1 '- ·i.~· · 2019. 7. 31. · hot spot fluid temperature (4 inch) core power ... 1975...
TRANSCRIPT
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11-_p,f.ERGENCY CORE COOLING SYSTEM ANALYSIS
REQU:RED BY 10 CFR 50.46
·1,
I I I 1-
SEPTEMBER 1, 1974 REVISED APRIL 11 2 1975
REVISED JUNE 5, 1975
SURRY ~OWER STATION UNIT NOS. 1 AND 2
DOCKET NOS. 50-280 AND 50-281 LICENSE NOS. DPR-32 AND DPR~37
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I I I I I I I I I I I I I I I I I
!I .. ,
I.
IL
III.
IV.
v.
VI.
TABLE OF CONTENTS
List of Tables - Small Break List of Figures - Small Break List of Tables - Large Break List of Figures - Large Break
Introduction
Loss of Reactor Coolant From Small Ruptured Pipes or From Cracks in Large Pipes Which Actuate the Emergency Core Cooling System
Major Reactor Coolant System Pipe Ruptures (Loss of Coolant Accident)
Conclusions
References
Proposed Changes to the Technical Specifications
Page No.
i ii
.. iii iv
1
5
25
77
78
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I I LIST OF TABLES
SMALL BREAK
I TABLE NO. TITLE
I II-1 TIME SEQUENCE OF EVENTS
I II-2 RESULTS AND ASSUMPTIONS
I I I I I I I I I I I I I I i
/
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I I I I I I I I I I I I I I I I I I -,
FIG. NO.
II-1
II-2
II-3
II-4
II-5
II-6
II-7
II-8
II-9a
II-9b
II-lOa
II-lOb
II-lla
II-llb
LIST OF FIGURES SMALL BREAK
TITLE
Safety Injection Flowrate
Reactor Coolant System Depressurization Transient (4 inch)
Core Mixture Height (4 inch)
Clad Temperature Transient (4 inch)
Core Steam Flowrate (4 inch)
Rod Film Coefficient (4 inch)
Hot Spot Fluid Temperature (4 inch)
Core Power
Reactor Coolant System Transient (3
Reactor Coolant System Transient (6
Core Mixture Height (3 inch)
Core Mixture Height (6 inch)
Clad Temperature Transient (3 inch)
Clad _Temperature Transient (6 inch)
ii
inch)
inch)
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I I 'LIST OF TABLES
LARGE BREAK
I TABLE NO. TITLE
I III-1 TIME SEQUENCE OF EVENTS FOR DOUBLE ENDED COLD LEG GUILLOTINE BREAK (DECLG)
I III-2 ASSUMPTIONS AND RESULTS FOR DOUBLE ENDED COLD LEG GUILLOTINE BREAK (DECLG)
I III-3 CONTAINMENT DATA
I I I I I 'I I I I I I I -,.
iii
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I I I I I I I I I I 'I I I I I I I I I
FIG. NO.
III-la
III-lb
III-le
III-2a
III-2b
III-2c
III-3a
III-3b
III-3c
III-4a
III-4b
III-4c
III-Sa
III-Sb
III-Sc
III-6a
III-6b
II1-6c
III-7a
III-7b
III-7c
III-Sa
III-Sb
III-Sc
LIST OF FIGURES LARGE BREAK
DOUBLE-ENDED COLD LEG GUILLOTINE BREAK (DECLG)
TITLE
Fluid Quality - DECLG (CD= 1.0)
Fluid Quality - DECLG (C = 0.6) D
Fluid Quality - DECLG (CD= 0.4)
Mass Velocity - DECLG (CD= 1.0)
Mass Velocity - DECLG (CD= 0.6)
Mass Velocity DECLG (CD= 0.4)
Heat Transfer Coefficient - DECLG (CD= 1.0)
Heat Transfer Coefficient - DECLG (CD= 0.6)
Heat Transfer Coefficient - DECLG (CD= 0.4)
Core Pressure - DECLG (CD= 1.0)
Core Pressure DECLG (CD= 0.6)
Core Pressure DECLG (CD= 0.4)
Break Flow Rate DECLG (CD= 1.0)
Break Flow Rate - DECLG (CD= 0.6)
Break Flow Rate - DECLG (CD= 0.4)
Core Pressure Drop - DECLG (CD= 1.0)
Core Pressure Drop - DECLG (CD= 0.6)
Core Pressure Drop - DECLG (CD= 0.4)
Peak Clad Temperature - DECLG (CD= 1.0)
Peak Clad Temperature DECLG (CD 0.6)
Peak Clad Temperature DECLG (CD 0.4)
Fluid Temperature - DECLG (CD= 1.0)
Fluid Temperature - DECLG (CD= 0.6)
Fluid Temperature - DECLG (CD= 0.4)
iv
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I I I ,. I I
I I· I I 1· I I I I I I
FIG. NO.
III-9a
III-9b
III-9c
III-lOa
III-lOb
III-lOc
III-lla
III-llb
III-llc
III-12a
III-12b
III-12c
III-13a
III-13b
III-13c
III-14
III-15
TITLE
Core Flow - Top and Bottom - DECLG (CD= 1.0)
Core Flow - Top and Botton - DECLG .(CD= 0.6)
Core Flow - Top and Bottom DECLG (CD = 0.4)
Reflood Transient
Reflood Transient
Reflood Transient
DECLG (CD = 1.0)
DECLG (CD= 0.6)
DECLG (CD= 0.4)
Accumulator Flow (Blowdown) DECLG (CD = 1.0)
Accumulator Flow (Blowdown) - DECLG (CD= 0.6)
Accumulator Flow (Blowdown) - DECLG (CD= 0.4)
Pumped ECCS Flow (Reflood) - DECLG (CD= 1.0)
Pumped ECCS Flow (Reflood) DECLG (C = 0.6) D
Pumped ECCS Flow (Reflood) DECLG (CD 0.4)
Containment Pressure - DECLG (CD = LO)
Containment Pressure DECLG (CD= 0.6)
Containment Pressure DECLG (CD= 0.4)
Unit No. 1, Cycle 2, Maximum Peaking Factor .vs. Axial Core Height (+6, -9 Delta Flux Band)
Unit No. 2, Cycle 2, Maximum Peaking Factor vs Axial Core Height (+6, -9 Delta Flux Band)
V
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I I I I I I I I I I I I I I I I I I
! I
I. INTRODUCTION
The re-evaluation of ECCS cooling performance as required by the
Order for Modification of License for Surry Power Station, Unit Nos. 1
and 2, issued by the Atomic Energy Commission on December 27, 1974, and
as specified by 10 CFR 50.46, 1 "Acceptance Criteria for Emergency Core
Cooling Systems for Light Water Reactors" is presented herein. This
analysis was performed with the March 15, 1975 version of the Westing
house evaluation model which includes modifications as specified by the
Regulatory Staff in Reference 8. The implementation of these modifications
in the evaluation model is described in Reference 13. The analytical
techniques utilized in the analysis are in compliance with Appendix K to
10 CFR 50, and are described in the topical report, "Westinghouse ECCS
Evaluation Model-Summary," WCAP 8339, 2 dated July 1974. The Nuclear
Regulatory Commission Staff acceptance of the "Westinghouse ECCS Evaluation
Model" is presented in Reference 14. The results of the analysis show
that the calculated cooling performance of the emergency core cooling
system (ECCS) following postulated loss-of-coolant accidents conforms to
the criteria set forth in paragraph b of 10 CFR 50.46.
A loss-of-coolant accident may result from a rupture of the reactor
coolant system or of any line connected to that system up to the first
closed valve. Ruptures of very small cross section will cause explusion
of coolant of a rate which can be accommodated by the charging pumps.
Should such a small rupture occur, these pumps would maintain an opera
tional level of water in the pressurizer, permitting the operator to
execute an orderly shutdown. A moderate quantity of coolant containing
J ..
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I I I I I I I I I I I I I I I I I I I
such radioactive impurities as would normally be present in the coolant,
would be released to the containment.
Should a larger break occur, subcooled fluid is expelled from the
break rapidly reducing the pressure to saturation. Depressurization of
the reactor coolant system causes fluid to flow to the reactor coolant
system from the pressurizer, resulting in a pressure and level decrease
in the pressurizer. For a postulated large break, reactor trip is ini
tiated when the pressurizer low pressure setpoint is reached, while the
safety injection system (SIS) is actuated by coincident pressurizer low
pressure and low level. A high containment pressure signal serves as a
backup by initiating a safety injection signal which in turn trips the
reactor. The consequences of an accident are limited two ways:
1. Reactor trip and borated water injection supplement
void formation in causing rapid reduction of the
nuclear power to a residual level corresponding to
delayed fissions and fission product decay.
2. Injection of borated water also ensures sufficient
flooding of the core to prevent excessive clad
temperatures.
The safety injection system, even when operating on emergency power,
limits the cladding temperature to below the melting temperature of
Zircaloy-4 and below the temperature at which gross core geometry dis
tortion, including clad fragmentation, may be expected. In addition,
the total core metal-water reaction is limited to less than one (1) per
cent. This is valid for reactor coolant piping ruptures up to an in
cluding the double ended rupture of a reactor coolant loop. Conse
quences of these ruptures are well within .those for the hypothetical
-2-
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I I I I I 1: I I I I I I I I I I I I I
accident and are, therefore, well within the limits of 10 CFR 100.
Before the break occurs the unit is in an equilibrium condition,
i.e., the heat generated in the core is being removed via the secondary
system. During blowdown, heat from decay, hot internals and the vessel
continues to be transferred to the reactor coolant system. The heat
transfer between the reactor coolant system and the secondary system
may be in either direction depending on the relative temperatures. In
the case of continued heat addition to the secondary, secondary system
pressure increases and the main safety valves may actuate to reduce the
pressure. Make-up to the secondary side is automatically provided by
the auxiliary feedwater system. The safety injection signal stops nor
mal feedwater flow by closing the main feedwater control valves, trips
the main feedwater pumps and initiates emergency feedwater flow by
starting the auxiliary feedwater pumps. The secondary flow aids in the
reduction of reactor coolant system pressure. When the reactor coolant
system depressurizes to 600 psia, the accumulators begin to inject water
into the reactor coolant loops. The reactor coolant pumps are assumed
to be tripped at the initialization of the accident and effects of pump
coastdown are included in the blowdown analyses.
The water injected by the accumulators cools the core and subsequent
operation of the low head safety injection pumps supply water for long
term cooling. After the contents of the refueling water storage is
emptied, long term cooling of the core is accomplished by switching to
the recirculation mode of core cooling, in which the spilled borated
water is drawn from the containment sump by the low head safety injection
pumps and returned to the reactor vessel.
