ML13316B006
| ML13316B006 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 05/23/1988 |
| From: | Knighton G Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML13316B005 | List: |
| References | |
| DPR-13-A-102 NUDOCS 8806030232 | |
| Download: ML13316B006 (22) | |
Text
~PdR REG(,,0 o10 UNITED STATES 0
NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SOUTHERN CALIFORNIA EDISON COMPANY SAN DIEGO GAS AND ELECTRIC COMPANY DOCKET NO. 50-206 SAN ONOFRE NUCLEAR GENERATING STATION, UNIT NO. 1 AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 102 License No. DPR-13
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Southern California Edison Company and San Diego Gas and Electric Company (the licensee) dated August 29, 1977, October 20, 1978, May 8, 1984 and January 21, 1986 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
8806030232 860523 PDR ADOCK 05000206 P
-2
- 2. Accordingly, the license is amended by changes to the Technical Speci fications as indicated in the attachment to this license amendment, and paragraph 3.B. of Provisional Operating License No. DPR-13 is hereby amended to read as follows:
B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 102, are hereby incorporated in the license. Southern California Edison Company shall operate the facility in accordance with the Technical Specifications, except where otherwise stated in specific license conditions.
- 3. This license amendment is effective as of the date of its issuance and must be fully implemented no later than 30 days from the date of issuance.
FOR THE NUCLEAR REGUL TORY COMMISSION George W Knighton rector Project irectorate V Division of Reactor Projects -
- III, IV, V and Special Projects
Attachment:
Changes to the Technical Specifications Date of Issuance: May 23, 1988
May 23, 1988 ATTACHMENT TO LICENSE AMENDMENT NO. 102 PROVISIONAL OPERATING LICENSE NO. DPR-13 DOCKET NO. 50-206 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.
REMOVE INSERT Table of Contents (ii & iii)
Table of Contents (ii & iii) 10c 10c 10d 10d 11 11 12 12 13 13 14 14 16a 16a 16c 16c 16d 19 19 20 20 20a 23 23 24 24 39gg 39hh 60dd
Page 3.5.3 Control and Shutdown Rod Misalignment......
33b 3.5.4 Rod Position Indicating System.........
33g 3.5.5 Containment Isolation Instrumentation......
33j 3.5.6 Accident Monitoring Instrumentation.......
33n 3.5.7 Auxiliary Feedwater Instrumentation.......
33p 3.5.8 Radioactive Liquid Effluent Instrumentation.
33s 3.5.9 Radioactive Gaseous Process and Effluent Monitoring Instrumentation...
33v 3.5.10 Radiation Monitoring Instrumentation......
33y 3.6 CONTAINMENT 34 3.7 AUXILIARY ELECTRICAL SUPPLY 36 3.8 FUEL LOADING AND REFUELING...
38 3.9 CORE AVERAGE BURNUP 39 3.10 INCORE INSTRUMENTATION.................
39a 3.11 CONTINUOUS POWER DISTRIBUTION MONITORING......... 39c 3.12 CONTROL ROOM EMERGENCY AIR TREATMENT SYSTEM 39e 3.13 SHOCK SUPPRESSORS (SNUBBERS) OPERABILITY........
39g 3.14 FIRE PROTECTION SYSTEMS OPERABILITY 39i 3.15 RADIOACTIVE LIQUID EFFLUENTS.
39n 3.16 RADIOACTIVE GASEOUS EFFLUENTS 39q 3.17 DOSE..........................
39w 3.18 RADIOLOGICAL ENVIRONMENTAL MONITORING..........
39x 3.19 SOLID RADIOACTIVE WASTE.................
39ff 3.20 OVERPRESSURE PROTECTION SYSTEMS.............
39gg SECTION 4 SURVEILLANCE REQUIREMENTS..
40 4.1 OPERATIONAL SAFETY ITEMS.....
40b 4.2 SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEM..
45 4.3 CONTAINMENT SYSTEMS...................
47 4.4 EMERGENCY POWER SYSTEM PERIODIC TESTING.........
