ML20101S680

Rev 4 to Vol Xiii,Plant Emergency Procedure PEP-03.6.3, Estimate of Extent of Core Damage Under Accident Conditions

ML20101S680
Person / Time
Site:	Brunswick
Issue date:	01/31/1985
From:	Boyer L CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML20101S653	List: ML20101S651 ML20101S669 ML20101S680
References
PEP-03.6.3, PEP-3.6.3, NUDOCS 8502050587
Download: ML20101S680 (74)

Text

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~ ~. _. -, t 1.0 Responsible Individual and Objectives 3 The Radiological Control Director is responsible to the Site Emergency Coordinator for determining the magnitude of potential radioactive l releases to the environment. The Radiological Control Director may delegate the calculational aspects to the Plant Sampling and Analysis Team Leader. The Dose Projection Coordinator and the Accident Assessment Team Leader l should be familiar with this procedure and available for consultation as requested by the Plant Sampling and Analysis Team Leader. 1 [ 2.0 Scope and Applicability This procedure is to be implemented by 'the Site Emergency Coordinato'r or the Radiological Control Director whenever the potential for core damage exists and/or there exists a potential or actual radiological release to the environment (e.g., site or general emergency). This procedure provides information on inventories of reactor full-power radioisotopes in curies and gives methods for comparing actual radioactive liquid and gaseous samples with expected activity levels after a reactor accident based on cesium, noble gases, and iodines. There are several other plant parameters which are measured in the BWR which can provide sufficient informat' ion to confirm the initial core damage estimate based on radionuclide measurements. Containment radiation level provides a measure of core damage, because it { is an indication of the inventory of airborne fission products (i.e., noble gases, a fraction of the halogens, and a much smaller fraction of the particulates) released from the fuel to the containment. Containment hydrogen levels, which are measurable by the PASS or'the containment gas analyzers, provide a measure of-the extent of metal water reaction which, in turn, can be used to estimate the degree of clad damage. [ Another significant parameter for the estimation of core damage is reactor vessel water level. This parameter is used to establish if there 'has been an interruption of adequate core cooling. Significant periods l with the core uncovered, as evidenced by reactor vessel water. level readings, would be an indicator of a situation where core damage is - likely.- Water level measurement would be particularly useful in distinguishing between bulk core damage situations caused by loss of F' adequate cooling to the entire core and localized core damage situations caused by a flow blockage in some portion of the core. i i There are other parameters which may provide an indication that a core i. damage event has occurred. These are main steam line radiation level and reactor vessel pressure. The usefulness of main steam line radiation measurement-is limited because the main steam line radiation monitors are t' l O .S,yo1. x m az,.03. 3 1 mev.. -m, = H% r, - - - - " 1 -. -~_-'"~~~~_"_~v'"MM"V'=, _,. -, ..__..,_m., w-mm:y a v - -y,

v-7 w..._. _ _. f + i p downstream of the main steam isolation Valves'(MSIVs) and would be () unavailable following vessel isolation. Reactor vessel pressure measurement would provide an ambiguous indication of core damage, because, although a high reactor vessel pressure may be indicative of a core damage event, there are many nondegraded core events which could also result in high reactor vessel pressure. There are other measurements besides radionuclide measurements which are obtainable using the PASS which would further aid in estimating core damage. Detection of such elements in the reactor coolant as Sr, Ba, La, and Ru is evidence of fuel melting. These indications could be factored 1 into the final core damage estimate. [. l l 3.0 Actions and Limitations 3.1 Summary of Method i Liquid and gaseous samples will be obtained from the Postaccident Sampling System (PASS)--Liquid from the reactor coolant and/or suppression pool and gaseous'sr.mples from the primary and/or I secondary containment. The samples will be quantitatively analyzed on the appropriate equipment. The results of the above analysis, in addition to containment radiation level, hydrogen analysis, and the core water level history, will be used in the estimation. This procedure follows the General Electric procedure NED0-22215, August

1982, ol] -

List of Exhibits 3.6.3-1 Sequence of Analysis for Estimation of Core Damage 3.6.3-2 Relationships Between Concentration in the Primary Coolant t and the Extent of Core Damage in Reference Plant 3.6.3-3 BSEP to Reference Plant Parameters 3.6.3-4 Core Inventory of Major Fission Products in a Reference j. Plant 3. 6. 3 -5 Hydrogen Concentration for Containment as a Function of Metal-Water Reaction

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3.6.3-6 Percent of Fuel Inventory Airborne in the Containment l 3.6.3-7 Computer Inputs for the PASS Program j 3.6.3-8 Verification of PASS List of Attachments Attachmer.t A Plant Parameter Correction Factors [ Attachment B Inventory Correction Factor Attachment C . Comparison with Reference Plant Data Attachment D Integration on Containment Atmosphere Hydrogen Measurement Into Core Damage Estimate Attachment E Integration of Containment Atmosphere Radiation Measurement Into Core Damage Estimate f i \\ BSEP/Vol.'XIII/ PEP-03.6.3 2 Rev. 4 t 7"Sp?M w ~ ~* ' " ~ ~ ~ ~ * ~ ~ ""'" ~ Y " " ""~ T " ~ ~~?' M v n, --,y z.

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. - _ ~ -. - )- l i ). List of Worksheets i t Worksheet Al Calculation of Isotopic Concentrations Worksheet A2 Calculation of Inventory Correction Factor Worksheet A3 Calculation of Normalized Isotopic Concentrations Worksheet A4 Estimate of Fuel / Cladding Damage Worksheet B1 Determination of Clad Damage from Hydrogen Monitor Reading Worksheet B2 Determination of Fuel Inventory Release Based on Containment Radiation Monitor Realing 3.2 Limitations 3.2.1 Analysis of PASS samples for concentrations of Ba, Sr, La, and Ru and consideration of the relati"e amounts of fission products would indicate if any fuel melt has occurred. 3.2.2 The selection of a sample location should account for the type of event which will determine where the fission products will concentrate. j I 3.2.3 The recommended sampling locations are as follows: Gaseous Event Type Sample Location s I) Nonbreaks Suppression pool atmosphere (e.g., MSIV) -Small breaks Drywell (before depressuri-zation); suppression pool atmosphere (after depressur-ization) L 6 Large breaks (liquid Drywell or steam) in primary [ containment l Large breaks outside Suppression pool atmosphere primary containment f 3.2.4 The recommended sampling location for liquid for all events is the jet pumps as long as there is sufficient j reactor pressure (normally > 50 psig) to provide a sample from that location. If there is not sufficient reactor pressure to allow a sample to be taken from the jet pumps, the sample should be taken from the sample points on the RHR System. 3.2.5 If a jet pump liquid sample is requested at low (< 1%) power conditions for a small break or nonbreak event, recommend to Operations that the reactor water level be i BSEP/Vol. XIII/ PEP-03.6.3 3 Rev. 4 4 """"R " ~ ^ ' ""' ~ ~ ' ~ * ~~

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raised to the level of the moisture separators. This will 7 ) fully flood the moisture separators and will provide a thermally induced recirculation flow path for mixing. 3.3 Actions 3.3.1 Evaluations of Liquid and Gaseous Samples NOTE: The extent of core damage can be determined by comparing the measured concentrations of major l j fission products in either the gas or water samples, after appropriate norealization, with the reference plant data. t 3.3.1.1 The Plant Sampling and Analys'.s Teae Leader should request samples fro the PASS. NOTE: Step-3.1.1.2 through 3.3.1.7 c.m be accomplished using PASS, a compu?.er program developed for use on the Dose Projection Team's IBM Personal Computer. To use the program, the Plant Sampling and Analysis Team Leader should complete Exhibit 3.6.3-7, Computer Inputs for the PASS Program, and give the completed exhibit to the Dose Projection ) Coordinator who will run the program and return the results. Exhibit 3.6.3-8 provides example test cases which can be used to verify that the computer program PASS is working properly. Expected results for known computer inputs are given. These test h: cases should be used to demonstrate the validity of PASS each time the program is initially used. 3.3.1.2 Obtain the samples from the PASS and determine i the concentration of the fission product i (Cyg in water or C in gas as determined in Attachment A using data provided in Exhibit p 3.6.3-3). i 3.3.1.3 Correct the measured concentration for decay to the time of reactor shutdown. Ensure that the measured gaseous activity concentration has been corrected for temperature and pressure difference in the sample vial and the containment (torus) gas phase. NOTE: This is normally included in the quantitative analysis results. BSEP/Vol. XIII/ PEP-03.6.3 4 Rev. 4 t = n-..-. .,~ -..,. -. - n

c_ w c,:. w. ~. 9 v'1 3.3.1.4 Calculate the f'ission product inventory I ib correction factor F per Attachment B and g record on Worksheet A2. '3.3.1.5 Calculate the C and C sing the information yg Ei obtained in Step'3.3.1.2 and the methods in Attachment A and record on Worksheet A1. l 3.3.1.6 Using the correction factors, determined in g Attachments A and *, calculate the normalized I Re.. Ref g, par Attachment C and l } concentration, C,f or C e record on Worksheet A3. 3.3.1.7 Use Exhibit 3.6.3-2 to estimate the extent of Ref fuel or cladding damage using C for Cs-137 and g Ref I-131 and C for Xe-133 and Kr-85. Record data on Worksheet A4. 3.3.2. Evaluation of Metal-Water Reaction and Inventory Release 3.3.2.1 Use Attachment D to determine the percent l 8 . O metal-water reaction. Record data on Worksheet Bl. 3.3.2.2 Use Attachment E to determine the fuel inventory release to the containment. Record data on I Worksheet B2. L f@ 3.3.3 Application of Other Significant Parameters to Core Damage 4: Estimay Section 3.3.1 provides an estimate of core damage based on I5 radionuclide measurements. Based on Step 3.3.1.7, an h

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initial assessment of core damage is made. Based on a ~ clarification provided by the NRC, that assessment would-appear in a matrix as follows: Degree of Minor Intermediate Major f: Desradation (< 10%) (10% - 50%) (>-50%) j i

No fuelldamage

-( 1- ) Cladding failure-2 3 4 'uel overheat - 5 6 7 Fuel melt 8 9 10 As recommended by the NRC, there are four general classes of damage and three degrees of damage within each of the Y' ~ classes except for the "no fuel damage" class. J -\\ BSEP/Vol. XIII/ PEP-03.6.3 5 Rev. 4 ~

) %au m___1.....-..m.--- , ~1 Consequently, there are a total of ten possible damage x_) assessment categories. For example, Category 3 would be i-descriptive of the condition where between 10% and 50% i of the fuel cladding has failed. Note that the conditions 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. The initial core damage assessment based on radionuclide measurement will provide one or several candidate categories which most likely represent the actual in-plant condition. The other parameters should then be evaluated (as identified in Section 3.3) to corroborate and further l refine the initial estimate. 4 For example, fission product measurement using PASS may indicate Category 4 core damage and, additionally, the potential for fuel overheat and fuel melt (i.e., Categories 5 through 10). Measurement of hydrogen in j containment and use of the hydrogen correlation provided in Attachment D is used to verify that extensive clad l damage had occurred. Use of the containment radiation monitor reading along with the correlation provided in l l Attachment E would verify that a significant fission product release to the containment had occurred, further r1 verifying the initial assessment. ^ () Further analysis of the PASS samples for concentrations of Ba, Sr, La, and Ru and consideration of the relative amounts cf fission products released would indicate if any fuel melt had occurred. L Exhibit 3.6.3-1 indicates how the analysis of the other significant parameters relates to the estimation of core [J damage based on radionuclide measurements. i 3.3.4 Consult with the Dose Projection Coordinator and the Radiological Control Director when results of this procedure are determined and repeat this procedure as necessary. 4.0 References + ? Lin, C. C., " Procedure for the Determination of the Extent of Core Damage Under Accident Conditions," NED0-22215, 1982. Letter and Attachment from Mr. D. K. Smith, Service Supervisor - Nuclear, General Electric to Mr. A. C. Tollison, Jr., General Manager, Brunswick Steae Electric Plant, dated November 9, 1979,

Subject:

Radiation Source Term Information. m ~~] BSEP/Vol. XIII/ PEP-03.6.3 6 Rev. 4 i

u 6-c :.:a -. w Letter and Attachments form Mr. T. J. Dente, dhairman - BWR Owner's Group I '{d to Mr. D. G. Eisenhut, Licensing Director - USNRC, dated June 17, 1983,

Subject:

Transmittal of Generic Procedures for Estimation of Core Damage Using Postaccident Sampling System. j 4 I t I i 1 s i

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b -{ EXHIBIT 3.6.3-1 E SEQUENCE OF ANALYSIS FOR I ESTIMATION OF CORE DAMAGE a g e m mb-p Hyd rogen YES Containment YES Water YES-NORMAL OPERATION Analysis Radiation Level MINOR CLAD DAMAGE fConfIrm) fConfIrol IConfIrm) f V H L H LOW { H D NO NO NO - [ M 7' oW Dete rm i ne Core Damage i on Optimum Estimate Sample From PASS

• W Point HIGH NO NO NO

[ t, i Hyd rogen YES Containment YES Weter YES l Ana lysis For Analysis Radiation Level i Ba. Sr. La. Ru m (Contirol fContirm) fContirol MAJOR CLAD DAMAGE YES De te rmi na t ion I FUEL OVERHEAT Or Fission FUEL MELT Product Ratio I f NO t CLAD DAMAGE I POSSIBLE FUEL OVERHEAT NO CORE HELT 1 i 1 4 ? I e .. _...... 7,.

