ML20213E615
| ML20213E615 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 10/05/1983 |
| From: | Knight J Office of Nuclear Reactor Regulation |
| To: | Houston R, Muller D, Rubenstein L Office of Nuclear Reactor Regulation |
| References | |
| CON-WNP-0658, CON-WNP-658 NUDOCS 8310180024 | |
| Download: ML20213E615 (17) | |
Text
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, DISTRIBUTION OCT 5 1983
/ Central File Q QB Reading ~F'i1e MEMORANDUM FOR:
R. Wayne Houston, Assistant Director for Reactor Safety Division of Systems Integration Lester S. Rubenstein, Assistant Director for Core and Plant Systems Division of Systems Integration Daniel R. Muller, Assistant Director for Radiation Protection Division of Systems Integration FROM:
James P. Knight, Assistant Director Components & Structures Engineering-Division of Engineering
SUBJECT:
JUSTIFICATION FOR INTERIM OPERATION, WPPSS NUCLEAR PROJECT NO. 2
Reference:
Memo for R. Wayne Houston, etc. from James P. Knight dated July 29,1983; subject as above.
The Equipment Qualification Branch, Division of Engineering has received from WPPSS additional justifications for interim operation with safety-related electrical equipment which is not fully qualified for the harsh environment created by accidents.
Pursuant to the request for assistance made in the referenced memorandum, the attached justifications are being forwarded for review.
A written response is required by October 21, 1983 in order not to adversely affect the plant licensing schedule.
Those branches which have not responded to our previous request should,
address all the iustifications'in their response.
X4-8310180024 831005 1
4 ADOCK 05000g 7 OD James P. Knight, Assistant Director Components & Structures Engineering Division of Engineering
Attachment:
As stated I
cc:
See next page
Contact:
A. Masciantonio, NRR X28205
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E0JIPMENT JUSTIFICATION INDEX JIO tb.
EPN(s) 1 CIA-PROG-1A 2
CIA-PS-22A 3
CIA-RLY-21A 4
CIA-RLY-22A 5
CIA-TDS-1A~
6 C45-A7-1, 3 7
C4S-LE-3A, 38 8
C4S-RMS-HTP 71 9
OiS-TE-1, 2, 26, 27, 29, 30 10 CiS-TS-4A, 48, 4C, 40, SA, 58, SC, 50 11 EDR-P05-V/20 12 FDR-POS-V/4 13 HPCS-FT-5 14 LPRM-Oetectors 15 LPRM-Connectors MS-RE-3A, 38 16 4
17 ROA-POS-V/1 18 SRM-DET-10 19 SW-M0-187A, 188A 20 3W-PS-1014 21 SW-V-201,204,212,213 l
22 TIP-V-1, 2, 3, 4, 5 No $c :
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EOUIPMENT JUSTIFICATION INDEX i
i l
JIO No.
, EPN(s) -
23 CIA-SPV-1 A thru 15A 24 MS-LITS-26A t
25 RIE-V-75A,B Rev.2 26 RRA-RMS-FN/2 27 SRM-CONN-04 i
l 28 SRM-EAMP-10 1
ii I
(
e6 7358f j
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E0JIPMENT JUSTIFICATION #23 1.0 COMPONENT IDENTIFICATION EPN : CIA-SPV-1 A, 2A, 3A, 4A, 5A, 6A, 7A, 8A, 9A,10A, ll A,12A,13A, 14A, 15A
==
Description:==
0.5" Solenoid Pilot Valve on N2 Bottle Discharge Component Type:
Solenoid Valve Manufacturer /Model:
Marotta/MV252-3 20 ACCIDENT CONDITIONS
^
Temoerature Reiatiye Humidity Accident Profile:
- 4
- 21X Use Code:
1 Operability Time:
4320 Hours Radiation Zone:
R4410 l
Zone Cose:
8 2x103 Rads i
3.0 COMPONENT SAFETY FUNCTION These solenoid valves are mounted on their respective lines from the nitrogen bottles (CIA-TK-1A through 15A). These valves control the flow of nitrogen from the bottles to the Containment Instrument Air System.
