ML20058G886
| ML20058G886 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 07/28/1982 |
| From: | Fay C WISCONSIN ELECTRIC POWER CO. |
| To: | Clark R, Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.F.2, TASK-TM NUDOCS 8208030440 | |
| Download: ML20058G886 (28) | |
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Wisconsin Electric powea couesur 231 W. MICHIGAN, P.O. B0X 2046, MILWAUKEE. WI 53201 July 28, 1982 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Mr. Robert A. Clark, Chief Operating Reactors Branch #3 Gentlemen:
DOCKET NOS. 50-266 AND 50-301 REPLY TO NRC REQUEST FOR INFORMATION REACTOR VESSEL WATER LEVEL INSTRUMENTATION SYSTEM POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 Attached to this letter is our reply to your letter of June 19, 1982, in which you requested additional information on the proposed Reactor Vessel Water Level Instrumentation System for the Point Beach Nuclear Plant, Units 1 and 2. Please contact us if you have any additional questions regarding this information.
Very truly yours, y,
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,c c - r Assistant Vice President C. W. Fay Attachment Copy to: NRC Resident Inspector
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. i REPLY TO NRC REQUEST FOR ADDITIONAL INFORMATION REACT 0E VESSEL WATER LEVEL INSTRUMENTATION SYSTEM POINT BEACH NUCLEAR PLANT
, JULY 28, 1982 i
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- 1. According to your submittal dated October 20, 1981 and responses to questions to Westinghouse submittal (your attachment #2) (specifically responses to question numbers 12,21,22,and 23), your system is substan-tially different from the Westinghouse system. Provide a detailed des-cription with respect to the itemized documentation requirements per NUREG-0737, Item II.F.2.
The system description of the Reactor Vessel Water Level Instrumentation System was provided by our submittal dated October 20, 1981. The schedule for installation of this system and other TMI items was provided in our submittal dated April 26, 1982. A description of the thermocouple upgrade is included with this transmittal as Attachment 19A. An analysis of system accuracy is expected to be available by the end of September 1982. Appli-cable generic procedures are discussed in the response to Question 4.
The analog Reactor Vessel Water Level Instrumentation System using the Foxboro SPEC 200 racks is the system designed to meet the water level indicating requirements of NUREG-0737. In addition to the analog computation of water level, water level will be computed in two independent areas of the plant computer system. Water level is computed in the supermultiplexers and displayed on command on a twelve line by forty characters per line LED display located on the Auxiliary Safety Instrumentation Panel. The computation performed in the supermultiplexers is independent of the computer system CPUs.
Water level will also be computed using the Safety Assessment System CPUs.
This computed water level is for displayed on any of the Safety Assessment
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Syst.em or Plant Process Computer System CRT displays. These CRTs are located in the control room, computer room, Technical Support Center and Emergency Operating Facility.
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A block diagram showing one of two redundant trains of the overall analog and digital computation of Reactor Vessel Water Level is included as attachment 1A.
In Attachment 1A, the signal paths are as follows:
- 1. Five thermocouples in each of two trains in Unit 1 and four thermo-couples in each of two trains in Unit 2 are connected to their res-pective multiplexers.
- 2. One wide range and one narrow range differential pressure transmitter are connected to the Foxboro SPEC 200 analog racks in each of two trains.
- 3. One wide range pressure transmitter is connected to the analog racks in each of two trains.
- 4. Nineteen core exit thermocouples in one train and twenty core exit thermocouples in another train are connected to their respective multiplexers.
- 5. The multiplexer provides two temperature signals to the analog racks in each train. One is a weighted average of the sensing line temper-atures and the other is the average of selected core exit thermo-couple temperatures.
- 6. The analog racks drive a Narrow Range Reactor Vessel Water Level and a Wide Range Reactor Vessel Water Level display in each of two trains. These indicators are located on the Auxiliary Safety Instru-mentation Panels in the control room.
- 7. The analog rack outputs are connected as inputs to the multiplexer.
These signals include Wide Range Reactor Coolant System Pressure, Wide t
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Range Reactor Vessel Differential Pressure, Narrow Range Reactar Vessel Differential Pressure, Wide Range Reactor Vessel Water Level and Narrow Range Reactor Vessel Water Level in each of two trains.
- 8. There are four fiber optics cables, one from each multiplexer, to each of two supermultiplexers.
- 9. There are four fiber optics cables, one from each multiplexer, to each of four CPUs in the plant computer system.
