ML20206H885
| ML20206H885 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 11/15/1988 |
| From: | Quay T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20206H889 | List: |
| References | |
| NUDOCS 8811230426 | |
| Download: ML20206H885 (58) | |
Text
'[ '
- 4: S UNITED STATES s
- > s.
NUCLEAR REGULATORY COMMISSION
%L cAseiNO ton. O. C. 205s5 CONSUMERS POWER COMPANY PALISADES PLANT DOCKET NO. 50-255 AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No.ll8 License No. OPR-20 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The applications for amendment by Consumers Power Company (the licensee) dated March 25 and September 1, 1988, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; 8.
The facility will operate in conformity with the applications, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the her'.th and safety of the public; and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and i
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
l l
$5N*12Wggj11:
l f#5
Os s
I r i
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 3.B. of Provisional Operating License No.
DPR-20 is hereby amended to read as follows:
i Technical Specifications
[
The Technical Specifications contained in Appendices A and B, as revised through Amendment No.118, are hereby incorporated in the license.
The licensee shall operate the facility f
in accordance with the Technical Specifications.
3.
This lice.1se amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COM415$10N b*
W}
Theodore Quay, Acting Director Project Directorata 111-1 Division of Reactor Projects - III, IV, V
& Special Projects
Attachment:
Chariges to the Technical Specifications Date of Issuance:
November 15, 1988
ATTACHMENT TO LICENSE AMEN 0 MENT N0.118
[
PROVISIONAL OPERATING LICENSE NO. OPR-20 t
DOCKET N0. 50-255 t
i i
Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages.
The revised pages are identified by l
the captioned amendment number and contain marginal lines indicatir:g the area of change.
i RENOVE INSERT i
i i & it i & 11 y
y 1-2 1-2 1-2a b
2 2-9 2 2-9 2-10 2-11
[
2-12 f
2-13 3-Ib 1d 3-lb 1d i
3-2 & 3-3 3-2 & 3-3 i
3-3a 3-3a 3-25 3-25 t
3-39 3-39 I
3-58 3-58 i
3-59 I
l 3 3-64 3 3-64 3-66a & 3-66b 3-66a & 3-66b i
3-66c 3-67 3-67 3-68 i
3-77 & 3-78 3-77 a 3-78 I
3-81a & 3-81b 3-81a & 3-81b 3-103 105 3-103 105 3-106 4
3-107 - 3 112 3-107 - 3 112 4 4-5 4 4-5 i
4-10 & 4-11 4-10 & 4-11 4-70 4-70
+
4 4-85 4 4-85
)
s I
i 4
2
l....c2.. '.:...... _.
......s...u 9 '..,'. Q_ r.. 4 PALISADES PLANT TECHNICAL SPECIFICATIONS TABII 0F CONTENTS - APPENDIX A SECTION DESCRIPTION PAGE' NO 1.0 DEFINITIONS 1-1 1.1 REACTOR OPERATING CONDITIONS 11 1.2 PROTECTIVE SYSTEMS 1-3 1.3 INSTRUMENTATION SIRVE!I.IANCE 1-3 1.4 MISCELLANEOUS DEFINITIONS 1-4 2.0 SAFETY LIMITS AND LIMITING SATITY SYSTEM SETTINGS 2-1 2.1 SAFETY LIMITS - REACTOR CORE 2-1 2.2 SATITY LIMITS - PRIMARY COOLANT SYSTEM PRISSURE 2-3 2.3 LIMITING SAITTY SYSTEM SETTINGS - REACTOR PROTECTIVE SYSTEM 2-4 Taile 2.3.1 Reactor Protective Systen Trip Setting Limits 2-5
/
3.0 LIMITING CONDITIONS FOR OPERATION 31 l
3.0 APPLICABILITY 3-1 3.1 PRIMARY C00LAN7 SYSTEM 3 lb 3.1.1 Operable Components 3-lb Tigure 3-0 Reactor Inlet Temperature vs Operating Pressure 3-3a 3.1.2 Heatup and Cooldown Rates 3-4 Figure 3-1 Pressure - Temperature Limits for Heatup 3-9 Figure 3-2 Pressure - Temperature Limits for Cooldown 3-10 Figure 3-3 Pressure - Temperature Limits for Hydro Test 3 11 3.1.3 Minimum Conditions for Criticality 3-12 3.1.4 Marious Primary Coolant Radioactivity 3-17 3.1.5 Primary Coolant Systee Leakage Limits 3-20 3.1.6 Maximus Primary Coolant oxygen and Halogens Concentrations 3-23 3.1.7 Primary and Secondary Safety Valves 3-25 3.1.8 Overpressure Protection Systems 3-25a 3.2 CIEMICAL AND VOLUME CONTROL SYSTEM 3-26 3.3 EMERGENCY CORI COOLING SYSTEM 3-29 l
3.4 CONTAINMENT COOLING 3 34 3.5 STEAM AND IT.EDWATER 3YSTLM 3-38 3.6 CONTAINMENT SYSTEM 3-40 3.7 ELECTRICAL SYSTL*tS 3-41 1
3.3 RU1ELING OPERATIONS 3-46 3.9 ETTLUENT RELEASE (DEIITED) 3-50 i
Amendment No. 31, II, 37, 83,198,118 i
f
, ~ - - -.. - - -. - - - -., - -., -. - - - - - - - -..
n,-
--,---.,,_-,,n
n-
_.--.-.-.y.
PALISADES PLANT TECHNICAL SPECITICATIONS TABLE OF CONTENTS - APPENDIX A SECTIO,N, DESCRIPTION PAGE NO 3.0 LIMITING CONDITIONS FOR OPERATION (Continued) 3.10 CONTROL 200 AND POWER DISTRIBUTION LIMITS 3-58 3.10.1 Shutdown Margin Requirements 3-58 3.10.2 c)eleted )
3-58 /
3.10.3 Part-Length Control Rods 3-58 3.10.4 Misaligned or Inoperable Control Rod or Part-Length Rod 3-60 3.10.5 Regulating Group Insertion Limits 3-60 3.10.6 Shutdown Rod Limits 3-61 3.10.7 Low Power Physics Testing 3-61 3.10.8 Center Control Rod Misalignment 3-61 Tigure 3 6 Control Rod Insertion Limits 3-62 3.11 POWER DISTRIBUTION INSTRLHENTATION 3-65 3.11.1 Incore Detectors 3-65 3.11.2 Escore Power Distribution Monitoring System 3-66a Figure 3.11-1 Axial Variation Bounding Condition 3-66d 3.12 MODERATOR TEMPERATURE COETTICIENT OF REACTIVITY 3-67 3.13 (Deleted) 3-69 3.14 CONTROL ROOH VENTILATION 3-70 3.15 REACTOR PRIMARY SHIELD C00LIN0 SYSTEM 3-70 3.16 ENGINEERID SATETT TEATURES SYSTEM INITIATION INSTRUENTATION SETTINGS 3-71 Table 3.16.1 Engineered Safety Features Systes Initiation Instrument setting Linits 3-75 3.17 INSTRLMENTATION AND CONTROL SYSTLMS 3-76 i
Lble 3.17.1 Instrumentation Operating Requirements for Reactor Protective System 3-78 Table 3.17.2 Instrumentation Operating Requirements for Engineered Safety Testure Systems 3-79 Table 3.17.3 Instrument Operating Conditions for Isolation Tunctions 3-80 Table 3.17.4 Instrumentation Operating Requirements for Other Safety Teature Functions 3-81 1
3.18 (Deleted) 3 82 3.19 IODINE REMOVAL SYSTEM 3-84 j
3.20 SHOCK SUPPRESSORS (SNUBBERS) 3-88 3.21 MOVEMENT OT E AVT LOADS 3-92 i
3.22 FIE PROTECTION SYSTEM 3-96 3.22.1 Fire Detection Instrumentation 3-96 Table 3.22.1 Tire Detection Instrumentation - Minimus Instruments Operable 3-97 i
11 i
i AmendmentNo.37,33,ff,SJ,Ig, Jff,JJf,))),
i
PALISADE 3 PLANT TECHNICAL SPECITICATIONS TABLE OP CONTENTS - APPENDIX A SECTION DESCRIPTION PAGE NO 4.0 SURVE!!.!ft'CE REQUIREMENTS (Continued) i Table 4.11-3 Detection Capabilities for Environmental Sample Analysis 4-57 4.11.1 Bases for Monitoring Program 4-59a 4.11.3 Bases for Land Use Census 4-59a 4.11.5 Bases for Interlaboratory Comparison Program 4-59a 4.12 AUGMENTED INSERVICE INSPECTION PROGRAM TOR HIGH ENERGY LINES OUTSIDE OF CohTAINMEhT 4 60 Tig. 4.12 A Augmented Inservice Inspection Program - Main Steam Welds 4-63 Fig. 4.12 8 Augmented Inservice Inspection Program - Teedwater Line Welds 4-64 4.13 REACTOR INTERNALS VIBRATION MONITORING (DELETED) 4-65 4.14 AUGMENTED INSERVICE INSPECTION PROGRAM TOR STEAM GENERATORS 4-68 Table 4.14.1 Operating Allowances 4-68d Table 4.14.2 Maximum Allowable Degradation 4-69 4.15 PRIMARY SYSTEM TLOV MEASURIMENT 4-70 4.16 INSERVICE INSPECTION PROGRAM TOR SHOCK SUPPRESSORS (SNUBBERS) 4-71 4.17 TIRI PROTICTION SYSTEM 4-75 4.17.1 Tire Detection Instrumentation 4-75 4.17.2 Tire Suppression Water System 4-76 4.17.3 Tire Sprinkler Systes 4-78 4.17.4 Tire Hose Stations 4-79 4.17.5 Penetration Tire Barrier:
4-80 4.18 PobTR DISTRIBUTION INSTRUMENTATION 4 81 4.18.1 Incore Detectors 4-81 4.18.2 Excore Monitoring System 4 82 4.19 POWER DISTRIBUTION LIMITS 4-83 4.19.1 Linear Neat Rate 4-83 4.19.2 Radial Peaking Tactors 4-84 4.20 Moderator Temperature Coefficient (MTC) 4-85/
4.21 (latentionally Left Blank) 4-86 4.22 (Intentionally Lef t Blank) 4-87 4.23 (Intentionally Left Blank) 4-88 (Intentionally Left Blank) 4-89 4.24 RADIOLOGICAL ITTLLT.NT RELEASES 4-90 4.24.1 Radiological Liquid Effluent Monitoring Instrumentation 4-90 4.24.2 Radiological Caseous Effluent Monitoring Instrumentation 4-90 4.24.3 Liquid Effluent Concentration 4-90 4.24.4 Liquid Effluent Dose 4 90 4.24.5 Caseous Effluent Dose 4-90 v
Amendment No. 37, 48, p), $7, 88, 83,,144/,118
1.1 REACTOR OPERAT7NC CONDITIONS (Ccatd) i Low Power Phystes Testing Testing performed under approved written procedures to determine i
control rod worths and other core nuclear properties. Reactor power during thfse tests shall not exceed 2% of rated power not including l
decay heat and primary system temperature and pressure shall be in the range of 260'T to 538'T and 415 psia to 2150 psia, respectively.