These solenoid valves normally are in the closed position and their opening is sequenced by thh step programmer controller (CIA-PROG-1 A).
4.0 QJALIFICATION STATUS l
4.1 Sumary of Oualification Status The environmental qualification test program is still in progress.
These valves are qualified for all post-LOCA conditions by materials analysis.
However, they are not qualified for relative humidity l
resulting from a HELS.
Therefore, the following justification is provided.
i 6956f/1
- 4. 2 Parameter Recuiring Justification Humi dity.
5.0 JUSTIFICATION FOR INTERIM OPERATION The ADS is required to reduce reacter vessel pressure if the itish j
Pressure Core Spray (HPCS) is not maintaining the proper reactor vessel water level. This allows the Low Pressure Coolant Injection (LPCI) mode of the Residual Heat Removal (RER) System and the Low Pressure Ccre Spray (LPCS) System to provide make-up water to the reacter core.
ADS safety / relief valves can be actuated by either the Nitrogen Supply System, controlled by CIA-iROG-1 A, or by the charged air accumulator tank provided for each safety / relief valve (SRV).
In the event that the program controller fails, the accumulator tank is capable of providing at least one SRV actuation.
Once actuated, these SRV's will reduce the reactor vessel pressure so that the LPCS and LPCI systems can provide core cooling.
Failure of the solenoid pilot valves (SPVs) will not inhibit ADS operation. These SPVs are bi-stable devices; that is, they can only assume a fully open or closed position.
If they fail in the 'open position, they will provide the necessary path for the nitrogen to reach the ADS and maintain system operability.
If the SPVs fail in the closed position, then an alternate pneumatic supply is available from the remote nitrogen station.
Upon receipt of a low pressure indication in the Control Rcom from the nitrogen supply header pressure sensor, CIA-PT-21 A, the operator will manually initiate charging of the CIA system from the reste nitrogen bottle station (CIA-TK-20A). This station, located in the corridor 1
between the reactor building and the diesel generator building, is l
accessible under post-LOCA conditions.
t 6.0 CONCLUSION Interim operation is justified on the following basis:
I 1.
The SRV acumulator tanks provide alternate means to initiate reactor vessel depressurization to allow the LPCS and LPCI to function.
2.
The operater can manually charge the CIA system for an indefinite period of time frcm a remte nitrogen station, if subsequent ADS actuations are needed.
I 6956f/2
EOUIPMENT JUSTIFICATION #24 1.0 COMPONENT IDENTIFICATION EPN: MS-LITS-26A
==
Description:==
RPV Level Indicator Congonent Type:
Level Indicating Transmitter Switch Manufacturer /Model:
Barton/760 2.0 ACCIDENT CONDITIONS Temoerature Relatiye Htznidity keident Profile:
- 4
- 21X Use Code:
1 Operability Time:
4320 Hours Radiation Zone:
RS22K i
Zone Dose:
9.1x103 Rads i
(6 no. accident dose plus 10 yr.
normal dose) 3.0 COMPONENT SAFETY FUNCTION MS-LIT 3-26A provides the Control Room operator with the wide-range RPV level indication. This indication is used to verify that the top of the reactor core is covered withNater. Regulatory Guide 1.97 requires that this indication be provided as part of the plant's long-term post-accident monitoring capability.
4.0 00ALIFICATION STATUS i
ei l
4.1 Surmary of Qualification Status This transmitter switch is qualified for a post-accident pressure, tenverature and radiation environment.
Hcwever, qualification documentation is not available to demonstrate switch operacility during exposure to 100% relative humidity.
Therefore, a justification for interim operation is provided.
