ML20211N939

Forwards Pages 30-39 & 97 of Final Rept on Calvert Cliffs Unit 1 Nuclear Plant Reactor Pressure Vessel Surveillance Program:Capsule 263, Dtd 801215,detailing Methodology Used to Determine Vessel Wall Fluence

ML20211N939
Person / Time
Site:	Calvert Cliffs
Issue date:	12/09/1986
From:	Tiernan J BALTIMORE GAS & ELECTRIC CO.
To:	Thadani A Office of Nuclear Reactor Regulation
References
REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR NUDOCS 8612180348
Download: ML20211N939 (14)

Text

{{#Wiki_filter:g, l w. l BALTIMORE GAS AND ELECTRIC p CHARLES CENTER

• R O. BOX 1475 BALTIMORE, MARYLAND 21203 Joseph A.TleRNAN Vice PRaslotNT NUCLEAR ENEROY December 9,1986 U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, D. C. 20555 ATTENTION:

Mr. Ashok C. Thadani, Director PWR Project Directorate #8 Division of PWR Licensing-B

SUBJECT:

Calvert Cliffs Nuclear Power Plant Unit Nos.1 & 2; Docket Nos. 50-317 & 50-313 Pressurized Thermal Shock Rule

REFERENCES:

(a) Letter from Mr. J. A. Tiernan (BG&E), to Mr. A. C. Thadani (NRC), dated August 29,1986 (b) Letter from Mr. S. A. McNeil (NRC), to Mr. 3. A. Tiernan (BG&E), dated August 21,1986 Gentlemen: This letter clarifies information submitted by Reference (a) in response to Reference (b). This clarification was initiated by an October 31,1986 conference call between your S.A. McNeil and L. Lois, of the NRC, and our staff; wherein, specific information was requested relating to the methodology used in determining the End of Life (EOL) fluence. The calglation submijted with Reference (a) identified an estimated fluence value of 4.34x10 neutron /cm at the vessel wall. Reference was made to a Battelle, Columbus Laboratories report for Unit I surveillance capsule 263. This report was previously submitted in February 1981. Copies of selected pages (pp. 30-39, and 97) from the Battelle report are enclosed which detail the methodology used to determine the vessel wall fluence. The DOT 3.5 computer code was used in this determination as tabulated in Table 4D of the enclosure (p31). The balance of the calculation submitted by Reference (a) establishes tb: derivation of total EOL fluence. \\ 8612180348 861209 OD PDR ADOCK 05000317 P PDR a

O' Mr. Ashok C. Thadani December 9,1986 Page 2 Additionally, we indicated that the next surveillance capsule to be sampled (approximately 1990) will also be benchmarked to the DOT code. We are confident that the level of conservatism presented in the submitted calculation for capsule 263 will bound the next surveillance capsule. We trust this provides the requested clarification. Should you have any further questions, please do not hesitate to contact us. Very truly yours, e 3AT/AM/ dim Enclosure cc: D. A. Brune, Esquire

3. E. Silberg, Esquire S. A. McNeil, NRC T. Foley, NRC

a-(, .., 4 ' ? ' ~ FINAL REPORT on CALVERT CLIFFS UNIT NO. 1 NUCLEAR PLANT REACTOR PRESSURE VESSEL SURVEILLANCE PROGRAM: CAPSULE 263 to BALTIM0RE GAS AND ELECTRIC COMPANY December 15, 1980 by J. S. Perrin, E. O. Fromm, D. R. Farmelo, R. S. Denning, and R. G. Jung + BATTELLE Columbus Laboratories 501 King Avenue Columbus, Ohio 43201 Battelle is not engaged in researen fer advertising, sales promotien, or publicity purposes, and this report may not be reproduced in full or in part for such purposes.

