ML18031A429

Revised Std Practice Instruction TI-88, Procedure for Estimation of Extent of Core Damage Under Accident Conditions.

ML18031A429
Person / Time
Site:	Browns Ferry
Issue date:	12/23/1985
From:	LEWIS R L TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML18031A428	List: ML18031A427 ML18031A429
References
TAC-61613, TAC-61614, TAC-61615, TI-88, NUDOCS 8605280166
Download: ML18031A429 (36)

Text

7r rmeSSee Jr![!ey AuthOri i, BroL'ns Ferry Huc!e lr Plant STA jt)A!lD PM(,T[CE/PERHANENT

[HSTRUCT Standard Pr.!ctice/'in" truction~itle Procedure for Estimation of the Extent of Conditions t A re or2'rPar" e Form BF-5!'F l.2[ON CH."s4(rE[NFOR!".AT IGir Core Damage Under Accident PER!fA"'LzVI!S'IL.uY.

ION C.'!ANGE[s this a d'or!(pian-initiated change'?Yes (Mori(plan Does thi" implement, delete, or change a L.eference to a vendor manual?:"" yeso"CC Vendor t!anual Coordinator signature is required.No)';:o.x Yes X'io Vendor Manual Coocdinator Does this implement, delete, or change an SI data cover sheet IE yes, Planning and Scheduling supervisor signature is requi Date Yes X No red./P l ann i ng 6 Schedu!, i ng Date IE item added, deleted, or Erequency changed on IE (or CSSC)PM requirements, complete BF-ill.ST~%DARD PR KCT[CE ChANGE..QA reviev required p.ior to issuance?PGR(.'evieLI required prior to issuance'(See'""ac t:Lent..0 stall(!ard pL.actice OF-!Qua[i(y.-'.s:!rance~co ram requirements have Yes Yes 2.)been prooer[y'ncluced iio Quality Assu ance Supv.Date Intent.:h.:n e ill:>cocedural detail oE FSAR or other licensin.document?LreS::o X Net/Ills~'c"-on'=s X No':=ves<<o either<<E t!Lese questions, A.USQD i'equired.

'.;thi'hange in response to an LE!?, IE bulletin, NRC inspection manageŽnt/supervi.;or inspection, COAB'.udit, etc'?IE yes, spec'nder'ca>>on Eoc re I!sion.JAS i.h.'hanlee made tO r,:eet an N!(C Ciltr11[tl!lent".Yes If ye>>'.o ei(tier question, refer tr Of:-:!.3 tor proper'dentit'icat Doe, t!Lis L(.vision implement a so(LL'<..locument?".es If ye>>, lrtach:orm BF-r'e.Due d;lt(e otocm BF-~e i" report, Ey docu..ent e~o%'.O ion oi ta!e cha'nge:io ire!otectiofl Nn Fire!'.otection System invol:ed'?

/Engineer Date Security System invol;ed?....+0.""'Lblic Safety Supervisor Date Rerieued by See BF-[.2 Eor standard J.L., Turner Submitted by praC t leer~~EN'22/.18/85/..26!il.Sacr!oo Dere 8!'.o.".e!4v rre:::7m/g3%Supv~Date The requir~>cents herein ,C+'Jc'J'r'pcz/Q PORC C!la 1':.an"Rev'L s 1 0(l are consistent wi//=Sorry f)ate th saF;E ie 850425-22 Df'SCU Job No.operation.

8b052801bb Bbo 259 PDR ADOCK 05000 P Page 2 Form BF-5 cBF 1.2&2.3 April 10, 1985 T ssee Valley Author'sty Browns Ferry Nuclear Plant STANDARD PRACTICE/PERMANENT INSTRUCTION CHANGE INFORMATION (Continued)

REVISION LOG Standard Practice/Instruction BF TI-88 APPROVAL DATE PAGES AFFECTED REASON FOR REVISION Qpg p3>g95 All New Procedure.

(85O425-22)

DP&CU Job No.850425-22 Total Number of Pages in Procedure:

34 Retention Period: Lifetime Responsibility:

Revision DP&CU Supervisor OE4 4 0 l303'Page 1 BF TI-88 TI-88 PROCEDURE FOR ESTIMATION OF THE EXTENT OF CORE DAMAGE UNDER ACCIDENT CONDITIONS

1.0 Statement

of A licabilit 1.1~Sco e This procedure is intended to provide the information required to permit the assessment of the degree of reactor core damage from measured fission product concentrations in either the water or gas samples taken from the reactor systems and by the interpretation of other plant parametric data under accident conditions.

1.2 Summar

of Method 2.0 The procedure involves the calculation of fission product inventories in the core and the release of this activity into.the reactor systems under postulated LOCA conditions.

