtain environmental samples.
- 7. Brief and dispatch EOF field teams in accordance with procedures, as needed.
POSITION 64 RAD ENGINEER EOF Radiological Engineer
- 1. Ensure EOF field teams are briefed and dispatched in accordance with procedures.
- 2. Review plant effluent monitor data on EPIC and/or 708 system and inform Rad Support Coordinator and Dose Assessment Coordinator of status.
- 4. Coordinate the location and type of environmental sampling that is needed.
- 5. Interface with Rad Data Coordinator to obtain environmental samples.
- 6. Compare field data with model results and inform Rad Support Coordinator of differences.
- 7. Operate EDAMS as required.
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EMERGENCY PLAN ASSIGNMENTS SAP-20 POSITION 65 RAD SUPPORT CLERK
- 1. Assist Dose Assessment Coordinator as directed.
- 2. Update status boards as directed.
POSITION 66 RADIO OPERATOR "EOF Dispatcher"
- 1. Ensure all equipment is operational.
- 2. Monitor and log locations of any team already dispatched from the TSC.
- 3. Interface with Rad Data Coordinator to continually update and move field teams.
- 4. Keep Rad Data Coordinator informed of locations and data collected by field teams.
- 5. Periodically brief teams regarding plant conditions and significant events.
POSITION 68 PUBLIC INFORMATION TECHNICAL ASSISTANT "Public Information Technical Assistant" 1 Promptly relay current information from the EOF/TSC to the JNC.
2 Respond to questions from the JNC on various aspects of the incident (such as plant status, accident management or dose assessment).
3 Ensure all offsite agency news releases are telecopied from the JNC for Emergency Director review and subsequent posting.
4 Securing review of news releases by the Emergency Director to assure technical accuracy.
5 Relay verified information to Joint News Center.
6 Post all JAF news release on EOF status board after issuance from JNC.
7 Provide interpretation of technical information to the public information officer.
8 Review plant data using computer.
Also complete the following:
- 1. Assist in facility set-up.
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22 Page 33 of 55
EMERGENCY PLAN ASSIGNMENTS SAP-20 POSITION 68 PUBLIC INFORMATION TECHNICAL ASSISTANT "Public Information Technical Assistant" "Continued"
- 2. Review plant status logs and information to update Joint News Center on emergency status.
- 3. Draft news releases or review JNC drafts prior to getting Emergency Director's approval.
- 4. Telecopy approved news releases to JNC.)
- 5. Maintain communication with JNC.
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EMERGENCY PLAN ASSIGNMENTS SAP-20 POSITION 70 TECHNICAL BRIEFER The position of Technical Briefer will be responsible for providing more detailed technical information to news media in order to supplement the functions of Entergy's spokesperson.
Together, the responsibilities of the Technical Briefer will include:
- 1. Assuring technical accuracy of information received at the JNC and used by the Entergy Nuclear Northeast spokesperson or other personnel.
- 2. Providing technically accurate information on the incident and plant operations to the news media.
- 3. Assist in the preparation of news releases to ensure technical accuracy.
POSITION 71 JNC DIRECTOR The position of JNC Director will be filled by the plant Manager of Communications or his designee when circumstances warrant establishment of the position.
Upon direction by the Director, Public Information, he will supervise and direct those operations of the JNC which are involved with the flow of information from the plant to the staff at the JNC.
The responsibilities of the JNC Director include:
- 1. Supervising the preparation of statements and news releases and distribution to the press and to public officials.
- 2. Maintaining communications between the JNC and other emergency facilities and assuring the appropriate flow of information.
- 3. Maintaining communications and coordinating the activities between the JNC and ENN-Office.
- 4. Coordinating information and briefings with federal, state and local emergency preparedness groups and others located at the JNC.
- 5. Supervising the activities of the JNC Administrative Manager who will be directly responsible for all administrative functions not involved in the immediate flow of information from the plant to the news media at the JNC.
- 6. Coordinating information with Technical Consultants at the JNC.
- 7. Coordinating the Inquiry Response and Rumor Control Programs with the respective team leaders.
