ML20205S124

Amend 93 to License DPR-6,modifying Tech Specs by Replacing Requirements to partial-stroke Test Reactor Depressurization Sys Depressurizing Valves Quarterly W/Requirement to full- Stroke Test for Depressurizing Valves Each Outage

ML20205S124
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	11/04/1988
From:	Quay T Office of Nuclear Reactor Regulation
To:
Shared Package
ML20205S119	List: ML20205S117 ML20205S124 ML20205S127 ML20206C184
References
NUDOCS 8811100257
Download: ML20205S124 (11)

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CONSUMERS POWER C0fdANY BIG ROCK POINT PLANT DOCKET NO. 50-155 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 93 License No. OPR-6

1. The Nuclear Regulatory Comission (the Comission) has found that:

A. The appl.ication for amendment by Consumers Power Company (the licensee) dated July 5, 1988, as modified October 10, 1988, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulatic,1s set forth in 10 CFR Chapter I;

8. The facility will operate in conformity with the application, the provisions of the Act, and the rul?s and regulations of the Comission; 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 Comission's regulations; D. The issuance of th9 amendment will not be inimical to the common

. defense and sr;urity or to the health and safety of the public; and 1

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.

8811100257 SS1104 PDR ADOCK 05000155 P PDC

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2. Accordingly, the license is amended by changes to the Technical i Specifications as indicated in the attachment to this Itcense amendment and Paragraph 2.C.(2) of Facility operating License No. OPR-6 is hereby amended to read as follows:

Technical Specifications The Technical Specifications contained in Appendix A, as t revised through Amendment No. 93 , are hereby incorporated  ;

in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment became effective October 14, 1988.  !

. FOR THE NUCLEAR REGULATORY C0tHISSION j

C% ,R heodc r Qua),

f-)/usfc-tin.g Direct'dr '

Proje t Di1 ecto ate Iljyi .

Div ion o" R ctor Projects - III, IV, V Special rojects

Attachment:

Changes to the Technical Specifications Date of Issuance: November 4, 1988 1

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ATTACHMENT TO LICENSE AMENDMENT NO.93 FACILITY OPERATING LICENSE NO. OPR-6 AOCKET NO. 50-155 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the trea of change.

REMOVE INSERT 3-8 3-8 11-7 11-7 11-8 11-8 11-9 11-9 11-10 11-10 11-11 11-11 11-12 11-12 11-13 11-13 11-13a -

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,y __ __ _ _ _ _ _ _ - _ _ _ _____ _ __

(f) If tra consecutive integrated leak rate tests fail to treet the specifications contained in this section, then an ILRT shall be performed at each plant shutdoun for refueling or approximately 18 months, whichever occurs first, u.itil two .

consecutive ILRTs meet the acceptance criterie. After the '

above special retest requirement is satisfied, then the testing schedule outlined in 3.7.E may be resumed from the date of the last special test (i.e., 3-1/3 years after l completion of the second consecutive satisfactory special test). i (g)

All leakane rites the applic3bic desinndetemined by(a test prest.ure contair.. ent desion pressure Icss, than or design basis ucident) shall be ccrrected using the follo, ting formln:

Lt

• Le (Pg/Pe)l/2 Lt = % maxir.cn allouable leakage rate, at test pressure.

Le = % leakage rate, et extrapolated pressure. . .

Pt=Testpressure(PSIG).

Pe = Extrapolated pressura (PSIG).

Acceptance criteria on allowable leakage for th: ILitT is

.75 Lt*

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

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3-8 Amendm:;t 16. f), 93

DERGENCY CORE COOLING SYSTEMS REACTOR DEPRESSURIZATION SYSTEM LIMITING CONDITION FOR OPERATION 11.3.1.5 The Reactor Depressurization System (KDS) shall be OPERABLE with

a. Pour RDS valve trains including pilot valves.

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

c. Pour Uninterruptible Power fupplies (UPS) as described in Specification 11.3.5.3.

d. All mechanical snubbers in service.

_ATJ ?.ICAnILIT_Y : POVER 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 e period not to exceed seven (7) days, providing the actuating circuitry for the remaining channele 14 demonstrated to be OTERABLE 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 OPERA 3LE status within seven (7) days, a normal orderly shut down shall be initiated within one (1) hour and the reactor shall be SRUTDOWN as described in Section 1.2.5(a) within twelve (12) hours and '

SRUTDOVA as described in Section 1.2.5(a) and (b) within the i folleving 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 SHUTDOVN 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 SHUTDCVN condition within the folleving 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c.

