ML20247E597
| ML20247E597 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 03/20/1989 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Northeast Nuclear Energy Co (NNECO), Connecticut Light & Power Co, Western Massachusetts Electric Co |
| Shared Package | |
| ML20247E603 | List: |
| References | |
| DPR-65-A-139 NUDOCS 8904030175 | |
| Download: ML20247E597 (62) | |
Text
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UNITED STATES g
NUCLEAR REGULATORY COMMISSION 3.
- j WASHINGTON, D. C. 20555
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i NORTHEAST NUCLEAR ENERGY COMPANY THE CONNECTICUT LIGHT AND POWER COMPANY THE WESTERN MASSACHUSETTS ELECTRIC COMPANY DOCKET N0. 50-336 MILLSTONE NUCLEAR POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 139 License No. DPR-65 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The applications for amendment by Northeast Nuclear Energy Company, et al. (the licensee), dated November 15, 1988 and February 1, 1989 comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in'10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The i'.suance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
8904030175 880320 PDR ADOCK 05000336 P
~ 2..
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-65 is hereby amended to read ss follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.139, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of issuance, to be implemented within 30 days of issuance.
FOR THE NUCLEAR REGU ATORY COMf1ISSION J n
. Stolz, Directo rojpct Directorate 4
iM.sion of Reactor Projects I/II
~
Office of Nuc1 ear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: March 20, 1989 e
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DEFINITIONS i
ENGINEERED SAFETY FEATURE RESPONSE TIME (Continued) performing its safety function i.e positions, pump discharge pressu(res., the valves travel to thhN required reach their required values, etc.).
Times shall include diesel generator starting and sequence loading dela where applicable.
PHYSICS TESTS 1.28 nuclear characteristics of the reactor core and related
- 1) described in Chapter 13.0 of the FSAR, 2) authorized under the provision'i of 10 CFR 50.5g, or 3) otherwise approved by the Connission.
UNRODDED INTEGRATED RADIAL PEAKING FACTOR - F,.
1.2g pin power to the average pin power in an unrodded co SOURCE CHECK 1.30 when the thannel sensor is exposed to radiation.A SOURCE CH RADIOLOGICAL EFFLUENT MONITORING AND 0FFSITE 1.31 A RADIOLOGICAL EFFLUENT MONITORING MANUAL shall be.a ma the site and environmental sampling and analysis programs for measurements of radiation and radioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential radiation exposures to individuals from station operation.
be a manual containing the methodology and parameters to be use culation of offsite doses due to radioactive gaseous and liquid effluents and alarm / trip setpoints.in the calculation of gaseous and liquid effluent mo a-tion 6.16.
Requirements of the REMODCM are provided in Specifier-RADI0 ACTIVE WASTE TREATMENT SYSTEMS 1.33 RADIOACTIVE WASTE TREATMENT SYSTEMS are those liquid waste systems which are required to maintain control over r,adioactive material gaseous and solid in order to meet the LCOs set forth in these specifications.
PURGE - PURGING 1.34 PURGE or PURGING is the controlled process of dischargingEr or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.
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MILLSTONE - UNIT 2 1-6 Amendment No. 75,104 j
j
DEFINITIONS AXIAL SHAPE INDEX 1.23 The AXIAL SHAPE INDEX (Y ) used for normal control and indication is g
the power level detected by the 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,)
used for the trip and pretrip signals in the reactor protection system ts the above value (Ye) modified by an appropriate multiplier (A) and a constant (B) to determinh the true core axial power distribution for that channel.
Y E " L+U YI - AYE+0 1.24 Deleted.
ENCLOSURE BUILDING INTEGRITY 1.25 ENCLOSURE BUILDING INTEGRITY shall exist when:
1.25.1 Each door in each access opening is closed except when the access opening is being used for normal transit entry and exit, and 1.25.2 The enclosure building filtration system is OPERABLE.
REACTOR TRIP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until electrical power is interrupted to the CEA drive mechanism.
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 l
MILLSTONE - UNIT 2 1-5 Amendment No. JE 139
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and maxi-mum cold leg coolant temperature shall not exceed the limits shown on Figure 2.1-1.
APPLICABILITY: MODES 1 and 2.
ACTION:
Whenever the point defined by the combination of maximum cold leg temper-ature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.
APPLICABILITY: MODES 1, 2, 3, 4 and 5.
ACTION:
MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2750 psia, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
MODES 3,;4and5 Whenever the Reactor Coolan, System pressure has exceeded 2750 psia, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.
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2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SETPOINTS 2.2.1 The reactor protective instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.
APPLICABILITY,; AS SHOWN FOR EACH CHANNEL IN TABLE 3.3-1.
ACTION:
With a reactor protective instrumentation setpoint less conservative than
_ the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1.1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
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o, 0
0.2 OA 0.6 0.s too FRACTION OF RATED THERMAL POWER p
FIGURE 2e2e1 Local Power Density - High Trip Setpoint Part 1 (Fraction of RATED THERMAL POWER Versus OR I 2
e f
2-6 MIt.CSTONE - UNIT 2 e.
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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 rate. 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 temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure 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 been 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, is 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 conditions.
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 for reactor coolant inlet temperatures less than or equal to 580*F. The dashed line at 580*F coolant l
inlet temperatures is not a safety limit; however, operation above 580'F is l
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 THERMAL POWER levels higher than 112% of RATED THERMAL POWER is prohibited by the high power level trip setpoint specified in Table 2.2-1.
The area of safe operation is below and to the left of these lines.
1 1
MILLSTONE - UNIT 2 B 2-1 Amendment No. 7, 52,91, 139
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.e THIS PAGE INTENTIONALLY LEFT BLANK.
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MILLSTO'NE - Ohlt 2 8 2-2 Amendment No. 52.61 g W $" e
____p
i SAFETY LIMIT BASES 4
The conditions for the Thermal Margin Safety Limit curves in figure 2.1-1 to be valid are shown on the, figure.
The reactor 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 integrity 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 III of the ASME Code for Nuclear Power Plant Components which permits a maximum transient pressure of 110% (2750 psia) of design pressure.
The Reactor Coolant System piping, valves and fittings, are designed to ANSI B31.7, Class I which permits a maximum transient pressure of 110% (2750 psia) of compon'ent design pressure. The Safety Limit of 2750 psia is therefore consistent with the design criteria and associated code requirements.
The entire Reactor Coolant System is hydrotested at 3125 psia to demon-strate integrity prior to initial operation.
