ML19093A561

Non-LOCA Accidents Safety Evaluation for Higher Levels of Steam Generator Tube Plugging

ML19093A561
Person / Time
Site:	Surry
Issue date:	08/09/1977
From:	Virginia Electric & Power Co (VEPCO)
To:	Office of Nuclear Reactor Regulation
References
Download: ML19093A561 (44)

Text

50-280 NON-LOCA ACCIDENTS SAFETY EVALUATION FOR

'HIGHER LEVELS OF STEAM GENERATOR TUBE PLUGGING Rec I d wth ltr dtd 8/9/77 *. ~. MCN 772230048 NOTICE -

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.e e

e NON-LOCA ACCIDENTS SAFETY EVALUATION FOR HIGHER LEVELS OF STEAM GENERATOR TUBE PLUGGING August 8, 1977 Virginia Electric and Power Company

e e

e

1.0 INTRODUCTION

The current Surry 1 and 2 safety analyses are valid for steam generator tube plugging levels,l)f up rto 19% on Unit 1 and up to 20% on Unit

2.

Above these levels, the current LOCA~ECCS accident analysis as well as some of the current non-LOCA accident analyses are not valid because:

(1) the RCS flow rate is reduced,to below the thermal design value assumed in the current LOCA-ECCS and applicable non-LOCA analyses, (2) the adverse impact of higher steam generator plugging levels on the blowdown and reflood phases of LOCA-ECCS analysis has not been analyzed, (3) the reduction in RCS volume (from the plugged steam generator tubes) can have an impact on some of the current non-LOCA analyses and. must no~J hP. explicitly considered, and (4) the pump coast down characteristics are more severe than those assumed in the current loss of flow analysis.

It is the purpose of this report to present an evaluation of the

• applicable non-LOCA accidents, considering the above factors, to demonstrate that with appropriate Technical Specifications changes, Surry 1 and 2 can be

• operated safely from a non-LOCA accident standpoint with up to 40% of the steam generator tubes plugged.

A LOCA-ECCS accident reanalysis for higher than 20% steam generator tube plugging and wh~ch considers the above factors is currently being performed, but is not included. It will, however, be

--submitted as a supplement to thi_s report in the near future.

The evaluation provided in this report was conducted as follows:

(1)

Determine the RCS flow rate associated with 40% steam generator tube plugging.

I e

(2}.aluate the impact of this tub.ugging and the associated RCS flow rate on those significant parameters which influence the results of the applicable non-LOCA accident analyses.

(3)

Reanalyze those non-LOCA accidents which are either most limiting or.most 1sensitive to the impacts resulting from 40% tube plugging level.

The remainder of this report is organized as follows:

Conservative flow rates versus level of steam generator tube plugging are developed in Section 2.

The applicable non-LOCA accident evalua-tions and reanalyses are provided in Section 3.

The required changes to the Technical Specifications are summarized in Section 4.

Conclusions are given in Section 5.

References are provided in Section 6.

e 2.0 FLOW RATE VERSUS LEVEL OF STEAM GENERATOR TUBE PLUGGING

~

Flow measurements have been taken at the Surry Power Station for e

several levels of steam generator tube plugging.

These data were than com-pared to the flow rates obtained from the analytical model used to calculate the best-estimate flow rate( 3 )_

Deviations between the model prediction nnd the measurement data points were conservatively accounted for by sub-tracting a constant bias (equal _to the largest deviation between the measure-ment data and the design prediction) from the model prediction curve of flow rate versus steam generator tube plugging level.

This measurement bias corrected curve was then further reduced by a factor of 1.02 to account for measurement instrumentation uncertainty (see Table 1).

The result{ng curve of flow rate versus level of steam generator tube plugging is provided in Figure 1.

This curve indicates that a tube plugging level of 40% will conservatively result in a flow rate of no more than 10% below the thermal design flow rate of 88,500 gpm per loop.

This value, 79,650 gpm per loop, was then used, along with the tube plugging level of 40%, as the basis for the non-LOCA accident evaluation.

e 3.0 ACCIDENT ANALYSIS 3.1 Introduction The impact of higher steam generator tube plugging levels of up to 40% on the non-LOCA accident analyses presented in Chapter 14 of the FSAR has been assessed.

The basic approach used was to identify th2 important para-meters for each accident, determine which of these parameters were affected by the higher steam generator tube plugging levels, and then determined how the impacted parameters affected the accident analysis.

The resulting impacts were determined by either evaluating the accident to qualitatively demonstrate that the accident is not limiting or reanalyzing the affected accident (if the accident was found to be limiting or very sensitive to the impact of higher steam generator tube plugging levels).

The evaluations were consistent with the following assumptions:

Thermal design flow, gpm/loop S. G. tube plugging,%

• . Maximum allowed power, Mwt (102% of)

T at 100% power, °F (+ 4°F)

Afvit 100% power, °F FrH Fq maximum Used In This Report

.79,650 40 2441 574.4 69.1 1.55 2.0 Used in the Currently Approved Analyses "88,500 19 (Unit 1), 20 (Unit 2) 2441 574.4 62.8

1. 55 2.0 In general, reanalysis and evaluation techniques were based on the "assumptions and methods employed in the FSAR.

Exceptions to this policy are noted in the text.

-The evaluation of the non-LOCA accidents is presented in _Section 3.2.

