GUIDELINES FOR DETERMINING FAILED FUEL

AT OCONEE NUCLEAR STATION

Don Anthony Assistant Engineer

Duke Power Company

April, 1984

9409050333 840830 PDR ADOCK 05000269 P PDR

TABLE OF CONTENTS

Page

I. OBJECTIVE 1

II. TECHNICAL BASIS 2

III. CASE I NON-OVERHEATING CONDITION 4

A. Assumptions 4

B. Theory 4

C. Initial Conditions 10

D. Parameters To Check 10

E. Power Changes 10

F. Iodine Spiking 10

G. Water Density Correction Factor 13

And Final Equations For Failed Fuel

IV. CASE II FUEL OVERHEATING WITHOUT FUEL MELT 15

A. Theory 15

B. Parameters To Check 20

C. Power Corrections Factor 20

D. Core Inventory 21

E. Sampling Locations 25

F. Decay Correction 27

G. Failed Fuel Estimate Based on 31

Radio Chemistry Data

H. Area Monitors 32

I. Hydrogen Concentration In The 39

Containment Building

V. CASE III OVERHEATING WITH FUEL MELT 42

A. Theory 42

B. Initial Conditions 46

TABLE OF CONTENTS

Page

C. Parameters To Check 46

D. Failed Fuel Based On Radio Chemistry 46

Samples

E. Area Monitors 48

VI. CONCLUSIONS 51

VII. REFERENCES 52

VIII. APPENDIX 53

g/

.. OBJECTIVE

The purpose of this document is to give guidelines for a procedure to determine core damage under accident conditions or at normal operating conditions. This document emphasizes the use of radio chemistry data from the RCS, sump and containment for iodine and noble gas isotopes. However, other plant data shall be utilized to give indications of core damage.

1

II. TECHNICAL BASIS

Based upon experimental evidence, the amount of fission products released from a fuel pellet is proportional to the temperature of that fuel pellet. There are five release mechanisms; burst release, diffusional release from the pellet-to-cladding gap inventory, grain bounda p)release, diffusion from the U02 grains and release from molten material. 1 Each of these release mechanisms occur at a different temperature and indicates severity of damage to the core. (See Table 1) This document establishes three categories - No overheating, overheating but no fuel melt*, and overheating with fuel melt.

*Note: Overheating is defined as any condition in which RCS temperature and pressure indicate coolant is superheated.

2

TABLE 1

EXPECTED FUEL DAMAGE CORRELATION WITH FUEL ROD TEMPERATURE

Fuel Damage Temperature oF*

No Damage < 1300

Clad Damage 1300 - 2000 Ballooning of zircaloy cladding > 1300 Burst of zircaloy cladding 1300 - 2000 Oxidation of cladding and hydrogen generation > 1600

Fuel Overtemperature 2000 - 3450 Fission product fuel lattice mobility 2000 - 2550 Grain boudary diffusion release of fission 2450 - 3450

products

Fuel Melt > 3450 'Dissolution and liquefaction of U02 in > 3450

the Zircaloy - Zr0 2 eutectic Melting of remaining U02 5100

These temperatures are material property characteristics and are non-specific with respect to locations within the fuel and/or fuel cladding.

3

III. CASE I NON-OVERHEATING CONDITION

A. Assumptions

1. All Iodine and Xenon isotopes are at equilibrium.

2. All Iodine isotopes in the RCS pass through a 90% efficient demineralizer at the rate of one cooland volume per day.

3. There is no plate out of iodine in the RCS.

4. The noble gases are equally mixed throughout the RCS and - consideration is not given to noble gases that may be in the letdown storage tank or pressurizer.

5. The reactor is operating at 100% power - 2568 MWT.

B. Theory

The production of fission products in the fuel for Iodine can be described by the following equations.

Rate of change = production - loss

dnF. = F Y -X . N Fi - y. N (1) ___ R i i L Fi dt

Where

dn_. = Change in the number of atoms

dt of isotope i in the fuel

FR = Fission rate (fissions/sec)

Y. = Fission yield (atoms/fission) 1

X = Decay constant (sec1) of i th isotope

y = Leak rate coefficient (seJ1) of i th isotope

N = Number of atoms of i th isotope in the fuel F

If the reactor is at equilibrium then dnF =0 dt

and equation 1 becomes

0= F Y. -A. N -yN R 1 1 Fi i Fi

or

4.

N .= F Y. N Fi R 1 i (2)

1 i

Equation 2 shows the relationship between the number of atoms of fission products in the fuel and fission rate. Since power is a function of fission rate, a correction factor can be added to Equation 2 to correct for power level.

N-. = PF Y (3) Ex R i Ux. + -Y )

Where P = Percent power/100 Fissions FR = Fission rate at 100% power (8.025 x sso

R sec

If it is assumed that the number of atoms leaking from the fuel to the RCS is proportional to the fraction of power required to produce these atoms, then Equation 3 becomes

N Fi =:FR PY (4) (A i)

Where = Fraction of power used to produce fission products in the failed fuel

N is redefined as the number of atoms of isotope i produced in the faxled fuel.

The equation for the number of atoms leaking to the RCS of the i th isotope of iodine is as follows:

dN . . N . - A. N . - BE. N . (5) ci yi Ft 1 ci *i ci. dt

Where Nci = The number of atoms of the i th isotope in the RCS

BE = Removal rate of i th isotope from the RCS due to filtration

If dNi =0, then Equation 5 becomes

dt

0= yi NF - y. N. - BE. N ci 6

5

Substituting Equation 4 into Equation 6

_(X + BE)M(A + Y) Nc (7) yPF R C

Since it is difficult to measure the number of atoms, N c is converted to activity by the relationship

a(t) = AN (t)

Where a(t) = Activity (dis/sec)

Equation 7 becomes

(A 7 +BE)( +y) AN iyP R FY C

or

5.= yPX FRY c(8

It should be remembered that this is a simplified example of one isotope. However, most isotopes are not only produced by the fissioning process but also by the decay of other isotopes. If the special case of iodine isotopes decaying to xenon isotopes is taken, then Equation 1 becomes

dXeF =SPFR Y + A I - A - y Xe - o Xe (9) F__ R xe I F xeF xe F t axe F dt

Where Xe = Xenon atoms in the failed fuel

I = Iodine atoms in the failed fuel

= Thermal Flux

aaxe = Microscopic absorption cross section of Xenon

Assuming dXeF = 0 Then

dt

0 =PF R Yxe I F xe Xe F yxe XeF t aaxe Xe (10)

From Equation 4

IF =-P FRY I (X I + Y I)

and Equation 10 becomes

6

0=SPFR Yxe + PFR xe F xe XeF t axe Xe

FR xe I I xe xe +t axe) Xe

Xe =PF xe II (11)

