ML20206J225
| ML20206J225 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 11/15/1988 |
| From: | NORTHEAST NUCLEAR ENERGY CO. |
| To: | |
| Shared Package | |
| ML20206J223 | List: |
| References | |
| NUDOCS 8811280103 | |
| Download: ML20206J225 (38) | |
Text
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lifEEX DEFINITIONS E8.GE SECTION Enclosure Building Integrity................................
1-5 Reactor Trip System Response Time...........................
1-5 Engineered Safety Feature Response Time.....................
15 Physic.s Tests...............................................
16 Unrodded Integrated Radial Peaking Factor - F..............
1-6 r
Souwce Che'ck................................................
1-6 Radiological Effluent Monitoring and Offsite Dose Calcul ation Manual (REM 0DCM)..............................
1-6 Radioactive Waste Treatment Systems.........................
1-6 Purge -
Piping..............................................
16 Venting.............................
1-8 Membe r ( s ) o f t h e Publ i c.....................................
1-8 Site Bounoary............
18 l
Unrestricted Area...........................................
1-8 Storage Pattern.............................................
1-8 8811280103 881115 PDR ADOCK 05000336 P
PDC MILLSTONE - UNIT 2 11 Amendment No. 79,4, JJJ, JJJ
INDEX LIMITING CONDITIONS FOI. OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 LINEAR HEAT RATE.......................................
3/4 2-1 3/4.2.2 Deleted T
3.4.2.3 TOTAL INTEGRATED RADIAL PEAKING FACTOR - F............
3/4 2 9 7
3/4.2.4 AZIMUTHAL POWER TILT...................................
3/4 2-10 3/4.2.5 Deleted 3/4.2.6 DNB MARGIN.............................................
3/4 2 13 3/ M INSTRUMENTATIQN 3/4.3.1 REACTOR PROTECTIVE INSTRUdENTATION....................
3/4 3-1 3.4.3.2 EMINEERED SAFETY FEATURE AC10AI!ON SYSTEM INSTRUMENTATION......................................
3/4 3 10 3/4.3.3 MONITORING INSTRUMENTATION.............................
3/4 3-26 Radiation Monitoring...................................
3/4 3-26 Incore Detectors.......................................
3/4 3-30 Seismic Instrumentation................................
3/4 3-32 Meteorological Instrumentation.........................
3/4 3-36 Chlorine Detection Systems.............................
3/4 3-42 Fire Detection Instrumentation.........................
3/4 3 43 Accident Monitoring....................................
3/4 3 46 Radioactive Liquid Efficent Monitoring Instrumentation.
3/4 3-50 Radioactive Gaseous Effluent Monitoring Instrumentation 3/4 3-56 3.4.4 REACTOR COOLANT SYSTEM 3/4.4.1 COOLANT LOOPS AND (%LANT CIRCULATION..................
3/4 4-1 Startup and Power (peration...........................
3/4 4-1 Hot Standby............................................
3/4 4-la Shutdown...............................................
3/4 4-lb MILLSTONE - UNIT 2 V
Amendment No. 75, 38, $$,
pp. 99, 19.8
INDEX BASES SECTION PJ_q[
3.4.0 APPLICABillTY.........................................
B 3/4 0-1 3.4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTR0L..................................
B 3/4 1 1 3.4.1.2 BORAT10N SYSTEMS..................................
B 3/4 1-2 3/4.1.3 MOVABLE C0NTROL ASSEMBLIES........................
B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 LINEAR HEAT RATE..................................
2 3/4 2-1 3/4.2.2 Deleted T
3/4.2.3 TOTAL INTEGRATED RADIAL PEAKING FACTOR - F.......
B 3/4 2-1 7
3/4.2.4 AZIMUTHAL POWER TILT..............................
B 3/4 2-1 3/4.2.S Deleted 3/4.2.0 DNB HARGIN......................................
B 3/4 2-2 144 3 INSTRUMENTATION 1
s 1
3/4.1.1 FROTECTIVE INSTRUMENTATION........................
I 3/4 3-1 3/4 3.2 ENGINEERED SAFETY FEATURE INSTRUMENTATION.........
I 3/4 3-1 3/4.3.3 MONiiORING INSTRUMENTATION........................
I 3/4 3-2 t
)
MILLSTONE - UNIT 2 XI Amendment No. J), /9, Jp/
l' DEFINITIONS AXIAL SHAPE INDEX 1.23 The AXIAL SHAPE INDEX (Y used for normal control and indication is the nower level detected by thb) lower excore nuclear instrument detectors (L) less the power level detected by the upper excore nuclear instrument detectors (U) divided by the sum of these power levels. The AXIAL SHAPE INDEX (Y )t used for the trip and pretrip signals in the reactor protection system is the above determin5)thetruecoreaxialpowerdistributionforthatchannel. modified by an appropriate value (Y E*L YI - AYE+0 Y
L+V 1.24 Deleted.
l ENCLOSURE BUILDING INTEGRITY 1.25 ENCLOSURE BUILDING INTEGRITY shall exist when:
1.25.1 Each door in each access opening u clesed except when the access opening is being used for normal transit entry and exit, and l
1.25.2 The enclosure building filtration system is OPEPABLE.
REACTOR TRIP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval J
from when the monitored parameter exceeds its tri) setpoint at the channel sensor until electrical power is interrupted to tie CEA drive mechani:m.
ENGINEERING SAFETY FEATURE RESPONSE TIME 1.27 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of e
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4 MILLSTONE - UNIT 2 1-5 Amer.dment No. JS Y
i5 TABLE 2.2-1 I~
REACTOR PROTECTIVE INSTRUMENTATION TRIP xTPOINT LIMITS N
FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 1.
Manual Reactor Trip Not Applicable Not Applicable
" 2.
Power Level-High Four Reactor Coolant Pumps s 9.6% above THERMAL POWER, s 9.7% Above THERMAL POWER, Operating with a minimum setpoint of with a minimum of $ 14.7%
1 ;4.6% of RATED THERMAL of RATED THERMAL POWER, and POWER.
i maximum of s 106.7% of RATED THERMAL POWER.