!1, 8337f/1
- 4. 2 Parameters Recuiring Justification Humidity 5.0 JUSTIFICATION FOR INTERIM OPERATION RPV level indication is most critical during a LOCA when the operatcr must concentrate on making up coolant to the RPV.
During a LOCA, the effects of relative humidity on +he Reactor Building are negligible (i.e., 657.) and operability of MS-LITS-26A can be assured.
During a HELB, the need for wide range monitoring is not as critical.
System isolation features will ensure that the reactor coolant pressure boundary and RPV water inventory is maintained under HELB conditions.
MS-LITS-26A also has a redundant, physically separated, and electrically independent backup, MS-LITS-260.
If MS-LITS-26A were to fail, the operate would be able to detect the error by conparing the two transt :tcr ications. The two redundant and separated instrument tr ain'. ;r ;v ir 2ssurance that a single failure will not mislead the operator ur caase loss of level indication.
In the event that both MS-LITS-26A and 260 fail, there are additional backup RPV level indicators. MS-LIIS-44A and 448 are capable of nonitoring water level in the reactor vessel. This capability satisfies the intent of Reg. Gaide 1.97 and provides a reliable means of recording water level following an accident.
Failure of the transmitter portion of MS-LITS-26A will not interfere with RPS logic operations since its signal only provides Contro. Room indication. The switch portion of MS-LITS-26A is isolated frem the RPS by fuse KlA and is not used as part of the logic scheme.
Tnerefcre any short-to-ground failure will not incapacitate the RPS.
l
6.0 CONCLUSION
Interim operation is just'ified on the following basis:
1.
Redundant level indicating instrumentation exists to assure long-term monitoring of RPV water level.
2.
Failure of the transmitting switch can not effect the RPS logic and hence has no adverse impact on any safety.related equipment.
l l
l l
8337f/2 l
EOUIPMENT JJSTIFICATION #25 1.0 COMPONENT IDENTIFICATION EPN:
RHi-y-75A, 8
==
Description:==
Solenoid Valve for RHR Water Sampling and Pressure Boundary Isolation Component Type: Valve Manufacturer /Model:
Marotta/MV36RP-H3 2.0 ACCIDENT CONDITIONS Temoerature Relative Humidity Accident Profile:
- 4
- 21X Use Code:
2 Operability Time:
4320 Hours Radiation Zone:
R548N, J Zone Dose:
3.1x106 Rads 3.0 CCMPONENT SAFETY FUNCTION Solenoid valves RHi-V-75A and 758 are used to sample suppression pool water taken from either loop of the RHR system. However, these valves are not part of the Post Accident Sar.pling System.
During accident conditions, this valve serves solely to maintain RHR pressure boundary and thus must remain in the closed position for 6 months following the accident.
4.0 QU ALIFICATION STATUS 4.1 Sumnary of Qualification Status The environmental qualification program is still in progress.
These valves are qualified for all post-UDCA conditions by materials analysis. However, they are not qualified for relative humidity resulting frem a HELS.
Therefore, a justification for interim operation is provided.
(
-4.2 Parameters Reauirino Justification Humi dity.
5.0 JUSTIFICATION FOR INTERIM OPERATION The only safety-related function of RHR.-V-75A is to maintain the RHR System pressure boundary.
It is not required for isolation of the primary containment, and long-term operation of the valve is not needed.
These valves are normally closed during operation and opened only when a sample is drawn.
In the rare event that an accident occurs simultaneous to the opening of these valves, an automatic isolation signal will immediately close the valves.. This will enable the valves to assune their fail-safe position long before a post-accident environment is experienced.
If these valves fail to close when required, a second set of series isolation valves (RHR-V-60A and 608) would provide a redundant means of maintaining pressure boundary.
In the event that the operator fails to close this solenoid valve, pressure boundary of the RHR system is still maintained by the sample lines and downstream sample station sampling valve. Since there is no direct path for the sample fluid to leak into the Reactor Building, and i
use sample station is not accessible following an accident, failure to close RHR-V-75A will not impair safe shutdown.
