ML20141J751
| ML20141J751 | |
| Person / Time | |
|---|---|
| Issue date: | 08/20/1997 |
| From: | Joshua Wilson NRC (Affiliation Not Assigned) |
| To: | Reda R GENERAL ELECTRIC CO. |
| References | |
| TAC-M97490, TAC-M99069, NUDOCS 9708210200 | |
| Download: ML20141J751 (20) | |
Text
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _
August 20, 1997 Mr. Ralph J. Reda, Manager Fuel and Facility Licensing GE Nuclear Energy P. O. Box 780, MC J26 3901 Castle Hayne Road Wilmington, NC 28402
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION FOR GE TOPICAL REPORTS NEDC-32601P (TAC NO. M97490) AND NEDC-32694P (TAC NO. M99069)
REFERENCES:
1.
GE Topical Report, NEDC-32601P, " Methodology and Uncertainties for Safety Limit MCPR Evaluations,"
GE Nuclear Energy, December 1996.
i 2.
GE Topical Report, NEDC-32694P, " Power Distribution Uncertainties for Safety Limit MCPR Evaluations,"
GE Nuclear Energy, January 1997.
Dear Mr. Reda:
l
_The staff is reviewing your submittals dated December 13, 1996 and' June 10, o
1997, regarding the methodology and uncertainties for safety limit MCPR evaluation and the power distribution uncertainties for safety limit MCPR
-evaluation, respectively. The staff concludes that additional information is needed before it can complete its review. contains a request for additional information (RAI) related to the staff's review of NEDC-32601P,
" Methodology and Uncertainties for Safety limit MCPR Evaluations." Enclosure 2 contains an RAI related to the staff's review of NEDC-32694P, " Power Distribution Uncertainties for Safety Limit MCPR Evaluations."
You are requested to provide responses to these RAls within 30 days of the date of-this letter.
If you need further clarification concerning this request, please contact Dr. Tai Huang at (301) 415-2867-.
Sincerely, 4
James H. Wilson, Senior Project Manager Generic-Issues and Environmental Projects Branch Division of Reactor Program Management Office of Nuclear Reactor Regulation J
Enclosure. As stated oW cc w/ atts: See next page
- oEl DISTRIBUTION
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Docket File PUBLIC SRXB r/f PGEB r/f MCase DMatthews GCENTRAL5 GHolahan TCollins LPhillips JHWilson THuang 4 -Document Name: GE-MCPR.RAI
- See previous concurrence OFC PGEB O % J (A)SC:PGEB SRXB/DSSA SC:SRXB/DSSA C:SRXB/DSSA NAME-JHWils[n:sw-MJCase h THuang*
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DATE-08/h /97 08/20/9Y k 08/19/97 08/19/97 08/19/97 OFFICIAL RECORD COPY
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s August 20, 1997 Mr. Ralph J. Reda, Manager Fuel and Facility Licensing GE Nuclear Energy P. O. Box 780, MC J26 3901 Castle Hayne Road Wilmington, NC 28402
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION FOR GE TOPICAL REPORTS NEDC-3260lP (TAC NO. M97490) AND NEDC-32694P (TAC NO. M99069)
REFERENCES:
1.
GE Topical Report, NEDC-3260lP, " Methodology and Uncertainties for Safety Limit MCPR Evaluations,"
GE Nuclear Energy, Deceraber 1996.
2.
GE Topical Report, NEDC-32694P, " Power Distribution Uncertainties for Safety Limit MCPR Evaluations,"
GE Nuclear Energy, January 1997.
Dear Mr. Reda:
1 The staff is reviewing your submittals dated December 13, 1996 and June 10, 1997, regarding the methodology and uncertainties for safety limit MCPR evaluation and the power distribution uncertainties for safety limit MCPR evaluation, respectively. The staff concludes that additional information is i
needed before it can complete its review.
Enclosure I contains a request for additional information (RAI) related to the staff's review of NEDC-32601P,
" Methodology and Uncertainties for Safety limit MCPR Evaluations." Enclosure 2 contains an RAI related to the staff's review of NEDC-32694P, " Power Distribution Uncertainties for Safety Limit MCPR Evaluations."
You are requested to provide responses to these RAls within 30' days of the date of this letter.
If you need further clarification concerning this request, please contact Dr. Tai Huang at (301) 415-2667.