-..w.. ..,.a u ~ ~.~...~.--.._...;:- 2~^.= 2-... i I t 1 EXHIBIT 3.6.3-2 i [. ruft utLTDowN LPPER RELEASE DulT DEST ESTsuATE / / Lowan AstzAst uu T / so' / / / / / / / / / / / l / / l,/ s / / / / } 'E s',/ 7 / w / / p l. / / / / / / / / / i ys / / / i / g 2 1o I, / = I ,/ / I ~ O / / CLAOpeNC rAILURE j i f j vre R RsLaAsa uuir f / sesTesTwAT LOWER RELEAst uutT / / / / i; I / moRuAL smnoonw CONCENTRATION i l / IN REACTOm WATER j / / UPPER uulT: 39A pCWs 800MINAL: 0.7 pCWs / / t; ~ / r- , s.,,,1 ,,,,1 .....1 8.1 1A 10 m l ', c scLAno NG FAILURE r' SA to Soo s pueLusLToowN n' Relationship Between I-131 Concentration in the Primary Coolant (Reactor Water + Pool Water) and the Extent of Core Damage in Reference Plant O BSEP/Vol. XIII/ PEP-03.6.3 9 Rev. 4 ...,r--,,-:--.-..,-! . ~,.. .-_g....

.1 ,e. - A _ R i ___. y, - " 7.m.,Iu e.1. 4. 4 EXHIBIT 3.6.3-2 (Cont'd) (] f.' ._.~...a. .F.U.E L WELTDOWN UPPER AE'LE SE LIMIT ~ SEST ESTIMATE / / LOWER RELEASE LWIT I / / 10 - ,/ / / / / / / / / / / p l / / r i / / / / / / i I .80' / / / l ~::- / / / / / / j / / / l / / / ~ / / 3 t 1 / / / / /,. ' - / / E. 50 :- / f I ] / f / / l ? / / / . ci.AeoiNo e A LumE ,e } f / uprER RELEASE LMIT [, i j f sEsT EsTWATE . f j f / lower RELEASE Lassi f / / / l / / WonuALsNuToowN F / CONCENTRATION N nE AcToa wavun / ./ urFER La m OJ,Cire i / NOMINAL. 0A3pCUs / /- / j ,,,,,,1 r ,,,,,,,1 ,,,,1 ,,,,t ,,-r 0.1 1A 10 100 5 i ~' ^% CLADDING F AILURE 's i r 1A 10 100 l l l- % FUEL MELTDOWN - Relationship Between Cs-137 Concentration in the Primary Coolant (Reactor Water + Pool Water) and the Extent of Core Damage in Reference Plant BSEP/Vol. XIII/ PEP-03.6.3 10 Rev. 4 i ---+%WN MM-@4,- g y-._., _ gg.g g

pws:- x.s - .r=cz.=_w.,_.a.._..,,... -w I / V EXHIBIT 3.6.3-2 (Cont'd) i. d. FutL ME1TDowW ~ uFPEn RELEAss uusi j ~ / sesT EstruATE j LowtR RELEASE LIMIT / ~ d / e / // ~ l. //- / ~ // / ~ ~ // / // / // / // / Sp :* / '/ .l / / i ~ / / / / / / ,/ / / / f, / / / l f l~ f C / / l1A.. p p / eLApoiNo F AilunE f %J. / / / UPPER RELEASE LIMIT / l. / SEST ESTie4 ATE {j d I / LOWER RELEASE LIMIT l 's F /' r f ~ I / / / WORMAL OPERATING .1 y./ CONCENTRATION

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IN ORYWELL l g UPPER LIMIT if * #CWes woMiNAu tr5 cus ,o-2 ......I ,,....I ,......I ...f ~ i,..... o.s 54 to too y % CLADDtNG F AILURE I % FUEL ME LTDOWN Relationship Between Xe-133 Concentration in the Containment Gas (Drywell - Torus Gas) and the Extent of Core Damage in Reference Plant O BSEP/Vol. XIII/ PEP-03.6.3 11 Rev. 4 I i.. .rw w,- ws y,-. -,e e . N ram w w w __ -. .wew-; ,*t

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- m _ = _ - . ~.. _ i ; .. m m.. _..m l EXHIBIT 3.6.3-2 (Cont'd) .a f. d: _ _... FUE LMELTDOWN UPPER RELEASE LustT sEST ESTa' MATE ' LOWER 1tELEASE LIMIT j / l / i / / / i' i / / L 7 / / / sa / / / / ,/ .5 / h l / / / ~ p / / g 7 / / so-5 / ./ / / / / .lT / / m l~ / 8 CLAODING FAILUME O ~ / / UPPER RELE ASE LIMIT /. - / so-2 / I i / / sEsT EsTiMAT: Q.- -p / / LOWER RELE ASE LIMIT / / / [ ~ y-3 7 / NORMAL OPER ATION CONCENTRATION y eN ORYWE LL .j f UPPSR LIMIT: 4 m to" pC/es NOMINAL. 4m10 pGJac i ......i s ,o 100 o.1 1A 10 'r % CLADDING FAILURL - g 14 to too Z % FUEL MELTDOWN Relationship Between Kr-85 Concentration in the Containment Gas (Drywell Torus Gas) and the Extent of Core Damage in Reference Plant BSEP/Vol. XIII/ PEP-03.6.3 12 Rev. 4

L _ _ _._ -,__. m m_s_-_. m _ m. =-m...__--- ,/ t N i-V's ATTACHMENT U y Plant Parameter Correction Factors T e, Fission products measured together for reactor water and suppression pool water or'drywell gas and torus gas. F= BSEP total coolant mass (2.69 x 10' g) w-p. reference plant coolant mass (3.92 x 10' g) p. Q = 0.68622 Pp. Lr, w* hre g I .F= BSEP total containment gas volume (8.11 x 10' cc) g reference plant containment gas volun.e (4 x 10" cc) t 4,.. 0.20275 = i Fission products measured beparately for reactor water and suppression pool b' water or drywell gas and torus gas. L C,g = (cone. in Rx wtr) (Rx water mass) + (cone. in pool) (pool wtr mass) '[ ) reactor water mass + pool water v =-(conc. in Rx water)(2.14 x 10' g) + (cone. in pool)(2.48 x 10' g) ~ 2.69 x 10' g C = (conc. in d m ell) (j m ell gas vol.) + (cone. h tons) (toms gas vold gg drywell gas volume + torus gas volume = (cone. in drywe11)(4.65 x 10' cc) + (conc. in torus)(3.46 x 10' cc) h 8.11 x 10' cc I' e k l (~~) v i BSEP/Vol. XIII/ PEP-03.6.3 13 Rev. 4 l ? t i .m -n.,

p-www3m am. - p. q r / U [ e. t% ""} .W ATTACIDfENT B ~ bb. U. ?? ' 1 l Inventory Correction Factor I,T,Y. ' :_l. inventon in reference plant g F = 7g inventory in operation plant p, -1095x g = 3651 1-e 3 - g% s;; -1 T) -1 T) 1 1 1-e ,e I P) j where: l P) average steady reactor power operated in period j (Wt). ]4 = p T) duration of operating period j (day). j = rs time between th's and of operating period j and the time of the p. T = ~ 3' last reactor shutdown-(day). t 3651 = reference plant Wt. j;f If ths unit operating history is not readily available, use the k following F values (based upon Brunswick plant operations under the same 7 operational constraints): {.' W Nuclide Conservative F x (day -1) 7 I-131 1.34 0.0862 i-Cs-137 1.39 -6.29 x 10 ' l Xe-133 1.46 0.1320 ~ Kr-85 1.51' 1.77 x 10 " I Oh BSEP/Vol. XIII/ PEP-03.6.3 14 Rev. 4 i 1 ( d --~ ~ ~ -.a

^ %5,s..c..awa_ _ __a.a-3. _ p h. 1: H Y... ATTACHMENT C [ f^g ' Q Ml gTQ Comparison With Reference Plant Data

S The extent of core damage can be estimated from the measured fission product l

f concentrations in either the gas or water samples, as described for the reference plant. However, the measured concentration must be corrected for the differences in operation power level, time of operation, primary coolant mass, and containment gas voluma. Ref At

g.

g = e xF x F, ( C,g C,g 7g -OR ~ Ref At j g i xF xF- = C e Ci g gi Il g i Ref-C,g Concentration of isotope i in the reference plant coolant = (yCi/g). 6,- Ref F: C = Concentration of isotope i in the reference plant containment [. gas (yCi/cc). I C,g- = Measured concentration of isotope i in BSEP's coolant (yCi/g). [ See Attachment A. l ~ Measured concentration of isotope i in BSEP's containment gas f..' C = l {. j. (yCi/cc). See Attachment A. c. At k [ i l' I Decay correction to the time of reactor shutdown. e =- Decay constant of isotope 1 (day-1). i = Time between the reactor shutdown and the sample time (days). i-t- =- V Inventory correction factor for isotope i. See Att'achment B. .F = 7f I Containment gas volume correction factor. See Attachment A. F = F, Primary coolant mass correction factor. See Attachment A. l = l BSEP/Vol. XIII/ PEP-03.6.3 15 Rev. 4 n U i V- .-em.--.__._.,-.,-.w&----

yn; - m w w c.a. .,n. i V '( EXHIBIT 3.6.3-3 C. gs C) p'i.c-BSEP TO REFERENCE PLANT PARAMETERS .,r.. p Reference Plant BSEP Reactor Thermal Power 3651 Wt 2436 Wt i Number of Fuel Bundles 748 bundles 560 bundles Total Primary Coolant Mass 3.92 x 10' g 2.69 x 10' g t. (reactor water plus suppression pool water) k. Total Drywell and Torus Gas Space Volume 4.0 x 10 ' cc 8.11 x 10' cc 1 '~ Reactor Water 2.46 x 10' g 2.14 x 10' g Suppression Pool 3.67 x 10' g 2.48 x 10' g Drywell Gas Volume 7.77 x 10' cc 4.65 x 10' cc l Torus Gas Volume 3.25 x 10 ' cc 3.46 x 10' cc f" 2 i r ( ) i v t ht,.

