ML20093J549
| ML20093J549 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 07/13/1984 |
| From: | Stone I SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML14188B047 | List: |
| References | |
| XN-NF-83-72, XN-NF-83-72-R02, XN-NF-83-72-R2, NUDOCS 8407300172 | |
| Download: ML20093J549 (37) | |
Text
f X N-N F-83-72 REVISION 2 H. B. ROBINSON UNIT 2, CYCLE 10 SAFETY ANALYSIS REPORT r
ULY 1984 RICHLAND,WA 99352 ERON NUCLEAR COMPANY,INC.
CK O 1
p PDR
r; T
XN-NF-83-72 Revision 2 ISSUE DATE 7/13/84 H. B. ROBINSON UNIT 2, CYCLE 10 SAFETY ANALYSIS REPORT A
Written by:
f I[e*
I.L.Stojie, Nuclear Engineer Reviewed by: W[F f Jac @
F.~B.~5kogen, Manager G PWR Neutronics Prepared by:
[,
,1 [/r 49
~
~
/
s H. E. Williamson, Manager Neutronics and Fuel Management Prepared by: [L JM
/dTU44/t/
R. 8. St'out, Manager Licensing and Safety Engineering 7[et/tr Approved by:
'G. J. Susselman, Manager Fuel Design hpprovedby:
WW
// duc E V G. A. Sofe, M ger //
Fuel Engin ng and4 Technical Services C4/0/aM lZ M fy Concurred by:
J. N. Morgan, Manager Proposals and Customer Services Engineering ERON N UC _EA R CO V 3AN Y,
\\ C.
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NUCLEAR REGULATORY COMMISSION DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This technical report was derived through research and development programs sponsored by Exxon Nuclear Company, Inc. It is being sub-mitted by Exxon Nuclear to the USNRC as part of a technical contri-bution to facilitate safety analyses by licensees of the USNRC which utilize Exxon Nuclear fabricated reload fuel or other technical services provided by Exxon Nuclear for liaht water power reactors and it is true and correct to the best of Exxon Nuclear's knowledge, information, and belief. The information contained herein may be used by the USNRC in its review of this report, and by licensees or applicants before the USNRC which are customers of Exxon Nuclear in their demonstration of compliance with the USN RC's regulations.
Without derogating from the foregoing neither Exxon Nuclear nor any person acting on its behalf:
A. Makes any warranty, express or implied, with respect to the accuracy, completeness, or usefulness of the infor-mation contained in this document, or that the use of any information, apparatus, method, or process. disclosed in this document will not infringe priva ely owned rights; or S.
Assumes any liabi:it.es with respect to the use of, or for darrages resulting from the use of, any information, ao-paratus, method, or process disclosed in this document.
I F
5' i'
XN-NF F00,766 e
e
L XN-NF-83-72 Revision 2 TABLE OF CONTENTS Section Page
1.0 INTRODUCTION
1 2.0
SUMMARY
2 3.0 OPERATING HISTORY OF THE REFERENCE CYCLE............
3 4.0 GENERAL DESCRIPTION......................
6 5.0 MECHANICAL DESIGN.......................
11 6.0 NUCLEAR CORE DESIGN......................
13 6.1 PHYSICS CHARACTERISTICS..................
14 6.1.1 Power Distribution Considerations........
14 6.1.2 Control Rod R'eactivity Requirements.......
15 6.1.3 Isothermal Temperature Coefficient Considerations. 16 6.2 POWER DISTRIBUTION CONTROL PROCEDURES...........
17 6.3 ANALYTICAL METHODOLOGY 17 7.0 THERMAL-HYDRAULIC DESIGN....................
24 8.0 ACCIDENT AND TRANSIENT ANALYSIS................
25 8.1 PLANT TRANSIENT AND ECCS ANALYSES.............
25 8.2 R00 EJECTION ANALYSIS...................
25 8.3 RADILOGICAL ASSESSMENT OF POSTULATED ACCIDENTS......
26
9.0 REFERENCES
29
dk icc
_"jV
=
ii XN-NF-83-72 Revision 2 LIST OF TABLES Table Page 4.1 H. B. Robinson Unit 2 Cycle 10, Fuel Assembly Design Parameters. 8 6.1 H. B. Robinson Unit 2, Neutronics Characteristics of Cycle 10 Compared with Cycle 8 Data..................
18 6.2 H. B. Robinson Unit 2, Control Rod Shutdown Margin and Requirements for Cycle 10...................
19 8.1 H. B. Robinson Unit 2, Cycle 10 Ejected Rod Analysis, HFP...
27 8.2 H. B. Robinson Unit 2, Cycle 10 Ejected Rod Analysis, HZP...
28 LIST OF FIGURES Figure Page i
3.1 H. B. Robinson Unit 2, Critical Boron Concentration versus Exposure for Cycle 8 at Both Ncminal and Reduced Temp. Cond...
4 3.2 H. B. Robinson Unit 2, Cycle 8 PDQ Calculated to Measured
,,I Power Distribution Comparison..................
5 4.1 H. R. Robinson Unit 2 Cycle 10, Reference Loading Pattern for an E0C9 Exposure of 10,637 MWD /MT..............
9 l
q 4.2 H. B. Robinson Unit 2, 80C10 and E0C10 Quarter Core Exposure i
Distribution and Region ID for E0C9 = 10,637 MWD /MT.......
10 il 6.1 H. B. Robinson Unit 2, Cycle 10, HFP Critical Boron Concentration versus Cycle Exposure for 2,300 MWt........
20 6.2 H. B. Robinson Unit 2 Cycle 10, Assembly Power Distribution for A Cycle Exposure of 100 MWD /MT...............
21 6.3 H. B. Robinson Unit 2 Cycle 10, Assembly Power Distribution for a Cycle Exposure of 5,000 MWD /MT................
22 6.4 H. B. Robinson Unit 2 Cycle 10, Assembly Power Distribution for a Cycle Exposure of 10,820 MWD /MT
...............23
---i---m
1 XN-NF-83-72 Revision 2 H.B. ROBINSON UNIT 2, CYCLE 10 SAFETY ANALYSIS REPORT e
1.0 INTRODUCTION
The results of the safety analysis for Cycle 10 for the H.B. Robinson Unit 2 nuclear ~ plant are presented in this report. The Cycle 10 analysis reflects plant operation at 2,300 MWt. The topics addressed herein include operating history of the reference cycle, power distribution considera-tions, control rod reactivity requireinents, temperature coefficient con-siderations, and the control rod ejection analysis.
