ML20039G798
| ML20039G798 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 01/15/1982 |
| From: | OMAHA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML20039G786 | List: |
| References | |
| NUDOCS 8201190155 | |
| Download: ML20039G798 (14) | |
Text
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I 2.0 LIMITING CONDITIONS FOR OPERATION I
2.1 Reactor Coolant System (Continued) 2.1.2 Heatup and Cooldown Rate (Continued) l (a)
The curve in Figure 2-3 shall be used to predict the increase in transition temperature based on inte-grated fast neutron flux.
If measurements on the l
irradiation specimens indicate a deviation from this curve a new curve shall be constructed.
j 1
I (b)
The limit line on the figures shall be updated for a J
new integrated power peric' as follows: the total 1
integrated reactor thermal power from startup to the j
end of the new period shall be converted to an equiva-lent integrated fast neutron exposure (E > 1 MeV).
For this plant, based upon surveillance materials tests, the predicted surface fluence at the reactor l
vessel belt-line weld material for 40 years at 1500 2
I MWt and an 80% load factor is 4.4 x 1019 n/cm. The predicted transition temperature shift to the end of the new period shall then be obtained from Figure 2-3.
I (c)
The limit lines in Figure 2-1A through 2-2B shall be moved parallel to the temperature axis (horizontal) in the direction of increasing temperature a distance l
equivalent to the transition temperature shift during l
the period since the curves vere last constructed.
l The boltup temperature limit line shall remain at l
820F as it is set by the NDf7 of the reactor vessel i
flange and not subject to fast neutron flux. The j
icwest service temperature shall remain at 1620F i
because components related to this temperature are also not subject to fast neutron flux.
(d)
The minimur temperature at which the 1000F/hr cool-1 down rate curve may be used is defined by the LPSI I
pumps outlet pressure to provide for protection j
against low temperature overpressurization per Techni-cal Specification 2.3(3). The Technical Specifi-cation 2.3(3) shall be revised each time the curves of Figures 2-1A through 2-2B are revised.
i Basis All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to reactor coolant system temperature and pressure changes.Cl) These cyclic loads are introduced by normal unit load transients, reactor trips and startup and shutdown operation.
During unit startup and shutdown, the rates of temperature and pressure changes are limited. The design number of cycles for heatup and cooldown in based upon a rate of 1000F in any one hour period and for cyclic operation.
8201190155BNi17
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MAMEEA PDRADOCKOS000g P
.. c WHEREFORE, Applicant respectfully requests that Sections 2.1.2 and 2.3 and Figures 2-1A, 2-1B, 2-2A, and 2-2B of Appendix A to Facility Operating License No. DPR-40 be amended in the form attached hereto as Attachment A.
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OMAHA PUBLIC POWER DISTRICT
.7 By W. C. Jones Division Manager Production Operations Subscribed and sworn to before me I
this.U9.. day of January,1982.
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Notary Public m em-si.i. d a*aaa l J. T. GLEAsON
_... _ My Comm. Em M 26. W2 i un
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i 2.0 LIMITING CONDITIONS PJR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 Heatup and Cooldown Rate (Continued)
The maximum allowable reactor coolant system pressure at any temperature is based upon the stress limitations for brittle fracture considerations. These limitations are derived by using the rules contained in Section III(2) of the ASME Code including Appendix G, Protection Against Nonductile Toughness Require-ments. This ASME Code assumes that a crack 10-11/16 inches long and 1-25/32 inches deep exists on the inner surface of the a
vessel. Furthermore, operating limits on pressure and temper-at ce assure that the crach does not grow during heatups and cooldowns.
The reactor vessel belt-line material consists of six plates.
The nilductility transition temperature (TNDT) f each plate was established by drop weight tests.
Charpy tests were then performed to determine at what temperature the plates exhibited 50 f t/lbs. absorbed energy and 35 mils lateral expansion for the longitudinal direction. NRC technical position MTEB 5-2 was used to establish a reference temperature for transverse di-rection (RTNDT) of -12 F.
Similar testing was not performed on all remaining material in the reactor coolant system. However, sufficient impact testing was performed to meet appropriate design code requirements (3) and a conservative RTNDT of 500F has been established.