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MILLSTONE - UNIT 2 B 2-3 Amendment No. 7/$2,61,139
- 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES
~
2.2.1 REACTOR TRIP $ET POINTS The Reactor Trip Setpoints specified in Table 2.2-1 are'the values at j
which the Reactor Trips are set for each parameter.
The Trip Values have J
been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits. Operation with a Trip j
Setpoint less conservative than its setpoint but within its specified Allowable Value is acceptable on the basis that each Allowable Value is equal to or less than the drift allowance assumed to occur for each trip used in the accident analyses.
Manual Reactor Trip The Manual Reactor Trip is a redundant' channel to the automatic protective instrumentation channels and provides manual reactor trip capability.
~
Power Level-High The Power Level-High trip provides reactor core protection against reactivity excursions which are too rapid to be protected by a Pressurizer Pressure-High or Thermal Margin / Low Pressure trip.
The Power Level-High trip setpoint is operator adjustable and can be set no higher than 9.6% above the indicated THERMAL POWER level.
Operator j
action is required to increase the trip setpoint as THERMAL POWER is increased.
The trip setpoint is automatically decreased as THERMAL POWER decreases.
The trip setpoint has a maximum value of 106.6% of RATED THERMAL POWER and a minimum setpoint of 14.6% of RATED THERMAL POWER. Adding to this maximum value the possible variation in trip point due to calibration and instrument errors, the maximum actual steady-state THERMAL POWER level at which a trip would be actuated is 112% of RATED THERMAL POWER. which is the value used in the accident analyses.
Reactor Coolant Flow-Low The Reactor Coolant Flow-Low trip provides core protection to prevent DN3 in the event of a sudden significant decrease in reactor coolant fl ow.
Provisions have been made in the reactor protective system to permit MILLSTONE - UNIT 2 B 2-4 Amendment No. 61 c Eb -
e
LIMITING SAFETY SYSTEM SETTINGS BASES Reactor Coolant Flow-low (Continued) operation of the reactor at reduced power if 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 pumps operating, the Reactor Coolant Flow-Low trip setpoints, the Power Level-High 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-Hiah The pressurizer Pressure-High trip, backed up by the pressurizer code safety valves and main steam line safety valves, provides reactor coolant system protection against overpressurization in.the 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 relief. valves avoids the undesirable opera-tion of the pressurizer code safety valves.
Containment Pressure-Hiah 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-Low 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. The setting of 680 psia is sufficiently below the full-load operating point so as not to interfere with normal opera-tion, but still high enough to provide the required protection in the event of excessively high steam flow.
This setting was used with an uncertainty factor of 22 psi in the accident analyses.
MILLSTONE - UNIT 2 B 2-5 Amendment No. 52, 61, 139
LIMITING SAFETY SYSTEM SETTINGS BASES i
Etaam 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 j
the steam generators at the time of trip to provide a margin of more than 10 j
minutes before auxiliary feedwater is required.
Local Power Density-Hiah The Local Power Density-High trip, functioning from AXIAL SHAPE INDEX monitoring, is provided to ensure 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 maldistributions. 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 channel s.
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 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 occur during any anticipated ' operational occurrence prior to a Power Level-High trip is assumed.
Thermal Marcin/ Low Pressure The Thermal Margin / Low Pressure trip is provided to prevent operation when the DNBR is less than 1.17.
l MILLSTONE - UNIT 2 B 2-6 Amendment No. 38,41,52,61,139
LIMITING SAFETY SYSTEM SETTINGS BASES Thermal Marain/ Low Pressure (Continued)
The trip is initiated whenever the reactor coolant system pressure signal drops below either 1850 psia or a computed value as described below, whichever l
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 POWER 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 anticipated 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 l
processing error, and time delay cssociated 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 POWER.
This trip provides turbine protection, reduces the severity of the ensuring transient and helps avoid the lifting of the main steam line safety valves during the ensuing transient, thus extending the 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 setting is required to enhance the overall reliability of the Reactor Protec-tion System.
MILLSTONE - UNif 2 B 2-7 Amendment Nn. 3B/5d/139
LIMITING SAFETY SYITEM SETTINGS BASES Underspeed - Reactor Coolant Fumos The Underspeed - Reactor Coolant Pumps trip provides core protection to prevent DNB in the event of a sudden significant decrease in reactor coolant pump speed (with resulting decrease in flow) on all four reactor coolant pumps. The trip setpoint ensures that a reactor trip will be generated, considering instrumec'; errors and response times, in sufficient time to allow the DNBR to be maintained above 1.)? following a 4 pump loss of flow event.
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i MILLSTONE - UNIT 2 B 2-8 Amendment No. 32,61,139
3/4.1 REACTVITY CONTROL SYSTEMS 3/4.I.1 REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN -
T____ > 200*F avy LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be 13.60% AK/K.
APPLICABILITY: MODES 1, 2*, 3 and 4 ACTION:
With the SHUTDOWN MARGIN < 3.60% Ak/k, within 15 minutes initiate and continue l
boration at 2 40 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 1 3.60% AK/K:
a.
Immediately upon detection of an inoperable CEA.
If the inoperable CEA is. immovable or untrippable, the SHUTDOWN MARGIN, required by Specification 3.1.1.1, shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable CEA.
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.
c.
Prior to initial operation above 5% RATED THERMAL POWER after each refueling, with the CEA groups at the Transient Insertion Limits of Specification 3.1.3.6.
I
- See Special Test Exception 3.10.1 MILLSTONE - UNIT 2 3/4 1-1 Amendment No. 3.61,72 74,139
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REACTIVITY CONTROL SYSTEMS SURVEiLLANCEREQUIREMENTS(Continued)
' ??
d.
When in MODES 3 or 4, at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by consider-ation of the following factors:.
1.
Reactor coolant system boron concentration.
2.
CEA position.
3.
Reactor coolant temperature,.
4.
Fuel burnup based on gross thermal energy generation, 5.
Xenon concentration, and 6.
Samarium concentration.
4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within + 1.0% Ak/k at least once per 31 Effective Ful1 Power Days. This comparison shall consider at least those factors stated in Specification 4.A 1.1.1.d, above.
The predicted reactivity values shall be adjusted (normalized) to correspond-to the actual core conditions prior to excegding Afuel burnup of 60 Effactive Full Power Days after each refueling.