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.... - w m_. .. a .a _ w.._ -.--- ~~.~n.- .L 1r^ EXHIBIT 3.6.3-4 h ' N,)' ff > p}?> ~ Core Inventory of Major Fission Products in a Reference Plant

)J Operated at 3651 MWt for Three Years ta v ~

P' Half-Inventory Major Gamma Ray Energy-Chemical Group Isotope Life

• 10 Ci Intensity - kev (T /d) 8 l

. Noble Gases Kr-85m 4.48 h 24.6 151 (0.753) Kr-85 10.72 y 1.1 514 (0.0044) Kr-87 76.00 m 47.1 403 (0.495) Kr-88 2.84 h 66.8 196 (0.26), 1530 (0.109) Xe-133 5.25 d 202.0 81 (0.365) ['0 I J Xe-135 9.11 h 26.1 250 (0.899) Halogens I-131 8.04 d 96.0 364 (0.812) I I-132 2.30 h 140.0 668 (0.99), 773 (0.762) I-133 20.80 h 201.0 530 (0.86) I-134 52.60 m 221.0 847 (0.954), 884 (0.653) I I-135 6.59 h 189.0 1132 (0.225), 1260 (0.286) Alkali Metals Cs-134 2.06 y 19.6 605 (0.98), 796 (0.85) Cs-137 30.17 y 12.1 662 (0.85) Cs-138 32.20 m 178.0 463(0.307), 1436 (0.76) Tellurium Group Te-132 78.00 h 138.0 228 (0.88) Noble Metals Mo-99 66.02 h 183.0 740 (0.128) Ru-103 39.40 d 155.0 497 (0.89) f Alkaline Sr-91 9.52 h 115.0 750 (0.23), 1024 (0.325) Earths Sr-92 2.71 h 123.0 1384 (0.9) r N Ba-140 12.80 d 173.0 537 (0.254) l . k'_',) Rare Earth Y-92 3.54 h 124.0 934 (0.139) La-140 40.20 h 184.0 487 (0.455), 1597 (0.955) l ~ Ce-141 32.50 d 161.0 145 (0.48) Ce-144 284.40 d 129.0 134 (0.108) r-Refractories Zr-95 64.00 d 161.0 724 (0.437), 757 (0.553) l 1. Zr-97 16.90 h 166.0 743 (0.928) l j 'c O t; h = hour d = day m = month y = year I-r; y ? ?-; r ./^ 4 6 BSEP/Vol. XIII/ PEP-03.6.3 17 Rev. 4 1. o

e - - - =. # % % g e. - m, t,.., _ _. a Y ', l [s i t' ) ATTACHMENT D e r i h %-I \\ Integration of Containment Atmosphere Hydrogen Measurement ? Into Core Damage Estimate {[ {;i The extent of fuel clad damage as evidenced by the extent of metal-water reaction can be estimated by determination of the hydrogen concentration in [ the containment. That concentration is measurable by either the containment p hydrogen monitor or by the Postaccident Sampling System. p'. p A correlation has been developed which relates containment hydrogen A concentration to the percent metal-water reaction for Marks I and II type [G containments. That correlation is shown in Exhibit 3.6.3-5. Note A to that exhibit indicates the major assumptions used in developing the correlation. [2 Note B indicates the method by which Brunswi'ck plant can use the correlation to determine the extent of clad damage. l' I n L s nW, I L: / r, q;' f. bi-( r. I t ./e-, ) BSEP/Vol. XIII/ PEP-03.6.3 18 Rev. 4 i '~' i s l-

1-h, p-ATTACHMENT D (Cont'd) l '- O r IO - EXHIBIT 3.6.3-5 FG.: p.< - t' : t JR g ee - "/ E. ~ j go g '? 0 f :. Ol y g* 2 a g p9 ( 3 e. i y g <0 w g. (a ]- ,. =. t ~ 5 .l 24,- + if y y - 1 i. gg t: V a 4 t I t I t i I I i ~ 0 9 10 20 30 <0 30 ,80 70 00 90, 100 8m m E**F87 5 MAL--MTER REACTMN V } 2 2 l Hydz: ogen Concentration for Marks I and II Containments as a Function of Metal-Water Reaction O BSEP/Vol. XIII/ PEP-03.6.3 19 Rev. 4 T i ? W -: ~' PF *

• a A-

'-m' s 'g' # lh U ' e,, i y' #Q eg' _I 7~ "IP'J T," F E 8 % g j4 3q','f, ~,,' ' " ' ' ' '")

p, ~#.m. m. ~ m %2_m_,i c. m m.m_____., l-t 1-i. ATTACHMENT D (Cont'd) 9" Note A to Exhibit 3.6.3-5 i. Analytical Assumptions [ !' (For Marks I and II Containments) 8 1. Containment Volume =-350,000 ft '2. Number of Bundles = 500 3. Fuel Type = 8 x 8 R 4.~ A11~ hydrogen from metal-water reaction released to containment. 5. Perfect.miding in containment. 6. No depletion of hydrogen (e.g., containment leakage). 7. . Ideal gas behavior in containment. l I i i I LO P h[- F 'b b' s h v ( e k- 'O BSEP/Vol. XIII/ PEP-03.6.3. 20 Rev. 4 <( t c' ,4 s s A i j e j.;

• ^

l j

... ~fy. m i 6 ATTACHMENT D (Cont'd)' [ \\p .) I: Note B to Exhibit 3.6.3-5 Determination of Clad Damage From Hydrogen Monitor Reading Step-1. Obtain containment hydrogen monitor reading in percent. Step 2. Using the curve in Exhibit 3.6.3-5, determine the metal-water reaction for the reference plant, MWR f. Step 3. The metal-water reaction from the actual in-plant conditions (MWR) in determined from the following equation: ~% MWR ='(MWRref) x 500 x V N 350,000 where: N = Number of Bundles = 560 5 V = Total Containment Free Volume, ft' = 2.86 x 10 b .t

,~

l . l: i; i \\ -V- .BSEP/Vol. XIII/ PEP-03.6.3 21 Rev. 4 mwL - - n --,- ,..m.....,n,~ 1

we <- 7.,._. _=_=_ k /% ATTACHMENT E tg Integration of Containment Atmosphere Radiction Measurement 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. This attachment contains a correlation of the containment l radiation monitor dose rate to the percent of fuel inventory airborne in the containment. The purpose of this attichment is to present that correlation i . and provide a method to use that corre:lation to determine the degree of core i . damage-V 1 Exhibit 3.5.3-6 p,covides the results cf a correlation performed for the [. Monticello plant. The key parameters which ' impact the containment dose rate are reactor power and containment volt.me. The method whereby individual plants can apply this correlation is provided in Note A to Exhibit 3.6.3-6. I i.- gm ^ L) i. I. I-p I s. I t ,(~\\ BSEP/Vol. XIII/ PEP-03.6.3 22 Rev. 4 ' ~ ~ ' i f I r,'- m 4= :' y - ry m,- aus jm - - = UT*' 7' F-O ^

w =- - m.:-g m_ zu= _.. _. _ 2,,.. -,,,.a.; _o ; _. ___;___.._.. _ _., _.,. _ l l i ATTACHMENT E (Cont'd) p %.J EXHIBIT 3.6.3-6 Percent of Fuel Inventory Airborne in the Containment D] 2004 Psal 2nventary a 2004 Nehle Cases-

• 254 sodine

? In']

• 2L par.iculates C

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• i a mnio' I

Fuel bN b f~ Inventory Released Approximate Source and Damage Estimate l 100.00 100% TID-14844,100% fuel damage, potentir1 core melt. 50.00 50% TID noble gases, TMI source. l 10.00 10% TID, 100% NRC gap activity, total clad failure, l partial core uncovered. g- [ 3.00 3% TID, 100% WASH-1400 gap activity, major clad failure. t 1.00 1% TID, 10% NRC gap, maximur 10% clad failure. j l' O.10 0.1% TID, 1% NRC gap, 1% clad failure, local beating of t 5-10 fuel assemblies. 0.01 0.01% TID, 0.1% NRC gap, clad failure of 3/4 fuel element (36 rods). ~8 10 0.01% NRC gap clad failure of a few rods. ~ 10 ' 100% coolant release with spiking. 5 x 10 100% coolant inventory release. (3 10 ' Upper range of normal airborne noble gas activity in s j containment. t BSEP/Vol. XIII/ PEP-03.6.3 23 F.av. 4 t k-i r,e m. y.;,,.,- . -...-~.----:----- . e~ ~ ~-w n~~~, --r w i nv -, - - - i

y, p,n own m==.=. uwe m _w = = - - - - ~. - - -,. l ATIACHMENT E (Cont'd)' i I NOTE A to Exhibit 3.6.3-6 Determination of Clad Datage From Containment Radiation Monitor Reading .The' procedure for determination of fraction of fuel inventory released to the ' containment is as follows: Step 1: Obtain containment radiation monitor reading, [R) in rem /hr. Step 2: Determine elapsed time from plant shutdown to the containment [ radiation monitor reading [t] in h,ours. j Step 3: Using Exhibit 3.6.3-6, determine the fuel inventory release for the reference plant.[I),,g -in percent.

Step 4:

. Determine tho' inventory release to the containment [I] using the r i following formula: [I] = [I]ref 1670\\ V I A / (237, 450/ L P where: r.- t...

k]'

P = reactor power level W TBSEP = 2436 Wth)* I th .v, V = total containment free volume, ft" (BSEP = 286, 370 ft ). I 8 I NOTE:

Monitor location within the containment is assumed to have an insignificant' impact on dose rate due to fuel inventory airborne in containment.

~ l4, l J t. i ? j.. l + a BSEP/Vol.:XIII/ PEP-03.6.3 24 Rev. 4 I L -- n m - c, = - 3

m w..-._s~- i H' EXHIBIT 3.6.3-7 A, y Computer Inputs for the PASS Program i Concentration of I-131 in Reactor Water (yCi/ml) Concintration of I-131 in Suppression Pool (yCi/ml)* Concentration of Cs-137 in Reactor Vater (yCi/ml) Concentration of Cs-137 in' Suppression Pool (yCi/ml)* l r Concentration of Xm-133'in Drywell (yCi/cc) f ~ Concentration of Xe-133 in Torus (yCi/cc)** [ Concentration of Kr-85 in Drywell (yCi/cc) Concentration of Kr-85 in Torus'(yCi/cc)** [ g i Time between Reactor Shutdown and Sample Time (days) If time' and. availability permits, attach information necessary for the [; g calculation of_ Inventory Correction Factors (see Attachment B); otherwise, F . conservative de, fault correction factors will be used. [t -Plant Samplina and-Analysis Team Leader: Give completed exhibit to Dose Projection Coordinator. Dose Projection' Coordinator: Enter data into PASS computer program and . provide results to Plant' Sampling and Analysis Team Leader. F. h

• If: unavailable, assume suppression pool activity = 0 pCi/al.-
  • If-unavailable,. assume torus concentration equal to drywell in uCi/cc.

E h. t f L ?. i ,..n. 'BSEP/Vol.'XIII/ PEP-03.6.3 '25 Rev. 4.' ,s 7,, g-y+ a

p7mm www gyy-we--ng.

~. p gy-77., ,.r,. - p -,y w cg g.. ar-m v 7w % ;yp < rpK*#,g;4 ~

w _x c - w. a-_:- ~..~.~.~.w-l1 i. l Y EXHIBIT 3.6.3-8

@)

g V ~K ~ VERIFICATION OF PASS W 'l -f (A Computer program for estimating core damage based on Postaccident Sampling System results) . This exhibit is-intended to provide a means to ensure that PASS, a core damage estimate program designed for the IBM Personal Computer, is working properly. 7 - ' This is demonstrated-by duplicating expected results of known computer inputs. These results can be validated by. comparison to manual calculations for the i .same input. -Sy c'hro different, test cases are presented so that a number of alternate path.s within the program can be tested. The test ' cases with their expected results

follow.

