L-PI-03-041, Emergency Plan Implementing Procedures
| ML031130026 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 04/14/2003 |
| From: | Solymossy J Nuclear Management Co |
| To: | Document Control Desk, Office of Nuclear Security and Incident Response |
| References | |
| L-PI-03-041 | |
| Download: ML031130026 (60) | |
Text
Committed to NuclearExcellen Praire Island Nuclear Generating Plant Operated by Nuclear Management Company, LLC L-PI-03-041 1 OCFR50.4 April 14, 2003 U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT DOCKETS 50-282 AND 50-306 LICENSE NOS. DPR-42 AND DPR-60 PRAIRIE ISLAND EMERGENCY PLAN IMPLEMENTING PROCEDURES Furnished with this letter are the recent changes to the Prairie Island Nuclear Generating Plant Emergency Plan Implementing Procedures F3. This submittal includes the following documents:
INDEX:
Emergency Plan Implementing Procedures TOC REVISIONS F3-17 F3-22 F3-23.1 Core Damage Assessment Prairie Island Radiation Protection Group Response to a Monticello Emergency Emergency Hotcell Procedure Rev. 11 Rev. 17 Rev. 13 DELETIONS:
None TEMPORARY CHANGE DELETIONS:
2003 0080 F3-23.1 Emergency Hotcell Procedure INSTRUCTIONS:
Instructions for updating the manual are included.
This letter contains no new commitments and no revisions to existing commitments.
1717 Wakonade Drive East
- Welch, Minnesota 55089-9642 Telephone: 651.388.1121 ADq5
USNRC NUCLEAR MANAGEMENT COMPANY, LLC L-PI-03-041 Page 2 As per 10 CFR 50.4, two copies have also been provided to the NRC Region IlIl Office and one to the NRC Resident Inspector. If you have any questions, please contact Mel Agen at 651-388-1121 Extension 7210.
h M. Solymossy
-Sfe Vice President, P ie Is4nd Nuclear Generating Plant CC Steve Orth, USNRC, Region III (2 copies)
NRC Resident Inspector-Prairie Island Nuclear Generating Plant (w/o attachment)
Attachment 1717 Wakonade Drive East
- Welch, Minnesota 55089-9642 Telephone: 651.388.1121
Mfst Num:
2003 -
0259 Date
- 04/02/03 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:
F3-22 17 PRAIRIE ISLAND RADIATION PROTECTION GROUP TO A MONTICELLO EMERGENCY F3-23.1 13 EMERGENCY HOTCELL PROCEDURE F3-17 11 CORE DAMAGE ASSESSMENT Temporary Change Deletions:
2003 0080 F3-23.1 EMERGENCY HOTCELL PROCEDURE UPDATING INSTRUCTIONS Place this material in your Prairie Island Controlled Manual or File. Remove revised or cancelled material and recycle it.
Sign and date this letter in the space provided below within ten working days and return to Bruce Loesch or Mary Gadient, Prairie Island Nuclear Plant, 1717 Wakonade Drive E.,
Welch, MN 55089.
Contact Bruce Loesch (ext 4664) or Mary Gadient (ext 4478) if you have any questions.
Received the material stated above and complied with the updating instructions Date
I PRAIRIE ISLAND NUCLEAR l
Title:
GENERATING PLANT l Emergency Plan Implementing Procedures TOC I l Effective Date : 04/02/03 NOTE: This set may contain a partial distribution of this Document Type. Please refer to the CHAMPS lApproved By:
A6&t41 C
X Module for specific Copy Holder Contents.
WPS Suptl Document #
Title Rev F3-1 ONSITE EMERGENCY ORGANIZATION 19 F3-2 CLASSIFICATIONS OF EMERGENCIES 32 F3-3 RESPONSIBILITIES DURING A NOTIFICATION OF UNUSUAL 18 EVENT F3-4 RESPONSIBILITIES DURING AN ALERT, SITE AREA, 28 OR GENERAL EMERGENCY F3-5 EMERGENCY NOTIFICATIONS 21 F3-5.1 SWITCHBOARD OPERATOR DUTIES 8
F3-5.2 RESPONSE TO FALSE SIREN ACTIVATION 9
F3-5.3 RESPONSE TO RAILROAD GRADE CROSSING BLOCKAGE 8
F3-6 ACTIVATION & OPERATION OF TECHNICAL SUPPORT CENTER 16 F3-7 ACTIVATION & OPERATION OF OPERATIONAL SUPPORT 16 CENTER (OSC)
F3-8 RECOMMENDATIONS FOR OFFSITE PROTECTIVE ACTIONS 20 F3-8.1 RECOMMENDATIONS FOR OFFSITE PROTECTIVE ACTIONS FOR 13 THE ON SHIFT EMERGENCY DIRECTOR /SHIFT MANAGER F3-9 EMERGENCY EVACUATION 18 F3-10 PERSONNEL ACCOUNTABILITY 19 F3-11 SEARCH & RESCUE 8
F3-12 EMERGENCY EXPOSURE CONTROL 14 F3-13 OFFSITE DOSE CALCULATION 15 F3-13.3 MANUAL DOSE CALCULATIONS 11 F3-13.4 MIDAS METEOROLOGICAL DATA DISPLAY 7
F3-13.5 ALTERNATE METEOROLOGICAL DATA 5
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I I
Page I of 3
PRAIRIE ISLAND NUCLEAR GENERATING PLANT Title : Emergency Plan Implementing Effective Date : 04/02/03 Procedures TOC Document #
Title F3-13.6 F3-14.1 F3-14.2 F3-15 F3-16 F3-17 F3-18 F3-19 F3-20 F3-20.1 F3-20.2 F3-21 F3-22 F3-23 F3-23.1 F3-24 F3-25 WEATHER FORECASTING INFORMATION ONSITE RADIOLOGICAL MONITORING OPERATIONS EMERGENCY SURVEYS Rev 11 11 9
RESPONSIBILITIES OF THE RADIATION SURVEY TEAMS DURING A RADIOACTIVE AIRBORNE RELEASE RESPONSIBILITIES OF THE RADIATION SURVEY TEAMS DURING A RADIOACTIVE LIQUID RELEASE CORE DAMAGE ASSESSMENT THYROID IODINE BLOCKING AGENT (POTASSIUM IODIDE)
PERSONNEL & EQUIPMENT MONITORING & DECONTAMINATION DETERMINATION OF RADIOACTIVE RELEASE CONCENTRATIONS DETERMINATION OF STEAM LINE DOSE RATES DETERMINATION OF SHIELD BUILDING VENT STACK DOSE RATES ESTABLISHMENT OF A SECONDARY ACCESS CONTROL POINT PRAIRIE ISLAND RADIATION PROTECTION GROUP RESPONSE TO A MONTICELLO EMERGENCY EMERGENCY SAMPLING EMERGENCY HOTCELL PROCEDURE RECORD KEEPING DURING AN EMERGENCY REENTRY 22 17 11 10 8
18 9
9 10 17 18 13 7
8 F3-26.1 F3-26.2 F3-26.3 F3-29 F3-30 OPERATION OF THE ERCS DISPLAY RADIATION MONITOR DATA ON ERCS ERDS -
NRC DATA LINK EMERGENCY SECURITY PROCEDURES TRANSITION TO RECOVERY 7
7 1
18 6
Page 2 of 3
PRAIRIE ISLAND NUCLEAR GENERATING PLANT Title : Emergency Plan Implementing Effective Date : 04/02/03 Procedures TOC Document #
Title Rev F3-31 RESPONSE TO SECURITY RELATED THREATS REVIEW OF EMERGENCY PREPAREDNESS DURING OR AFTER NATURAL DISASTER EVENTS 7
F3-32 2
Page 3 of 3
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE Ni 2,
ti 4.
- Procedure segments may be performed from memory.
Use the procedure to verify segments are complete.
- Mark off steps within segment before continuing.
- Procedure should be available at the work location.
D.C. REVIEW DATE:
Q21sol C1OcZ-
.^...=n.
uwI'4L11:
EFFECTIVE DATE
/V-e2-0,3 l
M. Werner Page 1 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT 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 3.1 The numbers obtained using thins procedure are at best, estimates only.
