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Post-TMI Evaluation Task 2,Conceptual Design for Reactor Vessel Level Monitoring Sys. Sixteen Oversize Drawings Encl
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Issue date:	12/21/1979
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C-E POST TMI EVALUATION TASK 2 CONCEPTUAL DESIGN FOR A REACTOR VESSEL LEVEL MONITORING SYSTEM DECEMBER 21, 1979

ABSTRACT In March 1979 an incident occurred at the Three Mile Island Unit 2 Nuclear Power Plant which resulted in considerable damage to the reactor core. As a result of investigations into the cause of this incident, areas where improvements in nuclear power plant administration, design and operation are necessary ~ave been identified. With respect to nuclear power plant design, the Nuclear Regulatory Cormnission (NRC) has requested licensees to evaluate the use of a reactor vessel level measurement system as a means of providing additional useful* information to power plant operators or as a -means for directly determining the adequacy of core cooling~ Subsequently, the C-E Owners' Group authorized Combustion Engineering to develop the functional requirements and a conceptual design for a system to monitor reactor vessel level. This report provides the result of the C-E investigation of reactor vessel level monitoring system designs. The report includes a description of the design basis and functional requirements for a Reactor Vessel Level Monitoring System (RVLMS}; a description of the designs and design configuration recommended by C-E; and a brief discussion of other designs that were-evaluated * . The recommended design will not require major plant structural changes and has a high probability of being installed by January 1, 1981. In developing the functional requirements and the conceptual design recommended in this report, it was assumed that the RVLMS is technically required to provide a valid indication of reactor vessel level or the adequacy of core cooling. This is an assumption upon which the design is based - analyses have not been performed to substantiate such an assumption. P"ge 1 of. 43

r TABLE OF CONTENTS Section Title Page No. ABSTRACT 1 TABLE OF CONTENTS 2 LIST OF TABLES AND FIGURES 3 . l.O PURPOSE 4 . *2.0 SCOPE 4

3.0 REFERENCES

4

4.0 BACKGROUND

5

5.0 DESCRIPTION

OF ASSUMPTIONS, DESIGN BASIS AND 7 FUNCTIONAL REQUIREt1ElffS 5.1 ASSUMPTIONS 7 5.2 DESIGN BASIS 7 5.3 FUNCTIONAL DESIGN 8 5.4 FUNCTIONAL REQUIREMENTS 9 6.0 RECOMMENDATION 16 6.1 RECOMMENDED DESIGN 16 6.2 RECOMMENDED SYSTEM CONFIGURATION 18 7.0 OTH~R DESIGNS CONSIDERED 20

Table No. I Figure No. l 2 3 4 5 6 7 8 9 10 11 12 13 14 llST OF TABLES AND FIGURES Title of Table RVLMS DESIGN BASIS EVENTS Title of Figure RVLMS FUNCTIONAL BLOCK DIAGRAM HEATED JUNCTION THERMOCOUPLE INSTALLATION HEATED JUNCTION THERMOCOUPLE INSTALLATION SYSTEM CONFIGURATION FOR A RVLMS R-F PROBE INSTALLATION FLOATING SOURCE INSTALLATION FLOATING SOURCE INSTALLATION FIXED NEUTRON SOURCE AND DETECTOR FIXED NEUTRON SOURCE AND DETECTOR FLOATING DIPSTICK FLOATING SPHERES ULTRASONIC PROBE BUOYANT FORCE INDICATION PROBE EXTERNAL STANDPIPE WITH FLOAT SENSOR OR DP CELL Page No. 29 Page No. 30 31 32 33 34 35 36 37 38 39 40 41 42 43 Pqge 3 of 43

I 1. 0 f PURPOSE The purpose of this report is to document C-E's research and to recommend a Reactor Vessel Level Monitoring System Design. The design effort was requested by the C-E Owners Group in response to a Nuclear Regulatory Commission Lessons Learned Task Force requirement. 2.0 SCOPE This report provides C-E's conceptual design for a Reactor Vessel Level Monitoring System. The system functional requi.rements, description of the des_igns considered, a design recommenctation, and the recommended system configuration comprise the report.

3.0 REFERENCES

3.1 TMI-2 Lessons Learned Task Force Status Report And Short Term Recommendations 3.2 United States Nuclear Regulatory Commission, "Follow-up Actions Resulting From The NRC Staff Reviews Regarding The Three Mile Island Unit 2 Accident 11 , To All Operating Nuclear Power Plants, Dated September 13, 1979. Page 4 of 43

4.0 . BACKGROUND As a result of reviewing the TI1I incident, the NRC Lesson Learned Task Force issued a status report and short term recommendations document (Reference 3.1). This document consisted of recommendations to be implemented to further insure the public health and safety. One of these recommendations was associated with instrumentation for detection of inadequate core cooling in PWR's and BWR's. In its assessment of instrumentation adequacy, the task force concludes that sufficient instrumentation existed at TMI-2 to provide adequate information to indicate reduced reactor vessel coolant level, cor~ voiding and deteriorated core thermal conditions. However, in order to preclude failure to recognize such conditions in the future, the task force proposed to address this problem in two stages. The first stage further evaluated the use of installed instrumentation. The second stage was to study and develop system modifications that would not require major structural changes to the plant and that could be implemented *in a relatively rapid manner to provide more direct indication of certain plant parameters than that available with present instrumentation. This second stage includes the study of PWR vessel level detectors. In stating its position with respect to this recorrunerid~tion, the task force required licensees to provide a descrip-tion of any additional instrumentation or controls (primary or backup) proposed for the plant to supplement the currently installed instrumentation including a description of the functional design requirements. In reference 3.2 the NRC imposed the Lessons Learned Task Force's recommendations upon all Operating Nuclear Power Plants and provided a final schedule for implementation. The Owners Group subsequently authorized C-E to develop the functional requirements and conceptual design for a system to monitor reactor vessel level~ This report documents the results of that study. In "developing the functional requirements and the conceptual design P~ge 5 of 43 .,,,J~ .... !E:m!Jt!!'l'l ...JP.l!!ZUl'l;l.mi:ll!l.~.6""'"°"';;:~~.11r:'!.~¥,!4~1WC--.:,,,.,."11!'".'!"h'...i. t-....!'<11:&!111<"5..,.&lJaS!!!i/i~,Jimo_;;ua"""u"""""",,_'..,.*"'""" ---co-*.-=-=o-Ef u---*.wwww -~ I"'"!'.... ,.""'**..---'"'l:ltt;p'_..,

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recommended in this report, it was assumed that the RVLMS is techni-ca'lly required to provide a val id indication of reactor vessel 1 evel or the adequacy of core cooling. This is an assumption upon which the design is based - analyses have not been performed to substantiate such an assumption. In addition to clearly understanding what the results of this task are, it is important to understand what is not provided as part of this task.

1. The described system does not monitor reactor vessel level below the bottom of the hot leg, i.e. into the active core region.

Various options appear to be open in this area~ It is believed that already*installe~ instrumentation--the incore thermocouples and self po_wered neutron detectors--can be used to provide an indication of leve1 in the core. Further analyses are required to confirm this. Alternatively, the heated junction thermocouples described in this report could be adapted to extend down into the core region.

2.

No analyses or studies have been underta_ken as part of this task to correlate the output of the described RVLMS to an "unambiguous indication of inadequate core cooling". The system monitors and displays reactor vessel level above the core, only. Further analyses would be required to show how the indication of reactor vessel level relates to core cooling adequacy.

3.

Ho procedures or guidelines have been developed as part of this task to show how the RVLMS indication should be used.

4.

No analyses h~ve been done as part of this task to determine if a system to directly measure reactor vessel water level is required. L-------------*--*-----------------------

5.0 DESCRIPTION

OF ASSUMPTIONS, DESIGN BASIS AND FUNCTIONAL REQUIREMENTS 5.1 ASSUMPTIONS The following assumptions were made in the conceptual design effort of the RVLMS.

1. The reactor vessel inventory between the bottom of the hot leg to the top of the reactor vessel head is to be monitored. This range was specified at the beginning of the task and represents that area of the reactor vessel where no instrumentation exists which can be used to measure vessel inventory.

In addition; the response of exi~ting in-core instrumentation may be useful in m~asuring core inventory.

2.

The RVLMS will be useful to the plant operator, therefore, it must be reliable throughout the operating.cycle. 5.2 DESIGN BASIS A Reactor Vessel Level Monitoring System (RVLMS) will supplement the existing. NSSS instrumentation that gives an indication of the adequacy of core cooling. The NSSS design addresses the need to supply water to the reactor coolant system during an unlikely event that affects reactor coolant system inventory. However, under these conditions, the RVLMS will inform the plant operator via its alarm that the reactor vessel level inventory is affected, specifically, and_ additional effort to assure the adequacy of core cooling may be necessary. The continuous recording of reactor vessel level will provide both a permanent record of this parameter and a means of evaluating the results of evolutions which may affect reactor vessel level by indicating trends. In addi-tion, the recording of reactor vessel inventory, in conjunction with other plant instrumentation, will provide the plant operator with a means of determining core cooling adequacy. The reactor vessel level alarm setpoint will be such as to alert the plant operator in time to allow for evaluation of overall NSSS response to the cause for the alarm initiation. The alarm setpoint will also be such as to preclude spurious alarms during plant operation. PC\\ge 7 of 43

5.3 5.3.l In the event the system should suffer a loss of electrical power or a major failure, a unique indication shall be provided to infonn the plant operator that the system is not monitoring reactor vessel level and it should not be used to make decisions which could affect core cooling adequacy. The RVLMS Design Basis is as follows:

1. Sense reactor vessel level and continuously record the parameter in the control room.

2. Provide an alarm and annunciation in the control room when reactor

vessel water level decreases to a preset value to alert the plant operator prior to further degradation of reactor vessel inventory.

3. Monitor and record reactor vessel water level during the following plant operatJng modes:

a. Power operation

b. Start-up

c. Hot standby

d. Hot shutdown

e. Cold shutdown

4. Monitor and record reactor vessel water level during post accident conditions where reactor vessel inventory may be lost. The Design Basis Events under which the system must meet its design require-ments are listed in Table I.

FUNCTIONAL DESIGN General Figure 1 provides a functional block diagram of the Reactor Vessel Level Monitoring System _(RVLMS) and its interfaces. The RVLMS monitors changes in the reactor vessel bulk water level. This input is utilized to perfonn* two basic functions by the RVLMS; Display and Alarm. Pa~e 8 of 43

5.3.2 The first function is to display the reactor vessel water level to the plant operator by employing a strip chart recorder. The second function is to provide an alarm to the Annunciation System when-reactor vessel water level reaches a low preset level. Display Function The reactor vessel water level is the input to the. RVLMS. The RVLMS continuously displays the reactor vessel water level on a strip chart recorder in the control room. 5.3.3 Alarm Function The RVLMS provides its output signal to the plant Annunciator System. The plant Annunciator System will initiate an alarm when the sensed reactor vessel water level reaches a low preset value. SA FUNCTION REQUIREMENTS 5.4.1 This section specifies the requirements to be met*in the design of the RVLMS. Design Criteria The following Regulatory Guides, IEEE Standards and 10 CFR 50 require-ments will be met in the design of the RVLMS. 5.4.1.1 Regulatory Guides

1. 1.97 Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident - Revision 1, August 1977.

2. 1.29 Seismic Design Classification - Revision 3, September 1978.

3 *. 1.100 Seismic Qualification of Electric Equipment for Nuclear Power* Plants - Revision 1, August 1977. Pqge 9 of 43

4. 1.89 Qualification of Class lE Equipment for Nuclear Power Plants -

Original, November 1974.

5. 1.75 Physical Independence of Electric Systems - Revision 2, September 1978.

6. 1.53 Application of the Single-Failure Criterion to Nuclear Power Plant Protection Systems - Original, June 1973.

7~ 1.118 Periodic Testing of Electric Power and Protection Systems - Revision 2, June 1978.

8. 1.70 Standard Format ~nd Content of Safety Analysis Reports for Nuclear Power Plants - Revision 3, November 1978.

9. Qraft Revision 2 to 1.97 Instrumentation for Light-Water-Cooled Nuclear Power Plants to-Assess Plant Environs Conditions During and Following an Accident.' This includes the draft standard ANS-4. 5 Functional Requirements for Post Accident ~-1oni tori ng Ca pa bi 1 i ty for the Control Room*Operator of a Nuclear Power Generating Station.

5.'4.1.2 IEEE Standards

l. 323 Qualifying Cl ass lE Equipment for Nuclear Power Generating Stations.

2. 338 Criteria for the Periodic Testing of Nuclear Power Generating Station Safety Systems.

3. 344 Seismic Qualification of Class lE Equipment for Nuclear Power Generating. Stations.

4. 384 Standard Criteria for _Independence 6f Class lE Equipment and Circuits *

. 5. 379 Standard Application of the Single-Failure Criterion to .Nuclear Power Generating Station Class lE Systems.

These items are included to represent those requirements which should be considered in the design but are not presently required.

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5.4.l.3 10 CFR 50 Appendix A

1.

CRITERION 1 - Quality Standards and Records

2.

CRITERION 2 - Design Bases for Protection Against Natural Phenomenon

3.

CRITERION 3 - Fire Protection

4.

