ML20205N391
| ML20205N391 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 10/31/1988 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20205N389 | List: |
| References | |
| NUDOCS 8811040125 | |
| Download: ML20205N391 (42) | |
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i SAFETY EVALUATION OPERATION OF THE CALLAWAY PLANT WITH A POSITIVE MODERATOR TENPERATURE COEFFICIENT, INCREASED RWST/ ACCUMULATOR BORON CONCENTRATIONS, AND INCREASED SAT SODIUM HYDROXIDE CONCENTRATION October 1988 WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230
- utonh?il 31$$g3 P
6 i
Table of Contents Section 11111 EARt 1.
Introduction 1
2.
Positive MTC Safety Evaluation 2
2.1 Introduction 2
2.2 Transtants Not Affecced by a Positive MTC 5
2.3 Transients Sensitive to a Positive MTC 13 f
2.4 Sumary and Conclusions 22 3.
RWST/Acr.umulator Boron Concentration Increase f
Safety Evaluation 23 i'
3.1 Introduction 23 I
3.2 Non-LOCA Analyses 24 3.3 LOCA Analyses 31 3.4 LOCA Related Design Considerations 33 3.5 Sumary and Conclusions 37
(
4.
Balance of Plant Considerations 38 l
5.
References 40 1
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page 1 SECTION 1 INTRODUCTION l
The following information provides the safety evaluation for l
operation of the Callaway Plant with a positive moderator temperature coefficient.
In conjunction with this change, the boron i
concentrations in the Refueling Water Storage Tank and in the Safety Injection accumulators have been increased.
Section 2 is the evaluation of potential areas that relate to a positive moderator temperature coefficient. Section 3 is the evaluation of potential areas that relate to operation at an increased RWST/ Accumulator boron concentration.
Section 4 addresses the proposed changes as they may l
affect other balance of plant areas, including the increased sodium hydroxide concentration in the Spray Additive Tank.
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a page 2 SECTION 2 POSITIVE MTC SAFETY EVALUATION 2.1 Introduction This safety evaluation has been performed to support the proposed Technical Specification change for the Callaway Plant which would allow a positive moderator temperature coefficient (MTC) to exist at the beginning of cycle life (BOL) for power levels below 70% nominal rated thermal power, with the maximum positive value decreasing linearly to zero between 70% nominal power and 100% nominal power.
This evaluation is performed with respect to analyses documented in References 1, 2 and 3.
The results of this evaluation show that the proposed change can be accommodated such that the Reference 1, 2 and 3 results and conclusions are confirmed or are otherwise superceded by this report.
The present Callaway Plant Technical Specifications require the MTC to bi zero or negative at all tinies while the reactor is critical.
This requirement is overly restrictive, since a small positive MTC at reduced power levels would have a minor effect on the FSAR accident analyses. Safety analysis justification of a positive MTC supports reductions in fuel cycle costs by reducing burnable poison inventory, particularly for long cycles which require a large number of burnable poisons to control MTC at the beginning of cycle life.
The proposed MTC Technciai Specification change allows a
+5 pcm/'F MTC below 70% of nominal rated thermal power at BOL.
At 70% power the coefficient begins to decrease linearly from +5 to 0 pcm/'F at 100% nominal rated thermal power.
A power level dependent MTC was chosen to minimize the effect of the Technical Specification change on postulated accidents at high power levels.
Moreover, as the power level is raised, the average coolant temperature becomes higher as allowed by the plant programmed average temperature controller, tending to make the MTC more negative. Also, the boron concentration can be reduced as xenon builds into the
page 3 core. Thus, there is less need to allow a positive MTC as full power is approached. As fuel burnup is achieved, boron is further reduced and the MTC will eventually become negative over the entire operating power range.
The effect of a positive MTC has been evaluated for the following Callaway Plant non LOCA and LOCA safety analysis design bases.
Non-LOCA Safety Analysis Design Bases:
1.
FSAR Chapter 15 - Reference 1
- 11. FSAR Chapter 6 Steamline Break Mass / Energy Releases Inside Containment - Reference 1 iii. Steamline Break Mass / Energy Releases Outside Containment Reference 2 iv. Non-LOCA Analyses Supporting the Implementation of the Steam Generator low Low Level Reactor Trip Time Delay and Environmental Allowance Modifier in the Callaway Plant -
Reference 3 LOCA Safety Analysis Design Bases:
1.
FSAR Chapter 15 - Reference 1
- 11. Rod Ejection Mass Releases for Dose Calculations 111. Reactor Vessel and Loop Blowdown Forces iv. Post-LOCA Long term Core Suberiticality Requirements
page 4 Those incidents which were found to be sensitive to positive or near-zero HTCs have been analyzed; results of these analyses are documented in References 1, 2, 3 and in this report.
These transients are discussed specifically with respect to positive HTC in this report.
In general, these incidents are limited to transients which cause the reactor coolant temperature to increase.
Accidents which do not require analysis for a positive HTC include those resulting in excessive heat removal from the Reactor Coolant System; in these cases a large negative HTC assumption produces more limiting results. Also unaffected are those transients for which heatup effects following reactor trip are not sensitive to the HTC.
Section 2.3 discusses the design basis calculations which are sensitive to the Technical Specification for maximum allowable positive HTC at a given power level.
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page 5 2.2 Transients Not Affected by a Positive MTC Analysis assumptions for the following transients are not affected by a positive HTC since these transients either result in a reduction in Reactor Coolant System (RCS) temperature, and are therefore sensitive to a negative HTC, or are otherwise negligibly affected by a positive MTC.
Feedwater Systen Malfunctions that Result in a Decrease in Feedwater Temoerature and an Increase in Feedwater flow (FSAR Sections 15.1.1 and 15.1.2)
The addition of excessive feedwater or the reduction of feedwater temperature are excessive heat removal incidents and result in a reduction in primary coolant temperature.
Thus, they are most sensitive to a negative MTC. Results presented in the FSAR for these Condition II accidents are based on a negative MTC and represent the limiting cases. Therefore, the FSAR analysis assumptions and conclusions are not affected by the proposed MTC Technical Specification limit.
Excessive Increase in Secondary Steam Flow (FSAR Section 15.1.3)
An excessive load increase incident, in which steam load exaneds core power, result 3 in a decrease in RCS temperature. With the reactor in manual control, the analysis presented in the FSAR for this Condition
!! accident shows that the limiting case assumes a large negative MTC.
If the reactor is in automatic control, the control rods are withdrawn to increase power and restore the average temperature to the programmed value. The FSAR analyses assume a large and a small negative MTC, respectively.
The FSAR results show that the minimum DNBR is not sensitive to MTC. Therefore, reanalysis of this accident with positive MTC assumptions is not necessary and the FSAR conclusions remain valid for the proposed HTC Technical Specification limit.
page 6 Inadvertent Ooenina of a Steam Generator Relief or Safety Valve and Steam System Pinina Failure -(FSAR Sections 15.1.4 and 15.1.5)
Since the rupture of a main steam line is a temperature reduction transient, minimum core shutdown margin is associated with a large negative MTC. The worst conditions for-this Condition IV accident are, therefore, those anlayzed in the FSAR.
This is also true for the accidental depressurization of the Main Steam System (Condition III accident) which is bounded by the steam system piping failure.
Therefore, the FSAR analysis assumptions and conclusions remain valid for the proposed MTC Technical Specification limit.
