ML20215E785
| ML20215E785 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 12/11/1986 |
| From: | Thadani A Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20215E782 | List: |
| References | |
| NUDOCS 8612230120 | |
| Download: ML20215E785 (16) | |
Text
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((
UNITED STATES g
NUCLEAR REGULATORY COMMISSION
-l WASHINGTON, D. C. 20655
'...../
I MAINE YANKEE ATOMIC POWER COMPANY DOCKET NO. 50-309 MAINE YANKEE ATOMIC POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 91 License No. DPR-36 1.
The Nuclear Regulatory Comission (the Comission) has found thet:
A.
The application for amendment by Maine Yankee Atomic Power Company (thelicensee),datedJanuary 29, 1986, as supplemented by letters dated July 29 and August 28, 1986, complies with the standards and re as amended (the Act)quirements of the Atomic Energy Act of 1954,
, and the Commission'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 issuance 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 anendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
0612230120 861211 PDR ADOCK 05000309 P
. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment,andparagraph2.B(6)(b)ofFacilityOperatingLicense No. DPR-36 is hereby amended to reod as follows:
(b) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.
, 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 its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
~f GA hOv' W Ashok. Thadani, Director PWR P oject Directorate #8 Division of PWR Licensing-B
Attachment:
Changes to the Technical Specifications Date of Issuance:
December 11, 1986
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4 ATTACHMENT TO LICENSE AMENDMENT NO. 91 FACILITY OPERATING LICENSE NO. DPR-36 DOCKET N0. 50-309 Revise Appendix A as follows:
Remove Pages Insert Pages 4
4 3.6-1 3.6-1 3.11-4 3.11-4 3.14-1 3.14-1 3.14-3 3.14-3 3.15-1 3.15-1 3.22-2 3.22-2 3.24-1 3.24-1 4.1-8 4.1-8 4.2-2 4.2-2 4.4-2 4.4-2 4.6-2 4.6-2 5.8-1 5.8-1
MISCELLANEOUS DEFINITIONS Operable A system, subsystem, train, component or device shall be operable or have operability when it is capable of performings its specified function (s).
Implicit in this definition shall be the assumption i
that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, sub-system, train, component or device to perform its function (s) are also capable of performing their related support function (s).
Operating A system or component is operating if it is performing its safeguard or operating functions.
Control Element Assemblies All full-length shutdown and regulating control element assemblies (CEA's).
Partial-Length Control Element Assemblies Control element assemblies (CEA) that contain neutron absorbing material only in the lower quarter of their length.
Fire Suppression Water System A fire suppression water system shall consist of: A water source (s);
gravity tank (s) or pump (s); and distribution piping with associated sectionalizing control or isolation valves.
Such valves shall include yard hydrant curb valves, and the first valve ahead of the water flow alarm device on each sprinkler, hose standpipe or spray system riser.
Amendment No.$1,6$, 91 3.6 ' EMERGENCY CORE COOLING AND CONTAINMENT SPRAY SYSTEMS Applicab11ity:
Applies to the operating status of the emergency core cooling and containment spray systems.
Objective:
To define the conditions under which components of the emergency core cooling and containment spray systems must be operable.
Specification:
A.
The following equipment must be operable wher.ever the reactor coolant system temperature and pressure exceed 210 F and 400 psig:
1.
Two safety injection tanks set for automatic initiation.
Each tank shall contain 11,200 + 500 gallons of water borated to at least 1720 ppm and pressurTzed with nitrogen to 230 psig + 10 psi, - 25 psi.
2.
One operable ECCS train consisting of the following subsystems of the train.
Each subsystem includes the manual valves that are aligned and locked in the position required for safeguards operation, the automatically operated valves set for automatic operation or aligned and locked in the position required for safeguards' operation, the controls set for automatic initiation in accordance with Specification 3.9, and a pump powered from l
an engineered safeguards bus, a.
One service water pump subsystem b.