3.2
-When making core damage calculations as per this procedure, considerations should be given to oth6i plant indicators, for example:
3.2.1 Incore Thermn6coules.
3.2.2 Reabto C66olant Loop Radiation Monitors (R70/71).
3.2.3 Containment Radiation Monitors (R48/49).
3.2.4 Hydrogen Concentrati6n 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 NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
CORE DAMAGE ASSESSMENT' F3-17 j~REV; 1
5.0 DISCUSSION The appr6ach utilized in this methodology6 of core daiMage psessmrbit. is measurement of fission productconcenitrations i'n'th"e primary'do-lant svstem, and containment, when
-applicable, utilizing 'the post aczcidert'~sarhplin' s tem.i' Certain nuclides have been selected to be associated with each paritidclar core damage' state, i.e., clad damage, fuel overheat and fuel melt. These nuclides reach equilibrium quickly within the fuel cycle. Once equilibrium cohditidr ree'reached, 'a fixd irivbntory 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 pe consicpd,oe
,no e
wepri thde distribution of the nu'clides in the gap is not in thb same proportion as the fuel pellet inventory.since temrnigration of.each nucliqei into tpaqapis.dependet on its particular diffusion rate. The ieI6 ive riti6s of t1hfie piiless4;sqJ
,d epurindtg oqnjaq,$ident 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:,gapxgase.-or fuelpellet.
Clad damage is characterized Py the re~lase. 9fthew,,fission.prooucts, i.e..isotopes of the noble gases; iodine, and cesium which hafve accumulated in the gap and during the operation of the plant. When the cla~dqinr ruptures,, it is.assumedthat the fission product gap inventory of thfe damaged fuel rods is instantaneously released to the primary system. For this methodologyjti;s assumed that the,n9ble gases will escape through the break of the prirmary system boundary to the containment atmosphere and the iodines will stay in solption and tray el itp4 system water during the accident.
system wat Fission pro'duct release associated-with bvperjeMpepature fuel conditions.arises initially from the portion of the noble casesium and iodinie inventories that was previously accumulated in grain boundaries. In addition, small amounts of the more refractory elements, barium-lanthanum, and strontium are also released.. -.f' Fuel pellet jmeltirig leads t&orapid r'lease"6t'ra7in'rioble'gases, halides, 'and 8esiums remaining in the fuel after overheat'condihions!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 Z.j
'X '
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-+ "j 4',
con E PLANT SAFETY PROCEDURE
',AM,.GE A.aSSESSMENT Ei NUMBER:
F3-17 REV:
11 6.0 PREREQUISITES An emergency of an Alert, Site Alert, or General Emergency has been declared.
7.0 PROCEDURE T
[he program B80DAMASS may be used whenever core OT, I5 1
damage estimates are desired.
l 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 Thormoccuple Map 7.2.2 Containment Pressure 7.2.3 Containment Temperature a
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
! I I
K-
PRAIRIE ISLAND NUCLEAR GENERATING PLANT Kl W
I the computer is not available, perform the followIng TTE.manual calculations to obtain core damage estimates.
v 7.4 Decay correct the specific activities determined by the sample analysis, back to the time of reactor shutdown, as follows:
i The decay correction may have been accomplished by the icomputerduring the spectrum analys. lTherefeore, this step Frni'ay not need to be completed.
1
- --j A(A ao,
/e
.,A l
-_.rl r'
'o
-:~ ~~~~~~~
.7:rr
- v n
Where:
~
A
=
measured specific activity, [tCi/gm or jiCi/cc
}=
decay constant of isotope i, se'c' t
=
time elapsed frldr-bedt&rsh Nddw' to time ofsaimpling, sec.
AO =
decay corrected specfic ap v ity g
ju,Irm or RGqVcp 7.5 If a parent-daughter relationship exists for a specific isotope, the following steps should be followed to calculate the fraction of-th'stnme'as-ired'activity'due to the decay of the daughter that was released and then to calculate the activity of the daughter released at shutdowns-
.i ;-,i 1
7.5.1 Calculate the hypothetical dauqhtericpnceIntration (QB) at the time of the sample analysis a~suiirig I00pe'rehit'rblease'of the-parent and daughter source inventory.-
Ki Q01 ete-;Bt+Q B
QB (t)
=
K; AB; QAI(e e
Q e
XB -XAJ Where:
I, A
QOA
=
100% source inventory (Ci) of parent i; Table 2 or Table 4..
Q0B
=
100% source inventory (Ci) of daughter, Table 2 or Table 4.
QB (t)
=
hypothetical daughter activity (Ci) at sample time.
K!
=
if parent has 2 daughters, K, is the branching factor, Table 3.
iXAi
=
decay constant of parent i, sec B
=
daughter decay constant, sec -'
t
=
time period from shutdown to time sample, sec.
Page 5of 39;
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE 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:
Fr
=
Q0 B eXBt 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 AO Where: Ao =
decay corrected specific activity (gCi/gm or giCi/cc) as determined by the analysis.
7.6 Determine the total volume or rnmss of the medium which was sampled.
7.6.1 Containment Voiurne V
=
containment free volume (cc's)
=
3.74 X 1010 cc's 7.6.2 Liquid Mass:
A.
Liquid temperature < 200OF Mass (gms)
=
volume (ft3) X PSTP X 28-3Xl10cc vVhere: PSTP
=
water density at STP = 1.0 gm/cc B.
Liquid temperature > 200OF Mass (gms)
=
volume (ft3) X p
(2) X PSTP X 3
PsTp Where:
P (2)
=
water density ratio at medium PST?
temperature, from Figure 1 PSTP
=
water density at STP = 1.0 gm/cc Page 6 of 39
PLANT SAFETY PROCEDURE I
l I 111 MPnS=n I
lJ__
KI t
e I
7.7 Determine the total activity of each isotope in each medium.
7.7.1 Containment Atmosphere:
Total containment Ao (lCcc) X V (cc's) X xc urie Ci
.Activity (curies) 4 Where:
=
Specific'activity of containment atmosphere
,-,(Ci!cc),
decay corrected to time of reactor
,, *shutdown and temperature/pressure corrected.
V'-
cor.laiarnent free volume (cdos) -
=
3.74 X 10",Utc's<,4x' 7.7.2 Liquid Sample: '-
>'! S a.~
~~.4., A Total Liquid Activity (Curies)
Where:
=
4
=
Liquid X A (ltLP~o)
"XI~~C MASS (gms)
- J He ri I, 4 Specific activity of liquid sample ([iCigm), decay
..,corrected to time of reactor shutdown.
7.8 The approximate total activity of each isotope in the liquid samples can now be
-calculated.
4 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:
- - wr,,4
__arTV f
Total Activity
- Released, Total Water +
Activity Containment Atmosphere Activity
- .I i
Page 39>"
PRAIRIE ISLAND NUCLEAR GENERATING PLANT U:, '
Or
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C M..w -
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o=%
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I PLANT SAFETY PROCEDURE 7.10 Utilizing the total activity of each isotope released, calculate the activity ratios of the released fission products.
7.10.1 Noble Gas Ratio 7.10.2 Iodine Ratio
=
Noble Gas Activity Xe -133 Aciivity 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.
I 7.11 Determine the power history prior to reactor shutdown.
7.12 Using the power history, determine a power correction factor for each accordance with the following guidelines:
isotope, in Steady state power condition is assumed where the power does not vary by more than +/- 10% of rated power level from time averaged value.
I 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 Power Correction Factor = Average Power Level (Mwt) for Prior 1 year Rated Power Level (Mwt)
Page 8 of 39
PRAIRiE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
l F3 CORE DAMAGE ASSESSMENT E3-17 7.12.2'* Trahsient power history in which the 'power has not rermained constant prior to reactor shutdown.
For the majority of the selected nuclides, the 30-day power NTE Z
historyp~rior to shutdown is sufficient to calculate a'power correction factor.
A.
Power Correction Factor
=
.TP,.,.-,,,averagqpower leveJ (Mwt).during operating period t, S
,a z
.eotv I
=
rated power 'level of the core r(Mwt) operatilignperiod in dayspo er P, where.power does not vary 'orre" khan'1i0 percent power of rated power
,, levelfomtime aeeragerd r _ I rit F.: r VoiiuA ] ' 'r;§'! -
=
decay constant of nuclide i in inverse days.
d to -=
time between end'of-priodd-j'Ahd time of reaQtor,,
- h't'doWh in, days.<'- ;
i B.
