ML19345G001

Amend 52 to License DPR-36,changing Tech Specs & Implementing Program to Reduce Leakage from Sys Outside Containment That Could Contain Radioactive Fluids During Serious Accident

ML19345G001
Person / Time
Site:	Maine Yankee
Issue date:	02/06/1981
From:	Clark R Office of Nuclear Reactor Regulation
To:
Shared Package
ML19345F994	List: ML19345F993 ML19345G001 ML19345G005
References
NUDOCS 8102190706
Download: ML19345G001 (13)

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. UNITED STATES NUCLEAR REGULATORY COMMISSION g

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MAINE YANKEE ATOMIC POWER COMPANY DOCKET NO. 50-309 MAINE YANKEE ATOMIC POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 52 License No. DPR-36 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Maine Yankee Atomic Power Company, (the licensee) dated September 11, 1980, 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 f acility 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 (i) 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 conpliance 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 amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

810219o 700

2-2.

Accordingly, Facility Operating License No. DPR-36 is hereby amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and by revising and adding i

sections to paragraph 2.B.(6) as follows:

(b) Technical Specifications i

The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 52, are i

hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical

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Specifications.

(e)

Systems Integrity The licensee shall implement a program to reduce leakage from systens outside containment that would or.could contain highly radioactive fluids during a serious transient or accident to as low as practical levels.

(f)

Iodine Monitoring i

The licensee shall implement a program to ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions.

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION t

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Robert A. Clark, Chief Operating Reactors Branch #3-Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: February 6, 1981

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.l ATTACHMENT TO LICENSE AMENDMENT NO. 52 i

j TO FACILITY OPERATING LICENSE NO. DPR-36 DOCKET N0. 50-309 Revise Appendix A as follows:

Remove Page Insert Page 4

3.3-1 3.3-1 i

~ 3.3-2 3.3-2 i

3.3-3 3.9-1 3.9-1 3.9-2 3.9-2 2

3.9-5 4.1-11 4.1-11 4.1-12 i

4.2-5 4.2-5 4

5.3-1 5.3-1 i

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1 i.1 REACTOR COOLANT SYSTEM OPERATIONAL COMPONENTS Applicability:

Applies to the operating status of the reactor coolant system equipment.

Objective:

To specify conditions of reactor coolant system components i

for reactor operation.

Specification:

A.

At least one re' actor coolant pump or one low pressure safety injection pump operating in the residual heat removal mode. shall be in operation providing flow through the reactor when the reactor coolant system boron concentration is being reduced.

B.

At least one pressurizer code safety valve shall be operable whenever fuel is in the reactor and the reactor coolant system is isolated from the residual heat removal system and the head is on the vessel.

C.

At least two pressurizer code safety valves shall be operable whenever.the reactor is critical.

D.

At least one reactor coolant pump shall be in operation providing flow through the core with its steam generator capable of performing its heat transfer function whenever the reactor is in a critical condition.

E.

At least three reactor coolant pumps shall be in operation providing flow through the core with their steam generators performing _their heat transfer function whenever the reactor is in a power operation condition.

F.

At least three reactor coolant pumps shall be in operation providing flow through the core and with their steam generators performing their heat transfer function whenever the reactor power level condition-exceeds 63.0%.

G.

Minimum pressurizer spray flow must be operable whenever the reactor is critical.

4 Exception:

The requirement of,D and E may be modified j

during initial testing to permit' power levels not to exceed _10% of rated power with j

three loops, operating on natural circulation.

H.

One power operated relief valve (PORV) and its associated block valve shall be operable whenever the reactor coolant system temp. is greater than 210 F.

3.3-1 Amendment No. 52 4

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In the event that a PORV or its t lated block valve

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bec ome s inoperable, either restor e PORV un' block valve to operable status or close t. - associated block valve.and remove power from the block valve; otherwise, be in at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

J.

The pressurizer shall be operable with at least one bank of proportional heaters and a water level during normal system operation between 28 and 60 percent whenerer the reactor coolant system T is greater ay, than 5000F.

Basis:

When reactor coolant boron concentration is being reduced, the process must be uniform throughout the reactor coolant system volume to prevent stratification of reactor coolant at a lower boron concentration which could result in a reactivity insertion.

Sufficient mixing of the reactor coolant is assured by one low pressure safety injection (LPSI) pump operating in the RHR mode. When operated in this mode it will circulate the reactor coolant system volume in less than 12 minutes.

The

.pressur zer vo ume is relatively inactive; therefore, it i

l will tend to have a boron concentration higher than the rest of the reactor coolant system during a dilution operation.