Measured concentrations of major fission products in either gas or water samples (after appropriate normalization) are-compared with the reference plant data to estimate the extent of core damage.Containment structure radiation levels, hydrogen concentrations in containment atmosphere, and the core water level history can then be used to confirm the initial estimate which was based on radionuclide measurements (see Appendix A).References This procedure is based on, and follows the general principles presented in, the publication NED0-22215, dated August 1982, by the General Electric Company.0045E DEC 23 1985 Page 2 BF TI-88 The dedicated postaccident sampling systems should be employed as described in CI-1302, CI-1303 and CI-1305 procedures for reactor water, drywell atmosphere, and torus atmosphere sampling, respectively.

The torus water will be sampled with the normal sampling procedure CI-464.7.Samples so obtained shall have been analyzed qualitiatively and quantitatively for radionuclide content by appropriate procedures.

See reference procedures BF-CI, Procedures 701, 708, 709, 715, 716, 719, and 720.4.0 Precautions The conclusions from this TI are meaningful only when the samples that were analyzed for their radionuclide concentrations are representative of the systems from which they were sampled.5.0 M~ain Bnd 5.1 Plant Samples Collection and Analyses 5.1.1 Determine sample locations per Appendix B.5.1.2 Request information for Data Sheets A and B of Appendix C be collected.

Use the time of reactor shutdown, when the ND6600 gamma ray spectroscopy system requests the sample time so that all results are decay corrected to the time of reactor shutdown.5.1.3 Correct measured gas sample concentrations for temperature and pressure differences in the sample vial and containment (drywell)gas phases, calculate the total coolant and containment gas concentrations, per Appendix D and record on Worksheet 1.0045E GEC Zi i995 Page 3 BF TI-88 Plant Samples Collection and Analyses (Continued)

5.1.4 Calculate

the fission product inventory factor, Fi>, per Appendix E and record on Worksheet 1 (for I-131, Cs-137, Xe-133, and Kr-85 only).5.1.5 Complete Worksheet 1 to normalize concentrations to the reference plant.5.2 Estimation of Core Damage (See analysis sequence in Appendix A.)5.2.1 Use Figures 1 through 4 to estimate the extent of fuel/cladding damage and record on Worksheet 2.5.2.2 Use Appendix F to determine the extent of metal-water reaction based on hydrogen concentration in the containment and record on Worksheet 2.5.2.3 Use Appendix G to determine the fuel inventory release to the containment and to establish a c mage estimate.Record damage estimate on Worksheet 2.5.2;4 Review reactor vessel water-level history to establish if there has been an interruption of adequate core cooling.Record conclusions on Worksheet 2.5.2.5 Review Data Sheet A of Appendix C to determine if concentrations of Sr-92, Ba-140, and Ru-103 are unusually high.Record findings on Worksheet 2.5.2.6 Use Appendix H to determine fission products ratios and record results on Worksheet 2.5.2.7 Determine degree(s)of damage per Appendix I and record on Worksheet 2.5.2.8 Repeat as needed for updated determination of core post-accident status.0045E OEC 23$9b Page 4 BF TI-88 TABLE 1 Reactor Thermal Power No.of Fuel Bundles Total Coolant Mass (reactor water+suppression pool water)Total Drywell and Torus Gas Space Volume Reactor Water Mass Suppression Pool Water Mass Drywell Gas Volume Torus Gas Volume Reference Plant 3651 MW~748 bundles 3.92 x 10'4.00 x 10'm 2.46 x 10 g 3.67 x 10 g 7.77 x 10'm'25 x'.0 cm BFNP 3293 MW~764 bundles 3.72 x 10 g 7.87 x 10 cm 2.39 x 10 g 3.48 x 10 g 4.50 x 10'm'.37 x 10 cm'045E DEC i3 m85 Page 5 BF TI-88 TABLE 2 Chemical Group Major Gamma-Ray Energy, KeV Intensit (/d)CORE INVENTORY OF MAJOR FISSION PRODUCTS IN A REFERENCE PLANT OPERATED AT 3651 MW, FOR THREE YEARS Isotope'alE-Life Inventory'lo'i)Noble Gases Halogens Alkali Metals Tellurium Group Noble Metals Alkaline Earths Rare Earths Kr-85m Kr-85 Kr-87 Kr-88 Xe-133 Xe-135 I-131 I-132 I-133 I-134 I-135 Cs-134 Cs-137 Cs-138 Te-132 Mo-99 Ru-103 Sr-91 Sr-92 Ba-140 Y-92 La-140 Ce-141 Ce-144 4.48 h 10.72 y 76.3 m 2'4 h 5.25 d 9.11 h 8.04 d 2.3 h 20.8 h 52.6 m 6.61 h 2.06 y 30.17 y 32+2 m 78:2 h 66.02 h 39.4 d 9.5 h 2.71 h 12.8 d 3.54 h 40.2 32.5 d 284.3 d 24.6 1.1 47.1 66.8 202.0 26.1 96 140 201 221 189 19.6 12.1 178.0 138 183 155 115 123 173 124 184 161 129 lsl (o.7s3)s14 (o.oo44)403 (0.495)196 (0.26)1530)0.109)81 (0.365)2so (o.899)364 (0.812)668 (0.99)773 (0.762)s3o (o.86)847 (0.954)884 (O.6S3)1132 (0.225)1260 (0.286)605 (0.98)796 (0.85)662 (0.85)463 (O.3O7)1436 (O.76)228 (0.88)74O (O.128)497 (0.89)7so (o.23)1024 (O.32S)1388 (0.9)537 (o.2s4)934 (0.139)487 (o 4ss)ls97 (o.9ss)145 (O.48)134 (O.1O8)Refractories Zr-95 Zr-97 64.0 16.9 h 161 166 72'O.437)7s7 (o.ss3)743 (O.928)NOTE: See footnotes On next page.0045E UEC 23$95 Page 6 BF TI-88 TABLE 2 (Continued)

