ML19262A401

Forwards Responses to NRC 761217 Request for Info Re Reactor Vessel Surveillance Program

ML19262A401
Person / Time
Site:	Crane
Issue date:	01/20/1977
From:	Arnold R METROPOLITAN EDISON CO.
To:	Reid R Office of Nuclear Reactor Regulation
References
GQL-0066, GQL-66, NUDOCS 7910290475
Download: ML19262A401 (30)

Text

NRC FORM 195 U.S. NUCiE AR REC'JLA I

ilON DOCKET Ns *18E R J O,2 g 9 i2 7ei NRC DISTRIBUTION FoR PART 50 DOCKET MATERI AL T6 MR Reid FROM: Metropolitan Ediscn Co

"~

1-20-77 Reading, Pa DATE RECEIVE D R C Arnold l-25-77 3 LETTER O N OTO R IZ E D PROP INPUT FORM NUMBER OF COPIES RECEIVED 2 ORIGIN AL YUNC LASslFIE D OcoPY One signed DESCRIPTroN ENCLOSURE Ltr re our 12-17-76 request....trans the Respenses to NRC questions concerning following:

Reaccor surveillance prggram.........

(1 cy encl rec'd) l 15 a y

EfD'7 ?.sgQ

'$ Nn Dn 01.,4 PLANT NAME:

Tnree Mile Island #1 0r 1484 069 SAFSTY FOR ACTION /INFORMATION rwTun 1-26-77 ehf ASSIGNED AD:

AsqTPwn An.

4 BRANCH CHIEFt fe e d-5)

RMANPM PMTFT*

/

PROJECT MANAGER:

- - 2 w. a ft i s PROJECT MANAGER:

- V LIC. ASST. :

- l es,e N t e LIC. ASST. :

INTERNAL DISTRIBUTION l

TRE'G FIII i

SYSTEMS SAFETY I

PLANT SYSTEMS SITE SAFETY &

I

/ NRC PDR I HEINEMAN TEDESCO ENVIRO ANALYSIS

/I & E (2./

SCHROEDER BENAROYA DENTON & FUT T M i

/ OELD IATNA9

- 2 GOSSICK & STAFF ENGINEERING I

IPPOLITO ENVIRO TECH.

I MIPC i MACARRY l

KIRKWOOD ERN E I

CASE KNIGHT BALLARD I

HANAUER SIUWEIL OPERATING REACTORS SPANGLER HARLESS PAWLICKI STELLO I

SITE TECH.

PROJECT MANAGEMENT I

REACTOR SAFETY OPERATING TECH.

CAMMILL t

BOYD ROSS

/

EISENHUT STEPP I

P. COLLINS NOVAK

/

SHA0 HULMAN HOUSTON ROSZTOCZY

/

BAER PETERSON CHECK

/ BlTILER SITE ANALYSIS i

MELTZ

/

GRIMES VOLLMER HELTEPIS AT & I BUNCH SKOVHOLT SALT 2 MAN

/

J. COLLINS RUTBERG KREGER EXTERN AL DISTRIBUTION CONTROL NUMBER LPDR:d % 6.

/s NAT. LAB:

BROOKHAVEN MAT. I.ATL TIC:

REG V.IE ULRIKSON (ORNL) f 4 tiSIC:

LA PDR

!ASLB:

l CONSULTANTS: _

1,

,IACRS /s CYS McEDING/ BENT AS d#7 4//>b /I77 I

I I I

NRCFORM 196(2 768 4 91 o g o V76 4

2 YYE ner a nomss METRCPOLil~AN EDISON COMPANY POST OFFICE BOX 542 RE ADING. PENNSYLVANI A 19603 TELEPHONE 215 - 929 601 January 20, 1977

\\g 7

% 00t0

/

\\g N

5L

& @on 25; p e r % '@%

k G e 50.,y -

Director of Nuclear Reactor ? setor Regulation L

7 Qf'y, 4.

.Ii Attn: R.

'.f.

Reid, Chief 3

Cperating Reactors Branch No. h

-/

"\\

Washington, DC 20555 b%

~f/

U. S. Nuclear Regulatory Ccnsission

/

/

+ D-

Dear Sir:

Tnree :lile Island Nuclear Station Unit 1 (3!I-1)

Docket No. 50-289 Cperating License DFR 50 Enclosed please find responses to questions forwarded with your

' letter of December 17, 1976.

Sincerely, r,'i

'/

a. C. 2nold VicePrejl,8ent

s.. N RCA:JJM: rk

/e J

\\,.

~ ~.

Encicsure

\\

\\ ' '3

%@o

!Q'

~'

ou s

o w

l ySI..)h\\ '", Y -

~'

l!-

-- i

=,

.e

- z.!

m s

e / i-, 'x b2(i 1484 070

RESFCUSES TO 3RC QUESTIONS CU REACTOR VESSEL SUFJEILLANCE FRCGRAM CUESTICU 1.

Provide your centingency plans for assuring that your surveillance program vill not be jeopardized by ar extended outage of any other reactor (s) from which you expect to receive data. What time limits vill you place on the hest reactor (s) for a given cutage and justify these limits.

