ML14183A264
| ML14183A264 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 07/29/1994 |
| From: | Matthews D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML14183A265 | List: |
| References | |
| DPR-23-A-149 NUDOCS 9408040317 | |
| Download: ML14183A264 (9) | |
Text
-2 B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 149, are hereby incorporated in the license. Carolina Power & Light Company shall operate the facility in accordance with the Technical Specifications.
- 3.
This license amendment is effective as of the date of its issuance and shall be implemented within 60 days.
FOR THE NUCLEAR REGULATORY COMMISSION David B. Matthews, Director Project Directorate II-1 Division of Reactor Projects -
I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: July 29, 1994 9409040317 940729 PDR ADOCK 05000261 P
ATTACHMENT TO LICENSE AMENDMENT NO.149 FACILITY OPERATING LICENSE NO. DPR-23 DOCKET NO. 50-261 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.
Remove Pages Insert Pages v
v 3.1-21 3.1-21 3.1-21a 3.1-22 3.1-22 3.1-22a 3.1-4 3.1-4 3.1-5 3.1-5 3.1-6 3.1-6 3.1-7 3.1-7
LIST OF FIGURES Figure Title Page 1.1-1 Plant Site Boundary and Exclusion Zone 1-8 2.1-1 Safety Limits Reactor Core, Thermal, and Hydraulic Three Loop Operation, 100% Flow 2.1-4 3.1.4-1 Percent of Rated Thermal Power 3.1-15a 3.1-1 Reactor Coolant System Heatup Limitations Applicable Up to 24 EFPY 3.1-21 3.1-2 Reactor Coolant System Cooldown Limitations Applicable Up to 24 EFPY 3.1-22 3.10-1 (DELETED) 3.10-20 3.10-2 Shutdown Margin versus Boron Concentration 3.10-21 3.10-3 (DELETED) 3.10-22 3.10-4 (DELETED) 3.10-23 3.10-5 (DELETED) 3.10-24 6.2-1 Offsite Organization for H. B. Robinson 2 Management and Technical Support 6.2-3 6.2-2 Conduct of Operations Chart 6.2-4 v
Amendment No. 8,747,149
3.1.2 Heatup and Cooldown 3.1.2.1 The reactor coolant pressure and the system heatup and cooldown rates (with the exception of,the pressurizer) shall be limited in accordance with Figure 3.1-1 and Figure 3.1-2 (for vessel exposure up to 24 EFPY).
These limitations are as follows:
- a.
Over the temperature range from cold shutdown to hot operating conditions, the heatup rate shall not exceed 60aF/hr. in any one hour.
- b.
Allowable combinations of pressure and temperature for a.
specific cooldown rate are below and to the right of the limit lines for that rate as shown on Figure 3.1-2. This rate shall not exceed 100aF/hr. in.any one hour. The limit lines for cooling rates between those shown in Figure 3.1-2 may be obtained by interpolation.
- c.
Primary system hydrostatic leak tests may be performed as necessary, provided the temperature limitation as noted on Figure 3.1-1 is not violated. Maximum hydrostatic test pressure should remain below 2350 psia.
- d.
The overpressure protection system shall be operable whenever the RCS temperature is below 350aF and not vented to the containment. One PORV may be inoperable for seven days. If the inoperable PORV has not been returned to service within 7 days, or if at any time both PORVs become inoperable, then one of the following actions should be completed within 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />s:
- 1.
Cooldown and depressurize the RCS or 3.1-4 Amendment No. 89j,149
- 2.
Heatup the RCS to above 350aF.
- e.
Operation of the overpressure protection system to relieve a pressure transient must be reported as a Special Report to the NRC within 30 days of operation.
3.1.2.2 The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the vessel is below 120aF.
3.1.2.3 The pressurizer shall neither exceed a maximum heatup rate of 100aF/hr nor a cooldown rate of 200*F/hr. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320 0F.
3.1.2.4 Figures 3.1-1 and 3.1-2 shall be updated periodically in accordance with the following criteria and procedures before the calculated exposure of the vessel exceeds the exposures for which the figures apply.
- a.
At least 60 days before the end of the integrated power period for which Figures 3.1-1 and 3.1-2 apply, the limit lines on-the figures shall be updated for a new integrated power period utilizing methods derived from the ASME Boiler and Pressure Vessel Code,Section III, Appendix G and in accordance with applicable appendices of 10CFR50. These limit lines shall reflect any changes in predicted vessel neutron fluence over the integrated power period or changes resulting from the irradiation specimen measurement program.
- b.
The results of the examinations of the irradiation specimens and the updated heatup and cooldown curves shall be reported to the Commission within 90 days of completion of the examinations.
3.1-5 Amendment No.97,77,
steels such as ASTM A302 Grade B parent material of the H. B. Robinson Unit No. 2 reactor pressure vessel are well documented in the literature.
Generally, low alloy ferritic materials show an increase in hardness and other strength properties and a decrease in ductility and impact toughness under certain conditions of irradiation. Accompanying a decrease in impact strength is an increase in the temperature for the transition from brittle to ductile fracture.
A method for guarding against fast fracture in reactor pressure vessels has been presented in Appendix G, "Protection Against Non-Ductile Failure," to Section III of the ASME Boiler and Pressure Vessel Code. The method utilizes fracture mechanics concepts and is based on the reference nil-ductility temperature, RTNDT.
