Text

Amend 42 to License DPR-61,changing Tech Specs to Incorporate Certain TMI Lessons Learned Category a Requirements
	ML20031F197
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png; (DPR-61-A-042, DPR-61-A-42)
Issue date:	10/08/1981
From:	Crutchfield D; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20031F187	List: ML20031F186; ML20031F197; ML20031F201; ML20031F207;
References
	NUDOCS 8110190339
	Download: ML20031F197 (24)
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UNITED STATES l

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(,,i NUCLEAR REGULATORY COMMISSION 5Q f;E WASHING TON, D. C. 2FS6 e

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J THE CONNECTICUT YANKEE ATOMIC POWER COMPANY DOCKET NO. 50-213

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HADDAM NECK PLANT j

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 42 License No. DPR-61 1.

The Nuclear Regulatory Comission (the Comissioni has found that:

A.

The application for amendment by the Connecticut Yankee Atomic Power Company (the licensee) dated September 16, 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 facility will operate in conformity with the application, the previsions 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 condurc id in compliance with the Comission'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 Comission's regulations and all applicable requirements have been satisfied.

i 2.

Accordingly, the license is amended by changes to the Technical Specifi-l cations as indicated in the attachment to this license amendment, and O

8110190339 811008 PDR ADOCK 05000213 P

PDR

-2' i

paragraph 2.C.(2) of Facility Operating License No. DPR-61 is hereby amended to read as follows:

b

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(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 42 are hereby incor-a porated in the license. The licensee shall operate the l

facility ir accordance with the Technical Specifications.

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1 3.

Thi: license amendment is offective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMNISSION

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. Qief Dennis M. Crutchf eld, Operating Reactors Branch #5 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: October 8, 1981 e

e ATTACHMENT TO LICENSE AMEhDMENT NO. 42 FACILITY OPERATING LICENSE N0. DPR-61 DOCKET NO. 213 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.

Delete Pages Insert Pages v

v vi 3-3 3-3 3-4 3-4c 3-4c a

3-4d 3-4d 3-14 3-14 3-15 3-15 3-15a (Tabic 3.9-1) 3-15a 3-15b 3-15c 3-19 3-19 3-20 3-20 3-20a 3-20b 3-20c 4-2c 4-2c 4-2d*

6-4 6-4 6-5 6-5 6-25 6-25 6-26 6-27

Previously 4-2c; only involves a pagination change

INDEX ADMINISTRATIVE CONTROLS SECTION 6.0 ADMINISTRATIVE CONTROLS PAGE 6.5.2.4 Consultants.......................

6-9 6.5.2.5 Meeti n g F requ ency....................

6-9 6.5.2.6 Quorum.........................

6-9 6.5.2.7 Review.........................

6-10 6.5.2.8 Audits 6-11 6.5.2.9 Authority........................

6-11 6.5.2.10 Records.........................

6-12 6.5.2.11 Qualifications 6-12 6.6 REPORTABLE OCCURRENCE ACTION..............

6-12 6.7 SAFETY LIMIT VIOLATION.................

6-13 6.8 PROCEDURES 6-13 3.;

REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS.....................

6-15 6.9.2 REPORTABLE OCCURRENCES 6-17 6.9.3 SPECIAL REPORTS.....................

6-22 6.10 RECORD RETENTION....................

6-22 6.11 RADIATION PROTECTION PROGRAM..............

6-24 6.12 RESPIRATORY PROTECTION PROGRAM (DELETED) 6.13 HIGH RADIATION AREA...................

6-24 6.14 ENVIRONMENTAL QUALIFICATION...............

6-25 v

Amendment No. JJ, 25, 37, 42

INDEX ADMINISTRATIVE CONTROLS SECTION 6.0 ADMINISTRATIVE CONTROLS PAGE 6.15 SYSTEMS INTEGRITY..................... 6-25 6.16 IODINE MONITORING..................... 6-25 vi Amendment No. 42

,.i 3.3 REACTOR COOLANT SYSTEM OPERATIONAL COMPONENTS Applicability:

Applies to the operating status of the reactor coolant system.

Objective:

To specify those limiting conditions for operation of the reactor coolant system which must be met to insure safe reactor operation.

Specification:

A.

At least one pressurizer code safety valve shall be in service whenever the reactor is suberitical and the reactor coolant system is above 375 F or 350 psig except during hydrostatic tests.

B.

