surry, units 1 & 2 - reactor core thermal hydraulic analysis … · 2019. 11. 12. · a thermal...

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VEP-FRD-23 AUGUST, 1978

REACTOR CORE THERMAL HYDRAULIC ANALYSIS MODEL USING

LYNXl AND LYNX2 COMPUTER CODES

FUEL RESOURCES DEPARTMENT

VIRGINIA ELECTRIC AND POWER COMPANY

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VEPCO REACTOR CORE THERMAL HYDRAULIC ANALYSIS MODEL.USING LYNXl ANDLYNX2

COMPUTER CODES

NUCLEAR FUEL ENGINEERING GROUP FUEL RESOURCES DEPARTMENT_

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA

AUGUST, 1978

VEP-FRD-23

Recommended for Approval:

~LJL;tt M. L. Smit~ .

Nuclear Fuel Engineer

Approved:

/11~ ~- ~{;~~1 M. L. Bowling, Direct Nuc_lear Fuel Engineering

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CLASSIFICATION/DISCLAIMER:

The data and analytical techniques desc.ribed · in· this report have . ~ .

· be·en prepared ·s-pecifically for application. by_ the Virginia Electric and

·Power Company·.. The Virginia Electric and Power Company _makes no claim as·

to the accuracy of the data or techniques contained in this report if used

by· other organizations. . Any use of. this report or any part thereof must

have the prior written appro_val of the Vir~inia Electric and .. Power Company ..

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ABSTRACT

The Virginia Electric and Power Company (Vepco) has developed the

capability to perform steady state core thermal-hydraulic analysis of the

Surry Nuclear Power Station. The purpose of this capability is to 1) deve

lop expertise in the thermal-hydraulic area, 2) eval1.1ate fuel designs from

various suppliers, 3) support reactor operation and 4) provide input into

the reload core design and licensing process. A thermal and a hydraulic

model have been develo~ed to explicitly

S . • h L· YN. Xl ( 1) . d ( 2 ) tation using t e an LYNX2

represent the Surry Nuclear Power

computer codes developed by Babcock

and Wilcox for thermal-hydraulic analysis. The Vepco developed thermal and

hydraulic modeh, used in conjunction with the LYNX! and LYNX2 codes,

have been verified by comparison of results calculated, with this model to

the results from several thermal hydraulic calculations used in the design

and licensing of the Surry Nuclear Power Station.

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TITLE PAGE.

CLASSIFICATION/DISCLAIMER

ABSTRACT.

TABLE OF CONTENTS.

LIST OF·TABLES

LIST OF FIGURES

..

TABLE OF CONTENTS

. . .- •.

. . ..

. . .. • ..

SECTION 1 - INTRODUCTION ·•

. . .

· SECTION 2. - OVERVIEW. OF THE LYNX! AND LYNX2 COMPUTER CODES· ·

..

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iii

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.. . . vi

.• .. •· 1-1

. . 2-1

• 2-1 2.1 INTRODUCTION

2~2 LYNX! ••• 2-1

2~3 LYNX2 . . . . ·. . . . . . SECTION. 3 -HYDRAULIC MODEL FOR SURRY UNITS .1 AND 2

3. 1 INTRODUCTION . •. • -. . • . •

3. 2 AXIAL LOCATIONS AND LENGTHS ·• . . . . . . .

•.

.• •. .• .•.

. · ... 3. 3 LOSS. COEFfICIEN'i'S AND SURF.ACE ROUGHNESS .• . .... . ... 3.4 RADIAL FLOW AREAS •••••..•

SECTION 4 . - THERMAL MODEL FOR SURRY UNITS 1 AND 2 ·

4. 1 INTRODUCTION • •

.. 4. 2 · SYSTEM .PARAMETERS .. . 4.3 REACTOR CORE POWER AND INLET. FLOW DISTRIBUTIONS

4. 4 UNCERTAINTY FACTORS • • • • ·• • ·• • • • . • • • •

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4. 5 SURRY FSAR AND. CYCLE 1 ANALYSES - CASES .Al AND A2 ·•

4.6 SURRY DENSIFICATION ANALYSIS - CASE B. ,

4. 7 SURRY·. 90 PERCENT FLOW ANLAYSIS - CASE C •

SECTION 5 - RESULTS. AND ·COMPARISONS . . . . . . •.

5 • 1 INTRODUCTION • • • •.

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2-2.

3,...1

3-1

3-1- .

3-3

4-1

4-1

4-1

4-2

4-2

4-3

4-4

4-5

5-1

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5.2 SURRY FSAR COMPARISON - CASE·Al ....... .

. 5.3 SURRY UNIT 1, CYCLE 1 COMPARISON - CASE A2.

5.4 SURRY DENSIFICATION COMPARISON - CASE.B . . . : . · 5. 5 SURRY 90 PERCENT FLOW COMPARISON - CASE C ...

..

SECTION 6 - SUMMARY AND CONCLUSIONS

-SECTI0N·7 - REFERENCES ...•... ..

. 5-1 ·

5-2

5-2

5-3

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. 7-1

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Table

3-1

4-1

4-2

4-3

4-4

5-1

5-2

5-3

5,...4

5-5

5-6

. 5-7

5-8

5-9

5-10

5-11

5-12

5-13

5-14

5-15

5-16

LIST OF TABLES

TITLE

Hydraulic Model Parameters Values for Surry Fuel Assembly

Case Al - Reactor Conditions

Case A2 - Reactor Conditions

Gase B - Reactor .Conditions

Case C - Reactor Conditions

Case Al - Core Av~rage Exit Conditions

Case Al - Highest Power Assembly Average Exit Conditions

Case Al - Highest Power Subchannel Exit Conditions

Case Al - Conditions at Location of Minimum DNBR

Case A2 - Core Average Exit Conditions

Case·A2 - Highest Power Assembly Average Exit Conditions

Case A2 Highest Power Subchannel Exit Conditions

Case A2 - Conditions at Location of Minimum DNBR .

Case B - Core Average Exit Conditions

Case B - Highest Power Assembly Average Exit Conditions

Case B - Highest Power Subchannel·Exit Conditions

Case B - Conditions at Location of Minimum DNBR

Case C - Core Average Exit Conditions

Case C - Highest Power Assembly Average Exit Conditions

Case C - Highest Power Subchannel Exit Conditions

Case C - Conditions at Location of Minimum DNBR

V

Page No.

3-5

4-8

4-9

4-11

5-5

5-5

5-6

5-6

5-7

5-7

5-8

5-8

5-9

5-9

5-10

5-10 ·

5-11

5-11

5-12

5~12

--------------------------------------,--~-------~-------~-

I I FIGURE

I 3-1

.I 3-2

3-3

·1 3-4

4-1:

I 4-2

I 4-3

4-4-

I 4-5

I 4-6

4-7

1· 4-8 ·

5-1

I ·5-2.

5-f ·

I ·5-4

I. 5-5

5_;6

I 5-7

5-'-8

I 5-9

5-10

I 5-11

I 5-12

5,-13, -

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LIST OF 'FIGURES

TITLE PAGE NO.

Side View of Surry Fuel Assembly_ 3-6

Cross Sectional View of Surry Fuel'Assemblies 3-7.

Unit Cell Subchannel 3-8

Thimble Cell Subchannel 3-9

Relative Assembly Flew Distrib-ution for all Cases 4-1_2 -

Relative Assembly Distribution for.Cases Al and A2 4-13

Relative Pin Power Distribution·for Case Al 4-14

· Relative Pin Power DistJ:1ibti.tion for Case A2 4-15

ReiativeAssembly Power Distribution for Case_B 4-16

· Relative Pin Power Distribution for Case· B 4-17 . . . ' . . . . ..

. .

Relative Assembly Power.Distribution for Case c 4-18

-.. Relative :P_in- Pqwer .Distribution, for Case C 4-19-

Case Al - Hot Assembly Mass Velocity. 5-13

Case Al -' Hot Assembly Enthalpy __

Case Al - Hot.Unit Cell Mass Velocity

Case·.Al Hot Unit. Cell Enthalpy·

Case Al - Hot Unit Cell- DNBR

Case A2. - Hot ·Assembly Mass Velocity

Case A2 - Hot Assembly Enthalpy

Case A2 - Hot Assembly-Void Fraction

· Case A2 - Hot Thimble· Cell Mass Velocity

Case .A2 - Hot Thimble· Cell Enthalpy

· Case A2 -·Hot Thimble Cell, __ Void ·Fraction_

Case A2 - Hot Thimble Cell DNBR

Case B - Hot Assembly Mass -Velocity

vi

5-14

5-15

5"'."16

5-17

5-18

5-19

5-20.

5-21.

·. 5-22

5-23

5-24

5-25-

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FIGURE

5-14

5-15

5-16

5-17

5-18

5-19

5-20

5-21

5-22

5-23

5-24

·s-25

Case

. Case

Case

Case

Case

Case

Case

Case

Case

Case

Case

Case

LIST OF FIGURES (CON' T)

TITLE PAGE NO.

