ML20005D862

Amend 35 to License NPF-57,revising Tech Specs by Deleting Refs to Conditions Permitted During Startup Testing Program

ML20005D862
Person / Time
Site:	Hope Creek
Issue date:	12/18/1989
From:	Butler W Office of Nuclear Reactor Regulation
To:
Shared Package
ML20005D863	List: ML20005D862 ML20005D864
References
NUDOCS 9001020102
Download: ML20005D862 (32)

Text

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UNITED STATES

'F NUCLEAR REGULATORY COMMISSION n-j R

WASHINGTON. D. C,20555 5

q e... +,s PUBLIC SERVICE ELECTRIC 8 GAS COMPANY ATLANTIC CITY FLECTRIC COMPANY I

DOCKET NO. 50-354 HOPE CREEK GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE j

Amendment No. 35 -

License No. NPF-57 1.

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

A.

'The application for amendment filed by the Public Service Electric &

Gas Company (PSE8G) dated September 25, 1989 complies with the standards and requirements of the Atomic Energy Act of.1954, as amended (the Act), and the Commission's rules and reculations 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 Commission; 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 conducted in compliance with the Consnission's regulations set forth in 10 CFR Chanter I; 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 reculations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-57 is hereby amended to read as follows:

(2) Technical Soecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.35, and the Environmental Protection Plan contained in Appendix R, are hereby incorporated in the license.

PSEAG shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

9001020102 891218 DR ADOCK 0500 4

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3.

This license amendment is effective as of its date of issuance and shall

- be implemented within 60 days of its date of issuance.

FORTHdNUCLEARREGULATORYCOMMISSION Valter R. Butler, Director Droject Directorate I-2 Division of Reactor Projects I/11 Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

December 18, 1989 r

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ATTACHMENT TO LICENSE AMENDMENT NO. 35 FACILITY OPERATING LICENSE NO. NPF-57 00CKET NO. 50-354 Replace the following pages of the Appendix "A" Technical Specifications with the attached papes. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. Overleaf pages provided to maintain document cor'pleteness.*

Remove Insert 3/4 3-5 3/4 3-5 3/4 3-6 3/4 3-6*

3/4 3-21 3/4 3-21*

3/4 3-22 3/4 3-22 3/4 3-23 3/4 3-23 3/4 3-24 3/4 3-24 3/4 3-25 3/4-3-25 3/4 3-26 3/4 3-26*

3/4 3-47 3/4 3-47 t

3/4 3-48 3/4 3-48*

1 3/4 3-65 3/4 3-65 3/4 3-66 3/4 3-66*

l 3/4 4-7 3/4 4-7*

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

3/4 4-10 3/4 4-10 3/4 4-17 3/4 4-17*

3/4 4-18 3/4 4-18 3/4 5-5 3/4 5-5 3/4 5-6 3/4 5-6*

3/4 10-1 3/4 10-1*

3/4 10-2 3/4 10-2 2/4 10-3 3/4 10-3*

3/4 10-4 3/4 10-4 3/4 10-5 3/4 10-5 3/4 10-6 3/4 10-6*

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. ATTACHMENT TO LICENSE AMENDMENT NO. 3s FACILITY OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354 Remove Insert 3/4 10-7 3/4 10-7*-

3/4 10-8 B'3/4 10-1 B 3/4 10-1 l

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-4 TABLE 3.3.1-1 (Continued)

' REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS i

(a) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system

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is monitoring that parameter.

(b) This function shall be automatically bypassed when the reactor mode switch is in the Run position.

(c) Unless adequate shutdown margin has been demonstrated per Specifica-tion 3.1.1, the " shorting links" shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn *,

(d) The non-coincident NMS reactor trip function logic is such that all channels go to both trip systems.

Therefore, when the " shorting links" are removed, the Minimum OPERABLE Channels Per the Trip System are 4 APRMS, 6 IRMS and 2 SRMS.

(e) An APRM channel is inoperable if there are less than 2 LPRM inputs per level or less than 14 LPRM inputs to an APRM channel.

(f) This function is not required to be OPERABLE when the reactor pressure vessel _ head-is removed per Specification 3.10.1.

(g) This function shall be automatically bypassed when the reactor mode switch is not in the Run position.

(h) This function is not required to be OPERABLE when PRIMARY CONTAINMENT Ilff EGitITY is not required.

(i) With any control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

(j) This function shall be automatically bypassed when turbine first stage pressure is < 159.7 psig equivalent to THERMAL POWER less than 30% of l

RATED THERMAE POWER.

To allow for instrument accuracy, calibration, and drift, a setpoint of 5 135.7 psig is used.

l (k) Also actuates the EOC-RPT system.

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

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l-HOPE CREEK 3/4 3-5 Amendment No. 35 l

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TABLE 3.3.1-2 8

REACTOR PROTECil0N SYSTEM RESPONSE

• TIMES l

o m

n 23 i

RESPONSE IIME m*

FUNCTIONAL UNIT (Seconds) 1.

