ML20209H835
| ML20209H835 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 04/30/1987 |
| From: | Johnson B CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NUDOCS 8705040131 | |
| Download: ML20209H835 (43) | |
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Consumers
@rewramus ASKMAN'E PREGRE55 Power oeneral officos: 1945 West Parnail Road, Jackson, MI 49201 . (517) 788-055o April 30, 1987 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT -
1986 ANNUAL REPORT OF FACILITY CHANGES, TESTS AND EXPERIMENTS Attached is Consumers Power Company's Annual Report describing the Facility Changes (FC), Specification Changes (SC), tests and experiments performed at the Palisades Plant during 1986. This report is submitted in accordance with the provisions of 10CFR50.59(b). A delay in the submittal had been requested per telephone conversations with NRC Washington prior to the normal March 31 due date. The report is divided into the following five areas:
- 1. Facility Changes, Specification Changes, Tests and Experiments Declared Operable in 1984, Closed in 1986.
- 2. Facility Changes, Specification Changes, Tests and Experiments Declared Operable in 1985, closed in 1986.
- 3. Facility Changes, Specification Changes, Tests and Experiments Declared Operable in 1986, closed in 1986.
- 4. Facility Changes, Specification Changes, Tests and Experiments Declared Operable Prior to December 31, 1986, Package Not Closed Out.
- 5. Open Facility Changes and Specification Changes Reportable on the Annual Report, Not Declared Operable or Closed Out, m ~ w c)
Brian D Johnson Staff Licensing Engineer i
CC Administrator, Region III NRC NRC Resident Inspector - Palisades Attachment ,.[ l OC0487-0050-NLO2 8705040131 870430 PDR ADOCK 05000255 R PM /
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FACILITY CHANCES, SPECIFICATION CHANCES, TESTS AND EXPERIMENTS j DECLARED OPERABLE IN 1984 i
CLOSED IN 1986 i
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I Page 2 Palisades Plant j~ Facility Changes, Specification Changes Tests and Experiments
- Declared Operable in 1984 Closed in 1986
- FC-459-2 i
This Facility Change covered the removal of the Urea Formaldehyde Liquid Rad j Waste Solidification System as part of the Rad Waste Extruder modification.
SAFETY EVALUATION
SUMMARY
i j The Ursa Formaldehyde System was not in use at the time of removal. The prob-l ability of leakage from any of the related process piping was no greater after j the removal than before the removal. The Auxiliary Building will have the same j'
j type of liquid waste materials af ter the equipment is removed as before. The ability of the facility to contain the liquid in the event of a rupture will not
! be impacted by this removal. .
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! TL.refore the probability of occurrence or the consequences of an accident or l' malfunction of equipment important to safety either previously evaluated in the ,
FSAR, or of a different type, is not increased. Since this Facility Change covers !
l removal of significant pieces of equipment including Urea Formaldehyde Tank j support beams, and does not encompass any modification to floors or walls, etc, which constitute the main structure of the Auxiliary Building, the loads applied to the members are expected to be less, such that the margin of safety should
- increase.
4 i FC-516-2 i
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This modification covered the addition of a third pump on the Auxiliary Feedwater System, and an automatic isolation system. **The addition of a third auxiliary feedwater pump located in a separate room, single failure modes identified in a reliability study, prohibiting auxiliary feedwater flow to the steam generators i were eliminated, increasing the system reliability. Automatic isolation and actu-ation eliminates operator actions for system initiation or isolation during an i accident situation. Therefore, the probability of occurrence or the _ consequences of an accident or malfunction of equipment important to safety previously evaluated
[ in the FSAR is decreased.
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! the third pump was not taken credit for during an accident condition with loss I
of off-site power. Removing the third pump from the HPSI system decreases the system flexibility, but not its safety function. The margin of safety defined ,
in the Technical Specifications can still be met by the two remaining pumps.
1 The margin of safety for the Auxiliary Feedwater System is increased due to the j increase in system reliability.
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- The third HPSI pump was removed and used as the third auxiliary feedwater pump.
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Page 3 FC-568 This modification installed a second channel CCW Surge Tank Level Indication.
This new level indication will activate the same annunciator window as the existing level indicator. This system is redundant to the existing level indicating system, and is seismically qualified to IEEE-344, 1975.
SAFETY EVALUATION
SUMMARY
1 The new system provides an additional level indication circuit for monitoring the CCW Surge Tank. The CCW Surge Tank level is an important parameter to measure because it gives anticipatory indication of a possible condition of loss of CCW water, which might lead to a loss of system function. A redundant system enables an operator to compare readings and take appropriate actions if necessary and reduces the possibility of misleading indication.
1 The two circuits will be powered from different power sources and cabling routed in physically separate raceways, reducing the possibility of a common mode fail-ure. It will be powered from a preferred AC Power Bus.
Because of this modification, the probability of occurrence of an accident or malfunction of equipment important to safety previou' sly evaluated in the FSAR is not increased.
FC-570 This modification covered the installation of two spray deflectors for each of the Service Water Pump Motors. They were designed to protect against possible spray from a break in other piping in the area. Such sources could be circulating water, dilution water, and fire protection water systems. The deflectors protect against a vertical downward spray, a horizontal spray and to some extent a vertical upward spray.
SAFETY EVALUATION
SUMMARY
This modification provides additional protection to the service water pump motors.
Disabling the service water pumps in the intake structure is a common mode failure that could eliminate the ultimate heat sink (Lake Michigan). The ability to cool vital components within the plant, such as the reactor coolant pump seals, by the Component Cooling Water System, which is cooled by the Service Water System, could ,
be affected, and the ability to perform a safe shutdown would be seriously impaired.
If the reactor coolant pump seals failed because of inadequate cooling, the Primary Coolant System boundary could be violated and a LOCA could occur.
I Also, cooldown of the plant, Hot and Cold shutdown, depends to a great extent on the Service Water System. Attainment of Ilot Shutdown depends on the Auxiliary Feedwater System whose water source is the condensate storage tank. In case the condensate storage tank is disabled, the service water or fire pump in the intake structure may have to be used. In addition, service water is needed to attain cold shutdown because the regenerative heat exchangers are cooled by component cooling water. The addition of spray deflectors will protect against possible spray from a break in other piping systems in the area and reduce the possibility of a conson mode failure.
MIO387-0046A-TC07
s Page 4 Because of this modification, the probability of occurrence of an accident impor-tant to safety previously reported in the FSAR is not increased, and the margin of safety as defined in the basis for any technical specification is not reduced.
FC-576 This modification was made to install a manual isolation valve on the outside of containment on Containment Penetration #33. This valve was installed to increase the reliability of the penetration, reducing the number of isolation valves from seven to two.
SAFETY EVALUATION
SUMMARY
The installation of the manual valve will decrease the probability of occurrence and the consequences of an accident, and is essentially upgrading the penetration in case of an accident. The modification will increase containment isolatability, therefore the margin of safety as defined in the basis for any technical specifi-cation is not reduced.
FC-593 Under this Facility Change the refueling machine bridge cable festoon was modi-fied. The existing rigid towing bar between the bridge and the wheeled carrier at the end of the festoon was replaced with a spring assembly. The spring assembly includes a limit switch to provide an alarm if the towing load is excessive.
Strain relief cables were added in the festoon itself so that if a jam were to occur the power cables would not be loaded.
SAFETY EVALUATION
SUMMARY
This modification was designed to prevent jamming of the refueling machine bridge cable festoon. It does not increase the probability of occurrence or the conse-quences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR.
FC-613 This modification was performed to modify the Auxiliary Feedwater Nozzle within each steam generator to remove the auxiliary feedwater sparger rings, and install a thermal liner assembly with an inverted J-tube at the nozzle discharge into the steam generator.
SAFETY EVALUATION
SUMMARY
The auxiliary nozzle thermal liner discharge system replaces the previously in-stalled sparger system. The new line system will perform functionally in a similar manner to the previously installed system.
The probability of occurrence of an accident or malfunction of equipment important to safety would remain the same or be decreased relative to that which existed with no sparger system. An ASME Section III Code type evaluation of the liner discharge system was completed by Westinghouse prior to declaring the steam gene-rator operable to demonstrate the adequacy of this component.
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Page 5 SC-84-024 Under this Specification Change the electrohydraulic personnel air lock door closure units located inside containment near the air lock, and inside the air lock itself, were removed. The units were designed to allow personnel to close opposite end air lock door from the other exterior end of the air lock. The electrical components and hydraulic tubing lines were also removed. Parts for the originally installed equipment are totally obsolete, and would have to be replaced in their entirety. Functional benefits of remotely operated electro-hydraulic door mechanism is negligible.
SAFETY EVALUATION
SUMMARY
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
The elimination of the door closure units have no impact on increasing the con-sequences of an accident as shown below.
The electrohydraulic units are useful only for closing the auxiliary building side door from inside containment, or for closing the containment side door from the auxiliary building side of the air lock. There is little need for these electrohydraulle door closure units because of other means available to close the opposite side door, including the use of phones located on both sides of the air lock to communicate the need for door closure and also the use of the emer-gency air lock for alternate personnel access.
It should be noted, however, there is a little chance that an air lock door will be inadvertently left open. During plant operation, access through the personnel air lock into containment is limited and is monitored by the control room, a re-quirement of the technical specification testing (DWO-13), required within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of any personnel air lock door opening. This control helps to prevent in-advertent use of the air lock, decreasing a chance for leaving the door open.
Aspects of the air lock design itself help to ensure personnel closing doors be-hind themselves. These include door interlock system, door closure lights, and
" hump" design during handwheel operation of the door. Overcoming " hump" during door closure signifies to personnel that the door is closed and sealed.
1 SC-84-032 I
l This Specification Change was initiated to provide a more accurate indication of the component cooling water Surge Tank (T-3) level during all modes of operation.
The flange mounted level transmitter was referenced to the atmosphere. Under this change it was modified to reference tank pressure.