52 ii Amendment No. 00, 102
Page 4.5 RADIOACTIVE LIQUID EFFLUENTS...........
54 4.6 RADIOACTIVE GASEOUS EFFLUENTS..............
56 4.7 INSERVICE INSPECTION REQUIREMENTS............
58 4.8 REACTIVITY ANOMALIES...................
60 4.9 REACTOR VESSEL SURVEILLANCE PROGRAM...........
60-A 4.10 AUGMENTED INSERVICE INSPECTION OF HIGH ENERGY LINES OUTSIDE CONTAINMENT..
60-B 4.11 CONTROL ROOM EMERGENCY AIR TREATMENT SYSTEM....
60e 4.12 MISCELLANEOUS RADIOACTIVE MATERIALS SOURCES.......
60g 4.13 TURBINE DECK LOAD BEARING TEST AND VISUAL INSPECTION.
60-H 4.14 SHOCK SUPPRESSORS (SNUBBERS) SURVEILLANCE........
60-I 4.15 FIRE PROTECTION SYSTEMS SURVEILLANCE...........
60-L 4.16 INSERVICE INSPECTION OF STEAM GENERATOR TUBING......
60-0 4.17 DOSE.............
60w 4.18 RADIOLOGICAL ENVIRONMENTAL MONITORING..........
60x 4.19 SOLID RADIOACTIVE WASTE................
60cc 4.20 OVERPRESSURE PROTECTION.................
60dd SECTION 5 DESIGN FEATURES 5.1 SITE DESCRIPTION.
61 5.2 CONTAINMENT.......................
62 5.3 REACTOR.........................
65 5.4 AUXILIARY EQUIPMENT...................
67 SECTION 6 ADMINISTRATIVE CONTROLS.
6-1 6.1 RESPONSIBILITY...................... 6-1 6.2 ORGANIZATION....................... 6-2 FIGURE 6.2-1....................
6-5 FIGURE 6.2-2....................
6-6 TABLE 6.2-1..
6-7 iii Amendment No.
- 91.
102
-loc
- 1. Two residual heat removal (RHR) trains shall be operable*
and at least one RHR train shall be in operation**
- 2. With less than the above required RHR trains operable, immediately initiate corrective action to return the required RHR trains to Operable status as soon as possible.
- 3. With no RHR train in operation, suspend all operations involving a reduction in boron concentration of the reactor coolant system and immediately initiate corrective action to return the required RHR train to operation.
I. A reactor coolant pump shall not be started with the RCS pressure
< 400 psig unless:
(1) the pressurizer water level is less than 80%, or (2) the potential for having developed reactor coolant system temperature gradients has been evaluated.
Basis:
One pressurizer safety valve is sufficient to prevent over pressurizing when the reactor is subcritical, since its relieving capacity is greater than that required by the sum of the available heat sources, i.e., residual heat, pump energy and pressurizer heaters.
- One RHR train may be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing provided the other RHR train is operable and in operation.
- The RHR pump may be de-energized for up to one hour provided (a) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and (b) core outlet temperature is maintained at least 400 F below saturation temperature.
Amendment No.77, 102
-10d Prior to reducing boron concentration by dilution with make up water either a reactor coolant pump or a residual heat removal pump is specified to be in operation in order to provide effective mixing. During boron injection, the operation of a pump, although desirable, is not essential.
The-boron is injected into an inlet leg of the reactor coolant loop.
Thermal circulation which exists whenever there is residual heat in the core and the reactor coolant system is filled and vented, will cause the boron to flow to the core.
Lack of further mixing cannot result in areas of reduced boron concentra tion within the core. Prior to criticality the two pressurizer safety relief valves are specified in service in order to conform to the system relief capabilities.(1)
The plant is designed to have all three reactor coolant loops operational during normal power operation (Modes 1 and 2).
Under these conditions, the DNB ratio will not drop below 1.30 after a loss of flow with a reactor trip.(2)(3) With one reactor coolant loop not in operation, this specifi cation requires that the plant be in at least Hot Standby within one hour. However, exception is taken whenever reactor power is less than 10%
of full power. Heat transfer analyses show that reactor heat equivalent to 8% of full power can be removed with natural circulation only; hence, for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> the specified upper limit of 10% of full power with 1 or 2 reactor coolant pumps operating provides a substantial safety factor.
