ML19322A217

Tech Spec Change Request 88 in Support of Request to Change once-through Steam Generator Pressure Temp Limit Per App a of License DPR-50

ML19322A217
Person / Time
Site:	Three Mile Island
Issue date:	01/04/1979
From:	METROPOLITAN EDISON CO.
To:
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ML19322A216	List: ML19322A215 ML19322A217
References
NUDOCS 7901150072
Download: ML19322A217 (9)

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METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LICitT COMPANY AND PENNSYLVANIA ELECTRIC COMPANY TilREE MILE ~ ISLAND NUCLEAR STATION UNIT I Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request No. 88 This Technical Specification Change Request is submitted in support of Licensee's request to change Appendix A to Operating License tio. 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 07 ,

Sworn and subscribed to me this _ [/

day of _ h /97fp D

Notary Pt lic

[1 UCRCC J. TnorFER Nof:ry (m, nnen y,,g,, ca, 94 Comm!n :n Fypwn e,,,, y ym 790115007>

2

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION IN THE MATTER OF DOCKET NO. 50-289 LICENSE NO. DPR-50 METROPOLITAN EDISON COMPANY

'This is to certify' that a copy of Technical Specification Change Request No. 88 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. Weldon B. Arehart Mr. Harry B. Reese, Jr.

- Board of Supervisors of Board of County Commissioners Londonderry Township of Dauphin County R. J. fl, Ceyers Church Road Dauphin County Court House Middletown, Pennsylvania 17057 Harrisburg, Pennsylvania 17120 METROPOLITAN EDISON COMPANY I

By

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Dated: 90

= - - -

, G. ' ,' '.

LThree Mile Island Nucicar Station, Unit I (TMI-1) e Operating License No. DPR-50

_ Docket No. 50-289 Technical Specification Change Request No. 88 The' licensee requests that the attached pages vi, 3-3, 3-4, and 3-5 replace like pages in the existing Technical Specifications, that the attached Figure 3.1-1 replace the current Technical Specification Figure 3.1-1, and that the current Figure 3.1-2 be deleted.

Reason for Change Request Metropolitan Edison Company (Met-Ed) was informed by the NSSS vendor, Babcock and Wilcox (B&W) that weld wire atypical of the submerged arc weld filler wire used .

by B6W in_the construction of reactor pressure vessels may have been used in the

. construction of the TMI-1 Reactor Vessel. It is uncertain whether the atypical material has actually been used in the construction of the TMI-1 Reactor Vessel.

. Based on the assumption that the atypical material is present in the welds, B&W proposed the attached changes to Tech. Specs, which govern heatup and cooldown '

rates and pressure-temperature limits. . The commitment for this change request was made in LER 78-24/01T, dated August 21, 1978.

While the reactor coolant system heatup and cooldown limits were revised, it was determined that it was necessary to change the once-through steam generator (OTSC) pressure-temperature limit.

p Safety Evaluation Justifying Change There are two weld ' locations in the THI-1 Reactor Vessel that may contain the atypical material: upper shell to nozzle helt and lower.shell to the Dutchman.

Both are circumferential welds. The latter weld location is subject to low neutron fluence (nvt <1.0 x 1017 n/cm2 , E > 1 MeV) and the applied streuses are

- no greater than those applied in the beltline region.

A f racture mechanics evaluation ha- demonstrated that the structure integrity of the vessel' is uncompromised by th< aossible presence of the atypical material.

However, welds of the atypical material exhihit a higher unirradiated reference temperature, RTNDT. The f act that- RTNDT may have been underestimated would indicate that the ' current P-T limits would not have adequate margins of safety against non-ductile failure.

However, the applied pressure stress normal to the upper shell to nozzle belt weld is' approximately half that employed in calculating P-T limits, i.e., P-T limits were calculated assuming a defect oriented normal to the maximum stress direction.

-Thus, for circumferential welds,- the P-T limit curves have an added safety factor of 2. 'fith this safety factor and the safety factor of 2 applied on the stress intensi::y factor due to primary stresses (as required by' ASME Section III Appendix C),

the overall safety f actor is approximately four.-

s wt.

Aino, weldments of the atypical material have dif ferent silicon and nickel contents than the weldmenta using the atypical Mn 'fo-Ni weld wire. The copper content of the atypical weldments in about the name an the copper content of curveillance weldment s of Oconee Station, ifnits 2 and 3. Those nurveillance weldments exhibit a lower HTNDT than predicted by Reg. Guide 1.99 (Rev. 1).

The neutron induced embrittlement of the atypical weld meterial in expected to be equal to or less than the surveillance specimen weldmentn; equal because of the name copper content, possibly lower becaune of the lower nickel content. The effect of high at11 con is currently being evaluated by !!&W.

