ML20235P803
| ML20235P803 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 02/10/1989 |
| From: | Morris K OMAHA PUBLIC POWER DISTRICT |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-87-12, GL-87-17, LIC-89-045, LIC-89-45, NUDOCS 8903020372 | |
| Download: ML20235P803 (16) | |
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s Omaha Public Power District 1623 Hamey Omaha. Nebraska 68102 2247 y
402/536 4000
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February 10, 1989 LIC-89-045 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Station P1-137 Washington, DC 20555
References:
1.
Docket No. 50-285 2.
Generic Letter 87-12, " Loss of Residual Heat Removal (RHR)
WhiletheReactorCoolantSystem(RCS)isPartiallyFilled",
dated July 9, 1987 3.
Generic Letter 88-17, " Loss of Decay Heat Removal," dated October 17, 1988 4.
Engineering Evaluation of Fort Calhoun Station Loss of Shutdown Cooling at Mid-Loop Conditions, dated September 1988 5.
Letter from 0 PPD (K. J. Morris) to NRC (Document Control Desk) dated January 4, 1989 (LIC-88-1106)
Gentlemen:
SUBJECT:
Response to NRC Generic Letter 88-17 On January 4, 1989, Omaha Public Power District (0 PPD) provided a description of activities to address expeditious actions recommended in Generic Letter 88-17 " Loss of Decay Heat Removal."
This letter provides a summary of actions presented in Reference 5 and a description of OPPD's programmed enforcement.
Numerous procedural revisions and a modification to the existing Reactor Coolant System level instrumentation have already been completed. Additional modifications and procedural reviews are planned, as discussed in the attachment, which will ensure that OPPD has the necessary controls to preclude or mitigate the consequences of a loss of shutdown cooling event.
If it is necessary to enter mic loop conditions which requires the operation of shutdown cooling equipment prior to the installation of the modifications, OPPD will adhere to the procedural controls which were discussed in Reference 5.
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U. S. Nuclear Regulatory Commission LIC-89-045 Page 2 Detailed activities necessary to achieve compliance are provided in the attachment. A summary of these activities include:
1.
Discuss the Diablo Canyon and Related Events OPPD has provided training to Operations personnel, Shift Technical Advisors and NRC Licensed Plant Management on the Diablo Canyon and other related events. Related training will be included in the Technical Staff and Systems Engineer Training Programs. Refer to page 4 of reference 5 for additional information.
2.
System Analysis OPPD completed an analysis in September 1988, to further investigate the system response and equipment interaction during a loss of shutdown cooling event. This analysis provided the basis for procedure changes and proposed modifications to prevent or mitigate the consequences of such an event.
Refer to page 8 of the attachment for additional information.
3.
Procedures and Administrative Controls 0 PPD changed procedures and developed checklists to address Decay Heat Removal related events. Refer to page 4 of the attachment for additional information.
4.
Provide Two Independent. Continuous Temperature Indications I
Continuous temperature monitoring is provided by the core exit thermocouple when the reactor vessel head is on, except while the reactor vessel head removal is in progress. OPPD will provide non-environmentally qualified cables to provide temperature indication during that period.
Refer to page 2 of the attachment for additional information.
5.
Provide Two Independent Continuous RCS Water level Indicator When the System is in a Reduced Inventory Condition OPPD has one level RCS level system installed and currently uses tubing as a second refueling level instrument. A second level system will be installed.
Refer to page 2 of the attachment for additional information.
OPPD believes the actions described in this letter and the attachment will provide for complete compliance with Generic Letter 88-17, Decay Heat Removal.
OPPD is evaluating the methods available to accomplish the actions described herein. An implementation schedule will be provided within 60 days of this letter.
During this time OPPD will be processing the modification stated in this letter through our nuclear planning process. This effort includes prioritization of resources to develop a detailed schedule.
U. S. Nuclear Regulatory Commission LIC-89-045 Page 3 l
On November 30, 1988, and December 1, 1988, Mr. Warren Lyon and Mr. Harry Balukjian, Nuclear Reactor Regulation, visited the Fort Calhoun Station.
During the visit a copy of the analysis (Reference 4), which is requested in item 4 of the attachment, was given to them.
This response is being submitted under oath pursuant to the provisions of Section 182a of the Atomic Energy Act of 1954 as amended.
