ML19317F770
| ML19317F770 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 03/02/1977 |
| From: | Roe L TOLEDO EDISON CO. |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8001290661 | |
| Download: ML19317F770 (9) | |
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u.s. NUCLEAR REGULATCRY COMMISSl!N DOCKE T NUM T E Q NOCact.u 195 j
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NRC DISTRI,8UTiON PcR PART 50 DOCKET MATERIAL TO: Mr J Fht FROM: Toldeo Edison Co
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one signed DESC RIPT'ON ENCLOSU RE Ler trans the following:
Addl info re questions raised about the.reactc :
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TOLEDO C"
9 EDISON Docket No.50-34t, i
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LOWELL E. ROE p~
March 2, 1977 v'c' **d'a' vJ 1 Facihtees Deve'ooment (4191 259-5242
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Director of Nuclear Reactor Regulation A
Attn:
Mr. John F. Stolz, Chief y
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- b Light Water Reactors Branch No. 1 N g T rear Division of Project Management
.a U.S. Nuclear Regulatory Commission m.
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Washington, D.C.
20555
Dear Mr. Stolz:
During a recent discussion with members of your staff, several questions were raised about the reactor surveillance specimen program for the Davis-Besse Unit 1 and the specimen withdrawal schedule shown on Table 4.4-5 of the Draft Technical Specifications. This letter should answer these questions, clarify the surveillance specimen program and provide better Technical Specification information.
The Reactor Vessel Surveillance Program (RVSP) for Davis-Besse Unit 1 is designed to be in compliance with Appendix H to 10 CFR Part 50 and is generally described in BAW-10100A. The recer.t change in design and installation of surveillance specimen holder tubes in the Davis-Besse No. I reactor vessel, and results of recent analyses of surveillance samples 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 and 5.4.7 reflect these changes and will be incorporated in the FSAR.
The redesigned and installed s'arvei.11.ance specimen holder tubes are describcd in Supplement 1 to BAW-10051. There are six of these surveil-lance specimen holder tubes installed with three being used for the Davis-Besse Unit 1 RVSP. The neutron flux lead factor for the specimen capsules in the new capsule holders is different than the lead factor stated in BAW-10100A and are shown in the attached Table I.
The calculated fluence values for exposure time have been updated using a somewhat different analytical model combined with analytical predictions of the effect of refueled core configurations on relative power distribu-tion.
This analytical model has been verified and refined by comparison with surveill.nce capsule specimen analyses recently removed from several B&W 177FA reactors. A fiaure showing this calculated fluence is attached and 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
t-O The surveillance specimen withdrawal schedule has been revised and expressed in removal intervals based on expected fuel cycles. This withdrawal schedule includes prevision for obtaining data on material properties at the reactor vessel inner surface as well as the 1/4T location. The inner wall data may be required to perform analyses of the reactor vassel for potential emergency and faulted ccnditions. This revised withdrawal schedule is attached and should be use.' as Table 4.4-5 of the Technical Specifications, replacing Table 5-25 ('.avision 21) of the FSAR.
A revised bases for irradiation surveillance specimens and withdrawal frequencies for inclusion in the Technical Specifications, first paragraph of page B 3/4 4-12 to correspond with the above is attached. The fifth and sixth capsules are standby capsules to provide for additional analyses or evaluation of in-place annealing, if required.
Very truly yours, d
Enclosures:
Davis-Besse Unit 1 FSAR Section 4.3.2.10 Davis-Besse Unit 1 FSAR Section 5.4.7 Table 1 Lead Factors for Fast Flux (E > 1 Mev)
Davis-Besse Unit i Technical Specification Bases Figure 4-1 Davis-Besse Unit 1 Technical Specification Table 4.4-5 Davis-Besse Unit i Technical Specification Reactor Coolant System Bases Pages B 3/4 4-lla and B 3'4 4-11b db c/3-4 i
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4.3.2.10 Neutron Flux Distribution Energy dependent neutron fluxes were determined by a discrete ordinate colution of the Boltzmann transport equation.
Specifically, ANISN, a one-dimensional code, and DOT, a two-dimensional code, were used.
In both codes, the system is modeled radially from the core out to the air gap outside the pressure vessel. The model includes the core with a time averaged radial power distribution, core liner, barrel, thermal shield, pressure vessel;, and water regions.
Inclusion of the internal compo ants is necessary to account for the distortions of the required energy spectrum by attenuation in these components.
The ANISN code uses the CASK 22-group neutron cross section set with an S rder of angular 6
quadrature and a P expansion of the scattering matrix. The problem is 3
run along a radius across the core flats. Azimuthal variations are obtained with a DOT r-theta calculation that models one-eighth of a plan-view of the core (at the core midplane) and includes a pin by pin, plant specific time averaged power distribution. The DOT calculation uses S6 qua rature and a Pg cross section set derived from CASK.
Fluxes calculated with this DOT model must be adjusted to account for lack 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 energy and radial-location dependent, whereas the latter is axial location dependent. A P /P correction factor is obtained by comparing two ANISN 3 g 1-D model calculations in which only the order to scattering was varied.