Certain deviations from normal operating practice which are i
necessary to enable performing some of these tests are permitted in accordance with the specific provisions therefor in these Technical Specifications.
J Shutdown Boron Concentratione Boron concentration sufficient to provide k 5 0.98 with all control rods in the core and the highest woI b control rod fully withdrawn.
I Refueling Boron Concentratieti
)
{
Boron concentration of coolant at least 1720 ppa (corresponding to a ahutdown margin of at least 52 40 with all control rods withdrawn).
Quadrant Power Tilt The difference between nuclear power in any core quadrant and the average in all quadrants.
Assembly Radial Peaking Tactor - T ^
r 1
The assembly radial peaking f actor is the maximum ratio of l
individus.1 fuel assembly power to core average assembly power I
integrated over the total core height, including tilt.
}
/
i
/
/
/
t Interior Fuel Rod i
Any fuel rod of any assembly that is not on that assembly's periphery.
i Total Interior Rod Radial Peaking Tactor - F,AI J
r The maximum product of the ratio of individual assembly power to
, I core average assembly power times the highest interior local peaking j
l factor integrated over the total core height including tilt.
j i
l 1-2 Amendment No. Jf. (J. Jd. 31, 68, 118
{
1 I
l l
e i
1.1
, REACTOR OPERATING CONDITIONS (Continued) i Axial Offset or Axial Shape Inde_x
/
l
\\
The power in the lower half of the core minus the power in the
/
upper half of the core divided by the sua of the powers in the lower half and upper half of the core.
j
/
r i
/
l
/
i
- )
I I
l l
I i
4 t
i t
i I
l j
i l
f i
r o
i d
k i
i i
i i
I i
s i
i l
i I
1-2a l
Amendment No. 118 1
l i
t i
J
5 l
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFITy LIMITS - REACTOR CORE l
Asslicability i
This specification applies when the reactor is in hot standby
/
condition and power operation condition.
/
Objective To maintain the integrity of the fuel cladding and prevent the release of significant amounts of fission products to the primary coolant.
i Specificationa The MDNRR of the reactor core shall be maintained greater than or equal
/
to 1.17.
/
l'
/
Basis To maintain the integrity of the fuel cladding acd prevent fission product release, it is necessary to prevent overheating of the cladding under normal operating conditions. This is accomplished by operating within the nucleate boiling regime of best transfer, wherein the heat transfer coefficient is large enough so that the clad surface I
temperature 'is only slightly greater than the coolant temperature. The upper boundary of the nu.a. ate boiling regime is termed "departure free nucleate boiling" (DN1). At this point, there is a sharp reduction of the best transfer coefficient, which would result in high-cladding i
temperatures and the possibility of cladding failure. Although DN3 is not an observable parameter during reactor operation, the observable paraanters of thermal power, primary coolant flow, temperature and pressure, can bt related to DN3 through the use of the XNB DN3
/
Correlation."II) The INE DK5 Cortelation has been developed to predict
/
r DN3 and the location of DK5 for axially uniform and nonunifors heat fluz distributions. The local DN3 ratio (DNER), defined as the ratio of the heat flux that would cause DNB at a parti?slac core location to the actual heat flua, is indicative of the margia to DNB. The minimum value of the DNER, during steady-state operatiot, moraal operational f
transients, and anticipated transients is lisidid to 1.17.
A DN3R of
/
1.17 corresponds to a 95% probability at a 95% confidence level that
/
(
f i
l 2-1 Amendment No 118
.c..<. - ___.__.__ -............... _.....__.:_..__._...._...
......_......u...
l t
I
~
\\
t 2.1 SAFETT LIMITS - REACTOR C,0R), (Contd)
DNB will met occur which is co aidered an appropriate marsia to DNR for'
/
all operating conditiona.II)
//
l
/
L
/
3 The reactor protective system is designed to prevent any anticipated
(
t combination of treasiest condittoes for primary coolant systes temperature, pressure and thermi power level that would result is a L
53)
DNBR of less than 1.17 The XNB DNB correlation has been shown to be
/
applicable to the Palisades Flaat in Reference 2.
/
/
References
/
l
/
(1) XN-NT-621(P)(A), Rev 1
/
l
.i (2) XN-NT-709
/
l l
(3) Updated TSAR, Section 14.1.
/
i l
l l
1 I
l i
4 e
1 a
i 2-2 l
Amendment No 3Ie 33 'I I
i i
i
\\
l
2.2 SATETY LIMITS - PRIMARY COOLANT SYSTEM PRESSURE Applicability Applies to the limit on primary coolant system pressure.
Objective To maintain the intsgrity of the primary coolant system and to prevent the release of significant amounts of fission product activity to the primary coolant.
Spect(teatten The primary coolant system pressure shall not exceed 2750 psia when there are fuel assemblies in the reactor vessel.
i Basis IU The prieary coolant system series as a barrier to prevent radionuclides in the primary coolant from reaching the atmosphere.
In the event of a fuel cladding f ailure, the primary coolant system is the foremost barrier against the release of fission products.
t Establishing a system pressure limit helps to assure the continued integrity of both the primary coolant system and ths fuel cladding.
The maximum transient pressure allevable in the primary coolant syste.
pressure vessel under the ASMI Code Section III, is 110% of design pressure. The maximum transient pressure allowable in the primary coolant system piping, valves and fittings under ASA Section 531.1 fr-120% of design pressure. Thuv. the safety limit of 2750 psia (110:
)
of the 2500 psia design pressure) has been established.(2) The settings and capacity of the secondary coolant system safety valves t
(985-1025 psig)I3). the reactor high-pressure trip (:2400 psia) and the primary safety valves (2500-2580 psia)I'I have been establisaed l
to assure never reaching the primary coolant system pressure safety l
limit. The initial hydrostatic test was conducted at 3125 psia (125%
of design pressure) to verify the integrity of the primary coolant system. Additional assurance that the nuclear steam supply system (NSSS) pressure does not exceed the safety limit is provided by I
setting the cecondary coolant system steam dusp and bypass valves
/
at 900 psia.
References (1) Updated TSAR. Section 4.
/
(2) Updated TSAR Section 4.3.
/
(3) Updated TSAR, Table 4-5
/
(4) Updated TSAR, Table 4-10
/
i
{
l 2-1 l
Amendeent No 25, 116 i
5 2.3 IIMITING SAFETT SYSTTM SETTINGS - REACTOR PROTECTIVE SYSTEM Applicability This specification applies to reactor trip settings and bypasses for instrument channels.
Objective To provide for automatic protective action in the event that the principal process variables approach a safety limit.
$necification The reactor protective system trip setting limits and the permissible bypasses for the instrument channels shall be as stated in Table 2.3.1.
The TM/1.P trip system monitors core power, reactor coolant maximum
/
inlet temperature. (T,).
ore coolant system pressure med axial shape
/
g index. The low pressure trip limit (Pvar) is calculated using the
/
following equation.
/
1 1
$0 * (0^}(0 1} +
- ( in} ~ '
P var i
wheret QR
= 0.412(Q) + 0.588 Q s 1.0 Q = core pcwer
/
g
=Q Q > 1.0 rated power
/
= -0.691(ASI) + 1.058
-0.653 < ASI < -0.156
/
i
= -0.521(ASI) + 1.085
-0.156 < ASI < +0.162
/
0.226(AS1) + 0.964
+0.162 < ASI < +0.544
/
=
l 1.085 when Q < 0.0625
/
=
The calculated limit (P,7) is then compared to a fixed icv pressure
/
y trip limit (P,gn). De auctionund highut of then signals buoan
/
trip.
P,,g, h compand to the asuund nactor
/
l the trip limit (P i
coolant pressure (P) and a trip si aal is generated when P is less
/
[
A pu-trip a am oas gnuate wnP
/
l than er equal to P trip.
is less than or equal to the pre-trip setting Ptrip + AP.
/
f i
l 2-4 Amendment No 113 i
i i
I A
~ '
TABLE 2.3.1
/-
Ai
/
i Reacto: Protect 1*6
.** sten Trip Setting Limits
/
l l
/
l t
cour Primary Coolant ThreePrimaryCoogt
/
7 Pumps Operatina pus'i Operating
/
l
/
i l
1.
Varin g High 5101 above core power, 5101 above core power
/
l Power with a minimum setpoint witt a minimum setpoint
/
of $30% of rated power of 5153 rated power
/
and a maximum of 5106.51 and a maximum of 5491
/
of rated power of rated power
/
/
2.
Primary 4951 of Primary Coolant 160% of P:1sary Cool-
/
Coolant Plow (2)
Plow With Pour Pumps ant Plow With Tcur
/
Operating Pumps Operating
//
3.
High Pressure 52255 Psta
$2255 Psta
/
Pressuriser
/
/
4 Thermal g g n/ Low P
t Applicable Limits Replaced by Variable
/
trip High Power Trip and
/
Pressure 1750 Psia Minimus L &-
/,
Pressure Sett*,eg
/
/
5.
Steam Oenerator Not Lower Than the Cen-Not Lower T' san the Cen-
/
Low Water Level ter Line of Teed-Water ter Line of Teed-Water
/
Ring Which Is Located Ring Which Is Located
/
6'-0" Below Normal 6'-0" Below Normal
/
Water level Water Level
/
/
6.
SteamGenerag 1500 Pata 2500 Psta
/
Low Pressure
/
l 7.
Containment High 53.70 Fats 53.70 Pais
/
Pressure
/
/
(1) The VHPT can be 30% of rated power for power levels s 20% of rated
/
power.
/
(2) May be bypassed below 10I of rated power provided auto bypass removal
-circuitry is operable. Por low power physics ter.s. thermal margin / low pressure, primary coolant flow and low steam generator pressure trips may
/
be bypassed until their react points are reached (approximately 1750 psia and5,0gpsia,respectively),providedautomaticbypassrenovalcircuitry at 10 I rated power is operable.
(3) Minimum trip setting shs11 be 1750 psia.
/
(4) Operation with three pumps for a maximum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is permitted to
/
provide a lipited time for repair / pump restart to provide for an orderly
/
shutdown or to provide for the conduct of reactor internals noise
/
sonitoring test sensurements.
/
2-5 Amendment No 3!, f f,118
... ?..... u J, i.... _ -...-
.Z
'..... ii.. n s...
e a
2.3 LIMITING SA*177 '>TSTEM SE_7 TINGS - REACTOR PROTECTIVE SYSTEM (Contd) ga,tjs The rea1 p-ative system consists of four instrument hannels to notator a ilast conditions which will cause a reactor trip if any of \\r itions deviate from a preselected operating range to the de-a safety limit may be reached.
1.
