ML20076K144
| ML20076K144 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 06/30/1983 |
| From: | James Smith LONG ISLAND LIGHTING CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| SNRC-916, NUDOCS 8307080269 | |
| Download: ML20076K144 (16) | |
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LONG ISLAND LIGHTING COMPANY h
SHOREHAM NUCLEAR POWER STATION wmw-a-v= xa P.O. BOX 618, NORTH COUNTRY ROAD e WADING RIVER. N.Y.11792 Direct Dial Number June 30, 1983 SNRC-916 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 Dynamic Qualification / Justifications Safety Evaluation Report Outstanding Item No. 8 Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322
Dear Mr. Denton:
As part of our effort to completely resolve outstanding issue num-ber 8 of the Shoreham Safety Evaluation Report, we hereby submit justifications for interim station operation (Enclosures lA through lE) for the following equipment which may not be fully qualified at fuel load which is presently scheduled for Aug. 31, 1983.
t A)
Radiation Monitoring Instrumentation Mark No. 1Dll*PNLil7A&B Mark No. 1Dll*P-126, 134 B)
SDV Vent and Drain Valves i
Mark No. 1C11*AOV81/82 Mark No. 1Cll*AOV50/51 C)
SDV Solenoid Valves Mark No. 1Cll*SOV044 Mark No. 1Cll*SOV052 D)
HPCI Turbine Mark No. lE41*TU-002 i
l E)
In Vessel Rack Mark No. 1F16*RAK-09 A few additional equipment items which are essentially qualified but not yet fully documented will have full qualification documen-tation transmitted from Stone & Webster or General Electric to the plant site permanent file prior to fuel load.
)
kh ll 8307080269 830630 PDR ADOCK 05000322 l}
SNRC-916 June 30, 1983 Page 2 Based on information we have received from individual vendors, our Architect-Engineer, and our Nuclear Steam System Supplier, we have reason to conclude that all but one of the equipment presented in these Enclosures will have all of its associated testing and analysis completed prior to exceeding 5% power.
Only formal completion of documentation packages will be outstanding at that time.
The only exceptions are the SDV solenoid valves and the HPCI turbine.
As stated in the justification for interim operation, a turbine similar to the Shoreham turbine was recently subjected to a dynamic test which enveloped Shoreham requirements.
G.E.
was given authorization by LILCO in April 1981 to perform the HPCI Turbine Qualification Program.
G.E.
Report NEDE-30123 transmitted to LILCO in May 1983 documents the results of this program.
The report demonstrates that Shoreham specific parameters were successfully enveloped.
Prior to the HPCI qualification testing, G.E.
initiated certain subcomponent changes to assure a higher probability of test success.
Following completion of the Qualification Program, G.E.
Informed LILCO that modifications to Shoreham's HPCI turbine would have to be made to establish similarity to the test turbine.
A list of required replacement parts are included in Enclosure A.
The test report also indicated that additional changes or verifi-cations are needed, as listed in Enclosure B.
At present, only two of the fourteen parts, namely items 1 and 2 on Enclosure A, have been changed out.
LILCO requested that G.E.
provide the balance of replacement parts to support a fuel load date of August 31, 1983.
However, because of lead time needed to procure the replacement parts, the balance of the modification package is not expected to arrive in time to complete the modifi-l cation prior to August 31, 1983.
Our current information is that the replacement parts will be available for installation no sooner l
than September 15, 1983 and it will take a minimum of two months to install and test the modifications.
We therefore have deter-mined that we cannot complete this modification prior to exceeding 5% power.
We are investigating the possibility of performing the modifications prior to the first refueling outage and will, if possible without violating our Technical Specification Limiting Conditions for Operation.
Given the extent and time needed for completion of these modifications, it appears unlikely that we will accomplish this task prior to the first refueling outage.
