ML20206U454
| ML20206U454 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 05/17/1999 |
| From: | Nolan M NRC (Affiliation Not Assigned) |
| To: | Hutchinson C ENTERGY OPERATIONS, INC. |
| References | |
| TAC-MA2398, NUDOCS 9905250225 | |
| Download: ML20206U454 (10) | |
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UNITED STATES C
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NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D.C. 3066H001 May 17, 1999 i
Mr. C. Randy Hutchinson Vice President, Operations ANO Entergy Operations, Inc.
1448 S. R. 333 Russellville, AR 72801
SUBJECT:
VARIOUS TECHNICAL SPECIFICATION BASES CHANGES FOR ARKANSAS NUCLEAR ONE, UNIT NO. 2 (TAC NO. MA2398) i
Dear Mr. Hutchinson:
By letter dated July 13,1998 (2CAN079803), as supplemented by letter dated May 10,1999 (2CAN059903), Entergy Operations, Inc. (EOI) submitted changes to Facility Operating License No. NPF-6, Appendix A - Technical Specifications (TS) Bases section. The first change adds a footnote to TS Bases Section 2.1.2, " Reactor Coolant System Pressure," to indicate that later editions to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code), Section Ill, may be utilized provided that these provisions are reconciled back to the original construction code. TS Bases Section 2.1.2 lists the year and addenda for the construction code of record for various reactor coolant system components. The footnote provides clarifying information to the TS Bases by recognizing the reconciliation process provided in the ASME Code,Section XI, to address the use of later editions of the ASME Code.
It should be noted that Title 10 of the Code of Federal Reaulations (10 CFR) Section 50.55a requires that the use of requirements set forth in editions and addenda of the ASME Code that have been issued subsequent to those editions and addenda incorporated by reference in 10 CFR 55.55a(b) is subject to Commission approval.
The change to TS Bases Section 2.2.1, " Reactor Trip Setpoints, Steam Generator Pressure -
Low," would remove the reference to the specific value for the nominal stearn generator pressure at full load. EOl indicated that nominal full load steam generator pressure can vary from cycle to cycle as primary and secondary plant parameters are optimized to best manage the performance of the steam generators. TS Bases Section 2.2.1 stillindicates that the steam generator pressure - low trip will be set sufficiently below the full load operating point so as not l
to interfere with normal operations, but still high enough to provide the required protection in the
/
event of excessively high steam flow.
The change to TS Bases Section 2.2.1," Reactor Trip Setpoints, Departure from Nucleate Boiling Ratio (DNBR) - Low," would raise the integrated radial peaking factor - high limit from M
4.28 to 7.00. EOl credited an evaluation demonstrating the acceptability of this change. In g) addition, the integrated radial pe ng factor - high limit is not credited in any safety analysis X
event.
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The change to TS Bases Section 3/4.2.1, Power Distribution Limits, Linear Heat Rate," and Section 3/4.2.4, " Power Distribution Limits, DNBR [ Departure from Nucleate Boiling Ratio]
Margin," would remove the specific discussion of the uncertainty factors used in the Core Operating Limit Supervisory System (COLSS) methodology and replace it with a more generic discussion. EOl indicated that these uncertainties are evaluation dependent and can vary from 9905250225 990517 DR ADOCK 050003 8
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C. R. Hutchinson
-2 May 17, 1999 cycle to cycle. Additionally, this change is consistent with the guidance contained in Revision 1 of NUREG-1432," Standard Technical Specifications for Combustion Engineering Plants."
The change to TS Bases Section 3/4.6.2.3," Containment Cooling System," would correct a typographical error introduced in Amendment No.154.
The change to TS Bases Section 3/4.6.3, " Containment Isolation Valves," would remove the specific reference to the plant procedure number for the location of the listing of containment isolation valves and replace it with a more generic reference to plant procedures. This change will provide flexibility to allow modifications and revisions to the plant procedure structure without effecting the technical content of the information contained in the TS Bases section.
EOl has eva!uated each change pursuant to 10 CFR 50.59 and determined that the changes do not involve an unreviewed safety question. The staff has no objection to these Bases changes.
