ML20039F110
| ML20039F110 | |
| Person / Time | |
|---|---|
| Site: | 05000561 |
| Issue date: | 05/04/1978 |
| From: | Cox T Office of Nuclear Reactor Regulation |
| To: | Rosztoczy S Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML111090060 | List:
|
| References | |
| FOIA-80-515, FOIA-80-555 NUDOCS 8201110873 | |
| Download: ML20039F110 (8) | |
Text
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D enci a, Occket File y4y,g g
LWR r3 File-J T. Coxw M. Rushbrook Docket flo. STN 50-5G1 2
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MEMORAllDUM FOR:
Z. Rosztoczy, Chief, Analysis Branch, Division of k
Systems Safety
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FROM:
T. Cox, Project Manager, Light Water Reactors Branch No. 3 Division of Project Management
SUBJECT:
BSAR-205 SER, SECTI0li 4.4, THERMAL AUD HYDRAULIC DESIGN
Reference:
Memo, D. F. Ross to D. B. Vassallo, " Evaluation of the
, {i Effect of a Change in the Flow Maldistribution Factor on DiB for Babcock & Wilcox 205 Fuel Assembly Plants",
yj April 27,1978 The referenced meno was transmitted in response to the scheduled input needed for completion (updating) of the BSAR-205 SER prior to a PDA issuance.
I have altered the contents of both the referenced memo and
]t Section 4.4 of the July 1977, Report to the ACRS on DSAR-205, in order to fit these separate reports into a final SER Section 4.4.
Tae result i
is enclosed.
L 4
Please review and coment on the Enclosure (SER Section 4.4) as soon as possible as it is needed for final SER issuance on BSAP.-205.
I Originc.! 3?qued bZ
]i Thomas h. Cox, Project Manager Light Water Reactors Branch No. 3 g
Division of Project flanagement
Enclosure:
1 As Stated cc w/o enclosure:
D. B. Vassallo y
D. Ross
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R. DeYoung
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- 0. Parr D. Tibbitts S. Burwell LWR #3: LPM
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't-' a L ".C m e: b f r t,.e P.n en current schedales will de Bellefonte Unit 1.
4.3.6 Analytical m thods e
BsDcock & Wilcos has sabmitted a series of topical reports (BAW-10111 to 10115, BAe-10117, and BAw 10124) which describe the analytical methods used in the nucleae design of reactoe cores. These reports describe the data base used and j
the manner in which the data are treated to obtain calculational parameters for use in the design codes PDQ07 and FLAME, to perform the design calculations.
Sufficient detail is given concerning data sources, physics of the calculations and calculation procedures to conclude that the methods employed are " state-of-the-art."
Vehaverjviewedthesereportsandhaveconcludedthattheyareacceptableas X
referenceginlicensingsubmittals. We conclude that these methods are acceptable X as used in the BSAR 205 cesign at the preliminary design review stage.
4.4 The mal and Hydeaulic Desion 4.4.1 Evaluation The principal criterion for the thermal-bydraulle design of a reactor is avoidance of thermally induced fuel damage during normal steady-state operation and during anticipated operational occurrences. Babcock & Wilcox uses the following design limits to satisfy this criterion:
(1) The fuel rod cladding, fuel pellets, and fuel rod internals must be designed so that the fuel to-Clad gap characteristics ensure that the maximum fuel temperature does not exceed the fuel melting limit at the 112 percent design overpower at any time during core life. The fuel melting temperature is 5080 degrees Fahrenheit at the beginning of core life and reduces linearly to 4800 degrees Fahrenheit at the end of core lifetime (43,000 megawatt-days per metric ' ton of uranium).
(2) The minimum allewable departure frem nucleate boiling ratio during steady-state operation and anticipated transients is 1.30, based on the BAW-2 correlation.
(3) h draulic stability is required during all steady state and operational transient conditions.
The shemal and hydraulic design parameters for the BSAR-205 and tre Bellefonte reactors (Cocket Nncers 50-433 and 50-433) are compared in Table 4.2.