As a result of fast neutron irradiation in the region of the core, there will be an increase in the TNDT 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 FSAR, except that the integrated fast neutron flux (E > 1 19 2
MeV) is 4.4 x 10 n/cm, including tolerance, over the 40 year design life of the vessel.(5)
Since the neutron spectra and the flux measured at the samples and reactor vessel inside radius should be nearly identical, the measured transition 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 flux magnitude.
The maximum exposure of the reactor vessel will be obtained from the measured sample exposure by application of the calibrated azimuthal neutron flux variation. The maximum integrated fast neutron (E > 1 MeV) exposure of the reactor vessel including tolerance is computed to be 4.4 x 1019 n/cm2 for 40 years oper-ation at 1500 MWt and 80% load factor.(5) The predicted TNDT shift for an integrated fast neutron (E > 1 MeV) exposure of 4.4 x 1019 n/cm2 is 344 F, the value cbtained from the curve shown l
in Figure 2-3.
The actual shift in TSDT will be re-established periodically during plant operation by testing of reactor vessel material samples which are irradiated cumulatively by securing them near the inside wall of the reactor vessel as described in Section 4.5.3 and Figure 4.5-1 of the FSAR. To compensate for any increase in the TNDT caused by irradiation, limits on the Amendment No. 22, 47 2-5
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2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) i l
2.1.2 Heatup and Cooldown Rate (Continued) l L
pressure-temperature relationship are periodically changed to l
stay within the stress limits during heatup and cooldown.
l Analysis of the first removed irradiated reactor vessel sur-l veillance specimen has shown that the fluence at the end of 6.1 l
Effective Full Power Years (EFPY) at 1500 MWt will be 8.4 x l
. 1018 n/cm2 on the inside surface of the reactor vessel.(5) This I
results in a total shif t of the RT f 2380F for the area of NDT greatest sensitivity (wo.ld metal) at the 1/4t location as deter-mined from Figure 2-3.
Operation through fuel cycle 7'will re-sult in less than 6.1 EFPY.
The limit lines in Figure 2-2A through 2-2B are based on the following:
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A.
Heatup and Cooldown Curves - From Section III of the ASME Code Appendix G-2215.
KIR = 2 Ktg + KIT KIR = Allowance stress intensity factor at temperatures related to RTNDT (ASME III Figure G-2110.1).
K g = Stress intensity factor for membrane stress (Pres-t sure). The 2 represents a safety factor of 2 on pressure.
l l
KIT = Stress intensity factor radial thermal gradient, i
The above equation is applied to the reactor vessel belt-line. For plant heatup the thermal stress is opposite in sign from the pressure stress and consideration of a heatup rate would allow for a higher pressure. For heatup it is therefore conservative to consider an isothermal heatup or KIT = 0.
For plant cooldown thermal and pressure stress are addi-tive.
KIM - MM PJt t
M3 = ASME III, Figure G-2214-1 l
P = Pressure, psia l
R = Vessel Radius - in.
t = Vessel Wall Thickness - in.
KIT =MATg T
l MI = ASME III, Figure G-2214-2 ATy = Highest Radial Temperature Gradient Through Wall at End of Cooldown Amendment No. 22. 47 2-6 j-.-. -
2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 Heatup and Cooldown Rate (Continued) is therefore calculated at a maximum gradient and is KIT considered a constant = A fer cooldown and zero fcr heatup.
R is also a constant = B.
M3 t
Therefore:
KIR = AP + B P = Kin - B A
is then varied as a function of temperature from KIR Figure G-2110-1 of ASME III and the allowable pressure calculated. Hydrostatic head (48 psi) and instrumentation errors (120F and 32 psi) are considered when plotting the curves.
B.
System Hydrostatic Test - The system hydrostatic test curve is developed in the same manner as in A above with the exception that a safety factor of 1.5 is allowed by ASME Ill in lieu of 2.
C.
Lowest Service Temperature = 500F + 1000F + 120F = 1620F.
for all material with the As indicated previously, an RTNDT exception of the reactor vessel belt-line was established l
at 500F. ASME III, Art. NE-2332(b) requires a lowest service temperature of RTNDT + 100 P for piping, pumps and valves. Below this temperature a pressure of 20 percent of the system hydrostatic test pressure (.20)(3125) 32 psi = 545 psia cannot be exceeded.
