ML20236D997

Proposed Tech Specs Providing Updated RCS Heatup & Cooldown Limits for Operation to 10 EFPY to Account for Irradiation Effects on Reactor Pressure Vessel NDTT as Determined by Analysis

ML20236D997
Person / Time
Site:	Three Mile Island
Issue date:	07/24/1987
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20236D974	List: ML20236D972 ML20236D991 ML20236D997 ML20236E015
References
NUDOCS 8707310175
Download: ML20236D997 (8)

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ATTACHMENT Technical Specification Changes i

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LIST OF FIGURES

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. 1 TMI-l Core Protection Safety Limit 2.. . TMI-l Core Protection Safety Limits 2.1 -3 TMI-l Core Protection Safety Bases

2. 3-1 TMI-l Protection System Maximum Allowable Set Points 2.3-2 Protection System Maximum Allowable Set Points for Reactor Power Imbalance, TMI-l j

.l 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) J i  ;

3.1 -3 Limiting Pressure vs. Temperature Curve for 100 STD cc/ Liter H0 2

3.5-2A Rod Position Limits for 4 Pump Operation from 0 to 30+10/-0 EFPD, TMI-l i

3.5-28 Rod Position Limits for 4 Pump Operation from 30+10/-0 to 25010 EFPD, TMI-1 -

l 3.5-2C Rod Position Limits for 4 Pump Operation af ter 25010 EFPD, TMI-l  !

3.5-2D Rod Position Limits for 3 Pump 0peration from 0 to 30+10/-0 l EFPD, TMI-1 3.5-2E Rod Position Limits for 3 Pump Operation from 30+10/-0 to 25010 EFPD, TMI-l

3. 5-2 F Rod Position Limits for 3 Pump Operation after 25010 EFPD, TMI-l 3.5-2G Rod Position Limits for 2 Pump Operation from 0 to 30+10/-0 EFPD, TMI-l 3.5-2H Rod Position Limits for 2 Pump Operation from 30+10/-0 to 25010 EFPD, TMI-l 3.5-2I Rod Position Limits for 2 Pump Operation after 25010 EFPD, TMI-l l- 3.5-2J Power Imbalance Envelope for Operation from 0 to 30+10/-0 EFPD, l TMI-l vii Amendment Nos. 11, 17, 29, 39, 45, 50, 59, 72, 106, 109, 120, 126

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3.1.2 PRESSURIZATION HEATUP AND C00LDOWN LIMITATIONS Applicability Applies to pressurization, heatup and cooldown of the reactor coolant system.

Objective To assure that tempe'rature and pressure changes in the reactor coolant system  !

do not cause cyclic loads in excess of design for reactor coolant system  !

components.

Specification 3.1. 2.1 Fcr 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 j Figure 3.1-1 and Figure 3.1-2 and are as follows:

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Heatup/Cooldewn 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.

Inservice Leak and Hydrostatic Testing I

Allowable combinations of pressure and temperature shall be to the '

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.

3.1.2.2 The secondary side of the steam generator shall not be pressurized above 200 psig if the temperature of the steam generator shell is 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 than 430*F. ]

3.1. 2. 4 Prior to exceeding ten effective full power years of operation, j Figures 3.1-1 and 3.1-2 shall be updated for the next service period in accordance with 10 CFR 50, . Appendix G,Section V.B. The highest predicted adjusted reference temperature of all the beltline 1 materials shall be used to determine 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.

3-3 Amendment No. 29 l

Bases All reactor cyclic loadscoolant system temperature due to system componentsand are pressure designedchanges.ll to withstgnjThese the effects cyclic of 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 maximym and cooldown rates satisfy stress limits for cyclic operation.l2) unit Theheatup l

200 psig pressure limit for the secondary side of the steam generator at a temper NDTT. JJ (gture less than 100*F satisfies stress levels for temperatures below the l'

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 i accommodate 15*F step changes at any time with the appropriate soak (hold) l 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.

l 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.

The predicted RT F was calculated using the respective neutron fluence af ter ten effective ful p)wer years of operation and the procedures defined in Regulatory Guide 9, Rev. 2. The analysis of the reactor vessel material contained in the second Three Mile Island Nuclear Station Unit i surveillance capsule in impactconfirmed properties that duethetocurrent techniques irradiation used for pr9gicting are conservative.t > the change I

Analyses of the activation detectors in the TMI-l surveillance capsules have 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 ef fective 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, establishedthe pressure-temperature in accordance limits of Figure with the requirements of 103.1-1 CFR and 3.1-2 have 50, Appendix G.Igqen 1

The methods and criteria employed to esf'3blish the operating pressure and l I temperature limits are as described in thW-10046A, Rev. 2. The protection 4 against nonductile failure is assumed bp maintaining the coolant pressure l i I below the upper limits of these pressure' temperature limit curves, j e .

3-4 l Amendment No. 29 l l l _

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The pressure limit lines on Figures 3.1-1 and 3.1-2 have been established j considering the following:

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a. A 25 psi error in measured pressure.

b. A 12*F error in measured temperature.

c. System pressure is measured in either loop. f 1

d. Maximum differential pressure between the point of system pressure  !

measurement and the limiting reactor vessel region for the allowable operating pump combinations.

The spray temperature difference restriction, based on a stress analysis of j spray line nozzle is imposed to maintain the thermal stresses at the  ;

pressurizer spray line nozzle below the design limit. Temperature I 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 l

(3) FSAR, Section 4.3.10.5  !

(4 ) BAW-1901, Analysis of Capsule TMI-lC, GPU Nuclear, Three Mile Island Nuclear Station - Unit 1, Reactor Vessel 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 i Surveillance Program, Supplement 1 Pressure - Temperature Limits. I i

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BAW-1901 Supplement 1 June 1987

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ANALYSIS OF CAPSULE TMll-C GPU NUCLEAR THREE MILE ISLAND NUCLEAR STATION-UNIT 1

-- Reactor Vessel Material Surveillance Program --

Supplement 1 PRESSURE-TEMPERATURE LIMITS 1

Babcock &Wilcox a McDermott company I