F axe F

0, xe +xe t axe xe t a

The modified equation for the number of Xenon atoms in the RCS is

dXe xe Xe + A I -X Xe a r Xe (12) c Ye F I c xe c t a c (2 dt

Where r = Fraction of time xe is in the core

Assuming dXef = 0 Equation 12 becomes

dt

0 = yxe Xef + Ic xe Xec ta Xec (13)

Substituting Equation 11 into Equation 13

0 = yxe PFR xe + I Y + X1 xe Xec (14) (Xxe +Yxe +ta)

-o r 1Xe t axe c

Solving for 5 equation 14 becomes

S "xe +xe + toaxe) (xe Xec I c t axe r Xec) (15) Y P F Y + A y xc R xe I I

Since activity = AN(t), Equation 15 becomes

= xe xe teaxe)( teaxe ae - ac (16)

xe 7 yxe P FR[Y + X I Y I

7

If X, y, and Y are assumed to be constants (see Table 2), then equations 8 and 16 simplify to the following equations:

= 1.826& x 10-5 (a - a )/P (17) xel33 xel33 1133 P(7

£ xe135 = (11.609) [(2.1024 x 10-5) + P(1.3180 x 10 4)](1.0 + 0.5669P)

axe1 35 - al1 351/P (18)

[1131 = 3.1975 x 10 (al1 3 1)/P (19)

&1133 = 2.2185 x 10 (a1 33)/P (20)

1135 = 4.6639 x 10 (a1135 )/P (21)

Where r = 0.09

ot = 5.0 x 1013 Neutrons/cm2 - sec

a = 2.64 x 10-18 cm2 axe135

NOTE: The to term for 1131, 1133, 1135, Xe133 was very small compare8 to the other terms in the equation.

8

TABLE 2

IODINE AND XENON CONSTANTS

Isotopes Y Half-Life X (sec'l) y (sec')

Xel33 0.0677 5.27 Days 1.522 x 10-6 1.0 X 10-7

Xe135 0.0672 9.20 Hrs. 2.0924 x 10-s 1.0 x 10-7

1131 0.0277 8.04 Days 9.9762 x 10-7 2.0 x 108

1133 0.676 20.80 Hrs. 9.2548 x 10-6 2.0 x 108

1135 0.639 6.70 Hrs. 2.8731 x 10-s 2.0 x 10-8

9

C. Initial Conditions

Normal operating conditions at any power level or shutdown with no unusual conditions prior to shutdown. Adequate core cooling is maintained.

D. Parameters to Check

1. Correlation of CET's and cladding temperature below the saturation region (Figure 1).

2. No significant loss of coolant or pressure.

a. Pressurizer level remains constant with heaters operating normally.

b. Make up remains constant.

c. Water level in sump or quench tank not increasing.

3. Little or no hydrogen in containment.

E. Power Changes

Since the equations in Section IIB assumes equilibrium, a period of four half-lives must pass at the new power level before radiochemistry samples are taken and counted. For example, a sample should not be taken and counted for 1-135 until about 27 hours after a power change. When possible, an activity count should be made on the following isotopes - 1131, 1133, 1135, Xe133, Xe135. Before a failed fuel estimate is made, an average activity for each isotope should be calculated. If possible, this average should be for 30 days of constant power. If this is not possible or there is a significant increase in activity then use as many days as possible after equilibrium is established or the activity level stabilizes. Always try to base failed rod estimates on at lease one iodine isotope and one Xenon Isotope, and as many radiochemistry samples as possible.

A general rule to remember is that the shorter the half-life of the isotope used to calculate failed rods, the greater the estimate of damage will be. For example 1135 will indicate more failed fuel than 1131.

F. Iodine Spiking

Whenever there is a change in power, an iodine spike will occur. It is believed that this spike is the result of a pressure difference between the RCS and gap. By comparing pressure to activity in the RCS, a spike begins when the pressure is decreased. The greater the pressure drop, the greater the spike. This spike should not be misinterpreted as an increase in failed rods. If an estimate of failed fuel is required after shutdown then radiochemistry data prior

10

to shutdown should be used. If any of the conditions of Section IID are exceeded or in the judgement of person using this procedure that major fuel damage has occurred then use the procedures in Case II fuel overheating without fuel melt.

0

11

FIGURE 1 (2)

CORRELATION OF CORE EXIT AND CLADDING TEMPERATURES

2600

2400

2200SUPER HEAT TCLAD > 1400F

REGION 2000

1800

(0 1600C', 0j

g 1400

cr. 1200- TCLAD > 1800F

1000

800

o,, 600.

400

200

400 500 600 700 800 900 1000 1100 1200 1300

CORE EXIT THERMOCOUPLE TEMPERATURE (F)

12

G. Density Correction Factor for Radiochemistry Samples

Since radiochemistry samples are measured in pCi/gr and total activity in the RCS is needed then a correction factor for density of water must be added to equations 17 thru 21. See Table 3 for correction factors.

&xel33 = (1.826 x 10- 5)(X)(Xe 33 - I 133)/P (22)

Sxel35 = (11.609)(X)((2.1024 x 105) + (1.3180 x 10 4 )P) x ((1.0 + 0.5669P) Xe - I)]/P (23)

1131 = (3.5219 x 103 )(X)(I1 3 1 (24) P

1133 = (2.349 x 103 )(X)(I1 33) (25) P

S 1135 = (4.8018 x 103 )(X)(I1 35) (26) P

Where 1131 133' 135, Xel33, and Xel35 is

given in (pCi/gr)

5 = Fraction of rods failed

X = Correction factor from Table 2

P = Power fraction (if reactor at 50% power then P = 0.5)

System volume = 3.3 x 108 cm3

An estimate of the number of rods failed can be calculated by the equation.

Failed Rods = (36,316)

The percent of failed fuel can be obtained by the equation % failed fuel = 100 x:5.

It should be noted that each equation will estimate a different number of failed rods. These equations should be used to determine a range of failed fuel rods. This range can be used to give a high estimate, a low estimate and an average estimate.

13

TABLE 3

Density Correction Factor, X, for NC Temperature Changes

Find the appropriate NC System average temperature at the time of accident. Find the approximate temperature at which the NC samples are taken. The intersection of both numbers is the density correction factor, X.

NOTE: Normal NC System sample temperature is approximately 90 OF. Use this temperature if no other information is available.

