ML20150C783

Application for Amend to License DPR-6,changing Tech Specs Re Reactor Depressurization Sys (Rds) Surveillance Testing, Eliminating Quarterly Partial Stroke Exercising of Rds Depressurizing Valves

ML20150C783
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	07/05/1988
From:	Hoffman D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML20150C786	List: ML20150C783 ML20150C792 ML20153B588
References
NUDOCS 8807130143
Download: ML20150C783 (19)

Text

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s CONSUMERS POWER COMPANY Docket 50-155 Request For Change to The Technical Specifications License DPR-6 For.the reasons hereinafter' set forth, it is requested that the Technical Specifications contained in Facility Operating License DPR-6 . Docket 50-155, issued.to Consumers Power Company on May 1, 1964 for the Big Rock Point Plant be changed as described in Section I below:

I. CHANGES A. Revise Section 11.0 of Table of Contents page numbers to show Emergency Core Cooling Systems as pages "130-134" and Reactor Depressurization System as pages "135-142".

B. Add new Section 3.7(h) to read as follows:

"The RDS containment pei.etration assemblies seal pressure shall be examined'et six-month intervals."

C. Replace entire Section 11.3.1.5/11.4.1.? and bases to read as follows:

8807130143 880705

, PDR ADOCK 05000155 P PNU MIO688-0044-NLO4

2 "EMERGENCY CORE COOLING SYSTEMS REACTOR DEPRESEURIZATION SYSTEM LIMITING CONDITION FOR OPERATION 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: POWFR 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 SHUTDOWN condition within the following 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 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 />.

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d. Should a pilot valve be isolated from service and removed, it shall be 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 aa OPERABLE statue 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 e per quarter the isolation valves shall be full stroke eised.

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 FUNCTIONAiuY 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 ara no visible indications of damage or impaired operability to the snubbers or their attachments.

4. A FUNCTIONAL TEST of 10% (2) of the thirteen mechanical snubbers on the RDS shall be perforced. 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.

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Table 11.3.1.5 Instrumentation That Initiates RDS Operation I Limiting Ccndition for Operation Surveillance Requirement Minimum Protective OPERABLE Instrument Instrument Channel Parameter Channels Set Point Trip Test Calibration Trip t

Low Steam Drum 3 Above or Equal Moathly Each Major- -

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

Fire Pump (s) 3 '.05 Psig 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 5 120 Seconds-Fol- Monthly Each Major -

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

Monthly - -

A Through D Output Channels 3 - - -

Monthly 1 Through IV

• Fire Pump Start 1 -

Monthly -

Monthly

• Reference Specifications 11.3.1.4 and 11.4.1.4 for Bases.

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

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S 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 break in the pritary system. This will allow core cooling with the objectiv; 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 thena 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, 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 depressurization 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 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 of operation for the RDS.

The objectives 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 (ii) to prescribe the trip settings required to assure adequate performance. To conduct the required luput 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.

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

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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 MIO688-0044-NLO4 l-

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I 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 channel. Each of the four plant vartables 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 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 2 enter line.

The low steam drum level signal initiates a two-minute delay which allows a containment 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 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 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 coolant 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 delayod 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) or 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-oct-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 feature is used to bypass a channel when the channel has failed to the "trip" state and/or when channel maintenance is required. Bypassing of an input channel in the "trip" state or for raintenance causes the coincidence trip condition of the input channel to be changed from 1

1-out-of-3 or 2-out-of-4, respectively, to 2-out-of-3. The input channel l bypassed condition is alarmed as "channel 'X' unavailable" and "bypassed."

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7 Should an output channel require maintenance or should a single fault cause an

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output channel subchannel trip (two independent subchannels operate in 2 of 2 coincidence), the output channel actuation capability can be disabled by removing the associated 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 ouput 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 manualJy reset. This reset can be accomplished only after the iritiating signals (ie, input channel 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 c-c Bus 2A. Normal station power to Buses-1A 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 transmission 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 i be performed on the UPS while the plant is in operation.

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) aa specified in Section 9.0 of thera Technical Specifications. Testing of the depressurization valve is only pr 21 on a Refueling Outage basis to minimize pilot valve degradation.

The modified depressurization valve design permits the isolation of a pilot valve aesembly for maintenance during POWER OPERATION. Following removal 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 stroke test with a compressed gas is used MIO688-0044-NLO4

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l foJ1owing reinstallation to establish operability. These tests fulfill l 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 designed 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 cavity 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 requirements as the snubbers do not contain any fluid, seals, etc."

II. DISCUSSION A. Description of Changes Change A revises the Table of Contents to reflect the correct page numbers. Because of the reformatting of the RDS Section in Change C an additional page is needed for all of the information.