-3-
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I I I. I 'I I I I I I I I I I I I I I I
The containment spray system and the recirculation spray system
operate to return the containment to subatmospheric pressure.
-4-
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I I I I I I I I I 1--I I I I I I I I I
II. LOSS OF REACTOR COOLANT FROM SMALL RUPTURED PIPES OR FROM CRACKS IN LARGE PIPES WHICH ACTUATE THE EMERGENCY CORE COOLING SYSTEM
Methods of Analysis
For breaks less than 1 ft 2 the WFLASH10 digital computer code is
employed to calculate the transient depressurization of the reactor.
coolant system, as well as to describe the mass and enthalpy of flow
through the break.
The WFLASH program used in the analysis of the small break loss
of coolant accident is an extension of the FLASH-4 code developed at
the Westinghouse Bettis Atomic Power Laboratory. The WFLASH program
permits a detailed spatial representation of the reactor coolant system.
The reactor coolant system is nodalized into volumes interconnected
by flowpaths. The broken loop is modeled explicitly with the intact
loops lumped into the second loop. The transient behavior of the system
is determined from the governing conservation equations of mass, energy
and momentum applied throughout the system.
The use of WFLASH in the analysis involves, among other things,
the representation of the reactor core as a heated control volume with
the associated bubble rise model to permit a transient mixture height
calculation. The multi-node capability of the program enables an ex
plicit and detailed spatial representation of various system components.
In particular it enables a proper calculation of the behavior of the
loop seal during a loss of coolant transient.
Safety injection flow rate to the reactor coolant system as a
function of the system pressure is used as part of the input. The
-5-
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I I I I I I I I I I I I I I I I I I I
safety injection system was assumed to be delivering borated water to
the reactor coolant system 25 seconds after the generation of a safety
injection signal.
For these analyses, the safety injection delivery considers pumped
injection flow which is depicted in Figure II-1 as a function of reactor
coolant system pressure. This figure represents injection flow from the
safety injection pumps based on performance curves degraded five (5)
per cent from the design head. The twenty-five (25) seconds delay in
cludes time required for the emergency diesel generator to assume its
load. The effect of low head safety injection pump flow is not considered
since their shutoff head is lower than reactor coolant system pressure
during the time portion of the transient considered. Minimum safeguards
emergency core cooling system capability and operability has been assumed
in these analyses.
Peak clad temperature analyses are performed with the LOCTA-IV4 code
which determines the reactor coolant system pressure, fuel rod power
history, steam flow past the uncovered part of the core and mixture
height history.
Results
The results of the limiting break size-are given in Table II-2.
The worst break·size (small break) is a 4 inch diameter break. The
depressurization transient for this break is shown in Figure II-2. The
extent to which the core is uncovered is shown in Figure II-3.
-6-
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I I I I I I I I I I I I I I I I I I I
During the earlier part of the small break transient, the effect of
the break flow is not strong enough to overcome the flow maintained by
the reactor coolant pumps through the core as they are coasting down
following reactor trip. Therefore, upward flow through the core is
maintained. The resultant heat transfer cools the fuel rod and clad to
very near the coolant temperatures as long as the core remains covered
by a two phase mixture.
The maximum hot spot clad temperature calculated during the transient
is 1898 degrees Fahrenheit including the effects of fuel densification
as described in Reference. 11. The peak clad temperature transient is
shown in Figure II-4 for the worst break size, i.e., the break with the
highest peak clad temperature. The steam flow rate for the worst break
is shown on Figure II-5. When the mixture level drops below the top of
the core, the steam flow computed in WFLASH provides cooling to the upper
portion of the core. The rod film coefficients for this phase of the
transient are given in Figure II-6. The hot spot fluid temperature for
the worst break is shown in Figure II-7.
The core power (dimensionless) transient following the accident
(r~lative to reactor scram time) is shown in Figure II-8.
The reactor shutdown time (3.4 sec) is equal to the reactor trip
signal time (1.2 sec) plus 2.2 sec for rod insertion. During this rod
insertion period the reactor is conservatively assumed to operate at
rated power.
-7-
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I I I I I I I I I I I I I I I I I I I
Additional Break Sizes
Additional break sizes were also analyzed. Figures ll-9a and ll-9b
present the reactor coolant system pressure transient for the three (3)
inch and six (6) inch breaks, respectively, and Figures 11-lOa and II-lOb
present the volume histroy (mixture height) for both breaks. The peak
clad temperatures for both cases are less than the peak clad temperature
of the four (4) inch break. The peak clad temperatures for both cases
are given in Figures 11-lla and 11-llb.
The time sequence of events for all small breaks analyzed is shown
in Table 11-1 and Table 11-2 presents the assumptions and results for
these analyses.
The results of several sensitivity studies are reported in WCAP-8342. 12
These results are for conditions which are not limiting in nature and
hence, are reported on a generic basis.
-8-
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----~---------~----
I I.O I
TABLE II-1
TIME SEQUENCE OF EVENTS FOR SMALL BREAKS
BREAK SIZE (INCHES)
EVENT
.Start of LOCA
Reactor Trip Signal
Top of Core Uncovered
Accumulator Injection Begins
Peak Clad Temperature Occurs
Top of Core Covered
3
TIME AFTER START
0.0
21.2
690
971
992
1011
4 6
OF LOCA (SECONDS)
o.o 0.0
13.4 8.0
333 108
530 265
574 276
853 315
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-------------~~----
I ...... 0 I
TABLE II-2
ASSUMPTIONS AND RESULTS FOR SMALL BREAKS
Results
Peak.Clad Temp. (°F)
Peak Clad Location (Ft.)
Local Zr/H2
0 Rxn (max) (%)
Local Zr/H2
0 Location (Ft.)
Total Zr/H20 Rxn (%)
Hot Rod Burst Time (sec)
Hot Rod Burst Location (Ft.)
Calculation
NSSS Power MWt 102% of
Peak Linear Power Kw/ft 102% of
Peaking Factor
Fuel region analyzed (Most Limiting)
Unit No. 1
Unit No. 2
3 IN
1702
11. 0
1.17
11.0
<0.3
N/A
N/A
Region
3
3
4 IN
1898
11.0
1.66
10.5
<0.3
N/A
N/A
2542
14.44
2.32
6 IN
1559
10.5
0.91
10.5
<0.3
N/A
N/A
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--- -
\
I I I I
FIGURE II,-1
SAFETY INJECTION FLOWRATE
I 2500
I r,
I 2000 ,_
I -<(
en· 0... - 1500 -
I LI.I 0::: :::::, en en LI.I 0:::
I 0...
en (.)
1000 -
0:::
I 500 -
I I
. -- r 'I I
500 0
0
I II
100 200 300 SI FLOW {LB/SEC)
I
~00
I I I I I I I
,J. .\ 1: ..
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I I I I I I
....--
I <t rr., c.. -UJ
I et:: ::::) rr., C1) UJ et:: c..
I rr., (...) a::
1. I I I I I I I I I
FIGURE II-2
RCS DEPRESSURIZATION TRANSIENT (4 inch)
2500
2000
1500
1000
500
0
0 250 500 Tl ME (SECONDS)
-·-···- - ·- -- - . -- - - ·- -- ------ .. _ --- ... -
750 I 000
I I
I
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I I I I I I I I I I I I I I I I I I I
FIGURE II-3
CORE MIXTURE HEIGHT (4 inch)
20
16 i
-f-,-u.. 12 -f-,-::c C!J
LU ::c
LU a:: 8 0 c:...>
0 0 250 500 750
TIM~ (SECONDS)
------------·~- ------ -- .. ~ --··-------------
'
1000
; I
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I I I I I I I I I I I I I I I I I I I
-LL 0 -Cl) UJ IX :::, ..... ~ IX UJ Q. :::E UJ ..... Q ~ ...J (.)
FIGURE II-4
. CLAD TBMPDATUBES TRANSIBliT ( 4 inch)
2000
1500
1000
500
300 ~00 500 600 700. 800 '900
Tl!ME (SECONDS)
' .... llji:., •
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I I I I I I I I I I I I I I I I I I I
-u UJ Cl) -al ....I -UJ I-<( cc:
3 0 ....I LL
::E <( UJ I-Cl)
~00
300
200
I 00
0
-100
-200
-300
-~00 0
FIGURE II-5
STRAM FLOW ( 4 inch)
-·
250 500
TIME (SECONDS)
750 I 000
; • .-- ~.!""-I'~--- -
/
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------- .. --·--------·-
-LL. 0
I CN I-LL.
I cc: :c -:::, I-al -I-z: w
.. \ c..:>
I ., .. ·..=.,· . LL.
°' ·- --~
f?.: ,. w
~ ·-· 0 J-:;,- c..:>
.. _ .. X _J
LL.
Q 0 cc:
I 03
8
6
lj.
2
102
8
6- --- I
'
lj.
2
10 1
300 .. ~00 500 600
TIME (SECONDS)
..:~~1L .. r·- -
ROD r)lti!1tmrit~ er,·~~)
800 900
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I I I I I I I I
'I I I ·I I I
-1 I I I I
-LL 0 -LU cc: ::,
~ cc: LU ~ ::E: LU I-
C
::, ....J LL
I-0
-~ Cl)
I-0 :c
FIGURE II-7
HOT SPOT FLUID TEMPDATUBE (4 inch)
2000
1500.
1000
500
0 300
' : j .
-- ifoo · · · · ·· --s-oo-··· -· Boo- 100
Tl~E (SECONDS)
--.-· .... - '·- ·-~. -·.
,. r.
___ , __
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-------------------
I I-' Q:) I
e::::
I oo 8 6
lj.
2
UJ 10-1 ~ 8 0...
I< II<( 3: -en I-
6
lj.
2
i IO-~ 6
lj.
2 '
tQ-1 2 4- 6 s I o0 2
TOTAL RESIDUAL HEAT (WITH 4-% SHUTDOWN)
4- 6 s to 1 , I
2 4- 6 s 102 2 4- 6 8 I 03
TIME AFTER SHUTDOWN ( SECONDS)
FIGURE II-8
CORE POWER
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I I
\ \
I FIGURE II-9a
RCS DEPRESSUJUZATION T&ANSIEBT (3 inch)
I I 2500
I 2000
! I -<(
Cl)
I a.. -L&J 0:::
1500
=> Cl)
I Cl)
L&J 0::: a.. 1000 Cl) (..)
I 0:::
500
I I
0 200 4-00 600 800 1000 1200
I TIME (SECONDS)
I I I I I I -19-
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I I I FIGURE 1I-9b
I RCS DEPRESSURIZATION TRANSIENT (6 inch)
I 2500
I I -
2000
<(.
I Cl) c.. - 1500 UJ a:::
I ::::, Cl) Cl) UJ a::: c.. 1000
I Cl)
u a:::
I 500
I ·o - .