One component cooling pump subsystem c.
One low pressure safety injection pump subsystem d.
One high pressure safety injection pump subsystem f.
One containment spray pump and RHR heat exchanger subsystem 3.
Station service power in accordance with Technical Specification 3.12.A supplying the same operable ECCS train as in (2) above.
4.
The refueling water storage tank and spray chemical addition tank are filled and available in accordance with Technical Specification 3.7.
5.
The fill header motor operated root valves to two non-isolated loops.
Exception: The requirements nay be modified with regard to the position of controls and valves during periods of hydrostatic testing.
Remedial Action:
Restore required limiting condition within four hours.
Amendment No. 59,55, 91 3.6-1
Remedial Action:
If the containment weight of air monitoring system is out of service for more than ten days with the reactor critical, the Commission must be notified of plans to restore the system operability.
2.
When the containment weight of air monitoring system indicates a daily air loss greater than the following, an evaluallun slidll Le f alli4 Led to deteraisine Llie validity of the indication, a.
Equivalent to 0.15 weight percent per day at 50 psig for seven consecutive days, or b.
Equivalent to 0.5 weight percent per day at 50 psig for four consecutive days, or c.
Equivalent to 1.0 weight percent per day at 50 psig for three consecutive days.
3.
The reactor shall be made subcritical within six hours if the evaluation required by G2:
a.
Results in identification of the source of the leak and a determination that the known containment leak rate exceeds the equivalent of 0.10 weight percent per day at 50 psig through the containment integrity boundary and cannot be isolated, or b.
Fails to identify the source of the leakage within ten days and the containment weight of air monitoring system indication persists at an average rate in excess of 0.15 weight percent per day at 50 psig.
Basis Specification A includes a limit of 210*F on reactor coolant tempera-ture assures that no steam will be generated in the unlikely event of a reactor coolant system rupture and hence no driving force to release any fission products from the containment. The shutdown margins are selected based upon the types of activities that are being carried out.
The higher value for refueling precludes criticality under all postulated incidents.
Specificatinn B assures that the containment pressure boundary is defined while pennitting maintenance of components necessary to integrity.
Specification 4.4 requires that the uncontrolled containment leakage conform to specified limits to assure that public exposure will be main-tained well within the guidelines presented in 10 CFR 100 for the
(
hypothetical accident described in Section 14.18 of the FSAR.
3.11-4 Amendment No. 65, 72, 8t/, 91 t
3.14 PRIMARY SYSTEM LEAKAGE Applicability:
Applies to limiting operation of the plant under varying rates and condi-tions of primary system leakage.
Objective:
l To specify primary plant operability with primary system leakage.
Specification:
A.
When the reactor is above 2% power, two reactor coolant leak detection systems of different operating principles shall be operating, with one of the two systems sensitive to radioactivity in the containment.
l l
Remedial Action:
If two reactor coolant leakage detection systems are i
operable but neither is sensitive to radioactivity, a system sensitive l
to radioactivity must be made operable within 43 hours4.976852e-4 days <br />0.0119 hours <br />7.109788e-5 weeks <br />1.63615e-5 months <br />.
B.
The reactor coolant system indicated leak rate shall be limited to t
I gpm or less.
l Remedial Action:
If the indicated leak rate should exceed 1 gpm, within four hours investigate the source and assess safety implications.
C.
Reactor coolant system leakage shall not exceed any of the Specifica-tions 1 through 5 below.
1.
Leakage into the reactor containment of any magnitude that has been determined to be an indication of a deterioration of primary
~
systen pressure boundary strength welds or material.
2.
Leakage into the reactor containment in excess of I gpm through bolted closures, valve packing, or other mechanical connections.
l 3.
Leakage in excess of 1 gpm that is unexplained or unaccounted for.
4.
Leakage in excess of 10 gpm to aerated or uncontained systems.
5.
