ML20235G851
| ML20235G851 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 09/25/1987 |
| From: | Cockfield D PORTLAND GENERAL ELECTRIC CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| GL-87-12, TAC-66264, NUDOCS 8709300257 | |
| Download: ML20235G851 (28) | |
Text
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M NUMM David W. Cockfield Vice President, Nuclear September 25, 1987 Trojan Nuclear Plant Docket 50-344 License NPF-1 U.S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washington DC 20555
Dear Sir:
Response to Generic Letter 87-12 Loss of Residual Heat Removal While the Reactor Coolant System is Partially Filled Information is provided in the attachment to this letter pursuant to your request under Title 10, Code of Federal Regulations, Part 50, Section 54(f)
[10 CFR 50.54(f)) to assess the safe operation of pressurized water reactors when the Reactor Coolant System water icvol is below the top of the reactor The information provided is specific to the design and operation of vessol.
the Trojan Nuclear Plant. It is our conclusion that the systems and proco-dures used for residual heat removal meet the requirements of General Design Critorion 34 (10 CFR Part 50, Appendix A) and the applicable Technical Specifications. Trojan's safe operation under the specified conditions is assured and'does not represent an unanalyzed event that would impact safety.
Sincerely, Attachment Mr. David Kish, Director c:
Mr. John B. Martin State of Oregon Regional Administrator, Region V U.S. Nuclear Regulatory Commission Department of Energy Mr. R. C. Barr NRC Resident Inspector Trojan Nuclear Plant Subscribed and sworn to before me this 25th day of September 1987.
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Trojan Nuclear plant Document Control Desk i
Docket 50-34 September 25, 1987 License NpF-1 Attachment Page 1 of.21 l
NRC INFORMATION ITEM NO. 1 l
A detailed description of the circumstances and conditions under which your plant would be entered into and brought to a drain-down process and operated with the Reactor Coolant System (RCS) partially fil)ed, including any interlocks that could cause disturbance to the system.
Examples of the type of information required are the time between full-power opera-tion and reaching a partially filled condition (used to determine decay heat loads); requirements for minimum steam generator (SG) levels; changes in the status of equipment for maintenance and testing and coordination of such operations while the RCS is partially filled; restrictions regarding testing, operations, and maintenance that could perturb the Nuclear Steam Supply System (NSSS).; ability of the RCS to withstand pressurization if the reactor vessel head and steam generator manway are in place; requirements portaining to isolation of Containment; the time required to replace the equipment hatch should replacement be necessary; and requirements portinent to reestablishing the integrity of the RCS pressure boundary.
RESPONSE
Operational Considerations The RCS at Trojan is considered in a partially drained condition for the accomplishment of maintenance activities when the level of the system is reduced below the top of a reactor coolant loop (hot log).
The specific plant elevations related to these conditions are en follows:
Top of Reactor Coolant Loop - Elevation 61 feet 6 inches Middle of the Reactor Coolant Loop - Elevation 60 feet 1-3/4 inches Maintenance activities that require operation of the RCS with the level drained below the top of the reactor coolant loop are as follows:
A.
Removal of the steam generator primary side manways for access to inspect or repair steam generator tubes.
B.
Maintenance on components that.are not isolable by a boundary valve.
Examples of these components include:
resistance temperature detector bypass loop isolation valves, safety injection system check valves, and pressurizer spray valves.
The removal of the steam generator primary manways requires lowering the RCS level to slightly abovo 60 feet 4 inches but never below 60 feet 2 inches. The icvol is maintained at this level until the steam genera-tor tubes have completed draining and then level is restored to the top of the reactor coolant loop. The level that is established for valve
Trojan Nuclear Plant Document Control Desk Docket 50-34 September 25, 1987 License NPF-1 Attachment Page 2 of 21 maintenance varies dependent upon the location of the piping tap off of the reactor coolant loop. Activities that require lowering the level of' the RCS below an elevation of 60 feet 2 inches can only be performed by procedure after the unloading of all fuel assemblies from the reactor vessel.
Operational experience at Trojan with the RCS at levels below the top of the reactor coolant loop has shown that satisfactory operation of the RHR pump with flow less 3500 gallons per minute can be achieved with levels as low as 60 feet 0 inches. As a result of the loss of RHR cooling event in 1984 at Trojan (Licensee Event Report 84-10, Revision 1) a margin of 2 inches was established to allow for unanticipated system upsets in level and results in the absolute minimum elevation by procedure for j
operation being set at 60 feet 2 inches.
l The time to establish operation at RCS levels below the top of the reactor coolant loop will vary dependent upon the type of outage that has been planned.
If the facility is required to go to a forced outage con-dition from normal 100 percent power operation for the purpose of con-ducting unscheduled maintenance, the RCS could be in a drained condition for maintenance in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
If the facility is taken into a normal scheduled outage and operation at reduced RCS levels is required, the average time to reach a drained condition could be as early as 7 days.