Sincerely, James H. Wilson, Senior Project Manager Generic Issues and Environmental Projects Branch Division of Reactor Program Management Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/ atts: See next page DISTRIBUTION:
Docket File PUBLIC SRXB r/f PGEB r/f MCase DMatthews CENTRAL GHolahan TCollins LPhillips JHWilson THuang Document Name: GE-MCPR.RAI
- See previous concurrence OFC PGEB O W )
(A)SC:PGEB SRXB/DSSA SC:SRXB/DSSA C:SRXB/DSSA NAME JHWils[n:sw MJCase h THuang*
LPhillips*
TCollins*
DATE 08/h /97 08/20/9h 08/19/97 08/19/97 08/19/97 0FFICIAL RECORD COPY
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t UNITED STATES j
NUCLEAR REGULATORY COMMISSION 2
WASHINGTON, D.C. 30666 4001 49*****
,o August 20, 1997 Mr. Ralph J. Reda, Manager Fuel and Facility Licensing GE Nuclear Energy P. O. Box 780, MC J26 3901 Castle Hayne Road Wilmington, NC 28402.
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION FOR GE TOPICAL REPORTS NEDC-32601P (TAC NO. M97490) AND NEDC-32694P (TAC NO. M99069)
REFERENCES:
1.
GE Topical Report, NEDC-32601P, " Methodology and Uncertainties for Safety Limit MCPR Evaluations,"
GE Nuclear Energy, December 1996, 2.
GE Topical Report, NEDC-32694P, " Power Distribution Uncertainties-for Safety Limit MCPR Evaluations,"
GE Nuclear Energy, January 1997.
Dear Mr. Reda:
The staff is reviewing your submittals dated December 13, 1996 and June 10, 1997, regarding the methodology and uncertainties for safety limit MCPR evaluation and the power distribution uncertainties for safety limit MCPR evaluation, respectively.
The staff concludes that additionai information is needed before it can complete its review.
Enclosure I contains a request for additional information (RAI) related to the staff's review of NEDC-3260lP,
" Methodology and Uncertainties for Safety limit MCPR Evaluations." Enclosure 2 contains an RAI related to the staff's review of NEDC-32694P, " Power Distribution Uncertainties for Safety Limit MCPR Evaluations."
You are requested to provide responses to these RAIs within 30 days of the date of this letter.
If you need further clarification concerning this request, please contact Dr. Tai Huang at (301) 415-2867.
Sincerely, Ja s H. Wilson, Senior Project Manager Generic Issues and Environmental Projects Branch Division of Reactor Program Management Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/ atts:
See next page
s GE Nuclear Energy cc:-
Gary L. Sozzi, Manager Technical and Modification Services GE Nuclear Energy 175 Curtner Avenue San Jose, CA 95125 George B. Stramback GE Nuclear Energy 175 Curtner Avenue San Jose, CA 95125 James F. Klapproth GE Nuclear Energy P.O. Box 780 Wilmington, NC 28402 l
l I
ENCLOSURE I RE0 VEST FOR ADDITIONAL INFORMATION FOR GE TOPICAL REPORT NEDC-32601P
" METHODOLOGY AND UNCERTAINTIES FOR SAFETY LIMIT MCPR EVALUATIONS" I.
Process Comouter Uncertaintiet 1.
What is the variation of the weighing coefficients, used in evaluating the uncertainty in the Equation (2-3) inlet subcooling, over reactor statepoint (e.g., flow, power, subcooling, pressure) and in what sense are these coefficients conservative?
2.
How does the channel flow uncertainty account for channel bulge and non-uniform crud / corrosion build up on the fuel rods?
3.
Provide justification for neglecting the bias in the core pressure drop calculation in Table 2.2.
4.
In Table 2.2, the fact that only the BWR6 data is negative suggests that the uncertainty is plant dependent. Also, the fact that 50%
of the data is outside the one-sigma interval suggests that the data is not normal.
Provide justification for treating this uncertainty as normally distributed in the SLMCPR analysis.
5.
How is the uncertainty in the bypass flow included in the uncertainty analysis?
II.
R-Factor Uncertainty 1.
What specific fuel designs were the TGBLA-to-HCNP pin power comparisons performed and how do these comparisons cover the intended range of GE BWR fuel designs?
2.
How is the uncertainty in the TGBLA exposure calculation accounted for in the determination of the local pin power peaking factor uncertainty?.
3.
Provide justification for the weight used to combine the standard deviations of Table 3.1 and determine the local peaking model uncertainty.
4.
Is the enrichment tolerance for rods with enrichment less than 3.5%
greater than for rods with enrichment greater than 3.5% and, if so, how is the resulting increased uncertainty accounted for in the local peaking model uncertainty?
5.
Provide justification for the assumption that the effect of rod position (e.g., due to rod bowing) has a negligible effect on the local power peaking.
e
. 6.
Describe the measurements of Table 2.2 and how they determine the single-phase friction factor uncertainty.
7.
The comparisons of Table 3.3 irdicate that the local power peaking factor uncertainty is larger at the top of the fuel bundle. How is this apparent spatial dependence of the peaking factor uncertainty accounted for in the SLMCPR evaluation?
