JPRS ID: 10238 TRANSLATION OPERATION MODES OF WATER-MODERATED WATER-COOLED NUCLEAR POWER REACTORS BY F. VA. OVCHINNIKOV, ET AL.

• Document Type: CREST
• Collection: JPRS
• Document Number (FOIA) /ESDN (CREST): CIA-RDP82-00850R000500020007-7
• Release Decision: RIF
• Original Classification: U
• Document Page Count: 107
• Document Creation Date: November 1, 2016
• Sequence Number: 7
• Content Type: REPORTS

APPROVED FOR RELEASE: 2047/02/09: CIA-RDP82-00850R000504020007-7
FOR OFFICIAL USE ONLY
JPRS L/ 10238 .
. 7 January 19~2
= Translation
,
OPERATION MODES OF WATER-MODERATED
WATL~R-COOLED NUCLEAR POWER REAG~TORS
By
- F. Ya. ~vchinnikov, et al.
1
~
FBIS FOREICaN BROADCAST If~FORMATION SERVICE
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R040500020007-7
~
NOTE
JPRS publications contain information primarily from foreign
newspapers, periodicals and books, but aiso from news agency
transmissions and broadcasts. Materials from foreign-language
sources are translated; those from English-language sources
are transcribed or reprinted, with the original phrasing and
other characteristics retained.
Headlines, editorial reports, and material enclosed in brackets
are supplied by JPRS. Processing indicators such as [Textj
or [ExcerptJ i.n the first line of each item, or following the
last line of a brief, indicate how the original information was

~ processed. Where no processing indicator is given, the infor-
mation was summarized or extracted.
;
Unfamiliar names rendered phonetically or transliterated are
enclosed in parentheses. Words or names preceded by a ques-
tion mark and enclosed in parentheses were not clear in the
original but have been supplied as appropriate in context.
Other unattributed parenthetical notes with in the body of an
item originate with the source. Times within items are as
given by source.
The contents of this publication in no way represent the poli-
cies, views or attitudes of the U.S. Government.
COPYRIGHT LAWS AND REGULATIONS GOVERNING OWNERSHIP OF
MATERIALS REPRODTJCED HEREIN REQUIRE THAT DISSEMINATION
OF THIS PUBLICATION BE RESTRICTED FOR OFFICIAL USE O~TLY.
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R040500020007-7
FOR OFFlC1AL USE ONLY
_ ~ JPRS L/10238
7 ~anuary 1982
;
OPERAT ION MODES i~F WATER-MODERATED
WATER-COOLE~ ~(~1CLEAR POWER REAGTORS
Moscow EKSPLUATATSIONNYYE REZHIMY VODO-VObYANYKH ENERGETICHESKIKH
'~ADERNYKH REAKTOROV in Russian 1979 (signed to press 3 Jul 79)
jChapters 2.4, 3.2, 3.3, 4.1, 4.2, 4.3, 5.3, 7.1, 7.2, and Append.ix,
8.4, 11.1, 11.2, 12.4, and table of contei~ts from book "OpPration �
Modes of Wat~r-Moderated Wa~er-Cool~d Nuclear Power Feactors", by
Fedor Yakovlevich Ovchinnikov, Lev Ivanovich Golube'v (deceased),
~~yacheslav Dmitriyevich Dobrynin, Victor Ivanovich 1Clochkov, Vladimi~.
Vladimirovich Semenov and Va7.entin Mikhaylovich Tsybenko, Atomizdat,
4,100 copies, 288 pages]
CONTENTS
2.4. Special Features of Neutron Physics Characteristics
of the Core of WER-1000 1
3.2. Coefficients of Reactivity of the Reactor 6
3.3. Requirements for the JVER Control and Safety System 15
4.1. Distribution of Energy Release in the Core 24
4.2. Changes in the Reactivity of th~ Reactor During Its Work
at Full Power 31
4.3. WER Control and Maneuverability 42
~.3. Permissible Power Level of Fuel E1en.znts, ~ssemblies, and
the Reactor 49
7.1. Arrangemene of Fuel Assemblies in t1~re Core 56
7.2. Calculation of the Neutron Physics ~haracteristics of a?teactor 65
Appendix. Example of the Calculat~on of Loading (Reloading) of Fuel
in the WER-440 Reactor 74
8.4. Studying Spent Nuclear Fuel in a Hot Chamber 81
11.1. Reactor Plant 86
11,2. Steam Turbine Plant 93
12,4. Utilization of Spent Fuel of WER 100
Table of Contents 103
- a- IT - USSR ~ K FOUO]
FOR OFFICIAL USE ONL.Y
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2047/02/09: CIA-RDP82-00850R000504020007-7
FOR OFFiCI:~L USE ONL1'
[Text ~
, 2.4. Special Features of Neutron Physics Character.istics of the Core of WER-1000
'r.he dimensions of the core of the WER-1000 [water-moderat~d water-cooled power
~ reactor] excePd substantially the dimensions of the core for VV~R-440. Fuel assem-
blies are larger than the fuel assemblies of WER-440 and contain 317 fuel elements
- each arranged with a somewhat larger lattice pitch. The basic desi~n parameters of
the core, assemblies and fuel elementa are given below:
Basic Desigr- Characteristics of the Core ~f VVER-1000
Equivalent radius of the core 156.0 cm
Core heigh~_ (working state) 355.0 cm
Core volum~~ 27.0 mj
katio of ttie moderato- area ~o the fuel area in the core
cross sec:tion 2.00
, Design Characteristics of WEF:-~1J00 Assembly of the V Block of NVAES [Novovoronezh-
skaya Nuc.lear Electric P ~wer S~tation]
"Box-wrench"' size of assembly 238 mm
Spacing of assPmblies 241 mm
Thickness of assembly wall 1.5 mm (with 25% surface perforation)
Height of assembly with a bunch of control rods or SVP [rods with a
burnable ab;;orber] 4665 mm
Number of fuel elements in the assembly 31.7
Spacing of fue.l elements 12.75 mm
Number of guiding ciiannels for regulating rods 12
Number of channels for energy release indicators 1
Dimensions of guiding channels and the channel for the energy
release indicator 12.6 X 0.85 mm
Size of the centi�al tube 10.3 X 0.65 mm
Material of guiding channels, the channel for energy release measure-
ments, and the ~:entral tube zirconium a11oy
Characteristics of Fuel E:~ements of Assemblies in VVER-1000
; Size of fuel elemer~t jacket 9.1 X 0.69 mm
Material o� the fuel element jacket zirconium alloy
Diameter of fuel pellet 7.6 mm
Diameter of axial hale in fuel pellet - 1.4 mm
1
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2047/02/09: CIA-RDP82-00850R000504020007-7
FOR OFFIC~AL USE ONi.Y
Material of pellet U02
Enrichment c~f make-up fuel,
in a two-year cycle 3.3
in a three-year cycle 4.4
Weight of U02.load in one fuel elemen~ 1575 g
Characteristics oF a Regulating Rod and Rods with a Burnable Absorber (SVP)
Dimensiuns of the envelope of the regulating rod and SVP 8.2 X 0.6 mm
Envelope material stainless steel
Diameter of the core of the regulating rod and SVP 7 mcm
Material of the core of the regulating rod Eu203 aluminum alloy
Material of SVP core boron in a zirconium matrix
Concentration of natural boron in the material of SVP 1%
The achievement of the prescribed burnup fraction of fuel with permissible specific -
thermal loads is determined entirely by the length of time the fuel elements stay
in the core. The operation time af fuel elements in ~NER-440 is three years. A
three-year run was also adopted for thE mair, fue? reloading mode in WER-1000. Due
to hi~h specifir_ loads in the core of this reactor in the above-mentioned mode, an
average calculated burnup fraction of 39,800 Mw X day/T U is achieved, which is ap-
proximately 30% greater than the average fuel burnup fraction achieved in WER-440.
For the first loadings of the corA of the first WER-1000, it is planned to use a
two-year-run mode. In this case, the fuel burnup fracrion will be 26,500 Mw�day/T
U and will not exceed the achieved limits.
' After an experimental study of part of the fuel elements kept in the reactor until
- a burnup of 45,000 Mw�day/T U, the V~TER-1000 reactor is changed to a three-,year xe ~s ao ar e~ a ea 05
. O6 61 8 3 i~ s i~ ~ sa i t ~ 4~ n ~3 06
DI - S! I6~ 9 7 16 6 S2 S! 34 f5 i6 -.S) 6d 6f ~1 - 07
~8 - da ~ d~ JI Z ~5 6 3 at 1 L'~ " ~ ' S7 ~ zo � 08
>>a s ia s n n t_ ~ ~d ~o ~ 09
09 - ~e~ n
10 0 , Is ' s _ R ~ J 2r 6 t! ~2 - 2 - r 5 -~d 10
11 ~ ~oJ e~ s s a t r e u it ~t ? ts - ta n ~e r~ 11
/Z C I Ln 9 4" . 10 d ~ 6. . n 1 ~ J J b 6 3 a ~o y- 9 0�,~a ~ 12. ~ 1)
~
1~ c ii io' ivr J9 f ! 2 n ti " bi ! 3 u 3- - od r~ a 13
14 - ~n, ~ o. s I ~e 1 s, ~ ~ J 3 i 6- i~2 ~ t. Q - os ~!4
I5 W n s r~ a n: ; 4~ n s-~ n~- 15
!s ~oo " ' s~ 6 s j ~ Z n 1! ~ e - I6
l7 ~ I,6 I a9 I~1, ~J !6 tS f IS Sa 1 1 ' - 17
!B . " ~ e i � ~ ~ i~s 2 ~ w s ~ , - - - I8
F9 ~ - - 19
uo > sd e 7 ,t7 f m__ ~
20 - ~ ~ 9 1 e 1f . - ~ 4 t ~ 0 - - ~ - - 10
~ .
21 - I ~ ;o i 12 ~i~ ~ r vi - - - - - 21
12 - I N~~Q ~ t , e ~ , 7 j - ~~p ~IO - - - 1.2
.23 ' ( I I I ~ 4 ' ~r . . .ie~-~-J Z~
~
?4 ?6 ZB .id 32 34 36 38 40 42 49 46 48 50 51 54 56 5d 60 62
Figure 3,9. Cartogram of the arrangement of SUZ assembly groups
in the WER-440 of the III block of NVAES.
Key: 1. Loop
17
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500024447-7
FOR OFFICIAL USE ONY.Y
14 76 7R i0 3? 34 36 3B 40 42 44 46 48 50 52 54 55 58 60 6Z
~
I .I I~I I
" ~ ~ ~ ~n~ n! n6 ni '
~
01 " ~~y i ,im s~ ~ no w it ~ -0 y - 02
03 - ~PQ ~e ~a ~se s~ a se sv ~oc ar ia - - 03
D4 - ~Q I tA w 9 9 Af 4 A4 d9 ~ S 9 01 4 9 91 90 . O~
05 - ~ sr ;a ~e e I~o ~3 is ii ;a ir bb i" et e~ e+ 0~
06 - r 8 So ~la 5 te n~ t. 61 ~ s at Z a~ ~ 6+ 4 " fl~ f!
Ol - 1 FI o9 !d 11 t6 I6 S ~ i4 Sf !6 9' SS 0 5i
OE - go ~ r~ do 3f 1 ti i3 6 S ai ~ of 3 i3 ~ -y 1f 9 g - 08
� n S9 ~t ~J6 t5 ~I~ 3i 7 ]1 JS a a ~ s 09
10 y, ia9 4170 !,se ~~'we !f ~t i3 3l ti 61i r! i, ItJ Z It~ 5 9 3a - JO
11 z,~ I4f Pi In9 st S i!4 13 $ 3 I ti it i! ~i - is li3 if 1e 19 '~i z 1f
1z C) 9I9t Ad !Ti I~F 3 ~ 6 ii ~it 2~l 4 6 6~ t e 9~ t0 ~ C~i ~1~
E
13 C ~ gi I1~ !1 Sf 4! 1 I 1 t ~i lir f~ ~!t ,d! ~y S qy �F~' 13
C
14 t ~ 101 S 9 ~)e 2 f4 1 j I It 5!j Jt ~ 4 if) ~ d1 7! 4!6 96 ~ .
15 NI 9 f) 37 i 4 13 Itl I7! f IS~ I6f 76 d> j97 r 1 ~5 .
16 � n 7~a as "~~a ~~a ~ s 6 r~ r n~ c ss 1` n u - f 6 .
~ -17
17 - i6 ~Oe jd) 7f ~6 IS S~!S S 67 ' 1 1 49 !
]8 ~ Y� a � � ~ ~ ~e ~ ~ ~ ~ i~~ ~ i~ si ~ ~ - 1~
i9 - - - - l9
i~~ a~ a a- ~ n p~ ~sn s~ eo ~ oi ~ m
20 0 9 id S Id 11 ~ s! 4~ . e1 9! ZO
- Z~ - OS r ~1 fl- 1 I 6D ~fl d'. ! _ 2f
` 6
22 . ~ _ . e~i' � e ~ ~d~ ~ !f - ~pP~ - - 22
23 . _ . Ati~ , _ I ; ~ ~ i..i- I ~I I I I . I ~ ~ .
. ~~~III1 60 62
7� 26 1d ,~0 31 94 36 ,~9 40 42 44 46 4B 50 51 54 36 58
Figure 3.10. Cartogram of the arrangement of SUZ assembly groL~s
in the WER-440 of the IV block of NVAES.
Key: 1. Loop
The effectiveness of SUZ assemblies depends chiefly on their location in the core
and on the core temperature. As the fuel in the core burns up, as well as during
the burnup of the loB isotope in the absorbers of the SUZ assemblies, their effec-
tiveness changes. When a reactor is designed, the optimal orc?er of successive set-
ting of SUZ groups is calculated for a definite number of ass~2mblies in a group.
The criterion for determining the order of the setting of SUZ groups is the insurance
uf minimal nonuniformity of energy release over the radius and height of the core.
It is taken into consideration that the reactivir.y excess for the burnup is compen-
- sated almost fully by boric acid, therefore, all SUZ groups are in the upper posi-
- tion with the exception of the last XII working group.
Figures 3.9-3.11 show cartofirams of the arrangement of SUZ groups in the cores of
WER-440 of the III and IV hlock of NVAES, as well as Ko1'skaya AES. The WER core
in the plane is divided into three computation sectors with a 120 degree angle at
the vertex. The neutron-physics computations of the WER are done,as a rule, for
one sector on the assumption that the properties are repeated with a periodicity of ~
120 degrees in the remaining sectors. In each computation sector, the design numbers
18
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00854R004500020007-7
~ NnR nNFI~'IAI. URh: (1Nf.Y
24 26 28 30 32 34 ~6 ~8 ~?0 42~44 46 48 50 52 54 56 58 60 62 ^
~ I I I n3 ub ~i I-I I
I I ' I _'_I
~ ~,roe ro~ ra, ~o~ io m ni ~o -~P~ I 1 I
02 - P7 .
_ 03 ~ e~pA Izo ~ ro 9s s6 ar oe ~d m ni ~ ?Ay~ 03 ,
I _
04 - ~ I i a4
m i9 4 as ee a~ M ds so n s~ 4+
as - ! -I 05
, ob z~ ~e e ~a ~u ~e i~ ie i4 ~ i b3 ee I
06 . 05
ei sb f_ te -~i s i ~S~ eo Z s e ei ~i ib 4 n ` i~ I~i ,
7
QI - ~a si ",9 ~aa 27 .~6 6 e � - . _ yi " se ' i � eo " - . i
I . 08
~8 . ~ - TZ . ~ JI ~t6 Il3 f S m 42 93 4~ ~ 07 -d .e . . i 3d S?
09 ~ E! 4 59 47 2. ii Ito 3 4. 1 JI f! J4 J! 2~ --1 ~
r~ ~ 9~ -!1 70 !A I 6 f ~L4 /S ! 1I T 7! ~ 2f i6 77 1! ~ b }r .
? 1 V! el 69 ~4! I,lb N tv' 1 11 ~ 13 I~ ~ U 16 17 ~A -h. h
. E.I ~ ~oz ~en 6 I e I~ 5~~ ~ x u~ ~ ~ ,f 3 d e 6~ I d ~ 1 ID C f2 ~ 1~
- ~ C' ~ � qi I9 ~67 j73 7? ~ 1 t t i 7 1 71 si ft 7 IA ~as ~~-~q~ c 13
i I 14
f4 - ~a ~nu vo ~e " is. (i .t n : ~ "st e+ i -
15 m J~ f~ii1- !9 r 5 i~ f ~'1- �m- - 15
16 - ii iio iaf l ie s a H a i - i~ e-' o- - 16
i . ' . .
_ . .
11 - ~6 ~ ~ h> > ~ .t 6 is _ . > > 1
18 - Ins 6 ~6 ~r 2_ ei 6- 3- 18
~9 - - 19
~ni io I+e s ~a n-Ir~ ~ ki i vi -
_ - - - _ _
. qp I� ~ j~ ~ r 20~
21 1 ~ro ~ ~3- ~ 4 ~ 1 - 21'
n - - '�~~,ry i ~na d6 ~2 ~
23- I I I~ y Q~~ . r I~ I I I -I I- 23
24 26 28 30 32 J4 ,~6 .~8 40 41 44 46 48 50 51 54 36' S8 60 62
Figure 3.11. Cartogram of the arrangement of SUZ assembly
groups in the WFR-440 of Kol'skaya AES.
Key: 1. Loop
of the assemblies are indicated. The cores of the reactors of the III and IV blocks
of NVAES contain 73 SUZ assemblies which are divided into 12 groups. In the care of
the WER-440 reactor of the III block, th~ controlling group is group No 12 which
contains six assemblies with design numbers24 and 42. Since the WER~440 of the IV
block uses nuclear fuel with a greater enrichment than in the III block, the composi-
tion of the controlling group No 12 was changed in order to reduce the nonuniformity
of energy release. SUZ assemblies with design numbers 1, 7, 68 from group No 10
were included in it, and assemblies with design numbers 24, 42 were transferred to
group No 10.
The core of the WER-440 of Kol'skaya AES has only 37 SUZ assemblies. Unlike'the
WER-210, which also has 37 of them, the assemblies are spaced farther apart and,
accordingly, they are displaced toward the periphery of the core. This is due to
the ~one loading of the WER-440 and a higher enrichment of the fuel. Tables 3.4
and 3,5 show the values of the effectiveness of the SUZ assembly groups at various
temperatures for WER-440 of NVAES.
19
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2047/02/09: CIA-RDP82-00850R000504020007-7
FOR OFFICIAL USE ONY.Y
Table 3.4 ~
Effectiveness of SUZ Groups of WER-440 of the III block of NVAES
(first loading, no boric acid in the coolant)
so �c ~ oo �c i eo �c zss �c:
llo4ep Pec~crude ~
e3noAxMOp xorepa I
- ~yYntis~ KeceCT P. % IQP. % P. % lon. % P. % I~P. ~~i P. ~~o ~n� %
3
Bc ~ yn- - -3,93 I - -6,18 - -9,22 r -15,94 -
nd e~ixsy
1 9G, fi4 -3,38 0,55 -~~i,62 0,56 -8,67 ~,55 -!19,02 1~12
2 2G, Gfi 0,74 4,12 1,48 4,14 -4,03 4,64 -8,25 5,77
3 3, 44 3,89 3,1~ 1,88 3,36 -0,267 3,763 -4,1~ 4,06 .
4 7Q, R6 4,94 1,05 3,01 1,13 1,009 1,276 -2,818 1,3i5�
5 2H, 90 5,81 0,87 4.02 1,01 2,072 1,063 - t,6 l~, I.`l04
6 5, 22 %.45 l,fi4 5,86 I,84 4,14 2,068 I,U51 2.GG~i
7 7`?, 110 R,O: Q,SG 6,4�1 0,55 4,71 0,57 1,59 (1:53~~
R ~i~, IO8 D.'l2 1,21 7,fi7 1,26 5,9'T l,2fi 2,26 O,C?7
!1 !)2 10,;"t6 1,34 9,Ifi 1.49 7,~iM1 1.57 3,$2 l.5fi
10 1. 7, (i8 12,11~i 1,441 10,84 l,fi8 9,4fi 1~)2 fi,41 2,59 ' .
11 4R, RR 13,7t) I,G5 12,,9 1,75 11.39 l.!~3 8A9 2~U8
19 ?q. q2 14.7O l,00 13,78 I.:� 12.71 1,3`l 10.295 I,SO~i�
Key: 1. Number of the group being set up
2. Design numbera of assemblies
3. All lower graups ,
Table 3.5
Effectiveness of SUZ groups of WER-440 of the IV block of NVAES
(first loading, no boric acid in coolant) ;
( ,2` 'l0 �C ] 00 �C 200 �C 285 �C
}I~l~c~ Pecveiuwc
e,nonHMOiI HaM~pa
- rpynna xaccer P. % IoP~ % % I~P~ �o P~ % ~P. �/s P. SP~ % .
~
iBce ~py~n- 3,I16 ~ 1,682 -1.,779 - -6,720 ~
~te~i nri?t3y
1 4fi, G4 3,196 0,08 1,765 0,083 --I,C>47 0,132 -G,495 0,275
2 2f,, G6 3,837 0,641 2,436 0,671 -0,478' 1,IG9 -~4,309 2,13'6 .