In modes other than Modes 1 and 2, functional redundancy in the core heat removal methods (not necessarily system redundancy) is specified to satisfy single failure considerations. Functional redundancy, as applied to the San Onofre Unit 1 power plant, includes use of diverse heat removal methods. Furthermore, single failure considerations apply only to active components.
In Mode 3, a single reactor coolant loop provides sufficient capability for removing decay heat; however, single failure considerations require that two loops be OPERABLE.
In Mode 4 and Mode 5 (reactor coolant loops filled), a single reactor coolant loop or RHR train provides sufficient capability for removing decay heat; but single failure considerations require that at least two methods (either RCS loop or RHR train) be OPERABLE.
In Mode 5 (reactor coolant loops not filled), a single RHR train provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of any of the steam gener ators as a heat removing component, require that at least two RHR trains be OPERABLE.
Amendment No.1/, 102
The operation of one reactor coolant pump or one residual heat removal pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control (4).
"The limitation on reactor coolant pump operation with the RCS pressure
< 400 psig ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50s.
A pres surizer water level of less than 80% ensures that the start of a reactor coolant pump, with a temperature differential of 1000 F will not result in 10 CFR Part 50 Appendix G limits being exceeded.
There are several means available for determining that there is not a temperature differential of > 500 F between the secondary and primary systems with < 400 psig primary system pressure. These methods may include but are not necessarily limited to the following:
- 1) Converting steam line pressure indication into maximum temperature of steam generator fluid.
- 2) Tagging RCP switches with shutoff temperatures.
- 3) Assuring adequate time for temperature gradients to dissipate.
- 4) Filling steam generators with water of known temperature.
References:
(1)
Final Engineering Report and Safety Analysis, Sections 9 and 10 (2)
Final Engineering Report and Safety Analysis, Paragraph 10.2 (3)
Supplement No. 1 to Final Engineering Report and Safety Analysis, Section 3, Question 9 (4)
NRC letter dated June 11, 1980 from D. G. Eisenhut to all operating pressurized water reactors.
(5)
Letter to A. Schwencer from K. Baskin dated October 12, 1977.
Amendment No. 43, it, 102
-12 3.1.3 COMBINED HEATUP, COOLDOWN AND PRESSURE LIMITATIONS Applicability:
Applies to heatup and cooldown of the reactor coolant system.
Objective:
To maintain the structural integrity of the reactor coolant system throughout the lifetime of the plant.
Specification:
A. Reactor pressure and heatup and cooldown of the reactor coolant system during the first 16 years of equivalent full power operation shall be limited in accordance with Figures 3.1.3a and 3.1.3b. Thereafter, limits shall be based on neutron exposure equivalent to not less than 16 years of full power operation, and Figures 3.1.3a and 3.1.3b shall be updated accordingly (by formal license amendment application).*
B. Figures 3.1.3a and 3.1.3b shall be updated in accordance with the following criteria and procedures:
(1) The methods of Appendix G, "Protection Against Nonductile Failure", to Section III of the ASME Boiler and Pressure Vessel Code shall be used to obtain the allowable pressure-temperature relationships for the reactor coolant system.
(2) The curves in Figure 3.1.3c shall be used in predicting the reference nil-ductility temperature increase ARTNpT, unless measurements on the irradiation s cimens show ART s greater than those predicted by the curves, in whyu case a new curve having the same slope as the original shall be constructed.
C. The pressurizer heatup rate of 100 0F/hour and cooldown rate of 200aF/hour shall not be exceeded.
D. The reactor shall not be brought to a critical condition until the pressure-temperature state is to the right of the criticality limit line as shown in Figure 3.1.3a.
Basis:
The initial Reference Nil Ductility Temperature (RT NDT) for all reactor vessel material based on Charpy V-notch data, drop weight tests, and conservative estimates** is 82oF or less. The RT at the 1/4 thickness location (location of Appendix G refereNRI flaw tip) increases as a function of cumulative neutron exposure up to approximately 240aF for the core region of the reactor vessel after 30 years of operation.