The results of this evaluation indicated that the following accidents

• • were limiting (i.e., the results of the accident analysis are close to

• safety limits) or most sensitive to the impa:ct of the higher steam generator tube plugging levels:

Uncontrolled Control Rod Withdrawal at Power Uncontrolled Boron Dilution e

e e

e Loss of Flow The results of the reanalyses of these accidents are provided in Section 3.3.

3.2 Evaluation The following accidents were* evaluated and found to have sufficient margin to the accident safety limits.

A.

Control Rod Withdrawal From a Subcritical Conditionf2]

A control rod assembly withdrawal incident when the reactor is subcritical results in an uncontrolled addition of reactivity leading to a power excursion (Section 14. 2.1. *of the FSAR).

The nuclear power response is characterized by a very fast rise terminated by the reactivity feedback of the negative fuel temperarture coeffi-cient.

The power excursion causes a heatup of the moderator.

However 1since the power rise is rapid and is followed by an imme-diate reactor trip, the moderator t~npe~ature rise is small.

. nuclear power response is primarily a function of the Doppler tem-perature coefficient.

The reduction in primary coolant flow is the primary impact which influences this accident.

The reduced primary coolant flow results in a decreased core heat transfer coefficient which in turn

• results in a faster fuel temperature increase than reported in the most recent analysis. [2]

The faster temperature increase would result in more Doppler feedback thus reducing the nuclear power heat flux

• excursion, as presented in Reference 2, which would partially com-pensate for the flow reduction.

Therefore, the nuclear transient is only moderately sensitive to the impact of steam generator tube plugging.

The most recent analysis[2] shows that for a 65 x 10-5 /::,k/sec reactivity insertion rate, the peak heat flux achieved is 59.4% of e

e*

e e

nominal.

This is conservativg_,,.fqr the higher plugging situation for the reasons stated above.:. "The resultant peak fuel average temperature was 747°F.

A 10% reduction in flow and the associated reduction in core heat transfer coetficient would degrade heat transfer from the fuel by a maximum of 10% and increase the rise in peak fuel and clad temperatures by a maximum of 10%.

Therefore, the fuel and clad temperatures would be less than 767°F a:1d 602°F, respectively, for the present evaluation.

These values are still significantly below fuel melt (5080°F) and zirconium-HzO reaction (1800°F) limits, and the impact of increased steam generator tube plugging, up to 40%,

would not result in a violation of safety limits.

B.

Control Rod Assembly Drop[l]

The drop of a Control Rod Assembly results in a step decrease in reactivity which produces a similar reduction in core power, thus redu~ini rhe roolent average f-c,TTIT)D,...~i-'t,-,.-,....

............ "'t',._.._.......................__.

ator temperature coefficient (-35 pcmi°F) assumed in the analysis results in a power increase (overshoot) above the turbine power run-back value causing a temporary imbalance between core power and secondary power extraction capability.

This analysis is potentially sensitive to steam generator tube pluggi?g due to the reduc~d flow.

The effect of a 10% reduction in initial RCS flow would be a smaller reduction in coolant average temperature.

Thus the power overshoot would be less than the value shown in Section 14.2.4 of the FSAR.

Based on the FSAR transient, statepoints were evaluated consistent with a 10% reduction in flow.

The results of this DNB evaluation showed that the DNBR limit-of 1.30 can be accommodated with margin.

Therefore, -the impact of increased steam generator tube plugging on the Control Rod Assembly Drop Accident analysis would not appreciably affect the margin to the safety limits.

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C.

Startup of an Inactive Reactor Coolant Loop[l]

~

An inadvertent startup of an idle reactor coolant pump with loop e

stop valves open results in the injection of cold water into the core.

This accident need not be addressed due to Technical Specifications restrictions which prohibit power operation with a loop out of service.

However, evaluation shows, that the results presented in the FSAR would be conservative for any impacts associated with increased levels of steam generator tube plugging.

D.

Excessive Heat Removal Due to Feedwater System Malfunction[!)

The addition of excessive feedwater and inadvertent opening of the feedwater bypass valve are excessive heat removal incidents which result in a power increase due to moderator feedback.

Increased levels of steam generator tube plugging would impact this analysis principally due to the reduced flow.

Section 14.2.7 of the FSAR presents two cases.

The first case assumes a zero moderator coefficient, which is used to d~monstrate inherent transient attenuation capability during a feedwater re-duction.

A reduction in flow will have a negligible effect on stability since the reactivity insertion is identical to the FSAR case due to the zero moderator temperature coefficient.

DNB is not a consideration for this case since DNBR's do not fall below the steady state value.

This is due to the relatively large reduction in Ta.vg*

The reduction in flow, however, will result in the initial steady state DNBR being reduced from 1.73 to ~1.51 (this value cor-responds to the DNBR which was calculated at time zero in the Loss of Flow reanalysis using THING).

Thus, adequate margin to safety limits is retained.

The second case assumes a large negative moderator coefficent, The impact of increased steam generator tube plugging (reduction in F

e e

e flow) will result in a slower cooldown and, therefore, a lower reactivity insertion rate than in the FSAR analysis.

The integral reactivity insertion due to moderator temperature reduction will be less than the FSAR case, thus producing a lower peak nuclear power.

Therefore, the reduction in DNBR from khe steady state value (~1.51 for increased steam tube plugging levels) would be no greater than that shown in the FSAR.

The FSAR shows a DNBR reduction of ~a.as.

Thus, the la% flow reduction will result in a minimum DNBR of ~1.45, and considerable margin to safety limits.

