ML20217F910

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Proposed Tech Specs Re Safety Limit Min Critical Power Ratio
ML20217F910
Person / Time
Site: Brunswick Duke Energy icon.png
Issue date: 03/27/1998
From:
CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML20217F907 List:
References
NUDOCS 9804010272
Download: ML20217F910 (13)
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Text

( . e 4

2.0 SAFETY LIMlTS AND LIMITING SAFETY SYSTEM SETTINGS

! 2.1 SAFETY LIMITS THERMAL POWER (Low Pressure or Low Flow) 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 800 psia or core flow less than 10% of rated flow.

l APPLICABILITY: CONDITIONS 1 and 2.

ACTION:

L With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel 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 (Hiah Pressure and Hiah Flow) 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.09* I with the reactor vessel steam dome pressure greater than 800 psia and core flow greater than 10% of rated flow.

APPLICABILITY: CONDITIONS 1 and 2.

ACTION:

With MCPR less than 1.09* and the reactor vessel steam dome pressure greater I than 800 psia a SHUTDOWN within.nd core flow greater than 10% of rated flow, be in at least HOT 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

l REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.

APPLICABILITY: CONDITIONS 1. 2. 3. and 4.

l l 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 l 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 />.

l

9804o10272 98032f E ApocK 0500 goa BRUNSWICK - UNIT 1 2-1 Amendment No.

(

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued)

'b. The core flow and core power adjustments for l

Specification 3.2.2.1.

c. The MINIMUM CRITICAL POWER RATIO (MCPR) for Specifications 3.2.2.1 and 3.2.2.2.
d. The rod block monitor upscale trip setpoint and allowable value t

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 i j described in the following documents.

a. NEDE-24011-P-A, " General Electric Standard Application for Reactor l Fuel" (latest approved version). '

l

b. The May 18, 1984 and October 22, 1984 NRC Safety Evaluation l Reports for the Brunswick Reload Methodologies described in:

l 1. Topical Report NF-1583.01 "A Description and Validation of

Steady-State Analysis Methods for Boiling Water Reactors,"

February 1983.

l 2. Topical Report NF-1583.02. " Methods of RECORD." February 1983.

l 3. To)ical Report NF-1583.03, " Methods of PREST 0-B," '

Fe]ruary 1983.

4. To)1 cal Report NF-1583.04 "Verificatica of CP&L Reference  !

l BW1 Thermal-Hydraulic Methods Using the FIBWR Code." May l 1983.

c. The NRC Safety Evaluation for Brunswick I!.iit 1 Amendment No. I j 6.9.3.3 The core operating limits shall be determined such that all a)plicable limits (e.g., fuel thermal-mechanical limits, core tiermal-hydraulic-limits. ECCS limits, nuclear limits such as shutdown margin, i transient analysis limits, and accident analysis limits) of the safety l

analysis are met.

6.9.3.4 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements shall be 3rovided, upon issuance for each reload cycle, to the NRC Document Control Desc with copies to the Regional Administrator and l Resident Inspector.

l BRUNSWICK - UNIT 1 6-23 Amendment No.

> ENCLOSURE 5 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO.1 DOCKET NO. 50-325/ LICENSE NO, DPR-71 SUPPLEMENT TO REQUEST FOR LICENSE AMENDMENT FUEL CYCLE 12 RELOAD LICENSING (NRC TAC NO. MA1044) l l

MARKED-UP TECHNICAL SPECIFICATION PAGES - UNIT NO.1 i

1 l

f' l-

7 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER (Low Pressure or low Flow)

! 2.1.1

! THERMAL POWER shall not exceed 25% of RATED THERMAL POWER reactor vessel steam dome pressure less than 800 psia or core flow less than 10% of rated flow. I

{

APPLICABILITY: CONDITIONS 1 and 2.

ACTION:

)

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor ve 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 />.

M RMAL' POWER (Hiah Pressure and Hioh Flow) I* M 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR shall not be less  ! than

! with the reactor vessel steam flow greater than 10% of rated flow.

dome pressure gr) eater than 800 psia and c l

APPLICABILITY: CONDITIONS 1 and 2.

ACTION:

With than 800MCPR psia andless core than f @ low greater than 10% 3ressure of ratedgreater flowI

  • and th SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

. Je in at least HOT l

l REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome. shall-not exceed 1325 psig.

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 coolarit 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 />.

2

-BRUNSWICK - UNIT 1 2-1 Amendment No. @

ADMINISTRATIVE CONTROLS l

, CORE OPERATING LIMITS REPORT (Continued)

b. The core flow and core power adjustments for Specification 3.2.2.1.

c.

The MINIMUM CRITICAL POWER RATIO (MCPR) for Specifications 3.2.2.1 and 3.2.2.2.

d. l The rod block monitor upscale trip setpoint and allowable value l for Specification 3.3.4.

l 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 described in thepreviously reviewed and approved by the NRC specifically those following documents.

a.

NEDE-24011-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 described in:
1. Topical Report NF-1583.01. "A Description and Validation of Steady-State Analysis Methods for Boiling Water Reactors."

February 1983.

2.

Topical 1983, Report NF-1583.02. " Methods of RECORD." February

3. To)ical Report NF-1583.03. " Methods of PREST 0-8."

Fe)ruary 1983.

4. Topical Report NF-1583.04. " Verification of CP&L Reference BWR Thermal-Hydraulic Methods Using the FIBWR Code." May 1983.

c.