RESPONSE

B&W has developed a diverse ec=bined program for irradiating surveillance specimens of welds of interest between operating reactors and test reactors. Such a diverse progrs= will offer protection against an extended cutage of the host reacter shculd this cecur.

Redundancy will be inecr-pcrated in the conbined program by ensuring that most of the representative velds to be irradiated in operating reactors vill also be irradiated in the test reactors.

The fluence levels in the test reacter prograns should be sufficiently high to ensure that the surveillance material stays ahead of the corresponding reactor vessel beltline region.

This, in itself, vill allow for senevhat other than nornal cutages at the host reactor. Also, there is redundancy incorporated in the operating reactor program so that if an outage occure at ene host reactor, at least one other hest reacter vill have representative veld metal in a neutron environment.

In si--a ry, the conbined surveillance program offers a double redundancy feature for the irradiation of representative veld metal should the host reactor suffer an extended cutage.

There is no time limitation on an outage at the hest reactor. The operaticns of this plant vill be =cnitored as discussed in response to question 6.

Shculd it be determined that an extended cutage has the pctential for allowing the fluence en the guest reactor vessel to apprcach the fluence on the surveillance capsules at the host reactor, a review of alternative sources of surveillance data vill be made, as discussed in respense to question 5 The appropriate corrective action vill be taken fclleving review by the XRC.

The time that the host reactor can renain out of service is, of course, a function of the prior service and after a few cycles of operation, it would essentially have to be retired frem service to seriously jeopardize the program.

Since TMI Unit #1 (guest) and TMI Unit #2 (host) are to be operated by Met-Ed, no special arrangenents are necessary to keep track of the host reactor perfernance.

1484 071 9

QUESTION 2.

Provide your program and schedule for installing the redesigned surveillance capsule holders in your reactor in the event this action becctes necessary.

_RESPOUS E Due to the availability of applicable surveillance data frc= altarnate sources, installation of surveillance specimen holder tubes (SSET) will not be necessary.

In order to re-install holder tubes on TMI-1, 3&W =ust first complete the development and testing of a substantial snount of required tooling. A tabulation of the required tcoling and its current status is given in Table 1.

B&W estinates that 26 months will be required to complete the development and testing of the above tcoling.

This 26 =cnths would have to be expended before holder tube installation can be initfated. Once the tooling is developed an estimated 3 months vould be required to install three surveillance capsule holders en an irradiated plant. This time estimate does not include any contingency for an inproperly installed tube or failure of any tooling to perform as planned i tested.

At this time, Met-Ed and B&W are not proceeding with centinued develcpment of tooling for the installation of surveillsace capsule holders or irradiated plant s.

As discussed in our September 9,1976 submittal of Technical Specification Change Request No. 38, Met-Ed does not consider that re-installation of holder tubes en TMI-l is a reascnable alternate, based on the costs and excessive personnel radiation expcsures which would result.

The B&W estinates en the time required to develop and test the necessary tooling further supports this position. 1484 072

CUESTIO" 3

What is the schedule for withdrawal of your capsules frem the host reactor (s)? Relate the schedule to predicted trends in adjusted reference tenperature and Charpy upper shelf energy. What arrangements have been made with the evners of the host reactors to assure that this withdrawal schedule vill be met.

RESPOUSE Table 2 lists the withdrawal schedule for the surveillance capsules, as related to the appropriate cycle at T E-2.

Table 3 presents the basis and j ustification for this withdrava.'. schedule. Table 3 relates the schedule to the actual and predicted trends in adjusted reference temperature and Charpy upper shelf enerr/ of the surveillance veld retal.

Since TMI Unit #1 and Unit #2 are bcth to be operated by Met-Ed, no outside arrangements to assure that this withdrawal schedule vill be met are required.

i48;4 073

- 3 _

m e.. re J c

mu h.

SpeciP/ the minimun and maximum radiation lead times for:

(a) surveillance specimens relative to the vessel beltline inner surface, and (b) surveillance specimens relative to the 1/h? position in vessel vall, which you will require for guest specimens exposed in the host reactor (s). Justify the values specified.

RESPONSE

In the near future, Met-Ed will submit for NRC approval a change to IMI-l Technical Specification 3.1.2.

This change vill provide modified heatup and cocidown limits for up to 6EFFY cf cf IMI-1, gperations is 3.2 x 10 gssel fluence predicted for 6EFPY TMI-l operations. The reactor y

"/c=" at the inner surface and 1.8 x 10^" "/cm# relative to the 1/kT position in the vessel vall.

For operations beycnd 6EFPY, surveillance data for the reactor vessel beltline velds of interest is required to assure that the requirements of 10CFR50, Appendix G, Section V.3 are cet for future service periods.

The predicted properties of the OII-l reactor vessel material and existing surveillance data indicate that all future pressure / temperature limitations en the reactor vessel vill be controlled by the irradiated properties of the beltline velds. The changes in properties of the other reactor vessel materials are expected to te insignificant compared to these of the velds.

Technical Specification Change Request No. 38, which was submitted on September 9,1976, requested approval of a TMI-1 and 2 site integrated reactor vessel surveillance program.

This action was required as a result of damage reported in our letters of I-hrch 18, 1976 and March 23, 1976 which necessitated renoval of the TMI-l reactor vessel surveillance holder tubes. At the tine Technical Specification Change No. 38 was prepared, several iters were unknown. First, it was not kncvn that additional data en the TMI-l beltline velds vculd be required as early as 6EFPY.

