ML20155E627
| ML20155E627 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 10/04/1988 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20155E029 | List: |
| References | |
| NUDOCS 8810120359 | |
| Download: ML20155E627 (16) | |
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'o UNITED STATES
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NUCLEAR REGULATORY COMMISSION
- E WASHINGTON, D. C. 20565 g
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.11 TO FAC,Ij.ITY OPERATING LICENSE NPF-68 GEORGIA P05!ER COMPANY, ET AL.
1 DOCKET NO. 50-424 V0GTLE ELECTRIC GENERATING PLANT, UNIT 1
1.0 INTRODUCTION
By letter dated May 19,1988 (Ref.1), Georgia Power Company (the licensee) made application to amend the Technical Specifications of Vogtle Electric Generating Plant, Unit 1 (Vogtle 1). 'ng (preposed changes would modify the The
- 1) the moderator temperature Technical Specifications (TS) concee coefficient (itTC) and (2) shutdown jin (SDM) requirements. The MTC change would allow a slightly positive MTC aelow 100 percent of full rated power. The principal benefit of this change is that it would facilitate the design of future reload fuel cycles.
TS changes are required to meet SDM requirements to acconnodate the positive MTC and future 18 month reload fuel cycles. To assure that subcriticality requirements are met following a postulated loss-of-coolant accident (LOCA), the boron concentration is increased for the refueling water storage. tank (RWST) and the accumulators.
An increase in the RWST water volume requirement for Modes 5 and 6 is also proposed. To meet subcriticality requirements for boron dilution et mts, the SOM limits for Modes 3, 4, and 5 and the high flux at shutdown alarm setpoint are changed.
The applicable safety analysis for Vogtle 1, which is in its initial cycle, is that documented in the Final Safety Analysis Report (FSAR). This safety analysis is based on a O pcm/deg F MTC at all times when thg reactor is critical.
(Note that a pcm is equal to a reactivity of 10" delta k/k.) The proposed change to the TS would allow a + 7 pcm/deg F MTC below 70 percent power, with a linear variation in the MTC of + 7 pcm/deg F at 70 percent power to O pcm/deg F at 100 percent power. The licensee has reevaluated the FSAR safety analysis using this positive fiTC as well as an increase in the boron concentration for the RWST to a concentration range of 2400-2600 ppm and an increase in the boron concentration for tte accumulators to a concentration range of 1900-2600 ppm.
ThisreevaluationisprovidedinEnclosure1(Ref.2) of Reference 1.
Additional information was submitted by letter dated 'ugust 12, 1988 (Ref. 14).
Included in this submittal was a change to the TS bases Section 3/4.9.1, ' Boron Concentration' which clarified the calculational uncertainty for reactivity conditions during refueling.
This small clarifying change to the bases did not substantially affect the amendment request as noticed or the staff's initial determination; therefore, the pendment was not renoticed.
8810120359 081004 PDR ADOCK 05000424 P
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o MditioM1 information was submitted by lettei dated October 3, 1988 (Ref.
- 15) and by telecon of October 4,1988 (Ref.16). This information described the licensee's intended implementation of this amendment. To implement this dmendment, a footnote 1s being added to Technical Specification 3.5.4, "Refueling Water Storage Tank" and Technical Specific 6 tion 3.1.2.6, "Borated Water Sources-Operating" that states, "Until concentration is initially raised to 2400 ppm from the maximum limit duthorized prior to Amendment No.
11, the minimun boron concentration limit is 2000 ppm."
This footnote prevents the licensee from being in immediate noncompliance when this Technical Specification amendment is implemented.
This footnote regarding in:plementation does not substantially affect the ar,endment request as noticed or the staff's initial determination; therefore, the amendment, was not renoticed.
The NRC stoff has reviewed the proposed amendment for Vogtle 1 and its evaluotion follows.
2.0 EVALUATION 2.1 Effect of Positive MTC on Transient and Accident Analyses The licensee has evaluated the effect of the proposed change to the MTC TS on the transients and accidents previously an41yzed in the FSAR. Events which were sensitive to e pusitive or near zero MTC were reana lyzed.
These events were the transients which cause the reactor coolant temperature to increase (or equivalently the density to decrease).
The events reanelyzed usea the identical analysis methods, computer codes, and dssumpt.icns used for the FSAR analysis. Any differences in the ptesent dnalysis to the previous FSAR analysis will be noted in the folluwing discussion.
A number of transients were not reanalyzed for the proposed change in the TS to a positive MTC. These transients are listed below, along with the applicable FSAR Sections, and the reason given by the licensee for not requiring a reenalysis.
Transi g FSAR Section Reason for not Reanalyzing 1.
Feedwater System Malfunctions (a) Decrease in Feedwater 15.1.1 Limiting case with negative Temperature MTC (b) Increase in Feedwater 15.1.2 Limiting case with negative Flow MTC 2.
Increase in Steam Flow 15.1.3 Analysis assumes large negative MTC, minimum DNBR not sensitive to MTC
3.