The submittal date of February 10, 1989, was discussed with Mr. P. D. Milano, NRC Project Manager, and members of my staff.
If you have any questions concerning this matter, please contact us.
Sincerely,
/W $9' lO f,t. J. Morris K
f Division Manager Nuclear Operations KJM/sa Attachment c:
LeBoeuf, Lamb, Leiby & MacRae T. E. Murley, Director, Office of NRR R. D. Martin, NRC Regional Administrator P. D. Milano, NRC Project Manager t
P. H. Harrell, NRC Senior Resident Inspecter
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_ UNITED STATES OF AMERICA NUCLEAR REGULATORY COM4ISSION l
In the' Matter of Omaha Public Power District Docket No. 50-285 (Fort Calhoun Station Unit No. 1)
'l W. G. Gates, being duly sworn, hereby deposes and says that he is'the Manager -
Fort Calhoun Station of the Omaha Public Power District; that as such he is duly authorized to sign and file with the Nuclear Regulatory Commission the attached information concerning the response to NRC Generic Letter 88-17; that he is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of his knowledge, information,.and belief.
AV. 5 NAha W. G. Gates Manager - Fort Calhoun Station I
STATE OF NEBRASKA ss COUNTY OF DOUGLAS Subscribed and sworn to before me, a Notary Public in and for the State of Nebraska on this //ra day of February, 1989.
/ MAL ~~
Notary Public GERAL NOTARY State of serska J.T. OLEASON
% bI4 42519 b
'U. S. Nuclear Regulatory Commission
- LIC-89-045 ATTACHMENT Programmed Enhancements Item 1 Instrumentation Provide reliable indication of parameters that describe the state of the RCS and the performance of systems normally used to cool the RCS for both normal and accident conditions. At a minimum, provide the following in the CR:
(a)twoindependentRCSlevelindications (b)atleasttwoindependenttemperaturemeasurementsrepresentativeofthe core exit whenever the RV head is located on top of the RV (We suggest that temperature indications be provided at all times.)
(c) the capability of continuously sonitoring DHR system performance whenever a DHR system is being used for cooling the RCS.
(d)visibleandaudibleindicationsofabnormalconditionsintemperature, level, and DHR system performance.
Response
RCS Level Indications A modification to the existing RCS refueling level instrumentation was performed in response to NRC Generic Letter 87-12 item 9, to provide low level alarm indication in the control room. The modification to LI-197 included replacement of the existing level transmitter with a more accurate level transmitter, an alarm function, and the ability to provide input into the plant computer for continuous readout and trending capability.
The modification as completed provides the operators with a method for establishing trends and obtaining a precise level indication.
It also provides early warning of potential low level by the audible and visual alarm which alarms at 2 inches below the centerline of the hot leg. The remote control room RCS refueling level instrument is LI-197 which spans between the bottom of the hot leg, to the bottom of the pressurizer, an indicating range of approximately 225 inches.
The new level instrument loop has the following instrument inaccuracies; LT-197 level transmitter of 2 1.98 inches, LA-197 level alarm of 1.7 inches, L197 computer level point of
- 1.98 inches, and LI-197 control 2
board level indicator of + 7.16 inches, - 7.06 inches.
The computer readout is in inches.
The control board instrument range of 0-100 percent corresponds to 0-240 inches.
During the 1988 refueling and maintenance outage, following full core off-load to the spent fuel pool, the RCS was drained to below the bottom of the hot leg to facilitate maintenance of items attached to the cold legs. To determine the point of pump vortexing, the low pressure safety injection pump was left running while draining down. The indications of vortexing appeared at 7 inches below centerline of the hot leg, as indicated by fluctuating motor amerage.
This is a point not commonly observed during refueling outages and is well below the alarm setpoint including instrument inaccuracies.
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'U. S.* Nuclear Regulatory Commission LIC-89-045 Currently the Fort Calhoun Station also uses tubing as a refueling level instrument.
The tubing is installed off a line at the bottom of the hot leg, (approximateelevation994 feet)andextendstoanelevationof1056 feet.' The end of the tubing is attached to a railing on the 1056 foot level.
A modification to the tubing installation methodology will be completed during the 1990 refueling outage.
