ML20195K248
| ML20195K248 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 11/30/1988 |
| From: | Charnley J, Marriott P, Plotycia G GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19295G771 | List: |
| References | |
| 23A5948, 23A5948-R01, 23A5948-R1, NUDOCS 8812050172 | |
| Download: ML20195K248 (22) | |
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O 23A5948 REVISION 1 CIAS$ I O-Nov5KSER 1988 O
l 23A5948 REV. 1 SUPPLEMENTAL RE14AD LICENSING SUSMITTAL O
r0R PERRY NUCLEAR POVER P1Ahi UNIT 1 O
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Prepar d
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J M akl lel Licens O
Verified by:
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w C. DT Plotycia /
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Tuel Licensing O
Approved by:
P. V, Nathott. Managar Licensing and Consulting Services l
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O NAlsedperEastyy l
O IM cvw Ava Se Jon. CA 99M 1/2 0
g 23A5948 REV. 1 IMPORTANT NOTICE RECARDING CONTENTS OF THis REPORT 3-Pt.FASE READ CARE W1.1.Y this report was prepared by the General Electric Ceepany (CE) solely 3
for Cleveland tiectric Illuminating Ceepany (Ct!) for Ct!'s use with the United States Nuclear 5.ogulatory Ceemission (USMRC) for amending CEI's operating license of the Perry Nuclear power Plant Unit 1.
The information contained in this report is believed by CE to be'an accurate and true 3
representation of the fact 3 known, obtained or provided to CE at the time this report was prepared.
The only undertakings of CK respecting informatten in this document O
'**"**'"**'"'h'**"*'******"""C""'""'***""'
related services for the nuclear system for the Perry Nuclear Power Plant Unit 1. and nothin5 contained in this de;usent shall be construed as changing said contract. The use of this information except as defined by O
'' **"*** ** '"# '" " ** **h** ***" **** '*' "h'*" **
'"**"d
is not authertsed; and with respect to any such unauthertsed use, neither CE nor any of the contributors to this document makes any rsyresentarien er warranty (express or implied) as to the completeness, accuracy or useful.
ness of the informatten contained in this document er that such use of such information may not infringe privately evned rights; nor de they usume any responsibility for liability or damage of any kind which may result free such use of such information.
O O
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LO 23^5s nav 1 ACKN0'iLEDCEMENTS O
The engineering and reload licensing analyses, which fore the technical basis of this supplemental Reload Licensing subetttal, were performed in the ruel Engin.oring seceton by r. A, x.hn and J. L. c. sill...
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23A5948 REV. 1
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1.
PLANT UNIOUE ITEMS (1.0i*
Appendix A: Analysis Conditions c(s Appendix B: Besim for Analysis of Loss of-Feedwater Heating Event Appendix C: Application of GEMINI Methods Appendix D: CEXL PLUS Correlation Appendix E: Analyzed Operating Domain I) 2.
RELOAD FUEL BUNBLES (1.0 and 2.0)
Fuel Tyne Cvele Loadgd Number
')
Irradiated BP85RB176 1
52 BP8 SRB 219 1
424 O
(
New BS301E 2
136
()
BS301F 2
_111_
Total 748 3.
REFERENCE CORE LOADING PATTERN (3.2.1)
O Nominal previous cycle core everage exposure at end of cycle:
10,070 MVd/MT Minitum previous cycle core averaga exposure
()
at end of cycle from cold shutdown considerations:
9,708 MVd/MT Assumed reload cycle core average exposure at end of cycle (all rods out, rated power):
15,598 mwd /MT Core loading pattern:
Figure 1 O
- ( ) Refers to area of discussion in General llectric Standard Application for Reactor Fuel, NEDE 24011 P A 9 (dated September 1988); a letter "S"
preceding the number refers to the United States Supplement.
' C) 7 0
1
23A5948 REV. 1 O'
4.
CAlfUIATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH-NO VOIDS. 20 DEC. C (3.2.4.1 AND 3.2.4.2) 8' Beginning of Cycle, k effective Uncontrolled 1.119 Fally Controlled 0.943 II Strongest Control Rod Out 0.971 R, Maximum Increase in Cold Core Reactivity with Exposure into Cycle, Delta k O.017 5.
STANDBY LIOUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.2.4.3)
O Shutdown Margin (Delta N) gas (20 der.C. Xenon Free) 660 0.038 4D L 6.
RELOAD-UNIOUE Anticiented Ooerational Occurrence (ADO) ANALYSIS INPUT (3.2.3 AND 4.3.1.2.3) e Values normally reported in this section are REDY inputs. There were no transients analyzed using REDY.
7.
RELOAD UNIOUE CETAB AOO ANALYSIS INITIAL CONDITION PARAMETERS (S.2.3)
Gl i Exposure: BOC2 TO EOC2 Fuel Peakina Factors R.
Bundle Bundle Flow Initial II.'
Desien Local Radial &Eial Factor Power (HWT)
(1000 lb/hr)
MCPR BP8x8R 1.20 1,54 1.40 1.051 7.211 115.0 1.16 CE8x8EB 1.20 1.54 1.40 1.051 7.195 116.3 1.17 9
9 1
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9
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23A5948 REV. 1 8.
SELECTED MARGIN IMPROVEMENT OPTIONS (S.5.11
)
Recire.i.ation Pump Trip:
Yes Rod Withdrawal Limiter:
Yes Thermal Power Monitor:
Yes Improved Scram Time:
No
}
Exposure Dependent Limits:
No Exposure Points Analyzed:
1 (EOC) 9.
OPERATING FLEXIBILITY OPTIONS (S.5.2) 3 Single Loop Operation:
Yes Load Line Limit:
No Extended Load Line Limit:
No Increased Core Flow:
Yes
)
Flow Point Analyzed, 1:
105 Feedvater Temperature Reduction:
Yes ARTS Program:
No Maximum Extended Operating Domain:
Yes l
10.
CORE VIDE ADO ANALYSIS RESULTS (S.2.21 Methods Used: GEMINI and GEXL PLUS 3
(% NBR) (% NBR) BP8x8R GE8x8EB Figure Exposure Range: BOC2 to EOC2 Pressure Regulator 146 104 0.06 0.06 2
Failure Downscale loss of 100'T Feedwater 0.11 0.11 Heating Feedwater Controller 204 112 0.09 0.10 3
Failure (143%)
lead Rejection Without 237 106 0.08 0.07 4
Bypass
- See Appendix B
- Limiting Values Listed.
See Appendix E.
9 1
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O 23A5948 REV. 1 11.
LOCAL ROD WITHDRAWAL ERROR (VITH LIMITING INSTRUMENT FAILURE)
SUMMARY
(S.2.2.1.5)
O-The generic bounding BVR/6 Rod Withdrawal Error analysis described in NEDE-24011 P A 9 US and GESSAR-II Appendix 15B is applied; the resulting delta CPR is 0.11.
The original generic ' analysis in GESSAR II was not applicable for control cell core operation; however, 9:
it was subsequmtly shown to be applicable for control cell core operation and GESSAR II was revised to reflect this application in Revision 21. A plant cycle specific verification was performed to show the validity of the delta CPR with GEXL PLUS.
O, 12.
CYCLE MCPR VALUES (4.3.1. and S.2.2)*
Non Pressurization Events O
Exposure Range: BOC2 to EOC2 BP8x8R GE8x8EB e
Loss of 100 F Feedwater Heating 1.18 1.18 Rod Withdrawal Error 1.18 1.18 Pressurization Events g
Exposure Range: BOC2 to EOC2 Ootion A BP8x8R GE8x8EB g
Fressure Regulator Failure Downscale 1.14 1.14 Feedwater Controller Failure 1.17 1.18 Load Rejection Vithout Bypass 1.15 1.15 g
- GEMINI ODYN adjuster.ent factors are provided in the letter from J.S.