3 44 f,179 2,3i42 4,864 2,,428 2,444 2,922 ~-0,934 3,375
4 70, 86 7,681 1,502 6,429 1,565 4,130 1,686 0,83~4, 1,768
~5 28, 90 8,683 1,002 7,478 1,04,9 5,323 1,193 2,175 1,341
. fi 5, 22 10,237 1,554 9,i17 1,639 7,335 2,012 4,633t 2,458
7 72. I 10 10,878 0,641 9,768 0,651 7,91�3 0,578 5,109 0,476 '
8 fi0, 108 12,225 1,347 11,154 1,386 9,2G2 1;,349 6,264 1,155
9 ~J, 92 13,947~ ~1,722 12.94! 0,787 11,225 1,963 8,289~ 2,025
~10 21, 42 15,207 1,260 1~1,281 1,340 13,087 1,853 ~10,825 2,536
11 ~R, R8 16,340 1,033 15,A52 1,171 14,317 1,239 12,102 1,277
12 I. 7, 68 17,772 1.432 16,949 1.497 16.OG6 I,749 14,104 2,002
,
~ Key: 1, Number of the group being set up ,
2. Design numbers of assemblies ;
3. All lower groups
20
FOR OFFICIAL USE ONLY '
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7
FOR OFFICIAL USE ONY.Y .
~
ao V' -t~ OO l~ Qf O c0 'r N ~3' c0
~n p~ N Q> ~A ~ M t0 f~ 00 tD 00
O M M O> ~D h ~p M ~ p p~ ~
. �D O C7 N CV N '
cr N N N CV GV h N h OO 1~
~ M y' O 00 O l~ tp GV Qf ~A M
N - cD ~n ~ cD O O T ~t tD tD O .
. ~ V ~ O O N CV CV Cj .-i Pj
~ ' .
�1-1 w a~o ~ O O~i O ~ M ~ N N ~ ~ O '
~ N O M N ~ N O r+ GN N r+ CV
7 ~
~ pp pp
~ p
1.1 G~+ N ~+~j ~ ti ~ ~ ~ ~ O ~ N d
V~ ~ O N M ~ N O N N GV
Q l 0
. ~ y~ v- O O ~ tD ~ a0 ~ u) M M O~~O ~
y 00 O O O N O N N N O ~ bO
W la t~ .y
O'ra ~y M N t0 l~ tA aD 1~ pf Npp O M
~ x 4-1 O h O~O_ t~D_ sT M ti C~V N ~D ti t~D' dJ OO
v U ~
~ V ~ ~ O O O N ~ ~ CV N O ~ w
O 1~ oo z
~ ,1] RS N a~o S N a~0 N a0 O GV p t~ t~ M M QI M1
rl ~ N ~ ~
~ c~+I~ ti O ~ N N 1~
O O N N ~.j ~
O
~ H U p GV cl tD M C`1 pp 'O~ M Mu~ O1 O~ ~ iJ
~ u o .-M. ~ O~ c~0. ~ O t~ M O~ a0 CV 1: ~1 Gl
i iJ ~ O ~ ~y p ' b0 V
Q
' C'~') O G'+ ~a v~i ~ ~c~ M G~O ' QMi GV N M p~~ G~`l ~ ~ �t' U
, �rl . O O O a0 Q c'~ a0 C'! O d;, O~ JJ '
Q �O O O O O N N cl O O '
c~ ~i ~U t0 P1 �.r~ ti `n� c~ ti ~w ~ M c~`'~v ~ N uo0~i ~ cd
H W ~ o o ch ~ o cv ~ c~ ~
o ~ vf O O O N O N N N O ~ V
V o s
~ ~ ~ 1~ �'1' h t0 N f~ CO t0 C~ ~Opp~ M ~ Fd
! (A O ~ O ~ N ~ CV N CMO~ ~ t0 ~ ~ zi Wa
, GL .a e~ O O N p
~ . .
N
H O
I b0 w v ~ o ~ ~ ~ o ~ c~ ~ ah~o ~ ~ r-~ N
~ o o ei o ~ ~
~ O ao tD OCV~ M N GV t~Dp ~ d' Q1 00 N ?~i
~ ~ O O O tr O N a0 O ~ ~
~ ~ ~ �D O O O ^ O GV CV Cl d ^ O
0 ~1
1J
(n GI . r,
N ~ N O ~ O ^ ~ N O Q~i GV ~ ~
~ ~ r..
~ V o o 0 o ci o N N o
N O �
, ~ v r n n a~ if~ ti ~
a0- o 0 0 � `A ~ oo ao ~n ~ c~i
.u o~ o o r.. c.~ .r o
U
~ N N GV d' f` N O M N
- 4-1 at' -Y O ~n 'c!' d' CV tD M M ~
4-1 0 ~ cD ch ~ O u) c0 M t~ N O -t~
W O O N O ^ ^
c
~ a~ ~
r+ c~ r~ -r co ti ao rn o ~ c~
v or~q ^ "
x,~~ �
~ ,
- 21
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7
FOR OFFIC[AL USE ONLY
The effectivene~s o� SUZ assembly groups dPpends also~ irteffectivenesslat various
acid in the reactor. Table 3.6 shows the values of t e
concentrations of boric acid for WER-440 of the IV block of NVAES.
iv . _ -
16 iR ?.0 Z? ~4 26 28 z0 37_ 34 36 dB 40
- ^f,5 N~'
6' N�!7
�>4 (3) ~ �
0~3 O rs re ~sn ~s
O1~ _ _ n- sr t~ ~ ~ ~a o u a N�18
01 ~~~r . ~ ~12 _ _ VIII XIl1 VaII O
yF 3~ i.~z ~V X 1 ~!V ~X ~ ~ N
03 - - - N�19
- 110 ~Zl ~ 1 1J f1 f25 tt. t7~ r7 ~
o~ - v~~~ xi x v vl~ x~v xr ?~r~ Cj N'~
N�11 ~oe +ov rru ~ ~n n2 ~ ru ao ~~a rra ~n n � nem~A ~ ~ ~
a5 ~ XIl1 IV X!V 1/1 Vl V! I~ XIV l XIII
93 96 9 4(i 16 iJr f0 !0 fJ nrh ~ ro: q N�2O
oe vni ~ vs~ vi xn ix xrr v~ v~~ ~v vir~ ~
B.t R4 B~ 9 67 AA P9 9 9f 9 ?J Ov
n~--- x xiv vi rx ~,~i i.~ d, xiv x
70 7f 'Z 7 ~S 7.5 ' 6 ~B 'r 31 9: ~ I!( .
J ne X! IN ,171 ll ll ~/!I x7 V o N~~ . ,
o.o c~,o~n~r ..,A ~ X X~v ~ ix ii ~ ri ii~ vi riv 59 d N~~ ~
s~ 3 54 ~ .fli 5'
- ~o 5 v~n iv vn �v~ x~~ i~ .r~~ vi vr~ i vb~r .
N~9 `J ~ S3 J4 d5 J 3 J8 ~ 4 4/ 4 y o
_ ~ p N�3 X~l I X/ v I!! V! v/ I!I .t~ I v XIII O N 2
~ sn~ N�8 U 2 2 z zs z z 2e zo ao s~ ~2 ' i
_ >2-- VIII X 1~ XIV VII XIV XI X VIlI o Np N�~
!1 13 l
4 1 l
6 f 7 f 8 1 1 2
1,~ V X IV b X V . ~
N�7 n � s s ~ e ~ to t> '~y, ,
r4 ---~Q ' V!l ZXlll iV~l4 0 ~ � ;
N 6 (n~~
>s - - o :
' Q Q . ~ -4 NP _6 . i
0~ N5 '
Q ~~~4 HQ -5 ~ i ,
~
Figure 3,12. Cartogram of the arrangement of the groups of SUZ
� absorbers and neutron meas~iring channels in the co:e and the
ionization chanbers of the SUZ of the WER-1000 reactor of the
q block of NtIAES:
1-- assemblies with SUZ absorbers filled to one half by absorb-
ing materials;2 assemblies with neutron measuring channels; 3
number of the group of SUZ absorbers; 4-- energy-range ioniza-
~ tion chambers; 5-- startup-range ionization chanbers; 6-- standby
startup-range ionization chambers; 7-~ intermediate-range ioniza- ;
tion chambers.
Key: 1. loop
i
The mechanical control and safety system in WER-1000 of the V block of NVAES con- !
sists of 109 drives each of which is capable of moving a bunch (clusterj of twelve ;
absorber rods (core material Eu203 in a m3trix of an aluminum alloy) inside the
� assembly within tne limits of the core. The SUZ drives combined into groups (Figure ~r
3.12) move the clusters simultaneously.
22
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPR~VED F~R RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7
' FOR OFFiC1AL USE ONLY
- The mechanical system of SUZ is intended for compensating rapid changes in the reac-
tivity (temperature, power, snd poisoning effects). Slow changes in the reactivity
(fuel burnup) are compensated by changing thE concentration of the boric a~id solu-
tion in the coolant. For emergency situations, there is a high-speed boron injec-
tion syster.?.
The total effectiveness of the mechanicai system of the SUZ of V'VER-1000 must be not
less than the sum of the following effect~:
~ ~ Doppler effect of fuel when the power of the reactor changes from 0 to 100% 0.013;
changes in the average water temperature of the first circuit when the power changes
- from 0 to 100% 0.014;
changes in the steam content in individual jets of the coolant in the core when
the power changes from 0 to 100% 0.002;
effective reserve for nonsteady-state xenon poisoning and leveling of energy re-
- lease 0.015;
effectiveness of a jammed bunch of absorbers not over 0.010;
- initial sub:.riticality after the triggering of the safety system 0.010.
In order to allcw for errors in the neutron-physics computations, the necessary ef-
fectiveness of SUZ is taken 20% higher than the total sum of the ahove reactivity
effects and must be not less than 0.077.
COPYRIGHT: Atomizdat, 1979
- 23
- FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7
FOR OFFICIAL USE ON?.Y
F
4.1. Distribution of Energy Release in the Core
The distribution~of energy release in the core is characterized by energy r.elease
variation factors which must be known for ~'etermining the permissible thermal power ;
of the reactor (see section 5.3) .
It is customary t o consider the energy release variation factors with respect to the
radius, height, and volume of the core. According to the definition, the energy re-
lease variation fa ctor in the i-th assembly over the radius of the core kr~i is
equal to .
k,.~ = Q~~Q, ~4.1>
where Qi power of the i-th assembly; Q-- average power of the assembly in the
core.
For boron-regulat ed reactors, the maximum value of kraX varies within the limiCs of
1.2-1.4, and for reactors with regulation by mechanical members of SUZ within the
limits of 1.5-2.1. '
The small variatian of energy release in boron-regulated reactors ma.kes it possible
- to remove a great amount of thermal power from the core. The variation factors of
energy release in the core are calculated on a computer for the entire run (see sec-
tion 7.2). Exper imental values of kr~i are determined on the basis of water temper-
= ature measurement s at the outlet from the assemb lies and the coolant temperature at
the inlet to the reactor. The values of the variation factors are computed by the
formula
m i++ ~
~edx _ `l ~nx G ~ ~ ~
l ~ ~ ~1 ! gl ~~F . 2~
k~,1 = ~ ~ m m n '
- c1 ~sax ~l I~ g~ rI z GJ I~ G, ,,~j gl
L
j 1 f- n
: ~
I ~
i
where tRh~x water temperature at the outlet from the i-th assembly; tBx
water temperature at the inlet to the reactor from the j-th circulation loop;3G3 -
ma.ss velocity in the j-th circulation loop; gi mass velocity in the i-th assembly;
m-- number of c irculation loops; n-- number of fuel assemblies in the core.
24
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R400540020007-7
' FOR OFFICIAL USE ONLY
For measuring the water temperature at the outlet from the assemblies, ~NER are
equipped with a temperature monitoring system. Temperature measurements are done
by thermocouples which are sufficiently efficient in a wide temperature range under
the conditions of neutron and gamma irradiation. Chromel-copel and platinum-plati-
norhodium (10% Rh) thermocouples are used most widely. In WER-440, the outlet wa-
ter temperature is monitored for about two thirds of the assemblies of the itire
- core, and in V`lER-1000 at the outlet of all assemblies. The water temperature
at the inlet to the core is measured in the circulation loops. Depending on the
steam load of the steam generators, the number of working loops, and the hydraulic
= resistance, certain deviations of the inlet temperature in the loops are possible.
In practical calculations, the water temperature at the inlet to the core is taken
to oe equal to the average value for all working loops of the reactor. In averaging,
differences in the water flow rate in the loops are taken into considerations.
The flow rate of the coolant through the assemblies of the core are determined on
the basis of the hydraulic characteristics of the assemblies (see section 5.2). The
power of the assemblies under operating conditions is calculated by the energy re-
lease variation factors at a known average power of the assemblies of the core. The
distribution of energy release over the height of the core is usually obtained by
calculations for the entire run (see section 7.2). Experimental distribution of en-
ergy release is determined in special measuring channels with the aid of the sensors
of the in-pile monitoring system. From 12 to 36 measuring channels are installed in
the core of WER-440. The measuring channel is a stainless steel tube with the lower
end plugged up passing through the lid of the reactor into the central tube of the
working assembly. In the WER-1000 assemblies, the detectors of the in-pile monitor-
ing system (VRK) are arranged not in the central tube, but in a special channel of
- energy release measurements. The WER-1000 reactor of the V block of NVAES has 31
measuring channels connected to the VRK system (see Figure 3.12).
The distribution of the neutron flux density and energy release along the height are
measured by activation and emission detectors, as well as by ionization chambers
[31]. Calibrated copper wires with a constant mass per unit length are used as ac-
tivation detectors. The wire is irradiated in the measuring channel in the course
of time sufficient for its saturation by the 64Cu isotope, after which it is removed
from the core and coolPd for a while for the decomposition of the short-lived cop-
per isotopes. The distribution of beta-activity through the length of the wire mea-
sured aftpr this will correspond to the distribution of the neutron flux density
;aith the height of the measuring channel at the moment of the irradiation of the
wire.
- In ~NER, emission detectors direct charge detectors (PrZ) are used widely.
The operating principle of DPZ is based on the appearance of an electric potential
in a detector consisting of an emitter and a collector during the beta-decomposition
of tl~e neutron-sensitive emitter. Rhodium and vanadium are usually used as DPZ emit-
_ ters. DPZ are small and sufficiently simple secondary devices. A drawback of the
- direct char~e detectoris their rather great inertia. In the series-produced WER-440,
each measuring channel has four rhodium DPZ 250 mm long and one vanadium DPZ 2500 mm
long. Rhodium detectors are intended for measuring the disrribution of the neutron
flux density with tlle height of the channel, and vanadium detectors are intended
- for measuring the total neutron power in the channel. Moreover, the distribution
of the neutron flux density with the height can be measured by moving the DPZ along
_ the length of the channel.
25
FOR OFFIC[AL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2407102/09: CIA-RDP82-00850R000500420007-7
FY)!t UFFtI'1:11. U~F: UN1.1.
For pracrical moni~orin~ working conditions of the fuel elements, it is necessary to ~
know the energy release distribution whose correlation with the density distribution
of the thermal neutron flux cnanges as the fuel burns up. At the initial moment of
burnup, the neutron flux density and the specific energy release are connected by
the relation
W = moaz5 Nss, (4 .3)
where 6~5 fission cross section of 235U; N25 initial concentration of 235U
nuclei.
Allowing for the accumulation of 239Pu and 241Pu and 235U burnup~ the relation (4.3)
assumes the form
W=`P ~a25N26 T Q39 N39 ~ Q11 N41)~ (4 .4) ~
where d39 and o41 235fis239n cross241ctions of 239Pu and 241Pu; N25, N39 and N41
concentration of U, Pu and Pu nuclei, respectively.
- In order to proceed from the measured density distribution of the thermal neutron
flux to the distribution of energy release, it is necessary to perform conversions
- whicti, as a rule, are done on the computer and allow for the burnup of 235U and accu-
mulation of plutonium isotopes. For this purpose, information from DPZ is loaded
- into the computer which promptly processes the data and informs the operator about
the results of ineasurements with consideration for the burnup of the DPZ emitter.
In order to simplify the computation programs of information (controlling) electronic
computers,� it is desirable to use, in addition to the detectors measuring the den-
~ sity of the thermal neutron flux, detectors whose indications characterize directly
the energy release in the fuel elements surrounding the channel. Energy release in
fuel elements is characterized unambiguously by the flux density of fast or resonance
neutrons which can be measured with the aid of modernized DPZ or ionization chambers.
For e~:ample, a DPZ with an emitter of silver surrounded by a cadmium jacket for cut-
ting off thermal neutrons registers chiefly the density of the flux of resonance neu-
- trons which, in the final analysis, is proportional to the energy release in the sur-
rounding fuel elements.
Reduction of the factors of variation along the radius kr and the height kz and the
volume factor of variability kp = krkZ is of great practical significance from the
view point of the possibility of increasing the reactor power and the fuel burnup
fraction. Therefore, the work on the leveling of WER power is being done continu-
ously.
In WER, a zonal principle of load assembling is used (see 5ection 7.1), which makes
it possible to level energy release along the radius of the core and, in combination
with boron regulation, to reduce the volume nonuniformity of energy release.
In the process of fuel burnup, there occurs additional self-leveling of energy re-
lease due to the nonuniformity of burnup proportional to energy release, and due to
the nonuniformity of the manifestation of the poisoning effect and Che power effect
of reactivity. For an example, Figure 4.1 shows the changes in the calculated and
experimental values of maximum variation factors of energy release kq X and kmTX
26
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7
FOR OFFICIAL USE ONLY
K~iner. ~ ~ ~ ~2~ IlXll
V cy3
, 11~~~ ~2~
- c 3 ~ CR'K?
_ Z00 9,fi e
~ ~ n e e ~
o m
~8.f50 e
~ oc c n
0,6 1,/ Kr+aKc
0,5 100 " ~ 4
0,4 f,3 a ~ � ~ � o � o 0
0~ 50 ~1,2 0 ~ 5) ~4)
0,9
a D 1,1 3m, y
- 0 50 f00 150 200 250 T c m~ru
Figure 4.1. Changes in the maximum energy release variation
factors during the work of the first loading of WER-440 of
the IV block (curves calculated data; dots experimental
data) .
Key: 1. Maximum 3, g/kg
2. SUZ 4. days
' S. Teffective
n% - -------nk-e,s4~io -
7,5b
/
1,6 . 6,46
~~5 , 5,40 .
4,J2 :
~'4
. J,14
~'3 2,16
1,2 1,Od ~
1,1 �
1~0
_ o o,~ 0,2 0,~ 0,4 o,s o,s o,~ o,e q9 i,o t,~
Relative height of withdrawal
Figure 4,2. Dependence of the height factor of energy release
variation on the height of withdrawal of inechanical members of
SUZ with different integral effectiveness = 1 for cosine
? /2
_ distributi~n of energy release along the height of the core) �
27
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7
FOR OH'FI('IAi. USN: ONI.Y
f.or the first fuel loading of tiie VVER-440 of the IV block of NVAES. The values of ,
~the variation fac~tors decrease in the course of the loading operation. Some increase
in the variation factors at the end of tY?e run is due to the gradua~l withdrawal of
the controlling group of SUZ from the core. There is a particularly strong depen- ,
dence of the height distribution of energy r~lease on the presence of partially
inserted SUZ assemblies. Figure 4.2 shows theoretical dependence of the height
variati~n of energy release of the core on the withdrawal height of the StJZ assem-
blies having different effectiveness [6]. The withdrawal of "light" groups creates
the least nonuniformity. ' ;
1,5 . _
2,0 .
~
= 1 - ,
0
(1)`^
o z--
~
~
E- �
c
b
E
0
~
a
_ ro 0,5 - ~
.
0 0,5 i,0 1,5
3~+epzoBbrdeneHUe, omN. ea. ~2~ .
Figure 4.3. Energy release distribution along the height of the_
- measuring channel of the WER-440 of the III block of NVAES (1 -
computation; 2-- measurement), Channel in the cell 13-30; power
in the reactor 55%; withdrawal height of the 12-th group 127 cm;
~H3B03 = 2.79 g/kg H2O.
Key: 1. Core height, m
2, Energy release, per unit value
rigure 4.3 shows the distribution curve of energy release obtained in the measuring
channel of the WER-440 of the III block.
In conclusion, let us note that the leveling of energy release increases the proba-
bility of the appearance of xenon oscillatiorn (see section 4.3). Xenon oscillations
are the effect of periodic redistribution of power over the core volume caused by
the feedback between the power and concentration of 135$e.
In WER-440, the probability of xenon oscillatio~is small and if such oscillations
occur, they are aperiodic in nature, have a small amplitude, and attenuate rapidly.
A. distinguishing characteristic of WER-1000 is the possibility of the occurrence of
spatial xenon oscillations of power in the volume of the core. The probability of
- the occurrence of xenon oscillations increases as the dimensions of the reactor in-
crease in the case of disturbances in the distribution of power. The greatest
28
FOR OFFIC[AL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2047/02/09: CIA-RDP82-00850R000504020007-7
1~(llt (ll~l~ll'IA1. Ittil? (1NI,Y
field disturbances in WER-1000 occur in the operation mode with changes in the
level of power, for example, the lowering of power from 100 to 50% for a while with
subsequent increase to 100% (Figure 4.4). The figure shows clearly the appearance
of deformation in the axial distribution of the neutron f lux caused by transitional
xenon processes and the shifting of the SUZ regulating membera.
