ML20028A412

Forwards Justification for Interim Operation of safety- Related Mechanical Equipment,In Support of Resolution of SER Outstanding Issue 9.Justification Assures Safe Operation Pending Completion of Environ Qualification Program

ML20028A412
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	11/19/1982
From:	James Smith LONG ISLAND LIGHTING CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
SNRC-797, NUDOCS 8211220098
Download: ML20028A412 (15)

Text

e r=-.

g 8 -

LONG ISLAND LIGHTING COM PANY

, SHOREHAM NUCLEAR POWER STATION

-=w----r- P.O. BOX 618. NORTH COUNTRY ROAD + WADING RIVER. N.Y.11792 Direct Dial Number November 19, 1982 SNRC-797 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Mechanical Equipment Environmental Qualification SER Outstanding Issue No. 9 Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322

Dear Mr. Denton:

Enclosed are forty (40) copies of the justifications for interim operation with safety related mechanical equipment for which outstanding items were identified by the review program described in letter SNRC-737 dated July 23, 1982 from LILCO (J. L. Smith) to the NRC (H. R. Denton). This submittal fulfills the LILCO commitment made in letter SNRC-767 dated September 9, 1982 from LILCO (J. L. Smith) to the NRC (H. R. Denton), Exhibit 3, Question 1. An Index of Submittal has been attached for your convenience.

These justifications for interim operation provide assurance that the Shoreham Nuclear Power Station can be safely operated pending completion of the mechanical equipment environmental qualification program.

If you have any questions regarding this matter, please contact this office.

"ir'tt'e'"""'

L.'Smi Manager, Special Projects Shoreham Nuclear Power Station JPE:mp ()

Enclosures cc: J. Higgins All parties

/ ([()

8211220098 821119 PDR ADOCK 05000322 E PDR F C-8 9 3 5.1

]

INDEX OF SUBMITTAL Justifications for Interim Operation of 2

Safety-Related Mechanical Equipment 4

Mark No. Equipment Name 1B21*AOV81A-D Main Steam Inboard Isolation Valves 1C11-AOV126 Scram Inlet and Outlet Valves 1C11-AoV127 1C11-IICU01 Hydraulic Control Unit Scram Accumulator 1C51*EV801A-D Explosive Shear Valve IT46*AOV038A Air Operated Butterfly Valves 1T46*AOV039A-B VCS-60X-2 Forged Stainless Steel Piston Check Valves 0

6 a

EQUIPMENT INTERIM JUSTIFICATION Mark No. Equipment Name:

IB21*AOV081A,B,C,D Main Steam Inboard Isolation Valve System Name: Manufacturer:

Main Steam System ,Rockwell International Model No.:

1612 1.0 SYSTEM AND EQUIPMENT FUNCTION 1.1 System Function The main steam isolation valves form part of the nuclear system process barrier for openings outside the primary containment, and part of the pressure barrier for nuclear system breaks inside the primary containment.

1.2 Equipment Function The equipment functions that are to be performed by the main steam isolation valves, when exposed to a LOCA or a PBOC:

a. Main Steam and Containment Isolation for a LOCA or Main Steam Isolation for PBOC.

b. Close the main steam lines within the time established by design basis accident analysis to limit the release of reactor coolant.

The Inboard Main Steam Isolation Valves will experience the postulated harsh environmental conditions resulting from a LOCA during which they must complete the above functions. Additionally, these valves must not fail in a manner detrimental to plant safety or accident mitigation subsequent to a LOCA or PBOC accident.

4 2.0 NON-METALLIC SUBCOMPONENT(S) REQUIRING INTERIM JUSTIFICATION 2.1 Identification of Subcomponent(s)

The non-metallic subcomponents requiring justification are made of viton. The parameter requiring justification is radiation.

The valve pneumatic and hydraulic control units contain viton seals whose function is to contain the air and hydraulic oil necessary for valve operation.

-, un- -- , -- -m - , - , - - - - , , . .- ,- , - . , - - - - - - - - - - , . - . -

2.2 Comparison of Postulated Environment and Documented Environment ThedosetotheinboardMSIVvalves,ifrequiredtowithstang 40-year normal plus accident radiation, would reach 2.7 x 10 rads. This accident dose is based on a required operating time plus margin of 70 minytes. The radiation tolerance of the viton seals is 1 x 10 3.0 JUSTIFICATION STATEMENT As stated above, the 40-year normal plus accident integrated 7

radiation dose would reach 2.7 x 107 which is in excess of the radiation tolerance of viton (1 x 10 ). However, the 2-year normgl radiation plus 70-minute accident integrated dose is 9.8 x 10 rads. Since viton is the limiting material contained within this device and its radiation tolerance is greater than the above 2-year normal radiation plus accident dose, these valves are expected to perform their safety function for at

{ least an interim period of 2 years.

