ML17264A240
| ML17264A240 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 11/22/1995 |
| From: | CONSOLIDATED EDISON CO. OF NEW YORK, INC. |
| To: | |
| Shared Package | |
| ML17264A239 | List: |
| References | |
| NUDOCS 9511270297 | |
| Download: ML17264A240 (52) | |
Text
R.E.
Ginna Nuclear Power Plant Core Operating Limits Report Cycle 25 Draft B
Note:
This report is not part of the Technical Specifications.
This report is referenced in the Technical Specifications.
9511270297 951122 PDR ADaCK, 05000244 P.
TABLE OF CONTENTS 1.0 CORE OPERATING LIMITS REPORT...................
2 2.0 OPERATING LIMITS 2.1 SHUTDOWN MARGIN.
2.2 MODERATOR TEMPERATURE COEFFICIENT 2.3 Shutdown Bank Insertion Limit.
2.4 Control Bank Insertion Limits.
2.5 Heat Flux Hot Channel Factor (Fo) 2.6 Nuclear Enthalpy Rise Hot Channel Factor (F"~)
2.7 AXIAL FLUX DIFFERENCE 2.8 RCS Pressure, Temperature, and Flow Departure from Boiling (DNB) Limits 2.9 Boron Concentration
~
~
~
0 Nucl cate 3
3 3
3 4
4 4
4 5
5 3.0 UFSAR CHAPTER 15 ANALYSIS SETPOINTS AND INPUT PARAMETERS.....
5
4.0 REFERENCES
6 FIGURE 1 - RE(VIREO SHUTDOWN MARGIN.
7 FIGURE 2 -
CONTROL BANK INSERTION LIMITS................
8 FIGURE 3 K(Z)
NORMALIZED Fo(Z)
AS A FUNCTION OF CORE HEIGHT o
~
~
~
~
9 FIGURE 4 - AXIAL FLUX DIFFERENCE ACCEPTABLE OPERATION LIMITS AND TARGET BAND LIMITS AS A FUNCTION OF RATED THERMAL POWER 10 TABLE 1 -
UFSAR CHAPTER 15 ANALYSIS SETPOINTS AND INPUT PARAMETERS.....
11 COLR Cycle 25, Draft B
lI h
A'
R.E.
Ginna Nuclear Power Plant Core Operating Limits Report Cycle 25 Draft b 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report (COLR) for Ginna Station has been prepared in accordance with the requirements of Technical Specification 5.6.5.
The Technical Specifications affected by this report are listed below:
3.1.1 3.1.3 3.1.5 3.1.6 3.2.1 3.2.2 3.2.3 3.4.1 3.9.1 "SHUTDOWN MARGIN (SDH)"
"MODERATOR TEMPERATURE COEFFICIENT (HTC)"
"Shutdown Bank Insertion Limit" "Control Bank Insertion Limits" "Heat Flux Hot Channel Factor (Fo)"
"Nuclear Enthalpy Rise Hot Channel Factor (F">>)"
"AXIAL FLUX DIFFERENCE (AFD)"
"RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits" "Boron Concentration" COLR Cycle 25, Draft B
'$k, J~
2.0 OPERATING LIMITS The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the following subsections.
These limits have been developed using the NRC approved methodologies specified in Technical Specification 5.6.5.
All items that appear in capitalized type are defined in Technical Specification l. 1, "Definitions."
- 2. 1 SHUTDOWN MARGIN (LCO 3. 1. 1)
(Limits generated using Reference 1) 2.1.1 2.1.2 The SHUTDOWN MARGIN in MODE 2 with K,<< < 1.0 and MODES 3 and 4 when the reactor coolant pumps are OPERABLE and in operation shall be greater than or equal to the limits specified in Figure 1.
The SHUTDOWN MARGIN in MODE 4 when the reactor coolant pumps are not OPERABLE and in MODE 5 shall be greater than or equal to the one loop operation curve of Figure l.
2.2 MODERATOR TEMPERATURE COEFFICIENT (LCO 3. 1.3)
(Limits generated using Reference 1) 2.2.1 The Moderator Temperature Coefficient (MTC) limits are:
The ARO/HZP -
MTC shall be less positive than +5.0 pcm/'F for power levels below 70%
RTP and less than or equal to 0 pcm/'F for power levels at or above 70% RTP.
The ARO/RTP -
MTC shall be less negative than -42.9 pcm/'F.
where:
ARO stands for All Rods Out HZP stands for Hot Zero THERMAL POWER RTP stands for RATED THERMAL POWER 2.3 Shutdown Bank Insertion Limit (LCO 3. 1.5)
(Limits generated using Reference 1) 2.3.1 The shutdown bank shall be fully withdrawn which is defined as
> 221 steps.
COLR Cycle 25, Draft B
E Il N
t
)1
2.4 Control Bank Insertion Limits (LCO 3. 1.6)
(Limits generated using Reference 1) 2.4.1 2.4.2 The control banks shall be limited in physical insertion as shown in Figure 2.
The control banks shall be moved sequentially with a 100
(+5) step overlap between successive banks.
2.5 Heat Flux Hot Channel Factor FQ (LCO 3.2. 1)
(Limits generated using References 1 and 2) 2.5.1 Fo(Z) s ~F J"K(Z)
P FQ(Z) s QF(J"K(Z) 0.5 when P
> 0.5 when P s 0.5 where:
Z is the height in the core, Fo = 2.32, K(Z) is provided in Figure 3, and THERMAL POWER P
RATED THERMAL POWER 2.6 Nuclear Enthal Rise Hot Channel Factor F"
(Limits generated using Reference 1) 2.6.1 F"~ s F" ~* (1 + PF~ * (1-P))
where:
F BTP PF~
= 0.3, and THERMAL POWER P
=
RATED THERMAL POWER 2.7 AXIAL FLUX DIFFERENCE (LCO 3.2.3)
(Limits generated using References 1 and 3) 2.7.1 2.7.2 The AXIAL FLUX DIFFERENCE (AFD) target band is + 5%.
The AFD acceptable operation limits are provided in Figure 4.