The Cycle 10 design requires the loading of sixty-five (65) fresh Exxon Nuclear Company (ENC) supplied fuel assemblies; forth-four (44) XN-7 assemblies, nine (9) XN-6 assemblies and twelve (12) Partial Length Shield Assemblies (PLSAs). The forty-four (44) fresh XN-7 fuel assemblies utilize natural uranium axial blankets (NUA8s) with thirty-six (36) of the assemblies also containing gadolinia-bearing fuel pins. One of the XN-6 assemblies also contain gadolinia-bearing fuel pins.
The twelve (12)
U PLSAs will reduce the fast neutron fluence reaching the pressure vessel wall.
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1
a m
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2 XN-NF-83-72 Revision 2 2.0
SUMMARY
Cycle 10 of the H.B. Robinson Unit 2 nuclear plant is designed to operate at 2,300 MWt in Cycle 10 beginning in the fall of 1984. The characteristics of the fuel and of the reload core result in conformance l
with required shutdown margins and thermal limits. Ihis document provides the neutronic analysis for the plant during Cycle 10 operation.
The ENC fuel mechanical design is presented in Reference 1.
The thermal-hydraulic analyses are provided in Reference 14.
The Plant Transient Analyses and the ECCS/LOCA Analyses are presented in References 15 and 16, respectively.
The generic Control Rod Ejection Analysis is
.provided in Reference 17. These ana' lyses are all applicable to the Cycle
.10 operating conditions at 2,300 MWt.
H.B. Robinson Cycle 8 has been chosen as th'e reference neutronics cycle due to the close resemblance of the overall neutronic character-istics of Cycle 10 to Cycle 8.
The results contained in this report.show that operation at a power level of.2,300 MWt can be safely achieved within the licensed limits.
1 s
1 k
c 4
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'Y_
s
l i
3 XN-NF-83-72 Revision 2 3.0 OPERATING HISTORY OF THE REFERENCE CYCLE H. 8. Robinson Cycle 8 has been chosen as the reference neutronics cycle due to the close resemblance of the overall neutronic character-istics of Cycle 10 to Cycle 8.
The Cycle 8 plant operation at nominal core average temperature, and power (5780F, 2,300 MWt) started in October 1980 and continued until November 1981. In early December 1981, operation at a reduced core average temperature and power was initiated to improve steam generator tube performance. Cycle 8 ended with an accumulated exposure of 10,383 MWD /MT; 2,300 MWD /MT of which was achieved at reduced temperature and power.
The measured power peaking factors remained below the Technical N
Specification limits for Cycle 8.
The total nuclear peabing factor, F g, andtheradialnuclearpinpeakingfactor,FfH,remainedbelow1.96and 1.49 respectively. Cycle 8 operation was typically rod free with Control Bank D positioned in the range of 200 to 220 steps; 228 steps being fully withdrawn. It is anticipated that similar control bank insertions will be seen in Cycle 10.
i The critical boron concentration as calculated by ENC for Cycle 8 has f
agreed well when compared to the observed values (see Figure 3.1). A power distribution calculated with the PD0 model is compared to measured values shown in Figure 3.2.
The comparison is made at a Cycle 8 exposure of 7,083 MWD /MT for a' core power of 96% of 2,300 MWt.
g.
i a
y.
b K-
'.._ y v
s L
1400 --
r m
=
i r
[:
1200
[
/
g I
~
.m g'
1000 7
i p
f f
ENC Calculdt ion (539"F, s 1746 'HWt )
'y 8
~
c ' ~ 800 -
m i
ENC Calculatton (578,F, 2300 MWt) 88 t
8' g%og o
Measured g
o E
600 N
O r
.g
.u 400 v.
'E
- qo f
u 200
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i E
p-
^
^
^
~'
O 1
2 3
4 5
6 7
8 9
10 11 0
4 Cycle Exposure, GWD/MT 5
- u s i
Figure.3.1 H.B. Robinson Unit 2, Critical Boron Concentration versus Exposure for Cycle 8 at Both Nominal and Reduced Temperature Conditions
&T 5
E$
"L
~~
m k
g
-________= _ __
5 XN-NF-83-72 Revision 2 H
G F
E D
C B
A 0.750 1.115 1.013 0.982 1.168 1.139 0.935 0.804 8
0.730 1.085 0.982 0.961 1.163 1.136 0.947 0.828 2./
2.8 3.2 2.2 0.4 0.3
-1.3
-2.9 1.111 0.996 1.184 1.033 1.180 0.975 1.141 0.668 c
1.087 0.975 1.159 1.016 1.175 0.978 1.161 0.689 z
2.2 2.2 2.2 1.7 0.4
-0.3
-1.7
-3.0 1.003 1.182 1.006 1.182 1.001 1.081 0.971 10 0.984 1.161 0.992 1.173 1.001 1.089 0.983 1.9 1.8 1.4 0.8 0.0
-0./
-1.2 0.972 1.027 1.184 1.118 u.970 1.127 0.698 11 0.963 1.016 1.174 1.104 0.971 1.138 0.708 0.9 1.1 0.9 1.3
-0.1
-1.0
-1.4 1.148 1.168 1.011 0.977 0.940 0.731 reasured 12 1.163 1.174 1.001 0.971 0.941 0.738 PDQ
-1.3
-0.5 1.0 0.6
-0.1
-0.9 M-P X 100 P
~
1.099 0.962 1.092 1.137 0.733
~
1.136 0.976 1 088 1.136 0.737 13
-3.3
-1.4 0.4 0.1
-0.5,
Measured Calculated 0.921 1.131 0.974 0.705 14 0.948 1.161 0.982 0.708 FNH 1.30 1.33
-2.8
-2.6
-0.8
-0.4 N
F 1.45 1.45 Q
0.806 0.670 15 0.329 0.689
-2.8
-2.8 Figure 3.2 H.B. Robinson Unit 2, Cycle 8 PDQ Calculated to Measured Fawer Distribution Comparison, 96% of 2,300 MWt 7,083 MWD /MT f or D-Bank at 215 Steps ul"
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6 XN-NF-83-72 yS.
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4.0 GENERAL DESCRIPTION r9 m l1..
The H.B. rod 1nson reactor consists of 157 assemblies, each having a gj 7.
15x15 fuel rod array.
Each assembly contains 204 fuel rods, twenty RCC g..
g' y
guide tubes, and one instrumentation tube.
