ML20236D160
| ML20236D160 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 03/13/1989 |
| From: | Martin R Office of Nuclear Reactor Regulation |
| To: | Hunger G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| References | |
| TAC-67190, TAC-67191, NUDOCS 8903220492 | |
| Download: ML20236D160 (6) | |
Text
._
March 13, 1989 Dockets Nos. 50-277/278 Mr. George A. Hunger, Jr.
Director-Licensing Philadelphia Electric Company Correspondence Control Desk P. O. Box 7520 Philadelphia, Pennsylvania 19101
Dear Mr. Hunger:
SUBJECT:
QUESTIONSONREPORTPECo-FMS-0005(TACNOS. 67190/67191)
RE:
PEACH BOTTOM ATOMIC POWER STATION, UNIT NOS. 2 AND 3 The staff has developed questions as a result of its review of the report entitled PECo-FMS-0005, " Methods for Performing BWR Steady State Reactor Physics Analyses." These questions, as attached, were developed by our contractor, the Brookhaven National Laboratory.
Please provide us with a firm commitment to a schedule for the submittal of your response so'that we may schedule our review.
Sincerely,
/s/
Robert E. Martin, Project Manager Project Directorate I-2 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/cnclosure:
See next page
. DISTRIBUTION (Docket File NRC PDR & Local PDR PDI-2 Reading Glainas/BBoger WButler/M0'Brien RMartin/RClark OGC EJordan/BGrimes WHodges HRichings
@~ 0 ACRS (10) b
'\\
[ HUNGER]
tr kB t
]//J/89
} / g 89
!!R 888 85885377 l
p PNU
~
[+f
'o,,
UNITED STATES NUCLEAR REGULATORY COMMISSION 3
o
. r, a
w AsHINGTON, D. C. 20666
%.....,/
March 13, 1989
]
4 Dockets Nos. 50-277/278 i
Mr. George A. Hunger, Jr.
l Director-Licensing i
Philadelphia Electric Company j
Correspondence Control Desk P. O. Box 7520 Philadelphia, Pennsylvania 19101 i
f
Dear Mr. Hunger:
SUBJECT:
QUESTIONS ON REPORT PECo-FMS-0005-(TAC NOS, 67190/67191)
RE:
PEACH BOTTOM ATOMIC POWER STATION, UNIT NOS. 2 AND 3 i
The staff has developed questions as a result of its review of the report J
entitled PECo-FMS-0005, " Methods for Performing BWR Steady State Reactor y
l Physics Analyses." These questions, as attached, were developed by our i
contractor, the Brookhaven National Laboratory.
Please provide us with a firm commitment to a schedule for the submittal of your response so that we may schedule our review.
i Sincerely, i
5 b
b
' Robert E. Martin, Project Manager Project Directorate I-2 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation l
Enclosure:
As stated j
1 l
cc w/ enclosure:
See next page I
i
Mr. George A. Hunger, Jr.
Peach Bottom Atomic Power Station, Philadelphia Electric Company Units 2 and 3 cc:
Troy B. Conner, Jr., Esq.
Single Point of Contact 1747 Pennsylvania Avenue, N.W.
P. O. Box 11880 Washington, D.C.
20006 Harrisburg, Pennsylvania 17108-1880 i
Philadelphia Electric Company Mr. Thomas M. Gerusky, Director ATTN: Mr. D. M. Smith, Vice President Bureau of Radiation Protection Peach Bottom Atomic Power Station Pennsylvania Department of i
Route 1, Box 208 Environmental Resources i
Delta, Pennsylvania 17314 P. O. Box 2063 Harrisburg, Pennsylvania 17120 J
H. Chris Schwemm Vice President, Production Mr. Albert R. Steel, Chairman Atlantic Electric Board of Supervisors P.O. Box 1500 Peach Bottom Township 1199 Black Horse Pike R. D. #1 Pleasantville, New Jersey 08232 Delta, Pennsylvania 17314-Resident Inspector Mr. Gary Mock U.S. Nuclear Regulatory Commission P. O. Box 09181 Peach Bottom Atomic Power Station Columbus, Ohio '43209 P.O. Box 399 Delta, Pennsylvania 17314 Delmarva Power an'd Light Company c/o Jack Urban Regional Administrator, Region I General Manager, Fuel-Supply J
U.S. Nuclear Regulatory Commission 800 King Street 475 Allendale Road P. O. Box 231 King of Prussia, Pennsylvania 19406 Wilmington, DE 19899 Mr. Bryan W. Gorman Mr. Tom Magette Manager - External Affairs Power Plant Research Program Public Service Electric & Gas Company Department of Natural Resources P. O. Box 236, N28 B-3 Hancocks Bridge, New Jersey 08038 Tawes State Office Building Annapolis, Maryland 21401 Mr. Roland Fletcher Department of Environment 201 West Preston Street Baltimore, Maryland 21201 l
~,
EHCLOSURE
)
bestions concernina _the PECO Steady-State Physics Nethods (PEC0=FNS-0005) i b
1.
How is the bias in the SIMUI. ATE-B k gg critical data relative to j
k
= 1.0 secounted fort hat is the standard devietton of the j
8 ALTE-1 k gg predictions relative to k,gg = 1.,07 e
2.