Trojan's Technical Specifications and procedures do not specify a minimum water level for the steam generators when the facility is in Modes 5 and 6.
However, steam generators under cold shutdown conditions in Modes 5 and 6 are normally maintained in a wet layup status in order to optimize the chemistry control of the steam generators.
Individual steam generators may be drained for cleaning or secondary side maintenance.
It would be a very unusual situation for all four steam generators to be in a drained condition at the same time. The actual time in the drained condition is minimized in order to echieve the optimum chemistry control conditions with the steam generators full in a wet layup condition.
Therefore, under normal outage conditions, removal of heat generated by decay heat could be transferred to the steam generator water inventory when RHR cooling is interrupted.
There is only one interlock in the drained condition that could interrupt the operation of the RHR System. This interlock is the automatic closure of the RHR suction isolation valves off Reactor Coolant Loop 4 when a high pressure condition (600 psig) is sensed in the RCS.
When vented and drained below the top of the reactor coolant loop in Hodos 5 and 6, the i
interlock is defeated by removing control power to the valve motor opera-tors. This action, taken by procedure, is intended to preclude the inadvertent actuation of this interlock from interrupting RHR cooling.
4 Trojan Nuclear Plant Document Control Desk Docket 50-34 September 25, 1987 License NPF-1 Attachment Page 3 of 21 Maintenance and Equipment Status Maintenance activities and equipment operational status are controlled by procedures contained in the Plant Operating Manual.
Specific procedures that apply to the Plant for the condition of being drained below the top of the reactor coolant loop, are as follows:
Operating Instruction OI-3-2, Draining the RCS, requires Shif t Supervisor approval for evolutions, such as operating tests or instrument calibrations, that may result in the unavailability of the RHR System until the reactor has been shut down greater than 10 days. Prior to granting permission, the Shift Supervisor must consider, by procedure, the decay heat load, duration of the evolution, and the available backup decay heat removal mechanism.
Administrative Order AO-3-14, Safety-Related Outages, assures the availability of mode-related components required by the technical specifications.
This procedure provides the operator with the controls needed to assure that essential safety-related equipment is operational. With the RCS drained below the top of the reactor coolant loop, this procedure would preclude the removal of train-related equipment necessary in supporting the operation of redundant RHR pumps.
In addition, maintenance outages require the approval of a senior reactor operator prior to beginning the work. This approval, required by Administrative Order A0-3-9, Maintenance Request, helps ensure that equipment required in Modes 5 and 6 is operable. This procedure applies to all maintenaneo and is in addition to the control of safety-related equipment as described above.
RCS Pressurization and Containment A pressurization of the RCS following a loss of RHR cooling when drained to a level below the top of a reactor coolent loop would be controlled by the procedures for the restoration of core heat removal.
Initially, pressurization of the RCS caused by the heatup of the core and subsequent boiling is precluded by requiring redundant means for the heat removal capability. Redundant RHR pumps are required and level is maintained by procedure above the icvols at which vortexing and resultant air intake becomo the only common mode failure for the RHR pumps.
In addition, Off-Normal Instruction ONI-47, Backup Core Heat Removal With Open RCS, provides means to remove decay heat should the RHR pumps tail.
See the response to Information Item 5 for details of these heat removal processes. Should pressurization occur during the heatup of the RCS, some venting will take place from the pressurizer and reactor head vents.
Additional venting of the RCS, should it be required, could be accomplished by manually opening the pressurizer power-operated relief
o
' Trojan Nuclear Plant -
Document Control Desk Docket 50-34 September 25, 1987 License NPF-1 Attachment Page 4 of 21 valve.to the pressurizer relief tank.
In all cases, the venting would be to the Containment environment which can be both monitored and isolated.
See our response to Information Item 6 for a discussion of Containment integrity.
Procedures have not been issued for reestablishing RCS pressure boundary, integrity, as.it could only be accomplished after the decay heat removal is reestablished and Containment access restored. The exact operator or maintenance actions required are dependent upon the precise system configuration at the time of the event.
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i Trojan Nuclear Plant Document Control Desk Docket 50-34 September 25, 1987 License NPF-1 Attachment Page 5 of 21 i
NRC INFORMATION ITEM NO. 2 A detailed description of the instrumentation and alarms provided to the operators for controlling thermal and hydraulic aspects of the Nuclear Steam Supply System (NSSS) during operation with the Reactor Coolant System (RCS) partially filled. You should describe temporary connec-
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tions, piping, and instrumentation used for this RCS condition and the quality control process to ensure proper functioning of such connections, piping, and instrumentation, including assurance that they do not con-tribute to loss of RCS inventory or otherwise lead to perturbation of the NSSS while the RCS is partially filled. You should also provide a description of your ability to monitor RCS pressure, temperature, and icvel after the RHR function may be lost.
