ML17277A624

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Draft Procedure 9.3.30, WNP-2 Core Damage Evaluation
ML17277A624
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Issue date: 06/23/1983
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9.3.30, NUDOCS 8307050141
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WNP-2 PLANT PROCEDURES~

9.

NUCLEAR PERFORMANC~ALUATION 9.3.30 WNP-2 CORE DAMAGE EVALUATION 1.0 PURPOSE The purpose of this procedure is to provide a method of estimating damage to the core following a reactor accident.

2.0 DISCUSSION Core damage evaluation is based on several plant parameters including reactor water level, containment atmosphere radiation levels, contain-ment hydrogen concentrations and fission product radionuclide concen-trations in the reactor coolant and the containment atmosphere.

The estimate of core damage is determined by comparing the levels and con-centrations against those which co d result from 100/ core degradation at WNP-2.

The fission product inventWis ased on 1095 days of operation at 3323 NWt or 100Ã rated e

calculated by General ETectric for concentrations of radi onuc > es in the coolant or the containment atmo-sphere.

The radiation levels in the containment atmosphere are based'on the same core power level but it is assumed to have operated only 80Ã of the time or 295 days per year for three years.

These dose rates were calculated by ISOSHLD a point kernel shielding code utilizing RIBD to develop the source terms.

The two codes produce nearly identical source terms because of the relatively short after decay periods being considered.

Other plant parameters are:

Number of fuel bundles Reactor Primary Coolant mass Total Coolant (Primary plus suppressi on pool )

Containment Atmosphere (free volume)

Drywell Atmosphere

= 764 bundles

= 2.74 x 10 g

= 3.17 x 10 g

9

= 9.83 x 10 cc

= 5.75 x 10 cc 9

8307050i4i 830623 FDR ADQCK 05000397 E

PDR

vp P.

l l

! Ii

The procedure looks at water level history.

If the core has been uncovered this would be a first indication that there may be fuel or fuel cladding damage.

It looks at containment atmosphere radiation levels which will be quickest estimation of core damage in case of a LOCA. It also looks at, the hydrogen present in the containment atmo-sphere.

If hydrogen is present it will be indicative of Che extent of metal/water reaction that has occurred with the'ircalloy cladding sur-rounding the active fuel pellets.

Gas and water samples may be taken of the reactor coolant and/or the containment atmosphere to be analyzed for fission product concentrations and/or hydrogen concentration.

The presence of radioactive halogens and cesiums in the coolant and/or noble gases in the atmosphere will be indicative of core damage.

The presence of less volatile fission products such as barium, lathenium or strontium would indicate fuel melting.

Also, the ratio of short lived noble gas isotopes to Xe-133 and iodine isotopes to I-131 will assist in distinguishing between gas gap releases by cladding failure from core releases by fuel melt.

3.0 REFERENCES

1.

H. A. Careway "Calculation of Fission Product Inventor'y and Spectra RADC101 Program",

NBDO-25176 (October 1980).

2.

J. Greenborg et.al.,

"ISOSHLD - A Computer Code for General Purpose Isotope Shielding Analysis" BNML-236 (June 1966).

3.

C. C. Lin, "Procedures for the Determination of the Extent of Core Damage Under Accident Conditions" NEDO-22215 (August 1982).

4.0 PROCEDURE 4.1 Reactor Water Level Determine the reactor water level history from the Fuel Zone Level monitor, recorder LR-R615 on panel P-601.

Full Scale

='-117" Top of active core

= -167" Bottom of active core

= -317" From these readings determine how much, if any, of the core has been uncovered, for how long and when during the course of the accident.

4.2 Fission Gases i'n Containment Take readings on the Containment Radiation LOCA monitors (located on panels RAD22 and RAD23). If radiation levels are high enough to be recognizable, utilizing Appendix A, determine the fraction of fission gases that have been released to containment.

4.3 Hydrogen Gas in Containment Take readings from the containment Hydrogen Analyzers, located on panels K-I and K-II.

Evaluate metal-water reaction in accordance with Appendix B to estimate the fraction of core cladding damaged.

Top scale on the Analyzer is 10/ H2 which equates to 5/ of the cladding.

If Analyzer is off-scale, hydrogen concentration will have to be determined'from a containment atmosphere sample analyzed in the chem lab.

Use Appendix B to evaluate extent of cladding damage.

4.4 Post Accident Sample Under accident conditions when the WNP-2 Health Physics/Chemistry, manager or designee has decided to take a sample:

4.4.1 Obtain reactor coolant and/or containment atmosphere

samples, consistent with Appendix C, from the Post-Accident Sampling System.

