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u.s. NUCLEAR REGULATCRY COMMISSl!N DOCKE T NUM T E QNOCact.u 195 j n JR- 3 H &Wm **' ,l .-"''"""2'"NRC DISTRI,8UTiON PcR PART 50 DOCKET MATERIAL

TO: Mr J Fht FROM: Toldeo Edison Co ~~OATE OF DOCUMENTToledo, Ohio-

L E Roe DATE RECEIVED-,

CLETTER ONOTORIZED PROP INPUT FORM NUMBER OP COPIES RECEl% 08 ORIGIN A L ' bNCLASSIFIEDO cop Y - one signed

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DESC RIPT'ON . ENCLOSU RE'

Ler trans the following: Addl info re questions raised about the.reactc :ve ssel surveillance specimen grog' ram. . . . . . . . . .

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SAFETY, FOR ACTION /INFORMATICN rwynn 3-10-77 - eht +

/ ASSIGNED AD: %t 554 // O [4.4I AnnTcve'n An. I(JBriCJLCJ'TEF ? 5Y o / 2s nnANmr mtTrv. I/ , PROJECT !.ANAGER: EMqlo PROJECT MANiGER: I/ LIC. ASST. : FlClfen LIC. ASST. : I

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INTERNAL DISTRIBUTION/ W C FIL2 SYSTEMS SAFETY FLANT SYSTEMS SITE SAFETv &[ NRC FDR HEINEMAN / TEDESCO [LM ENVIRO ANALYSTS[ I & E (M SCHROEDER /' BENAROYA DENTON & Mt'T T rn |

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COSSICK & STAFF ENGINEERING / IPPOLITO ENVIRO TECH.MIPC MACARRY KTRK'400D ERNSTCASE / BOSNAK BALLARDRANAUER / SIINEIL OPERATING REACTORS SPANGLERHARLESS / PAk'LICKI STELLO

SITE TECH.PROJECT MANAGEMENT REACTOR SAFETY OPERATING TECH. / GA}etILL [Z)BOYD / ROSS (LfQ EISENHITI STEPP I

/ P COLLINS / NOVAK SHA0 HULFAN ,g HOUSTON /. ROSZTOCZY BAER '

PETERSON / CllECK BtfrLgg SITE ANALYSISMELTZ CRIMES / VOLLMER (i te)

7 HELTEMES AT & I /. BUNCH' J. COLLINS i,SKOVHOLT SALTZMAN

RUTBERG / KRECER _EXTERNAL DISTRIBUTION CONTROL NUMBER

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C" 9 EDISON iDocket No. 50-34t,

i Mk, 7 LOWELL E. ROEp~

March 2, 1977 '; v'c' **d'a',

vJ 1 Facihtees Deve'ooment- (4191 259-5242, .f .,

Serial No. 232

a .- , ,Director of Nuclear Reactor Regulation AAttn: Mr. John F. Stolz, Chief ygb g *bLight Water Reactors Branch No. 1 *

N g T rearDivision of Project Management .a ,U.S. Nuclear Regulatory Commission m. 7QWashington, D.C. 20555

Dear Mr. Stolz:

During a recent discussion with members of your staff, several questionswere raised about the reactor surveillance specimen program for theDavis-Besse Unit 1 and the specimen withdrawal schedule shown on Table4.4-5 of the Draft Technical Specifications. This letter should answerthese questions, clarify the surveillance specimen program and providebetter Technical Specification information.

The Reactor Vessel Surveillance Program (RVSP) for Davis-Besse Unit 1 isdesigned to be in compliance with Appendix H to 10 CFR Part 50 and isgenerally described in BAW-10100A. The recer.t change in design andinstallation of surveillance specimen holder tubes in the Davis-BesseNo. I reactor vessel, and results of recent analyses of surveillancesamples from several B&W reactors, has resulted in some modifications to

the RVSP for the Davis-Besse Unit 1 from that described in BAW-10100A.The attached revisions to Davis-desse Unit 1 FSAR Sections 4.3.2.10 and5.4.7 reflect these changes and will be incorporated in the FSAR.

The redesigned and installed s'arvei.11.ance specimen holder tubes aredescribcd in Supplement 1 to BAW-10051. There are six of these surveil-lance specimen holder tubes installed with three being used for theDavis-Besse Unit 1 RVSP. The neutron flux lead factor for the specimencapsules in the new capsule holders is different than the lead factorstated in BAW-10100A and are shown in the attached Table I.

The calculated fluence values for exposure time have been updated usinga somewhat different analytical model combined with analytical predictionsof the effect of refueled core configurations on relative power distribu-tion. This analytical model has been verified and refined by comparisonwith surveill.nce capsule specimen analyses recently removed from severalB&W 177FA reactors. A fiaure showing this calculated fluence is attachedand should be used as Bases Figure 4-1 of the Technical Specification,replacing Figure 5-24 (Revision 21) of the FSAR.