One or more reactor coolant pumps or the residual heat removal system shall be in operation when changes are made in the boron concentration cI the reactor coolant.

C.

The reactor shall not be critical unless the following conditions have been satisfied:

(1) Three self-actuated, spring loaded safety valves, having a combined relieving capability of 720,000 #/hr. shall be in service and shall be in accordance with Section VIII of the ASME Boiler and Pressure Code.

(2) Above 1 percent of Nominal Operating Pcwer Level, at least one reactor coolant pump operating.

(3) Above 10 percent of Nominal Operating Power Level, at least three reactor coolant pumps operating.

(4) Above 65 percent of Nominal Operating Power Level.

Four reactor coolant pump operating.

(5) Two steam generators are capable of performing their heat transfer function.

(6) Two puwer operated relief valves (PORV's) and tneir associated block valves shall be operable except that:

a.

With one or more PORV(s) inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV(s) to OPERABLE status or close the associated block valve (s) and remove power from the block valve (s);

otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3-3 Amendment No. 30, 33,' 42

l b.

With one or more block valve (s) inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the block valve (s) to OPERABLE status or close the block valve (s) and remove power from the block valve (s); otherwise, be in at least HOT STAND 3Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTIDWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

(7) The pressurizer shall be operable with at least 150 KW of pressurizer heater capacity. The pressurizer. level shall be within + 5% of its programmed value during periods of normal operation, s.

With the pressurizer inoperable due to

~

the inoperability of both emergency power supplies to the' pressurizer heaters either restore the inoperable emergency power supply within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . With the pressurizer otherwise inoperable, be in at least HOT STANDBY vith the reactor trip breakers open within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in the HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

_ w.,

b.

During transient operations (startup, power level changes, trips, etc.) the pressurizer level may be outside the 5%

band for periods not to exceed one hour.

D.

Each steam generator in a non-isolated reactor coolant loop shall be restored to operable status 0

prior to increasing Tave above 200 F.

When starting a reactor coolant pump, and the reactor coolant cold leg temperature in any non-isolated loop is at or below 340 F, the secondary water temperature of i

i each non-isolated steam generator shall not be more than 200F higher than the water temperature of each of the non-isolated reactor coolant cold legs.

E.

The RCS Overpressure Protection System (OPS) shall be in operation when the RCS temperature is below 340 F unless the RCS is vented through a minimum

(

opening of three (3) inches (nominal diameter) or I

its equivalent.

If one or more of the relief trains is taken out of service and the RCS is not vented, the following actions shall be taken:

l l

3-4 Amendment No. 42

Basist

$chf35e ~pressurizis'UEcdemYety~4hTves3Fidesigned"foYriilf&e',

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240,000 lbs per hr. of saturated: steam at the-valve set @oint.

They are described more fully in FDSA Section 5.2.7.

Below '375'F and 350 psig in the reactor coolant system, the residual he t removal system can remove decay heat and thereby control systen temperature and. pressure.

If no decay heat were removed by any of the means available, the amount of steam which could be generated at safety valve relief pressure would be less than half the valves' capacity. One valve therefore provides adequate defense against over-pressurization.

When the boron concentration of the reactor coolant system is to be changedJ the process must be uniform to prevent sucden reactivity changes in the reactor. Mixing of the reactor coolant will be sufficient to maintain a unifonn boron concentration if at least one reactor coolant pump or one residual heat removal pump is running while the change is taking place. The residual heat renoval pump will circulate the primary system volume in approxi.aately one-hal f hour.

All pressurizer code safety 7alves are to be in service prior to criticality to permit the design relieving flow to occur if required.

Part C of the specification requites that a sufficient nur.ber of reactor coolant pumps be operating to provide core cooling in the event loss of flow occurs. The flow provided in each case vill keep DN3 vell above 1.30 as discussed in FDSA Section 10.3.2.

Therefore, cladding damage release of fission products to the reactor coolant cannot occur.

By limiting the temperature differential between the primary and secondary sides to twenty (20) degrees in-Part D, the resulting pressure transient will be prevented by the RCS OPS (See Reference 1) from exceeding the li=its in Specification 3.4.

As described in Reference (1), the RCS OPS, in conjunction with administrative controls, prevents exceeding the tempers ure and pressure limits in Specification 3.4 while RCS temperature is under 3400F. Considerations have been incorporated to provide for the inoperability--of_;one or more relief trains (relief valve, motor operated isolation valve, and associated instrumentation) when the RCS OPS is required to be operable.

