ML20059C833

Amend 203 to License DPR-59,making Miscellaneous Corrections to App a TSs & App B Radiological Effluent TSs

ML20059C833
Person / Time
Site:	FitzPatrick
Issue date:	12/29/1993
From:	Capra R Office of Nuclear Reactor Regulation
To:
Shared Package
ML20059C836	List: ML20059C833 ML20059C841
References
NUDOCS 9401060123
Download: ML20059C833 (20)

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UNITED STATES

[ [' $ ".j NUCLEAR REGULATORY COMMISSION l

f WASHINGTON. D.C. 20555-0001

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7 POWER AUTHORITY OF THE STATE OF NEW YORK DOCKET NO. 50-333 JAMES A. FITZPATRICK NUCLEAR POWER PLANT AMENDMENT TO FACILITY OPEPATING LICENSE Amendment No. 203 License No. DPR-59 1.

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

A.

The application for amendment by Power Authority of the State of New York (the licensee) dated September 24, 1993, 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; 8.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.. The>

is reasonable assurance (i) that the activities authorized by tnis 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 Commission's regulations; D.

The issuance of this 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 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specificatioas as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-59 is hereby amended to read as follows:

9401060123 931229 PDR ADOCK 05000333 p

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1 (2) Technical Soecifications

-The Technical Specifications contained in Appendices Aland B, as revised through Amendment No. 203, are hereby. incorporated in the-

-J 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 to be-implemented within 30 days.

t FOR THE NUCLEAR REGULATORY COMMISSION

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g.y) -

1 Robert A.-Capra, Director Project Directorate I-I-Division of Reactor Projects - I/II Office of Nuclear. Reactor Regulation

Attachment:

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Changes to the Technical Specifications L

Date of Issuance:

December 29, 1993 1

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ATTACHMENT TO LICENSE AMENDMENT NO. 203-FACILITY OPERATING LICENSE NO. DPR-59 DOCKET NO. 50-333 Revise Appendix A as follows:

Remove Paaes Insert Paaes 55 55 56 56 64 64 96 96 116 116 142 142 144 144 150 150 162a 162a 186 186 192 192 197 197 247 247 Revise Appendix B as follows:

Remove Paoes Insert Paoes 23 23 33 33 37 37-66 66

JAFNPP 3.2 BASES Besides reactor protection instrumentation which initiates a reactor The low water level instrumentation set to trip at 177 in. above the scram, additional protective instrumentation is also provided. This top of the active fuel closes all isolation valves except those in protective instrumentation initiates action to mitigate the Group 1. Details of the isolation valve grouping are given in Section consequences of accidents which are beyond the operator's ability 7.3 of the updated FSAR. For valves which isolate at this level, this to control, or terminates operator errors before they result in serious trip setting is adequate to prevent uncovering the core in the casa consequences. This set of specifications provides the limiting of a break in the largest line.

conditions of operation for the primary system isolation function, initiation of the Core Cooling Systems, Control Rod Block and The low-low reactor water level instrumentation is set to trip when Standby Gas Treatment Systems. The objectives of the reactor water levelis 126.5 in, above the top of active fuel. This specifications are to assure the effectiveness of the protective trip in;trumentation when required, even during periods when portions eof such systems are out of service for maintenance, and to prcscribe the trip settings required to assure adequate performance.

When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.

Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety. The set points of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal cperating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.

Actuation of primary containment valves is initiated by protective instrumentation shown in Table 3.2-1 which senses the conditions ior which isolation is required. Such instrumentation must be cvailable whenever primary containment integrity is required.

The instrumentation which initiates primary system isolation is connected in a dual bus arrangement.

Amendment No.1)d,1[, }/3, 203 55

JAFNPP l

3.2 BASES (cont'd) l initiates the HPCI and RCIC systems and trips the recirculation Venturis are provided in the main steam lines as a means of pumps. The low-low-low reactor water levelinstrumentation is set measuring steam flow and also limiting the loss of mass inventory.

to trip when the water levelis 18 in. above the top of active fuel.

from the vessel during a steam line break accident. The primary This trip activates the remainder of the ECCS subsystems, closes function of the instrumentation is to detect a break in the main the main steam isolation valves, main steam line drain valves and steam line. For the worst case accident, main steam line break rcactor water sample line isolation valves, and starts the emergency outside the drywell, a trip setting of 140 percent of rated steam di:sel generators. These trip level settings were chosen to be high flow in conjunction with the flow limiters and main steam line valve enough to prevent spurious actuation but low enough to initista closure, limits the mass inventory loss such that fuel is not ECCS operation and primary system isolation so that post-accident uncovered, fuel temperature peak at approximately 1,000"F and l

cooling can be accomplished and the guidelines of 10 CFR 100 will release of radioactivity to the environs is below 10 CFR 100 l

not be exceeded. For large breaks up to the complete guidelines. Reference Section 14.6.5 of the updated FSAR.

