Text

Amend 10 to License NPF-36,changing Definition of Core Alteration in Tech Specs to Include Certain Exceptions & Footnotes in Tech Specs to Be Consistent W/New Definition
	ML20196E578
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	11/30/1988
From:	Butler W; Office of Nuclear Reactor Regulation
To:	;
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ML20196E570	List: ML20196E564; ML20196E578; ML20196E596;
References
	NUDOCS 8812120040
	Download: ML20196E578 (15)
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n a aseg( ~ UNITE 3 STATES

! c NUCLEAR REGULATORY COMMISSION W ASHING TON. D. C. 20666

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LONG ISLAND LIGHTINr. COMPANY DOCKET N_0. 50-322 SHOREHAM NUCLEAR P0hTR STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.10 License No. NPF-36 L

1. The Nuclear Regulatory Comission (the Comission) has found that A. The application for amendment by Long Island Lighting Company (the licensee), dated September 4,1987 and supplernented November 19, 1987, cortplies 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 provisions of the Act, and the rules and regulations of the Comission; C. There is reasonable assurance (i) that the activities authorized by this arrendment can be conducted without endangering the health and safety of the putslic, and (ii) that such activities will be conducted in compliance with the Commissinn'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.

2. Accordingly, the license is amended by changes to the Technical Specifications

as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF 36 is hereby amended to read as follows

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised '

through Amendment No. 10 are hereby incorporated into this license. !

Long Island Lighting Company shall operate the facility in !

accordance with the Technical Specifications and the Environmental Protection Plan.

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8812120040 GG1130 i PCR ADOCK 05000322 P PDC j

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, . 3. This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION i

4 Walter R. Butler, Director !

Project Directorate I-2 :

Division of Reactor Projects I/!! ;

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Attachment:

Changes to the Technical Specifications 7

. Date of Issuance: November 30, 1988

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ATTACHMENT TO LICENSE AMENDMENT NO.10 FACILITY OPERATING LICENSE NO. NPF-36 DOCKET NO. 50-322 l

Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by Amendment number and l contain vertical lines indicating the area of change. Overleaf pages are provided to maintain document completeness.*

Remove Insert 1-1 1-1*

1-2 1-2 3/4 1-1 3/4 1-1 3/4 1-? 3/4 1-2 3/4 1-5 3/4 1-5*

3/4 1-6 3/4 1-6 4

3/4 3-3 3/4 3-3*

3/4 3-4 3/4 3-4 i I

3/4 9-3 3/4 9-3 3/4 9-4 3/4 9-4* !

3/4 9-7 3/4 9-7 1

3/4 9-8 3/4 9-8*

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1.0 DEFINITIONS The following terms are defined so that uniform interpretation of these specifications may be achieved. The defined terms appear in capitalized type j and shall be applicable throughout these Technical Specifications.

ACTION 1.1 ACTION shall be that part of a Specification which prescribes remedial measures required under designated conditions, j AVERAGE PLANAR EXPOSURE l 1.2 The AVERAGE PLANAR EXPOSURE shall be applicable to a specific planar height and is equal to the sum of the exposure of all the fuel rods in the specified bundle at the specified height divided by the number of fuel r rods in the fuel bundle.

AVERAGE PLANAR LINEAR HEAT GENERATION RATE 1.3 The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) shall be applicable 4

to a specific planar height and is equal to the sum of the LINEAR HEAT

GENERATION RATES for all the fuel rods in the specified bundle at the 1 specified height divided by the number of fuel rods in the fuel bundle.

CHANNEL CALIBRATION 1

j 1.4 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel

output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION j shall encompass the entire enannel including the sensor and alarm and/or i trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or 1

total channel steps such that the entire channel is calibrated.

! CHANNEL CHECK 1.5 A CHANNEL CHECK shall be the qualitative assassment of channel behavior during operation by observation. This determination shall include, where Dossible, comparison of the channel indicat'on and/or statu. with other i

indications and/or status derived from indelendent instrument channels j measuring the same parameter.

CHANNEL FUNCTIONAL TEST 1.6 A CHANNEL FUNCTIONAL TEST shall be:

1 a. Analog channels - the injection of a simulated signal into the channel j as close to the sensor as practicable to verify OPERABILITY including i alarm and/or trip functions and channel failure trips,

b. Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

I The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is tested.

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l SHOREHAM - Ud!T 1 11 1

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i OEFINITIONS CORE ALTERATION 1.7 CORE ALTERATION shall be the addition, removal, relocation or movement of fuel, sources, incore instruments or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel.

Normal movement of the SRMs, IRMs, or TIPS in their driving system is not considered a CORE ALTERNATION. Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe conservative '

position. l CRITICAL POWER RATIO 1.8 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly which is calculated by application of the GEXL correlation to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

DOSE EQUIVALENT I-131 1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries per gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, 1-132, I-133, I-134, and I-135 actually present. :

The thyroid dose conversion factors useo for this calculation shall be those listed in Table III of TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites."