LILCO's Equipment Qualification (IK)) Group, after informal l
discussions with G.E.
and after a preliminary review of the test results, concluded that we could not fully establish seismic qualification of the HPCI turbine without replacing components as J
r 1
SNRC-916 June 30, 1983 Page 3 listed in Enclosures A and B, with the possible exception of items 12-14 (wiring and conduit) on Enclosure A.
Another conclusion reached as a result of these discussions is that a number of these G.E.
modifications have been motivated primarily by environmental rather than seismic considerations.
However, it is our position that we have provided adequate justification for interim operation without these modifications and, considering the similarity between the Shoreham design and other recently licensed BWR designs, we need not complete these as a prerequisite to full power operation.
In accordance with R.
L.
Tedesco's letter to LILCO dated January 28, 1981, four copies of this submittal with enclosures are being forwarded to Dr. Morris Reich at Brookhaven National Laboratory.
Should you have any questions regarding this matter, please do not hesitate to call this office.
Very truly yours, v
J.
L.
Smith Manager, Special Projects Shoreham Nuclear Power Station GJG/ law S3 Attachment Enclosures cc:
J.
C. Higgins Dr. Morris Reich, BNL (4)
All Parties Listed on Attachment 1
]
e-SNRC-916 June 28, 1983 Enclosure A Item Name Quantity 1.
Woodward EGri Control Box (9903-103) 1 2.
Woodward RGSC (9903-091) 1 3.
Woodward EGR Hydraulic Actuator (DR 9903-099) 1 4.
Woodward Remote Servo (9903-060) 1 5.
Electro. Corp. Speed Sensor (725967) 1 6.
Terry Overspeed Test Controller 1
7.
Skinner Trip Solenoid (L2LB5150) 1 8.
Fisher Viton Diaphragm (1F702202402) (for Fischer PCV oSSES) 1 9.
Robertshaw Buna-N Diaphragm (25471-A1) (for Robertshaw Valve VC210) 1 10.
Amot Thermo Valve (2BFC-130-11) 1 11.
Square D Temperature Switch (9025-BFW-42W1) 1 12.
EJ Steven Wire #5007-1B 22 ft 13.
EJ Steven Wire #S5007-3 275 ft 14.
Anaconda Conduit #NWC 25 ft
SNRC-916 June 28, 1983 Enclosure B RESOLUTION FOR NOTICE OF ANOMALY Notice #
Three areas of the turbine assembly were identified as 1
requiring upgrade in order to support reasonable nozzle load definition:
a)
Increase size of coupling and pedestal dowel pins from #11 to #12 pins, b)
Weldment for governor end pedestal guide blocks must be 3/8" minimum fillet.
c)
Allowable anchor bolt stress (upset and faulted condition) must equal 12,000 psi minimum.
3 Additional piping / tubing supports should be installed in the following locations as per G.E.
Test Report 21021-1 and VPF2763-313-1:
a)
Oil supply to driven equipment b)
Oil supply to overspeed trip valve c)
Hydraulic tubing between the Woodward Actuator and Remote Servo.
In addition, two other areas require supports as per Turbine Vendor Piping Layout Dwg. #125777E-Rev.
A, 5
sheets and VPF 2763-312 (1 thru 5).
d)
Auxiliary oil pump discharge line, e)
Instrument sensing line.
RESOLUTION FOR NOTICE OF DEVIATION 7
Inspection of bonnet of Fisher PCV 655ES, Pressure Con-trol Valve. Current design is provided with an internal support ring which assures operator pressure across the surface area of the diaphragm.
If bonnet is obsolete design, install current model.
EWCLOSURE lA SHOREHAM NUCIEAR POWER STATIM - UNIT 1 INIERIM JUSTIFICATICN Mark No:
1Dll*PNL ll7A&B Catpcnent Name: IE Cabinets & Internals System Nane: Radiation Monitoring Vendor: Kaman Instrumntation Spec. No:
SH1-475 Model No: Various Quantity:
2_
1)
LGJIPMENT Operation /
Hot Standby @ Cold Shutdown C Both C Neither C Other IX_1 Post Accident 2)
QUALIFICATIm SUlHARY here are nineteen (19) modules located in the two Ccntrol Rocm Cabinets.