3.
Reactor Coolant Flow -
'?
Low (1) 2 91.7% of reactor coolant 2 90.1% of reactor coolant flow with 4 pumps operating *.
with 4 pumps operating.
4.
Reactor Coolant Pump 1 830 rpm 2 823 rpm Speed - Low y 5.
Pressurizer Pressure - High
$ 2400 psia 5 2408 psia E
i 6.
Containment Pressure - High
$ 4.75 psig 5 5.24 psig 3"
7.
Steam Generator Pressure -
2 680 psia 1 672 psia p
Low (2)
(5)
- 8.
Steam Generator Water 2 36.0% Water Level - each 2 35.2% Water Level - each Level - Low (5) steam generator steam generator
% 9.
Local Power Density -
Trip setpoint adjasted to not Trip setpoir, adjusted to O
High (3) exceed the limit lines of not exceed the limit lines Figures 2.2-1 and 2.2-2 (4).
of Figures 2.2-1 and
_2 2.2-2 (4).
p
- Design Reactor Coolant flow with 4 pump operating is 340,000 gpm.
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TABLE 2.2-1 r.n REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS I
c:
FUNCTIONAL UNIT TRIP SETPOINT All0WABLE VALUES
[
- 10. Thermal Margin / Low Pressure (1)
Four Reactor Coolant Pumps Trip setpoint adjusted to not Trip setpoint adjusted to not j
Operating exceed the limit lines of exceed the limit lines of Figures 2.2-3 and 2.2-4 (4).
Figures 2.2-3 and 2.2-4(4).
11.
Loss of Turbine--Hydraulic 1 500 psig 2 500 psig TABLE NOTATION (1) Trip may be bypassed below 5% of RATED THERMAL P0'JER; bypass shall be automatically removed when THERMAL POWER is + 5% of RATED THERMAL POWER.
o 5.
(2) Trip may be manually bypassed below 780 psia when all CEAs are fully inserted; bypass shall be automatically removed at or above 780 psia.
(3) Trip may be bypassed below 15% of RATED THERMAL P0kER; bypass shall be automatically removed when THERi4AL POWFx IS.>_15% of RATED THERMAL POWER.
(4) Calculations of the trip setpoint includes measurements, calculational and processor uncertainties, and dynamic allowances.
(5) Each of four channels actuate on the auctioneered output of two transmitters, one from each steam generator.
I 2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate at or less than 21 kw/ft. Centerline fuel melting will not occur for this peak linear heat rt.te. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface te:nperaturo is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could i
result in excessive cladding temperatures because of the onset of departure 1
from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have t sen related to DNB through the XNB correlation.
The XNB DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions.
The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, u indicative of the margin to DNB.
The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.17.
This value corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conoitions.
The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and maximum cold leg temperature with four Reactor Coolant Pumps operating for which the minimum DNBR is no less than 1.17.
The limits in Figure 2.1-1 were calculated fo" reactor coolant inlet l
temperatures less than or equal to 580*F.
The dashcd line at 580*F coolant inlet temperatures is not a safety limit; however, operation above 580'F is not possible because of the actuation of the main steam line safety valves which limit the maximum value of reactor inlet temperature. Reactor operation at THrRMAL POWER levels higher than 112% of RATED THERMAL POWER is prohibited l
by the high power level trip setpoint specified in Table 2.2-1.
The area of 1
safe operation is below and to the left of these lines.
MILLSTONE - UNIT 2 B21
SAFETY LIMIT BASHS The conditions for the Thermal Margin Safety Limit curves in figure 2.1-1 to be valid are shown on the figure.
The rer.-tor protective system in combination with the Limiting Conditions for Operation, is designed to prevent any anticipated combination of transient conditions for reactor coolant system temperature, pressure, and thermal power level that would result in a DNBR of less than 1.17 and preclude the existence of flow instabilities.
2.1.2 REACTOR COOLANT SYSTEM pESSURE The restriction of this Safety Limit protects the inte;rity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
The reactor pressure vessel and pressurizer are designed to Section Ill of the ASME Code for Nuclear Power Plant Components which permits a maximum transient pressure of 110% (2750 psia) of design pressure. The Reactor j
Coolant System piping, valves and fittings, are designed to ANLI B31.7, Class I which permits a maximum transient pressure of 110% (2750 psia) of t
component design pressure.
The Satety Limit of 2750 psia is therefore l
consistent with the design criteria and associated code requirements.
The entire Reactor Coolant System is hydrotested at 3125 psia to demon-l strate integrity prior to initial operation.
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4 MILLSTONE - UNIT 2 B 2-3
LIMITING SAFETY SYSTEM SETTINGS BASES Reactor Coolant Flow-low (Continued)
' operation ~of the reactor at reduced power ;f one or two reactor coolant pumps are taken out of service.
The low-flow trip setpoints and Allowable Values for the various reactor coolant pump combinations have been derived in consid-eration of instrument errors and response times of equipment involved to maintain the DNBR above 1.17 under normal operation and expected transients.
For reactor operation with only two or three reactor coolant pum)s operating, the Reactor Coolant Flow Low trip setpoints, the Power Level Hig1 trip setpoints, and the Thermal Margin / Low Pressure trip setpoints are automatical-ly changed when the pump condition selector switch is manually set to the desired two-or three pump position.
Changing these trip setpoints during two and three pump operation prevents the minimum value of DNBR from going below 1.17 during normal operational transients and anticipated transients when only two or three reactor coolant pumps are operating.
Pressurizer pressure Hioh The pressurizer Pressure-High trip, backed up by the pressurizer code safety valves and main steam line safety valves, 3rovides reactor coolant system protection against overpressurization in tie event of loss of load without reactor trip. This trip's setpoint is 100 psi below the nominal lift setting (2500 psia) of the pressurizer code safety valves and its concurrent operation with the power-operated relisf valves avoids the undesirable opera-tion of the pressurizer code safety valves.
Containment Pressure Hioh The Containment Pressure High trip provides assurance that a reactor trip is initiated concurrently with a safety injection.