These reactors are physically identical. The BSAR-2C5 reactor is designed to be operated at a hig*er maxim power leve' ind the incevased heat will be removed by a Mgher -enctor coolant rhw rate. The Sydrsulic analysis was based in part 4-21
r tan E 4.2 CCW AR!$CN CF THERMAL AND WDilAgdC PARAMETERS Bellefonte BSA8-205 1&7 Reactor Core Meat Cutput, the-mal megawatts 3800 3600 System Pressure, pounds per square inch, absolute 2250 2250 Minimum Ce;4rture from Nucleate Boiling Ratio, at Design Power 1.B3 1.82 Minteum Departure from Nucleate Boiling Ratio, at Design C,erpo.er, 112 percent 1.44 1.4 Total Reactor Coolant Flow, million pounds per hour 158.63
.150.5 Average Velocity of Core Coolant, feet per second 16.9 16.2 Coolant Temperature, degrees Fahrenheit Design Inlet 569.0 572.3 I
Average Rise in Core 57.0 56.5 Total Heat Transfer Surface in Core, square feet 63,991 63,991 Average Heat Flux, British Thermal Units per hour per square foot 197.151 186,800 Maximum Heat Flux, British Thermal Units per hour per square foot 523,083 507.000 l
Maximus Thersal Output, kilowatts per foot, at design power 15.2 14.74 i
Maximum Thermal Output at 112 percent Operpower, kilowatts per foot '
17.0 16.51 Maximum Fuel Central Temperature, degrees Fahrenheit, at beginning of core life 100 percent pcwer 4280 3670 112 percent power 4540 3970 l
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ine o. scriptia ano analysis of the vesset moel flow test was sMitte3 fer our review as Eabcock & W'lcos Topical Repcrt !Mter 3AW-10025P, "Peactor ','essel "odel Flow Tests for the 205 Fuel Assembly Core," August 1976. We intend to complete the review of ',his report prior to the sutmittal of the first Final Safety Analysis Report to derence the BSAR-206 design.
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The margin to departure from nucleate boiling at any point in the core is expressed in terms of the departure frcm nucleate boiling ratio. This ratio is defined as the heat flux required to produce departure from nucleate boiling at the calculated x
local coolant conditions divided by the actual local heat flux.
The margin to departure from nucleate boiling will be chosen to provide a' 95 percent probability, with 95 percent confidence, that departure from nucleate boiling will not occur on fuel rods having the (calculated) minimum departure from nucleate l
boiling ratio during normal operation and anticipated operational occurrences.
For the BSAR-205 Preliminary Safety Analysis Report, Babcock & Wilcox demonstrated the effect of various parameters on the departure from nucleate boiling ratio using the following analytical methods:
i (1) BAW-10021. " TEMP-Thermal Inthalpy Mixing Program". April 1970, and
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l (2) BAW-10110. "CHATA-Core Hydraulics and Thermal Analysis", February 1976.
BAW-10021 was approved by the staff, but BAW-10110 is still under review. Any limitations on the thermal hydraulic design resulting from the topical report reviews can be compensated for by appropriate operating restrictions at the operating license review of plants referencing BSAR-205.
The departure from nucleate boiling correlation used for the design of this core is the BAW-2 correlation. The BAW-2 correlation was derived from data on six-foot long heated rods which simulated the rod diameter and spacing of 15 x 15 fuel assembl ies. Babcock & Wilcox has stated that they will perform tests on full-length heated rods of the'17 x 17 geometry with uniforin and nonaniform axial heat flux.
i Results from these tests and the statistical analysis of the results must confirm the thermal-hydraulle design prior to issuance of an operating license ~ for plants referencing the BSAR-205 design.
Prevention of departure from nucleate boiling for steady-state operation and anticipated transients will assure that the hot spot of the fuel cladding will be maintained at a temperature only slightly greater than that of the coolant, which will not lead to a loss of cladding integrity. Babenck & Wilcox has calculated that at the beginning of core life, at 100 percent power, with a linear heat generation rate of 15.2 kilowatts per foot, the fuel centerline temperature w'1:
be 4.280 degrees Fahrenheit. The peak power density that would occur prior to a reactor trip at 112 percent (maxiraum) overpower is 17.0.kilcwatts per foot. For a linear Oc:t ger.cration rate of 17.0 kilowatts per foot. Babcock & Milcox calculated a centerline temperature of approximately 4540 degrees Fahrenheit (at the teginning of core life), thus indicating no fuel melting.
