ML18227B004

Non-Loca Accidents Safety Evaluation for Higher Levels of Steam Generator Tube Pugging, NAD-QR-46, June 1978

ML18227B004
Person / Time
Site:	Turkey Point
Issue date:	06/30/1978
From:	Florida Power & Light Co
To:	Office of Nuclear Reactor Regulation
References
Download: ML18227B004 (101)

Text

650 2400 psia 640 630 2250 psia 620 2100 psia 610

+

0 A

600 1900 psia

~ 5 590 580 i 570 Note: These curves are applicable with steam generator tube plugging

>19% and <25%.

560 550 540 0 20 40 60 80 100 120. 140 RATED POWER (PERCENT) c Figure 2.l-lb. Reactor Core Thermal and Hydraulic Safety Limits, 3 Loop Operation

Reactor Coolant Tem erature Overtempera-hT < bTo tK1 0. 0107 (T-574) + p- ppp453 (P-2235) f (A ) ]

hT0 Indicated hT at rated power, F CR Average temperature, F Pressurizer pressure, psig f(a,q)- a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during startup tests such that:

(q, qb) within +10 percent and -14 percent where qt and qb are the percent R

power in the top and bottom halves of the 0 core respectively, and qt +

total core power f(aq) - 0.

qb in percent of rated power, Por each percent'that the magnitqde of {qt qb ) exceeds +10 percent, the Delta-T trip setpoint shall be automatically reduced by 3.5 percent

/

of its value at interim power'.

Por each percent that the magnitude of (q qb) exceeds -14 percent, the Delta-T trip setpoint shall be automatically reduced by 2 percent of its value at interim power.

(Three Loop Operation) = 1.095*

{Two Loop Operation) 0.88

I Over-power AT < AT 111"- K 1

dt K2 (T T ) f ~Aq bT 0 ~ Indicated AT at rated power, F T ~ Average temperature, F T' Indicated average temperature at nominal conditions and rated power, F Kl . 0 for decreasing average temperature, 0.2 sec./F for increasing average temperature K2 ~ 0.00068+for T equal to or more than T';

0 for T less than dT T'ate dt of change of temperature, F/sec f (6q) ~ As def ined. above

'ressurizer Low Pressurizer, pressure equal to or greater than 1835 psig.

High Pressurizer pressure equal to or less than 2385 psig-High Pressurizer water level equal to or less than

'2% of full scale.

Reactor Coolant Flow Low reactor coolant flow equal to or greater than 90% of normal indicated flow Low reactor coolant pump motor frequency equal to or greater than 56.1 Hz Under voltage on reactor coolant pump motor bus equal to or greater than 60% of normal voltage Steam Generators Low-low steam generator water level equal to or greater than 5% of narrow range instrument scale

• This factor is 1.11 for steam generator tube plugging < 15 percent This-factor is 1.10 for steam generator tube plugging > 15 percent and < 19  %

This factor is l.08 for steam generator tube plugging >198 and

<25%.

"This factor is 0.00106 for steam generator tube plugging >19> and <25%.

2%33

6~ ~ DNB PARAMETERS The following DNB related parameters limits shall be maintained during power operation:

a- Reactor Coolant System Tavg < 578.2 F

b. Pressurizer Pressure > 2220 psia+

c. Reactor Coolant Flow > 268,500 gpm+

With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce thermal powex to less than 5% of rated thexmal power using normal shutdown pxocedures.

Compliance with a- and b. is demonstrated by verify-ing that each of the parameters is within its limits at least once each 12 hours.

Compliance with c. is demonstrated by verifying that the parameter is within its limits after each refuel-ing cycleimit not applicable during either a THERMAL POWER ramp increase in excess of (5%) RATED THERMAL POWER per minute or a THERMAL POWER step increase in excess of (10%) RATED THERMAL POWER.

+ Reactor Coolant Flow > 268,500 gpm for steam generator tube plugging

< 15%.

Reactor Coolant Plow >'63,130 gpm for steam generator tube plugging

> 15% and < 19%.

Reactor Coolant Flow > 255,075 gpm for steam generator tube plugging

> 19% and < 25%.

3.1-7

NAD-1813 NON-LOCA ACCIDENTS SAFETY EVALUATION .

FOR HIGHER LEVELS OF STEAM GEI'AERATOR TUBE PLUGGING NAD-QR-46 JUNE, 1978 FLORIDA POWER AND LIGHT COMPANY

I

1.0 INTRODUCTION

The current Turkey Point 3 and 4 safety analyses are valid for steam generator tube plugging levels of up to 19% on Units 3 and 4.

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

l. 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 cur-rent non-LOCA analyses and must now be explicitly considered, and

4. the pump coastdown characteristics are more severe than those assumed in the current loss of flow analysis.

It is purpose of this report to present. an evaluation of the applicable non-LOCA accidents, considering the above factors, to demonstrate that with appropriate Technical Specification changes, Turkey Point 3 and 4 can be operated safely from a non-LOCA accident standpoint with up to 25% of the steam generator tubes plugged. A LOCA-ECCS accident reanalysis for higher than 19% steam generator tube plugging and which considers the above factors is currently being peformed, but is not included. It will, however, be submitted as a supplement to this report. in the near future.