Startuo of an Inactive Reactor Coolant Pumo at an incorrect Temoerature (FSAR Section 15.4.4)
An inadvertent startup of an idle Reactor Coolant Pump, as postulated in the FSAR, is a Condition II accident which results in a decrease in core average temperature. As the most negative MTCs produce the greatest reactivity addition, the analysis reported in the FSAR represents the limiting case.
Therefore, the FSAR analysis assumptions and conclusions are not affected by the proposed MTC Technical Specification limit.
Inadvertent loadino and Ooeration with a Fuel Assembly in Imoroper_
Position (FSAR Section 15.4.7)
This Condition III event addresses the possi*ility and consequences of one or more fuel pellets having the wrong enrichment or the loading of a fuel assembly without the prescibed amount of burnable poisons. The FSAR concludes that any significant perturbation from the intended core inventory would be detectable due to the resulting effects on power distribution.
The MTC does not affect the ability of core instrumentation to detect unexpected power shapes.
Additionally, the analyses performed for the FSAR are not dependent on assumptions related to MTC.
Therefore, the FSAR analysis assumptions and conclusions are not affected by the proposed MTC Technical Specification limit.
page 7 Inadvertent Ooeration of the ECCS Durina Power Ooeration (FSARSection15.5.1)
The inadvertent actuation of the ECCS is a Condition II accident which results in a reduction in power, temperature and pressure, as shown in the FSAR. Since temperature is decreasing, a positive MTC will result in the addition of negative reactivity. This will cause power to decrease even more rapidly and will, therefore, be a DNBR benefit. Therefore, the FSAR analysis assumptions and conclusions are not affected by the proposed MTC Technical Specification limit.
CVCS Malfunction that Increases Reactor Coolant Inventory (FSAR Section 15.5.2)
The FSAR CVCS malfunction accident is a Condition Il event which results in a negative reactivity excursion d e to injected boron,
- h. ctor power decreases and the resulting primary to secondary power mismatch causes a drop in vessel average temperature. Although the FSAR analysis assumes a least negative MTC, the FSAR documents that this transient is relatively insens;tiva to a change from minimum to maximum reactivity feedback assumptions.
Therefore, the FSAR analysis assumptions and conclusions are not significantly affected by the proposed MTC Technical Specification limit.
Steam Generator Tube Ruoture - SbTR (FSAR Section 15.6 S)
SGTR - Stuck ARY Scenario (SLNRC-86-01, 1-8 86)
In this SGTR scenario, a single tube rupture leads to a slow depressurization of the RCS and subsequent reactor trip.
Prior to trip, there is no change in RCS temperature and, therefore, retetivity feedback is not important. After reactor trip, the fatIted SG atmospheric relief valve (ARV) continues to relieve steam for 20 minutes.
Since the post-trip primary side conditions are a funttion of decay heat and safety injection flow, the analysis conciusions are not affected by the use of positive MTC.
page 8 SGTR - Overfill Scenario (VLNRC-1518, 5-27 87)
In this SGTR scenario, a reactor trip is assumed to occur coincident with SGTR. Again, spice the post-trip primary side conditions are a function of decay heat Ond safety injection flow, there factors are unaffected by positive HTL and the analysis conclusions remain valid.
Mass and Enerav Release Analysis for Postulated Secondary Pioe Ruotures Inside Containment (FSAR Section 6.2.1.4) and Steamline Break Mass and Enerav Releases for Ecuioment Environmental Qualification Outside Containment (WCAP-10961-P)
As described for the limiting steamline break core response transients (FSAR Sections 15.1.4 and 15.1.5), a reduction in primary RCS temperature is characteristic of a steamline break. Whether analyzing for minimum DNBR or maximum mass and energy releases, conservative transient assumptions minimize core shutdown marg %.
Therefore, the limiting steamline break analysis assumptions for mass and energy release include a large negative HTC. Therefore, the FSAR and WCAP-10961 P, analysis assumptions and conclusions remain valid for the proposed HTC Technical Specification limit.
Loss of Coolant Accidents Resultina from a Soectrum of Postulated -
Pioe Breaks within the Reactor Coolant Pressure Boundary (FSAR Section 15.6.5) a) Small Break LOCA In the standard small break LOCA analysis methodology, core kinetics calculations are not explicitly performed.
Instead, the core power is maintained at initial conditions until the reactor trip setpoint is reached and a delay time has passed.
The delay accounts for signal processing and the time it takes for the rods to reach the bottom of the core.
This delay results in the generation of an additional few full power-seconds of heat by not accounting for partial rod worth while the rods are falling into the core and the
page 9 shutdown effect of voiding for a core with a negative HTC. After trip, the power is calculated by interpolation from a table of power versus time which has been conservatively derived to envelope all plants. This generic power decay curve is composed of three parts:
residual fission heat, fission product decay, and actinide gamma decay.
The residual fission term is based upon the exponential decay of the fission power for a low shutdown margin and with the core full of hot water. This is conservative for essentially all small break LOCAs, since some net voiding occurs coincident with reactor trip due to the sudden depressurization.
An evaluation has determined that the excess core power generation which may be expected from explicitly modeling a positive HTC of 0
+5 pcm/ F would be much less than 1 full power-second. This excess power could impact the transient in two areas; 1) the effect of increased power on the time of reactor trip and safety injection (SI) initiation signals, and 2) the influence of increased heat generation on peak clad temperature (PCT). Applicable calculations and sensitivity studies confirm that a positive HTC has only a small, third order effect on PCT.
The smal) core power excursion induced in the initial few seconds of the trar.sient will slow the depressurization negligibly, delaying reactor trip and SI initiation only slightly. These delays, plus the small excess power, in turn have a small influence on loop seal clearing and subsequently the core bolloff uncovery transient, hundreds of seconds into the accident, during which time PCT occurs. However, there will be virtually no direct influence on decay heat generation during the clad temperature excursion.
The existing analysis of record, performed with the Westinghouse small break LOCA code, NOTRUMP, shows that a very substantial margin exists between the results calculated in the analysis of record and 0
the 10CFR50.46 limit of 2200 F.
Any impact from operation with a positive HTC is very small compared to this margin. Thus 0
implementation of a positive HTC of +5 pcm/ F fro:n 0% power to 70%
0 power and a linear ramp down to O pcm/ F at 100% power does not I
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page 10 alter the conclusions of the FSAR small break LOCA analysis, and meets the acceptance criteria in 10CFR50.46.
b) Large Break LOCA For large break LOCA analyset., a positive MTC currently can only be modeled during blowdown. Once voids form during blowdown, the negative reactivity added by the decrease in moderator density and the increase in neutron leakage from the core is substantial compared-to the reactivity added during the brief period during blowdown where positive reactivity feedback occurs. During refill and reflood, the core will remain shutdown due to the voids, and the effects of a positive MTC will be negligible. On this basis, PMTC is currently not modeled during the reflood and refill phases of a LOCA.
Therefore, the implementation of a positive MTC will not alter the results or conclusions of the FSAR large break LOCA analysis and the requirements of 10CFR50.46 are met.
c) Post-LOCA Long Term Core Cooling An increased boron concentration in the RWST is required to offset the increased core reactivity due to the positive MTC which has been evaluated in Section 3.3 of this report. The increased boron concentration will add negative reactivity to offset that added by the positive MTC. Confirmation that the boron concentration increase will provide enough margin to keep the core suberitical for the post-LOCA long-term core cooling is provided through the normal Reload Safety Analysis Checklist (RSAC) evaluation process.
page 11 Looo and Vessel LOCA Hydraulic Forcina Functions (FSARSection 3.6.2.2.1.5 and Section 3.9(N).2.5)
The primary factors affecting LOCA forces analyses are pressure, temperature and density. A positive MTC c.ould potentially result in an increase in RCS pressure or temperature resulting in changes in the fluid density. However, the transient is essentially over with before any feedback from a positive MTC, which could alter the results of the forces analysis, could occur. Therefore, the proposed positive MTC will not alter the results of the LOCA forces analysis.
