ML17164A611
| ML17164A611 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 02/22/1995 |
| From: | Russell W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17164A612 | List: |
| References | |
| NUDOCS 9503070347 | |
| Download: ML17164A611 (47) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 205550001 P NNSYLVANIA POWER 8L LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE INC.
DOCKET NO. 50-387 SUS UEHANNA STEAM ELECTRIC STATION UNIT 1
MENDMENT TO FACILITY OPERATING LICENSE Amendment No. 143 License No. NPF-14 1.
The Nuclear Regulatory Commission (the Commission or the NRC) having found that:
A.
The application for the amendment filed by the Pennsylvania Power Light Company, dated July 27,
- 1994, as supplemented October 27,
- 1994, and February 3,
- 1995, 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'pplicable requirements have been satlsfied.
9503070347 950222 PDR ADOCK 05000387 P
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2.
Accordingly, Facility Operating License No.
NPF-14 paragraph 2.C.(1) is hereby amended to read as follows:
(I) Maximum Power Level Pennsylvania Power and Light Company (PPKL) is authorized to operate the facility at reactor core power levels not in excess of 3441 mega-watts thermal in accordance with the conditions specified herein and in Attachment 1 to this license.
The preoperational
- tests, startup tests and other items identified in Attachment 1 to this license shall be completed as specified.
Attachment 1 is hereby incorporated into this license.
3.
Further, 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 Fdcility Operating License No.
NPF-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 No.
143 and the Environmental Protection Plan contained in Appendix 8, are hereby incorporated in the license.
PPIIL shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
4.
This license amendment is effective as of its date of issuance and is to be implemented prior to startup in Cycle 9, currently scheduled to occur in May 1995.
FOR THE'NUCLEAR REGULATORY COMMISSION Attachments:-.
1.
Page 3 of License 2.
Changes to the Technical Specifications Date of Issuance:
February 22, 1995 William T. Russell, Director Office of Nuclear Reactor Regulation
- Page 3 is attached, for convenience, for the composite license to reflect this change.
(3)
PP&L, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to
- receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor
- startup, sealed neutron sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)
PP&L, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to
- receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)
PP&L, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to
- possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This license shall be deemed to contain and is subject to the conditions-specified in the Commission's regulations set forth in 10 CFR Chapter I
and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and's subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level Pennsylvania Power
& Light Company (PP&L) is authorized to operate the facility at reactor core power levels not in excess of 3441 megawatts thermal in accordance with the conditions specified herein and in Attachment 1 to this license.
The preoperational
- tests, startup tests and other items identified in Attachment 1 to this license shall be completed as specified.
Attachment 1 is hereby incorporated into this license.
(2)
Technical S ecifications and Environmental Protection Plan The'echnical Specifications contained in'ppendix A, as revised through Amendment No.
and the Environmental Protection Plan coiltained in Appendix B, are hereby incorporated in the license.
PP4t. shall oper ate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
Amendment No. 5, 143
TTACHMENT TO LICENSE AMENDMENT NO. 143 FACILITY OPERATING LICENSE NO. NPF-14 DOCKET NO. 50-387 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
~REHOV 1-6 2-1 2-4 8 2-la 8 2-7 3/4 1-20 3/4 2-2 3/4 3-4 3/4 3-5 3/4 3-17 3/4 3-18 3/4 3-19 3/4 3-20 3/4 3-24 3/4 3-26 ~
3/4 3-48~
3/4 3-54 3/4 4-la 3/4 4-lb 3/4 4-lc INSERT 1-6 2-1 2-4 8 2-la 8 2-7 3/4 1-20 3/4 2-2 3/4 3-4 3/4 3-5 3/4 3-17 3/4 3-18
. 3/4 3-19 3/4 3-20 3/4 3-24, 3/4 3-26 3/4 3-42 3/4 3-54 3/4 4-la 3/4 4-lb 3/4 4-lc 4'
TTACHM NT TO ICENSE AMENDMENT NO.
ACI ITY OPERATING LICENSE NO. NPF-14 DOCKE NO. 50-387 Replace the following pages of the Appendix A Technical Specifications with
'nclosed pages.
The revised pages are identified by Amendment number and contain vertical lines indicating'the area of change.
fgMO~V INSERT 3/4 4-1f 3/4 4-3 3/4 4-5 3/4 4-7 3/4 4-21 3/4 5-4 B 3/4 3-3 B 3/4 4-8 B 3/4 4-9 B 3/4 5-1 B 3/4 6-3 5-6 5-8 6-20b 3/4 4-1f 3/4 4-3 3/4 4-5 3/4 4-7 3/4 4-21 3/4 5-4 B 3/4 3-3 B 3/4 4-8 B 3/4 4-9 B 3/4 5-1 8 3/4 6-3 5-,8 6-20b
DEFINmONS 1 33",
RATEDTHERMALPOWER shall be a total reactor core heat transfer rate to the reactor coolant'of'3441 MWT.
I 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time Interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energizatton of the scram pilot valve solenoids.
The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.
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, -I 1.35 1.36 A REPORTABLE EVENT shall be any o 10 CFR Part 50.
I ROD DENSITY shall be the number of c total number of control rod notches.
Al DENSITY.
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f those conditions specified in Section 50.73 to ontrol rod notches inserted as a fractlo'n of. the>-."-'
rods fully inserted is equivalent to 100% ROLE.--;
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1.37 SECONDARY CONTAINMENTINTEGRITYshall exist when:.
a.
All secondary containment penetrations required to be dosed during accident conditions are either:
1.
Capable of being closed by an OPERABLE secondary containment automatic isolation system, or r
2.
Closed by at least one m'anual valve, blind flange, or deactivated automatic damper secured in its closed position, except as provided in Table 3.6.5.2-1 of Specification 3.6.5.2.
'.', '.-',::~j~~~:Att:secondary) containment hatches and blowout panels are closed and sealed.
- "; +;-; '.,<'g~;~itindby gas treatment system is OPERABLE pursuant to Specification
'." At least one door in each access to the secondary containment is closed.
e.
The sealing mechanism associated with each secondary containment penetration, e.g.,
- seats, or O-rings, is OPERABLE.
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f.
The pressure within the secondary containment is less than or equal to the value required by Specification 4.6.5.1.a.
SUSQUEHANNA - UNIT 1 1-6 Amendment No. N~ 143
2.0 SAFETY UMITS AND LIMmNGSAFETY SYSTEII SETTINGS 2.1.1, THERMALPOWER shall not exceed 25% of RATED THERMALPOWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10 million Ibm/hr.
OPERATIONAL CONDITIONS 1 AND 2.
ElGIIQH'ith THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core fiow less than 10 million Ibm/hr., be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements ot Specification 6.7.1.
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2.1.2 The MINIMUMCRITICAL POWER RATIO (MCPR) shall not be less than 1.06 with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10 million Ibm/hr.