C ] 30 isthen4.74x1M9 n/cm. This compares reasonably well with the Technical I9 2 Specification value of 3.44 x 10 n/cm. The fluence at the 1/.1 T position is 0.574 that at the inner wall and the fluence at the 3/4 T position is 0.132 that at the inner wall. These values are sumarized in Table 4D. Analytical Methods The determination of the neutron flux at the capsule, and subse-quently in the pressure vessel wall, requires the completion of three proce-dures. First, the disintegration rate of the product isotope per unit mass of the flux monitor must be determined. This has been discussed earlier under experimental procedures. Second, in order to find a spectrum-averaged reaction cross section at the capsule location, the neutron energy spectrum must be cal-culated for that identical location. Third, the neutron flux at the capsule must be found by calculations involving the counting rate data, the spectrum averaged cross sections, and the operating history of the reactor. The energy and spatial distribution of neutron flux in the reactor were calculated using the DOT 3.5 computer program (21) 00T solves the Boltzman transport equation in two-dimensional gecmetry using the method of discrete ordinates. Balance equations are solved for the density of particles moving along discrete directions in each cell of a two-dimensional spatial mesh. Anistotropic scattering is treated using a Legendre expansion of arbitrcry order. The two dimensional geometry that was used to model the reactor is shown in Figure 11. As seen there are 19 circumferential meshes and 52 radial meshes. Capsule 263 includes circumferential meshes 13,14, and 15 and radial meshes 36, 37, and 38. Third order scattering was used (P ) and 48 angular 3 directions of neutron travel (24 positive and 24 negative) were used (58 quad-rature). Neutron energies were divided into 22 groups with energies from 14.9 MeV to 0.01 eV. The 22 group structure is that of the RSIC Data Library DLC/ CASK (22), and neutron absorption, scattering, and fission cross sections used are those supplied by this library. The core shroud and the core support barrel are type 304 stainless steel. The capsule is also modeled as a solid piece of 304 stainless steel. The reactor pressure vessel wall is A5333 steel. The reactor core was mocked ,g 'b

f';. - ~ 31 TABLE 4D.

SUMMARY

OF FAST NEUTP.ON FLUX 'ND FLUENCES AT VARIOUS LOCATIONS Fast Fluence (E > 1.0 MeV)(a) Predicted From Maximum (a) After 2.94 After 32 Technical 2 Location Flux, n/cm /sec EFPY EFPY Specifications W 263 Surveillance 10 18 19 Capsule 6.7 x 10 6.2 x 10 6.8 x 10 Pressure Vessel 10 18 I9 I9 Wall Inner Diameter 4.7 x 10 4.4 x 10 4.7 x 10 3.4 x 10 1/4 Thickne:s of 10 18 19 Vessel Wall 2.6 x 10 2.4 x 10 2.6 x 10 3/4 Thickness of 9 17 18 Vessel Wall 6.1 x 10 5.6 x 10 6.1 x 10 4 (a) Note from test that the largest fast flux (E>l MeV) at the bottom of the ~ capsule was used for all estimates. The actual measured center flux was 7.2% lower than the bottom value. See Figure 6A. 6

• m e I

A

e-o 32 15, 34. i3 'II 27t ME ES 1 4,15 70 9 220.50 6 U 1. El ,o J 3 r =w N, 1!L _; p g ,,lg ,O

1. 4 273 T r1J 2/

o. 93

.r ( -l I y '#g, $ 355 g3 wumme mum s 140 ummxN

• e u 135.05 114 I

e i f s l l sf ,%,#+ q, -{ 100 \\ \\, / p l -Q %*#e l r4 93.50 g 3 g NUMBERS IN BOX REPRESENTS THE RELATIVE POWER IN CORE I I 'k ],/ f 41.53 1

o h,

I I I I I I O 20 40 60 80 100 120 140 160 180 DISTANCE FROM CORE CENTER, em ,(' FIGURE 11. CALVERT CLIFFS GEOMETRY USED IN DOT RUN. g. tL

33 up as homogenized fuel and water having the densities found in the operating reactor. The water in the core region has a density consistent with the average coolant temperature in the core (570* F) at the operating pressure of 2250 psia. The water in the downcomer region outside the core barrel had a I density consistent with the inlet coolant temperature (543 F) and operating pressure of 2250 psia. All the water carried boron in solution at the concen-tration specified for full power operation, i.e., 725 grams of beton per 6 ~ 10 grams of water. Finally, the fuel was a source of neutrons having a U-235 fission energy spectrum. The relative powe.r in the assemblies nearest the capsule, during the interval the capsule was in the reactor, is shewn in Figure 11(23). The neutron spectrum at the capsule center, as calculated by DOT, is shown in Figure 12. Also shown for comparison is the fission spectrum. Both spectra have been normalized to contain one neutron above 1.0 MeV. As can be seen, the capsule spectrum is very much harder than the fission spec-trum. This is caused by travel through water. The DOT calculated values of " spectrum-averaged" cross-section, c, differ from the fission spectrum-R averaged cross sections by as much as a factor of 2. Based upon the fluxes calculated by D0T at r mesh 37 and e mesh 13, 14, and 15 (the three radial centered meshes used to represent the capsule and the region in which the flux monitors were placed), effective cross-sections og (E>0.1 MeV) and eR (E>l.0 MeV) defined as 1 e I e (E)e(E)dE R(E>Ec)= c i I 4(.E)dE Ec were calculated for iron, nickel, copper, titanium, and uranium in each of the three e meshes. The results are shown'in Table 5 for cR (E > 1.0 MeV) which is of most interest. These values differ from center to edge of the capsule because the fast spectrum is being depressed by the capsule presence as was shown earlier and because the neutron spectrum is being modified by the stainless steel of the capsule. Using the results of Table 5 and the geometry shown in Figure 98, the cross-section appropriate to each of the monitors can be interpolated. These values and other nuclear constants needed in the third step of the flux-finding procedure are given in Table 6. ~