I 0 500 600 100 200 300 ~00
TIME (SECONDS)
I I I I I I -20-
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--- -~---
' I I I FIGURE II-lOa
I CORE MUTUBE HEIGHT ~3 inch)
I 20
I I
16
I -I-LL 12 -I-
I :c c:,
UJ :c
UJ 8
I 0.::: 0 c..>
I 4-.
I 0
.I 0 200 4-00 600 800 IOOO 1200 TIME (SECONDS)
I I I I I I --21-
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I I I FIGURE II-lOb
CORE MIXTURE HEIGHT (6 inch)
I I 1· 16
I -I
I-LL. -I-
12
:::c <:!'
I LU :::c LU
8 0::: 0
I (.)
I I 0
·o .100 200 300 4-00 500 600
I TIME (SECONDS)
I I I I I I -22-
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I I I I I I I I I I I I I I I I I I I
-LL 0 -LJ.J 0:: ::, I-<( 0:: LJ.J 0...
i:fi I-a <( _J (.)
FIGURE II-lla
CLAD TEMPERATURE TRANSIENT (3 inch)
2000
1500
1000
500
600 700 800 900 1000 1100 1200
TIME (SECONDS)
-23-
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I I I I I I I I I ·I I I ·1 I I I I I I
u. 0 -w 0:::: => I-< 0:::: w Q.. :::: '
w I-
0 < _J c:..:,
FIGURE II-llb
CLAD TEMPB2ATURE TRANSIENT (6 inch)
2000
1500
1000
500
0 -~· --- .
.. 0 100 200 300 ~00 500 600
TliME (SECONDS)
-24-
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I I I I I I I I I I I I I I I I I I I
III. MAJOR REACTOR COOLANT SYSTEM PIPE RUPTURES (LOSS OF COOLANT ACCIDENT)
Method of Thermal Analysis
The description of the various aspects of the LOCA analysis is
given in WCAP-8339. 2 This document describes the major phenomena modeled,
the interfaces among the computer codes and features of the codes which
maintain compliance with the Acceptance Criteria. The SATAN-IV, WREFLOOD,
COCO, and LOCTA-IV codes used in their analysis are described in detail
in WCAP-8306, 3 WCAP-8171, 5 WCAP-8326 6 and WCAP-83054 , respectively. The
containment parameters used in the containment analysis code to determine
the ECCS backpressure are presented in Table III-3.
Results
Table III-2 presents results for the double ended cold leg break for
three (3) large break sizes (discharge coefficients (CD)). This range of
discharge coefficients was determined to include the limiting case for
peak clad temperature from sensitivity studies reported in WCAP-8356. 7
The analysis of the loss of coolant accident was performed at 102 per
cent of 2542 megawatts thermal. The peak linear power and core power used
in the analyses are given in.Table III-2. The equivalent core peaking
factor at the license application power level (2441 MWth) is also shown
in Table III-2.
The limiting power peaking envelope applicable to this analysis,
along with the actual power peaking values expected during operation of
-25-
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I I I I I I I I I I I I I I I I I I I
each Surry unit, are shown in Figures III-14 and III-15. The actual
power peaking values were determined using the procedures described
in WCAP-8385 9 assuming a +6, -9 per cent delta flux (~I) band. The
limiting envelope was determined on the basis that:
1.
2.
3.
Extensive sensitivity studies presented in WCAP-83567
and WCAP-8472 15 have shown that the cosine is the
worst power shape as long as the change of FQ with
elevation is maintained as shown on Figures III-14
and IIi-15 in the region from 0.0 feet to 11.0 feet.
The computed peak clad temperature does not exceed
2200°F for the case of a cosine power shape with
FQ = 2.10.
The location of the line segment between 11.0 feet
and 12.0 feet is the same as that presented in
WCAP-8356. 7 This line segment was confirmed by the
small break results presented in Section II of this
report.
For results discussed below, the hot spot is defined to be the
location of maximum peak clad temperature. This location is given in
Table III-2 for each of the three (3) discharge coefficients used for
the double-ended break.
The time sequence of events for the three (3) discharge coefficients
used in the large break analysis is shown in Table III-1.
Figures III-1 through III-13 present the transients for the principal
parameters for the double ended cold break for the three (3) discharge
coefficients used. Each of these parameters is enumerated below:
-26-
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------
I I
1.
I I I
2.
I I I 3.
I I I
4.
I I 5.
I I I 6.
I I I
Fluid Quality - Figures III-la, band c show the fluid
quality at the clad burst and hot spot locations
(location of maximum clad temperature) on the hottest
fuel rod (hot rod) for the three (3) discharge co-
efficients used.
Mass Velocity - Figures III-2a, band c show the mass
velocity at the clad burst and hot spot locations on
the hottest fuel rod for the three (3) discharge co-
efficients used.
Heat Transfer Coefficient - Figures III-3a, b ~nd c
show the heat transfer coefficient at the clad burst
and hot spot locations on the hottest rod for the
three (3) discharge coefficients used. The heat
transfer coefficient shown was calculated by the LOCTA
IV code.
Core Pressure - Figures III-4a, band c show the cal
culated pressure in_the core for the three (3) dis
charge coefficients used.
Break Flow Rate - Figures III-Sa, band c show the
·calculated flow rate out of the break for the three (3)
discharge coefficients used. The flow rate out the
break is plotted as the sum of both ends for the guillo-
tine break cases.
Core Pressure Drop - Figures III-6a, band c show the
cal~ril~ted core pressure drop for the three (3) dis
charge coefficients used. The core pressure drop
-27-
I '·
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I I I I I I I I I I I I I I I I I I I
7.
shown is from the lower plenum, near the core, to the
upper lenum at the core outlet.
Peak Clad Temperature - Figures III-7a, band c show
the calculated hot spot clad temperature transient and
the clad temperature transient at the burst location for
the three (3) discharge coefficients used.
8. Fluid Temperature - Figures III-Sa, band c show the
calculated fluid temperature for the hot spot and burst
locations for the three (3) discharge coefficients used.
9. Core Flow - Figures III-9a, band c show the calculated
core flow, both top and bottom, for the three (3) dis
charge coefficients used.
10. Reflood Transient - Figures III-lOa, band c show the
calculated reflood transient for the three (3) dis
charge coefficients used.
11. Accumulator Flow - Figures III-lla, band c show the
accumulator flow for the three (3) discharge co
efficients used. The accumulator delivery during blow
down is discarded until the end of bypass is calculated.
Accumulator flow, however, is established in refill re
flood calculations. The accumulator flow assumed is the
sum of that injected in the intact cold legs.
12. Pumped ECCS Flow (Reflood) - Figures III-12a, band c
show the calculated flow of the emergency core cooling
system for the three (3) discharge coefficients used.
-28-
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I I I I I I I I I I I I I I I I I I I
13. Containment Pressure - Figures III-13a, band c show
the calculated pressure transient for the three (3)
discharge coefficients used. These pressure transients
are based on the data given in Table III-3.
The clad temperature analysis is based on a total peaking factor
of 2.10. The hot spot metal reaction reached is 5.6 per cent, which is
will below the embrittlement limit of 17 per cent, as required by 10
CFR 50.46. In addition, the total core metal-water reaction is less
than 0.3 per cent for all breaks as compared with the 1 per cent criterion
of 10 CFR 50.46.
The results of several sensitivity studies are reported in WCAP-8356. 7
These ~esults are for conditions which are not limiting in nature and
hence are reported on a generic basis.
-29-
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----~--------------
I w 0 I
TABLE 111-1
TIME SEQUENCE OF EVENTS FOR DOUBLE ENDED COLD LEG GUILLOTINE BREAKS (DECLG)
DISCHARGE COEFFICIENT (CD)
EVENT
Start of LOCA
Reactor Trip Signal
Safety Injection Signal
Accumulator Injection
End of Blowdown
Pump Injection
Bottom of Core Recovery
Accumulator Empty
(GD=l.O) (CD=0.6)
TIME AFTER START OF LOCA
0.0 p.o
0.63 0.65
1.43 1. 78
11.5 14.4
23.9 23.3
26.5 26.8
37.3 36.9
43.4 45.8
(CD=0.4)
(SECONDS)
0.0
0.66
2.2
· 18. 0
30.6
27.2
40.99
: 49. 6
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- - - - - - - - - - - - - - - - - - ·-
I (.;.) I-"' I
TABLE III-2
ASSUMPTIONS AND RESULTS FOR DOUBLE ENDED COLD LEG GUILLOTINE BREAKS (DECLG)
Results
Peak Clad Temp. (OF)
Peak Clad Location (Ft.)
Local Zr/H2o Rxn (max) (%)
Local Zr/H2o Location (Ft.)
Total Zr/H2o Rxn (%)
Hot Rod Burst Time (sec)
Hot Rod Burst Location (Ft.)
Assumptions
Power. (MWt) 102% of
Peak Linear Power (Kw/ft) 102% of
Peaking Factor Used
Accumulator Water Volume (Ft3)
Fuel region analyzed (Most Limiting)
Unit No. 1
Unit No. 2
DECLG ( CD= 1. 0)
1986
7.5
3.90
7.5
<0.3
48.2
6.0
Cycle
2
2
DECLG (CD=0.6)
1995
7.00
4.19
7.25
<0.3
32.4
6.00
2542
13.12
2.10
975.0
Region
4
4
DECLG (CD=0.4)
2090
6.75
5.60
7.0
<0.3
27.2
5.75
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I I I I I I I I I I I I I I I I I I I ~
TABLE III-3
CONTAINMENT DATA
NET FREE VOLUME
INITIAL CONDITIONS Pressure Temperature RWST Temperature Service Water Temperature Outside Temperature
SPRAY SYSTEM I Number of Pumps Operating Runout Flow Rate (each) Actuation Time
SPRAY SYSTEM II RECIRCULATION SPRAY FROM SUMP
Number of Pumps Operating Runout Flow Rate (each) Actuation Time Heat Exchanger UA (per pump) Service Water Flow (per exchanger)
-32-
1.863 X 106 ft3
9.35 psi 75 40 35.0 9.0
2 3200
20.0
2 3500
125 3.5
6100
OF oF OF OF
gpm secs
gpm secs
x 106 BTU/hr-°F gpm
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I I I I I I I I I I I I I I I I I I I
STRUCTURAL HEAT SINKS
Thickness (In)
Concrete 6.
Concrete 12.
Concrete 18.
Concrete 24.
Concrete 36.
Carbon Steel 0.375 Concrete 54
Carbon Steel a.so Concrete 30.
Concrete 24.