1

.circumferential break of a 24 in. recirculation line and with the trip setting given above, ECCS initiation and primary system isolation cr3 initietod in time tt, meet the above criteria. Reference paragraph l

6.5.3.1 of the updated FSAR.

The high drywell pressure instrumentation is a diverse signal for malfunctions to the water level instruraentation and in addition to initiating ECCS, it causes isolation of Groups B and C isolation v lves. For the breaks discussed above, this instrumentation will generally initiate ECCS operation before the low-low-low water level instrumentation; thus the results given at ove are applicable here clso Details of the isolation valve closure group are given in l

Section 7.3 of the updated FSA:1. The water levelinstrumentation initiates protection for the full spectrum of loss-of-coolant accidents.

Amendment No. f,((,1)d,1[3, d

203 56

.3 JAFNPP

-Table 3.2-1 INSTRUMENTATION THAT INITIATES PRIMARY CONTAINMENT ISOLATION Minimum No.

cf Operable Total Number of Instrument t

Instrument Channels Channels Per Provided by Design

. Trip System (1) -

Instrument Trip Level Setting for Both. Trip Systems

' Action (2) 2 (6)

Reactor Low Water Level-2 177 in. above TAF 4

A-l 1

Reactor High Pressure s 75 psig -

2 D

.l' i

(Shutdown Cooling isolation) 2 Reactor Low-Low-Low Water Level -

..h 18 in. above the TAF 4-

.A l-2 (6)

High Drywell Pressure s 2.7 psig 4

A.

.l 2

High Radiation Main s 3 x Normal Rated 4

8

-l-Steam Line Tunnel Full Power Background (9) 2 Low Pressure Main Steam Line 2 825 psig (7)

'4 B

l 2

High Flow Main Steam Line ~

s 140% of Rated Steam Flow 4

B.

-l' 2

Main Steam Line Leak.

s 40*F above max ambient 4

~B.

.l Detection High Temperature 4

Reactor Cleanup System Equipment '

s 40*F above max ambient 8

C l-Area High Temperature 2

, Low Condenser Vacuum -

2 8" Hg. Vac (7)(8) 4

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_ Closes MSIV's

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, J/di, 203 d

64

' Ame'ndment.No.

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JAFNPP 3.3.C (cont'd) 4.3.C (cont'd) 2.

The average of the scram insertion times for the three 2.

At 16-week intervals,10 percent of the operable control fastest operable control rods of all groups of four control rod drives shall be scram timed above 950 psig.

rods in a two-by-two array shall be no greater than:

Whenever such scram time measurements are made, an evaluation shall be made to provide reasonable assurance Control Rod Average Scram that proper control rod drive performance is being Notch Position Insertion Time maintained.

Observed (Seconds) 46 0.361 38 0.977 24 2.112 e

04 3.764 3.

The maximum scram insertion time for 90 percent 3.

All control rods shall be determined operable once each insertion of any operable control rod shall not exceed operating cycle by demonstrating the scram discharge l

7.00 sec.

volume drain and vent valves operable when the scram test initiated by placing the mode switch in the SHUTDOWN position is performed as required by Table 4.1-1 and by verifying that the drain and vent valves:

a.

Close in less that 30 seconds after receipt of a signal for control rods to scram, and b.

Open when the scram signal is reset.

l Amendment No. [, pd,J,[,1)[5, 203 d

96

JAFNPP 3.5 (cont'd) 4.5 (cont'd) 2.

Should one RHRSW pump of the components required in 2.

When it is determined that one RHRSW pump of the 3.5.B.1 above be made or found inoperable, continued components required in 3.5.B.1 above is inoperable, the reactor operation is permissible only during the remaining components of the containment cooling modo succeeding 30 days provided that during such 30 days all subsystems shall be verified to be operable immediately remaining components of the containment cooling mode and daily thereafter.

subsystems are operable.

