ML20073S391
| ML20073S391 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/23/1994 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Metropolitan Edison Co, Jersey Central Power & Light Co, Pennsylvania Electric Co, GPU Nuclear Corp |
| Shared Package | |
| ML20073S392 | List: |
| References | |
| DPR-50-A-184 NUDOCS 9406030028 | |
| Download: ML20073S391 (14) | |
Text
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3 UNITED STATES 5'
NUCLEAR REGULATORY COMMISSION 1
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WASHINGTON, D.C. 20555-0001
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METROPOLITAN EDIS0N COMPANY JERSEY CENTRAL POWER & LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY l
l GPU NUCLEAR CORPORATION i
l DOCKET NO. 50-289 THREE MILE ISLAND NUCLEAR STATION. UNIT NO. I i
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 184 License No. DPR-50 l
1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by GPU Nuclear Corporation, et al.
(the licensee), dated November 30, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the l
Act), and the Commission's rules and regulations set forth in 10 CFR I
Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission-C.
There is reasonable assurance (i) that the activities author 1 zed by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part.51 of the Commission's regulations and all applicable requirements have been satisfied.
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9406030028 940523 PDR ADOCK 0500 9
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2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.c.(2) of Facility Operating License No.
DPR-50 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.184, are hereby incorporated in the license.
GPU Nuclear Corporation shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance, to be implemented within 30 days of issuance.
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FOR THE NUCLEAR REGULA ORY COMMIS$ ION i
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t ohn T. Stolz, Director r ect Directorate I-4 i
vision of Reactor Projects - I/II l
Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications
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Date of Issuance: May 23, 1994 t
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ATTACHMENT TO LICENSE AMEN 0 MENT N0.184 r:ACILITY OPERATING LICENSE NO. DPR-50 DOCKET NO. 50-289 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
Remove Insert vii vii 2-1 2-1 2-2
- 2 2-3 2-3 2-4b 2-4b 2-4c 2-4c 2-5 2-5 2-6 2-6 2-7 2-7 2-10 2-10 2-12 2-12 l
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LIST OF FIGURES FIGURE TITLE-PAGE 2.1-1 Core Protection Safety Limit THI-l 2-4a 2.1-2 DELETED 2.1-3 Core Protection Safety Bases THI-l 2-4c 2.3-1 TMI-1 Protection System Maximum Allowable Setpoints 2-11 2.3-2 DELETED 3.1-1 Reactor Coolant System Heatup/Cooldown Limitations 3-!a (Applicable thru 10 EFPY) l 3.1-2 Reactor Coolant Inservice Leak and Hydrostatic Test 3-5b (Applicable thru 10 EFPY) 3.1-2a Dose equivalent I-131 Primary Coolant Specific Actual 3-9b Limit vs. Percent of RATED THERMAL POWER l
3.1-3 Limiting Pressure vs Temperature Curve for 3-18b l
100 STD cc/ Liter H 0 3
3.5-2A thru DELETED 3.5-2M 3.5-1 Incore Instrumentation Specification 3-39a l
Axial Imbalance Indication 3.5-2 Incore Instrumentation Specification 3-39b Radial Flux Tilt Indication 3.5-3 Incore Instrumentation Specification 3-39c l
3.11-1 Transfer Path to and from Cask Loading Pit 3-56b 4.17-1 Snubber Functional Test - Sample Plan 2 4-67 5-1 Extended Plot Plan TMI N/A l
l 5-2 Site Topography 5 Mile Radius N/A 5-3 Gaseous Effluent Release Points and L'iquid Effluent N/A 0utfall Locations 5-4 Relationship Between Initial Enrichment and Acceptable 5-7a Fuel Burnup (Spent Fuel Pool A - Region II) vii Amendment Nos. JJ, J7, 79, 39, (5, 59, 59, 72, 199, JSP, J29, JES, J M,
JfE, 159, 19%, 197, 7%#, 184, 1
l
2.
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS. REACTOR CORE Aprilicability Applies to reactor thermal power, axial power imbalance, reactor coolant system pressure, coolant temperature, and coolant flow during power operation of the plant.
Objective To maintain the integrity of the fuel cladding.
Specification 2.1.1 The combination of the reactor system pressure and coolant temperature shall not exceed the safety limit as defined by the locus of points established in Figure 2.1-1. If the actual pressure / temperature point is below and to the right of the line, the safety limit is exceeded.
2.1.2 The combination of reactor thermal power and axial power imbalance (power in the top half of core minus the power in the bottom half of the core expressed as a percentage of the rated power) shall not exceed the protective limit as defined by the locus of points (solid line) for the specified flow set forth in the Axial Power Imbalance Protective Limits given in the Core Operating Limits Report (COLR). If the actual-reactor-thermal-power / axial-power-imbalance point is above the line for the specified flew, the protective limit is exceeded.