RESPONSE
A.
Level Indication System The function of the Shutdown Reactor Vessel Level Indication System is to provide the operators in the control room with indication of vessel level during Plant operating Modes 5 and 6.
The system consists of two separate tubing / standpipe designs which include RCS taps at flow transmitters FT-425 and FT-435, and two 25 foot long tempered glass standpipes, with vent paths to the pressurizer spray i
vent line and pressurizer sampling system vent line, respectively, (see Figures 1 and 2).
The system provides:
l.
Remote (inside Containment) visual indication of reactor vessel level from the 55 foot elevation to 80 foot elevation.
2.
Visual indication of level in the control room over a limited range (16 inches from the camera's center of focus). A camera mounted inside Containment is focused on the standpipes (LG-1040B and 1040C) at RCS loop conterline (60 foot 1-3/4 inch).
The signal from the camera is transmitted to a 9 inch TV monitor in the control room.
3.
An electronic level transmitter is connected to the FT-435 tubing run.
This level transmitter (LT-1040) is powered from a non-vital AC bus that can be powered from either emergency diesel generator.
The signal (4 to 20 mA) is recorded on a chart recorder and roads out on a level indicator (LIS-1040) with operator adjustable setpoint alarms (see Figure 4).
Due to the fact that this system is only in service during Plant operating Modes 5 and 6, a portable cabinet, designated CO2A, has l
been designed to house the monitor and level indicator.
CO2A will set on Control Panel CO2 when in use.
This design places the level l
indicating equipment in one centralized location (see Figure 3).
Trojan Nuclear plant Document Control Desk Docket 50-34 September 25, 1987
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License NpF-1 Attachment i
page 6 of 21 l
The level indicating system is designed for operation in a borated water system with the maximum pressure being 100 psia and maximum temperature of 400 degrees fahrenheit.
By procedure, the system is calibrated and inspected for leakage when being placed into service 1
prior to draining the RCS in Modes 5 and 6.
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B.
pressure and Temperature Indication System RCS pressure is not monitored when the RCS level is below the top of a reactor coolant loop with the system vented. Pressure under low-pressure conditions in the RCS can be monitored by the indicated pressure in the pressurizer relief tank which can be connected to the system from the pressurizer through an open pressurizer power-operated relief valve. This pressure indication method is used only during the initial draining of the system or during the final stages of the refilling of the system to a solid condition.
When operating in a drained condition the power-operated relief valve j
is shut and the system is vented directly to atmosphere through J
filtered vent paths. During a loss of the residual heat removal capability with the RCS in a drained and vented condition, the increase in system pressure that would result from a core boiling.
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condition is not monitored by the pressurizer relief tank pressure j
indication (0 to 100 psig) because the power-operated relief valve is
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- shut, e
l Should the filtered vent paths from the pressurizer vent and the head J
vent to'the Containment atmosphere be insufficient, the operator could manually open the pressurizer power-operated relief valve and provide an additional vent path from the RCS to the pressurizer relief tank. This vent path would allow equilibrium conditions to establish between the RCS and the pressurizer relief tank. As long as RCS pressure stabilizes above the nitrogen cover gas pressure (5 psig) of the pressurizer relief tank, this could be a valid indication of system pressure.
The need to have a capability to mor.itor RCS pressure under this condition will be evaluated in further detail when the Westinghouse Owner's Group thermal-hydraulic analysis of the event is completed.
Other instruments for monitoring the pressure of the RCS, such as RHR pump discharge pressure, and wide-range RCS pressure are considered inadequate for monitoring pressure under the low-pressure conditions anticipated when in the drained and vented condition.
The temperature of the RCS when drained and vented is monitored with the temperature indicators in the RHR System and the core exit thermocouple system. These instrument systems are described in UFSAR 1
l
Trojan Nuclear Plant Document Control Desk Docket 50-34 September 25, 1987 License NPF-1 Attachment Page 7 of 21 Section 5.4.6 and Figure 5.4.8 for the RHR System and in UFSAR Sec-tion 4.4.6 and Figure 4.4-19 for the core exit thermocouple system.
The RRR System temperature indication is the principle means for monitoring temperature under the drained and vented system condi-tions. However, should the RHR System fail to function, the incore exit thermocouple remain functional and are a valid indication of l
the reactor vessel / core temperature. A minimum of two incore exit l
thermocouple are kept in service and are readable from the control room whenever the RCS is drained below the top of the reactor coolant loop.
System Summary The systems described above for monitoring the level, pressure and temperature of the RCS when in a drained and vented condition provide the Plant operator sufficient information to safely operate the Plant in this configuration.