4 -TEST-CASE 1-l /

Computer' Prompt Expected Input j

En$er The Concentration of the Fission Products

. Concentration of-I-131 in Reactor Water (pCi/ml) 1.72E + 3 r -Concentration of I-131 in' Suppression' Pool (pCi/al) 1.49E +.2 ' Concentration of Cs-137 in Reactor Water"(yci/ml) 6.55E + 2 1 Concentration of Cs-137 in Suppression Pool.(UCi/ml) 5.70E + 1 i ? Concentration of Xe-133 in Drywell (pCi/cc) 1.82E + 2 4. ~ 2.41E + 2 -p.- { Concentration.of.Xe-133inTorus(yCi/cc) .V

Concentration of'Kr-85'in Drywell (pCi/cc)-

1.43E + 0 [ 1 ~ Concentration of Kr-85 in Torus (yCi/cc) 1.90E + 0 0 For th's inventory correction factor do you want to use the conservative ' default values'which are bases upon BSEP's operations under the same ~ -operational constraints (YES or NO)? M Enter time between the reactor shutdown and the Sample Time (Days) 1 s .The results'should resemble the printout on the following page. If they do

not,: carefully check your inputs and try the test again.

If the results still ..are not similar, try a backup copy of the program.- If that fails, then seek eprogramming help. Rev. 4' ( 'i 2 BSEP/Vol. XIII/ PEP-03.6.3 : 26- ,N: a Q ', o M'r:L - s Si 3,w - 'N L'&~~ ~":*~"&*WW '"~"~~~~WYn!=.'- ' s' m._,.

EXHIBIT 3.6.3-8 (Cont'd)* ESfIMATE THE EXTENT OF CORE DAMAGE UNDER ACCIDENT CONDITIONS DATE: 0? 28-1984 TIME: 7,:21:27 The concentration of the fission products are: I-131 in Reactor Water 1.72E + 3 pCi/ml I-131 in Suppression Pool 1.49E + 2 pCi/ml .r Cs-137 in Reactor, Water 6.55E + 2 pCi/ml Cs-137 in Suppression Pool 5.70E + 1 pCi/ml Xe-133 in Drywell Air. 1.82E + 2 pCi/cc Xe-133 in Torus Air 2.41E + 2 pCi/cc Kr-85 in Drywell Air 1.43E + 0 pCi/cc Kr-85 in Torus Air 1.90E + 0 pCi/cc Time between the reactor shutdown and the sample time is: 2 days 0 The Concervative Default values of the Inventory Correction Factors were used. Estimate of fuel / cladding damage Primary Coolant Analysis t' Nuclide CwREF (pCi/ml) % -Cladding % Fuel Failure Meltdown I-131 3.00E + O2 69.00 1.35 i Cs-137 1.00E + O2 64.54 4.27 t Containment Gas Analysis Y Nuclide CwREF (pCi/ml) % Cladding % Fuel Failure Meltdown Xe-133 7.99E + 01 53.26 1.84 Kr-85 5.00E - 01 56.35 1.92 i BSEP/Vol. XIII/ PEP-03.6.3 27 Rev. 4

, ~ - --- ~.. _ _ -,,~. . _ g-Mn m

a. -.. ua =.

w.: x - _. a. i EXHIBIT 3.6.3-8 (C'ont'd)' ,- m ( TEST CASE 2 Computer Prompt Expected Input Enter The Concentration of the Fission Products Concentration of I-131 in Reactor Water (pCi/ml) 1.35E + 3 Concentration of I-131 in Suppression Pool (pCi/ml) 1.18E + 2 Concentration of Cs-137 in Reactor Water (pCi/ml) 1.17E + 2 Concentration of Cs-137 in Suppression Pool (pCi/ml) 1.02E + 1 Concentration of Xe-133 in Drywell (UCi/cc) 1.84E + 2 Concentration of Xe-133 in Torus (pCi/cc) 2.45E + 2 Concentration of Kr-85 in Drywell (VCi/cc) 2.91E - 1 Concentration of Kr-85 in Torus (pCi/ce) 3.86E - 1 For the inventory correction factor do you want to use the conservative default values which are bases upon BSEP's operations under the same i operational constraints (YES or NO)? NO Enter time between the reactor shutdown and the Sample Time (Days)? 2 Enter number of Operating Periods from the unit operating history? 3 [ For period number (1) enter: Average steady reactor power operated in this period (MWT)? 1000 Duration of this operating period (days)? 60 Time between the end of this operating period and the time of the most recent reactor shutdown (days)? 254 i. l-For period number (2) enter: f, Average steady reactor power operated in this period (MWT)? 2000 f Duration of this operating period (days)? 200 Time between the end of this operating period and the time of the most recent reactor shutdown (days)? 44 For period number (3) enter: Average steady reactor power operated in this period (MWT)? 3000 Duration of this operating period (days)? 14 Time between the end of this operating period and the time of the most recent reactor shutdown (days)? 0 The results should resemble the printout on the following page. If they do not, carefully check your inputs and try the test again. If the results still are not similar, try a backup copy of the program. If that fails, then seek '] programming help. BSEP/Vol. XIII/ PEP-03.6.3 28 Rev. 4 i .n. ---..-----,.--.m._

-c ---,.rw.___....._.-.---- p. \\ [1 EXHIBIT 3.6.3-8 (C'ont'd)' g [I'~ ESTIMATE THE EXTENT OF CORE DAMAGE UNDER ACCIDENT CONDITIONS y DATE: 03-28-1984 TIME: 13:27:17 ] (+- The concentration of the fission products are: I-131 in Reactor Water 1.35E + 3 pCi/ml I-131 in Suppression Pool 1.18E + 2 pCi/ml Cs-137'in Reactor Water 1.17E + 2 pCi/ml Cs-137.in Suppression Pool 1.02E + 1 pCi/ml Xe-133 in Drywell, Air 1.84E + 2,yCi/cc }- Xe-133 in Torus Air 2.45E + 2 pCi/cc Kr-85 in Drywell Air 2.91E - 1 pCi/cc Kr-85 in Torus Air 3.86E - 1 pCi/cc L . Time between the reactor shutdown and the sample time is: 2 days The Inventory Correction Factors were calculated from the following: f

• Period No.

Operation Time Time Between-Period Average Power r (days) & Last Shutdown (days) (MWt) P ( i.^ -V 1 '60 254 1000 ). 2-200 44 2000 3 14 0 3000 i. t Estimate of Fuel / Cladding Damage Primary Coolant Analysis [3 p-Nuclide CwREF (pCi/ml) % Cladding % Fuel Failure Meltdown [ I-131 3.00E + O2 69.02 1.35 I Cs-137 9.99E + 01 64.49 4.27 ) l'. -Containment Gas Analysis Nuclide CwREF (pC1/ml) % Cladding % Fuel Failure ' Meltdown Xe-133 8.00E + 01-53.30 1.84 l Kr-85 5.00E - 01 56.40 1.92 'x l BSEP/Vol. XIII/ PEP-03.6.3 29 Rev. 4 I .. n n -. --~-~.- n.--

L t e WORKSHEET Al' t.1c ~ CALCULATION OF ISOTOPIC CONCENTRATIONS IN PRIMARY WATER AND SUPPRESSION POOLWATER(Cw)ANDDRYWELLGASANDTORUSGAS(Cgy References Section 3.3.1.2 Section 3.3.1.5 Attachment A f Exhibit 3.6.3-3 g (pCi/ml) = (Concentration Rx H O)g (0.08') + [ Cw 2 (Cs18', I281)- (Concentration Suppression Pool H O)g (0.92) l 2 = + (pCi/ml) pCi/m1 = Cs 181 pCi/ml g and = y i Cgg (pCi/ml) = (Concentration Drywell)g (0.57) + (Concentration Torus)g (0.43) { (Xe*88, Kr") ,s = + (pCi/cc) pCi/cc l = Xe pCi/cc and = Kr F L i i \\ BSEP/Vol. XIII/ PEP-03.6.3 30 Rev. 4 l ~ ' ~ ~ - ~ ~ - -*~"~~~~: -~

• x; ' r ' ' - *** ~: ~ ~ ~

.? -*?vr: m 1 *:~s r= 19 tvev=~~r ..-g.,w-e -

t..p t.==i== m p hg t>. WORKSHEET Ai p- })g ,s, (j ' j CALCULATION OF INVENTORY CORRECTION FACTOR (FI 1 hNh References Section 3.3.1.4 Attachment B g Exhibit 3.6.3-4 (: 'P = MW T.= Days j. thermal j 1 f< Tj = Days Ag= Days r FI = 36'51'(1 - e ~1"' i) f i AT I) P) (1 - e - 1 j)(e-1 T ') ij (Cs ) -f 1 = o i-(I181) (Xe!'8) 3_ .;- p' 'N. /. (Kr) [ t i f

5..

b r 4, T a

t.,

i. t-e BSEP/Vol. XIII/ PEP-03.6.3 31 Rev. 4 ,q \\ l s..e I i .,.~ __ _.._.__.-_.,_..m

LJ w ks u ' bm 2 D- .e_ a .m it' -s w. r 1_ i WORKSHEET A3' L: p .b[^.L E6., CALCULATION OF NORMALIZED ISOTOPIC CONCENTRATIONS IN PRIMARY WATER AND W* "I SUPPRESSION POOL VATER (Cw " ) AND DRWELL GAS AND TORUS GAS (Cg l j '., References j Section 3.3.1.6 NOTE: For BSEP, i"I Attachment C Fw = 0.68622 Worksheet Al Fg = 0.20275 Worksheet A2 6-t- [' ~ Ref. lit = Cw e x FI x Fw Cwg ^ g g (Cs287, 181) pCi/m1 = 187 Cs pCi/mlyis2 Ref lit Cg e x FI x Fg ( Cg = g g f (Xe188,- Kr') l pCi/cc Xe '8 l = VCi/cc Kr' r - 1,. ,L i: - p k r A 4 (] .BSEP/Vol. XIII/ PEP-03.6.3 32 Rev. 4 \\_.) "*'I

• NT 2N"M1 8 t'S e-%e*"*'
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t___ n WORKSHEET A4 ESTIMATE OF FUEL / CLADDING DAMAGE References Section 3.3.1.7 Exhibit 3.6.3-2 Worksheet A3 Primary Coolant Analysis % Cladding % Fuel Isotope Cw, "I(pCi/ml) Failure Meltdown l ^ 881 + 1 1 i Cs [ 18' i Containment Gas Analysis % Cladding % Fuel Isotope Cg Ref(vCi/ml) Failure Meltdown 188 Xe 88 Kr r 4 9 s BSEP/Vol. XIII/ PEP-03.6.3 33 Rev. 4 .,v ._;-n -. ..,,.,.r.nn--,~~--------,-------.--..,-,~.c. v~~-

-.._,.,..m _y.-.-- ~ l-' f m WORKSHEET B1 [ -( j!:;-4 DETERMINATION OF CLAD DAMAGE FROM HYDROGEN MONITOR READING 1 . :7 References Section 3.4.1 Attachment D ' Exhibit 3.6.3-5 m Containment Hydrogen Monitor Reading: MWR ref { t Calculate % MWR: l'. i % MWR = (MWR ref)(0.73) T- 'I h-i I' 4 Y- ~ BSEP/Vol. XIII/ PEP-03.6.3 34 Rev. 4 \\ -., - ~, -

a r ..~ - - - ~. ~ WORKSHEET B2 DETERMINATION OF FUEL INVENTORY RELEASE BASED ON CONTAINMENT RADIATION MONITOR READING References Section 3.4.2 Attachment E Exhibit 3.6.3-6 Containment Radiation Monitor Reading: rem /hr Time from Shutdown to Monitor Read'ing: hrs [I] ref (Reference Fuel Inventory Release, %, from Exhibit 3.6.3-6) I (Actual Fuel Inventory Release) = [I] ref

• 0.827

= f l Li f i BSEP/Vol. XIII/ PEP-03.6.3 33 Rev. 4 J ..;,,,.,n. r-..,v,n - -, - - ~ - - - - ~ ~ - - -, - - - h

.:,-~3 - m ne sn.r m n.._,:_. x,. - . a a.a w - -a... -.-~~. - - - f e i I h ..p E CAROLINA POWER & LIGHT COMPANY BRUNSWICK STEAM ELECTRIC PLANT UNIT 0 L { l: ESTIMATE OF THE EXTENT OF CORE DAMAGE UNDER ACCIDENT CONDITIONS PLANT EMERGENCY PROCEDURE: PEP-03.6.3 i VOLUME XIII i t/ Rev. 004 { ,.c e i-4 1 f Approved By: Date: I 29 IS' GeneralManapf/ / Director - Administr Kive Support 4 ( one. .e-, .-e-~~~.---.n~-*.