Footnotes:

'Only the representative isotopes which have relatively large inventory and are considered to be easy to measure are listed here.At the time of reactor shutdown.0045E OEc 23~%5 Page 7 BF TI-88 WORKSHEET 1 CONCENTRATION CORRECTIONS AND NORMALIZATION TO REFERENCE PLANT 1.Gas Concentrations Corrected for Non-Standard Pressure and Temperature and decay from Appendix D (uCi/cm'):

Nuclide Xe-133 Kr-85 Kr-85m Kr-87 Kr-88 Cs-137~De el 1 Torus 2.Total Coolant and Containment Gas Concentrations from Appendix D Corrected for Decay to Time of Reactor Shutdown.Nuclide Reactor Water (uCi/g)Nuclide Containment Gas (uCi/cm')I-131 I-132 I-133 I-134 I-135 Cs-137 1 Xe-133 Kr-85 Kr-85m Kr-87 Kr-88 3.Fission Product Inventory Correction Factors, F<>, from Appendix E: Nuclide Fii I-131 Cs-137 Xe-133 Kr-85 0045E m QEC 2 3 1985 Page 8 BF TI-88 WORKSHEET 1 (Continued) 4, Calculate the Normalized Concentrations to the Reference Plant Using Results 2 and 3 Above: (~i)(Ii)(g)()()(0.949)(2)(3)C~=(CGi)(FIi)(FG)REF=()()(0197)(2)(3)Nuclide REF C (uCi/g)i1i Nuclide REF C (uCi/cm)Gi I-131 Cs-137 Xe-133 Kr-8S NOTE: The calculations have ignored the"presence of I-131 and Cs-137 in the gas samples and the presence of Xe-133, and Kr-8S in the liquid samples.If these assumptions are found to be invalid, corrections

~sill have to be made.0045E DEG 23 Soi Page 9 BF TI-88 WORKSHEET 2 ESTIMATE OF FUEL/CLADDING DAMAGE l.Reactor Water Analysis: Nuclide C (uCi/g)Wi X Cladding X Fuel Meltdown Failure I-131 Cs-137 Containment Gas Analysis: Nuclide C (uCi/cm'Gi X Cladding X Fuel Meltdown Failure ,r Xe-133 Kr-85 2.Per Cent Metal/Water Reaction from Appendix F: 3.Damage Estimate Based on Containment Radiation Measurement from Appendix G: 4.Water Level History Indicates Interruption of Adequate Core Cool-ng?(yes)~(No)5.Concentrations of Sr-92, Ba-140, La-140 and Ru-103 Unusually High, Inferring Some Degree of Fuel Melting?(yes)(No)6.Results of Fission Product Activity Ratios Calculated in Appendix H Indicates Some Degree of: a).Fuel Melting (No)b).Fuel Overheating (No)7.Core Damage Assessment Categories Applicable per Appendix I: Minor Intermediate Major 0045E SEQUENCE OF AtlALYSIS FOfl ESTIMATlnt OF CpnE DAMAG LOW Ilyd rogon Ana lysis f rm YES Con ta Inmon t YES zT'ar YES HOAMAL OPEAATIOtl Aadl a t Ion Conf I rn Cont'Irm HO ItO IIO Oo t c rn I no Optinum Samp la Po if 1 Coro Damago Estlmato f~rom SS II I GII HO HO HO lly rogon=Ana lysis conrtr YES Conta Inmont YFS Wator Aad la t Ion Lovo I Conf I rn Co fl n YES Ana lys I s for Aa Sr la An MAJOR CLAD DAMAGE YES FUEL OVEflllEAf

~IIII IIII T Dotormlna t ton Of F lsslon Prodffc Aa lo NO Y CLAD DAMACTE POSSIDLE FUEL OVEAHEAT NO COAE MELT tjtt~fTI (ITt f-I e~CO O~~CO CJl DEC 23 1985 Page 11 BF TI-88 APPENDIX B Determination of Sampling Location: In order to ensure a representative sample which reflects the actual incore conditions, care must be taken in selecting a suitable sampling locations For gas sampling, the recommended sampling locations are as follows:~Event 1 e Non-Breaks (e.g., MSIV Closure)Sam lin Location Suppression Pool Atm.Small Breaks Drywell (before depress.)Suppression Pool Atm, (after depress.)Large Break (Liguid or, steam in containment)

Large Breaks Outside Containment Drywell Suppression Pool Atm.For liquid sampling, the optimum sample point for all events is the jet pumps as long as there is sufficient reactor pressure to provide a sample from that location..If there is insufficient reactor pressure to allow a sample to be taken from the jet pumps, then the sample should be taken from the sample point in the RHR system.In order to ensure a representative liquid sample from the jet pumps at low ((1'L)power conditions for small break or non-break events, the reactor water level should be raised to the level of the moisture separators.