COLR Cycle 25, Draft B
0 S
2.8 RCS Pressure Tem erature and Flow De arture from Nucleate Boilin
~Olla Ii l<<(LCG 3.4.1)
(Limits generated using Reference 4) 2.8.1 2.8.2 2.8.3 The pressurizer pressure shall be z 2205 psig.
The RCS average temperature shall be s 577.5 F.
The RCS total flow rate shall be ~ 170,200 gpm.
2.9 Boron Concentration (LCO 3.9. 1)
(Limits generated using References 1 and 5) 2.9. 1 The boron concentrations of the hydraulically coupled Reactor Coolant System, the refueling canal, and the refueling cavity shall be a 2000 ppm.
3.0 UFSAR CHAPTER 15 ANALYSIS SETPOINTS AND INPUT PARAMETERS The setpoints and input parameters for the UFSAR Chapter 15 accident analyses arepresented in Table 1.
The values presented in this table are organized based on system and major components within each system.
The failure of a component or system to meet the specified Table 1 value does not necessarily mean that the plant is outside the accident analyses since:
(1) an indicated value above or below the Table 1
values may be'ounded by the Table 1 values, and (2) the setpoint or parameter may not significantly contribute to the accident analyses final results.
The major sections within Table 1 are:
1.0 Reactor Coolant System (RCS) 2.0 Hain Feedwater (HFW) 3.0 Auxiliary Feedwater (AFW) 4.0 Hain Steam (HS) System 5.0 Turbine Generator (TG) 6.0 Chemical and Volume Control System (CVCS) 7.0 Emergency Core Cooling System (ECCS) 8.0 Containment 9.0 Control Systems 10.0 Safety System Setpoints 11.0 Steam Generators COLR Cycle 25, Draft B
1.>
4.0 REFERENCES
2.
3.
4, 5.
WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, July 1985.
WCAP-9220-P-A, Westinghouse ECCS Evaluation Model-1981 Version, Rev.
1, February 1982.
OR WCAP-10054-P-A and WCAP-10081, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUHP Code," August 1985.
WCAP-10924-P-A, Volume 1, Rev.
1, and Addenda 1,2,3, "Westinghouse Large-Break LOCA Best-Estimate Methodology, Volume 1:
Model Description and Validation," December 1988.
WCAP-10924-P-A, Volume 2, Rev.
2, and Addenda, "Westinghouse Large-Break LOCA Best-Estimate Methodology, Volume 2:
Application to Two-Loop PWRs Equipped with Upper Plenum Injection," December 1988.
WCAP-10924-P-A, Rev.
2 and WCAP-12071, "Westinghouse Large-Break LOCA Best Estimate Methodology, Volume 2:
Application to Two-Loop PWRs Equipped With Upper Plenum Injection, Addendum 1:
Responses to NRC questions,"
December 1988.
WCAP-10924-P, Volume 1, Rev.
1, Addendum 4, "Westinghouse LBLOCA Best Estimate Methodology; Model Description and Validation; Model Revisions,"
August 1990.
WCAP-8395, "Power Distribution Control and Load Following Procedures-Topical Report," September 1974.
WCAP-11397-P-A, "Improved Thermal Design Procedure",
April 1989.
WCAP-11596-P-A, "qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores,"
June 1988.
COLR Cycle 25, Draft B
3
~O d 2 t9 g1 0
(1600, 1A5)
(1500. 1')
ACCEPTABLE OPERATION ONE LOOP OPERATION nVOLOOP OPERATION UNACCEPTABLE OPERATION (0, 2.45)
(0, AS) 1000 500 COOLANTBORON CONCENTRATION (ppm)
FIGURE I REQUIRED SHUTDOWN MARGIN COLR Cycle 25, Draft 6
I P
yt pL
220
~> 200
~~ 180 5 <6O
~ 140 izo
~o
$ 00 o
80 60 40
<o 20 0
0 8 Bank (100, 1')
G Bank 0 Bank
$ 0 20 30 40 50 60.
70 80 SO
'l00 Core Power (Percent of 1520 lVlVVT)
The fullywithdrawn position is defined as ~222 steps.
FIGURE 2 CONTROL BANK INSERTION LIMITS COLR Cycle 25, Draft B
$.25 0.75 Total Fq = 2.320 0.25 Core Height 0.00 6.00 50.80
$ 2.00 K(Z)
'1.000
$.000 0.940 0.647 0
0 4
6 8
CORE HEfeHT {ft) 10 12 FIGURE 3 K(Z) " NORMALIZED FQ(Z)
AS A FUNCTION OF CORE HEIGHT COLR Cycle 25, Draft B
~
~ ~ OO ~
~1t
~
~
( 11 80)
DO NOT OPERATE IN THIS AREA WITHAFD OUTSIDE'1HE TARGEr BAND (11/0) 80 0
60 40 ACCEPl'ABLE OP ERAT)ON VATHAFD
'-- -'- -----OurSIOEmc TARGET BAND N)THc1 HR PENALTY 9EMATION TIME
~
tt (41,60)
P1$ 0)
ACCEPfABLE OPERATION 20 0
-20
-10 0
10 20 30 AXIALFLUXDIFFERENCE (%)
FIGURE 4 AXIAL FLUX DIFFERENCE ACCEPTABLE OPERATION LIMITS AND TARGET BAND LIMITS AS A FUNCTION OF RATED THERMAL POWER COLR 10 Cycle 25, Draft B
I II rt t,
Item 4 Xtem/Name Value Remarks:
1.0 Reactor Coolant System (RCS)
Upper head volune, ft Upper Plenun volwe, ft Top of fuel voiune, fts Inlet nozzle(s) volune, total of two, ft'utlet nozzle(s) voiune, total of two, ft Active fuel volwe, fts Bottom of fuel volune, ft Lower Plenun voiwe, ft Downcomer volune, above bottom of cold leg, fts Downcomer, lo~er core plate to elevation of the bottom of the cold leg voiune, ft'arrel baffle, lower core plate to upper core plate volwe, ft Total volune, ft Hot leg pipe volune per loop voiwe, ft Cold leg voiwe per loop + cross over, fts RC Ixmp voiune per pump, ft Cold leg pipe ID, in./Pump suction ID, in.