Reference (1) D. C. Switzer letter to A. Schwencer, dated September 7, 1977.

3-4c Amendment No. 33, 39, C, 42

The power operated relief valves (PORVs) operate to relieve RCS pressure below the setting of the pressurizer code safety valves. These relief valves have remotely operated block valven to provide a positive shutoff capability should a relief valve become inoperable.

The electrical power for both the relief valves and the block valves is capable of being supplied from an emergency power source to ensure the abil1ty to seal this possible RCS leakage path.

The requirement that (150) kw of pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that,these heeters can be energized during a loss ' offsite power condition to maintain natural circulation at HOT STANDBY.

In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single

.ailure considerations require that two loops be' 0PERABLE.

In MODES 4 and 5, a single reactor ccolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops.be OPERABLE.

Thus, if the reactor coolant loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE.

The bperation of one Reactor Coolant Pump or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron conceritration reductioni in the Reacter Coolant System. The reactivity change rate associated with boron reduction will therefore, be within the capability of operator recognition and control.

The provisions of Specification 3.3.H arE' desirable on a one time basis for the purpose of demonstrating the reactor's ability to transfer decay heat and be cooled down while in natural circulation. Since at least two -nolant loops will be available and since this procedure is o plar.ned maneuver, there is no undue hazard to the plant or public safety.

3-4d Amendment No. 39, (J, 42

339 nPERATInNAL SArETY TNSTRTENTATION AND Ctt: TROL SYSTEMS nplic<bilitv Applico to tha op3rability of plant in:trunentction and control systens required for reactor safety.

Obiective:

To specify the limits imposed on plant operation by thesa instrunentatien and centrol systens required for reactor safety.

Sne ci fi cat i en :

Specifications 3.9.A,t3.9.B and 3.9.C relate to the items listed in Table 3.9-1.

The use of the table is illustrated in tne basis. Specifications 3.9.D and 3.9.E relate to the items listed in Table 3.9-2.

A.

Plant opetation at a rated power shall be per=itted to cen-

.tinue with the logic stated in colurn I of Tabic 3.9-1 for on-line testing or in the event a subsysten instrurentr.-

tion channel f ails.

B.

In the eve..c the nurber of channels of a particular sub-system in service falls belew the lir.it riven in Table 3.9-1, colarn I, so that the required locic cannot be net, plant operatien shall be limited accerding to the require-raent shown on Table 3.9-1, column 11.

C.

Neutron nenitoring instrunentatien shall be proviaed to centinuously eenitor neutron flux intensitics fron the fully shutdeen conditien to 120" of fuki power. Neutron monitors in each range shall be in centinuous eperatien until at least one decede of reliable indicatien is veri-fied on the next range of instrumentation.

D.

The accident monitcring instrumentation channels shcwn in Table 3.9-2 shall be operable whenever the reactor is critical.

E.

In the event the number of operable channels of a particular instrument is less than the minimum channels operable requirement, actions shall be taken per Table 3.9-2.

Basis:

In eclurn I of Table 3.9-1, the First fieure is the nu-l'cr of cperatienci channels which ust sense the abnorral ecnditien in crder to cause a reacter trir te c: cur, The secend firute is the mini =un nunber of operat.cnal chtnnels required for the

~

reactor to be at full power.

As an exanple, censider the nuclear overpever reactor trip.

Ivo channels must sense a nuclear pewer equal te or creater than the overpeuer tri-settine for a reactor trip te cecur, A ninitum of three channeli rust be sperable and in service te sense nuclear overpever fer the reactor to be at full power, Reactor safety, censistent with relichility of operati:n, is provided by the reactor protectien syster, which aute-aticalt-initiates appropricte action te prevent exceedinn ertab1:shed limits. This specifict: ion outlines limitine conditionr for operation necessary te preserve the effectiveness of the reactor control and protection system when any one of the channels is out of service.

3-14 Amendment No. 42

l The operability of the accident monitoring instrumentation listed in Table 3.9-2 ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and following an accident.

The term " operator surveillance" used in Table 3.9-1 means that when prescribed an operator shall continuously he watching the specified instruments.

Reference:

(1)

FDSA Section 7.2 i

l l

^u 3-15 Amendment No. 42 m

TABLE 3.9-1 MINIMUM INSTRUMENTATION OPERATING CONDITIONS Column I Column II Logic Required For Required Operating Action If At Power Requirements Full Power Operation Column I Logic Cannot Be Met 1.

