ML20204F832

Proposed Tech Specs Change Re Partial Stroke Testing Following Main Valve Seat Leakage Repairs

ML20204F832
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	10/10/1988
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML20204F836	List: ML20204F832 ML20207L189
References
NUDOCS 8810210651
Download: ML20204F832 (11)

Text

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .

3

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 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 OPERABIE:

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 1.ogic 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 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 performed. FUNCTIONAL TESTE shall be used to verify that the force that initiates free movement of the snubber rod, in tension and compression, in 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.

d. Should a pilot valve be isolated from service and removed, the /

replacement pilot valve shall be functionally tested prior to /

installation and return to service. /

OO10210 PDR ADOCK W %0h55 ppg P

TSB1088-0163-NLO4

6 pressure. These variables are jointly processed by the four independent actuation syste'n 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 variables is monitored by four separate sensors. One sensor in each of the four variableo 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 :: tate 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 center 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 sta;c signal. Verification of a fire pump start and thus verif. cation that a source of core spray water is available at the core spray valves is i obtained when the pressure switch associated with the input chantel at either fire pump discharge has tripped, corresponding to a pressure eqtal to or exceeding 100 psig. This variable is used as an enabling input to the actuation system to prevent depressurizing the reactor coolant system whoi the cource of coolant required to cool the core is not available. A low reac'or water level signal is generated when the input channel reactor water level ainsor indicates a level 2 2'8" above the top of active fuel. Low reactor water livel is confirmation of the LOCA and with the other two inputs present (tit. delayed low drum level and core spray wator availability) causes the automate trip of the input channel. These trip level settings were chosen to be low enou # to prevent spurious actuation but high enough to initiate RDS operation so that post-accident cooling can be accomplished. I 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-out-of-4 to 2-out-of-3. A power supply failure associated with an output channel /

results in loss of that channel. / t

, 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 maintenance causes the coincidence trip condition of the input channel to be changed from The input channel 1-out-of-3 or 2-out-of-4. respectively, to 2-out-of-3.

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

TSB1088-0163-NLO4 l

b _ _ _ _

7 St.ould an output channel reqc.re maintenance or should a single fault cause an output channel subchannel trip (two independent subchannels operate in 2 of 2 coincidence), the output crannel actuation capability can be disabled by removing the associated 12s 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 a: ar-22 at 'enannel 'X' unavailable." ,

Since 3-out-of-4 output eiannels are required to assure design requirements are met (one output channel ot erstes one depressurf zing valve and one isolation

+

valve), the f ailure of or i output channel will not preclude achieving the required rate of depressur 'zation. This redundancy also enables maintenance to '

be performed on one ouput i hannel while the plant is in operation. l Once the RDS actuation system output channels are enabled (at least two input I chansels 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 channel trips or j uanual trip) have been restored to levels at which EDS 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 i cempletion of the depressurization. Each of the divisions (1, 2, 3 and 4) is ,

supplied with electrical power from one of leur 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 cnd 125 V d-c. Divisions 1 and 2 normally receive power from the existing 480 V a-c Bus 1A. Divisions 3 and 4 [

are supplied by 480 V a-c Bus 2A. Normal station power to Buses 1A and 2A can  !

be provided by one of three sources: (1) The scation 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 powe_ [

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 cuput 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 l preclude achieving the required rate of depressurization. This redundancy  !

also enables maintenance to be performed on the UPS while the plant is in operation, j i

Technical Specifications in this part, a.so 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 Outage basis to minimize pilot vrive degradation.

The modified depressurization valve design permits the isolation of a pilot valve assembly 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. These tests fulfill the post /

TSB1088-0163-NLO4

8 maintenance testing requirements of Section XI of the ASME Code (Subsection IWV-3200) for replacement pilot valies. / .

Four new containment penetration as semblies are used in transmitting electrical power, control and instrumentation signals between equipment located inside the containment building and facilities located external to the containment hinilding. 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 assenblies is designed and fabricated to the requirements of Subsection NE, Class MC vessels, of Section III of the ASME Code. . The penetration assemL.ies include a single aperture seal and a double 1

electrical conductor seal and .re 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. TPs 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 PDS 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 onubbers 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 penetration assemblies from the RDS section to the containment leakage section. This requirement pertains to containment leakage and not RDS operability and therefore is more logically located with the other containment leakage requirements.

TSB1088-0163-NLO4

1

, . l l

i

. 9

> l l

Change C reformats the entire RDS section of the Technical Specifica-tions' to the Standard Technical Specification format. This is done to provide clarity of the operability requireuents associated with the RDS system and to reduce the prtential for misinterpretation of the requirements, This changt also replaces the quarterly require-ment to test operate the RDS depressurizing values with a requirement to full stroke exercise all of the valves each refueling outage. /

Also, Sections 11.3.1.5.1 and 11.3.1.5.3 have been combined into one action statement. With the reformatting, consistent usate of defined terms and the capitalization of defined terms has been incorporated as is tne practice in Standard Technical Specifications.