Variab_1(._High Power - The variable high power trip (VHPT) is
/
incorporated in the reactor protection system to provide a reactor
/
trip for transients e:hibiting a core power increase starting from
/
any initial power level (such as the boron dilution transient).
/
The VHPT system provides a trip set 11oint no more than a
/
predeteraired amour.t above the indicated core power. Operator
/
action is required to increase the setpoint as core power is
/
ine.> eased; the setpoint is automatically decreased as core power
/
decreases. 7s6 visions have been made to select different set points
/
for three pump and four pump operations.
/
During normal plaat operation with all primary coolant pumps operating, reactor trip is initiated when the reactor power level reaches 106.5% of indicated rated power. Adding to this the possible variation in trip point due to calibration and instrument errors, the maximas actual st sady state power at which a trip would be actuated is 112%, which w., used for the purpose of safety analysis.(I)
/
/
/
/
/
2.
, Primary Coolant System Low Flow - A reactor trip is provided to
/
protectthecoreagainsg3gNBshouldthecoolantflowsuddenly l
decr1ase significantly.
Flow in each of the four coolant lonps
/
is determined from a measurement of pressure drop from inlet to outlet of the steam generators. The total flow through the reactor core is measured by summing the loop pressure drops across the steam generators and correlating this pressure sum with the pump calibration flow curves. The percent of normal core flow is shown in the following table:
/
4 Pumps 16h.0%
3 Pumps 74.7%
/
During four-pump operation, the low-it'ow trip setting of 95%
l insures that the reactor cannot operate when the flow rate is L
errors.{g)93%ofthenonimalvalueconsideringinstrument less th
/
i 2-6 Amendment No 32,113
---w-.-
w
- ... -. s --
. '. u. l
- .l..'.=..
+.
e 2.3 LIMITING SATETY SYSTEM SETTINGS - REACTOR PROTECTIVE SYSTEM (Contd)
Basis (Contd)
Provisions are made in the reactor prottetive system to permit
/
operation of the reactor at reduced power if one coolant pump is
/
taken out of service. These low-flow and high-flux settings have
/
been derived in consideration of instrument errors and response
/
times of equipment involved to assure that thermal margin and flow
/
stability w{jj be maintained during normal operation and anticipated /
transients For reactor operation with one coolant pump
/
inoperative, the low-flow trip points and the overpower trip points
/
must be manually changed to the specified values for the selected
/
pump condition by means of set point selector switches. The trip
/
points are shown in Table 2.3.1.
/
3.
High Pressurizer Pressure - A reactor trip for high pressurizer pressure is provided in conjunction with the primary and secondary safety valves to prevent prinary system overpressure (Specification 3.1.7).
In tha event of lo.3s of load without reactor trip, the temperature e.J pressura of the primary ce41aat system would increase due to the reduction in the h st removed from the coolant via the steam generators. This setting is consistent with the
/
trip point assumed in the accident analysis.(II) i I
l 2-7 Amendment No 31, 118 l
l 1
l
2.3 LIMITING SAFETY 3YSTEM SETTINGS - REACTOR PROTECTIVE SY e
Basis (Continued) 4.
Thermal Margin / Low-Pressure Trip The TM/LP trip set points are derived from the 4 pump operation
/
core thetual limits through application of appropriate allowances
/
for measuresent uncertainties and processing errors.
A pressure allowance of 165 psi is assumed to
//
account for:
instrument drift in both power and inlet temperatures;
/
calortsetric power measurement; inlet temperature seasurement; and
/
primary system pressure seasurement.
Uncertainties accounted for
/
that are not a part of the 165 psi ters include allowances for:
/
assembly power tilt; fuel pellet manufacturing tolerances; core
/
flow seasurement uncertainty and core bypass flow; inlet temperature
/
measurement time delays; and ASI measurement. Each of these
/
allowances and uncertainties are included in the development of
~
/
the TM/LP trip set point used in the accident analysis.
/
For three-pump operation, continued power operation is restricted.
/
During thi: ade'ei opstativu, the high power level trip in
/
conjunction with the TM/LP trip (minimum set point = 1750 psia) and the secondary system safety valves (set at approximatggy 1000 psia) assure that adequate DNB margin is maintained
/
5.
Low Steam Generator Water Level - The low steam generator water level reactor trip protects against the loss of feed-water flow accidents and assures that the design pressure of the primary coolant system will not be exceeded. The specified set point assures that there will be sufficient water inventory in the steam generator at the time of trip to allow a safe and orderly
/
plant shutdown and to prevent steam generator dryout assuming
/
minimum auxiliary feedwater capacity.(9)
/
The setting listed in Table 2.3.1 assures that the heat transfer surface (tubes) is covered with water when the reactor is critical.
2-8 Amendment No 31, 82, 118
(
.' m :
~
s
~
0 2.3 LIMITING SAFETY SYSTEM SETTINGS - REACTOR PROTECTIVE SYSTEM (Contd)
Basis (Contd) 6.
Low Steam Generator Pressure - A reactor trip on low steam generator secondary pressure is provided to protect against an excessive rate of heat extraction from the steam generators and subsequent cooldown of the primary coolant. The setting of 500 psia is sufficiently below the rated load operating point of 739 psia so as not to interfere with normal operation, but still high enough to provide the required protection in the event of excessively high steam flow. This setting was used in the accident analysis.(8) 7.
Containment Hish Pressure - A reactor trip on containment high pressure is provided to assure that the reactor is shut down before
/
theingtionofthesafetyinjectionsystemandcontainment
/
spray.
/
8.
Low Power Physics Testina - For low power physics tests, certain tests will require the reactor to be critical at low temperature (1 260'F) and low pressure (1 415 psia). For these certain tests only, the thermal cargin/ low prusure, primary coolant flow and low - /
steam generator pressure trips say be bypassed in order that reactor power can be increased for improved data acquisition. Special operating precautions will be in effect during these tests in accordance with approved written testing procedures. At reactor power levels below 10"I7,of rated power, the thermal margin / low-pressure trip and low flow trip are not required to prevent fuel
/,
rod thermal limits from being exceeded. The low steam generator pressure trip is not required because the low steam generator pressure will not allow a severe reactor cooldown, should a steam line break occur during these tests.
References (1) ANT-87-150(P), Volume 2, Table 15.0.7-1
/
(2) deleted
/
(3) Updated FSAR, Section 7.2.3.3.
/
(4) ANT-87-150(P), Volume 2, Section 15.3
/
(5) XN-NF-86-91(P)
/
(6) deleted
/
(7) deleted
/
(8) XN-NT-77-18, Section 3.8
/
(S) ANT-87-150(P), Volume 2, Section 15.2.7
/
i l
(10) Updated FSAR, Section 7.2.3.9.
/
(11) ANT-87-150(P), Volume 2, Section 15.2.1
/
(12) ANT-87-150(P), Volume 2, Section 15.0.7.2
/
//
10 2-9 Amendment No 3Z,118
=.
3 3.1 PRIMARY COOLANT SYSTEM Applicability Applies to the operable status of the primary coolant system.
Objective To specify certain conditions of the primary coolant system which must be met to assure safe reactor operation.
Specifications 3.1.1 Operable components a.
At least one primary coolant pump or one shutdorn cooling pump
/
with a flow rate greater than or equal to 2810 spa shall be in
/
operation whenever a change is being made in the boron
/
concentration of the primary coolant and the plaat is
/
vyerating in cold shutdown or above, except durit g an emergency
/
loss of coolant flow situation. Under these circiastances, the
/
boron concentration may be increased with no priau ry coolant
/
pumps or shutdown cooling pumps running.
//
/
b.
Four primary coolant pumps shall be in operation wienever the
/
reactor is operated above hot shutdown, with the following
/
axceptions
/
/
Before removing a pump from service, thermal power shall be
/
reduced as specified in Table 2.3.1 and appropriate corrective
/
action implemented. With one pump out of service, return the
/
pump to service within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (return to four-pump operation)
/
or be in hot shutdown (or below) with the reactor tripped (from
/
the C-06 panel, opening the 42-01 and 42-02 circuit breakers)
/
within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Start-up (above hot shutdown) with
/
less than four pumps is not permitted and power operation with
/
less than three pumps is not permitted.
/
/
The measured four primarg coolant pumps operating reactor vessel
/
c.
flow chall be 124.3 x 10 lb/hr or greater, when corrected to
/
532'F.
/
/
/
\\
d.
Both steam generatirs shall be capable of performing their heat transfer function whenever the average temperature of the primary coolant is above 325'F.
Maximum primary systes pressure differentials shall not exceed e.
i the following:
l (1) Maximum steam generator operating differential of 1380 psi.
/
l 3-1b Amendment No 31, 83, 118 l
t
3.1 PRINARY COOLANT SYSTEM (Centinu:d)
?
3.1.1 Operrbin Compon nts (Ccatinu d)
(2) Hydrostatic tests shall be conducted in accordance with i
applicable paragraphs of Section XI ASME Boiler & Pressure Vessel Code (1974).
Such tests shall be conducted with sufficient pressure on the secondary side of the steam generators to restrict primary to secondary pressure differential to a maximum of 1380 psi. Maximum hydrostatic test pressure shall not exceed 1.1 Po plua 50 psi where Po is nominal operating pressure.
1 (3) Primary side leak tests shall be conducted at normal
[
operating pressure. The temperature shall be consistent-i with applicable fracture toughness criteria for ferritic materials and shall be selected such that the differential pressure across the steam generator tubes is i
not greater than 1380 psi.
[
(4) Maximum secondary hydrostatic test pressure shall not exceed 1250 psia. A miniaua temperature of 100*F is required. Only ten cycles are permitted.
i (5) Maximum secondary leak test pressure shall act exceed l
1000 psia. A minimum temperature of 100'F is required.
(6)
In performing the test; identified in 3.1.1.e(4) and 3.1.1.e(5), above, the secondary pressura shall not exceed the primary pressere by more than 350 psi.
f.
Nominal primary system ope.ation pressure shall not exceed 2100 psia.
l g.
The reactor inlet temperature (indicated) shall not exceed the value given by the fo11owin6 equation at steady state power operations
/
i i
T 3 543.3 +.0575(P-2060) + 0.00005(P-2060)**2 + 1.173(W-120) -
/
l inlet.0102(W-120)**2
/
i i
= reactor inlet temperature in F' l
Where T I"I'* P = nominal operating pressure in psia l
6 W = total recirculating mass flow in 10 lb/h i
t i
corrected to the operating temperature i
conditions.
6 i
i When the ASI exceeds the limits specified in Figure 3.0, within
/
15 minutes, initiate corrective actions to restore the ASI to
/
l l
the acceptable region. Restore the ASI to acceptable values
/
l within one hour or be at less than 70% of rated power within
/
+
the following two hours.
/
If the measured primary coolant system flow rate is greater
/
than 130 M lba/hr, the maximum inlet temperature shall be
/
less than or equal to the T LCO at 130 M lba/hr.