Evaluation has shown that sufficient margin is available to the safety limits for the Feedwater System Malfunction Accident for increased levels of steam generator tube plugging.

In addition, further protection is assured fot b~th casei via the revised Over-temperature ~T protection setpoints.

E.

Excessive Load Inc*rease ll, 2 J An excessive load increase event, in which the steam load exceeds the core power, results in a decrease in reactor coolant system temper-ature, which is very similar ta the feedwater malfunction analysis. As in the Feedwater Malfunction Accident, reduced flow is the principal impact on this accident due to increased levels of steam generator

.tube plugging.

Four Excessive Load Increase cases are presented in FSAR Section 14.2.8.

BOL and EOL in manual control.

BOL and EOL in automatic control.

All four cases show considerable margin to safety limits.

The applicable results indicate that reactor trip is not encountered and a minimum DNBR of 1.53 is calculated.

The impact of increased steam generator tube plugging on the F.

e e

e Excessive Load Increase Accident is an overall reduction in DNBR of approximately 13%.

Applying this general reduction to the Excessive Load Increase Accident transient shown in the I

FSAR yields a minimum DNBR of ~1.33.

Thus there is margin to safety limits.

In addition, the revised Overtemperature ~T trip protection setpoints afford additional protection. - The adequacy of this protection was verified in the Rod Withdrawal at Power Accident reanalysis.

The reactiv:i.ty insertion rate associated with the Excessive Load Increase Accident (7 pcm/sec) is less limiting than the reactivity insertion rate which results in the minimum DNBR for the revised Overtemperature ~T protection setpoint equation (see Figure 2).

Consequently, the minimum DNBR resulting for the Excessive Load Increase Accident would be g1-i::ai.er i.l1a.11 i.:lre mluimulll DN:OR achieved for che Roci Withdrawal at Power Accident reanalysis which is greater than the DNBR safety limit.

These results indicate that margin is availabl~

. to the safety limits for the Excessive Load Increase Accident.

Locked Rotor Ul The FSAR (Section 14.2.9) shows that the most severe Locked Rotor Accident is an instantaneous seizure of a reactor coolant pump rotor at 100% power with three loops operating.

Following the incident, reactor coolant system temperature rises until shortly after reactor trip.

The impact on the Locked Rotor Accident of increased steam genera-tor tube plugging will be primarily due to the reduced flow.

These impacts will not affect the time to DNB since DNB is conservatively assumed to occur at the beginning of the transient.

The flow coast-down in-the affected loop due to the Locked Rotor is so rapid

.e e

that the time of reactor trip (low flow setpoint is reached) is essentially identical to that presented in Reference 2.

Therefore, the nuclear power and heat flux responses will be the same as shown.

However, the reduction in flow would result in slightly higher cal-culated system pressures and fuel and clad temperatures.

The currently applicable analysis[ 2) shows a 2alculated peak fuel temperature of 3100°F and a peak clad temperature of 1794°F.

Peak temperatures are highly sensitive to the initial hot spot values assumed in the analysis.

The above analysis was based on a hot spot heat transfer calculation which employed heat flux and fuel temperatures based on an Fq of 2.55.

LOCA considerations rcqu~re that an Fq limit of 22.0 (value assumes axial stack height and spike penalties) be used.

This results in a 20% reduction in total. energy input to the hot spot which will more than compensate for the 10% reduction in flow.

Consequently, the expected peak fuel and clad temperatures would remain below the results of the currently applicable analysis.

It-is estimated that the peak system pressure will increase 'vl50 psia above the previous value, however, the maximum calculated value was 2626 psia.

This is significantly below the pressure at which vessel stress limits are exceeded ('v500 psia exists to this limit),

thus, considerable margin exists to absorb any slight pressure increase.

(It should be noted that the 40% reduction in the number of steam gene-rator tubes would result in approximately a 15% reduction:in primary coolant mass which would decrease the heat capacity of the RCS by the same amount.

This would not result in higher peak temperatures or pressures, however, since the peak values are reached in considerable less than one loop transport time constant.)

Therefore, operation at reduced flow ~ill not cause safety limits e

e to be exceeded for a Locked Rotor Accident.

G.

Loss of External Electrical Load[2)

The result of a loss of load is a core power level*which momen-tarily exceeds the secondary system power extraction causing an increase in core water temperature.

The impact of increased levels of steam generator tube plug-ging would be again principally due to the reduced flow and the decreased RCS mass inventory.

Two cases, analyzed for both begin-ning and end of life conditions, are presented in Section 14.2.10 of the FSAR:

1.

Reactor in automatic rod control with operation of the pres-surizer spray and the pressurizer power operated relief valves; and

2.

Reactor in manual rod control with no credit for pressurizer tit.

spray or power operated relier valves.

e The most recent analysis[ 2] results in a peak pressurizer pressure of 2250 psia following reactor trip and a minimum DNBR of 1.45.

A reduction in loop flow and.RCS mass inventory will result in a more rapid pressure rise than is currently shown.

The effect will be minor, however, since the reactor is tripped on high pressurizer pressure.

Thus, the time to trip will be decreased which will re-

.sult in a lower total energy input to the coolant.

Therefore, although the initial margin to DNB will be reduced, the minimum

• transient DNBR will be only slightly affected and the margin to the safety limits will be maintained.

In addition, the revised Over-

... temperature l!.T setpoints will assure adequate margin to DNB.

{

H.

Loss of Normal Feedwater/Station Blackout[l)

This transient is analyzed to determine that the peak RCS pressure does not eNceed allowable limits and that the core remains covered e

e I.

e with water.

