ML20132F506

Amends 177 & 164 to Licenses DPR-39 & DPR-48,respectively, Allowing Comed to Control Reactor Coolant Sys Pressure & Temp Limits for Heatup,Cooldown,Low Temp Operation & Hydrostatic Testing

ML20132F506
Person / Time
Site:	Zion  File:ZionSolutions icon.png
Issue date:	12/20/1996
From:	Shiraki C NRC (Affiliation Not Assigned)
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ML20132F504	List: ML20132F498 ML20132F506 ML20132F511
References
NUDOCS 9612240307
Download: ML20132F506 (27)

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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20666 4 001 4 ,o s

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' l COMONWEALTH EDISON COMPANY DOCKET NO. 50-295 ZION NUCLEAR POWER STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 177 License No. DPR-39

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

A. The application for amendment by Commonwealth Edison Company (the licensee) dated July 26, 1996, as supplemented on September 3, 1996, September 18, 1996, two submittals dated October 14, 1996, October 22, 1996, two submittals dated November 8, 1996, and December 17, 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, at amended (the Act) and the Comission's rules and regulatioas 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 (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 Commission'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 Commission's regulations and all applicable requirements have been satisfied.

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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. DPR-39 is hereby amended to read as follows:

9612240307 961220 PDR ADOCK 05000295 P PDR

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(2) Technical Snecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 177 , 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 and shall be implemented within 60 days.

I FOR THE NUCLEAR REGULATORY COMISSION j'

xM W &

Clyde Y. Shiraki, Sen roject Manager Project Directorate II-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: December 20, 1996 l

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M atuq g k UNITED STATES t j j

2 NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20056-0001 o

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00PMONWEALTH EDISON COMPANY DOCKET NO. 50-304 ZION NUCLEAR POWER STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.164 License No. DPR-48

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

A. The application for amendment by Commonwealth Edison Company (the licensee) dated July 26, 1996, as supplemented on September 3, 1996, September 18, 1996, two submittals dated October 14, 1996, October 22, 1996, two submittals dated November 8,1996, and December 17, 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; 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 (i) 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 Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. 1

2. Accordingly, the license is amended by changes to the Technical l Specifications as Odicated in the attachment to this license amendment, i and paragraph 2.C.(2) of Facility Operating License No. DPR-48 is hereby amended to read as follows:

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(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 164 , 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 and shall be implemented within 60 days.

FOR THE NUCLEAR REGULATORY COMISSION l .

CldeY.Shiraki,SeniorProjectManager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: December 20, 1996 l

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4 ATTACHMENT TO LICENSE AMENDMENT NOS. 177 AND 164 FACILITY OPERATING LICENSE NOS. DPR-39 AND DPR-48 DOCKET NOS. 50-295 AND 50-304 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identifieri by the captioned amendment number and contain marginal lines indicating the area of change.

Remove Paaes Insert Paaes 5 5 6 6 79 79 80 80 82 82 83 83 83a 83a 83b 83b 84 84 85 85 86 86 87 87 88 88 89 89 90 90 91 91 92 92 l

93 93 93a 93a l 94 94 105 105 316b 316b 1

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1.0 DEFINITIONS 1.28 OPERATING temperature limits, including heatup and cooldown rates, and the power operated relief valve (PORV)  !

OPERATING is defined as performing the intended lift settings and enable tamperature associated with function in the intended manner. the Low Temperature Overpressure Protection (LTOP)

System, for the current reactor vessel fluence ,

1.29 OPERATING CYCLE period. These pressure and temperature limits shall i be determined for each fivence period in accordance The OPERATING CYCLE shall be the interval between with the Specification 6.6.1.G. Plant operation with  !

the end of one major refueling outage and the end these operating limits is addressed in individual of the next subsequent major refueling outage per Specifications.

unit. .

1.33 PROCESS CONTROL PROGRAM (PCP) '

l.30 OPERATIONAL MODE - MODE i

' The PROCESS CONTROL PROGRAM (PCP) shall contain  !

An OPERATIONAL MODE (i.e. MODE) shall correspond the current formulas, sampling, analyses, tests ,

to any one inclusive combination of core and determinations to be made to ensure that the t reactivity condition, power level, and average processing and packaging of solid radioactive wastes reactor coolant temperature specified in Table based on demonstrated processing of actual or t 1.1, when fuel assemblies are present in the simulated wet solid wastes will be accomplished in reactor vessel. such a way as to assure compliance with 10 CFR Parts 20, 61 and 71, State regulations, burial ground 1.31 PHYSICS TESTS requirements and other requirements governing the -

disposal of radioactive waste.

PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of 1.34 PROTECTION LOGIC CHANNEL the reactor core and related instrumentation and

1) described in Chapter 14.0 of the FSAR, 2) A PROTECTION LOGIC CHANNEL shall be an authorized under the provisions of 10 CFR 50.59, arrangement of relays, contacts or other or 3) otherwise approved by the Commission. components which operate in response to INSTRUNENT CHANNEL outputs to produce a decision 1.32 PRESSURE BOUNDARY LEAKAGE output. The decision output is the initiation of a protective action signal. At the system '

PRESSURE BOUNDARY LEAKAGE shall be leakage (except level, the decision output is the operation of a steam generator tube leakage) through a sufficient number of ACTUATION DEVICES and the ,

non-isolable fault in the Reactor Coolant System associated ACTUATED EQUIPMENT as required to component body, pipe wall, or vessel wall. place or restore the Nuclear Steam Supply System  ;

to a design safe state. The channel is deemed 1.32a PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) to include the ACTUATION DEVICES.

The PTLR is the unit specific document that provides the reactor vessel pressure and 5 Amendment Nos. 177 and 1e

1.0 DEFINITIONS 1.35 PROTECTION SYSTEM 1.40 REFUELING CYCLE OR OUTAGE The PROTECTION SYSTEM shall consist of both the When REFUELING CYCLE or OUTAGE is used to Reactor Protection System and the Engineered designate a surveillance interval, the ,

Safeguards System. The PROTECTION SYSTEM surveillance shall be performed at least once encompasses all electric and mechanical devices every 18 months as allowed by general and circuitry (from sensors through ACTUATION requirement 4.0.2.

DEVICES) which are required to operate in order '

to place or restore the Nuclear Steam Supply 1.41 REPORTABLE EVENT System to a design safe state.

A REPORTABLE EVENT shall be any of those 1.36 PURGE - PURGING conditions specified in Specification 6.6.2 or Section 50.73 of 10 CFR Part 50.

PURGE OR PURGING is the controlled process of discharging air or gas from a confinement to 1.42 SHUTDOWN MARGIN maintain temperature, pressure, humidity, concentration or other operating condition, in SHUTDOWN MARGIN shall be the instantaneous such a manner that replacement air or gas is amount of reactivity by which the reactor is required to purify the confinement. subcritical or would be subcritical from its present condition assuming all control and .

1.37 OUADRANT POWER TILT RATIO shutdown banks are fully inserted, except for the single rod cluster assembly of highest QUADRANT POWER TILT RATIO shall be the ratio of reactivity worth which is assumed to be fully the maximum upper excore detector calibrated withdrawn.

output to the average of the upper excore detector calibrated outputs, or the atio of the maximum 1.43 SITE BOUNDARY

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lower excore detector celibrated output to the  !

average of lower enve detector calibrated The SITE BOUNDARY shall be that line beyond outputs, whichever is greater. which the land is not owned, leased or otherwise i

controlled by the licensee. t 1.38 RATED THERMAL POWER '

l.44 DELETED RATED THERMAL POWER shall be a total steady state reactor core heat transfer rate to the reactor  ;

coolant of 3250 MWt.

1.39 REACTOR PRESSURE The REACTOR PRESSURE shall be the pressure in the steam space of the pressurizer.

6 Amendment Nos. 177 and 164

LINITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.3.2 PRESSURIZATION AND SYSTEM INTEGRITY 4.3.2 PRESSURIZATION AND SYSTEM INTEGRITY A. Heatup and Cooldown A. The reactor coolant temperature and pressure shall be determined to be within RCS Pressure, RCS Temperature, and RCS the limits at least once per 30 minutas heatup and cooldown rates shall be during system heatup, cooldown, and maintained within the limits specified in inservice leak and hydrostatic testing the PTLR. operations.

APPLICABILITY: At all times.

ACTION: With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes-perform an engineering evaluation to determine the effects of the out of limit condition on the structural integrity of the Reactor Coolant System,  ;

determine that the Reactor Coolant System remains acceptable for continued operation or be in at least N00E 3 ,

within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce RCS T ,and pressure to less than 200*F and 500 ,

psig, respectively, wi?.hin the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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79 Amendment Mos.177 and 164

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.3.2 (Continued) 4.3.2 B. Not Applicable B. Not Applicable C. The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the primary and secondary coolant is below 70*F. C. Not Applicable ,

D. The pressurizer heatup rate shall not exceed 100*F/hr and the pressurizer cooldown rate not exceed 200*F/hr. The spray shall not be used if the temperature difference between the D. Not Applicable pressurizer and the spray fluid is greater than 320"F. ,

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E. Hydrostatic Testing

1. System inservice leak and hydrotests shall  !

be performed in accordance with E. Not Applicable the requirements of ASME Boiler and Pressure Vessel Code,Section XI and applicable addenda; except as stated in ,

Specification 4.3.4.C.l.