CRITERION 13 - Instrumentation and Control 5.4.1.4 10 CFR 50 Appendix B 5.4.2 Interface Requirements

The f o 11 owing interface requirements s ha 11 be accommodated:

5.4.2.1 Input Requirements Electrical Power Source The RVLMS power sources shall be from two Class IE power buses. One power source shall supply electrical power to one of the RVLMS channels. The other, separate, power source shall supply electrical power to the other, separate, RVLMS channel. 5.4.2.2 Output Requirements Annunciation and Alarm The RVLMS shall provide output signals to the Plant Annunciator system when the sensed reactor vessel water level reaches a low preset value. A separate output signal shall be provided for each separate RVLMS channel. These signals shall alarm separate indications on the Plant Annunciator system. The Plant Annunciator shall provide an audible and visible indication to the plant operators of the RVLMS low reactor vessel water level condition. Pa~~ 11 of 43

5.4.3 5.4.4 Sensor Requirements Reactor Vessel Level The RVLMS shall provide signals via reactor vessel level transmitters which represent the reactor vessel bulk water level. The range of the signals shall be at least from the top of the reactor vessel head to the bottom of the reactor vessel outlet nozzle. Sensor Requirements Locations The RVLMS sensors may be located internal to or external to the reactor vessel. Sensor Requirements Range The RVLMS sensors shall monitor the range from the bottom of the reactor vessel outlet nozzle to the top of the reactor vessel head, as a minimum. Number of Channels The RVLMS shall have at least two redundant channels for sensing reactor vessel level. System Accuracy and Response Time Requirements System Accuracy The RVLMS system accuracy shall be better than +/-10% of the required range discussed above. This accuracy applies to both the alarm set- .Point and to the indication to the plant operator as read on the RVLMS recorders

Pag~ 12 of ~3

5.4.5 5.4.6 System Response Time The RVLMS system response time shall be at most 6.0 seconds from the time a change in reactor vessel level is sensed until. the change is indicated at the recorder. It shall. also be at most 6.0 seconds from the time the reactor vessel water level reaches the low preset setpoint until. the alarm signal is present at the output of the RVLMS. Display Requi~ements Recording The RVLMS shall continuously monitor and simultaneously record the signal from each reactor vessel level sensor. Separate recorders shall be provided for each separate RVLMS channel. The recorders shall be a permanently mounted component located in the control room and uniquely identified as post accident monitoring equipment. Alarm and Annunciation Requirements The RVLMS shall provide alarm signals to the Plant Annunciator System upon determining that reactor vessel level has reached a low preset value. A separate alarm signal ~hall be provided for each RVLMS channel. This setpoint will be determined later during the setpoint effort

r~~: 13 of 43

5.4.7 Testing Requirements 5.4.8 Instrument Channel Checks The RVLMS channel outputs shall be displayed such t~at channel operability may be verified by comparing readi_ngs between channels. functional Tests The RVLMS channels shall have the capability to ?e tested by injecting a 1 test signal as close to the sensor as is practical to assure that each channel is capable of performing its design function. This shall include having the capability of testing the accuracy of the setpoint used for the alarm function as well as testing the accuracy of the display available to the operator.. The ability to test the system response time shall also be provided. Sensor Calibration Verification Tests The RVLMS sensors shall have the capability to be tested such that with a known precision input the instrument gives an output signal of the required accuracy and response time. Operational Requirements The RVLMS is required to fimction within the required accuracy and time response limits during the modes of plant operation specified in the Design Basis section. p"~~ 14 of 43 '"-az4; Wi?"Jil

5.4.9 Post Accident Monitoring Requirements The RVLMS is required to perform its required functions with the required accuracy and time response limits during post accident monitoring conditions where reactor vessel level may be affected. 5.4.10 Qualification Requirements 5.4.11 The RVLMS shall be qualified to perform its required functions with the required accuracy and time response before, during and after the most adverse environmental conditions associated with the Design Basis Events listed in Table I. With. respect to-seismic qualification, the RVLMS shall continue to perform its required functions with required accuracy and time response following the seismic event. Indication of Loss of Power-or System FaHure Requirements An indication of loss of electrical power or a system failure to each RVLMS channel shall be provided. 5.4.12 Accessibility & Mounting Requirements The RVLMS recorders shall be permanently mounted in the control room and the recorders shall be clearly displayed for operator use. Pa~~ 15 ~f 43 -~"""---*"*!"'--~~*--------------------------

6.0 RECOMMENDATION This section of the report provides Combustion Engineering's recommenda-tion of the best means by which to measure reactor vessel level and the reason for the recommendation. 6.1 RECOMMENDED DESIGN 6.1.1 Heated Junction Thermocouple The heated junction thermocouple sensi_ng means mea*s:ures the change in thermocouple output voltage as a result of the difference in the thermal conductive properties between steam and water. A series of heated junction thermocouples will be located at different axial positions above the core. Ffgures 2 and 3 depict heated junction thermocouple installations in the reactor vessel. Each discrete location would have two thermocouples connected in series but in electrical opposition to each other. One of these thermocouples will be heated, thereby establishing a reference differential voltage output which is a function of the heater input. When the water surrounding the heated junction thermocouple is replaced by steam or air the voltage generated by the heated thermocouple will change because the heat generated will no longer be removed from the area by the water. Thus the temperature of the heated thermocouple will increase relative to the unheated thermocouple. This change will be used to detennine the steam water interface in the reactor vessel. 6.1.1.1 Evaluation of the Heated Junction Thermocouple The heated junction thermocouples are adaptable to the operating reactors by insertion through an In Core Instrument (ICI) nozzle port or a Part Length CEA location. This method of determining vessel level is presently in use in a research facility in Idaho. The resolution of the system is adjustable over the range of measurement based upon the distance between thermocouples. The acc*uracy of the system is expected to exceed the + 10% of span . Pa~e 16 of 43

presently required. The heated junction thermocouples may be compatible with existing core designs for measuring level in the core.

However, this application has not been thoroughly evaluated. Since the application of heated junction thermocouples for monitoring vessel level has been demonstrated, there is a high probability of system installation by the January 1981 deadline imposed by the Nuclear Regulatory Commission.

Three disadvantages associated with the use of heated junction thermo-couples are as follows:

l. Use of the ICI nozzle e.e>rt for thermocouple ins~rtion will require removal of in-core instrumentation. With respect to the Technical Specifications, the plant will then be operating with a certain.

percent~ge of unavailable in-core-instrum~nts based upo~ the in-core-instrument uncertainty analysis.

2. Use of a Part Length CEA location will require removal of a Part Length CEA in order to accommodate the heated junction thermocouples.

3. The resolution of the system is restricted by the number of electrical leads which can physically be accommodated by the ICI nozzle port and/or spare containment electrical penetrations

Page 17 of 43

6*. 2 RECOMMENDED SYSTEM CONFIGURATION 6.2.l Description of the Recommended System 6.2.2 figure 4 shows the recommended system configuration for the Reactor Vessel L.evel Monitoring System as discussed below. The recommended system configuration will require that four Part Length CEA's be removed to provide locations for the sensing devices. Two Heated Junction Thermocouple level 5ensor groups will comprise one channel of vessel level indication. Each group of Heated Junction Thermocouples will be inserted into. a vacated Part-Length CEA location. The.output*signal will be sent to a recorder located in the control room where each output signal can be continuously_ recorded. A selector switch may be provided to allow the average of the two signals or any one or both of the two level sensor outputs to be recorded at a time. The remaining two Part-Le.ngth CEA locations will also contain a Heated Junction Thermocouple level sensor in each location. This second channel will be identical tb the first channel. The second Heated Junction Thermocouple channel will record its sensor output on a separate recorder. Reasons for the Recommended System T.he advantage of the specific type of sensor recommended has previously been discussed. Following is a discussion of the reasons for the recommended system configuration. It must be emphasfzed that.in developing the functional requirements and conceptual design recommended in this report, it was assumed that the RVLMS is technically required to provide a valid indication of reactor vessel level or the adequacy of core cooling. This assumption was required to be made in order that the task could proceed. No analyses PC\\~e 18 of 43

have been performed as part of this task to substantiate such an assumption. Making the assumption that the RVLMS is technically required, it is believed that the recommended design is the optimum method of per-forming its required functions. A system of lesser complexity (i.e. no redundancy) may be acceptable from a licensing standpoint. However, if a RVLMS is not required for a technical reason, it is recommended that no system be installed. During the course of developing a conceptual design for a Reactor Vessel Level Monitoring System, certain basic criteria emerged as being necessary or very d~sirable for such a system. Meeting these criteria will go a long ways toward ensuring that the system will not only* indicate level to the required degree of accuracy but will also have sufficient reliability and clarity so as to be genuinely useful to a plant operator. The criteria and how the recommended system configuration meets the criteria is discussed pelow. Redundancy Redundancy is a desirable design feature to incorporate because it is a means by *which the availability of the system can be assured throughout the operating cycles. If one has a system of two separate channels, one has redundancy. If one sensor fails, the redundant sensor is still operable. However, it is very desirable to have more than two degrees* of redundancy in any system that is to be relied upon. If there are two channels indicating different levels, which one is correct? It is thus desirable to have at least three degrees of redundancy to alleviate this problem. If there are four degrees of redundancy, the system can continue operating reliably with one redundant indication system failed. This still leaves the capability to identify a failure that may develop as a result of the event the system. is *needed for. The recommended system has four degrees of redundancy. There are two redundant channels, each of which has two separate sensors. This is page 19 of 43

7.0 the optimum system from a redundancy standpoint and is easily imple-mented because the designs require removal of part length CEAs and they are arranged in symmetric groups of four. Hon-Hydraulic During the events that the RVLMS is required to operate, the hydraulic forces in the area of the reactor vessel head can be extremely dynamic and unpredictable. Although this concern may not be insurmountable, much analytical and design work would be required to ensure proper indication of level using a sensor system that worked on a hydraulic principle. The recommended design does not utilize this principle to determine level. No Moving Parts A level system that relies on physically moving components should be avoided if possible. Historically, moving components tend to fail by virtue of their mover.ient.. The recommended design uses no moving parts to sense vessel level. OTHER DESIGNS CONSIDERED This section provides a brief description of other designs considered for use in measuring Reactor Vessel Level along with the advantages and disadvantages associated with each design. 7~1 Radio Freq~ency (RF) Probe The R-F probe design (Figure 5) consists of a Time Domain Reflectometry unit which sends a voltage step along the length of the R-F Probe. The pulse is reflected ha.ck to the Time Domain Reflectometry unit when it encounters a discontinuity along the probe such as water. The time between pulse initiation and reflection is measured by the circuitry and translated into a level signal. page 20 of 43

7.1.1 Advantages 7.1.2

1. It is highly feasible since the design was used successfully in the C-E pot boiler blowdown test conducted in 1976.

2. Test Report, TR-ESE-164, "R-F Probe for Two Phase Liquid Level Measurements Description and Calibration", dated September 29, 1977, indicated this system is suitable for measurement of reactor vessel 1 evel. Accuracy +/-1. 5"; Response Time 60 msecs.

3. It is adaptable to operating operators.

4. There are no moving parts.

5.* Corrections to problems are documented in the test report. Therefore, development time is reduced making the January 1981 installation date realistic

Disadvantages

1. Requires removal of part length CEA.

2.

May impact existing refueling procedures (probe must be withdrawn).

3. Probe and accessories must be manufactured.

4. The effect of radiation, coolant temperature changes and coolant chemical concentration changes on system stability and reliability must be further evaluated.

ra~e 21 of 43

7.2 FLOATING SOURCE The floating source concept is shown in two variations on Figure 6 and 7. It consists of a low strength Berylium-Plutonium neutron source double encapsulated within a thin wall stainless steel sphere. The sphere is pressurized with helium to minimize the shell thickness and for detecting leaks. In one variation (Figure 6) the sphere is 7 11 in diameter and is contained within a CEA shroud. The only addition required to the upper guide structure (UG~) is the removal of CEA guide brackets and the addition of a "top hat" which is mechanically joined to the top of the CEA shroud. This arrangement of course means that a part length CEA must be removed. The second variation (Figure 7) utilizes a 3" dia sphere contained within a perforated sleeve~ The sleeve is installed within the CEA shroud and mechanically locked in place with no machining or cutting required. During normal operation, the sphere is pinned against the top of the sleeve by the buoyant force of the coolant. When the water level drops, the sphere follows the water level. A neutron detector mounted on the R.V. closure head directly above the source sleeve measures the position *of the sphere as a function of neutron reception. 7.2.1 Advantages

1. It is feasible as determined by review of existing technology.

2.

It is expected to operate under normal and accident conditions.

3. It is adaptable to operating reactors by modifying vessel components.

4.

It has minimum impact on existing refueling procedures.

5.

Its accuracy is expected to meet the functional requirements. pqge 22 of 43

7.2.2 Disadvantages

1. Requires removal of part length CEA and modification to CEA guide tube.

2.

Design must ensure that ball and/or source remain in CEA guide tube.

3.

Floating ball movement against the CEA guide tube may wear through ball surface.

4.

Problems associated with design development may prevent system installation by January 1981.

5. Source, ball and CEA guide tube modifications must be manufactured.

6.

Vessel Head Removal is required to correct any mechanical problems. 7.3 FIXED NEUTRON SOURCE AND DETECTOR The.fixed neutron source and detector concepts are shown on Figures 8 and 9 and utilize the ICI routing conduit. The two schemes under consideration include a fixed separate source conduit and two assemblies of s~all neutron detectors axially located within an adjacent conduit (Figure 8),* and an alternat~ scheme where the source and detectors are located in the same conduit (Figure 9). The detectors measure the neutron flow from adjacen*t sources and, therefore, accurately predict water level and steam quality above the water layer. 7.3~1 Advantages

1. It is feasible as determined by review of existing technology.

2. It is expected to operate under normal and accident conditions.

3. It is adaptable to operatfog reactors by inserting detector string through ICI port.

4.