D.
Boltup Temperature = 100F + 600F + 120F = 820F. At pres-oure below 545 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 100F NDTT of the reactor vessel flange, which is not subject to radiation damage, plus 600F data scatter in NDTT measurements, plus 120F instrument error.
E.
Reactor Critical Heatup and Cooldown Figures. During low physics testing, the rear.cor may be made critical at re-duced temperature and pressure. To provide for heatup and cooldown during testing, Appendix C requires that the RCS temperature be increased an additional 400F beyond heatup and cooldown curves for the non-critical reactor. Also, Appendix G requires that the RCS temperature must be greater than the minimum temperature, 4170F, required for l
the 3125 psia hydrostatic testing to 125% of the 2500 psia RCS design pressure.
Amendment No. 22, 47 2-7
2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued)
- 2.1.2 lieatup and Cooldown Rate (Continued)
F.
Minimum Temperature for 100 F/hr Cooldown Rate = 1580F.
This limit provides protection against low temperature overpressurization during operation of the LPSI pumps.(6)
This temperature corresponds to a pressure of 206 psia on the 100 F/hr curve, which is the LPSI pump dead head and 0
minimum flow pressure.
For temperatures of 158 F or less, a cooldown rate of 200F/hr maximum will allow unrestricted operation of the LPSI pumps so that shutdown cooling may be utilized.
References (1) FSAR, Section 4.2.2 (2) ASME Boiler and Pressure Vessel Code,Section III (3) FSAR, Section 4.2.4 (4) FSAR, Section 3.4.6 (5) Omaha Public Power District, Fort Calhoun Station Unit No.
1, Evaluation of Irradiated Capsule W-225, Revision 1, August, 1980.
(6) Technical Specification 2.3(3) 2-7a
RCSPRESS-TEMPLIMITSHEATUP 6.1 EFPY 1500m EACI@ NOT GITIX RESS$1IG PESS PSIA) 3200 l
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50 100 150 200 250 300 350 400 450 500 E E D M Tc FORT CALHOUN TECHNICAL FIGURE SPECIFICATIONS 2-18 Amendment No. 22, 47
RCSPRESS-TEMPl.IMITSHEATUP 6.1EFPY eo m EXIG GITICA.
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FIGURE SPECIFICATIONS 2-2A Amendment No. 22, 47
.-..1 RCS PRESS-TEMP LIMITS C00LD0',itl 6.1 EFPY so m 1
EACTOR CRITICAL RES3Jt!BIESS FSIA) 3200 3000 2500 M
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50 100 150 200 250 300 350 400 450 500 AC LtET TDP DES fl Ic FORT CALHOUN TECHNICAL FIGURE SPECIFICATIONS 2-28 l
l Amendment No. 22, 47
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2.0 LIMITING CONDITIONS FOR OPERATION 2.3 Emergency Core Cooling System (Continued)
(3)
Protection Against Low Temperature Overpressurization The following liniting conditions shall be applied during scheduled heatups and cooldowns. Disabling of the HPSI pumps need not be required if the reactor vessel head, a pressurizer safety valve, or a PORV is removed.
Whenever the reactor coalant system cold leg temperature is below 3370F, at least one (1) HPSI pump shall be disabled.
l Whenever the reactor coolant system cold leg temperature is below 3270F, at least two (2) HPSI pumps shall be disabled.
l Whenever the reactor coolant system cold leg temperature is below 2930F, all three (3) HPSI pumps shall be disabled.
l Whenever the reactor coolant system cold leg temparature is below 1580F, the coolaown rate of Figure 2-1B, Technical Specification 2.1.2, shall be limited to a maximum rate of 200F/hr.
In the event that no charging pumps are operable, a single HPSI pump may be made operable and utilized for boric acid injection to the core.