It was also not kncvn that the TMI-2 holder tube design would be changed to those currently installed in Eavis-Besse and Crystal River and thus that higher lead factors would be available. Also, it was assumed that all new upgraded DE-1 surveillance ca;sules possibly could be required.

Current plans are to install two surveillance capsules containing three (3) different heats of veld retal in addition to the existing TMI-1 and 2 surveillance caps ules.

1484 074 4,

RESPONSE to Question h (continued)

Table 7 demonstrates that a site integrated surveillance progran to irradiate both TMI-1 and TMI-2 capsules in the TFE-2 reactor vessel can provide the data required on the TMI-l beltline welds to support operations of TMI-1 after 6EFPY. Please note that Table 7 basically updates the information provided in Table 2 of our Technical Specification Change Request No. 38.

As a result, the following conservative assumptions were also included in the preparation of Table 7-1.

Start-up of TMI-2 is delayed frct 12/77 to 7/79 2.

TMI-1 cperates at. a 0.8 capacity facter.

3 TLE-2 operates at a 0.6 capacity factor.

In addition, Table 7 assu=es that 13 capsules must be irradiated whereas Change Request No. 38 only assumed 11 capsules required irradiaticn.

The results presented in Table 7, therefore, demonstrate that, even given a delay in TMI-2 operations and pocr TMI-2 performance, the radiation lead times which will be available in TMI-2 are justified.

As discussed in response to question 6, a review of the radiaticn lead times which exist en TMI-l and 2 will be undertaken by early 1979 This review will determine if timely surveillance data exists or will exist to support further modification of the TLE-1 heatup and cocidown limits for periods in excess of 6 EFFY. Should this review show that corrective action as discussed in response to Question 5 is required, at least 3 calendar years will be available to obtain the necessary data and prepare the Technical Specification amendment prior to reaching 6EFFY cn TSE-1.

Thus, with the conservative assumptions of Table 7 and the review to be conducted of the conservatism in these assumptions, it is not necessary to specify minimum radiation lead times at present.

5hximum lead times are not specified, since the withdrawal schedule discussed above will provide the required surveillance data at the proper intervals of service life regardless of the oporational status of IMI-1.

Should TMI-1 operations be interrupted for an extended period of time with continued operation cf IMI-2, the effect would be the ability ta verify reactor vessel material properties for a longer period cf service than would otherwise be pcssible.

e 1484 07o

~-y.,

wus01.sa 5

Indicate the corrective action to be undertaken at the guest reactor if the limits specified in response to Question h, above, cannct be met.

If the correc;ive action dces not involve reacter shutdevn, justify the proposed alternative.

RESPOUSE Since there are several alternatives, the corrective action vill nct involve reactor shutdown. As discussed in the answer to Questien 1, 3% has developed a synergistic surve111snee program in which several

': elds vill be irradiated in three operating reactors (Davis Besse 1, Crystal River 3, and Three Mile Island 2).

In addition, data for several of the same velds will be obtained in at least two test reactor irradiation programs.

The velds to be irradiated and the test reacter programs are described in the ansvers to the questions en " Traceability of Welds".

The proposed B&W progrs= assures that applicable data for the 177 FA, B&W design reactor vessels vill be available through the design service life of the vessels.

In the event that the TMI-2 reactor is delayed in startup cr has an extended cutage which is of sufficient duration to endanger the timeliness of the data availability, several possibilities exist that would minimice the i= pact of such an outage. Such possibilities or alternatives are:

1.

An evaluation of the applicability of the available data (frc=

that reactor er other reactors, including test reactors) to SII-l could be made.

Such evaluation nay indicate TMI-l capsules do not need to be irradiated within the expected time of TMI-2 cutage.

2.

The capsules which vill generate applicable data for the TMI-l reactor can be renoved frc= and inserted into another hest reactor that is in operation.

3 The specimens which will generate applicable data for IMI-l can be re=cved frc= TMI-2 and inserted inte a test reactor.

h.

The pressure-temperature limit curves of 3II-l could be developed with material properties conservatively assumed until applicable data is available.

The best alternative can only be chosen at the time at which the extended cutage occurs, since all the above options require evaluation,cf the data which is or vill be available in a timely =anner. gg4 076

OUESTION 6.

Describe how the cperating staff of the guest reacter vill keep informed of the exposure status of the guest specimens at the hest reactor (s) relative to the limits specified in response to Question h, above.

?ES?CNSE 3cth TMI-l and TMI-2 are to be operated by Met-Ed.

Therefore, the operating staff and =anagsnent of Met-Ed will always be aware of the operational status Of both of these units. As discussed in response to Questien h, even assuming a 1-1/2 years delay in operation of TMI-2 folleved by peor TMI-2 perfor ance, a site integrated sme111ance program is still feasible. As a result periodic =cnitoring of the operational status of T.C-1 and I::I-2 is not required.

Ecvever, prior tc Nhrch 1979, Met-Ed vill review the radiatien lead times which exist en TMI-l and TMI-2 and access whether any of the alternate acticns discussed in response to Question 5 are required.