Nuclear Overpower Reactor Trip 2/3 Mclutain Shutdown Canditions 2.

Pressurizer Vcriable Low Pressure Reactor Trip 1/2 Maintain Load Below PJ% Full Power (F.P.)

3.

Pressurizer Fixed High Pressure Reactor Trip 1/2 Maintain Shutdown.Conditicas 4.

Pressurizer High leater Level Reactor Trip 1/2 Maintain Load Below 10% F.P.

5.

Low :

iant Flow Reactor Trip - 4 Loop Operation 1/4 Maintain Load Below 84:!, F.P.

- 3 Loop Operation 1/3 Maintain Load Below 10% F.P.

6.

Pressurizer Low Pressure Signal (For Safety Injection Trip) 1/2 Maintain Shutdown Conditions 7.

Presiurizer Low Water Level Signal (For Safety Injection Trip) 1/2 Maintain Shutdown Conditions 8.

Manual Trip 1/1 Maintain Hot Shutdown Conditions 9.

Steam-Feedwater Flow Mismatch. Coincident Continuous Operator Surveillance of with Low Steam Generator Level-Reactor Trip 1/4 Steam and Feedwater Flow Recorder and Steam Generator Water Level of Affected Steam Generator

10. Hich Steam Flow Isolation Valve Trip - 4 Loop Operation 2/4 Isolate Corresponding Loop

- 3 Loop Operation 2/3 Maintain Shutdown with 1/2

11. Containment High Pressure Signal (For Safety Injection Trip) 2/3 Maintain Shutdown Conditions Startup Requirement Intermediate Range SUR Reactor Trip 1/1 Maintain Shutdown Conditions Source Range SUR Rod Stop 1/1 Maintain Shutdown Conditions

, Refueling Requirement 1

Shutdown High Neutron Level Alarm 1/1 No Changes in Core Geometry Permitted 3-15a Amendment No. 42

TABLE 3.9-2 ACCIDENT MONITORING INSTRUMENTATION

~

MINIMUM TOTAL NO.

CHANNELS INSTRUMENT OF CHANNELS OPERABLE ACTION 4

1.

Pressurizer Water Level 3

2 1

2.

Auxiliary Feedwater Flow Rate 1/S. G.

1/S. G.

1 3.

RCS Subcooling Margin Monitor 1

1 2

4.

PORV Position Indicator Acoustic Flow Monitor 1/ valve 1/ valve 3

us 2.

5.

PORV Block Valve Position EP Indicator 1/ valve

'/ valve 3

6.

Safety Valve Position Indicator Acoustic Flow Monitor 1/ valve 1/ valve 3

5 8

8n t

e 4

e

TABLE 3.9-2 (CONTINUED)

ACTION 1 - With the number of OPERABLE channels less than required by Table 3.3-11, either restore the inoperable channel (s) to OPERABLE status within 30 days or be in H0T STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 2 - With the subcooling margin monitor IN0PERABLE, determine the subcooling margin once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 3 - With any individual valve position. indicator inoperable, obtain quench tank temperature, level and pressure infor-mation, and monitor. discharge pipe temperature once per shift to determine valve position. This action is not required if the PORV block valve is closed with power removed in accordance with Specification 3.3.C.(6).

i I-3-15c Amendment No. 42 I

(2)

Containment purgs ccpability may be rendsred inoperable when the reactor is critical by placing a blank flange on the 42-inch purge air exhaust penetration inside the reactor containment for a period of seven days.

If the blank flange can not be removed within seven days, then the reactor shall be shut down within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

G.

The containment isolation valves specified in Table 3.11-1 shall be operable while in Modes 1, 2, 3 and 4 or:

With one or more of the isolation valve (s) specified in Table 3.11-1 inoperable, maintain at least one isolation valve OPERABLE in each af fected penetration that is open and either:

Restore the inoperable valve (s) to OPERABLE a.

status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or b.

Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one deactivated automatic valve secured in the isolation position, or Isolate each affected penetration within c.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one closed manual j

valve or blind flange, or d.

Be in at least HOT STANDBY within the next I

6, hours and in COLD SHlTfDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

H.

Containment Isolation The Containment Isolation actuation system shall be operable with the following trip setpoints:

Containment Pressure - HI < 5 psig Pressurizer Pressure a L0f,>lJ00 psig Basis:

A containment leakage rate of 0.3 weight percent of the con-tained air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at an internal pressure of 40 psig under hypothetical accident conditions with 3 of 4 air recircu-lation units operating will maintain public expesure well below 10 CFR 100 values.

