ML18227C692
| ML18227C692 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 09/09/1975 |
| From: | Robert E. Uhrig Florida Power & Light Co |
| To: | Boyd R Office of Nuclear Reactor Regulation |
| References | |
| Download: ML18227C692 (62) | |
Text
P.O. BOX 013100 MIAMI,FLA. 33101
+lip~
Q FLORIDA POWER 5 LIGHT COMPANY September 9,
1975 L-75-435 Roger S.
Boyd, Deputy Director Division of Reactor Licensing Office of Nuclear Reactor Regulation U.
S. Nuclear Regulatory Commission washington, D.
C.
20555
Dear ~w. Boyd:
Re:
Turkey Point Plant Unit 3
Docket No.
50-250 Cycle 3 Reload Fuel Submittal and P oposed Amendment to Facility One"atin-License DPR-3i i ~
Florida Power 6 Light Company,,hereby, submits a'escription of the uel which will'be loaded into turkey Point Plant Unit 3.
=or CycLe 3 operation.
A proposed Technical Specificat'on amendment is also submitted,to incorporate, an increase in nominal system pressure from 1900 psia to 2100 psia.
Tn
~accordance with 10 CPR 50.30, three (3), signed originals and forty (40) conformed copies of'he, request to amend Appendix A of Facility Operating License DPR-31 are submitted.
Very truly yours, Robert E. Uhrig Vice P esident REU: iLfAS:nch Attachments cc; Hr. Norman C. Hoseley Jack R.
Newman, Esquire HELPING BUILD FLORIDA
7
STATE OF FLORIDA
)
)
ss COUNTY OF DADE,
)
ROBERT E.
UHRIG, being first duly sworn, deposes and says:
That he is a Vice President of Florida Power
& Light Company, the Licensee herein; That he has executed the foregoing instrument, that the statements made in this said instrument are true and correct to the best of his knowledge, information and belief; and that he is authorized to execute the instrument of said Licensee.
Subscribed and sworn to before me this day of
~:;
< n;.,;.',
1975 Notary Public xn and for the State of Florida at Large Ry Commission expires gY CONg~ecJ~~ ~~pgeg ~p~.
' 'i~
WADED T}6U~" " NSUaAiCE UND:RwetHRS
0
RELOAD SAFETY EVALUATION TURKEY POINT PLANT UNIT 3 ~
CYCLE 3
FLORIDA POWER AND LIGHT COMPANY
TABLE OF CONTENTS Title
~acae
1.0 INTRODUCTION
AND SUMi41ARY 2.0 3.0 4.0 REACTOR DESIGN 2.1 Mechanical Design 2.2 Nuclear Design 2.3 Thermal and Hydraulic Design POWER CAPABILITY AND ACCIDENT EVALUATION 3.1 Power Capability 3.2 Accident Evaluation 3.3 Incidents Reanalyzed TECHNICAL SPECIFICATIONS
- 4. 1 Proposed Amendments 4.2 Revised Technical Specification'ages
/
0
LIST OF TABLES Table Title Fuel Assembly Design Parameters Kinetics Characteristics Shutdown Requirements and Margins Results of Rod Ejection Analysis'
0
LIST OF FIGURES Ficiure Title Loading Pattern Control Groun Insertion Limits for Unit 3 Three Loop Operation Control Group Insertion Limits for Unit 3 Two Loop Operation
1
INTRODUCTION AND
SUMMARY
Turkey Point Unit 3 achieved initial criticality in
- November, 1972 and is now in its second cycle of operation.
The Unit is scheduled for a refueling shutdown on October 20, 1975, with Cycle 3 startup planned for early December.
The Turkey Point 3 Cycle 3 core loading pattern is shown as Figure 1.
Forty-eight of the Region 2 assemblies will be removed and replaced by twenty-four Region SA assemblies and twenty-four Region 5B assemblies (see Table 1).
An additional change in plant'peration for Cycle 3 is an increase in nominal system pressure from 1900 psia to 2100 psia.
lt This report presents an evaluation for Cycle 3 which demon-strates that the core reload will not adversely affect. the safety of the plant.
It is not the purpose of this report to present a reanalysis of all potential incidents.