This will fully flood the moisture separators and will provide a thermally induced recirculation flow path for mixing.0045E DEC 2 3$85.Page 12 BF TI-88 APPENDIX C DATA SHEETS A.CHEHICAL 1.Containment Gas Concentrations (uCi/cm)decay corrected to time of reactor shutdown.Nuclide Drywell CGi Containment P2, T2 7P Vial 1Xe-133 2Kr-85 Kr-85m Kr-87 Kr-88 Nuclide Torus CGi Containment P2 T2 Vial Pl Tl Xe-133 2Kr-85 Kr-85m Kr-87 Kr-88 2.Reactor Mater Concentrations (uCi/g)decay corrected to time of reactor shutdown.Nuclide Rx tJater CMi Supp.Pool Mater CMi 1I-131 I-132 I-133 I-134 I-135 2Cs-137 3.Other Nuclides in Rx Mater Unusually High (relative to normal)Sr-92 RU-103 Ba-140 La-140'Zes Yes Yes Yes No No No No 4, Concentration of Hydrogen in Containment Atmosphere:

mole%.(Continued on next page)0045E DEC 23 1985 Page 13 BF TI-88 APPENDIX C (Continued)

Footnotes:

P, T: Pressure and temperature

'Most important nuclides initially Not likely to be measureable until shorter-lived nuclides have decayed.NOTE: Concentrations from both drywell and to'rus needed for gaseous samples, and concentrations from both Rz water and suppression pool needed For water samples.B.OTHER 1.Date/Time of Shutdown: 2.Power Level Prior to Shutdown: MW, 3.Containment Monitor Read ng: Time: Rem/hr Distance of Detector from Reactor Biological Shield Wall: feet 4.Reactor Water Level History (attach): 5.Operating Power History: a.Obtain the power history from the Monthly Operating Reports.Plot power history for the last 80 days.Detail is required since this plot will be used for short-lived nuclides.Attach data sheet with tabular information if possible.3 IC~~lf 4)t QO 70 b.Plot power history for the present and last two cycles.Plot average power and ignore short term outages since this data will be used for long-lived nuclides.This plot may nr t be required if longer-lived nuclides have not been detected./GC pgO~O 0045E'i 0 (Q rrC~~<~(O gCC DEC 23 1985 Page 14 BF TI-88 APPENDIX D CORRECTION OF CONCENTRATIONS FOR PRESSURE AND TEHPERATURE DIFFERENCES 1.Correct gas concentrations for pressure-temperature differ'ences in sample~vial (Pl, Tl)and containment (P2, T2).Record on Worksheet 1: P2 Tl CG (corrected)

=CGi(vial)------

--'--1 Pl T2 NOTE: Use temperature values on absolute scale, e.g.: oK oC+273.2, or oR=oF+459 2.Calculate total coolant and containment gas concentrations and record on Worksheet 1: Tot 1 Wi (Rx water conc.)(Rx water mass)+(Su.ool conc.)'(ool mass)(Rx water mass+pool water mass)=(Rx water conc.)(2.39 x 108)+(Su.ool onc.)(3.48 x 109)3.72 x 109 g Total Gi (Dr ell conc.)(Dr well as Vol.)+(Torus Conc.)(Torus as vol.)(Drywell gas vol.+(Torus gas vol.)(Dr ell conc.)(4.5 x 109 cm+(Torus conc.)(3.37 x 10 cm 7.87 x 109 cm If the torus water cannot be sampled due to high radiation levels the equations above may still be used provided that the assumption can be made that the Suppression pool concentrations are equal to the Reactor water concentrations.

