ML19210A407

Tech Specs Change Request 40 to Amend DPR-50,App a Re Reactor Vessel Surveillance Program

ML19210A407
Person / Time
Site:	Three Mile Island
Issue date:	10/29/1976
From:	Arnold R METROPOLITAN EDISON CO.
To:
Shared Package
ML19210A399	List: ML19210A398 ML19210A407
References
NUDOCS 7910290630
Download: ML19210A407 (11)

Text

,

METROPOLITAN EDISOI COMPANY JERSEY CEIITRAL PO'4ER & LIGHT COMPANY AND PE:iNSYLVANIA ELECTRIC COMPANY THREE MILE ISLAND !!UCLEAR STATION U'iIT 1 Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Recuest No. 15 0 This Technical Specification Change Request is submitted in support of Licensee's request to change Appendix A to Operating License I!o. 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 EDISO:; COMPA!Y e

By '

Vice Pres'ident-Generation Sworn and subscribed to me this 2.4 day of , 1976.

!!otary Pu@ic

- m y ,,7,,3

.e *y, 73 l't '33 1469 N 791029 0

Metropolitan Edison Company (Met-Ed)

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

Operating License No. DPR-50 Docket No. 50-289 Technical Srecification Change Recuest No.h0 The Licensee requests that the attached changed pages (v, 3-h, 3-5, h-11, 4-13, and table 4.1-2) replace pages v, 3-4, 3-5, h-11, and h-13 of the existing technical epecifications.

Reason For Proposed Change As a result of damage reported in our letters of March 18, 1976, and March 23, 1976, the TMI-l reactor vessel surveillance holder tubes were removed. TMI-1 Cycle 2 operations began in accordance with Amendment 15 and the exemption to 10CFR50 Appendix H issued on May 1h, 1976, in response to our requests of March 23, 1976.

Amend =ent 15 requires that the TMI-l surveillance specimens be reinstalled prior to TMI-1 Cycle 3 operation. In order to achieve this goal, new surveillance holder tubes =ust be installed. Sabcock and Wilcox (3&W) atte=pted to design a new holder tube that could be installed without re= oval of the core barrel.

(The design in concept was similar to the original holder tube design except that no push rod assembly was used. ) The B&W design did not prove to be 2dequate since prototype flow testing indicated that unacceptable wear would be experienced.

Since an acceptable design,which could be installed without removing the core barrel, could not be found, other designs were explored. Installation of all these other designs involves core barrel removal and underwater machine work or velding on an irradiated thermal shield as well as nu=erous other complex operations. We feel that quality assurance and inspection of these operations vould be very difficult, that the operations have a high potential for creating significant unforeseen proble=s, and that the expected man-REM exposure vould be excessive. In addition, it is extremely probable that the Cycle 3 refueling outage vould be extended by several =onths. Therefore, we have concluded that a TMI-1 and 2 site integrated reactor vessel surve.. lance progra= permitted by 10CFR5u \ppendix H paragraph II.C.h is the only reasonable alternative.

As explained above, this change request is necessary to avoid an extended and costly outage and to avoid excessive personnel radiation exposure, while assuring that changes in the reactor vessel toughness properties are appropriately

=enitored.

Safety Evaluations The purpose of the reactor vessel surveillance preg:a= is to =onitor changes in the fracture toughness properties to permit the determination of the conditions undar which the vessel can be operated with adequate =argins of safety throughout service life. To acco=plish this objective, ;.ppendix H to 10CFR50:

}kb9

=e he

. u .' *

- 2 -

1. requires surveillance speci=en capsules to be irradiated at fast neutron flux levels one to three ti=es the level existing at the vessel inner surface.

2. specifies the withdrawal require =ents for the first four capsules.

3. requires irradiation of a standby capsule.

h. and requires provisions be =ade for additional surveillance tests to monitor the effects of annealing and subsequent irradiation, if re qv. ired.

In addition, fro = an operating standpoint it is desirable to collect data prior to the vessel having reached a given exposure in order to avoid unnecessarily restricting plant operations and to ensure i=portant changes in fracture toughness properties are foreseen.

