ML20012C125
| ML20012C125 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 03/06/1990 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Northeast Nuclear Energy Co (NNECO) |
| Shared Package | |
| ML20012C126 | List: |
| References | |
| NPF-49-A-048 NUDOCS 9003200148 | |
| Download: ML20012C125 (14) | |
Text
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. h, UNITED STATES.
3 NUCLEAR REGULATORY COMMISSION wAsMiwoToN. D. C. 20656
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NORTHEAST. NUCLEAR. ENERGY. COMPANY..ET.AL.
DOCKET.NO. 50-423 MILLSTONE. NUCLEAR POWER. STATION. UNIT NO. 3 1
AMENDMENT.T0. FACILITY.0PERATING. LICENSE Amendment No. 48 License No. NPF,
L 1.
The Nuclear Regulatory Comission (the Comission) has found that:
The app (lication for amendment by Northeast Nuclear Energy Company, A.
et al. thelicensee)datedOctober 20, 1989 complies 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 'n conformity with the application, the h
provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by
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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 issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public;-and l
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 L
of the Comission's regulations and all applicable requirements have been satisfied.
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l 900320014e 900306 PDR ADOCK 05000423 Mp P
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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. NPF-49 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.
48, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date of its issuance, to tp implemented within 30 days of issuance.
F0 THE NUCLEAR LAT RY COMMISSION JohiT.Stolz, Director Pr jec't Directorate'I-4 ision of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical
-Specifications Date of Issuance:'
March 6, 1990
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ATTACWNENT.TO LICENSE. AMENDNENT.N0_. 48.
FACILT!Y.0PERATING-LICENSE.NO..NPF-49 DOCKET.NO. 50 423 Replace the following pages of the Appendix A-Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. The corresponding overleaf pages are provided to maintain document completeness.
Remove Insert xiv xiv 3/4 4-36 3/4 4-36 B3/4 4-B B3/4 4-B B3/4 4-11 B3/4 4-11 83/4 4-12 B3/4 4-12 B3/4 4-13 B3/4 4-13 B3/4 4-14 B3/4 4-14' B3/4 4-15 B3/4 4-15 B3/4 4-16 j
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INDEX
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BASES I
SECTION PAGE 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 3/4.1.3 MOVABLE CONTROL ASSEMBLIES................................
B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS...................................
B 3/4 2-1 3/4.2.I' AXIAL FLUX DIFFERENCE.....................................
B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CFANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR.........
B 3/4 2-2 FIGURE B 3/4.2-la TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL POWER FOR FOUR LOOP OPERATION.....................
B 3/4 2-3 FIGURE B 3/4.2-lb TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL POWER FOR THREE LOOP 0PERATION....................
B 3/4 2-4 3/4.2.4 QUADRANT POWER TILT RATI0.................................
B 3/4 2-6 3/4.2.5 DNB PARAMETERS............................................
B 3/4 2-7 3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATION and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION...........................................
B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION................................
B 3/4 3-3 3/4.3.4 TURBINE OVERSPEED PROTECTION..............................
B 3/4 3-7 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION.............
B 3/4 4-1 3/4.4.2 SAFETY VALVES.............................................
B 3/4 4-2 3/4.4.3 PRESSURIZER...............................................
B 3/4 4-2 3/4.4.4 RELIEF VALVES.............................................
B 3/4 4-2 3/4.4.5 STEAM GENERATORS..........................................
B 3/4 4-3 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE............................
B 3/4 4-4 3/4.4.7 CHEMISTRY.................................................
B 3/4 4-S 3/4.4.8 SPECIFIC ACTIVITY.........................................
B 3/4 4-S 3/4.4.9 PRESSURE / TEMPERATURE LIMITS...............................
B 3/4 4-7 MILLSTONE - UNIT 3 xiii I
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BASES SECTION EASE TABLE B 3/4.4-1 REACTOR VESSEL FRACTURE TOUGHNE05 PROPERTIES......
