ML20115A863
| ML20115A863 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 10/05/1992 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20115A859 | List: |
| References | |
| NUDOCS 9210150138 | |
| Download: ML20115A863 (7) | |
Text
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SAFETY EVALVATION BLJHE OFFICE OF NUCLEAR REACTOR REGVLATit>d RLLATED TO AMENDf1ENT N0.138 TO FACILITY OPERATING LICENSE NO. NPF-6 ENTER 6Y OPERATIONS. INC..
ARKANSAS NUCLEAR ONE. UNIT NO. 2 DOCKET NO. 50-368
1.0 INTRODUCTION
By [[letter::0CAN079208, Informs That,Per 10CFR50.71 & 10CFR50.59(a)(3),Amend 10 to Arkansas Nuclear One Units 1 & 2 & Rev 15 to QA Manual Operations Shipped Under Cover Ltr (1CAN079204, 2CAN079210 & 0CAN079210)|letter dated July 22, 1992]], as supplementau by le'. tars dated September 11 and 14, 1992, Entergy Operations, Inc. (the licensee) submitted a request for changes to the Arkansas Nuclear One, Unit No. 2 (ANO-2) Technical Specifications (TSs).
The requested changes would increase the albwable pressurizer pressure range and also allow a lower low pressurizer pressure setpoint for reactor trip, safe'.y injection, and containment cooling. The revisions would change Technical Specification (TS) 3.2.8 and associated Bases to allow plant operation with pressurizer pressure between 2025 and 2275 psia.
TS 2.1.1 Bases md be clarified with regards to the applicatien of the peak linear heat rate (PLHR) limit to anticipated operational occurrences analysis results.
Also, this revision would lower the ANO-2 TS Table 2.2-1 (and associated Bases) reactor protection low pressurizer pressure trip setpoint and allowable values to 1717.4 and 1686.3 psia respectively.
The safety injection and containment cnoling actuation system (CCAS) trip setpoi'.it and allowable values given in ANO-2 TS Table 3.3-4 would be lowered to 1717.4 and 1586.3 psia respectively by this revision.
Tue September 11 and 14, 1992, letters provided clarifying information that did not change the initial proposed no significant hazards consideration determination.
2.0 BACKGROUND
The ANO-2 plant in the past has had small amounts of safety valve leakage below the 1 uable TS limit 'or peactor Coolant System (RCS) leakage, which was consi@r,d insignificant.
Also, continuous plant operatiun runs were sufficientb short and plant shutdowns and forced outages due to other reasons allowed the safety valves to be replaced or repaired if necessary.
The ANO-2 plant performance has improved as indicated by longer continuous plant operating times.
Recently, this problem alone has resulted in plant shutdown for valve replacement or repair. Although modifications have been made to the safety valves and the valves' discharge piping in an attempt to minimize or eliminate valve simmering / leakage, these changes have not eliminated the problem.
Initially, valve simmering is characterized by low volume, high 921015013e 92100S PDR ADOCK 0500 P
2 velocity, saturated steam leakage acrsss the _ valve seats.
prolonged simmering tends to allow an increased steam volume. Continued eAgosure at the higher volume and_ velocity will typically cause seat damage, and once seat damage occurs, the valve cannot be reseated into a leak tight condition. Once seat damage occurs the leakage rate tends to increase with time in an exponential manner, eventually crusing forced plant shutdown with the attendant economic impacts and associated safety concerns.
By reducing the operating pressurizer pressure for a short time period when the valve starts to simmer, it is expected that the safety valve simmering problem can be substantially curtailed or eliminated, thus eliminating subsequent valve leakage.
By operating the RCS at reduced pressures, the safety valves are given a chance to reach a thermal equilibrium point at a pressure with sufficient margin to the valve lift setpoint (2500 +1,-3% psia) to avoid simmering.
15 3.2.8 specifies the RCS operating pressure bounds and maintains only an approximate 10% margin to the safety valve lift setpoint.
Small perturbations in the valve thermal equilibrium point can initiate valve simmering.
It is postulated that when the perturbations occur, if_the RCS pressure is -aduced for a sufficient titt period to allow the valve to reestablish an equilibrium point, sir.mering and valve leakage can be terminated.