A site integrated surveillance progra= to irradiate both TMI-l and TMI-2 capsules in the TMI-2 reactor vessel achieves the above objectives. The sequence of re= oval / insertion of the various capsules and their attendant exposure is shown in table II. Several conservative assumptions shown in table I were made to demonstrate that, even given a delay in TMI-2 operations and poor TMI-2 pez'ormance cc=bined with excellent TMI-l perfor=ance, the above surveillance progra=

i objectives can be met, i

! As indicated by Tables I and II, the average effective capsule exposure for

' both TMI-l and TMI-2 is one to three times that of the reactor vessel at the kt location. In addition, all of the required test data vill be obtained and available prior to the vessels reaching approxi=ately one-half of their service life. An e.* fort has also been =ade to =inimize the number of capsule insertions /re= ovals while at the sa=e time irradiating two capsules for each unit to lead the vessel prior to reaching h of service life (capsule exposures can be equalized i at any time). This assures that both the standby capsule provision and the provision for =onitoring the effects of annealing and subsequent irradiation are met, i

It should be noted that due to the si=ilarity of TMI-l to TMI-2, no significant l differences in the fast neutron flux spectru=, operatibg te=peratures or other environ = ental conditions vill exist for the TMI-l speci= ens in TMI-2. As a result, a site integrated surveillance program for the irradiation of both TMI-1 and TMI-2 capsules in TMI-2 will have no adverse effect on the quality of the ,

data obtained.

In su==ary, the proposed TMI site integrated Reactor Vessel surveillance progra=

provides assurance that both the TMI-1 and TMI-2 reactor vessels will be =enitored such that vessel operating limits can be established and de=cnstrated to be conservative throughout service life. Therefore, this change does not represent undue risk to the health and safety of the public.

1469 132 5

6

.A 8 TABLE I SCHEDULING DATA FOR THE TMI SITE INTEGRATED RV SURVEILLANCE PH0 GRAM TMI-1 TMI-2 Notes BV Service Life 32 EFPY 32 EFPY (1)

Assumed Capacity Factor 0.8 0.6 Ther=al Power (MWt) 2535 2772 Assumed 1st Cycle Length (EFPY) 1.3 (Actual) 1.20 Assumed 2nd Cycle Length (EFPY) 0.8 0.8 Assu=ed Subsequent Cycle Length (EFPY) 7 7 Surveillance Speci=en Lead Factor

@ ht in TMI-2 30 2.8 (8)(9)

Specimen Equivalent ".xposure to Date (EFPY) 0 0 (10)

Nu=ber of Capsules to be Irradiated 5 6 (2)

Req. Withdrawal Times in EFPY Specimen Exposure 3, 10, 17, 24 3, 10, 17, 2h (3) (h) (5) (6)

Unit Startup Date NA 6/79 (7)

Notes:

(1) k0 yrs. at 0.8 capacity factor = 32 EFPY exposure (2) First TMI-l capsule subjected to destructive testing (3) Predicted shift of the adjusted reference te=p. approximately 50 F0 is 3EFPY

, (k) 3/4 service life is 2hEFPY (5) Time interval between first and fourth capsule withdrawal is 21 EFPY therefore 1/3 and 2/3 of this interval is 7 and ik EFPY respectively which yields a withdrawal at 10 and 17 EFPY l (6) Fifth and sixth capsules are spares to monitor subsequent anealing/ irradiation

. (7) Startup on 12/77 except 1 year delay assuned i (8) 2772/2535 x 2.8 = 3 0 (9) Analysis of TMI-1 capsule 1 and other capsules frc= other B&W-177 plants I

showed an actual lead factor of 2.8 for the 177 fuel asse=bly plants (10) Assu=ed to be zero for TMI-1 i

i 1469 133

r

" ~ ' I

?. ' + Ai n D9)bLA(WbMONfl( -

1 i

- -- . . . . - - _ _~ ,

TABIE II

• /

""E S"E INTFC9ATED RV SimVEILY.ANCE SCHEDUT.R

x en[! start 2 Cycle .h'o.

Date 1

7/79 l

.2 .

3/4

/61 A 1/62 1/8!.