MILLSTONE - UNif a 3/4.1-2 l
w
J REACTIVITY CONTROL SYSTEMS tiODERATOR TEMPERATURE COEFFICIENT (MTC)
]
LIMITING CONDITION FOR OPERATION (Continued) l 3.1.1.4 The moderator temperature coefficient (MTC) shall be:
Less positive than 0.7 x 10~4 Ak/k/*F whenever THERMAL POWER is 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 RATED THERMAL POWER.
c.
APPLICABILITY: MODES 1 and 2*#
ACTION:
With the moderator temperature coefficient outside any one of the above limits, be in at least HOT STANDBY within 6 h'ours.
SURVEILLANCE REQUIREMENT 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.
- With K 2 1.0.
eff
- See Special Test Exemption 3.10.2.
MILLSTONE - UNIT 2 3/4 1-5 Amendment No. 79,4),7f.130
.=
REACTIVITY CONTROL SYSTEMS SdRVEILLANCEREOUIREMENTS(bontinued)
~
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4.1.1.4.2 The MTC shall be determined at the following frequencies and THERMAL POWER conditions during each fuel cycle:
l a.
Prior to initial operation above 5% of RATED THERMAL POWER, after each refueling.
h.
At any THERMAL POWER, within 14 EFPD a'fter each fuel loading i
at equilibrium boron concentration.
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l MILLSTO'NE - UNIT 2 3/4 1-6 Amendment No.)S,74
3/4.2 POWER DISTRIBUTION LIMITS LINEAR HEAT RATE l
LIMITING CONDITION FOR OPERATION (Continued) 3.2.1 The linear heat rate, including heat generated in the fuel, clad and moderator, shall not exceed:
a.
15.1 kw/ft when the reactor coolant flow rate measured per Specification 4.2.6.1 2 340,000 gpm.
b.
14.5 kw/ft when the reactor coolant flow rate measured per Specification 4.2.6.1 2 325,000 gpm r.nd < 340,000 gpm.
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 the limit 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 the limit and either:
l.
a.
Restore the linear heat rate to less than or equal to the limit 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 />.
During 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 eycore detector monitoring system or with the incore detector monitoring system.
MILLSTONE UNIT 2 3/4 2-1 Amendment No. J7, JE, E2, 99 139
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EQWER DISTRIBUTION LIMITS
{
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SURVEILLANCE REQUIREMENT (Continued) 4.2.1.2 Excore Detector Monitorina Syrts - The excore detector monitoring system may be used for monitoring the core power distribution by:
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 State Insertion Limits of Specification 3.1.3.6.
b.
Verifying at least once per 31 days that the AXIAL SHAPE INDEX I
alarm setpoints are adjusted to within the allowable limits of l
j Figure 3.2-2.
4.2.1.3 Incore Detector Monitorina System - The incore detector monitoring system may be used for monitoring the core power distribution by verifying that the incore detector Local Power Density alarms:
a.
Are adjusted to satisfy the requirements of the core power distribution map which shall be updated 'at least once per 31 days.
b.
Have their alarm setpoint adjusted to less than or equal to the limit when the following factors are appropriately included in the l
setting of these alarms:
- l.
Flux peaking augmentation factors as shown in Figure 4~.2-1.
[
2.
A measurement-calculational uncertainty factor of 1.07, 3.
An engineering uncertainty factor of 1.03,
- 4.
A linear heat rate uncertainty factor of 1.01 due to axial fuel l
densification and thermal expansion, and 5.
A THERMAL POWER measurement uncertainty factor of 1.02.
- These factors are only applicable to fuel batches "A" through "L" MILLSTONE UNIT 2 3/4 2-2 Amendment No. 77, #, M, #
139
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FIGURE 3.2-1 LEFT BLANK INTENTIONALLY l
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l MILLSTONE - UNIT 2 3/4 2-5 Amendment No. JE, E2, 79, sp, 97,99,177 139
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i MILLSTONE UNIT 2 3/4 2 *)
Amendment No. 139 u-_-_---_---__---_--------.____
THIS PAGE LEFT BLANK INTENTIONALLY l
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MILLSTONE - UNIT 2 3/4 2-8 Amendment No. 9), 139
POWER DISTRIBUTION LIMITS TOTALINTEGRATEDRADIALPEAKINGFACTOR-Ff LIMITING CONDITION FOR OPERATION T
T 3.2.3 The calculated value of F defined as F =Fr (1+T ), shall be r
r q
limited to:
a.
0.90 < PF 1 1.00 Ff1(11.73-PF)((1.24x10-7 x FL)+0.108) b.
0.70 < PF 1 0.90 Ff1(3.50-PF)((5.18x10-7 x FL) + 0.449)
T
-7 c.
PF s 0.70 F 5 1.75 ((8.28 x 10 x FL) + 0.718) y where:
PF = THERMAL POWER divided by RATED THERMAL POWER FL - The lesser of either:
1)
The reactor coolant flow rate measured per Specification 4.2.6.1 down to a minimum of 325,000 gpm, or 2) 340,000 gpm APPLICABILITY: MODE 1*.
ACTION:
T With F exceeding its limit,within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> either:
l 7
R9duceTHERMALPOWERtobringthecombinationofTHERMALPOWERand a.
F to within the limit and withdraw the full length CEAs to or l
bEyondtheLongTermSteadyStateInsertionLimitsofSpecification 3.1.3.6; or b.
Be in at least HOT STANDBY.
SURVEILLANCE REQUIREMENT 4.2.3.1 The provisions of Specification 4.0.4 are not' applicable.
T T
be determined to be within its limit at the followin6 intEr 4.2.3.2 F shall be calculated by the expression F
=F a.
Prior to operation above 70.nercent of RATED THERMAL POWER after each fuel loading, b.
At least once per 31 days of accumulated operation in Mode 1, and c.
Within foui hours if the AXIMUTHAL POWER TILT (T ) is > 0.020.
q T
4.2.3.3 F shall be determined each time a calculation of F is required by using the iEcore detectors to obtain a power distribution map with all full r
length CEAs at or above the Long Term Steady State Inserticn Limit for the existing Reactor Coolant Pump Combination.
T 4.2.3.4 T
shall be determined ea9 time a calculation of F is required and h
thevalueofT used to determine F shall be the measured vafue of T.
q r
q
- See Special Test Exception 3.10.2 MILLSTONE - UNIT 2 3/4 2-9 Amendment No. JE, E, 79, 99, 99, JU, 139
POWER DISTRIBUTION LIMITS AZIMUTHAL POWER TITL - T
~
q LIMITING CONDITION FOR OPERATION 3.2.4 The AXIMUTHAL POWER TILT (T ) shall not exceed 0.02.
q APPLICABILITY. MODE I above 50% of RATED THERHAL POWER *.