- 10. One fiber optics cable connects each supermultiplexer to its digi-tal display located in the Auxiliary Safety Instrumentation Panel.
- 11. Fiber optics cables connect the plant computer system to the Safety Assessment System CRT displays. Four SAS CRT displays are located in the control room, one in the Technical Support Center and one in the computer room.
- 12. Fiber optics cables connect the plant computer system to the Plant Process Computer System CRT displays. Four PPCS CRT displays are located in the control room, two in the Technical Support Center, one in the computer room, and one in the Emergency Operations Facility.
- 2. Since your system is different from the Westinghouse system, the devel-opment of this level measurement system should include a test program to verify how the system works. Provide information about what to observe?
When will this data be available?
Although the hardware is different, the Wisconsin Electric system measures differential pressure in a manner similar to the Westinghouse differential pressure system. No separate test program will be provided for dynamic condi-tions. It is planned to perform a static verification test during filling and venting of the reactor vessel after refueling. During this test, the
water level indicated by the system will be compared to the water level indicated by the level system used during refueling.
Since the reactor vessel water level system uses the computer system MUXs to provide part of its compensation, the system will not be fully operational until early 1984. During the first refueling in 1984, the initial static testing of the system will be performed. Information on this test will be available after the test data have been evaluated.
- 3. Provide an analysis of system accuracy, listing the separate contribu-tions from each component and hcw they are combined for the estimate of the total system accuracy.
An analysis of system accuracy has not been completed as yet. The analysis is expected to be completed and will be transmitted to the NRC by the end of September 1982. Included in the overall analysis will be a static error analysis which includes such items as transmitter and amplifier calibration accuracy and indicator reading accuracy. Also included will be an analysis -
of accuracy under accident conditions which will consider the effects of the containment conditions on the differential pressure transmitters. Environ-mental testing of the transmitters is presently in the post LOCA phase and preliminary results of the LOCA testing are expected to be available at that time. Also an analysis of accuracy after an ICC event will be included.
- 4. Your response to Question 7 indicated that your level measurement instru-mentation will not be used for the operator to make decisions to initiate action to recover the plant from the accident and prevent damage to the core. What then is the basis for operator decisions? Please relate your responses to various size breaks.
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The response to Question 7 stated that the Reactor Vessel Water Level System does not dictate operator response to large break or small break LOCAs as provided in Emergency Operating Procedures. The response also stated that the operator uses other parameters for decisions and uses the level system only to provide additional information and as an indicator of the possibility of inadequate core cooling.
The response that was given to Question 7 can be clarified with the following discussion and attached material. In order to understand how the reactor vessel water level system will be used, the Emergency Response Guidelines need to be studied to ascertain when and how the operator will use this information. The Emergency Response Guideline set was sent to the NRC by the Westinghouse Owner's Group in letter OG-64, dated November 30, 1981, Robert Jurgensen to D. G. Eisenhut. The statements made in the response to Question 7 stated that the operator uses other parameters for decisions, this means that in dealing with a LOCA the operator does not use reactor vessel water level alone in making decisions. The Emergency Response Guidelines do not use reactor vessel water level as a parameter the operator uses to determine how l to mitigate a design basis accident, i.e., which procedure is to be followed.
l l Nor is reactor vessel water level used to terminate or reinitiate safety injection. If for some reason, known or unknown, the guideline the operator l is following does not accomplish its objective of providing core cooling and restoring primary system inventory, the reactor vessel water level indication
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l is 'available for use as a diagnostic aid for response to inadequate core l
l cooling.
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The two key indicators of inadequate core cooling, or the approach to it, are core exit thermocouples and reactor !essel water level. Even if the reactor vessel water level system gives completely erroneous information, the core exit thermocouples will indicate when inadequate core cooling has started and the appropriate emergency response guideline provides the steps necessary to mitigate the situation. In addition, vessel water level by itself is not used by the operator to make decisions or take actions. The operator must have abnormal containment conditions along with the low indication on reactor vessel water level before he can take any mitigating actions. Attachments 4A, 48, 4C, and 4D summarize the key actions taken in procedures E-0, E-1, E-2, and E-3 and the parameters used by the operator to make the decision to take these actions. A review of those attachments will show that reactor vessel water level is used as a symptom of inadequate core cooling and water level indication is used in conjunction with other parameters.
