ML20209G472

Rev 1 to Qualification of Reactor Physics Methods for Application to Prairie Island Units

ML20209G472
Person / Time
Site:	Prairie Island
Issue date:	12/31/1983
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML20209G419	List: ML20209G415 ML20209G472
References
NSPNAD-8101-A, NSPNAD-8101-A-R01, NSPNAD-8101-A-R1, NUDOCS 8702050373
Download: ML20209G472 (252)

Text

. g-PRAIRIE ISLAND  ;'

NUCLEAR POWER PLANT g 9_

1, j  ;

. QUALIFICATION OF REACTOR PHYSICS METHODS $

FOR p

. APPLICATION TO..PI-UNITS T NSPNAD-8101-A F Rev.1- b

__=

h 3-2 y

s 1

E December 1982 T ke k

f

}

^

~ T Sl7a. $'-

ig/ g se.

_V NORTHERN STATES POWER COMPANY  %

NUCLEAR ANALYSIS DEPARTMENT 35

) .

k ,4 3

- . . . . , , , , l ?

4

~g - , - -

. .a . -

.w. _::.. -

.~ r.e , hB 2 21983 ,

[

g.

o UNITED STATES NUCLEAR REGULATORY COMMISSION 2 I wAsmnavon. o. c.20ssa k..... -

FEB 17 1983

. Docket Nos. 50-282 '

and 50-306 y Mr. D. M. Musolf Nuclear Support Services Department Northern States Power Company A 414 Nicollet Mall - 8th Floor Minneapolis, Minnesota 55401

Dear Mr. Musolf:

The staff has completed the review of Technical Reports NSPNAD-8101P Rev.1 and NSPNAD-8102P Rev.1 (except for the C08RA-IIIC/MIT and a portion of the DYNODE P/3 Code) that was submitted by letter dated February 18, 1982 and revised by letter dated December 13, 1982. Our review of the COBRA-IIIC/MIT has progressed sufficiently to permit you to use the code for evaluating the thennal-hydraulic hot channel parameter for the Unit 2 Cycle 8 reload analysis. However, the mass and energy release data from the DYN00E P/3 Code serving as input to the CONTEMPT-LT Code for analyzing the containment pressure transient during a steam line break accident is not qualified since we have not reviewed OYN0DE for this application. ,

, We find the revised reports to be an acceptable method for analyzing future reloads and operation for the Prairie Island Nuclear Generating Plant, Unit Mos. I and 2 subject to these limitations and others listed

! in-the conclusion of our safety evaluation.

A copy of the Safety Evaluation is enclosed.

Sincerely, h

h obert A. Clark, Chief cw % ~

Operating Reactors Branch #3 Division of Licensing

Enclosure:

Safety Evaluation cc: See next page l

l W - -- 4 . e>geg w -e -= ee-g.-e+ese oese e e =*w a ese e..ee ye.e.,,..

---_-,er. -

y , . - - - - _ . - - - -

-p,,

~

Northern States Power Company cc:

Gerald Charnoff, Esquire Shaw, Pittman, Potts and Trowbridge 1800 M Street, N.W.

Washington, D. C. 20036 Mr. Louis J. Breimburst Mr. R. L. Tanner y Executive Director County Auditor Minnesota Pollution Control Agency Red Wing, Minnesota 55066 2 1935 W. County Road B2 *

. Roseville, Minnesota 55113 U. S. Environmental Protection Agency Federal Activities Branch Region V Office ATTN: Regional Radiation _

Representative 230 South Dearborn Street' Chicago, Illinois 60604 Mr. E. L. Watzl, Plant Manager Prairie Island Nuclear Generating Plant i Northern States Power Company Route 2 Welch, Minnesota 55089 Jocelyn F. Olson, Esquire .

Special Assistant Attorney General . .

! Minnesota follution Control Agency 1935 W. County Road B2

Roseville, Minneosta 55113

! U.S. Nuclear Regulatory Commission Resident Inspectors Office Route #2, Box 500A Welch, Minnesota 55089 Regional Administrator Nuclear Regulatory Commission, Region III Office of Executive Director for Operations 799 Roosevelt Road 4 Glen Ellyn, Illinois 60137 L

. ~ _ . _ _ _ , .

7 et

~ '

• p #" %

S UNITED STATES

'1 NUCLEAR REGULATORY COMMISSION

, wAsHtNGTON, D. C. 20555

/

%' .' .O...

1 '

1.

SAFETYEVALUATIdRBYTHEOFFICEOFNUCLEARREACTORREGULATION OF THE REACTOR PHYSICS AND RELOAD SAFETY EVALUATION METHODS TECHNICAL llEPORTS NSPNAD-8101P AND -8102P FOR THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT NOS. 1 ND 2 4

1

\.

r ,

e TABLE OF CONTENTS

. Page No.

1. Introduction 1

2. Summary of Reports 1 ,

f (A)' Qualification of Reactor Physics Methods 1 2

for Application to Prairie Island Units

. NSPNAD-8101P Rev. 1 (B) Reload Safety Evaluation Methods 3 for Application to Prairie Island Units _

NSPAD-8102P Rev. 1

3. Discussion and Evaluation 6 (A) Qualification f Reactor Physics Methods 6 for Application to Prairie Island Units NSPNAD-8101P Rev. 1 ,

(B) Reload Safety Evaluation Methods ,

10 for Application to Prairie Island Units ,

NSPAD-8102P Rev. I

1. General Physics Methods (Section 2 11 NSPNAD-8102P Rev. 1)

2. Safety Evaluation Methods (Section 3 12 NSPNAD-8101P Rev. 1) 4 Conclusion 25

5. Tables 27 .

l 29 s

6. Figures

7. References .

80 I . . .

1. Introduction Northern States Power Company (NSP) the licensee, submitted two Technical Reports titled " Qualifications of Reactor Physics Methods for Application 3 to PI Units NSPNAD-8101P Rev.' 1" and " Reload Safety Evaluation Methods for Application to PI Units NSPNAD-8102P Rev.1" by letter dated February 18, '

1, 1982 and revised by letter dated December 13, 1982. The licensee proposes to use these reports as guides to analyze future reloads and operations starting with Cycle 8 of Unit 2 (i.e. , scheduled in August 1983) at the Prairie Island Nuclear Generating Plant, Unit Nos.1 and 2. Report No.

NSPNAD-8101P addresses the qualification of reactor physics methods that cover reactor model description, qualification and quantification of -

reliability factors as applied to operation and fuel reloads for the Prairie Island Nuclear Generating Plant, Unit Nos.1 and 2. Report No. NSPNAD-8102P addresses methods for calculating specific physics parameters and their comparison to bounding values used in the postulated accident analysis.

This report also gives the licensee's experience on the safety analysis and calculational results for the Prairie Island units. The Revtsion 1 of both reports issued by the licensee's letter dated December 13, 1982 includes all of the staff's comments which resulted from the review of the initial versions of these reports. The purpose of our review is to ,

assure that the licensee has established acceptable methods for analyzing future reloads and operations at the P~rairie Island Nuclear Generating Plant, Unit Nos. 1 and 2.

2. Summary of Reoorts LA.' Oualification of Reactor Physics Methods for Acolication to Prairie Island Units NSPNAD-810lP Rev. 1 This report describes the reactor physics methods used by Northern States Power Company (NSP) for their analysis of the Prairie Island (PI) Nuclear Reactors. The report addresses the reactor model description, the quali-fication and quantification of reliability factors, and the application of the physics methods to both reactor operations and to reload safety evaluations.

The computer model used to analyze Prairie Island is the Advanced Recycle

!' Methodology Program (ARMP) developed by Nuclear Associates International Corporation (NAI) under the sponsorship of Electric Power Research l

v . ,

2-Institute (EPRI). The ARMP computer system is composed of individual computer programs including the following physics related ones:

1. EPRI-CELL, a spectral code which is used to generate initial and

~

burnup dependent nuclide concentrations and few group neutron cross ,

7-sections.

PDQ7/ HARMONY, a two-dimensional diffusion-depletion code which is 2 2.

used to calculate local

• peaking factors as well as for generating input data for the three-dimensional nodal code.

3. OP5, a three-dimensional nodal code which is a derivative of the -

EPRI-NODE-P code and is used'to obtain ' core power distributions as well as core physic,s, parameters such as control rod worths and reactivity coefficients.

4. CPM, a multi-groupicollision probability transport code which is used to provide lumped absorber data for burnable poisons and control rods.

5. CASMO II, a spectra.1 code which is used to generate initial and ~

burnup dependent nuclide concentrations and few group. neutron cross sections.

Two models based on ARMP are used to analyze the Prairie Island reactors.

The first is used for Cycles 3 through 6. The second model is identical to the first with the exception that the EPRI-CELL program is replaced with CASMO II which was developed by Studsvik Energiteknik AB. This model is used for Cycles 5 and 6 and will be used to analyze future cycles.

These models are verified by benchmarking against Prairie Island measure-ments for several cycles of operation. Reliability factors, describing the allowances to be used in safety related calculations to assure con-servatism, and uncertainty factors, describing the actual model. accuracy, are evaluated by direct comparison of calculations to experimental data.

The following physics parameters are addressed:

1. control rod worths (and critical boron concentrations),

2. isothermal temperature coefficient, ,

3. Doppler coefficient, hem um' u mMumu u

4. local, nodal and assembly power distributions,

5. burnup dependent isotopic compositions, and

' 6. delayed neutron parameters. ,

A For each parameter addressed, the data base is presented, including com-parisons between calculations and measurements as well as the uncertainty and reliability factors of the calculation. Conclusions are drawn re-garding the suitability of the model to perform the calculations.

B. Reload Safety Evaluation Methods for Aoolication to Prairie Island _

Units NSPNAD-8102P Rev. 1 This report describes the'c'alculational methods for the specific physics parameters used in each cycle reload for accident analyses. Section 2 of this report describes the procedures of these calculational methods for each specific physics parameter. These specific physics parameters and their comparison to bounding values used in the accident analysis are as follows: ,

Moderator Temperature Coefficient

1. ,

2. Power Reactivity Coefficient

3. Doppler Reactivity Coefficient 4 .' Boron Reactivity Coefficient

5. Shutdown Margin

6. Scram Reactivity

7. Nuclear Heat Flux Not Channel Factor i - 8. Nuclear Enthalpy Rise Hot Channel Factor l

9. Effective Delayed Neutron Fraction l,

l

10. Prompt Neutron Lifetime i

I

f Section 3 of the report describes the methods in which these physics parameters are applied to each accident or transient in order to deter-mine whether or not any reanalysis is required. For each accident or transient, the following are considered in the application of these physics parameters: , f

1. Definition of Accident - a brief description of the causes and 2 consequences. ,

2. Accident Analysis - a brief description of the methods employed and discussion of the sensitive physics parameters. Included is a list of the acceptance criteria.. --

3. Licensee's Safety ,Arla,1ysis Experience - a brief summary of the licensee's calculational experience and results of the comparison of their models to .the Prairie Island Final Safety Analysis Report.

4. Cycle Specific Physics Cal &ulations - a description of the ' specific physics calculations performed each cycle for the purposes of a re-load safety evaluation.

5. Reload Safety Evaluation - A description of the comparison of the cycle specific physics characteristics and the bounding values used in the safety analysis. Specific application of the model reliability.

factors and biases are also addressed.

t In addition, Section 3 of the licensee's report (NSPNAD-8102P Rev.1) addresses the specific accidents and transients which are considered in the safety evaluation methods. These accidents and transients are as follows:

- Uncontrolled Rod Cluster Control Assembly (RCCA) Withdrawal from Subcritical

- Uncontrolled RCCA Withdrawal at Power ,

- Control Rod Misalignment

! - Dropped Control Rod Uncontrolled Boron Dilution l

i l

l

- .- - _ , - . _ - _ _ . _ - -_.x. -

\

-S-

- Startup of an Inactive Loop

- Feedwater System Malfunction 3 - Excessive Load Increase .

g

- Loss of External Load

- Loss of Normal Feedwater

- Loss of Reactor Coolant Flow (Pump Trip)

Loss of Reactor Coolant Flow (Locked Rotor)

- Fuel Handling Accid'eIn ' '

- Main Steam Line Break .

- Ejected Rod .

- Loss of Coolant

- Fuel Misloading The licensee's report also describes the computer programs that are used to simulate the response of the nuclear steam supply system (NSSS) and the thermal-hydraulic response of the hot coolant channel and hot spot in the core for the above listed transients and accidents when applicable.

~

The computer programs that are used by the licensee in the analysis of these accidents and transients are as follows.

(a) DYNODE-P/3 . .

The DYNODE-P Version 3 program is used to analyze the transient response of the Nuclear Steam Supply System (NSSS). This program

- provides a simulation of the core average power, the core average fuel temperature and the core average coolant channel thermal-hydraulic responses.

- _ , - . _- . . . . _ _ , , - . . _ _ .. . . -_,_____,-.,-.c. - . _ . - . - - _ _ - - . , - _ . -

b (b) COBRA-titC/MIT '

The COBRA-IIIC/MIT is u thermal-hydraulic rehot channel in the core. sed to analy rods within the core.sponse of coolant on of the channe (c) CONTEMPT-LT c ated fuel e c

The CONTEMPT-LT program is used t 2 response of the containment during a o compute the temperature-pressure (d) T_000EE2 steam ifne break accident.

The T00DEE2 program.'is 'used to co rod and the associtted if the COBRA-IIIC nnels.

coolant chaof the hot fue ents This which program include is usedthe onlyhot fuel 3.

sponding to the 95% probability arture fromlimitnuclea ue corre-A.

Discussion and Evaluation

~ at 'a 95% confidence level.

