ML20247G048
| ML20247G048 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 03/30/1989 |
| From: | Loflin L CAROLINA POWER & LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20247G053 | List: |
| References | |
| NLS-89-059, NLS-89-59, TAC-71112, TAC-71113, NUDOCS 8904040110 | |
| Download: ML20247G048 (18) | |
Text
{{#Wiki_filter:__. - i CASA Carolina Power & Light Company NAR 3 01989 SERIAL: NLS-89-059 10CFR50.90 88TSB04 United States Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-325 & 50-324/ LICENSE NOS. DPR-71 & DPR-62 SUPPLEMENT TO REQUEST FOR LICENSE AMENDMENT PRESSURE / TEMPERATURE LIMITS (TAC NOS. 71112 AND 71113) Gentlemen: On October 26, 1988, Carolina Power & Light Company submitted a request for license amendment concerning reactor coolant system pressure / temperature limits for the Brunswick Steam Electric Plant, Units 1 and 2. On February 13, 1989, the NRC requested additional information to support review of this request. The Company's response to the NRC's request is provided in Enclosure 1. Information concerning Charpy V shelf energies is also provided separately in Item 7 of. Please refer any questions regarding this submittal to Mr. Stephen D. Floyd at (919) 546-6901. Yours very truly, r Leona d Loflin Ma ger Nuclear Licensing Section BAB/bab (\\cor\\88tsb04s)
Enclosures:
1. Response to Request for Information 2. Reactor Cavity Neutron Dosimetry Program for Brunswick Unit 2 cc: Mr. Dayne H. Brown Mr. S. D. Ebneter Mr. W. H. Ruland I I Mr. E. G. Tourf',ny I \\ 411 Fayetteville Street
- P O Box 1551
- Raleigh N C. 27602
~ " " 8904040110 890330 PDR ADOCK 05000324 P PDC' 1
ENCLOSURE 1 BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 NRC DOCKETS 50-325 AND 50-324 OPERATING LICENSCS DPR-71 AND DPR-62 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REACTOR COOLANT SYSTEM PRESSURE / TEMPERATURE LIMITS RESPONSE TO REQUEST FOR INFORMATION 1. NRC QUESTION Provide detailed calculations of neutron fluence and reference nil-ductility transition temperature (RTNDT) used in pressure / temperature (P/T) limit curves as shown in Figures 3.4.6.1-1, 2, 3a, 3b. and 3c. A. The neutron fluence calculation should include a graph of neutron fluence versus effective full power years (EFPY) and a discussion of neutron transport analysis. The licensee should also submit the Westinghouse Report, " Reactor Cavity Neutron Dosimetry Program for Brunswick Unit 2," WCAP-10903, which was referenced in the October 26, 1988 submittal. B. The RT calculation for the vessel material that was used in the NDT P/T curves should include the following: reactor vessel beltline thickness chemistry factor a fluence factor a unirradiated RT = NDT increase in RT due to neutron irradiation NDT safety margin e adjusted RT a NDT CP&L RESPONSE A. The graph of fluence versus effective full power years is shown on Figure 1. A discussion of the neutron transport calculations as performed for BSEP-2 but equally applicable to BSEP-1 is provided as follows: Analytical Methodology In performing the fast neutron exposure evaluations for the BSEP-2 reactor pressure vessel, two distinct sets of neutron transport calculations were carried out. The first computation, a two-dimensional analysis in R, O geometry, was utilized to provide baseline data characteristic of the average exposure for BSEP-2 Cycle 6. This baseline two-dimensional calculation provided the relative neutron energy spectra at key locations within the pressure vessel and the reactor cavity as well as relative radial distributions of exposure parameters and dosimeter reaction rates through the vessel wall. 1 240CRS/che)