- ' 2,0 ~ 2~~ 2'~ T=5,3a
~ ~ f0,6
~ ~ ~ o > 0 T-Ov
,
s, 4r~~ 15,8
~ 0 0,5No.,,,anH.ea 0 0,5NQ3,om~~ea 0 0,5NQ,9,omy.ea ~
a a 6 b ec
Figure 4.4. Axial density distribution of the neutron flux in
W~R-1000 under the condition of changing loads: a-- reactor
power 100%, regulating rods are withdrawn from the core; b--
reactor power lowered to 50%, regulating rods are lowered to the
- height of the core equal to 0.4; c-- reactor power increased to
- . 100% after working at a power level of SO% in the course of 5.3
hours.
Key: 1. pei� unit va~ue
1
Cmep~rNU ynpo6naroweu zpynr.~ ,
! ,CmepMNU 9 ~2~
1 1 ~ ?
~ ~ 1 ~
a ~ . ~ - . .
x - -
o _
a . - -
o - - -
~ ' -
(3) Y ~ -
T
,
- .
Padoma ~+a noulHormu t.8y � t�84
I~O~~O HOfIUNU/IAN01f - _.__....._i ~
Po6om~l Ha noWHOCmu Paboma Na noWHVCmu
~ 5~ 50% ~~onuNVner+ou ~4~ ~009'o NOMUHanoNOu
Figure 4.5. Movement scheme of control rods Y for suppresaing
spatial xenon oscillations along the height of the core of WER-100.
Key: 1. Rods of the controlling group
2. Rods Y 4. Work at 100% nominal power
3. Core 5. Work at 50% nominal power
6. t = 8 hours
29
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2047/02/09: CIA-RDP82-00850R000504020007-7
- FOR (1FN1('IAI, l~`F (1N1.1'
Evidently, in ~ number of cases, in order to reduce the dimensions of disturbances
in power distribution, it is expedient to lower the regulating rods completely, and
to compensate their influence by changing the concentration of boric acid during the
entire operation time of the reactor at a lower power.
A sufficiently developed system of in-pile monitoring with results processed by a
compu*.er is a component part of the WER-1000 control system. The operational infor-
mation about the distribution of power over the core makes it possible to correct
the developing deformations of the field in time. Radial and azimuthal deformations
of power distribution can be corrected by withdrawing or lowering certain groups of
absorbers. Height deformations can be corrected with the aid of a special group of
control rods Y with a half-height of the absorbing substance by changing their posi-
tiun along tt~e height of the core (Figure 4.5).
COPYRIGHT: Atomizdat, 1979
30
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7
' FOR OFFICIAL USE ONLY
4.2. Changes in the Reactivity of the Reactor During Its Work at Full Power
When a reactor works at full power, besides the changes in the reactivity described
in section 3.2, its reactivity changes as a result of the effects of poisoning and
- slagging of the core.
As a result of the fission of uranium and plutonium nuclei, there form various fis-
sion fragment nuclei and product nuclei of the radioactive decay of fragments. It
is possible to isolate two main groups (see Chapter 1) among fission fragments and
products of their decay whose accumulation in the core affects substantially neutron-
physics characteristics of the reactor. The first group includes 135ge and 149Sm
nuclei which have large thermal neutron absorption cross sections. The decrease in
� the reactivity of a reactor as a result of the accumulaCion of 135Xe and 149Sm nuc-
lei is called poisoning. Poisoning by 135Xe is particularly important in transi-
tional processes, because the half-life periods of 135Re and its predecessor 135I
are relatively small. The second group of fission fragments and products of their
radioactive decay includes stable and long-lived isotopes which have relatively small
neutron absorption cross sections. The decrease in the reactivity of a reactor
caused by the appearance of fission products during the burnup of fuel is called
rea~tor slagging.
Reactor Poisoning, Poisoning of a Reactor by 135Xe. Chapter 1 gives the chain of
the formation of 135Xe from 135I. A small portion of 135I does not form directly
during fission, but as a result of the radioactive decay of another fragment, 135Te.
The half-life of 135Te is very short, therefore it is assumed in calculations that
- the entire 135I forms directly during fission. Allowance is made in calcu~.ations
that a small part of 135Xe nuclei forms directly during fiesion. The balance of 135I
and 135Xe nuclei in the reactor is described by the system of diffe~ential equations:
dNI - E ~ ~,,N,; (4.5)
dT ~t JT Pr -
dNX~ _ ~+Xe ~IT~PT -I- a'INI - (6;ceV~T ~Xe) N;c~, (4.6~
r!T
where NI. Nge concencration of iodine and xenon nuclei, respectively, cm-3;
235 239 1~
~X e-- yield of iodine and xenon per one fission of a heavy isotope ( U, Pu,
241Pu~~ Z+fT macroscopic fuel fission cross section, cm'1; ~T current den-
sity of thermal neutrons, neutrons/(cm2~sec); ~li, ,~1. Xe decay constants of
iodine and xenon, sec'1; 6 Xe microscopic absorption cross section of neutron
by the isotope 135Xe, cm2.
31
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7
FOR OFF[CIAL USE ONY.Y
= 0.061� Y = 0.003;
Xe
For VVER, Liie followin}; xenon process constants are taken: Z ~
~ I= 0.287�10-4 sec-1; a ge � 0.207�10"4 sec-1.
In a rigorous examination, the values of ~fT and 6Xe musC be averaged over the
spectrum of thermal neutrons in the reactor which is determined by the temperature
of the fuel and coolant, concentration of boric acid in the coolant, layout of fuel
assemblies in the core, concentration of 135ge nuclei, etc. As a rule, the neutron
spectrum is calculated in one or another degree of approximation on a computer (see
Chapter 7).
~ const i.e., under ~he condition of
Under the condition that ~fT = constp ~T q ~ 4.6 can be solved
constancy of the power and the neutron s ectrum e uations (4.5)-( )
analytically: (4.7)
- N~ = Noc [ 1 - exp 7~iT )l ~
~ equilibrium concentration of 135I nuclei.
where N~i = ~ i ~~T ~T/~~
~ Equilibrium concentration of 135ge nuclei is dPtPrmined fron the expression
(YI'~- YXe~ ~JT`PT (~F.B~
Noxe = ~,Xe + QXeTr .
As follows from the above formulas,steady-state (equilibri135ge~concentrationfdepends
and ~35Xe nuclei depends on the neutron flux density, and
on it nonlinearly. .
-o,ns~ -----N
>on�o ~
(1) - - ' -0,04
0 04 - 75 �
- ~ ' _ ~J n~ -O,Od '
. . _ -
_ ~I,Od _ 25 t .
F - - - - - ~ -0,02 - -
~ (],02 - - ~ 10 q
~ -n,n, ~ . -o,o> -
--zU - 30 40 l,v 0 50 Nm~~
Figure 4.6._ Steady-state xenon paison- xenonepoisoningpofdWER-440SOnathermale
ing of WER 440.
Key: 1. Per unit vaiue power.
Figure 4.6 shows curves of steady-state 135$e poisoning of the core of WER-440.
Wt?en the reactor works at a steady power of 100%, the maximum steady-state poison-
ing equal to 4% is established approximately after 40 hours. At this moment, equili-
brium is established in the formation of 135Xe nuclei from 135I and their disappear-
ance as a resundenceroflsteadVestatey135gebpoisoningtof nWERr440fonXitsFpower.4.7
shows the depe Y
When the power of the reactor changes from N1 to N2, the balance of 135I and 135$e
nuclei is distu-rbed, which causes transient processes accompanied by changes in the
reactivity of the reactor. When the power decrF~.:.ses, the reactivity of the reactor
also decreases, because, as a result of the decrease in the neu~ron flux density,
32
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2047/02/09: CIA-RDP82-00850R000504020007-7
� ~
the burnup of xenon by neutrons decreases, while its yield from 135I~ whose amount
at the initial moment is determined by the former power level, does not change,
which leads to an increase in the concentration of 135Xe nuclei and increased poi-
soning, This phenomenon is called iodine pit,
The greatest depths of the iodine pit (4.5%) occurs when the load of the reactor is
reduced from 100% to zero (Figure ~+.8 a), which is achieved nine hours after the
_ drop of load. Her_ceforth, as the number of 135I nuclei and, consequently, the num-
ber of the forming nuclei of 135Xe decrease, the reactivity increases. When ~he
power is increased, the poisoning of the reactor by xenon first decreases (depoison-
ing takes place) due to intensive burnup by the increased neutron flux of 135Xe,
whose yield f rom 135I remains for some time corresponding to a lower level of power.
Then, the increased yield of xenon nuclei from iodine compensates the released reac-
tivity and brings additional poisoning due to increased concentration of 135Xe nu-
clei (see Figure 4.8d). The maximum depoisoning can reach 0.6-0.7% reactivity.
In plotting the curves in Figure 4.8, it was assumed that, before changing the power,
- the reactor operated in a steady-state mode for a long time (2-3 days). The origin
of coordinates was selected as an initial point for all curves, and for determining
the total poisoning of the reactor by xenon, it is necessary to shift a11 points of
curves in the direction of negative reactivities to a corresponding steady-state
poisoning. If the cl~ange in the power in the reactor occ~.urred before the establish-
ment of steady-state poisoning, then the following should be done for determining
~ the total poisoning of the reactor by 135ge: to determine from curves in Figure 4.6
the poisoning of the reactor by xenon at a given power at a given moment of time
and, using Figure 4.7, to find to what steady-state power level the obtained value
of poisoning corresponds. Then, at the established value of the steady-state power,
the parameters of the transient xenon process of interest to us are determined from
Figure 4.8.
If the reactor was stopped for a longer time (more than 1.5-2 days), then practical-
ly the entire iodine and xenon decay, and curves in Figure 4.6 must be used in cal-
culating the poisoning of the reactor after its startup. If the reactor stopped
for less than 1.5 days, then for estimating its xenon poisoning after subsequent
power increase, it is sufficient to have in mind that the total poisoning of the re-
actor tends to the steady-state poisoning at a given power. A method for a more ac-
curate estimation of poisoning for such cases is given in work [32].
The time of attaining the maximum depth of the iodine pit depends on the percentage
of the lowering of the reactor power. For example, if the total l0d% load is drop-
ped, the maximum of poisoning is attained after nine hours, but if the load is drop-
ped from 100% to 50%, it is attained after five hours (see Figure 4.8a). The total
time of 135Xe processes is approximately equal to 40-50 hours.
Let us give an example of calculating the poisoning of a reactor by the 135Xe iso-
tope.
Problem. WER-440 worked at 100% power (1375 Mw therm) in the course of 10 days.
As a result of the triggering of the scram system, the power of the reactor dropped
to 25%. After three hours, power was increased to 75%. Xenon poisoning of the re-
_ actor to be determined.
33
FOR OFFICIAL USE ONLY �
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2447/02/09: CIA-RDP82-00850R000500424447-7
N'OR UN'H7('IAt. t~SH: l1Nl.V
i
O,~AT - - -_.Y . . - N ~O ~
N~ f00% ~ 1n
o ~
0~07 - - - - - - - Zy-'--i ( ,
0,01 - . - - Sp ~
N n - 30 41.~ ; 4
(1) E n,o~ - - - -
p,~) - - - - - _ .i ~
:
0,03 . - - -
0,04 . _ _-~1 . ~
~ �
n,n.~ . _ . _ : - ,
n,a~ ~
_ _ -
rv,-n
N~ ~ � R5 : , ~n^~ ~
~ l~n~ . ~ _ f-
~ 15
q,01 - - - - I -5~ ~
= 70 _ y
~ 0 - - -
f00 ~
0,01 - I
- - ll0? - - - - _ _ _ _ _ _ -
~ , '
. 0~~~ . . - _ . _ .
lI,O,~ - - - - - M11_Q ~
N~ ~ -50Y
- . -f0%
n,o2 - - - .
qor - - - . - - 25
0 10 30 40 T v ~
~ ~ !0
-ao~ ~
e
-0,02
qo,~ - -
Nr~�25% N1e~
0,02 =
>0;6
~ 0,0f - -
_ i
E
~ o !0 20 .~0 40 50 T,v
-0,01 75 '
f00 a
-qn~ -
Figure 4.8. Nonsteady-state xenon poisoning of WER-440 when
power drops from 100 (a), 75 (b), 50 (c) and 25% (d) levels. '
Key: 1. Per unit value . '
34
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R040500020007-7
FOR OFFICIAL USE ONLY
Solution. Before the shutdown of the reactor, steady xenon poisoning equal to -4%
became stabilized (see Figure 4.6). Three hours after the drop of the loud to 25%,
the depth of the iodine pit reached -1.890 (see Figure 4.8a), and before the reactor
- power was increased to 75%, xenon poisoning was equal to(*4%), +(-1.8%) _-5.8% of
reactivity. After 1.5-2 days of power operation of the reactor, xenon poisoning
dropped to the steady-state value at this power, i.e., to -3.75%.
Transient xenon processes have a substantial effect on the maneuversbi.litq of the
~ AES.
Modern water-moderated water-cooled power reactors work, as a rule, in the boron re-
gulation mode, which ma.kes it possible in the steady-state mode to withdraw almost
all SUZ assemblies from the core. Among the SUZ groups, only the controlling group
remains in a semi-inserted state in the core, which compensates reactivity distur-
bances connected with the maintaining of the neceasary power level of the reactor.
The effective reactivity excess of the controlling group of the SUL is equal to
- N0.5%. When the load variations are small, the effective reactivity excess is
quite sufficient for compensating the temperature and power effects and the poison-
ing effect in the transient xenon processes. When the reactor power drops greatly,
the depth of the iodine pit becomes greater, which can lead to the necessity of im-
mediate removal of boric acid from loop I to compensate poisoning.
For an example, let us examine the above problem in a somewhat modified form.
Problem. VVER-440 operated at 100% power in the course of 10 days; the concentra-
- tion of boric acid in the coolant was 2 g/kg, and the effective excess of the SUZ
control group was 0.4%. As a result of the triggering of the scram system, the
power of the reactor dropped to 25%. After three hours, power was increased to 75%
of the nominal power. What actions must taken by the operating personnel for chang-
ing.the reactivity in order to ensure the operation of the reactor at the above-men-
tioned power levels?
Solution. Steady-state 135Xe poisoning and the power effect were compensated during
the operation at 100% power by a boric acid solution and by the control group. When
the load drops from 100 to 25%, a power effect equal to +1.2% is released (see sec-
tioil 3.3). However, this not enough for compensating xenon poisoning which will
reach -1.8% after three hours (an iodine pit maximum equal to -2.25% is attained
after six hours d~iri.ng operation at 25% power see Figure 4.8a). In addition, it
is ne~essary to release the effective excess of the control group equal to +0.4%,
and also +0.2% reactivity by removing boric acid from the reactor. Moreover, at the
expense af the removal of boric acid, it is necessary to releas~e +1.2% reactivity
three hours later for compensating the power effect during the subsequent increase
of power (if there are no other conditions, it is necessary to count on a power of
100%) and +0.45% of reactivity for fully compensating the iodine pit (-2.25%), since,
as a rule, the exact moment of the emergence of the reactor to a higher power level
is not known, and it is quite probable that this will occur at the moment when the
maximum of the iodine pit is attained. Thus, in order to compensate +0.2+1.2+0.45%
=f-1.85% reactivity, it is necessary to lower the concentration of boric acid in the
~ reactor by 1.38 g/kg (ap/acH3BO3 =-3�10'2 kg/g) by delivering 67 m3 of pure water. to
loop I[calculated by formula (4.20)]. When three feed pumps with a total capacity
of 14 m3/h are turned on, the delivery of such a volume of a volume of.water requires
35
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00854R004500020007-7
- 1~OR (1H1~1('IAI, l~~l? (1N1.1' ,
4,g ;~~urs, in whicll case the concentrati~n Wi~l becreleasedlwhichrisstemporar8ly/kg .
after three hours and +1.26/0 of reactiv y
~ompensated by inserting SUZ assemblies into the core.
Since the power of the xeactor will incremoval of5ob~oriceacidrwill~besused.foo~r comhe
reactivity released as a result of the oisonin The effective
pensating the power e�fect and 0.2% for compensating p g�
26%.
excess of the control group wil'. be +1.26-0.8�0�2%
Reactor Poisoning by 149Sm. Chapter 1 gave the chain of the formation of 149Sm from
149pm directly during the fissionl49 uranium and plutonium nuclei and, additionally,
during the disinti ration of the Ndef ardedninncalculationsadue~tohitsrshortehalf-
the intermediate 1~9Nd isotope is disr g
life (1.8 h) .
The change in the concentration of 149pm and 149Sm is described by the following
- differential equations: .
c 'vPm ~'/T(~T - ~1'Pm NPinr ~~F .9~
c17~
dNs,,, ~ (4.10)
d7' - ~,pm NPm - QSm~T Nsm+
where Npm, Nsm concentrations of promethium and samarium nuclei, re~_s,pfTtivelmacro-
~m 3~ yield of promethium per one fission of heavy isotope;
Y pm CPT average density of the therma.l neu-
scopic fuel fission cross section, ~m 1' romethium decay constant, sec'1; 6 Sm
tron flux, neutron/ (cm2 X sec) ;.)1, pm ' P
microscopic cross section of the absorption of neutrons by the 149Sm isotope, cm2.
For WER, the physical constants have the following values: Y
pm = 0.011; ~.pm =
0.357�10'S sec'1.
= const (reactor power constancy condition), it is possible to obtain analyt-
If ~ T for N
ical solution of equations (4~10) and (4.9) Sm~
~'Pm ~NOPm NnPm~ (EXp A,'pmT~ -
N5m (T ) � Nnsm Ex~ QSm~PrT~ + ~Pm - ~PTQSm (~F .11~
- exp QS~,~~PT~~ IVosm ~ ~ - eXP UStn~PTT~~i
_ YP,"~~T _ equilibrium concentration of 1~9Sm nuclei;
- -
where Nosm"
QSm
Y~',"~fi mT equilibrium concentration of 149pm nuclei; N n gm and Nn, pm "
Norn, =
~`Pm
concentrations of Sm and Pm at the moment of transition.
36
FOR OFFIC[AL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00854R004500020007-7
l FOR OFFICIAL USE nNLY
P,amNed (1) _
-0,008
' - 0007
' N=>oo io
- 0,006 ~s
-0,005 s~
15 -
-0,004
- 0,003
- g002
_ -0,001
. 0 10 ?0 d0 T, cymn~ ( 2) '
Figure 4.9. Steady-state poisoning of VVER-440 by samarium.
Key: 1. Per unit value ~
2. Effective days
Thus, the equilibrium concentration of the nuclei of samarium No Sm does not depend ~
on the neutron flux and, consequently, an the reactor power. However, the reactor
, power determines the time of the attaining of the equilibrium concentration of 149gm.
Figure 4.9 shows the curves of the poisoning of the WER-440 core by samarium for
. the first startup from which it follows that the equilibrium concentration of 149g~
nuclei and steady-state (equilibrium) poisoning, which is 0.82% for WER-440, is
attained in the course of 30 effective days when operating at a steady power. The
~ time of the onset of steady-state 149Sm poisoning can also be ~estimated from the re-
lation
, l CT84 ~ 1Qlfi/TTt C4 ~ 12,
where T CT ~II" time of the onset of steady-sLate poisoning by sama.rium during power
operation of the reactor, days; ~T average density of the thermal neutron flux
= during reactor's operation at a steady power, neutron/(cm2�sec).
The change in the reactor power from N1 to N2 (Figure 4.10) causes 51ow transient
processes connected with the changes in the numbers of 149Pm and 149Sm nuclei in the
core. The phenomenon of the lowering of the reactivity of the reactor when its
power docreases due to the disturbance of the balance of 149pm and 149Sm nuclei, by
analogy with the iodine pit, is called promethium pit. The greatest depth of the
promethium pit (of the order of -0.570) is attained when the power of the reactor de-
creases from 100% to zero (see Figure 4.10a). and the tota~. transformation of the
formed 149pm into 149Sm occurs approximately in 15 days after the drop of the load.
During this time, the samarium nuclei formed from promethium during the shutdown
time are added to the samarium nuclei accumulated during the time of power operation.
The 149Sm isotope is stable, therefore, at the zero power of the reactor, the number
of samarium nuclei remains constant. When the power drops partially, the depth of
the promethium pit is smaller, because part of the accumulated samarium is burned up
by neutrons.
37
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-04850R000500020007-7
F(lit (1F1~1C'IA1, lltil~ t1NA,\'
10 20 d0 r~p~.c mKU 1)
0
Np=75% 50 N~f 100%
-D,ODf - -25
-0,002 - -
TO
-0,003
- -0,004 - - - - -
. -0,005 - - Nt~o .
( 2) omN~eB. '
a
~ ~ ~
~ ) 0 75 % ~
2 p~omN.eB. N2=1009'al ~~i f 1
~ Z .~o r, ~xu ( )
0
- ~50
' -0~001 25 - - - i
-0,002 - ,
~10
-0J003 - ~N ,
2-
-0~004 b , f
b
' '~2) P,omH.eB.
N~ ~ 50 �/a .
N2=100% - "
0,001 75
_ . 0 Zp 30 , 3~� cymnu
25 . . - .
-0,001 - - -
10
-0~002 ,
- N *O
-0,003 8-~ . -
Figure 4.10 a, b, c ,
_ Key: 1. Ef.fective days
2. Unit per value
- The burnup of the "excess" of the samarium formed from the promethium after partial
power drop gradually brings its concentration to a steady-state value which, as has
been menEioned, does not depend on the power level. It should be kept in mind that
38
FOR OFFIC[AL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7
t~c~R ~t~~rr~ ~i.~ ~ . t +c~' (1NA ,Y
- (1) on~,,.r,i. , . . . I
_ a' N2-1n0% N~~{25%
0,0015 .75 . .