Basis The normal procedure for starting the reactor is to first heat the reactor coolant to near cperating temperature by running the reactor coolant pumps. The reactor is then made critical by withdrawing CEA's and diluting boron in the reactor coolant. With this mode or. start-up, the energy stored in the reactor coolcut during the approach to criticality is substantially equal to that during power operation and therefore all engineered safety features and auxillcry cooling systems are required to be fully operable.
During low power physics tests at low temperatures, there is a negligible amount of stored energy in the reactor coolant; there-fore, an accident comparable in severity to the design basis accident is not possible and the engineered safeguards systems ate not required.
The SIRW tank contains a minimum of 283,000 gallons of usable water containing at least 1700 ppa boron.(1) This is sufficient boren concentration to provide a shutdown margin of 5%, including allowances for uncertainties, with a 1 control rods withdrawn and a new core at a temperature of 60 F. 2)
The limits for the safety injection tank pressure and volume assure the required amount of water injection during an accident and are based on values used for the accident analyses. The 3 and minimum 116.2 inch level corresponds to a volume of 825 ft the maximum 128.1 inch level corresponds to a volume of 895.5 ft3 Prior to the time the reactor is brought critical, the valving of the safety injection system must be checked for correct alignment and appropriate valves locked. Since the system is Amendment No. 17, 39, 43, 47 2-22
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2.0 LIMITING CONDITIONS FOR OPERATION 2.3 Emergency Core Cooling System (Continued)
With respect to the core cooling function, there is functional
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redundancy over most of the range of break sizes.(3)(4)
The LOCA analysis confirms adequate core cooling for the break spectrum up to and including the 32 inch double-ended break assuming the safety injection capability which most adversely i
affects accident consequences and are defined as follows. The entire contents of all four safety injection tanks are assumed to be available for emergency core cooling, but the contents of one of the tanks is assumed to be lost through the reactor coolant system.
In addition, of the three high-pressure safety injection pumps and the two low-pressure safety injection pumps, for large break analysis it is assumed that two high pressure and one low pressure operate while only(one of each type is assumed to operate in the small break analysis 5); and also that 25% of their com-bined discharge rate is lost from the reactor coolant system out of the break. The transient hot spot fuel clad temperatures for i
the break sizes considered ara shown on FSAR Figures 1-19 (Amend-ment N'. 34).
Inadverter.t actuation of three (3) HPSI pumps and three (3) l charging pumps, coincident with the opening of one of the two i
PORV's, would result in a peak primary system pressure of 1190 psia.
1190 psia corresponds with a minimum permissible temper-ature of 337 F on Figure 2-1B.
Thus, at least one HPSI pump is l
disabled at 3370F.
Inadvertent actuation of two (2) HPSI pumps and three (3) charging pumps, coincident with the openic.s of one of the two PORV's, would J
result in a peak primary system pressure of 1040 psir4 1040 psia corresponds with a minimum permissible temperature of 327 F on l
Figure 2-1B.
Thus, at least two HP3I pumps will be disabled at 3
327 F.
l 1
I Inadvertent actuation of one (1) HPSI and three (3) charging pumps, coincident with opening of one of the two PORV's, would i
result in a peak primary system pressure of 685 psia. 685 psia corresponds with a minimum allowable temperature of 2930F on l
Figure 2-1B.
Thus, all three HPSI pumps will be disabled at 2930F.
l The operation of either or both ifSI pumps, with or without three charging pumps, coincident with the opening of one of two PORV's, would result in a peak primary system pressure of 206 psia. This is the LPSI pump dead head and minimum flow pressure.
206 psia corresponds with a minimum allowable temperature of 1580F on the 1000F/hr cooldown rate curve of Figure 2-1B. (6) Since it is necessary that the LPSI pumps be available'for shutdown cooling, they cannot be disabled. Thus, whenever the cold leg temperature is less than or equal to 15807, a maximum cooldown rate of 20 F/hr shall be required.
Amendment No. 39, 47 2-23a
2.0 LIMITING CONDITIONS FOR OPERATION 2.3 Emergency Core Cooling System (Continued)
Inadvertent actuation of three (3) charging pumps, coincident with the opening of one of two PORV's, would result in a peak primary system pressure of 160 psia.
160 psia would correspond with a minimum allowable temperature of 121 F on the 100 F/hr cooldown rate curve of Figure 2-1B.
