ML20199L909

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Rev 0 to Technical Data Book (Tdb) TDB-IX, FCS RCS Pressure-Temp Limits Rept (Ptlr)
ML20199L909
Person / Time
Site: Fort Calhoun Omaha Public Power District icon.png
Issue date: 12/31/1997
From:
OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML20199L879 List:
References
TDB-IX, TDB-IX-R, TDB-IX-R00, NUDOCS 9802100066
Download: ML20199L909 (19)
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Text

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NT 4

Fort Calhoun Station-Unit No.1 TDBIX TECHNICAL DATA BOOK

Title:

RCS PRESSURE TEMPERATURE LIMITS REPORT (PTLR)

FC-68 Number:

Reason for Change:

Contact Person:

4 ISSUED: -97 9:30 am RO

FORT CALHOUN STATION TDB IX TECHNICAL DATA BOOK Pcge 1 of 18

. Table of Contents

1. I NT R O D U CTI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
2. GL 96-03 (Reference 3.1) PROVISION REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . . 2 2.1 Neutron Fluenco Values . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 2.2 Reactor Vessel Surveillance Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.3 LTO P System Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.3.1 R3 actor Coolant System (T.S. 2.1) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 2,3.2 Chemical and Volume Control System (T.S. 2.2) . . . . . . . . . . . . . . . . . . . 5 ,

2.3.3 Emergency Core Cooling System (T.S. 2.3) . . . . . . . . . . . . . . . . . . . . . . . 6 2.4 Beltline Material Adjusted Reference Temperature (ART) . . . . . . . . . . . . . . . . . . . 7 2.5 Pressure-Temperature Limits using limiting ART in the P-T curve calculation . . . 8 2.5.1 Heatup and Cooldown Rate (T.S. 2.1.2) . . . . . . . . . . . . . . . . . . . . . . . . . . 8 2.6 Minimum Temperature Requirements in the P-T curves . . . . . . . . . . . . . . . . . . . 13 2.7 Surveillance Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 2.7.1 Reactor Coolant System and other components subject to ASME Section XI Boller and Pressure Vessel Code Inspection and Testing Surveillance (T. S. 3. 3 ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14

3. R E F E R E N CES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 4.-- O P E RATI N G F I G U RES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 4.1 RCS Pressure - Temperature Limits for Heatup . . . . . . . . . . . . . . . . . . . . . . . . . 16 4.2 RCS Pressure - Temperature Limits for Cooldown . . . . . . . . . . . . . . . , . . . . . . . 17 4.3 Predicted Radiation Induced NDTT Shift . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 R0

FORT CALHOUN STATION TDBIX TECHNICAL DATA BOOK Pcge 2 of 18

1. INTRODUCTION l

This PTLR for Fort Calhoun Station (FCS) Unit No.1 contains Pressure Temperature  !

(P-T) limits corresponding to 20 Effective Full Power Years (EFPY) of operation. In addition, this report contains Low Temperature Overpressure Protection (LTOP) specific .

requirements which have been developed to protect the F T limits from being exceeded during the limiting LTOP event.

The Technical Specifications affected by this report are listed below and are separated into the appropriah category; P T limits or LTOP requirements.

i' l

2. C L 96-03 (Reference 3.1) PROVISION REOulREMENTS 2.1 Neutron Fluence Values  !

The reactor vessel beltline neutron fluence has been calculated for the critical

- locations using the NRC accepted methodologies as described in Reference 3.2, Appendix B.

