ML19325E089

Proposed Tech Specs Re Reactor Vessel Pressure/Temp Limits

ML19325E089
Person / Time
Site:	Millstone
Issue date:	10/20/1989
From:	NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML19325E088	List: ML19325E087 ML19325E089
References
NUDOCS 8911010172
Download: ML19325E089 (8)

Text

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INDEX 1 BASES o ,

SECTION f%QE TABLE B 3/4.4-1 REACTOR VESCEL FRACTURE TOUGHNESS PROPERTIES...... B 3/4 4-9 1 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E*1MeV) 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 t

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 b/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 9 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 l

51k 3/4.7.5 ULTIMATE HEAT SINK........................................

3/4.7.6 FLOOD PROTECTION..........................................

B 3/4 7-3 B 3/4 7-4 l.D 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM................. B 3/4 7-4 L

h 3/4.7.8 CONTROL ROOM ENVELOPE PRESSURIZATION SYSTEM............... B 3/4 7-4

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o< 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM.......................... B 3/4 7-4 "a. 3/4.7.10 SNUBBERS.................................................. B 3/4 7-5 l mo con a. .

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TABLE 4.4-5' ~

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h REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WIT)0RAWAL SCHEDULE -

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$ CAPSULE VESSEL LEAD

. NUMBER LOCATION FACTOR WITHDRAWAL TIME (EFPY) f 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 w

1

! a) Plant specific evaluation

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REACTOR COOLANT SYSTEM BASES l

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, i

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 ,

3200F, and

5. System preservice hydrotests and inservice leak and hydrotests shall be performed at pressures in accordant.e with the requirements of ASME Boiler' and Pressure Vessel Code, Section_XI.

The . fracture toughness testing of the ferritic materials in the reactor vessel were performed in accordance with the 1973 Summer Addenda to Section !!!

of the ASME Boiler and Pressure Vessel Code. These properties are then evaluated in accordance-with the NRC Standard Review Plan.

Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RT NDT, at the end of 10 Effective full power years (EFPY) of service life. The 10 EFPY service life period is chosen such that the limiting RT NDT at the 1/4T location in the core region is greater than the RTNDT of the limiting unirradiated material.

The selection of such a limiting RT NDT assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

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-ti$T;and resultant neutron irradiation can cause an increase in the RT 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 Eactor 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 adjustments for possible errors in theYressure N

and temperature sensing instruments.

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hEACTOR COOLANT SYSTEM l

L BASES I PRESSURE / TEMPERATURE LIMITS (Continued)

Values of ART NDT determined in this manner may be used until the results ,

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 NDT determined from the surveillance capsule exceeds the calculated ART NDT for the equivalent capsule radiation exposure.

Allowable pressure-temperature relationships for various heatup and l

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 i paragraphs.

The general method for calculating heatup and cooldown limit curves is '

based upon the principles of the linear elastic fracture mechanics (LEFM) technology. In the calculation prccedures a semic111ptical surface defect with a depth of one-quarter of the wall thickness, T, and a length of 3/2T is assumed to exist at the inside of the vessel wall as well as at the outside of L the vessel wall. The dimensions of this postulated crack, referred to in 1 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.

L To assure that the radiation embrittlement effects are accounted for in the i calculation of the limit curves, the most limiting value of the nil-ductility reference temperature, RTNDT, is used and this includes the radiation-induced l shift, ARTHDT, corresponding t the end of the period for which heatup and cooldown curves are generated.

l.

1 The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, j Kg , for the combined thermal and pressure stresses at any time during heatup i

or cooldown cannot be greater than the reference stress intensity factor, K for the metal temperature at that time. K IR is obtained from the referedEe, I fracture toughness curve, defined in Appendix G to the ASME Code. The K IR l curve is given by the equation:

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MILLSTONE - UNIT 3 B 3/4 4-11 l

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kEACTOR' COOLANT SYSTEM BASES I

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PRESSURE /TEMPERATURELIMITS(Ccntinued)  !

i KIR = 26.78 + 1.223 exp (0.0145(T-RTNDT + 160)] (1)  ;

Where: K IR is the reference stress intensity factor as a function of the gtal temperature T and the metal nil-ductility reference temperature RTNDT' us, the governing equation for the heatup-cooldown analysis is:

CKgg + kit iK IR (2)

Where: Kgg = 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 NDT f the material, 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 correspondirg thermal stress intensity factor, K IT, for the reference flaw 'is computed. From Equation (2) the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.

C00LDOWN  !

For the calculation of the allowable pressure versus coolant temperature during cooldown, the Code reference fisw 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.

Allowabic 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 IR at the 1/4T l location i l

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, itEACTOR COOLANT SYSTRi ,

BASES' PRESSURE / TEMPERATURE LIMITS (Continued)

.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. .

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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.

HEATUP Three separate calculations are required to determine the limit curves -

for finite heatup rates. As is done in the cooldown analysis, allowable 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 2 tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K for the 1/4T crack during conditionsheatup is lower at the same than coolant the K@ature.for temp During heatup, the 1/4T cracdRduring especially at the steady end of the transient, conditions may exist such that the effects of compressive heatup rates do thermal stresses not offset eachand different other and the Kyhressure

's for 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.

'The second portion of the heatup analysis concerns the calculation of f 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, l 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 stresses, of course, are dependent on both 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.

i MILLSTONE - UNIT 3 8 3/4 4-13

itEACTOR COOLANT SYSTEM BASES' l

PRESSURE / TEMPERATURE LIMITS (Continued) j Following the. generation of pressure-temperature curves for both the '

steady-state and finite heatup rate situations, the final limit curves are i produced as follows. A composite curve is constructed based on a -

point-by-point comparison of the steady-state and finite heatu) rate data. At 1

.any given temperature, the allowable pressure is taken to be tie 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 exist such that over the course of the heatup ramp the controlling condition switches from the inside i to the outside and the pressure limit must at all times be based on analysis I of the most critical criterion, j 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 i there is reason for concern of nonductile failure, . operating limits are '

provided to assure compatibility of operation with the fatigue analysis I 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 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 PORY Setpoint which can occur as a result of time delays 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.

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.,. REACTOR COOLANT SYSTEM BASES 3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2, 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)(1).

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 ani Addenda through Winter except i where specific written relief has been granted pursuant to 10 CFR 50.55a(g)(6)(1).

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 vent path from the reactor vessel head and the pressurizer steam space ensures '

that the capability exists to perform this function. i The valve redundancy of the Reactor Coolant System vent paths serves to minimize the probability of inadvertent or irreversible actuation while 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.

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