The procedures are structured independent of break size and the operator responds based on symptoms which are independent of break size. The proce-dures have been analyzed for large and small break LOCAs, including isolat-ible LOCAs, and they have been written to properly mitigate all break sizes with the same procedure. The analysis shows that the parameters will have varying degress of response in magnitude and timing, but the symptom-based procedures properly deal with these differences.
The reactor vessel water level indicating system will also be used to indi-t _
l cate bhen a bubble exists in the upper head of the reactor vessel. _ If no other major action is in progress, e.g., responding to an accident, the .
l operator is directed to a procedure which determines if the bubble can be 1
collapsed. If not, the bubble is considered non-condensible and the operator will have to decide if he should vent the reactor vessel head using the gas l
vent system. ,
i The use of the reactor vessel water level system, as described above, occurs ;
at a time in the transient when the system parameters have settled down. In
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case of a 1arge LOCA, blowdown is over and reflooding is occurring when the reactor vessel water level indication may first be utilized by the operator.
If for some reason reflooding does not occur, symptom set III of attachments 4A through 4D would probably be the first indicator of inadequate core cooling.
Even for this case, it is tens of minutes into the transient before the symptoms of set III occur before and the vessel level indication would be used by the operator. For the small break LOCA it will take even longer to lose enough inventory to approach or result in inadequate core cooling and the dynamic effect of fluid flow on the accuracy of water level indication during this slow transient is minimal. A general conclusion can be reached concerning the fluid flow effect on the accuracy of the vessel water level indication; whenever there is enough fluid flow through the vessel to cause a flow induced AP error in water level indication, that flow is sufficient to cool the core and inadequate core cooling will not exist.
! 5. Your response to Question 8 indicates expected response time of fractions
! of a second. Show a calculation of the expected response time using typical line diameters, lengths and differential pressure transmitter volumetric displacement for given pressure change. This response time should be measured and reported as a response to this question (using
, the specific transmitter, line diameters and line lengths). What is the required response time for your system?
The response time, to 90% of input, of the differential pressure transmitters to a step input of 50% of span is approximately 0.5 to 0.7 seconds. This information was provided by the manufacturer.
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There are approximately 172 feet total of tubing connecting the upper and I lower reactor vessel level fluid connections to the transmitters. This tubing has an internal diameter of 0.245 inches. Per manufacturer information, the volumetric displacement of a transmitter for full span indication is less than one cubic centimeter. The response tine of the transmitter with tubing attached is expected to be less than one second. We have no plans to attach long lengths of tubing to a transmitter for a bench test. The response time.
to 90% of input, of the indicator to a 50% of full scale step input is approxi-mately one second.
The required response time for the system need not be any further than 30 seconds.
- 6. Your response to Question 11 indicates that overranging "is not expected" to affect d/p transmitter perfortnance. This question should be answered with data from test programs or past performance (well documented) for the specific transmitter type intended for this system. Provide the basis for "is not expected".
The narrow range transmitter is a Foxboro type NE13DH which will be calibrated to a span somewhat less than 500 inches of water. With reactor coolant pumps operacing, an additional 40 psi differential pressure (approximately 1100 inches water) will be added causing an overrange condition on the transmitter. The capsule used in this transmitter is rated for and will not be damaged with differential pressures up to 3000 psi. Overrange pressures up to 3000 psi can cause a zero shift, but will otherwise not affect calibration or operation of the transmitter. As part of the manufacturing process, each transmitter is pv.erranged to 1500 psi in each direction with the span set at approxi-mately 200 inches water. The zero shift must be less than 3% of span between overranging in different directions for the transmitter to continue in the manufacturing process. The effect of the zero shift is inversly proportional
l to the span so the equivalent effect on a transmitter set to a span of 400 inches i
would be 1 %. We do not have test program data or documented performance data for the magnitude of overrange that is expected to occur. Since the overrange expected is small compared to what is allowed and to what has been tested, the zero shift, if any, is expected to be small, less than 0.5%.
This conclusion is based on discussions with manufacturer engineering personnel.
- 7. Provide your expected error range for your temperature and pressure com-pensation of differential pressure measurements. Describe how this error range will be factored into system operation.
An error analysis, which will. include temperature and pressure compensation of the differential pressure measurements, will be submitted by the end of September, 1982. See the response to Question 3.
s The resulting error, or uncertainties, will be added to the value specified in the generic emergency response guideline. As an example, refer to the attachments to Question 4. The footnote for response to inadequate core s cooling Symptom Set III deals with the value for reactor vessel water level indication, including uncertainties, i
l 8. Please answer Question 15, what is the source of the tables or relation-ships used to calculate density. corrections for the level system?