Does the FECO physics methodology require three SIELATE/ measurement l
normalisatione during a burnup cycle (30C, WC, 50C)?
Eince this is not typical industry practice why are these additional normalisations i
requiredt het parameters will be adjusted and will this information be j
used to update the precalculated eyele safety analyses such as those of j
Appendices A-D7 j
i f
3.
Wat is the effect on the PECO calculation uncertainty estimates when only the PICO calculations are included in the benchmark comparisons 4.
Are the KINO-II and CASMO calculation independent? Por example, ars the nuclear eroes sections used in the two calculations differenet 5.
How do the uncertainties derived fron, the EPRI and AB Atomenergi Doppler coefficient comparisons account for actual core conditions such as fuel burnup, fuel rod geometrical changes, fuel temperature uncertainty, spec-tral effecte due to the presence of voids, etc.?
6.
How is the expression for (&K/K) pop (p. 4-209) derived?
Wat is the relation between Koop and the CA8MO*-1 cross sections?
7.
How are the rodded and unrodded void and Doppler reactivity coefficient uncertainties combined to determine the core reactivity coefficient l
uncertainties 8.
What is the sensitivity of the Doppler and void coefficients to changes in the core flux distribution during a transient?
How is this uncer-tainty accounted fort j
i 9.
Iow do the Peach Botton-3 Cycle-7 (for example) Doppler and void coeffi-cients and control rod worth, calculated by 81MULATE-E, compare with the fuel vendor valueet 14 How is the dependence of tha void coe fficient on changes in control rod insertion and void fraction during a transient accounted forf Wat un-e artainty is introduced by the treatment of these effectst l
f l
11.
How is the reduced leakage probability for delay \\
ed neutrons calculatedt i
12.
How is the highest worth rod determined in the calculation of core shut-l down margint
- 13. How are the 'proj ected cold critical eigenvalues" determined from the "date base of cold critical projections" (p. 5-32)? How are the project =
ed eigenvaluvos' represented by a polynomials i
4
4.
14 4w is the increased uncertainty in the shutdown margin calculation of k gg for the state with the highest worth rod withdrawn accounted fort e
15.
hat are the calculational uncertainties when all asasured TIF signals are included in the calculation /aeasureaant comparisons (e.g.,
- ICFR, MAFLEGR, FFLEGR, etc.)?
16.
In the analysis of the rod withdrawal event, how is the error-rod yield-ing the minimum REM setpoint determined?
Eow is the uncertainty intro-duced by this procedure accounted for?
17.
In the analysis of the rod withdrawal event, how are rods with less than four adjacent LPRM strings treated considering worst-case LPIM failurast
- 18. Bow is the miseriented fuel bundle treated?
19.
What is the sensitivity of the ACPn regression fit of Figure 3-1, for the aislocated bundle loading error, to the core operating conditions (power, flow, rod pattern, menon, etc.)?
How is the resulting uncertainty accounted forf
- 20. In the loss of feedwater heater event, is the feedwater flow increased as
{
a result of the increased power leve17 If not, how is the resulting in-j eresse in power and axial peaking accounted for in the analysis?
j 21.
What is the sensitivity of the loss of feedwater heater event to the core i
j c6aditions (power, flow, menon, rod pattern, exposure, pressure, initt subcooling), and how is the resulting uncertainty accounted fort j
22.
The PEC0 loss of feedwater heater analysis assumes a fixed core power shape during the transient.
In fact, the asial power distribution be-comes more botton-peakad during the transient, resulting in an additional reduction in CPR margin in the bottom of the core.
Iow is this effect accounted for in the FECO methodology?
t
{
23.
The 4.1% EMS arter in the 8IMULAR assembly integral power calculation is I
bated en the elimination of the top and bot 10s 18 inches of'the core from I
tha statistics. W at is the effect of this deletion on the calculational uncertainty? Are these regions ever limitingt j
i 24.
Describe the fuel loadings of 79-2, Cycles 5 and 6 and 75-3, Cycles 4, 5 and 6 which were included in the benchmarking of SIMULATE. Does FECO in-tend to see fuel designs and loadings which are not represented in the benchmarking? If so, what are they and how will they affect the qualifi-cation of the FECO esthodst
{
- 25. 'the het critical sigenvalue results for F5-2 and PR-3 shown in Fi l
3.1.1 through 31.9 show a pronounced upward trend with esposure. gures This j
trend is clearly seen in the multicycle plot of Figures 3.1.6 and 317.
What is' causing this esposure dependent bias, and how is it accounted for j
in the SIMULATE predictions?
i i
~
- 16..With specific SIMULATI-E normalization parameters are adjusted from one cycle to the next?
27.
tuplain the systematic underprediction of the core average axial power distribution near the bottom of the cors.for BF-3 Cycle 6.
28.
Describe the peccedures used in correcting for temperature and reactor period in the calculation of the critical tests.
29.
Have any few-rod criticals been evaluated with SIMULATE-E and, if so, how do these calculations compare with the esasurements?
- 30. Are the system variables such as pressure, feedwater flow, staan flow, etc., assumed constant during the loss of feedwater heater transient?
If so, provide the basis for this assumption.
Can changes in these variables result in a limiting ACFR. during the transient, making - the
'j final-state ACPR calculation not bounding?
l l
l l
1 l
l l
l l
1 i
-