L-PI-03-014, Emergency Plan Implementing Procedures
| ML030570011 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 02/18/2003 |
| From: | Solymossy J Nuclear Management Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-PI-03-014 | |
| Download: ML030570011 (64) | |
Text
N MC__
Committed to NuclearEXCe/ e Prairie Island Nuclear Generating Plant Nuclear Management Company, LLC 1717 Wakonade Dr. East 9 Welch MN 55089 L-PI-03-014 10 CFR 50.4 February 18, 2003 US Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT DOCKET NOS. 50-282 AND 50-306 LICENSE NOS. DPR-42 AND DPR-60 PROCEDURES PRAIRIE ISLAND EMERGENCY PLAN IMPLEMENTING to the Prairie Island Nuclear Furnished with this letter are the recent changes Procedures. This submittal Generating Plant Emergency Plan Implementing includes the following documents:
INDEXES:
of Contents Emergency Plan Implementing Procedures Table REVISIONS: Activation and Operation of Operational Rev. 16 F3-7 Support Center (OSC)
Rev. 10 F3-17 Core Damage Assessment Rev. 8 F3-19 Personnel and Equipment Monitoring and Decontamination TCN # 2003-0080 TEMPORARY CHANGE ADDITIONS:
F3-23.1 Emergency Hotcell Procedure DELETIONS:
None
L-PI-03-0014 Nuclear rmanagemmu, .,ui* ,pjj .y, Page 2 INSTRUCTIONS:
Please post changes in your copies of the Prairie Island Nuclear Generating Plant Emergency Plan Implementing Procedures (F3). Procedures, which have been superseded or deleted, should be destroyed. Please sign and return the acknowledgment of this update to Bruce Loesch, Prairie Island Nuclear Generating Plant, 1717 Wakonade Drive East, Welch, MN 55089.
As per 10 CFR 50.4, two copies have also been provided to the Regional III Office and one to the NRC Resident Inspector. The Nuclear Management Company has in not made new or revised existing Nuclear Regulatory Commission commitments at this letter or the attachments. If you have any questions, please contact Mel Agen 651-388-1121 Extension 4240.
iý cPr de rie unea os M.S&mos 0 -r dul lear Generating Plant Isice Presiden airie CC Steve Orth, USNRC, Region III (2 copies)
NRC Resident Inspector-Prairie Island Nuclear Generating Plant (w/o attachment)
Mfst Num: 2003 - 0074 Date : 01/30/03 FROM : Bruce Loesch/Mary Gadient Loc : Prairie Island TO : UNDERWOOD, BETTY J Copy Num: 515 Holder : US NRC DOC CONTROL DESK SUBJECT : Revisions to CONTROLLED DOCUMENTS Procedure # Rev Title Revisions:
F3-17 10 CORE DAMAGE ASSESSMENT F3-7 16 ACTIVATION & OPERATION OF OPERATIONAL SUP CENTER (OSC)
F3-19 8 PERSONNEL & EQUIPMENT MONITORING & DECONT Temporary Change Additions:
2003 0080 F3-23.1 EMERGENCY HOTCELL PROCEDURE UPDATING INSTRUCTIONS Island Controlled Manual or File. Remove Place this material in your Prairie recycle it. Sign and date this letter revised or cancelled material and and return to Bruce in the space provided below within ten working days Island Nuclear Plant, 1717 Wakonade Drive E.,
Loesch or Mary Gadient, Prairie Welch, MN 55089.
4478) if you have any Contact Bruce Loesch (ext 4664) or Mary Gadient (ext questions.
Received the material stated above and complied with the updating instructions Date
PRAIRIE ISLAND NUCLEAR
Title:
Emergency Plan Implementing Procedures TOC GENERATING PLANT II Effective Date : 01/30/03 I
jApproved By: e PS Supt Rev Document # Title 19 F3-1 ONSITE EMERGENCY ORGANIZATION 31 F3-2 CLASSIFICATIONS OF EMERGENCIES 18 RESPONSIBILITIES DURING A NOTIFICATION OF UNUSUAL F3-3 EVENT SITE AREA, 28 F3-4 RESPONSIBILITIES DURING AN ALERT, OR GENERAL EMERGENCY 21 F3-5 EMERGENCY NOTIFICATIONS 8
SWITCHBOARD OPERATOR DUTIES F3-5.1 9
F3-5.2 RESPONSE TO FALSE SIREN ACTIVATION BLOCKAGE 8 K.> F3-5.3 RESPONSE TO RAILROAD GRADE CROSSING SUPPORT CENTER 16 F3-6 ACTIVATION & OPERATION OF TECHNICAL SUPPORT 16 F3-7 ACTIVATION & OPERATION OF OPERATIONAL CENTER (OSC) 20 RECOMMENDATIONS FOR OFFSITE PROTECTIVE ACTIONS F3-8 ACTIONS FOR 13 F3-8.1 RECOMMENDATIONS FOR OFFSITE PROTECTIVE
/SHIFT MANAGER THE ON SHIFT EMERGENCY DIRECTOR 18 F3-9 EMERGENCY EVACUATION 19 F3-10 PERSONNEL ACCOUNTABILITY 8
F3-11 SEARCH & RESCUE 14 F3-12 EMERGENCY EXPOSURE CONTROL 15 F3-13 OFFSITE DOSE CALCULATION 11 F3-13.3 MANUAL DOSE CALCULATIONS 7
F3-13.4 MIDAS METEOROLOGICAL DATA DISPLAY 5
F3-13.5 ALTERNATE METEOROLOGICAL DATA Page 1 of 3
Title : Emergency Plan Implementing Procedures TOC PRAIRIE ISLAND NUCLEAR Effective Date : 01/30/03 GENERATING PLANT Rev Title Document #
11 WEATHER FORECASTING INFORMATION F3-13.6 F3-14.1 ONSITE RADIOLOGICAL MONITORING 9
F3-14.2 OPERATIONS EMERGENCY SURVEYS 22 SURVEY TEAMS F3-15 RESPONSIBILITIES OF THE RADIATION DURING A RADIOACTIVE AIRBORNE RELEASE 17 TEAMS F3-16 RESPONSIBILITIES OF THE RADIATION SURVEY DURING A RADIOACTIVE LIQUID RELEASE 10 F3-17 CORE DAMAGE ASSESSMENT 10 THYROID IODINE BLOCKING AGENT (POTASSIUM IODIDE)
F3-18 8
PERSONNEL & EQUIPMENT MONITORING & DECONTAMINATION F3-19 17 F3-20 DETERMINATION OF RADIOACTIVE RELEASE CONCENTRATIONS 9
F3-20 .1 DETERMINATION OF STEAM LINE DOSE RATES 9
STACK F3-20.2 DETERMINATION OF SHIELD BUILDING VENT DOSE RATES 10 CONTROL POINT F3-21 ESTABLISHMENT OF A SECONDARY ACCESS 16 GROUP RESPONSE F3-22 PRAIRIE ISLAND RADIATION PROTECTION TO A MONTICELLO EMERGENCY 18 F3-23 EMERGENCY SAMPLING 12 F3-23.1 EMERGENCY HOTCELL PROCEDURE 7
F3-24 RECORD KEEPING DURING AN EMERGENCY 8
REENTRY F3-25 7
F3-26.1 OPERATION OF THE ERCS DISPLAY 7
F3-26.2 RADIATION MONITOR DATA ON ERCS 1
ERDS - NRC DATA LINK F3-26.3 18 F3-29 EMERGENCY SECURITY PROCEDURES 6
F3-30 TRANSITION TO RECOVERY Page 2 of 3
Title : Emergency Plan Implementing Procedures TOC PRAIRIE ISLAND NUCLEAR GENERATING PLANT Effective Date : 01/30/03 Rev Document # Title 6
F3-31 RESPONSE TO SECURITY RELATED THREATS 2
F3-32 REVIEW OF EMERGENCY PREPAREDNESS DURING OR AFTER NATURAL DISASTER EVENTS Page 3 of 3
EMERGENCY PLAN IMPLEMENTING PROCEDURES PRAIRIE IS1 AND NUICLIEAR GENERATING PLANT NUMBER:
ACTIVATION AND OPERATION F3-7 OF OPERATIONAL SUPPORT CENTER REV: 16 0 Proceduresegments may be performed from memory.
"* Use the procedureto verify segments are complete.
"* Mark off steps within segment before continuing.
"* Procedureshould be availableat the work location.
OWNER: EFFECTIVE DATE O.C. REVIEW DATE:
M.Werner Page 1 of 9
V' PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES
~ACTIVATION AND OPERATION F3-7 .
OF OPERATIONAL SREV: SUPPORT CENTER 16 1.0 PURPOSE This procedure provides instruction for the activation and monitoring requirements of the Operational Support Center (OSC).
2.0 APPLICABILITY This instruction SHALL apply to the OSC Coordinator and all other plant personnel who may report to the OSC.'
3.0 PRECAUTIONS 3.1 Only those personnel designated by this instruction or as requested by plant supervisors, should assemble in the Operational Support Center. All other personnel in Records Room should evacuate when the OSC is activated.
- J 3.2 All personnel assigned to the OSC should remain in the OSC unless.directed to report elsewhere: DO NOT congregate in the Contr'ol Room.
3.3 The OSC is provided with makeup and exhaust ventilation from the Unit 1 turbine deck. This ventilation is equipped with fire dampers and motor dampers that fail shut on fan shutdown. The fans are controlled by local switches on the west wall of the OSC. These fans should be operated together and may be shutdown to mitigate introduction of smoke or airborne contamination fromthe turbine deck into the OSC.