These criteria are assured by applying the more strin-gent requirement that the pressurizer must not be filled with water.

Increased steam generator tube plugging levels would impact the, ac,cident principally due to reduced flow.

The effect of these impacts would be a larger and more rapid heatup of the primary system.

The resulting coolant density change would increase the volume of water in the pressurizer.

The analysis results presented in Sections 14.2.11 and 14.2.12 of the FSAR show that considerable margin is available.

This analysis shows that the peak pressurizer volume reached is 1055 ft 3 on less than a 100 ft 3 change in volume.

This result was due.to a 10°F change in coolant average temperature.

Using the highly conservative assumption that the average tsnperature change would double due to flow.reductions, this would result in a maximum increase of less than 200 fc3 in liquid volume.

rhis is still below the 13uu ft3

-capacity of the pressurizer thus no reanalysis is necessary.

-In addition, due to the relatively long duration of the tran-sient following trip, the results are highly sensitive to residual (decay) heat generation.

Residual heat generation is directly pro-portional to initial power level preceding the trip.

The accident

.. assumed the power to be at 102% of the maximum turbine rating 2546 Mwt.

Thus 1 the total energy input to the system would be ~4.3% less than assumes in the FSAR.

Therefore, the results of this reevaluation indicate that the impact of increased levels of steam generator tube plugging will allow sufficient margin to* be maintained to the safety limit asso-ciated with the Loss of Normal Feedwater/Station Blackout Accident.

Rupture of a Control Rod Drive Mechanism Housing, Control Rod

• Ej ec t ion l 5 ]

The rupture of a control rod drive mechanism housing which allowed e

e e

e a control rod assembly to be rapidly ejected from the core would result in a core thermal power excursion.

This power excursion would be limited by the Doppler reactivity effect as a result of the increased fuel temperature and would be terminated by a reactor trip activated by high nuclear power signals.

The rod ejection transient is analyzed at full power and hot standby for both beginning and end of life conditions (Section 14.3.2 of the FSAR).

Reduced core flow is the primary impact resulting from increased levels of steam generator tube plugging.

This impact would result in a reduction in heat transfer to the coolant which would increase clad and fuel peak temperatures.

The current analysis[S]

results and inputs are summarized in Table II.

As is shown, all cases have significant margin to fuel failure limits.

The effect of reduc-ing flow by 10% is to primarily increase the peak clad temperatures by *ulOO"r.

The <.:urrenL i.J.ualys.i.s shuws L11i:iL [o:.:- *a.ll - -

~- - 1 *. -

~ +

L..Cl.LIC.~

c..&.

V c.a....1....u.C..

-at least 200°F can be accommodated before peak clad limits are reached (2700°F).

The fuel temperatures will also increase, however, they will increase much less than the clad increase due to the rapid nature of

-the transient.

In addition, there is a significant degree of conservatism in the inputs.

The ejected rod worths and post ejection peaking factors are

~10% above the calculated Surry reload values.

Also for the full

-power cases, the initial hot spot fuel temperatures were calculated

-assuming an Fq of 2.55.

Due to LOCA considerations, the Fq limit

• will be 'v2,0.

This results in more than a 175°F reduction in _initial fuel temperature which translates into a 'v75°F reduction in peak transient fuel temperatures which will compensate for the reduction in thermal design flow.

Therefore, the ~mpact of increased levels of steam generator e

e tube plugging on the Rod Ejection* Accident will not significantly reduce the margin to the safety limit due to the conservative inputs and large margin to the limits.

J.

Steamline Break[6]

The steamline break transient is analyzed for hot zero power, end of life conditions (Section 14.3.2 of the FSAR) for the following cases:

-Hypothetical Break (steam pipe rupture)

Inside Containment with and without power Outside Containment with and without power

-Credible Break (Dump valve opening)

A steamline break results in a rapid depressurization of the stea~ generators which causes a large reactivity insertion to the core via primary cooldown.

The accept.ance* criteria for this accident

-is that no DNB must occur following a retm;n to power.

This limit, however, is highly conservative since steam line break is c.Lassi:fieci

.. as a Condition III event.

As such, the *occurrence of DNB in small regions of the core (1\\,5%) would not violate NRC acceptance criteria.

-The impact of increased levels of steam generator tube plugging

-.would affect the accident principally due to the reduced flow, re-

.sduced RCS inventory, and reduced heat transfer coefficient.

These

.impacts would result in changed cooldown and feedback reactivity characteristics such that the return to power as shown in the

-*"Previous analysis [6] would be slightly conservative with respect to

• 'the lower initial flow conditions.

In addition, the time of Safety Injection actuation would be unaffected by flow conditions for the

• Hypothetical Breaks.

This coupled with,the slight+Y slower return to power would result in a reduction in peak ave~age power for the cases with and without power and indicate results conser*vative with respect to the current analysis.

e e

e e

Thus the impact of increased levels of steam generator tube plugging will not result in a violation of Westinghouse or NRC safety limits.

3.3 Reanalysis The following accidents were reanalyzed because they were either limiting or were sensitive to the *impact of increased steam generator tube plugging.

A.

Uncontrolled Control Rod Assembly Withdrawal at Power[ 2]

An uncontrolled control rod assembly withdrawal at power pro-*

duces a mismatch in steam flow and core power, resulting in an increase in reactor coolant temperature.

