ML20059M967
| ML20059M967 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 10/27/1993 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| References | |
| LZP-1130-2, NUDOCS 9311300026 | |
| Download: ML20059M967 (22) | |
Text
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,- LAP-820-3 . e, Revision 23-August 2, 1991 9
ATTACHMENT C OFF SITE PROCEDURE NOTIFICATION FORM DATE: Wednesday, November 10, 1993 DISTRIBUTE TO: US NRC/ WASHINGTON,0.C.(3 HOLE) Plsase REMOVE the following pages from your controlled copy of the LaSalle
-County Station Procedures Manual. INSERT the new pages as indicated and REM 0'd and DESTROY the superseded pages. $1CN this transmittal form in the space provided for Manual holder below. RETURN this signed sheet to:
Central File Supervisor LaSalle County Station Station Manager ;
.LaSalle County Station t
SET NUMBER: 97 - MANUAL HOLDER SIGNATURE DATE , I (If NEW HOLDER, PLEASE ADVISE.) REMOVE INSERT DOCUMENT REV # REV # DATE INDEX OLD NEW LZP-1130 2 0 1 10/27/93 i i I w 9311300026 931027:-- I PDR ADOCK 05000373 . F _ PDik j [
\
]
;, 11/10/93' LZP Ilipjpg :Ptg3.2' .e PROC. NO. TITLE REV. DATE I LZP-530-2 DELETED 02 10/80 ,
V LZP-540-1 DELETED 01 10/80 , LZP-700-2 DELETED 01 11/80 LZP-710-1 DELETED 02 10/80 LZP-810-1 DELETED 01 10/80 LZP-820-1 DELETED 01 10/80 LZP-830-1 DELETED 01 10/80 LZP-850-1 DELETED 01 10/80 LZP-860-1 DELETED 01 10/80 LZP-880-1 DELETED 01 11/80 LZP-1110-1 STATION DIRECTOR (ACTING STATION DIRECTOR) 17 3/93 IMPLEMENTING PROCEDURE LZP-1110-2 ASSISTANT STATION DIRECTOR IMPLEMENTING 04 2/93 PROCEDURE [ LZP-1120-1 OPERATIONS DIRECTOR IMPLEMENTING PROCEDURE 07 2/91 .\ /. LZP-1120-2 OPERATIONAL SUPPORT CENTER DIRECTOR IMPLEMENTING 05 9/93 PROCEDURE LZP-1120-3 OPERATIONAL SUPPORT CENTER SUPERVISOR 03 9/93-IMPLEMENTING PROCEDURE LZP-1130-1 TECHNICAL DIRECTOR IMPLEMENTING PROCEDURE 10 2/93 LZP-1130-2 CORE DAMAGE ASSESSMENT 01 10/93 LZP-1135-1 COMMUNICATOR IMPLEMENTING PROCEDURE 02 7/93 LZP-1135-2 ELECTRONIC STATUS BOARD RECORDER IMPLEMENTING 00 6/93 PROCEDURE LZP-1140-1 MAINTENANCE DIRECTOR IMPLEMENTING PROCEDURE 04 2/91
.LZP-1150-1 STORES DIRECTOR IMPLEMENTING PROCEDURE 04 2/91.
LZP-11tiO-1 ADMINISTRATIVE DIRECTOR IMPLEMENTING PROCEDURE 09 2/93 LZP-1170-1 SECURITY DIRECTO'R IMPLEMENTING PROCEDURE 12 2/93 LZP-1170-2 ASSEMBLY'AND ACCOUNTABILITY OF PERSONNEL 03 9/92 () LZP-1180-1 RADIATION PROTECTION DIRECTOR IMPLEMENTING 09 6/93 PROCEDURE LZP-1180-2 DELETED 05 3/86
.ZWINDEX
LzP-1130-2
,i Ravision'1.
October 27,'1993 1 COR'c DAMAGE ASSESSMENT , i A. EEEECSE The purpose of this procedure is to determine the extent of core damage under accident conditions. B. REFERENCES P - 1. Procedures for the Determination of the Extent of Core Deraage Under Accident Conditions, C.C. Lin, General Electric NEDO-22215. -
- 2. BWR Owner's Group document,'" Integration of Other Plant Parameters Into Core Damage Estimate", Attachment 2 of Dente to Eisenhut transmittal dated 2 Tune 17, 1983. r
- 3. Westinghouse Owner's Group Post-Accident Core Damage Assessment Methodology, February, 1984.