[n.,, ...:~_ , -.w ..n.-- -u - a w " - ~~ ~ - ~ - - - - ~ - - - ry L t /m LIST OF EFFECTIVE PAGES. p - ti 1j y ,s E!c:, ", PEP-03.6.3 -g. k: ' Page(s) Revision l3 1-35 4 i kn. i 6 t t 73 1 t x,_./ 1 i i t, l. t ( ) \\~ ' '/ BSEP/Vol. XIII/ PEP-03.6.3 i Rev. 4 ,w

m_ g . =- -c-1.0 Responsible Individual and Objectives i-e The Radiological Control Director is responsible to the Site Emergency i-Coordinator for determining the magnitude of potential radioactive [ releases.to the environment. The Radiological Control Director may ' delegate the calculational aspects to the Plant Sampling and Analysis Team Leader. The Dose Projection Coordinator and the Accident Assessment Team Leader l should be familiar with this procedure and available for consultation as requested by the Plant Sampling and Analysis Team Leader. .2.0 -Scope and Applicability This proceliure is to be implemented by 'the Site Emergency Coordinator or the Radiological Control Director whenever the potential for core damage exists and/or there exists a potential or actual radiological release to the environment (e.g., site or general emergency). This procedure provides information on inventories of reactor full power radioisotopes in curies and gives methods for comparing actual radioactive liquid and gaseous samples with expected activity levels after a reactor accident based on cesium, noble gases, and iodines. Thors are several other plant parameters which are measured in the BWR which can provide sufficient informat' ion to confirm the initial core damage estimate based on radionuclide measurements. t Containment radiation level provides a measure of core damage, because it f. ~ is an indication of the inventory of' airborne fission products (i.e., noble gases, a fraction of the halogens, and a much smaller fraction of the particulates) released from the fuel to the containment. Containment hydrogen levels, which are measurable by the PASS or'the containment gas at.alyzers, provide a measure of the extent of metal water reaction which, in turn, can be used to estimate the degree of clad damage. l./, 4 Another significant parameter for the estimation of core damage is reactor vessel water level. This, parameter is used to establish if there has been an interruption of adequate core cooling. Significant periods I with the core uncovered, as evidenced by reactor vessel water level i; readings, would be an indicator of a situation where core damage is likely. Water level measurement would be particularly useful in distinguishing between bulk core damage situations caused by loss of L' L

adequate cooling to the entire core and localis.ed core damage situations caused by a flow blockage in some portion of the core.

J. There are other parameters which may provide an indication that a core damage event has occurred. These are main steam line radiation level and reactor vessel pressure. The usefulness of main steam line radiation measurement is limited because the main steam line radiation monitors are. ? O .BSEP/Vol. XIII/ PEP-03.6.3 1 Rev. 4 m - -gr pgrg,.., m. ,, e r'3T;:ia.- 7 ~ ~ ; v m,~ m 1,

• • awr 7 en Ct';; *" n* *-"*'---~,Tt

-2 1 o _m g p downstream of the main steam isolation v'alves'(MSIVs)'and would be Q unavailable following vessel isolation. Reactor vessel pressure {- measurement would provide an ambiguous indication of core damage, because, although a high reactor vessel pressure may be indicative of a l.' core damage event, there are many nondegraded core events which could also result in high reactor vessel pressure. There are other measurements besides radionuclide measurements which are obtainable using the PASS which would further aid in estimating core damage. Detection of such elements in'the reactor coolant as Sr, Ba, La, and Ru is evidence of fuel melting. These indications could be factored into the final core damage estimate. l-3.0 Actions and Limitations j 3.1 Summary of Method Liquid and gaseous samples will be obtained from the Postaccident Sampling System (PASS)--Liquid from the reactor coolant and/or suppression pool and gaseous' samples from the primary and/or i secondary containment. The samples will be quantitatively analyzed on the appropriate equipment. The results of the above analysis, in addition to containment radiation level, hydrogen analysis, and the core water level history, will be used in the estimation. This procedure follows the General Electric procedure NEDO-22215, August 1982. i-List of Exhibits 3.6.3-1 Sequence of Analysis for Estimation of Core Damage 3.6.3-2 Relationships Between Concentration in the Primary Coolant and the Extent of Core Damage in Reference Plant

F 3.6.3 BSEP to Reference Plant Parameters

.3.6.3-4 Core Inventory of Major Fission Products in a Reference Plant l 3.6.3-5 Hydrogen Concentration for Containment as a Function of' [ Metal-Water Reaction t 3.6.3-6 Percent of Fuel Inventory Airborne in the Contair. ment .l 3.6.3 Computer Inputs for the PASS Program i-3.6.3-8 Verification of PASS List of Attachments TAttachment A Plant Parameter Correction Factors Attachment B-Inventory Correction Factor Attachment C Comparison with Reference Plant Data Attachment-D Integration on Containment Atmosphere Hydrogen Measurement Into Core Damage Estimate Attachment E Integration of Containment Atmosphere Radiation . Measurement Into Core Damtge Estimate G i k d ' BSEP/Vol. XIII/ PEP-03.6.3 2 Rev. 4 ~ l .=w mv==r=mw& -.~ .,.n -m. x~/-~n,.mm .+

w. t } i I List of Worksheets Worksheet Al Calculation of Isotopic Concentrations Worksheet A2 Calculation of Inventory Correction Factor Worksheet A3 Calculation of Normalized Isotopic Concentrations Worksheet A4 Estimate of Fuel / Cladding Damage Worksheet B1 Determination of Clad Damage from Hydrogen Monitor Reading Worksheet B2 Determination of Fuel Inventory Release Based on Containment Radiation Monitor Reading

3.2 Limitai

ions i I 3.2.1 Analysis of PASS samples for concentrations of Ba, Sr, La, j and Ru and consideration of the relative amounts of fission products would indicate if any fuel melt has occurred. 3.2.2 The selection of a sample location should account for the type of event which will determine where the fission products will concentrate. l 3.2.3 The recommended sampling locations are as follows: Gaseous Event Type Sample Location i Nonbreaks Suppression pool atmosphere (e.g., MSIV) t Small breaks Drywell (before depressuri-zation); suppression pool atmosphere (after depressur-I ization) j Large breaks (liquid Drywell j or steam) in primary containment l~ Large breaks outside Suppression pool atmosphere primary containment 3.2.4 The recommended sampling location for liquid for all f events is the jet pumps as long as there is sufficient reactor pressure (normally > 50 psig) to provide a sample from that location. If there is not sufficient reactor pressure to allow a sample to be taken from the jet pumps, the sample should be taken from the sample points on the RHR System. 3.2.5 If a jet pump liquid sample is requested at low (< 1*.) power conditions for a small break or nonbreak event, recommend to Operations that the reactor water level be s BSEP/Vol. XIII/ PEP-03.6.3 3 Rev. 4 .g--- t 1 p g % 'k 8 %. g-

i I ( n raised to the level of the mois'ture separators. This will () fully flood the moisture separators and will provide a r thermally induced recirculation flow path for mixing. 3.3 Actions. 3.3.1 Evaluations of Liquid and Gaseous Samples NOTE: The extent of core damage can be determined by comparing the measured concentrations of major l fission, products in either the gas or water samples, after appropriate normalization, with i the reference plant data. { 3.3.1.1 The Plant Sampling and Analysis Team Leader should request samples from the PASS. NOTE: Steps 3.3.1.2 through 3.3.1.7 can be accomplished using PASS, a computer program developed for use on the i Dose Projection Team's IBM Personal l' Computer. To use the program, the Plant Sampling and Analysis Team Leader should complete Exhibit 3.6.3-7, Computer Inputs for the PASS Program, and give the completed exhibit to the Dose Projection o ) Coordinator who will run the program and return the results. Exhibit t 3.6.3-8 provides example test cases which can be used to verify that the computer program PASS is working properly. Expected results for known {; computer inputs are given. These test cases should be used to demonstrate the validity of PASS each time the program is initially used. i~ i } 3.3.1.2 Obtain the samples from the PASS and determine i the concentration of the fission product i (Cd in water or C in gas as determ hed in gg Attachment A using data provided in Exhibit 3.6.3-3). 3.3.1.3 Correct the measured concentration for decay to the time of reactor shutdown. Ensure that the measured gaseous activity concentration has been corrected for temperature and pressure difference in the sample vial and the containment (torus) gas phase. NOTE: This is normally included in the quantitative analysis results. BSEP/Vol. XIII/ PEP-03.6.3 4 Rev. 4 f re. -**rg> geni.isgepepnpow.we ., eo-t a e o w + ewe =9. e===e e mes e gar m4 ,4 4 wa <. p.s ys, s e w m. Wg*9* g ' Ngt. '

+ww a w w=.. ~ _; + ~.., h-6,g e d.9 3.3.1.4 Calculate the fission product inventory -correction factor F per Attachment B and l } yg record on Worksheet A2. p_ 3.3.1.5 Calculate the C and C using the information. yg gg obtained in Step 3.3.1.2 and the methods in Attachment A and record on Worksheet A1. g 1 3.3.1.6 Using the correction factors, determined in g Attachments A and B,-calculate,the normalized I i Ref Ref gg, per Attachment C and l concentration, C,g or C t record on Worksheet A3. 3.3.1.7 Use Exhibit 3.6.3-2 to estimate the extent of l Ref fuel or cladding damage using C,g for Cs-137 and Ref 1, I-131 and C for Xe-133 and Kr-85. Record data. gg on Worksheet A4. g 3.3.2 Evaluation'of Metal-Water Reaction and Inventory Release L 3.3.2.1 Use Attachment D to' determine the percent (' metal-water reaction.. Record data on [ Worksheet Bl. -l 313.2.2 Use Attachment E to determine the fuel inventory release.to the containment. Record data on g Worksheet B2. 3.3.3' Application of Other Sinnificant Parameters to Core Damane p Estimate 'Section 3.3.1 provides an estimate of core damage based on radionuclide measurements. Based on Step 3.3.1.7, an -initial assessment of core damage is'made. Based on a 5 clarification provided by the NRC,that assessment would appear in a matrix as follows: Degree of Minor Intermediate Major Dearadation' (< 10%) (10% - 50%) (> 50%) No fuel' damage ( 1-Cladding failure 2 3 4 ' Fuel overheat 5 6 .7 Fuel melt 8 9 10 As ree-adM by the NRC, there are four general classes of damage sad three degrees of damage within each of the hJ ' classes except for the."no fuel damage" class. l t h '38EP/Vol.:XIII/ PEP-03.6.3' 5 Rev. 4 p1 -

_ ___m _ am..% ae w_

p. om%sh

( Consequently, there are a total of ten possible damage assessment categories. For example, Category 3 would be [ J descriptive of the condition where between 10% and 50% of the fuel cladding has failed. Note that the conditions 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. The initial-core damage assesserent based on radionuclide ( measurement will provide one er several candidate categories which most likely represent the actual in-plant condition. The other parameters should then be evaluated (as identified in Section 3.3) to corroborate and further l refine the initial estimate. .i For example, fission product measurement using PASS may indicate Category 4 core damage and, additionally, the potential for fuel overheat and fuel melt (i.e., Categories 5 through 10). Measurement of hydrogen in l containment and use of the hydrogen correlation provided in Attachment D is used to verify that extensive clad l damage had occurred. Use of the containment radiation monitor reading along with the correlation provided in Attachment E would verify that a significant fission product release to the containment had occurred, further verifying the initial assessment. ( Further analysis of the PASS samples for concentrations of Ba,' Sr, La, and Ru and consideration of the relative i I amounts of fission products released would indicate if any fuel melt had occurred. l b. Exhibit 3.6.3-1 indicates how the analysis of the other significant parameters relates to the estimation of core j damage based on radionuclide measurements. 3.3.4 - Consult with the Dose Projection Coordinator and the Radiological Control Director then results of this procedure are determined and repeat this procedure as necessary. '4.0 References p I l Lin, C. C., " Procedure for the Determination of the Extent of Core Damage Under Accident Conditions," NEDO-22215, 1982. Letter and Attachment from Mr. D. K. Smith, Service Supervisor - Nuclear, General Electric to Mr. A. C. Tollison, Jr., General Manager, Brunswick ' Steam Electric Plant, dated November 9,1979,