Increased steam generator tube plugging will impact the analysis principally due to the influence of the reduced flow, the elevated outlet temperature, and the incn~ased loop transit time.

The first two impacts will result in less initial margin to DNB.

The third impact requires new values of lead/lag time constants to be determined for the Overtemperature

~T setpoint equation.

Additionally, to assure adequate core pro-

. tection, the Reactor Core Thermal and Hydraulic Safety Limits have been recalculated consistent with the reduction in RCS flow.

Based on these new protection lines, the Overtemperature and Overpower ~T setpoint equations constants have been recalculated consistent with the new Core Limits.

This accident ha*s been reanalyzed to verify

• the protection setpoints and the lead/lag time constants.

The transient was reanalyzed employing the same digital computer

.code and assumptions regarding initial conditions and instrumentation

.and setpoints errors used for the FSAR, including:

1.

Power levels equal to 102%, 62%, and 12% of 2441 MWT; 2.. Inlet temperature 4°F above Tin corresponding to the initial power level; e

e e

e

3.

Pressure (2220 psia) 30 psia below nominal;

4.

Reactor trip on high nuclear flux at 118% of nominal power, with trip delay time of 0.5 Seconds; and

5.

The setpoints for the Overtemperature tT reactor trip function are those which presently appear in the Technical Specifications section of this report, with allowances for instrumentation errors.

A trip delay time of 2.0 seconds was used.

6.

Nominal flow is 79,650 gpm/loop.

Figures 2 through 4* show the minimum DNBR as a function of reactivity insertion rate.

The limiting case for DNB margin is a reactivity insertion rate of 6.2 x 10-5 tk/sec from full,power initial conditions which results in a minimum DNBR of 1.31.

These results demonstrate that the conclusions presented in the FSAR are still valid.

That is, the core and rea~tor coolant system are not adversely affected since nuclear flux and Overtem-perature tT trips prevent the core minimum DNB ratio from falling below 1.30 for this incident.

Thus the setpoint equation changes have adequately compensated for the reduction in thermal design flow.

B.

Boron Dilution[l]

Increased levels of steam generator tube plugging can impact cer-tain boron dilution accident analysis cases primarily due to the re-

-duction in the RCS volume.

For a boron dilution incident during re-fueling or startup, while the reactor is subcritical, it is necessary that the operator have sufficient time to identify the problem and terminate the dilution before the reactor becomes critical.

Section

• • 14.2.5 of the FSAR shows that the operator has sufficient time to satisfy this criteria.

Tube plugging has no affect on the analysis at refueling condi-tions since only the reactor vessel and RHR system volumes are considered.

For dilution during startup, however, the effective I

e e

• e

'I e

e volume of primary coolant in the steam generator tubes has been reduced by ~40% (936 ft 3).

Thus the volume of the reactor coolant is reduced from 7600 ft 3 to 6664 ft3.

The minimum time required for the reactor to return critical is reduced from 98 minutes to at least 82 minutes.

Thus adequate time is available for the oper-ator to recognize and terminate the dilution flow from startup conditions.

For dilution at power, it is necessary that the time to lose shutdown margin be sufficient to allow identification of the problem and termination of the dilution.

As in the dilution during startup case, the RCS volume reduction due to steam generator tube plugging must be considered.

Assuming a total shutdown margin of 1% 6k/k, as in the FSAR analysis, and accounting for the RCS volume reduction, a reduction in the time to lose shutdown margin from 16.3 minutes to 13.8 minutes would result when the reactor is in automatic control.

The Te~hnical Specifications will be changed to require a total shutdown margin of 1.77% 6k/k at the beginning of a cycle.

Analyzing the accident with this shutdown margin and the same RCS volume reduction results in an operator action time of 24 minutes which is sufficient.

For dilution at power in which the reactor is in manual con-trol, the reduction in RCS volume would increase the effective reactivity insertion rate of the dilution which sould slight'ly

. *reduce the time to the Overtemperature 6T trip setpoint.

However,

• *when the accident is reanalyzed for a shutdown margin of 1.77 percent 6k/k, the operator action time is 22 minutes which is sufficient.

The FSAR (Section 14.2.6) also shows that the operator has sufficient time to terminate a boron dilution during startup of an inactive loop with loop stop valves closed.

This incident is caused e

e by violation of adminstrative procedures which require that boron concentration in the inactive loop be checked prior to opening the loop stop valve.

This accident need not be addressed due to

)

Technical Specification restrictions which do not allow operation with a loop out of service.

However, analysis would show similar results as in the power dilution cases.

Thus operator action time would increase from that shown in the FSAR.

The limiting boron dilution accident cases have been reanalyzed consistent with the impact of increased steam generator tube plugging levels.

For the Technical Specifications change in shutdown margin, the results indicate that sufficient operator action time is main-tained.

C.

Loss of Reactor Coolant Flow[2]

As demonstrated in the FSAR, Section 14.2.9, the most severe loss 0.[ Iluw L.x:auslem:: is caused by the simultaneous loss of elec-trical power to all three reactor coolant pumps.

This transient was reanalyzed to determine the effect of steam generator tube plug-

-ging on the minimum DNBR reached during the incident.

Tube plug~

ging will result in a significant decrease in margin to safety limits due to the following effects:

-Higher loop resistances result in a more rapid flow coast-

• down.

-Lower initial flows result in less margin to the 1.30 DNBR limit.

Thus reanalysis is required.

Analysis methods and assumptions used in the reevaluation were 0 consistent with those employed in the FSAR.