B - Hot Assembly Enthalpy 5-26

B - Hot. Assembly Void Fraction 5-27

B - Hot Assembly Thimble Cell Mass Velocity 5-28

B - Hot Thimble Cell Enthalpy 5-29

B - Hot Thimble Cell Void Fraction 5-30

B - Hot Thimble Cell DNBR 5-31

C - Hot Assembly 'Mass Velocity 5-32

C - Hot Assembly Enthalpy 5-33

C - Hot Thimble Cell Mass Velocity 5-34

C - Hot Thimble Cell Enthalpy 5-35

C - Hot Thimble Cell Void Fraction 5-36

C - Hot Thimble Cell DNBR 5-37

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SECTION 1 - INTRODUCTION

The Virginia Electric and Power C~mpany (Vepco)has developed the

·capability· to perform steady state core thermal-hydrau'iic analysis.· The . . ..

docµmentation and verification of· this capability is provid.ed iri this report .

. The purpo.se _of this capability .is to 1) develop expertise in the thermal-hydraulic

area, 2) evaluate fuel designs from various suppliers; 3) :support reactor opera

tion and 4) pr~vide input i.nto the reload core designandlicensing process. ' .

The thermal hydraulic analysis capability documented· in this.report . .

consists ~f thermal and hydraulic models which have been·de.;,elop~d to expli-. . - .

citly represent· the Surry Nuclear Power·station. (The same modeling techni-

·q~es will be used for the North Anna Power Station). These·Vepco developed

model~ are. used· with the ~YNXl (1) ·and LYNX2( 2) coinputer codes developed by

Babcock and Wilcox for thermal.,-hydraulic ·analysis. The method of solution,,· ., ·;- . . . . .

correlations used; input .irequirements, and accuracy of. th!:! LYNX!. and LYNX2 .

computer codes have be.en documented by Babcock .and wi'1cox in References. 1 and

2.

The hydraulic.part of the Vepco d~veloped thermal and hydraulic • I • •

model for the Unit 1. and 2 reactor cores of the Surry Nuclear Power Station.·

describes the-physical characteristics which are important in calculating

coolant.flow. The thermal part of the model describes the core power distri.,

bution, system parameters and core inlet coolant conditions which are import

ant in calculating coolant flow and determining local coolant conditions and

heat fluxes.

The LYNX! computer code calculates · overall core c.oolant flow on an

assemhlywise basis, based on the thermal and hydraulic model which explicitly

represents the Surry Units.No. 1 and 2 reactor cores. LYNXl also calculates·

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axially dependent crossflo.w boundary conditions. for the highest power assembly

which are-input.to LYNX2.· Usirigthese boundary conditions and the thermal and

hydraulic model for Surry, the LYNX2 c_ode ·is.used to cal~ulate the local cool

ant conditions and heat'flux for each subchannel in the highest power assembly. . . . -

.Based on the co~lant conditions,. heat flux,· and a user selected critical heat

flux c.orrelation, the. LYNX2 code calculates the local coolant· properties which

are used to.determine the minimum Departure from Nucleate Boiling Ratio (DNBR).

Several. comparison case.s, based. on ·calculations performed for the

design and licensing of the Surry Nuclear Power· Station,".have been formulated to

verify re_sults calculated ~ith the Vepco developed therl!lal and hydraulic model

. -used in conjunction with t~e LYNX! and LYNX2 computer codes. Results of thes.e

calculations· will be presente·d · and compared to restil ts of calculations. used

in the design and licensing of Sur;y Uni.ts No. 1 and 2.

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---~------- ---~-~~~--' ,,. ...... ·

SECTION 2 - OVERVIEW OF THE LYNX! AND LYNX2 COMPUTER CODES·

·2.1 INTRODUCTION

The LYNX~(!) and LYmC2< 2 >- computer codes have been developed by Bab- .

cock and Wilcox for thermal hydraulic analysis of _PWR re.actor cores. The

LYNX! computer code is used. to _determine the coolant_ flow and conditions in

· the reactor· core. ·_ Each fuel assembly is represented· as a single flo~- channel. . '

. and calcul~tions: are perf~rmed '~hich -consider 'the effect of intera.ssembly

mass and. energy exchange. - The interassembly cross fiows for the highest power . '---

a_ssembly, which· are calculated .by LYNX!, are. used- as bounda,ry conditions by

_the LYNX2 code. ·. LYNX2 is used to d~termine the coolant conditions and DNBR

v_alues in each subchannel of the highest power assembly .based on the_ cross flow.

conditions calculated by LYNX!-~ The e:ffects- of intersubcha~nel diversion arid .

tU:~bulent c:~ossflow on: subchannel coolant conditions are included by LYNX2 in·

the calculation· of local coolant. conditions.·

2~2 LYNX! , .·· '' -. .· . .

.The-LYNX! computer code is·based'.on a one dimensional (axial)·solu-

tion o( _the conservation eq~at:ions. for mass,. momentum and energy with a simpli-. ' . .

. . .. .

fied equation for the conservation of transverse· momentum. . The equation -of . . - . . ·.

state for water used in LYNX! i.~ based on.the 1967,ASME steam-water property

. (3) correlations •.

LYNXl uses the forward finite difference method to solve· a boundary

value'problem based on an input core inlet c~olarit velocity- profile. anc;I an . . ·· .. •/'.,·._,· .... _

- input core. exit pressure profile •. At points between the inlet and exit· boun-

. dary coridit_io_ns-,. coolant p_roperti~s are based on the_ conservation_ equations . . . . '

for mass, energy, and momentum and the equation of state for water. The·

LYNX! computer code use~·. empirical correiations for determining. clad surface·

temperature, coolant void·fraction, flow regime, frictional pressure drop, and

two-phase multipliers for frictional and form loss pressure drops. These cal-

. culations- are discussed in detail.. in References 1 through 5.

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The LYNX! code also- calculates interassembly diversion crossflows-

at- each axial. level. . These. cross flows are then used as_ boundary conditions.

for the LYNX2 .computer code·. The primary use for LYNX! is the calculation

.. of· these. interassembly_ crossflows; since the core minimum DNBR must be. cal

culated· on a subchannel basis_by LYNX2.

The interbundle diversion crossflow boundary conditions·calculated

by LYNX!, as wen· as the conservation equations. for mass; energy', and m6'(11entum,_

·and the equation a'£ state for.water,' are used by LYNX2 to determine the coolant

· conditions- in, e_ach subcha_nnel in the fuel assembly which contains. the minimum·

DNBR subchannel. LYNX2 uses a backward finite difference method to solve an .

initial value probi~ .basecLon inlet flow, pressur_e and entl:talpy. ·. LYNX2 - .

iterates ayer each axial increment until differences in diversion crossflow .

. a:·t every cross flow boundary. satis~y an input convergence _criteria~: . . . ·. . . . . - '

·The LYNX2 co~ptiter_ '.c<><ie. uses. the same· empiricai cor.relatiotis as

_LYNX! for the.clad surface temperature, coolant void fraction,·flow regime, - . . . .

·. frictional pressure· drop, and tw~:-:-phase ni.ultipli~rs for~ frictional and form . ·

loss pressure drops. Once the LYNX2 computer code has solved for the coolant

conditions.in _each subchannel, LYNX2uses this data to determine the critical

heat.flux distribution 'for each fu~l· rod in the assembly. ·rhe critical heat

flux calculatio-r,. is-based on a user selected·critical heat flux correlation.

and the, appropriate non-uniform- heat flux co:rrect,ion factor for the selected

criti.cal heat flux correlation .. Based on.these .;~lues and the actual"heat . . . '

. fluxes, LYNX2 · caiculates the DNB ratio for each fuel · rod at each axial incre-:-

ment' in order to -determine the minimum DNBIL

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SECTION 3. :..,. HYDRAULIC MODEL FOR SURRY UNITS 1 AND 2

3 . 1 .INTRODUCTION

· The hydraulic portion· of the thermal and hydraulic model which has. .. .·

. been formulated at Vepc() for analysis of the Surry Units. No ... 1 and 2. reactor

cores will be.presented in this section. The hydraulic model describes the

geometry. and physical charac~eristics of the Surry reactor core which are.

important: in d.etermining' the coolant flow through the reactor core. The data

req·uired to develop this· hydraul~c model cons is ts of:

1. Axial locations of the fuel assemblytop·and bottom nozzles' grids,. and. bottom and top. of the rdds.

2. Form loss coefficients .for the top and bottom nozzles and grids·, and surface roughness of the rods

3. Radial flow area asa function, of axial position

This data (or information- necessary to calculate this data) has. been obtained· . . . ·• . . . . . . ~ . ' . - . . .

from ·References .. 6 ·through 10. The reactor .core of Surry Units No. 1 a:nd' 2··

c(lrr.eni:ly consists of. 15 :ic 15 · rod array fuel :asseniblies nianuf~ctur~i by West..;.

inghouse. A. detailed ~escription of· the reacfor core and this . fuel is given ·

in.Reference· 6 .

. 3. 2 Axial Locations and ·Lengths . ' . . . .

A side view of a 15 :ic 15 fuel assembly us.ed. in .the Surry reactor

·cores is provided·in: Figure 3-1. The following dimensions (at cold conditions)

for this fueLassembly were ·obtained from Refere·nces 7 and 8:

1. · Overall as.semhly l~ngth

2. Overall fuel rod length ·

.3. · Active fuel length

· 4. Bottom enci plug length

5. Top end plug and holddown spring length

6. Bottbm nozzle length.

7~ Top nozzle length

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3·.3 Loss Coefficients- and Surface Roughnes,s . .