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80 Amendment Nos. 177 and 164 l

LINITING CONDITION FOR OPERATION SURVEILLANCE REQUIRENENT 3.3.2.G. Low Temperature Overpressure Protection 4.3.2.G. Low Temperature Overpressure Protection

1. At least one of the following low 1. Surveillance and testing of the low temperature overpressure protection methods temperature overpressure protection methods shall be available: shall be performed as follows- ,

a. Two power operated relief valves a. Each PORV shall be demonstrated as  ;

(PORVs) shall be OPERABLE with lift OPERABLE by:

settings within the limits specified in the PTLR, or 1. Performance of a CHANNEL FUNCTIONAL  ;

TEST, but excluding valve i operation, on the PORV actuation channel within 31 days prior to  !

entering a condition in which the PORV is required OPERABLE, and at least once per 31 days thereafter when the PORV is required OPERABLE.

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2. Verifying the PORV backup air supply is charged, within 31 days  ;

prior to entering a condition in ,

which the PORV is required OPERABLE, and at Teast once per 31 days thereafter ~'en the PORV is required OPERABLE. -

3. Performance of a CHANNEL  !

CALIBRATION on the PORV actuation channel at least once per refueling l outage.

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82 Amendment Nos.177 and 164 .-

LINITING CONDITION FOR OPERATION SURVEILLANCE REQUIRENENT i 3.3.2.G. Low Temperature Overpressure Protection 4.3.2.G. Low Temperature Overpressure Protection (Continued) (Continued) j

4. Verifying each PORV's isolation

, valve is open at least once per '

shift when this method is being used for low temperature ,

overpressure protection.

5. Testing pursuant to Specification  ;

4.0.5.

b. The Reactor Coolant System (RCS) b. The RCS pressure shall be verified to pressure shall be less than 100 be less than 100 psig, and pressurizer  ;

psig, and the pressurizer level level shall be verified to be less than '

less than 25%, or 25% at least once per shift, when this i method is being used for low  !

temperature overpressure protection.

c. The RCS is depressurized and one c. Verifying one PORV and it's isolation  !

PORV and it's isolation valve are valve are open at least once per shift, open. when this method is being used for low  !

temperature overpressure protection.

2. A maximum of one* charging pump, 2. At least two of the three charging pumps,* ,

aligned for in.in : ion into the RCS, and all accumulators, and all safety i and no accumulawes and no safety injection pumps, shall be verified to be  !

injection pumps shall be OPERABLE. incapable of injecting into the RCS prior l to entering a condition in which they are i required to be inoperable, and at least i once per shift thereafter while they are required to be inoperable. t o For short durations of time during pump switchover, two charging pumps may be OPERABLE  !

for the purpose of maintaining seal injection i flow to the reactor coolant pumps. i t

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83 Amendment Nos.177 and 164

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.3.2.G. Low Temperature Overpressure Protection 4.3.2.G. Low Temperature Overpressure Protection (Continued) (Cor.tinued)

3. When starting a reactor coolant pump, when 3. Not applicable.  !

no reactor coolant pumps are running, the .

temperature in the steam generator b. With one PORV inoperable in MODES 5 or '

secondary side in any unisolated RCS loop 6, restore the inoperable PORV to t shall be less than 50'F higher than the operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or RCS temperature. within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> either;  !

1. Depressureize the RCS to less than APPLICABILITY: Mode 4 when the temperature of any 100 psig and lower pressurizer .

RCS cold leg is less than or equal to level to less than 25%, or l the LTOP erule temperature specified in the PTLR, MODE 5 and MODE 6 with 2. Depressurize the RCS and open at }

the reactor vessel head on. least one PORY and its block  !

valve.  !

ACTION: a. With one PORY inoperable in MODE 4, ,

restore the inoperable PORV to  !