Its accuracy is expected to exceed the functional requirements. 5a There are no moving components.

6.

Detectors exist.

7.3.2 Disadvantages

1. Requires use of ICI Nozzle Port. If there are no spare ports, ICI removal *is necessary.

2. There is a history of failure associated with the detector.

3. Problems associated with design development may prevent system installation by January 1981.

4.

Source must be manufactured. 7.4 FLOATING.DIP STICK - FLOATING SPHERES The floating dip stick concept is shown in Figure 10 and consists of large diameter thin walled stainless steel spheres connected to a shaft. The top of the shaft contains a magnet which can be located by a reed switch a*ssembly in the sam.e manner as CEA position is determined. The large diameter spheres ser.ve as a floatation device for the stick. This concept requires that the density of the coolant be constantly monitored by means of temperature recording equipment. An alternate scheme to the floating dip stick concept is shown in Figure 11 and consists of a long column of 1-1/2 11 diameter thin walled spheres. The column of spheres is supported by the buoyant force acting on a*bout the 15 lower spheres. The entire column is enclosed within a perforated sleeve. The upper sphere contains a magnet similar to that used in the floating dip stick scheme. This scheme is much simpler than the floating dip stick when it comes to special refueling procedures because the upper containment sleeve and spheres are handled along with the closure head, i.e., the column of spheres are held in place by battery operated electromagnetic means. 7.4.1 Advantages

1. It is feasib1e as determined by review of existing technology.

2. It is expected to operate under normal and accident conditions.

3.

It is adaptable to operating reactors by removal of a part length CEA.

4.

Its accuracy is expected to meet the functiona 1 requirements.

5. The existing reed switch position transmitter may be ~serl.

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7.4.2 Disadvantages

1. Requires removal of part length CEA and modification to CEA guide tube.

2.

Accuracy is dependent upon water density.

3.

More than one floating ball is required.

4.

Movement of the floating ball agair.st the CEA guide tube may wear through the surface of the ball.

5.

Floating balls, dipstick and CEA guide tube modifications must be manufactured.

6.

Vessel Head Removal is required to correct any mechanical problems.

7.

Problems associated with design development may prevent system installation by January 1981.

8. It may impact existing refueling procedures (dipstick must be removed).

7.5 ULTRASONIC PROBE The ultrasonic probe design (Figure 12) consists of a pulse transmitter which sends a pulse dm'ln a stepped shaft wave guide which is located within a Part Length CEA guide tube. The wave guide is reduced in circumference (stepped) at regular intervals to provide a distinct interface upon which the signal will act. The pulse will reflect back when it encounters the discontinuity in the wave guide. The amplitude of the back reflection will be larger if a step is covered with steam than with water. The difference in the amplitude of reflected signals is analyzed by the circuitry for water level determination. 7.5.1 Advantages

1.

2.

3.

It is feasible as determined by review of existing technology. It is expected to operate under normal and accident conditions. It is adaptable to operating reactor by removal of part length CEA.

4.

There are no. moving parts. s.* Head removal not necessary for repair. f ,~- I I I. I l I t l ! l I I ~ i l i I l lI t

7.5.2 7.6 Disadvantages

1.

Requires removal of part length CEA and modification to CEA guide tube.

2.

Problems associated with design development may prevent system installation by January 1981.

3.

M~y impact existing refueling procedures (wave guide must be withdrawn).

4.

Wave guide and CEA guide tube modification must be manufactured.

5.

Steam bubbles may collect on shaft steps thus causing false level indication. BUOYANT FORCE TRANSMITTER The buoyant force transmitter is shown in Figure 13 and consists of a 15 feet long - 2" di~meter light weight hermetically sealed stainless steel tube. The shaft is connected to a linear voltage differential transmitter (LVDT) cell which monitors change in weight of,the tube as a function of the buoyant force. The shaft is positioned within a CEA shroud. I . I

i I

i 7.6.1 Advantages 7.6.2

1. It is.feasible as determined by review of existing technology.

2.

It i~ e~pected to operate under normal and accident conditions.

3. It is adaptable to operating reactors by removing part length CEA.

4. Its accuracy is expected to meet the functional requirements.

Disadvantages

1.

2.

3.

4.

5

Requires removal of part length CEA.

Output must be compensated for fluid density changes to meet accuracy functional requirements. Problems associated with design development may prevent system installation.by January 1981. Probe and accessories.must be manufactured. Vessel head removal is required to correct any mechanical problems. page 26 of 43 41AP:W:

6.

Receiving an accurate indication of. the buoyant force on the LVDT. cell is complicated by the system pressure, the dynamic forces on the tube and the frictional resistances. 7.7 EXTERNAL STANDPIPE WITH FLOAT SENSOR OR DP CELL The ~xternal standpipe approach (Figure 14) would make use of existing connections on the primary system. The standpipe would be connected

between the head vent and a flange connection on the bottom of the hotleg piping.

The standpipe, which is filled with reactor coolant, would contain a-float mechanism which contains a source or a magnet. A suitable detector would be located alongsid~ the standpipe to detect the position of the float. The position of the float would.reflect the reactor vessel water level. A DP cell (~igure.14) may be added to or substituted for the standpipe. 7.7.1 Advantages

1. It is feasible as determined by existing technology.

2. It is expected to operate under normal and accident conditions.

3. It is adaptable to operating reactors by connecting to the head vent and flange on the hotleg.

4.

Installation will not require modification of reactor internals or* removal of the vessel head.

5.

DP cells are used to measure level and inaccuracies associated with level measurement are understood.

6. *Accuracy associated with float in the standpipe is expected to meet the functional requirements.

7. System is accessible.

Page 27 of 43 I,*

Disadvantages

1. External standpipe must be removed and stored during refueling.

2.

DP cell must account for flow in the reactor vessel, changes in reactor coolant temperature, and the difference between reference leg temperature and reactor coolant temperature.

3.

DP cell reference leg flashing effects on accuracy must be considered.

4.

Standpipe layout problems are plant specific.

5.

Increases primary system piping which can experience leakage.

6.

Leakage at flanged connections will affect system accuracy.

7. Problems associated with standpipe design and installation may prevent system installation by January 1981.

pa~e 28 of 43

TABLE I RVLMS DESIGN BASIS EVENTS The RVLMS shall be designed to function during the following Design Basis Events (including the environmental effects associated with these events).

1. Loss of Coolant Accident (LOCA) up to and including a double-ended rupture of the largest pipe of the Reactor Coolant Pressure Boundary {RCPB) as defined in 10 CFR 50.2.

11. Steam System pipe rupture up to and including a _double""'.ended rupture of the largest pipe in the system.

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0 MP R ASSOCIATES, INC.

December 4, 1979 P-490 Mr. R. A. Vincent Consumers Power Company P~lisades Nuclear Power Plant Route 2 - Box 154 Covert, Michigan 49043

Subject:

Palisades Systems Outside Containment which Could Contain Radioactive Fluids

Dear Mr. Vincent:

At your request, we have performed a review to determine which of the systems located outside containment would or could contain highly radioactive fluids during a serious transient or accident. In particular, this review identifies the systems included in the leakage reduction program being implemented* in accordance with Nuclear Regulatory Commission Recommendation 2.1.6.a in "TMI-2 Lessons Learned Task Force Report (Sho~t Term)," NUREG-0578. The results of our rev.iew indicate that Consumers Power Company compliance with the NRC recommendation should be accomplished as follows:

1.

A series of plant modifications would be made to reduce the number of systems outside containment which would be contaminated following an accident.

2.

A leakage reduction program will be implemented on the remaining systems which may be required to process con-taminated fluids following an accident, specifically the engineered safeguards systems and the sampling systems. These results were discussed with Consumers Power Company on November 9, 1979, and it was agreed this was a technically desirable approach. The specific modifications agreed to, together with the pertinent background and basis for the modi-fication, are described in Attachment A. For convenience, the recommended modifications are summarized below: 1140 CONNECTICUT AVENUE. N. W, WASHINGTON, 0. C. 20036 . 202-659-2320

MPR ASSOCIATES, INC. Mr. R. A. Vincent 2 - December 4, 1979 0 0 0 0 Reactor Coolant System Degassing Modifications will pe made to permit venting the pressur-izer effluent to the containment atmosphere, most likely via the quench tank. These modifications would permit degassing of the primary coolant system within contain-ment, if required following an accident. Currently, this pressurizer effluent is cooled in the sampling system and routed to the volume control tank or to the vacuum degasi-fier. Sampling Systems Sampling system effluents will be returned to con-tainment, rather than to chemical and volume control system and waste gas system components. Sampling lines from engineered safeguards equipment to the existing liquid sampling panel will be deleted or isolated within the safeguards rooms. A single sampling line from the engineered safeguards pump discharge to the new liquid sampling system will be considered. This line would enable a reactor coolant sample to be obtained when reactor coolant system pressure is low following a major accident. Safeguards Room Sump Effluent The sump effluent will be returned to containment in the event of a serious accident. The effluent is currently routed to the dirty waste tank outside the safeguards room. An alternate route is to the auxiliary building equipment drain tank. Either of these routes could cause contamination of numerous additional auxiliary building components and will not be used following an accident. Safeguards Equipment Relief Valve Discharge The discharges from engineered safeguards component relief valves will be routed to the safeguard room sumps. These discharges are currently routed to the auxiliary building equipment drain tank located outside of the safeguards room

M PR ASSOCIATES. INC.

Mr. R. A. Vincent 3 - December 4, 1979 0 0 High Pressure Air Supply to Safeguards Equipment Valve Operators A modification will be made to prevent air from the safe-guards rooms to enter the turbine building via the intertie between the high pressure air compressor located in the east safeguards room and a similar compressor located in the turbine building. Letdown Line Isolation based on Failed Fuel Monitor Plant personnel indicated that they are considering an additional possible modification to ensure that the let-down line is isolated whenever significant quantities of radioactive material are present in the reactor coolant. This would involve adding one additional isolation signal derived from the failed fuel monitor. In addition to Attachment A, this letter contains the following two attachments: 0 0 Attachment B -- A checklist of each line which penetrates containment. This checklist was used to identify each system which could potentially contain radioactive material in the event of a severe accident, presuming no plant modi-fications are accomplished. Attachment C -- A checklist of each potentially radioactive system identified in Attachment B. This checklist was prepared based on review of system P&ID's, plant operating procedures, and detailed discussions with Consumers Power Company personnel. It presumes that plant modifications are made, and is used to identify the systems or portions of systems which could contain radioactivity. This check-list also identifies various functions which must be per-formed following an accident, and how they would be accomplished. As part of developing the plan outlined in Attachment A, we performed several evaluations which will be transmitted by separate correspondence. These include: 0 0 An evaluation of pressurizer degassing capability during potential accident conditions~ An evaluation of the off-site doses resulting from opera-tion of the steam generators with a tube leak

MPR ASSOCIATES. INC.

Mr. R. A. Vincent 4 - December 4, 1979 0 An evaluation of the off-site doses resulting from leakage of contaminated water from safeguards equipment. It is our understanding that various plant modifications described in Attachment A are currently being detailed by vari-ous Consumers Power personnel or by o~her organizations. We would be pleased, if you wish, to review these designs, and to assist you in the design of specific modifications. Finally, as we have discussed in recent telephone conversa-tions, you will establish a mutually convenient date for us to visit Palisades to continue with our efforts on the leakage reduction program. If you have any questions concerning this letter or its attachments, please do not hesitate to contact us. cc: R. B. DeWitt CPCo, Jackson, MI J. G. Lewis CPCo, Covert, MI Sincerely, rh-:

-DRAFT-Attachment A REVIEW OF SYSTEMS OUTSIDE CONTAINMENT OF THE PALISADES PLANT WHICH COULD CONTAIN RADIOACTIVE FLUIDS IN THE EVENT OF SIGNIFICANT CORE DAMAGE A review has been performed to determine which of the systems located outside containment would or could contain highly radio-active fluids during a serious transient or accident. In par-ticular, this review identifies the systems included in the leakage reduction program being implemented in accordance with Nuclear Regulatory Commission Recommendation 2.1.6.a in "TMI-2 Lessons Learned Task Force Report (Short Term)," NUREG-0578, and subsequent clarifications of NUREG-0578 contained in the NRC letter dated October 30, 1979, "Discussions of Lessons Learned Short Term Requirements." The review also identifies the systems which will not be in-eluded in a leakage reduction program. It is considered that the delineation of systems to be covered in the leakage program is in full compliance with th.e Nuclear Regulatory Commission recommendations. The type accidents cori~idered for this review include a TMI-2 type of accident in which the initial path for loss of reactor coolant was isolated. Also considered were non-isolable breaks in the reactor coolant pressure boundary, either large or small in size, which would eventually require recirculation of re-actor coolant from the containment sump to achieve reactor

cooling and containment spray. For each type of accident, the functions which must be performed by various plant systems outs~de of containment to achieve a safe shutdown condition were considered. In summary, the review indicates that a number of systems out-side containment could contain radioactivity following an accident, based on their current configuration. Some of these systems, such as the emergency core cooling systems, would clearly be required under certain accident conditions, and cannot reasonably be modified to avoid contamination. Accord-ingly, such systems will be included in the leakage reduct.ion program. The systems in this category are identified in Para-graph A below. Contamination could be avoided in other plant systems, such as the chemical and volume control system, by implementing plant modifications. It is considered that such modifications, which permit radioactive materials to be kept inside containment, are the technically desirable means for minimizing off-site expo-sures. Such modifications would eliminate significant off-site exposure as a result of leakage, and would als6 avoid potential exposures as a result of operator error (e.g., valve misalign-ments), or equipment malfunctions (e.g., decay tank overpres-surization). Such modifications would also minimize exposure of on-site personnel because they significantly reduce the potential exposure from radioactive components -- pumps, tanks, 2 -

piping, etc. Accordingly, it is*planned, where practicable, to implement appropriate modifications. Clearly, leakage from such systems will be kept as low as practical as a matter of normal plant practice, in order to minimize waste processing requirements and personnel exposure during operation and main-tenance. The specific modifications which will be made are identified in Paragraph B below. Paragraph C summarizes the systems which can avoid contamination as a result of the modi-fications, and identifies.the means which would be used to accomplish various functions which may be required following an accident. A. Systems Included in Leakage Reduction Program

1.

