ML19350B963
| ML19350B963 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 03/18/1981 |
| From: | Crouse R TOLEDO EDISON CO. |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| 693, NUDOCS 8103240501 | |
| Download: ML19350B963 (14) | |
Text
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TOLEDO EDISON Docket No. 50-346 j,",, cme
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I.icense No. NPF3
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Serial No. 693 4
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issca gj Director of Nuclear Reactor Regulation g,
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Attention:
Mr. Robert W. Reid, Chief S.
y,/p' Operating Reactors Branch No. 4 N/j Division of Operating Reactors United States Nuclear Regulatory Conunission Washington, D. C. 20555
Dear Mr. Reid:
This is in response to your letter dated January 14, 1981, requesting information on the accident induced neutron flux error issue. Enclosed, please find the Toledo Edison response to your request as it relates to Davis-Besse Nuclear Power Station Unit 1.
Yours very truly, k kMrL-cc:
NRC Davis-Besse 1 Resident Inspector pp/bb la II THE TOLEDO EDISCN COMPANY EDISON PLA7A 300 MADISON AVENUE TOLEDO. DHIO 436b2
'8103240501 f
Docket No. 50-346 License No. NPF-3 Serial No. 693 March 18, 1981 Response to the NRC Request for Information on Accident Induced Neutron Flux Errors Item 1 Determine if the high flux trip setpoint for your plant is affected by the accident-induced neutron flux errors discussed above. Provide us with information establishing that your present accident and transient analyses are valid and that the present Technical Specification limits provide as a minimum the original protective margin derived from the safety analyses; if not, provide the following information:
a.
Confirm that only the two non 205FA plant concerns discussed in the B&W letter of October 29 affect your plant; namely, small overcooling events including a small steam line break, and a rod ejection accident.
Response
3 mall overcooling events (such as a stuck open main feedwater control valve or a bypassed high pressure feedwater heater) and a rod ejection accident will affect the Reactor Protection System (RPS) high flux trip setpoint in light of the accident-induced neutron flux errors at Davis-Besse Nuclear Power' Station Unit 1 (DB-1). The small steam line break accident was not considered in the licensing basis for DB-1.
The effect of this accident, however, is similar to the effects of the overcooling events mentioned above.
Item I b.
Provide the effects of the error on your plant, supported by-appropriate analyses.
Response
Following is a description of the effects of the accident induced neutron flux error on the DB-1 FSAR accident analyses.
Overcooling Accident Analyses The current flux error
- assumptions in the FSAR are:
2.0% Heat Balance.
2.0% Steady State Neutron Measurement-2.0% Transient-Induced Neutron Measurement 0.5% Instrumentation 6.5% Total
- -The errorsnlisted in this letter are % full ~ power i
s Docket No. 50-346 License No. NPF-3 Serial No. 693 March 18, 1981 j
The 2.0% transient induced neutron measurement error is sufficient to accommodate all transient induced errors excluding those which are the subject of this letter. The additional transient induced error was discovered during a study for the WPPSS - WNP 1/4 FSAR, a 205 plant.
Although no transient analysis has been performed specifically for Davis-Besse 1, an engineering evaluation based on the WPPSS analysis has shown that the maximum transient induced error for moderate frequency overcooling transients will be approximately 13% (also see Attachment 1).
Therefore, for moderate frequency overcooling transients only, the instrumentation error that should be considered is:
2.0% Heat Balance 2.0% Steady State Neutron Measurement 0 to 13% Transient Neutron Measurement Dependent on Coolant Temperatures 0.5% Instrumentation 17.5% Total To justify full power operation, one must demonstrate that operation up to 123% full power is acceptable during these overcooling transients.
This power level is based on a high flux trip setpoint of 105.5% full power plus a total error of 17.5%.
It should be remembered that this power level could only be reached during certain overcooling transients that provided specific core conditions.
The analysis of induced flux errors during overcooling transients led to the quantification of the ratios of indicated power to actual core power as a function of downconer fluid temperature and core average coolant temperature. The primary concern is to determine the conditions that would permit the actual core power to exceed 112% full power without a reactor trip occurring. The error calculations were used to determine 4
the maximum actual core power as a function of temperature for the case where the indicated power would be 105.5% full power (Figure 1) which is
-the high flux trip setpoint. A series of heat balance calculations were then performed, using'the minimum licensed RCS flowrate-(387,200 GPM),
to determine the corresponding core operating conditions. When the heat balance is superimposed on the curves of Figure 1, the result is' a single line, for any given pressure, which defines the actual ~ core power as a function of coolant inlet and average temperature,' consistent with the assumed l constant indicated power. level of 105.5% full power. This line.islshown (dashed) on Figure 1 plotted against downcomer temperature and again on Figure-2 plotted against Tavg.