The RCC guide tubes and the
- q[.
- 9
.. p
., viJ instrumentation tube are made of decaloy.
Each ENC assembly contains
' j,. n
- e.. '.h
,i 7 ',,
seven zircaloy spacers with Inconel springs; six of the spacers are located A.. q '.
4
~ 7 within the active fuel region. The fuel rods consist of slightly enriched
' t.. ^g -
.3
,s-s.u:.
U0p pellets inserted into zircaloy tubes.
/
e-7.
The Cycle 10 design reflects the loading of sixty-five (65) fresh ENC yd y
yh-
= v-
[W
'.T.
supplied fuel assemblies.
This core design contains the first reload of
.. i..
NI axially blanketed fuel in H. B. Robinson Unit 2, XN-7, and twelve (12)
Y tu
.c J-4-
r
?'..+
special fuel assemblies which will reduce the fast neutron ' fluence b E
- u. am
- e.
Nl ".
reaching the pressure vessel wall.
The latter are denoted as Partial
- ..y.7,
r 3-
~~
J,i Length Shield Assemblies (PLSAs).
This core design is the second H.
B.
'f.
.h Robinson reload fuel design utilizing 4 w/c gadolinia. Thirty-six (36) of M..'.
a.4 t -
the forty-four (44) fresh Region 13 (XN-7) fuel assemblies and one of the g:
4
),U fad
...Y nine (9) fresh Region 12 (XN-6) assemblies contain gadolinia-bearing pins.
. g.
- @a
, i b~
The design for Batch XN-7, Region 13, includes natural uranium axial
!p.7 '
- p.
ty. !
.L blankets in the top and bottom six (6) inches of the active fuel region.
h; g
[jl.;f k.'
The batch average enrichment for the blanketed assemblies is 3.08 w/o U-6
- f 235.
This average enrichment is achieved by using a central axial zone 3 : '"
N y ~;, -
enrichment of 3.34 w/o in fuel pins which contain no gadolinia cnd 2.37 w/o j d.;.:
Q in the gadolinia-bearing fuel pins.
In twelve (12) Region 13 assemblies, M.[
.4 4
n
-,' Q.
2 l,,1
- e k/,
je r x.
,f
. an
,~.'[
-); _,5 -
l& ',
-s,
+ i,
- h M s.j % y * %.'; W.7 7.7.; V 4.f f. p g:3 n.. p g.a j.1 w. y f.y 7..
..; m. g ;g 7 4
J, ' %. y a > x,,,. y 7 _ 7.y ;.. ;_.
_-.:.. v ;.,.. > 3 w
- .y.,x, y..,.
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7 XN-NF-83-72
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Revision 2 y '5 s
q' _ RE f.<
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d4 twelve (12) fuel pins per assembly will contain 4 w/o gadolinia.
In b-2.-
<w.M another twenty-four (24) assemblies, eight (8) pins of 4 w/o gadolinia per 3,,..
7g :l rh assembly will be utilized.
In the remaining eight (8) Region 13 (XN-7) f.
.c+$
....q fuel assemblies no gadolinia pins will be used.
In addition to the forty-
.;J p ;:,
. :n.
. [.3 four (44) blanketed assemblies loaded, nine (9) fresh XN-6 assemblies 3[:
6 7;
' ?-
containing 2.85 w/o enrichment will be used in Cycle 10. One fresh (1) XN-lyy
- g..g.,.('
6 assembly will contain twelve (12) pins of 4 w/o gadolinia and will be
[;%. s
- p; 4
/,
.l
- N.
loaded in the center core location. An enrichment of 2.20 w/o U-235 is used q
p qj p in the XN-6 gadolinia-bearing fuel pins. Thus, the total number of f
3.j l gadolinia pins required for Cycle 10 is 348.
%(
s.
Twelve (12) PLSAs aie being loaded on the core periphery (the " flats")
.I. 1"f
[
3 ;;
- A1 as part of the program to reduce the fast neutron fluence to the pressure V ::.> -
%g' vessel wall.
In the active fuel region, the top six (6) inches contein
.M.
s
.e ej.p natural uranium, the next ninety-six (96) inches contain uranium enriched
- ,5 '
Jy 4-to 1.24 w/o, and the bottom forty-two (42) inches contain 304 stainless y
- t. o J.- (
q.t OI4
' N steel.
g7
-s p
~e.
l,. V....
The projected Cycle 10 loading pattern is shown in Figure 4.1 with the
)paf N.$,
-h..<
' d assemblies identified by assembly tabrication ID and by their core
[.'.
y:%
f &g
??.@
location in the previous cycle or by fresh fuel region. 80C10 exoosures,
' e.c.,
r y/p 4
based on an EOC9 exposure of 10,637 MWD /MT, along with Region ID's, are g
r, g
'.f.
J k..
shown in Figure 4.2.
The initial enrichments of the various regions are ay
. r:: :
v listed in Table 4.1.
Also included in Table 4.1 are the peak assembly
.q
.* 3.2 y.
,sy g
. ;,j-exposures by region and fuel type.
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.m
.n,.
kA A'5
,q
,.rc 3.:;
. D?.
8,.'
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Yg l'Q
- * +1 g y:
. ;.m
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(h-[
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v E
b b
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vd' p'.; c.j '.- ' - ' ?j, -j :
yyNp;
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Table 4.1 H.B. Robinson Unit 2 Cycle 10, Fuel Assembly Design Parameters
,g
~i Region PLSA 11 12 12 12*
12**
12***
12***
13 13**
13***
XN Number XN-7 PLSA 5
6 6
6 6
6 6
7 7
7 Number of Assemblies 12 48 16 8
12 8
8 1
8 24 12 Pellet Density, ETD 94.0 94.0 94.0 94.0 94.0 94.0 94.0 94.0 94.0 94.0 94.0 Pellet to Clad Diametral Gcp, Mil 7.5 7.5 1.5 7.5 7.5 7.5 7.5 7.5 7.5 7.5 7.5 go Initial Enrichment (w/o U-235)
Upper 6 inches 002
.71 2.90 2.85 2.85 2.85 2.85 2.85 2.85
.71
.71
.71 UO -Gd 023 2.20 2.20 2.20 2.20 2
Central 132 inches UO2 1.24+
2.90 7 85 2.85 2.85 2.85 2.85 2.85 3.34 3.34 3.34 UO -Gd 023 2.20 2.20 2.20 2.20 2.37 2.37 2
Lower 6 inches 2.85 2.85 2.85 2.85 2.85
.71
.71
.71 UO2 3045STL 2.90 2.85 UO -Gd 023 2.20 2.20 2.2G 2.20 2
Average 0.86 2.90 2.85 2.85 2.84 2.82 2.81 2.81 3.12 3.09 3.02 4
4 4
4 4
4 0
Initial Gd2 3, w/o i.