6.0 CONCLUSION Interim operation is justified on the following basis:
1.
RHR-V-75A is normally closed.
This is the fail-safe position for maintaining RHR pressure boundary.
2.
RHR-V-75A is opened only fcr the short duration when a sample is being drawn.
In the event that the valve is not returned to the closed position, pressure boundary is still maintained by the sanple lines and downstream lample station sampling valve.
7071f/2
EQUIPMENT JUSTIFICATION #26 1.0 COMPONENT IDENTIFICATION EFN: RRA-RMS-FN/2
==
Description:==
Centrol Switch for RRA-fN-2 Component Type:
Remote Manual Switch Manufacturer /Model:
Square D/KYC-1 20 ACCIDENT CONDITIONS Temperature Relative Htnidity Accident Profile:
- 4
- 21X Use Code:
2 Operability Ti.me:
4320 liours Radiation Zone:
R441G Zone Dose:
9.9x105 Rads 3.0 COMPONENT SAFETY FUNCTICN This cogonent is not required to perform an active safety function following a design basis accident. The switch is used for n:anual testing of the air handling unit fan, RRA-FN-2.
4.0 OUA!. FICATION STATUS
=
4.1 Sucrary of Qualification Status Tne switch is qualified for all post-LOCA conditicns.
- However, it is not qualified for the relative humidity folicwing a HELS.
Therefere, the following justification for interim operation is provided.
- 4. 2 Parameters Recuirine Justification Hurr.idity e
8336f/1
5.0 JUSTIFICATION FOR INTERIM GPERATION This remote manual switch can assume three positions, namely "On-Auto-Off."
The switch assembly consists of one set of normally open centacts and one set of normally closed contacts. When the switch is in the " Auto" position, the set of.ncrmally closed contacts remain closed and the set of normally open contacts remain open. When the switch is in the "On" position, the set of normally open contacts are closed and the set of normally closed centacts are open. When the switch is in the "Off" position both sets of contacts are in the open position. RRA-RMS-FN/2 is ncrmally in the " Auto" position (i.e., normally closed contacts).
In the AJto mode, RRA-FN-2 is interlocked with the Residual Heat Removal (RM) Pump 2A and is automatically started with initiation of the R$
Pump. There are two possible switch failure nodes cue to 100% humidity.
l.
Shorting of the normally closed contact (primary failure mode).
2.
Shorting to the cround through the switch enclosure (secondary, infrequent failure mode).
Shorting across the normally closed contacts does not have any adverse effect because the circuit will assume its intended configuration.
Shorting to the ground through the switch enclosure is highly unlikely due to low voltage (120 V AC) and large air gap between the contacts and switch enclosure. Additionally, the switch contacts are housed in a waterproof NEMA Type 4 enclosure. The waterproof construction will protect the switch from the high humidity environment.
2 auld an accident occur simultaneous to fan testing -(i.e., the switch is in the "On" position or closed) the circuit configuration will be in its intended state.
In the rare event that a short-to-ground does inhibit RRA-FN-2 operation, only one RHR pump room will be affected. Therefore failure of the switch will not fail the R$ system since there are two redundant R$ trains.
6.0 CONCLUSION
Interim operation is justified on the following basis:
1.
RRA-EMS-FN/2 is enclosed in a waterproof housing wnich will prevent moisture from affecting the switch.
2.
The only failure mode with potential adverse impact has a low probability of occuring. That is, a snort-to$round through the switch enclosure cannot occur due to the low voltage and 1arge air gap between the contacts and switch enclosure.
3.
If a short-to-ground does take place, only one RHR pump is impacted.
Since there are two trains of the RHR system, safe shutdown of the plant is not compromised.