For the few nuclides with half-lives around one year or longer, a power correction factorwhich 'rtios effective full power days to total calendar days of cycle operation is applied.
Powr CActual Operating EFPD of equilibrium cycle Power Correction Factor =-
A.
Total expected EFPD of equilibrium cycle operation three (3) cycles of core operation Where: Equilibrium Cycle =
(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 PLANT SAFETY PROCEDURE NUMBER:
FR 3-17 REV:
i 1 7.14 Determine the percentage of inventory released, for each isotope.
Release Total Activity Released (Ci) X 100 Correctedinventory (Ci)
Percentage (%)
7.15 The results of radlonuclide analysis may now be used to determine an estimate of the extent of core 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 re:aiive ratios 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 Conlta;nment Hydrogen Concentration:
A.
Obtain the containment hydrogen concentration (%).
i :~
Within the accuracy of this methodology, it is assumed that recombiners will have an insignificant effect on the T
qgX~t^rhydrogen concentration when it is indicated that extensive 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 I 0 of 39
PRAJRIE ISLAND NUCLEAR GENERATING PLANT PLANTSAFETY PROCEDURE NUMBER:
wa CORE DAMAsGE ASSESSMENt
'F3-17 REV:
11 7.16.2 Core Exit thermocouple Readings:
t '
A.
Obtain as many core exit-thermocouple readinbs as possible for evaluation of coreetemp6rature conditions.
If a thermocouple reads greaterthan 1650°Fror is reading c onsiderably different than'neighb9rjfng therrpocouples, thermocouple failure should be considered.
1
.t r
t..
4Rr; 1!.
B --"Compare thedtherirf'cbbple readih'gs WIth those in Table 6 to confirm the core damage estimate.."- '
Radiation Monitors in containmentP yY0xxp erience errors dOuTE ringfirst 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> afteirait-b LtA A Au'e to thermally Induced errors.' See~Att'acimriiij' foY%'ore infjrmation.
8
~
1 *_*4x
- J' (
Itbff 41 s iS'
)'J0 la' I-
. 4t 7.16.3 -. -Containment Radiation Monitor:
- ii-;:.'
te" e',i X
A.
Obtain t~he containment domejlrdnitorrreadings, R/Hr, from R-48 and/ or R-49.
B.
From Figure 17, verify core damage estimate. The exposure rate in Figure -17 isbase'd on the rele§se of only hoble'gases t-the contaihmeint.';`Haldgenrs and otherfission products were not bonsidered to'be s~i nfica'nt cointribuforsto the containrhent monitor reading. :!
X':.- 1 f
- 2 1
7.16.4 Reactor Coolant Loop Radiation MonitQr:
A.
Obtain the reactor coolant loop radiati7n monitor readings, RtHr, 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 1 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE Table 1 Suggested Sampling Locations Principal Sampling Locations Other Sampling Locations Scenario Small Break TCOCA Reactor Power > 1.*
Reactor Power < 1%*
RCS Hot Leg, Containment Atmosphere RCS Hot Leg**
Large Break LOCA Reactor Power > 17.*
Reactor Power < 1%*
Containment Sump, Containment Atmosphere, RCS Hot Leg Containment Sump, Containment Atmosphere Steam Line Break RCS Hot Leg, Containmen:
Atmosphere Steam Generator Tube RCS Hot Leg, Secondary Rupture System Indication of Signif-Containment Sump, Containment icant Ccu'alnrment Sump Atmosphere Inventory Containment Building Radiation Monitor Alarm Containment Atmosphere, Containment Sump Safety Injection Actuated RCS Hot Leg RC.S Hot Leg Indication cf High Radiation Level in RCS i
Assume operating at that level for some appreciable time.
If a RCS hot leg sample is unavailable and the RHR system is operating, obtain a RIR system sample.
However, for a RIR systcm sample to be a good representation of the RCS, the primary water should be circulating through the system.
Page 12,of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT I
PLANT-SAFETY PROCEDURE
_NUMBER:
i
, <' P3-17 REV:
11 Nuclide Kr 85m Kr 87 Kr 88 Xe 131m Xe 133 Xe 133m-t Xe 135- -_
I 131 I-132 1 133 I 135 ----
Rb 88 Table 2 Fuel Peliet Inventory' Fuel Pelle't Inventory*
Half Life Inventory Curies**
.1[
- 3.
4 4.4 1.0 x Id I
I
.76m 1.85-x 0 2.8 2.69 x1D7.
<1.8 d,
,2.94,x-10 5 V., I, 7,*\\
5.27 d 9.26 x 1.0 2.26 d--
1.35 x 10
_- 9.14-h
1.77 x 107
- 8. 05 4.54 x 10
-2.26-h
6.665-x-1 7
20.3 h r---,.
r
-9+26 x 10' Cs Cs Te Te 134 137 129 132 6.68 h i-17.8 m -
2 yr 1
30 yr 68.7 m '-
77.7-h - - -
I I
_8.33 x 10 s ~ W t
- 2* 6 9 -
- 1 0,;
7
.. P9 x-,10 4.96 x 10 1.51 x 107.
6.65 x.107 ----
Sr Sr Ba La La Pr 89 5i.7 d '
3.70 x 107 90
' 2
- -..28 yr r3-\\
rL J336 C 10 140 12.8 d
,7.91 x.107.
140 40.22 h 8.33 x 107 142 92.5 m 7.07 x 107 144 17.27 m 5.81 x 107 Inventory based on ORIGEN run for equilibrium, end-of-life core.
Westinghouse, 2-Loop, 1650 Mwt Plant Page 13 o6 39.1
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDU:RE TablQ 3 Parent.-Daughter Relationships Parent Parent Half-Life*
Kr-88 2.8 h 1-131 8.05 d 1-133 20.3 h 1-133 20.3 h Xe-133m 2.26 (1 I-135 6.68 h Xe-135m 15.6 in 1-135 6.68 h Te-132 77.7 h Sb-129 4.3 h Te-129m 34.1 d Sb-129 4.3 h Ba-140 12.8 d Ba-142 11 m Ce-144 284 d Table of Tsoropes, Lederer,
'* Branching decay factor Daughter Rb-88 Xe-131m Xe-133m Xe-133 Xe-133 Xe-135 Xe-135 Xe-135m I-132 Te-129 Te-129 Te-129m La-140 La-142 Pr-144 Daughter Half-Life*
17.8 m 11.8 d 2.26 d 5.27 d 5.27 d 9.14 h 9.14 h 15.6 m 2.26 h 68.7 m 68.7 m 34.1 d 40.22 h 92.5 m 17.27 m K**
1.00
.008
.024
.976 1.00
.70 1.00
.30 1.00
.827
.680
.173 1.00 1.00 1.00 Hollander, and Perlman, Sixth Edition Page 14 of 39
PRAIRIE, ISLAND NUCLEAR GENERATING PLANT A
.N I
PLANT SAFETY PROCEDURE Table 4 Source Inventory of Related Parent Nuclides Nuc lide Xe-135m Sb-129 Te-129m Ba-142 Ce-144 Half-Life
-15.6-m 4.3 h 34.1.d 11 m
' 284-d J (
- nz,.
Inventory, Curies 1.97 x 10 1.49 x 10 3.74 x 106 7.65 x 107 L
7 4.83 x 10 C
- v V I I
j
,' 8;~ CI I
I
-_ r -1
,r
.,v I\\
I I
(S
-r, I. - ! -:
r S t i
V I
I i
I
_I t
Page 15 of 39 '
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE Table 5 isotopic Activity Ratios of Fuel Pellet and Gap Isotopic Activity Ratios of Fuel Pellet and Gap*
Nuclidc Kr-85m Kr-87 Kr-88 Fuel Pellet Activity Ratio 0.11 Xe-131m Xe-133 Xe-133m Xe-135 0.22 0.29 0.004 1.0 0 0.14-0.19 1.0 1.5 C.. I Gap Activity Ratio 0.022 0.022 0.045 C).004 1.0 0.096&
0.051-1.0 C1.17 0.7' 1-131 1-132 1-133 1-135 1.9 0.39 Noble Gas Ratio =
Iodine Ratio =
Noble Gas Isotope Inventory Xe-133 Inventory Iodine Isotope Inventory 1-131 Inventory The measured ratios of various nxiclides 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
Core Damage Category Pel I
a.Pe
, ani ofI
' I
'I Prc Re!
