ML20040A408
| ML20040A408 | |
| Person / Time | |
|---|---|
| Site: | 05000561 |
| Issue date: | 05/28/1976 |
| From: | Phillips L Office of Nuclear Reactor Regulation |
| To: | Deyoung R Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML111090060 | List:
|
| References | |
| FOIA-80-515, FOIA-80-555 NUDOCS 8201210033 | |
| Download: ML20040A408 (5) | |
Text
_ - _ __
- - ~ -
- - - ~
.E.
T[#
UNITED STATES M 8lbT Id NUCLEAR REGULATORY COMMISSION g*
,hl j
WASHINGTON, D. C. 20555 e/
NAY 2 81976 Richard C. DeYoung, Assistant Director for LWRs, DPM THRU:
Zoltan R. Rosztoczy, Chief, Analysis Branch, DSS 7 BSAR-205 - FIRST ROUND QUESTIONS Plant Name: BSAR-205 Licensing Stage:
CP Docket No. or Project No. : P-566 Milestone No.: 05-24 Responsible Franch and Project Manager:
LWR-1, T. Cox Systems Safety Branch Involved: Analysis Branch Description of Review: First Round Questions Requested Completion Date: May 24, 1976' Review Status: Awaiting Additional Information Enclosed please find the first round questions for ESAR-205.,
f I
I f ~ _..,. e j j m w& W i,yo l
n' Laurence E. Phillips, Section ader i.
Reactor Analysis Section Analysis Branch
[
Division of Systems Safety l
Enclosure:
Questions cc:
S. Hanauer R. Heineman D. Ross M
L. Phillips Z. Rosztoczy
^ W. Hodges G. Kelly l
/
A i
8201210033 810403 PDR FDIA MADDEN 80-515 PDR
~
ROUND ONE QUESTIONS - BSAR 205 220.0 Analysis Branch 221.1 Provide a reference or a description of the digital computer (4.4.2.2) code used for calculating clad temperature and other thermal-hydraulic conditions.
Clarify the relationship of this code to other calculations of thermal-hydraulic conditions.
What design limits are evaluated by each code?
221.2 Provide a discussion of the correlation of the effects of (4.4.2.3) densification on 17 x 17 fuel vs. 15 x 15.
Include analytical and experimental work, and describe where you plan additional work or feel additional work is required.
Indicate this with a schedule of future availability of information.
221.3 How is the safe operating level with a flux tilt determined?
-(4.4.2.4)
Describe the operational procedures and the safety margins vs.
flux tilt determination procedures which you will propose for technical specifications. Does the operator have override power over the ICS's control of rod position?
221.4 Provide assurance that the greater than average fuel assemblies
,(4.4.2.6) will receive more than average flow.
Include specific references
(
to the design data and vessel flow test which justify the flow distribution.
221.5 Provide a detailed breakdown or references to quantify and justify (4.4.2.6) the flow reduction components described in items 2 thru 4, page 4.4-4.
In general, provide sufficient information to substantiate the 85.9% figure.
221.6 Provide a reference for the experience from the 177-FA core design (4.4.2.6) and show the optimization technique for the fixed core flow dis-tributors. When will the topical report pertaining to the partial loop operation tests be'availabic? Was 99% minimum flow achieved for 177-FA design?
1 221.7 You have previously stated that there is a net downward force in (4.4.2.7) excess of 100 pounds at all times during narmal operation for similar cores. Now you have deleted reference to the actual force.
What is the new value? Provide the substantiation.
Include calcu-lational, tolerance, experimental, and scaling errors in the sub-stantiation.
221.8 The total steady-state unrecoverable pressure drop is stated to be (4.4.2.7) from one inch below the lower support grillage up to and including the upper grid assembly.
This is a change in the description from previous submittals.
Compare the pressure drop profile vs. elevation s _
for this core to that of Greene County.
i I
~
220-2 l
221.9 Equation 4.4-1 is not an energy equarion. Equation 4.4-2 would (4.4.2.8.1) normally be obtained by taking a balance over an increment of length AZ.
If equation 4.4-1 is truly integrated, please show the steps and assumptions. Discuss the error involved in ap-plying this equation in regions of variable flow area due to spacers or bowed fuel pins. What equations are used for two phase flow r'egions?
9 221.10 Provide a schedule of submittal or a reference for the value of (4. 4. 2. 8.1)
Ck in equation 4.4-7.
Justify the use of this single phase equation for two phase flow.
The references provided should also include pressure drop behavic: and hydraulic forces as mentioned in Section 4.4.4.
221.11 The range for mass velocity in the BAW-2 data base has been 6
(4.4.2.8.2) expanded from a previous value of 0.95 x 106 lb/hr ft2 to 0.77 x 10,
Justify.the expanded range.
221.12 What code was used to calculate equation 4.4-23?
(4.4.2.8.3) 391.13 What is the source of Equation 4.4-24? Why is this different
's. 2. 8. 4 )
from your proposed equation in TACO?
/
221.14 You assume the fuel is always in contact with the clad when the (4.4.2.8.4) fuel diameter is smaller than the clad diameter.