A modification to install a second control room RCS refueling level instrument will also be completed.
It is intended that this modification be a narrow range instrument with a span from the bottom of the hot leg to the top of the hot leg, with more accurate instrumentation than currently installed. The modification, due to limited piping penetrations on the bottom of the hot leg, will be required to utilize the same instrumentation taps as LI-197. Although this approach does not provide full independence, it is not desirable to create any more penetrations in the RCS pressure boundary. The procedures which govern the use of the two level instruments will include guidance concerning instrument line plugging. An alarm function will be provided with input to the plant computer for trending purposes. This modification will be installed during the 1990 refueling outage.
RCS Temperature Indications The reactor vessel is instrumented with 28 core exit thermocouple (CET)
'3 installed with the reactor incore instrumentation strings.
The CET indications can be accessed at either the Qualified Safety Parameter Display System (QSPDS) display panels, 2 channels installed, or at any CRT used with the plant
- computer, j
The CETs are equipped with an alarm function through the plant com) uter.
If CETs are used when the reactor coolant level is below the top of tie hotleg and shutdown cooling is in service, two representative CET temperatures will be recorded by operators in the control room on regular hourly intervals e.g.,
routine hourly readings. An alternate temperature reading device will be available if the computer is inoperable.
This will be incorporated into procedures prior to entry into mid loop operations.
The configuration of the RV and support equipment does not readily facilitate the use of CETs while RV head removal is in progress.
It is necessary to remove the cable trays, which contain the environmentally qualified cables, prior to removal of the RV head. The cable tray removal may proceed RV head removal by several days and its removal disables the CET readout capability.
In order to ensure core temperature indication is available during this transition aeriod, OPPD will procure two non-environmentally qualified CET l
cables whic1 will be used for control room temperature instrumentation while RV head removal preparations are progressing, whenever the RCS level is less than the top of the hot leg.
The two non-environmentally qualified CET cables will be installed prior to the removal of the cable trays and remain installed until RV head removal preparations have been completed. The two non-environmentally qualified CET cables will be reinstalled following RV head reinstallation and remain installed until ready for cable tray replacement. OPPD will procure the two non-environmentally qualified cables for use during the 1990 refueling and maintenance outage.
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U. S. Nuclear Regulatory Commission 1
LIC,89-045 OPPD will also evaluate by the end of the 1990 refueling outage, the feasibility of installing temporary temperature instrumentation while the RV i
head is removed, as suggested in Generic Letter 88-17.
CHR System Performance The instrumentation currently installed to monitor shutdown cooling operation includes the RCS level instrumentation described earlier, shutdown cooling flow, safety injection and containment spray (pump motor amperage, shutdown cooling outlet temperature, CET indications if installed), and safety injection header pressure, all of which is displayed in the control room.
Safety injection and containment spray discharge local pressure guages can be used to verify pump discharge pressure if required by Operations. The monitoring of these parameters from the control room combined with the auxiliary building operators routine inspections of operating equipment provide assurance that the system is performing as required.
Visible and Audible Indications The alarms currently installed to monitor shutdown cooling operation include the Refueling Level Low Alarm described earlier, Low Pressure Safety Injection (LPSI) Pump Off Normal, Safety Injection Pump Valves Off-Normal, Shutdown Cooling Valves Off-Normal, and Shutdown Cooling Valves Closed Sig Fail or Violation. The LPSI Pump Off-Normal alarm indicates a breaker-to-control switch mismatch. This condition may be present if the pump has tripped due to either an electrical or mechanical problem. The Safety Injection Pump Valves Off-Normal alarm indicates that any one of the suction or discharge valves are in the closed position. The Shutdown Cooling Valves Off-Normal alarm indicates a
HCV-347 and I
when any(of the following valves are in the closed position:ShutdownCoolingContai HCV-348 CoolingHeatExchangerFlowControlvalve),andFCV-326(ShutdownCoolingHeat ExchangerBypassvalve).
The Shutdown Cooling Valves Closed Sig Fail or Violation alarm indicates that the autoclosure interlock installed on HCV-347 and HCV-348 has received a signal to close the valves.
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U. S. Nuclear R::gulatory Commission LIC-89-045 Ites 2 Procedures Develop and implement procedures that cover reduced inventory operation and that provide an adequate basis for entry into a reduced inventory condition.