OPERABLE status within 7 days, or within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> either; ,

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1. Depressurize the RCS to less l-than 100 psig and lower

' pressurizer level to less than 25%, or

2. Depressurize the RCS and open at least one PORV and it's block valve.

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b. With one PORV inoperable in MODES 5 or 6, restore the inoperable PORV to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> either;

1. Depressurize the RCS to less than  ;

100 psig and lower pressurizer level to less than 25%, or 83a Amendment Nos.177 and 164 i

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LINITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.3.2.G. Low Temperature Overpressure Protection 4.3.2.G. Low Temperature Overpressure Protection (Continued) (Continued) i

2. Depressurize the RCS and open at least one PORV and its block valve.

c. With both PORV's inoperable, within  :

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

1. Depressurize the RCS to less than 100 psig and lower pressurizer >

level to less than 25%, or

2. Depressurize the RCS and open at least one PORV and its block '

valve.

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d. In the event that a PORY is used to '

mitigate an RCS pressure transient, a SPECIAL REPORT shall be prepared ,

and submitted to the Commission t pursuant to Specification 6.6.3.B.

The report shall include the following information: -

1. A description of the circumstances initiating the transient, and

2. The effect of the PORV's on the -

transient, and

3. The corrective action necessary to prevent reoccurrence.  ;

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e. The provisions of Specification  ;

3.0.4 are not applicable. '

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89 Amendment Nos.177 and 164 i

Bases 3.3.2 & 4.3.2 FRACTURE TOUGHNESS PROPERTIES All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.

The PTLR contains P/T limit curves for heatup, cooldown, inservice leak and hydrostatic (ISLH) testing, and data l for the maximum rate of change of reactor coolant temperature.

Each P/T limit curve defines an acceptable region for normal operation which has been detemined in accordance with Reference 2. The usual use of the curves is during heatup or cooldown maneuvering, when pressure and temperature  ;

indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.

The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCP8). The vessel is the component most subject to brittle failure, and is as such, the limiting component for which the limitations are based. The limits do not apply to the pressurizer because of its different design characteristics and operating functions. Pressurizer operational limitations are addressed separately within the Technical Specifications.

10 CFR 50, Appendix G, requires the establishment of P/T limits for specific material fracture toughness i requirements of the RCP8 materials. 10 CFR 50, Appendix G requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use of the American Society of Mechanical Engineers (ASME) Code,Section III, Appendix G. ,

The neutron embrittlement effect on the material toughness is reflected by increasing the nil ductility reference temperature (RT ) as exposure to neutron fluence increases.

The actual shift the irradiated in thevesse reactor RT,,*1 material specimens, in accordance with ASTM E 185 and Appendix H of 10 CFR The 50.o operating P/T limit curves will be adjusted, as necessary, based on the evaluation findings and recommendations of Regulatory Guide 1.99.

The P/T limit cures are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit.

The heatup curve represents a different set of restrictions than the cooldown curve because the directions of the thermal gradients through the vessel wall are reversed. The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls during heatup and cooldown, respectively.

90 Amendment Nos. 177 and ma

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4 The criticality limit curve includes the requirement that it be 2 40*F above the heatup curve or the cooldown curve, and not less than the minimum permissible temperature for ISLH testing. However, the criticality curve is not operationally limiting; a more restrictive limit exists in LCO 3.2.1.C, " Reactivity Control & Power Distribution."

The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident. In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components. The ASME Code,Section XI, Appendix E, provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.

The P/T limits are not derived from Design Basis Accident (DBA) analyses. They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, which is an unanalyzed condition. Reference 2 establishes the methodology for determining the P/T limits.

Verification that operation is within the PTLR limits is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes. This Frequency is considered reasonable in view of the control room indication available to monitor RCS status. Also, since temperature rate of change limits are specified in hourly increments, 30 minutes permits assessment and correction for minor deviations within a reasonable time.

, PRESSURIZER LIMITS Although the pressurizer operates in temperature ranges above those for which there is reason for concern of nonductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

INADVERTENT SAFETY INJECTION In the event of an inadvertent safety injection actuation, the affected reactor will trip isenediately, placing the reactor in the hot shutdown condition. After 60 seconds safety injection may be reset and injection terminated as required. An inspection of the primary system while at hot shutdown will prevent possible degradations in the primary system from undergoing further immediate thermal shock imposed during a cooldown. If degradations in the primary system are discovered, an orderly controlled cooldown will be planned to minimize the effects of thermal shod on these degradations on the affected unit.