Furthermore, the current, and attached, P-T limits are applicable to only 5 EFPY.

Early in the life of the reactor, most of the P-T limita are controlled by the clonure head region during heatup and by the steam generator during con 19wn (see BAW 10046A, Rev. 1).

The revined operational heatup and cooldown limitn were calculated by assuming the atypical material van present in circumferential welda, and superimponing those points on exinting heatup and couldown curven. The operational limits include a safety factor of 2.0.

The operational 11mitu curve, with its safety f actor of 2.0, han been extrapolated to 2500 puin and will be used for hentup and cooldown for Innervice henk and liydrontatic tents as well.

The new limit for the OTSG was alno calculated uning proceduren dencribed in BAW-10046A , Rev. 1.

The temperature limit in baned on the highent RTNDT of the nteam generator (40"F) plun denign margin (60"F) plun a margin for inntrument error (12"F) for a total of 112"F. The preunure limit in based on 20% of the preoperational hydrostatic ten t prennure (1312 pntg) minun a margin for instrument error (24 pulg) for a total of 238 pnig.

The revised reactor coolant system heatup and cooldown limitu do not nubstantially differ from the current limit s, al though they are nomewhat more connervat ive.

Daned on the assumption that the atypical material may be present in nome weldn, the revised limits reflect the procedural requirements of BAW-10046A, Rev. 1. The revised reactor coolant system heatup and couldown 11mitu and OTSC prenuure-temperature limits will permit continued nafe operation of TMI-l in compliance with current regulations.

hicenne Amendmqnt Fee (10 CFl(17_0.22)

Because thin change involven a ningle innue, the licennee han determined that thin in a Clasu Ill Amendment and therefore the proper remittance in $4,000.

LIST OF FICURES Figure Title 2.1-1 Core Protection Safety Limit 2.1-2 Core Protection Safety Limits 2.1-3 Core Protection Safety Basis 2.3-1 Protection System Maximum Allowabic Set Points 2.3-2 Protection Syutem Maximum Allowable Set Points 3.1-1 Reactor Coolant System lleatup/Cooldown Limitations 3.1-2 ' Deleted 3.1-3 Limiting Pressure vs. Temperature Curve for 100 STD cc/ Liter Il20.

3.5-1 Incore Instrumentation Specification Axial Imbalance Indication 3.5-2 Incore Instrumentation Specification Radial Flux '

Tilt Indication 3.5-2A- Rod Position Limits for 4 Pump Operation Applicable from 0 to 125+ EFPD; TMI-1, Cycle 4 3.5-2B Rod Position Limits for 4 Pump Operation from 125 + 5 EFPD to 265 i 15 EFPD; TMI-1, Cycle 4 3.5-2C Rod Position Limits for 2 and 3 Pump Operation Applicable During the Period from 0 to 125 i 5 EFPD; TMT-1, Cycle 4 3.5-2D Rod Position Limits for 2 and 3 pump Operation from 125 1 5 EFPD to 265 i 15 EFPD; TMI-1, Cycle 4

.3.5-2E Power Imbalance Envelop for Operation from 0 to 125 i 5 EFFD, TMI-1, Cycle 4 vi L

3.1.2 PRESSURIZATION HEATUP AND COOLD0h'N LIMITATIONS Applicability Applies to pressurization, heatup and cooldown of the reactor coolant system.

Objective To assure that temperature end pressure changes in the reactor coolant system coolant system do not cause cyclic loads in excess of design for reactor components. * .

Specification ,

3.1.2.; For operations until five ef fective fuil power years, the reactor coolant pressure and the system heatup and cooldown rates (with the exception of tha pressurizer) shall be limited in accordance with '

Figure 3.1-1 for normal operations and for inservice leak and hydrostatic testing.

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 in Figure 3.1-1.

3.1.2.2 The secondary side of the steam generator shall not be pre surized is above 238 psig if the temperature of the steam generator shell below 112 F.

exceed 1000F 3.1.2.3 The pressurizer heatup and cooldown rates shall not in any one hour. The spray shall not be used if the temperature ~

dif f erence between the pressurizer and the spray fluid is greater than 4300F.

3.1.2.4 Prior to exceeding five effective full power years of operation, Figure 3.1-1 shall be updated for the next service period [

The in accordance with 10 CFR 50, Appendix G, Section V.B.

highest predicted adjusted reference temperature of all the belt-line 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.