Containment Intearity (FSAR Section 6.2)
For containment integrity analyses, like the large break LOCA, a positive HTC can currently only be modeled during the blowdown phase of the LOCA mass and energy release analysis and in the steamline break mass and energy release analysis.
The LOCA mass and energy release analysis was performed based upon operation at 100% power.
The proposed positive MTC will not affect the analysis since the positive MTC is zero at 100% power. The steamline break mass and energy release analyses are conservatively modeled using a negative HTC. The negative HTC is conservative because the steamline break transient is a cooldown of the RCS. Therefore it is concluded that the implementation of a positive MTC will have a negligible effect on the containment integrity analyses.
Hotleo Switchover to Prevent Boron Precipitation (FSAR Sections 6.3.2 and 15.6.5)
The implementation of a positive MTC requires that the RWST boron concentration be increased. The results of increasing the boron concentration have been evaluated in Section 3.3 of this report.
A positive MTC affects the time for hot leg switchover and the switchover time has been recalculated to reflect the increased boron I
concentration. The implementation of positive MTC is acceptable with respect to the new time determined for hot log switchover.
{
page 12 Rod Eiection Mass Releases for Dose Calculations (FSAR Section 15.4.8 and Figure 15.4-27)
The rod ejection mass release analysis is sistlar to the small break LOCA analysis and those same arguments presented for the small break apply here.
The effect of the positive MTC will be negligible and will not alter the FSAR results.
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l page 13 l
2.3 Transients Sensitive to a Positive MTC 1
l Loss of External Electrical load and/or Turbine Trio (FSAR Sections 15.2.2 and 15.2.3) 1 l
Two turbine trip cases, analyzed at both beginning and end of life (BOL, EOL) conditions, are presented in the FSAR for this Condition t
!! accident. One case assumes full credit for the operation of the pressurizer spray and the pressurizer power operated relief valves; the other case assumes no pressurizer control.
EOL conditions assume maximum reactivity feedback. This iacludes a iarge negative MTC.
For BOL conditions, the two turbine trip cases assume minimum reactivity feedback. The analysis assumption for MTC is a constant
+ 5 pery" F.
l The result of a loss of load is a core power level which momentarily i
l l
exceeds secondary system power removal causing an increase in core water temperature.
The consequences of the reactivity addition due l
to a positive MTC are small increases in both peak nuclear power and pressurizer pressura.
For both BOL cases (with and without pressurizer control), the transient is mitigated by reactor trip on i
I high pressurizer pressure. The FSAR results illustrate that RCS l
pressure is maintained below 110% of the design value. The steam
[
generator safety valves open and limit the secondary steam pressure increase and there is margin to the DNBR safety analysis limits.
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Loss of Non emeraency AC Power to the Station Auxiliaries and i
Loss of Normal Feedwater Flow f
(FSAR Sections 15.2.6 and 15.2.7 and WCAP-ll884)
Loss of offsite power and loss of normal feedwater are ANS Condition
[
!! accidents analyzed to demonstrate the ability of the auxiliary feedwater system to remove decay heat from the RCS.
Following initiation of these incidents, the RCS tesperature rises prior to
[
reactor trip due to reduced heat transfer in the steam generators.
Thus, the assumption or s positive HTC results in a reactivity
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page 14 insertion and resultant increase in core power prior to reactor trip. This in turn increases the amount of heat that must be removed following reactor trip, resulting in a more severe transient.
A constant MTC of +5 pcm/'F was assumed for these analyses reported in the FSAR.
The FSAR results show that the capacity of the auxiliary feedwater system is adequate to provide sufficient heat removal from the RCS.
The pressurizer does not fill with water, assuring that the integrity of the core is not adversely affected.
For the case without offsite power, the results verify the natural circulation capacity of the RCS to provide sufficient heat removal capability to prevent overpressure and fuel or clad damage following Reactor Coolant Pump coastdown.
Therefore, the FSAR analysis assumptions and conclusions remain valid for the proposed MTC Technical Specification limit.
These events have also been analyzed for the Callaway Plant to support implementation of the Steam Generator Low Low level Reactor Trip Time Delay (TTD) and Environmental Allowance Modifier. The concepts for these protection system modifications and the supporting safety analysis methodology are documented in NRC approved topical reports (References 4 and 5).
Loss of offsite power and loss of normal feedwater analyses were performed, consistent with the approved methodology, to provide a revised safety analysis limit for the steam generator loy low level trip setpoint and to generate new safety analysis limits for programmed trip time delays at specified part-power levels. Analysis results are reported in Reference 3.
As documented in Reference 3, the Callaway Plant TTD design is consistent with the Reference 4 approved conceptual design.
Consistent with the approved conceptual design and supporting safety analysis methodology, the Reference 3 part-power loss of normal feedwater analyses support a positive MTC for power levels less than 20% of rated thermal power fc-h the TTD imposes a programmed delay on the steam generator vel protection functions.
page 15 The full power loss of offsite power and loss of normal feedweter FSAR transients were reanalyzed (Reference 3) to lower the safety analysis limit for the steam generator low-low level trip setpoint to 0% of span.
The Reference 3 reanalysis assumption for HTC at full power is O pcm/'F.
This was verified to be more limiting than reduced power and higher positive HTCs within the proposed Technical Specification limit.
Therefore, the Reference 3 analysis assumptions and conclusions remain valid for the proposed HTC Technical
~
Specification limit.
Feedwater System Pioe Break (FSAR Section 15.2.8 and WCAP ll884)
Hain feedwater system pipe break is an ANS Condition IV accident i
analyzed to demonstrate the ability of the secondary system auxiliary feedwater to remove decay heat from the RCS.
Following initiation of the incident, RCS temperature rises prior to reactor trip due to reduced heat transfer in the steam generators. Thus, the assumption of a positive MTC results in a reactivity insertion and results in an increase in core power prior to reactor trip. This in turn increases l
the amount of heat that must be removed following reactor trip, resulting in a more severe transient.
A constant HTC of +5 pcm/'F was assumed for the FSAR transients, 1
j These transients, performed for cases with and without offsite power available, assume full power initial conditions. The FSAR transient results show, for the case with offsite power, that the capacity of the auxiliary feedwater system is adequate to provide sufficient heat l
removal from the RCS to prevent overpressure or core uncovery.
For the case without offsite power, the results verify the natural circulation capacity of the RCS to provide sufficient heat removal capability to prevent overpressure and fuel or clad damage following j
Reactor Coolant Pump coastdown.
l As reported in Reference 3, a part-power feedline break transient was l
analyzed to verify the acceptability of the TTD for the Callaway Plant.
The analysis assumed a constant positive HTC of +5 pcm/'F. All applicable safety analysis acceptance criteria are
page 16 met for this part-power case and the current FSAR full power cases are verified to be limiting. The Reference 3 and FSAR analysis assumptions and conclusions rr. main valid for the proposed MTC Technical Specification limit.
Partial and Comolete loss of Forced Reactor Coolant Flow (FSAR Sections 15.3.1 and 15.3.2)
The incidents analyzed for the FSAR are loss of two Reactor Coolant Pumps (Condition !! event) and the more severe complete loss of forced reactor coolant flow (Condition !!! event).
The imediate effect of loss of coolant flow is a temperature increase.