SAFETY EVALUATION
SUMMARY
Under normal operating conditions, the component cooling water surge tank is vented to atmosphere. However, if a high radiation signal is received from the component cooling water radiation monitor, the tank becomes exposed to a nega-tive pressure. LT-0917 was formerly referenced to atmosphere and would not MIO387-0046A-TC07
t Page 6 respond correctly under emergency conditions. This modification allows LT-0917 to respond correctly to tank level changes under both normal and emergency conditions. Thus, the probability of occurrence, or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
SC-84-033 This Specification Change war initiated to provide a more accurate indication of component cooling water surge tank (T-3) level during all modes of operation.
The flange mounted differential pressure level transmitter was originally refer-enced at 95.6 of tank level. Under this change it was modified to reference tank pressure at a point above 100% tank level.
SAFETY EVALUATION
SUMMARY
The upper sensing line of LT-0920 was located at 94.6% of tank level. The purpose of this modification provides accurate level indication of the tank over its entire span, which was accomplished by relocating the sensing line to a point over 100% of tank level. This modification does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR.
SC-84-089 This Specification Change covers the installation of rubber type shims in areas where needed on the personnel air lock door compression sesis (cross section of 1-1/8" x 1-3/16"). The shims are to be located at the base of the door seal slot.
The gasket slots are oversized in some areas, and the use of rubber type shims help fill up the slot volume so that the personnel air lock door gaskets can be compressed. Compression of gasket creates gasket seal at the sides of the slot, and also prevents metal to metal contact of the air lock door and frame during latch bracket adjustment.
SAFETY EVALUATION
SUMMARY
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased because the use of the rubber type shim helps to provide an effective seal for the personnel air lock doors. These shims are especially required at areas where the gasket slots are oversized.
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DECLARED OPERABLE IN 1985 ,
CIASED IN 1986 l
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l Palisades Plant Facility Changes, Specification Changes Tests and Experiments Declared Operable in 1985 I FC-445-2 l This modification covered the installation of motor operators in the MSIV bypass valves. This allows selected operation of these valves from the main control
[ room rather than at the local valve position.
i l SAFETY EVALUATION
SUMMARY
The MSIV bypass valves are not specifically addressed in the FSAR but are part of the main steam system and designed to Class 2 requirements. The modification I
l does not af fect the mode of operation f rom manual handwheel to motor actuated.
, The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
The changes will not affect the valves ability to be operated, although they will be primarily motor operated. In the event of a motor operator malfunction or loss of power, the valves will still have a manual handwheel from which emergency oper-ation can be achieved. Therefore, the possibility of an accident or malfunction of a different type than previously evaluated in the FSAR is not created.
The technical specifications require closure of the MSIV's in five seconds for a main steam line break accident. This modification will not affect closure time on the HSIV's. Thus the margin of safety as defined in the technical specifica-l tions is not reduced.
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FACILITY CHANCES, SPECIFICATION CHANCES, TESTS AND EXPERIMENTS DECLARED OPERABLE IN 1986 CthSED IN 1986
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Page 2 Palisades Plant 1986 Facility Changes, Specification Changes Tests and Experiments Declared Operable in 1986 FC-419 This modification was performed so that the pressure on T-102 (Hydrazine Tank) would be increased to 10 psig. This was accomplished by installing hand switches and solenoid valves which close on a CHP.
SAFETY EVALUATION
SUMMARY
The pressure increase allows hydrazine to be injected because there is no head differential between T-102 and the Safety Injection Refueling Water Tank. The increase in head also allows T-102 to drain into tLe SIRW Tank if control valves fail open. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
A possibility of an accident or malfunction of a different type than any evaluated in the FSAR is not created due to the fact that if T-102 drained too fast (shut off valve on nitrogen failed open) there would be a greater initial concentration of hydrazine, which would taper off quicker, but would be pi ad up again, and be more evenly distributed when containment sumps start.
The margin of safety as defined in the basis of technical specifications is in-creased because of the faster injection of hydrazine.
FC-459-6 I
This modification covered the installation of a new load center and transformer and it's termination of wiring to the asphalt extruder system.
SAFETY EVALUATION
SUMMARY
The load center will be used to supply station power to the asphalt extruder system.
The equipment is not safety related.
FC-459-7 A number of required civil related modifications to the structure of the asphalt extruder system building are covered by this Pacility Change includings a) Removal of part of the Mezzanine and its structural supports.
b) Removal of nonstructural walls and partitions.
c) Removal of stairs. .
d) Holes cut through reinforced concrete walls for pipes, cables, HVAC and doors.
e) Removal of blocks for door openings and miscellaneous penetrations.
f) Hiscellaneous floor and wall core borings for various penetrations, tie-ins and relocations.
r) Roughing of concrete floors and walls, installing anchors / dowels, installing rebar, pouring concrete pads, and installing partition / shield walls.
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a Page 3 SAFETY EVALUATION
SUMMARY
The structure modifications were subjected to rigorous structural analysis.
The results were compared with design allowable acceptance criteria and assure that design af ter modifications comply with required design allowables. Thus, the probability of occurrence, or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
The incident loads which the structure must-resist after modification are the same as prior to the modification, therefore no new failure types are created.
FC-459-9 -
Miscellaneous electrical installations for radwaste extruder project was covered by this Facility Change.
SAFETY EVALUATION
SUMMARY
The probability of occurrence or the consequences of an accident or malfunction l of equipment important to safety previously evaluated in the FSAR was not increased because of the modification. Appropriate construction and design controls were implemented to assure the system would not create any new accident or malfunction scenarios.
FC-459-10 This Facility Change covered miscellaneous extruder mechanical process equipment installations for the radwaste solidification project.
SAFETY EVALUATION
SUMMARY
The function of the equipment installed under this project is non-Q, and as such any defects or failures thereof should have no impact on the health and safety of the public. All installations were covered by plant reviewed and approved con-struction work packages to assure proper installation details were incorporated.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. I I
No new type of accidents or malfunctions were created as a result of this modifi-cation. ;
FC-459-11 This Facility Change covered the HVAC equipment installed for the Asphalt Ex-truder System Building. The equipment installed is functionally non-Q, in that ,
HVAC is not required to operate under any modes of reactor operation. However, {
HVAC interfaces with Q components.
l SAFETY EVALUATION
SUMMARY
The complete failure of the HVAC System will not degrade the radweste building with respect to Seismic Class I criteria and will not endanger the health and MIO387-0046A-TC07
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- i Page k welfare of the public. There is no risk of off-site dose due to the failure of this system.
The safety related interfaces aret
- 1) The HVAC attached to the Seismic Class I Radwaste Auxiliary Building
- 2) HVAC will penetrate the structure and will require appropriate fire barrier protection.
- 3) The HVAC attachment to the structure may be by way of welding.
l The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
FC-514-1 This Facility Change replaced the existing containment area radiation monitoring system with an environmentally qualified system. The replacement system is qual-ified in accordance with NUREC 0588 and IEEE Standard 323-1974, and seismically qualified per IEEE 344-1975. The new readout / alarm modules are located in control panel C-11 and provide isolation signals on containment high radiation and local alarms and sealed sources.
SAFETY EVALUATION
SUMMARY
j This major modification involwd the replacement of the containment area radiation monitoring system, which provides the same function as the one it replaced, and l is environmentally qualified, and provides inputs to the Critical Function Moni-
' toring System. The replacement system does not feature internal radiation check sources; however, sealed sources have been mounted externally on the detectors to provide a sufficient background signal to prevent an alarm on loss of detector signal. Each of the four instrument loops is to be continuously monitored for circuit f ailure which will alarm on loss of power, loss of high voltage or loss of signal from the detector. The replacement system is capable of being checked electronically from the C-11-5 panel which only verifies proper operation of the amplifier circuitry and not the detectors and cable.
The probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased since the continuous circuit monitoring feature of the new system provides a higher probability of detecting equipment failure than the monthly use of the radiation check sources installed in the existing detectors. The installation of background sources to RE-1806, RE-1807 and RE-1808 does not increase the prob-ability or consequences of an accident.
l With the exception of not having internal radiation check sources and the instru-l ment scale being 10 2 R/hr to 10 4R/hr, the replacement system will provide the same l function as the one it replaces. The failure trip monitoring circuitry ensures i that the possibility of an accident or malfunction of a different type than eval-l uated previously in the FSAR is not created.
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)' An event in which the externally mounted sealed source is detached from the detector would ' result in a lower indication and would possibly provide a circuit
- failure light. While this would move the detector farther from the alarm and trip points, the 1% shift in output would not result in any significant change
- in the safety analysis.
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The new system does not have a radiation check source but- does have an equiva-i . lent loss of signal detection circuitry. The change in radiation level is not
} significant and does not reduce the margin of safety.
4 FC-563
- This modification installed a power operated valve in series with CV-2083 outside j containment on the primary coolant pump controlled bleed-off line adjacent to k
penetration #44. The project included valve procurement, installation, seismic j analysis, Electrical Equipment Qualification for mild LOCA environment, piping supports, control switches and indication lights in the control room, and mis-i cellaneous cabling and hardware.
j NUREC 0820, Palisades 1A Report, dated April 1982, based on Palisades SEP Topic '
l VI-4 committed the plant to install a second remotely operated valve in series
- with CV-2083 during the 1983 refueling outage.
i j SAFETY EVALUATION
SUMMARY
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] The valve installed is remotely operated from the control room or closed by a j
Containment Isolation Signal and is used in place of a manual valve already in the line. This remote /CIS closing capability will' decrease the probability of i
occurrence and the consequences of an accident and is essentially upgrading the penetration in case of an accident. Installation and valve specifications are j comparable to or better than those for CV-2083.
The probability of occurrence or the consequences of an accident or malfunction- ,
- of equipment important to safety previously evaluated in the FSAR is not increased due to this modification.