8.
Do the TGBLA-to-HCNP comparisons of Table 3.1 indicate a larger uncertainty in the high powered fuel rods and, if so, how is this accounted for in the SLMCPR?
9.
How is the uncertainty in the fuel density accounted for in determining the manufacturing uncertainty?
- 10. How is the uncertainty in the local voids and exposure accounted for in the local peaking factor uncertainty?
11.
In the improved R-Factor method, how is the uncertainty in the bundle-average void and exposure distributions used in performing the integration of the local R-Factor accounted for in the determination of the R-Factor ut. certainty?
- 12. The local power. peaking modeling uncertainty provided on p. 3-2 is greater than the value given in the first sentence of Section 3.1.4.
Please explain this apparent incunsistency.-
- 13. The power peaking uncertainties in neighboring fuel rods are generally correlated and, consecuently, can not be taken to be independent and random as assumec in the-calculations of Section 3.2.
Provide an estimate of the R-Factor uncertainty that' recognizes the correlation of the uncertainties in neighboring fuel rods.
III. SLMCPR Evaluation Methodoloav 1.
In Section 4.3, it is stated that the SLMCPR values of Table 4.1 are for the revised methodology and the present (larger) uncertainties, while Table 4.1 indicates that the revised methodology is evaluated with the revised uncertainties.
Please explain this apparent inconsistency.
2.
Provide the basic mathematical definition of Wcore (rather than the mathematical result of the integration given in Section 4.2).
3.
Provide the definition of the MCPR Importance Parameter (MIP) of Figure 4.4.
4.
What pin power distribution is assumed in the definition of Wcore and what is the effect of this assumption?.
e
. 5.- Describe the selected 100 nominal control rod pattern cases of Figure 4.4.
How do these cases and the Figure 4.4 comparisons of the nominal and limiting MIPs accommodate off-nominal operating statepoints. Provide justification for the assumption that these 100 patterns bound the operating statepoints.
6.
Provide specific details describing how the limiting control rod pattern will be selected in a typical reload core determination of the SLMCPR. What quantitative criteria will be used to select the limiting pattern? What statepoints (e.g., power, flow, exposure) will be included in the determination of the limiting pattern? How will the Figure 4.4 comparisons be used to confirm the limiting pattern selection? Will the data base of these 100 nominal cases be expanded or updated?
Will the Wcore parameter be monitored by the process computer to 55ure that the design SLMCPR limiting power distribution bounds operating power distribution?
all the MIP calculations of Figure 4.4 assume the same o
l uncertainties? If not, provide the nominal-to-limiting MIP l
comparisons separately for each set of uncertainties and explain the effect of this inconsistency on the conclusions drawn from the i
Figure 4.4 comparisons.
9.
The elimination of fuel bundles from the SLMCPR calculation using the criteria P, < AP results in an underestimate of the number of c
rods in boiling transition. Provide an estimate of the effect of this nonconservatism.
-10.
The proposed SLMCPR methodology differs from the presently approved
- generic methodology. The new method appears to be less conservative with respect-to:
(1) the selection of the initial CPR distribution, (2) determination of the limiting control rod
- pattern, (3) termination of the search for maximum SLMCPR and (4) the use of an equilibrium rather than a xenon-free xenon-distribution. These specific concerns were identified and described in the NRC Inspection Report No. 99900003196-01 (Letter U.S. NRC to C. P. Kipp (GE), dated Se)tember 10,1996).
Provide justification for these changes that lave been included in the proposed SLHCPR methodology.
- 11. The. revised methodology in which the power distribution model uncertainty is assigned on an individual rather than a four-bundle basis results in a (non conservative) decrease in the SLMCPR. This revision is based on the assumption that the modeling uncertainty in neighboring fuel bundles is uncorrelated.
In order to justify this revision, provide benchmark comparisons for the nodal bundle powers demonstrate that the modeling error in adjacent fuel bundles is not correlated..
s_
m ENCLOSURE 2 REQUEST FOR ADDITIONAL INFORMA"10N FOR GE TOPICAlf REPORT NEDC-M694P
" POWER DISTRIBUTION UNCERTAINTLES FOR SAFLT,Y LIMlT MCPR EVALUAMONS"
- 1.. Three-Dimensional Model Descrintion 1.
Under certain conditions, the 3-D MONICORE system rejects the TIP and LPRM measurements and uses calculated values in place of these measurements. When this occurs, are the TIP or LPRM measurements replaced by calculated values in any of the other BWR surveillance, monitoring or safety systems?
l 2.