/
l Inlet t
l 3-Ic Amendment No 3I.II.8Iell1 II8 l
t i
i
3.1 PRIMARY COOLANT SYSTEM (Cont'd) 3.1.1 Operable Components (Cont'd) h.
During initial primary coolant pump starts (i.e., initiation of forced circulation), secondary system temperature it. the steam generators shall be < the PCS cold leg temperature unless the PCS cold leg temperature is > 450'F.
1.
The PCS shall not be heated or maintained above 325'T unless a minimum of 375 kW of pressurizer heater capacity is available from both buses ID and IE.
Should heater capacity from either
' bus ID or 1E fall below 375 kW, either restore the inoperable heaters to provide at least 375 kW of heater capacity from both buses ID and 1E within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Basis When primary coolsat *uoron concentration is being changed, the process must be uniform throughout the primary coolant system volume to prevent stratification of primary coolant at lower boron concentration which could result in a reactivity insertion.
Sufficient mixing of the primary coolant is assured if one shutdown cooling or one primary coolant pump is in operation.( )
The shutdown cooling pump will circulate the primary system volume in less than 60 minutes when operated at rated capacity. By imposing a
/
minimum shutdown cooling pump flow rate of 2810 gpm, stifficient time
/
is provided for the operat
/
asymmetricflowconditions.[6)oterminatetheborondilutionunder The pressurizer volume is relatively
/
inactive, therefore will tend to have a boron concentration higher 3
than rest of the primary coolant system during a dilution operation.
Administrative procedures will provide for use of pressurizer sprays to maintain a nominal spread between the boron concentration in the pressurizer and the primary system during the addition of boron.( )
The FSAR safety analysis was performed assuming four primary coolant
/
pumps were operating for accidents that occur during reactor
/
operation. Therefore, reactor startup above hot shutdown is not
/
permitted unless all four primary coolant pumps are operating.
/
Operation with three primary coolant pumps is permitted for
/
a limited time to allow the restart of a stopped pump or for
/
l reactor internals vibration monitoring and testing.
/
Requiring the plant to be in hot shutdown with the reactor tripped
/
from the C-06 panel, opening the 42-01 and 42-02 circuit breakers,
/
assures an inadvertent rod bank withdrawal will not be initiated
/
by the control room operator. Both steam generators are required
/
to be operable whenever the temperature of the primary coolant is
/
greater than the design temperature of the shutdown cooling system
/
to assure a redundant heat removal system for the reactor.
/
l l
l Amendment No 67, #J, 118 l
l l
{
t
(
..-...u.,
s..
O 3.1 PRIMARY COOLANT SYSTEM (Contd)
Basis (Contd)
Calculations have been performed to demonstrate that a pressure differential of 1380 psi (3) can be withstood by a tube uniformily
/
thinned to 36% of its original nominal wall thickness (64% degradation), while maintaining:
(1) A f actor of safety of three between the actual pressure differential and the pressure differential required to cause bursting.
(2) Stresses within the yield stress for Inconel 600 at operating temperature.
(3) Acceptable stresses during accident conditions.
/
//
Secondary side hydrostatic aud leak testing requirements are consistent with ASME BPV Section XI (1971). The differential maintains stresses in the steam generator tube walls within code allowable stresses.
The minimum temperature of 100'T for pressurizing the steam generator secondary dide is set by the NDTT of the mayway cover of + 40*F.
The transient analyses were performed assuming a vessel flow at hot zero power (532'F) of 124.3 x los 1b/hr minus 6% to account for flow
/
measurement uncertainty and core flow bypass. A DNB analysis was
/
performed in a parametric fashion to determine the core inlet
/
temperature as a function of pressure and flow for which the
/
sinimum DNBR is equal to 1.17.
This analysis includes the
/
following uncertainties and allowances: 2% of rated power for power
/
measurement; 10.06 for ASI measurement; 150 psi for pressurizer
/
pressure; 27'F for inlet temperature; and 3% measurcaent and 3%
/
bypass for core flow.
In addition, transient biases were included in
/
the derivatiog43f the following equation for limiting reactor inlet
/
temperatv s
/
/
T,g,g 1 543.3 +.0575(P-2060) + 0.00005(P-2060)**2 + 1.173(W-120) -
/
g
.0102(W-120)**2
/
/
The limits of validity of this equation ares
/
1800 < Pressure < 2200 Psia
/
s 100.0"x 108 i Vessel Flow 1 130 x lo Lb/h
/
ASI as shown in Figure 3.0
/
With measured primary coolant system flow rates > 130 M lba/hr,
/
limiting the maximum allowed inlet temperature to the T,g,g LCO
/
y
/
at 130 M lba/br increases the margin to DNB for higher PCS flow rates.
3-2 Amendment No 20, )), IId
d 3.1 PRIMARY COOLANT SYSTEM (Cont'd)
Basis (Cont'd)
The Axial Shape Index alarm channel is being used to monitor the
/
ASI to ensure that the assumed axial power profiles used in the
/
development of the inlet temperature LCO bound measured axial power
/
profiles. The signal representing core power (Q) is the
/
auctioneered higher of the neutron flux power and the Delta-T power.
/
The measured ASI calculated from the excore detector signals and
/
adjusted for shape annealing (Y ) and the core power constitute an
/
7 ordered pair (Q,Y ).
An alarm signal is activated before the
/
7 ordered pair exceed the boundaries specified in Figure 3.0.
/
The requirement that the steam generator temperature be < the PCS temperature when forced circulation is initiated in the PCS ensures that an energy addition caused by heat transferred from the secondary system to the PCS will not occur.
This requirement applies only to the initiation of forced circulation (the start of the first primary coolant pump) when the PCS cold leg temperature is < 450*F.
At or above 450'F, the PCS safety valves prevent the PCS pressure from exceeding 10CFR50 Appendix G limits.
References (1) Updated FSAR, Section 14.3.2.
/
(2) Updated FSAR, Section 4.3.7.
/
(3)
Palisades 1983/1984 Steam Generator Evaluation and Repair
/
Program Report, Section 4. April 19, 1984
/
(4) ANF-87-150(P), Volmte 2. Section 15.0.7.1
/
(5)
(Deleted)
(6) A NF 108
/
i I
3-3 Amendment No 2f, 51,121,118 L
t:
j i.
- [ -
FIGURE 3-0 l
ASI LCO FOR Tinlet FUNCTION l l-I, I
1.15 i
l-UNACCEPTABLE OPERATIONS
~
1.00 j
2 3
eW
~
b l
2 0.85 j
o i
m l
BREAK DOINTS a-y=
l j
't u.
0.70 I
O 1
1.
.300, 0.7 I
i, z
(
o
(.
2.
.080, 1.0 0.55 4
ACCEPTABLE E
OPERATIONS
- 3. +. 6, 1.0 i
t 0.40
?r H
s.
h I
l l
I I
l l
I i
0.25
- e_
.n
-0.4
-0.2 0
0.2 0.4 C.6 t
Y p
AXIAL SHAPE INDEX
,--,n
_,_c
<-,n
PRIMARY COOLANT SYSTEM (Centd) 3.1 3.1.7 Primary and Secondary Safety Valves Specifications
- a. The reactor shall not be made critical unless all three pressurizer safety valves are operable with their lift settings maintained between 2500 psia and 2580 psia (* 1%).
- b. A minimum of one operable safety valve shall be installed on the pressurizer whenever the reactor head is on the vessel.
- c. Whenever the reactor is in power operation, a minimum of 23 secondary system safety valves shall be operable with their lift settings betveau 985 psig (2 30 psis) and 1025 (2 3%) psig.
Basis The primary and secondary safety valves pass sufficient steam to limit the primary system pressure to 110 percent of desigt (2750 psia) following a complete loss of turbine generator load without sinultaneoua reactor trip while operating at 2650 MW g g.
The reactor is assumed to trip on a "High Primary Coolant System Pressure" signal. To determine the maximum steam flow, the only other pressure relieving system assumed operational is the secondary system safety valves. Conservative values for all system parameters, delay times and core moderator coefficient are assumed. Overpressure protection is provided to the portions of the primary coolant system which are at the highest pressure considering pump head, flow pressure drops and elevation heads.
If no residual heat were removed by any rif the means available, the amount of steam which could be generated at safety valve lift pressure would be less than half of one valve's capacity. One valve, therefore, provides adequate defense against overpres-surization when the reactor is suberitical.
Thetotalregiefcapacityofthe24secondarysystemsafetyvalves is 11.7 x 10 lb/h. This is based on a steam flow equivalent to an l
NSSS power level of 2650 MW, at the nominal 1000 psia valve lift pressure.
At the power rating of 2530 MW. a relief capacity of less than t
6 11.2 x 10 lb/h is required to prevent overpressurization of the l
secondary system of loss of load conditions, and 23 valves provide 0 lb/h.U' relieving capability of 11.2 x 10 The overpressurization analysis for the loss of load event ( }
l supports the specified secondary safety valve lift pressure tolerance. ASME B&PV Code, 1986 edition Section XI, subsection IWV-3500, specifies ANSI /ASME OM-1-1981 requirements which allow the specified tolerances in the lift pressures of the safety valves.
References (1) Updated FSAR, Section 4.3.4 and 4.3.9.4
/
(2) ANF-87-150(NP). Volume 2 Section 15.2.1 3-25 Amendment No JI ll6. II8 l
l
.L.,N'a,- _..$.
'. 2 m.......,.a 3.5 STEAM AND FEEDWATER SYSTEMS (Cont'd)
BASIS The Steam and Power Conversion System is designed to receive steam-from the NESS and convert the ster.a thermal energy into electrical energy. A closed regenerative cycle condenses the steam from the main turbine and returns the condensate as heated feedvater to the steam generators. Normally, the capability to supply feedvater to the steam generators is provided by operation of the turbine-driven main feedwater pumps.
A reactor shutdown from power requires removal of core decay heat.
Immediate decay heat removal requirements are normally satisfied by the steam bypass to the condenser, or by steam discharge to the valves. g ia the main steam safety valves or power operated relief atmosph If the main feedvater pumps are not operating, any one auxiliary feedwater pump can supply sufficient feedwater for removal of decay heat from the Plant. The Plant is provided with two motor driven auxiliary feedwater pumps (P-%, P-8C) and one turbine driven auxiliary feedvater pump (P-85). The Auxiliary Feedvater System is designed so that an automatic start signal is generated to the auxiliary feedvater pumps upon lov secondary side steam generator level.
Upon low secondatry side steam generator level, auxiliary feedvater pump P-8A would be the first auxiliary feedvater pump to receive an automatic start signal. If pump P-8A failed to start or establish flow within a specified period of time, auxiliary feedvater pump P-8C would receive an automatic start signal. If both pump P-8A and pump P-8C failed to start or establish flow within each pump's specified period of time, auxiliary feedvater pump P-8B vould receive an automatic start signal. All three auxiliary feedvater pumps normally tskea suction from the condensate storage tank. The minicium amount of water in the condensate storage tank and primary coolant system makeup tanks combined is the amount needed for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of auxiliary fradvater pump operation. If the outage is more than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Lake Michigan water can be used, by utilizing a fire pump to supply water to the auxiliary feedvater
(
pumps P-8A and P-85, or by utilizing a service water pump to supply i
vater to auxiliary feedvater pump P-8C.
