ML19260A121

Tech Spec Change Request 50 Supporting Licensee Request to Change App a of License DPR-50 Re Limits for Operation Beyond Two Effective Full Power Yrs,Accounting for Shift in Unirradiated Ref Nil Ductility Temp of Reactor Vessel

ML19260A121
Person / Time
Site:	Three Mile Island
Issue date:	02/23/1977
From:	Arnold R METROPOLITAN EDISON CO.
To:
Shared Package
ML19260A119	List: ML19260A118 ML19260A121 ML19260A122
References
NUDOCS 7910290738
Download: ML19260A121 (10)

Text

e 6 METROPOLITA I EDISOIT COMPA iY JERSEY CE:ITRAL PO'a?R & LIGHT COMPA!IY AND PEITIISYLVANIA ELECTRIC COMPA?IY THREE MILE ISLA lD MUCLEAR STATION UIIIT 1 Operating License No. DPR-50 Docket Iio. 50-289 Technical Srecification Chanste Reauest !!o.50 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 I.uclear Station Unit 1. As a part of this request, proposed replacement pages for Appendix A are also included.

LJETROPOLITAN EDISON COMPANY r n By 8v

'Vice FT6sident M

Sworn and subscribed to me this day of / 1977

/

Totary P'f/lic m .

, , . . - s e3 b 1480 030

@ 910 290 73g

e s U:!ITED STATES OF AMERICA ITUCLEAR REGULATORY COF"!ISSIO:I IN THE MATTER OF DOCKET :IO. 50-289 LICENSE N0. DPR-50 METROPOLITAN EDISON COMPANY Sis is to certify that a copy of Technical Specification Change Request :lo.

50 to Appendix A of the Operating License for Three Mile Island Nuclear Station Unit 1, has, on the date given below, been filed with the U. S. Nuclear Regulatory Commission and been served on the chief executives of Londonderry Township, Dauphin County, Pennsylvania and Dauphin County, Pennsylvania by deposit in the United States mail, addressed as follows:

Mr. '4eldon B. Arehart Mr. Harry 3. Reese, Jr.

Board of Supervisors of Board of County Commissioners Londonderry Township of Dauphin County P. D. #1, Geyers Church Road Dauphin County Court House Middletown, Pennsylvania 17057 Harrisburg, Pennsylvania 17120 METROPOLITAN EDISON COMPANY A

! ! D By 'I Vibe President Dated: February 23 , 1977 1480 031

. s Metropolitan Edison Company (Met-Ed)

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

Docket No. 50-289 Operating License DPR-50 Technical Specification Change Request No. 50 The licensee requests that the attached changed pages 3-3, 3-h, 3-5, 3-6 and Figures 3.1-1 and 3.1-2 replace the corresponding present pages.

Reasons for Change Recuest i This change request provides new technical specification limits for operation

{ beyond 2 EFPY to account for any shift in the unirradiated reference nil l ductility temperature (RTNDT) of the reactor vessel. This change request is

~j required to comply with the requirements of TMI-1 technical specification 3.1.2.h concerning pressurization, heatup and cooldown li=itations.

I t Safety Analysis Justifyine Chance 1 The new technical specifications are based upon the analysis of capsule TMI-IE, I

as described in BAW-lh39 January 1977 (enclosed). The shift in the nil ductility temperature of the controlling beltline region vere based upon an as yet unpublished revision to Regulatory Guide 1.99 This technique is basically the same as

! that now contained in Regulatory Guide 1.99 except that the upper limit curve I has been raised. The techniques used for predicting change in nil ductility

. temperature due to irradiation have been shown to be conservative since the

, data for the surveillance veld metal was close to the predicted value.

!* Due to the demonstrated conservatism of these techniques, this change is determined to not cause a threat to the health and safety of the public.

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e s 3.1.2 PRESSt;RIZATION HEATUP AND C00LDOWN LIMITATIONS Applicability Applies to pressurization, heatup and cooldown of the reactor coolant system.

Objective To assure that temperature and pre:sure changes in the reactor coolant system do not cause cyclic loads in excess of design for reactor coolant system co=ponents.

Specification 3.1.2.1 For operations until six effective full power years, the reactor coolant pressure and the system heatup and cooldown rates (with the exce-tion of the pressurizer) shall be limited in accordance with Figure 3 1-1 and Figure 3 1-2 and are as follows:

Heatup/Cooldown Allowable co=binations of pressure and temperature shall be to the right of and below the lir .t line in Figure 3 1-1. Heatup an?

cooldown rates shall not exceed those shown on Figure 3.1-1.

, Inservice Leak and Hydrostatic Testing

{ Allowable combinations of pressure and tempersture shall be to the right of and below the limit line in Figure 31-2. 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 cocidown rates shall not exceed 100 F in any one hour. The spray shall not te used if the temperature difference between the pressurizer and the spray fluid is creater than h30 F.

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e s 3.1.2.h Within six 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 (1}

effects of cyclic loads due to system temperature and pressure changes.

g These cyclic loads are introduced by unit load transients, reactor trips, and l '

unit heatup Lnd cooldown operations. The number of thermal and loading cycles used for design purposes are shown in Table h-8 of the FSAR. The maxi =um unit heatup and cooldown rate of 100 F in any one hour satisfies stress limits for cyclic operation. (2) The 200 psis Ivessure limit for the secondary side of

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- the steam generator at a temperature less than 100 F satisfies stress levels for temperatures belov the DTT.(3) l The heatup and cooldown rate limits in this specification are not intended a

t to limit instantaneous rates of temperature change, but are intended to limit j

l temperature changes such that thers exists no one hour interval, in which a temperature change greater than the limit takes place.