' Qualification of Reactor Phys _

TsTand Units N57NAD;BI0IP .I 'Rev.1cs Methods for Acolicati on to Prairie We have reviewed the informati ARMP computer system was used tomethods ns and experiments. onal and com detailed code descriptions analyze the are Prairieareport Island emphasizes coresAlthough

, the thethe

" Advanced ProjectRecycle Methodology Pro Research 118-1 the ARMP system, per se, September 1977.ystem Documentation,"vailable in the EPRI ARMP d gram S ation, CCM-3

\

determining Cores. physics para, but rather metersthe f qualification of itTherefore, we did s use in or application to the Prairie Island Many of the computer programs i

industry-wide include th codes and, used therefore in the ARMP system are acceptabl

• e GAM, THERMOS, , requireand no CINDER additional c

programs review. e These whi h

\ form the neutron cross i

i i

l i

e e

~

section generator, EPRI-CELL, as well as the two-dimensional diffusion-depletion program, P0Q7/ HARMONY. We also find that the three-dimensional nodal code, DP5, is acceptably verified by comparisons to measured Prairie Island data over four cycles.of operation. CASMO II is a suitable replace-3 ment for EPRI-CELL for future cycles although the data base for Cycles 5 ,

and 6 is somewhat limited.

L Control rod worth value's are obtained from k-effectives and control rod positions computed by the nadal code. These are compared to worths obtained by measuring critical boron concentrations at various stages of rod insertion.

The standard deviation of the differences between measured and calculated critical boron concentrations, hence control rod worths, is 1.0% for the -

! CASM0/P00/DP5 model. For conservatism, however, a rod worth reliability factor of 10% is assumed . ,This is acceptable and is consistent with currently approved design methods.

Isothermal temperature

• coefficients are computed by varying the qore average temperature in three-dimensional nodal calculations while holding all other parameters constant. These are compared to isothermal tempera-ture coefficient measurenents made with the plant reactivity computer.

. Based on these comparisons, we conclude that the NSP ARMP model predicgs the isothermal temperature coefficient with a reliability of 2.9 x 10-Ak/k per degree F when the appropriate bias is included. This represeqts a 95% probability at the 95% confidence level that the measured isothermal temperature ccafficient is bounded by the predicted value. This is accept-

' able and is consistent with currently approved design methods.

T'he three-dicansional nadal code is used to calculate power coefficients as a function of power and exposure. The Doppler coefficient is then calculated by removing the moderator temperature coefficient component from the power coefficient. However, the uncertainty in the measure-ment of the Doppler coefficient with the plant reactivity computer is

too large to quantify a' reliability factor. However, U-238 resonance integrals calculated by CPM and by EPRI-CELL have been compared with experimental values obtained by AB Atomenergi and were well within the 4% experimental uncertainties associated with the measurements. Com-i parisons of temperature coefficients are also within the 10% experimental

uncertainties associated with the measurements. A 10% reliability

!- factor for the Doppler coefficient and a 20% reliability factor for the Doppler defect is, therefore, defined by the licensee. This is l

acceptable since it is consistent with currently approved design methods.

4 Also, comparisons of EPRI-CPM and CASMO calculations with experiments conducted in the KRITZ high temperature critical facility in Sweden corroborate the 10% reliability factor on the Doppler coefficient for - t j the CASM0/PDQ/DP5 model.

Either of the spectral codes, EPRI-CELL or CASMO II, produces initial j nuclide concentratiions, depletion and fission product chain dat4, and tables of microscopic and macroscopic cross sections varying with burnup.

Based on the good agreement between isotopic compositions calculated by both EPRI-CELL and CASMO and spent fuel isotopic data from Yankee Rowe and Saxton, we conclude that either method is acceptable for deter-mining isotopic compositions.

! Comparisons to measured. data spanning Cycles 3 through 6 of Prairie ,

Island operation are used to determine the uncertainties applicable i

to nodal power distributions. Since the reactor power distribution is not directly observable, the licensee's ARMP model has been used to cal-

. culate the actual incore detector signals in, each of the instrumented locations. These measured reaction rates are local values measured at

. 61 axial locations in each instrument thimble. The three-dimensional l calculational model using EPRI-CELL as the cross section generator were compared to reaction rate measurements over Cycles 3 through 6 for full power conditions representative of beginning, middle, and end of cycle.

Calculations using CASMO II as the cross section code were compared i

to reaction rate measurements over Cycles 5 and 6 for full power condi-i tions representative of beginning, middle, and end of cycle. The nuclear heat flux hot channel factor, F , is defined as the maximum local fuel 0

rod linear heat generation rate (LHGR) divided by the average fuel rod LHGR. It is computed by multiplying peak nodal values from the three-dimensional nodal DPS results by the two-dimensional P0Q7 pin-to-box ratios. The calculated and measured reaction rates are local values and are, therefore, proportional to F . The staff also agrees that the uncertainty determined from comparisoks of measurements and calculations l

for the instrument thimble and detector is the same as the uncertainty that would be determined for a fuel pin if that pin could be instrumented. -

j Therefore, the observed differences between calculated and measured flux

detector reaction ratos can be assumed to be the same as that which would '

be obtained between the calculated and measured local F values and can be used to determine the calculational reliability fact r for Fg.

I i

l l

l r

e ,

. .~ ~

. g.

The calculational reliability factor for Fg is defined as a single value, aFQ , such that the calculated value of FQ at any core location plus AFQ has a 95% probability at a 95% confidence level of being conservative with respect to the measured Fa at that location. A bounding reliability factor based on the value obtained at higher reaction rates has been used ,

3 by the licensee. The reliability factors determined are .067 with a .007 bias using the EPRI-CELL model and .062 with a .015 bias using the CASMO II model. The derivation and use of this bounding calculational reliability b factor for FQ is acceptable.

I The nuclear enthalpy rise hot channel factor, Fay, is defined as the ratio i of the integral of linear power along the rod on which the minimum DNBR _

occurs to the core average integral power. It is calculated by multiplying the nodal assembly average power values by the PDQ7 peak pin-to-box ratios.

As was the case with FQ,, the calculation reliability factor for Fay is based on comparisons of measured and predicted in-core flux detector signals. For FaH, however, the measured rate is obtained by inte. grating the 61 measured axial locations for each instrument thimble. Unlike the case for F 0, the Fag values were found to exhibit no dependence on power.

The reliability factors determined are .035 using the EPRI-CELL model and .044 using the CASMO II model. No calculational bias was cbserved.

The reliability factor is. applied as an additive factor to the calculated Fay since it is derived from the differences of absolute reaction rates. .

We find the derivation and the use of the calculational reliability factor for FaH is acceptable. -

The determination of the reliability factors for the effective delayed ,

neutron fraction, Seff, is somewhat different than that for previous parameters since these values can be calculated but not measured. The licensee determines the calculational uncertainty for Seff yb considering the experimental and calculational uncertainties in its major components.

The staff agrees with the assuned uncertainties and with the important components of Seff. Ve, therefore, find the presented reliability factor l of 4% for Seff acceptable and appropriate for other delayed neutron para-meters as well.

. In conclusion, we have reviewed the report within the guidelines provided by Section 4.3 of the Standard Review Plan. Included in our review was the description of the experimental data base, tne calculations performed,

• and the comparisons made to support the conclusions that the licensee's AR!iP computer model is adeouate to calculate steady-state physics para-

• meters for Prairie Island reload cores.

l l

l i

e w., - .,--,.- e- .m4 .i.- . , + - - . - - - - - ,- , _ -,, e-, -

Based on our review we conclude that the ARMP reactor physics methods used by the licensee and benchmarked against Prairie Island measurements over several cycles of operation are acceptable to be used in Prairie Island safety related calculations for the following parameters:

. t ,

a) control rod worths J

b) temperature coefficients c) Doppler coefficients d) burnup dependent isotopic compositions _

e) power -distributions .. . .

f) critical boron concentrations g) delayed neutron parameters We also find the detailed discussion and evaluation of the uncertainty and reliability associated with each physics parameter acc20 table. .The

' application of the reliability factors and biases for each parameter is

~

shown in Table 1.

We requested that the licensee continue the comparison of measured and calculated physics parameters during each cycle in order to provide con-tinued assurance of the model applicability. This request was discussed with and agreed to by the licensee.

Based on the above evaluation, we find the report titled " Qualification

(

of Reactor Physics Methods for Application to PI Units NSPNAD-8101P Rev.1" l

submitted by the licensee is an acceptable method for performing operations

' and fuel reload analyses for Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2.

B. Reload Safety Evaluation Methods for Acolication to Prairie Island Units NSPNAD-8102P Rev. 1 The licensee, in conjunction with Nuclear Associates, has developed various -

safety analyses for the Prairie Island Nuclear Generating Plant, Unit l

i l

t

~

~

Nos. 1 and 2. The licensee's report (NSPNAD-8102P Rev.1) shows represen-tative calculations applicable for operations and reload evaluations. Our evaluation covers the licensee's methodology used for performing general physics calculations shown in Section 2 titled " General Physics Methods" 9^ and the methodology for analyt!ng plant transients for 'a variety of reactor ,

and balance-of-plant malfunctiuns shown in Section 3 titled " Safety Evalua-tion Methods". Specifically our review covers. evaluations of the plant

-' accidents listed in Section 2B of this evaluation (Page 3) and the computer codes used to analyze these accidents. More than 100 graphical comparisons in Section 3 of the licensee's report showing the results of calculations using Westinghouse FSAR models versus the results of the licensee's computer codes were also reviewed by the staff. Our specific discussion and evalua- -

tions of the topics ccvered in the licensee's report NSPNAD-8102P, Rev.

1 and listed in this evalya, tion, Section 28 (Page 3), is as follows:

1. General Physics Met ads (Section 2 NSPNAD-8102P Rev.1)

The calculational methods used to determine the important cycle s'pecific safety-related physics parameters and the parameters themselves are des-cribed in Section 2, NSPNAD-8102P Rev.1. The physics parameters described are: ,

a. Moderator Temperature Coefficent .

b. Power Reactivity Coefficient .

c. Doppler Reactivity Cpefficient

d. Boron Reactivity Coefficient

e. Shutdown Margin .

i

f. Scram Reactivity - .

g. Nuclear Heat Flux Hot Channel Factor ,

h. Nuclear Enthalpy Rise Hot Channel Factor

- 1. Effective Delayed Neutron Fraction J. Prompt Neutron Lifetime

...,s ,---.,--,--.,-,,----..,,----,,------,.,-,.-,,-rr ,,,w.--, _-,,.,-.,,,m-,, .

-,w--.- . , -. .-- ,--..=.. ,,,r ,-.-.m-

r ..

We have reviewed the definitions of these parameters and find them con-sistent with both the staff and industry-wide practice and, therefore, are acceptable. The calculational procedures used to calculate these physics parameters are consistent with currently approved methods and (

are acceptable. .

2. Safety Evaluation Methods (Section 3, NSPNAD-8102P Rev. 1) j b

(a) Uncontrolled Rod Cluster Con' trol Assemblies (RCCA) Withdrawal from a Subcritical Condition 4

=

The uncontrolled RCCA withdrawal .from a subcritical condition is an An-ticipated Operational Occurrence for which the fuel integrity must not be violated. Therefore,.nos centerline melt nor DNB must occur. The event is analyzed using . input assumptions consistent with the Prairie r Island FSAR and the licensee's DYN0DE-P/3 computer code. The NSSS simu- L 1ation incorporates. poi'nt kinetics (including delayed neutrons and decay {

heat), fuel, clad, and gap heat conduction, and channel coolant themal- _

hydraulics. Reactivity effects due to moderator and fuel temperature (Doppler) variation as well as those due to control rod scram are in-

' cluded. A comparison with the FSAR results show that the licensee's model predicts higher temperatures while the nuclear power between the two calculations agrees well. Based on these results, we find the licen-see's methods acceptable for calculating the uncontrolled RCCA withdrawal transient from a subcritical condition if a cycle specific reanalysis is

, required.

Cycle specific physics calculations are performed at the most limiting core conditions during the cycle for the following parameters: (1) Doppler coefficient, (2) moderator temperature ' coefficient, (3) maximum reactivity insertion rate, (4) scram reactivity, and (5) effective delayed neutron fraction. Each of these parameters is adjusted to include the model reliability factors and biases derived in NSPNAD-8101P Rev. I which we find acceptable (note Section 3A of this SER). We find the key physics parameters selected to be appropriate for the transient. We also find -

the specific physics calculations perfomed each cycle to determine these key physics parameters acceptable. In addition, the detemination of whether each parameter should be greater than or less than its cor- -

responding value used in the reference analysis in order to be considered bounded, is acceptable.

.- ~_.____ _

- ~

(b) Uncontrolled Rod Cluster. Control Assemblies (RCCA) Withdrawal at Power The uncontrolled RCCA withdrawal from power is also an. anticipated opera-3 tional occurrence and, therefore, must not violate the fuel integrity. ,

The reactivity insertion rate determines which protection system trip b will initiate termination of the accident. Therefore, both a fast rate, which results in a high power trip, and a slow rate, which results in either an overtemperature trip or a high pressurizer level trip, are analyzed and compared to the Prairie Island FSAR results. For the fast reactivity rate, the licensee's calculation indicates a more severe tem-perature and pressure response but, in general, shows the same trends -

as the FSAR results. For the slow reactivity rate, the licensee's results predict a slower power. r. amp and slower pressure and temperature increases although the maximum values differ by only 3% or less. The minimum DNBR calculated by the licensee, however, is about.17% higher and, hence, less conservative, thaS the FSAR value. A further evaluation of the acceptability of the licensee's method for predicting minimum DNBR for this transient is made in cur review of the COBRA-IIIC/MIT code discussed below (Note Page 23 of this SER).

Cycle , specific physics calculations are performed at the most limiting core conditions during the cycle for the following parameters: (1) Doppler coefficient, (2) moderator temperature coefficient, (3) maximum reactivity insertion rate, (4) scram reactivity, and (5) nuclear enthalpy rise hot channel factor (F3 y). Each of these parameters is adjusted to include the model reliabiTity factors and biases derived in NSPNAD-8101P Rev. I which we find acceptable (note Section 3A of this SER). We find the key physfcs parameters selected to be apprcpriate for the transient. We also find the specific physics calculations performed for each cycle to determine these key physics parameters acceptable. In addition, the determination of whether each parameter should be greater than or less than its cor-responding value used in the licensee's analysis in order to be considered bounded, is acceptable.

