ML20215N030
| ML20215N030 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 10/30/1986 |
| From: | Mcneil S Office of Nuclear Reactor Regulation |
| To: | Tiernan J BALTIMORE GAS & ELECTRIC CO. |
| References | |
| NUDOCS 8611040188 | |
| Download: ML20215N030 (11) | |
Text
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October 30, 1986 Distribution Docket Nos. 50-317 f"DocketsFile2
.JPartlow and 50-318 NRC PDR TBarnhart (4)
Local PDR WJones PBD-8 Reading EButcher FMirt. glia.
NThompson Mr. J. A. Tiernan _
PKreutzer WRegan Vice President-Nuclear Energy SAMcNeil ACRS(10)
Baltimore Gas & Electric Company OGC-Bethesda OPA P. O. Box 1475 LHarmon-EWW360A LFMB Baltimore, MD 21203' EJordan BGrimes
Dear Mr. Tiernan:
Gray File 3.2a Your submittal dated July 31, 1986, requested a change in the Departure from Nucleate Boiling Ratio (DNBR) limit cited in the Technical 5pecification (TS)
Bases sections B 2.1 and B 3/4.2.5 for Calvert Cliffs Unit 1.
Your request would change _the DNBR from 1.23 to 1.21.
The value of 1.23 was an interim value that was placed in the TS Bases of Calvert Cliffs Unit 1 while the review of the Combustion Engineering, Inc.
(CE) Topical Report CENPD-207, "C-E Critical Heat ~ Flux:
Critical Heat Flux Correla. tion for C-E Fuel Assemblies with Standard Spacer Grids; Part 2 - Non-
-uniforn Axial Power Distribution," was in progress.
The Commission, in its letter to CE dated November 2, 1984, found this report to be acceptable.
This change in the DNBR limit was made by the Commission in its letter to you dated November 21, 1985.
This modification of the DNBR limit reflects only' changes in the Unit 1 core-physics that were previously approved for cycle 7 of Unit 1 in Amendment No. 104 to Facility Operating License No. DPR-53, issued on May 20, 1985.
Additionally, this change has no effect upon any Unit 1 TS safety limit, safety margin, limiting condition for operation or surveillance requirements as it solely modifies the TS Bases.
Therefore, this
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change only represents.an administrative change of the TS Bases of Calvert Cliffs Unit 1 and as such, does not require the issuance of an Amendment to Facility Operating License No. DPR-53.
Corrected Unit 1 Technical Specification pages with corresponding overleaf pages are enclosed for your convenience.
Sincerely,
/s/
86110 $ $ h h 17 Scott A. McNeil, Project Manager
{DR PDR PWR Project Directorate #8 Division of PWR Licensing-B y [1 j(
Enclosures:
_n TS pages B 2-1 through B 2-6 and B 3/4 2-2 g.
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i cc:
See next page
/f PBD-h:
l PBD-8:
PBD-8:
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'tzer SA%il:jch AThadani P.//86 1 / 9 86 10/3l ll & &q6 f
1 10/f7/86 4
Mr. J. A. Tiernan-Baltimore Gas & Electric Company Calvert Cliffs Nuclear Power Plant cc:
Mr. William T. Bowen, President Regional Administrator, Region I Calvert County Board of U.S. Nuclear Regulatory Commission Commissioners Office of Executive-Director Prince Frederick, Maryland 20768 for Operations 631 Park Avenue D. A. Brune, Esq.
King of Prussia, Pennyslvania 19406 General Counsel Baltimore Gas and Electric Company P. O. Box 1475 Baltimore, Maryland 21203 Jay E. Silberg Shaw, Pittman, Potts and Trowbridge 2300 N Street, N.W.
Washington, DC 20037 Mr. M. E. Bowman,' General. Supervisor Technical Services Engineering Calvert Cliffs Nuclear Power Plant MD Rts 2 & 4, P. O. Box 1535 Lusby, Maryland 20657-0073 Resident Inspector c/o U.S. Nuclear Regulatory Commission P. O. Box 437 Lusby, Maryland 20657-0073 Bechtel Power Corporation ATTN: Mr. D. E. Stewart Calvert Cliffs Project Engineer 15740 Shady Grove Road l
Gaithersburg, Maryland 20760 t
Combustion Engineering, Inc.