Seismic test reports for ten (10) modules have not been received. Seismic test reports for eight (8) of the modules have been reviewed and approved by SWEC. The remaining module is Category II and is isolated frcm the Category I equipmnt. Eight of the ten pieces of equiptent requiring seismic reports are digital and analog isolaticn modules. The test report for the modules will ecnsist of two parts; the first part deals with the circuitry and the second part deals with the structure. The circuitry section of the report will be based on test results of the High Range Gas Miu-yaters (lDll*PNL 135&l37),
the Safety Related Monitoring System (SRMS), a module not purchased by LILCO but qualified by Kaman as part of their generic qualification program, and the Ratemet.ars (IDll*RIS 085A&B). The test report for the microccuputers has been reviewed and approved as revised by SWEC. The coments are minor and will not affect the seismic qualificaticn. Discussions with Kaman seismic perscnnel revealed that the SRMS module has been successfully tested to the following levels:
3.0g's ZPA, vertically and horizcntally, frcm 1 to 100 Hz, multifre-quency, biaxially. These levels well exceed those required at Shoreham. The ratemeter was tested and qualified at Kaman's facilities in Colorado Springs.
The test report has been reviewed and approved. The structural section of the report will be based cn the test results of the NIM Bin, the unit which houses the isolaticn module. The NIM Bin was tested at Acton Labs in January 1983 and discussicns with Kaman revealed that it had been qualified to the same levels as the SRMS module, again, well exceeding the Shoreham requirements.
The r-ining two items are Recorder Power Supplies which are very similar to the Rateneter Power Supplies. The test report for these items will consist of an analysis which will be based on test results of the Rateneter Power Supplies (lD21*E/S ll7B&D) and the High Range Gas Miu-yuters (lDll*PNL 135&l37), all of which have been seismically tested to the Shoreham levels and approved by SWEC.
Reports for the Isolation Modules are expected to be subnitted by late July 1983 and reports for the Recorder Pcwer Supplies are expected to be submitted by the end of August 1983. Final qualification documentation is expected to be in Permanent Plant Files by September 1983.
1 of 4
- 3) DESCRIPTION OF COMPONENT SAFETY FUNCTION Components in these cabinets are not used to achieve safe-shutdown. Monitor components located in these cabinets are used to estimate and evaluate the real or potential radioactive effluent releases and implement LILCO's commitment for Reg. Guide 1.97 (Rev. 2) monitors. Components for the High Range Area Monitors and the Post Accident Effluent Monitors are located in these cabinets.
COMPONENT MARK... JER CABINET NAME ID11*PNL ll7A A
1D21*RIS 085A Ratemeter 1Dil-RIS 127A Keric IDll*RIS 134A Keric IDll*RR 502 Recorder 1Dil*E/S ll7A Power Supply (For Recorder) 1D21*E/S ll7B Power Supply (For Ratemeter) 1Dil*XE 134X Isolation Module Digital ID11*XE 134Y Isolation Module Analog ID21*XE 085AX Isolation Module Digital ID21*XE 085AY Isolation Module Analog ID11*PNL 117B B
ID21*RIS 085B Ratemeter ID11*RIS 126A Keric 1Dil*RR 503 Recorder IDll*E/S ll7C Power Supply (For Recorder) 1D21*E/S ll7D Power Supply (For Ratemeter) 1D11*XE 126X Isolation Module Digital IDil*XE 126Y Isolation Module Analog ID21*KE 085BX Isolation Module Digital ID21*XE 085BY Isolation Module Analog
- 4) FAILURE CONSEQUENCE ANALYSIS Failure of a component in one channel will not affect the operation of other channels located within the cabinet. Each channel is mechanically isolated by barriers and elec-trically separated with individually fused Class IE fuses.
Independent component / channel fusing protects IE power from damage in the event of short-circuiting within the components.