Enclosed are the affected Bases pages, B 2-2, B 2-4, B 2-7, B 3/4 2-1, B 3/4 2-3, and B 3/4 6-4.
Sincerely, ORIGINAL SIGNED BY M. Christopher Nolan, Project Manager, Section 1 Project Directorate IV & Decommissioning l
Division of Licensing Project Management Office of Nuclear Reactor Regulation
(
Docket No. 50-368 DISTRIBUTION Docket File ACRS
Enclosure:
Corrected Bases pages PUBLIC OGC PDIV-1 Reading G. Hill (2) cc w/ encl: See next page M.C.Nolan(2)
K.Brockmari,RIV L.Hurley,RIV J.Kilcrease,RIV W.Beckner S.Richards J.Zwolinski/S. Black DOCUMENT NAME: G:\\ANO2\\LTRA2398.WPD
- See Previous Concurrence To retoive a copy of this document, Indicate in the box: "C" = Copy without enclosures "E" = Copy with enclosures "N" = No copy OFFICE PDIV-1/PM.
E PDIV-1/LAtw SRXB/BC E
EMEB/BC DSQ
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NAME CNolan #
LBerry W
JWermiel*
Elmbro @
X 05/ { !99
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04/06/99 05/// /99 7 /99 N DATE 05/ // /99 OFFICE PDIV-1/SC 6.
NAME RGramm (L&
DATE 05/ n /99 OFFICIAL RECORD COPY i
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Arkansas Nuclear One ec:
Executive Vice President Vica President, Operations Support
& Chief Operating Officer Entergy Operations, Inc.
Entergy Operations, Inc.
P. O. Box 31995 P. O. Box 31995 Jackson, MS 39286-1995 Jackson, MS 39286-199 Wise, Carter, Child & Caraway Director, Division of Radiation P. O. Box 651 Control and Emergency Management Jackson, MS 39205 Arkansas Department of Health 4815 West Markham Street, Slot 30 Little Rock, AR 72205 3867 Winston & Strawn 1400 L Street, N.W.
Washington, DC 20005-3502 Manager, Rockville Nuclear Licensing Framatone Technologies 1700 Rockville Pike, Suite 525 Rockville, MD 20852 Senior Resident inspector U.S. Nuclear Regulatory Commission P. O. Box 310 London, AR 72847 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 County Judge of Pope County Pope County Courthouse Russellville, AR 72801 May 1999
4 BAFETY LIMITS AND LIMITIN 3 SAFETY SYSTEM SETTIN"5 RASES Limiting safety system settings for the Low DNBR, High Local Power Density, High Logarithmic Power Level, Low Pressurizer Pressure and High l!
Linear Power Level trips, and limiting conditions for operation on DNBR and kw/ft margin are specified such that there is a high degree of confidence i
that the specified acceptable fuel design limits (i.e., DNBR and centerline fuel melt temperature) are not exceeded during normal operation and design basis anticipated operational occurrences.
2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
The Reactor Coolant System components are designed to Section III of the ASME Code for Nuclear Power Plant Components. The reactor vessel, stema generators sad pressurizer are designed to the 1968 Edition, Summer 1970 Addendas piping to the 1971 Edition original issue; and the valves to the 1968 Edition, Winter 1970 Addenda"8.,Section III of this code permits a l
maximum transient pressure of 110% (2750 psia) of design pressure. The safety Limit of 2750 psia is therefore consistent with the design criteria and associated code requirements.
The entire Reactor Coolant System is hydrotested at 3125 psia to demonstrate integrity prior to initial operation.
2.2.1 REACTOR TRIP SETPOINTS The Reactor Trip setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety Limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.
The DNBR - Low and Local Power Density - High are digitally generated trip setpoints haced on Limiting safety system settings of 1.25 and 21.0 kw/ft, respectively. Since these trips are digitally generated by the core Protection Calculators, the trip values are not subject to drifts common to trips generated by analog type equipment. The Allowable values for these trips are therefore the same as the Trip setpoints.