FG and FW are normalization factors to reference plant for containment gas volume and reactor water mass, respectively:

FG=BFNP total containment gas vol., 7.87 x 10 cm Reference plant containment gas vol., 4'x 1010 cm3 0.197 BFNP total Rx water mass, 3.72 x 109 g FW=Reference plant Rx water mass, 3.92 x 109 g 0.949 DEC~S tOSS Page 15 BF TI-88 APPENDIX E SAMPLE CALCULATION OF FISSION PRODUCT INVENTORY CORRECTION FACTOR FIi 3651 FIi=3293 where: Inventory of nuclide i in reference plant Inventory of nuclide i in operating plant (1-exp.(-1095 Xi))gK WK gj Pj (1-exp (-Xi Tj)(exp (-Xi Tj))then: 3651 HWt=average operating power for the reference plant.3293 MWt=rated operating power for each BFNP unit.1095 days='ontinuous operation time for the reference plant.fractional steady reactor power operated in period j, (MWt 3293).decay constant of nuclide i, (day 1)-see Appendix D.duration of operating period j, (days)time between the end of operating, period j and time of current reactor shutdown, (days).'the weighting factor for cycle K assuming n is the present cycle, n-1 is the previous cycle, and n-2 is the cycle before the last cycle Wn=1 because all of the present core was irridated during Cycle n Wn-1=2/3 because 2/3 of the present core was irridated during Cycle n-1 Wn-2=1/3 because 1/3 of the present core was irridated during Cycle n-2 WK accounts for fuel discharge 0045E DEC"-)IgII5.BF TI-88'10 P U E L ME LTOOWN UPPER RELEASE LIMIT BEST ES i IMATE LCWEA RELEASE LIMIT 5 O vl U i z 4t Z I Z 0 V 0 I I la>L7 0 EJ Pl Pl X 10'1.0 r rl rr II CLAOOING P AILURE UPP'EA RELEASE LIMIT PEST ESTIMATE, LOWER AELEhSE LIMIT-~~~-.C.I NORM AI OPE A AT I NQ CONCENTRATION IN QAYWELL UPPER LIMIT.NOMINAI.4 I 0 Ci/cc 10 5 aCsicc 10 0.1'I 0~I~i i~~..~.~I 10~~~I f~<CLAOQINQ FAILURE 1,0 10 100 I i ruas MeivacwN-M I Figure 3.Relationship Between Xe-133 Concentration in the Containment Gas (Dry'well+Torus Gas)and the Extent of Core Damage in Reference Fiant 0045E UEt: 23 1995 Page 17 BF Tj-88 10 F U E I.M E L TD OW H 10 U~ER RELEASE LIMIT BEST ESTIIAATE LOWER RELEASE I.IMI T 1.0 M vs U I Z I O 10 Z O I C I z O U 10 l X CLAOOING F AILURE UFI'ER RE I.EASE LIMIT BEST ESTIMAT'E LOWER RIELEASE LIMIT 10 HCRM*L OFERATION CONCENTRATION IN ORYWEI L UPPER LIMIT HOMIHAI.-5 4 a 10 i Cilcc-6 4<10~ci/cc 10 0.1 1.0 4 CLAOOINC F AILURE 10 10'4 FUEL MELTDOWN Figure 4.Relationship Between Kr-85 Concentration in the Containment Gas (Drywell+Torus Gas)and the Extent of Core Damage in Reference Plant 0045E DEC 23 1985 Page 18 BF TI-88 APPENDIX E (Continued)

Assume the reactor has been operating for 40 days in the current fuel cycle (Cycle n), the previous refueling outage was 90 days and the power history for the last 40 days was as shown in Figure 5.Prior operating history was 16 months at an average 86%power (Cycle n-2)followed by a 9 month outage followed by 12 months at an average 93%power (Cycle n-1)with a 90 day outage leading to the current fuel cycle.One third the core was replaced at each outage.This power history is plotted by month in Figure 6.Calculate the factors FIi for 20.8 h I-133, 5.25 d Xe-133 and 10.72 year Kr-85.Step 1.Determine the values of the decay constants for the nuclides of interest from Xi=ln2/half-life.

For consistency, this example uses units of days->.However, any consistent time units could be used.Nuclide Half-Life da s-1 I-133 Xe-133 Kr-85 20.8 h 5.25 d 10.72 y 0.800 0.132 1.77 E-4 Step 2.Draw a power plot with values of Ti and ti as in Figures 5 and 6.The details of the power plot and the duration of the plot should be commensurate with the half-lives involved: It is not necessary to plot more than 5, or at most 6 half-lives, of data.b.The longer the half-life the less the detail which is required.Therefore, the plot showing all 3 fuel cycles (Figure 6)does not require detail.Step 3.Calculate the values of FIi.I-133 Five half-lives is approximately 5 days so it is necessary only to consider the final 8 day power interval.From Eq.2: FI I 133 (3651/3293) x 1/[0.90 (1-exp (-0,800 x 8))]=1.11 x 1/(0.90 x 0.998)=1.23 Only fJK=1 is required due to the short half-life.