It can be seen by examination
- of either Figure 1 or Figure 2, that core operation under conditions
.below or to.the left of the power vs temperature line would be less restrictive (lower power _and temperature). Also, operation above and to the'right of this line:would be prevented because the_ indicated power level would be. greater than.105.5% full power, thus resulting in a reactor. trip.
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Docket No. 50-346 i
License No. NPF-3 Serial No. 693 March 18, 1981 In order to quantify core thermal margin for the conditions corres-ponding to-operation at an indicated power level of 105.5% full power, DNBR calculations were performed using the CHATA and TEMP codes. Minimum DNBR.rd power vs average temperature are plotted on Figure 2 for 2200 and 2000 PSIA. These pressures correspond to the nominal operating pressure and the lowest pressure allowed by the RPS low pressure trip function. At all points, corresponding to operation with indicated power equal to 105.5% full power, the minimum DNBR is well above the i
1.30 B&W-2 correlation design limit.
At 2000 PSIA, the variable low pressure trip function provides a trip when outlet temperatures exceed 606.7 F.
However, the high flux trip at 105.5% full power is more restrictive than the pressure-temperature trip above approximately 110% real power, as shown by Figure 3.
Any operation at the right of these limits as plotted is prevented by the Reactor Protection System. In order to quantify DNBR margin along this line, a parametric study was performed to determine the effect of coolant temper-ature variation on DNBR at 2000 PSIA and constant power. Figure 2 showed that the 2000 PSIA case is more restrictive than the 2200 PSIA case. These calculations showed that a line corresponding to a constant DNBR of 1.30 ' falls outside of the range of operation permitted by the high flux trip function for power in excess of 150% full power as shown on Figure 3.
In addition, a constant DNBR line corresponding to a 1.433 DNBR also falls' outside the range of permitted operation for actual core power levels up to 150% full power, as shown on Figure 4.
The calculations described.above are based upon reference design peaking conditions (1.714 radial x. local peak and a 1.5 core midplane axial peak)..The applicability of the design peaking to' core power shapes is ensured by the use of Maximum Allowable Peaking (MAP) curves. The MAP limits define the maximum peak allowed over a range of core elevations and axial peaks. These maximum allowable peak combinations represent radial and axial peaking for which the calculated minimum DNBR is the
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same as for the design peaking conditions. The MAP limits are used in the maneuvering' analyses as an acceptance criteria for the development of power imbalance limits (normal operating or RPS limits). Three-dimensional power distribution calculations were performed to assess the core power distribution perturbation and to determine the margins to centerline fuel. melt (CFM) and departure from nucleate boiling (DNB) limits. Identical calculations were generated from normal steady-state operation at 100% full power and from operation at 125% full power with ~
a 16 F inlet temperature reduction. All power distribution calculations i
were initiated from within or near the normal rod index, APSR, and axial.
' mbalance limitsfoi operation, such thatithe core behavior over the i
entire allowable operating range was examined.
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Docket No. 50-346 License No. NPF-3 Serial No. 693 March 18, 1981 The largest BOC peaking increase (4%) between 100% full power to 125%
full power was calculated with regulating rod index and APSR positions of 300 and 12.6 respectively; at MOC and EOC, the largest peaking increase (9.3%) occurred with rod index and APSR positions of 290.3 and 22.3 respectively. However, peaking decreases with increasing burnup such that the maximum core total peak decreased from 1.76 at BOC to 1.67 at EOC.
CFM and DNB margins were computed to determine if core safety limits would be violated during an overcooling transient. Since all calcu-lations were performed from near steady-state conditions, peaking factors of 5% and 7.36% were included in the 125% full power margins calculations to account for potential peaking increases due to transient xenon and quadrant tilt, respectively. Maximum allowable peaking curves for Davis-Besse 1, cycle 2, based on a 2772 MWT power level and the BAW-2 critical heat flux correlation were used to evaluate the DNB margins.
The applicability of these 112% full power MAP curves at 125% full power was verified by DNBR analyses performed for the latter power level at conditions corresponding to operation with an indicated power of 105.5%
full power (Tin = 535 from Figure 1).
Following conclusions were drawn from the above analysis for overcooling transients:
The minimum margins occurred at BOC-2 as expected, due to higher peaking early in the cycle. Minimum calculated CFM and DNB margins are shown in Table 1.