In the event of failure of the High Range Are: Jadiation Monitor components located in the IE cabinets, samples of the containment atmospizere can be obtained using the Post Accident Sampling System and analyzed to estimate exten; of core damage. In the event of failure of the Post Accident Effluent Monitors' components located in the IE Cabinets, local indication and/or grab samples with analysis can ia used to estimate releases.
- 5) JUSTIFICATION
SUMMARY
The seismic failure of the components on these cabinets will not degrade the safety function of any other component required for safe-shutdown or for LOCA mitigation and will not mislead the operator.
In addition, the Post Accident Sampling System can be used in assessing the extent of core damage and local indication and/or grab samples can be used for radiation release assessment.
Based on these considerations, interim plant operation is justified.
L 2 of 4
SHGEHAM NUCIEAR POWER STATIW - UNIT 1
{
mrERIM auSTIFICATIN Mark No: ID11*P-126, 134 Ccmponent Name: Auxiliary PuTo Skid System Name: Radiation bbnitoring Vendor: Kaman Instrumentation Spec. No: SH1-475 fbdel No: KPM Quantity: 2 1)
EQUIPMENT REwlNrMNTS Hot Standby @ Cold Shutdown C Soth U Neither [ Other
@ Cperation/
Pcst Accident 2)
QUALIFICATIN SLM1ARY 2e Auxiliary Putp Skids are to be qualified by analysis. Ccxuponents on the skids are identical to those located on the High Range Gas bbnitors, 1Dll*PNL 126&l34. W e above manitors have been seismically tested at Acton Labs, to the following levels; 6.0 g's ZPA, vertically and horizontally frcm 1 Hz,to 100 Hz., multifrequency biaxially. Rese levels well exceed those i
required at Shoreham. Seismic test reports for these monitors have been reviewed and approved by SWEC. 'Ihe remaining item on the skids requiring analysis, the base plate on which the pump is mounted, will be analyzed using an ANSYS program.
'Ihe final test report is expected by late August 1983. Final qualificatiut documntation is expected to be in Permanent Plant Files by September 1983.
3)
DESCRIPTICN Or CWPWENT SAETIY EUNCTIN me putps mounte. on these putp skids are not used to achieve safe shutdown.
d
'Ihey are used to supply air sanples frcm both the Station Vent and Reactor Building Standby Vent to the Post Accident Station Vent and Reactor Building Standby Vent Monitors, respectively.
The monitors are required by LIICO's emmitment to inplement Reg. Guide 1.97 (Rev. 2) and are used to estimate and evaluate radioactive effluent releases frcxn the vents.
4)
FAIIURE CCNSEQUENCE ANALYSIS In the event of failure of the Auxiliary Pump Skids, noble gas grab sanples can be obtained frca the normal range Reactor Building Standby Vent bbnitors and a Continuous Air Monitor (CAM) which can be connected at the sample inlet to the High Range Station Vent Monitor. Rese samples can then be used to estimate releases and provide input into the emergency plan. We Reactor Building Standby Vent bbnitors,1Dll*PNL 021&O22, have been seismically qual-ified. 'Ihe Continuous Air Monitor does not require seismic qualification since it will not be installed and utilized until after the seismic event has occurred.
3 of 4
5)
JUSTIFICATIm SQtRRY me seismic failure of the Auxiliary Punp Skids will not degrade the safety of any other ccmpcnent required for safe shutdown or IIX2 mitigaticn and will not mislead the operator. Failure of punps will initiate the " PUMP FAIIURE" alarm cn the KERIC's,1Dll*RIS 126M134A, which are located cn the Ccntrol Roca Cabinets,1Dll*PNL ll7A&ll7B, thus alerting an operator to the problem. Se normal range Reactor Building Standby Vent Mcnitors and a Continuous Air Mcnitor can then be utilized to provide noble gas grab sanples.
Based on these considerations, interim plant operaticn is justified.
