ML20039E067
| ML20039E067 | |
| Person / Time | |
|---|---|
| Site: | 05000561 |
| Issue date: | 05/25/1977 |
| From: | Ross D Office of Nuclear Reactor Regulation |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML111090060 | List:
|
| References | |
| FOIA-80-515, FOIA-80-555 NUDOCS 8201060489 | |
| Download: ML20039E067 (9) | |
Text
_
u6
//
" L I' y '
/
.[ 'T
^
t.,
nucuAn R LAT cot.V.* t Ss t O N vy?:cin]#4 4 E..v
,l!
W A 5mN G T ON. p.
f,.
20555
-2 gT'
\\.....
It.AY 2 5 W Doctet flo. 0-566 i:Lf10PA!!DUM F0k:
Domenic B. Vassallo, Assistant Director for LWRs, DPM FOM:
D. F. Ross, Jr., Assistant Director for Reactor Safety, DSS St. RJECT:
BSAR-205 SER IliPUT I:cnt !;tns:
CEAP.-205 Licensing Stege:
[
le cl et ib.:
P-566 Milestone :;o.:
24-22 I4sannsible Branch and Project l'anager:
LUR-1,T. Cox Systems safety Branch Involved: Analysis Branch D scription of Review: SER Input e.euuested Completion Date:
1/31/77 Peview Status:
Complete at lillestone, Outstanding Issues fioted teclosed please find the SER input for BSAR-205, Section 4.4.
pp]
}ss(,Jr.,Assi~santDirector D. ;f-for Reactor Safety Division of Systems Safety Eiiclosure:
DER Input t.c : S. Hanauer R. Heineman D. Ross T. Cox
- 2. Rosztoczy L. Phillips W. Hodges G. Kelly
.ntact:
!;RR: G. Kelly Ext. 27588
'ns a
S 9201060489 810443 POR FOIA sa t'ADDEN80-515
- PDR 1.
T
~
' 4.4 Tidrmal and Ilydraulic Design.
The principal criterion for the thermal-hydraulic design of a
/-
p.
.~
~
reactor is avoidance of thennally 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
~
r.ust be designed so that the fuel-to-clad gap characteristics J
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 5030*F at BOL and refutes linearly to 4800*F at E0L (43,000 ! bid /mtu).
2.
The minimum allo'. table departure from nucleate boiling ratio. (DNBR) during steady-state operation and anticipated transients is 1.30 with the BAU-2 correlation.
3.
Flow stability is required during all steady-state and operational transient conditions.
l The thermal and hydraulic design parameters for the reactors are listed in Table 4.4-1.
A comparison of these parameters with those of Bellefonte-is given in the table. The hydraulic analysis was based on vessel model flow tests for a 205 fuel assembly design.
Review of the 205 fuel' assembly vessel model flow tests (BAW-10025P)
L applicable to the design of' BSAR-205, will be completed by the final riesign review stage to confirm the acceptability of the thermal-hydrnlic calculations.
?
- References at end of saction.
j.,
):p
' ~
r
=
TABLE 4.4-1
-e
- sr - ~
THERfML AND HYDRAULIC PARAMETERS BSAR-205 Bellefonte 1&2 Reactor Core Heat Output (MWt) 3800 3600 System Pressure, nominal (PSIA).
2250 2250 i'.i r.i nut D'3R at design power 1.83 1.82 Minimum DNSR at design overpower (112'4) 1.44 1.4.
6 Total Reactor Coolant Flow (10 lb/Hr) 158.63 150.5 Core coolant average velocity (Ft/s) 16.9 16.2 Coolcnt tamperature (F) design nominal inlet 569.0 572.3 average rise in : ore 57.0 56.5 2
Total heat transfer surface in core (Ft )
63,991 63,991 Average heat flux (BTU /Hr/Ft )
197,151 186,800 2
Maximum heat flux (BTU /Hr/Ft )
523,083 507,000 Maximum thennal output (KW/Ft) 15.2 14.74 Maximum thermal output at overpower (KW/FT) 17.0 16.51
!!aximum fuel central temperature (F) 100% power 4280 3670 s
112% power
-4540 4470
1 -
- -Preventior, of DNB for steady-state operation and anticipated transients 2 -
,w.
. will assure that the hot spot of the fuel cladding will be maintaind at a temperature only slightly greater than that of the coolant, which will not lead to a loss of cladding integrity.
B&W has calculated that at the beginning of core life at 100% power, with a linear heat generation rate of 15.2 tilewatts per foot, the fuel centerline temperature will be 4280 degrees Fahrenheit.
The peak power density that would occur for a reactor trip at 112 percent raaximum over power trip is 17.0 kilowatts per foot.
At a linear heat generation rate of 17.0 kilouatts per foot, BSW calculated a centerline temperature of approximately 4540 degrees Fahrenheit at BOL, tnus indicating no fuel melting.
The margin to DNB at any point in the core is expressed in terms of the departure frcm nucleate boiling ratio (DNBR). The DNBR is defined as the ratio of the heat flux required to produce departure from nucleate boiling at the calculated local coolant conditions to the actual local heat flux.
The DNB correlation used for the design of this core is the RAW-2 correlation. The BAW-2 correlation was derived from data on 6 ft.
/
heaters which sinulated the rod diameter and spacing of 15x15 fuel assemblies.
BW has comitted to tests on full length heaters of the 17x17 geometry with uniform and non-uniform axial heat flux. Results from these tests and the statistical data analysis must confirm the thermal-hydraulic design prior to approval of the final design report for BSAR-205 plants.
