ML20058G266
| ML20058G266 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png (DPR-39-A-151, DPR-48-A-139) |
| Issue date: | 11/29/1993 |
| From: | Dyer J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20058G257 | List: |
| References | |
| GL-90-06, GL-90-6, NUDOCS 9312090158 | |
| Download: ML20058G266 (15) | |
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UNITED STATES i
l' NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D.C. 20555-0001 j
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COMMONWEALTH EDISON COMPANY DOCKET NO. 50-295 ZION NUCLEAR POWER STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE i
Amendment No.151 l
License No. DPR-39 l
1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Comonwealth Edison Company (the licensee) dated July 8, 1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in l
10 CFR Chapter I; l
B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the l
Comission; l
C.
There is reasonable assurance (1) that the activities authorized t
j by this amendment can be conducted without endangering the health i
and safety of the public, and (ii) that such activities will be i
t conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be. inimical to the common defense and security or to the health and safety of the public and E.
The issuance of this amendment is in accordance with 10 CFR l
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, l
and paragraph 2.C.(2) of Facility Operating License No. DPR-39 is hareby.
i amended to read as follows-i i
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(2)
Technical Specifications l
The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 151, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
l 3.
This license amendment is effective as of the date of its issuance and I
shall be implemented within 30 days.
FOR THE NUCLEAR' REGULATORY COMMISSION ames E. Pyer, Director Project. Directorate III-2 Division of Reactor Projects - III/IV/V-Office of Nuclear Reactor Regulation
Attachment:
l Changes to the Technical Specifications i
i Date of Issuance: November 29, 1993 i
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o, UNITED STATES 4
8 NUCLEAR REGULATORY COMMISSION r.
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COMMONWEALTH EDISON COMPANY i
l DOCKET NO.' 50-304 ZION NUCLEAR POWER STATION. UNIT 2 l
AMENDMENT TO FACILITY OPERATING LICENSE i
i Amendment No. 139 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 8,1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the i
Act) and the Commission's rules and regulations set forth in l
10 CFR Chapter I; l
B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the i
Commission; l
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 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 l
Part 51 of the Commission's regulations and all applicable
- i requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical j
Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-48 is hereby:
l amended to read as follows:
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(2)
Technical Soecifications The. Technical Specifications contained in Appendices A and B, as l
revised through Amendment No. 139, are hereby incorporated in the license. The licenseeishall 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 30 days.
f FOR THE NUCLEAR REGULATORY COMMISSION
<.f.p I
mes E. Dyer, Director l
Project' Directorate III-2.
.i Division of Reactor Projects - III/IV/V
-i Office of Nuclear Reactor Regulation
Attachment:
I Changes to the Technical l
Specifications Date of Issuance: November 29, 1993 1
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ATTACHMENT TO LICENSE AMENDMENTS i
AMENDMENT NO.151 TO FACILITY OPERATING LICENSE NO. DPR-39 AMENDMENT NO.139 TO FACILITY OPECTING LICENSE NO. DPR-48 DOCKET NOS. 50-295 AND 50-304:
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Revise the Appendix A Technical Specifications by removing the pages i
identified below and inserting the attached pages.. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.
1 Remove Peggi Insert Paaes j
77 77 77a 77a l
77b' 77b 77c 77c I
77d 78a 78a 78b 82 82 83 83
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83a 83a 83b 83b 94 94 1
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1 1
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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.3.1.E.3 (Continued) 4.3.1.E.3 (Continued) f.
The loop Stop Valve Interlock System of f.
The volume of water needed to refill the the isolated loop is not required to be isolated loop shall be measured to OPERABLE when opening reactor coolant ensure that the 25% add.tlonal boron loop isolation valves provided that the concentration is maintained taking into reactor is in COLD SHUTDOHN condition account the residual water in the with sufficient boron concentration for drained loop. The required boron 70*F operation, and the isolated loop concentration shall be verified during has been drained and refilled, borated refilling by sampilng the refilling to a baron concentration of 25% greater water, prior to opening the isolation than that required for COLD SHUTDOHN.
valves by sampilng the loop, and by sampilng the loop every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to 3.3.1.F.