Should the heat removal capability be inter-rupted during system operation, the ability to monitor level and temperature independent of RHR System operation are sufficient to indicate system status until forced flow cooling is restored. It is recognized that system pressure monitoring is limited when the RCS is in a drained and vented condition. However, under those low-pressure conditions, the capability to monitor system level and temperature are the key parameters needed to monitor and restore the system to normal operation from postulated upset conditions.
Trojan Nuclear Plant -
Document Control Desk
' Docket 50-34 September 25, 1987 License NPF-1 Attachment Page 8 of 21 NRC INFORMATION ITEM NO. 3 Identification of all pumps that can be used to control Nuclear Steam Supply System (NSSS) inventory.
Include:
(a) Pumps you require be operable or capable of operation (include information about such pumps that may be temporarily removed from service for testing or maintenance);
(b) other pumps not included in Item a above; and (c) an evaluation of Items a and b (above) with respect to applicable Technical Specification requirements.
RESP 0NSE Trojan's operating procedures require that the following pumps that can be used to control the Nuclear Steam Supply System (NSSS) inventory be operable or capable of operation during Plant operations with the RCS drained to mid-loop level, 1.
The following pumps are required to be operable per Trojan procedures and Technical Specifications.
a.
One charging pump (Technical Specification 3.1.2.2.),
- EE -
One boric acid transfer pump (Technical Specification 3.1.2.2).
Only one flow path is required; the other may be removed for maintenance.
b.
Both RHR pumps (Technical Specifications 3.4.1.3 and 3.9.8.2).
l Neither RHR pumps may be removed from service while operating I
at mid-loop.
2.
The following pumps may be operable.
a.
At least one Spent Fuel pool cooling pump (cooling only).
Only one pump at a time is removed from service.
b.
Safety Injection pumps.
Both pumps may be removed from service.
c.
Primary makeup water pumps.
Both pumps may be removed from service.
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4 Tecjan Nuclear Plant Document Control Desk Doc.ket 50-34 September 25, 1987 License NPF-1 Attachment Page 9 of 21 3.
The following pumps may be operable to increase the water inventory on the secondary side of the steam generators.
a.
Auxiliary.!feedwater pug)s (electric-driven and diescl-driven).
t Hoth pumps may be renkved f rom service.
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l Trojan Nuclear Plant Document Control Desk Docket 50-34 September 25, 1987 License NPF-1 Attachment Page 10 of 21 NRC INFORMATION ITEM NO. 4 A description of the containment closure conditions you require for f.he conduct of operations While the Reactor Coolant System (RCS) is partially filled.
Examples of areas of consideration are the equipment hatch, per-t sonnel hatches, containment purge valves, steam generator (SC) secondary-side condition upstream of the isolation valves (including the valves),
piping penetrations, and electrical penetrations.
RESPONSE
Guidance for Containment isolation to establish integrity in the event of a prolonged loss of decay heat removal while drained down to loop center-line (Elevation 60 feet 2 inches) in Mode 5 is provided by temporary procedure, Administrative Orier A0-T-28, Containment Integrity for RCS Drain Down. The procedure is applicable when drained to a level less than the top of the loop (Elevation 61 feet 6 inches).
The procedure providec for prompt restoration of modified Containment integrity following loss of RMR forced cooling.
Modified Containment integrity is defined as meeting the integrity requirements specified in Trojan Technical Specif) cation 3.9.4.
The status of each penetration that could provide direct access between the containment and outside atmosphere if open is maintsined on a Containment penetration integrity restoration plan workshoot required by A0-T-28.
These worksheets are reviewed each shift by the Shift Supervisor.
Penetrations included in this procedure include both Type B and C penetrations and other paths that may be opened as a result of maintenance activities. The removal of stsam generator manways would require closure of the main steam isolation valves and power-operated relief valves.
The procedure requires availability of tools and equipment for shutting the penetration and the manpower to accomplish the task in the event forced cooling is lost. The penetrations are allowed to be open during the drained-down condition only if the time to accomplish isolation is less than the time predicted from loss of forced circulation to core uncovery for the given core power history and time after shutdown, as specified on Figure 5.
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' Trojan Nuclear Plant Document Control Desk Docket 50-34 September 25, 1987 License NPF Attachment-Page 11 of 21 NRC INFORMATION ITEM NO. 5 Reference to.and a summary description of procedures in the control room of your plant'Which describe operation While the Reactor Coolant System (RCS) is. partially filled. Your response should include the analytic basis you used for procedures development. We are particularly inter-ested in your treatment of drain-down to the condition where the RCS is partially filled, treatment of minor variations from expected behavior such as caused by air entrainment and de-entrainment, treatment of boiling in'the core with and without RCS pressure boundary integrity, calculations of approximate time from loss of residual heat removal (RHR) to core damage, level differences in the RCS and the effect upon instru-mentation indications, treatment of air in the RCS/RHR system, including the impact of air upon Nuclear Cteam Supply System (NSSS) and instrumen-tation response, and treatment of vortexing at the' connection of the RHR suction line(s) to the RCS.