The source of the tables used to calculate density corrections is the ASME Steam Tables,, Fourth Edition, 1978.
9'. ~ What are the provisions to insure that the impluse lines remain full? '
How can an empty line be detected? What are the corrective actions?
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A seal chamber is located at the high point of the system. The seal chamber is a water reservoir which will refill any void in the impulse line tubing leading from the seal chamber to the transmitters. The seal chamber is sized such that one-half its volume will refill a completely voided impulse line.
4 Refer to system diagram provided in Attachment 9A. The impulse line from the bottom of the vessel to the transmitters will remain filled because the transmitters are located at an elevation slightly below the bottom of the core. It is expected that the water level in the vessel will always be greater than the bottom of the core.
The impulse line cannot empty unless the seal chamber is also empty. The seal chamber is designed to prevent an empty line from occurring. An empty or partially filled impulse line cannot be detected. Should a partially filled impulse line occur, it would result in an additional error in the indicated reactor vessel water level. The Emergency Response Guidelines are written in a manner that protects against operator errors even if the vessel level system has greater than expected errors. This includes errors as great as off-scale high or off-scale low. Refer to the response to Question 4 on how the operator will use reactor vessel water level indication in the Emergency i
! Response Guidelines.
The following situations are postulated to evaluate the operator response with an erroneous reactor vessel water level indicating system.
Case 1 - Actual level sufficient to cool the core, indicated level off-
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scale high. This could be caused by a broken upper impulse line, an impulse line with a large fraction of voids, or a s
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system break that results in lower pressure in the head area.
If this were to occur, the wide range level indicator would be off-scale high and the operator would suspect a malfunctioning system because the wide range instrument is designed not to go off scale high. Even if the operator did not realize this, he would still respond to the situation in the same manner because his actions are not based on reactor vessel water level indica-tion. The only thing he might miss would be a bubble in the upper head and should the operator not deal with that, no serious consequence would result. It is interesting to note that an upper head break or broken impulse line would vent that bubble anyway.
Case 2 - Actual level is not sufficient to cool the core, indicated level in off-scale high or reading high. The causes for this erroneous reading are the same as Case 1. For smaller posi-tive errors, lesser amount of voids in the sensing lines or vessel flow induced AP are needed. In this case, the vessel level system would not provide indication of inadequate core cooling at the expected time but the core exit thermocouples would. The emergency response guidelines require that the operator respond to inadequate core cooling on core exit thermocouples alone. Refer to attachments to Question 4, Sympton Set I.
Case 3 - Actual level is a full vessel, indicated level is off-scale low. The only causes of this would be a failed lower impulse line or a large reverse vessel AP. Since the vessel is never expected to be empty, the operator would look at core exit thermocouples and if the reading is 700*F or less, the operator would probably consider the system malfunctioning. If the core exit thermocouples are reading greater than 700*F, he probably would respond to inadequate core cooling. Should the core exit thermocouples go above 1200*F, the operator would definitely respond to inadequate core cooling.
Case 4 - Actual level is a fuli vessel, indicated level shows a bubble.
Malfunction of system compensation could cause this to occur.
The bubble in the head procedure would first call for an attempt to collapse the bubble by increasing pressure. With no change in bubble size, the operator might suspect instrument compensation error. Should the operator not recognize the problem at this time and proceed with venting of the bubble, no serious consequence would result. Liquid would be vented instead of gas and the operator should realize that he has an instrumentation problem when the bubble does not decrease in size.
Case 5 - Actual level is sufficient to cool the core, indicated level is less than setpoint for inadequate core cooling. This could be caused by the reason given in Cases 3 or 4. Should this l occur, the operator would not take actions to mitigate inade-l quate core cooling until the core exit thermocoupler exceed
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700 F. Reliance on core exit thermocouple readings prevents I
the operator from taking action based solely on reactor vessel water level.
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- 10. In addition to your response to Question 20 indicating that the RVWLS will have an ambiguous response describe under what conditions would the response be ambiguous? What operating procedures are provided to alert operator of possible ambiguous indications.
Refer to the response given to Question 9. As described in the response to Question 9, the emergency response guidelines are written to accommodate and deal with errors in the reactor vessel water level indicating system. Refer also to the response to Question 4.
