surry, units 1 & 2 - reactor core thermal hydraulic analysis … · 2019. 11. 12. · a thermal...
Post on 30-Oct-2020
3 Views
Preview:
TRANSCRIPT
I
"
1. I)
1
1 1· I,
'I:· '1-
I_
I )
.I I I I I ,,~' I~ ,, I
' <
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
I I I I
..
I I-I, I 1-I I I 1·
-·------·-···-- -
I I
.,
-1---I I ·.1
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
I. I I.
I·
I I 1. I I I I ,, .1
I I I
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 ..
l.
I I I
I
I I I I I I I I I I I I I
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.
ii
I I II I I I I I I
· I :.·
I I I I ·I .1 I I
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· ·
..
i
ii.
iii
. . . .~ V
.. . . 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 • • • • ·• • ·• • • • . • • • •
..
..
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 • • • •.
iii
. .
. .
. .
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
5-1
I· ·I ,. I. I I , .. ·· I I
.. , I I I I I I. 1 .. ·
I I
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
6::-1
. 7-1
I I I I I I I I I I I I I I I I I I I
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, -
I I
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-
·1
I I I I I I I I I I I 1-I I I I I I
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
vii
1·
I ., I I I I I I
I I I
I I
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·
1-1
I .,_ .. ··
·.,
I I
I I I I I I I I
I I I
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.
1-2
I
I I I I I I I I I
-1-
-I I I I I I
---~------- ---~-~~~--' ,,. ...... ·
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.
2-1
I
I I I I I I I I . I
I I
-I
I I
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
2-2
I
I I. ·I I I
I I
I
I I I I I
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
3,""'.'l
I I
II I I
1-·I I I
I I
-., I I
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 .fuel- assembly. · 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
::. /'~.::·"
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
-I I
I I I
I I
·1 ,' '
I I I I_
I I
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
I I I I I I I I I I I I I I I I I I I
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
.1 · 1.
I I I I I I
i,I: 1.·.·.
. , I I I
I I
·.,
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.
-I .
. ·I. ·1 '' -I -· .
. I .. ' '
~ . . . .-
. 1 .. ·-'1- :-
1 .:, ·
:1 ,' ._ "'< . ·, . :·' -.,· · .
. I' ,·I·· :.:1_.
:1· .. · .1 . .. ,··: :·1
·. ·· . . · --~., Ii ii 1111 D U ffl R DU II· 111111 . , .- . I . . ..
r.- c-: '' , .
. ~- -T
3-6.
',. -'·:· .
.. '·( :_ ' '
FIGURE 3-2 -
CROSS SECTIONAL VIEW OF SURRY FUEL ASSEMBLIES
~, , .... ~ ,, ,,. ..
. ... /
r;. o._
. . ~
--- --- .41§,(T¥P,J_ - / 0(
7 ; .. , •• , .... ,.,,r,.,,,.~ ~ •c;,
11 _, I . "'1 I ,7 ' oa'
i-~~-~gf 00000000_ o .·x~qciooooooo\ ··p1
_ ~ 0 . __ ,., 0000 0 ·00000000 10: ) C O I O 000 0010 :, 0000000 ~ ~- [O)i · 0 ZD O 000 0 · © O 1,0\1 , . O_ 0 C:·~;(';(; I01 [,-@ 0 01\ · C. 0 OD01
1
1 0' Pl ® O!l C O ". O" ; - · !Iv · I ;:
0 0 ·. 0 OHO O Q) 0 o:: Qooo, .. 0 1
0 Oji"
lo o o ol o o o 01! o o o ol o - o e 01
10 0 0 0 0 OI 'O O O C O O O - 0 0 @ 0 0.1 CJ · 0 0 0 1
·000000000000000 000000000000000! I -
000000000000000 000000000000():C ,J,_ 1 -
0 . _ _ 0 ,IO 000 . ·: "·tH'Y'l 000 .. @ ® 0 0_10 0 0 00 ._, i . •o ---.o~OJ (Q) o. 0 0 0 0 : o e -.Ol 0
1
0-.0~03001 o Q) _ , () oj o o oo,1
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
Thermal Power Level (% of nominal 2441 Mwt)
Average Fuel 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)
Total System Flow Rate (106 lbm/hr)
Reactor Core Bypass Fraction
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
I I I I I I I I I I I I I I I I I I I
TABLE 4-3
CASE B-REACTOR CONDITIONS
Thermal Power Level (% of nominal 2441 Mwt)
Average Fuel Rod Surface Heat Flux (106 BTU/hr-ft2 )
Fraction of Heat Generated in Fuel
Core Inlet Enthalpy (BTU/lbm)
Core Inlet Temperature (OF)
System Pressure (psia)
Total System Flow Rate (gpm)
Total System Flow Rate (106 lbm/hr)
Reactor Core Bypass Fraction
Average Core Mass Velocity (106 lbm/hr-ft2)
4-10
112
0.2177
0.974
552.4
554
2200
265,500
99.2
0.045
2.273
I I I I I I I I I I I I I I I I I I I
TABLE 4-4
CASE C-REACTOR CONDITIONS
Thermal Power Level (% of nominal 2441 Mwt)
