ML20003G241

Smud Task 170 - N.I. Calibr Error,Final Rept

ML20003G241
Person / Time
Site:	Rancho Seco
Issue date:	03/31/1981
From:	BABCOCK & WILCOX CO.
To:
Shared Package
ML20003G240	List: ML20003G239 ML20003G241
References
12-1125041, TAC-43264, NUDOCS 8104280520
Download: ML20003G241 (15)

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Sacramento Municioal Utility District Task 170 - N. I. Calibration Error Final Report March 1981 Document Ntsnber 12-1125041 l

l

INTRODUCTION A concern raised during certain overcoolino transients and rod ejection transients is that a transient-induced neutron flux error greater than that assumed in the FSAR could exist. This concern was presented to the Owners in a group meeting in Lynchburg on October 23 and 24,1980. The purpose of Master Services Task 170 - NI Calibration Error is to identify plant safety margins available to offset the additional NI errors, thereby providing justification for full power operation of Rancho Seco. Cycle 5.

SUf*ARY AND CONCLUSIONS A concern was raised that overcooling events could result in an induced neutron flux measurement error which might increase as reactor coolant temperature decreases. This error could result in the actual core power level exceeding the assumed 112% overpower condition before a reactor trip occurs. However, an evaluation has shown that the increase in core power is offset by the beneficial effect of the temperature decrease on core thermal marains. DNBR analyses performed for the most limiting condition (indicated power at high flux trip limit of 105.5%, RC pressure at low pressure trip limit of 1900 psig) demonstrate that the minimum DNBR will be greater than 1.30 for conditions under which a reactor trip would be initiated at core power levels up to 135% of rated power.

In addition, cycle 5 calculated core power distributions at all allowable rod index and Axial Power Shaping Rod (ASPR) positions for normal full power operation were examined and margin existed for both DNB and CFM assuming an actual core power of 125%.

It is therefore concluded that the induced flux measurement error does not compromise the safe operation of Rancho Seco, Cycle 5 during overcooling events initiated from anywhere within the allowable operating range..

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%

k/k.

Under these conditions. Current models would show unacceptable results (peak fuel enthalpy > 280 cal / gram). Although no reanalysis has been performed, an enoineering evaluation of the conservatisms in the orioinal analysis (such as adiabatic heatuo) has led to the conclusion that a reanalysis using realistic assumptions, will show that the peak fuel enthalpy will not exceed 280 cal / oram. Therefore. this concern is not considered to compromise the safe operation of Rancho Seco, Cycle 5.

A further evaluation of the small worth ejected rod accident is provided in.

OVERC00 LING ACCIDENT ANALYSIS 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 l

0.5% Instrumentation 6.5% Total The 2.0% transient induced neutron measurement error is sufficient to l

accomodate all transient induced errors excluding those which are the topic of this report.

The additional transient induced error was discovered durino a study for the WPPSS - WP 1/4 FSAR. a 205 plant. Attachment 2 l

provides an evaluation of the induced NI error during certain overcooling events.

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Although no transient analysis has been performed specifically for Rancho Seco, an engineerino evaluation based on the WpPSS analysis has shown that the maximum transient induced error for moderate frecuency overcooling transients will be approximately 13%. Therefore, for moderate frecuency 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%

?nstrumentation 17.5%

Total To justify full power operation, one must demonstrate that operation uD 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 durino 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 downcomer fluid tenperature and core averaae coolant temoerature.

The crimary concern is to determine the conditions that would permit the actual core power to exceed 112% without a reactor trip i

occurring. The error calculations were used to determine the maximum I

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actual core power as a function of temperature for the case where the indicated power would be 105.5% which is the high flux trip setpoint (Figure 1). A series of heat balance calculations were then performed, using the minimurn licensed RCS flowrate (387, 710 GPM). to determine the corresponding core operating conditions. When the heat balance is superirnoosed on the curves of Fioure 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 constant indicated power level of 105.5%. This line is shown (dashed) on Figure 1 plotted against downcomer temperature.