Hot leg pipe ID, in.
Design pressure, psig Design temperature, F
Cold Leg and Hot Leg Centerline Elevation 300.0 580.2 50.3 43.2 37.4 367.6 11.0 514.3 138.4 278.2 128.5 2449.1 78.7 cross over = 140.7
~
cold leg -"46.8 192 27.5/31 29 (28.969) 2485 650 246'0" Above upper support plate.
Bottom of upper core plate to top of upper support plate.
Includes outlet holes in the barrel.
Top of active fuel to bottom of upper core plate, inside barrel baffle.
Includes nozzle forging protrusion into vessel.
Does not include mating hole in barrel, this is included in the Upper Pienun vol+ac.
Bottom of fuel to top of fuel Top of lower core plate to bottom of active fuel.
Below top of lower core plate Above bottom of cold leg elevation to bottom of upper support plate Top of lower core plate to elevation of bottom of cold leg Top of lower core plate to bottom of upper core plate.
Includes nozzles Reactor Coolant Pmp Head-Capacity and HPSH curves for reactor coolant pumps/Homologous Curves Rated RC pump head and flow/ ft 8 gpll Rated RC pump torque and efficiency 0 rated head/flow, ft-lb, fraction RCP Punp Rated Power (hot, 556 degrees F)
RCP Hotor Rated Speed, RPH Homent of inertia of pwp and motor, lb-ft RC pump power, Hut (max/min)
See HSSL 252; 90,000 84K efficiency at hot conditions 4842 BHP 1189 80,000 10, 8 Homologous Curves are available in RETRAH Pwp power varies with RCS tery from approx 8 Hut to 10 Mwt 1.2 Core COLR 12 Cycle 25, Draft B
hhh
I..
Table 1:
UFSAR Chapter 15 Analysis Setpoints and Input Parameters Item g Item/Name Value Remarks:
Rated power, NlJt Reactor power uncertainty, X RTP Bypass prior to Thimble Plug Deletion, X
Upper head bypass, X
Upper head temperature, degrees F
Neat transfer area, ft Average core heat flux,Btu/hr-ft'520 4.5 M proprietary 596 26,669 189,440 Heeds to be updated to account for thimble plug deletion Temperature can be used to back out the upper head bypass flow 1.3 1.3.1 Fuel Asseaht ies Height Total, inches (length from bottom of assembly to top nozzle)
Fuel Rod Length, inches (length from bottom of pin to top of pin)
Active, inches 159.935 149.138 141.4 1.3.2 Fuel Assembly Geometry Ness of fuel, ibm Ness of clad, ibm Nwher of fuel pins per fuel assembly (FA)
No. of Fuel Assemblies Fuel pin pitch, in.
Bottom nozzle weight and volune Top nozzle, w/ insert, weight and volune Fuel Assembly resistance
[core dP f(flow)], psi f(lb/hr)
Fuel Assembly free flow area, in'05,500 25,927 179 121 0.556 9.1 lbs.
31.5 in'8.15lbs.
62.9 in core delta P = 20.4 psi 0 flow = 186400 gpn 34.75 Note:this value includes thimble plugs.
Thimble plugs are being removed.
No plug values are 20.0 at 170,200 single assembly 1.3.3 Fuel pin geometry Pellet diameter, in.
Clad 00/ID, in./in.
0.3444 0.400/0.3514 1.3.4 Control Rod 8 Instrunent Guide Tubes No. of control rod guide tubes Ho. of instrunent guide tubes Control Rod Guide tube upper part OD/ID, in./in.
Instrunent Guide tube OD/ID, in./in.
Guide tube lower part OD/ID, in./in.
Control Rod Drop Times, maximms, sec.
16 0.49/0.528 0.395/0.350 0.4445/0.4825 Hon-LOCA 2.4 LOCA 3.0 Allowances are added to the Tech Spec el lowable value.
COLR 13 Cycle 25, Draft B
0 0
~
0 Table 1:
UPSAR Chapter 15 Analysis Setpoints and Input Parameters Item 4 Item/Name Value Remarks:
Control rod maximla withdrawal rate, in./min.
Control rod maxim(m insertion rate, pcm/sec.
Control rod insertion limits Hot channel radial peaking factor Heat Flux Hot channel factor FO 45 90 See COLR 1.66 2.32 1.4 Pressurizer Code safety valve flow capacity, ibm/hr 288,000 Rating at 2485 psig plus 3X accuml at 1 on Code safety valve open time Code safety valve setpoint Spray valve n(snber Spray valve flow capacity, gpm/valve Spray valve setpoint-start open/full open Spray valve time constant, sec.
PORV number PORV flow capacity, ibm/hr PORV Cv 200 2260/2310 5
2 179,000 50 gpm/(psid)1/2 Proportional Assumed value Steam flow at 2335 psig Rating is for liquid relief. Valve characteristic is quick opening see Copes Vulcan Selecting and Sizing Control Valves 8/75, page 8, Table 18 for Cv vs travel curve.
0.8 sec seal clearing time Crosby Model HB-BP-86, size 4K26 2485 psig Tolerance is a 3X.
2 PORV open time PORV close time PORV setpoint
[normal] open/close, psig PORV setpoint tLTOP) open/close, psig PORV blowdown characteristic Heater capacity w/ bank capacity and setpoints, kw Control banks Backup Heaters Minimis heater capacity required for LOOP, kW Heater bank controller type 1.87 sec
+ transmitter 3.95 sec
+ transmitter 2335/2315 424 800 0 kw at 2250 psig and 400 kM at 2220 psig Full on at 2210 psig and resets at 2220 psig 100 proportional 400 kw LTOPs transmitter is Foxboro E11GM-HSAE1, with a time response of 1 sec (time to 90K of final value for step input)
LTOPs transmitter is Foxboro E11GM-HSAE1, with a time response of 1 sec (time to 90K of final value for step input) 1.4.1 Pressurizer vol(see(s)
(100X / OX power)
- Mater, fthm (100K / OX power)
- Steam, fthm (100X / OX power)
Total, ft Pressurizer IO, ft-in 396/199 404/601 800 83.624 in / cladding thickness is 0.188 in COLR Cycle 25, Draft B
N..