This involves a significant pressure reduction to approximately 2025 psia for short durations.
To ensure-the valve (s) remain leaktight, it may be appropriate to maintain continuous operation at a small pressure reduction (to approximately 2150 psia). By avoiding the simmering, valve seat damage is also precluded which enhances the valve reliability and lengthers the life of the valve.
3.0 TVALVATION To avoid simmering of the safety valve, the licensee preposed.to revise the ANO-2 TS 3.2.8, Pressurizer Pressure Limiting Condition for Operation (LCO) to allow plant operatiun in Mode 1 with pressurizer oressure between 2025 psia and 2275 psia (current TS range of values is between 2225 psia and 2275 psia).
These lower pressure limits are co'aistent with other CE plants (Palo Verde, San Onofre, Waterford).
In additaa, this proposed change sould reduce the low pressurizer pressu c ^< actor Protection System (RPS), Safety Injection Actuation System (SIAS), and Containment Cooling Actuation System (CCAS) trip setpcint and allowable values to 1717.4 and 1686.3 psia, respectively (current TS values are 1766 psia and 1712.757 psia respectively).
The evaluation is covered in three parts as follows; 1) the pressurizer pressure reduction justification - based on the Safety Analysis Report (SAR)
Chapter 15 evaluations and plant safety system Core Protection Calculator l
(CPC) and Core Operating Limit Supervisory System (COLSS) range verification,
- 2) the clarification to the PLHR TS Bases, and 3) the proposed low pressurizer pressure RF3, SIAS, and CCAS trip sespoint and allowable value changes.
3.1 Pressurizer Pressure Reduction The safety analyses supporting the Chapter 15 FSAR presently bound plant operation with pressurizer pressure between 2200 and 2300 psia. These l
l.
3 analyses were reviewed by the licensee to identify which would be adverself affected by plant operation at a lower pressurizer pressure.
Four events were found to be affected by lower pressurizer pressure:
- 1) Loss of External Load / Turbine Trip, 2) Uncontrolled Control Element Assembly (CEA) Withdrawal,
These events were reanalyzed assuming an iiitial pressurizer pressure of 2000 psia.
The plant re m ses to these events were simulated using thc NRC approved CESEC-Ill compuer code.
Departure f rom nucleate boiling ratio (DNBR) analyscs were performed based on the TORC computer code, the CE-1 critical heat flux correlation, and the CETOP code for which the Cycle 5 DNBR limit of 1.25 is applicable for the current anal p es.
The results, ding approved codes and critical heat correlation, were verified to be within the accepti.nce criteria including the specified acceptable fuel design limits (SAFDLs).
No fuel cladding damage is predicted for any event, therefore. no changes to the ra'iological doses were calculated.
The results of the four events analyzed are presented below.
Loss of External load / Turbine Trip The loss of External Load / Turbine trip bounding analysis was performed using an initial RCS prenure of 2000 psia.
The event was reanalyzed with con ervative assumptions.
The results included a peak RCS pressure of 2744 psia and str.am generator pressure of 1160 psia.
These values are within 110 percent of the design limit pressores of 2750 psia and 1210 psia for the RCS and steam ger.erator, respectively, and are therefore acceptable.
Uncontrolled CEA Withdrawal The uncontrolled CEA bank withdrawal event from both subcritical and 1% power s
was evaluated at an RCS pressure of 2000 psia.
It is noted that tne 100X power case is not adversely affected by the lower RCS pressure due to credit taken for the CPC low DNBR trip which is valid over the proposed pressurizer pressure range.
Ti.e subcritical CFA sank withdrawal evaluation uses conservative assumptions for the Cycle 10 analysis.
The acceptable fuel design limits (DNBR equal or greater than 1.25 and fuel centerline temperature below 4900 degrees f) were easily met with a great deal of margin.
Conservative assumptions for the CEA withdrawal fror 1% power event were used in the analysis.
The Variable Over Poser Trip (V0PT) is the first trip encountered and terminatos the reactor power excursion at a lower level than prc. Husly calculated.