4.4.2 Analyze the samples by galena spectroscopy to determine the concentrati ons of radionuclides in each sample.

Also analyze the atmosphere sample for hydrogen if required.

4.4.3 Normalize the concentration to a full power core (3323MMt) irradiated for 1095 days at decay time zero and for gas samples also to normalize to containment temperature and pressure in accordance with Appendix D correction factors.

4.4.4 Using normalized concentrations of:

Kr-85 Xe-133 I-131 Cs-134 Kr-88 Xe-135 I-133 Cs-137 enter the figures in Appendix E to determine the percent cladding failures or percent fuel melt.

4.4.5 If after entering figures in Appendix E the results could indicate either a gross cladding failure or a partial fuel melt, determine the amount of low volatile fission products present in the coolant sample.

The absence of low volatile fission products would indicate gross cladding failures rather than melting.

4.4.6 To distinguish between cladding failures and melting, deter-minee the ratio of Kr-87, Kr-88, Kr -85m to Xe-133 and/or I-132,

133, 134, 135 to I-131 and compare to ratios given in Appendix F.

APPENDIX A CONTAINMENT RADIATION LEVELS Four radiation monitors monitor the radiation levels inside the Drywell.

Two of the detectors are located in the bioshield wall at elevations 522'nd 525'nd azimuth 60 (Figure 1) and 297 (Figure 2).

They are in the best location to monitor the released fission gases in the drywell for the first two days following a reactor accident while radioactive gas energies are high.

They also minimize the background from other sources such as plate-out and recirculation lines.

Two other detectors are located inside containment at elevation 515" and azimuth 51.5 (Figure 3) and 290 (Figure 4).

These monitors will measure the long term and low energy radiation created by Xe-133 (80 KeV) and radiation from the solid radionuclides plated out on the interior surfaces.

The dose rate curves in the figures are based on the release of:

N.G.

= 100Ã noble gas core (3323MW) inventory.

N.G.+I

= 100K noble gas plus 50K halogen core (3323 MW) inventory.

To estimate the release of noble gases into the Drywell:

(1)

Read the dose rate (D1) for radiation monitors CONTAINMENT LOCA RAD-AZ-60 and CONTAINMENT LOCA RAD-AZ-297, located on RAD-Boards 22 and 23 in the Control Room.

(2)

On corresponding figures 1 and 2 enter the decay time since plant shutdown on the abscissa and determine meter reading (D

) for the N.G. curve.

If power level differed from 3323 MW, then correct P

Corrected C

=

m x (actual power level).

3323

0 I

4l

(3)

Fraction (FD) of noble gas core inventory released to the Drywell is:

F

'm Example:

o At 19 minutes decay, the reading on LOCA RAD AZ 60 is 35.R/hr.

Past operating level

= 2700 MW.

o Enter figure 81 at 19 mins., to find a meter reading (0

) of 2200R/hr.

o Corrected D

=

x 2700

= 1788 2200 m

3323 1

35 o

F

= 0.020 or 2.0X core inventory.

d 0

J./88 (4)

If the dose rate D1 exceeds the N.G. curve then other radio-nuclides are present in significant quantity and core damage is more severe than just releasing the volatile noble gases.

The same method should be followed for the inside containment monitors using graphs on figures 83 and 84.

These monitors should be utilized after 3 days

. decay.

A-2

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APPENDIX B

CONTAINMENT HYDROGEN LEVELS The extent of metal-water reaction for the zirconium fuel cladding can be estimated from the hydrogen content in the containment atmosphere.

Contain-ment hydrogen concentration can be determined from the containment hydrogen-oxygen analyzers up to 10%.

Analysis of concentrations greater than 10/

H2 requires analysis of samples taken by the post accident sampler..

To determine the extent of. zirconium metal-water reaction:

(1)

Enter Figure 8-1 with the percent hydrogen and from the curve determine the percent metal water reaction.

ANALYTICALASSUMPTIONS (1)

Containment volume

= Drywell - 200,450 ft

+ Wetwell 142,500 ft 3

3

= Containment total - 343,000 ft3 (2)

Active zirconiumonly that portion surrounding fuel pellets o

Length

= 150" o

I.D.

= 0.419" Wal 1

= 0.032" Density

= 0.236 lb/in3 62 tubes/fuel assembly 764 assembli es (3)

Zr + 2 H20

'Zr02 2H2~

(4) 1001 metal-water reaction produces 7.56 x 10 moles H2 6

(5)

Amount of N2 in containment when isolated assuming that the Drywell is at 135 F and 30Ã humidity and the Wetwell is at 90 F and 100K humidity.