The transients are analyzed to demonstrate the reactor protection system response to terminate the temperature increase and thus prevent DNB and subsequent fuel damage. The effect of a positive MTC is limited to the initial part of the transient, prior to reactor trip.
Both the complete and partial loss of flow analyses reported in the FSAR assume a constant MTC of +5 pcm/'F.
The FSAR results illustrate that all applicable safety analysis acceptance criteria are met for the partial and complete loss of flow transients assuming a +5 pcm/*F MTC.
Therefore, the FSAR analysis assumptions and FSAR conclusions remain valid for the proposed MTC Technical Specification limit.
Reactor Coolant Pumo Shaft Seizure flocked Rotor) or Break (FSAR Sections 15.3.3 and 15.3.4)
The FSAR presents the analysis and results for an instantaneous seizure of a Reactor Coolant Pump (RCP) rotor at full power with four loops operating (Section 15.3.3). The consequences of this Condition
!Y event bound those of the RCP shaft break (Sectici 15.3.4). During a locked rstor event, RCS temperature rises until shortly after reactor trip. A positive MTC will not affect the time to ONB since DNB is conservatively assumed to occur at the beginning of the transient. A positive MTC assumptior, however, may affect the
page 17 nuclear power transient and thus peak RCS pressure, peak fuel clad temperatures and calculated percentage of fuel rods in DNB.
The FSAR results for peak RCS pressure and fuel clad temperatures reflect the analysis assumptions of full power initial conditions and a constant positive HTC of +5 pcm/'F.
This case conservatively illustrates the impact on the transient of a positive HTC, since the HTC will actually be zero or negative at full power. The calculation of percentage of rods in DNB assumes full power initial conditions and a O pcm/'F HTC.
This was verified to be more limiting than reduced power and higher positive HTCs within the proposed Technical Specification limit.
With these assumptions, the FSAR analyses demonstrate that the peak RCS pressure reached during the transient is less than that which would cause stresser to exceed faulted condition stress limits and the peak clad temperature is less than 2700*F. The calculated percentage of fuel rods in DNB is within the safety analysis limit of 5%.
Therefore, the FSAR analysis assumptions and conclusions remain valid for the proposed HTC Technical Specification limit.
Uncontrolled RCCA Bank Withdrawal From a Sui. critical or low Power Startuo Condition (FSAR Section 15.4.1)
A control rod assembly withdrawal incident when the reactor is suberitical is a Condition !! event which results in an uncontrolled additian of reactivity leading to a power excursion.
Reactor trip is initiated due to a high neutron flux low setpoint signal.
The nuclear sower response is characterized by a very fast rise terminated by the reactivity feedback of the negative fuel temperature (i.e. Doppler) coef ficient.
The poler excursion causes a heatup of the moderator and fum1. The reactivity addition due to a positive HTC results in increases in peak heat flux and peak fuel and clad temperatures.
page 18 The FSAR analys b 2... m a +5 pcm/*F HTC at zero power nominal average temperature which becomes less positive for higher temperatures. This is necessary since the TWINKLE computer code, on which the analysis is based, is a diffusion theory code rather than a point kinetics approximation and the moderator temperature feedback cannot be artificially held constant with temperature.
With this assumption, the FSAR concludes that in the event of an RCCA withdrawal accident from a subcritical condition, the core and the RCS are not adversely affected since the minimum DNBR remains above the safety analysis limits.
Uncontrolled RCCA Bank Withdrawal at Power (FSAR Section 15.4.2)
I An uncontrolled control rod assembly withdrawal at power is a Condition II event which produces a mismatch in steam flow and core power resulting in an increase in reactor coolant temperature. A positive HTC would contribute to the power mismatch and potentially reduce the margin to DNB. The FSAR transient is analyzed for a range l
of power levels and reactivity insertion rates assuming minimum and
{
maximum reactivity feedback.
The FSAR minimum feedback cases were performed assuming a constant positive HTC of +5 pcm/*F.
This is a conservative assumption for the full power case since the proposed Technnical Specification limits the maximum HTC at full power to be zero or less. The most limiting case in terms of minimum DNBR is a full power transient assuming minimum feedback.
For all cases it is shown that the applicable safety analysis acceptance criteria are met.
Therefore, the FSAR analysis assumptions and FSAR conclusions remain valid for the proposed HTC Technical Specification limit.
RCCA Hisoperation (FSAR Section 15.4.3)
The RCCA drop case presented in the FSAR is affected by a positive HTC.
The limiting case for this transient assumes the reactor is operating at 100% of nominal power and is in automatic rod control.
The dropped RCCA is assumed to be of small enough worth to not cause a reactor trip. While the HTC is still required to be less than or
1 page 19
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equal to o pcm/'F at 100% power, it may not be as negative relative to a core with a 0 pcm/*F limit at 0% power. This transient has been reanalyzed using MTC input at hot full power j
consistent with the allowed positive MTC at reduced powers and the i
results demonstrate that the DNBR limits are still met for this transient and that the conclusions in the FSAR remain valid.
l CVCS Malfunction that Results in a Decrease in Boron Concentration in l
the Reactor Coolant (FSAR Section 15.4.6) l Analysis of inadvertent boron dilution is primarily dependent on RCS f
volume assumptions, dilution flow rates. Technical Specification j
minimum shutdown margin requirements, assumed protection features.
l boron concentrations and boren worths.
Core designs utilizing the j
flexibility of positive MTCs are likely to predict higher values for critical boron concentrations. Changes to the magnitudes of boron concentrations and the minimum change in boron concentration associated with the loss of shutdown margin may require the reanalysis of this FSAR Condition !! transient to assure that the NRC Standard Review Plan minimum time requirements are available for automatic or operator mitigation of the transient prior to loss of shutdown margin.
The Callaway Plant FSAR Section 15.4.6 boron dilution transient has been reanalyzed for this report.
New analysis assumptions, in consideration of the proposed Techncial Specification MTC limit, are made for the initial and final boron concentrations for the Modes 1, 2, 3, 4 and 5 transients.
Boron dilution in Mode 6, as stated in the FSAR, is assumed to be administratively precluded. The results of these calculations are documented in the marked up pages in of this submittal.
It is concluded that the applicable safety analysis acceptance criteria continue to be met in consideration of the proposed MTC Technical Specification limit.
The continued validit,v of the new boron concentration assumptions is checked for future cycles as part of the reload verification process.
page 20 Soectrum of RCCA E.iection Accidents (Rod Eiection)
(FSAR Section li 4.8)
The FSAR rod ejection transient is a Condition IV event which is analyzed at hot full power and hot zero power for both BOL and EOL conditions in the Callaway Plant FSAR. Since the moderator temperature coefficient is zero or negative at EOL, only the BOL casos are affected by a positive MTC.
The high nuclear power levels and hot spot fuel temperatures resulting from a rod ejection may be increased by a positive MTC. The FSAR analysis assumes a +5 pcm/'F MTC at zero power nominal average temperature which becomes less positive for tagher temperatures. This is necessary since the TWINKLE compute' code, on which the analysis is based, is a diffusion theory code rathat than a point kinetics approximation and the moderator temperature feedback cannot be artificially held constant with temperature.
i The analysis results indicate that the acceptance criteria described in Section 15.4.8 of the FSAR are met. These criteria ensure that i
there is no danger of sudden fuel dispersal into the coolant or i
consequential damage to the primary coolant loop. Therefore, the FSAR analysis assumptions and conclusions remain valid for the proposed MTC Technical Specification limit.