Page 2 of 9
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES
~NUMBER: ACTIVATION AND OPERATION F3-7 OF OPERATIONAL-ZUPPORT CENTER REV: 16 4.0 RESPONSIBILITIES 4.1 The OSC Coordinator is responsible to implement .the actions directed in this procedure.
4.2 The Radiation Protection Specialists are responsible to provide oversight of radiation monitoring of personnel and facilities in the OSC and Control Room.
4.3 Operators, I&C Staff, Electricians, Mechanical Mdinte'nanc6 Staff and Nuclear Plant Service Attendants are responsible to assist in the repair and operation of plant equipment as directed by their supervisors or OSC Coordinator.
5.0 DISCUSSION The Operational Support Center (OSC) is located in the plant Operating Records Room adjacent to the Contfol Room.
"TheOSC pro'vfde'sa central location to assemble the necessary Operations Staff, Radiation Survey Teams, I&C Supervisors and technicians, Electrical & Mechanical Supervisors and staff, and Nuclear Plant Service Attendants. These teams support the operations of the plant during emergency conditions. lf,morepspace is needed for OSC personnel waiting for emergency work, the Operators Lounge area may be used.
The OSC SHALL' be 'activated whenever an Alert, Site Area Emergency, or General Emergency is declared.
Monitoring of the OSC for direct radiation and airborne radioactive materials (particulates and iodine) SHALL be performed to ensure habitability of the OSC.
6.0 PREREQUISITES An Alert, Site Area, or General Emergency has been declared.
Page 3 of 9
J'\ \.
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES
&-CACTIVATION AND OPERATION F3-7
- SOF OPERATIONAL SUPPORT CENTER REV: 16 7.0 PROCEDURE 7.1 Activation and Operation of the OSC 7.1.1 The Operational Support Center SHALL be activated whenever an Alert,
- .. . Site Area or General Emergency is declared.
A.. During normal work hours, the following personnel should immediately report to the Operational Support Center, whenever an Alert, Site Area or General Emergency is announced over the Public Address System.
- 1. Operations personnel onsite, but not assigned to the on-shift crew.
- 2. Mechanical & Electrical Maintenance Supervisors.
- 3. Lead Electricians, Lead Machinists, Lead Riggers.
- 5. Radiation Survey Teams (unless directed otherwise by the Emergency Director or Radiological Emergency Coordinator).
- 6. Nuclear Plant Service Attendants.
- 7. Anyone as requested by their supervisor or the Emergency Director.
Page 4,of 9
EMERGENCY PLAN IMPLEMENT~ING PROCEDURES PRAIRIE ISLAND NUCLEAR 'GENERATING PLANT NUMBER:
ACTIVATION AND OPERATION F3-7 OF OPERATIONAL SUPPORT CENTER REV: 16 B. If activation occurs during off-normal work hours, the Emergency Director SHALL direct the Shift Emergency Communicator (SEC) to notify and activate the emergency organization in accordance with F3-5. The following personnel should report to the OSC to establish an initial compliment of support personnel to assist in the emergency:
- 1. 'Mechanical & Electrical Maintenahce Supervisors
- 3. Designated Electricians
- 4. Radiation Survey Teams-,
- 5. Nuclear Plant Service Attendants
- 6. ' Designated Purchasing & Inventory Control personnel
- 7. Designated operations personnel and extra operators considered necessary by the Shift Supervisor.
- 8. Anyone called in by their supervisor or Emergency Director.
7.1.2 All nonessential personnel should evacuate the Records Room when the OSC is activated.
7.1.3 Additional personnel may augment the Operational Support Center staff as deemed necessary.
7.1.4 The Operational Support Center should remain activated until the emergency has been terminated or as otherwise directed by the Emergency Director.
7.1.5 The Operational Support Center Coordinator should ensure proper activation and operation of the OSC by completing the duties listed on PINGP 574, OSC Coordinator checklist. " Page 5 of 9
PRAIRIE ISLAND NUCLEAR GENERATING PLANT " EMERGENCY PLAN IMPLEMENTING PROCEDURES ACTIVATION AND OPERATION NME: F3-7 OF OPERATIONAL SUPPORT CENTERRE:
1 7.2 Radiological Monitoring Described below are those actions taken to ensure radiological monitoring in the OSC and Control Room.
7.2.1 Verify operation of R-65, OSC Area Monitor.
7.2.2 If R-65 fails, or is not working, set up the AM-2 for monitoring:
A. Obtain the AM-2 from the OS3 Emergency Locker.
B. Plug the AM-2 in and verify the green power light is on.
C. Source check the AM-2 with the button source in the OSC Locker apq verify an upscale reading.
D. If the AM-2 fails (power loss, incorrect reading, etc.) contact the Radiation Protection Group for additional radiation monitors.
7.2.3 At about 15 mR/hr, consider evacuating all nonessential personnel to a low dose rate areas, such as the Operators Lounge area or first floor TSC.
7.2.4 Establish operation of the OSC/Control Room CAM.
A. The CAM is in a hot standby condition with the electronics energized and the blower, chart, and filter paper off. Perform the following:
- 1. Turn the blower switch (located next to the recorder) to the ON position to start the blower, strip chart recorder, and the filter paper.
- 2. Adjust the blower flow rate to 3 SCFM using the toggle switch located on the right side of the CAM. Then adjust suction flov,..
so that 50% is from OSC and 50% is from Control Room.
Page 6 of 9
4 PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES
~NUMBER: ACTIVATION AND OPERATION F3-7 OF OPERATIONAL SUPPORT CENTERRE: 1
- 3. Verify the CAM is in operation (i.e.,'verify the blower, filter, and strip charts are operating, meters are on scale, etc.).
- 4. If the CAM fails to operate properly contact the Radiation Protection Group for additional sampling.
B. Periodically monitor the CAM for airborne particulate and iodine activity.
C. Recommend the following Protective Actions, based on readings from the'CAM.
S1. CAM- Particulate
- lx10" 9 gCi/cc - no protective action necessary.
> lx1i0 establish program of regular portable air samples by the Radiation Protection Group.
If portable air sample results > 1 DAC - consider evacuation of unnecessary personnrel'and limit ex'posures to less than 40 DAC
- hours/week if possible. '
If 0SC portable air sample results > 10 DAC - consider evacuation of OSC personnel to the Control Room. Surplus OSC personnel may go to first floor TSC.
If Control Room air sample results > 10 DAC - respiratory protection for Control Room personnel is required.
SPage 7 of 9
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPILEMENTING PROCEDURES NUMBER:
ACTIVATION AND OPERATION F3-7
- OF OPERATIONALSUPPORT CENTER R EV: 16
- 2. CAM - Iodine If the CAM alarms for iodine (5 x 10-9 itCi/cc) - establish program of regular portable air samples by the Radiation Protection Group.
If portable air sample results > 1 DAC - consider evacuation of unnecessary personnel and limit exposures to less than 40 DAC - hours/week if possible.
If OSC portable air sample results > 10 DAC - consider evacuation of OSC personnel to the Control Room. Surplus OSC personnel may go to first floor TSC.
If Control Room air sample results > 10 DAC - respiratory protection for Control Room personnel is required.
The Radiological Emergency Coordinator (REC) should recommend the use of potassium iodide pills (thyroid blocking agent) if the projected thyroid exposure approaches 25 REM CDE. See F3-18, Thyroid Iodine blocking Agent (Potassium Iodide), for determining projected thyroid exposures. I ,-
7.3 Dosimetry Issue 7.3.1 If the event is a radiological event, an RPS should issue dosimetry to each individual in the OSC and Control Room and log initial dosimeter readings on PINGP 652, Emergency Center Activation Exposure Records.
7.3.2 Refer to F3-12, Emergency Exposure Control, for the Administrative Control and Documentation of Exposure, in regards to work teams dispatched from OSC.
Page 8 of 9
At PRAIRIE ISLAND NUCLVAF. GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES
~NUMBER: ACTIVATION AND OPERATION 17,3-7 OF OPERATIONAL SUPPORT CENTER RV" 1 7.4 Operations Lounge Area The Operations Lounge Area may be used by OSC personnel, as necessary, for some of the following reasons:
7.4.1 Need for more space for large number of OSC personnel on standby for emergency work.
7.4.2 Dose rates in the Operations Lounge Area are lower than dose rates in the primary OSC.
7.5 Considerations during loss of AC power in the OSC
7.5.1 Equipment
A. Phones will continue to operate.
B.' Emergency lights will operate.
C. Flashlights are stored in emergency locker.
D. Dosimetry and radiation meters will continue to operate.
E. The OSC access door will continue to operate.
7.5.2 Re-evaluate the number of craft-workers needed in OSC and Operations Lounge.
7.5.3 Consider sending extra personnel back to work, to Maintenance Lunchroom, Old Admin Lunchroom, NAB Lunchroom, or PITC depending on radiological conditions.
Page 9 of 9
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE
"* Proceduresegments may be performed from memory.
"* Use the procedure to verify segments are complete.
"* Mark off steps within segment before continuing.
"* Procedureshould be availableat the work location.
I K. I O.C. REVIEW DATE: OWNER: EFFECTIVE DATE I L Z <2 0 d 3 C. M. Werner /-30-03 Page 1 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANIT SAFETY PROCEDURE 1.0 PURPOSE The purpose of this procedure is to provide a means to best estimate the degree of reactor core damage from the measured fission product concentrations in water and gas samples taken for the primary system and containment under accident conditions.