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EMERGENCY PLAN ASSIGNMENTS POSITION 72 COMMUNI CATIONS /WRITERS The position of the writer will be filled by an Information Specialist or another individual designated by the JNC Spokesperson.
The primary responsibility of the writer is to draft news releases based on information received at the JNC.
POSITION 73 INQUIRY RESPONSE & RUMOR CONTROL
- 1. The Media Inquiry Response Team will include members of the Public Affairs staff designated from Entergy Nuclear Northeast in coordination with the state and county.
Functions will include responding to inquiries from the media, providing accurate responses or referring inquiries as required.
A team leader will be appointed by the JNC Director and state spokesperson to coordinate Media Inquiry Response activities.
Each team member will be supplied with the information and materials need to handle inquiries.
Team members will read the prepared statements and give the standard answers provided.
They will be authorized to give facts about Entergy Nuclear Northeast and plant which are in their data and fact sheet news releases and annual report if requested.
In addition, times and locations of press conferences and briefings, as well as names and telephone numbers of appropriate contacts in other agencies, may be supplied to the media.
The responsibilities of the Media Inquiry Response team will include:
- a. Logging all contacts including time of inquiry, identity, affiliation and telephone number of the caller, and nature of the inquiry and response.
- b. Providing standard response when appropriate.
- c. Referring inquiries requiring further elaboration or special response to the appropriate source.
- d. Returning phone calls as soon as feasible with consideration given to deadlines of individual media.
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EMERGENCY PLAN ASSIGNMENTS SAP-20 POSITION 73 INQUIRY RESPONSE & RUMOR CONTROL (continued)
- 2. The Rumor Control team will identify and correct inaccurate or misleading information.
This will be accomplished by monitoring news broadcasts on radio and television, reviewing newspapers, and through telephone lines which can be used to provide answers to questions or confirm information.
Off-air monitoring and Rumor Control telephone equipment is installed at the JNC.
- 3. The Rumor Control team will include individuals assigned by the JNC Director from Entergy's Public Affairs staff.
State and county representatives, as well as Nine Mile-Point staff, may also be assigned to the team.
A team leader will be appointed by the JNC Director or alternate in coordination with the New York State spokesperson.
- 4. The responsibilities of the Rumor Control team will include:
- a.
Monitoring radio and television broadcasts and newspapers to identify incorrect, inaccurate or misleading information.
- b.
Bringing such information to the attention of the JNC Director for correction.
- c.
Producing taped messages for use on the Rumor Control telephone lines.
- d.
Logging and responding to inquiries from emergency workers or the general public.
- e.
Bringing significant information learned from inquiries (such as recurrent misinformation or trends which are identified) to the attention of the JNC Director.
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EMERGENCY PLAN ASSIGNMENTS POSITION 74 ADMINISTRATIVE MANAGER The position of Administrative Manager will direct all activities and functions at the JNC not directly involved with the flow of information from the plant to the news media.
The responsibilities of the Administrative Manager or alternate will include:
- 1. Supervising administrative functions such as:
- a.
Registration (media, visitors, and participants).
- b.
Clerical services.
- c.
Security.
- d.
Setup and maintenance of JNC facilities.
- e.
Distribution/stocking of news releases.
- 2. Supervising videotape and photo services, including off-air monitoring.
- 3. Coordinating auxiliary services such as flights, lodging and food services.
POSITION 76 CLERICAL Clerical support personnel assigned to the JNC will perform the following functions as assigned by the Administrative Manager:
- 1. Typing/word processing for news release activities.
- 2. Photocopy/telecopy support for JNC staff.
- 3. Distribution of news releases/supporting materials.
- 4. Registration Registration personnel will perform the following functions as directed by the Administrative Manager.
- a.
Verifying proper identification of all staff, media, and visitors entering the JNC.
- b.
Registering all personnel entering the JNC.
- c.
Issuing proper color coded identification badges to all individuals.
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EMERGENCY PLAN ASSIGNMENTS SAP-20 POSITION 78 VIDEO/PHOTO SERVICES At the JNC, photographic and video services will be provided by Entergy photographers with assistance from Entergy Nuclear Northeast Public Affairs personnel and Constellation photographic and video services personnel.