Table 6 Characteristics of Categories of Fuel Damage Containment Fcent:
Radlogas I Type Monitor Cori Fission Fission R/hr 10 hrs
,Thei
)ducts',
. Product after
- Real leased Ratio"*I" I shutdown**-
(Deg r Exit rmocouples.
ding,
I F-)
Core Uncovery Indication -
Hydrogen Monitor (Vol % H2)
No clad damage 0-50% clad damage 50 - 100% clad damage I
0 - 50% fuel pellet overtemperature 50-100% fuel pellet overtemperature 0 - 50% fuel melt 50 - 100% fuel melt KR-87
<1X104 Xe`133> <1x104 1-131
<1X103 1-133- *- c1X104 Kr.87 104-0.01 Xe-133 104-0.1.
', 131 104-0.3 133 104-_0. 1.
- Kr-87 0,01 - 0.02 Xe-133- 0.1-0.2
.131
-Q;3-0.5 1-133:
0.1-0.2 Xe-Kr, Cs, i 1-20. o Sr-Ba' b-0.1 XeKr, Cs, I 20 - 40,.:
SrBa - 0.1-0.2 Xe'Kr, Cs, I 40-70 Sr-Ba 0.2 - 0.8 Pr 0.1-0.8 Xe, Kr, Cs, I, Te
^
- 970' Sr, Ba > 24 Pr>0.8 No Applicable
< 750 No uncovery Kr-87 = 0.022 0-50 750 - 1300 Core uncovery 1-133 = 0.71 I.
t-1
.II Kr-87 = 0.022 50 to 100 1300- 1650 1-133 = 0.71 Coro uncovery C
.u v
1(
Core uncover-y Negligible I
0-6
-13
-13" 6 - 13 1
6-13 I
I-6 - 13
!' -- 13 - -'
I Kr-87 = 0.22 I
I.
I I,
1-133=2.1.
o Kr-87 = 0.22 1-133 = 2.1 r1Il00 to 1.1 5E4.
E t C-4 1.15E4 to 2.3124
> 1650
> 1650 Core uncovery
> 1650 Core uncovery Kr-87 = 0.22 2.3E4 to 2.7E4 1-133 = 2.1 Cc Kr-87 = 0.22
> 2.7E4
> 1650
)re uncover/
I 1-133 = 2.1 Characteristics of Cateaories of Fuel Damaae*
SThis table is Intended to indicate whether there Is fuel damage.
_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 />.
- -'Kr-87 1-133
,xe -133 1-131 I
.. I t
Page 17 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE Table 7 Expected Fuel Damage Correlation With Fuel Rod Temperature For Information Only - See Note Below I
Fuel_ Damage Temperature OF*
No Daniage
< 1300 Clad Damage Ballooning of zircaloy cladding Burst of zircaloy cladding Oxidation of Cladding and hydrogen generation Fuel Overtemperature Fission product fuel lattice mobility Grain boundary diffuision release of fission producis Fuel Melt Dissolution and liquefaction of UO2 in rhe Ztxcaloy - ZrCG eutectic Melting of remaining U0 2 1300 - 2000
> 1300 1300 - 2000
> 1600 2000 -
3450 2000 - -2550 2450 - 3450
> 3450
> 3450 5100 These temperatures are material property characteristics and are non-specitic with respect to locations within the fuel and/or fuel cladding.
Page 18 of 39
PRAIRIEISLAND NUCLEAR GENERATING PLANT P
I.,
- I PLANT SAFETY PROCEDURE NUMBER
C O
R
- IDAMAE, A
4 I
CORE DAMAGiC~E ASSESSMENT I'
IF I
REV:
11l
- ' Figure 1 'Water Density Ratio'(Temperdturd vs.'STP)'
- I
.0O 4-
' a2-V.
I-I 1
I-1.
'20D.
j 0
-1,I o- 0 0a c,
- r A
(
I I
- n 1
- i.
I 0/p STP Page 19 of 39'
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE Figure 2 Power Correction Factor For CS-134 Based on Average Power During Operation
-j
-c LI I-L.)
00 LC-
- VLI C-,
'0~
LI)
- a t9 J
03 0z 0:
0>
- 0) 0; o
Page 20 of 39
PRAiRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE I NUMBER:
l Figure 3 Relationship bf /6 CladDamage With% Core Inventory Reledsed of XE-1 33 Z........
0.1-C ILi U~)
1.1 Iti L:
0 z
w a:
o I.
in Legend AVERAGE-I LOW BURNUP' H'GH 8URNUP
~~~~..
_ k...
- j _..
~~~~~~.......-..
- ~
':2 T
- 3' -
t..
.r 7
I
- r j...
o
.--. 7'.-'
4E..
t r.!-'-
. 4.r 1
/ r- '
ir-;./j.,
, -,*. T-- w
........... H :{.1
' /
2
. j.
- r, y
.. _::z:
1 J,-
' r
. § *,r..,r. r
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)
I r
r 7
- 4. ;
r r, r S,r, f r: r r.
8 j
t,.vV 1Ir- __
, I I
i-i; i
i 0.1 I
. CLAD DAMAGE (7,)'
61
, i 11 Q I.. - I:.
Page.2 of 39 '
PRAIRIE ISLAND NUCI EAR GENERATING PLANT PLANT SAFETY PROCEDURE Figure 4 ReIationship of %o Clad Damage With % Core Inventory Released of 1-131 I
A Ii
.~ ~
Legend AVERAGE LOW BURNUP HIGH BURMuPW C,-)
1-J LJi 0
z Lii 0
C)>
0.1-0.01
/
./............
- .,F x --.
.7
.I..
I t *...
. I...........,
II Flfl01-4,'A
.7
.7.
-/
.1 7..
A
.I 0.
vsArs 1JL)tj1-+--
4--
- I
I.
I
.-- T--rTll 0,1i I0 CLAD DAMAGE (%0) i6o Page 22 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT, PLANT SAFETY PROCEDURE NUMBER:;
I F3-17
____ -REV:
_ 11_ I Figure 5 Relationship of % C'ad Dar.ia'ge With %O Cor'elnivrito¶'Ielefased of 1-131 W/Spiking I.
I i
I I
II I
0 0
LiW Uj
-J LJ c __
a:
0 Z:
Li m
LO w
0 U
1.1 0.1-
.... * -**^-'
Le
.. _Lgend; AVERAGE.
LOW BURNUP _
HIGH BURNUP
..............,:j,,;.
..................-'-i'--
u
.......-: i:-i:i
..,-;b P
/ar
- i
/ ~ ;*@*e*
7.O*sw-wXe
....... i.. j4._.
..... l.w/._
____/,
t...
II
-O n
I I
7:_
n nrs I
R-
^..__.....
.*,{
I0.00014 e
z z
w g
s I
10 CLAD, DAMAGE (%-)
.j O Pagfe' Of W:'_'
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE Figure 6 Relationship of % Clad Damage With % Core Inventory Released of KR-87 n l V -..
-I Legend AVERAGt.
LOW BURNUP HIGH BURNUP 1-1-1 0
0 LLJ U')
Li
-j LA.)
Dtf 114 0
z LO z
Ili w
(D U
0.01-0 oo1~
J i
j
'
/
- '
Z I
/
X,,
1-- -- -
- t
-s ts.
I f
-
/f-.;
1-/:
/
,/
,9/
,...._...e z
/
/:
/:
n
-nan U.V
.1-_
. j
--r--- m ?
- i -r -r-r 17
.4
-1.
I I.
I I-..-.
.....................7.
100 0.0O0vO1 t
- 0...
.......I-----I----
CLAD DAMAGE (7.) U Page 24 of 39 1 1
PRAIRIE ISLAND NUCLEAR GENERATING PLANT I.
PLANT SAFETY PROCEDURE Figure 7 'Relationship of% Clad Damage Wfith %i Core Inventory Relaased of XE-i 31M VI 0
w V) 1.1
-JLa m
w 0
Z LA Z
LO w
0 U
0.1-
................ _.:JC.
t,I AL'
) I 1m I
7:
'v 11 P
Z.
V
I o-
- s A_..,%....-........