Those incidents analyzed and reported in the FSAR which could potentially be affected by fuel reload have been reviewed for the Cycle 3 design described herein.
The increase in the nominal system pressure from 1900 psia to 2100 psia has also been evaluated.
The results of new analyses have been included and the justification for the applicability of.
previous results from the remaining aaalyses is presented.
been concluded that the Cycle 3 design does not cause the previously acceptable safety limits for any incident to be exceeded.
This conclusion is based on the assumption that:
(1) Cycle 2 operation is terminated after 8700+ 1000 MWD/MTU and (2) there is adherence to plant operating limit-ations discussed later in recommended modifications to the Technical Specifications.
It has Nominal design parameters for Cycle 3 are 2200 Mwt rated
- power, 2100 psia system pressure, 539 F core inlet temper-
- ature, and 5.56 kw/ft, average linear fuel power density (based on 144" active fuel length).
0 I
2.0 REACTOR DESXGN 2.1 Mechanical Desi n
The mechanical design of Region 5 fuel is dimensionally the same as Region 4 fuel.
Region 5 fuel has different enrichments as noted in Table l.
Other physical design aspects of Region 5 are the same as Region 4, except that the initial prepressurization level of the fuel rods has been decreased by 30 psi-One Region 5
assembly will contain Zr02-Y203 marker pellets.
Clad flattening time is predicted to be 23,500 EFPH for the limiting region (Regigq
- 3) using the current Westinghouse evaluation model<~).
Therefore, Region 3
has a nominal Cycle 3 allowed residence time of 6500 EFPH (assumes a Cycle 2 lifetime of 6700 EFPH and a Cycle 1 lifetime of 10,300 EFPH).
This analysis considers Cycle 3 operation at 2100 psia.
Region 5A contains some fuel made from UO2 powder which meets Westinghouse specifications but was fabricated by a process which differs from the standard Westinghouse powder process.
The Draft Regulatory Guide on Densifica-tion issued July 29, 1975, will be 'used to determine the amount of densification in Region 5A fuel.
Two fuel rods in one assembly of Region 5A will contain Zr02-Y203 marker pellets to permit the gathering of in-pile fuel stability data.'hese rods will be posi-tioned adjacent to the instrument tube in an assembly containing fuel prepared by the new Westinghouse powder process, in order that the location of the Zr02-Y20~ marker pellets may be monitored by the
'ovable ancore neutron flux detector.
Equivalent Zr02-CaO marker pellets have previously been suc-cessfully used in the Region 4 test assembly in Point Beach 1'(~).
The pellets are chemically inert and will have no adverse effect on the performance of the fuel rods
~
Westinghouse has had considerable experience with Zircaloy-clad fuel.
This experience is extensively described in WCAP-8183, "Operational Experience with Westinghouse Cores"(3).
This report is updated about every six months,.
2.2 Nuclear Desi n The Cycle 3 loading pattern results in a maximum F of 2.32 under normal operating conditions.
Table 9 provides a comparison of the Cycle 3 core kinetics characteristics with the current limit based on
I
2.2 Nuclear Desi n (Continued) previously submitted accident analyses.
Xt can be seen from the table, that most of the Cycle 3 values fall within the current limits.
These parameters are evaluated in Section 3.0.
Table 3 provides the control rod worths and requirements.
The required shutdown margin is Pased on previously submitted accident. analysis~4>.
The available shutdown margin exceeds the minimum required.
The control rod inseition limits have been revised to provide one set of insertion limits which are applicable for the entire cycle.
Revised control rod insertion limits for three loop and two loop operation are provided in Figures 2 and 3, respect-ively.
2.3 Thermal and H draulic Desi n No significant variations in thermal margins will result from the Cycle 3 reload.
The present DNB core limits have been found to be conservative.
3.0 POWER CAPABILITY AND ACCIDENT EVALUATION 3.1 Power Ca abilit This section reviews the plant power capability considering the consequences of those incidents examined in the FSAR using the previously accepted design bases.
It is concluded that the core reload will not adversely affect the ability to safely operate at 100% rated power during Cycle 3.
For the overpower transient, a maximum local rod power limit of 20.4 kw/ft corresponds to the fuel centerline temperature limit of 4700 F for Region 3 fuel.
