ML20213G471
| ML20213G471 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 05/11/1987 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20213G475 | List: |
| References | |
| NUDOCS 8705180307 | |
| Download: ML20213G471 (12) | |
Text
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UNITED 5TATES NUCLEAR REGULATORY COMMISSION
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PHILADELPHIA ELECTRIC COMPANY DOCKET NO. 50-352 LIMERICK GENERATING STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 4 License No. NPF-39 1.
The Nuclear Regulatory Comission (the Comission) has found that I
A.
The application for amendment by Philadelphia Electric Company (the licensee) dated February 11, 1987, 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 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 cenpliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
Accordingly, the license is amended by changes to the Technical Specifications 2.
as indicated in the attachment to this license amendment, and paragraph 2.C.(2)
I of Facility Operating License No. NPF-39 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B as revised through Amendment No. 4
, are hereby incorporated in'to this license. Philadelphia Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protectinn Plan.
P B705180307 870511 PDR ADOCK 05000352 P
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
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Walter R. Butler, Director Project Directorate I-2 Division of Reactor Projects I/II-
Attachment:
Changes to the Technical Specifications Date of Issuance: May 11, 1987 PDI-?/Df-l I
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This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR RFGULATORY COMISSION Walter R. Butler, Director Pro.iect Directorate I-2 Division of Reactor Projects I/II
Attachment:
Changes to the Technical Specifications Date of Issuance: May 11, 1987 l
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1 I
ATTACHMENT TO LICENSE AMENDMENT NO. 4 FACILITY OPERATING LICENSE NO. NPF-39 DOCKET N0. 50-352 Replace the following pa9es of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. Overleaf page(s) provided to maintain document completeness.*
Remove Insert 3/4 3-57*
3/4 3-57*
3/4 3-58 3/4 3-58
~
3/4 3-5f 3/4 3-59 3/4 3-60*
3/4 3-60*
3/4 9-3 3/4 9-3 3/4 9-4 3/4 9-4 B 3/4 9-1 B 3/4 9-1 B 3/4 9-2*
B 3/4 9-2*
6 O
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N.
INSTRUMENTATION 3/4.3.6 CONTROL R00 BLOCK INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.6.
The control rod block instrumentation channels shown in Table 3.3.6-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.6-2..
APPLICA81LITY: As shown in Table 3.3.6-1.
ACTION:
a.
With a control rod block instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.6-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
b.
With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, take the ACTION required by Table 3.3.6-1.
SURVEILLANCE REQUIREM' 'S 4.3.6 Each:bf he above required control rod block trip systems and instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.6-1.
O LIMERICK - UNIT 1 3/4 3-57
f TABLE 3.3.6-1 E
CONTROL ROD BLOCK INSTRUMENTATION r-m5 MINIMUM APPLICA8LE OPERABLE CHANNELS OPERATIONAL R
TRIP FUNCTION PER TRIP FUNCTION
' CONDITIONS ACTION 1.
R00 BLOCK MONITOR (a) 4 i
E a.
Upscale 2
1*
60 Z
b.
Inoperative 2
1*
60 s
c.
Downscale 2
1*
60 2.
APRM a.
Flow Biased Neutron Flux -
Upscale 4
1 61 b.
Inoperative 4
1,2,5 61 c.
Downscale 4
1 61 d.
Neutron Flux - Upscale, Startup 4
2, 5 61 3.
SOURCE RANGE MONITORS ***
Detector not full in(b)
R a.
3 2
61 i
2 5
61 J,
b.
Upscale (c) j l
Inoperative (c) j c.
6 d.
Downscale(d)
\\
3 l
4.
INTERMEDIATE RANGE MONITORS a.
Detector not full in 6
2, 5 61 b.
Upscale 6
2, 5 61 c.
Inoperati d.
Downscale }
6 2, 5 61 l
6 2, 5 61 5.
Water Level-High 2
1, 2, 5**
62 6.
REACTOR COOLANT SYSTEM RECIRCULATION FLOW a.
Upscale 2
1 62
.n, b.
Inoperative 2
1 62 c.
Comparator 2
1 62 7.
REACTOR MODE SWITCH SHUTOOWN POSITION 2
3, 4 63 l
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TABLE 3.3.6-1 (Continued)
CONTROL R00 WITHDRAWAL BLOCK INSTRUMENTATION ACTION STATEMENTS ACTION 60 Declare the RBM inoperable and take the ACTION required by Specification 3.1.4.3.
ACTION 61 With the number of.0PERABLE channels one or more less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within one hour.
ACTION 62 With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within one hour.
With the number of OPERABLE channels less than required by the ACTION 63 Minimum OPERABLE Channels per Trip Function requirement, initiate a rod block.