Since the time of the performance of the analyses supporting the 20 EFPY limits in References 3.3,3.14 and 3.15 several significant operating and analysis changes . l have occurred at FCS. The Reference 3.3 (based upon References 3.14 and 3.15) 20 EFPY limit analysis was performed in 19g0 at a time when low radial leakage fuel managernent was being used as well as the ENDF/B IV nuclear cross section  !

library. The fast neutron fluence (E> 1 MeV) at 20 EFPY was projected to be ,

1.501 x 10 n/cm', to the limiting reactor vessel material (i.e., the 3-410 welds at i 60' and 300' and up to approximately 40% of core height) It should be noted that ,

although there is a 3 410 weld at 180' the fluence at this location is significantly less than at the 60' and 300' locations and it is therefore not limiting.- In Cycle 14 ,

extreme low radial leakage fuel management was implemented to furt.h sr reduce the reactor vessel fast neutron flux and insure operation to August g, 2013 without exceeding the 10CFR50.61 PTS screening criteria. Use of the ENDF/B-VI cross section library in a 1994 updated fluence analysis (Reference 3.6) in .

conjunction with methods described in Reference 3.4 (and its predece'isor Reference 3.5) resulted in a revised projection at 20 EFPY of 1.150 x 10'? n/cm'.  !

Since the original analysis remains conservative with respect to the actual projected >

fast neutron fluence the Reference 3.3 results are utilized herein. It should be .

^

noted that further conservatism exists in the reactor vessel weld chemistry factor of 234.5'F used in References 3.3, 3.14 and 3.15 versus the most recent!y derived value of 231.06'F (per Reference 3.7). '

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FORT CALHOUN STATION TDBIX TECHNICAL DATA BOOK Pcge 3 of 18 2.1 In References 3.3,3.14 and 3.15 the peak value(s) of fast neutron fluence (E > 1 MeV) at the vessel clad interface was used as input to the Adjusted Reference Temperature (ART) calculations for FCS. The fluence corresponding'to the limiting 3-410 welds located at 00* and 300' for 20 effective full power years (EFPY) is 1.501 x 10" neutrons per square centimeter (n/cm 2) with an associated 9 a-Sty of less than i 20%. In the updated Reference 3.6 analysis the uncensinty was shown to be less than i 13%.

2.2 Reactor Vessel Surveillance Program The reactor vessel surveillance program was developed in accordance with ASTM E-185-66. The surveillance capsule withdrawal details.and schedule are described in Appendix B, Reference 3.2 and Reference 3.8. The reports describing the post irradiation evaluation of the survelliance capsules are contained in References 3.9,3.10 and 3.11. Each removed capsule has been evaluated in accordance with the testing requirements of the version of ASTM E-185 in effect at the time of capsule removal. Reference 3.12 describes the pre irrt 11ation materials baseline surveillance program.

2.3 LTOP System Llmits The LTOP requirements have been developed bv making a comparison between the peak transient pressures and the appropriate Aro',ndix G pressure-temperature -

limit curves. The acceptabsty criterion regarding each particular transient is that the peak transient pressure does not exceed the applicable Appendix G pressure limit. These requirements for LTOP have been established based on NRC accepted methodologies and are described in Appendix B, Reference 3.2.

The affected Technical Specification Limiting Conditions for Operation (LCO's) which ensure adequate LTOP are:

T.S. 2.1.1 Reactor Coolant System - Operable Components T.S. 2.1.2 Reactor Coolant System - Heatup and Cooldown Rate T.S. 2.1.6 Reactor Coolant System - Pressurizer and Main Steam Safety Valves T.S 2.2.1 Chemical and Volume Control System - Boric Acid Flow Paths - Shutdown T.S. 2.2.2 Chemical and Volume Control System - Boric Acid Flow Paths - Operating R0

FORT CALHOUN STATION TDB Ir TECHNICAL DATA BOOK Pcge 4 of 18 2.3 T.S. 2.2.3 Chemical and Volume Control System - Charging Pumps - Shutdown T.S. 2.2.4 Chemical and Volume Control System - Charging Pumps - Operating LS. 2.2.5 Chemical an'd Volume Control System Boric Acid Transfer Pumps - Shutdown T.S. 2.2.6 Chemical and Volume Control System - Boric Acid Transfer Pumps - Operating T.S. 2.2.7 Chemical and Volume Control System Borated Water -

Source - Shutdown T.S. 2.2.8 Chemical and Volume Control System Borsted Water Sources - Operating T.S. 2.3(3) Emergency Core Cooling System Protection Against Low

. Temperature Overpressurization The LTOP specific requirements for each T.S. LCO, when applicable, are presented in the following subsections. .

