ML20150C792

Proposed Tech Specs Re Reactor Depressurization Sys Isolation Valves

ML20150C792
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	07/05/1988
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML20150C786	List: ML20150C783 ML20150C792 ML20153B588
References
NUDOCS 8807130147
Download: ML20150C792 (14)

Text

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ATTACHMENT 1 Consumers Power Company Big Rock Point Plant Docket 50-155 PROPOSED TECHNICAL SPECIFICATION PAGE CHANGES July 5, 1988 P

R khyg O7 880705 5

10 Pages M10688-0044-NLO4

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t CONTjt"Sjf (Contd) .

.i Page No ,

8.0 Research and Development Program (Deleted) . . . . .... 79 -

91 9.0 Inservice Inspection and Testing . . . . .. . . . .... 92 - 102 9.1 Applicability . . ... .............. 92 9.2 Objective . . . . . . . .-. . . . . . .. . . . . . .- 92 9.3 Specifications. . . . .. . .............. 92 9.4 Basis . . . . . . .

. .. .............. 92 10.0 (Section 6.0) Administrative Controls . . . . . . . .... 103 - 126 6.1 Responsibility . . ...... . . . . . . . .... 103 6.2 Organization . . . ....... . . . . . . .... 103 - 106 6.3 Plant Staff Qualifications . . . . . . . . . .... 107 6.4 ' Training . . . . . .... . . . . . . . . . .... 167 6.5 Review and Audit . ... .............. 107 - 113 6.6 Deleted . . . . . . ............. .... 113 6.7 Saf ety Limit Violation ............... 113 6.8 Procedures . . . . ... .............. 113 - 114 6.9 Reporting Requirements ............... 114 - 120 6.10 Record Retention . ...... . . .. . . . .... 120 - 122 6.11 Radiation Protection Program . . . . . . . . .... 122 6.12 High Radiation Area ..... . . . . . . . .... 123 - 123a 6.13 Environmental Qualification . . . . .. . . . .... 124 6.14 Process Control Program . .............. 125 6.15 offsite Dose calculation Manual (ODCM) . . . .... 125 ,

11.0 (Section 3.1.4/4.1.4) Emergency Core Cooling System ._. .. 130 - 134 (Section 3.1.5/4.1.5) Reactor Depressurization System . .. 135 - 142 (Section 3.3.4/4.3.4) Containment Spray System . . .... 143 - 145 (Section 3.5.3/4.5.3) Emergency Power Sources . . . .... 146 - 149 12.0 Fire Protection Program . .. .............. 150 - 157 (Section 3.3.3.8/4.3.3.8) Fire Detection Instrumentation . 150 - 151 (Section 3.7.11.1/4.7.11.1) Fire Suppression Water System . 152 - 153 (Section 3.7.11.2/4.7.11.2) Fire Spray / Sprinkler Systems . 154 (Section 3.7.11.5/4.7.11.5) Fire Hose Stations . . .... 155 (Section 3.7.12/4.7.12) Penetration Fire Barriers . .... 156 - 157 13.0 Radiological Effluent Technical Specifications . . .... 158 - 186 13.1 Radiological Effluent Releases . . . . . . . .... 158 - 173 13.2 Radiological Environmental Monitoring . . . . .... 174 - 186 11 Amendment No. 63, 77 August 26, 1985 TSB1184-0081A-NLO4

11 3.7 (Contd)

(g) All leakage rates determined by a test pressure less than the applicable design pressure (containment design or design basis accident) shall be corrected using the following formula:

L = L, (P /P,) !

L = maximum a wa e ea age rate, at test pressure.

t L, = % leakage rate, at extrapolated pressure.

Pg = Test pressure (PSIG).

P, = Extrapolated pressure (PSIG).

Acceptance criteria on allowable leakage for the ILRT is .75 Lg .

(h) The RDS containment penetration assemblies seal pressure shall be examined at six-month intervals.

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l Proposed BRP TS SECTION 3

4 4 s EMERGENCY CORE COOLING SYSTEMS I REACTOR DEPRESSURIZATION SYSTEM LIMITING CONDITION FOR OPERATION M*

11.3.1.5 The Reactor Depressurization System (RDS) shall be OPERABLE with:

a. Four RDS valve trains including pilot valves.

b. Four input and output channels including instrumentation given in Table 11.3.1.5.

c. Four Uninterruptible Power Supplies (UPS) as described in Specification 11.3.5.3.

d. All mechanical snubbers in service.

