ML20203H048
| ML20203H048 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 02/23/1998 |
| From: | CAROLINA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML19317C912 | List: |
| References | |
| NUDOCS 9803030144 | |
| Download: ML20203H048 (14) | |
Text
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ENCLOSURE 8 IIRUNSWICK STEAM ELECTRIC PLANT, UNIT NO.1 DOCKET NO. 50425/ LICENSE NO. DPR 71 REQUEST FOR LICENSE AMENDMENT FUEL CYCLE 12 RELOAD LICENSING TYPED TECIINICAl, SPECIFICATION PAGES - UNIT NO. I
- BR R8sM 31888 6 P PDft
. ' 2. 0 SAFETY LIMilS AND LIMITING SAFETY SYSTEh SETTINGS 2,1 SAEETY LIMilS I THERMAL POWER (tcw Pressure or low Flow) i J
2.1.1 THERMAL POWtR shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressura less than 800 psia or core flow less than 10% of rated flow.
f APPLICABILITY: CONDITIONS 1 and 2.
ACTION:
a With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel i
. steam dome pressure less than 800 psia or core flow less than 10% of rated
- flow, 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 />.
THERMAL POWER (Hioh Pressure and Hiah Flow) 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.09 I i
- with the. reactor vessel steam dome pressure greater than 800 psia and core flow greater than 10% of rated flow.
APPLICABILITY: CONDil!ONS 1 and 2. l ACTION:
With MCPR less than 1.09 and the reactor vessel steam dome pressure greater l than 800 psia and core flow greater than 10% of rated flow, be in at least HOT SHUfDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
REACTOR /J0LANT SYSTEM PRESSURE
, 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel '
steam dome, shall not exceed 1325 ps19 APPLICABILITY: CONDITIONS 1, 2, 3, and 4.
ACTION:
With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure s 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
, I
' BRUNSWICK.'-: UNIT 1 '2 Amendment No.
ADMlblSiPXIIVE CD'ilROL5
[0RE OPERATltn LIMlls RFPORT (Continued)
- b. The core flow and core poner adjustments for Specification 3.2.2.1.
- d. The rod block monitor upscale trip setpoint and allowable value for Specification 3.3.4.
and shall be documented in the CORE OPERATING LIMITS REPORT.
6.9.3.2 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 documents.
- a. NEDE-240ll P-A. " General Electric Standard Application for Reactor Fuel" (latest approved version).
- b. The May 18. 1984 and October 22. 1984 NRC Safety Evaluation Reports for the Brunswick Reload Methodologies describri in:
- 1. Topical Repe ' NF-1583.01. "A Description and Validation of Steady-Statt ,aalysis Methods for Boiling Water Reactors."
February 1983.
- 2. Topical Report NF-1583.02 " Methods of RECORD." february 1983.
- 3. To31 cal Report NF-1583.03. " Methods of PRESTO-B."
Fearuary 1983.
- 4. .ical Report NF-lf 83.04 " Verification of CP&L Reference d Thermal-Hydraulic Methods Using the FIBWR Code." May 1383.
- c. Deleted. l 6.9.3.3 The core operating limits shall be determined such that all a3plicable limits (e.g.. fuel thermal-mechanical limits, core tiermal-hydraulic limits. ECCS limits, nuclear limits such as shutdown margin.
transient analysis limits, and accident analysis limits) of the safety analysis are met.
6.9.3.4 The CCCc OPERATING LIMITS REPORT. including any mid-cycle revisions or supplements shall be )rovided, upon issuance for each reload cycle, to the NRC Document Control Dest with copies to the Regional Administrator and Resident inspector.