NCS Sample Temperature o

80 90 100

100 .996 .998 1 150 .983 .985 .987 200 .966 .968 .970

E 0 250 .945 .947 .949 300 .921 .923 .924 350 .894 .895 .897 400 .862 .864 .865

o 450 .827 .828 .830 500 .787 .788 .790 550 .739 .740 .741 560 .728 .729 .731 570 .717 .718 .719 580 .706 .708 .708 590 .693 .694 .695 600 .680 .681 .683

14

IV CASE II FUEL OVERHEATING WITHOUT FUEL MELT

A. Theory

As discussed in Section II, there are five release mechanisms for fission products from the fuel. In a fuel overheating without fuel melt situation, three of the five release mechanisms are involved. The first release mechanism will be the Ballooning and bursting of the Zircaloy cladding. This condition will begin at a cladding temperature of about 1300 0 F to 1400 0F. Relating this temperature to CET temperature, th Teactor coolant would be in the superheat ,region of Figure 2 . 2 As long as the reactor coolant stays on or below the saturation line of Figure 2, no damage should occur to the reactor core. Once Figure 2 indicates that the cladding temperature is greater than 1400 0 F but less than 1800 0F, then the rate of ballooning and bursting of zircaloy cladding will increase. As the cladding temperature approaches 1800 0 F the zirconium metal-water reaction will begin. The higher the temperature the greater the reaction rate. When Figure 2 indicates that the cladding temperature is past 18000F, then it can be assumed that there is major damage to the cladding and possible damage to the fuel.

Figure 3 is a plot oz ercent inventory released from fuel versus temperature of fuel. This figure shows that the inventory released increases rapidly with temperature increase. For this reason, it is difficult to determine which release mechanisms are in process. Therefore, this document will assume that once overheating begins, all three release mechanisms are involved. It is believed that during the Three Mile Island accident, all the fuel rods were damaged but little fuel actually melted. In follow-up test at TMI, it was determined that 40 to 70 percent of total halogens and noble gases were released from the gap and fuel pellets. Therefore, this document will assume a linear relationship between failed fuel and percent of fission products released. If 100% of cladding fails then 40 to 70 percent of the halogens and noble gases are released. If 1.0% of cladding fails then .4 to .7 percent of the hal glns and noble gases will be released. (See Table 4 and Figure 4.)

15

FIGURE (2)

CORRELATION OF CORE EXIT AND CLADDING TEMPERATURES

2600

2400

2200

SUPER HEAT TCLAD > 1400F REGION

2000

1800

(.3 1600C,, 0j

2: 1400

Q. 1200- CLAD >

(n 1800F o -J

1000

800

600

400

200

400 500 600 700 800 900 1000 1100 1200 1300

CORE EXIT THERMOCOUPLE TEMPERATURE (F)

16

FIGURE 3

GAP INVENTORY VS. TEMPERATURE

100

9590

85 *

800. 75

0 7065

a w 60

w 55 - TC 131 w 50

>- 45ac XC 135 0 40z w 35

2 30

252015- KR 87

105 - -XC 135

0-1

C4 N C4 C4 N4(N N C4 N C4 (N C (NC

TEMPERATURE (oK)

BURNUP= 30,000 MWD/MTU TIME 420 DAYS

17

TABLE 4

PERCENT ACTIVITY RELEASE FOR 100 PERCENT OVERTEMPERATURE CONDITIONS

Nuclear Min.* Max.* Nominal** Min.*** Max.***

Kr-85 40 70

Xe-133 42 66 52. 40 70

1-131 41 55

Cs-137 45 60

Sr-90 0.08****

Ba-140 0.1 0.2 0.15 0.08 0.2

* Release values based on TMI-2 measurements.

Normal value is simple average of all Kr, Xe, I, and Cs measurements.

*** Minimum and maximum values of all Kr, Xe, I and Cs measurements. .*** Only value available.

18

FIGURE 4 (4)

100. I I i l II I 1 1 7 0 .. 70..

30 .

20.. .

* //

0.57

0.2.

Fuel Overtemperature()

RELATIONSHIP OF % FUEL OVERTEMPERATURE WITH % CORE INV7ENTORY RELEASED OF XE, K(R, I, OR CS

19

7 ?

B. Parameters to Check

1. Figure 2 in the superheat region or higher.

2. Pressurizer water level changing (either up or down).

3. High radiation in any of the following

a. Containment Building

b. Quench tank

c. Letdown storage tank

4. Water level increasing in the sump or quench tank.

5. Hydrogen concentration increasing in the containment building.

C. Power Correction Factor

Since all core inventories were calculated assuming 100% power, a methodology was developed by Westinghouse to correct for power changes. The equations for correcting power are as follows:

1. Steady state power prior to shutdown.

a) Half-life of nuclide < 1 day

Power Correction Factor = Average Power Level (Percent) for Prior 4 days Rated Power Level (100)

b) Half-life of nuclide > 1 day

Power Correction Factor = Average Power Level (Percent) for Prior 30 days Rated Power Level (100)

c) Half-life of nuclide 1 year

Power Correction Factor = Average Power Level (Percent) for Prior 1 year Rated Power Level (100)

Steady state power condition is assumed where the power does not vary by more than ±10 percent of rated power level from time averaged value.

An example of the power correction factor is as follows:

Assume Oconee 1 has run at 55 percent power for the past 60 days.

55 Power correction factor = - 0.55 100

Note that this correction factor is good for nuclides with a half life less than one year.

20

2. Transient Power History

-A.t. -X.to. Power Correction Factor = 1. P. (1-e 1) e 1 3

.71t

RP (1-e j 3)

Where

P. = Average power level (Mwt) during operating period t.

RP = Rate power level 'of the core (Mwt)

t = Operating period in days at power P. where power does not vary more than ±10 percent power of rated power level from time averaged value (P.)

Xi = Decay constant of nuclide i in inverse days.

to. = Time between end of period j and time of reactor shutdown in days.

If the total period of operation is greater than four half-lives of the nuclide being considered, the power correction is as follows. This is within the accuracy of this methodology.

0.t. > .0693

-ct. -. to Power Correction Factor = I P. (1-e 1 J) e 1 3

RP

For the few nuclides with half-lives around one year or longer, a power correction factor which ratios effective full power days to total calcendar.days of cycle operation is applied.

Power Correction Factor EFPD total calendar days of cycle operation

It should be noted that all equations were derived assuming 100 percent power and that this correction factor is a method of determining core damage when the reactor has operated at less than 100 percent power.

D. Core Inventory

In order to determine a conservative core inventory for Oconee, three LOR2 computer runs were made. (See Appendix A) All three runs assumed an enrichment of 3.3%. Each run represents a different burnup region of the core. (i.e., one run assumes fuel used for 3

21

cycles, another run assumes fuel used for 2 cycles and the last run assumes fuel used.for 1 cycle.) Each region assumed 59 assemblies. Table 5 gives activity level for one fuel assembly for each region. Table 6 gives total activity in the core and compares these values to FSAR values. Most of the core values are close to FSAR values except for Xel33m and Xe135. It is possible that this difference is the result of the higher enrichment value used in the LOR2 runs.

