Text

Amend 105 to License DPR-35,modifying Tech Specs to Ensure Safe Operation During Cycle 8 Following Core Reload 7
	ML20237K895
Person / Time
Site:	Pilgrim
Issue date:	08/31/1987
From:	Nerses V; Office of Nuclear Reactor Regulation
To:	;
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UNITED STATES 8 ' ;,.s r ",'g NUCLEAR REGULATORY COMMISSION

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l WASHINGTON, D. C. 20555 c

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i BOSTON EDIS0N COMPANY i

DOCKET NO. 50-293 PILGRIM NUCLEAR POWER STATION AMENDMENT'TO FACILITY OPERATING LICENSE 1

Amendment No.105 License No. DPR-35 1.

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

1 A.

The application for amendment by Boston Edison Company (the licensee)

{

dated May 22,1Q87 as supplemented by letter dated July 22, 1987, complies with the standards and requirements of the Atomic Energy j

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 confonnity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the attivities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the 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 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 3.8 of Facility Operating License No. DPR-35 is hereby amended to read as follows:

P

' (2) Technical Specifications-

'The Technical Specifications contained in Appendix A as revised through. Amendment No.105, are hereby incorporated in the~ license. The' licensee shall operate the facility in accordance with the Technical Specifications.

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

)

FOR THE NV LEAR REGULATORY COMMISSION

)

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}

Victor Nerses, Acting Director-Project Directorate I-3 Division of Reactor Projects I/II

)

Attachment:

Changes to the Technical Specifications Date of Issuance:

August 31, 1987 l

4 ATTACHMENT TO LICENSE AMENDMENT NO.105 FACILTIY OPERATING LICENSE NO. DPR-35 DOCKET NO. 50-293 Replace the' following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. The corresponding.

overleaf pages are provided to maintain document completeness.

Remove Pages Insert Pages l

7 7

i 8

8 I

46' 46 53 53 59a 59a 6R 68 203 203 205A-1

-205A-1 2058-2 2058-2 205C-3 2050-3 205E-1 205E-1 205E-2 205E-2 205E-3 205E-3 205E-4 205E-4 205E-5 205E-5 205E-6 205E-6 205F 205F' 206m 206m

1 1.1 SAFETY LIMIT 2.1 L1MITING SAFETY SYSTEM SETTING D;

Whenever the reactor is in the In the event of. operation-with a cold shutdown condition with-maximum fraction of limiting' power irradiated fuel in the reactor densi'ty (MFLPD) greater than the vessel, the water level shall not fraction of' rated power (FRP). the I

be less than 12 in, above the top setting shall be modified as

'{

of the normal active fuel zone, follows:

l FRP S 1 (0.58W + 62%)

MFLPD 2 Loop i

Where,

_1 FRP - fraction of rated thermal I

~

power (1998,MWt) l MFLPD = maximum fraction of.

limiting power density where the limiting. power density is 13.4 KW/ft for all fuel.

-l The ratio of FRP.to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value=will be j

used.

For no combination of loop,

recirculation flow rate and core thermal power shall the APRM flux scram trip setting be allowed to exceed 120% of rated thermal power.

b. APRM Flux Scram Trip Setting (Refuel or Start and Hot Standby Mode)

When the reactor mode switch is'in the REFUEL or STARTUP position, the APRM i

I scram shall be set at less than or equal to 15% of 1

rated power.

c. IRM The IRH flux scram setting shall be 1120/125 of s.cale.

B.

APRM Rod Block Trip Setting The APRM rod block trip setting shall be:

I S...$ 0.58W + 50% 2 Loop 7

Amendment No. 42,72,105

__ - _ - _a

1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING i

Where, S.

Rod block setting in I

percent of rated thermal power (1998 MWt)

W Percent of drive flow l

required to produce a l

rated core flow of 69 l

Mlb/hr.

In the event of operating with a maximum fraction limiting power density (MFLPD) greater than the fraction of rated power (FRP), the I

setting shall be modified as follows:

{

FRP S 1 (0.58W + 50%)

MFLPD 2 Loop i

Where, l

FRP - fraction of rated thermal power j

MFLPD = maximum fraction of limiting power density where the limiting power density is 13.4 KW/ft for I

all fuel.