The Wisconsin Electric reactor vessel water level indication system utilizes vessel differential pressure to determine water level. The Wisconsin Electric system uses the same principles as the Westinghouse RVWLS. The response of vessel AP as an indicator of water level during transients and stable conditions will be the same for the Wisconsin Electric system as it is for a Westinghouse system. The only differences between the Wisconsin Electric and Westinghouse system occur in how the AP is sensed and processed electronically.
- 11. Your response to Question 23 is inadequate. How does your system respond
, to voids in the vessel? What tests are to be run or have been run under these conditions?
l The narrow range instruments respond directly to voids in the fluid. For l
example, with reactor coolant pumps off, if the voids amount to 10% of the height of the fluid within the measurement range, the narrow range display will indicate a water level equal to 90% of its normal display. With reactor l
coolant pumps running the presence of voids in the fluid will cause a decrease l in the level indicated by the wide range instruments.
After installation of the system, during filling and venting of the reactor vessel after refueling, it is planned to test the system under static condi-tions. See the response to question 2.
- 12. Provide an analysis in response to Question 24 dealing with expected accuracy after an ICC event.
Expected accuracy after an ICC event will be included in the analysis of sys-tem accuracy to be transmitted to the NRC by the end of September 1982.
Please see the response to question 3.
- 13. Justify that the single computer for level calculation meets the 99%
availability requirement of NUREG-0737.
The Reactor Vessel Water Level Instrumentation System is essentially an analog system. The only portion of this system that is dependent on a digi-tal computer are the thermocouple inputs. Only the multiplexer portion of the computer system is required to provide these inputs, the CPU's are not required. The computer multiplexers as well as the analog electronics are redundant. With redundant multiplexers, the availability of these tempera-ture signals is greater than 99%. The computer system will also be used to compute water level to augment the analog calculation.
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See the response to question 1.
- 14. Justify that the location of your pressure transmitters inside the con-tainment will provide satisfactory operation during transients which
~ ~ degrade containment environment. Describe how this impacts the mainten-ance, calibration, or replacement of the transmitters and their ability to continue to provide reliable information when containment access is not possible.
The differential pressure transmitters are being tested during an extensive environmental test program conducted per IEEE 323-1974 by the Utility Trans-mitter Qualification Group. The LOCA/HELB testing was recently completed and the testing program is in the post-accident phase. This environmental test program will demonstrate that the transmitters will provide satisfactory opera-tion during and after transients which degrade containment environment. The transmitters will provide reliable information without maintenance, calibra-tion, or replacement when containment access is not possible.
- 15. In your response to Question 3, you describe calibration and testing of the electronic modules. What provisions will be provided for calibra-tion and testing of the differential pressure transmitters?
The differential pressure transmitters will be calibrated once each year during refueling outages. Associated with each transmitter is a valve mani-fold which allows a transmitter to be isolated from its normal process connec-tions and allows the introduction of precision pressure inputs for calibra-tion. The transmitters will be calibrated by applying various pressure inputs measured by precision test gauges and using a digital voltmeter in conjunction with a precision resistor to measure the electrical output.
i 16. What effect will ambient conditions inside containment (i.e., temperature) have on the RVWLS response?
t The ambient temperature inside containment will have little effect on system accuracy because of the temperature compensation provided. Thermocouples i
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are mounted on the vertical portions of fluid lines leading to the transmitters and the outputs of these thermocouples are used to compensate for water j density changes in the fluid lines as the temperature of the lines changa due to containment ambient temperature changes.
- 17. In your system descripticn you refer to " weighted average temperature signals" on Page 5 second paragraph. Please define this term.
It is possible that there may be different numbers of thermocouples in each quadrant monitored by the RVWLS and which are used as inputs to this system.
In such a case it is desirable that the temperature of each quadrant be averaged rather than computing a straight average of all of the thermocouple readings. The proportionality of the thermocouples in the quadrant with the fewest thermocouples would be increased to obtain a value more representative of the overall quadrant average. Another possibility is that some thermocouples may be judged to represent the temperature of a larger portion of the core than others. In this case it may be desirable to increase the proportionality of those thermocouples representing the larger areas to obtain a value more representative of the core average temperature.
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! 18. In your equation on Page 6 you assume constant temperature, pressure and j density through the vessel for your calculations. We know that these parameters vary, please justify your assumptions.
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! The temperature input is a time varying average of several temperature measure-ments made using the core exit thermocouples. Since there is no mechanism for heat' input above the core these thermocouples represent, on an
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average basis, the highest fluid temperature within the reactor vessel.