Is there.any systematic trend in the types of fuel-bundles for which the TIPS ~and LPRMs are rej cted which suggests that the 3-D MONICORE calculated values are in error rather than the measurements? For example, are measurements more likely to be rejected if they-are adjacent'to (1) the core periphery (2)__ rodded bundles (3) part-length rods or (4) high-burnup fuel bundles?
3.
What is the recommended 'value for the rejection parameter o and is
-the same value used for both the TIP and LPRM rejection criterion?
Is this value consistent with the value of the rejection parameter a used.in the benchmark comparisons of Section-37 4.
How doer the 3-D MONICORE-thermal-hydraulic model differ from.the thermal-hydraulic model used in P-1,. and are the approved-SLMCPR i
uncertainties applicable to the 3-D MONICORE model?
5.
The correction for nodes that_are not adjacent to a TIP may be determined by either of two methods:-(1) reflecting values from symmetric nodes or (2) using averige values Which method was used in the beachmark comparisons (p.22-3, 1-3).
of Tables-3.1 ani 3.3?
What._ls-the dependence of the power _-distribution uncertainty on the
. selection of the method used to-determine-the buckling correction?
6.-~Are Equations (2-1) through (2-6) identical to the steady-state equations of-Reference-l?: If there are differences, are the comparisons' of Tables 3.1 and 3.2 applicable to the. 3-D MONICORE-calculation and what is the effect on the uncertainty estimates of Section-3?
7.
The 3-D MONICORE System rejects the-TIP measurements when the calculated and measured TIP readings indicate-large differences.
What evaluation and/or_ tests are performed to insure that calculation errors resulting from design, operational and fuel performance anomalies are identified?
w
- x....
- 8.- In the case of a rejected TIP or LPRM, the corrections for the nodes-
. adjacent to the rejected instrumentation are determined by planar average corrections. This approximation may introduce significant
-errors in cases where the local-axial power distribution differs significantly from the core-average' axial (e.g., in the case of part
-length fuel rods,. partially controlled fuel bundles and axially zoned Gd bundles).
Do the radial bundle power comparisons of Tables 3.1 and 3.3 include the effects of these types of situations? If
._not, how is the increased uncertainty in'these cares included in the uncertainty analysis?
l 9.
Should the flux, d, on the right hand side _pf Equation (2-17) i multiplying the LPRM leakage correction be (,7 If not, justify the difference between the TIP adaptive Equation (2-10) and the LPRM adaptive Equation (2-17)?.
- 10. The 3-D MONICORE TIP rejection criterion of Equation (2-25) rejects TIP readings that are in_ good _ agreement with the calculated values (i.e.,R
- 00)..
where la < R,P mislocation is present, the measured values areIn view of the rge TI considered to be more accurate, justify this approach.
II. Power Distribution'and TIP Instrument Uncertainties 1.
It is stated in NEDC-32694P (p. 3-8) that for most applications a TIP acceptance criterion of a, s 6% -is established.
Provide justification for the 6% TIP uncertainty value for cases in which this-acceptance criterion is not established.
- 2. -The comparisons _ of the :TIP. integral data of Table' 3.1 were determined using the " core tracking" predictions, rather than the-3-D MONICORE calculations. = What are the differences between the 3-D MONICORE and core tracking calculations? Can these differences result-in bette.- agreement between the TIPS and the core tracking predictions, than with the 3-D MONICORE calculations?
3.
Each of the seven cycles of TIP comparisons presented in Table 3.1 indicate that the 3-D MONICORE radial bundle power uncertainty increases with cycle exposure. What is-. causing this increase in uncertainty at the end-of-cycle (EOC)?- Is the 3-D MONICORE adjusted at beginning-of-cycle to improve agreement-with the TIPS? Provide
. justification fcr not using the larger EOC uncertainty?-
4..
Tables 3.1, 3.2 and 3.3. provide the 3-D MONICORE benchmarking data base for BWR core surveillance applications..What specific fuel designs are included in these tables?- Provide justification for the application of 3-D MONICORE to fuel designs not included in this-data' base..
- 5.-2The process computer monitors peak kw/ft and MAPLHGR. While MCPR
- depends primarily on the radial bundle power distribution, peak kw/ft and MAPLHGR depend on the bundle axial power distribution and, consequently,-are significantly more sensitive to the 3-D L
MONIC0RE replacement of the TIP/LPRM axial-power distribution.
Provide an uncertainty analysis for the-3-D MONICORE prediction of peak kw/ft and MAPLHGR.
6.
The' text (p. 3-5) states that the maximum RMS difference in bundle power due to missing TIP data for Case-4 is larger than the value given in Table.3.3.
Please explain this apparent inconsistency.
7.
Do the bundle power comparissns of Table 3.3 include the thermal-hydraulic feedback effects due to variations-in the axial power distribution, resulting from the TIP and LPRM replacements?