Batch Average Barnup at 80C10, MWD /MT 0
21,534 8,348 0
10,899 11,043 12,646 0
0 0
0 4
Peak Assembly Burnup
>c at EOC10, MWD /MT 3,705 34,705 21,003 10,465 22,820 22,975 25,349 13,936 7,909 12.099 13,756 goje flE5 Fuel with 4 pins of 4 w/o gadolinia per assembly vs e 2
Fuel with 8 pins of 4 w/o gadolinia per assembly g* 00 Fuel with 12 pins of 4 w/o gadolinia per assenbly s e
+
Lower 36 inches of central 132 inches contains 304SSTL no pj PLSA Part Length Shield Assembly
- lr -'
9 XN-NF-83-72 Revision 2 R
P N
M L
K J
W G
F E
D C
B A
PLSA PLSA PLSA 1
N45 N49 N53 12 13 7
13 12 N01 M18 N21 N25 N29 M19 NO2 K13 3
F13 L3 L2 H7 E2 E3 L52 N41 L10 M01 LO2 M02 L13 N42 L50 N10 4
M3 J4 K2 H14 F2 G4 D3 C10 L31 N37 MOS LO4 M41 M49 M42 LO9 M06 N38 L23 13 K9 C4 E4 L4 13 N4 F9 N05 N33 M09 L33 N09 LO6 N17 LO7 N10 La2 M10 N34 N06 5
D7 C5 12 G10 G14 F5 J14 J10 12 NS M7 M22 Ll4 Ll6 N13 L48 M36 L41 M35 L45 N14 L19 L15 M23 6
PLSA PLSA PS P6 MS B9 012 H5 M12 P9 05 B6 B5 N56 N32 M13 M46 L23 M40 M32 M17 M31 M38 L24 M51 M14 N22 N46 7
PLSA P8 G8
+
E8 E10 L10 L8 PLSA J8 88 i 8 N52 N28 L26 M48 N20 L18 M26 M53 M27 Ll7 N18 M50 L27 N26 N50 l PLSA ia P11 P10 Mll B7 04 H11 M4 P7 011 810 Bil PLSA N48 N24 M16 M4 5 ++ L29 M39 H30 M28 M29 M37 L30 MS2 M15 N30 N54 12 N11 M9 G6 G2 F11 J2 J6 D9 Cll 12 33 M25 L38 1.34 N16 LO8 M34 L11 M33 LOS N15 L37 L39 M24 13 K7 C12 E12 N12 F7 L12 13 N08 N36 M12 L21 N12 L46 N19 L47++
N11 L20 Mll N35 N07 N6 M13 J12 K14 H2 F14 G12 Cl3 C6 L25 N40 M08 L44 M44 M47 M43 L49 M07 N39 L22 K3
,a L13 Ll4 H9 E14 E13 F3
^~
LO3 N44 L40 M04 L51 M03 L43 N43 LO1 12 13 la 13 12 N04 M21 N31 N27 N23 M20 NO3 S pins of 4 w/o gadolinia per PLSA PLSA PLSA assembly in Region 12 fuel -
N55 N51 N47 15
- 12 pins of 4 w/o gadolinia per assembly in Region 13 f eel PLSA Part lengtn Shielo Assembly
_y NumDer of gadolin13 pins per assembly 12 pins of 4 w/o gadoliaia per ICycle 9 core !ccat:an or egior. numter
+
assembly in Region 12 fuel
-=a As semb l y f abr i c a t i on 'O This assembly contains one (1)
++
l inert zirconium rod Figure 4.1 H.B. Robinson Unit 2 Cycle 10, Ref erence.0ading Pattern f or an EOC9 Exposure of 10,637 "WC,vT r
M
.__._.~
l
=
10 XN-NF-83-72 Revision 2 l
H G
F E
D C
B A
O 13,186 24,048 0
0 0
11,179 21,094 E
8 13,936 25,349 34,507 13,745 24,140 32,093 12,089 3,705 12***
12 11 13***
12 11 11**
PLSA 0
[
13,155 12,121 11,072 24,067 10,522 7,909 0
3,083 9
25,324 24,292 22,975 34,698 22,820 20,549 11,792 e
12 12 12 11 12 12 13**
PLSA 24,045 11,014 22,612 0
20,446 19,555 0 10 34,510 22,924 33,404 13,602 31,277 30,102 10,457 11 12 11 13***
11 11 12 E
11 24 070 0
22 634 8 791 0
0 13,756 34,,705 13,609 33,,492 20,,990 11,832 7,903 13***
11 13***
11 12 13**
13 L
i E
11,636 10,498 20,406 8,790 0
15 984 12 24,086 22,815 31,256 21,003 11,730 25,,654
[
12 12 11 12 13**
11 13 21,077 7,902 19,538 0
19,962 32,092 20,557 30,099 11,842 25,637 11 12 11 13**
11 L
80C10 Exposure, MWD /MT 14 0
0 0
0 12,099 11,802 10,465 7,909 E0C10 Exposure, MWD /MT 13**
13**
12 11 Region 10 15 0
0 3,707 3,040 r-PLSA PLSA I
Iuel with 8 pins of 4 w/o gadolinia per assembly T
Fuel with 12 pins of 4 w/o gadolinia per assembly i
PLSA Part length Shield Assembly E
Figure 4.2 H.B. Robinson Unit 2, B0C10 and EOC10 Quarter k
Core Exposure Distribution and Region ID for E0C9 = 10,637 MWD /MT r
N u
t-
.i
11 XN-NF-83-72 Revision 2 5.0 MECHANICAL DESIGN The 44 XN-7 Reload fuel assemblies are mechanically identical to the previous reload assemblies, with the exception of the fuel rods. The 144 inch fuel column includes a six (6) inch column of natural U02 pellets at each end. The natural UO2 pellets have a length-to-diameter ratio of 1.5 and a total dish volume of 0.7% as compared to 0.8 and 1.0%, respectively, for the enriched pellets. In addition, the fuel column insulator discs are no longer used.