The unirradiated reference nil ductility temperature (RT NDT) for the

surveillance region =aterials were determined in accordance with 10CFR50, I

g Appendixes G and E. For other beltline region materials and other reactor coolant pressure boundary materials, the unirradiated impact properties were

,~ estimated using the methods described 4- '"3'onh6.

As a result of fast neutron irr?.diation in the beltline region of the i

core, there vill be an increase in the RTNDT vith accumulated nuclear operations. The adjusted reference temperatures have been calculated by sdding the predicted radiation-induced RTNDT and the unirradiated RTNDT for each of the reactor coolant beltline materials.

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7ne predicted RT was calculated using the respective neutron fluence after ET six effective full power years of operation and copper and phosph.,rus content.

The analysis of the reactor vessel material contained in the first surreillance capsule removed from Three Mil, Island Nuclear Station Unit 1 confirmed that the current techniques used f: predict'.ng the change in impact properties due to irradiation are conservative.

Analysis of the activation detectors contained in the first surveillance capsule indicates that the average fast flux during cycle 1 was 1.h5 x 10 10

"/cm2 -sec :sximum at the pressure vessel vall. Extrapolation of the cycle 1 flux based on predi.:ted fuel reload and burnup conditions indicates that the m uimum average fast neutron (EDL Mev) flux during six full power years of operation vill be 1.68 x 1010 n/cm2 - see at the reactor vessel vall and 9.33 x 10 9U /en see at the 1/4T location. The fast neutron exposure during six effective full power years of operation, therefore, is 1.8x1018 n/cm at the U

1/h T Iveation and h.hx10 /cm at tne 3/h T location.

Based on th: predicted RTNDT after six effective full power years of opa ation, the pressure-temperature limits of Figure 3 1-1 and 3.1-2 hav' heen established in acccrdance with the requirements of 10CFR50, Appendix G. Th-f methods and criteria employed to establish the operating pressure and te=perat".re limits are as described in BAW-100h6. The protection against nonductile failure is assumed by maintaining the coolant pressure belov the upper limits of these pressure temperature limit curves.

The pressure limit line en Figure 3.1-1 and 3 1-2 have been established considering the following:

a. A 25 psi error in neasured pressure

b. a 12 F error in meacured temperature 3_3 1480 035 t

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c. System pressure is measured in either loop.

d. Maximum differential pressure between the point of system pressure measurement and reactor vessel inlet for all operating pump combinations.

The spray temperature difference restriction, based on a stress analycis of spray line no::le is imposed to maintain the thermal stresses at the pressurizer ,

sprsy line no::le below the design limit. Temperature requirements for the steam generator correspond with the measured NDTT for the shell.

. REFERENCES (1) FSAR, Section h.l.2.h (2) ASME Boiler and Pressure Code,Section III, N-h15 (3) FSAR, Section h.3.10 5 (k) BAW-lh39, Analysis of Capsule StI E From Metropolitan Edison Corpany, Three Mile Island Nuclear Station - Unit #1, Reactor Vessel Materials Surveillance Program.

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l Assumed RT , F i

Beltline 1/4 T lh5 i Beltline 3/h T 82

! Closure Head Region 60 Outlet Uczzle 60 2 30C'. Point Temu Press.

E A 75 105 2200 - B 125 h10

, C 175 h85 D 275 h85 2000 E 320 2275 RC Pump Combinations A11cvable 1800 - Above 195 F All i

' Belov 195o F 1-A,1-B,0-A,1-B, 1-A,0-3 e 1600 -

2 1) When Decay Heat Removal System (DH) ,

ai lh00 _ is operating without any BC pumps

= operating, indicated DH return tesp. .

O to the Reactor Vessel shall be used.

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M 1200 i E 2) Heat-up and Cooldown rates shall not g exceed 100 F in any one hour

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o i Indicated Reacter Coolant Temperature, F l I i

! Reactor Ccolant Syste.m Heat-up/Cooldown Linitaticns (Applicable to 6 EFFY)

Figure 3.1-1

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Assumed RTg,F Beltline Region 1/h T 1h5 Beltline Region 3/h T 82 Closure Head Region 60 Outlet Nozzle 60 Point Tern Press. E 2h00-- -. A 75 175 3 115 h05 C 160 h85 2200-- D 260 h85 E 287 2500 2000 -

RC Pump Combinations Allevable 1 1800-- Above 195 F All l* Below 195 F 1-A,1-B,0-A,1-3; 1-A,0-B g 1600 -

[ E A 1) When Decay Heat System (DH) la g 1h00-- is operating without any RC pumps g operating, indicated DH Return Te:p g to the Reactor Vessel shall be use y 1200--

A 2) Heat-up and Cooldown Rates shall n t exceed 50 0 F in any one hour.

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50 100 15b 2cb 25b 300 IIDICATED REACTCR COOLA:C TEMPERATURE, F Reactor Coolant Systen Inservice Leak & Rydrsstatic Test Limitations

{ Figure 3.1-2

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-=q EF3ATA

1. Figures 5-1, 5-2, 5-3, and 5-5: The data sur=r_ry values of Cy-USE should be in units of ft-lbs and not degrees F.

2. Table 7-2: The last set of data indicat,e " Decrease" rather than

" Increase" in Charpy USE.

3 Page 4-1 ccaelusion 2 should read "The fast fluence of 1.8 x 10 18 n/cm2 (E>1MeV) vill increase the RT NDT of the pressure vessel core region shell materials to a maxi =um of ih50 f.

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