• Should one isolation valve or depressurising valve become inoperable in the open position the plant shall be brought to the COLD SRUTDOWN (1sdition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

11-7 Amendment No. 10, 5f. fl. 18 63h

EMIRGENCY CORE COOLING SYSTEMS REACTOR DEPRESSURIZATION SYSTEM ACTION: (Continued)

d. 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 SEUTDOVN 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 dee.iared OPERABLE. If the plant is already in a COLD SKVTDOWN condition, it shall not be started up until all snubbers are OPERABLE.

SURVEILI)NCE REQUIREMINTS 11.4.1.5 The Reactor Depressurization Systen shall be verified to be OPERABLE:

a. At least once per month the instrumeetation and system logie shall be TUNCTIONALLY TESTED as indicated in Table 11.3.1.5.

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

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

1. The four depressurization valses shall be full streke exercised.

2. The instrumentation and system logic shall be CALIBRATED, CHECKED, and TUNCTIONALLY 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 visible indications of damage or impaired operability to the snubbers or their attachments.

4. A FUNCTIONAL TEST of 101 (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 the snubber rod, in tension and compression, is less than the vendor specified maximum drag force. Activation restraining action shall be achieved within the vendor specified range of velocity or acceleration in both tension and compression.

4 d. Should a pilot valve be isolated from service and removed, the replacevent pilot valve shall be functionally tested prior to installation and return to service.

11-8 Amendment No.10. II .18. 93

_ _ - - . . . - - -= .--- --- -

Tablo 11.3.1.5 Instrumentation That Initiates RDS Operation Limiting Condition for Operation Surveillance Requiremen';

Minimum OPERABLE Protective Instrument Instrument Cher.nel -

Parameter Channels Set Fofnt Trip Test Calibration Trip Low Steam Drum 3 Above or Equal Monthly Level Each Major -

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

1 Fire Pump (s) 3 105 Fais 2 Monthly Each Major -

! 5 Psig

! Refueling Low n'esctor 3 2 2'9" Above Top Water Level Monthly Each Major -

of Active Fuel Refueling (Tolerance Limit t

-1")(1) 120-Secome 3 5 120 Seconds Fol- Monthly Each Major -

Time Delay loving Low Steam Refueling Drum Level Signal

Input Channels 3 -

Monthly - -

i a Through D Ou.put Channels 3 -

I.Through IV Monthly

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• Fire Pump Start 1 -

Monthly -

Monthly 4

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ference Specifications 11.1.1.4 and 11.4.1.4 for Bases.

(1) Level instrument set points shall be as specified. Level instrument calibration shall be based on normal operating temperature (582*F) and pressure (1350 psia) end instrument reference leg temperature of 250*F or louer as measured j

to maintain an actual trip level greater than that assumed in accident analyses (2'8" above top of active fuel for reactor vessel level trips, and 25" below steam drum center line for RDS actuation).

I 11-9 Amendment No. 19. 31, 34, 43, 93

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11.3.1.5 Bases:

The RDS provides for both manual and automatic depressurisation of the primary system to allow injection of the core spray following a small-to-intermediate site break in the primary systes. This will allow core cooling with the objective of preventing excessive fuel elad 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. l Operation and Performance Analysis submitted August 15, 1974 to the Directorate of Licensing USAIC, demonstrated that these conditions result in adequate safety margins for both the fuel and the system pressure. Performance analysis l of the RDS is considered only with respect to its depressurising effect in conjunction with core spray. Therefore, no 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 depressurisation of the system is 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 is restered 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 staan at 1350 psis 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 of operation for the RDS.

The objectives of the specifications are (1) a assure the effectiveness of the protective instrumentation when tequired evet during periods when portions of such systems are out of service for maintenance and (11) to prascribe 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 test 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.

11-10 Amendment No. 1063

11.3.2.5 Basest (Contd)

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

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

sensors. One sensor in each of the four variables is assoc.iated with each of

{ the four input channels. The actuation of the RDS is enabled when two of the

four it.put channels are in a tripped state.

! The input channel is in a tripped state upon coincidence and subsequent l processing of the following inputs: (1) Low steam drum level (delayed for two 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 i generated when the steam drum level sensor associated with the input channel

! indicates a level of 25" below steam drum center line.