Another parameter that must be accounted for in the thermal-hydraulic J
design of the core is the rod-to-rod bowing within fuel assemblies.
During the'0conee 1 refueling, six fuel assemblies were examined visually and e
- ~
3.
dimensionally. The water channel and line scan measurements indicated a maximum rod bow of approximately 30 mils.
B&W states that the observed rod bow is accommodated within the current design and that B&W is pursuing a program to demonstrate this.
B&W plans to develop bow correlations and predictive techniq0es to analyze the data and the predicted bow from a thermal-hydraulic standaoint in the near future and has cormitted to provide results of the localized effect on DiiB.
Operating restrictions can be applied at the final design review stage if rod-to-rod bowing proves to reduce the r.argin to 9liB for the 17x17 fuci assembly design.
Ct.rrently Rancho Seco is the only operating B&W plant with a rod bow penalty.
The penalty is expected to be reduced or removed when sufficient rod bow data become available.
parallel channel flow stability analysis is performed with the HYTRAN 3
computer code, in coniunction with the CitATA code.
Topical reports'CHATA and HYTRAN are presently under review by the Staff.
Any limitations to the thermal-hydraulic design resulting from the review of IIYTRAN (BAW-10109) and CHATA (BAW-10110) will be compensated for by appropriate eperating restrictions; however, no operating restrictions are anticipated, in Section 4.4.5.9.1 BSAR-205 commits to provide a loose Parts Monitoring System (LPitS) to detect the presence of loose parts in the reactor core and primary coolant system. Additional information on the loose parts monitoring system is to be supplied in the license applications for individual BSAR-205 plants. The staff considers the BSAR-205 commitment as an interface requirement and will review the acceptability of the LpMS
'" a bl.m t -ta ul a n t h.v. i s. 1:e will recaire that the LP!15 be acorocriataly
.o...,_.
~x.
identified as an interface on the BSAR-205 docket.
The 0.99 flow distribution factor proposed by B&W was justified based I*
on statistical analysis of the data'from the Vessel Model ' Flow Test for the 205 assembly plants.
Subject to our acceptance of the 205 assembly model-flow tests now under review by the staff, we believe the justification presented in BSAR-205 provides reasonable assurance' that the minimum flow 3
factor is 0.99.
The staff finds the 0.99 flow factor acceptable with the stated proviso. The operating license applications for BSAR-205 plants should include as-built decwings of the core flew distributor plate for j
ccmparison to the plate dimensions to be included in the-Vessel Model Flow I
Test topical report.I*
Part of the protective mechanism to prevent DNB in t ie core is provided by the Reactor Protection System's calculating module.
The' design and in.plementation of this protective software used to guard against DNB.
must be reviewed and approved by the Staff prior to approval of the final report for BSAR-205 plants. A topical report 4* on'the Reactor Protection l
System (RPS) has been submitted to'the staff for review.
Conclusions i.
On the basis of our review we conclude that the thermal-hydraulic' design of the core is acceptable for Preliminary Design Approval-of the reference system.
Ilowever, we will require that several items be resolved prior to the Final i
,rDesign Approval; these are:
1 1.
Development of data with full length heaters for the BAW-2 correlation.
This data base should include both uniform and r.on-uniform axial heat flux tests on full length heaters with the 17x17 rod diameter and
,y
.r
r i
spacing.
If necessary, the BAW-2 correlation should be nodified to agree with the data.
2.
Statistical analysis of the Dr$ data to verify the minimum DNBR limit for compliance with the 95/95 design criterion.
3.
Revicw of.the flow model tests for the 205 fuc1 asser.bly configuration (BAU-10025P)I 4.
Determination of extent and effect of red-to-rod bowing.
5.
Review and approval of the HYTR/W flow stability couputer code (BAU-10109)2 and the CliATA code (BAU-10110)3 6.
Review and approval of RPS calculating module (BAW-10085P, Rev. 2).
h" y
i
i
}
i i
References l. ]. -. '. : ; ;-- -
1.
R. M. Gribble, " Reactor Vessel hiodel Flow Tests for 205-Fuel E
Assembly Core," BAW-10025P, Babcock and Wilcox Company, (August 1976)..
2.
H. S. Kao, W. R. Cardwell C.D. Morgan, " Hydraulic Transient Code for Investigating Channel Flow Stability (liYTRAN)," BAW-10109, Babcock and Wilcox Company, (January 1976).
3.
J. M. Alcoren, R. H. Wilson, " Core Hydraulics and Thernal Analysis (Cl!ATA)," CAU-10110, Babcock and Wilcox Company, (Janusry 1976).
4.
D. B. Fairbrother, D. R. Vincent, L. M. Lesnaik, E. G. Orgera,
" Reactor Protection System - Revisions 2," BAW-10085P,.Rev. 2.,
(Proprietary version), Babcock and Wilcox Company, January 1977.
4 i
t e
- -l ACRS Generic Items - Analysis Branch
)3l : --
Group I - Resolved Generic Issues Item 9 - Section 3.12.20 of BSAR-205 comnits to a comprehensive test program in compliance with regulatory guide 1.20 - Vibration Measurements on Reactor Intervals (revision 2, May 1976).
Group II - Resolution Pending Item 5 - Loose Parts Moritoring
.See Section 4.4 4
Group IIA - Resolution-Pendino - Items since December 18, 1972 Item 7 - Steam Generator Tube Leakage System design to preclude or mitigate the. consequences of steam.
1 generator tube failure is considered in the applicants scope and will be reviewed when the applicant references B5AR-205.
9 i
i i
r 1
i e
1 e