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I FORT CALHOUN STATION TDBIX i TECHNICAL DATA BOOK Page 5 of 18 2.3.1 Reactor Coolant System (T.S. 2.1)

A. Operable Components (T.S. 2.1.1)

The following requirements apply:

1) Reactor coolant system leak and hydrostatic tests shall be conducted within the limitations of PTLR Figure 4.2.
2) -l' no reactor coolant pumps are operating a reactor pump shall not be started while T, is below 385' F (LTOP enable >

temperature) unless at least one of the following is met:

(i) A pressurizer steam space of 53% by volume (50.6% or less actuallevel) exists, or (ii) The steam generator secondary side temperature is less than 30'F above that of the reactor coolant system cold leg. . Startup of the first RCP with " cold steam generators", under these conditions may result in a cooldown which exceeds the 10'F/ hour cooldown limit of Figure 4.1, however per the analysis of Reference 3.13 this does not result in exceeding the ASME Section lll (and XI) or -

10CFR50 Appendix G criteria. Thus, startup of the first RCP, which is considered to be a normal plant evolution, with a resultant step cooldown of up to 50'F is acceptable, provided the pressurizer pressure remains less than 275 psia. Reference 3.13 (page BS) shows the allowable RCS pressure to be 369 psia which is reduced to 275 psia to include allowances for instrument uncertainty, elevation head effects, reactor vessel to surge line -

AP, and analysis margin.

B. Heatup and Cooldown Rate (T.S. 2.1.2)

The RCS (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on PTLR Figures 4.1 and 4.2 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing.

2.3.2 Chemical and Volume Control System (T.S,2.2)

A. Boric Acid Flow Paths - Shutdown (T.S. 2.2.1)

The flow path from the SIRWT to the RCS via a single HPSI pump shall only be established if:

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FORT CALHOUN STATION TDBIX TECHNICAL DATA BOOK Page 6 of 18 2.3.2A 1) The RCS pressure boundary does not exist (i.e., mode 5 with the reactor vessel head removed or a 0.94 square inch or larger vent area per T.S. 2.1.6(4)), or

2) No charging pumps are operable with flow paths defined in T.S.

2.2.1a through 2.2.ic and the RCS heatup and cooldown rates shail be limited to those in PTLR Figure 4.2.

B. Charging Pumps - Shutdown (T.S. 2.2.3)

The flow path from the SIRWT to the RCS via a single HPSI pump shall be established only if:

1) The RCS pressure boundary does not exist (i.e., mode 5 with the reactor vessel head removed or a 0.94 square inch or larger vent area per T.S. 2.1.6(4)), or
2) No charging pumps are operable and the RCS heatup and cooldown rates shall be limited to those in PTLR Figure 4.2.

2.3.3 Emergency Core Cooling System (T.S. 2.3)

A. Protection Against Low Temperature Overpressurization (T.S. 2.3(3))

The following limiting conditions shall be applied during scheduled heatups and cooldowns. Disabling of the HPSI pumps need not be required if the RCS is vented through at least a 0.94 square inch or larger vent, consistent with T.S. 2.1.6(4).

1) Whenever the reactor coolant system cold leg temperature is below 385'F, at least one (1) HPSI pump shall be disabled.
2) Whenever the reactor coolant system cold leg temperature is below 320*F, at least two (2) HPSI pumps shall be disabled.
3) Whenever the reactor coolant system cold leg temperature is below 270*F, all three (3) HPSI pumps shall be disabled.

In the event that no charging pumps are operable when the reactor coolant system cold leg temperature is below 270'F, a single HPSI pump may be made operable and utilized for boric acid injeciton to the core, with flow rate restricted to no greater than 120 gpm.

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FORT CALHOUN STATION TOBIX l' TECHNICAL DATA BOOK Page 7 of 18 2.3.3A The normal procedure for stailing the reactor is to first heat the reactor coolant to near operating temperature by running the reactor coolant pumps. The reactor is then made critical by withdrawing CEA's and diluting the boron concentration of the reactor coolant.