The evaluation provided in this report was conducted as follows:

1. Determine the RCS flow rate associated with 25% steam generator tube plugging.

2. Evaluate the impact of this tube plugging 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 sensitive to the impacts resulti'ng from 25% 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 evaluations 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.

2.0 FLOM RATE VERSUS LEVEL OF STEAM GENERATOR TUBE PLUGGING Flow measurements have been taken at the Turkey Point power station for several levels of steam generator tube plugging.

These data were then compared to the flow rates obtained from the analytical model used to calculate the best-estimate flow rate. Deviations between the model prediction and the measurement data points were conservatively accounted for by subtracting a constant bias (equal to the largest deviation between the measurement data and the design prediction) from the model prediction curve of flaw rate versus steam genera-tor 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 I).

The resulting curve of flow rate versus level of steam generator tube plugging is provided in Figure l. This curve indicates that a tube plugging level of 25% will conservatively result 'in a flow rate of no more than 5% below the thermal design flow rate of 89,500 gpm per loop. This value, 85,025 gpm per. loop, was then used, along with the tube pl'ugging level of 25%, as the basis for the non-LOCA accident evaluation.

3.0 ACCIDENT ANALYSIS

3.1 INTRODUCTION

The impact of higher steam generator tube plugging levels of up to 25% on the non-LOCA accident analyses presented in Chapter 14 of the FSAR has been assessed.

The basic approach used,was to identify the important parameters for each accident, determine which of these parameters were affected by the higher steam generator tube plugging levels,. and then determine how the im-pacted 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:

0 Used In This Report Used A

in the Currently roved Anal ses Thermal design flow, gpm/ 85,025 89,500 loop S.G. tube plugging, 25 19 Maximum allowed power, MN 2200 2200

'P avg at 100% power, T 574.2 574.2 bT at 100% power, 'F 58.9 55.9 N 1.55 (1.75 1.55 FSAR) hH F maximum 2.05 2.55 (non-LOCA)

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.

0

The evaluation of the non-LOCA accidents is presented in Section 3.2. The results of this evaluation indi-cated that the following accidents were limiting (i.e.,

the results of the accident analysis are close to safety limits) or most sensitive to the impact of the higher steam generator tube plugging levels:

Uncontrolled RCCA Withdrawal at Power Loss of Reactor Coolant Flow (Flow Coastdown Accident)

Chemical and Volume Control System Malfunction 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.

l. Uncontrolled RCCA Withdrawal From a Subcritical Cond'ition (1)

A control rod assembly withdrawal incident when the reactor is subcritical results in an uncontrolled addition of reactivity leading to a power excursion (Section 14.1.1 of the FSAR). The nuclear power response is characterized by a very fast rise termi-nated by the reactivity feedback of the negative fuel temperature coefficient. The power excursion causes a heatup of the moderator.. However, since the power rise is rapid and is followed by an immediate reactor trip, the moderator temperature rise is small. Thus, nuclear power response is primarily a function of the Doppler temperature 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 FSAR analysis (1) . The faster temperature increase would result in more Doppler feedback thus reducing the nuclear power heat flux excursion, as presented in Reference 1, which would partially compensate for the flow reduction.

Therefore, the nuclear transient is only moderately

, sensitive to the impact of steam generator tube plugging.

-5 The FSAR analysis (1)

~

shows that for a 60 x 10 Bk/sec reactivity insertion rate, the peak heat flux achieved is 46.8% of nominal. This is conservative for the higher plugging situation for the reasons stated above. The resultant peak fuel average temperature was 951'F. A 5% reduction in flow and the associated reduction in core heat transfer coefficient would degrade heat transfer from the fuel by a maximum 5% and increase the rise in peak fuel and clad temperatures by a maximum of 5%.

Therefore, the fuel and clad temperatures would be less than >1000'F and >630'F, respectively, for the present evaluation. These values are still signi-ficantly below fuel melt (5080'F) and zirconium-H 2 0 reaction (1800'F) limits, and the impact of increased steam generator tube plugging, up to 25%, would not result in a violation of safety limits.

2. Mal ositionin of the Part Len th Rods (1)

A malpositioning of a part length rod accident need not be addressed due to'Technical Specification

restrictions which prohibit power, operation with part length rods in the core.

3. Rod Cluster Control Assembl (RCCA) Dro (1)

The drop of a Control Rod Assembly results in a step decrease in reactivity which produces a similar reduction in core power, thus reducing the coolant average temperature. The highly negative moderator temperature coefficient (-'35 pcm/'P) assumed in the analysis results in a power increase (overshoot) above the turbine power runback value causing a temporary imbalance between core power and secondary power extraction capability. This analysis is potentially sensitive to s'team generator tube plugging due to the reduced flow. The effect of a 5% reduction in initial RCS flow would- be a smaller reduction in coolant average temperature.

Thus the power overshoot would be'ess than the value shown in Section 14.1.4 of the PSAR. Based on the PSAR transient, statepoints were evaluated consistent with a 5% reduction in flow. The results of this DNB evaluation showed that the DNBR limit 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.

