ML17037C192
| ML17037C192 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  (DPR-063) |
| Issue date: | 10/03/1975 |
| From: | Rhode G Niagara Mohawk Power Corp |
| To: | Lear G Office of Nuclear Reactor Regulation |
| References | |
| Download: ML17037C192 (30) | |
Text
NRC DISTRIBUTION FOR PART 50 DOCKET MATERIAL (TEMPORARY FORM)
CONTROL NO:
FILE'
- ROM, Niagara Mo awk Power Co
- Syracuse, N,Y ~
Gerald K. Rhode PDATE OF DOC 10-3-75 DATE REC'D 10-7-75 LTR TWX RPT
- OTHER, TO:
NRC CLASS UNCLASS PROPINFO DESCR IPTION:
ORIG 1-signed INPUT CC OTHER NO CYS REC'D 1
ENCLOSURES:
SENTNRC PDR SENT LOCAL PDR DOCKET NO:
50>>220 Ltr re our 9>>17-75 ltr
~ ~ ~ ~ furn addi info concerning the Core Cycle. 4 Ana'lysis
~ ~ ~
trans the following:
PLANT NAME:,Nine Mile Point ffl Responses to Questions concerning the Core Cycle 4 Analysis
~ ~ ~ ~ ~ ~
<:( 1'y enc' rec'd)
ACTINO...
BUTLER (L)
W/ Copies CLARK (L)
W/ Copies PARR (L)
W/ Copies KNIEL (L)
W/ Copies ZlEMANN (L)
W/ Copies DICKER (E)
W/ Copies KNIGHTON (E)
W/ Copies
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YOUNGBLOOD W/ Copies SCHWENCER (L)
W/ Copies STOLZ (L)
W/ Copies VASSALLO (L)
W/ Copies PURPLE (L)
W/ Copies REGAN (E)
W/ Copies
~AR (L)
W/(Copies SPIES W/ Copies (E)
LPM W/
opuses INTERNALDISTRIBUTION FOR ACTION/INFORMATION 5 JGB REID(L)
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COPIES RC PDR OGC, ROOM P-506A GOSSI CK/STAFF CASE G IAMBUSSO BOYD MOORE (L)
DEYOUNG (L)
SKOVHOLT (L)
GOLLER (L) (Ltr)
P. COLLINS D
ISE G OPR FILE 5 REGION (2)
MIPC TECH REVIEW SCHROEDER
~ACCARY KNIGHT PAWLICKI SHAO MELLO HOUSTON
~VAK ROSS IPPOLITO
~DESCO SNCOLLINS LAI NAS BENAROYA VOLLMER DENTON LIC ASST G R IMES R. DIGGS (L)
GAMMILL H. GEAR IN (L)
KASTNER E. GOULBOURNE (L)
BALLARD P. KREUTZER (E)
SPANGLER J. LEE (L)
M. RU3HBROOK(L)
ENVIRO S. REED (E)
- MULLER, M. SERVICE (L)
DICKER S. SHEPPARD (L)
KNIGHTON M. SLATER (E)
YOUNGBLOOD H. SMITH (L)
R GAN g8. TEETS (L)
OJECJ LDR G. WILLIAMS(E)
Al4'. WILSON (L)
AIV..ESS R. INGR '"5 (L)
,M. DUNCAN EXTERNALDISTRIBUTION A/T IND
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BRAITMAN SALTZMAN ME LTZ PLANS MCDONALD CHAPMAN DUBE (Ltr).
E. COUPE PETERSON HARTFIELD (2)
KLECKER EISENHUT WIGGINTON
~ LDCALPDR~Oswe 0, N.Y.
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(1)(2)(10) NATIONALLABS W-NSIC (BUCHANAN) 1 W. PENNINGTON, Rm E-201 GT 1 ASLB 1
CONSULTANTS 1 Newton Anderson NEWMARK/BLUME/AGBABIAN.
ACRS lWI88$RSISENT 1 P DR-SAN/LA/NY 1 BROOKHAVEN NAT LAB 1 G. ULRIKSON ORNL
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~~>ry NIAGARAMOHAWKPOWER CORPORATION NIAGARA a
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300 ERIE BOUI EVARD, WEST SYRACUSE, N. Y. 13202 5
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/~ I october 3, 1975
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"4,r ll Director of Nuclear Reactor Regulation Attn:
Mr. George Lear, Chief Branch
//3 U.
S. Nuclear Regulatory Commission Washington, D. C.
20555
'0 Re:
Nine Mile Point Unit Docket No. 50-220 C-
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OCT g 1975~h 1III~Ihh hFr'II'ICRI C0>aaaaa cry aI ag IhhII thtIIoh
Dear Mr. Lear:
Your letter dated September 17, 1975 requested additional information concerning the Core Cycle 4 Analysis for Nine Mile Point Unit 1.
The enclosed information addresses itself to the attachment of your letter.
Sincerely, NIAGARA MOHAWK POWER CORPORATION erald K. Rhode Vice President Engineering Ir'SZ Enclosure ro >4~
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1.
QUESTION Provide the value for the delayed neutron fraction applicable to the beginning and the end of Cycle 4.
RESPONSE
The'value of the delayed neutron fraction applicable to the beginning of Cycle 4 is 0.006261; the value applicable to the end of Cycle 4 is 0.005361.
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2.
QUESTlON Provide the design conservatism factors for scram reactivity, Doppler coefficient, and void coefficient (include both values used in Table 6-1).
RESPONSE
The attached Table 2.1 provides the design conservatism factors for scram reactivity, Doppler coefficient and void coefficient.
Table 2.1 Design Conservatism Factors Parameter Percent of Nominal Scram R'eactivity Doppler Coefficient Void Coefficient 80~o 90~o Negative reactivity transients Positive reactivity transients 90~o 125~
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3.