E-AVERAGE DISINTEGRATION ENERGY 1.10 E shall be the average, weighted in proportion to the concentration of each radionuclide in the reacter coolant at the time of sampling, of the ,

sum of the average beta and gamma energies per disintegration, in MeV, for isotopes, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME ,

1.11 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time }

interval from when the monitored parameter exceeds its ECCS actuation set- ;

point at the channel sensor until the ECCS equipment is capable of performing l its safety function, i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc. Times shall i include diesel generator starting and sequence loading delays where applicable. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME i 1.12 The END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be i that time interval to complete suppression of the electric are between t the fully open contacts of the recirculation pump circuit breaker from i initial movement of the associ3*.d:

a. Turbine stop valves, and

b. Turbine control valves.

The response time may be measured by any series of sequential, overlapping -

or total steps such that the entire response time is measured.

SHOREHAM - UNIT 1 1-2 Anendment No. 10

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUT 00WN MARGIN LIMITING CONDITION FOR OPERATION l

3.1.1 The SHUTDOWN MARGIN shall be equal to or greater than:

a. 0.38% delta k/k with the highest worth rod analytically determined, or

b. 0.28% delta k/k with the highest worth rod determined by test.

APPLICABILITY: OPERATIONAL CONDITION! 1, 2, 3, 4, and 5.

ACTION:

i With the SHUTDOWN MARGIN less than spec' fled:

a. In OPERATIONAL CONDITION 1 or 1, reestablish the required SH'JTDOWN -

MARGIN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

b. In OPERATIONAL CONDITION 3 or 4, immediately verify all insertable control rods to be inserted and suspend all activities that could

reduce the SHUTDOWN MARGIN. In OPERATIONAL CONDITION 4, establish l SECONDARY CONTAINMENT INTEGRITY within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

I c. In OPERATIONAL CONDITION 5, suspend CORE ALTERATIONS

and other activities that could reduce the SHUTDOWN MARGIN and insert all insertable control rods within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Establish SECONDARY CONTAIN-MENT INTEGRITY within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . -

SURVEILLANCE REQUIREHENTS I

4.1.1 The SHUTDOWN MARGIN shall be determined to be equal to or greater than specified at any time during the fuel cycle:

a. By measurement, prior to or during the first startup after each refueling,

b. By measurement, within 500 HWD/T prior to the core average exposure at which the predicted SHUTDOWN MARGIN, including uncertainties and l calculation biases, is equal to the specified limit,

c. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after detection of a withdrawn control rod that is immovable, as a result of excessive friction or sechanical inter-forence, or is untrippable, except that the above required SHUTDOWN t MARGIN shall be verified acceptable with an increased allowance for the j withdrawn worth of the immovable or untrippable control rod.

! *Except movement of special movable detectors.

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i SHOREHAN - UNIT 1 3/4 1-1 A.- endment No.10 l

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REACTIVITY CONTROL SYSTEMS ,

3/4.1.2 REACT!VITY ANOMALIES t'

LIMITING CONDITION FOR OPERATION l 3.1.2 The reactivity equivalence of the difference between the actual ROD :

DENSITY and the predicted ROD DENSITY shall not exceed 1% delta k/k. '

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

i j ACTION: !

With the reactivity equivalence difference exceeding LE delta k/k: !

a. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months perform an analysis to determine and explain the cause ;

of the reactivity difference; operation may continue if the difference ,

is explained and corrected, i i

b. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . !

4 SURVEILLANCE REQUIREMENTS I L

4.1.2 The reactivity equivalence of th'e difference between the actual ROD

. DENSITY and the predicted R00 DENSITY shall be verified to be less than or equal to 1% delta k/k

a. During the first startup following CORE ALTERATIONS,* and

b. At least once per 31 effective full power days during POWER OPERATION. (

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j Except movement of control rods with their normal drive system. i I

SHOREHAM - UNIT 1 3/4 1-2 Anendment No.10 l 1

REACTIVITY CONTROL SYSTEMS SURVEILLANCE PigU!REMENTS (Continued) 4.1.3.1.4 The scram discharge volume shall be determined OPERA 8LE by demonstrating:

a. The scram discharge volume drain and vent valves OPERA 8LE, when control rods are scram tested from a normal control rod configura-tion of less than or equal to 504 ROD DENSITY at least once per 18 months, by verifying that the drain and vent valves:

1. Close within 30 seconds after receipt of a signal for control rods to scram, and

2. Open when the scram signal is reset.

b. Proper level sensor response by performance of a CHANNEL FUNCTIONAL TEST of the scram discharge volume scram and control rod block level j instrumentation at least once per 31 days, i

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i SHOREKAM - UNIT 1 3/4 1-5

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_ REACTIVITY CONTROL SYSTEMS '

CONTROL ROD MAXIMUM SCRAM INSERTION TIMES

.! i LIMITING CONDITION FOR OPERATION I

( 3.1.3.2 The maximum scram insertion time of each control rod from the fully !