Net N2

= 3.53 x 10 moles 5

(6)

Analysis of H2 is based on a dry sample.

(7)

All hydrogen from metal water reaction is released into containment.

(8) It is mixed uniformly throughout drywell and wetwell.

B-2

68

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APPENDIX C

OBTAINING A REPRESENTATIVE SAMPLE FOR ESTIMATING THE CORE'S CONDITION POST-ACCIDENT (1)

Coolant samples may be obtained from the:

o Reactor vessel at jet pump 810 or 820 instrument tap o

RHR downstream of Pump A or B

o Suppressi on pool (2)

Atmosphere samples may be obtained from the:

o Drywell above seal plate (hydrogen sampling) o Drywell below seal plate o

Wetwell Before withdrawing a sample consi der the following (this is not an all inclusive list):

(1)

Have Safety Reli ef Valves

(~RVs) lifted?

(2)

Temperature and pressure in Drywell and Wetwell.

(3)

Is HPCS operating'?

(4)

Is LPCS operating?

(5)

Are RHR pumps operating?

Which mode?

Shutdown cooling (normal ) mode.

Shutdown cooling (LPCI) mode.

Suppression cooling mode.

Drywell or wetwell spray mode.

(6)

Drywell radiation levels.

(7)

Reactor vessel water level (fuel zone level monitor).

(8)

If recirculation pumps are not operating it is necessary to raise.

water level so the flow through the core -overflows the moisture separators

(+51" or 15" above normal water level) to be able to obtain a representative sample of.the jet pump.

(One exception is

. for a large liquid line break where there is a reverse flow through core.)

C-2

OBTAINING A REPRESENTATIVE SAMPLE OF THE CORE COOLANT TO ESTIMATE CORE DAMAGE Break Cate or

  • RCS pressure No break or small break Sam le Location to be Used Jet Supp.

RHR Drywell Wetwell

~Pum Pop 1

~Pum ATMS ATMS Other Instructi ons Hi press.

Lo'press.

Yes Yes Yes o

RHR in shutdown

, cooling mode Large Break Li uid line Hi press.

Lo press.

Large Break Steam Line Hi press.

Yes Yes o

RHR in Suppres-sion pool cooling o

RHR in shutdown cooling and/or suppression pool cooling modes.

o RHR in suppres-sion pool cooling AlOde Lo press.

Yes Yes o

RHR in shutdown cooling mode

~

Notes:

A - Use if SRVs are not vented B - Use if SRVs vent to suppression pool C - Use if make-up water is 50/ of core D - Use if make-up water is 50Ã of core flow flow C-3

~

~

APPENDIX D This Appendix contains:

o Reference Plant Parameters o

-Baseline fission product (f.p.) concentrations in Reactor Coolant and Drywell atmosphere under normal conditions o

Core Fission Pro'duct Inventory o

Normalization of Measured Fission Product Concentrations REFERENCE PLANT PARAMETERS (WNP-2)

Power Level Fuel bundles 3323MWt

= 764 Mass Reactor Coolant

= 2.74 x 10 g

Coolant

+ Suppression Pool

= 3.44 x 10 g Drywell Atmosphere Core operating time

= 1095 days Total containment Volume

= 5.68 x 10 cc

= 9.71 x 10 cc 9

Table D-1 Base Line FISSION PRODUCT CONCENTRATIONS IN REACTOR COOLANT AND DRYWELL ATMOSPHERE DURING REACTOR SHUTDOWN UNDER NORMAL CONDITIONS Reactor Coolant Ci/

Dr ell Atmos here Ci/cc Isotope Upper Limit Nominal Upper Limit Nominal I-131 29 0.7 Cs-137 Xe-133 Kr-85 0 3 0.03 a

10-4 a

4 x 10 b

10 6

4x10 aObserved experimentally, in an operating BWR-3 with MK I containment, data obtained from GE unpublished

document, DRF 268-DEV-0009.

bAssuming 10%%d of the upper limit values.

Release of Cs-137 activity would strongly depend on the core inventory which is a function of fuel burnup.

D-2

Table D-2, CORE INVENTORY OF MAJOR FISSION PRODUCTS IN WNP-2 PLANT OPERATED AT 3323 MWt FOR THREE YEARS 1095 DAYS Chemical Grou

~?seto e*

Half-Life Hours Major Gamma Ray Energy

( Intensi ty) 10 Ci KeV (//d Noble Gases Halogens Alkali Metals Kr-85m Kr-85 Kr-87 Kr-88 Xe-133 Xe-135 I-131 I-132 I-133 I-134 I-135 Cs-134 Cs-137 Cs-138

4. 48 9.39 x 104 1.27
2. 84 126.