Bases To maintain the 10tugrity of the fuel cladding and to prevent fission product release, it is necessary to prevent overheating of the cladding under normal operating conditions.
This is accomplished by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is large enough so that the clad surface temperature is only slightly greater than the coolant temperature.
The upper boundary of the nucleate boiling regime is termed, departure from nucleate boiling (DNB).
At this point there is a sharp reduction of the heat transfer coefficient, which could result in excessive cladding temperature and the possibili o of cladding failure. Although DNB is not an observable parameter during reactor operation, the observable parameters of neutron power, reactor coolant flow, temperature, and pressure can be related to DNB through the use of a critical heat flux (CHF) correlation.
The BAW-2 (Reference 1) and BWC (Reference 2) rorrelations have been developed to predict DNB and the location of DNB fo* 6xially uniform and non-uniform heat flux distributions.
The BAW-2 correlatL n applies to Mark-B fuel and the BWC correlation applies to Mark BZ type fuel.
The local DNB ratio (DNBR), defined as the ratio of the heat flux that would csuse DNB at a particular core location to the actual heat flux, is indicative of the margin to DNB.
The minimum value of the DNBR, during steady-state operation, normal 2-1 Amendment No. J7, JR, J57,184,
operational transients, and anticipated transients is limited to 1.30 (BAW-2) and 1.18 (BWC). A DNBR of 1.30 (BAW-2) or 1.18 (BWC) corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur; this is considered a conservative margin to DNB for all operating conditions.
The difference between the actual core outlet pressure and the indicated reactor coolant. system pressure has been considered in determining the core protection safety limits.
The curve presented in Figure 2.1-1 represents the conditions at which the minimum allowable DNBR or greater is predicted for the limiting combination of thermal power and number of operating reactor coolant pumps.
This curve is based on the nuclear power peaking factors given in Reference 3 and the COLR which define the reference design peaking condition in the core for operation at the maximum overpower. Once the reference peaking condition and the associated thermal-hydraulic situation has been established for the hot channel, then all other combinations of axial flux shapes and their accompanying radials must result in a condition which will not violate the previously established design criteria on DNBR. The flux shapes examined include a wide range of positive and negative offset for steady state and transient conditions.
These design limit power peaking factors are the most restrictive calculated at full power for the range from all control rods fully withdrawn to maximum allowable control rod insertion, and form the core DNBR design basis.
4 The Axial Power Imbalance Protective Limits curves in the COLR are based on the l
more restrictive of two thermal limits and include the effects of potential fuel densification and fuel rod bowing:
The DNBR limit produced by a total nuclear power peaking factor a.
consisting of the combination of the radial peak, axial peck, and position of the axial peak that yields no less than the DNBR limit.
b.
The maximum allowable local linear heat rate that prevents central fuel melting at the hot spot as given in the COLR.
Power peaking is not a directly observable quantity and therefore limits have been established on the basis of the axial power imbalance produced by the oower peaking.
2-2 Amendment No. J7, 59, 99, J2$, JJR, 157, 184,
The specified flow rates for curves 1, 2, and 3 of the Axial Power Imbalance Protective Limits given in the COLR correspond to the expected minimum flow rates with four pumps, three pumps, and one pump in each loop, respectively.
The curve of Figure 2.1-1 is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in Figure 2.1-3.
The curves of Figure 2.1-3 represent the conditions at which the DiB limit is predicted at the maximum possible thermal power for the number of reactor coolant pumps in operation or the local quality at the point of minimum DNBR is equal to 22 percent, (BAW-2), or 26 percent (BWC) whichever condition is more restrictive.
The maximum thermal power for each reactor coolant pump operating condition (four pump, three pump and one pump in each loop) given in the COLR is due to a power level trip produced by the flux-flow ratic multiplied by the minimum flow rate for the given pump combination plus the maximum calibration and instrumentation error.
Using a local quality limit of 22 percent (BAW-2), or 26 percent (BWC) at the point of minimum DNBR as a basis for curves 2 and 3 of Figure 2.1-3 is a conservative criterion even t5ough the quality at the exit is higher than the quality at the point of mininium DNBR.
The DNBR as calculated by the BAW-2 or BWC correlation continually increases from the point of minimum DNBR, so that the exit DNBR is always higher and is a function of the pressure.
For each curve of Figure 2.1-3, a pressure-temperature point above and to the left of the curve would result in a DNBR greater than 1.30 (DAW-2) or 1.18 (BWC) or a local quality at the point of minimum DNBR less than 22 percent (BAW-2), or 26 percent (BWC) for the particular reactor coolant pump situation.
Curve 1 is more restrictive than any other reactor coolant pump situation because any pressure / temperature point above and to the left of this curve will be above and to the left of the other curves.