4 "I
Use of a later ASME Section III Code is acceptable, provided the Code section(s) is reconciled in accordance with section XI.
ARKANSAS - UNIT 2 B 2-2 Amendment No. 44.44.-79,446,-
Revised by NRC Imtter dated May 17, 1999 mm+
e m w e
l SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES Pressurizer Pressure-High The Pressurizer Pressure-High trip, in conjunction with the pressurizer safety valves and main steam safety valves, provides reactor coolant system protection against overpressurization in the event of loss of load without reactor trip. This trip's setpoint is at 5 2370.887 psia l
which is below the nominal lift setting (2500 psia) of the pressurizer safety valves and its operation avoids the undesirable operation of the pressurizer safety valves.
Pressurizer Pressure-Low l
1 The Pressurizer Pressure-Low trip is provided to trip the reactor and l
to assist the Engineered Safety Features System in the event of a Loss of Coolant Accident.
During normal operation, this trip's setpoint is set at l
l 2 1686.3 psia. This trip's setpoint may be manually decreased, to a i
adnimum value of 100 psia, as pressurizer pressure is reduced during plant shutdowns, provided the margin between the pressurizer pressure and this trip's setpoint is maintained at 5 200 psis this setpoint increases automatically as pressurizer pressure increases until the trip setpoint is reached.
l Containment Pressure-High 4
The Containment Pressure-High trip provides assurance that a reactor trip is initiated concurrently with a safety injection. The setpoint for this trip is identical to the safety injection setpoint.
Steam Generator Pressure-Low The Steam Generator Pressure-Low trip provides protection against an I
excessive rate of heat extraction fram the steam generators and subsequent cooldown of the reactor coolant. The setpoint is sufficiently below the full load operating point so as not to interfere with normal operation, but l
still high enough to provide the required protection in the event of excessively high steam flow. This trip's setpoint may be manually decreased as steam generator pressure is reduced during plant shutdowns, prov;Aed the margin between the steam generator pressure and this trip's setpoint is maintained at 5 200 psi; this setpoint increases automatically as steam generator pressure increases until the trip setpoint is reached.
/=d=ut No. 49, H8 ARKANSAS - UNIT 2 3 2-4 Revised by NRC Letter dated May 17, 1999 l
. =. - - - -,
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS RASE 5 q
a.
RCS Cold Leg Temperature-Low 2490*F b.
RCS Cold Leg Temperature-High 5585*F j
c.
Axial Shape Index-Positive Not were positive than +0.6 i
d.
Axial Shape Index-Negative Not more negative than -0.6
)
e.
Pressurizer Pressure-Low s1785 psia f.
Pressurizer Pressure-High
$2415 psia g.
Integrated Radial Peaking Factor-Low 2 1.28 h.
Integrated Radial Peaking Factor-High
$7.00 l
1.
Quality Margin-Low 50 Steam Generator Level - High The Steam Generator Level - High trip is provided to protect the turbine from excessive moisture carry over.
Since the turbine is automatically tripped when the reactor is tripped, this trip provides a reliable means for providing protection to the turbine from excessive moisture carry over. This trip's setpoint does not correspond to a Safety Limit and no credit was taken in the accident analyses for operation of this trip.
Its functional capability at the specified trip setting is required to enhance the overall reliability of the Reactor Protection System.
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Amad=:6t No. 24, 42, ??, 99 j
ARKANSAS - UNIT 2 5 2-7 Revised by NRC Letter dated May 17, 1999 l
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3/4.2 POWER DISTRIBUTION LIMITS EASES l
3/4.2.1 LINEAR HEAT RATE The limitation on linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 2200*F.
Either of the two core power distribution monitoring systems, the Core operating Limit Supervisory System (COL 55) and the Local Power Density channels in the Core Protection Calculators (CPCs), provide adequate monitoring of the core power distribution and are capable of verifying that the linear heat rate does not exceed its 10mits. The COLSs performs this function by continuously monitoring the core power distribution and calculating a core power operating limit corresponding to the allowable peak linear heat rate.