Relief Valves initial pump operation.
The refilling water sample and the isolated loop l
Both power operated relief valves (PORVs) sample shall be taken and analyzed and their associated block valves shall be independently by different personnel.
4.3.1.F.
Relief Valves A_PP_IdCABILITY: Modes 1,
2, 3 1.
In addition to the requirements of ACI1QN:
a.
With one or both PORVs inoperable Specification 4.0.5, each PORV shall be because of excessive seat leakage, demonstrated OPERABLE at least once per within I hour either restore the 18 months by:
PORV(s) to OPERABLE status or close the associated block valve (s) with power a.
Operating the PORV through one maintained to the block valve (s);
complete cycle of fc11 travel during otherwise, be in at least H00E 3 within MODES 3 or 4, the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MODE 4 within-i the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The provisions b.
Operating air control valves and of LCO 3.0.4 are not applicable to check valves on associated air PORVs which are inoperable due to accumulators in the PORV control excessive seat leakage.
system through one complete cycle of full travel, and c.
Performance of a CHANNEL CALIBRATION of the actuation instrumentation.
77 Amendment Nos. 151 and 139
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t LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT b.
With one PORV inoperable due to causes 2.
Both block valve (s) shall be other than excessive seat leakage, demonstrated OPERABLE at least onca per within I hour, either restore the PORV quarter by OPERATING the valve tht hgh to OPERABLE status or close its one complete cycle of full travel
.iless associated bloc' <alve and remove power the block valve is closed in order to from the block valve; restore the PORV meet the requirements of ACTION b or c to OPERABLE status within the following in Specification 3.3.1.F.
'72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in MODE 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 in MODE 4 within the 3.
The standby AC on-site power supply following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
(diesel generators) for the PORV and block valves shall be demonstrated c.
Hith both PORVs inoperable due to OPERABLE at least once per 18 months by causes other than excessive seat transferring motive and control power leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore from the normal to,the standby AC at least one PORV to OPERABLE status or on-site power supply and OPERATING the close the associated block valves and valves through a complete cycle of full remove power from the block valves and travel.
be in MODE 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 in M0GE 4 within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
d.
With one or both block valves Inoperable, within 1-hour restore the block valves (s) to OPERABLE status or verify its associated PORV(s) is closed with power removed. Restore at least one block valve to OPERABLE status within the next hour if both block valves are Inoperable; restore any remaining block valve to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; otherwise be in at least MODE 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 in MODE 4 within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
77a Amendment Nos. 151 and 139
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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.3.1 4.3.1 G.
Reactor Vessel Head Vent System (per Unit)
G.
Reactor Vessel Head Vent System At least one reactor vessel head vent path Each reactor vessel vent path shall be consisting of at least two valves in series demonstrated OPERABLE at least once per 18 powered from an emergency bus shall be months by:
OPERABLE and closed.*
1.
Verifying all manual isolation valves in
&EEllCABJL_LTl:
Modes 1, 2, 3 and 4.
each vent path are in the open position.
6CIIQN:
2.
Cycling each valve in the vent path through at least one complete cycle of a.
Hith both reactor vessel head vent paths full travel # rom the control room during inoperable, STARTUP and/or P0HER OPERATION MODE 5 or MODE 6.
may continue provided the inoperable vent paths are maintained closed with power 3.
Verifying flow through the reactor removed from the valve actuator of all the vessel head vent paths during MODE 5 or valves in the inoperable vent paths; MODE 6.
restore at least one Inoperable vent path to OPERABLE status within 30 days, or, be in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MODE 5 within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Power may be removed from the valves to prevent inadvertent actuation.
77b Amendment Noss 151 and 139
Bases:
3.3.1 and 4.3.1 Opeta_ tion _aLComp_onstItts The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR equal to or greater than the applicable design limit ONBR (1) during all normal operations and anticipated transients.