The valves, if they fail, do so in a closed position and are not required to reopen or reshut for the duration of the accident. Furthermore, the outboard MSIVs are qualified for

the 40-year plus accident integrated radiation dose.

Based on these considerations, interim plant operation is justified.

l

~

.t BOLLJMENT INTERIM JUSTIFICATION ,

Neck No.: Eautoment Name:

IC11-AOY126 Scram Inlet and Outlet

, IC11-AOV127 Yalves -

System Name: Manufacturer:

Control Rod Drive Robertshaw Model No.:

88470-Al 83460-B2 1.0 SYSTEM AND EQUIPMENT FUNCTION 1.1 System Function The scram inlet and outlet valves are important for the operation of the Control Rod Drive Hydraulic System. During normal operation, the inlet and outlet scram valves are held closed by control air pressure supplied to the tops of the diaphragm actuators by the scram pilot valves. Upon the receipt of a scram signal, the scram inlet valve opens to supply pressurized water to the bottom of the drive piston.

The scram exhaust valve opens slightly before the scram inlet valves, exhausting water from above the drive piston. The differential pressure across the drive piston causes the control rod to insert.

1.2 Eauipment Function The inlet and outlet scram valves are quick opening globe valves operato$ by an internal spring and system pressure. The scram outlet valve opens slightly before the inlet valve because of a larger spring in the valve operator.

The inlet and outlet scram valves are an essential part of the i Hydraulic Control Unit. It is an extremely rapid operating I system, fully activating within 4 seconds of receiving a scram signal and 6 seconds following the most limiting accident.

2.0 NON-METALLIC SUBCOMPONEMT(S) REQUIRING INTERIM JUSTIFICATION l 2.1 Identification of Subcomponent(s) l The non-metallic subcomponents requiring justification are made of teflon. They are the packing set for both valves and the seat ring for the 1C11-A0V127. The parameter requiring justification is radiation.

1

i t 9

~

2.2 Comparison of Postulated Environment and Documented Environment a

The inlet and outlet scram valves are exposed to a postulated 40-year normal and 6-month accident integrated radiation dose of 5.75 x 106 rada. The radiation tolerance of the " weak link" subcomponent teflon is 3.4 x 104 rads.

3.0 JUSTIFICATION STATEMENT The inlet and outlet scram valves are expected to perform their safety function at least up to the radiation tolerance of the weak link (teflon), which as stated above, is 3.4 x 104 rads. This value is less than the postulated value for the 40-year normal and accident integrated dose. However, the scram valves are fast acting and they fully perform their required safety function within 6 seconds following an accident. This activation time is far within the margin of 12 minutes (0.2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />), which is the acceptable span of time in which the scram valves can perform their function without being affected by radiation. The postulated dose to the scram valves feca 40-year normal and 12-minute accident radiation is 3.22 x 104 rads.

Upon initial operaton of the inlet and out3et scram valves, the control rods will be fully inserted and latched secure within 6 seconds following an accident. Thus, any suosequent failure of the scram salves is not detrimental to plant safety or accident mitigation.

Therefore, since the scram valves need function only once, immediately following accident initiation, and then are subsequently not required, the radiation dose seen by the scram valves is far less than the tolerable doso. Thus, the devices are expected to perform their safety functica as required.

Based on these considerations, interim plant operation is justified.

/

ROUIPMENT INTERIM JUSTIFICATION Mark No.: Eaulement Name:

IC11-HCUO1 Hydraulle Control Unit Scram Accumulator System Name_: Manufacturer:

Control Rod Drive General Electric

. '.l}nWof'*No . :

921DS95G 1.0 SYSTEM AND EQUIPhENT FUNCTION 1.1 System Function The purpose of the Hydraulic Control Unit Scram Accumulator is to provide a backup drive source for the control rods in the event that reactor pressure is low and/or the Control Rod Drive Hydraulic System (CRDHS) fails.

1.2 Eaulement Function The scram accumulator is an hydraulic cylindor with a free floating piston. The piston separates the water on top from the high pressure nitrogon below. Nitrogen is used as the source of stored energy to drive the water in tho accumulator.

The scram accumulator stor9s sufficient energy to fully insert a ctatrol rod at lower reactor vessel pressures. Once inserted, the control rod is restrictod from withdrawing by a latch which hclds the rod in place. At higher reactor vessel pressures, tra accumulator pressure is assisted by reactor vessel pressure. There are a total of 137 accumulators (one for each control rod).