It can be seen by examination of Figure 1 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 richt of this line would be prevented because the indicated power level would be greater than 105.5%.

thus resulting in a reactor trip.

In order to quantify core thermal margin for the conditions corresponding to operation at an indicated power level of 105.5%. DNBR calculations were performed using the CHATA and TEMP codes. All 2200 psia RCS pressure points allowed by the RPS and corresponding to operation with indicated power equal to 105.5%. lie well above the Tech Spec minimum DNBR of 1.30 for the B&W-2 correlation. For low pressures corresponding to the RPS low pressure setpoint (1900 psig), the variable low pressure trip provides a trio if the RCS outlet temperature exceeds 597.80 l

F.

This trip function then provides protection to a minimum DNBR of 1.3 for core power levels up to7123%FP(Figure 2).

In addition. the high flux trip provides protection to the minimum DNBR of 1.3 for power levels uo to 135% FP.

Any operation at the richt of these limits as plotted is prevented by the Reactor Protection System.

In order to quantify DNBR margin along this line. a carameter study was performed to determine the effect of coolant temperature variation on DNBR at 1900 psig and constant power

_a.

The calculations described above are based uoan reference design peaking conditions (1.714 radial x local peak and a 1.5 core mid-plane axial peak).

The applicability of the design peaking to core power shapes is ensured by the use of Maximum Allowable Peakina (MAP) curves The MAP limits define the maximum peak allowed over a rance of core elevations and axial peaks.

These maximum allowable peak combinations represent radial and axial peaking for which the calculated MDNBR is the sane as for the design peakinc conditions.

The MAP limits are used in the maneuvering analyses as an acceptance criterion for the development of power imbalance limits (normal operatino or RPS limits).

Three-dimensional power distribution calculations were perforned to assess the core power distribution perturbation at 125% FP due to an overcooling transient, and to determine the margins to centerline fuel melt (CFM) and departure from nucleate boilino (DNB) limits.

Identical calculations were cenerated from normal steady-state operation at 100% FP and from operation at 125% FP with a 160F inlet temperature reduction. All power distribution calculations were initiated from wi.nin or near the normal rod index, APSR and axial imbalance limits of operation, such that the core behavior over the entire allowable operating range was examined.

5-

CFM and DNB margins were comouted for the 125% FP cases to determine if core safety limits would be preserved during an overcooling transient.

Since all calculations were performed from near steady-state conditions.

appropriate peaking f actors were included in the 125% FP margins calculations to account for potential peaking increases due to transient xenon and cuadrant tilt. Maximum allowable peakinc curves for Rancho Seco.

Cycle 5. were used to evaluate the DNB margins. The applicability of these curves at 125% FP was verified by DNBR analyses performed for the limitino cases.

RESULTS The increased power resulting from certain overcoolina transients is accommodat J within the cycle 5 Normal Operating Limits.

The analysis of the desian peaking distribution yielded DNBR marcins in excess of 6.7% DNBR relative to the 1.30 DNBR limit. Additionally, analysis of the most limiting peakinc distribution identified by maneuverina analyses yielded DNBR peaking margins as shown in Table 1.

The high flux trip provides DNBR protection to the minimum DNBR limit up to core power levels of 135%FP.

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T,ABLE,1 1

RANCHO SECO. Cycle 5 Minimum CFM DNB Peaking Margins (Most Limiting Conditions)

CFM DNS Cycle Burnuo Rod Index APSR Index Margin

• Margin *

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(%)

(%) _

(

,_ EFPD) _

(

125 300 12 19.93 4.68 125 287 12 17.39 7.05 244 300 100 15.17 8.55 244 273 100

.54 10.23

{

• Margin is quoted in terms of relative peaking compared to the limit value.

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ATTACHMENT 1 Since the ejected-rod accident analyzed in the FSAR for.65% tik/k worth rod is terminated by a high flux trip, concern has been expressed that if no high flux trip occurred during ejection of rods with worths less than

.2% /i k/k, that fuel pin f ailure might occur. 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 > 280 cal /om).