0 Table 1:
UFSAR Chapter 15 Analysis Setpoints and Input Parameters Item 4 Item/Name Value Remarks:
Surge line ID, in.
Spray line ID, in.
Surge line volune,ft'.75 3.062 18.4 Surge line is 10 in schedule 140 1.4.2 Pressurizer Level Lower level tap elevation Upper level tap elevation Pressurizer level vs X power Distance Hot Leg Centerline to Lower Tap, ft Haxinun level allowed for steam bubble, X
257' 275' Xpower Level 0 X 19.5X 100 X 49X 10.750 87 Pressurizer level is ramped linearly between these points.
1.5 RCS Flows, Temperature and Prcssurcs Total reactor coolant flow, gpm (15K plugging)
Total reactor coolant flow, gpm (15K plugging)
Average reactor coolant temperature, degrees F
(Full power/H2P)
Reactor coolant pressure, psig Reactor coolant flow uncertainty, X nominal Reactor coolant temperature uncertainty, degrees F
Reactor coolant pressure uncertainty, psi DHB Limit (safety analysis) 170,200 173,800 573 '/547 2235 3.1 s 30 1.52 typical cell 1.51 thimble cell Use for non DNB Use for statistical DNB 1.6 Low Temperature Overpressurc Protection (LTOP)
Hininun RCS vent size, square inches No. of S1 ixlnps capable of injection (PORVs/vent)
Maxinun pressurizer level for RCP start, X
0/1 38 1.7 Fuel Handling/Dose Calculations Maxinun reactor coolant gross specific activity Haxinun reactor coolant dose equivalent 1-131 Haxinun secondary coolant dose equivalent 1-131 Mininun reactor coolant boron concentration, ppm Hininun reactor coolant level Hininun spent fuel pool level Hininun spent fuel pool boron concentration, ppm Spent fuel pool temperature, degrees F (min/max)
Hininun spent fuel pool charcoal filter efficiency, X iodine removal Mininun post accident charcoal filter efficiency, X iodine removal 100/b pCi/gm 1.0 pCi/gm 0.1 pCi/gm 2000 23 ft above flange 23 ft above fuel 300 50/180 90 90 COLR 15 Cycle 25, Draft B
Table 1: UFSAR Chapter T5 Analysis Setpoints and Input Parameters Item g
Item/Name Value Remarks:
Hinimm control room charcoal filter efficiency, X iodine removal Hfnfszza time between reactor criticality and fuel movement, hrs.
Source Terms used for dose calculations Haxfnua Gas Decay Tank Xenon-133 concentration, C'0 100 TID 14844, MCAP-7823 100,000 2.0 Hain Fecdwater (HFM)
Feedwater temperature versus load Power Temperature 102X 425 F
70X 385 F
30X 322 F
OX 100 F
100X design temp fs 432 degrees F
Feedwater Suction Temperature vs Power, nominal Feedwater Suction Pressure vs Power, nominal Power 98X 70X 50X 30X Po~er 98X 70X 50X 30X Terperature 345 F
319 F
295 F
259 F
Pressure 277 psig 282 psfg 305 psig 370 psig 2.1 3.0 Head-Capacity and HPSH curves Head. Capacity and HPSH curves for main feedwater pumps Hain Feedwater fxmp - Rated Head Hain Feedwater pump - Rated Torque Hain Feedwater pump - Homent of inertia Elevation of steam generator inlet nozzle Elevation of main feedwater fxmp, ft Elevation of condensate pump, ft HFM regulating valve open time on demand, sec HFM regulating valve close time on demand, sec HFM regulating valve Cv, full stroke Low load HFM regulating valve Cv, (bypass valves)
HFM Heater resistance (delta P)
Auxiliary Fccdwater (AFM)
Hfnfsxan design temperature of the water source service water / CST (degrees F)
Haxfaxia design tecpcraturc of the water source service water / CST (dcgrees F)
See HSRL 21501 289.612 257.75 250.833 10 990 48.7 see HSRL 32(*), 50 80, 100 Selected flow splits are provided for model validation.
Elevation is at center of shaft HFM transients use 20 sec stroke time Assumed value.
Actual value
= 493.6.
Effective Cv: includes bypass line Desfgn data on the High Pressure Heaters (2 in parallel) is provided initial AFM water source are the CSTs located in the Service Bldg. Safety Related source is the Service Mater system (lake).
- Value different for CHHT integrity.
fnitfal AFM water source are the CSTs located in the Service Bldg. Safety Related source is the Service Mater system (lake).
COLR 16 Cycle 25, Draft 8
e..
e Table 1: UPSAR Chapter 15 Analysis Setpoints and Input Parameters Item/Name Value Remarks:
Startup time for the auxiliary feedwater
- pumps, sec Mininxm delay for AFM start, sec Haximan delay for AFW start, sec AFW control valve open time on demand, sec AF'W control valve Cv[flow is f(dP)]
TDAFMP, maxinasa flow, gpm AFW, minimis flows, both generators
- intact, gpm Hinisaia delay for standby AFW start, min TDAFM - 0(
IIDAFM -
1 HOAFM - 47, TDAFM at LO Level both SGs M/A 600 TDAFMP 200/SG MDAFMP 200/SG 10
<<TOAFW starts on LO level (17X) in both gens or UV on both unit 4Kv busses.
HOAFM starts on SI (seq), or LO level either SG, or trip of both HFP or AHSAC HDAFM acceleration time test results show approximately 1.5 s.
For HDAFW, LOOP on sequencer is 47 sec.
TDAFW starts at nominal 17K in both SGs HDAFM control valves are normally open and throttle closed to control flow between 200-230 gpm HDAFMP valves are 3 Rockwell model ¹ A4006JKHY stop check valves.