These assumptions include:

1.

Initial operating conditions most adverse with respect to the margin to DNBR, i.e., maximum steady state power level (102%

of nominal), minimum pressure (2220 psia), and maximum tem-perature (578.4°F);

-is..:.

e e

2.

Maximum Doppler power and temperature coefficient and most positive moderator temperature coefficient;

3.

Time from loss of power to all pumps to the initia.tion of control rod assembly motion (reactor trip) of 1.2 seconds; and

4.

4% ~k trip rea~tivity from full power using frozen fe~dback!

5.

Loop resistances equivalent to 40% tube plugging.

The flow coastdown was calculated by.the PHEONix[3]

code, and the resultant system transient was simulated using the LOFTRAN(4]

code.

The calculation was performed for a zero constant moderator coefficient value.

The minimum value of DNBR presented in the FSAR for this incident was 1.33.

The DNB evalu2tion performed in the reanalysis was based on the evaluation model described in WCAP-8013.

The results have also incorporated the effect of rod bowing.

Figures 5 through 8 show the flow coastdown, minimum DNB ratio vs. time.

nuclenr.,...._T.T,.....,...

pvvvL-.a..,

flux, ::ind Steam generator tube plugging appreciably affects the results of the complete loss of flow transient; however, the minimum DNBR

-remains above 1.30 for this incident.

The three pump case was re-

-analyzed since it is the most limiting -one presented in the FSAR.

Loss of a single pump with all loops in service or with a single loop out of.service and the loop stop valves open or closed were less limiting.

WThe reactivity rate versus control rod* insertion following a reactor scram used in this analysis is slightly less (i.e., more conservative) than that used in the previously *applicable analyses *.

(Please note that the conclusion of the previously applicable analyses are not affec-ted by the use of the new reactivity rate insertion curve),

4.0 TECHNICAL SPECIFICATION CHANGES

~

As indicated in Section 3.0, three Technical Specifications e

changes are required:

Reactor Core Thermal and Hydraulic Safety Limits Overtemperature and Overpower 6T Protection Setpoints*

Shutdown Margin The first two changes are required to reflect the effects of reduced RCS flow and to protect against overpower and overtemperature transients.

The adequacy of these changes was demonstrated by a reanalysis of the Uncontrolled Control Rod Withdrawal at Power Accident.

The third change is required to offset the effect of a reduced RCS volume on operator action time during a Boron Dilution Accident.

The specific Technical Specification changes are given in Attachment 2 to this submittal.

• The setpoints were calculated in accordance with the methods outlined in WCAP-8745;' Reactor Design Bases Fdr The Thermal Overpower 6T and Thermal Overte.mperature L'lT Trip Functions\\ March 1977.

I e

e e

5.0 CONCLUSION

S Based on an analytical and empiricai study of the Surry Units 1 and 2 flow characteristics, a 4C~ steam generator plugging level was con-servatively determined to result in a flow reduction of 10% below thermal design flow.

The impact of this higher steam generator tube plugging level on the non-LOCA accident analyses applicable to Surry Units 1 and 2 has been assessed.

The reevaluation has indicated that the impacts associated with increased steam generator tube plugging can be accommodated with mar-gin to the safety limits for the following accident analyses:

Control Rod Withdrawal from a Subcritical Condition Concrol Rod Assembly Drop Startup of an Inactive Reactor Coolant Loop Excessive Heat Removal Due to Feedwater System Malfunction Locked Rotor Loss of External Electric Load Steamline Break Several accidents were determined to be either limiting or most

• • sensitive to the impact of higher steam generator* tube plugging levels.

• • The 'following accidents were rea..rialyzed:

Uncontrolled Control Rod Withdrawal at Power

• Uncontrolled Boron Dilution Loss of Flow

• Based on these reanalyses, it was determined that the results meet the

• **appropriate safety limits with the* conservative Technical Specifications

.. *changes to the Reactor Core Thermal and Hydraulic Safety Limits, Over-

':temperature and Overpower t,.T protection setpoint constants and the shut-

. ****down margin.

Therefore, it is concluded that the impact of increased steam

,-***.)..,

generator plugging levels, up to 40%, can be accommodated without adversely

'affecting the safe operation of Surry Units 1 and 2.

e

6. 0 REFERENCES -.*

1.

"Final Safety Analysis Repor,t - S,urry Power Stations Units 1 and 2,"

Dockets Nos. 50-280, 50-281.

2. Letter, C. M. Stallings (Vepco) to K. R. Goller (NRC), Amendment to Operating Licenses DPR-32 and DPR-37 Technical Specifications change No. 30 Surry Power Station - Unit Nos. 1 and 2, Serial No. 553, June 5, 1975.

3.

"Calculation of Flow Coastdown After Loss of Reactor Coolant Pump (PHOENIX Code)", WCAP-7551, August, 1970.

4.

LOFTRAN Code Description, WCAP-7878, March, 1972.

5. Letter, C. M. Stallings (Vepco) to B. C. Rusche (NRC), Amendment to Operating Licenses DPR-32 and DPR-37 Technical Specifications Change No. 39 Surry Power Station - Units Nos. 1 and 2, Serial No. 936, March 11, 1976.

6.

Letter, C. M. Stallings (Vepco) to B. C. Rusche (NRC), Amendment to the Operating License Technical Specification Change No. 47 Surry Power Station Units 1 and 2; Serial No. 256, September 27, 1976.