Although spot temperatures in the core may be higher, the average fluid temperature will be lower.
A single time varying pressure measurement is used as an input to the system.
Two transmitters are used, one for each redundant loop. This pressure is measured at one of the main coolant loops with the pressure transmitter located at an elevation corresponding to approximately mid core. Because of the weight of the fluid in the reactor vessel, the pressure at the bottom of the vessel will be greater than the pressure at the top. Until the equivalent water level falls to approximately mid core, the pressure used as a system input will be somewhat higher than the actual pressure of the steam bubble.
This pressure effect is minor. For example, a 20 foot head of water in the reactor vessel results in approximately an 8 psi pressure difference. .
The effects of these temperature and pressure differences will be detailed further in the error analysis to be submitted in September 1983. See the response to Question 3.
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- 19. Your design used one line each for the high pressure side and low pres-sure side; these two lines are connected to four differential pressure transmitters. Please indicate line lengths, line diameters (inside and outside) and routing. A single line leak or break would make inoperable all four d/p transmitters. Please explain how this design satisfies the single failure requirement of N'JREG-0737.
The lines from the upper and lower fluid connections on the reactor vessel to the-transmitters consist of stainless steel tubing with an outside diameter of 0.375 inch and an inside diameter of 0.245 inch. The line length from the upper tap to the transmitters is approximately 92 feet and the line length
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from the lower tap so the transmitters is approximately 80 feet. The general arrangement for Unit 2 is shown in Attachment 9A.
Instrumentation for the detection of inadequate core cooling consists of two systems, core exit thermocouples and the reactor vessel water level indica-tion system. Therefore, the failure of this single impulse line will make the vessel water level system read erroneously, but it does not prevent the operator from recognizing and properly responding to inadequate core cooling using core exit thermocouples. The interaction of the use of core exit thermocouples and the reactor vessel water level system is discussed in the responses to Questions 4 and 9.
In order to provide a fully qualified system for the detection of inadequate core cooling, we are also upgrading the core exit thermocouple system. A description of this upgrade is provided in Attachment 19A.
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f0f, b ) CclaPuTTR D15PLF YS SVSTEM THEEMD-C.DllPLEs Pf1200*F - > 700* F l 2. Containment Condition - ABNORMAL ABNORMAL I 3. RCP Status - ANY ON ALL OFF l 4. RVLIS - ' <100% NR < di % NR
- 5. 5YMPTOMS FOR FR.N.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK Go to FR-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK, If AFW Pow is NOT AVAILABLE.
(1) Litre piant specsfic vehar dertved from backgroeund document. Q) Enter pient weepe shutoff head presan of Mgh-head sipwnpr pha instnunent uncertexntw or 2000 psug, whachever d lower.
- 0) Enter swn of temperenn and perwure measurewst system errors consisted into temperenn us.ng set.uetaan gebdes.
(4) Enser piant spec @c no load value. (3) Enter piant spec $c wede range dane wMeh is above top of steam senerator (J-tuber. I:b Enter piant specpc vehar dersved from background document.
- 47) Enter piant spectfic undue for shutoff head pressure of kish-head $1paunpr or low pressurxzer presan SI serpount. whachever is lower.
13 Enter piant spec @c vehan whach is 3H feet above bottom of actne)%et in core wnth :ero woadfrecsson, pha uncertsanssas. 9 of 9 1 _ ~_ _
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FOLDOUT FOR E 1 AND ES-1 GUIDE . RCP TRIP CRITIRIA O Trip any RCP if component cooling water to that pump A79)cggpr is lost. h8 O Trip all RCPs if BOTH conditions listed below are met: -
- a. Si is ON.
- b. RCS pressure - EQUAL TO OR LESS THAN m PSIG.
B 51 TERMINATION CRITERIA FOLLOWING LOS$ OF RE L
- o. Terminate SI when ,A_LL porometers listed below m (2) RCS Subcooling - GREATER THAN m*F (3) Pressurizer level - GREATER THAN S0%
(4) Heat Sinks (a) SG Level- GREATER THAN fg % NR
-OR-(b) AFW Flow - GREATER THAN m GPM
- 3. 51 REINITIATION
- o. Reinitiate Si if ANY ONE CRITERIA of the parametersFOLLOWING listed below occurs: L055 0F RE PSIG (1) RCS Pressure - LESS THAN 3 (2) RCS Subcooling - LESS THAN 3*F (3) Pressurizer Level- LESS THAN 20%
- 4. COLO LEG RECIRCULATION SWITCH 0VER C COLD LEG RECIRCULATION FOLLOWING LOSS OF R i ANY OE of the
- 5. STMPTOMS FOR FR.C.1, RESPONSE TO IN .
following symptom sets occur: STMPTOM SET 11 Ill 1 PARAMETER:
- 700*F
>1200*F ABNORMAL
- ABNORMAL 1.TCs ANY ON ALL OFF
- 2. Containment Condition <100% NR < m % NR
- 3. RCP Status
- 4. RVLIS NR i NOT AVAILABLE.