The COLs3 calculated core power and the COLS3 calculated core power operating limits based on linear heat rate are continuously monitored and displayed to the operator. A COLSS alamm is annunciated in the event that the core power exceeds the core power operating limit. This provides adequate margin to the linear heat rate operating limit for normal steady state operation. Normal reactor power transients or equipment failures which do not require a reactor trip may result in this core power operating limit being exceeded.
In the event this occurs, COLSS alarms will be annunciated.
If the event which causes the COLS5 limit to be exceeded results in conditions which approach the core safety limits, a reactor trip will be initiated by the Reactor Protective Instrumentation.
The COLSs calculation of the linear heat rate limit includes appropriate uncertainty and penalty factors necessary to provide a 95/95 confidence level that the maximum linear heat rate calculated by CoLss is greatar than or equal to that existing in the core. To ensure that the design margin to safety is maintained, the COLSS computer program includes uncertainties associated with planar radial peaking measurement, engineering design factors, state parameter measurement, software algorithm modeling, computer processing, rod bow, and core power measurement.
Parameters required to maintain the operating limit power level based on linear heat rate, margin to DNB and total core power are also monitored by the CPCs. Therefore, in the event that the COLSS is not being used, operation within the limits specified in the CORE OPERATING LIMITS REPORT can be maintained by utilizing a predetermined local power density margin and a total core power limit in the CPC trip channels. The above listed uncertainty and penalty factors are also included in the CPCs.
Amendment No. 24, 79, M7 ARKAN5A3 - UNIT 2 5 3/4 2-1 Revised by NRC Letter dated May 17, 1999
.-- p..
POWER DISTRIBUTION LIMITS RASES Ptilt/Puntilt is the ratio of the power at a core location in the presence of a tilt to the power at that location with no tilt.
3/4.2.4 DNBR MARGIN The limitation on DNBR as a function of AXIAL SHAPE INDEX represents a conservative envelope of operating conditions consistent with the safety analysis assumptions and which have been analytically demonstrated adequate to maintain an acceptable minimum DNBR throughout all anticipated operational occurrences.
Operation of the core with a DNBR at or above this limit provides assurance that an acceptable minimum DNBR will be maintained in the event of any anticipated operational occurrence.
l Either of chs two core power distribution monitoring systems, the Core Operating Limit Supervisory System (COLSS) and the DNBR channels in the Core Protection Calculators (CPCs), provide adequate monitoring of the core power distribution and are capable of verifying that the DNBR does not violate i
its limits. The COLSS performs this function by continuously monitoring the core power distribution and calculating a core operating limit corresponding i
I to the allowable udnimum DNBR. The COLSS calculation of core power operating limit based on DNBR includes appropriate uncertainty and penalty factors necessary to provide a 95/95 confidence level that the core power at which a DNBR of less than 1.25 could occur, as calculated by COLSS, is less than or equal to that which would actually be required in the core. To ensure that I
the design margin to safety is maintained, the COLSS computer program includes uncertainties associated with planar radial peaking measurement, I
engineering design factors, state parameter naasurement, software algorithm modeling, computer processing, rod bow, and core power measurement.
Parameters required to maintain the margin to DNS and total core power are also monitored by the CPCs. Therefore, in the event that the COLSS is not being used, operation within the limits specified in the CORE OPERATING LIMITS REPORT can be maintained by utilizing a predetermined DNBR as a function of AXIAL SHAPE INDEX and by monitoring the CPC trip channels. The above listed uncertainty and penalty factors are also included in the CPC.
A DNBR penalty factor has been included in the COLSS and CPC DNBR calculations to accommodate the effects of rod bow. The amount of rod bow in each assembly is dependent upon the average burnup experienced by that assembly.
Fuel assemblies that incur higher average burnup will experience a greater magnitude of rod bow.
Conversely, lower burnup assemblies will experience less rod bow.