Technical Specification 3.20.A(1) should be reevaluated for continued applicability of the low pressure PORV overpressure setpoint at any time the heatup and cooldown curves are changed.
- NRC Standard Review Plan Branch Technical Position MTEB 5-2.
Amendment No. t$, $8, 102
-13 A sixteen (16) equivalent full power year service period was chosen for the operational limits given in this specification because at the end of this period the limiting RT of the reactor vessel at the 1/4 thickness location is a oximately 2170 F in the core region. This RT is at least 500 F above the RTNDT of all other regions in the OTmary reactor coolant system.
The highest RT of the core region material is determined by adding the rad lion induced ART T for the applicable time period to the original RT shown in t TTable 3.1.3.1. The fast neutron (E > 1Mev) fluenceNRT 1/4 thickness and 3/4 thickness vessel loca tions is given as a function of full power service life in Figure 3.1.3d. Using the applicable fluence at the end of the year period and the copper content of the material in question, the ARTNDT is obtained from Figure 3.1.3c.
Values of ART may continue to be determined in this manner unless measure nts on the irradiation specimens show ART s
greater than those predicted by the curves for the equiva 7 t capsule exposure.
Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from non mandatory Appendix G in Section III of the ASME Boiler and Pressure Vessel Code, and discussed in detail in Reference 1.
The results of these calculations are provided in Reference 2.
The design heatup and cooldown rates for the pressurizer are 1000 F/hour and 2000 F/hour, respectively.
The vertical line portion of the criticality limit given in Figure 3.1.3a is at the minimum permissible temperature for the 2485 psig in-service hydrostatic test as required by Appendix G to 10 CFR Part 50. The non-vertical portion of the criticality limit is shifted 400 F to the right of the heatup curve as required by Appendix G to 10 CFR Part 50.
References:
(1) "Pressure Temperature Limits" Section 5.3.2 of Standard Review Plan, NUREG-751087, 1975 (2) S. E. Yanichko, et al., "Analysis of Capsule F from the Southern California Edison Company San Onofre Reactor Vessel Radiation Surveillance Program," WCAP 9520, May 1979 Amendment No. If, 30, 102
-14 This page intentionally blank Amendment No. 14, 102
16a 400 300 200 150 a100 s6o 60 400O0%
CU BASE. 0.15% CU WELD 40 0.15% Cu BASE, 0.20% CU WELD OM0% CU aUSE, OJS cu WEtD 20
-0.15%
CU BASE, 0.10% CU WELD 0.10% CU BASE, 0.DS% CU WELD 0
10 2
5 101 2
5 10t FLUENCE (N/CM E >1 MEV)
FIGURE 3.1.3c Effect of Fluence and Copper Content on ART For Reactor Vessel Steels Exposed to IrradiatiofDt 550OF Amendment No. 14, 102
TABLE 3.1.3.1 REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED)
Minimum Average 50 ft-lb/35 mil RT Upper Shelf Material Cu P
NDTT Temp(oF)
NDT Energy (ft-lb)
Component Code No.
Type
(%)
(%)
(OF)
Long.
Trans.
(OF)
Long.
Trans.
Cl. Hd. Dome W7604 A302B 60(a) 112 132 72 72.5 Peel Segment W7605-1 A302B
-10 114 134 74 70.5 Peel Segment W7605-2 A302B
-10 90 110 50 122 Peel Segment W7605-3 A302B
-10 108 128 68 85 Peel Segment W7605-4 A302B
-10 120 140 80 74 Peel Segment W7605-5 A302B
-10 26 46 10 109 Peel Segment W7605-6 A302B
-10 102 122 62 88 Hd. Flange W7602 A336 mod 60(a) (b) 60 Ves. Flange W7603 A336 mod 60(a) (b) 60 Inlet Nozzle W7611-1 A336 mod 60(a) (b) 60 Inlet Nozzle W7611-2 A336 mod 60(a) (b) 60 Inlet Nozzle W7611-3 A336 mod 60(a) (b) 60 Outlet Nozzle W7610-1 A336 mod 60(a) (b) 60 Outlet Nozzle W7610-2 A336 mod 60(a) (b) 60 Outlet Nozzle W7610-3 A336 mod 60(a) (b) 60 Upper Shell W7601-3 A302B 0.15 0.014
-10 48 68 8
98.5 Upper Shell W7601-6 A302B 0.16 0.012
-30 64 84 24 104 Upper Shell W7601-7 A302B 0.15 0.014
-20 52 72 12 95.5
- a. Estimated per NRC Standard Review Plan Branch Technical Position MTEB 5-2.