The neutron spectral information was necessary for the interpretation of the dosimetry withdr vn from the reactor cavity following the BSEP-2 Cycle 6 irradiation a. well as for the determination of important exposure parameter ratios, i.e., dPa/d (E > 1.0 MeV), within the pressure vessel geometry. The second set of analyses performed for the BSEP reactors consisted of a series of one-dimensional calculations designed to investigate the impact of several operational and geometric parameters on the interpretation of cavity neutron dosimetry and on the extrapolation of those measured results to the pressure vessel proper. In particular, the one-dimensional studies were used to investigate the effect of void fraction in the reactor core, of the amount of steel present in the jet pump region of the downcomer, and of the amount of icon present in the sacrificial shield. Clearly, each of these variables has e marked impact on the calculated neutron flux magnitude at the inner diameter and through the thickness of the pressure vessel shell. However, with cavity neutron dosimetry available to establish absolute magnitude, the calculations are required only to produce accurate relative neutron energy spectra and relative radial distributions. The intent of the one-dimensional studies was to confirm the insensitivity of this relative data to the key operating parameters noted above. The combination of the analyses described in this section and the measurement results discussed in Section 5.0 allowed an evaluation of the absolute exposure of the pressure vessel and of the exposure gradients within the vessel wall with a minimum uncertainty. Two-Dimensional Analysis A plan view of the BSEP reactor geometry at the core midplane is shown in "igure 2. Since the reactor exhibits 1/8th core symmetry, only a 0* - 45* sector la depicted. In addition to the core, reactor internals, pressure vessel, and sacrificial shield, the geometry also includes representations of the jet pumps located internal to the pressure vessel. Development of the geometric model shown in Figure 2 made use of nominal design dimensions throughout. The jet pumps were modeled as homogeneous zones characteristic of the pump geometry opposite the axial mid lane of t the reactor care. For the purposes of this baseline neutron transport calculation, the coolant void fraction associated with the reactor core was assumed to be 30 percent. The small aluminum capsules containing the multiple foil sensor sets that were positioned along the 0* and 45* axial traverses were not explicitly modeled in the two-dimensional transport analysis. These capsules were designed to minimize perturbations in the fast neutron flux and, thus, to provide free field data at the measurement locations. This assumption of an insignificant perturbation in the fast neutron field due to the presence of the aluminum holders is supported by the close agreement between iron gradient measurements and iron measurements from foils contained within the irradiation capsules. ) 2 (240CRS/che )