- 4. LaSalle Annex to the GSEP Manual.
C. PREREQUISITES i ( 1. Post accident high radiation sampling system (HRSS) or other suitable means of obtaining the necessary samples. D. PRECAUTIONS i
- 1. None.
E. LIMITATIQHS AND ACTIONS
- 1. This procedure is to be implemented by the Technical Director with the assistance of the Chemistry Director w',en core damage is indicated.
- 2. The calculations in this procedure are based on a reference plant _
(BWR-6/238, MARK III) and extrapolated for LaSa11e's. BWR-5, MARK.II.
- 3. Measurements of Cs-137 and Kr-85 activities may be diffi 7 nit to !
measure until most of the shorter-lived isotopes have decayed.~
- 4. It is recommended that both the water and gas phase' samples be measured in order to reduce the uncertainty in core damage estimations.
'N 4
zWLze i s k N -
LZP-ll30-2 Revision 1 October 27, 1993 2 7 ., ! ) el
- 5. If water smnple results show unusually high concentrations of some less volatile isotopes such as Sr-92, Ba-140, La-140,-Ru-103 and metastable isotopes of Tellurium, some degree of fuel oxidation or melting may be inferred.
- 6. The fission product inventories in the core are calculated based on three years (1095 days) of continuous operntion at 3651 MW, or 102%
of rated power for the reference plant. Inese parameters were used to formulate Attachments A through D.
- 7. If the concentration of a fission product in reactor water or drywell, corrected for decay to ti.s P Lne of reactor shutdown, is measured to be higher than the baseline concentration shown in the lower right hand corner of Attachments A throagh D then the extent of fuel or cladding damage can be determined f rom the curves in Attachments A through D based on I-131, Cs-137, Xe-133, and Kr-85.
F. EEQCEDUEE
- 1. Core damage assessment can be done by one of the following steps:
- a. Steps 2 thru 8 using chemistry sample results,
- b. Step 9 using uncorrected containment radiation monitor readings.
- 2. From the chemistry sample results, pick out the results for the desired fission product. Identify the fission product as Cy g - for a fission product from a liquid sample Cgi - for a fission product from a gas sample NOTE In case the fission product concentrations are measured separately for th_e reactor water and suppression pool water or the drywell atmosphere and the suppression pool atmosphere, the measured concentrations Cgy or CGI would be averaged from the separate measurements:
C,1 = (Conc in Rx water)(Rx water mass)+(cone in SP)(SP water masil Rx water mass + SP water mass Cwi = LCRAc in Rx watgr)(2.92E8 a)+(cong_ ira SP)(3.6019 4 2.92E8 g + 3.60E9 g Cgi = icgng_in DW){DW cas vol)+(conc in SP atmosphere)(SP atmosphere ons vnll DW gas voJuma + SP gas volume 9 ZWLZP
r-L. LZP-1130-2 Revision 1 October 27, 1993 3
. ,-si Cg g = feone in DW)f6.79E9 cc)+fconc in SP)(4.36E9 cc) i ) - %,,,/
(1.12E10 cc) HQII i ti and t must be in the same units.of time.
- 3. Decay correct sample results to the time of reactor shutdown. ,
e AID where A1 = h.93 ti ti = half '.ife of fission product i t = time from reactor shutdown to sample time
- 4. Correct the measured' gaseous activity concentrations for temperature C and pressure differences ir the sample vial and the containment .F (drywell/ suppression pool a tmosphere) gas phase. l t
NOTE The following correction for the measured concentration is needed if-j3 the temperature and pressure in the sample vial (T 1 , P 1 ) are different from that in the containment (T 2 , P 2 )* () T Cgt = Cgi(vial) x 21 PT y2
- 5. Calculate the inventory correction factor (F71) (See Attachment M for ex ample ) ,
Fy g = Inventory in reference plant Inventory in operating plant- , 3651 (1-e-1095Ag ) B E -A T- -A T 0 t j P3 (1-e. I i) e I$ . Where P$ = steady reactor power operatica-in period J (MWt) l i A3 = decay constant of isotope i T3 = duration of operating period j_(days)' il y
/ g h b
\~ 4 ,
4 ZWLZP l 1
, .LZP-1130-2
, Rsvision 1 October 27, 1993 4
Q:
-Q
-0 J T3 = time between the and of operating period j and time of the last reactor sautdown (days) ,
NOTE In each period, the variation of steady power should be limited to-120%.