Carbon Steel 0.366
Stainless Steel 0.426
TABLE III-3
CONTAINMENT DATA
Area (Ft 2 ),
-33-
including uncertainty
6972
57,960
40,470
10,500
4410
46,887
25,075
11,284
167,165
3399
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I I ·1 ·1 I I I I I I I I I I ·1
I I I I
, .... 18111
1--
-0 •r- . JISOII '.:l
..--LL
4-. C
>. +-' •,-r-/Cl ::I O'
. 2500G
o.,
0 ••
6.oc/
8 8 -
FIGURE III·- la
(J\t lrnllct-;l ru'-'1 rud)
8 8 C\,I
Time (Sec)
8 8 .. ..
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I I 'I I I I t ·-
I t.nal
I ··-I ·o ......
~ r--LL
I· 4--0
>, +->
I •.-,--<t.l :::s c::r
I .... I •••
I I .I I I 1·
f.H:UI!E l'Ti-· lh
FL C TD c, !J /o.i, T' 1 'Y ·-· ·--·· -· - ... ~ . - ..... .
(At hottest fll(~l rod)
...... _._.. ........ ~-------+-------+------c---·-+---------11
I-toll-------:-+----------- --·-·-· ·-·------..J....------~ ------~ 6.oo--' y-_ I
~7.00 . ,
t--------+-·---------~- ~- ----·-·-------·--'-·--- ---1--------- .
-------,---+------·------· -----------·-------..--------
•• •-8 8 ....
Time (Sec)
DECLG ( c1r·O. G)
. -35-
8 8 -a i ' I
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I I ·1 I .I I I I I -c
.,.-::; r-Li..
I I 4-. C
>, .:-, •,-
I
I r-res ::I er
I I· I I I I I I
, ....
., ... .1'50IO
,,..
...
ff.PJD _quM,I'l'i'
(At hottest fuel rod)
. '
' lt--------+------.._----------t-------l~-----1
. .
..__ _____ . --+-·----···•' --·-----·--····-! -. -·· --·---------t----------+-------t !
' . I . I ! _j_
'-------+-·----. -·-·-1-· -------·· . -------+------1
. I I . •• ~
I Ii ..
L. . I i
Time (Sec)
-36-
I • ' I
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I· I I I 1. I (.)
CJ (/)
IN~ 1::: .c ,... ..
I ---~ t·· •r·
(._l
I 0
r-· C..!
> (/)
I (,1
C1J =~
I I I I 1·
I i I I I
nG DO
ICI0,00 ----1~. ---1-----1-----t------
100,DO
D.0
-tao oo
-200,00
-no.oo
----·-······---·- -- -· ...... ----·· . . l-.·--· f
··-----·--- ---··---·-··-·---- ~·-----t
----------·--·--· -·-·-··--···-·- ·-····------- ~----·----·-------t
V 7.50 1
v6.co'
-----------·-- ------- r----·----------------
• • 8 8 ...
Time (Sec)
T' ''('- (' , (' 1 ()) J.L:..,L .. ~·1/"'···.
-37-
8 i
8
' • •
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,------
I I I I
I
I. I
I I I I .--..
u (1)
V1
I ~4J ll-......_ s:.-:
15 .--
I ---· > . . p ..... CJ
I 0 r-Q)
::-=-(()
Vi
I ct)
~
I I I I I I I I I
••••
tcill .•
O.Q
. -sm •
-mJ.IIO
FTCUHE U l···?h
~1/\.<:s v1,:1,nc1 TY -·-···- - -- --- .. . . ... - -
--------r---------· --·-···-···· ··--·--·--·----=-·--'--·--------.
1-------+---:·-----'-·-------'--~---··--·- - .. --------
.o •
,
8 8 -
1 . I.
----__ -_ . . ----_ ----T--
8 8 "'
Time (Sec)
-38-
8 8 4"'
8 8 •
8 i
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I ·-.... I I
) I I I ....
... (.)
><I N
~
I I
til H
H
ol
H
,-..,, ;~
l ~
>i
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Time (Sec)
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I I I I FIGURE III-4a
I CORE PRESSURE
I I 25
I 2000
-I
<( - 1500 Cl)
0... ·-UJ
I c,::
. :::, Cl) 1000 Cl) UJ c,:: 0...
I 500
I I 0 5 10 15 20 25
TIME {SECONDS)
1· I DF.CLG (~•LO)
I ·1 ------------.L.-.------
I I -43-
--- --- -
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~---
I I I I
FIGURE 1II-4b
I CORE PRESSURE
I I 2500
I 2000
I ......... < Cl) Ii.
1500
I -LU c:::: =::, . Cl)
1000
I Cl) LU c:::: Ii.
I 500,
I (}·
0 5 10 15 20 25
1· TI ME (SECONDS)
·1 I DECLG (CDa0.6)
I I I -44-
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I I I I
FIGURE III-4c
I, CORE PRESSURE
I I
2500
I 2000
I -<(
1500 rr., a..
I LJ.J cc: :::, rr.,
1000 rr.,
I LJ.J cc: a..
I 500
I 0 -
I' 0 10 20 30 ~o TIME (SECONDS)
'I I DECLG (CD•0.4)
I I I -45..'.,
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I I I I I I I I I I' I I I -, 'I I I I I
FIGURE III-Sa
BREAK FLOW RATE
1--,-,------_;_ __ ....;._ __ ~
r-
u 5 UJ Cl) -::E: ~ ij
5 10 15 20 25 TIME (SECONDS)
DECLG (~ml.O)
..;.46-
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I I I I FIGURE III-Sb
I BREAK FLOW RATE
I 7 -
I 6
I, - 5 ::t" 0 -
I -u 4, UJ (/) -:=:: al
I ....J 3 -3: 0 ....J
I u,. 2 :=.::: <t UJ 0::: al
I 0
I 0- 5 10 15 20 25
I TIME. ( SECONDS)
'I I
DECLG (~•0.6)
I I I -47-
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:I
I I I FIGURE III-Sc
I BREAK FLCN BATE
I 7
I 6
I <:j-
0
5 -(.)
I I.LI en -::E:
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I -3 0 ....J
3 LL
I ::..::: <( I.LI 0::: a:i
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I .I
01
I .. 20 30 ~o 0 10
TIME (SECONDS)
I I DECLG (CD•0.4)
I I I
-48-
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I I I I I I I I -
en c... -
I c... 0 a::: Cl
a::: c...
I UJ a::: 0 u
I I I I I I I I I
· FIGURE 111-6&
CORE PRESSURE DROP
50 -..---------------------
25
0
-25
-50
-75 · 0 5 10 15 20 25
TIME (SECONDS)
-49'-
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I I I I I I I I I I I I I I I I I I I
-en c... -c... 0 0:: Cl
0:: c... UJ 0:: 0 u
0 5
FIGURE III-6b
CORE PRESSURE DROP
10 15
TIME (SECONDS)
DECLG (C -0.6) D
-50-
' - - ..... ; .
20 25
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----- l ·1 I I I
FIGURE III-6c
I CORE PRESSURE DROP
I I 50
I I - 25
Cl)
0.. -~1 0..
0 0 0::: Cl
0::: 0..
I UJ 0::: 0 -25 (.)
I I
-50
0 10 20 30 lJO
TIME (SECONDS)
I I I
DECLG (~ca0.4)
I I I -51-
~---
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-Time (Sec)
DECLG (C ,0 J .O) D
-52-
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8 8 ..
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I I I I I I I I I I I I I I I I I I I
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r
Time (Sec)
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i 5,75
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I I I I FIGURE II1-9a
I CORE FLOW - TOP AND BOTTOM
I 7500·
I 5000
I 2500 -(..) I LU
Cl) -::::E al
0 ....J
I -3: 0 ....J u. -2500
BOTTOM
I LU 0::: 0 (..) -5000
I -7500 -
I -10000
I 5 25 0> 10 15
TIME (SECONDS) 20
I I
DECLG (~•1.0)
I I
•• ~.-=. ...
I -58-
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I I I I FIGURE III-9b
I CORE FLOW - TOP AND BOTT~
I 75
I .......... 50
N
I 0 ..... ..__. -u
LU 25
I en -::E: IJl .....J -
I 3: 0 .....J LL
0
LU 0::
I 0 u
-25
I -50
I -75 0 5 10 . 15 20 25
I TIME (SECONDS)
I I I I I -59-
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-- - - -
I I I FIGURE III-9c
I CORE FLOW - TOP AND BOTTOM
I 7500
I 5000
I I
2500
-(..) I
UJ Cl) -::E CD
0 _J -
I 3: 0 _J
u... :-2500 UJ
I a::: 0 (..)
I -5000
I -7500 I
I -10000
I· 0 10 20
TIME (SECONDS)
30
·1 I I I
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I I I I I
··- t'0.000
·I l WiC 17 ~00
I 3.0GOO 15 000
I r.sa ~ 12.:;oo
!5 u
I > ....
2.0000 ID 000
I t 5000 7 5000
I 1 0000 5.0000
I :,uc::J 2.5000
I 0 C o.o
I I I I I I
.r r I- l Oa
-----. -----------·---+·- ...... ·---·-·. ·-· "-·· ""l" ---- ---- .. ···--··-+-----····-··
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........................ ~- ·-·- .............. _. _________ J ________ i--------1
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1 I
l i ... ----····1 ·--·· .... - ... j
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i
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Time (Sec)
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g ...
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8 8 •
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I I I I I ,.ooao lO 000
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-:., CJ C) '\·
Time (Sec)
DECLG ( Crt·O . r,)
-62.-
0 0
8 .., 8 g ,,,
8
i
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I I I I I I I I I I I I I I I I I I I
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3.0000
Q Z.5000 ....
u ... 2 0000
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15.000
tl'SOO
to.ooo
7.5000
5.onoo
l.5000
•••
• ci
FIGTHrn lJI--10c
Time (Sec)
DECLG (C ,·(). l1) . l)
-63-
DOWN<>( fYIER Lt\JE.L
8
! 8 I
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I I I I I I I I I I I I I I I I I I I
~~~-~----
7000 -
6000
5000 -(..) UJ Cl) lJOOO -:::E in -' - 3000 3: 0 -' LL
2000
IOOO
FIGURE IIl-lla
ACCUMULATOR FLOW (ILOWOOWN)
5 I (j 15
TIME (SECONDS)
DECLG (~11:11.0)
-64-
20 25
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(SONO:l3S) ll~ll
9l Ol 91 . 01 9
I.
0 ----0
0001
OOOl
000€
oooti
I ooog.
0009
-OOOL
"T1 r 0 :::e:
-r gJ
3:
-en IT1 (")
-
I I I I I I I I I I I I I I I I I I I
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----
---
----
---
' i i I I '
li i: ll t' r! !l ii fl II '
·rt-
' i
FLOW
(LB
M/S
EC)
;~
1 I
1· i
o I
s:
-----
:---
----
--::
,---
--,-
,--"
'T""
'I'-
.;._
__
.. · .... --·--·
···----
--..-
----
-....
