ML20084R865

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Rev 1 to Procedure Bzp 380-A9, Example of Core Damage Assessment
ML20084R865
Person / Time
Site: Byron, 05000000
Issue date: 04/30/1984
From:
COMMONWEALTH EDISON CO.
To:
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ML20084R841 List:
References
BZP-380-A9, NUDOCS 8405230293
Download: ML20084R865 (31)
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.BZP 380-A9 p . ,. , ; g.

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(:],,t , U p )_ f Revisinn i l 1

EXAMPLE OF CORE DAMAGE ASSESSMENT ,

l The following hypothetical example is intended to illustrate the use of BZP l 380-19,." Core Damage Assessment."

Simulated Accident Scenario The Station has experienced a reactcr accident during which the core ,

temporarily became uncovered and safety injection has initiated. Some degree of fuel damage is likely to have occurred. All indications are that a large break LOCA has taken place. The temperature of the reactor coolant is now

  • 200*F. Samples are requested 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the reactor was shutdown.

I Samplina and Samole Activities From page Ib of BZP 380-A8, " Suggested Sampling Locations," (page 5 of this appendix) a reacter coolant sample is requested at the containment sump sample point as the containment sump contents are now providing core cooling. Also the containment atmosphere sample is requested. All samples are analyzed 2 '

hours af ter they are drawn.

The results of the analysis are as follows:-

Reactor Coolant Containment Atmosphere

(Containment Sump)

Kr 87 6.91E-6 Ci/gm Kr 87 2.79E-6 Ci/cc Xe 133 9.95E-4 Ci/gm .Xe 133 4.01E-4 Ci/cc 1 I 131 1.97E-3 Ci/gm . I 131 4.07E-6 Ci/cc .

I 132 4.88E-4 Ci/gm I 132 4.00E-6 Ci/cc i Ba 140 9.99E-4 Ci/gm

[

All sample activities reported represent the activity of the sampl3 at the time of the analysis and have not undergone a decay correction back to time of sampling.

4 Specific activities from the chemistry gamma isotopic analysis report are j recorded in the measured specific activity column of the RCS and Containment Atmosphere Activity Worksheets (pages 6 and 7 of this appendix).

i Time elapsed from reactor shutdown to time of analysis (6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />) is recor11ed t l in column 2.

. Ingrowth and decay correction factors are calculated using the equations found on page 10 of BZP 380-A8, (page 8 of this appendix).

l I The corrected specific activity is then determined by multiplying the measured -

specific activity by the correction factor. The corrected specific activity is recorded in column 5 (pages 6 and 7).

APPRoyED 8405230293 840504 PDR ADOCK 05000454 APR 3 01984

! A PDR

)

B. o, s, g*

l. (4013P)

P u... --

BZP 380-A9 Revisicn 1 Initial Assessment From a brief review of the nuclides and activities present in samples and BZP 380-A8 page 8, Selected Nuclides for Core Damage Assessment (page 9 of this appendix), it is likely that some degree of fuel melting has occurred. The presence of a large concentration of Ba-140 in the reactor coolant sample is an i.,dication of a fuel melt condition.

Total liquid Mass Determination The RCS mass and all RCS mass additions are identified and calculated using the Estimate of RCS Mass' worksheet (page 10 of this appendix). Reactor coolant specific gravity is. determined from the " Specific Gravity of Water vs Temperature graph (page 11). Reactor coolant specific gravity is approximately 1.0 at 200*F. The total liquid mass is recorded on the RCS Activity Worksheet (page 6).

Total Activity Released From Fuel The total activity released to the coolant and containment atmosphere is calculated from the decay and ingrowth corrected specific activities, reactor coolant system mass and the containment atmosphere free volume. The total activities released are then recorded in column 7 of the activity worksheets

. (page 6 and 7) and columns 2 and 3 of " Release Activity / Percent Rel, eased (page 12 of this appendix)". The sum of the liquid and containment atmosphere 3

activity is recorded in column 4 (page 12).

