ML17158A143
| ML17158A143 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 02/09/1994 |
| From: | Chris Miller Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17158A144 | List: |
| References | |
| NUDOCS 9403020052 | |
| Download: ML17158A143 (12) | |
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~O 0~*~4 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 20555-0001 PENNSYLVANIA POWER
& LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE INC.
DOCKET NO. 50-388 SUS UEHANNA STEAM ELECTRIC STATION UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
License No. NPF-22 1.
The Nuclear Regulatory Commission (the Commission or the NRC) having found that:
A. The application for the amendment filed by the Pennsylvania Power 8I Light Company, dated August 19,
- 1992, as supplemented by letters dated May 18, and October 7,
- 1993, 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 set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
940iO2OO5Z 94OiO9 PDR ADOCK 05000388, P,
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2.
Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-14 is hereby amended to read as follows:
(2) Technical S ecifications and Environmental Protection Plan The Technical Specific~/,jons contained in Appendix A, as revised through Amendment No. "'nd the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.
PPRL shall operate the facility in accordance with the Technical Specifica-tions and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance and is to be implemented prior to startup in Cycle 7, currently scheduled for Hay 20, 1994.
FOR THE NUCLEAR REGULATORY COMMISSION A< Pw~
Charles L. Hiller, Director Project Directorate I-2 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
February 9,
1994
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ATTACHMENT TO LICENSE AMENDMENT NO.
101 FACILITY OPERATING LICENSE NO. NPF-22 DOCKE NO 50-388 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.
The overleaf pages are provided to maintain document completeness.*
REMOVE 3/4 3-27 3/4 3-28 3/4 3-29 3/4 3-29a 3/4 5-5 3/4 5-6 INSERT 3/4 3-27*
3/4 3-28 3/4 3-29*
3/4 3-29a 3/4 5-5 3/4 5-6*
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INSTRUMENTATION 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3 The emergency core cooling system (ECCS) actuation instrumentation channels shown in Table 3.3.3-1 shall be OPERABLE with'heir trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.3-2 and with EMERGENCY CORE COOLING SYSTEM RESPONSE TIME as shown in Table 3. 3. 3-3.
APPLICABILITY:
As shown in Table 3.3.3-1.
ACTION:
a ~
b.
With an ECCS actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.3-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
With one or more ECCS actuation instrumentation channels inoperable, take the ACTION required by Table 3.3.3-1.
SURVEILLANCE RE UIREMENTS 4.3.3.1 Each ECCS actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK,,CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION opera+ions for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.3.1-1.
4.3.3.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automa+ic operation of all'hannels shall be performed at least once per 18 months.
4.3.3.3 The ECCS
RESPONSE
TIME of each ECCS trip function shown in Table 3.3.3-3 shall be demonstrated to be within the limit at least once per 18 months.
Each test shall include at least one channel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundan+
channe',s in a specific ECCS trip system.
SUSQUEHANNA - UNIT 2 3/4 3-27
rj)
C r/)0C x
zz I
z TRIP. FUNCTION "',
,'INIMUM"OPERAQLE C)IANNELS.'
'ER TRIP SYSTEM '- ':
tC APPUCABLE OPERATIONAL
'ONDmONS:-
CORE SPRAY SYSTEM a.
Reactor Vessel Water Level - Low Low Low, Level 1
b.
Drywell Pressure - High c.
Reactor Vessel Steam Dome Pressure
- Low (Permissive) 2(a) 2(a) 2(a) 1,2,3,4,5 1,2,3 1, 2, 3, 4,5 TABLE3.3.3-1 EMERGENCY CORE COOLING SYSTEM ACTUATIONINSTRUMENTATION ACTION 30 30 31 32 d.
Manual Initiation 1/subsystem 1, 2, 3, 4
, 5 33 2.
LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM
. a.
Reactor Vessel Water Level - Low Low Low, Level 1 b.
Drywell Pressure
- High c.
Reactor Vessel Steam Dome Pressure
- Low (Permissive)
- 1) System Initiation
- 2) Recirculation Discharge Valve Closure d.
Manual Initiation 2(a) 2(a) 2(a) 2(a) 1/subsystem 1,2,3,4,5 1, 2, 3 1~2, 3 4,5 1,2,3 4,5 1,2,3,4,5 30 30 31
- 32 31 32 33 O.
HIGH PRESSURE COOLANT INJECTION SYSTEM a.
Reactor Vessel Water Level - Low Low, Level 2 b.
Drywell Pressure - High c.
Condensate Storage Tank Level - Low
~ t ~
d.
Suppression Pool Water Level - High e.
Reactor Vessel Water Level - High, Level 8 I
IvldllcldlIllltlduon 2(a) 2(a) 2(a)(b) 2(a) 2(c) 1/system 1, 2, 3 1, 2, 3 1, 2, 3 1, 2, 3
- 1. 2. 3 1,2,3 30 30 34 34 31 33
CA DC X
zz z
MNNlUMOPERASLE CHANNELS PER TRP SYSTEM APPUCASLE OPERATIONAL CONDITIONS 4.
AUTOMATICOEPRESSURIZATION SYSTEM a.
Reactor Vessel Water Level - Low Low Low, Level 1
2lll
- l. 2, 3 TABLE3.3.3-1 tContinued)
EMERGENCY CORE COOLING SYSTEM ACTUATIONINSTRUMENTATION ACTION 30 CD CD hl CD b.
Orywell Pressure
- High c.
ADS Timer d.
Core Spray Pump Discharge Pressure
- High (Permissive) e.