The effect of the t;op and bottom nozzles on core flow .can be modeled

· using a· form fo~s coefficient and the dimensions from References_ 7 and 8 along . . .

with the_appropri.ite axially depertdent radiaL flow areas. However, ail alter-

"native procedure which: simplifies the hydraulic model can be employed. .In ' - - , . . ., ., . . - . .·

this_ procedure, a constant radial flow area is useci o_ver the entire assembly

length, and the form loss coefficients are increased to yield pressure drops

in the region of the top and bottom nozzles equivalent to those·calculated

with the flow area variations. This alternative procedure has been used in

the Vepc~ hydraulic model of the St1:rry 15 :ic 15 .fuelassembly. · Form loss coef

ficients· are also_ used to repr~sent the effect of tq.e assembly grids. _

The form -losscoefficients·for the top.and bottom nozzles; and

grids (both ·with and w~thout flow mixing vanes) were. developed fr_om pressure

d'rop: d~ta for top· and bottom nozzles :and grids given in References .9 . ~nd iO.

The .form 'loss coeffici.ents were then calculated using· the following :equati,on: (H)

K = · LlP(2p~c) -. G .

· where K = form loss coefficient· -

· Llp _ = pressure drop across nozzle -or grid

p = water density

= 32 ~ 174 lbm--ft 2 lbf-,.sec

G·= mass. flux

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The above equation is appropriate for use, since ·the hydraulic model,does

'not .represent .flow area-variations in the top and bottom nozzle areas. . . . . . - ' .

In addition to modeling fo.rm _loss coeffi~ients - in LYNX! and LYNX2 ,.

the frictional pressure drop across ·the fuel assembly is model_ed. The sur

facE?_ roughness of the fuel rod~ is. required for calculation.of the frictional

pressure· drop._- Refe~ence 9 provides a best_ estimate frictional pressure drop

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which-: can be, used to, calculate an ass'embly average frictional factor. fo,r the

d b d h f 1, 1,' ., ' ' . (11)

ro s ase on t e o ~wing equation~

f = ap(D~) (2 ·gcP) . L(G2 )

' where f = friction factor

ap -- frictional pressure ·drop

D = equivalent diameter e

-p = water density

g· = 32.i74 ·1bm - ft c-lbf - sec;

L = 'fuel· _rod length_ ·

G =.mass flux

2

.Toe. friction factor calculated with.the above formula is used along.w:i.th. the·

applicable Reynolds number and Moody. friction factor chart from Reference 'i1

- to determine the surface roughness.

3..4 .Radial. Fl~w Are~~-, . .,

As-discussed,in the pl'.'evious section, the radial flow-areas repre-

_sented in the hydr~ulic model are the· flow areas in the rodd-ed (Le., fuel

rodded) region of the fuel assembly. _ Figure 3-2 is a cross _sectional view of

the 'rodded region of four Surry fuel assemblies. A core wide calculation -

_reer.~aezj.t~ng ~ach f~eJ :as'semb'ly; as 'a sit?-gle flow' channel is 'performed with

'LYNXl. . For. these LYNX! calc~lations, the assembly average flow area, wetted· -

-. perimeter and heated perimeter are required. These parameters are pr~vided·

in Table 3-1 for_the-Surry core for cold conditions.

The LYNX2 code is then used fora subchannel analysis_of the hot . ' . . '

assembly based on cross flow boundary conditions_· calculated .by LYNX!. Cross-. . . .

·, s~ctionai vie,ws' of a unit cell subchannel and a thimble cell sub~hannel u~ed

_in the hydraulic model are given in Figures 3:-3 and 3-4. As- can be- seen from -

the figures, a unit cell is a flow region bounded by four fuel rods, -and a

thimble cell is a flow region bounded by three fuel rods and one guide tube.·

3-3


The values of the flow area, wetted perimeter, and heated perimeter for the

unit and thimble cells are also given in Table 3-1 for the Surry cores.

3-4

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TABLE 3-1

HYDRAULIC .. MODEL PARAMETER VALUES. FOR· SURRY FUEL ASSEMBLY*

Pe:ta:ineter ; ···. . ·. .

· Fuel Rod Diameter. (inches)

. Guide· Tubes/Instrumentation · · .. Thimble Diameter ·(inches)\ .

Rod• Pitch (inches) ·.

Number. Fuel Rods/Assembly·.

Number Guide Thimbles/Assembly

·Number Instrume~tation.Thimbles/Assembly

. Assembly Flow Area (square inche.s)

Assembly Wetted. Perimeter (inches)

· Assembly Heated Pe~iuieter Cinches).

Unit Cell Flow Ar.ea (.square: inches).·.

• Un:i.t Cell W~tted/He~ted .Peri~etei: '(inche~) . . .

Thimble .Cell Fl_ow Area ·(squ~r~ inches) .

Thimble Cell Heated.Perimeter (inches)

Thimble Cell Wetted Pez:imeter. (inches) ·

*Dimensions obtained from Reference 6 for 70 ·-~.

3-5

Value

0.422

.o .546

0.563

204

20

1

38.·22

306.47

0 ~ 1771

.·1.3258.

0 ~1535

0.9943

· I .4231.

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·. ·· . . · --~., Ii ii 1111 D U ffl R DU II· 111111 . , .- . I . . ..

r.- c-: '' , .

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3-6.

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FIGURE 3-2 -

CROSS SECTIONAL VIEW OF SURRY FUEL ASSEMBLIES

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1iO ~ ...... ~--- "1 0, 0- 0 0 @ . Q () 0 ·o Q --· ~ I 0 -·-~·~-•-•e, -:- • .~, - 0 0 o1 ~'!!"S?."' cO-TltOc l")D

o 0 : , C) o o o ·r'""" .. , ... o. -·"o Oii 0 ® .. © 0 O O I O o\ 0 © o.o O 01

'8 ~ .... ~ .j .. ~ '""~ 8f 8 0 0 ! 0 0 8! OQQOOOOCDOOOOqoo OOOOOOOCDOOOOOO i

...... ,:,:·., ..... , .. ,.,i,: I ·--- - -- Jl£t.:;-;1 .. e,.o...:..... _ __J 'ult. •oo cc_ •~2 r- ·

