ML17304B406
| ML17304B406 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 08/25/1989 |
| From: | Conway W ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 161-02219-WFC-K, 161-2219-WFC-K, NUDOCS 8908300018 | |
| Download: ML17304B406 (38) | |
Text
o o REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:8908300018 DOC.DATE: 89/08/25 NOTARIZED: NO DOCKET 4s FACIL:STN-50-528 Palo Verde Nuclear Station, Unit 1, Arizona Publi 05000528 AUTH.NAME AUTHOR AFFILIATION CONWAY,W.F.
Arizona Public Service Co.
(formerly Arizona Nuclear Power RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control: Desk)
SUBJECT:
Forwards revised Section 7 to Cycle 3 reload analysis rept for plant.
DISTRIBUTION CODE:
IE26D COPIES RECEIVED:LTR ENCL SIZE:
TITLE: Startup Report/Refueling Report (per Tech Specs)
NOTES':
05000528 RECIPIENT ID CODE/NAME PD5 LA CHAN,T INTERNAL: ACRS NRR CQA~ERTON gXG FILE OR RCH2 DRSS/EPRPB EXTERNAL: LPDR NSIC NOTES:
COPIES LTTR ENCL 1
0 2
2 5
5 1
1 1
1 1
1 1
1 1
1 1
1 RECIPIENT ID CODE/NAME PD5 PD DAVIS,M.
IRM TECH. ADV NUDOCS-ABSTRACT RGN5'ILE 01 NRC PDR COPIES LTTR ENCL 1
1 2
2 1
1 1
1 1
1 1
1 TOTAL NUMBER OF COPIES REQUIRED:
LTTR 21 ENCL 20
Qr IO
Arizona Public Service Company P.O. BOX 53999
~
PHOENIX. ARIZONA85072-3999 WILLIAMF. CONWAY
'XECUTIVE VICEPRESIDENT NUCLEAR 161-02219-WFC/KLMC August 25, 1989 Docket No.
STN 50-528 Document Control Desk U.
S. Nuclear Regulatory Commission Mail Station Pl-37 Washington, D.
C.
20555
Reference:
Letter to the NRC from D. B. Karner, ANPP, dated January 18, 1989;
Subject:
Reload Analysis Report for Unit 1, Cycle 3 (161-01620)
Dear Sirs:
Subj ect:
Palo Verde Nuclear Generating Station (PVNGS)
Unit 1 Revised Section 7 of Reload Analysis Report for Unit 1, Cycle 3
File: 89-E-056-026 Attached is a revision to Section 7 to the Unit 1, Cycle 3, Reload Analysis Report.
The Unit 1, Cycle 3,
Reload Analysis Report was transmitted by the Reference.
The revised Section 7 pages should be incorporated into the Reload Analysis Report and the previously transmitted Section 7 pages discarded.
The revision contains the reanalysis of the lnadvertant Opening of an Atmospheric Dump Valve with a concurrent Loss of Off-Site Power.
The results of this re-analysis are bounded by the reference cycle.
Sincerely, WFC/KLMC/jle Attachment cc:
J.
B. Martin T. L. Chan T. J. Polich A. C. Gehr M. J.
Davis (all w/attachment)
I
7.0
. NON-LOCA.SAFETY ANALYSIS 7.0.1 Introduction This section presents the results of the Palo Verde Nuclear Generating Station Unit I (PVNGS-I), Cycle 3 Non-LOCA safety analyses at 3800 HWt.
The Design Basis Events (DBEs) considered in the safety analyses are listed in Table 7.0-1.
These events are categorized into three groups:
Hoderate Frequency, Infrequent, and Limiting Fault events.
For the purpose of this report, the Hoderate Frequency and Infrequent Events will be termed Anticipated Operational Occurrences.
The DBEs were evaluated with respect to four criteria:
Offsite Dose, Reactor Coolant System (RCS) Pressure, Fuel Performance (DNBR and Centerline Helt SAFDLs),
and Loss of Shutdown Hargin.
Tables 7.0-2 through 7.0-5 present the lists of events analyzed for each criterion.
All events were re-evaluated to assure that they meet their respective criteria for Cycle 3.
The DBEs chosen for analysis for each criterion are the limiting events with respect to that criterion.
7.0.2 Hethods of Analysis The analytical methodology used for PVNGS-} Cycle 3 is the same as the Cycle 2 (Reference Cycle) methodology (References 7-1, 7-2 and 7-9) unless otherwise stated in the event p'resentations.
Only methodology that has previously been reviewed and approved on the PVNGS dockets (References 7-10 and 7-11), the CESSAR docket (Reference 7-2), or on other dockets is used.
7-1
ll
H.-thematical Hodels The mathematical models and computer -codes used in the Cycle 3
Non-LOCA safety analysis are the same as those used in the Reference Cycle analysis (References 7-1, 7-2 and 7-9).
Plant response for Non-LOCA Events was simula.ed using the CESEC III computer code (Reference 7-3).
Simulation of the fluid conditions within the hot channel of the reactor core and calculation of DNBR was performed using the CETOP-D computer code described in Reference 7-4.
The TORC computer code was used to simulate the fluid conditions within the reactor core and to calculate fuel pin DNBR for the RCP Shaft Seizure and Sheared Shaft event, The TORC code is described in References 7-6 and 7-7.
The number of fuel pins predicted to experience clad failure is taken as the number of pins which have a
CE-1 DNBR value below 1.24.
The only exceptions are the Shaft Seizure and Sheared Sha t events for which the sta.istical convolution method, described in Reference 7-8, was used.
Reference 7-8 has been approved by the NRC and has been used in References 7-1, 7-2 and 7-9.
The HERNITE computer code (Reference 7-5} was used to simulate the reactor core for analyses which required more spatial detail than is provided by a point kinetics model.