REFERENCES (1) UFSAR, Section 3.2.3.1.1
" Fuel Assembly Heat Transfer Design" (2) BWC Correlation of Critical Heat Flux, BAW-10143P-A, Babcock & Wilcox, Lynchburg, Virginia, April 1985 (3) UFSAR, Section 3.2.3.1.1.3
" Nuclear Power Factors" 4
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2-3 Amendment No. J7, 29, JP, 59,129, JES, JJE, JSP, J57,184,
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l 120
(-43.8.112) 1 (37.8.112) 3 i
ACCEPTABLE i
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OPERATION 100
(_ -43.8,89.3) 2 (37.8,89.3)
AC EPT
(-58.5.30.4) 80 (53.3.30.C l
ERATION
(_ -43 0) 3 (37.3,62.0)
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ACCE N LE 60 (53.0,57.3) 2,3, & 4 PU OPERATION 4 40 i
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(-58.5.30.4)
(53.0.30.4) i j
- 20 4
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- 70 50 30 10 0
10 20 30 40 50 60 70 80 1
l Axial Power Irnbelance, 5 i
l Curve Reactor Coolant Flow (1b/hr) 6 1
139.8 x 10 6
2 104.5 x 10 0
i 3
68.8 x 10 1
i CORE PROTECTION SAFETY LIMITS TMI-1 Figure 2.1-2 2-4b i
Amendment No. J7, ES, SS, (5, 59, J29, JES, JJE, 157, 184, J
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Power Pumps Operating (Type of Limit) 1 139.8 x 10' (100%)*
112%
Four Pumps (DNBR Limit) 2 See COLR See COLR Three Pumps (Quality Limit) 3 See COLR See COLR One Pump in Each Loop (Quality Limit)
- 106.5% of Cycle 1 Design Flow f
II CORE PROTECTION SAFETY BASES THI-1 Figure 2.1-3 2-4c Amendment No. $9, JJ%, J92, Jp7,184, i
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2.3 LIMITING SAFETY SYSTEM SETTINGS. PROTECTION INSTRUMENTATION Applicability Applies to instruments monitoring reactor power, axial power imbalance, reactor coolant system pressure, reactor coolant outlet temperature, flow, number of pumps in operation, and high reactor building pressure.
Objective i
To provide automatic protection action to prevent any combination of process i
variables from exceeding a safety limit.
Specification 9
2.3.1 The reactor protection system trip setting limits and the permissible bypasses for the instrument channels shall be as stated in Table 2.3-1 and the Protection System Maximum Allowable Setpoints for Axial Power i
Imbalance as given in the COLR.
Bases The reactor protection system consists of four instrument channels to monitor each of several selected plant conditions which will cause a reactor trip if any one of these conditions deviates from a pre-selected operating range to the i
degree that a safety limit may be. reached.
1 The trip setting limits for protection system instrumentation are listed in i
Table 2.3-1.
These trip setpoints are setting limits on the setpoint side of the protection system bistable comparators.
The safety analysis has been based upon these protection system instrumentation trip' set points plus calibration and instrumentation errors.
l Nucler.r Overpower 1
i A reactor trip at high power level (neutron flux) is provided to prevent damage to the fuel cladding from reactivity excursions too rapid to be detected by j
pressure and temperature measurernents.
During normal plant operations with all reactor coolant pumps operating, reactor trip is initiated when the reactor power level reaches 105.1% of rated power.
Adding to this the possible variation in trip set points due to calibration and instrument errors, the maximum actual power at which a trip would be actuated could be 112%, which is the value used in the safety analysis (Reference 1).
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2-5 i
Amendment No. JJ,17, ES,12%, JJ2, J57,184, l
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a.
Overpower trip based on flow and imbalance The power level trip set point produced by the reacto coolant system i
flow is based on a power-to-flow ratio which has been established to accommodate the most severe thermal transient considered in the destgr, the loss-of-coolant flow accident from high power. Analysis has demonstrated that the specified power to flow ratio is adequate to prevent a DNBR of less than 1.30 (BAW-2) or 1.18 (BWC) should a 16w flow condition exist due to any malfunction.
The power level trip set point produced by the power-to-flow ratio provides both high power level and low flow protection in the event the reactor power level i
increases or the reactor coolant flow rate decreases.
The power level trip set point produced by the power to flow ratio provides overpower DNB protection for all modes of pump operation. For every flow rate there is a maximum permissible 4
power level, and for every power level there is a minimum permissible low flow rate. Typical power level and low flow rate combinations for the pump situations of Table 2.3-1 are given in the COLR.
The flux / flow ratios account for the mtximum calibration and instrumentation errors and the maximum variation from ti,e average value of the RC flow signal in j
such a manner that the reactor protective system receives a conservative indication of the RC flow.
No penalty in reactor coolant flow through the core was taken for an open core vent valve because of the core vent valve surveillance program during each refueling outage.
For safety analysis calculations the maximum calibration and instrumentation errors for the power level were used.