The scram accumulator forms an integral part of the Hydraulic Control Unit. It is an citromaly rapid operating system, fully activating within 4 seconds of receiving a sceam signal and 6 seconds following the most limiting accident.

2.0 NON-METALLIC SUBCOMPONENT(S) REQUIRING INTERIM JUSTIFICATION 2.1 Identification of Subcomponent(s)

The non-metallic subcomponents requiring justification are made of teflon. They are the wiper rings and the O-ring seal. The parameter requiring justification is radiation.

• O s

2.2 Comparison of Postulated Environment and Documented Environment The accumulator is exposed to a postulated 40-year normal and 6-month accident integrated radiation dose of 5.75 x 106 rade. The radiation tolerance of the " weak link" subcomponent teflon is 3.4 x 10 4 reds.

3.0 JUSTIFICATION STATEMENT The accumulator is expected to perform its safety function at least up to the radiation tolerance of the weak link (teflon), which as stated above, is 3.4 x 104 rads. This value is less than the postulated value for the 40-year normal,and. accident integrated dose. However, the accumulator is a TasC sit'ing system and it fully performs its required safety function within 6 seconds fo11'owing an accident. This is far within the margin of 12 minutes (0.2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />),

which is the accoptable span of timo in which the accumulator can perform its function witheat being affected by radiation. The postulated dose to the accumulator from 40-year normal and 12-minute accident radiation 15 3.22 x 10 4 rada.

Once the accumulators have functionod, the control rods will be fully insertod and latched secure within 6 seconds following an accident. Thus, any rubsequent failure of the accumulators is not detrimental to plant safety or accidont mitigation.

Therefore, the device is expected to perfona its safety function as required.

Based on these considerations, interim plant operation is justified.

O e

e S

EQUIPMENT INTERIM JUSTIFICATION Mark No.: Equipment Name:

IC51*Ev801A, B, C, D Explosive Shear Valve System Name: Manufacturer:

Traversing In-Core Probe (TIP) System Consolidated Controls Corp.

Model No.:

73074 1.0 SYSTEM AND EQUIPMENT FUNCTION 1.1 System Function The purpose of the Traversing In-Core Probe (TIP) System is to provide a highly accurate flux signal used to calibrate the Local Power Range Monitoring (LPRM). While in service, it provides.

continuous line plots of the axial flux distribution at 31 locations in the core. The TIP equipment is designed to maintain the integrity of the primary containment if isolation is required.

1.2 Equipment Function The explosive shear valve is manually activated in an emergency situation. It is used only if containment isolation is required when the TIP is beyond the ball valve and power to the TIP system fails.

The shear valve is located on the detector guide tube between the ball valve and the drive mechanism. The shear valve is a safety device designed to cut the guide tube and drive cable and to seal the guide tube if a leak develops in the reactor coolant system when l the drive cable is traversing the guide tube. An explosive charge, detonated by an electric current that is controlled by a key switch mounted on the control panel, provides the force to drive a chisel-shaped slug through the guide tube and drive cable, thereby cutting the cable and scaling the reactor end of the guide tube.

2.0 NON-METALLIC SUBCOMPONENT(S) REQUIRING INTERIM JUSTIFICATION 2.1 Identification of Subcomponents The non-metallic subcomponents requiring justification are made of teflon. They are the seals, packing material, insert (guillotine seal) , and teflon-coated electrical wire. The parameter requiring justification is radiation.

b 2.2 Comparison of Postulated Environment and Documented Envirci.nont The shear valve is exposed to a postulated 40-year nogmal and 6 month accident integrated radiation dose of 5.26 x 10 rads.

The radigtion tolerance of the " weak link" subcomponent teflon is 3.4 x 30 rads.

3.0 JUSTIPFCATION STATEMENT The TIP shear valve is expected to perform its safety function up to at least the radiation tolergnue ol -ile %cak link (teflon) , which, as stated above, is 3.4 x 10 rads. Thus, the integrity of the reactor coolant pressure boundary is assured up to this radiation dose. In comparison, the postulated 40-year gormal and 6-month accident integrated radiation dose (5.26 x 10 rads) is slightly greater than the radiation tolerance of the weak link. Thus, it is doubtful whether any significant degradation will occur in the seals. Additionally, several seals must be penetrated in order for the reactor coolant to pass the boundary.

If a containment isolation signal is present while the TIP system is in use (dctector is inserted) , an automatic withdrawal of the detector trave cable would occur and the ball valve would close to complete the containment isolation function. With a concurrent loss of 120V and 208V AC power to the TIP system, the ball valve, if open, would close and a jam of the drive cable will result.