With this mode of start up, the energy stored in the reactor coolant during the approach to criticality is essentially equal to that during power operation and therefore all engineered safety features and auxiliary cooling systems are required to be fully operable.

B. The LTOP enable temperature is 385'F. Below this temperature the system is enabled while above it the high pressurizer pressure trip serves to actuate the PORVs.

2.4 Beltline Material Adjusted Reference Temperature (ART)

The calculation of the adjusted reference temperature (ART) for the reactor vessel beltline region has been performed using the NRC accepted methodologies as described in Appendix B, Reference 3.2. Provision 7 (Section 2.7, Application of Surveillance Data) was conservatively not used to refine the chemistry factor and the margin term. As noted in PTLR Section 2.3 a chemistry factor of 234.5'F and a fast neutron fluence of 1.50 x 10" n/cm2 was us*1 for the limiting 3-410 welds. Use of Reference 3.7 has resulted in the determination of a revised best estimate value of 231.06'F, although not credited in these analyses.

The limiting ART values in the beltline region for FCS corresponding to the 20 Effective Full Power Years (EFPY) for the 1/4t and 3/4t locations (based on References 3.3,3.14, and 3.15) are:

Location ABI Material 1/4t 242.23' F Welds 3-410 at 60' and 300' 3/4t 184.97* F Welds 3-410 at 60' and 300' The projected RTm value for FCS which is currently calculated in accordance with 10 CFR 50.91 is 267' F (Reference 3,7 value refined from end of cycle value to license expiration on August 9,2013) which corresponds to the 3-410 welds at 60' and 300* (i.e.180' not limiting as previously noted in Section 2.1) for a tandem arc weld of weld wire heat numbers 27204/12008. Provision 7 (Section 2.7, Application of Surveillance Data) was not used to refine the chemistry factor and the margin term. ,

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FORT CALHOUN STATION TDBolX TECHNICAL DATA BOOK Pogo 8 of 18 2.5 Pressure Temperature Limits using limiting ART in the P T curve calculation 2.5.1 Heatup and Cooldown Rate (T.S. 2.1.2)

The limits for T.S. 2.1.2 are presented in the subsection that follows. The analytical methods used to develop the RCS pressure temperature limits for heatup and cooldown are based on NRC secepted methodoh ,ies and discussed in Reference 3.2, Appendix A. The methodology is discussed below.

- Before the radiation exposure of the reactor vessel exceeds the exposure for which they apply, PTLR Figures 4.1 and 4.2 shall be updated in .

accordance with the following criteria and procedures: -

A. The curve in PTLR Figure 4.3 shall be used to predict the increase in transition temperature based on integrated fast neutron flux. -If measurements on the Irradiation specimens indicate c significant deviation from this curve, a new curve shall be constructed, if the embrittlement correlation (l.a. R.G.1.99 or ASTM E900) changes, the figure shall be reviewed for assessment of the need of revision and revised if necessary.

B. The limit line on the figures shall be updated for a new integrated .

power period as follows: The total integrated reactor thermal power from startup to the end of the new period shall be converted to an equivalent integrated fast neutron exposure (E>1 MeV). The

- predicted transition temperature shift to the end of the new period shall then be obtained from PTLR Figure 4.3.

C. The limit lines in PTLR Figures 4.1 and 4.2 shall be moved parallel to .

the temperature axis (horizontal) in the direction of increasing temperature a distance equivalent to the transition temperature shift during the period since the curves were last constructed. The boltup temperature limit line shall remain at 82*F as it is set by tne NOTT of the reactor vessel flange and not subject to fast neutron flux. Tha lowest service temperature shall remain at 182'F because -

components related to this temperature are also not subject to fast

  • neutron flux.

D. The PTLR requirements associated with Technical Specification 2.3(3) shall be reviewed and revised as necessary each time the PTLR curves of Figures 4.1 and 4.2 are revised.