QUESTION Provide a statement verifying that the Neutron Effective Void model was used in the calculation of the void coefficient used for Cycle 4.
RESPONSE
The core transient data used in the transient analysis for Cycle 4 is based on the Neutron Effective Void (NEV) model for calculating void coefficients.
For the abnormal operating transient
- analyses, the void coefficient calculated by the NEV model is converted to units of r/':.
4.
QUESTNN Provide the control rod location used in the development of Figures 6-12 and 6-13 of NED0-20772, "GE-HER Reload-5 Licensing Submittal for NMP-1 Nuclear Power Station, Unit 1", for the Rod Withdrawal Error transient analysis.
RESPONSE
The control rod at position 22, 35 was used to develop information presented in Figures 6-12 and 6-13 of NED0-20772.
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5.
QUES'I'ION Provide the time of Cycle 4 to which the curves in Figures 6-2 through 6-5, NED0-20772, correspond.
RESPONSE
The curves provided in Figures 6-2 through 6-5 of NEDO-20772 correspond to the beginning of Cycle 4 conditions.
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6.
QUESTION The analysis of the fuel loading error accident (page 6-7 of NEDO-20772) shows that a localized MCPR of 1.05 would occur in a misplaced bundle.
Provide an explanation of the consequences of operating with a MCPR less than the Safety Limit of 1.06.
Include the radiological consequences, if any.
RESPONSE
The results of the fuel loading error accident -as presented in NEDO-20772 were calculated using overly conservative initial conditions.
Particularly, the initial MCPR's were below the cycle 4 operating limits.
The results of a re-analysis of this accident based upon operating limits show that the resulting MCPR of a misplaced bundle will be 1.097.
The MCPR value which determines boiling transition for the fuel loading error accident is 1.0.
Therefore, there are no radiological consequences'
7.
QUESTION The GETAB transient analysis initial condition parameters, Table 4-4 of NEDO-20772, indicate that the initial condition MCPR values used in the transient analyses are lower than the operating limits derived from the analyses.
Provide a discussion of how this affects the conservatism of the analyses.
Include an explanation of the relationship between initial MCPR and dMCPR.
RESPONSE
The effect of initial condition MCPR values used in transient analyses i<as discussed in Response 9 of a January 20, 1975 letter from C. H.
,Frauenholz (General Electric) to A. J. Ignatonis (NCR).
The transient analysis as presented in NEDO-20772 has been reanalyzed to reflect proper initial conditions.
Revisions to Tables 4-3 and 4-4 of NEDO-20772 are attached.
No changes to MCPR operating limits are required.
Table 4-3 Revised LIMITING PRESSURE AND POi'/ER INCREASE TRANSIENTS Event Maximum r CPR 7x7 8x8 Turbine Type w/o Bypass Trip Scram, 94 percent Power 100 percent
.13
.16 Rod withdrawal Error 0.30 0.32
Table 4-4 Revised GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAhKTERS Peaking factors (local, radial and axial)
R-Factor Bundle Power, hNt Non-fuel Power Fraction Core Flow, Mlb/hr Bundle Flow, 10 lb/hr Reactor Pressure, psia Inlet Enthalpy, Btu/lb Initial hfCPR jx7 1.30, 1.47, 1.40 1.100
- 4. 707
- 0. 035
- 67. 5 114.1 1048.2 526.1 1.36 8x8 1.22, 1.612, 1.40 1.102 5.158 0.035 67.5 101.4 1048.2 526.1 1.38
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8.
QUESTS)N Provide an explanation of the differences in the GETAB transient analysis initial condition parameters as presented in Table 4 of your June 30, 1975 submittal and Table 4-4 of NED0-20772.
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RESPONSE
The information presented in our June 30, 1975 submittal relates to cycle 3 operation whereas the NEDO-20772 document addresses cycle 4
operation.
The differences in the GETAB transient analysis initial condition parameters are due to differences in core transient response Because of a more negative void coefficient and a "worsened" scram reactivity curve, the transient response for cycle 4 is more severe.
Therefore, initial MCPR's'were necessarily higher.
Additionally, the increased severity of the transient response is reflected in the
'reduction of allowable end of cycle power level for cycle 4 to 94 percent rated.
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9.
QUESTION For the Rod Withdrawal Error transient analysis, provide a curve of APRM channel reading
(~o of initial level) versus control rod position (ft. withdrawn) for the case where no LPRM's are bypassed.
RESPONSE
The APRM response versus rod position curves presented in NEDO-20772 are for the worst permitted bypass conditions.
The numerical data for other bypass conditions (including the case where no LPRM's are bypassed) have been reviewed to. assure that the worst case curves are those previously submitted.
The standard rod withdrawal error analysis for the unique APRM system at Nine Mile Point Unit l does not create the computer generated graphs normally submitted with other plant analyses.
As a result, the APRM channel reading curve with no LPRM bypasses is not readily available.
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10.
QUESTION Provide an explanation for the difference in the void fraction values in Tables 5-1 and 6-1, NED0-20772.
RESPONSE
The void fraction in Table 6-1 of NEDO-20772 is calculated from a void map which is a function of subcooling, exit quality, pressure and core flow.
This value is used in the dynamic code from which transient calculations are performed (for further description see NED0-10802, "Analytical Methods of Plant Transient Evaluations for the General Electric Boiling l(ater Reactor",
issued in February, 1973).
The voi:d fraction value presented in Table 5.1 of NEDO-20772 was calculated from a nuclear code and is typical of average expected conditions for cycle 4.
This value is not used in safety analyses.
The intent historically was that the dynamic code should provide more conservative transient results.
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