9.09 193.

2. 29 20.8

.877 6.59 1.80 x 104 2.64 x 105

0. 537 22.

1.0 43.

61.

184.

24.

87 127.

183.

201.

172.

18; 11.

162.

151(0.753) 514(0.0044) 403(0.495) 196(0.26),

1530(0.109) 81(0.365) 250(0.899) 364(0.812) 668(0.99),773(0.762) 530(0.86) 847(0.954),884(0.653) 1132(0.225),1260(0.286) 605(0.98),796(0.85) 662(0.85) 463(0.307),1426(0.76)

Noble Metal s Mo-99 RU-103 Tellurium Group Te-132 66.0 946.

167.

141.

78.0 125.

228(0.88) 740(0.128) 497(0.89)

Alkaline Earths Rare Ear ths Refractor ies Sr-91 Sr-92 Ba-140 Y-92 La-140 Ce-141 Ce-144 Zr-95 Zr-97

9. 52
2. 71 307.

1406.

40.2 780.

6830.

1104 16.8 105.

112.

157.

107.

167.

147.

117 147.

151.

750(0.23),1024(0.325) 1388(0.9) 537(0.254) 934(0.139) 487(0.455),1597(0.955) 145(0.48) 134(0.108) 724 (0. 437), 757(0. 553) 743(0.928)

  • Only the representative isotopes which have relatively large inventory and considered to be easy to measure are listed here.

~ At the time of reactor shutdown.

D-3

NORMALIZATION OF MEASURED FISSION PRODUCT CONCENTRATIONS The measured isotope concentrations in the coolant and airborne activity. as received from the chemical lab must be normalized to referenced plant conditions for use in Appendix E to determine core damage.

The referenced plant core fission product inventory is WNP-2 operating at 3323 MWt for 1095 days.

The coolant mass and the contained air volume are given under Plant Parameters (page D-1).

The following four correction factors as determined in steps (1), (2), (3) and (4) are applied to the measured individual isotope concentrations as shown in step (5) to give concentrations normalized to the reference plant.

(1)

Correction for Core 0 eratin Histor (FI)

The inventory of fission product isotopes given in Table D-2 is for WNP-2 core operated at 3323MW for 1095 days at zero decay.

o For short-lived radionuclides if the last operating period equals more than 5 half-lives then correct for power 3323 F IP P

where P

= operating power level.

o For longer lived radionuclides correct for actual length of operating period and the power level.

Where g5

=

1 e 51095 3223 0.693 and t

= half life (days)

Operating period in days.

If the operating period exceeds 5 half lives this correction is not needed.

(E.g., if the operating period (T) exceeds 40 days no correction needed for I-131 (t1 2

= 8.0 days) or any isotope with a shorter half life.)

D-4

o For very long-lived radionuclides (i.e., Cs-137, Kr-85, Sr-90)

This takes into account the full core history of operating periods, power levels and shutdowns (see example calc.).

3323 1-e Z

P 1e J

e ~

J J

= summation of all operating periods Pg

= Power level during operating period J

(MWt)

Tg

= Operating time for period 0 (days)

T

= Time since end of operating period J

J SAMPLE CALCULATION FOR OPERATING HISTORY Assuming a reactor has the following power operation history:

Operation Period 1A 1B Days Since Startup 1-60 61 - 70 Operation Time T> (day) 60 TO J

254 Average Power P~

(M~~t) 1000 0

2A 2B 71 - 270 271 -'00 200 44 2000 0

301 - 314 14 3000 o

For I-131

= ~~

= 0.0862 day 0.693

-1 F

3323 ( 1

-0. 0862xl095

-e I(I-131) 1000(1

-0.08 x60) -0. 862x25

+ 2000(1

-0.0862x20

)

-0.0862x44 (1

-0.0862x 4) -0.08 2xO 3651 107 D-5

o For Cs-137

= 6.29 x 10 da I(Cs-137)

-5 3323 1

-6.29xlp x1095

-6.29xl0 x60i -6.29xlp x254 Rppp(1

-6.29xlp x200) -6.29xlp x44

-6.29xlp x14i -6.29xlp xp 243.16 3.