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FORT cal.HOUN STATION TDBIX TECHNICAL DATA BOOK Pcga 9 of 18 2.5.1 The maximum allowable reactor coolant system pressure at any temperature is based upon the stress limitations for brittle fracture consideration. These limitations are derived by using the rules contained in Section lll(and XI) of the ASME Code including Appendix G, Protection Against Nonductile Failure, and the rules contained in 10 CFR 50, Appendix G, Fracture Toughness Requirements, The ASME Code assumes that a crack 1-25/32 inches deep and 10-11/16 inches long exists (i e.1/4t location crack with an aspect ratio of 1:6) on the inner surface of the vessel. Furthermore, operating limits on pressure and temperature assure that the crack does not grow during heatups and cooldowns.

The reactor vessel beltline material consists of six plates. The nil-ductility transition temperature (Tuo7) of each plate was established by drop weight tests. Charpy tests were then performed to determine at what temperature the plates exhibited 50 ft lbs. absorbed energy and 35 mits lateral expansion for the longitudinal direction. NRC technical position MTEB 5-2 was used to establish a reference temperature for the transverse directicn (RTwor) of -12'F.

. Similar testing was not performed on all remaining material in the reacter coolant system. However, sufficient impact testing was performed to meet appropriate design code requirements and a conservative RTuo, of 50'F has been established, e

' The initial RT Nor value for the Fort Calhoun submerged arc vessel weldments was determined to be -56'F consistent with 10 CFR 50.61(2)(1) with a standard deviation of 17'F The adjusted reference temperature (RTnot) was determined through the use of Regulatory Guide 1.99, Rev 02 and 10CFR50.61 methods by summing the -56'F initial RTnor (10CFR50.61(2)(i)), the margin term of 66*F (10CFR50.61(2)(ii)) and the ARTuor (100FR50.61(2)(ill)).

As a result of fast neutron irradiation in the region of the core, there will be an increase in the Tuor with operation. The techniques used to predict the integrated fast neutron (E>1 MeV) fluxes of the reactor vessel are described in Section 3.4.6 of the USAR.

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FORT CALHCUN STATION TDBIX TECHNICAL DATA GOOK Page 10 of 18 2.5.1 Since the neutron spectra and the flux measured at the samples and reactor vessel inside radius should be nearly identical, the measured trantition shift for a sample can be applied to the adjacent section of the reactor vessel for later stages in plant life equivalent to the difference in calculated flu:: magnitude. The maximum exposure of the reactor vessel will be obtained from the measured sample exposure by application of the calibrated stimuthat neutron flux variation. To compensate for any increase in the Tm, caused by irradiation, limits on the pressure tompernture relationship are periodically changed to stay within tM stress timus during haatup and cooldown. Analysis of the second removen irradinted reactor vessel sonraillance specimen (Reference 3.10) combinad y(ith wela enemical composition dat arid reduced fluence core Icading designs it itiated in Cycle 8, indicated tnat the fluence at the end of 20.0 Effective Full Power Years (EFPY) at 1500 MWt will be 1.50 x 10' nicm8 on the inside surface of the reactor vessel. Operation through fuel Cycle 19 will result in less than 20.0 EFPY.

j The limit lines in PTLR Figures 4.1 and 4.2 are base' on the following:

a. Hea:vp and Cooldown Curves - From Section 111 of the ASME Code, Appendix G-2215.

Kn = 2 Ku + N Kn = Allowance stress intensity factor at temperature related to RTer (ASME Ill Figure G-2110.1).

K= Stress intensity factor for membrane st:ess (pressure). The 2 represents a safety factor of 2 on rxessure.

Kn = Stress intensity factor radial thermal gradient.

The above equation is applied to the reactor vessel beltline. For plant' heatup the reference stress intensity is calculated for both the 1/41 and 3/4t locations. Composite curves are then generated for each heatup rate by combining the most restrictive pressure-temperature limits over the complete temperature interval.