This can be accommodated with margin in the Cycle 3
core.
The time dependent densification model~2>
was used for this evaluation.
The LOCA limit is met by maintaining F~ at or below 2.32.
3.2 Accident Evaluation The effects of the reload on the design basis and postulated incidents analyzed in the FSAR~4~
have been examined.
In most cases it was found that the effects can be accommodated within the conservatism of the initial assumptions used in the previous applicable safety analysis.
For that incident which was reanalyzed, it was determined that the applicable desi'gn basis limits are not exceeded, and therefore the conclusions presented in the FSAR are still valid.
This reload can typically affect accident analyses input parameters in three major areas:
kinetics characteristics, control rod worths and'ore peaking factors.
Cycle 3 parameters in each of these three areas were examined as discussed below to ascertain whether new accident analyses are required.
The effect of the increase in pressure from 1900 psia to 2100 psia is to increase DNB margin; however, the accident evaluation was performed at the more limiting (1900 psia) pressure.
Technical Specification changes recommended for the increase in pressure are given in Section 4.
Kinetics Parameters
A comparison of Cycle 3 kinetics parameters with current limits is given in Table 2.
Most of the Cycle 3 coefficients remain within the bounds of current limits.
The small changes in core physics parameters have a neglible effect on transient analysis.
Therefore, no additional accident analysis is required due to changes in these parameters.
Control Rod Worths Changes in control rod worths may affect shutdown margin, differential rod worths, and
0
3.2 Accident Evaluation (Continued) ejected rod worths.
Table 3 shows that Cycle 3
shutdown margin is adequate.
Table 2 shows that the reactivity insertion rate due to control rod withdrawal is not greater than was previously analyzed.
Cycle 3 ejected rod worths are less than the current limits.
Core Peakin Factors Evaluation of peaking factors for the rod out of position and dropped RCCA incidents shows that DNBR is maintained above 1.3.
For the dropped bank incident, the turbine runback setpoint is sufficient"to prevent a
DNBR less than 1.30.
Peaking factors following control rod ejection were less for Cycle 3 than the current limits.
3.3 Incidents Regnal zed The end of cycle full power rod ejection incident was re-evaluated since the average fuel temperature conservatively assumed at the initial hot spot linear power density exceeded that previously used in this incident* by approximately 260 F.
Since the peak fuel and clad temperatures obtained for the previous analysis were well below the limiting criteria(5),
the results with the 260oF increase (shown in Table 4) still do not cause the criteria to be exceeded.
The limiting case for the rod ejection incident actually occurs at the beginning of life for which the parameters are as described in the previous analysis(5).
- This fuel temperature was used only for the end of life ejected rod analysis.
Fuel temperatures previously used in other incidents are unaffected.
0
4.0 TECHNICAL SPECIFICATIONS 4.1 Pro osed Amendment The required changes to the Technical Specifications for the operating pressure change and the reload fuel scheme are as described below and as shown in the accompanying page changes bearing the date of this letter in the lower right hand corner.
Unit 3 fuel residence time is revised from 23000 EFPH
.to 23500 EFPH.
The overtemperature hT used to derive the formula is changed such that the factor (P-1885) is revised to read (P-2085).
This Limiting Safety System Setting (LSSS) change is consistent with operation at 2100 psia.
Pa e 2.3-3 The Low Pressurizer Pressure LSSS is changed such that the factor 1715 psig is revised to read 1915 psig.
This LSSS change is consistent with operation at 2100 psia.
Sections 3.2.l.b, 3.2.1.c, and 3.2.1.d are revised to differentiate between Units 3 and 4.
For Unit 3, Cycle 3, one control rod insertion limit is valid for the entire cycle.
Fi ures 3.2-1 and 3.2-1(a)
These figures are revised to apply to Unit 4 only.
Fi ures 3.2-1(b) and 3.2-1(c)
These figures are added to provide control group inser-tion limits for Unit 3, Cycle 3.
The basis for specification 3.2.1 is amended to dif-ferentiate between Units 3 and 4.
For Unit 3, Cycle 3, one control rod insertion limit is valid for the entire cycle.
~Pe ev The list of figures is amended to indicate that Figures 3.2-1 and 3.2-la apply to Unit 4.