Intermediate Range Monitors:

a.

Neutron Flux - High NA b.

Inoperative NA 2.

Average Power Range Monitor *:

a.

Neutron Figx - Upscale, Setdown NA i

b.

Flow Biased Simulated Thermal Power - Upscale

< 0.09**

c.

Fixed Neutron Flux - Upscale 7 0.09 d.

Inoperative iilA e.

Downscale NA R

3.

Reactor Vessel Steam Dome Pressure - High

< 0.55 4.

Reactor Vessel Water Level - Low, Level 3' E 1.05 T

5.

Main Steam Line Isolation Valve - Closure 7 0.06 6.

Main Steam Line Radiation - High, High IIA 7.

Drywell Pressure - High NA j

i 8.

Scram Discharge Volume Water Level - High NA i

a.

Float Switch MA 1

b.

Level Transmitter / Trip Unit NA 9.

Turbine Stop Valve - Closure

-< 0.06 10.

Turbine Control Valve Fast Closure, Trip 011 Pressure - Low

< 0.08#

11.

Reactor Mode Switch Shutdown Position IIA 12.

Manual Scram NA

" Net. tron detectors are exempt from response time testing.

Response time shall be seasured from the detector output or from the input of the first electronic component in the channel.

• • Not including simulated thermal power time constant, 6 1 0.6 seconds.

Oneasured free start of turbine control valve fast closure.

4 b

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TABLE 3.3.2-1 (Continued)

A ISOLATION ACTUATION INSTRUMENTATION n

A W

TABLEfi0TATION TRIP FUNCTION VALVES CLOSED BY SIGNAL 7.

RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION a.

. Reactor Vessel Water Level Low, Level 3 3 (HV-F008, HV-F009 HV-F015A & B, HV-F022. HV-F023) b.

Reactor Vessel (RHR Cut-in Permissive) Pressure - High 3 (HV-F008, HV-F009, HV-F015A & B. HV-F022, HV-F023) c.

Manual Initiation R.

3 (HV-F008, HV-F009, HV-F015A & B, HV-F022. HV-F023)

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TABLE 3.3.2-2 Es ISOLATION ACTUATION INSTRUMENTATION SETPOINTS

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ALLOWABLE n

Ni TRIP FUNCTION TRIP.SETPOINT VALUE 1.

PRIMARY CONTAINMENT ISOLATION a.

Reactor Vessel Water Level 1)

Low Low, Level 2

> -38.0 inches *

> -45.0 inches b.

Drywell Pressure - High 31.68psig

-i -136.0 inches 2)

Low Low Low, Level 1 5 -129.0 inches *

$ 1.88 psig c.

Reactor Building Exhaust Radiation - High 5 1x10 3pCi/cc 5 1.2x10 3pCi/cc l

d.

Manual Initiation NA NA 2.

SECONDARY CONTAINMENT ISOLATION a.

Reactor Vessel Water Level -

Low Low, Level 2 3 -38.0 inches

• 3 -45.0 inches g,

b.

Drywell Pressure - High 5 1.68 psig i 1.88 psig

{l4 c.

Refueling Floor Exhaust Radiation - High 5 2x10 3pCi/cc 5 2.4x10 3pCi/cc l

d.

Reactor Building Exhaust Radiation - High 5 1x10 3pCi/cc

$ 1.2x10 3pCi/cc l

e Manual Initiation NA NA 3.

MAIN STEAM LINE ISOLATION a.

Reactor Vessel Water Level -

Low Low Low, Level 1 3 -129.0 inches

• 2 -136.0 inches b.

Main Steam Line

< 3.0 X full power

< 3.6 X full power-Radiation - High, High###

Sackground 6ackground 2>

c.

Main Steam Line g

Pressure - Low 2 756.0 psig 3 736.0 psig d.

Main Steam Line Flow - High 5 108.7 psid

$ 111.7 psid

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2 TABLE 3.3.2-2 (Continued)

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ISOLATION ACTUATION INSTRUMENTATION SETPOINTS

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o ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE x

MAIN STEAM LINE ISOLATION (Continued) e.

Condenser Vacuum - Low 2 8.5 inches Hg vacuum 3 7.6 inches Hg vacuum f.

Main Steam Line Tunnel Temperature - High 5 160 F

$ 172 F g.

Manual Initiation NA NA 4.

REACTOR WATER CLEANUP SYSTEM ISOLATION a.

RWCU a Flow - High 5 56.3 gpm 5 61.3 gpm b.

RWCU A Flow - High, Timer 45.0 seconds 5 t 5 47.0 seconds 45.0 seconds 5 t 5 47.0 seconds

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c.

RWCU Area Temperature - High 5 160 F, 140 F or 135 F***

$ 172 F, 152 F or 147 F***

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RWCU/ Area Ventilation A Temperature - High 5 60 F

$ 70 F e.

SLCS Initiation NA NA f.

Reactor Vessel Water Level -

Low Low, Level 2 1 -38.0 inches *

'1 -45.0 inches g.