Change B relocates the existing surveillance requirement for the RDS containment penttration assemblies from the RDS section to the containment leakage section. This requircment pertains to containment leakage and not RDS operability and therefore is more logically located with the other containment leakage requirements.

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9 Change C reformats the entire RDS section of the Technical Specifications to the Standard Technical Specification format. This is done to provide clarity of the operability requirements associated with the RDS system and to reduce the potential for misinterpretation of the requirements. This change also replaces the quarterly requirement.to test operate the RDS depressurizing values with a requirement'to full stroke exercise all of the valves each refueling outage. It retains the existing partial stroke exercise test for the situation where the main portion of the depressurizing valve is disassembled for repair of main seat leakage. Also, Sections 11.3.1.5.1 and 11.3.1.5.3 have been combined into one action statement. With the reformatting, consistent usage of defined terms and the capitalization of defined terms has been incorporated as is the practice in Standard Technical Specifications. A portion of the note on the instrumentation table which referenced a revised Technical Specification section which is no longer applicable was deleted. The intent of the deleted portion of the note is now included within the 10CFR50.72/50.73 reporting requirements.

B. Background Attachment A provides a brief Reactor Depressurization System (RDS) description. The system is similar to but differs in that the Automatic Depressurization System (ADS) in newer BWRs discharges to a torus instead of directly to containment. Existing evidence suggests the current RDS testing practices have placed the Plant in a cycle which results in cooldown and heatups placing additional stress on all plant systems and equipment. The cycle is initiated by a depressurizing valve pi. lot valve leak. The Plant unidentified leak rate increases and it is shut down to repair the leaking pilot valve.

The ADS designs can tolerate pilot valve leakage unlike the Big Rock Point RDS in which the leakage contributes to primary system leak rare. Because the depressurizing valve must be completely disas-sembled to repair the leaking pilot, a post-maintenance partial stroke exercise is required by ASME Code to verify correct reassembly of the repaired valve. If it has been greater than MIO688-0044-NLO4

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_' 10 three months since the last partial stroke exercise on any one of the other three valves, Technical Specifications require the valves to be partial stroke exercised. Generally, if a pilot valve does develop a leak initially, one occurs soon after startup. The unidentified leak rate gradually increases, due to a leaking pilot valve, and the Plant _is again required to shut down and repair the leak. Frequently, the leaking pilot valve is in one of.the three valves that was exercised. Pilot valves which aren't partial stroke exercised during

-the previous shutdown generally don't leak, but when the Plant is forced to shut down for repair of a leaking pilot valve shortly after the three-month point, the exercising / leaking cycle is repeated.

This cycle has been .ery difficult to break and is a major contri-butor to reduced availability and unneeded Plant startups and shutdowns which challenge overall Plant safety. Approval of these proposed changes will eliminate routine performance of partial stroke exercising.

Current Surveillance Testing Requirements All RDS depressurizing valves are partial stroke exercised and all pilot valves are full stroke exercised each cold shutdown, except where it has been less than three months since the lest test. The depressurizing valves are ASME Code category B, Class 1. Because full stroke exercising of the depressurizing valves is not possible during power operation or without removing them from the system, relief from this code requirement had been requested in the past. At present, these valves are partial otroke exercised using a compressed  !

gas trapped in the spool piece between the isolation valve and the depressurizing valve. Evidence is available to show that this test is a significant contributor to chronic pilot valve leakage. 3 Proposed Surveillance Testing Requirements I

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11 All RDS depressurizing valves will be full stroke exercised in

-accordance with code requirements each refueling outage. Initially, it is anticipated this will be accomplished by sending-the valves off site to be full stroke exercised with steam. We are alsol researching methods which will allow conduct of the full stroke exercising on site and/or in place. In addition to the full stroke exercising surveillance test, we intend to disassemble and visually inspect one depressurizing valve each refueling outage as a preventive measure.

If results of the inspection indicate corrective repairs (other than scheduled preventive maintenance) to the valve are required for it to continue to perform its function, additional valves will be disassembled and visually inspected to ensure the concern is not generic.

i Basis for Proposed Surveillance Testing Requirements Depressurizing valve pilot valve leakage has been a chronic Plant problem since the RDS system was first installed. Reducing pilot valve leakage is imperative, since leakage can cause erosion leading to inadvertent valve actuation and a primary system blowdown into the containment. Several attempts to correct this problem have been made by the Plant staff and Target Rock Corporation, the valve manufacturer.

Attachment B lists previous and current attempts to reduce leakage.