3.

When one containment cooling subsystem becomes 3.

Should one of the containment cooling subsystems inoperable, the redundant containment cooling subsystem become inoperable or should one RHRSW pump in each shall be verified to be operable immediately and daily I.

subsystem become inoperable, continued reactor thereafter. When one RHRSW pump in each subsystem operation is permissible for a period not to exceed 7 becomes inoperable, the remaining components of the days.

containment cooling subsystems shall be verified to be operable immediately and daily thereafter.

4.

If the requirements of 3.5.B.2 or 3.5.8.3 cannot be met, the reactor shall be placed in a cold condition within 24 hr.

5.

Low power physics testing and reactor operator training shall be permitted with reactor coolant temperature

< 212'F with an inoperable component (s) as specified in 3.5.B above.

Amendment No.[,

,1[8, 1, 1[3, 1 [1, 203

JAFNPP 3.6 (cont'd) 4.6 (cont'd) 5.

With the Primary Containment Sump Monitoring System 3.

Drywell Continuous Atmosphere Radioactisity Monitoring (Equipment Drain Sump Monitoring or Floor Drain Sump System instrumentation shall be functionally tested and Monitoring) inoperable, restore the system to operable calibrated as specified in Table 4.6-2.

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status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or 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 the cold condition within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

6.

With the Primary Contairtment Atmosphere Radioactivity Monitoring System (gaseous) or the Primary Containment Atmosphere Radioactivity Monitoring System (particulate) inoperable, operation may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Amendment _ No. /, jid, [12,[2,203 142

JAFNPP 3.6 (cont'd) 4.6 (cont'd)

F.

Structural Intearity F.

Structural Intearity The structural integrity of the Reactor Coolant System shall be 1.

Nondestructive inspections shall be performed on the maintained at the level required by the original acceptance ASME Boiler and Pressure Vessel Code Class 1,2 and 3 standards throughout the life of the Plant.

components and supports in accordance with the requirements of the weld and support inservice inspection program. This inservice inspection program is based on an NRC approved edition of, and addenda to Section XI of the ASME Boiler and Pressure Vessel Code which is in effect 12 months or less prior to the beginning of the inspection interval.

2.

An augmented inservice inspection program is required for those high stressed circumferential piping joints in the main steam and feedwater lines larger than 4 inches in diameter, where no restraint against pipe whip is provided. The augmented in-service inspection program shall consist of 100 percent inspection of these welds per inspection interval.

3.

An Inservice inspection Program for piping identified in -

the NRC Generic Letter 88-01 shall bo implemented in accordance with NRC staff positions on schedules, methods, personnel, and sampie expansion included in this Generic Letter, or in accordance with attemate l

measures approved by the NRC staff.

G.

Jet Pumos G.

Jet Pumos Whenever the reactor is in the startup/ hot standby or run Whenever there is recirculation flow with the reactor in the modes, o!! jet pumps shall be operable. If it is determined that startup/ hot. standby or run modes, jet pump operability shall be a jet pump is inoperable, the reactor shall be placed in a cold checked daily by verifying that the following conditions do not condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

occur simultaneously:

,1[4, [0, [0, 203 Amendment No.

144

JAFNPP 3.6 and 4.6 BASES (cont'd) than 100,000 lb/hr, a more restrictive limit of 0.1 ppm has During startup periods, which are in the category of less than been established to assure the chloride-oxygen combinations 100,000 lb/hr, conductivity may exceed 2 pmho/cm because of Fig. 4.6-1 are not exceeded. At steaming rates of at least of the initial evolution of gases and the initial addition of l

100,000 lb/hr, boiling occurs causing deaeration of the reactor dissolved metals. During this period of time, when the water, thus maintaining oxygen concentration at low levels.

conductivity exceeds 2 pmho/cm (other than short-term spikes). samples will be taken to assure the chloride When conductivity is in its proper normal range, pH and concentration is less than 0.1 ppm.

chloride and other impurities affecting conductivity must also be within their normal ranges. When and if conductivity The conductivity of the reactor coolant is continuously -

becomes abnormal, then chloride measurements are made to monitored. The samples of the coolant which are taken every determine whether or not they are also out of their normal 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> will serve as a reference for calibration of these operating values. This is not necessarily the case.