...1
I ,
• .
.,,..,
O: i '
~ ~ .....
~ __ .
..... __
____
...,. ___
____
____
___ ....
..,. __
_. ~
--
--
---
·---
--
-·-
--
--
--··
-
.
~·' T'
, "
r-j '"
' I ! R
,.. ~ . I
, \
I) r
'
!1. Q
--f'
'
1 .L
i:. !
::i '-
1\;1
.,.
"ct 1
'~uN
,.ucR
i-"
.,.!-J
"-
·I I I 1 . . J
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I I I I I I I I I I I I I I I I I I I
FIGURE ITT-12a
PlJHPED ECCS FLO\~ - (REFLOOD)_
-67-
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I I I I I I I I I I I I I I I I I I I
<..) LJ..I Cl) --"'t L~)~J;;~~~;
FIGURE III-12h
I
i, .
.JI
DECLG (CD"'0.6)
-68-
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I I I I I I I I I I I I I I I I I I
~-
<...) w C/) -M f-LL ..__
3': 0 _J
ll.
FIGURE III-12c
~,1t;+~tlj1~~~~1•~~~=iJ••,.~fi!~;is;~~ ···r 0···~1 ·*' ··· '[;-' I "'''\.="'j' .,,,,,y•· I
,,~0c,I. ·· · ···r, lr,·E,~·,·~ ,~- · i'iiCf -+· ·· I ·£ · 1'· · · · ·· ·. ·•
=····· ..... j .1.+1±1 Jf l-' "1···''·1~~~~1 +' .
·:_~JJ~4~lr[~~ff 4'.f~~\,Y~~: .. ~~~~f8 ··'"·~•L. ,.·. ·' +•·,, , ... , ..... '··t'M( (SEC)·•"""· l+··1 ~1 ~ . I ~~s-: ~~L.!~~ 1:~:-r=·T~:r.-:-~-l-::-::~ J :-.~---,-- :,-:)f~)~l:~~,:~~~ ::\r-~~n:;:~ ·:_::}~~-:: :=~~--~-·
DECLG (Cn"'0.4)
-69-
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I I I
FlGUH.E lll-13a
I CONTA INW.C:NT PRESSURE ---- - - --·--------- -------·---
I I I j
0.. ..........
I Q) ~ ~ V)
t·1 (l)
I ~-
CL
I I I I I I ---
I I J
I -70-
J
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I I I FH;Ulrn 111-Uh
I CO!\TAINMENT PRESSURE
I,
I I ct
•r-u,
CL .........
I Cl) s... ::l u, u, Cl)
I .S... CL
I I I I I I
. ---A_
I DECLG (Cn""0.6)
I I I ~
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------
I I I
FICURE ll] -1 Jc
CON'l.'A rNNENT ]'lrn::smrn -------·-·-------~---- -·- ··--------
I I I I &:
•r-(./') • -· 1- •
0.. ...._~
I ClJ !... ::i (./') (./')
ClJ !...
I CL
I I I I I Time (Sec)
I -J\--
DECLG (C0==0.4)
I I I -72- ,
I
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I I I I I I I I I I I ;-._
N -..J
O'
I ~
I I I I I I I
2.4
2.2
2.0
1. 8
1.6
1.4
1. 2
1.0
0.8
0.6
0.4
0.2
0.0
FIGURE III-14
SURRY 1, CYCLE 2 MAXIMUM PEAK.ING FACTOR vs AXIAL CORE HEIGHT
(+6, -9% Delta Flux Band)
::: : ': '. '. ::::::::;:!I !\J.' '. 111 :11J1·f H1J tli-L;\: rJt tlUT::lL11 iJ;-: ~:i;_1r:; __ lln !!.U. :;u ii~'. :: 1
: :··· c:..:_;_:_/---'...:..'..>-'-....;.:.'·-· i,•, -1·-···· ., i+111 r-,,,1+!1 --,--11--, rl:-t +l+W-1---ti11n1-1,:-r 1-:1---+:-r,+11-·-,'1-'·--d. !, ..• : - : 1 : ·:: i,: - , . , - .1,; L
1-J ,- I;!. 1 I 1J :1 t.i" 1-JrJ: _l:H_E- -1+1:1- Jffi':--'fH-H 1-1-1-1.:1 +-H, 1-1-+1 1 11-1- --,· u-11-H ,.1n '-+111- -1-1-1+ +,+• .. ·",
;lc\l .. , 1 ,1,! 1:1: ,
111 , , __ •• 1\-+J-'"-i',j·:ffi \+.J:l lt+H--f 1--t+!_! +j-r'.,:±:'.""t:\. •J··ti: ___ ._,·c+.t-~IJ41_]J"t'~-+T'Y~-:t_r:rJ·::·:.::;
:·! t •• , , :- i- . ; • i 1., i ., L i !-: k --,- 1.t..:L .... -·- J_,_~ ! ...1 _j_ 1i:t:11, _j r. -l -r..t t-t.., .. 1...!--+--J.I -1 .. ; .• _ .. rr. _, 1=-t r:1Jt ·-,·
0 2 4 6 8 10 12
CORE HEIGHT, ft.
-73-
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I I I I I I I I I I
,....._ N 1:0
I I I I .I I I I
2.4
2.2
2.0
1.8
1.6
1.4
1.2
1.0
0.8
0.6
0.4
0.2
0.0
;;.11
FIGURE III-15
SURRY 2, CYCLE 2 MAXIMUM PEAKING FACTOR vs AXIAL CORE HEIGHT
(+6, -9% Delta Flux Band)
•11 i1: ,, ' 1i'- l!.1 1 11,1.·1. !I,!! ! l,r 11 1.1-,.1_11111-1-1- :t·-1+-- - 1-H + I -ill-+ -ij·-~-j+H-~lt-f: + -0 +1+ --1+~-.-rL_ •:t"·+ : __ i:_·,. 1,i._1'._ -!_::_·-.:_! .' .• !•.'.' _1 __ .1 .. · , -lj-1-- 11 -iT;.,:1Tl p: -,, _q:1_-1: 1:+:1:lit ++:H:t ·:rj.t!t .. ,:·1_::J=:[i.:LTJ
ill'···: ,. :.:., j;, -f-r-1--H-t i, fJ -r!--l--r1-Jlr-·· r-H~+f--/-!-,gi -rr~--h+--!-l--+ 13~--- :_::, :, :_]-:_--:.~,-----!_-_:_--~,--:. ~-·-_:,·. +--i-:..-:·-1-i~ii ~. / l ii:; j I iii-I ii =--IH-ll=\-"-1 - -lt ·:t+1,Hl 1+-. -tf:!1 :\ =i':: :l \:ttl·\+l-::·! t!j.· \±1-l:hlt lj:: . : , , : ; : :_ ~~ :::: .. ,_, -t: , 1 , : ; :±r tr --:1tttl:tt -· .tr·- - 1: :1:;. 11- :r: I · i i"t-F ~I :t :j~/:Ji::1~1:t-: · 1Ji.r. L--.r:rh:H- ~i "
0 2 4 6 8 10
CORE HEIGHT, ft.
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I I I I I I I I I I I I I I I I I I
IV. CONCLUSIONS
For breaks up to and in~luding the double ended severance of a
reactor coolant pipe, the emergency core cooling system meets the
Acceptance Criteria as presented in 10 CFR 50.46. Specifically,
1.
2.
3.
4.
s.
Peak ·Cladding Temperature
The calculated peak fuel element clad temperature
provides margin to the requQrement of 2200 degrees F. l
Cladding Oxidation
The clad temperature transient is terminated at a time
when the core geometry is still amenable to cooling.
The cladding oxidation limits of 17 per cent are not
exceeded during or after quenching.
Hydrogen Generation
The amount of fuel element cladding that reacts
chemically with water or steam does not exceed 1 per
cent of the total amount of Zircaloy in the reactor.
Coolable Geometry
The core remains amenable to cooling during and after
the break.
Long-term Cooling
The core temperature is reduced and decay heat is re
moved for an extended period of time, as required by
the long-lived radioactivity remaining in the core
by the safety injection system.
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I I I I' I I I I
·1 I I I I I ·I I
I
V. REFERENCES
1. "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors" 10 CFR 50.46 and Appendix K of 10 CFR 50. Federal Register, Volume 39, Number 3 January 4, 1974.
2. "Westinghouse ECCS Evaluation Model-Summary" WCAP-8339, Bordelon, F.M., Massie, H.W., and Zordan, T.A., July 1974.
3. Bordelon, F.M., et al., "SATAN-IV Program: Comprehensive Space-Time Department Analysis of Loss-of-Coolant," WCAP-8306, June 1974.
4. Bordelon, F.M., et al., "LOCTA-IV Program: Loss-of-Coolant Transient Analysis," WCAP-8305, June 1974.
5. Kelly, R.D., et al., "Calculational Model for Core Reflooding after a Loss-of-Coolant Accident (WREFLOOD Code)," WCAP-8171, June 1974.
6. Bordelon, F.~. and Murphy, E.T., "Containment Pressure Analysis Code (COCO)," WCAP-8326, June 1974.
7. Buterbaugh, T.L., Johnson, W.J. and Kopelic, S.D., "Westinghouse ECCSPlant Sensitivity Studies," WCAP-8356, July 1974.
8. Federal Register, "Supplement to the Status Report by the Directorate of Licensing in the matter of Westinghouse Electric Company ECCS Evaluation Model Conformance to 10 CFR 50, 'Appendix K, 111 ,November 1974.
9. Morita, T., et al., "Power Distribution Control and Load Following Procedures," WCAP-8385, September 1974.
10. Esposito, V.J., et al., "WFLASH-Computer Program for Simulation of Transients in a Multi-Loop PWR," WCAP-8261, July 1974.
11. WCAP-8219, "Fuel Densification Experimental Results and Model for Reactor Applications, ,i October 1973.
12. WCAP-8342, "Westinghouse Emergency Core Cooling System Evaluation Model-Sensitivity Studies," July 1974.
13. Bordelon, F .M., et al., "Westinghouse ECCS Evaluation ModelSupplemental Information," WCAP-8472, April 1975.
14. Letter from D.B. Vassallo of the Nuclear Regulatory Commission to C. Eicheldinger of Westinghouse Electric Corporation dated May 30, 1975.
15. WCAP-8472, "Westinghouse ECCS Evaluation Model, Supplementary Information."
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I I I I I I ·1 I I 'I I 'I I I I ·1 I 'I I
VI. PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS
The proposed Technical Specification changes contained herein
are designated as Change No. 29. The proposed changes are denoted by
a heavy black line in the righthand margin. The operation of Unit Nos.