In the event that alte nate action is required, the NEC will be informed at this time of our intended course of corrective action.

1484 077

~Lmem n 4 W

ww

.v 7

Submit amended proposed Technical Specifications that reflect the apprcpriate portions of your responses to Questions 3, k, 5 and 6 abcve.

RESPONSE

Eased on our respense to Questicns 3, L, 5 and 6, the onl-; Technical Specification Changes which are considered necessary are:

1.

revised heat-up and cocidown ll=itations for operations of TMI-l up to 6EFPY.

2.

and either a change to Technical Specification 'hange h.2 to allow operations of TMI-l during Cycle 3 vithout surveillance capsules installed or a change to specification h.2 to reflect the planned integrated surveillance program such as was submitted in Change Request No. 33.

These Technical Specification Changes will be submitted for NEC approval in the near future.

i484 078 FLUENCE ESTIMATES QUESTION 1.

Describe analytical techniques that you plan to use to esticate the fluence expected a* the varicus velds of the be tline of your vessel.

Mcv much uncertainty do you expect there to be 10 the fluence estimates?

RESPONSE

Energy dependent neutrcn fluxes are determined by a discrete ordinate solution of the 3citerann transport equation. Specifically, ANISN, a one-dimensional ccde, and DCT, a two-dimensional ecde, are used to calculate the flux at the detector positicn.

In both codes, the system is modeled radially from the core cut to the air gap outside the pressure vessel. The model includes the core with a time averaged radial pcVer distribution, core liner, barrel, thermal shield, pressure vessel, and water regions.

Inclusion of the internal conponents is necessary to account for the distortions of the required energy spectrum by attenuation in these cc=penents. The ANISN code uses the CASK 22-group neutron cross section set with ;n S-order of angular qu.drature and a P, expansicn of b

s the scattering = atrix.

The problem is rua along a radius across the core flats. Acimuthal variations are obtained with a DOT r-theta calculation that =odels one-eighth of a plan-view of the core (at the core midplane) and includes a pin by pin, plant specific ti=e averaged pcVer distribution.

the DCT calculation uses S6 quadrature and a F1 cross section set derived frcm CASK.

Fluxes calculated with this DCT =cdel must be adjusted to acccunt for lack of P cross section detail in calculations of anisotrcpic scattering, a pert $rbation caused by the presence of the capsule, and the axial pcVer distribution. The first two ite=s are both energy and radial-location dependent whereas the latter is axial location dependent. A P /P, correcticn q

factor is obtained by comparing tvc ANISN 1-D :odel calculaticf.s in which only the Order of scattering was varied. The capsule perturbation factor is obtained from a ec=parisen of two DCT x-y redel calculaticns, one with a capsule explicitly =cdeled - SS3Ch cladding, Al filler region, and carben steel specimens--and the other with water in those regions.

The effect cf axial pcVer distribution is determined frc= plant specific burnup calculations as a function of axial location for the outer rows of fuel assemblies. The net result from these parameters studies is a flux adjustment factor K which is applicable to the appropriate desireters in the 1TT-FA surveillance pro grams. 1484 079

Fluence Estimates (continued)

The calculation described above provides the neutron flux as a functicn of energy at the dosimeter position. These calculated data are used in the fc11cving equations to cbtain the calculated activities used for cc=parison with the experimental values.

The basic equation for the activity D (pCi/gn) is given as follows:

un

=

C:I f E c (E) $ (E) E i j,-Y (,,-Tjs

~y s

D i

( y_,~ )

y A 3.T x 10*

E n

"j

~

i j=1 where C = normalizing constant, ratic of measured to calculate flux

I = Avagadrc 's number, A = atc=ic veight of target i, f = either weight fraction of target isotope in nth

=aterial or fission yield of desired isotope, c (E) = group-averaged cross sections for material r.,

$(E) = group-averaged fluxes calculated by DCT analysis, F = fraction of full pcVer during jth time interval, t j

3, u

A = decay constant of ith raterial, 1

t = interval of pcVer history, j

T = sun of total irradiation time, e.e., residual time in reactor, and vai'

-a between reactor shuticvn and ccunting, t = cumulative time frca reactor startup to end cf jth time pericd, i.e.,

T _j j - I t,s k=1 The nornalizing constant C can be obt. tined by squating the right side of the above equation to the reasured activity. '4ith C specified, the neutron fluence greater than 1 Mev can te calculated frca 15 Mev M

• (

>. 0 7p,

+r 4 ( ~e )

r r

e,

+

~

o 4 '

r=.1 o=-

where M is the number of irradiaticn time intervals; the other values are defined abcVe. 1484 080

Fluence Estimates (centinued)

The analytical model described above, for calculating fast fluence at the surveillance capsule includes the pressure vessel rt eion.

Thus each cal-culation produces fluence data at the veld position as well as the capsule loc at ion.

Since analytical results currently being documented cenpare within + 15". to dosineter ceasurements frcm surveillance capsules fren 5 reactors to date, calculated data at the nearby veld positions should have similar reliability.

Dosimeter data ecmparisons from surveillance capsules irradiated at the host reactor vill provide further ce=parisons with the analytical scdel. Because of the similarity of the hest and guest reactors, these cenparisens vi.~.1 also be applicable to the pressure vessel fluence calculation for the guest reactor since it uses the same analytical mo del.