ACTION:
With the indicated AZIMUTHAL POWER TILT determined to be 1 0.02 but a.
1 0.10, either correct the power tilt within two hours or determine withinthenext2hoursandatleastoncepersubs9quent8 hours, that the Total Integrated Radial Peaking Factor (F ) is within the limit of Specification 3.2.3.
r b.
With the indicated AZIMUTHAL POWER TILT determined to be > 0.10, operation may proceed for up to 2 h urs provided that the Total 9
Integrated Radial Peaking Factor (F ) is within the limits of Specification 3.2.3.
SubsequentopErationforthepurposeof measurement and to identify the cause of the tilt is allowable provided the THERMAL POWER level is restricted to 1 zur. of the maximum allowable THERMAL POWER level for the existing Reactor Coolant Pump combination.
SURVEILLANCE REQUIREMENT 4.2.4.I The provisions of Specification 4.0.4 are not applicable.
4.2.4.2 The AZIMUTHAL POWER TILT shall be determined to be within the limit by:
Calculating the tilt at least once per 7 days when the Channel High a.
Deviation Alarm is OPERABLE,
- See Special Test Exceptien 3.10.2.
l l
l Millstone Unit 2 3/4 2-10 Amendment No. 7E, JE, pp, 139
POWER DISTRIBUTION LIMITS DNB MARGIN LIMITING CONDITION FOR OPERATION 3.2.6 The DNB margin shall be preserved by maintaining the cold leg temperature, pressurizer pressure, reactor coolant flow rate, and AXIAL SHAPE INDEX within the limits specified in Table 3.2-1 and Figure 3.2-4.
APPLICABILITY: MODE 1.
ACTION:
With any of the above parameters exceeding its specified limits, restore the parameter to within its above specified limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or
~
reduce THER:!AL POWER to < 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE RE0VIREMENTS 4.2.6.1 The cold leg temperature, pressurizer pressure, ans \\XIAL SHAPE INDEX shall be determined to' be within the limits of Table 3.2-1 and Figure 3.2-4 at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The reactor coolant flow rate shall be determined to be within the limit of Table 3.2-1 at least once per 31 days.
4.2.6.2 The provisions of Specification 4.0.4 are not applicable.
t MILLSTONE - UNIT 2 3/4 2-13 Amendment No. JE.90 ll3
BBLE 3.2-1 DNB MARGIN LIMITS Four Peactor Coolant l
Parameter Pumos Operatino Cold Leg Temperature s 549'F l
Pressurizer Pressure 1 2225 psia
- Reactor Coolant Flow Rate 1 340,000 gpm**
l AXIAL SHAPE INDEX Figure 3.2-4
- Limit not applicable during either a THERMAL POWER ramp increase in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step increase of greater than 10% of RATED THERMAL POWER.
T
- Flow reductions to 325,000 gpm are compensated for by reducticas in the F limit (Specification 3.2.3).
r l
I MILLSTONE - UNIT 2 3/4 2-14 Amendment No. JE, JZ, 79,
- ,)T3,139
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TABLE 3.3-1 (Continued)
TABLE NOTATION
- Wi'h'the protective system trip breakers in the closed position and the CEA driv system capable of CEA withdrawal.
(a) Trip may be bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER 1:: 2 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.
(c) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed wher. THERMAL POWER is 215% 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 equal to the refueling concentration of Specification 3.9.1.
(e) Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3.
(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 ACTION 1 -
With the number of channels OPERABLE one less than required by 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 breakers.
ACTION 2 -
With the number of OPERABLE channels one less than the Total Number of. Channels and with the THERMAL POWER level:
a.
1 5% of RATED THERMAL POWER, immediately place the inoperable channel in the bypassed condition; restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 5% of RATED THERMAL POWER.
b.
2 5% of RATED THERMAL POWER, operation may continue with the inoperable channel in the bypassed condition, provided the following conditions are satisfied:
MILLSTONE - UNIT 2 3/4 3-4 Amendment No. jl, JE, 77, JJE, 139
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INSTRUMENTATION INCORE DETECTORS LIMITING CONDITION FOR OPERATION 3.3.3.2 The incore detection syster, shall be OPERABLE with at least one OPERABLE detector segment in each core quadrant 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 values at each of the four axial elevations containing incore detectors.
b.
For recalibration of the excore neutron flux detection system:
1.
At least 75% of all detector segments, 2.
A minimum of 9 OPERABLE incore detector segments at each detector segment level, 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, c.
For monitoring the UNRODDED INTEGRATED RADIAL PEAXING FACTOR or the linear heat rate:
1.
At least 75% of all incore detector locations, 2.
A minimum of 9 OPERABLE incore detector segments at each detector segment level, and 3.
A minimum of 2 OPERABLE detec7.or segments in the inner 109 fuel assemblies and 2 OPERABLE segments in the outer 108 fuel assemblies at each segment leval.
An OPERABLE incore detector segment shall consist of an OPERABLE rhodium detector constituting one of the segments in a fixed detector string.
An OPERABLE incore detection location shall 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. EE, /E, 139
a.
INSTRUMENTATION LIMITING CONDITION FOR OPERATION (Continued)
An OPERABLE quadrant symmetric incore detector segment group shall consist of o
a minimum of three OPERABLE rhodium incore detector segments in 90 symmetric fuel assemblies.
APPLICABILITY: When the incore detection system is used for:
a.
Monitoring the AZIMUTHAL POWER TILT, b.
Recalibration of the excore neutron flux detection system, or c.
Monitoring the UNRODDED INTEGRATED RADIAL PEAXING FACTOR or the linear heat rate.
ACTION:
With the incore detection system inoperable, do not use the system for the above applicable monitoring or calibration functions.
The provisions of specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENT 4.3.3.2 The incore detection system shall be demonstrated OPERABLE:
a.
By performance of a CHANNEL CHECK within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to its use and at least once per 7 days thereafter when required for:
1.
Monitoring the AZIMUTHAL POWER TILT.
2.
Recalibration of the excore neutron flux detection system.
3.
Monitoring the UNRODDED INTEGRATED RADIAL PEAKING FACTOR or the linear heat rate.
b.