E 74

THE TOLEGO EDISON CCMPANY EDISON PLAZA 300 MADISCN AVENUE TOLEDO, OHIO 43652

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The surveillance specimen withdrawal schedule has been revised andexpressed in removal intervals based on expected fuel cycles. Thiswithdrawal schedule includes prevision for obtaining data on materialproperties at the reactor vessel inner surface as well as the 1/4Tlocation. The inner wall data may be required to perform analyses ofthe reactor vassel for potential emergency and faulted ccnditions. Thisrevised withdrawal schedule is attached and should be use.' as Table 4.4-5 of the Technical Specifications, replacing Table 5-25 ('.avision 21) ofthe FSAR.

A revised bases for irradiation surveillance specimens and withdrawalfrequencies for inclusion in the Technical Specifications, first paragraphof page B 3/4 4-12 to correspond with the above is attached. The fifthand sixth capsules are standby capsules to provide for additional analysesor evaluation of in-place annealing, if required.

Very truly yours,

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Enclosures:Davis-Besse Unit 1 FSAR Section 4.3.2.10Davis-Besse Unit 1 FSAR Section 5.4.7Table 1 Lead Factors for Fast Flux (E > 1 Mev)Davis-Besse Unit i Technical Specification Bases Figure 4-1Davis-Besse Unit 1 Technical Specification Table 4.4-5Davis-Besse Unit i Technical Specification Reactor CoolantSystem Bases Pages B 3/4 4-lla and B 3'4 4-11b

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4.3.2.10 Neutron Flux Distribution

Energy dependent neutron fluxes were determined by a discrete ordinatecolution of the Boltzmann transport equation. Specifically, ANISN, aone-dimensional code, and DOT, a two-dimensional code, were used. Inboth codes, the system is modeled radially from the core out to the airgap outside the pressure vessel. The model includes the core with atime averaged radial power distribution, core liner, barrel, thermalshield, pressure vessel;, and water regions. Inclusion of the internalcompo ants is necessary to account for the distortions of the requiredenergy spectrum by attenuation in these components. The ANISN code usesthe CASK 22-group neutron cross section set with an S rder of angular

6quadrature and a P expansion of the scattering matrix. The problem is

3run along a radius across the core flats. Azimuthal variations are,obtained with a DOT r-theta calculation that models one-eighth of aplan-view of the core (at the core midplane) and includes a pin by pin,plant specific time averaged power distribution. The DOT calculation

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uses S6 qua rature and a Pg cross section set derived from CASK.

! Fluxes calculated with this DOT model must be adjusted to account forlack of P cross section detail in calculations of anistotropic scattering,a perturbation caused by the pressence of the capsule (where applicable)and the axial power distribution. The first two items are both energyand radial-location dependent, whereas the latter is axial locationdependent. A P /P correction factor is obtained by comparing two ANISN3 g1-D model calculations in which only the order to scattering was varied.-The capsule perturbation factor is obtained from a comparison of two DOTx-y model calculations, one with a capsule explicitly modeled -SS304

; cladding, Al filler region, and carbon steel specimens--and the otherwith water in those regions. The effect of axial power distribution hasbeen determined from a typical 177 FA plant burntp calculation as afunction of axial location for the outer' rows of fuel assemblies. Thenet result from these parameter studies is a flux adjustment factor Kwhich is applicable to calculated data from the D.I model. ~

Plant specific analytical results based on the calculation techniquedescribed above were converted to activities and then compared withdosimeter measurements from surveillance capsules from 5 similar reactors.All these comparisons are within + 15% of the dosimeter measurement.,

; The generic fluence predictions used for the Davis-Besse Unit I reactor- vessel surveillance program are presented in the technical specification

bases 2. the reactor vessel surveillance program (Bases Figure 4-1).These precictions included an estimated power distribution for an equil-ibrium-fue!. cycle.

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*5.4.7 MATERIAL SURVEILLANCE PROGR g

The Davis-Besse i reactor vessel material surveillance program.(RVSP) meets the requirements of Appendix H tc 10 CFR Part 50.The materials selected for surveillance monitortag are thoseidentified as being the materials wnich will first control theoperating limitations of the reactor vessel during its servicelife. The RVSP materials are listed in Table 5-24. The selec-tion procedure for these materials is described in B&W TopicalReport BAW-10100A.

There is a total of six (6) specimen capsules provided in theDavis-Besse 1 surveillance program. The type and number ofspecimens is as described for the modified program in B&W Topi-cal Report BAW-10100A. Five of tne specimen capsules areirradiated beginning with initial operation, while the remain-ing capsule is scheduled for insertion in the surveillancespecimen holder tube (SSHT) following the second Davis-Besse 1surveillance capsule withdrawal. The detailed withdrawalschedule is provided in the technical specifications.