The setpont for this trip is identical to the safety injection setpoint.
Steam Generator Pressure tow The Steam Generator Pressure Low trip provides protection against an excessive rate of heat extraction from the steam generators and subsequent cooldown of the reactor coolant. Tho setting of 680 psia is sufficiently below the full-load operating point so as not to interfere with normal opera-tion, but still nigh enough to provide the required protection in the event of excessively high steam flow. This setting was used with an uncertainty factor of i 22 psi in the accic'ent analyses.
MILLSTONE - UNIT 2 8 2-5
LIMITING SAFETY SYSTDLSETTINGS
- BASES, c
Steam Generator Water level - Low The Steam Generator Water level-Low Trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity and assures that the design pressure of the reactor coolant system will not be exceeded. The specified setpoint provides allowance that there will be sufficient water inventory in the steam generators at the time of trip to provide a margin of more than 10 minutes before auxiliary feedwater is required.
Local Poggr Density-Hioh The Local Power Density High trip, functioning from AXIAL SH/PE INDEX monitoring, is provided to ene that the peak local power density in the fuel which corresponds to fuel centerline melting will not occur as a conse-quence of axial power ma1 distributions. A reactor trip is initiated whenever the AXIAL SHAPE INDEX exceeds the allowable limits of Figure 2.2-2.
The AXIAL SHAPE INDEX is calculated from the upper and lower ex-core neutron detector l
channels.
The calculated setpoints are generated as a function of THERMAL POWER level.
The trip is automatically bypassed below 15 percent power.
The maximum AZIMUTHAL POWER TILT and maximum CEA misalignment permitted t
for continuous operation are assumed in generation of the setpoints. In addition, CEA group sequencing in accordance with the Specifications 3.1.3.5 and 3.1.3.6 is assumed.
Finally, the maximum insertion of CEA banks which can i
occur during any anticipated operational occurrence prior to a Power Level-High trip is assumed.
Thermal Marain/ Low Pres 3u e i
The Thermal Margin / Low Pressure trip is provided to prevent operation when the DNBR is less than 1.17.
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MILLSTONE - UNIT 2 B 2-6
LIMITING SAFETY SYST'M SETTINGS BASES Thermal Marain/ Low Pressure (Continuedl The trip is initiated whenever the reactor coolant system pressure signal drops below either 1750 psia or a computed value as described below, whichever is higher.
The computed value is a function of the higher of AT power or neutron power, reactor inlet temperature, the number of reactor coolant pumps operating and the AXIAL SHAPE INDEX. The minimum value of reactor coolant flow rate, the maximum AZIMUTHAL F0WER TILT and the maximum CEA deviation permitted for continuous operation are assumed in the generation of this trip function.
In addition, CEA group sequencing in accordance with Specifications 3.1.3.5 and 3.1.3.6 is assumed.
Finally, the maximum insertion of CEA banks which can occur during any anticip6ted operational occurrence prior to a Power Level-High trip is assumed.
Thermal Margin / Low Pressure trip setpoints are derived from the core safety limits through application of appropriate allowances for equipment response time measurement uncertainties and processing error. A safety margin is provided which includes:
an allowance of 5% of RATED THERMAL POWER to compensate for potential power measurement error; an allowance of 2'F to compensate for potential temperature measurement uncertainty; and a further allowance of 72 psi to compensate for pressure measurement error, trip system processing error, and time delay associated with providing effective termina-tion of the occurrence that exhibits the most rapid decrease in margin to the safety limit.
The 72 psi allowance is made up of a 22 psi pressure measure-ment allowance and a 50 psi time delay allowance.
loss of Turbine A Loss of Turbine trip causes a direct reactor trip when operating above 15% of RATED THERMAL PdWER.
This trip provides turbine prot?ction, reduces the severity of the ensuring transient and helps avoid the lifting of the main steam I!ne safety valves during the ensuing transient, thus extending the i
service life of these values. No credit was taken in the accident analyses for operation of this trip.
Its functional capability at the specified trip settia] is required to enhance the overall reliability of the Reactor Protec-tion System.
t MILLSTONE - UNIT 2 8 2-7 n
4 0
i LIMITING OAFETY SYSTEM SETTINGS BASES Underspfad - Reactor Coolant pub 21 I'
The Underspeed - Reactor Coolant Pumps trip provides core protection to prevent DNB in the event of a sudden significant decrease in rea; tor coolant 4
pump speed (with resulting decrease in flos) on all four reactor coolant pumps. The trip setpoint ensures that a reactor trip will be generated, y;
considering instrument errors and respotise times, in sufficient time to allow the DNBR to be maintained above 1.17 following a 4 pump loss of flow event.
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t MILLSTONE - UNIT ?
B 2-8
i 3/4.1 Ri&CTVITY CONTROL SYSTEM 1 3/4.1.1 REACTIVITY CONTROL SYSTEMS SliQlDOWN MARGIN - T
> 200'F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be 13.60% AK/K.
APPLICABILIII:
MODES 1, 2*, 5 and 4 ACTION:
With the SHUTDOWN MARGIN < 3.60% Ak/k, within 15 m.nutes initiate and continue boration at 140 gpm of boric acid solution at or greater than the required refueling water storage tank (RWST) concentration (ppm) until the required SHUTDOWN MARGIN is reached.
SURVEILLANCE REQUIREMENT 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be 13.60% AK/K:
i a.
Immediately upon detection of an inoperable CEA.
If the inoperable CEA is immovable or untrippable, the SHUTDOWN MARGIN, required by l
S)ecification 3.1.1.1, shall be increased by an amount at least equal to tie withdrawn worth of the immovable or untrippable CEA.
l t
b.