(a)InadvertentOpeningofa 15.1.4 Limiting case with negative Steam Generator Relief or MTC Safety Valve (b) Steam System Piping Failure 15.1.5 Limiting case with negative MTC (c) Steamline break mass / energy 6.2.1.4 Limiting case with negative release inside containment MTC 4
Feedwater System Pipe Break 15.2.8 Analyzed with negative MTC, not sensitive to positive MTC since reactor trip occurs near beginning of transient before reactor coolant temperature increases Trorisient FSAR Section Reason for r.ot Reanalyzing 5.
Control Rod Misoperation 15.4.3 Misalignment cases are not affected by a positive MTC, single rod withdrawal analyzed at steady-state conditions until the reactor trips on the overtemperature delta-T signal and is not dependent on a positive MTC, dropped rod limiting analysis unaffected since MTC must be close to zero or negative at 100% power.
6.
Startup of an Inactive Loop 15.4.4 Limiting case with negative at an Incorrect Temperature MTC 7.
Inadvertent operation of the 15.5.1 A positive MTC is less ECCS During power Operation limiting than FSAR analysis.
Positive MTC causes power to decrease uore rapidly and increases mergin to DNB.
8.
a Small Break LOCA 6.2 Positive MTC has negligible b Large Break LOCA 15.6.5 impact on small break LOCA; c LOCA forces during large break LOCA a d Containment Integrity positive MTC will not be Analys is significant due to cure voiding during blowdown period; peak LOCA forces occur before any impact from positive MTC occurs;
containment integrity analysis is performed at 100% power where the analysis is based on o zero MTC; the steamline break mass and energy analysis is based on a negative MTC.
9.
Steam Generator Tube Failure 15.6.3 Effect of a positive or Rupture (SGTR)
HTC was determined using 4 LOFTRAN result; minimum DNBR remains above limits; increase in reactor power reduces j
reactor trip time and increases primary to Transient FSAR Section Reason for not Reanalyzing secondary leakage but doses remain well below NRC limits of a small fraction of 10 CFR 100.
I WCAP-11731 Revised SGTR analysis was (Ref. 3) perforn=d with Vogtle positive MTC and boron requiren,ents with acceptable results.
- 10. Fuel Misloading Error 15.4.7 Analysis is performed with steady-state methods and is not affected by a l
positive MTC.
The NRC staff has reviewed the evaluation presented by the licensee for each of the transients listed above and concurs with the licensee's assessment that these transients do not require a reanalysis.
The transient analyses for those events that were reanalyzad used 'the same coaputer codes that were used for the Vcgtle 1 FSAR analyses. These codes are LOFTRAN (Ref. 4), TWINKLE (Ref. 5), FACTRAN (Ref. 6), and THINC (Refs. 7 and 8).
The initial conditions, instrument errors, and setpoint errors are essentially the sarne as those found in Chapter 15 cf the Vogt'e 1 FSAR.
The allowance on pressurizer pressure given in FSAR Section 15.0.3.2 has been 4
changed from 2 30 psi to a more conservative 45 psi.
Also, the pressurizer and steam generator water levels uncertainties have been increased from 5% to 6.6% to bound calculated increases in associated transmitter uncertainties, a
4
+--m
The LOFTRAN code was used to reanalyze all the events affected by a positive MTC except for the rod ejection accident and rod withdrawal from a suberitical condition which were analyzed using the TWINKLE code.
The analyses with the LOFTRAN code conservatively used a + 7 pcm/deg F HTC over the entire power range except for the locked rotor accident.
The MTC was conservatively assumed to remain constant for variations in temperature for all transients.
The analyses with the TWINKLE code used at least a + 7 pcm/deg F MTC at zero power nominal overage temperature conditions.
This coefficient becomes less positive at higher temperatures since the TWINKLE formulation does not allow r
the moderator temperature feedback to artificially be held constant.
The evaluation of each transient that was reanalyzed for a positive MTC is discussed below, I
l Uncontrolleo Rod Bank Withdrawal From a Suberitical or Low-Power Startup l
Cor.dition (F5AR 5ection 15.4.1) l i
The uncontrolled rod bank withdrawal from a subcritical condition transient leads to a power excursion.
This power excursion is terminated, after a fast power rise, by the negative Doppler reactivity coefficient of the fuel, and a reactor trip on source, intermediate or power range flux, or high positive neutron flux rate.
The power excursion results in a heatup of the moderator / coolant and the fuel.
The positive MTC causes an increase in the i
rete of reactivity addition, resulting in an increase in peak heat flux and peak fuel and clad terrperature. The analysis used the same reactivity 1
Insertion rate of 60 pcm/sec as the Vogtle 1 FSAR.
This reactivity insertion rate is greater than two sequential control banks withdrawing at the maximum speed of 45 inches / minute.
The neutron flux overshoots the nominal full power value; however, the peak heat flux is much less than the full power nominal value because of the inherent thermal lag in the fuel.
