ML20236T861
| ML20236T861 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 11/18/1987 |
| 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 | |
| ML20236T863 | List: |
| References | |
| DPR-50-A-134 NUDOCS 8712020135 | |
| Download: ML20236T861 (9) | |
Text
..
, ff UNITED STATES y
9, NUCLEAR REGULATORY COMMISSION l
n
- j WASHINGTON, D. C. 20555
)
-\\...../
METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER AND LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY GPU NUCLEAP. CORPORATION DOCKET NO. 50-289 THREE MILE ISLAND NUCLEAR STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.134 License No. DPR-50 l
1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The application for amendment by GPU Nuclear Cor) oration, et al.
(the licensee) dated July 24, 1987, complies wit 1 the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I;
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B.
The facility will cperate in conformity with the application, the provisiens of the Act, and the rules and regulations of the Comission;
)
C.
There is reasonable assurance (1) that the activities authorized by this anendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issusnee o.* this amendment is in accordance with 10 CFR Part 51 of the Comiss!on's regulatio.is and all applicable requirements have been atisfied.
8712020135 871118' PDR ADOCK 05000289 p
PDR-
o 8
l P.
Accordingly, the license is amended by chances to the Technical Specifications as indicated in the attachment to this license
)
amendment, and paragraph 2.c.(2) of Facility Operatino License
]
No. DPR-50 is hereby amended to read as follows:
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Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.134, are I
hereby incorporated in the license.
GPU Nuclear i
Corporation shall operate the facility in accordance with the Technical Specifications.
l 3.
This license amendment is effective as of its date of issuance, i
FOR THE NUCLEAR P.EGULATORY COMMISSICM l
se/
8lML/Jff JohnF.Stolz,Diredo'r Pr6fect Directorate I-4 Division of Reactor Projects I/II i
Attachment:
Changes to the Technical Specifications Date of Issuance: NOV2stgg i
ATTACHfiENT TO LICENSE AMENDMENT NO. 134 FACILITY OPEP.ATING 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 3-3 3-3 3-4 3-4 3-5 3-5 Figure 3.1-1 Figure 3.1-1 Figure 3.1-2 Figure 3.1-2 l
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LIST OF FIGURES Figure Title 2.1 -1 TMI-1 Core Protection Safety Limit 2.1 -2 TMI-1 Core Protection Safety Limits 2.1-3 TMI-1 Core Protection Safety Bases
- 2. 3 -1 TMI-1 Protection System Maximum A11on'able Set Points 1
2.3-2 Protection System Maximum Allowable Set Points for Reactor Power I
Imbalance, THI-1 3.1 -1 Reactor Coolant System Heatup/Cooldown Limitations (Applicable to 10 EFPY) l 3.1 -2 Reactor Coolant System, Inservice Leak and Hydrostatic Test Limitations (Applicable to 10 EFPY) t l
3.1 -3 Limiting Pressure vs. Temperature Curve for 100 STD cc/ Liter-Hgo 3.5-2A Rod Position Limits for 4 Pump Operation from 0 to 30+10/-0 EFPD, TMI-1 l
3.5-20 Rod Position Limits for 4 Pump Operation from 30+10/-0 to 25010 l
EFPD, THI-1 1
3.5-2C Rod Position Limits for 4 Pump Operation after 25010 EFPD, TMI-1 3.5-2D Rod Position Limits for 3 Pump Operation from 0 to 30+10/-0 EFPD, TMI-1 3.5-2E Rod Position Limits for 3 Pump Operation from 30+10/-0 to 25010 EFPD, TMI-1 3.5-2F Rod Position Limits for 3 Pump Operation after 25010 EFPD, TMI-1 L
3.5-2G Rod Position Limits for 2 Pump Operation from 0 to 30+10/-0 EFPD, TMI-1 3.5-2H Rod Position Limits for 2 Pump Operation from 30+10/-0 to 25010 EFPD, TMI-1 3.5-21 Rod Position Limits for 2 Pump Operation after 25010 EFPD, TMI-1 3.5-2J Power Imbalance Envelope for Operation from 0 to 30+10/-0 EFPD, TMI-1 vii Amendment Nos. Pf, F/, $5, )dI, $$, yd, pd, J,19,1J 9,1/0,1l[6, d
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l?4
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l 3.1.2 PRESSURIZATION HEATUP AND C00LDOWN LIMITATIONS Applicability Applies to pressurization, heatup and cooldown of the reactor coolant system.
Objective
)
To assure that temperature and pressure changes in the reactor coolant system do not cause cyclic loads in excess of design for reactor coolant system components.
Specification l
3.1. 2.1 For operations until ten effective full power years, the reactor j
coolant pressure and the system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with l
Figure 3.1-1 and Figure 3.1-2 and are as follows:
Heatup/Cooldown Allowable combinations of pressure and temperature shall be to the right of and below the limit line in Figure 3.1-1.
Heatup and cooldown rates shall not exceed those shown on Figure 3.1-1.
l Inservice Leak and Hydrostatic Testing Allowable combinations of pressure and temperature shall be to the l
right of and below the limit line in Figure 3.1-2.
