ML20040A036
| ML20040A036 | |
| Person / Time | |
|---|---|
| Site: | 05000561 |
| Issue date: | 11/05/1976 |
| From: | Parr O Office of Nuclear Reactor Regulation |
| To: | BABCOCK & WILCOX CO. |
| Shared Package | |
| ML111090060 | List:
|
| References | |
| FOIA-80-515, FOIA-80-555 NUDOCS 8201200237 | |
| Download: ML20040A036 (6) | |
Text
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7 UNITED 3TATES 4
NUCLEAR REGULATORY CCMMISSION
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? U3 Docket No. STN 50-561 Babcock & Ullcox Company ATTN:
Mr. Kenneth E. Suhrke Manager, Licensing Nuclear Power Generation P. O. Box 1260 Lynchburg, Virginia 24505 Gentlemen:
ROUND 2 POSITIONS AND REQUEST FOR MORMATION As a result of our continuing review of the Babcock & Wilcox Standard Safety Analysis Report BSAR-205, your response to certain Regulatory staff positions and requests for information is required. The specific information required is detaileddn the Enclosure. - Regulatory Positions are identified by (RSP) underneath the position number shown in the Enclosure. We are prepared to meet with ou to discuss further any of our positions to assure complete understanding of the factors at issue and the bases for our positions; however, we do not believe extended or iterative debate would be useful.
In reviewing Sections 4.4 and 15 of the BSAR-205, we have identified certain technical information regarding your design that must be approved by the staff prior to a favorable decision by the staff on issuance of an /
operating license for a plant using the BSAR-205 system.
Such approval is not required for us to complete the current preliminary design review.
The required information is described here for your future reference:
(1)
Your heat transfer correlation, BAW-2, has not been approved for Mark C 17X17 fuel assemblies, and such approval depends in part on critical heat flux data you have committed to provide.
(2)
Approval is required for the following individual codes and their combined, systematic application to the design basis analyses of the BSAR-205 system:
TACO, IIYTRAN, CHATA, TRAP, ATUS codes (BAW-10099, BAW-10101) and POWERTRAIN.
r201200237 810403 PDR FOIA MADDEN 80-SIS PDR
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i UNITED STATES NUCLEAR REGULATORY COMMISSION W ASHIN GTON, D. C. 20555 301-492-7000 FACSIMILE SERVICE REQUEST DATE: // d7 7[
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MESSAGE TO:
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804-394 - 777 3 AUTOMATIC:
No VERIFTCATION NUMBER 804-3&4-2 742.-
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PLUS INSTRUCTION SHEET STATE & CITY M.
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M TELECOPY NUMBER 301 492-8110 RAPIFAX AUTOMATIC 301 492-7617 3M VRC AUTOMATIC VERIFICATION NUMBER 301-492-7371 BUILDING OFFICE PHONE 3.'72.4 3 MAIL STOP
/2.8 CLASS OF SERVICE:
Overnight 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 2 hour 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Immediate SPECIAL INSTRUCTIONS:
Received / Tim -Date Transmitted / Time-Date 1
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l Babcock & Wilcox Company <.y 7 g73 d
If you have not already submitted data for, review concerning the above items, you should plan on making such submittals no later than as part of the operating license application of the first applicant to i
reference BSAR-205 in such an application, or in your own application i
for final design approval of BSAR-205, whichever is earlier.
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I foresee a potential extension of the licensing review' schedule outlined in our letter of April 22, 1976 (R. Boyd to K. Suhrke). To keep such an extension to a minimum, we need your complete responses to the Enclosure i
items by November 30, 1976. Please inform us within seven days after
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receipt of this letter of the date on which you plan to respond so that we may revise our schedule as necessary.
If you plan to appeal to licensing management on any of the positions in the Enclosure, please advise us of your intentions within two weeks.
1
.t Please contact us if you have ahy questions.
i Sincerely, t
k
%,k Olan D. Parr, Chief Light Water Reactors Branch No. 3 Division of Project Management
Enclosure:
Positions and Requests for Additional I' formation n
ec: See next page f
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's ENCLOSURE POSITIONS AND REOUESTS FOR INFORMATION BSAR-205: DOCKET NO. STN 50-561 220.0 ANALYSIS 221.42 Since Equations in Hytran assume no heat generation in the (4.4.2.2) fluid, justify your statement that ye assume 97.3% of the heat generation occurs within the fuel rod with the remaining 2.7% generated in the coolant.
l 221.43 You list three references for cladding conductivity. Since i
(4.4.2.2) they are not all identical, detail the equation you use to calculate conductivity and explain any deviation from the references.