When in MODES 1 OR 2, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that CEA group withdrawal is within the Transient Insertion Limits of Specification 3.1.3.6.
l Trior to initial operation above 5% RATED THERMAL POWER after each refueling, with the CEA groups at the Transient Insertion Limits of I
Specification 3.1.3.6.
l I
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- See Special Test Exception 3.10.1 1
MILLSTONE - UNIT 2 3/4 1-1
REACTIVITY CONTROL SYSTEMS
' MODERATOR TEMPERATURE COEFFICIENT (MTC)
LIMITING CONDITION FOR OPERATION (Continued) 3 1.1.4 The moderator temperature coefficient (MTC) shall be:
Less positive than 0.1 x 10'4 Ak/k/'F whenever THERMAL POWER is I
a.
s 70% of RATED THERMAL POWER, b.
Less positive than 0.4 x 10'4 Ak/k/'F whenever THERMAL POWER is
> 70% of RATED THERMAL POWER, and Less negative than -2.8 x 10'4 Ak/k/*F at RATEU THERMAL POWER.
c.
APPLICABILITY: MODES I and 2*#
ACTION:
I' With the moderator temperature coefficient outside any one of the above limits, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
l SURVEILLANCE REQUIREMENT t
4.1.1.4.1 The MTC shall be determined to be within its limits by confirmatory measurements. MTC measured values shall be extrapolated and/or compensated to permit direct comparision with the predicted values.
I
- With K,f7 2 1.0.
- See Special Test Exemption 3.10.2.
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MILLSTONE - UNIT 2 3/4 1-5
(
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j 3/4.2 POWER DISTRIBUTION LIMITS LINEAR HEAT RATE LIMITING CONDITION FOR OPERATION (Continued) 1 3.2.1 The linear heat rate, including heat generated in the fuel, clad and moderator, shall not exceed 15.1 kw/ft.
APPLICABILITY: MODE 1.
ACTION:
During operation with the linear heat rate being monitored by the Incore Detector Monitoring System, comply with the following ACTION:
With the linear heat rate exceeding 15.1 kw/ft, as indicated by four or more l
coincident incore channels, within 15 minutes initiate corrective action to reduce the linear heat rate to less than or equal to 15.1 kw/ft and either:
l a.
Restore the linear heat rate to less than or equal to 15.1 kw/ft l
within one hour, or b.
Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
Durin3 operation with the linear heat rate being monitored by the Excore Detector Monitoring System, comply with the following ACTIONS:
With the linear heat rate exceeding its limit, as indicated by the AXIAL SHAPE INDEX being outside of the power dependent limits on the Power Ratio Recorder, either:
a.
Restore the AXIAL SHAPE INDEX to within the limits of Figure 3.2-2 within I hour from initially exceeding the linear heat rate limit, or b.
Be in at least HOT STANDBY within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENT 4.2.1.1 The linear heat rate shall be determined to be within its limits by continuously monitoring the core power distribution with either the excore detector monitoring system or with the incore detector monitoring system.
MILLSTONE UNIT 2 3/4 2-1 Amendment No. J/, M, M, 99
H-v.
A POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENT IContinued) 4.2.1.2 Excore Detector Monitorino System - The excore detector monitoring system may be used for monitoring the core power distribution by:
j a.
Verifying at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that the full length CEAs are withdrawn to and maintained at or beyond the Long Term Steady 1
State Insertion Limits of Specification 3.1.3.6.
b.
Verifying at least once per 31 days that the AXIAL SHAPE INDEX alarm setpoints are adjusted to within the allowable limits of Figure 3.2 2.
I 4.2.1.3 Incore Detector Monitorina Systtk - The incore detector monitoring system may be used for monitoring the core power distribution by verifying i
j that the incore detector Local Power Density alarms:
)
a.
Are adjusted to satisfy the requirements of the core power j
distribution map which shall be updated at least once per 31 days, b.
Have their alarm setpoint adjusted to less than or equal to 15.1 kw/ft when the following factors are appropriately included in the setting of these alarms:
}
- 1.
Flux peaking augmentation factors as shown in Figure 4.2-1.
l 2.
A measurement calculational uncertainty factor of 1.07, 3.
An engineering uncertainty factor of 1.03, j
+4.
A linear heat rate uncertainty factor of 1.01 due to axial fuel l
j densification and thermal expansion, and j
5.
A THERMAL POWEK measurement uncertainty factor of 1.02.
- These factors cre o'.iy ap>1tcable to fuel batches "A" through "L" I
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i MILLSTONE UNIT 2 3/4 2-2 Amendment No. 27, JE, E2, 99 i
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FIGURE 3.2-1
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l MILLSTONE - UNIT 2 3/4 2-8 Amendment No. 99
0 0
NIER DISTRIBUTION LIMITS T
10TAL INTEGRATED RADIAL PEAKING FACTOR - F y LIMITING CONDITION FOR OPERATION l
3.2.3 The calculated value of F defined as F -F7 (1+T ), shall be T
T 7
I 9
limitied to s 1.61.
APPLICABILITY: MODE 1*.
ACTION:
With FT > 1.61, within 6 heurs either:
7 R duce THERMAL POWER to bring the combination of THERMAL POWER and I
a.
9F to within the limits of Fiqure 3.2 3 and withdraw the full length CEAstoorbeyondtheLongTermSteadyStateInsertionLimitsof Specification 3.1 ' 6; or i
b.
Be in at least HOT STANDBY.
I i
SURVEILLANCE REQUIREMENT 4.2.3.1 The provisions of Specification 4.0.4 are not applicable, i
l T
T T
4.2.3.2 F shall be calculated by the expression F
-F 1 T ) and F shall
{
bedeterminEdtobewithinitslimitatthefollowin6intErv(a+ls g
r a.
Prior to operation above 70 percent of RATED THERMAL POWER after each fuel loadiqg, r
b.
At least once,tr 31 days of accumulated opuration in Mode 1, and c.
Within four hours if the AXIMUTHAL POWER TILT (T ) is > 0.020.
q 4.2.3.3 F shall be determined each time a calculation of F is required by I
T using the IEcore detectors to obtain a power distribution map with all full r
length CEAg at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Pump Combination.