3&W intends to docu=ent the uncertainty based on the contributing factors in both the calculation and the measurements from the present capsule evaluations. This documentation vill be available folleving ec=pletion of the surveillance capsule results and submittal to :IRC in the form of a Topfcal Repo rt is expected by June 1977 1484 081

Fluence Estirates (centinued)

Question 2.

Describe any dosimetry checks that you plan to make on the analytical results.

Restense Desimeter measurements from Oconee I (cycle 1 and 2), Oconee II (cycle 1),

Oconee III (cycle 1), TMI-l (cycle 1) and A'!O-1 (cycle 1) have been ec= pared to the analytical model. A nominal difference of + 15% vas noted in the fast flux (E>l tel). Multiple dcsimeters in surveillance capsules will be in the host reacters and also in subsequent 3&W plants to startup in the 1980's. When each capsule is removed dosimeter activitier vill be teasured and then cc= pared to the plant specific analytical result.

'"his will provide data for further vertification ec=parisons with the analytical technique which will be used for plant specific fluence calculations at both the host and guest reacters. ?Ic check is considered necessary for calculated data at the veld locations as noted in response to question #1.

2 _

1484 082 7

_F_luen c e EstL=ates (continued)

Cuestion 3

What differences in neutron energy spectra and dose rate do you predict for your reactor beltline and your surveillance specimens, wherever they are to be irradiated? Describe the corrections, if any, that vill be made to the predicted radiation damage at your beltline velds as a result of these differences.

Possible corrections include differences in specimen irradia-tien temperatures, differences in neutron spectra arising frc= differences in reactor gec=etry cr a different type of fuel (e.g. mixed exides), and differences in dose rate if sc=e test reactor data are used.

Respense For the same fuel type (e.g. Icv enriched uranium', relative neutron energy spectrum is a function of only the internals cc=penents (gecretry and materials).

The internal ec=penents design is the same for both guest and host reacters as discussed in response to the question under

" Similarity of Guest and Ecst Reactors".

Thus the relative energy spectrum at the same spatial location should not vary between reactors. (and consequently dose rate vill vary directly as the fast flux). The analytical

=cdel is a multigroup calculation with the same intervals arrangement using plant specific core parameters as discussed in response to question 1.

Consequently no correction is required between plants since the significant variables are already accounted for in the calculation. The use of mixed oxide fuel vould harden the spectrum sotevhat but any effect on dose rates should be within the analysis uncertainty limits. Possible ccrrections in using data f c= test reactors will depend en the design of the test re acte r e

program, which is nct final. Since i= pact, tensile and fracture data on many of the same materials, vill be obtained both frc= test reacters and the surveillance programs, a basis for ec=parison vill ce available.

Such ec=parison vill determine if correction would be needed.

)k04 SIMILARITY OF GUEST AND HCST REACTCRS Questien 1.

Provide a ec=prehensive tabulation fcr the guest reactcr and each hc :

reactor, of the values cf all parameters, including ecnstruction and operating characteristics, that may affect the fracture toughness of.he reactor vessel material as it is irradiated. Discuss how all differences in these para.=eters are accc==cdated in the integrated surveillance program.

RESPOUSE The reactor parameters which could possibly affect the material prcperties as the vessel is irradiated are 1) the neutron flux energy spectrum,

2) the irradiation rate, 3) the irradiatica tenperature, and h) the material type and initial properties.

Each of these is addressed belev.

Energy Spectrum - As discussed in the response to the questions under

" Fluence F.sti=ates", the relative neutron energy spectrum is primalily a function of the gec=etry and materials of the reactor internals components. As shcvn in Table h, the dimensions and caterials of both TMI-l and TMI-2 are essent ? 2117 identical. Thus, there is no difference to be accon=cdated.

Irradiation Rate - Any signiricant difference in dose rate obtained at the TMI-l and IMI-2 vould be due to the variations in power level and pcVer distribution. Since the licensed power levels are ec= parable, the only difference is the variation in 1 cad svings as the plant maneuvers.

When time averaged ever multiple fuel cycles, the varia icn in pcVer level and pcVer distribution due to maneuvers is expected to be ec= parable betweea plants. The ec=carability of reacter vessel surveillance results frc= a number of plants presently available supports this.

Thus, there are no significant differences to be acec==cdated.

Irradiacion Terrerature - There are two differences in irradiation terperature ceasidered.

The TMI-l reactor vessel beltline inner surface and the surveillance specimens in TMI-2 vculd be exposed to reacter coolant at essentially inlet conditions. The temperature distribution in the sur-veillance specimens and capsules is centrolled primarily by the terperature of the reactor ecolant. This is due to the goed heat trannfer characteristics of the specimen / capsule configuration.

Thus, the variation in reacter coolant inlet tenperatures due both to design difference and the variation as the plant is maneuvered must be considered.

The variation due to design differences between the hest and guest reactors is insignificant as shcvn on Table h.

Between parcial ( 15%) and full lead conditions, the inlet a

te=perature vill vary by about 20 F as an inverse function of pcwer level.

Figure h-9 in the FSAR shcws this variation. The duration of this variation due to =aneur ' ring is expected to be ec= parable between plants over time This is supported by the cc=parability of reacter vessel surveillance results presently available frc= a number of plants.

