ML20039G403
| ML20039G403 | |
| Person / Time | |
|---|---|
| Site: | 05000481 |
| Issue date: | 09/18/1974 |
| From: | Stello V US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Moore V US ATOMIC ENERGY COMMISSION (AEC) |
| Shared Package | |
| ML111090060 | List:
|
| References | |
| FOIA-80-515, FOIA-80-555 NUDOCS 8201180164 | |
| Download: ML20039G403 (7) | |
Text
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UNITED STATES rp - f, ATOMIC ENERGY COMMISSION
/ d WASHINGTON. D.C. 20543
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Docket io 5 -481 SEP 1 8 1374 I
V. A. Moore, Assistant Director for Light Water Reactors Gro0p 2 L INITIALQUESTIONSFORBSAR(Chapters 4,7,and15)
Plant Name:
B-SAR-241
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Licensing Stage:
Q1 Docket Nc.:
50-481 Responsible Branch LWR 2-3 and Project Manager:
D..K. Davis Technical Review Branch Involved:
Core Perfonnance Branch Requested Completion Date:
September 13, 1974
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Description of Review:
Initial Questions Enclosed are first round questions of the Reactor Physics Section of the Core Performance Branch relating to Chapters 4, 7, and 15 of BSAR.
M Victor Stello, J., Assistant Director for Reactor Safety
, Directorate of Licensing
Enclosure:
Questions cc:
S. Hanauer F. Schroeder A. Giambusso W. Mcdonald A. Schwencer D. Davis D. Ross
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L. Rubenstein P. Check D. Houston R. Bottimore D. Basdekas l
L. Kopp-l L. Chandler l
E. leins l
l W. Lanning khkII8fj MADDEkso64 e10403 325 ppg
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242-1 242.0 REACTOR PHYSICS SECTION - CORE PERFORMANCE BRANCH 242.1 Specify the fuel enrichment for each of the fuel regi'ons shown (4. 3.2.1 )
in Figure 4.3-1.
242.2 Specify the poison content of the Burnable Poison Rod Assemblies (4.3.2.1)
(BPRA) shown in Figure 4.3.-2.
242.3 Specify the poison content of the Control Rod Assemblies (CPA)
(4. 3.2.1 )
and the Axial Power Shaping Rod Assemblies (APSRA) used in the analyses of Sections 4.3 and 15.
242.4 Specify the mass corresponding to 100% total plutonium in (4.3.2.1 )
Figure 4.3-4, and plot plutonium-239 content as a functiun of burnup.
242.5 Show the axial and radial power sha (4.3.2.2.1) of-life (BOL) and end-of-life (EOL) pes expected at beginning-for normal transients such as load following, xenon buildup, xenon decay or redistribution, and xenon oscillation control.
242.6 Show the expected axial power shapes for various power levels (4.3.2.2.1),
at BOL and EOL.
242.7 Discuss the design basis power maneuver.
Show that it is (4.3.2.2.1) the worst case power maneuver allowed.
Show the axial and
' radial power profiles for this maneuver.
Discuss the behavior of the peaking factors during this maneuver.-
242.8 Present the limiting nadial and axial power distributions used (4.3.2.2.2) to define the offset and total peaking factor limits.
Describe the reactor conditions corresponding to these limiting power.
distributions.
Discuss any uncertainties or errors which may be associated with the calculated limiting distributions.
Present experimental data which confirms the analysis.
242.9 Present plots of peaking factor versus axial offset, and core (4.3.2.2.3) power imbalance versus reactor power, showing the envelopes of allowable operation.
4 242.10 Describe the procedure used to produce the envelop lines on (4.3.2.2.3) the peaking factor offset map.
Provide the criteria and technique used.
242.11 Provide a list of power maneuvers that were analyzed to provide (4.3.2.2.3) the peaking factor offset plot.
Provide the assumptions upon which these maneuvers are based (e.g., rod position, fuel burnup, power level, soluble poison concentration, etc.).
Indicate the limiting maneuvers.
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242-2 242.12 Under what conditions, if any, is operator action considered (4.3.2.2.3) in calculation of the peaking factor-offset map? To the extent that Axial Power Shaping Rods are taken into account in the peaking factor map, are favorable or unfavorable actions of the operator considered and to what extent?
If unfavorable actions are.not included, why are,they not?
242.13 Discuss all adjustments to be made to the calculated offset (4.3.2.2.3) limits to account for instrumentation uncertainty, calibration, excore detector bias, etc.
Give the values of the offset trip set points and explain how they were determined.
242.14 Show the limiting axial and radial power shapes that are (4.3.2.2.6) anticipated during normal operation.
Show how this data supports des'ign axial and radial peaking factors of 1.55.