Charnley (CE) to M.W. Hodges (NRC), "GEMINI ODYN Adjustment Factors for BVR/6," dated July 6,1987.
g 10 0
23A5948 REU. 1 13.
OVERPRESSURIZATION ANALYSIS
SUMMARY
(S.3)
Steam Line vessel
)
Pressure Pressure Plant Transient (nsiri h
Rennense MSIV Closure 1235 1266 Figure 5 (Flux Scram) 14.
LOADING ERROR RESULTS (S.2.2.3.7)
Loading Error Results are not applicable for BVR/6 plants. NRC approval of the non applicability of leading Errors to BVR/6 plants is
)
documented in Section S.2.2.3.7 of NEDE 24011 P A 9 US.
". 5. CONTROL ROD DROP ANALYSIS RESULTS (S.2.2.3.1)
Banked Position Withdrawal Sequence is utilized at the Perry Nuclear Power Plant Unit 1; therefore, the bounding Control Rod Drop Analysis (CRDA) described in NEDE 24011 P A 9 US is applied. NRC approval of the bounding analysis is given in the letter to J.S. Charnley (GE),
"Acceptance for Referencing of Licensing Topical Report NEDE 24011, Revision 6 Amendment 9 'GESTAR II General Electric Standard Application for Reactor Fuel'," January 25, 1985.
()
16.
STABILITY ANALYSIS RESULTS (S.4) l i
l CE SIL 380 recommendations have been included in the Perry Nuclear l
Power Plant Unit 1 Technical Specifications; therefore, no stability
)
analysis is required as documented in the letter Cecil 0. Thomas (NRC) to H. C. Pfefferlen (CE), "Acceptance for Referencing of Licensing Topical Report NEDE 24011 Rev. 6, Amendment 8, ' Thermal Hydraulic Stability Amendment to CESTAR II'," April 24, 1985.
)
11 i
23A5948 RM.1 0
17.
IDSS-OF COOIANT ACCIDENT RESULTS (S.2.2. 3.2)
LOCA Method Used: SAFE /REFLCOD (see Perry Nuclear Power Plant Unit 1 Updated Safety Analysis Report) 4:
Fuel Type: BS301E Average MAPMGR (kW/ft)
Planar Exposure Most LRA31 0xidation O
(CVd/MT)
Limitine Limitina PCT ( F)
Fraction 0.0 12.4 12.4 2105
<.06 1.1 12.4 12.4 2100
<.06 5.5 12.8 12.9 2097
<.06 8.8 13.2 13.3 2146
<.06 11.C 13.5 13.5 2182
.062 13.8 13.5 13.5 2187
.063 22.0 12.5 12.5 2041
<.06 0
38.6 10.4 10.4
<1800
<.06 49.6 8.9 8.9
<1660
<.06 55.1 6.9 6.9
<1570
<.06 Fuel Type:
BS301F OI Average MAPMGR (kW/ft)
Planar Exposure Most Least Oxidation (CVd/MT)
Limitine Limitine PCT ( F)
Fraction 0.0 12.6 12.6 2131
<.06 g,
5.5 12.9 13.0 2121
<.06 7.7 13.2 13.2 2135
<.06 11.0 13.5 13.5 2176
.061 13.8 13.5 13.5 2184
.063 g:
22.0 12.5 12.5 2037
<.06 27.6 11.8 11.8 1950
< 06 38.6 10.4 10.4
<1800
<.06 49.6 8.9 8.9
<1660
<.06 i
g 55.1 6.9 6.9
<1570
<.06
- MAPMCR multiplier for single loop operation is 0.80 9:
12 i
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- "l'M C - BP8 SRB 219 Figure 1.
Reference Core leading Pattern
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Figure 2.