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FORT CALHOUN STATION TDB IX TECHNICAL DATA BOOK Pcge 11 of 18 2.5.1 in accordance with the ASME Code Section lll Appendix G ,

requirements, the general equations for determining the allowable pressure for any assumsd rate of temperature change during Service Level A and B operation are:

2Ku + Kn < Kn where, -

Ku = Allowable pressure stress intensity factor, Ksi/in Kn = Thermal stress intensity factor, Ksl/in Kn = Reference stress intensity, Ksl/in The pressure temperature limits provided in this report account for the ,

temperature differential between the reactor vessel base metal and the reactor coolant bulk fluid temperature Correction for elevation and RCS flow induced pressure differences between the reactor vessel beltline and pressurizer, are included in the development of the '

pressure-temperature limits. Consequently, the P-T limits are provided on coordinates of pressurizer pressure versus indicated RCS temperature, b i 4

4 0 t

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FORT CALHOUN STATION TDBIX TECHNICAL DATA BOOK Page 12 of 18 2.5.1 The pressure correction factors are based upon the differential pressure due to the elevation difference between the reactor vessel wall adjacent to the bottom of the active core region, and the pressurizer pressure instrument nozzle. This term of the pressure correction factor is equal to 27.6 psl. The pressure correction factors are also based upon flow induced pressure drops across the reactor core through the hot leg pipe up to the surge line nozzle. This term of the pressure correction factor has two values which are dependent

upon the Reactor Coolant Pump (RCP) combination utilized during operellon. At temperatures of T, < 210*F, the flow induced pressure j

drop le based upon the RCS flow rates resulting from two operating -

RCPs and is equal to 27.82 psla. At the temperatures of T, a 210*F, the induced pressure drop is based upon the RCS flow rates resulting l

from three operating RCPs and is equal to 33.79 psl. Consequently, two pressure correction factors are utilized in correcting the reactor vessel beltline region pressure to pressurizer pressure depending upon the cold leg temperature. The following pressure correction factors have been utilized:

T. ('F) PRESSURE CORRECTION FACTOR (PSI) 2210'F 61.5 psl

<210*F 55.5 psi Gy explicitly accounting for the temperature differential between the flaw tip base metal temperature and the reactor coolant bulk fluid temperature, and the pressure differential between thc beltline reglon of the reactor vessel and the pressurizer pressure measurement nozzle, the P-T limits are correctly represented on coordinates of pressurizer pressure and cold leg temperature. A temperature correction factor of 16'F has been utilized in the calculation of the P-T limits to account for temperature measurement uncertainties.

Pressure instrument loop uncertainties have not been included in the Pressure-Temperature limits since these uncertainties have been included in the setpoints of the Low Temperature Overpressure Protection (LTOP) system which has an enable temperature of 385'F.

Including the pressure insturment loop uncertainties in the P-T limits therefore would have resulted in a redundant summation of these uncertainties which would be overly conservative.

b. Inservice Hydrostatic Test - The inservice hydrostatic test curve is developed in the same manner as in PTLR Section 2.5.1.a above with the exception that a safety factor of 1.5 is allowed by ASME Ill in lieu of 2.

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, ..o FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK Page 13 of 18 2.5.1 c. Lowest Service Temperature = 50'F + 120'F + 12'F = 182'F. As indicated previously, an RT,er for all material with the exception of the reactor vessel beltline was established at 50'F.10 CFR Part 50, Appendix G, IV.a.2 requires a lowest service temperature of RT,o7+120'F for piping, pumps and valves. Additionally 12'F is added to account for insturment error. Below this temperature a pressure of 20 percent of the system hydrostatic test pressure cannot be exceeded. Taking into account pressure correction factors for elevation and flow, this pressure is (.20)(3125) - 56 = 569 psia, where 56 psi is the hydrostatic head correction factor,

d. Boltup Temperature = 10'F + 60'F + 12*F = 82'F. At pressures below 569 psia, a minimum vessel temperature must be maintained to comply with the manufacturer's specifications for tensioning the vessel head.