- 6. SYMPTOM 5 FOR FR.N.1 RESPONSE TO ~
000 pstg. whichenr k tower. II) Enter ;%ent specsfic value derrvedfrom background d docun
- 13) Enter sum of temperature and pressure **esurement system el des allowance for no h r is lower.
(4)reference Enter plant spenfic dnarrow range value whsch sne ure of high. head sipump seg process errors. (3) Enter plant spectfle value for shusoff hea pressu RWsT (6) Enter pient speqfic value corresponding to swstchover eierm sn pient specsfic units.bcttom of (7) Enter plant sportfic w wksch o 3H feet above R of 8
FOLDOUT FOR E-2 AND ,- ES-2 GUIDELI l
. RCP TRIP CRITERIA
$4Tr Cg/gw7 yg l o Trip any RCP if component cooling water to that pump is lost.
O Trip oil RCPs if BOTH conditions listed below are met w
- u. Si is ON.
- b. RCS pressure - EQUAL TO OR LESS THAN ffL PSIG.
- 2. $1 REINITIATION CRITERIA FOLLOWING LOSS OF SECON 8
- a. Reinitiate Sl if ANY g of the parameters listed below occurs:
(1) RCS Pressure - DECREASES BY 200 PSI AFTER Si TE (2) RCS Subcooling - LESS THAN Ift*F (3) Pressurizer level drops by 10% ofte Si termination
- 3. AFW $UPPLY SWITCHOVER CRITEltl0N
%, THEN switch to alternate AFW water supply.
IF CST level less than ts)
- 4. COLD LIG RECIRCULATION SWITCH 0VER CRITERl0N%,
IF.RWST level less than 3 RECIRCULATION FOLLOWING LOSS OF SECONDARY COOLANT,
- 5. $YMPT0M5 FOR Fit.C.1, RESPONSE To INADEQUATE ECORE t m setCOOLING Go to FR.C.1, RESPONSE TO INADEQUATE CORE COOLING occurs
~ SYMPT 0M SET 11 lit PARAMETER: I i
~
- > 700*F
>1200*F ABNORMAL ABNORMAL
- 1. TCs -
ALL OFF ANY ON
- 2. Containment Condition -
<n> % NR
- < 100% NR
- 3. RCP Status
- 4. RVLIS
- 6. SYMPTOMS F0R FR.N.1, RESPONSE TO LOSS OF SECONDARY BLE. HE Go to FR H.1 RESPONSE TO LOSS OF SECONDARY HEAT S aunt saturetwn sabins.
i!) Enter pient specsfic value denved from bechground documen pecsfic units. (3) Enter pient spenfic low level setposnt.14) Enter plant spenfk value correspo IS) Enter pient speppc' value which a 3n feet above bottqm of activefd n c , 9M9
\
k1~[8Cl}ha/VT D FOLD 0UT FOR E-3 AND ES-3 GUIDELINES -
- 1. RCP TRIP CRITERIA O Trip any RCP if component cooling water to that pump is lost.
o if a controlled cooldown is not in progress, then trip all RCPs when BOTH condit mett a
- o. 51 is ON
- b. RCS pressure - EQUAL TO OR LESS THAN g PSIG
- 2. 51 REINITIATION CRITERIA FOLLOWING STEAM GENERATOR TUBE R Reinitiate 51if ANY ONE of the parameters listed below occurs:
(1) RCS subcooling - LESS THAN al PSIG (2) Pressurizer level - LESS TNAN 20%
- 3. STMPTOMS OF LOSS OF REACTOR COOLANT DURING STEAM Go to E-1, LOSS OF REACTOR COOLANT, if abnormal containment cond to failure of PRT rupture disc.
- 4. STMPTOMS OF PRIMART TO SECONDART LEARAGE DURING RE Charging and letdown flows should be compared to deteinnno if Iwakoge ruptured stecnj generator exists.