RTNOT is defined as the greater of:
- 1) the drop weight nil-ductility transition temperature (NDTT per ASTM E-208) or 2) the temperature 60OF less than the 50 ft-lb (and 35 mils lateral expansion) temperature as determined from Charpy specimens oriented in a direction normal to the major working direction of the material.
The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (KmR curve) which appears in Appendix G of Section III of the ASME Boiler and Pressure Vessel Code. The KR curve is a lower bound of dynamic, crack arrest, and static fracture toughness results obtained from several heats of pressure vessel steel.
When a given material is indexed to the KmR curve, allowable stress intensity factors can be obtained for this material as a function of temperature.
Allowable operating limits can then be determined utilizing these allowable stress intensity factors.
The Certified Material Test Reports (CMTRs) for the original steam generators provided records of Charpy V-notch tests performed at +10*F. Acceptable Charpy V-notch tests of +100F indicate RTNOT is at or below this temperature.
The steam generator lower assemblies were replaced in 1984 and the material tests results indicate the highest RTNOT is 60aF or below.
The ASME code recommends that hydrostatic tests be performed at a temperature not lower than RTNDT plus 60Fo, thus the pressurizing temperature for the steam generator shell is established at 120OF to provide protection against nonductile failure at the test pressure.
3.1-6 Amendment No. 01,149
V-notch 30 ft-1b temper ure (A RTNOT) due to irradiation is added to the original ARTNDT to adjust the RTNOT for radiation embrittlement.
This adjusted RTNDT (RTNOT initial + ARTNDT) is utilized to index the material to the KIR curve and in turn to set operating limits for the nuclear power plant which take into account the effects of irradiation on the reactor vessel materials.
Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods (2) derived from Appendix G to Section III of the ASME Boiler and Pressure Vessel Code. The approach specifies that the allowable total stress intensity factor (K,) at any time during heatup or cooldown cannot be greater than that shown on the KR curve in Appendix G for the metal temperature at that time. Furthermore, the approach applies an explicit safety factor of 2.0 on the stress intensity factor induced by pressure gradients.
Following the generation of pressure-temperature curves for both the steady state and finite heatup rate situations, the final limit curves are produced in the following fashion. First, a composite curve is constructed based on a point-by-point comparison of the steady state and finite heatup rate data. At any given temperture, the allowable pressure is taken to be the lesser of the two values taken from the curves under consideration. The composite curve is then adjusted to allow for possible errors in the pressure and temperature sensing instruments.
The use of the composite curve is mandatory in setting heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling analysis switches from the 0.0. to the I.D.
location; and the pressure limit must, at all times, be based on the most conservative case. The cooldown analysis proceeds in the same fashion as that for heatup, with the exception that the controlling location is always at the I.D. position. The thermal gradients induced during cooldown tend to produce tensile stresses at the I.D. location and compressive stresses at the 0.D.
position. Thus, the I.D. flaw is clearly the worst case.
As in the case of heatup, allowable pressure temperature relations are generated for both steady state and finite cooldown rate situations.
Composite limit curves are then constructed for each cooldown rate of 3.1-7 Amendment No. $,149
MATERIALS PROPERTIES BASE Controlling Material Lower Circumferential Weld Curves applicable for heatup Copper Content 0.20 wt.%
rates up to 60'F/Hr for the Nickel Content 1.06 wt.%
service period up to 24 EFPY.
RTNDT Initial
-80*F Includes +10*F and -60 PSIC RTNDT After 24 EFPY 1/4 T. 207.83'F allowance for instrumentation 3/4 T. 137.15'F error.
3000 2500 I
v1 2000 Leak Test Limit (II II II T o 6 0I'F / H rII II I 20 Le 1 50 0 0
Q I I CrV i
[ti l Limi (21000 hdottc ts O
temperature (327F)
.Z
.for the service period 500 II-I 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE
('F)
H.. Robbson Unit *2 Reactor Coolant System l I.
FIGURE Heatup Limitations CAROLINA POWER & LIGHT COMPANY Technical Specifications Applicable Up To 24 EFPY 3.1-21 Amendment No. $2,X,149
MATERIALS PROPERTIES BASE Controlling Material Lower Circumferential Weld Curves applicable far cooldown Copper Content 0.20 wt.%
rates up to 100F/Hr for the Nickel Content 1.06 wt.%
service period up to 24 EFPY.
RTNOT Initial
- -80'F Includes +10'F and -60 PSIG RTNDT After 24 EFPY 1/4 T. 207.83F allowance for instrumentation 3/4 T. 137.15'F error.
3000 2500 2000 I
LIJ 1500 0
1000 50 III100150 20 2501 30 350 400 450 500 IIIIII e
ooldown Rotes Lmao 1IGURE
('F/Hr)11 1
11 1
1 1
Techica Speifcaton Appial Up1 To11 241 EFP 311.1-21.
11 3 1 2 100 No.-----
50
.100 150 200 250 300.
350 400 450 500 INDICATED TEMPERATURE (7F)
H.B. Robinson Unit #2 Reactor-Coolant System FIGURE Cooldown Limitations CAROLINA -POWER & LIGHT COMPANY Technical Specifications Applicable Up To 24 EFPY 3.
3.1-22 Amendment No. 9Z,Y13, 149