Responsibilities will include:
- 1. Videotape recording or photography of all new briefings at the JNC to provide a permanent record.
- 2. Providing duplication and playback capability for videotapes of earlier briefings.
- 3. Assisting off-air monitoring of radio and television news broadcasts and bulletins concerning the emergency.
POSITION 81 RAD ENGINEER SUPPORT
- 1. Assist in facility activation.
- 2. Assist Rad Engineer as directed.
- 3. Assist Rad Engineer in performing the following:
- a. release rate calculations.
- b. dose calculations.
- c. PAR determination.
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EMERGENCY PLAN ASSIGNMENTS POSITION 82 NEW YORK STATE LIAISON
- 1. Report to the New York State Emergency Operations Center (NYS EOC) and contact the JAF EOF/TSC for conditions briefing.
- 2. Assist NYS personnel in the interpretation o plant data.
- 3. Assist NYS personnel in reconciling any apparent discrepancies in plant data.
POSITION 83 PARAMETER ASSESSMENT ADVISOR
- 1. Determine EPIC and instrument availability
- 2. Obtain and trend parameter data.
- 3. Forecast parameter data.
POSITION 84 SYSTEM ASSESSMENT ADVISOR
- 1. Conduct system assessments.
- 3. Assist with forecasting parameter trends.
Rev. No.
22 Page 40 of 55 SAP-20
Page 1 of 7 ORGANIZATION CHARTS AND LEGEND EMERGENCY AUGMENTED FACILITY LEADS EOF 1
DW, D24 Req.
E.D. Aide LEGEND IT Reporting Location
'A Drill Requirement*
43 Activation Requirement*
Designated Alt.*
Z IiA
- Explanation found in SAP-20 C.
EMERGENCY DIRECTOR (1)
TSCIEOF 2
DW, D24 3
Req.
TSC Manager EMERGENCY PLANT MANAGER TSC
- A DW, D24 4
Req.
TSC Manager JI EMERGENCY DIRECTOR AIDE (2)
TSCIEOF DW, D24 NR EOF Manager OPERATIONS COORDINATOR (3)
CRITSC D, D24 Req.
Shift Manager
-'+
.':'21-slS TSC MANAGER (4) 0 TSC DW, D24 g Req.
Emergency Plant Mgr. 2 OSC MANAGER (5)
I i
~OSC It DW, D24 Req.
Maint. Supervisor 2 I
I &C Supervisor "
EOF MANAGER (6)
ENTERGY SPOKESPERSON (7)
JNC 1
DW, D24 4 Req.
Recovery Mgr.
Emergency Organization Title SAP-20 ATTACHMENT 1 Rev. No.
22 EMERGENCY PLAN ASSIGNMENTS Page 41 of 55 it I
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WR DTW, D24 As Assigned 1
PARAMETER ASSESSMENT ADVISOR (83) (AM7)
DTW, 024 I WR As Assigned FIELD SUPPORT SUPERVISOR/STA (8) 1 As Ass gned REACTOR ENGINEERING (28) (*AMT) 9 D, 02 NRAsAsne LEGEND EMERGENCY DIRECTOR (1)
TSC/EOF OW, 024 Req.
TSC Manager I
EMERGENCY PLANT MANAGER 1.!4 TSC
]
DW, D24 f
Req.
TSC Manager OPERATIONS COORDINATOR (3) (AM7)
CRITSC D, D24 Req.
Shift Supervisor SHIFT MANAGER (9)
.n Req.
As Assigned t-I CONTROL ROOM SUPERVISOR (11)
SD I
Req.
As Assigned
.N
~~~~~~~,,.......
SENIOR NUCLEAR OPERATOR (12)
Req As Assigned 7j IJ X -. _
Lmergency Organization Title
(*AMT) Designates part of the Accident Management Team Drill Requirement 2, Activation Requirement Designated Alt.
- j NUCLEAR PLANT OPERATOR (13)
SD I
Req.