Total leakage through all steam generator tubes shall not exceed l
1.0 gpm.
l Remedial Actions:
1.
If the leakage specified in C.1 above has been determined to be a deterioration of primary system pressure boundary strength welds or material, then the provisions of Specification 3.0. A.2 and 3 apply.
2.
If reactor coclant system leakage exceeds any of the Specifications C.2 through C.5 above, the reactor shall be shut down within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
l Amendment No. $$,53,$5,$$, g1 3.14-1
where:
e C = Secondary coolant sample activity 0.1 micro ci/cc = 0.1 l
Ci/MJ V = Water volume in three steam generator = 131 M3 at standard conditions B (t) = Breathing rate (3.47 x 10-4m3 sec)
/
X/Q = 6.48 x 10-4 sec/m3 (corresponding to Pasquill F stability and 1 m/sec wind speed)
DCF = 1.48 x 106 rem /Ci I-131 inhaled The resulting thyroid dose is less than 1.5 rem.
s Amendment No. fl. 91 3.14-3
i r
I f
3.15 REACTIVITY ANOMALIES Applicability:
Applies to potential reactivity anomalies.
Objective:
To require evaluation of reactivity anomalies within the reactor.
Specification:
Following a normalization of the computed boron concentration as a function of burnup, the actual boron concentration of the reactor coolant shall be periodically compared with the predicted value.
The difference between the observed and predicted steady-state concentrations shall be less than the equivalent of 1% of reactivity.
Remedial Action:
The Nuclear Regulatory Commission shall be notified and an evaluation as to the cause of the discrepancy shall be made and reported to the Nuclear Regulatory Connission in accordance with Technical Specifica-tion 5.9.1.7.
Basis:
To eliminate possible errors in the calculations of the initial reactivity of the core and the reactivity depletion rate, the predicted relation between fuel burnup and the boron concentration, necessary to maintain adequate control characteristics, must be adjusted (normalized) to accurately reflect actual core conditions.
When full power is reached initially, and with the CEA groups in the desired positions, the boron concentration is measured and the predicted curve is adjusted to this point. As power operation proceeds, the measured boron concentration is compared with the predicted concentration and the slope of the curve relating burnup and reactivity is compared with that predicted. This process of normalization should be completed after about 10% of the total core burnup. Thereafter, actual boron concentration can be compared with prediction and the reactivity status of the core can be continuously evaluated, and its occurrence would be thoroughly investigated and evaluated. The methods employed in calculating the reactivity of the core vs. burnur, and the reactivity worth of boron vs. burnup, are given in the FEAR.
Amendment No. 19,$$,65,91 3.15-1
4 The system valves are aligned to provide flow to each steam generator following system actuation upon low steam generator water level signal from any one of the three steam cenerators. However, for a steam generator depressurization event, such as a steam line break, receipt of a low steam generator pressure signal initiates closure of the control and isolation valve (s) feeding the depressurized steam generator (s). This limits excessive reactor coolant system cooldown and the resultant reactivity insertion produced by excessive feedwater flow to a depressurized steam-generator. Flow will continue to steam generators remaining pressurized.
Flow to a depressurized steam generator will be reestablished by reopening the control and isolation valves after repressurization e.g., by isolation from the steam line break.
Operability of the system assures that the reactivity attributable to reactor coolant system cooldown due to feedwater addition to steam generators after a main steam line break is within the limits established in the steam line break safety analysis.
If the feedwater trip system is discovered to la inoperable, the best course of action is to restore its operability prompdy, thus avoiding challenges to plant systems that result from perturbing steady state operation. A two-hour time period presents low risk of a main steam line break yet allows enough time for deliberate restoration of system operability through maintenance actions.
If operability cannot be restored the reactor must be shut down. Six hours provides ample time for an orderly controlled shutdown.
If operability cannot be restored by that time, the reactor coolant system must be borated to hot shutdown concentration within an additional six hours. Twelve hours pemits an orderly shutdown while assur'ing that the risk of a main steam line break during the period is very low.
