ML20196F555
| ML20196F555 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 06/22/1999 |
| From: | PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | |
| Shared Package | |
| ML20196F553 | List: |
| References | |
| NUDOCS 9906290243 | |
| Download: ML20196F555 (9) | |
Text
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ATTACHMENT 2 Limerick Generating Station Units 1 and 2 l
Docket Nos. 50-352 50-353 License Nos. NPF-39 NPF-85 l
Technical Specifications Change Request No. 99-01-0 l
List of Affected Pages Unit 1 Unit 2 3/44-2 3/4 4-2 6-18a 6-18a 9906290243 990622 PDR ADOCK 05000352 P
POR
w REACTOR' COOLANT SYSTEM SURVEILLANCE REQUIREMENTS
~4.4.1.1.1 DELETED 4.4.1.1.2 DELETED 4.4.1.1.3 Establish a baseline APRM and LPRM** neutron flux noise value within the regions for which monitoring is required (Specification 3.4.1.1, ACTION c) within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of entering the region for which monitoring is required unless baselining has previously been performed in the region since the last refueling outage.
4.4.1.1.4 With one reactor coolant system recirculation loop not in operation, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> verify that:
a.
Reactor THERMAL POWER is s 76.2% of RATED THERMAL POWER, b.
The recirculation flow control system is in the Local Manual mode, and c.
The speed of the operating recirculation pump is s 90% of rated pump speed.
d.
Core flow is greater than 39% when THERMAL POWER is within the restricted zone of Figure 3.4.1.1-1.
4.4.1.1.5 With'one reactor coolant system recirculation loop not in operation, within 15 minutes prior to either THERMAL POWER increase or recirculation loop flow increase, verify that the following differential temperature requirements are met if THERMAL POWER is s 30% of RATED THERMAL POWER or the recirculation loop flow in the operating recirculation loop is s 50% of rated loop flow, a.
s 145'F between reactor vessel steam space coolant and bottom head drain line coolant, b.
s 50*F between the reactor coolant within the loop not in operation and the coolant in the reactor pressure vessel, and c.
s 50*F between the reactor coolant within the loop not in operation and the operating loop.
The differential temperature requirements of Specification 4.4.1.1.5b. and c.
do not apply when the loop not in operation is isolated from the reactor pressure vessel.
1
LIMERICK - UNIT 1 3/4 4-2
i i
ADMINI5TRATIVECONTROLS CORE OPERATING LIMITS REPORT l
6.9.1.9 Core Operating Limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the CORE OPERATING LIMITS REPORT for the following:
)
a.
The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for
]
Specification 3.2.1, j
b.
MAPFAC(P) and MAPFAC(F) factors for Specification 3.2.1, c.
The MINIMUM CRITICAL POWER RATIO (MCPR) for Specification 3.2.3, d.
The MCPR(P) and MCPR(F) adjustment factors for specification 3.2.3, e.
The LINEAR HEAT GENERATION RATE (LHGR) for Specification 3.2.4, f.
The power biased Rod Block Monitor setpoints and the Rod Block Monitor MCPR OPERABILITY limits of Specification 3.3.6, g.
The Reactor Coolant System Recirculation Flow upscale trip setpoint and allowable value for Specification 3.3.6, h.
DELETED 6.9.1.10 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those
' described in the following document:
a.
NEDE-24011-P-A " General Electric Standard Application for Reactor Fuel" (Latest approved revision).
- 6.9.1.11 The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, transient analysis limits, and accident analysis limits) of the safety analysis are met.
j 6.9.1.12 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the i
NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.
- For Cycle 8, specific documents were approved in the Safety Evaluation dated (5/4/98) to support License Amendment No. (127).
)
LIMERICK - UNIT 1 6-18a
REACT 0kCOOLANTSYSTEM SURVEILLANCE REQUIREMENTS 4.4.1.1.1 DELETED 4.4.1.1.2 DELETED 4.4.1.1.3 Establish a baseline APRM and LPRM** neutron flux noise value within the regions for which monitoring is required (Specification 3.4.1.1, ACTION c) within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of entering the region for which monitoring is required unless baselining has previously been performed in the region since the last refueling outage.
4.4.1.1.4 With one reactor coolant system recirculation loop not in operation, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> verify that:
a..
Reactor THERMAL POWER is s 76.2% of RATED THERMAL POWER, b.
The recirculation flow control system is in tha Local Manual mode, and c.
The speed of the operating recirculation pump is s 90% of rated pump speed.
d.
- Core flow is greater than 39% when THERMAL POWER is within the restricted zone of Figure 3.4.1.1-1.