4. Startu of an Inactive Reactor Coolant Loo (1)

An inadvertent startup of an idle reactor coolant pump with loop stop valves open results in the in-jection of cold water into the core. This accident need not be addressed due to Technical Specification

restrictions which prohibit power operation with a loop out of service. However, evaluation shows that the results pr'esented in the FSAR would be conservative for any impacts associated with in-creased levels of steam generator tube plugging.

5. Reduction in Feedwater Enthal Incident (1)

The addition of excessive feedwater and inadvertent opening of the feedwater bypass valve are excessive heat removal incidents which result in a power in-crease due to moderator feedback. Xncreased levels of steam generator tube plugging would impact this analysis principally due to the reduced flow.-

Section 14.1.7 of the FSAR presents two cases. The first case assumes a zero moderator coefficient, which is used to,demonstrate inherent transient attenuation capability during a feedwater reduction.

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 Tavg . The reduction in flow, however, will result in the initial steady state DNBR being reduced to <1.47 (this value corresponds to the DNBR which was calculated at time zero in the Loss of Flow reanalysis with rod bow penalty included). Thus, adequate margin to safety limits is retained.

The second case assumes a large negative moderator coefficient. The impact of increased steam generator

tube plugging (reduction in 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 the steady state value (<1.47 for increased steam tube plugging levels) would be no greater than that shown in the FSAR. The FSAR shows a DNBR reduction of ~0.04.

Thus, the 5% flow reduction will result in a minimum DNBR of <1.43, and considerable margin to safety limits.

Evaluation has shown that sufficient margin is available to the sa fety limits for the Feedwater System Malfunction Accident for increased levels of steam generator tube plugging. In addition, further protection is assured for both-cases via the identified margin in the Overtemperature hT protection setpoints (see Section 3.3.1).

6. Excessive Load Increase Incident (1)

An excessive load increase event, in which the steam load exceeds the core power, results in a decrease in reactor coolant system temperature, which is very similar to 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.

Two Excessive Load Increase cases are presented in FSAR Section 14.1.8.

-10% step load increase with manual control

-10$ step load increase with automatic control The worst case results (automatic control) indicate that with no trip actuation, steady state conditions are reached with a minimum DNBR of >1.30. Margin is available in that the F>H used in the FSAR analysis was 1.75 as compared 'to the current value of 1.55.

This results in a DNB benefit of >30%. The impact of increased steam generator tube plugging on the Excessive Load Increase Accident, is an overall reduction in DNBR of approximately 5%. Thus there is considerable margin available given all of the assumptions in the original analysis remain valid.

Tube plugging does result in, small changes to the initial conditions, however, these changes tend to be in the conservative direction.

Another area of margin available to offset. the flow reduction penalty is in the Overtemperature hT trip which is 'used for protection in the case with automatic control. Margin has been identified in the Over-temperature BT trip protection setpoints (see Section 3.3.1). As stated in the FSAR, the adequacy of this protection was verified in the Rod Withdrawal at Power Accident (see FSAR Section 14.1.2).

7. Loss of Reactor Coolant Flow (Locked Rotor Accident) (2)

The FSAR (Section 14.1.9) shows'hat 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.

10

The impact on the Locked, Rotor Accident of increased steam generator 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 coastdown in the affected loop due to the Locked.

Rotor is so rapid 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 high calculated system pressures and fuel and clad temperatures.

The currently applicable analysis (2) shows a calculated peak fuel temperature of 2940'F and a peak clad temperature of l500'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 F< of 2.55.

LOCA conside'rations require that an F limit of <2.05 (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 5% 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 +80 psia above the previous value; however, the maximum calculated value was 2720 psia. This is significantly below the pressure at which vessel stress limits are exceeded (+400 psia exists to this

limit), thus, considerable margin exists to absorb any slight pressure increase. (Xt should be noted that the 25% reduction in the number of steam generator tubes would result. in approximately a 10%

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 considerably less than one loop transport time constant.)

Therefore, operation at reduced flow will not cause safety limits to be exceeded for a Locked Rotor Accident.

8. Loss of External Electrical Load (1)

The result of a loss of'oad is a core power level which momentarily exceeds -the secondary system power extraction causing an increase in core water temperature.

The impact of increased levels of'team generator tube plugging would be again principally due to the reduced flow and the decreased RCS mass inventory. Two cases, analyzed for both beginning and end of life conditions, are presented in Section 14.1.10 of the PSAR:

a. Reactor in automatic rod control with operation of the pressurizer spray and the pressurizer power operated relief valves; and

b. reactor in manual rod control. with no credit for pressurizer spray or power operated relief valves.

The FSAR analysis (1) results in

~

a peak pressurizer 12

S pressure of 2517 psia following reactor trip and a minimum DNBR of 1.61. 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.result in a lower total energy input to the coolant. Therefore, although the inital 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 identified margin in the Overtemperature bT setpoints (see Section 3';3.1) will assure adequate margin to DNB.

9. Loss of Normal Feedwater (1)

This transient is analyzed to determine that the peak RCS pressure does not exceed allowable limits and that the core remains covered with water. These criteria are assured by applying the'ore stringent requirement that the pressurizer must not be filled with water.

Increased steam generator tube plugging levels would impact the accident 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 Section 14.1.11 of the FSAR show that considerable margin is available.