5

/S5 6

5/g 7

7/ a.,

8

/ na 9

1/ 89 10 1/ni 11

/o2 12'

/93 13

/ oi, 14 / g

/ 9/ g #h T" -2 Cyala Length 1,3 c,3 c,7 3,7 0.7 0.7 0. 0.7 0.7 0.7 0.7 0.7 0.7 0.7 07 0.WY)/ Elapsed Time 124 138 152 166 180 19h 208 2h 40 54 68 82 96 110 222_

Prnm S/U D'as. )

FN Ex}m ura T!4I-1 3. C, ( l ' 5.5 6.6 75 8.4 9.h 30.3- 11.2 12.2 13.1 14.0 15.0 15.9 16.3 17.[Q nt be b.1 .h.8 - .6.2 65 7.6 6.3 9.0 97 10.4 (LFPYIinCycle TMI-2 ~0 1.2 2 2.7 3.4 s.5 TMI-1 Capsula l' NA CAPSUIS TEST 2D Effcetive 6.0 10.2g NA4 TESTED 2 0** 3.6 8.1 e

' Exposure at g 8.1 10.2 12.3 lb.h' 16.5 Bedi a Cycle 3 0** i 1. 6 6.0 -

3.0 o o.. 4.5 6.6 '8.7 10.8 12.9 17.1 19.2 21.3 -

23.48 NA T WTl D.

~Lced Factc:. g , 2.4 15.0

( 0 on , b.2 6.3 8.4- 10 5 12.6 15.7 17.8 19.9 22.0(2) 5 }O O O . g,y 22.0(2) 0 h.2 8.h 19.6 15.7 17.8 19.9 6 0 lC 0 o.. 2.1 g, m-f t

LMI-2 Capsde 1 0 3.k' NA CAPSIT.E TESTER Effective 0' 4h '5. 6 7.6, 9.5* NA TmTED

~2

- Exposure at 5.6 7.6 95 11.5 13.h - 15 h NA TETED 1 0  ! 3.4 17 h*

Gecin Cycle 95 15.h 17.h 19 3 21.3 23.28 NA TR TED 4 0** 3.h 5.6 7.6 11 5 13.h g ; g t - 2.8 , , 0 0 0 0 0 0.. 2.0 3., 5.9 7.e ,.e ,1.8 , , . ,n3 5 9.8 0 0 0 0 2.0 3.9 59 7.8 11.M 21

, 6 0 0 o o g.,

T & l capscle !;cs. W 7 (2)

Installed 2.3 2.3.h

• 3.h.5.6

• h.5.6-5.6 1,2,'3,4 2,3,4 ,, p 3,4 >- 3 . 13 , 5 _ h,5.6 . r 5,6 e (2) y

' (1) Erpc ns 7/79, L.3 from "MI-1 Cyc' e_1 plus 3.25 yra 9 0.8 caiacity (3/76-7/79) ' 3 9 EFP(

1. (2) Caps sure ules 5begi{6& for both units p' anned to be removei and sto) ed when e xposed to 32'EFPY l

• l .

^

( y w ,

• Rewved

• • Inse rted ,

.C>

i LIST OF TABLES Table Title Pa~e 2.3-1 Reactor Protection System Trip Setting Limits 2-9 1

3.5-1 Instruments operating Conditions 3-29 h.1-1 Instruments Surveillance Require =ents h-3 h.1-2 Minimum Equipment Test Prequency 4-8 h.1-3 Minimum Sampling Fn quency h-9 h.2-1 Instrument Surveillance Program h-14 h.2-2 Three Mile Island Nuclear Station Site Integrated Reactor Vessel Surveillance Program 4-27a k.h-1 Selected Tendons and Corresponding Inspection h-35a Periods h.h-2 Tendons Selected for Tendon Physical Condition h-36 j Test I

k.h-3 Ring Girder Surveillance h-36g i

h.15-1 Radioactive Liquid Waste Sampling and ,aalysis h-59 4.15-2 Radioactive Gaseous Waste Sampling and Analysis h-63

) 6.12-1 Protection Factors for Respirators 6-23 1

1469 iM i

.I I

f I

loads are introduced by unit load transients, reactor trips, and unit heatup and cooldown operations. The number of thermal and loading cycles are 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 satis"les stress limits for cyclic operation.(2) The 200 psig pressure limit for the secondary side of the stea=

generator at a te=pe a ure less than 1000 F satisfies stress levels for tempera-tures below the DTT.13 The reactor vessel plate material and welds have been tested to verify conformity to specified requirements and a maximum NDTT value of 30 F has been determined based on Charpy 7-notch tests. The maximum NDTT value obtained for the steam generator shell material and welds was 40 F.