4 of 4
ENCLOSURE 1B JUSTIFICATION FOR INTERIM OPERATION NAMEg '
SDV Vent and Drain Valves MPL/ MARK NO.:
C11-F010/F011/1C11*A0V081/82 c11-F180/F181/lc11*NA050/51 SAFETY FUNCTION:
To close, isolating the scram discharge volume from the radwaste drain system.
FAILURE MODES:
Fail Open Fail Closed X
Loss of Power Loss of Air X
Loss of Pressure Integrity Loss of Structural Integrity Distortion of Mounting i
These valves are air operated. Therefore, " loss of power" does not apply.
" Loss of air" causes the valves to close, subordinating this
[
failure mode to " fail closed".
Other failure modes identified which could cause the valve to fail open have been evaluated and judged by I
General Electric to be not credible under worst case dynamic loading conditions.
FAILURE EFFECT:
(.
P.
A.
Effect on Primary Use I-If the valves fail closed, there will be SDV water accumulation.
- h~
B.
Secondary Effect None l
l DISCUSSION AND CONCLUSION:
I SDV vent and drain valves identical to those on the Shoreham Plant were l
4 recently dynamically tested to conditions which envelop Shoreham require-l c ments.
The valves performed satisfactorily both during and following the dynamic tests. All that remains is to document the test results and complete the Shoreham SQRT forms.
![!;.
Interim plant operation with the SDV valves not fully documented is justified since the dynamic test program has been successfully concluded.
RWH:rm/A060927-1 6/9/83
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t ENCLOSURE 1C JUSTIFICATION FOR IN'IERIM OPERATICN 1@JE:
SDV Solenoid Valve NIAMRK NO.:
Cll-F009/1Cll*SOV044 ii Cll-F182/lCll*SOV052 SAITIY FUNCTICN:
'Ib open, permitting air to vent fran the air supply header, thereby closing the CRD vent and drain valves.
FAIIURC MDDES:
Fa.il Open (deenergized)
X Fail Closed (energized) i Ioss of Power X
Ioss of Air t
Ioss of Pressure Integrity Ioss of Structural Integrity Distortion of } bunting i
An evaluation of the solenoid valve, and the contrcl rod drive hydraulic control systen in which it is used, shtr.c that the only credible failures f
which affect this normally closed valve's safety functions are " fail L
l open" and " loss of power". Other failures are judged by General Electric to not be credible under worst case dynamic loading conditions, or are subordinate to " fail op n".
- p FAIIURE IznCr:
i-A.
Effoct on Prim' Use Valve fails open, and hence safe, en loss of po mr, permitting air to vent frun the air supply header. This closes the SDV vent and 4
drain valves.
L.
B.
Secondary Effect Water accunulation in the SDV instrument volume results in an auto-matic SCRAM when the water level reaches a predetennined set point.
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DISCUSSICN AND CONCLUSICNi' The SIN soleroid valve is designed to fail safe cm loss of power facilitating closure of the vent and drain valves. This is considered to be the only credible failure node which the solenoid
~
valve can experience. Any water accunulation in the SDV instrtunent
!I volume will be detected by level switches nounted on the instrument volume. If the level exceeds a predetennined setpoint, the reactor will scram autcznaHeally.
Therefore, interim operation with this valve does not pose a safety hazard.
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e Enclosure ID JUSTIFICATION FOR INTERIM OPERATION NAME:
HPCl Turbine MPL/ MARK NO.:
E41-C002/lE41*TU-002 SAFETY FUNCTION:
To supply high pressure emergency cooling water to the reactor pressure vessel, in order to maintain reactor core temperatures within specification limits.
FAILURE MODES:
Fail Open Fail Closed Loss of Power X
Loss of Air Loss of Pressure Integrity Loss of Structural Integrity X
Distortion of Mounting X
" Fail open or closed" and " loss of air" are failure modes which are not, by their nature, applicable to this high pressure steam turbine.
