ML19340E440
| ML19340E440 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 01/05/1981 |
| From: | Clayton F ALABAMA POWER CO. |
| To: | Varga S Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8101140507 | |
| Download: ML19340E440 (9) | |
Text
Atb;ma Power Company 600 North I'th Street Post Offica Box 2641 Birmingham Alabama 35291 Telephone 205 250-1000 m
F. L. CLAYTON, JR.
Senior Vice President Alabama Power the souther's electrc system January 5, 1981 Docket No. 50-348 Director of Nuclear Reactor Regulation i
U. S. Nuclear Regulatory Commission Phillips Building, Room 316 7920 Norfolk Avenue Bethesda, Maryland 20014 Attention:
Mr. S. A. Varga JOSEPH M. FARLEY NUCLEAR PLANT - UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING R00 B0W PENALTY Gentlemen:
Enclosed is our response to your request of December 12,1980, for additional infomation regarding the J. M. Farley Nuclear Plant - Unit 1 rod bow penalty. Should you have any further questions, please do not hesitate to call.
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Yours very tr y,
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(% i AF. L. Clayton, Jr.
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Enclosures cc: Mr. R. A. Thomas Mr. G. F. Trowbridge Mr. E. A. Reeves Mr. W. H. Bradford 7
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ENCLOSURE RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION (DATED 12/12/80, S. A. VARGA TO F. L. CLAYTON)
Question 490.1 Describe what margins were used to offset the reduction in DNBR due to fuel rod bowing. Has NRC approved these margins for Farley applications? If not, please provide the necessary justification.
Response
The following margins were used to offset the reduction in DNBR due to rod bow penalty.
9.1% - Generic margin in the DNBR analysis for 17 x 17 plants.
l This margin was used and approved by the NRC on the J. M. Farley Nuclear Plant-Unit 1 Technical Specifications Amendment 8.
This margin is composed of the following components:
DNB Maroin, %
Design Limit DNBR of 1.28 vs.1.30 1.6 DNB Correlation Multiplier, 0.865 vs. 0.88 1.7 Thermal Diffusion Coefficient, 0.038 vs. 0.051 1.2 Pitch Reduction 1.7 Eight vs. Seven Grids
2.9 Total
9.1 %
3% - Margin in the difference between the 1.3 DNBR safety limit and the minimum DNBR calculation for the Complete Loss of Flow Accident. This margin was approved by the NRC for the J. M. Farley Nuclear Plant-Unit 2 Technical Specifications.
Both units are identical with respect to thermal hydraulic analysis.
Question 490.2 Are the margins used to offset rod bov*ng DNBR penalties employed solely for this purpose? If not, please prsvide justification for using these margins more than once.
Response
Th@ margins shown in response to Question Q490.1 are solely used to offset the DNBR rod bow penalty. However, at high burnup these margins will not be completely sufficient to offset the DNBR rod bow penalty. Therefore, the penalty shown in proposed Figure 3.2-3 is used to compensate for this difference at high burnup.
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ENCLOSURE RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION Page 2 Question 490.3 The reduction in DNBR due to rod bowing for the loss-of-flow tranrient and N-1 loop operation is greater than the reduction in DNBR for other conditions.
Is additional available margin used, or is the calculated limiting DNBR sufficiently large to offset this incremental penalty?
Response
No additional margin is used to offset the penalty in the analysis. The calculated limiting DNBR is sufficiently large to offset the incremental penalty.
Question 490.4 State the reasons or justification for discontinuing rod bov:ing penalty calculations at 33,000 MWD /MTV.
Response
It is our position that by the time the fuel attained the burnup of i
r l-33,000 MWD /MTV, it is not capable of achieving the limiting peak factor, F9H, due to the decrease in fissionable isotopes and buildup of fission product.
Question 490.5 Amend the basis of the technical specifications to identify each generic or plant-specific margin that has been used to offset the reduction in DNBR due to rod bowing. Also reference either the source or approval of each generic margin.
Response
The attached proposed draft technical specifications have been revised to include the information requested.
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SAFETY LIMITS BASES N
The curves are based on an enthalpy hot channel factor, F and a reference cosine with a peak of 1.55 for axial power shak., of 1.55 An N
allowance is included for an increase in F at. reduced power based on.
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These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the taximud allowable' control red insertion assuming the axial power imbalance is within the limits of the ff (AI) function of the Overtemperature trip. When the 3.__
axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature AT trips will reduce the setpoints
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to provide protection consistent with core safety limits.
2.1.2" REACTOR COOLANT SYSTEM PRESSURE
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The restriction of this Safety Limit protects the integrity. of the Reactor Coolant System from overpressurization and thereby prevents.the release of radionuclides contained in the reactor coolant from reaching the containment a1:mosphere.