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With MCPR less than 1.06'nd the reactor vessel steam dome press'ure greater than 785 psig and core flow greater than 10 million Ibm/hr., be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
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and comply with the requirements of Specification 6.7.1.
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I V I'.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.
- OPERATIONALCONDITIQNS 1, 2, 3 AND4.
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With th'e";reactor'~eo'oIant'system
- pressure, as measured in the reactor vessel steam dome, above 1326'.,psig~M,'lij,ot-'.least HOT SHUTDOWN with reactor coolant system pressure less than or equal to"..)3&psig wlthtn,"2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.
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Sae Specification 3.4.1.1.2.a for single loop operation requirement.
SUSQUEHANNA - UNIT 1 2-1 Amendment No.
$ti, 143
' RFACTOR PROTECAON SYSTEM::lNSTRUINENTATION;SETPOIMTS~$"'.;."':'~::;-:,"~",~-."":.'.;.-","-:;.,::.-.',,;-:
Ch roDCmX R
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Intermediate Range Monitor, Neutron Rux-High 2.
Average Power Range 54onitor "
a.
Neutron 'wn b.
Flow Biased
'lormal Power-Upscale 1)
Row 8lased, 2)
High Flow CIamped 5 120/1 26 divisions of full scale S 122/1 26 divisions of full scale 5 16% of RATED THERMALPOWER 5 20% of RATED THERMALPOWER;~--".'~',";
1
~ 0.68 W+58%',
with a maximum of S 113.6% of RATED THERMALPOWER S 0.68 W+82%,
with a maximum of 5 116,6% of RATED THERMALPOWER
- TNP SHUNT'~"':-""'-'
". '""ALLONARSVANES ~
c.
Neutron Rux-Upscale d.
Inoperative 5 118% of RATED THERMALPOWER 5 120% of RATED THERMALPOWER NA 3.
Reactor Vessel Steam Dome Pressure - High 4.
Reactor Vessel Water Level - Low, Level 3 5 1087 psig 2 13.0 inches above instrument zero S 1093 psig 2 11,6 inches above instrument zero 3
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5.
Main Stean Line Isolation Valve - Closure 6.
Main Steam Line Radiation - High 7.
Drywall Pressure - High 8.
Scram Discharge Volume Water Level - High a.
Level Transmitter b.
Boat Switch S.
Turbine Stop Valve - Closure
- 10. Turbine Control Valve Fast Closure, Trip Oil Preset' Low 11.
Reactor Mode Switch Shutdown Position
- 12. JUlanual Scram 5 10% closed 5 7.0 x fullpower background 5,1.72 psig 5 88 gallons 5 88 gallons S 6.6% closed h 500 pslg NA C 11% dosed 5 8.4 x fullpower background 5 1.88 pslg 5 88 gallons 5 88 gallons 5 7% closed I480 psig NA 0
Sea Speal~soon 3 4 I I 2.a lor single loop operation requirement.
BASES,'~
(Continued) 25%
peaki peaki of 3.35 Mwt corresponds to a bundle which is significantly higher than the exp thermal power a bundle power ng factor of approximately 3.0 ng factor.
radial ected
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- Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressures below 785 psig is conservative.
SUSQUEHANNA - UNIT 1 B 2-1a Amendment No. 72> 143
BASES (Continued) 9.
The turbine stop valve closure trip anticipates the pressure, neutron fiux, and heat flux increases that would result from closure of the stop valves.
With a trip setting of 5.5% of valve closure from full open, the resultant increase in heat flux is such that adequate thermal margins are maintained during the worst case transient assuming the turbine bypass valves operate.
10.
This function is not required when THERMAL POWER is below 30% of RATED THERMAL POWER.
The Turbine Bypass System is sufficient at this low power to accommodate a
turbine stop valve closure without the necessity of shutting down the reactor.
This function is automatically bypassed at turbine first stage pressures less than the analytical limit of 147.7 paid, equivalent to THERMAL POWER of about 30% RATED THERMAL POWER.
Turbine first stage pressure of 147.7 psig is equivalent to 22% of rated'turbine load.
~8 The turbine control valve fast closure trip anticipates the pressure, neutron flux, and heat flux increase that could result from fast closure of the turbine control valves due to load rejection coincident with failure of the turbine bypass valves.
The Reactor Protection System initiates a trip when fast closure of the control valves is initiated by the fast acting solenoid valves and in less than 30 milliseconds after the start of control valve fast closure.
This is achieved by the action of the fast acting:solenoid valves in rapidly reducing hydraulic trip oil pressure at the main turbine control valve actuator disc dump valves.
This loss of pressure is sensed by pressure switches whose contacts form the on~ut-of-two-twice logic input to the Reactor Protection System.
This trip setting, a faster closure time, and a different valve characteristic from that of the turbine stop valve, combine to produce transients which are very similar to that for the stop valve.
Relevant transient analyses are discussed in Section 15.2 of the Pinal Safety Analysis Report.
This function is not required when THERMAL POWER is below 30% of RATED THERMAL POWER;:,~<The,;Turbine Bypass System is sufficient at this low power to accommodate a
,;,',,'turbine",control'valve closure without the necessity of shutting down the reactor.
This
'functl+h'.automatically bypassed at turbine first stage pressures less than the analytical
'Iim>~447,.'7.'pslg',
equivalent to THERMAL POWER of about 30% RATED THERMAL
~ POWER";=,'Turbine.first stage pressure of 147.7 psig is equivalent to 22% of rated turbine
" load'.;-"~:" " '"--
12.
The reactor mode switch Shutdown position is a redundant channel to the automatic protective instrumentation channels and provides additional manual reactor trip capability.
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I The Manual Scram is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.
SUSQUEHANNA - UNIT 1 8 2-7 Amendment No.4$, 143
I'I I I'URVEILLANCEREQUIREMENTS (Continued)
- b. At least once 'per 31 days by;
- 1. Verifying the continuity of the explosive charge.
2.
Determining that the available weight of sodium pentaborate is greater than or equal to 5500 Ibs and the concentration of boron in solution is within the, limits of Rgure 3.1.5-2 by chemical analysis.
- 3. Verifying that each valve, manual, power operated or automatic, in the fiow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
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- c. Demonstrating that, when tested pursuant to Specification 4.0.5, the minimum flow requirement of 41.2 gpm at a pressure of greater than or equal to 1224 psig is met.
- d. At least once per 18 months during shutdown by; F
1.
Initiating one of the standby liquid control system loops, Including an explosive'alve, and verifying that a flow'ath from the pumps to the reactor pressure vesse~F..
is available by pumping demineralized.
water into the reactor vessel; The::-.
replacement charge for the explosive valve shall be from the same mariufactured';
batch as the one fired or from another batch which has been certified by having one of that batch successfully'fired.
Both Injection loops shall be tested ln 36 months.
~ 0 2.