TDAFM control valves (4297, 4298) are 3 Fisher
¹470-MS'BLOCA asswes 300 gpm per SG with the failure of one DG 4.0 Nein Stems System (HS)
Location (and elevation) of condenser dip valves and atmospheric relief valves Full load steam line pressure drop, psi MS Isolation valve close time [full open to full close] close time, sec MS Isolation valve Cv [flow is f(dP)]
CSD - elev 256'.875 ARV - elev 289'.563 approx 45 HSIV - 5.0 check valve. 1.0 HSIV - 23500 check valve - 17580 This estimate, to the governor valves, is provided for comparison purposes only.
The check valve is assigned to close in 1 sec under reverse flow.
4.1 Nein Stems Code Safety Valves Mumber of valves (4 per line)
Valve flow capacities
- Total, ibm/hr Valve Flow vs SG pressure (psia),
total per bank (4 valves),
ibm/sec.
8 6621000 1110 1115 1120 1125 1131 1136 1141 1151 1161 1166 1173 1181 1190 1200 1205 1209 1211 0
40 91 141 191 222 223 225 227 228 342 494 646 799 859 920 931 Rated flow (3X accunulation per ASME,Section III):
1085 psig
~ ~.........
~.... 797,700 ibm/hr (each) 1140 ps'ig
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
837g600 ibm/hr (each)
COLR 17 Cycle 25, Draft 8
0
/1~1 4
'I I'
Table 1:
UFSAR Chapter%5 Analysis Setpoints and Input Parameters Item g
Item/Name Value Remarks:
Number of valves in bank Valve setpoint(s), (first/last three),
- nominal, psig Valve blowdown characteristic 4
10BS/1'140 15K maxiaam Valves are Crosby PHA-65 6R10 Setpoint tolerance is -1X / +3K.
Nodel valve setpoint at 1.01 (nominal),
and full flow at 1.04 (nominal).
4.2 Atmospheric relief valves No. Atmospheric relief valves Atmospheric relief valve setpoint/Air-operated, ps 1 g Atmospheric relief valve setpoint/Booster, psig Atmospheric relief valve capacity, ibm/hr 2
1050 1060 313550 at 1060 psig During Hot Standby operation setpoint is lowered to control no load Tavg Nax flow is 380000 5.0 5.1 Turbine Generator (TG)
Condenser No. of condenser dump valves Condenser dump valve open time, sec Condenser dump valve close time, sec Condenser dump valve setpoint(s)
Condenser dump valve Cv [flow is f(dP)]
For TT: Tavg>555 4 valves,
>563 4 valves; no TT: Tref +12 4 valves, Tref+20 4 valves Ass+ning close time ~ opening time On TT valves control open at 6.7X/F (PiD) above 547 with full open sctpoints as described.
On 10K step load decrease same ratio with a 6F deadband from Tref Design Cv (240) from design conditions (302,500 ibm/hr sat steam at 695 psig) 6.0 Chemical and Volcan Control System (CVCS)
CVCS capacity/pwp CVCS minimum/pump, gpm Type of controller (e.g.,
P + i) and gains 3 pumps, 60 gpm max each 15 PID 100K,180 sec,10 sec Normal ops:
2 charging pips - one is manual at 15-20 gpm and the other in automatic.
Charging ixips are PDPs w/ 46 gpm total - 8 gpm to seals - 3 gpm leakage
+ 5 gpm into RCS.
40 gpm letdown 6.1 Reactor Nak~ Mater System (RÃM)
RHM capaci ty/pump 2 pwT>s, 60 gpm each 7.0 7.1 7.1.1 Emergency Core Cooling System (ECCS)
ECCS Delivery vs RCS Prcssure Residual Heat Removal (RHR) Delivery vs RCS Pressure COLR 18 Cycle 25, Draft B
0..
~
Table 1: UPSAR Chapter 15 Analysis Setpoints and Input Parameters Item/Name Value Remarks:
Hinimm RHR Delivery, train failure Hininun RHR Delivery, two punps running, one line blocked RCS Pressure (psia) 155 152 150 140 120 100 80 60 40 20 14.7 RCS Pressure (psia) 155 154 152 150 140 120 100 80 60 40 20 14.7 Delivery (gpm) 0 0
0 250 648 836 985 1115 1232 1338 1365 Delivery (gpm) 0 0
160 252 516 830 1056 1243 1406 1552 1686 1720 LOCA Appendix K case.
Train failure results in one pump running with 10K degradation with one line blocked.
LOCA Appendix K case (offsite power available).
Two pcs running with 10K degradation with one line blocked.
7.1.2 Safety Injection (SI) Delivery vs RCS Pressure Hininun SI delivery, 2 pumps operating, one line spilling Hininam Sl delivery, 3 pumps operating, non-LOCA Press (psig) 1375 1300
'1200 1100 1000 900 800 700 600 500 400 300 200 100 0
Press (psia) 1390 1315 1215 1115 1015 915 815 715 615 515 415 315 215 115 15 Spi l I (gpn) 465 465 465 465 465 465 465 465 465 465 465 465 465 465 465 y
(gpm)
Loop 19 97 163 214 257 295 329 360 390 418 444 469 493 516 538 Delivery (gpm)
. 0.0 62 125 167 201 229 253 273 289 305 321 336 352 368 394 Deliver Loop 'A I8 I 16 87 147 193 231 266 297 325 352 377 400 423 445 465 485 LOCA Appendix K case.
Train failure results in two pumps running with 5X degradation with one line spilling to contaiwent.
Used for non-LOCA transients, 5X purp degradation COLR 19 Cycle 25, Draft B
,0 l,
0..
~
Table 1:
UFSAR Chapter 15 Analysis Setpoints and Input Parameters Item 4 Item/Name Value Remarks:
Hinimaa SI delivery, 2 punps operating non-LOCA Press (psia) 1390 1315 1215 1115 1015 915 815 715 615 515 415 315 215 115 15 Delivery (gpm)
Loop 'A'oop IBI 8
8 69 71 121 126 162 169 197 206 228 239 255 269 281 296 305 322 328 346 350 369 370 391 390 412 409 432 427 452 Used for non-LOCA transients, 5X puIp degradation.