The NRC Safety Evaluation for Brunswick Unit 1 Amendment No. @. l 6.9.3.3 The core operating limits shall be determined such that all a)plicable limits (e.g., fuel thermal-mechanical limits. core tiermal-hydraulic limits. ECCS limits. nuclear limits such as shutdown margin.

transient analysis areanalysis met. limits and accident analysis limits) of the safety 6.9.3.4 The CORE OPERATING LIMITS REPORT. including any mid-cycle revisions or supplements shall be )rovided upon issuance for'each reload cycle. to the NRC Document Control Desc with copies to the Regional Administrator and l Resident Inspector.

i BRUNSWICK - UNIT.1 6-23 AmendmentNo.4@

l i.

ENCLOSURE 6 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO.1 DOCKET NO. 50-325/ LICENSE NO. DPR-71 SUPPLEMENT TO REQUEST FOR LICENSE AMENDMENT FUEL CYCLE 12 RELOAD LICENSING (NRC TAC NO. MA1044) 4 3

l TYPED PAGE REVISION TO PREVIOUSLY SUBMITTED IMPROVED TECHNICAL SPECIFICATION (ITS) CONVERSION - UNIT NO.1 I

.l I I l'

l

I SLs

. 2.0 l

. 2,0 SAFETY LIMITS (SLs) 2.1 SLs s

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:

THERMAL POWER shall be s 25% RTP.

2.1.1.2 --------------------------NOTE---------------------------

MCPR SL values are only applicable for Cycle 12 operation.

With the reactor steam dome pressure a: 785 psig and core flow a: 10% rated core flow:

MCPR shall be a: 1.09 for two recirculation loop operation or at 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; ano 2.2.2 Insert all insertable control rods.

~

\

I Brunswick Unit 1 2.0-1 Amendment No.

i i l 1

Reactor Core SLs B 2.1.1 g3 .

BASES 1

APPLICABLE 2.1.1.3 Reactor Vessel Water Level SAFETY ANALYSES (continued) During NODES 1 and 2 the reactor vessel water level is

required to be above the top of the active irradiated fuel' L

" 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 during periods when the reactor is shut down, consideration must k 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 during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and cirA perforation in the event that the  !

water level becomes < 2/3 of the core height. .The reactor vessei 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 LIMITS 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  !

resultant clad perforations.  !

i The MCPR SL values are based on an NRC approved methodology that uses cycle specific input parameters. As a result, SL 2.1.1.2 is modified by a Note which restricts use of the l

MCPR values in SL 2.1.1.2 to Cycle 12 operation only.

l APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.

SAFETY LIMIT 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 B 2.0-4 Amendment No.

R:psrting R:quirements 5.6

,5.6 . Reporting Requirements- (continued)'

5.6.5 CORE OPERATING LIMITS REPORT (COLR)

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 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. NEDO-32339-A, " Reactor Stability Long Term Solution:  !

Enhanced Option I-A," July 1995. J

3. NEDC-32339-P Supplement 1, " Reactor Stability Long Term Solution: Enhanced Option I-A ODYSY Computer Code," i March 1994 (Approved in NRC Safety Evaluation dated i January 4, 1996). l
4. NEDO-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. NRC Safety Evaluation for Brunswick Unit 1 Amendment No. [ ).

(continued)

Brunswick Unit 1 5.0-19 Amendment No.

ENCLOSURE 7 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO.1 DOCKET NO. 50-325/ LICENSE NO. DPR-71 SUPPLEMENT TO REQUEST FOR LICENSE AMENDMENT FUEL CYCLE 12 RELOAD LICENSING (NRC TAC NO. MA1044)

MARK-UP FOR REVISION TO PREVIOUSLY SUBMITTED IMPROVED TECHNICAL SPECIFICATION (ITS) CONVERSION - UNIT NO.1 f

w . .

+

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Beactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10% rated core flow:

THERMAL POWER shall be s 25% RTP. 82 2.1.1.2 --------------------------NOTE--------------- ----------

MCPR SL values are~ only applicable for Cycle 1 operation.

With the reactor steam dome pressure at 785 psig and core flow at 10% rated core  :

1 01 MCPR shall be at .10 for two recirculation loop operation or at . for ng e recirculation loop operation.

1.10 2.1.1.3 Reactor vesse 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.

i Brunswick Unit 1 2.0-1 Amendment No.

Reacter Core SLs 8 2.1.1

, BASES APPLICABLE 2.1.1.3 Reactor Vessel Water Level SAFETY ANALYSES (continued) During MODES I 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 1 for protective system action to prevent exceeding acceptable j 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 requires.ents due to the effect

of decay heat. . If the water level should drop belew the top f

of the active irradiated fuel during this period, the l ability to. remove decay heat is reduced. This reduction in  ;

l cooling capability could lead to elevated cladding i temperatures and clad i water level becomes <2/3 perforation in the of the core eventThe height. thatreactor the 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.

l l SAFETY LIMITS 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 resultant clad perforations.

The MCPR SL values are based on an NRC approved methodology that uses cycle specific input parameters. As a result, j SL 2.1.1.2 is modified by a Note which restricts use of the -

MCPR values in SL 2.1.1.2 to Cycle ion only.

APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.

SAFETY LIMIT 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 B 2.0-4 Revision No.

e o Rep:rting Requirements 5.6 5 .'6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR)

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 SirJ1ated 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-24011-P-A, " General Electric Standard Application for Reactor Fuel" (latest approved version).
2. NEDO-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. NRC Safety Evaluation for Brunswick Unit 1 Amendment No. .

b 3 (continued)

Brunswick Unit 1 5.0-19 Amendment No.

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