At least once per 18 months by performance of a CHANNEL CALIBRATION operation which exempts the neutron detectors but includes all electronic components. The neutron detectors shall be calibrated prior to installation in the reactor core.
MILLSTONE - UNIT 2 3/4 3-31 smandment No 45, 139
\\
E-INSTRUMENTATION SEISMIC INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.3 The seismic monitoring instrumentation channels.shown in Table 3.3-7 shall be OPERABLE.
APPLICABILITY: ALL MODES.
ACTION:
a.
With the number of OPERABLE seismic monitoring channels less than required by Table 3.3-7 restore the inoperable channel (s) to OPERABLE status within 30 days. The provisions of Speci-fications 3.0.3 and 3.0.4 are not applicable.
b.
With one or more seismic monitoring channels inoperable for more than 30 days,. prepare and submit a Special Report to the Commission pursuant tc Specification 6.9.'2 within the next 10 days outlining the cause of the sm1 function and the plans for restoring the system to OPERABLE status.
s SURVEILLANCE REQUIREMENTS 4.3.3.3 Each of the above seismic monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECX, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the fre-quencies shown in Table 4.3-4.
MILLSTONE - UNIT 2 '
3/4 3-32
- af
h__.
.....u 3/4.10 SPECIAL TEST EXCEPTIONS SHUTOOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of CEA worth and shutdown margin provided reactiv.ity equivalent to at least the highest estimated CEA worth is available for trip insertion from OPERABLE CEA(s).
APPLICABILITY: M'JDES 2 and 3.
ACTION:
With any full leng'th CEA not fully inserted and with less than the a.
a"bove reactivity equivalent available for trip. insertion, within 15 minutes initiate and continue boration at > 40 gpm of boric acid solution at or greater than the required refueling water storage tank (RWST) concentration (ppm) until tha SHUTDOWN MARGIN required
' by Specification 3.1.1.1 is restored.
b.
With all full length CEAs inserted and the reactor suberitical by less than the above reactivity equivalent, immediately initiate and continue boration at > 40*gpm of boric acid solution at or greater than the required refueling water storage tank (RWST) concentration (ppm) until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.
SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full length CEA required either partially or fully withdrawn shall be determined at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
4.10.1.2 Each.CEA not fully inserted shall be demonstrate'd capable of full insertion when tripped from at least the 50% withdrawn position within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1.
MILLSTONE - UNIT 2 3/4 10-1 Amendment No. 52 ST. 7?
e l
l L______________________._____l_.___
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, 3.1.3.5, 3.1.3.6, 3.2.3 and 3.2.4 may be suspended during the performance of PHYSICS 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 POWER 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 F
3.3.3.2 during PHYSICS TESTS above 5% of RATED THERMAL POWER in which the I
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.
l l
MILLSTONE - UNIT 2 3/4 10-2 Amendment No. 18,52,139
=____
a 3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 B0 RATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made suberitical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.
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 E0L, with T at no load odYr a. ting temperature, and is 9
associated with a postulated aM.eam line break accident and resulting uncontrolled RCS cooldown.
In the analysis of this accident, a minimum SHUTDOWN MARGIN of 3.60% Ak/k is initially required to control the reactivity transient. Accordingly, the SHUTDOWN MARGIN required by Specification 3.1.1.1 is based upon this limiting condition and is consistent with FSAR accident analysis assumptions.
For earlier periods during the fuel cycle, this value is conservative. With T s 200*F, the reactivity transients resulting from any postulated accident We minimal and a 2% Ak/k shutdown margin provides a
adequate-protection.
3/4.1.1.3 BORON DILUTION AND ADDITION l
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 :t least 3000 GPM will circulate an equivalent Reactor Coolant System volume of 10,060 + 700/-0 cubic feet in approximately 30 minutes.
The reactivity change l
rate associated with boron concentration changes will be within the capability for operator 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 accident and transient analyses remain valid through each fuel cycle.
The surveillance requirements for measurement of the MTC during each fuel cycle are adequate to confirm the MTC value since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.
The confirmation that the measured MTC value is within its limit provides assurance that the coefficient will be maintained within acceptable values throughout each fuel cycle.
MILLSTONE - UNIT 2 B 3/4 1-1 Amendment No. 139
REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1.5 MINIMUM TEMPERATURE FOR CRITICALITY The' MTC is expected to be slightly negative at operating conditions.
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 lower temperatures, this specification is provided to restrict reactor operation when T,yg is 'significantly below the normal' operating tempera-ture.
3/4.1.2 BORATION SYSTEMS The boron injection system ensures that negative reactivity control is available during each mode of facility operation.
The components required to perform this function include 1) borated water sources, 2) charging pumps, 3) separai.e flow paths, 4) boric acid pumps, and 5) an emergency power supply from OPERABLE diesel generators.
With the RCS average temperature above 200*F, a minimum of two separate and redundant 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 the Boric Acid Storage Tanks are achieved through Boric Acid Pumps, gravity feed lines and Charging Pumps.
Redundant flow paths from the Refueling Water Storage Tank are achieved through Charging Pump. flow path guaranteed by Technical Specification 3.1.2.2 and the HPSI flow path guaranteed by Technical Specification 3.5.2 and 3.5.3.
Allowable out-of-service periods ensure that minor component repair or corrective action may be completed without undue risk to overall facility safety from injection system failure.s during the repair period.
The minimum boration capability is sufficient to provide a SHUTDOWN MARGIN of 3.6% Ak/k at all temperatures above 200*F.
The maximum boration capability requirement occurs at E0L from full power equilibrium xenon conditions and requires an equivalent of 4900 gallons of 3.5% boric acid solution from the boric acid tanks plus 15,000 of 1720 ppm borated water from the refueling water storage tank.
The refueling water storage tank 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 minimum contained volume of 370,000 gallons of berated water in the refueling water storage tank ensures the capability for borating the RCS to the desired level.
The specified quantity of borated water is consistent with the ECCS requirements of Specification 3.5.4.
Therefore, the larger vclume of borated water is specified here too.
d MILLSTONE - UNIT 2 B 3/4 1-2 Amendment No. 38.33,139
s BASES 3/4.1.2 B0 RATION SYSTEMS (Continued)
The analysis to determine the boration requirements assumed that the Reactor Cooiant System is borated concurrently with cooldown.
In the limiting situation when letdown is not available, ti e cooldown is assumed to be f
initiated within 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> and cooldown to 200 F is completed in the next 28 l
hours.