The capsule perturbation factor is obtained from a comparison of two DOT x-y model calculations, one with a capsule explicitly modeled -SS304 cladding, Al filler region, and carbon steel specimens--and the other with water in those regions. The effect of axial power distribution has been determined from a typical 177 FA plant burntp calculation as a function of axial location for the outer' rows of fuel assemblies. The net result from these parameter studies is a flux adjustment factor K which is applicable to calculated data from the D.I model. ~
Plant specific analytical results based on the calculation technique described above were converted to activities and then compared with dosimeter 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.
l DB-1
r 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 those identified as being the materials wnich will first control the operating limitations of the reactor vessel during its service life. The RVSP materials are listed in Table 5-24.
The selec-tion procedure for these materials is described in B&W Topical Report BAW-10100A.
There is a total of six (6) specimen capsules provided in the Davis-Besse 1 surveillance program. The type and number of specimens is as described for the modified program in B&W Topi-cal Report BAW-10100A. Five of tne specimen capsules are irradiated beginning with initial operation, while the remain-ing capsule is scheduled for insertion in the surveillance specimen holder tube (SSHT) following the second Davis-Besse 1 surveillance capsule withdrawal. The detailed withdrawal schedule is provided in the technical specifications.
The method of neutron fluence (E > 1 MZV) calculation used to determine the surveillance specimen withdrawal schedule and current predictions of reactor vessel irradiation is described in Section 4.3.2.10.
As discussed in Section 4.2.2.2, Subsection 6, a total of six S'3HT's are installed. These SSHT locations have two lead factors depending on their azimuthal location with respect to geactor vessel axis (see Figure 4-3C) which is either 11 or 26.5. The two 26.5 locations were chosen as being in the lowest available azimuthal flux location to minimize the flux lead factor available for specimen capsules. The original flux lead factor calculations using the methods and model de-scribed in B&W Topical Report BAW-10100A showed the flux lead factor to be 2.3 from the specimen capsule to the location of highest azimuchal flux on the reactor vessel inner wall for the 26.5 location and 4.1 for the 11 location. The updated fluence model used for the surveillance specimen withdrawal schedule shows the lead factors for these SSHT locations to be 3.9 and 5.4, respectively, from the specimen capsule to the reactor vessel inner wall location of highest flux. This difference from the original calculations is attributed to the calculational model changes, particularly in the inclusion of the capsule purturbation effect, the use of realistic thermal conditions and the use of an updated azimuthal flux variation.
The actual inservice lead fact r for the 26.5 SSHT location will vary depending on the actual azimuthal peaks for.he particular core cycle design. Plant specific fluence analysis, accounting for the actual core design, will be performed when the Davis-Besse 1 surveillance capsules are tested.
DB-1
C TABLE 1 Davis-Besse Nuclear Pcwer Station Unit 1 Lesd Factors for Fast Flux (E > 1 MEV)
BAW-10100A BAW -10100A Model & Capsule Holder Tubes XW & ZY (26.5 )
Model Perturbation Effects Capsule to RV Surface 2.3 3.9 Capsule to 1/4 T 4.1 6.9 Holder Tubes WZ, YZ, YX, WX (11 )
Capsule to RV Surface 4.1 5.4 Capsule to 1/4 T 7.2 9.6 a
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i TAB',E 4. 4-5 Reactor Vessel Material Irradiation Surveillance Schedule ( }
F Capsule Adder AccumulatedNeutron{luence(
Removal
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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:
19 h
TEl-A WX (11 )
1.3 x 10 End of 5th Cycle I
TEl-C YX (11 )
2.34 x 10 End of 9th Cycle I
TEl-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 and correspond 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 shift 5
of 50 F in the surveillance weld metal.
(4) These are the refueling outages more closely approaching the withdrawal schedule defined in
, i BASES 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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A. SURVEILANCE CAPSULE (11*)
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8 16 24 32 EXPOSURE TIME, EFFECTIVE FULL FCWER YEARS Bases Figure 4-1 Fast Neutron Fluence (E > I Mev) as a Function of Full Power Service Life
' DAVIS -BESSE, UNIT I B 3/4 4-7
2 REACTOR COOLANT SYSTEM BASES The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 4.4-5.
The withdrawal schedule is based on four considerations:
(a) uncover possible technical anomalies as early in life as they can be detected (end of first fuel cycle), (b) define the material properties needed to perform the anal s1= required by Appendix G to 10 CFR 50, (c)
J reserve two capsules for evaluation of the effectiveness of thermal annealing in the event the inplace annealrag becomes necessary, (d) provide material property data corresponding to the reactor vessel line surface conditions at the end of service. With these considerations, the withdrawal schedule's general 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 neutron fluence (E > 1 Mev) corresponds to an intermediate value between tho:e of the first and third capsules.
Third Capsule At the time when the capsule's accumulated neutron fluence (E > 1 Mev) correspends to that at 1/4T reactor vessel wall location at approximately the end of vessel's design service life.
Fourth Capsule At the time when the capsule's accumulated neutron fluence (E > 1 Mev) corresponds to that of the reactor vessel inner wall location at approximately the end of vessel's design service life.
Fifth and Standby.
Sixth Capsule The withdrawal schedule is specified to assure compliance with the requirements of Appendix H to 10 CFR 50.
The actual withdrawal intervals have been modified to meet the criteria described above. The withdrawal schedule will be reviewed for the need for revision following testing of each specimen capsule.
l Davis-Besse Unit 1 B 3/4 4-lla
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