4 221.44 Justify your statement that failure to include expanded flow l
(4.4.2.6) areas around bowed assemblies will result in less than 0.03%
3 error in the hot bundle flow. Detail these calculations completely and provide the basis and assumptions of your l
analysis.
221.45 In Section 4.4.2.7 correct your Section reference for a (4.4.2.7) discussion of core pressure drops and hydraulic loads during
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accident conditions.
" '.46 Why is there more than a 10% pressure drop difference between
. 4.4-30)
Greene County and BSAR-205 at the 157 inch icvel?
221.47 Provide values of K in equation 4.4-7.
Give a reference for (4.4.2.8) the values of K and indicate what experiments they were 1
derived from.
221.48 Equation 4.4-26 has,a division sign missing. Define the radiation (4.4.2.8) gap conductance (hrad) f r equati n
.4-28.
i 221.49 Amendment notation was omitted in this section. Provide a schedule (4.4.2.8) for completion of the Oconee 1 fuel exam and the submittal of the data reduction.
221.50 Provide a discussion of the role of pump monitors in preventing (16.2.3) core DNBR from decreasing below 1.3 if there were a loss of the reactor coolant pump (s),
1 221.51 Which transient is most limiting for DNBR? Provide che values (15.1) with respect to time of the DNBR for all transients which reach a DNBR of less than 1.39.
i 221.52 Does the 2.5% thermal uncertainty on the peak radial power (4.4.2.10) determination correspond to f "?
If not, indicate where and q
how this 2.5% uncertainty is applied.
13 Provide a sample case for TEMP and CHATA.
Include input to
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and output from the codes.
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.54 The.99 flow maldistribution factor for 205-fuel assembly
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edP) cores must be justified by experimental evidence. Topical Report BAR-10025P, received 9/29/76, is under review,.but i
our initial assessment is that this report does not justify a.99 flow maldistribution factor. The staff requires a e
I conservative value for flow maldistribution and the staff
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requires that the.95 value (used in previous B&W plant 1
applications) be used until a higher number is justified based on vessel model flow test data.
Our position is that in order to complete our review for 1
Preliminary Design Approval, you must justify in appropriate detail that your thermal-hydraulic analysis is acceptably conservative even with a.95 flow maldistribution factor.
t 221.55 Provide che data which indicates how fluid from each reactor
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i (4.4.4) coolant inlet pipe (in 4 or 3 pump operation) is distributed I
to each assembly. For example, will one assembly receive 73%
of its flow from reactor coolant pump 1 and 9% flow from each of the other three reactor coolant pumps.
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i 221.56 Section 4.4.1 states that flow stability is required during all 4
(Chapter 15) steady-state and operational transient conditions.
Examine each transient (excluding LOCA or other occurrences for which DNB is permitted) for possible flow instabilities.' Using HYTRAN, demonstrate the presence or absence of instability.
In parcicular, provide analyses for Steam Line Break and Excessive Heat Removal.
For those transients not analyzed by HYTRAN, provide a stability map such as presented in ANL-76-23 by M. Ishii as justification i
for not analyzing for stability.
Provide a comparison of HYTRAN to data over HYTRAN's range of application. There should be sufficient data points to insure j
a meaningful analysis.
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b 221.57 In Section 4.4.1 you state that the maximum temperature in a (Chapter 15) fuel pin will not exceed the melting temperature of that fuel at 1127. power during the life of the plant.
Provide the maximum t
fuel centerline temperature for any anticipated transients where the fuel centerline temperature may approach or exceed its limit.
For any transients not analyzed, provide a justification for their exclusion.
Include Steam Line Break in your analyses.
221.58 Provide a description of the method used to mount each of the 18 Reactor Diagnostic System sensors (e.g., clamped, drilled &
boted, magnetic, etc.).
221.59 Define " active" and " passive" sensors and note all differences.
(4.4.5.9)
Provide the justification for having onl/ passive sensors on the Reactor Coolant Pumps. Provide the justification for having only one active rather than two active sensors on the CRD mechanisms.
221.60 Provide a commitment to (1) fully define the gap closure (rod bow) rate for your prototype bundles, (2) determine by experiments the DNB effect that bounds the gap closure from
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(1), and (3) provide application of separate-effects bundle-bow 4
data to reactor transient analysis, i
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.21.61 Describe in detail your determination of the hot channel factors fq and fq", including a list and values for all physical parameters u.ed to obtain the hot channel factors. Give the basis for each
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I input (e.g., reference the fuel specifications or any other docu-mentation which provides manufacturing tolerance limits consistent with the values used in your derivation). How will you assure that the design values r'emain valid or that changes, if any, due to as-built deviations from design are properly accounted for in j
the hot channel factors?
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