Plant Response to Pressure Regulator Failure Downscale (BOC2 to EOC2)
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Plant Response to Teodwater Controller Failure l
(BOC2 to EOC2)
)
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O Figure 4 Plant Response to Generator load Rejection Without Bypass (BOC2 to EOC2) 0 16 0
o 23A5948 REV. 1 m/
l
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1 i = Us em,.U.
i niin eqtii eiucesis 2 a t t SJIf 4:1 844T FLut 2 SAflf' 54 W1 FLOT I
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Plant Response to MSIV Closure (Flux Scram) 0 17/18 O
23A5948 REV. 1 U
APPENDIX A ANALYSIS CONDITIONS
-Q.
To accurately reflect actual plant parameters, the values shown in Table A 1 were used instead of the values reported in NEDE 24011-P A 9-US, September, 1988 for A00 a.. lyses,
'O l
Table A 1 PLANT PARAMETER O
Parame'/h, Analysis Value Thermal Power, MVt 3579 6
Rated Steam Flow, 10 lb/hr 12.57 6
O Core Flow, 10 lb/hr 109.2 Dome Pressure, psig 1005 Turbine Pressure, psig 969 Inlet Enthalpy, Btu /lb 512,4 N n Fuel Power Fraction 0.038 O
Dual Mode Safety / Relief Valves
"""b
' "*1"**
17 O
Relief Mode Lowest Setpoint, psig 1143*
Safety Mode Lowest Setpoint, psig 1177 Capacity,1b/hr 840,448 lO (R8f Pr*88ur8 Psis)
(1080) i l
l l
- O
- Assumes two lowest setpoint Safety / Relief Valves out of service,
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19/20 i
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23A5948 REV. 1 O
APPENDIX B BASIS FOR ANALYSIS OF IDSS OF FEEDWATER HEATING EVENT O
The Loss of Feedwater Heating event was analyzed with the 3D BWR Simulator code described in NEDE 24011 P A 9 US.
The tr'ansient analysis inputs normally reported in Section 6 of the licensing submittal are O
internally calculated in the 3D BVR Simulator code and in ODYN. Tha transient plots, flux, and Q/A normally reported in Section 10 are not outputs of the 3D BVR Simulator code; therefore, these items are not included in this document.
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23A5948 REV. 1 O
APPENDIX C APPLICATION OF CEMINI METHODS O
The GEMINI system of methods are used to perform the licensing analyses of Perry Nuclear Power Plant Unit 1 (PNPP) Reload 1.
The CEMINI system of methods is described in Reference 1; NRC approval of these methods is documented in Reference 2.
In Reference 3, the application of nV CEMINI methods in licensing analyses is describt.d. Pressurization events that could establish the Cr.ating Limit MCPR are analyzed at the 100%
power level.
Power level uncertainties specified in Regulatory Guide 1.49 are accounted for by adding adjustment factors to the calculated delta CPR.
.O NRC approval of this procedure is provided in Reference 4.
The CEMINI system of methods have been incorporated into the approved CESTAR II licensing topical repert, Reference 5.
O Rod Withdrawal Error l
l As described in Section 11, the generic Rod Withdrawal Error analysis for BWR/6 plants is applied to PNPP Reload 1.
An evaluation of the impact O
of CEMINI methods on the generic analysis indicates that the results of the generic analysis continue to be conservative and bounding.
Overoressurization Analysis
'O The MSIV Closure (Flux Scram) analysis is performed using CEMINI methods at the 102% power level to account for the power level uncertainties specified in Regulatory Guide 1.49.
The analysis was O
performed with 13 highest setpoint safety valves operational.
Control Rod Droe Accident O
The NRC approved bounding Control Rod orop Accident analysis for banked Position Withdrawal Sequence plants (such as PNPP) described in Reference 5 is applied to PNPP Reload 1.
The impact of CEMINI methods on the results of the generic analysis is negligible.
O 23 0
23A5948 REV. 1 O
Stability The NRC approved generic stability approach described in Section 16 is applied to PNPP Reload 1.
The use of CEMINI methods does not impact the
~
generic analysis.
References 9
1.
Letter, J.S. Charnley (CE) to C.O. Thomas (NRC), "Amendment 11 to CE LTR NEDE 240ll P A," February 27, 1985, 2.
