ML20244B409

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Proposed Tech Specs Modifying APRM Flow Biased Rod Block & Scram Limits to Allow Up to 100% Power Operation W/Core Flow Reduced to as Low as 87% of Rated W/Up to 50 F Reduction of Rated Final Feedwater Temp
ML20244B409
Person / Time
Site: Fermi DTE Energy icon.png
Issue date: 05/31/1989
From:
DETROIT EDISON CO.
To:
Shared Package
ML20244B402 List:
References
NUDOCS 8906130100
Download: ML20244B409 (8)
v • d • e


Text

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ATTACHMENT 2

SUMMARY

OF PROPOSED TECHNICAL SPECIFICATION / BASES CHANGES 8906130100 890531 PDR ADOCK 05000341 P' PDC

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4 ATTACHMENT 3 TECHNICAL SPECIFICATION AND BASES CHANGES 1

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(. .. POWER DISTRIBUTION LIMITS l

3/4.2.2 APM SETPOINTS l!MITINGCONDITIONFORJPERATION 3.2.2 The APM flow biased neutron flux-Mgh scram trip setpoint ($) and flow biased neutmn flux-high control rod block trip setpoint (5g) shall be i established according to the following relationships: ,'

{

TRIP SETP0 INT ALLOWABLE VALUE j

$< 5< +

us(0.h+h)T, (0.g . g a s(0 s.y where:

5 and Se are in percent of RATED THERMAL POWER, d W = Looparecirculation flow as a percentage of the loop recirculation j flow which produces a rated core flow of 100 million Ibs/hr, at 200K of RATED THERMAL POWER T = Lowest value of the ratio of FRACTION OF RATED THERMAL POWER divided by the MAXIMLM FRACTION OF LIMITING POWER DENSITY. T is applied only if legs than or equ21 to 1.0 -

APPLICABI .ITY: OPERATIONAL CONDITION 1, when THERNAL POWER is greater than or equal to Jan of RATED THERMAL POWER.

ACTION:

With the APRM flow biased neutron flux-high scram trip setpoint and/or the

( flow biased neutron flux-high control rod block trip setpoint less conservative than the value shown in the Allowable Value column for 5 or Sg , as above $

determined, initiate corrective action within 15 minutes and adjust 5 and/or

  • Sg otbe consistent with the Trip Setpoint value* within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> or reduce THERMAL POWER to less than 2SK of RATED THENWH HlWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS

' 4.2.2 The FRTP and the WLPD for each class of fuel'ahall be detemined, the

' value of T calculated and the most recent actual APRM flow biased neutron
  • V flux-high scras and flow biased neutron flux-high control rod block trip aetpoints verified to be within the above limits or adjusted or.the APRM gainreadingsshallbeverifiedasindicatedbelow,*asrequIred:
a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,
b. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion' of a TERMAL POWER incrisase of at least 15K of RATED THERMAL POWER, and

_,_ c . Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with E LPD greater than or er,ual to FRTP. i

d. The provisinns of Specification 4.0.4 are not applicable.

nWith WLPD greater than the FRTP during power ascension g to 90K of RATED THERMALPOWER;ratherthanadjustingtheAPRMsetpoints,theAPRMgainanybe  :

adjusted such that APRM readings are greater than or m t provided that the adjusted APRM reading does not exe. -100% pel to 200K of RATED times W LPD, THERMAL

i POWER and a notice of . adjustment is posted on the res.cor control panel.

FERMI - UNIT 2 3/4 2-5 Amendment No. g

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POWER DISTRIBUTION LIMITS BASES 

                      . 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady-state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR of 1.06, and an analysis of abnormal operational transients. For any abnormal operating transients analysis evaluation with the initial condition of the reactor being at the steady state operating limit, it-is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given'in Specification 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded i during any anticipated abnormal operational transient, the most limiting j transients have been analyzed to determine which result in the largest I reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the lar st delta MCPR. When added to the Safety Limit MCPR of 1.06, the required e inum operating limit MCPR of Specification 3.2.3 is obtained and presented in f Figure 3.2.3-1. The evaluation of a given transient begins with the system initial 

                     - parameters shown in FSAR Table 158.0-1 that are input to a GE-core dynamic behavior transient computer program. The code used to evaluate pressurization
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events is described in NED0-24154I3) and the program used in nonpressurization events is described in NEDO-10802(2) . The outputs of this program along with i the initial MCPR form the input for further analyses of the thermally limiting bundle with the single channel transient thermal hydraulic TASC code described in NEDE-25149I4}. The principal result of this evaluation is the reduction in MCPR caused by the transient. The purpose of the gK faci.or of Figure 3.2.3-2 is to define operating limits at other than rated core flow conditions. At less than 100% of rated flow the required MCPR is the product of the MCPR and thegK factor. The K f q factors assure that the Safety Limit MCPR will not be violated during a flow increase transient resulting from a motor generator speed control failure. The Kg factors may be applied to both annual and automatic flow control modes. { The K factor values shown in Figure 3.2.3-2 were developed generically and are app,licable to all BWR/2, BWR/3, and BWR/4 reactors. The Kg factors were c derived usin the f1 control line corresponding t RATED THERMAL POWER at rated core f ow,a ugk tkg we. WpHs4 tie. se(s.Jad apereHg MghW.. For the manual flow control mode, the K g factors were calculated such that for the maximum flow rate, as limited by the pump scoop tube setpoint and the corresponding THERMAL POWER along the rated flow control line, the limiting bundle's relative power was adjusted until the MCPR changes with different core flows. The ratio of the MCPR calculated at e given point of core flow, , , divided by the operating limit MCPR, determines the Kg. 8 FERMI - UNIT 2 8 3/4 2-4 A

, 1. Enclosure 1 GE' Nuclear Energy analysis by : H. .T. Kim, " Maximum Extended-Load Line Limit and Feedwater Heater Out of Service Analysis for Enrico Fermi Atomic Power Plant Unit 2", dated December 1987, NEDC-31515/DRF A00-03119/ Class II/EAS 108-1187-I _ _ __ - _- _ _ i}}

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