Demonstrating that all heat traced piping is unblocked by pumping from the storage tank to the test tank and then draining:and flushing the discharge piping and test tank with demineralized water.
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~ 0 This test shall also be performed anytime water or boron is added to the solution or when the solution temperature drops below the limitof Figure 3.1.6-1.
This test shall also be performed whenever both heat tracing circuits have bean found tO be inoperable and may be performed by. any series of sequential, overlapping or total flow pijth steps such that the entire flow path is included.
SUSQUEHANNA - UNIT 1 3/4 1-20 Amendment No. f35, 143
y UNITINGCONDmON FOR OPERATION 3.2.2 The APRM flow biased simulated thermal power-upscale scram trip setpoint (S) and flow biased neutron flux-upscale control rod block trip setpoint (Sss) shall be established according to the following relationships:
S5 (0.68W + 59%) T Sea 5 (0.68W + 50%) T SS (0.68W + 62%) T Saeva(0.58W + 63%) T Loop, recirculation flow as a percentage of the loop rec produces a core flow of 100 million Ibs/hr, Lowest value of the ratio of FRACTION OF RATED THERM the MAXIMUMFRACTION OF UMITINGPOWER DENSITY.
is the actual LHGR divided by the UNEAR HEAT GENERAT Setpoints limitspecified in the CORE OPERATING UMITS RE W
irculation flow, which AL POWER divided by ~ v,'~~ '~,
ION RATE for,APRIL/I'-
J' "i>'"i T is always less than or equal to 1.0.
25% of RATED THERMALPOWER.
where:
S and Sea are in percent of RATED THERMALPOWER, gLQIIQ5:With the APRM flow biased simulated thermal power-upscale scram trip setpoint and/or the fiow biased neutron flux-upscale control rod block trip setpoint less conservative than the value shown in the Allowable Value column for S or Sas, as determined
- above, initiate corrective action within 15 minutes and adjust S and/or Sea to be consistent with the Trip Setpoint value within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMALPOWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
'=".g~With'MFLPD;greater'han the FRTP during power ascension up to 90% of RATED THERMAL
"~i.IPO~iither;than. adjusting the APRM setpoints, the APRM gain may be adjusted such that
'" reeSng~do'es not'exceed 100% of RATED THERMAL POWER, the required gain adjustment increment doei not exceed 10% of RATED THERMALPOWER, and a notice of the adjustment is posted on the reactor control panel.
See Specification 3.4.1.1.2.a for single loop operation requirements.
SUSQUEHANNA - UNIT 1 3/4 2-2 Amendment No.Egg, ging, 143
TABLE3.3.1-1 (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION STATEMENTS ACTION 1 ACTION 2 ACTION 3 ACTION4 ACTION 5 Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Verify all insertabie control rods to be inserted in the core and lock the reactor mode switch in the Shutdown positiori within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Suspend all operations involving CORE ALTERATIONSand insert all insertable control rods within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Be in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
Be in STARTUP with the main steam line isolation valves closed within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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"Yt'Y" ACTION 6 ACTION 7 ACTION 8 ACTION 9 v i Initiate a reduction in THERMALPOWER within 15 minutes, and reduce THERMALPOWER to less than.3096 of. RATED THERMAL'OWERwithin 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Verify all insertabie control rods to be inserted within 1
- hour, Lock the reactor mode switch in the Shutdown position within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Suspend all operations involving CORE ALTERATIONS, and insert all insertable control rods and lock the reactor mode switch in the SHUTDOWN position within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
PY YF SUSQUEHANNA - UNIT 1 3/4 3R Amendment No. gg, 1.43
(a)
'b)
(c)
(d)
(e)
(g)
(h)
TABLE3.3.1-1 (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLENOTATIONS A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 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.
Upon determination that a trip setpoint cannot be restored to within its specified value during performance of the CHANNELCALIBRATION,the appropriate ACTION, 3.3.1a or 3.3.1b, shall be followed.
This function is automatically bypassed when the reactor mode switch is in the Run position.
The "shorting links" shall be removed from the RPS circuitry prior to and I
0 during the time any control rod is withdrawn and shutdown margin demonstrations performed per Specification 3.10.3.
The norH:oincident NMS reactor trip function logic is such that all channels go to both trip systems.
Therefore, when the "shorting links are removed, the Minimum OPERABLE Channels Per Trip System is 4 APRMS and 6 IRMS.
An APRM channel is inoperable ifthere are less than 2 LPRM inputs per level or less than 14 LPRM inputs to an APRM channel.
This function is not required to be OPERABLE when the reactor pressure vessel head is unbolted or removed per Specification 3.10.1.
This function is automatically bypassed when the reactor mode switch is not in the Run position.
This function is not required to be OPERABLE when PRIMARY CONTAINMENTINTEGRITYis not required.
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With any control rod withdrawn.
';Th(sr:function shall not be automatically bypassed when turbine first
- .sta>>ge pressure is greater than an allowable value of 136 psig.
',,Also actu'ates the EOC-RPT system.
".ThJi",function is required to be OPERABLE only prior to and during
'shutdown'margin demonstrations as performed per Specification 3.10.3.
Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.
SUSQUEHANNA - UNIT 1 3/4 3-S Amendment No.++~ ~~ ~ ~4-
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- 1. PRIMARYCONT
.NOLITION
- a. Reactor Vessel Water Level 1l Low, Level 3
- 2) Low Low, Level 2;". -;-
3l Low Low Low, Level 1 2 13.0 inches 2-38.0 inches" 2-129 inches 2 11.6 inches 2 <6.0 inches 2-138 inches
,-..-;g~~~~~g r.ISOLATIONACTUATIQMIQQTRUMENTA'nON;.SETPQlNTS.':,':;.:;~l~;:;.-.:.:::-';.,";:.":'"~",~ ~ '.- ',~;:-', =.,
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- b. Drywall Pressure - High
- c. Manual Initiation
- d. SGTS Exhaust Radiation - High e.
Main Steam line Radiation - High
- 2. SECONDARY CONTAINIIENTISOLATION
- a. Reactor Vessel Water Level-Low Low, Level 2
- b. Drywall Pressure - High
- c. Refuel Roor High Exhaust Duct Radiation - High
- d. Railroad Access Shaft Exhaust Duct Radiation-High
- e. Refuel Roor Wa0 Exhaust Duct,Radiation - High
- f. Manual Initiation
- 3. MAINSTEAM ONE ISOLATION
- a. Reactor Vessel Water Level - Low Low Low, Level 1
- b. Main Steam Une Radiation - High
- c. Main Steam Une Pressure - Low
- d. Main Steam Une Flow - High S 1.72 psig NA S 23.0 mR/hr ~
S 7.0 x fullpower background 2-38.0 inches S 1.72 psig S 2.6 mR/hr S 2,6 mR/hr S 2.6 mR/hr NA 2-129~>>
S 7.0 x fWIpower background 2 881 pslg S 113 paid>>>>
S 1.88 psig NA S 31.0 mR/hr S 8.4 x fWl power background 2 %6.0 inches S 1.88 psig S 4.0 mR/hr S 4.0 mR/hr S 4.0 mR/hr NA 2-138 Inches S 8.4 x fullpower background
>841 psig
<121 paid>>>>
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RCIC Equipment Room Tenperature - High e.