Subject:

Radiation Source Ters Information. O 88EP/Vol. XIII/ PEP-03.6.3 6 Rev. 4 I -t ~.

pz. v. Letter and Attachments form Mr. T. J. Dente, Caairman - BWR Owner's Group to Mr. D. G. Eisenhut,- Licensing Director - USNRC, dated June 17, 1983,

Subject:

Transmittal of Generic Procedures for Estimation of Core Damage Using Postaccident Sampling System. f O t !i' I 'D 'b BSEP/Vol. XIII/ PEP-03.6.3 7 Rev. 4 l _. -. - - -.... - -. -. - - - - - - ~ ~ ~ - - - - - - -

g C(,r 8,) s EXHIBIT 3.6.3-1 SEQUENCE OF ANALYSIS FOR f ESTIMATION OF CORE DAMAGE f r Cn V m? t Hydrogen YES Containment YES Water YES NORMAL OPERATION Analysis Radiation Level MINOR CLAD DAMAGE i iConfIrei iconfI rm) fConfIrol [ H H LOW H Q M M M m7 i o F Determine Core Damage 'e m Optimum Estimate Sample From PASS W Point t HICH NO NO NO Hyd rogen YES Containment YES Water YES l Analysis For l ~* Analysis Radiation Level i Ba. Sr. La. Ru I fConrirei iconfIrm) fConfIrm) m 1 r k MAJOR CLAD DAMACE YES Determination FUEL OVERHEAT of Fission FUEL MELT Product Ratio 3 i NO CLAD DAMAGE POSSIBLE FUEL OVERHEAT NO CORE MELT 6 b. I n l .k 9 I i ~ --.-g. -e ,.n,- ,.-,w

wuu.n.Uw - wa.. : ?... ^ ~ ^ ^ .. ~ -. - >. EXHIBIT 3.6.3-2 k FUtt utLTDoww 1 AELEASE UM87 l SEsf ESTIMATE / / toWEn n4 LEASE UMIT / n" / / / / i ~ / ? t / / / / / / s' / / f / / p/ / / f 1# /,/ 't 7 i / / / / / p / / / t / / / \\ / / / I / / E / / t .i g to r ,e

• /

I ~ i ) / I ~ O / cLA00mC F AlLUME / j Iy j veren metaAS,8 UMET f / sesta:Ti.4AT / LOWen naLaAsa uurt F- / f / / / I f wonuAL swurpoWw + coweswTnATsow y m anActom wavam f /- uppen LasrT: seaacus woumau e.7,cus / / / t-l ...A.... t .......I .....I ......1 l e,, e.1 ts to too ', c 5 CLAcomG PAILunt r' ta to too E PUGL MELTDoWw L' Relationship Between I-131 Concentration in the Primary Coolant (Reactor Water + Pool Water) and the Extent of Core Damage in Reference Plant O BSEP/Vol. XIII/ PEP-03.6.3 9 Rev. 4 'y w - g a3g '""W

'2-L - 14.% m _ w -s E. EXHIBIT 3.6.3-2 (Cont'd) ~ so" = LMPEREE'LE SE LIMIT SE5T ESTIMATE / ~ / = t.owER RELEASE UMIT 3 / i-10 / / / / / / ~ / / / p / / c.' / p p ~ / / r / / / n3 / / / . 8'~:- / / / / / / ~ / / / / / / / / / i / / f h / / y / / -5 /. / ,/ ( E 10 -l ~ :- / ,/ - p ~ l/ a /^

{

/ ~ i Ct. ADDING F A LUME O8 j uprER nELEAst Liuit f f sEst Est TE f / i.owER RELEASE uu T f / / ~ / / wonuALswurooww f E' / cowcEwTnAviou fN RE ACTom WATER / i / l UPPER LIMIT: e.3 pOfs l / WOMINAL. 023 pCUs l / l / l / !? ,,,,,,1 ,,,,,,,1 ,o-2 ,,,,1 ,,,,,I i o.1 1.0 10 100 I , :s cLAnotwo rAiLuRE la 10 100 I l ', c s putt uELTooww -- - -- l Relationship Between Cs-137 Concentration in the Primary Coolant (Reactor Water + Pool Water) and the Extent of Core Damage in Reference Plant

o BSEP/Vol. XIII/ PEP-03.6.3 10 Rev. 4 l

i 6

pw p TNj**A N. De.. n-i EIHIBIT 3.6.3-2 (Cont'd) Y. Ef FUEL MELTDoWW ,/ urrEr,REtaASE uM T l DEST ESTIMATE / j / l LOWER RELEASE LIMIT / / l ,,;s / = t / L. // ~ // / ~ // / // / // / // / // / Sp :- // / I // / i ~ l / ,/ / e l l / / ~ / / i ', l so / l / / 5 / l ~ E ,/ y / ~ f f / CLAODING F Alluf":E / 7 / UPPER RELE ASE LIMIT / 2 ""/ ,/ 5# sEst EsTiuaTE .,/ LOWER RELEASE LIMIT l ? j' r / / ~ s ./ / WORM AL OPER ATING ?.1 ::= / I "/ CONCENTRATION IN ORYWELL f UPPER LIMIT: 30"#pCWee NOMINAL: 50- 5,gg,, l I.. - ,,,,,,,1 ,.....I ,,,,,,1 ,,,,,,,1 {' so-2 ,,,,,,i, as sa to sas b % CLADDING F AILURE

I 5

I i sa to 180 I ,2 E FUEL MELT 0oWN Relationship Between Xe-133 Concentration in the Containment Gas (Drywell Torus Gas) and the Extent of Core Damage in Reference Plant

O BSEP/Vol. XIII/ PEP-03.6.3 11 Rev. 4 t

r. __m,. I

w a._ - -,:. =., - =. .. _ n - _ =.. .._ __. _. =,, _ e 1 1 EXHIBIT 3.6.3-2 (Cont'd) s O ll", f gf, _..,... FUE L MELTDOWN UPPER RELEASE LlutT l BEST ESTs'uATE ~ LOWER RELEASE LIMIT j / SO :- / / / / \\ / / / t p / / $ th / / f / / / E / p i- / / p / / / / / 5 / / to-5 i / / lT / / ~ / / f' / 3 CLAODING F A1LUME p / / UPPER RELE ASE LIMIT 10-2 / / ? / / sEsT E TiMATE ~ "~ / / ~ / LOWER RELE ASE LIMIT 1 / / / t.. ~ 10"3 / NORMAL OPERATION / CONCENTRATION ll / eN DRYWELL g .. j / UPPCR LIMIT: 4a10-5seCi/se NOMINAL. 4 a10 asCuse .I' ,,,,I ,,,,,,,1 ,,,,,,1 t .,,,,,f to d 0.1 1.0 10 100 --{* 'T % CLADDING FAILURE - 14 to too j '= % ruEL uELTDoWN - Relationship Between Kr-85 Concentration in the Containment Gas (Drywell Torus Gas) and the Extent of Core Damage in Reference Plant O BSEP/Vol. XIII/ PEP-03.6.3 12 Rev. 4 9 J ~.-

py m, ow g - ,yr. i 0 ATTACIDfENT A Plant Parameter Correction Factors J# 1 ;'I Fission products measured together for reactor water and suppression pool .' water or drywell gas and torus gas. F,= ~BSEP total ~ coolant mass (2.69 x 10' a) reference plant coolant mass (3.92 x 10' g) =.0.68622 g i. i s s BSEP total containment gas volume (8.11 x 10' cc) F = 8 reference plant containment gas volume (4 x 10" cc) i = ~0.20275' Fission products measured separately for reactor water and suppression pool water or drywell gas and torus gas. p C,g = (conc. in Rx wtr) (Rx water mass) + (cone. in pool) (pool wtr mass) [ reactor water mass + pool water t, i -= (conc. in Rx water)(2.14 x 10' a) + (conc. in pool)(2.48 x 10' a) [. 2.69 x 10' g p .C = (cone. in drywell) (drywell nas vol.) + (cone. in torus) (torus nas vol.) g drywell gas volume + torus gas volume i. = (conc. in drywe11)(4.65 x 10' cc) + (cone. in torus)(3.46 x 10' cc) !~' 8.11 x 10' cc t l. I I' i i C. 7 i BSEP/Vol.-XIII/ PEP-03.6.3 13 Rev. 4 f i i m.

.L ~ - p._ . _ _ _ ~

9,p. k p. {@ ATTACHMENT B G Inventory Correction Factor t;- (M , r-inventory in reference plant i F = g inventory in operation plant -1095k g = 3651 1-e o e ij ij h ~ 1-e ,e p I P) j where: average steady reactor power operated in period j (Wt). P = 3 r T). duration of operating period j (day). l = T) time between the end of operating period j and the time of the = last reactor shutdown (day). 3651 = refere. ace plant Wt. p ~V If the unit operating history is not readily available, use the following F values (based upon Brunswick plant operations under the same 7 operational constraints): } I' O Nuclide Conservative F 1 (day -1) 7 I-131 1.34 0.0862 ~ Cs-137 1.39 6.29 x 10 ' Xe-133 1.46 0.1320 I ~ Kr-85' 1.51 1.77 x 10 ' e k nV BSEP/Vol. XIII/ PEP-03.6.3 14 Rev. 4 I t i .. m w e ee w.estw : ^- - :n~ ~ n=v ; - -m -, -. ~-.,. .c = - ~ - n-

y-wa s:=mn. nw..mmm . s u.= --~.--.-_n . ~.... - - \\ t t-l h; p ATTACHMENT C p' 'O h Comparison With Reference Plant Data bB~ The extent of core damage can be estimated from the measured fission product l (( concentrations in either the gas or water samples, as described for the reference plant. However, the measured concentration must be corrected for the differences in operation power level, time of operation, primary coolant mass, and containment gas volume. Ref At g = e xF x F, i -C,g C,g g OR Ref 1t xF xF C e C = g 7g g l Ref p Concentration of isotope i in the reference plant coolant C = yg (yCi/g). ) Ref Concentration of isotope i in the reference plant containment C = 3 'N gas (yCi/cc). (d J., Measured concentration of isotope i in BSEP's coolant (yCi/g). C = yg See Attachment A. l t Measured concentration of isotope i in BSEP's containment gas { C = (yCi/cc). Sea Attachment A. L F At L g (- e. = Decay correction to the time of reactor shutdown. t Decay constant of isotope 1 (day-1). l 1 r i = l Time between the reactor shutdown and the sample time (days). i l: t = t Inventory correction factor for isotope i. See Attachment B. f F = 7g I I F . Containment gas volume correction factor. See Attachment A. = l 8 F, Primary coolant mass correction factor. See Attachment A. l l = l t l BSEP/Vol..XIII/ PEP-03.6.3 15 Rev. 4

n v

6 9 , y- .- w ~~ -~c ~~~~~~~~m ,I i )

..c, m r n w n u a w_.,.. a--- I i; f I: t EXHIBIT 3.6.3-3 i' '( U BSEP TO REFERENCE PLANT PARAMETERS fc p: Reference Plant BSEP Reactor Thermal Power 3651 MWt 2436 MWt Number of Fue1~ Bundles 748 bundles 560 bundles Total Primary Coolant Mass 3.92 x 10' g 2.69 x 10' g (reactor water plus suppression pool water) l i ^ Total Drywell and Torus Gas Space Volume 4.0 x 10 ' cc 8.11 x 10' cc 1 Reactor Water 2.46 x 10' g 2.14 x 10' g Suppression Pool 3.67 x 10' g 2.48 x 10' g I Drywell Gas Volume 7.77 x 10' cc 4.65 x 10' cc l -Torus Gas Volume 3.25 x 10 ' cc 3.46 x 10' cc 1 1 P O g 1 p I i l l l t BSEP/Vol. XIII/ PEP-03.6.3 16 Rev. 4 i ~m,m - n. c- -, <. n n, y