Heatup and cooldown rates shall not exceed those shown on Figure 3.1-2.
l l
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3.1.2.2 The secondary side of the steam generator shall not be pressurized j
above 200 psig if the temperature of the steam generator shell is l
below 100*F.
3.1.2.3 The pressurizer heatup and cooldown rates shall not exceed 100*F in any one hour.
The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater l
than 430*F.
3.1.2.4 Prior to exceeding ten effective full power years of operation, l
Figures 3.1-1 and 3.1-2 shall be updated for the next service period l
in accordance with 10 CFR 50, Appendix G, Section Y.B.
The highest I
predicted adjusted reference temperature of all the beltline materials shall be used to detennine the adjusted reference temperature at the end of the service period.
The basis for this prediction shall be submitted for NRC staff review in accordance with Specification 3.1.2.5.
3.1.2.5 The updated proposed technical specifications referred to in 3.1.2.4 shall be submitted for NRC review at least 90 days prior to the end of the service period.
Appropriate additional NRC review time shall be allowed for proposed technical specifications submitted in accordance with 10 CFR 50, Appendix G, Section V.C.
Amendment M, /A,134
Bases All reactor coolant system components are designed to withst the effects of.
cyclicloads'duetosystemtemperatureandpressurechanges.gnjThesecyclic
\\1 loads are introduced by unit load transients, reactor trips, and unit heatup and cooldown operations.
The number of thermal and loading cycles used for design purposes are shown in Table 4-8 of the FSAR.
The maxim
} unit heatup and cooldown rates satisfy stress limits for cyclic operation.
The l
200 psig pressure limit for the secondary side of the steam generator at a temperature less than 100*F Satisfies stress levels for temperatures below the NDTT.(31 The heatup and cooldown rate limits in this specification are based on linear heatup and cooldown ramp rates which by analysis have been extended to accommodate 15'F step changes at any time with the appropriate soak (hold) i times. Also, an additional 15*F step change has been included in the analysis with no additional soak time to accommodate decay heat initiation at approximately 252*F.
The unirradiated reference nil ductility temperature (RT NDT) for the surveillance region materials were determined in accordance with 10 CFR 50, Appendixes G and H.
For other beltline region materials and other reactor coolant pressure boundary materials, the unirradiated impact properties were estimated using the methods described in BAW-10046A, Rev. 2.
As a result of fast neutron irradiation in the beltline region of the core, there will be an increase in the RT NDT with accumulated nuclear operations.
The adjusted reference temperatures have been calculated by adding the predicted radiation-induced RT NDT and the unirradiated RT NDT for each of the reactor coolant beltline materials.
i
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f The predicted RT NDT was calculated using the respective neutron fluence after ten effective full power years of operation and the procedures defined in i
Regulatory Guide 1.99, Rev. 2.
The analysis of the reactor vessel material contained in the second Three Mile Island Nuclear Station Unit 1 surveillance capsule confirmed that the current techniques used for pr9licting the change in impact properties due to irradiation are conservative.t J I
Analyses of the activation detectors in the TMI-1 surveillance capsules have j
provided estimates of reactor vessel wall fast neutron fluxes for cycles 1 through 4 (Reference 4).
Extrapolation of reactor vessel fluxes, and corresponding fluence accumulations, based on predicted fuel cycle design
{
conditions during 10 effective full power years of operation are described in Reference 4.
Based on the predicted RT NDT after ten effective full power years of l
operation, the pressure-temperature limits of Figure 3.1-1 and 3.1-2 have established in accordance with the requirements of 10 CFR 50, Appendix G. genl The methods and criteria employed to establish the operating pressure and temperature limits are as described in BAW-10046A, Rev. 2.
The protection
[
against nonductile failure is assumed by maintaining the coolant pressure below the upper limits of these pressure temperature limit curves.
3-4 Amendment No. pl,134
The pressure limit lines on Figures 3.1-1 and 3.1-2 have been established considering the following:
1 a.
A 25 psi error in measured pressure.
1
)
b.
A 12*F error in measured temperature.
)
i c.
System pressure is measured in either loop.
1 d.
Maximum differential pressure between the point of system pressure measurement and the limiting reacter vessel region for the allowable-operating pump combinations.
The spray temperature difference restriction, based on a stress analysis of spray line nozzle is imposed to maintain the thermal stresses at the pressurizer spray line nozzle below the design limit. Temperature requirements for the steam generator correspond with the measured NDTT for the shell.
References (1 ) FSAR, Section 4.1.2.4 (2) ASME Boiler and Pressure Code,Section III, N-415 (3) FSAR, Section 4.3.10.5 (4 ) BAW-1901, Analysis of Capsule TMI-1C, GPU Nuclear, Three Mile Island Nuclear Station - Unit 1, Reactor Yessel Materials Surveillance Program (5) BAW-1901, Supplement 1, Analysis of Capsule TMI-10, GPU Nuclear, Three Mile Island Nuclear Station - Unit 1, Reactor Vessel Materials Surveillance Program, Supplement 1 Pressure - Temperature Limits.
3-5 Amendment'No.2/.134
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