(c) Control Rod Misalignment In this anticipated operational occurrence, one or more RCCAs are assumed to be di;placed from the normally allowed position. Full power operation under these conditions could lead to a reduction in DNBR. The licensee has analyzed this occurrence using input consistent with the Prairie

/

~

Island FSAR and predicts a minimum DNBR of 1.47, about 6.5*. higher than the FSAR result. Based on these results, we find the licensee's methods acceptable for calculating RCCA misalignment events when a cycle specific reanalysis is required. ,

t The important cycle specific physics. parameter is the nuclear enthalpy rise hot channel factor (Fag) and the maximum value is conservatively a calculated with Bank D fully inserted and one RCCA of Bank D fully Q withdrawn.

We have reviewed the accident definition, the analytical methods and -

assumptions used, the important physics parameters selected, the specific physics calculations performed each cycle, and the bounding values used in the safety analysis,,and find them to be acceptable.

~

(d) Control Rod Drop .., .

For this anticipated operational occurrence a single RCCA is assumed to drop into a fully inserted position in the core. The drop may or may not result in a reactor trip. If the reactor is brought to full power'with the fully inserted RCCA, a reduction in core thermal margins

• could result because.of an increated hot channel peaking factor. The licensee's analysis shows a minimum DNBR of 1.934 as compared to the FSAR calculated value of 1.9. Although the licensee's results agree well with the FSAR analysis, they do not consider the concerns involving several aspects of the RCCA drop event which presently require operating restrictions on many Westinghouse-designed, reactors including the P.rairie Island units. This operating restriction, based on analyses of control rod drop events, is addressed in our letter dated November 28,1979 to all affected licensees (Reference 1).

The most important cycle specific physics parameter is the resulting F w for the ONB evaluation, which is adjusted to include the model reTiability factor derived in NSPNAD-8101P Rev. I which we find accep-table (note Section 3A of this SER). .

He have reviewed the accident definition, the analytical methods and assumptions used, the important physics parameter selected, the specific ~

physics calculations performed each cycle, and the bounding values used in the safety analysis, and find them acceptable with the above mentioned operating restrictions in effect.

j

I (e) Control Rod Ejection The RCCA ejection accident is analyzed by the licensee and compared to the results reported by Westinghouse (Reference 2) for.zero power (80L 9 and EOL) and full power (BOL and E0L) initial conditions. A point kinetics a model is used with reactivity feedbacks corrected by weighting factors to account for the spatial dimensions not included in the model. The thermal-

' hydraulic model includes a multi-nodal radial model of fuel, gap, and clad conduction, and a multi-nodal axial model of the coolant channel.

The comparison shows that the licensee's results agree well for the zero power cases and are more conservative for the full power cases. We, therefore, find the licensee's method for analyzing this event to be -

acceptable.

(f) Fuel Handling Accident The licensee has addrefsed the fuel handling accident related to sudden releases of gaseous fission products held in the voids between the fuel pellets and the cladding of one fuel assembly. If a fuel handling ac;.ident were to occur, the release would be in the containment or auxiliary building. -

Such a release would activate the special ventilation system containing .

an absolute and a charcoal filtration system. The licensee stated tha*:,

when calculating offsite exposure fran a potential accident, the discharge to atmosphere is assumed to be at ground level from the auxiliary building to maximize the offsite doses. We find this acceptable. Other~ aspects ,

of the methods used by the licensee for analyzing a potentional fuel handling ac.cident were reviewed by the staff.

RadioTogical Consequences of Accidents For Which Fuel Failure is Predicted Based on this review, the licensee has not furnished sufficient information related to extended fuel burnup and increased specific power of the fuel assemblies. Increases in burnup and fuel specific power beyond 'the tradi-tional ranges covered in the Regulatory Guides and Standard Review Plan could affect the radiological consequences of some or all of the accidents

. for which fuel failures are predicted by changes in fuel failure rates, changes in total inventory and mix of radioisotopes in the fuel, the frac-tion of isotopes accumulated in the fuel-clad gaps, iodine spiking be-havior, and the effect of fuel rod gas pressure on decontamination factors assumed for fuel handling accidents. The consequences of these effects due to burnup and fuel specific power must be demonstrated to be within the Standard Review Pian guidelines. The licensee stated that in order O

6 l

L

• l' to demonstrate that the offsite dose guidelines are met, the methods in NSPNAO-8102P Rev.1 assume that the calculations used for the off-site doses in the FSAR are valid. These methods were reviewed and found accep-table by the staff prior to granting the operating license for those con-ditions of burnup and specific power expected. In addition, the licensee by letter dated December 13, 1982, has committed to evaluate the releases i for burnups up to and beyond 37,000 MWD /MTU batch average exposure. The 4 I licensee at this time will also evaluate any increases in fuel assembly specific power levels from those previously analyzed. The results of these evaluations will be submitted for NRC staff review prior to June 1,1983 and the licensee will include batch average burnup and specific power as _

bounding parameters in their methodology documents for every accident where fuel failure might be predicted to occur. The licensee does not plan to use the methodo-logy in NSPNAD-8102P Rev. 1 before the Unit 2 Cycle 8 reload which is scheduled to begin in August 1983. Based on the licensee's commitment, we find tha), the methods for evaluating radiological conseguences of accidents for which fuel failure can occur in NSPNAD-8102P Rev.1 are adequately addressed and are acceptable.

(g) Nuclear Steam Supply System Transients and Malfunctions

~

The licensee, in conjunction with Nuclear Associates International (NAI),

developed models for performing various safety analyses involving the plant transients and malfunctions for the Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2. These analyses include representative calculations for potential accidents of importance for a typical reload evaluation.

- The licensee's report (NSPNAD-8102P Rev.1) addresses the methodology for analyzing plant transient behavior based on these models for a variety of reactor and balance of plant accidents. We have performed a detailed review of these results. The reactor and balance of plant accidents that appear in the licensee's report are as follows:

Uncontrolled Boron Dilution Startup of an Inactive Loop

- Feedwater System Malfunction

~

Excessive Lead Increase Loss of External Load Loss of Normal Feedwater Loss of Reactor Coolant Flow - Pump Trip l

~ .

- Loss of Reactor Coolant Flow - Locked Rotor t - Main Steam Line Break -

- Loss of Coolant .

1 In performing the transient analyses, the licensee utilized the DYNODE-P s code under a license agreement with NAI, the developer of the DYN0DE-P/3 code. Results calculated by the licensee using the DYNODE-P/3 code were compared with previous Westinghouse evaluations as documented in the Prairie Island FSAR. The licensee provided more than 100 graphical comparisons to illustrate the validity of the DYNODE-P/3 code. -

~

Although the licensee did.n.ot include the analysis of the steam generator tube rupture event as part of the submittal in NSPNAD-8102P Rev.1, a separate analysis was performed ta gain confidence in the use of DYNODE-P/3 code.

The report titled "Usidg RETRAN-02 and DYN0DE-P to Analyze Steam Senerator Tube Breaks", NSAC-47, issued in May 1982, was used as a basis for qualifying the DYN0DE-P/3 code for analyzing a steam generator tube rupture event for the Prairie Island Plant. The staff considers the NSAC-47 report to be a significant milestone in establishing confidence in a methodology capable of performing post-accident analysis of an actual steam generator tube rupture event (a severe transient event for a PWR). The report shows that

, good agreement can be achieved between the analytical results and actual plant data obtained during the transient that occurred at Prairie Island Unit 1 on October 2, 1979. In addition, RETRAN-02 was also used in our evaluation of the DYN0DE-P/3 code. The RETRAN code is a derivatica of th'e RELAP code that the staff has found to be an adequate tool for audit calcuiations when properly used to model selected reactor plant transient behavior. RELAP has been used in both " conservative" and "best estimate" modes and the staff has confidence that RETRAN, because of its derivation, is also an adequate too,1 when properly used. ,

Based on the above, the staff reviewed the results from the NSAC-47 report and the results from RETRAN-02 and DYN0DE-P/3 in assessing the steam generator tube rupture event. The objective of the staff was (1) to establish an analytical link between the results of RETRAN-02 and DYN0DE-P/3 codes by means of the NSAC-47 report and (2) to use the NSAC-47 results to provide a link between both codes and the data obtained during the Prairie Island steam generator tube rupture event. The comparison of D

~

l . .

the DYN00E-P/3 code results with RETRAN-02 results and with plant data

• for the Prairie Island steam generator tube rupture event is shown in Figures 1 thru 14. This comparison of results and data for key operating parameters illustrates the qualification process used for DYN00E-P/3. In some cases the DYN00E-P/3 results are more conservative than the results from RETRAN-02. . f This was as expected since RETRAN-02 was developed to less conservative conditions (i.e., "best estimate") as compared to DYNODE-P/3. ~Based on the comparisons, we conclude that the DYN0DE-P/3 results reasonably agree ,

with the results given by RETRAN-02 and with the actual plant data obtained during the steam generator tube rupture event.

Folieving the review of the capability of the DYNODE-P/3 code to analyze -

the steam generator tube break event, a comparison audit of the DYN0DE-P/3 code was mace for eight. selected malfunctions appearing in the licensee's report NSDNAD-8102P. This comparison audit compared the DYN00E-P/3 results generated by the licensee with the RETRAN-02 results generated by the NRC st

• f f. The staff used 'the same initial system input conditions for RETRAN-02 as thusa used by the licensee for the DYNODE-P/3 code. The principal input parameters were reactor power level, coolant inlet enthalpy, core exposure (beginning or end of life) and trip set points (i.e., flow, temperature andpressure). ,

=

The eight accidents chosen for comparative technical audits were considered to be accidents which would encompass the range of technical severity of the types of calculations for which the licensee proposed to apply the DYNODE-P/3 code. The evaluation of DYN0DE-P/3 for eight accidents that were audited by the staff is illustrated by the results shown in Figures 15 thru 50. In these examples the figures also show the FSAR results.

Our evaluations of the transients (eight acc'idents) are as follows:

(1) Loss of Feedwater Heaters, Beginning of Life, Without Reactor Controllers Operating - Figures 15 thru 18 -

In this transient, the loss of feedwater heating causes an associated drop in core inlet temperature. In response to the drop in cora average temperature, the doppler reactivity feedback attempts to drive the core -

power up until the core power matches the load demand and a new steady state level is reached. The RETRAN results show a higher increase "Ine plant data were obtained from the Prairie Island plant computer which accumulated key data approximately 200 seconds preceding and following the reactor trip.

s in power than the FSAR or DYN00E results for approximately the same decrease in core average temperature and the same doppler reactivity coef ficient. This is due to the RETRAN calculations use a "best estimate" fuel cladding gap coefficient which is much larger than the conservatively 3

small value used in the FSAR calculations. This larger coefficient causes ,

a much faster response of the fuel temperature to the drop in the coolant temperature and hence a faster increase in power. Note the scaling on

' the plotted comparisons is somewhat deceiving and that the associated difference in peak power between the RETRAN and DYN00E calculations is only about 0.5%. The rest of the system parameters show reasonably good comparisons to the DYN0DE results, given the difference in the power response. In this transient, the core power increase effect on DNBR is outweighed by the effect of the decrease in coolant temperature causing an increase in DNBR. Therefore, the minimum DNBR (MDNBR) occurs at the start of the transient and even though the RETRAN results are slightly more conservative than the DYNODE calculations, there will be no significant effect on MDNBR. On thNs basis, we conclude that DYNODE-P/3 is an ac-ceptable tool for predicting the transient behavior of key parameters during this accident. ,

(2) Turbine Trio, Beginning of Life, with Reactor Controllers Ooerating

- Figures 19 thru 23 The DYN0DE results for-this transient show good comparison to the FSAR results until the pressurizer relief valve setpoint is reached. The FSAR assumes an instantaneous opening of the relief valves which hold the pressure below the high pressure trip setpoint for approximately 10 seconds more. In the DYN0DE analysis the reliefs are more realisti-cally'modeled with a 0.3 second opening delay and a two second stroke time. Note that eveq though the model is more realistic than the instantaneous opening used in the FSAR, the setpoints used in DYN0DE, i.e., delay and stroke times, are conservative estimates of the actual times. This causes the' DYNODE calculations to show a much faster pres-surization rate after the relief valve setpoints are reached. The reactor  ;

trip setpoint is reached in approximately 2.5 seconds. The RETRAN results show a much slower initial pressurization rate which we attribute to the additional heat sink of the metal heat conductors. Therefore, the relief l valve opening setpoint is reached approximately six seconds later in the

- transient. However, since the relief valves are nodelled in RETRAN in the same way as they are in DYNODE, the time from when the relief stitpoint l

l l

l

1 ,

.is reached to when the trip setpoint is reached, 3-4 seconds, is approx-imately the same as in the DYN00E analysis, as opposed to the ten seconds shown in the FSAR. The system behavior in general shows comparable trends between the RETRAN and DYN00E analysis. On this basis, we conclude that DYN00E-P/3 is an acceptable methodology for predicting the transient , r behavior of the key parameters during this accident.

, (3) Turbine Trio, End of Life, With Reactor Controllers Ooerating - Figures 24 thru 28 This transient show similar behavior as in the beginning of core life with control case. In this case,, however, both the RETRAN and FSAR -

analyses predict that the reactor does not trip whereas the DYN00E analysis shows a reactor trip on. hi.g,h pressure. In the FSAR, the pressure is held down due to the faster response of the relief valves. In the RETRAN r

. case, the pressure is held below the trip setpoint due to the smaller i initial pressurization rate. The results in general show comparable trends. We have concluded that DYN0DE-P/3 adequately predicts the transient behavior for this accident. .