1 (lAO "'•oc.•wr,1._ 014!. · ~

~~~ :~~:~;~ ... ::_~·~;~:_ :·'-· ~···: . . /:-~c·-~:\; :,;.~··. -WOT':': .... LL Qi,,~ _(OflllUCTl D "TO &.&"'Ff: z.•

3-7

•1.,

.I I

· 1.··· ..

I I ·1 1-. . I I I I I ·I ·I I I I·.-·

I

UNIT CELL SUBC!-IANNEL·

t 0.422"·,(F'UEL ROD) .

j_'

r-.0.563" ~

(ROD PITCH)

1·

·I ·.I I I

·1 I I I·-

_· I--· . ,. .--1---·

I .1,/·-.:-··

:I 1·

I .I 1·

THIMBLE .CELL SUBCHANNEL

-~ 0.563" . -, .

("ROD PITCH)·

FIGURE 3-4

3-9

T -· .

_ 0~ 546•. (GQlrit T~IMBL£} . . -

i-

. T . . 0.422" (FUEL ROD)

··i

·1 ;· .

·I I

-I

1. -1·

I I I I I I I I

-1

. SECTI_ON 4'. - THERMAL MODEL FOR SURRY UNITS 1 AND· 2

4. 1 INTRODUCTION .

· The thermal portion of the thermal and hydraulic models which has

been formulated at Vepco for. the. Surry ·units No~ rand 2 re.actor .cores will:

be presented in this.section~ A thermai model consists of the description . . . . . . .

of the reactor system par~eters, core p~wer distribution, c~olant conditio~s . .

at the core inlet,.arid thermal hydraulic analysis uncertainty-factors.- The

· thermal _modeT, along with th_e hydraulic model. described .in Sectio~ 3,. is· then

. used with. the· LYNXi. and ~YNX2. computer codes to solve· for. local c_oolant prdperti~s, .··.

critical. heat; flux, .·and- DNB ratio ...

·. Since the. hydz:aul:i.c model describes physical charapteristics of. the

Surry _reactor cores, values used:for the parameters in· the hydraulic model are

generally affecte~. only by significant :chan~~s :i.n th'e · mechanical design: of the ·

· fueL.. Ho~ever, the values. of pa;ani~i:e~s: ·U:Sed in th~ thermal tnodei are. affected . .

by changes in· power :~n:d· flow: distributions' power peaking 'factors, and reactor .

system -parameters; · . Therefore, the values·. of some parameters requ'ired · tb spedfy

tl:lethermal ~odei may change for each thermal hydraulic analysis .performed ... The. . . . .

parameters required tci:_specify' the thermal ·model will be described' and then the \.

·values of ;hese, para.meters which were used' iri 'establishi~g several thermal hyd'ra~...;

'lie· a~alysis _cases_will be· given. The cases which ha.ve been selected-include

.the thermal hydraulic analysis origfoaily used to license the Surry cores.as well

as sub'sequent updates.

4.2 System Parameters.

The system parameters which must be specified for the thermal model

consist of-:

l) ·Thermal power level

2) Direct moderator heating fraction

3) Coolant.inlet erithalpy

4) ~ystem pressure

I . . I·. ·1: . .

I I 1·

.'1. . .

I I I I

-I I I I

5) Total· system flow rate

6). Fraction, of system fiow which. bypp.sses the reactor core.

4.3 Reactor Core Power and Inlet Flow Distributions

:The. core power distribution and c'ore inlet. flow distribution must

be· specified in ·the foihiulation of the thermal model. Th.e assembly relativ~.

inlet fiow distribution which.has been used in all cases is given in Figure 4-1.

The 5% reduction in relativ~ inl~t flow· shown in Figure 4-1 for the highest

power assembly has been used for all licensing calculations fo·r Surry; This

· 5%, flo:w reduction was· coµserva'tively distributed to. the peripheral fuel, ass em~

blies.

Both the axial and· radial core power distr.ibution must. be spe.cified

in the thermal model.. The average relative .radial power .must b~ specified for. . .

each.assembly.in the,reactor core. For the highest po:wer assembly; the relatiye

powe~_ for each fuel pin in. the ass~mbly must als·o,he specified .• The p·ower

··· ... 'distributions. used are- conservative thermal hydraul.ic' design power shapes~

· 4.4 Uncert~inty:-Fact·ors

· Since ·nuclear fuel pellets, rods, and assemblies can only be :fab:ricated

• -to within a· certain tolerance about: the spe·~ific design criteria~ uncertainty fac,-.

tors must be .used on. the thermal model to account for. limiting actual conditions~

The uncertainty factors used in the thermal model .are d~lineated below and ensure

a conservative representatibn of the actual fuel condition.

1) A re.duced fuel pin pitch for the highest power, subchannel

is-applied to account for the.effect onDNBR of variations

from nominal channel dimensions. of the manufactured.fuel

assembly~

E .2)- . An engineering factor . (FtiH) due to enr.ichment · and density·

4-2

1.· 1·

I . I I I·

I I I

. . I.

I ·1 .I

. . .

variations between fuel rods· is applied to increase the ..

relative·power·of the highest power rod.'

3) An engineering factor (F~) due to diameter~ density and

enrichment variations.between pellets and ctad, eccentricity

· ·variations is· appli'~d. as a power spike on· the highest· power

r.od at the axial h,cation of uiiniint.im DNBR. . . : . . .

4), A·redU:ction in inlet flow to the highest power assembly.is

. applied to account for non-uniform inlet flow. d'ist~ibutions. . .

(See Figur~ 4-1) .

· Other factors. have_ been applied to the thermal model to accommodate

fuel·dens1.ficationor_fuel rod-bowing.· These.factors will be discussed as . . . .

. appropirat:e in the cases presented in the following sections. ·Nuclear uncer-. .

·. tainty factors on the calculated relative power distributioQs are already•-

included in the<pow~r .distributi~n used in the thermal model for each case.

. 4 .. 5· Surry FSAR. and Cycle L.A0:alys~s· -· Ca~ea Al and A2

· A miniril~ DNBR (MDNBR) · for a unit cell subcha~nel is documented .in

· the FSAR (6)~ for .. a ~he~al hydraulic analysis ~'t the nom.inal core power of·

2441. Mwt .. · A. thermal model case, · denoted .as. Case Al, has, been ·developed, for

input into the Vepco Core Thermal-Hydraulic Model in order to compare ·to this

· ·. FSAR analys_is .. · The system .parameters used, based on Reference 6, are. given in

. Table 4-1~ and the assembly average power distribution used is· given in Figure.

4-2 .. The power ,.level of the highest power assembly is given· in Reference 12.

.· The -local power distribution used in the highest power· assembly is given in

Figure 4-3~- The highest power channei for this case is a unit. cell subchannel . .

which .is located in the ce~ter fuel assembly of the core.·. The peak relative· . . . .

radial p~wer,. includ~s a ·ri.ucieai- uncertainty of 10% oil the power distribution,

and the axiai power· distribu·tion used is a i, 72 cosine: shape. 0 3) , A reduc

tion in the critical heat flux calculated with the W-3 correlation is used in·

I I 1-·1· I

I 1·-

1 -I I I I

I I

Case Al .. This i::eduction was applied in. the· o:dgina:l FSAR analysis because -only

· limited _DNB test dat~ at high pressure .was- available a_t that time. (It should ·

also be noted that the _critical heat flux correlation·used.in the·dev'elopment_

of Case Al did not include :a grid spacer factor. The spacer· factors. for the . -

W-3 correlation wer.e developed· after the completion of· Reference 6). ·

A second. FSAR analysis, denoted as Case A2 ,: ,has been develop~d for- a

· point··on: the reactor core .thermal' and hydraulic safety cu;rves which were estab

lished. for: the first cycle of Surry Unit °I. The selected point on the thermal._

limit- curves· from-Se~tion 2.1 of Reference 14 was calculated to result in a. ' . .

minimwri: DNBR-slightly above 1.30. (The development and purpose of the thermal

limit curves are _des~ribed · in:.Reference 13 and Section :2 .·t of Reference 14).. ' . .·· . . -._ . . ' .

. The th~rmal ~ode1 parameter v:atueS tised for Case A2. are the ·same as Case Al ·

except ·for' the' system parameters as indicated in Tab~e·. 4_:z and the_ local _power . . . .. . . '. ..

-.distribution. used in the: highest p~we~ assembly as indicated in Figur~- 4.:.::3·.

Th_e highest po~er channel- fo_r. Case A2 is a thimble cell s~bchannel which ·is ' . .

located- in the center fu~l as~enibly _of th_e core. ' . . . . .

4.6 SurryD~nsification Analysis·- Case B