Show the form of the analysis utilized for this case and compare typical results to the case in which the fuel is centered in the clad.
221.15 A factor of 12 and other changes have been made to Equation 4.4-28
~~ -
(4.4.2.8.4) compared to Equation 4.4-31 of the Pebble Springs PSAR submittal.
Explain the differences. Are the brackets placed properly in the definition of F,2?
1 221.16 Provide a detailed breakdown of the uncertainty components (4.4.2.10.3) comprising the 2.5% thermal uncertainty.
221.17 Why isn't a 5% local peaking factor included in the determination (4.4.2.10.3) of radial peaks as was done for Pebble Springs?
221.18 You are using a.008 inch cold diametral gap.
This doesn't appear (4.4.2.10.3) conservative in comparison with previous submittals such as Pebble Springs. Please quantify effects of this change and provide justi-fication.
221.19 You state that the uncertainty associated with total peaking factor
4.2.10.4) is 7.5%, but in 4.4.2.10.3, you state it is 10%.
Explain.
s_
l 1
~
I-
[~
g' 220-3 x
221.20 Justify the uncertainty in the values of core pressure and (4.4.3.4.1) inlet temperature which are given as -45 PSIA and +2*F, respectively.
Include a detailed breakdown of all uncertainty components. Why was the +3*F uncertainty used in Pebble Springs reduced to +2*F?
In item 5, what is the area reduction factor used?
3 221.21 Do you have experimental evidence that the pressures in the (4. 4. 3. 4. 2) plena are the same on horizontal planes adjacent to the core?
Justify your assumptions.
221.22 Provide the basis for selection of your hot channel factor.
(4.4.3.4.3).
At the top of page 4.4-27 you reference a 99% confidence.
Is this correct or should it be 95%? When will the measurements and analytical techniques be described for the 17 x 17. fuel?
221.23.
Could channels other than the hot channel exhibit unstable (4.4.3.5) behavior? If so, justify your assumption of a constant pressure drop.
221.24 What are the values of rated partial power for partial pump (4.4.3.5) operation?
.25 Have you determined limit values which lead to instability for (4.4.3.5) these parameters? Have anticipated transients been considered?
J 221.26 Provide information or a schedule for submittal of a topical (4.4.3.7) report discussing in detail the effects of fuel rod bowing.
221.27 Correct the referenced section number for discussion of maximum (4.4.4.3.7) internal rod pressure and nominal coolant pressure.
221.28 Justify the statement that the range selected more than covers (4.4.3.8) the expected DNB claddin~g temperature. Were rod temperatures higher in ATWS calculations?
221.29 Provide information on or a schedule for submittal of crossflow (4.4.4) codes.
Provide a reference for the Battelle work. When will the VMFT report be available?
221.30 Provide a schedule for submittal of information or a topical (4.4.4) report giving results of the crud buildup evaluation.
221.31 The discussion on loose parts monitoring is inadequate.
Provide (4.4.5.9.1) your design documentation on the loose parts monitorine system, including but not limited to, the design criteria, design description, current development status, future development requirements and schedule for completion.
s
4.
=
.-- ~-x~~~-
A f',
220-4'
~
1
~
221.32 Provide design documentation s'imilar to that reque,sted in question (4.4.5.9.2) 221.31 for the Neutron Noise Monitoring System.
221.33 Provide design documentation similar to that requested in question (4.4.5.9.3) 221.31 for the Vibration Monitoring System.
221.34 Some of the tabulated volumes have changed from those given (Tables 4.4-4 for Pebble S'prings, but a cursory. investigation of the figures and 4.4-5) and other tables does not show how they are consistent with the changes. Lengths appear to be the same in some of the cases, flow areas are the same, but volumes have changed. Explain and provide sufficient dimensional information that we can check for consistency.
221.35 BAW-2 gives an inflection point near the upper end of the. core (Figure 4.4-4) that did not appear on the Pebble Springs submittal. What are the physical reasons for this?
221.36 This figure is identical to the Pebble Springs submittal.
Is (Figure 4.4-8) this consistent with plant differences in design core heat out-put, average teat flux, average thermal output, hot spot max /
average heat flux ratio, and total reactor coolant flow?
.37 Why has the shape of this curve changed from that of Pebble
..gure 4.4-20) Springs?
221.38 Jhe temperature spike of this curve is different from that of (Figure 4.4-25)Greene County and Pebble Springs.
Explain why.
221 39 In comparison of this figure with Greene County and Pebble Springs, (Figure 4.4-29)we find that the conditions at small blockages are quite different but that the minimum DNB occurs at about the same blockage.
Physically, why is this the case?
221.40 Satisfactory completion and documentation of the tests described (None) in subsections 1.5.1, 1.5.2.1.1, 1.5.2.1.3 and 1.5.2.1.4 with suitable inclusion of the results in the design consideration will be required in the FSAR. Provide a commitment in the PSAR to supply these results for use in the FSAR review.
221 41 A full length bundle test has been scheduled by B&W. Completely (None) describe this test and the incorporation of the test results into.
your system design'.
s _
e
,_