Therefore, the shear valve would be manually activated only if a containment isolation signal is present and the detector cannot be withdrawn.

i In addition, the probability of a LOCA in a location that will subject the TIP valve to radiation combined with a loss of power during the time in which the TIP system is operating (plant procedures require the TIP system to operate for only four hours during a month) is quite small.

Thus, it is very unlikely that the integrity of the reactor coolant pressure boundary will be violated due to radiation-induced failure of the TIP shear valve.

Based on these considerations, interim plant operation is justified.

l t

l l

l l

SHEET IJ-172-1 EQUIPMENT INTERIM JUSTIFICATION PREPARED BY: S. POLLER /

W. DROOKS MARK NO. 1T46*AOV038A 1T46*AOV039A IT46*AOV039B* -

'**EgGTPmdiF'NAME: AIR OPERATED BUTTERFLY VALVE SYSTEM NAME: REACTOR MANUFACTURER: FISIIER COffrROLS BUILDING STANDBY VENTILATION SYSTEMS (RBSVS) MODEL NO.: 9220

1. Equipment Function The RBSVS is initiated during an accident or abnormal condition. When the RBSVS is initiated, the reactor building secondary containment is automatically isolated from the primary containment by the closing of several valves. Valve IT46*AOV038A is a primary containment purge isolation supply valve. Valve IT46*AOV039A is a primary containment purge isolation exhaust valve. Both valves are part of the primary containment purge system and are normally closed and not required to operate during normal plant operation or accident conditions. These valves perform no other safety function during an accident than to remain closed for the duration of the accident.

2. Non-Metallic Sub-Components Requiring Interim Justification

! T-Ring (EPT-Ethylene / Propylene Terpolymer, Sulphur Cured)

Radiation Threshold = 8.77 x 10 Rads Maximum Service Temperature = 300 F An adjustable T-Ring seat is contained in the valve disc. The adjusting set screws and compression ring force the T-ring against the body bore seating surface, which provides a bubble tight shut-off feature.

• This valve has been found to be acceptabic based on S&W calculation No. SNPS-1-URB-24-T.

SHEET IJ-172-2 i

These two purge isolation valves are located within the Drywell where the postulated 40 year gormal plus 6 month harsh environment ragiationlevelsare1.2x10 rads for valve IT46*AOV038A and 1.8 x 10 rads for valve IT46*AOV039A and the postulated maximum temperature for both valves is 340 F. These radiation levels and the maximum temperature exceed the radiation threshold level and the maximum service temperature of the EPT material.

3. Justification Statement Valve IT46AOV038A This valve is the inboard isolation valve of the inboard and outboard isolation valve combination of IT46*AOV038A & B. The outboard isolation valve (IT46 *AOV038B) is located in zone H-10 where the postulated 40 yeag normal plus 6 month harsh environment radiation level of 5.3 x 10 rads and the postulated maximum temperature of 177 F are well below the threshold radiation level and maximum service temperature of the EPT material. The outboard isolation valve will therefore withstand the harsh environment and will perform its safety function throughout the accident period.

Based on the fact that the outboard isolation valve (IT46*AOV038B) is a redundant acceptable component, the required safety function will be accomplished and thus interim operation is justified.

Valve IT46*AOV039A This valve is the inboard isolation valve of the inboard and outboard isolation valve combination of IT46*AOV039A & B. The outboard isolation valve (IT46*AOV039B) is located in an area of zone K-15 where the postulated 40 year gormal plus 6 month harsh environment radiation level of 1.8 x 10 rads and the postulated maximum temperature of 158 F are well below the threshold radiation level and maximum service temperature of the EPT material. The outboard isolation valve will therefore withstand the harsh environment and will perform its safety function throughout the accident period. Based on the fact that the outboard isolation valve (IT46*AOV039B) is a redundant acceptable component, the required safety function will be accomplished and thus interim operation is justified.

EQUIPMENT INTERIM JUSTIFICATIOM PREPARED BY: K. MENON MARK NO. VCS-60X-2 EQUIPMENT NAME: FORGED STAINLESS STEEL PISTON CHECK VALVE SYSTEM NAME: NUCLEAR BOILER (1B21) MANUFACTURER: VELAN ENGINEERING INSTRUMENT & SERVICE AIR (lP50) MODEL NO: WO-203B-13MN General: Piston check valves of description nwmber VCS-60X-2 located in the i primary containment, associated with the Safety Relief Valve

accumulators and containment isolation function are addressed in '

this interim justification.

l. Equipment Function:

A) Eleven 3/4" piston check valves are used in the Nuclear Boiler System (1B21) as check valves to prevent backflow and loss of pressure in the nitrogen supply lines to each of the accumulators for the short term operation of the individual SRV's.