~ .50
0,0010 _
' 0~0005 . . _ _I _ . .
i
~
_ ~ 10 ZO 30 T, 3q~. cymKU ( 2)
j
-00005 ~ . .
-0,0010 I ~ - _ _ ' �
Nz�0
_ Z d ~
-0,0015
Figure 4.10. Nonsteady-state poisoning of WER-440 by samarium
when thermal power changes from 100 (a), 75 (b), 50 (c) and 25%
(d) levels.
- hey: 1. ~Per unit value
2. Effective days
the curves in Figure 4.10 presuppose that steady-state poisoning by 149g~ is attained
_ before the change in the reactor power.
The decrease in the time of the onset of the maximum of the promethium pit as the ~
values of the power N2 increase in Figure 4.10 should not be misleading, because
the effective time is in question, and not the calendar time. The calendar time of
the onset of the maximum of the promethium pit remains constant (15 days). As the
power of the reactor increases, a samarium overshoot (reactivity increase) is observ-
ed, which is explained by the change in the rate of the burnup of sama.rium by neu-
trons and the rate of its accumulation from promethium. The maximum sama.rium over-
shoot can reach 0.25% during the time of the order of five hours after the rise of
the reactor power from zero to 100%, assuming that the reactor was shut down for 15
~ days and the concentration of samarium stabilized as constant.
- Let us give an example of the computation of the poisoning of a reactor by the 149Sm
isotope.
Problem. After 40 effective days of operation WER-440 was shut down for two days.
Before the shutdown the reactor operated in the course of 20 days at a 75% power.
- After tlie startup, the reactor was brought to a power of 5070. The poisoning of
the reactor by samarium has to be determined.
Solution. Directly before the shutdown of the reactor, steady-state samarium poison-
- ing was established in the core, since the transient process connected wi'th the em-
- ergence of the reactor to the 75% power level occurred 20 days before the shutdown
of the reactor (more than 15 days) and, moreover, in 40 effective days, steady-state
~ 39
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-04850R000500020007-7
FOIt OHFIC'IAL USE ONLY
poisoning by 149Sm became established. From Figure 4.9, we determine the steady- i
state poisoning by samarium equal to -0.82%. From Figure 4.lOb, we find thatld5ref- ,
/ 100�/ '
ing the two calendar days of the shutdown of the reactor daThe tot~al poisoning of i
fective days) the depth of the promethium pit reached -0.2/0.
the reactor by samarium directly before the startup is equal to -.82% +(-0.2%) _
-1.02%. When the reactor reaches the 50% power level, samarium burns up and approx-
imately after 30 effective days of operation (30 eff. days�100%/50% = 60 days) the
poisoning drops to the steady-state value of -0.82%.
_ Reactor Slagging. The composition of stable and long-lived fission fragments
slags forming during the fission of 235U nuclei is given in Chapter 1. The accu-
mulation of slags in the fuel is proportional to the neutron flu~ncy and is an energy
characteristic of nuclear fuel.
The unit of ineasurement of fuel burnup is the amounda S ttherma~nothergfrequently
from 1 t of fuel in terms of inetal in 24 hours (Mw y/ Ln �
used unit is the slag mass accumulating in 1 t of uranium (kg/tU). Fuel burnup is
directly connected with the number of nuclear fissions of the fissionable isotope:
1 ~ 2~ (4 .1~)
Mmll�f.~RtKll `>,:3~_
~_~~Z' aen/mU,
1 _ ~
mU ~
Key: 1. Mw� day
_ 2. Fission/mW
E-- energy reieasedin one isotope fission event, Mev: 5.3916�1023 coefficient of
conversion of inegaelectron volts into megawatt per day.
The energy carried away by neutrinos is not included in the value of the energy of
one fission event in calculations, since it is nof theereactor~the volume of the re-
actor and is not a part of the thermal power
The recalculation between the burnup measurement units is done by a simple calcula-
tion of the mass of the s~lit fuel nuclei with consideration of the fact that a ma.ss
of 1.6599�10'27 kg corresponds to the atomic mass unit:
r 1 (4.14)
1 Mern�cy~~u _ A�I,fu99�10-a~.5,391G�102' rct uin}+~c{s
mU L mU
Key: 1. Mw�day
2, kg of sZag
where A-- mass of the nucleus of the fissionable isotope, in atomic mass units.
Table 4.1 shows nuclear fission ene=�gies of 235U, 239Pu, 241Pu, as well as the num-
ber of fissions Nf and the mass of slags forming during the release of energy in 1
Mw�days/mU.
The relation of burnup units using the data of Table 4.1. can be writted in the form
1 kg slag/mU ~ k"1 Mw X days/mU (4.15)
w
- 40
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R040500020007-7
FOR OFFiC1AL USF. ONLY
Table 4.1
Energy and Mass Characteristics
of the Burnup Process
~2~ N 10'~ k~,
_ > ~
Naoron E. M3n ~tA 3 ~ o' S`
az wnaKOe ~ !
- ~~~y~nrcu Mem�ctimKU ~4~
xa6U 195 [9) 2.76 1,078
asoPu 202 [9J 2,67 1,059
a~lpu 205 [ 16J 2,63 1,052
_ Key: 1. Isotope 3. Fission
2. Mev 4. Mw�day
5. kg of slag
In WER, a considerable release of energy occurs due to the fission of the accumulat-
ing plutonium isotopes 239Pu and 241Pu in which case the ratio between the concentra-
tion of the isotope 235U and content of the fissioning plutonium isotopes changes.
Therefore, for a more accurate calculation of the accumulation of slag, it is neces-
sary to consider the changes in the coefficient lcw in the course of the run. At the
beginning of the run, when working with pure 235U, the value of lcw is equal to 1.078
�10"3, while during the burnup of 28 kg/T with the corresponding accumulation of
plutonium isotopes, its value decreases to 1.072�10'3.
COPYRIGHT: Atomizdat, 1979
41
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2447/02/09: CIA-RDP82-00850R000500424447-7
hY1R ()b'F'It'1:~1 l~`b' 11N1
~
4.3. VVER Control and Maneuverability
In WER, power control is facilitated by the negative temperature effect of reactiv-
ity. The temperature effect of reactivity in a number of practically important cases
is capable of compensating power perturbation during emergency situations without
the interference of control devices.
First, let us examine the processes in the reactor after a load perturbation without
consideration for the influence of the power effect of reactivity. As the load of
the turbo-generators increases at a constant power of the reactor, the flow rate of
steam per turbine increases, and the steam pressure in the steam generators decreases.
The decrease in the steam pressure and, consequently, of the saturation temperature '
increases the thermal head of the steam generator and the removal of heat from the I
- I circuit, which, in turn, low~rs the average temperature of the coolant in the core.
The reactivity released during the lowering of the coolant temperature is spent on
increasing the power of the reactor. After the termination of the transient process,
the average temperature of the co~~lant in the core becomes equal to the original val-
ue, and the power of the reactor f~tlls in line with the increased removal of heat
from the steam generator~. The self-reg~llation process has the nature of damped os-
cillations whose amplitude and period depend on the scale of changes in the load and
the value of the negative temperature ef~ect.
The negative power effect of reactivity, which appears when power increases, compen-
sates to some degree the effact of the temperature effect. Therefore, the stabili-
zation of the parameters of the reactor plant actually occurs at a lower medium tem-
_ perature of the coolant than before the transient process and a power corresponding
to the amount of heat removed from the steam generator. The change in the average
water temperature in loop I is determined by the ratio of the temperature and power
coefficients of reactivity. Figure 4.11 shows the curves of changes in the WER-
365 parameters in the process of self-regulation when one turbine is disconnected.
During th~~ self-regulation of WER, the power of the reactor eventually falls in
line with t.he load of the turbo-generators. However, the process of the establish-
_ ment of i~he steady state continuing for a relatively long period of time is aperiod-
ic with large initial deviations of the parameters. Therefore, sufficisntly simple
and reliable forced control systems are installed in V~1ER.
WFR power control is accomplished according to the programs shown schematically in
Figure 4,12.
42
FOR OFFICIAL USE ONLY ~
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00854R004500020007-7
FOR OF FI('IAI. LISF ONL�Y
~ !40 - a
~ 120
~ 100
_ ~ ~SS - - --T - --6
~
~~oo
2~ 50 - "
0 -
6 -
8
~
~2~ ~ 4 ~ _ .
~ 2 _ _ '
. 4 z -
V O
' 1 2 3 4 5 T, r~uN 3~
-1
B
~ 2
~ ~ i Z
~
3 4 5 T, MttH
-4
Figure 4.11. Changes in the parameters of WER-365 in the
case of a sharp load drop;
a-- position of the SUZ control group during the adjustment
of a load perturbation by the regulator of the ARM [automatic
power control]; b-- electrical power of the unit; c-- steam
pressure in steam generators; d-- average coolant temperature
in the reactor; in the self-regulation process (1
~H3B03 = 2.3fg/kg; apia4 =-1.7�10-4 1/�C; 2-- Cg3g03 = 0 ~
g/kg; dp/dt =-4.3�10 1~C; during regulation by ARM
regulator.
Key: 1. Nelectric~ MW
2. Kilogram-force/cm2
3. Minutes '
_ The maintenance of a constant average coolant temperature in the core while the load
is decreasing is accompenied by a rise of steam pressure in the steam generators.
The positive aspects of the program of maintaining a constant average temperature of
circuit I are the maximum utilization of the self-regulation property of the reactor,
less rigid requirements for the volume compensation system of circuit I, and insig-
nificant changes in the amount of heat accumulated in circuit I. The latter is par-
ticularly valuable for dependable operation of AES in the mode of variable loads.
The negative aspect of the program of maintaining a constant average coolant temper-
ature is the necessity of manufacturing heavier steam generators designed for the
- saturation pressure at the average temperature of circuit I, which is 15-20 kg-f/cm2
higher than the nominal pressure.
43
FCR OFF[CIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2047/02/09: CIA-RDP82-00850R000504020007-7
FOR OFFICIAL USE ONLY .
. '
;
i
~ i
~~~n ~
I
!
~ i
i
L~r ~
S
~
. '
o a nr ioo%
~ - i~K-
t~~
S
~ - ...N . 100%
t t1K
Cnr
_ s r`
~ i i
~ 6 N 100�/r
t . tIK
nr
-------^~~t S
0 Z N 100%
Figure 4.12. Programs of WER power control:
a-- with constant average water temperature in the reactor;
b-- with constant steam pressure in circuit II; c-- stepped ~
regulation by the average water temperature in the reactor;
d-- compromise program of regulation with ma.intaining con-
~ stant steam pressure in circuit II at small loads and a con-`
stant average temperature in circuit I at large loads;
average water temperature in the reactor; saturation
water temperature of circuit II in the steam generator.
The advantage of the constant steam pressure program is the use of the least expen-
sive casings of steam generators, as well as easier temperature conditions of the
operation of circuit I at a lower power. However, this regulation program is char-
acterized by the greatest changes in the thermal potential of circuit I as the load
of the steam generator changes: when the load changes from 0 to 100%, the average
temperature of the coolant of circuit I increases by 20-30 degrees C.
The disadvantages of the programs for maintaining a constant average temperature of
the coolant in circuit I and a constant steam pressure in steam generators are par-
tially eliminated in the compromise regulation programs.
WER working in the basic load mode are regulated by using the regulation program
with maintenance of a constant steam pressure in the steam generators. For example,
44
.
FOR OFIFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2047/02/09: CIA-RDP82-00850R000504020007-7
FUIt UH FICIAL USE UNLY .
W~R-365 and WER-440 use a two-pulse reactor power control system with an analog
power control of the ARM (IRM) type [36]. Figure 4.11 shows the curves of changes
- in the WER-365 parameters when the load drops from 120 to 60 Mw and the ARM regu-
lator adjusts the perturbation. As can be seen from the curves, the ARM regulator
steadily and rapidly adjusted the power of the reactor to the load.
During the regulation of the reactor, it is permissible to change the load of the
unit at a rate of about 3-S% a minute. With respect to their regulation character-
istics, AES are close to hydr.oelectric power stations and, in principle, can operate
in the mode of variable loads. However, when a reactor is controlled by changing
the concentration of boric acid, the rate of possible changes in the reactivity is
considerably lower than when it is regulated by mechanical elements of the SUZ. The
concentration of boric acid is changed by diluting the coolant of the reactor with
pure water or, on the contrary, by feeding a highly concentrated boric acid solution
into circuit I. Naturally, the time of the removal of boric acid from the coolant
- of circuit I and, consequently, release of reactivity, is much longer than the time
of the withdrawal of SUZ assemblies, which limits Che possibility of changing the
reactor power in nonsteady-state transient xenon processes.
Let us estimate the rate of changes in the concentration of boric acid by solving
the differential equation of acid balance in circuit I:
VydC = CQOnn9uonnYnonndT - CqnpyIIydT, (4.16)
where V-- volume of circuit I of the reactor without volume compensators, m3; -
water density at the average coolant temperature in circuit I of the reactor, kg/m3;
C-- concentration of boric acid in the water of circuit I,g/kg; C noRn concen-
tration of boric acid in the feed water of the reactor, g/kg; q o n- volumetric
flow rate of the feed water of the reactor, m3/h; Y no~n ~ee~ water density,
kg/m3; q np volumetric flow rate of the blast water of the reactor, m3/h;
Y nP blast water density, kg/m3; T-- operation length of the feed pumps, h.
Having solved equation (4.16), we obtain the expression
C~~~) C~~~,o~ f~~~'~~, ;1~, r 1- exp 9n~Y P 7'/ I{ (4.17 )
~!n iYnp ~ \
Coexp r- 9�~1'~~~ T1
\ ~V ~
Having used the condition of the material balance of the feed water and blast water
_ in circuit I, �
9nonnYuonn = 9np~np~ (4.18)
we obtain
~~T) = Cnuttn I 1- exp 9i~~Vy 7�1'~ C~ exp 9~ivln~ Tj , (4.19)
~ i ~
Expression (4.19) can be used for calculating the changes in the concentration of
boric acid in the coolant when circuit I is fed with water with a high boric acid
content and with pure water. For pure water, formula (4.19) is simplified:
45
- FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7
FOR OFFICIAL USE ONLY
4ul~ynP .
c (r~ = co eX~ --~v ~ (4 . Zo~ ~
If Che initial concentration of boric acid in the coolant is equal to zero (pure
water), then the changes in the concentration of boric acid in the case of feeding
with boric water is described by the expression
L 9nPYnP J J ~~F . 21~
C(T CnoAu 1- exp - VY T.
Formula ~4.17) also holds true for the case when the safety boron system is included
into operation and when water from circuit I leaks. In this case, a high concentra-
tion solution of boric acid is delivered by high-efficiency safety feed pumps.
For maneuverability estimation, the concept of a rela*ive rate of the lowering o�
the boric acid concentration with the dilution of the c~olant t~f circuit I with
water is used: ,
I dC (T) Qnp1'np
_ ~ - C ~T~ dT Vy ' (4.22)
The maneuverability of WER changes in the course of the run depending on the con-
centration of boric acid in the coolant (Figures 4.13 and 4.14).
(1)
R'/. ~-C(~) Z N,~80,~ - C=6 Z
N=~~- ~ 1 ~ .
~ ?r2 HZO ~ KZ N20
4,5 1,5 . _ - . _ '
5
4
,~0 1~0 - -
3
2
1,5 0,5 - - - -
1
~ 25
F �
0 0 1,0 1,0 d,0 4~0 T, 4~2~
- Figure 4.13. Changes in the concentration of boric acid in the
coolant and the release of reactivity when circuit I is fed with
- pure water. The flow rate of the feed water is 13 m3/h, and the ,
volume of circuit I is 200 m3. '
~ Key: 1. gHgB03/kgH2O
2. T, h ,
46
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7
Fc~R c~~Ft~t.~~. t~~~. c~Nt.v
~ ~1~ -
~ ~-CnoBn \ - -
Go~noOn
\ ~
1 ~ P 6nJ/y
, ~ \ ` qn
_ a� ; \ ~ . __s~`
~
; ~ ~ ld
~ \ 12
r00 ~ ~
n ` 10 ZO 30 ; v
Figure 4.14. Relative changes in the concentration of boric
acid in the coolant of circuit I of VL1ER-440 at various flow
rates of pure feed water: hot state (t=260�C);
cold state (t=100�C).
Key: 1. Cfeed
As it follows from Figures 4.13 and 4.14, the withdrawal of boric acid from circuit
I occurs slowly, which, however, does not create any inconveniences in the operation
of the reactor at a steady-state power. When the power of the reactor changes,
- there occur difficulties in raising its power which are connected with a rapid de-
crease in the reactivity excess due to the poisoning of the core by 135Xe (iodine
pit). In such cases, provision can be made for switching on the standby ionite fil-
ter cleansed of boric acid which greatly increases the rate of boric acid removal
[37]. Figure 4.15 shows the characteristics of the maneuverability of a unit with
WER-440 with consideration for the presence of the controlling group of SUZ assem-
blies in the core for various relative rates of the decrease of boric acid concen-
tration. It can be seen from the curves that the maneuverability of the reactor
gets worse toward the end of the run.
The maneu~~erability of a reactor depends on the speed of bringing it to i.ts full
power after a shutdown. For example, if a reactor is brought rapidly to its full
power, the accumulated xenon is burned out intensively by the neutron flux, and the
additional formation of Xe from iodine does not yet have any substantial effect on
the reactivity. This fact can be used for increasing the power of the station tem-
porarily, for example, to cover peak loads of the power system at the end of the
reactor run when the reactivity excess is small. By lowering the power during the
hours of low needs in energy, it is possible to accumulate a reactivity excess neces-
sary for maintaining the fu11 power of the reactor during peak hours. At the NVAES,
this mode of operation at the end of the run was tested repeatedly on units II and
III. Calculation show that it is possible to select an optimal mode of power vari-
ations ensuring the fulfillment of the daily load schedule of a power system.
47
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7
FOR OFFICIAI. USF. ON1,Y
;
,
~
- ~
I
NT'2'~ Nrf=f00%
BO - - -
. ;
- - ~
40 > ~ , 2
.
- ~
p p,2 0; f 0,6 O,B f,0 '
reactor run, per unit value ~
Figure 4.15. Characteristics of the ma.neuverability of an AES
with WER-440 with consideration for the presence of the con-
trolling group of SUZ in the core:
reactor is maintained at the rated power, and attains
the rated power in the case of a shutdown of the reactor for not
more than one hour; reactor is maintained at a lower power
and can attain the rated power at any time; 1-- ~3 � 0.05 1/h;
- 2 f~ = 0.20 1/1i.
The probl~ms of the optimization of transient xenon-related processes are treated
in detail in works [34, 35]. The basic mathematical method of the optimization of
- these processes is the Pontryagin maximum principle.
COPYRIGHT: Atomizdat, 1979
~ 48 ~
FOR OFF[CIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R040500020007-7
FOR OFFICIAi. t?SF. ONi.Y
~
5~.3. Permissible Power Level of Fuel Elements, Assemblies, and the Reactor
The main generally accepted initial prerequisites of determining the permissible
level of the thermal power of a reactor are: a) impermissibility of the melting of
the fuel even in individual, the most energy-intensive fuel elements; b) necessity
of the absence of a heat exchange crisis on the surface of the fuel elements of the
~nost energy-intensive assemblies in the steady-state and any transient mode of oper-
ation of the reactor.
The conditions of the absence of fuel melting were analyzed in work [3], where,
among other things, it was shown that, for the fuel elements of WER-440, in which
slightly enriched uranium dioxide is used as fuel, the maximum linear load is ~-500
w/cm. This load for the height distribuCion variation factor kZ = 1.57 corresponds
to a maximum power of fuel elements of 0.08 Mw.
; permissible Power of Fuel Elements. The maximum power of fuel elements with re-
spect to the heat-exchange crisis which is characterized by the appearance of cri-
tical phenomena even in one section of the fuel elements at prescribed physical and
geometrical characteristics of fuel elements is determined by the flow rate, temper-
ature, and pressure of the coolant. The condition of safe operation of a fuel ele-
ment can be ensured if the heat flow in the steady-state and emergency modes cioes
not exceed the critical flow over the entire height of the core. This condition de-
, termines the permissible power of the fuel element with respect to the heat-exchange
crisis. Critical values of heat flows are calculated by the empirical formula (5.9).
- Finally, the permissible power of the fuel element N A~n is established as the low-
est of the maximum powers with respect to the melting o~fuel and the heat-exchange
' h a definite factor of assurance k(arbitrary to a certain degree due to
, crisis wit tors .
insufficient understanding of the processes of heat removal in operating reac )
Possible deviations of physical and geometrical characteristicsT 3~ uel elements are
allowed for by a special factor which is called mechanical k Me~ . With consid-
~ eration for these tw~ coefficients, the permissible power of fuel ele-
- ments is equal to
NTn n ` NTnin~kMex~k+ ~5.30~
where N npe~ the lowest of the maximum powers of the fuel element with re-
TA 3JI
spect to the melting of fuel and the heat-exchange cri'sis.
The factor of assurance k for WER-440 is taken to be 1.1.