A portion of the note on the instrumentation table which referenced a revised 'sechnical Specification section which is no longer applicablo 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 briet Reactor Depressurization System (RDS) description. The system is similar to but differs in that the Automatic Depressurization System (ADS) in neser BWRs discharges to a torus instead of directly to containment. Existing evidence suggests the current RDS resting 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 pilot 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 rate. Because the depressurizing valve must be completely disas-seca.ed 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 grecter than TSB1088-0163-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 disturb-ed during the process of replacing a pilot valve. /

In sunmary, the elimination of partial stroke exercising in justi*ied 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 paths being capable of passing one third of the total required flow rate.

III. ANALYSIS OF NO SIGNIFICANT HAZARDS CONSIDERATION These proposed changes eliminate the requirements for partial stroke exercising the reactor depressurizing valves when in cold shutdown following three months of operation and reformat the RDS section of Technical Specifications to be more consistent with Standard Technical Specificationn. The partial stroke exercise is rtplaced with a full stroke exercise eech refueling outage. Reformatting of the RDS section has been done to clarify the limiting conditions for operation, action statementa 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 TSB1088-0163-NLO4

)

l 14 for an entire operating cycle, it would be permissible to not partial stroke exercise any of the depressurizing salves until the end of the cycle. Because existing evidence suggests partial stroke exercising is a significanc contributor to Icakage through the depressurizing valve pilot valves, it is expecced fewer Plant shutdowns will be ,

required to repair pilot valves. It fewe.- Plant shutdowns are realized, an overall decrease in the probability /f an accident or malfunction of equipment would result, as systems and equipment would not be subjected to additional cooldown end heatup cycles. The full stroke exercising on a refueling outage basis demonstraces 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 ths valves.

The proposed surveillance testing change has the same margin of ,

safety as would occur if the Plant op1 rated continuously without any shutdowns for a full cycle; ie, no partial stroke exercises would be pe r f o rmed . Consequently, these pre?' sed surveillance testing changes do not involve a significant har..;ds consideration.

The proposed reformatting of the RDS section of Technical Specifications is adminstrative in nature. The reformatting consiste of organir.ing the section into a Standard Technical Specification format 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 Specifications because it more logically fits there. To be consistent with ASME Code requirements /

TSB1088-0163-NLO4 a $

15 a surveillance requiring functional testing of a pilot valve prior te /

returning it to service has been added. These proposed administrative changes cla.lfy the limiting conditions for operation, action statemente 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.

i VI. CONCLUSION The Big Rock Point Plant Review Committee has reviewed this Technical Specification 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 l 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.

CONSUMERS POWER COMPANY j By D P Hoffman (Signed)  !

D P Hoffman, Vice President Nuclear Operations Sworn and subscribed to before me this 5th day of July, 1988.  !

Elaine E Buehrer (Signed) [ seal]

Elaine E Buehrer Notary Public i Jackson County, Michigan  ;

My Commission expires October 31, 1989 i

i TSB1088-0163-NLO4

EMERGENCY CORE COOLING SYSTEMS REACTOR DEPRESSURIZATION SYSTF.M ACTI_0N : (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 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 l declared OPERABLE. If the plant is already in a COLD SHUTDOWN condition, I it whall not be started up until all snubbers are OPERABLE. (

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

j

a. At least once per month the instrumentation and system logic  ;

shall be FUNCTIONALLY TESTED as indicated in Table 11.3.1.b.

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 depressurfeation valves shall be full atroka  ;'

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 visible indications of damage or impaired operability to the snubbers or their i attachments.

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

i i  !

(

J s

136 Proposed I

! TSB1088-0163-NLO4

11.1.4.5 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 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 minutes), (2) high fire pump discharge pressure (either diesel- or motor-driven) and (3) low reactor water level. A low stoau drum level s1Fnal is generated when the steam drum level senser 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 allows a containment evacuation interval prior to syutem 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 lire 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 delayed low drum level and core spray water availability) causes the automatic trip of the input channel. These trip level settings were chosen te 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 af f ected 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 chennel 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 140 Proposed TSB1088-0163-NLO4

. 11 3.1.5 Bases: (Contd) l T, :hnical 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 Outage basis to minimize pilot valve degradation.

The modified depressuritation valve design permits the isolation of a pilot valve assembly for maintenance during F0WER OPERATIOh. Following removal and repair, the pilot valve as*>eebly is 'unctionally tested and leak checked prior to reinstallation to ensure operability. Following installation, actuation circuit continuity checks are performed. These tests fulfill the post maintenance testirg requirements of Section XI of the ASME Code (Subsection IWV-3200) for replacement pilot valves.

Four new containment penecration 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 tN1ded into spare containment penetration sleeves. The penetration assemblies are designed in accordance with IEEE 317 and are sei mically and environmentally qualified to the RDS design conditions.

l The pressure retaining oortion 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 electiical conductor seal and are designed to operate with the internal cavity pressorized 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 inspcetion 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 Laintained during and following a seismic or other event initiating dynamic loads. There are thirteen mechanical snubbers en 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 i fluid, seals, etc.

142 Propcsed TSB1088-0163-NLO4

_ _ _ _ _