(4) Turbine Trio, Beginning of Life, Without Reactor Controllers Ooerating

- Figures 29 thru 33 In this case, the OYN00E results show good comparison to the FSAR since the relief valves are ignored in the analysis. The RETRAN results again show a smaller initial pressurization rate which causes the high pressure trip to occur slightly later. However, the system behavior shows similar trends. Also the RETRAN calculations do not predict an opening of the pressurizer safety valves. We have concluded that DYN0DE-P/3 adequately predicts the transient behavior for this accident.

(5) Turbine Trio, End of Lifs. Without Reactor Controllers Ooerating

- Figures 34 thru 33 This transient shows behavior similar to beginning of life without reactor controllers operating (Item 4 above) and the same concluding remark applies .

for this transient.

I 1

' ~ i

, j i

6) Pumo Trio,1 of 2 Reactor Coolant Pumos - Figures 39 thru 42 and

7) Pumo Trio. 2 of 2 Reactor Coolant Pumos - Figures 43 tnru 45 The following comments apply.to both the 1 and 2 pump trip cases. The

• 1 RETRAN results show very good comparison to the DYNODE. calculations. The -

only differences are that in both cases the core power drops off much faster after the reactor trip signal in the RETRAN case. This is due to the fact that the RETRAN calculations use a best estimate trip curve with a greater total worth than the DYN00E or FSAR calculations. The OYN00E results in this case are therefore more conservative than the RETRAN results, as would be expected. We have concluded that DYN00E-P/3 adequately predicts the transient behavior for the key parameters for -

this accident.

~

' (8) Main Steamline Break"- Figures 46 thru 50 The RETRAN results show an associated cooldown rate of less than palf of that predicted by DYN0DE for the same break' size. This slower cool-down is due to the fact that the RETRAN calculations include the effects of metal heat conductors. This large source of metal stored heat causes a much slower cooldown of the primary system. The pressurizer does not empty and the core just barely reaches criticality. The core goes critical much later in the transient and is quickly turned around by the high boron concentration of safety injection water causing a much lower peak heat flux and hence a larger margin of DNB. The DYN0DE calculations are clearly much more conservative than the RETRAN results. On this basis, we have concluded that DYN00E-P/3 adequately predicts the tr'ansient behavior pertainng to the reactor and primary coolant system for the key parameters for this accident.

Based on our review, the qualification of the DYN00E-P/3 code is summarized as follows: .

1. The license to operate the Prairie Island Nuclear Generating Plant, Unit Nos.1 and 2 was predicated in part upon our acceptance of the t

' . transient results prepared by Westinghouse which appear in the Prairie Island Final Safety Analysis Report (FSAR). This acceptance of the transient results still applies since the plant has not been

- modified nor has the mode of operation changed in a manner that would affect the transient results.

- - - - --e-  %-- . __. . . . , . . , , , , , - _ _ _ _ . - _ . - _ _ - _ - . ,_,,, , -- , , , . . _ . -. ,_ _-.-., -

w, .-.4.- ,.._,_,_--.yy

r- -

2. The licensee's report NSPNAD-8102P Rev.1 shows many comparisons of the transient results of the FSAR versus DYN00E-P/3 results. The DYN00E-P/3 code results show trends similar to those appearing in the FSAR for parameters important.to reactor plant safety. . Based on our review of these results, we have concluded that the DYN00E-P/3 code gives '

r results that are either equivalent to or greater than the level of conservatism used in the FSAR for parameters important to reactor '

. plant safety. (

3. The staff spent many years developing and improving the RELAP codes and has used various versions of RELAP for audit calculations. As a result of this experience, we have gained confidence in the capa- -

bility of the RELAP codes, when properly used, to model selected reactor plant transient behavior.' Depending on the input assumptions and sub-routines used RELAP. has been used in both " conservative" and "best estimate" modes. This level of confidence in RELAP also extends to RETRAN-02, which is' derived from RELAP, and which has been designed for "best estimate" use by utilities. Our use of the RETRAN-02 code, t as a means of auditing the results from the DYN00E-P/3 code shown in NSPNAD-8102P Rev. I was based on the staff's experience with.and confidence in the RELAP and.RETRAN codes. Our audit has shown that in all but one case, the DYN00E-P/3. code gives results that are

~

equivalent to or greater than the level of conservatism given by RETRAN-02. In the case that exhibited a reduction in the level of conservatism, we determined that it has no safety significance to the reactor plant.

Based on the above evaluation we conclude that the DYN00E-P/3 code is an acceptable means for analyzing transients caused by a single steam generator tube rupture in a Westinghouse 2-loop reactor such as those existing at the Prairie Island facility. In addition, the DYN0DE-P/3 code is acceptable for analyzing of transients caused by accidents de-scribed above and in the licensee's report, NSPNAD-8102P Rev.1.~ This acceptance is contingent upon the completion of an audit of the licensee's administrative control procedures for the computer code quality assurance.

The responsibility for this quality assurance audit is assigned to the .

NRC Region IV office. The Region IV audit will be initiated during January 1983 and is scheduled for completion by March 1983. .

'_----- J

(h) Reactor Core Thermal-Hydraulic Hot Channel Analysis (COBRA-IIIC/MIT)

As shown in the NSPNAD-8102P Rev.1 report, the licensee uses the COBRA-IIIC/MIT code to analyze the transient response of the _ hot channel in 9 the core. The COBRA-IIIC/MIT code simulates the thermal-hydraulic response ,

in the coolant channel of fuel rods within the core. The thernal-hydraulic hot channel analysis is performed for each transient to assure that the

' mininum DNBR is not reached. While we have not completed our review of the licensee's submittal dealing with the application of COBRA-IIIC/MIT for the analysis of the transient responses of the hot channels in the core, our review has progressed sufficiently to indicate that the results from COBRA-IIIC/MIT code gives a reasonable approximation, if not a con- _

servative analysis of the limiting fuel assembly in the core. Consequently, we conclude that the licensae may use COBRA-IIIC/MIT for evaluating the thermal-hydraulic characteristics for the Prairie Island Unit 2 Cycle 8

- reload analysis' of the hot channel in order to assure that the minimum DNBR limit is not viola'ted. ,

(i) Reactor Fuel Hot Soot Analyses for Certain Accidents" - T00DEE2 Code

+ The T000EE2 code developed by the NRC staff is a two dimensional, time dependent fuel element thermal analysis program. The primary object of this orogram is to calculate fuel element thermal response during post-LOCA refill and reflood in a PWR such as the facilities at Prairie Island. However, this code may also be used for certain non-LOCA tran-sients with appropriate assumptions and input data. The code provides, as output, the fuel and cladding temperatures and the energy deposited into the fuel during transients.

The licensee's methodology in NSPNAD-8102P Rev.1 applies the T000EE2 code for analyzing non-LOCA transients including the postulated locked

'Totor and rod ejection accidents. The publicly-available version of the T00DEE2 code (NUREG-75/057) is used by the licensee to compute the transient temperature response of the hot channel for a number of tran-sients and accidents. The code is used to demorstrate that coolability

- criteria continues to be satisfied when the hot channel condition analyzed by the COBRA-IIIC/MIT code yields an unfavorable DNBR. According to SRP 4.2, the fuel coolability criteria for non-LOCA accidents must in-

• clude cladding embrittlement and violent expulsion of fuel. For cladding embrittlement, the Prairie Island FSAR uses a cladding maximum temperature

~

=

t:

(

f limit of 2700*F for some transients, e.g., locked rotor. For violent expulsion of fuel, a radially averaged enthalpy limit of 280 cal /g de-posited into fuel is observed.

Since the T000EE-2 code was originally not intended to be used for non- ,

r LOCA analyses, the licensee demonstrated that the results from T000EE2 for limiting transients are indeed compatible to those previously analyzed 4

. in the FSAR.

For the two accidents analyzed in this manner, the results shown in NSPNAD-8102P Rev.1 (Figures 3.12-6 for locked rotor; 3.15-2 and 3.15-4 for rod ejection) indicate that T000EE2 calculations are consistent with the ~

original calculations supporting the FSAR analyses. The initial and boundary conditions were,.taken from the FSAR. The transient co.41tions for power, core inlet flow, core inlet temperature, and presst :-dependent coolant. saturation temperature were obtained from the DYN00E-P/3 code, =

which we find acceptable (note 3.B.2.(g)). For the cladding surface heat transfer coefficient, the licensee calculated the lowest value (with respect to time and axial location) of the Sandberg correlation used in the FSAR through an iterative process. When the lowest value of heat transfer coefficient was determined, this value was input to the T000EE2 code as a constant for all time and axial locations to calculate the cladding surface temperature. This procedure was described in the '

licensee's letter dated January 4,1983 (O. M. Musolf to NRR). Because of the assumption of cosine axial power shape, the highest cladding temperature always occurs at the core midplane of the hot channel, which reaches DNB first.

We have reviewed the NSP methodology of using T000EE2 for non-LOCA transients and accidents identified in NSPNAD-8102P Rev.1. Although only two accidents (locked rotor and rod ejection) were analyzed for benchmarking, we believe that T000EE2 can be used for other similar events with conservative assumptions and appropriate initial as well as boundary conditions as described in the submittals. We, therefore, find the licensee's application of T000EE2 acceptable for analyzing transient temperature response and energy deposited into the fuel at a hot channel

• for non-LOCA accidents such as those described in NSPNAD-8102P Rev.1 in order to satisfy fuel coolability limits. .

e e

e .

(j) Containment Temoerature-Pressure Resoonse During a Steam Line Break (CONTEMPT-LT)

The licensee calculated the containment response to postulated main steam line break (MSLB) accidents using the computer code CONTEMPT-LT/026 which 1 requires the mass and energy release input data obtained from the DYN00E-P/3 ,

computer code. The heat transfer correlations used in CONTEMPT-LT/026 by the licensee were taken from the Westinghouse Topical Report WCAP-8327, Contain-ment Pressure Analysis Code (C0CO) which we find acceptable for benchmark comparison. However, the correlations are more suitable for minimum contain-ment pressure analyses following loss of coolant accidents rather than for the peak containment pressure analysis following MSLB accidents. Current guidelines in SRP Section 6.2.1.1.A, which refers to NUREG-0588 (Appendix B) _

recommends that the Uchida heat transfer correlation should be used for MSLB accidents while in ,th,e, condensing mode, and that a natural convection heat transfer coefficien.t should be used in the non-condensing mode.

The CONTEMPT-LT/026 code calculates the containment pressure, temperature, thermodynamic state, and heat transfer as a function of time, following postulated pipe break accidents inside containment. As stated in Standard Review Plan (SRP) Section 6.2.1, the CONTEMPT-LT code has been found acceptable by the staff for use in containment' analysis. However, our

• review of the mass and energy release data from the DYNODE-P/3 code .

that serve as input data to the CONTEMPT-LT code is not complete at this time. Consequently, we cannot make a finding that the DYN00E-P/3 code is qualified for use in conjunction with the CONTEMPT-LT code to perform the containment pressure transient analyses during the main steam line break accident described in the licensee's report, NSPNAD-8102P Rev. 1 Section 3.14. Therefore, staff acceptance of the licensee's method for analyzing the containment pressure transient resulting from the main steam line break accident is deferred until the staff has had an opportunity to complete its review.

4. Conclusion -

Based on the above evaluation, we find the reactor model description,

. qualification, and quantification of reliability factors addressed in the licensee's report, NSPNAD-8101P Rev. I to be acceptable. Therefore, the methods described in NSPNAD-8101P Rev.1 can be used by the licensee in performing operations and fuel reload analyses for the Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2.

Based on our evaluation, we find that the general physfes calculational methods in Section 2 of the licensee's report, NSPHAD-8102P Rev. 1 are acceptable and that the licensee may use these calculational methods for performing operations and fuel reload analyses for 'ae Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2.

,~

Based on our evaluation, we also find analyses of postulated accidents described in Section 3 of the licensee's report, NSPNAD-8102P Rev. I acceptable contingent upon.the following limitations.

r

1. Acceptance of the computer codes referenced,in the' licensee's report, .

NSPNAD-8102P Rev.1 is limited to analyzing only the type of accidents listed and described in the report. Therefore, any use of these codes ,

for licensing analysis beyond the type of accident discussed in the l licensee's report, NSPNAD-8102P Rev. I must be accompanied by the  !

t applicable qualification information.

2. A.cceptance of the control rod drop accident does not cover the _

analysis of the control rod drop events addressed in our letter dated November 28,;1979.

- 3. Acceptance of the methods of analysis for radiological consequences '

i of any accident with predicted fuel failures is predicated.up'on the licensee submitting information on the release rates and effects on i the mix of nuclides for burnups up to and beyond 37,000 MWD /MTU batch

! average exposure which is acceptable to the staff. This information -

will also include any increases in fuel assembly specific power from those previously analyzed.

4 Acceptance of the DYN00E-P/3 code is contingent upon achieving a i

a satisfactory resolution of findings resulting from audit of the licensee's administrative control procedures for the computer code i quality assurance. However, an interim use of tne DYNODE-P/3 code

! for the Prairie Island Unit 2 Cycle 8 reload analysis is acceptable since our. audit findings revealed only minor discrepancies in the administrative control procedures for the computer code quality

- assurance. These discrepancies have no adverse effects on the safety level of plant operations.

l ,

S. The mass and energy release data from the DYN00E-P/3 code that l serves as input data to the CONTEMPT-LT code is not qualified at this time for use in performing the containment pressure transient -

L analysis for the main steam line break accident.

Date: FEB 17 1553 J

l -

i

?