8236f/2
EQUIPMENT JUSTIFICATION #27 1.0 COMPONENT IDENTIFICATION EPN: sri-CO NN-04
==
Description:==
Electrical Connector for SRM detector SRM-DET-lO Component Type:
Electrical Connecter Manufacturer /Model: GE/ PPD # ll485388G004 2.0 ACCIDENT CONDITIONS
[
Temperature Rel ative Hunidity Accident Profile:
- 1
- 2 Use Code:
2 Operability Time:
4320 Hours Radiatioa Zone:
Containment Zone Dose:
7x107 Rads 30 CCNPONENT SAFETY FUNCTION This SRM connector is part of a system which monitors reactor core power between 10-6% and 10% of full power.
The system is required to assure that reactivity has been controlled during an emergency shutdown. The connector is an electrical junction for the SRM detector signal cable.
The SRM System is required by Reg. Guide 197 to provide long-term post-accident monitoring capability.
4.0 QUALIFICATION STATUS 4.1 Sunnary of Oualification Status Qualification data is not available.
Design data is per GE specification. Therefore, a justification for interim operation is provided.
- 4. 2 Parameters Recuirino Justification Pressure, temperature, humidity, and radiation dose I
8344f/1
5.0 JUSTIFICATION FOR INTERIM OPERATION The SRM system is required by Reg. Gaide 1.97 to verify that a SCtAM has been successfully accomplished.
The intent is to ecnfi.m that reactor core activity has been effectively controlled during an emergency snutdown.
Four independent SP!4 channels monitor icw power ccre activity.
For the SRM system to fail, all four of these channels would have to fail.
Each SRM channel has an individual connector which serves as an electrical cable junction for the retractable detector.
These ccnnectors have metallic screw fittings on both ends which seal the detector from the harsh drywell environment. Per GE drawing 1148S888, the non-metallic materials are composed of alumina (ceramic), rexolite and polyethylene.
Each material is resistant to a radiation threshold of at least lx108 rads.
This is well above the mest conservative integrated dose for a pcst-LOCA environment.
l The small guide sleeve portion of the connector is compcsed of teflon and is susceptible to radiation.
However, review of GE documents indicates that degradation of the teficn sleeve will not interfere with detec:cr retraction /inserticn.
Furthermere, the sleeve is not used to isolate the detector dry tube frcm the containment atmosphere. Therefore the connecter's safety function is not compromised.
In the rare event that all four SRM channels fail, alternate means of confirming reactivity centrol are available. The Full Core Display Rod Position Indication. the Process Computer Red Position Log and the Select Rod Position Display provide the Centrol Rocm operator with a verification that the SCRAM has been accocolished.
In addition, the APRM and IRM downscale displays give indications of icw power core activity.
The Regulatory Guide 1.97 requirements fer these ecmoonents have been evaluated in depth. The Supply System's positicn en this issue has been transmitted to the NRC in Docket No. 50-397, the letter from Mr. G.C.
Sorensen to Mr. A. Schwencer addressing "Nelear Project tb. 2 Source Range Mohitor Qualification, RE: Regulatory Guide 1.97".
6.0 CONCLUSION
4 4
Interim operation is justified on the following basis:
1.
A materials analysis of the SRM connecters based en the GE design specification indicates that tne connectors will not fail due to post-accident conditions.
i 2.
SRM failure would require the failure of all four independent SRM t
channels at the same time, which is very unlikely due to GE design l
considerations.
3.
In the event of SRM connec:cr (and system) failure, alternate SCRAM verification instrumentation is available.
I 4.
This model of connector has been supplied by GE and is being used en operating SWR's.
8344f/2
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EOUIPMENT duSTIFICATION #28 i
10 COMPONENT IDENTIFICATION EPN: SM-EMP-10
==
Description:==
Pulse Pre-amplifier for SM-DET-10 Component Type:
Electrical Signal Pre-amplifier Manufacturer /Model: GE/112C2276G001 20 ACCIDENT CONDITIONS Temperature Relative Hunidity Accident Profile:
- 4
- 21X Use Code:
1 Operability Time:
4320 Hours Radiation Zone:
R501K Zone Oose:
1.4x104 Rads 3.0 COMPONENT !AFETY FUNCTION SM-EMP-10 pre-amplifies the electric signal from SM-DET-10.