Previous actions have not'been very successful.

l Following refurbishmer. of a leaking pilot valve, it is bench tested for leak tightness. The valve manufacturers standard acceptance criteria is 32 bubbles per minute. To help reduce cur leakage problem, standard Plant practice for several years has been to obtain a O bubble per minute leak rate before accepting a pilot valve for se rvice . Because a partial stroke exercise has been required to verify correct valve reassembly, approximately 407,of the valves develop pilot valve leaks almost immediately upon returning to power.

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12 A review of the surveillance test history and operating records shows 12 cases where depressurizing valves did not leak prior to a Plant shutdown, did not require exercising during the shutdown, and were not exercised during the shutdown. Not one of these valves developed a pilot valve leak during or soon after Plant startup. There are also seven cases where no maintenance was performed on the depressurizing valves, but they were partial stroke exercised as required by Technical Specifications prior to Plant startup. Upon return to power, in four of seven cases or about 57% of the time the pilot valves leaked. These results clearly show a relationship exists between pilot valve leakage and the number of exercise cycles the pilot valve experiences. By eliminating partial stroke exercising the probabil'ity of pilot valve leakage can be reduced.

During the 1988 refueling outage at Big Rock Point, the depressurizing valve tops have been modified and now have removable pilot valve assemblies. This modification consists of the installation of two isolation valves and a bolting flange between the depressurizing valve top and the pilot valve inlet and a bolting flange betweer the depressurizing valve top and the pilot valve outlet (See Figure 2 of Attachment A). This will allow the isolation and removal of the pilot valve assembly while the plant is in power operation. The combination of eliminating the partial stroke exercise surveillance test and the installation of removable pilot valve assemblies will l

reduce the potential for developing pilot valve leakage and, if leakage does occur, will allow leaks to be isolated and repaired without forcing the plant through shutdown /startup cycling and thereby improve overall plant safety and reliability.

The full stroke exercise on a refueling outage basis in lieu of the code requirement to full stroke exercise every three months is intended to demonstrate operability of the valves. To demonstrate operability of the va)ves following maintenance involving only pilot 4

valve removal and reinstallation, 1) pilot valve lift voltage will be measured, 2) pilot valve main disc lift will be measured, 3) pilot M10688-0044-NLO4

13 valve will be full stroke exercised, and 4) pilot valve will be leak tested. All of these actions will be accomplished prior to installation of the pilot valve on the depressurizing valve. After installation,

1) pilot valve solenoid electrical continuity will be checked, 2) pilot valve isolation valves will be verified to be open, and 3) pilot valve inlet bolting flange leakage will be checked by using system operating pressure. No exercising of the main valve will be required because the main valve internals will not have been disturbed during the process of repiccing a pilot valve. In situtations where the main valve internals are disturbed (i.e. the main valve :op is unbolted) partial and/or full stroke exercising will be used to demonstrate operability.

In summary, the elimination of partial stroke exercising is justified based on the observed relationship between depressurizing valve partial stroke exercising and pilot valve leakage and the built-in redundancy provided by each of the four blowdown ;.ths being capable of passing one third of the total required flow rate.

III. ANALYSIS OF NO SIGNIFICANT HAZARDS CONSIDERATION These proposed changes eliminata the requiremer.cs for partial stroke exercising the reactor depressutizing valves when in cold shutdown following three months of operation and reformat the RDS section of Technical Specificationa to be more consistent with Standard Technical Specifications. The partial stroke exercise is replaced with a full stroke exarcise each refueling outage. Reformatting of the RDS section has been done to clarify the limiting conditions for operation, action statements and surveillance requirements thereby reducing the probability of misinterpreting the requirements which ensure operability of the system.

Without significant design changes, stroke exercising of these valves is not possible with the Plant operating at power. Under the current Technical Specification requirements, if the Plant operated at power HIO688-0044-NLO4

14 for an entire operating cycle, it would be permissible to not partial stroke exercise any of the depressurizing valves until the end of the cycle. Because existing evidence suggests partial stroke exercising is a significant contributor to leakage through the depressurizing valve pilot valves, it is expected fewer Plant shutdowns will be required to repair pilot valves. If fewer Plant shutdowns are realized, an overall decrease in the probability of an accident or malfunction of equipment would result, as systems and equipment would not be subjected to additional cooldown and heatup cycles. The full stroke exercising on a refueling outage basis demonstrates the ability of the depressurizing valves to perform their design function.

In an ideal situation (ie no shutdowns during operating cycle) the partial stroke exercise would be performed once, at the end of the operating cycle as is the case with the proposed changes. The full stroke exercise is also a more reliable indicator of valve operability.

Therefore, an accident or malfunction of a different type is not created. Reducing testing frequency from quarterly to refueling outage could be considered as a reduction in the depressurizing valves margin of safety. However, the number and capacity of valves is not altered and the changed test method will reduce the probability of pilot valve leakage, increasing the reliability of the valves.