The minimum Departure frcra Nucleate Boiling Ratio (DNBR) remains above the limiting value at all times throughout the transient.
Thcrefore, the conclusions of the FSAR remain valid.
Uncontrolled Rod Bank Withdrawal at Power (FSAR Section 15.4.2)
The uncontrolled rod bank withdrawal from a power condition transient leads to a power increase. The transient results in an increase in the core heat flux and an increase in the reactor moderator / coolant temperature.
This transient could result in departure from nucleate boiling (DNB) unless terminated by l
manual or automatic action.
The Power Range High Neutron Flux and j
Overtemperature Delta-T reactor trips are assumed in the analysis to provide protection against DNB. The positive MTC causes an increase in the rate of reactivity addition.
The minimum reectivity feedback cases presented in the j
F:AR were reanalyzed with the positive MTC.
The maximum reactivity feedback caset Were not rednelyzed because these assumed a large negative MTC.
The licensee presented results for the minimum reactivity feedback case for power lj levels of 10%, 60%, and 100% power for a range of reactivity insertion rates.
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1 The results indicate that the departure from nucleate boiling ratio (DNBR) limit is met for all the cases that were analyzed and the conclusions of the FSAR remain valid.
4 Loss of Forced Reactor Coolant Flow (FSAR Sections 15.3.1 and 15.3.2)
The loss of ficw transient causes the reactor power to increase until the reactor trips on a low flow trip signal or a reactor coolar nump power supply undervoltage signal. The reactor power increase causes a reactor moderator / coolant tempera *.ure increase. This initial coolant temperature increase causes a positive reactivity insertion because of the positive MTC.
The licensee reanalyzed both a partial loss of flow (loss of 2 pumps with four coolant loops in operation) transient and a complete loss of flow transient.
For the partial loss of flow transient, DNB does not occur.
The average fuel 5
and clad temparatures do not increase significantly above their initial values j
because the primary coolant maintains its ability to remove heat from the fuel.
For a partial loss of flow the reactor will stabilize at a condition of 4
power, flow rate, and pressure from which a normal plant shutdown may proceed.
For the complete loss of flow transient DNB does not occur.
The average fuel and clad temperatures do not increase significantly above their initial values because the primary coolant maintains its ability to remove heat from the fuel.
For this complete loss of flow, the flow will coastdown until natural circulation flow is established and normal plant shutdown may proceed. The results indicate that the DNBR limit is met for the partial and complete loss of flow events and the conclusions of the FSAR remain valid.
Reactor Coolant Pump Shaf t Seizure (FSAR Section 15.3.3)
The FSAR analysis for the reactor coolant pump seizure event assumed DNB to occur at the beginning of the avent.
Consequently, the positive MTC will not affeet the time to DNB.
Existing sensitivity studies, which were used for the FSAR analysis, show that a zero MTC at full power is more limiting than a i
positive MTC at lower power. The stnsitivity results covered the proposed positive MTC for the Vogtle 1 The amount of fuel in DNB that is assumed to fail is, therefore, plant.
the same as that previously assumed for the FSAR analysis and the FSAR radiological consequences evaluation remains valid. The reactor coolant pum) seizure event was reanalyzed with a positive MTC to evaluate the effect of t1e power transient on peak reactor coolant system pressure and clad temperatures. This reanalysis was performed both
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with and without offsite power.
For both of these cases, reactor trip occurs I
when the low flow trip setpoint is reached.
The peak reactor coolant system pressure reached during both cases is less than that which would cause stresses to exceed the faulted condition stress limits.
The peak clad surface temperature for both cases is about 1750 deg F so that the amount of zirconium-water reaction is small.
These two cases demonstrate that the
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conclusions of the FSAR remain valid with respect to peak pressure and clad i
temperature for the reactor coulant pump shaft seizure event, j
Turbine Trip Events (FSAR Sections 15.2.3,15.2.4,and15.2.5) 1 1
r 4
A turbine trip event is raore limiting than other events which lead to a turbine trip. The minimum reactivity feedback cases were analyzed for the positive MTC. These cases occur at beginning-of-cycle (BOC). The maximum reactivity feedback cases were not reanalyzed because they occur near end-of-cycle (EOC) and assume a negative MTC.
For one case, full credit is ta ke: in the analysis for the pressurizer spray and pcwer operated relief valves (PORV).
For the other case, no credit is taken in the analysis for the i
operation of the pressurizer spray or PORVs.
Both cases assume the availability of the pressurizer safety valves.
For both pressure control cases, the reactor trips on reaching the High Pressurizer Pressure Trip l
setpoint.
The results show that the primary system pressure remains below the 110% design value and that CNBR r3 mains well above its limit. This transient i
remains the limiting Condition 2 transient with respect to peak pressure.
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These two cases demonstrate that the conclusions of the FSAR remain valid with respect to peak pressure and DNBR for the turbine trip event.
In addition, l
the conclusions of the Overpressure Protection Report remain valid because the system pressure remains below 110% of the design value.