Inadvertent Openina of a Pressurizer Safety or Relief Valve (FSARSection15.6.1) l The inadvertent opening of a pressurizer safety or relief valve is a Condition 11 event which causes an accidental depressurization of the RCS, Accidental depressurization causes a reduction in the RCS density. A positive MTC can also be considered as a negative density coefficient. As such, the density reduction due to the RCS l
depressurization causes a positive reactivity insertion and an increase in nuclear power. This may result in a DNBR penalty.
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page 21 Full power initial conditions and a constant MTC uf +5 pcm/'F are l
assumed in the FSAR transient. The positive MTC causes poser and I
temperature to increase as pressure decreases until a reactor trip j
occurs on an overtemperature delta T trip signal.
The ONBR decreases
(
until the reactor trip occurs and then increases.
All applicable safety analysis acceptance criteria are shown to be met for the FSAR
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Therefore, the FSAR analysis assumptions and conclusions remain valid for the proposed MTC Technical Specification limit.
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page 22 l
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l 2.4 Sumary and Conclusions i
In order to assess the effect on the accident analyses of plant t
operation with a positive MTC, a safety evaluation has been performed
[
[
on the non LOCA and LOCA design basis calculations. Transients
(
sensitive to a positive MTC are identified to be; l
i I
Loss of External Electrical load and/or Turbine Trip, i
Loss of Non emergency AC Power to the $tation Auxiliaries.
[
l Loss of Normal Feedwater Flow, l
l Feedwater System Pipe Break, l
}
Partial and Complete Loss of Forced Reactor Coolant Flow, l
Reactor Coolant Pump Shaft Seizure or Break, f
j Uncontrolled RCCA Bank Withdrawal from a Suberitical or low
[
l Power Startup Condition.
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Uncontrolled RCCA Bank Withdrawal at Power,
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RCCA Misoperation,
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l CVCS Malfunction that Results in a Decrease in Boron j
l Concentration of the Reactor Coolant, j
j Spectrum of RCCA Ejection Accidents.
l Inadvertent Openhg of a Pressurizer Safety or Relief Yalve.
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l Except as noted in this report, these analyses employed a constant l
MTC of +5 pc.e/'F, independent of power level.
The results 1
demonstrate that the safety criteria are met fu the proposed
[
]
Technical Specification limit of lest than 0r equal to +5 pcV'F l
from 0 to 70% nominal rated thernal power, thereafter decreasing j
linearly to o pcm/'F at 100". nominal rated thermal power. A i
revised analysis of the FSAR Section 15.4.6 boron dilution transient I
is provided in this report.
The FSAR Chapter 6 and 15 safety analysis conclusicas are verified to remain applicable as are the conclusions of iteferences 2 and 3.
In conclusion, this report j
provides the safety analysis justification for the proposed MTC l
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Technical Specift:ation limit, t
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page 23
- 3. RhST/ ACCUMULATOR BORON CONCENTRATION INCREASE SAFF.TY EVALUATiCN 3.1 Introduction
'ueling Water 3torage Tank (RWST) Technicai 5pecification boron concentration range has been increased from 2000 2109 to 2350 2500 ppm, and the accumulator Technical Specification boron concentration range has been increased from 1900 2100 to 2300 2500 ppm.
The purpose for these changes is to allow for a longer core cyc?< 2nd implementation of a positive HTC.
l It has also been confirmed that the existing 2000 ppin minimum refueling mode RCS boron concentratito remains suffi'i.nt to provide l
l the required Mode 6 shutdown margin.
These chang (s do n6t require an increase in the boron concentration in the sp!v. 'uel pool.
The accident analyses reported in the FSAR were evaluated for the abe, increased boron concentrations.
The Non LOCA and LOCa evaluations are contained in the following see: ions. The results of the evaluation indicate that there is no adverse effect on the FSA.i results for any of the accident analyses as a roule of increasing tne J
RWST and accumulator boron concentration operating bands.
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page 24 3.2 Non-LOCA Analyses The RWST, accumulators and the Safety Injection System (SIS) are subsystems of the Emergency Core Cooling System (ECCS). Upon actuation of the SIS, borated water from the RWST is delivered to the Reactor Coolant System (RCS) in order to provide adequate core cooling as well as provide sufficient negative reactivity following steamline break transients to prevent excessive fuel failures.
The accumulators are a passive system and provide borated water to the RCS when the system pressure drops below approximately 600 psig.
The only non-LOCA safety analyses in which boron from the RWST or accumulators is taken credit for, or assumed to be present, are those in which the SIS is actuated. These analyses are:
Inadvertent Opening of a Steam Generator Relief or Safety Valve (FSAR Section 15.1.4)
Steam System Piping Failure (FSAR Section 15.1.5)
Feedwater System Pipe Break (FSAR Section 15.2.8)
Spectrum of Rod Cluster Control Assembiy Ejection Accidents (FSAR Section 15.4.8)
Inadvertent Operation of the Emergency Core Cooling System During Power Operation (FSAR Section 15.5.1)
Stearr Generator Tube Ruptara (FSAR Section 15.6.3)
Mass and Energy Releast..nalysis for Postulated Secondary System Pipe Ruptures (FSAR Section 6.2.1.4 and WCAP-10961-P)
The effect of the recomended Technical Specification revisions to the minimum RWST and accumulator boron concentrat.ons on each of these transients is discussed below.
page 25 Inadvertent Ooenino'of a Steam Generator Eafety or Relief Valve (FSAR Section 15.1.4)
An accidental depressurization of the Main Steam System due to the inadvertent opening of a steam generator safety or relief valve is classified as an ANS Condition II event. The accident results in a cooldown of the RCS which, in the presence of a negative MTC, causes a' positive reactivity excursion. Borated water from the RWST onters the core following actuation of the SIS on low pressurizer pressure or low steamline pressure.
The FSAR demonstrates that the negative reactivity provided by the borated water from the RWST limits the return to power to an acceptable level so that the minimum DNBR remains above the safety analysis limits. As the transient proceeds and more water from the RWST reaches the RCS, the boron concentration in the RCS gradually increases, ultimately causing the core to become subcritical.
The proposed increase to the RWST boron concentration changes only the safety analysis assumption for SIS boron conceretration. No accumulator actuation is assumed and all other analysis assumptions are verified to remain applicable.
If the minimum RWST boron concentration is increased from 2000 ppm to 2350 ppm, more negative reactivity would be available to terminate the return to power earlier and at a reduced peak power level. Thus, the maximum core heat flux reached will be reduced. Additionally, the core would become subcritical earlier in the transient.
The minimum DNBR would be higher than for the case currently analyzed which assumes the minimum RWST boron concentration of 2000 ppm.
Therefore, in consideration of the proposed changes, the conclusion is that the i
Condition 11 safety analysis acceptance criteria continue to be met and the conclusions in the FSAR remain valid.
page 26 Steam System Pinina Failure.(FSAR Section 15.1.5)
A major rupture of a ?ain steam line is classified as an ANS Condition IV event. This accident is more severe than the inadvertent opening of a steam generator relief or safety valve (Section 15.1.4) and results in a rapid cooldown of the RCS which, in the presence of a negative MTC,'causes a positive reactivity excursion.
Borated water from the RWST enters the core following actuation of the SIS on low steam line pressure or low pressuMzer pressure. The FSAR demonstrates that the negative reactivity provided by the berated water from the RWST limits the' return to power to an acceptable level so that the minimum DNBR remains above the safety analysis limits. As the transient proceeds and more water from the RWST reaches the RCS, the boron concentration-in the RCS gradually increases, ultimately causing the core to become subcritical.
i The proposed increases to the RWST and accumulatar boron l
concentrations change only the safety analysis assumptions for SIS i.
and accumulator boren concentrations, respectively. All other analysis assumptions are verified to remain applicable.