As a result· _of t;he. fuel densification .phenomenon, the values of a

.number of pir~ameters :i~ the thermal model used in Re_fere.nce 6 for the Surry cores .

were modified~ -- . The fuel ·ciensificat~on phenomenon, which caused gaps to form in . . . . .

the fuel stack and a reduction in the fuel stack, height, impacted the thermal . . •. .

model in two w~ys. First, gaps in the fuel stacR: resulted in local_ po,~e_:i:. spi~es1:_:,:})~',i!:

in. adjacent fuel .rods (due to' the increased local fuel to moderator r~tio), and _.

. sec~ndli,' the reduced. fuel st~ck height resulted in-. an. fa1creased linear_ power _ ..

densi_ty. - Ananalysis~ ·_-which conservatively ~ccommod;ted·_ the impact ·of ·fu~l

densification, for the initial core: Surry fuel is do'cumen~ed in Reference- 13.

The Case B -thermal model parameter values which will be described in this sec-. .

tion are for an c1nalysis of a·DNBR limiting·point on the thermal limit curves

which were developed-from the thermal model documented in Reference 13-and pro-

·4-4 -

1 · · . vided in Reiference 16. The point on' the reactor core· thermal- and hydra~lic

1-.,

-I I I

I I 1. I

I

I I I

· safety limit curves chos,en for· analysis is a point wh.ich was -calculated to have

a minimum _DNBR ~ligh~ly aboye: 1.30.

The system parameters .used for Case B are provided in Table 4~3, (l3 ) .··

and the'assemb1y:pow~r distributi~n used is -given:i~ Figur~ 4-5.· The power

level.· of· the. highest power asse~bly is given .in Reference 12. The local power

· distribution used in the highest power .assembly is given in Figure 4-6 .. · The

relative power of. the hi~hest power fuel. rods is .based on Reference._ 13. · The

highest power £µel.rods are .located aro~nd a thimbleceil subchannei in·the

center fu~l>assembly of the core._ The axial power distribution· for Case B

. 1 55· h. . d. - . . . . - h - . ( 13 ). . 1s a . c. oppe · cosine s ape~. · . ·

., ·, -

The. changes in the power distribution used in the thermal model for-. -

Surry between ~a:ses Al ~n:dAz':and Case B.resulted: froin ~perating. data and.design.

~xperience ·de~elop~d .·betwee'n th~ time the' analyses d6cumented·.in Refer~~ces 6 ·' . . .

and u we:re performecL: This ~xperience supported the r·educl:ioU in the. peak- ·

radial and .. axia:l powers used between.Cases Al and A2.and Case B .. Additional

, fact:ors imp~sed on Case B to accominodate the effect .of fuel densification

-. .1 . (13) inc uded.: : .··

1) A power spike to conservatively acconunodate the gaps·in

. the relatively. low density· initial· core Surry fuel.. This

factor was multiplied by the F~ factor and the product was.

supplied as a power spike on the.highest p~wer rod :at.the

·. · axial location of minimum· DNBR.

· - 2) · A reduction in the fuel stack height from the. nominal 144

. inches J and· a·. corresponding. increase in linear power density. .

. For Case B; .the W--3 correlation was augumented to reflect the use. of

t;he L-grid factor. (with a TDC of 0.019) .. -This applica_tion was c.onsistent with the

analysis documented in Reference 13.

4-5·

I I. I I I I

I

I I I I I

I I

4. 7 . Surry . 90· Percent ,·Flow ·Analysis ""'.·Case . C

Subsequent to the Surry fuel densification thermal analysis, sever.al

. events. have. ocurred which . again changed _the. value.s of . some of the parameters in . . .

. the. thermal- modei. These c:hanges ·resulted from: ·

1) Test data which:demonstrated that.large local power spikes

do not result i~ .·a reduction in critical heat- flux ..

2) Plugging of steam ge_nerator t~bes a,t Surry I1nits 1 ~nd 2.

3). Fuel: ex,amination data which indica.ted the _observance of· fuel

rod bow. in excess of that• accommodated by tli(;!''_uncertainty .•.

.factor· for reduced fuel pin pitch.

A thermal.hydraulic analysis which reflected·these changes.was per-:

formed and is documented' in Reference 15. , The Case. C thermal model w~s deve·l~ped

from Reference 15.

The system· parameters used. f<>r Case· C -are given in .'J;able 4'."'4 •. · As. ' . ,-

, indica·t~d iin Tabr~ 4-4,. the core :a~er·~ge. now rate used··i'n'this ~nalysis is_.·

90% of. the value used iri Case:A and Case B; This reduced flow rate was con'."'

servatively selected to-represent the effect on .flowfroin steam generator tube

_ plugging._ : . . . . '

The ai1sembly: average. power distribution used for· Case C 1.s given in

E:igure 4-7 ~·. The .relative power. of the highest power· assembly is based on Refer

ence 12 .. The local power distribution used·in the·h1ghest power assembly· is given

in Figure 4-8 .· The highes·t power fuel rods . are i~cated aro~~d a thimble. cell sub-..__::;;:;_:·,_ - ' .. / --,':"~:;-:._':.:·.

channel .for the c~nter fuel assembly in the core a~d> the. reiative radial power. of . . . - .

these rods is. 1.55. · Based on the- power spike critical heat fltix tests:mentioned ·

above, the power spike penalty us.ed in the Case B thermal model was eliminated

in the Case C thermal model. Howeve.r, a reduction of approximately )% was applied

to the DNBR.valuecalculated in Case C in order to make it consistent with the

4-6

I , .. ·. I .I. I I I I I ., I I I I I I .1 1:

·1

analysis reported in. Reference 15; . {In the·. Reference ls analysis:; ·a, generic.·

reduction.in DNBR of approximately 7% due to fuel. densification was maintained.

and used to·offset-the impact of rod bow which is. riot explicitly represented

in the.analysis).

4-7

I I I I I

I I I I I I

I I I I I I

TABLE 4-1

CASE Al-REACTOR CONDITIONS

Thermal Power Level (% of nominal 2441 Mwt)

Average Fael Rod Surface Heat Flux (106 BTU/hr-ft2)

Fraction of Heat Generated in Fuel

Core Inlet Enthalpy (BTU/lbm)

Core Inlet Temperature (°F)

System Pressure (psia)

Total System Flow Rate (gpm)

Reactor Core Bypass Fraction

Average Core Mass Velocity(l06 lbm/hr-ft2)

4-8

100

0.1911

0.974

538.6

543

2250

265,500

0.045

2.308

I -1

I I I I I I I' I I I I I I I I I I

TABLE 4-2

CASE A2-REACTOR CONDITIONS


Average Fuel Rod Surface Heat Flux · (106 BTU/hr-ft2)


Core Inlet Enthalpy (BTU/lbm}


System Pressure (psia}


Total System Flow Rate (106 lbm/hr)


Average Core Mass Velocity (106 lbm/hr-ft2)

4-9.

112

0. 2140

0.974

548.6

551

2200

265,500

99.6

0.045

2.283


TABLE 4-3

CASE B-REACTOR CONDITIONS


Average Fuel Rod Surface Heat Flux (106 BTU/hr-ft2 )



Core Inlet Temperature (OF)






4-10

112

0.2177

0.974

552.4

554

2200

265,500

99.2

0.045

2.273


TABLE 4-4

CASE C-REACTOR CONDITIONS


Ave6age Fuel R~d Surface Heat Flux (10 BTU/hr-ft )









4-11

102

0.1955

0.974

543.6

547

2220

2.38,950

90.2

0.045

2.066


0.95

1.000

1.000

1.000

1.000

1.000

1.000

1.001

FIGURE 4-1

RELATIVE CORE INLET FLOW DISTRIBUTION FOR ALL CASES

1.000 1.000 1.000 1.000 1.000 1.000 1.001

1.000 1.000 1.000 1.000 1.000 1.001 1.001

1. 000 1.000 1.000 1.000 1.000 LOOI

1.000 1.000 1.000 1.000 1.000 1.001

1.000 1.000 1.000 1.001 1.001

1.000 1.000 1.000 1.001

1.001 1.001 1.001 ,.

1.001

4-12


FIGURE 4-2

RELATIVE ASSEMBLY POWER DISTRIBUTION FOR CASES Al AND A2

1.432 1.432 1.400 1. 300 1.200 0.800 0.700

1.432 1.432 1.400 1.300 1.100 0.900 0.700

1.400 1.400 1.200 1.300 1.200 0.800 0.700

1. 300 1. 300' 1.300 1.200 1.000 0.800 0.764

1.200 1.100 1.200 1.000 1.000 0.800

0.800 0.900 0.800 0.800 0.800

0.700 0.700 0.700 0.764

0.500 0.600

4-13

0.500

0.600


1. 36

1.46

1.50

Thimble Tube

1.56

1.54

' 1.50

Inst Tube

FIGURE 4-3

RELATIVE PIN POWE.R DISTRIBUTION FOR CA!;iE Al

1. 30 1. 32 1. 28 1. 28 1.276 1.26 1.26

..

1.40 1.42 1.40 1.36 1. 35 1. 32

1.50 Thimble 1.46 1.42 Thimble Tube Tube

1.58 1.58 1.56 1.54

Thimble 1.58 1.58 Tube

1.56 1.56

1.50

Center of Assembly

£,-14


1.36

1.46

1.50

Thimble Tube

1.56

1.54

1.50

Inst Tube

FIGURE 4-4

RELATIVE PIN POWER DISTRIBUTION FOR CASE A2

1. 30 1. 32 1.28 1.28 1.276 1.26 1.26