RCIC Equipment Room h Temperature - High f.
RCIC Pipe Routing Area Temperature - High g.
RCIC Pipe Routing Area h Temperature - High h.
RCIC Emergency Area Cooler Temperature - High i.
Manual Initiation j.
Drywall Pressure - High 6.
HIGH PRESSURE COOLANT NJECTION SySTEM ISOLATION a.
HPCI Steam Line Row - High b.
HPCI Steam Supply Pressure - Low c.
HPCI Turbine Exhaust Diaphragm Pressure - High d.
HPCI Equipment Room Temperature - High e.
HPCI Equipment Room h Temperature - High f.
HPCI Emergency Area Cooler Temperature - High g.
HPCI Pipe Routing Area Temperatwe - High 5,167'F 5 89'F S 1674FNf S 894FNt 5 1670F 5 1.72 psig 5 387 inches H>O I 104 pslg S 10 psig 5 167'F
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5 88'F 5 1670F 5 167'FfS
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5 174'F 5 98'F S 174'Ftf 5 984FSf 5 174'F NA 5 1,88 psig 5 400 inches HsO h 80 psig 5 20 pslg 5 1740F 5 88'F 5 ]74'F 5 174'FNf
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"..::; ',j>~>.'.=';",','..';', ".Q+'JOIE ACTUATlOlgINSTRUlNENTATlOQSETi?QllblTS':- '; ---::.
... TRIPSETPOINT.=='.:--:=:
'-..=-
-:-;-':.-'.:,,'-";-ALi.aWruaEViLUE'.:.:.;:;,':,
Ca)
D la)
O h.
HPCI Pipe Routing Area 4 Temperature - High I ~
Manual Initiation j.
Dryweli Pressure - High 7.
RHR SYSTEM SHUTDOWN COOUNQfHEADSPRAY IIODE ISOLATION a.
Reactor Vessel Water Level - Low, Level 3 b.
Reactor Vessel {RHR Cut-in Permissive) Pressure - High c.
RHR Row - High d.
Manual Initiation e.
Drywall Pressure - High 5 894F NA 5 1.72 psig 2 13.0 inches 5 98 psig S 26,000 gpm S 1.72 psig 5 984F NA S 1.88 psig 2 11.S inches 5 108 psig S 28,000 gpm 5 1.88 psig R,3 z0 I
lA
-\\
~dl~lk ~k
+ >>'N>>Wl
~ ~ ~
~
co co0 R
X
=I TRI'FUNCTION'.<-"."
CHANNEL CHECK CHANNEL FUNCTIONAL TEST OPERATIONALCONDITIONS FOR WHICH SURVEILLANCERE+IIIED~~:.q (Continued) d.
HPCI Equipment Room-Temperature,"-'igh e.
HPCI Equipment Room'4"Temperaturi'- High f.
HPCI Emergency Area'Cooler Temperature - Higlt NA" Q
Q 1,2,3 1,2,3 1,2,3 TABLE4.3.2.1-1 (Ccwttlnued)
- ISOLATIONACTUATIONINSTRUMENTATIONSURVEILLANCEREQUIREMENTS Ca)
O Col lb 01 g.
HPCI Pipe Routing Area Temperatwe - High h.
HPCI Pipe Routing Area 4 Temperatwe - High i.
Manual Initiation j.
Drywall Pressure - High NA" 7.
RHR SYSTEM SHUTDOWN COOUNQ/HEAD SPRAY MODE ISOLATION M
R Q
1.2.3 1,2,3 1,2,3 1.2,3 tttt a.
Reactor Vessel Water Level - Low, Level 3 b.
Reactor Vessel (RHR Cutm Permlss)ye)
Pressure - High c.
RHR Row - High d.
Manual Initiation e.
Drywall Presswe - High NA S
M R
R 1,2,3 1,2,3 1, 2, 3 1,2,3 1,2,3 4
CL3 z0 EC1 lv t ~
~ t ~
tttt When handling Irradiated fuel in the secondary containment and during CORE ALTERATIONSand operations with a potential for draining the reactor vessel When any twblne stop valve Is open.
When VENTING or PURGING the dryweg per Specification 3.11.2.8.
This trip function need not be OPERABLE from October 19, 1989 to January 19, 1890,.
v s-
~
p~k~M r TABLE3.3.4.2-1 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION TRIP FUNCTION MIIIIMUMOPERABLE CHANNELS PER TRIP SYSTEM 1.
Turbine Stop Valve - Closure (b) 2.
Turbine Control Valve - Fast Closure Ib)
(a)
(b)
A trip,system 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 provided that the other trip system is OPERABLE.
This function shall not be automatically bypassed when turbine first stage pressure is greater than an allowable value of 13B pslg.
SUSQUEHANNA - UNIT 1 3/4 3%2 Amendment No. 143
coC coDCmX z
R a.
Upscale f8 b.
Inoperative c.
Downscale n 'iks I"4 S0.83 W + 41%
NA R 6/126 divisions of full scale S 0.83 W + 43%
NA h 3/126 of divisions full scale sc
- e
. "'" QQQTR(}L'Rap QLQQK INS~ggQffg'egg Qgg?QPffs'~~-"".4'-~."os "~"~".'~%~'.:a>,
'::.4:"$~:"r,'P "n8 f
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4
- ".
- :.'f::::,;";>'LLONABgVflLLVR;.;.'>>'"',:<-:;<'";~,',
a.
Row Biased Neutron Rux Upac'aMP 1)
Row Biased 2)
High Row Camped b.
Inoperative C.
Downscale d.
Neutron Rux - Upscale Startup S0.68 W+ 60%
S 108% of RATED THERMALPOWER NAI 6% of RATED THERMALPOWER S 12% of RATED THERMALPOWER S0.68 W + 63%
S 111% of RATED THERMALPOWER NA 2 3% of RATED THERMALPOWER S 14% of RATED THERMALPOWER CD O
Cal crfD a.
Detector not full in b.
Upscale c.
Inoperative d.
Downscale NA S2 x 10'ps NA 20.7 cps NA S4 x 10'ps NA h 0.6 cps a.
Detector not full in b.
Upscale c.
Inoperative d.
Downscale NA S 108/126 divisions of fug soils NAI6/126 divisions of fug scale NA S 110/126 divisices of full scale NA 2 3/126 divisions of fullscale 6.
8.
a.
Water Level - High R
O
'6l a.
Upscale b.