The results show that the minimum DNBR remains above 1.0 md the maximum linear heat rate remains below 17 kw/f t.
We find this acc >able as these values are each within the acceptan e criteria of 1.25 and 21 k,.;f t, respectively.
CEA Ejection I
The CEA ejection events from Hot full Power (HFP) and Hot Zet o Power (HZP) were both reevaluated utilizing the new lower RCS pressure limit of 2000 psia.
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4 The STRIKIN-II computer program was used to simulate the heat conduction within a reactor fuel rod (ad its associated surface heat transfer.
Ccnservative assumptions were used in the CEA ejection analysis.
The maximum centerline enthalpy decreased for both cs.es.
However there was a slight increase in the number of fuel pins (0.32%) having incipient centerline melting for the HFP case, but no fuel pins were calculated as having clad dam se or fully molten centerline.
Therefore we consider the results from this evaluation as acceptable.
Single Part length CEA Drop A single part length CEA (PLCEA) drop incident was reevaluated by the licensee to determine the effects of a reduction of RCS pressure to 2000 psia.
Only positive reactivity insertions resulting from a PLCEA drop are of concern, With the PLCEA insertion limits impnsed by Rection 3.1.3.7 of the ANO-2 Technical Specifications, positive reactivity iracrtier.s can only be postulated for PLCEA drops below 50% power. A reduction in the initial pressure can delay the high pressurizer pressure trip, thereby allowing a greater power increase, and a correspondingly larger decrease in the fuel thermal margin. However, sufficient initial thermai margin will be preserved by the COLSS, which is verified every cycle in the reload analyses, to assure that the DNBR SAFDL is met thrcughout the PLCEA drop event.
The licensee used conserutive assumptions for the PLCEA drop, which produce the maxirr.am power.
increase that avoids the high pressurizer pressure trip.
lhe data and algorithms or the CPCs were verified by the licensee to be valid for a range of pressurizer pressures which cover the proposed. operating pressure range.
An uncertainty factor is applied in the COL 5S calculation for DNBR to account for instrument uncertainty on the measured parameters used as inputs to the COLSS colculations.
The 9ncertainty factors were reviewed by the licensee and verified to be conservative over the proposed expanded pressurizer pressure rang? down to 2000 psia.
3.2 PLHR Clarification The results of the uncontrolled CEA bank withdrawal from subcritical conditions ana"ysis by the licensee indicated a transient pea! linear heat rate (PLHR) in excess of the 21 kw/ft limit given in TS 2.1.1.2.
The calculated value was a PLHR of less than 28 kw/tt which exceeded 21 kw/ft for less than one second. The limit of 21 kw/ft is specified based on steady state operation fuel centerline melting temperat' ares. - Therefore, higher linear heat rates can occur under transient conditions without resulting in fuel melting.
The fuel centerline melt temperature was acceptable for the subcritical CEA bank withdrawal as indicated previously in a paragraph above.
The licensee has proposed a clarification to the Bases of TS 2.1.1.2 to ensure the appropriate application of the peak linear heat rate limit.
The linear heat rate and analysis results are typically given in kw/ft for use for monitoring by the CPCs. Therefore, the peak linear heat rate liinit of 21 kw/ft is appropriate for most situations. However, for anticipated operational occurrencas with transient peak linear heat rates the more apprcpriate limit is tN specified acceptat..e fuel design limit centerline melting temperaturt. which is the basis for tne peak linear heat rate.
We
~
5 find this clarification to be acceptable es it is consistent with other CE-plant interpretations of this Technical Specification (Maine Yankee, Waterford) and is also consistent with the interpretation utilized for the CPCs as documented in the methodology and software manuals.
3.3 Lower low Pressurizer Pressure Setpoint The effects of operating with a lower low pressurizer pressure setpoint were examined relative to taadvertent operation of the Emergency Core Cooling System (ECCS)_during power operation. Oparating with the pressurizer pressure-below approximately 2150 pria may result in an undesirable SIAS following a.
reactor trip from significant power levels. The probability of an inadvertent actuation of an SIAS during power operation is not increased, but the likelihood of receiving an SIAS following a reactor trip is increased during reduced pressure operation.