Manual Initiation NA NA 5.

REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION a.

RCIC Steam Line a Pressure (Flow) - High

< 598" H O

< 611" H O l

2 2

b.

RCIC Steam Line A Pressure 3.0 seconds < t < 13.0 seconds 3.0 seconds < t < 13.0 seconds (Flow) - High, Timer

,l c.

RCIC Steam Supply Pressure - Low 3 64.5 psig 3 56.5 psig e

d.

RCIC Turbine Exhaust Diaphragm Pressure - High 5 10.0 psig 5 20.0 psig 5

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TABLE 3.3.2-2 (Continued)

ISOLATION-ACTUATION INSTRUMENTATION SETPOINTS z

o ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION (Continued)

RCIC Pump Room Temperature - High 5 160 F 5 172 F e.

f.

RCIC Pump Room Ventilation Duct A Temperature - liigh 5 70 F

< 80 F g.

RCIC Pipe Routing Area Temperature - High

< 160 F, g

h.

RCIC Torus Compartment

~< 172 F g

g Temperature - !!igh 5 128 F

~< 140 F i.

Drywell Pressure - High

< l.68 psig 5 1.88 psig j.

Manual Initiation NA NA

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6.

HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION w

a.

HPCI Steam Line a Pressure 4

(Flow) - High

-< 1032 inches H O

< 1064 inches H O 2

2 a

b.

HPCI Steam Line A Pressure 3.0 seconds < t < 13.0 seconds 3.0 seconds < t < 13.0 seconds (Flow) - High, Timer

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c.

HPCI Steam Supply Pressure - Low

>-100.0 psig

> 90.0 psig d.

HPCI Turbine Exhaust Diaphragm Pressure - High 5 10.0 psig

< 20.0 psig e.

IIPCI Pump Room Temperature - High 5 160 F

_ 172 F f.

HPCI Pump Room Ventilation Ducts a Temperature - High 5 70 F

$ 80 F F

g.

HPCI Pipe Routing Area gg gg Temperature - High

< 160 F

< 172 F k

h.

HPCI Torus Compartment 128 F

'140 F

1. 1. ~

g Temperature - liigh y

i.

Drywell Pressure High 5 1.68 psig i 1.88 psig j.

Manual Initiation NA NA w

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TABLE'3.3.2-2;(Continued)

ISOLATION ACTUATION INSTRUMENTATION SETPOINTS g

ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE m

x 7.

RHR SYSTEM SHUIDOWN COOLING MODE ISOLATION-a.

Reactor Vessel Water Level -

Low, level 3 3 12.5. inches

• 1 11.0 inches b.

Reactor Vessel (RHR Cut-in Permissive) Pressure - High

< 82.0 psig i 102.0 psig-c.

Manual Initiation NA NA Y

^See Bases Figure B 3/4 3-1.

• • • These setpoints are as follows:

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160 F - RWCU pipe chase room 4402-140 F - RWCU pump room and heat exchanger rooms m

135 F - RWCU pipe chase room 4505

1. 30 minute time delay.

1. 1. 15 minute time delay.

1. 1. 1. The hydrogen water chemistry (HWC) system shall not be placed.in service until reactor power reaches 20%

of RATED THERMAL POWER.

After reaching 20% of RATED THERMAL POWER, and prior to operating the HWC system, the normal full power background radiation level and associated trip setpoints may be increased.

to levels previously measured during full power operation with hydrogen injection.

Prior to decreasing below 20% of RATED THERMAL POWER and after the HWC system has been shutoff, the background level and associated setpoint shall be returned to the normal full power values.

If a power reduction event occurs so that the reactor power is below 20% of RATED THERMAL POWER without the required setpoint change, control rod motion shall be suspended (except for scran or other emergency actions) until the necessary p

setpoint adjustment is made.

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TA8LE 3.3.2-3 (50LAT10N SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Secomes)r 1.

PRIMARY CONTAINMENT !$0LAT!0N a.

Reactor Vessel Water Level 1)

Low Low, Level 2 NA 2)

Low Low Low, Level 1 NA b.

Drywell Pressure - High-NA c.

Reactor Building Exhaust Radiation a High NA d.

Manual Initiation NA 2.

SECONDARY CONTAINMENT ISOLATION Reactor Vessel Water Level-Low Low, a.

Level 2 NA b.

Drywell Pressure - High NA c.

Refueling Floor Exhaust Radiation -

< 4.0 High(b)

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d.

Reactor Sullding Exhaust

< 4.0 Radiation - High(b)

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e.

Manual Initiation NA j

3.

MAIN STEAM LINE ISOLATION Reactor Vessel Water Level - Low Low Low, a.

Level 1

< 1.0*/c 13(a)..

' Main Steam Line Radiation - High, High(a)(b) 7 1.0"/I 13 *),,

b.

I c.

Main Steam Line Pressure - Low i 1.0"/7 13 *)..

I d.

Main Steam Line Flow-High I 0.5*/I 13 *)..