Reference 7-5 has been approved by the NRC and has been used in References 7-1, 7-2 and 7-9.
HERMITE was also used to generate input to the CESEC point kinetics model by partially crediting space-time effects so that the CESEC calculation of core power during a reactor scram is conservative relative to HERMITE.
7-2
~
0
-II
~
y'
IQ
7.0.4 n ut Parameters and Analvsis possum tions Table 7.0-6 summarizes the core parameters assumed in the Cycle 3
transient analysis and compares them to the values used in the Reference Cycle.
Specific initial conditions for each event are tabulated in the section of the report summarizing that event.
Tech Spec changes are described in Section 10.
The effects of these changes were considered for each DBE and were included as appropriate.
For some of the DBEs presented, certain initial core parameters were assumed to be more limiting than the actual calculated Cycle 3 values.
Such assumptions resulted in more adverse consequences.
Events which have credited CPC trip protection have assumed instrument channel response times which are conservative relative to the Cycle 3 Technical Specifications.
7.0.5 Conclusion All DBEs have been evaluated for PYHGS-1, Cycle 3 to determine whether their results are bounded by the Reference Cycle.
ii
Table 7.0-1 PVNGS Unit 1 Desion Basis Events Co.sidered in tne Cycle 3 Safety Analysis 7.1 Increase in Heat Removal'y the Secondary System 7.1.1 i.i.2 7.1.3 7.1.4 71 "*
Decrease in Feedwater Temperature Increase in Feedwater Flow Increased Main Steam Flow Inadvertent Opening of a Steam Generator Safety Valve or Atmospheric Dump Valve Steam Svs-em Piping Failures
'.2 Decrease in Heat Removal by the Secondary System 7.
2..'.2.2 7.2.3 7.2.4 7.2.5 7.2.6" Loss o
External Load Turbine Trip Loss of Condenser Vacuum Loss of Normal AC Power Loss of Normal Feeowater Feedwater System Pipe Breaks 7.3 Decrease in Reactor Coolant F)owrate 7.3.i 7
3'x Total Loss of Forced Reactor Coolant Flow Single Reactor Coolant Pump Shaft Seizure/Sheared Shaft 7.4 Reactivity and Power Distribution Anomalies 7.4.i 7.4.2 7.4.3 7.4.4 7.4.5 7.4.6*
Uncontrolled CEA 'k'ithdrawal from a Subcri ical or Low Power Condition Uncon-rolled CEA Mi hdrawal a
Power C.A Nisopera ion Even".s CVCS Nalfunction (Inadvertent Boron Dilution)
Star-up o
an inac ive Reactor Coolant System Pump Control Element Assembly Ejection 7.5 increase in Reactor Coolant System inventory 7.5.i 7.5.2 CVCS Yialfunc ion Inadvertent Opera ion of the ECCS During Power Operation
- Categorized as Limiting Fault
=vents
i~
Table 7.0-1 (continued) 7.6 Decrease in Reactor Coolant System 1nventory 7.6.1 7.6.2" 7.6.3" Pressurizer Pressure Decrease
'Events Small Primary Line Break Outside Containment Steam Generator Tube Rupture 7.7 Miscellaneous 7.7. 1 Asymmetric Steam Generator Events
" Cateaorized as Limiting Fault Events 7-0,
ii 10
Table 7.0-2 DBioe Breaks Tne results are bounded by the Reference Cycle.
7.3 DECREAS IN REACTOR COOLANT FLOWRATE 7.3.1 Loss of Forced Reactor Coolant The resulis are bounded by the Reference Cycle.
7.3.2 Single Reac or Coolant Pump Shaft Seizure/Sheared Sha t ihe resul.s are bounded by the Reference Cycle.
7.4 REACTI'PITY AND POk'ER DISTRIBUTION ANOMALIES 7.4.1 Uncon.rolled CEA V'drawal from a Subcri.ical or Low Power Condition The results are bounded by ihe,Reference, Cycle.
o<i 7.4.2 Uncontrolled CEA Vi hdrawal a
Power Tne resulis are bounded by the Reference Cycle.
7.4.3 CEA Misooeration
- Even, The results are bounded by the Reference Cycle.
7-14
i lk
S-artuo of an Inac.ive Reactor Coolan:
Pump Evert The results are bounded by the Reference Cycle.
7.4.6 Control Element Assembly Eiection The results are bounded by the Reference Cycle.
INCREASE IN R.ACTOR COOLANT SYSTEN INVENTORY 7.5.1 CVCS Malfunction The results are bounded by the Reference Cycle.
7.5.2 Inadverteni Opera-.ion of the ECCS Durinc Power Operation The results are bounded by the Refer ence Cycle.
7.6 DECREASE IN R"ACTOR COOLANT SYSTEYi INVENTORY 7.6.1 Pressurizer Pressure Decrease Events The resulis are bounded by the Reference Cycle.
7,6,2 Small Primary Line Pine Break Outs ide Coniainment The results are bounded by the Reference Cycle.
7.6.3 Sieam Genera-or Tube Rupture The resul is are bounded by the Reference Cycle.
7-15
(i
7.7 I;!SC=~ LAu:OuS 7.7.j Asymmetric Steam Generator Events The resu1ts are bounded by the Reference Cyc1e.
7-16
i
-0. 5
-1,0
-1 5 ALL0HAt3Lt;
-2.0 Z 5
-3.0
-3.5 20 00 100 I
COOr; VOet:.O LrVLL, X Ot-Wlrn tltOWAL t OH+
Fato Verde Nutkrr Cmecatbg Station hygro)g/Nfl-: 1(fC HOOKS 1 ht(0 2 t'Igttre 7.0-/
iy C
0'