This signal is used to monitor reactor low power range between 10-6% to 10%
of full power. This pre-amplifier is part of the equipment required by Reg. Guide 1.97 to provide long-term post-accident monitoring capability, t
to 4.0 OUALIFICATION STATUS 4.1 Sumary of Qualification Status Qualificaticn data is not available.
Design data is per GE specificaion. Therefore, a justification for interim operation is provided.
4.2 Parameters Reauirino Justification Pressure, temperature, humidity and radiation dose.
i l
l 8339f/1
5.0' JUSTIFICATION FOR INTERIM OPERATION
(
The SRM system is required by Reg. Guide 1.97 tc ccnfirm that reactor core activity has been effectively controlled during an emergency shutdcwn.
The SRM system performs this function by verifying that a scram has been successfully acccmplished.
Four independent SRM channels monitor icw power ccre activity.
In crder to fail the SRM system, all four of these channels wculd have to fail.
Eacn SRM channel has an individual pulse pre-amplifier whicn transmits a power level signal from one of the in-core neutron detectors to tne Control Rocm.
It is highly unlikely that all four channels (or that all four pulse pre-amplifiers) f ail at once.
The pre-amplifiers are physically located in different enes of the Reacter Building and will therefore not experience the same accident environment.
The requirement to monitor lcw power core activity is most critical during a LOCA.
It is essential that the operatcr assure that a SCRAM has been successfully implemented by the time the ECCS are operating. The effects of a LOCA on this pre-a@lifier will be well within GE design limits.
HELS accidents will create less severe conditicns in the pre-a@lifier lccation than a LOCA, with the exception of relative humidity.
However, the anticipated effects of the HELS will have minimal i@act on system performance because of GE design specifications.
The six month post-accident radiation dose experienced by the pre-a@lifiers is 1.4 x 104 rads, which is almost mild.
Since the
(
pre-amplifiers will provide verif.ication of a successful scram within minutes folicwing an accident, they will have performed their intended function icng before receiving a significant radiation dose.
GE document GEX 9871 states that the pre-amplifiers are designed to operate for a sustained period at 166*F (60*C). This is censiderably above the worst case post-accident temperature (130*F) experienced by any of the pre-amplifiers.
Similarly, this equipment is designed to operate for relative humidity conditions of at least 98%.
The anticipated
~
relative humidity in the Reacter Building due to a LOCA is only 65%;
therefore, operability for LOCA is assured.
If tne relative humidity exceeds 98% (during a HELB) the protective casing would ensure system operability.
In the rare event that all four SRM channels fail, alternate means of confirming reactivity centrol are available.
The Full Ccre Display Fod Position Indication, the Prccess Co@ uter Rod P4sition Lcg and tne Select Rod Position Display provide the Centrol Room operatcr with a verification that the SCR/M has been acccmplished.
In accition, the APRM and IRM downscale displays indicate low pcwer reacter ccre activity.
I
The Regulatory Guide 1.97 repuirements for these components nave been evaluated in depen.
The Supply System's position on this issue has been t,
transmitted to the NRC in Docket No. 50-397, the letter frca Mr. G.C.
Sorensen to Mr. A. Senwencer accressing " Nuclear Project No. 2 Source Range Monitor Qualification, RE: Regulatcry Guide 1.97".
6.0 CONCLUSION
Interim operation is justified on the following basis:
1.
Failure of all four SRM pre-amplifiers is very unlikely due to desigi and physical separation of the equipment.
2.
In the event of SRM system f ailure, alternate SCRAM verification instrumentation is available.
3.
This preamplifier has been supplied by GE and is being used on operating BWR's.
6 0
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-