T 4.2.3.4 T shall be determined ea h time a calculation of F is required and thevalueofT used to determine F shall be the measured vafue of T.
q q
- See Special Test Exception 3.10.2 MILLSTONE - UNIT 2 3/4 2-9 Amendment No. 3852,79,90, 99,113
~
1.4 M
o 12-n.
g UNACCEPTAnLE 0"ERATIO'I REGION g
(1.61,1.0)
E-*
1: :
4g 1.625,0.9) rz.
O 0.e -
i ZO 1.75,0.7)
Mb 0.6 -
<M AccEaTant,E ODERATIO" PECInN M
0.4 -
1 A4ic O
O2-
.3 O
c, 1.52 1.58 1.64 1.70 1.75 1.76 1.82 FIGURE 3.2-3 Allowable Fraction of RATED THERMAL PER vs. Total Integrated Radial Peaking Factor (F I r
l l
MILLSTONE - UNIT 2 3/4 2-9a
==
_ POWER DISTRIBUTION LIMITS AZ1MUTHAL POWER TITL - Tq LIMITING CONDITION FOR OPERATION 3.2.4 The AXIMUTHAL POWER T!L1 (T ) shall not exceed 0.02.
q APPLICABILITY: MODE I above 10% of RATED THERMAL, POWER *.
ACTION:
a.
With the indicated AZIMUTHAL POWER TILT determined to be 2 0.02 but s 0.10, either correct the power tilt within two hours or determine 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 per subsequent 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, that the Total Integrated Radial Peaking Factor (F') is within the limit of Specification 3.2.3.
b.
With the indicated ' IMUTHAL POWER TILT determined to be > 0.10, operation may proceed for up to 2 h urs provided that the Total Integrated Radial Peaking Factor (F ) is within the limits of Specification 3.2.3.
Subsequert op ration for the purpose of measurement and to identify the cause of the tilt is allowable provided the THERMAL POWER level is restricted to 1 20% of the maximum allowable THERMAL POWER level for the existing Reactor Coolant Pump combination.
SURVEILLANCE REQUIREMENT 4.2.4.1 The provisions of Specification 4.0.4 are not applicable.
4.2.4.2 The AZIMUTHAL POWER TILT sh111 be determined to be within the limit by:
a.
Calculating the tilt at least once per 7 days when the Channel High Deviation Alarm is OPERABLE,
- See Special Test Exception 3.10.2.
Millstone Unit 2 3/4 2 10 Amendment No. JE, J2, 10
l TABLE 3.3-1 (Continued) l TABLE NOTATION
[
- With the protective system tri) breakers in the closed position and the CEA
],
drive system capabb of CEA witidrawal.
(a) Trip may t's bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed when THERRAL POWER is 1 5% of RATED THERMAL POWER.
(b) Trip may be manually bypassed below 780 psia when all CEAs are fully inserted; bypass shall be automatically removed at or above 780 psia.
l (c) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall te automatically removed when THERMAL POWER is 2 15% of RATED THERMAL POWER.
(d) Trip does not need to be operable if all the control rod drive mechanisms are de-energized or if the RCS boron concentration is greater than or i
equal to the refueling concentration of Specification 3.9.1.
(e) Trip may be bypasskd during testing purscant to Special Test Exception 3.10.3.
l (f) AT Power input to trip may be bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is 2 5% of RATED THERMAL POWER.
ACTION STATEMENTS l
(
With the number of channels OPERABLE one len than required by L
ACTION 1 the Minimum Channels OPERABLE requirement, restore the inoperable channel 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 HOT STANDBY within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and/or open the protective system trip l
breakers.
ACTION 2 -
With the number of OPERABLE channels one less than the Total f
Number of Channels and with the THERMAL POWER level:
a.
15% of RATED THERMAL POWER, immediately placc the inoperable channel in the byp'.ssed condition; restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER l
above 5% of RATED THERMAL POWER.
l b.
2 5% of RATED THERMAL POWER, operation may continue with the t
incperable channel in the bypassed condition, provided the following conditions are satisfied:
i i
[
i
{
MILLSTONE - UNIT 2 3/4 3-4 Amendment No. 9, JS, 71, JJ6 L
L
INSTRUMENTATION INCORE DETECTORS LIMITING CONDITION FOR 0?ERATION 3.3.3.2 The incore detection system shall be OPERABLE with at least one OPERABLE detector segment in each core quadrcnt on each of the four axial elevations containing incore detectors and as further specified below:
a.
For monitoring the AXIMUTHAL POWER TILT:
At least two quadrant symmetric incore detector segment groups at each of the four axial elevations containing incore detectors in the outer 184 fuel assemblies with sufficient OPERABLE detector segments in these detector groups to compute at least two AZIMUTHAL POWER TILT valaas at each of the four axial elevations containing incore detectors, i
b.
For recalibration of the excore neutron flux detection system:
1 At lust 75% of all detector segments, 2.
A minimui;: of 9 OPERABLE incere detector segments at each 1
detector segicent level, and l
3.
A minimum of 2 OPERABLE detector segments in the inner 109 fuel assemblies a?d 2 OPERABLE segments in the outer 108 1
fuel assemblics at each segment level, l
c.
For mocitoring the UNR000ED INTEGRATED RADIAL PEAKING FACTOR or the linear hstt rate:
1.
At least 75% of all incera detector locations, 2.
A minimum,f 9 OPEMBLE incore detector segmsnts at each detector segmet hvel, and 3.
A minimum of 2 OPERABLE detector segments in the inner 109 fuel assemblies and 2 OPERABLE segments in the outer 108 fuel assemblies at each segment level.
An OPERABLE incore detector segment shall consist of an OPERABLE rhodium detector constituting one of the segments in a fixed deter. tor string.
An OPERABLE incore detection location shal' consist of a string in which at least three of the four incore detector segments are OPERABLE.
MILLSTONE - UNIT 2 3/4 3 30 Amendment No. 25,,45
INSTRUMENTATION LIMITING CONDITION FOR OPERATION (Continued)
An OPERABLE quadrant symmetric incore detector segment group shall, consist of a minimum of three OPERABLE rhodium incore detector segments in 90 symmetric fuel assemblies.