C
TABLE 3.1.3.1 (continued)
REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED)
Minimum Average 50 ft-lb/35 mil RT Upper Shelf Material Cu P
NDTT Temp(oF)
NDT Energy (ft-lb)
Component Code No.
Type
(%)
(%)
(oF)
Long.
Trans.
(OF)
Long.
Trans.
Inter. Shell W7601-1 A302B 0.17 0.013 0
57 120(a) 60 94 75 Inter. Shell W7601-8 A302B 0.18 0.012 10 93 100(a) 40 97 79 Inter. Shell W7601-9 A302B 0.18 0.014 0
64 115(a) 55 102 72 Lower Shell W7601-2 A302B 0.17 0.013
-20 74 94 34 97 Lower Shell W7601-4 A302B 0.14 0.014
-10 91 111 51 94 Lower Shell W7601-5 A302B 0.14 0.014 10 122 142 82 87.5 Bot. Hd. Peel W7607 A302B
-20 62 82 22 91 Bot. Hd. Dome W7606 A302B 60(b) 99 119 60 86 Weld 0.19 0.017 0(b) 29(a) 0 90 HAZ 0(b)
-14(a) 0 101
- a. Actual not estimated
- b. Estimated per NRC Standard Review Plan Branch Technical Position MTEB 5-2.
O0 C+
CD
-19 3.2 CHEMICAL AND VOLUME CONTROL SYSTEM Applicability:
Applies to the operational status of the chemical and volume control system.
Objective:
To identify those conditions of the chemical and volume control system necessary to ensure safe reactor operation.
Specification:
A. When fuel is in the reactor, the following chemical and volume control system conditions shall be met:
(1) One charging pump or the test pump shall be operable.
However, when the RCS pressure is < 400 psig and pressurizer water level is greater than 50%, a maximum of one of the two centrifugal charging pumps shall be operable. The inoperable centrifugal charging pump shall have the motor circuit breaker removed from the electrical power supply circuit and shall be condition tagged.
(2) One boric acid transfer pump or the boric acid injection pump shall be operable.
(3) A solution of at least 3450 pounds of boric acid in not less than 3500 gallons of water at a temperature of 140'F or higher, with at least one heater operable, shall be in the boric acid tank.
(4) System piping and valves shall be operable to the extent of establishing two flow paths for boric acid tanks.
(5) During periods when borated water is in the refueling cavity, the requirements in A.1 through A.4 may be waived provided that an alternate source of borated water is available to establish at least one flow path to the core for boric acid injection which can be initiated from the control room. The minimum capability for boric acid addition shall be equivalent to that supplied by a charging pump from the refueling water storage tank.
B. The reactor shall not be made critical unless the following additional conditions are met:
(1) One additional charging pump or test pump operable.
(2) One additional boric acid transfer pump or boric acid injection pump operable.
(3) Electrical heat tracing for boric acid piping operable.
Amendment No. 102
-20 C. After criticality is achieved, maintenance on item B-3 will be allowed providing the boric acid temperature does not fall below 1400 F.
Basis:
The Chemical and Volume Control System(1 ) provides control of the reactor system boron concentration. This is accomplished by using either one of the two charging pumps or the test pump (Chemical and Volume Control System test pump installed in parallel with the charging pump) to inject concentrated boric acid solution into the reactor coolant system. There are two sources of borated water available for injection through three different paths as follows:
- 1. The boric acid injection pump can deliver the boric acid tank contents to the charging pump and/or test pump.