49
FOR OFF[CIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7
FOR OFFICiAL USE ONi.Y
The main limitations on the permissible power of fuel elements in WER are imposed
in analyzing emergency situations connected with d~creases in the flow rate of water
through the core (see section S.5). In analyzing emergency conditions of this kind,
it is necessary to calculate the law of changes of the flow rate of the coolant in
the heated channel and the power of the fuel elements by using the formulas of sec-
tion 5.1 which correspond to the appearance of a heat-exchange crisis at the initial
and lowered values of the flow rate. Then the actual change in the energy release
in the fuel e~ements after the operation of the safety system caused by the lowering
of the f~low rate of the coolant is calculated.
By comparing the curves of the changes in the flow rate of the coolant and the actual
energy release in the fuel elements (see Figures 5.7 and 5.8), the initial level of
power of the fuel elements is determined in the steady-state made which makes it pos-
sible to prevent the heat-exchange crisis in the course of the entire transient pro-
cess.
The actual power of the fuel elements and its changes in thz course of the run are
determined by calculations using, for example, programs BIPR-4 (45] and "Hexahedron"
[46]. Due to the absence of a system for measuring this power during the operation
of the reactor, the permissible rated power of fuel elements is decreased by the
accuracy of calculations, i.e.,
N783ApaCy - � � TD3 ~~3(T83l1)~
(5.31)
where k~3(TB 3Jt) " factor of assurance for a gossible deviation of the actual value
of the relative power of the fuel element in the reactor from the calculated value.
It is determined by the accuracy of the accepted calculation methods and programs,
specifically, for the programs BIPR-4 and "Hexahedron" k~~TB 3~~ 1.14.
In WER, the condition of the absence of the melting fuel is fulfiZled in the major-
ity of cases of operational modes and, therefore, the ma.ximum power of the fuel ele-
ments is determined practically only by the danger of critical phenomena of heat re-
moval.
Permissible Power of Fuel Assemblies. Knowing the permissible power of fuel elements,
it is possible to determine the permissible power of fuel assemblies (clustErs). It
is determined by the condition of the absence of the heat-exchange criais even on
the most energy-intense fuel element.
For the convenience of the evaluation of the critical power of the assemblies, the
concept of critical power Nkp of an assembly with mean physical and geometrical
characteristics and equally intense fuel elements is introduced. Critical power is
calculated with semiempirical formulas obtained by processing experimental data in
bundles of heat-producing rods (see, for example, work [47]). Specifical~y, the
critical power of a WER-440 assembly at constant heat production along the height
is determined by formulas
NH~' G,9 ( I- 0,323xQN'~~~'aQK'~ Mem;
- Nu ~~niax ~~z~ +2~~/8G0 Mem; (Mw) (5.32)
H ~l
NK ~ = N,;,
50
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R040500020007-7
~ FOR OFFICIAL USE ONY.Y
where i i coolant enthalpy at the inlet and outlet of the assembly,
BX ~ BbIX
` respectively, kcal/kg; x-- mass steam content at the outlet from the assembly; Qk
water flow rate through the assembly, t/h. The water flow rate through.the as-
sembly can be determined if we know its hydraulic characteristics (see section 5.2). ~
The maximum permissible power of an assembly when there is no boiling of the water
(XB~X = 0) is found by the formula '
(5.33)
~IKp�n�" - (i' - io:) QK~860 MQm.
In real assemblies, power is distributed nonuniformly amorag the fuel elements.
This nonuniformity is characterized bp the coefficient kk which is equal to the ra-
tio of the maximum power of the fuel element of the assembly to the mean power.
This coefficient depends on the enrichment of the fuel, the position of the assem~ily
in the core, the position of the regulating elements, and the concentration of the
liquid absorber of neutrons (boric acid) in the coolant. The value of kk is deter-
mined through calculations (see section 7.3).
With consideration for the mechanical coefficient kM ex and the factor of assur-
. ance k, the expression for the permissible power of the assembly will assume the
following form:
NK~~, _ NK''/kKkMea~. ~5.34)
TB 3J!
Here, k~RPx mechanical coefficient which, just as k a,~e x , allows for devia-
tions in the physical and geometrical characteristics of the fuel elements from
the mean values and, moreover, for daviations in the lattice pitch of the fuel ele-
ments. Consequently, ~~K ~hT~3n
Mez n+ex '
OII
The connection of N x~~ with NTB3.~ is characterized by the relation
' non _ Nre
nnTn~nkMex~ NTn9n nrn~n
N = ~ (5.34')
~c 1~MexkK kMexk~~k
where nT~3~ the number of fuel elements in the assembly. For VVER-440 as-
semblies, n TB3~ = 126. '
In determining the permissible power of assemblies, sometimes it is taken into con-
sideration that, due to the agitation of the coolant flow, the most intense fuel ele-
ments of the asseml~ly are in better cooling conditions. The agitation effect is
allowed for by the coefficients k n which depends on kk. For example, for V~1ER-440
assemblies, kn = 0.95 at kk = 1.15.
Rorr. pac~
The permissible rated power of assemblies N K , just as the permissible
Aon. pac~
rated power of fuel elements N TA 3J! , is reduced by the accuracy of calcula-
tions, i.e.,
51
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2047/02/09: CIA-RDP82-00850R000504020007-7
FOR OFFiC1AL USE ONLY
NKO~,.pa~~~ _ NKO~,.~k,~k,~~. (5.35)
where k 3(K9;C) " factor of assurance for the deviation of the actual power of the
assemblies from the rated power. F~r the BIPR-4 program, characterizes~the inaccu~~
1.1. In this case, the relation k~~TB3~~ ~k 3~xaC)
racy of the calculated determination of the relative power of fuel elements in the
assembly with a prescribed power, for example, for the program "Hexahedron"
h~(�rnan)I~lapcac)= / ~
]~I KP = f I~~C~c) andNKOn. pacv ~~21(~~c) - found by
In~practical calculations, the relations
formulas (5.32) and (5.35) are plotted on the chart of the family of the hydraulic
characteristics of assemblies Nk = fo(Qk) (for various values of pressure difference
The intersection points of the curves make it possible to
in the core Q pa. 3)�
determine the critical and permissibie rated conditions of work of the ma.ximally
intense assemblies .
In the process of operation, the the oolantWat thetoutletlfromeindividual assemblies
by measuring the temperature of the c ower of
and average temperature at the inlet to the core. The permissible measured p
the assemblies is corrected with consideration for measurement errors.
The above evaluations are done in the following mnn.ner. The relative power of the ' ,
fuel assemblies kq is calculated witl. the relatio
/ f )
! G!- ~ ltedx~ ~ ex �M.
(5.36) ;
K� C?KOxrP 1 ~ ~tae+x~ - i ~~ex)
m ~.r
i~ ~
Here, K~, _ K q pa~q /S average xated relative power of the fuel assembly in a '
1=~
sample having a temperature controof~fueleassembliesthavinguaetemperatureacontrol~in
number of the design numbers relative
at least one of the symmetry sectors (see Figures 3.9-3.11); Kq p aC~I
rated power of a fuel assembly in a cell of the core with ~-th design number; m--
- coolant enthalpy
number of fuel assemblies having a temperature control; i(t~~x ~
at the outlet from the j-th assembly determined by the measured temperature t~
kcal/kg; i(t s x)-- coolant enthalpy at the inlet to the reactor core having t~ie
temperature t~X , kcal/kg; G~/GxoxTp ratio of the rate of coolant flow through
the j-~h fuel assembly to the mean flow rate for the assemblies have a temperature
~ control.
The value of t R X is determined either by the direct measurements or by the rela-
tion '
rn
nt c
I l pK
l�x = - tAwx - KM~ (5,37)
m h~,p~KOErrn
-
~
52
FOR OFFICIAL US~ ONiY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7
- FOR OFFICIAL USE ONLY
where A tp average measured heating of the coolant in the reactor, �C; GkIG~OHTp
ratio of the average rate of ~oolant flow through all fuel assemblies to the
mean f1Qw rate far assemblies having a temperature control; k np coefficient
- of the rate of coolant ~low through fuel asaemblies (see section 5.2).
The advantage of using the relation (5.37) is the high accuracy of ineasurement of
Q,tp due to the differential scheme of ineasuremen*_s which eliminates the error in
_ the compensation of the cold~junctions of the thermocouples. Howev~r, in this case
~ it is necessary to know the Gk/G x OHTp ratio.
- Due to the absence of m~asurements of G3 in reactors not using shaped washers in the
fuel assemblies with a temperature control, it is assumed that Gj/G lOHT~J - 1�
Moreover, for a suff iciently representative sample of fuel assemblies with tempeta-
ture control, KM 1. In this case, the calculation of Kq by the relation (5.36)
is simplified considerably.
The maximum relative error Kq can be determined by the relation
( t! 1 ~
1S/? da I~ewx~ b~ t nx~ ~Kq_" ~
� b~k~] lU~ IvJ/vH01ITpI ~ 2-F- ~ 2 ~ (5.38)
~ ~ l - ~nx~leax~ ~ ~ - ~edx~~gx~
where ~IKa)~ s[G;/~%~;~~~~T,,], S[l p~X ~S(t,,,] ~ximum relative errors of the values of
Kq. G;/Gr.nnTp, ima~' lnx.
According to [124], the error in the determination of temperature with the aid of
chromel-copel thermocouplesused in WER-440 reactors is (without consideration for
the errors of the measuring scheme) 2.6�C.
With consideration for the errors of the measuring.circuits (0.3% for the tHx mea-
suring system on each of the loops; 0.9% for the t~~X measuring system with allow-
ance for an additional error of the two-step circuit for compensating the temperature
- of the cold junction of the thermor,ouple), the corresponding maximum relative errors
= are equal to ~~t""~ ~'25~~0 and b{t�7~ 1'~3~~0 . Due to the nonuniform mixing
of the coolant at the inlet to the reactor'which is responsible for the nonuniform-
ity of input temperatures, we assume that ~S~~na~=8{t�x]. � The value of ~r~~~;,~~�TPJ ,
- according to the available data, constitutes 2%.
The values of the maximum relative error S[Kq] for typical operating conditions of
the WER-440 reactor ~ a "t r~
f i -1r,~ c; ~ t~ 1~~~ - 2s,~ c)
~�ax
t-~
are given below.
li~. ~~Ti~. ~n. (1) 0,~ ~,G 0,8 I,0 1,1 1,2 1,3 l,4
g~/(~I, 32,3 20,9 15,~i 12,f 11,6 10,9 1Q,3 9,8
~ . KPy: 1. per unit value
53
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-04850R000500020007-7
FnR nFFI('1A1. USF, nN?.Y
It can be seen from these data that the error for the maximum values of Kq occurring
during the operation of reactors (1.25-1.35) is ~ 10%.
Permissible Power of a Reactor. The permissible power of a reactor is determined by
the absence of critical phenomena even on the most intense fuel element of the most
intense assembly of the core in a steady-state mode and in modes when the flow rate
of water in circuit I decreases with simultaneous activation of the safety system.
It is evident that the permissible power of a reactor can be increased if the non-
uniformity of energy release is reduced. Nonunifo~-mity of energy release in any as-
sembly of the core is characterized by the calculated coefficient kq~i, where i is
the number of the assembly in the core. The maximum value of the coefficient
"1z,t=~z~a"~ characterizes energy release in the most intense assembly.
The coeff icient of the nonuniformity of energy release of the most energy-intense
fuel elements 1c~qK~ can be represented in the form
\ (5.39)
kM�~o = ~ka~ i ku/MOKCkwez .
Since the coefficient kk is not monitored during the operation of a reactor, th~s
representation of k M~x ~ is not very convenient. Therefore it is usually assumed
that ;
k _ hNaKCkKkwex ~ ( 5 . ~F~~
MnItC Q
where kk calculated value of kk for an assembly with the maximum product (kQ,! hX)Ma,~c. ;
Thus, the critical power and the maximum permissible power of the reactor will be ~
def ined as
N~r = 1VK~'n~~/~zMa~cc~ ~5.41) .
Nnpcn�Au~ _ npell.llon ' I ~S.~F2~
P NK 1 ZW ~M9NC~
where nk the number of fuel assemblies in the core of the reactor.
The permissible power of the reactor must be reduced by a factor of assurance k.
Moreover, in order to ensure safe operation of the reactor, in addition to k3(~H 3.1t
the following additional factors of assurance are introduced: kg~N~-- allowing for
deviations in the real power of the reactor from the design power; k~(TP) - allow-
ing for the deviations of the power of the reactor from the prescribed level due to
variations in the loads of the turbogenerators; k 3~gx~ - allowing for the de-
gree of nonmixing of the water at the inlet to the core and the resulting macronon-
uniformity of the temperature field at the inlet to the core of the reactor.
54
FOR OFFIC[AL USE OIVLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7
F(1R (1FF1('IA1, ITti~ (1'VZ \
For WER-440, the above factors are taken to be: ~z~cN~-~,U'~; lz:~~Tr>=1,03; ka~oa>=1,01=1,U3
(the higher value is taken when the thermal power of the reactor is less than 50%
- of the rated power).
With considerati_on of all correction coefficients, the permissible rated power of
the reactor is
- Np�O'r",'' = Nl~pli~~/ItMaKCkaan~ (5.43)
where h;,;,,, = It1 ~~rn~n)ha (N)ka (Tr>>z;i (n~t)~Z the aggregate factor of as surance . For WER-440,
k 3 a~t = 1.36. The total correction coefficient in determining the permissible power
of the reactor with consideration for the mechanical coefficient k Mex = 1.1, is 1.5.
It should be mentioned that the permissible power of the reactor also depends on the
temperature (enthalpy) of the coolant at the inlet to the core. In calculating the
critical therma.l flow by formula (5.9), a proportionally greater channel power cor-
responds to the same critical flow (to the same output steam content) when the inlet
temperature (enthalpy) drops. The change in the permissible power of the reactor
when the inlet temperature (enthalpy) changes is determined by the relation~
L~Nnon - mCi~i~BG~ kn~arcr,kaa~t
kw, where G-- flow rate of-the coolant through the core of
' the reactor, kg/h; i-- change of the coolant enthalpy at the inlet to the core,
kcal/kg; coefficient allowing for the change in the f low rate of the coolant
in transient modes. The coefficient ~Q is selected as a ratio of the coolant flow
- rate during transient modes corresponding to the moment of the maximum reactor power-
- flow r~;te ratio~to the initial cool.ant flow rate. For VVER-440 with a regular GTsN
[main circulation pump] feed circuit, the worst transient (emergency) mode is the
loss of the productivity of two GTsN without a change in the power (see section 5.7).
_ In this case, ~ 0.7.
The increase of the permissible power of the reactor during the lowering of the in-
let temperature (enthalpy) of the coolant is limited by the critical heat f low at a
- zero steam content at the outlet from the channel with the highest thermal stress.
This flow must not be exceeded in any section of the heated channel (which, apart
- from the power of the channel, is determined by the distribution of energy release
along the channel). Moreover, the lowering of the inlet enthalpy is limited by the
undesirable lowering of the parameters of the second (steam-turbine) circuit, poten-
tialities of the heat exchangers (steam generators), and the growth of thermal
stresses in the reactor vessel.
The values of the above-mentioned factors of assurance are taken and introduced ar-
bitrarily in many respects due to insufficiently accurate measurements or calcula-
tions of the core parameters, Making them more accurate is a potential reserve for
increasing the power of WER which can be realized after conducting the necessary
complex of studies and implementation of a number of technical measures.
' COPYRIGHT: Atomizdat, 1979
'*When the permissible power of the reactor is determined by equation (5.42)
55
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00854R004500020007-7
~ F(1R (1~~1('IA1. 1 iC~ (1NZ Y
7.1. Arrangement of Fuel Assemblies in the Core
- The operating economy of an AES [atomic electric pac~Ter station] is determined to a
considerable degree by the effectiveness of the utilizatio� of nuclear fuel which is
characterized by the attained burnup fraction. It follows from the theoretical anal-
ysis conducted in the work [50] that thp burnup fraction for a prescribed level of
energy productior~ depends on the mode of fuel reloading. The evaluation of the ef-
fectiveness of various reloading modes is done by introducing the concept of an
ideal mode in which the depleated fuel is constantly replaced by fresh fuel with
_ constant mixing in the volume of the core so that the burnup fraction would be iden-
tical for a11 assemblies being unloaded.
The WER design makes it impossible to accomplish the ideal reloading mode, however,
by comparing the selected real mode with the ideal mode, it is possible to evaluate
its effectiveness and the degree to which it approaches the ideal mode.
Ideal Reloading Moae. The burnup fraction of the reloaded fuel in the ideal mode
can be determined by introducing the concept of the mean factor of neutron multipli-
cation in a core of infinite dimensions:
~Maxc pn+aKc
mm m~
/~R, ~Pmn~ = f k~ ~Puu~~dPmn .f ~?nm (7.1)
~ o
~axc
where fuel burnup fraction for the moment of time being examined; p~
~ maximum burnup fraction at which fuel is unZoaded from the reactor in the ideal
mode; k~ (P~u~ neutron multiplication constant for an infinite reactor at
the burnup fraction which corresponds with a sufficient degree of accuracy to
the asymptotic multiplication constant ko(p~~~ ) calculated by the programs POP and
UNIRASOS (see section 7.2).
The characteristics of k~(P~~~ ) for various degrees of enrichment of WER-440
fuel are given in Chapter 2. Ttiese characteristics in the absence of burnable poi-
sons are close to linear, and the rate of changes in k oo (p ) depends weakly on
the enrichment of the fuel at an identical uranium-water ratio. The nonlinearity
56
- FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2047/02/09: CIA-RDP82-00850R000504020007-7
FOR OFFICIAL USE ON~.Y
~ of the characteristics of k~(p ~u,n ) in the first approximation can be disre-
- garded, then ~ A'
~~ro ~1'IUJt~ _ ~l~ - !1r'LLIJir ~7 ~ 2~
where k oo multiplication constant of the hot fuel lattice poisoned at the initial
moment of power operation of the reactors; p,~~ amount of the slags at the mo-
ment of time being examined, kg/mU; A-- rate of changes in the characteristics of
k oo ~ p ,II,n ) ~ ~kg /mU) -1,
After substituting the linear relation (7.2) in the integral of expression (7.1),
we obtain
� hp, ~Pmn~ /Z~ - A(?"n~c~2. , (7 .3~
As was mentioned in section 2.1, the reactor is critical when the mean multiplica-
_ tion constant of an infinite lattice analogous to the core of the reactor in its
composition is equal to k'~�~'~ .
k~ IP~~~~~) = /i~~a~ = kro - Ap~,;; `/2.
_ . (7.4)
Hence, the maximum burnout fraction of the entire unloaded fuel is
(~uin ~ = 2 ~~Z~o - /Zi~oGx~/~. . S~
The value of A which is determined from the graphical re~3tions in Figure 2.6 is
- approximately equal to 0.01 (kg/mU)'1. For more accurate cal~~ulations of the value
Maxc
of P i~T~ , it is necessary to consider the nonlinearity of the character_�istics of
k,~ (pl~ ) which is represented in the form of polynomials. The values of the
burnup fraction attained in the ideal mode without con~ideration for the nonlinear-
ity of the characteristic of k~(P,u~ ) are given in Table 7.1.
r~axc Table 7.1
Burnup Fraction P~~~~ of the Fuel Being Unloaded in the ~deal Reloading.
Mode for Assemblies of the WER-440 Reactor for Various 235U Content
Kz/m U 10 32 I 50
, ~mn I
a~hU, ~o I 1,6 I 2.4 I 3,6
, Real or close-to-real fuel reloading modes are compared with the ideal modes by means
~ of the burnup fraction disadvantage factor k~:
fic - ~1un ~I Pu~n ~ ( ~ . 6 ~ .
where p,k~ finite burnup fraction of *_l:a ~:uel being unloaded in a mode dif-
ferent from the ideal.
57
� FOR OFFIC[AL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2047/02/09: CIA-RDP82-00850R000504020007-7
FOR OFFICIAL USE ONLY
Periodic Reloading Mode. The WER design calls for periodic fuel reloading on a
stopped reactor. It is evident that in order to increase the burnup fraction it is
desirable to reload the reactor more often, approaching the ideal mode. On the
other hand, frequent shutdowns of the reactor lower the production of electric en-
ergy and, therefore, are undesirable. Moreover, even with continuous reloading and
radial mixing of the fuel during the power operation of the reactor, the ideal burn-
up fraction cannot be attainedoWSefrom thenformula [50~]:of energy release along the
height of the core, which foll
Iz~ = ~k:11') ~ 1 -1!olHz) , ~7 .7)
~'~;~;u~~'Mn` nonuniformity factor of the burnup along the height of
where
p~� - average burnup
the core (index 0 refer~ to the center of the ~eactor)height, cm; Ho critical
in the assemblies being unloaded, kg/mU; H-
height of the reactor, cm, determined by the formula
(7 .8) .
~ _ 2 ~
f~p--7L~7(z p.
Here, x=.o-- xn - x: axial component of the material pa.rameter, cm'2;
z = 2 405/R)2
c,ir 2; xo -(k,r, n- 1) /.11~' n?aterial parameter of the core, cm'2; x~
radial component of the material parameter, cm'2; R~- core radius, cm.
The parameter Y(in formula (7.77) is equal to
MAI(C
/1 Pwn.~ ~7.9~
,Y _ Mn
o 1
2
where M2 migration area, cm2.
When kZ = 1.20 - 1,30, which can take place in a real burnup situation, the dis-
advantage factor~k~ = 1.10 1.20, i.e., even in the mode of continuous reloading
of the fuel assemblies with continuous rearrangement in the core (the fuel is not
mixed along the height) the burnup fraction is approximately 20% smaller than in the
ideal mode.