The two-dimensional calculation for the reactor model shown in Figure 2 was carried out in R,0 geometry using the DOT two-dimensional discrete ordinates code and the SAILOR cross-section library. The SAILOR library is a 47 group ENDFB-IV based data set produced specifically for light l water reactor applications. In the BSEP-2 analysis, anisotropic scattering was treated with a P3 expansion of the cross sections, and the rder.of angular quadrature. angular discretization was modeled with an S8 Plant-specific reactor core power distributions for use in the neutron transport analyses were provided by CP&L. For the BSEP-2 Cycle 6, R,9 analyses, the assembly wise power distribution representative of midcycle was employed. Relative assembly powers for this point in the BSEP-2 Cycle 6 burnup history are illustrated in Figure 3. In addition to the BSEP-2 Cycle 6 relative assembly power fractions, axial distributions of both core void fraction and relative power density were used. An examination of these data indicated that the use of an axial peaking factor of 1.20 was consistent with the assumption of a 30 percent core void fraction in the peripheral fuel assemblies and, further that this calculational model should be representative of an axial location near core midplane. In order to compute the absolute magnitude of neutron radiation levels within the pressure vessel geometry, a design basis core power level must be chosen. For the BSEP reactors, all data, both analytical and experimental, were referenced to a reactor core power level of 2436 MWt. One-Dimensional Analysis The one-dimensional discrete ordinates neutron transport calculations were N transport code. In performing the carried out using the ANISN S one-dimensional analyses, the reactor core was inodeled as an equivalent volume e iiudes ulch the remaining reactor and cavity components modeled j as cylindrical annuli as illustrated in Figure 2. The presence of jet pumps was approximated by including stainless steel as a homogeneous mixture in the downcomer region. All calculations were carried out in 47 neutron energy groups using the SAILOR library with a P3 scattering rder of angular quadrature. cross-section approximation and an S8 To complete the parameter evaluation of core void fraction, stainless steel in the jet pump region, and iron content in the sacrificial shield, the following analytical cases were run. Case ID Core Void % Downcomer SS % Iron in Shield % A 0.0 0.0 0.0 B 30.0 0.0 0.0 C 60.0 0.0 0.0 D 90.0 0.0 0.0 E 30.0 10.0 0.0 F 30.0 20.0 0.0 C 30.0 30.0 0.0 H 30.0 0.0 15.0 I 30.0 0.0 30.0 3 (240CRS/che)
) From an interpretive viewpoint, the key outputs from these one-dimensional computations are the neutron spectra in the reactor cavity that are used j in the derivation of the exposure parameter (E > 1.0 MeV) from the j measured reaction rates and the ratios of exposure parameters at the l vessel inner. radius to those at the cavity dosimetry locations that are ^ employed to relate cavity measurements to the exposure of the vessel itself. -Absolute comparisons of calculation and measurement were not an l intended goal of these one-dimensional studies. B. The requested data input for BSEP-1 are as follows: Beltline Thickness 5.28" minimum Chemistry Factor CF = Maximum Plate (Cu = 1.5 di e.54) CF = 106.7 Maximum Weld (Cu =.05 Ni <) CF = 68 f Surface Fluence = 17 2 At 8 EFPY f = 4.84 x 10 n/cm 17 n/cm2 At 10 EPPY f = 6.05 x 10 17 2 At 12 EFPY f = 7.27 x 10 n/cm 17 2 At 16 EFPY f = 9.68 x 10 n/cm RTNDT(I) Unirradiated RT = NDT Maximum Plate (201) RTNDT(i) 10"F = -56'F Maximum Weld RTNDT(i) = 2,a 2 Margin = 2 ci Maximum Plate (201) 2/0+172 = 34 2[172 + 282 = 66 Maximum Weld f x e.24t/4 1/4T Fluence g = At 8 EPPY f' =.0353 x 1019 "/cm2 > 1 MeV At 10 EFPY f" =.0404 x 1019 "/cm2 > 1 MeV At 12 EFPY f" =.0529 x 1019 "/cm2 > 1 MeV At 16 EFPY f =.0705 x 1019 "/:m2 > 1 MeV f, (.28 -0.1 logf ) Fluence Factor at 1/4T ff s = = .0353 ((.28 -0.1 log.0353) .2411 At 8 EFPY ff = = .0404 (.28 -0.1 log.0404) .2732 At 10 EFPY ff = = .0529 ( 28 -0.1 log.0529) .3017 At 12 EFPY ff = = 28 -0.1 log.0705) .3506 .0705 At 16 EFPY ff = = l 1 4 (240CRS /che >
l l MAXIMUM PLATE CALCULATION Increase in 1/4T RT CF X f f ART = = NDT NDT 25.7'F (106.7) (.2411) At 8 EFPY ART = = NDT 29.2*F (106.7) (.2732) At 10 EFPY ART = = NDT 32.2*F (106.7) (.3017) At 12 EFPY ART = = hDT 37.4*F At 16 EFPY ART (106.7) (.3506) = NDT Adjusted 1/4T RTNDT = RTNDT (i) + ARTNDT + Margin At 8 EFPY 10 + 25.7 + 34 = 70*F At 10 EFPY 10 + 29.2 + 34 = 73*F At 12 EFPY 10 + 32.2 + 34 = 76*F At 16 EFPY 10 + 37.4 + 34 = 81*F MAXIMUM WELD CALCULATION Increase in 1/4T RT CF X ff ART = = NDT NDT (68) (.2411) 16.4*F At 8 EFPY ART = = NDT M.6*F (68) (.2732) At 10 EFPY ART = = NDT 20.5'F (68) (.3017) At 12 EFPY ART = = NDT 23.8'F (68) (.3506) At 16 EFPY ART = = NDT Sigma Delta (Reduced) = o A 8.2*F At 8 EFPY a = 3 9.3*F At 10 EFPY 0 = 3 10.3*F At 12 EFPY = a 11.9'F At 16 EFPY o = g 2 g32 Margin 2 ot = At 8 EFPY 2 172 + 8.22 37.8'F = ? )172 + 9.32 38.8'F At 10 EFPY = 2/172 + 10.32= 39.8'F At 12 EFPY At 16 EFPY 2 172 + 11,92= 41.5*F RTNDT(i) + ARTNDT + Margin Adjusted 1/4 T RT = NDT -1.8*F At 8 EFPY -56 + 16.4 + 37.8 = 1.4*F At 10 EFPY -56 + 18.6 + 38.8 = l 4.3*F At 12 EFPY -56 + 20.5 + 39.8 = 9.3*F At 16 EFPY -56 + 23.8 + 41.5 = Therefore, Plate 201 is controlling through 16 EFPY. l ( 5 (240CRS /che ) L_--
The requested data and calculations for BSEP-2 are as follows: Beltline Thickness 5.28" minimum Chemistry Factor CF = Maximum Plate (Cu =.19, Ni =.58) CF = 139.8 Maximum Weld (Cu =.06, Ni =.87) CF = 82 Surface Fluence f = 17 2 At 8 EFPY f = 4.84 x 10 n/cm 17 2 At 10 EFPY f = 6.05 x 10 n/cm 17 n/cm2 'At 12 EPPY f = 7.27 x 10 17 n/cm2 At 16 EFPY f = 9.68 x 10 RT II} Unirradiated RT = NDT NDT Maximum Plate RTNDT(i) 10*F = Maximum Weld RTNDT(i) -56*F = f x e.24t/4 1/4T Fluence g = = At 8 EFPY f* =.0353 x 1019 "/cm2 > 1 MeV At 10 EFDY f' =.0404 x 1019 "/cm2 > 1 MeV At 12 EFPY f =.0529 x 1019 U/cm2 > 1 MeV At 16 EFPY f" =.0705 x 1019 n/cm2 > 1 MeV f (.28 -0.1 logf,) Fluence Factor at 1/4T gg = = g .2411 At 8 EFPY ff = At 10 EFPY ff .2738 = .3017 At 12 EFPY ff = .3506 At 16 EFPY ff = MAXIMUM PLATE CALCULATION l Increase in 1/4T RT CF X ff ART = = NDT NDT (139.8) (.2411) 33.7'F At 8 EFPY ART = = NDT (139.8) (.2732) 38.2*F At 10 EFPY ART = = NDT 42.2*F (139.8) (.3017) At 12 EFPY ART = = l NDT (139.8) (.3506) 49.0*F At 16 EFPY ART = = NDT Adjusted 1/4T RTNDT = RTNDT (i) + ARTNDT + Margin At 8 EFPY 10 + 33.7 + 34 = 77.7'F At 10 EFPY 10 + 38.2 + 34 = 82.2*F At 12 EFPY 10 + 42.2 + 34 = 86.2*F At 16 EFPY 10 + 49.0 + 34 = 93.0*F 6 f (240CRS /che ) L __ ____.