=n _ c__. m- .... =. - - - -- ~. - - - - / ( ?: EXHIBIT 3.6.3-4 y j'"$ _ s: V [f:l ^ Core Inventory of Major Fission Products in a Reference Plant X." Operated at 3651 MWt for Three Years In Half-Inventory Major Gamma Ray Energy-k Chemical Group Isotope Life

• 10'Ci Intensity - kev (T /d) l Noble Gases Kr-85m 4.48 h 24.6 151 (0.753)

Kr-85 10.72 y 1.1 514 (0.0044) Kr-87 76.00 m 47.1 403 (0.495) Kr-88 2.84 h 66.8 196 (0.26), 1530 (0.109) Xe-133 5.25 d 202.0 81 (0.365) I I Xe-135 9.11 h 26.1 250 (0.899) Halogens I-131 8.04 d 96.0 364 (0.812) I I-132 2.30 h 140.0 668 (0.99), 773 (0.762) I-133 20.80 h 201.0 530 (0.86) 1-134 52.60 m 221.0 847 (0.954), 884 (0.653) I-135 6.59 h 189.0 1132 (0.225), 1260 (0.286) Alkali Metals Cs-134 2.06 y 19.6 605 (0.98), 796 (0.85) Cs-137 30.17 y 12.1 662 (0.85) l Cs-138 32.20 m 178.0 463(0.307), 1436 (0.76) g l Tellurium Group Te-132 78.00 h 138.0 228 (0.88) Noble Metals Mo-99 66.02 h 183.0 740 (0.128) Ru-103 39.40 d 155.0 497 (0.89) Alkaline Sr-91 9.52 h 115.0 750 (0.23), 1024 (0.325) Earths Sr-92 2.71 h 123.0 1384 (0.9) r'~s Ba-140 12.80 d 173.0 537 (0.254) (_-) Rare Earth Y-92 3.54 h 124.0 934 (0.139) La-140 40.20 h 184.0 487 (0.455), 1597 (0.955) l Ce-141 32.50 d 161.0 145 (0.48) Ce-144 284.40 d 129.0 134 (0.108) Refractories Zr-95 64.00 d 161.0 724 (0.437), 757 (0.553) l Zr-97 16.90 h 166.0 743 (0.928) l h = hour d = day a = month .y = year ( l' l l l (D BSEP/Vol. XIII/ PEP-03.6.3 17 Rev. 4 l b - -weeg= sme= _ w - - e, w w. l mm

y:N" "*" ~N,_.a whu, ..a__m _..w_sa __ u.. 1,_.. ,/ t i. I 1 ATTACHMENT D L ,'j. g Integration of Containment Atmosphere Hydrogen Measurement (( Into Core Damage Estimate The extent of fuel clad damage as evidenced by the extent of metal-water reaction can be estimated by determination of the hydrogen concentration in the containment. That concentration is measurable by either the containment hydrogen monitor or by the Postaccident Sampling System. l i. A correlation has been developed which relates containment hydrogen concentration to the percent metal-water reaction for Marks I and II type [' containments. That correlation is shown in Exhibit 3.6.3-5. Note A to that exhibit indicates the major assumptions used in developing the correlation. Note B indicates the method by which Brunswi'ck plant can use the correlation to determine the extent of clad damage. 1 i FN } ) i e I l i i I ,o BSEP/Vol. XIII/ PEP-03.6.3 18 Rev. 4 j

.... -... ~ - -. r ? / f; ATTACHMENT D (Cont'd) I t' O y";J-EXHIBIT 3.6.3-5 by.., Nl~ t.' 1

• 8

?- p g. so 3,., [ L, 4 I se ,/ h, U ^ ~- l 52 l t l i w g-j 22 [ y s 3 o ON l24 m i I to 4 12 s a I i t I i t i t i s o e to ao ao do , so .co 70 80 80 100 l sounte4Fst 5 hsETAt.-hTsR REACT 90eg Hydrogen Concentration for Marks I and II Containments as a Function of Metal-Water Reaction BSEP/Vol. XIII/ PEP-03.6.3 19 Rev. 4 1

• [. ',* " ' ~M *:

.,WT' 9w'7eW M Y **** K :*"'T C'Y, , n i sW w w ..r y.. r -( -. W in. 9,5 m,- ew tzs n 1- ;, T n ww

• r i

pg., n,m wc-y=m u m,--.=~...:. w Vl G ATTACHMENT D (Cont'd) Note A to Exhibit 3.6.3-5

97~

Analytical Assumptions f.[: (For Marks I and II Containments) 0-E. ; 8 1. Containment Volume = 350,000 ft ey '2. Number of Bundles = 500 3. Fuel Type = 8 x 8 R t,,.. s p,s w 4. All hydrogen from metal-water reaction released to containment. p 5. . Perfect imiding in containment. 6. No depletion of hydrogen (e.g., containment leakage). 7. Ideal gas behavior in containment. a / O r k i k,' L t (' t i. i BSEP/Vol. XIII/ PEP-03.6.3 20 Rev. 4 1 -l. i i

g__ _...... ~. ~..-. i ii 3 ATTACHMENT D (Cont'd) k' fN h, J Note B to Exhibit 3.6.3-5 7t~ m L' !:'n. 3 s. Determination of Clad Damage From Hydrogen Monitor Reading 69 Step 1. Obtain containment hydrogen monitor reading in percent. t' Step 2. Using the curve in Exhibit 3.6.3-5, determine the metal-water i~ reaction for the reference plant, MWR ref* Step.. The metal-water reaction from the actual in plant conditions (MWR) i. in determined from the following equation: f, $ ' b. % MWR ='(MWRref) x 500 x V N 350,000 i where: } N = Number of Bundles = 560 k 8 2.86 x 10' V = Total Containment Free Volume, ft =

• I I

i.- Y p ?. l t x/ [ Y-[ 3 4-3(. s ) 7" l s BSEP/Vol. XIII/ PEP-03.6.3 21 Rev. 4 e f r ,m .-~,-n,, . m r - ~ ~ ~.~e m-->-- .a

u_. - - -. - - I - ~ ATTACHMENT E v Integration of Containment Atmosphere Radiation Measurement

• ?

Into Core Damage Estimate An indication of the extent of core damage is the containment radiation level I which is a measure of the inventory of fission products released to the containment. This attachment contains a correlation of the containment radiation monitor dose rate to the percent of fuel inventory airborne in the containment. The purpose of this attachment is to present that correlation and provide a method to use that correlation to determine the degree of core I damage. g t p Exhibit 3.5.3-6 p,covides the results of a correlation performed for the Monticello plant. The key parameters which ' impact the containment dose rate are reactor power and containment volume. The method whereby individual plants can apply this correlation is provided in Note A to Exhibit 3.6.3-6. i' f l~ f i I ,m i, ) i !~ t i k I' i i I I ~ BSEP/Vol. XIII/ PEP-03.6.3 22 Rev. 4 iI m /* m ersw=. ..rt==.-ee-n e, ---r-- ~.-- * - - r m &*w-r e e =.:o w,r : *r= w w.e-r=r*. :

mw m .,. u_a m. uw -. : = - = ~ ~.--- p i 4 A'ITACHMENT E (Cont'd) ti Ox k" EXHIBIT 3.6.3-6 [M g S-Percent of Fuel Inventory Airborne in the Containment p.o e p,' g 1 ....,es2 2_.rr.... =e

• 256 2adine W

f' b']

• 24 particulasas p.:

te } 5N.+ s. oE w o l .e / ]' I C t l 1 i &W ss .1 g k

i. kN

~ e.1, [ l/ .. 1 i i. O l. i L, 1 *, - t j d 3 p r e ) A A b' W1tf. I A A II)8Il0 I F' ) ) 4 hb) Wig ).) i h b>W)$ ) F 1 t b@hM % Fuel Inventory Released Approximate Source and Damans Estimate f;y-t E 100.00 100% TID-14844, 100% fuel damage, potential core melt. 50.00 50% TID noble gases, THI source. 10.00 10% TID, 100% NRC gap activity, total clad failure, g l partial core uncovered. l 3.00 3% TID, 100% WASH-1400 gap activity, major clad failure. 1.00 1% TID, 10% NRC gap, maximum 10% clad failure. l J 0.10 0.1% TID, 1% NRC gap, 1% clad failure, lo.a1 beating of 5-10 fuel assemblies, r l 0.01 0.01% TID, 0.1% NRC gap, clad failure of 3/4 fuel element l (36 rods). ~8 10 0.01% NRC gap clad failure of a few rods. 10

• 100% coolant release with spiking.

5 x 10 100% coolant inventory release. ~8 10 Upper range of normal airborne noble gas activity in containment. i BSEP/Vol. XIII/ IMP-03.6.3 23 Rev. 4 f; g ~

a a. .e >w._ = ---_,m_.-.- I l- ' ~ '. KITACHMENT E (Cont'd) r. O p y' NOTE A to Exhibit 3.6.3-6 Determination of Clad Damage From Containment Radiation Monitor Reading The procedure for determination of fraction of fuel inventory released to the containment is as follows: u. Step 1: Obtain containment radiation monitor reading, [R] in rem /hr. [j fi. Step 2: Determine elapsed time from plant shutdown to the containment ' j, " isdiation monitor reading [t] in h,ours. w Step 3: Using Exhibit 3.6.3-6, determine the fuel inventory release for the l-reference plant [I],,f in percent. t-Step 4: Determine the inventory release to the containment [I] using the [' following formula: g _. [I) = [I) 1670 V j g 1 P 237, 450 e k where: i. V ) P = reactor power level Wth (BSEP = 2436 Wth)* V = total containment free volume, ft (BSEP = 286, 370 ft ). [ 8 8 I NOTE: Monitor location within the containment is assumed to have an insignificant impact on dose rate due to fuel inventory } airborne in containment. t V p i 9 4 f^ f BSEP/Vol. XIII/ PEP-03.6.3 24 Rev. 4 1 .. _ -. - -,- - --. - --..... - - ~. I

.mo____._ i l k l. I EXHIBIT 3.6.3-7 [ s r Computer Inputs for the PASS Program j I~ Concentration of I-131 in Reactor Water (pCi/ml) Concentration of I-131 in Suppression Pool (VCi/ml)* Concentration of Cs-137 in Reactor Water (pCi/ml) Concentration of Cs-137 in Suppression Pool (pCi/ml)* l Concentration of Xe-133 in Drywell (pCi/ce) ? Concentration of Xe-133 in Torus (pCi/cc)** i Concentration of Kr-85 in Drywell (WCi/cc) t Concentration of Kr-85 in Torus (pCi/cc)** l l Time between Reactor Shutdown and Sample Time (days) l t If time and availability permits, attach information necessary for the calculation of Inventory Correction Factors (see Attachment B); otherwise, } ~s conservative default correction factors will be used. t i Plant Sampling and Analysis Team Leader: Give completed exhibit to Dose { Projection Coordinator, j Dose Projection Coordinator: Enter data into PASS computer program and i provide results to Plant Sampling and Analysis Team Leader. i. I

• If unavailable, assume suppression pool activity = 0 pCi/ml.
  • If unavailable, assume torus concentration equal to drywell in pCi/cc.