0045E o C)C W 0 III Jags s jl'I: sl.s~I ss~li": L 20 s~I~-Tq-II I s~s'ij I I;I.ji 1 I s s J s sssl 5 I I s 30 40 s s'ZII 4-gS s I s I I I&Jsssls s I s s T.8 LJs~I s ilj'I II]li I jIj sI IIs~~s 50 s I;;i!: Is s:,I Qs I~~~I II:I~~60 I J gs Figure 5 Example Pover Ilistory Plot by Days.1IIP 80 60 40 20 0:I~I~~II~~s s I~s s~~'~~s~i!II s~,~II jli i~(s s sj I II s s s~s~s s s s III:,I,;ALII I;s I s~~s'I~~)as.!I s'.~st li:I~I~I~II s I s~'~~~~~~~':Is~i.I:,.I I:: I,.~!I~s j I I I s s s~~s s~I Is',lj"1I's~s I s'.: I i!s}:~Is~s's:I I I s'II'!~~~s s s n 0 p~s C Q s s I I 10 Ilsl sssssn'I4-~~s s 20 s s s I 30 25 sssss JIIS 4Q 50 s s s s-I assn S-s s.~I..-JI sssos IIss Ils s s I moni<60 n nnlnS I ssI Q fsI H e IIII I-I Co sCI~I CO OO CJl Figure 6 Example Power Iiistory Plot by Honths.

Page 20 BF TI-88 APPENDIX E (Continued) b.Xe-133 Five half-lives is approximately 26 days.This takes the calculation back to the beginning of the 80'L power interval.From Eq.2: FZ Xe 133=(3651/3293)x 1/(0.80[1-exp (-0.132 x ll)][exp (-0.132 x 13)]+0.90[1-exp (-0,132 x 8)])=1.11 x 1/{0.80[0.766][0.180]

+0.90[0.652])=1.11 x 1/(0.110+0.587)=1.59 Only MK=1 is required due to the short half-life.

c.Kr-85 Long half-lives such as Kr-85 and Cs-137 require special consideration.

Designate the 16 month irradiation fuel cycle as Cycle n-2, the 12 month irradiation cycle as Cycle n-l, and the current cycle as Cycle n.Only 1/3 the fuel originally present in Cycle n-2 is still in core while 2/3 the fuel from Cycle n-1 is still in core, It is assumed that none of the fuel loaded at the beginning of Cycle n-1 was removed at the beginning of Cycle n, only fuel from Cycle n-2.Because of the long half-life it is necessary to use all three t4 terms.The numerator has the valve of K[1-exp (1095 x 1.77 E-4)]or 0.176.For simplicity the denominator will be divided into 3 parts, one for each cycle.Equation 2 takes into account fuel discharge.

Note, however, that each term in the summation is activity remaining from each irradiation.

Therefore, each prior irradiation period should be multiplied by the fraction of the fuel load still present at the beginning of Cycle n.0045E OEC 23 3985 Page 21 BF TI-88 APPENDIX E (Continued)

At 30.5 days per month, the Cycle n-2 contribution is: 1/3 x 0.86[1-exp (-1.77 E-4 x 16 x 30,5)][exp (-1.77 E-4 x 25 x 30.5]=0.287[0.0828][0.874]

=0.0208 The Cycle n-1 contribution is: 2/3 x 0.93[1-exp (-1.77 E-4 x 12 x 30.5)][exp (-1,777 E-4 x 4 x 30.5)]=0.62[0.0627][0.979]

=0.0380 t The effective Cycle n irradiation is 907 power for 14+11+8=33 days.(Because of the long half-life it is possible to use the average power and ignore short-term outages.)The Cycle n.contribution is 1 x 0.90[1-exp (-1,77,.E-4 x 33)]=90[5.82 E-4]=5.24 E-3 The value for KI Kr 85 is-therefore 3651 x 0.176 3293 o+s+2.47 0045E DEC 23 1985 Page 22 BF TI-88 APPENDIX F INTERPRETATION OF CONTAINMENT ATMOSPHERE HYDROGEN GAS MEASUREMENT AS INDICATOR OF CORE DAMAGE The extent of fuel cladding damage as evidenced by the amount of metal-water reaction can be estimated by determination of the hydrogen concentration in the containment atmosphere.

The hydrogen concentration is measurable by either the containment hydrogen monitor or by samples taken with the post-accident sampling system.A correlation has been developed which relates containment hydrogen concentration to the percent metal-water reaction for Mark I containment designs.That correlation is shown in Figure 1 of the Appendix F.Analytical Assumptions:

1.Containment Volume=350,000 ft (for Mark I 6 II).2.No.of fuel bundles=500 (for Mark I 6 II).3.Fuel type: 8 x 8 R.4.All hydrogen from metal-water reaction released to containment volume.5.Perfect mixing of hydrogen in containment gas volume.6.No depletion of hydrogen, e.g., containment leakage.7.Ideal gas behavior for containment gases.0045E OEC 23 1905 Page 23 BF TI-88 APPENDIX F (Continued)

Procedure:

l.Obtain containment hydrogen'onitor reading or postaccident sample analysis value for hydrogen concentration, in percent H<.2.Using the curve in Figure 1 of this Appendix, determine the percent metal-water reaction for the reference plant.3.The extent of the metal-water reaction for the actual BFNP in-plant conditions is determined from the following equation: 500 V 7o Metal-Water