Explain how your analytic basis supports the following as pertaining to your facility:
(a) procedural guidance pertinent to timing of opera-tions, required instrumentation, cautions, and critical parameter; (b) operations control and communications requirements regarding opera-tions that may perturb the NSSS, including restrictions upon testing, maintenance, and coordination of operations that could upset the condi-tion of the 11SSS; and (c) response to loss of RHR, including regaining control of RCS heat removal, operations involving the NSSS if RHR cannot be restored, control of effluent from the containment if containment was not in an isolated condition at the time of loss of RHR, and operations to provide containment isolation if containment was not isolated at the time of loss of RHR (guidance pertinent to timing of operations, cautions and warnings, critical parameters, and notifications is to be clearly described).
RESPONSE
In May of 1984 Trojan experienced a loss of RHR cooling due to vortexing at the pump suction [Licensco Event Report (LER) 84-10, Revision 1).
The analysis of the root causes of this event prompted PCE to review care-fully the drain-down process, instrumentation required, and previous knowledge gained through operating experience.
This review resulted in both system modifications and procedure revisions that were fully implemented during the 1986 refueling outage.
In addition to the procedure changes and system modifications, analyses were completed by PGE to provide the operator with heatup rate informa-l tion and time margin information for core boiling and core uncovery when operating in the drained to the hot leg centerline condition.
This information (sco Figure 5 and Figure 6) has been made available to the operator in the procedures for operating the Plant in the drained condition.
L_ _ _______.____._ _
Trojan Nuclear plant Document Control Desk Docket 50-34 September 25, 1987 License NpF-1 Attachment i
page 12 of 21 1
The following are portinent steps contained in Operating Instruction (01)-3-2, Revision Draining the RCS, that reflect guidance related to control of the RCS level to prevent loss of RHR.
1.
The draining of the RCS is accomplished per two different sections of OI-3-2.
Section I provides for draining to the vessel. flange,Section II provides the more delicate operation of draining to the loop centerline.
2.
Both the pressurizer vent'and reactor vessel head vent must be open.
3.
RHR Inlet isolation valves MO-8701 and MO-8702 must be open and their respective control switches must be in LOCKOUT position.
Inadvertent closure could damage both RHR pumps. With the RCS open, there would then be no way to remove decay heat.
4.
NOTE:
It is desirable to expeditiously restore RHR flow to preclude the approach to boiling. This is particularly relevant when the core has only been shut down for a few days and residual heat is substantial.
5.
Both RHR trains must remain operable while drained down and one train must be running. All RHR may be stopped for up to one (1) hour j
providing the RCS is not diluted and core temperatures remain <200*F.
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6.
Critical levels associated with draining RCS:
Reactor Head Flange 67 feet 1-3/4 inches Top of Seal Table 67 feet 0 inches j)
Top of Loop 61 feet 6 inches Loop Centerline 60 feet 1-3/4 inches Top of Fuel 54 feet 9 inches j
l 7.
Monitoring pRT pressure on the computer may aid in determining if a j
negative pressure is being developed in the RCS during draining.
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8.
The RCS standpipes are made of tempered glass and rated for 100 psia and 400*F.
Do not valve in standpipe gauge glasses unless RCS is depressurized.
9.
The RCS standpipe couplings use a Teflon gasket that is susceptible to radiation-induced aging.
Check for leaks when valving-in standpipes.
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Trojan Nuclear Plant Document Control Desk Docket 50-34 September 25, 1987 l
Licerise NPF-1 Attachment l
Page 13 of 21 10.
NOTE:
One charging pump must remain operable to provide a boration flow path. The caution tag should alert the operator of this requirement and that a suction flow path and recirculation must be aligned before starting the pump.
11.
OPEN 8094. pressurizer vent valve.
Place the RCS standpipe systems in service as follows:
Contact maintenance to make following tubing connections:
Connect tubing from pressurizer spray line vent valves to RC-20.
Connect tubing to RC-13 from LG-1040C.
Connect tubing to RC-14 from LG-1040B.
CLOSE drain valves in system.
Drain downstream RC-13.
Drain downstream RC-14.
Drains from LT-1040 (2 valves).
OPEN root valves from LG-1040 B&C.
RC-13, LG1040C lower root.
RC-14, LC1040B lower root.
RC-20, LG1040C upper root.
RC-19, LC1040B upper root.
OPEN pressurizer spray line vent and root block valves to RC-20.