In MODES 1 and 2 with one reactor coolant loop not in operation this specification requires that the plant be in at least HOT SHUTDOWN within I hour.
In MODE 3, a single reactor coolant loop provides sufficient heat removal capability to remove decay heat from the reactor.
However, for Hot Zero Power (HZP) conditions operation is bounded by the following three accidents:
1) steamline break, 2) rod ejection, and 3) bank withdrawal from subtritical.
These accidents all assume four reactor coolant pumps in operation.
For the steamline break and the rod ejection accidents, the operation of only a single reactor coolant pump will not impact the conclusions presented in the FSAR.
But with less than four reactor coolant pumps in operation, the margin of safety may be reduced during a bank withdrawal accident.
(
Reference:
Hestinghoue letter, H. J. Johnson to J. S. A*1el, dated June 6, 1984 (CHE-84-579)).
To ensure that the plant is maintained in an analyzed condition it is necessary to either maintain all four reactor coolant loops in operation, or prevent the bank withdrawal accident from occurring by ensuring the reactor trip breakers are open or the rod drive motor generator (H.G.) sets are de-energized.
Single failure considerations require two reactor coolant loops OPERABLE for the purpose of decay heat removal.
In MODES 4 and 5, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE.
Thus, if the reactor coolant loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE.
The operation of one Reactor Coolant Pump or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reduction in the Reactor Coolant System.
The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.
During Mode 5, one reactor coolant loop capable of providing natural circulation may also be substituted for an OPERABLE RHR loop. At least one RHR loop must be OPERABLE at all times.
When the Unit has been in H0DE 5 for an extended period of time, decay heat addition to the reactor coolant is very slow.
During these periods, the requirement for an operating RHR pump may be relaxed to require two RHR pumps to be OPERABLE. Operation with no forced coolant flow is considered to be an unusual and undesirable mode of plant operation.
If it becomes necessary to interrupt forced coolant flow for testing, maintenance, or any other reason, the Technical 77c Amendment Nos. 151 and 130
e 4
Specification will assure that the onset of potential bolling can be de:ected and core cooll'ng initiated before bolling occurs.
The rate of reactor coolant system heatup with no forced flow is estimated to be less than 6*F per minute after I day, 5'F per minute after 4 days, and 3'F per minute after I month of plant shutdown.
This estimate is based on no natural circulation flow through the coolant loops and reactor vessel water level at the centeritne of the nozzles.
The reactor vessel head vent system is provided to exhaust noncondensible gases
- rom the reactor vessel that could inhibit natural circulation core cooling.
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j' 77d
~ Amendment Nos. 151 and 139,
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The OPERABILITY of the PORVs and block valves is Testing of the PORVs in MODE 3'or MODE 4 is required in datermined on the bases of their being capable of order to simulate the temperature and pressure performing the following functions:
environmental effects on the PORV.
A.
Manual control of the PORV to control reactor coolant system pressure.
This is.a function that is used for the steam generator tube rupture accident for plant shutdown.
B.
Maintaining the integrity of the reactor coolant pressure boundary. This is a function that is related to controlling identified leakage and ensuring the ability to detect unidentified reactor coolant pressure boundary leakage.
C.
Manual control of the block valves to: (1) unblock an isolated PORV to allow it to be used for manual control of-reactor system pressure (Item A), and (2) isolate a PORV with excessive seat leakage (Item B).
D.
Manual control of a b'eck valve-to isolate a stuck-open PORV.
BQcause the plant can meet the above functions while in Action a., continued plant operation'is allowed in this Action. A LCO 3.0.4 exception is provided for Action
- a. for this_ reason.
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Surveillance Requirements provide the-assurance that the PORV and block valve can. perform'their functions.
The: block valves are exempt:from the surveillance rCquirements to cycle;the valves when they have.been closed to comply with'the ACTION requirements.
This precludes the need to cycle the valve with full _ system-differential pressure or when maintenance is'being performed to restore an inoperable PORV to operable status.