-8. 'The 3-D MONICORE bundle power uncertainty is based, to a large extent --on the TIP comparisons of Table 3.1.
Are the 3-D MONICORE calculations of Table 3.1 adjusted or normalized to dve. improved agreement with the measured TIPS? -For example, (1) Are the measured TIPS used in performing-the 3-D MONICORE calculation -
(2) Are the exposure, void or void-history distriM tions adjusted based on the TIP measurements or signal-to-power correlations been a(3) Have the 30 l'011 CORE TIP djusted based on.he TIP measurements? Also, is the resulting RMS difference typical of current BWR cycles?
III. Anolication of Revised Uncertainties to SLMCPR-Evaluation Methodoloav 1.
What code isLused to perform the Monte Carlo SLMCPR analysis for plants using 3-D MONICORE? -If.3-D MONICORE is not used, provide justification for using this alternate code for performing the SLMCPR Monte Carlo analysis.
2.
Is the methodology used to determine the SLMCPR for plants using the-P-1 surveillance methodology-affected by the changes implemented for the 3-D MONICORE System? What-uncertainties will be used for plants that-use the P-1 approach?
3.
Provide justification for the assumption made in the SLMCPR calculation that the power distribution uncertainties are normally distributed.
l.
N A
August 20, 1997 Mr. Ralph J. Reda, Manager Fuel and Facility Licensing GE Huclear Energy P. O. Box 780, MC J26 3901 Castle Hayne Road Wilmington, NC 28402
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION FOR GE TOPICAL REPORTS NEDC-3260lP (TAC NO. M97490) AND NEDC-32694P (TAC NO. M99069)
REFERENCES:
1.
GE Topical Report, NEDC-3260lP, ' Methodology and Uncertainties for Safety Limit MCPR Evaluations,"
GE Nuclear Energy, December 1996.
2.
GE* Topical Report, NEDC-32694P, " Power Distribution Uncertainties for Safety Limit MCPR Evaluations,"
GE Nuclear Energy, January 1997.
Dear Mr. Reda:
The staff is reviewing your submittals dated December 13, 1996 and June 10, 1997, regarding the methodology and uncertainties for safety limit MCPR evaluation and the power distribution uncerisinties for safety limit MCPR evaluation, respectively. The staff concludes that additional information is needed before it can complete its review. contains a request for additional information (RAI) related to the staff's review of NEDC-3260lP,
" Methodology and Uncertainties for Safety limit MCPR Evaluations." Enclosure 2 contains an RAI related to the staff's review of NEDC-32694P, " Power Distribution Uncertainties for Safety Limit MCPR Evaluations."
You are requested to provide responses to these RAls within 30 days of the date of this lotter.
If you need further clarification concerning this request, please contact Dr. Tai Huang at (301) 415-2867.
Sincerely, James H. Wilson, Senior Project Manager Generic Issues and Environmental Projects Branch Division of Reactor Program Management Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/ atts:
See next page DISTRIBlITION:
Docket: File' PUBLIC SRXB r/f PGEB r/f MCase DMatthews CENTRAL GHolahan TCollins LPhillips JHWilson THuang Document hame: GE-MCPR.RAI
- See previous concurrence OFC PGEB O J (A)SC:FufB SRXB/DSSA SC:SRXB/DSSA C:SRXB/DSSA NAME JHWils[n:sw MJCase h THuang*
LPhillips*
TCollins*
DATE 08/A /97 08/gu/9fM 08/19/97 08/19/97 08/19/97 0FFICIAL RECORD COPY
pm04 q}t UNITE 3 STATES p
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}
NUCLEAR REGULATORY COMMISSION 2
WASHINGTON, D.C. 300eM1001 fd August 20, 1997 Mr. Ralph J. Reda, Manager Fuel and Facility Licensing GE Nuclear Energy P. O. Box 780, MC J26 3901 Castle Hayne Road Wilmington, NC 28402
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION FOR GE TOPICAL REPORTS I
NEDC-3260lP (TAC NO. M97490) AND NEDC-32694P (TAC NO. M99069)
REFERENCES:
1.
GE Topical Report, NEDC-32601P, " Methodology and Uncertainties for Safety Limit MCPR Evaluations,"
GE Nuclear Energy, December 1996.
2.
GE Topical Report, NEDC-32694P, ' Power Distribution Uncertainties for Safety Limit MCPR Evaluations,"
GE Nuclear Energy, January 1997.
Dear Mr. Reda:
The staff is reviewii.g your submittals dated December 13, 1996 and June 10.
1997, regarding the methodology and uncertainties for safety limit MCPR evaluation and the power distribution uncertainties for safety limit MCPR evaluation, respectively. The staff concludes that additional infomation is needed before it can complete its review.