I e.

Condenser Vacuus - Low HA

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f.

Main Steam Line Tunnel Temperature - High NA g.

Manual Initiation NA 4

REACTOR WATER CLEANUP SYSTEM !$0LATION a.

RWCU A Flow - High NA b.

RWCU A Flow - High, Timer NA RWCU Area Temperature - High NA c.

RWCU Area Ventilation & Temperature - High

. NA d.

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e.

SLC5 Initiation NA f.

Reactor Vessel Water Level - Low Low, Level 2 NA g.

Manual Initiation NA 5.

REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION RCIC Steam Line A Pressure (Flow) - High NA a.

b.

RCIC Steam Line A Pressure (Flow) - High, Timer NA c.

RCIC Steam Supply Pressure - Low NA "IC Turbine Exhaust Diaphragm Pressure - High NA c.

l HOPE CREEK 3/4 3-26 1

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TABLE-3.3.4.2-1 END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION MINIMUM TRIP FUNCTION OPERABLECHANNE[j)

PER TRIP SYSTEM 1.

Turbine Stop Valve - Closure 2(b) 2.

Turbine Control Valve-Fast Closure 2(D)

M (a)A trip system may be placed in an inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillance provided w

I that the other trip system is OPERABLE.

(b)This function shall be automatically bypassed when turbine first stage pressure is 1 159.7 psig equivalent l to THERMAL POWER less than 30% of RATED THERMAL POWER.

To allow for instrument accuracy, calibration and drift, a setpoint of $ 135.7 psig is used.

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TABLE 3.3.4.2-2 5

A END-0F-CYCLE RECIRCULATION PUMP TRIP SETPOINTS 2

m ALLOWABLE TRIP FUNCTION

TRIP SETPOINT VALUE 1.

Turbine Stop Valve-Closure

< 5% closed-

< 7% closed 2.

Turbine Control Valve-Fast Closure

> 530 psig

- > 465 psig w

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TABLE 3.3.7.1-1 (Continued)

F RADIATION MONITORING INSTRUMENTATION F

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ACTION

- i i

' ACTION 71 With one of the required m:nitors inoperable, place the inoperable

'i a.

channel in the tripped condition within one hour; restore.the inoperable channel to OPERABLE status within 7 days, or, within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, initiate and maintain operation of the control

[

room emergency filtration system in the pressurization mode of i

i operation.

b.

With both of the required monitors inoperable, initiate and b

maintain operation of the control roum emergency filtration g-System in the pressurization mode of operation within one hour.

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L ACTION 72 With the required monitor inoperable, perform area surveys of the monitored area with portable monitoring instrumentation at i

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

ACTION 73 With the required monitor inoperable, obtain and analyze at-1 least one sample of the monitored parameter at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 74 -

With the number of channels OPERABLE less than required by Minimum Channels OPERABLE requirement, release (s) via this' pathway may continue for up to 30 days provided:

t a.

The offgas system is not bypassed, and b.

Gret' samples are taken at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and analyzed within the following 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; Otherwise, be in at least HOT SHUTDOWN witnin 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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HOPE CREEK 3/4 3-65 Amendment No. 35 m

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C TABLE 4.3.7.1-1 5

A RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS n5 OPERATIONAL CHANNEL CONDITIONS FOR CHANNEL FUNCTIONAL CHANNEL

)#fICH SURVEILLANCE INSTRUMENTATION CHECK TEST CALIBRATIO.*J REQUIRED 1.

Control Room i

Ventilation Radiation Monitor 5

M R

1, 2, 3, 5 and

• 2.

Area Monitors a.

Criticality Monitors 1)

New Fuel Storage S

M R

Vault 1

2)

Spent Fuel Storage 5

M R

Pool e

b.

Control Room Direct 5

M R

At all times Radiation Monitor 3.

Reactor Auxiliaries Cooling S

M R

At all times Radiation Monitor 4.

Safety Auxiliaries Cooling S

M R

At all times Radiation Monitor 5.

Offgas Pre-treatment 5

M R

Radiation Monitor O

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REACTOR COOLANT SYSTEM i

i 3/4.4.2 SAFETY / RELIEF VALVES i

SAFETY /REllEF VALVES LIMITING CONDITION FOR OPERATION l

3.4.2.1 The safety valve function of at least 13 of t coolantsystemsafety/reliefvalvesshallbeOPERABLE*gefollowingreactor with the specified code safety valve function lift settings:'*

4 safety-relief valves @ 1108 psig 21%

5 safety-relief valves @ 1120 psig 21%

5 safety-relief valves @ 1130 psig 21%

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

With the safety valve function of two or more of the above listed fourteen a.

safety / relief valves inoperable, be in at least HOT SHUTDOWN within 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 next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.

With one or more safety / relief valves stuck open, provided that suppression pool average water temperature is less than 110'F, close the stuck open safety relief valve (s); 'f unable to close the stuck open valve (s) within 2 minutes or if suppression pool average water temperature is 110'F or greater, place the reactor mode switch in the Shutdown position.