APPLICABILITY: POWER OPERATION ACTION:

a. Should one RDS valve train, one input channel, one output channel, or UPS Power Supply become inoperable in the closed position, the reactor may remain in POWER OPERATION for a period not to exceed seven (7) days, providing the actuating circuitry for the remaining -

channels is demonstrated to be OPERABLE within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and at least once each 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> until the components are restored to OPERABLE status. If these components are not returned to OPERABLE status within seven (7) days, a normal orderly shut down shall be initiated within one (1) hour and the reactor shall be SHUTDOWN as described in Section 1.2.5(a) within twelve (12) hours and SHUTDOWN as described in Section 1.2.5(a) and (b) within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

b. Should two or more RDS valve trains including input channel, output channel or UPS Power Supply become inoperable the plant shall be brought to the SHUTDOWN condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to the COLD SHUTD0"N condition within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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c. Should one isolation valve or depressurizing valve become inoperable in the open position the plant shall be brought to the COLD SHUTDOWN condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, s

135 Proposed BRP TS SEC 11

A EMERGENCY CORE COOLING SYSTEMS REACTOR DEPRESSURIZATION SYSTEM ,

ACTION: (Continued) i 3

d. Should a pilot valve be isolated from service and removed, it shall be t functionally tested prior to installation and returned to service.

e. Should a main valve be repaired to correct seat leakage, a partial stroke test shall be performed to establish operability.

f. If the RDS is declared inoperable because of a snubber defect and is not returned to an OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the plant shall be brought in a normal and orderly manner to a COLD SHUTDOWN condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and be maintained in COLD SHUTDOWN until RDS can be declared OPERABLE. If the plant is already in a COLD SHUTDOWN condition, it shall not be started up until all snubbers are OPERABLE.

SURVEILLANCE REQUIREMENTS 11.4.1.5 The Reactor Depressurization System shall be verified to be OPERABLE:

a. At least once per month the instrumentation and system logic shall be FUNCTIONALLY TESTED as indicated in Table 11.3.1.5.

b. At least once per quarter the isolation valves shall be full stroke exercised.

c. At each refueling outage, not to exceed 18 months;

1. The four depressurization valves shall be full stroke exercised.

2. The instrumentation and system logic shall be CALIBRATED, CHECKED, and FUNCTIONALLY TESTED as indicated in Table 11.3.1.5.

3. A visual inspection of 10% (2) of the thirteen mechanical snubbers on the RDS shall be performed. Visual inspections shall be used to verify that there are no virible indications of damage or impaired operability to the snubbers or their attachments.

136 Proposed BRP TS SEC 11

EMERGENCY CORE COOLING SYSTEMS REACTOR DEPRESSURIZATION SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

4. A FUNCTIONAL TEST of 10% (2) of the thirteen mechanical snubbers on the RDS shall be performed. FUNCTIONAL TESTS shall be used to verify that the force that initiates free movement of .he snubber rod, in tension and compression, is less than the vendor specified caximum drag force. Activation restraining action shall be achieved within the vendor specified range of velocity or acceleration in both tension and compression.

137 Proposed BRP TS SEC 11

Table 11.3.1.5 -

.' H Instrumentation Thet Initiates RDS Oparation ,

Limiting Condition for Operation Surveillance Requirement .!

Minimum Protective OPERABLE Instrument Instrument Channel Parameter Channels Set Point Trip Test Calibration Trip Low Steam Drum 3 Above or Equal Monthly Each Major .-

Level to 17" Below Refueling Center Line (Tolerance Limit -5")(1)

Fire Pump (s) 3 105 Psig i Monthly Each Major -

5 Psig Refueling Low Reactor 3 2: 2'9" Above Top Monthly Each Major -

Water Level of Active Fuel Refueling (Tolerance Limit

-1") (1) 120-Second 3 $ 120 Seconds Fol- Monthly 'Each Major -

Time Delay lowJng Low Steam Refueling Drum Level Signal Input Channels 3 -

Monthly - -

A Through D Output Channels 3 - - -

Monthly I Thrt, ugh IV

• Fire Pump Start 1 -

Monthly -

Monthly

• Reference Specifications 11.3.1.4 and 11.4.1.4 for Bases.