! FC-564 I l
l Under this Facility Change a new auxiliary hot shutdown panel (C-150A) was installed in the south penetration room (250) includingt i 1) Start up source range flux monitor channel j
- 2) SIRW Tank level indication
). 3) Primary system hot leg and cold leg temperature indication j 4) Steam Generator pressure indication i
j SAFETY EVALUATION
SUMMARY
! The intent of this new panel is to provide additional monitoring capabilities
'i to the existing system (C-150 panel) which is important to the safe shutdown of the plant during a fire which disables the present control ' room. This modi-4 1
MIO387-0046A-TC07 i
Page 6 fication does not affect the functional design or the operation of any equipment important to safety. Questions with respect to seismic loading of Q-structures, loading of Class IE electrical systems or raceways were evaluated prior to per-forming the individual jobs. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evalu-ated in the FSAR is not increased. Because none of the equipment specifically identified on the new panel is covered under the present technical specifications, the margin of safety as defined in the basis for any technical specification is not reduced.
FC-638 This modifica' tion is in response to NUREC 0737, Item II.K.3.25 to provide cooling water to the Primary Coolant Pump after the loss of offsite power. The modifi-cation shall perform analysis and change wiring as necessary to add the CCW pump motors to the normal shutdown sequencer (NSD). This analysis required a change in the setting on Auxiliary Feedwater Pumps P-8A and P-8C on the NSD sequencer from 19 seconds to 45 seconds.
SAFETY EVALUATION
SUMMARY
This modification adds the CCW pumps to the normal shutdown sequencer (NSD).
Spare contacts in the designed sequencer were used and no additional changes made. The channel of each contact matches the channel of the CCW pump which it is wired to. A transient load study was performed to show that the startup power system is capable of handling these additional loads. This ensures the integrity cf the reactor coolant pump seals during a loss of offsite power event.
In order to assure cooling to the RCP seals, the CCW pumps will be added to the l normal shutdown sequencers. The pumps will start automatically upon a loss of offsite power, thus omitting the potential for RCP seat failure and the subse-quent loss of PCS integrity. An analysis was performed to show that if the A and B pumps are started at 20 seconds and the C (backup) pump at 37 seconds, the transient and dynamic results will be acceptable. In order to ensure, accepta-bility the two motor driven auxiliary feedwater pumps (P-8A and P-8C) have been changed to start at 45 seconds. Although this is changed from the previous 19 seconds it will have no effect on any of the auxiliary feedwater analysis as t
credit has always been taken for a sequencing start time of 45 seconds. The auxiliary feedwater pump start times on the DBA sequencers are at 45 seconds.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
All worst scenarios were analysed for diesel generator loading combinations, therefore the possibility of an accident or malfunction of a different type than evaluated previously in the FSAR is not created.
I Tne margin of safety is not reduced. Starting the component cooling water pumps will reduce the possibility of loss of primary system integrity through the PCS pump seals. Starting the AFW pumps at 45 seconds is within the system analysis.
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FC-639 This Facility Change covered the installation of isolation switches in breaker l cubicles 152-103,107 and 110 on the 1C switchgear, and in the gauge panel and control panel for the 1-1 emergency diesel generator.
SAFETY EVALUATION
SUMMARY
This project marginally increases the probability of occurrence or the conse-quences of an accident to equipment important to safety. The replacement of the slide links installed inside safety-related equipment for safe hot shutdown per 10CFR40.48, Appendix R with new isolation / transfer switches (total contact 63; 15 open, 48 closed) requires performance of additional pieces of equipment in safety-related circuitry. These isolation / transfer switches (total of five) will allow local / remote control of Service Water Pump P-7B (152-103), Load Center
- 13 (152-110), and Emergency Diesel Generator 1-1 and its associated breaker (152-107).
This marginal increase in probability is not a safety concern for the following reasons:
- 1) The new isolation / transfer switches are qualified to IEEE-344-1975 and are the same type used throughout the industry.
- 2) Replacement of the existing links will be done following an approved install-t ation procedure.
- 3) Following construction activities, the individual equipment will be tested to verify operability and conformance to 10CFR50.48 and Appendix R.
The control handles for these isolation / transfer switches are a special remove-able type. They will be administrative 1y controlled to prevent human error in switching positions from remote to local.
FC-653 9
This Facility Change covered the replacement and installation of proper fusing to provide short circuit overcurrent protection, and replace existing undersized I cables with proper sized cables. (DC Panel and Cable Short Circuit Protection).
SAFETY EVALUATION
SUMMARY
l This modification was installed as a result of Event Report E-PAL-83-162A, iden- l tifying a deficiency with short circuit overcurrent protection, after a previous l modification (FC-497) installed larger capacity station batteries.
The modification brings the plant back in conformance with FSAR requirements in Section 8.5.2 which defines limits for conductor temperature upon initiation of a fault.
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Page 8 Therefore, this project will decrease the probability of occurrence or the conse-quences of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR.
l The modification reduces the available short circuit current to previously analyzed and acceptable values. Therefore, this modificatton will prevent the possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR.
FC-657 Prior to this modification, Component Cooling Water flow to containment isolated on a SIS coincident with low CCW pump discharge header pressure. This modification modified the isolation logic to isolation valves CV-0910, CV-0911 and CV-0940 so that isolation occurs when a Containment High Pressure signal is present.
A key-operated switch was placed in series with the CHP contacts to allow operator override.
Logic circuitry associated with SIS / low CCW pressure isolation was removed.
SAFETY EVALUATION
SUMMARY
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is increased because This change replaces the former method for isolating CCW from containment (SIS coincident with low CCW pump discharge header pressure) with Containment High Pressure. The proposed CHP actuated logic is considered as effective as an SIS actuated logic in restricting the influx of contaminants into the CCW system which may result from a LOCA or MSLB event. Civen a LOCA or MSLB which results in containment pressurization (events and conditions considered as requisite for a break of the CCW system from missiles), the CCW system will automatically iso-late at 4 psig, a substantial margin below the normal minimam CCW system pressure of 21 psig in the return piping near the containment exit penetration.
Capability will exist to manually bypass the isolation signal. Procedural con-trols will allow the operator to place a bypass switch (one switch per isolation l valve) to the "open" position, thus opening the associated isolation valve, only i if the following conditions are met
- 1) Containment pressure is 5 20 psig.
- 2) CCW surge tank level is above the low level alarm setpoint
- 3) Instrument air is available for future valve closures. In addition, the pro-cedures will identify the specific instrumentation to be used to assess these conditions.
l MIO387-0046A-TC07
e Page 9 Once the bypass is initiated, the operator can reisolate the CCW system by placing the bypass switch to the " auto" position or, if the CHP condition no longer exists, the switch can be placed to the "close" position in which case the close signal is applied by a circuit path parallel to that for automatic isolation. Explicit procedures will require that the operator periodically monitor conditions il and 2 above to ensure that reisolation is effected if either condition is not met.
A review of emergency operating scenarios using procedures being upgraded in accordance with NUREC 0737 requirements shows that manual bypassing or reisolating will not be required when operator attention would be diverted by more important actions.
The bypass switches are of proven design, and are key-locked and located in the control room which minimizes the potential for the switches to be mispositioned.
l The switches will be seismically qualified for IE service.
l l Although diverse signals will not be employed to automatically actuate isolation as in the case of SIS (which responds to either low pressurizer pressure or CHP),
four redundant and independent CHP measuring channels are combined in redundant 2/4 logic matrices to reliably isolate redundant isolation valves when required.
I For this and the reasons provided above, it is concluded that the probability of occurrence or the consequences of an accident or malfunction of equipment i important to safety previously evaluated in the FSAR is not increased.
1 l
Civen the intent of the isolation system is to restrict the influx of contam-ination into the CCW system, the proposed isolation method, CHP, will protect the CCW system as effectively as an SIS actuated scheme for LOCA or MSLB's.
Operator actions relating to operation of the bypass switch will be controlled by procedure such that important plant parameters are noted before a bypass can be in effect.
An accident of a different type than that explicitly evaluated in the FSAR could potentially occur if an intersystem leak developed between the PCS and the CCW system leg (eg, gas caused by a tube failure in the letdown heat ex-changer, the PCS pump seal heat exchangers, control drive mechanisms or the reactor shield cooling heat exchanger). In such a case, containment pressuri-zation would not occur and thus the CCW isolation valves would not automatically close. In the current and original designs for the valves' actuation circuitry, SIS initiation would attempt to close the valves whenever pressurizer pressure dropped approximately 400 PSIC to an SIS set point of 1605 PSIC.
Although automatic isolation would not occur for this condition, diverse indi- :
cation is present in the control room to alert the operator. High and low CCW .
I surge tank level alarms exist as does a CCW system high radiation alarm which is sensed across the CCW pumps. This equipment, in addition to existing operating procedures which require that operators attempt to locate and isolate leaks into or out of the CCW system based on these indications, serve to ensure that isolation will be effected when needed.
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l MIO387-0046A-TC07
Page 10 The above provisions are consistent with Begulatory Guide 1.45 which indicates that substantial intersystem leakage from the PCS boundary to other systems across passive barriers is not expected. Regulatory cuide 1.45 also indicates that the monitoring of water radioactivity and tank water levels in the system connected to the PCS boundary should be performed in the event that such leakage
, would occur.
Reliance on manual actions to isolate such a leak is simply an extension of original design and operating practice. The CCW system was never fully auto-mated to respond to intersystem leaks. Even if an SIS signal were used, success-ful isolation could not be guaranteed as the isolation valves are not rated for PCS pressure.
FC-662 This Facility Change covered Core 7 reload, Batch K, which permits power pro-duction for 12,250 MWD /NTU. The changes in Cycle 7/ Batch K reload ares l 1. Expected cycle burnup is 12,250 MWD /HTU for Cycle 7 as opposed to 10,589 MWD /MTU for Cycle 6.
l
! 2. Cycle 7 will contain 16 more Cadolinia fuel assemblies than Cycle 6. Batch K contains 20 such assemblies. Batch J was also made up of 20 Cadolinia assem-blies.