The proposed surveillance testing change has the same margin of safety as would occur if the Plant operated continuously without any shutdowns for a full cycle; fe, no partial stroke exercises would be performed. Consequently, these proposed surveillance testing changes do not involve a significant hazards consideration.

The proposed reformatting of the RDS section of Technical Specifications is admitstrative in nature. The reforaatting consists of organizing the section into a Standacd Technical Specification formac with the capitalization of defined terms and the consistent use of terminology throughout the section. Also, the surveillance requirement for the containment penetrations has been relocated to the containment leakage section of the Technical c ecifications because it more logically fits there. The existing partial stroke exercise surveil-

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lance test requirement has been retained to address operability testing for the main section of the depressurizing valve when it is disaasembled for seat leakage repair. It is intended to be used only in this situation. To be consistent with ASME Code requirements,'an action statement requiring functional testing of a pilot valve prior to returning it to service has been added. These proposed administrative changes clarify the limiting conditions for-operation, action statements and surveillance requirements thereby reducing the probability of misinterpreting the requirements which ensure operability of the RDS.

Therefore, these changes do not involve a significant increase in the probability or consequences of an accident previously evaluated, create the possibility of a new or different kind of accident or involve a significant reduction in a margin of safety. Consequently, these proposed changes also do not involve a significant hazards consideration.

VI. CONCLUSION The Big Rock Point Plant Review Committee has reviewed this Technical Specificction Change Request and has determined this change does not involve an unreviewed safety question and, therefore, it involves no significant hazards consideration. This change has also been reviewed under the cognizance of the Nuclear Safety Board. A copy of this Technical Specification Change Request has been sent to the State of Michigan official designated to receive such Amendments to the Operating License.

CONSUNERS POWER COMPANY By -

D P Hoffm President Nuclear 0 tions Sworn and subscribed to before me this 5th day of July, 1988.

/h tO ladu Elaine E Buehrer, Notary Public Jackson County, Michigan My Commission expires October 31, 1989 MIO688-0044-NLO4

4 ATTACHMENT A Description of System The purpose of the RDS i8 to provide a means to depressurize the primary system following a small break loss of coolanc accident so that the core apray system can provide core cooling. The RDS valves are not used for overpressure protec-tion and do not have a means of actuating in response to high system pressure.

The system is actuated as a result of low level signals combined with core spray system availability. The following pages show a simplified system piping diagram (Figure 1) and a photograph of the modified depressurizing valve top (Figure 2).

The Reactor Depressurization System (RDS) consists of four depressurization paths.- Each path has a depressurizing valve (pilot operated solenoid valve) and an isolation valve (split wedge gate valve with air operator) branching from a common header. A meane of bypassing steam around the isolation valves is provided by a remotely operated bypass isolation valve. It branches into four lines, each containing a manual bypass valve, which tie into the depressur-ization paths between the isolation valve and the depressurizing valve. Both the depressurizing valves and the isolation valves are normally closed. The bypass valvec are designed to be open during power operation. This allows the piping between the isolation valves and the depressurizing valves to be pressur- ,

ized. The bypass line allows the ieolation valves to be stroked each 90 days and reduces the possibility of inadvertent opening of the depressurizing valve if the isolation valves inadvertently open (due to loss of air, etc) during power operation.

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FIGURE 1 REACTOR DEPRESSURIZATION SYSTEM Depressurizing Isolation Valves Valves g

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p  := _

Bypass Isolation gg CV484 Valve fh

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Valves

{ i n i t i r 1 r Exhaust to D Steam Dru:n Euclosure r

I MODIFIED DEPRESSURIZING VALVE F

F

.q g TW .^ ~

sf5 '

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l ATTACHMENT B c

Actions Taken to Reduce Pilot Leakage  ;

1. Increased seating force by addition of spring preload washer.  ;

(November 79)

2. Improved disc seat alignment by using closer tolerance pilot discs.

(November 79)

3. Eliminate "crud" trap for seat contaminants by chamfering the pilot valvt inlet line socket weld connection. (February 85)

4. Change from carbon steel valve top flanges to stainless steel in order to reduce sources of possible corrosion products. (March 87)

5. Increase seating force of pilot disc by enlarging seat diameter and doubling preicad spring. A larger solenoid coil was required to compensate 1

for this modification. (March 87)

6. Added a sleeve between the pilot valve discs to eliminate the direct impingement of steam onto the pilot disc after passing thru the main pilot disc orifice. (March 87)

7. Provide a softer pilot valve disc seating surface by nickel plating its seat. This change is currently being tested on one of the four depres-surizing trains and has not yet been evaluated. (November 87)

8. Installation of removable pilot valve assemblies (June 88).

MIO688-0044-NLO4