- 6. The plant parameter correction factors (Fw, Fg) are Fw = Operating _ plant....cnolant mass = 1.663 Reference plant coolant mass 1
Fg = Opar.n.tlng plant containm_9 Mas volume = 0.278 Reference plant containment gas volume
- 7. Calculate the reference plant equivalent concentration Cwl = Cwi (Step F.2)xelit (Step F.3)x Fgg(Step F.5)xFw(Step F.6) l Cgi = Cgl'(Steps F.2 and 4)xe lit .(Step F.3)x F yg(Step F.5)xFg(St'ep -l-
{ F.6)
'u\ l
- 8. Use Attachments A through L to estimate the extent of fuel or .
cladding damage. Estimate the percentage of fuel inventory airborne in the containment
~
9. from the containment radiation monitor in the following manners llDIE The calculation of core damage assessment using the containment. radiation monitor readinge is only valid for the first ten hours following reactor shutdown.
- a. OBTAIN the uncorrected containment. radiation monitor _ reading (R/Hr). Plot this reading on Attachment'E to'obtain the percent of fuel clad damage or Attachment F to obtain the' percent of ,
core damage. - ZWLZP i
~. - _ - - - _ _
>f LZP-1130 Ravision 1-0ctober 27, 1993-5 p- .
V'
- 10. Obtain information partaining to reactor water level history (ie, degree of care uncovery and duration). If cora uncovery was prolonged, it is likely that some degree of clad oxidation, fuel pellet overheat, or fuel melt has taken place. Note that if ,
increased hydrogen and reactor coolans. containment air radionuclide concentrations have acconpanied na_Lhange in reactor water level, then damage is more likely to be restricted to a few fuel assemblies-due to a localized coolr.at flow blockage. )
- 11. Should reactor coolant and containment air samples be found to contain unusually high copeentration1 of radiocesium, tellurium, ,
ruthenium, barium, and/or lanthauum, then'it is likely that fuel overheat has occurred. Selected radionuclides are presented in Attachment J. The presence of no single rrdioisotope can provide positive indication tLet fuel melt nas occurred. Rather, the degree of pellet failure (cuerheat versus melt).is better determined ny the , concentratiou of tho e radioactive species in the coolant system. and containment. , The LaSalle FSAR specifies containment activity due to iodine and noble gas species released from the core following a large break LOrA using the assumption basis of NRC Regulatory Guide 1.3. This basis ' is thst 100% of the core equilibrium inventory of noble gases and 25%
,,, cf the core equilibrium inventory radiolodines are released to the
( ) containment. Containment activity.is specified in FSAR Table U 15.6.5-2, herein presented-as Attachment G. The derived Attachment'I-provides core equilibrium (steady-state operation at full reactor power) inventorias of radiolodine and noble gas for LaSalle. [ Gap and core activity ratios are presented in Attachment-H. The activity ratloc for each radionuclide' are given relative to the Xe 188~ concentration for noble gas species and relative to the I l81 concentration for radiolodines.- It can be seen that.the fuel. gap F activity ratios significantly differ with. core inventory activity- , ratios. - Examine the results of radionuclide analysis for reactor - coolant and containment air samples. to evaluate whether the ' noble gas . - and ludius activity. ratius present att taure 'tefleullve of a gap release or pellet release. Note that radiolodines can form organic iodides which tend-to adhere to containment surfaces. Therefore, evaluations should more heavily weigh noble. gas activity ratio = profiles, rather than radioiodine. , t J ZWLZP i i
i LZP-1130 ! R3 vision 1 October 27, 1993 6 . 2 I 's_- !