The heatup and cooldown rate limits in this specification are not intended to limit instantaneous rates of temperature change, but are intendad to limit temperature changes such that there exists no one hour interval, in which a temperature 6hange greater than the limit takes place.

Figures temperature3.1-1and3.1-2containthelimiting{eactorcoolantsystempressure-relationship for operation at DTT h) and below to assure that i

stress levels are lov enough to preclude brittle fracture. These stress levels and their bases are defined in Paragraph 4.3.3 of the FSAR.

I as a result of fast neutron irradiation in the region of the core, there vill be an increase in the NDTT with accumulated nuclear operation. The pre maximumNDTTincreaseforthe40-yearexposureisshownonFigure4-10.picted 4) The

, actual shift in NDTT will be determined periodically during plant operation by testing of irradiated vessel =aterial samples in accordance with specification k.2.2. The results of the irradiated sample testing vill be evaluated and compared tc the design curve (Figure 4-11 of the FSAR) being used to predict the increase in transition temperature.

Thedesignvalueforfastneutron(E>1MeV)exposureofthereactorvesselis 3.1 x 10 0 n/c=2 2 exposure of 3.0 x seeatthereferencedesignpowero{5)568MWtandanintegrated 10 19 n/cm2 for 40 years operation. The calculated maximum l values are 2.2 x 1010 n/cm2 see and 2.2 x 1019 n/cm2 integrated exposure for LO years operation at 80 percent load.(h) Figure 3.1-1 is based on the design value which is considerably higher than the calculated value. The DTT value i for Figure 3.1-1 is based on the projected NDTT at the end of the first two j effective full power years of operacion. During these output has been conservatively estimated to be 17 x 10gvo years,megawatt thermal the energy days, I which is equivalent to 655 days at 2568 MWt core power. The projected fast neutron exposure to the reactor vessel for the two years is 1.7 x 1010 n/c=2 which is based on the 1.7 x 106 thermal megawatt days and the design value for f' fast neutron exposure l The actual shift 'n NDTT vill be established periodically during plant operation l by testing vessel =aterial sa=ples which are irradiated by securing them I periodically near the inside vall of the TMI-2 reactor vessel in the core l area to achieve an averag9,gffective exposare between 1 and 3 times that of I

the reactor vessel at ht. W To ec=pensate for the increases in the NDTT

! caused by irradiation, the limits on the pressure-temperature rel" ionship 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 sensitive to integrated reactor power and not to instantaneous power level.

Figures 3.1-1 and 3 1-2 are applicable to reacter core thermal ratings up to 2568 MWt.

!, 3h )4h9 l3b

The pressure limit line on Figure 3.1-1 has been selected such that the reactor vessel stress resulting fro ir.ternal pressure will not exceed 15 percent yield strength considering the following:

a. A 25 psi error in =easured pressure

b. System pressure is measured in either loop

c. Maxi =um differential pressure between the point of system pressure

=easurement and reactor vessel inlet for all operating pu=p conbinations For adequate conservatism, in lieu of portions of the Operational Require =ents of Appendix G to 10 CFR 50, a maximum pressure of 550 psig and a maximum heatup rate of 500F in any one hour has been i= posed belov 275 F as shown on Figure 3 1-1.

The spray te=perature difference restriction, based en a stress analysis of the

. spray line nozzle is i= posed to maintain the thermal stresses at the pressurizer spray line nozzle below the design li=it. Te=perature requirements for the

steam generator correspond with the =easured NDTT for the shell.

l REFERENCES (1) FSAR, Section h.1.2.h (2) ASME Boiler and Pressure Code,Section III, N-kl5

! (3) FSAR, Section h.3.10 5 5

(h) FSAR, See" -1 h.3.3 (5) FSAR, Sections 4.1.2.8 and h.3.3

. (6) BAW-10100A t

i I

l, 1469 137

. 3-5 t

I

h.2 REACTOR COOLANT SYSTEM INSERVICE INSPECTION Aeolicabilitz This technical specification applies to the inservice inspection of the reactor coolant system pressure boundary and portions of other safety oriented system pressure boundaries as shown on Figure h.2-1.

