ML20203N589
| ML20203N589 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 04/21/1986 |
| From: | Rubenstein L Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20203N591 | List: |
| References | |
| TAC-59902, NUDOCS 8605050528 | |
| Download: ML20203N589 (13) | |
Text
,
UNITED STATES O
NUCLEAR REGULATORY COMMISSION y
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ALABAMA POWER COMPANY DOCKET NO. 50-364 JOSEPH M. FARLEY NULLEAR PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 55 e
License No. NPF-8 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Alabama Power Company (the licensee) dated September 30, 1985, as supplemented March 27, 1986, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's
. rules and regulations set forth in 10 CFR Chapter I; E
The ficility will operate in conformity with the application, as
- ' amended, the provisions of the Act, and the regulations of the Commission; C.
There is reasonable assurance: (i) that the activities authorized by
~'
this amendment can be conducted without endangering the health and s
safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this license amendment will not be inimical to the common 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 e' '
o Lthe Commission's regulations and all applicable requirements have been satisfied.
^
2.
Accordin' gly, 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-8 is hereby amended to read as follows:
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8605050528 860421 PDR ADOCK 05000364 P
PDR A ^
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., - (2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. S5, 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 its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION S-Lula.a Lester S. R'uSenstein, Director PWR Project Directorate #2 Division of PWR Licensing-A Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: April 21,1986
ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. 55 FACILITY OPERATING LICENSE NO. NPF-8 DOCKET NO. 50-364 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages as indicated.
The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
The corres-ponding overleaf pages are also provided to maintain document completeness.
Remove Pages Insert Pages 3/4 4-29 3/4 4-29 3/4 4-30 3/4 4-30 B 3/4 4-6 B 3/4 4-6 8 3/4 4-7 8 3/4 4-7 B 3/4 4-8 8 3/4 4-8 8 3/4 4-9 B 3/4 4-9 8 3/4 4-10 B 3/4 4-10 B 3/4 4-14 B 3/4 4-14 L
0
REACTOR COOLANT SYSTEM BASES 3/4.4.8 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion.
Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System cver the life of the plant.
The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent.
Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State. Limits, up to the Transient Limits, for th2 specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System.
The time interval permitting continued operation within the restrictions of the Transient Limits pravides time for taking corrective actions to restore the contaminant concen-trations to within.the Steady State Limits.
The surveillance require'ments provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.
3/4.4.9 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure
..that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an
.Eppropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM.
The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations.
These values are conservative in that specific site parameters of tne Farley site, such as site boundary location and meteorological conditions, were not considered in this evaluation.
The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 1.0 cicrocuries/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occ'ur following changes in THERMAL POWER.
Operation with specific activity levels exceeding 1.0 microcuries/ gram DOSE EQUIVALENT I-131 but Within the limits shown on Figure 3.4-1 must be restricted to no more than 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> per year (approximately 10 percent of the unit's yearly operating time) since the activity levels allowed by Figure 3.4-1 increase the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid dose at the site boundary by a factor of up to 20 following a postulated steam generator tube rupture.
The reporting of cumulativ'e operating time over 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> in any 2 consecutive calendar quarters period with greater than 1.0 microcuries/ gram DOSE EQUIVALENT I-131 will allow sufficient time for Commission evaluation of the circumstances prior to reaching the 800 hour0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> limit.
FARLEY-UNIT 2 B 3/4 4-5
l REACTOR COOLANT SYSTEM BASES
-ESE to less than 500*F prevents the release of activity should a steam Reducing Tavo generator tube rupture since the saturation pressure of the primary coolant is The below the lift pressure of the atmospheric steam relief valves.
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 permissible if justified by the data obtained.
3/4.4.10 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 Yessel Code,Section III, Appendix G as required per 10CFR Part 50 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 f or the first full-power service pe riod, 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.
b) Figures 3.4-2 and 3.4-3 define limits to assure prevention of nonductile f ailure only.
For normal operation, other inherent plant characteristics, e.g., purp heat addition and pressurizer heater capacity, rey limit the heatup and cooldewn 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 tenperature of the steam generator is below 70*F.