- 12. Attachment K lists the predicted release fractions for several ,
radi.oisotope classes under various states of fuel degradation. Referring to the calculated (in Step F.5) core inventory or the steady-state full power equilibrium core inventory or.the steady-state full power equilibrium core inventory-given in Attachment I, compare the measured activity fraction of the full core-inventory with those release fractions contained in Attachment K in order to identify whether the mode of release is due to clad rupture, cicd oxidation, or fuel melt. Should the measured release exceed tha clad oxidation release fraction, yet be below the fuel melt release fraction profile fuel pellet overheat is possible. Pellet overheating would be accompanied by hydrogen generation; the overheat temperature regime is a minimal 400*r above the temperatures at which zircalloy oxidation occurs (1600*F, refer to Attachment L). If significant quantitles of hydrogen are detected in the containment (7.92 standard cubic fect of hydrogen per pound of zirconium reacted), and the measured release fraction profilce are at a level of at least au approximate 30% of the core inventory, an2 unusually high concentrations of radiocesimn, telluriums, ruthenim.), barium, ' and lanthanum are detected in reactor coolant / containment air samples, then it may be concluded that pellet over-temperature or melt is possible to have occurred. G. CHEC EIEIS v
- 1. None.
H. TECIDUfAL SPECIFICATIHLEEEEEEliCES
- 1. None.
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, 'LZP-1130-2 RevisionL1 '
October'27, 1993' 7 Ug . ATTACHMENT A IS I Fu?t vet- .w . UPPER RELLASE UWIT
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Relationship Between I-131 Concentration in the Primary Coolan (Reactor Water + Pool Water) and the Extent of Core Damage in Reference Plant. '{ l ZWLZP
LZP-1130-2 R9 vision 1 October 27, 1993 8
,, ATTACIDiENT_J / \ ! ) 4; 'd i evet urt w %
[ UP*ER RELIAst WWIT
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'_ - % CLADDING F AILuRE -{
1A
- 15 . ' 100t .
% PUEL MELTDOWN Relationship Between Cs-137 Concentration in the Primary Coolant (Reactor Water + Pool Water) and the Extent of Core Damage in Reference Plant.
("r x ,) ZWLZP
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LZP-1130-2 .
. Revision'1
' October 27,'1993 9
ATTACHMENT C ~ 10' %.
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gA _ le ~100 1- ECLAODINEPA4 LURE ' 1 - 1, - ,i 1.8 10 -151 + '
., s PUEL ' MELTDOWN _- -
Relationship Between Xe-133 Concentration in the-Containment Gas I (Drywell + Torus Gas) and the Extent of Core Damage in Reference Plant. ~ t O ZWLZP >
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- 'LZP-1130 .- Revision 1 October.27,-1993 10 AIIACllHEliI D O. tal. ~~
put L we'LTDcwa. * , WPPER RELIA &E LIW4T . SE.ITISTtW ATE LOWER RELEAEE UMIT f so - /.
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[- Ct.ADOtNG F AILUME , h #- / uppen RataAssuust . , a sera _ f f S E f f assrssTunaTu ' i -p p '
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escestNAL. 4a lef CJus Gees.
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3l ta"# ' S.1 la la ten ;
' % CLAcotNG 7AILURE ', _
'im ,so -- iso ' )
.jh % FUEL MELTDOWN
-{._
Relationship Between Kr-85 Concentration in'the Containment; Gas. !
-(Drywell i Torus Gas) and'the Extent of Core Damage.in Reference. Plant.
- a. , ;
4 ZWLZP !
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+
LZP-1130-2 Revision 1 October-27, 1993 :! . 11 l u O ATTACHMENT E LaSalle County Station Percent Clad Damage Versus Containment , Radiation Readings (Uncorrected). - N " " 10' - 5
, 3 d6V/A ,1 10'
,J A& /'// !
1 d(f// -i 10
; 4hv// .: J O 'j
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10, 2- 'S 10, f* . 5 .103 2 5 10 3
- 2. 5.-
.10
2 SL 10 8 Containment monitor reading (rad / hour) Clad Damage Assessments LOCA with Noble Gas Release Only; O ZWLZP .; I
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- . . . . , . . . - ,, . - b
LZP-1130-2 Revision 1 October 27, 1993 12 ATTACliMEliT_E LaSalle County Station i Percent Core Damage Versus Containment ! Radiation Readings (Uncorrected)
.]