Also, Figures 3.2-1b and 3.2-1c are'dded to the list.
0
4.2 Revised Technical S ecification Pa es Thxs section contains the following revised Technical Specification pages and figures:
Page v
Page 1-6
'.Page 2.3-2 Page 2.3-3 Page 3.2-1 Figure 3.2-1 Figure 3.2-1(a)
Figure 3.2-1(b)
Figure 3.2-1(c)
Page B3.2-1 Page B3.2-la
I
LIST OF FIGURES
'P~iu re
- 2. 1-1 Title Reactor Core Thermal and Hydraulic Safety Limits, Three Loop Operation 2.1-2 3.'l-l 3.1-2
- 3. 2.-1 Reactor Core Thermal and Hydraulic Safety Limits, Two Loop Operation
.Reactor Coolant System Pressure Limits Radiation Induced Increase in Transition Temperature for. A302-B Steel Control Group Insertion Limits for Unit 4, Cycle 2,
. Three Loop Operation 3.2-la'.2-lb Control Group Insertion Limits for Unit 4, Cycle 2, Two Loop Operation Control Group Insertion Limits for Unit 3, Cycle 3; Three Loop Operation
- 3. 2-3;c 3 ~ 2 2
3 ~ 2 3
3.2-4 4.12-1 Control Group Insertion Limits for'Unit 3, Cycle 3, Two Loop Operation
'equired Shutdown Margin
-'. Hot Channel Factor Normalized Operating Envelope II Maximum Allowable Local KW/FT Sampling Locations
.6.1-1'-
'. 1-2
- 6. 1-3 B3.2-1 B3. 2-2 Management. Organization Chart Plant Organization Chart Organization of Opeiating Support Groups Target Band on Indicated Flux Difference as a
function of Operating Power Level.
Permissible, Operating Band on Indicated Flux Difference as a Function of Burnup.
V 9/9/75
1.16
'INTERIM LIMITS 1.16.1 Fuel Residence Time Limit The fuel residence time fqr Unit 3 shall be limited to 23,500 effective full power hours
('EFPH) under low pressure operating conditions.
The fuel residenc'e time for Unit'4 shall be.limited to 30,000 EFPH.
1.16 '
Reactor Coolant Pumps Operation The reactor 'shall not.be operated with less than three reactor'coolant pumps in operation.
1.17'OW POWER PHYSICS TESTS Low power physics tests are tests below a nominal 5X of rated power which measure fundamental characteristics of the reactor core and I
related instrumentation.
1-6 9/9/75
0
'nit No.
3 Reactor Coolant Tem erature
. ~
'"'vertempera-ture.hT
< LT 'l- 0.0174(T-566.6)
+ 0.000976(P-2085) - f(hq)
I hT Indicated hT at rated power, F
T
~ Average temperature,'
P Pressurizer
- pressure, psig I, ~
f(5q) ~ a function of the indicated difference between top and bottom. detectors of the power-range nuclear ion chambers; with gains to be selected
~m'h' based on measure/
instr'ument response during.
startup tests such that:
I
~
I ~
~
For (q q ) wi'thorn +10 percent and -14 percent
'here q
and q
are the percent
'power in'the top and bottom hal'ves-of the core respectively, and.
q
+ qb is total 'c'ore.power in percent of rated.
power, f(hq) 0; I
~
~
\\
r.
'I For each percent.,that the magnitude of (q - q )
t b
exceeds
+10 percent, the Delta-T trip set point shall be automatically reduced by 3.5 percent of" 4
its value at. interim power.
1 For each percent that the magnitude of (q - q )
t b
~ 'xceeds
-14 percent, the Delta-'
trip set point shall be automatically'educed by 2 percent of.'ts value at interim power.
1 Xl (Three Loop Operation)
,1.120; (Two Loop Operation) '.88 1
P, ~
'I I I.
2 ~ 3 2 Unit No.
3 9/9/75
0 I
~ l Unit No.