These include:
(a) procedures that cover normal operation of the MSSS, the containment, and supporting systems under conditions for which cooling would normally be provided by DHR systems.
(b) procedures that cover emergency, abnormal, off-normal, or the equivalent operation of the NSSS, the containment, and supporting systems if an off-normal condition occurs while operating under conditions for which cooling would normally be provided by DHR systems.
(c) administrative controls that support and supplement the procedures in items (a), (b), and all other actions identified in this conamnication, as appropriate.
Response
Normal Operatino Procedures In response to the analysis described in response to item 4, OPPD has revised the following procedures. Specific details of the changes were included in Reference 5.
A.
OperatingProcedureOP-6,(HottoColdShutdown),isutilizedtocontrol major plant evolutions to manipulate the plant from the hot shutdown to a cold shutdown condition.
B.
Operating Instruction, 01-C0-4 " Establishing Modified Containment Integrity", is used for containment penetration isolation during refueling operations, RV closure head lift for removal or replacement, and mid-loop refueling containment integrity.
C.
Operating Instruction 01-RC-2A, " Reactor Coolant Fill Instruction," is utilized to make level changes in the RCS which result in inventory increases.
D.
Operating Instruction 01-RC-4, " Reactor Coolant System Normal Shutdown",
is utilized to bring the RCS from hot shutdown to cold shutdown condition in a controlled manor.
E.
Operating Instruction 0I-RC-5, " Reactor Coolant System Draining", is utilized to make RCS level changes which result in inventory decreases.
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'U. S.' Nuclear R:gulatory Commission LIC-89-045 F.
Operating Instruction 0I-SC-1, " Initiation of Shutdown Cooling", is utilized in establishing and maintaining snutdown cooling flow.
G.
Surveillance Test ST-SDC-1, " Shutdown Cooling Valve Interlock Test", is utilized to test the functioning of the autoclosure interlock feature of the shutdown cooling system. The Fort Calhoun Station is currently evaluating the need for this interlock in conjunction with the Combustion Engineering Owners Group.
H.
MP-N0ZDAM-1, " Steam Generator Nozzle Dam Installation Procedure",
requires either the pressurizer manway or reactor vessel head to be removed prior to installation of the nozzle dams.
The procedure also requires that the nozzle dams be installed in the cold legs first, followed by the hot legs.
I.
MP-N0ZDAM-2, " Steam Generator Nozzle Dam Removal Procedure", requires that the nozzle dams be removed from the hot legs first, followed by the cold legs.
J.
MP-RC-2-1-A, " Steam Generator Primary Manway Removal", recuires that the hot leg steam generator manways be removed first, followec by the cold leg steam generator manways.
K.
MP-RC-2-1-B, " Steam Generator Primary Manway Replacement", requires that the cold leg steam generator manways to be installed first, followed by the hot leg steam generator manways.
A review of the procedures which govern shutdown cooling operation to the i
I criteria set forth in Generic Letter 88-17 revealed the lack of a requirement for a high pressure injection source.
Procedures OP-6 and 01-SC-1 will be revised to require that one high pressure injection source with a capacity greater than 55 gpm, be available when shutdown cooling is required to be operable. Also identified was a lack of concise flow paths to accomplish actions saecified in A0P-19. Procedures outlining the flow paths needed to accomplis 1 the actions specified in A0P-19, and the requirement for a high pressure injection source, will be written and implemented prior to the 1990 refueling outage or prior to mid loop operation if required before refueling.
The procedures will include the flow paths for the following.
1.
The charging pumps from the boric acid storage tanks or the safety injection refueling water tank to the RCS.
2.
The high pressure safety injection pumps from the safety injection and refueling water tank or the containment sump to the RCS.
3.
The containment spray pumps from the safety injection and refueling water tank or the containment sump to the RCS.
Emaraency/ Abnormal Operatina Procedures Abnormal Operating Procedure A0P-19, Loss of Shutdown Cooling, is utilized to recover from and to mitigate the consequences of a loss of shutdown cooling.
This was upgraded, including a complete technical and human performance review.
Changes resulting from the technical review included those actions required to restore shutdown cooling to operation, along with the actions required to initiate and maintain an alternate shutdown cooling mechanism.