A. *.,'.
Is II njini-J-I CLAD DAMAGE (7.)
lb 1
r I
I DO 1-L (-- e - -- C-Page 25 bf 39 ' '
PRAIRIE ISLAND NUCLEAR GENERATING PLANT r,,f r
r
-'t-
_' Io+
't ev f gl-PLANT SAFETY PROCIFDIJPE Figure 8 Relationship of % Clad Damage With % Core Inventory Released of 1-132 0.11 0.01-0 LLJ
-JC LJ LX 0
LIJ Li 0C)
... -- --'1~'
LOW BURNUP HIGH BUR-up
.f
.t T--.
_ z F.
t;e
.s
/.X.,
v.
_ /;
e.
P So
//
/
/
'/
t z
_.1.
wf.
_ s
/vE, /-
__/^
' _ I f
_ w b
_ _ /
tz
.//.
__ s s _
- /,
I'
//
0.0001 I
I O OOO01 t C.
I
-lb 100 CLAD DAMAGE (7-)
Page 26 of 39
PRAIR16SLAND NUCLEAR GENERATING PLANT
. I.
E T PN PLANT SAFETY PROCEDURE - -
t I
F I IIt I I
Figure 9 'Relationship of % Clad Damnage With % Core ln'v'entoir'y R"~ae of1-F,.
r I
I LI 0 0w -
in LI
-J W
ix ir, R
-1 z
I w
z w
m 0'
I
,0.1-I..;....-
'Legend AVERAGE LOW PURNUP HIGH BURNUP
........,t,
~~~.........
-g.*-
....... : ::.:.:..I
.Iewo#eo-s-X a.e s eXe*¢
@@@@*s
'7 94*P; ns.
.I Ul-0.001-_
...:..:..A:
I...
w;.--ri-r>.:
.@/o
- w
,tb 4..z b t, _
s
.7.....
4.....
=-
.~r' 1 i 0.0001~
I
... V.
I 1
..,-.. t.
'_:j_.'..
I v.ttU301,-
0.1 1111 1-16
. I CLAD DAMAGE (7-)
l T
T T
X l I
IO()
I r,
Page' 2"7' 'o;"'
f3b - '
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEPURE Figure 10 Relationship of % Clad Damage With % Core Inventory Released of 1-135 I ;-T
.;I 0,1-, --
Legend
- j AVERA(;E LOW BURNUP HIGH BURNIlP
. I
.1-1 0
0 Li V)
LJ jDM w
PZ LA 2.
LO a,
O Q
...2 I,
.,_I__
- --. II
- -- : "I
-I
-I
,I..
Z 7....
/
,,/:
of
,}_._:
5
- t,
,/:
H::
/...
I
/
-'j D.
,\\n.
nn1--
I-T 7
I 100 0
I l
q E
1a1 0 1
.Ic CLAD DAMAGE (%)
Page 28 Qf 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PRARI ILAD UCEA GNEATNGPLNT-PLANT SAFETYPROCEDURE NUMBER:
I F3-17
_ REV:
iii1 Figure 11' RelIatior sh'ip' of % Fuel Over Temper ature' With~ %;Coe'C 1I6vn-Ventor~y' Released of XE, KR, I, or CS 100 *1
w LI) w
-J ir LO 0
0-10-
-Lege'nd NOMINAL MINIMUM MAXIMUMA..
............ 7........
7 7..
7...
.7..
7.
?....
- r....
tt7 L
1:......
~~~~~~~:......:...'...
L...........
I 0:I -
I 10 FUEL OVERTEMPERATURE (7.)
100 Page 29 of 39-
PRAIRIE ISLAND NUCLEAR GENERATING PLANT Figure 12 Relationship of % Fuel Over Temperature With % Core Inventory Released of BA or SR I.,.
Z..........
..........................-----r--:
--t.
0.1 1-)
0LJ V:
LAJ LLJ 0
m) 0.00 0.000 I
Legend NOMINAL
_1MNIU MiNIMUM MAXIMUM
... I......
................_.;f.,-_
I
,1
/;
,/~~~~.,......:;.
~~~~~~~~~.
.;,/,..,,;..
.. I
............,-f
~~~~~~~~~......,-,--- --- :---r.-
,////
/
.~.
1_/.-:
I.
I-I
....... I......-.........
............................ I.".
I.
't"'
r l
a I
r 10 FUEL OVERTEMPERATURE(%)
100 Page 30 of 39
PRAIRIE-ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
3 CORE DAMAGE ASSESSMENT F3-17 R
=te--s j
X
?'+
.REV:
Figure 13 Relationshia of % Fuel Melt With !0 Core InVenr'tory Released r
, 4vw-0 Lii
-J cr_
0 z
Wii z
0 Q-10-Legend NOMINAL MAXIMUM
.I.......
.................. ----t;:..'-:':'
- .;y.X
~~~~~~~~~............-----
- -*_-p,-,*
-- -v t ; 7.,
- : /,. /
- j~ if./
. /-/* *7 1,'
-7..... ~.
.S;.i.,/.... /...-._
g<--r,.1
,/
~.s 7/Yr:
- 1 t
[
- !t:,
/j/.lG'itt.-
~.........
5 '
T
.\\..
-s;^-.i I
zt i
tiI
.4.
I.
t00 In
-0.
1 I
T 1
FUEL MELT;(-) -
Pabb316f3
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE Figure 14 Relationship of % Fuel Melt With O/% Core Inventory Released of BA or SR 1 11 1_
g ti,...................
z.......
egend ~..........................
..egend..
NOMINAL I
10-
I Lii U) tLii 0Ir-:Z 0
C)
_._._S
...... tMINIMUM.....
._.L.
MAXIMUM
---f - - - ~-
I---
2 A r_,#ot-.
w
-A --.
- 1. I 0.01 -
r
-q I
I I
10 rFEL MELT(%)
100 Page 32 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE Figure 15'Relationship of % Flel Melt With % Core Inventory Released of PR 4^^
1 z } [ ]
- r
.1 I
~~............
~.
.i.
- -*-rr~r6 lo:
- ' -. NOMINAL--
... -MINIMUM-.
MAXIMUM 1-0 C-I LJ V)
LLJ I
LLJcr m0 Z
Ljj m
LJ m
0 U
I I
II I I
II I 0.
II I
1.
.....................,-.,,isl
'.. s...,:
~~~..........
..,*t'.'
I 1.
r
, I_
~~~~~~..............., : :,
..... _ z:.
O.nl-eUS-I0.
^ft1
.uu-l-
.1
.I 10 TUEL MELT N 100 I -1 I:. _
Pagd 93 -of 35'
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE Vie 7-7;,,'
CORE DAMAGE ASSESSMENT F3___17 fi^$-
n rll REV:
1 Figure 16 Containment Hydrogen Concentration Based on Zirconium Water Reaction 7..............
2 -
- =
1-.
Z:.......
....../
0 Z.
I/.....
-/
S -.
/
~~~~~~~~~~~~~~~~.........
Z.......
Go 0
o 20 40 S
80s 100 ZiRCONILJM WATER REACTION (0 Page 34 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT Figure 17 Percent Noble Gases in Containinent'"
1000000. *..
.s.....
...... 3 ;................
..............; ii:
Z.........
.eleae..
1000.........
- -a@
-b......................
10000D0 S2Z Noble
~~~~~~..
OEML Gas Release
_s................... !-
.0 0CORE MDLT:
32Z No REGIME cr
-I' L~i 0.
CLJ
........:.:..,t..,, :..
'.'i1-:1.s.........
... :..;..:,::::.::::.....':..s
'.FUEL PELLET OVERTEMPERATUiE jREGION
.~..
10 0 0.-:.l}Z'
>t'.
1............1...
soCA A UEl@t*{.
l;..........
4*r.......
l00t I.
Z.:.:..:'.::..
- @--*-~;-;t4t-@-----*---;---. v*@a.e*......
>+.
r4.......
- 0. %
l Ga.s 0 _e Noblease
.CA.
F.AU R.
.. i...j.
Rele se REGION 10
. _},,
,i s,,
>o@---w-;-t-v4 j.;-ii4 i -*s---;----i--;
nev
-1~ *1
_f ANS 18.
Normal Operating, Noble Gas Rele~A&A
- i.
0..
0....NORMAL RCS
.4ACTIVITY REGION 0.0.....1...