Rod Ejection Accident (FSAR Section 15.4.8) i The rod ejection accident is analyzed at full power and hot standby conditions for both BOC and EOC in the FSAR.
The EOC cases are those with large negative MTCs. Therefore, the reanalysis was performed for the positive MTC for BOC conditions. The reanalysis used ejected rod worths and transient peaking factors that are conservative with respect to the actual values for the current fuel cycle.
In the analysis, reactor trip occurred when the power range high neutron flux setpoint was reached.
The peak hot spot clad everage temperature was reached in the hot zero power case.
The peak hot spot value of 2490 deg F was below the limit specified in the FSAR.
The maximum fuel temperature and enthalpy occurred for the hot full power case.
For both cases, the peak fuel enthalpy was well below the staff criterion of 180 cal /gm.
For the hot full power case, the peak fuel centerline temperature at the hot tjot exceeded the melting temperature, but the extent of melting was l
less than the innermost 10% of the fuel pellet.
Because fuel and clad temperatures and the fuel enthalpy do not exceed the lirnits in the FSAR, the conclusions of the FSAR remain valid.
Loss of Normal Feedwater Flow / Loss of Honemergency AC Power to the Plant 4
Auxiliaries (F5AR 5ections 15.2.7 and 15.f.6)
FSAR Section 15.2.7 presents the analysis of a loss of normal feedwater with offsite power available.
FSAR Section 15.2.7 presents the analysis of a loss of nurnal feedwater which assumes offsite power is lost.
This event was reanalyzed for B0C conditions for the positive MTC.
The reanalysis used a conservative core decay heat model based on the ANSI /ANS-5.1-1979 decay heat l
standard (Ref 9).
The pressurizer preuure control system, sprays and PORVs were osumed to be available since a Icwer yressure results in a greater system expansion.
For the cases with and without offsite power, reactor trip occurred when the low-low steam generator water level trip setpoint was i
redched.
For the case with the loss of offsite power assumption, power was 1
assumed lost to the reactor coolont pumps following control rod motion.
The reanalysis showed that a loss of normal feedwator does not adversely affect the reactor core, the reoctor coolant system, the steam system, and that the auxiliary feedwater system is sufficient to prevent water relief tr rough the pressurizer relief or safety valves.
For the case without offsite power dveilable, the natural circulation capability of the reactor coolant system is sufficient to remove decay heat following a reactor coolant pump coastdown to prevent fuel or clad damage.
For both cases, the pressurizer does not fill and, therefore, the conclusions of the FSAR remain valid.
Inadvertent Opening of a Pressurizer Safety or Relief Valve (FSAR 5ection 15.6.1)
The inadvertent opening of a pressurizer safety or relief valve event causes a moderator / coolant density reduction.
Because a positive MTC can be considered to be a negative density coefficient, that is, equivalent to a temperature increase, the density reduction due to a reactor coolant system depressurization causes a positive reactivity insertion.
For this event, the nucleer power increases until a reactor trip occurs when the overtemperature delta-T trip setpoint is reached.
The analysis is performed with the control rods in the manual mode so that control rod insertion does not compensate for the reactivity insertion caused by the coolant density reduction.
The results indicate that the DNBR remains above the limit of 1.30 throughout the transient and, therefore, the conclusions of the FSAR re.uin valid.
Chemical and Volume Control System Malfunction that Results in a Decrease in the Baron Concentrdtion in the Reactor Coolant (FSAR 5ections 15.5.2 and 15.4.6)
The baron dilutiun event for Mode 1 assumes manual rod control and is dependent upon the results of the Uncontrolled Rod Bank Withdrawal at Power Ana ly si s.
A reactivity insertion rate is calculated for this tiode 1 boron dilution event and compared to those analyzed for the rod withdrawal at power event.
This analysis shows that a minimum of 16.9 minutes are available for the operator to terminate the event from the time of reactor trip which is the first indication of the event. The analysis for Modes 3, 4, and 5 was performed in the same manner as the FSAR analysis.
The high flux at shutduwn f
dlarm setpoint was revihed to 2.3 times background, and the makeup flow t
control valve setpoint was changed to 100 gpm.
The change to a 100 gpm setpoint allows a lower dilution flow of 110 gpm for the analysis compared to the previous analysis.
The analysis results in curves of required shutdown margin as a function of reactor C0olant system boron Concentration.
Meeting I
the shutdown margin requit ements will provide at least 15 minutes for operator action from the high flux at shutdown alarm to the time when shutdown margin is lost.
These shutdown margin curves include the increased reload boron concentration and will be placed in the Technical Specifications as Figures 3.1-1 and 3.1-2.
For Mode 6, administrative procedures require that certain velves be locked closed to prevent a boron dilution event.
The Mode 2 boron i
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i dilution event was not reanalyzed because it is ' lot affected by a positive MTC. The results provided by the licensee demonstrate that sufficient time is available to the operator to terminate a boron dilution event in the i
reanalyzed modes and, therefore, the conclusions of the FSAR remain valid.
