ML20113G350
| ML20113G350 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 01/16/1985 |
| From: | Hukill H GENERAL PUBLIC UTILITIES CORP. |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20113G351 | List: |
| References | |
| 5211-84-2300, NUDOCS 8501240325 | |
| Download: ML20113G350 (18) | |
Text
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GPU Nuclear Corporation g
gf Post Office Box 480 Route 441 South Middletown, Pennsylvania 17057 0191 717 944 7621 TELEX 84 2386 Writer's Direct Dial Nurnber:
January 16, 1985 5211-84-2300 Office of Nuclear Reactor Regulation Attn:
J. F. Stolz, Chief Operating Reactor Branch No. 4 Division of Licensing U.S. Nuclear Regulatory Commission Washington, DC 20555
Dear Mr. Stolz:
Three Mile Island Nuclear Station Unit 1 (TMI-1)
Operating License No. DPR-50 Docket No. 50-289 TMI-1 Safety Parameter Display System By letter dated October 9,1984, NRC transmitted requests for additional information with respect to the THI-1 Safety Parameter Display System.
Attached are the GPU Nuclear Corporation responses to the questions raised by NRC.
Sincerely, H. D. Hukill Director, TMI-1 HDH:SK:RS:spb cc:
R. Conte J. Van Vliet 5 8501 6 85012 05000 89 Attachment PDR A DR F
$0 GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation
RESPONSE TO NRC REQUESTS FOR ADDITIONAL INFORMATION Question 1 - Isolation Devices a.
For each type of device used to accomplish electrical isolation, describe the specific testing performed to demonstrate that the device is acceptable for its application (s). This description should include elementary diagrams where necessary to indicate the test configuration and how the maximum credible faults were applied to the devices.
b.
Data to verify that the maximum credible faults applied during the test were the maximum voltage / current to which the device could be exposed, and define how the maximum voltage / current was determined.
c.
Data to verify that the maximum credible fault was applied to the output of the device in the transverse mode (between signal and return) and other faults were considered (i.e. open and short circuits).
d.
Define the pass / fail acceptance criteria for each type of device.
e.
Provide a commitment that the isolation devices comply with the environmental qualifications (10 CFR 50.49) and the seismic qualifications which were the basis for plant licensing.
f.
Provide a description of the measures taken to protect the safety systems from electrical interference (i.e. Electrostatic Coupling, EMI, Common Mode and Crosstalk) that may be generated by the SPDS.
Response
a) Appendix 1 is a list of the SPDS parameters at TMI-1 along with a brief discussion of the isolation device employed to buffer the instrumentation from the plant computer. Appendix 2 provides available qualification data for the isolation device.
Note that the plant computer is used to generate the SPDS and that certain input signals were part of the original construction. Operating experience over the years in this configuration has proved satisfactory.
Further, certain of the parameters in this category have no automatic safety function but are used in an advisory capacity.
b) At TMI-1120VAC/125VDC are most likely to come into contact with instrumentation circuits.
For the purpose of this analysis it will be assumed that 480VAC can be applied. This is considered unlikely and conservative.
The Appendix 2 data for the Foxboro 2A0-VAI Voltage to Current Converter indicates that 600 VAC was applied between both output leads tied together and ground and that the units remained operable.
Reactor building temperature RTD's are buffered from the computer by a Rochester SC 1374 Temperature Transmitter. The manufacturer indicates on the product specification that 600 VAC/1000 VCC isolation capability exists. -
i Incore th rmocouple inputs to the computer via the backup incore readout system (BIRO) are isolated from the computer by Electroswitch series 24 switch contacts. Appendix 2 indicates 2200 VRMS, 60 HZ was applied between open circuit contacts, closed circuit contacts and non current carrying pa'rts.
-Monolithic integrated circuits are used by the Nuclear Instrumentation System. -The nature of this device makes it current limiting. We know of no test data for this module. However, conversations with the vendor indicate that calculations were performed to verify fault isolation capability.
c).From the available test data from Foxboro for the 2AO-VAI Voltage to Current Converter, the 600 VAC test potential was applied between the output leads and ground. The output terminals were grounded during another phase of the. isolation test. The test is described and the results presented in Appendix 2.
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-The Electroswitch series 24 switch used to isolate backup incore readout system from the computer was tested as described in (b) above.
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The Rochester model SC1374 and SC1302 transmitters used to buffer reactor _ building temperatures and RC pump power from the computer provide the isolation in (b) described above. We currently have no infomation on the actual tests. However, the vendor has agreed to
- supply the test results which should demonstrate acceptable fault isolation capability.
No. test data is available on the Bailey Meter Company isolation amplifiers. However, as indicated above for the Nuclear Instrumentation System, calculations were performed.
d) The pass / fail criteria for_the isolation devices-for which ' test data -is-available are as follows:
Electroswitch Series 24--When 2200 volts rms is applied between contacts there shall be no. arcing, breaking of insulation or damage. Maximum.
i allowable leakage is 100 microamps.
