ML19262A184
| ML19262A184 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 03/29/1974 |
| From: | John Miller METROPOLITAN EDISON CO. |
| To: | Anthony Giambusso US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 7910260511 | |
| Download: ML19262A184 (9) | |
Text
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AEC DIS
.BUTION FOR PART 50 DCCKET MAH
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FROM:
DATE OF DOC DATE FEC'D LTR MEMO PST O-~.-2 R Metropolitan Edison Company Reading, Penn. 19603 Mr. J.G. Miller 3-29-74 4-1-74 X
TO:
ORIG CC OTHER SENT AEC PDR XXX SENT LOCAL POR XXX A. Giambusso 1 siened CLASS UNCLASS PROP INFO INPUT NO CYS REC'D DOCKET NO:
XXX 1
50-289 DESCRIPTION:
ENCLOSURES:
Ltr adv of a change in the analysis concerning Plux/ Flow Trip Delay Time Increase.
the delay time of the flux / flow trip....
trans the following....
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@ -..$YS METROPOLITAN EDISON COMPANY POGT OFFICE EOX 542 READING,PENNSYLVANI A 19603 TELEPHONE 215 - 929 4601 March 20, lo7h g.
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Dear Mr. Giambusso:
SUBJECT:
THREE MILE ISLAND NUCLEAR STATION UNIT 1 DOCKET NO. 50-289 FLUX / FLOW TRIP DELAY TIME INCREASE Our NSS supplier, Babcock & Wilcox, has informed us of a change in our analysis as stated in the Three Mile Island Unit 1 FSAR. This change in the analysis is that the delay time of the flux / flow trip in the Faactor Protection System has proven to be 1.4 seconds rather than the original assumption of 0.65 seconds.
Attached is an analysis which shows that this increase in trip delay time has not infringed upon plant safety.
Very truly yours,
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J. G.# jMiller 4
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c:gtu Attachment
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Flux /Flev Trip Delay Time Increase I.
3ACKGROUND Reactor coolant flow measurements are obtained by pressure drop taps installed in the hot leg of all 177 fuel assembly plants.
Oconee I test data have shown that the average a? measurement produces an accurate indication of the reactor coolant flow. However, the "As-Built" sensing string has a measured time constant of 1.h seconds versus the 0.65 seconds assumed in the FSAR.
It is the purpose of this report to demonstrate that the calculated value of the flux / flow trip ratio is still conservative and adequate.
II.
THERMAL-HYDRAULIC METHODS To determine the flux /flcw trip setpoint that is necessary to meet the hot-channel DNB ratio criteria, several calculational steps are required. These steps involve such things as the determination of steady-state operating conditicns, fuel densification effects and transient calculations.
A.
Thermal-Hydraulic Conditions Durine Normal Oteration The hot channel thermal hydraulic conditions are calculated for design conditions at 10E% of rated power. The power level of 108% includes operation at 102% of rated pcwer plus a maximum pcwer level measurement error of o%
(h5 neutron flux error and 2% heat balance error). This serves as the bench-
= ark calculation frc= which the densification penalty and the transient effects can be determined. The steady state analysis is performed using the TDIP computer code (BA'4-10021) with the appropriate hot channel factors, coolant inlet te=perature and system pressure errors, and a 5% hot assembly flow =aldistribution factor applied.
"'hese censervatisms are consistent with the calculational techniques employed in the FSAR analyses. The 0
design flow rate of 131.32 x 10 #/hr was used. The het assembly power distribution consisted of a 1.78 radial-local nuclear factor (F AH) with a 1 5 cosine axial flux chape. The primary cutput of the calculation is the minimum hot channel DN3 ratio as calculated by either tl.e '4-3 or the EA'4-2 ( 3A'4-10000 )
correlations.
1482 175
O 3.
Densificatien Effects The fuel densification penalty applied to the het channel was determined by the methods discussed in the Cconee I Fuel Densification Report, 3AW-1395, June 1973, page A-5.
A conservative slumped and spiked 1.83 outlet peaked axial pcVer chape was used in cenju: ction with a 1.k9 : adial-local factor to deterdne the maximum fuel densification effect en DITE ratic. Tnis reduced hot channel DN3 ratio is the basis for establishing the initial conditions for the transient calculations.