With the RCS temperature below 200*F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity
(
condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable.
l The boron capability required below 700*F is based upon providing a 2% A k/k SHUTDOWN MARGIN at 140*F after xenon decay.
This condition requires either 3750 gallons of 2.5% boric acid solution from the boric acid tanks or 57,300 gallons of 1720 ppm borated water from the refueling water storage tank.
The maximum boron concentration requirement (3.5%) and the minimum temperature requirement (55'F) for the Boric Acid Storage Tank ensures that boron does.not precipitate in the Boric Acid System.
The daily surveillance requirement. provides sufficient assurance that the temperature of the tank will be maintained higher than 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 analysis assumptions for boron dilution incidents and other transients using the RWST as a borated water source as well as the analysis assumption to determine the boration requirement to ensure adequate shutdown margin.
3/4.1.3 MOVEABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power 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.
The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original criteria are niet.
The ACTION statements applicable to an immovable or untrippable CEA and to a large misalignment (120 steps) of two or more CEAs, require a prompt, shutdown of the reactor since either i
MILLSTONE - UNIT 2 B 3/4 1-3 Amendment no 38,67 d,116,33 139
i 1
~
BASES s
~
3/4.1.3 MOVEABLE CONTROL ASSEMBLIES (Continued) of these ccnditions may be indicative of a possible loss of mechanical functional capability of the CEAs and in the event of a immovable or untrippable CEA, the loss of :'iUTDOWN 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 3
LCOs and LSSS setpoints for DNBR and linear heat rate, 2) a small effect on the tir..e dependent long term power distributions relative to those used in generating LCOs and LSSS setpoints for DNBR and linear heat rate, 3) a small effect on the available SHUTDOWN MARGIN, and 4) a small effect on the ejected 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 may be made to 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 provided to protect the core in the event of a large misalignment (1 20 steps) of a CEA.
However, this misalignment would i
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 temperature) may not be sufficient to cause trips, it is possible that the reactor could be operating with process variah'es less conservative than those assumed in generating LCO and LSSS setpotats.
Therefore, the ACTION statement associated with the large misalignment of a CEA requires a prompt and significant reduction in THERMAL POWER prior to attempting realignment of the misaligned CEA.
The ACTION statements applicable to misaligned or inoperable CEAs include requirements to align the OPERABLE CEAs in a given group with the inoperable CEA.
Conformance 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 lead 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 MILLSTONE - UNIT 2 B 3/4 1-4 Amendment 'n. JE, JJJ,139
3/4.2 POWER DISTRIBUTION LIMITS BASES 3/4.2.I 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.
Either 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 Power Ratio Recorder are less than or equal to the limits of Figure 3.2-2.
In conjunction with the use of the excore monitoring system and in establishing the AXIAL SHAPE INDEX limits, the following assumptions are made:
- 1) the CEA insertion limits of Specifications 3.1.3.2, 3.1.3.5 and 3.1.3.6 are satisfied,
- 2) the flux peaking augmentation factors are as shown in Figure 4.2-1, 3) the AZIMUTHAL POWER TILT restrictions of Specification 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 measure of the peaking factors and the alarms which have been established for the individual incore detector segments ensure that the peak linear heat rates will 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 augmentation factors as shown in Figure 4.2-1, 2) a measure-ment-calculational uncertainty factor 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 THERMAL 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 FACTORS-F AND AZIMUTHAL r
POWER TILT - T q T
The limitations on F and T are provided to 1) ensure that the assump-l tions used in the analysis for % establishing 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 l
assumptions used in the analysis establishing the DNB Margin 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 l
limitations, operation may continue under the addit 5cnal 9 restrictions imposed by the ACTION statements since these addithal 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 LCOs and LSSS
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POWER DISTRIBUTION LIMITS BASES setpoints remain valid. An AZIMUTHAL POWER TILT > 0.10 is not expected and if it should occur, subsequent operation would be restrictr.d 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 cif F arHf T do not T
exceed the assumed values.
Verifying F after each fuel foading Srior to exceeding 75% of RATED THERMAL POWER prc[vides additional assurance that the core was properly loaded.
l 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, operation must be maintained within the speci-fied limits for the accident and transient analyses to remain valid.
MILLSTONE - UNIT 2 B 3/4 2-2 Amendment No. EE, JE, JJJ 139 m
3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 COOLANT LOOPS AND COOLANT CIRCULATION 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 MODES 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.
In MODES 4 and 5, a single reactor coolant loop or shutdown cooling loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require 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 shutdown cooling 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 with boron reductions will, therefore, be within the capability of operator recognition and control.
The restrictions on starting a Reactor Couant 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 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 Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into or (2) by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 43*F (31'F when measured by a surface contact instrument) above each of the RCS' cold leg temperatures.
3/4.4.2 SAFETY VALVES 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 to relieve 296,000 lbs per hour of saturated steam at the valve setpoint. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which 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. Ep, JJ, E9, 139
q REACTOR COOLANT SYSTEM RASES During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2750 psia.
The combined relit./ capacity of these valves is sufficient to limit the Reactor Coolant System pre!,ure to within its Safety Limit of 2750 psia following a complete loss of turbine generator load while operating at RATED THERMAL POWER and assuming no reactor trip until the first Reactor Protective System trip setpoint (Pressurizer Pressure-High) is reached (i.e., no credit is taken for a direct reactor toip on the loss of turbine) and also assuming no operation of the pressurizer' power operated relief valve or steam dump valves.
3/4.4.3 RELIEF VALVES The power operated relief valves (PORVs) operate to relieve RCS pressure below the setting of the pressurizer code safety valves.
These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable.
The electrical power for both the relief valves and the block valves is capable of being supplied from an emer-gency power source to ensure the ability to seal this possible RCS leakage path.
3/4.4.4 PRESSURIZER
~
An OPERABLE pressurizer provides pressure control for the reactor coolant systen during operations with both forced reactor coolant flow and with natural circulation flow.
The minimum. water level in the pressurizer assures the pressurizer heaters, which are required to achieve and maintain pressure control, remain covered with water to prevent failure, which occurs if the heaters are energized uncovered.'.The~ maximum water level in the pressurizer ensures that this paramter is maintained within the envelope of operation assumed in the safety analysis.
The maximum waten level also ensures that the RCS is not a hydraulically solid system and that a steam bubble will be pro-i vided to accommodate pressure surges during operation.