Haximia Si delivery, 3 punps operating, SGTR Press (psig) 1375 1300 1200 1100 1000 900 800 700 600 500 400 300 200 100 0
Loop A (gpm) 76 128 180 221 258 290 320 348 374 398 421 443 464 485 504 Loop B (gpm) 84 141 198 245 285 321 354 385 413 440 466 490 514 536 558 The KYPIPE model asswes no punp degradation.
Loop A and B pressures are set equal.
Used for SGTR.
COLR 20 Cycle 25, Draft B
o..
e Table 1:
UFSAR Chapter 15 Analysis Setpoints and Input Parameters Item 4 Item/Name Value Remarks:
7.3 Accumulators Hunber of accumiators Total volune, each, ft Liquid volme, ft
- min/max Liquid volune, fts - Best Estimate Initial pressure, psig - Miniate / Maximun Initial temperature, F
Boron concentration, ppm (min/max) 2 1750 1126/1154 1140 700/790 105 1800/2900 LBLOCA Mote -
EQ analyses use a maximun concentration of 3000 ppm 7 4 RWST RWST Teaperature, mfn / max, degrees F
Minlmm RWST volune, gal RWST boron concentration, ppm (min/max) 60 / 80 300,000 2000/2900 Mote -
EQ analyses use a maxirxin concentration of 3000 ppm 8.0 Contaiaaent Initial contairwent pressure, psia Initial containment temperature (LOCA/SLB) degrees F
Initial relative henidity, X
SW temperature min/max, degrees F
Maxiaun contaiment
- leakage, wtX/day min - 14.5 max - 15.7 90/120 20 35/80 0.2 Minimus is used for LOCA analysis.
Maximia is used for the contaiwent integrity cases (SLB).
LOCA temperature lower for PCT calculations.
SLB higher for contairgent integrity 8.1 Contairioent Heat Sinks Listing of Passive Heat Sinks, quantities, materials, and configurations see HSRL 8.2 Concrete density, conductivity, capacity Densities, Thermal Conductivities and Heat Capacities of Heat Sinks Insulation density, conductivity, capacity 6.67.ibm/ft 0.0208 BTU/hr F ft 2.0 BTU/ft F
141 ibm/ft to 150 ibm/ft 0.73 to 0.81 BTU/hrFft 0.2'I BTU/ibm F note: mininun conductivity corresponds to maxiaxia density, and maxiaaia conductivity corresponds to minimrn density.
Steel density, conductivity, capacity Stainless steel density, conductivity, capacity Contaiwent free volune, min / max, cu. ft.
Ground Temperature (degrees F)
Outside Air Temperature, min / max, degrees F
490 ibm/f t'8 to 30 BTU/hrFft 0.111 BTU/ibm F 496 ibm/ft 15 BTU/hrFft 0.11 BTU/ibm F 1,000,000 / 1,066,000 55
-10 / 100 below grade tenperature COLR 21 Cycle 25, Draft 8
II Il g
g II I
IA'i If M
e e
Table 1: UFSAR Chapter 15 Analysis Setpoints and Input Parameters Item 4
Item/Name Value Remarks:
HTC for outside surfaces Containment fan cooler performance Contaiwent spray flow, min'/ max, each, gpm Teap (deg F) 120 220 240 260 280 286 Hin Hax (X106BTU/hr) 2.05 4.55 35 F 1 99.2 40.8 113.8 46.8 129.3 52.9 145.5 54.7 150.4 1300 / 1800 1.65 BTU/hr ft degrees F
8.3 Delays for CRFCs and Spray Pumps CRFC delay, offsite power available, seconds CRFC delay, offsite power not availabLe, seconds Contairment Spray, 1300 gpm each
- pump, maxiazan
- delay, sec Contairment
- Spray, 1800 gpm each pump, mininun
- delay, sec Contairment Design pressure, psig Distance Basement floor to Springline, feet Distance Springline to top of dome, feet 34 44 28.5 - one punp 26.8 - two imps 9 / (14 w LOOP) 60 95 52.5 includes 2.0 sec SL delay includes 2.0 sec SL delay This delay is from the time Contaiment Hi-Hi setpoint is reached.
Lt includes instrunent delay and spray line fill time.
This delay is from the time of break.
8.4 Contaireant Smp Nininun wtX of HaOH Tank 30 9.0 Control Systems (Reactor, FM, Przr Level, Turbine, AFM)
Tavg versus po~er Pressurizer pressure and Level algorithms SG secondary level algorithm H/A H/A H/A Tavg ramps linearly from 547 degrees F at OX power to 573.5 degrees F at 100'ower Pressurizer pressure setpoint is constant at 2235 psig
. Pressurizer level ramps from 19.5 X to 49K for 0 to 100 X power (547 - 573.5 degrees F).
Level ramps from 39K at OX power to 52K at 20K power and remains constant at 52K to 100K power.
(Power from turbine 1st stage press.)