ATTN: Mr. W. R. Horlacher, III Project Manager P. O. Box 500 1000 Prospect Hill Road Windsor, Connecticut 06095-0500 Department of Natural Resources Energy Administration, Power Plant Siting Program ATTN:
Mr. T. Magette Tawes State Office Building Annapolis, Maryland '21204
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r l-October 30, 1986 Distribution Docket Nos. 50-317 Docket File JPartlow E
and 50-318 NRC PDR TBarnhart (4)
Local PDR WJones PBD-8 Reading EButcher FMiraglia NThompson
'Mr. J. A. Tiernan PKreutzer.
WRegan Vice President-Nuclear Energy SAMcNeil ACRS(10)
Baltimore Gas & Electric Company OGC-Bethesda 0PA P. O. Box ~14?5 LHarmor. EWW360A LFMB Baltimore, MD 21203 EJordan BGrimes
Dear Mr. Tiernan:
Gray File 3.2a Your submittal dated July 31, 1986, requested a change in the Departure from Nucleate Boiling Ratio (DNBR) limit cited in the Technical Specification (TS)
Bases sections B 2.1 and B 3/4.2.5 for.Calvert Cliffs Unit 1.
Your request would change the DNBR from 1.23 to 1.21.
The value of 1.23 was an interim value that was placed in the TS Bases of Calvert Cliffs Unit 1 while the review of the Combustion Engineering, Inc.
(CE) Topical Report CENPD-207, "C-E Critical Heat Flux: -Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids;!Part 2 - Non-uniform Axial Power Distribution," was in progress.
The Commission, in its letter to CE dated November 2, 1984, found this report to be acceptable.
This change in the DNBR limit was made by the Commission in its letter to you dated November 21, 1985.
This modification-of the DNBR limit reflects only changes in the Unit 1 core physics that were previously approved for cycle 7 of Unit 1 in Amendment No. 104 to Facility Operating License No. DPR-53, issued on May 20, 1985.
Additionally, this change has no effect upon any Unit 1 TS safety limit, safety margin, limiting condition for operation or surveillance requirements as it solely modifies the TS Bases.
Therefore, this change only represents an administrative change of the TS Bases of Calvert Cliffs Unit 1 and as such, does not require the issuance of an Amendment to Facility Operating License No. DPR-53.
Corrected Unit 1 Technical Specification pages with corresponding overleaf pages are enclosed for your convenience.
Sincerely,
/s/
Scott A. McNeil, Project Manager PWR Project Directorate #8 Division of PWR Licensing-B g.j[
Enclosures:
,u TS pages B 2-1 through B 2-6 and B 3/4 2-2 93 T-ri o
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cc:
See next page
/T PBD-8:
PBD-8:
CG PBD-3:
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AThadani PKr)e(/86 10/
10/19/86 1
86 10/}l/6 d
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2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding and possible cladding perforation which could result in the release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the steady state peak linear heat ~ rate at or less than 22.0 kw/ft.
Centerline fuel melting will not oc:ur for this peak linear heat rate. Overheating of the fuel cladding is prevented by rastricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling reoime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient.
DNB is not a directly measurable parameter during operation and therefore THERIML POWER and Reactor Coolant Temper-ature and Pressure have been related to DNB through the CE-1 correlation.
The CE-1 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distri-butions. The' local' DNB heat flux ratio, DNBR, defined as the ratio off I
the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.
The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.21.
I This value corresponds to a 95 percent probability at a 95 percent con-fidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.
I j
The curves of Figures 2.1-1, 2.1-2, 2.1-3 and 2.1-4 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and maximum cold leg temperature of various pump combinations for which the minimum DNBR is no less than 1.21 for the family of axial shapes and I
corresponding radial peaks shown in Figure B2.1-1.