BRUNSWICK - UNIT 1 6-23 Amendment No.
l ENCLOSURE 9 IIRUNSWICK STEAh! ELECTRIC l'LANT, UNIT NO.1 DOCKET NO. 50 325/ LICENSE NO. DPR 71 REQUEST FOR LICENSE AMENDMENT FUEL CYCLE 12 RELOAD LICENSING M ARKPD UP TECilNICAl, SPECIFICATION PAGliS - UNIT NO.1
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,., ;; . , ,.< s ,_ ,. , , . _
211 lid R"/J l On! R A 1 mt e ci 0 - f i- 't m , E J F. . i . t n
- n, .
reactor .'er.sel ;t vi de m pr onue - 'Mr :w D : s , i r <- 'E.,'. , + nan 10% of rated fira APPL}CMUlllY. C(4NDlil0% 1 and 2 ACT 10ti:
With THERMAL POWLR erceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome preuure less than 800 psia or core flon len than 10% of rated flos, be in at least HOI 5HUIDOWN within ? hours IHEPM/i P0 Q R (H3 h Pre g re 6nd High fird l, #
2.1.2 The MINIMUM CRITICAL POWER RAllo (MCPR) r. hall not be lez than h l with the reactor vezel steam dome pres-:. ore greater than 600 pua and tore flow greater than 10% of rated flow.
APPL IC ARil l iY : CCNDITIONS 1 and ?
ACTION I' M With MCPR less than - J and the reactor vessel steam dome pressure greater i than 800 psia und core t low greater than 10% of rated flo,v. 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 />.
PfACTOR_ COOLANT SYM [M PR[SSUR[
2.1.3 The reactor coolant system pressure. as measured in the reactor vessel steam dome, shall not exceed 1325 psig.
APPLICABilllY: CONDITIONS 1. ? 3 and 4.
ACT10%
With the reactor coolant system pressure, as measured in the reactor vessel steam dome above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure s 1326 psig wtthin 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
/ r /_
- CPP val.es
, i To no1.a1 ",)ect 1ca ion /.1 / a'e ar 11c ale .nly for Cyie/
ce atir n _
BRUNSWICK - UNIT 1 2-1 Amendment Na. #
l ALMINISTRAll;E ":NT ' , '
- . = a =.= = = .=======:===_.===== . . = . = = . = = - -
E;.7 r;i :: Ai p6 iMi' . ;F T g io n-
- b. The core flo., and core p ..,er :iajustents 'v Specification 3 2.2.1
- c. The MINIMUM CRITICAL PC.sER RATIO (MCPR) for 5recific3tions 3 2.2.1 and 3.2.2.2.
- d. The rod block monitor upscale trip setpoint and allowable value for Specification 3.3 4.
and shall be documen in the CORE OPERATING LIMITS REPORT.
6.9 3.2 The anal, al methods used to determine the core operating 11 nits shall be those orevi isly reviewed and approved by tt NRC. specifically those described in the foli 'ing documents.
- a. NEDE-240ll-P- A. " General Elec;ric Standard Application f or Reactor fuel" (latest approved version).
- b. The May 18. 1984 ana October 22, 1984 NRC Safety Evaluation Reports f or the Brunsuck Reload Methodologies descntsed in:
- 1. Topical Report NF-1583 01. "A Description and Validation of Steady-State Analysis Methods for Bolling Water Reactors."
February 1983.
2 Topical Report NF-lS83.02. " Methods of PECORD." February 1983.
- 3. Topical Report NF-1583.03. " Methods of PRESTO B."
February 1983
- 4. Topical Report NF-lS83.04. " Verification of CP&L Reference BWR Thermal-Hydraulic Methods Using the FIBWR Code." May 19R3.
h' eleted,)- wa rm 1,qt3cn <r h, u i t. 1 +'%~ maw "-
C.
]& l 6.9.3.3 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 l'mits, and accident analysis limits) of the safety analysis are met.
6.9.3.4 The CORE OPERATING LIMITS REPORT. Including any mid-cycle revisions or supplements shall be provided, upon issuance r each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident inspector.
l BRUNSWICK - UNIT 1 6-23 Anendment No. #
I!NCl.OSulti! 10 liitUNSWICK STl!AM lill!CTitlC Pl ANT, UNIT NO. I DOCKl!T NO. 50 325/LICl!NSli NO. Di'it 71 Ill!QUliST l'Olt 1,1CliNS!! AMIINDMl!NT 1:Ul!L CYCL.l! 12 Illil,OAD 1,1Cl!NSINO s
D'I'l!D l'ACdi.itliVISION TO l'ltliVIOUS1,Y SUllM((1E12 IMI'ItuYJilEUi.CIINICAL SI'l!Cll!JCATION (ITS) CONVI!!(SiON - UNI'IWal
. SLs
! 2.0 2.0 SAFETY LlHITS (SL')
2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10% rated core flow:
THERHAL POWER shall be s; 25% RTP.