Under real operating conditions, the following periodic reloading modes can b~e accom-
plished,
Mode A. During each reloading, 1/n part of the core assemblies is replaced. In the
course of the entire operation time of the assemblies T(total operating ~eriod of
the fuel), n partial reloadinp~ are done. The assemblies in the core are not rear-
ranged during the reloadinga. The disadvantage facCor for this mode is estimated by
the formula
- k~ - 0 (1 l/n)~ (7.10)
where 9-- nonuniformity function of energy release over the volume of the core.
The value of 9 in the general case is determined by numerical methods. In the
58
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2047/02/09: CIA-RDP82-00850R000504020007-7
FOR OFFICIAL USE ONY.Y
extreme case, when the radius of the core is very large, or when the fuel assemblies
are continually rearranged, the value of 8 coincides with the value of k~ determined
by formula (7.7). In another extreme case, when the reactivity excess compensating _
the burnup is very small, A has a maximum value equal to 2.25, i,e., k~ex~�=2,25(1-I-1/n).
The maximum value of the disadvantage fa~tor (at n= 1) is 4.5.
Mode B. Part of the assemblies are rearranged during fuel reloading for equalizing
the breeding properties of the core over the radius. During the total operating
time of the fuel T, in each of the n reloadings of the core consisting of N assem-
blies, N/n of the most burned-out assemblies staying in the reactor during the T
time are unloaded. The same number of N/n fresh assemblies are loaded in the corea
The disadvantage factor for this mode (see work [50]) is
k~ - 20t/nB m, (7.11)
where 0z- (k,~~) (1--110/112) . The product nB cP is calculated by the formula
ii(3 ~p 0,415 [ti'~n - 2)z O,G92 8 (rt 1) - (n - 2) 0,831] . (7.12)
The above mode is accomplished only at n 7j 3.
Mode C. The fuel asaemblies in the core are not rearranged. Periodic reloading of
the assemblies is done in proportion to the radial component of the energy-release
field. The core consisting of N assemblies is divided into m circular subzones.
Each circular subzone is divided into sections containing Zi assemblies each (i
number of the subzone, 1~ i< m); the number of assemblies in each section of
- the subzone is the same, and the number of these sections is proportional to the num-
ber of reloadings in a given subzone. The portion of the assemblies being reloaded
, in each zone 1/ Ci is proportional to the radial component of the energy-release
field in the subzone. Then, the number of the assemblies being reloaded, for exam-
ple, in the first subzone can be found if the number of assemblies reloaded in the
i-th subzone is known;
N,ll, ~p~
~ (7.13)
Nr/~? ~Pt
where ~1, ~Qi average densities of the neutron fluxes in the first and the
i-th subzone; N1~ Ni number of assembliPS in the first and the i-th subzone.
From expression (7.13) it is possible to find the total number of the assemblies be-
ing reloaded N/n in the entire core: N/n
q~N
m ~ ' ~7. Z~F~
~ ~PtNt ~ (Ntl~t)
t -1 i-1
where CP average density of the neutron flux in the core; n-- number of reload-
ings during the run.
_ The disadvantage factor for this reloading mode is equal to
59 ~
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2047/02/09: CIA-RDP82-00850R000504020007-7
FOR OFFICIAL USE ONY.Y
k~ Oz (1 0,/n), (7.15)
where 0: (k= -1- ~)/2; 0, ~(k, -f- 1)I2; kZ nonuniformity factor af energy release
along the height of the core; kr nonuniformity factor of energy release along the
radius of the core.
Under real conditions, mode B is used most widely because it makes it possible to .
improve the distribution of energy release over the core and thus to reduce the dis- ;
advantage factor k~ ~.nd to increase the fuel burnup fraction. ~
a Selection of ~uel Loading iTnder Real Operating Conditions. When selecting the fuel
reloading mode under real conditions, it is necessary to consider the technical po-
tentialities of the WER. As a rule, the reloading of a reactor is done simultane-
ously with routine maintenance of the equipment of the AES unit which takes approxi-
mately one m~nth. The necessity of prolongzd shutdown of a reactor for reloading
lowers the utilization factor of its installed capacity, therefore, the number of
shutdowns for reloading must be minimal. The needs of the power system which exclude
possible shutdowns of AES units for reloading during the fall and winter energy load
maximums should be considered as an additional condition. The mode of operation of
~ES units with one shutdown of the reactor a year for reloading during the spring
floods when energy needs are satisfied by hydxoelectric power stations is the most
favorable for the energy system.
On the other hand~theenumberdofgthedreactorereloadingsfduringtabf llpfuelCOperating
which increases as
period increases.
_ With consideration for these conditions, the nsuresftheroeerationlofutherunitdbeg$
~ during the run was accepted for WER, which e p In this case as a
tween reloadings in the course of one year (see section 12.2). ~
_ rule, the number of the assemblies reloaded during each reloading time is close to
one third of all the core assemblies, but it can also deviate from this number in
individual units depending on the planned tasks of the AES as a whole.
In the first reactors WER-210 (unit I of the NVAES) and WER-70 (AES "Reinsberg",
GDR) the loaded assemblies were distributed evenly in the core [3, 51]. In the
new reactors V~TER-440 and V~1ER-1000 which are equipped with boron control systems
a zonal arrangement of the core is adopted, when the energy-release field is ad-
ditionally leveled along the radius.
The special characteristics of the zonal method of fuel arrangement are as follows:
1) fresh fuel is arranged only at the core periphery; 2) the burned out fuel is in
the central part of the core.
The fuel reloading mode with zonal arrangement is accomplished in the following ma.n-
ner: a) highly burned out fuel which remained in the reactor for three periods is
removed from the central part of the core (approximately one third of all the assem-
blies); b) assemblies from the periphery and the its adjacent area which rema.ined
in the reactor for one or two periods are moved to the central part of the core; c)
_ fresh fuel is loaded in the peripheral part of the core.
- 60
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2047/02/09: CIA-RDP82-00850R000504020007-7
b'OR OH'h'1(YA1. 1?1h: ONI.Y
Ttie zonal arrangemeTiL- of ttic core m~ikes it possible to achieve a considerable burnup
L-raction wtiich is d~ce ~o tlie lowering of the nonuniformity of radial energy release.
The loading of the fuel is determined by calculation according to special programs
on a computer, the following requirements being fulfilled: 1) ensurance of the neces-
sary length of reactor operation until the next reloading; 2) ensurance of the opera-
tion of the core at the rated power with minimum possible values of the nonuniform-
ity factors over the radius and volume of the core; 3) ensurance of the necessary
subcriticality of the stopped reactor in the cold state.
yy pg 2g Jp .iZ 34 36 38 40 42 44 46 48 50 51 54 5d 58 60 62
r~"--- i I .i Q~
11 , I v I~~ fl f l_ pZ
dZ I'S r. ' o ' I~ ~ 7`�oi ~ I z
_ - -
~3 i `\~,~,~`p t i - ' a~ p :i . . i9 ~ ~s ~ . f : ptF
05 ; ~ s, . ~ , e 1~ n ~a ~ 7 ~ ~ � ~ i. _ ~ ' OS
- 96 ~ ( ~ " ~ ~ , ~ . ~ - =
= ~ ' I
oi
DI ~ ~ - - . ; e - - ~s, 08
08 ~ n � : {t~t_,,~ ~ . s - - J9
99 ~ 9 . 4~ .4~s a ~ y . ~ _ _i~ i0
~0 I F'~ro . ; _ - . _ _ _ ~.t ` ~ z 11
_ ~ ; . ~ :
~ " N !rr ' N . :i.~ . I.. , ~c ' f ~~11 i:~_ � - - ~ Q ~2
' ~ i r _ .I. - ~1.~ : ~ _k - ~i ie d. C
~ZRI _ , z ; ~ ~13
:
~ 1~ 13 ~ r~. ~ . ~ a ~a . _ ;J~ ' ~ - r ~ 14
C ~
~~f 'i m ' , _ ~a : ~ 15
i5 ~ n :J. _~it ~ 'A' ~ ~ i_~ _ ; .
' ~ 6
16 ~n ' , ~'i ' ' , ~ ~ ~ ~ - - - - 17
; ~
j n o ~i. 'f.s t l8
18 ~ ;,f I,~ ~ k 6 - - ~ 19
19 . l ti ' ) _ ' , _ '
6 _ 20
20 il"P~ ~ 21
I - 4
Z~ ,I l , l i Ol,~qy~- � i ~ _ .~0~r- - - ~I- 22
23 ~ ~ i ~ , I '`J _ r--}~ - r I-~ - 23
I ~ ~ ~ I I I I I ~ ~ I ~ 1
Q
24 Z6 Z8 30 32 3~i ,~6 ,~8 40 42 44 46 48 50 52 54 5& 58 60 62
--1% ~-1,5% ~-Z% ~'-3,3%
~ ,
Figure 7.1. Cartogram of the first fuel loading of the WER-440
of the III unit of the NVAES with assemblies with various degrees
of fuel enrichment.
Key: 1. Loop
61
- FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2047/02/09: CIA-RDP82-00850R000504020007-7
_ FOR OFFICIAL USF. ONi.Y
24 26 1B JO ~2 34 dG 38 40 4Z 44 46 48 5~1 52 34 56 5d 60 6Z :f
' ~ ~ .
~ , ~ ~ - ~ , ,
01 ' pr~ � r ia ' Q�
Cl. 4~3 ..I ? n u n] ~c~,~' 0
3
OJ ~oF~ aa ;o os ~~a ar s~ is ~ v~ ~ Qyf 04
Oti i9 9 Idt ~ d1 9n o~ ~l 07 ' ~5
AS . St , t9 ~ d 74 N. ~'d 77 Td ' 7s dD ~ al �1 M' I I Os
i6 r - 9 I,y . y`�f - as r s~ : n~ i
~i . ~ n~
~i ~ ~ : ~a ~ - rs a , ~s s~ ~ sa .
ci s_l . ~s a s U~
(b Gr . ~i b - ~s - _ _ (~9
09 s~ ~ 6�s ie q ri r~ a s c ~ 10
.1~' a; _ . _ . ~ 9 ~
9V ~f / E !3 ~ ~ 'L~t ~ ~ ~
11 U IM a , y I~ r u- r~ , id ' IB i7 i 1~ 2 1~
IDR
11 ~1 0 .ia~ - i~._r i',~;, a..6 n e ~13
s~ I - d �1) : ~ ` i~i S 4i
i�_ Pd o.~ t~- i~1 -N~ 5 ! i t d r -f _ 14
. ~y ~ if - d~ ~iA ~ ~4 _ f~ i a r, uY a rs
15
15 p, jci T 6 ~s ~ s- 1~! i i
I ~ ~ 16
16 , - ia ' L,� " s - - ' _ - n ~ f
11
I7 - ~e - > >s e �a s ' ~ - . 18
= ~
. . . ~
!8 ; - ~ ~ ~r. 's ~ " _
_ - 19
19 n oa p~ a - ~ . . . ~ ~ - _ . 10
20 r{ ~ ' ~ _ - 2i
11 - I~P~ io a ' 9 f ~~i ! 6 12 �
12 - i ; ~ ~e~~P 23
z3 ~ , . ii11_ :
i ' ~
I ~ I~ r 56 58 6D . 62
14 26 1B d0 ,~Z ,~4 .~6 ,iB 40 42 4k ~r6 46 ~0 5~ ~f _
- ~-16% ~ -�24% ~-,~6�/t
,
Figure 7.2. Cartogram of the first fuel loading of the WER-440
of the IV unit of NVAES with assemblies with various degrees of
fuel enrichment.
1. Loop
Special requirements are imposed upon the first fuel loading. In order to imitate
in it a steady-state loading of the core containing partially burned out fuel main-
ly of the initial enrichment, assemblies with various degrees of enrichment are used.
By using assemblies with 3-4 enrichments in the first loading, it is possible to
achieve an acceptable nonuniformity of radial energy release ensuring the operation
of the reactor at the rated thermal power. During further reloadings, the assemblies
with the initial enrichment below the design enrichment are unloaded, and after 2-3
partial reloadings a mode is achieved when the reactor has assemblies only with the
initial design enrichment. Figures 7.1 and 7.2 show cartograms of the first load-
ings of the WER-~s40 at NVAES, and Table 7.2 shows their ~omposition.
62
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2047/02/09: CIA-RDP82-00850R000504020007-7
Table 7.2
Composition of the First Fuel I~oadings of WER-440 at NVAES
i~omivecrno Kaccer o nepooll Kom~vecr~ Kaccer n nepeoil
(3, ~1~ sarpy3Ke ` , ~1~ aarpy3ee
1 I I ~to~ t v CucoK 3 r ~ i~~~K I 6nex
OGoranlc. I 06oraule.
iiNe T~m- HHe ion-
_ (4) (S) M (4) (5)M ,"�'ii (4) ~5)m �
a" ~ u d = ca.~ a,
�-'~l m ~ s m m~
S ~ a ~Cu iq~r x~ ~0~~
. $ S S 1-~ v $ u S
L ~ u ~p u C u c
V
i u~ ~ u v
V C V V V
f1.Y F T Y dY Fo~'Y O.Y 1-~bW'Y O.Y F~X ~ .
1.0 24 25 2.4 - 78 49
1.5 48 l8 - 3.3 108 12
_ 1.6 - 114 24 3.6 - - 84 -
2.0 96 18 - -
Key: 1. Number of assemblies in the first loading
2. tTnit
3. . 235U fuel enrichment, %
4. Working assemblies
' S. Fuel parts of SUZ assemblies
i In WER-440 reactors at NVAES, Rol'skaya AES, and other AES, fuel of two composi-
tions 2.4 and 3.6% enrichment are used in the steady-state fuel makeup mode.
1
Cartograms of the first fuel loading af WER-1000 of the V unit at NVAES are shown
i in Figures 2.10 and 2.11. In the mode of a three-year fuel cycle, the core is made
~ up with a fuel with an initial enriclunent of 4.4%, and in the mode of a two-year
I cycle with an initial enrichmeat of 3.3%.
' As a rule, the selection of the first loadings o� a WER are chQCked and sometimes
I refined by phys~ical experiments on critical assemblies in the Institute of Nuclear
Energy imeni I. V. Kurchatov [24, 50, 52]. Such experiments make it possible to
correct the calculation programs used in selecting subsequent loadings of reactor
cores. The necessary experiments are also conducted during the startup and opera-
~ tion of WER. .
I .
; Physical experiments make it poasible to determine: 1) critical positions of groups
of SUZ assembliea and their effectivenesa; 2) boric acid effectiveness, which is
~ particularly important for WER-440 in which the necessary subcriticality of the
~ core is not ens::red without boric acid in the coolant of circuit I; 3) temperature
and power coefficiente of reactivity, which is neceasary for evaluating the self-
regulation of reactora; 4) power diatribution over the core assemblies; 5) 135Xe
and 149Sm tranaient proceasee.
The loading of the core of WER ie symmetric over the azimuth with an order of sym-
~ metry of not lesa than 3(sector of symmetry with a 120~ angle in the plane). This
makea it possible to calculate only one third part of the core when selecting the
next loading of the reactor. However, in practice, it is possible to have cases
when during the reloading of the fuel from the reactor, it is necessary for various
i
63
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2447/02/09: CIA-RDP82-44850R444544424447-7
N'UR UFI~I(:IAL USE UNLY
reaaons to unload from some sector of symmetry one or several assemblies which are
not sub~ect to replacement. In thia case, the symmetry of the loading cartogram
is disturbed, because the empty cell is loaded with a fresh assembly of the same
(or, perhaps, different) enrichment, or a burned-out~ assembly, but with a different
slag content. In such cases, it is neceseary to calculate the burnup of the entire
core (360� in the plane), or be content with averaging the properties of the assem-
blies which are symmetric with respect to their asimuthal position, but are not iden-
tical with respect to their neutron-physics characteristics. The asymmetry in slag
content can appear also in a symmetrically loaded core,ifin the course of the burnup
process the symmetrically arranged mechanical controls of the SUZ have different de-
grees of insertion into the reactor.
In conclusion, let us mention that at NVAES, where reactors with various specific ,
pawers are operating, it is possible, in principle, to increase the attainable burn-
up fraction by burning assemblies which are not completely depleted in a higher
power reactor in a reactor of a lower power [53], This is possible because the mul- ,
- tiplication constant of a burned-out assembly is increased when it is placed in a
reactor of a lower power resulting~from a partial release of th~ power and tempera-
ture effects of reactivity and a decrease in the effect of xenon poisoning. Since
an assembly reaches its maximal posaible burnup fraction iz a low.power reactor, it
is possible to unload a certain amount of not completely burned out assemblies from
a high~-power reactor.. In other words, it is possible to reload more than one third
asaembliea in a high-power reactor and thus to increase the operation time between
reloadings. The economical advantanges of this combined use of fuel are discussed
in more detail in section 12.4.
COPYRIGHT: Atomizdat, 1979
,
,
64
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7
FOR OFFICIAI, t I.SE nNI.Y
7.2. Calculation of the Neutron Physics Characteristics of a Reactor
As was mentioned before, the selection of fuel loading is done with the aid of spe-
cial calculation programs. The following basic programs are used in designing WER
and in predicting their characteristics in the process of operation*:
1) UI~IIRASOS (or ROR) a four-group program for calculating neutron-physics proper-.
ties of low-enrichment uranium-water fuel lattices and changes in these properties
~ as the nuclear fuel burns out [24, 54];
2) BIPR a single-group diffusion program for calculating a three-dimensional re-
actor as a whole [45];
~ 3) RAGU a one-dimensional four-group diffusion program for calculating the neu-
tron flux density on the external boundary of the core with a reflector and on the
surfaces of the absorbers of the SUZ assemblies;
4) KR a single-group program using the perturbation theory fos calculating the
reactivity factors of a reactor, the lifetime of prompt neutrons, the effective frac-
- tion of delayed neutrons, and the time constant of delayed neutrons (this program
works only in conjunction with the BIPR program) [25];
5) TWEL a program for calculating the temperature field and the margin until the
melting of the fuel and pressure of gaseous fission products in the fuel elements of
WER reactors [55];
6) GDKH a program for calculating the lnydrodynamic characteristics of fuel assem-
blies ; ~
7) RASKhOD a program for analyzing emergency situations of the reactor in the
case of a partial lowering of the coolant flow;
8) P-002 a program for solving the kinetic equations of the reactor with consider-
~ ation for automatic control.
*The programs mentioned here were prepared by the following members of the Institute
of Atomic Energy imeni I. V. Kurchatov: A. N. Novikov, V. D. Sidorenko, D. M. Petru-
nin, Ye. D. Belyayeva, V. S. Ionov, D. F. Strelkov, G. A. Bogachev, V. F. Ostashenko _
(deceased), Ye. V. Vinokhodov, V. D. Borisov= A. I. Belyayev, and others.
65
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7
FOFF-nFFI('lAl. IlfiF: (1N1.Y
- Uther programs lor some concrete problems of WER operation are also being used.
The following programs are used for calculating the distribution of energy release
fields within assemblies (for more detail see section 7.3):
1) ShESTIGRAI~TNIK [He~:ahedron] (or its variant TREUGOL'NIK [triangle]) a two-
group diffusion program for calculating the distribution of energy release in indivi-
dual fuel elements of hexahedral assemblies [46];
2) MIKRO a program for a simplified evaluation of specific thermal loads of in-
dividual fuel elements in an assembly (by the results of calculations using BIPR and
ShSTIGRANIVIK programs) .
Simultaneous combined use of several programs ma.de it possible to develop a method
for physical calculation of the core and a method of thermohydraulic and dynamic cal-
culations of reactors.
Let us examine the ideology of some of the,above-r~entioned programs used in the se-
lection of cartograms of core loading of WER.
UNIRASOS (ROR) Program. The four-group program ROR is used widely in designing WER
and in calculating regular fuel reloadings. This program makes it possible to deter-
mine the neutron-physics properties and the burnup of homogeneous low-enrichment fuel
lattices of WER [54]. The ROR program makes it possible to calculate lattices with i
1nca,1 nonuniformities, when a small aart of fuel rods is replaced with abosrbing ele-
ments (PEL), or when boron is contained in th~ jackets of ~the assemblies. It is pos- I
I
sible to estimate indirectly the effect of the boric acid dissolved in the coolant
of the first circuit. At the present time, this program is supplemented by a con- i
siderably improved program UNIRASOS. �
The UNIRASOS (universal calculation of states) progrdm makes it possible to calculate !
- with the aid of computer small-group constants and their changes in the process of ,
fuel burnup for a uniform fuel lattice (the presence of a small number of PEL is per-
mitted) composed of identical assemblies, cells, or fuel elements, i.e., for a homo-
genized lattice of a given composition and geometry. This program makes it possible
to caiculate a series of individual states of the lattice, obtain and store (on a
magnetic tape or on punched cards) the dependence of the neutron-physics constants
of the lattice on the temperature of the coolant and the fuel, density of the moder-
ator, power, and other parameters. The moderator can be in the f orm of a mixture of
- H20 and D20 and can contain a boric acid solution. The fuel is a mixture o� isotopes
from 231Pa to 244Cm of any composition which may also contain nonfissionable ele- .
Ments. The covering material of the fuel elements and PEL, assembly jackets, and
PEL rods may contain, besides 10B, 16 other burnable isotopes, including those with
strong isolated resonances.