3
~ Over-power AT
< hT 1 ~ 09 K "
K (T
T' f(5q)
.;:dT 1.
dt 2
5T Indicated 5T at rated
- power, F
0 T
~
Average temperature, F
Indicated average temperature at nominal conditions and rated power, F
Kl 0 for decreasing aveiage temperature, 0.2 sec./F.for increasing average t'emperature KZ 0; 00134 for T 'qual to or more than
.T'
'0 for T less than T'T dt Rate of change
'of temperature F/sec 7
f(hq) ~
As defined above
.Pressurizer Low.Pressurizer pressure '- equal:to or greater than 1915 psig.
High Pressurizer pressure equal to or less than
- 2385 psig.
High Pressurizer water level equal to or less than i
92% of.full scale.
'e'actor'oolant Flow
.. Low reactor coolant flow equal to or greater than
."90% of normal indicated flow Low reactor coolant pump motor frequency equal to or greater than 56.1 Hz '.."."
Under voltage on reactor coolant pump motor'us <<equal
. to or greater than 60% of normal voltage Steam Generators Low-low steam generator water'evel equal to or greater than 5% of narrow range instrument scale 203-3 Unit No.
3 9/9/75
3.2 CONTROL ROD AN WER DXSTRIBUTXON LIMITS distribution limits.
e
~Ob ective:
To ensure (1) core subcriticality after.a react'or trip,
..(2) a li'mit on potential reactivity insertions from a hypo-thetical control rod ejection, and (3) an acceptable
- core, power distribution 'during power operation.
'I
. a.
Whenever the reactor is critical, except for physics tests and control rod exercises, the shutdown control rods shall be fully withdrawn.
b.
Por Unit 4, Cycle 2, whenever the reactor is critical, except for physics tests and control.rod
.exercises,'he control group rods shall'be
'no further inserted t
than the'limits shown by the solid lines on Pigure 3.2-1 for three loop operation and on. Figure 3.2-1(a) for two loop operation.
Control rod insertion limits for Unit 3, Cycle 3, are
'shown on Figures 3.2-1(b) and 3.2-1(c)..',
For Unit 4, Cycle 2, subsequent cycles as shall be adjusted as the end-of-core life Pigure 3.2-1.
k after 70% of the second and defined by burnup, the limits a linear function of burnup'oward as shown by the dotted lines on d.
The Unit 4, Cycle 2 end-of-core life limit shown on Figure 3.2-1 and the Unit 3, Cycle 3 control. rod insertion limits shown on Figure 3.2-1(b) may be revised on the basis of physics calculations and physics data obtained during startup and subsequent operati'on.
e.
Part length rods shall not be permitted in the core except for low power physics tests and for axial offset calibration tests performed below 75% of rated power.
9/9/75
I
UNIT 4, CYCLE 2.
CONTROL GROUP INSERTION'LIMITS FOR THREE LOOP. OPERATION 100
- . LIMIT UP TO 705 OF 2ND CYCLE LIP'E ROD POSITION LIHIT AT END OF 2ND.CYCLE B
80 60.
h 40 20 D
I tl 0 0'0 40 60 SO PONER LEVEL,'ERCENT RATED.
100
'IGURE 3.2-1.
t 9/9/75
I
~
~
I I
~
s
~
'.UNIT 4, CYCLE 2
- CONTROL'ROUP 'INSERTION LIMITS FOR TWO LOOP OPERATION 100
~ ~ I ~
.:11: ill.l I",'lll!j'i':;ll t
~t t I
I fjfi ILII
, I 4
.b ->>tt
~ s
~
L i~
Me I a
~
r jl fq.
qt f -f:;! r 1
~
~
~
I prt l>>fl aI'g e i ~.t I vr ~
t ~ '
~
sW 4 t'ien
'b
~
4
~
~ J i I t
~
~ I jl'jj'tjjjjj;
~ ~
~
~ ~
~ ' I'l
~
~ 4>>LL lr'4
~ ~ ~
~
I tb~
s>>
~ '
I
~ ~
I'e LL
~
b
~
~
11.
jr ~
ski.
~>>I I 11' L
Iq
>> ~ 4 s,j 1 I!~
W>> r
~ ~
'j'
~ a tQ C$
S QJ O
60 40 jr j l.[ 1 I ~
i ~
jr fbi t>>
r
! Ig.
L ~ l lit jt
~
~
"i" j'[)!
i '- ll r tie "er
~
I
~ ~ ~
VS !4rt pr
~
.Ll
':!i jijj j iliB!