This temperature is based on previous NDTT methods. This temperature corresponds to the measured 10'F NDTr of the reactor vessel flange, which is not subject to radiation damage, plus 60*F data scatter in NDTT measurements plus 12'F instruritent error,

e. The temperature at which the heatup and cooldown rates change in PTLR Figures 4.1 and 4.2 reflects the setpoint at which the limiting heatup and cooldown rates with respect to the inlet temperature (T,)

change.

2.6 Minimum Temperature Requirements in the P-T curves The minimum temperature requirements *pecified in Appendix G to 10 CFR 50 are applied to the P/T curves using the NRC secepted methodologies as described in Appendix B, Reference 3.2. .

The minimum temperature values applied to the P/T curves for FCS corresponding to 20 Effective Full Power Years (EFPY) as described above in Sections 2.5.1.d and 2.5.1.c, respectively are:

Loc.ation Min Temoerature Boltup 82'F Lowest Service Temperature 182*F R0

FORT CALHOUN STATION TDB>lX TECHNICAL DATA BOOK Page 14 of 18 2.7 Surveillance Requirements 2.7.1 Reactor Coolant System and other components subject to ASME Section XI Boiler and Pressure Vessel Code Inspection and Testing Surveillance (T.S. 3.3).

The FCS post-irradiatiori ,veillance capsule test program was established in accordance with ASTM-E185 66 and compiles with 10CFR50 Appendix H. The test results are c'ven in Reference 3.9 through 3.11. The test results do meet the credib!hty criteria of Regulatory Guide 1.99 Revision 2, with the exception of item A because the surveillance weld does not represent the controlling reac'.or vessel beltline weld material.

The criteria are summarized below:

A. The surveillance program plate or wold duplicates the controlling reactor vessel beltline materialin terms of ART:

B. Charpy data scatter does not cause ambiguity in the determination of the 30 ft lb shift; 1

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1 C. The measured shifts are censistent with the predicted shifts; D. The capsule irradiation temperature is comparable to that of the vessel; and E. Correlation monitor data are available and are consistent with the known data for that material.

The credible surveillance data have not been used to refine the chemistry factor and the margin term.

3. REFERENCES 3.1 NRC GL 96-03, " Relocation of Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits". January 31,1996.

3.2 CE NPSD-683, Rev 02, " Development of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits and LTOP Requirements from the Technical Specifications, CEOG Task 942", December 1997.

3.3 FCS Technical Specifications, Amendment No.161.

3.4 Draft Regulatory Guide DG-1053, " Calculational and Dosimetry Methods for Determining Pressure Vesoci Neutron Fluence", June 1996.

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FORT CALHOUN STATION TDBIX TECHNICAL DATA BOOK Pcge 15 of 18 I

3.5 Draft Regulatory Guide DG 1025," Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence", Septemt ir 1993. i 3.6 SE REA-95-003," Fast Neutron Fluence Evaluations for Fort Calhoun Unit 1 Reactor Pressure Vessel", Nov6mber 1995.

3.7 Letter from OPPD (S.K. Gambhlr) to NRC (Document Control Desk), dated October 13,1997 (LIC-97-0159),

3.8 FCS USAR Table 4.5-4," Capsule Removal Schedule" 3.9 Letter from OPPD (W.C. Jones) to NRC (H.R. Denton), dated January 23,1981 (LIC-81-0011). Enclosuro: TR-O-MCM-001, Rev. 01, " Evaluation of Irradiated Capsule W 225".

3.10 Letter from OPPD (W.C. Jones) to NRC (D.G. Eisenhut), dated April 25,1984.

Enclosure; TR-O-MCM 002," Post-Irradiation Evaluation of Reactor Vessel Surveillance Capsule W 265".

3.11 Letter from OPPD (T.L. Patterson) to NRC (Document Control Desk), dated December 9,1994.

Enclosure:

BAW 2226 " Evaluation of Irradiated Capsule W 275".

3.12 TR-O MCD-001,"OPPD FCS Unit No.1 Evaluation of Baseline Specimens Reactor Vessel Mrarlais Surveillance Program", March 22,1977.

