ML20216J411

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Proposed Tech Specs,Removing cycle-specific Min Critical Power Ratio Safety Limit Restriction
ML20216J411
Person / Time
Site: Brunswick Duke Energy icon.png
Issue date: 09/28/1999
From:
CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML20216J407 List:
References
NUDOCS 9910050118
Download: ML20216J411 (7)
v • d • e


Text

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ENCLOSURE 5 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO. I DOCKET NO. 50-325/ LICENSE NO. DPR-71 -

REQUEST FOR LICENSE AMENDMENT - REMOVAL OF A CYCLE-SPECIFIC MINIMUM CRITICAL POWER RATIO SAFETY LIMIT RESTRICTION .

TYPED TECHNICAL SPECIFICATION PAGES - UNIT 1

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7. .

SLs 2.0 i

j .c 2.0 SAFETY LIMITS (SLs) 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:

l THERMAL POWER shall be s 25% RTP.

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

MCPR shall be a 1.09 for two recirculation loop operation or k 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.

t 2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be s 1325 psig.

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2.2 SL Violations -

I 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:

l 2.2.1 Restore compliance with all SLs; and l

2.2.2 Insert all insertable control rods.

Brunswick Unit 1 2.0-1 Amendment No. I L

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Reporting Requirements  !

5.6 l

! 5.6 Reporting Requirements (continued) -

i 5.6.5 CORE OPERATING LIMITS REPORT (COLR) l

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).

l 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,"

March 1994 (Approved in NRC Safety Evaluation dated January 4, 1996).

4. NE00-32339 Supplement 3, " Reactor Stability Long Term Solution: Enhanced Option I-A Flow Mapping Methodo%gy," August 1995 (Approved in NRC Safety Evalua; ion dated May 28,1996).

I (continued) 1 1

Brunswick Unit 1 5.0-19 Amendment No. l l

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ENCLOSURE 6 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO. I DOCKET NO. 50-325/ LICENSE NO. DPR-7i REQUEST FOR LICENSE AMENDMENT- REMOVAL OF A CYCLE. SPECIFIC MINIMUM CRITICAL POWER RATIO SAFETY LIMIT RESTRICTION MARKED-UP TECHNICAL SPECIFICATION PAGES - UNIT 1 l

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SLs

! 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs J 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.

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. . . . . . . . ./:

2.1.1. 2 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 a 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.

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! Brunswick Unit 1 2.0-1 Amendment No. @

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLM

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 I 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; l 1
3. The Allowable Value for Function 2.b, APRM Flow Biased I 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. 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,"

March 1994 (Approved in NRC Safety Evaluation dated January 4, 1996). {

4. NED0-32339 Supplement 3, " Reactor Stability Long Term l Solution: Enhanced Option I-A Flow Mapping Methodology," August 1995 (Approved in NRC Safety l Evaluation dated May 28,1996). 1

. C fet ati n c n td J/

(continued)

Brunswick Unit 1 5.0-19 Amendment No. @

Reactor Core SLs B 2.1.1 BA.SES .

l l

L APPLICABLE 2.1.1. 3 . Reactor Vessel Water Level SAFETY ANALYSES I

.(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 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 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 active irradiated fuel to provide a point that can be i monitored and to also provide adequate margin for effective l action.

SAFETY LIMITS The reactor core SLs are established to protect the 1 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 1 design criteria. SL 2.1.1.3 ensures that the reactor vessel j water level is greater than the top of the active irradiated fuel in order to prevent elevated clad temperatures and resultant clad perforations, h MCP S va ue ar b ed a NRC app ved met odol gy th t es yc}e ec fic inp pa ame ers As a r sul ,

S 2. .l. 11f if ed ya ote whi h r tri ts se th PR val es n L .l. 2 Cy e 2 op rat on nly.

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. O

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