4-23
A'ta- +:e 6sd compt ated ou-preliminary review of t*e ??AC.??' thermal.bydrauties
- c;*gn S..%)/ '977. Bab ces 1 wit
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1977, tnat tre v.intm2m ficw tettor tcr U.e core hot chant:el, as catain65 fruen tre
<ctsel mcdel f10. tests (reported in 8AW-1002f?), should be reduced from 0.99 to 0.965. The minimum flow factor utilized by Cabcock & Wilcox was based on statistical analysis of the data from the vessel model flow test for the 205 fuel assembly plants.
Using the CHATA/ TEMP analyses, the reduced minimum flow factor reported by Babcock & Wilcox has the effect of reducing the previously calculated minimum departure from nucleate boiling ratio (1.31) by 0.08.
The CHATA/ TEMP analyses are based on the conservative assumption of no cross. flow between flow channels in the reactor core. Babcock & Wilcox has also developed, and submitted to us for review, an advanced analytical method based on the use of codes which account for cross. flow between adjacent flow channels.
I The methods are:
(1) BAW.10129, " LYNX 1 CODE", October 1976 (2) BAW.10130, " LYNX 2 CODE", October 1976 Using the LYNX 1/ LYNX 2 analyses, the effect of the reduction in minimum flow 4
factor is a smaller reduction in the minimum departure from nucleate boiling ratio, one of 0.01.
Babcock & Wilcox described these analyses in a letter dated February 6,1978, from J. Taylor to S.,Varga. The analyses in the letter of February 6 demonstrate that although the BSAR.205 core thermal analysis was done with an incorrect minimum flow factor for the hot channel, this error is almost entirely offset by the conservatism to the use of the (closed channel)
CHATA/TEMD codes. We espect that reanalysis at the Final Safety Analysis Re rt ed stage, using the LYNX 1/ LYNX ' codes and the correct minimum flow factor, result in approximately the same minimum departure from nucleate boiling ratio as the CHATA/ TEMP analyses repcrted in the Preliminary Safety Analysis Report.
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Anct.her significant effect on the analyses is that of the upper plenum radial I
pressure gradient. Upper plenum pressure gradients were measured in the vessel l
model flow tests and in the Oconee 1 reactor. A pressure profile which bounds the existing test data was then used in the LYNX analyses to determine the effect of the pressure gradient on the minimum departure from nucleate boiling ratio.
The result was that the minimum departure from nucleate boiling ratio was reduced by 0.04 using the BAW.2 correlation.
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The total redaction to wirdmm ecarture from nuclette boiling ratio then, usfrg tM.tU m ? m, it ' Ci fcr tv f % 9c*or r9dur_ tion *, 0.g55, Inc a fur! ~ 0.04 fo< !.r af % t ^,f tna >; - ; 4 v re1ul p m s n y ad'ent (necessary in tr.c M channel LYNX analyses). The BSAR-205 design as originally presented in the Preliminary Safety Analysis Report had a margin of
,, h between the minimum departure from nucleate bolling ratio and the allowable i '
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minimum value of 1.30 which we had already ap;;eeved for the BAW-2 correlation.
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This margin is larger than the combined effect (using LYNX 1/ LYNX 2 methods)
( of the inlet flow error correction and the upper plenum pressure gradient effect.
- [ Therefore, in our judgement, the preliminary design calculations are sufficient t/*
to indicate an acceptable minimum departure from nucleate boiling ratio, at the
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Preliminary Design Approval stage of review.
I The LYNX 1 and LYNX 2 codes are under review. We may identify a need for additional safety margin over what is now incorporated in the reports submitted by Babcock & Wilcox. However, the effect on plant operation can be bounded by considering the effect of the upper plenum radial pressure gradient with a closed channel code; our calculations show this effect to be an approximately 10% reduction in the departure from nucleate boiling ratio. Even this calculated amount could be offset by a future reduction in radial power peaking, by a core inlet temperature reduction, or both. Adjustments in plant parameters, if necessary, will be addressed in the final design review of BSAR-205.
The Final Safety Analysis Report for BSAR-205 should include both thgorrect minimum flow factor and the upper plenum pressure gradient.
We wil1 require 4
4 the final design description of the SSAR-205 design to include as-built drawings of the core flow distribution plate for comparison to the plate dimensions in the vessel model flow test topical report.