In any case, the inlet ccndition temperature vill also vary abcut h0 ? between the hct :ere power condition and partial load.

This variation is a direct functicn of pcVer level (0-155) and again is not significant due to the lov te=perature and the expected comparability in duratien over the Icng term. 1484 084

Similarity of Guest and Host Reactars (continued)

Materia' Sfre and Initial Preterties - 3cth the hest and ~uas+ reac*-"s are constructed of similar materials as discusse 1 in ecnj ction -$1+

+v e neutron spectrum e nsideratien. Thus, there is no differenca~ to be accc= cdated.

c

_ ~ -

1484 085

mn cr G.- rm..

e r,S nt C r.

e asA-a H~~r Cuestion 1.

Identify the heats of veld wire ar.d flux used in all beltline welds, and give specific locaticns where each is used.

F.esronse The heats of veld wire and flux used in all beltline region velds, including the surveillance veld, and their specific locations are given in Table 5 1484 086

.O

Trseeabilitv cf 'delds Icentinued)

Questien 2.

State which veld or velds is expected to be centrolling with regari tc radiation damage and why, i.e. give expected neutren flux, initial RT.p,,

Charpy upper shelf energ, and chemical ec= position fer the centrclling' velds.

Response

Table 5 also lists the unirradiated RT..e,,, and Charpy upper shelf energ (CV-USE), the weight percent of the pe'.41nent elements, the expected end of service neutron fluence at the 1/h T vessel vall iccatien, the predicted shift and adjusted RT

, and the predicted drop and adjusted Cv-USE.

As shown in Table 5 vcid' a 70 has the highest adjusted RT..

and the lovest adjusted CV-USE of all the beltline region velds.

EcweYO, velds '4F 25, and SA 1526 also have the pctential cf being the certrelling veld because their predicted irradiated properties are similar to those predicted for veld '4F 70.

The surveillance veld, '4F-25, is considered to be representative of the other tvc controlling velds.

The predicted irradiated properties fer the surveillance veld are similar to those predicted for '4F 70 and SA 1526 at the same fluence value.

Note that the unirradiateu properties af the surveillance veld were determined by testing and these for '4F 70 and SA 1526 are estimated.

1484 087 Traceability of '4 elds (centinued)

Ouestien 3

'4hich velds are represented in the surveillance capsules irradiated in your reactor?

Fespense Weld WF-25 is the surveillance veld centained in the existing surveillance capsules.

Cc= pact fracture specimens for velds V h, W-5, and '4-7 (See Table 6) are proposed to be included in the two new capsules.

-a

- ^ -

1484 088

Traceability of Welds fecntinued) nwuesticn L.

7nich vblds, if any, are represented in surveillance prcgrams for Other reae crs?

Respense Table 6 lists all the velds that are censidered representative which will be irradiated as part of the surveillance program of ?his and other 177FA B & W design pcVer plants. The velds of Table 6 are c 'nsidered representative of the beltline region velds.

1484 089 Traceability of '4 elds (centinued)

Questicn 5

List any test reactor programs en radiaticn damage in which your veld metals are represented.

Res;cns e Presently there are two test reacter programs in which representative welds will be studied. Rese progre.:s are:

1.

HSST Irradiation Studies Program.

2.

IRC-:!EL Inplace Annealing Studies Program Eata from these programs is, of course, readily available to :IEC.

s

%e

%1 I

3 o-1484 090

Traceability of Welds (continued)

Question o.

any etner test reactor and surveillance prcgrams in which velds that Lis:

are expected to be in the sane category as yours from the standpoint of radiation sensitivity are represented, which you intend to utilize.

Response

Other than the progra=s outlined in response to Questions k and 5, 3 L W is investigating the possibilities of irradiating similar veld metals in an FFBI program to be initiated prior to late 1977 1AB4 091

_ m

TABLE 1 STATUS OF REQUIRED SSilT IRRADIATED PLANT INSTALLATION TOOLING

- Tool Status Comments 1.

Boring Mill Completc with Backup 2.

Pintle Renoval Tool Complete, no Backup No backup Necessary l

3.

Drill and Tapper a) Ecsic tool 98% com-I

plete, b) Backup drills and taps must be sealed water-tight and test-cd.

c) Drilling and tapping

_ n_t_other than.pintle locations has not been developed.

t 4..

Thread Inspection Concept only Tool i

5.

Spot Face Tool Basic tool 20% complete 6.

Spot Face Inspec-Concept only tion Tool 7.

S.S.H.T.

Ins *.alla-Concept only l

tion Tool I

8.

Verification of Concept only Bracket Contact Inspection Tool 9.

Crimping Tool Concept only 10.

Free Path Inspec-Complete, no Backup tion Tool 1484 092 i

1 1

_....-~...