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withdrawn position to notch position 5, based on deenergization of the :

scram pilot valve solenoids as time zero, shall not exceed 7.0 seconds. !

i l APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2, 1

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ACTION:

l l; l a. With the maximum scram insertion time of one or more control rods exceeding 7 seconds: !

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1. Declare the control rod (s) with the slow insertion time -

inoperable, and

2. Perform the Surveillance Requirements of Specification 4.1.3.2.c.

1 at least once per 60 days when operation is continued with three !

! or more control rods with maximum scram insertion times in excess !

of 7.0 seconds.

l Otherwise, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

f I b. The provisions of Specification 3.0.4 are not applicable.

I SURVEILLANCE REQUIREMENTS :

i 4.1.3.2 The maximum scram insertion time of the control rods shall be demon- i strated through measurement with reactor coolant pressure greater than or l 1

equal to 950 psig and, during single control rod scram time tests, the control l rod drive pumps isolated from the accumulators:

l i l a. For all control rods prior to THERMAL POWER exceeding 40% of RATED I THERMAL POWER following CORE ALTERATIONS

or after a reactor I

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shutdown that is greater than 120 days, j 1

! b. For specifically affected individual control rods following I l

maintenance on or modification to the control rod or control rod !

drive system which could affect the scram insertion time of those l specific control rods, and

c. For at least 10% of the control rods, on a rotating basis, at least I once per 120 days of POWER OPERATION.

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l l l *Except movement of special movable detectors or normal control rod movement.

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SHOREHAM - UNIT 1 3/4 1-6 Amendment No.10 l l

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TABLE 3.3.1-1 (Continued) b REACTOR P20TECTION SYSTEM INSTRUIENTATION

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x APPLICABLE MINIMUM I '

OPERATIONAL OPERABLE CHAf81ELS E FUNCTIONAL UNIT CONDITIONS PER TRIP SYSTEN (a) ACTION

~ 6. Main Steam Line Radiation -

High 1,2(f) 2 5

7. Primary Containment j Pressure - High 1, 2(h) 2 1

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8. Scram Discharge Volume Water 1 Level-High l a. Level Transmitter 1, 2 2 1 w 5(i) 2 3 k

w b. Float Switch 1, 2 2 1 J. 5(i) 2 3 i

i 9 Turbine Stop Valve - Closure 1(j) 4(k) 6 Turbine Control Valve Fast Closure, Valve Trip Oil Pressure - Low 1(j) 2(k) 6

11. Reactor Mode Switch Shutdown Position 1, 2 2 1 3, 4 2 7 5 2 3

12. Manual Scram 1, 2 2 1 3, 4 2 8 5 2 9

TABLE 3.3.1-1 (Continued)

.i REACTOR PROTECTION SYSTEM INSTRUMENTATION c ACTION STATEMENTS I ACTION 1 -

Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . '

ACTION 2 -

Verify all insertable control rods to be inserted in the core !

and lock the reactor mode switch in the Shutdown position f within I hour, i

ACTIG 3 -

Suspend all operations involving P. ORE ALTERATIONS

and insert all insertable control rods within I hour.

ACTION 4 -

Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 5 -

Be in STARTUP with the main steam line isolation valves closed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . ,

h ACTION 6 -

Initiate a reduction in THERMAL POWER within 15 minutes and ;

reduce turbine first stage pressure to < 109 psig,"* equivalent '

to THERMAL POWER less than 30% of RATED THERMAL POWER, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

ACTION 7 -

Verify all insertable control rods to be inserted within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . !

ACTION 8 -

Lock the reactor mode switch in the Shutdown position within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 9 -

Suspend all cperations involving CORE ALTERATIONS,* and insert l

all insertable control rods and lock the reactor rods switch in ;

the SHUTDOWN position within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . :

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Except movement of special movable detectors, or replacement of LPRM strings f provided SRM instrumentation is OPERABLE per Specification 3.9.2. l

- Initial setpoint. Final setpoint to be determined during the startup test program, j

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SHOREHAM - UNIT 1 3/4 3-4 Amendment No.10

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REFUELING OPERATIONS 3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 At least 2 source range monitor * (SRM) channels shall be OPERABLE and inserted to the normal operating level with:

a. Continuous visual indication in the control room,

b. One of the re;uired SRM detectors located in the quadrant where CORE ALTERATIONS are being performed and the other required SRM detector located in an adjacent quadrant, and

c. The "shorting links" removed from the RPS cigcuitry prior to and during the time any control rod is withdrawn and shutdown margin

.' demonstrations are in progress.