This is less than the minimum 1580F temperature required for use of the 100 F/hr cooldown rate curve.
th-200F/hr cooldown rate curve is controlling and does not
- inus, limit the ty_ ration of the char ing pumps.
o l
Removal ot the reactor vessel head, one pressurizer safety valve, or one FORV provides sufficient expansion volune :o limit any of the design basis pressure transients. Thus, no additional relief capacity is required.
Technical Specification 2.2(1) specifies that, when fuel is in the least one flow path shall be provided for boric acid reactor, at injection to the core.
Should boric acid injection become neces-sary, and no charging pumps are operable, operation of a single HPSI pump would provide the required flow path.
References (1)
FSAR, Section 14.15.1 (2)
FSAR, Section 6.2.3.1 (3)
FSAR, Section 14.15.3 (4)
FSAR, Appendix K (5)
Omaha Public Power District's Submittal, December 1, 1976 (6)
Technical Specification 2.1.2, Figure 2-1B Amendment No. 47 2-23b v
DISCUSSION These changes are required to allow for the safe opercticn of the reactor and associated primary coolant system beyord the 'i.2 Equivalent Full Power Years (EFPY) of operation to which the present Technical Specifications am written. The specifications are revised to allow operation through 6.1 EFPY. This will provide operating limits through the end of fuel cycle 7.
The EFPY of operation corresponds to a neutron fluence received by the reactor vessel. This fluence causes the nil-ductility transition reference temperature (RTNDT) of the reactor vessel steel to increase. The amount of RTNDT shift is predicted using procedures detailed in Regulatory Guide 1.99.
The fluence value for the reactor vessel belt-line weld material used for getermjning the RTnDT 1
e The fluence shif t through 6.1 EFPY of operation is 8.4 x 10 - n/cr.
vahe for 6.1 FFPY was calculated using the end-of-life predicted fluence of 4.4 x 1019 n/cm2 which was calculated and approved by the Commission for cycle 6 operation using the Fort Calhoun Station first surveillance capsule test data, as reported in the Combustion Engineer-ing document " Evaluation of Irradiated Capsule W-225", Revision 0, dated May 1979.
It should be noted that the CE report, " Evaluation of Ir-radiated Capsule W-225," Revision 1, dated August 1980, using imprgyed calculational techniques determined the EOL fluence to be 4.2 x 10 0 n/cm2 Accordingly, the E0L value and 6.1 EFPY fluence values used in the proposed Technical Specifications are considgred conservative. The calculated RTNOT t tal shift for 8.4 x 1018 r/cm' is 2380F for the bel t-line weld material. The heatup and cooldown rate pressure-temperature limit curves were then adjusted according to 10 CFR 50, Appendices G and H, to ensure that adequate fracture toughness is maintained through all conditions of normal operation, including anticipated operational transients and system hydrostatic tests. The beginning-of-life RTNDT value for weld materials used for developing the heatup and cooldown limit curves was 00F in accordance with Branch Technical Position MTEB 5-2.
The disabling of HPSI pumps in order to ensure protection of the RCS against low temperature overpressurization is dependent upon the pemissible 1000F/hr cooldown rate which is determined from the shift in RTNDT and plotted on Figure 2-18.
Thus, Technical Specification 2.3(3) is modified to maintain adequate overpressurization protection.
In addition, it has become apparent that the LPSI pump shutoff head will not allow operation on the 1000F/hr cooldown rate curve of Figure 2-1B at temperature at or below 1580F.
It becomes necessary to limit oper-ation at that temperature and below to a cooldown rate of 200F/hr.
The present Technical Specifications, valid through 5.2 EFPY, will provide operating limits for a period of 47 days of full power operation (71,662 MW-HR) after initial criticality of fuel cycle 7.
Therefore, Commission approval of the proposed Technical Specifications by no later than February 8,1982 is requested.
ATTACHNENT B 9
FEE JUSTIFICATION The proposed Facility License Amendment is deemed to be a Clas:,
III Amendment within the meaning of 10 CFR 170.22. This determination is made in that it involves only a single safety issue and does not involve a significant hazards consideration.
l 1
ATTACHMENT C
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