TABLE 2 INSERT AND WITilDRAWAL SC11EDULE OF INTEGRATED PROGRAM AT TIIREE MILE ISLAND THIT 2 CYCLES Lead Tube at 1/4T Location Capsule 0

1 2

_3 4

5

_6_._

7 8

9

_1.0 11

_17 13 IIolder Factor h

1.4 1

9.6 Upper THI-L1 x

)

TMI1B

t 2.5 THIID 15 Lower THI2B x

2.25 3

TMIlC

.35 2.3 lj 2

6.9 Upper TMI2A x

THI2F i

1.3 4.5 Lower THI2D

-P-TMI2E 19

.84 3

6.9 Upper TMIIA x

2.5 p6 1

TMI2C 2.08

-- ^

2.5 4

Lower TMI-L2 x

o3

>h O

TMIlF 2

CD The assumed EFPD per cycle are 450 days for the first cycle and 250 days for the others.

w The values to the right of o (Identification of Withdrawal) in the predicted "best estimate" accumulated u

neutron fluence x 10 (n/cm E > 1 Mev) at the capsule location of the host reactor.

8 2

fracture specimens of weld material.

TMI-L1 & Tl!1-12 are new surveillance capsules containind compact PC, 2D, 2E, und 2F ure the TMf-1 A, lb,1C,1D & ]F are the existing TMI-l surveillance capsules and 'ITfT-2A, 2B.

existing TMI-2 survel.11ance capsulea.

x Capuuie Innertion 0 - Capsule Wi thdrawa]

TABLE 3 1

SCHEDULE FOR WITHDRAWAL OF TMI-l's REACTOR VESSEL SURVEILLANCE CAPSULES FROM TMI-2 1

Predicted Impact Properties f Surveillance W Id Metal Approximate Neutron Fluence to be Accumulated by Capsule RTNDT Cv-USE Capsule Time of Withdrawal (E > 1 Mev, n/cm )

(F)

(Ft-Lbs) 2 g

Unirradiated 0

-14 81 TMllE lias been withdrawn for testing 1.1 x 10:e (2a) 103 (3a) 64(3a) fMIIA Following the 2nd cycle at TMI-2 8.4 x 10 (2b) 256(3b) 48(3b) 18 IMI1B To be withdrawn at the time when 1.4 x 10 ' (2c)

No weld metal No weld metal 3

the capsule's accumulated neutron fluence (E > 1 Mev) correspond to that at 1/4 of THI-2 reactor vessel wall location at approxi-mately the end of vessel's design ro" service li fe.

TMI1C To be withdrawn at the time when 2.25 :: 10 ' (ec) 316 (3b) 42 (3b) 3 the capsuic's accumulated neutron fluence (C > 1 Mev) corresponds to that of TMI-2 reactor vessel Inner wall location at approxi-h mately the end of vessel's design service. life i

>2.'5 x 1038 (2c)

No weld metal No weld meta.

TM11D Standby

>2.5 x 10 ' (2e)

No weld metal No weld metal 1

1TMI1F Standby RI) Withdrawal schedules may be modified to coincide with those refueling outages or plant shutdown of TMI-2 most closely approaching the above withdrawal schedule. The schedule may also be modified, if necessary, af ter the evaluation of each capsule.

a) Measured value using doalmeter data.

b) Predicted neutron fluence value for the capsule in the identified location of Table 1. The assumption made on predicting t.he fluence value are given in Table 1.

(2c) Predicted neutron fluence values for TMI-l's vessel. They are measured values extrapolated based on predicted power j

distribution leakage flux, and fuel handling procedures. Values contain a 1.2 safety factor.

(3a)Mcasured values.

RT tused on 50 ft-lb charpy data.

(3b) Predicted values.

TA3LE h COMPARISON OF TMI-l AND TMI-2 Paraceter TMI-l TMI-2 Design Heat Output (Core), fit 2568 2772 Design Cverpcwer. 5 D~ sign Pcwer 112 112 2,200 2,200 System Pressure, Nccinal Coolant Flev Eate, lb/m x 10~g/GPM 131 3/352,000 137.8/369,600 Ccclant Temperatures ( F)

Nominal Inlet 55h 556.5 Avg. Eise in Vessel k8 51 Avg. in Vessel 579 582 Fuel Asse blies, No.

177 177 Fuel Assemblies, ?/pe MK3(15x15)

'KB (15x15)

Core Barrel, ID/CD, in.

141/1h5 lhl/lh5 Therral Shield ID/0D, in lh7/151 lh7/151 Core Structural Characteristics Cere Diemeter, in (equivalent) 128.9 128.9 Core Height, in (active fuel) 1hh lkh Reflector Thicknesses and Cc= position Top (Water plus Steel), in 12 12 Botten (Water plus Steel), in 12 12 Side (Water plus Steel), in 18 18 Reactor Vessel Design Parameters Principal Material SA-533, SA-533 GR.3 Gr.3 Design Pressure, Psig 2500 2500

esign Te perature, "F 650 650 ID of Shell, in 171 171 OD Across No

:les, in 2h9 2h9 Overall Height of Vessel and Closure Head (over CRD and Inst. Nozzles ), ft/in h0/8-3/k h0/8-3/h Cere Barrel and Ther al 30h SS 3Ch SS Shield Principal Material e 1484 09b

s,

\\

f

%.s -

t I

N g

i

\\

l

. + du A 4 21fli Ul1 f!d c'a al.w.. on M d > e ta2 b 1,'

'J'#I' u! m -

a Wi L % h t 1 %/' L'n hi PC1-

-~

TdBLE5 WELDMETALINFORMATIOdANDDATA (TMI-1) 1/4T EOL OL Loc.ation Unirradiated Impact Chemistry Composition Neut. Fluence Shift in Adj sted USE Adjusted Weld Metal in reactor Da6a, Transverse E > 1 Mev RTNDT RTNDT Reduction USE Ident.