Safety Injection and Containment Spray The following systems located outside containment would be required to operate in the recirculation mode (draw potentially radioactive water from the containment sump and return it to the reactor coolant system or containment) following a non-isolable break in the reactor coolant pressure boundary. Accordingly, these systems will be included in the leakage reduc-tion program: 0 High Pressure Safety Injection (HPSI) 0 0 Low Pressure Safety Injection (LPSI) Containment Spray 3 -

2.

Shutdown Cooling Operation of the LPSI system in the shutdown cooling mode (draw potentially radioactive water from the reactor coolant system and return it to the reactor coolant system) would be employed as a backup to the steam generators to achieve core cooling following an isolable break in the reactor coolant system pres-sure boundary. Accordingly, the additional portions of LPSI associated with the shutdown cooling mode will be included in the leakage reduction program.

3.

Sampling The systems employed for obtaining samples of reactor coolant and of the containment atmosphere will be re-

quired following an accident and, accordingly, will be included in the leakage reduction program.

In this regard, these systems will be modified in accord-ance with the TMI-2 Lessons Learned Task Force Recom-mendation 2.1.8.a. New sample lines and panels will be employed. Accordingly, the new systems, rather

than the existing sampling systems, would be included in the leakage reduction program.

Modifications to sampling systems which minimize the likelihood that radioactive sample materials could escape to the environment are identified in Para-graph B below

4 -

B. Plant Modifications Plant modifications to minimize transport of radioactive materials to systems outside containment are identified below.

1.

Degassing of Primary Coolant Degassing could be required as it was at TMI-2 to remove excess hydrogen from the reactor coolant. The current degassing method employed at Palisades during normal shutdowns is to withdraw steam and excess hydrogen from the pressurizer gas space via a sampling line to the sample system outside contain-ment. The fluid is cooled by sampling system coolers and is then returned to the volume control tank in the chemical and volume control system. The cooled fluid may also be returned to the vacuum degasifier. In either case, the gas is then passed to the waste gas surge tank, compressed, and stored in the gas decay tanks. This degassing method will be modified to avoid un-necessary contamination of portions of the chemical and volume control system, the vacuum degasifier, and waste gas system components. In particular, effluents from the pressurizer gas space during degassing will be vented to the containment atmos-

phere, most likely via the quench tank.

This modification will be accomplished in conjunction with 5 -

the installation of a pressurizer vent in accordance with TMI-2 Lessons Learned Task Force Recommendation 2.1.9, Reactor Coolant System Venting.

2.

Sampling Systems Modifications to the new and existing sampling sys-tems are as follows: 0 0 0 The new sampling systems will contain provisions to return liquid and gas sampling system ef flu-ents to containm~nt rather than to systems out-side containment. Sampling lines from engineered safeguards equip-ment to the existing liquid sampling panel will be deleted or isolated within the safeguards rooms. A single sampling line to the new liquid sam-pling system from the engineered safeguards pump discharge will be installed. This line would enable a reactor coolant sample to be obtained when reactor coolant system pressure is low.

3.

Safeguards Room Sump Effluent The two rooms which contain the HPSI, LPSI, and con-tainment spray systems each contain a sump which collects any leakage from the equipment. As cur-rently designed, effluent from these sumps is pumped 6 -

automatically on high sump level to tanks outside of the safeguards room. The normal route is to the dirty waste tank, which is vented via the auxiliary building gas collection header to the plant stack. An alternate sump discharge path, not normally used, is to the equipment drain tank, which is vented to the waste gas surge tank. In either case, noble gases from sump pump effluent, as well as the liquid, would be transported to numerous additional auxiliary building components. Accordingly, a modif i_cation will be made to return safeguards room sump effluent to containment, and thus avoid extensive contamination of other auxiliary building components.

4.

Safeguards Equipment Relief Valve Discharge The relief valves for various components of the engineered safety systems currently discharge to the equipment drain tank which, as noted above, is located outside the safeguards room_, and is vented to the waste gas surge tank. A modification will be made to reroute these discharges to the safeguards room sumps. In conjunction with the sump effluent modi-fication described above, these discharges would be returned to containment. 7 -

5.

High Pressure Air Supply to Safeguards Equipment Valve Operators The high pressure air supply to various control valves for engineered safeguards equipment is cur-rently intertied with a similar system in the turbine building. The current system allows the air corn-pressers located in the turbine building and the safeguards room to serve as backups to each other. The turbine building compressor can supply engineered safeguards system valves and, alternatively, the safeguards room compressor can supply valves in the turbine building. With such an arrangement, contaminated air from the safeguards room could enter the turbine building. Accordingly, a modification will be made to prevent flow of air from the safeguards room to the turbine building. C. Systems which Avoid Contamination The modifications identified above prevent contamination of various plant systems as noted above and, accordingly, it is considered that each of these systems need not be included in the formal leakage r.eduction program. These are: 0 The Chemical and Volume Control System 0 The Liquid Waste System 0 The Vacuum Degasif ier System 8 -

0 The Waste Gas System, including the waste gas compressors and decay tanks, and the auxiliary building gas collection header These systems perform various functions during normal operation, some of which would be required in the event of a serious accident. The modified method of performing various required functions following an accident is sum-marized below. In the following discussion, "design basis" accident refers to breaks in the reactor coolant system which cannot be isolated, and "TMI type 11 accident refers to breaks which can be isolated.

1.

Chemical and Volume Control System (CVCS)

a.

Letdown of Reactor Coolant During normal operation, letdown of -reactor coolant is accomplished in order to (1) purify the reactor coolant in the purification ion exchangers, (2) deborate the reactor coolant in the deborating ion exchangers during plant start-ups, and (3) remove excess reactor coolant to maintain pressurizer water"ievel. Following a major accident, the letdown line would be iso-lated by high containment radiation (CHR) or high containment pressure (CHP). Operator action would be required to re-establish the letdown flow. This action will not be needed during or immediately following an accident. In particular: 9 -

Purification: Purification of the primary cool-ant would not be required during the course of an accident or immediately following an accident. It is expected that, as at TMI, specialized sys-tems would be required to achieve eventual cleanup of highly contaminated fluids. These systems would utilize highly shielded process components, special spent waste handling equip-ment, etc., and would be thoroughly leak tested before use. Reactor coolant purification would not rbe accomplished in the eves. Deboration: This function would not be required fol;towing a design basis or TMI type of accident. Excess ~oolant Removal: Removal of coolant would not be required following a design basis or TMI type of accident. During both a design basis accident and a TMI type accident, there is a requirement to add water to the PCS to replace coolant which has been lost, rather than to letdown. If an unforeseen requirement to re-lease fluid from the PCS arises, this may be accomplished through the pressurizer relief or vent valve to the. quench tank or to the contain-ment sump

10 -

b.

Reactor Coolant. Pump Seal Leakage Collection During normal operation, reactor coolant pump seal leakage is diverted to the volume control tank. On high containment pressure or radiation, the leakoff line is isolated and the leakage is automatically directed to the primary system drain tank inside containment. This flow path would be maintaine.d following an accident so as to keep the radioactive fluids inside contain-ment. (Note that there is no seal injection system at Palisades such as at TMI.)

c.

In this regard, it is understood that the cur-rent Combustion Engineering recommendation is tq keep two pumps operating in the event of an accident. If this recommendation is implemented, the leakage flow would be approximately 1/2 to 1 gpm (about 700 to 1,400 gallons per day). Depending on the duration of pump operation, this could cause release of drain tank contents to the containment sump. This is considered preferable to permitting transport of the leak-age flow outside of containment. Degassing and Sampling The degassing and sampling effluents during normal operation are routed to the volume 11 -

control tank. As discussed in Paragraph B above, plant modifications will be made to ensure that these effluents are kept within containment.

d.

Reactor Coolant Makeup

e.

Makeup of reactor coolant via the charging pumps would be performed following a design basis or TMI type of accident. However, the sources of makeup water would not be radioactive. These sources include the primary system makeup stor-age tank or, alternatively, the safety injection and refueling water (SIRW) storage tank. Reactor Coolant System Boration Boration of the reactor coolant system by eves equipment could be required following a TMI type accident to achieve cold shutdown. (This function would be accomplished following a design basis accident by safety injection compo-nents drawing borated water from the SIRW tank.) Boration would be accomplished by supplying highly borated water from the concentrated boric acid tanks. These tanks are required by the Technical Specifications to contain at least 6,540 gallons at 10,936 ppm, which is adequate to achieve cold shutdown. The borated water 12 -

would not be significantly radioactive as a result of the accident. It is also noted that addition of the water would not impose a letdown requirement on the eves. Reactor coolant shrink as a result of cooldown amounts to about 22,000 gallons, which is significantly greater than the volume of borated water.

f.

Boron Concentration Monitoring A boronometer is employed during normal opera-tion to monitor the boron concentration of the reactor coolant. This function would be per-formed following an accident by the new sampling system.

2.

Liquid Waste System During normal operation, this system collects liquids from various equipment and sumps both inside and outside containment, and processes those liquids as required to permit their discharge or re-use in the plant. Our review indicates the following: 0 In the event of a serious accident, each line from collection tanks or sumps inside contain-ment is isolated by CHR and/or CHP, and need not be reopened. 13 -

0 Based on the modifications discussed in Para-graph B above (rerouting of engineered safe-guards room sump effluent, engineered safeguards component relief valve discharge, and liquid sampling system effluents back to containment), none of the equipment outside of containment which could contain highly radioactive fluids would transmit fluids to the liquid waste system. Accordingly, the liquid waste system would not be contaminated following an accident and its use would not be required. As noted above, it is expected that eventual cleanup of highly contaminated fluids would be accomplsihed by specialized equipment such as at TMI so that the liquid waste system would not be required for this function.

3.

Vacuum Degasifier System This system is normally employed to remove gases from various contaminated liquids before they are transported to liquid waste storage tanks. These include liquids from equipment drain tanks, both inside and outside containment, and sample system liquid discharges. Our review indicates: 0 The line from the primary system drain tank inside containment is isolated by CHP and/or CHR following a serious accident, and need not be reopened. 14 -

4.

0 The line from the auxiliary building equipment drain tank will not contain highly radioactive materials, based on the modifications to the engineered safeguards room sump effluent routing and relief valve discharge routing. 0 The line from the sampling system will not con-tain highly radioactive fluid, based on rerouting the sampling system discharge effluent. Accordingly, the vacuum degasifier system will not be contaminated following an accident. Gaseous Waste System This system contains two subsystems for collection of potentially radioactive waste gases. One sub-system consists of the waste gas surge tank, com-pressers, and decay tanks. The other subsystem consists of the auxiliary building gas collection header. Inputs to these subsystems are as follows: Waste Surge Tank Inputs 0 0 0 0 0 0 Containment vent header, which isolates on CHP and/or CHR. Nitrogen addition. Dirty waste drain tank relief valve discharge. Auxiliary building equipment drain tank vent. Vacuum degasifier gas discharge. Containment atmosphere' sampling gas discharge. 15 -

0 0 0 0 0 eves purification and deborating demineralizer vents. Radwaste evaporator vents. eves volume control tank vent. Gas decay tank relief valve discharge. Waste gas compressor discharge header relief valve effluent. Gas Collection Header 0 0 0 0 0 0 0 0 0 0 Containment clean waste receiver tank vent, which isolates on CHP and/or CHR. Dirty waste tank vent. Controlled chemical lab tank vents. Radwaste demineralizer vents. Filtered waste monitor tank vent. Treated waste monitor tank vent. Volume control tank N2 and H2 supply header relief valve vent. Component cooling water surge tank vent. Charging pump seal loop tank and well vents. Radwaste building vent gas collection header. Based on our review, each input from containment is isolated following a serious accident. As a result of the plant modifications discussed above, none of the other inputs would be radioactive. Accordingly, the waste gas system will not be contaminated follow-ing an accident

16 -

In summary, our review has indicated that a relatively few systems outside of containment could become highly radioactive following a serious accident. These include the engineered safeguards systems and the sampling systems, which will be included in a formal leakage reduction program. In addition, the specific plant modifications discussed above will be made to ensure that highly radioactive fluids will be kept within containment. This is considered a technically desirable approach to minimize both off-site and on-site radiation expo-sures. It is considered that this appro~ch is fully r~sponsive to the Nuclear Regulatory Commission recommendation. 17 -

Attachment B REVIEW OF CONTAINMENT PENETRATIONS A review was made of each line which penetrates contain-ment to determine, for.the current plant configuration, which systems outside containment could potentially contain radio-active material in the event the reactor coolant or containment atmosphere were highly contaminated. The containment penetration configurations have been re-cently verified, as described in MPR Report 639, "Palisades Nuclear Plant - Independent Review of Containment Penetrations," dated November 1979'. The penetration configurations described in MPR-639 were employed for the current review. The results of that review are documented in the enclosed checklist, which indicates for each penetration: 0 0 0 0 0 The containment penetration number. The system name and associated P&ID number. The signals, if any, which would isolate the line. An indication* of whether or not fluids in the line would be potentially radioactive. A conclusion as to whether or not the associated system outside containment would potentially contain highly radioactive materials in the event of a serious accident, and the basis for that conclusion. In summary, the review indicates that, without any plant modifications, the following systems outside containment could contain radioactive materials: 0 0 0 0

o.