APPL 1CABIQTX: When the incore detection system is used for:
T a.
Monitoring the AZIMUTHAL POWER TILT, b.
Recalibration of the excore neutron flux detection system, or c.
Monitoring the UNRODDED INTEGRATED RADIAL PEAKING FACTOR or the linear heat rate.
ACTION:
i r
i With the incore detection system inoperable, do not use the system for the above applicable monitoring or calibration functicis. The provisions of specifications 4
3.0.3 and 3.0.4 are not applicable.
f' SURVEILLANCE REQUIREMENT I
4.3.3.2 The incore detection system shall be demonstrated OPERABLE:
l a.
By performance of a CHaANEL CHECK within N hours prior to its l
l use and at least once per 7 days thereafte" when required for:
i i
1.
Monitoring the AZIMUTHAL POWER TILT.
1 i
2.
Recalibration of the excore neutron flux detection system.
3.
Monitoring the UNRODDED INTEGRATED RADIAL FEAKING FACTOR or the l
linear he..: rate.
~
b.
At least once per 18 months by performance of a CHANNEL i
CALIBRATION operation which exempts the neutron detectors but
}
includes all electronic components. The neutron detectors shall be i
calibrated prior to installation in the reactor core.
l r
t i
f MILLSTONE - UNIT 2 3/4 3-31
SPECIAL TEST EXCEPTIONS GROUP HEIGHT AND INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2 The requirements of Specifications 3.1.1.4, 3.1.3.1, 3.1-3.0, 3.1.3.6, 3.2.3 and 3.2.4 may be suspended during the performance of PHYSlwS TESTS provided:
a.
The THERMAL POWER is restricted to the test power plateau which shall not exceed 85% of RATED THERMAL POWER, and b.
The limits of Specification 3.2.1 are maintained and determined as specified in Specification 4.10.2 below.
APPLICABILITY: MODES I and 2.
ACTION:
With any of the limits of Specification 3.2.1 being exceeded while the requirements of Specifications 3.1.1.4, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.3 and 3.2.4 are suspended, immediately:
a.
Reduce THERMAL PCWER sufficiently to satisfy the requirements of Specification 3.2.1 or b.
Be in HOT STANDBY within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
SURVEILLANCE REQUIREMENT 4.10.2.1 The THERMAL POWER shall be determined at least once per hour during PHYSICS TESTS in which the requirements of Specifications 3.1.1.4, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.3 or 3.2.4 are suspended and shall be verified to be within the test power plateau.
4.10.2.2 The linear heat rate shall be determined to be within the limits of Specification 3.2.1 by monitoring it continuously with the incore Detector Monitoring System pursuant to the requirements of Specifications 4.2.1.3 and 3.3.3.2 during PHYSICS TESTS above 5% of RATED THERMAL POWER in which the requirements of Specifications 3.1.1.4, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.3 or 3.2.4 are saspended.
MILLSTONE - UNIT 2 3/4 10-2
c o
i i
3/4.1 REACTIVITY CONTROL SYSTEMS BASES l
3/4.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARS @
t A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made 5
subcritical from all operating conditions, 2) the reactivity transierts associated with postulated accident conditions are controllable within i
acceptable limits, and 3) the reactor will be maintained sufficiently i
subcritical to preclude inadvertent criticality in the shutdown condition.
I SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T The most restrictive condition occurs at EOL, with T"heamat no load oiN A. ting temperature, an associated with a postulated line break accident and resulting i
uncontrolled RCS cooldown.
In the analysis of this 9ccident, a minimum l
SHUTDOWN MARGIN vf 3.60% Ak/k is initially required to control the reactivity t
transient. Accordingly, the SHUTDOWN MARCIN required by Specification 3.1.1.1 l
1s based upon this limitirg condittor, and is consistent with FSAR accident r
analysis assumptions.
For earlier periods during the fue', cycle, this salue 6
is conservative. With T 1200'F the reactivity transients resulting from any postulated accident *We minimal and a 2% Ak/k shutdown margin provides i
adequate prottction.
F 3/4.1.1.3 BORON DILUTION AND ADDITION i
A minimum flow rate of at least 3000 GPM provides adequate mixing, prevents stratification and ensures that reactivity changes will be gradual during I
boron concentration changes in the Reactor Coolant System.
A flow rate of at least 3000 GPM will circulate an equivalent Reactor Coolant System volume of 10,060 + 700/-0 cubic feet la approximately 30 minutes. The reactivity change
{
rate associated with boron concentration chane,es will be within the capability l
for co"ator recognition and control.
3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT (MTC)
The limitations on MTC are provided to ensure that the assumptions used in the j
accident and transient snalyses remain valid through each fuel cycle.
The l
surveillance requirements for measurement of the HTC during each fuel cycle are adequate to confirm the MTC,alue since this coefficient changes slowly due i
principally to the reduction in RCS boron concentration associated with fuel burnup.
The confirmation that the measured MTC value is within its limit I
providts assurance that the coefficient will be maintained within acceptable i
values throughout each fuel cycle, j
l
[
h MILLSTONE - UNIT 2 B 3/4 1-1 l
l
Q REACTIVITY f0NTROL SYSTEMS BASES 3/4.1.1.5 MINIMUM TEMPERATURE FOR CRITICAllTY t
The HTC is expected to be slightly negative at operating conditions.
l However, at the beginning of the fuel cycle, the MTC may be slightly positive at operating conditions and since it will become more positive at l
lower temperatures, this specification is provided to restrict reacto.
operaticn when T is significantly below the normal operating tempera-ture.
gg 3/4.1.2 BORAT10N SYSTEMS The boron injection system ensures that negative reactivity control is available during each mode of facility operation.
The components requir'd to perform this function include 1) borated water sources, 2) charging 1
pumps, 3) separate flow paths, 4) boric acid pumps, and 5) an emergency power supply fron OPERABLE diesel generators.
i With the RCS average temperature above 200'f, a minimum of two l
separate and.edundant boron injection flowpaths are provided to ensure single functional capability in the event an assumed failure of a pump or valve renders one of the flowpaths inoperable.