Enclosure I contains a request for additional information (RAI) related to the staff's review of NEDC-32601P,
" Methodology and Uncertainties for Safety limit MCPR Evaluations." Enclosure 2 contains an RAI related to the staff's review of NEDC-32694P, " Power Distribution Uncertainties for Safety Limit MCPR Evaluations."
I-You are requested to provide responses to these RAIs within 30 days of the date of this letter.
If you need further clarification concerning this request, please contact Dr. Tai Huang at (301) 415-2867.
Sincerely, b
Ja s H. Wilson, Senior Project Manager Generic Issues and Environmental Projects Branch Division of Reactor Program Management Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/ atts: See next page
l'.
GE Nuclear Energy--
- CC:
Gary L. Sozzi, Manager i
. Technical and Modification Services GE-Nuclear Energy 175 Curtner Avenue San Jose, CA' 95125 George B. Stramback GE Nuclear Energy 175 Curtner Avenue
. San Jose, CA 95125 James F. Klapproth GE Nuclear Energy.
i P.O. Box 780 H
Wilmington, NC 28402 l
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ENCLOSURE l' RE00EbT-FOR ADDITIONAL IhFORMATION FOR GE TOPICAL REPORT NEDC-326GlP
" METHODOLOGY AND UNCERTAINTlES FOR SAFETY LIMIT MCPR EVALUATIONS" I.
Process Comouter Oncertainties 1.
What_is the variation of the weighing coefficients, used in evaluating the uncertainty in the Equation-over reactor statepoint (e.g., flow, power,(2-3) inlet subcooling, subcooling, pressure)-
andnin what sense are these coefficients c~onservative?
2.
How does the channel flow uncertainty account for channel bulge and i
non-uniform crud / corrosion build up on the fuel' rods?
3.
Provide justification for neglecting the bias in the core pressure drop calculation in Table 2.2.
l 4.
In Table-2.2, the fact that only.the BWR6 data is negative-suggests i
that the uncertainty is plant dependent. Also, the fact that 50%
!=
of the data is outside the one-sigma interval-suggests that:the data is not normal.. Provide justification for treating this uncertainty as-normally. distributed -in the SLMCPR analysis. -
5.
How is'the uncertainty-in the bypass flow included in the uncertainty analysis?
II. 'R-Factor Uncertainty-
' l.- What specific fuel designs were the TGBLA-to-MCNP pin power:
comparisons performed and how do thes:t comparisons cover the intended range of GE BWR fuelf designs?-
2.
How is-the uncertainty in the TGBLA exposure' calculation accounted for -in:the determination of the local pin power peaking factor uncertainty?-
-3.
Provide ~ justification for the weight used to combine the standard-deviations of Table 3.1 and determine the local peaking model.
uncertainty.-
- 4. -Is the enrichment tolerance for rods with enrichment less:than'3.5%
greater than for rods with enrichment greater than 3.5% and, if so,
-how is the resulting increased uncertainty accounted for in the -
local peaking model uncertainty?
5.
Provide justification for-the assumption.that the effect of _ rod position-(e.g., due to rod _ bowing) has a negligible effect on the -
local power peaking.
O
6.
Describe the measurements of Table 2.2 and how they determine the single-phase friction factor uncertainty.
~
7.
The comparisons of Table 3.3 indicate that the local power peaking factor uncertainty is larger at the top of the fuel bundle. How is this apparent spatial dependence of the peaking factor uncertainty accounted for in the SLHCPR evaluation?
8.
Do the TGBLA-to-MCNP comparisons of Table 3.1 indicate a larger uncertainty in the high powered fuel rods and, if so, how is this accounted for in the SLMCPR7 9.
How is the uncertainty in the fuel density accounted for in determining the manufacturing uncertainty?
10.
How is the uncertainty in the local voids and exposure accounted for in the local peaking factor uncertainty?
11.
In the improved R-Factor method, how is the uncertainty in the bundle-average void and exposure distributions used M performing the integration of the local R-Factor accounted for in the determination of the R-Factor uncertainty?
- 12. The local power peaking modeling uncertainty provided on p. 3-2 is greater than the value given in the first sentence of Section 3.1.4.
Please explain this apparent inconsistency.
- 13. The power peaking uncertainties in neighboring fuel rods are generally correlated and, consequently, can not be taken to be independent and random as assumed in the calculations of Section 3.2.
Provide an estimate of the R-Factor uncertainty that recognizes the correlation of the uncertainties in neighboring fuel rods.
III.
SLMCPR Evaluation Methodoloov 1.
In Section 4.3, it is stated that the SLMCPR values of Table 4.1 are for the revised methodology and the present (larger) uncertainties, while Table 4.1 indicates that the revised methodology is evaluated with the revised uncertainties.