C.
Transient Hot Channel Conditions During a Less of Flcw Tha flux / flow trip setpoint is derived to protect the core during a one puap coastdcwn. A one pump coastdown is analyced because redundant pump
=cnitors are provided which will provide D:i3 protection for all other pump coastdowns including coastdowns while the plant is in partial pump operation.
The pump =cnitor logic will not cause a reactor trip for the 1 css of ca2 pum; frc= four pump operation.
The thermal-hydraulic response of the hot channel is calculated by RADAR cc=puter code (3AW-10069)('). The initial hot channel DN3 ratio is set equal to the steady state value with densification effects included. The RADAR cutput in the form of Ect Channel DN3 ratic versus time is the basis for establishing the flux /flev ratio trip setpcint.
II
I. PROCEDURE
FOR DETERMI:lI:iG FLUX / FLOW SETPOI:IT The determination of the flux / flow setpoint is acec=plished in four basic steps.
The result of these steps is designed to yield a value of the flux /ficw ratic that will prevent the mini =un hot channel D ER frc= going below 1.3 (W-3) or 1.32 (BAW-2) for the coastdown for which protection is required.
These steps are as followc :
A.
Total Time Determination From a plot of minimum D:ER versus time find the time that yields a D:ER of 1.3 for the naximum pcwer level (10S%) for the maximum number of pumps lost for which the flux /ficw trip =ust provide protection (one pump for TMI-1, although the original Technical Specifications were based en a two pump coastdown since redundant pump =cnitors were installed subsequent to the original calculaticns. )
1482 176
B.
Cc,astinz Time Determination The total time to reach a DNER of 1.3 minus a conservative value of the total trip delay time gives the maximum allevable coasting time prior to trip initiation.
C.
Mini =:n Flew Determination Frc= a plot of ficv versus time for the coastdown of interest, the percent flow for the =aximum allowable coasting time is found. This yields the ficw at which trip must be initiated.
D.
Flux / Flow Ratic Calcul1tien The naximum allowable flux / flow ratio is the maximum real power level of interest (108%) minus the power level measurement error (65) divided by the =ini=um ficv.
IV.
CALCULATIONAL RESULTS Figure 1 shows the flow versus time that is the design basis for the determination of the flux / flow latio.
Figure 2 shows the calculated DNER versus time with the effects of densification included. From this figure it is seen that a DNBR cf 1.3 is reached at about 3.35 seconds. Using figure 1 and the technique explained previcusly, this yields a flux /flcw ratio of 1.08.
This is the value presented in the FSAR Technical Specificaticns for densified fuel. Figure 3 shows DNER versus time for a TMI-l one pump coastdown using B&W's standard analysis methods thich includes the use of the EAW-2 corralation.
Frc= Figure 3 using the previous method and a trip delay time of 1.h0 seccnds rather than 0.65 seconds, it is seen that a flux /ficv ratio of 1.12 is cbtained.
V.
CONCLUSIONS Due to differences in the assumed and "As-Built sensing strings, the trip delay time for TMI-1 has been changed frc= 0.65 seccnds to about 1.h secends. The analysis presented in this report has de=cnstrated that when using 3&W's standard techniques including the BAW-2 correlation, the technical specification value of the flux / flow ratio of 1.08 is concervative even with the delay time increased frc= 0.65 seconds te 1.ho seconds. Since the 1.08 value is conservative, the te chnic a' specification value does not need to be changed and the analyses in the F3AR and Fuel Censification Report where the flux / flow ratic is used, do not need to be revised since the results are conservative.
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REFERENCES (1) BAW Topical Report BAW-10021, "TDiP-Thermal Enthalpy Mixing Program,"
April 1970.
(2) B&W Topical Report BAW-10000, " Correlation of Critical Heat Flux in a Bundle Cooled by Prec3urized Water,"
March 1970.
(3) B&W Topical Report BAW-10069, " RADAR-Reactor Thermal and Hydraulic Analysic During Reactor Flow Coastdown," July 1973.
1482 178
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