Inoperative c~ Comparator S 114/126 divisions of fug scale
'NA s
S 10% flow deviation S 117/126 divisions of fullscale NA S 11% flowdeviation The Average Power Range Manitor rod btack furiadan is v'arled as'afuncllonof!
Ik'in<loop tloW fttg.:;:loha'trig 'aattrogOftffs'kwcdantrtusIba PrOVided Signal tO-nOiee ratiO iS h 2. OtherWiae, 3 Cpa aa trip Setpcint and 2.8,qiktcr'tICfWible'Vastu'ed@< Iv','i. ~v~Y".".,';::;.<i:"'@AT)%S,.'>','.-)k~>j'j<<'<g"p',';.
See SpeCI fICatIOn 34.1,1,2,e fOr SIngle IOOp'Operaticl) reftuiremente;:,,"':;:.g~o'~<A",pbbs~',~p~e~jjjrsj
<j'>.+t.""j;;.,,~:y4,,'::4<igs@g+$ j~",.<Og.'".j~@: ~>tf~~,.<"
b.
In OPERATIONAL CONDITION 2 with no reactor coolant sy'tem recirculation loops in operation, return at least one reactor coolant system recirculation loop to operation, or be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
4.4.1.1.1.2 Each pump MG set scoop tube electrical and mechanical stop shall be demonstrated OPERABLE with overspeed setpoints less than or equal to a core flow of 109.5 million Ibm/hr and 110.5 million Ibm/hr respectively, at least once per 18 months.
',~
SURVEILL'ANCE,REQUIREMENTS V'~
- >'.i ;~"."
hQIlg5: (Continued) )
- 3. With less than 50% of the required LPRM upscale alarms
- OPERABLE, follow ACTION a.1.d upon entry into Region II of Figure 3.4.1.1.1-1.
t 1
- c. With any pump discharge valve not OPERABLE remove the associated loop from operation, close the valve and comply with the requirements of Specification 3.4.1.1.2.
I
- d. With any pump discharge bypass valve not OPERABLE close the valve and verify closed at least once per 31 days.
4$,
- -.j",:-'.4.1.1.1.1 Each pump discharge valve and bypass valve shall be demonstratecP OPERABLE by cycling each valve through at least one complete cycle o5',.'.,"g,>,'~-*,.-.,
fulltravel during each startup prior to THERMALPOWER exceeding 25%
of RATED THERMALPOWER.
F 4.4.1.1.1.3 At least 50% of the required LPRM upscale alarms shall be determined OPERABLE by. performance of the following on each LPRM upscale alarm:
1)
CHANNELFUNCTIONALTEST at least once per 92 days, and
., 2)
CHANNELCALIBRATIONat least once.per 184 days.
~ ~
If not performed within the previous 31 days.
SUSQUEHANNA - UNIT 1 3/4 4-1a Amendment No. 9$, 143
Q o4
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~
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40 4$
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Core Flow (MiHionIbm/hr)
Bgure BA.1.1.1-1 THEIMALPOWER STABILHYRESIRICHONS
~
0g SUSQUEHANNA - UNIT 1 3/4 4-1b Amendment No. II8, 718; 14-
1 UMITINGCONDITION FOR OPERATION a
3I f'
<<>>.'<<<<>>(
'$ >>t4
~ <<>>or$
( <<. $.
t 3.4.1
~ 1.2 One reactor coolant recirculation loop shall be in operation with the pump speed 5 80% of the rated pump speed and the reactor at a THERMAI POWER/core flow condition outside of Regions I and ll of Figure 3.4.1.1. f;1,'and
- a. the following revised specification limits shall be followed:
~ >>
'I.
Specif(cation 2.1.2: the MCPR Safety,L(mit shall be increased to 1.07; 2.
Table 2,2.1-1 the APRM Flow-Biased'cram: Trip. Se'tpoints shall be as follows:
';,";~,. $~>':;,",;',.'Trfp Setpoht' "j
50.58W + 54%
.',;3.'5 0.68W + 67%.
~,,( <<.0 $>>.<<(.l 3.
Specification 3.2.2: the APRM Setpoints shall be as follows:
I
,(irk +>><<ai>>1>>a
( i
)>>
i
>>>>t(/jhow<<. (im~ >><<<<
t'r Ag: "x$) $$($'~'g+":+Nd~Q +8::::@4&wc '
>>(>> ~4>>>>'rd~<'k~~f~" +'~re ss(o.saw
+ 64%) T SS(0.68W + 67%) T SRg 5'(0.58W + 46%) T
(
~
>>/ i I "<'SRB'S (0,68W'+ 48%)'T>
4.
Specification 3.2.3: The MINIMUMCRITICAL POWER RAT(O (MCPR) shall be greater than or equal to the applicable Single Loop Operation MCPR limit as specified in the CORE OPERATING LlhlllTSREPORT; io<
5.
Specification 3.2.4:
The LINEAR HEAT GENERATION RATE (LHGR) shall be less than or equal to the applicable Single Loop Operation LHGR limit as specified in the CORE OPERATING LIMITSREPORT.
6.
Tablei3.3.6-2:
the RBM/APRQContro)..Rod,,B(ock, Setpoints,shall be as
,follows:.
r( -<<>>
,a.
RSM - Upscale yVr~Y'>>'j'tR'". '<<
b., APRM-Row Biased ip 50.83W + 36%
50.83W + 37%
, Trip,Setpo(rit,"",:.,',:", :-.'."; "., 'Alowibio'.Vetue'.
'o.saw+
46%,,
~o'.caw.,+ 48%
- -':<:->>:."'~>> THp Set paht"'""-<"'$"': k;,.':, Alowebfe.Vatue(
EulaalaX:
operation.4
<<(g y o<<p
<<<<<<<<t N
<< ~
SUSQUEHANNA - UNIT 1 3/4 4-1 c Amendment No. LN fW ~~~
f
'(ll e
2 r+)
I LIMmNGCONDITION FOR OPERATION 1
I J ll 3.4.1.3 Recirculation pump speed mismatch shall be maintained within:
a.
5% of each other with core flow greater than or equal to,75 million Ibm/hr.
b.
10% of each other with core flow less than 75 million Ibm/hr.
I
- OPERATIONAL CONDITIONS 1 AND 2" when both recirculation loops are in operation.
hGlM:
I I
With the recirculation pump speeds different by more than the specified limits, either:
N a.
Restore the recirculation pump speeds to within the specified limit within 2>.
hours, or
'I b.
Declare the recirculation loop of the pump with the slower speed not in operation and take the ACTION required by Specification 3.4.1.1.1.
II I
. "Iy
'A~
E.
},
J
',, Er
,';Qj>t'}"-
},.~ j glP j SURVEILLANCEREQUIREMENTS 4.4.1.3 Recirculation pump'speed mismatch shall be verified to be within the limits at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
I I
r,
'll See Special Test Exception 3.10.4.