3.13 Letter 0-PENG 97-012 from ABB/CE (J. Ghergurovich) to OPPD (K Holthaus),

dated August 22,1997. "OPPD Interim Report on ASME Appendix G Evaluation of -

Step Changes in RCS Temperatures (CEOG Task 1004)".

3.14 Letter 0-MPS90-043 from ABB/CE (A.A. Ostrov) to OPPD (KC. Holthaus), dated June 14,1990.

3.15 Letter 0-MPS 90053 from ABB/CE (A.A. Ostrov) to OPPD (KC. Holthaus), dated July 20,1990. ,

4. OPERATING FIGURES The referenced operating Figures follow.

4 R0

~

, . . . . v - _

-_y .

,.__m_. . , _ _ - ,.e , - . -- ,

p g. . -. - ._._ - - - , _ .. _ _ _

TDB-IX Page 16 of13 i I d

l FORT CAlllOUN STATION UNIT 1 P/T LDdITS. 20 EPPY a

2500 t 2500

c0 Fmn-75'F/HR%

i 2000 l

2000 l

l U

IN ,

1 1500 i SERVICI:

i

g. TEMPIBATURE
j. 182'F % ' ALLOWABLE HEATUP RATES '

j 1000 .

mp ty.F hm'FM

! s335 75 i

l > 335 100 t- .

75"F/HR 500 - 500 10C 'F/HR M. BOLTUP TEMP. B2'F' 0 0 0- 100 200 . 300 . 400 500 600 T. INDICATED REACTOR COOLANT SYSTEM 17.MPERA'IURE. 'F RCS Pmssure-Tempersare Omaha Public Power District Figure Limiu for Heamp. For: Calhoun Station- Urdt No.1 4,1

, i46 TDB-IX Page 17 or13 FORT CALHOUN STATION UNIT 1 P/T LIMITS,20 EFPY COOLDOWN AND INSERVICE TEST 2500 7 INSERVICE HYDROSTATIC 'IESTg 2

! 2000 4 000 I

4 U

' d 1500 h [ 400'F/HR E 'IEMPERATUhH i f- 182'F ALLOWABLE COOLDOWN RATE l TEMR t furr 'F RAM'FMR 1000 ,

<135 10

  • 135-285 30 *

, y > 285 100 l

f r

500 g

500 L

1 'F/HR ' " # /

30'F/HR 100'F/HR y MIN. BOLTUP TEMP, 82'F 0

200 300 400 0 0 100 ,

500 g

. T, INDICATED REACTOR COOLANT SYSTEM TEMPERATURE, 'F

  • Startup of the first RCP is a routine plant evolution and exceeding these cooldown limits is acceptable as analyzed in Reference 3.13, which demonstrates compilance with the crite-ria of 10CFR50 Appendix G and ASME Section III Appendix G.

t RCS Pressurs-Tempernmre Omaha Public Power District Figure

- Limiu for Cooldown Fort Cnihoun Station- Unit No.1 4.2

_~. _. _ _ . _ . _ _ . . _ _ . _ _ _ . _ = . - _ _

- , - g --- --- -. - - -------- ---- --- _ _ - _ - _ _ _ - .,

TDIl-IX Page M of 18 O

k Predicted Radiation Induced NDTT Shift Fort Calhoun Reactor Vessel Beltline ARTndt

  • I ,
i g i e i

' j l

l

)

450 '

l '

~

I I I l ,# '

!.D. SHIFT l Including Margin l

/g  !

l

[/g I t/4t SHIFT

/1 Including Margin 230 #

3/41SHET

[ /r IncludingMargin

/

/

/

/

t r

150 ,

100 0.0 0.3 1.0 1.5 2.0 2.5 3.0 3.3 4.0 4.5 3.0 Neutron Fluence, IE19 n/cm2 Predicted Radiation Induced Omaha Public Power District Figure

, NDTT Shltt _, Fort Calhot'n Station-Unit No.1 4.3 l r _-_o

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