The UNIRASOS program, just as ROR, is based on splitting the neutrons spectrum inico
four groups (see section 2.3). With an appropriate approach, the three upper groups
are combined into one group of epithermal neutrons. The f irst groups contains prac-'
, tically the entire fission spectrum. A noticeable portion of f ission occurs on
238U.� The lou~,er boundary of the group corresponds to ~he vanishing of the cross
section of ~~8U fission. The neutron spectrum within the second group is close to ~
66
FOIt OFFIC[AL U~E ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2047/02/09: CIA-RDP82-00850R000504020007-7
FOR (1FF1('L~1. IiSF' ON1.1'
the Fermi spectrum, ~ind elie resonance capture is weak. Resonances of the heavy
isotopes are concentrated almost entirely in the third group: 232Th, 233U, 235p~238U~
- 239pu~ 240pu~ 241pu~242pu, In the four~h (thermal) group, a large portion of the
fission takes place. In describing the processes in this group, the thermal motion
of the nuclei of the medium is taken into consideration (approaching of a heavy gas-
eous moderator) .
Within the above energy groups, effective group constants are practically constant
for all reasonable variants of the properties of the medium. In a homogenized medium
of a prescribed composition, the group constants are expressed in terms of effective
_ microscopic cross sections calculated in the P1-approxima.tion. The homogenizing of
the equ~valent cell is done for each of the four groups.
The distinctive characteristic of WER is the distribution of the fuel of a given
degree of enrichment over the assemblies whose dimensions are suff iciently great in
comparison with the characteristic path lengths of neutrons. Even in the hot state,
- the diffision length of thermal neutron is such (L ~ 3 cm) that the dimentions of
. the assemblies are 4-S L. Moreover, the dimensions of the assemblies are much
greater than the moderation length. Therefore, eacti assembly can be considered to.
be isolated both with respect to thermal neutrons, and with respect to epithermal
neutrons. When the enrichment is small, the spectrum of epithermal neutrons depends
weakly on the enrichment, and it is taken to be identical in all assemblies. In all
variants of the geometry of the lattice composition occurring in practice, the migra-
tion area M2 even in the hot state is of the order of 60-80 cm2, i.e., M= 8 = 9
cm. Since the dimensions of the assemblies are 2-3 M, it is sufficient to limit one-
- self to the P1-approximation for determining the angular dependence of the neutron
flux density.
With all of the above-mentioned characteristics of the fuel lattice, it is possible
to consider that in the greater part of the assembly an approximation holds true at
which the spatial distribution of the neutron flux density is identical for all en-
ergy groups. The length of the fuel elements of WER is many times greater than
their diameter, therefore, the conditions of the diffusion of neutrons along the ax-
is of the rods and at right angles to them are different. The following system of
equations is solved in the UNIRASOS program with consideration for the neutron dif-
fusior: anisotropy:
D:i ~z m~ Dnr BR -I- ~Pi = k~ ( ~i vl~i ml~ -I~
E~'" ~~Pr-~ ; i- i, 2, 3, 4, .16)
where DZi, DRi coefficients of diffusion along and across the axis of a fuel ele-
ment (by groups); BZ, B~ size-shape factor along and across the axis of a fuel
- element; ~i group density of the neutron ~lux; i-- macrosc~pic absorption
- cross section for each group of neutrons; ~ iB macroscopic cross ~ection of the
removal of neutrons from a given group to the next group; ~ i-- macroscopi~ f ission
67
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7
FOR OFFICIAL U~E ONLY
cross section by groups; v~ number of neutrons produced in one fission event;
~(,i share 4f fission neutrons per given group; k~~ effective multiplication
constant of the lattice; i(j) neutron group index.
By solving this system at prescribed values of BZ and BR, it is possible to deter-
mine k3~ and vice versa, if we take k 3~ = 1, it is possible to determine the size-
shape factor in the critical state. The system of equations (7.16) ma.kes it possible
to determine the neutron-physics characteristics of the fuel lattice also in a single-
group and a two-group approximations. The determination of the heterogeneous effects
is accomplished during the homogenizing of the equivalent cell.
The UNIRASOS program has a number of innovations in comparison with the ROR program
which make it possible to take into account a number of effects more accurately and
expand the potentialities of the program.
A list of fission products which are taken into consideration is given below in the
order adopted in the standard variant of the UNIRASOS program:
i-4. B~Kr-BSKr 51--52. 12aXc-~~Xe
5. "SRb 128'CC
6--7. "8Sr-�iSr 5,}. ~xsi
- 8. ~[~r �`'i5. ~~'Xe
g. e7Rb 56. i:soTe
]0. 88Sr 57-58. iaiXe-~asXe
~ ~ evy 59-60. ~~'Cs-'~Cs '
I'l. 90Zr 61. 13dXe
13. 0�Sr 6'l. 1J5Cs
- 14-17. '~7.r-"~Zr fi3-64. '~aBa-~'rf3a ,
18-19. 95Mo-98Mo GS. "~Xe
~ 20. ~'l.r 6fi. ~s~Cs
2 I-`L2. 97 MO -9BM0 67. I:Iq ~ ~ ~
23. 99"['c 68. '~OLa
ioo 69. ~~oCe
24. Ru
25. 10�Mo 70. Pr
2t'~-27. ~o~Ru-~ozRu 7l. ~,zNd
- ?8. iosRh ~ 72. i~zC~
29. ~o,['d 73-76. ~~'Nd--1/eNd
30. ~o~RU 77. ~aPm
31--32. ~osPd--106I'd 78. ~~eStn
33. ia;Rtt 79. t~e~~ ~
:i4 --35. to~p~__inepd ~0-8i. i~9Sui-jOSni
36. ~o,q~ 82. ~SONd
.}~--~n. ~~~c~i--~~~~~ R3-84. isiSm--issSm
41. '~'~"I,n 85--8G. isal:u-is~[:i~
,}1-43. iisSn--~nSn 87. ~''~Sm
- 44 -4(i. i~a�~�~._iss�~�~ 88. ~ssEu '
47. ~s~tin 89-90. ~ss~~{-.~s~Cd
aK. ~z6st~ ~i. ~59~r~,
- ,~g isc�fc 9'l-9r~ ~copy-~sspY
- ~ill. ixt~
BIPR Program. This program is the basic program for the three-dimensional calcula-
tion of the WER core. BIPR makes it possible to determine the reserves and reac-
_ tivity factors of the core, to perform calculatio.zs of the integral and differential
effectiveness of ehe control elements, to determine critical positions of the con-
trol elempnts and critical values of boron concentrations in the coolant, to obtain
the three-dimensional distribution of the energy release fields in the core, and to
calculate the fuel burnup in the core and 135Xe and 149Sm transient processes.
_ 68
~OR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2047/02/09: CIA-RDP82-00850R000504020007-7
~1R f1F!'ICI,~I, 1~~~ (1N1,1'
In tl~e 13IPh pro~;r~iu?, ~l~e reactor core Ls represented in the plan by several symmetric
sectors in each of which the arrangement of assemblies with identical calculated
neutron-physics properties and the same design number n repeats symmetrically (1 1
- n G 117 for three syRUnetry sectors for WER-440).
This program makes it possible to calculate symmetry sectors with anglea of 30, 60,
120 and 360 degrees. Due to the presence of symmetXy, calculations are done for
only one sector, which substantially simplifies and speeds up the calculation of the
loading cartograms. Moreover, the "joining" conditions of the neutron flux densities
on the inner boundaries of the symmetry sector are observed closely, i.e., the re-
sults of calculations are true for the entire core.
For mathematical description of physical processes, the real core in which the assem-
blies are arranged in plan over a triangular lattice is represented by a mathema.ti-
cal modzl, where continuous changes in the properties in the core volume are replaced
by discrete changes over the points in which all physical properties averaged over
*he cross section of the assemblies are concentrated. Along the height of the assem-
blies; m cross sections (points, 1~ m~ 10) are taken and, thus, the core is re-
preser,�.ted in the form of a spatial lattice. At each lattice point with coordinates
_ (n, m), a number of characteristics changing in time during the burnup are detera?ined,
breeding properties (k ao ) depending on the kind of fuel, local power effects, ef-
fects of poisoning by samarium and xenon, fuel burnup depth, neutron flux density
field, energy release field, etc.
The working assemblies are represented as a stationary spatial lattice of points,
the real lattice of SUZ assemblies is represented as a mobile spatial lattice of
points. The movement of control elements is taken to b~e discrete with the movement
pitch ~ hg~ equal to the distance "~etween the points along the height. In the
- process of the calculation of one s~:~te, the points of the mobile and stationary lat-
tices always coincide. Unlike the points of the stationary lattice, the points of
the mobile lattice c~n have either the properties of absorbers, or of the fuel, de-
pending on the degree of the withdrawal of the SUZ assembly from the core.
The simulation of the process of power control by a liquid absorber is also accepted
to be discrete ( Q cg).
The initial prerequisites of the BIPR program are as follows.
~ 1. The distribution of the neutron flux density over an inhomogeneous core is found
from the solution of the following one-group equation:
c1~~ ~r) -I- x2 ~P ~r) = 0. (7 .17)
(Transition to discreteness is accomplished by replacing the radius vector r of the
lattice point under consideration by its coordinates n, m). This equation can be
reduced to the Poisson equation or to the diffusion equation (depending on the state-
ment of the problem). It is solved by the iteration method according to the finite-
difference scheme in a nine-point spatial lattice around each point.
- Boundary conditions are given on the external boundaries of the core and on the sur-
faces of the SUZ assemblies:
69
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2047/02/09: CIA-RDP82-00850R000504020007-7
~OR OFFICIAL USE ONLY
i dT I .
q~ dl d,~ ' (7.18)
where d3~ reciprocal value of the effective logarithmic derivative of the neu-
tron flux density taken along the normal ro external boundary of the region being
examined; 1-- direction of the normal.
2, The values of k oo and MZ are calculated by the UNIRASOS (or ROR) program by the
~ormulas ~
~ v:.�~ cr:, ~(c. r3o) ~r:
~ ; (7.19)
( ~r.�) f (e. d~
MZ _ v(L�') f(1i , rt~) dE (7 . 20) .
~~;.,.,~~L~E~~. ~~o)Jc
3. It is taken (and is confirmed experimentally) that all assemblies in the reactor
are uniform with respect to their moderating properties, resonance capture and 238U
, multiplication; M2 ~'L and is constant over the core.
4. Material parameter '
xz (r~ k~~i~) ( kk~r) - 1) / M', (7 . 21)
where M2 the value of square migration Zength averaged over the core; k oo(r)
multiplication factor of~an infinite lattice with fuel of a given grade E; k 3~
effective multiplication factor.
S. The multiplica~ion factor k oo (r) at the points has its own definit~ value which
depends on: a) enrichment of the fuel (grade E) in the assemblies; b) fuel burnup
fraction at a given point p~JJ ; c) density Y and temperature of the moderator-
coolant t; d) boran concentrati.on in the coolant cg; e) power of the asaembly
f~ 135Xe and 149Sm poisoning, respectively, ~e, pSm�
- Thus, k oo (r) is a composite function of many variables:
k~ (r) = k~ (n, ~n) _ (1 -1- ako ~c~ - Ak,,,,, ~n,m> -
- Aksrn (~~.?n> - ~k~p ~n.n,>l nB ne ~ nx~ � ~7 .22)
Here Oka~~%~--h~c~.~~.>~ ~ excess multiplication factor of a fresh, unpoisoned fuel
_ lattice of grade E without power; ~~,,,,,~,,,m> slagging effect; ,~(;5m(n,m~
149Sm poisoning effect; n~i+DCn.+~~> power effect; n.~e 135Xe poisoning
effect; n~t temperature effect; TI g-- effect of poisoning by 10B dissolved
in the coolant-moderator.
70
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2047/02/09: CIA-RDP82-00850R000504020007-7
FOR OPFI('IAL l1,CE ~1NI.Y
Neutron flux density ~(r) depends on the same parameters as k co (r);
y, ~P (n~ nt~ E, Pmn~ Ptim . pX~ ~ At, CB~ .
6. In the BIPR program, the following defi~iition of relative energy release is
taken with a certain approximation;
(n, m) = h~ (~)'P ~r) = h~ (n~ m) m m). (7 . 23 )
7. It is assumed that slag formation is proportional to energy release in the seg-
ment of time when the characteristics of the reactor are practically constant:
T~
nP~~�~ ~T) k~~~n ~ ~~xlT. (7.24)
Ti '
Here, k 1~ _(0.40=0.45) 10'3 kg slag�1/(t�kw�day) conversion factor.
8. Nonsteady-state poisoning by xenon and samarium is taken iato consileration.
A special control unit simulates the control of the reactor during the movement of
the SUZ assemblies, which shows on the changes of the curvature of the neutron field
CQ (r) with the changes in the concentration of boric acid allowed for in the
value of the material parameter X 2 and in combined use of boCh control methods.
The BIPR program delivers the following information in the process of computations:
1) input characteristics af the core as a whole, as well as by the t;~?pes of working
assemblies and SUZ elements; 2) current moment of time T, eff. days; 3) average slag
content in the core p~~~ , kg slag/t U; 4); k 3~ of the core; 5) nonuniformity fac-
tor of volume energy release/r~a"~ -(~~m,n)~,~~~~~V'~ 6) number of the calculated point with
k
VX ; 7) nonuniformity factor of energy release of the assemblies ~
hoa"` _((~~~dz) M,K~/oll'I'~1z; ; 8) number of the assembly with k
qX; 9) boron concentra-
tion cB, gB/kg H2O; 10) withdrawal height of the working group ~f SUZ assemblies, cm;
11) number of the assembly in the operating group of SUZ; 12) reactivity in the given
state p= 1-1/k 3fi ; 13) average content of slags pn, of energy release kqn, and
heating ~ tn over the assemblies; 14) distribution of the slag content pm~n, of neu-
tron flux density ~P m~n, and energy release 1~J m~n over the entire volume of the
core; 15) concentration matrix of slag for all assemblies at ten points along the
height; 16) concentration matrix of 149Sm for all assemblies at ten points along the
height; 17) concentration matrix of 149pm for all assemblies at ten points along the
- height.
In spite of the simplicity of its mathematical model, the BIPR program makes it pos-
sible to simulate witi~ practically acceptable accuracy the work of a reactor in time
and to obtain the necessary characteristics of the fuel burnup process in the core.
RAGU Program. This program makes it possible to determine the following in a small-
group (up to four groups) diffusion approximation or P1-approximation: 1) effective
71
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00854R004500020007-7
FOR Oh'FI('IAL USl: UNLY
_ multiplication factor af a multizonal (up to 10 zones) medium�; 2) neutron flux den-
sity as a function of the coordinate r; 3) logarithmic derivative of the neutron
flux density on the boundary surface of the SUZ absorber with the surrounding fuel
part of the zone or the core with the reflector.
In order to perform calculations on SUZ absorbers, conversion is performed from a
real hexahedral geometry to an equivalent cylindrical cell. The values of k 3~ ,
Br, dlog are determined, after which inverse conversion to the real geometry is per-
formed, and single-group effective reciprocal value of the logarithmic derivative
d 3~, suitable for using in the BIPR program is delivered.
Analogous operations are performed for determining the effective logarithmic deriva-
tive for a Uoundary with a reflector, but the entire core in this case is represented
in the form of one equivalent cell.
The following is the initial system of equations in the program:
I ~ ~~a ~i~ ~o ~po = ~o~ -f - So ;
r�C dr ~
(7.25)
d ~ �
1 Ei ~Di = f tj 'i' S~
3 dr
with boundary conditions for the core ~
7.26)
~ ~OIr=O - ~OIr=R - ~ ~
and for the absorber cell
D~pl~-p = ~i ~~OIr=R - ~ ~7.27~
and the conditions of joining at the boundaries of the zones:
r
~pi = ~i+i ~ Di ~~Pi - ~i�?-i ~Wi+i � (7.28) .
Here, i-- zone index; j-- number of the neutron group; Ql geometry parameter
(plane, cylinder, sphere); r-- generalized coordinate; R-- size of the absorber
cell or the core; neutron flux density; ~
~o = ~a -I- ~oyH+i ~ E~ _ ~ir EyBJ+i ~
S~ and Si isotropic and anisotropic sources of ne.utrons;
~pf k~ Q F' ~oyn;/ t+ r . 29~
II~ ~ `j;yAM~(P~-'~ Q =~(y~~)" ~Q
T
(the same symbols as used earlier).
72
FOR OFF7CIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2047/02/09: CIA-RDP82-00850R000504020007-7
FOR OF FICIAL USE ONLY ~
The initial constants for the fuel are found by the L]NIRASOS (or ROR), and for other
ma.terials by the RAGU program. The RAGU program has been checked on a considerable
amount of experimental material and showed that, in principle, it is applicable.
KR Program. The input constantsfor this program are calculated by the i1NIRASOS (or
ROR) and RAGU programs. The necessary data on three-dimensional density distribution
of a single-group neutron flux in the core volume are taken from the BIPR progra~n
which operates jointly with the KR program.
The expression for the reactivity factor is represented in the form
z ~ dh12 \ , zdV
_ r t dx ~~zdV - (1 - k J ~ + 7 .30
dx ~ ka,p ) dx Mz dx ~ ~ ~
~ d (~t.,,~~) ~~z~~ (~~~M2~ k~ cp2dVl-~ , '
f z ~X ~ , ~
.S d.N~~ y
where y~ 2-- material parameter of a homogeneous lattice; MZ average (indepen-
- dent of the coordinates) area of neutron migration; d~-- effective inverse value
of the logarithmic derivative (for a single-group flux~, x-- used to determine the
changes in the reactivity factor (change in the coolant density, coolant temperature
without changes in the density, average fuel temperature, coolant temperature includ-~
ing changes in the dansity and changes in the reactor power).
Other cha:acteristics calculated by the KR program, their expression in terms of the
determined par3meters of the reactor, their caclLlation and representation in thE
final form are given in work [25].
COPYRIGHT: Atomizdat, 1979
73
- . FOR OFFICiAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7
FOR OFFICIAL USE ONLY
'
~
_ ~
Appendix. Example of the Calculation of Loading (Reloading) of Fuel in the WER-440
Reactor
The neutron energy spectrum averaged over th~ ~ross section of the assembly is found
to be close to the asymptotic spectrum and is determined, primarily, by the proper-
ties of the assembly itself, which makes it possible to separate the problem ot the
calculation of small-~roup cross sections of fuel lattices and changes in the iso-
topic composition during th~ burnup of fuel from the calculation of the reactor as
a whole. Thus, the final results of calculations by the UNIRASOS (ROR) program are
the input parameters for the BIPR program.
Let us illustrate the selection of the cartogram of the loadirig (reloading) of nucle- ~
ar fuel on an example of the calculation of the first loading of the WER-440 of the ,
IV unit of NVAES.
The three-dimensional calculation of the r~actor is performed in the following se-
quence: ~ .
1. Calculation of the neutron-physics characteristics of the fuel assemblies to be
used in loading, by the UNIRASOS (ROR) program.
2. Calculation of the reciprocal values of the logarithmic derivatives of the neu-
tron flux density on the boundary of the core with a reflector and on the boundary
of the absorbers of the SUZ with fuel by the RAGU program.
3. Calculation of the subcriticality (supercriticality) of the core in the cold
state by the BIPR program.
4. Calculation of the distributions of the densities of neutron fluxes and energy
- releases, as We1i as of the fuel burnup over the volume of the core by the BIPR
program. ~
5. Calculation of the effectiveness of the control elements by the BIPR program.
6. Calculation of the coefficients of the reactivity of the reactor by the KR
and BIPR programs.
7. Thermohydraulic analysis of permissible operati.on modes by the GDKh, RASKhOD,
ShESTIGRANNIK, TWEL programs.
74
FOR OFFICIAL USE OI~iLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7
F'UR uF F1C'!AL U~E: ONLl'
Let us examine the process of calculations step by step.
1. Fuel assemblies of three degrees of enrichment were used in the reactor: 1.6,
- 2.4, and 3.6% (see Table 7.2). Consequently, it is necessary to perform.three com-
- piete calculations of the burnup of such fuel lattices, as well as several short
calculations of individual state for each fuel lattice (see Section 2.3) for vari-
ous values of specific power, temperature of the coolant-moderator, and the boron
concentration in it.
Calculations were done by the ROR program. The input parameter for this program are
given below on the example of a fuel lattice of 3.6% enrichment.
kr = 0.866 coefficient of the geometry of a hexagonal lattice;
hT - 14,7 cm assembly spacing with consideration for the water gap;
dT ~ 0.775 cm diameter of a fuel rod in a fuel element with con-
sideration for the gas gap;
S ~T = 0.065 cm wall thickness of the fuel element jackets;
~ k= 14.4 cm "box wrench" dimensions of the assembly;
~k = 0.15 cm thickness of assembly wall;
n� 126 number of fuel elements in the assembly;
dUp2 - 0.755 cm diameter of the fuel pellet;
YU02 " 10.2 g/cm3 uranium dioxide density in the pellet;
TH2O ~ 293 degrees K-- temperature of the moderator (for an example,
the cold state of the fuel lattice i.s taken) ;
~H2O = 1.0 g/cm3 density of the moderator at Tg2p;
mZr ~ 2-- index indicating the material of the wa11 of the fuel element
_ and the assembly (Zr);
TUp2 = 293 degrees K-- average temperature of the fuel (cold etate);
mUp2 = 0-- index indicating the type of fuel in the fuel element (U02);
wXe = 0.064 parameter indicating 135Xe poisoning;
WSm = 0.011 parameter indicating 149Sm poisoning;
w= 84.3 kw/1 specific power in the volume of the core; �
p25 = 0.036 initial concentration of 235U;
p28 = 0,964 initial concentration of 238U;
Pi = 0-- initial concentration of 236U, 239pu~ 240pu~ 241pu~ 237Np~
lOg~ 242pu, and slag.