4.4.1.1.5 With one reactor coolant system recirculation loop not in operation, within 15 minutes prior to either THERMAL POWER increase or recirculation loop flow increase, verify that the following differential temperature requirements are met if THERMAL POWER is s 30% of RATED THERMAL POWER or the recirculation loop flow in the operating recirculation loop is s 50% of rated loop flow, a.
s 145*F between reactor vessel steam space coolant and bottom head drain line coolant, b.
s 50*F between the reactor coolant within the loop not in operation and the coolant in the reactor pressure vessel, and c.
s 50*F between the reactor coolant within the loop not in operation and the operating loop.
The differential temperature requirements of Specification 4.4.1.1.5b. and c.
do not apply when the loop notsin operation is isolated from the reactor pressure vessel.
i
LIMERICK - UNIT 2 3/4 4-2
ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT 6.9.1.9 Core Operating Limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the CORE OPERATING LIMITS REPORT for the following:
a.
The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for Specification 3.2.1, b.
MAPFAC(P) and MAPFAC(F) factors for Specification 3.2.1, c.
The MINIMUM CRITICAL POWER RATIO (MCPR) for Specification 3.2,3, d.
The MCPR(P) and MCPR(F) adjustment factors for specification 3.2.3, e.
The LINEAR HEAT GENERATION RATE (LHGR) for Specification 3.2.4, f.
The power biased Rod Block Monitor setpoints and the Rod Block Monitor MCPR OPERABILITY limits of Specification 3.3.6.
g.
The Reactor Coolant System Recirculation Flow upscale trip setpoint and allowable value for Specification 3.3.6, h.
DELETED 6.9.1.10 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following document:
a.
NEDE-240ll-P-A " General Electric Standard Application for Reactor Fuel" (Latest approved revision).
6.9.1.11 The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, transient analysis limits, and accident analysis limits) of the safety analysis are met.
6.9.1.12 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.
LIMERICK - UNIT 2 6-18a
PAGE 2 et REACTOR CCOLANT SYSTEM SURVEfLLANCE REOUIREMENTS 4.4.1.1.1 DELETED / DELETED 4.A.1.1.2 Each % M6 set scoop tube b enanical (demonstrtedOPERhhLEwithovrspeedsatoointsless nd elect ical stop an be han or a ual to the ke flow, at etpoints
. as noted n the COR( OPERATING LIMITS REPORT, as a pe centage f rated co Qeastone per 24 moh g 4.4.1.1.3 Establish a baseline APRM and LPRM** neutron flux noise value within the regions for which monitoring is required (Specification 3.4.1.1, ACTION c eithin 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of entering the region for which monitoring.is required unless) baselining has previously been perfomed in the region since the last refueling
)
outage.
4.4.1.1.4 With one reactor coolant system recirculation loop not in operation, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> verify that:
Reactor THERMAL POWER is s 76.2% of RATED THERMAL POWER, a.
b.
The recirculation flow control system is in the Local Manual mode, and The speed of the operating recirculaticn pump is s 90% of rated pump c.
speed.
d.
Core flow is greater than 3g5 when THERMAL POWER is within the restricted zone of Figure 3.4.1.1-1.
4.4.1.1.5 With one reactor coolant system rectreulation loop not in operation, eithin 15 minutes prior to either THERMAL POWER increase or recirculation loop flow increase, verify that the following differential temperature requirements are met if THERMAL POWER is s 30% of RATED THERMAL POWER or the recirculation loop flow in the operating rectreulation loop is s 50% of rated loop flow, s 145'F between reactor vessel steam space coolant and bottom head a.
drain line coolant, b.
s 50*F between the reactor coolant wMhin the loop not in operation and the coolant in the reactor pressure vessel, and s 50'F between the reactor coolant within the loop not in operation c.
1 and the operating loop.
The differential temperature requirements of Specification 4.4.1.1.5b. and c.
do not apply when the loop not in operation is isolated from the reactor pressure vessel.
i LIMERICK - UNIT 1 3/4 4-2 Amencment No. 3.33.7J,76,77.105 FEB121996
RAG _d_
10w1NTSTGsTIVr CONTROLS
t0:~ OP' ING L IMITS :~ 0RT 6.S.1.9 Core Operating Limits shall be established prior to each reicad cycle, er Orior to any remaining portien of a reload cycle, ano shaii ce cocumente: in the CORE OPERATING LIMITS REPORT for the following:
The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHG a.
. Specification 3.2.1, b.
MAPFAC(P) and MAPFAC(F) factors for Specification 3.2.1, The MINIMUM CRITICAL POWER RATIO (MCPR) for Specification 3.2.3, c.
d.
' The MCPR(P) ~and MCPR(F) adjustment factors for~ specification 3.'2.3, The LINEAR HEAT GENERATION RATE (LHGR) for Specification 3.2.4, e.
f.