1.40 1.42 1.40 1. 36 1.35 1.32

Thimble Thimble 1.50 Tube 1.48 1.42 Tube

1.56 1.58 1.58 1.54

Thimble 1..56 1.58 Tube

1.56 1.56

1.50

Center of Assembly

4-15

I I 1·

I I I I I I 1. I I I I I I I I I

1.476

1.476

1.400

1. 300

1.200

0.829

0.700

0.500

;FIGURE 4-5

RELATIVE ASSEMBLY POWER DISTRIBUTION FOR CASE B

1.476 1.400 1. 300 1.200 0.829 0.700 0.500

1.476 1.400 1.300 1.100 0.900 0.700 0.600

1.400 1.200 1. 300 1.200 0.800 0.700

1.300 1.300 1.200 1.000 0.800 0.700

1.100 1.200 1.000 1.000 0.800

0.900 0.800 0.800 0.800

0.700 0.700 0.700

0.600

4-16

I I I I I 1· I I I I I I I I I I I I I.

1.450

1.480

1.530

Thimble Tube

1.530

1.520

1.500

Inst Tube

FIGURE 4-6

RELATIVE PIN POWER DISTRIBUTION FOR CASE B

1.420 1.420 1.410 1.390 1.390 1.390 1.386

1.460 1.460 1.460 1.440 1.420 1.400


1.540 1.550 1.550 1.540


1.520 1.520

1.500

Center of Assembly

4-17

I I I I I I I I

I I I I I I I I

1.475

1.475

1.400

1. 300

1.200

0.831

0.700

0.500

FIGURE 4-7

RELATIVE ASSEMBLY POWER DISTRIBUTION FOR CASE C

1.475 1.400 1. 300 1.200 0.831 0.700 0.500

1.475 1.400 1. 300 1.100 0.900 0.700 0.600

1.400 1.200 1.300 1.200 0.800 0.700

. . I

1. 300 1. 300 1.200 1.000 0.800 0.700

1.100 1.200 1.000 1.000 0.800

0.900 0.800 0.800 0.800

0.700 0.700 0.700

-

0.600

4-18


1.445

1.480

1.530

Thimble Tube

1.530

1.520

1.500

Inst Tube

FIGURE 4-8

RELATIVE PIN POWER DISTRIBUTION FOR CASE C

1.420 1.420 1.400· 1.400 1. 390 1. 380 1.380

1.460 1.460 1.460 1.440 1.420 1.400


1.540 1.550 1.550 1.540

,


1.520 1.520

1.500

Center of Assembly

4-19

. I.

I I I I I I I I I I I I I I I I I I

SECTION 5 - RESULTS AND COMPARISONS

5.1 Introduction

The thermal hydraulic modeling described in Sections 3 and 4 has been

used with the LYNXl and LYNX2 computer codes to perform thermal hydraulic

calculations for Surry Units No. 1 and 2. Results of these analyses will be

presented in this section and the minimum DNBR calculated in these analyses will

be compared to the minimum DNBR obtained from calculations used in the design

and licensing of the Surry cores.

5.2 Surry FSAR Comparison - Case Al

As noted in Section 4-5, the Surry Unit 1 and 2 FSAR presents a thermal

hydraulic analysis for a unit cell at the nominal core power of 2441 Mwt. The

Case Al thermal model presented in Section 4-5 and the hydraulic model presented

in Section 3 have been used to perform a thermal hydraulic analysis with the

LYNXl and LYNX2 computer codes.

The key LYNX! results for the highest power assembly are presented in

Tables 5-1 and 5-2 and Figures 5-1 and 5-2. Core average exit conditions are

giv~n in Table 5-1, and conditions at the exit of the highest power assembly

are given in Table 5-2. Plots of the mass velocity and enthalpy in the highest

power assembly are given in Figures 5-1 and 5-2, respectively. The key LYNX2

results for the highest power assembly are presented in Tables 5-3 and 5-4, and

Figures 5-3 through 5-5. Exit conditions for the highest power subchannel are

given in Table 5-3, and conditions at the point of minimum DNBR are given in

Table 5-4. Plots of the mass velocity, enthalpy, and DNBR for the highest

power subchannel are given in Figures 5-3 through 5-5, respectively.

The minimum DNBR calculated by LYNX2 for this analysis was 1.94. The

minimum DNBR reported in the Surry Unit 1 and 2 FSAR under these same conditions

is 1.97 which indicates very good agreement between these two calculations.

5-1


5.3 Surry Unit 1, Cycle 1 Comparison - Case A2

The Case A2 thermal model presented in Section 4.5 and the hydraulic

model presented in Section 3 have been used to perform a thermal hydraulic ana

lysis with the LYNX! and LYNX2 computer codes for a DNBR limited point on the

thermal limit curves for Cycle 1 of Surry Unit No. 1.

The key LYNXl results for the highest power assembly are presented

in Tables 5-5 and 5-6 and Figures 5-6 through 5-8. Core average exit conditions

are given in Table 5-5, and conditions at the exit of the highest power assembly

are given in Table 5-6. Plots of the mass velocity, enthalpy, and void fraction

in the highest power assembly are given in Figures 5-6 through 5-8, respectively.

The key LYNX2 results for the highest power assembly are presented in Tables

5-7 and 5-8, and Figures 5-9 through 5-12. Exit conditions for the highest

power unit cell are given in the Table 5-7, and conditions at the point of mini

mum DNBR are given in Table 5-8. Plots of the mass velocity, enthalpy, void

fraction, and DNBR for the highest power unit cell are given in Figures 5-9

through 5-12, respectively.

The minimum DNBR calculated by LYNX2 for this analysis was 1.29. The

point chosen from the thermal limit curves is one which would be determined to

have a minimum DNBR of 1.30. This indicates very good agreement between LYNX

and initial Surry core licensing calculations documented in Reference 14.

5.4 Surry Densification Comparison - Case B

As discussed in Section 4.6, the effect of fuel densification on

the Surry reactor cores was 7valuated and reported in Reference 13. The Case

B thermal model presented in Section 4.6 and the hydraulic model presented in

Section 3 have been used to perform a thermal hydraulic analysis with the

LYNXl and LYNX2 computer codes.

The key LYNX! results for the highest power assembly are presented

in Tables 5-9 and 5-10 and Figures 5-13 through 5-16. Core average exit "

5-2


conditions are given in Table 5-9, and conditions at the exit of the high-

est power assembly are given in Tabl~ 5-10. Plots of the mass velocity, enthalpy, '-

and void fraction in the highest power assembly are given in Figures 5-13 through

5-16, respectively. The key LYNX2 results for the highest power assembly are

presented in Tables 5-11 and 5-12, arid Figures 5-16 through 5-19. Exit conditions

for the highest power unit cell are given in the Table 5-11, and conditions at -

the point of minimum DNBR are given in Table 5-12. Plots of the mass velocity,

enthalpy, void fraction, and DNBR for the highest power unit cell .are provided

in Figures 5~16through 5-9, respectively.

The minimum DNBR calculated by LYNX2 for this analysis was 1.31.

The point chosen from the thermal limit curves is one which would be determined

to have a minimum DNBR of 1.30. This indicates very good agreement between

LYNX and licensing results provided in Reference 13.

5.5 Surry 90 Percent Flow Comparison - Case C

As discussed in Section 4.7, the impact of a reduction in core flow

caused by steam generator tube plugging at Surry was calculated and reported

in Reference 15. The Case C thermal model presented in Section 4.7 and the

hydraulic model presented in Section 3 have been used to perform a thermal

hydraulic analysis with the LYNX! and LYNX2 computer codes.

The key LYNX! results for the highest power assembly are presented

in Tables 5-13 and 5-14 and Figures 5-20 and 5-21. Core average exit condi

tions are given in Table 5-13 and conditions at the exit of the highest power

assembly are given in Table 5-14. Plots of the mass velocity and enthalpy in

the highest power assembly are given in Figures 5-20 and 5-21, respectively.

The LYNX2 results for the highest power assembly are presented in Table 5-15

and 5-16, and Figures 5-23 through 5-26. Exit conditions for the highest power

subchannel are given in Table 5-15 and conditions at the point of minimum DNBR

5-3


are given in Table 5-16. Plots of the mass velocity, enthalpy, and DNBR for

the highest power subchannel are given in Figures 5-22 thr,ough 5-24, respectively.

The minimum DNBR calculated by LYNX2 for this analysis was 1.51.

The minimum DNBR given in Reference 15 for these conditions was 1.50 which

indicates very good agreement between these two calculations.

5-4

I I I I ,I I I I I I I I I I I I I I I

TABLE 5-1

