ML19341A200

Proposed Changes to Tech Specs 3.4.9 & 4.4.9 Re Heatup & Cooldown Limit Curves

ML19341A200
Person / Time
Site:	Farley
Issue date:	01/09/1981
From:	ALABAMA POWER CO.
To:
Shared Package
ML19341A198	List: ML19341A197 ML19341A199 ML19341A200
References
NUDOCS 8101220365
Download: ML19341A200 (13)

Text

.

I OReA'CTORC00tANTSYSTEM . .

3/4.4. 9 PRESSURE / TEMPERATURE LIMITS

. REACTOR COOLANT SYSTEM -

LIMITING CONDITION FOR 0PERATION ] '

3. 4. 9 .1 The Reactor Coolanc System (except the pre'ssurizer) temperature 'and pren.ure shall be limited in accordance with the limit lines shown on

. Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice .

leak and hydrostatic testing with- -

I

a. A maximum heatup of 100 F in,any one hour period. . . ,.

b. A maximum cooldown of 100 F in any one hour period. -

~

c. A maximum temperature change of less than or equal to 10*F in any one hour period during inservice hydrostatic and leak testing operations above the heatup and ccoldown limit curves.

APPLICABILITY: At all times. .

ACTION: ,

i F -

With any of the above limits exceeded, -restore the temperature and/or pressure <

to withirt the limit with#r 30 minutes; perform an engineering evaluation or inspection to determine the effects of the out-of-limit condition on the fracture toughness of the Reactor Pressure Vessel; determine that the Reactor

' Pressure Vessel remains acceptable for continued operation or 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 /> and reduce the RCS T and pressure to lessthan200Fand500psig,respectively,withinthef8Y?owing30 hours.

SURVEILLANCE REQUIREMENTS 4.4.9.1,1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per hour during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in mattirial properties, as required by 10 CFR 50, Appendix H in accordance with the schedule in Table 4.4-5. The results of these examinations shall be used to update Figures.3.4-2 and 3.4-3.  :

810122f365 m, , ,y , -

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'FARLEY-UNIT 1 3/4 ~ 4-25

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MATERI AL PROPERTY BASIS CONTROLLING M ATE RI AL: WELD METAL  ;

COPPER CONTENT: 0.24 WT%

PHOSPHORUS CONTENT: 0.011 WT%

0 RTNDT INITIAL: 0F .  ;

RTNDT AFTER 7.7 EFPY: 1/4T, 1850F 3/4T 1180F CURVE APPLICABLE FOR HEATUP RATES UP TO 600F/HR FOR THE SERVICE ACCEPTABLE PERIOD UP TO 7.7 EFPY AND CONTAINS MARGINS OF 100F AND 60 PSIG FOR REGION FOR 3000 POSSIBLE INSTRUMENT ERRORS HYDR 0 STATIC TESTING OPERATIONS ,

6 f LEAK TEST LIMIT E

a UNACCEPTABLE -

h n.

2000 -

OPERATION 8

HEATUP RATES ACCEPTABLE iE UP TO 600 F/HR OPERATION 1000 -

f "' ' ^

r BASED ON

_ INSERVICE HYDROSTATIC TEST .-

TEMPERATURE (3250F) FOR THE SERVICE PERIOD UP TO 7.7 EFPY 0

0 100 200 300 400 500 INDICATED TEMPERATU RE (OF)

Ficure 3,4-2 Farley Unit 1. Reactor Coolant System Heatup Limitations Applicable For The First 7.7 EFPY .

FARLEY-UNIT 1 3/4 4-26 e

l 000 MATERIAL PROPERTY BASIS CONTROLLING MATERIAL: WELD METAL i COPPER CONTENT: 0.24 WT%

PHOSPHORUS CONTENT: 0.011 WT%

RT NDT INITIAL: 0 F ,

RT NDT AFTER 7.7 EFPY: 1/4T,1850F 3/4T, 1180F CURVE APPLICABLE FOR COOLDOWN RATES UP TO 100 F/HR FOR THE SERVICE PER!OD UP TO 7.7 EFPY AtiD CONTAINS MARGINS OF 100F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS 2000 -

.,,f. .

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!b.

E D

_ M UNACCEPTABLE OPEPATION E

c.

8 H

5 s

3 1000 -

ACCEPTABLE OPERATION COOLDOWN RATES OF/HR ,;

O N

20 %

40 -

60 100 0

0 100 200 300 400 INDICATED TEMPERATURE ( F)

Figure 3.4-3Farley Unit 1 Reacto Coolant System Cooldown Limitations Applicable For The First 7.7 EFPY FARLEY-UNIT 1 3/4 4-27

TABLE 4.4-5 ,

h REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM-WITHDRAWAL SCHEDULE Z -

VESSEL LEAD .

CAPSULE LOCATION FACTOR , WITHDRAWAL TIME Y 3430 3.5 1st Refueling Outage .

U 1070 3.5 3 EFPY. .

X 2870 3.5 6 EFPY

~

W 1100 2.9 11 EFPY R

[ V 2900 2.9 20 EFPY k

Z 3400 2.9 STBY e

9 4

- 9

REACTOR COOLANT SYSTEM BASES Reducing T,yg to less than 500 F prevents the release pf activity should c steam generator tube rupture since the saturation pressure of the primary c':olant is below the lift pressure of the atmospheric steam relief valves.

The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissib'e 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.