0 22

TABLE 5

ACTIVITY PER FUEL ASSEMBLY

1 Cycle 2 Cycles 3 Cycles Isotope (Curies) (Curies) (Curies)

Kr85 2.102(3)* 3.272(3) 4.525(3)

Kr87 2.264(5) 1.433(5) 1.550(5)

Kr88 3.206(5) 2.030(5) 2.194(5)

Xe133 3.483(5) 6.335(5) 3.161(5)

Xel33m 1.610(5) 9.016(4) 1.164(5)

Xe135 4.714(5) 3.973(5) 4.499(5)

Xel35m 1.610(5) 1.255(5) 1.669(5)

1131 3.982(5) 3.075(5) 4.066(5)

1133 8.469(5) 6.317(5) 8.134(5)

1135 7.869(5) 5.879(5) 7.603(5)

Ba139 7.608(5) 5.561(5) 7.051(5)

Ba140 7.429(5) 5.432(5) 6.331(5)

Bal41 6.955(5) 5.073(5) 6.392(5)

Pr145 4.432(5) 3.177(5) 3.950(5)

Pr146 3.437(5) 2.537(5) 3.200(5)

* 2.102(3) = 2.102 x 103

23

TABLE 6

TOTAL CORE ACTIVITY

LOR2** FSAR Isotope (Curies) (Curies) % A

Kr85 5.8405(5)* 5.84(5) 0.0

Kr87 3.0958(7) 4.00(7) -29.207

Kr88 4.3837(7) 5.60(7) -27.775

Xe133 1.355(8) 1.28(8) 7.800

Xel33m 2.1686(7) 3.07(6) 85.84

Xel35 7.7798(7) 2.19(7) 71.85

Xel35m 2.6751(7) 3.31(7) -23.73

1131 6.5626(7) 7.42(7) -13.065

1133 1.3523(8) 1.28(8) 5.340

1135 1.2597(8) 1.27(8) -0.82

Ba139 1.193(8)

Bal4o 1.165(8)

Ba141 1.087(8)

Pr145 6.820(7)

Pr146 5.442(7)

* LOR2 - FSAR LOR2 x 100 * LOR2 = 59 (cycle 1 + cycle 2 + cycle 3)

where cycle 1, cycle 2, cycle 3 is on Table 5

*** 5.8405(5) = 5.8405 x 10s

NOTE: FSAR values assume 400 EFPD and LOR2 values assume 421 EFPD

24

E. Sampling locations

In most accidents, fission products will be contained in three areas reactor coolant system, sump and containment building atmosphere. There are two other situations which would cause other sampling locations to be considered. One of these situations is a steam generator tube rupture. If a steam generator tube rupture is suspected then a radiochemistry sample should be taken from the secondary system. The other situation would be a stuck open PORV. If it is suspected that the PORV on top of the pressurizer has been stuck open for a long period of time, then some attempt should be made to estimate activity in the quench tank. A suggested list of sampling locations is given in Table 7.

During Crisis Management drills, the problem has arisen of not being able to obtain a representative radiochemistry sample in the RCS due to steam or hydrogen bubbles. If this is the Case and a LOCA is assumed to be the cause and the pressurizer PORV is not stuck open, then a failed fuel estimate can be made based on the noble gases in the containment atmosphere. This statement is based on the assumption that noble gases are always going to concentrate at the highest possible elevation and that noble gases are not soluble in water.

25

TABLE 7(4)

Suggested Sampling Locations

Principal Other Scenario Sampling Locations Sampling Locations

Small Break LOCA Reactor Power > 1%* RCS Hot Leg, Containment RCS Pressurizer

Atmosphere Reactor Power < 1r* RCS Hot Leg** RCS Pressurizer

Large Break LOCA Reactor Power > 1%* Containment Sump, Containment

Atmosphere, RCS Hot Leg Reactor Power < 1%* Containment Sump, Containment

Atmosphere

Steam Line Break RCS Hot Leg RCS Pressurizer Containment Atmosphere

Steam Generator Tube RCS Hot Leg, Secondary Containment Rupture System Atmosphere

Indication of Signifi- Containment Sump, Containment cant Containment Sump Atmosphere Inventory

Containment Building Containment Atmosphere, Radiation Monitor Alarm Containment Sump

Safety Injection RCS Hot Leg RCS Pressurizer Actuated

Indication of High RCS Hot Leg RCS Pressurizer Radiation Level in RCS

* Assume operating at that level for some appreciable time. If a RCS hot leg sample is unavailable and a RCS cold leg sample is available, obtain a RCS cold leg sample. However, for a cold leg sample to be a good representation of the RCS, the primary water should be circulating through the system.

26

Activity of liquid samples can be calculated by the following

equations:

RCS Activity (Ci) = Specific Activity (pCi/gr) x 108 Ci/pCi x

RCS Volume (cm3) x water density (gm/cm 3 )

RCS Activity = Specific Activity (pCi/gm) x density (gm/cm3 ) x 330

Where volume is at system temperature and pressure

Sump activity (Ci) = specific activity (pCi/gm) x water mass in sump

Sump activity (Ci) = specific activity (pCi/gm) x 106 Ci/pCi x 2.79 x 104 cm2 x Height (cm) x density (gm/cc)

Where Density is at sump temperature

If the sump has been flooded due to the injection of water from the

BWST then two different methods can be used to calculate failed

fuel. The first and easiest method is to assume that all the noble

gases from the failed fuel are in the containment air. A failed

fuel estimate would be made based on an air sample for the Xe

isotopes. The second method is to assume that a large break LOCA

has occurred and that any water injected into the RCS will leak to

the Reactor Building floor. An estimate of the volume of water from

the BWST injected into the RCS should be made. An estimate of the

volume of water leaked from the RCS should also be made. The

containment floor activity can be calculated by the following

e q u a t i o n : ( ~ / M 0 6 C /~

Containment Floor (Ci) = specific activity (pCi/gm) x 10 ci/pCi

Activity of sump

x Density of (gm/cc) x Volume of + Volume of RCS

Water LBWST injected water leaked

The density is at sump temperature

The containment air sample can be calculated by the following

equation:

Air sample activity (Ci) = specific activity (pCi/cm3) x Containment

Volume (cm3 ) x correction factor x 106 Ci/pCi

Air sample activity = specific activity (pCi/cm3 ) x 5.188 x 1010 cm3

x 10-6 Ci/pCi x P2 P1 T 2

Air sample = specific activity (pCi/cm 3 ) x 2 1 x 5.188 x 10

P1 T2

27

Where T1 , P1 = sample temperature & pressure

T2, P2 = containment atmosphere temperature and pressure

Temperature is in oR and pressure is in PSIA

Total Activity = RCS activity + sump activity + containment activity

F. Decay Correction

The specific activity of a sample is decay adjusted to time of

reactor shutdown using the following equation.

Specific activity at shutdown = Specific activity (measured) -X.t

1 e

Where:

A. = Radioactive decay constant, 1/sec t = Time period from reactor shutdown to time of sample analysis,

sec.

Since this correction may also be performed by some analytical equip

ment, care must be taken to avoid duplicate correction. Also, consi

deration must be given to account for precursor effect during the

decay of the nuclide. For this methodology, only the parent-daughter

relationship associated with the methodology. The decay scheme of the

parent-daughter relationship is described by the following equation.