(1)Lesel instrument set points shall be as specified. Level instrument calibration shall be based on normal operating temperature (582*F) and pressure (1350 psia) and instrument reference leg temperature of 250*F or lower es sacasured to maintain an actual trip level greater than that assumed in accident analyses (2'8" above top of active fuel for reactor v-ssel level trips, and 25" below steam drum center-line for RDS actuation).

138 Proposed BR15 TS SEC 11

11.3.1.5 Bases:

The RDS provides for both manual and automatic depressurization of the primary system to allow injection of the core spray following a small-to-intermediate size breek in the primary system. This will allow core cooling with the objective of preventing excessive fuel clad temperatures. The design of this system is based'on the specified initiation set points described in Table 11.3.1.5. Transient analyses reported in Section 6 of the RDS Description, Operation and Performance Analysis submitted August 15, 1974 to the Directorate of Licensing USAEC, demonstrated that these conditions result in adequate safety margins for both the fuel and the system pressure. Performance analysis of the RDS is considered only with respect to its depressurizing effect in conjunction with core spray. Therefore, ro credit is taken for steam cooling of the core which provides further conservatism to the emergency core cooling system.

These specifications ensure the operability of the RDS under all conditions for which the automatic or manual depressurization of the system !s an essential response to the transient described above.

One RDS valve can remain out of service in the closed position for seven days because of redundancy, provided the actuating circuitry for the remaining RDS valves is tested within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and at least once each 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> until the valve ic restored to operable status. When conditions for the actuation on the depressurizing system are reached, all the valves in the four blowdown paths are opened. Each blowdown path is designed to pass 144 lb/second of steam at 1350 psig which is a third of the required total flow rate. Therefore, failure of one flow path to open upon actuation does not preclude achieving the required rate of depressurization.

In addition to reactor protection instrumentation, which initiates a reactor scram, protective instrumentation has been provided for the RDS which initiates action to mitigate the consequences of the Loss of Coolant Accident. This set of specifications provides the limiting conditions >f operation for the RDS.

The objeutives of the specifications are (1) to assure the effectiveness of the protective instrumentation when required even during periods when portions of such systems are out of service for maintenance and (11) to prescribe the trip settings required to assure adequate performance. To conduct the required input channel maintenance or functional tests and calibrations, any one channel may be bypassed. If the input channel is not bypassed when functional tests and calibrations are performed, actual trip signals supersede tect and calibration conditions.

The minimum functional testing frequency used in this specification is based on a frequency that has proven acceptable and conforms to that of the existing reactor protection system.

i 139 Proposed BRP TS SEC 11 l

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11.3.4.5 Bases: (Contd)

Four plant variables are monitored and used as inputs to the actuation system.

These are (1) steam drum water level, (2) reactor water level, (3) motor-driven fire pump discharge pressure and (4) diesel engine driven fire pump discharge

! pressure. These variables are jointly processed by the four independent actuation system input channels which are physically and electrically isolated from one another. A failure in one channel cannot propagate into another l channel. Each of the four plant variables is monitored by four separate

! sensors. One sensor in each of the four variables is associated with each of the four input channels. The actuation of the RDS is enabled when two of the four input channels are in a tripped state.

The input channel is in a tripped state upon coincidence and subsequent processing of the following inputs: (1) Low steam drum level (delayed for two l minutes), (2) high fire pump discharge pressure (either diesel- or motor-driven) and (3) low reactor water level. A low steam drum level signal is generated when the steam drum level sensor associated with the input channel indicates a level of 25" below steam drum center line.

The low steam drum level signal initiates a two-minute delay which allowe a contsinment evacuation interval prior to system blowdown and also permits the incorporation of operator input to the system initiation logic specified in the design basis (Reference Section 3.3.D of the August 15, 1974 RDS Description.