B 3/4 4-9 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E*1Mei) AS A FUNCTION OF FULL POWER SERVICE LIFE..................................
B 3/4 4-10 3/4.4.10 STRUCTURAL INTEGRITY.....................................
B 3/4 4-15 3/4.4.11 REACTOR COOLANT SYSTEM VENTS.............................
B 3/4 4-15 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS..............................................
B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS...............................
B 3/4 5-1 3/4.5.4 REFUELING WATER STORAGE TANK..............................
B 3/4 5 2 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT.......................................
B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS......................
B 3/4 6-2 3/4.6.3 CONTAINMENT ISOLATION VALVES..............................
B 3/4 6-3 3/4.6.4 COMBUSTIBLE GAS CONTR0L...................................
B 3/4 6-3 3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM....................
B 3/4 6-3 3/4.6.6 SECONDARY CONTAINMENT.....................................
B 3/4 6-4 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE.............................................
B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION...........
B 3/4 7-3 3/4.7.3 REACTOR PLANT COMPONENT COOLING WATER SYSTEM..............
B 3/4 7-3 3/4.7.4 SERVICE WATER SYSTEM......................................
B 3/4 7-3 3/4.7.5 ULTIMATE HEAT SINK........................................
B 3/4 7-3 3/4.7.6 FLOOD PROTECTION..........................................
B 3/4 7-4 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM.................
B 3/4 7-4 3/4.7.8 CONTROL ROOM ENVELOPE PRESSURIZATION SYSTEM...............
B 3/4 7-4 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM..........................
B 3/4 7-4 3/4.7.10 SNUBBERS..................................................
B 3/4 7-5 MILLSTONE - UNIT 3 xiv Amendment No. 48
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MATERi&L PeopCRiv SAsis CONTROLLING MATERi&L PLATE METAL CO*PER CONTENT CONSERVAf avC' Y A55vwE0 TO SE 010 wt *s.
PwC5P=CRV5 CONTENT
. 0 010wf't.
RT INITIAL 60*F NOT RT AMER 10 Epf
.iMT.122'F NOT 34 T.101'F CURVE APPLICABLE FOR C00LDowN RATES UP TO 100*F/MR FOR TME SERvCE PERIOD UP TO 10 EFPY AND CONTAIN5 MARGIN 5 0F 10*F AND to P5lG FOR P055lSLE INSTRUMENT ERROR $
3000 0 20000 1
+
r i
10000 a'
f C00LOOwN RATES.-
(* F/MR) 20 _
'e r0 s
100 -
-00 00 100 0 2000 300 0 4000 5000 INDICATED TEMPERATURE (CEG F)
FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS - APPLICABLE UP TO 10 EFPY MILLSTONE - UNIT-3 3/4 4-35
TABLE 4.4-5 x.
.P REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITIORAWAL SCHEDULE:
'G Uj'i CAPSULE VESSEL LEAD NUNBER LOCATION
[ ACTOR WITHDRAWAL TIME (EFPY) b U
58.5*
3.98(a)
First Refueling (1.3 EFPY actual)
Y 241*
3.74 9
Y 61*
3.74 16 W
121.5*
4.01 STANDBY.
X 238.5*
4.01 STANDBY Z
301.5*
4.01 STANDBY M.
a) Plant specific evaluation M
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REACTOR C00LAN.T SYSTEM I.:
BASES SPECIFIC ACTIVITY (Continued)
Based upon the above considerations for excluding certain radionuclides from the sample analysis, the allowable time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> between sample taking and completing the initial analysis is based upon a typical time necessary to perform the sampling, transport the sample, and perform the analysis of about 90 minutes.
After 90 minutes, the gross count should be made in a reproducible geometry of sample and counter having reproducible beta or gamma self-shielding properties.
The counter should be reset to a reproducible efficiency versus energy.
It is not necessary to identify specific nuclides.