I In summary, the NRC staff has reviewed each of the resnalyzed events discussed above and concludes that the licensee's assessments and results obtained are r
dCCeptable.
i In meetings with the staff on the ATWS MTC, Westinghouse and the Westinghouse l
Owners Group have presented data for their plants which show that there has been no doverse trend in the ATWS MTC.
The data include data from plants q
with extended fuel cycles and positive HTCs similar to the planned future operaticn of Vogtle 1.
Because ATWS is considered for a class of plants and the Vogtle 1 class of pidnts have ATWS HTCs which are not more positive than i
the previous ATWS analysis basis HTC, the staff concludes that ATWS need not 4
be considered in this review of a change to a more positive HTC for Vogtle 1,
[
2.2 RWST and Accumulator Boron increase t
The implementation of a positive MTC and extending operating cycles to 18 l
months require changes to the boron concentration of the refueling water storage tank (RWST) and the accumulators.
The licensee states that these changes to the boron concentration are required to meet the long term core cooling requirements of 10 CFR 50.46.
The maximum boron concentration of the RWST and accumuldtors has been increased to 2600 ppm.
The minimum boron concentration of the RWST has been increased to 2400 ppm.
The minimum boron concentration of the occumulators remains unchanged at 1900 ppm.
The licensee has considered the effect of these boron concentration changes in a number of areas.
Non-LOCA Transient and Accident Analyses which Model the RWST i
The various steamline break accidents discussed in FSAR Sections 15.1.4, 15.1.5, and 6.1.2.4 all take credit for the RWST boron injected into the reactor ccolant system by the Safety injection (SI) system.
Increasing the i
minimum RWST boron concentration leads to 4 less limiting analysis than provided in the FSAR which used a smaller value for the boron concentration, i
The increased minimum boron concentration will insert more negative reactivity into the core for these events, thus providing less limiting results. The FSAR analysis will remain bounding and the conclusions in the FSAR remain valid.
The feedwater system pip: break (FSAR Sectica 15.2.8) analysis models the St i
1 sy stem.
This transient is not sensitive to the RWST boron concentration.
Increasing the RWST baron concentration would, however, give less limiting
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results.
The purpose of the $1 is to provide a source of cool water to aid in cooling the primary system and to help ensure that the core remains covered.
l i
The FSAR onelysis remains bounding and the FSAR conclusions remain valid.
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The minimum RWST boron concentration has been considered for boron dilution events (FSAR Seccion 15.4.6) for Modes 3, 4, and 5 in the reanalysis with a positive MTC discussed above. The shutdown margin limits have been defined for the possible higher reactor coolant system boron concentrations allowed by the increasei RWST boron concentrations.
New Technical Specification limits for the shutaown margin as a function of reactor coolant system boron are being proposed for Modes 3, 4, and 5.
The inadvertent operation of the ECCS during power operation event (FSAR Section 15.5.1) would result in an increased boron concentration in the SI water.
This would cause the nuclear power to decrease at a somewhat faster rate.
The core average temperature and pressure would decrease at a somewhat fastar rate.
The decreasing power and temperature inore than offset the decrease in pressure so that the trend of increasing DNBR for this event will i
not change.
The FSAR analysis remains bounding and the FSAR conclusions remain valid.
The only non-LOCA transients which take credit for the accumulator boron concentration are the steamline break event.
The mininum boron concentration is codeled for conservatism. There is no change in the steamline break l
analyses because the minimum accumulator boron concentration has not been changed.
The NRC staff has reviewed the effect of the proposed changes to the RWST and accumulators boron concentrations on affected non-LOCA transients and concludes that the licensee's assessments, as discussed above, are acceptable.
The licensee requeJted 60 days to implement this amendment. As discussed in reference < 15 and 16, the licensee intends to corrence implementation of this amenoment prior to shutdown for refueling in order to assure that coolant is adequately mixed to ensure uniform borun concentration for refueling cperations. As a result, RWST boron concentration will be increased above the i
old TS limit of 2000-2100 ppm and yet will still be below the new TS limit of 2400-2600 ppm for a period of time.
Therefore, ty telecon of October 4,1988 (Ref.16), a footnote was agreed to be added to TS to allow the minimum boron concentration to be 2000 ppm until concentration is initially raised to 2400 ppa.
The increase in RWST boron concentration will be accomplished by utilizing a gravity feed of approximately 100 gpm from the RWST to the spent fuel pool with a return to the RWST via a spent fuel pool cooling pump bypass loop. During this operation, RWST level will be maintained approximately 36,000 gallons above the minimum TS limit; however, if a small break LOCA occurs, operators will secure this operation long before 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> pass which is the time that the 36,000 ga)1cns is based on. Also, prior to exceeding 2100 ppm, the )rocedure for hot leg recirculation switchover will be changed to occur at 11 1ours post accidot.