r.- t y*. -e

• 34 1-m.

i i i i i i 3 i 3-FISSION SPECTRUM - - - - CAPSULE SPECT. RUM I i S j' l I g, --i h 9 4 ,? s I 3 o= v - p t 5 5 .i.3 i .s x g k t -y g 1 1 m m l i I i i O l r z y, Lu I m. l' If1 1 m l i I e 1 e 1 i I e I e B 2 4 6 8 10 12 14 NEUTRON ENERGY (MEV) FIGURE 12. COMPARISON OF DOT SPECTRUM AT CAPSULE WITH FISSION SPECTRUM. 5. h . 'Y.;. t ,r_--

Y 4-35 0 TABL'E 5. CROSS-SECTIONS FOR THE FLUX MONITORS (E>l.0 MeV) IN NINE CAPSULE MESHES e Core 4

• 0.

.00148 .00142 .00148' 0 Copper Cross-Sections (barns) p . [, Corep ?* .170 .165 .170

ll)

Nickel Cross-Sections (barns) 3 l" Corep li h' .134 .129 .134 Y Iron Cross-Sections (barns) t' Core h [' I 4, .0225 .0216 .0225 Titanium Cross-Sections (barns) .1 i Core A 4 j'- I.423 .418 .423 j id C I ! >l Urenium Cross-Sections (barns) \\ h H'

t,. p-36 e e e. C 4 e TABLE 6. CONSTANTS USED IN DOSIMETRY CALCULATIONS .e -40 Cross-Sections I' Isotopic Threshold (Barns)

Target, Abundance,

Energy, Product E>1.0 MeV

• o i

Reaction (MeV) Half-Life E20.1 Mey

t*

Fe 54(n.p)Mn 54 99.865Fe 5.82 1.5 314d .131 .0688 3 Cu63(n,a)Co60 99.999Cu -69.17 5.0 5.25y .00147 .00078 7 Ni S8(n.p)Co S8 99.951Ni 67.77 1.0 71.3d .166 .087 7 Ti46(n.p)Sc46 99.793Ti 7.93 2.5 83.8d .0216 .0113 4i. U238(n,f)Ce144 100.0(a)g gg,g7(b) 0.8 284.1d .420 i .221 ,7 3 t. l o. (a) The target was considered 100.0% uranium since the oxide powder was dissolved and analyzed as mg U/ml solution. (b) The uranium monitor was analyzed to contain 0.03% U-235 (300 ppm). r >1$ t t

C m-o O 37 In the third step the full power flux at the capsule location is determined from the radicactivity induced in the monitor foils, the spectrum-averaged cross-sections calculated for the monitor elements, and the power history of the reactor during capsule exposure. The fluence at the capsule is then calculated from the integrated power output of the reactor during the exposure interval. The activity A induced into an element irradiated for a time t in a constant neutron flux is given by j -At A = N [/,e (E)e(E)dE] (1-e I ). where c(E) = the differential cross section for the activation reaction

(E) = the neutron differential flux N = the atom density of the target nuclei (atoms /g)

A = the decay constant of the product atom (sec-I). If the sample is permitted to decay for a time t between exposure and g s counting, then the activity when counted is 6 - Att ),-At = A = N [/ e (E);(E)dE] (1-e w A "spectrun-averaged cross section" may be defined as /,e (E):(E)dE /"4(E)dE and the integrated flux as u. t c = /oc (E)dE Then 4 i. / e (E)e(E)dE ./,e (E)c(E)dE = - /[: (E)dE = c: / ; (E)dE L. l' ' J,

s. y r 38 50 that the activity A may be written as - At -At A = N c(1-e $e The flux is then computed from the measured activity as ^ += - I) e-u