2.1.1.2 With the reactor steam dome pressure 2: 785 psig and cora flow a: 10% rated core flow:
HCPR shall be a: 1.09 for two recirculation loop operation or 2: 1.10 for single recirculation loop operation.
2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.
2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be s; 1325 psig.
2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:
2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.
Brunswick Unit 1 2.0-1 Amendment No.
- Reactor Core SLs B 2.1.1 BASES
} APPLICABLE 2.1.1.1 Reactor _ Vessel Water level SAFETY ANALYSES (continued) During MODES 1 and 2 the reactor vessel water level is required to be above the top of the active irradiated fuel to provide core cooling capability, in conjunction with LCOs, the limiting safety system settings, defined in LCO 3.3.1.1 as the Allowable Values, establish the threshold for protective system action to prevent exceeding acceptable limits, including this reactor vessel water level SL, during Design Basis Accidents. With fuel in the reactor vessel dur".g periods when the reactor is shut down, consideration must be g# ien to 'sater level requirements due to the effect of decay nat. If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level becomes < 2/3 of the core height. The reactor vessel water level SL has been established at the top of the acthe irradiated fuel to provide a point that can be monitored and to Liso provide adequate margin for effective action.
SAFETY LlHITS The reactor core SLs are established to protect the integrity of the fuel clad barrier to prevent the release of radioactive materials to the environs. SL 2.1.1.1 and SL 2.1.1.2 ensure that the tore operates within the fuel design criteria. SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent elevated : lad temperatures and resultant clad perforations.
APPllCABILITY 3L ; 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.
SAFETY llHIT Exceeding an SL may cause fuel damage and create a potential VIOLATIONS for radioactive releases in excess of 10 CfR 100, " Reactor Site Criteria," limits (Ref. 2). Therefore, it is required (continued)
Brunswick Unit 1 8 2.0-4 Revision No. l
Responsibility 5.1 5.6 Reporting Reo"irements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR)
- a. Core operating limits shall be established prior to each roload cycle, or )rior to any famaining portion of a reload cyi.lo, and shall )e Acumented in the COLR for the following:
- 1. The AVERAGE PLANAk LINEAR HEAT GENERATION RATE (APLHGR) for Specification 3.2.1;
- 2. The MINIMUM CRITICAL POWER RATIO (MCPR) for Specification 3.2.2;
- 3. The Allowable Value for Function 2.b, APRM Flow Biased Simulated Thermal Power-High, for Specification 3.3.1.1; and
- 4. The Allowable Values and power range setpoints fer Rod Block Monitor Upscale Functions for Specification 3.3.2.1.
- b. 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 documents:
- 1. NEDE-240ll P-A, " General Electric Standard Application for Reactor fuel" (latest approved version).
- 2. NED0-32339-A, " Reactor Stability long Term Solution:
Enhanced Option I-A " July 1995.
- 3. NEDC 32339 P Supplement 1, " Reactor Stability Long Term Solution: Enhanced Option I A ODYSY Computer Code,"
Harch 1994 (Approved in NRC Safety Evaluation dated January 4, 1996).
- 4. NED0-32339 Supplement 3, " Reactor Stability Long Term Solution: Enhanced Option I-A Flow Mapping Methodology," August 1995 (Approved in NRC Safety Evaluation dated May 28,1996),
(continued)
Brunswick Unit 1 5.0 19 Amendment No.
ENCLOSUltE11 IIRUNSWICK STEAhi ELECTRIC l'LANT, UNIT NO. I DOCKET NO 50 325/ LICENSE NO. DPR 71 REQUEST FOR LICENSE AMENDh1ENT FUEL CYCLE 12 RELOAD LICENSING hiAltK-UP FOR REVISION TO l'RE' IOUSI,Y SUHhilTTED lhil' ROVED TECilNICAl. SI'ECIFICATION (ITS) CONVERSION - UNIT NO. l
l SLs .
2.0 l i
2.0 SAFETY llHITS (SLs) 2.1 SLs
- 2.1.1 Reactor Core SLs 1
- 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10% rated core flow
~
l THERMAL POWER shall be s 25% RTP.