) MAXIMUM WELD CALCULATION l Increase in 1/4T RT X ff ART e = NDT NDT 20*F (82) (.2411) At 8 EFPY ART = = NDT (82) (.2732) 22*F At 10 EFPY ART = = NDT (82) (.3017) 25*F At 12 EFPY ART = = NDT 29'F (82) (.3506) At 16 EFPY ART = = NDT ' # ;;; Oclia (nedrad) = o i g 10*r At 8 EFPY = A ll*F At 10 EFPY o = g 12.5*F At 12 EFPY = A 14.5*F At 16 EFPY o = g /o;2,"A Margin 2 = 2/172 + 102 39,4 7 At 8 EFPY = 2/17 2 2 40.6*F
- 11 At 10 EFPY
= 2 ) 172 + 12.52= 42.2*F At 12 EFPY At 16 EFPY 2 172 + 14.52= 64,7.p = RTNDT(i) + ARTNDT + Margin Adjusted 1/4 T RTNDT 3*F At 8 EFPY -56 + 20 + 39 = 7'F At 10 EFPY -56 + 22 + 41 = ll*F At 12 EFPY -56 + 25 + 42 = 18"F At 16 EFPY -56 + 29 + 45 = Therefore, Plate 351 is controlling through 16 EFPY. 2. NRC QUESTION Identify the limiting (controlled) beltline material in the RTNDT calculation and in Figures 3.4.6.1-1, 2, 3a, 3b, and 3c. CP&L RESPONSE The controlling materials through 16 EFPY are as follows: Brunswick 1 Plate 201 Brunswick 2 Plate 351 Unirradiated controls are noted on heatup-cooldown curvec. 3. NRC QUESTION Provide material and fracture toughness properties (e.g., copper and nickel contents, unirradiated RTNDT' upper shelf energy, etc.) and material identification (e.g., heat number and ASME designation) for all Brunswick Units 1 and 2 reactor beltline materials. 7 (240CRS/che )
CP&L RESPONSE Material and fracture toughness properties of the beltline materials are j shown in Tables 1 and 2. 1 4. NRC QUESTION Provide the unirradiated RT E reactor vessel closure flanges and HDT nozzles (feedwater and recirculation). Identify the limiting flange and nozzle. CP&L RESPONSE The following information is applicable to both BSEP-1 and BSEP-2: Closure flange and adjacent material = 10"F maximum (ASTM E208) Recirculation Nozzles = 40*F maximum (ASTM E208) Feedwater Nozzles = 40*F maximum (ASTM E208) InstrumentationNozzgesNigAandBareexposedtoaneutronfluxof approximately 4 x 10 n/cm -sec maximum in 16 years but do not become controlling for calculations. 5. NRC QUESTION The licensee proposed three separate (8, 12, and 14 EFPY) P/T curves for hydrostatic tests, but only one P/T curve for normal operation (16 EFPY). Provide an explanation for this inconsistency. CP&L RESPONSE Three separate curves (8, 10, and 12 EFPY) were drawn in order to limit the time to reach the hydrostatic test temperature as much as possible by using as low as possible RT temperatures as time progresses. There is NDT no point in drawing curves for longer periods since capsules will be will be required before 16 EFPY. withdrawn and a new appraisal of RTNDT Only one rurve (at 16 EPPY) is drawn for full heatup and cooldown because the curves are not limiting to the BSEP operating procedures. In this case it is best for operator training purposes to keep changes in procedures as few as possible. 6. NRC QUESTION Provide detailed calculations showing how the pressure versus temperature data points on the P/T curves were obtained. If a computer code was used, provide a sample calculation to show the algorithm. i 8 (240CRS/che )
CP&L RESPONEE Pressure versus temperature data points for P/T Curves - The methods of the Standard Review Plan were used and checked by a computer program developed by SIA for EPRI. Since that program has not yet been verified by quality assurance methods, it cannot be part of the formal BSEP Technical Specification submittal. Conservative criteria were used in all Cases. Representative calculations are given below for BSEP-1: PR + P Pressure Stress = t 2 9600 psi 450 x 110 + 450 Pressure Stress at 450*F = = 5.28 2 16,000 psi Pressure Stress at 750*F 750 x 110 + 750 = = 5.28 2 21,760 psi Pressure Stress at 1020*F = 1020 x 110 + 1020 = 5.28 2 .24t/4 fluence (surface) x e Fluence at 1/4 t = (.24 x 5.28) 17 4 17 3.5 x 10 n/cm Fluence at 8 EFPY = 4.84 x 10 x e = (.24 x 5.28) 1 4 1 7.1 x 10 n/cm Fluence at 16 EFPY = 9.68 x 10 x e = [CF]f (.28 .10 logf) ART at 1/4 t = NDT l [100.7] [.2411] = 25.7*F ART at 8 EFPY = NDT [106.7] [.3506] = 37.4*F ART at 16 EFPY = NDT ART = RTNDT(i) + ARTNDT + Margin 10 + 25.7 + 30.7 = 66.4*F ART at 8 EFPY = 10 + 37.4 + 34 = 81.4*F ART at 16 EFPY = Leak and Hydrotest Temperature Determination Kip = 1.5 x M,x Pressure Stress 72,499 psi /in 1.5 x 2.2 x 21,969 Kip at 1030 psig = = 78,468psijin 1.5 x 2.22 x 23,564 K at 1105 psig = = ip 9 (240CRS/che )
l 1 1 IR = -26.777 = 1.233e.014493 (T-RTNDT + 160) l K T at 1030 psig @ 8 EFPY = 155.6*F T at 1105 psig @ 8 EFPY = 164.1*F Heatup or Cooldown f rom Leak or flydrotest KlR*2*M x Pressure Stress + K m 1T 10,667 + 2000* = 48,508 psi / in at 500 psig Kg= 2 x 2.18 x 104.2" T = 17,067+2000*=76,992 psi)in at 800 psig K1= 2 x 2.197 x 159.2* T =
- K taken from WRC Bulletin 175.