Redundant flow paths from th) Bo.*ic Acid Storage Tanks are achieved through Boric Acid Pumps, gravity
)
feed lines and Charging Pumps, pedundant flow paths from the Refueling
)
Water Storage Tank are achieved through Charging Pump flow path guaranteed by T;chnical Specification 3.1.2.2 and the HPSI flow path guaranteed by Techn' cal Specification 3.5.2 and 3.5.3.
Allowable out-of service periods i
ensure that minor component repair or corrective action may be completed t
without undue risk to overall facility safety from injection system t
)
]
failures during tre repair period, f
i 1
The minimum boration capability is sufficient to provide a SHUT 00WN MARGIN of 3.6% Ak/k at all temperatures above 200'F.
The maximum boration l
capa'vility requirement occurs at EOL from full power equilibrium xenon j
conditions and requires an equivalent of 4910 gallons of 3.5", boric acid solution from the boric acid tanks plus 15,000 of 1720 ppm borated water j
from the refueling water storage tank.
The refueling water storage tank
~i can also be used alone by feed and bleed using well under the 370,000 gallons of 1720 ppm borated water required.
The requirements for a ininimum contained volume of 370,000 gallons of borated water in the refueling water storage tank ensures the capability for borating the RCS to the desired level.
The specified quantity of i
borated water is consistent with the ECCS requirements of Specification 5.5.4.
Therefore, the larger volume of borated water is speciflad here too.
4 i
MILLSTONE - UNIT 2 B 3/4 1 2 l
BASES i
3/4.1.2 BORAT10N SYSTEMS (Continued)
The analysis to determine the boration requirements assumed that the Reactor Coolant System is borated concurrently with cooldown.
In the limiting situation when letdown is not available, the cooldown is assumed l
to be inititited within 26 hourt, and cooldown to 200'F, is completed in the l
next 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br />.
With the RCS temperature below 200'F, one injection system is accept-able without single failura consideration un the basis of the stable l
reactivity condition of the reactor and the additional restrictions prohib-iting CORE ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable, j
The boron capability required below 200'F is based upon providing i 2% A k/k SHUTDOWN MARGIN at 140*F after xenon decay.
This condittoi requires either 3750 gallons of 2.5% boric acid solution from the bort:
acid tanks or 57,300 gallons of 1720 ppm borated water from the refueling water storage tank, i
i The maximum boron concentration requirement (3.5%) and the minimum j
j temperature requirement (55'F) for the Boric Acid Storage Tank ensures that i
baron does not precipitate in the Boric Acid System.
The daily surveil-i lance reg'Jirement provides sufficient assurance that the temperature of the J
tank will be maintained higher that 55'F at all times.
A minimum boron concentration of 1720 ppm is required in the RWST at all times in order to satisfy safety ar,alysi s assumptions for boron i
dilution incidents and other transients using the RWST as a borated water i
source es well as the analysis assumption to determine the boration i
j r,!quirement to ensure adequate.hutdown margin.
3/4.1.3 MOVEABLE CONTROL ASSEMBLIES l
The specifications of this section ensure that (1) acceptable power f
distribution limits are maintained. (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the potential effects of a CEA ejection accident are limited to acceptable levels.
r The ACTION statements which permit lin4tted variations from the basic requirements are accompanied by additional restrictions which ensure that i
the original criteria are met.
l The ACTION statements applicable to an immovable or untrippable CEA f
and to a large misalignment (120 steps) of two or more CEAs, require a prompt shutdown of the reactor since either l
l 1
k l
ll
(
L MILLSTONE - UNIT 2 B 3/4 1-3 A endment No. 38, fl, 72, lif,133 l
s
~
BASES 3/4.1.3 MOVEABLE CONTROL ASSEMBLIES (Continued) of these conditions may be indicative of a possible loss of mechanical functional capability of the CEAs and in the event of a immovable or untrippable CCA, the loss of SHUT 00WN MARGIN.
For small misalignments (< 20 steps) of the CEAs, there is 1) a small degradation in the peaking factors relative to those assumed in generating LCOs and LSSS setpoints for DNBR and linear heat rate, 2) a small effect on the time dependent long term power distributions relative to thoss used in generating LCOs and LSSS setpoints for DNBR and linear heat rate, 31 a small effect on the available SHUTDOWN MARGIN, and 4) a small effect on the e;ected CEA worth used in the safety analysis.
Therefore, the ACTION statement associated with the small misalignment of a CEA permits a one hour time interval during which attempts way be m.de t.s restore the CEA to within its alignment requirements prior to initiating a reduction in THERMAL POWER.
The one hour time limit is sufficient to (1) identify causes of a misaligned CEA, (2) take appropriate corrective action to realign the CEAs and (3) minimize the effects of xenon redistribution.
Overpower margin is pro *ided to protect the core in '.he event of a large misalignment (2 20 steps) of a CEA.
However, this misalignment would cause distortion of the core power distribution.
The reactor protective system would not detect the degradation in the radial peaking factor and since variations in other system parameters (e.g.,
pressure and coolant tertperature) may not be sufficient to cause trips, it is possible that the reactor could be operating with process variables less conservative than those assumed in gen 9 rating LCO and LSSS setpoints.
Therefore, the ACTION statement associated with the large misalignment of a CEA requires a prompt and significant reduction in THERMAL POWEl! prior to attempting realignment of the misaligned CEA.
The ACTION statements applicable to misaligned or inoperable CEAs include requirements to align the OPERABLL' CEAs in a given group with the inoperable CEA.
Confermane.e with these alignment requirements bring the core, within a short period of time, to a configuration consistent with that assumed in generating LCO and LSSS setpoints.
However, extended operation with CEAs significantly inserted in the core may Icad to pertur-bations in 1) local burnup, 2) peaking factors and 3) available shutdown margin which are more adverse than the conditions assumed to exist in the i
MllLSTONE UNIT 2 B 3/4 1-4 Amendment No. 38, 133
3/4.2 POWER DISTRIBUTION LIMITS BASES 3/4.2.1 LINEAR HEAT RATE The limitation on linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 2200'F.
t'ither of the two core power distribution monitoring systems, the Excore Detector Monitoring System and the Incore Detector Monitoring System, provide adequate monitoring of the core power distribution and are capable of verify-ing that the linear heat rate does not exceed its limits.
