ML17156A243
| ML17156A243 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 05/03/1985 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17156A244 | List: |
| References | |
| NUDOCS 8505160353 | |
| Download: ML17156A243 (11) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 PENNSYLVANIA POWER
& LIGHT COMPANY ALLEGHENY ELECTRIC CO PERATIV IH DOCKET NO. 50-387 I III A A~TE<< <<ATION I IT I AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 43 License Ho.
NPF-14 1.
The Nuclear Regulatory Commission (the Commission or the NRC) having found that:
A.
The application for an amendment filed by the Pennsylvania Power
& Light Company, dated April 9, 1985 as supplemented on April 25, 1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's
.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 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.
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 the Facility Operating License No. HPF-14 is hereby amended to read as follows:
(2)
Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment Ho. 43, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.
PP&L shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
850516035~
05000387 PDR ADDCK PDRQ P
3.
This amendment was,effective April 30, 1985 FOR THE NUCLEAR REGULATORY COMMISSION
Attachment:
Changes to the Technical Specifications Date of Issuance:
~x 3.
~9B~
A. Schwencer, Chief Licensing Branch No.
2 Division of Licensing
ATTACHMENT TO LICENSE AMENDMENT NO.
F C L Y OPER TING LICENSE NO. NPF-14 0
K N
. 50-3 Replace the following pages of the Appendix "A" Technical Specifications with enclosed pages.
the revised pages are identified by Amendment nUmber and contain vertical lines indicating the area of change.
REMOVE 3/4 3-51 3/4 3-52 3/4 3-53 3/4 3-54 3/4 9-3 3/4 9-4 8 3/4 9-1 8 3/4 9-2 INSERT 3/4 3 3/4 3-,52 3/4 3-53 3/4 3-54 3/4 9-3 3/4 9-4 8 3/4 9-1 8 3/4 9-2
INSTRUMENTATION 3/4.3.6 CONTROL ROD 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.
APPLICABILITY: As shown in Table 3.3.6-1.
ACTION:
a.
b.
With a.control rod block instrumentation channel trip setpoint less cbhservative than the value shown in the Allowable Values column of
~" Table 3.3.6-2, declare the channel inoperable unti.l.the channel is t'estored to OPERABLE status-with its trip setpoint adjusted....
consistent with the Trip Setpoint value.
With the number of OPERABLE channels less than required by the Minimum OPERABtE Channels per Trip Function requirement, take the ACTION required by Table 3.3.6-1; SURVEILLANCE RE UIREMENTS 4.3.6.
Each of the 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.
1 SUSQUEHANNA - UNIT 1 3/4 3-51
TRIP FUNCTION 1.
ROD BLOCK MONITOR a.
Upscal e b.
Inoperative c.
Downscale TABLE 3. 3. 6-1 CONTROL ROD BLOCK INSTRUMENTATION MINIMUM APPLICABLE OPERABLE CHANNELS OPERATIONAL PER TRIP FUNCTION CONDITIONS 1 A'A 1*
~
ACTION 60 60 60 2.
APRM a.
Flow Biased Neutron Flux-Upscale b.
Inoperative c.
Downscale d.
Neutron Flux Upscale, Startup 3.
SOURCE RANGE MONITORS a.
Detector not full in
. (b) b.
Upscale~
~
c.
Inoperative d.
Downscale 4.
INTERMEDIATE RANGE MONITORS a.
Detector not full in b.
Upscale c.
Inoperati~~~
d.
Downscale 5.
Water Level-High 6.
REACTOR COOLANT SYSTEM RECIRCULATION FLOW a.
Upscale b.
Inoperative c.
Comparator 3
2 3
2 3
2 3
1 1,
1 2
2 2
2 2
2, 5
5 55.
5 5.
2 5A'*
61 61 61 61 61 61 61 61 61 61 61 61 61 61 61 61 62 62 62 62
TABLE 3.3.6-1 (Continued)
CONTROL ROD BLOCK INSTRUMENTATION ACTION ACTION 60 Declare the RBM inoperable and take the ACTION required by Specification
- 3. 1.4.3.