The remaining failure modes, however, are credible under worst case dynamic loading conditions.
FAILURE EFFECT:
A.
Effect on Primary Use Failure could cause flow rate of emergency cooling water to be less than required to maintain reactor core temperatures within specification limits under high pressure accident conditions.
B.
Secondary Effect None DISCUSSION AND CONCLUSION:
A HPCI turbine, similar to the Shoreham turbine, has recently been subjected to a dynamic test which envelops Shoreham requirements (see Figure 1).
The tested turbine performed within specification 1 of 3
r limits both during and after the dynamic test events.
- However, certain modifications will be made to the Shoreham turbine to make it conform exactly to the tested turbine.
Operation of the Shoreham Plant prior to completion of these modifications poses no safety hazard because a redundant, single failure proof equipment r.ath exists which provides safe shutdown without HPCI for those accidents for which HPCI is designed to operate.
This alternate safe shutdown path to provide core cooling would be accomplished by vessel depressurization through ADS coupled with the low pressure LPCI mode of RHR operation and/
or core spray.
The LPCI and core spray systems are redundant to one another.
Either system, by itself, can provide adequate core cooling under all postulated accident conditions.
Sections 6.3.2.2.2, 6.3.2.2.3 and 6.3.2.2.4 of the FSAR describe 1
operation of this alternate shutdown path.
In addition, Sections 7.3.1.1.2, 7.3.1.1.3 and 7.3.1.1.4 of the FSAR discuss the con-trols and instrumentation in detail.
Section 6.3.2.2.2 summarizes as follows:
"In case the capability of the reactor feed pumps, control rod drive pumps, RCIC System, and HPCI System is not sufficient to maintain the reactor water level, the ADS functions to reduce the reactor pressure to a value low enough to allow the LPCI and core spray systems to pump water to the reactor vessel in time to cool the core consistent with the design bases.
To ensure proper cooling of the core under all circumstances, the ADS meets the single failure criteria.
This ensures that the LPCI and core spray systems can be actuated with a HPCI failure and one additional single failure."
Section 6.3.3.7 of the FSAR discusses the worst case small break accident assuming HPCI failure.
This analysis determined that peak clad temperature remained well below the Appendix K limits utilizing the alternate shutdown path described above.
The combination of ADS, LPCI, and core spray systems ensures that a single failure proof equipment path exists which provides safe shutdown without HPCI for those accidents for which HPCI is designed to operate.
2 of 3
NEDE-3012J GENERAL ELECTRIC COMPANY
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y ~ ENCLOSURE lE JUSTIFICATIO' FOR INTERIM OPERATION N NAME: In-Vessel Rack 5MPumRK tb. i' F16-E006 /1F16*FAK-09 _ SAFETY FUNCTION: Support fuel bundles inside reactor during refueling. FAILURE MODES: Fail Open Fail Closed Loss of Power Loss of Air Loss of Pressure Integrity loss of Structural Integrity X Distortion of Mounting X The in-vessel rack is a passive frame type structure. It does not open, '_i close, use power and air, or retain pressure. The only failure modes that are credible under worst case dynamic loadings are loss of structural integrity and distortion of mounting. FAILURE EFFECT: A. Effect on Primary Use Loss of structural integrity, or dist'ortion of mounting, coul'd permit up to four fuel bundles to fall down upon the top of the reactor core during refueling. B. Secondary Effect [ None sj' DISCUSSION AND CONCLUSION: i The in-vessel rack is used during refueling only, as a convenient il in-vessel parking place for fuel bundles. It is not used during the initial fuel loading. Ample time is available before the first refueling outage to perfom the required nonlinear analysis to qualify ij the Shorcham in-vessel rack to the SQRT criteria. Even if it is not qualified by the first refueling outage, refueling could proceed without i the use of this rack. ~ Therefore, interim operation before the in-vessel rack is qual.ified l poses no safety hazard. RWH: 1m:im/13U-9 12/1/82 )( . = - - - - ~-3 -}}