A review of LOCA, non-LOCA, and system related t ansients was performed regarding implementation of the amendment prior to Mode 5.
It has been l
e i
determined that conclusions regarding dmendment acceptability reached in previous licensee submittals remain valid.
From this, the NRC staff concludes that implementation of this amenduent to increase boron concentration in the RWST may commence prior to shutdun and that therefore the footnote allowing the minimum boron concentration to be 2000 ppm is acceptable.
The submittal of October 3,1988 (Ref.15) requested discretionary enforcement from the previous TS upper boron concentration limit of 2100 ppm to allow the boron concentration to be raised to 2400 ppm for this amendment. Since this i
amendment, a footnoted, covers raising boron concentration, discretionary enforcement is not necess6ry.
LOCA The small break LOCA analysis (SBLOCA) does not explicitly model core reactivity or account for baron provided by the Emergency Core Cooling Syftem. The analysis assumes that the reactor remains subcritical following control rod insertion.
The andlysis further dssumes that the boron provided by the ECCS will keep the reactor subcritical for the all rods in minus 2 (ARI-2) condition (one rod out for rod ejection caused SBLOCA and one ru out for stuck rod assumption).
The additional RWST and accumulator boron concentrations would make the SBLOCA iess limiting with regard to core reactivity and, therefore, the FSAR conclusions regarding the SBLOCA analysis remain valid.
The large break LOCA onelyses do not take credit for the negative reactivity of the toren in the ECCS water.
During a time to just beyond the time of peak clad temperature (PCT), core voiding keeps the reactor subcritical.
The FSAR conclusions remain valid regarding the large break LOCA for an increased RWST and accumulator boron concentration.
For the post-LOCA long term cooling (FSAR Section 15.6.5) tne increase in the RWST minimum boron concentration gives an increase of about 280 ppm of boron in the post LOCA reactor coolant system / sump boron concentration.
The licensee states that this increased boron concentration is snough to offset the effect of a positive MTC.
The licensee states that LOCA hydraulic vessel an',
- es (FSAR Section 3.6) are not affected by the increase in the mini:
gu oron concentration because the maximum loads are generated within the 1 y ew seconds of a break initiation.
For this reason, the ECCS, includin,
,e PWST, is not modeled when considering LOCA forces.
For the FSAR analysis of the steam generdtor tube rupture (SGTR) event (FSAR Section 15.6.3), sufficient shutdown margin is assumed to be available initially because of control rod insertion following a reactor trip and adequate shutdown margin is assumed to be maintained for the long term by the l
-- -i
t borated safety injected water. An increased RWST minimum boron concentration will result in mort negative reactivity insertion for this event and will, therefore, have no adverse impact on the FSAR analysis.
For the revised SGTR anslysis ir. WCAP-11731, operator uctions are rrtdeled in the anal sis.
It is f
assuined that sufficient shutdown margin will be provided initially by the insertion of control rods on reactor trip and will be maintained during the reactor coulant system cooldown by the borated safety injection water.
The increased negative reactivity insertion rate will, therefore, have no adverse impact on the WCAP-11731 SGTR analysis, t
For the containment Integrity analysis (FSAR Section 6.2), the short term mass and energy subcompartment pressure analyses are not affected by the increase in the RWST baron concentration because, for the short duration of the transient, safety injection flow frum the RWST is not considered. The long t
term inoss and er.argy release and containment response calculations following a L
LOCA do not take credit for the soluble boron in the safety injected water from the RWST and, therefore, the increased RWST boron concentration will have no L
effect on these analyses.
For secondary system pipe ruptures inside containment, the RWST is rrodeled in the rnass and energy analysis.
The increased RWST boron concentration will insert more negative reactivity in the cure and result in less limiting mass and energy releases.
The licensee states
- hat the conclusions presented in the Vogtle FSAR remain valid.
For the increased concentration of boron in the RWST, the licensee established the titre for hut leg twitchover from cold leg injection (FSAR Section 6.3.2).
This switchover is required to prevent boron precipitation from occurring for a cold leg break.
The licensee states that 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> from the start of a LOCA, hot leg switchover of RWST burated water injection must be initiated before loss of solubility of borun in the water.
This hot leg switchover time covers the cun'plete break spectrum.
The rod ejection accident (FSAR Section 15.4.8) mass and energy releases use similar asnn.pticos in nedelin and large break LOCA analyses.g the RWST safety injection flow as the mall Therefore, the increased RWST boron concentration will have no aaverse effect on the FSAR rod ejection accident.
it af '
s '? has reviewed the effect of the proposed increases to the RWST and baron concentrations on LOCA and concludes that the licensee's l
atm asses =,,s.ts, as discussed above, are acceptable, j
Effect of RWST and Accurnulator Boron Concentrations on Fluid Systems and LOCA Radiological Consequences i
The licensee evaluated other effects of the increased borun concentrations in the RWST and accumulators.
The volume of boric acid solution required in i,he RWST during Modes 5 and 6 is increased, as well as the boron concentration, to meet the reqeirement: of an 18 month cycle.