0 High pressure safety injection system in the recirculation mode. The containment spray system in the recirculation mode. Low pressure safety injection system in the recirculation mode. The containment air sampling system. The chemical and volume control system. Reactor coolant sampling system.

0 The vacuum degasif ier system. 0 The liquid waste system. (1) 0 The gaseous waste system. (1) (1) These systems are not directly contaminated by fluids which escape through containment penetrations, but would be contaminated by processing fluids from other systems outside containment. 2 -

~W OF LINES WHICH PENETRATE CONTAINMENT AS POTENTIAL SOURCES OF LEAKAGE FROM SYSTEMS OUTSIDE CONTAINME T CONTAINMENT PENETRATION NUMBER SYSTEM NAME (P&ID NO.) 1 Purge air supply (M-218) 2 Main steam line (M-205, M-207) 3 Main steam line (M-205, M-207) 4 Purge air exhaust (M-218) 4A Purge air exhaust sample line (M-218) 5 6 7 8 9 10 SG(A) bottom blow (M-207/ M-226) SG(B) bottom blow (M-207/ M-226) Feedwater to SG(A) (M-207) Feedwater to SG(B) Spare Service air (M-212)

CHP Containment High Pressure CHR Containment High Radiation ISOLATION SIGNAL Blower shuts on CHP* and CHR*; valves normally closed Steam Generator A or B low pressure Steam generator A or B low pressure CHP and CHR Locked closed CHP and CHR (also normally closed)

CHP and CHR (also normally closed) Open Open Locked closed PAGE 1 OF 7 POTENTIALLY RADIOACTIVE No No (except if tube leak at high pressure) No (except if tube leak at high pressure) Yes Yes CONCLUSION Not a source (not radio-active) Not a source (can isolate if a tube leaks - evalua-tions indicate doses due to short-term operation with a leak are acceptable) Not a source (can isolate if a tube leaks - evalua-tions indicate doses due to short-term operation with a leak are acceptable) Not a source (isolates) Not a source (locked closed) No (except if tube leak Not a source (normally non-at high pressure) radioactive and closed; also isolates) No (except if tube leak Not a source (normally non-at high pressure) radioactive and closed) No No No Not a source (not radio-active) Not a source (not radio-active) Not a source (not radio-active; locked closed)

REVIEW OF LINES WHICH PENETRATE CONTAINMENT AS POTENTIAL SOURCES OF LEAKAGE FROM SYSTEMS OUTSIDE CONTAINMENT CONTAINMENT PENETRATION NUMBER SYSTEM NAME (P&ID NO.) ISOLATION SIGNAL 11 Condensate to shield cool-CHP and CHR ing surge tank (M-221) 12 Service water supply Remain open (M-208) (to containment air coolers) 13 Service water return Remain open (M-209) 14 Component cooling water in Remain open (M-209) (letdown HX, RC 15 16 17 18 pump HX) Component cooling water out Remain open (M-209) SG(A) surface blowdown (M-207 and M-226) Containment pressure instruments (M-218) Fuel transfer tube CHP and CHR Locked open Locked closed 19 Personnel lock Double isolation doors 20 Spare 21 Spare PAGE 2 OF 7 POTENTIALLY RADIOACTIVE CONCLUSION Yes (when clean fluid Not a source (isolates) not being pumped in) No Not a source (not radio-active) No Not a source (not radio-active) No Not a source (not radio-active) No Not a source (not radio-active) No (unless a tube leak) Not a source (isolates) Yes Is a source, but tested during containment ILRT Yes Not a source (locked closed) Yes Not a source (double doors)

REVIEW OF LINES WHICH PENETRATE CONTAINMENT AS POTENTIAL SOURCES OF LEAKAGE FROM SYSTEMS OUTSIDE CONTAINMENT CONTAINMENT PENETRATION NUMBER SYSTEM NAME (P&ID NO.) ISOLJl,TION SIGNAL 22 Redundant HP safety injec-Open on SIS tion (M-203 and M-204) 23 Redundant HP safety injec-Open on SIS tion (M-203 and M-204) 24 Spare 25 Clean waste receiver tank CHP and CHR ~ent to stack (M-210) 26 Nitrogen to quench tank (M-222) CHP and CHR 27 ILRT fill line Blind flange during normal operation 28 29 30 31 32 tontainment air sample (M-224) Spare

Containment spray pump discharge (M-203)

Containment spray pump discharge (M-203) LP safety injection Normally closed solenoid valves Opens on high containment pressure Opens on high containment pressure Opens on SIS PAGE 3 OF 7 POTENTIALLY RADIOACTIVE CONCLUSION Yes (in recirculation Is a source mode) Yes (in recirculation Is a source mode) Yes Not a source (isolates) Yes (if at low pressure) Not a source (isolates) Yes Yes Yes (in recirculation mode) Yes (in recirculation mode) Yes (in recirculation mode) Not a source (blind flange) Is a source (would need for sampling) Is a source Is a source Is a source

REVIEW OF LINES WHICH PENETRATE CONTAINMENT AS POTENTIAL SOURCES OF LEAKAGE FROM SYSTEMS OUTSIDE CONTAINMENT CONTAINMENT PENETRATION NUMBER 33 34 35 36 37 38 39 40 41 42 SYSTEM NAME (P&ID NO.) ISOLATION SIGNAL Safety injection tank drain Locked closed (M-204) Spare Shutdown cooling return (M-204) from primary loop 2 Can be opened Letdown purification ion exchanger (M-202) Primary system drain tank pump return (M-210) Condensate return from steam he~ting units (M-215) ~ontainment heating system steam supply (M-215) Reactor coolant sample line (M-219) (pressurizer vapor and liquid, loop 2 hot leg, quench tank vapor and liquid) Degassif ier pump discharge (M-210) Demineralizer water to quench tank (M-204) CHP and CHR CHP and CHR CHP and CHR CHP and CHR, also spool piece CHP and CHR CHP and CHR CHP and CHR PAGE 4 OF 7 POTENTIALLY RADIOACTIVE No Yes Yes Yes No No Yes Yes, from sampling system No CONCLUSION Not a source (locked closed) Is a source Not immediately, could be if use letdown Not a source (isolates) Not a source (isolates, not radioactive) Not a source Is a source (isolates initially, but needed for sampling) Is a source Not a source (isolates, not radioactive)

REVIEW OF LINES WHICH PENETRATE CONTAINMENT AS POTENTIAL SOURCES OF LEAKAGE FROM SYSTEMS OUTSIDE CONTAINMENT CONTAINMENT PENETRATION NUMBER SYSTEM NAME (P&ID NO.) ISOLATION SIGNAL POTENTIALLY RADIOACTIVE CONCLUSION 43 Spare 44 Controlled bleedof f from CHP and CHR Yes Not a source, unless con-RCPs (M-202) elude controlled bleedoff to letdown instead of primary coolant drain tank is needed 45 Charging pump discharge Remains open No, unless use letdown Not a source, unless use (M-202) letdown 46 Containment vent header CHP and CHR Yes Not a source (isolates) (M-211) 47 Primary system drain tank CHP and CHR Yes Not a source (isolates) pump suction (M-210) 48 Containment pressure .Stays open Yes Is a source, but covered instrumentation* by containment ILRT 49 Clean waste receiver tank CHP and CHR Yes Not a source (isolates) circulating pump suction (M-210) 50 . Emergency access Double doors Yes Not a source (double doors) 51 Equipment door Closed Yes Not a source (closed) 52 Containment sump drain to SIS and CHR Yes Not a source (isolates) dirty waste tank (M-211) PAGE 5 OF 7

REVIEW OF LINES WHICH PENETRATE CONTAINMENT AS POTENTIAL SOURCES OF LEAKAGE FROM SYSTEMS OUTSIDE CONTAINMENT CONTAINMENT PENETRATION NUMBER SYSTEM NAME (P&ID NO.) 53 Containment spray pump suction (M-204) 54 Containment spray pump suction (M-204) 55 SG(B) surface blowdown (M-207 and M-226) 56 57 58 59 60 61 62 63 64 65 Spare Spare Spare Spare Spare Spare Spare Spare Reactor cavity fill and recirculation (M-221) Instrument air (M-212) ISOLATION SIGNAL Opens on SIRW tank low level Opens on SIRW tank low level CHP and CHR Locked closed Stays open 66 ILRT instrument line Locked closed 67 Clean waste receiver tank CHP and CHR pump recirculation return, (M-210) PAGE 6 OF 7 POTENTIALLY RADIOACTIVE CONCLUSION Yes, in recirculation Is a source mode Yes, in recirculation Is a source mode No, unless SG tube leak Not a source No No Yes Yes Not a source (locked c],osed) Not a source (not radio-active) Not a source (locked closed) Not a source (isolates)

REVIEW OF LINES WHICH PENETRATE CONTAINMENT AS POTENTIAL SOURCES OF LEAKAGE FROM SYSTEMS OUTSIDE CONTAINMENT CONTAINMENT PENETRATION NUMBER SYSTEM NAME (P&ID NO.) ISOLATION SIGNAL POTENTIALLY RADIOACTIVE CONCLUSION 68 Air supply to air room CHP and CHR No Not a source (M-218) 69 Clean waste receiver tank CHP and CHR Yes Not a source pump suction (M-210) 70 Spare 71 Spare 72 Reactor refueling cavity Locked closed Yes Not a source (locked drain (M-221) closed) 73 Spare (with capped pipe in place) PAGE 7 OF 7

Attachment C REVIEW OF POTENTIALLY P.ADIOACTIVE SYSTEMS A review was made of each potentially radioactive system outside of containment to determine the specific systems or portions of systems which would contain radioactive materials following a severe accident. This review also identified the various functions which must be performed following an accident, and the means for accomplishing these functions with available equipment. The results of the review are documented in the enclosed checklist. This checklist presumes that the plant modifica-tions described in Attachment A will be implemented. The checklist includes each potentially radioactive system identi-fied in Attachment B, and for each includes: 0 0 0 0 The system name and associated P&ID number. The subsystem involved, e.g., specific lines, or portions of the system which perform a unique function. The review results for a design basis accident, which is considered to be a break in the reactor coolant system pressure boundary, either large or small in size, which cannot be isolated. The results indicate whether sub-system components would be radioactive, and defines how various functions are accomplished. The review results for a TMI type accident, which is con-sidered to be an isolable break in the reactor coolant pressure boundary. The results of the review indicate that the following systems could contain radioactivity following a severe acci-dent: 0 0 0 0 The high pressure safety injection system. The containment spray system. The low pressure safety injection system. The reactor coolant and containment atmosphere sampling systems

SYSTEM Safety Injection, Containment Spray Shutdown Cooling (M-203, M-204)

IDENTIFICATION OF SYSTEMS OUTSIDE CONTAINMENT WHICH COULD CONTAIN RADIOACTIVE FLUID FOLLOWING AN ACCIDENT SUBSYSTEM HP safety injection: from containment sump through HP pumps to HP safety injection con-tainment penetrat~on 23 Redundant HP safety injection: from HP pumps to redundant HP safety injection con-tainment penetration 22 LP safety injection: from containment sump through LP pumps througl shutdown cooling heat exchangers to LP injec-tion containment pene-tration 32 Containment spray: from containment sump through containment spray pumps, through shutdown heat exchangeri to containment spray containment penetrationi 30 and 3l Containment spray re-circulation line from shutdown heat exchanger~ to SIRW tanks DESIGN BASIS ACCIDENT Contents will be radioactive: operated in recirculation mode (suction from containment sump) following depletion of SIRW tank. Contents will be radioactive: may be required to operate in recirculation mode. Contents will be radioactive: operation in re-circulation mode may be required for long-term cooling of primary coolant system. Contents will be radioactive: operation in re-circulation mode may be required for reduction of containment pressure and for iodine reduction in containment atmosphere. Contents not radioactive: this line is used for periodic pump test but is normally locked closed and would be closed during and following an accident. PAGE 1 OF 10 TMI-TYPE INCIDENT Contents not expected to be radio-active; recirculation from contain-ment sump is not required. Contents not expected to be radio-active; recirculation from contain-ment sump is not required. Contents may be radioactive: operation in.recirculation mode may be required for long-term cooling. Contents not expected to be radio-active; recirculation from contain-ment sump not required. Contents not radioactive (same as for OBA)

SYSTEM Safety Injection Containment Spray Shutdown Cooling (M-203, M-204) (continued) IDENTIFICATION OF SYSTEMS OUTSIDE CONTAINMENT WHICH COULD CONTAIN RADIOACTIVE FLUID FOLLOWING AN ACCIDENT SUBSYSTEM Connecting lines from discharge of shutdown heat exchangers to suc-tion of HP safety injec-tion pumps DESIGN BASIS ACCIDENT Contents radioactive: operation in recirculation mode may be required for long-term cooling of primary coolant system PAGE 2 OF 10 TMI-TYPE INCIDENT Contents not expected to be radio-active: recirculation from contain-ment sump not required. Recirculation (minimum flow) lines from dis-charge of HP, LP and containment spray pumps to SIRW recirculation inlet valves. Contents are radioactive up to SIRW recirculation Contents are radioactive (same as SIRW tank inlet valves. for OBA). Contents not radioactive, SIRW. inlet lines are isolated when recirculation taken from contain-ment sump. Contents not radioactive (same as for OBA)