Please explain this apparent inconsistency.
2.
Provide the basic mathematical definition of Wcore (rather than the mathematical result of the integration given in Section 4.2).
3.
Provide the definition of the MCPR Importance Parameter (MIP) of Figure 4.4.
4.
What )in power distribution is assumed in the definition of Wcore and w1at is the effect of this assumption?,
A
i;g 1-F 5.
Describe the selected 100 nominal control rod pattern cases of Figure 4.4.
How do these cases and the Figure 4.4 comparisons of the nominal-and limiting MIPs accommodate off-nominal operating statepoints.- Provide justification for-the assumption that these i
100 patterns _ bound the operating statepoints.
l 6.
Provide specific details describing how the limiting control rod pattern will be selected in a typical reload-core determination of i
the SLMCPR. What quantitative criteria will be-used to select the F
limiting pattern? What statepoints _(e.g., power, flow, exposure)
F will-be included in the determination of the limiting pattern? How i
will the Figure 4.4 comparisons be used to confirm the limiting c
pattern selection? Will the data base ~or'these 100 nominal cases
- be expanded or updated?
7.
Will the Wcore parameter be monitored by the process computer to j
insure that the design SLMCPR limiting power distribution bounds -
- the operating power distribution?'
l:
- 8. _Do all the MIP calculations of Figure 4.4 assume the same uncertainties? If not, provide the nominal-to-limiting MIP L
comparisons separately for each set of uncertainties and explain
[
the effect of-this inconsistency on the conclusions drawn from the i
Figure 4.4 comparisons.
l j
9.
The elimination of fuel bundles from-the-SLMCPR calculation using i
the criteria P < AP results.in an underestimate of the number of rods in-boilin,g transition. Provide an: estimate of the effect of-c j.
this nonconservatism..
- 10. The proposed SLMCPR methodology differs from the presently approved generic methodology. The new method appears-to-be less conservative-with respect to:
(1) the selection of the initial CPR distribution,-(2) determination of the limiting control-rod-
~
L pattern, :(3)- termination of the search for maximum SLMCPR-and (4) the use of an equilibrium rather than a. xenon-free xenon
' distribution. These specific concerns were identified and described in'the NRC. Inspection Report No._99900003196-01 (Letter
- U.S.:NRC to C. P.. Kipp (GE), dated September:10,1996).- Provide justification for these changes that have been included in the
[
proposed SLMCPR methodology.
11.-'The revised methodology in which the power distribution model uncertainty is assigned on-an individual rather than a four-bundle basis results in a~(non conservative) decrease in the SLMCPR. This revision-is based on the assumption that the modeling uncertainty in-neighboring fuel bundles 1s.uncorrelated.
In order to justify this revision, provide benchmark comparisons for the nodal bundle powers demonstrate that the modeling error in adjacent fuel bundles-j
.is.not correlated. l l:
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ENCLOSURE 2 RE0 VEST FOR ADDITIONAL INFORMATION FOR GE TOPICAL REPORT NEDC-32694P
" POWER CISTRIBUTION UNCERTAINTIES FOR SAFETY LIMIT MCPR EVALUATIONS" I.
Three-Dimensional Model Descriotion 1.
Under certain conditions, the 3-D MONICORE system rejects the TIP and LPRM measurements and uses calculated values in place of these measurements. When this occurs, are the TIP or LPRM measurements replaced by calculated values in any of the other BWR surveillance, monitoring or safety systems?
2.
Is there any systematic trend in the types of fuel bundles for which the TIPS and LPRMs-are rejected which suggests that the 3-D MONICORE calculated values are in error rather than the measurements? For example, are measurements more likely to be rejected if they are adjacent to (1) the core periphery (2) rodded bundles (3) part length rods or (4) high burnup fuel bundles?
3.
What is the recommended value for the rejection parameter a and is the same value used for both the TIP and LPRM rejection criterion?
Is this value consistent with the value of the rejection parameter a used in the benchmark comparisons of Section-37 4.
How does the 3-D MONICORE thermal-hydraulic model differ from the thermal-hydraulic model used in P-1, and are the approved SLMCPR uncertainties applicable to the 3-D MONICORE model?
5.
The correction for nodes that are not adjacent to a TIP may be determined by either of two methods:- (1) reflecting values from symmetric nodes or (2) using average values (p. 2-3, 1-3). Which method was used in.the benchmark comparisons of Tables 3.1 and 3.3?
What is the dependence of the power distribution uncertainty on the selection of the method used to determine.the buckling correction?
6.
Are Equations (2-1) through (2-6) identical to the steady-state equations of Reference-1? If there are differences, are the comparisons of Tables 3.1 and 3.2 applicable to the 3-D MONICORE calculation and what is the effect on-the uncertainty estimates of Section-3?