SAFETY EVALUATION
SUMMARY
The thermohydraulic performance of Batch K will be identical to Batch J. From a mechanical standpoint the reload Batch K design is identical to Batches I and J.
Cycle 7 of the Palisades Plant is designed to operate at 2530 MWT starting in February 1986. The characteristics of the fuel and of the reload core result in conformance with required shutdown margins and thermal limits.
Exxon Nuclear Co's Palisades Cycle 7 Safety Analysis Report (XN-NF-85-94) indi-cates a refueling boron concentration of 2 1770 ppm is required to achieve a Keff 5 0.95 with all rods out at 70 degrees F and Batch K loaded which results in a conservatively high ppm value relative to the 60 degrees F condition refer-red to in Technical Specification 3.3 Basis. Per changes to COP 11, SOP 3, SOP 28 and appilcable chemistry procedures a Keff 5 0.95 (all rods out assumed) is assured for the length of the refueling. In addition, this new concentration has been noted in a standing order which will be reviewed by all operating shifts prior to their moving fuel. This preserves the assumptions used in the Safety Analysis. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. The characteristics of the Batch K fuel and of Cycle 7 reload core result in conformance with existing technical specification limits regarding power peaking, shutdown margins and moderator coefficients. Refueling boron concentrations required is sufficient to maintain Keff 5 0.95 (assuming all rods out, 70*F).
Therefore the margin of safety as defined in the basis for any technical specifi-cation is not reduced.
M10387-0046A-TC07
o Page 11 FC-675 This modification added a backup Nitrogen supply to the auxiliary feedwater tur-bine inlet valve. The original standby 2,400 psig nitrogen bottle system with manifold and pressure reducers were located in the auxiliary feedpump room. The purpose of this new system provides a backup to operate P-8B turbine inlet valves CV-0522B and PCV-0521A in the event that a fire disables the plant air system.
The nitrogen backup system had failed to provide pressure required during a C-150 and C-150A safe shutdown panel surveillance test. This modification proposes to repair the leakage in the system, as well as add three nitrogen bottles to the i
system. The system will be relocated from the auxiliary feedpump room to the north i
end of the 590' elevstion of the turbine building.
l SAFETY EVALUATION
SUMMARY
In the event of fire causing the loss of normal plant instrument air, the standby nitrogen bottle system will supply control nitrogen to the auxiliary feedpump turbine inlet valves (CV-0522B and PCV-0521A) for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The addition of three nitrogen bottles to the axiliary feedwater turbine inlet valves backup nitrogen
! supply will increase the reliability of this system. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased by this modification.
The original nitrogen backup system was designed with seismic, environmental and missile considerations. This modification is installed due to fire protection requirements as specified in Appendix R. Appendix R does not require systems to
)
meet seismic or environmental criteria. The new system is designed to withstand a one half operating basis earthquake which is the same as the instrument air system which the nitrogen systems is tied into. A check valve and isolation valve were installed prior to the tie in to the air system to isolate the nitrogen system in the event of an accident. Because of the above, the possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created.
FC-679 This modification covered the addition of two hangers on an interconnecting pipe i
to the SIRW tank (SIRW tank recirculation discharge piping), which had experienced l
galvanic corrosion. It was determined that with the addition of two pipe supports the only stresses left on the pipe would be due to pressure.
I SAFETY EVALUATION
SUMMARY
l l
The two pipe supports were designed to ASME II Class 2 piping specifications.
! If pipe failure were to occur, leakage through the pipe would be minimized with l the addition of supports. This modification provides additional support to a !
degraded pipe, therefore, an accident or malfunction of a different type than any l l evaluated previously in the FSAR will not occur because of this modification. The modification will allow continued operation of the SIRW tank recirculation system, l j
and the margin of safety will not be reduced. 1 l
l HIO387-0046A-TC07 I
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, Under this Specification Change a number of tubes in Steam Generator B were re-moved from service using a welded interference tube plug. These tubes are listed on the Palisades 1985 Plugging Candidata List, and were removed to return the steam generator- to an operable conditioJ .
1 SAFETY EVALUATION
SUMMARY
, ~
. Tubes removed from service exceeded the Palisades plugging criteria as specified in the 1985 Steam Generator Eddy Current Test Outage Plan. By removing defective tubes-from service, the possibility of a primary to secondary leak is greatly decreased. Since qualified plugs and qualified welding procedures were used, the
- probability of occurrence or the consequences of an accident or malfunction of l equipment important to safety previously evaluated in the FSAR is decreased. Pro-ceduralizing the welding process insures that correct practices and materials be utilized and minimizes the potential for a weld failure.
I SC-86-007 l This Specification Change covered the replacement of the Incore Detector assmebly i with an optional incore restraint assembly installed in the reactor vessel instru-mentation flange and connected to an incore detector cable to allow an irradiated incore detector cable to be cut and restrained until the subsequent refueling outage and to replace the normal incore detector assembly as the pressure boundary.
I' SAFETY EVALUATION
SUMMARY
The design of the in-core restraining device conforms to the operating and
! seismic requirements of existing seal plugs found on in-core instrument cable j assemblies. They are fabricated from the same materials, or with materials I with comparable mechanical properties, as used for the in-core instrument seal j pluss, and are machined to the same dimensions as those found on the seal plug.
i Fasteners used to retain the in-core cable are fully captured so they cannot become loose parts during reactor operation or during postulated seismic con-3 ditions. Because of the high initial tightening torque, the fasteners which
]i secure the in-core detector cable within the restraining device are expected to remain tight during operation and postulated seismic conditions. Therefore, I this device does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously eval-uste in the FSAR.
5 l SC-86-216 1 .
This modification covered the equalizing valve cam on the containment escape air lock. The original design of the equalizing valve cam caused the valve to open when the opposite door of the air lock was opened, providing a leak path from containment to atmosphere. This is a result of the cam lobe swinging back to the point where the can roller / lever rides up on the cam rather than staying in the
- closed position or on the small radius of the cam. The can was modified by grind-Ing the cam lobe to assure the equalizing valve remain closed while the opposite i
4 MIO387-0046A-TC07 1
Page 13
.i door is opened. As a final test of operability, the valves were pressurized to verify the valves remain closed when the opposite door is opened, and that they equalize properly. The cam itself sees no significant stresses due to the air lock operation, therefore this modification will not reduce the integrity of the material.
SAFETY EVALUATION
SUMMARY
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased because this modification will bring the air lock operation into compliance with the FSAR and technical specifications regarding containment integrity.
Special Test T-95 This Special Test provides surveillance for the initial approach to criticality after a core alteration.
SAFETY EVALUATION
SUMMARY
The operational steps performed within this test are bounded by present safety
' analysis for the original and subsequent new cores. The procedure is used in conjunction with approved operating procedures and the result is additional sur-veillance and control of the critical approach. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
i Special Test T-181 i
The purpose of this Special Test is to demonstrate operability of the Auxiliary l Feedwater Turbine (K-8) and Pump P-8B when normal supply steam pressure'is less than 82 psig. It is intended that this procedure be used following maintenance on the turbine and/or pump when normal steam pressure is not available to demon-
! strate operability.
}
SAFETY EVAULATION
SUMMARY
1 l This test is designed to demonstrate that the turbine / pump will perform as de-l signed. The test will be conducted using low pressure steam. The turbine test l will be conducted with its design. The probability of an occurrence or the con-I sequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. The test is to be performed when the i
primary system is above 325*.
The test procedure places the automatic actuation of flow control valves in one
! train (to A and B Steam Generators) in manual. This is required to control flow.
l The proposed change to the technical specification allows up to two pumps to be in manual for testing. Proposed valve alignment T-181 is considered to be con-sistant with this criteria. This test places the valve train in manual which is i no different. The basis for the technical specification is not reduced by this j testing.
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- Special Test T-192
- The purpose of this test is to demonstrate performance capabilities of the aux-111ary feedwater Pump P-8A. It is intended that this procedure be 'used following maintenance that could affect performance. It is expected.(but not necessarily required) that this procedure be used prior to exceeding a primary system temp-
! erature of 325'F. 'Other technical specification tests are. required to demonstrate overall operability to Pump P-8A c d associated system. .
i SAFETY EVALUATION
SUMMARY
I
! This procedure-demonstrates the performance of the Auxiliary Feed Pump P-SA. The test will be conducted within equipment design. Testing reduces the probability of a malfunction of equipment by' demonstrating the pumps performance. The
,[ testing does not introduce the possibility of an accident or malfunction of a different type than previously evaluated in the FSAR.
!- Special Test T-193 1
The purpose of this test is to provide the seat leakage integrity of the safety '
injection tank fill and drain valves, post maintenance, and to provide a means
] for operations to identify a failed fill and drain valve when leakage is suspected.
l
} SAFETY EVALUATION
SUMMARY
1 i The test is performed with the plant in either hot or cold shutdown, and the 2
reactor is not critical during the test activities. The test helps to verify i
acceptability of the safety injection tank fill and drain line valves, thus helping to ensure improved operation of the safety injection tanks due to po-j tential boron dilution or tank level problems. The test incorporates appropriate
- precautions to protect safety related equipment.
1 l The probability of an accident or its consequences are not increased as a result j of completing the test for safety injection tank fill and drain line valves.
j Special Test T-194 i
i The purpose of this test is to verify the seat leakage integrity of safety in-
- jection tank pressure control valve following valve maintenance, and to provide j a means for operators to identify a failed pressure control valve when leakage i is suspected.
1 SAFETY EVALUATION
SUMMARY
This test is performed with the plant in hot shutdown condition, thus the reactor
] will not be critical'.
1 J
I.