7,

-"Fuel Densification Surry Power Station, 11 WCAP 8013, December, 1972,

/

I II I

~

'\\

e e

TABLE 1 REACTOR COOLANT FLOW MEASUREMENT UNCERTAINTY Parameter Uncertainty Feedwater Flow Feedwater Temp.

Steam Pressure THot Tcald RCS Pressure

+/-1.25%

H.0°F

+/-30 psi

+/-0.5°F

+/-0.5°F

+/-50 psi RC Loop Flow Uncertainty

+/-1. 25%

+/-0.2%

+/-0.1%

+/-0.9%

+/-0.9%

+0.2%

+/-1.8%.fi:2

'I e

TABLE II

SUMMARY

OF ROD EJECTION ANALYSIS PARAMETERS AND RESULTS BOL BOL EOL EOL Power Level,%

102 0

102 0

Ejec~ed rod work, %~k

.30

• . 90

.35

.90 Delayed neutron fraction, %

.55

.55

.48

.48 Feedback reactivity weighting 1.20 2.73 1.2 3.0 Trip rod shutdown, %~k 5

2 4

2 Fq before rod ejection 2.55 2.55 Fq after rod ejection 5.46 15.2 5.5 17.96 Number of operating pumps.

3 2

3 2

Initial fuel temperature, OF 2820 547 2820 547

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e.

e CHANGE NO. 57 TO THE TECHNICAL SPECIFICAT~ONS SURRY POWER STATION UNITS NO. 1 AND 2 AUGUST 8, 1977

.i

e e

e TS 2.1-1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMIT, REACTOR CORE Applicability Applies to the limiting combinations of thermal power, Reactor Coolant System pressure, coolant temperature and coolant flow when a reactor is critical.

Objective To maintain the integrity of the fuel cladding.

A.

The combination of reactor thermal power level, coolant pressure, and coolant temperature shall not:

1.

Exceed the limits shown in TS Figure 2.1-1 when 90% of design flow from three reactor coolant pumps exist.

2.

Exceed the limits shown in TS Figure 2.1-2 when full flow from two reactor coolant pumps exist and the reactor coolant loop stop valves in the non-operating loop are open.

3 *. Exceed the limits shown in TS Figure 2.1-3 when full flow from two reactor coolant pumps exist and the reactor coolant loop stop valves in the non-operating loop are closed.

I

TS 2.1-3

• ~niform and non-unif~ heat flux distributions.

Thetltcal DNB heat flux ratio, defined as the ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of the margin to DNB.

The minimum value of the DNB ratio (DNBR) during steady state operation, normal operatioJai tiansients and anticipated transients, is limited to 1.30.

A DNBR of 1.30 corresponds to a 95% probability at a 95%

/

confidence level that DNB will not occur and 1.s chosen as an appr-opriate margin to DNB for all operating conditions. (1)

The curves of TS Figure 2.1-1 which show the allowable power level decreas-ing with increasing temperature at selected pressures for constant flow (three loop operation) represent limits equal to, or more conservative than, the loci. of points of thermal power, coolant system average tempera-ture, and coolant system pressure for which the DNB ratio is equal to 1.30 or the average enthalpy at the exit of the core is equal to the sat-uration value.

The area where clad integrity is assured.is beiow t:hese lines.

The temperature limits are <:onsiderably more conservative than would be required if they were based upon a minimum DNB ratio of 1.30 alone but are such that the plant conditions required to violate the limits are precluded by the self-actuated safety valves on the steam generators.

The three loop operation safety limit curve has been revised to allow for heat flux peaking effects due to fuel densification and to apply to 90%

of design flow.

The effects of rod bowing are also considered in the DNBR analyses.

The curves of TS Figures 2.1-2 and 2.1-3 which* show the allowable power

• level decreasing.with increasing temperature at selected pressures for

~

constant flow (two loop operation), represent limits equal to, or more conservative,

1,,,

670 660 650

µ.i 0......,

640

-A

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(:-1

+

630

"(:-1 0

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(:-1 620 N

r-l 610 r.,

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~

580 570 560 550 0

e TS Figure 2.1-1 e

I 2000 l's14 l.855 l?s11i.

"\\

-~-

' ~\\

10 20 30 40 50 60 70 80 90 100

• -POWER (PERCENT OF RATED)

FIGURE 2.1-1 REACTOR CORE THERMAL & HYDRAULIC SAFETY LIMITS-THREE LOOP_ OPERATION, 90% DESIGN FLOW j

110 120

. l i

l1 fj

~1 fj

t l '.

.e (b)

High p1:essurizer pressure - <2385 psig.

(c)

Low pressurizer pressure - >1860 (d)

Overtemperature flT h.T<!J.T0 [K

- Kz( 1 '-J-*T1S l

1 + T2S ) (T -

where

!J.T0 = Indicated !J.T at rated thermal power, °F T = Average coolant temperature, °F T' = 574.4°F P = Pressurizer pressure, psig P' = 2235 psig Kl = 1.07 Kz = 0.0095 psig.

T') +

K3 = 0.0005 for 3-loop operation K..

.L =- 0.951 TS 2.3-2 e

K3 (P - P') - f(flI)]

Kz =

K3 =

0.01012 0.000554 for 2-loop operation with loop stop valves open in inoperable loop Kl =

. Kz =

K3 =

1.026 0.01012 0.000554 for 2-loop operation with loop stop valves closed in inoperable loop

!J.I = qt - qb, where qt and qb are the percent power in the top and bottom halves of the core respectively, and qt+ qb is total core power in percent of rated power f(!J.I) = function of A.I, percent of rated core power as shown in Figure 2.3-1

{e)

-r1 = 30 second!?