The method of neutron fluence (E > 1 MZV) calculation used todetermine the surveillance specimen withdrawal schedule andcurrent predictions of reactor vessel irradiation is describedin Section 4.3.2.10. As discussed in Section 4.2.2.2, Subsection6, a total of six S'3HT's are installed. These SSHT locationshave two lead factors depending on their azimuthal location with

respect to geactor vessel axis (see Figure 4-3C) which is either11 or 26.5 . The two 26.5 locations were chosen as being inthe lowest available azimuthal flux location to minimize theflux lead factor available for specimen capsules. The originalflux lead factor calculations using the methods and model de-scribed in B&W Topical Report BAW-10100A showed the flux leadfactor to be 2.3 from the specimen capsule to the location ofhighest azimuchal flux on the reactor vessel inner wall forthe 26.5 location and 4.1 for the 11 location. The updatedfluence model used for the surveillance specimen withdrawalschedule shows the lead factors for these SSHT locations to be3.9 and 5.4, respectively, from the specimen capsule to thereactor vessel inner wall location of highest flux. Thisdifference from the original calculations is attributed to thecalculational model changes, particularly in the inclusion ofthe capsule purturbation effect, the use of realistic thermalconditions and the use of an updated azimuthal flux variation.The actual inservice lead fact r for the 26.5 SSHT locationwill vary depending on the actual azimuthal peaks for .heparticular core cycle design. Plant specific fluence analysis,accounting for the actual core design, will be performed whenthe Davis-Besse 1 surveillance capsules are tested.

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TABLE 1,

Davis-Besse Nuclear Pcwer Station Unit 1Lesd Factors for Fast Flux (E > 1 MEV)

BAW-10100ABAW -10100A Model & Capsule

Holder Tubes XW & ZY (26.5 ) Model Perturbation Effects

Capsule to RV Surface 2.3 3.9Capsule to 1/4 T 4.1 6.9

Holder Tubes WZ, YZ, YX, WX (11 )Capsule to RV Surface 4.1 5.4Capsule to 1/4 T 7.2 9.6

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Reactor Vessel Material Irradiation Surveillance Schedule ( }

F Capsule Adder AccumulatedNeutron{luence( Removal )'.

& Capsule Location (E) IMev. n/em ) Interval *T 18 (3)E TEl-F ZY (26.5 ) 3.5 x 10 End of 1st. Cycle: 18

TEl-B ZY (26.5 ) 8.0 x 10 End of 3rd. Cycle*c:

19h TEl-A WX (11 ) 1.3 x 10 End of 5th Cycle

ITEl-C YX (11 ) 2.34 x 10 End of 9th Cycle

ITEl-1) XW (26.5 ) 2.34 x 10 End of 12th Cycle

gg (5)TEl-E(5) ZY (26.5 ) 2.34 x 10

(1) The schedule may be modified, if necessary, after the evaluation of each capsule.

(2) They are the "best estimate" fluence values as defined in Figure 4.1 of BASES 3/4.4.9 andcorrespond to the fluence values of the withdrawal schedule described in the same BASES.

# (3) The neutron fluence expected to be accumulated by the capsule at the end of the first fuel cycle.i' The assumed EFPD for the first cycle are 450 days. This fluence is expected to cause a shift5 of 50 F in the surveillance weld metal.

(4) These are the refueling outages more closely approaching the withdrawal schedule defined in , iBASES 3/4.4.9.

(5) This capsule to be inserted into the reactor vessel surveillance holder at end of 3rd cycle.

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Bases Figure 4-1 Fast Neutron Fluence (E > I Mev) as a Functionof Full Power Service Life

' DAVIS -BESSE, UNIT I B 3/4 4-7

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REACTOR COOLANT SYSTEM BASES.

The number of reactor vessel irradiation surveillance specimens andthe frequencies for removing and testing these specimens are providedin Table 4.4-5. The withdrawal schedule is based on four considerations:(a) uncover possible technical anomalies as early in life as they can bedetected (end of first fuel cycle), (b) define the material propertiesneeded to perform the anal s1= required by Appendix G to 10 CFR 50, (c)Jreserve two capsules for evaluation of the effectiveness of thermal annealingin the event the inplace annealrag becomes necessary, (d) provide materialproperty data corresponding to the reactor vessel line surface conditionsat the end of service. With these considerations, the withdrawal schedule'sgeneral guidelines are:

First Capsule At the time when the highest predicted RTNDT

shift of all the capsule =nterials is approxi-mately 50 F .

Second Capsule At the time when the capsule's accumulated neutronfluence (E > 1 Mev) corresponds to an intermediatevalue between tho:e of the first and third capsules.

Third Capsule At the time when the capsule's accumulated neutronfluence (E > 1 Mev) correspends to that at 1/4Treactor vessel wall location at approximately theend of vessel's design service life.

Fourth Capsule At the time when the capsule's accumulated neutronfluence (E > 1 Mev) corresponds to that of thereactor vessel inner wall location at approximatelythe end of vessel's design service life.

Fifth and Standby.Sixth Capsule

The withdrawal schedule is specified to assure compliance with therequirements of Appendix H to 10 CFR 50. The actual withdrawal intervalshave been modified to meet the criteria described above. The withdrawalschedule will be reviewed for the need for revision following testing of Ieach specimen capsule.

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