Q BA Q A(e -e ) + Q e B AB- A A B

Where:

Qo = Activity (Ci) or specific activity (pCi/gm or pCi/cc)

of the parent at shutdown

Q = Activity (Ci) or specific activity (pCi/gm or pCi/cc) of the daughter at shutdown

QB = Activity (Ci) or specific activity (pCi/gm or pCi/cc) of the daughter at time of sample

xA = Decay constant of the parent, sec1

AB = Deacy constant of the daughter, sec

t = Time period from reactor shutdown to time of sample analysis, sec.

28

Since the activity of the daughter at sample time is due to the decay of the parent and the decay of the daughter initially released at shutdown, an estimation of the fraction of the measured activity at sample time due to only the decay of daughter is required. To

use the above equation to determine the fraction, an assumption is

made that the fraction of source inventory released of the parent and the daughter at time of shutdown are equal (for the nuclides

used here within a factor of 2). The following steps should be followed to calculate the fraction of the measured activity due to the decay of the daughter that was released and then to calculate

the activity of the daughter released at shutdown.

1. Calculate the hypothetical daughter concentration (QB) at the time of the sample analysis assuming 100 percent release of the

parent and daughter source inventory.

SAt t

Q =Q 0(e A - e + Q e B XB A A B

Where:

Q = 100% source inventory (Ci) of parent, Table 6

Qo 100% source inventory (Ci) of daughter, Table 6

QB(t) Hypothetical daughter activity (Ci) at sample time

K = If parent has 2 daughters, K is the branching factor, Table 6

xA = Parent decay constant, sec 1

AB = Daughter decay constant, sec 1

t = Time period from shutdown to time of sample, sec.

2. Determine the contribution of only the decay of the initial

inventory of the daughter to the hypothetical daughter activity

at sample time.

oe B

QB (t)

3. Calculate the amount of the measured sample specific activity

associated with the decay of the daughter that was released.

MB = Fr x measure specific activity (pCi/gm or pCi/cc)

4. Decay correct the specific activity (MB) to reactor shutdown.

MB B -At

e B

29

TABLE 8

PARENT-DAUGHTER RELATIONSHIPS

Parent Daughter Parent Half Life* Daughter Half Life* K**

Kr-88 2.8 h Rb-88 17.8 m 1.00

1-131 8.05 d Xe-131m 11.8 d .008

1-133 20.3 h Xe-133m 2.26 d .024 1-133 20.3 h Xe-133 5.27 d .976 Xe-133m 2.26 d Xe-133 5.27 d 1.00

1-135 6.68 h Xe-135 9.14 h .70 Xe-135m 15.6 m Xe-135 9.14 h 1.00 1-135 6.68 h Xe-135m 15.6 m .30

Te-132 77.7 h 1-132 2.26 h 1.00

Sb-129 4.3 h Te-129 68.7 m .827 Te-129m 34.1 d Te-129 68.7 m .680 Sb-129 4.3 h Te-129m 34.1 d .173 O Ba-140 12.8 d La-140 40.22 h 1.00

Ba-142 11 m La-142 92.5 m 1.00

Ce-144 284 d Pr-144 17.27 m 1.00

Table of Isotopes, Lederer, Hollander, and Perlman, Sixth Edition ** Branching decay factor

30

G. Failed fuel estimate based on radiochemistry data.

After all radiochemistry sample have been taken and corrections made due to decay, the following equations can be used to estimate core' damage.

1. Iodine and Xenon

Total Activity for Isotope low - (0.7)(Power correction factor)(Isotope core inventory)

Total Activity for Isotope -..high (0.4)(Power correction factor)(Isotope core inventory)

= Il131 low 13

1131 (Y) (4.5938 x 107)

i high I131

1131 (Y) (2.6250 x 10')

31

=o 1133

Iow3 (Y) (9.4661 x 107)

.i=h1133

1gh (Y) (5.4092 x 107)

low 1135 1I35 (Y) (8.8179 x 10')

high 1135 =1135 (Y) (5.0388 x 107)

low = Xe133

Xe133 (Y) (9.4850 x 107)

high= Xe133

Xel33 (Y) (5.4232 x 107)

lo= Xe135

Xel35 (Y) (5.4459 x 107)

h= Xe135 Xe135 (Y) (3.1119 x 107)

=low Bal40 Ba140 (Y) (2.33 x 10')

high Bal40 (Y) (2.6250 x 107)

Where 1131, 1133, 1135, Xe133, Xe135, Ba140 is the total activity (Ci) for each isotope*.

Y = power correction factor

NOTE: Isotope core inventory values are list in Table 6

* Total activity can be calculated using Section IIE Sampling locations.

H. Area monitors

Generally, a radiochemistry sample will give a more accurate indication of core damage than area monitors in the containment building. However, radiochemistry samples take a long time to

32

evaluate, where as area monitors give results immediately. This section will attempt to make some simplifying assumptions and give a rough estimate of failed fuel versus dosae ratein containment. It will be assumed that only noble gases are in the containment atmos phere.* The noble gases are also assumed to be equally distributed throughout the containment building.

X = (2.62 x 10s) X Ey x 3600 sec/hr

X = (9.432 x 10X) x Ey R/hr

Where-: X = Ci/cm3

Ey = Average energy of all y-rays per disintegration

X = Dose rate (R/HR)

Table 9 list the average gamma energy level for the most prominent noble gas isotopes. Table 10 shows the methodology for calculating total noble gas dose rate. Table 11 gives a high estimate and low estimate of dose rate for 100% failed fuel without fuel melt at various times during shutdown. Figure 5 is a plot of dose rate from Table 10 as the noble gases decay.

An approximation of failed fuel can be determined by the Equation:

m Xm

(Y) X(t)

Where: Xm = Area monitor reading in the containment (R/HR)

X(t) = Dose rate from Figure 5 (R/HR) at the appropriate time after shutdown

Y = Power correction factor

Where Y = Average Power for Prior 30 days Rated power level

m = Fuel failure fraction according to area monitors

It should be noted that this equation assumes a "PUFF" release of noble gases. If a small break LOCA occurs then the failed fuel estimate of m will be low. One possible method for using this equation during a small break LOCA is to wait until the.monitor dose rate peaks and starts to decline. Remember to use Figure 5 to account for decay.

33

*It is understood that more isotopes than noble gases are released to the containment. However, modeling which isotopes and their activity is difficult. Therefore only noble gases are considered. This will give a conservative estimate of failed fuel.

34

TABLE 9

Ey Isotope (Mev) Half-Life

Kr85m 0.151 4.4 Hrs.

Kr85 0.00211 10.76 Yrs.

Kr87 1.37 76.0 Min.

Kr88 1.74 2.79 Hrs.

Xel33m 0.326 2.26 Days

Xe133 0.030 5.27 Days

Xel35m 0.422 15.70 Min.

Xe135 0.246 9.20 Hrs.