The power-imbalance boundaries are established in order to prevent reactor thermal limits from being exceeded.
These thermal limits are either power peaking Kw/ft limits or DNBR limits. The axial power imbalance (power in the top half of the core minus power in 2-6 Amendment No. J), J7, ES, ES, JP. 49, JES, JJE, J57, 184, l
l the bottom half of core) reduces the power level trip produced by the power-to-flow ratio so that the boundaries of the Protection System Maximum Allowable Setpoints for Axial Power Imblance in the COLR are produced.
b.
Pump Monitors The redundant pump monitors prevent the minimum core DNBR from decreasing below 1.30 (BAW-2) or 1.18 (BWC) by tripping the reactor due to the loss of reactor coolant pump (s). The pump monitors also restrict the power
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level for the number of pumps in operation.
c.
Reactor coolant system pressure During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure trip setpoint is reached before the nuclear overpower trip setpoint. The trip setting limit shown in Figure 2.3-1 for high reactor coolant system pressure ensures that the system pressure is maintained below the safety limit (2750 psig) for any design transient (Reference 2). Due to calibration and instrument errors, the i
safety analysis assumed a 45 psi pressure error in the high reactor coolant system pressure trip setting.
1 As part of the post-THI-2 accident modifications, the high pressure trip setpoint was lowered from 2390 psig to 2300 psig.
(The FSAR Accident Analysis Section still uses the 2390 psig high pressure trip setpoint.)
The lowering of the high pressure trip setpoint and raising of the setpoint for the Power Operated Relief Valve (PORV), from 2255 psig to 2450 psig, has the effect of reducing the challenge rate to the PORV l
while maintaining ASME Code Safety Valve capability.
l-A B&W analysis completed in September of 1985 concluded that the high reactor coolant system pressure trip setpoint cedld be raised to 2355 psig with negligible impact on the frequency of opening of the PORV during anticipated overpressurh ion transieni.s (Reference 3). The high pressure trip setpoint was subsa.ently raised to 2355 psig.
The potential safety benefit of this action is a seduction in the frequency of reactor trips.
The low pressure (1800 psig) and variable low pressure (11.75 T.,-5103) trip setpoint were initially established to maintain the DNB ratio greater than or equal to 1.3 for those design accidents that result in a pressure reduction (References 4, 5, and 6). The B&W generic ECCS analysis, however, assumed a low pressure trip of 1900 psig and, to establish conformity with this analysis, the low pressure trip setpoint has been raised to the more conservative 1900 psig. Application of the B&W 2-7 Amendment No. 17, 2), 77, (5, 75,12%, JJ5, JJ2, J57,184, l
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2.3-1 REACTOR PROTECTION SYSTEM TRIP SETTING LIMITS (5)
Four Reactor Coolant Three Reactor Coolant One Reactor Coolant Pumps Operating Pumps Operating Pump Operating in (Nominal Operating)
(Nominal Operating)
Each Loop (Nominal Shutdown Power - 100%
Power - 75%
Operating Power - 49%1 Bypass
- 1. Nuclear power, max.
105.1 105.1 105.1 5.0(2)
% of rated power i
- 2. Nuclear power based on Power / Flow Setpoint Power / Flow Setpoint Power / Flow Setpoint Bypassed flow (1) and imblance in COLR times flow in COLR times flow in COLR times flow v
max. of rated power minus reduction due minus reduction due minus reduction due to imbalance to imbalance to imbalance
- 3. Nuclear power based NA NA 55%
Bypassed (4) on pump monitors max. % of rated power
- 4. High reactor coolant 2355 2355 2355 1720(3) system pressure, psig max.
- 5. Low reactor coolant 1900 1900 1900 Bypassed system pressure, psig min.
- 6. Reactor coolant temp.
618.8 618.8 618.8 618.8 F., max.
- 7. High Reactor Building 4
4 4
4 pressure, psig max.
(1)
Reactor coolant system flow, %.
(2)
Administratively controlled reduction set during reactor shutdown.
(3)
Automatically set when other segments of the RPS (as specified) are bypassed.
(4)
The pump monitors also produce a trip on:
(a) loss of two reactor coolant pumps in one reactor coolant loop, and (b) loss of one or two reactor coolant pumps during two-pump operation.
(5)
Trip settings limits are limits on the setpoint side of the protection system bistable connectors.
Amendment No. O, 7S, M, Uli, pp, W,184, 2-10
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, 70 50 30 10 0 10 20 30 40 50 60 70 80 Axial Power Imbalance, 1 PROTECTION SYSTEM MAXIMUM ALLOWABLE SETPOINTS FOR l
AXIAL POWER IMBALANCE THI-1 Figure 2.3-2 2-12 Amendment No. 17, 29, D, 9, H, 59, J29, JES, JJE, 197. 184,
.