The 12 Partial Length Shield Assemblies (PLSAs) are mechanically identical to the Reload XN-7 assemblies, with the exception of replacing the bottom 42 inches of the fuel column with inserts of type 304 stainless steel rod used for neutron shielding.
A supplementil mechanical design analysis for the PLSAs is provided in Reference 3.
A description of the basic Exxon Nuclear supplied fuel design and design methods is contained in Reference 1.
In addition, mechanical design analysis of the Reload XN-7 fuel and the resident XN-5 and XN-6 fuel with current methodology is contained in Reference 2.
Several of the XN-5 assemblies are e'xpected to exceed the original design burnup of 33,000 MWD /MT. The reference analysis encompasses the expected conditions, such that the relevant mechanical criteria will not be exceeded.
Due to the low power of the PLSA fuel, the design conditions are generally enveloped by previous design analyses. However, the differences in physical properties of the stainless steel relative to uranium dioxide required confirmation of certain aspects of the design.
y
-F 12 XN-NF-83-72 555 Revision 2
=Ft j'
o The thermal and mechanical behavior of the fuel rod was a
evaluated for thermal expansion ettects of the stainless steel.
The fuel rod response was determined to be bound by the higher
]
A power fuel rods.
]
~
The loss in assembly holddown margin resulting from the lower o
weight fuel rods was determined to be within the capability of
('
the holddown springs to prevent hydraulic liftoff under normal E-operating conditions.
=:
o The fuel assembly was determined to be no more sensitive to l;
irradiation-induced bowing than previous design.
a o
The seismic analysis was determined to be valid for the lower
{-
~
weight of the assemblies.
One (1) once burnt and one (1) twice burnt assembly containing one
'nert rod each will be loaded irito the core for Cycle 10. The. inert rods were designed to be mechanically compatible with the fuel assembly.
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13 XN-NF-83-72 Revision 2
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q,.9 6.0 NUCLEAR CORE DESIGN Nk'
- c Y l.
The H. B. Robinson Unit 2, Cycle 10, Region 13 reload design has
- s. Ag!.
A ;-
- .y a i-been developed in accordance with the following requirements
V.r:
J 1.
The Cycle 10 reload shall contain sixty-five (65) new fuel
- fe. p a4
'.m o
.j assemblies; nine (9) XN-6 (Region 12) assemblies, forty-four 9[..
,1 (44) XN-7 (Region 13) asemblies, and twelve (12) Part length Shield,tssemblies.
Q F~- l 2.
The length of Cycle 10 shall be maximized.
e :.;,.,1
, a...
....[
3.
The rated power for Cycle 10 shall be 2,300 MWt.
- 41. ',(
..,'. d 7f The length of Cycle 10 shall be determined based on an actual h.
4.
a.
. a.,
n.
E0C9 exposure of 10,637 MWD /MT.
1*
t,2 J.. C
'E 6.
Cycle 10 Operation is anticipated to be base loaded; how-
' ) 'l: I s,-
e ever,the reload fuel shall be designed to accommodate load igy.'
./. ;
following operation between 50% and 100% of rated power while
. -[.'
,..s ir not precluding the current ramp and step change bases as set 4
g),5, l 3
forth in the FSAR.
e-1 s.
'n 6.
In accordance with plant Technical Specifications, the control 7.J.!
, w.:
rod worth requirements shall be met.
Y
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7.
The loading pattern shall be designed to produce acceptable
';.p f:;
...,y..
.s j power distributions. The design F g, including uncertainties,
, M..
T 9[
shall be less than 2.32 at 2,300 MWt. The integrated peak to
,{.
average pin power, FaH, including measurement uncertainties,
',y[
' n '.,e shall be less than 1.65 at 2,300 MWt.
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14 XN-NF-83-72 Revision 2 8.
The loading pattern shall be designed to accommodate the PLSAs used to reduce the neutron fluence at the pressure vessel.
The neutronic design methods utilized in the analyses are consistent with those described in References 4 through 7.
6.1 PHYSICS CHARACTERISTICS The neutronic characteristics of the Cycle 10 core are compared to those of Cycle 8 in Table 6.1. The data presented in the table indicates the neutronic similarity between Cycles 8 and 10.
The reactivity coefficients of the Cycle 10 core are bounded by the coef t icients used in the safety analysis. The safety analysis f or Cycle 10 is based on physics characteristics representative of those expected f or a Cycle 9 length of 10,637 MWD /MT.
The boron letdown curve for Cycle 10 operation at 7,300 MWt is shown in Figure 6.1.
As shown, the B0C10 no xenon, hot f ull power (HFP )
critical boron concentration is predicted to be 1,002 ppm. At 100 MWD /MT,
equilibrium xenon, the critical boron concentration at HFP is 713 ppm. The Cycle 10 length is projected to be 10,820 MWD /PR (312 EFPDs) with no boron at E0C at a power level of 2,300 MWt.
6.1.1 Power Distribution Considerations At a power level of 2,300 MWt at equilibrium xenon conditions,100 MWU/MT, the calculated XTG peak F6H is 1.48 including a 4%
T measurement uncertainty.
At the same exposure the peak Fg js 2,1g T.
including a 3% engineering f actor, a 5% measurement oncer tainty, K(Z) consideraions, and an 11% allowance for operation with PDC-II and +5%
target bands.
i------i
15 XN-NF-83-72 Revision 2 T
The peak F3H and Fg in Cycle 10 occur at a cycle exposure of 5,000 MWD /MT.
The predicted value of FaH at this exposure is 1.56 T
including the 4% measurement uncertainty. The peak F g, again including V(Z) and K(Z) considerations and appropriate uncertainties, is 2.18.
The quarter-core radial power distributions are pre-i sented in Figures 6.2 through 6.4 for Cycle 10 exposures of 100 MWD /MT, 5,000 MWD /MT, and (E0C) 1.0,820 MWD /MT, respectively.
=
With the normalized axial dependence factor K(Z) in-T cluded, the expected peak power, F g, versus core axial elevation has been T
All F g values are within Technical Specification limits. The evaluated.
limiting case is found to be at 5,000 MWD /MT near the core mid-plane with T
F g being about 6% below the allowable unit for operation with target bands of + 5% at 2,300 MWt.
For Cycle 10 (as in Cycle 9), the total allowable power peaking factor of 2.32 at 2,300 MWt includes the local power peaking due to fuel densification and uncertainties related to engineering factor, measurement, and analysis.
This power peaking limit assures that the linear power density remains below the limiting value3 of the fuel rod linear power density, thus meeting the LOCA and the overpower limit criteria.