ACTION 61 With the number of OPERABLE Channels:
a ~
One less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel
.,to OPERABLE status within 7 days or place the inoperable channel in the tripped condition within the next hour.
ACTION 62 b.
Two 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 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
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 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
'a 0 b.
C.
d.
e.
NOTES With THERMAL POWER > 30K of RATED THERMAL POWER.
With more than one control rod withdrawn.
Not applicable to control rods removed per Specification
- 3. 9. 10. 1 or 3. 9. 10.2.
Not required when eight or fewer fuel assemblies (adjacent to the SRMs) are in the core.
The RBM shall be automatically bypassed when a peripheral control rod is selected or the reference APRM channel indicates less than 30K of RATED THERMAL POWER.
This function shall be automatically bypassed if detector count rate is
> 100 cps or the IRM channels are on range 3 or higher.
This function is automatically bypassed when the associated IRM channels are on range 8 or higher.
This function is automatically bypassed when the IRM channels are on range 3 or higher.
This function is automatically bypassed when the IRM channels are on range l.
SUSQUEHANNA - UNIT 1 3/4 3"53 Amendment No. 43
TABLE 3.3.6-2 CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS TRIP FUNCTION 1.
ROD BLOCK MONITOR a.
Upscale 1) 106X 2) 108XA'.
Inoperative c.
Downscale 2.
APRM a.
Flow Biased Neutron
-Flux - Upscale b.
Inoperative c.
Downscale d.
Neutron Flux - Upscale Startup 3.
SOURCE RANGE MONITORS TRIP SETPOINT
< 0.66 W + 40X
< 0.66 W + 42X NA
> 5/125 divisions of full scale
< 0.58 W + 50&
NA
> 5X of RATED THERMAL POWER
< 12X of RATED THERMAL POWER ALLOWABLE VALUE
< 0.66 W + 43X
< 0.66 W+ 45X I
NA
> 3/125 of divisions full scale
< 0.58 W + 53X" NA
> 3X of RATED THERMAL POWER
< 14X of RATED THERMAL POWER a@
b.
C.
d.
Detector not full in Upscale Inoperative Downscale NA
< 2x10 cps NA
> 0.7 cps*~
NA<4x10 cps HA
> 0.5 cps**
4.
INTERMEDIATE RANGE MONITORS a.
Detector not full in b.
Upscale c.
Inoperative d.
Downscale 5.
< 108/125 divisions of full scale NA
> 5/125 divisions of full scale NA
< 110/125 divisions of full scale NA
> 3/125 divisions of full scale
< 44 gallons a.
Water Level - High
< 44 gallons REACTOR COOLANT SYSTEM RECIRCULATION FLOW 6.
B (D
CL B
c+
a.
Upscal e
< 108/125 divisions of full scale b.
Inoperative NA c.
Comparator
< 10X flow deviation
< ill/125 divisions of full scale NA.
< llX flow deviation O
GJ The Average Power Range Monitor rod block function is. varied as a function of recirculation loop flow (W).
The trip setting of this function must be maintained in accordance with Specification 3.2.2.
""Provided signal-to-noise ratio is >2.
Otherwise, 3cps as trip setpoint and 2.8cps for allowable value..
N1ay be used. when the associated MCPR requirements in Specification 3.2.3 are satisfied.
REFUELING OPERATIONS 3/4'. 9. 2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 At least 2 source range monitor* (SRM) channels shall be OPERABLE and inserted to the normal operating level with:
a.
Continuous visual
- ndication in the control
- room, b.
C.
At least one with audible alarm in the control
- room, Ohe of the required SRM detectors located in the quadrant where CORE ALTERATIONS are being=-performed and the other required SRM detector lqcated in an adjacent
- quadrant, and d.
