ML20012E791
| ML20012E791 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, Catawba, McGuire  |
| Issue date: | 03/20/1990 |
| From: | Hood D Office of Nuclear Reactor Regulation |
| To: | Tucker H DUKE POWER CO. |
| References | |
| TAC-73769, TAC-73770, TAC-73771, TAC-73772, NUDOCS 9004060302 | |
| Download: ML20012E791 (10) | |
Text
.
March 20, 1990 Dockets Nos. 50-369, 50 370, 50 413 and 50 414 Mr. H. B. Tucker, Vice President Nuclear Production Department Duke Power Company 422 South Church Street Charlotte, North Carolina 28242
Dear Mr. Tucker:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION ON BAW.10173, MARK.BW RELOAD SAFETY ANALYSIS FOR MCGUIRE AND CATAWBA (TACS 73769/73770/73771 AND73772)
The NRC staff is reviewing the Cabcock and Wilcox topical report BAW.10173 which you submitted for application to the reload safety analyses for McGuIre and Catawba Nuclear Stations.
Enclosed is a request for additional information as a result of our review of the report.
Please respond to the enclosure expeditiously in order that we may ccaplete the review to the previously established schedule.
If you have questions, contact me at (301) 492-1442 or K. Jabbour at (301) 4921496.
The reporting and/or recordkeeping requirements contained in this letter affect 1
fewer than ten respondents; therefore, OMB clearance is not required under P.L.
96 511.
Sincerely, Darl S. Hood, Project Manager Project Directorate 113 Division of Reactor Projects - 1/II e.4 Office of Nuclear Reactor Regulation b
Enclosure:
, ogg As stated e> o 08 cc w/ enclosure:
t eo See next page mi i
88 DISTRIBUTION l
oo
'WENTTTF Ringram l ;o" NRC & Local PDRs DHood PDII.3~Rdg. File KJabbour oo CATAWBA PLANT FILE OGC(forinfo.only) p 1,g }
MCGUIRE PLANT FILE EJordan t-l SVarga ACRS(10)
Glainas g
(
.f:r1.....Rf #fr.t......,g P D g[l. 3
(
I.3 OFC
- LA:PD I-3
- PM:PDIl-3
- P
.V n..
,NAME :R ngram
- DHood: bid
- KJabbour
. Matt ews i
0 CIAL OR OPY
' Document Name:
CATAWBA /MCGUIRE RAI
Mr. H. B. Tu cker Duke Power Conpany ticGuire Nuclear Station i
CC:
Mr. A. V. Ca rr, Esq.
Dr. John M. Barry Duke Power Company Depa rtnent of Environnental Health P. O. Box 33189 Pecklerburg County 422 South Church Street 1200 Blythe Boulevard Charlotte, North Ca rolina 28242 Charlotte, North Carolina 28203 County Manager of Mecklenburg County Mr. Dayne H. Brown, Dimetor 720 East Fourth Street Departnent of Environmental, Charlotte, North Carolina 28202 Health and Natural Resources Division of Radiation Protection P.O. Box 27687 Mr. J. S. Warren Raleigh, North Carolina 27611-7687 Duke Power Company fluclear Production Departnent Mr. Alan R. Herdt, Chief P. O. Box 33189 Project Branch #3 Charlotte, North Carolina 28242 U.S. Nuclear Regulatory Comission 101 Marietta Street, NW, Suite 2900 J. Michael McGarry, III, Esq.
Atlanta, Georgia 30323 Bishop Cook, Purcell and Reynolds 1400 L Street, N.W.
lis. Ka ren E. Long Washington, D. C.
20005 Assistant Attorney General N. C. Departnent of Justice Senior Resident inspector P.O. Box 629 c/o U.S. Nuclear Regulatory Comission Raleigh, North Carolina 27602 Route 4. Box 529 l
Hunterville, North Carolina 28078 Regional Administrator, Region II U.S. Nuclear Regulatory Comission 101 Marietta Street, N.W., Suite 2900 Atlanta, Georgia 30323 l
Ms. S. S. Ki 1 born l
Area fianager, Mid-South Area ESSD Projects Westinghouse Electric Corporation NNC West Tower - Bay 239 P. O. Box 355 Pittsburgh, Pennsylvania 15230 l
i i
J
p
[
Mr. H. B. Tucker Catawba Nuclear Station rule Power Company CC:
A. V. Carr, Esq.
Cuke Power Company North Carolina Electric Membership Corp.