\\
FARLEY-UNIT 2 B 3/4 4-6 AMENDftENT NO. 55
3 REACTOR COOLANT SYSTEM BASES 4)
The pressurizer heatup and cooldown rates shall not exceed 100*F/hr and 200*F/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 preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.
The fracture toughness properties of the ferritic materials in the reactor
. vessel are determined in accordance with ASTM E185-82, and in accordance with additional reactor vessel requirements. These properties are then evaluated in accordance with Appendix G of the 1976 Summer Addenda to Section III of the ASME Boiler and Pressure Vessel Code and the calculation methods described in WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves, April 1975."
.Heatup and cooldown limit curves are calculated using th,e most limiting value of the nil-ductility reference temperature, RTndt, at the end of 8 effective full power years of service life. The 8 EFPY service life period
-is chosen such that the limiting RTndt at the 1/4T location in the core region is greater than the RTndt of the limiting unirradiated mater.ial.
The selection of such a limiting RTndt 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 RTndt; the results of these tiasts are shown in Table B 3/4.4-1.
Reactor operation and resultant f ast neutron (E greater than 1 MEV) irradiation can cause an increase in the RTndt. Theref ore, an adjusted reference 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 RTndt at the end of 8 EFPY.
FARLEY-UNIT 2 B 3/4 4-7 AMEN 0 MENT NO. 55
REACTOR COOLANT SYSTEM BASES determined in this manner may be used until the next Values of ARTndt results from the material surveillance program, evaluated according to ASTM E185, are available. Capsules will be removed in accordance with the requirements of ASTM E185-82 and 10 CFR 50, Appendix H.
The surveillance specimen withdrawal schedule is shown in Table 4.4-5.
The heatup and cooldown curves must be recalculated when the A RTndt determined f rom the next surveillance capsule exceeds the calculated ARI ndt 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 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 WC AP -7924 -A.
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 surf ate 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 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 capabilicies of inservice inspection techniques.
Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient saf ety margins for protection against non-ductile f ailure.
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 shif t, ARTndt, corresponding to the end of the period for which heatup and cooldown curves are generated.
s i
FARLEY-UNIT 2 B 3/4 4-8 AMEN 0 MENT NO. 55
1 TABLE D3/4.4-1 p
U REACTOR VESSEL TOUGilNESS DATA Si G
Average Upper N
Shelf Energy Normal to Principal Principal Working Working Cu P
NI T
EI Direction Direction NOT NOT Component Code No, Grade L11 (M
LK1
( *[},
(*[},
(ft-lb)_
(ft-lb)
CL. 110. Dome 87215-1 A533,8,CL.1 0.17 0.010 0.49
-30 16(a) 83(a) jpg CL.110. Flange 87201-1 A508,CL.2 0.14 0.011 0.65 60(a) 60(a)
>56(8)
>86(C) vtS. Flange 87206-1 A508,CL.2 0.10 0.012 0.67 60(3) 60(a)
>7)(a)
>109 0.010 0.68 50(8) 50(a) to3(a) j$g E
Inlet Noz.
87218-2 A508,CL.2
)
Inlet Noz.
87218-1 A508,CL.2 0.010 0.71 32(3) 32(a) 112(a) 172 inlet Noz.
87218-3 A500,CL.2 0.010 0.12 60(3) 60(a) 98(a) 150 E
Outlet Noz.
87217-1 A508,CL.2 0.010 0.73 60(a) 60(a) 100(a) 154 e
Outlet Moz.
87217-2 A50b,CL.2 0.010 0.12 6(a) 6(a) 108(a) 167 Outlet Noz.
