ML20147D938

Responds to Request for Addl Info Re 871021 Application to Revise Tech Specs Concerning Fuel Thermal Limits & Refueling Operations.Revised Pages of Unit 2 Tech Specs Incorporating Changes Approved by Amend 87 Encl

ML20147D938
Person / Time
Site:	Hatch
Issue date:	01/14/1988
From:	Gucwa L GEORGIA POWER CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
SL-3934C, TAC-66524, TAC-66525, NUDOCS 8801200383
Download: ML20147D938 (4)

Text

___________ - ____

• Georgia Fbwor Company

'. hant , a ga 3 3 8 Teepnone 401526-0526 st f 54 5 Atlanta, Georgia 30302 Georgia Power ,

L T. Oucwa '* * * "6~

• V

• Manager Nuclear Safety and Ucensing SL-3934c 1809C X7GJ17-H600 January 14, 1988 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Hashington, D. C. 20555 PLANT HATCH - UNITS 1, 2 NRC DOCKETS 50-321, 50-366 OPERATING LICENSES DPR-57, NPF-5 REQUEST FOR ADDITIONAL INFORMATION REFUELING OPERATIONS Gentlemen:

This letter responds to your request for additional information *

(L. P. Crocker/J. D. Heidt teieconference of December 15, 1987) ,

concerning Georgia Power Company's (GPCs) submittal of October 21, 1987 -

"Request to Revise Technical Specifications: Fuel Thermal Limits and i

Refueling Operations". Specifically, the information requested concerns refueling practices at Plant Hatch Units 1 an6 2 including the loading of fuel bundles around Source Range Monitors (SRMs),

i Refueling outages at both Plant Hatch units typically involve the l

' complete offload of fuel bundles using a spiral officading strategy starting at the periphery and progressing toward the center of the core. '

This fuel offload practice assures that the previous cycle Shutdown Margin (SDH) calculations and deecnstrations renin valid so that Technical Specifications (TS) SDH requirements are satisfied.

Refueling the core begins by temporarily loading two to four exposed fuel bundles from the previous cycle around the SRMs (as required by the current TS) to obtain the TS required channel count rate. The I temporarily loaded bundles will be less reactive than the bundles '

scheduled for these locations in the upcoming cycle. I Refueling proceeds by spiral loading of the core, from the center outward, until the core is completely loaded. The core configuration present at this time would be different from the scheduled full core i loading due to the temporarily loaded bundles around the SRMs. The SDH for this intermediate configuration would be bounded by the SDH calculated for the final loading configuratioa since the temporarily loaded bundles are less reactive than the bundles scheduled for these locations. Fuel loading is completed by removing the temporary bundles y and inserting the scheduled bundles in their place.

880120o383 880114 PDR I P

ADOCK 05000321 DCD 31 l '

. l Georgialbwer d  !

U. S. Nuclear Regulatory Commission January 14, 1988 Page Two The SRM loading practices proposed in the October 21, 1987 submittal would allow the loading of any combination of up to four fuel bundles around each SRM to obtain the required count rate during refueling. As described in the October 21, 1987 submittal, the subtriticality of any such configuration will be assured by the bounding 2x2 spent fuel pool analysis described in the Hatch 2 Final Safety Analysis Report.

Consistent with current practice, any intermediate core loading configuration will be evaluated to ensure that the TS required SDM exists. Therefore, although the proposed change will allow the loading of "any" four assemblies around each SRM during refueling, the actual bundles loaded will be, overal3, subject to the current Technical Specifications SDM requirements.

Per the L.P. Crocker/J. D. He10t teleconference of January 5,1988, attached are revised pages to the Phnt Hatch Unit 2 TS, that were previously transmitted to the NRC in our October 21, 1987 submittal.

These TS pages (3/4 2 formerly 3/4 2-4g, and B 3/4 2-2) were modified to incorporate changes recently approved by TS Amendment 87. These '

changes are purely administrative in nature and in no way alter the ,

discussion presented in our October 21, 1987 submittal.