1 and 2, Surry Power Station, in accordance with these specifications
will assure the validity of the ECCS analysis presented herein.
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I I I I I I I 1· I I I 'I I I I I ·1 I ·1
L.
TS 1.0-7
Low Power Physics Tests
Low power physics tests are tests conducted below 5% of rated power
which measure fundamental characteristics of the reactor core and
related instrumentation.
(Deleted)
CHANGE NO. 29
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I I I I I I I ·I I I I I I I I I ·1 ·1 I
TS 3.3-1
3.3 SAFETY INJECTION SYSTEM
Applicability
Applies to the operating status of the Safety Injection System.
Objective
To define those limiting conditions for operation that are necessary to
provide sufficient borated cooling water to remove decay heat from the
core in emergency situations.
Specifications
A. A reactor shall not be made critical unless the following conditions
are met:
1.
2.
3.
The refueling water tank contains not less than 350,000 gal. of
borated water with a boron concentration of at least 2000 ppm.
Each accumulator system is pressurized to at least 600 psig and
contains a minimum of 975 ft 3 and a maximum of 989 ft 3 of
borated water with a boron concentration of at least 1950 ppm.
The boron injection tank and isolated portions of the inlet and outlet
piping contains no less than 900 gallons of water with a boron con
centration equivalent to at least 11.5% to 13% weight boric acid solution
CHANGE NO I 29
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I I I I I I I I I I I I I I I I I I 1
3.12 CONTROL ROD ASSEMBLIES AND POWER DISTRIBUTION LIMITS
TS 3.12-1 12-27-74
Applicability
Applies to the operation of the control rod assemblies and power distribution
limits.
Objective
To ensure core subcriticality after a .reactor trip, a limit on potential re
activity insertions from hypothetical control rod assembly ejection, and an
acceptable core power distribution during power operation.
Specification
A. Control Bank Insertion Limits
1. Whenever the reactor is critical, except for physics tests and
control rod assembly exercises, the shutdown control rods shall
2.
be fully withdrawn.
Whenever the reactor is critical, except for physics tests and
control rod assembly exercises, the full length control rod
banks shall be inserted no further than the appropriate limit
determined by core burnup shown on TS Figures 3.12-lA, 3.12-lB,
3.12-2, or 3.12-3 for three-loop operation and TS Figures 3.12-4A,
3.12-4B, 3.12-5, or 3.12-6 for two-loop operation.
CHANGE NOO 29
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I I I I I I I I ·1 I I I I I I I I I I
3.
TS 3.12-2 12-27-74
The limits shown on TS Figures 3.12-lA through 3.12:-6 may be
revised on the basis of physics calculations and physics data
obtained during unit startup and subsequent operation, in
accordance with the following:
a,
b.
c.
The sequence of withdrawal of the controlling banks, when
going from zero to 100% power, is A, B, C, D.
An overlap of control banks, consistent with physics
calculations and physics data obtained during unit
startup and subsequent operation, will be permitted.
The shutdown margin with allowance for a stuck control rod
assembly shall exceed the applicable value shown on TS
Figure 3.12-7 under all steady-state operation conditions,
except for physics tests, from zero to full power, including
effects of axial power distribution. The shutdown margin
as used here is defined as the amount by which the reactor
core would be subcritical at hot shutdown conditions
(T >547°F) if all control rod assemblies were tripped, avg - .
assuming that the highest worth control rod. assembly re-
mained fully withdrawn, and assuming no changes in xenon,
boron, or part-length rod position.
CHANGE NO. 29
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I I I I I I I
I I I I I I I I I I
4.
5.
6.
TS 3.12-3
Whenever the reactor is subcritical, except for physics tests,
the critical rod position, i.e., the rod position at which
criticality would be achieved if the control rod assemblies were
withdrawn in normal sequence with no other reactivity changes,
shall not be lower than the insertion limit for zero power.
Operation with part length rods shall be restricted such that
except during physics tests, the part length rod banks are with
drawn from the core at all times.
Insertion limits do not apply during physics tests or during
periodic exercise of individual rods. However, the shutdown
margin indicated· in TS Figure 3.12-7 must be maintained except
for the low power physics test to measure control rod worth and
shutdown margin. For this test the reactor may be critical with
all but one full length control rod, expected to have the highest
worth, inserted and part length rods fully withdrawn.
B. Power Distribution .Limits
1. At all times except during physics tests, the hot channel factors
defined in the basis must meet the following limits:
F (Z) < (2.10/P) x K(Z) for P > .5 Q
FQ(Z) < (4.20) x K(Z) for P < .5
F~H :5._ 1.55 (1 + 0.2(1 - P))
CHANGE NO. 29
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I I I I I I I I I
I I I I I I I I I
2.
TS 3.12-4
where Pis the fraction of rated power at which the core is operating,
K(Z) is the function given in Figure 3.12-8, and Z is the core
height location of FQ.
Prior to exceeding 75% power following each core loading, and
during each effective full power month of operation thereafter,
power distribution maps using the movable detector system, shall
be made to confirm that the hot channel factor limits of this
specification are satisfied. For the purpose of this confirmation:
a.
b.
The measurement of total peaking factor, F~eas, shall be
increased by three percent to account for manufacturing
tolerances and further increased by five percent to account
for measurement error.
N The measurement of enthalpy rise hot channel factor, FliH'
shall be increased by four percent to account for measure-
ment error.
If either measured hot channel factor exceeds its limit specified
under 3.12.B.1, the reactor power and high neutron flux trip set
point shall be reduced until the limits under 3.12.B.1 are met.
If the hot channel factors cannot be brought to within the limits
FQ 2_ 2.10 x K(Z) and F~H ~ 1.55 within 24 hours, the overpower LiT
and overtemperature LiT trip setpoints shall be similarly reduced.
CHANGE NO. 29
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I I I I I
II
I I I I I I I I I I I I I
3.
4.
TS 3.12-5
The reference equilibrium indicated axial flux difference (called
the target flux difference) at a given power level P0
, is that
indicated axial flux difference with the core in equilibrium
xenon conditions (small or no oscillation) and the control rods
more than 190 steps withdrawn. The target flux difference at
any other power level, P, is equal to the target value of P
multiplied by the ratio, P/P0 • The target flux difference shall
be measured at least once per equivalent full power quarter.
The target flux difference must be updated during each effective
full power month of operation either by actual measurement, or by
linear interpolation using the most recent value and the value
predicted for the end of the cycle life.
Except during physics tests, during excore detector calibration
and except as modified by 3.12.B.4.a., b., or c. below, the
indicated axial flux difference shall be maintained within a
+6 to -9% band about the target flux difference (defines the
target band on axial flux difference).
a. At a power level greater than 90 percent of rated power,
if the indicated axial flux difference deviates from its
target band, the flux difference shall be returned to the
target band, or the reactor power shall immediately be
reduced to a level no greater than 90 percent of rated
power.
CHANGE NO. 29
i
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I I I I I I I I I I I I I I I I I I I
b.
TS 3.12-6
At a power level no greater than 90 percent of rated power,
(1) The indicated axial flux difference may deviate from
its +6 to -9% target hand for a maximum of one hour
(cumulative) in any 24 hour period provided the flux
difference does not exceed an envelope bounded by -18
percent and +11.5 percent at 90% power. For every 4
percent below 90% power, the permissible positive flux
difference boundary is extended by 1 percent. For every
5 percent below 90% power, the permissible negative flux
difference boundary is extended by 2 percent.
(2) If 3.i2.B.4.b. (1) is violated then the reactor power
shall be reduced to no greater than 50% power and the
high neutron flux setpoint shall be reduced to no
greater than 55% power.
(3) A power increase to a level greater than 90 percent
of rated power is contingent upon the indicated axial
flux difference being within its target band.
c. At a power level no greater than 50 percent of rated power,
(1) The indicated axial flux difference may deviate from
its target band.
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I I I I I I I I I I I I I I I
I I I
5.
TS 3.12-7.
(2) A power increase to a level greater than .50 percent
of rated power is contingent upon the indicated axial
flux difference not being outside its target band for
more than two hours (cumulative) out of the preceding
24 hour period. One half of the time the indicated
axial flux difference is out of its target band up to
50 percent of rated power is to be counted as con
tributed to the one hour cumulative maximum the flux
difference maximum deviate from its target band at a
power level less than or equal 90 percent of rated
power.
Alarms shall normally be used to indicate the deviations from the
axial flux difference requirements in 3.12.B.4.a and the flux
difference time limits in 3.12.B.4.b. If the alarms are out of
service temporarily, the axial flux difference shall be logged,
and conformance to the limits assessed, every hour for the first
24 hours, and half-hourly thereafter.
The allowable quadrant to average power tilt is
T = 2.0 + 50 (1.435/F - 1) < 10% xy
where F is 1.435, or the value of the unrodded horizontal plane xy
peaking factor appropriate to FQ as determined by a movable incore
detector map taken on at least a monthly basis; and Tis the per
centage operating quadrant tilt limit, having a value of 2% if
F is 1.435 or a value up to 10% if the option to measured F xy xy
is in effect.
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I I I I I I I I I I I I I I I I I I I
6.
7.
TS 3.12-8
If the quadrant to average power tilt exceeds a value T% as
selected in 3.12.B.5, except for physics and rod exercise
testing, then:
a.
b.
c.
The hot channel factors shall be determined within 2 hours
and the power level adjusted to meet the specification of
3. 12. B. 1, or
If the hot channel factors are not determined within two
hours, the power and high neutron flux trip setpoint shall
be reduced from rated power, 2% for each percent of quadrant
tilt.
If the quadrant to average power tilt exceeds ±_10% except
for physics tests, the power level and high neutron flux
trip setpoint will be reduced from rated power, 2% for each
percent of quadrant tilt.
If after a further period of 24 hours, the power tilt in 3.12.B.6
above is not corrected to less than +T%:
a. If design hot channel factors for rated power are not
exceeded, an evaluation as to the cause of the discrepancy
shall be made and reported as an abnormal occurrence to the
Nuclear Regulatory Commission.
CHANGE NO. 29
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I I I I I I I I I I I I I I I I I I I
,d,
C.
b.
c.
TS 3.12-9
If the design hot channel factors for rated power are exceeded
and the power is greater than 10%, the Nuclear Regulatory
Commission shall be notified and the nuclear overpower, over
power ~T and overtemperature ~T trips shall be reduced one
percent for each percent the hot channel factor exceeds the
rated power design values.
If the hot channel factors are not determined the Nuclear
Regulatory Commission shall be notified and the overpower ~T
and overtemperature ~T trip settings shall be reduced by the
equivalent of 2% power for every 1% quadrant to average power
tilt.