(See Section 10.4 of the FDSA).

The reactor coolant system conditions of 300.psig and 2000F assure that no steam will be formed and hence there would be no pressure buildup in the containment if a reactor coolant I

system rupture were to occur.

The shutdown margins are selected based on the type of activities that are being carried out.

The 8% shutdown margin during refueling precludes criticality under any circumstances, even though fuel is being moved. When the reactor head is not to be removed, the specified shutdown margin of 3% 6 k pre-cludes criticality bn any occurrence.

3-19 Amendment No,42

Regarding internal pressure limitations, the containment design pressure of 40 psig would not be exceeded if the internal pressure before a major loss-of-coolant accident is maintained in accordance with Technical Specificatioi 3.12.C.

However, 3 psig maximum is sufficient for operations of the tuntinuous leakage monitoring system. The containment is designed to withstand an internal vacuum of.7.5 psig. The 2.0 psig vacuum is <,pecift;d as an operating limit to avoid any difficulties with motor cooiing.

The design air recirculation ilow rate gith 4 fans operating onder saturated conditions of 40 psig and 261 F is 200,000 CFM. The system is designed to perform its function with only 3 of 4 units in operation.

The air filtration system is discussed in detail in FSDA Section 3.6.

The containment spray system in itself can control the containment pressure.

It, therefore, provides a backup to the air recirculation system.

Containment post accident hydrogen venting can be accomplished by two methods. One uses the containment air particulate monitoring system and the other uses the containment purge exhaust system. These methods are not required in any short time frame after an accident; it is expected that months may elapse.

In any event the systems used if not operable for maintenance reasons can be readily made operable providing access into the centainment is not required.

Centainment purge is utilized as a backup means of venting hydrogen from the centainment following a loss-of-coolant accident. The containment air particulate monitoring syste= provides the primary means of purging because it provides adequate purge flow to prevent an explosive mixture buildup while allowing fine control of the release of radioactivity during purges. When necessary to effect repairs to the containment purge or purge bypass isolation valves, a blank flange must be applied to the 42" purge air exhaust penetration inside the reactor containment so that the containment re=ains leak tight. This renders the purge syste= inoperable for a finite time.

Seven days is considered a reasonable length of time for repair parts to be received, installed and the system retested for leak tightness and returned to service.

3-20 Amendment No. J7, 42

The OPERABILITY of the contaiment isolation valves ensures that i

the containment atmosphere will be isolated from the outsiAfe**

environment in the event of a release of radie.ctive material to the containment atmosphere or pressurization of the containment.

Containment isolation ensures that the release of-radioactive material to the environment will be consistent with the assumptions

'used in the analyses for a LOCA.

Reference:

(1)

FDSA Section 3.6 (2)

D. C. Switzer (CYAPC.) letter to A. Schwencer (NRC),

dated June 27, 1977, Attach 1ent No. 2.

3-20a Amendment No. 42

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TABLE 4.2-1 (CONTINUED)

Channel Action Minimum Frequency

25. Auxiliary Feedwater Calibrate Each Refueling Flow Rate Check Each Month

26. Reactor Coolant System Calibrate Each Refueling Subcooling Margin Monitor Check Each Month

27. PORV Position Indication Calibrate Each Refeeling (Acoustic Monitor)

Check Each Month

28. PORV Block Valve Calibrate Each Refueling Position Indication

29. Safety Valve Position Calibrate.

Each Refueling Indication (Acoustic Check Each Month Monitor) 4-2c Amendment flo. 42

Table 4.2-2 Minimum Equipment Check and Sampling Frequency Check Frequency 1.

Reactor Coolant Samples Radio-chemical analysis 5 days / week 2.

Reacter Coolant Boron Boron concentration 5 days / week 3.

Refueling Water Storage Baron concentration Weekly Tank Water Sample 4.

Control Rods Check rod drop times of Each Refueling s.11 rcds to be less than 2.5 seconds 5.

Control Rod Partial movement of Every 2 weeks all rods 6.

Pressurizer Safety Valves Check set point Each refueling 7.

Main Stean Safety Valves check set point Each refueling 8.

Main Steam Isolation valves Check Functioning Each refutling 9.

Reactor Containment Check Functioning Each refueling Trip Valves 10.