6.1.2 Control Rod Reactivity Requirements vetailed calculations of shutdown margins for Cycle 10 are compared with Cycle 8 data in Table 6.2.
A value of 1,770 pcm is used at E0C in the evaluation of the shutdown margin to be consistent with the Technical Specifications. The Cycle 10 analysis indicates excess shutdown
16 XN-NF-83-72 Revision 2 margins of 1,911 pcm at the B0C and 461 pcm at the EOC. The Cycle 8 analysis indicated excess shutdown margins for that cycle of 1,554 pcm at the B0C and 565 at the E0C. The Cycle 10 excess shutdown margins are seen to be similar to the Cycle 8 values.
The control rod groups and insertion limits for Cycle 10 will remain unchanged from Cycle 8. The control rod shutdown requirements in Table 6.2 allow for a HFP D Bank insertion equivalent to 600 pcm for B0C and 400 pcm for E0C, to bound the Cycle 10 control rod worths.
6.1.3 Isothermal Temperature Coefficient Considerations The Cycle 10 isothermal temperature coefficients are shown in Table 6.1 for HFP and HZP conditions at both BOC and E0C.
At 80C10, following a nominal E0C9 shutdown exposure of 10,637 mwd /MT, the HFP isothermal temperature coefficient is projected to be -5.1 pcm/0F at a -
critical baron concentration of 1,002 ppm. The corresponding HZP critical boron concentration is 1,134 ppm with the isothermal temperature coef-ficient being -0.7 pcm/0F.
The Technical Specification for the moderator temp-erature coefficient for Cycle 10 allows for a +5.0 pcm/0F (Reference 15) at HZP conditions.
With a calculated value of + 1.0 pcm/0F at HZP, the Technical Specifications at this condition is expected to be met, with no control rod insertion anticipated for ascension to HFP conditions. Simi-larly, at HFP conditions, the calculated moderator temperature coefficient for HFP, shown in Table 6.1 for no xenon, is well below the regirement of 0 pcm/0F.
.. - s.
17 XN-NF-83-72 Revision 2 6.2 POWER DISTRIBUTION CONTROL PROCEDURES ine control of the core power distribution is accomplished by following the procedures for " Exxon Nuclear Power Distribution Control for Pressurized Water Reactors, Phase II"(8,9,10). These procedures, denoted PDC-II, have been generically approved by the NRC for application to Westinghouse type PWRs.
6.3 ANALYTICAL METHODOLOGY The metnoas usea in tne Cycle 10 core analyses are described in References 4 through 7.
In summary, the reference neutronic design analysis of the reload core was performed using the XTG(II) reactor simulator system. The fuel shuffling between cycles was accounted for in the calculations. The P0Q/ HARMONY (12,13) code package will be utilized to monitor the power distribution.
Calculated values of Fg, Fxy, and FoH were studied with the twenty-four (24) axial node XTG reactor model.
The thermal-hydraulic feedback and axial exposure distribution effects of power shapes, rod worths, and cycle lifetime are explicitly included in the analysis.
~
. =..
18 XN-NF-83-72 Revision 2 Table 6.1 H. 8. Robinson Unit 2, Neutronics Characteristics of Cycle 10 Compared with Cycle 8 Data Cycle 8 Cycle 10 BOC EOC BOC EOC Critical Baron HFP,AR0,(ppm) 1,217 20 1,002 0
1,134 HZP, AR0, No Xenon (ppm) 1,305 Moderator Temp. Coefficient HFP,(pcm/0F)
-1.25
-29.5
-3.8
-31.3
+1.0
-21.0 HZP,(pcm/0F)
+3.7 Isothermal Temp. Coefficient HFP,(pcm0F)
-2.5
-30.9
-5.1
-32.8
-0.7
-22.8 HZP,(pcm0F)
+2.0 Doppler Coefficient (pcm/0F)
-1.2
-1.4
-1.3
-1.5 Power Defect (Moderator +
Doppler),pcm 1,359 1,904 1,532 2,030 Boron Worth, (ppm /103 pcm)
HFP
-106
-96
-107
-96
-105
-92 HZP l
Promp Neutron Lifetime (usec) 24.8 24.0 25.7 24.2 Delayed Neutron Fraction 0.0060 0.0053 0.0061 0.0053 Control Rod Worth'of All Rods In Minus Most Reactive Rod,HZP,(pcm) 5,626 5,765 6,326 5,901 Excess Shutdown M3rgin, (pcm) 1,554 565 1,911 461 XN F00.553
'~ ~
19 XN-NF-83-72 Revision 2 Table 6.2 H.B. Robinson Unit 2, Control Rod Shutdown Margin and Requirements for Cycle 10 (pcm)
BOC 8 E0C 8 B0C 10 E0C 10 Control Rod Worth AR1 6,926 7,065 7,626 7,201 N-1 5,626 5,765 6,326 5,901 (N-1)*.9 5,063 5,189 5,693 5,311 Reactivity Insertion Power Defect (Moderator
+ Doppler) 1,359 1,904 1,532 2,030 Full Power D Bank Insertion 500 300 600 4)0 Flux Redistribution 600 600 600 600 Void 50 50 50 50 Total Requirements 2,509 2,854 2,782 3,080 Shutdown Margin (N-1)*.9 - Total Require-ments 2,554 2,335 2,911 2,231 Required shutdown Margin 1,000 1,770 1,000 1,770 Excess Shutdown Margin 1,554 565 1,911 461
20 XN-NF-83-72 Revision 2 M
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21 XN-NF-83-72 Revision 2 H
G F
E D
C B
A
- +
O 1.265 1.244 1.029 1.219 1.243 1.078 1.051
.283 PLSA i
9 1.245 1.258 1.190 1.002 1.232 1.271 1 054
.236 PLSA W
+
10 1.030 1.191 1.013 1.162 1.034 1.029 1.034 11 1.221 1.004 1.163
.991 1.158 1.069
.765 12 1.248 1.235 1.037 1.160 1 031
.496 13 1.082 1.275 1.031 1.071
.497 Relative Assembly Power Peak Assembly = 1.28 (G13)
PeakFfH
= 1.43 (H8)
+
34 1.05' 1.056 1.036
.767 Peak F
= 1.52 (G9) xy N
Peak F
= 1.82 (G14) g PLSA PLSA 8 pins of 4 w/o gadolinia per assembly 12 pins of 4 w/o gadolinia per assembly
+
Region 12 fresh fuel loaded at 80C10 PLSA Part Length Shielding Assembly Figure 6.2 H.B. Robinson Unit 2 Cycle 10, Assembly Power Distribution f or A Cycle Exposure of 100 MWD /MT 3-D XTG analysis at 2,300 MWt
22.