Ave6age Fuel R~d Surface Heat Flux (10 BTU/hr-ft )
Fraction of Heat Generated in Fuel
Core Inlet Enthalpy (BTU/lbm)
Core Inlet Temperature (°F)
System Pressure (psia)
Total System Flow Rate (gpm)
Total System Flow Rate (106 lbm/hr)
Reactor Core Bypass Fraction
Average Core Mass Velocity (106 lbm/hr-ft2)
4-11
102
0.1955
0.974
543.6
547
2220
2.38,950
90.2
0.045
2.066
I I I I I I I I I I I I I I I I I I I
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
I I I I I I I I I I I I I I I I I I I
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
I I I I I I I I I I I I I I I I I I I
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
I I I I I I I I I I I I I I I I I I I
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
Thimble Thimble 1.520 Tube 1.520 1.500 Tube
1.540 1.550 1.550 1.540
Thimble 1.540 1.550 Tube
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
I I I I I I I I I I I I I I I I I I I
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
Thimble Thimble 1.520 Tube 1.520 1.500 Tube
1.540 1.550 1.550 1.540
,
Thimble 1.540 1.550 Tube
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
I I I I I I I I I I I I I I I I I I I
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
I I I I I I I I I I I I I I I I I I I
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
I I I I I I I I I I I I I I I I I I I
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
I I I I I I I I I I I I I I I I I I I
--------
\
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
Heat Flux (106 BTU/hr-ft2)
Mass Velocity (106 lb /hr-ft2)
5-8
1.286
94
662.46
-0.070
0.022
0.5256
1.937
I I I I I I I I I I I I I I I I I I I
\
TABLE 5-9
CASE B - CORE AVERAGE EXIT CONDITIONS
Exit Enthalpy (BTU/lbm) 652.1
Exit Temperature (°F) 624. 2 ·
Exit Void Fraction 0.009
Mass Velocity (106 lbm/hr-ft2) 2.272
Pressure Drop (psi) 17.5
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
Heat Flux (106 BTU/hr-ft2)
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 Enthalpy (BTU/lbm) 642.6
Exit Temperature (OF) 618.3
Exit Void Fraction 0.001
Mass Velocity (106 lb /hr-ft2) 2. 06 m
Pressure Drop (psi) 15.0
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.
I I I I I I I I I I I I I I I I I 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 .
Heat Flux (106 BTU/hr-ft2)
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
AXIAL POSITION IN INCHES
- - - - - - -- - - - - - - - - - - - -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
AXIAL POSITION IN INCHES
--~·--;:.--.~.,.~-.. ----.•:::·-:-:-:.· .·-:,--.:-.··-~---:--~--~·--;;: . ...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
/
AXIAL POSITION IN INCHES
-------------------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
AXIAL POSITION IN INCHES
- - - - - - -- - - - - - - - - - - - -
~
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
AXIAL POSITION IN INCHES
- - - - - - ·- - - - - - - - - - - - -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
AXIAL POSITION IN INCHES
- - - - - ·- ·- - - - - - - - - - - - -
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
AXIAL POSITION IN INCHES
- - - - - - ·- - - - - - - - - - - - -
'""' 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
AXIAL POSITION IN INCHES
-------------------
,....,. 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
AXIAL POSITION IN INCHES
- - - - - - ·- - - - - - - - - - - - -
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
AXIAL POSITION IN INCHES
- - - - - - ·- - - - - - - - - - - - -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
AXIAL POSITION IN INCHES
- - - - - - ·- - - - - - - - - - - - -
,-,.
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
I I I I I I I I I I I I I I I I I I I
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 I I I I I I I I I I I I I I I I I
.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
;°\
~ NOTICE -THE ATTACHED FILES ARE OFFICIAL RECORDS OFT.HE
· DIVISION OF DOCUMENT CONTROL. THEY HAVE BEEN CHARGED TO YOU FOR A LIMITED'TIME PERIOD AND MU.ST BE RETURN.ED TO THE RECORDS FACILITY BRANCH 016. PLEASE DO NOT SEND DOCUMENTS CHARGED OUT THROUGH THE MAIL. REMOVAL OF ANY PAGE(S) . FROM DOCUMENT FOR REPRODUCTION MUST BE REFERRED TO FILE PERSONNEL. .
DEADLINE RETURN DATE
RECORDS FACILITY BRANCH
I I I I I I I I I I I I I I I I I I I
top related