The limits in Figures l
2.1-1, 2.1-2, 2.1-3 and 2.1-4 were calculated for reactor coolant inlet temperatures less than or equal to 580 F.
The dashed line at 580*F coolant inlet temperature is not a safety limit; however, operation above 580*F is not possible because of the actuation of the main steam line safety valves which limit the maximum value of reactor inlet temperature.
Reactor operation at THERMAL POWER levels higher than 110% of RATED THERMAL POWER is prohibited by the high power level trip setpoint specified in CALVERT CLIFFS - UNIT 1 B 2-1 Amendment No. 33,fS,7J,$$,10/30/86 1
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SAFETY LIMITS BASES Table 2.1-1.
The area of safe operation is below and to the left of these lines.
The conditions for the Thermal Margin Safety Limit curves in Figures 2.1-1, 2.1-2,- 2.1-3 and 2.1-4 to be valid are shown on the figures.
The reactor protective system in combination with the Limiting Conditions fcr Operation, is designed to prevent any anticipated combination of transient conditions for reactor coolant system temperature, pressure, and THERMAL POWER level that would result in a DNBR of less than 1.21 and preclude the existence of flow instabilities.
l 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 pressure vessel and pressurizer are designed to Section III,1967 Edition, of the ASME Code for Nuclear Power Plant Components which permits a maximum transient pressure of 110% (2750 psia) of design pressure. The Reactor Coolant System piping, valves and fittings, are designed to ANSI B 31.7, Class I, 1969 Edition, which permits a maximum transient pressure of 110% (2750 psia) of component design pressure.
The Safety Limit of 2750 psia is therefore consistent with the design criteria and associated code requirements.
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The entire Reactor Coolant System is hydrotested at 3125 psia to demonstrate integrity prior to initial operation.
1 l
I CALVERT CLIFFS - UNIT 1 B 2-3 Amendment No. 33,39,fE,7J,10/30/86
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2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 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 paramt.ter. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits. Operation with a trip set less ceaservative than its Trip Setpoint but within its spect-fied Allowable Value_is acceptable on the basis that the difference between the trip setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.
Manual Reactor Trip The Manual Reactor Trip -is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.
Power Level-High The Power Level-High trip provides reactor core protection against reactivity excursions which are too rapid to be protected by a Pressurizer Pressure-High or Thermal Margin / Low Pressure trip.
The Power Level-High trip setpoint is operator adjustable and can be set no higher than 10% above the indicated THERMAL POWER level. Operator action is required to increase the trip setpoint as THERMAL POWER is increased. The trip setpoint is automatically decreased as THERMAL power decreases. The trip setpoint has a maximum value of 107.0% of RATED THERMAL POWER and a minimum setpoint of 30% of RATED THERMAL POWER.
Adding to this' maximum value the possible variation in trip point due to calibration and instrument errors, the maximum actual steady-state l
THERMAL POWER level at which a trip would be actuated is 110% of RATED THERMAL POWER, which is the value used in the safety analyses.
Reactor Coolant Flow-Low The Reactor Coolant Flow-Low trip provides core protection to prevent DNB in the event of a sudden significant decrease in reactor coolant Provisions have been made in the reactor protective system to permit flow.
A CALVERT CLIFFS - UNIT 1 B 2-4 Amendment No. As,71
-. =. - - - -... - -
LIMITING SAFETY SYSTEM SETTINGS BASES operation of -the reactor at reduced power if one or two reactor coolant pumps are taken out of service. The low-flow trip setpoints and Allowable Values for the various reactor coolant pump combinations have been derived in consideration of instrument errors and response times of equipment involved to maintain the DNBR above 1.21 under normal operation and expected transients.1 For reactor operation with only two or three reactor coolant pumps operating, the Reactor Coolant Flow-Low trip setpoints, the Power Level-High trip set-points, and the Thermal Margin / Low Pressure trip setpoints are automatically changed when the pump condition selector switch is manually set to the desired two-or three-pump position. Changing these trip setpoints during two and three pump operation prevents the minimum value of DNBR from going below 1.21 l
during normal operational transients and anticipated transients when only two or three reactor coolant pumps are operating.