10.0 10.1 10.1.1 Safety System Setpoints Reactor Protection System Power range high neutron flux, high setting nominal accident analysis delay time, sec 1.09 1 ~ 18 0.5 10.1.2 Power range high neutron flux, Low setting nominal 0.250 COLR 22 Cycle 25, Draft B
~ 'r'
~
~
e l
V
o e
Table 1:
UFSAR Chapter 15 Analysis Setpoints and Input Parameters Item 4 Item/Name Value Remarks:
accident analysis delay time, sec 0.350 0.5 10.1.3 Ovcrtcmperaturc delta T nominal accident analysis delay time, sec 10.1.4 Overpower delta T nominal accident analysis delay time, sec Variable Variable 6.0 Variable Variable 2.0 Total delay time - from the time the temperature difference in the coolant loops exceeds the trip setpoint until the rods are free to fall Not explicitly modelled in safety analysis 10.1.5 High pressurizer pressure nominal, psig accident analysis, psia delay time, sec 2377 2410 2'
10.1.6 Low pressurizer prcssure nomfnal, psig accident analysis, psia delay tine, sec 1873 1760 (non-LOCA),
1715 (LOCA) 1890 (SGTR) 2.0 10.1.7 Low reactor coolant flow nominal accident analysis delay time, sec 91K of normal indicated flow 87X per loop 1.0 10.1.8 Low-low SG level nominal accident analysis delay time, sec 17K of the narrow range level span OX of narrow range level span 2.0 While trip setpoint could be as low as
- 16K, AFW initiation limits to 17K 10.1.9 Turbine Trip (low fluid oil prcssure) nominal, psfg accident analysis delay time, sec 45 N/A 2.0 Not explicitly modeled in safety analysis COLR 23 Cycle 25, Draft 8
I Table 1: UFSAR Chapter 15 Analysis Setpoints and Input Parameters Item g Item/Name Value Remarks:
10.1.10 Undervoltage
- nominal, V
accident analysis delay time. sec 3150 N/A N/A Not explicitly modeled in safety analysis 10.1.11 Under frequency
- nominal, Hz accident analysis delay time 57.7 N/A N/A Not explicitly modeled in safety analysis 10.1.12 Intermediate range nominal safety analysis delay time, sec 0.31 N/A N/A Not explicity modeled in safety analysis 10.1.13 Source Range nominal, cps accident analysis, cps delay time, sec 1 'E+5 1.0E+
2.0 10.1.14 High Pressurizer level nominal accident analysis delay time, sec 0.90 0.938 2 '
10.2 10.2.1 10.2.1.1 Engineered Safety Features Actuation System Safety Injection System High contaimant pressure Nominal setpoint, psig Accident Analysis setpoint, psig Delay time, sec 4.0 6.0
- 34 44 w/ LOOP
~only modeled in accident analysis for start of containment fan coolers.
Time delays are for start of contairment fan coolers.
10.2.1.2 Lou pressurizer pressure Nominal setpoint, psig Accident Analysis setpoint, psig Oelay time, sec 1723 1715 2.0 10.2.1.3 Lou steam line pressure Nominal setpoint, psig Accident Analysis setpoint, psig 514 372.7 See NSRL COLR Cycl e 25, Draft B
0 II e
Q Table 1:
UFSAR Chapter 15 Analysis Setpoints and Input Parameters Item 4 Item/Name Value Remarks:
Delay time, sec 2.0 See HSRL 10.2.2 Contairnnent Spray Hominal Setpoint, psig Accident analysis setpoint, psig Delay time, sec 28 32.5 28.5 See NSRL Delay time includes time to fill lines.
See HSRL 10.2.3 AFM System Low-low stean generator water level Nominal Setpoint Accident analysis setpoint Delay time, sec 17 X of narrow range instrunent span each steam generator 0 X of narrow range instrunent span each steam generator 2.0 A positive 11K error has been included to account for the SG level measurement system at a contairment temperature of 286 F
10.2.4 Stean Line Isolation 10.2.4.1 High contaiInnent pressure Hominal Sctpoint, psig Accident analysis setpoint Delay tine 18 H/A H/A Hot explicitly modeled Not explicitly modeled 10.2 4.2 High stean flow, coincident with low Tavg and SI Nominal Setpoint Accident analysis setpoint Delay time 0.49E6 lb/hr equivalent steam flow at 755 psig and Tavg < 545 F
H/A H/A Hoten flow setpoint is below nominal full power flow and therefore this portion of. logic is made up at power Not explicitly modeled Hot explicitly modeled.
Steam line isolation is assumed concurrent with SI (i.e.
2 s delay + 5 s valve stroke) 10.2 4.3 High-high stean flow, coincident SI Nominal Setpoint Accident analysis setpoint Delay time 3.6E6 lb/hr equivalent steam flow at 755 psig H/A N/A Hot explicitly modeled Hot explicitly modeled.
Steam line isolation is assumed concurrent with SI (i.e.
2 s delay + 5 s valve stroke) 10.2.5 Feedwater isolation 10.2.5.1 High stains gencl atol'atcl'evel Nominal Setpoint 67X of the narrow range instrInnent span each SG COLR 25 Cycle 25, Draft 8
~ ~I l
H y d
g wk II
%g e
k g
II I
< ~
w 4
I U
J
I
~
Table 1:
UFSAR Chapter 15 Analysis Setpoints and Input Parameters Item 4 Item/Name Value Remarks:
Accident analysis setpoint Delay time 100X of the narrow range instrurent span each SG 2.0 Instrwent loop only 11.0 Original Stoma Generators OSG secondary outlet pressure at OX full power, psig Steam tenperature at OX full power, F
OSG collapsed Liquid level at OX full power, X
NRS OSG total liquid mass per SG at OX full power, Lbm OSG secondary outlet pressure at 100X full power, psig Steam temperature at 100X full power, F
OSG collapsed Liquid level above tube sheet at 100X full power, ft OSG total Liquid mass per SG at 100X full power, ibm Neat load per SG, Btu/hr Primary flow per SG, Lb/hr - Design Steam flow per SG, Lb/hr - Design Secondary design pressure, psig Secondary design teaperature, F
No. of tubes per SG Tube OD, in.
Tube average wall thickness, in.
Haxiaxin moisture carryover, X
Secondary heat transfer area, ft per SG Primary heat transfer area, ft per SG Tube length(s)
Haximia, ft Hiniaun, ft Average effective length, ft Overall OSG bundle height, ft Narrow range level tap Locations (elevations),
ft Wide range level tap locations (elevations), ft Secondary nozzle to nozzle dP 9 full power, psi Primary nozzle to nozzle dP with no plugged tubes Secondary volunc, fta (water volune 9 1525/0 HWt) - nominal 1005 547 39 130,120 815 522.5 38.75 85,410 2602000000 33600000 3290000 1085 556 3260 0.875 0.05 0.25 44,430 39,406 71.365 57.146 59.5 elevation - 286.549 ft or 33.031 ft above bottom of tube sheet 287.474/299.401 256.349/299.401 16.5 32.3 psi g flow "- 33.64E6 Lb/hr 1681/2821 ASD setpoint to control Tavg at 547 F ~
Assigned
= to Tavg Nominal value, analysis rmst justify assuned error band CIRC run using new SG conditions with water level at 55.5X NRL (38.75'bove tubesheet)
This value is for OX plugging and a fouling factor of.00002.