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REFERENCES

1. ASME Boiler and Pressure Vessel Code,Section III, 1976 Summer Addenda.

2. WCAP-14040-NP-A, " Methodology Used To Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," January 1996.

93a Amendment Mos.177 and 164

dasss: Low Temperaturo Ovsrpressure Prct:cticn '

3.2.2.G There are 3 means of protecting the RCS from overpressurization by a pressure transient at low temperatures.

& The first type of protection is ensured by the operation and surveillance of the power operated relief valves 4.2.2.G with lift settings within the limits specified in the PTLR. A single power operated relief valve (PORV) will relieve a pressure transient caused by 1) a mass addition into a solid RCS from a charging pump or 2) a heat input based on a reactor coolant pump being started in an idle RCS and circulating water into a steam generator whose temperature is 50*F greater than the RCS temperature. (1) ,

The second means of protection is ensured by a PORY being open. It will have the same relieving capabilities as mentioned above.

The third r.:e.s of protection limits the pressurizer level to 25% and the pressurizer pressure to 100 psig.

A pressure transient caused by the inadvertent mass addition from a charging pump running for 10 minutes will  ;

be relieved by the large gas volume and low pressure present in the pressurizer as mentioned above. i Maintaining the pressurizer level below 25% will also make the hi pressurizer level deviation alarm available to the operator during a mass addition accident. +

In the event that a single PORV becomes inoperable in MODE 4, the repair period of 7 days is based on allowing sufficient time to effect repairs using safe and proper procedures and upon the operability of the redundant PORV. Industry experience has shown that the potential for an overpressure transient is greatest in MODE 5 or MODE 6 with the reactor vessel head on. Therefore, a reduced repair period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is specified in these MODES.

In the event that both PORV's become inoperable, the condition is more serious than for a single inoperable PORV, therefore every attempt should be made to depressurize the RCS in a controlled manner as rapidly as possible. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time period to reach the restrictive conditions in the pressurizer represents a reasonable amount of time to meet these conditions under an expedited circumstance.  ;

This LCO has been provided a LCO 3.0.4 exception. This exception is justified because with an inoperable PORY  :

it may be necessary to enter the Applicability of this LCO in order to achieve the conditions necessary to repair the inoperable PORV. ,

The Low Temperature Overpressure Protection System must be tested on a periodic bases consistent with the

, need for its use. A CHANNEL FUNCTIONAL TEST shall be performed prior to enabling the overpressure protection system during cooldown and startup.

The limitations and surveillance requirements on the ECCS equipment provides assurance that a mass addition i pressure transient can be relieved by the operation of a single PORV or the limiting conditions placed on the  !

pressurizer. l The restrictions for startup of a RCP limits the heat input accident to within the relieving capabilities of I a single PORV. 1 (1) Pressure Mitigating Systems Transient Analysis Results July 1977 Westinghouse Owners Group on RCS

Overpressurization.

94 Amendment Mos. 177 and 164 ,

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LINITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.3.4 4.3.4.0. Not Applicable 105 Amendment Nos. 177 and 164

4 I 6.6.1.F.2. (Continued) 6.6.1.G Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE

11) NFSR-0033, Commonwealth Edison Document, LIMITS REPORT (PTLR)

"VIPRE/WRB-1 DNBR Thermal Limit for Westinghouse OFA Fuel," dated October 14, a. RCS pressure and temperature limits for heatup, 1988, by James C. Boerger, et al., approved by cooldown, low temperature operation, criticality, and the NRC SER dated February 13, 1990. (Thermal hydrostatic testing, as well as heatup and cooldown Hydraulic DNBR Safety Limit.) rates shall be established and documented in the PTLR for the following:

3. The core operating limits shall be determined such that all applicable limits (e.g., fuel LCO 3.3.2.A, "Heatup and Cooldown"; and thermal-mechanical limits, core thermal-hydraulic LCO 3.3.2.G, " Low Temperature Overpressure limits, ECCS limits, nuclear limits such as Protection".

shutdown margin, and transient and accident analysis limits) of the safety analysis are met. b. The analytical methods used to determine the RCS pressure and temperature limits shall be those

4. The CORE OPERATING LIMITS REPORT, including any previously reviewed and approved by the NRC, mid-cycle revisions or supplements thereto, shall specifically those described in the following be provided upon issuance, for each reload cycle, documents: NRC letter dated October 16, 1995, to the NRC Document Control Desk with copies to " Acceptance for Referencing of Topical Report the Regional Administrator and Resident WCAP-14040, Revision 1, ' Methodology used to Develop Inspector. Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curve,'" and safety evaluation dated December 20, 1996.

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluency period and for any revision or supplement thereto.

316b Amendment Nos.177 and 164

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