Wire M

vessel RTNDT USE Cu P

S Ni n/cm (1)

AF F

Ft-lbs 2

WF8 8T1762 8632 L1 (20)

(66)

.20

.009.009

.61 1.1 x 10

173 193 35 43 8

WF25 299L44 8650 C2,5W

-14 81

.34

.015.013.71 1.2 x 10 '

290 276 44 45 88 WF67 72442 8669 C3 (20)

(66)

.27

.014.017

.57

>7.0 x 10

< 21

< 41

<10 59 ra WF70 72105 8669 Cl,C3 (20)

(66)

.27

.014.011

.46 1.2 x 10

290 310 42 38 os SA-1494 8T1554 8579 L3 (20)

(66)

.14

.015.012.4$

9.0 x 10 (2) 128 148 27 48 SA-1526 299L44 8596 L3 (20)

(66)

.36

.016.012

.60 9.0 x 10

275 295 42 38 (1) Measured values extrapolated based on predicted power distribution flux leakage, and fuel handling p

procedures. Values contain a 1.2 safety factor.

g (2) Ueld conservatively assmned to be exposed to 1/hT fluence, even though materint is not within the 1/hT region or

.p.

NOTE: L1 = Upper longitudinal weld the reactor vessel.

O L2 = Middle longitudinal weld w

L3 - Lower longitudinal weld C0 = liigher-upper circumferential weld C1 = Upper circum, weld C2 = Middle -ircum. weld C3 = Lower circum. weld SW = Surveillance weld

(

)= Estimated per 15AW-10046P

TABLE 6

!!ATERIAL PP.CPERTIES OF PIPPISE::TATIC UE1.DS TO BE 1RPADIATED I:; SURVEILLN:CE PT.00RA?:S OF 177 F. A. ELU DESIG: PIACTOP VESSELS Weld USE(Ft-lbs)

RTiDT (

)

Desip, nation Cu P

v i

W1

.40

.020 67

+65 W2

.22

.024 65 0

1

{

W3

.24

.016 78

+10 I

W4

.36

.011 74

.'O W5

.35

.015 72

+10 P

W6

.24

.022 70

+10 W7

.34

.015 81.3

+9 WF25 * ~

.29

.019 81

+9 WF112 *

.22

.024 65 0

WF182-1 *

.18

.014 83

+15 WF193

.19

.016 66

+15 WF209-1 *

.30

.020 66 443 f

• hterial include in existing surveillance capsules for B&~4 reactor vessels.

e e

i 4

4

1484 0.97

i TABLE T TMI SITE INTEGRATED SURVEILLANCE SCHEDULE c[6/80 TMI-2 Cycle No. &

1 2

3 4/

5 6

7 9

10 11 12 t3

[.

8/89 9/90 11/9.1 - 1/93 2/ 9h

/8

/85

/86 4/8, Start Date 7/79 7/81 9/82 f

TMI-2 Cycle Length 189 [ g (EFPD) & Elapsed Time From S/U (MDS.)

214 38 52 66 80 93 107 121 134 1h8 162 175 1

.21

.28

.34

.38

.44

.49 53 58

.614

.67 72 78

.83

'cMI-1 RV Inner 2xposure Surface N CM2 x 10D T

.12

.16

.19

.22 0.25 0.28 0.30 0.33 0.36 0.38 0. 14 1 0.t!4 0.17 4

0

.09

.14 0.19 0.25 0.26

.34

.39

.44

.h9 53 56

.64 TMI-2 RV Inner surrr.ce Exposure

%T o

.05

.08

.11

.14

.16

.19

.22

.25

.28

.30

.33

.36 n/CM2 x 1019 2

.11**

.46

.66

.8h*

N/A Remov sd For ' 'es ting Existing TMI-l Surveillance 3

.11**

14 0

.69 97 1.2 1.4*

N/A Removed For Te: ting N CapsuleExgosure, n/CM x 10 9

.11**.40

.69

.97 1.2 1. 14 1.7 1.9 2.25*

N/A 2

5

.11**

)

Re'tove i After Cycle l'.

^

6

.11**

temoved 11'ter Cyr le

"< i s ti ng TMI-2 1

0**

.35*

N/A Remc red Fi r Test ing

""~i'.'"""

sere, 2

o" ao n

1 o*

ne ~avr T-ti" n/CM2 x 10 9 3

0*

.35

.55

.TJ 97 1.10 1.3*

N/A Remov d For 'i esting g

5 I#

,t Removed Aftera ycle 13_

0*

CD 5

0**

I Remove (

After ( yele 16 6

0**

)

Removed After ( ycle 19 New Weld Metal 0**

14 0

• N/A Remo red F< r Test ng 1

Surveillance Capaule Exposure, 2

0**

.35 55 76 97 1.10 1.3 1.5 1.7 1.9 2.08*

N/A Ret aved for l'es t ing n/CM2 x 10 9 1

s

• • Inserted

• llemoved

.