The test helps to verify acceptabllity of the safety injection tank pressure control valves, thus helping to ensure improved operation of the safety injection tanks i
due to potentia 1 ' boron dilution or tank level problems.- The test incorporates I appropriate precautions to protect safety related equipment. The probability of l l
an accident or its consequences are not increased as a result of completing this I
{ test for-the safety injection tank pressure control valves. I j
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1 MIO387-0046A-TC07
~
.: w Pags 15 Special Test T-195 The purpose of this Special Test is to determine a PCS hot full power _ mass flow
] rate for Cycle 7.
SAFETY EVALUATION
SUMMARY
The purpose of Special Test T-195 'is to accurately determine the primary system flow and the corresponding limit for the reactor inlet temperature at rated conditions. -
Technical Specification 4.15 requires that the primary system flow be measured before the reactor is made critical after each refueling outage or af ter plugging ten or more steam generator tubes. This is required to ensure that the PCS flow is consistent with the basis of Technical Specification 3.1.1.c.
Under Technical Specification Surveillance Procedure RT-57, this flow measurement 1
is performed in hot shutdown by measuring the differential pressure across.each i
of the PCP pumps, then converting to flow using the PCP pump curve, then converted 1 to 532*F from the test temperature. If this measured flow is less than 126.9 X 10s Ibe/hr, the TM/LP setpoints are reduced by 1*F in inlet temperature'for each
, 1% of reactor flow deficiency as required by Technical Specification 3.1.1.c.
{ Special Test T-195;is not intended to change this requirement.
1 i The maximum allowable T-inlet at steady state 100% power operation is defined in s Technical Specification 3.1.1.g. The parameter "W" in the equation is an input '
to this equation and is given as the total recirculating mass flow corrected to s
the operating conditions. " Operating temperatures conditions" is interpreted to
, mean the desired average PCS temperature at 100% rated power. .Before the average PCS temperature is increased, a revised "W" will be determined for the new desired operating condition. The revised "W" will then be used to calculate the new max-
- imum allowed T-inlet temperature to det*rmine if the desired operating conditions are acceptable. , a.
In the past, the value for "W" has been calculated by correcting the.PCP flow l
measured in Technical Specification Surveillance Pthcedure RT-57. It is believed l
that this flow is lower than actual and hence overly conservative. Therefore, Special Test T-195 was written and uses the heat balance method for calculating mass flow. Power is determined by a secondary calorimetric measurement and mass flow is then determined from the equation. This particular flow measurement <
methods is used by most of the Combustion Engineering plantr. l t
The heat balance method requires the determination of Enik average temperatures l in the cold and hot legs of the reactor. Because of incomplete mixing in the l reactor upper plenum and hot legs, temperature stratification may occur in the i hot legs. Therefore, a correction factor is required to adjust the indicated hot '
leg temperatures to a bulk average temperature. "
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Combustion Engineering will provide the necessary correction-factor after they have completed their analysis for the Palisades reactor. The correction factor must be supplied in order to complete Special Test T-195.
The basis for Technical Specification 3.1.1.g states that the equation for T-inlet includes a 3% flow measurement uncertainty. The heat balance flow measurement method includes uncertainties in the secondary heat balance, the stratification correction factor and in the temperature and pressure readings. The total measure-ment uncertainty for the heat balance method will be determined and included in T-195. Any measurement uncertainty over the 3% assumed in the basis of Technical Specification 3.1.1.g is subtracted off from the heat balance flow prior to cal-culating the maximum allowed T-inlet. .
The basis of Technical Specification 3.1.1.g also states that the T-inlet equation has been reduced by 3% for core flow bypass. Since the heat balance flow is total vessel flow, including bypass flow, it is consistent with the assumptions in the T-inlet equation.
, The probability of occurrence or the consequences of an accident or malfunction of' equipment important to safety previously evaluated in the FSAR is not increased by this Special Test.
Special Test T-196 This Special Test demonstrates the operability of the motor driven Auxiliary Feed Pump (P-8C) over a specified period of time. The pump will be operated continuously for a minimum of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to comply with a commitment to the NRC (NUREC 0737).
SAFETY EVALUATION
SUMMARY
This special test was written so that it may be performed at any plant condition.
The Auxiliary Feedwater Pump P-8C is_ run on minimum flow (recirculation flow) recirculating back to the condensate storage tank. The test takes temperature, vibration and discharge pressure data and compares them to acceptance cri~teria.
4 S
This procedure demonstrates the endurance of the Auxiliary Feed Pump P-8C and is conducted within equipment design. Testing reduces the probability of a malfunction of equipment by demonstrating the pump's endurance. It does not introduce the possibility of an accident or malfunction of a different type than previously evaluated in the FSAR.
Special Test T-222 ,
This Special Test provides instructions for periodic testing of auxiliary feed-wtter check valves for back leakage to assure leakage does not become excessive which could lead to stress binding of an auxiliary feedwater pump. l
\
SAFETY EVALUATION
SUMMARY
This Special Test involves opening the vent / drain valve (s) on the auxiliary ,
- feedwater lines upstream of the particular check valve under test, resulting l in the application of maximum AP across the valve under the test. Maximum AP l MIO387-0046A-TC07 w
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1 Page 17 across the valve increases the probabil.ity of the valve seating (if it is not fully seated prior to testing). Also, one- train of auxiliary feedwater will be available at all times during the test. Therefore, the probability of occurrence or-the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
In the unlikely event that the 1" or 3/4" vent / drain valve (s) used.in,the test were unable to be reclosed, accompanied by the simultaneous or subsequent failure of the check valve under test, the accident or malfunction created would not-exceed the main feed / main steam line break outside containment previously analyzed in the FSAR. Therefore, this test would not create the possibility of.a different type of accident or malfunction than any previously evaluated in the FSAR.
During this test, one train of feedwater at a time will be removed from service.
- Technical Specification 3.5, Paragraphs 3.5.1 and 3.5.2 allow one auxiliary feed-
! water pump (motor driven) to be inoperable (steam driven pump not required prior to criticality) for a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> providing one free pump is operable. A
) conservative estimate would be each auxiliary feedwater train out of service for one shift. Therefore, the margin of safety as-defined in the technical specifi-cation is not reduced.
j t Special Test T-223 i
The purpose of this Special Test is to balance the Component Cooling Water System for a DBA condition such that the shutdown heat exchangers have a minimum flow of' 5000 gpa and to balance the CCW system in normal shutdown condition such that
! the flow through the CCW heat exchangers is less than 4000 gpm each.
SAFETY EVALUATION
SUMMARY
This test will be performed during cold shutdown. PCS temperature will be main-tained between 70-150*F. Spent fuel pool temperature will be maintained below 130*F. The probability of an accident is not increased because the equipment in the component cooling system will be operated within their design limits. The
) consequences of an accident are reduced because the plant is in cold shutdown and adequate CCW flow to safety related equipment is verified..
The test will verify that a minimum of 5000 gpm is flowing to the shutdown heat P
3 exchangers, 61 gpm is flowing to engineering safeguards pumps 41 and gpm is
, flowing to the charging pumps in the post DBA condition.
The test will verify that a msximum of 4000 gpm is flowing through each CCW heat exchanger with a minimum of 5000 gpm to the shutdown heat exchangers, 61 gpm is flowing through the engineering safeguard pumps, 41 gpm to the charging pumps and
- a minimum of 300 gpm to the spent fuel pool heat exchanger for normal plant shut-down. The other CCW loads will be monitored to show adequate CCW flow.
Engineering safeguard heat exchangers, engineering safeguard pumps and spent-fuel pool will be set lower than FSAR Table 9-7 requirements.- These are justified by Combustion Engineering Analysis, revised vendor information, and fuel pool tem-perature rise without cooling data.
i Temporary gauges will be installed on the CCW system to support flow verifications.
t MIO387-0046A-TC07 1
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FACILITY CHANCES, SPECIFICATION CHANGES, TESTS AND EXPERIMENTS DECLARED OPERABLE PRIOR TO 12-31-86 PACKAGE NOT CI4 SED OUT 4
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MIO387-0046A-TC07
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FC-567-1 i The Facility Change upgraded 16 incore thermocouple signals to meet the require-4 ments of NUREC 0737 as follows:
a) Provide qualified connectors on 16 incore detectors.
b) Provide qualified cable inside containment for 16 incore detectors.
. -c) Provide qualified containment side penetration connectors for 16 incore
! detectors.
4 SAFETY EVALUATION
SUMMARY
The in-containment electrical connectors and cabling of 16 incore detectors are being upgraded to meet environmental qualifications. The Facility Change' modifies equipment to provide instrumentation having increased reliability, thus the prob-ability of occurrence or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. The upgrade of the thermocouple increases the reliability of instrumentation used by control room operators during post accident conditions and thus potentially reduces- the consequences of an accident.
! Under this Specification Change a number of tubes from the Steam Generator A 1
were removed from service by mechanically plugging them in order to restore the steam generator to an operable condition.
A SAFETY EVALUATION
SUMMARY
Westinghouse Model 27 mechanical plug was used to remove defective tubes from service, resulting in a decrease in the possibility of a primary to secondary
' leak. The plug has been qualified and approved for use in 3/4" OD tubing by I the NRC as shown in Westinghouse Test Report SE-SP-40(80) Revision 1. The plugging criteria used is 51% as determine by the 1983/1984 Steam Generator
' Evaluation and Repair Report, NRC Docket #50-255. This 51% through-wall IGA-
' as detected by the 4C4 Eddy Current probe is comparable to the present technical specification limit of 64% wastage through-wall defect. By plugging to this criteria, the level of safety is neither increased or decreased, but is main-tained. The possibility of occurrence or the consequences of an accident or i malfunction of equipment important to safety previously evaluated in the FSAR
{ is not increased.
4 i
The possibility of an accident or malfunction of a different type than any eval-i usted previously in the FSAR is not created because the plugging of defective i
steam generator tubes using a qualified plug will be done within the limits of-j existing technical specifications. No design or operating parameters are violated.