~
~ ~
~
a
~ I ~ LI ~ 4. ~
't ~ I
~
1 4 ~ 44 ~ ~
~
i tL!
1
~ l s lb
~ p
~
4 Tf
~
~ st t
'I I
a ~ ~
beni
- i, < f!j. 1!jr.
~ It I
~ ~ I1 t1.'
~ 1fr l.~ 4 j.I, lt I i ~
~
~ I
~ l
~
~ 1'll 1".>1
~ 1 >>
~ i I I srj>>i 1
~ ~
e I
~
~ ',
~ ~
- elj I ~
~ I!la
,lb
~
~ ~
~ ~i ~
r:11
~ ~ I ~
~ '
~ ~
~ I 'I T ~ r
~ ~
f)j~ sill Cf)j '
~ ~
I
~ ~
s
'I' r
~ b br e'
C r>>44 4
~
I I ~
4 I I 4
~ ~
I
~ It!
~
~ rL j pr st 4
I ~
~ t i ~ ~ ~
~ I ~
~ I I ~ 't
~ I'4 l i
~ ~
't>>
~>> 44 4
'4 E+
s>>
~
I
~'I>>
4 pi
~
4
~
~ 4
~
4+ I 4+i TT 4 ~ I tt 4 ~
~>>
~
4
>>I t
1
~,
~P
~
I I'P>>
~
i ~,
4 ~
~tr P
>>rt p tir4' 4 '
~ Li
- b. ~
t~
Ls
~
~s
~i!.'
a It
~ 'P
~ I 4
~jh
~ ~ bt rh
~.I L ~
4 ~
st r a',
a
.Pj&I 1
QP I
i rj ~
Lt fij
~
~
I
"~ 's I>>IT I
't'4 t ~
' '1 li:,.,
I ~
~ ~
~ 4
~ I I 4
I ~
4
~
~
~
~
I
~
~
~
~ I
~ t
- fr.
r>>1 ~
' t't I T
~
~ I Pi I j+LL
+
i ~ ~
'1 sr Lt I
~ s
~
~
~
~
I
~
~ ~ '
4
~
~ 4 ~
$ ~
~
~4b a ~ ir>>.i s, ~
~ l s',.=p
~
~
~
~
e 1'-jl 'f~tjj
~'-!
I'
~ ~
iL!I
~ t Ltf
~ Il
~ e ~
'ei i
~ ~
!D I tl ~
La ~
- I' t ~ k ~
Pt r'
~ r
~
t I
~t 4 ~ 4 ~
~
r ~ 1 jl~ l 20
~
~ I tilt b ~ ~
~ l ~
it j.! j'.1t 9'.:',I jll', a.,',
ej l I
!!4 ~
~
~
4 ~ t
~ ~ I t r
~ 4
~
~ ~
Lt
~ I ~ ~
14+
4 ~
I>> '
~
~ l
~
1 s'>>I
>>f)
!L;; '.:
~ ~ II
~ ~ ~ I
~
~ i
~
~
~
~ ~
t I
~ i
~
~
~
~
s I.i
~ stb"t
'L!4..
i ~ I Pg!.
I ~
~
I'
'I r IL I
~ PL ~
r>>r>>>>s
~
b4 ir; LL~
I t
I'
>>s 4'j
~ I
~
r
. p41 ts l4'r'I 4>> 4 s&t I ~
4 l ~
~
. I rtI
~
~ ~ ~
~ ~
~
1 ~
~
~ 1 VI 444 ptr s>>f 44 e
e
~
~ I
~
~ ~
~ t I 'll
~j"'
~ ~
~ s
~
I!4 'l
~ '
~ i
~ rl
~
~
ib'f
~ 'iT ti rl I-'Lr la ~ '
i l ~
~
1 ji lt,'
I 4"
I
~'itI
~
e ~
I' b
rr>>
L>> ~
~ r ~ ~ j. ~ ~ 4
~ ~
~
4Li 00'0 40
..60 80 100 POWER LEVEL, PERCENT RATED FIGURE 3. 2-1(a)
e 0
UNIT 3, CYCLE 3
CONTROL GROUP INSERTION LIMITS'OR THREE LOOP OPERATION l00 h
I
~
80 '
hilt~
L L
Bank Cg:"
Lf W~r T 60,
~,l>>
40 Ll,i f= Bank 0 20 0'
~
~
~
~
20 40 60 80 100
'OWER LEVEL, 'PERCENT RATED FIGURE 3.2-1t'b) 9/9/75
I I'
le
, ~
4
'4 UNIT 3 CYCLE 3
CONTROL GROUP INSERTIOi'l LIHITS FOR TWO LOOP OPERATION
~ 4 5~
100 l
5 l5 80 Bank C
~ 5 60
~
4 QL 40 Bank 0 4'0 4 ~
'l
]'4
'0 4J 20 4
40 60 80
.100
'(
~
I
,'I POWER LEVEL, PERCENT RATED 4
~
FIGURE 3.2-1{c)'/9/75
0
B3. 2 BASES FOR LIMITING CONDITIONS FO."