Objective The objective of this inservice inspection program is to provide assurance of the continuing integrity of the reactor coolant system while at the same time minimizing radiation exposure to personnel in the performance of inservice

, inspections.

I Specification l

l h.2.1 The inservice inspection program to be followed is outlined in Table h.2-1. Except as provided for in tnis Table and as discussed herein, the inservice inspection program is in accordance with the ASME Code, i Section XI, Rules for Inservice Inspection of Nuclear Reactor Coolant I

Systems, dated January 1, 1970, as mcdified by the Winter 1970 Addenda.

Prior to initial plant operation a pre-operat ial inspection of the

. plant vill be performed of at least the areas listed in the ASME Code, provided accessibility and the necessary inspection techniques are available for each of these areas. The only exception to this vill be areas where the necessary base line data is already available and has been obtained by the same techniques as vill be used during inservice inspection.

, 4.2.2 Reactor vessel irradiation capsules are planned to be withdrawn for testing at specimen exposures (E > 1Me\) equivalent to 3, 10, 17, and

~

24 effective full power years. Withdrawal schedules for testing may be modified to coincide with those refueling outages most closely

, approaching the testing withdrawal schedule and may be adjusted l following evaluation of data from each withdrawal in accordance with 10 CFR 50 Appendix H paragraph II.C.3.g. Speci=en capsules shall be irradiated in the TMI-2 reactor vessel in accordance with the schedule given in table h.2-2.

l

, k.2.3 The accessible portions of one reactor coolant pump motor flywheel 8

assembly vill be ultrasonically inspected within 3-1/3 years, two within 6-2/3 years, and all four by the end of the 10 year inspection interval. However, the U.T. procedure is developmental and vill be used only to the extent that it is shown to be meaningful. The extent of coverage vill be limited to those areas of the flywheel which are accessible without motor dicassembly, i.e., can be reached through the access ports. Also, if radiation levels at the lover access ports are prchibitive, only the upper access ports vill be used.

1469 1 M

b. The vessel specimen surveillance program is based on specinen equivale +5) exposure years These times of 3,10,17, were selectedand 2h EFPY to meet referenced the requirements to 1/h t .

of Appendix H to 10 CFR 50 for the 32 EFPY service life of the reactor vessel.

The planned withdrawal schedule is based on a TMI-l specimen lead factor of 3 0 when installed in the TMI-2 reactor vessel surveillance capsules holder tubes. The schedule provides periodic exposure of the various capsules such that a lead factor between 1 and 3 as required by Appendix to 10 CFR 50 is obtained. Provision has also been made to monitor the effects of anaealing and subsequent exposure should it ever be necessary.

Therefore, the surveillance program provides sufficient data to substantiate the conservatism of the pressurization, heatup, and cooldown limits.

l

c. The reactor coolant pump motor flywheel ultrasonic test procedure is being developed to detect flaws of a scall enough size to provide assurance of continued integrity, based upon a conservative I fracture mechanics evaluation.

l REFEMENCE t

(1) FSAR, Section b.h l

i (2) BAW-10100A i

i I

\

)h(0 l

l L-13

l -

TABLE h.2-2  !

TilREE MILE ISLAND NUCLEAR STATION SITE INTF.'. GRATED

• REACTOR VESSEL CURVEILLANCE PROGRAM t

I Hequired ,

Capsule exposure Capsule No. Capsule No.

When with itawn TMI-l INI-2 '

inserted withdrawn inserted withdrawn ,

O EFI'Y(1) 2,3 1,2,3,4 iG 43 EFPY h 1 i

% 10 EFFY 5,6 2 2

% l'( EFPY 3 5,6 3

% 24 EFPY h h h N 32 EFPY 5.6 5,6 (1) initial Installation at the beginning of THI-2 cycle 1 4

h-2 ~{a C