The volume of boric acid solution in the RWST required to bring)the plant to cold shu*down conditions is increased (affects a TS Bases.
No changes are requiecd to the Boric Acid I
i
Storage Tank.
The Reactor Makeup Control System automatic flow rate is decreased frorn 120 to 100 gpm so that the makeup control system will be able to blend approxirnately 35 gpm of 7000 ppm boron solution from the boric acid storage tank with 65 gya of reactor makeup water to deliver 100 gpm of 2500 ppm liquid to the reactor.
The current licensing basis for Vogtle 1 limits the long term surp pH to between G.5 and 10.5 and the containment spray pH to less than 11.0.
The 4
increased RWST and accumulators boron concentrations reduce the long term sump pH.
The licensee established that the minimum sump pH is 8.15.
This value of the sump pH is less than the current licensing basis of 8.5.
The licensee evaluated the effect of this change in the minimum sump pH. The licerisee determined that the increase in the corrosion rate between a pH of 8.0 and 8.5 is not significant for the :onstruction materials in containment dnd, therefore, the source of post-LOCA hydrogen generation is not increased by a reduction in the minimum sump pH.
The licensee states that equipment i
qualification is not affected by the reduction from a pH of 8.5 to a pH of 8.0 l
because equipment is qualified at a pH of 10.7 and any move towards a pH of 1 0 would provide a more neutral environment.
For chloride induced stress corrosion cracking of stainitss steel Westinghouse reconnends o minirnum sump pH of 7.5.
The NRC recomends in Reference 10 that i
the minimum equilibrium sump pH should be between 7.0 and 9.5, with the higher values providing greater assurance that no stress corrosion cracking will occur.
The minimum pH of 8.0 is consistent with both the Westinghouse and NRC recommendations.
The licensee evaluated the effect of the reduction in minimum sump pH to 8.0 or. (OCA thyroid duses.
Reduced conservatisms were assumed for (1) deposition removal of elemental iodine from the containment (Ref.11), (2 spray rernoval i
of particulate iodine from the containmenc atmosphere (Ref. 11, and (3) rate of unfiltered inleakage into the control rooru (Refs.12 anu 13.
Revised perfomance for the Control Roum Emergency HVAC was also ned in the rea nu lys is.
The revised LOCA doses, including effects of the reduced conservatisms and increased RWST boron concentration, meet the acceptance criteria.
j The NRC staff has reviewed the effect of the proposed increase in the RWST boron concentration on other fluid systems and LCCA doses and concludes that the licensee's assessments, as discussed above, are acceptable.
2.3 Technical Specifications TS chariges are required to incorporate the changes to a more positive MTC and increas-d buron concentr tions of the RWST and accumulators. All of these TS
]
changes are acceptable. per the discussion in the preceding evaluation section.
These chariges are discussed belcw.
i i
.=
-_=,
O Specification 3/4.1.1.2 New shutdown margin curves as a function of reactor coolant system boron t
i concentration have been generated for Modes 3, 4, and 5.
These new curves, given by Figures 3.1-1 and 3.1-2, are based on a higher reactur coolant system borun concentration, a reduction in the assumed dilution flow rate, and a reduction in the high flux at shutdown alarm setpoint.
Specification and Basis 3/4.1.1.3 The MTC is changed tr,be less positive than + 7 pcm/deg F for the all rods withdrawn, beginninr, of life condition for power levels up to 70% of rated i
thermal power with a linear ramp to O pcm/d.eg F at 100% rated thermal power.
Specifications and Bases 3/4.1.2.5 and 3/4.1.2.6 The RWST miniram and maximum boron concentrations are changed, the minimum is increased fecm 2000 to 2400 ppm, the maximum is increased frcm 2100 or 2200 to 2600 ppm. ihe minimum RWST volume is increased from 70,832 to 99,404 gallons for Modes 5 and 6.
The Ba %s are changed to require the RWST to have 118,182 gallons of borated 4
water at a boron concentration of 2400 ppm for Modes 1 through 4.
For Modes 5 and o, the Bases are changed to req..re the RWST to have 41,202 gallons of berated water with a boron concerntration of 2400 ppm.
The lower limit pH of the recirculated borated water solution during a LOCA is changed from 8.5 to 8.0 in the Bases.
j Specification and Basis 3/4.3.1 r
i The Source Range High Flux at Shutdown Alarm Setpoint in Table 4.3-1, Footnote 9 was changed from 3.16 to 2.3 times background.
A stater:ent was added to the Bases to state that this setpoint is an assumption of the boron dilution event in Modes 3, 4, and 5.
Specification and Basis 3/4.5.1 I
)
The maximum boron concentration for the accumulators is increased from 2100 to l
2600 ppm.
i
)
A statement was added to the Bases to state that the minimum boroa concentration of the accumulators is required to maintain the reactor i
suberitical during the accumulator injection period of a small break LOCA.