XN-NF-83 Revision 2 H
G F
E D
C B
A
- +
8 1.360 1.141
.981 1.323 1.163 1.031 1.158
.349 PLSA 9
1.141 1.135 1.114 1.005 1.150 1.182 124
.285 PLSA
+
10
.982 1.115 1.021 1.315 1.020
.989
.982 11 1.325 1.005 1.316 1.032 1.152 1.125
.744 12 1.166 1.152 1.022 1.153 1.119
.537 13 1.032 1.183
.990 1.126
.537 Relative Assembly Power Peak Assembly = 1.36 (H8)
PeakFfI
= 1.50 (H8)
+
14 H
1.159 1.125
.982
.744 Peak F
= 1.61 (H8) xy N
Peak F
= 1.82 (H14) g 15
.350
.285 PLSA PLSA 8 pins of 4 w/o gadolinia per assembly 12 pins of 4 w/o gadolinia per assembly Region 12 fresh f uel loaded at 80C10
+
PLSA Part length Shielding Assembly Figure 6.3 H.B. Robinson Unit 2 Cycle 10, Assembly Power Distribution for a Cycle Exposure of 5,000 MWD /MT, 3-0 XTGPWR Analysis at 2,300 MWt
23 XN-NF-83-72 Revision 2 l
l H
G F
E D
C B
A c**
+
8 1.282 1.075
.963 1.314 1.124 1.025 1.214
.402 PLSA 9
1.075 1.073 1.076 1.002 1.114 1.148 1.167
.327 PLSA
+
10
.963 1.076 1.018 1.330 1.019
.988
.969 11 1.314 1.002 1.330 1.041 1.142 1.155
.746 12 1.125 1.115 1.020 1.143 1.162
.579 l
13 1.025 1.148
.988 155
.579 Relative Assembly Power Peak Assembly = 1.33 (Fil)
N
+
Peak F
- 1.42 (E10)
AH 14 1.213 1.167
.969
.746 Peak F
= 1.60 (E10) xy N
Peak F
= 1.68 (E10) g 15
.401
.327 PLSA PLSA 8 pins of 4 w/o gadolinia per' assembly 12 pins of 4 w/o gadolinia per assembly
+
Region 12 fresh fuel loaded at B0C10 PLSA Part Length Shielding Assembly Figure 6.4 H.B. Robinson Unit 2 Cycle 10, Assembly Power Distribution for a Cycle Exposure of 10,820 MWD /MT, 3-D XTGPWR Analysis at 2,300 MWt
___ _--_ _______ _ J
1 l
24 XN-NF-83-72 Revision 2 7.0 THERMAL-HYDRAULIC DESIGN The Cycle 10 core consists of ENC supplied fuel assemblies as shown in Table 4.1 The hydraulic design of all fuel types are identical, thereby ensuring compatibility. The hydraulic design was analyzed and the results of the analysis documented in Reference 14.
The thermal hydraulic performance of the H.B. Robinson Unit 2 core has been analyzed. A bounding core assembly power distribution at the exposure which exhibited the worst radial peaking was used in the analysis.
The maximum technical specification limif (F[g = 1.65) assembly was assumed. A boundingly low local rod power distribution was used in the analysis of the maximum radially peaked assembly to provide low departure from nucleate boiling ratio (DNBR) results.
The analysis uses a lower plenum flow maldistribution factor of 5% to the limiting fuel assembly, 4.5% core flow bypass and a low estimate of total recirculating primary coolant flow rate for 6% steam generator tube plugging with a further 3% reduction to account for flow measurement uncertainty. Therefore, the analysis of the core as constituted considered all fuel types in a bounding manner.
The results of the thermal-hydraulic performance analyses are used as bases for the analyses of anticipatdd operational occurences(15). The results of the plant transient analyses show that the specified acceptable fuel design limits for each event as defined in the plant licensing bases are met. The results of the analysis are reported in Reference 15.
25 XN-NF-83-72 Revision 2 8.0 ACCIDENT AND TRANSIENT ANALYSIS 8.1 PLANT TRANSIENT AND ECCS ANALYSES The plant transient (15) and ECCS(16) analyses are performed to support operation at a power level of 2,300 MWt with 6% of the tubes plugged in the replacement steam generators. Confirmation that operation with an Fpg of 1.65 and an Fg of 2.32 meets the acceptance criteria for each event T
as defined in the licensing basis for H. B. Robinson Unit 2 is provided in References 15 and 16.
8.2 R0D EJECTION ANALYSIS A Control Rod Ejection Accident is defined as the mechanical f ailure of a control rod mechanism pressure housing, resulting in the ejection of a Rod Cluster Control Assembly (RCCA) and drive shaf t.
The consequence of this mechanical f ailure is a rapid reactivity insertion together with an adverse core power distribution, possibly leading to localized fuel rod damage.
The rod ejection accident has been evaluated with the procedures developed in the ENC Generic Rod Ejection Analysis, Reference 17.
The ejected rod worths and hot pellet peaking f actors were calculated using the XTG code. No credit was taken for the power flattening effects of Doppler or moderator feedback in the calculation of ejected rod worths or resultant peaking factors. The pellet energy deposition resulting from an ejected rod was conservatively evaluated explicitly for 80C and E0C conditions.
The HFP pellet energy deposited was calculated to be 165 cal /gm at B0C and 172 cal /gm at E0C. The HZP pellet energy deposition was calculated to be less than 40 cal /gm for both BOC and E0C conditions.
The rod ejection
26 XN-NF-83-72 Revision 2 accident was found to result ir: an energy deposition of less than the 280 cal /gm limit as stated in Regulatory Guide 1.77.
The significant parameters for the analyses, along with the results, are summarized in Tables 8.1 and 8.2 8.3 RADIOLOGICAL ASSESSMENT OF POSTULATED ACCIDENTS With the increased fuel exposures planned for H.B. Robinson Unit 2 fuel, the potential radiological consequence of the postulated accidents have been assessed to verify that the 10 CFR 100 limits continue to be satisfied. This cssessment assumes that the reactor is operated at 2,300 MWt with a peak assembly exposure of 44,000 MWD /MT.