This analysis shows that the peak pressurizer volume reached is 1100 ft3 on an approximate 300 ft3 change in volume. This result was due to a 26'F change in coolant average temperature. Using the highly con-servative assumption that the average temperature 13

delta would increase by 50% due to flow reductions, this would result in a maximum increase of less than 150 ft in liquid volume. This is still below the 1300 ft capacity of the pressurizer thus no reanalysis is necessary.

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 associated with the Loss of Normal Feedwater Accident.

10. Loss of AC Power (1)

This transient is analyzed to show that upon loss of all AC power to the unit auxiliaries, the auxiliary feedwater system is sufficient. to remove stored and n

residual heat without water release through pressurized relief valves.

As in a Loss of Normal Feedwater Accident, increased steam generator tube plugging would impact the Loss of AC Power transient. primarily due to reduced flow.

Upon the loss of power to the reactor coolant pumps, coolant flow necessary for core cooling and the removal of residual heat is maintained by natural circulation in the reactor coolant. loops. In the FSAR analysis, the natural circulation flow was calculated using an analytical method based on the condition of equilibrium flow and maximum loop resistance. The reduction in natural circulation flow rates due to the 5% reduction in thermal design flow is negligible.

In addition, due to the relatively long duration of the transient following trip, the results are highly sensitive to residual (decay) heat genera-tion. Residual heat generation is directly propor-tional to initial power level preceding the trip.

The accident assumed the power to be at 102% of the maximum turbine rating 2300 MN . Thus, the total energy input to the system would be +4.3% less than assumed in the FSAR.

This reevaluation indicates that the impact of in-creased levels of steam generator tube plugging -will not adversely impact the Loss of AC Power Accident.

Sufficient margin to the safety limits are maintained.

(3)-.,

11. Ru ture of a Steam Pi e The steamline break transient is analyzed for hot zero power, end of life conditions (Section 14.2.5, of the FSAR) for the following cases:

-hypothetical break (steam pipe rupture) inside containment with and without power outside containment with and.wi'thout power

-credible break (dump valve opening)

A steamline break results in a rapid depressuriza-of the steam generators which causes a large I'ion reactivity insertion to the core via primary cool-down. The acceptance criteria for this accident is that no DNB must occur following a return to power.

This limit, however, is highly conservative since steamline break is classified as a Condition III event. As such, the occurrence of DNB in small regions of the core (+5%) would not violate NRC 15

acceptance criteria.

The impact of increased levels of steam generator tube plugging would affect the accident principally due to the reduced flow, reduced RCS inventory, and reduced heat, transfer coefficient. These impacts would result in changed cooldown and feedback reac-tivity characteristics such that the return to power as shown in the previous analysis (3) 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 slightly slower return to power would result in a reduction in peak average power for the cases with and without power and indicate results conservative with respect to the current analysis.

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

12. Ru ture of a Control Rod Drive Mechanism Housin RCCA E'ection (2)

The rupture of a control rod drive mechanism housing which allowed a control rod assembly to be rapidly ejected from the core would result in a core thermal power excursion. This power'xcursion 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 14.2.6 of the FSAR). Reduced core flow is the primary impact resulting from in-creased 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 (2)

~

results and inputs are summarized in Table II. As is shown, all cases have significant margin to fuel failure limits. The effect of reducing flow by 5%

is to primarily increase the peak clad temperatures by +50 F. The current analysis shows that for all cases a value of at least 300'F can be accommodated before peak clad limits are reached (2700 P). 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 con-servatism in the inputs. The ejected rod worths and post ejection peaking factors. are >5% above the calculated Turkey Point reload .values. Also for the full power ca'ses, the initial hot spot fuel temperatures were calculated assuming an F~ of 2.55. Due to LOCA considerations, the F limit will be <2.05. This results in more than a 175'F reduction in initial fuel temperature which, translates into a +75'P re-duction in peak transient fuel temperatures which will compensate for the reduction in'hermal design flow.

Therefore, the impact of increased levels of steam generator tube plugging on the Rod Ejection Accident will not significantly reduce the margin to the 17

safety limit due to the conservative inputs and large margin to the 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.

l. Uncontrolled Control Rod Assembl Withdrawal at Power An uncontrolled control rod assembly withdrawal at power produces 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. increased loop transit time.

As a result, there will be less initial 'margin to DNB.

To assure adequate core protection, the Reactor Core Thermal and Hydraulic Safety Limits have been recalculated consistent with the reduction in RCS flow. The overtemperature and overpower bT set-point equation constants have also been recalculated consistent with the new core limits and the methods outlined in Reference 5. Table III gives a comparison of the original (FSAR) and the recalculated constants.

For the overtemperature equation, the FSAR values are significantly more limiting than the recalculated values. The reason for this is that the FSAR analyses were based on considerably higher core peaking factors N N (F< = 1.75) than the current analysis (F> = 1.55).

18

The FSAR overtemperature aT trip constants will there-fore be maintained in the Technical Specifications (section 4.0).

Table III shows that the overpower hT setpoints have become more limiting with the flow reduction (see

,Reference 5 for a discussion of this function), so that a Technical Specification change, as shown in Section 4.0, will be required., However, no addi-tional analysis is required since the overpower 5T trip function is not explicitly used in any of the FSAR transients analyzed. Rather this trip function is considered a diverse means of trip for the Power Range Nuclear Instrumentation System.

