ML19262A435

Tech Spec Change Request 52 Supporting Licensee Request to Change App a of License DPR-50 Re Presssurization,Heatup & Cooldown Limitations.Certificate of Svc Encl

ML19262A435
Person / Time
Site:	Three Mile Island
Issue date:	03/09/1977
From:	Arnold R METROPOLITAN EDISON CO.
To:
Shared Package
ML19262A432	List: ML19262A431 ML19262A435 ML19262A437
References
NUDOCS 7910290781
Download: ML19262A435 (4)

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METROPOLITAN EDISON COMPANY JERSEY CENTRAL PCWER & LIGHT COMPANY AND PENNSYLVANIA ELECTRIC CCMPANY THREE MILE ISLAND NUCLEAR STATION UNIT 1 Operating License No. DPR-50 Docket No. 50-289 Technical Suecification Change Reauest No. 52 This Technical Specification Change Request is submitted in support of Licensee's request to change Appendix A to Operating License No. DPR-50 for Three Mile Island Nuclear Station Unit 1. As a part of this request, proposed replacement pages for Appendix A are also included.

METROPOLITAN EDISON COMPANY By /s / p . c . ar-ev Vice President Sworn and subscribed to me this oth day of W-e w , 1977

/s / vav14, c  %""a" Tr .

Notary Public 1480 295

?9102o0 7g

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 h-8 of the FSAR. The maximum unit heatup and c o down rate of 1000F in any one hour satisfies stress limits for cyclic operation. 2},

The 200 psig pressure limit for the secondary side of the steam generator at a tempa DTT.(ryturelessthan1000Fsatisfiesstresslevelsfortemperaturesbelowthe 3 The reactor vessel plate material and welds have been tested to verify conformity to specified requirements and a maximus NDTT value of 30CF has been determined based on Charpy V-notch tests. The maximum NDTT value obtained for the steam generator shell material and welds was 40cF.

The heatup and cooldown rate limits in this specification are not intended to limit instantaneous rates of tempr .sture change, but are intended to limit tempera-ture changes such that there exists no one hour interval, in which a temperature change greater than the limit takes place.

Figures 3.1-1 und 3.1-2 contain the limiting ctor coolant system pressure-temperature relationship for operation at DTT and below to assure that stress levels are low enough to preclude brittle fracture. These stress levels and their bases are defined in Paragraph h.3.3 of the FSAR.

As a result of fast neutron irradiation in the region of the core, there will be an increase in the NDTT with accumulated n telear operation. Th NDTT increase for the h0-year exposure is shown on Figure h-10. bpredicted maximum

/ The actual shift in 'DTT will be determined periodically during plant operation by testing of Ir! eunted vessel material samples located in this reactor vessel.(5) The results or the irradiated sample testing will be evaluated and compared to the design curve (Figure h-ll of the FSAR) being used to predict the increase in transition temperature.

The desi n value for fast neutron (E > 1 MeV) exposure of the reactor vessel is 3.1 x 10 0 n/cm2 see at the reference design power of 2568 MWt and an integrated exposure of 3.0 x 1019 n/cm2 for ho years operation (6) The calculated maximum values are 2.2 x 1010 n/cm2 see and 19 n/cmb integrated exposure for h0 yearsoperationat80percentload.({)2x10 Figure 3.1-1 is based on the design value which is conuderably higher than the calculated value. The NDTT value for Figure 3.1-1 is based on the projected NDTT at the end of the first two effective full power years of operation.

The actual shift in NDTT vill be established periodically during plant operation by testing vessel material samples which are irradiated by securing them periodically near the inside wall of the vessel in the core area to achieve an average effective exposure between 1 and 3 times that of the reactor vessel inner surface. To compensate for the increases in the NDTT caused by irradiation, the limits on the pressure-temperature relationship are periodically changed to stay within the established stress limits during heatup and cooldown.