The radiochemical determination of nuclides should be based on multiple counting of the sample within typical counting basis following sampling of less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, about I day, about 1 week, and about 1 month.
Reducing T to less than 500*F prevents the release of activity should asteamgeneratN9 tube rupture since the saturation pressure of the reactor coolant is below the lift pressure of the atmospheric steam relief valves.
The Surveillance Requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action.
A reduction in frequency of isotopic analyses following power changas may be permissible if justified by the data obtained.
3/4.4.9 PRESSURE / TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G.
Also, the 10 CFR 50, Appendix G rule which addresses the metal temperature of the closure head flange and vessel flange regions is considered.
This rule states the minimum metal temperature of the closure flange regions should be at least 120'F higher than the limiting RT f r these regions when the pressure exceeds 20% of the NDT preservice hydrostatic test pressure (636 psia). The minimum ~ temperature of the closure flange and vessel flange regions is 150*F since the limiting RT is 30 F (See Table B 3/4.4-1).
The heatup curve shown in Figure 3.4-2 is NDT not impacted by the 10 CFR 50 rule.
However, the cooldown curve shown in Figure 3.4-3 is impacted by the rule.
1.
The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3 for the service period specified thereon:
a.
Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown.
Limit lines for cooldown rates between those presented may be obtained by interpolation; and b.
Figures 3.4-2 and 3.4-3 define limits to assure prevention of I
non-ductile failure only.
For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater MILLSTONE - UNIT 3 8 3/4 4-7
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j REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) ca)acity, may limit the heatup and cooldown rates that can be ac11eved over certain pressure-temperature ranges.
2.
These limit lines shall be calculated periodically using methods provided
- below, 3.
The secondary side of.the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 700F, 4.
The pressurizer heatup and cooldown rates shall not exceed 1000F/h and 2000F/h, respectively.
The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than L
3200F, and 5.
System preservice hydrotests and inservice leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and. Pressure' Vessel Code,Section XI.
l The fracture toughness testing of the ferritic materials in the reactor vessel were performed in accordance with the 1973 Summer Addenda to Section III of the ASME Boiler and Pressure Vessel Code. These properties are then evaluated in accordance with the NRC Standard Review Plan.
l Heatup and cooldown-limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RT at the end of i
The '10 EFPY service life period is chosen such that the limiting RT at the 1/4T location in i
NDT the core region is greater than the RT of the limiting unirradiated material.
NDT The selection of such a limiting RT assures that all components in the NDT Reactor Coolant System will be operated conservatively in accordance with p
applicable Code requirements.
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The reactor vessel materials have been tested to determine their initial RT the results of these tests are shown in Table B 3/4.4-1.
Reactor opera-tihT;and resultant neutron irradiation can cause an increase in the RT E
Therefore, an adjusted reference temperature, based upon the fluence, cohr.
content, and nickel content of the material in question, can be predicted using Figure B 3/4.4-1 and the largest value of ART computed by Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Ector Vessel Materials."
The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RT at the end of 10 EFPY as well as l
adjustments for possible errors in the7ressure and temperature sensing N
l instruments.
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MILLSTONE - UNIT 3 B 3/4 4-8 Amendment No. 48 j
-BASES L
PRESSURE /TEMPERATURELIMITS(Continued)
Values of ART determined in this manner may be used until the results NDT from the material surveillance program, evaluated according to ASTM E185, are available.
Capsules will be removed in accordance with the requirements of ASTM E185-73 and 10 CFR Part 50, Appendix H.
The surveillance specimen with-drawal schedule is shown in Table 4.4-5.
The lead factor represents the rela-tionship between the fast neutron flux density at the location of the capsule and the inner wall of the reactor vessel.
Therefore, the results obtained from the surveillance specimens can be used to predict future radiation damage to the reactor vessel material by using the lead factor and the withdrawal time of the capsule.
The heatup and cooldown curves must be recalculated when the ART determined from the surveillance capsule exceeds the calculated NDT ART for the equivalent capsule radiation exposure.
' Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Sec-tion III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50, and these methods are discussed in detail in the following paragraphs.
l The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) i l
technology.
In the calculation procedures a semielliptical surface defect I
with a depth of one-quarter of the wall thickness, T, and a length of 3/2T is l
assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall.
The dimensions of this ' postulated crack, referred to in Appendix G of ASME Section.III as the reference flaw, amply exceed the current capabilities of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against nonductile failure.
To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil-ductility reference temperature,.RTNDT, is used and this includes the radiation-induced shift, ARTNDT, correspondirig to the end of the period for which heatup and L
cooldown curves are generated.
The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K, for the combined ~ thermal and pressure stresses at any time during heatup g
or cooldown cannot be greater than the reference stress intensity factor, K i
for the metal temperature at that time.
K is obtained from the referebbe, 1
IR fracture toughness curve, defined in Appendix G to the ASME Code.
The K IR curve is given by the equation:
MILLSTONE - UNIT 3 8 3/4 4-11 Amendment No. 48
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REACTOR COOLANT SYSTEM BASES PRESSURE /TEMPERATURELIMITS(Continued)
KIR = 26.78 + 1.223 exp (0.0145(T-RTNDT + 160)]
(1)
Where:
K is the reference stress intensity factor as a function of the IR gtal' temperature T and the metal nil-ductility reference temperature RTNDT' us, the governing equation for the heatup cooldown analysis-is:
CKgg + kit IK (2)
IR Where:
Kjg = the stress intensity factor caused by membrane (pressure)
- stress, kit = _the stress intensity factor caused by the thermal gradients, KIR = constant provided by the Code as a function of temperature relative to the RT of the material, NDT C = 2.0 for level A and B _ service limits,- and C = 1.5 for inservice hydrostatic and leak test operations.
At any time during the heatup or cooldown transient, K is determined by IR the metal temperature at the tip of the postulated flaw, the appropriate value for RTNDT, and the reference fracture toughness curve.
The thermal stresses resulting from temperature gradients through the vessel wall are calculated and then the corresponding thermal stress intensity factor, K for the IT, reference flaw is computed.
From Equation (2) the pressure stress intensity factors are obtained and, from thcse, the allowable pressures are calculated.
COOLDOWN For the calculation of the allowable pressure versus coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall.
During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates.
Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate - situations.
From these relations, composite limit curves are constructed for each cooldown rate of interest.
-The use of the composite curve in the cooldown analysis-is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel ID. This condition, of course, is not true for the steady-state situa-tion.
It follows that at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of K at the 1/4T IR location MILLSTONE - UNIT 3 B 3/4 4-12 Amendment N0. 48
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REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)
L for finite cooldown rates than for steady state operation.
Furthermore, if conditions exist such that the increase in K exceeds Kit, the calculated allowable pressure during cooldown will be heater than the steady-state value.
The above procedures are needed because there is no direct control on temperature at the 1/4T location; therefore, allowable pressures-may unknowingly be violated if the rate ' of cooling is decreased at various intervals along a cooldown ramp.
The use of the composite curve eliminates this problem and-assures conservative operation of the system for the entire cooldown period.
HEATVP Three separate calculations are required to determine the limit curves for finite heatup rates.
As is done in the cooldown analysis, allowable l
pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the vessel wall.
The thermal gradients during heatup produce compressive stresses at the inside of the wall that alleviate the l
tensile stresses produced by internal pressure.
The metal temperature at the crack tip lags the coolant temperature; theref' ore, the K f r the 1/4T crack during heatup is lower than the K for the 1/4T cracERduring steady-state conditions at the same coolant temphature.
During heatup, especially at the end of the transient, conditions may exist such that the effects of l
compressive thermal stresses ' and different Kg 's for heatup rates do not offset each other and the hressure steady-state and finite temperature curve based on steady-state conditions no longer represents a lower bound of all-similar curves for finite heatup rates when the 1/4T flaw is considered.