SUSQUEHANNA - UNIT 1 3/4 4-3 Amendment No. $lt~
~43
LIMITINGCONDITION FOR OPERATION 3.4.2 The safety valve function of at least 12 of the following reactor coolant system safety/relief
)
valves shall be OPERABLE with the specified code safety valve function liftsettings:
2 safety-relief valves 9 1175 psig ~1%
6 safety-relief valves 9 1195 psig ~1%
8 safety-relief valves 9 1205 psig a196 OPERATIONALCONDITIONS 1, 2 and 3.
i Ai ~
1
%I <<
<<\\
<<1 i,
hGIN5',
b.
C.
With the safety valve function of one or more of the above required'safety/relief valves inoperable,ibe in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
<<<<g With one or more safety/relief valves stuck open, provided that suppressio'n'ool~'verage water temperature is less than 106'F, close the stuck open relief valve(s); if-"".
unable to close the open valve(s) within 2 minutes or if suppression pool water temperature is 106'F or greater, place the reactor mode switch in the Shutdown-"
position.
With one or more safety/relief valve acoustic monitors inoperable," restore the inoperable monitor(s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, Il i '"i
<< ~ <<j SURVEILLANCEREQUIREMENTS
<<<<'.4.2 The acoustic monitor for. each safety/relief valve shell be demonstrated OPERABLE with the setpoint verified to be 0.26 of the full open noise level by performance of a:
- a. CHANNELFUNCTIONALTEST at least once per 31 days, and a b.Calibration in.accordance with procedures prepared In.conjunction with its manufacturer's
".,':.;",'.,I'acorn'mendations'iit least once per 18 months.
The'lift'settfng'press'ure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.
Up to 2 inoperable valves may be replaced with spare OPERABLE valves with lower satpoints until the next refueling.
The provision 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.
t I
i' P
~E SUSQUEHANNA - UNIT 1 3/4 4-6 Amendment No. 88~
) 43
ger>>
LIMITING.CONDITION FOR OPERATION 3.4.3.2 a.
No PRESSURE BOUNDARYLEAKAGE, b.
5 gpm UNIDENTIFIEDLEAKAGE.
~>>'eactor'oolant system leakage shall be limited to:
>>-p Wpg t
f$,',
h>> 5,A,,$
d.
1 gpm leakage at a reactor coolant system pressure of 1035 k 10 psig from any reactor coolant system pressure isolation valve specified ln Table 3A.3.2-1.
hGIlQH',,c' i'.vr-F, c.
25 gpm total leakage averaged over any 24-hour period.
qt,'ir e.
2 gpm increase in UNIDENTIFIED LEAKAGE within any 24-hour period,, in;'.: "
.j";",j"':,','PERATIONAL CONDITION 1.
a.
With any PRESSURE BOUNDARY LEAKAGE, be ln at least HOT SHUTDOWN ",g;: ~~V~~ I within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b.
With any reactor coolant system leakage greater than the limits in b and/or c, above, reduce the leakage rate to within the limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
c.
With any reactor coolant system pressure isolation valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one closed manual or deactivated automatic valves, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
d.
With one or more of the high/low pressure interface valve leakage pressure monitors shown in Table 3.4,3,2-1 inoperable, restore the inoperable
-..':.,-;."=;" '.monitor(s).to OPERABLE status within 7 days or verify the pressure to be less
'-, -g<-,"-.,~Q,<thiri',the alarm pressure at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; restore the inoperable
';,~~a'-.'-'p.monitor(s)=,to OPERABLE status within 30 days or be in at least HOT
.',-, " ~~C<NHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the
=.'".'~-:~.,following:24 hours.
'"-"e." With'any reactor coolant system UNIDENTIFIED LEAKAGE.!ncrease greater than 2 gpm within any 24-hour period, in OPERATIONAL CONDITION 1 only.
identify the source of leakage increase as not service sensitive Type 304 or 318 austenitic stainless steel within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
II Pk SUSQUEHANNA - UNIT 1 3/4 4-7 Amendment No. jig, 'gg~
143
i'dr-'.tsar ',
crore e'
LIMmNGCONDITION FOR OPERATION 3.4.6.2The pressure in the reactor steam dome shall be less than 1050 psig.
Bauucr KIMH:
With the reactor steam dome pressure exceeding 1050 psig, reduce the pressure to less than 1050 psig within 15 minutes or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
, l.
SURVEILLANCEREQUIREMENTS 4.4.5.2 The reactor steam dome pressure shall be verlfled to be less than 1050 palp at least i.
" ~t";::. ",'l once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
I 4
'pet Not applicable during anticipated transients.
SUSQUEHANNA - UNIT 1 3/4 4-21 Amendment No. 143
SURVEILLANCaREQUIREINENTS 4.5.1 The emergency core cooling system shall be demonstrated OPERABLE by:
f~t
- a. At least once per 31 'days:
1.
For the CSS, the LPCI system, and the HPCI system:
~ PJ J,
1 sly.'I
"">'i).
a)
Verifying that the system piping from the pump discharge valve to the system isolation valve is filled with water by:
1.
Venting at the high point vents 2.
Performing a CHANNEL FUNCTIONAL TEST of the condensate transfer pump discharge low pressure alarm instrumentatlon..
2.
3.
4.
b)
Verifying that each valve, manual, power-operated, or automatic, in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
For the CSS, performance of a CHANNEL FUNCTIONAL TEST of the core spray header dP instrumentation.
For the LPCI system, verifying that at least one LPCI system subsystem cross-tie valve is closed with power removed from the valve operator.
For the HPCI system, verifying that the pump flow controller is In the correct position,"g:
b.
Verifying that, when tested pursuant to Specification 4.0.6:
1.
The two CSS pumps in each subsystem together develop a total flow of at least 6350 gpm against a test line pressure of greater then or equal to 269 psig, corresponding to a reactor vessel steam dome pressure of 2 105 psig.
0-2.
3.
Each LPCI pump in each subsystem develops a flow of at least 12,200 gpm against a test line pressure of I 204
- psig, corresponding to a reactor vessel to primary containment differential pressure h 20 psid.
The HPCI pump develops a flow of 'at least 6000 gpm against a test line pressure of 2 1140 psig when steam is being supplied to the turbine at 920, +140, -20 psig, c.
At least'once per 18 months:
C
,",;:.";a'.."; ~,) '
For-the CSS; the LPCI system, and the HPCI system, performing e system functional
.,>.~g"",.:-"-..'~I.~~.test;which. Includes simulated automatic actuation of the system throughout its zz,-.~ ~~<~~',emergency.'perating sequence and verifying that each automatic valve in the flow
."path'actuates to its correct position. Actual injection of coolant into the reactor vessel
,".-..+..'~.'- -'-miy be excluded from this test.
The provision of Specification 4.0A 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.
Except that an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may be in position for another mode of operation.