The NDTT shift and the magnitude of the thermal and pressure stresses are sensitivr:

to integrated reactor power and not to instantaneous power level. Figures 3.1-1 and 3.1-2 are applicable to reactor core thermal ratings up to 2568 MWt.

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

Ob.iective To assure that temperature and pressure changes in the reactor coolant sys+.em do not cause cyclic loads in excess of design for reactor coolant system components.

3.1.2.1 For the first two effective full nover years (EFPY) the reactor l

coolant pressure and the system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figure 3.1-1 and Figure 3 1-2 and are as follows:

Heatuo:

A11ovable combinations of pressure and temperatur shall be to the right of and below the limit line in Figure 3 1-1. Heatup rates shall not exceed those shown on Figure 3.1-1.

Cooldown:

A11ovable cotbinations of pressure and temperature for a specific cooldown shall be to the left of and below the limit line in Figure 3.1-2. Cooldown rates shall not exceed those shown on Figure 3.1-2.

Hydro Tests:

For isothermal system hydrotests during the first two years of operations, the system may be pressurized to the limits set forth in Specification 2.2, when there are fuel assemblies in the vessel and to ASME Code Section III limits when no fuel assemblies are present if the system temperature is_2150 F or greater. The system may be tested to a pressure of 1150 psig provided system temperature is 175cF or greater. Initial system hydrotests prior to criticality may be conducted if the reactor coolant system temperature is 1180F or greater.

3.1.2.2 The secondary side of the steam generator shall not be pressurized I above 200 psig if the temperature of the steam generator shell is below 1000F.

3.1.2.3 The pressurizer heatup and cooldown rates shall not exceed 1000 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 430F.

3.1.2.h Within two effective full power years of operation, Figure 3.1-1 and 3.1-2 shall be updated in accordance with criteria acceptable to the NRC.

Bases All reactor coolant system components are' designed to withstand the effects of cyclic loads due to system temperature and pressure changes.(1) These cyclic 1480 297 3-3

Metropolitan Edison Company (Met-Ed)

Three Mile Island Nuclear Station Unit 1 (TMI-1)

Docket No. 50-289 Operating License DPR-50 Technical Snecification Change Request No. 52 The licensee requests that the attached changed pages 3-3 and 3-h replace the corresponding present pages. This change request is unnecessary upon issuance of Technical Specification Change Request No. 50 (Submitted February 23, 1977).

Reasons for Chance Requesi Amendment 15 (issued May 1b, 1976) changed Technical Specification 3.1.2.4 such that Figures 3.1-1 and 3.1-2 need be updated within two effective full power years (EFPY) rather than two years. This created an inconsistency with Technical Specificaticn 3.1.2.1 which specifies that present heatup and cool-D down limgts apply for the first 17 x 10 thermal megawatt days of operation.

1.7 x 10 thermel megawatt days is somewhat less than two effective full power years. These teei:nical specifications, and the corresponding bases are therefore changed to make them consistent.

Safety Analysis Justifying Change The conservatism of the present figures 31-1 and 3.1-$8have giready been demonstrated up to,a fast neutron exposure of 1.7 x 10 n/cm#. This is the result of 1.7 x 10D thermal megawatt days at the design flux value of 3.1 x 1010 n/cm2 see at the reactor vessel inner vall. Extrapolation of measured cycle 1 flux based on predicted fuel reload and burnup conditions indicates that the maximum average fast neutron (E>lMgV) flux during six full power years of operation vill be 1.68 x 10 10 n/cm see at the inner vall and 9.33 x 109 n/cm2 seeatthe1/hTbocatin. This indicates a maximum average exposure after 2 EFPY of 1.06 x 10l n/cm at the vessel vall and 5 9 x 1017 n/ cme at the 1/hT location. Based upon the above, the present figures 3.1-1 and 3.1-2 are conservative and therefore, this change constitutes no threat to the health and safety of the public.

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