Total Core Inventory

' The reactor power history is obtained from the control room and used to calculate the core inventory of the nuclides used in the assessment. The reactor power history for the period prior to the accident for this example is as follows:

The Plant has been operating at 100% reactor power for approximately 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br /> prior to experiencing abnormal conditions.

Reactor power was reduced to 90% for a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Reactor power was reduced to 80% for the period of 30 minutes prior to the reactor trip.

The reactor power history is recorded and core inventory determined for e6ch nuclide in the Power History / Data Sheets (pages 13 through 22 of this appendix) . Average power levels are recorded in column 1 of the Power History Sheet. The duration at each power level is recorded in column 2. The time between the end of each power level and the reactor shutdown is recorded in column 3. The duration of the power history that must be considered for core inventory determinations varies with each nuclide and is specified on the Power History Sheet.

APPROVED APR 3 01984 (4013P)

BZP 580-A9 RGvision i Factors Ai and Bi are determined from the graphs on the Data Sheet following each Power History sheet. Ai and 84 are recorded in columns 4 and 5 respectively. The product of A i, Bi and Pi is recorded in column 6.

The corrected core inventory is then calculated from the Power Correction Factor and the equilibrium core inventory given.

The total corrected inventory is recorded in column 5 of the Release Activity / Percent Released data sheet (page 12).

Release percentages are then calculated and recorded in column 6.

Damage Estimates From Nuclide Release Percentages With the nuclide release percentages determined, the core damage graphs may be used.

The percent released for each nuclide is used with the appropriate graph (pages 23 through 28) to determine the category and estimate the extent of damage.

Estimates are entered in the appropriate column of the " Core Damage Assassment Summary Sheet" (page 30 of this Appendix).

Non-Radiological Indicators of Core Damage The non-radiological indicators of core damage are determined to be the following:

Containment atmosphere hydrogen 7%

Containment High Range Monitor 7E4 R/hr Core Exit Thermocouple Readings 1700*F The above values are compared to values in table on page 31 " Characteristics of Categories of Fuel Damage" and recorded in the appropriate column of the

" Core Damage Assessment Summary Sheet" (page 30).

Radionuclide Ratios A radionuclide ratio is examined and compared to ratio's in the

" Characteristics of Categories of Fuel Damage" (page 31) of the appendix table. The Kr87/Xe133 ratio is recorded in the appropriate column of the i Summary Sheet. Page 29 of this appendix lists radionuclide ratios in the fuel

  • gap and fuel pellet. These ratios help to indicate the category of fuel damage.

The Kr87/Xe133 ratio of 0.18 is recorded under the fuel melt category (pas,e 30).

l APPROVED l APR 3 01984 l B.O.S.R.

l (4013P)

I j _ _ , , _ _ _ _

+

BZP 380-A9 R& vision 1

, Final Assessment All data collected on the " Core Damage Assessment Summary Sheet" is now evaluated to make the final assessment.

Nuclide release percentages indicate that a large amount of clad damage has occurred. This is supported by the following:

1. It was determined frcm reactor level instrumentation that the core had been uncovered at some point in the accident.
2. Core thermocouple temperatures reached 1700*F.
3. The containment atmosphere contained 7% hydrogen which indicates a 54%

zirconium cladding reaction with water (from page 28).

The damage to the core is beyond the extent that can be attributed to clad failure alone. Nuclide release percentages indicate that between 28% and 47%

of the fuel had reached an overtemperature condition. This estimate is supported by the following:

1. The Kr87/Xe133 release ratio approximated the ratio typical of fuel cellets (see page 29).
2. The containment high range monitcrs indicated dose rates typical of the fuel overtemperature conditions.

The Ba-140 release percentage (5%) corresponds to a 20% fuel-melt condition.

Other nuclide release percentages indicate a 15% to 27% pellet melt. The non-radiological damage indicators (uncovered core, core exit thermocouple readings, and containment hydrogen measurements) support this estimate.

Therefore, for this example, the final fuel damage assessment is:

1. Greater than 50% fuel clad failure.

. 2. A maximum of 50% fuel overtemperature.