As Assigned H
E'IV t.Si,>41}gE,,;
SAP-20 ATTACHMENT 1 Rev. N <
22 EMERGENCY PLA SSIGNMENTS Page 42 55 4
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I I l H R TSC/EOF DW, D24 l
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Page 3 of 7 EMERGENCY PLANT MANAGER Na TSC DW. D24 r
Req.
I DW, D24 I.
Req.
Emergency Dir.
EMERGENCY MAINTENANCE COORDINATOR 116)
COMMUNICATORS (25) 0 DW, D24 OSC Manager I N MMUNICATIONS21 l
(
NR Communicator DW, D24
_d DW. 024 NR As Asslgned NR As Assigned n
EMERGENCY LOG KEEPER PLANT COMPUTER
- 29)
OPERATOR (301 l7 SECURITYGUARDS (31)
I" W,W12 Req.
As Assigned l 1
'z2;2s,j.
v%-L LEGEND RAD SUPPORT COORDINATOR
- 19) (-AMT)
DW, D24 Req.
Rad Engineer I.,-t;t,;H,,,i, s,s RADIO DISPATCHER (32)
- I~
~~~~~~~t HR DW. D24 NR As Assigned
\\
,S EMERGENCY DIRECTOR (1) 1;4 TSC/EOF DW, D24 Req.
TSC Manager EMERGENCY SECURiTY COORDINATOR 117)
DW, D24 7n NR Sec. Shift SupervisorI "I
tM KJ TECHNICAL COORDINATOR (I 8) l DW, D24 l
I Req.
Plant Engineer
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'Jt TELEPHONErrELECOPYI ACCOUNTABILtTY CLERK (26)
-SECURITY SHIFT SUPERViSOR (221 I
0~W,D24 I Req.
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r l
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TSC SUPPORT (96)
IZII HR A
- gDW, D24 lNR As Assigned l DW, NO NR As Assigned I
.,f 4
RAD ENGINEERS (24)
OW. D24 a
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Rad Support Coordinator I
t' -
DW. D24 As Assigned NR l
PLANT ENGINEERS (23) 1i
~~~~~~~~~~~~~~~~~~I... e1 DW, D24 WR AsAssigned l
RAD ENGINEER SUPPORT (81)
OW, D24 NR Rad Engineer Emergency Organization Title
(*AMT) Designates part of the Accident Management Team -
Drill Requirement Activation Requirement Designated Alt.
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SAP-20 EMERGENCY PLAN ASSIGNMENTS ATTACHMENT 1 Rev. No.
22 Page 43 of 55 I
I COMMUNICATIONS AND E
RECORDS COORDINATOR 11 )
DW, D24I Req.
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ORGANIZATION CHARTS AND LEGEND OSC Emergency Augmented Staff PAGE 3 OF 7 gii,l CHEMISTRY SUPERVISOR R
RP Supervisor g RAD PROTECTION SUPERVISOR 44)
OW, D24 RES Tech.
Chem. Supervisor
'4 MAINTENANCE SUPERVISOR 38)
DW. D24 i Req.
Mechanic/Electrician ISC Supervisor
,I_
s MAINTENANCE ENGINEER s
(35)
~
WR Plant Engineer
=1.441tl.t, MECHANICS 411 ELECTRICIANS (423 DW, ND IDW,N WR As Assigned WR Asssqe I
W 1;
~~~~~,S~
i; EMERGENCY MAINTENANCE COORDINATOR (16)
>i OSC MANAGER (5)
DW, D24 1 Req.
Maint. Supervisor i I&C SuPervisorJ I'X
<'tI.I.8i:-
I I&C SUPERVtSOR 36)
DW. D24 -i Req.
I&C Technician Q Maintenance Supervisor 1 CHEMISTRY TECHNICIANS RADIATION TECHNICIANS (49) DW (50)
W.2 Deq, D2 i;
R DW, D24 Req.
RES Tech eA Req.