Provide a summary of recent operating experience on measured values of radial peaking' factors, axial peaking factors, and tilt for a variety of operatirg conditions.
242.15 Derive a chart that depicts ;he time variation of peaking (4.3.2.2.6) factor on return to full power from a typical maneuvering transient.
Indicate the time at which a limiting peaking factor is achieved during and following the return to power.
Present any experimental verification available relative to the time variation of peaking factor after return to full power.
242.16 Discuss the effect that a 9-inch deviation in CRA position (4. 3. 2.2. 7) compared to group position would have on the axial flux shape at various power levels.
Is this effect included in the axial flux shapes considered in the accident analyses?
242.17 How much does an APSRA have to get out of alignment with (4.3.2.2.7) the group average before an alarm occurs?
Is this effect included in the axial flux shapes used in accident analyses?
242.18 What is the change in radial power distribution due to a (4.3.2.2.7).
dropped rod in the reactor? Show radial power distributions for the cases of single dropped rods at various radial positions (including the center of the reactor and for varying fuel burnups).
Present any measured data available l
to substantiate the calculation.
Show excore and incore instrumentation response to the dropped rods.
242.19 Describe the maximum, non-detectable changes in the design power l
(4.3.2.2.7)
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distribution which could occur as a result of fuel misloading.
Discuss whether or not these changes will exceed fuel design limits.
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x.u a.waaaww 242-3 242.20 Discuss quadrant power tilt including tilt l'mits and alarms, i
(4.3.2.2.7) excore and incore instrumentation response.
Describe,the technique used to calibrate excore detectors to incore detectors.
242.21 Discuss calculation of excore detector responses for various (4. 3. 2. 2.7 )
normal and transient reactor conditions including representative rod complements, power levels, fuel burnup, and xenon conditions (stableandtransient).
Show that the excore detector response will protect the fuel and reactor for normal and anticipated transient conditions.
Discuss the computer code (or calculational technique) used to calculate the excore detector responses.
Describe all parameters (geometric, material etc.) considered in the calculation.
242.22 Present comparisons betweer. measured and calculated' excore (4. 3. 2. 2. 7 )
detector responses for the various reactor conditions given in Question 242.21.
How does excore detector response change with reactor coolant inlet temperature in the power ope. rating range?
Discuss the effect of rod shadowing on excore detector response.
242.23 Which moderator coefficients are being shown in Figures 4.3-24 (4. 3. 2. 3) -
and 4.3-25? What are the coolant boron concentrations and the CRA configurations for Figures 4.3-24 and 4.3-257 242.24 Explain fully the meaning of the " threshold value of moderator (4.3.2.3) temperature coefficient for azimuthal instability" for 50%
flatness and for 25% flatness in Table 4.3-5.
242.25 Based on the Oconee 14 2 startup tests, what value for the (4. 3. 2. 3) uncertainty can be applied to calculated moderator temperature coefficients, power coefficients, control rod worths, and power distributions?
242.26 Su ply the foilowing graphs:
(4.3.2.3)
(1 Doppler coefficient vs fuel temperature for BOL and E0L.
(2 Moderator temperature coefficients vs noderator temperature for BOL and E0L, with and without equilibrium Xe and Sm.
for representative boron concentrations and rod configurations.
(3) Moderator pressure coefficient vs coolant pressure for BOL and E0L.
(4) Power coefficient vs reactor power for BOL and EOL. -
9 (5) Fuel thermal expansion coefficient vs fuel temperature at BOL and E0L.
(6) Moderator void coefficient vs percent void for BOL and EOL.
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242-6 242.45 Discuss the analtyical method used to calculate control rod (4. 3. 3) worth.
Discuss the allowance for isotopic resonance interference and overlap in the control' rod.
242.46 Discuss in detail the analytical technique used to calculate (4.3.3) the a fuparameters of the Barnable Poison Rod Assembly (BPRA) as nction of fuel burnup.
Discuss any comparisons made between calculations and experiments.
242.47 Discuss the thermal feedback effects modification to the (4.3.3)
PDQ07 computer code.
242.48 Compare calculated and measured rod configurations and boron (4.3.3) concentrations for criticality in operating power reactors.
Compare calculated and measured values for operating power reactors for a) hot and cold ejected and stuck CRA worths, b) CRA group worths.
242.49 Compare calculated and measured radial and axial power distributions (4.3.3) for opertting) reactors for a) various power levels, including full power, b various rod complements and soluble poison concentrations, c) and various fuel burnups, including both BOL and E0L.
242.50 Describe any comparisons with experimental data that serve (4. 3. 3) to validate the power dist.ribution calculations within the fuel assemblies.
Discuss measurement uncertainties.
242.51 Compare calculated and measured reaction rates for uranium-235, (4. 3. 3) uranium-238, and plutonium-239.
Discuss how these results, or similar rates, are..used in the B&W Standard 241 reactor design.