With one or more of the above required safety / relief valve acoustic moni-c.

tors inoperable, restore the inoperable monitors to OPERABLE status within 7 days or be in at least HOT SHUTOOWN 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 SHUTOOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, i

• SRVs which perform as ADS function must also satisfy the OPERABILITY requirements of Specification 3.5.1 ECCS-Operating.

• • The lift setting pressure shall correspond to ambient conditions of the valves I

at nominal operating temperatures and pressures.

1. 5RVs which perform a low-low set function must also satisfy the OPERABILITY requirements of Specification 3.2.2, Safety / Relief Valves Low-Low Set Function.

1 HOPE CREEK 3/4 4-7 1

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS t

-4.4.2.1.The acoustic monitor for each safety / relief valve shall be demonstrated OPERABLE with the setpoint verified to be 5 30% of full open noise level **

by performance of a:

a.

CHANNEL FUNCTIONAL TEST at least once per 31 days, and a b.

CHANNEL CALIBRATION at least once per 18 months *.

4.4.2.2 At least 1/2 of the safety relief valves shall be removed, set pressure tested and reinstalled or replaced witn spares that have been previously set pressure tested and stored in accordance with manufacturer's recommendations at least once per 18 months, and they shall be rotated such that all 14 safety relief valves are removed,-set pressure tested and reinstalled or replaced with spares that have been previously set pressure tested and stored in accordance with manufacturer's recommendations at least once per 40 months.

• The provisions of Specification 4.0.4 are not applicable provided the Surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.

• • Initial setting shall be in accordance with the manufacturer's recommendations.

Adjustment to the valve full open noise level shall be accomplished after the initial noise traces have been analyzed.

HOPE CREEK 3/4 4-8 Amendment No. 35

REACTOR COOLANT SYSTEM SAFETY / RELIEF VALVES LOW-LOW SET FUNCTION LIMITING CONDITION FOR OPERATION 3.4.2.2 The relief valve function and the low-low set function of the following reactor coolant system safety / relief valves shall be OPERABLE with the following settings:

Low-Low Set Function Setpoint* (psio) 12%

Valve No.

Open Close F013H 1017 905 F013P 1047 935 APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

With the relief valve function and/or the low-low set function of one of a,

the above required reactor coolant system safety / relief valves inoperable, restore the inoperable relief valve function and low-low set function to OPERABLE status within 14 days 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 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 />, b.

With the relief valve function and/or the low-low set function of both of the above required reactor coolant system safety / relief valves inoperable, be in at least HOT SHUTDOWN within 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 next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.2.2.1 The relief valve function and the low-low set function pressure actuation instrumentation shall be demonstrated OPERABLE by performance of a:

CHANNEL FUNCTIONAL TEST at least once per 31 days, a.

b.

CHANNEL CALIBRATION, LOGIC SYSTEM FUNCTIONAL TEST and simulated automatic operation of the entire system (ext'uding actual valve actuation) at le6st once per 18 months.

"The lift setting pressure shall correspond to ambient conditions of tne valves at nominal operating temperatures and pressures.

l HOPE CREEK 3/4 4-9

l f-'

REACTOR COOLANT SYSTEM 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS l

LIMITING CONDITION FOR OPERATION 3.4.3.1 The following reactor coolant system leakage detection systems shall be OPERABLE:

The drywell atmosphere gaseous radioactivity monitoring system, l

a.

b.

The drywell floor and equipment drain sump monitoring system, c.

The drywell air cooler condensate flow rate monitoring system, d.

'The drywell pressure monitoring system, and e.

The drywell temperature monitoring system.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

With only four of the above required leakage detection systems OPERABLE, opera-tion 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 /> when the re-quired drywell atmosphere gaseous radioactivity monitoring system, the drywell pressure monitoring system, the drywell temperature monitoring system and/or the drywell air cooler condensate flow rate monitoring system is inoperable; 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 />.

SURVEILLANCE REQUIREMENTS 4.4.3.1 The reactor coolant system leakage detection systems shall be demonstrated OPERABLE by:

a.

Drywell atmosphere gaseous radioactivity monitoring system performance of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, a CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CALIBRATION at least once per 18 months.

b.

The drywell pressure shall be monitored at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and the drywell temperature shall be monitored at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c.

Drywell floor and equipment drain sump monitoring system performance of a CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CALIBRATION TEST at least once per 18 months.-

d.

Drywell air coolers condensate flow rate monitoring system performance of a CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CALIBRATION at least once per 18 months.