The intended function of the feedwater trip system can be accomplished under conditions of partial system inoperability provided all main feedwater system l
pumps and valves tripped by the system which are operating can be tripped by the operable portions of the trip system. Pumps which cannot be tripped by the trip system due to partial trip system inoperability can be shut down to assure functional capability.
When the reactor coolant system is at hot shutdown boron concentration, the steam line break cooldown cannot cause sufficient reactivity insertion to cause a return to critical, so the feed trip system is not required to function.
Amendment No. $2, ##, 91 3.22-2
I' 3.24 SECONDARY COOLANT ACTIVITY Applicability:
Applies to measured maximum activity in the secondary coolant system.
Obiective:
To ensure that the secondary coolant activity does not exceed a level commensurate with the safety of the plant personnel and the public.
Specification:
The specific activity of the secondary coolant system shall be less than or equal to 0.10 micro Ci/ gram DOSE EQUIVALENT I-131.
Basis:
The limitations on secondary system specific activity insure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in a steam line rupture. This dose includes that contributed by a 10 spm primary to secondary tube leak in the l
steam generator of the affected steam line.
}loje: The secondary coolant activity surveillance requirements are given in Table 4.2-1 Item 7.
Amendment No. 38,4/(,91 3.24-1
l Table 4.1-2 (Continued)
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Surveillance Channel Description Function Frequency Surveillance Method s
o 8.
Manual Containment Isolation Test R
Manual switch test.
5 Initiation 9.
Manual Initiation Containment Test R
Manual switch operation.
g Spray 3
- 10. Refueling Water Tank Level Test R
Fluid removed from level transmitters RAS Initiation to verify actuation of valves.
- 11. Safety Injection Tanks Level
- a. Check 5 (3)
- a. Verify level and pressure.
and Pressure
- b. Calibrate R
- b. Known pressure and differential pressure applied to pressure and level sensors.
T'
- a. Check S (3)
- a. Compare four independent steam Circuits generator pressure indications.
- b. Calibrate R
- b. Sinulated. signal. applied to meter relays to verify trip points, logic operation, solenoid valve operation.
- 13. High Pressure Safety Injection
- a. Check M (3)(4)
- b. Calibrate R
- b. Known pressure applied to sensors.
- 14. Low Pressure Safety Injection
- a. Check M (3)(4)
- b. Calibrate R
- b. Known pressure applied to sensors.
Table 4.2-1 Minimum Frequencies for Sampling Tests Test Frequency
- 1. Reactor Coolant Gross Activity Determination At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Samples Isotopic Analysis for DOSE EQUIVALENT I-131 concentra-tion.
I per 14 days Radiochemical for T Determination 1 per 6 months (a)
Isotopic Analysis for todine, Until the specific activity including I-131 I-133, and of the primary coolant system I-135.
is restored within the limits, once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, whenever the DOSE EQUIVALENT I-131 exceeds 1.0 uC1/ gram, and One sample between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a THERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a one-hour period.
Chemistry (chloride 3 times per week andfluoride)
Hydrogen Weekly (d) 0xygen Prior to heatup above 250*F, and 3 times per week when above 250'F.
Process Radiation Continuous (b)(d)
Monitor
- 2. Reactor Coolant Boron Concentration 3 times per week Boronometer Continuous (b)(d)
- 3. Refueling Water Boron Concentration Monthly I')I9)
Tank Water Sample i
t Amendment No. M,91 4.2-2
2.
Type C containment leakage rate tests will be performed in accordance with the provisions of 10 CFR 50.54(o) and 10 CFR 50 Appendix J.
This requirement is effective following the Cycle 8/9 refueling outage.
It will be implemented as practicable prior _ to that time.
3.
The Type 8 air lock tests will be performed in accordance with the provisions in 10 CFR 50.54(o) and Appendix J, at a test 3ressure of not less than 50 psig. The allowable leak rate for tie air lock test is 30 lbm per day or 0.05 L,.