monitors and is considered adequate to assure accurate Conductivity could be high due to the presence of a neutral readings of the monitors. If conductivity is within its normal salt; e.g., Na,SO., which would not have an effect on pH or range, chicrides and other impurities will also be within their chloride. In such a case, high conductivity alone is not a normal ranges. The reactor coolant samples will also be used cause for shutdown. In some types of water-cooled reactors, to determine the chlorides. Therefore, the sampling frequency conductivities are,in fact, high due to purposeful addition of is considered adequate to detect long-term changes in the additives. In the case of BWR's, however, where no additives chloride ion content. Isotopic analyses of the reactor coolant are used and where neutral pH is maintained, conductivity required by Specification 4.6.C.1 may be performed by a provides a very good measure of the quality of the reactor gamma scan.

water. Significant changes therein provide the operator with a warning mechanism so he can investigate and remedy the condition causing the change before limiting conditions, with D.

Coolant Leakaos respect to variables affecting the boundaries of the reactor coolant, are exceeded. Methods available to the operator for Allowable leakage rates of coolant from the Reactor Coolant correcting the condition include operation of the Reactor System have been based on the predicted and experimentally Cleanup System, reducing the input of impurities and placing observed behavior of cracks in pipes and on the ability to make the reactor in the cold shutdown condition. The major benefit up Reactor Coolant System leakage in the event of loss of of cold shutdown is to reduce the temperature dependent off-site a-c power. The normally expected background leakage -

corrosion rates and provide tirne for the Reactor Water due to equipment design and the detection capability for Cleanup System to reestablish the purity of the reactor determining system coolant.

1[J,1[0, 203 Amendment No.

150

s e :s n-s JAFNPP-Table 4.6-2 Minimum Test and Ca!ibration Freauency for Drywell Continuous Atmosphere Radioactivity Monitorina System inst. Channel Inst. Functional Test Calibration Sensor Check-1.

Air Particulate Analyzer.

None Once 'l 3 mos.

once / day

'l-

'1 2.

Gaseous Activity Analyzer None Once / 3 mos.

once / day.

3,.

lodine Analyzer None Once / 3 mos.

once / day' L

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JAFNPP 3.7 BASES- (cont'd) of the containment. Closure of one of the valves in each line A list of containment isolation valves, including a brief-would be sufficient to maintain the integrity of the Pressure

description of each valve is included in Section 7.3 of thei

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Suppression System. Automatic initiation is required to updated FSAR.

.ll minimize the potential leakage paths from the containment in the event of a loss-of-coolant accident.

The containment isolation valves on the containment vent and purge lines may be open for safety related reasons. Safety related reasons include, but are not limited to, the following:

inerting or de-inerting primary containment; maintaining containment oxygen concentration; maintaining drywell'and suppression ' pool atmospheric pressures; and maintaining the differential pressure between the drywell and suppression pool. These valves have been modified to limit the maximum angle of opening as shown in 3.7.D.1.

Nine remote manual isolation valves have bet.n added to the Reactor Budding Closed Loop Cooling Water System IRBCLCWS) in order to comply with.10 CFR BO Appendix A GDC 57; These valves are air operated (with solenoid pilot -

valves), normally openi and are designed to fail."open"on loss =

of electrical power or "as is" upon loss of instrument air.- Each -

ACV is provided with a. Seismic Class I accumulator tank to allow operation of the valves uoon loss of instrument air up to 2 full valve cycles.-The fail-open design permits continued operation of the system to' supply water to the recirculation pump-motor coolers and drywell coolers during normal:

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- operation and as necessary under accident conditions. If there is a postulated accident, and indications of leakage from

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RBCLCWS appear, the operator will selectively close the AOV's affected to provide containment isolation.

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- JAFNPP 4.7 BASES (cont'd) operability results in a more reliable system.

The main steam line isolation valves are functionally tested on a more frequent interval to establish a high degree of -

reliability.

The primary containment is penetrated by several small

- diameter instrument lines connected to the reactor coolant

-system. Each instrument line contains a 0.25 in. restricting orifice inside the primary containment and an excess flow-

-check valve outside the primary containment.

5 The RBCLCWS valves are' excluded from the quarterly

- surveillance requirements because closure of these valves will eliminate the_ coolant flow to the drywell air and recirculation pump-motor coolers. - Without cooling water, the drywell air.

and equipment temperature willincrease and may.cause damage to the equipment during normal plant operations.