As demonstrated in the FSAR, the RCS pressure does not decrease to the point of accumulator injection. Any accumulator injection at the h..:er boron concentration would, however, tend to improve transient results. Therefore, in consideration of the proposed changes, it is concluded that the Condition IV safety analysis acceptance criteria continue to be met and the conclusions in the FSAR remain valid.
Feedwater System ploe Dreak (FSAR Section 15.2.8)
A major feedwater itne rupture is classified as an ANS Condition IV event. This accident is defined as a break in a feedwater line which is large enough to prevent the addition of sufficient feedwater to maintain shell side fluid inventory in the steam generators. Rapid decrease in secondary pressure following a feedline rupture may cause the low steamline pressure SIS signal to be reached. The negative reactivity insertion due to the addition
page 27 of borated SIS water is not required to maintain the reactor in a subcritical condition following a feedline break; although, the cold SIS water serves to reduce the RCS temperatures and pressures. The proposed increase to the RWST boron concentration changes only the safety analysis assumption for SIS boron concentration.
No accumulator actuation is assumed and all other analysis assumptions f
are verified to remain applicable. An increase in the minimum RWST boron concentration from 2000 to 2350 ppm, will increase the negative reactivity insertion rate without significantly affecting the reduction of RCS temperatures and pressures. Thus, an increase in the minimuin RWST boron concentration to 2350 ppm will have no adverse impact on the feedwater line break ar,alysis.
Therefore, in consideration of the proposed changes, it is concluded that the Condition IV safety analysit acceptance criteria continue to be met and the conclusions in the FSAR remain valid.
This conclusion is equally valid for the feedlinn break verification study reported in Reference 3.
Spgetrum of RCCA E.iection Accidqn11 (FSAR Section 15.4.8)
Mechanical failure of a control rod mechanism pressure housing resulting in the ejection of an RCCA and drive shaft is classified l
as an ANS Condition IV event. This accident results in a rapid l
positive reactivity insertion and system depressurization together with an adverse core power distribution, possibly leading to localized fuel rod damage.
Following the ejection of a control rod, the rapid nuclear power excursion causes the RCS to experience a i
large pressure rise due to the energy released into the coolant.
RCS pressure then drops as fluid inventory is lost through the break (maximum of 2 square inches) in the control rod housing. As the RCS pressure continues to drop, actuation of the SIS on low pressurizer pressure will inject borated water from the RWST into the RCS.
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page 28 The effect of the RWST boron concentration increase for the rod ejection accident from a LOCA standpoint is addressed in the safety evaluation of the proposed changes on the small break LOCA transient (FSAR Section 15.6.5). No consideration is given to potentially beneficial effects due to SIS or accumulator actuation for the calculations of the acceptance criteria related to fuel / core damage and RCS and secondary integrity as defined in FSAR Section 15.4.8.
Therefore, an increase in the minimum RWST boron concentration from 2000 ppm to 2350 ppm does not affect the rod ejection accident analysis calculations performed for FSAR Section 15.4.8.
In consideration of the proposed changes, it is concluded that the Condition IV non-LOCA safety analysis acceptance criteria continue to be met and the FSAR conclusions remain valid.
Inadvertent Oceration of the ECCS Durina Power Ooeration (FSAR Section 15.5.1)
Spurious actuation of the ECCS while at power is classified as an A!IS Condition II event. The accident results in a negative reactivit, excursion due to injected boron from the RWST.
The decreasing reactor power causes a drop in the enre average temperature and coolant shrinkage.
If reactor trip on SIS actuation l
is assumed not to occur, the reactor will ultimately trip on low pressurizer pressure.
DNBR does not drop below the initial value.
l The proposed increase to the RWST boron concentration changes only l
the safety analysis assumption for SIS boron concentration. No l
accumulator actuation is assumed and all other analysis assumptions are verified to remain applicable.
concentration is increased from 2000 ppm to 2350 ppm, the negative f
reactivity excursion would occur at a faster rate causing a more j
rapid drop in the core average temperature and coolant shrinkage.
The reactor will trip on low pressurizer pressure as before, though at an earlier time in the transient. As before, the DNBR will not decrease below the initial value.
--_,...- -,,. ---_,_,-,.,,_.,,- _,__, _ -,_.- c __ _ - m y
page 29 Therefore, in consideration of the proposed changes, the conclusions of the evaluation are that-the-Condition II safety analysis acceptance criteria continue to be met and the conclusions in the FSAR remain valid.
Steam Generator Tube Ruoture (FSAR Section 15.6.3)
SGTR - Stuck Open Atmospheric Relief Valve Scenario (SLNRC-86-01,1-8-86)
In this SGTR scenario, a single tube rupture leads to a slow depressurization of the RCS and subsequent reactor trip and safety injection (SI).
In order to maximize offsite radiological con >:.quences, SI flow rates and temperatures were assumed which maximize secondary steam production and primary to secondary break flow.
Credit was taken for sufficient boron in the SI flow to maintain subcriticality. Therefore, increasing the boron concentration will have no effect on the analysis conclusions.
SGTR - Overfill Scenario (ULNRC-1518, 5-27-87)
In this SGTR scenario, a reactor trip is assumed to occur coincident with the tube rupture.
After reactor trip, the RCS continues to depressurize until SI flow is initiated.
The consequences of thir, scenario (potential overfill and offsite release) are predominantly dependent upon the time until SI termination. Credit was again taken for sufficient boron in the SI flow to maintain subcriticality.
Therefore, increasing the boron concentration will have no effect on the analysis conclusions.
Egip anu Enerav Release Analysis for Postulated Secondary System-M Ruotures (FSAR Section 6.2.1.4 and WCAP-10961, Rev.1)
A major rupture of a main steam line results in a rapid cooldown of the RCS which, in the presence of a negative HTC, causes a positive reactivity excursion. The calculation of steamline break mass and energy releases for use in determining peak containment pressure and
page 30 temperature (FSAR Section 6.2.1.4) assumes that borated water from the RWST enters the core following actuation of the SIS on low steamline pressure, low pressurizer pressure, or Hi-1 containment pressure.
FSAR results demonstrate that the negative reactivity provided by the borated water from the RWST limits the return to power to an acceptable level so that the minimum DNBR remains above the safety analysis limits. Additionally, by limiting the return to power, the borated RWST water reduces the total energy that is dissipated via steam release through the ruptured steamline.
As the transient proceeds and more water from the RWST reaches the RCS, the boron concentration in the RCS gradually increases, in time causing the core to become subcritical.
The proposed increase to the RWST boron concentration changes only the safety analysis assumption for SIS boron concentration. No accumulator actuation is assumed and all other analysis assumptions are verified to remain applicable.
An increase in the minimum RWST boron concentration from 2000 to 2350 p p would add more negative reactivity to terminate the return to power earlier and reduce the peak power level.
Thus, the maximum core heat flux reached will be reduced and the core would become subcritical earlier in the transient.
Over the course of the transient, the reduced peak power and earlier return to subcriticality would reduce the integral mass and energy releases as a function of time relative to those for cases analyzed assuming 2000 ppm boron in the RWST. Therefore, the cJnclusions in the FSAR remain valid. This conclusion applies equally to the Reference 2 calculations, applicable to the Callaway Plant, of steamline break mass and energs releases for equipment qualification outside containment.
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3.3 LOCA Analyses This evaluatica will determine the effect of increasing the RWST and accumulator boron concentrations on LOCA related accidents in the Callaway FSAR.