1 7.
The 3-D MONICORE System rejects the TIP measurements when the celculated and measured TIP readings indicate large differences.
What evaluation and/or tests are performed to insure that calculation errors resulting from design, operational and fuel performance anomalies are identified?
.. 8.
In the case of a rejected TIP or LPRM, the corrections for the nodes adjacent to the rejected instrumentation are determined by planar average corrections. This approximation may introduce significant errors in cases where the local axial power distribution differs significantly from the core-average axial (e.g., in the case of part length fuel-rods, partially controlled fuel bundles and axially zoned Gd bundles). Do the radial bundle power comparisons of Tables 3.1 and 3.3 include the effects of these types of situations? If not, how is the increased uncertainty in these cases included in the uncertainty analysis?
9.
Should the flux, di, on the right hand side _pf Equation (2-17) multiplying the LPRM leakage correction be d ?
If not, justify the i
difference between the TIP adaptive Equation (2-10) and the LPRM adaptive Equation (2-17)?
- 10. The 3-D MONICORE TIP rejection criterion of Equation (2-25) rejects TIP readings that are in good agreement with the calculated values
( i. e., R < R, - ea).
In view of the fact that, except for cases where la,rge TIP mislocation is present, the measured values are considered to be more accurate, justify this approach.
l II.
Power Distribution and TIP Instrument Uncertainties l.
It is stated in NEDC-32694P (p. 3-8) that for most applications a l
TIP acceptance criterion of a s; 6% is established.
Provide y
justification for the 6% TIP unc,ertainty value for cases in which this acceptance criterion is not established.
2.
The comparisons of the TIP integral data of Table 3.1 were determined using the " core tracking" predictions, rather than the 3-D MONICORE cclculations. What are the differences between the 3-D MONICORE and core tracking calculations? Can these differences result in better agreement between the TIPS and the core tracking predictions, than with the 3-D MONICORE calculations?
3.
Each of the seven cycles of TIP comparisons presented in Table 3.1 indicate that the 3-D MONICORE radial bundle power uncertainty increases with cycle exposure. What is causing this increase in uncertainty at the end-of-cycle (E0C)? Is the 3-D MONICORE adjusted at beginning-of-cycle to improve agreement with the TIPS? Provide justification for not using the larger EOC uncertainty?
4.
Tables 3.1, 3.2 and 3.3 provide the 3-D MONICORE benchmarking data base for BWR core surveillance applications.
What specific fuel designs are included in these tables? Provide justification for the application of 3-D MONICORE to fuel designs not included in this data base.,
i 5.
The process computee monitors peak kw/ft and MAPLHGR.
While MCPR depends primarily on the radial bundle power distribution, peak kw/ft and MAPLHGR depend on the bundle axial power distribution i
and, consequantly, are significantly more sensitive to the 3-D MONICORE replacement of the TIP/LPRM axial power distribution.
Provide an uncertainty analysis for the 3-D MONICORE prediction of peak kw/ft and MAPLHGR.
6.
The text (p. 3-5)ing TIP data for Case-4 is larger than the value
. states that the maximum RMS difference in bundle power due to miss given in Table 3.3.
Please explain this apparent inconsistency.
7.
Do the bundle power comparisons of Table 3.3 include the thermal-hydraulic feedback effects due to variations in-the axial power-distribution, resulting from the TIP and LPRM replacements?
8.
The 3-0 MONICORE bundle power uncertainty is based, to a large extent, on the TIP comparisons of Table 3.1. - Are the-3-D MONICORE calculations of Table 3.1 adjusted or normalized to give improved agreement with the measured TIPS? For example (1) Are the measured TIPS used in performing the 3-D MONICORE calculation (2) Are the exposure, void or void-history distributions adjusted based on the TIP measurements-or -(3) Have-the 3-D MONICORE TIP signal-to-power correlations been adjusted based on the TIP measurements? Also, is the resulting RMS difference typical of j
current BWR cycles?
!!!. Aonljeation of Rev,titA Uncertainties to SLMCPR Evaluation Methodoloav 1.
What code is used to perform the Monte C;rlo SLMCPR analysis for plants using 3-D MONICORE7 If 3-D MONICORE is not used, provide justification for using this alternate code for performing the SLMCPR Monte Carlo analysis.
2.
Is the methodology used to determine the SLMCPR for plants using the P-1 surveillance methooology affected by the changes implemented for the 3-D MONICORE System? What uncertainties will be used for plants that use the P-1 approach?
3.
Provide justification for the assumption made _in the SLMCPR calculation that the power distribution uncertainties are normally distributed.
(.
_j p