=(X M-W,o p>(---)(N 350,000 Where N=No.of fuel bundles=764 V=Total containment free volume=278,000 ft 4.Record X Metal-.Water-reaction from Figure 1 of this Appendix on Worksheet 2, item 2.0045E 3~-*Cf"0 O'~V M~IX Q m IIi+I m 4J I K 0 0.0 0 LI gr V 0 r'II bO 0 20 60 70%METhL-tTATErl flEACTIOH MJAKO&rll l3E'C 2 3 f985 Page 25 BF TI-88 APPENDIX G INTEGRATION OF CONTAINMENT AREA RADIATION INTENSITY MEASUREMENTS INTO CORE DAMAGE ESTIMATE An indication of the extent of core damage is the containment radiation level which is a measure of the inventory of fission products released to the containment volume.This appendix contains a correlation method for relating the~containment radiation monitor dose rate value to the percent of fuel inventory rendered airborne in the containment volume by the accident.The purpose of this appendix is to present that correlation and to provide a method for use in determining the degree of core damage based on that correlation.

Figure 1 of this Appendix G provides the results of a correlation performed for the Monticello plant.The key parameters which affect the postaccident containment dose rate are reactor power history, containment volume, and monitor detector location within the containment volume.0045K Note A Determination of Cladding Damage from Containment Radiation Monitor Readings DEC 23 iSOS Page 26 BF TI-88 The procedure for determination of the fraction of the fuel fission product inventory released to the containment is as follows: 1.Obtain containment radiation monitor reading, R, in Rem/hr.2.Determine elapsed time from plant shutdown to the containment radiation monitor reading, t, in hours.3.Using Figure 1, this Appendix G, determine the fuel inventory release for the reference plant, I,,<~in percent.4.Determine the inventory release to the containment, I, using the following formula: 1670 V 6 I=I, (---)(-----)(--)P 237,450..Dwhere P=reactor power level, MWat shutdown V=total containment free volume=278,000 ft D=distance of monitor detector from reactor biological shield wall, ft.5.Determine the estimate of core damage from Table 1 of this Appendix G./Record on Morksheet 2, item 3.0045E FIGURE DEC 23 1985 Page 27 BF TI-88 percent of toel$~~ASzbc~e M the Contafzxsent 100l Foal Znven ory o 100l Xohle%sec e 25l Toddle 1%pax=lcala=as 100l o o lo lol L.o 0 lgg 0.1 C 2 0 i l 0 0.01 l 0.001l lp',l z is S>~V'\Q'l i h Slei~P))c 3 l>el'>P)a i 5 C toe'~Py Pirnc After Shul"~~(Hrs)l roel Re'cased ApprŽx~c-ec 5oc"ca and Pa=age Zar~t,o 100.50.10.5xlo C C 100\D-]4844, OOl fuel doge, pc ea core 50'L YM noble gasas, aomco.10%.:3, 100l NRC ga-ac-fifty,~":a'lad fa'la=a, par.lal core c"~ercda 100l WE58-1400 gap ac=w', na,or clad f alla"a.10l)at gap,~.10l clad fal'"~.ll C~~, ll KC gap, ll clad f41-o, heat.'aq of 5-10 fuel ass~ides.~010.W,.1a'gap, cl ad f aM"~of 3/4><<l el~ac (36 rods)..Oll AC qep, clad failu"e of a fev red'.1ool coolant release vent-0 apA~)ool coolant fnvenccry release~@~r ra~>>of normal aŽho~c oahkc~s activity~oonm~ns,.ppr 5E HEI," 238N Page 24 BF TI-88 APPENDIX H DETERHINATION OF THE SOURCE OF FISSION PRODUCT RELEASE FROH FISSION PRODUCT NUCLIDE RATIOS The ratios of the fission product nuclides released from either the fuel gap or the molten fuel are significantly different as shown in Table 1 of this Appendix H.Therefore, by the measurement of ratios of fission product nuclide activities in either the postaccident gas or water samples, the source (gap or fuel pellet)of the release may be identified, 1.Calculate fission product nuclide ratios using corrected concentrations of Section 3 of Uorksheet 1: Noble Gases Iodines hluclide Ci/CXe 133 Nuclide Ci/CI 131 Kr-87 Kr-88 Kr-85m Xe-133 1.0 I-134 I-132 I 135 I-133 I-131 1.0 2.Compare ratios above with Table 1 of this Appendix H and note closest inventory ratio agreements:

Nuclide Core Inventor Ratio Fuel Ga Ratio Kr-87 Kr-88 Kr-85m I-134 I-132 I-135 I-133 Yes Yes Yes Yes Yes Yes Yes No hlo No hlo No Wo No Yes'es Yes Yes Yes Yes Yes No No No No No No No 3.Based on the results of Step 2 above, identify dominant core damage mode and record on Plorksheet 2, item 6.hlote: Heasured nuclide r atios similar to those given in Table 1 for"Core Inventory" would be indicative of severe fuel damage (melting), while measured ratios close to that given in Table 1 for"Fuel Gap" would represent a distribution of nuclides typical of the inventory in the inter-pellet fuel gap spaces.0045E OEC 2 3 3985 Page 29 BF TI-88 APPENDIX H (Continued)