Inoperable Control Rods
1.
2.
3.
A control rod assembly shall be considered inoperable if the
assembly cannot be moved by the drive mechanism, or the assembly
remains misaligned from its bank by more than 15 inches. A full
length control rod shall be considered inoperable if its rod drop
time is greater than 1.8 seconds to dashpot entry.
No more than one inoperable control rod assembly shall be permitted
when the reactor is critical.
If more than one control rod assembly in a given bank is out of
service because of a single failure external to the individual rod
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I I I I I 1·
I I I I I I I I I I I I I
4.
5.
6.
7.
TS 3.12-10
drive mechanisms, i.e. programming circuitry, the provisions of
3.12.C.l and 3.12.C.2 shall not apply and the reactor may remain
critical for a period not to exceed two hours provided immediate
attention is directed toward making the necessary repairs. In the
event the affected assemblies cannot be returned to service within
this specified period the reactor will be brought to hot shutdown
conditions.
The provisions of 3.12.C.l and 3.12.C.2 shall not apply during
physics test in which the assemblies are intentionally misaligned.
If an inoperable full-length rod is located below the 200 step
level and is capable of being tripped, or if the full-length rod
is located below the 30 step level whether or not it is capable
of being tripped, then the insertion limits in TS Figure 3.12-2
apply.
If an inoperable full-length rod cannot be located, or if the in
operable full-length rod is located above the 30 step level and
cannot be tripped, then the insertion limits in TS Figure 3.12-3
apply.
No ·insertion limit changes are required by an inoperable part
length rod.
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I I I I I I I I I I I I I I I I I I I
D.
8.
TS 3 .12-11
If a full-length rod becomes inoperable and reactor operation is
continued the potential ejected rod worth and associated transient
power distribution peaking factors shall be determined by analysis
within 30 days. The analysis shall include due allowance for non
uniform fuel depletion in the neighborhood of the inoperable rod.
If the analysis results in a more limiting hypothetical transient
than the cases reported in the safety analysis, the unit power
level shall be reduced to an analytically determined part power
level which is consistent with the safety analysis.
If the reactor is operating above 75% of rated power with one excore
nuclear channel out of service, the core quadrant power balance shall
be determined.
1.
2.
Once per day, and
After a change in power level greater than 10% or more than 30
inches of control rod motion.
The core quadrant power balance shall be determined by one of the
following methods:
1. Movable detectors (at least two per quadrant)
2. Core exit thermocouples (at least four per quadrant).
CHANGE NO ?.9
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I I I I I I I I I I I I I I I I I I I
E.
F.
TS 3.12-12
Inoperable Rod Position Indicator Channels
1.
2.
If a rod position indicator channel is out of service then:
a.
b.
For operation between 50% and 100% of rated power, the position
of the RCC shall be checked indirectly by core instrumentation
(excore detector and/or thermocouples and/or movable incore
detectors) every shift or subsequent to motion, of the non
indicating rod, exceeding 24 steps, whichever occurs first.
During operation below 50% of rated power no special monitoring
is required.
Not more than one rod position indicator (RPI) channel per group
nor two RPI channels per bank shall be permitted to be inoperable
at any time.
Misaligned or Dropped Control Rod
1. If the Rod Position Indicator Channel is functional and the
associated part length or full length control rod is more than
15 inches out of alignment with its bank and cannot be realigned,
then unless the hot channel factors are shown to be within design
limits as specified in Section 3.12.B.1 within 8 hours, power
shall be reduced so as not to exceed 75% of permitted power.
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I I I I I I I I I I I I I I I I I I I
2.
Basis
TS 3;12-13
To increase power above 75% of rated power with a part-length or
full length control rod more than 15 inches out of alignment with
its bank an analysis shall first be made to determine the hot channel
factors and the resulting allowable power level based on Section
3.12.B.
The reactivity control concept assumed for operation is that reactivity changes
accompanying changes in reactor power are compensated by control rod assembly
motion. Reactivity changes associated with xenon, samarium, fuel depletion,
and large changes in reactor coolant temperature (operating temperature to
cold shutdown) are compensated for by changes in the soluble boron concentration.
During power operation, the shutdown groups are fully withdrawn and control of
power is by the control groups. A reactor trip occurring during power operation
will place the reactor into the hot shutdown condition.
The control rod assembly insertion limits provide for achieving hot shutdown
by reactor trip at any time, assuming the highest worth control rod assembly
remains fully withdrawn, with sufficient margins to meet the assumptions used
in the accident analysis. In addition, they provide a limit on the maximum
inserted rod worth in the unlikely event of a hypothetical assembly ejection,
and provide for acceptable nuclear peaking factors. The limit may be determined
on the basis of unit startup and operating data to provide a more realistic
limit which will allow for more flexibility in unit operation and still assure
compliance with the shutdown requirement. The maximum shutdown margin require-
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I I I I I I I I I I I I I I I I I I I
TS 3.12-14
ment occurs at end of core life and is based on the value used in the analysis
of the hypothetical steam break accident. The rod insertion limits are based
on end of core life conditions. Early in core life, less shutdown margin is
required, and TS Figure 3.12-7 shows the shutdown margin equivalent to 1.77%
reactivity at end-of-life with respect to an uncontrolled cooldown. All other
accident analyses are based on 1% reactivity shutdown margin.
Relative positions of control rod banks are determined by a specified control
rod bank overlap. This overlap is based on the consideration of axial power
shape control.
The specified control rod insertion limits have been revised to limit the
potential ejected rod worth in order to account for the effects of fuel
densification.
The various control rod assemblies (shutdown banks, control banks A, B, C, and
D and part-length rods) are each to be moved as a bank, that is, with all
assemblies in the bank within one step (5/8 inch) of the bank position. Position
indication is provided by two methods: a digital count of actuating pulses
which shows the demand position of the banks and a linear position indicator,
Linear Variable Differential Transformer, which indicates the actual assembly
position. The position indication accuracy of the Linear Differential Trans
former is approximately ±_5% of span (±_7.5 inches) under steady state conditions.
The relative accuracy of the linear position indicator is such that, with the
CHANGE NO. 29
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I I I I I I I I I I I I I I I I I I I
TS 3.12-15
most adverse errors, an alarm is actuated if any two assemblies within a bank
deviate by more than 14 inches. In the event that the linear position indicator
is not in service, the effects of malpositioned control rod assemblies are
observable from nuclear and process information displayed in the Main Control
Room and by core thermocouples and in-core movable detectors. Below 50% power,
no special monitoring is required for malpositioned control rod assemblies with
inoperable rod position indicators because, even with an unnoticed complete
assembly misalignment (part-length of full length control rod assembly 12 feet
out of alignment with its bank) operation at 50% steady state power does not
result in exceeding core limits.
The specified control rod assembly drop time is consistent with safety analyses
that have been performed.
An inoperable control rod assembly imposes additional demands on the operators.
The permissible number of inoperable control rod assemblies is limited to one
in order to limit the magnitude of the operating burden, but such a failure
would not prevent dropping of the operable control rod assemblies upon reactor
trip.
Two criteria have been chosen as a design basis for fuel performance related
to fission gas release, pellet temperature and cladding mechanical properties.
First, the peak value of linear power density must not exceed 21.1 kw/ft for
Unit No. 1 and 20.4 kw/ft for Unit No. 2. Second, the minimum DNBR in the
core must not be less than 1.30 in normal operation or in short term transients.
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I I I I I I I I I I I I I I I I I I I
TS 3.12-16
In addition to the above, the peak linear power density must not exceed the
limiting Kw/ft values which result from the large break loss of coolant accident
analysis based on the ECCS acceptance criteria limit of 2200°F on peak clad
temperature. This is required to meet the initial conditions assumed for the
loss of coolant accident. To aid in specifying the limits on power distribution
the following hot channel factors are defined.
Fq(Z), Height Dependent Heat Flux Hot Channel Factor, is defined as the maxi
mum local heat flux on the surface of a fuel rod at core elevation Z divided
by the average fuel rod heat flux, allowing for manufacturing tolerances on
fuel pellets and rods.
FE, Engineering Heat Flux Hot Channel Factor, is defined as the allowance Q
on heat flux required for manufacturing tolerances. The engineering factor
allows for local variations in enrichment, pellet density and diameter, surface
area of the fuel rod and eccentricity of the gap between pellet and clad.
Combined statistically the net effect is a factor of 1.03 to be applied to
fuel rod surface heat .flux.
F~H' Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the
integral of linear power along the rod with the highest integrated power to the
average rod power.
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I I I I I I I I I I I I I I I I I I I
TS 3.12-17
It should be noted that F~H is based on an integral and is us~d as such in
the DNB calculations. Local heat fluxes are obtained by using hot channel
and adjacent channel explicit power shapes which take into account variations
in horizontal (x-y) power shapes throughout the core. Thus the horizontal
power shape at the point of maximum heat flux is not necessarily directly
N related to F6H.
An upper bound envelope of 2.10 times the normalized peaking factor axial
dependent of TS Figure 3.12-8 has been determined from extensive analyses
considering all operating maneuvers consistent with the technical specifi
cations on power distribution control given in Section 3.12.B.4. The results
of the loss of coolant accident analyses are conservative with respect to the
ECCS acceptance criteria as specified in 10 CFR 50.46.
When an FQ measurement is taken, both experimental error and manufacturing
tolerance must be allowed for. Five percent is the appropriate allowance
for a full core map (~ 40 thimbles monitored) taken with the movable incore
detector flux mapping system and three percent is the appropriate allowance
for manufacturing tolerances.
In the specified limit of F~H there is an eight percent allowance for un
certainties which means that normal operation of the core is expected to.
result in F~H ..::_ l.55/1.08. The logic behind the larger uncertainty in this
case is that (a) normal perturbations in the radial power shape (e.g. rod
misalignment) affect F~H' in most cases without necessarily affect FQ, (b)
the operator has a direct influence on F through movement of. rods, and can Q
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I I I I I I I I I I I I I I I I I I I
TS 3.12-18
limit it to the desired value, he has no direct control over F~, and (c) - · LlH
an error in the predictions for radial power shape, which may be detected
during startup physics tests can be compensated for the FQ by tighter axial
control, but compensation for F~H is taken, experimental error must be allowed
for and four percent is the appropriate allowance for a full core map (2:_ 40
thimbles monitored) taken with the movable incore detector flux mapping
system.
Measurement of the hot channel factors are required as part of startup physics
tests, during each effective full power month of operation, and whenever ab-
normal power distribution conditions require a reduction of core power to a
level based on measured hot channel factors. The incore map taken following
core loading provides confirmation of the basic nuclear design bases including
proper fuel loading patterns. The periodic incore mapping provides additional
assurance that the nuclear design bases remain inviolate and identify operational
anomalies which would, otherwise, affect these bases.