4 ,

m 9

27 -

O

5. Tables 1 ,

b 99 O S 6

4 9

9 e

e a

TABLE 1 l

RELIABILITY FAC: TORS FOR PRAIRIE ISLAND i-RELIABILITY PARAMETER FACTOR BIAS USED TO COMPARE TO SAFETY VALUE -r N

Fq RFpg = 0.042 Off N FqN , (pq (Model)*RFpq*Blas F q )*(1*T) <

F AH RF F4H = .0 H .O Fgg = (F4H( el)*RFFAH

+Blas F4H)*(1-T) -

Rod Worth 0 RFRod = 0.10 ,

, yrod=C4Pg(Model) - Bras) * (1:R7%)

Tamperature RFga 2,9 .03 My

=

a "g(bel)

• Blang a RFy Coe#icient (pcm/*F) .

Doppler- 0 #

CoeMicient RFD = 0.10 D*WC D o!)

• Bla:D) * (1 : RFD)

Doppler 0 RFDD = 0.20 Defect DD = (D0 (Mel)

• Bla:DD) * (1 : RFDD)

Coren 0

• RFB = 0.10 g a F g(Wal)"+ Blas'g) * (1 : RF3) .

Worth Delayed RF 0 p = 0.04 0,g = (6,g(Model)

• Blasg) * (1 : RFd )

Neutron Parameters RF J , s 0.04 0 2* = (R*(Model)

• Blasg ) * (1 : RFj ,) -

4 e

5 g

. 1.

e ,

• 8 m .m . iii ..

O e

9

6. Figures i

4 s

een 4

98 9 6 4

,d 8 4

e e

6 e

e 9

8 6

o e4 . *

,i .

30 -

hbE'j' vtr 2700 0 Ereak initiation O Plant data 1 10% Icad reduction R E N42 2 Charging pump 2 on .

r.

3 Charging pump 3 on .- ---- DYNCDE P 2500 I I e oSf*a*e*ty injection en '

6 Reactor ccctant pump 1 trip 7 Reactor cociant pump 2 trip e

3 2300 m . . . . ,

3 o -

e \ -

.c 5 .

$ N 6. *

. E .

C- 2100 -

t 7

• 2 Sj#

1900 -

)4 /

e** Coo

/

/

!  !  ! I 1700 O 100 200 300 400 '500 Transient Time (s) ,

8 Figure 1. Reacter ceclant system pressure.

\

- .-__-_-__-____________-________u__

m%; g*./. 4 =!-

s ,.

d 1.1 0 Plant data 3 1.0 _ M RETRAN-02 ' ,

Leadj i ---

.DY, NODE-P 6 0.e.

reduction p_ m -- -

0.8 -

Reactor ,

trio *

. 0.7 - '

R .

o

$ 0.6 -

a

> .. ~ ~

sw

]0.5 -

g ,

0.4 -

, s .-

0.3 -

0.2 -

0.1 -

l \Y ' '

l o O

l 100 l

200

%?300 m -' e.OO .

SCO Transient Time (s)

Figure 2. Norma!i:ed system p0wer. ., ,

{

l 1 .

l

. e

.: ; *i," :s Cl. g'...,*.-

gu t 4

. r 4 ( .

Lead O Plant data a

~'N reduction R_TTRAN-02

• O <

o ----

DYNOCE-P 3 -

000 .

, ~ '

\ -

G v--- '

g . . . .

. Reacter l s j2 -

trip

- t. l 6 , , , .

.o. l I -

h.

.:_:-g - i .

1 g

I

n l l 1

I I M Y 0 '- -

0 100 700 300 400 500 '

Transient Time (s) .

..~ '.

Figure 3. Leco A feedwater flow.

6

- e S

r DRM* r A

) ,

4 g C Plant cata R=iMAN 02

~ ---- DYNODE-P 3 - Lead ,

reduction ._

• • • 54 Reactor EE Mp gg . . . . .

o- .

~ '- .'

3@ 2 . .-

D2 e e.

.I

• • .t*

5 g

u. .

- t .

, 1

\ '

- I  ! '-'-

O o 100 200 300 400 500 Transient Time (s) .

Figure 4 Loep B steam generator feedwater'f$w. .'

l e

i.* .

h

l .. __ ,

f f

.a

• f 610 ,

O Plant data

. RETRAN42

---

DYNODE-P '

C sgo A b

• t -

a _ _ .

o ---gD ' *

5.

• Loadj.... 00CO d

~

5

. reduc'Jon .

Reactor _/

/ . -

(

g e.

trip

& syn _ .

E

a. -

3_ . "o K~ s o.

g o 550 -

o a

\ coCgCOCoo g

I I 1 I 530 ~

O 100 200 300 400 500 .

Transient Time (s) ,

Figure 5. Loep A hot leg temperature. ,

a e

e e

h==sameeimumi im- m- i e . m O . ....-..--ie

I

\

s

.I$' a

? s fi:%n p-. ,

610 . ,.

O Plant data Load -- P.ETRAN 02 reduction

---

DYNCOE P sgo _ a ----,

C occe ,

g . . . . . Reactor trip o .

- B .

3 570 - -*. -

g . . ..

8 w *

. . . O

. .' ' . O 550 -

- - .' l Oc3DC::::DCt30

\ .

Y l l l l 333 O 100 200 300 400 500 Transient Time (s) .

Figure 6. Hot leg temperature (Loop B).  ;

S ,

e e

e i ,- -

9 6

~

, 5

• kl f 550 e O Plant data ,

METRAN42 OYNODE-P l 550 -

. .e _

s C

y . . . .

o

• 5 7 540 -

6 gv QOOON o

g. 00  % -

, ~ '

• o- ' " O O  %*==, .

~

/0oho O, .

530 -

O e'

f-- kload reduction Reactor trip 520  !  !  !  !

0 100 200 3C0 400 500 Transient Time (s) ,

Figure ,7. teep A cold leg tem;:erature. -

+

q . g .

4 e

1 .

-- u .~--.r_-, . . _ , . - - ~ . , , ._,~-,w.- - - - . __-.,,_______.__.r.,__ . _ , . _ , . . , , , , . . _ . _ . ,

~

l@Ar**

It; g s v. i 550 '

O P: ant cata RETRAN 02

~ ~ - ~ CYNCDE P 550 -

+

~

D . . . . .

o

= -

5_ 540 g .

e  %

• U

_O co o _ , ,,, 3

.'--*~ /

530 -

/ Mg o O 0 06 Q

4 O

OO o

- wJ g ( Load Reactor reduction trip 520  !  !

. 0 100 200 300 400 500 Transient Time (s)

Figure 3. Leco B ccid leg temperature- .

l I

~

I

[ .

s '

~l - 38 N. -

.M i t

Y.!}n.r!'%

~

Ih [

~

s .

s r

570

, e' o Plant cara t ,e g

• recue:ica j RsisAN42

---

DYNCOE-P 560 -

y -\

) E

• • eo' . -

!g .

ne.ctor wp E 550 -

9_ .

o N g

I M A i -

- . is. g ss 5

gi ' ' '

Go 540 --

Y CCC:XDocco s

N l l l_ l 530 _

O 100 200 300 s. 400 ,

.- 500 t

Transient Time (s) -

Figure 9. Leco A average temg:erature. . .

e 6

1 y 6 .

s

^ O ,

t,..

. ,S f?N

'< =u ;!'f s

570 ,

s o Plant cata P.ETRAN42

---

DYNCDE-P

~

. 560 7_, y

' ~ 0] ,

Lead "

reduction .

{ Reactor  !

o -

trio

• 3 - .

a 550

- L -

(

i t- .

I

." I OC3 '

540 -

Mc::coccco _

' ~-

w

$30 l \ 'l l O 100 200 300 400 500 Transient Time (s) ,

Figure 10. Leep 3 average tem::erature. i O

• 5 4

6 s.

G num i ei - -

e

.. b...

f.

s 80 ,

i .

o Plant data 70 - Lead

• racuction R iMAN42

-- m, ----

DYNCDE P 60 -

e W

• _ _ _

t-- 'O v y q _.

I5 50 -

g . . . . .

. Reactor

• b trip

.o 40 2c: . ..

g 30 '

h 20 '

l l '

9 10 -

\ .m O

O I

100 200 i 5 <I 300

-sh( t00 s.

500 Transient Time (s) .

.~

. Figure 11. Loop A delta T. .

• e 0

F e ,

0 e

l e

., g e

$ g

" ' * - - - - - -++-4 ,m- .= , _ . g _g, , , , , , _$ __

~

r*h*.b{ f.

~;;-

50 ,

' O Plant cata recuction FETRAN-32 70l l- [ Lead ~ ~- - DYN005 P 60 - ,,

c t

O 000 *

~

6 so -

Reacter 5 . . . . .

3a 40 -

s.

5 . . .

p, i 30 -

. Er 5

- . s- e- W s'~~~

10

~

Oh I I '- *v I O

, O 100 200 300 40'J 500 Transient Time (s)

Figure 12. Loop B delta T. .

b .

6

. Dhr On.nrT 1000 '

0 Plant cata ,

,- RETRAN 02

/*

---

DYNODE P ~

900 - s l 's

$ SCO .

a s. '

E o n s . .

w

/

. . coo /

. .. 700 *-W ~

Lead Tur".,i..e

. and .

reduction reactor trip I I I I 500 O 100 200 300 t00 500 Transient Time (s)

Figure 13. Steam generator pressure (Leco A). -

t M

e I

i l s .- .' .

i l

[

I i

~ fl kIlkC,t 1000 O Plant data ,

s' RETRAN 02 DYNODE P 900 -

w g - _ __

Turbine and N 3s reacter trip - -

a - -

300 __

5 -

~

8 o i' E O -

00

/-_ w o

^/ . .

A w ( Lead 700 .

reduction soo 1 I I l 0 100 200 300 e.00 soo

, Transient Time (s)  :

~

I Figure 14 Steam generator pressure (Leon S). .

h 6 . .

9 -

- . 44 c *

. e a m

. [ . . . . . . . . * *

.O 8

~I d

. e .

g ( .

m no .e . . . .

<eg *;*

. .7""  :  :

.4  :  :  :  :

,C

. ............................................*"ar.............**.................................:..ono."""""

u. I ** * . . . .

-C ee e .

31 L.. - .

. . g . . . .. . . .

F;* Ig  :

• a .

.

• r t

,c

......g.........;...................g....................g...................

t  : .

...................a.....................

1 .

. , , 1.e.3 t

g

• g . . .

. . . t t . .

. . * . g . . . .

r

. I t .

. .c

.........;.........t.......o.:nn..."*,.un..g.........t.......i.........t.........i.........t.........i......." . =c

.m I g . . . .

f* f" f;5""  :.  :.

E  :. l .  :.

".g, a.%.. f

.. t .

• • * .a. f.r ,i E r*.'

. r. .

. f . r t

.O

.......~,t.........t...........:..........:...... ..g.........:.........g.........

.........g.........:.........t ......

-m

. r.

.v

. F.

. s .

. t= .

. r. .

T.

. r.

.. =c g . .

.............(........

. ........................ ....... ........................ c r .- .

. 3. .

. .T

. . . . s.. . . . .

/... . .

p.

. . 6, . . . . .

y. . .

. . r. . . . .C

' (A

. . . . . . . . . .t. . . . . . . . . .

.................... ........ .......... .................... . m

. . e *J u.p p.

l g

g .-*

g

. C a p

. 3 .

g i

=.=  :  :  :  :  : i  :  :  :  : .

e e,,} .

u O . . . . . . . . .. .

4...... . . . . . . . . . .... ........1.g .l......... . . . . . . . . . . . ................. . . ...................

. . . . . . . ~ .

- c ,-) Q

. L, .

e .

g . .

.C g .. . . . . .

. s . . .

C* . .

. , t . . .

- . . . . . . s . . .

l .

. .c

""\................................

..... .. .. ..t, . . a

. s.

.........t..........

.................... -m

. . .N

s.  %.. .

O.

u .

. s.% .

. g . . .

s

s. .

.c

.........:.........f......: .u.. < .......:..... .t.a...'s.:.........).........:.........}...... .:......... -o s

.N

. s. s

g. .

O. .

s..

O.

. .c

• IT .~......i........ .. ...{.a...... ; .........Ia a..~ -mp 6 .........{.........

.  :.........l.........t.....a a.4......

C .

. g,_ e . . . .

v =.

e e.

O .O . . . .

.= . . .

g .

P .

s g .

,O DaJ sm., y .......................................

........."..................... ..................... . % ..., 4..y........". -c Q .t g., .

g.

g

= v.

c .

OCC

. %.. . a. .

.  %. g

>e. .

g .

O .

C e m.

s..

c i

O ......... ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ....

. g3

l . . . . .

(, . . . . . . .

Q..*O.c , e  ;

. 9. -. :.

Em y v W d c3 N

. N. g c. c. v. . g C. v. N. o y .v , r3 o "o o e N N N N e o O o O C O O O

- o.

, - - - - ,O. - - -

puguoN tuscaed .

*. ;"% f..ism. 5, . .

I 6 3. f a g. e.

5- , t

1. e p. L...;.(. a.. .

C

• o e l a t . . . . . *

• 8

=l  ;  ;  ;  ; .  :. ,

.C.

. e e . . . . . .

s<

!e< * .r"u .

. . . . g. ..........

7) ..... ..i..........i..~ a .............i..........i..........i..........i..........!;.........,:.........,,,,

! e

, g g, g, ,

i s t: : C  :. I  :.  :.  :  :.  : i  : 1. <::

. . s . .

3 .

m . * . . . s.

4;  ;  :*  :  :  :. ,

= * . .

g

. t,. .

.e

. . . . . . . . . . . . . . . , . . ....... ...........,...................(.................... ........:........ 4.........;,......., ,, g

.W l  :  : .  :  :  :

. I .

.  ; =

g . . .

3 .. . . . . . .

• g .

I .

.........;......... ..... .. .........g.........;........

t < . . . .C 3.........;.........:..................p..................,

g

. g . . . . . igg 3 . . . .

. . l .

.g .

.......[...................g.......... . .. .. . . . g . . .

. . .C

.........{......... .........g.. ..... o n .o . .. . g oo. . . . . ggy

. e

, v.

g .

s

. s . .

. s. .

......... ......... ........ ............. . . . . .%..s. . ..s.........