The steam bubble also protects the pressurizer code safety valves and power operated relief valve against water relief. The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reactor Coolant System pressure and establish and maintain natural circulation.
l The requirement that 130 kW of pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency bus
{
provides assurance that these heaters can be energized during a loss of off-site power condition to maintain natural circulation at HOT STANDBY.
j 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the stea i generator tubes ensure that the structural integrity of this portion of the RCS will be l
l maintained.
The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.
Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is MILLSTONE-UNIT 2 B 3/4 4-2 Anendment No. 22, 27 E2, 66,97
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3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRIT ensures that the release of I
radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses. This restriction, in conjunc-tion with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR 100
- during accident conditions.
3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure of 54 psig, P, leakage rate is further limited to 1 0.75 L As an added conservatism, the measured overall integrated during performance of the periodic tests to account for possi$le degra-dation of the containment leakage barriers between leakage tests.
The surveillance testing for measuring leakage rates are.
consistent with the requirements of Appendix "J" of 10 CFR 50, with the option of the use of the mass poinc method for perform-ing leakage calculations.
3/4.6.1.3 CONTAINMENT AIR LOCKS l
The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and leak rate given in Specifications 3.6.1.1 and 3.6.1.2.
The limitations on the air locks allow entry and exit into and out of the containment during operation and ensure through the surveillance testing that air lock leakage will not become excessive through continuous usage.
MILLSTONE - UNIT 2 B 3/4 6-1 i
I
CONTAINMENT SYSTEMS BASES 3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that the contain-ment peak pressure d)es not exceed the design pressure of 54 psig during LOCA conditions.
The maximum peak pressure obtained from a LOCA event is 53.8 psig.
The limit of 2.1 psig for initial positive containment pressure will limit the total pressuro to less than the design pressure and is consistent with the accident analyses.
3/4.6.1.5 AIR TEMPERATURE The limitation on containment air temperature ensures that t,'e contain-
, ment peak air temperature does not exceed the design temperature of 289'F
(
during LOCA conditions.
The containment temperature limit is consistent with the accident analyses.
3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment vessel will be maintained comparable to the original design standards for the life of the facility.
Structural integrity is required to ensure that the vessel will withstand the maximum pressure of 53.8 psig in the event of a LOCA.
The measurement of containment tendon lift off force, the visual and metallurgical examination of tendons, anchorages and liner and the Type A leakage tests are sufficient to demonstrate this capability.
The surveillance requirements for demonstrating the containment's struc-tural integrity are in compliance with the recommendations of Regulatory Guide 1.35 " Inservice Surveillance of Ungrouted Tendons in Prestressed Concrete Containment Structures."
L 1
HILLSTONE - UNIT 2 B 3/4 6-2 Amendment No. 13,72,139-
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M LLSTONE UNIT 2 5-3
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DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment buildir.' is designed and shall be maintained for a maximum internal pressure of 54 psi; and a temperature of 289'F.
l PENETRATIONS 5.2.3 Penetrations through the reactor containment building are designed and shall be maintained in accordance with the original design provisions contained in Section 5.2.8 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.
5.3 REACTOR CORE FUEL ASSEMBLIES l
5.3.1 The reactor core shall contain 217 fuel assemblies with each fuel assembly containing 176 rods.
Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 3.7 weight percent of U-235.
CONTROL ELEMENT ASSEMBLIES 5.3.2 The reactor core shall contain 73 full length and no part length control element assemblies.
The control element assemblies shall be designed and maintained in accordance with the original design provisions contained in Section 3.0 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.
5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed cad shall be maintained:
In accordance with the code requirements specified in Section 4.2.2 a.
of the FSAR with allowance for normal degradation pursuant of the applicable Surveillance Requirements, i
b.
For a pressure of 2500 psia, and I
f For a temperature of 650*F except for the pressurizer which is c.
700*F.
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?fc MILLSTONE-UNIT 2 5-4 Amendment No.35, 139
ATTACHMENT TO LICENSE AMENDMENT NO.139 FACILITY OPERATING LICENSE NO. DPR-65 DOCKET NO. 50-336 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.
The revised pages are identified by amendmert number and contain vertical lines indicating-the areas of change.. The corresponding overleaf pages are provided to raintain document completeness.
Remove Insert II II V
V XI XI 1-5 1-5 2-2 2-2 2-4 2-4 2-5 2-5 2-7 2-7 B2-1 B2-1 B2-3 B2-3 B2-5 B2-5 B2-6 B2 6 B2-7 B2-7 B2-8 B2-8 3/4 1-1 3/4 1-1 3/4 1-5 3/4 1-5 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2 3/4 2-3 3/4 2-3 3/4 2-3a 3/4 2-4 3/4 2 a 3/4 2-5 3/4 2-5 3/4 2-6 3/4 2-6 3/4 2-7 3/4 2-7 3/4 2-7a 3/4 2-8 3/4 2-8 3/4 2-9 3/4 2-9 3/4 2-10 3/4 2-10 3/4 2-14 3/4 2-14 3/4 3-4 3/4 3-4 3/4 3-30 3/4 3-30 3/4 3-31 3/4 3-31 3/4 10-2 3/4 10-2 B 3/4 1-1 B 3/4 1-1 B 3/4 1-2 B 3/4 1-2 B 3/4 1-3 B 3/4 1-3 B 3/4 1-4 B 3/4 1-4 B 3/4 2-1 B 3/4 2-1 B 3/4 2-2 B 3/4 2-2 B 3/4 4-1 B 3/4 4-1 B 3/4 6-2 B 3/4 6-2 5-4 5-4 i
t INDEX
{
DEFINITIONS 1
i t
E8El SECTION i
{
1 Enclosure Building Integrity................................
1-S' Reactor Trip System Response Time...........................
1-5 3
Engineered Safety Feature Response Time.....................
1-5 Physics Tests...............................................
1-6 Unrodded Iritegrated Radial Peaking Factor - F..............
1-6 r
Source Check........,.......................................
1-6 Radiological Effluent Monitoring and Offsite Dose Cal cul at ion Manual (REM 0DCM)..............................
1-6 Radioactive Waste Treatment Systems.........................
1-6 Purge - Piping..............................................
1-6 Venting.....................................................
1-8 Membe r( s) of the Publ i c.....................................
1 Site Boundary...............................................
.~1 Unrestricted Area...........................................
1-8 Storage Pattern.............................................
1-8 1
MILLSTONE - UNIT 2 II Amendment No. 79f, JJJ, JJJ 139
INDEX PEFIAITIONS SECTION PAGE 1.0 DEFINITIONS De f i n e d Te rns.................................................