The power biased Rod Block Monitor setpoints and the Rod Block Monitor MCPR OPERABILITY limits of Specification 3.3.6, The Reactor Coolant System Recirculation Flow upscale trt:
9 and allowable value fer Specification 3.3.6, setpoint DELETEn
- h. b [The etir lati MG et meckanical \\nd elehrical\\overspsed stopj Qetpo'nts r Sp ifi tion 4.\\4.1.1.2\\f 1
5.9.1.10 The analytical methods used to determine the core operating limits
.1 shall be those previously reviewed and approved by the NRC, specifically those described in the following document:
NEDE-24011-P-A " General Electric Standard Application for Reactor a.
Fuel" (Latest approved revision).
l 6.9.1.11 The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core
{
thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN analysis are met. transient analysis limits, and accident analysis limits) of the safety 6.9.1.12 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions' i NRC Document Control Desk with copies to the Regional A Resident Inspector.
SpECIAL rep 0RTS 6.9.2 Regtenal Office of the NRC within the time period specified 1
For Cycle 8, specific occuments were approved in the Safety Evaluati:n cated (5/4/98
) to support License Amencment No. (127 ),
MAY 1 31998 n, n, a LIMEF.!CX - UN!T 1 6-1Ba Amencment N:. 127 i
2" SURVEfLLENCE PEOUTREMENTS j
4.4.1.1.1 DELETEDycEO 4.4.1.1.2bfEacnpumpM3 set.icooptube\\mecnanicalanoelec oemonstr ted OPERABLE with oveqspeed setpoints isss than or\\rical stop shal n the CORE \\0PERATING L,IMITS REPORT, as a :ercentage\\ qual to t e.
setpoints as noted of rated re flow, at least once per 24 months.f 4.4.1.1.3 Establish a baseline APRM and LPRM** neutron flui noise value within the regions for which monitoring is required (Specification 3.4.1.1, ACTION c) within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of entering the region for which monitoring is required unless baselining has previously been perfomed in the region since the last refueling outage.
4.4.1.1.4 With one reactor coolant system recirculation loop not in operation, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. verify that:
a.
Reactor THERMAL POWER is s 76.2% of RATED THERMAL POWER, l
b.
The recirculation flow control system is in the Local Manual mode,
{
and c.
The speed of the operating recirculation pump is s 90% of rated pump speed.
d.
Core flow is greater than 39% when THERMAL POWER is within the restricted zone of Figure 3.4.1.1-1.
4.4.1.1.5 With one reactor coolant system recirculation loop not in operation, within 15 minutes prior to either THERMAL POWER increase or recirculation loop flow increase, verify that the following differential temperature requirements are met if THERMAL POWER is s 30% of RATED THERMAL j
POWER er the recirculation loop flow in the operating recirculation loop is i
s 50% of rated loop flow.
a.
s 145'F between reactor vessel steam space coolant and bottom head drain line coolant, b.
s 50*F between the reactor coolant within the loop not in operation and the coolant in the reactor pressure vessel, and c.
s 50*F between the reactor coolant within the loop not in operation and the operating loop.
The differential temperature requirements of Specification 4.4.1.1.5b. and c.
do not apply when the loop not in operation is isolated from the reactor pressure vessel.
- Detector levels A and C of one LPPJi string per core octant plus detectors A and C of one LFRM string in the center of the core should be monitored.
LIMERICK - UNIT 2 3/4 4-2 Amende.ent No. 27, 33, 51 FEB 1 6 rm
, v. o.; i - i i i w i i r u m
m _
CooE MERATING tiMTTS PEPORT l
4.9.1.9 Core Operating Limits snall be established prior to eacn reload I
cycle. or prior to any remaining portion of a reload cycle, anc shall ce cocumented in the CORE OPERATING LIMITS REPORT for the following:
a.
The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for Specification 3.2.1, b.
MAPFAC(P) and MAPFAC(F) factors for Specification 3.2.1, The MINIMUM CRITICAL POWER RATIO (MCPR) for Specification 3.2.3, c.
d.
The MCPR(P) and MCPR(F) adjustment factor for specification 3.2.3, The LINEAR HEAT GENERATION RATE (LHGR) for Specification 3.2.4, I
e.
f.
The power biased Rod Block Monitor setpoints and the Rod Block Monitor MCPR OPERABILITY limits of Specification 3.3.6.
g.
The Reactor Coolant System Recirculation Flow upscale trip setpoint l
j and allowable value for Specification 3.3.6, OELETED i
h.bheReci ulati MG Nset mAchankcal a d elh.ctrical setooNnts for 50 cifitation\\4.4.(.1.2f.
overspeecs3cg l
~
j 6.9.1.10 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following document:
a.
NEDE-24011-P-A " General Electric Standard Application for Reactor Fuel" (Latest approved revision).
i 6.9.1.11 The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, i
transient analysis limits, and accident analysis limits) of the safety analysis are met.
6.9.1.12 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
SPECIAL REPORTS 6.9.2 Special reports shall be suomitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.
~
LIMERICK - UNIT 2 6-ISa Ameno. men:.No. 4, 33, 43 JM 3 i 102 e