Letter, C.O. Thomas (NRC) to J.S. Charnley (GE), "Acceptance for l
Referencing of Licensing Topical Report NEDE 24011 P A, Rev. 6, gg Amendment 11, ' General Electric Standard Application for Reactor Fuel'," November 5, 1985.
3.
Latter, J.S. Charnley (CE) to H.N. Berkov (NRC), "Revised
]
Supplementary Information Regarding Amendment 11 to GE Licensing Topical Report NEDE 24011 P A," January 16, 1986.
gg 4.
Letter, G.C. Lainas (NRC) to J.S. Charnley (GE), "Acceptance for l
Referencing of Licer*ing Topical Report NEDE 24011 P A, 'GE Generic Licensing Reload Repot.', Supplement to Amendment 11," March 22,1986.
5.
"GESTAR II, General Electric Standard Application for Reactor Fuel,"
NEDE 24011 P A 9.JS, September 1988.
gg I
GD '
1 I
GD l d>
er 24 0
23A5948 REV. 1 g
APPENDIX D GEXL PWS CORRELATION The analyses performed for this cycle utilized the,CEXL PWS thermal correlation. The cycle 1 analyses used the CEXL correlation. The CEXL PWS correlation is described in Reference 1.
NRC approval of this 4
correlation is documented in Reference 2.
Reference 3 describes the application of the CEXL-PWS correlation which was approved by the NRC in Reference 4.
This application requires an adjustment to the MCPR values g
for bundle flows below 0.5 M1b/sq ft hr and incorporation of a 31 O
adjustment factor if inlet subcooling.xe..ds 70 stu/1da. The CExL PWS correlation has been incorporated into Rsference 5.
I References O
l 1.
latter, J. S. Charnley (CE) to C. O. Thomas (N"C), "Amendment 15 to l
l General Electric Licensing Topical Report NEDE 44011 P A," January 23 I
l 1986.
l 2.
Letter, Ashok C. Thadani (NRC) to J. S. Charnley (CE), "Acceptance for lO Referencing of Amendment 15 to General Electric Licensing Topical Report NEDE 24011 P A, ' General Electric Standard Application for Reactor Fuel'," March 14, 1988.
l 3.
Letter, J. S. Charnley (CE) to M. W. Hodges (NRC), "Application of CESTAR II Amendment 15," March 22, 1988.
O 4
Letter, Ashok C. Thadani (NRC) to J. S. Charnley (CE), "Acceptance for Referencing of Application of Amendment 15 to General Electric l
Licensing Topical Report NEDE 24011 P A.
' General Electric Standard Application for Reactor Fuel'," May 5, 1988.
5.
"GESTAR II, Cennal Elutric Standard Appilcation for Reactor Fuel,"
O NEDE 24011 P A 9 and NEDE 24011 P A 9 US, September 1988 l
O 0
25/26 0
O 23A5948 REV. 1 APPENDIX E O
ANALYZED OPERATING DOMAIN The core wide abnormal operational occurrence (A00) analysis results reported in Section 10 are the most limiting values over the entire
- b
- perating rar.ge. This' range covers the following operating O
options:
1.
Standard 100% power / flow map;
'O 2.
End of cycle power coastdown; l
3.
MEOD with 100% poser flow range from 75% to 105% of rated; and IO i
4.
Partial feedvater heating to 320*F durgns the cycle with finsi l
feedwater temperature reduction to 250 F after "All Rods Out' at l
l end of cycle.
1
,0 Limiting events and e notet n analyzed an based on Refennee 1 and l
the USAR analytical results. The reload 1, cycle 2. analyses were l
perforned assuming all four turbine control valves in a full are mode of operation. This is conservative for partial arc configuration.
O The Single Loop Operation analysis was performed for the standard power / flow map with normal feedwater temperature.
O Refereneu 1.
"General Electric Standard Application for Reactor Fuel,"
NEDE.24011.P A 9.US, September 1988.
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27/28 (final)
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8 4
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