To demonst'rate the availability of margin to DNB, even when the rod bow penalty is included, four cases of rod withdrawal were reanalyzed, corresponding to those analyzed in the FSAR, Section 14.1.2.

1. Fast RCCA Withdrawal at Full Power Reactivity insertion rate 5. 625*10 -4 5k/sec

2. Slow RCCA Withdrawal at Full Power Reactivity insertion'ate 2. 5*10 -5 6k/sec

3. Fast RCCA Withdrawal at, 60% Power Reactivity insertion rate 3*10 -4 5k/sec

4. Slow RCCA Withdrawal at 60% Power Reactivity insertion rate 0.8*10 6k/sec The full power analysis was performed with initial conditions of 102% of 2200 MW core power, system pressure at 2220 psia (nominal 30 psi), core inlet 19

temperature at 550.2'F (nominal + O') and system flow at 85025 gpm per loop (95% of design flow rate).

'Correspondingly appropriate conditions were used at, 60% power.

The calculational procedure followed the one .

described in the report on.FPL Safety Analysis Methods (6) , which shows close agreement between the results obtained with these procedures and those shown in the FSAR. The system behavior was obtained with the DYNODE code (6i' '7) with proper adjustments for .primary loop inventory and steam generator coolant. flow area and heat transfer surface. The fast rod withdrawal cases tripped on high neutron power while the slow rod withdrawal cases tripped on overtemperature hT. The transient curves for the full power case are shown in Figures 2 through 9. The DNB ratios were calculated by means of the COBRA-IIIC/

'ode, MIT (6,8) modified for the Westinghouse "L" grid correction, taking into account cold wall effects.

N The hot channel enthalpy rise factor, F was 1.55 and the axial profile a chopped cosine with a peak to average ratio of 1.55. Reduction in stack height due to densification was'aken .into account. The procedure is described in detail in Reference 9, Section 3.2.2, Reanalysis Model.

The results of the rod withdrawal cases are summarized in Table IV. When the rod bow penalty of 28.9% for high burnup fuel( is included., the DNB ratios are all above the DNBR limit of 1.24. Actually, there is considerably more margin to DNB than shown here, because the heat flux peaking factor, F<, on 20

which these analyses are based is 2.32, whereas the maximum allowable peaking factor for 25% tube plugging is less than 2.05.

2. Loss of Reactor Coolant Flow (Flow Coast-Down Accident) (2)

As demonstrated in the FSAR, Section 14.1.9, the most severe loss of flow transient is caused by the simultaneous loss of electrical power to all three reactor coolant pumps. This transient was reanalyzed to determine the effect of steam generator tube .plugging on the minimum DNBR reached during the incident. Tube plugging will result in a decrease in margin to safety limits due to the following effects:

, Higher loop resistances result in a more rapid flow coastdown.

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

-As for the rod withdrawal cases, the initial power was 102% of 2200 MWt with minimum pressure (2220 psia) and maximum inlet temperature (550. 2 F) . The initial flow was 95% of design flow. As in the FSAR, the maximum Doppler coefficient and most positive moderator coefficient were used in the analysis and the time from loss of power to all pumps to the initiation of control rod assembly motion (reactor trip) was taken as 1.6 seconds.

The flow coastdown was calculated by the RETRAN (6,4) code for a loop resistance equivalent to 25% tube plugging. The system transient was simulated using 21

the DYNODE 'ode. The DNBR'was calculated by the COBRA-IIIC/MIT 'ode.

(6,8)

The results of the loss of flow calculation are shown in Figures 10 through 14. The minimum DNBR, with the 28.9% rod bow penalty included is .1.29.

Thus,'the DNB limit of 1.24 is got. violated. (9)

3. Chemical and Volume Control S stem Malfunction Section 14.1.5 of the Turkey Point Units 3 and 4 FSAR, shows that for a boron dilution event the operator has sufficient time to identify the problem and terminate the dilution before the reactor returns critical'or loses shutdown capability. The standard acceptance criteria and FSAR calculated values for operator action are summarized below:

FSAR Acceptance Mode (minutes ) Criteria Refueling >120 30 Startup >240 15 Power

a. Manual Control >15 15

b. Auto Control 21 15 Steam generator tube plugging has no affect on the analysis at refueling conditions since only the reactor vessel volume is assumed active. The coolant loop volume is conservatively assumed stagnant.

For dilution during startup and at power, there is 22

an effect due to the reduction in primary coolant volume. The effective volume of primary coolant in.

the steam generator tubes is assumed to be reduced by 25% (+510 ft 3

). Thus the total volume assumed in the analysis has been reduced from 7800 3 to ft 7290 ft3

. This translates into approximately a 7%

reduction in the originally calculated dilution-time from startup conditions (240 minutes). The result is still significantly greater than the required operator action time, therefore no safety concerns exist.

For dilutions during power operation a highly

-5 conservative reactivity insertion. rate of 1.1*10 5k/sec was assumed in the FSAR consistent with an initial boron concentration of 1200 ppm. FSAR Figure 14.1.5-1 (Reactivity Insertion Rate vs Boron Concentration) has been recalculated consistent with the lower primary value and is shown in Figure

15. The results show that the reactivity insertion rate assumed in the FSAR is still conservative.

Therefore no additional analysis is required. 'It should be noted, however, that the FSAR analysis is still highly conservative with respect to the current cycles since the analysis assumed that only 1% shutdown margin is available. The Turkey Point units have been designed such that >2.5% shutdown margin is always available for BOL conditions. The result is that operator action times would be >70 minutes with a more realistic value.