2.0 APPLICABILITY This procedure SHALL'apply to the Nuclear Engineering Staff.
3.0 PRECAUTIONS - -1"-*
3.1 The numbers, obt~ifned using this procedure are at best, estirmtes 6nly.
3.2 When making core damage calculations as per this procedure,,considerations should be given toother plant indicators, for example:
3.2.1 Incore Thermocouples.
3.2.2 Reactor Coolant Loop Radiation Monitors (R70/71).,
3.2.3 Containment Radiation Monitors (R48/49).
3.2.4 Hydrogen Concentration in the Containment Atmosphere 3.3 Spiking may occur after a shutdown or significant power change, usually during the 2 to 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> period following the power change. Iodine spiking is a characteristic of the condition where an increase in the normal primary coolant activity is noted, but no damage to the cladding has occurred.
4.0 RESPONSIBILITIES The Nuclear Engineering Group is responsible to estimate the degree of reactor core damage according to the guidance provided in this procedure.
Page 2 of 39
PRAIRIE ISLAND NUCLIEAR GENERATING PLANT PLANT SAFETY PROCEDURE 5.0 DISCUSSION The approach utilized in this methodology of core damage a'ssessment is measurement of fission product concentrations in the primary coolant system, and containment, when applicable, utilizing the post accident sampling 'system.'
Certain nuclides have been selected to be associated with each particular core damage state, i.e., clad damage, fuel overheat and fuel melt. These nuclides reach equilibrium quickly within the fuel cycle. Once equilibrium condition are reached,'a fixed inventory of the nuclides is assumed to exist within the fuel pellet. For these nuclides which reach equilibrium, their relative ratios within the fuel pellet can also be considered to be constant. During operation, certain volatile fission products collect in the gap. The relative ratios in the gap can also be considered to be constant, however, the distribution of the nuclides in the gap is not in the same proportion as the fuel pellet inventory since the migration of each nuclide into the gap is dependent on its particular diffusion rate. The relative ratios of the nuclides analyzed during an accident may be compared to the predicted relative ratios existing in the gap and fuel pellet to determine the source of the fission product release, i.e., gap release or fuel pellet.
Clad damage is characterized by the release of these fission products, i.e., isotopes of the noble gases, iodine, and cesium which have accumulated in the gap and during the operation of the plant. When the cladding ruptures, it is assumed that the fission product gap inventory of the damaged fuel rods is instantaneously released to the primary system. For this methodology it is assumed that the noble gases will escape through'the break of the primary system boundary to the containment atmosphere and the iodines will stay in solution and travel with primary system water during the accident.
Fission product release associated with overtemperature fuel conditions arises initially from the portion of the noble gas, cesium and iodine inventories that was previously accumulated in grain boundaries. In addition, small amounts of the more refractory elements, barium-lanthanum, and strontium are also released.
Fuel pellet melting leads to rapid release of many noble gases, halides, and cesiums remaining in the fuel after overheat conditions. Significant release of the strontium, barium-lanthanum chemical groups is perhaps the most distinguishing feature of melt release conditions.
Auxiliary indicators such as core exit thermocouples, reactor vessel water level, reactor coolant loop radiation monitors, containment radiation monitors, and the containment hydrogen concentration are available for estimating core damage. These indications should confirm the core damage estimates which in turn are based on the radionuclide analysis.
Page 3 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLAUCT-SAFETY PROCEDURE NUJiMBER:
CORE DAMAGE ASSESSMENT F3-17 R EV: 10 6.0 PREREQUISITES An emergency 6f an Alert, Site Alert, or General Emergency has been declared.
7.0 PROCEDURE The program B80DAMASS may be used whenever core damage estimates are desired.
7.1 Request the Radiation Protection Group to obtain the applicable samples to enable an adequate assessment of core damage. See Table 1 for suggested sampling locations.
7.2 Obtain the following plant data at the approximate sample time:
7.2.1 Incore Thermocouple Map 7.2.2 Containment Pressure 7.2.3 Containment Temperature 7.2.4 Containment Hydrogen Concentration 7.2.5 Containment Radiation Level 7.2.6 Containment Sump Level 7.2.7 RVLIS Level 7.3 Perform B80DAMASS according to the instructions in SWI-NE-5 (23) to obtain core damage estimates. Continue with Step 7.15 of this procedure when the B80DAMASS run is complete.
Page 4 of 39
PRAIRIE ISLAND NUC"EAR'GCENERATING PLANT PLANT SAFETY PROCEDURE
- *', * *NUMBER:
CORE DAMAGE ASSESSMENT F3-17 REV: 10 If the computer Is not available, perform the following W4' m manual calculations to obtain core damage estimates.
7.4 Decay correct the specific activities determined by the sample analysis, back to the time of reactor shutdown, as follows:
fThe decay correction may have been accomplished by the INOTE:
.computer during the spectrum analysis. Therefore, this step
[*. **w] may not need to be completed.
. , -- A 0 "
A. i : ,
Ao= e Where: .
A = measured specific activity, pCi/gm or lC/cc X. = decay constant of isotope i, sec" t = time elapsed from reactor shutdown to time of sampling, sec.
A0 = decay corrected specific activity gCi/gm or giCi/cc 7.5 If a parent-daughter relationship exists for a specific isotope, the following steps should be followed to calculate the fraction of the measured activity due to the decay of the daughter that was released and then to calculate the activity of the daughter released at shutdown.
7.5.1 Calculate the hypothetical daughter concentration (QB) at the time of the sample analysis assuming 100' percent release of the parent and daughter source inventory:
()1XB QOA,(e-xt, e-;Bt)+ QOB eXBt QB(t) = KB - 'Ai Where:
'Q°Ai = 100% source inventory (Ci) of parent i, Table 2 or Table 4.
Q0 = 100% source inventory (Ci) of daughter, Table 2 or Table 4.
QB (t) = hypothetical daughter activity (Ci) at sample time.
Ki = if parent has 2 daughters, K1 is the branching factor, Table 3.
XA1 = decay constant of parent i, sec
?1B = daughter decay constant, sec t = time period from shutdown to time sample, sec.
Page 5 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLW w ETY'PROCEDURE I
NUfB El CORE DAMAGE ASSESSMENT F3-17 I
REV: 10 7.5.2 Determine the contribution of only the decay of the initial inventory of the daughter to the hypothetical daughter activity at sample time:
QOB e-Xat Fr QB (t) 7.5.3 Calculate the amount of decay corrected sample specific activity associated with just the daughter that was released.
MOB = Fr X A0 I Where: A0 = decay corrected specific activity (lCi/gm or jiCi/cc) as determined by the analysis.
7.6 Determine the total volume or mass of the medium which was sampled.
7.6.1 Containment Volume:
V = containment free volume (cc's)
= 3.74 X 1010 cc's 7.6.2 Liquid Mass:
A. Liquid temperature < 200OF 28.3 X 103CC Mass (gms) = volume (ft3) X PSTP X ft 3 Where: PSTP = water density at STP = 1.0 gm/cc B. Liquid temperature > 200OF Mass (gms) = volume (ft) X3 POSTPft 28.3 Xl1O 3 CC PSTP (2) X PSTP X ft3 Where: P (2) water density ratio at medium PSTP temperature, from Figure 1 PSTP = water density at STP = 1.0 gm/cc Page 6 of 39
PRAIRIE ISLAND NUCLEAH GENERATING PLANT PLANT SAFETY PROCEDURE 7.7 Determine the total activity of each isotope in each medium.
7.7.1 Containment Atmosphere:
Total containment A0 (liCi/cc) X V (cc's) X Curie I!. 1 X1061Ci Activity (curies),
Where: Ao = Specific activity of "containmentatmosphere
([tCi/cc), decay corrected to time of reactor shutdown and temperature/pressure corrected.
- containment free volume (cc's)
- 3.74 X 1010 cc's , ,
7.7.2 Liquid Sample:
CI irin Total Liquid = Liquid x Ao (gCVcc) x I XCurie 106 Ci Activity (Curies) MASS (gms)
Where: A0 = Specific activity of liquid sample (gCi/gm), decay corrected to time of reactor shutdown.
7.8 The approximate total activity of each isotope in the liquid samples can now be calculated.
Total Water Activity = RCS Activity + Sump Activity + Activity Leaked to Secondary System.
7.9 Now the total activity of each isotope released at the time of the accident can be determined:
Total Activity = Total Water + Containment Released Activity Atmosphere Activity Page 7 of 39
I PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANTl"SAFETY PROCEDURE 7.10 .Utilizing the total activity of each isotope released, calculate the activity ratios of the released fission products.
7.10.11". Noble Gas Ratio Noble Gas Activity Xe - 133 Activity 7.10.2, Iodine Ratio Iodine Activity 1-131 Activity Steady state power conditions may be assumed where power does not vary by more than +/- 10% of rated power level from time averaged value.
7.11 Determine the power history prior to reactor shutdown.
7.12 Using the power history, determine a power correction factor for each isotope, in accordance with, the following guidelines:
Steady state power condition is assumed where the power does not vary by more than +/- 10% of rated power level from time averaged value.
7.12.1 Steady State power prior to shutdown.
A. Half-life of nuclide < 1 day Power Correction Factor = Average Power Level (Mwt) for Prior 4 Days Rated Power Level (Mwt)
B. Half-life of nuclide > 1 day Power Correction Factor = Average Power Level (Mwt) for Prior 30 Days Rated Power Level (Mwt)
C. Half-life of nuclide - 1 year Average Power Level (Mwt) for Prior1 year Power Correction Factor =
Rated Power Level (Mwt)
Page 8 of 39
PRAIRIE ISLAND NUCLgAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
CORE DAMAGE ASSESSMENT F3-17 SREV: 10 7.12.2 Transient power history in which the power has not remained constant prior to reactor shutdown.
~For the majority of the selected nuclides, the 30-day power history prior to shutdown is sufficient to calculate a power correction factor.