The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual j

operating value is less than the design value of 1.0, in which case l

the actual operating value will be i

used.

C.

Reactor low water level scram I

setting shall be > 9 in. on level instruments.

I I

D.

Turbine stop valve closure scram j

settings shall be 1 10 percent valve closure.

E.

Turbine control valve fast closure l

setting shall be 2150 psig l

control oil pressure at acceleration relay.

F.

Condenser low vacuum scram setting shall be 2 23 in. Hg. vacuum.

G.

Main steam isolation scram setting l

shall be i 10 percent valve l

closure.

Amendment No. #2, 7//,105 8

l l

l

NOTES FOR TABLE 3.2.A 1.

Whenever Primary Containment integrity is required by Section 3.7, there shall be two operable or. tripped trip systems for each function.

2.

Action If the first column cannot be met for one of the trip s,ystems, that trip system shall be tripped.

If the first column cannot be met for both' trip systems, the appropriate action' listed below shall be taken.

A.

Initiate an orderly shutdown and~have the reactor in Cold Shutdown Condition in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

B.

Initiate an orderly load reduction and have Main Steam Lines isolated within eight hours.

C.

Isolate Reactor Water Cleanup System.

D.

Isolate Shutdown Cooling.

3.

Instrument set point corresponds to 128.26 inches above top of active fuel.

4.

Instrument set point corresponds to 77.26 inches above top of active fuel.

5.

Not required in Run Mode (bypassed by Mode Switch).

6.

Two required for each steam line.

7.

These signals also start SBGTS and initiate secondary containment isolation.

8.

Only required in Run Mode (interlocked with Mode Switch).

9.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months orior to the planned start of hydrogen injection test with the reactor power at greater than 20% rated power, the normal full power radiation background level and associated trip setpoints may be changed

':ased on a calculated value of the radiation level. expected during th?

test The background radiation level and associated trip setpoints may be adjusted during the test based on either calculations or measurements of actual radiation levels resulting from hydrogen injection.

The background radiation level shall be determined and associated trip setpoints shall be set within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of re-establishing normal radiation levels after completion of hydrogen injection and prior to establishing reactor power levels below 20% rated power.

Amendment No. 86, 105 46

NOTES FOR TABLE 3.2.8 l.

Whenever any CSCS subsystem is required by Section 3.5 to be coerable, there shall be two (Note 5) operable trip systems.

If the first column cannot be met for one of the trip systems, that system shall be repaired or the reactor shall be placed in the Cold Shutdown Condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after this trip system is made or found to be inoperable.

2.

Close isolation valves in RCIC subsystem.

3.

Close isolation valves in HPCI subsystem.

4.

Instrument set point corresponds to 77.26 inches of active fuel.

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5.

RCIC and HPCI have only one trip system for these sensors.

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l Amendment ho. 105 53

E PNPS TABLE 3.2-G INSTRUMENTATION THAT INITIATES RECIRCULATION PUMP TRIP AND ALTERNATE ROD INSERTION Minimum Number of Operable or Tripped Instrument Channels Per Trip System (1)

Trip Function Trip Level Setting 2

High Reactor Dome 1175 g 15 PSIG Pressure 2

Low-Low Reactor 2 77.26 :nches l

Water Level above the top of the active fuel Actions (1)

There shall be two (2) operable trip systems for each function.

(a)

If the minimum number of operable or tripped instrument channels for one (1) trip system cannot be met, restore the trip system to operable status within 14 days or be in at least hot shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(b)

If the minimum operability condi' ions (l.a) cannot be met for both (2) trip systems, be in at least hot shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l

)

Amendment No. 42, EZ, 105 59a

l BASES:

3.2 In addition to reactor protection instrumentation which initiates a j

reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequences.

This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the core cooling systems, control rod block and standby gas treatment systems.

The objectives of the Specifications are (i) to assure the effectiveness of the protective instrumentation when required by preserving its capability to tolerate a single failure of any component of such systems even during periods when portions of such systems are out of service for maintenance, and (ii) to prescribe 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 j

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 operating 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.A which senses the conditions for which isolation is required.

Such instrumentation must be available j

whenever primary containment integrity is required.