Pressurizer Pressure-High The Pressurizer Pressure-High trip, backed up by the pressurizer code safety valves and main steam line safety valves, provides reactor coolant system protection against overpressurization in the event of loss of load without reactor trip. This trip's setpoint is 100 psi below the nominal lift setting (2500 psia) of the pressurizer code safety valves and its concurrent operation with the power-operated relief valves avoids the undesirable opera-tion-of the pressurizer code safety valves.
Containment Pressure-High The Containment Pressure-High trip provides assurance that a reactor trip is initiated prior to, or at least concurrently with, a safety injection.
Steam Generator Pressure-Low The Steam Generator Pressure-Low trip provides protection against an excessive rate of heat extraction from the steam generators and subsequent cooldown of the reactor coolant. The setting of 685 psia is sufficiently below the full-load operating point of 850 psia so as not to interfere with normal operation, but still high er.ough to provide the required protec-tion in the event of excessively high steam flow. This setting was used with an uncertainty factor of + 85 psi which was based on the main steam line break event inside containmentT l
CALVERT CLIFFS - UNIT 1 B 2-5 Amendment No. 37,#$,77,$E, yy7,10/30/86 l
i LIMITING SAFETY SYSTEM SETTINGS BASES Steam Generator Water Level The Steam Generator Water Level-Low trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity and assures that the pressure of the reactor coolant system will not exceed its Safety Limit. The specified setpoint in combination with the auxiliary
.feedwater actuation system ensures that sufficient water inventory exists in both steam generators to remove decay heat following'a loss of main feedwater flow event.
Axial Flux Offset The axial flux offset trip is provided to ensure that excessive axial peaking will not cause fuel damage. The axial flux offset _is determined from the axially split excore detectors. The trip setpoints ensure that_neither a DNBR of less than 1.21 nor a peak linear heat I
rate which corresponds to the temperature for fuel centerline melting will exist as a consequence of axial power maldistributions. These trip setpoints were derived from an analysis of many axial power shapes with allowances for instrumentation inaccuracies and the uncertainty associated with the excore to'incore axial flux offset relationship.
Thermal Margin / Low Pressure The Thermal Margin / Low Pressure trip is provided to prevent operation when the DNBR is less than 1.21.
l The trip is initiated whenever the reactor coolant system pressure signal drops below either 1875 psia or a computed value as described below, whichever is higher. 'The ' computed value~ is ~a-function' of the higher of AT power or neutron power, reactor inlet temperature, and the number of reactor coolant pumps operating. The minimum value of-reactor coolant flow rate, the maximum AZIMUTHAL POWER TILT and the maximum CEA deviation permitted for continuous operation are assumed in the genera-tion of this trip function.
In addition, CEA group sequencing in accor-dance with Specifications 3.1.3.5 and 3.1.3.6 is assumed.
Finally,-the maximum insertion of CEA banks which can occur during any anticipated operational occurrence prior to a Power Level-High trip is assumed.
CALVERT CLIFFS - UNIT 1 B 2-6 Amendment No. 33,39,A$,7LBE,10/30/86
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l 3/4.2. POWER DISTRIBUTION LIMITS BASES 3/4.2.1 LINEAR HEAT RATE The limitation on linear heat rate ensures that in the evgnt 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 Excore Detector Monitoring System and the Incore Detector Monitoring System, provide 1
adequate monitoring of the core power distribution and are capable of verify-ing that the linear heat rate does not exceed its limits. The Excore Detector i
Monitoring System performs this function by continuously monitoring the AXIAL SHAPE INDEX with the OPERABLE quadrant symmetric excore neutron flux detectors and verifying that the AXIAL SHAPE INDEX is maintained within the allowable
+
limits of Figure 3.2-2.