- 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 -

a) Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of 4he limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation.

b) Figures 3.4-2 and 3.4-3 define limits to assure prevention of non'-

ductile failure only. For normal operation,.other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved 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 70 F.

O FARLEY-UNIT 1 8 3/4 4-6 .

_ REACTOR C00LAf1T SYSTIM BASES

4) The pressurizer heatup and cooldown rates shall not exceed 100 F/ hr and 200 /hr respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320 F.

5) System and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of AS!!E Boiler and Pressure Yessel Code,Section XI.

The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with ASTM E185-73, and in accordance with additional reactor vessel requirements. These properties are then evaluated in accordance with Appendix G of the 1976 Summer Adden'da to Section III of the ASME Boiler and Pressure Vessel Code. .

Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNDT, at the end of 7.7 effective full power years of service life. The 7.7 EFPY service F-! life period is chosen such that the limiting RT NDT at the 1/4T location in the core region is greater thar. the RTf1DT f the limiting unirradiated material. The selection of such a limiting RT assures that all t1DT 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 RTNDT; the results of these tests are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron (E greater than 1 MEV) irradiation can cause an increase in the RTt1DT. Therefore, an adjusted reference l temperature, based upon the fluence and copper content of the material in question, can be predicted using Figure B 3/4.4-1 and the recommendations of Regulatory Guide 1.99, Revision 1, " Effects of Residual Elements on Predicted Radiation Damage to Reactor 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 NDT at the end of 7.7 EFPY (as well as adjustments for possible errors in the pressure and temperature sensing instruments).

FAlLEY-Ut1IT 1 B 3/4 4-7

. REACTOR COOLANT SYSTEM BASES 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 require-ments of ASTM E185-73 and 10 CFR 50, Appendix H. The surveillance specimen withdrawal schedule is shown in Table 4.4-5. The heatup and

. coolcown curves must be recalculated when the ARTNDi determined from the surveillance capsule exceeds the calculated ART NDT f r the equivalent capsule radiation exposure. ,

Allowable press.ure -temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in -

Section 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, g 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 procedures a semi-elliptical surface defect with a depth of one quarter of the wall thickness, T, und a length of 3/2T is 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 non-ductile 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, ARTHDT, c rresponding to the end of the period for which heatup and cooldown curves are generated FARLEY-UNIT 1 B 3/4 4-8

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FARLEY-UllIT 1 B 374 4-10 1

i REACTOR COOLANT SYSTEM ,  !

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BASES

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The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K3 , for the combined thermal and pressure stresses at any time during heatup cr cooldown cannot be greater than the reference stress intensity factor, K IR' fcr the metal tecperature at that time. k* is obtained from the reference  ;

.IR fracture toughness curve, defined in Appendix G to the ASME Code. The K IR curve is given by the equation: l KIR = 26.78 + 1.223 exp [0.0145(T-RTNDT + 160)] (1) ,,

where K IR is the ref.erence stress intensity factor as a function of the metal temperature T and the metal nil ductility reference temperature RTl4DT. Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

~

CKyg + kit IKIR (2)

Where, K yg is the stress intensity factor caused by membrane (pressure) stress.

K It is the stress intensity factor caused by the thermal gradients.

K IR is provided by the code as a function of temperature relative to the RT NDT f the material.

1 C = 2.0 for level A cnd B service limits, and C = 1.5 for inservice hydrostatic and leak test operations.

At any time during the heautp or cooldown transient, K IR is determined by l the metal temperature at the tip of the postulated flaw, the' appropriate I value for RTilDT, and the reference fracture toughness curve. The thermal stresses resulting from temperature gradients through the vessel wall are '

FARLEY-UllIT 1 B 3/4 4-11 .

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REACTOR COOLANT SYSTEM BASES

. calculated and then the corresponding thermal stress intensity factor, KIT' fer 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 flaw is assumed to exist at the inside of the vessel wall. During cooldown, the controll;ng 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 be-cause 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 situation.

It follows that at any given reactor coolant temperature, the delta T developed ,

during cooldown results in a higher value of K IR at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist such that the increase in KIR exceeds kit, the calculated allowable pressure during cooldown will be greater 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. j l

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FARLEY-UNIT I B 3/4 4-12 . ,

=

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(tEACTOR COOLANT SYSTEM BASES HEATUP f

_ Three sen ,te 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 '.iall. The thermal gradients during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K f r the 1/4T crack IR during heatup is lower than the K f r the 1/4T crack during steady-state IR conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and different K IR 's for steady-state and finite heatup rates p do not offset each other and the pressure-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 concern's the calculation of p

'ressure-temperature limitttions for the case in which a 1/4T deep outside surface flaw is assumed. 1)nlike 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 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.

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 taken to be the lesser of the three values taken from the curves under consideration.

FARLEY-UNIT 1 B 3/4 4-13 '

REACTOR COOLANT SYSTEM BASES '

~

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 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 non-ductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

The OPERABILITY of two RHR relief valves or an RCS vent opening of greater

~

than or equal to 2.85 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 310 F.

Either RHR ief valve has adequate relieving capability to protect the RCS from overpr ' ation when the transient is limited to either (1) the start k'~

of an idle RL. with the secondary water temperature of the steam generator less than or equal to 50 F above the RCS cold leg temperatures or (2) the start of 3 charging pumps and their injection into a water solid RCS.

3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASNE 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 wit 5 Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by

- 10 CFR Part 50.55a(g) except where specific written relie.' has been granted by the Commission pur:" ant to 10 CFR Part 59.55a (g) (6) (i).

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