RHR LPCI Mode Pump Discharge Pressure
- High IPer missive) f.
Reactor Vessel Water Level - Low, Level 3 IPermissive) g.
AOS Oryweg Presswe Bypass Timer h.
Manual Inhibit i.
Manual Initiation 2III 2ldllll 2ld)tei(II 2(Ii 1/valve 1, 2, 3 I, 2, 3 1,2,3 1.2,3 1,2,3 I, 2, 3 1,2,3 1,2,3 30 3I 31 31 31 31 33 33 TOTAL NO.
Of CHANNELS CHANNELS TO TISP aNNNsUM CHANNELS OPEltASLE APPUC/NLE OPERATIONAL CONDITIONS ACTION S.
LOSS OF POWER 3
b Ea o
I c-a.
4.16 kv ESS Bus Under-voltage (Loss of Voltage, < 20%)
b.
4.16 kv ESS Bus Under-voltage
{Degraded Voltage, < 6S%)
c.
4.16 kv ESS Bus Under-voltage IDegraded Vohage, < 93%)
d.
460V ESS Bus OBS65 Under. voltage IDegreded Vottege. < 65%)
e 44OV fbb eve Oeb4b Undec voltage IoetP eded Votreee c 92%I See footnotes on r>>xt page.
1/bus 2/bus 2/bus 2lbus 1/bus 2/bus 2/bus 1/bus 2/bus 2/bus 2/bus 2/bus 2/bus I, 2, 3, 4
, 5 1,2,3,4,5
- 1. 2. 3, 4, 5 1.2.3.4
.5
- 1. 2, 3, 4, 5 35 36 36 36 36
U)C lO0Cm zz Cz TABLE 3.3.3-1 (Continued)
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION (a)
A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.
k (b)
One trip system.
Provides signal to HPCI pump suction valves only.
(c)
Two out of two logic.
(d)
Either 4d or 4e must be satisfied.
The ACTION is required to be taken only if neither is satisfied.
A channel is not OPERABLE unless its associated pump is OPERABLE per Specification 3.5.1.
4>
4)
CO (e)
Within an ADS Trip System there are two logic subsystems, each of which contains an overall pump permissive.
At least one channel associated with each of these overall pump permissives shall be OPERABLE.
(f)
A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillance testing provided that all channels in the other trip system are OPERABLE.
When the system is required to be OPERABLE per Specification 3.5.2 Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 150 psig.
Required when ESF equipment is required to be OPERABLE.
¹¹ Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 100 psig.
Required to be OPERABLE only when Diesel Generator E is either aligned to the Class 1E system or not aligned to the Class 1E system but operating on the Test Facility.
O.
D z0 The automatic transfer of HPCI pump suction from the condensate storage tank to suppression pool on high suppression pool water level occurs only when HPCI injection valve is open.
EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS Continued 2.¹ For the HPCI system, verifying that the system develops a flow of at least 5000 gpm against a test line pressure of greater than or equal to 245 psig when steam is being supplied to the turbine at 150 a 15 psig 3.
Performing a CHANNELCALIBRATIONofthe CSS header 4P instrumentation and verifying the setpoint to be 6 1 psid.
4.
Verifying that the suction for the HPCI system is automatically transferred from the condensate storage tank to the suppression chamber either on a suppression chamber water level-high signal when HPCI injection valve is open, or on a condensate storage tank water level - low signal.
5.
Performing a CHANNEL CALIBRATION of the condensate transfer pump discharge low pressure alarm instrumentation and verifying the low pressure alarm setpoint to be > 113 psig.
d, For the ADS:
1
~
At least once per 31 days, performing a CHANNEL FUNCTIONALTEST of the accumulator backup compressed gas system low pressure alarm system.
2.
At least once per 18.months:
a) Performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence, but excluding actual valve actuation.
b) Manually opening each ADS valve'when the reactor steam dome pressure is greater than or equal to 100 psig and observing that either:
1)
The control valve or bypass valve position responds accordingly, or 2)
There is a corresponding change in the measured steam flow.
'he provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.
ADS solenoid energization shall be used alternating between ADS Division 1 and ADS Division 2.
For the startup following the Third Refueling and Inspection Outage, this surveillance shall read as follows:
For the HPCI System, verifying that the system develops a flow of at least 4850 gpm against a test line pressure of 600 psig when steam is being supplied to the turbine at 150 a psig.
SUSQUEHANNA - UNIT 2
'/4 5-5 Amendment No.
EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE IRENENTS
- Continued c)
Performing a CHANNEL CALISRATION of the accumulator backup ccepressed gas system low pressure alars systems and veri-fyfng air alarw setpoint of 2070 + 35 psig on decreasing pressure eo At least every 18 aanths the following shall be accomplished by any series of sequential, overlapping or total channel steps such that the entire channel is tested:
1.
A functional test of the interlocks associated with LPCI and CS puap starts in response to an autoaatic initiation signal in Unit 1 followed by a "False" autoaatic initiation signal in Unit 2.
2.
A functional test of the interlocks associated with LPCI and CS pump starts in response to an autoeatic initiation signal in Unit 2 followed by a "False" autoaatic initiation signal in Unit 1.
3.
A functional test of the interlocks associated with LPCI and CS puap starts in response to simultaneous occurrence of an autoaatic initiation signal in both Unit 1 and Unit 2 and a Loss-of-Offsfte-Power condition affecting both Unit 1 and Unit 2.
SUQQKHANA - NlT 2 3'-6 Aea~nt No.81 JUNj P 1991