4.
Containment purge supply, exhaust and bypass valve testing require-ments for on line purging are as follows:
a)
The containment purga supply, exhaust and bypass valves will be leak tested within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the termination of on-line purge, except when the valves are being used for multiple cyclings, then at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, b)
The containment purge supply, exhaust and bypass valves will be leak tested at six month intervals.
Basis A leakage rate value of 0.10 weight percent per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> will, under the most adverse design basis accident conditions, maintain public exposure well below 10 CFR 100 values in the event of the hypothetical accident. The tests at reduced pressure will assure the continued ability of the containment to perfom its function.
The air lock leak test is a measure of the operability.of the air lock, however, l
the air lock is still considered to be operable in the event of exceeding air -
lock leak test acceptance criteria as long as one hatch is determined to be properly closed and sealed in accordance with Technical Specification 3.11 and totalcontainmentleakagefromallpenetrationsandisolationvalves(including the air lock) has been evaluated to be within L.
p L in pounds mass per day = (0.6)(0.1%) (609,000 lbm) = 365 lbm per p
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 50 psig.
The seal is determined to be operable by an appropriate pressure test.
References:
FSAR Sections 5 and 14.18 10 CFR 50.54. Paragraph (o), and 10 CFR 50, Appendix J. Reactor Containment Testing Requirements Amendnent No. 38/12/82/ 91 4.4-2
One LPSI purr and one CS pump shall be flow tested at 100 psi discharge he 2.
During these tests flow distribution thru the HPSI and LPSI flow i
orifices will be checked.
Acceptance performance shall be that the pumps and orifices attain flow values used in the safety analysis.
Alternate pumps will be tested at each refueling interval, so that all pumps will be tested within any five year period.
b.
ECCS Valves:
All automatically operated valves and the motor operated fill header root valvas shall be exercised through their full travel in conjunction with the actuation signal testing set forth in Table 4.1-2 of Technical Specifications.
c.
Safety Injection Tanks:
Each safety injection tank will be flow tested by opening the tank isolation valve sufficient to verify check valve operation.
d.
The correct position of each electrical and mechanical position stop for the following throttle valves shall be verified:
- 1) Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following completion of maintenance on the valve when the HPSI system is required to be operable.
- 2) At least once per 4 months Valve Numbers HSI-M-11 HSI-M-12 HSI-M-21 HSI-M-22 HSI-M-31 HSI-H-32 c.
A flow balance test, as described in 4.6.A.2 above, shall be performed during shutdown to confirm the injection flow rates assumed in the Safety Analyuis following completion of HPSI or LPSI system modifica-tions that alter system flow characteristics.
f.
ECCS Check Valves The check valve barriers defined in Technical Specification 3.19.A.4 shall be determined to be intact by leak testing.
l Amendment No. Y7.11,71,71, 91 4.6-2
=
5.8 PROCEDURES 5.8.1 liritten procedures shall be established, implemented and maintained covering *. lie activities referenced below:
a.
The applicable pr cedures recommended in Appendix "A" of Regulatory Guide 1.33, (Rev. 2), February 1978.
l b.
Refueling operations.
c.
Surveillance and test activities of safety related equipment.
d.
Security Plan implementation.
e.
Emergency Plan implementation, f.
Fire Protection Program implementation.
- 5. 0. 2 Each procedure of 5.0.1 above, and changes thereto, shall be reviewed by the PORC and approved by the Plant Manager prior to implementation and reviewed periodically as set forth in administrative procedures.
5.8.3 Temporary changes to procedures of 5.8.1 above may be made provided:
a.
The intent of the original procedure is not altered.
b.
The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License.
c.
The change is documented, reviewed by the PORC and approved by the Plant Manager within 14 days of implementation.
Amendment No. #,59,77, 91 5.8-1