CASE Al - CORE AVERAGE EXIT CONDITIONS

Exit Enthalpy (BTU/lbm) 625.4

Exit Temperature (OF) 607.5

Exit Void Fraction 0.00

Mass Velocity (106 lbm/hr-ft2) 2.31

Pressure Drop (psi) 17.5

TABLE 5-2

CASE Al - HIGHEST POWER ASSEMBLY AVERAGE EXIT CONDITIONS

Enthalpy (BTU/lbm)

Quality

Void Fraction

Temperature (°F)

5-5

663.3

-0.091

0.001

632.0

\

I I I I I I I I I I I I I I I I 1·

I I

TABLE 5-3

CASE Al - HIGHEST PQWER SUBCF.ANNEL. EXIT CONDITIO~S

Enthalpy (BTU/lbm)

Quality

Void Fraction

TABLE 5-4

675.4

-0.062

0.00

CASE Al - CONDITIONS AT LOCATION OF MINIMUM DNBR

Minimum DNBR

Location (inches)

Enthalpy (BTU/lbm)

Quality

Void Fraction

Heat Flux (106 BTU/hr-ft2)

Mass Velocity (10 6 lbm/~- ft2)

5-6

1.94

96

643.0

-0.140

0.003

0.4564

2.209

\

I I I I I I .I I I I I I I I I I 1·

I· I

\

TABLE 5-5

CASE A2 - CORE AVERAGE EXIT CONDITIONS

Exit Enthalpy (BTU/lbm)

Exit Temperature (OF)

Exit Void Fraction

Mass Velocity (106 1bm/hr-ft2)

Pressure Drop (psi)

TABLE 5-6

647.13

621.98

0.002

2.28

17.52

CASE A2 - HIGHEST POWER ASSEMBLY AVERAGE EXIT CONDITIONS

Enthalpy (BTU/lbm)

Quality

Void Fraction

Temperature (°F)

5-7

690.78

-0.011

0.028

646.91


--------

\

TABLE 5-7

CASE A2 - HIGHEST POWER SUBCHANNEL EXIT CONDITIONS

Enthalpy (BTU/lb) m

Quality

Void Fraction

TABLE 5-8

704.56

0.021

0.112

CASE A2 - CONDITIONS AT LOCATION OF MINIMUM DNBR

Minimum DNBR

Location (inches)

Enthalpy (BTU/lbm)

Quality

Void Fraction


Mass Velocity (106 lb /hr-ft2)

5-8

1.286

94

662.46

-0.070

0.022

0.5256

1.937


\

TABLE 5-9

CASE B - CORE AVERAGE EXIT CONDITIONS


Exit Temperature (°F) 624. 2 ·


Mass Velocity (106 lbm/hr-ft2) 2.272


TABLE 5-10

CASE B - HIGHEST POWER ASSEMBLY AVERAGE EXIT CONDITIONS

Enthalpy (BTU/lbm)

Quality _

Void Fraction

Temperature (°F)

5-9

702.0

0.015

0.083

649.4

I I I I I I I I I I I I I

: I I I I I I

TABLE 5-11

CASE B - HIGHEST POWER SUBCHANNEL EXIT CONDITIONS

Enthalpy (BTU/lbm)

Quality

Void Fraction

TABLE 5-12

711.8

0.040

0.192

CASE B - CONDITIONS AT LOCATION OF MINIMUM DNBR

Minimum DNBR

Location (inches)

Enthalpy (BTU/lbm)

Quality

Void Fraction


Mass Velocity (106 lbm/hr-ft2)

5-10

1.31

96

669.4

-0~061

-0.025

0.4742

1.912

I I I I I I I I -1 I I I I I I I I I I

TABLE 5-13

CASE C - CORE AVERAGE EXIT CONDITIONS


Exit Temperature (OF) 618.3


Mass Velocity (106 lb /hr-ft2) 2. 06 m


TABLE 5-14

CASE C - HIGHEST POWER ASSEMBLY AVERAGE EXIT CONDITIONS

Enthalpy (BTU/lbm)

Quality

Void Fraction

Temperature (OF)

5-11

690.9

-0.016

0.009

647.1

I.


TABLE 5-15

CASE C - HIGHEST POWER SUBCHANNEL EXIT CONDITIONS

Enthalpy, BTU/lbm

Quality

Void Fraction

TABLE 5-16

7 00.1

0.006

0.078

CASE C - CONDITIONS AT LOCATION OF MINIMUM DNBR

Minimum DNBR

Location (inches)

Enthalpy (BTU/1:bm)

Quality

Void Fraction .


Mass Velocity (106 lb /hr-ft2)

'i-1.2

1.51

94

645.0

-.104

0.006

0.4385

1.824

-------------------

..-..

N I-LL

I a:::: I

' (I) _J :E: .__,

U1 ~ I I-...

u) 1-f

u a _J

w > CJ) CJ)

a: E

LD er) .

0 er) .

LD (\I . (\I

0 (\I . . (\I

LD .... . (\I

0 .... . (\I

ID 0 . (\I

0 0

f

FIGURE 5-1 ttOT ASSEMBLY MASS VELOCITY

CASE Al

"4------.------r-----,------,----.-----.------,-----.-----, ~-00 20.00 40.00 so.oo ao.oo 100.00 120.00 140.00 160.00 100.00

AXIAL POSITION IN INCHES

-------------------0 o· . 0 ('I') ['

0 0 . 0 0 ['

0 0 . 0 I:' (D

,.......,

::E CD _J

' 0 ::)

...,. I-

(D

CD .....,

0 u,

~ 0 I . .... CL 0 ~

_J -cc (D

I I-z 0 w 0 .

0 (X) u,

0 0 . 0 u, u,

0 0 . 0 -N

tna.oo I 20.00

I 40-00

I

r· I GURE 5-2 HOT ASSEMBLY ENTHALPY

CASE Al

I so.oo I 00.00 100.00

I 120.00

AXIAi. POSITION IN INCHES

I 140.00

I 160.00 180-00

-------------------

.,.....,. N I-LL

I 0::: :c

' (D _J

l:: .......,

VI >-I I-,_.

VI 1-f

u 0 _J

w >

en (I)

a: L

tn f') . N

0 er) .. N

UJ N . N

0 N

N

ID -N

0 -N

LI)

b .. N

0 0

FIGURE 5-3 HOT UNIT CELL MASS VELOCITY

CASE Al

"-+-----.----~-------r-----,....-----~----r-------,-------.-------, "tJ . 00 20.00 so.oo 00.00 100.00 120.00 140. 00 LGO .oo 100.00


- - - - - - -- - - - - - - - - - - - -0 FIGURE 5-4 0 . 0

HOT UNIT CELL ENTHALPY (I')

r-- CASE Al

0 0 . 0 0 r--

0 0 . 0 r--(0

,....... :I: CD 0 ..__J 0 . ' 0 ::> ,q-

I-(0

CD .....,

0 u, 0 I >-..... . °' a... 0

_j .... a: CD

I I-z 0 w 0 .

0 a:> lJ)

0 0 . 0 lD ID

0 0 . 0 N

~.oo 20,00 40,00 so.oo ao.oo 100,00 120,00 140 .oo 160.00 180,00


--~·--;:.--.~.,.~-.. ----.•:::·-:-:-:.· .·-:,--.:-.··-~---:--~--~·--;;: . ...Jo~,·--·.·.•.. . ~ a··· ". -· .. ·. . . ·.--------------------

0::: CD

u, z I 0 ......

-..J

0 0 . I.I)

0 I.I)

0 0 . ~

0 I.I) . er,

0 0 . ,.,

0 LI) . N

0 0 . N

0

FIGURE 5-5 HOT UNIT CELL DNBR

CASE Al

LJ?_L __ ...:__.....----.....-----.-----.-----,-----.-------~-:-::-:----:-,:-:-::---:-1 120. 00 14-0. 00 160. 00 180. 00 "o ,00 20,00 40,00 60,00 eo.oo 100,00

qXIAL POSITION IN I~CHES

-------------------

~

C\I I-LL

I ~ :::r:'. ......... en _J

E .._,

V1 >-I I-.....

00 -u 0 _J

w > (J) (J) C( E

l/) (t') .

0 (t')

.• N

IJ)

N . N

0 N ... N

LI) .... . N

0 .... •

N

l/) 0 •

N

0 0

FIGURE 5.6 HOT ASSEMBLY MASS VELOCITY

CASE A2

f' (

"-+''-----~----...-----.------,.....-----.------,-------,.-----~-----, "tJ ,00 20 ,00 40 ,00 60 .oo 80 ,00 100 .oo 120 .oo 140 .oo 160 .oo 180 .oo

/


-------------------0 0 • 0 (T) ['""

0 0 ~

0 0 r"

0 0 • 0 ~ (D

'"'"' ~ (D 0

0 ..._J ~

' 0 ::)

...,. I-

(D

(D ........

V, 0 I >- ~ ......

\0 (L 0

~ ~

<D

I-z 0 w 0

• 0 CD lt)

0 0 • 0 lD 1./)

0 0 • ~ "b . 00 ZO. 0-0 40.00

FIGURE 5.7 HOT RSSEMBL Y ENlHALPY

CRSE R2

60 .oo 00.00 L00.00 LZO .oo

AKIAL POSITION IN fNCHES

140 .oo 160.00 LB0.0-0

-------------------

z 0 1-1

I-u cr et=

\J1 LL I

N D 0 1-1

0 >

co 0 -~ 0

LI)

q 0

'QT

0 ~

0

(\')

0 h

(.)

N 0 . 0

-0 . 0

0 0

FIGURE 5.8 HOT ASSEMBLY VOID FRACTION

CASE A2

·-&-V-~'----W..~-M------41!----M---M--~'""-.....,a;:: __ __,.-------r-------r------.-------r-----, °o .oo 20.00 40.GO so.oo eo.oo 100.00 120.00 140.00 160-00 100.00


- - - - - - -- - - - - - - - - - - - -

~

N I-LL.,

I ct:: I

' CD L...J l:i ........

\JI >-I I-

N 1--'

....... u 0 w w > CJ) CJ)

a: E

. N

0 er> . N

0 N . N

0 .... . N

0 0 . N

0 m . ...

0 (l) . ....

0 r--

FIGURE 5-9 HOT THIMBLE CELL MASS VELOCITY

CASE A2

· ;-----,-----.----.------,----r----~---......----------C) .oo 20.00 40,00 60,00 ao.oo 1,00. 00 LZO. 00 L40 .oo 1,60,00 Lao.oo


- - - - - - ·- - - - - - - - - - - - -0 0

FIGURE 5-10 . 0

HOT THIMBLE CELL ENTHALPY CV) CASE A2 ['

0 0 . 0 0 r.-

0 0 . 0 ['

to

~

:I: (I) 0

_J 0 . "- 0 ::::)

..,. t- co (I) .........

0 V1 >- 0 I .

N CL 0 N _J -a: to

:c I-z 0 w 0

0 (X) LI)

0 0 . 0 LO

.>f--*" I.J)

0 0 . 0 (\I

~.JO 20.00 40.00 60.00 ao.oo 100.00 120.00 140.00 160.00 100.00

A:<IAL POSITION IN INCHES -------

- - - - - - ·- - - - - - - - - - - - -

z 0 1-1

I-u a: 0:: LL

\J1 I

N D w 1-1

0 >

.... (.'J . 0

CD .... . 0

LO .... . 0

(.'J .... . 0

0) 0 . 0

CD 0 . 0

('()

0

0

FIGURE 5-11 HOT THIMBLE CELL VOID FRACTION

CASE A2

0

~~-~~h--*-*---rlf----M---M-rd~~--,,----,-------.------.-------.----, 9J .oo w.oo 40.00 100.00 120.00 140.00 160.00 1,eo.00


- - - - - ·- ·- - - - - - - - - - - - -

Ct: (D

z \J1 0 I N .p.

D (0 . N

D ..,. . N

D N . N

0 D . N

0 ([) . .....

0 (0 . .....

0 -.t .

0 N

FIGURE 5-12 HOT THIMBL~ CELL DNBR

CAS~ A2

·:r:----r-----.-----.----.-------.------y-------,.....-----.-----"""b.oo 2.0. 00 4D .oo 60.00 ao.oo 100.00 12.0.00 140,00 160.00 100.00


- - - - - - ·- - - - - - - - - - - - -

'""' N I-Lu

)

g§

' en L.J l:i .._,

u, S-I I-N

u, .:--. u 0 L--1 w > (/) (f)

a: ~

U) (11 .

0 (I')

~

II) (.\,I . (.\,I

0 (.\,I . (.\,I

U) .4 . cJ

s (.\,I

l/i)

0 . (.'J

0 0

F fGURE 5-1..3 HOT ASSE.HBLY MASS VELOCtTY

CRSE. B

