ML20090E159

Ao:On 741110,fuel Channel Deflections Indicated Increase in Core Bypass Flow Over Calculations Assumed in Original Design.Channels Will Be Replaced.Finger Springs Will Be Used to Control Leakage

ML20090E159
Person / Time
Site:	Nine Mile Point
Issue date:	12/10/1974
From:	Schneider R NIAGARA MOHAWK POWER CORP.
To:	Goller K Office of Nuclear Reactor Regulation
References
12527, NUDOCS 8303110582
Download: ML20090E159 (3)

Text

. - .. ,

O O so u ;g NIAGARA MOHAWK POWER CORPORATION NtAGARA MOHAWK 300 ERIE DOULEVARD. WEST ,

SYRAcust. N. Y, t 3202 g3 December 10, I974 gh

.I ' -)

1

- v.DEC];; y74x y 9, s.cp mjf[*,q f'

/ tc.,yfQf Mt. Karl R. Gotter nJ I v' Assistant Director of Operatin5 Reaetchs c, ' fQay '

Dircetonitte of Licensing U.S. Atomic Energtj Comission

Wasitington, D.C. 20545 Dear Mt. Gotteri In accordance saltit tite Tecitnical Specifications for the Nine Mile Poin.t Nuctcar Sta, tion Unit HI, tite enetooed Unusual Event is be,btg submitted. This is in the accepted fonmnt as detailed in Regulatott) Guide 1.16, Rev.1.

VeAtj tratt) yours, '

/

if ? / . g)

,!k C. G La.usy

' V R.'R. c'hneider Vice Presidestt Electr.ic Operations TJ0/m Enc.

.{s303h0 7M No '-^

}D* .

. PPR- gg 1252;7

. ) /EttdL Y ' '

. COPY SENT REdiON-

• 2

\ . ,__

.w -

= a

sh

(

f3,/

NIAGARA MOH AWK POWER CORPORATION NI AG AR A MOHAWK NINE MILE POINT NUCLEAR PLANT UNUSUAL E. VENT

1. Docket No. 50-220 DPR-11 2a. Date December 10, 1974 2b. Occurrence Date November 10, 1974

3. Facility NY NMP #1

4. Identification of Occurrence Nine Mile Point Unit I has been informed by General Elcetric Co.,

car fuel s;.pplier, that their calculations of fuel __e channel deflections indicate an increase in core bypass flow over that' assumed on the original design. _

5. Conditions of Plant The reactor is pre'sently 2200 MKD/ST into the operating cycle with a targeted cycle orpnsure potential of 5500 MWD /ST.

6. Description of Occurrence '

"o .

Nine Mile Point Uisit #1 has had a channel measuring and evaluation program for the past several years. Primarily the interested channel deficctions are (1) elastic deficction due to differential pressure across the channel wall and (2) permanent deficction due to relaxation of fabricated stresses and inservice creep of irradiated zircalog.

These deficctions could potentially cause three problems: (1) an increase in core leakage flow (less water flow thru the fuel assembly) (2) less cicarance between the channel and the control rod (control rod interference) (3) less clearance between the corner fuel rod and the inside channel wall. The results of the studies and measurements indicated no control rod interference or corner rod to channel decrease clearance. The channel deficction study did show that an increase in bypass Icakage flow has ,c-curred. This concurs with a letter from Mr. J.A, Ilinds (G.E. Co.)

to the Atomic Energy Commission dated November 28, 1973. In this letter. It was stated that the major effect of channel deformation is a decrease in in-channel flow.

y .

~~-.

Pcg3 2 N s

Unusual Event

- NY:NMP #1:

75- Analysis and Corrective Action Changes in core bypass flow only.slightly affect the void coefficient and operational transient analyses. Because a boundary analysis

. approach was taken in performing the transient analysis, the affects of increased core-bypass flow are well within the applicability cf

~

the transient analyses and will remain so throughout this cycle. In-the future, channels will be replaced on an accelerated schedule and finger springs will be used on reload fuel to control leakage.

~

The thcrmal margins during normal operation will be maintained.

Minimum critical heat flux ratio will be maintained above 1.9 for al1~ fuel and the minimum critical power racks will be maintained.

above 1.24 and 1.29 for 7x7 and Ex8 fuel respectively. Conservative adjustments will be applied to our calculational procedures to account for reductions in in-channel flow.

The locus of operating conditions for which the most limiting rod has a minimum critical heat flux ratio equal to 1.0 as described by Figure 2.1.1 of the Technical Specifications is essentially unef-fccted. That curve assumes. an initial MCHFR of 1.9 at rated power and total core flow conditions. Because the reduction in.in-channel flow is accounted for in the calculations of ~ operation conditions,_

the ' initial MCHFh reamins at 1.9 frated power and cora flow),.and :-

~

the offect on the safety limit line is insignificant.

I Another' potential effect of channel bowing is the possibility of fuel channel-control- rod interference. Any. interference would be gradual and channels will be replaced well before:any expected problems. Scram times and other parameters will-be monitored to detect any ' onset- of. fue1 ~ channel-control- rod interference ~.

i i -

• 16 ,* -

(

w n d

u e d

9 t

- ,s * *:

.. i ~_,

c l

~

Q b > -

4- . . _

- , .c; ,. . 4 ,,__m