I i

HOPE CREEK 3/4 4-10 Amendment No. 35

TABLE 3.4.4-1 5

A REACTOR COOLANT SYSTEM r,

CHEMISTRY LIMITS 5

9 OPERATIONAL CONDITION CHLORIDES CON 00CTIVITY (pehos/cm 925'C) 3-1 5 0.2 ppm 51.0 5.6 5 pH $ 8.6 2 and 3

$ 0.1 ppe

$ 2. 0 5.6 5 pH $ 8.6 At all other times 5 0.5 ppa

$ 10.0 5.3 $ pH $ 8.6 o

i 0

REACTOR COOLANT SYSTEM 3/4.4.5 SPECIFIC ACTIVITY I

LIMITING CONDITION FOR OPERATION 3.4.5 The specific activity of the primary coolant shall be limited to:

-Less'than or equal to 0.2 microcuries per gram DOSE EQUIVALENT I-131, a.

and b.

Less than or equal to 100/E microcuries per gram.

e APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3 and 4.

3 ACTION:

In OPERATIONAL CONDITIONS 1, 2 or 3 with the specific activity of a.

the primary coolant; 1.

Greater than 0.2 microcuries per gram DOSE EQUIVALENT I-131 but less than or equal to 4.0 microcuries per gram DOSE EQUIVALENT i

1-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or greater than 4.0 microcuries per gram DOSE EQUIVALENT I-131, be in at least HOT SHUTDOWN with the main steam line isolation valves closed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

2.

Greater than 100/E microcuries per gram, be in at least HOT SHUTOOWN with the main steam line isolation valves closed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b.

In OPERATIONAL CONDITIONS 1, 2, 3 or 4, with the specific activity of the primary coolant greater than 0.2 microcuries per gram DOSE EQUIVALENT I-131 or greater than 100/E microcuries per gram, perform the sampling and analysis requirements of Item 4a of Table 4.4.5-1 until the specific activity of the primary coolant is restored to within its limit, c.

In OPERATIONAL CONDITION 1 or 2, with:

1.

THERMAL POWER changed by more than 15% of RATED THERMAL POWER in one hour, or 2.

The off gas level, at the SJAE, increased by more than 10,000 microcuries per second in one hour during steady state operation at release rates less than 75,000 microcuries per second, or 3.

The off ges level, at the SJAE, increased by more than 15% in one hour during steady state operation at release rates greater than 75,000 microcuries per second, perform the sampling and analysis requirements of Item 4b of Table 4.4.5-1 until the specific activity of the primary coolant is restored to within its limit.

HOPE = CREEK 3/4 4-18 Amendment No. 35 a

w t.

f

' EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 2.

For the HPCI system, verifying that:

i a)

The system develops a flow of at least 5600 gpm against a i

test line pressure corresponding to a reactor vessel pressure i

of > 200 psig, when steam is being supplied to the turbine

(

at 200 + 15, -0 psig.**

6 '

b)

The suction is automatically transferred from the condensate '

storage tank to the suppression chamber on a condensate storage tank water level - low signal and on a suppression chamber - water level high signal.

3.

Performing a CHANNEL CALIBRATION of the CSS, and LPCI system.

j discharge line " keep filled" alarm instrumentation.

r 4.

Performing a CHANNEL CALIBRATION of the CSS header AP instrumenta-tion and. verifying the setpoint-to be i the allowable value of 4.4 psid.

j.

5.

Performing a CHANNEL CALIBRATION of the LPCI header AP instrumen-tation and verifying the setpoint to be 1 the allowable value of-1.0 psid.

d.

For the ADS:

1 1.

At least once per 31 days, performing a CHANNEL FUNCTIONAL TEST of the Primary Containment Instrument Gas System low-low pressure alarm system.

2.

At least once per 18 months:

a)

Performi b a system functional test which includes simulated i

automatic" actuation of the system throughout its emergency operating sequence, but excluding actual valve actuation, b)

Manually opening each ADS valve when the reactor steam dome pressure is greater than or equal to 100 psig** and observing that either:

1)

The control valve or bypass valve position responds accordingly, or 2)

There is a corresponding change in the measured steam

flow, c)

Performing a CHANNEL CALIBRATION of the Primary Containment Instrument Gas System low-low pressure alarm system and verifying an alarm setpoint of 85 + 2 psig on decreasing pressure.

• • The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.

1 HOPE CREEK 3/4 5-5 Amendment No. 35 L-

P 1

EMERGENCY CORE COOLING SYSTEMS i

3/4 5.2 ECCS - SHUT 00WN LIMITING CONDITION FOR OPERATION 3.5.2 At least two of the following shall be OPERABLE:

l Core spray system subsystems with a subsystem comprised of:

a.

1.

Two OPERABLE core spray pumps, and 2.

An OPERABLE flow path capable of taking suction from at least one of the following water sources and transferring the water through the spray sparger to the reactor vessel:

a)

From the suppression chamber, or b)

When the suppression chamber water level is less than the limit or is drained, from the condensate storage tank i

containing at least 135,000 available gallons of water.

b.

Low pressure coolant injection (LPCI) system subsystems each with a subsystem comprised of:

1.

One OPERABLE LPCI pump, and 2.

An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water to the reactor vessel.

APPLICABILITY:

OPERATIONAL CONDITION 4 and 5*.

ACTION:

a.