The remaining cons~ants are tabular in nature and are not given here.
. The dynamics of the fuel burnup is described by the following system of equations
~PtIdT =St-~~PPt-~tPt~ (7.35)
where pi concentration of the i-th isotope; T-- time; Si source of the appear-
ance of the i-th isotope (from the preceding isotope as a~esult of neutron capture
or radioactive decay of some isotope); d~ one-group absorption cross section of
the i-th isotope; {Q neutron flux density; ,)Li radioactive decay constant,
Fission products are combined into a fictitious isotope (slags) with changes in the
absorption cross section according to the 1/v law in the thermal region and having
a resonar~ce integral equal to 200.barns. Slags do not include isotopes 135Xe and
- 75
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7
FOR OFFICIAI. USE ONLY
149Sm whict~ are allowed for separately. The initial data are transferred to punched
cards and are introduced into the computer. After completing machine cQmputations,
extensive output information is printed. The results of computations are shown
graphically in Figures 2.5-2 9. .
- The next step is the processing of the obtained results in a form convenient for the
BIPR prgram. For this purpose, the dependence of the multiplication constant of the
fuel lattice of grade E on various effects of reactivity [see formula (7.22)] is ap-
proximated as follows: .
8
OkWn = ~j QiE Pwn ~ (7.3d)
6
. nk~ _ ~ bIE (7.35)
~-i
7.36
:1kSm eSm~t Sm' ~ )
r~Xe I -~-eXe,E X~~ (7.37)
nsr I_~ d,~ ~1 dt~ ~t_; (7.38)
~ -F. ~s~.~8 VH,o , (7.39)
nB i -t- c,~ B ~vH,o
Here, P~~ slag concentration in the fuel lattice� w-- power at which the fuel
lattice works; Sm, Xe concentration of 149Sm and 1~35}Ce nuclei in the fucoolantice;
a t-- heating of the coolant over the length c~f the assembly; 1~H20
density (depends on the temperature); cg boron concentration in the coolant-mo-
derator; ai, b~, cl, c2, dl, d2, ege, eSm aPProxima.tion coefficients (Table 7.6).
2, The determination of logarithmic derivatives is particular in nature and is dnne .
one time for each reactor. Results for the cold and hot states of WER are given
on page 41, as well as below.
3. For the three-dimensional calculation of fuel burnup in the core: the symmetry
sector of 120 degrees containing 117 assemblies. Preparations for computations
according to the BIPR program begin with the selection of an arrangement scheme of
assemblies in the core with various initial fuel enrichment levels and various de-
grees oi burnup fractions. The initial prerequisites and the ideology~of combining
fuel assemblies are discussed in section 7.1. For this case, the loading cartograms
are shown in Figure 7.2, its composition is given in Table 7.2.
Initial data for the BIPR program are given below:
.M2 ~ 64.5 cm2 migration area;
hr ~ 14.7 cm distance between centers of assemblies (in plane);
m~ 10 number. of calculation point~ along the height;
H~ 250 cm core height;
- V 3~ 16,270 1-- core volume;
2c~k = 17,36 cm; doubled reciprocal values of logarithmic derivatives for
2dg ~ 19.96 cm the radial and the end reflectors;
76
FOR OFFtCIAL USE ONLY �
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00854R004500020007-7
FOR OFFICIAL USE ONLY
Table 7.6
Approximation Coefficients for Fuel Lattices of WER-440___
/2\ ~HA9CHNC K03~HI~HCIiTB I1pN OG0~8I11,tHHH TOi1THB9. ~/y~
I \ l I .
~ y(o~x~wcnr z. G I. 2. 4 1.6
f11
Ak -}-0,341800000 -{�0,248550000 -}-0,137180000
Q~ � -0,9228016G0� 10-z -0.680990342� 10-' -I-0,~~9283� 10-=
u� -0 , 88.~545897 � 10-' -0 , 209788752 � 10-' -0 , 543949452 � 10-s
n.~ -{-0, I'15842000 � 10-' -}-0, 3G572G45G � 10-3 -f -0 ,112442249 � 10-~
Q~ -0,8;39037485� 10-6 -Q.32893G284� 10-~ -0,1313853fi7� 10-~
n., -}-0.32273G125�10-6 -~_0,175778354�10-6 -~-0,934A33fi58�10-6
~a -0,7337810(il � IQ~-" -0,5620230$2� 10-4 -U,400432270� 10-~
n~ -{�0,919482135� 10--~0 -{-0,993~i92319~ 10-9 -{-0,949278745� 10-R
R p -0 , 489('~8971 h� 10-1z -p , 7454.53 I I I� 10-11 15~1(i3000(1 � 10-~~
b -0,1707500(H1� 10-3 --0, ifi8030000� 10-3 0.25203000b� 10-~
bs -}-0,30804(~00� 10-v -}.0,3.3G040000� l0-' +0.000O00000
h; U, 00000(x100 0, 000!)OQ000 ~~~p~00(~0
ti O,b00p000UU O,OOOOq0000 0,00000000(1
b, . 0, 00(~00()00
- he 0,0(i0000000 O,OOOOOOOUO
� c -0,737fi000(x)�10-t __p,~nlG8cxf00 -0.138I30000
r~ ~�0 , 40~80000(1 � 10-a 0,G88100000 ~ 10-a -~-0,1120900~x) � 10-~
d ~ -U , ~Si41 J00(xf � 10 -0 , 7~i9350b00 � l U-~ -0 .479170000 � 10-~ .
/t2 -n,l!~i~~x)� ~l)--M1 .--~,1421x~~' ~~-~6 ~G~~~~� ~~-6
eX~ --(1,1 I 5(H1~000 � 10-F ~ --0 ,18~i000000 � 1 W-1 -0 , 2G5f~'~000 � 10+1
- ~Sm -0 ,12 l (H3000U � 10--1 -0 , I G8000000 � 10-1 .-p , 218000(1100 ~ 10-~
Key: 1. Coefficient �
2. Coefficient value for fuel enrichment, % .
2dr s 14.86 cm; the same for the lateral side and the end part
2dr = 37.4 cm of the controls;
1 4 ' 10 ~
n5 ; ng s 11, number of.assemblies in each of the 12 horizontal
ng = nlp = 10, rows of the cartogram in the symmetry sector of .
, nll = 9 ; n12 = 4 120 degrees ~i~ _ ~ _
i
hXjj 125 cm height af the initial withdrawal of the working group
of the regulating rods (rods No 1, 7, 68);
a Cg ~ 0.01 gV/kg H20 step of the change in the boron concentration
in the coolant during burnup;
gi m 1.0 relative flow rates of the coolant through the assemblies
(117 numbers);
- Q a 39,000 m3/h total flow rate of the coolant through the core;
'~~H2O = 0.75�103 kg/m3 average coolant density in the core;
cp s 1,249 kcal/(kg�degrees) average specific heat of the coolant
at a constant pressure;
~~pm = 0.011 149pm fission yield;
Ysm ~ p__ 149Sm fission yield;
pm = 0.357�10-5 sec-1 149pm radioactive decay constant;
77
FOR OFFLCIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2047/02/09: CIA-RDP82-00850R000504020007-7
FOR f?FF'ICIAL USE ONLY
�tiM. ~ ~
-O.202�10'~ barns
~cm. 2 "
~~3;i7.~pR barns effective microscopic cross sections of 1 9Sm ~
~`m z- capture in the lattices 3.6, 2.4, 1.6%;
- i?,4ti4� 106 barns ~ .
pU o 40.91 t-- uranium charge in the reactor;
~U02 � 10�2 t~m3 uranium dioxide density;
NT = 1375�103 kw total thermal power of the reactor;
~ T~ S days burnup time spacing ;
Ako, t-., atF, b~P,
c,,z, r�., ~~.2.F. approximation coefficients for a fuel lattice of grade E:
C~m, F.. PXr, 1.6, 2.4, 3.6% (values are taken from Table 7.6) ;
T= 0 effective days beginning of burnup.
It is also necessary to give the values of 149Pm, 149Sm and slag concentrations in
all 1170 points of the symmetry sector of the core. If the buT ePinitialrparameters
poisoned core is calculated, these values are equal to zero.
are introduced into the computer andlsue~inteduoutnat definitemtime intervaZslsdisl
culated. The following information p
= tributiun of slags, relative enalculatedasasawelleasrthefaveragesdistributionhofen-
tire volume of the core being c ~
slags, energy releases and heating over the assemblies in the form of cartograms or
the symmetry sector. For a typical case, some results of calculations are shown
graphically in Figure 4.1. The diagrams reflect the peculiarities of conductingogi-
the burnup mode: first, with thkinld rou orandathenlafter thelwithdrawal ofXboron
tion of the controls of the wor g S P~
from the reactor, with the aid of SUZ assemblies.
The following are some characteristics of the burnup mode prinr.ed by the ~~overethe
before the end of the boron regulation mode: TX~ ?1~447fattaVpo ntsn,m~ 41.3; k
qX
core) = 8. 1 4 6 k g s l a g/ t U; k 3 = 0 90905 k~.k H 0.
e1.259 in an assembly of n' 4~; cg g g 2
On completion of computations, the computer pri149d comIiavin lsupplemented the ini-
149Sm and Pm.
core and Funched networks for slags,
tial data with some constants, it is p4ssible to calculate 1 SXe transient proceases
when the powex rises or lowers.
If the computed characteristics of the burnup in tea CberinningtwithPthetchangerind
are unsatisfactory, the entire procedure is repea , S
the loading cartogram.
4. Due to the fact that the temperature effect of WER reacti:vity is negat'iv~intsof
section 3.2), the cold state of the reactor is the most dangerous from the p
nuclear safety. :The determination of the subcriticalit.y (supercriticality} in the
78 �
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-04850R000500020007-7
FOR OFFICIAL USE ONLY
cold state is particularly important for VVER-440 in which the total reactivity ex-
cess of the loaded (reloaded) core (see section 3.3) is sufficiently great and does
not make it possible to operate it without boric acid in the coolant. Subcritical-
ity (supercriticality) of the selected load is calculated by the BIPR program as a
computation of one state of the reactor with initiat data corresponding to the cold
state. ~
The following results were obtained for tre case in question for the cold state:
keff - 1.2161 all upper SUZ assemblies, cg ~ 0; keff ' 1.03214 aIl lower SUZ
assemblies, cg = 0; keff = 10000-- a11 lower SUZ assemblies, cg ~ 0.174 g V/kg H2O;
keff ' 10000-- all upper SUZ assemblies, cg ~ 1,347 g V/kg H2O; keff = 0.9221 -
all upper SUZ assemblies, cg = 2.1 g V/kg H2O; keff � 0.8129 all lower SUZ as-
semblies, cB = 2.1 g V/kg H2O. ' .
The last value of keff characterizes the allowed stopped condition of the reactor
in the cold state.
5. It is impartant to know the differential and integral effectiveness of the con-
_ trols (see section 3.3) for the eval�ation of the capacity of the SUZ system (as a
whole or for individual groups of as~emblies) to compensate successfully various re-
activity effects under any operating conditions, both steady-state, and nonsteady- ~ ~
state (transient and emergency).
The effectiveness of controls depends on the composition and the combination method
of fuel assemblies in the core. Therefore, for a new reloaded (and in a number of
cases for a reloaded).core, it is necessary to determine the effectiveness of the
controls for various states of the reactor. Calculations are done by the BIPR pro-
gram in a special mode of its operation. The results o� the calculations are shown
in Tables 3.5, 3.6 and in ~igures 3,7, 3.8,
6. Com~,utations of reactivity coefficients of the core (see section 3.2) character-
izing the dynamics ~f changes in reactivity ~in nonsteady-states of the reactor are
connected very directly with everything that was said above.
The necessary initial data prepared by the UNIRASOS (ROR) prcgram together with the
KR program are introduced into the computer for computations using the BIPR program.
Thus, the determination of the coefficients of reactivity is combined with computa-
- tions of individual states of the reactnr or fuel burnup. Information on both pro-
- grams is presented simultaneously, which is very convenient.
The results of the computatictns of coefficients of r eactivity for various modes of
WER are shown in Tables 3,1, 3,2, and in Figures 3.3-3-6.
7, The selection of the cartogram of core loading is concluded by a thermohydraulic
analysis of permissible operation modes of the reactor with this load for creating
safe conditions for the fuel elements eliminating the melting of the fuel and the
occurrence of a heat exchange crisis under any nonsteady-state conditions (see Chap-
ter 5). This is a very important stage of calculations which can introduce definite
corrections in the operating mode of the reactor.
79
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7
FOR OFFICIAL USE ONLY
- The principles of the thermohydraulic analysis are explained in section 5.3. The
initial materials are: results of calculation of ttie fuel load burnup by the BIPR
program (see Figure 4,1), hydrodynamic characteristics of the assemblies and the
reactor obtained by the GDKh program (see Figures 5.1, 5.2), charts for emergency
reduction of the coolant flow rate when several GTsN are turned off (see section 5.5,
_ Figure 5.10), coefficients of micrononuniformity of energy release for the fuel ele-
ments within the assemblies calculated by the programs ShESTIGRANNIK and MIKRO (see
section 7,3, Figures 7.4-7.7, Tables 7.4, 7.5), anc? the method for calculating the
margin to the melting~of the fuel usir.g the TWEL program.
~ The maximally energy-intensive fuel element is fo~snd in the core, and it is assumed
that it limits the power level of the reactor. The maximum permissible power of the
fuel element, the assembly, and the reactor are determineci in successive order. As
a resu?.c of these computations, a chart is obtained for the maximum permissible power
of the reactor in the course of the run and a table of permissi~le power levels of
the reactor depending on the numbPr of operating GTsN and power supply circuits of
- GTsN (see Table 5.7) are obtained.
COPYRI~HT: Atomizdat, 1979
:
~4
i
= 80 '
FOit OF'FIC[AL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-04850R000500020007-7
l~l1R 11NF11'U1. l?CF (1NI.1'
- 8.4. Studying Spent Nuclear Fuel in a Hot Chamber
The condition of fuel elements which reached the necessary burnup level and were un- .
loaded from the reactor are studied in detail in a hot chamber. In the process of
studies, the condition of spent fuel elements and other parts of the assembly is
evaluated, defective fuel elements and causes of the appearance of the defect s are
revealed, possibilities of achieving an above-plan burnup level are evaluated, and
ways of improving the assemblies are determined, Moreover, gamma-spectrometrics
studies of spent fuel elements make it possible to determine experimentally the ab-
solute valuE of fuel burnup and burnup distribution over the cross section and height
of the assemblies and to obtain information about the migration of fission f ragments,
- which makes it possible to evaluate indirectly the operating temperature of the fuel
in the fuel elements.
WER fuel elements work with high temperature differences between the central part
of Che fuel elements and the jacket, creating thermal stresses. Neutron fluences
affecting fuel elements reach values of the order of 1t~21 neutrons/cm2. Unde r these
conditions of fuel element operation, the fuel swells, g aseous fission fragments
accumulate under the jacket, and the strength properties of the fuel element jackets
change.
Experimental studies of spent assemblies in a hot chamber made it possible t o con-
clude that the design of the WER assemblies and fuel elements and the technology of
- their manufacturing ensure a sufficient working capacity of the fuel to the rated
" burnup level and higher.
When stuciying spent assemblies in a hot chamber, they are first examined visually
in order to reveal possible defects and to evaluate the nature of the deposits of
corrosion products, The equipment of tt?e hot chamber makes it possible to measure
the diameter and length of the fuel element, pressure of gaseous fission products
under the jacket, ultimate stretigth; and relative lengthening of the fuel element
jackets and to determine the chemical composition of the detected deposits.
= For an example, Table 8.h givPS the experimental data about the changes in the dia-
met:r and length of the fuel elements of WER-365 assemblies of the unit II of NVAES
[67]. Changes in the fuel Plement diameter are within the limits of the manufactur-
- ing tolerance, i,e., thEre is actually no transverse swelling of the jackets. The
increase in rhe fuel element langth is somewhat higher than the manufacturing toler-
_ ance. The design of the assemb~lies make~ it possible to compensate the temp eraturE
- elongation of the fuel elements, and the absence of their bending indicates a normal
compensation of such elongations,
~
~ 81
FO~t OFFICiAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2407/02/09: CIA-RDP82-00850R000500420007-7
r�~R 0.33 0,24
Key: 1, Burnup, Mw�day/mU � 10+3
2, Changes in the dimensions of fuel elements, %
3. Diameter
4. Length
The pressure of gaseous fission fr.agments under the fuel element jacket is determined
by punctur.ing the jacket. In the VVER-365 assemblies which were studied, the gas ~
pressure under the fuel elemenr jackets under operating conditior.~ is about 10 kg-
force/cm2, This value of the pressure under the jacket indicates that the tempera-
ture of the fuel pellets during the operation of the fuel elements does not exceed
1600 degrees C [b8].
Special attention is given to the study of fuel elements with detected defects in
the jackets.
The followir~g are the possible causes of damages of fuel element jackets: 1) local
overheating; 2) cracking connected with stresses or fatigue, as well as with the ef-
fect of thermal cycles of the core and jacket of the fuel elemenC during rapid and
considerable changes in the power of the assemblies and the reactor as a whole; 3)
swelling or excessive elongation of the fuel elements caused by the accumulation of
gaseous f.ission products or structural changes in the fuel pellets; 4) development
of microdefects in the jackets of fuel elements which were not detected in the pro-
cess of the manufacturing of assemblies at the plant,
The main cause of the appearance of damages in fuel element jacket is, evidently,
the development of hidden defects which are not revealed during inspection at the
plant.
The amount of d~posits of corrosion products on the fuel element surface is insig-
nificant [69J and cannot lead to serious disturbances in the heat removal, This is
explained to a considerable degree by the absence of stagnant ~ones in the assemb-
- lies. The detected deposits have a dark brown color against Che background of a
dark grey oxide film of the fuel element jackets and are removed easily with a wad
of cotton. Stagnant ~ories appear in the gaps between the covers of the assemblies.
- Therefore, there are more corrosive deposits on thQ surfaces of the covers of assem-
blies which are activated in the neutron f1ux. If the wat~r regime is disturbed,
the activated corrosion products deposited on the covers of the assemblies and de-
vices within the vessel can spread over the entire I circuit and increase the radio-
active contamination of the equipment.
82 .
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2047/02/09: CIA-RDP82-00850R000504020007-7
FOR OFFIC[AL USE ONLY
Studies of spent assemblies make it possible to obtain information about the work-
ing conditions of thP fuel in the reactor.
' " ~~o:�e i.a;:~ ~6;,0;
o..�Y:CI ~ ~ 4~:0,.:~.Q:
\ 7 / e�r~~
t
3~
\ g ~ .
- - - - - - - - - ,
~ ~ .
.o~
, a o�o:d;,
o
%!ar'":~
?~.'~6i
\ g;~o:.y,.ly
` ~ , �b. '/o.. 0/
��Q
~e a ~.p~. ~ ~ ..~Q/:n:i.p
F.i'~� � 7 f ' .
� . o:d~�: , . \ ? a...~�.e J
�:;p;e,~%;~ � ~o 0
~ O . o� o ~ ~ , p ' e./y.
Figure 8.3. Diagram of a Device for Gamma-Spectrometic
Studies c~f Fuel Burnup:
1-- Ge(Li)-detector; 2-- capture; 3-- fuel element;
4-- collimator; 5-- table for separating the assemblies.
Gamma-spectrometric studies of spent fuel elements make it possible to determine the
absolute values of fuPl burnup, burnup distribution over the cross section and the
height of the assemblies, and a number of other characteristics [70-72]. The m~thod
of gamma-spectrometry of fission fragments znakes it possible to conduct burnup mea-
surements without preliminary radiochemical reprocessing of the irradiated fuel.
For the burnup evaluation, fission fragments and the products of their decay having
a high yield and a sufficiently large half-life are selected (Tab1e 8,7),
In the hot chamber of NVAES, gamma-spectra of fission fragmen.ts are determined by a
semiconductor germanium-lithium detector with the use of a multichannel analyzer.
When an assembly is taken apart, f.uel elements are installed in front of the colli-
mator and are moved along the height in relation to the detector (Figure 8.3). By
u~ing the multichannel analyzer, it is possible to resolve well gamma-lines of 513
kev from 85Kr and 106Ru; 605 and 796 kev 134Cs, 622 kev 106Ru, 724 and 757 kev 85Zr,
and others.
Nuclear fuel burnup is determined by using gamma-lines of isotopes 137Cs and 106Ru.
Isotope 137Cs has a large yield during the fission of 135U dnd 239Pu. 106Ru has a
preferential yield from the fission of 239Pu and 24~Pu nuclei, The absolute burnup
_ is determii~~d by comparing the intensity of the gamma-line of 137Cs of the fuel be-
ing studied and the standard cesium source.
Table 8.8 shows an example of the results of ineasurements of fuel burnup in assem-
blies of WER-210 and WER-3h5. The error in the determination of 235U burnup does
not exceed +10%, and ;:hat of 239Pu +15%.
The gamme-spectrometric mFthod of burnup determination is checked by the mass-spec-
trometric mett~od,
- 83
FOR OFFICIAL USE ONLY
APPROVED FOR RELEASE: 2007/02/09: CIA-RDP82-00850R000500020007-7

APPROVED FOR RELEASE: 2047/02/09: CIA-RDP82-00850R000504020007-7
1~