ML20132C627
| ML20132C627 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 09/17/1985 |
| From: | DAIRYLAND POWER COOPERATIVE |
| To: | |
| Shared Package | |
| ML20132C616 | List: |
| References | |
| NUDOCS 8509270102 | |
| Download: ML20132C627 (4) | |
Text
i 4.1 GENERAL 4.1.3 During periods when the reactor is in Condition 3, 4 or 5, either Channel 1 or 2 of the Nuclear Instrumentation System shall be in operation and shall be monitored by the operator.
4.1.4 Whenever the reactor contains one or more fuel elements, any operations from points outside the control room of equipment which may affect the reactor shall be conducted under the direction, or with the knowledge, of the control i
room operator.
4.1.5 If the plant is operational during a tornado warning, the Shift
, Supervisor on duty shall keep informed of the actual tornado activity which may approach the plant. In the event that reports indicate an imminent tornado strike at or near the LACBWR plant, the Shift Supervisor shall reduce
- reactor power to a level which permits prompt reduction of power generation to station load. However, the Shift Supervisor shall be instructed to disconrinue plant operation if, in his judgment, this action is required to ensure plant safety.
4.2 OPERATING LIMITS 4.2.1 Reactor Building
)
4.2.1.1 CONTAlt! MENT INTEGRITY shall be maintained in Conditions 1, 2, 3 and during:
i (a) CORE ALTERATIONS, (b) handling of irradiated fuel, or (c) there is fuel in the reactor and any control rod is withdrawn.
4.2.1.2 Gasketed closures and ventilation system closures which have been c subjected to maintenance, repair or other operations which might affect their performance shall, before any subsequent operation for which containment integrity is required, be tested for leak tightness using the soap-bubble technique (or other method of equivalent sensitivity). This test shall be performed using a pressure differential no less than 0.5 psi and the results shall be used as a guide in evaluating leakage.
Reference 4 also affects this page 8509270102 850917 Amendment No. 34 PDR AEK}CK 050(N)409 <
p PENT WPl.6.8
4.2.1.3 The number of containment vessel electrical penetrations may be varied, as may the containment vessel piping penetrations, provided the new penetrations are equivalent in design to the existing penetrations. New penetrations a,re defined as conduit or piping requiring attachment to the containment vessel shell. Af ter any such penetration changes, an integrated reactor building leak test shall be performed at approximately 52 psig. This test must demonstrate that the containment vessel meets the leak rate specified in Sec. 5.2.1.
4.2.1.4 Existing containment vessel penetrations may be removed from or placed in service, provided the containment vessel is not affected and the closures are equivalent in strength and tightness to those previously installed. After any such change the penetration, exclusive of the containment shell connection, shall be tested for leak tightness at a pressure no less than 52 psig using the soap-bubble technique (or other method of equivalent sensitivity) or by determining the rate of pressure loss of a test chamber. The penetration leakage rate shall not exceed 1.0 percent of the containment vessel leak rate Lp specified in Sec. 5.2.1.
4.2.1.5 The reactor building vacuum breakers shall be set to relieve at a differential (external-over-internal) pressure not exceeding 0.5 psi.
4.2.1.6 The reactor building ventilation system isolation dampers shall close upon loss of control air or loss of electrical power supply, and they shall be automatically closed by abnormal conditions as specified in Table 1.
4.2.1.7 The main steam line, the reactor building vent header, and the decay heat system blowdown line shall be isolated automatically by abnormal conditions as specified in Table 1.
4.2.1.8 The containment building steel shell temperature shall be greater than 0*F during reactor operation or during any integrated leak rate test performed with a pressure exceeding 10.4 psig.
4.2.1.9 The containment building shall be isolated whenever the spent fuel storage well contains irradiated fuel which has decayed less than 43* days after exposure in a critical reactor and a shipping cask for irradiated fuel is being moved by the crane on the 701 foot level or located within one cask length of the top of the spent fuel storage well or is within the spent fuel storage well. During cask movement near or at the FESW the water level in the FESW must be at least 16 ft. above the top of the fuel storage rack (no more than 7 feet below the top of the FESW).
- 43 days for off loading less than one half of the core, i.e. less than 36 fuel elements. 51 days for off loading more than 36 fuel elements.