NOTES With THERMAL POWER > 30% of RATED THERMAL POWER.
With more than one control rod withdrawn.
Not applicable to control rods l
removed per Specification 3.9.10.1 or 3.9.10.2.
These channels are not required when sixteen or fewer fuel assemblies, adjacent to the SRMs, are in the core.
(a) The RBM shall be automatically bypassed when a peripheral control rod is selected or the reference APRM channel indicates less than 30% of
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RATED THERMAL POWER.
(b) This function shall be automatically bypassed if detector count rate is
> 100 cps or the IRM channels are on range 3 or higher.
(c) This function is automatically bypassed when the associated IRM channels are on range 8 or higher.
(d) This function is automatically bypassed when the IRM channels are on range 3 or higher.
l ie) This function is automatically bypassed when the IRM channels are on range 1.
i LIMERICK - UNIT 1 3/4 3-59 Amendment No. 4
I TABLE 3.3.6-2 r-
- i' 9
CONTROL R00 BLOCK INSTRUMENTATION SETPOINTS
~n 7
TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE E
1.
R00 BLOCK IGNITOR l
0 a.
Upscale 1.
flow biased
<?0.66 W + 40%, with a
< 0.66 W + 43%, with a
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maximum of, iiiaximum of,
- 11. high flow clamped
< 106%
< 109%
b.
Inoperative R.A.
H.A.
c.
Downscale
> 5% of RATED THERMAL POWER
> 3% of RATED THERMAL POWER 1
2.
APRM a.
Flow Blased Neutron Flux - Upscale
< 0.66 W + 42%*
b.
Inoperative R.A.
-< 0.66 W + 45%*
N.A.
c.
Downscale
> 4% of RATED THERMAL POWER
> 3% of RATED THERMAL POWER 1
d.
Neutron Flux - Upscale, Startup 312%ofRATEDTHERMALPOWER
{14%ofRATEDTHERMALPOWER 3.
SOURCE RANGE MONITORS a.
Detector not full in N.A.
N.A.
b.
Upscale
< 1 x 105 cps
< 1.6 x los ep, i
c.
Inoperative R.A.
R.A.
d.
Downscale
> 3 cps **
> 1.8 cps **
4.
INTE W OIATE RANGE SONITORS a.
Detector not full in N.A.
M.A.
b.
Upscale
< 108/125 divisions of
< 110/125 divisions of Tull scale Tull scale-l c.
Inoperative N.A.
M.A.
d.
Downscale
> 5/125 divisions of full
> 3/125 divisions of full icale icale E
@ k" 5.
SCRAN OISCHARGE VOL N to a.
Water Level-High
$ 257' 5 9/16" elevation ***
$ 257' 7 9/16" elevation w P.
a.
Float Switch 2
8 (AD
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3/4.9 REFUELING OPERATIONS BASES 3/4.9.1 REACTOR MODE SWITCH Locking the OPERABLE reactor mode switch in the Shutdown or Refuel position, as specified, ensures that the restrictions on control rod withdrawal and refueling
. platform movement during the refueling operations are properly activated.
These conditions reinforce the refueling procedures and reduce the probability of inadvertent criticality, damage to reactor internals or fuel assemblies, and exposure of personnel to excessive radioactivity.
3/4.9.2 INSTRUMENTATION The OPERABILITY of at least two source range monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.
The minimum count rate is not required when sixteen er fewer fuel assemblies are in the core.
During a typical core reloading, two, three or four irradiated fuel assemblies will be loaded adjacent to each SRM to produce greater j
than the minimum required count rate.
Loading sequences are selected to provide for a continuous multiplying medium to be established between the required oper-able SRMs and the location of the core alteration.
This enhances the ability l
of the SRMs to respond to the loading of each fuel assembly. During a core un-loading, the last fuel to be removed is that fuel adjacent to the S Ms.
I 3/4.9.3 CONTROL R00 POSITION The requirement that all control rods be inserted during other CORE ALTERATIONS ensures that fuel will not be loaded into a cell without a control rod.
3/4.9.4 DECAY TIME The minimum requirement for reactor subcriticality prior to fuel movement ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products. This decay time is consistent with the assump-tions used in the accident analyses.
3/4.9.5 COMUNICATI0NS The requirement for communications capability ensures that refueling station personnel can be promptly infomed of significant changes in the facility status or core reactivity condition during movement of fuel within the reactor pressure vessel.