.n., 4 $ s.s . s s . :
5 lfff ** i A ff// - Man 1e i MH// O la i say A ff// u i@V//
~
yy l 81818 I Ill l I:lill Illi l- Itill Illi 8 8 3 a =3 s
.g s ,4 w,. g, Containment. monitor reading (rad / hour)~
Core Damage' Assessment: LOCA with Noble Gas Release Only ZWLZP ;
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LZP-113022 Revision 1 ' October 27, 1993. 13 ATTACHMENT G TABLE 15.6.5-2 LOSS-OF-COOLANT ACCIDENT PRIMARY CONTAINMENT ACTIVITY (CURIES) (Design (NRC) Basis) ,
' ISOTOPE 1 Minute 30 Minutes 1 Hour 2' Hours 4 hours 8 Hours- 12 Hours 1 Day 4 Days 30 Days
- L I-131- -2.2E 07 2.2E 07 2.2E 07 2.2E 07 2.1E 07 2.1E 07 2.1E 07 2.0E 07 1.5E 07 1.4E 06 I-132 3.3E 07 2.8E 07- 2.4E 07 1.8E 07 9.8E 06 2.9E 06- 8.6E 05 2.2E 04 6.9E 06 0.
I-133- 4.9E 07 4.8E 07 4.7E 07 4.5E 07 4.2E 07 3.7E 07 3.2E 07 '2.2E 07 1.9E 06 1.5E 03-I-134. 5.6E 07' -3.8E 07 2.6E 07 1.2E 07 2.4E 06 1.0E 05 4.2E 03 3.1E 01 0. O. I-135 4.4E 07 4.2E 07 4.0E 07 3.6E 07 2.9E 07 1.9E 07 1.2E 07 3.5E 06 1.EE 03 0. TOTAL I 2.0E 08 -1.8E 08 1.5E 08 1.3E 08 1.0E 08 8.0E 07 6.6E 07 4.5E 07 1.7E 07 1.4E 06-KR-83M 1.4E 07- '1.2E 07 9.9E 06 6.8E 06 3.2E 06 7.2E 05 1.6E 05- 1.8E 03 3.3E 09 0. KR-85M 4.5E 07' 4.2E 07 3.8E 07 3.3E 07 2.4E 07 1.3E 07 7.0E 06 1.1E 06 1.5E 01- O. KR-85 1.4E 06- 1.4E 06 1.4E 06 1.4E.06 1.4E 06 1.4E 06 1.4E'06 1.4E 06 1.4E 01 1.2E 06 KR-87 8.0E 07- 6.1E 07 <4.7E 07 2.7E 07- 9.0E 06 1.0E 06 1.1E 05 1.6E 02 0. O. KR-88 1.1E 08 9.8E 07 8.6E.07- 6.7E 07 4.1E'07 1.LE 07 5.6E 06 2.9E 05 5.0E 03 0.
- KR-89 1.1E 08 1.9E 05 2.6E 02 4.9E 94 .0. O. O. O. O. O.