- 2. Boric acid transfer pumps can deliver the boric acid tank contents to the charging pumps and/or test pump.
- 3. The charging pumps and the test pump can take suction directly from the refueling water storage tank (3750 ppm boron)..
The quantity of boric acid in storage from the above two sources is sufficient to borate the reactor coolant in order to reach cold shutdown at any time during the core life.
Furthermore, if the letdown capability from the primary coolant system to the chemical and volume control system should be impaired, the pressurizer void space volume is sufficient to accommodate the required injection. This free volume will accommodate sufficient concentrated boric acid solution such that the reactor coolant water can reach a concentration of about 400 ppm above the required level to shut the plant down.
Amendment No. 25, 102
-20a The above system assured that for Specification A, continuous borated water supply is provided to maintain the core sub critical.
In Specification B, redundancy is provided for borated water injection during reactor operations.
The limitation for a maximum of one centrifugal charging pump to be operable with an RCS pressure < 400 psig with pressurizer water level greater than 50%, provides assur ance that a mass addition pressure transient can be relieved by operation of the overpressurization mitigating system assuming a single failure of one PORV and no operator action for 10 minutes.
Tagged, as it applies to the 'inoperable charging pump, means tagged in accordance with current Southern California Edison procedures for tagging of equipment which must not be operated.
Reference:
(1) Final Engineering Report and Safety Analysis, Paragraph 3.6.
Amendment No. 2%, 102
3.3.2 SHUTDOWN STATUS Applicability: Applies to piping connections between the feedwater condensate and safety injection systems and the reactor coolant system.
Objective:
To preclude injection of feedwater condensate into the reactor coolant system when the reactor is shut down and to preclude the potential for overpressurization when water solid.
Specification: A. When reactor fuel assemblies are in the vessel and the reactor coolant pressure is less than 500 psig, two "posi tive barriers" shall be provided between the feedwater condensate systems and the piping connections to the reac tor coolant system. Additionally, when the reactor coolant system is water solid at less than 500 psig, two positive barriers shall be provided between the safety injection system and piping connections to the reactor coolant system.
A positive barrier is defined as follows:
(1) Motor Operated Valves When closed and tagged with safety switches open, except that power may be restored during no-flow tests of the safety injection system (Specification 4.2).
For MOV 850C, the three (3) pole double throw switch must also be connected to the inverter with the inverter output breaker open.
(2) Pneumatic/Hydraulic Operated Valves When closed and the condition tagged with the respec tive hydraulic block valve closed except that they may be opened during no-flow tests of the safety injection system (Specification 4.2).
(3) Manually Operated Valves When closed and condition tagged.
(4) Feedwater Pump (Overpressurization Protection Only)
When shutdown with the breaker in the racked out condition.
Amendment No.
5, 102
-24 Basis:
Under normal conditions, system operational interlocks assure that injection of feedwater condensate into the reactor by the Safety Injection System cannot occur. (1) These interlocks include:
- 1. Actuation of the safety injection relay which de-energizes the condensate and heater drain pumps and closes the flow path for condensate, thereby preventing injection of feedwater into the coolant system.
- 2.
Interlocks between the condensate isolation valves at the feedwater pump suction and the safety injection header isolation valves at the pump discharge which prevent the opening of the one valve unless the other is closed.
Below 500 psig the Safety Injection System may be removed from service. Below 400 psig, the Feedwater System may be removed from service. During these low pressure shutdown reactor coolant system conditions, the interlocks may be overridden for maintenance and/or test of components of these systems. However, it is still necessary to prevent intrusion of feedwater condensate or safety injection water into the reactor coolant system. Injection of feedwater has the potential to dilute the system and create a potential for a reactivity excursion. Injection of either safety injection water or feedwater, especially during water solid operations, creates the potential for pressurizing above limits established by 10 CFR 50 Appendix G and as reflected in Technical Specification 3.1.3.