For normal operation, it is not necessary to measure these quantities. Instead
it has been determined that, provided certain conditions are observed,
the hot channel factor limits will be met; these conditions are as follows:
1. Control rods in a single bank move together with no individual
rod insertion differing by more than 15 inches from the bank
demand position. An indicated misalignment limit of 13 steps
precludes a rod misalignment no greater than 15 inches with
consideration of maximum instrumentation error.
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I I I I I I I I I I I I
·1 I I I I I I
2.
~-
4.
TS 3.12-19
Control rod bariks are sequenced with overlapping banks as
shown in Figures 3.12-lA, 3.12-lB and 3.12-2.
The full length and part length control bank insertion limits
are not violated.
Axial power distribution control procedures, which are given
in terms of flux difference control and control bank insertion
limits are observed. Flux difference referes to the difference
in signals between the top and bottom halves of two-section
excore neutron detectors. The flux difference is a measure
of the axial offset which is defined as the difference in
normalized power between the top and bottom halves of the core.
The permitted relaxation in F~H with decreasing power level allows radial
power shape changes with rod insertion to the insertion limits. It has been
determined that provided the above conditions 1 through 4 are observed, these
hot cha~nel factors limits are met. In specification 3.12.B.1 Fq is
arbitrarily limited for P < .5 (except for physics tests).
The procedures for axial power distribution control referred to above are
designed to minimize the effects of xenon redistribution on the axial power
distribution during load-follow maneuvers. Basically control of flux
difference is required to limit the difference between the current value of
flux difference (~I) and a reference value which corresponds to the full
power equilibrium value of axial offset (axial offset= ~I/fractional power).
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I I I I I I I I I I I I I I I I I I I
TS 3.12-20
The reference value of flux difference varies with power level and burnup,
but expressed as axial offset it varies only with burnup.
The technical specifications on power distribution control given in 3.12.B.4
assure that the FQ upper bound envelope of 2.10 times Figure 3.12-8 is not
exceeded and xenon distributions are not developed which at a later time,
would cause greater local power peaking even though the flux difference
is then within the limits specified by the procedure.
The target (or reference) value of flux difference is determined as follows.
At any time that equilibrium xenon conditions have been established, the
indicated flux difference is noted with the full length rod control bank more
than 190 steps withdrawn (i'.e. normal full power operating position appropriate
for the time in life, usually withdrawn farther as burnup proceeds). This
value, divided by the fraction of full power at which the core was operating
is the full power value of the target flux difference. Values for all other
core power levels are obtained by multiplying the full power value by the
fractional power. Since the indicated equilibrium value was noted, no allowances
for excore detector error are necessary and indicated deviation of +6 to -9% ~I
are permitted from the indicated reference value. During periods where extensive
load following is required, it may be impractical to establish the required
core conditions for measuring the target flux difference every month. For this
reason, the specification provides two methods for updating the target flux
difference.
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I I I I I I I I I I I I I I I I I I I
TS 3.12-21
Strict control of the flux difference (and rod position) is not as necessary
'during part power operation. This is because xenon distribution control at
part power is not as significant as the control at full power and allowance
has been made in predicting the heat flux peaking factors for less strict
control at part power. Strict control of the flux difference is not
possible during certain physics tests or during required, periodic, excore
detector calibrations which require larger flux differences than permitted.
Therefore, the specifications on power distribution control are not applied
during physics tests or excore detector calibrations; this is acceptable
due to the low probability of a significant accident occurring during these
operations.
In some instances of rapid "unit power reduction automatic rod motion will
cause the flux difference to deviate from the target band when the reduced
power level is reached. This does not necessarily affect the xenon dis
tribution sufficiently to change the envelope of peaking factors which can
be reached on a subsequent return to full power within the target band, how
ever, to simplify the specification, a limitation of one hour in any period
of 24 hours is placed on operation outside the band. This ensures that the
resulting xenon distributions are not significantly different from those
resulting from operation within the target band. The instantaneous
consequences of being outside the band, provided rod insertion limits are
observed, is not worse than a 10 percent increment in peaking factor for the
allowable flux difference at 90% power, in the range +14.5 to -21 percent
(+11.5 percent to -18 percent indicated) where for every 4 percent below
rated power, the permissible positive flux difference boundary is extended
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I I I I
I I I I I I I I I I I I I I
TS 3.12-22
by 1 percent, and for every 5 percent below rated power, the ~ennissible
negative flux difference boundary is extended by 2 percent.
As discussed above, the essence of the procedure is to maintain the xenon
distribution in the core as close to the equilibrium full power condition
as possible. This is accomplished, by using the boron system to position
the full length control rods to produce the required indicated flux difference.
At the option of the operator, credit may be taken for measured decreases in
the unrodded horizontal plane peaking factor, Fxy· This credit may take the
form of an expansion of permissible quadrant tilt limits over tilt limits
over the 2% value, up to a value of 10%, at which point specified power re-
ductions are prudent. Monthly surveillance of F bounds the quantity because xy
it decreases with burnup. (WCAP-7912 L).
A 2% quadrant tilt allows that a 5% tilt might actually be present in the core
because of insensitivity of the excore detectors for disturbances near the core
center such as misaligned inner control rods and an error allowance. No
increase in FQ occurs with tilts up to 5% because misaligned control rods
producing such tilts do not extend to the unrodded plane, where the maximum
FQ occurs.
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I I
TS FIGURE 3.12-8
I HOT CHANNEL FACTOR NORMALIZED
I OPERATING ENVELOPE
SURRY POWER STATION
I UNIT NOS. 1 AND 2
I I
1.0
I I ,,..... .8
N '-'
O' Ii-<
I A ~ N H .6 i--l
I ~ 0 z
I ,,..... .4 N '-' ~
I . 2
I 0
I 0 2 4 6 8 10 12
CORE HEIGHT (ft.)
I I I I
(' 1 i ~ \"Cr NO, .,,0 .,,,. ri. j J .Jt
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I I I I I I I I I I I I I I I I I I I
TS 4.10-1
4.10 REACTIVITY ANOMALIES
Applicability
Applies to potential reactivity anomalies.
Objective
To require evaluation of applicable reactivity anomalies within the reactor.
Specification
A.
B'.
Following a normalization of the computed boron concentration as a
function of burnup, the actual boron concentration of the coolant shall
be compared monthly with the predicted value. If the difference between
the observed and predicted steady-state concentrations reaches the
equivalent of one percent in reactivity, an evaluation as to the cause
of the discrepancy shall be made and reported to the Nuclear Regulatory
Commission per Section 6.6 of these Specifications.
During periods of power operation at greater than 10% of power, the hot
N channel factors, FQ and FL'IH' shall be determined during each effective
full power month of operation using data from limited core maps. If these
factors exceed values of
F Q (Z) < (2 .10/P) x K(Z) for p > .5
F (Z) < Q
(4. 20) x K(Z) for P < .5
FN < t,H - 1. 55 (1 + 0.2 (1 - P))
CHANGE NO. :29
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I I I I I I I I I I I I I I I I ,I I I
TS 4.10-2
(where Pis the fraction of rated power at which the core is operating,
K(Z) is the function given in Figure 3.12-8, and Z is the core height
location of FQ), an evaluation as to the cause of the anomaly shall be
made.
Basis
BORON CONCENTRATION
To eliminate possible errors in the calculations of the initial reactivity
of the core and the reactivity depletion rate, the predicted relation between
fuel burnup and the boron concentration necessary to maintain adequate control
characteristics, must be adjusted (normalized) to accurately reflect actual.
core conditions. When full power is reached initially, and with the control
rod assembly groups in the desired positions, the boron concentration is
measured and the predicted curve is adjusted to this point. As power operation
proceeds, the measured boron concentration is compared with the predicted
concentration, and the slope of the curve relating burnup and reactivity is
compared with that predicted. This process of normalization should be completed
after about 10% of the total core burnup. Thereafter, actual boron concentration
can be compared with prediction, and the reactivity status of the core can be
continuously evaluated. Any reactivity anomaly greater than 1% would be· un
expected, and its occurrence would be thoroughly investigated and evaluated.
The value of 1% is considered a safe limit since a shutdown margin of at
least 1% with the most reactive control rod assembly in the fully withdrawn
position is always maintained.
CHANGE NO. 29
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I I I I I I I I I I I I I I I I I I I
TS 4.10-3
PEAKING FACTORS
A thermal criterion in the reactor core design specifies that "no fuel
melting during any anticipated normal operating condition" should occur.
To meet the above criterion during a thermal overpower of 118% with additional
margin for design uncertainties, a steady state maximum linear power is
selected. This then is an upper linear power limit determined by the maxi
mum central temperature of the hot pellet.
The peaking factor is a ratio taken between the maximum allowed linear power
density in the reactor to the average value over the whole reactor. It is of
course the average value that determines the operating power level. The peaking
factor is a constraint which must be met to assure that the peak linear power
density does not exceed the maximum allowed value.
During normal reactor operation, measured peaking factors should be significantly
lower than design limits. As core burnup progresses, measured designed peaking
factors are expected to decrease. A determination of FQ and F~H during each
effective full power month of operation is adequate to ensure that core re
activity changes with burnup have not significantly altered peaking factors in
an adverse direction.
G HANGE NO. 29
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I I I I I I I I I I I I I I I I I I I
-----------~--------------~
3.
4.
5.
6.
TS 5.3-2
Reload fuel will be similar•in design to the initial core. The
enrichment of reload fuel w.ill not exceed 3.60 weight per cent of
U-235.
Burnable poison rods are incorporated in the initial core. There
are 816 poison rods in the form of 12 rod clusters, which are
located in vacant control rod assembly guide thimbles. The burnable
poison rods consist of pyrex glass clade with stainless steel.
There are 48 full-length control rod assemblies and 5 part-length
control rod assemblies in the reactor core. The full-length control
rod assemblies contain a 144 inch-length of silver-indium-cadmium
alloy clad with stainless steel. The part-length control rod
assemblies contain a 36 inch-length of silver-indium-cadmium alloy
with the remainder of the stainless steel sheath filled with Al 2o3
.
The initial core and subsequent cores will meet the following per
formance criteria at all times during the operating lifetime.
a. Hot channel factors:
F (Z) < (2.10/P) x K(Z) for P > .5 Q
FQ(Z) < (4.20) x K(Z) for P < .5
N F~H _::_ 1.55 (1 + 0.2 (1 - P))
where Pis the fraction of rated power at which the core is
operating, K(Z) is the function given in Figure 3 .12.-8, and Z
is the core height location of FQ.
CHANGE NO. 29
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