Thus for all cases it has been shown that steam generator tube plugging up to 25% will not. affect the safety conclusions of the FSAR'or boron dilution.

23

4.0 TECHNICAL SPECIFICATIONS This section contains, the technical content of proposed changes to the Technical Specifications. These changes are consistent with the plant operation necessary for the design and safety evaluation conclusions stated previously to remain valid. This report is contingent upon approval of these changes.

Change the following pages of the Technical Specifications:

1. 2.3-2, overtemperature hT

2. 2.3-3, overpower dT

3. 3.1-7, DNB parameters Add the following figure:

2.1-lb, Reactor Core Thermal and Hydraulic Safety Limits, 3 Loop Operation The reactor core thermal and hydraulic safety limits shown (9) in Figure 2.1-1b are based on a DNB limit of 1.24 and a rod (9) bow penalty of 28.9%. They were generated with the COBRA-IIIC/

MIT code with the same methods described in Section 3.1, N

except that the F<H factor was taken to be F~H

= l. 55 [1 + 0. 2 (1-P) ] for P<1.

24

650 2400 psia 640 630 2250 psia 620 2100 psia 610 600 1900 psia 590 580 570 Note: These curves are applicable with steam generator tube plugging

>19% and <25%.

560 550 540 0 20 40 60 80 100 120 140 RATED POWER (PERCENT)

Figure 2.l-lb. Reactor Core Thermal and Hydraulic Safety Limits,'3 Loop Operation 25

Reactor Coolant Tem erature Overtempera-T < K1 0.0107 (T-574) + 0.000453 -(P 2235) f(Q )]-

I hT Indicated h,T at rated power, F 0

T Average temperature, F Pressurizer pressure, psig f(hq) a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during startup tests such that:

qt - qb) within +10 percent and -'14 percent where t and qb are the percent power in the top aud bottom halves of the core respectively,, and q : +

total core power in percent of rated power, f(~) - 0.

For each percent'hat the.magnitude of (qt qb ) exceeds +10 percent, the Delta>>T trip setpoint shall be automatically reduced by 3.5 percent of its value at interim power.

For each percent that the magnitude of {q qb) exceeds -14 percent, the Delta-T trip setpoint shall be automatically reduced by 2 percent of its value at interim power.

(Three Loop Operation) ~ 1.095+

(Two Loop Operation) ~ 0.88 Kl ~ 1.095 for steam generator tube plugging < 25 percent 2~3 2 26

Over-power AT < AT 1.11

• K1 dT dt K2 (T - T') f (Aq) b,T 0

~ Indicated 5T at rated power, F T Average temperature, F T' Indicated average temperature at nominal conditions and rated power, F Kl ~, 0 for decreasing average temperature, 0.2 sec./F for increasing average tempe'rature K2 0.00069+for T equal to or more than T';

0 for T less than T'T Rate of change of temperature, F/sec dt f(5q) ~ As defined above Pressurizer Low Pressurizer. vressure equal to or greater than 1835 psig.

High Pressurizer pressure equal to or less than 2385 psig.

High Pressurizer water level 'qual to or less than 92% of- full scale.

Reactor Coolant Flow Low reactor coolant flow equal to or greater than 90% of normal indicated flow Low reactor coolant pump motor frequency - equal to or greater than 56.1 Hz

'nder voltage on reactor coolant pump motor bus equal to or greater than 60% of normal voltage Steam Generators Low-low steam generator water level equal. to or greater than 5% of narrow range instrument scale

• This factor is 1.11 for steam generator tube plugging < 15 percent This factor is 1.10 for steam generator tube plugging > 15 percent and < 19  %

Thxs factor is 1.08 for steam generator. tube plugging >19% and

<25%.

This factor is 0.00106 for steam generator tube plugging >19% and <25%.

27 2~3 3

4 t

6~ ~ DNB PARAMETERS The following DNB related parameters limits shall be t

maintained during power operation:

a Reactor Coolant System Tavg < 578.2 F

b. Pressurizer Pressure > 2220 psia+

c. Reactor Coolant Flow > 268,500 gpm+

With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce thermal power to less than 5X of rated thermal power using normal shutdown procedures.

Compliance with a- and b. is demonstrated by verify-ing that each of the parameters is within its limits

. at least once each, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Compliance with c. ip demonstrated by verifying that the parameter is withi'n its limits after each refuel-ing cycle-'

Limit not appli'cable during either a THERMAL POWER ramp increase in excess of (5X) RATED THERMAL POWER per minute or a THERMAL POWER step increase in excess of (10%) RATED THERMAL POWER.

+ Reactor Coolant Flow > 268,500 gpm for steam generator tube plugging

< 15X.

Reactor Coolant Flow > 263,130 gpm for steam generator tube plugging

> 15X and < 19X Reactor Coolant Flow > 255,075 gpm for steam generator tube plugging

> 19X and < 25%.