Post-trip pressure response from reduced initial pressurizer pressure is expected to be comparab;e to higher pressure trips,-
with the minir 7 post-trip pressure being correspondingly lower and closer to the SIAS setp
.it.
As a result, a lower low pressurizer pressure trip s;tpoint of 1717.4 psia (reduced from 1776 psia) is proposed.
Safety analyses identified above as being adversely impacted by the reduced pressurizer pressure were reevaluated down to 2000 psia and found acceptable.
The proposed limit of 2025 psia is based on the analysis assuration of 2000 psia plus 25 psi which bounds pressure measurement uncertainties.
The basis for the 25 psi error was explained in a letter dated September 11, 1992.
Due to the potential for an undesirable SIAS actuu on following reactor trip when i
operating below 2150 psia, operation in the range between 2025 psia to 2150 psia for short durations (24_hnurs) will be administratively controlled.
This will ellow operator flexibility when attempting to reseat simmering pressuriier code safety valves and yet minimize the exposure in an undesirable SIAS actuation.
We find this acceptable based on the supporting analyses and.
the permissible administrative control.
The licensee also proposed reouctions in the low pressurizer pressure RPS and Engineered Safety Feature Actuation-System (LSFAS) trip setpoint and allowable values. These reductions are to help prevent an undesirable SIAS following a reactor trip when operating at reduced pressures.
The 1ew low pressurizer pressure setpoints and allowable values were based on new instrument crror c icu'.ations; the safety analysis setpoint assumptions were not changed. The calculations support the proposed low pressurizer pressure setpoint of 1717.4 psia and the allowable value of 1686.3 psia for the.RPS, SIAS and CCAS trip functions.
Regarding the new calculations, the licensee used the statistical method of the square root of the sum of squores (SSRS) to determine the sum of randem errors in irdividual components, and in the complete loop.
The licensee combined non-random errors algebraically with the sum of random errors to establish the total uncertainty of the instrument loop.
The licensee removed from the ' calculation the terms for all seismic errces and for the cumulative effects of background radiation.
The licensee statrd that the concurrent occurrence of an accident and a seismic activity is beyand the
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design basis of the plant.
The licensee also implemented comprehensive procedures at the plant to assess the effects of seismic activity immediately after it occurs and tnat the plant operators are trained for these procedures.
The licensee stated that cumulative effects of background radiation is calitrated out during each calibration and effects of cumulative dose of the background,adiation far a period bctween the two successive calibrations is very negligible.
Therefore, the licensee did not consiler the cumulative eff ects of background radiation in the calculation. The staff found these I
explanations acceptable.
lhe licensee used a currently accepted methodology for calculating setpoints.
In addition, the licensee usos this methodology for calculation ?l-EQ~2002-02, Revision 0, " Loop Error, Setpoint, and Time Response Analysis for Narrow Range Containment Building Pressure ESFAS and PSS Trip Functions," which the staff recently approved for another TS amendment. The licensee has not committed to strict compliance with the guidance in ISA-S67.04-1988 "Setpoints for Nuclear Safety-Related Instrumentation Used in Nuclear Power Plants." However, the licensee considered the guidelines of this ISA standard in calculating the 100; orrors, periodic test errors, and allowable vcl'v3 associated with the low pressurizer pressure setpoints.
The calculation methodology is acceptable to the staff.
4.0 STATf_(ONSULTATION In accordance with the Commission's regulations, the Arkansas State official was notified of the proposed issuance of the amendment.
The State official had no comments.
5.0 ENVIRONMt NT,ALinNSIDERAT10N The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 23. -The NRC staff has determined that the amendment involves no signi'icant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that thera is no significant increase in individual or cumu'ative occupational radiation exposure.
The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has Deen nc public comment on such finding (57 FR 37567). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based un the considerations discussed above, that:
(1) there is reasonable assurance that the !.ealth and safety of the public will not be endangered by operation in the proposed manner, (2) suc5
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7 activities will te conducted in compliance with the commission's regulations, and-(3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: ;1, Balukjian, f;RX8
. Athavale, SICB Date: October 5, 199g F
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