Discuss measurement uncertainties.
242.52 Discuss the agreement between calculated and measured fuel (4. 3. 3) depiction.
Discuss any measurement uncertainties.
242.53 Discuss the change in boron reactivity worths as a function of (4.3.3) fHb burnup and boron concentration.
Compare measured and calcuMted changes in boron reactivity worths for operating reactors. Discuss measurement uncertainties.
242.54 Describe fully th'e methods used for calculating the scram (4. 3. 3) reactivity as a function of time-after-scram signal including consideration for Technical Specification scram times, stuck red (s), power level and shape, time in cycle, and any other parameter important to CRA group worth and axial flux shape.
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v 242-7 242.55 Describe the computer calculation of the deoletion function (7.8.2.2) for the incore detectors.
Describe all adjustments made to incore detector signals by use of the computer.
Justify the statement that calibration of incore detectors is not required, including any verification measurements made.
242.56 Describe all corrections made to the excore power range (7.8.2.2) instrumentation signals before the signals are used to determine axial offset and tilt.
242.57 Describe the bases for regrouping control rods during core life.
(7.7.1.3)
Discuss the possible errors which could result, especially the grouping of CRA(s) with the wrong group.
Present safety analyses of the possible errors, and prove that they can be detected and' safely controlled.
242.58 How have unc'ertainties been included in the moderator and (15.1, 15.1.3, Doppler coefficients used in the control rod accidents?
15.1.18)
Are these coefficients flux-volume weighted?
242.59 The paragraph on page 15.1-3 refers to a fuel damage criterion (15.0) of 210 cal /gm for the rod ejection accident.
This appears to be inconsistent with the criterion of 280 cal /gm stated in Section 15.1.18.1.
Explain.
242.60 Discuss the derivation of the BOL and EOL rod worths in (15.1)
Table 15.1-2.
How have uncertainties been included?
242.61 Show the changes in minimum DNBR and maximum kw/ft for the (15.1.1, 15.1.2 control rod withdrawal and misoperation accidents.
15.1.3) 242.62 Describe the startup and full power control rod withdrawal (15.1.1, accidents at the BOL.
Discuss the time-in-cycle when the 15.1.2) limiting cases occur.
242.63 Discuss how the power distributions or peaking factors change (15.1.1,15.1.2, with time in the control rod withdrawal and misoperation 15.1.3) accidents.
242.64 Since the operator has provisions to manually control single (15.1.3) rod motion, the single rod withPawal accident should be analyzed as an anticipated event.
Specifically, provide information justifying the selection of worst-case conditions of time-in-cycle.
- power level, power distribution, peaking factors, control rod worth, control rod position, etc.
Justify any differences between these selected values and those used in the rod ejection.
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2A2 8 242.65 Provide a plot of maximum fuel centerline temperature as (15.1.3) function of time for the control rod misoperation accident.
Demonstrate that the worst-case initial conditions have been analyzed.
242.66 Justify that the following control rod mis' operation accidents (15.1.3) do not have to be analyzed:
(1) Inadvertent withdrawal of two or more contro] rods at the
.:__..... same time.
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(2) Leaving one or more rods behind (i.e., stuck rods) during rod bank withdrawal.
(3) Insertion of a rod bank with one or more bank rods stuck.
242.67 Show the effects of fuel loading errors on power distribution (15.1.15) and stability indices.
In addition to the fuel loading errors described in Section 15.1.15, analyze :
(1) interchanges of fuel assemblies with and without burnable (2)poisonrods, the loading of a, fuel assembly with the wrong number of burnable poison rods or the wrong poison content in the burnable poison rods, and (3) the leading of a fuel assembly with empty or partially empty Zircaloy tubes.
Discuss the ability of the nuclear instrumentation to detect each postulated misloading.
Describe the startup tests which will specifically identify loading errors.
242.68 Provide the names of and describe the computer code used to (15.1.18) analyze the rod ejection accident.
Compare the control rod ejection analysis to that discussed in Regulatory Guide 1.77, specifically discussing any differences between the two analysis techniques.
242.69 Describe how the total nuclear peaking factor used in the rod (15.1.18) ejection analysis was derived with respect to Table 4.3-3 and 4.4-1.
Is any nuclear uncertainty included? Justify the use of the pre-transient total peaking factor for the entire ' rod ejection analysis.
242.70 Describe in more deta'il the compariso of space-dependent (15.1.18) and. point kinetics results on fuel enthalpy given in Table 15.1.18-3.
. Include any differences in initial assumptions and in calculated fuel enthalpies.
From the trend of the table, it appears that a 0.65". rod may give more conservntive results using WIGL2 than point kinetics.. Is this true?
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