28 3 0 l 7

5. 0 CONCLUSIONS Based on an analytical and empirical study of the Turkey Point Units 3 and 4 flow characteristics, a 25% steam generator plugging level was conservatively determined to result in a flow reduction 5% below thermal design flow.

The impact of this higher steam generator tube plugging level on the non-LOCA accident analyses applicable to Turkey Point Units 3 and 4 has been assessed. The reevaluation has indicated that the impacts associated with increased steam generator tube plugging can be accommodated with margin'to the safety e limits for the following accident analyses:

Uncontrolled RCCS Mithdrawal From a Subcritical Condition Malpositioning of a Part Length Rod Rod Cluster Control Assembly (RCCS) Drop Startup of an Inactive Reactor Coolant Loop

- Reduction in Feedwater Enthalpy. Incident Excessive Load Increase Incident Loss of Reactor Coolant Flow (Locked Rotor Accident)

Loss of External Electrical Load Loss of Normal Feedwater Loss of AC Power Rupture of a Steam Pipe Rupture of a Control Rod Mechanism Housing 29

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 reanalyzed with the rod bcw penalty included:

Uncontrolled Control Rod Withdrawal at Power Loss of Reactor Coolant Flow (Flow Coastdown Accident)

Chemical and Volume Control System Malfunction Based on these reanalyses, it was determined that the results meet the appropriate safety limits with the conservative Technical Specification changes to the Reactor Core Thermal and Hydraulic Safety Limits and =the overpower 'hT.

E Therefore, it is concluded that the impact, of increased steam generator plugging levels, up to 25%, can be accommodated without adversely affecting the safe operation of. Turkey Point Units 3 and 4.

30

6. O'EFERENCES

1. "Final Safety Analysis Report, Turkey Point Plant, Units 3 and 4", Docket Nos. 50-250, 50-251.

2. "Reload Safety Evaluation, Turkey Point Plant, Unit 4, Cycle 4", January, 1977.

3. "Reload Safety Evaluation, Turkey Point Plant, Unit 4, Cycle 5", June, 1978.

4. K. V. Moore, et al, RETRAN A Program for One Dimensional Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems, Volume 1: Equations and Numerics, EPRI NP-408, January, 1977.

5. "Desgin Basis for the Thermal Overpower hT and Thermal Overtemperature hT Trip Functions", WCAP-8745 (Proprietary) and WCAP-8746 (Non-Proprietary), March, 1977.

6. Safety and Fuel Management Analysis Methods, Florida Power and Light Company, NAD-1483, Volume I, June, 1978.

7. R. C. Kern and D. Hodges, "DYNODE-.P, Version 2:, A Nuclear Steam Supply System Transient Simulator for Pressurized Water Reactors User Manual", NAI-76-67, Revision 5, April 19, 1978.

8. R. W. Bowring and P. Moreno, COBRA-IIIC/MIT Computer Code Manual, March, 1976. (Available. from Electric Power Research Institute.)

9. Margins in Turkey Point Units 364 Safety Analysis to Offset the Effects of Fuel Rod Bowing, FPL Report NAD-QR-25, March, 1977.

31

TABLE I REACTOR COOLANT FLOW MEASUREMENT UNCERTAINTY Parameter Uncertaint RC Loo Flow Uncertaint Feedwater Flow +1.25% +1.25%

Feedwater Temp. +1.0 F +0. 2%

Steam Pressure +30 psi +0. -l%

hot +0 5F +0. 9%

cold +0.50F +0.9%

RCS Pressure +50 psi +0.2%

Total +1.8%*

• RMS uncertainty 32

TABLE II

SUMMARY

OF ROD EJECTION ANALYSIS PARAMETERS AND RESULTS BOL BOL EOL EOL Power level, 102 0 102 0 Ejected rod worth, %6k .35 .71 .30 .84 Delayed neutron fraction, .50 .50 .44 .44 Feedback reactivity weighting 1.20 1.30 1.2 2.6 Trip rod shutdown, Sdk 4 F before rod ejection 2.55 2 55 F after rod ejection t

5.48 7.0 5.52 14.3 Number" of operating pumps 3 2 Initial fuel temperature, 'F 2660, 547 2475 " 547 Maximum fuel pellet average temperature, 'F 3990 2040 3490 3080 Maximum clad average temperature, oF 2370 1580 2125 2270 Fuel pellet melting, 0 0 0 0 33

lg TABL'E III BT PROTECTION SETPOINTS OVERTEMPERATURE hT' TURKEY POINT TECHNICAL SPECIFICATION SECTION 2 SETPOINT < T [Kl A (T-574) + B (P-2235) f (hq) ]

FSAR ANALYSIS CURRENT LIMIT RECALCULATED VALUES VALUES (19% S.G. PLUGGING) (25% S.G. PLUGGING)

Kl 1.095 1.08 1.18

.0107 .0107 .00959

.000453 .000453 .000420 OVERPOWER bT TURKEY POINT TECHNICAL SPECIFICATION SEC%'ION 2 SETPOINT < T [K -K o o K2 (T-T')

dT ldt f(hq)]

K 0 1.11 1.10 1.08 Kl 0.2 0.2 0.'