h I l' 4 BSEP/Vol. XIII/ PEP-03.6.3 25 Rev. 4 i ---.-l - - - + - - - - c-- g- -r:. w,- n-~- ;~,-- r-, ww c c -r r-, y - -

n .n+ = - ~. - g 1-~ i EXHIBIT 3.6.3-8 ( - ~ ~ ~ - s D- ~ VERIFICATION OF PASS P (A Computer program for estimating core damage based on Postaccident Sampling System results)' This exhibit is intended to provide a means to ensure that PASS, a core damage estimate program designed for the IBM Personal Computer,.is working properly. This is demonstrated by duplicating expected results of known computer inputs. These results can be validated by comparison to manual calculations for the same input. 'No different, test cases are presented so that a number of alternate paths within the program can be tested. The test " cases with their expected results follow. {_ TEST CASE 1 Computer Prompt Expected Input I' _ Enter The Concentration of the Fission Products [ Concentration of I-131 in Reactor Water (DCi/al) 1.72E + 3 Concentration of I-131 in Suppression Pool (pCi/el) 1.49E + 2 b / -t Concentration of Cs-137 in Reactor Water (DCi/el) 6.55E + 2 - l,' Concentration of Cs-137 in Suppressiott Pool (pC1/el) 5.70E'+ 1 L Concentration of Xe-133 in Drywell (DC1/cc) 1.82E + 2 l Concentration of Xe-133 in Torus (pci/cc) 2.41E + 2 Concentration of Kr-85 in Drywell (pC1/cc) 1.43E + 0 Concentration of Kr-85 in Torus'(sci /cc) 1.90E + 0 [ 1 For the inventory correction factor do you want to use the conservative i default values which are bases upon BSEP's operations under the same YES operational constraints (YES or NO)? Enter time between the reactor shutdown and the Sample Time (Days) 2 The results should resemble the printout on the following page. If they do not, carefully check your inputs and try the test again. If the results still are not similar, try a backup copy of the program..If that fails, then seek programming help. O". BSEP/Vol. XIII/ PEP-03.6.3 26 Rev. 4 Pm1 /.T MgQg gyrpPW.WM*P T97'7[ "*?7W"sr yt".ry****;;fyy,3ya* - qq*;*7*+ My,r=e*.7

l EXHIBIT 3.6.3-8 (Cont'd) } ESTIMATE THE EXTENT OF CORE DAMAGE UNDER ACCIDENT CONDITIONS DATE: 03-28-1984 TIME: 13:21:27 The concentration of the fission products are: I-131.in Reactor Water 1.72E + 3 pCi/ml I-131 in Suppression Pool 1.49E + 2 pCi/a1 Cs-137 in Reactor, Water 6.55E + 2,pC1/ml Cs-137 in Suppression Pool .i.70E + 1 pCi/mi Xe-133 in Drywell Air, 1.82E + 2 pCi/cc Xe-133 in Torus Air 2.41E + 2 pCi/cc j Kr-85 in Drywell Air. 1.43E + 0 pCi/cc { Kr-85-in Torus Air 1.90E + 0 pCi/cc /. O Time between the reactor shutdown and the sample time is: ' 2 days + The Concervative Default values of the Inventory Correction Factors were used. Estimate of fuel / cladding damage Primary Coolant Analysis [ i'k. r Nuclide CwRET (pci/al)' % Cladding % Fuel I, Failure Meltdown J.:. J -I-131 3.00E + 02 69.00 1.35 i Cs-137 1.00E + 02 64.54 4.27 r Containment Gas Analysis F i' Nuclide CwREF (pci/ml) % Cladding % Fuel i Failure Meltdown Xe-133 7.99E + 01 53.26 1.84 Kr-45 5.00E - 01 56.35 1.92 i BSEP/Vol. XIII/ PEP-03.6.3 27 Rev. 4 t I f," $# 5 ' ~- 0 E I C,* NT --f;' y V-rh y j,- ' 1 -Y .A

L ._ __ _ _.-. _ _ _ _.~ _ _ n EXHIBIT 3.6.3-8 (Cont'd) U TEST CASE 2 Computer Prompt Expected Input Enter The Concentration of the Fission Products Concentration of I-131 in Reactor Water (uCi/ml) 1.35E + 3 Concentration of I-131 in Suppression Pool (pCi/ml) 1.18E + 2 Concentration of Cs-137 in Reactor Water (uCi/ml) 1.17E + 2 Concentration of Cs-137 in Suppression Pool (pCi/ml) 1.02E + 1 Concentration of Xe-133 in Drywell (pCi/cc) 1.84E + 2 Concentration of Xe-133 in Torus (DCi/cc) 2.45E + 2 Concentration of Kr-85 in Drywell (uC1/cc) 2.91E - 1 Concentration of Kr-85 in Torus (pCi/cc) 3.86E - 1 For the inventory correction factor do you want to use the conservative 7 default values which are bases upon BSEP's operations under the same operations 1 constraints (YES or NO)? NO Enter time between the reactor shutdown and the Sample Time (Days)? 2 Enter number of Operating Periods from the unit operating history? 3 O For period number (1) enter: V Average steady reactor power operated in this period (MWT)? 1000 Duration of this operating period (days)? 60 Time between the end of this operating period and the time of the most recent reactor shutdown (days)? 254 i. k, For period number (2) enter: f. Average steady reactor power operated in this period (MWT)? 2000 Duration of this operating period (days)? 200 l f Time between the end of this operating period and the time of the most recent reactor shutdown (days)? 44 For period number (3) enter: f, Average steady reactor power operated in this period (MVT)? 3000 [ Duration of this operating period (days)? 14 Time between the and of this operating period and the time of the most recent reactor shutdown (days)? O The results should resemble the printout on the following page. If they do not, carefully check your inputs and try the test again. If the results still are not similar, try a backup copy of the program. If that fails, then seek programming help. BSEP/Vol. XIII/ PEP-03.6.3 28 Rev. 4 J -. - -. -,.. ~.. - ~.

1 EXHIBIT 3.6.3-8 (Cont'd) -p Q ESTIMATE THE EXTENT OF CORE DAMAGE UNDER ACCIDENT CONDITIONS DATE. 03-28-1984 TIME: 13:27:17 The concentration of the fission products are: -I-131,in Reactor Water 1.35E + 3 pCi/ml I-131-in Suppression Pool 1.18E + 2 pCi/ml Cs-137 in Reactor Water 1.17E + 2 pCi/ml Cs-137 in Suppression Pool 1.02E + 1 pCi/mi Xe-133 in Drywell, Air 1.84E + 2,y,Ci/cc Xe-133 in Torus Air 2.45E + 2 pCi/cc Kr-85 in Drywell Air 2.91E - 1 DCi/cc 'Kr-85 in Torus Air 3.86E - 1 DCi/cc Time between the reactor shutdown and the sample time is: 2 days The Inventory Correction Factors were calculated from the following: I. Period No. Operation Time Time Between Period Average Power (days) & I.ast Shutdown (days) (MWt) 1 60 254 1000 1 2 200 44 2000 .3 14 0 3000 t i Estimate of Fuel / Cladding Dacage Primary Coolant Analysis b I Nuclide CwREF (DCi/al) % Cladding % Fuel Failure Meltdown- [; t I-131 3.00E + 02 69.02 1.35 l Cs-137 9.99E + 01 64.49 4.27 j. Containment Gas Analysis Nuclide CwREF (uci/ml) -% Cladding % Fuel Failure Meltdown Xe-133 8.00E + 01 53.30 1.84 g Kr-85 5.00E - 01 56.40 1.92-BSEP/Vol. XIII/ PEP-03.6.3 29 Rev. 4 i

.._n_....-...--. . - -... - - - - - ~ - - -. ~ WORKSHEET A1 i CALCULATION OF ISOTOPIC CONCENTRATIONS IN PRIMARY WATER AND SUPPRESSION POOL WATER (Cw ) AND DRYWELL GAS AND TORUS GAS (Cg 1 f Reference _s Section 3.3.1.2 Section 3.3.1.5 Attachment A Exhibit 3.6.3-3 g (pCi/ml) = (Concentration Rx H O)g (0.08') + Cw 2 (Cs18', I182) (Concentration Suppression Pool H O)g (0.92) 2 = + (pCi/ml) DCi/ml = Cs 6 i 881 pCi/ml i and = y Cgg (pCi/ml) = (Concentration Drywell)g (0.57) + (Concentration Torus)g (0.43) (Xe " 8, Kr') = + (pCi/cc) J DCi/cc,188 = y pCi/cc and = Kr r l BSEP/Vol. XIII/ PEP-03.6.3 30 Rev. 4 __,t, m

~.. _. .a _.:_ _ __. 1 ~~ WORKSHEET A2 ) CALCULATION OF INVENTORY CORRECTION FACTOR (FI 1 References Section 3.3.1.4 Attachment B Exhibit 3.6.3-4 MW T). = Days P) = t al Days'1 T) = Days 1 = FI = 3651'(1 - e ~1"' i) g 1T I) P) (1 - e - i j)(e-1 T ') ij (Cs18') ( = (I 588) i. (Xe288) -s I ) (Kr") t c. i I, BSEP/Vol. XIII/ PEP-03.6.3 31 Rev. 4 _ _i ... _ _ - ~ _. _

y m w ~_ = ,..= -x. =, - - -. .a.ac - = ~ - . ~. -. - WORKSHEET A3 f CALCULATION OF NORMALIZED ISOTOPIC CONCENTRATIONS IN PRIMARY WATER AND SUPPRESSION POOL WATER (Cw Ref) AND DRYWELL GAS AND TORUS GAS (Cg " f), g g References Section 3.3.1.6 NOTE: For BSEP, Attachment C Fw = 0.68622 Worksheet Al Fg = 0.20275 Worksheet A2 Cw = Cw e x FI x Fw g g g (Cs18', i181) yCi/m1 18' = Cs yCi/mi :: { g f Cg = Cg a x FI x Fg g g g (Xe188, Kr) 188 pCi/cc Xe = pCi/cc Kr 4 i i L l i , p BSEP/Vol. XIII/ PEP-03.6.3 32 Rev. 4 v l .~ ~ - >. ~ ~ ~ ~ ~ ~ ~ ~ - - - < ~m ~ ~ ~ ~ - ~. =;: =,e ,,e-.r..,-,,. - w n, ~ ~ v,--- w ~~.~

4 1,_. t' i WORKSHEET A4 ,q . tg. ESTIMATE OF FUEL / CLADDING DAMAGE References Section 3.3.1.7 Exhibit 3.6.3-2 Worksheet A3 Primary Coolant Analysis % Cladding % Fuel Isotope Cw,Ref(yCi/ml) Failure Meltdown l ^ Inst 6 I l Cs18' Containment Gas Analysis [ C._ I % Cladding % Fuel h g *I(DC1/el) Failure Meltdown Isotope Xe888 88 Kr f' t I If I L i 6 \\u BSEP/Vol. XIII/ PEP 03.6.3 33 Rev. 4 c -..v w.. - m =~ w,-.-, -+>--+1--.. row..- c -e- + - r - a --

p.,w, ~w.- --.- - -~ -~ ( 1 WORKSHEET B1 p y c.; DETERMINATION OF CLAD DAMAGE FROM HYDROGEN MONITOR READING h: 1 References Section 3.4.1 Attachment D Exhibit 3.6.3-5 n Containment Hydrogen Monitor Reading: MWR ref f Calculate % MWR: i % MWR = (MWR ref)(0.73) = 0 ? 1 i I l I i i l O l BSEP/Vol. XIII/ PEP-03.6.3 34 Rev. 4 I l l ._..,m,... =- ~n -~. - ~ ~.,,n ~ ~-,.m- ~~' w

Lg_ t, -x WORKSHEET B2 i DETERMINATION OF FUEL INVENTORY RELEASE BASED ON CONTAINMENT RADIATION MONITOR READING References Section 3.4.2 Attachment E Exhibit 3.6.3-6 Containment Radiation Monitor Reading: rem /hr Time irom Shutdown to Monitor Read'ing: hrs [I] ref (Reference Fuel Inventory Release, %, from Exhibit 3.6.3-6) I (Actual Fuel Inventory Release) = [I) ref

• 0.827 i

= /^N i s I m )' BSEP/Vol. XIII/ PEP-03.6.3 35 Rev. 4 - - - -,-. ~, m J ---v,,v,.c.-.,,-.-.-------- i}}