XE-131M 9.0E 05 9.0E 05 8.9E 05 8.9L 05 ' 8.9E 05 8.8E 05 8.7E 05 8.4E 05 6.9E 05 1.3E 05 XE-133M ~4.8E 06 4.8E 06 ~4.7E 06 4.7E 06 4.5E 06 4.3E 06 4.1E 06 3.5E 06 1.3E 06- 3.5E 02 XE-133 1.9E 08 -1.9E 08 1.9E 08 1.9E 08 -1.9E 08 .1.9E 08 1.8E 08 1.7E 08: 1.1E 08- 3.1E 06 XE-135M: ~5.1E 07- 1.4E 07 3.6E 06- 2.3E 05' 1.0E 03 1.9E 02 3.7E 07. O. -0. O. i XE-135- 1.9E 08 1.8E 08 -1.7E 08- 1.6E 08- 1.4E 08 1.0E 08 7.5E 07' 3.0E 07 1.3E 05 O. XE-137 1.5E 08 .7.8E.05 3.5E'03 6.8E 02 2.6E 11' O. O. O. O. O. XE-138 1.6E 08 3.8E 07 8.8E 06 '4.7E 05 1.3E 03 1.1E 02 8.8E 08' O. O. O. TOTAL' NG : 'l~1E 09 . 16.4E108 5.7E 08 4.9E 08 4.1E 08 3.2E 08 2.8E 08 2.1E 08 .1.2E'08 4.4E 06 ZWLZP _..____.x -.___ __ . . - . ._ _ . . - w. _ _ _ . , , ,
t LZP-1130-2 Esvision 1 October 27, 1993 14 y ATIACIDiENT H ; RATIOS OF ISOTOPES IN CORE INVENTORY AND FUEL GAP Activity Ratio
- in Activity Ratio
- in lag,topa Half-Life Core Inventory Fuel Gap Kr-87 76.3 m 0.233 0.0234 Kr-88 2.84h 0.33 0.0495 Kr-85m 4.48h 0.122 0.023-Xe-133 S.25d 1.0* 1.0* >
I-134 52.6 m 2.3 0.155 I-132 2.3 h 1.46 0.127 I-135 6.61h 1.97 0.364 I-133 20.8 h 2.09 0.685 I-131 8.04d 1.0* 1.0*
,-s
- Ratio = Noble Gas Isntooe ConcentratioD
[ b Xe-133 Concentration For Noble Gases
%)
= 12 dine Isotopg_.C2ncentration I-131 Concentration For Iodines +
l l r 4 i
,l' 1 ZWLZP l
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. I
^
LZP-1130-2 Revision 1 October 27, 1993 15
.% )
ATTACHMENT I CORE INVENTORY ACTIVITY FOR RADIOIODINES & NOBLE GAS ISOTOPE DECAY CONSTANT (day ~1) CORE INVENTORY (C1) Kr 87 13.08 8.0 x 10 7 Kr 88 5.86 ~1.1 x 10 8 Kr 85m 3.71 4.5 x 10 7 Ke 133 0.13 1.9 x 10 8 I 134 18.98 2.24 x 10 8 I 132 7.23 1.32 x 10 8 I 135 2.51 1.76 x 10 0 I 133 0.80 1.96 x.10 8
\~- - I 131 0.086 8.8 x 10 7'
.. i s
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1 i-1 ZWLZP I
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_ ._ m . . . . _ . . LZP-1130-2 Ravision 1 October 27, 1993 16 U anAcmist1I_J SELECTED NUCLIDES FOR CORE DAMAGE ASSESSMENT Core Damage __ Stage tincli.de IIalf-Life
- EIndQminant Ganunas (Kev) Yield (%1*
I Clad Failure Kr-85m** 4.4 h 150(74), 305(13) Kr-87 76 m 403(84), 2570(35) Kr-88** 2.8 h 191(35), 850(23), 2400 (35) , Xe-131m 11.8 d 164(2) Xe-1133m** 2.26 a 233 (14) Ke-135** 9.14 h 250(91) I-131 8.05 d 364(82) I-132 2.26 h 773(89), 955(22), 1400(14)' I-133 20.3 h 539(90) I-135 6.68 h 1140(37), 1280(34), 1460(12), 1720(19) ' Rb-88 17.8 m 898(13), 1863(21) Fuel Overheat Cs-134 2 yr 605(98), 796(99) Cs-137 30 yr 662(85) Te-129 68.7 m 455(15) Te-132 77.7 h 230(90) 4 Fuel Melt Sr-89 52.7 d (beta emitter) Sr-90** 28 yr (beta emitter) Ba-140 12.0 d 537(34) La-140 40.22 h 487(40), 815(19), 1596(96) La-142 92.5 m 650(48), 1910(9), 2410(15), 2550(11) Pr-144 17.27 m 695(1.5) Ru-105 4.44 h 730, 480, 670, 320 Ru-107 4.2 m 195, 370, 480, 860 j Ru-97 '2.92 216.
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- Values obtained from Inble of Isotopgit Lederer, Hollander, _and Perlman, Sixth Edition.
** These nuclides are marginal with respect to selection criteria for candidate.
.nuclides; they have been included on the possibility that they'may be detected and thus utilized in a manner analogous to the candidate nuclides.