• N (1-e If it is desired to find the flux for neutrons with energies above a given energy level, E, the cross section corresponding to this c

energy level is defined as /": (E)o(E)dE c(E>E ) = c =f c(F)dE c where c(E>E ) " # c E c The n /~s (E)c(E)dE /~c(E);dE = /E ; (E)dE /* 9(E)dE c c = c (E>E ) 4 (E>E ) c c and the activity A may be written as A = N:(E>E ) o(E>E )(I-* )' c c In case that the neutron flux is not constant, the dosimeter activity at the time of removal from the reactor is A = N (E>E )t(E>E )C c c s O>=

t. 39 where J -AT C= I f (1-e d )e-A(T-t. ) J j =1 j J = number of time intervals of constant flux f) = the fractional power level during the time interval j T) = the time length of interval j t) = the elapsed time from beginning of irradiation to end of interval j 'T = the time from begfr>ning of irradiation to counting. Then o(E>E ) " NME E )C c g This is the equation used to find fluxes based on surveillance dosimeter activations. The time intervals are taken as one month each and average power during the month is used for values of f. Calculations of the flux and fluence were made with the DECAY code. The reactor power history was supplied in a private communication (24) The fluences can be changed to fluxes by dividing by the number of seconds 7 in 2.94 years which is 9.28 x 10. These results are summarized in Table 40 as discussed earlier. Disolacements per Atom (dca) Analysis One measure of neutron radiation damage is the number of times, on the average, that an atom has been displaced during an irradiation. The number of dpa associated with an irradiation depends on the amount of energy deposited in the material by the neutrons; hence depends on the neutron spec-trum and the neutron fluence. If the spectrum is constant over the duration of the irradiation, then: dpa-fe(Z}*IEIdE"i1(d}i'i d 1 0 m eW:V'. 7-

y .,. r g ' 97 (12) "Detemining Neutrcn Flux, Fluence, and Spectra by Radioactivation Technioues", ASTM Designation =251-77, Annual Book of ASTM Standards, ( Part 45. 1 ( ~ (I3) "Ceter-ining Thermal Neutron Flux by Racicactive Techniques." ASTM Designaticn E252-77, Annual Book of ASTM Standards, Part 45. (14) " Determining Fast-Neutron Flux by Radioactivation cf Iron", ASTM Designation E253-77, Annual Book of ASTM Standards, Part 45. (15) " Determining Fast-Neutron Flux by Radioactivation of Nickel", ASTM Designation E264-77, Annual Book of ASTM Standards, Part a5. (16) " Measuring Fast-Neutron Flux Density by Radioactivation of Cooper, ASTM Designaion 523-76, Annual Book of ASTM Standards, Part 45. (17) " Measuring Fast-Neutron Flux by Radioactivation of Titanium", ASTM Designation E525-76, Annual Book of ASTM Standards, Part 45. (18) " Determining Fast Neutron Flux Density of Radioactivation of Uranium-238." ASTM Designatien E704-79, Annual Book of ASTM Standards, Part 45. (19) Perrin, J. S., Fromm, E. 0., and Lowry, L. M., " Remote Disassembly and Examination of Nuclear Pressure Vessel Surveillance Capsules", Proceedings of the 25th Conference on Remote Systems Technology, American Nuclear Society (1977). { (20) " Mechanical Testing of Steel Products", ASTM Designation A370-77, Annual Book of ASTM Standards, Part 10 (1979). pp 28-83. I (21) RSIC Computer Code Collection, DOT 3.5-Two Dimensional Discrete Ordinates Radiation Transport Code, Radiation Shielding Information Center, Oak Ridge National Laboratory, Oak Ridge, Tennessee. 1 (22) RSIC Data Library Collection, DLC-23/ CASK, 40 Group Coupled Neutron jv - and Ganma-Ray Cross Secticn Data, Raciation Shielding Information }, Center, Oak Ridge National Laboratory, Oak Ridge, Tennessee. (23) Private Communication, Lippold, W. J., to Farmelo, D. R., A'. gust 5, 1980. (24) Private Communication, Titland, L. E., to Fromm, E. O., July 17, 1950. l-(25) Characterizing Neutron Exposures in Ferritic Steels in Tems of Displacements per Atom (DPA), ASTM Designation E693-79. (26) " Rules for Construction of Nuclear Power Plant Components", ASME Boiler and Pressure Vessel Code, Section III, A,merican Society of Mechanical Enginee-s, (1974 Edition). _. _ _ _ _ _ _ -. _ _ _ _ _. _. _ _. _ _ _ _ _ _ _. _}}