SYSTEM Chemical and Volume Control [M-202] IDENTH,ICATION OF SYSTEMS OUTSIDE CONTAINMENT trIBICH COULD CONTAIN RADIOACTIVE FLUID FOLLOWING AN ACCIDENT SUBSYSTEM Letdown system: from letdown heat exchanger through containment penetration 36 through purification filters to volume control tank Volume control tank Line from volume controJ tank through charging pumps to containment penetration 45 (to regenerative heat exchangers) Boronometer piping PAGE 3 OF 10 DESIGN BASIS ACCIDENT TMI-TYPE INCIDENT Not radioactive: these lines are isolated upon Not radioactive: letdown not re-high radiation signal and would not be re-opened quired following TMI-type acci-as long as primary coolant is highly radioactive. dent. Not required Not radioactive: although charging pumps can. be actuated by safety injection signal, suction will be from non-radioactive supply. Boron addition is accomplished by the engineered safeguards equipment. Not radioactive: boron concentration of primary coolant will be determined by primary coolant sample analysis (see table for sample system) Volume control tank is normally used in degassing and sampling primary system and collecting primary coolant pump seal leakage. Plant modifications will be made to allow: Degassing to be accomplished in pressurizer with venting to containment atmosphere (addi-tion of remotely operated con-trol valve probably required for pressurizer). Sampling to be accomplished with special sample system (see table for sample system) In addition, primary coolant pump seal leakage will be allowed to collect in primary coolant drain tank and containment sump if required. Not radioactive as long as volume control tank is not radioactive (see above). Boron addition is accomplished by the high concentra-tion boric acid tanks. Not radioactive: (same as for (OBA)

SYSTEM Chemical and Volume Control (M-202) (continued) SUBSYSTEM Purification Ion Exchangers IDENTIFICATION OF SYSTEMS OUTSIDE CONTAINMENT WHICH COULD CONTAIN RADIOACTIVE FLUID FOLLOWING AN ACCIDENT DESIGN BASIS ACCIDENT Not required Debora ting Ion Exchangei Not required PAGE 4 OF10 TMI-TYPE INCIDENT Not required Not required

SYSTEM Sampling System (M-219 and M-224) IDENTIFICATION OF SYSTEMS ou*rsIDE CONTAINMENT WHICH COULD CONTAIN RADIOACTIVE FLUID FOLLOWING AN ACCIDENT SUBSYSTEM (M-219) NSSS sample line from containment penetration 21 through sample coolers to volume con-trol tank (M-202) or vacuum degasifier tank (M-210) Miscellaneous liquid sample lines from the following: containment spray pump discharge, LP injection pump dis-

charge, HP injection pump dis-charge; SI tanks drain Purification ion ex-changer inlet purification filter outlet, purification ion exchanger outlet SIRW tank recirculation (M-224)

Containment building atmosphere sample line from containment pene-tration 28 through vacuum pumps and moisture separators, n2 and 02 analyz~rs to waste gas surga tank (M-211) and equipment drain tank (M-210) DESIGN BASIS ACCIDENT Samples for radioactive primary coolant will be required. New sample lines and panel will be installed for use with highly radioactive fluid. Sample fluid will be returned to containment. These lines coula be radioactive to the first isolation valve (near sample panel). Modifica-tions will be made to eliminate these lines or isolate within engineered safeguards rooms. One special sample line at pump discharge should be considered for use in the event that PCS pressure is too low to provide PCS sample. Not radioactive: sources will not be contaminat-ed (see table on chemical and volume control system) Samples of radioactive containment atmosphere will be required* New sample lines and panel will be installed for u*se with highly radioactive gas. Sample fluid will be returned to contain-ment. PAGE 5 OF 10 TMI-TYPE INCIDENT Samples of *radioactive primary coolant will be required (same as for OBA). New sample panel will be used. These lines could be radioactive to the first isolation valve (near sample panel). Modifications will be made to eliminate these lines or isolate within engineered safeguards rooms. These lines will not be used following TMI-type accident. Not radioactive: (same as for OBA) Samples of radioactive contain-ment atmosphere will be requir-ed; (same as for OBA)

SYSTEM Sampling System (M-219 and M-224) (continued) IDENTIFICATION OF SYSTEMS OUTSIDE CONTAINMENT WHICH COULD CONTAIN RADIOACTIVE FLUID FOLLOWING AN ACCIDENT SUBSYSTEM Miscellaneous gas sample lines from the follow-ing: Volume control tank Waste gas surge tank Waste gas decay tanks DESIGN BASIS ACCIDENT These lines will not be radioactive since the sources will not be contaminated (see ta.bles on chemical.and volume control system and waste gas system) PAGE 6 OF 10 TMI-TYPE INCIDENT These lines will not be radioactive (same as for DBA)

SYSTEM SUBSYSTEM IDENTIFICATION OF SYSTEMS OUTSIDE CONTAINMENT WHICH COULD CONTAIN RADIOACTIVE FLUID FOLLOWING AN ACCIDENT DESIGN BASIS ACCIDENT (OBA) Waste Gas - Vacuum 0 Input from primary sys-PNot radioactive -- the line is isolated by CHP Degasifier (M-210) tern drain tank (located or CHR, and does not need to be reopened. inside co~tainment) PAGE 7 OF 10 TMI-TYPE INCIDENT 0 Same as OBA. 0 Input from auxiliary building' equipment drain tank 0Not radioactive -- modifications will be made to 0 Same as OBA, prevent radioactive inputs from: 0 Letdown from chemical and volume control sys-tem upstream of volume control tank (letdown can go to either the VCT or vacuum degasifier 0 Nitrogen addition line 0 Nuclear steam supply ~ampli~~ sys£em dis-charge line Poegasifier tank to degasif ier pump suction 0 Degasifier pump dis-charge to containment -- Safeguards room sump, and -- Safeguards equipment relief valves PNot radioactive -- the letdown system will not be used following a OBA. PNot radioactive. PNot radioactive -- modifications will.be made to return sample system effluent to containment. PNot radioactive, based on modifications noted above. 0 Not radioactive, based on modifications noted above. PVacuum pump and associ-PNot radioactive, based on modifications noted ated seal coolant above. circuit 0 Same as OBA. Note that reactor coolant degasification will not require use of the letdown system. This will be accomplished by venting the pressurizer inside containment. 0 Not radioactlve. 0 Same as OBA. 0 Same as OBA. 0 Same as OBA. 0 Same as DBA.

SYSTEM Waste Gas - Surge Tank SUBSYSTEM IDENTIFICATION OF SYSTEMS OUTSIDE CONTAINMENT WHICH COULD CONTAIN RADIOACTIVE FLUID FOLLOWING AN ACCIDENT DESIGN BASIS ACCIDENT (DBA) °Containment vent header 0 Not radioactive. The header isolates on CHP input to surge tank or CHR, and does not need to be re-opened. 0 Nitrogen addition line 0 Not radioactive. input to surge tank PAGE ti OF 10 TMI-TYPE INCIDENT 0 Same as DBA. 0 Same as DBA. 0 Dirty waste drain tank relief valve discharge to surge tank 0 Not radioactive. A modification will be made to 0 Same as DBA. ensure safeguards room sump effluent is not 0 Equipment drain tank vent to surge tank 0 Vacuum degasif ier vacuum pump discharge to surge tank 0 Auxiliary building gas analyzers input to surge tank 0 Purif ication and de-borating demineralizer vents to surge tank 0 Radwaste evaporator vent 0 Volume control tank vent to the surge tank 0 Waste gas relief valve input to surge tank 0 Waste gas compressor discharge header relief valve input to surge tank routed to the dirty waste tank. 0 Not radioactive. A modification will be made to 0 Same as DBA. ensure safeguards room sump effluent is not routed to the equipment drain tank. 0 Not radioactive, based on sampling system modifications. 0 Not radioactive, based on sampling system modifications. 0 Not significantly radioactive; normally closed. 0 Not significantly radioactive. 0 Not radioactive, based on sampling system modifications. 0 Not radioactive. 0 Not radioactive. 0 Same as DBA. Also modifications to permit coolant degasification inside containment will elimi-nate this source. 0 Same as DBA. 0 Same as DBA. 0 Same as DBA. 0 Same as DBA. 0 Same as DBA. 0 Same as DBA.

SYSTEM Waste Gas - Surge Tank (continued) Waste Gas - Gas Collection Header IDENTIFICATION OF SYSTEMS OUTSIDE CONTAINMENT WHICH COULD CONTAIN RADIOACTIVE FLUID FOLLOWING AN ACCIDENT SUBSYSTEM DESIGN BASIS ACCIDENT (OBA) 0 Surge tank to waste gas 0.Not radioactive. compressor suction 0 Waste gas compressors and decay tanks °Clean waste receiver tank vent 0 Dirty waste drain tank vent 0 Not radioactive. 0 Not radioactive. Isolates on CHP or CHR 0 Not radioactive. A modification will be.made to ensure safeguards room sump effluent is not routed to this tank. °Controlled chemical lab 0 Not significantly radioactive. tank vents I * ) PAGE 9 OF 10 TMI-TYPE INCIDENT 0 Same as OBA. 0 Same as DBA. 0 Same as DBA. 0 Same as DBA. 0 Same as OBA. 0 Radwaste demineralizer 0 Normally closed. Not significantly radioactive. 0 Same as OBA. vents °Filtered waste monitor 0 Not significantly radioactive. tank vents 0 Treated waste monitor 0 Not significantly radioactive. tank vent 0 Volume control tank 0 Not radioactive. N2 + H2 supply h~ader relief valve vent 0 cornponent cooling surge 0 Not radioactive. tank vent 0 Same as OBA. 0 Same as OBA. 0 Same as DBA. 0 Same as OBA.

SYSTEM Waste Gas - Gas Collection Header (continued) IDENTIFICATION OF SYSTEMS OUTSIDE CONTAINMENT WHICH COULD CONTAIN RADIOACTIVE FLUID FOLLOWING AN ACCIDENT SUBSYSTEM DESIGN BASIS ACCIDENT (OBA) °Charging pump - seal 0 Not significantly radioactive. loop tank vents and well vents 0 Radwaste addition vent 0 Not significantly radioactive unless waste water gas collection header transferred to radwaste. PAGE 10 OF 10 TMI-TYPE INCIDENT 0 Same as DBA. 0 Same as OBA.

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--'-*---* ~*
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..: ~:......

. i:1*cH111.* COMPANY* &WO 8428 PALiSADES PLANT*

... :CONSUMERS -:POWER COMPANY.

~ -~

),.. i ' l i ' ' \\ \\ \\ RELAY* CONTACl Na svx-o Q522B' FUNCTION ~. ~;~** -~ I DWG. I l N~ AUX FEEDWATER. i ~ I PUMPPB-B AuroE-92.38 3

. 7 TR IP ALARIV\\ j (QJSH.3 5P,.\\RE

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RE.L.Ay PSX-0741A 'oN. PAt>.\\EL.:_

C 0 I Tc BE R.E TAG-<iED AS PS~-0741 A. Rote~. l"'*i ~ ~* ~

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h~

9'000 ~eries for t ~t d*1~~*:r..~c1=i.... f t hi:.:. proiec~ a-nd will revert. *to. the *)~* i.;; i r.al.;iumber "1hen ~ 1 u!ssu~d !~r Recnrj" rP~isi 1.~r~ i~ issued. _1 l

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Cl2.R I P21 1N21,21-~

(ENCLOSE IN 29' SP(lH) (ENCLOSE IN FLEXIBLE CONDUIT) FLEXIBLE CONDUIT) I J545"

,23,25(1Z~): ~

123,2.5(122)~ lfil :!P21,2.2(122): J543 I

.11-23, 25, SP, SP, SP (225) (CABLE IN LEFT [_A____,_l I--j CHANNEL) ---.2.6,27,P21 2B,N21,24, SP,SP, SP (249) I _fQ!_ I NOTES: I. FOR CONTACT DEVELOPMENT OF AUX. IC?ELAVS PSY-0741 _ PSY*P6 P5l-P8 WOl4) W0/4 W0/4 & 5YX-C/05-22B WOl4 SEE RELAY TABLE E-9238 (Q)SH. 8 ON DWG. REFEIC'ENCE DWG5: l. 1 LOGIC DIAGIC'AM: SK-JLG-96. 2 COt\\HJECTION DIAGRAMS:~~ AUXILIARY FEEDWATER PUMP TURBINE. STEAM VALVE 5V*0522B CONTROL-AUXILIARY RELAYS SC.HEME-W0/4 SCALE ~ DESIGNED I... /C. !DRAWN J,COLPEAN WORK ITEM""3 ~:ote: c:..;o: SYSTEMS MODIFICATIONS PER. NURE.G 057B This -::rawing is tc::iporarily rer:urr~b~re~ in the 9PCJO series for the *:._!urati<.,n of the project ~H?d -..;i:l revert to t!*~e original numt:er whe1~ the !ss1~e~ f0r ~ecc.rd ~e~'~sion is issued~ BECHTEL COMPANY GWO B42B PALISADES PLANT CONSUMERS POWER COMPANY SCHEMATIC DIAGRAM

MAIN STEAM

l SOLATION VALVES

~ JOB No. DRAWING No. REV. ~ 12.447-039 E-92.38(0.)SH. 6 A

... ~ -~

i

1 *'

. r* ..,... \\ CL0.S£

'© '

/ " / ' AUTO C.LO~E 1 OPEN AUTO <F.V.l OPEN I 1.3 I 13 h..Lo hl.o 0,0:010 0 o:o 0 z. 14. trLo:~ 2 4 0 010. 0 0 00 0 I I 5 17 5 17 hln'hl.o o oro o c:i_:>: ~ I* ~!~ 18 6 f, 16 0 -010 0 0 010 . I. I I CONTROL SWITCH - Hs-oszas 'MAINTAINED CONTACTS. TURBINE DRIVEN AUX. FEED WATER PP AUTO ~.S.TART SELECTOR. LOC-C.01 (5.E E NOT£ I) 0 tts-osaa & CO.NT CLOSE 'AUTO

- NO.

t.

