ML20206H658

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Amend 127 to License NPF-1,changing Tech Specs to Revise Reactor Vessel Matl Irradiation Surveillance Schedule & pressure-temp Limits
ML20206H658
Person / Time
Site: Trojan File:Portland General Electric icon.png
Issue date: 04/09/1987
From: Varga S
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20206H643 List:
References
NUDOCS 8704150410
Download: ML20206H658 (9)
v • d • e


Text

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o uc oq UNITED STATES

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"j WASHINGTON, D. C. 20555

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f PORTLAND GENERAL ELECTRIC COMPANY THE CITY OF EUGENE, OREGON I

PACIFIC POWER AND LIGHT COMPANY ,

DOCKET NO. 50-344 TROJAN NUCLEAR PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment flo127 License No. NPF-1

1. The Nuclear Regulatory Commission (the Comission) has found that:

A. The application for amendment by Portland General Electric Company, et al., (the licensee) dated October 31, 1986, complies with the staiidards and req)uirements of the Atomic Energy Act of 1954, a forth in 10 CFR Chapter I; f,

B. The facility will operate in conformity with the application, the provisions of the Act, and the fules and regulations of the Commission; -

C. There is reasonable assurance (i) that the activities authorized '

by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be l'

) - conducted in compliance with the Comission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 1 51 of the Comission's regulations and all applicable requirements -

have been satisfied.

8704150410 DR s7o4o9 ADOCK 05000344 PDR ,

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-1 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A l and B, as revised through Amendment No.127 , are hereby incorporated in the license. The licensee'shall operate the facility in accordance with the Technical Specifica-tions, except where otherwis.e stated in specific license conditions.

3. This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION 0/ b

. ' Direct PWR Project Directorate .3 Division of PWR Licensing-A, NRR

Attachment:

Changes to the Technical Specifications ,

Date of Issuance:

April 9, 1987 e

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52 E:f Co 8 UNITED STATES

[ 0,h NUCLEAR REGULATORY COMMISSION 3

WASHINGTON, D. C. 20555

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ATTACHMENT TO LICENSE AMENDMENT NO.127 TO FACILITY OPERATING LICENSE.NO. NPF-1 DOCKET NO. 50-344 Revise Appendix A as follows:

Remove Paoes Insert Paaes TS 3/4 4-25 ~ TS 3/4 4-25 3/4 4-?6 3/4 4-26 3/4 4-27 3/4 4-27 B 3/4 4-5 B .3/4 4.-5

B 3/4 4-Sa B 3/4 4-9 B 3/4 4-9 o

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CATERIAL PROPERTY CASIS:

- CONTROLLING MAT,ERIAL : R.V. LOWER SHELL COPPER CONTENT  : 0.16 WT%

PHOSPHORUS CONTENT : 0.012 WT%

INITIAL RT , :10*F '

RTwy AFTER 10 EFPY  : %T.111*F

uT. 55'F iil l4iieiiiiiIiliIIiiiiiiiiiiiIiiiiiiiilllii l

. CURVE APPLICABLE FOR HEATUP FiATES UP TO 60*F/HR FOR THE SERVICE PERIOD UP TO 10 EFPY AND CONTAINS MARGINS OF 10*F AND SO PSIG FOR POSSIBLE INSTRUMENT ERRORS.,;

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il l I  ! lll i' 00 OO 100 0 200.0 300.0 d.)0 0 500 0 AVERAGE REACTOR COOLANT SYSTEM TEMPERATURE (*F)

REACTOR COOLANT SYSTEM PRESSURE -~ l FIGURE 3.4-2 TEMPERATURE LIMITS VERSUS 60'F/ HOUR HEATUP RATE -

CRITICALITY LIMIT AND HYDROSTATIC TEST LIMIT .

l TROJAN-UNIT 1 3/4 4-25 Amendment No.127 l

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CATERIAL PROPERTY BASIS:

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CONTROLLING MATERIAL : R.V. LOWER SHELL COPPER CONTENT  : 0.16 WT%

PHOSPHORUS CONTENT : 0.012 WT%

INITIAL rte, :10*F RT , AFTER 10 EFPY  : %T.111*F

%T. 55'F .

I 3000o ... ,

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' CURVES APPLICABLE FOR COOLDOWN RATES UP TO 100*F/HR FOR !I l THE SERVICE PERIOD UP TO 10 EFPY AND CONTAINS MARGINS OF 10*F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS.

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'Fs H R 0 ' ' E. Esw/ l 20 7 ]- g/ l 6 M [ 100 #~ l l l1 I I i l oo 00 100 o 200.0 300.o aco o soo o AVERAGE REACTOR COOLANT SYSTEM TEMPERATURE'(*F) FIGURE 3.4-3 REACTOR COOLANT SYSTEM PRESSURE - I TEMPERATURE LIMITS VERSUS , l COOLDOWN RATES 1 TROJAN-UNIT 1 3/4 4-26 Amendment No. 127 

  . - - _ . _ - . . ..             , . _ _ - - - _ -                                                   - - _              - , - .                - - _ _ ,            l__-         - _ . _ _ _ - . . - _ -

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a TABLE 4.4-5 REACTOR VESSEL MATERIAL IRRADIATION SURVEILLANCE SCHEDULE SPECIMEN REMOVAL INTERVAL

1. U 1.23 EFPY (Actual)
2. X 4.R8 EFPY (Actual)
3. V 8 EFPY
4. Y 15 EFPY
5. W standby
6. Z standby a

l .i i TROJAN-UNIT 1 3/4 4-27 l l ,

a g. 