OPEN 8094, pressurizer sample line vent to RC-19.
I contact I&C to calibrate LT-1040 and place in service.
Place portable control panel CO2A, on control panel CO2 and plug in.
WHEN the standpipe level indication systems have been placed in service and functional checks are completed, THEN restart draining operations.
WHEN the water level (as indicated by the standpipe systems) is about the level of the reactor vessel flange (67 feet 1-3/4 inches) perform the following:
STOP draining.
Trojan Nuclear Plant Document Control Desk Docket 50-34 September 25, 1987 License NPF-1 Attachment Page 14 of 21 NOTE:
Some change in the standpipe level indications may occur due to the release of vacuum in the reactor vessel head when RC-02 is opened.
Slowly OPEN RC-02, reactor vessel head vent.
WHEN standpipe levels have stabilized and the reactor vessel head has completed venting:
START draining the RCS.
WHEN RCS level is about four inches below the reactor vessel flange (66 feet 10 inches).
STOP draining the RCS.
12.
Both pressurizer vent and reactor vessel head vent must be open to atmosphere when lowering the water level to the centerline of the hot and cold leg nozzles.
RHR inlet isolation valves MD-8701 and MO-8702 must be open and their respective control switches must be in LOCKOUT position.
Inadvertent closure could damage both RHR pumps. With the RCS open, there would then be no way to remove decay heat.
13.
If it is necessary to lower the reactor coolant level below the center line of the hot and cold les nozzles, decay heat can no longer be removed and all fuel must be removed from the reactor vessel.
Never drain below 60 feet 2 inches unless the core is unloaded and RHR is off.
At center line control icvel at 1 0 feet 4 inches.
6 Initiate a check of the standpipes to check the level by looking for air bubbles in standpipes and checking for leaks once per shift.
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Evolutions such as Periodic Instrumentation and Control Tests (PICTs) or Periodic Operating Tests (POTS) which may result in the unavailability of the RHR System to remove decay heat, SHALL NOT be permitted until the reactor has been shut down >10 days.
Performance of these evolutions at any time while drained down requires Shift Supervisor approval.
Before granting approval, the following thould be considered:
What is the decay heat load? Figure 3.3a (attached as Figure 6) in the Control Room Curves and Tables manual should
Trojan Nuclear Plant Document Control Desk Docket 50-34 September 25, 1987 License NPF-1 Attachment page 15 of 21 i
l be used to provide an estimate of RCS heatup rate without RHR cooling.
What is the duration of the evolution?
What backup decay heat removal and mechanisms are available?
14.
Prior to starting an RHR pump, ensure RHR heat exenanger outlet valve (HCV-606/607) is CLOSED.
15.
CAUTION:
Closely monitor standpipe level indications during draining operations.
If Loop B and C gauge glass indications disagree by >2 inches or gauge glass indi-j cators to LT-1040 disagrees by >3 inches, STOP draining until the cause of disagreement has been resolved.
16.
WHEN the water level reaches the top of the hot and cold les nozzles (61 feet 6 inches), the level will temporarily stop decreasing until the steam generator tubes are emptied.
The water may come out in slugs, causing erratic level indication.
If this occurs, stop draining periodically to allow the system to stabilize.
j 17.
CAUTION:
During the final stages of RCS draining, closely monitor RHR pump motor current for indication of possible air entrainment (vortexing) in the RHR pump suction. The level at which vortexing occurs is dependent upon RHR flow, but is normally not expected at water levels >60 feet 0 inches or RHR flows <3500 gpm.
IF this should occur IMMEDIATELY stop draining and reduce RHR flow.
IF indication of vortexing continues, STOP the RHR pump and go to Step 2.3.7 for actions to to restore RCS level and RHR pump operability.
18.
Before the water level reaches the centerline of the hot and cold leg nozzles, stop draining. Level should be maintained in a band slightly above 60 feet 4 inches, and never below 60 feet 2 inches.
At approximately 60 feet 0 inches, the RHR suction is very close to entraining air.
19.
In the event the RHR pump has been stopped due to indication of vor-toxing or cavitation, RCS level and RHR flow should be restored in an expeditious manner to preclude the approach to core boiling. Judg-ment must be exercised in determining the method used to restore level.
IF MD-8812 is opened to raise level, its travel may be stopped by placing its control switch in the " LOCKOUT" position.
Trojan Nuclear Plant Document Control Desk Docket 50-34 September 25, 1987 License NpF-1 Attachment Page 16 of 21 Fully stroking MO-8812 will result in a level increase of approxi-mately three feet. IF the S/G manways are open, this will result in flooding through the open manways. After RCS level is restored, the RHR pump suction must be vented prior to starting to prevent pump damage. Alternate methods of decay heat removal may be found in ONI-47, Backup Core Heat Removal With an Open RCS.