35

TABLE 10

Activity in Containment

Isotope At Shutdown Ey X. X

Kr85m 4.8405 (5)* 0.00211 9.3302(-6) 18.569

Kr87 2.0958 (7) 1.370 4.0397(-4) 5.22(5)

Kr88 3.0686 (7) 1.740 5.9148(-4) 9.71(5)

Xe133 9.3558 (7) 0.030 1.8034(-3) 5.10(4)

Xe133m 1.5180 (7) 0.0326 2.9260(-4) 9.00(4)

Xe135 5.4459 (7) 0.246 1.0497(-3) 2.43(5)

Total dose rate at shutdown = 1.88(6) R/HR

* 4.8405(5) = 4.8405 x 105

36

TABLE 11

DOSE RATES AFTER DECAY

Time After High Estimate Low Estimate Shutdown Dose Rate Dose Rate

(HRS) (R/HR) (R/HR)

0 1.88(6)* 1.08(6)

6 6.32(5) 3.61(5)

12 4.00(5) 2.29(5)

24 2.49(5) 1.42(5)

* 1.88(6) = 1.88 x 106

37

FIGURE 5

DOSE RATE VS. TIME FOR FUEL OVERHEATING

WITHOUT FUEL MELT

2.0

1.8

1.6

1.4

x 1.2

2

1.0

0.8U1 0

0.6

0.4

0.2

0.0 I I

0 6 12 18 24

TIME (HOUR)

38

I. Hydrogen concentration in the containment building

At approximately 16000 F zirconium reacts with water to produce hydrogen. The greater the temperature the faster the reaction rate. During the zirconium - water reaction heat is also released which raises the cladding temperature which increases the reaction rate. If the hydrogen concentration is constant or increasing slightly without recombiners on, then the cladding temperature is probably around 1600 0F or less. If hydrogen concentration is increasing rapidly (with or without recombiners) then the clad temperature is above 1600aF. A rough estimate of core damage can be made, based on hydrogen concentration in the containment if the following assumptions are made.

1. All hydrogen produced in the RCS is released to the containment building.

2. All hydrogen in the containment building comes from the zirconium - water reaction*.

3. The recombines have not been turned on (i.e., no hydrogen has been burned).

The equation for the zirconium - water reaction is

Zr + 2H20 -+ Zr0 2 + 2H2

or

Two moles of hydrogen in the containment building are produced by the reaction of one mole of zircunium in the core.

At STP 1 mole of hydrogen has a volume of

22.4 . or 0.79 ft3

Volume of hydrogen in containment = hydrogen concentration

(volume precent unit) X containment free volume

or

H2 H Vcontainment where VH2 is the volume of hydrogen in containment as a percent of atmosphere.

Tl1 T2

There are other sources of hydrogen, but assuming all hydrogen is produced by the zirconium - water reaction will give a

* conservative estimate.

39

*V = P V T

PSTP cH, TP

PSTP Tc

V = P T F V STP c STP H2 c T P c STP

Where VSTP STP, STP = Volume, temperature, and pressure at STP

T *=4920 R STP

PP = 14.7 PSI

V = Containment free volume = 1,832,033 ft3

V = P 492 (1,832,033) (F ) STP c -- 14.7 c

VSTP = c F (6.1317 x 107 )

T c

The total amount of hydrogen moles in the containment = VSTP Volume of one mole

MH P F 6 .1317 x 10 = P F (7.7616 x 107) HT 0.79 T TT

c c

Since it take 1 mole of Zr to product 2 mole of H2 then the number of zirconium moles reacting with hydrogen is 1/2

or

=Zr = 1/2 MH = 1/2 PC H (7.7616 x 107)

T c

The zirconium mass that reacts can be calculated by the equation

Zr = MZr x W

Where W = gram - Atomic Weight = 91.22 gr/mole

Z = (MZ) (91.22) = (P F ) (3.5401 x 109)

T

*rTZ

40

The fraction of zirconium that reacts with water is calculated by

H2 tot

Where Zrtot = Total amount of zirconium in the core = 8.1204 x 10 gm

H2 c H 3.5401 x 10 = P F 2 (43.594) 100T 8.1204 x 10T - --- ,

c c

Where H = Fraction of core damage 11 2

P = Containment pressure (PSIA)

T = Containment temperature (OF + 460)

FH2 = Percent of hydrogen in containment atmosphere

It should be noted that when estimates of core damage are made using radiochemistry samples, area monitors and hydrogen concentration that the results can be greatly different. Whenever possible, all three methods should be used and their combined results used as an indication of core damage.

41

V. CASE III OVERHEATING WITH FUEL MELT

A. Theory

In a fuel melt condition all five release mechanisms discussed in Section II are involved. As fuel melts, up to 99% of the halogens and noble gases will be released. There will also be a significant release of barium and praseodymium. As in Case II, a linear relationship between failed fuel and isotope activity will be assumed. (See Figures 6 and 7.)

The major difference between Case II and Case III is the percent of fission product inventory released from the fuel. The methodology for correcting isotopic decay and reactor power remains the same. The methodology for using hydrogen concentration to estimate core damage remains the same. The main changes will be in the radiochemistry method and area monitor method.

42

TABLE 12

PERCENT ACTIVITY RELEASE FOR 100 PERCENT CORE MELT CONDITIONS

Large* Small* Nominal** Min.*** Max.*** Species LOCA Transient* LOCA Release Release Release

Xe 88.35 99.45 78.38

Kr 88.35 99.45 78.38

87 70 99

I 88.23 99.44 78.09

Cs 88.55 99.46 78.84

Te 78.52 94.88 71.04

Sr 10.44 28.17 14.80 24 10 44

Ba 19.66 43.87 24.08

Pr 0.82 2.36 1.02 1.4 0.8 2.4

* Calculated releases for severe accident scenarios without emergency safeguard features, taken from draft NUREG-0956

** Normal release are averages of Xe, Kr, I, Cs, and Te groups, or Sr and Ba groups.

*** Maximum and minimum releases represent extremes of the groups.

43

FIGURE 6

100.0 I 1

10.0

CU

1.0

0.1

0.01

1.0 10.0 100.0

Fuel Melt (%)

RELATIONSHIP OF % FUEL MELT WITH % CORE INVENTORY RELEASED OF BA OR SR

44

FIGURE 7

1000 70-'7 70.

50.

* 3.7

0.5.~

74

2. 0. o

0.1lo

. 77

0.70

,* * . . .7 .1

Fuel Mel t(%

RELATIONSHIP OF % FUEL MELT WITH % CORE 0INVENTORY RELEASED OF XE, KR, I, CS, OR TE

45

B. Initial Conditions

1. Abnormal shutdown of reactor.

2. Belief that core uncovered for a long period of time.

3. Case II calculations indicate more than 100% failed fuel

C. Parameters to Check

1. Figure 2 indicates cladding temperature past.18000F.

2. Very high radiation in any of the following.

a. Containment building atmosphere

b. Quench tank

c. Letdown storage tank

3. High water levels in the sump.

4. Very high concentration of hydrogen in the containment atmosphere.

5. High levels of barium and praseodymium in radiochemistry samples.

D. Failed fuel based on radiochemistry samples.

The following equations can be used for core damage estimates based of iodine and xenon.