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

The low water level instrumentation set to trip at 128.26 inches above I

the top of the active fuel closes all isolation valves except those in Groups 1, 4 and 5.

Details of valve grouping and required closing times are given in Specification 3.7.

This trip setting is adequate to prevent core uncovery in the case of a break in the largest line assuming a 60 second valve closing time.

Required closing times are less than this.

The low low reactor water level instrumentation is set to trip when reactor water level is 77.26 inches above the top of the active fuel l

(-49" on the instrument).

This trip closes Main 5 team Line Isolation Amendment No. 105 68

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4 o

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

2. The SRM shall /" ave a minimum Spiral Reload of 3 cps excer,t as specified

-in 3 and 4 below.

During spira'i reload, SRM operability'will be' verified by

3. Prior to spiral unloading, the using a portable external source SRM's shall have an initial every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months until the required

~

count rate of >3 cps.

During amount of fuel is loaded to spiral-unloading, the count maintain 3 cps., As an rate on +.he SRM's may drop alternative to the above, up to below 3 cys, two fuel assemblies will be

~

loaded in different cells

4. During spiral reload, each containing control blades around 3

I control cell shall have at-each SRM to obtain the required 3 least one assembly with a cps. Unti'J.these assemblies have

.s minimum exposure of 1000 loaded, the. cps requirement is t

MHD/ST.

not necessary.

1 1

C.

Spent fuel Pool Water Level C. Spent Fuel Pool Water Level a

l Whenever irradiated fuel is

'HbeneverLirradiated fuel is stored in the spent fuel pool, stored in the spent fuel pool,,,.

I the pool water level shall be 1

the water level shall be maintained at or above 33 feet recorded daily.

n.

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(

D.

Multiple Ccntrol Rod Removal D. Multiple Control Rod Removal Any number 4f control rods and/or Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to the control rod drive mechanisms may start of removal of control be removed from the reactor,

rods and/or control rod drive pressure vessel provided that at mechanisms from the core q.

'~

least the fd lowing requirements and/or reactor pressura vpdel l

are satisfied until all control.

and at least once per 24' hours rods and control rod drive thereafter until all control s

mechanisms are reinstalled and rods and contyrsl roj drive.

all control rods are fully mechanisms are reinsta!Ied'and inserted in the core.

all control rods are fully inserted in the core verify-

a. The reactor mode switch is that:

operable and locked in the Refuel position per

a. The reactor mode switch is operable and locked in the Specification 3.10.A, except 4

that the' Refuel positioh "one Refuel position per rod out" interlock may be Specification 3.10.A.

bypassed, as required, for those control rods and/or

b. The SRM r.hannels are I

control. rod drive mechanisms operable per Specification to be removed, after the fuel 3.3.B.4.

assemblies have been removed as specified below.

c. The Reactivity Margin i

requirements of j

b. The source range monitors Spectf.ication 3.3. A.1 are (SRM) are operable per satisfied.

Specification 3.3.B.4.

c. The Reactivity Margin requirements of Specification 3.3.A.1 are satisfied.

2 2

1 203 Amendment flo. U,105, y

+ 1

4 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS B.

Linear Heat Generation Rate (LHGR)

B.

Linear Heat Generation Rate (LHGR)

?

During reactor power operation The LHGR as a function of cbre the linear heat generation rate height shall be che&ked daily (LHGR) of any rod in any fuel during reactor operation at asser$!) at any. axial location

>25% rated thermal power.

Shall not exceed 13.4 kw/ft for all fut).'

l If at any time during operation it is determined oy cormal surveillance that the limiting value for LHGR is being exceeded, action shall be ini'ttated within e

15 minutes to restore operation to within the prescribed limits.

l If the LHGR is not returned to within the prescribed limits seithin two (2) hours, the reactor shall be brought to the Cold Shutdown condition within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . Surveillance and corresponding action shall l

continue untti reactor c;eration is witnin the prescribed limits.

i 1

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1 205A-1 Amer, rims,t No. /t2,105 1

J.

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I TABLE 3.11-1

{

OPERATING LIMIT MCPR VALUES A.

MCPR Operating Limit from Beginning of Cycle (BOC) to BOC + 7,513 MWD /ST.