422 South Church Street 3400 Sumner Boulevard Charlotte, florth Caroitna 28242 P.O. Box 27306 Raleigh, North Carolina 27611 O liichael McCarry,111. Esq.
Bishop, Ceck, Purcell and Feynolds Saluda River Electric Cooperative.
la00 L Street, N.W.
Inc.
Washington, D. C.
20005 P.O. Box 929 Laurens, South Carolina 29360 North Carolina MPA-1 t
Suite 600 Senior Resident inspector Reute 2, Box 179N 3100 Smoketree Ct.
York, South Carolina 29745 P.O. Box 29513 Raleigh, t! orth Carolina 27626-C513 Regional Acministrator, Region II Ms. S. S. Kilborn U.S. Nuclear Regulatory Comission 101 Marietta Street, NW, Suite 2900 Area Manager, Mid-South Area Atlanta, Georgia 30323 ESSD Projects Westinghouse Electric Corp.
Mr. Heyward G. Shealy, Chief MNC West Tcwer - Bay 239 Bureau of Radiological Fealth P.O. Box 355 South Carolina Department of Health fittsburgh, Pennsylvania 15230 and Environnental Control 2600 Bull Street County Manager of York County Colunt>ia, Scuth Carolina 29201 York County Courthouse York, South Carolina 29745 Ms. Karen E. Long Assistant Attorney General Richard P. Wilson, Esq.
N.C. Department of Justice Assistant Attorney General P.O. Box 629 S.C. Attorney Gener&l'5 Office Raleigh, North Carolina 27602 P.O. Box 11545 Columbia, South Carolina 29211 Mr. Robert G. Morgan Nuclear Production Department Pieomont Municipal Power Agency Duke Power Company 100 Memorial Drive P.O. Box 33189 Greer, South Carolina 29651 Charlotte, Mcrth Carolina 28241 Mr. Alan R. Herdt, Chief Project Branch #3 U.S. Nuclear Regulatory Commission 101 Mariette Street, NW, Suite 2900 Atlanta, Ceorgia 10323
^
I p
ENCLOSURE RE0 VEST FOP ADDITIONAL INFORMATION ON BAW-10173P 1.
Many places in topical report BAV-10173P have statements indicating certain calculations are performed on a cycle by cycle basis, and, i
therefore, no result is provided.
Please confirm that a cycle-specific analysis, including the first cycle using the Mark-BW fuel, will be submitted.
2.
The Mark-BW fuel design differs from the OFA fuel in flow area and the mixing vane spacer orid pressure drop, and having a non-mixing grid in the first intennediate grid.
In a mixed core with both Mark-BW and 0FA fuel assemblies, the combined effect of these hydraulic design differences results in a flow diversion into and out of the Mark-BW fuel assemblies at elevations below and above the core midplane. respectively.
Section P.3 indicates that the flow diversion is less than 2 ft/sec, which is used as a maximum crossflow criterion for design evaluation.
(a) Discuss how the magnitude of isothermal diversion crossflow resulting from the hydraulic characteristic differences between the two fuel designs is obtained.
Is the maximum flow diversion of 2 ft/sec a bounding value of the isothermal crossflow?
(b) Diseass how crossflow enhancement by this isothermal flow diversion effect is treated in the thermal-hydraulic design evaluation.
3.
In discussing the hydraulic compatibility between the Mark-BW and the OFA fuel designs. Section 2.2.2 indicates that the TFTR test demonstrates the total pressure drop of the Mark-BW fuel to be approximately 5 percent less than that of the OFA fuel.
It further indicated that specific transition core effect will be addressed on a cycle by cycle basis.