Chemistry Tech
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1 t&C TECHNICIANS (40)
DWC ND S i
WR I&C Supervisor tl LEGEND PLANNERS (45)
L 14, i
Ops (34)'
B&G (37)
GA (39)
W/H (48)
Nurse (47)
Fire Prot (46)
Clerk Support FuncUon (0)
Tool Room Attendants DW N WR AsAssigned SO
'Req As Assigned Emergency Organization Title D rill Requirement d
Activation Requirement Designated Alt. S kac EZ'-~E
= =M0 SAP-20 ATTACHMENT 1 Rev. N 22 EMERGENCY PLAN--SSIGNMENTS Page 44 5S Izt^
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Page 5 of 7 EMERGENCY DIRECTOR AIDE (2).>
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CLERK 53)COMPUTER COMMUNICATORS OSWEGO COUNTY PURCHASING/
l TECHNICAL LIAISON NATOR E51 KEEPER 52G CLERK (53)
I OPERATORS (54)
( (55) 1 LIAISON (56)
ACCOUNTING (57) g 1 (58)
DW024 Ls Assigned WR As Assgned WR As Assigned 9 NR A Assigned l
Req.
As Assigned e
- J Req.
Plant Engineer NEW YORK STATE EOF SECURITY EOF RAD SUPPORT NRC NIAGARA MOHAWK HER LIASONS LWASON (82)
W COORDINATOR (59) l COORDINATOR (60) l REPRESENTATIVES LIAISON lj O
L WR D
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E COORDINATOR (63)
RAD ENGINEERS (4)
ENGINEER Req. TSC Rad Eng~
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RAD ENGINEER SUPPORT (81)
R Rd E WR A As Dose Assessment Coordinator DW.
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4 HFMAITRY Z.
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,DW, D24 o VR As Assigned WR As Assigned W R As Assigned
(_
LEGEND Emergency Organization Title Drill Requirement Activation Requirement Designated Alt.
'!¶ U-5-.:.>-3 -5.S II Fio 1.- I SAP-20 EMERGENCY PLAN ASSIGNMENTS ATTACHMENT 1 Rev. No.
22 Page 45 of 55 STAF COORDIN lWR A
ORGANIZATION CHARTS AND LEGEND Page 6 of 7 RECOVERY MANAGER VII JNC Emergency Augmented Staff NOTE: Staffing fr JNC I
personnel per WPO Public Relations I-I 1
.w........
- - - - - - - - - - - - - - - - - - - - - - - I Listed staff may report to TSC or EOF I
~~~~~~~~~~~~~~~~~~~~~~
INFORMATION INFORMATION I
'TECHNICAL ASST (68)
TECHNICAL ASST (68)
NR As Assigned NR As Assigned
- 3
[ 1 ] Indicates staff from WPO, other projects or agencies LEGEND I
Emergency Organization Title Ni Drill Requirement 1
Activation Requirement Designated Alt.
SAP-20 ATTACHMENT 1 Rev. N ffi_
EMERGENCY P qS2 Page 46 oN T
55 I.
-9 I
4t I-_--_--_--_--_--_-__
__ __ __ __ _J
ORGANIZATION CHARTS AND LEGEND Page 7 of 7 Drills and Walk-Thrus DW, D24
=
Drill and walk-thru before functioning in position, drill at least once every 24 months thereafter.
D, D24
=
Drill before functioning in position; drill at least every 24 months thereafter.
DW, ND
=
Drill and walk-thru before functioning in position; no periodic drill requirement.
W, W12 Walk-thru before functioning in position, walk-thru at least every 12 months thereafter.
=
Simulator drills for initial and requalification.
DTW, DT24 =
Drill or tabletop and walk-thru before functioning in position, drill or tabletop at least every 24 months.
Reporting Locations CR
=
=
=
=
TRNG =
SEC
=
Control Room Technical Support Center Operational Support Center Emergency Operations Facility Joint News Center Training Building at JAF Security Post at JAF Activation Requirements Req = Required for facility to be staffed WR = As needed by facility manager only NR = Not required but preferred for facility to be declared staffed SAP-20 EMERGENCY P ASSIGNENTS ATTACHMENT 1 Rev. No.