The "shorting links" removed from the RPS circuitry prior to and derring the time any control rod is withdrawn*" and shutdown margin demonstrations are in progress.
APPLICABILITY:
OPERATIONAL CONDITION 5.
ACTION:
Mith the requirements of the above specification not satisfied, immediately suspend all-operations involving CORE ALTERATIONS and insert all insertable control rods.
SURVEILLANCE RE UIREMENTS 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 quadrant where CORE ALTERATIONS are being performed and another is located in an adjacent quadrant.
- The use of special movable detectors during CORE ALTERATIONS in place of the nor'mal SRM nuclear detectors is permissible as 'long as these special detectors are connected to the normal SRM circuits.
These channels are not required when eight or fewer fuel assemblies (adjacent to the SRMs) are in the core.
""Not required for control rods removed per Specification 3.9. 10. 1 and 3.9. 10.2.
SUS(UEHANNA - UNIT 1 3/4 9"3 Amendment No. 43
REFUELING OPERATIONS SURVEILLANCE RE UIREMENTS Continued) b.
Performance of a CHANNEL FUNCTIONAL TEST:
1.
Mithin 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.
c.
Verifying that the channel count rate is at least 0.7 cps:~~~
l.
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 that the RPS circuitry "shorting links" have been removed 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:
1.
The time any control rod is withdrawn, or 2.
Shutdown margin demonstrations.
"""Provided the signal-to-noise ratio is > 2; otherwise, 3 cps.
%Not required for control rods removed per Specification 3.9.10.1 or
- 3. 9. 10. 2.
SUS(UEHANNA - UNIT 1 3/4 9"4 Amendment No. 36
3/4. 9 REFUELING OPERATIONS BASES 3/4.'9.
1 REACTOR MODE SMITCH 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.'he minimum count rate is not required when eight or fewer fuel assemblies are in the core because calculations for SSES have shown that eight fuel assem-blies at'aximum reactivity conditions and worst possible core geometry are subcritical.
During a typical core reloading two irradiated fuel assemblies will.be loaded around each SRM to produce greater than the minimum required count rate.
Loading schemes are selected to provide for a continuous multi-plying medium to be established between the required operable SRMs and the location of the core alteration.
This enhances the ability of the SRMs to respond to the loading of each fuel assembly.
During a core unloading the last fuel to be removed is that fuel adjacent to the SRMs.
3/4.9.3 CONTROL ROD POSITION
- The requirement that,all.control rods be inserted during other CORE Al TERATIONS 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 subcritica]ity 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 assumptions used in the accident analyses.
3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity condition during movement of fuel within the 'reactor pressure vesse).
SUSQUEHANNA " UNIT 1 8 3/4 9-1 Amendment No.
43
REFUELING OPERATIONS BASES 3/4.9.6 REFUELING PLATFORM The OPERABILITY requirements ensure that*(l) 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 f'r handling fuel assemblies and control rods, and (3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.
3/4.9.7 CRANE TRAVEL " SPENT FUEL STORAGE POOL The restriction on movement of loads in excess of the nominal weight of a fuel assembly 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 singl'e fuel assembly, and 2) any possible distortion of fuel in the storage racks will not result in a critical array.
This assumption is consistent with the activity release assumed in the safety analyses.
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 99K of the assumed lOX iodine gap activity released from the rupture of an irradiated fuel assembly.
This minimum water depth is consistent with the assumptions of the safety analysis.
3/4.9. 10 CONTROL ROD REMOVAL These specifications ensure that maintenance or repair of control rods or control rod drives will be performed under conditions that limit the probability of inadvertent criticality.
The requirements for simultaneous removal of more 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) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140 F as required during REFUELING, and 2) suffi.-
cient coolant circulation would be available through the reactor core 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.
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 provided to initiate alternate methods capable of decay heat removal or emergency procedures to cool the core.
SUS(UEHANNA " UNIT 1 B 3/4 9"2 Amendment No. Z9