'OPERATION, CONTROL AND POWER DISTRIBUTION LIMITS I
Reactivity changes accompanying changes in.reactor powe'r are compensated by control rod motion.
Reactivity changes associated with xenon,
- samarium, fuel depletion, and large changes in reactor coolant temperature (operating temperature to cold shutdown) are compensated by changes in the soluble boron concentration.
During power operation, the shutdown groups are fully withdrawn and control of reactor power is by. the control groups.
A reactor.
trip occurring during-power'peration'ill put the reactor into the hot shutdown condition.
The control rod insertion limits provide for achieving hot shutdown by reactor trip at, any time, assuming the highest wozth control rod remains, fully withdrawn, with sufficient margins to meet the assumptions used in the
'accident analysis.
In addition, they provide a limit on the maximum (1) inserted rod worth in the unlikely event of a hypothetical rod e5ection, and provide for acceptable nuclear peaking factors.
The solid line shown on Figure 3.2-1 meets the shutdown requirement for the first 70% of second
. and subsequent cycles for. Unit 4, except for two loop operation.
The Unit 4 end-of-c'ore life limit may be more restrictive, as shown by the conservative estimate represented by the dotted line.
Figures 3.2-1(b) and 3.2-1(c) meet the shutdown requirements of Unit 3, Cycle 3.
The Unit 4 end-of-core-life limit and the Unit 3, Cycle 3, rod insertion, limits may be determined on the basis, of startup and operating data to provide a more realistic limit which will allow for more flexibilityin'peration and still
~assure compliance with the shutdown requirement.
Figure 3.2-1(a) shows the shutdown requirements for Unit 4 second cycle two loop operation.
The 15 maximum shutdown margin requirement occurs at end-of-core life and is based on the value used in analysis of the hypothetical steam break accident.
Early in core life, less shutdown margin is required, and Figure 3.2-2 shows the 1
shutdown margin equivalent to 1.77% reactivity at end-of-core-life with respect to an uncontrolled cooldown.
All other accident analyses are based on 1% reactivity shutdown margin.
B3.2-1 9/9/75
0
The overlap between successive control banks is allowed because the control rod worth.is lower near the top and bo'ttom of the core than in the center.
Positioning of the part-length rods is governed by the'requirement to main-tain the axial 'power shape within specified limits or to accept an automatic cutback of the overpower hT and overtemperature hT set points (see Specification 2.3).
Thus, there is no need for imposing a limit on the physical positioning of the part-length rods.
B3.2-la
/9 75
4
Table 1
TURKEY POINT 3 - CYCLE 3 FUEL ASSEMBLY DESIGN PARAtiETERS
~Re ioo Enrichment (w/o U 235)
. Density (4 Theoretical)*
Number of Assemblies Approximate Burnup at Beginning of Cycle 3
(HMD/f1TU) 1.86 2.56 3.11
- 2.56 93.8 92.8 92.0' 94.6 4
52 '2 131 00 23500 21000 7600 5A
~
e 4 2.60 2.90 95.0 95.0 24 24 1
~A11 regions except Reg..c..
5 are as-built values; Region.5 is the nominal
- value, however, an average density of 94.5$ theoretical was used in therma1 evaluations.