.% . .......... ........ ...................a

.C

......... O s.

.v

. 6. . . .

. . . . s .

. . s. . . ,, C

......... . . . . . . . . . . . . . - . . . . ................................. ...S........ m ..................... .................... o *C3

. N. '- s  : . : . . C'3 l3 to

. .s ~

.n . . .

. . s . .

. %",44 s . .

Q g

. s . . . .

Q em . .  %.. .

i s.

g,,a s.

........4.g.........Y.a................ ... ....

.c .

Q O

.gc.3 W

. . . s j . .

g .

. . . . s . . .

.O.e g

g.

-  :  :  :  :  :  :  :  %  :  : .c a

. .........:.........i.~ ....~.:... ....i...

. . ...I......... 4...

....."A . ....i....s.....*. .

...... -g3

. . . s .N

. . .g O. O

. .i

, w a. C

= .

- . . . . # s

.u) )= ,.. . . . . . p.. s .

.-#...r#..:........g..) ............ '< .O

...................).................

a .... ...,......... l................. .

.... c

.N g .

. . s,s .

. ,,.w e . a .

. ,s .

3 . . . . . . . 3 IT , g

.......{.........

.........(.....au . .

a ..au.1. =a..

a. . a l . ,. ~ .a u .a . .a ( . .. a u . .

gs . ...........I.......... - g3

. 6 .,,s w

OO c . . ,

- w  :.  :  : ,  :.  :  :.  :  :

,e . . .

. C

@O6 ,

g

.. ............... ~u.-

. g aan ~~ a; a............au..a.......:.........:......~.:......... -c o.* a .

.t

. e.

. .c 3

OCO #

y .. . . . .

O

, /* . .

yC .:l: ......... ......... . , . ,<., . :. . . . . .. .r :a

.aa:. .. . .. a . : = .a. .u a .. a a a :a a o a . : ~ . a ~ ~ .n. a a . .:.........

c.3 g,

D g  :. #  :  :  :.  :.  :.  :.  :  : -

<. J 4 . . .

3 . .

g .

O c It . .

- e *I e e . . . . i a . .C

. e e .

e o o .. o e o e o e o o N " '" C i N. c."J v, - O, C, N. O. O. -

. . e

VlSd -

46 - I.:" *~ L. *, ., !"W*:

• * ;.a .

g *g .

M L* C 8 * .

W s ,L .

. C t == 3 . . . . . .

t <, . .

t a. . .

8 . . . . . . . .

<, H  : .

c

ws .:,:

. l

-C

. C 3 L, -, . . . .

. . . 8 .

. . . t. . .

m  :.  :  :  : e:  :  :.  : =

w .

.g

. . 3

< . . 8 . .

. . c

........... .....................'....n.....n.n...

. S. . .

n...g.......................

. n g

. . . 3 . .

. . . . . =

. . 3 . . .

. . . . . e

. 8. .

.  ! . t . .

.

• 8.

. . C

. t.........i......T......t...........i.........:.)...........s..................

...s...........

O

. . . 3 . . . .

s. .

1 .

.f

. t.. . .

3 . .

. l.

. O

...........l. ..... ..

....................{..-.....g..g...........{.....................4............

m 8 . .

. T. _

.. . . . t . . . .

e.

s.. .

J . . . .

. s.. .

- C t

......;...........;...........; . .. . . . . . y ... . . . . . . . . . . . .;............. . . . . . . . . . . . ;.....,.o.. . =c

- 8 . .

.. T

. g .

. . . I . .

. . 8 f .

. . t .

. . . t . . . .

t.

. . . . C C

1..............n.

e:  :*

: e* D

. C"3 r%

- . . . - .  :.s:  :.  ;

l. s :. .

C

• t .

. . c

. . . t .

r> w r . . .

. s

. .. =

C

.. ..  :  :  :  :  :  :  :  : - o

- nc (f) 4,a ........ .................. ............................ ..g.....

t J

. . . f =

/ . . . .

. 8 . . . .

o

= ...

l.  :

,8

..f....i...........:...........(...............

...... -c e

a H .

. t

. .N O. e t

. . g f

e ,..A . . . . .

. . g . .

O . . . g . .

C . e . . . . .

.C .

w .

M. ..........)..................s

... ..... ............)....... g....}............ .........)... .......,........... .c 3 . .

. s .

. . . . N c .

c m

o

s s: s

g. .

~ . . . < . . . . .

C .E.  :.  : . .

2  :.

.  :. .c Q o ........ .g. ........ . . . . . . .

s . . .. .. . . . t .gg.........(...........t...........<......... e b  :

.  : g

. v=

, u . . s . .

t w .c .

C. .

=

wc .c l . . . . . . . . . .. ..............

6 o . . .

. ,C-Cl) e. as .

s s.,

OewO . . . . . .

s . . . . .

o

/

. ..s..-

O  ;  ; '  : * *

.  :. a. .

g .

.c .............. ........ ........... ........... ........... ........ ...............:...........

=

...

• g. .

-c c . ... _ . . . . 4.

<J . .

s . .

. c3 .

. . . . .  % o -

C . .

. . .. . = = = =.- \ .

I -

O g I

. . g (Q 3  !

N '" O

. c. o. .e. .N.. cc. c. T. N.

?"  ?" ** T* O C Q C Usuue.:qad see Sea 1_

L

~ ... ,

6

\ - 47 - l.W e6.W 9 ..

e e

i  :.  :.  :  :. .

:.  :. . .  !  : .c

:  : m: >  :  :  :  :*  :  :* .

. e,.r< < e,;)....... ................

.....=.2.. .....u . ..................................,.......:........;,......,

,,i,,,, ., e 2, u.:.C  :

.e

.v . . . .

.i

........$........ . . . . . . . $ . . . . . .v.6........................;,................y....... ........q..........,.... . .e.,

g .n

. t .

. . 3 1 .

s. . .

s ,

. c

. . . . . . ..:. . . . . . .. t. . . n . n 4. n u o . .g . u n . u. n n ot.u n n o:nf n.o. o.o.o;.....o....o.ogo... .

.y.o... 4.,

.!..... g

, ., . .c

. . . . . . . e .

. . . . . . . s

. 3..

, . . . . . . . , e

.e

........v...............).............)...... .......(.................<...............e,..................;..g.... g

. .T

.s .

..s

. r .

s

.C

....... ;........ .. ...o........;.,.:.............. ...g.....

........ ............... ................ ........ . J. ..g. c

. .t

. .T

. s ..

I < .

6. . s . .

8 8

. .I

. .C g C3 t

.# .. . f

.O w

. .J

.e=

.C.D

. . J .

.i O c U 6.

r

o g . .

1. *C C

~~

e ..... .. ... ..km......... :. .O

a. ....... ....,..... ... . ....... ....

... . /... ....

( -c

.n @ 1 .

. g..

. . . . p. .

g . . . . . . . . .

O. - .

J..

s

= .

g.. . . .

~.

L . .

. . p

. .Q s

4........ .......s

................ ....... . . . . . .P. . .......

........................).s ... . . . . . . . - c

. a . . . . .

.N s .

O. C sa . .

".3 4 . ... . .

. . . .C g . .

.... ..........).....s.. -c

.....f.t....... .......<......  ;....... ................s

....... ...... q....... ........

O ..

s.. .N

- . . , .e O, . - .

g V.

. . .o

4. "*~ ~ '. . g 4 *""* ~ .* ~ .- ~ (". " ~ ~; a . " a <" ~ a a t a a a ". " ~ . " .-c 1

.If e -.~ ~.)~..."t.-ana)." .

a.t". . . . .

i . .

a, . . . . . .

...,.e W. . . . . . . . . . .

v .

• OO =

e.

. C 4 . .

MO . .

p ..a u 7. ..... o ;..n... .

...an.a.....~.**..a*...n... -c, a

-.m C

- . . g. .nu - .... . .

.=

. =.

i g , .

g * . . .

O .

O y

e. . . .

O C .= . . . . . . . .

................ ..............................................~ =c

~

... ......... ..e.. *....~..........".e n n

= . . . . .

! o . .

,,,) a .

[-

C .

g .

.l 1

~ .,O . . . . . . .

c a n, a, s, a, e -

- u n v

. n. .

. , e. .. a. . . , , ,

naqueaysy seeJoea

.#.s 1 fte s

. 48 f' l .. ..i. hg r.a

. ., e ,

• • m W Q S&

y e g  ;- C

. . I 6 . . .

. s . L':

$I

.-= . . .

g .

.s ps =

f."* .

a C

..g.. .. . . . . . . . . . . . . . . . . . . . . . . .

............t c I

t l

w 9

F

, s

o  :  :.  :.  :  :  :. .

s  :

J. L i

l a. .

. . -.. .- i r

(/") O a  :  :  :  :  :  :  :  : -

g .g  :  :

:  : ti,.  : l

.v o e a o .

e e u o c i

i o

o a e a e r- e- e o e

, o e e e o e e e o e c 1!.G U.U9)U.W S 9 9. J 6 9 G - . .

e

• - .- e s --s L

.L ..p1 * . ; .0.W. ;b k:Tu l

• o I

t a f -

W I.  ; **

p*C s

(.

• =* 'e

.l.

(se .*.".,,".g e - .

1 .18 C. e . .

W q*

i. i e: .

i

i i i.. . . .. .. .. . . f T

.. r . . .

a

. . s .

. s . . . . . .

. t r

. t . . . . .

s . . . .

............c j

............ ............ ....,8.... .

' "T* .

t .

...I............. .

I .T e

n . . . . . .

l 1 . .

. r .

,3 .

. . g . . . .

g .

.l . . . . . . . .

l s .

, . . . . =

. . .s . . . . . .

g

. . . . . \ a. ..

c 8 . . . . .

. . . t

. . o -

. r .. .

5 o

g. s s

c .

c

. o 6... . . * .

e . .

. g .

. r . . .

.s

,. o .

. . g1

. .so

., *Q o

. .A .

cv 2

...sen..t...v.....:...........*...........:...........:.........:...........:.......................:...........C lN Q g s

t , .

o.

Q ..

O t . . . . .

. C

.a5 . . m at. .

ws .

f.=

.S g . . . .

s.

)

. s . . .

. s

............ ......%., . . . . ............. ....X ....... ... ......... . . . . . . . . . . . .. . . . . . . . . . ... . . . . . . . . . . . .. . . . . . . . . . . . ............c N

s.

. s. g. . ,

s s . .

............ . . . . . . . . .....s........,:

............ M. *,,.,.....

. l, . = .

c .

c .

n .. . .

g w%

l M .

s. .

31

. I.

s L. .-= . .

... ....... c C . . . . . . . . . . . ..................................................* * *

...... r..-................... ...*....... .

I"

'O C  ::,,

I d"

. . " n. - . . .

i . .

e i n. 6  :

. t

. (

o a

g

. . P I-

[  :.  :.  :

M"" Q-C .......................:...........;.s.........:..................................:......%..:........................-m g

• l F. -

F I

f. 3 -

0 .

).

===* :l> .. . .

.c .

<. J

, C .

8 O

w .

o .

. * ;b

{3_ C3 o -

o e

e o e .o e o e o e o l

c c t.3 C *:* T M M e t t=.

o e o o - m e n . . e o e e

. 'v.GU.UBJyE.J l SG 9 J5 ea .

c

't

.cJ. .

.r, :,w, e .

t  :

m Li \..., \

1 i

. ,

• o e ~ ,- i e.

e <..

s ie i

,o .

M i . . . .

8.

• I < -

-  ;<  :  : . l'*

i

• I c..d W

$ 8- ..> * .

~.

i '

o. i

. .: i e

.i..

i

_ ........... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .......,... t .

............. ....................... ...........u.4 J  :

v

. . . r

,d .

l

.< . .e

. . . .. . e

,r .

,c

............. .................p..

• a.

.....r==,.,............................................. .

. . . v

. ,s .

.t g

g

. . e. .

.a .

o o-

. o . . .

. . . r .. f

. . . . . . . . . . . . ........... m

....._t................

g .

n r

. g . . .

F

.= .

. .e

. e . .

a . . . .

r .

.p .Q

................................a.1...

..e8.l .......... . . . . ... .. . . . .. . . . . . . . .*.............

. C1 I

. . c1 e

"O ,,,.e

. 8-*

== N em o

............ . . . . . . . . . . . . ..... ..... c,

{n

- o  %

.s

.r

,a... .

o o.

Q s..

A......... ..... ...... . . . . . . . . . . .. . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . .

e 3 eq g

i . . .

s. .

. A . .

( , .. .

. ~. . .  %. . . .

. s

. o.,,. ...........

................... ...................................... ........... m

. . . ....... T-

& n .

.- > i  : 3' . . . ..i  :

..- o . .

. . i  :'.

i l

g . - ,. . . Y

... ..................................................... ....... . . ... ..~ .............. ,'.. ................

Ie C' O. o .

i . ..  : .  :

~

r.

. .6 >  :. .

r

- - -- .s  :  :.  :  :  : '. . l-nO l

• g

.- s_  :.  : -

: g

s. . .%. .

.. . . . s

.A J  :*  : .  :  :  :*  :

\

~e

• 4 - w ...........

.................').................'............

.. @. ...I- C Y

o,,,  : .

N _\s 5 t

-e 3 ( ' e .,,,,  : *

.""~  !  :  : g s ,. i

. s.

g uI=

g u.

. . . . . . sb C, ,O.O o  :. i '..  :

i  !  :

m .

!. i. $

.C i' *

., o

. . . . . s o . o e o o o e o e o o o o a w n o a c w n a c N N N N N O C C e o

.  ! G G .J 010 n'") . . .

• • .m e.w

.- . a. .

n.

m.-.%.

.t

. a. L. s... . a . .

~

I. e d=  ; .

C

-, a * . a 4

'3

. . . . . . 8 l

s. 6 8

r

. c g .......... . . . . . . . . . . . . . . . . . . . .