B) Four (two 1 " and two 3/4") piston aheck valves are associated with the containment isolation function of the N supply system.

2

2. Non-Metalic Sub-Components Requiring Interim Justification Soft Ring, Neoprene Disc Sgat i Radiation Threshold = 8x10 rads Maximum Service Temp. = 194 F

! These check valves are all located within the Primary Containment where

the postulaged 40 year normgl plus 180 day LOCA radiation levels range from 1.0x10 rads to 1.8x10 rads and the postulated maximum temperatures range from 225 F to 340 F. The postulated radiation levels and the maximum temperatures exceed the radiation threshold level and the maximum service temperature of the neoprene material.

Exposure to this postulated harsh environment may result in the deterioration of neoprene which could pote,tially cause these valves to lose their scaling characteristics.

I j 3. Justification Statement i

! A) Safety Relief Valves:

There are four modes of supplying nitrbgen to the SRV's (normal operation, short term supply, intermediate term supply, and long term supply).

During normal operation, nitrogen is supplied to the short term and f intermediate term accumulators. Nitrogen is retained in these accumulators hf check valves so they remain pressurized and are available for operation of SRV's in case the normal nitrogen supply fails during normal operation.

In the event of an accident the nitrogen supply header that normally supplies nitrogen to the drywell SRV accumulator headers is automatically isolated by a LOCA signal by environmentally qualified motor operated valves (lP50*MOVll3A&B) located outside containment.

In addition, two other environmentally qualified motor operated valves (lP50*MOV105A&B), located inside the drywell, are automatically closed to divide the supply header piping. The closing of these additional valves also isolates other lines from the now divided accumulator supply headers inside the drywell.

Isolation of the piping system in this manner will eliminate the possibility of pressure loss in the accumulators caused by any backflow leakage due to deteriorated check valve neoprene seats.

Intermediate term supply to the SRV accumulators can be accomplished by two intermediate term nitrogen accumulators each capable of supplying sufficient nitrogen for a minimum of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of operation. This backup of intermediate nitrogen supply to the divided header is made available by automatic opening (on a LOCA signal) of environmentally qualified motor operated valves (1P50*MOV114A&B) located in the connecting piping downstream of valves IP50*MOV113A&B that have isolated the N normal supply 2

header. This backup supply of nitrogen further ensures that the nitrogen headers serving the SRV's remains pressurized and thus prevents any backflow out of the short term accumulators resulting from any deterioration of the neoprene seats of the piston check valves.

Beyond 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, for long term operation of the SRV's, the pneumatic system uses connections located outside the reactor building to

custain SRV operability indefinitely.

Based upon the redundant supply of nitrogen to the SRV accumulators, the required safety function will be accomplished and thus interim operation is justified.

B) Containment Isolation Function of VCS-60X-2:

Drywell - Two 1 " piston check valves located on the nitrogen supply header inside the drywell function, on reverse flow, as containment isolation valves.

In the event of an accident with the postulated harsh environment, the neoprene valve seat may deteriorate which could cause.these valves to lose their sealing characteristics. If a subsequent loss of nitrogen supply to the drywell occurs,

• these piston check valves may experience backflow leakage out of the nitrogen supply header. This loss of nitrogen would be detected by environmentally qualified pressure transmitters IP50*PTll6A&B and environmentally justified pressure switches IP50*PS105A&B. At that time an environmentally qualified valve, either IP50IMOV103A or B, could be remote manually closed to provide containment isolation.

• As previously stated in the SRV justification, the possibility of loss of pressure in the accumulators and associated nitrogen supply header piping is extremely remote.

- Wetwell - Two 3/4" piston check valves located on the nitrogen cupply header inside the wetwell function, on reverse flow, as containment isolation valves.

The supply header in the wetwell is used only to test the downcomer vacuum breaker valves. In the event of a LOCA signal, environmentally qualified motor operated valves (lP50*MOV104 and 1P50*MOV106) would close to isolate the conteinment. Therefore, even if the neoprene valve seats were to deteriorate because of the postulated harsh environment, containment isolation would not be jeopardized. Furthermore, these environmentally qualified motor operated valves are normally closed, eliminating the possibility of containment leakage.

Based on the redundancy of the environmentally qualified motor operated isolation valves, the required j safety function will be accomplished and thus interim operation is justified, s

i f

i

)

. .._ _ _, _ . . . _ . , . . . _ .