3-3 1

. 3.1.2.5 Tha upacecd proposed technical sprcifications referred to in 3.1.2.4 ch211 be cubmLtted for NRC review at least 90 days prior to the end of the cervice psriod. Appropriate additional NRC review time shall be allowed for proposed technical specification submitted in accordance with 10 CFR 50, Appendix G, Section V.C.

Bases All reactor coolant system components are designed to withstand the ef fects of cyclic loads due to system temperature and pressure changes. (1) These cyclic 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 maximum temperature dependent heatup and cooldown rates are given below.

Max Heatup Rates Max Cooldown Ratas Temp 0F Max Rate Temp 0F Max Rate _

700 - 5500 500 F/hr Operating Range - 280 100'F/hr 280 - 150 500F/hr 150 - 2505/hr 1

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

The unirradiated reference 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. 1.

As a result of f ast neutron irradiation in the beltline region of the core, The 1 there will be an increase in the RT NDT with accumulated neutron fluence. I adjusted reference temperatures have becc. calculated by adding the predicted .

radiation-induced ART NDT and the unirradiated RT NOT for each of the reactor coolant beltline materials.

The predicted RT NDT was calculated using the respective neutron fluence af ter five ef f ective f ull power years of operation. The upper limit of Figure 1 of Regulatory Cuide 1.99, Revision 1, was utilized for determining ART NDT. Fluence values are based on conservative analysis and extrapolation of actual measured neutron doses measured after the first fuel cycle to 5 ef fective full power years.

Based on the predicted RT NDT after five effective full power years of operation, the pressure-temperature limits of Figure 3.1-1 have been established in accordance with the requirements of 10 CFR 50, Appendix G. The methods and criteria employed to establish the operating pressure and temperature limits are as described in BAU-10046A, Rev. 1. The protection against nonductile failure is assured by maintaining the coolant pressure below the upper limits of these pressure-temperature limit curves.

3-4

Th2 prassure limit lins on Figure 3.1-1 has been establishe@ considuring

'ths following:

a. System pressure is measured in the loop with the pressurizar.

b. Maximum differential pressure between the point of system pressure measurement and reactor vessel controlling regions for all operating pump combinations.

c. No adjustment for instrument error, which is included in the procedure.

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.

The limitation on steam generator pressure and temperature provide protection against nonductile failure of the secondary side of theof the steam gene steam generators. The highest estimated R + 60 F (100 F), the is 40 F. At metal temperatures lower than eved by limiting the protection against non-ductile failure isThe acsecondary coolant pressu highest estimated '

system hydrostatic test pressure of 1312 psig.of the steam RT The generato is gassumed to be equal to or greater than 0 the coolant temperature.

limits of 112 F0 and 238 psig include 12 F and 24 psig for possible in-strument error.

REFERENCES (1) FSAR, Section 4.1.2.4 0 (2) ASME Boiler and Pressure Code,Section III, N-415 (3) FSAR, Section 4.3.10.5 (4) BAW-1439, Analysis of Capsule TMI-IE from Metropolitan Edison Company, Three Mile Island Nuclear Station - Unit 1, Reactor Vessel Materials Surveillance Program.

,89 4

3-5

'r*.

l.. .

When Decay Heat Removal System is Assumed RTNDTe I Beltline 1/4T 238 -

operating without any RC pumps oper-Beltline 3/4T 177 -

ating, indicated Dit return temp. to Clocure Head Region 60 the Reactor Vessel shall be used. 60 Outlet Nozzle RC Pump Combinations Allowable:

Above 195 F All

!!ax Heatup Rate j 50 F/hr Below 195 F 1-A, 1-B; 0-A, 1-B; 1-A, 0-B Cooldown Rate:

Temp. C.D. Max Rate Operating Range-280 1000F/hr 280 -150 500 F/hr 150 - 250F/hr

. Point Press Temp.

A '105 75 410 125 G 2500 - B 550 210 A maximum step temperature C

D 550 275 change of 75 F is allowable 1500 300 F when removing all RC pumps

$! E

• F 2250 368 from operation with the G 2500 391 DHR system operating. The gj 2000 -

step temperature change is defined as th+. RC temp Dl gj (prior to stopping all RC

• pumps) ninus the DHR return temp (after stopping all RC h

• pumps). The above F tabulated ramp decrease is

% 1500 -

y allowable both before and after the step change.

S O

d a

Ej 1000 - '

0 n.

C D' 8 500 -

u B o

t; -

ef M A 100 200 300 400 500 REACTOR COOLANT TEMPERATURE, c T,F Reactor Coolant System Heatup/Cooldown Limitations (Applicabic to 5 EFPY)

Figure 3.1-1

t' 4 4 g

e h

FIGURE 3.1-2 DELETED

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gs .e e G