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n 78a Amendment Nos. 151 and 139
Bases:
3.3.1 S_ team _Gemrators 4.3.1 The Surveillance Requirements for inspection of primary-to-secondary leakage less than this limit during the steam generator tubes ensure that the operation will have an adequate margin of safety to structural integrity of this portion of the RCS withstand the loads imposed during normal operation and will be maintained.
The program for inservice by postulated accidents. Operating plants have inspection for steam generator tubes is based demonstrated that primary-to-secondary leakage of 500 on a modification of Regulatory Guide 1.83, gallons per day per steam generators can readily be Revision 1.
Inservice inspection of steam detected by radiation monitors of steam generator generator tubing is essential in order to blowdown.
Leakage in excess of this limit will require maintain survelliance of the conditions of the plant shutdown and an unscheduled Inspection, during tubes in the event that there is evidence of which the leaking tubes will be located and plugged.
mechanical damage or progressive degradation due to design, manufacturing errors, or Hastage-type defects are unlikely with proper chemistry inservice conditions that lead to corrosion.
treatment of the secondary coolant.
However, even if a Inservice inspection of steam generator tubing defect should develop in service, it will be found also provides a means of characterizing the during scheduled inservice steam generator tube nature and cause of any tube degradation so examination.
Plugging or repairs will be required for that corrective measures can be taken.
all tubes with imperfections exceeding the 2 40% limit of the tube nominal wall thickness.
Steam generator The plant is expected to be operated in a tube inspections of operating plants have demonstrated manner such that the secondary coolant will be the capability to reliably detect degradation that has maintained within those chemistry limits found penetrated 20% of the original tube wall thickness.
to result in negligible corrosion of the steam generator tubes.
If the secondary coolant Hhenever the results of any steam generator tubing chemistry is not maintained within these inservice inspection fall into Category C-3, these limits, localized corrosion may likely result results will be promptly reported to the Commission in stress corrosion cracking.
The extent of pursuant to Specification 6.6.2 prior to resumption of cracking during plant operation would be plant operation. Such cases will be considered by the limited by the limitation of steam generator Commission on a case-by-case basis and may result in a tube leakage between the primary coolant system requirement for analysis, laboratory examinations, and the secondary coolant system tests, additional eddy-current inspection, and revision (primary-to-secondary leakage - 500 gallons per of the Technical Specifications, if necessary, day per steam generator). Cracks having a 78b Amendment Nos. 151 and 139
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 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.
Survelliance and testing of the low l
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 a lift OPERABLE by:
setting of 435 psig, or 1.
Performance of a CHANNEL FUNCTIONAL TEST, but excluding valve 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.
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 least once per 31 days thereafter when the PORV is required OPERABLE.
3.
Performance of a CHANNE'l CALIBRATION on the PORV actuation channel at least once per refueling outage.
4 i
82 Amendment Nos. 151 and 139-
LIMITING CONDITION "0R OPERATION SURVEILLANCE REQUIREMENT 4.3.2.G Low Temperature Overpressure Protection 3.3.2.G Low Temperature Overpressure Protectie r 4
(Continued)
(Continued) 4.
Verifying each PORV's isolation valve is open at least once per shift when this method l' ig used for low temperature overp
.are I
protection.
a 5.
Testing pursuant to Specification b.
The Reactor Coolant System (RCS) 4.0.5.
pressure shall be less than 100 psig, and the pressurizer level less than b.
The RCS pressure shall be verified to be 25%, or lest than 100 psig, and pressurizer level shall be verified to be less than 25% a least once per shift, when this method is being used for low temperature
4 i
and its isolation valve are open.
c.
Verify one PORV and its isolation valve are open at least once per shift, when this method is being used for low 2.
A maximum of one* charging pump or safety temperature overpressure protection.
injection pump, aligned for injection into the RCS, and no accumulators shall be 2.