The "two positive barriers" required by this specification provide protection of the Reactor Coolant System against boron dilution and overpressurization when in the low pressure and low temperature conditions. Two positive barriers are provided in each potential path between the Feedwater Condensate System, Safety Injection System and the RCS. During periods of no-flow testing, an exception is provided on two of the positive barriers to allow the components involved in the test to perform their test functions while the remaining positive barriers (nos. 3 and 4) remain in effect.
Tagged, as used above, means tagged in accordance with current Southern California Edison Company procedures for tagging of equipment which must not be operated.
Reference:
(1) Final Engineering Report and Safety Analysis, Paragraph 5.1.
Amendment No. 102
-39gg 3.20 OVERPRESSURE PROTECTION SYSTEMS Applicability:
Applies to operability of the overpressurization protection systems.
Objective To preclude the potential for exceeding 10CFR50, Appendix G, in the event of a pressure transient while water-solid.
Specification:
A. When the RCS pressure is < 400 psig* and pressurizer water level is greater than 50%, at least one of the following overpressure protection systems shall be operable:
(1) Two power operated relief valves (PORVs) with a lift setting of < 500 psig,** or (2) A reactor coolant system vent(s) of > 1.75 square inches.
B. With one PORV inoperable when required in accordance with Specification A above, either restore the inoperable PORV to operable status within seven days or depressurize and vent the RCS through a 1.75 square inch vent(s) within the next eight hours; maintain the RCS in a vented and tagged condition until both PORVs have been restored to operable status.
C. With both PORVs inoperable when required in accordance with Specification A above, depressurize and vent the RCS through at least a 1.75 square inch vent(s) within eight hours; maintain the RCS in a vented and tagged condition until both PORVs have been restored to operable status.
- The placing in service of the OMS at < 400 psig is intended to assure that protection is provided whenever temperature is below 3600 F. The alarm to arm the OMS being keyed to pressure assures that inadvertent opening of the PORV's does not occur due to placing the OMS into service with RCS pressure above the 500 psig initiation setpoint.
- The 500 psig setpoint is based on the current heatup and cooldown curves for 16 EFPY. The setpoint requires reevaluation for accept ability any time the curves are changed.
Amendment No. 102
-39hh D. In the event either the PORVs or the RCS vent(s) are used to mitigate a potential RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances indicating transient, the effect of the PORVs or vent(s) on the transient and any corrective action necessary to prevent recurrence.
Basis:
The operability of two PORVs or an RCS vent opening of greater than 1.75 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10CFR Part 50 when the initial RCS pressure is ( 400 psig and the pressurizer water level is greater than 50%, assuming a single failure of one PORV and no operator action for 10 minutes. Either PORV has adequate relieving capability to protect the RCS from overpressurization due to a design basis transient as described in submittal to the NRC dated October 12, 1977.
Tagged as it refers to the RCS vent, means tagged in accordance with current Southern California Edison procedures for tagging of equipment which must not be operated.
Amendment No. 102
-60dd 4.20 OVERPRESSURE PROTECTION SYSTEMS Applicability:
Applies to operability of the overpressurization protection systems.
Objective:
To verify that the overpressure protection systems will respond promptly and properly if required.
Specification:
A.
Each power operated relief valve (PORV) shall be demonstrated operable by:
(1) Adjusting the pressure control bistable setpoint such that the PORVs are actuated and the annunciators alarm within 31 days prior to returning to a water-solid condition following a cold shutdown with the RCS depressurized.
(2) Performance of a channel test within 31 days prior to enabling the low pressure overpressure mitigation setting of the pressurizer PORV's on cooldown.
(3) Performance of a channel calibration on the PORV actuation channel at least once per 18 months.
(4) Verifying that position indications on the PORV isolation valves indicate that the valves are open at least once per week when the PORVs are being used for overpressure protection.
Basis:
The surveillance requirement to verify operability of the PORVs provides assurance that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10CFR Part 50 when the initial RCS pressure is < 400 psig. Either PORV has adequate relieving capability to protect the RCS from overpressurization due to a design basis transient as discussed in Reference 1.
- 1. Letter to A. Schwencer from K. Baskin dated October 12, 1977.
Amendment No. 102