Operation and Performance Analysis) . For the latter, the operator is provided with manual timer reset capability for each of the four input channels at the control panel. The icv steam drum level signal is also used to generate a fire pump start signal. Verification of a fire pump start and thus verification thct a source of core spray water is available at the core spray valves is obtained when the pressure switch associated with the input channel at either fire pump discharge has tripped, corresponding to a pressure equal to or exceeding 100 psig. This variable is used as an enabling input to the actuation system to prevent depressurizing the reactor coolant system when the source of coolpnt required to cool the core is not available A low reactor water level signal is generated when the input channel reactor water level sensor indicates a level 2 2'8" above the top of active fuel. Low reactor water level is confirmation of the LOCA and with the other two inputs present (time delayed low drum level and core spray water availability) causes the automatic trip of the input channel. These trip level settings were chosen to be low enough to prevent spurious actuation but high enough to initiate RDS operation so that post-accident cooling can be accomplished.

l l Upon failure of an uninterruptible power supply (UPS) or a channel power supply, the affected charnel fault condition is alarmed as "channel 'X' unavailable." Power failures associated with input channels cause the coincidence trip conditions for the input channels to change from 2-out-of-4 to 2-out-of-3. The output channel actuation coincidence reverts to 3-of-3 upon failure of an output channel power supply.

Input channel bypass capability is provided to permit bypassing any one input channel at a time. The bypass feate2e is used to bypass a channel when :he ,

channel has failed to the "trip" state and/or when channel maintenance is l 140 Proposed BRP TS SEC 11

11.3.1.5 Bases (Contd) required. Bypassing of an input channel in the "trip" state or'for maintenance causes the coincidence trip condition of the input channel to be changed from 1-out-of-3 or 2-out-of-4, respectively, to 2-out-of-3. The input channel bypassed condition is alarmad as "channel 'X' unavailable" and "bypassed." j Should an output channel require maintenance or should a single fault cause an outpat channel subchannel trip (two independent subchannels operate in 2 of 2 '

coincidence), the output channel actuation capability can be disabled.by removing the ar.sociated 125 V DC supply. The 125 V DC supply to an output channel is disabled via a circuit breaker in its respective UPS. The disabling of an output channel is alarmed as "channel 'X' unavailable."

Since 3-out-of-4 output channels are required to assure design requirements are met (one output channel operates one depressurizing valve and one isolation valve), the failure of one output channel will not preclude achieving the required rate of depressurization. This redundancy also enables maintenance to be performed on one opput channel while the plant is in operation.

Once the RDS actuation system output channels are enabled (at least two input channels are in a tripped state or a manual trip is initiated) and tripped, they remain in that condition until they are manually reset. This reset can be accomplished only after the initiating signals (ie, input ci.annel trips or manual trip) have been restored to levels at which RDS operation is not required.

Separate independent one-hour sources of electrical power are provided, through four divisions, to accomplish the detection of the LOCA and the completion of the depressurization. Each of the divisions (1, 2, 3 and 4) is supplied with electrical power from one of four independent uninterruptible power supplies (UPS) consisting of a battery charger, a battery and an inverter.

Each UPS has output of 120 V a-c, 60 Hz and 125 V d-c. Divisions 1 and 2 normally receive power from the existing 480 V a-c Bus IA. Divisions 3 and 4 are supplied by 480 V a-c Bus 2A. Normal station power to Buses IA and 2A can be provided by one of three sources: (1) The station turbine generator, (2) the 138 kV transmission line or (3) the 46 kV tranemission line. Should none of these sources be available, provision is included for supplying input power from the 480 V a-c Bus 2B which is tied to the emergency diesel. If all 480 V a-c power is lost, the UPS is capable of sustaining its ouput for one hour.

Since only 3 out of 4 blowdown paths are required to assure adequate depressurization, the single system failure of one UPS division will not preclude achieving the required rate of depressurization. This redundancy also enables maintenance to be performed on the UPS while the plant ic in operation.

l 141 Proposed BRP.TS SEC 11

11.3.i.5 Bases: (Contd)

Technical Specifications in this part, also include action statements and surveillance requirements to comply with the required tests of Section XI of the ASME Code (Subsection IWV) as specified in Section 9.0 of these Technical Specifications. Testing of the depressurization valve is only practical on a Refueling Octage basis to minimize pilot valve degradation.