Thrie fire pumps are provided, one motor driven and two diesel driven, each capable of delivering 1500 spa at 125 psig. Three service water punrps are provided, all of which are motor driven, each capable of delivering 8000 gpm at 60 psig.
RITERENCES l
(1) Updated FSAR, Section 10.2.1
/
(2) ANF-87-150(P), Volume 2. Section 15.2.7
/
/
3-39 Amendment No 62,96,118 l
l
~
3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Applicability Applies to operation of' control rods and hot channel factors during operation.
Objective To specify limits of control rod movement to assure an acceptable power distribution during power operation, limit worth of individual rods to values analyzed for accident conditions, maintain adequate shutdown margin after a reactor trip and to specify acceptable power limits for power tilt conditions.
Specifications 3.10.1 Shutdown Margin Requirements With four primary coolant pumps in operation at hot shutdown a.
and above, the shutdown margin shall be 2%.
b.
With less than four primary coolant pumps in operation at hot shutdown and above, boration shall be immediately initiated to increase and maintain the shutdown margin at 3 3.75%.
c.
At less than the het shutdown condition, with at least one
/
primary coolant pump in operation or at least one shutdevn
/
cooling pump in oneration, with a flow rate 3 2810 gpm, the
/
boron concentrat.
shall be greater than the cold shutdown
/
boron concentration for normal cooldswns and heatups, ie.
/
nonemergency conditions.
/
During nonemergency conditions, at less than the hot
/
shutdown condition with no operating primary coolant pumps
/
and a primary system recirculating flow rate < 2810 gpm
/
but 3 650 gpm, then within one hour either
/
1.
(a) Establish a shutdown margin of 2 3.5%
/
and
/
(b) assure two of the three charging pumps are
/
electrically disabled.
/
I' OR
/
2.
At least every 15 minutes verify that no charging pumps
/
are operating. If one or more charging pumps are
/
determined to be operating in any 15 minute surveillance
/
period, terminate charging pump operation and insure that
/
]
the shutdown eargin requirements are not and meintained.
/
t i
3-58 Amendment No. 2.43,51,68,10,118
3.10 CONTROL ROD AND POWER DISTRIBUTION 1.IMITS (Continued)
/
~
3.10.1 Shutdown Margin Requirements (Continued)
/
During nonemergency conditions, at less than the hot shutdown
/
condition with no operating primary coolant pumps and a
/
primary system recirculating flow rate less than 650 gpm,
/
vithin one hour:
/
(a) Initiate surveillance at least every 15 minutes to verify
/
that no charging pumps are operating.
If one or more
/
charging pumps are determined to be operating in any
/
15-minute surveillance period, terminate charging pump
/
operation aniinsure that the shutdown margin requirements
/
are met and tafatained.
/
d.
If a control rod cannot be tripped, shutdown margin shall be increased by boration as necessary to compensate for the worth of the withdrawn inoperable rod.
e.
The drop time of each control rod shall be no greater than 2.5 seconds from the beginning of rod motion to 90%
insertion.
3.10.2 (Deleted)
/
3.10.3 Part-Length Control Rods The part-length control rods will be completely withdrawn from the core (except for control rod exercises and physics tests).
3-51' Amendment N>.
118
4 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS (Centd) 3.10.6 Shutdown Rod Limits a.
All shutdown rods shall be withdrawn before any regulating rods are withdrawn.
b.
The shutdown rods shall not be withdrawn until normal water level is established in the pressurizer.
c.
The shutdown rods shall not be inserted below their exarcise limit until all regulating rods are inserted.
3.10.7 Low Power Physics Testing Sections 3.10.1.a. 3.10.1.b. 3.10.3, 3.10.4.b, 3.10.5 and 3.10.6
/
may be deviated from during low powar physics testing and CRDM exercises if necessary to perform a test but only for the time necessary to perform the test.
i 3.10.8 Center control Rod Misc.lignment The requirements of Specifications 3.10.4.1, 3.10.4.a. and 3.10.5 may be suspended during the performance of physics tests to determine the isothermal temperature coefficient and power coefficient provided that only the center control rod is misa11gned and the limits of Specification 3.23 are maintained.
Basis l
Sufficient control rods shall be withdrawn at all times to assure that the reactivity decrease from a reactor trip provides adequate shutdown margin. The available worth of withdrawn rods must include the reactivity defect of power and the failure of the withdrawn rod of highest worth to insert. The requirement for a shutdown eargin of 2.0% in reactivity with 4-pump operation, and 1
of 3.75% in reactivity with less than 4-pump operation, is consistent with the assumptions used in the analysis of accident conditiens (including steam line break) as reported in Reference 1 and 2 and additional analysis. Requiring the boron
/
concentration to be at cold shutdown boron concentration at
/
less than hot shutdown assures adequate shutdown margin exists
/
to ensure a return to power does not occur if an unanticipated
/
cooldown accident occurs. This requirement applies to normal
/
operating situations and not during emergency conditions where
/
it is necessary to perform operations to mitigate the
/
consequences of.an accident.
By imposing a minimum shutdown
/
l cooling pump flow rate of 2810 gpm, sufficinnt time is provided
/
t for the operator to terminate a boron dilution under asymmetric
/
l conditions.
For operation with no primary coolant pumps operating
/
and a recirculating flow rate less than 2810 spa the increased
/
-hutdown margin and controls on charging pump operability or
/
alternately the surveillance of the charging pumps will ensure
/
that the acceptance criter1
/
eventwillnotbeviolated.93foraninadvertentborondilution The change in insertion limit
/
with reactor power shown on Figure 3-6 inseres that the shutdown 3-61 Amendment No 11,54451,68,118
.~,.a..
.... a
.a..... :..
..+.........
N MAP OPERADON l
90 to C
IRA'X9828'hR LEVEL x
I,,
N l
yn m
4 10 I
0 80 30 0
20 40 40
=== @ 00 80 00 0
20 N O, 40,,,gto 0
E C00ffROL A00 WWERT1001 PEltCENT fouR Mar OPERATION 10 0 90
- so I
N N
R 7'
N N
\\
B" 4
" 40
\\
30 s'
N 10 s
N h.o N b
,0 20 4
40 t
80 10 0 0
0 20 0
2'O 4'0 s'o do CONTMOL ROD 94SERTCt. M14CEKT m L s A cets riount CONTROL ROO INSERTKW L20ffS TECHPSCAL SPECtFK:ATEN 3-6 AmendmentNoff.118 l
3.10 CONTROL ROD AND POUER DISTRIBUTION LIMITS (Continued)
Basis (Continued)
/
margin requirements for 4-pump operation 'is met at all power levels.
The2.5-seconddroptimespecifiedforthecgrol rods is the drop time used in the transient analysia
/
/
The insertion of part-length rods into the core, except for rod exercises or physics tests, is not permitted since it has been demonstrated on other CE plants that design power distribution envelopes can, under some circumstances, be violated by using part-length rods.
Further infomation may justify their use.
Part-length rod insertion is permitted for physics tests, since resulting power distributions are closely monitored under test conditions, Part-length rod insertion for rod exercises (approximately 6 inches) is pemitted since this amount of insertion has an insignificant effect on power distribution.
For a control rod misaligned up to 8 inches from the rerainder of the banks, hot channel factors will be well within design limits.
If a control red is misaligned by more than 8 inches, the maximum reactor power will be reduced so that hot channel factors, shutdown margin and ejected rod worth limits are met.
If in-core detectors are not available to measure power distribution and rod misaligneer.ts
>8 inches exist, then reactor power must not exceed 75% of rated power to insure that het channel conditions are met.
Continued operation with that rod fully inserted will only be pemitted if the hot channel factors, shutdown margin and ejected rod worth limits are satisfied.
In the event a withdrawn control rod cannot be tripped, shutdown margin requirements will be maintained by increasing the boron concentration by an amount equivalent in reactivity to that control rod. The deviations pemitted by Specification 3.10. 7 are required in order that the centrol rod worth values used in the reactor physics calculations, the plant safety analysis, and the Technical Specifications can be verified. These deviations will only be in effect for the time period required for the test being perforced. The testing intervel during which these deviations will be in effect will be kept to a minimum and special operating precautions will be in effect during these deviations in accordance with approved written testing procedures.
3-63 Artendment No. 31.43.51,68. 110
J 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS (Continued)
/
Basis (Continued)
/
Violation of the power dependent insertion limits, when it is necessary to rapidly reduce power to avoid or minimize a situation harmful to plant personnel or equipment. is acceptable due to the brief period of time that such a violation would be expected to exist, and due to the fact that it is unlikely that l
core operating limits such as thermal margin and shutdown margin would be violated as a result of the rapid rod insertion. Core i
thermal margin will actually increase as a result of the rapid rod insertion.
In addition, the required shutdown margin will most likely not be violated as a result of the rapid rod insertion because present power dependent insertion limits result in shutdown margin in excess of that required by the
/
safety analysis.
i References (1) XN-NT-77-18
/
(2) ANF-87-150(NP), Volume 2
/
l (3) ANF-88-108
/
I f
i l
l l
l l
l l
l i
3-64 L
Amendeert No. 68, 118 l
t.-
'. J,.
.y._ I.......R.._E i w _...
r...
.I
.J POWER DISTRIBUTION INSTRUMENTATION 3.11.2 EXCORE POWER DISTPTBUTION MONITORING SYSTEM LIMITING CONDITION FOR CPERATION The excore monitoring system sha.'.1 be operable eiths a.
The target Axial Offsac (AO) and the Excore Monitoring Allowable Power Level (APL) deter.sined within the previous 31 days using the incore detectors, and ths measured A0 not deviated from the target AD by more tha: 0.05 in the prevfous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b.
The A0 measured by the excore detectors calibrated with the A0 measured by the incore detectors.
c.
The quadrant tilt measured by the excore detectors calibrated with the quadrant tilt measured by the ircore detectors.
APPLICABILITY:
(1) Items a., b. and c. above are applicable when the excore etectors are used for monitoring LHR.
(2) Item c. above is applicabic when the excore detectors are used for monitoring quadrant tilt.
(3) Item b., above is applicable for each channel of the TM/LP trip and
/
the Axial Shape Index (ASI) alage.
/
ACTION 1:
With the excore monitoring system inoperable, do not use the system for monitoring LHR.
ACTION 2:
If the measured quadrant tilt has not been calibrated with the incores, do not use the system for monitoring quadrant tilt.
ACTION 3:
/
When the measured A0 uncertainty is greater than specified in Specification
/
4.18.2, the TM/LP trip function and the ASI alars setpoints shall be
/
conservatively adjusted within twelve (12) hours or that channel shall be
/
declared inoperable. The operability requirements for TM/LP and ASI are
/
given in Table 3.17.1 and 3.17.4, respectively.