Specification and Basis 3/4.5.4 J
The RWST minimum and maximum boron concentrations are changed, the minimum is increased from 2000 to 2400 ppm, and the maxinom is ircreased from 2100 to 2600 ppm.
f.
4 The lower limit pH of the recirculated borated water solution during a LOCA is changed from 8.5 to 8.0 in the Bases.
Basis 3/4.6.2 The Bases lower the pH value from 8.5 to 8.0 of the recirculated solution within containnent af ter a LOCA.
B_ asis 3/4.9.1 The licensee provided a clarification of the change to this basis in Reference 14 This change is acceptable.
2.4
SUMMARY
The NRC stalY has reviewed the submittal for the operation of Vogtle 1 with a more positive MTC and with increased bo.sn concentrations for the RWST and aCCumulaturs.
Based on its review, the NRC stoff Concludes that appropriate material was submitted and that the transients and accidents that were evaluated and reanalyzed are acceptable.
The TS changes submitted for this license amendment suitably reflect the necessary rrodifications for the operation of Vogtle 1 for extended cycles with a more positive HTC.
Therefore, the hRC staff finds that the propored amendment is acceptable.
3.0 EhVIRONMENTAL CONSIDERATION This amendment involves a change in the use of a facility component located within the restricted area as defined in Part 20 and changes in surveillance requirements. The staff has determitiea that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents tnet may be released offsite, and that there is no significant increase in individual or curnulative occupational exposure.
The NRC staff has made a determination that the emendment invulves no significant hazards consideration, and there has Leen no public comment en such finding. Accordingly, the anendment neets the eligibility criteria tur categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environa. ental impact statement or environmental assessment need be prepared in cor.nuction with the issuance of the ameidaent.
4.0 CONCLUSION
The Corraission made a proposed determination that the amendment involves no significant hazards consideration which was published in the Federal Register on June 29, 1988 (53 FR 24509), and consulted with the state of Georgia.
No
~
public comments were received, and the state of Georgia did not have any corraents.
The staff has canclu#.d. based on the considerations discussed above, that: (1) there is reasonabic ' assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulat. ions, and the issuance of this a a ndment will not be inimical to the conmon defense and security or to the health and safety of the public.
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5.0 REFERENCES
1.
Letter (SL-4682) from R. P. McDcnald (GPC) to USNRC, dated May 19, 1988.
2.
"Positive Moderator Temperature Coefficient and RWST/Accumclator Boron Concentration Increase Licensing Report for Vogtle Electric Generating Plant Units 1 and 2," Westinghouse Electric Corporation, April 1988 (this report was provided as Enclosure 1 to Reference 1).
I 3.
Letter from C. Rossi (NRC) to A. Ladieu (Westinghouse Owners Group),
dated March 30, 1987 (letter approved methodology of WCAP-11731 for performing SGTR analysis).
4 Burnett, T.W.T., et al., "LOFTRAN Code Description," WCAP-7907-P-A (Proprietary), WCAP-7907-A (Non-Proprietary), April 1984.
5.
Risher, D.H., Jr. and Barry, R.F., "TWINXLE - A Multidimensional Neutron Kinetics Computer Code," WCAP-7979-P-A (Proprietary) and WCAP-8028-A (Non-Proprie ta ry). 1975.
6.
Hargrove, ii.G., "FACTRAN - A Fortran -IV Code for Thermal Transients in a U0 Fuel Rud," 1972 (Westinghouse Topical Report).
2 7.
Hochreiter, L.E., Chelemer, H. ; and Chu, P.T., "Subchannel Thermal Analysis of Rud Bundle Cores," WCAP-7015, Revision 1,1969.
8.
Chelemer, H.; Weisman J.; and Tong L.S., "Subchannel Thermal Analysis of Roo Bundle Cores," WCAP-7015, Revision 1, 1969.
9.
"American National Standard for Decay Heat Power in Light Water l
Reactors," ANSI /ANS-5.1-1979, August 1979.
- 10. Branch Technical Position MTEB 6-1, "pH for Emergency Coolant Water for PWR's."
11.
"Technological Bases for M9 del of Spray Washout of Airborne Contaminants in Containment Vessels," NUREG/CR-009, October 1978.
12.
Georgia Power letter to NRC (Bailey to Youngblood), GN-808, February 19, 1986, 13.
"Assumptions for Evaluating the Habitability of a Nuclear Power Plant Control Room Ouring a Postulated Hazardous Chemical Release," Regulatory Guide 1.78, June 1974, 14.
Letter from W. G. Hairston, III (GPC) to NRC, dated August 12, 1988.
15.
Letter from W. G. Hairston, III (GPC) to NRC, dated October 3, 1988,
- 16. Telecon between J. Swartzwelder (GPC) and J. Hopkins (NRC) on October 4, 1988.
Principal Contributors: Jon B. Hopkins, PDI 3/ORP-!/II Daniel B. Tieno, SRXB/ DEST DateJ: October 4, 1968