2.

x 3 GE..C.R2,40 4 x TYPE 5 x US i!03K G 1( GE, CR2940 7 x TYPE 8 x U82.03 L.

DENOTES SPARE (SEE NOTE' I),

OPEN x x

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rn 't'V$ TESTTDAFPP,AUTOSTART *

- - .:_*.*~*....

/'!!), ( LOC.. FRONT OF PNL.C.H L).*.. **.... ss*~a/p:a:a. 'ry/ TEST TDAF PP AUTO STARi ( F.V.) ,..'-V,(LOC.. REAR OF' PHL.CllR) CONT. : --.*.- ... *tt ............ AL TEST NOR~AL TEST .,_N_o_. _____

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0--'-+--*-*~**-*--i hl.d6..L.a I 13 I 13 0 .0~ 1 ;.i_o 0 OtO I ~!&Lo 2 14 2 I .0 QrO 0 0 010 0 I I I !b '1..Lci 01 hl..olhi.c h.L.o'7 10 0 ~:a Is 6

V-u 10 0 0

010 0 0 01 I I TEST SELECTOR SWITCH ss-l/P88 SPRING RETURN TO NORMAL KEY REQUIRED TO TURN TO TE.$T . LOC. REAR OF PANEL Cll'.'"R GE. CR2.940 I X ~<: .,_~----~--......._--"""""" _______ _ TYPE

2.

x USZ03A' GE. CRZ.940 TYP£ US2.03E AUXILIARY FEED WATER PP Tl)RBINE. STEAM VALVE-SV05'2.2. 8 CONT1\\0L NOTE~ :THIS CONTROL SWITCH REPLACES THE . EXISTING CLOSE-OPEN SWITCH HS-0522.B (GE TYPE. 5BM) ON PANEL-COi. CONTROL.SWITCH DEVELOPMENT SC.HEME WOl4-Note: WORK ITEM *3 GWO: 84H.SYSTEMS MODIFICATIONS PE.R. NUREG..$S7B Thi~ drawing is temporarily renumbered in the 9000 series for the duration of the project and will rever~ to the original number when. the ~Is~u!?d for R~ .. cord" revision is issued

SCALE -

IQ&,IGllllO 1,..

... BECHTEL COMPANY '*..

' GWDS4U

... PAL1SADEs PiANT

~~ERS ~" COMP~'11Y.

II

- 11.1*;

!t .. ~..... * ~. r.** RELA'< TASLE 62-2/P*B8 $SX-2/P8B SSY-2/P8B SVY-0 /W0~4- ,052.2 B WOl4 W014 Wnt4-RELA'Y DWG. RELAY OWG. RELAY OWG. RELAY DWG. CONTAC.T FUNCTION f'.12. CONTAC.1 FUNCTION N2 CONTAGT FUNCTION N2 CONTAGT FUNCTION Na NQ N9. N2 N2 ENERG1ZE AU)(. E-,238 SEAL-tN E-C)238 ENER&IZE TIME E_,238 SEAL-IN E-9238 ,~rs RELAY SVY-0 <QlSH,5 ,.,~7 ~)5H.5 '~ f-.1 DELAY RELAY (Q)SH.5 1~~7 CQ>:SH.5 '(( . 052ZB G2-2./P8B . 0P£ N. SV-0~22.S: r:: SPARE ENER&f7E AUX. E-,238 l~OICATION E-92.38 2.~ 1--8 ENER'1ZE SV*0522B E-9.23$ z~ f--e RELAY* QlSH.5 ~1--8 (TEST TDAFPP <Ql5H6 C START TOAFP~ 'Q)SH.3 SVY"':O/ ()52. 2. B AUTO START} P-86) PSY-0741 LWOl4-P~V-PSlWOt4 svx-c /W014 0522.B --- P5Z*PBLWOl4 RELAY DWl:ii. RELAY DWGi. RELAY OWG. RELAY DW6. CONTACT FUNCTION N2. CONTACT FUNCTION Ne CONTACT FUNCTION N2 CONTACT FUNCTION N! N2 N2 N.R. N2 E'NERGIZ.E AUX E-?238 SEAL-IN E*,238 BLOGK TEST E-9238 ';m SPARE' ,~~7 RELA'l p5y-.pa <O>SH.'- ji~7 ~Q>SH.&. 3*7 CQ}SH.,- 3 7 ENERGIZE AUX ~,238 ENER&l?E AUX E-9238 BLOC.KOPEN £-9238 DEENER61'2E 2~r-8 RELA\\' P5Y-P8 ~QlSH.(, 2~r--8.RELAY ~vx-c/ lQ)5H.'1 ~8 SV-05226 (Q)SH.5 5V-05Z2B E-9238 (ENERGIZE AUX. 05228 (TRIP RELAY S\\JY-D) 4*8 <TRIP TDAFPP, P~B) KQISH.3 TDAFF'.P f*8B) OS22B AU'lCILIAR'l RELA"l TABLE-AUXILIARY FEEDWATER PUMP TURBINE STEAM VAL\\/E-SV0522B SC HE ME WOl4-TDAFPP-TURBINE DRIVEN AUXILIARY FEEDWAT.ERPOMP WORK ITEM N9 3 NOTE:GWO 8428 SYSTEMS MODIFICATION PER NUREG 0578 This drawing :lS tempox;;n:ily. renumbered.in* the 9000. *~er:f.e.s for the duration of th~ project .. and Will revert to the original *::nll"1!ber.when.* the.

~Isa.ue*f for Record" tevis.1o.n i13.issued.

SCALE 'V I DESIGNEOI.* ....... DRAWN. "'" ti. IC.A "' BECHTEL COMPANY* . *.. &we::a42e.....

SCHEMATtc:.DIAGRAM-~ -

MAir*t.S:t'~AM - d ISOLATION '.\\/AL \\/ES * *. peNci> >*;_\\; ;l)~N6~:.... :MY. 1244~03~ '>(~~2~CQlSH.a'.:>.:. :A.

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vAt.VE VACUVt*A CLOSED FA1LUR£~*V1BR/ffJ'CJF *

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t:iir~cuc. *'*.1..

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REF:

CONTACT FUNCTION DWG. N !!. START E-jl9" m MDAFPP P-8A CQ)SH.2 -m START E-9238 TDAFPP P-8B CQlSH.5 RELAY TABLE' TOA FPP-Tl,IRBINE DRIVEN AUXILIARY FEEDWATER PUMP MDAFPP-MOTOR DRIVEN AUXILIARY FEEDWATER PUMP SCALE A.J (DESIGNEDL..KJ\\'L1Jt'WJl5AMllDRAWN WI ILL<AN BECHTEL *coMPANY G.WO 8428 PALISADES PLANT CONSUMERS POWER COMPANY SCHEMATIC DIAGRAM FE EDWATER TURBINE CONTROL ~ JOB No. DRAWING No. ~ 12447-039 E-~187CQ) Sl-1.Z REV.

L r ~:...* :

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I-

,I 'I I 2 152-104 - I 15'2*104 cs-* --cs-AJ_/ I.. O.SUCf"I()~ P2SS. PS 0142 .,.. '*.-..* *.. *e*. _ l!YeJ:Jn te.ec_.. ~ -- wt:(;.~~c~. AUX.FOWTR

PUAIP(M()ToR)

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efLAY CllL J543

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- TEST AUX.

RELAY TEST MDAFPP _AUTO ~1:ART REAR Of'CllL I *FRO NT OFC:llL MAIN F'EEDWATEIC MAIN FEED PUMP FLOW LOW T£5T TURBINE TRIP AUTO MANUAL .. ~

NOTES=.

.. 0 ~-P/l~o----------------~""-~---------------......__~-......~1----------....1------------------e------......---:-a----------------------1------~ . I. F"OR B*L,:OCI( DIAGRAMt, S~E.: E*919b-!Q)_SH~. P -CHL 5A -.~'>--....,_----1t---------------i--....,..----e-2 J5423 SH*2) CONT'D ON FSX-P8 SSX-1 KX-7 E~9J9b (Q)SH.3. J 543 Pll~E-9187(~)

z. FOR cdNTAC_T DEV£*LOPMENT

.N. CllL Pll Cl2R Cl 2L Pl I u---.----.. 9 FS-0709 FS*071o* . fSX* SS* I

10 (E*90b8(Q) 10.(£*90'8 (Q). 7 0707l

. *1 PSA. 7 51-1. I)

*SH. I)... *

. r!":SX (OPEN ON (OPE:.N ON. . r

C 12L LOW FLOW)

LOW FLOW) 7 l1nO. --. II Cl2R IZ CllL CllL o Cl2L o -11 . --12 J543 Cl IL o --. 12 J543 SSX-1 PSA 3)(-P8A. (E-91'1'1(Q) SH. I) 5 5 FSX-0709 FSX-0710 WOl2 WOl2

5 JS43 14 CllL 7

a P8A 1 CJ..LJ. J543 J54~ 5 J542 cotL_,.. 15 l..,. -"* l HS-~A HS-~P&A CO' 1 4 ~UTO) I (MANUA~ HS-PgA iT" (AUTO) COfo A'll 16 Af.!D 1>7 67 152-105/a. 152-106/a. foB (E*913l(Q))68 (£*9132(Q) SH. I) Afl All COi 'J 17 J543

CJ 62-l/P8A WOl2.

W A OF AU)(:J~*ilA'-l(S....

F~~~o-ro.9:~; F'sx~o1io, F~':PB *. ssx-1/. *

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62.:1/e&A~i)~~*L~Y TABLE* O~~~ '"91'4

. WOil.. -:..... * .. * ;. '.CQl SH.5 "* F="OR sw1rcw 6Evi(oPME.~T* oi::' * *_ H~-P8A) S!;r;l/WAJ S£E -.E.::9*%(~)SH* + . MDAFPP~Mo.ToR**oR:1vE.N**AOX:** *: .*. '..REEOWATER P\\JMp;* TOPU;..f:n~E DEL:*A~ON,PtC.K.IJP. ~*EF'EReNC£ OWG:

1. LooJc>o1J\\aRAM. *.sk*JLG~,,

-:~-.. 8 ~ :.. -*~~ '.. ~*. ' 6 6 6 L2 TDPU SCALE (20- 200 SEC.) u------,---i--------------i----------------~-----..i-----------~-----------------1--------~------------------------------~2 MOTOe DRiVEN AUXJLIARY r£EDWATER PUMP P-SA CONTROL-AUXILIARY RELAYS . SCHEME W012 CONT'D ON E-9/9h(Q) SH.*3: WORK IT£M f\\JS2 3 Note: GWO: 8428 SYSTEMS MOD. PE~ NUIU.G- 0578 Tnis draw:lng is temporarily.renlllilbered in t-he 9000 series for the. durntion of t*he ~t:oject*

and will revert to the orig:t:nal'riumbe.r when *th~
.,"Issued foi Record" rev:isi~n is is:s~d..:,......

:-;.. *:,...,~~~-.-. -~:*:**

- .***::*:;:\\**.~*.**:*. :,.

~- ~.. AUXILIARY RELAYS-LOW SUCTION PRESSURE 2~ LOGIC LOW SUCT. PRESSURE '. ~*. 1 AUXILIARY RELA'(S I* ---...---------~ .: 1 I I 2 . PSX-074/A I

I PS)(-

PSX- [

rr~IA 7

07418 __ 8. -:.7 I

2.

I I I PSX *0741B* 1 PSX-0741C. 'c9

J 5'43 I

7 'I 18

.PSX *0141.

7

1 CE"-~19,CQlSH.G)

I

L__ -----*-----*-""-.J COi 2

5 PSK-0741 W012. L52-104/CS. (OPEN WHEN O. cs 15.INCLOSE OR TRIP) -19. J543 ' ,.PSX-P8

- wo12

~ P.SX*PS 15/lil6Jl9 ISP(22.5) Pl I,Ntr{IZl) PSX-0741 J54.3 CllL COi ...... ' ~** ~ : ".,,-:.*.-'.:: ':' :: __ Pl I 15C122) CE,NCLOSE iN F'LEXIBLE *.,. ',, CONDUIT W11HIN BOXES)'> J542. Pl I, 11, 12, 13, Nll)4 -.-J5PS(249) Pll 11 12. ZSP<2ZS) I I ) Cl2L . I,, 17, 3 SP(225) All RE.FERENCE DWGS~ .;:._c . 1. LO"C *DIAGRA-M SK-::JL G-:,s'*,

2. CONNE.Cno~iD,W6S::t~4./'

~..