I SUSQUEHANNA - UNIT 1 3/4 5R Amendment No. Ã9, 143
t
The anticipated transient without scram (ATWS) recirculation pump trip system provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an anticipated transient.
The response of the plant to this postulated event falls within the envelope of study events in General Electric Company Topical Report NEDO-10349, dated March 1971 and NED0-24222, dated December 1979.
hl PP 'P h (
<<4hh' The endef-cycle recirculation pump trip (EOC-RPT) system is a part of the Reactor
'rotection System and is an essential safety supplement to the reactor trip. The purpose of the EOC-RPT is to recover the loss of thermal margin which occurs at the end-of-cycle.
The physical phenomenon involved is that the void reactivity feedback due-to-a pressurization transient can add positive reactivity to the reactor system at a faster'ate
'han the control rods add negative scram reactivity.
Each EOC-RPT system trips both,.'ecirculation pumps, reducing coolant flow in order to reduce the void collapse in the core-
~
',",,:.'-";."~""-'uring two of the most limiting pressurization events.
The two events for which the EOC..
~,,-".;-'-;;:;
RPT protective feature will function are closure of the turbine stop valves and fast closure'",';,"~"'-';,~
of the turbine control valves.
P A fast closure sensor from each of two turbine control valves provides input to the EOC-RPT system; a fast closure sensor from each of the other two turbine control valves provides input to the second EOC-RPT system.
Similarly, a position switch for each of two turbine stop valves provides input to one EOC-RPT system; a position switch from each of the other two stop valves provides input to the other EOC-RPT system. 'or each EOC-RPT
- system, the sensor relay contacts are arranged to form a 2wut-of-2 logic for the fast closure of turbine control valves and a 2wutwf-2 logic for the turbine stop valves.
The operation of either logic willactuate the EOC-RPT system and trip both recirculation pumps.
This function is not required when THERMAL POWER is below 30% of RATED THERMAL POWER.
The Turbine Bypass System is sufficient at this low power to accommodate a
turbine stop valve or control valve closure without the necessity of tripping the reactor recirculat(on.;;pumps...
This function is automatically "bypassed at turbine first stage pressures"Ieii.,th~;th'e.analytical limit of 147.7. psig, equivalent to THERMAL POWER of about~3Q-'RATEO'.THERMAL POWER.
Turbine first stage pressure of 147.7 psig is equivaleiitW22%'f. rated turbine load.
Each-..=EOC-RPf system may be manually bypassed by use of a keyswitch which is administratively controlled.
The manual bypasses and the automatic Operating Bypass at less than 30% of RATED THERMALPOWER are annunciated in the control room.
The EOC-RPT response time is the time assumed in the analysis between (nit(at(on of valve motion and complete suppression of the electric arc, i.e., 175 ms.
li I
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.
SUSQUEHANNA - UNIT 1 B 3!4 3-3 Amendment No. ~43
Ch UlD mZ RR C
Plate SA-633 GR B C2433-1 CL.1 BASES TABLEB 3/4.4.6-1 REACTOR VESSEL TOUGHNESS 0.10 0.63
+ 18 39.3 79.4 67.3 Weld N/A; -
8298161 L320A27AG 0.04 0.99
-60 32.6 103.7
-1 7.6 59?g:
" Table updated with surveillance test results reported in PLA<127, dated May 19, 1994.
W D
P Shell Ring Bottom Head Dome Bottom Head Torus Top Head Dome SA-633 GR B CL.1 C1 232-2 C9942-2 C9942-2 C9220-2
+20
+10 Top Head Torus C9366-1
+10
~0 R
O 4LA Top Head Flange Vessel Flange Feedwater Nozzle Recirculation Inlet Nozzle Weld Closure Studs SA-608, CL.2'A-640 GR B24 J ~
~
~
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N/A N/A Q2049W Q2Q49W No CNVS Available 82662
+10
+10
-18 0
+70 '
8 7.6x 10 I32 EFPY 4Q Yya Operating 7
+O x
6 S
4 c
O C
Z 0
10 20 Sclvlce Life lYeare')
Fast Neutron Fluence lR>> 1 lHev) at l.D. Surface*as n Function of Service Ufee Bases Hgure B 3/4.4AM
~ ~ ~
A Based on Power Uprate conditions as described in Pl%<127, dated'Miy 18,.-1884
~'.
1
The core spray system (CSS) is provided to assure that the core is adequately cooled foliowing a loss-of-coolant
- accident, and together with the LPCI mode of the RHR system, provides adequate core cooling capacity for all break sizes up to and including the double-ended reactor recirculation line break, and for smaller breaks following depressurization by the automatic depressurization system (ADS).
/
The CSS is a primary source of emergency core cooling after the reactor vessel is depressurized and a source for flooding of the core in case of accidental draining.
The surveillance requirements provide adequate assurance that the CSS. will be OPERABLE when required.
Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a,
complete functional test requires reactor shutdown.
The pump discharge piping I@
maintained full to prevent water hammer damage to piping and to start cooling at thW earliest moment.
//~ /.+ ///
The low pressure coolant injection (LPCI) mode of the RHR system is plovided to assure that the core is adequately cooled following a loss-of-coolant accident.
Two subsystems, each with two pumps, provide adequate core flooding for all break sizes up to and including the double-ended reactor recirculation line break, and for small breaks following depressurization by the ADS.
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/ '4j~f I The surveillance requirements provide adequate assurance that the LPCI system will be OPERABLE when required..Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a
complete functional test requires reactor shutdown.
The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the
, earliest moment.
The'high!j'iresiure'.,coolant'injection (HPCI) system.is provided to assure that the reactor core is,'a'f4qultiily cooled to limit fuel clad temperature in the event of a small break in the.-.rihactot>coolant-'s/ystem and loss of coolant which does not result in rapid depressuriiation of.",the reactor vessel.
The HPCI system permits the reactor to be shut down'-while',maintaining sufficient reactor vessel water level inventory until the vessel is depressunzed.
The HPCI system continues to operate until reactor vessel pressure is below the pressure at which CS system operation or LPCI mode of the RHR system operation maintains core cooling.
/
The capacity of the system is selected to provide the required core cooling.
Thy HPCI pump is designed to deliver greater than or equal to 5000 gpm at reactor pressures between 1187 and 150 psig.
Initially, water from the condensate storage tank is used
(
instead of injecting water from the suppression pool into the reactor, but no credit is taken in the safety analyses for the condensate storage tank water.
SUSQUEHANNA - UNIT 1 B 3/4 5-1 Amendment No. 29~ 14>
BASES The specifications of this section ensure that the primary containment pressure will not exceed the design pressure of 53 psig during primary system blowdown from full operating pressure.
The suppression chamber water provides the heat sink for the reactor coolant system energy release following a postulated rupture of the system.
The suppression chamber water volume must absorb the associated decay and structural sensible heat released during reactor coolant system blowdown from 1053 psia.