Therefore,; testing of these valves would only be conducted in the cold condition.

A list of containment isolation valves, including a brief -

. description of each valve is included in Section 7.3 of the

- updated FSAR.

- [3,203..

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r JAFNPP 6.0 ADMINfSTRATIVE CONTROLS Administrative Controls are the means by which plant operations are subject to management control. Measures specified in this section provide for the assignment of responsibilities, olant organization, staffing qualifications and related requirements, review and audit mechanisms, procedural controls and reporting requirements. Each of these measures are necessary to ensure safe and efficient facility operation.

6.1 RESPONSIBILITY The Resident Manager is responsible for safe operation of the plant. During periods when the Resident Manager is unavailable, one of the three General Managers will assume this responsibility. In the event all four are unavailable, the Resident l

Manager may delegate this responsibility to other qualified supervisory personnel.

The Resident Manager reports directly to the Executive Vice President-Nuclear Generation.

6.2 ORGANIZATION i

6.2.1 Facility Manecement and Technical Suonort Onsite and offsite organizations shall be established for plant operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities that affect the safety of the nuclear power plant.

1. Lines of authority, responsibility, and communication shall be established and defined for the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organization charts, functional descriptions of department responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the Updated FSAR.

2. The Resident Manager shall be responsible for overall plant operation, and shall l

have control over those onsite activities that are necessary for safe operation and maintenance of the plant.

3. The Executive Vice President - Nuclear Generation shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining,

[

and providing technical support to the plant to ensure nuclear safety.

4. The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.

6.2.2 Plant Staff The plant staff organization shall be as fo!!ows:

1. Each shift crew shall be composed of at least the minimum shift crew composition shown in Table 6.2-1:

Amendment No. pd,Jid,)d,1[,1

,1[247, [8, 203

r-i JAFNPP NOTES FOR TABLE 3.2-1 (continued)

(d)

Main stack gaseous sampling and analysis shall also be performed following shutdown, startup, or a thermal power change exceeding 20% of rated thermal power in one hour.

1. This requirement applies only if:

Analysis shows that the dose equivalent 1-131 concentration in the primary coolant has increased more than a factor of 3; and The noble gas monitor shows that effluent activity has increased more than a factor of 3; and Correction for increases due to changes in thermal power level have been made in both cases.

(e)

Main stack iodine and particulate sampling shall also be performed daily fo!!owing each shutdown, startup or thermal power change exceeding 20% of rated thermal power in one hour.

1. Daily saa i

'3 s not required for thermal power changes if the off gas charcoal filters are,n service.

2. In addition, this requirement applies only if:

Analysis shows that the dose equivalent 1-131 concentration in the primary coolant has increased more than a factor of 3; and The noble gas monitor shows that effluent activity has increased more than a factor of 3; and Corrections for increases due to changes in thermal power level have been made in both cases.

3. Daily sampling shall be performed until two consecutive samples show no increase in concentration but not to exceed 7 consecutive days.

4. LLDs may be increased by a factor of 10 for analysis of daily samples.

5. Analysis of daily and weekly samples shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of changTng.

(f)

Incinerated oil may be discharged via points other than the main stack and building vents (i.e., auxiliary boiler). Release shall be accounted for based on pre-release grab sample data.

(g)

Samples of incinerated oil releases shall be co!!ected from and representative of filtered oil in liquid form. Whenever oil samples cannot be filtered such as No. 6 bunker fuel oil, raw oil samples shall be collected and analyzed.

l Amendment No. [,1)d, 203 23

JAFNPP LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS treatment system undes; the following conditions:

1.

An instrument check shall be performed daily when the offgas treatment system is in operation.

1.

The offgas dilution steam flow instrumentation shall alarm and automatically isolate the offgas recombiner 2.

An instrument channel functional test shall be performed I

system at a low flow setpoint greater than or equal to once per operating cycle.

6300 pounds per hour and at a high flow setpoint less than or equal to 7900 pouiids per hour.

3.

An instrument channel calibration shall be performed once per operating cycle.

2.

The offgas recombiner inlet temperature sensor shall alarm and automatically isolate the offgas recombiner I

system at a temperature setpoint of greater than or equal to 125'C.

l 3.

The offgas recombiner outlet temperature sensor shall alarm and automatically isolate the offgas treatment I

system at a temperature setpoint of greater than or equal to 150'C.

c.