In addition, for the cases in the Table, peak fuel assembly power distributions were analyzed directly to verify the DNB margins.
The results show DNB peaking margin in excess of 6%.
The smallest DNB margin found was for a rod index of 300 although the power distribution with a rod index of 280.6 yielded a higher total peak. This result arises from the shift in axial imbalance to the bottom half of the core when control rods are inserted. This action lowers the local linear heat rate in the top half of the core where the coolant is hottest and DNB margin increases even though the maximum peak is higher.
It is therefore concluded that the induced flux measurement error does not compromise the safe operation of DB-1 during overcooling events initiated from anywhere within the operating range of rod index. The net result is that all cases show acceptable margins.
Control Rod Ejection Analysis Past analyses of ejected control rod used larger ejected control rod worths and a high flux trip was reached without violating acceptance criterion using the overly conservative adiabatic heat up assumption.
4 Dockst No. 50-346 License No. NPF-3 Serial No. 693 March 18, 1981 The concern on the rod ejection transient is that the high flux trip may not be activated for an ejected rod of small worth (less than.2% Ak/k).
Current models used for FSAR analysis do not consider energy transfer from the local pin to the reactor coolant, therefore, if used for analysis of the small worth rod ejection accident could show unacceptable results (peak fuel enthalpy greater than 280 cal /gm).
For feed and bleed operation of a B&W core, the highest worth ejected rod at power is extremely low (.1%Ak/k). This is because the trans-ient bank of rods (Group 7) is normally operated 80% withdrawn from the core. Therefore, for an evaluation of this low worth rod ejection accident, the most important parameters for consideration are the sub-channel rate of power increase and the total local power generation.
The combination of these two parameters will determine the heat transfer and therefore, fuel-rod enthalpy. Contrary to the ejected rod analysis in the FSAR, where the local peaking factor doubles or triples in less than.3 seconds, the small rod ejection causes a very small local peaking change. This results because the power is depressed to the bottom of the core in the ejected-rod location before the rod ejection. After rod ejection, the axial flux shape actually flattens resulting in mini-mal percentage changes in both radial and total power peaking.
For the average-channel power response, the small ejected rod accident without trip on high flux provides a comparison with the large ejected rod analyzed in the FSAR. First, the peak power is considerably lower than the FSAR ejected-rod case and secondly, the local power rate of change is relatively small.
An illustration for the above justification cons'.ders the power rate of change for two ejected rod cases (.2 and.45% akfk) and the resulting energy balance as a function of the time the pow er is above 100% full power. For a.2% A k/k ejected rod case, the pea't power is 135% full power and decreases to 112% full power in 11 seconds whereas for a.45%
ok/k ejected rod the peak power is 245% full powe r at.15 seconds and decreases to 100% full power at 1.4 seconds. Therefore, the total stored energy in the fuel pin is the difference between thermal power generation and heat removal rate.
Since the power response after a small rod-ejection is slow, and the average-channel power is relatively low with an insignificant total ~ peaking change, the heat transferred out of the pin to the reactor coolant within the 11 seconds should be sig-nificant and therefore, the peak pin fuel enthalpy will be less than 280 cal /gms.
Using the above observation and maintaining normal RCS flowrate, it is conservative to compare che amount of power' excursion between small worths and FSAR ejected-rod transients. The existing data iadicates that the stored energy should be between 200 to 250 cal /gm which is highly conservative yet still below the limiting value of 280 cal /gm.
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Dockst No. 50-345 License No. NPF-3 Serial No. 693 March 18, 1981 Item 1 c.
Provide your program and schedule for mitigating the effects of the error.
Response
As discussed above (see pages 1 and 2 of this letter), the transient (accident induced) neutron measurement error could be as high as 13%
depending on coolant temperatures. At DB-1, the operating procedure for calibration of the nuclear instrumentation (NI's) call for always calibrating the NI's above the heat balance value. The conservatisms associated in the DB-1 calibration procedures have increased the allow-able transient neutron measurement error from 2% (FSAR assumption) to 4%
and have reduced the steady state error from 2% (FSAR assumption) to 0%.
Thus, the over-cooling events of concern for DB-1 here will have addi-tional effect of 9% transient error (as compared to 13% or, an addi-tional error of 11% as noted in the analysis above). The effects of the error as analyzed above, therefore possess an added conservatism.
To mitigate the above effects of the accident induced neutron flux errors, plant procedures have been modified at DB-1 to eliminate regu-lating rod operation in the " operation restricted" zone. The " operation restricted" zone has been conglomerated into the " unacceptable operation" zone. The operation of these regulating rods in the " unacceptable operation" and " operation restricted" zone is already limited by DB-1
' Technical Specifications (see Section 3.1.3.6).