- 5. STMPTOMS FOR FR.C.1, RESPONSE 70 INADEQUATE CORE COOLING Go to FR-C.1, RESPONSE TO INADEQUATE CORE COOLING, when ALL sy l
following symptom sets occurs: STMPT0M SET PARAMETIR: 1 il lli
- > 700*F
>1200*F
- 1. TCs - ABNORMAL ABNORMAL
- 2. Containment Condition - ANY ON ALL OFF
- 3. RCP Status - <100% NR < g % NR
- 4. RVLIS
- 6. STMPTOMS FOR FR.H.1, RESPONSE TO LOSS OF SECONDART HEAT SINR Go to FR-H.1, RESPONSE TO LOSS OF SECONDARY NEAT SINK, if AFW NOT (1) Enter plant spectl1c value dertwd from background document.Q) Enter sum t fractson plus uncertesntses.
(3) Enter plant spectfic value whach is 3 H feet above bottom of acttwfuel in core wuth zero vou 16 d 16
I ATTACHMENT 9A
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hn GENERAL ARRAN3EMENT _ REACTOR VESSEL WATER _ l Seefhe LEVEL INDIDAT' ON SYST EM _
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.. l ATTACHMENT 19A l CORE EXIT THERM 0 COUPLE UPGRADE The present core exit thermocouple system consists of 39 thermocouples which are somewhat evenly divided between two reference junction boxes in the containment. Chromel-Alumel cable is routed from stalks on the reactor vessel head through a single cable tray to a pull box and then through conduit to the cold junction boxes. From the junction boxes, Cu-Cu cable is routed to pene-tration IQ06, CU-Cu cable is routed back to 1C121 in the control room and from IC21 to the computer 1C10 in the computer room. Presently the Incore Thermo-couples are not channel related.
In the present system, the connectors at the vessel head, the cable and the cold junction boxes are non qualified. The following actions will be taken to upgrade the system:
- 1. The connectors at the vessel head will be replaced with environ-mentally and seismically qualified connectors.
- 2. The thermocouples will be split into redundant separated channels (WHITE and YELLOW). The proposed distribution was selected to maxi-mize the separation of the redundant channel, especially at the T/C stalk location, and still provide adequate distribution within each core quadrant for each channel. Each channel has a minimum of four T/Cs in each core quadrant for use in the Safety Assessment System (SAS) and Plant Process Computer System (PPCS) computer displays.
The split between channels includes separate routing for the WHITE and YELLOW T/C channels with physical separation also between T/C cable. The cable on the reactor vessel head lifting assembly will be permanently routed in seismically-mounted conduit and/or trays. The cables fo WHITE and YELLOW channels will be routed to the . refueling cavity wall via two separate moveable trays.
- 3. A second set of qualified connectors will be attached to the T/C cables near the refueling cavity wall. This allows the T/C cables on the trays to be coiled back onto the reactor vessel head lifting l assembly during reactor vessel head removal.
- 4. The cold junction boxes will be bypassed by routing qualified Chromel-Alumel cable from the vessel head directly to penetrations.
l a. All cable will be Anaconda #16 AWG, single pair, Chromel-Alumel I with shield, qualified to requirements of IEE 363-1971, and 323-1974.
~ ~~
- b. The 19 YELLOW thermocouples will be routed from the stalk on the vessel head to "new" penetration IQ21 and from the penetra-tion to "new" computer input rack C179.
- c. The 20 WHITE thermocouples will be routed from the stalks on the vessel head to "new" penetration 1Q22 and from the pene-tration to "new" computer input rack C177.
1
- d. Cable and Raceway separation will be provided.
- 5. The Chromel-Alumel Cable routed to the computer input racks C177 (WHITE) and C179 (YELLOW) vill be terminated on Uniform Temperature Reference (UTR) Plate Assemblies.
- 6. The computer input racks will be seismically qualified.
- 7. Microprocessor based multiplexer in racks C177 and C179 will compen-sate the thermocouple signals for the UTR temperatures and provide the compensated signals to the SAS and PPCS computers. They will also provide a calculated value of average incore temperature to the Spec 200 racks.
- 8. Redundant seismic supermultiplexers will be used to pole the computer input multiplexers and display the thermocouples and other Regulatory Guide 1.97 variables on demand on digital displays on the Auxiliary Safety Instrumentation Panel. The supermultiplexers and digital displays will be seismically qualified.
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