ML17341A824

Submits Response to NRC Questions Re Safety Analysis for Positive Moderator Temp Coefficient Operation. Reload Safety Evaluation,Turkey Point Plant Unit 4,Cycle 6,Revision 1 Encl

ML17341A824
Person / Time
Site:	Turkey Point
Issue date:	01/20/1982
From:	Robert E. Uhrig FLORIDA POWER & LIGHT CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
Shared Package
ML17341A825	List: ML17341A824 ML17341A826
References
L-82-24, NUDOCS 8201260368
Download: ML17341A824 (8)

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FLORIDA POWER & LIGHT COMPANY January 20, 1982 L-82-24 Office of Nuclear Reactor Regulation Attention: Mr. Darrell G. Eisenhut, Director

. Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555 8 Pyric,

Dear Mr. Eisenhut:

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Re: Turkey Point Unit 3 S 4 Docket Nos. 50-250 and 50-251 Proposed License Amendment Moderator Tem erature Coefficient On December 10, 1981 we sent you the subject proposed amendment with an attached safety evaluation (FPL letter L 517). Mr. Ron Frahm requested clarification concerning this submittal which was discussed by telephone on January 18, 1982. A formal response is submitted with this letter. Since the response refers to the Reload Safety Evaluation of Turkey point Unit 4 Cycle 6, a copy of it is also attached.

Very truly yours, Robert E. Uhrig Vice President Advanced Systems S Technology REU/SKM/jc cc: Mr. J. P. O'Reilly, Region TX Harold F. Reis, Esquire Attachment 82012gpg~

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FPL Response to the NRC Questions on the Safety Analysis for Positive Moderator Temperature Coefficient Operation for Turkey Point Units 3 E 4 Question  : 1 The analysis of the Rod Ejection indicates a peak hot spot clad temperature of 2 2 1 0-F in the write up while a value of 23 6 7 -F is indicated in Table I II . Which is the correct value2 Response :

The value of 23 6 7 -F indicat ed in Table' I I is the correct value .

Question  : 2 The analysis of the Rod E j ection concludes that the fuel and clad temperature limits specified in the FSAR are not exceeded . Have the limits specified in WCAP 7 5 88 been veri fied?

Response  :

In the pre s ent, analysis , the fuel performance values are below the limits specified in WCAP 7 5 88 .

Que stion : 3 In the Locked Rotor analysis , the FSAR quote s a value of 2 5 1 0-F for peak average pellet temperature . What is the corresponding value for the pre s ent ana lysis 2 Response  :

The peak average pellet t emperature during the t rans ient calculated for the present analysis is 2 1 3 7 -F .

Question : 4 For the Uncontrolled Rod Withdrawal , peak heat flux in the present analysis is much higher than the FSAR value .

Expla in the reasons for this behavior and'he cons equence s .

Response  :

The FSAR analysis is based on a reactivity ins ert ion rat e of 6 OX1 0-5 delta -k/ s ec while the present analysis is based on a reactivity insertion rate of 7 5Xl0- 5 de 1 ta -k/ s ec .

Also the doppler power defect and prompt neutron lifetime are d ifferent from the FSAR values ( see Unit 4 , Cycle 6 RSE ) . The following description of this acc ident analysis is provided to adequate ly address the que s tion . The description refers to figures 2, 3 4 of the sub j ect S

submittal .

0 Control Rod Withdrawal From a Subcritical Condition I INTRODUCTION A control rod assembly withdrawal incident when the reactor is subcritical results in an uncontrolled addition of reactivity leading to a power excursion (Section 14.1.1 of the FSAR). The nuclear power response is characterized by a very fast rise terminated by the reactivity feedback of the negative fuel temperature coefficient. The power excursion causes a heatup of the moderator and fuel. The reactivity addition due to a positive moderator coeffic-ient could result in increases in peak heat flux, peak fuel, and clad temperatures. The time the core is critical before a reactor trip is very, short so that the coolant temperature does not increase significantly.

Hence, the effect of a positive moderator coefficient is small.

II METHOD OF ANALYSIS The analysis was performed in Unit 4 Cycle 6 RSE for a reactivity insertion rate of 75X10-5 delta-k/sec. This reactivity insertion rate assumed is greater than that for the simultaneous withdrawal of the combination of the two sequential contxol banks having the greatest combined worth at maximum speed (45 inches/minute). A constant moderator temperature coefficient of +5 pcm/degrees-F was used in the analysis. The digital computer codes, ini-tial power level, and reactor trip instrument delays and setpoint errors used in the analysis were the same as used in the FSAR, subsequent safety analyses, and WCAP-8284 Rev. 2, (Florida Power and Light Turkey Point Units 3 and 4 Precautions, limitations, and Setpoints).

III RESULTS AND CONCLUSIONS The nuclear power, coolant temperature, heat flux, fuel average temperature, and clad temperature versus time for a 75 X 10-5 delta-k/sec insertion rate are shown in Figures 2 through 4. This insertion rate, coupled with a positive moderator temperature coefficient of +5 pcm/degrees-F, yields a peak heat flux which does not exceed the nominal value.

Taking into account the conservative assumptions with which the accident has been analyzed, it is concluded that in the unlikely event of a control rod withdrawal acci-dent, the core and reactor coolant systems are not adversely affected since the thermal power reached is less than the nominal value. and the core water 'temperature reached is 564 degrees-F compared to 577 degrees-F for the nominal conditions. These conditions (subcooled coolant,,

less than nominal heat flux) show that the minimum DNBR is well above 1.30. No damage could occur to the cladding due to low temperature (less than 650 degrees-F) if com-pared to the melting point (greater than 3200 degrees-F).

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