DO NOT start the second RHR pump until the cause of the loss of the first pump is determined and corrective actions have been taken.
The analytical basis for the information in the curves (Figures 6 and 7) was developed from calculations performed by PGE.
These calculations were performed in response to the Trojan loss of residual heat removal cooling event, Licensee Event Report 84-10 dated June 1, 1984. This information was made available to the Plant operators following completion of the calculations in 1984.
The basic assumptions in the calculations are as follows:
1.
Decay heat generation is based on the NRC Branch Technical Position, Auxiliary Systems Branch 9-2.
Assumptions include:
core average burnup at 16,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, which is approximately 25,000 megawatt days per metric ton uranium, heavy element decay heat is included, and 100 percent power operation at 3411 megawatts thermal prior to shutdown is assumed.
2.
The heated water volume includes water in the core and the reactor vessel. Water volume in the loops is not included in the calcula-tion.
Partial credit is taken for certain water volumes in the reactor vessel that are not indirect communication with the core, for example, the water in the downcomer region.
3.
Heat absorption by structural materials is included.
Credit for heat absorption ic weighted with respect to geometry and exposure to heated water.
For example, zero credit is taken for lower plenum materials because of assumed thermal stratification and partial i
credit is taken for upper internals because half of the upper internals will be above the water level.
Control of effluent from the containment, if containment was not in an iso-lated condition at the time of loss of RHR, would be through condensation of steam at the Containment air coolers, with subsequent Containment sump trans-I for to the Dirty Waste Drain System.
Gaseous releases would be monitored via Reactor and Auxiliary Building process radiation monitors.
In the unlikely event that RHR is lost for an extended period of time, RCS cooling would be reestablished per the Trojan Off-Normal Instruction (ONI-47).
j Initially, the level in the RCS would be restored by direct injection using a charging pump, boric acid transfer pump, or by gravity drain from the refuel-ing water storage tank, if conditions permit. The RHR System suction piping i
I
l Trojan Nuclear Plant Document Control Desk Docket 50-34 September-25, 1987 License NPF-1 Attschment Page 17 of 21 would be vented to prevent air binding of the backup pumps.
Backup cooling methods would then be placed in service, if the RHR pumps were to remain inoperable, as follows:
Cooling Using RHR Heat'Exchangers This method aligns the safety injection or charging pumps not removed from service due to maintenance, to take a suction from the RCS through the RHR pumps and discharge back to the RCS via their normal injection' pathways.
This path provides for flow through the RHR heat exchangers, which will remove the decay heat.
Cooling Using Spent Fuel Pool Cooling System and the Refueling Water Storage Tank The second heat removal path utilizes the versatility of the Spent Fuel Pool Cooling and RHR Systems which allows for cross-connecting the systems to allow Spent Fuel Pool cooling to remove RCS decay heat using the Spent Fuel Cooling System heat exchangers.
Spent fuel pool cooling pumps take suction from the RHR System connection off of RCS Loop No. 4.
The RCS water is then discharged to the refueling water storage tank via the spent fuel pool cooling heat exchangers.
The available charging =and safety injection pumps are aligned to take suction from the RWST and discharge to the RCS, thus completing the coolant flow loop.
1 l
1 l
l l
l i
Trojan Nuclear Plant Document Control Desk Docket 50-34 September 25, 1987 License NPF-1
-Attachment Page 18 of 21 NRC INFORMATION ITEM NO. 6 A brief description of training provided to operators and other affected personnel that is specific to the issue of operation while the Reactor Coolant System (RCS) is partially filled. We are particularly interested in such areas as maintenance personnel training regarding avoidance of perturbing the Nuclear Steam Supply System (NSSS) and response to loss of decay heat removal while the RCS is partially filled.
RESPONSE
Operation of the RHR System is included in initial license training and in the biannual Licensed Operator Retraining. Proper operator actions with the RCS partially drained are included in this training. Related procedures, Off-Normal Instruction (ONI)-13, Malfunction of Residual Heat Removal, and ONI-47, Backup Core Heat Removal with RCS Open, are included in this training. During the course of the discussion, Trojan Licensee Event Report 84-10 is examined in detail to ensure understanding of causes, effects and resolutions to the problem.
Maintenance Personnel Training regarding avoidance of perturbing NSSS and response to loss of decay heat removal while the RCS is partially filled is not part of the Trojan Training Program. All Maintenance activities at Trojan are approved by a Senior Reactor Operator who is familiar with the current Plant configuration. The operator is, as stated above, trained in RHR requirements and potential effects upon the system from Maintenance-related activities. The operator is responsible for eliminating Maintenance-initiated perturbation of the NSSS.
l 1
L______________
Trojan Nuclear Plant Document Control Desk Docket 50-34 September 25, 1987 License NPF-1 Attachment Page 19 of 21 NRC INFORMATION ITEM NO. 7 Identification of additional resources provided to the operators while the RCS is partially filled, such as assignment of additional personnel with specialized knowledge involving the phenomena and instrumentation.