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OPEN FACILITY CHANGES AND SPECIFICATION CHANGES REPORTABLE ON THE ANNUAL REPORT NOT DECLARED OPERABLE OR CLOSED OUT i
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FC-500 This Facility Change covers the modification of Palisades Control Room and Technical- Support Center ventilation system. The modifications include an extended Control Room suction line to a location 100 meters from containment, replaces intake and discharge dampers, and modifies intake duct and dampers to allow for 100% make-up, and other miscellaneous items.
4 SAFETY EVALUATION
SUMMARY
1 The Control Room HVAC is being upgraded to alleviate the design deficiencies that were originally present in the original system. The - probability of occur-
' rencesor the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is much less than with the original system due to the addition of the redundant system and the seismic qualifica-tions of the system.
FC-510-5 1
Note: Part of this modification was declared operable by Operations in 1982. ,
However, the Facility Change was not reported on the Annual Report due !
to the fact that the replacement of a 10 hp fan was delayed and the I Facility Change remains open in it's entirety as of this date.
I l
The completed portion of this modification covered the replacement of the i
purge system with two 8" lines. Valves CV-1805, CV-1806, CV-1807 and CV-1808 ;
were replaced with 8" valves. Penetration #4 was grouted and sealed. CV-1813 and CV-1814 were replaced with new redesigned 12" valves (NUREC 0578).
The incomplete portion of the Facility Change was replacement of V-35 with a 10 hp fan which is not reportable in the annual report. The installation of the fan was delayed at the time the above was performed until the next Refuel-ing Outage. At that time it was determined that the in'tallation s of the fan
' was not a requirement or a safety concern, nor did it effect the operability of any other system, and was put on an indefinite hold.
SAFETY EVALUATION
SUMMARY
48" ducks are being replaced with two 8" lines which will be grouted. Valves CV-1813 and CV-1814 are redesigned to close against containment pressure of 55 pounds. This system will be designed to meet plant seismic requirements. The system will not degrade containment integrity, therefore the probability of oc-currence or the consequences of an accident or malfunction of equipment-important i
to safety previously evaluated in the FSAR is not increased.
The margin of safety as defined in the basis for any technical specification is not reduced since leakage will be equal to or less than specified in the technical specifications.
- MIO387-0046A-TC07
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FC-567-1 Part of this Facility Change is on Page 28 of the Annual Report for FC's that
-were completed in 1986. This covers,the balance of the modification. The mod-ification upgrades the Subcooled Margin Monitor (SMM) as follows:
i
- 1. Replace 2 pressurizer pressure transmitters to increase range to 0-4000 psi
- 2. Replace 16 primary coolant transmitters to increase range 50-700*F 1
- 3. Reprogram the SMM to accept the increased input ranges and output subcooled margin over the range from 200*F subcooled to 35'F superheat.
SAFETY EVALUATION
SUMMARY
This modification provides new pressurizer pressure and primary coolant temper-ature transmitters as inputs to the subcooled margin monitor (SMM) to expand the range of the SMM below the present lower limit of 515'F. This facility change modifies equipment and components currently describ td in the FSAR to provide in-strumentation having increased reliability. The increased range of the SMM, along with the upgrade of the incore thermocouples, increases the reliability of
- instrumentation used by the control room operators during post accident condi-tions, and thus potentially reduces the consequences of an accident.- The proba-4 bility of occurrence or malfunction of equipment important to safety previously evaluated in the FSAR will not be increased.
The margin of safety as defined in the basis for any technical specification is not reduced, and the possibility of an accident or malfunction of a type not pre-viously identified in the FSAR is not created.
FC-567-2 This Facility Change provides reactor vessel level monitoring to meet NRC re-quirements provided in Ceneric Letter 82-28 including:
, 1. Install 2 level probes in existing guide tubes
- 2. Provide level indication and recorders on the C-11A Panel I
- 3. Provide recorders for monitoring temperature of 16-qualified incores on the
- e C-11A panel. 1
- 4. Provide reactor vessel level indication and incore thermocouples readouts-_in the critical function monitor. .
SAFETY EVALUATION
SUMMARY
This modification provides redundant instruments to monitor reactor vessel level from near the top of the head to near the top of the fuel. The instruments are inserted into existing incore detector guide tubes through existing instrument
, flanges mounted on the reactor vessel head. Cabling exits the reactor vessel
~
! MIO387-0046A-TC07 i
Page h f
1 through seal plugs having a design similar to the incore instrument seal plugs. ,
As the reactor vessel level monitoring system is similar to the incore detectors presently described in the FSAR, the probability of occurrence or malfunction of equipment important to safety previously evaluated in the FSAR will not txt
'; increased. The reactor vessel level information provided to the control room operators sids in their decision making process during post accident conditions and thus potentially reduces the consequences of. an accident.
The possibility of an accident or malfunction of a type not previously identified in the FSAR is not created. The components and equipment are qualified in ac-4 cordance with applicable requirements of IEEE' standards.
The margin of safety as defined in the basis for any technical specification will not be reduced, as the reactor vessel level monitoring system is not pre-sently included, but is to be included in the technical specifications following completion of the modification.
FC-609-1 i
This modification upgrades the Critical Function Monitor to provide the Safety Parameter Display System required by NUREC-0737, Supplement 1. The modification
- consists of the following:
i 1. Procure and install a Perkin Elmer 3210 computer in CFM Trailer for use as s
- software development system.
- 2. Upgrade the 3220 Perkin Elmer computer to contain 4 megabytes of main core memory, j
i
, 4. Implement CFM sof tware developed Item 3. l 1
SAFETY EVALUATION
SUMMARY
l i
j The Facility Change modifies the Critical Function Monitor System (CFMS) to meet !
the NUREC 0737 Supplement I requirements for a Safety Parameter Display System.
The computer hardware is non-class lE and it's' failure cannot increase the proba- l j
l bility or consequences of an accident or malfunction of equipment important to i safety as all safety grade inputs to the computer. system are isolated from the i
CFMS to the requirements of IEEE-384-1977. The SPDS consolidates important plant parameters in a single location. Failure of the SPDS would not jeopardize plant safety as the information required to access plant status is available on j other instrumentation in the Control Room.
The CFMS computer system is not required by the technical specifications. In-strumentation required by technical specifications which input to the CFMS computer system are isolated such that failures of the CFMS computer will not impact the input instrumentation.
4 MIO387-0046A-TC07
Page-5 i
i FC-609-2 3
-This modification provides additional parameters as input to the CFMS. These additional inpr.ts are required to support development of the Safety Parameter Display being performed by Facility Change 609-1.
SAFETY EVALUATION
SUMMARY
- This-modification adds additional inputs to the critical function monitor (CFM) computer from existing safety grade circuits. Isolation devices are provided to prevent malfunctions in the non-safety grade circuits from causing unacceptable influences in the safety grade circuits. The isolation devices utilized meet the criteria of IEEE-384-1977. Utilizing this isolation criteria assures that the probability of occurrence of an accident or malfunction of equipment important
{ to safety previously evaluated in the FSAR will not be increased.
The incorporation of these additional inputs to the CFM computer provides the plant operators with additional information necessary to quickly access the safety status of the plant. Providing this additional information in a manner which can be readily and quickly accessed by the operator during post accident conditions potentially reduces the consequences of an accident. The isolation
- capacity assures that the possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. The margin of safety as defined in the technical specifications is not affected because the addition of the new inputs is designed to have no effect on the operability of existing systems.
FC-628 4
This modification will add an axial shape index alarm ( ASIA) and a variable high i
power trip (VHPT) to the Reactor Protective System, will upgrade the thermal i
margin / low pressure calculators, will modify the High Rate Trip Bypass circuitry to receive the adjusted nuclear power signal and will modify the Rate Trip Channel
. Alarm to provide trip " enabled" versus " disabled" indication.
SAFETY EVALUATIfJ
SUMMARY
Installation of the Thermal Margin Calculator:
'he thermal margin / low pressure trip is designed to protect the fuel against
- f. ' ,re due to the loss of heat transfer capability between the fuel and the i cooling water. Poor heat transfer conditions can be caused by increased heat flux (power level) or decreased inlet subcooling. If the power level rises
+
rapidly above the high flux trip set point, the reactor will be tripped and the fuel will be protected. If the power level increases slowly with increasing core inlet temperature, thermal margin can be quickly lost and the fuel could
- experience a boiling crisis if a TM/LP trip was not provided.
The reason for replacing the original TM/LP trip system is-to:
- 1. Obtain current technology equipment
- ~
- 2. Procure a system capable of measuring both thermal and nuclear power and j
I selecting the higher value of the two 4
- MIO387-0046A-TC07
Page 6 1
~
- 3. Provide a means of protecting the fuel at all power levels without placing excessive restrictions on full power operations.
The VHPT and the ASIA are designed to. reduce the conservative assumptions which establish the TM/LP trip set points. Because of the ability to monitor actual plant operating conditions, fewer " worst case" assumptions must be made during j
the transient analysis. The VHPT system will limit power increases to approxi-mately 10% above the initial power level by continuously calculating a VHPT set point. Resetting VHPT establishes a new set point at a fixed level above the existing power level, and. is repeated until reaching a maximum allowable set point. The ASIA will aid the operator in limiting local power increases caused by axial power shif ts while maintaining a constant core power level.
A Thermal Margin Calculator (TMC) will be installed which will monitor and display the nuclear reactor's coolant temperature as well as reactor power. This TMC provides trips which are based on functions of reactor power and temperature.
These trips act to protect the reactor by providing safety control signals to i the RPS. The TMC alarm system provides positive operator. control for enhanced safety and plant efficiency.
This modification will install one TMC in each of the four safety channels, each {
- powered from its respective safety related instrument BUS. !
The failure of any input to a TMC will not affect the other independent channel TMC's nor will the failure of one of the TMC's. Thus a single failure will not initiate or prevent a safety action from occurring.
Panel C-27 which will contain the four TMC channels will be seismically stiffened to reduce the magnitude and frequencies the panel would experience if exposed to an earthquake.