Since all of the gases in the
)
drywell are purged into the suppression chamber air space during a loss of coolant accident, the pressure of the liquid must not exceed 53 psig, the suppression chamber maximum pressure.
The design volume of the suppression chamber, water and air, was obtained by considerirfg that the total volume of reactor coolant and to be considered is discharged to the suppression chamber and that the drywell volume is purged to the, suppression chamber.
P Using the minimum or maximum water volumes given in this specification, containmenF@
-'ressure during the design basis accident is approximately'45.0 psig which is below the+"
design pressure of 53 psig.
Maximum water volume of 133,540 ft results in a downcomer submergence of 12 feet and the minimum volume of -122,410 ft results in a submergence approximately 24 inches less.
The majority of the Bodega tests were run with a submerged length of four feet and with complete condensation.
Thus, with respect to the downcomer submergence, this specification is adequate.
The maximum temperature at the end of the blowdown tested during the Humboldt Bay and Bodega Bay tests was 170'F and this is conservatively taken to be the limit for complete condensation of the reactor coolant, although condensation would occur for temperatures above 170'F.
Should it be necessary to make the suppression chamber inoperable, this shall only be done as specified in Specification 3.5.3.
Under full power, operating conditions, blowdown from an initial suppression chamber water temperaturo"-'of~90;F" results in a water temperature of approximately 128'F immediately foliowlrig;;64w'dow'ii'which is below the 170'F used for complete condensation via T-quencherj~yjcee.',=',~At',this temperature and atmospheric
- pressure, the available NPSH exceedsitflet"roqul lcd'b'y both the RHR and core spray pumps, thus there is no dependency on coritainment".,oyerpressure during the accident Injection phase.
If both RHR loops are used for "containment cooling, there is no dependency on containment overpressure for post-LOCA operations.
Experimental data indicates that excessive steam condensing loads can be avoided if the peak local temperature of the suppression pool is maintained below 200'F duripg any period of relief valve operation.
Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially high suppression chamber loadings.
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SUSQUEHANNA - UNIT 1 B 3/4 6-3 Amendment No. 79~
>4>
C 0
DESIGN FEATURES 5.3.1 5.3.2 The reactor core shall contain 764 fuel assemblies.
Each assembly consists of a matrix of Zircaloy clad fuel rods with an initial composition of non-enriched or slightly enriched uranium dioxide as fuel material and water rods.
Limited substitutions of Zirconium alloy filler rods for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used.
Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff-approved codes and methods, and shown by test or analyses to comply with all fuel safety design bases.
A limited number of lead use assemblies that have not completed representative testing may be placed in non-limiting core regions.
Each fuel rod shall have a nominal active fuel length of 158 inches.
Reload fuel shall have a maximum average enrichment of 4.0 weight percent U-235.
The reactor core shall contain 185 cruciform shaped control rod assemblies.-~-
The control material shall be boron carbide powder. (B4C),'nd/or Hafnium metal.
The control rod shall have a nominal axial absorber length of 143 inches.
Control rod assemblies shall be limited to those control rod designs approved by the NRC for use in BWRs.
5.4.1 The reactor coolant system is designed and shall be maintained:
a.
In accordance with the code requirements specified in Section 5.2 of the
", <.FSAR, with allowance for normal.degradation pursuant to the applicable
,. -.~:.'...Stiiveillance.'Requirements,
,:-b,",',;-'For,'a', j'ressure, of:
":,.t".'.1260 psig on the suction side of the recirculation pumps.
- 2. 1500 psig from the recirculation pump discharge to the Jet pumps.
c.
For a temperature of 575'F.
5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 22,400 cubic feet at a nominal Te~ of 532'F.
SUSQUEHANNA - UNIT 1 5-B Amendment No. DS~ EN~ ~43
.>>>>>>=
J COMPONENT Reactor
>>-r CYCLIC OR TRANSIENT UMIT 120 heatup and cooldown cycles 80 step change cycles 180 reactor trip cycles I
1 30 hydrostatic'pressure and leak tests OESIGN CYCLE OR TRANSIENT 70'F to 661'F to 70'F Loss of feedwater heaters 100% to 0% of RATED THERMALPOWER Pressurized to 2 930 psig and'1260 psig.
TABLE 5.7.1-1 COMPONENT CYCUC OR TRANSIENT LIMITS r
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lAkP>>
- *w"-".'4' I
'r '>>'>>
>>*>> E SUSQUEHANNA - UNIT 1 5-8 Amendment No. ~+~ i43
ADMINISTRATIVECONTROLS ra reel e E
(Continued)
- 10. PLA-2728, "Response to NRC Question: Seismic/LOCA Analysis of U2C2
- Reload, Letter from H.W. Keiser (PP&L) to E. Adensam (NRC), September 25, 1986.
9.
XN-NF-84-97, Revision 0, "LOCA-Seismic Structural Response of an ENC 9x9 Jet Pump Fuel Assembly," Exxon Nuclear Company, inc., December 1984.
A i re r
, q ~
- 11. XN-NF-82-06(P)(A), Supplement 1, Revision 2, "Qualification of Exxon Nuclear Fuel for Extended Burnup Supplement 1
Extended Burnup Qualification of ENC 9x9 Fuel," May 1988.
- 12. XN-NF-80-19(A), Volume 1, and Volume 1 Supplements 1 and 2,,"Exxon Nuclear Methodology for Boiling Water Reactors:
Neutronic Methods for Design and Analysis," Exxon Nuclear Company, Inc., March 1983.
~
~
'. ~:,';->'k ". I r
- 13. XN-NF-524(Al, Revision 1, "Exxon Noaleer Critlael Power Methodology far/:;':,';-;:.:::.:;:
Boiling Water Reactors, Exxon Nuclear Company, lnc., November 1983.
'.,=-.~".-;;"".'4.
XN-NF-512-P-A, Revision 1 and Supplement 1, Revision 1, "XN-3 Critical Power Correlation," October, 1982.
15.
NEDC-32071P, "SAFER/GESTR-LOCA.
Loss of Coolant Accident Analysis," GE Nuclear Energy, May 1992.
- 16. NE92-001A,.Revision 1, "Ucensing Topical Report for Power Uprate With Increased Core Row," Pennsylvania Power
& Light Company, December 1992.
r',
r, t
.,7.
NRC SER on PP&L Power Uprate LTR (November 30, 1993).
6.9.3.3<:The:.'core.'.oeperiting limits shall be determined such that all applicable limits
,:~,,;(ay,~.fuel'thermrahmechanical limits, core thermal4ydraulic limits, ECCS limits,
- '~<~ejiiQiar Ifmits such as shutdown margin, transient analysis limits and accident
'.-;, an~si3 limits) of the safety analysis are met.
SUSQUEHANNA - UNIT 1 6-20b Amendment No. fl'lls 143
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