A. Power Correction Factor = P (i - e 1"t') e-'-'t° RP Pj,= average power level (Mwt) during operating period t, RP = rated power level of the core (Mwt) tj = operating period in days at power Pj where, power does not vary more than +/-10 percent power of rated power level 'from time averaged value (Pj).
X = decay constant of nuclide i'in inverse days:
t° = time between end of period j and time of reactor shutdown in days.
B. For the few nuclides with half-lives around one year or longer, a power.correction factor which ratios effective full power days to total calendar days of cycle operation is applied.
Power Correction Factor = Actual Operating EFPD of equilibrium cycle Total expected EFPD of equilibrium cycle operation Where: Equilibrium Cycle = three (3) cycles of core operation (approximately 1050 EFPD) 7.12.3 For Cs-134, Figure 2 is used to determine the power correction factor.
To use Figure 2, the average power during the entire operating period is required.
7.13 The total inventory of fission products available for release at reactor shutdown are calculated by applying the power correction factors to the equilibrium, end-of-life core inventories.
Equilibrium Inventory at Power Corrected Inventory = end - of - life (Ci) X Correction (Table 2) Factor Page 9 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLAN'J"SAFETY PROCEDURE 7.14 Determine the percentage of inventory released, for each isotope.
Release- Total Activity Released (Ci) X 100 Corrected Inventory (Ci)
Percentage (%)
7.15 The results of radionuclide analysis may now be used to determine an estimate of the extent of' cre damage.
7.15.1 From Figure 3 thru 15, estimate the extent of core damage by categorizing the percentage of clad damage, fuel over-temperature, and fuel melt.
7.15.2 Compare the calculated activity ratios with those listed in Table 5.
Measured relativeratios greater than the gap activity ratios listed in Table 5 are indicative of more severe failures, e.g., fuel overheat.
7.16 To verify the conclusion of the radionuclide analysis, other indicators should now be used to provide verification of the estimate of core damage.
7.16.1 Containment Hydrogen Concentration:
A. Obtain the containment hydrogen. concentration (%).
- .... Within the accuracy of this methodology; it is assumed that NOTE: *: , recombiners will have an insignificant effect on the hydrogen concentration when it is indicated that extensive I
- zirconium-steam reaction could have occurred.
B. From Figure 16, determine the percentage (%) zirconium water reaction.
C. Table 6 can be used to validate the extent of core damage estimate.
Page 10 of 39
PRAIRIE ISLAND NUCLEARGENERATING PLANT PLANT SAFETYPFROCEDURE NUMBER:
CORE DAMAGE ASSESSMENT F3-17 REV: 10 7.16.2 Core Exit thermocouple Readings:
A. Obtain as many core exit thermocouple readings as possible for evaluation of core temperature conditions.
If a thermocouple reads greater than 1650°F or is reading.
considerably different than neighboring thermocouples, thermocouple failure should be considered. '
B., Compare the thermocouple readings with those in Table 7 to confirm the core damage estimate.
Radiation Monitors in containment may experie'nce errors N TLi
- I during first 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after a DBA LOCA due to thermally 1-3j'k * , induced errors. See Attachment 1 for more information.
7.16.3 Containment Radiation Monitor:
A. Obtain the containment dome monitor readings, R/Hr, from R-48 and/ or R-49.
B. From Figure 17, verify core damage estimate. The exposure rate in Figure 17 is based on the release of only noble gases to the containment. Halogens and other fission products were not considered to be signif icant contributors to the containment monitor reading.
7.16.4 Reactor Coolant Loop Radiation Monitor:
A. Obtain the reactor coolant loop radiation monitor readings, R/Hr, from R-70 and/or R-71.
B. From Figure 18, determine estimated core damage.
7.17 All indicators should confirm any core damage estimates. If radio-nuclide analysis and auxiliary indicators do not agree on core damage estimates, then recheck of indications may be performed, or certain indicators may be discounted, based on engineering judgment.
Page 11 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLAINTSAFETY PROCEDURE Table 1 Suggested Sampling Locations Principal Other Scenario Sampling Locations Sampling Locations Small Break LOCA Reactor Power > 1lýA RCS Hot Leg, Containment Atmosphere Reactor Power < l%* RCS Hot Leg**
Large Break LOCA Reactor Power > l%7* Containment Sump, Containment Atmosphere, RCS Hot Leg Reactor Power ( 1%* Containment Sump, Containment Atmosphere Steam Line Break RCS Hot Leg, Containment Atmosphere Steam Generator Tube RCS Hot Leg, Secondary Rupture System Indication of Signif- Containment Sump, Containment icant Containment Sump Atmosphere Inventory Containment Building Containment Atmosphere, Radiation Monitor Alarm Containment Sump Safety Injection RCS Hot Leg Actuated Indication of High RCS Hot Leg Radiation Level in RCS
- Assume operating at that level for some appreciable time.
Page 12 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE Table 2 Fuel Pellet Inventory '
Fuel Pellet Inve'ntory*
Nuclide Half Life Inventory Curies**
1.0 x 107 Kr 85m 4.4 1.85 x 107 Kr 87 76 m Kr 88 2.8 h 2.69 x 107 5
Xe 131m 11'.8 d 2.94 x 10 .
Xe 133 5.27 d 9.26 x 107 Xe 133m 2.26 d 1.35 x 10 Xe 135 9.14 h 1.77 x 107 I 131 8.05 d "4.54 x 107 1 132 2.26 h 6.65 x 107 1 133 20.3 h 9.26 x 107 1 135 6.68 h 8.33 x '107 Rb 88 17.8 m 2.69 x 107 Cs 134 2 yr 1.09 x 107 6
Cs 137 30 yr 4.96 x 10 Te 129 6R.7 m 1.51 x 107 107 Te 132 77.7 h 6.65 x Sr 89 52.7 d 3.70 x 107 90 28 yr 3.36 x 106 Sr 12.8 d 7.91 x 107 Ba 140 40.22 h 8.33 x 107 La 140 92.5 m 7.07 x 107 La 142 Pr 144 17.27 m 5.81 x 107 Inventory based on ORIGEN run for equilibrium, end-of-life core.
Westinghouse, 2-Loop, 1650 Hwt Plant Page 13 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE Table 3 Parent-Daughter Relationships Parent Daughter Parent Half-Life* Daughter Half-Life* Kw*
Kr-88 2.8 h Rb-88 17.8 m 1.00 1-131 8.05 d Xe-131m 11.8 d .008 1-133 20.3 h Xe-133m 2.26 d .024 1-133 20.3 Xe-133 5.27 d .976 Xe-133m 2.26 h Xe-133 5.27 d 1.00 1-135 6.68 h Xe-135 9.14 h .70 Xe-135m 15.6 In Xe-135 9.14 h 1.00 1-135 6.68 h Xe-135m 15.6 m .30 Te-132 77.7 h 1-132 2.26 h 1.00 Sb-129 4.3 h Te-129 68.7 m .827 Te-129m 34.1 d Te-129 68.7 m .680
- b-1Z9 4.3 h Te-129m 34.1 d .173 Ba-140 12.8 d La-140 40.22 h 1.00 Ba-142 11 m La-142 92.5 m 1.00 Ce-144 284 d Pr-144 17.27 m 1.00
- Table of Lsotopes, Lederer, Hollander, and Perlman, Sixth Edition
', Branching decay factor Page 14 of 39
PRAIRIE ISLAND NUCMEAF" GEwNERATING PLANT PLANT SAFETY PROCEDURE Table 4 Source Inventory of Related Parent Nuclides Nuclide Half-Life Inventory, Curies Xe-135m 15.6 m 1.97 x 107' Sb-129 4.3 h 1.49 x 107 Te-129m 34.1 d 3.74 x 106 Ba-142 11 m 7.65 x 107 Ce-144 284 d 4.83 x 107 Page 15 of 39
PRAIRIE ISLAND NUCLEAR C iEiirAa INrIN*'I.AP PLANTSAFETY PROCEDURE
- NUMBER
CORE DAMAGE ASSESSMENT F3-17 REV: 10 Table,5 Isotopic Activity Ratios of Fuel Pellet and Gap Isotopic Activity Ratios of Fuel Pellet and Gap*
Nuclide Fuel Pellet Activity Ratio Gap Activity Ratio Kr-85m 0.11 0.022 Kr-87 0.22 0,022 Kr-88 0.29 0.045 Xe-131m 0.004 0.064 Xe-133 1.0 1.0 Xe-133m 0.14 0.096 Xe-135 0.19 0.051 1-131 1.0 1.0 1-132 1.5 0.17 1-133 2.1 0.71 1-135 1.9 0.39 Noble Gas Ratio = Noble Gas Isotope Inventory Xe-133 Inventory Iodine Ratio = Iodine Isotope Inventory 1-131 Inventory The measured ratios of various nuclides found in reactor coolant during normal operation is a function of the amount of. "tramp" uranium on fuel rod cladding, the number and size of "defects" (i.e., "pin holes"), and the location of the fuel rods containing the defects in the core. The ratios derived in this report are based on calculated values of relative concentrations in the fuel or in the gap. The use of these present ratios for post accident damage assessment is restricted to an attempt to differentiate between fuel overtemperature conditions and fuel cladding failure conditions. Thus the ratios derived here are not related to fuel defect levels incurred during normal operation.