Table 1'I RATIOS OF ISOTOPES IN CORE INUENTORY AND FUEL GAP Isotope Half-Life Activity Ratio in Core Inventory Activity Ratio in Fuel Gap Kr-87 Kr-88 Kr-85m Xe-133 76.3 m 2.84 h 4.48 h 5.25 d 0.233 0.330 0.122 1.0 0.0234 0.0495 0.0230 1.0 I-134 I-132 I-135 I-133 I-131 52.6 m 2.3 h 6.61 h 20.8 h 8.04 d 2.30 1.46 1.97 2.09 1.0" 0.155 0.127 0.364 0.685 1.0'Ratio=-noble gas isotope concentration for noble gas nuclides Xe-133 concentration iodine isotope concentration b Rat xo-------------for xodxne nuclxdes 1-131 concentration 0045E 4~(~t v Page 90 BF TI-88 APPENDIX I APPLICATION OF OTHER SIGNIFICANT PARAMETERS TO CORE DAHAGE ESTIMATES Based on radionuclide measurements, an initial assessment of core damage is>possible.

Based on a clarification provided by the NRC, that assessment would appear in a matrix as follows: Dama e Cate or Desi nation Matrix Degree of Degradation Hinor ((10'L)Intermediate (10%-SO%)Major (>SO'L)No fuel damage Cladding failure Fuel Overheat Fuel Helting 10 As recommended by the NRC;, there are four general classes of damage and three degrees of damage within each of the classes except for the no-fuel-damage class.Conseguently, there are a total of 10 possible damage assessment categories.

For example, Category 3 would be descriptive of the condition where between 10 and SO percent of the fuel cladding has failed.Note that the condi'tions representative of more than one category could exist simultaneously.

The objective of the final core damage assessment procedure is to narrow down, to the maximum extent possible, those categories which apply to the actual in-plant situation.

004SE SE4<<em Page 31 BF TI-88 APPENDIX I (Continued)

The initial core damage assessment based on radionuclide measurements will provide one or several candidate categories which most likely represent the actual in-plant conditions, The other parameters should then be evaluated to corroborate and further refine the initial estimate.For example, fission product measurement using the Post-Accident Sampling System (PASS)may indicate the likelihood of Category 4 core damage and, additionally, the potential for fuel overheat and fuel melting (i.e., Categories 5 through 10).Heasurement of hydrogen concentration in the containment volume and the use of the hydrogen correlation provided in Appendix F could be used to verify that extensive cladding damage had ,.occurred.

-Use-of the containment radiation monitor reading along with the correlation provided in Appendix G would verify that a significant fission product release to the containment had occurred, further supporting the initial conclusions.

Additional analysis of the PASS samples for concentrations of Ba, Sr, La, and Ru nuclides and consideration of the relative amounts of these non-volatile fission products released'would indicate if any significant fuel melting had occurred.The flow diagram in Appendix A indicates how the analysis of the other significant parameters relates to the estimation of core damage based on radionuclide measurements.

004SE DEC 23~485 Page 32 BF TX-88 10'0 FUEL MELTDOWN UPPER RCLEAS'E LIMIT SEST ESTIMATE LOWER RELEASE LIMIT J'r/r s'0 10 CLAOOING F AILUAE UP9ER RELEASE LIMIT 0 EST E ST IM a TE LOWER RELEASE LIMIT NOAMAI.SHUT@OWN CONCE N TAATION IN A E ACTOR WA I'E A UI'I'ER LIMIT NOMINA I"9,0~Cilia 0.'I~CiIg 0 I 0.1 ,I~j IO~I~I~~~I s~a CLAOI:Iso f AILURE 10 IO 100 a PU L Afoot C'NN I Figure 1.Relationship Between I-131 Concentration in the Primary Coolant (Reactor Water+Pool Water)and the Extent OE Core Damage in Reference Plant 0045E DEC 23$985 Page 33 BF TI-88 10'Ip FUEL MELTDOWN UPPER REI.EASE LIMIT BEST ESTIMATE LOWER RELEASE I.IMIT V C C tJ<<C I x I tJ 2 EJ i 10 I.p/r r r r r CLACOING F A<LURE l UPPER RELEASE LIMIT OESi ESTIMATE LOWER RELEASE LIMIT 2~~NORMAL SHUTDOWN CQHC EN TRA TIQN IN REACTOR WATER UPPER LIMIT NQMINAI.0.3 I<C<III 0.03 i Cilg 0~0 I~.i~~I l.p~s~i~~I Ip~I i~I~~~4 CL*OOIN<i F AII.UIIE lp I IO'<<FUEL MEI.TDQWN Figure 2.Relationship Between Cs-137 Concentration in the Primary Coolant (Reactor Water+Pool Water)'and the Extent of Core Damage in Reference Plant 0045E a'l 4W~