In design calculations, the penalty for each batch required to compensate for rod bow is determined from a batch's maximum I
average assembly burnup applied to the batch's maximum integrated planar-radial power peak. A single not penalty for COLS$ and CPC is then determined 1
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knendnent No. 94, 96, 32, 66, 79, M7 ARKANSAS - UNIT 2 B 3/4 2-3 Revised by NRC Letter dated May 17, 1999
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i CONTAINMENT SYSTEMS l
aASES The SR 4.6.2.2.b requirement to dissolve a representative sample of TSP in a sample of borated water provides assarance that the stored TSP will i
dissolve in borated water at the postulated post-LOCA temperatures.
Testing must be performed to ensure the solubility and buffering ability of the TSP after exposure to the containment environment. A representative sample of 3.00 i 0.05 grams of TSP from one of the baskets in containment is submerged in 1.0 i 0.01 liter of water at a boron concentration of 3000 t.'O ppm and at a temperature of 12015'F. The solution is allowed to stcnd for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> without agitation. The liquid is then decanted from the solution and mixed, the temperature adjusted to 77 i 2'F and the pH measured. At this point, the pH must be 2 7.0.
The representative sample weight is based on the minimum required TSP weight of 6804 kilograms, which at manufactured density corresponds to the minimum volume of 278 cubic ft, and assumed post LOCA borated water mass in the sump of approximately 5284102 lbm normalized to buffer a 1.0 liter sample. The boron concentration of the test water is representative of the maximum possible boron concentration corresponding to the maximum possible post LOCA sump volume. Agitation of the test solution is prohibited, since an adequate standard for the agitation intensity cannot be specified. The test time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is necessary to allow time for the dissolved TSP to naturally diffuse through the sample solution.
In the post LOCA containment sump, rapid mixing would occur, significantly decreasing the actual amount of time before the required pH is achieved.
This would ensure compliance with the Standard Review Plan requirement of a pH 2 7.0 by the onset of recirculation after a LOCA.
3/4.6.2.3 CONTAINMENT COOLING SYSTEM The OPERABILITY of the containment cooling system ensures that 1) the containment air temperature will be maintained within limits during normal operation, and 2) adequate heat removal capacity is available when operated in conjunction with the containment spray systems during post-LOCA conditions.
The containment cooling system 3cd the containment spray system are redundant to each other in providing post accident cooling of the containment atmosphere. As a result of this redundaner in cooling capability, the allowable out-of-service time requiremen s for the containment cooling system have been appropriately adjusted.
However, the allowable out of service time requirements for the containment spray system have been maintained consistent with that assigned other inoperable ESF equipment since the containment spray system also provides a mechanism for removing Iodine from the containment atmosphere.
The addition of a biocide to the service water system is performed l
during containment cooler surveillance to prevent buildup of Asian clams in the coolers when service water is pumped through the cooling coils.
This is performed when service water temperature is between 60'r and 80*F since in this water temperature range Asian clams can spawn and produce larva which could pass through service water system strainers.
Aerdnt No. H4,194 ARKANSAS - UNIT 2 B 3/4 6-5 Revised by NRC Letter dated May 17, 1999
. CONTAINMENT SYSTEMS BASES 3/4.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment.
Containment isolation within the time limits specified ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA. The containment isolation valves have been relocated to plant procedures.
l The opening of locked or sealed closed manual and deactivated automatic containment isolation valves on an intermittent basis under administrative control includes the following considerations:
(1) stationing an operator, who is in constant communication with control room, at the valve controls, (2) instructing the operator to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to close the valves and that this action will prevent the release of radioactivity outside containment.
3/4.6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will be available to maintain the hydrogen concentration within containment below its flammable limit during post-LOCA conditions.
Either recombiner unit is capable of controlling the expected hydrogen generation associated with 1) zirconium-water reactions, 2) radiolytic decomposition of water, and 3) corrosion of metal within containment. These hydrogen control systems are consistent with the reconnendations of Regulatory Guide 1.7 " Control of Combustible Gas Concentrations in Containment Following a LOCA", March 1971.
The containment recirculation units are provided to ensure adequate mixing of the containment atmosphere following a LOCA. This mixing action will prevent localized accumulations of hydrogen from exceeding the flammable limit.
Amendment No. 694 ARKANSAS - UNIT 2 B 3/4 6-6 Revised by NRC Letter dated May 17, 1999
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