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'U. S. Nuclear Regulatory Commission LICa89-045 Incorporated into the procedure were the required injection flow rates to maintainthecorecooled(575gpm)andcovered(55gpm). The procedure also includes time to boil curves which are based on plant specific decay heat values. A listing of critical plant elevations is included as reference to installed instrumentation. The procedure also provides guidance for containment closure following a loss of shutdown cooling.
Following the technical review, Senior Reactor Operators ana System Engineering, conducted a plant specific systems review to validate the procedure.
A review of the procedure de srmined that additional guidance would be beneficial to ensure that per sonnel are evacuated from containment in a more timely and orderly fashion. Fcrsonnel safety must be included when the potential exists for steam to be released from the RCS. T!ae procedure will be revised prior to the 1990 refueling outage or prior to mid-loop operation if required before the refueling outage.
Administrative Controls The Fort Calhoun Station is in the process of developing procedural controls governing the equipment required when operating in a reduced inventory mode.
These additional controls will be implemented prior to the 1990 refueling outage.
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'Uc S. Nuclear Regulatory Commission LIC-89-045 Item 3 Equipment (a)
Assure that adequate operating, operable, and/or available equipment of high reliability is provided for cooling the RCS and for avoiding a loss of RCS cooling.
l (b)
Maintain sufficient existing equipment in an operable or available status so as to mitigate loss of DHR or loss of RCS inventory should they occur. This should include at least one high pressure injection pump and one other system. The water addition rate capable of being provided by each equipment item should be at least sufficient to keep the core c e red.
(c)
Provide adequate equipment for personnel communication that involve activities related to the RCS or systems necessary to maintain the RCS in a stable and controlled condition.
Response _
Adeauate/ Sufficient Eauipment Available The specific equipment required to prevent or mitigate the consequences of a loss of shutdown cooling are administratively controlled by the procedures described in response to item 2, and were discussed in detail in Reference 5.
Plant Communication A review will be conducted to determine the adequacy of the installed Fort l
Calhoun Station communications. Communications will be upgraded where appropriate.
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'U. S.* Nuclear Regulatory Commission LIC+89-045 Item 4 Analyses Conduct analyscs to supplement existing information and deYelop a basis for procedures, instrumentation installation and response, and equipment /NSSS interactions and response. The analyses should encompass thermodynamic and physical (configuration) states to which the hardware can be subjected and should provide sufficient depth that the basis is developed. Emphasis should be placed upon obtaining a complete understanding of NSSS behavior under nonpower operations.
Response
OPPD, in the response to NRC Generic Letter 87-12 Item 9, describes a site specificanalysiswhichhasbeenprovidedbyCombustionEngineering(reference 4). The analysis was designed to give the Fort Calhoun Station a basis for procedures, administrative control, modifications, and other actions which reduce the probability of a loss of shutdown cooling or mitigate the consequences of the loss of shutdown cooling. Results of the analysis were discussed in Reference 5 and are summarized below.
I Core Uncoverv The )ossibility of Core Uncovery is a concern whenever the RCS temperature reacles an elevated level for the operating condition the plant, or allowed to reach saturation.
The analysis determined that in the wrong configuration, the
)otential exists to pressurize the upper plenum, force water through a cold leg 1 ole, and result in core uncovery and fuel damage. The analysis also demonstrated that the RCS can achieve saturated conditions in less than 30 i
minutes and without a sufficient vent, core uncovery could result.
To preclude this condition the pressurizer manway will be removed prior to any operation with reduced RCS inventory (i.e. 3 feet below the top of the RV flange), unless a steam generator is available. The manway provides sufficient venting capability to preclude core uncovery if inventory replenishment capability is maintained. The steam generator will ensure that the capability to remove decay heat is available.
Dose Assessment Dose rates were calculated based on 10 CFR 100 criteria to predict the consequences of a loss of shutdown cooling event. The radioactive source terms were calculated by multiplying)the one hundred percent power RCS specific activity, (with 1% failed fuel times the assumed RCS volume at mid-loop. A containment leakage rate of 15.1% per day, equal to the rate of boiloff, was assumed by using decay heat present 1 day after shutdown. The results predicted dose rates of 12.18 Rem (Thyroid) and 0.0774 Rem (Whole Body) at the siteboundaryand1.07 Rem (Thyroid)and 0.00746 Rem (Whole Body) at the Low Population Zone.