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O O' LZP-1130 O- - Revision 1 October 27, 1993- , 17 ATTACHMENT K r BEST-ESTIMATE FISSION PRODUCT RELEASE FRACTIONS
' GAP RELEASE MELTDOWN RELEASE- QZIDATION RELEASE VAPORIZATION RELEASE Lower Upper Lower Upper .' Lower . Upper Lower Upper Nominal Limit Limit Nominal Limit Limit ITominal Limit Limit Nominal Limit Limit 1
Noble Gases '0.030 0.010' O.12 0.873 0.485 0.970 0.087 0.078 0.097 0.010 0.010 0.010 (Ke, Kr)
-Halogens' O.017 0.001 0.20 0.885 0.492 0.983 0.088 0.078 0.098 0.010 0.010 0.010
+ (I,-Br)
- -Alkali Metals' 'O.050 0.004 -0.30 0.760- 0.380 0.855 0.190 0.190 0.190
-- (Cs , ' Rb )
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LZP-1130-2' Ravision 1.. October 27, 1993 18 A 1 ATTACHMENT L EXPECTED FUEL DAMAGE CORRELATION WITH FUEL ROD TEMPERATURE IO) Illf4_ DAMAGE 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 A I h Gain Boundary Diffusion Release of
! 2450 - 3450 Fission Products Fuel Melt > 3450 Dissolution and Liquefaction of UO2 I" '
the Zircaloy - ZrO2 Eutectic Melting of Remaining UO2 5100
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- These temperatures are material property characteristics and are non-specific with respect to locations-within the' fuel and/or fuel cladding.
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LZP-1130-2
,- Ravision'1 October 27, 1993
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O Amcmimu SAMPLE CALCULATION OF FISSION PRODUCT INVENTORY CORRECTION FACTOR D F yi = Inventory of anc11de i in reference plant Inventory of nuclide-1 in operating plant 3651 (1-e-1095Xi ) E -A T -A T o j P3 (1-e I i) e I$ Where: P 3 = steady' reactor power operation in period J (MWt) Xi = decay constant of isotope i
?$ = duration of operating period j (days) 0 Tj t time between the end of operating period j and time of the last reactor shutdown (days).
3651 = ave operation power (in MW t) for the reference plant. 1095 = continuous operation time.(in day) for the-reference plant.. ! Assuming a reactor has the'following power operation history: Operation Operation Time 'O Average Power-Period Days Since Startup: T3 (day) Tj ~ Pj(MWt): 1A 1 - 60 61 255 1000 18 61 --70 --- --- 0 2A 71 - 270 200 ~45- 2000 2B 271 - 300 --- --- .0 3 301 - 314' 14 0 3000l For I-131 (11 = -0.086 day 11 ,
.j es -1095 x 0.086
= 3651 (1-e i
!g_j/. FI (i=I-131) j .P) (1-e I 3) e I)
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LZP-1130-2
- Rsvision 1 October 27, 1993 20 FINAL-1 AH ACHMENT M (Continued)
WHERE: o E (P (1-e~ 1 j)e~ j }
= 1000(1-e-0.086 x 61)(,-0.086 x 255) + 2000(1-e-0.086 x 200)4,-0.086 x 45)
-0.086 x 14 o
+ 3000(1-e
= 2.9 x 10" + 41.6 + 2100.7 = 2142.3 F7 ,131 = 3151(1-e-1095 x 0.086) = 1.7 2142.3 Enr Cs-137 (A1 = 6.29 x 10~ day ~ )
F -6.29 x 10~ x 1095 ) I(I=Cs ) = 1651(1-e j)e~ ] E)(P (1-e-
-6. x 0 x 61)(,-6.29 x 10~ x WHERE: E. = 1000(1-e 255) 3
-6.29 x 10
[ + 2000(1-e x 200)(e-6.29 x 10" x 45 )
-6.29 x 10~
+ 3000(1-e x 14) 0
= 3.80 + 24.93 + 2.64 = 31.37
-6.29 x 10~ x 1095 )
I(I=Cs ) = 3651(1-e - 243.15
= 7.75 j)e~ 31.37 I [P (1-e~ J y
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