For a postulated LOCA, the ECCS is designed to limit the consequences of an accident to meet the acceptance criteria of 10CFR50.46. The LOCA analyses take credit for pumped safety injection from the RWST and passive injection of accumulator water to prevent or mitigate the resulting clad temperature excursion.
Also, post-LOCA long term core cooling takes credit for the available water in the RWST and accumulator in determining the post-LOCA RCS/ sump boron concentration and the hot leg switchover time to prevent boron precipitation.
The effect of an increase in t
the RWST boron concentration Technical specification range from 2000-2100 ppm to 2350-2500 ppm, and an increase in the accumulator boron cencentration range from 1900-2100 to 2300-2500 ppm is discussed in this evaluation.
Small Break LOCA l
The small break LOCA analysis is described in FSAR Section 15.6.5.
l The current FSAR small break LOCA analyses were performed with the NOTRUMP Evaluation Model, which assumes the reactor core is brought to a subcritical condition by the trip reactivity of the control rods. There is no assumption requiring the presence of boron in the ECCS water or the need for negative reactivity provided by the l
f soluble boron.
Thus the changes in the RWST and accumulator boron
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concentrations do not alter the conclusions of the FSAR small break LOCA analysis.
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page 32 Larae Break LOCA
= The large break LOCA analysis is described in FSAR Section 15.6.5.
The current large break LOCA analysis was performed with the NRC Approved 1981 Evaluation Model with BASH. Currently, the large break LOCA analysis does not take credit for the negative reactivity introduced by the soluble boron in the ECCS water in determining the reactor power during the early phases of a postulated LOCA.
The large break LOCA also does not take credit for the negative reactivity introduced by the control rods. During a large break LOCA, the reactor is brought to a subcritical condition by the presence of voids in the core.
Since credit is not taken for the 4
soluble boron that is present in the core, a change in the RWST or accumulator boron concentrations will have no effect on the current FSAR large break LOCA analysis.
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page 33 3.4 LOCA Related Design Considerations LOCA Short and Lona Term Mass and Enerav Releases The containment analyses are described in FSAR Section 6.2 and consider containment subcompartments and mass and energy releases for postulated LOCAs.
for containment subcompartment analyses, an increase in the RWST and accumulator baron concentrations would have no effect on the calculated results, since the short duration of the transient
(< 3 seconds) does not consider any safety injection flow taken from the RWST. The long term mass and energy release calculations do not take credit for the soluble boron present in the safety injection from the RWST supplied to the RCS.
This is similar to the LOCA analyses assumptions, and therefore an increase in-ECCS water boron concentration would-have no effect on the long term mass and energy releases calculated.
Post-LOCA Lona-term Corc J olina Long-tera core cooling is discussed in FSAR Section 15.6.5.
The Westinghouse licensing position for catisfying the requirements of 10CFR50.46 Paraaraph (b) Item (5) "Long Term Cooling" is defined in WCAP.8339 (Reference 6, pp. 4 22). The Westinghouse Evaluation Model commitment is that the reactor will remain shutdown by borated ECCS water residing in the RCS/ sump after a LOCA (Reference 7).
Since credit for control rods is not taken for a large break LOCA, the borated ECCS water provided by the accumulators and the RWST must have a boron concentration that, when mixed with other water sources, will result in the reactor core remaining suberitical assuming all control rods out.
The effect on the post-LOCA RCS/ sump boron concentration as a result of chinging the miaicum Technical Specification boron concentration from 2000 to 2350 ppm for the RWST, and from 1900 to 2300 ppm for the accumulators is an increase of about 100 ppm in the RCS/ sump boron concentration.
It has been confirmed that this proposed increase will provide enough margin I
to keep the core subcritical during Cycle 4 operation for the post-LOCA long-term core cooling requirement.
Following cycles will be verified through the normal RSAC evaluation process.
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page 34 Hot Leo Switchover to Prevent Boron Precipitation 4
A discussion of hot leg switchover time is presented in FSAR Sections 6.3.2 and 15.6.5.
A hot leg recirculation switchover timo analysis has been performed to determine the time following a LOCA that hot leg recirculation should be initiated.
This analysis addresses the concern of boron precipitation in the reactor vessel following a LOCA and has been performed to support an increase in the Technical Specifications for RWST and accumulator boron concentrations.
During a large break LOCA the plant switches to cold leg recirculation after the RWST switchover setpoint has been reached.
If the break is in the cold leg there is a concern that the cold leg injection water will fail to establish flow through the core.
Safety injection (SI) entering the broken loop will spill out the break, while SI entering the intact cold legs will circulate around the downcomer and out the break. With no flow path established through the core the fluid in the core remains i
stagnant.
Steam is produced in the core from decay heat.
In the analysis, it is conservatively assumed that the boron associated with the steam remains in the vessel.
Thus, as steam is boiled off with no circulation present in l
the core the boric acid concentration increases in the vessel.
The boron i
concentration in the vessel will increase until the solubility limit of the baric acid solution is reached at which time boron will begin to precipitate. The boron precipitate may plate out on the fuel rods, which woulj adversely affect their heat transfer characteristics. The purpose l
of the hot leg recirculation switchover time analysis is to provide a time l
at which hot leg recirculation can be established such that boron precipitation in the core can be prevented.
The calculation considers the increase in boric acid concentration in the vessel during the long term cooling phase of a LOCA. The analysis assumes that following a LOCA the steam boiloff from the core does not carry any boron. A constant volume of liquid in the vessel is assumed so that as steam is boiled off and the boron is left behind, the boric acid
page 35 concentration of the vessel increases. The time when the boric-acid solution reaches the solubility limit less 4 weight percent is when hot leg recirculation should be initiated. The solubility limit less 4 weight 0
percent at a solution temperature of 212 F has been established as 23.53%.
Thus, when the boric acid solution concentration reaches 23.53%
hot leg recirculation should be initiated.
Hot leg recirculation provides an injection path into the core which dilutes the boron solution and prevents the further build up of boron.
An analysis has been perfmned to determine the time following a LOCA that switchover to hot leg recirculation should be initiated to prevent boron precipitation in the reactor vessel. This time has been determined to be 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />.
The analysis considers the increase in boric acid concentration in tho reactor vessel during the long term cooling phase of a LOCA, assuming a conservatively small effective vessel volume.
This volume includes only the free volumes of the reactor core and upper plenum below the bottom of the hot leg nozzles.
This assumption conservatively neglects the mixing of boric acid solution with directly connected volumes, such as the reactor vessel lower plenum. The calculation of boric acid concentration in the reactor vessel considers a cold leg break of the RC! in which steam is generated in the core from decay heat while the boron associated with i
the boric acid solution is completely separated from the steam and remains in the effective vessel volume.
The results of the ana' dis show that the maximum allowable boric acid corcentration of 23.53 weight percent established by the NRC will not be l
exceeded in the vessel if hot leg recirculation is initiated 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> after the LOCA inception. The operator should reference this switchover time against the reactor trip /S! signal.
The typical time interval l
between the accident inception and the reactor trip /SI actuation sir 'l i s negligible when compared to the switchover time.
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page 36 r
I Procedure philosophy assumes that it would be very difficult for the operator to differentiate between break sizes and locations. Therefore, one hot leg switchover time is used to cover the complete break spectrum.
Rod E.iection Mass Releases for Dose Calculations The Rod Ejection mass releases are presented in FSAR Figure 15.4-27. Dose releases are discussed in Section 15.4.8.3.