                                                                                         /

REACTOR COOLANT SYSTEM ' i BASES During heatup, the thernal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner w311 to tensile at the outer wall. These thernal induced compressive Jtresses tend to alleviate the tensile stresses induced by the internal pressure. Therefore, a pressure-temperature curve based on steady state conditions (i.e., no thermal stresses) represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location. The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location. The thernal gradients established during heatup produce tensile stresses at the outer wall of the vessel. These stresses are additive to the pressure induced tensile stresses which are already present. The thernal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of , heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be

defined. Consequently, for the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate of el interest must be analyzed on an individual basis.

The heatup limit curve, Figure 3.4-2, is a composite curve which was prepared by determining the most conservative case, with either the insida or outside wall controlling, for any heatup rate up to 60*F per hour. The cooldown limit curves Figure 3.4-3 are composite curves which were prepared based upon tha teme type Pr21ysi; with Uw exceptien that the controlling location is always the inside wall where the cooldown thermal gradients tend to produce tensile stresses while producing compressive stresses at the outside wall. The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of 10 EFPY and reflect the analysis of Capsule X contained in WCAP-10861. The knee in the heatup and cooldown rate pressure temperature limit curves is based on the May 31, 1983 revision to Section IV.A.2 of 10 CFR 50, Appendix G. This revision requires that when the core is not critical, and pressure exceeds 20 percent of the preservice system hydro-static test pressure (620 psi), that the temperature of the closure flange regions must exceed the reference temperature of their.aterial in those regions by at least 120*F for nornal operation. The' reference temperature of the head flange is limiting in this case (20*F)' and it, [ . taken from Technical Specification Table B 3/4.4-1. ,'> TROJAN-UNIT 1 B 3/4 4-5 9 

                                                                                              -r

i REACTOR COOLANT SYSTEM BASES . The reactor vessel materials have been tested to determine their initial RTNDT; the results of these tests are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron (E>l Mev) irradiation will cause an increase in the RTNDT. Therefore, an adjusted reference temperature, based upon the fluence and copper content of the material in question, can be predicted using Figures B 3/4.4-1.and B 3/4.4-2. The heatup and cooldown limit curves Figures 3.4-2 and 3.4-3 include predicted adjustments for this shif t in RTNDT at the end of 10 EFPY, l as well as adjustments for possible errors in the pressure and temperature sensing instruments. 1

  • ( 4 .

I TROJAN-UNIT 1 B 3/4 4-Sa s 

        .                            REACTOR COOLANT SYSTEM BASES The actual shift in NDTT of the vessel unterial will be established periodically during operation by removing and evaluating, in accordance i

with ASTM E185-82, reactor vessel noterial irradiation surveillance 1 specimens installed near the inside wall of the reactor vessel in the core area. Since the neutron spectra at the irradiation samples and-vessel inside radius are essentially identical, the measured transition i shif t for a sample can be applied with confidence to the adjacent section of the reactor vessel. ' The heatup and cooldown curves must be recalculated when the ART determined from the surveillance capsule is different fromthecal!0IatedARTNDT for the equivalent capsule radiation exposure. The pressure-temperature limit lines shown on Figure 3.4-2 for reactor criticality and for inservice leak testing have been provided to l 4 assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50. ) l 

 ;                                       The number of reactor vessel irradiation surveillance specimens and                                                           l the frequencies for removing and testing these specimens are provided in                                                          ;

Table 4.4-5 to assure compliance with the requirements of Appendix H to I 10 CFR Part 50. The limitations imposed on pressurizer heatup and cooldown and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis perforned in accordance with the ASME Code requit ements. The OPERABILITY of two PORVs or an RCS vent opening of greater than 3.40 square inches ensures that the RCS will b,e. protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are 5 290*F. Either PORV has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with 

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the secondary water temperature of the steam generator $ 50*F above the RCS cold leg temperature or (2) the start of one safety injection pump, two centrifugal charging pumps, and one positive displacement pump and their simultaneous injection into a water solid RCS. The RCS vent opening is based on the flow area of the open PORV. An initial ressurizer cientsurgevo$umetopreclteamvolumeff>200fkngwater-solidduring3 will provide suffi-ude the RCS rom becom  ; the transient where a reactor coolant pump is started in an idle RCS with - ! one or more steam generator secondary side temperatures > 50*F above the , RCS cold leg temperatures. ' l The current heatup and cooldown curves are applicable for the first 

!                                    10 EFPY. The value of 290*F used for the overpressure mitigation system                                                      l    l l                                   1s similarly dependent on the irradiation of the reactor vessel and is                                                            ,

j applicable only for the first 10 EFPY. l l 1 l TROJAN-UNIT 1 8 3/4 4-9 Amendment No. 57, 78, 127 I i i _ ~. . . _ _ _ . - - - - - _ , - _ _ _ - _ _ , , __ , , _ . - , _ _ - _ _ _ -}}

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