APPLICABILITY: OPERATIONAL CONDITION 5. -

ACTION:

With the requirements of the above specification not satisfied, imediately suspend all operations involving CORE ALTERATIONS ** and insert all insertable control rods.

SURVEILLANCE REQUIREMENTS 4.9.2 Each of the above required SRM channels shall be demonst.*ated OPERABLE by:

a. At least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s:

1. Performance of a CHANNEL CHECK,

2. Verifying the detectors are inserted to the normal operating level, and

3. During CORE ALTERATIONS, verifying that the detector of an OPERABLE SRM channel is located in the core quadrant where CORE ALTERATIONS are being performed and another is located in an

. adjacentquadrant.

The use of special movable detectors during CORE ALTERATIONS in place of the normal SRM nuclear detectors is permissible as long as these special detectors are connected to the normal SRM circuits.

- Except movement of special movable detectors.

Not required for control rods removed per Specification 3.9.10.1 and 3.9.10.2.

SHOREHAM - UNIT 1 3/4 9-3 Amendrent No.10

REFUELING OPERATIONS ,

SURVEILLANCE REQUIREMENTS (Continued)

b. , Performance of a CHANNEL FUNCTIONAL TEST:

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1. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to the start of CORE ALTERATIONS, and

2. At least once per 7 days,

c. Verifying that the channel count rate is at least 0.7 cps #:

1. Prior to control rod withdrawal,

2. Prior to and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during CORE ALTERATIONS, I and

3. At least once per 24 h ers,

d. Verifying, within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during, that the RPS'cirs'aitry "shorting links" have been removed during:

1. The time any control rod is withdrawn,## or

2. Shutdown marg?n dersnstrations.

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Provided signal to-noise ratio is 1 2. Otherwise, 3 cps.

1. Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.

l SHOREHAM - UNIT 1 3/494 l

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. I REFUELING OPERATIONS f

3/4.9.5 COMUNICATI0NS LIMITINGCONDITIONFOROPER[ ION ,,_ _ ,

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1 3.9.5 Direct communication shall be maintained between the control room and

{ refueling platform personnel, j l APPLICABILITY: OPERATIONAL CONDITION b, during CORE ALTERATIONS.* !

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ACTION: ,

j When direct communication between the control room e d refueling platform !

personnel cunnot be maintained, immediately suspend CORF ALTERATIONS.* l L

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l SURVEILLANCE REQUIREMENTS l-j 4.9.5 Direct communication between the control room and refueling platform i personnel shall be demonstrated wichin one hour prior to the start of and at i 1 east once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during CORE ALTERATIONS.* !

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Except movement of control rods with their normal drive system.

SHOREHAM - UNIT 1 3/4 9 7 Amendm nt No. 10

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REFUELING OPERATIONS .

l 3/4.9.6 REFUELING PLATFORM !

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LIMITING CONDITION FOR OPERATION r 3.9.6 The refueling platform shall be OPERABLE and used for handling fuel .

assemblies or control rods within the reactor pressure vessel.

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APPLICABILITY: During handling of fuel assemblies or control rods within the !

reactor pressure vessel, f

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ACTION: l i

With the requirements for refueling platform OPERABILITY not satisfied, suspend f use of any inoperable refueling platform equipment from operations involving ,

the handling of control rods 4 i fuel assemblies within the reactor pressure j vessel after placing the load in a safe condition. .

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, SURVEILLANCE REQUIREMENTS l 1

, I 4.9.6 Each refueling platform crane or hoist used for handling of control rods :

or fuel assw blies within the reactor pressure vessel shall be demonstrated

OPERABLE w '.nin 7 days prior to the start of such operations with that crane l or hoist by: ,

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, a. Demonstrating operation of the overload cutoff on the main hoist when

the load exceeds 1200 2 50 pounds.

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b. Demonstrating operation of the overload cutoff on the frame mounted ,

q and monorail hoists when the loLd exceeds 1000 2 50 pounds.

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c. Demonstrating the operation of the uptravel mechanical cutoff on the l

) frame mounted and monorail hoists when uptravel brings the top of ;

I active fuel assembly to 7 feet below the normal fuel storage pool !

j water level. ;

f i d. Demonstrating operation of the downtravel mechanical cutoff on the !

main hoist when grapple hook down travel reaches 2 inches above the l 1

top of the fuel guide, f l

e. Demonstrating operation of the slack cable cutoff on the main hoist l

when the load is less than 50 2 25 pounds. ,

) i i f. Demonstrating operation of the loaded interlock on the main hoist I when the load exceeds 485 1 50 pounds. l 1

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g. Demonstrating operation of the leadeo interlock on tne auxiliary {

hoist when the load exceeds 400 t 50 pounds, j J

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i SHOREHAM - UNIT 1 3/4 9-8 l

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ML20196E578

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