ML20151L414

Errata to Amend 151 to License DPR-62,consisting of Tech Spec Pages 3/4 3-39 & B 2-2 Re Control Rod Withdrawal Block Instrumentation & Thermal Power (High Pressure & High Flow), Respectively

ML20151L414
Person / Time
Site:	Brunswick
Issue date:	04/18/1988
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20151L391	List: ML20151L384 ML20151L414
References
NUDOCS 8804220052
Download: ML20151L414 (2)

Text

INSTRUMENTATION 3/4.3.4 CONTROL ROD WITHDRAWAL BLOCK INSTRUNENTATION LIMITINC CONDITION FOR OPERATION 3.3.4 The control rod withdrawal block instrumentation shown in Table 3.3.4-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.4-2.

APPLICABILITY As shown in Table 3.3.4-1.

ACTION!

a. With a control rod withdrawal block instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.4-2, declare the channel inoperable until the channel is restored to CPERABLE status with its Trip Setpoint adjusted consistent with the Trip Setpoint value.

b. With the requirements for the minimum number of OPERABLE channels not satisfied for one trip system. POWER OPERATION may continue provided that eithert . .

1. The inoperable channel (s) is restored to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or

2. The redundant trip system is demonstrated OPIRABLE within 4 hetes and at Isast once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> until the inoperable channel is restored to OPERABLE status, and the inoperable channel is restored to OPERABLE status within 7 days, or ,

3. For the Rod Block Monitor only, THERMAL POWER is limited such l that the MCPR will remain above 1.04, assuming a single error l that results in complete withdrawal of any single control rod that is capable of withdrawal.

4. Otherwise, place at least one trip system in the tripped condition within the next hour.

c. With the requirements for the minimum number of OPERABLE channels not I satisfied for both trip systems, place at least one trip system sn '

the tripped condition within one hour.

d. The provisions of Specification 3.0.3 are not applicable in OPERATIONAL CONDITION 5.

SURVEILLANCE REQUIREMENTS 4.3.4 Each of the above required control rod withdrawal block instrumentation channels shall be demonstrated OPERABLE by the performance of a CHANNEL CHECK, CHANNEL CALIBRATION, and a CHANNEL FUNCTIONAL TEST during the OPERATIONU CONDITIONS and at the frequencies shown in Table 4.3.4-1.

BRUNSWICK - UNIT 2 3/4 3-39 Amendment No. 119, 151 8804220052 000410 4 PDR ADOCK 05000 P

SAFETY LIMITS BASES (Continued) _ _ _ _

2.1.2 THERMAL POWER (High Pressure and High Flow)

The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Sinco the parameters which result in fuel damage are rot directly observable during reactor operation, the ther.nal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure frem nucleate boiling would not necassarily result in da= age to BWR fuel rods, the critical power at which boiling transition is cal:ulated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power, result in an uncertainty in the value of the critical pcwer. Therefore, the fuel cladding integrity safety limit is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are .

expected to avoid boiling transition considering the power distribution within the core and all uncertainties.

The Safety Limit MCPR is determined using a statistical medel that

' ccmbines all of the uncertainties in operating parameters c.nd the procedures used to calculate critical power. The probability of the occurrence of boiling transiti:n is determined using an approved critical pcwer c:rrelation. Details of the fuel cladding integrity sa(ety limit calculation are given in Reference 1 and 2.

l Uncertainties used in the determination of the fuel cladding integrity safety li=it and the bases of these uncertainties are presented in Reference 1 l and 2. 1 The power distribution is based on a typical 764 assemoly core in which the rod pattern was arbitrarily chosen to produce a skewed power distribution having the greatest number of assemblies at the highest power levels. The worst distribution in Brunswick Unit 2 during any fuel cycle could not be as severe as the distribution used in the analysis. The pressure safety limits are arbitrarily selected to be the lowest transient overpressures allowed by the applicable codes, ASME soller and Pressure Vessel Code,Section III, and USAS Piping Code, Section B31.1.

Reference

1. "Ceneral Electric Standard Application for deactor Fuel,"

NEDE-24011-P-A, Revision 8.

l

2. "Ceneral Electric Standard Application for Reactor Fuel,'"

NEDE-24011-P-A, Amendment 14.

i BRUNSWICK - UNIT 2 B 2-2 Amendment No.151 4

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