The results of the assessment (Reference 17) indicate that the whole body and thyroid dose received from the postulated loss-of-coolant accident (LOCA), fuel hand-ling accident (FHA), and other events with high burnup fuel are conser-vatively shown to be well below values prescribed in 10 CFR 100.
3m mW m
Table 8.1 H.B. Robinson Unit 2, Cycle 10 Ejected Rod, Analysis, HFP BOC E0C Contribution (3) to Contribution (a) to Energy Deposition, Energy Deposition, Value (cal /gm)
Value (cal /gm)
A.
Initial Fuel Enthalp; (cal /gm) 86.5 102.5 8.
Generic Initial Fuel Enthalpy (cal /gm) 40.8 40.8 C.
Delta Initial Fuel Enthalpy (cal /gm) 45.7 45.7 61.7 61.7 D.
Maximum Control Rod Worth (pcm) 100 121 130 125 E.
Doppler Coefficient (pcm/0F)
-1.3(e) 0.99(b)
-1.5 (e) 0.89(b)
F.
Delayed Neutron Fraction,8 0.0061 1.00(b) 0.0053 1.03(b)
G.
Power Peaking Factor 2.7 4.1 H.
Power Peaking Factor Used(c) 4.0 5.0 Total 165.0(d) 172,0(d)
(a) The contribution to the total pellet energy deposition is a function of initial fuel enthalpy, x
maximum control rod worth, Doppler coefficient, and delayed neutron fraction. The energy gz deposition contribution values and factors are derived from data calculated in the " Generic
<2 Analysis of the Control Rod Ejection Transient..." document.
{7 (b) These values are multiplication f actors applied to (C + D).
S[
(c) The energy deposition dua to maximum control rod worth is a function of the power peaking factor.
(d) Total pellet energy deposition (cal /gm) calculated by the equation Total (cal /gm) = (C+D)(E)(F)
(e) For this Doppler coefficient, conservative values of -1.1 and -1.4 were assumed at BOC and E0C, respectively.
Table 8.2 H.B. Robinson Unit 2, Cycle 10 Ejected Rod Analysis, HZP B0C E0C Contribution (a) to Contribution (a) to Energy Deposition, Energy Deposition, Value (cal /gm)
Value (cal /gm) 16.7 A.
Initial Fuel Enthalpy (cal /gm) 16.7 B.
Generic Initial Fuel Enthalpy 16.7 (cal /gm) 16.7 C.
Delta Ini'.ial Fuel Enthalpy (cal /gmi 0.0 0.0 0.0 0.0 D.
Maximum Control Rod Worth (pcm) 600 33 600 33 E.
Doppler Coefficient (pcm/0F) 1.7(e) 1.04(b)
-1.8(e) 0.73(b)
F.
Delayed Neutron Fraction,8 0.0061 1.00(b) 0.0053 1.13(b) 6.8 G.
Power Peakir.g Factor 4.9 13.0 H.
Power Peaking Factor used(c) 13.0 TOTAL 34.0(d) 27.0)d)
(a) The contribution to the total pellet energy deposition is a function of initial fuel enthalpy, maximum control rod worth, Doppler coefficient, and delayed neutron fraction. The energy g
deposition contribution values and f actors are derived from data calculated in the " Generic ro.
Analysis of the Control Rod Ejection Transient..." document.
}y (b) These values are multiplication f actors applied to (C + D).
{y mU (c) The energy deposition due to maximum control rod worth is a function of the power peaking factor, (d) Total pellet energy deposition (cal /gm) calculated by the cquation Total (cal /gm) = (C+D)(E)(F)
(e) For this Doppler coefficient, conservative values of -1.0 and -1.4 were assumed at BOC and E0C, respectively.
29 XN-NF-83-72 Revision 2
9.0 REFERENCES
1.
XN-75-39, " Generic Fuel Design for 15x15 Reload Assemblies for Westinghouse Plants", September 1975.
2.
XN-NF-83-55, " Mechanical Design Report Supplement for H.B.
Robinson Extended Burnup Fuel Assemblies", August 1983.
3.
XN-NF-83-71, " Mechanical Design Report Supplement for H.B.
Robinson Part length Shielding Assemblies", September 1983.
4.
XN-75-27(A), Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors", Exxon Nuclear Company, June 1975.
5.
XN-75-27, Supplement 1, September 1976.
6.
XN-75-27, Supplement 2, December 1977.
7.
XN-75-27, Supplement 3, November 1980.
8.
XN-NF-77-57(A), " Exxon Nuclear Power Distribution Control for Pressurized Water Reactors - Phase II", Exxon Nuclear Company, January 1978.
9.
XN-NF-77-57(S), Supplement 1, June 1979.
10.
XN-NF-77-57(A), Supplement 2, September 1981.
11.
XN-CC-28, Revision 5, "XTG - A Two Group Three-Dimensional Reactor Simulator Utilizing Coarse Mesh Spacing (PWR Version)",
Exxon Nuclear Company, July 1979.
12.
WAPD-TM-678, "P0Q7 Reference Manual", Westinghouse Electric Corporation, January 1967.
13.
WAPD-TM-478, " HARMONY:
System for Nuclear Reactor Depletion Computation", Westinghouse Electric Corporation, January 1965.
14.
XN-75-38, "H.B. Robinson Unit 2, Cycle 4 Reload Fuel Licensing Data Submittal", August 1975.
15.
XN-NF-84-74, " Plant Transient Analysis for H.B. Robinson Unit 2,at2,300MWtwithIncreasedFNH",ToBeIssued.
16.
XN-NF-84-72, "H.B. Robinson-2 Limiting Break ECCS/LOCA Analysis with Increased Enthalphy Rise Factor, To Be Issued.
XN F00.553
4 XN-NF-83-72 i
Revision 2 ISSUE DATE 7/13/84 H. B. ROBINSON UNIT 2, CYCLE 10 SAFETY ANALYSIS REPORT DISTRIBUTION FT Adams G.l'Busselman JC Chandler LJ Federico NF fause TJ Helbling JS' Holm WV Kayser MR Killgore CE Leach JN Morgan FB Skogen GA Sofer IZ Stone RB Stout
.TT Tahvili HE Williamson Document Control (5)
CP&L (82)/TJ Helbling 4
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ATTACIDiENT 10 XN-NF-84-72 H. B. ROBINSON UNIT 2 LARGE BREAK LOCA-ECCS ANALYSIS WITH INCREASED ENTHALPY RISE FACTOR
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