In lieu of continuous hydrogen or oxygen monitoring, the c.

With condanser offgas treatment system recombiner in condenser offgas treatment system recombiner effluent shall service, in lieu of continuous hydrogen or oxygen monitoring, be analyzed to verify that it contains less than or equal to 4%

the hydrogen content shall be verified weekly to be less than hydrogen by volume.

or equal to 4 % by volume.

d.

With the requirements of the above specifications not in the event that the hydrogen content cannot be verified, satisfied, restore the recombiner system to within operating operation of this system may continue for up to 14 days.

specifications or suspend use of the charcoal treatment system within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Amendment No. [.1)d,1 d, 203

)

33

JAFNPP Table 3.101 RADIATION MONITORING SYSTEMS THAT INITIATE AND/OR ISOLATE SYSTEMS Minimum No.

cf Operable Total Number of Instrument instrument Channels Channels Trip Function Trip Level Settings Provided by Design Actions 1(a)

Refuel Area Exhaust Monitor (b) 2 (c) or (d!

1(a)

Reactor Building Area Exhaust Monitors (b) 2 (d) 1(a)

SJAE Radiation Monitors

.g_ 500,000 pCi/sec 2

(e) 1(a)

Turbine Building Exhaust Monitors (b) 2 (f) 1(a)

Radwaste Building Exhaust Monitors (b) 2 (f) 1(a)

Main Control Room Ventilation s4 x 108 cpm" 1

(g) l (h)

Mechanical Vacuum Pump Isolation 53 x Normal Full 4

(h)

Power Background NOTES FOR TABLE 3.10-1

(;)

Whenever the systems are required to be operable, there shall be one operable or tripped instrument channel per system. From and af ter the time it is found that this cannot be met, the indicated action shall be taken.

(b)

Trip level setting is in accordance with the methods and procedures of the ODCM.

(c)

Cease operation of the refueling equipment.

(d)

Isolate secondary containment and start the SBGTS.

(:)

Bring the SJAE release rate within the limit within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in hot standby within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

(f)

Refer to Appendix B LCO 3.1.d.

(g)

Control room isolation is manually initiated.

(h)

Uses same sensors as primary containment isolation on high main steam line radiation. Refer to Appendix A Table 3.2-1 for minimum -

number of operable instrument channels and action required.

(i)

Conversion factor is 8.15 x 10' cpm - 1pCi/cc.

Amendment No. pd,1[7, 203 37

l JAFNPP 7.0 ADMfNISTRATIVE CONTROLS 7.1 RESPONSIBILITY

a. The Resident Manager shall have direct responsibility for assuring the operation of the James A. FitzPa* ck Plant is conducted in such a manner as to provide continuing protection to the environment. During periods when the Resident Manager is unavailable, one of the three General Managers will assume this responsibility. In the event all four are unavailable, the Resident Manager may delegate this responsibility to other qualified supervisory personnel.

b. Implementation of the Radiological Effluent Technical Specifications is the responsibility of the General Manager - Operations, with the assistance of the

[.

plant staff organization.

7.2 PROCEDURES Written procedures and administrative policies shall be established, implemented and maintained that meet or exceed the requirements and recommendations of Section 5

" Facility Administrative Policies and Procedures" of ANSI 18.7-1972 and Regulatory Guide 1.33, November 1972, Appendix A.. In addition, procedures shall be established, implemented and maintained for the PCP, ODCM, and Quality Control Program for effluent and environmental monitoring using the guidance in Regulatory Guide 4.1, Revision 1.

7.3 REPORTING REQUfREMENTS

a. Planned Liouid and Gaseous Releases The limits for radioactive materials contained in liquid and gaseous effluents are contained in Specifications 2.3,3.3 and 3.4.

b. Environmental Samoles Exceedino Limits of Table 6.1-2 When the limits of Table 6.1-2 are exceeded, refer to Specification 6.1.b for reporting requirements.

c. Semiannual Radioactive Effluent Release Reoort Routine Radioactive Effluent Release Reports covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The period of the first report shall begin with the date of initial criticality.

1. The Radioactive Effluent Release Report shallinclude a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit using as guidance Regulatory Guide 1.21, Revision 1, June 1974, " Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants", with data summarized on a quarterly basis following the format of Appendix B thereof.

d Amendment No. J, 203 66 j