LIMERICK - UNIT 1 B 3/4 9-1 Amendment No. 4
d REFUELING OPERATIONS BASES 3/4.9.6 REFUELING PLATFORM The OPERABILITY requirements ensure that (1) the refueling platform will be used for handling control rods and fuel assemblies within the reactor pressure vessel, (2) each hoist has sufficient load capacity for handling fuel assemblies and control rods, (3) the core internals and pressure vessel are protected from cxcessive lifting force in the event they are inadvertently engaged during lifting operations, and (4) inadvertent criticality will not occur due to fuel being loaded into a unrodded cell.
1 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE P00L The restriction on movement of loads in excess of the nominal weight of a fuel assembly and associated lifting device over other fuel assemblies in the storage pool ensures that in the event this load is dropped 1) the activity release will be limited to that contained in a single fuel assembly, and 2) any pessible distortion of fuel in the storage racks will not result in.a critical erray.
This assumption is consistent with the activity release assumed in the safety analyse.s.
3/4.9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL and WATER LEVEL -SPENT FUEL STORAGE POOL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. This minimum water depth is censistent with the assumptions of the accident analysis.
3/4.9.10 CONTROL'RDD REMOVAL These specifications ensure that maintenance or repair of control rods or i
control rod drives will be performed under conditions that limit the probability of inadvertent criticality.
The requirements for simultaneous removal of more 1
than one control rod are more stringent since the SHUTDOWN MARGIN specification provides for the core to remain subcritical with only one control rod fully withdrawn.
3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal loop be OPERABLE or that an alternate method capable of decay heat removal be demonstrated and that an alternate method of coolant mixing be in operation ensures that 1) suf-
-ficient cooling capacity is available to remove decay heat and maingpin the water in the reactor pressure vessel below 140*F as required during REFUELING, cnd 2) sufficient coolant circulation would be available through the reactor care.to assure accurate temperature indication and to distribute and prevent stratification of the poison in the event it becomes necessary to actuate the standby liquid control system.
i The requirement to have two shutdown cooling mode loops OPERABLE when there is less than 22 feet of water above the reactor vessel flange ensures that a single failure of the operating loop will not result in a complete loss of resid-ual heat removal capability. With the reactor vessel head removed and 22 feet of water above the reactor vessel flange, a large heat sink is available for core cooling.
Thus, in the event a failure of the operating RHR loop, adequate time is p ovided to initiate alternate methods capable of decay heat removal or ernergency procedures to cool the core.
LIMERICK - UNIT 1 B 3/4 9-2 1
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4 REFUELING OPERATIONS 3/4.9.2 INSTRUMENTATION LIMITING CONOITION FOR OPERATION 3.9.2 At least two source range monitor (SRM) channels
- shall be OPERABLE and inserted to the normal operating level with:
Continuous visual indication in the control room, a.
b.
At least one with audible alarm in the control room, One'of the required SRM detectors located in the quadrant where CORE c.
ALTERATIONS are being perfomed and the other required SRM detector located in an adjacent quadrant, and d.
Unless adequate shutdown margin has been demonstrated, the shorting links shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn.**
APPLICABILITY: OPERATIONAL CONDITION 5.
ACTION:
With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS and insert all insertable control rods.
SURVEILLANCE REQUIREMENTS i
4.9.2 Each of the above required SRM channels shall be demonstrated OPERABLE by:
~
a.
At least once per 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />s:
1.
Performance of a CHANNEL CHECK, 2.
Verifying the detectors are inserted to the normal operating level, and 3.
During CORE ALTERATIONS, verifying that the detector of an OPERABLE SRM channel is located in the core quadtant where CORE ALTERATIONS are being performed and another is located in an adjacent quadrant.
i
- These channels are not required when sixteen or fewer fuel assemblies, ad-l jacent to the SRMs, are in the core.
The use of special movable detectors during CORE ALTERATIONS in place of the normal SRM nuclear detectors is per-missible as long as these special detectors are connected to the normal SRM circuits.
)
- Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.
l LIMERICK - UNIT 1 3/4 9-3 Amendment No. 4 D
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REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued) b.
Performance of a CHANNEL FUNCTIONAL TEST:
1.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the start of CORE ALTERATIONS, and 2.
At least once per 7 days.
Verifying that the channel count rate is at least 3.0 cps:*
c.
1.
Prior to control rod withdrawal, 2.
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 /> during CORE ALTERATIONS, and 3.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
d.
Verifying, within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 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 /> during, that the RPS circuitry " shorting links" have been removed during:
1.
The time any control rod is withdrawn,** or i
2.
Shutdown margin demonstrations.
J
- May be reduced to 0.7 cps provided the signal-to-noise ratio is > 2.
channels are not required when sixteen or fewer fuel assemblies,, adjacent to These the SRMs, are in the core.
- Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.
LIMERICK - UNIT 1 3/4 g-4 Amendment No. 4
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