...............................:.........9............,.....;.........,g..........g e* .

t

.  ; . i lp. .

.r=. . . . . I". T s t.

. . . . . . s .

t .

. . . . . . =

. 12".L....

m g5

. . . . 1 s v.

.. e i

3 g

e .-

. . . . . s

. . . . s .

. . s -a

. f s

.. g

, s u

........ .......... . . . . . . . . . .......... .......... .......... ....t .....'.e.p....... .

g .

;  ;  ;

. "f" C

3. . . . .

. . t .

. . . .. .e t .

. e, t *

... . . . ,. 4 .

. 8

.g

. .C,'*

. t w N 8 -

. . . s a, .e N 3...... . . . . . . . . . . ......... ......... ... ........... ..

.. ..... ......... ..fm,.....

f

. . , .., ..... n O g

. . . . c4 w t

O

,8 O . o..

W

,t .

, L.

s .

. . . . . g y

,e .,

. . . . t .f

. . . . . . . . . . ......................t

...........t

.....=0.a.......,.C

. n

. . . . . . a .

.  :  : ='=" * ;

• ..a=='""".,.

. . t .

I

: .  :. . p.

.j...** ..  : L

............s....***...

,, p

. .; . .  :.. I

.........,na=="............

~. ......c .

Ts 3 . . . . .. .

v. .

. . l s

. . . . t

• t =

• o .

.r s . . . . . .

e= 8

, 3: .  :  :  :  :  :

• I"** C t..

. l w

t 3 . . . . . .

- l.........6..s.......... . . . . . . . . . . . .

C O 2 6

Hs '

. . . . . . . . . . - . . . . . . ~ . . . ~ . . . . . .......... . - . . . ~ ................~..........C

!~

i

c. -

I;,

i .r

.f"". - > .s .

1c.

O w t *.(

... C I 2

. 4

w. . . . .

eJ g .

1.

["*"" O

,,,e l.......T"t..........:........4..........a...-~~..*...-.-*-.....J--.-J~.-..*...-...*.-~~~

t .2, g

t

. .e .e .

-c Y, C = . .

. L o 4

< _3:

-g e I

s . =

gy et

• O

,3 1 C I.*

. . O.; 1

. ~- * *

. _. .. . . . . . . e .

C O c C C C *. O C c C C C C 9 -p c Q Q  ?. C C T C3 N '"

pugucN :UGOJBd .

,. u r.

s .  ?.

52 - E .>i*r;

. .. .t.} rr t .m - e .

1. - 8 .h g

e

. . . . e  ;

• ~.

4 . . . .. . f

1. . '

. , I

' 1. . s g

*8 (

i, ,,.<8 3: )  : .

8 ...

== i .

s e

n. ,. , #  !  :. .

3

.l n.= . . .

• A

. \

w

..,,,,,,:e,,,,,,,,,

t yg t l, c .

. . p.

. . . p .

.t .

. E g. .

. a. .

..................... . . . . . . . . ..................... ........................ m, ............. .. ..............I.............

u

s. c j . . . . - * .

y a 6 . . .

g .

. . g . . .

. . . g .

. s. .

. . . f . .. .

3 .

. t ... .

. r .

s .

.....r.... ..... .............

4 .

...s......

.E.

n g .

. t.

. . s .

. . . .g. . ...

8

.o s

. . .. .. . . . . . . . ........... ,a. .

cc.g

. a

.. .o

. t. . .

. . . a . .

. . . o .

T*

,o . .

e .*

=

.w

~

m

. s s. .= cu

.........t,. l

. ............................ . . . . . . . . . . . . . . . . . . . ............r.

........... . . . . . . . . ........... . N O .. ,

I

. &w. - . .

C -

C o

. . /, * . . .

w) . y

• o.  :.  :.  :. .

. . s, **. # *

. . . . . . . . . . --. _. ,........... .................n..

. . c a..,*..../ ....... .................... N

.,..e.,,

c..

) .

. ,..s. ..,. .

) . . . . . .

..........s..,.o.... .................... .... ................

. . 4

.g . . . . . . v- .

. g. .

O c .

g [ .

.. . .. c. a. . . . . . (

e~ .

.s - %

s .

m -

=

4) 7

> ...............p....A,nn'.$-.~~.~.~~a.....;."".~.".~."4".~.....~~"Q....~...:="...."pC t

s r

< - s . g :.  :

==

.u. g I.

t: .

O . g. .

C a

o g g.

.n

. w . ,

.g . .

.) . . . s .

..onn.a."..".".".****.

["""".* w O

. . ~o o n . .

.""~a.

" a a a * . .. . . " a . . . ". ~ . . . . " .e- t:"J

. . - t u .

r r*. ~ ~3 g: ,-

<' -.J U g i i  % i i  : i .  : i f

i  :  :

CO O e  :  :  :  :- i i gg gj .

i. .

[,

c c c c O .C C C C C C C C C v3 c n C t.*3 C C C W C W C t:: v v O M N N *- ** C c G G N N N N N N N .N N N N " *~

^

. + i ..s

- * . f

. g:j.

., - . .s..6 m.* i.,

. g C

I, g . . 8  : """ . 4 g 'er.-

t

<* 8 ,

.I

• .  : . 8  : .

.t

"")

. 8 . .

I,. gr .'

.8 8

.=. g

,f e.'"

e=

. . '. 8

. . g 31 ,, . . . .

. g .

8 -. . . s . . .

=

. . .t . . . . .

.......................!*........... "D.......*......*... .:...........:............*.. .......:. ......... .: vC O  :  : ,I

.. 3 . . . .

g g ,

. . .n .. .

. I . S

. . i 8 .

L .

.......,8. . .

I s ............ ...................... . . . . . . . . . . . . .............C 8...............a.u........

. g .

. . I.. .

F..

. t .,

g.. .

g g

t- .

,t .

g .

........... .u ~ ................... m

.........................s........y.. ........... .... ..... ........... ...........

l l t l -r* .  : C'2

. t . . i . .

. g .. g  ; . .

. g .

. . .. 1 . . . -

... . . t . . . . *

. g . .

}. .

. g . . . . . .

\

g..g. ..

......................... .........................e g

........................a... .. . . . . . . . . . .

c g .

. . l. C7 g

,o . . g .. .

g g. .

. g . . . . .

g g .

.C

. . . g

. w y 3

• cs

.. ................. ............ g p

...........I"."

td

. . . . . . . . .. . . . ............. m N O =

. 3 .

• .i U h.

. .:. g  :. "a  : -

. g

. f,O ) ..

. g \ . . . . -

m. .

. t . %, . .

g .

. . g . .

. s . . .

.. . . . . . . . . . . . ..............g

............ . . . . . . . . . . . . . .v g 6....... ........

... . . . . . . . . . . . ............... . . . . . . . . . C

. .g . .

.............N a

. s . . .

.g .

. . . g .  % . .. . . .

l . * .. g .

. . s g C *  :.  : s  :. *

- g.

- g .

s .

. . g . . . .

t- .......................

s,.........................b..

..........................*......**.* = "v.*

n  :  :. N  :.  :  :  :  : =

t

%. g g

. - . I-O_ .

s

. . . s. . .

e === 0  :.  ; *

%.s

• 2 l  : . . *  % ; .  :

n

. . . . . . . . . . . .. . . . . . . . . . . . .............................s......... . . . .w. . . . .. . . . . . . . . . . .. . . . . . ..'. . . " . . . . . . . . . . . ............*c

.  % r . .

ts . . . .

.e - . . s .

l l .

I.'." - , ,

. 5.

4 .- ,, g . . . .

n 1.= . . .

. -. O

~ . * . . .

.~~ m.

g. s . . .

,a - -

.-~~..aa".;a."".

l g"""., C

......................:..............~.~.........~aan.~

. . . .s s . . ~ ~ .

y

" .-- C e

a -

[.

p 4 .

~  : . .

j . . . ss .

4 n .

,O n .

n . .

gy * .

{ y *t e . . . , . .

C c N c v C C N C T C C O C C "C C N N C C C Q Q C C C C 4 C C C D

= ==  % as i }e

.- m C p gwg g g.se'")g .g =ll='. =.e 3 C C .e .e .- *:* ,

- f~ *.7., s!* V.

i

. W. . .

, r.. . .OL. yc t

.g g

, r .

e. ~.

. e s 3.

==s s .

. s

' ,- l g <g ee H ',

8 3

, .., 3 . .

.g gi 1.: t C.. .

g D.

8 i

..................g................. 3 .

. . ...... ........... ................................................, . ....... ,,, c

. . y

.=: .

0 3 . .

. g , .

. g . . . . . .

s s. .

. 8 . . .

3 . . . . . .

g 3. .

g .

........... .....,...fg..................................... . . . . . . . . . . . ..... .................................... ....... gTe ,

. g .

g . . . .

. g . .

. 3 . . . .

. . g . . . . .

. 3 .

3 g. . . .

g . *

. . . . . 8

. g . . .

g . . ..

............... .....a *

aO

. g  !" . . . .

3

. 3 . . .

. 3 . . . . .

. g . .

~

3 .; .

g . * .

g . . .. . . . . .

..- ........s....g.......w . .. ........ . . . . . . . . . . ......................................

.... O g g g .* .

. . * * . . . = s.

s g . . . . . . . . .

. . g .

g . .

3 . . . . . . .

• g .

y m N

. g . .

~ .'  : 1 a

.............g...... .....................................................................................t**'C g ,..... .

.C

. . g . .

. g . . . . . ..-

g .. . . .

. g . . .

i . .

. s.g .

t  % .

g t . . .

=

t g . . .

g i,

. . . . . . . . . . . ........ g

.. ............. . . . . . . . . . .............. ........................ . . . . . .. . . . . . .. . . . . . . . . . . . . ........... .N

... g . . . g . . .

. g.. .

.a .

4s . .

... .g

• s .

g .

g .

s. .

....... ............s .......... ............ ...........

... ...... e

• 1=

c

.

• s . .

. . s s .

e== c .

F a . * .

.=.......s...................................

O ............. . . . . . . . . . . . ............. . . . . . . . . . . . . .............

. ...... C l *

• l e h $ 5 .

f. ,

s h.

, .- w .

O .. . . . .

• O C.

c .

O *

.e

= ........................ ............

.................. v......................... -c

> .O . . . . . .

y s .c . .

. . . . . .<3 .

O.  % . . . . .

O *

.:0  : ,.

O c O to c c O W O

.c to C C c N N C . Cl3 C C T T m o e e o e o e o e c

. 3 T l.a- U 9 ) E.J. -=ce - e .

~

~

55 - i 3I)I O [n E4

.e.

r m t.n 3 c

,. g .

,"3

. 3 <e g

. 3  :  :

g .

.8

• 5 . i . -

<g H .

s . .

:.  ::  :.  : 4  :.  :  :.  : -.

. ,.., 3 He

== .

.g g

L ::. g C  :. '

' s * -

. . .l:

...........:.......... 1..........:...........:....

g .

y  :

@ ....*...........:... ......:..........3...........:. ........... .,, b'3

g 8 . . .

..g . .

, . .a .

. . g . . .

.. g . . . . . .

3 . .

.I . .

............ ........... 3,...........:................ 44 .....

. . c

=. g . . .

8

. l .

. . . . . a

. . 8 . 8 . . . .

g. .

. g g

. f. . .

. g

. .g .

. .l .

....................... , .g....................... . . . r a.,6.... . . . . . . . . . . .....

g .

.....................................3 g

g*

. . a . . . . .

. l . .

g .

g . . .

. . g . . .

. 1 g .

. . l . . .

. g .

. g . ,

............................J,........... .............. . . . l,i~ 4. . . . . . . . . . . . . . . .... . . . . . . . . ....................... ........... -

c eg g . g . . .

s =

a

. 8 g

g. .

. . s . . . . . .

g .

i

.M l

g . . .

w u.,

. t . . g

. . e. . cu

.....l..................s .....s ,................ . . . ........ ...........

............ ........... g.S O c:

. . . N 8

g .

=

C 6.

. s .

. . . C g (f). .,

l

. . .:s.*

. g  :

. -+ s . .

. . g . .

., s, . . .

c g .............

.C

_a g.....................

n

. s .

,r .

s. .

. s.s

. s . .

. .g . .

s .

s. .

s . .

O .

• g,9

..I.....................

.........%.s. ........... .. .......

=

g..

........... I.............

T=

.  %. . s. . .

H . .

. .e, .

n .

. . . .i .

. ._,- c sg. . .

I a . . .

n- -

c.

W-O . .

c .  : .  : . '.: .

. r L.  :>  : .

., = .=

. *

s O  :::  :  :

. . N. .\

Q N.

. _C* .

e. J f"" -

.J 6.

. j

......................................................................................................'...........h...

j j j Q + g3

. . - I w 3 . .

• \ t

. - . e F. .n .

5 I

- -  :.  :  : -  :  :  :  :  :. a

<m y. =

=. g bisr*I n

g=

g rg JO C . .. . * .

.i e

. .. =.

. . .e L.= C. C C C O C ..C C C C C C C c C C v N O C C v N C c N N N N N C C C C C O N

• t .s c 9 OI O w

1 *

, . . ., - . . . . . . . . . .- . ---w ~ - -

- 30 -

• C

• =re *

e -

.. W3

- p.a <t 'e -

-.t.t

$ =* ,

. . . . B . . .

s .

. . . S . . .

.~. .

8

. 8 . .

. , . .e te L E

t e

g .e

. . ,, , *W1 L

. ...............................'.....................;........;.....e'C.*,J...................:..........*.. . ......

6,,, .

. 8 .

. . . . 4 L . . . .

g g

. . . {

. t .

g s..... .................................... .

. . . 3 . .

. . . e , .

. . 3 . .

. .. g..

m .

.. 3

.,,,,.n.

. f s .

. .t . .

.t . . . . .

b.M,. "

.= . ,

....................... . . . . . . . . . . . ......... . . . . . . . . .. s . .... ap.........  !.