"-/
Page 16 of 39
C C T
M Containment M Percent Radlogas M and Type Monitor Core Exit Core of Fission Hydrogen Fission s/hr 10 hrs Thermocouples Core Monitor Damage Products Product after Readings tUcovery Category Released Ratio"** shutdown** (Deg F) J'ndicstien (Vol %1) z M- .7 I x e -3 Wn eirsdl d ag Not Appl1cable , 7,0 4o uncovery Neglilflblc Xe-!33 ! t x 3~ z C:
1-131 C I x -30 1-133 C 1 x ~
30 Kr-87 10-3 . 0 O1 0-50% clad dame Xe-133 103. 0.1 Xr-87 = 0.022 0 - SO 750 - 1100 Core uncovery 0 - 6 0~ z 1-331 I0:3 - 0.3 1-133 - 0.71 1-133 0"3 - 0.1 50-100% clad damage Kr-87 0.01 - 0 02 Kr-87 = 0.022 50 to 100 1300 - 1650 Core uncovery 6 - 13 0, Xe-133 0.1 - 0.2 1-131 0.3 - 0.5 1-133 = 0.71 1-133' 0.1 - 0.2 C-)
r-5 0-50% fuel Pell et Xe-Kr, Cs, I Kr-a7 = 0.22 100 to 1.15E4 > 1650 Core uneovery 6 - 13 overtemperature 1 - 20 0D Sr-B& 0 - 0.1 1-133 a 2.1 0
30-100% fuel pe liet Xe-Kr, Cs, I Kr-87 = 0.22 1.1E4 to 2.3E4 > 1650 Core uoeovery 6 - 13 overtemperature 20 - 40 CD Sr-Ba 0.1 - 0.2 1-133 = 2.1 Wi
=m 0-o50 fuel melt Xe, Kr, Ce, I 40-70 1r-?7 z 0.22 2.3E4 to 2.7E4 > 1650 Core uncovery 6 - 13 Sr-Bf 0.2 - 0.8 Pr 0.1 - 0.2 1-133 = 2.1 W!,
S0-100% fuel meit Xe, Xr, Cs, 1, Te Kr-87 = 0.22 > 2.7E4 > 1650 Core uncovery 6 - 13
> 70 Sr, Be 3 24 1-133 u 2.1 Pr ) 0.8 Characteristics of Categories of Fuel Damage
- This table is'intended to supplement the methodology outlined in this report and should not be used without referring to this report and vithout considerable engineering Judgement.
/ These values are from Figure 17 and should be revised for times other than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, -
w Kr-87 1-133 Xe-133' 1-131 "-n
-L CD 0
-,4 0 m
0 Co C
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANBSAF ETY'PROCEDURE NUMBEF CORE DAMAGE ASSESSMENT F3-170 REV: 10 Table 7 Expected Fuel Damage Correlation With Fuel Rod Temperature Fuel. Da.ge Temperature *F*
No Damage < 1300 Clad Damage 1300 - 2000 Ballooning of zircaloy cladding ) 1300 Burst of zircaloy cladding 1300 - 2000 Oxidation of cladding and hydrogen generation > 1600 Fuel Overtemperature 2000 3450 Fission product fuel lattice mobility 2000 2550 Grain boundary diffusion release of fission 2450 3450 products Fuel Melt > 3450 Dissolut.ion and liquefaction of U02 in > 3450 the Zircaloy - ZrO eutectic 2
Melting of remaining UO 2 5100
- These temperatures are material property characteristics and are non-specific with respect to locations wirhin the fuel and/or fuel cladding.
Page 18 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SArETY PROCEDURE Figure 1 Water Density Ratio(Temperature vs:STP)-,
goo-.
4 P j-0 0 0- 0 0 STP Page 19 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLAN]'T SAFETY PROCEDURE Figure 2 Power Correction Factor For CS-134 Based on Average Power During Operation
~aJ
,-3 La.
0 0' fl 1 0 0 0 0 0 0 o C) 0 H U'"
C, Page 20 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE Figure 3 Relationship of % C!ad Damage With % Core Inventory Released of XE-1 33 1 - -
S.... .........
.............. ...... !.-' "--i' "*' '
S........... ........ ....... ......
............ * "! ........... i. . . .. ........
S........... ...... '....-..L.-.*i Legend "AVERAGE ..........i......
, LOW BURNUP HIGH BURNJUP 0.1" I
...... ...... . . ' S.... ..... ..... :: . .- - . .
(/d Ed I.
S. ;..... ...... ....
z L~LJ 7 :
z 0 1 o 0.0 *~~ I"* _
I - - . - :
II IJIJl*
- eV V l I I I I . . ... * ' . I .'g _
0.1 100 CLAD DAMAGE (7)
Page 21 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT_ SAFETY PROCEDURE Figure 4 Relationship of % Clad Damage With % Core Inventory Released of 1-131
- l S..... .... .. :.......;.....:. S.......... *:......: ..... *-:--:L
..........*...... ..- .:. -f -
S.... .......... .. . ; . . . . .
........ ".. .i :
Legend AVERAGE I.- -
0.1 * ......
"." .. ..- ...... 3 ..:'
LOW BURNUP ' * " . :" . .."....
... HIGH BURNUP
.. ... .o ................
S....... . .... .
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....... .... I.... ....... ..
/... -. .. . ...... ...
._J ... .".... ..
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o "°_°
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t... ....... . ..: . ." - - - ' 'l A
-S
-n-1 llliltl* ' T - w t "'* . . . .
- I I I t
. ý 11
,i 0.*l 10 i00 CLAD DAMAGE (,-)
Page 22 of 39
6S Jo S3 e6ed oot I I I I
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-':.':--'!-i-- ........ .......
- J
{
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE Figure 6 Relationship of % Clad Damage With % Core Inventory Released of KR-87 U. I-,
0.1 1 ...... :........ .... ; ... t..;...;. ,I 9S........................ - .- I-
~
.. . .. ~
...... S........... :.............. ...
..... -... . - ------ S............*..... . ...**°.... * *; t S.......... ...... .... '...?. .:..=
S......... i... .". .. . '. . - . .
0.01-Legend AVERAGE I m I 71 S.......... ...
LOW BURNUP S..........*..... ."- - -- - - -. . . . . . ... ... ..- .
HIGH BURNUP
................. .. .. :..'.. .......... "......i "! "" ":
0 .-... ...
LiJ (n
LJ
-J Lai
.. ..7...
0 I)1*1.
Urj-1 I .d LIJ Lix
.......... Z z
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........ ....... ;..... .... *--..'.;L
.... .. . .... ..... .. *. ... ........ °.......:*..
.1 lr- rr r r y T
- ir ,---r-r-.--
. , 4 U.UUUUI-r . .. 1" .... T- I i 60 0.1 to CLAD DAMAGE (%)
Page 24 of 39
Figure 7 Relationship of % Clad Damage With % Core Inventory Released of XE-1 31 M S.......... . -. . .. . . . . S........... :...... ;..... . . --. -.....
..... ."... .... .;...:. S.......... .-...... .'---}*-. -÷*. " -*
'2 S.......... -.. ... .. _ : . S........... .......... ......
i... ... '
2
... ..... ... ..... ... ?. ? ....... ,,........ .. o..... , ...°*....
.. ...... " "? ....... ....... .... ...,: . : . °....
S.......... -.. ...... .* i! !
. ... ..' ... %..:i"ii S............i.........- --...:'..
Legend S........... i. . . . . -..
"AVERAGE LOW BURNUP
'.-. 0.1 HIGH SURNUP i -
S.. ...... "......" -- "-"T K...
LLJ LU, z . ...
.. ...... ..... S....... ... . . }....}...I T...:..*.:
I-.
z LJ M 0.01 f ,. " ,
- m _1 m F.. >..... . .. S.. ......... "..... - -: i
. . . ....- . ...T.......................-" ' .-'........
!i ...........
S.....
" ' ........ S..
S.
0.001 I 0.1 10 100 CLAD DAMAGE (%)
Page 25 of 39
66 JO 92: aeud Wz) 3sVVo aVID 001 at
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HunGfaOOd MndVS .LXNd I INId DNLLVU3N-u1 uv3onN aGNfSI t31HlVUd
PRAIRIE ISLAND NUCLEAR G ENERATING PLANT PLANT.SAFETY PROCEDURE NUMBER:
CORE DAMAGE ASSESSMENT F3-17 REV: 10 Figure 10. Relationship of % Clad Damage With % Core Inventory Released of 1-135 4-
............. ...... " ....i---- -:*
S.....
S........... .... . *-...'.....i
................. - .... .. S.... ..... ."...........
."T!i
.. ...... S........... ......... : -- -....... '
. - -- '. ....... l S.... . . . !....;
- i i 0.J 77Legend I
.......... I . .......
AVERAGE S.......... *...... ..-....-- ............. ...
- - -. i
- LOW BURNUP ................ ........
"HIGH BURNUP t-i LI U)
I LiL
,..J S.......... ,...... :... ° " °-"I .
LIA S
I-,
z . . . ....
LLJ z
. -. 4 I v ull "
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0kc .. ... ...............
7 ........ . ..... .....
.7 :; *: i 0.0001 - --
~
iii
......... ..... .. .. .. ....... S.. ........... I. °. ,*
S......
...... ... .i S..
0 I 0.1 10 100 CLAD DAMAGE (-/)
Page 28 of 39
PRAIRIE ISLAND NUCL&AR'GENERATING PLANT PLANT SAFETY PROCEDURE Figure 11 Relationship of % Fuel Over Temperature With % Core Inventory Released of XE, KR, I, or CS 1I)(1 1UU *......... ...... *......... ....... ".. .. ' .- . .*. S.................
i ,i *1 I S......... ....... .......... *...... .: ....-.. - . .. . . S................. *;......... :....... : ..-...-... :.*.. -.
......... ....... ": ......... : ...... ...... :...;... .I..... S................. ......... !...... !..... ....... '..*..
S......... ...... i..........* ...... *....i '.. -'.-. -'.