We expect plants utilizing the BSAR-205 design to be capable of full power operation even if adjustments in final thermal-hydraulic design parameters are I
required.
Another parameter that influences the thermal-hydraulic design of the core is the rod-to-rod bowing within fuel assemblies. During the Oconee 1 refueling, six 15 x 15 fuel assemblies were examined visually and dimensionally. The water r
channel and line scan measurements indicated a maximum rod J
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ctse ved acd bew is accowodated withir the cur-ent 17 x 17 cesign and that they are pursuing a program to demonstrate this. Batcock & Wilcox plans to develop bow correlations and predictive techniques to analyze the data and the predicted bow from a
- Jet :al-hydraulic stand;oint in the near fut:.re, and has co=mitted to provide results of the localized effect on the deoarture from nucleate boiling.
Operating restrictions will be applied at the final design review stage if rod-to-rod bowing proves to reduce the margin to departure from nucleote boiling for the 17 x 17 fuel assembly design. As discussed in Section 4.2.1.3 of this report, we conclude that the information regarding rod bowing now available from Babcock & Wilcox is acceptable at the preliminary design stage of review.
Rancho Seco (Occkat Number 50-312) is the only operating Babcock & Wilcox plant with a rod bow penalty. The penalty is expected to be red 0ced or remov'ed when sufficient red bow data become available.
Parallel channel flow stability analyses are performed with two computer codes which have been submitted for staff review as topical reports:
(1) BA= 19109, " Hydraulic Transient Code for Investigating Channel Flow Stability (HYTRAN)," January 1976, and (2) Bra-10110. " Core Hydraulics and Thermal Analysis (CHATA)," January 1976.
We are reviewing these reports. Any limitations to the thermal-hydraulic design resulting from the topical report reviews can be compensated for by appropriate operating restrictions applied during the operating license review.
!w. minimum flow stribution fa-*
util! Zed by Babcock Ilcox was justi-The ed based on istical analy of the data from the ssel model flow test f the 205 f assemely pla subject to our ace.ance of the topical e rts on this est ncy under ew, we conclude tha e justification pres ed in AR-205 provid reasonable assurance the preliminary de
,n review stage, that the a mum flow factor is 0.
We will require e rating license applica-tions forencing the BSAR-2C.esign to include a uilt drawings of th re w distributor plate f? comparison to the
- e dimensions in th essel modp a flow test topical re t.
Protective action to prevent departure from nucleate boiling in the core is pro-vided in part ey the reactor protection system's calculating module. We will review the design and implementation of the protective soft.are used in the calcu-lating module prior to issuance of an operating license for an application refer-encing BSAR-205. Staff review of the reactor trotection system (RPS-II) is discussed in Section 7.2 of this report.
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- .....>..2 On the basis of our review of the themal-hydraulic analytical technigt.es and available supporting esperimental data, we conclude that there is reasonable assurance that (1) the ; opesed tharral-hydraulic design will account for depar-ture from nucleate boiling and fuel' centerline temperature limitations in a satisfactory manner, and (2) $ conservatism in the thermal-hydraulic design X
procedures can be verified. Therefore, we conclude that the presently available information on the preliminary thermal-hydraulic design of the BSAR-205 reactor is acceptable for issuance of a construction permit or Preif ainary Design Approval.
However, we will require that several items be resolved prior to a Final Design Approval or operating license issuance on an application referencing BSAR-205.
These are:
(1) Development of critical heat flux cata with full-length heaters for the SAW-2 correlation. This data base should include both uniform and nonuniform axial heat flux tests on full-length heaten with the 17 x 17 rod diameter and spacing. If necessary, the BAW-2 correlation should be modified to agree with the data.
(2) Statistical analysis of the critical heat flux data to verify that the mini-mum departure from nucleate boiling ratio complies with the 95/95 design criterion.
(3) Review and approval of the vessel model flow test topical report for the 205 fuel assembly configuration.
(a) Dete*mination of extent and effect of rod-to-red bowing.
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(5) Review and approval of the HYTRAN g CHATAj odes.
(6) Review and approval of the reactor protection system protective software.
In the event that the analytical methods are determined not to be conservative during the final design review, appropriate restrictions on operation can be established at the operating license stage for applications referencing the BSAR*205 nuclear steam supoly system.
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