The TMC has been designed, constructed and tested to meet IEEE standards for
' Class 1E electrical equipment, seismic qualification and for safety systems. The digital sof tware has been tested per a joint ANSI /IEEE standard to verify the proper execution of it's logic and the operation of it's outputs to safety or trip conditions on loss of power and on failure of the operating system. The installation of the TMC will not increase the probability of occurrence or the consequences of an accident previously discussed in the FSAR.
4 Modifying the high rate trip bypass input signal will not increase the probability of occurrence or the consequences of an accident previously evaluated in the FSAR.
The modification to the high rate trip disabled alarm will reduce the number of i
' alarms during power operation, and therefore will not increase the probability I of occurrence or consequences of an accident previously evaluated in the FSAR. l The possibility of an accident of a different type than any evaluated previously in the FSAR is not created by any one of the three modifications covered in this Facility Change.
J MIO387-0046A-TC07 i
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The probability and confidence level of DNB occurring is not reduced by the installation of the thermal margin calculator, the margin of safety as defined
, in the basis for the technical specifications is not reduced.
The modification of the high rate trip bypassed input signal will ensure that the margin of safety as defined in the basis for the technical specifications
! 'is not reduced by ensuring that the set point for the bypass is adjusted each time the nuclear instrumentation is adjusted.
The modification of the high rate trip disabled alarm does not affect the tech-nical specifications.
FC-687 i This modification added a local / remote transfer switch to Breaker 152-106. This
! is an extension of the Appendix "R" isolation switch project, resulting in the modification of the control circuit for Breaker 152-106 (Startup Transformer 1-2).
SAFETY EVALUATION
SUMMARY
This project replaces slide links installed inside' safety related equipment (152-106) for safe hot shutdown per 10CFR50 and Appendix R with a new isolation /
transfer switch. This switch will allow local / remote control of start-up power to the 1C bus in case of severe fire in various areas of the plant (ie, Control ,
Room, Cable Spreading Room, Corridor 106).
4 This modification does not increase the probability of occurrence or the conse-quences of an accident or malfunction of equipment important to safety previously i evaluated in the FSAR because
?
- 1. This new isolation / transfer switch is qualified to IEEE 344-1975'and is the t
' same as those already installed on plant safety related equipment (1-lDC, Breakers 152-103, 107 and 110).
) 2. Replacement of existing links will be done following technically reviewed j' plant drawings.
4 I
- 3. Following construction activities, the individual breaker will be tested to '
verify operability and conformance to 10CFR50 and Appendix R.
- 4. The control handle for this switch is a special removable type. This handle ;
will be administrative 1y controlled to prevent human error in switching posi- l l tion from remote to local.
! i 1
In order to prove that the possibility of an accident is not created, a test will be performed showing that it is possible to operate the equipment per it's design j
4 basis. Sections 8.3.2 and 8.4 of the present FSAR do not address any type of local / remote isolation switch for the IC switchgear. These sections will be up-
- dated to reflect this modification.
The margin of safety as defined in the basis for any technical specification will not be reduced since the project will follow the requirements listed in Section 3.7.lc of the present technical specifications in removal of equipment from service.
, MIO387-0046A-TC07
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l FC-689 This modification replaces the cooling water sulfuric acid scaling inhibitor system with an organic scaling inhibitor system.
This modification replaces the cooling water sulfuric acid system (which pro-hibits calcium carbonate buildup on the main condenser tubes) with an organic scaling inhibitor system which will perform the same function at a much lower cost. Also, the new chemical PCL-1 is not as hazardous to the environment as sulfuric acid and will eliminate spill cleanup costs. PCL-1 is much less harm-ful to the steam generator tubes in the event of a condenser tube leak.
The new system will have a containment wall. built around it.
SAFETY EVALUATION
SUMMARY
The only safety related interface of installing the new cooling water inhibitor system is running tubing through an Appendix R rated fire stop. The tubing from l the chemical tank to the cooling water pipe will be routed through the fire stop separating the turbine building and the service water room. The probability of j occurrence or the consequences of an accident or malfunction of equipment impor-tant to safety previously evaluated in the FSAR is not increased by this modifi-cation.
Section 10.2.4.2 of the FSAR will be changed as a result of this modification. i Presently sulfuric acid can be added to the circulating water system to prevent scaling and biological growth on the condenser tubes. This modification will permit the addition of a copolymer, PCL-1. Because this change does not affect any safety related equipment or equipment important to safety, the probability of an accident or malfunction of equipment important to safety previously evalu-ated in the FSAR is not increased.
FC-705 This modification covered the installation of new fire detection equipment in the intake structure.
SAFETY EVALUATION
SUMMARY
This modification resolves a finding resulting from Appendix R inspection by the NRC (Report No 86-022). It adds a fire detection system to the intake structure using new infrared or ultraviolet detection equipment. It has a battery capable of providing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of backup service. This new system is designed for alarm annunciation in the control room.
The safety-related portion of this modification is the seismic mounting of the equipment (conduit, J-boxes and panel) such that neither the seismic capability of the walls is degraded nor the possibility of impact to safety-related equipment located nearby is created. Construction follows E-42 specifications for conduit and J-box supports, as well as design specifications seismically-approved speci-fically for this modification for control panel mounting. With the addition of this fire detection system we have increased the reliability of the fire detec-'
MIO387-0046A-TC07 i
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- Page 9 l tion systma in this area end decreased the probability of occurrence of an acci- t i
dent or malfunction of equipment important to safety previously evaluated in the FSAR.
The function of this system is to detect fires in the intake structure and to annunciate this condition for control room operators. Creation of an accident or as1 function of a type of different than any evaluated previously in the FSAR
- is not expected. A preoperational acceptance test was performed to verify that ,
j the new system operates per its design basis. , The FSAR will be updated to reflect !
I this modification. i
! ' At present, technical specifications do not address fire detection' in the intake structure. Therefore, the margin of safety as -currently considered in the basis i for any technical specification is not reduced. Technical' Specification Table 3.22.1 1
on fire detection instrumentation will be revised to reflect this modification. l FC-707 1
} This modification is to the RAS logic circuitry. It will replace the current utwo out of four actuation logic with a one out of two taken twice actuation logic j for the SIRW tank low level.
!- SAFETY EVALUATION
SUMMARY
1 i
The Recirculation Actuation System (RAS) receives its actuation signal from redundant SIRW tank low level two out of four logic matrices. This method of actuation has been determined to be in conflict with the requirements of 10CFR50, j Appendix A.
4 I
This system (RAS) has two safety related positions, the first one being lined up to the SIRW tank, and the second one being lined up to the containment sump. A
! two out of four logic matrix vill not encompass the requirements of a two mode j
j safety function system (such as the RAS) when exposed to the initiating event / l single failure survivability requirement. The one out of two taken twice logic '
matrix incorporates both reliability against inadvertent actuation and surviva-4
!< bility necessary in a dual safety function system.
Since the two out of four logic matrix concept is deficient, installing a one out of two taken twice logic matrix would not increase the probability of occur-rence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR.
Since this modification will only change the SIRW tank low level actuation to the RAS, there will be no parameter or equipment actuation changes. Therefore, the possibility of an accident or malfunction of a different type than any evaluated i previously in the FSAR has not been created.
- The margin of safety as defined in the basis 'for any technical specification i has not been reduced since this modification will eliminate the single failure
^ mechanism which could lead to the failure of all ECCS pumps. Current technical specification provisions will be restricted by this modification to ensure that undesired inoperable channel conditions are precluded, eg, two channels declared inoperable channel placed in trip.
i
- MIO387-0046A-TC07 4
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.Page'101 1
-FC-708 i :This Facility Change covered the addition of sprinklers to be installed which would be dedicated specifically to the diesel oil transfer pumps, P18A and P18B.
This modification was recommended by the NRC during the Appendix R audit. The sprinklers will be installed eight feet above each pump and will be supplied from existing sprinkler system in the screen house.
j SAFETY EVALUATION
SUMMARY
q f . .
?
This modification adds two sprinkler heads to the Fire Protection System in the
! screen house, dedicating a sprinkipr to each Diesel Fuel Transfer Pump, P18A and P188. The design of this system is in accordance with NFPA-13, required by the
{ FSAR. The installation requires taking the sprinkler system out of. service in
, the screen house, thus a fire watch will be established as required by the tech-nical specifications.
Therefore, the probability of occurrenc*e or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is i not increased. Because a fire watch will be adhered to during installation of f the modification, the margin of safety as defined in the basis for the technical 1 specifications will not be reduced.
.I FC-713 i
, This Facility Change covers the modification of the valve on the service water outlet from containment air cooler VHX-4, changing the valves safety function
- position from fail open to fail closed.
i SAFETY EVALUATION
SUMMARY
i j
The Service Water Outlet Valve (CV-0867) will fait closed on an SIS signal and j
V-4A will continue to operate as always, however, it will not provide cooling
- during accident conditions. The cooling capacity of VHX-4 is not required for i
accident conditions, however, the cooler may be made available by jumpering
- around the SIS signal which will allow the operator control of CV-0867 from '
3 the Control Room or Panel C-33.
I 1
To maintain the proper service water flows during accident conditions to each i
of the remaining three coolers (VHX-1, 2 and 3) flow to VHX-4 must be shut off.
If during credible accident conditions CV-0867 failed to close, VHX-1, 2 and 3 would receive the minimum flow required by the FSAR because.both DC's would be available as well as three service water pumps. Postulating a DBA with loss of 1 off site power, an inoperable DC and failure of CV-0867 is not considered a cred-ible accident.
I j The probability of occurrence of an accident or malfunction of equipment impor-tant to safety is not increased by modifying CV-0867 because the V-4A cooler is considered to be excess equipment and not necessary to mitigate the containment 1
pressure for loss of coolant accidents. MSLP analysis resulting from LER 80-003
' has shown the one containment spray pump on the 1-1 DC is ample to prevent con-tainment overpressurization. The bypass valve CV-0843 which provides flow around j
CV-0867 will be failed closed. CV-0843 provides no safety function.