J P8x8R/BP8x8R For all values of 1 1.36

)

B.

MCPR Operating Limit from BOC + 7,513 MWD /ST to End of Cycle.

T P8x8R/BP8x8R 11 0 1.39 0.0 <t1 0.1 1.40 I

0.1 <t 1 0.2 1.40 0.2 < T 5 0.3 1.41 0.3 <t 1 0.4 1.41

)

0.4 <t 1 0.5 1.42 0.5 <t 1 0.6 1.42 0.6 <t 1 0.7 1.43 0.7 <t1 0.8 1.43 0.8 <t1 0.9 1.44 0.9 <t 1 1.0 1.44 Amendment No. 54, 70. 78, 105 205B-2

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[pC MINIh$pCR!TICAL PflhdR 4ATIO (MCPR[

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'Operatity Limit MCPR 4

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s 3 1

l For any ar/n.o,imaU operating transient 'enalysis evaluation with the initial condition of the reactor being at the steady state operating limit, it is required that thEs,resulting MCPR does not detresse below the Safety Limit n

s' MCPR at any tirre during the transient assuming i;nstruwnt trip setting I

given in Specification 2.1.

y s

Lie differenkShnen the specified Operating Limit flCPR in Specification fliC and the Safety Limit MCPR in'Specif5 cation 1.lA defines the,lVr;es t reductivhit critical power ratio (CM) permitted

, sduring any anti::ipated abnormal operating transient.

To ensure that this

< reduction is not exceeded, the most limiting traisients arcyanalvzed for

' earp reload and fuel type tKdetermine that transient which yielt s the j

lar.;est value rif 3CPR.

This value, when added to the Safety Lidt MCPR tcust be less than the n.inime:r operat':ng limit'MCPR's of Specification.3.11.C.

The result of 'tMy evaluation is documented in the s

'Suppleaentai Reload Licens'ng Sut,mittal" for the current reload.'

s

['

.Tfe evaluation of a given transient ' eg'.ns with the system input v

'l pa'rweters shown in Taus 5-4, 5-6 and 5-6 of NEDE-240)l-P(,

Supplercerited by reload unique inputs given in the current Supplemental i

Reload '.lcensing Submittal.

Tr.'esa values are input-to a GE core dynamic

'behvio* transtent computer program described in NEDO-10802 58'.

The trarsfer.t code used for all pressurization events is described in f EC244154-P (Peterence 5).

The MCPR analysis for pressurization events; is done in accc,rdance with"the precidures given in Reference 6.

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l 5.0 MAJOR QESJGN FEATURES 5.1 SITE FEATURES

~

Pilgrim Nuclear Power Station is located on the.Hestern Shore of Cape Cod Bay in the Town of Plymouth, Plymouth County, Massachusetts. ' The site'is located at approximately 41'51' north latitude and-70'35' west longitude on the Manomet Quadrangle, Massachusetts, Plymouth County 7.5 Minute Series (topographic) map issued by U.S. Geological Survey. UTM coordinates are 19-46446N-3692E.

The reactor (center line) is located approximately 1800 feet from the

')

nearest property boundary.

5.2 REACTOR A.

The core shall consist of not more than 580 fuel assemblies.

B.

The reactor core shall contain 145 cruciform-shaped control rods. The control materials shall be either boron carbide powder (B C) compacted to approximately 707. of theoretical 4

density or a combination of boron carbide powder and solid

{

hafnium.

J

-]

5.3 REACTOR VESSEL The reactor vessel shall be as described in Table 4.2.2 of the FSAR.

The applicable design codes shall be as described in-Table 4.2.1 o' the FSAR.

I 5.4 CONTAINMENT A.

The principal desigr} parameters for the primary containment shall be as given in Table 5.2.1 of the FSAR.

The applicable J

design codes shall be as described in Section 12.2.2.8 of the

]

FSAR.

B.

The secondary containment shall be as described in Section 5.3.2 of the FSAR, C.

Penetrations to the primary containment and piping passing through such penetrations shall be designed in accordance with standards set forth in Section 5.2.3.4 of the FSAR.

j Amendment No. #2, 78, 98,105 206m I

ML20237K895

Text

Attachment:

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