At least four of the five pumps (charging OPERABLE.
pumps and safety injection pumps), and all l
accumulators, shall be verified to be incapable of injecting into the RCS prior to entering a condition in which they are required to be inoperable, and at least once
- For short durations of time during. pump switchover, per shift thereafter while they'are required two charging pumps may be OPERABLE for the purpose of to be inoperable.
maintaining seal-injection flow to the reactor 1
coolant pumps.
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83 Amendment Nos, 151 and 139
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1 LlHITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT d
3.3.2.G Low Temperature Overpressure Protection 4.3.2.G Low Temperature Overpressure Protection 1
(Continued)
(Continued) 3.
When starting a reactor coolant pump, when 3.
Not applicable, no reactor coolant pumps are running, the temperature in the secondary side of the steam generator in the loop in which the reactor coolant pump is to be started shall he less than 50*F higher than the RCS temperature.
APEllCAB_III_IX:
Mode 4 when the temperature of any RCS cold leg is less than or equal to 250*F, MODE 5 and MODE 6 with the reactor vessel head on.
&CIlQN:
a.
With one PORV 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; 1.
Depressurize the RCS to less~than 100 ps'.g and lower pressurizer leval to less than 25%, or l-2.
Depressurize the RCS and open at least one PORV and its block valve.
b.
With one PORV;lnoperable'in MODES'S or 6, restore the inoperable PORV to i
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;
.l.
Depressurize the RCS to less.than i.
100 psig and lower pressurizer l
level to less than 25%, or 22.Depressurize the RCS and open at least one PORV and its block valve.
t 83a Amendment Nos. 151 and 139 i
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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.3.2.G Low Temper.iture Overpressure Protection 4.3.2.G. Low Temperature Overpressure Protection (Continued)
(Continued) 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 257., or 2.
Depressurize the RCS and open at least one PORV and its block valve.
l d.
In the event that a PORV is used to mitigate an RCS pressure transient, a SPECIAL REPORT shall be preMred and submitted to the Commission cursuant to Specification 6.6.3.B.
The report shall include the fol!owing 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.
l e.
The provisions of Specification 3.0.4 are not applicable.
t
.l 83b
' Amendment Nos. 151 and 139
i Bases:
Low Temperature Overpressure Protectn'+
[
3.2.2.G There are 3 means of protecting the RCS from overpressurization by a pressure transie'nt at low temperatures
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(below 250*F).
The first type of protection is ensured by the operation and surveillance of the power
- 4. 2.2.G operated relief valves with a lift setting of 435 psig.
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 PORV being open.
It will have the same relieving capabilities as mentioned above.
i The third means of protection ilmits 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 i
i relieved by the large gas volume and low pressure present in the pressurizer as mentioned above. Maintaining the pressurizer level below 25% will also make the hl pressurizer level deviation alarm available to the operator during a mass addition accident.
j In the event that a single PORV becomes inoperable in MODE 4, the repair period of 7 days is based on allowing l
sufficient time to effect repairs using safe and proper procedures and upon the operability of the redundant PORV.
j Industry experience has shown that the potential for an overpressure transient is greatest in MODE 5 or MODE 6 with j
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 PORVs become inoperable, the condition is more serious than for a single inoperable PORV, l
i therefore every attempt should be made to depressurize the RCS in a controlled marner as rapidly as possible.
The-
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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 tol-meet these conditions under an expedited circumstance.
t This LCO has been provided a LCO 3.0.4 exception.
This exception is justified because with an inoperable PORV lt may be necessary to enter the Applicability of this LCO in order to achieve the conditions necessary to repair the j
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.
4 The limitations-and surveillance requirements on the ECCS equipment provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV or the limiting conditions placed on the pressurizer.
The restrictions for startup of a RCP limits the heat input accident to within the relieving capabilities of a j
single PORV.
t (1) Pressure Mitigating Systems Transient Analysis Results July 1977 Westinghouse Owners Group on RCS l
j Overpressurization.
i i
94 Amendment Nos. 151 and 139
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