Administrative procedures will require use of pressurizer spray to maintain a nominal spread between the boron concentration in the pressurizer and the reactor coolant system during the addition of boron.

Without residual heat removal, the amount of steam which could be generated a.t safety valve lift pressure with the reactor suberitical would be less than half of one valve's i

capacity.

One valve, therefore, provides adequate defense against overpressurization when the reactor is suberitical.

Overpressure protection is provided for all critical conditions. The safety valves are sized to relieve steam at a rate equivalent to the peak volumetric pressure surge rate.

For this purpose one safety valve is s6fficient; however, a minimum of two safety valves is required by Sectio'n III of the ASME Code.

3.3-2 Amendment No. 52 9

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• ti Reactor coolant pump flow and steam generator heat ~ transfer capabilities are specified to assure adequate core heat trans fer capability under all operating conditions from criticality to full power. Three loop operation is specified to assure plant operation is restricted to conditions considered in the LOCA analyses.

The exception permits testing to determine decay heat removal capabilities of the primary system prior to higher power operation while on natural circulation.

Following a loss of offsite power, stored and decay heat from the reactor would normally be^ removed by natural circulation using the steam generators as the heat sink.

Water supply to the steam generators is maintained by the auxiliary feedwater system.

Natural circulation cooling of the primary system requires the use of the pressuriaer heaters or high pressure safety injection pumps to maintain a suitable overpressure on the reactor coolant system.

Alternatively, in the event that natural circulation in the reactor coolant system is interrupted, the feed and bleed mode of reactor coolant system operation can be used to remove decay heat from the reactor. This method of decay heat removal requires the use of the emergency core cooling system (ECCS) and the power-operated relief valves (PORVs) in the pressurizer The PORVs can be operated either manually or automatically in the Maine Yankee design.

Block valves are provided upstream of the relief valves to isolate the valve in the event that a PORV valve fails.

Re ferences:

FSAR, Sections 4 and 9.

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l 3.3-3 Amendment No. 52 t

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3.9 OPERATIONAL SAFETY INSTRUMENTATION, CONTROL SYSTEMS, AND g

ACCIDENT MONITORING INSTRUMENTATION I

Applicab ili ty :.

Applies to, plant instrumentation system.

Ob je c tive:

To delineate the conditions of the plant ins trumentation and control systems necessary to ensure reactor safety.

Specification:

The operability of the plant instrument and control systems'shall be i~n accordance with Tables 3.9-1, 3.9-2 and 3.9-3.

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A.

Plant operation at rated power shall be permitted to continue with the limits as stated in the column entitled " Minimum Operable' Channels" except as conditioned by the column entitled j

" Bypass Conditions".

B.

In the event the number of operable channels or i

initia tion circuits falls below the limit given in Table 3.9-1, the plant shall be placed in a hot shutdown condition.

C.

In the event the number of operable sensors of a particular subsystem falls below the limit given Table 3.9-2, the plant shall be placed in a hot

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shutdown condition.

One subsystem can be removed from service during periods of maintenance or on-line testing for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D.

In the event the number of operable accident monitoring instrumentation channels falls below rSe Minimum Channels Operable requiraments in s

i Table 3.9.3, ei'ther restore the inoperable channel (s) to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least hot shutdown condition in the next i

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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l Basis:

Reactor safety is assured by the instrumentation channels, logic circuitry, trip modules, and other equipment necessary in the reactor protective-system.

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Selected nuclear steam supply system conditions are l

monitored and a rapid reactor shutdown lis inititated if any one or a combination of conditions deviates i

from a pre-selected range. This system automatically.

A initiates appropriate action to. prevent exceeding established sa fety limits.

Safety is'not compromised i

by continuing operation with certain instrumentation channels or initiation circuits oat of service since provisions were made for this in the plant design.

This specification outlines limiting conditions for opera tion necessary to preserve the e ffectiveness of the reactor protection system when~any one or more of the channels or circuits are out of service.

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3.9-1 j

Amendment No. 52 4

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In the reactor protective system, four independent and redundant channels monitor each safety parameter.

If any one of the four channels deviates from a pre-selected range, a trip signal is initiated. For any safety parameter, a trip signal from any two of the four protective channels will cause a reactor trip.

If one of the four channels is taken out of service for maintenance, the protective system for that parameter is changed to a two out of three coincidence for a reactor trip by bypassing the I

removed channel. When a second channel is taken out

- of service, the trip module for that channel is placed in the trip mode, and the resultant logic for that parameter is one out'of two.