Refueling System Interlocks Check Functioning Each refueling 11.

Boric Acid Pumps Test Run Pu=ps Weckly 12.

RCS Overpressure Protection Check Functioning Once/ Cold Shutdowr.

System Isolation Valve Interlocks and Alarms 13.

RCS Overpressure Protection Verify valves are Once/72 hours when Isolation Valves open OPS is required overpressure protection 14.

RCS Vent (s)

Verify vents are

0nca /12 hours when open vent (s) is (are) required for over-pressure protection

Except when the vent pathway is provided with a valve that is locked, sealed, or otherwise secured in the open position.

4-2d Amendment No. )WI, 77, 42 1.

?

.f.

TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION I Personnel Number Condition Category Required of Unit

SOP 1

Normal Operating R0P 2

Condition Except AOP 2

Cold Shutdown STA 1

SOP 1

Cold Shutdown Conditions ROP 1

I AOP 2

^

Ebreviations: SOP - Licensed Senior Reactor Operator ROP - Licensed Reactor Operator A0P - Additional Operator STA - Shift Technical Advisor

Qualified in Radiation Protection Procedures

Shift crew composition may be one less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to. accommodate unexpected absence of on-duty shif* crew members,provided immediate action is taken to

, restore the shift crew composition to within the minimum requirements of Table 6.2-1.

This provision does not permit an, shift crew position to be unmanned upon shift change due to an ' oncoming shift ' crewman being late or absent.

~

6-4 Amendment No, H,42

. o e

' o.'

e ADMINISTRATIW CONTROLS 6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions.

6.3.1.1 The position of Health Physics Supervisor shall meet the following minimum qualifications:

Academic degree in an engineering or science field or l

equivalent as p,er Section 6.3.1.1.c.

l b.

Minimum of five years professional technical experience in the area of radiological safety, three years of which shall be in applied radiation work in a nuclear facility dealing with problems similar to those encountered in a i

nuclear power reactor, Technical experience in the area of radiological safety c.

beyond the five year minimum may be substituted on a one-for-one basis towards the academic degree requirement (four years of technical experience being' equivalent to a four year academic degree).

d.

Academic and technical experience must total a minimum of nine years.

6.3.1.2 The position of the Shift Technical Advisor 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.

6.4 TRAINING f

l A retraining and replacement training program for the facility l

6.4.1 staff shall be maintained innder the direction of the Training f

[

Coordinator assigned program responsibility and shall be in accordance with Section 5.5 of ANSI N18.1-1971 and App'endix "A" I

of 10 CFR Part 55.

t A fire brigade training program shall be maintained under 6.4.2 the direction of the training department'and shall meet or exceed the intent of Section 27 of the NFPA Code-1976 except.

that drills / training shall be conducted at least quarterly.

The effective date of this specification is March 1, 1978.

6.5 REVIEW AND ALD E U

PLANT OPERATIONS R571EW COMMITTEE (PORC) 6.5.1 FUNCTION The Ponc shall function to advise the Station Superintendent 6.5.1.1 on all matters related to nucicar safety..

6-5 Amendment No. RJ, 75, 42

ADMINISTRATIVE CONTROLS 6.14 ENVIRONMENTAL QUALIFICATION A.

By no later than June 30, 1982.all safety-related electrical equipment in the facility shall be qualified in accordance with the provisions of Division of Operating Reactors " Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors" (DOR Guidelines); or, NUREG-0588 " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment", December 1979.

Copies of these documents are attached to Order for Modification of License DPR-61 dated October 24, 1980.

B.

By no later than December 1, 1980, complete and auditible records must be available and maintained at a central location which describe the environmental qualification method used for all safety-related electri-cal equipment in sufficient detail to document the degree of compliance with the D0R Guidelines or NUREG-0588. Thereafter, such records should be updated and maintained current as equipment is replaced, further tested, or otherwise further qualified.

6.15 SYSTEMS INTEGRITY The licensee shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. This pregram shall include the following:

1.

Provisions establiching preventive maintenance and periodic visual inspection requirements, and 2.

Integrated leak test requirements for each system at a frequency not to exceed refueling cycle intervals.

6.16 IODINE MONITORING The licensee shall implement a program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:

1.

Training of personnel, 2.

Procedures for monitoring, and 3.

Provisions for maintenance of sampling and analysis equipment.

6-25 Amendment No. 42 l

..

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