With one of the above required subsystems inoperable, restore at least two subsystems to OPERA 3LE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or suspend all operations with a potential for draining the reactor vessel, j

b.

With both of the above required subsystems inoperable, suspend CORE i

ALTERATIONS and all operations with a potential for draining the reactor vessel.

Restore at least one subsystem to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or establish SECONDARY CONTAINMENT INTEGRITY within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, l

• The ECCS is not required to be OPERABLE provided that the reactor vessel head is removed, the cavity is flooded, the spent fuel pool gates are removed, and water level is maintained within the limits of Specification 3.9.8 and 3.9.9.

l HOPE CREEK 3/4 5-6 1

3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.10.1 The provisions of Specifications 3.6.1.1, 3.6.1.3 and 3.9.1 and Table 1.2 may be suspended to permit the reactor pressure vessel closure head and the drywell head to be removed and the primary containment air lock doors to be open when the reactor mode switch is in the Startup position during low power PHYSICS TESTS with THERMAL POWER less than 1% of RATED THERMAL POWER and reactor coolant temperature less than 200'F.

t APPLICABILITY: OPERATIONAL CONDITION 2, during low power PHYSICS TESTS.

ACTION:

With THERMAL POWER greater than or equal to 1% of RATED THERMAL POWER or with the reactor coolant temperature greater than or equal to 200'F, immediately place the reactor mode switch in the Shutdown position.

SURVEILLANCE REQUIREMENTS l

4.10.1 The THERMAL POWER and reactor coolant temperature shall be verified to be within the limits at least once pe' iour during low power PHYSICS TESTS.

l HOPE CREEK 3/4 10-1

[

[

SPECIAL TEST EXCEPTIONS I

3/4.10.2 R00 SEQUENCE CONTROL SYSTEM

. LIMITING CONDITION FOR OPERATION 3.10.2 The' sequence constraints imposed on control rod groups by the rod worth minimizer (RWM) per Specification 3.1.4.1 and by the rod sequence control system (R$CS) per Specification 3.1.4.2 may be suspended by means of bypass switches for the following tests provided that control rod movement prescribed for this testing-is verified by a second licensed operator or other technically qualified member of the unit technical staff present at the reactor console:

a.

Shutdown margin demonstrations, Specification 4.1.1.

b.

Control rod scram, Specification 4.1.3.2.

L c.

Control rod friction measurements.

APPLICABILITY:

OPERATIONAL-CONDITIONS 1 and 2.

ACTION:

.With the requirements of the above specification not satisfied, verify that the RWM and/or the RSCS is OPERABLE per Specifications 3.1.4.1 and 3.1.4.2, respectively.

SURVEILLANCE REQUIREMENTS

.4.10.2 When the sequence constraints imposed by the RSCS and/or RWM are bypassed, verify:

a.

That movement of the control rods from 75% R0D DENSITY to the RSCS low power setpoint is limited to the approved control rod withdrawal sequence during scram and friction tests, b.

That movement of control rods during shutdown margin demonstra-tions is limited to the prescribed sequence per Specification 3.10.3.

c.

Conformance with this specification and test procedures by a second licensed operator or other technically qualified member of the unit technical staff.

HOPE CREEK 3/4 10-2 Amendment No. 35 i

a

SPECIAL TEST EXCEPTIONS 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS LIMITING CONDITION FOR OPERATION 3.10.3 The provisions of Specification 3.9.1, Specification 3.9.3 and Table 1.2 may be suspended to permit the reactor mode switch to be in the Startup position and to allow more than one control rod to be withdrawn for shutdown margin demonstration, provided that at least the following requirements are satisfied.

The source range monitors are OPERABLE with the RPS circuitry shorting a.

links" removed per Specification 3.9.2.

b.

The rod worth minimizer is OPERABLE per Specification 3.1.4.1 and is programmed for the shutdown margin demonstration, or conformance with the shutdown margin demonstration procedure is verified by a second licensed operator or other technically qualified member of the unit technical staff.

The " rod-out-notch-override" control shall not be used during c.

out-of-sequence movement of the control rods, d.

No other CORE ALTERATIONS are in progress.

APPLICABILITY:

OPERATIONAL CONDITION 5, during sheldown margin demonstrations.

ACTION:

With the requirements of the above specification not satisfied, immediately place the reactor mode switch in the Shutdown or Refuel position.

{

E SURVEILLANCE REQUIREMENTS 4.10.3 Within 30 minutes prior to and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> dering the performance of a shutdown margin demonstration, verify that; The source range monitors are OPERABLE per Specification 3.9.2, a.

The rod worth minimizer is OPERABLE with the required progrLm per b.

Specific tion 3.1.4.1 or a second licensed operator or other techni-cally qualified member of the unit technical staff is present and verifies compliance with the shutdown demonstration procedures, and No other CORE' ALTERATIONS are in progress.

c.