1 - 1 Thermal Power.................................................
1-1 Ra ted The rma l P owe r....... ;...................................
1 - 1 Operational Mode..............................................
1-1 Action...........
3.........................s.................. 1-1 O p e ra bl e - O pe rab i l i ty........................................
1 Re p o rt a bl e Eve n t..............................................
1 - 1 Con ta i nmen t Integ ri ty.........................................
1 - 2 Channel Calibration.........
1-2 Ch a n n el Ch e c k.................................................
1 - 2 Channel Functi onal Tes t.......................................
1-2 Co re Al te ra ti o n...............................................
1 - 3 S h u td own Ma rg i n...............................................
1 - 3 Identi fi ed Leakage............................................
1-3 Uni den ti fi ed Leakage..........................................
1-3 Pressure Bounda ry Leakage.....................................
1-3 Controlled Leakage............................................
1-3 Az i mu th a l P owe r T11 t..........................................
1 - 4 Do s e Eq u i va l e n t I-131.........................................
1 - 4 E-Average Di si ntegrati on Energy...............................
1-4 S ta g g e red Tes t Bas i s..........................................
1-4 Frequency Notation............................................
1-4 Axial Shape Index.............................................
1-5 l
MILLSTONE - UNIT 2 I
Amendment No. 9,38,J94,111
INDEX LIMITING CONDITIONS FOR 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 l
TOTALINTEGRATEDRADIALPEAKINGFACTOR-Ff............
3/4 2-9 3.4.2.3 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 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION.....................
3/4 3-1 3.4.3.2 ENGINEERED SAFETY FEATURE ACTUATION 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 Meteorol ogical 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 Effluent 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 COOLANT CIRCULATION..................
3/4 4-1 Startup and Power 0peration............................
3/4 4-1 Hot Standby............................................
3/4 4-la Shutdown...............................................
3/4 4-lb MILLSTONE - UNIT 2 V
Amendment No. JE, JE, EE, pp,SS,195 139
INDEX 1
LIMITING CONDITIONS FOR OPERATION _AND SURVEILLANCE REQUIREMENTS j
1 SECTION
,P_Ag I
A 3/4.4.2 SAFETY VALVES............................................
3/4 4-2 l
3/4.4.3 RELI EF VALVES............................................ 3/4 4-3 l
3/4.4.4 FRESSuR!zER.............................................. 3/4 4-4 3/4.4.5 STEAM 8ENERATORS.........................................
3/4 4-5 3/4.4.6 REACTOR C0OLANT SYSTEM LEAXAGE...........................
3/4 4-8 J
Leakage Detection Systems................................
3/4 4-8 Reactor Coolant System Leakage...........................
3/4 4-9 3/4.4.7 CHEMISTRY................................................
3/4 4-10 3/4.4.8 SPECIFIC ACTIVITY........................................ 3/4 4-13 3/4.4.9 PRESSURE / TEMPERATURE LIMITS...................r. u....... 3[4 4-17 Reactor Cool ant System................................... 3/4 4-1)
Pres suri zer.......................................'....... 3/4 4-21 Overpressure Protection Systems..........................
3/4 4-21a 3/4.4.10 STRUCTURAL INTEGRITY.....................................
3/4 4-22 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 SAFETY INJECTION TANXS...................................
3/4 5-1, 3/4.5.2 ECCS SUBSYSTEMS - T,yg2,300*F...........................
3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - T,yg < 300*F...........................
3/4 5-7 3/4.5.4 REFUELING WATER STORAGE TANK............................. 3/4 5-8 MILLSTONE - UNIT 2 VI Amendment No. 59,72,104 l
INDEX
~
BASES SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 COOLANT LOOPS AND C0OLANT CIRCULATION....................
8 3/4 4-1
)
3/4.4.2 SAFETY VALVES................................'............
8 3/4 4-1 4
3/4.4.3 RELI EF YALYES............................................ 8 3/4 4-2 3/4.4.4 PRESSURIZER..............................................
8 3/4 4-2 1
3/4.4.5 STEAM GENERATORS.........................................
8 3/4 4-2 3/4.4.6 REACTOR C0OLANT SYSTEM LEAKAGE...........................
8 3/4 4-3
~
3/4.4.7 CHEMISTRY................................................
8 3/4 4-4 3/4.4.8 SPECI FIC ACTIVITY..............'.......................... S -3/4 4-4 3/4.4.9 PRESSUTi/ TEMPERATURE LIMITS..............................
8 3/4 4 "
3/4.4.10 STRUCTURAL INTEGRITY..................................... 8 3/4 4-7 l
3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 SAFETY INJ ECTION TANKS.................................... B 3/4 6-1 j
3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS...............................
8 3/4 5-1 3/4.5.4 REFUECING WATER STORAGE TANK (RWST)....................... B 3/4 5-2 3/4.6 CONTAINMENT SYSTEMS I
3/4.6.1 PRIMARY CONTAINMENT......................................
8 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS...................... B 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES.............................
8 3/4 6-3
{
3/4.. 4 C0.uSTatE GAS e0NTR0t..................................
8 3/4 6-4 3/4.6.5 SECONDARYCONTAINMENT................................C...B3/468 i
MILLSTONE - UNIT 2 XII Amendment No.56,5),72,104 4
m..
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m.
.l_N.0,EX BASES SECTION E_AqE A
3.4.0 APPLICABILITY.........................................
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 B0 RATION SYSTEMS..................................
B 3/4 1-2 l
3/4.1.3 MOVABLE CONTROL ASSEMBLIES........................
B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 LINEAR HEAT RATE..................................
B 3/4 2-1 3/4.2.2 Deleted 3/4.2.3 TOTALINTEGRATEDRADIALPEAKINGFACTOR-Ff.......
B 3/4 2-1 3/4.2.4 AZIMUTHAL POWER TILT.................
B 3/4 2-1 3/4.2.5 Deleted 3/4.2.6 DNB MARGIN.............-.
B 3/4 2-2 3/4.3 INSTRUMENTATION 3/4.3.1 PROTECTIVE INSTRUMENTATION........................
I 3/4 3-1 3/4 3.2 ENGINEERED SAFETY FEATURE INSTRUMENTATION.........
I 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION........................
I 3/4 3-2 MILLSTONE - UNIT 2 XI gendmentNo.7),/S,J9/
_ _ - _ _ -