T2 = 4 seconds Overpower h.T I

T*s 2.3-3

• ,where AT0 = Indicated AT at rated thermal power, °F T = Average coolant temperature, °F T' = Average coolant temperature'measured at nominal conditions and rated power, °F K4 = A constant = 1. 07 K5 = 0 for decreasing average temperature A constant, for increaseing average temperature 0.02/°F K6 = 0 for T < T'

= 0.0011 for T > T' f(bI) as defined in (d) above,

'!3 = 10 seconds (f)

Low reactor coolant loop flow -.:,.90% of normal indicated _loop rlow as measured at elbow taps in each loop

• (g)

Low reactor coolant pump motor frequency -

~ 57.5 n2

• (h)

Reactor coolant pump under voltage -.::_ 70% of normal voltage

3.

Other reactor trip settings (a)

High pressurizer water*level -

< 92% of span (b)

Low-low steam generator water level -

> 5% of narrow range instrument span.

(c)

Low steam generator water level -

> 15% of narrow range instrument span in coincidence with steam/feedwater mismatch flow -

< 1.0x106 lbs/hr (d)

Turbine trip (e)

Safety injection - Trip settings for Safety Injection are detailed in TS Section 3.7.

~ ' -

e TS 2.3-5 e

e and source range high flux, high setpoint trips provide additional

• protection against uncontrolled startup excursions.. As power level increases, during startup, these trips are blocked to prevent unnec-essary plant trips.

) '

The high and low pressurizer pressure reactor trips limit the pressure range in which reactor operation is permitted.

The high pressurizer pressure reactor trip is also a backup to the pressurizer code safety valves for overpressure protection, and is therefore set lower than the set pressure for these valves (2485 psig).

The low pressurizer pressure reactor trip also trips the reactor in the unlikely event f

1 f

1 "d

. (3) o a

oss-o -coo ant acci enc.

The overtemperature b.T reactor trip provides core protection against and axial power distribution, provided only tha*:: the transient is slow with respect to piping transit delays from the core to the tem-perature detectors (about 3 seconds), and pressure is within the range between high and low pressure reactor trips.

With normal axial power (2) distribution, the reactor trip limit, with allowance for errors, is always below the core safety limit as shovm on TS Figure 2.1-1.

If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactcr limit is automatically reduced. (4 ) (S)

The overpower and overtemperature protection system setpoints have been revised to include effects of fuel densification on core safety limits and to apply to 90% of design flow.

The revised setpoints in the Technical Specifications will ensure that the combination of power, temperature, and pressure will not exceed the revised

e and TS 3.12-2 physics data~btained during unit startup an~ubsequent operation, will be permitted.

c.

The shutdown margin with allowance for a stuck control rod assembly shall be greater than or equal to 1.77% reactivity under all steady-(

state operation conditions, except for physics tests, from zero to full power; including effects of axial power distribution.

The shut-down margin as used here is defined as the amount by which the reactor core would 'be subcritical at hot shutdown conditions if all control rod assemblies were tripped, assuming 0

(T

>547 F) avg-that the highest worth control rod assembly remained fully withdrawn, and assuming no changes in xenon, boron, or part-length rod position.

4.

Whenever the reactor is subcritical, except for physics tests, the critical rod position, i.e., the rod position at which criticality would be achieved if the control rod assemblies were withdrawn in normal sequence with no other reactivity changes, shall not be lower than the insertion limit for

5.

zero power.

Operation with part length rods shall* be restricted such that except during

-physics tests, the part length rod banks are withdrawn from the core at all times.

6.

Insertion limits do not apply during physics tests or during periodic excerise of individual rods.

However, the shutdown margin indicated above

• must be maintained except for the low power physics test to measure control rod worth and shutdown margin.

For this test the reactor may be critical

.. with all but one full length control rod, expected to have the highest worth, inserted and part length rods fully withdrawn.

l

TS 3.12-12 still assure complia. with th.e shutdown.requirement*T~e maximum shut-down margin requirement occurs at end of core life and is based on the

~

value used 1n the analysis of the hypothetical steam break accident. The rod insertion limits are based on end of core life conditions.

The shut-e down margin

{.

for the entire cycle length is established at 1.77% reactivity.

All other accident analyses with the exception of the chemical and volume control system malfunction analysis are based on 1% reactivity shutdown margin.

Relative positions of control rod banks are determined by a specified control rod bank overlap.

This overlap is based* on the consideration of axial power shape control.

The specified control rod insertion limits have been revised to limit the potential ejected rod worth in order to account for the effects of fuel densification.

The various control rod assemblies (shutdown banks, control banks A, B, C and D and part-length rods) are each to be moved as a bank, that is, with all assemblies in the bank within one step (5/8 inch) of the bank

• position.

Position indication is provided by two methods:

a digital count of actuating pulses which shows the demand position of the banks and a linear position indicator, Linear Variable Differential Transformer, which indicates the actual assembly position.

The position indication accuracy of the Linear Differential Transformer is approximately ~5% of span

(+7.5 inches) under steady state conditions.

The relative accuracy of

• the linear position indicator is such that, with the most adverse errors, an alarm.is actuated if any two assemblies within a bank deviate by more than 14 inches.

In the event that the linear position indicator is not in service, the effects of

.. e

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e TS FIGURE 3.12-7 DELETE