However. Section 2.2.3 states that the total pressure drop difference between the two designs is sufficiently small (5 percenti that no transition core penalty is indicated.
i l
2 l
la) Clarify what is intended to be done regarding transition mixed core penalty.
(b)
In light of the hydraulic differences discussed in Question 2, a mixed core penalty should be assigned to the Mark-BW or 0FA fuel assemblies if the analysis of the transitional mixed core is performed with a homogeneous core assumption.
Provide a detailed i
description on the specific method used to determine the magnitude of the mixed core penalty, and how it will be applied in license calculation. Also provide the specific mixed core penalty value to be applied to the Mark-BW or OFA fuel assemblies.
4 Provide specific information to support the statement in Section 2.0 that the nuclear design analysis shows that the Mark-BW fuel has neutronic behavior similar to the STD assemblies, with essentially the same moderator. coefficient, and that the differences between the Mark-BW and the OFA are slight.
j 5.
The RELAPS/ MOD 2-B&W and LYNXT codes used in the safety analyses have a number of user inputs and option selections.
To ensure that each code is used within its capabilities and with proper input, information justifying e
i all selected options and input data, including defaults, should be presented. List all options to be used for analyses of the transients and accidents, including steamline break, turbine trip, main feedline rupture, complete loss of forced reactor coolant flow, locked rotor, uncontrolled RCCA withdrawal, and RCCA ejection, etc.
The list of options should be complete and the bases for each chosen option clearly provided for each transient or accident.
6.
(a)
In the RELAPS/M002-Bt.W modeling for the steamline break analysis, I
what are the values used for the junction flow area mixing and faulted /unfaulted reactivity feedback weighting factor?
l l
. (b) Provide a description of the core modeling using LYNXT for the subchannel DNB analysis for the steam system piping failure.
7.
Section 4.1.5.1 on the analysis of steam system piping failure indicates that a 3-D statepoint analysis is used to confirm that the reactivity employed in the kinetics analysis was larger than the reactivity would be when calculated including the effects of void formation, etc.
It further indicated that the 3-D statepoint is evaluated for each cycle to confirm the conservatism.
(a) Have the 3-D statepoint analysis and comparison been done for the cycle reloaded with Mark-BW fuel? What is the result?
(b) What would be done if the cycle-dependent calculation does not confirm the conservatism stated?
8.
(a) What is the worst single failure assumed in the steam system piping failure analysis? What is the basis for this choice?
(b) The main steamline break analysis described in Section 4.1.5 assumed offsite power to be available. This is inconsistent with the guideline of SRP 15.1.5 which requires the assumption of loss of offsite power.
Provide justification for your assumption of offsite power availability in the analysis, or perform a reanalysis of the main steamline break with loss of offsite power and worst single failure. Provide the reanalysis resultr, of all important parameters including minimum DNBR vs. time and a comparison with the results of the existing analysis assuming offsite power available.
9.
Without giving actual analysis results Section 4.2.3 indicates that the result of the turbine trip analysis is markedly less than that produced by nore limiting DNBR events in the reduced coolant flow category. Since turbine trip is a limiting transient in the decrease in secondary heat removal category, provide a detailed analysis with the Mark-BW fuel loading design for staff review including inputs, assumptions and results on important parameters, such as power level, pressurizer pressure and
~
\\;
water volume, core inlet and average temperatures, flow and DNBR as a function of time.
10.
Section 4.2.8 indicates that the main feedline rupture evaluation for the McGuire/ Catawba units with B&W reload fuel need only consider those aspects of the transient that could be affected by differences in fuel and fuel cycle design, and that so long as the decay heat levels are within the bounding values represented in the reference analyses, the acceptance criteria applicable to the main feedline break will continue to be met for the reload core.
(a) List fuel dependent parameters which could affect the consequence of a main feedline break, and provide the bases justifying the decay heat level to be the only parameter affecting the analysis.
(b) Provide data which ascertain the decay heat levels of the reload cores with the Mark-BW fuel to be within the bounding values in the reference analyses or (c) Perform a reanalysis of main feedline rupture for the new fuel L
loading design.