.......Legend S................. i......... *...... i.......-... ~
- S.............. ............ i..:...*. ..... . .*.
... Legend NOMINAL
.. ................. ......; ..ii. .E.... ._V
.,- [*. ....MINIMUM .. ... .... :....:.... ..; . .
MAXIMUM ______
10-Li
.1. ;.. ..... ..... -. Y;....I; ;.-.....
- '.* * ':...............:...ii'**
Cl) S.... -.. ................ . .... ....
S....... ....... !.. ......- i-i. .. .... ... .. ..... ...
0 Iz S......... .... .i.;.-.. * .
........... y Y......
2:
Li z
...... ..... .... iiiii~i tLi 0
0
- 1
.,z.::p../~.:....:-... ...._:i ::i:::
F". ....... *-.... !.... !........
' ..7<: ----
I L I
................. i..........i ...... .........-.-..
7.. .)....... ..'. .... TT
.... . ............ ...... ....- . .:- ~ '
0.1 ,,,I , -
- S....
I 10 100 FUEL OVERTEMPERATURE (%)
Page 29 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANTSAFETY, PROCEDURE Figure 12 Relationship of % Fuel Over Temperature With % Core Inventory Released of BA or SR
... . .... ............ ... ÷. . .. . -. . ... . ...... ' ' '.......
S................. ".. S..
.... ... ................... !. .... +. o ;. . . . . S... S
... . . ... .- .. ......... ...... i" S................. ......... i...... *...*-:i ° S....... ..... ... ...... ..........z..... , .. .. . . .. . S........... ....
... ...... ... T!
.............. .... .. °... .. .. .. . ... ... . ... .. ...
0.1
. . . . .. . . ; S MINIMUM ...
...MAXIMUM ... ... ............ ...:...:........ ".....-.. ..... ,,,. .
~........ ..........
.. J . .. ...
-1, 0.01 0 ... .. ... *....... .. :......t,........... ..-...-.... S
........ *......... *.... .. ..... . -. . . --.--. S z /........ . S iI ............. .. .... :..... ... ... :.. .
ct:
0.001 + m _ * . * . !
I
- ..... .. S
.'.. . . ....... : ...-....... I .......... ..
S................... *. ...... * ..... *......! i : '
S... .....
S................ .......... i.. . . . . . - -. . .
. ...........
- o °
....... . l .
S................ ;......... ;......... ..... ... ;.. . . .
. ...... .. .} - ...-
S.. --.--
..-- .S........... ..... ......... ....... ...... ....... :.....:
S..... ..... ....... ........ .... .. .;. .. . . .
0.0001 10 I 00 I
FUEL OVERTEMPERATURE(7.)
Page 30 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT -PLANT SAFETY PROCEDURE
~NUMBER:
CORE DAMAGE ASSESSMENT F3-17 REV: 10 Figure 13 Relationship of % Fuel Melt With % Core Inventory Released' oI XE, KR, I, CS or TE 100 .......................... ... ....... .... ....... . . . . . . ...-...-.... '.
.. o. ; . . . . . .... .. . . .. . . . ....
- ........... ...... ; .......... -j .
.... S............ .......
S ............ .. ...o. ........ *........ . ................. ° . .......
Legend
... L*;o ! .....
i l i ............. . ..... ..
NOMINAL ..... ...........
"~MINIMUM o 1.-.....
.. . 07_ IVCXIMUM:.-: , . .'..... - ..... . ....
_____ , ........i........ :i...... :..... .---- '........
S~~............ .. ...:............................. .
LAJ . * ............ ........ , ................... i.. . . . . . . ... ; :
Lii 10
-- ... ,... /
/_- :
0 ... * .. ..- ...... . .: ...... : .. : ......: .... .......... ....... ......... ;......":...... ."!............
o 1 .......... ..... ........ .. .. .......... . . .
.. ......... ........... . t. .. . .
o0.1 - ' 10 100 FUEL MELT (%)
Page 31 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT Figure 14 Relationship of % Fuel Melt With % Core Inventory Released of BA or SR ItUU S............. "......... *... .. 4... . . .4 . .. . . * ................... ....... . .
S..
. .. . .. . .. ,t........ .. . .... ;. .,.... .....
- ......... i.......i..... ..-.
---.-.- S
.. .. . .... "........ z. ...... : . .; .... ; .. ;
S.. ; ....... ..... ....:....*.*? S................ . ......... -...... .............. i..e*
. .. ......i . ..... ... ..... :..... :........ ! + *......... !...... .......*.... ......S................. .
Legend ".-. .ii" ... .. .... ... . . I ....
10-
..... .... ..-..... .i : - - --- -. :* S MINI* UM ... ... =.... . .. ...... . ".*... . .. ".* . . .. . ..
.. ... ........ ... *... ..... , 4 .... ... . S
"... MAXIMUM .. ...
.. ......... :.""? '. - " ": ' S............
-j
-J LaJ r)
.. ...... ./ . >...
.... .. ......... *... ... -.-. -.. S............
LaJ S-. . .i. .. ... .......... . ..... *...... *........ .. * ... ;.
LU L 4 S.. . ........
i~i S................ ......... ! .......... .,. : . .:. . . .: '..
0
.4--
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S.... .......... .....................-. .. . : ... .
0.O1 S. . .. . . . " ' rw ' = . J 1 10 100 FUEL MELT(%)
Page 32 of 39
PRAIRIE ISLAND NUCbLEAR GENERATING PLANT -PLANT SAFETY PROCEDURE Figure 15 Relationship of % Fuel Melt With % Core Inventory Released of PR '
dAA I l 'll 'l .....
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...... o°oo.o.. o.. ........ ........ °°....... ... °°o MINI__ _MU__M S................* ......... *.......; ...... *....,...°.. *... .
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S..~~ ~.... .......... ........ ........ . . .......-.
. o o .
-0.001 I 100 I 10 100 FUEL MAELT (7.)
Page 33 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANTSAFETYPROCEDURE NUMBEF CORE DAMAGE ASSESSMENT F3-17 REV: 10 Figure 16 Containment Hydrogen Concentration Based on Zirconium Water Reaction Ica'l 20 0
z 0 15 5
z 0
Lii 10-U.
0 T, . ... . . . . . . . . . . . . ...
5-0 I 0 20 40 60 80 I(00 ZIRCONIUM WATER REACTION (%)
Page 34 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFE'TY PROCEDURE Figure 17 Percent Noble Gases in Containment 1000000, i *. :...: i ili! ".. ..
S.....
- ........ i +'""* "-"l*-..........."I..... . ..... I =";=! '...............
o----.......
S- .... ..... .... -
M. O...M . "selease.:..::. ........... ... ....... " ...
S.....
- -"" * "--"""-" .*L* . ............... .:.... .........
S....." ... * " .:* * ',*........--.-:-. -,,-- OR MELTI *-'--
Gas Rl~eease .' j_ * "il*" :
1000Di : :.......... " "*
================ ...............................
S..... IFUEL OVERTEMPERATURE 1000-* ! .......
..... .. t ......: *- .............*. ...1 .......
.T.-. . 'L .... -.. ............
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~........ .' *-
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, ~ , *..... =; "'"...... ......
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. .. S
................-...................... ..... 0.Z Noble Gas Release
.C AILURE . S..... ..... ... .
10 ...
I. .. .....
~-ANS ~8INormal Operating "Noble Cas Reles .................
.;.':7... : :: :... . - ....- ..*.;. . :. .. . .
.. . . .... ........ ., "... .. . ;-- - -. '-:.-i
- 0. NORMAL RCS :;::":*::::'::"'
o~o , -S.:...
1 0 o100 1000 TIME AFTER ACCIDENT (HOURS)
Page 35 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE Figure 18 RCS Dose Rate vs. RCS Activity Concentrations 1 Hour After Shutdown 1.0E405 DBA: WOG 100% DBA 100% NG RELEASE 50%oIdines ii Gap: W OG ..
.OE+04 100% Radionuclides in Fuel Gap fil
-7 w ,p B1.OE+03 mI Z I II 1'1.0E+02 0~ - :ii 100 % GAP I
- "RELEASE ::
1.OE+01 I HIM fi 1 I 1 I .... I:: 1. Monitor response based on 0.025 R/hr Sper uCiuml.
.. 2. RCS Vol. = 1.8E8 cc 1.IE 0
1.OE+02 1.0E+03 1.OE+04 1.0E+05 1.0E+06 1.0E RCS CONCENTRATION (Total uCilml)
Page 36 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE S CORE S, ' DAMAGE ASSESSMENT NUMBER:" F3-17 REV: 10 Attachment 1 Thermally Induced Current Errors in Containment Radiation Monitors
- 1. R-48/R-49 & R-70/R-71Thermally Induced Errors R-48/49 or R-70/71 signals may experience errors during the first 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after a DBA LOCA. Industry testing of high range radiation monitor (HRRM)'syster-hs,has revealed that signal errors or~the loss of signal are the result of thermally induced current (TIC) and/or moisture intrusion into the coaxial connectors. Based on the EPRI Plant Support Engineering study, worst case estimated errors are summarized below:
Time After Postulated- Estimated Errors, in DBA., Readings,
- 1 minute > 3000 R/hr
- 2 minutes - 100 R/hr
- 8 minutes - 15 R/hr
- 2 to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> - - 9 R/hr
> 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> No Effect from TIC More background information concerning thermally induced current in high range, radiation monitors is described in Section II1.