This value is for new SG conditions This is a maxiaxin value, used to generate this mass value below.
A miniaun value would be 35.5 XNRL.
Value considered steam generator new conditions, with water level at 55'X narrow range (38.75~
above tubesheet)
Hiniaam wall thickness not specif ied",
Includes tube sheet (2*22)
Includes tube sheet (2*22)
Above tube sheet Tube sheet thickness is 22 inches Estimate value COLR 26 Cycle 25, Draft B
~ arh
0 Table 1:
UFSAR Chapter 15 Analysis Setpoints and Input Parameters Item 4 Item/Name Value Remarks:
Secondary volune, fts (steam volune 9 1525/0 Hilt) - nominal Primary total volune per SG, ft Hot leg head volune per SG, ft'old leg head volwe per SG, ft Tube primary volune per SG, ft Downcomer level versus downcomer volune profile Circulation ratio (100K power)
Total volune versus level SG Primary Head Cladding Thickness Haximm SG tube leakage, gpm 2898/1758 942.3 133.4 133.4 675.5 See HS&L 4.4 See HS&L 5/16 0.5/SG An ATHOS model of the OSG's was used.
The Circulation ratio is the downcomer flow divided by the outlet flow. Confirmed with ClRC run.
Actual value limited to 0.1 gpm/SG due to stress concerns COLR 27 Cycle 25, Draft B
Attachment III Revised Administrative Controls Program
eporting Requirements
~
~
~5.6 5.6 Reporting Requirements (continued) 5.6.3 Radioactive Effluent Release Re ort 5.6.4 The Radioactive Effluent Release Report covering the operation of the plant shall be submitted in accordance with 10 CFR 50.36a.
The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the plant.
The material provided shall be consistent with the objectives outlined in the ODCM and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.I.
Monthl 0 eratin Re orts 5.6.5 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the pressurizer power operated relief valves or pressurizer safety valves, shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.
CORE OPERATING LIMITS REPORT COLR
'a ~
Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload
- cycle, and shall be documented in the COLR for the following:
LCO 3.1.1, LCO 3.1.3, LCO 3.1.5>
LCO 3.1.6, LCO 3.2.1, LCO 3.2.2, LCO 3.2.3>
LCO 3.4.1, LCO 3.9.1, "SHUTDOWN MARGIN (SDM)";
"MODERATOR TEMPERATURE COEFFICIENT (MTC)";
"Shutdown Bank Insertion Limit";
"Control Bank Insertion Limits";
"Heat Flux Hot Channel Factor (Fo(Z))";
"Nuclear Enthalpy Rise Hot Channel Factor (F"~)";
"AXIAL FLUX DIFFERENCE (AFD) "
'RCS
- Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits"; and "Boron Concentration."
(continued)
R.E. Ginna Nuclear Power Plant 5.0-19 Draft B
flak 1 j k
eporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COLR (continued) b.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
1.
WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Hethodology," July 1985.
(Methodology for LCO 3. 1. 1, LCO 3. 1.3, LCO 3. 1.5, LCO 3. 1.6, LCO 3.2. 1, LCO 3.2.2, LCO 3.2.3, and LCO 3.9.1.)
2.
WCAP-9220-P-A, "Westinghouse ECCS Evaluation Model-1981 Version," Revision 1, February 1982.
(Methodology for LCO 3.2. 1.)
3.
WCAP-8385, "Power Distribution Control and Load Following Procedures
- Topical Report," September 1974.
(Methodology for LCO 3.2.3.)
4.
WCAP-8567-P-A, "Improved Thermal Design Procedure,"
February 1989.
(Methodology for LCO 3.4. 1 when using ITDP.)
5.
WCAP 11397-P-A, "Revised Thermal Design Procedure,"
April 1989.
(Methodology for LCO 3.4. 1 when using RTDP.)
6.
WCAP-10054-P-A and WCAP-10081, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code,"
August 1985.
(Methodology for LCO 3.2. 1) 7.
WCAP-.10924-P-A, Volume 1, Rev.
1, and Addenda 1,2,3, "Westinghouse Large-Break LOCA Best-Estimate Methodology, Volume 1:
Model Description and Validation," December 1988.
(Methodology for LCO 3.2. 1) 8.
WCAP-10924-P-A, Volume 2, Rev.
2, and Addenda, "Westinghouse Large-Break LOCA Best-Estimate Methodology, Volume 2:
Application to Two-Loop PWRs Equipped with Upper Plenum Injection," December 1988.
(Methodology for LCO 3.2. 1)
(continued)
R.E.
Ginna Nuclear Power Plant 5.0-20 Draft B
'f
)f I lr'
~
~
eporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 5.6.6 COLR (continued) 9.
WCAP-10924-P-A, Rev.
2 and WCAP-12071, "Westinghouse Large-Break LOCA Best Estimate Methodology, Volume 2:
Application to Two-Loop PWRs Equipped With Upper Plenum Injection, Addendum 1:
Responses to NRC guestions,"
December 1988.
(Methodology for LCO 3.2. 1) 10.
WCAP-10924-P, Volume 1, Rev.
1, Addendum 4, "Westinghouse LBLOCA Best Estimate Methodology; Model Description and Validation; Model Revisions," August 1990.
(Methodology for LCO 3.2. 1) c.
The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDH, transient
'analysis limits, and accident analysis limits) of the safety analysis are met.
d.
The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
Reactor Coolant S stem RCS PRESSURE AND TEMPERATURE LIMITS REPORT PTLR a ~
b.
RCS pressure and temperature limits for heatup,
- cooldown, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:
LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits" The power operated relief valv'e lift settings required to support the Low Temperature Overpressure Protection (LTOP)
- System, and the LTOP enable temperature shall be established and documented in the PTLR for the following:
MODE 4";
MODE 5, Loops Filled";
LCO 3.4. 10, "Pressurizer Safety Valves";
and LCO 3.4. 12, "LTOP System."
(continued)
R.E.
Ginna Nuclear Power Plant 5.0-21 Draft B
4
~
~r