Total Activity for Isotope low (Power correction factor) (Isotope core inventory)

Total Activity for Isotope J high (0.7) (Power correction factor) (Isotope core inventory)

S 1131 low (Y) (6.5626 x 107)

1131

1131 high (Y) (4.5938 x 107)

1131

1133 low (Y) (1.3523 x 101)

1133

1133 high (Y) (9.4661 x 107) 1133

46

1135 low (Y) (1.2597 x 108)

1135

1135. £ high ( (8.8179 x 10')

1135

Xel33 low (Y) (1.3523 x 108)

Xel33

Xel33 high- (Y) (9.4906 x 10')

Xel33

Xe135 low (Y) (7.7798 x 1O)

Xe135

Xe135 high ( (5.4453 x 107) Xel35

For Barium, the equation will be:

Total Activity for Isotope

low - (0.44) (Y) (Isotope core inventory)

Total Activity for Isotope S high (0.10) (Y) (Isotope .core inventory).

BAl39 low (Y) (5.2492 x 107)

BA139

BA139 high (Y) (1.193 x 710)

BA139

BA140 low (Y) (5.126 x 10')

BA139

BA140 high (Y) (1.165 x 10')

BA140

BA141 low (Y) (4.7828 x 10')

BA141

47

BA141 high (Y) (1.087 x 10')

BA141

The equations for praseodymium are as follows

Total Activity for Isotope low (0.024) (Y) (Isotope core inventory)

Total Activity for Isotope high (0.008) (Y) (Isotope core inventory)

Pr145 Pr146 low (Y) (1.6368 x 106) low (Y) (1.306 x 106)

PR145 Pr146

Pr145 Pr146 high (Y) (5.456 x 10b) high (Y) (4.3536 x 10b)

Pr145 Pr146

Where 1131, 1133, 1135, Xe133, Xe135, = Total Activity of Isotope (Ci) BA139, BA140, BA141, Pr145, Pr146

Y = Power Correction Factor

E. Area Monitors

The methodology for using area monitors to determine core damage due to fuel melt is the same as in Section II I. However, dose rate is based on 70 to 100 percent release of noble gases instead of the 40 to 70 percent used in Section II I. The dose rates after decay is listed in Tabel 13. Figure 8 shows a plot of dose rate versus time for 100% failed fuel. An approximation of failed fuel can be determined by the equation:

S m - m

(Y) (X(t))

Where X = Area monitor reading in the containment (R/HR)

X(t) = Dose rate from Figure 8

Y Power correction factor -Average power for prior 30 days Rated power level

m = Fuel failure fraction

48

TABLE 13

DOSE RATES AFTER DECAY FOR FUEL MELT

High Estimate Low Estimate Time After Dose Rate Dose Rate Shutdown (R/HR) (R/HR)

0 2.708 x 106 1.896 x 106

6 9.029 x 10s 6.320 x 10s

12 5.725 x 105 4.007 x 105

24 3.562 x 10s 2.493 x 105

49

FIGURE 8

DOSE RATE VS. TIME FOR FUEL MELT

2.8

2.6

2.4

2.2

2.0

1.8

Co 0 1.6x

1.4

uj

O 1.2

1.0

0.8

0.6

0.4

0-2

0.0

0 6 12 18 24

TIME (HOUR)

50.

VI. CONCLUSIONS

It is believed that utilizing the methods set forth in this document will result in a rough estimate of core damage. At this time, there are many areas where little is known about the release mechanisms of fuel and the transport of fission products. Some areas where more work should be done are; calculating background activity in RCS due to tramp uranium, iodine spiking, mechanisms affecting leak rate coefficients, and transport mechanisms of fission products in the containment building. With more work and more information the simplified equations of this document could be solved assuming non-equilibrium conditions and developed into a computer program which would produce more accurate results.

0

51

VII REFERENCES

1. "Technical Basis for Estimating Fission Product Behavior During LWR Accidents", NUREG-0772, June 1981.

2. "Operator Training - Degraded Core Recognition and Mitigation", TRC-81-3, Babcock & Wilcox.

3. "Methods for Calculating the Fractonal Release of Volatile Fission Products from Oxide Fuel", ANSI/ANS-5.4 - 1982.

4. "Post Accident Core Damage Assessment Methodology", Westinghouse, February, 1984.

5. "Introduction to Nuclear Engineering", John Lamarsh, 1977.

52

VIII. APPENDIX

ISOTOPIC DATA FROM LOR2

53

-~ .i Aw.,

.414

v7 A---- - - ~

.4 _ _ _ _ _ _ _ _ _ _ _ - O -- O ---- -............3. .T

7 - YC ' UED4 S EMrL~~~' - -~ T.- -1--E - - -. P "-"-- D" -T- -* __ _ __ _ __ _ __ __. .,., ~ ~ 6 * " ~ w .. ~ ~ ._ 7;7~

.

11--11112 2 2 22 2 3 3 3 3 3 4 4 4 4 4 5 5 , a 2

3= '7.7 - - - -

lo----..r--~ -.-.-j.- -.-.,...---- -~ --.~.,

1t 0-- , i,2 0 5 ,E 2 0 2 8 1 00. 0., 0 * * * * * * - 0 ....

73 .y*~ * r - . . ..... -t:...

2 2_ _ _ _ _ _ ..- - - - - - -.... .. .. ...

... .. ... .. .. ... .. ... .. Y ..-~ -- -~- . . ._---..,.. ..b. ____________________________________ --. 3...-:7 7 = 7 r. . . . a

4_ _ _ _ _ _ _ _ _ _ _ .. ,. . . - . - . 1 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _' 1'. . . . , . . . ~ 1 ~ ~ i- - -

4. 8..77. . . . - 0 = 3 .

S3 .... ...

.. . .

... .. . .. ..- ....... .

-, ....... ....

.300~L -,_.7_ r.~u

.............. K.. .. . . . . . . . -

I-oN

L0R2-VER~-1~2-------------1---W--i- 04105/8TERU40/0/8

vr I . E+43N/CM2 7E..