Therefore, both cases have to be analyzed in order to assure that at any coolant-temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.
1 1
The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed.
Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present.
These thermal stresces, of course, are dependent on both l
the rate of heatup and the time (or coolant temperature) along the heatup ramp.
Furthermore, since the thermal stresses at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined.
Rather, each heatup rate of interest must be analyzed on an individual basis.
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MILLSTONE - UNIT 3 8 3/4 4-13 Amendment No. 48 i*
q REACTOR COOLANT SYSTEM-f BASES PRESSURE / TEMPERATURE LIMITS (Continued)
Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the' final limit curves are produced as follows.
A composite curve is constructed based on a
point-by point comparison of the steady-state and finite heatup rate data.- At any given temperature, the allowable pressure is takei to be the lesser of the three values taken from the curves under consideration.
The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to axist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.
Finally, the composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves.
Although the pressurizer operates in temperature ranges above those for which there is reason for concern of nonductile failure, operating limits are
'provided to assure compatibility of operation with the fatigue analysis performed.in accordance with the ASME Code requirements.
COLD OVERPRESSURE PROTECTION The OPERABILITY of two PORVs or an RCS vent opening of at least 5.4 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 3500F.
Either PORV has adequate t
relieving capability to protect the RCS from overpressurization when the transient is limited to either:
(1) the start of an idle RCP with the secondary water temperature of the steam generator.less than or equal to 500 above the RCS cold temperatures, or (2) the start of a charging pump and its injection into a water-solid RCS.
The Maximum Allowed PORV Setpoint for the Cold Overpressure Protection System (COPS) is derived by analysis which models the performance of the COPS assuming various mass input and heat input transients.
Operation with a PORV Setpoint less than or equal to the maximum Setpoint ensures that Appendix G criteria will not be violated with consideration for a maximum pressure overshoot beyond the PORV Setpoint which can occur. as a result of time delays I
in signal processing and valve opening, instrument uncertainties, and single l
failure. To ensure that mass and. heat input transients more severe than those assumed cannot occur, Technical Specifications require lockout of all but one safety injection pump and all but one centrifugal charging pump while in MODES 4, 5, and 6 with the reactor vessel head installed and disallow start of an RCP if secondary temperature is more than 500 above primary temperature.
The Maximum Allowed PORV Setpoint for the COPS will be updated based on the results of examinations of reactor vessel material irradiation surveillance specimens performed as required by 10 CFR Part 50, Appendix H, and in accordance with the schedule in Table 4.4-5.
MILLSTONE - UNIT 3 B 3/4 4-14 Amendment No. 48 i
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BASES 3/4.4.10~ STRUCTURAL INTEGRITV L
L.
-The inservice inspection and testing programs for ASME Code Class 1, 2, l
and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant.
These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i).
Components of the Reactor Coolant System were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 80 Edition and Addenda through' Winter except where specific written relief has been granted pursuant to 10 CFR 50.55a(g)(6)(i).
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3/4.4.11 REACTOR COOLANT SYSTEM VENTS Reactor Coolant System vents are provided to exhaust noncondensible gases and/or steam from the Reactor Coolant System that could inhibit natural circulation core cooling. The OPERABILITY of least one Reactor Coolant System i
vent path from the reactor vessel head and the pressurizer steam space ensures that the capabi-lity exists to perform this function.
The valve redundancy of the Reactor Coolant System vent paths serves to minimize the probability of inadvertent or irreversible actuation while j-ensuring that a single failure of a vent valve, power supply, or control system does not prevent isolation of the vent path.
The function, capabilities, and testing requirements of the Reactor Coolant System vents are consistent with the requirements of Item II.B.1 of NUREG-0737, " Clarification of TMI Action Plant Requirements," November 1980.
i MILLST'ONE - UNIT 3 B 3/4 4-15 Amendment No. 48 m
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