For feed and bleed operaticn of a B&W core, the hiohest worth ejected rod at power is extremely low <.1% Ji k/k, this is because the transient bank of rods (Group 7) is normally operated > 80% wd from the core Therefore, for an evaluation of this low worth rod ejection accident, the most important parameters for consideration are the subchannel 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 peakino factor doubles or triples in less than.3 seconds, the snall rod ejection causes a very small local peaking chance. 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 resultino in minimal percentaoe chances in both radial and total power peakino.

For the average-channel power response, the snall e.iected rod accident without trip on high flux crovides a comparison with the large ejected rod analyzed in the FSAR. First, the peak power is considerably lower than the i

FSAR ejected-rod case and secondly, the local power rate of chance is relatively small.

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ATTACHMENT 1 An illustration for the above justification considers the power rate of change for two ejected rod cases.2 and.45%

k/k and the resulting energy balance as a function of the time the power is above 100% FP. For a

.2%

k/k ejected rod case. the peak power is 135% FP and decreases to 112%

FD in 11 seconds whereas for a.45%

k/k ejected rod the peak power is 245% FP at.15 seconds and decreases to 100% FP 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 insianificant total peaking change, the heat transferr!d out of the pin to the reactor coolant within the 11 seconds should be significant and therefore, the peak pin fuel enthalpy will be less than 280 cal / cms.

Using the above observation and maintaining normal RCS flowrate. it should be conservative to compare the amount of power excursion between small worths and FSAR ejected-rod transients.

The existina data indicates the stored energy should be between 200 to 250 cal /am which is hiohly conservative yet still below the limitina value of 280 cal /cm.

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,A_TTACHMENT 2 Since the concern of induced neutron flux error during overcooling events was raised. the important question has been to Quantify the macnitude 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 boundino moderate frecuency overcooling transient and estimating the resulting induced neutron flux error for that case.

The most comolete data available to evaluate was from WPSS 205 FSAR analysis. Based on the WPSS analysis, the most severe moderate frecuency overcooling transient which is terminated by a high flux trip signal is a failure of the turbine bypass system (atmospheric on/off valves). This f ailure consists of a single atmospheric on/off valve on each of the four steam lines openino at power. The induced " break" results in an increased steamflowof74.6x106 lbs./Hr. The real flux (thernal power) increases to $122% FP with an induced flux error of $13% FP.

Since the indicated power does not exceed 105.5% FP. the high flux trip will not terminate the transient. This transient causes a reduction in the cold leg l

or downcomer temp. by g 160F. a reduction in core averace temperature by

$100F and very little change in RCS pressure.

This transient has been analyzed on Dower Train IV with moderator coefficients for BOC and EOC covering a rance of 0.0 x 10 # 4 k/k/0F to -3.37 x 10 4 0

4 k/k/ F and with varying feedwater temperatures l

l The change in flux at the out-of-core detector locations due to changes in cold lea temperature were determined using the ANISN transoort theory program. The induced neutron flux error was obtained by using the Power Trah IV outout (real power Thot. Tave) and the results of the A<ISN analysis.

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ATTACHEENT2 Since the Atmospheric Valve f ailure 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 the small overcooling. moderate frecuency event of interest.

Usina the WPPSS Atmospheric Valve failure for determination of 177 FA plants induced neutron flex error results, involves the following ass umptions -

1) The specific transient response for 177 FA plants including the event timino is similar to WPPSS 205 FA plant

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 maanitude of the induced neutron flux error for 177 FA plants should be similar or conservative because o The total bypass flow in 177 FA plants is similar or less than the 205 plant. Also a specific sincie f ailure can be identified to cause all TBV's to open.

o The failure of a sincie steam generator bypass valve can also occur l

in a a 177 FA plant: however the resulting steam flow increase will be 50 to 70% less than the WPPSS steam flow increase.

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Figure 1.

ACTUAL CORE POWER VERSUS INLET TEWERATURE RCS PRESSURE = 1900 PSIG ACHIEVABLE CONDITIONS BASED ON ASSUMED TEWERATURE COMBINATIONS i

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