/
Basis The excore power distribution monitoring system ccasists of Power Range Detector Channels 5 through 8.
The operability of the excore monitoring system ensures that the III assumptions esployed in the PDC-II analysis for determining A0 limits that ensure operation within allowable LHR limits are valid.
3-66a Amendment No 13,50,68, 118
t POWER DISTRIBUTION INSTRUMENTATION 3.11.2 EXCORE POWER DISTRIBUTION MONITORING SYSTEM LIMITING CONDITION FOR OPERATION Basis (Contd)
Surveillance requirements ensure that the instruments are calfbrated to agree with the incore measurements and that the target A0 is based on the current operating conditions.
Updating the Excore Monitoring APL ensures that the core LHR limits are protected within the 20.05 band on AO.
The APL considers LOCA based LER limits, and factors are included
/
to account for changes in radial power shape and LHR limits over the calibration interval.
The APL is determined from the following:
LI!R(Z)
APL = [ LHR(Z) Max x ated Power x V(Z) x 1.02 j
Min Where (1)LHR(2)f3 is the limiting LHR vs Core Height (from Section 3.23.l (2) LHR(Z)g,Neight, is the measured peak th? including uncertainties vs Core (3) V(Z) is the function (shown in Figure 3.11-1),
(4) The factor of 1.02 is an allowance for the effects of upburn,
/
(5) The quantity in brackets is the minimum value for the entire
/
core at any elevation (excluding the top and bottom 10% of core) considering limits for peak rods.
If the quantity in
/
breckets is greater than one, the APL shall be the rated power level.
References
/
(1) XN-NF-80-47 (2) ANF-88-107
/
l r
l 3-66b Anerdment No 68,118 (Next page is 3-66d)
- 9jal.Le.'-.
.1.d..L ' u.'.
..h. i A~*...
- ..a......L a.
L... Y.
.-L,.
~
(
MODERATORTEMPERATURECOEFFICIENTOFREACTIViTY 3.12
^ AEEjicability Applies to the moderator temperature coefficient of reactivity for the core.
Objective To specify a limit for the positive moderator coefficient.
i Specifications The moderator temperature coefficient (MTC) shall be less
/
I
' positive than +0.5 x 10 Ap/*F at < 2% of rated power.
Bases Tha limitations on moderator temperature coefficient (NTC)
/
are provided to ensure that the assumptions used in the safety
/
I analysis (1) remain valid.
Reference (1) ANT-87-150(P), Volume 2, Section 15.0.5
/
3 3-67 Amendment No 118 (Next page is 3-69)
n.
{
.y.; *....
%'c )g w s.._.'.s;.w L.-
' ~ - '-
t 4
3.17 INSTRUMENTATION AND CONTROL SYSTEMS (Contd)
If the bypass is not effected, the out-of-service channel (Power Removed) assumes a tripped condition (except high rate-of-change power, variable high power and high pressuriser pressure),II} which results in a one-out-of-threa channel logic.
If, in the 2 of 4 logic system of either the reacter protective' system or the engineered safeguards s7stes, one channel is bypassed and a second channel manually placed in a tripped condition, the resulting logic is 1 of 2.
At rated power, the minimum operable variable high power
/
level channels is 3 in order to prcvide adequate flux tilt detection.
If only 2 channels are operable, tha reactor power level is reduced to 70% rated power which protects th.e reactor from possibly exceeding design peaking factors due to undetected flux tilts.
/
The engineered safeguards system provides ; ; out of 4 logic on the signal used to actuate the equipment connected to each of the 2 emergency diesel generator units.
e Two starteup channels are available any time reactivity changes are deliberately being introduced into the reactor and the neutron power is not visible on the log-range nuclear instrumentation or above 10-'1 of rated power. This ensures that redundant start-up instrumentation is available to operators to monitor effects of reactivity changes when neutron power levels are only visible on the start-up channels. In the event only one start-up range channel is available and the neutron power level is sufficiently high that it is being monitored by both channels of log-range instrumentation, a-startup can be perforned in accordance with footnote (d) of Table 3.17.4.
The Zero Power Mode Bypass can be used to bypass the low flow,
/
steam generator low pressure, and TM/LP trips ( ) for all four
/
Reactor Protective system channels to perform control rod testing
/
or to perform low power physics testing below normal operating
/
temperatures. The requirement to maintain cold shutdown boron
/
concentration when in the bypass condition provides additional
/
assurance that an accidental criticality will not occur. To allow
/
low power physics testing at reduced temporature and pressure, the
/
requirement for cold shutdown boron concentration is not required
//
and the allowed power is increased to 10-1 g,
References (1) Updated FSAR, Section 7.2.7.
/
(2) Updated FSAR, Section 7.2.5.2
/
3 s
3-77 Amendment No 118
- 1. '.y: 2; 3 Mti :.
1&.1a.i'.:.:-u..:.%.. :... -..:. -. i.."......T..
A Teble 3.17.1 Instrumentation Operatina Requirements for Reactor Protective System Minieus Minimus Fermissible 1
Operable Degree of Bypass No Functional Unit Channels Redundancy Conditions 7
Manual (Trip 1
None None Buttons) 2 Variable Eigh 2(b d) g(d) f Powe.' Level
/
% ') or I
3 Los Range 2
1 Below 10 Channels Aboveg2 Rated Power Except as Noted in (c)
Xf*of 2 *'I) 1 Below 10 I
/
4 Thermal Margin /
Low-Pressuriser Rated Power,
and
/
]
Fressure greater than cold
/
shutdown boron con-
/
centration.
/
O}
5 High-Fressuriser 2
1 None Pressure I) 6 Low Flow Loop 2
1 Below 10' %
of Rated Powet' and
/
greater than cold
/
5 shutdown boron con-
/
l centration.
/
7 Loss of Load 1
8 Low Steam Gen-2/Sga 1/ Steam None erator Water Gen Generator Level 9
Low Steam Cen-2/S g 1/ Steam Below 10 %
of erator Pressure Con Generator Rated Power and
/
greater than cold
/
I shutdown boron con-
/
centration.
/
I 10 High Containment 2
1 None O)
Fressure r
I t
(a) Bypass automatically removed.
[
(b) One of the inoperable chaneels must be in the tripped condition.
i (c) Two channels required if TH/LP, low steam generator or low-flow channels are bypassed.
load shall be reduced to (d)Ifonlytwochannelsareoperable,'%maybeincreasedto10'{0%orlessofratedpower.
(e) For low power phye Ms testing, 10'
%'and cold shutdr.,wn
/
boron concentration is not required.
/
(f) A0 operability requs.rements are given in Specification 3.11.2.
/
l 3-78 Amendment No 118 l
l i
i l
l
)
3 Trbia 3.17.4 (C:nt'd)
Minimus Minimum Permissible Operable Degree of Typass No Functional Unit Channels Redundancy Conditions 8.
Pressuriser Wide 2 (1, p, q)
None Not required in Range Water Level Cold or Refueling Indiustion Shutdown 9.
Pressuriser Code 1 per None Not Required Safety Relief Valves Valve below 325'F Position Indication (Acoustic Monitor or Temperature Indication) 10.
Power Operated Relief 1 per None Not required when Valves (Acoustic Valve PORV isolation valve Monitor or Temperature is closed and ita Indication) indication system is operable 11.
PORY Isolation Valves 1 per None Not required when Position Indication Valve reactor is depressurized and vented through a vent 21.3 sq.in.
12.
Subcooling Margin 1
None Not required Monitor below 515'F 13.
Auxiliary Feed Flow 1 per flow (h)
None Not required Rate Indication Control below 325'F Valve 14.
Auxiliary Feedwater 2persteg) 1 Not required Actuation System generator below 325'F Sensor Channels fb 15.
Auxiliary Fea:! water 2
1 Not required
.tetuatter. System below 325'F Actuation Channels IU 16.
Excore Detector I
None Not Required Below
/
Deviation Alarms 25% of Rated Power
/
II) 17.
Axial Shape Index 2
1 Not Required Below
/
Alarn 25% of Rated Power
/
(e)
Auxiliary Feedwater System Actuation System Sensor Channels contain pump auto initiation circuitry.
If two sensor channels for one steam generator are inoperable, one uf the steam generator low level bistable modules in one of the inoperable channels must be in the tripped condition.
3-81a Amendment No 67, 68, 96,Ill,118
~
3 Teblo 3.17.4 (Cont'd)
(f) With one Auxiliary Feedwater Actuation System Actuation Channel inoperable, in lieu of the requirement of 3.17.2, provide a second licensed operator in the control roon within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. With both inoperable, in lieu of following the requirements of 3.17.2, start and maintain in operation the turbine driven auxiliary feed pump.
(g) Calculate the Quadrant Power Tilt using the excore readings at
/
least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> when the excore detectors deviation alarms
/
are inoperable, or at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> using symmetric incore
/
detectors when the difference between the excore and the incore
/
measured Quadrant Power Tilt exceeds 2%.
/
(h) With two flow rate indicators inoperable for a ;iven control valve, the control valve shall be considered' inoperable and the requirements of 3.5.2(e) apply.
(1) AO operability requirements are given in Specification 3.11.2.
/
(j, k)
Blank j
(1) The provisions of Specification 3.0.4 are not applicable.
(m, n, o)
Blank (p) With one OPERA 3LE Pressurizer Wide Range Water Level Channel in lieu of the requirement of 3.17.2, restore the inoperable channel to OPERABLE status within 7 days or be in at least HOT SEUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
(q) With no CPERABLE Pressurizer Wide Range Water Level Channels in lieu of the requirements of 3.17.2, either restore at least one of the inoperable channels to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, t
i (next page is 3-82) 3-81b Arendment No 96, 98, JI),118 I
/.
o 3.23 POWER DISTRIBUTION LIMITS 3.23.1 LINEAR HEAT RATE (LRR)
LIMITING CONDITION FOR OPERATION The LER in the peak power fuel rod at the peak power elevation Z shall not exceed the value in Table 3.23-1 times F (Z) [the
/
g function F (Z) is sh wn in Figure 3.23-1).
/
A
/
/
APPLICABILITY: Power operation above 50% of rated power.
ACTION 1:
When using the inenre alare system to monitor LHR, and with four or more coincident incore alarus, initiate within 15 minutes corrective action to reduce the LHR to within the limits and restore the incore readings to less than the alarm setpoisats within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or failing this, be at less than 50% of rated power s.6,41n the following 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
ACTION 2 i
When using the excore monitoring system to monitor LHR and with the A0 deviating from the target A0 by more than 0.05, discontinue using the excore monitoring system for monitoring LHR. If the incore alarm system is 1soperable, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> be at 85% (or less) of rated thermal power and follow the procedure in ACTION 3 below.