The increase in the RWST and accumulator boron concentrations will have a negligible effect on the mass releases for the Rod Ejection accident.
The SI flow taken from the RWST is assumed at 0 ppm, which is the same assumption as in the large and I
small break LOCA analyses. On this basis, there will be no adverse effect on mass releases for the Rod Ejection accident.
LOCA Hydraulic Vessel and looo Forces The LOCA hydraulic forcing functions resulting from a postulated LOCA are considered in FSAR Section 3.9(N).2.5 and section 3.6.2.2.1.5.
The increase in the RWST and accumulator boron concentrations will have no effect on the LOCA hydraulic forcing fJnctions 3ince the maximum forces are generated within the first few seconds after break initiation.
For i
this reason the ECCS, including the RWST, is not considered in the LOCA hydraulic forces modeling and thus the increase in RWST and accumulator
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boran concentrations will have no effect on the results of the LOCA hydraulic forcing function calculations.
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page 37 R
3.5 Summary and Conclusions Non-LOCA Safety Analysis Desian Basis The proposed increases'in the RWST and' accumulator minimum boron l
concentrations.from 2000 to 2350 ppm and from 1900 to 2300 ppm,-
respectively, have been evaluated for the non-LOCA safety analysis design basis..It is concluded that these-changes have no adverse I
impact on the non-LOCA accident' analyses and that no changes to the associated FSAR sections (other than changes to the previous concentrations) are. required. :This conclusion applies as well for the replacement FSAR Section 15.4.6. analysis discussed in Section 2 of this report.
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LOCA Safety Analysis Desian Basis k
The increase in the RWST Soron concentration from a range of 2000 2100 ppa to a rar,se of 2350-2500 ppm and an increase in the I
accumulator boron concentration from a range of 1900 2100 ppm to a range of 2300 2500 ppm does not have a negative effect oa the FSAR
)
LOCA analyses, as previously described. Current margin to the post-LOCA shutdown requirement is increased, with continued I
conformance verified through the normal RSAC evaluation process, t
Tha higher boron concentrations decrease the allowable time for l
l operator action to initiate hot leg recirculation to 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />.
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Emergency Operating Procedure ES-1.4, Transfer to Hot Leg r
- rculation, will therefore be modified to reflect the new hot leg switchover time. The resulting time requirement of 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> is still more than adequate to assure that operato,' actions can be
[
accomplished.
In conclusion, there is no adverse effect on FSAR LOCA analyses for the proposed increases of the boron concentrations for the RWST and the accumulators at the Callaway Plant.
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page 38 4.0 BALANCE OF PLANT CONSIDERATIONS Post-Accident Containment Sorav DH The Containment Spray System, in conjunction with the Containment Fan Cooler System and the Emergency Core Cooling System, is capable of removing sufficient heat from the containment atmosphere following a LOCA to maintain the containment pressure below the containment design pressure.
In addition, the Containment Spray System reduces the iodine and particulate product inventories in the containment post-LOCA atmosphere. To enhance the iodine absorption capacity of the containment spray, the spray solution is adjusted to an alkaline pH to promote iodine hydrolysis. This is currently accomplished by adding 28% - 317.
concentration by weight Na0H solution in the spray. Approximately 2500 gallons of NaOH solution is presently required to be added to the containment sump to achieve a final post-LOCA containment sump pH of at least 8.5.
A minimum pH of 8.5 in the containment sump is necessary to ensure long-term retention of iodine in solution.
In order to ensure that the required amount of Na0H solution is added to the cump, spray additive eductor isolation valves in the NaOH solution supply headers are provided with an interlock from the tank level transmitters to preclude their I
closure prior to the addition of the required amount of NaOH solution.
These valves, once opened by the Containment Spray Actuation Signal (CSAS), can be closed only after a close permissive signal is given by low level switches installed on the Spray Additive Tank.
j As a result of the proposed implementation of a positive HTC, it has been determined that higher levels of boron concentration are needed to be maintained in the Refueling Water Storage Tank (RWST) and safety injection accumulators.
It has also been determined that higher sodium hydroxide j
(NaOH) concentrations (31% - 34% by weight) are required to achieve the minimum long term post LOCA containment sump pH of 8.5, based on the
(
increased boron concentrations.
l The 9tnimum pH value of the containment spray during the injection phase l
is slightly affected.
The new pH range for the containment spray during
page 39 the injection phase is 9.3-11.0 rather than the origin.'1 pH range of 9.5-11.0.
This new pH range of 9.3-11.0 has no adverse ing=ct on achieving the final containment recirculation sump solution pH of at least 8.5.
In addition, the revised pH range will not inpact the elemental or particulate iodine absorption coefficients used in the accident analysis; therefore, the radiological consequences of a LOCA as stated in FSAR Table 15.6 8 are not affected.
Environmental Oualification of Electrical Eau oment in Containment As described in FSAR Section 3.11(B).1.2.2, Accident Environments Irside containment, the containment spray pH ranges fr)m 4 to '.l.
The revisd NaOH concentration range will have no impact on the pH values utilized b.*
environmental qualification.
As described in Callaway Plant Technical Specification Bases 3/4.6.2.2, the lisits on Na0H volume (4340 to 4540 gallons) and concentration (31-34 percent by weicht) ensure a pH value of between 8.5 and 11 for the solution recire. lated within containmeat pott-LOCA. The subject modifications do not affect these regt.ieemen'.s, therefore the margin of safety is not red ced.
liydtqqtp__ Generation in Containment t
The capability of the Containment Spray Systen to perform its safety-related functions, as described in FSAR Section 6.2.2.1, is unaffected. heither the rate nor the total quantity of post-accident hydrogen generation, as discussed in FSAR Section 6.2.5.2.3, is adversely affected since the upper limit of pH for the Containment Spray remains unchanged.
Potential Precioitation of Boric Acid l
The increase in concentration of boron has been evaluated to assure that l
solubility limits :re not exceeded at the minimum temperature limits for the RWST ad SI accumulators.
The coincidence of maximum boron concentrations and minimum allowed temperatures would not result in precipitation of boric acid crystals.
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SECTION 5
.i REFERENCES 1.-
Union Electric Company, Callaway Plant, Final Safety Analysis Report.
Docket Number 50-483, Revision 0L-2.
l 2.
WCAP-10961-P, Steamliree Break Mass / Energy _ Releases for Equipment
'l Environmental Qualification outside Containment, Octnber 1985.
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3.
WCAP-11884, Implementation of the Steam Generator Low Low Level l
Reactor Trip Time Delay and Environmental A11owance' Modifier in the 4
Callaway Plant, August 1980, submitted by ULNRC-1822 dated August 30, 1988.
WCAP-11325-P-A, Rev. 1, Westinghouse Owners' Group Trip Reduction and 4.
"Assessment Program: Steam Generator low Water Level Protection System Modifications to Reduce Feedwater-Re'<ated Trips, February 1988.
l l
5.
WCAP-ll342-P-A, Rev. 1, Modification of the Steam Generator Low-Low
[
Level Trip Setpoint to Reduce Feedwater-Related Trips, April 1988.
t l
- 6..
WCAP 8339 (Non-Proprietary), Bordelon, F.M., et al, "Westinghouse ECCS Evaluation Model - Summary", June 1974.
{
l 7.
Westinghouse Technical Bulletin NSID TB 86 08 (Non-Proprietary),
i "Post LOCA Long Term Cooling: Boron Requirements", October 31, 1986, f
i 8.
SLNRC 86 01 dated January 8, 1986.
e l
9.
ULNRC-1518 dated May 27, 1987.
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