4

. . . 8 t .

.: g .

.8

) .

. .3

. .3

. . . . . . . . . . . . . ......................3......=....M.,.........:.........:..........,.:.. .......

e g . . ,

. gg ,

j

g. .

g -

. 8 . .

. . . l .

. . 8 . . ,,

e l.e ,

.e .

N

. . F.. . . .

ig3

..n...-. . n .. u a .. .......

...o e e O

........... .....y... .......... . . .. . ~ . . . . .....no. t. .. ........

N w

o. .

. .o S.

s ..

, . t o .

r .

s

. I*. I . .

s .

. . , c

.........W

. s.. . . .

3 .

1

. . . a . l.

~ .

op. . .

. . . r .

r .

,r

. s .

. . ...........9J .

f.......- ......... . ....... .

.......... s.

1=

(. .

,r eb

• r.  :  :.

1

- ......... . . . . . . . . . . . .. . . . . . . . *.. * . . . ... - ......... . . . . . . . . . ............. ..... ..........c w=

,e.

.s. .

o.'= . .= .

~ . ,. .

m .

q - > ,. o .

a;-= Q w  : , *.

i  : j  : j .

,a ,

~ 3 c  : ,. . .

!.,.,.,,,g 3  :

a -

.= 3 ..........,!.*..,./. ..........................................................~.........................n.

- V3 o .

u .u.

- c.. .

,3, t. . .

>a 3 8 s.

. o .

a e.. . .

O .-

42 . .

1W  : - - .

e s -. . . . . . . . s C C C C C C C C C C C O ""

C Q c ha O C .c ."'3 N pu;ujoN WWSd 9- , -~ , . ... . _ _ , ._ .o..._, __

e .______ _ _

57 - .

c "9' 8

• T.

sl

i. i i -

i. .

l *L '. ' 2'. -

i.

i.

. s .

.s e.

<l-t wa s ea .

.i

.i

=

1 t

i i

8

=8 -  :*  :  :  :. .

,b

. ....................................................................I t

. . g .

. . . * . 1 .

4 g .. 1 .

. . . g . . .

s . .

.

• g . t .. .

t . 3 .

=v

................ . . . . . . . . . . . . . . ................................. ......../...... 4 t

n. .s.........................................\C..

. y

(# ,

c"\

= . . . .

5 .

.g %=t. . . . , .

. . . s . s . .

= s r . *

... 1 m

........ .... ................ ............... .............. o............./ ............. . . . . . . . . . . . . ........

.t

. O

. f

1. r:  : . '

!. . s .

p .

. r. . '.: . .

.  ; ** *  :. i ;  :  : .

• s .

. . . . . . . . .c

.......... ... . . . . . . . . . . . . . . . C'3

- .  :.  :. s  :. =%:  :  :.

. r . .

. . t., .

1. . s. .

o . ..s- . ..

.r .

. .e g.

s. . .

v c3

• *  :  :  : . C". cv

.:  : i

,....... ............ - . . . . . . . . . .......r ,,.. ................ . ............... m O m.......... ........................ o s *. .

, .. . . N u

. c O -

. Q g

~~

i.

5 .., Y.  !.  !  !. .

(f) b.

i. ,g#

s .

.o *... *

• s

M. *

...............,"N

. c

. . . . . . . . . . . . . . ................................................* . . . . . . . . .; . . . ...........g..............

• o *Y .

. g

. t . . .

t .

. 3 . .

. ls. .

t . . .

................,g,.

.l......

?

i .

I g .

3 t g . .

O .

I g . . .

3 ,

,b=

==

}

. I *

• t -

6- a . l m .s,.. . . . .

. s- > ...............

.....% . . . . . . ..*. .i ..................e . . . . . . . . . . .. . .. .

C,=

m i

• ~

h  :.  !.

...~

(""

• C

~

- M. .

8 t.

t

.n g 3 . . . .

= C* . I  :  :  :  :  :

• C  :.  :.  :.

=

~ .= Q

- w #. 8 l

...............i.....*.......9.. .............u.........!.........................................................,

.d .u.

I u . a.

l.

o .

w . . .

m. . . . .

=*

. e"s  !.

! h

. N.% s .

i. .

=3 . .

O g .

Q . .

6

. . . s . . ,

t .

c t o.I w -

C C C C C C c .

c C C V3 C W c c. to O. . L3 N N

"'" " C v af C"3 C'3 N N M N N N N N N -

. . . _ _ . b. Q

~ -_ _ - - _ _ ._______________'.'l..M

e 58 - . . . .

r.

O e A .

I - . m s <'

8

. . I .

C' .s . e .

s

" s. eee

. }

. . 3 s

- s . . . .

--3 4 h3 .

. # . . e Za .:

- a . I Q .lla.cd  : C[6

. . . g .

. . t

1. $. d

. ...., . .........i......... .......... ...., , . y

. . t I .

s .

. . . . . 4 . # , .

. t  : .

.  : . . I ;e .  :,

. .  : {

a

.: ,g

. 4 .

.....c......................... ....... cT t

. g i e

. r* *"*  : .

8 s .

!. =

a

3. . .

hm .  :  :.  :. ,#; .

} 2 .

I g .

e .

. Is .

o .. .

. . . r . ~

........:....... .... .W

.........................................as.. .

gg

,8 .

. . =

. . . . o . . . .

g

. . g .

t .

. . ..==.. . . . r ,

t .

.. .t . .

. .......... .........c..

. . . . . . . . . . . . ........g4..... ... . . . . . . . . . . .e . . . . .. . . . . . . . . . . .. . . . . . . . . . ... .. .. . . . . . . . .

. o.3

. g

. s .g t *

. s. .

. . . c . .

s .. .

03

. . y . . .

s c .

V C4 N

C

1. .9 . e o . . . . . . . . . . ........... ........... . . - .........

. . . . . . . . ~ . ..~s v........ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ~ . ~ ~

mO N O

.i

.o O i y.

. p.

J Q g d

o .

Cf) =.= b t . p . =

o.

e . .

.t . . . . r . .

. . . p . .

. o .

. . . . . . . . .. . ................ ..... .... -. Q

............f........-6...................................

. . .o * .

. .e. N e .

o.. *e .

o . .e .

s. .

. . oo . .

.t* . . . .

O .

o .

. . o . . . . .

.= .

C I""

.i ,

M *.g -

. s......... ...........

. ~ . . . . . . . . . . . . . . . . . . . . ~ . . . . . . . . . . . . . . . . .. . . . . . . ... . . . . . . . . . .

se m  :  ; a  :.  :.  :  :.. .

y ,d .

g .

. s t

s .

s = -Q r=== ,,,

• . o g . . .

e

[" g,,,, .

• s .s s

l w s 3

. . -.~ p.~.~;~. . .

g ~~~~.~~.-~~~;g.-- ~ . ~ . . ~ . . ~ . ~ ~ ~ . - ~ ~ . ~ . . ~ . . ~ . .~ ~ ~. -~ .ns

' c,,. .

l f =* O

-=

~

s* ,, <..

I f*

[

e . m q s

.-o=,.*=.,s u  :.  :  :  :

g o .

O . . . * . . . . .

O . .s . .

t e

e  : *

e.  :  :  :

.,=

n o . . . . . . . . . .. . . . . . . . . . .. . . . . . . . . . ... . . . . . . . . . . . . . . . . . . . ~. . . ~ . * .

.~.~.- ~ ~ ~.~...~ .....- ~.- ~ ~ ~ ~.~ ~...

- to

<.*j

. w . . .

. Q .

. t .

. cs . .

O n Q

1Q .

i- ,e

, e a s o s , - . e a o e n o a as c O N P'= N O t.3 C C to e m LO 40 C C W Q m. fO l .

naqueayed S08J600 l .

3 . .

. _m . n _

e . 8

6

- - 59 -

• l l

l

- 1 I

C

-, 8 6 -

"T" 4

- .i . s

. . l a

'll 9  %' 3 1 3 . . .

8 Cs I

.s. i

._ [4[ g t

a.! Os g .

L..

• g .  ; .

QI L*,. s C,

• l8

.s .g * * *

.gg......

- 1......................e.=... .. o

.... . ........ ........ ........ .. ....... ,,, g s_. . . .g

• , .g . . T

. ,, .s . .

. .g . . .

, h

• p

. ,. E .

j

. . 3 .

. t . .

t, . .

t

.. s. ......g.......... . ..... ... ............... ............. .... .,... . ...... _ Q,

. 8 -

. . . . T t *

• f . .

. . y . .

• r**"8 *

• e,

. s- . . .

o.m., e.*L .

. p4 htrs 4 5 . .

. e g . . . . .

. s,

, C= .

. 8 . . . .

g. . . .

............ . . . . . . . . . . .. 4 ..4.e.. . . . . . . . . . ...................................... ........... ...........

m

. =

M t  :

t  : .

. 5 s. .

. 1

s. .

e .

. . s. . .. . .

s

. s.. . .

. . . . . . . . . . ........s.......,,s........................................ g .

. . .:J .

8s . . . . C'3 s,fs . ..

Is . .. . .

. ,e e

.O s

t c om C

. g

. . . . . . . . . . . . . .. . . . . . . . . . .m gy O rs cp i.

. . =

. s . . . =.

.  ; s . .

Q m

. m .

. s wg U=

.s

.s

. s

. s .s 9.............

s

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . . . . . . . . . . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . o cy

.s .

. .s

. s . .

t s .

. 6 . .

. g .

.. . g . . .

s .

. g . .

• . s . . . . ..

• s

. s. . .

. .......... . . . . . . . ~ . . . . . .. . . . . . . . . ..

.....a.a.a..~.a.a.....aa.... . . a . a a.a.a. . a a .

m

. T*

c ,

c . . ..

Q g .

e I"" .

! s [

O o .  :

's .  :*

["""" i .

4.

• O e g.......................................aa.aaaaan s w.a-.*,~~~..a.- a a..~.. ~~. .; a.......- C w

3 .

. "-s .

c-. O I

s  % * . *

• 1

,* .N i .. .

i .C .

.O u . .

o . .  : .

- .e 2.

e .- .

,s.  :.  :.

O c ..s ss o . . . .

_e w

t- ....................... ...........

. ........."..a...~~aa...a."=a.a.a"*a~.~..*.

. 9.~.a.v.aaa.".~a.a. = LO

.o . .

l; . . * .

.c . .

o  :

U) O d -

o .i

i  :

i

i 0 i Qg g . * .. . . .

I

,, . . . i e . - a e o e o .o e o e o e o o r .

, .a a m a v , e n sa @ 1D e c sn C @ tD C C

. 11ayueJye i sea 20a0

.____ .. ._n m m ^"c" '" > n < w ' "" "'

. . . j s .

t 9. . s .

4

,- . s . 1 '3 1 g . . * .

M s . .

• 8 C s.,.

' .C  :.

. i. .

8

.e ,

t . . . .

6, y 'A s L, 3e

• e r .

Os ta. s Cl

;: ,") -

'L i.

,._ . . . . . . . . ........ ... ..... .. . ...... .................... ... .. '........ . ... ...............,e..z_ . .

..... , , L e

. ei...., i e i r

=

e' t

=. .

C t

. c. .

. . i.

. . 4 .

. l  : t:

I e s,  : r

......................<...........f.......T'.;;.

'8

. . . . . .t._ C i.

g

4 T

g. '

o. m E.""O"

i f i i i i si s we.e.

. s . !.

8-e,. 18 I  :  :.  :.  : *

.  : e f* . .

g l. 45 *

. . . s

.a

....................,t........,..,........e

......... . . . . . . . . . . . . . . . . . . ................... , gg.3

. g

. . .- . 4

s. ...

. c.. .

. ~

. ,# 4 .l. .

i . .

. . p

. t

. .(

t

. .s...... ..................,.e ,

c

. . .g

. - C'3 .

.6 .

.e

. . . r ..

.o . .

. . cr

. . . . . c.. .

. ,o . .

. . . s .

g o .

m

......... . . . . . . . . . . . . . . . . . .................. ................... ..o........ . . . . . . . ............ ....... . . . . . . . g. .

o

• . s.  :  : .

N i

. .o .

. . a

. ,o .

. . p. . . , . . O

. W.

.l . .

. . . . o .

• .g . . ,. .

. o .

.. .......... . . . . . . . . . . . . . .* . . . . . . . . . .. . . . . . . . . ...........b..........<........ . . . . . . . .. . . . . . . . . . . . . . . . .. . . . . . . . . . c

. . . . . . p. . . N .

t

. o..

o .

.e .

. o. .

. . s .

.e ,

. . . ..o . . .

. o e . . .

r o.. . . .

. . . .o . . .

Q g l C ...... . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

...........oy g.... .

T i . . .. . . .

. e . .

. . o. ., . . . . .

n .

s. .

,o .

_ .e .

o $ . .

. 3 . . . . .

E[

. . g . . .

.w -

• .  : I a

.. ...........-......... . . . . . . . . . . . . . . ......................................a"1............................L.

"-- M-C.:.

r- O C 3

~  ;*.,~~,,

- [ ".

~

i

=,

(""1 .

l  : ,

:. _""4**.

<~ -

:. s,:

l 1

3 .C" .' .

. . g ,

l O J a .c w t.......... . . . . . . . . .. . . . . . . .; . . ............ . . . . . ..; . . . . . . . . . . . . . . ;. . . . . . . . ...................... .-.....

- its C ,  ;  ;  ;  ;  ; .

- N . . .

. N.*  :.

. s .

. . . . s . .

a a .

o .

, l3 3.

.C O O O O Q C O O O O O O O w c a c N C t.r3 v C'S N "

O O CD

,5 N O Q @ Q Q C C C O W 19e3 o[qno l

IftLL/lL.0194 l evy Aleast 40tddepeassade l

\

. - 61 - .

1 <

1 l . ~4= , - -

e I

c".2.

. .i a

<'i c

a _ fm

.