(
I I
3-103 Amendment No. 68,118
~
t POWER DISTRIBUTION LIMITS I
3.23.1 LINEAR HEAT _ RATE (LHR)
LIMITING CONDITION FOR OPERATION ACTION 3:
If the incore alara system is inoperable and the excore monitoring system is not being used to monitor LHR, operation
/
at less than or equal to 85% of rated power may continue provided that incore readings are recorded manually. Readings shall be taken on a minimum of 10 individual detectors per quadrant (to include 50% of the total number of detectors'in a 10-hour period) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and at least every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> i
thereafter.
If readings indicate a local power level equal to or greater than the alarm setpoints, the action specified in ACTION 1 above shall be taken.
I Basis e
i The limitation of LHR ensures that, in the event of a LOCA, the peak temperature of the cladding will not exceed 2200'F.(
/
i
/
/
1 Either of the two core power distribution monitoring systems (the incore alarm system or the excore monitoring system) provides adequate monitoring of the core power distribution and is capable of verifying that the LHR does not exceed its limits. The incore alarm system performs this function by continuously monitoring the local power at many points throughout the core and comparing the 3
measurements to predetermined setpoints above which the limit on l
LHR could be exceeded. The excore monitoring system performs this function by providing comparison of the measured core A0 with predetermined A0 limits based on incore measurements. An Excore i
Monitoring A11ovable Power Level (APL) which may be less than rated power, is applied when using the excore monitoring system to ensure that the A0 limits adequately restrict the LHR to less than the limiting values.( )
I If the incore alarm system and the excore monitoring system are I
both inoperable, power will be reduced to provide earsin between i
the actual peak LHR and the LHR limits and tne incore readings will be manually collected at the tirminal blocks in the control room utilizing a suitable signal detector.
If this is not feasible with the manpower available the reactor power will be i
reduced to a point below which it is improbable that the LHR limits could be exceeded.
i 3-104 Amendment No. 68. 82. 118 q
_ _. -. _.. _. _ _,~
v 1
POWER DISTRIBUTION LIMITS 3.23.1 LINEAR HEAT RATE (LRR) 1 LIMITING CONDITION FOR OPERATION 5
Basis (Contd)
The time interval of.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and the minimum of 10 detectors per j
quadrant are sufficient to maintain adequate surveillance of the i
core power distribution to detect significant changes until the conitoring systems are returnad to service.
To ensure that the design margin of safety is maintained, the
[
determination of both the incore alarm setpoints and the APL takes into account a measurement uncertainty factor of 1.10, an j
engineering uncertainty factor of 1.03, a thermal power t
measurement uncertainty factor of 1.02 and allowance for quadrant tilt.
References (1) ANFa88-107
/
i (2)
(Deleted)
/
[
(3)
(Deleted)
/
(4) XN-NT-80-47
/
i I
{
1 i
f I
i 1
I i
I I
l 3-105 Amendment No. ff, 118 I
(Next page is 3-107)
{
l
.w w n- ~ w o--w-w w w r ww,cm e m, __ _,
n,,, _
= - -
.}
TABLE 3.23-1 LINEAR REAT RATE LIMITS No. of Fuel Rods in Assembly
/
208 216 Feak Rod 15.28 kW/ft 15.28 kW/ft
/
i l
L TABLE 3.23-2 RADIAL PEAKING FACTOR LIMITS.F Feaking Tactor No. of Fuel Rods in Assembly i
j 208 216 Assembly 1.48 1.50
/
Interior Rod F,I A
1.70 1.73
/
3-107 Amendment No. 68. D 8
o f
I l
g u x
\\
a M
i D
I M
~ 11-UNACCEPTABLE l
I X
OPERATION i
l 2
l 1
A i
O am Z
o I
6O ACCEPTABLE 2
I i
sa..
OPERATION 88-E BREAK POINTS M
l 3
- 1. 0.6, 1.0 c
] e4_
- 2. 1.0. 93 i
A l
C.
>o L
a e
a i
< e.7 l
l om o.2 a.4 o.s os 1.o
=
l FRACTION OF ACTIVE FUEL HEIGHT M
FIGURE 3.23-1 ALLOMAstE LHR AS A FUIICTIO11 Of PEAK PCMER LOCATION
~
I
g
. e 1
FIGURE 3.23-2 (Deleted) 3-109 AmenJuant No. 68.118
7 i
IIGURE 3.23-3 (Deleted) i 3-!!0 Amendment No. 68,118
0 POWER DISTRIBUTION LIMITS 3.23.2 RADIAL PEAKING FACTORS LIMITING CONDITION FOR OPERATION A
A The radial peaking factorg F, and F, shall be less than or
/
equal to the value in Table 3.23-2 times the following quantity.
/
The quantity is [1.0 + 0.3 (1 - P)] for P 1 5 and the quantitiy
/
is 1.15 for P <
.5.
P is the core thermal power in fraction of
/
rated power.
APPLICABILITY: Power operation above 25% of rated power.
/
ACTION:
l '..For P < 50% of rated with any radial peaking factor
/
axceeding its limit, be in at least hot shutdown within
/
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
/
2.
For P 1 50% of rated with any radial peaking factor
/
exceeding its limit, reduce thermal power within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
/
to less than the lowest value oft
/
[1 - 3.33 (Fr - 1) ) x Rated Power
/
Ft isthemeasuredvalueofeitherF^,orFfandF
/
Where F g
r is the corresponding limit from Table 3.23-2.
Basis A
A The Itaitations on F, and F are pr vided to ensure that
/
r assumptions used in the analysis for establishing DN3 margin.
LHR and the thermal margin / low-pressura.and variable high-power
/
trip set points remain valid during operation. Data from the incore detectors are used for determining the measured radial peaking factors. The periodic surveillance requirements for determining the measured radial peaking factors provide assurance that they remain within prescribed limits. Determining the measured radial peaking factors after each fuel loading prior to exceeding 50% of rated power provides additional assurance that the core is properly loaded.
i l
l r
1 l
3-111 Amendment No if,III I
I-
p C h y;
- ffr L.u:.: 31..
'.....;. x
..l._'.....'.,~.'....
}
- r P0b7.R DISTRIBUTION LIMITS l
3.23.3 QUADRANT POWER TILT - T_
4 i
LIMITING CONDITION FOR OPERATION The quadrant power tilt (T ) shall not exceed 5%.
i g
APPLICABILITY: Power operstion above 25% of rated power.
/
ACTION:
1.
With quadrant power tilt determined to exceed 51 but less than or equal
/
to 10%.
/
j a.
Correct the quadrant power tilt within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after
/
exceeding the limit, or
/
b.
Deterwir.e within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and, at least once every 8
/
hours thereaf ter, that the radial peaking factors are within
/
the limits of Section 3.23.2, or
/
I c.
Reduce power, at the normal shutdown rate, to less than 85%
/
of rated power and determine that the radial peaking factors
/
are within the limits of Section 3.23.2.
At reduced power.
/
determine at least once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> that the radial
/
peaking factors are within the limits of Section 3.23.2.
/
s 2.
With quadrant power tilt determined to exceed 10%:
/
a.
Correct the quadrant power tilt within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after
/
exceeding the limit, or
/
b.
Reduce power to less than 50% of rated power within the next
/
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and determine that the radial peaking factors are
/
within the limits of Section 3.23.2.
At reduced power,
/
determine at least once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> that the radial peaking
/
factors are within the limits of Section 3.23.2.
/
i 3.
With the quadrant power tilt determined to exceed 151, be in at least hot standby within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Basis Limitations on quadrant power tilt are provided to ensure that design safety sargins are maintained. Quadrant power tilt is determined from encore detector readings which are calibrated ueing incore detector measurements.II) Calibration factors are determined from incore seasurements by performing a two-dimensional, full-core surface fit of deviations between measured and theoretical incore readings and integrating the fitting function over each core quadrant. Values of l
LHR and radial peaking factors are increased by the value of quadrant tilt.
3-112 Amendment No f 8.118
\\
f b.
Th3 PCS v:nt(s) sh:11 bo v rified to ba epea ct lecst on:o p:r 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> then the vent (s) is being used for overpressure protection except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days.
c.
When both open PORV pilot valves are used as an alternative to venting the PCS, then verify both PORV pilot valves and both FORV block valves are open at least once per 7 days.
Basis Failures such as blown instrument fuses, defective indicators, and faulted amplifiers which result in "upscale" or "downscale" indication can be easily recognized by simple observation of the functioning of an instrument or system.
Furthermore, such failures are, in many cases, revealed by alarm or annunciator action and a check supplements this type of built-in surveillance.
Based on experience in operation of both conventional and nuclear plant systems when the plant is in operacion, a checking frequency of once-per-shift is deemed adequate for reactor and steam system instrumentation. Calibrations are performed to insure the presentation and acquisition of accurate information.
The power range safety channels and AT power channels are
/
calibrated daily against a heat balance standard to account for errors induced by changing rod patterns and core physics parameters.
Other channels are subject only to the "drift" errors induced within the instrumentation itself and, consequently, can tolerate longer intervals between calibration. Process system instrumentation errors induced by drift can be expected to remain within acceptable tolerances if recalibration is performed at nach refueling shutdown interval.
Substantial calibration shifts within a channel (essentially a channel failure) vill be revealed during routine checking and testing procedures. Thus, minimum calibration frequencies of one-per-day for the power range safety channels, and once each refueling shutdown for the process system channels, are considered adequate.
The minimum testing frequency for those instrument channels connected to the reactor protective system is based on an estimated average unsafo failure rate of 1.14 x 10-5 failure / hour per channel.
This estimation is based on limitad operating experience at conventional and nuclear plants. An "unsafe failure" is defined as one which negates channel operability and which, due to its nature, is revealed only when the channel is tested or attempts to respond to a bona fide signal.
4-2 Amendme.,e No 15 J!,!!1, 118
2.
l '.
O.
4,
.=
TAB 12 4.1.1 Mialaus Frepies for Ncke, Calibrations and Testing of Beactor Protective Syntan(5)
S :.
F. '
Surveillance N._
Diemmel Description hinc tion Frequency Survettlance Method g
'l 1.
Feuer Rasse Safety N amelo e.
Nck S
a.
Comparisen of four pener eka==al readings.
$1 b.
Oneck(3)
D b.
Osammel adjustment to agree with best beleece h
calculation. Repeat idienever flus-AT pouer h
cooperator alerne.
c.
Test M(2) c.
Internal test signal.
/
f+,
d.
Calibrate (6)
R d.
Chemmel elignment through measurement /edjustment
/
f:
of internal test points.
y'.
(
2 Wide-Range 14seritimic a.
N ck 5
a.
famparteen of both wide-range readings.
g
{'
Meutres Moottore b.
Test P
b.
Internal test elemel.
l 6
3.
Reactor Coolant Flow s.
Dieck 5
a.
ra-partoon of four esperate total flew indications.
b.
Celibrate R
b.
r - differential preeeure applied to sensors.
/
c.
Test M(2) c.
51 stable trip teeter.(1)(4)
[
4.
Thermal Mergim/ low s.
Dieck:
S e.