22 Page 47 of 55
ERO TRAINING APPLICABILITIES Page 1 of 7 POSITIONS ~
I ~
I RAD EVAC I TSC~4
RAO
..I DI I STA.
l O
E S 8 DR&
E AD lC KH W WKTH WlTH lWKTH' T lFrlRs lR r6)
C.-CONTROL ROOM CONTROL ROOM XX XX xx SUPERVISOR SCBA SENIOR NUCLEAR XX XX
=
XX OPERATOR SCBA NUCLEAR PLANT XX XX XX XX OPERATOR SB OPS COORD2
_XX
=
XX
_X REACTOR ENG XX XX SHIFT MGR.
XX XX XX XX SCBA PARAMETER XX XX XX ASSESSMENT ADVISOR SYSTEM XX XX XX ASSESSMENT ADVISOR SHIFT TECH XX XX XX XX ADVISOR SCBA
',EMERGENCY OPERATIONS FACILITY (EOF)
CHEM/RP TECH XX XX XX CLERK XX X
COMMUNICATOR XX XX xx COMPUTER XXXX OPERATOR
-I DOSE ASSESS XX XX XX XX COORD EMERGENCY LOG XX XX KEEPER EOF MANAGER XX XXl XX SAP-20 ATTACHMENT 2 Rev. No 2
EMERGENCY P >
ASSIGNMENTS Page 4
f 55
C.
ERO TRAINING APPLICABILITIES Page 2 of 7
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_XX XX COORDINATOR NY STATE XX XX XX LIAISON OSWEGO co XX XXXX LIAISON PURCHASING/AC XX XX COUNT RAD DATA XX XX XX.
COORDINATOR RAD ENGINEER XX XX XX XX RAD ENG.
XX XX XX SUPPORT RAD SUPPORT XX XX CLERK RAD SUPPORT XX XX-XX XX COORDINATOR RADIO OPER XX XX XX STAFFING XX XX COORDINATOR TECHNICAL XX XX LIAISON EOF PUBLIC XX XX XX INFO TECH ASST
- '5 A
SIT E
_C ACCTY SUPV -
XX TRNG BLDG SECURITY XX XX SHIFT SUPERVISOR NUC SEC GUARD XX XX XX SAP-20 EMERGENCY PLAN ASSIGNMENTS ATTACHMENT 2 Rev. No.
22 Page 49 of 49 c-
ERO TRAINING APPLICABILITIES Page 3 OF 7 COMMUNITIONS Dl XlSp SPOKESPERSRS:N/
JNC DIRECTOR~~~~~~~~~~~~~~~iD N
L~~
~
~ ~ ~~~~~~~~~~~~~~~~~~~~~~~
7 ADMIN MANAGER XX XX XX CLERICAL XX XX XX WRITER XX XX XX INQUIRY XX XX XX RESPONSE/
RUMOR CONTROL JAF XX X
TD XX X
SPOKESPERSON/____
RP BRIEFER XX X
XX TECHNICAL XX XX PUBLIC INFO XX XX XX XX TECHNICAL ASST_
SERVICES XX B&G - TRADES Xx lr E
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-lXX r
xx xx MECHANICS SU PI IIV I
II II I
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l X ELECTRICIAN XX XX XX XX FIRE PROT XX XX XX XX SUPERVISOR x
x SC.A SAP-20 Rev.
No/,2 EMERGENCY PL~. ASSIGNMENTS ATTACHMENT 2 Page 5
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ERO TRAINING APPLICABILITIES CP o
Page 4 of 7 E),
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I&C - TECH XX XX XX XX I&C SUPERVISOR XX XX XX I&C - TOOL XX XX ROOM ATTEN.
IN-PLANT XX xx XX DISPATCHER MAINTENANCE XX XX XX ENGINEER MAINT SUPV XX XX XX ELECT/MECH MAINT TOOL XX XX ROOM ATTEND.
MECHANIC XX XX XX XX NURSE XX XX OSC MANAGER XX XX XX PLANNER XX XX XX Q.C.
XX XX XX SUPERVISOR X
RADl PROT XX XX XX SUPERVISOR RAD PROT XX XX XX XX XX TECHNICIAN SCBA WAREHOUSE
=
XX PERSON WAREHOUSE XX XX SUPERVISOR OSC SUPPORT XX XX SAP-20 ATTACHMENT 2 Rev. No.