87217-3 A500,CL.2 0.010 0.12 4u(a) 4g(a) 103(a) j$g Upper Shell 81216-1 A508,CL.2 0.010 0.13 30 30(a) gy(a) 349 140 Inter Shell 87203-1 A533,8,CL.1 0.14 0.010 0.60
-40 15 99 Inter Shell 87212-1 A533,8,CL.1 0.20 0.018 0.60
-30
-10 99 134 Lower Shell 87210-1 A533,8,CL.1 0.13 0.010 0.56
-40 18 103 120 Lower Shell 87210-2 A533,8,CL.1 0.14 0.015 0.57
-30 0
99 145 Trans. Ring 81208-1 A508,CL.2 0.010 0.13 40 40(a) 89(a) 13; Bot. HD. Dome 87214-1 A533,8,CL.1 0.11 0.007 0.48
-30
-2(a) 87(a) 134 Inter. Shell A1.46 SMAW 0.02 0.009 0.96 0(3)
O(a)
>j33 long Seams A1.40 SMAW 0.02 0.010 0.93
-60
-60
>106 Inter Shell to lower Shell G1.50 SAW 0.13 0.016
<.20(b)
-40
-40
>102 tower Shell g
tong Seams G1.39 SAW 0.05 0.006
<.20(h)
-70
-70
>126 E
S9 5
(a) [ stimate per NUREG 0000 "U5NRC Standard Review Plan Branch Technical Position Mif 8 5-2.
a (h) Estimated.
g (c) Upper shelf not available. value represents minimum energy at the highest test ternperature.
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tu Ab 20 25 30 35 FAP. LEY-UNIT 2 B3/4 4-10 AMENDMENT NO. 55
-~
REACTOR COOLANT SYSTEM BASES 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 tensile stresses produced by internal pressure. The metal temperature at the for the 1/4T crack crack tip lags the coolant temperature; therefore, the KIR f r the 1/4T crack during steady-state during heatup is lower than the KIR 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
's for steady-state and finite heatup rates IR 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 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 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 less'er of the three values taken from the curves.under consideration.
FARLEY-UNIT 2 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 rang the controlling condition switches f rom the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.
Finally, the 10 CFR Part 50, Appendix G Rule which addresses the metal temperature of the closure head flange and vessel flange must be considered.
This Rule states that the minimum metal temperature of the closure flange regions be at least 120'F higher than the limiting RTndt for these regions when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (621 psig for Farley Unit 2).
In addition, the new 10 CFR Part 50 Rule states that a plant specific fracture evaluation may be performed to justify less limiting requi reme nts.
Based upon such a f racture analysis for Farley Unit 2, the 8 EFPY heatup and cooldown curves are impacted by the new 10 CFR Part 50 Rule as shown on Figures 3.4-2 and 3.4-3.
Although the pressurizer operates in temperature ranges above those for wMeh
%here 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 requirerents.
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 f ron pressure transients which could exceed the limits of Appendix G to 10CFR Part 50 when one or more of the RCS cold legs are less than or equal to 310'F.
Either RHR relief valve has adequate relieving capability to protect the RCS from overpressurizaticn 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 50*F above the RCS cold leg temperatures or (2) the start of 3 charging pumps and their injection into a water solid RCS.
3/4.4.11 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 lif e 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 10CFR Part 50.55a(g) /except where specific written relief has been granted by the Commission pursuant to 10CFR Part 50.55a(g)(6)(1).
3/4.4.12 REACTOR VESSEL HEAD VENTS The OPERABILITY of the Reactor Head Vent System ensures that adequate core cooling can be maintained in the event of the accumulation of non-condensable gases in the reactor vessel.
This system is in accordance with 10CFR50.44(c )(3 )(iii ).
FARLEY-UNIT 2 B 3/4 4-14 AMENDMENT NO. 3E, 55 1
MamTAL Poort-v Gisis CONTRCt.LI)G MATERIAL :
R. V. INTT.D C IATE SET:L COPPER COhTENT
- 0.20 kT5 PHOSPMORUS C0tiTENT
- 0.0L8 k'T%
INITIAL RT
- -10 7 g
RT AFTER 8 m
- 1/4T, 145 F g
- 3/47, 83 7 CUMT.S APPLICABLE FOR HE.AWP RATES UP 70 60*F/HR FOR Ti.E SEr/ ICE PERIOD UP 70 8 m nu.e i
1 1
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FARLEY-UNIT 2 3/4 4-29 AMENCMENT NO. 55
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- 0.20 hi%
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IsolcAfte TinptaAtuas cats.F FIGURE 3.(-3 FARLEY UNIT 2 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE FOR THE FIRST B EFFY FARLEY-UN:7 2 3/4 4-30 AMENDMENT NO. 55 wm