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1 The Icw steam drus level signt' initiates a two-minute delay which allows a 1

containment evacuation interv . prior to systen blowdown and also permits the

! incorporation of operator input to the systen initiation logic specified in the j design basis (Reference Section 3.3.D of the August 15, 1974 RDS Description, j Operation and Performance Analysis). For the latter, the operator is provided a

with manual timer reset capability for each of the four input channels at the control panel. The low steam drum level signal is also used to generate a fire pump start signal. Verification of a fire pump start and thus verification that a sourev of core spray water is available at the core spray valves is obtained when the pressurs switch associated with the input channel at either fire pump discharge has tripped, corresponding to a pressure equal to or exceeding 100 psis. This variable is used as an enabling input to the actuation systen to prevent depressurizing the reactor coolant system when the source of coolant required to cool the core is not available. A low reactor watet level signal is generated when the input channel reactor water level sensor indicates a level t 2'8" above the top of active fuel. I.ow 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.

Upon failure of an uninterruptible power supply (UPS) er a channel power supply, the affected channel 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. A power supply failure associated with an output channel results in loss of that channel.

Input channel bypass capability is provided to permit bypassing any one input channel at a time. The bypass feature is used to bypass a channel when the channel has failed to the "trip" state and/or when channel maintenance is 11-11 Amendment No. H ,93

r-e t

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11.3.1.5 Bases: (Contd)

I 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-3e The input channel

. bypassed condition is alarmed as "channel 'X' unavailable" and "bypassed."

Should an output channel require maintenance or should a single favlt cause an

, output channel subchannel trip (two independent subchannels operato in 2 of 2 coincidence), the output channel actuation capability can be disabled by ramoving the associated 125 V DC supply. The 125 V DC supply to an output channel is disabled via a circuit breaker in its respective U?S. 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 e

met (one output channel operates one depressurising valve and one isolation valve), the failure of one output channel vill not preclude achieving the required rate of depressurization. Thie redundancy also enables maintenance to be performed on one cuput channel while the plant is in operation.

Once the RDS actuation system output chtnnels are enabled (at least two input channela are in a tripped state or a manual trip is initiated) and tripped, they remain in that ceaditien until they are n.anually reset. This reset can te accomplished only af ter the initiating signals (ie, input channel trips er manual trip) have been restored to levels at which RDS operation is not required.

Separate independent one-hour sources of electrical power are provided. l 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 uninterruptibic power supplies (UPS) consisting of a battery charger, a battery and an inverter.

Each UPS has output of 120 V a-c. 60 Et and 125 V d-c. Divisions 1 and 2 normally receive power from the existing 480 Y a-c Bus 1A. 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 sourcess (1) The station turbine generator, (2) the 138 kV transmission line o- (3) the 46 kV transmission line. Should none of these sources be available, provision is included for supplying input power from the 480 V a-c Bus 25 which is tied to the energency diesel. If all 480 V a-c power is lost, the U?S is capable of sustaining its oupur. for one hour.

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

11-12 A=endment No.10.93

l'1.3.1;5 Bnses (Contd)

Technical Specifications in this part. also include action statesents and surveillance requiremente 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 Outage basis to mini:1:e pilot valve degradation.

The modified depressuries: ion valve design permits the isolation of a pilot valve assembly for maintenance during POWEF. OPERATION. Following removal and repair, the pilot valve assembly is functionally tested and leak checked prior to reinstallation to ensure operability. Folluving installation, actuation circuit continuity checks are performed. These tests fulfill the post maintenance testing requirements of Section XI of the ASME Code (Subsection IWV-3200) for replacement pilot valves. ,

Tour new containment penetration assemblies are used in transmitting electrical power, control and instrumentation signals between equipment located inside the containment building and f acilities located extgrnal to the containment building. These electrical penetrations are velded into epare containment penetration sleeves. The penetration assemblies are designed in accordance with IEEE 317 and are seismically and environmentally qualified to the FIS design conditions.

The pressure retaining portion of the assemblies is designed and fabricated to the requirements of Subsection NE. Class MC vestals, 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 pressuris.d cavity licits. the intrusien of air which may degrade the life expectancy of the seals associated with containment isolation. The relatively maintenance-free seal assemblies dictate a einimum iaspection 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 iTS is maintained during and folleving a scismic or other event luitiating dynamic loads. There are thirteen sechanical snubbers on the RDS that are subject to required visual and functional testing. The snubbers have a 40-year service life and no maintenance requirements as the snubbers do not centain any flufd seals, etc.

11-13 Amendment No . 10, 93 I

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