WPl.6.8 _ _ _ _ _ _ - _ _
REACTOR BUILDING CONTAINMENT VENTIMTION DAMPERS LIMITING CONDITION FOR OPERATION 4.2.1.10 The Containment ventilation inlet and outlet dampers shall be OPERABLE with isolation times of less than or equal to 10 seconds.
APPLICABILITY: Whenever CORIAINMENT INTEGRITY (Specification 4.2.1.1) is required.
ACTION:
With one or more of the above ventilation damper (s) inoperable:
A. Operation may continue provided that at least one damper in each affected penetration is maintained OPERABLE and either:
- 1. The inoperable damper (s) is restored to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or
- 2. Each affected penetration is isolated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one automatic damper secured in the isolation position, or a blank flange; OR B. Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> unless the affected penetration is isolated.
C. The provisions of Specification 3.0.4 are not applicable if the affected penetration is isolated.
SURVEILLANCE REQUIREMENTS 5.2.1.10.1 The ventilation dampers shall be demonstrated OPERABLE prior to returning the damper to service after maintenance, repair or replacement work is performed on the damper or its associated actuator, control, or power circuit by performance of a cycling test, and verification of isolation time.
l l 5.2.1.10.2 The isolation time of each above damper shall be determined to be within its limit when tested pursuant to Specification 5.2.2.
j 5.2.1.10.3 The seat rings of the ventilation inlee and outlet dampers shall j be replaced at least once per 5 years.
BASES The Nuclear Regulatory Coumission requested that similar Technical Specifications per Generic Item B-24 and NUREG 0737 Item II.E.4.2 be submitted to help assure operabilty of containment ventilation dampers. DPC had previously committed to replace the containment ventilation inlet and outlet dampers' resilient sealing material at least once each 5 years until such time as additional in-situ data can be accumulated to justify a longer interval.
If in-situ data is accumulated which supports a longer seal replacement intervals, a change to Specification 5.2.1.10.3 may be requested.
Specification 3.0.4 is not applicable if the affected penetration is isolated since the safety function of the dampers is to close.
WPl.6.8 - 29a -
4.2.2 Rector Vessel, Coolant, and Auxiliary Syster7 4.2.2.1 Additional penetrations to the systems containing reactor coolant shall be designed, manufactured, and tested according to the provisions of the ASME Boiler and Pressure Vessel Code and the ASA Code for Pressure Piping applicable as of June 1962. These additional penetrations shall be limited to ,
instrument connections and piping connections, the latter being no larger than i 1-in. inside diameter.
4.2.2.2 The reactor coolant shall be light water and shall conform to the following requirements.
CONDITION 1 Normal Limit Maximum Limit Chloride concentration .2 ppm .5 ppm pH 5.3 - 8.6 NA Conductivity 3 pmho/cm 10 pmho/cm The time above 3 umho/cm at 70*F - 80*F and .2 ppm chloride should not exceed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> per incident, nor 2 weeks per year. When the single incident normal limit is exceeded an orderly shutdown shall be initiated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> unless returned to within the limits. When the maximum conductivity or chloride limits are exceeded an orderly shutdown should be initiatied immediately. If the pH is outside the limits for a period of greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> an orderly shutdown shall be initiated.
i CONDITION 2 & 3 Normal Limit Maximum Limit
! Chloride concentration .1 ppm NA pH 5.3 - 8.6 NA Conductivity 5 pmho/cm NA The time above 5 phmo/cm at 70*F to 80*F and .lppm chloride concentration is restricted to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> for any single occurrence during Condition 2.
When this time limit in Condition 2 is exceeded the reactor shall be brought to the hot shutdown condition (Condition 3) until the limits are restored. If the limits can not be restored in an additional 7 days, the reactor shall be taken to the cold shutdown condition (Condition 4).
CONDITION 4 & 5 Normal Limit Chloride concentration .5 ppm ph 5.3 - 8.6 Conductivity 10 phmo/cm The primary system chemistry parameters defined in this section shall be determined at least once every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in Condition 1, 2, and 3 and at least once every 7 days in Condition 4 and 5.
4.2.2.3 Deleted WPl.6.8 - 29b -
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