1 4 ~
F 0
. Tab1e 2
I
~
' i e,
- TURKEY POINT UNIT.3 '
KINETICS CHARACTERISTICS Cu.rrent Liinit 'QcIe 2
. 2250 sia 1900 sia Cycle 3'100sia
'oderator Temperature
~
Coefficient, (hp/'F)x10
'3.5 to +0.3+
-3.5 to 0
~ er t
-3.5 to 0 Dopp1er Coef(icient,
(~p/ F)x10
~
'Delayed Neutron Fraction, eff'
)
Prompt Neutron Lifetime
(>sec)
Maximum Differential Rod i<orth of Tio Banks Moving Together
..'t, HZP (pcm/in)~
.'50 to.'72
..$ 0 to
. 59
. 50 to
. 72 14 to.18 14 r
. 20 (max.)
80 80 80
-X.6 to -1.0
-2.6,to -1.0 -2.6 to -1.0
~
~
"The positive coefficient does not occur at operating conditions.
- +pcm,-=10
.,bp-
-5
0:
Tab]e 3
TURKEY POINT 3 - CYCLE 2 AND 3 SHUTDOWN RE(UIREHEHTS AHD HARGINS'ycle 2
EOC Contrbl Rod llorth
~ ho All Rods inserted Less Mo'rst Stuck Rod 6.01 (1).Less los...,..
5.41 Cycle 3
'OC EOC p
5.28
- 5. 57
.'.87 6.19 Control Rod Requirements 5 hp Reactivity Defects (Doppler, Tavg,
.. Void, Redistribution)
Rod Insertion Allowance (2) Total Requirements C
A
/
I Shutdown Margin';
1 2
X hp) 2.61
'..70 3.31 2.10
.50 2.14 3.14 2.52
.50 3.
02'.55-Re uired Shutdown Mar in 5 hp 1.77
'l.00 1.77
0
~
l 0
r
~ ~
Table. 4 RESULTS OF R00 EJECTION'ANALYSIS KGHMUM FUEL AID CLAD TEMPERATURES AT THE HOT SPOT EHO OF LIFE,.HOT FULL POWER Pr evious Anal sis 0
~CcIe 3 Fuel Average Temperature('F)'uel Centerline Te'~lperature('F'C1ad Average Temperature('F) 3180 4275
~ '1835 3440 2095
Figure. 1 TURKEY POINT'UNIT:3 CYCLE 3 LOAOING PATTERN R
.P.
N M'
K.
J
'H G
F E
8 z
g 9
9 3
9 5'
FUEL REGION
0 e
I ~
'\\
I Figure 2
UNIT 5 CYCLE CONTROL GROUP INSERTION LIMITS'OR THREE LOOP OPERATION
~
~
100 80'ank.C
~ \\
H
~
~
n5 lg S
+
"'Cl S
CLl O
60 40 Bank 0 20
.0 0
la'
~
~
~
20 40 60 80 100 POWER LEVEL, PERCENT RATED
E~
~ I+
F>guie 3
UNIT 3 CYCLE 3 COttTROL GROUP INSERTION LIHITS FOR TWO LOOP OPERATION
~
~ r
~
~ q Q ~
100
'80
~ ~
Bank C.
~
Bank D
20 0
I fi
'0 20 gp
~ ~
80 100 I
~ II.0 POWER LEVEL, PERCENT RATED
~ I
v 0
~"
REFEREHCES 20 3.
George, R.'A., et al "Revised Clad Flattening Model",
WCAP 8377 (Proprietary) and WCAP 8381 (Hon Proprietary}, July 1974
- Hellman, D. M. (Ed.), "Fuel Densification Experimental Results and Model for Reactor Operation",.
WCAP 8218-P-A, March 19?5 (Propnetary}
and'CAP 8219-A, March 1975 (Hon Proprietary)
I
- Plocsdo, V. J.
and Schreiber, R.
E. "Operational Experience with Westinghouse Cores",
WCAP 8183, Revision 2, Hovember 1974 Final Safety Analysis Report, Turkey Point Units Ho.
3 and 4 "Fuel Densification, Turkey Point Plant Unit Ho. 3" WCAP 8074
{Proprietary) and WCAP 8075 (Hon Proprietary),
February 1973 0
/'