22 EMERGENCY PLAN ASSIGNMENTS Page 51 of 55
(
ERO TRAINING APPLICABILITIES Page 5 of 7 P
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NO
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XX xx XX XX AIDE EMERGENCY XX XX XX XX DIR/TSC MGR ALT EMERGENCY LOG XX xx KEEPER EMERGENCY.
XX XX XX MAINTENANCE COORD NRC XX XX XX COMMUNICATOR PLANT XX xx ENGINEER/
ELEC/MECH SAP -20 ATTACHMENT 2 Rev.
No1 22
~~~EMERGENCY P
- ASSIGNMENTS Pg f5
ERO TRAINING APPLICABILITIES Page 6 of 7 D
X oszuIoINS::; I E8SP DR l
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PLANT ENGINEER XX XX
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RAD SUPPORT XX XX XX XX XX COORD RADIO XX XX XX DISPATCHER EMERGENCY SEC.
XX XX XX COORD TECHNICAL XX XX XX COORDINATOR TELEPHONE/
xx XX TELECOPY/
ACCOUNT TSC XX XX X
MGR/ EMERGENCY XX DIR ALT.
TSC MANAGER XX XX XX-AIDE TSC SUPPORT XX XX SAP-20 ATTACHMENT 2 Rev. No.
22 EMERGENCY PLAN ASSIGNMENTS Page 53 of 55
ERO TRAINING APPLICABILITIES ABBREVIATIONS & ACRONYM TABLE Page 7 of 7 ESS PERS ESSENTIAL PERSONNEL TRAINING EMER DIR/COORD EMERGENCY DIRECTOR & COORDINATOR TRAINING SNO/STA SENIOR NUCLEAR OPERATOR (LICENSED OPERATOR) & SHIFT TECHNICAL ADVISOR TRAINING E-COMM EMERGENCY COMMUNICATIONS TRAINING NPO NUCLEAR PLANT OPERATOR (NON-LICENSED OPERATOR TRAINING)
RAD ASSES RADIOLOGICAL ASSESSMENT PERSONNEL TRAINING EDAMS EDAMS COMPUTER APPLICATION RC&S RADIOLOGICAL CONTROLS AND SURVEYS TRAINING EVAC & ACCT EMERGENCY ACCESS CONTROL, EVACUATION AND ACCOUNTABILITY TRAINING ER&CA EMERGENCY REPAIR & CORRECTIVE ACTIONS TRAINING JNC JOINT NEWS CENTER TRAINING TSC WLKTH TECHNICAL SUPPORT CENTER WALKTHROUGH OSC WLKTH OPERATIONAL SUPPORT CENTER WALKTHROUGH EOF WLKTH EMERGENCY OPERATIONS CENTER WALKTHROUGH JNC WLKTH JOINT NEWS CENTER WALKTHROUGH SIML WLKTH SIMULATOR WALKTHROUGH SAM SEVERE ACCIDENT MANAGEMENT RESP Respiratory Protection Qualified SCBA Self Contained Breathing Apparatus Qualified SAP-20 E
ASSIGNMENTS ATTACHMENT 2 Rev. No(
22 E
Page 5
f 55 1~~~~~~~~~~~~~~~~~i~~~~~U--
NOTE (11( SAM training may include tabletop drills as training at a frequency to be determined by the EPC.
(2):
Operations Coordinators may attend either Emergency Director/Coordinator or SNO/STA training.
(3)
Successful completion of Emergency Director/Coordinator training satisfies SNO/STA training.
(4)
Additional walkthrough of CONTROL ROOM for SAM Team Members.
(5):
For NRC licensed personnel filling dual roles as Operations Coordinators and ED/TSC Manager Alternates, attendance at SNO/STA training satisfies the annual training requirement, with additional training being obtained through participation in drills/exercises in the ED role.
(6):
Respiratory qualification may be excluded for those individuals that are determined to be medically incapable of functioning in a respirator, or are unable to achieve a satisfactory seal due to extraordinary facial features (excluding beard).
FI-.
.Vk hK Rev. No.
22 EMERGENCY PLAN ASSIGNMENTS ATTACHMENT 2
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