RESPONSE
The procedural controls, described in response to NRC Information Item No. 5, in combination with the shutdown reactor vessel level indicating system, described in response to NRC Information Item No. 2, can be utilized by a normally staffed Operations crew to safely accomplish the drain-down evolution. pCE has no specific requirements for additional operators to specifically control the drain-down precess.
The need for additional operators to perform specific tasks that demand the complete attention of an operator is determined by the Shift Supervisor. Trojan has recognized that draining the RCS is a specific task that demands close operator attention.
It has been normal practice to assign an operator to the drain-down task.
In addition, it has been our normal practice to restrict control room access and to avoid con-current Plant operations that would distract operators attention from the drain-down evolution until it has been completed.
I i
1 1
L L
Trojan Nuclear Plant' Document Control Desk f
l Docket 50 September 25, 1987
)
License NPF-1 Attachment Page 20 of 21 j
i l
l NRC INFORMATION ITEM NO. 8 Comparison of the requirements implemented while the RCS is partially filled and requirements used in other Mode 5 operations.
Some require-i ments and procedures followed while the RCS is partially filled may not l
appear in the other modes. An example of such differences is operation with a reduced RHR flow rate to minimize the likelihood of vortexing and air ingestion.
RESPONSE
As described in the response to NRC Information Iten No. 5, OI-3-2.has two sections involving RCS drain-down.
Section I provides for drain-down to just below the reactor vessel flange.
Section II provides for drain-ing to the loop centerline.
Section II does the following:
1.
Provides for resolution of differences noted in icyc1 indicating systems.
2.
Warns the operator of effects of steam generator tubes draining and the resultant effect on RCS level.
3.
Provides cautions on level and flow known to be sufficient to prevent vortexing.
4.
Provides instructions on how to rapidly restore level.
5.
Pre-M' insi'setions to not start the second pump until the loss of the first pump is known.
6.
Restricts tests which may result in unavailability of the RHR System until the reactor has been shutdown more than 10 days.
In addition, as previously mentioned, Plant Administrative Order A0-T-28 is placed into offect when drained below the top of the reactor coolant loop. This procedure maintains control of all Containment penetrations for the sole purpose of being able to rapidly restore Containment integrity should RHR flow be interrupted.
During operation in Modes 5 and 6, the incore exit thermocouple system, is maintained operational with a minimum of two thermocouple that can be monitored in the control room. This provision applies only when the RCS level is drained below the top of the loop.
Trojan Nuclear Plant Document Control Desk Docket 50-34 September 25, 1987 License NPF-1 Attachment Page 21 of 21 NRC INFORMATION ITEM NO. 9 As a result of your consideration of these issues, you may have made changes to your current program related to these issues. If such changes have strengthened your ability to operate safely during a partially filled situation, describe those changes and tell when they were made or are scheduled to be made.
RESPONSE
PCE has been provided detailed information regarding the Loss of Residual Heat Removal Event at Diablo Canyon, Unit 2, in NUREG-1269, NRC Informa-tien Notice 87-23, and INPO Significant Event Report 15-87.
In addition, PGE has participated in two meetings with the NRC Region V Staff, one at Trojan on May 1, 1987, and the other in the NRC Region V office on July 9, 1987. As a result of these meetings and our review of the infor-mation provided, PGE placed into immediate effect a Containment integrity l
procedure (AO-T-28, described in other responses) and made the commitment that at least two incore thermocouple would remain in service when the RCS is drained to a level below the top of the coolant loop.
The evaluation of the information related to the Diablo Canyon event is continuing as part of PGE's operating experience review program.
Items being considered for further action are as follows:
- Improving the camera capability for monitoring the RCS level standpipes by installing a camera with a zoom lens and pan / tilt features.
- Reevaluation of the location of the level taps off of the RCS for optimum sensing of level under the partially drained conditions.
- Participating in the Westinghouse Owners Group proposal for a detailed thermal-hydraulic analysis of the event to ultimately optimize our capability to respond to a loss of residual heat removal event with the RCS in a partially drained condition.
+ Incorporating the Containment integrity procedures and the temperature monitoring into permanent Plant Operating Manual Procedures.
Th8s item will be completed prior to any future facility operation with the RCS drained below the top of the reactor coolant leop.
The evaluation of the camera and level taps will be completed by December 31, 1987. Design changes, if identified, would be implemented during outages in 1988 and 1989.
The schedule for the Owners Group analysis effort has not been determined; however, it will be discussed at a general meeting of the Owners Group on September 24 and 25, 1987.
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