Please note that errors in the range of +/-10 R/hr one, hour after a postulated DBA has minimal effect on our assessment of fission product release to containment when we are considering magnitudes of > 100 R/hr reading to be confirmation of fission product released to containment. ,
- 2. Background on Thermally Induced Current (TIC) on High Radiation Monitors
Background
Excerpts from: PINGP Response to High Range Radiation Monitor Cable Study: Phase II, Report No. TR-1 12582 November 2000.
Transient signal errors have been observed in industry testing of the high range radiation monitor (HRRM) system. At PINGP, these are plant radiation monitors RE-48 and RE-49.
The investigation into this issue revealed that signal errors or the loss of signal are the result of thermally induced currents (TICs) and/or moisture intrusion into the coaxial connectors. Information Notices, IN 97-45 and IN 97-45 Supplement 1, were issued by the NRC to alert licensees to these potential issues.
Page 37 of 39
PRAIRIE ISLAND NUCLEAR G!3ENERATING PLANT PLANT-SAFETY PROCEDURE NUMBER:
CORE DAMAGE ASSESSMENT F3-17 REV: 10 Attachment 1 Thermally Induced Current Errors in Containment Radiation Monitors EPRI Plant Support Engineering (PSE) was tasked to study the significance of this issue, which resulted in the issuance of.TR-1 12582, "'High Range Radiation Monitor Cable Study: Phase Il". This study'was focused on the therhially induced current phenomena since moisture intrUsibn issuesa're well understood within the industry and have more generic applications. Phase I'bfthe EPRI study confirmed that-TIC existed and was significant under thermal transients. Phase II of the study identifiepl the sources of the TIC and developed a mathematical model for cable responses to thermal transients.
r . 1/4. ,
Study Results and Ana lysis" ' " -
Using the developed profiles in the Phase II study, the actual-amplitude,-duration, and sign of HRRM signal errors to be expected could be determined. From this data, PINGP was able to ascertain:the. approximate expected signal error for the HRRMs during the postulated DBA. Theexpected radiation readings due to the TIC phenomena, based upon the worst case cable length; are as follows: . - "
50 seconds '-3872 R/hr 100 seconds 88 R/lr 500 seconds '13.2 R/hr 8000-15000 seconds ' -8.8 R/hr' ' - ,
>15000 seconds no effect fromTC -T From 8000 seconds until 15000 seconds, the HRRMs could provide a "fail" alarm, based on the required "keep alive" signal current of 1 E-1 1 amps since the current may drop to 8.8E-1 1 amps. It should be noted that any significant radiation releases would drive the current back up and the HRRMs would function properly, except for the-8.8 R/hr error that may be present. After 15000 seconds (4.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />), there would be no TIC effects on the HRRMs.
The installed HRRM cable at PINGP is the worst case tested cable, Rockbestos RSS-6-104, and is in greater lengths than were tested, 130 feet tested vs. 290 feet installed (worst case). Other variables that could significantly effect the TIC phenomena are, 1) the tested cable was not installed within conduit whereas the PINGP cable runs are installed entirely within conduit, 2) the temperature differential of the test samples, 100 degc, is greater than the temperature differential from the PINGP accident profile, 68 degc, 3) the EPRI mathematical model was developed based on hypothetical LOCA profiles, which are more severe than the PINGP LOCA profile, and 4) consideration regarding whether the test methodology of immersion of the test samples into a ice bath and then to a boiling water plunge is representative of what the cable would experience during an actual transient.
Page 38 of 39
PRAIRIE ISLAND NUCLEAR G ENERATING PLANT ,LAN- P I S r -I P-*OCED.,-URE NUMBER:
CORE DAMAGE ASSESSMENT F3-17 REV: 10 Attachment 1 Thermally Induced Current Errors in Containment Radiation Monitors PINGP Response to HRRM Signal Error During the initial phase of any postulated accident, it would not expected to see indication of actual fuel damage for the first 10-15 minutes. Ifindeed the alarms would come in for RE-48 and RE 49, Operations would be occupied with accident mitigation and monitoring tasks during this time period and this alarm, even though acknowledged, would be ignored during this period. Other parameters would be available for alarm validation, i.e., core exit temperatures, RVLIS, radiation monitors located in the Auxiliary Building, etc. Due to the nature of the TIC phenomenon, the radiation leyel readings, even if the alarms have come in, would be decreasing. Again, this is validation of an erroneous signal and not actual core damage.
For emergency plan response and possible SAMG considerations, the TIC phenomenon would no longer be affecting the radiation monitors and!or due to the earlier alarms and decreasing readings that were noted, it would be confirmed that no fuel damage had occurred and these were indeed erroneous readings. A general site emergency alarm would be activated at 1000 R/hr, but as cited previously, this is well after the expected error signal has been significantly reduced. Other variables would be available to verify possible fuel damage and any possible actions required within the emergency plan procedures would not occur until after the TIC phenomena has either passed or has been verified to be erroneous.
Page 39 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES PERSONNEL AND EQUIPMENT NUMBER:
MONITORING AND F3-19 DECONTAMINATION REV: 8 RE5,ER-E 'O
"* Proceduresegments may be performed from memory.
"* Use the procedure to verify segments are complete.
"* Mark off steps within segment before continuing.
"* Procedureshould be availableat the work location.
O.C. REVIEW DATE: OWNER: Effective Date
()I~2o II Sc M.Werner i-6-05 Page 1 of 7
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES PRAIRI LNUCEAPERSONNEL AND EQUIPMENT NUMBERO MONITORING AND F3-19 _
SDECONTAMINATION REV: 8 1.0 PURPOSE This procedure provides the guidance for contamination monitoring, contamination control, and decontamination procedures for personnel and equipment.
2.0 APPLICABILITY '
This instruction SHALL apply to all Emergency Directors (ED) and all members of the Radiation Protection Group (RPG)"
3.0 PRECAUTIONS 3.1 All personnel decontamination should be supervised by the RPG.
3.2 The safety of personnel SHALL take precedence over the monitoring of personnel and vehicles for radiation/contamination control purposes. Monitoring of personnel.
and/or vehicles SHALL be terminated (or not implemented) if such monitoring is known or suspected to be increasing the hazard to personnel during evacuation.
3.3 If any personnel are suspected to have received a biologically significant dose (dose exceeds twice the NRC Annual 10CFR20 Occupational Dose Limits), refer directly to the F3-12, Emergency Exposure Control.
4.0 RESPONSIBILITIES 4.1 The RPG has the responsibility to monitor personnel and equipment to determine if contaminated. When personnel or equipment is found contaminated, the RPG has the responsibility to document contamination levels and to coordinate the decontamination of personnel or equipment.
4.2 The Radiological Emergency Coordinator (REC) has the responsibility to authorize use of elevated contamination levels as listed in Attachment 1 under Emergency Guidelines.
4.3 The ED has the overall responsibility to ensure that radioactive contamination monitoring, control, and decontamination is being conducted throughout the emergency.
Page 2 of 7
PRAIRIE ISLAND NUCL,*EARAG ENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES PERSONNEL AND EQUIPMENT NUMBER:
MONITORING AND F3-19 DECONTAMINATION REV: 8 5.0 DISCUSSION During emergency conditions,. large areas of elevated surface contamination levels are probable within the plant boundaries: The REC should evaluate the contamination levels and determine if it would be beneficial to raise the contamination limits to the elevated guidelines in Attachment 1. The RPG should then control entry into the plant in accordance with these guidelines and monitor personnel and equipment exiting the plant per these guidelines. Decontamination of personnel and equipment to levels below these guidelines should be performed per applicable decontamination procedures.
6.0 PREREQUISITES The Prairie Island Nuclear Generating Plant has declared an Emergency classification.
7.0, PROCEDURE "
7.1 The RPG is responsible for contamination monitoring, control and decontamination.
1 7.1.1 All attempts should be made to maintain contamination levels below the normal guide!ines, as per Attachment 1. -"
7.1.2 During emergency conditions, elevated contamination limits may be authorized by the REC, as per Attachment 1.
7.2 Personnel Monitoring and Decontamination 7.2.1 Monitor personnel who evacuated directly out of the Radiological Controlled Area first.
7.2.2 Survey and document results on PINGP 985, Personnel and Vehicle Survey Log.
7.2.3 IF contamination is found, THEN initiate PINGP 915, Whole Body Survey Form.
7.2.4 Segregate monitored personnel into 3 groups using the following criteria.
Highly Contaminated > 5000 CCPM Contaminated > 100 CCPM NOT Contaminated < 100 CCPM Page 3 of 7
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IM~PLEMEtITIfNG PROCEDURES PERSONNEL AND EQUIPMENT NUMBER:
MONITORING AND F3-19 DECONTAMINATION REV: 8 7.2.5 Decontaminate Highly Contaminated personnel L> 5000 CCPM) first, followed by the Contaminated personnel L> 100 CCPM).
7.2.6 IF personnel contamination is found around the individual's mouth and nose THEN obtain a nasal smear.
7.2.7 IF the results are > 100 CCPM, THEN indicate Bioassay Required on PINGP 915, (See RPIP 1126, Contamination Monitor Alarm Response and Personnel Decontamination).
7.2.8 Attempt to reduce any contamination detected on an individual in accordance with RPIP 1126.
A. IF dose rates allow personnel habitability,,THEN use.Decon Showers at Access Control.
B. IF dose rates allow personnel habitability, THEN use old Admin Building shower facilities.
C. Use EOF Decon Shower at Prairie Island Training Center.
7.2.9 IF the Normal Guidelines are NOT achieved, THEN refer to the REC about using the Emergency Guidelines in Attachment 1.
7.2.10 IF contamination is coincident with injury, THEN follow procedures outlined in F4, Medical Support and Casualty Care.