•..J..----~-------.------..-----,---------.-----,------r-----, ~.oo :.,o.oo 40.00 60 .-00 ao.oo LOO .. QO iw.ao 140.00 160.00 100.00

AXrAL POSITION IN rNCHES

-------------------0 FIOURE 5-1.4 0 . HOT ASSEMBLY ENTHALPY lJ)

N CASE B ['

0 0 •

0 0 ['

0 0 . lJ) [' (0

,......, :I:: aJ g ..J

' 0

~ lf.)

I-(0

(D

'-'

\J1 0

I >- 0 ~

. Q\ a... lJ)

L..J N

~ (0

~ b w 0 .

0 0 (D

0 0 . lf.) [' lJ)

0 0 •

0 LO

"t.oo 20.00 40.00 so.oo ao.oo 100.00 120.00 140.00 160.00 180.00

AXIAL POSITION IN I NCHE.S

-------------------

z a 1---4

I-u cc 0:: LL

Ul I

N D -.J 1---4

a >

0

C\l .... . 0

0 .... . 0

(X) 0 . 0

co 0 . 0

'V 0 . 0

C\l 0 . 0

FIGURE 5-15 HOT ASSEMBLY VOID FRACTION

CASE B

0

~~:..__~--M--r--~~:-.-*---4,E--~~M:::::~::=;=-------,-------,-------r------r------, 9J ,00 20.00 40.00 60.00 ao.oo 100,00 120.00 140.00 160.00 180.00


-------------------

,....,. N I-LLJ

I 0::: ::r: ' en ...J E .....,

\J1 >-I I-N

00 1-1

u 0 ....J w > (/) U)

er E

0 ...,. . N

0 (T)

-N

0 N . N

0 -. N

0 0 . N

0 0) . -0 co . -0

FI OURE 5-1-6 HOT THIMBLE CELL MRSS VELOCITY

CASE B

r:..J_ ___ --r------,-----.----.-----.-----.,------,------.-----, """b.oo 20.00 40.00 so.oo eo.oo 100.00 120.00 140.00 160,00 100.00

RXIR.L POSITION IN INCHES

-------------------0 FIGURE 5-17 0 . HOT THIMBLE CELL ENTHALPY ll) N . CASE B ['

0 0 . 0 0 ['

0 0 . I.I) ['-(0

'""' E (I) (.)

-1 0 . ' 0 =, U)

I-(0

(I) "'-J

0 - Vl >- 0

I . t,.J o._ ll) \0

-1 N a: (0

I I-z 0 w 0 .

0 0 (D

0 0

ll) ['

U)

0 0 . 0 U)

U)J. 00 20.00 40.00 60,00 eo.oo 1-00.00 120.00 140,00 160,00 180,00


- - - - - - ·- - - - - - - - - - - - -

z 0 -I-u cc O::'. I.L.,

u, I

0 w 0 -0

>

(X) N . 0

0

0 N . 0

(!) .... •

0

N .... . 0

(X)

0 . 0

"It' 0 .. 0

0

flGUKE 5-18 HOT TH.I MBLE CELL VO ro fRR.CT ION

CASE B

~.l.w--¥------1""-~--¥---.4"-----W--~==:l!f::::::~:::::::;;=: ___ ,--___ ~ ___ --r-""'--------.-------,

20.00 40.00 so.oo 80,00 LOO.DO 120.00 140.00 160.00 180.00

AXIRL POSITION IN INCHES

- - - - - - ·- - - - - - - - - - - - -

0:: OJ z

V, 0 I c..., ......

0 (0

• N

(:) ~ • N

(:) N .

0 0 •

N

0 (X) . ....

(:) . <D • ....

....

(:) N

FIGURE 5-19 HOT TtlH1BLE CELL DNBR

CRSE B

"-+--------------.-----.........-----------.-----------------, 1).oo - 20.00 40-00 so.oo eo.oo 100.00 120.00 140-00 160 .. oo 180.00

AXIAL POSITfON IN INCHES

- - - - - --- - - - --- - - - - - - -

"""' N '-lJ...

) ~ r ......... t'.D L.J E ~

V,. t: I l,,J ...... N u

0 .....I w > (/)

~ E

0 N . N

0 -. N

0 0 . N

0 (J)

• -0 CD . -

.o l' • -

0 (D . -0 II)

FIGURE 5-20 HOT RSSEMBLY MASS VELOCITY

CASE C

. ....._ ___ ......-----..-----.-------,,-..-----,----.------r-------.-------. L).00 20.00 40.00 so.oo ao.oo 100.00 120.00 140.00 160.00 180.00


- - - - - - ·- - - - - - - - - - - - -0 FIGURE 5"""21 0

HOT ASSEMBLY ENTHALPY . 0 (I') CASE C r--

0 0 . 0 0 r--

0 0 . 0 {'

CD

'""" l:: CCI 0

.w 0 . ......... 0

=> -.I'

I- CD

CCI .....,

VI 0 I >- 0

LJ . LJ CL 0

L-1 .... a: CD

I

~ 0 .W 0 .

0 (X) LI)

0 0 . 0 ID LI)

0 0 . 0 C,J

lt.\J.oo 20.00 40.00 so.oo ao.oo 100.00 120.00 140-00 160.00 100.00


- - - - - - ·- - - - - - - - - - - - -

,-,.

N I-LL

I 0::: :J: ........ CD _J

l:: .......

>-\J1 I-I w -~ u

0 _J

w > (/) (/)

a: E

0 N .

0 -. N

0 0 . N

0 0) . -~

-b ['

ii -0 (0 . -0 lJi)

FIGURE 5-22 · HOT THIMBLE CELL MASS VELOCITY

CASE C

"-t------,-----r-------..------------.------,-------,-------.------, -o.oo 20.00 40.00 so.oo 80 .. oo LOO .oo ],ZQ.00 l40.00 LSO .oo 100.00

.RXIAL POSITION IN INCHES

- - - -·- - ·- - - - - - - - - - - - -· 0 FIOURE 5-23 0 . HOT THIMBLE CELL ENTHALPY 0 er, CASE C ['

0 0 . 0 0 r--

0 0 . 0 r--(0

-l:: CD 0 _J 0 . ' 0 ::J

...,. I-

(0

CD ......,

0 V, >- 0 I ..

w CL 0 V, -l .... a: (0

:::c I-z 0 w 0 .

0 CD lD

0 0 . 0 LO r.g

0 0 . 0 N

"b.oo 20.00 40 .-00 so.oo eo.oo 100.00 120.00 140.00 160.00 180.00

RXI A.L POSITION IN INCHES ....

-------------------

!Z 0 -I-u CI n:::

u, LL I

w 0 °' -0 >

0

CJ .... . 0

0 .... . 0

(X) 0 •

.o

CD 0 . 0

-.t" 0 •

0

CJ 0 . 0

0 0

FIGURE 5-2• HOT THIMBLE CELL VOID FRACTION

CASE C

·~--W---4'"----M---M-~"-----M--4f...-~~----,.------......----.....-----------------, 9).oo 20.00 40-00 so.oo eo.oo 100.00 120 • 00 · . 140 • 00 160.00 teo.oo

AXIRL POSITION IN INCHES

- - ·- - - - ·- - - - - - - - - - - ·- -

~ CD

V, z

I 0 w -..J

0 (X) . N

0 (!) . N

0 ...,. . N

0 N . cJ

0 0

cJ

-0 (D . ...I

0 ...,.

f I OURE . 5-25 HOT THI t1BLE. CELL DNBR

CRSE. C

•;----,------.-----,-----,-----"""'T"""---""""T""------.------,-----l).00 20.00 40.00 so.oo ao .. oo wo..oo 120.00 L40 .. oo iso.oo 1eo.oo

-RX I AL PO S I T I ON I N I NCH ES


SECTION 6 - SUMMA.RY AND CONCLUSIONS

Vepco has developed the capability to perform steady state core

thermal hydraulic analysis of the Surry Nuclear Power Station. These

analyses are being performed with the Babcock and Wilcox LYNX! and LYNX2

computer codes. The method of solution, correlations, and accuracy of the

LYNXl and LYNX2 computer codes have been documented by Babcock and Wilcox.

The specific thermal and hydraulic model developed by· Vepco for the analysis

of the Surry reactor cores has been documented in this report, and results

for Surry have been verified by comparison of Vepco calculations to calculations

which were performed for the design and licensing of the Surry Nuclear Power

Station. These comparisons are representative of the major thermal hydraulic

steady-state design and licensing cases that have been associated with the

Surry reactor cores. The results of these comparisons have been excelient

and indicate ·that the Vepco Reactor Core Thermal Hydraulic Model used in

conjunction with the LYNX! and LYNX2 computer codes can be conservatively

used to provide licensing, design, and operational support for both the Surry

and North Anna Nuclear Power Stations ..

6-1


.I

SECTION 7 - REFERENCES

1. J, M. Alcorn and B. R. llao, "LYNX! - Reactor Fuel Assembly Thermal Hydraulic Analysis Code", BAW-10129, Rev. 1 Babcock and Wilcox, November, 1976.

2. "LYNX2 - Subchannel Thermal Hydraulic Analysis Program", BAW-10130, Rev. 1, Babcock and Wilcox, April, 1977.

3. R. B. McClintock and G. J. Siverstri, "Formulation and Iterative Procedures for the Calculation of Properties of Steam", The ASME, United Engineering Center, New York, 1967. ·

4. J.M. Alcorn and R.H. Wilson, "CHATA:.. Core Hydraulic and Thermal Analysis", BAW-10110, Babcock and Wilcox, January, 1976.

5. "TEMP - Thermal Enthalpy Mixing Program", BAW-10021, Babcock and Wilcox, April, 1970.

6. Final Safety Analysis Report - Surry Power Station Units 1 and 2, Virginia Electric and Power Company, December 1969.

7. Surry No. 1 Fuel Assembly Outline and Reprocessing, Westinghouse Drawing No. 1189E68, Sub 3, February, 1976.

8. Surry No. 1 Fuel Rod Assembly Outline and Reprocessing, Westinghouse Drawing No. 271C861, Sub 4, February, 1976.

9. Private correspondence between Westinghouse and Vepco, FP-VP-393, October 12, 1~76.

10. J. Shefcheck, "Application of the THINC Program to PWR Design", WCAP-7395-L, Westinghouse Electric Corporation, August, 1969. (Proprietary)

11. L. S. Tong and J. Weisman, "Thermal Analysis of Pressurized Water Reactors", American Nuclear Society, 1970.

12. Private correspondence between Westinghouse and Vepco, FP-VP-494, June 20, 1977.

13. "Fuel Densification - Surry Power Station Unit l", WCAP-8012, Westinghouse Electric Corporation, December, 1972. (Proprietary)

14. Technical Specifications - Surry Power Stations Units 1 and 2, Virginia Electric and Power Company (original approved version).

15. Vepco to NRC letter dated August 9, 1977, (Serial No. 344), transmitting Change Number 47 to the Surry Technical Specifications.

16. Vepco to NRC letter dated March 19, 1973, transmitting Change Number 6 to the Surry Technical Specifications.

7-1

;°\

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