The modified depressurization valve design permits the isolation of a pilot valve assembly for maintenance during POWER OPERATION. Following removu: and repair, the pilot valve assembly is functionally tested and leak checked prior to reinstallation to ensure operability. Following installation, actuation circuit continuity checks are performed. Should a main valve be repaired to correct seat leakage, a partial stroka test with a compressed gas is used following reinstallation to e*tablish operability. These tests fulfill the post maintenance testing requirements of Section XI of the ASME Code (Subsection IWV-3200).

Four new containment penetration assemblies are used in transmitting electrical power, control and instrumentation signals between equipment located inside the containment building and facilities located external to the containment building. These electrical penetrations are welded into spare containment penetration sleeves. The penetration assemblies are designed in accordance with IEEE 317 and are seismically and environmentally qualified to the RDS design conditions.

The pressure retaining portion of the assemblies is designei and fabricated to the requirements of Subsection NE, Class MC vessels, of Section III of the ASME Code. The penetration assemblies include a single aperture seal and a double electrical conductor seal and are designed to operate with the internal cavity pressurized with nitrogen at approximately 27 psig. The pressurized c ity limits the intrusion of air which may degrade the life expectancy of the seals associated with containment isolation. The relatively maintenance-free seal assemblies dictate a minimum inspection frequency of twice annually, as specified in Section 3.7(h) of the Technical Specifications.

All snubbers are required OPERABLE to ensure that the structural integrity of the RDS is maintained during and following a seismic or other event initiating dynamic loads. There are thirteen mechanical snubbers on the RDS that are subject to required visual and functional testing. The snubbers have a 40-year service life and no maintenance requirerrnts as the snubbers do not contain any fluid, seals, etc.

142 Proposed BRP TS SEC 11

ATTACHMENT 2 Consumers Power Company Big Rock Point Plant Docket 50-155 RDS SYSTEM RELIEF REQUESTS July 5, 1988 2 Pages MIO688-0044-NLO4

1 RELIEF REQUEST System: Reactor Depressurization System'(Print A-203)

Valve: SV-4984, SV-4985, SV-4986 and SV-4987 Category: B ASME Class: 'l Function: Depressurizing Valve Test Requirement: Exercise valve for operability every three months.

Basis for Relief: The valve cannot be full stroke exercised unless continuous primary system pressure is applied. With the isolation valve _ closed,' pressure in the spool piece between the isolation valve and the depressurizing valve is relieved before the depressurizing valve reaches the full stroked position.. With the isolation valve open a continuous steam blowdown of the primary system to the containment would result. Partial ctroke exercising of the valve has been demonstrated to be a significant contributor to chronic pilot valve seat leakage.

. Alternate Testing: Full stroke exercise the depressurizing valve each refueling outage, but not to exceed eighteen months.

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l HIO688-0044-NLO4

v 2

RELIEF REQUEST System: Reactor Depressurization System (Print A-203)

Valve: SV-4984, SV-4985, SV-4986 and SV-4987 Category: B ASME Class: 1 Function: Depressurizing Valve Test Requirement: Demonstrate that the performance parameters which could be affected by replacement, repair, or. maintenance are within acceptable limits.

Basis for Relief: Previously accomplished quarterly partial stroke exercising of the valve has been demonstrated to be a significant contributor to chronic pilot valve seat leakage. To reduce pilot valve leakage problems the depressurizing valve has been modified to have isolable and removable pilot valves. This modification allows replacement of a leaking pilot valve assembly with the plant in power operation which eliminates the need to cycle the plant numerous times from full power to cold shutdown and back to full power and reduces the probability of the occurrence of an accident or equipment malfunction. It also allows leakage to'be terminated quickly. With removable pilots the main depreso cizing valve internals are not disturbed during the proce.s of replacing a pilot valve assembly. Therefore, it is not necessary to stroke the pilot and main valve as a unit to demonstrate post maintenance operability.

Alternate Testing: Following maintenance involving only pilot valve assembly removal and reinsta11ation, measure pilot valve lift voltage and disc lift, full stroke exercise and leak test pilot valve prior to installation on depressurizing valve.

After installation check pilot valve electrical continuity, verify pilot valve isolation valves open and check for pilot valve inlet bolting flange leakage using system operating pressure.

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HIO688-0044-NLO4

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