~

If you have any questions in this regard, please contact J. D. Heidt of my staff at (404) 526-4530.

Sincerely,

1. rM u L. T. Gucwa GDP/lc

Enclosure:

l c: Georaia Power Company Mr. J. P. O'Reilly, Sr. Vice President - Nuclear Operations ,

Mr. J. T. Beckham, Jr., Vice President - Plant Hatch ,

GO-NORMS ,

U. S. Nuclear Regulatory Commission. Washington. D. C.

Mr. L. P. Crocker, Licensing Project Manager - Hatch U. S. Nuclear Regulatory Commission. Region II Dr. J. N. Grace, Regional Administrator a Mr. P. Holmes-Ray, Senior Resident Inspector - Hatch 700775 __

s

• b 1

AVERAGE PLANAR LINEAR HEAT GENERATION RATE vs AVERAGE PLANAR EXPOSURE

=

• x; gg/ // 7,7 /';;/7 y /// :'"

13.0

' # 10*20 x -/ //,///////////I' /// ////.

//' / .

///

/ , eiI par e 4 p.7y / tz.4 p.7p UNACCEPTABLE OPERATION / e' O 12  ! '

9.42 m

h .0 l'l as.CD

• • O'Y n9 h ,

-l/ 'ff// l I H J g/ /'/ /-/' . C 63 LDW DPOSURE / //

l o axm ms -

5 o,

/.

7 w

s3 ( 11.0 y' i

~

x v , . jM x ao m ACCEfABLE OPERATION / / gx

$$ E n2

/ '

'. 3b

> h 10.0-U-

E - - - -

, , '/'

7.85  %

4@ CL no 12.5 tE$ / -

12.4 FUEL TYPE: BP8DRB284H,

[ l P8DRB284H AND 9X9 LFAs as par

- 80-MIL CHANNELS R 8.0 ----:----:-- -:---- ---- --- ---

6.28 0.0- - 5 .0--- 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0

$ AVERAGE PLANAR EXPOSURE (GWd/st)

• () - APLHGR LadTS FOR 9 X 9 LTAs

. m

Bases Table B 3.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS FOR HATCH-UNIT 2 #

Plant Parameters:

Core Thermal Power ..................... 2531 Mwt which corresponds to 105% of license core power

• Vessel Steam Output ..................... 10.96 x 10' lbm/h which corresponds to 105% of rated steam flow Vessel Steam Dome Pressure ............. 1055 psia Design Basis Recirculation Line Break Area For:

a. Large Breaks ................... 4.0, 2.4, 2.0, 2.1 and 1.0 ft8

b. Small Breaks ................... 1.0, 0.9, 0.4 and 0.07 ft*

Fuel Parameters:

PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL FUEL BUNDLE GENERATION RATC PEAKING POWER FUEL TYPE GEOMETRY (kW/ft) FACTOR RATIO Initial Core 8x8 13.4 1.4 1.18 A more detailed list of input to each model and its source is presented in subsection 6.3.3 of the FSAR.

For convenience, the APLHGR limits are reported in the units of kW/ft, which is the bundle planar power normalized to the number of fueled rods. Figure 3.2.1-2 shows that the 9x9 LFAs have the same planar power limits as the GE P80RB284H fuel; however, on a kW/ft basis, the APLHGR limits for the LFAs are 62/79 times the P80RB284H 1.iits.

• This power level meets the Appendix K requirement of 102%. The core heatup calculation assumes a bundle power consistent with operation of the highest powered red at 102% of its Technical Specification linear heat generation rate limit.

1. These are the initial core input parameters. For the updated Loss-of-Coolant Accident Analysis using SAFER /GESTR-LOCA, see Reference 4.

HATCH - UNIT 2 B 3/4 2-2 Proposed TS/0445q/223-87