I
-10 100
-TIME AFTER ACCIDENT- (HOURS) 1000 Page 35'b 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE Figure 18 RCS Dose Rate vs. RCS Activity Concentrations 1 Hour After Shutdown 1.OE+05 -
4i I I I
-I 4 I I I I DA: WOG l100%
DBA 100%
dnG RELEASE 1.E+04
- in Fuel Gap i_1.0E 00 -
& 1.0E+02 E0 E0
.E0
.E0
.E R1
.OE+0C --
L1 ml 0
_C ol =13_c 1.OE+00 1.OE+02 1.OE-+03 I.OE+04 I.OE+05 t.OE+06 tOEF RCS CONCENTRATION (Total uiCilmll)
Paae 36 of 39
PRA'RI, ISLAND NUCLEAR GENERATING PLANT I
I.
I
-
PLANT SAFETY PROCEDURE NUMBER:
I C.. I
..SE CORE DAMAGE ASSESSMENT, F3-17 11 l:
1
__-REV:
... ' Thermally Induced Current Errors in'Containment Radiation Monitors
- 1.
R-48/R-49 & R-70/R.71 Thermally Induce Erroi~s:..,,;,.:
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 /> tafter a DBA LOCA.' Industry testing of high rangebradiationrm6nitor (HRRM)'sjstbms has revealed that signal errors or the, loss of signal are the result of therrmally iridu6ed current (TIC) andlor moisture intrusion into the coaxial connectors. Based on the.EPRlPIant Support Engineering studyw,'worst case estirriated errors are summarized below:'
.,!i8 g
t)
Time After Postulated:
Estimated Err-6rs in,, i.
, dA Readings--,
-;1 minute,
> 3000'R/hr -,'
-2 minutes 100 R/hr -----
- f-t iL'
.8 rriinutes 15 Rhr i!,;i 2to 4 ho'urs-i, 9 R/hr-iiiL.
> 4-hours
!No-Effect-frorT m-
-i--t-i-r 1
___,__1l More' backgro'und information concerning thernmlI) Indlued current in' high range radiation monitors is described in S6ction 1I1.
i
!I Pleasenriote that errors in the range 'of *10 Rlhr'oneth-ioyr after~a postulatepBA has minimal effbct 6n ouir ass6s'smeritof fission product release-to-'on6tainmenUt-vhen we are considering magnhiudes of.> 100 R/hr-r6acding'jo be confirniati rifoffissioh product' releasedlo containment.
V t
Backgro~und on Thermally I nduc6ed Current, (TC) onib g
RMonitors Backqround i
.I t
Excerpts frorm:.PINGP Rqsponse to Hig h_'RangRe'-daiatioh Moiitor Cable Study: Phase lI, R'eport No: TR-1 12582 Novemrber-2000.- '. ',
1' Transient signal errors'have been obsdrved in; industry testing of the high range radiation monitor.(HRRM) system, At PINGP, these are'plant radiation nonitors RE-48 and RE-49.
Theiinvestigation into this issue re~6aled fhat signalrerrors or the loss of signal are the result of thermally Induced currents (T!Cs) and/or moisture intrusion into the coaxial connectors. Information Notices,,lN 97-45 and IN 97-45 Supplement 1, were issu6d by the NRC to alert licensees to these potential issues.
2.
3 -- -
Page ~37 of 839'
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
CORE DAMAGE ASSESSMENT F3-17 REV:
1 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 I'". This study was focused on the thermally induced current phenomena since moisture inti usi)n issues are well understood within the industry and have more generic applications. Phase I of the EPRI study confirmed that TIC existed and was significant under thermal transients. Phase II oi the study identified the sources of the TIC and developed a mathematical model for cable responses to thermal transients.
Studs Results and Anatlysis 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 aporoximate expected signal error for the HRRMs during the postulated DBA. The expected radiation readings due to the TIC phenomena. based upon the worst case cable length, are asjfollows:
50 seconds 3372 R/hr 1 n lsecords 38 R/hr 500 seconds 13.2 Rihr 8000-15000 seconds
,8.3 R/hi
>15000 seconds no effect from TIC 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 GENERATING PLANT
!--.4-iA4 A
S A 4C s
PLANTSAFETY PROCEDURE COR D
A S
, I S
CORE DAMAGE ASSESSMENT' NUMBER:
IF3.
-17
-11 I
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- __ _. __ REV: Thermally Induced Currenrt Errors' in'Containment Radiation'Monitors PINGP Response to HRRM Sidnal Error During the initial phase of any postulated accident, it would not-expected to s6e.:i indication of actual fueldamage for the first,10-15.minutes.-If indeed the alarms would come in for RE-48 and RE 49, Operations would be'odcupied with accident mitigation and monitoring tasks-during this time period and this alarm;;even though acknowledged, would be ignored during this period. Other-p'arameters-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 level reddings, -
even if the alarms have come in, would be decreasing. Again, this is validationof an j44t 5 s1 t]
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o F6r'emergency' plan response and possibI6 SAMG cbnsid65aftiohs tfie TIphenbr'enon would no longer beaffectingthe'radiation monitorsA d,
o the erlier alrn and decreasing readings that were noted,'iit"'wuldl'e~confirmed th'at no fuel damage had occurred and these were indeed erroneous readings&AWbneral site emergency-alarm would be activated at 1000 R/hr, but as cited previoUd1this 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 requirbd'within thetemergency plahnI procedures would not occur until after the TIC phenomena has either dp has.
been verified to be erroneous.
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT EEGNYPA MLMNIGPOEUE EMERGENCY PLAN IMPLEMENTING PROCEDURES PRAIRIE ISLAND RADIATION PROTECTION GROUP RESPONSE TO A MONTICELLO EMERGENCY NUMBER:
F3-22 REV:
17
- Procedure segments may be performed from memory.
- Use the procedure to verify segments are complete.
- Mark off steps within segment before continuing.
- Procedure should be available at the work location.
O.C. REVIEW DATE:
OWNER:
EFFECTIVE DATE M. Werner Page 1 of 8
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES J;
NUMBER:
PRAiRIE ISLAND RADIATION
- Xtq, PROTECTION GIROLUP RESPONSE F_-22 TO A MONTICELLO EMERGENCY REv:
17 1.0 PURPOSE When an Alert, Site Area, or General Emergency occurs at Monticello, the Prairie Island Radiation Protection Group SHALL be requested to respond with personnel and equipment to support the Monticello Radiation Protection Group. This Prairie Island support allows the Monticello personnel to concentrate their efforts in performing onsite sampling and monitoring and relieves them of the offsite monitoring requirements.
The purpose of this instrjction is to describe the personnel, equipment, and procedures required to respond to a Monticello ermergency.
2.0 APPLICABILITY This instruction SHALL apply to Shift Supervisors (SS), Shift Emergency Communicators (SEC), MNatager of Radiation Proteciion, and all Radiation Protection Group and Cffsite SUrVCy Tt::an, mr'enmbers.
3.0 PRECAUTIONS All meters should be source checked prior to departing the Prairie !slarld plant s.te.
4.0 RESPONSIBILITIES 4.1 The Shift Supervisor has the responsibility to ensure the Prairie Island Radiation Protection Group is notified of a Monticello request for assistance per this procedure.
4.2 The Shift Emergency Communicator is responsible to assist the Shift Supervisor as directed by the Shift Supervisor.
4.3 The Manager of Radiation Protection or designee is responsible to ensure Prairie Island provides adequate radiation protection staff in support of a Monticello emergency per this procedure.
4.4 Radiation Protection Group members are responsible to respond to a Monticello emergency per this procedure.
4.5 Offsite Survey Team members are responsible to respond to a Monticello emergency per this procedure.
Page 2,of-8
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES I-
-PRAIRIE ISLAND RADIATION j
NUMBER.
PR0TECTIOI d GROUP AESPONSE
.F3-22 TO A MONTICELLO EMERGENCY REV:
17 5.0 DISCUSSION When an Alert, Site Area, or General Ernergency 6cclrs at Monticello, thePrairie Island Radiation Protection Grop wl'lbe requeste'd toj(e'sjo with personnel a'nd equip'ment'to the Monticello EOF and i
if'ipp.e t'Moh ellAra'Pulic' Reception Cernter.
At least two (2) Offsite Survey Team members are required to respond to the Monticello EOF c6'sfaff twv'ofield survey tearms'and o't&(1')f adjation'Protectfori Specialist (RPS) to provide assistancet6o the'badiatib'irotectIon Support Spervisor (RPSS) in the EOF. If available, two (2) additional Offsite Survey Team members should respond to EOF to facilitate staffing two (2) Offsite Survey Team'merhbbrs per.
field survey team.