In conjunction with the use of the excore monitoring system and in establishing the AXIAL SHAPE INDEX limits, the following assump-tions are made: 1) the CEA insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are satisfied, 2) the AZIMUTHAL POWER TILT restrictions of Specifica-tion 3.2.4 are satisfied, and 3) the TOTAL PLANAR RADIAL PEAKING FACTOR does not exceed the limits of Specification 3.2.2.
The Incore Detector Monitoring System continuously provides a direct measure of the peaking factors and the alarms which have been established for the individual incore detector. segments ensure that the peak linear heat rates will be maintained _within the allowable limits of Figure 3.2-1.
The setpoints for these alarms include allowances, set _ in the conservative directions, for
- 1) a measurement-calculational uncertainty factor of 1.062, 2) an engineering uncertainty factor of 1.03, 3) an allowance of 1.002 for axial fuel densifica-tion and thermal expansion, and 4) a THERMAL _ POWER measurement uncertainty i
factor of 1.02.
3/4.2.2. 3/4.2.3 and 3/4.2.4 TOTAL PLANAR AND INTEGRATED RADIAL PEAKING 4
FACTORS-Fjy ANDFpANDAZIMUTHALPOWERTILT-T q
The limitations on FT an'd T are provided to ensure that the assumptions used in the analysis for E tablis0ing the Linear Heat Rate and Local Power Density - High LCOs and LSSS setpoints remain valid during operation at the various allowable CEA group insertion 1..aits. The limitations on FT and Tg are provided to ensure that the assumptions used~in the analysis establishing the DNB Margin LCO, and Thermal Margin / Low Pressure LSSS setpoi-t1 remain valigdu ng operation at the various allowable CEA group insertion limits.
If Fx,F or Tq exceed their basic limitations, operation may continue under J
i theadditonalrestrictionsimposedbytheACTIONstatementssincethese additional restrictions provide adequate provisions to assure that the assump-i tions used in establishing the Linear Heat Rate, Thermal Margin / Low Pressure and Local Power Density - High LCOs and LSSS setpoints rer:ain valid. An r
AZIMUTHAL POWER TILT > 0.10 is not expected and if it should occur, subsequent operation would not be restricted to only those operations required to identify the cause of this unexpected tilt.
i i
i CALVERT CLIFFS - UNIT 1 B 3/4 2 1 Amendment No. 33,39,104 i
--- w = - -
wu l-POWER DISTRIBUTION LIMITS BASES T
The value of T that must be used in the equation F*# = F*# (1 + T ) and q
9 F[ = F O + T ) is the measured tilt.
r q
i The surveillance requirements for verifying that FT FJandT afe withintheirlimitsprovideassurancethattheactualVIYu,esofFly)fuelloa F and T i
q do not exceed the assumed values.
Verifying FTyandFJaftereach prior to exceeding 75% of RATED THERMAL POWER provides additional assurance that the core was properly loaded.
3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the param-l eters are maintained within the normal steady state envelope of. operation
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assumed in the transient and accident analyses. The limits are consistent i
with the safety analyses assumptions and have been analytically demonstrated j-adequate to maintain a minimum DNBR of 1.21 throughout each analyzed transient. l
-In addition to the DNB criteria, there are two other criteria which set I
the specification in Figure 3.2-4.
The second criteria is to ensure that the existing core power distribution at full power is less severe than the power distribution factored into the small-break LOCA analysis. This results in a limitation on the allowed negative AXIAL SHAPE INDEX value at full power.
The third criteria is to maintain limitations on peak linear heat rate at i
low power levels resulting from Anticipated Operational Occurrences (A00s).
i-Figure 3.2-4 is used to assure the LHR criteria for.this condition because l
the linear heat rate LCO, for both ex-core and in-core monitoring, is set to l
maintain only the LOCA kw/ft requirements which are limiting at high power l~
(e.g., CEA withdrawal), tend to become more limiting than that for LOCA.
levels. At reduced power levels, the kw/ft requirements of certain A00s The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters through instru-ment readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.
The 18 month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis.
CALVERT CLIFFS - UNIT 1 B 3/4 2-2 Amendment No.M.M,55,77,Jpf,10/30/86 l
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