\
i MIO387-0046A-TC07 1
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Page il V-4A is not considered in.the FSAR accident, therefore, is excess capacity for the containment cooling and does not create an accident or malfunction of a different type previously evaluated in the FSAR.
A technical specification change dated 10/16/87 removed VHX-4 from technical specification requirements, therefore the margin _ of safety as defined in the.
technical specification basis has not been reduced.
FC-716 This Facility Change covered the installation of barriers for 1R doors in con-tainment. There are six access areas which lead to 1R areas in containment, and presently are not locked. The FSAR states these areas are to be either blocked off completly or are to be entered only through locked doors.
To prevent possible entry through these doors, a wire mesh gate will be installed.
The gates will have the ability to open from the inside, but will be locked on the outside.
SAFETY EVALUATION
SUMMARY
The six gates are to be installed in areas of containment which are over safety related equipment. To protect the safety related equipment, the anchorage of these gates will be analyzed for a safe shutdown earthquake as described in the FSAR.
Missiles in containment, as addressed in the FSAR, only considers components which are moving high velocity fluids or have a mechanical driving force. The-gates do not have the capability of becoming missiles with these types of forces.
However, to ensure the gates do not become missiles with gravitational force, mounting will be seismic as stated above.
FSAR Section 11.6.1.2 requires these areas to be either blocked off completely or locked. By installing these gates, this requirement will be met. The probability.
of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased becuase of this modification.
l Technical specifications require IR areas to be barriered or locked to prevent j access. By adding locked gates, compliance to the technical specification will be accomplished. Therefore, the margin of safety as defined in the basis for any l
-technical specification is not reduced.
FC-718 This modification adds a seal-in relay to the Recirculation Actuation System (RAS). This is being added to overcome a potential deficiency existing whereby the SIRW Tank Low Level switches could repeatedly open and close which could deplete the instrument air supply necessary to perform the RAS alignment. This modification will respond so that upon receipt of a RAS signal it will lock in until the signal is cleared and the circuit is reset.
MIO387-0046A-TC07
s s
Page 12 l
- SAFETY EVALUATION
SUMMARY
3
~
The RAS logic seal-in circuit addition will not change the intended operation I of any equipment already described in the FSAR or technical specifications. !
The modification will however, provide additional assurance that once RAS has I i
been actuated via a SIRW tank low level signal the suction for the HPSI, . LPSI I and CS pumps will not automatically shift back to the nearly drained SIRW tank. l l Pump suction will continue to be from the containment sump once RAS has been l initiated and can only be returned to the SIRW tank via direct operator action i of control switches. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated ]'
in the FSAR will not be increased because of this modification.
The existing FSAR and technical specifications applicable to this assume switch-ing of the LPSI, HPSI and CS pumps to the containment sump for suction _on SIRW tank low level. This modification assures that once this has taken place it will not inadvertently return to the SIRW tank, therefore the possibility of an accident or malfunction of a different type than any evaluated previously in
, the FSAR is not created.
4 FC-721 4
This modification is to provide isolation of the bypass valves around the service water outlet valves on a RAS isolation signal, and to change the current logic for the service water outlet valves.
SAFETY EVALUATION
SUMMARY
The isolation of the house service water to bypass the house service water valves from the CCW heat exchangers will not change the intended operation of any equipment described in the FSAR. This modification will however, provide ,
isolation of service water to these bypass valves on a RAS signal of either channel. It will also provide manual operation and indication on the Control Room Panel. The isolation system is designed such that loss of DC or air will cause the solenoid valves to close. The FSAR will be revised to show this
- isolation scheme.
The modification also changes the logic for the main service water outlet valves
' from the CCW heat exchanger. The current logic for opening these valves has one train opening CV-0823 and the other train opening CV-0826 on a RAS signal. Be-4 cause of this logic, a failure in one of the logic trains will result in one of the' heat exchanger valves not opening. This failure at the time of switch-over to recirculation will result in insufficient heat removal through the CCW heat exchangers. This failure must be identified by the operator and corrected prior to temperature in the CCW system increasing to an unacceptable value. To avoid this failure mode the logic is being changed such that actuation of either train of RAS will open both valves.
These changes are justified based on the fact that each CCW heat exchanger is rated for only 50% post DBA heat removal. The two heat exchangers are considered as one single heat exchanger. The reduction in margin for.the CCW heat exchangers is more than offset by the additional service water available for other service water loads.
{ MIO387-0046A-TC07 L
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c-t Page 13 The probability of occurrence or the consequences ~of an accident or malfunction of equipment important to safety previously ' evaluated in the FSAR is not increased.
LThe possibility of an accident or malfunction of 4 different type than any eval-usted previously in the FSAR is not created because the main service water valves will go. fully open for maximum cooling on actuation of either RAS chain, thus providing a higher margin of safety than the previous scheme.
FC-722 This Facility Change covers the installation of a backup nitrogen supply to existing air supply to pneumatic operated valves. This modification is in i response to findings of a recent Safety System Function Inspection (SSFI),
reference NRC Docket 50-255 on'11/20/86. The SSFI team identified a concern regarding reliability of a non-Q instrument air system in the event of a post accident situation.
SAFETY EVALUATION
SUMMARY
l The backup nitrogen system automatically powers the pneumatic operator- for each valve in question, in the event that normal air supply fails. Addition of the backup nitrogen supply to the instrument air system will increase reli-ability of the instrument air system.
i j
The check valves service description are as follows:
Containment Spray Isolation, Hydrazine Tank Outlet, SI Pump Recirculation',
HPSI Pump Subcooling, Service Water Containment Isolation Outlet, Service ;
] Water Containment Isolation Inlet, and Containment Instrument Air Isolation.
The check valves will be installed at the point of interface between nitrogen and air supply to prevent loss of adequate supply pressure to the valve operator in the event of loss of pressure in either of the two supply systems (nitrogen i
or instrument air). Ov,er pressure protection equipment will protect the nitrogen piping as well as downstream connected equipment. The modification will be made when the plant will be in a shutdown condition. Air supplied equipment impor-j tant to safety is designed to fail in the safe mode.
Therefore the probability of occurrence or the consequences of an accident of i malfunction of equipment important to safety previously evaluated in the FSAR
- is not increased .
The possibility of an accident or malfunction of. a different type than any evaluated previously in the FSAR is not created since the nitrogen bottle station and tubing supports are seismically designed to withstand an SSE.
The modification does not reduce the margin of safety as defined in the basis 1
for any technical specification because it adds reliability to the system.
i 1
MIO387-0046A-TC07 i
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SC-86-097 This Specification Change was written to document the exsitance of two second-
- ary manway cover plates on both steam generators relief valve assemblies. This is in accordance with E-PAL-86-019.
e SAFETY EVALUATION
SUMMARY
- The basis for acceptability and continued use of the' Steam Generator Secondary Manway Relief Valve Assemblies is as follows:
- 1. Seismic analysis indicates the present' assembly is acceptable.
- 2. Valves do not wet the primary or secondary systems, but relieve pressure 1
build up from between the secondary manway coverplate and inner manway door to containment atmosphere. The coverplate and valve assembly were added to prevent in-leakage into the steam generator and are considered "Q" for leak tightness only.
- 3. Technical specification test history documents that the present valves are
- acceptable, continuously leak-tight and reliable.
- 4. Corrective maintenance in accordance with E-PAL-86-019 placed the valve j
] assemblies in correct geometric configuration. '
I i
The relief valves were installed under the guidance of Combustion Engineering.
Since their installation, the assemblies have been tested during the performance i of an LLRT, and during containment ILRT. The assemblies have shown no indication !
of leakage with one exception. The valve was mispositioned and acted as a crud trap, disallowing proper valve operation. This exception was corrected per basis item #4 above.
' Analysis shows that the carbon steel piping can withstand a seismic event (SSE) with an acceptable factor of safety (3.7). Because the material was not pur-chased under our quality program, the analysis includes provisions for surface irregularities in calculating this factor of safety as well as very conservative material property value.
The above indicates that the relief valve assemblies are adequate to fulfill their safety function, and do not increase the probability of occurrence, or the con-sequences of an accident or malfunction of equipment important to safety as 4 previously evaluated in the FSAR. It is recommended that-the valve assablies be replaced by the end of REFOUT '88 with ASME Section 3, Class 2, qualified I
valves and piping, and will be tracked by E-PAL-86-019.
SC-86-271 This Specification Change replaced the Containment Air Cooler Service Water Flow Fischer & Porter Transmitter with CE/MAC square root converters with flow switch; to a Rosemount Flow Transmitter with linear output and Moore Industries deviation alarm with a time delayed output.
MIO387-0046A-TC07 l
t u
o
~Page 15 SAFETY EVALUATION
SUMMARY
The equipment is being replaced by functionally equivalent and more accurate equipment. Transmitter range was increased from 7000 CPM to 7500 CPM maximum.
This increase in range will prevent overranging the transmitters when two service water pumps are operating anit the service water inlet to Containment Control Valve CV-0824 is fully opened. (Flow is approximately 7400 CPM in this condition.) This increased range will allow flow to be monitored through its expected range and allow the flow switch for leak detection to be operable.
Due to inherent noise and low differential flow setpoint a time delay is being added to the output of the flow switch. Transmitter accuracy (0.2%) is suffi-cient to allow satisfactory monitoring of flow conditions using this increased range and no adverse effects will result due to the range change. The time delay is being used to eliminate nuisance alarms and will be set at less than one minute to allow leak detection in a timely manner, and no adverse effects will result from the time delay.
LE Classification in the FSAR was deleted. The basis for this is:
- 1. Current installation does not support IE requirements; and
- 2. Regulatory Guide 1.97 does not classify this instrument as category 1, 2 or 3.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not in-creased by this change of transmitters.
This equipment is used for post accident monitoring only. This change provides increased reliability via a functionally equivalent substitution.
MIO387-0046A-TC07