Provide the reanalysis results including the codes l
used in the analysis, assumptions and their bases, and results on important parameters such as power, core heat flux, pressurizer pressure and water volume, faulted and intact loop reactor coolant temperatures, end steam generator pressure and water mass.
- 11. The following items are related to the locked rotor analysis described in Section 4.3.3:
(a) No mention is made regarding the assumptions on the status of offsite power and unaffected reactor coolant pumps, single failure, and reactivity parameters such as the maximum or minimum values used for Doppler and moderator coefficients to ascertain that the limiting
p case is analyzed.
Please provide the information regarding assumptions and bases.
(b)
It is indicated that no " credit" is taken for pressure reducing effect of the pressurizer relief valves, pressurizer spray, steam dump or controlled feedwater flow after plant trip.
Since neolect of the pressure reducing effect is non-conservative with regard to the DNB consideration, they should be considered in the analysis.
Provide an analysis considering the pressure reducing effects and the assumptions in item a.
(c)
It is indicated that DNB is assumed to occur (initially?) in the core.
Provide the initial conditions for the average and hot channels.
(d) Provide a description of calculating the number of fuel pins in DNB.
(e) Clarify whether the cladding surface temperature shown in Figure 4.3.3-6 is the hot pin or core average pin temperature.
12.
Section 4.4.1 discusses the RCCA withdrawal from low power analysis and indicated that, at no time do the minimum DNBR fall below the limiting value, or the primary and secondary system pressures exceed the limits of general design criterion 15.
Please provide the time history curves for the minimum DNBR, and primary and secondary pressures.
13.
The analysis of RCCA withdrawal at power discussed in Section 4.4.2 assumed the maximum positive reactivity insertion rate of 75 pcm/see to be the worst case.
As stated in the staff SER on BAW-10169, Section 15.4.2 of SRP requires that the analysis should consider various reactivity insertion rates from very low to maximum possible for the control system and the fuel and moderator feedback reactivity coefficients covering the
)
F.
. r range expected throughout the cycle.
Provide analysis results for a range of reactivity cases.
14 Section 4.4.3.1 very briefly discussed the analysis method of the dropped RCCAs as (1) using the FLAME 3, NOODLE and P0007 to calculate dropped rod and control group worths, and power Doppler defect and power peaking, and (2) using LYNXT to calculate DNER.
It also provides a discussion of the results for 3 different conditions, and indicates that the ct.lculations are performed on a cycle by cycle basis.
(a) Provide a detailed description of method of analysis including code interface, inputs, assumptions, and bases of the assumptions for each code used.
(b) Provide the results of important parameters, such as nuclear power, i
core heat flux, coolant temperature, pressurizer pressure, and DNBR, etc., as a function of time.
(c)
Items (a) and (b) should be repeated for the following cases:
a l
dropped RCCA event in the automatic rod control mode, and a
(
i statically misaligned RCCA.
f I
i 15.
For the single RCCA withdrawal case discussed in Section 4.4.3.1, the i
discussion on the method of analysis is brief and not clear.
(a)
It indicated that the thermal calculations are perfor..ed with the l
LYNX code series.
Please be specific as to whether the steady state LYNX 1/ LYNX 2 codes or the transient LYNXT code is used.
I l
l l
(b) Provide a description of code interface, input and assumptions j
(including bases for the assumptions) for each code.
L i
L l
l
F t.
I I-7 l
(c) Provide the cuantitative analysis results of the important parameters, such as power, thermal flux, coolant terperature, pressure and DNBR as function of time. Also, provide the resulting percentage of fuel rods experiencing DND.
- 16. Section 4.4.4 indicates that since the bounding moderator and Doppler reactivity coefficients for the startup of an inactive coolant pump at incorrect temperature event are also used for the limiting accident (steam line break). the cycle-specific values of these parameters will continue to be compared to the bounding values.
This seems to imply that the inadvertent startup of the inactive pump event is bounded by the steamline break. However, the steamline break is a Condition IV event whereas this event is a Condition il event, and therefore the acceptance criteria are not the same. Justify why this event is bounded by the steamline break event or provide an analysis for this event.
i 1
.