+ 24.93

+ 2.6 D-6

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(2)

Correction for Radioactive Deca (FD) (Also see Figure D-1)

C F

D C

x T 0.693 C

= Ce

=

C x e 1/2 C

= Concentration, ~Ci/g or ~ Ci/cc, at time of shutdown C

= Concentration at time measured (t) after shutdown T = time sample concentration is measured (hours after shutdown).

t1/2 half life - hours R adi onuc 1 i de t 1/2 Half-life Hours Kr-85 Kr-88 Xe-133 Xe-135 1-131 I-13 Cs-134 Cs-137 9.39 + 4*

2. 84 126.

9.09 193.

20.8 1.80 + 4 2.64 + 5 7.38-6 0.244 5.50-3 0.0763 3.59-3 0.0333 3.85-5 2.63-6

  • Note:

9.39+4 is same as 9.39 x

104 (3)

Correction for Gas Sam le Vial Tem erature and Pressure, (FTP) cxTv TP P

x Tc v

c vTP C, P, T

= Concentration, Pressure and Temperature in in containment CyPyTyConcentrati on,PressureandTemperature 1 n vial D-8

(4)

Measured Concentrations c

Measured concentrations reported from the Chemical Lab include:

C

= reactor system coolant - aCi/g c

C

= suppressi on pool water - pCi/g s

Cd,= drywell air

- pCi/cc C

= Suppression Pool Air - aCi/cc sp When both the reactor coolant concentration (C

) and the suppression pool concentration (C

) are measured, they are to be averaged to give a total water concentration (CT).

(C x 2.74 x 10 g) + (C x 3.17 x 10 g)

C

=

c s

(3.44 x 10 g)

When both the drywell concentration (Cd) and the suppression pool free air concentration (C

) are measured, they are to be averaged to give a containment air concentration (C t).

(C x 5. 75 x 10 cc ) + (C x 4.08 x 10 cc )

C d

s ct (9.83 x 10 cc)

(5)

Normalized Concentration (CNW - coolant, CNG -,gas)

To determine the normalized concentration, multiply the measured concentration by the appropriate correction factors shown below:

CNW CT x FI x FD Reactor coola'nt

+ suppression pool

=

C x F> x FD Reactor coolant

=

C x Fi x FD Suppression pool

C~G

=

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=CdxF< xFDx TTP Containment (total )

0rywe 1 1

=Cs xF> xFTp Suppression pool Free Air volume D-10

APPENDIX E

The following figures are plots of the release fraction for individual isotopes from the gas gap or from a fuel melt condition for WNP-2 as if it had operated at 3323 NWt for 1095 days.

This serves as the normalized condition.

The solid line indicates the "best estimate"'

with dash lines indicating (3) the upper and the lower limit.'rom Appendix D part 5, enter the airborne"normalized concentration (CN<)

in Ci/cc or the coolant normalized concentration (C<G) in uCi/g along the ordinate and determihe the percent cladding failed or the percent fuel melt along the abscissa.

There are four sets of figures covering eight individual fission product radi onuc1 i des as follows:

Drywe1 1 5.68 x 10 cc Kr-85 Kr-88 Xe-133 Xe-135 Half Life 10 yr 2.84 hr 5.25 day 9.11 hr Containment 9.71 x 10 cc Kr-85 Kr-88 Xe-133

'Xe-135 Reactor Coolant 2:74 x 10 I-131 I-133 Cs-134 Cs-137 Half Life 8.04 day 20.8 hr 2.06 yr.

30 yr.

Reactor Coolant Plus Suppression

Pool, 3.44 x 10 I-131 I-133 Cs-134 Cs-137

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APPENDIX F If a cladding failure condition exists then only the fuel gap release would be expected. If there is fuel melting then the core inventory release would be expected. The ratios of specific short lived noble gas isotopes to Xe-133 and of specific short lived iodines to I-131 will be significantly different for. each condition. The expected ratios are given below in Table F-1. A comparison of actual measured ratios to these values will assist in determining which condition exists. Table F-1 RATIOS OF ISOTOPES IN CORE INVENTORY AND FUEL GAP ~Isoto e Kr-87 Kr-88 Kr-85m Xe-133 Half-Life 76.3 m 2.84 h 4.48 h 5.25 d Activity Ratio* in Core Inventor 0.233 0.33 0.122 1.0* Activity Ratio in Fuel Ga 0.0234 0.0495 0.023 1.0 I-134 I-132 I-135 I-133 I-131 52.6 m 2.3 h 6.61 h 20.8 h 8.04 d 2.3 1.46 1.97 2.09 1.0* 0.155 0.127 0.364 0.685 0%

  • Ratio Noble as isoto e concentration for noble a

Xe-33 concentration Iodine isoto e concentration for io I-concentration

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