7.2.11 Decontaminate personal clothing and shoes to Normal Guidelines.
7.2.12 IF Normal Guidelines CANNOT be obtained after reasonable efforts, THEN dispose of the items as contaminated waste OR the REC may authorize the use of Emergency Contamination Guidelines as specified in Attachment 1.
Page 4 of 7
PRAIRIE ISLAND NUCLEAR.GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES PDPERSONNEL AND EQUIPMENT NUMBERO iMONITORING AND F3-19 DECONTAMINATION REV: 8 7.3 Coordinated decontamination for Emergency Responise personnel remaining onsite, and conductirid emergency work activities.
7.4 Vehicles Monitoring and Decontamination
. SuVehicle eaonhtoering and decuntamonation shPuld beP peBormed as time allows depending on evacuation urgency.
- ~~MAJOR'VEHICLE CONTAMINATION MAY POSE A '
RADIATION HAZARD TO PERSONNEL CONDUCTING SURVEYS AND APPROPRIATE PRECAUTIONS SHOULD BE TAKEN OR SURVEYING SUSPENDED UNTIL LATER.
special attention to the air filter, tires, and radiators.
C. IF fuel damage is suspected, THEN smear areas where contamination is found with a cloth smear and save for alpha counting.
.D. Smear exterior of the Vehicle using two (2) masslins each covering 5 sq. ft. from, hood, roof, trunk or pick-up bed.
E. IF smears are > 100 CCPM, THEN log the vehicle as contaminated.
7.4.3 Initiate PING P 986, Vehicle Survey Form, for vehicles found contaminated.
7.4.4 Tape the PINGP 986 to the inside of the windshield with any saved smears.
Page 5 of 7
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES PERSONNEL AND EQUIPMENT NUMBER:
MONITORING AND F3-19 DECONTAMINATION REV: 8 7.4.5 Hold Contaminated vehicles in a designated area for later decontamination.
7.4.6 IF major vehicle contaminati6n exists, THEN evacuate personnel as
-. -. - quickly as possible using vehicles that meet the Guidelines of
"" Atta'chmentfl. Ou'tside assistance may. be requested as necessary.
7.5 Upon termination of emergency condition, survey the exterior, and interior surfaces of the vehicles.; Paying special attention to the air filter, tires, radiators, etc.
Contamination levels SHALL be returned to the Normal Guidelines, using approved decontamination procedures as outlined in F2, Radiation Safety, and D-13, Decontamination.-
Page 6 of 7
PRAIRIE ISLAND NUCLEAR.GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES PERSONNEL AND EQUIPMENT NUMBER:
MONITORING AND F3-19 i DECONTAMINATION REV: 8 Attachment 1 Contamination Limits
',-CONTAMINATION LIMITS NORMAL GUIDELiNES - EMERGENCY GUIDELINES REMOVABLE, LOOSE SURFACE ,
DPMf100 cm2
' Bycm. 100 DPM/1 00 cm 2 5000 DPM/100 cm 2 S10 DPM/100 cM2 500 DPM/100 cm 2 FIXED 100 CPM 500 CPM Based on Manual of Protective Action Guides and Protective Actions for Nuclear Accidents, EPA 400-R-92-001, May 1992, Table 7-7. Frisker response: 1mR/hr = 5000 CPM Cs 137.
- 1. Guidelines are based on using pancake probe.
- 2. By Portable survey instruments are located in all Emergency Centers and at both Assembly Points.
- 3. ax Portable survey instrument is located in the Hotcell Emergency locker.
Page 7 of 7
PINGP 1300, Rev. 5 TCN#. r- 007 0 Effective Date: - - 0g-.
Retention: 6 years Expiration Date:
Document Type: 1.190 Page 1 of t TEMPORARY CHANGE NOTICE C'AWT 1 1 1l'*
I . . ;;A I. CHANGE REQUEST Rev: Project ID #:
Procedure ,
WO #:
Title:
C.d"e.
C-i-I P/e-,
Description of Change:
(e.g.pagea#stepo#, :5 /" . 'p: 7 _r_,a4- o-C r C-tCa2.5 4, j.i?" rr-d-summary/reason) 40 te' qg.A a
. F Is this an OC reviewed procedure/critical WO?. Yes J No El Is a permanent procedure change needed ? Yes P No E] If Yes, submit PINGP 436 Originator. ,'fl, ,'e t>Employee #: Z* T* 6"*Date: /"1 '-' 3 I. INITIAL REVIEWS A. Is this a change in intent? Yes El No Wif No, go to II.B
- 1. OC Review Initial/Date (if procedure is OC reviewed/critical WO):
B. 10 CFR 50.59 Review
- 1. Is this change or procedure exempt from 10CFR 50.59 screening per 5AWI 3.3.5, App. A or B? Yes JR"No [I If Yes, go to II.C
- 2. Complete 50.59 screening per 5AWI 3.3.5/PINGP 1229. Screening/Evaluation #
C. Does this change affect Special Reviews? Yes El No :W If Yes, document Special Review(s) in CHAMPS D. Does this change affect plant operation or Tech Specs? Yes El No ZdIf Yes and work is in progress, SS notified by _ --(initial)
Yes No El If Yes, attach marked-up procedure pages to white copy E. Do controlled copies need to be updated?
F. All master/working copies updated? Yes El NAP G. Is training on temporary change needed'?-, Yes El No Wif Yes, PINGP 1268 or PINGP 1224 (Ops Only) issued III. REVIEW AND Date:
Reviewer.
Date:
Approver:
Forward white copy to Procedure Control II A)
IV. POST REVIEWS (required for OC reviewed procedures/critical WO; leave blank if reviewed/approved in OC Review Initial/Date: Procedure Approver: Date:
V. TEMPORARY CHANGE DELETION (leave blank if controlled copies not updated, refer to II.E)
Date A. Reviewer:
(Unit Management Staff)
Date:
B. Approver:
(Unit Management Staff, SRO for OC reviewed procedures/critical WO)
White Copy - Procedure Control ' Yellow Copy - Attach to procedure Form 17-5645 (4-02)
PRAIRIE ISLAND NUCLEAR GENERATING PLANT 0
O F EMERGENCY PLAN IMPLEMENTING PROCEDURE NUMBER:
EMERGENCY HOTCELL PROCEDURE F3-23.1 REV: 12 7.3 Gamma Analysis Preparation 7.3.1 Pipet 10 ml of diluted coolant sample from the 1 L volumetric to a 10 ml vial.
~Sample should be diluted to give a contact reading of under I milliren/hr contact. The diluted sample should NOT exceed 25 millirem/hr contacL 7.3.2 Verify that the indicated dose rate on the 10 ml vial is capable of being counted on extended geometry in EOF Countroom.
the 7.3.3 Label the vial with the sample point, date, time, and dilution factor to sample prior to sending to EOF Countroom.
7.3.4 Place the 10 ml vial in the shielded carier for transport to the EOF Countroom.
7.3.5 WHEN radioactive gas, charcoal, or particulate samples are received, THEN ensure all samples are labeled with date and time of sample, sample point, sample volume and/or correction factor, and flow rate.
7.3.6 Store all samples in the Hotcell Shielded Area until transported to the EOF Countroom.
7.4 Boron Analysis 7.4.1 Using the I L sample prepared in Step 7.1, Sample Preparation, analyze b .,.... =,lchericarSt
................ og r y.
in accordance with"t*.
- 4RPI P 3314, 93'-,
cjfrb f,,*, o,,, /roC.QoC 4 4fe...or 7.4.2 Log the results on PING0655, Post Accident Cheical Analysis Report.
7.4.3 Dispose of all radioactive waste according to Step 7.6, Post Accident Sample Waste Storage and Disposal.
Page 6 of 7
"ICr 9D03S-009O PRAIRIE ISLAND NUCLEAR GENERATING PLANT 6EMERGENCY PLAN IMPLEMENTING PROCEDURE EMERGENCY HOTCELL PROCEDURE NME F3-23.1]
7.5 Chloride Analysis
" * ~Chloride analysis SHALL be completed within 4 days of accident.I THE REACTOR COOLANT SAMPLES TAKEN IN AN ACCIDENT CONDITION HAVE THE POTENTIAL TO BE HIGHLY RADIOACTIVE. THIS MAY GIVE RISE TO DOSE RATES FAR INEXCESS OF WHAT WOULD NORMALLY BE ENCOUNTERED. THE ION EXCHANGE COLUMNS ON THE ION CHROMATOGRAPH COULD HAVE CONTACT READINGS OF UPTO 10 RIHR.
7.5.1 Using the 100 ml sample prepared in Step 7.1, Sample Preparation a,.,,,l,,..,,-,~nn
. . .A * ,L,;k r}~D o t4 . A -* _ IL- _- - -L -,,.,
analyze in- eaordanec with R.IP -- %A I3, An'il& Ly lullI llaII A-4 r~-os a :tV e a aP I~d~
-;&E e? #?.~~a/(dS5 az"r 7.5.2 Log the results on PINGP 655, Post Accident Chemical Analysis Report.
7.5.3 Dispose of all radioactive waste according to Step 7.6, Post Accident Sample Waste Storage and Disposal.
7.6 Post Accident Sample Waste Storage and Disposal
~Ensure samples are labeled. "TO BE SAVED" or '"TO BE DJPED" before storage In shielded area.
7.6.1 Place all capped or covered radioactive sample waste in the Hotcell Shielded Area.
7.6.2 IF additional waste samples are added to the Hotcell Shielded Area, THEN survey the Hotcell general area radiation levels. Add additional shielding, as necessary.
7.6.3 . IF making subsequent entries into Auxiliary Building, THEN return the sample waste to the Sample Room for disposal down the affected unit's Sample Hood Drain.
Page 7 of 7