3. Less than 50% fuel melt.

f APPROVED' APR 3 01984

s. o. S. R.

(4013P)

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O BZP 380-A9 Revision 1 .

i SCENARIO SAMPLING 10ChlIONS(S) SAMPLE PROCEDURES (M C0tMENIS i

Large break '

LOCA.

Rx Power >1% Containment sump NONE

, 1(2),PS156 ard l

Containment BZP 380-18 Atmosphere BCP 800-7 ard RCS llot Leg BZP 380-12/ If Containment Sump is providing cooling to

Loop 1 1(2)PS9351A or Diluted R/C and core, do not include hot leg sample activity L )op 2 1(2)PS93518 BXP 380-15 in damage assessment calculations.

R/C Stripped Gas Rx Power (11 Contiainment BZP 380-18 Atmosphere BCP 800-7 l AM -

Containment Sump NONE Provided Containment Sump supplying core 1 1(2)PS156 cooling or WilR System Provided RHR taking suction from Containment l

1(2)PS9353A or Sump.

i 1(2)PS93538 or RCS Ilot Leg Loop 1 1(2)PS93b1A or BZP 380-12, Diluted R/C

! Loop 2-1(2)PS93518 BZP 380-15 R/C Stripped l

Cas 18 APpRoyED APR 3 o ggg4 B. o S R. Page 5

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E X i X K1 Ml -f e v, m k. W l. - - 8, 2 -4, - {q,b APPROVED APR 3 01984 Page 7 B. O. S. R.

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BZP 380-A9 R1 vision 1 BZP 380-A8 Revision 1 CALCULATION 0? INGROWIM Me DECAY CORRETION FACTORS Kr-85 e0.157t Kr-67  : e0.547t = 6 0.S41(t0 ~ 30 b Kr-88  : e0 247t I

Xe-131M  : - -

-2.62e (3.54E-3)t + 3.62e (2.45E-3)t Xe-133 - -

-0.la k (3.41E-2)t -0.10e (1.28E-2)t+1.285e95.4as-3)t I

Xe-133M  : -0.le (3.415-2)t ,g,g, -(1.288-2)t 1

xe-135 - - -

-9.2k (1.04E-1)t -0.033e (2.66)t +10.293e (7.588-2)t

.I-131 > e .00359t 3 , e #'##359 k I

I-132 > 0.103e'(8*918~33* +0.897e (0.307)t h7 I-133 -  :- e0.0341t I-135 > e0.104t Cs-134 >1 1

. Te-129 -

> 1.09eMO.161)t +0.16e (8.47E-4)t -0.25e (0.605)t l Te-132 > e 0.00892t 0.00L LS(0) F /d/

Ba-140 - p. e .00225e 0 6 1

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l -0.14e (3. 8)t

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0.91e (1.04e4)t +0.09e (2.4)t where:

t = the maaber of hours between reactor shutdown and time of sample count (from column 2 of SZP 380-T4. pages 1 and 2).

(39569)

{

APPROVED i

APR 3 01984 B. O. S. R.

Page 8

e BZP 380-A9 Rsvision 1 i

SIP 380-A8 tevision 0 wt 't"TED 4te "t3 FCE C: RE OerseGE 2SSESSPENT

  • 3re Qamage State 1g Half-Life
  • W ominant Casanas (Mov) vtald (t)*

Clad Failure Kr-4Ses* 4.4 h 150(74), 30$(13) tr-47 76 m 403(84), 2S70(35)

(r-asse 1.8 n 191(35), 550(23). 1400(35) xe-lita 11.8 d 164(2) xe-113 S.27 4 81(37)

Xe-133ma* 2.26 d 233(14)

Xe=135*e 9.14 h 250(91)

I-131 S.05 d 164(82)

I-132 1.26 h 773(89), 9$5(22), 1400(14)

I-133 20.3 h $30(90)

I-135 6.68 h 1140(37), 1210(14), 1460(12), 1720(19)

RI> 48 17.8 m 898(13) 1863(21)

Fuel overheat Cs-134 2 ye 60$(98). 796(99)

Cs-137 30 yr 662(85)

Te-129 64. 7 a 4SS(15)

Te-132 77.7 h 230(90)

Fuel N lt SP-49 $2.7 d

$r-90se 23 yr Ba-140 12.S d $37(34)

La-140 40.22 h 447(40), 815(19), 1596(96) 1.a-142 92.5 3 650(44), 1910(9), 1410(15), 1550(11) .