Item 2 Provide justification for continued full power operation of your plant until your program to mitigate effects of the error is completed.
Response
The program to mitigate the effects of the error as specified in the response to item 1-c above has already been implemented at Davis-Besse.
In light of the response to item 1-b above,' continued full power operation at Davis-Besse is therefore justified.
pp bb/1-5.
Docket No. 50-346 License No. EPF-3 ATTACHMENT 1 Serial No. 693 hbrch 18,1961
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EVALUATION OF' INDUCED NI ERROR DURING OVERC00 LING EVENTS Since the concern of induced neutron flux error during overcooling events was raised, the important question has been to quantify the magnitude of the induced error and define the transient (s) that result in such an error.
To this end, a task was undertaken to review current available data, determining the bounding moderate frequency overcooling transient and estimating the resulting induced neutron flux error for that case.
The most complete data available to evaluate was from WPPSS 205 FSAR analysis.
Based on the WPPSS analysis, the most severe moderate frequency overcooling transient which is terminated by a high flux trip signal is a failure of the turbine bypass system (atrospheric on/off valves). This failure consists of a single atmospheric on/off valve on each of the four steam lines opening at 6
power. The induced " break" results in an increased steam flow of : 4.6 x 10 lbs/hr..The real flux (thermal pcwer) increases to : 122% FP with an induce flux error of : 13% FP.
Since the indicated power'dnes not exceed 1h5.5% FP, s,_
the high flux trip will not terminate the transient. This transient causes 0
a reduction in the cold, leg or downcomer temp. by : 16 F, a reduction in core average tempera'ture by : 100F and very little change in RCS pressure.
This transient has been analyzed on Power Train IV with noderator coefficients forLEOC and E0C covering a range of 0.0 x 10-4A k/K/F to -2.37 x 10-4a k/k/F and with varying feedwater temperatures.
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.1 The change in ' flux at' the out-of-core detector 'locaticos due to changes in cold
-leg temperature.were determined using the ANISN;transpcrt theory program. -The induced neutron flux error was obtained by using the Power Train in/ output (real power, Thot Tave)- and the results : of thel ANISM analysis.
i Dockst No. 50-346 Licente No. NPF-3 Serial No. 693 Ebech 18, 19&1 Since the Atmospheric Valve failure is considered to be the most probable and restrictive overcooling transient of concern, it is realistic to believe the 13% FP induced neutron flux error is valid for t'he smal1 overcooling, moderate frequency event of interest.
Using the WPPSS Atmospheric Valve failure for determination of 177 FA plants induced neutron flux error results, involves the following assumptions:
- 1) The specific transient response for 177 FA plants including the event timing is similar to WPPSS 205 FA plant.
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- 2) The failure of all bypass valves on'177 FA plants is similar to the failure of the Atmospheric valves event on WPPSS.
- 3) The magnitude of the. induced neutron flux error for 177 FA plants should be similar or conservative because:
The total bypass flow in 177 FA plants is similar or less than the. 205 plant. Also a specific single failure j
can be identified to cause all TBV's to-open.
~.-a, The failure of a single steam generator bypass valve can also occur in a 177 FA plant; however, the resulting stea,m flow increase will be 50 to 70% less than the WPPSS steam flow increase.
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Docket No. 50-346 License No. NPF-3 Serial No. 693 March 18, 1981 p
ACTUAL CORE POWER VS.
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Docket No. 50-346 Lfcence No. NPF-3 Serial No. 693 3
March 18, 1981
,j FIGURE 3 RPS AND ONBR LIMITS 1
P3yg = 2000 PSIA J
J 160 t
OPERATION 150 FORBIDDEN
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ASSUMING POWER E
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Docket No. 50-346 License No. IRF-3 Serial No. 691
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March 18, 1981 4
FIGURE 4 3
RPS AND ONBR LIMITS J
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150 OPERATION IN THIS ONBR = 1.433 (TECH SPEC LIMIT)
REGION ASSUMES
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C OPERATION FORBl00EN 130 BY RPS m
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Dockat No. 50-346 License No. NPF-3 Serial No. 693 March 18, 1981 Taule 1 p i DAVIS BESSE-1, CYCLE 2
!i MINIMUM CFM, DNB PEAKING MARGINS l1 4-EFPO - 1255 FP F; ;
(MOST LIMITING CON 0lil0NS) d 1
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R00 INDEX APSR POSITION CFM MARGIN DNB MARGINS (sw0)
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