K2 .00068 . 00068 .00106 34

TABLE IV UNCONTROLLED ROD WITHDRAWAL 25% STEAM GENERATOR TUBES PLUGGED MINIMUM DNB RATIO Rod Bow Penalt Included .

Slow Rod Withdrawal From Full Power 1.33 Fast Rod Withdrawal From Full Power 1.'4 Slow Rod Withdrawal From 60% Power 1.43

l. 43 35

FIGURE 1 TURKEY POINT UNITS 3'AND 4 RCS FLOW YS STEAM GENERATOR TUBE PLUGGING 1.00000' CD CD 90000 89500 85025.

80000 0 10 20 30 40 STEAM GENERATOR TUBE PLUGGING (X) tA .Best Estimate Calculated RCS Flow B Curve A Minus 4200 GPM: Lowest Expected Measurement C Curve B Ti'mes 0.98 0 Curve B Times 0.97 36

FIGURE 2 UNCONTROLLED RCCA WITHDRAWAL LOW FLOW ANALYSIS FAST REACTIVITY INSERTION AT FULL POWER (FRFP)

NUCLEAR POWER 1.4 0g C4 1.2 1.0 0

R

.8 u

.6

~j .4

.2 1.0 2.0 3.0 4.0 5.0 6.0 TIME, SECONDS 37

FIGURE 3 UNCONTROLLED RCCA WITHDRAWAL LOW FLOW ANALYSIS FAST REACTIVITY INSERTION AT FULL POWER (FRFP)

VESSEL AVERAGE TEMPERATURE 580 579 578 577 576 .

0 1.0 2.0 3.0 4.0 5.0 6.0 TIME, SECONDS 38

FIGURE 4 UNCONTROLLED RCCA WITHDRAWAL LOW FLOW ANALYSIS FAST REACTIVITY INSERTION AT FULL POWER (FRFP)

PRESSURE 2260 2240 2220

~i 2200 TIME, SECONDS 39

FIGURE 5 UNCONTROLLED RCCA WITHDRAWAL ACCIDENT, LOW FLOW ANALYSIS FAST REACTIVITY INSERTION AT FULL POWER,(FRFP)

MINIMUM DNBR VS TIME 1.0 2.0 3.0 4.0 5.0 6.0 TIME, SECONDS 40

FIGURE 6 UNCONTROLLED RCCA WITHDRAWAL LOW FLOW ANALYSIS SLOW REACTIVITY INSERTION AT FULL POWER (SRFP)

NUCLEAR POWER 1.2 1.0

."..8 g0

.6

.2, 0

10 20 . '0 40 50 60 TIME, SECONDS 41

'1 FIGURE 7 UNCONTROLLED RCCA WITHDRAWAL LOW FLOW ANALYSIS SLOW REACTIVITY INSERTION AT FULL POWER (SRFP)

VESSEL AVERAGE TEMPERATURE 586 582 0

gg 578 574 570 566 10 20 30 40 50 60 TIME, SECONDS 42

e 4

FIGURE 8 UNCONTROLLED RCCA WITHDRAWAL LOW FLOW ANALYSIS SLOW REACTIVITY INSERTION AT FULL POWER (SRFP)

PRESSURE 2300 2200 2100 2000 10 ,20 30 40 50 60 TIME~ SECONDS 43

FIGURE 9 UNCONTROLLED RCCA WITHDRAWAL LOW FLOW ANALYSIS SLOW REACTIVITY INSERTION AT FULL POWER (SRFP)

DNBR VS TIME 10 20 30 40 50 60 TIME, SECONDS 44

t FIGURE 10 LOSS OF FLOW (3 PUMP COASTDOWN) ACCIDENT LOW FLOW ANALYSIS CORE FLOW VS TIME

1. 0-p pH .6 u

4 8

p O

.2 1.0 2.0 3.0 4.0 5.0 TIME,, SECONDS 45

T FIGURE 11 LOSS OF FLOW (3 PUMP COASTDOWN)

LOW FLOW ANALYSIS NUCLEAR POWER VS TIME 1.0 2.0 3.0 4 0 5.0 TIME, SECONDS 46

FIGURE 12 LOSS OF FLOW (3 PUMP COASTDOWN)

LOW FLOW ANALYSIS HEAT FLUX VS TIME 1.0 2.0 3.0 4.0 5.0 TIME, SECONDS 47

C' FIGURE 13 LOSS OF FLOW ACCIDENT (3 PUMP COASTDOWN)

LOW FLOW ANALYSIS HOT CHANNEL HEAT FLUX VS TIME 1.0

.8 R

0 u .6 4

.2 0

TIME, SECONDS

FIGURE 14 LOSS OF FLOW ACCIDENT (3 PUMP COASTDOWN)

LOW FLOW ANALYSIS MINIMUM DNBR VS TIME 2.40

2. 00; H

g 1.60

1. 20 0 1.0 ~ 2.0 3.0 4.0 5.0 6.0 TIME, SECONDS 49

I (~

2.0 ~ ~ ~

Variation in Reactivity Inser tion Rate with Initial Boron Concentration for a Boron Dilution Rate of 230 GPH Q

Cl Q

A Cl cS 8 1.0 4

M

~

W 0

0 1000 2000 2500 Initial Boron Concentration, PPM Figure 15