The Excore Detector Monitoring System performs this function by continuously monitoring the AXIAL SHAPE INDEX with two OPERABLE excore neutron flux detectors and verifying that the AXIAL SHAPE INDEX is maintained within the allowable limits of Figure 3.2-2 using the Power Ratio Recorder. The power dependent limits of the Pc n Ratio Recorder are less than or equal to the limits of Figure 3.2-2.
In conjunction with the "se of the excore monitoring system and in establishing the AXIAL SHAPE INDEA limits, the following as umptions are made:
- 1) the CEA insertion limits of Specifications 3.1.3.2, 3.3.3.5 and 3.1.3.6 are satisfied,
- 2) the flux peaking augmentation factors are as shown in Figure 4.2-1, 3) the AZlHUTHAL POWER TILT restrictions of Specif, cation 3.2.4 are satisfied, and
- 4) the Total Integrated Radial Peaking Factor does not exceed the limits of Specification 3.2.3.
The Incore Detector Monitoring System continuously provides a direct mnsure of the peaking f actors and the alarms which have been established for the individual incore detector segments ensure that the peak lineer heat rates w:ll be maintained within the allowable limits of Figure 3.2-1.
The setpoints for these alarms include allowances, set in the conservative directions, for
- 1) flux peaking cugmentation factors as shown in Figure 4.2-1, 2) a measure-ment-calculathnal uncertainty factnr of 1.07, 3) an engineering uncertainty factor of 1.03, 4) an allowance of 1.01 for axial fuel densification and thermal expansion, and 5) a THERHAL POWER measurement uncertainty factor of 1.02.
Note the items (1) and (4) above are only applicable to fuel batches "A" through "L".
T 3/4.2.3 and 3/4.2.4 TOTAL INTEGRATED RADIAL PEAKING FACTORJ F AND AZlHUTHAL 7
POWER TILT - I q T
The ifmitations on F and T are provided to ensure 1) that the assump-tions used in the analysis for %stablishing the Linear Heat Rate and Local power Density - High LCOs and LSSS setpoints remain valid during operation at the various allowable CEA group insertion limits, and, 2) ensure that the assumptions used in the analysis establishing the DNB Hargin LCO, and Thermal Margin / Low Pressure LSSS setpoints remain valid during operation at the T
various allowable CEA group insertion limits.
If F or T exceed their basic ilmitations, operation may continue under the addit'ional 9estrictions imposed by the ACTION statements since these additional restrictions provide adequate provisions to assure that the assumptions used in establishing the linear Heat Rate, Thermal Margin / Low Pressure and Local Power Density - High tCOs and LSSS l
HILLSTONE UNIT 2 B 3/4 2-1 Amendment No. 38, 52, 112
00WER DISTRIBUTION LIMITS BASES setpoints remain valid. An AZIMUTHAL POWER TILT > 0.10 is not expected and if it should occur, subsequent operation would be restricted to only those operations required to identify the cause of this unexpected tilt.
T The value of T that must be used in the equation F =Fr (1 + T ) is the measured tilt, q
r q
T The surveillance requirements for verifying that F angT are within their limits provide assurance that the actual values if F a6 T do not T
exceed the assuined values.
Verifying F foading Srior to exceeding 75% of RATED THERMAL P06(R pro, after each fuel vides additional assurance that the core was properly loaded.
3/4.2.6 DNB MARGIN The limitations provided in this specification ensure that the assumed margins to DNB are maintained.
The limiting values of the parameters in this specification are those assumed as the initial conditions in the accident and transient analyses; therefore, obration must be maintained within the speci-fled limits for the accident and transient analyses to remain valid.
MILLSTONE UNIT 2 B 3/4 2 2 Amendment No. J8, E2, 122
e 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1__ COOLANT LOOPS AND COOLANT CIRCiLATION The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation, and maintain DNBR above 1.17 during all normal operations and anticipated transients.
A single reactor coolant loop with its steam generator filled above 10%
of the span provides sufficient heat removal capability for core cooling while in H0 DES 2 and 3; however, single failure considerations require plant cooldown if component repairs and/or corrective actions cannot be made within the allowable out of-service time.
t In MODES 4 and 5, a single reactor coolant loop or shutdown cooling loon provides sufficient heat re aval capability for removing decay heat; but si,1gle failure considerations requirr that at least two loops be OPERABLE.
Thus, if the reactor coolant loops are not OPERABLE, this specification requires two shutdown cooling loops to be OPERABLE.
The operation of one Reactor Coolant Pump or one snutdown coolirg pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.
The reactivity change rate associated vith boren
- r. ductions will, therefore, be within the capability of operator recognition and control.
The restrictions on starting a Reactor Coolant Pump during MODES 4 and 5 with one or more RCS cold legs 1275'F are provided to prevent RCS pressure transients, caused by energy additions from the secondary system, which could i
exceed the limits of Appendix G to 10 CFR Part 50.
The RCS will be protected against overpressure transients and will not exceed the limits of Appendir G by either (1) estricting the water volume in the f.essurizer and thtreby providing a volume for the primary coolant to expand into or (2) by i
l restricting starting of the RCPs to when the secondary water te:rperature of each steam generator is less than 43'F (31'F when measured by a sur' ace contact instrument) above each of the RLS cold leg temperatures.
3/4. 4. 2 _ S AF FU_y_Alili The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psia.
Each safety valve is designed ta relieve 296,000 lbs per hour of saturated steam at the valve setroint.
The relief capacity of a single safety valve is adequate to relieve any overpressure condition wnich could occur during shutdown.
In the event that no safety valves are OPERABLE, an operating shutdown cooling loop.
connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.
MILLSTONE - UNIT 2 B 3/4 4 1 Amendment No. H, 6. O 1
.