02ecks l*
Pressuriser Pressure (1) Temperature (1) Camperieen of four seperate calculated Input trip pressure set pelet indicatione.
(2) Fressure (2) Comperleen of four pressuriser pressure Input Radicatiese. Dame se 5(a) belem.)
g',
b.
Calibrate R
b.
Calibrate (1) Temperature (1) Enous resistance embetituted for RID cofact-l Input deot with knous pressure and youer 1*put.
/
(2) Pressure (2) Part of 5(b) beten.
g Input c.
Test M(2) c.
B! stable trip teater.(1)
/
[
5.
us e-tre..uriser cre..ure
.. O,c k s
.. Co-p.ri of four e.p.r.t. pre.eure n.dic.tions.
b.
Calibrate R
b.
Kmaun preneure applied to someos s.
/
c.
Test M(2) c.
Sistable trip tester.(1) 30,66, 118 4-3 m t Me i'.
e W
....,j.'..;5L.'.lLi.n.iI'i:..:. 5 nila. :~....".
....;...a...,.
. a.... :...
=ea
%N F"9 5
4 3
l1l I
I 2
l1 I
I 3
3
]
3 3 3 3
t 3
3 s
a 3l-I.
g 3
s-3 y
7 1
3
[
=
e a
su g
i i
]
I t
=
I t1 3
1 J g
I3 I g
II Ie 3
j 1:
3 8
2 3
- d g
1 y&
E.
t 1
S k
1 l
t Ad*%
5 m
I 2
gi 3
1 2
3
- d Ed iig$
I g 3j -[ j j j
i 8
i i j
}jE j
i
,,in f
3 ji 442d i
.dk 2
3
}
5 a
iiJ 4
i 4 4
i i
4 a
i dis f
t 3
3 a
a m I
=
=Y Y
m 6
'I
=
man i
m i
1 y
2 II 0
1 1
1 0
2 2
V e 2
=
e Ez
=
= *
=
x 3
v:
J J
d 4 0
0 A d
4 A
A A
A
$04
(
t a;a
- J t
i a
a I
I 3
=
l 3
u
.I.
.I I
E t
t E
l*
a 1
1 J
J
]
]
g
]3 1
A I
l I
I "3
1 3
5
.I I
3
!li a
a e
s i
a i
t
i I
1
\\.
.f
TA.StE 4.1.1 seistaman Fregissectee for Onecke, Cellbrettene and Testles of Seector Protective System (5) (Centd) 4.{
Surve1ilonce h l Seectlpties Ptsmetten Fregnency
$nsrwelltence ter*thed
- 14. hemel Iterste Calculater a.
Oneck Q
e.
Vertfy cemetente.
/
e.'
4 0:.
setY3:
(1)h bietehle trip teater 1ajecte e signal late the blotehle and providee a precialen reedset of the trip set point.
h.
ls (2)All anschly tests will be dame se only one of four channele et a time to present reacter trip.
}.
1:.
(3)Adjeet the emcleer romer er AT penser eentil reedout agrees with heet belance calculottene tenen above 1% of rated
/
ej
/
- t -
l4 pomer.
(4)Trty ee. ting for operettag pusy cashinatice only. Settings for other than operettes peep camhinettene meet he
[:
teeted during reestlee asethly teetles performed ideen shut dunen and with!s four haaste after ressentag aperation with p
a dif f arent pusy combinettee if the setting for that combinetten has met been tested within the preolese meeth.
[,
i.
(5)lt le set neceeeery to perform the specified test tag diaring prolonged periode le the refuelles shastesun canditten.
If Ef.le==e, estated teettag will be performed prior to seteratas the plant to service.
Y (6) Alee facindes teettag wortehle hip pauer fiancates in the humal Iserste Calculater.
/
IS i
Fweecy unArum r
I Itatatien Fe evnenc y
.b 5
At least once per 12 heure.
.t 1.e.t r,b,
r..
W At leset sace per F doye.
91 At least once per 31 doye.
Q At leset osace per 92 days.
SA At least esce per 6 meurthe.
3 At least once per 18 munths.
F Prior to each etert-up if met done previesse eseek.
NA esos applicable.
4-S a.e m.t me w,118 i
S
, ~,
g TABl.C 4.l.3 Minimum Frequencies for Checks. Calibrations and Testing of Miscellaneous Instrumentation and Controla I
i Survelliance Clannel Description Function Frequency Surveillance Method e
1.
Start-Up Rarge a.
Check S
a.
Comparison of both channel count rate Neutron Monitors indicativns when in service.
b.
Test P
b.
Internal test signals.
2.
Primary Rod a.
Check 3
a.
Comparison of output data with secondary Position RPIS.
Indication b.
Check it b.
Check of power dependent insertion limits System monitoring system.
c.
Calibrate R
c.
Physically measured rod drive position
/
used to verify system accuracy. Check rod position interlocks.
3.
Secondary Rod a.
Check 5
m.
Comparison of output data with primary RPIS.
j Position Indication b.
Check b.
Same as 2(b) above.
System c.
Calibrate H
c.
Same as 2(c) above, including out-of-
/
j sequence alarm function.
4.
Area Monitors a.
Check D
a.
Normal readings observed and internal Note: Process test signals used to verify instrument i
Monitor Surveil-opera tion.
l lance Requirements b.
Calibrate R
b.
Exposure to known external radiation i
are located in source.
l Tables 4.24-1 ano c.
Test H
c.
Detector exposed to remote operated 4.24-2 radiation check mource or integral electronic check source.
5.
Emergency Plan Radia-a.
Calibrate A
a.
Exposure to known radiation source.
tion Instruments b.
Test H
b.
Battery check.
6.
Environmental a.
Check H
a.
Operational check.
Monitors b.
Calibrate A
b.
Verify airflow indicator.
1 7.
Pressuriser Level a.
Check 5
a.
Comparloon of two wide and two narrow Instruments range independent level readings.
b.
Calibrate R
b.
Known differential pressure applied to i
sensor.
c.
Test H
c.
Signal to meter relay adjusted with teat device.
4-10 Amendment No. jd, jf. 38. $$. JJ),]}g l
l e
._,.-,-r--
h i'
p Y
TABLE 4.1.3 Minimum Frequencies for Checks, Calibrations and Testing of Miscellaneous Instrumentation and Controls (Continued) i-1I Serve 111ance Channel Description Function Frequency
- Surveillance Method N
- 8. Control Rod Drive System a.
Test R
a.
Verify proper operation of all manual
/
Interlocks red drive control system interlocks, p,
using simulated signals where necessary.
l b.
Test F
b.
Same as 8(a) above, if not done within three months, p*
- 9. Flux-AT Fower Comparator a.
Calibrate R
a.
Use simulated signals.
/
[h b.
Test M
b.
Use simulated signals.
/
g.g
- 10. Calorimetric Instrumentation a.
Calibrate R
a.
Known differential pressure applied to
/
feedwater flow sensors.
8 I.
II. Containment Building a.
Test R
a.
Expose sensor to high humidity p
Humidity Detectors atmosphere.
l.
i,
- 12. Interlocks - Isolation Valves a.
Calibrate R
a.
Known pressure applied to sensor.
on shutdown Cooling Line l
- 13. Service Water Break Detector a.
Test R
a.
Known differential pressure applied to
[-
in Containment Sensors.
.o
.r
/
I I
5
/
6 9
4-11 Amendment No Jg, jf, Jg, gg, pf,118 4
w.
u.
... -....-4.G." L%: J M.T. a.%.. t' !s.'.:.. %.. ':...
s 4.15 primary $7staa Tiov Measurement Applicability Applies to the measurement of primary system flow rate with four primary coolant pumps in operation.
Objective To provide assurance that the primary system flow rate is equal to or above the flow rate required in 3.1.1.c.
Specification Af ter each refueling outage, or af ter plugging 10 or more stsaa generator tubes, a primary system flow measuressat shall be made with four primary coolant pumps ia operation. This seasurement
/
shall be made within the first 31 days of rated powsr operation.
/
tasis This surveillance program assures that the reactor coolcat flow is consistent with that assumed as the basis for Specificatssa 3.1.lc.
i l
l 4-70 Amendment No 23 118 i
l l
I l
.6
.. u.
n.
f POWER DISTRIBtTTION INSTRIMDf7ATION
~
4.18.2 EXCORE MONITORING SYSTDi SURVgII. LANCE REQUIREMENTS 4.18.2.1 At least every 31 days of power operations a.
A target A0 and escore monitoring allowable power level shall be determined using encore and incore detector readings at steady state near equilibrium conditions.
b.
Individual encore chamael measured A0 shall be compared to the
/
total core Ao measured by the incores. If the difference is
/
greater than 0.02, the encore moaltoring system shall be recalibrated.
c.
The escore seasured Quadraat Power Tilt shall be compared to the incore measured Quadrant Power Tilt. If the difference is l
greater than 2%, the encore monitoring systes shall be recalibrated.
t i
l t
Amendment No $8, 118 l
. A..
..~
.. I '
...s..
l 4.19 POWUt DISTRIBUTION LIMITS 4.19.1 LIN0Ut HEAT RATES SURVEILIANCE REQUIREMDits 4.19.1.1 When using the incore alare system to monitor IJOL, prior to operation above 50% of rated power and every 7 days of power operation thereafter, incore alarna shall be set based on a measured power distribution.
4.19.1.2 Vt.en using the encore monitoring system to monitor IJDL:
a.
Prior to use, verify that the measured A0 has not deviated from the target A0 by more than 0.05 in the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for
/
each operable channel using the previous 24 hourly recorded
/
values.
/
b.
Once per day, verify that the measured Quadrant Power Tilt is less than or equal to 3%.
c.
Once per hour, verify that the power is less than or equal to the APL and not more than 10% of rated power greater than the power level used in determining the APL.
d.
Continuously verify that the seasured Ao is within 0.05 of the
/
established target A0 for at least 3 of the 4, 2 of the 3 or
/
2 of the 2 operable channels, whichever is the applicable case.
/
4 4
4-33 Amendment No gg,g
/ 4.19 POWER DISTRIBUTION LIMITS 4.19.2 RADIAL PEAKING FACTORS SURVEILI.ANCE REQUIRDENTS 4.19.2.1 The measured radial peaking factors (F. and TAR) j A obtained by using the incore detection #systes.#shall be determined to be less than or equal to the values stated in the LCO at the following intervals: a After each fuel loading prior to operation above 50% of rated power and b. At least once per week of power operation. F 4-84 Amendment No ff. 118
. - - -.,s. . -,... ~ ..~.;...---.. ? l 4.20 N000ATOR TDIPDATURE COEFTICIENT OffC) / i StlRVEII.! MCI MQUIRENDTS / 4.20.1 The NTC shall be determined to be within its limits by / ceafirestory measureneste prior to initial operation / above 21 oa' rated thermal power, af ter each refuelias." / 1 i i I t e i-f i (seat page is 4-90) 1 1 1 4 85 Amendment No 85. 118 l ,}}