In addition, another team,.consisting of-twb(2) RPSsantd.n'&3REG,(if available)'
)
should respond to the Monticello Area Publid ReceptiarnTCenterif it is abtivated.'
See Appendix A for a listing of required Monticello response equipment',
6.0 PREREQUISITES c-c 1
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C i The Shift Supervisor has received a request for radiation protection assistance in response to a Monticello emergency.
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDUPES 7.0 1-ha
-- 1>5J-Al -
PRARIE ISLAND RADIATION NUMBER:
F3 PROTECTION GROUP RESPONSE F3-22 TO A MONTICELLO EMERGENCY REV.
17 PROCEDURE 7.1 Initial Notifications Performed by the Shift Manager (SM) or Shift Supervisor (SS) he SM or SS may request the SEC to assist in notifying the Radiation Protection Group.
7.1.1 Using the RADIATION PROTECTION CALL LIST in the NUCLEAR EMERGENCY'C; PREPAREDNESS TELEPHONE DiRECTORY under Pi ERO, notify the Manager of Radiation P'rotection or designee and brief him on what is known of the following:
A.
EcneIgency Cfassification.
B.
1: there a radioactiie release.
C.
Has there been ofsite protective actions issued.
D.
Call Security Shift Lieutenant and notify him/her of Radiation Protection Personnel emergency call out and need for FFD screening.
7.1.2 WHEN the Offsite Survey Team notifies the Control Room of the expected departure time, THEN the SM or SS should notify the Monticello SEC (763) 295-3739 x1 216 or x1 072 and inform them of the Pi Radiation Protection Group's departure time.
Page 4 of 8
PRAIRIE ISLAND NUCLEAR GENERATING PLANT
-EMERGENCY PLAN IMPLEMENTING PROCEDURES PRAIRIE ISLAND RADIATION NUMBER 9
PROTECT:ON GROUP RESPONSE F3-22 TO A MONTICELLOEMERGENCY REV:
17 7.2 Response Team Activation Performed by the Manager of Radiation Protection or Designee 14 7.2.1 Using RADIATION PROTECTION CALL LIST in the NUCLEAR EMERGENCY PREPAREDNESS TELEPHONE DIRECTORY under PI ERO, or other list of qualified personnel, mobilize two (2) response teams to the Mortic6li EOF mad tip 6f at'least one (1) Offsite Field Team member perfield surveytam (two(2)'if available) and one (1) additional RPS to assist in the EOF.
. -7.2.2 During-off-normal viorkng'hodrs, s'crednR'Wl prsonnel for Fitness for
- Duty FFD).-
5 7.2.3 IF offsite protective action recommendations have been issued OR the classification is a Site Area: orIGenerabEmergency, THEN direct the dispatch of another response team to the Monticello Area Public Reception Center..The-teamrsbluldfbelmade up of two (2) RPSs and one (1) REC (if available). See Appendix B for directions to the i MonticelloAreaPublicEReceptiorn-ant-r.
n
-i Long term 6overage for Offsite'Sl' bems should be N
IrE "prvided by contract fie'alth 'Physics ersonn'el.
% Arrangements for contract services 11iblbe handled by the Monticello EOF Organization.
7.2.4 Augment additional Oersofnel as 'soon aslposbie to supply short term 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />sY pe day c r
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Page'-8'of 8.:"-'
PRAIR!E ISLAND NUCLEAr GENERNq.NG PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES PRAIRIE ISLAND RADIATION NUMBER:
PROTECTION GROUP RESPONSE F13-22 TO A MONTICELLO EMERGENCY REV:
17 7.3 Monticello EOF Response Team Activation 7.3.1 Personnel pick lip vehicles and equipment as listed on Appendix A.
7.3.2 Perform meter check on all meters prior to departing the plant.
7.3.3 Notify the Shift Supervisor of the expected departure time prior to departing the plant.
7.3.4 Notify the SS of any members that may need FFD screening and have those personnel report to Security Building (Guardhouse).
INOT: C g lThe Minnesota road map and Monticello survey map are
.. MS 1located in the field team survey kits.
7.3.5 Review the Minnesota road map and Monticello survey map for reference to routes to the Monticello EOF.
7.3.6 Depart for the Monticello EOF.
7.3.7 WHEN approaching the boundary line of the Monticello 10 mile EPZ, THEN attempt to contact the Monticello EOF using the portable radio.
Identify yourself as the Prairie Island Survey teams.
7.3.8 IF determined from the initial radio contact with the Monticello EOF that the plume may be encountered while enroute, THEN conduct a search for the plume, in accordance with F3-15 and proceed directly to the EOF.
7.3.9 Upon arrival at the Monticello EOF, contact the Radiation Protection Support Supervisor or EOF Coordinator at the Nearsite EOF for obtaining field team drivers and receiving further instructions.
7.3.10 Perform the requested offsite surveys as required by the Emergency Manager.
1N -7TE:
X All field sample analysis should be performed in the EOF
!.^ ICount Room by Monticello Radiation Protection Specialists.
7.3.11 Continue with all assigned duties until relieved by the Monticello Emergency Manager.
Page 6 of 3
PRAIRIE'ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES I
PRAIRIE ISLAND RADIATION NUMBER:
i PROTECFION GROUP 'RESPONSE F3-22 r
{
TO A MONTICELLO EMERGENCY REV:
17 Appendix A Monticello Emergency Support EQUIPMENT LIST. -.
- 1. Personnel TLD's and self-reading dosimeters (rezero)
- 2. Two (2) Vehilei t6 be used for-offsite monitoring purpioses.
- 3. Two (2) emergency survey team kits
- 4. Two (2) installed radios or Ooitab'f6 radios wifh'rrag-hioiihi' 'ntennas.
- 5. Two (2) dose rate meters
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- 6. Two (2) count rate meters -
- 7. Two (2) DC Air Samplers d
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES PRAIRIE ISLAND RADIATION NUMBER:
8' PROTECTION GROUP RESPONSE F3-22 TO A MONTICELLO EMERGENCY REV:
17 Appendix B Monticello Area Public Reception Center Support Directions to Osseo Jr. High School Option 1:
Proceed from plant to northwest side of Minneapolis via 1-94 or 1-694 to Highway 169 (Exit 31).
Exit onto Highway 169 (North).
Continue North on Highway 169 to 93rd Ave. N. (7th St. or Co. Rd. 30).
Turn left onto 93rd Ave. N. (7th St. or Co. Rd. 30).
Continue Westson 93rd Ave. No. (7th St. or Co. Rd. 30) to Osseo Jr. High School (10223 93rd Ave. N.).
Option 2:
If Proceed from plant to northwest side of Minneapolis via 1-494 to 1-94.
Continue West on 1-94 to Co. Rd. 109 (Exit 215).
Exit onto Co. Rd. 109 (East).
Continue East on Co. Rd. 109 to Highway 169.
Exit Co. Rd. 109 onto Highway 169 (North).
Continue North on Highway 169 to 93rd Ave. N. (7th St. or Co. Rd. 30).
Turn left onto 93rd Ave. N. (7th St. or Co. Rd. 30).
Continue West on 93rd Ave. No. (7th St. or Co. Rd. 30) to Osseo Jr. High School (10223 93rd Ave. N.).
Page 8 of 8
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLANIMPLEMENTING PROCEDURE C
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-Procedure segments may be performed from memory.
Use the procedure to verify segments are complete.
Mark off steps within segment before continuing.
Procedure should be available at the work location.
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O.C. REVIEW DATE:
OWNER:
SC-EFFECTIVE DATE M. Werner 4-/-
O l
Pa'ge i of 7
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PI-AN IMP!LEMENTING PROCEDURE 1.0 PURPOSE The purpose of this procedure is to provide instructions to the Radiation Protection Group on the use of the Hotcell, to include Hotcell setup, various chemical analysis evolutions and radioactive sample disposal techniques.
2.0 APPLICABILITY This Instruction is applicable to Chemistry Radiation Protection Specialists.
3.0 PRECAUTIONS 3.1 Monitor the general area of the Hotcell for direct radiation to ensure the habitability of the Hotcell.