The reactor pressure vessel pressurizer and the reactor coolant.
system piping and fittings are designed to Section III of the ASME Code.
for Nuclear Power Plant which pemits a inaximum transient pressure of 110% (.2735 psig) of design pressure.
The Safety Limit of 2735 psig is therefore consistent with the design criteria and associate'd code requirements.
The entire Reactor Coolant System is hydrotested at 3107 psig, l'25%
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of design pressure, to demonstrate integrity prior to initial operation.
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FARLEY - UNIT 1 B 2-2 Amendment No. 8
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POWERDISTRIBUTIONLIMITS
{V BASES 3/4.2.2and3/4'.2.3HEATFLUXANDNUCLEARENYHALPY'HOTCHANNELandPADIAL PEAKING FACTORS-F(Z),FfgandFxy (z) g The limits on heat flux and nuclear entialpy hot channel factor's ensure that 1) the design limits on peak. local power density and minimum
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DNBR are not exceeded and *2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F'ECCS acceptance criteria limit.
Each of these het channel factors are measureable but will non: ally only be determined periodically as specified in Specifications.4.2.2 and
- 4. 2. 3.
This periedic surveillance is sufficient to ensure that the hot channel factor limits are maintained provided:
- a. ' Control rods in a bank move together with no individual rod insertion differing by more than + 12 steps from the group demand position.
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b.
Control rod banks are sequenced with overlapping banks as described in Specification 3.1.3.5.
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The control rod insertion limits of Specifications 3.1.3.5 c.
and 3.1.3.6 are maintained
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d.
The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.
l igtheradialpowershap)He for all permissible red insertion lim The relaxation in F as a function of THERMAL F0WER allows chan F
wi ab$ve,.ll be maintained within its limits provided conditions a through d are maintained.
5 When an F measurement is taken, both experimental error and man-
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ufacturing tolhrance must be allowed for.
5 is the appropriate' allowance i
for a full core map taken with the incore detector flux mapping system and 3% is the appropriate allowance for manu.facturing tolerance.
N When F is measured, experimental, error must be allowed for and 4 H
is the appropriate allowance for a full core map taken.with the incere N
detection system.
The specified limit for F also cen'tains an 8 allow-aqce for uncertainties which mean that normakHoperation will result in FTtT$ns<_:l.55/1. 08.
The 8% allowance is based on the following considera-(!,
FARLEY - UNIT 1 8 3/4 2-4 Amendment No.10 T/
POWER DISTRIBUTION LIMITS g.
BASES Abnormal perturbations in ghe radial power shape, such as-from a.
rod misalignment, effect Fg more directly than F,
g b.
Although rod movement has a direct influence upon limiting F
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to with n its limit, such control is not readily available to limitF][H,and
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Errors in ~predi.ction for. control p'ower shape ' detected during
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startup physics tests can b~ compensated foi in F by ing axial flux distributions. -This compensation Ear Fh[estrict..
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Inse.rt The radial peaking factor, Fxy (z), is measured periodically to provide additional assurance that the hot channel factor, Fg (z), remains within' its Timit.
The Fxy (z) limits were determined frcm expected power control' '
maneuvers over the full range of burnup conditions iii the core.
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3/4.2.4 QUADRANT p0WER TILT RATIO l.
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..w-t g The quardrant power til.t ratio limit assures that the radial power distribution satisfies the dcsign values used in the power capability analysis.
Radial power distribution measurements are made during start-up testing and periodically during power operation.
The limit of.l.02 at which corrective action is required provides DNB and linear heat generation rate protection with x-y plane power tilts. A limiting tilt of 1.025 can be tolerated before the margin for l
uncertainty in F is depleted. The limit of 1.02 was selected to provide g
an allowance for the uncertainty associated with the. indicated power' tilt.
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.The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned' rod.
In the event such actien does not correct the tilt, the margin for uncertainty on F is reinstated by reducing the power by 3 percent for each percent of ti19 in excess of 1.0.
l B 3/4 2-5 Amendment No.10 FARLEY - UNIT 1 l
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s, Insert to Pace B 3/4 2-5 Fuel rod bowing reduces the value of DNB r,atio. Sufficient credit is available to offset this reduction. This credit comes,
from generic design margins totaling 9.1% and 3% margin in the difference between the 1.3 DNBR safety limit and the minimum DNBR calculated for the Complete Loss of Flow event. The penalties applied to Fh to account for Rod Bow (Figure 3.2-3)
as a function of burnup are consistent with those described in
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Mr. John F. Stolz's (NRC) letter to T. M. Anderson (Westinghouse) dated April 5,1979, ; ia WCAP-8691, Rev.1 (partial rod bow test,
data).
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