-- NUCLIDt-RAOIOACTIVITV,4CURIES :j- : *..~..-.332E+O~1.349.*5.3657+O4.23E+014.426E- .,4.87E303,.5..895E-O7,-7.136E-1 1 0. -0.. 0. -T14-- --.. E6.78E 0--1259, 2E02 3 78 0E-0 :5 E.286 E-1 0-1.80E 15 0.m-- 00

=!E13 6- 8 5 +05-400 0 00 -

IE13-0. 14 530E+047l 00

Is 139-0.-- '17 0.-. 5.1E+5- 79 055.8E0 530EO .. 3 +O...7EO,42 E0,.56 0.709+0..7 0 D

",1 0 C 40 . 0..0 ,113 77777 . ~~. 4 E 0 ' 0 08~ -a. -i 0. 0. 0. -- - - --- .........

p...!..-E141 - 0.-84 3 26E! 0.-4 -.- , Z .D 0l -1-10 s .02 27 14-"0- 0 'M- 0.., *.- 0......Fp0. 0.0 '3~~~E -f. -- 4! 10-:9w7r~

6"-1130 -O.------i 1.3 6E+03 -- ,259E03-; 1. r627E.2w'63-81E00.-+-.919E-0-03.-51E-02-.,----97E02-9018E--1-1.84-.030. ----- -- -.O XE~ - -80i- ZET -I 0,,-. -3-2 &?4-023ram,35.5?2+0-1 0!40725-O--.8E-4-:29Id . .2 0 5.5.7E0 .. 54+5536E0,503-0,472+5289E+05-.3.56E+85-7009E+u4u.27uE0? o..

IE04 12 6-0 - 0- 0. -'- -0. .

136 - : == -. ;

2 2

*~~~~~ ~~ 1CLE00E.FE.ASEMBi' L . Y*AE RUN 04/05/84".. .A tf i ....... fl

... ..- NUCLID-- RAIACTIV.CUIES-CRE.,..,,.... ....

KR~ -8 0. luD OD365.---------KR~ ~ 85-0.--20E Z-OI&-2-343-2.13o- OEt3.l3E+O I. 0E"~3 :2.103E*03-2 100203 -2.066E+03 1:972E+03 K5-4-6-2-9-E- 4--.---..-051 09E*O5,4 .4J11 E E4032-,966E+03-4.-61 9E+02-7-1-91 E+,01-8.-5 70E-- 1 2 0...-,....-(..-..

, - R - 8 6 - - 0 . 2 .L2 6 4 E + 0 5 % , . .1 8 7 .i + 0 5 - 8 . -5 8 4 ai + o 3 - 3 . 2 . 1 9 E 4 o - t 1 2 0 m + 0 1 f 2 : 5 2 6 .~.0 - - - - . . . . . . . . . . -:89=0 _ -8 957E-07 0 _ .......... a. K RE+O 8 53 5E+02-4...02E 01-2i.270E+0.66 10-.0........ KR-758E-03-8ii +0 .5.686E-2l'--0.-- 0.--

KR 90 -, 0 07.--.70.=

00 KR-9-. . O 0 -0........~20p

., RB 88 -. . . - . 3 2 E +O 2 O .6468 + O. & .L 4. 3 E o4.,1::7 . + 0L 41 9 E + 3 53 E 0 2 9 5 Ol S ,o 6 0 5 2 0. . ..... 23 R -9 ! 52 _0 '~0. .

: - : - . ---- - 649E% o+s-00 . 0 0 0. -to--- .*.. -'-'- 0.-- -.

44B-6 0 -0 .. k . . 0.-y 0 , ,0. 8-~~9d--~~-0 8 L.. .. 0__ _ _ _ 0 23,ti-0-2-l'0 0: .4 E 0 3-I I +0 1 62E-'-----.---. ----- ':,*I........

= RI8 '- . 4.453E!45 - 43 e6,4. .;-4-5 kf 6 43 A-5 .2E,06- 4 -41- I. 6"-40-... 4 4E 06 ;45 -061 4- 3 4 E!O-688 4E-616 8 .8 -0 89 0.-3.251E+0 5 2 64 6E L C)5 .14 3 E 4. 8, 99 -9.5 :x 91 .536E+00- . " -.- 0.4-L5 %.r 64 E 0 L -i9 0 09" 1 1E - 06 -;14. 6 8E 23 E 3 8 - 4 0 . 69 lE 50o .290- 2 0 .---- __.......

as___ a.I.-.G.4 0-? E. 020*

07

I. 42,

!r__ ItEou

0"v.g U.. '. __~ oITIi XE13-~0 ~ 7.21E05~i?6tE4~0 . ~~0. T, XE1 8-.. . . 07 Ill,- 01 31---.6#- 1 Ali114E-4-- . 2 - ,- ------ 0v -

XE13-0.--------5.494~+05 p.' ~p .~ ---~'-------1.1 0E+5-0.--~~- .0.0-"~-----.-.--~------.- 00.-------0-

XE14----0.------- ~ 0~- --- ~'---"0. ----- 0...........-. X~l6-.0. ___ ___ .91~ +0~.O .- 0..&.,~,,.~O...0..ol~ , ~~~~~~~~. ~ ~ ~ ~ 1 XE4--. 42.6~~~'- ~ ~ ---- o2

-0-i.--6 o,..-q*E01 -4&E22-44t22------3:2E03- 2

Si 37-27

SRU=?.0 139+W D: 0DA.- 0.- RU 04/05/80.4.

~~~ .......~ . . ~ .- . 8 * .. 0.0

= X 3 - 5 21 E~+f 0 43 U. 22T 0.O- 'o'~ 0 0. .3

.42 E2 S139- .7 ~ - -- ~ ~ 0 0 70 - - ~ -o ..... 0 ___ __ _ _ _ _ 0*1 08 0 T .040 + -0.. .- !' 0.

.. ,...8A 3 . - - 0. . . . : .. %- -*.0 8-s-s.---.- 30. . .. , ( . . 0. - j......0. . . .. 0.- 0........ ........

X 39 .- - - . . 030l+4 .i 3E 03 .0 84E4 -2, . 72e o0 .A 3 E-40- 3 E-05 0 . .. . . 0.

XE 1414k 052 05 0.5 1 1 at: 80 + 5 ; 0 . *-., . . . 0 0. - ~ ~ u ~

.05A066 E~ '00o 00

R.

44

?j %, ur. ______ __ Q:

~ .~ *. .- s

a,'. .HA~f---OIS-H~g~.~0"' ' .7~

Y U .

E-R E 1S4WH0--.- U R N U.. .....- *. +- 0 0.-R N 04 0 4....... 0.3066E0 mw

738+289910 .51

I: S14-0........ - 1~ S.83760..3E+3 0.i..Z 08-2... .O1-8...6E4 06E 6 0...........0 ~ 00 ~ ~~~~~~: 07; 0:911-.4 1 '-.*- ... ~ - 0.--..... ....... N81 31 - 0 *3 3 0~7-9~4~ 0 - - -- 9 ~ 3 8 1 3 3 0 * ~ ~

*~~~7 -,--'-.-..... 0 0. Sal 22 3e.--E+l,3:,!Lt10-275E0 -3.073E -2,23- E-02-2.232E-0--.11 E0 -. 3601- 0 .69 7E0

*43*~~0 .. .. -,. . .. . A... -. ~ . 0.

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