HOPE CREEK 3/4 10-3 i

SPECIAL TEST EXCEPTIONS e

i 3/4.10.4 RECIRCULATION LOOPS b

. LIMITING CONDITION FOR OPERATION s'

3.10.4 The requirements of Specifications 3.4.1.1 and 3.4.1.3 that recirculation loops be in operation with matched pump speed may be suspended for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the performance of:

PHYSICS TESTS,'provided that THERMAL POWER does_not exceed 5% of a-L.

RATED THERMAL-POWER.

I l

APPLICABILITY: OPERATIONAL CONDITIONS I and 2, during PHYSICS TESTS.

i, L-ACTION:

With-the above specified time limit exceeded, insert all control rods, a.

b'.

With the above specified THERMAL POWER limit exceeded during PHYSICS TESTS, immediately place the reactor mode switch in the Shutdown position.

SURVEILLANCE REQUIREMENTS 4.10.4.1_ The time during which the above specified requirement has been suspended shall be verified.to be less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at least once per hour during PHYSICS-TESTS.

l 4.10.4.2 THERMAL POWER shall be determined to be less than 5% of RATED THERMAL POWER at>1 east once per hour during PHYSICS TESTS.

5 m '

3f -

I HOPE CREEK 3/4 10-4 Amendment No. 35 4

+:.

.,s 7 -3 n-i,.~O",,

f

)

v SPECIAL TEST EXCEPTIONS

~

.j 3/4.10.5 OXYGEN CONCENTRATION ti O

7.;

LIMITING CONDITION FOR-OPERATION 5

-3 The: mate-iil1 originally contained in this Technical Specification was ' deleted.

O!

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E

.with the issuance of, Amendment No.35 tiowever, to maintain the historical i

J reference.to.this' specification, this section has beenzintentionally left d

blank..

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t HOPE CREEK 3/4 10-5 Amendment No. 35 5

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L SPECIAL TEST EXCEPTIONS i

3/4.10.6 TRAINING STARTUPS LIMITING CONDITION FOR OPERATION 3.10.6 The provisions of Specification 3.5.1 may be suspended to permit one RHR subsystem to be aligned in the shutdown cooling mode during training startups provided that the reactor vessel is not pressurized, THERMAL POWER is less than or equal to 1% of RATED THERMAL POWER and reactor coolant temperature is less than 200'F.

APPLICABILITY:

OPERATIONAL CONDITION 2, during training startups.

AC f!ON:

With the requirements of the above specification not satisfied, immediately place the reactor mode switch in the Shutdown position, i

SURVEILLANCE REQUIREMENTS 4.10.6 The reactor vessel shall be verified to be unpressurized and the THERMAL POWER and reactor coolant temperature shall be verified to be within the limits at least once per hour during training startups.

L l

l l

HOPE CD.EFK 3/4 10-6 l

r SPECIALTESTEkCEQ10NS 3/4.10.7 $PECIAL INSTRUMENTATICN - INITIAL CORE LOADINO t1MITIN0 CONDITION FOR OPERATION 3/4.10.7 with the issuance of Amendment ho.14.The material originally contained i t

reference to this section, Section 3/4.10.7 is intentionally left blankHoweve I

1 F

t t

HOPE CREEK 3/4 10-7 Amendment No.14 48 1W

t:

3/4.10 SPECIAL TEST EXCEPTIONS BASES 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY The requirement for PRIMARY CONTAINMENT INTEGRITY is not applicable during the period when open vessel tests are being performed during the low power PHYSICS TESTS.

L 3/4.10.2 ROD SEQUENCE CONTROL SYSTEM In order to perform the tests required in the technical specifications it is necessary to bypass the sequence re.straints on-control rod movement.

The additional surveillance requirements ensure that the specifications on heat L

generation rates and snutdown margin requirements are not exceeded during the period when these tests are being performed and that individual rod worths do not exceed the values assumed in the safety analysis.

3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS Performance of shutdown margin demonstrations during open vessel testing requires additional restrictions in order to ensure that criticality is properly monitored and controlled.

These additional restrictions are specified in this LCO.

3/4.10.4 RECIRCULATION LOOPS This special test exception permits reactor criticality under no flow conditions and is required to perform certain PHYSICS TESTS while at low THERMAL l

POWER levels.

3/4.10.5 0XYGEN CONCENTRATION The material originally contained in this Technical Specification was deleted with the issuance of Amendment No. 35 However, to maintain the.

historical reference to this specification, this section has been intentionally left blank.

3/4.10.6 TRAINING STARTUPS This special test exception permits training startups to be performed with the reactor vessel depressurized at low THERMAL POWER and temperature while controlling-RCS temperature with one RHR subsystem aligned in the shutdown cooling mode in order to minimize contaminated water discharge to the radioactive waste disposal system.

3/4.10.7 SPECIAL INSTRUMENTATION - INITIAL CORE LOADING The material originally contained in Bases Section 3/4.10.7 was deleted with the issuance of Amendment No. 14.

However, to maintain the historical reference to this section, Bases Section 3/4.10.7 is intentionally left blank.

HOPE CREEK B 3/4 10-1 Amendment No. 35