2.1.1.2 --- .- TC -- OT -- - --- -- -
P SL lue ar ii .y pp cab f C e I
o rat n.
-Y With the reactor steam dome pressure 2: 785 psig ani core I flow 2: 10% rated core flow: '
- l. oi NCPR s 11 be 2: 1.10 for wo recirculation loop operstion or 2: .1 for sing recirculation loop operation.
.o '
- 2.1.1.3 Reactor vess water level shall be greater than the top of active irradiated fuel.
c 2.1.2 Reactor Coolant System Pressure SL .
Reactor steam dome prc1sure shall be s 1325 psig.
_ = _
2.2 SL Violations
^
With any SL violation, the foliowing actions shall be completed wi, thin i 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:
2.2.1 Restore compliance with all SLst and i
2.2.2 Insert all insertable sontrol rods.
'l Brunswick Unit 1 2.0 1 Amendment No.
Reactor Core SLs B 2.1.1 BASES
/PPLICABLE 2.1.1.3 !Leactor Vessel Water level SAFETY ANALYSES (continued) During MODES 1 and 2 the reactor vessel water level is required to be above the top of the active irradiated fuel to provide core cooling capability. In conjunction with LCOs, the limiting safety system settings, defined in LC0 3.3.1.1 as the Allowable Values, establish the threshold for protective system action to prevent exceeding acceptable limits, including this reactor vessel water level SL, during Design Basis Accidents. With fuel in the reactor vessel during periods when the reactor is shut down, consideration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top of the active irradiated fuel durir.g this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures ano ciaa perforation in the eye.: that the water level becomes < 2/3 of the core height. The reactor vessel water level SL has been established at the top of the active irradiated fuel to provide a point that can be monitored and to also provide adequate margin for effective action.
SAFETY LlHITS The reactor core SLs are established to protect the integrity of the fuel clad barrier to prevent the release of radioactive materials to the environs. SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fuel design criteria. SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent elevated clad temperatures and resul+ ant clad perforations.
--The MCPR SL va h es-are-based-on :n 'RC 2ppr4ved methodology.
4trat uses cycle speeH4e-input-paremeters. M: r+ w it,.
4L--4dd,-2-45-mod 4f4ed-by-a-Hete-wh4eh-eestfiet1Hise-of-the-44CPR values-4n-St 2.1. M-te4yele-H-eperaHon-onlyt t
APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.
SAFETY LlHIT Exceeding an SL may cause fuel damage and create a potential VIOLATIONS for radioactive releases in excess of 10 CFR 100, " Reactor Site Criteria," limits (Ref. 2). Therefore, it is required (continuedl Brunswick Unit 1 8 2.0 4 Revision No. l
Reporting Requirements 5.6 5.6 Reporting Requiraments (contirbed) 5.6.5 [0RE OPER! STING LIMITS REPORT (C0tR)
- a. 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 COLR for the following:
- 1. The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for Specification 3.2.1;
- 2. The MINIMUM CRITICAL POWER RATIO (MCPR) for #
Specification 3.2.2;
- 3. The Allowable Value for Function 2.b, APRM Flow Biased l Simulated Thermal Power-High, for Specification 3.3.1.1; and
- 4. The Allowable Values and power range setpoints for Rod Block Monitor Upscale Functions for Specification 3.3.2.1.
- b. 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 documents:
- 1. NEDE 240ll-P-A, " General Electric Standard Application for Reactor Fuel" (latest approved version).
- 2. NE00-32339-A, " Reactor Stability Long Term Solution:
Enhanced Option I-A," July 1995.
- 3. NEDC-32339-P Supplement 1, " Reactor Stability Long Term Solution: Enhanced Option I-A ODYSY Computer Code,"
Harch 1994 (Approved in NRC Safety Evaluation dated January 4, 1996).
- 4. NED0-32339 Supplement 3, " Reactor Stability Long Term Solution: Enhanced Option I-A Flow Mapping Methodology," August 1995 (Approved in NRC Safety Evaluation dated May 28,1996).
- 5. PC Safety Evaluatien for Brunswick Unit ! %endmc 4 No. 102.
(contiroed)
I Brunswick Unit 1 5.0-19 Amandment No. I