Pe-144 17.27 a $95(1.5)

  • Values ontaines free Taale of tsotones, i.ederer, Hollanner, and Pert:nnn,

$1stn Edition.

  • These r:uclidee are . marginal with respect to telection criteria Per .

candiente ruclides; they Neve been included sn the :ossibility that they my be detected and tnus utilized in a menner analogous to the candidate nuclides.

i e

APPRoyED (1956P)

APR 3 01984 B. O. S. R.

Page 9

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B7P 380-A9 Ravicion 1 BZP 380-T4 Revision 2 Estimate of RCS Mass

1. Obtain the reector coolant volume additions for the following:

Estimated Voltano Maxisua Volume

. Tank Added Added (gallons)

a. Refueling Water Storage Tank /8d 400 495.000
b. Accumalator A 7i 1/7 7.217
c. Accumalator B 22/7 7.217 7,2/7 7.217
d. Accumulator C
e. Accumulator 0 ~7 2/7 7.217 Soric Acid Storage Tank "O" 48.000 f.
g. Residual Heat Removal System 8000 S.000
h. Other source MA "" O "

2 Il 8/eA Total

2. Convert gallons to grams as follous:

Total reector coolant system value added: ,

2J3>606 . gallons a 378Sgss/ gal = 8/80 es

3. Determine the Reector Coolant System Mass as follous:

3.2288 grams a system specific gravity * /O = D # 8 arans

4. Determine the Total Liquid Mass as follows:

acS Mess 13LK8 grams + Added Mass 8/E8 grame= l/ M Y atans

  • System Specific Cravity is determined from SZP 300-A8. page 2.

3 (39589) i APPROVED APR 3 01984 B. o'. S. R.

Page 10

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Page 11

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APPROVED APR 3 01984 B. O. S. R.

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APPROVEO APR 3 01984  ;

Page 13 B. O. S. R.

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APPROVED APR 3 01984 B. O. S. R.

Page 26

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8. o. s. R.

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BZP 380-A9 Rsvicion 1 SZP 340-48 Revision 0 Isotopic acTTvfTY teTIos 0F rttt settET ano cap Fuel Pellet activity eatio Gao activity tatio i

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APPROVED APR 3 01984 B. O. S. R.

Page 29 i

BZP 380-A9 Rsvision 1 I

82P 380-T4 Date: M8 Time: /33o Performed by: 7. Isa sem CORE DMGM ASSESSMENT SUN 0utY SHEET

Mode of Percent Clad Percent Percent Estimation n- lOvertemperature Puel Melt

<50% >50% <50% >50% <50% >50%

Kr 85m Kr 87 [oM 5:4 /e.) M7 27 Kr 88 - -

Xe 131m Xe 133 - - D O Xe L33m Xe 135 -

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(39589)

I APPRoyED APR 3 01984 Page 30 g, q g, g,

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-e Itze 380 A8 Revislosi 1 1

CIWuuCTERISTICS OF CAftOOttlES OF HJEl. DAMAGE PtialCENT CuterAIletElff j ate) TYPE HIGH SANGE C0kE Emit HYleM'm COHE 0 07 FISSION FISSION ARE TitEHprm H E CONE MONITolt 8WlAGE LWULEJCTS PHODUCT PK)NITOR. READINGS UNCOVtGetM I(2) Al P3 343 or 344 C_AJtG__Wj, __

, J R EA38D _ MIIO  !!hr* (f*C F) (YES/jl0_L (Vol. % Hy)

No Clad Damage u r -81<lu l0 -3 Kr-81 0.022 ----

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