ML20037A949

Application for Proposed Amend 15 to License DPR-2,App 2, Permitting Reactor Refueling W/O Cocked Rod Use During Fuel Insertion

ML20037A949
Person / Time
Site:	Dresden
Issue date:	01/17/1968
From:	Julie Hughes COMMONWEALTH EDISON CO.
To:
References
NUDOCS 8008250829
Download: ML20037A949 (5)

Text

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January 17, 1968

, '3 Dr. Peter A. Morris, Director Division of Reactor Licensing U. S. Atomic Energy Commission

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20545

Subject:

Proposed Change No. 15 to the Operating License DPR-2 as Atended - Okt 90-10 f-D

Dear Dr. Morris:

Pursuant to 10 CFR 50.59 and paragraph 3.a(4) of License DPR-2, as amended ("DPR-2"), Commonwealth Edison requests that Appendix A of DPR-2 be changed to allow the refueling of the Dresden reactor without the use of cocked rods during fuel insertion as required in "DPR-2."

Reference is made to Propased Change No. 14 of September 14, 1967 and as revised as of January 17, 1968, wherein the cocked rod refueling accident accident is evaluated.

Revise item "5.b."

of Section "D.

Power Operation" of Appendix "A" to DPR-2 to read in its entirety:

5.

Reactivity Limits b.

With the reactor in any condition, the following shutdown criterion shall be met:

" Stuck Rod" Criterion:

At every stage during loading and in the fully loaded configuration, the control rods must provide a shutdown control margin of at leat 0.012k with any rod wholly out of the core and completely unavailable.

Revise item "3" of Section "E.

Refueling and Maintenance" by deletion of the second sentence of said paragraph 3, and the addition of the following sentence:

The loading procedures shall require verification that the reactor is safely suberitical by with-drawing and re-inserting a control rod in the vicinity of the refueling activity before and af ter each fuel addition, q

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.m Dr. Peter A. Morris January 17, 1968 The justification for this proposed change was submitted originally as the answer to Question 2 in Supplement A to the Description and Safety Evaluation Report for Proposed Change No. 14 to DPR-2; however, in the interest of continuity and clarity the answer to Question 2 of said report is restated below:

"The fact that the Type VI-I fuel assembly reactivity is greater than for previous Dres. den fuel necessitated the re-examination of the refueling accident.

The accident postulated is the lowering of a fuel assembly at the maximum design hoist rate into a near-critical core during refueling.

The refueling accident has been hypothesized from a number of procedure violating circumstances and represants an accident of extremely low probability.

Conservative assumptions for this analysis are as follows:

1.

Two control rods, next to the fuel position to be loaded, have inadvertently been withdrawn to give a near critical 2 x 4 fuel assembly array witn one center fuel assembly missing.

The members of the refueling crew fail to note that these.two control rods are withdrawn and start to load a fuel assembly into the vacant position.

2.

The reactor operator in the control room fails to notice the indications from his instruments that the control rods are out and that the reactor is near critical prior to loading.

Procedures require him to observe this instrumentation and to be in communication with the members of the refuelina crew during all fuel loading operations.

3.

The assembly is inserted into the vacant fuel position at the maximum design rate of the hoist, 12 in/sec.

4.

The fuel assembly reactivity worth is 1.5% k.

Analysis i'ndicates that this is the maxinum potential reactivity worth for Type VI-I fuel in the reload array shown in Fig. 2 of the Safety Evaluation Report.

5.

The period scram circuitry f ails.

6.

The initial fuel and moderator temperature is 68'F.

The power level at initial criticality is 10"g times rated pcwer (i.e., 7 watts).

7.

The calculational model includes no negttive reactivity feedback effect from moderator or clad ?. eating or void formation.

Only negative Doppler feedback is considered.

Control rod motion is assumed not to start until 0.2 seconds after the scram signal of 120% of rated power.

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Dr. Peter A. Morris January 17, 1968 Based on the above assumptions the calculations indicate that the maximum radially averaged fuel 'emperature in any fuel rod would be less than 2800*F, and the corresponding maximum central fuel temperature would be less than 3200 F.

Clad temperatures will not exceed 2200*F.

No fuel or clad melting is expected in this accident.

The rapid power rise during the accident is terminated by the negative Doppler reactivity due to the fuel temperature rise, and the reactor is subsequently brought subcritical by the overpower scram.

It will be noted that this accident is less severe than the one reported in the Cycle 4 Safety Evaluation Report for III-F fuel because a smaller number of fresh fuel assemblies are loaded into the core at 30C 6 than at 30C 4, and because of the higher exposure accumulated in th'e remaining assemblies.

Analysis of the equivalent accident, but without scram,

- indicates that bulk fuel melting would occur in several rods of two fuel assemblies, but that the peak fuel enthalpy

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' rise would be insufficient to produce explosive fuel rod rupture."

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Tnerefore,, in consideration of the foregoing analysis, it is proposed that '

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The Ei ce,n s'es requirement for " cocked rods" during refdeling' operations be eliminated.

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The fuel'Ioading procedures be nodified to require l

verification that the reactor is safely suberitical by witlidrawing and re-inserting a control rod in the vicinity of the refueling activity before and L

after each fuel addition.

3)

Control rod withdrawal during movement of the fuel s

ovsr/or into the. reactor core be prohibited.

In our opinion Proposed Chan7,e No. 15 shall not result sin hazards which are greater ttan or different from, those analyzed p,

in the Hazards Sur. mary. Report, specifically there is (1) no increase in the probability of, or (2) no increase in the consequence of, M

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z-am 4-January 17, 1963 Dr. Peter A. Morris or (3) the creation of a credible probability of an accident different from, those accidents previous.y analyzed in the llazards Summary Report as amended or in connection with the amendments and. changes i

to Operating License DPR-2.

Very truly yours, COWL 0NWEALTil EDISON COMPANY ohn 11. Hughes Nuclear Licensing Administrator Subscribed and sworn to befgre me this /7 day ofV e @

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