Thus, with one or two channels removed from service for that parameter, protective action is initiated when required and the effectiveness of the reactor protection system is retained.

The operating requirements for the reactor protective system are shown in Table 3.9-1.

Redundant sensors and logic are provided for the initiation of all engineered safeguards systems.

In both the containment isolation and containment spray systems, two identical subsystems are used in each system.

In the safety injection actuation systems diverse sensors are used for the initiation of two identical subsystems.

Each of these three engineered safeguards systems may be operated as shown in Table 3.9-2 without jeopardizing safeguards initia tion.

One subsystem may be removed from service for a limited time for purposes of maintenance or testing.

Although no credit is taken for the high rate-of-change-of power channel in the Maine Yankee accident analysis, operability of this channel at low power levels provides back up assurance against excessive power rate increases.

Temperature feedback effects protect against excessive power rate increases at higher power levels.

The minimum number of operable channels for the accident monitoring instrumentation is given in Table 3.9-3. -The accident monitoring instrumentation is used to evaluate and aid in mitigating the consequences of an accident.

3.9-2 Amendment No. 19, 52 e

TABLE 3.9-3 ACCIDENT MONITORING INSTRUMENTATION Instrument Minimum Channels Operable 1.

Pressurizer Water Level 1

2.

Auxiliary Feedwater Flow Rate 1 per Steam Generator 3.

Reactor Coolant System 1

Subcooling Margin Monitor 4.

PORV Position Indicator 1/ valve

( Acoustic Flow Sensor) 5.

Sa fety Valve Position Indicator 1

(Acoustic Flow Sensor) 4 1

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Table 4.1+1 (Cont'd)

Channel Description Surveillance Function Frequency Surveillance Method 7.

Interlocks - Isolation Calibrate R

Apply known pressure to the Valves on Residual pressure sensors Hea t Removal Line 8.

Containment Pressure a.

Check S

a.

Verify pressure indication b.

Calibrate R

b.

Known pressure applied to sensor 9.

RHR HX Outlet a.

Check S

a.

Verify temperature indications, Temperature when the system is in operation b.

Calibrate R

Known resistance substituted for RTD 10.

Auxiliary Feedwater a.

Check ^

M(3)(4) a.

Perform internal self checking Flow Rate test b.

Calibrate R

b.

Apply simulated transducer EI s i r.na l

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11.

Reactor Coolant System a.

Check F(3)(4) a.

Comparison of monitor Subcooling Margin Monitor indication to existing pressure / temperature L,

relationships ro b.

Calibrate R

b.

Apply known pressure and temperature inputs into the processor 12.

PORV Positiori' Indicator a.

Calibrate R

Apply known frequency to sensors (Acoustic Flow Sensor) 4.1-11

Tcble 4.1-3 (Cent'4)

Channel Description Surveillance Function Frequency Surveillance Method 13.

Safety Valve Position a.

Calibrate R

'Appiy known frequency to sensor Indicator (Acoustic Flow Sensor)

14. _PORV Actuation a.

Check M(3)(4) a.

Bistable trip test Circuit b.

Calibrate R

b.

Apply known pressure to sensors (1) Not required unless the reactor is in the power operating condition.

(2)

Not required during plant startup and shutdown periods.

(3) Not required when plant is in the cold shutdown condition.

(4) Must be performed within 30 days prior to attaining a power operating condition.

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i Table 4.2-2 (Cont'd)

Test Frequency b)

System valves Verify operability Upon installation and within one month of startup from ear.h refueling shutdown.

c)

Flowpath Verify system flow Upon installation and capability by within one month of

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observing flow startup from each indication on the refueling shutdown.

system flowmeter 12.

PORV and PORV Block Valve operability test a)

Block Valves Verify operability At least once per 92 by operating the days.

valve (s) through one complete cycle of full travel b) PORVs Verify operability At least once per 18 by-manual actuation months.

of the control circuitry 13.

Pressurizer Verify Level ***

At least once per 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.4.

• Filters for containment and fuel storage building purging

• • Whenever the reactor coolant system is at or above operating pressure

• • • 0nly required when reactor coolant system T is greater than 500gye i

F.

1 4.2-5 Amendment No. 52-g

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5.3 FACILITY STAFF QUALIFICATIONS 5.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1) the Radiological Control Supervisor who shall meet or:

exceed the qualifications of Regulatory Guide 1.8, September 1975, and (2) the Shif t Technical. Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents.

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Amendment No. JJ, 59, 52 5.3-1 i

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