1T f 10 T on A 60* safety factor was added to the flange initial RTNDT the advice of Chicago Bridge & Iron Company. Normal Operation with Core not Critical at 16 EFPY - The feedwater nozzle data points were taken from sister plant submittats (ifatch 1 and Fitzpatrick) but corrected for differences in initial RTNDT. The recirculation suction which is representative of the beltline was calculated as follows: PR + P Pressure Stress = t 2 Pressure Stress at 600 psig = 600 x 110 + g = 12,800 psig 5.28 2 2 x 2.19 x 12,800 + 6400 = 62,500 psi in K = 1R In (62.5 - 26.777) +.014493(81.4) - 2.52833 T = .014493 153.7* T = 25,600 1200 x 110 + 1200 Pressure Stress at 1200 psig = = 5.28 2 129.3Ksijin 2 x 2.4 x 25,600 + 6400 K = = 1R 226.4*F in(129.3 - 26.777) - 1.3486 = T = .014493 K was taken from WRC Bulletin 175 by doubling the 50*/hr stress IT intensity. 10 (240CRS/che)
f-l l A safety factor of 20' was added to these calculations as offset for any impact that Revision 2 to the Regulatory Guide 1.99 might have on SRP assumptions. Normal operations with core critical 40*F was added to the noncritical calculations as required by Appendix C to 10CFR50. 7. Charpy V Shelf Energies These are unavailable in most cases; therefore, the Standard Review Plan will be followed. These energies will be determined on the materials in the first capsule removed from each plant. 11 (240CRS/che)
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e J ( 'l 1 Figure 2 Brunswick Unit 2 R,0 Reactor Geometry l li L ( esp =g oo .c o w l Insulation s { Pressure Vessel Jet Pumps. / G, Shroud g g f F c, .x. t ;,-l;, %. 3! i 4 N 9 / w I N / Core ,/ s l,, .s. / r 4 A j. ./ s / i i0 50 100 150 200 250 350 400 Radius (cm) (240CRS/en) -____-___--_-__-_______a
! ~ Figure 3 I Brunswick Unit 2 Radial Power Distribution j o l / / / / .442 .726 .671 .340 1.067 .910 ,875 .485 1.217 1.158 1.085 .987 .532 i 1.178 1.149 1.230 1.125 1.057 .911 .578 i .794 1.116 1.276 1.027 .980 1.091 .970 .805 .388 1 .780 .834 1.224 1.234' 1.019 .972 1.146 1.090 .948 .535 l 1.273 1.175 1.247 1.136 1.326 1.232 1.228 1.152 1.129 .982 .560 1.360 1.243 1.364 1.258 1.301 1.177 1.236 1.127 1.472 1.'124 .996 .583 .887 1.290 1.384 1.119 1.060 1.216 1.258 .862 .791 1.077 1.129 .991 .597 i (240CR$/cn)
ENCLOSURE 2 BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 NRC DOCKETS 50-325 AND 50-324 OPERATING LICENSES DPR-71 AND DPR-62 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION l REACTOR COOLANT SYSTEM PRESSURE / TEMPERATURE LIMITS REACTOR CAVITY NEUTRON DOSIMETRY PROGRAM FOR BRUNSWICK UNIT 2 l I (240CRS/che ) _ _ _ _ _ _ _ _ _ _ _ _}}