These are well within 10 CFR 100 limits; however, personnel working inside the containment could receive doses in excess of the 10 CFR 20 limits for exposure to airborne radioactive contaminants.
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V. S. Nuclear Regulatory Commission LIC 89-045 Containment Pressure The maximum containment pressure expected following a loss of shutdown cooling event was also calculated. Assuming the decay heat present I day after shutdown, a boiling rate of 7.5 lbm/sec would occur, which will pressurize the containment to approximately 1 psig in about 5 minutes once containment integrity is established. At this boiling rate, containment could reach 60 psig, containment design pressure, in about 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> without containment spray or containment cooling fans to reduce the pressure.
Inventory Replenishment The site specific analysis defined the flow rates needed to replenish inventory in the reactor vessel to keep the core covered. The required flow rate to replenish inventory due to boil off is 55 gpm, and 575 gpm to keep the core from boiling, decreasing with respect to decay heat load. These values are based on decay heat levels 1 day after shutdown.
The analysis provided the basis for the procedural controls discussed in response to item 2.
A0P-19, includes the use of the charging pumps, high ressure safety injection pumps (HPSI), low pressure safety injection Jumps p(LPSI), or the containment spray pumps for inventory replenishment. T1e flow paths are specified in general. The procedure directs operators to align the system to inject greater than 55 gpm to the RCS using one of the following line ups:
1.
The charging pumps from either the boric acid storage tanks or Safety Injection and Refueling Water Tank (SIRWT) to the RCS. Three charging pumps (positive displacement) are available and are designed to deliver water at a flow rate of 40 gpm each at a pressure of 2100 psia.
2.
The HPSI's from either the SIRWT or the containment sump.
Three HPSI pumps are available and are designed to deliver water at a rate of approximately 150 gpm each at a pressure of 1735 psig. (USAR values).
3.
The LPSI's from either the SIRWT or the containment sump.
4.
The containment spray pumps from either the SIRWT or the containment sump.
Three containment spray pumps are available and are desi deliverwaterataflowrateof1700gpmeach(injectionmode)gnedto at a head l
of 450 ft. of water. (USAR values).
Hot Leo injection As a possible alternate flow path for inventory replenishment, hot leg injection was evaluated. At Fort Calhoun, hot leg injection is accomplished utilizing the auxiliary spray line to inject water into the top of the pressurizer. Water then enters the hot leg through the pressurizer surge line.
If the pressurizer manway is removed to provide a vent path, the hot leg injection path would not be available. Following an extended loss of shutdown after the RCS reaches saturation the steaming rate would be approximately 7.5 lbm/sec, assuming shutdown decay heat levels present I day after shutdown.
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U. S. Nuclear Rsgulatory Commission i
LIC;89-045 At this steaming rate the steam would travel up through the pressurizer surge line with such velocity that the water ent? ring through the auxiliary spray line is entrained in the outgoing steam and does not reach the RCS. Based on this, hot leg injection is not a viable option, when the pressurizer manway is removed.
If the manway is installed, hot leg injection is a viable option.
The analysis showed, the manway removed is an acceptable configuration for mid-loop operation provided cold leg integrity is maintained. Cold leg injection with cold leg integrity provides more than adequate decay heat removal capability.
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U0 S, Nuclear Regulatory Commission LIC-89-045 Item 5 Technical Specificaife s, TechnicalSpecifications(TSs)thatrestrictorlimitthesafetybenefitofthe actions identified in this letter should be identified end appropriate changes should be submittc.d.
Response
A review was conducted of Technical Specifications and it was determined that there are no restrictions which would prohibit the actions specifically stated in this response from preventing or mitigating the consequences of a loss of shutdown cooling.
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. U. S. Nuclear Regulatory Commission LICJ89-045 Item 6 RCS Perturbations Item (5) of expeditious actions should be reexamined and operations refined as necessary to reasonably minimize the likelihood of loss of DHR.
Response
The Fort Calhoun Station is in the process of developing procedural controls governing the equipment required when operating in a reduced inventory mode.
These additional controls will be implemented prior to the 1990 refueling outage.
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