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~... *: NOTE: I. FOR CONACT ::lit'vE.LOfME.NT)or-AtlX. RELAYS PSX--~o74i:>.1 Psx:.P8 's'EE. R~LAY.. "-~*. i

wo12::--.** *:_ wo1a TASLf. ON oW~_~g~$tH(~)S~S 0
' ~*L*"

CONT'D FROM* }:-:91~,~Q) SH.2. MOTOR DRIVEN AUXILIARY FEEDWATER PUMP P-8A CONTROL-AUXILIARY RELA'<S SCHEMEW012 sf:H_E~"J'1c otA~_RA'tv(.<

MOTOA DRIVEN AUXJLl-ARY

.. FEEDWATE~ PUMP

' w.

.'. e

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/ / MANUAL AUTO \\ (F.V)

.:MANUAL AUTO*

I Io o : 3~*1 :. 2.~ I,4: I*

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I I GE-CR2~40 TYPE UB2036 GE-.CR2940 T'<PE UBZ.03~ MOOE' 5WITC.H

HS-P8A MAINTAINED

.*... *CONTACTS MOTOR DRl\\/E'N AUX. FEEDWATER . PUMP P-6A AUTO-START SELECTOR. LOC-COI . ~ " ~* ~.. CONTACT MANUAL AUTO

N!a I

x

2.

x 3 x 4 x 5 x b x 7 x 8 x -¥ DENOTES SPARE . ~~~~1 ~¥mhoc.r RONT OF' eNL.~llL)t-c ..... oN_N...... ~....,.~ 1 .... '-;:-*****-***.... N-OR_ .... _MA_ ... *-~+*-~-E~ST-t*** :i< / NORMAL> / l ~,. @),TE.ST MDAFP.P

2._"*_..**:_

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z.O 0 I 40 O I ~ '1 J~I~ i '~ 'eP o I CF, V.) NORMAL TEST I ~ I ~ I 3 ~ I : 6E-CR2940 2 I A' =TYPE I "I'

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US 0 A I

I -"w: 7w $wl 7~1~ _G&cR2940* ,o 0 I eP 0 ,o 0 I e. --T'<PE I .__ _____ !.._______ US 20.3 E'. TEST SELECTOR SWITCH ss-1 /Pg{\\ SPRIN& RETURN TO NORMAL, KEY REQUIRED TO TURN lt>TEST LOC. REAR OF' PNL.Cll-L

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<D.*. *.*. 'X _.a: ~:,.. *, MOTOR DRIVEN AUXILIARY FEEDWATER PUMP P-8A CONTROL CONTRO\\.. SWJTC.H DEVELOPMENT FV.=. FiiOf\\Jr :*v*IE'{V '...

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SCAL£ 1'V N0TE'6W084Z8WORK ITEM" 3

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FSX-070C/./WOI~. Fsx.,,0110/wo12 FSX-P 1./W012 3Y.-P8WW012 si~ci(i*g~::.

3. *.* *;:"~ ~,

.. RELAY

RELA'(

. RELA '< RE LA'Y RELAY *.. ~~:;'.., , g -*~~rACT FUNCTION ~"; 5 CO~ACT FUNCTION ~~ 6 ~~ACT FUNCTION ~~6. CO~~T FUNCTION ~~&. ~()~/~'!'~ -FUNCTION -~~Ci- . u 1. . A.:

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ENERGIZE AUX.

E*CjlCJ"

--.J l-. RELA'{_ FOR MAIN (Q)SH.2 f"'l 17 F'£EDWATE'R FLOW LOW, FSX-PI ALARM-MAIN E-9287

. 4e:ff
8 [~~WATER FLOW ~QlSH. 3 PSX-0741Lwo12

.RELAY DW<a. K;ONTAc.l FUNCTION ..Nt Nil ENERGIZE AUX. E:-CJl9" ,.,b RELA'r' PSX~PB . lQ>SH.3 . ENERGIZE AU>C. E-919'- -~~ RELA'r'.5 P.5X-P8.. Q)5H.3 ~* ' . _:,.-*:.~.... .},' ** f~: ~- ;~ ~J _i:*_ ~~~;U~':*.. *. ~:

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..... '*_. ~'.. ~. '.* f q;; J'.:[", !. *.*... t~*: '*;.;::*. ;,... **; -... ENERGIZE AU)(. E-91Cj&>

_ __JI _ RELA'r' FOR MAIN KQlSH.2 11 t97 FEEDWATER FLOW LOW I F'5X-P.1 ALARM-MAIN E-9287

~ 1Co~WATER FLOW ~QlSH4 PSX-P.6/WOl2. RELAY OW6. CONT'°'T FUNCTION Ne N9.. SEAL-IN E-~19" -rl~ 'Q)SH.3 TRIP.MDAFPP E-111')" ~rs P*&A CQ>.SH.I ENERGIZE TIM£ DELAY E-~19~ --tx___. RELAY 62-1/P-BA CQJSH.Z

r-fr"f (STAllT MOAFPP P*SA) t-----.,.-1~

NERGIZE TIME DELA~ E-CJ~~e

~ RELAY b2-2/P*8B CQ)SH.5 8 1(STAl?T TDAfPP P-8~ RELA'r' G2.-l/P-8A WOil CONTACT FUNCTION N2 CLOSE" 2400V _J L.._ ACS 152.-104 1115' (START MDAFPP P-8A) SPARE DWG. No. E-91~" (QlSH.I ENERGIZE TEST E*919~ _ 1.J-_ RELA'< SSX-1 /P8A CQ)SH.z. ~r-7 ~SPARE - MOTOR DRIVEN AUXILIAR*Y FEEOW"T£R PUMP P~8A CONTROL- ~UXILIARY RELA'c'S SCHEME WOl2. ~ ~.'.. 1'"";!'* *. *

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WORK lTE'M N*3 NOTE=GW08428.

svsTE'MS.MODIF'ICATlON PER NUREG0578

.* This drill-itng *is tempoi:'aJ"ilY. remtml;iered ~n t!te 9000 series fot tJie. durati,on oE -the* prqject *.* and wilL revert. tp the or-igj.nal n).llllbeJ: *when the qi:!>sued for ~ec~rd" revi~fon _is iss~d". SEAL-IN . e"'.CJl9b .* CQ>SH~Z. TDR~TtMEDELA'i RELA"1':,.. MDAF.PP~MOTP'fl'OFUV~NAUXILIARY.. FE-ibWATE-R~ PUMP. **...

TDA F#~*+~Ni;,.oR1vS'N.AuxtL 1ARv~
* ::#t.:\\:pw~Jc:fi-::~0..~P. * -* __ : *

~-* '; *,;'. .. ~-. I:. SCALE l"'\\J..J Diii .... **._; \\< ,)~TEL '(:o~~*t,tf*. ~./ '*'.-',

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- . "-' _:.~\\:if:.. Low' SUCTION. P/i?ESSURE:
.* <,::* : i:.. ".

.. PS*0741A AUX.*RELAYS LOW SUCTION. PRESSUJ?E PS-0741B AUX. RELAYS 120V AC PANEL,. V20 J~* 120V AC PANEL. YIO..., 1 -~-

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.. PS-o74iif PS-074181 P$*074,IA (SEE NOTE I) **

'PS*0741B

. PS. "0741AT (CLOSED ON PS*0741aT (CLOSED ON d **~.*. *... ~b. LOW Pl?ESSURE) cl ;-*.. *

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J543 11 -.JS4..2. Q J 543* <:~ ~ ~ > . PSX-PSY* PSX-PSY-0 ...J 0741A 0741A o...J 0741B /_ 0741B (\\JZ L NZ b a ~ 0 ~ J543 -=--N2 J542 v------ .-. -tJ2 CllR~ CllR -. -. - -~. 5A.. . ~*. ' LOW SUCTION PRESSW2E . P5~0741C AUX. RELAYS J20V AC. PA~EL Y 30

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PS-074/i PS-0741C

v PS-0741CI (CLO.SEO ON

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u (Y') 5 <( PSY* 2 ~ 0741C 0.. 6 J542 J543 3-- .-. -. -. N3 CllLJ CllL~ ~* 5A INTE~LOCK"S J_PSX-11PSX* if0741A 7J0741B 11Psx-7To141c 21Psx- [ 0741A 8 2 PSX- . aJ0741B TO MDAFPP TRIP CKT. -co.w&. 'E-9l9b (Q) SH. 3) 11 PSY* ii PSY-21 PS Y-7T0741A 1T0741B Sf 0741~ 11PSY* ~PSY* 11 ___ 0_74_,_c ___ e_.J 0741 B TO TDAFPP TRIP CKT. (DWG. E"9238 (Q) SH~) ...1 *..

PSY**.*.

2lps-2µ.- T _074M...,,,,. -:-.*... **n:tAIO" .B (SEE NOTE I) 8. ... ** ¥- ~,'*. CLOSES STEAM VALVE: CV"'."0521 (DWG.E.-9238CQ) SH.2) ';i" *:-.~ :.....->~~~:.... :* ~-* *.. *.... ~~..,.... ;_',~.!_'. . >/",,*.. >'>. /';:'..1:i? '.~.).'... . *...... *~<*:-.. **-.;>.1;. - I. EXISTING PRESSURE. S.'l.I *:ps.;;'074l-To:-&~'.} > NOTES: R£TAGGED AS PS-oi41A~*

2. EXISTING 2./C. CABLE. Ail()'t/c,Ql~*f!f~~

/I TO BE RETAGG:~D A.$.:~::wotl/CQJ;.(. PS074lf1 Rl'.fERE.NCE DW"S"

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. i:. I; I. LOGIC: DIAGRAM SK~JLG:;,f;:~/.. :. 2~ CONNECTION.DWGS: f~*:;.**> "

~...

AUXILIARY FEEDWATER PUMPS LOW SUCTION PRESSUICE. T121P-2/3 LOGIC AUX RE.LAYS SC~EME W013

.J*...

FLEXIBLE P2,12 (122) PS0741B. . RIGHT CHANNEL

*****-***--****-**-------*~*-* -

1'l'ote: WORK. ITE:M Nii 3 'c;Wo:'.8428 SYSTEMS. MODIFICATION Pt!R NtJREG.OB7a . This drawing '.is temporarily -r~Upibered *in 'the * *: . 9000 series for t:he duration ~* .. the project atid will r.evert. to -the origi,'nal *number when the. '"I.~1!.1,1ed for RecordM z:evision is issued,

MAIN FEED PUMP PIA DRIVE TURBINE K7A HP TRIP VALVE CLOSED POS*05io3 MAIN FEED PUMP PIA DRIVE TUi?Bll-JE k7A LP Tl?IP VALVE CLOSED POS*05io2 MAIN FEtD PUMP PIB DRIVE TUli?BINE K7B HP TRIP VALVE CLOSED POS*05io6 NAIN FEED PUMP PIS DRIVE TUl?SINE K7B LP TRIP VALVE CLOSED POS*05h5 MAIN f"EEDWATER FLOW LOW PUMP P*IA FS*0709 MAIN F"EEDWATEli? FLOW LOW PUMP P-IB FS-0710 Fli?OM THIS DWG TEST SELECTOR SWITCH SS*I MOAFPP P-tlA AUX. FtED PUMPS Pli!ESSURE LOW PS-0741A AUX.. FEED PUMPS SUCTION Pli!ESSUl'f LOW PS-0741B AUX. F'E[D PUMPS SUCTION PRESSURE LOW PS-0741C CLOSE {STA/i'T MOAF'PP P*8A) 152-104/CS OPEN (STAIH TDAFPP 'P*BB) HS* 052ZB (SE£ SK*JLG -96) 4 STATION POWER Tli!'ANSFOO:MER l!ICOMllJG ACS 152*105 CLOSED STARTUP TRANSFORMER INCOMING ACS 152*106 CLOS£D "AUTO" HS-X / 152-104/CS W TEST MOAF'PP / ,AUTO-STAk'T ill 1o2-1 (20*200 SEC.) CLOSE 2400v ACB ,52-104 START r.tOTOli? DRIVEN AUX. FEEi> PUH P*8A / MOTOR Dli?IVEN -~~M~A~N~U~A~L~'_H;.;..;::.S_-X""-~~r,;;\\ AUX FEEDWATEli? PUMP /\\2/, P*BA IN MANUAL COi POSITIOIJ MDAF'PP - MOTOR Cli?IVEN AUXILIARY FEEDWATER PUMP TDAFPP - TUl<BINE DRIVEIJ AUXILIARY FEEDWATER PUMP TO DWG. SK-JLG-96 TO THIS DWG MOTOR P20TEC TIOlll TRIP r--~---124oov AC B ISZ-104 "Tli!IP* 152-104/CS LOAD SHED 3 2 DATl -.) ' MODE SWITCH HS-x, MAINTAINED CONTACTS MOTOR DRIVEIJ AUX. FEEDWATEli! PUMP P*BA AUTO START SELECTOR CONTROL SWITCH SPRING RETURN 152-104/CS ~\\, >~ ~Qi\\ .s-> TEST SE'LECTOR SWITCH SS*I SPRING RETUR~.TO NORMAL KEY REQUIRED TO TURN TO TEST AfYISIONS DUOGH(D o. VASSALLO BECHTEL ANN ARBOR, MICHIGAN PALISADES PLANT CONSUMERS POWER COMPANY GWO 8428 NUREG. 0578 iMPLEMENTATION MOTOR DRIVEN AUXILIARY FEEDWATER PUMP P-BA .IOll No. DRAWING No. 12447-039 SK-JLG-95 LOGIC DIAGRAM D c REV

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ML18044A381

Text

3.0 REFERENCES

4.0 BACKGROUND

5.0 DESCRIPTION

3.0 REFERENCES

5.0 DESCRIPTION

Subject:

Dear Mr. Vincent:

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