3.2 The reactor coolant samples taken in an accident condition have the potential to be highly radioactive. this may give rise to dose rates far in excess of what would normally be encountered. All work involving these samples is to be performed in the Hotcell with the fume hood in operation and with remote handling tools, to minimize radiation exposure, until one of the following is deiermined:
3.2.1 The sample is determined not to have close rates in excess of normal values.
3.2.2 The sample has been diluted to the point where the diluted portion does not have dose rates in excess of normal values.
3.3 If a sample is determined to be of normal dose rate values, or is diluted to the point NOT to exceed normal dose rate values, the following should apply:
3.3.1 The instructions specified in this procedure may be completed in an area other than the Hotcell Hood.
3.3.2 Monitor the alteirnate area for direct radiation to ensure habitability.
3.3.3 Analyze the sample in accordance with the appropriate RPIP, as a normal chemistry sample for the analyte of interest 3.3.4 The instructions for Post Accident Sample Waste Storage and Disposal apply.
Page 2 of 7
PRAJR1i ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE 4.0 RESPONSIBILITIES
~ I 1:,.
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The Cheniistry Radiation-Protection Specialists are responsible to implementritlhis procedure.
a 5.0 DISCUSSION The Hot Chem Lab in tne Auxiliary Building "may ot'lbeavailable due to abn6ormal radiological conditions. Use of the Hotcell or Alternate Area would be necessary.
6.0 PREREQUISITES 6.1 Hotcell Set-up Procedure or Alternate Area I '
<& (The following procedure s1zpujd be pc dpleied priorto-.
l 5V~I pieipt~h lete Area r
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' ntr6ducing a hot s'arnO It
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6.1;. 1'- Ensure that all instrum'erntatfdn.is'tur'ned'tn, wa'rrnie'd'up and calibrated.
6.1.2 Fill a 1' L volumetricA6 the m'ark %With dnrhiinbialized water.
6.1.3 Fill a 100 ml volumetric to the mark with demineralized water.
6.1.4 Remove 1 ml of demineralized watdr'fro m each volumetric using a 1 ml pipet.
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6.1.5t Add a stir bar to'eachI volumetric.- 0" i j i 6.1.6 Turn ON thetwo:stir platesin the fume hood IF containment spray has been activated, consider buffering Cl-pH meter with 7 and 10 buffer.
6.1.7 Buffer the pH electrode.
6.1.8 Place a 250 ml beaker of water near the pH probe.
1Pagb 3 6f 7
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PR7CEDURE 7.0 Procedure 7.1 Sample Preparation The RPS Sample Team members SHOULD ensure all Xsamples are properly labeled with sample identification, t >-ZW sample sizelvolume, flowrates, pressures, and sample times, as appropriate to facilitate accurate analysis. As samples Big are diluted, split, or reduced; the appropriate information GINOTE T
needs to be included on nev' !abel3 attached to the newly created samples. Sample dose rate information should be
^
included on all sample labels, to help ensure personnel awareness of radiological consideration. For ALARA reasons, the sample containers should be prelabeied whenever possible.
7.1.1 Label all s~ssnples.
7.1.2 Verify postings and boundaries for expected radiation and contamination levels 7.1.3 Don a finger ring on each hand.
7.1.4 Ensure TLD and dosimeters are worn.
7.1.5 Place the 60 ml bottle shielded 3arrier in tMe fume hood near the pH probe.
AVOID PLACING HANDS OVER TOP OF OPEN SHIELDED C ARRIER.
7.1.6 IF radiation levels require, THEN use the remote handling tool.
7.1.7 Remove the lid from the 60 ml bottle shielded carrier.
7.1.8 Remove the stopper from the bottle.
7.1.9 Pipet 1 ml of coolant from the 60 ml bottle to the 1 L volumetric.
7.1.10 Cap the volumetric and agitate to mix.
7.1.11 Pipet 1 ml of coolant from the 60 ml bottle to the 100 ml volumetric.
Page 4 of 7
PRAIRISISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE The 100 ml volumetric is to be saved for the Chloride Analysis, which is to be completed within four days. The undiluted sample must also be saved for 30 days.
7.1.12
-Cap the volumetricb nd agitate to mix..
7.1.13: Label the volumetric with sample, date,time, and the number of mis of
- sample in.the volumetric.
7.1.14 Mark 's'aiple 'TO BE SAVEDW 7.1.15 Store the 100 ml v6lume tric irntheHotcel Shielded Area.
7.1.16 IF a pH Analysis is to be determined on the sample, THEN proceed to Step 7.2. IF NOT, THEN replace-tbeslopperion the 60 ml bottle.
., 7.1.17,. Replace the lead cover on 4he.slieldedcarrier, place the shielded carrier in the Hotoell Shielded Area and proceed to Step 7.3, Gamma Analysis Preparation.
7.2 pH Analysis - Using the Combination Methods
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The pH meter gives a digital readout of sample temperature and wilI auto-compensate for temperature.-
7.2.1-Insert the combination pH probe and temp probe into the 60 ml bottle and read pH and tep'erature of coolant.
7.2.2 Remove both probes and place in a beaker of demin water.
.- F.
,i 7.2.3 Log sample results on PINGP 655, Post Accident Chemical Analysis
,,Report.,,
7.2.4 IF radiation levels require, THEN use remote handling tools for handling the 60 ml bottle stopper and'shielded carrier Lid.
7.2.5 Replace the stoppe~ron the 60 ml bottle and thie lid on the 60 ml bottle shielded carrier.
7.2.6 Remove the shielded carrier and the beaker of rinse waterfrom the fume
'hood and store according toStiep 7.6, Post Accident Sample Waste Storage and Disposal.
P -'g 6 5o0f 7
PRDIPIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE 7.3 Gamma Analysis Preparation 7.3.1 Pipet 10 ml of diluted coolant sample from the 1 L volumetric to a 10 ml via.F Sample should be diluted to give a contact reading of under
-NO),TE~
- l1 miliirem/hr contact. The diluted sample should NOT exceed 25 rnill;rem/hr contact.
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.
7.3.3 Label the vial with the sample point, date, time, and dilution factor to the sample prior to sending to EOF Countroom.
7.3.4 Piace the 10 ml vial in -he shielded carrier for transport to the EOF Countroom.
7.3.5 WHEN radioactive gas, charcoal, or particulate samples are received, THEN ensure aii samples are labeled with date and tirne of sample, sample point, sample volume and/or correction factor, and flow rate.
7.3.6 Store all sainples in the Hoicell Shielded Area until transported to the EOF Countroom.
7.4 Boron Analysis 7.4.1 Using the 1 L sample prepared in Step 7.1, Sample Preparation, analyze in accordance ',,ith the appropriate Boron Analysis procedure.
7.4.2 Log the results on PINGP 655, Post Accident Chemical Analysis Report.
7.4.3 Dispose ot a.l radioactive waste according to Step 7.6, Post Accident Sample Waste Storage and Disposal.
Page 6 of 7
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE 7.5 Chloride Analysis Chloride analysis SHALL be completed within 4 days of accident.
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THE REACTOR COOLANT SAMPLES TAKEN IN AN ACCIDENT-CONDITION HAVE THE-POTENTIAL TO BE HIGHLY RADIOACTIVE. THIS MAY GIVE RISE TO DOSE G UTO RATES FAR IN EXCESS OF WHATWOULD NORMALLY BE; lENCOUNTERED. ;THE ION EXCHANGE COLUMNS ON THE
'ION CHROMATOGRAPH COULD HAVE CONkTACT READINGS OF UP TO 10 RIHR.
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- . S 7.5.1 Using the 100 ml sample prepared in Step 7.1, Sample Preparation analyze for, anions jAW the appropriatejanalysis procedure.
7.5.2 Log the results on PINGP 655, Post Accident Chemical Analysis Report.
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7.5.3 Dispose of all radioactive waste -according to Step 7.6, Post Adcident
-Sample Waste Storage and Disposal..P cv-e
-7.61-Post A-cident Sample Waste Storage and, Disposal-*
Ensure samples are labeled. "TO BE SAVED" or "TO BE PED before storage In shielded area.
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,7.6.1,, Place all capped or covered radioactive sample waste in the Hotcell Shielded Area.
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-1 1 I 7.6.2 IF additional waste samples are added to the Hotcell Shielded Area, THEN L
survey the Hotcell general area radiation levels. Add additional shielding, as necessary.
-I 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