Foxboro 2AO-VAI Isolation--Sufficient isolation to prevent damage propogation.
e) All of the isolation devices used in' connection with'SPDS are located in the control building at TMI-1 which is considered a mild environment.
Therefore in accordance with 10CFR 50.49 no special consideration to qualification need be given.
At TMI-1 seismic requirements are not invoked for the NNI or the reactor
" tripped" logic. Seismic requirements are invoked on the N!/RPS and for-signal conditioning equipment utilizing the 2AO-VAI. These requirements comply with the plant licensing basis.-
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f) At TMI-1 tha SPDS is an extensien of the plant computer system. Most of the input signals have been routed to the computer since the original construction with no deleterious effects due to crosstalk, EMI, electrostatic coupling, Common mode etc. New sigr.cls being routed to the computer are done so in a manner identical to existing signals and no detrimental consequences of this routing are expected.
At TMI-1 instrumentation circuits are routed independently of power cables. Low level analog signals are routed in Twisted Shielded Pairs grounded at one end. As stated earlier the plant computer has not contributed to any electrical interference to the plant instrument systems interfaced with it...
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.Qu2stion'2 - Human Factors Program
~ Provide a description of the display system, its human factors design and
.the methods used and results from a human factor program to ensure that the displayed information can be readily perceived and comprehended so as not to mislead the operator.
Response
The Safety Parameter Display System (SPDS) is an aid to the control room personnel in determining overall plant safety status during power operation and post trip along with identifying abnormal conditions. Since the SPDS provides an oierview of the plant safety status, the primary users have-been identified to be the Shift Supervisor and Shift Technical Advisor. The SPDS allows the user to obtain a minimum set of important parameters at one location. These parameters are organized into five (5) Critical Safety Functions and displayed to allow for easy and unambiguous interpretation of the information.
'The user will interact with the SPDS by means of the Plant Process Computer System. The computer. alarm processor will be used to alert.the user of an abnormal. condition identified by the SPDS logic.
The user will respond to SPDS alarms using the same human communication system and methods as all other process computer alarms.
The SPDS interface will consist of 10 additional points added to the alarm
. database. These 10 points will consist of 5 priority 1 alarms and 5 priority 2 alarms. Thus each critical safety function will have a priority.
I and priority.2 alarm associated with it. The priority 2 alarm is meant to be a warning condition while the priority 1 alarm will alert the user to a
.more severe condition.
Once the user receives an alarm from the plant computer alarm processor he/she should go the specific display for the critical safety function in alarm. -On the display the numerically displayed parameter (s) which are in alarm will be displayed in reverse video yellow for priority 2 alarms and reverse video red for priority 1 alarms. Graphically displayed parameters will be' identified as being in alarm by the parameter plot crossing over an
. alarm line on the graphical. display.
- The SPDS displays will use CRT hardware in the TMI-1 control room. A push button on the CRT console will provide access to the SPDS displays.
If a
- CSF is_ in alarm, the menu will shew the alarming CSF in reverse'. video yellow-for priority 2 and reverse video red for priority 1. alarms. When the user selects a display from the menu, the computer will decide whether the power
. operation or post trip display will be placed on the CRT.
A hard copy of any display should be able to be obtained upon request by the user..If a~CRT is displaying an SPDS display, this display shall not be automatically preempted by another non SPDS display. An SPDS display may be
. preempted by another SPDS display.
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2.
Methods and Results A committee was formed to develop the SPDS displays. Team members who contributed to this design process consist of Human Factors Engineers, STAS, Design Engineers, Computer Applications, and Shift Supervisors. A functional analysis, following the general guidance of NUREG 0700, was performed and used for parameter selection and generation of displays. Only the strict need of the user was considered in adding data to the display.
Preliminary displays were created and provided the foundation for the development of the final displays. NUREG 0700 guidance was followed in developing the preliminary displays. The users needs were evaluated with an initial survey and walkthrough. Final display configuration was generated after multiple walkthroughs had been performed by Human Factors Engineers with designated users.
During the course of this development program, more than one display was presented incorporating the same information.
In each case, the display was consistent with NUREG 0700 guidelines. This allowed for a number of different ideas to be presented and evaluated.
A more detailed survey was conducted using each set of displays.
Results of this survey were evaluated by Human Factors Engineers, Computer Personnel, and Design Engineers; then translated to modify the displays. Wal kthroughs were then scheduled including all team members.
All the control room users will be trained on the philosophy and use of the SPDS. The training will allow the user to utilize,the SPDS in determining whether the plant is responding in a normal or abnormal manner.
It will also allow the user to interpret the adequacy of the actions taken by the operators. The training department will comment on the displays based on these criteria.
All walkthrough coments and survey results were used by the display comittee to finalize the displays. Display criteria were generated and the final displays placed on the CRT for review. A final walkthrough was performed once the displays were coded. This walkthrough used transient data to show the response of SPDS to different situations.
The Human Factors Program is resulting in a consolidation of the number of displays, with a consistent formatting in terms of use of color, labels, and standarization of method of presentation. Principles of the NRC checklist (NUREG-0700, Section 6.7, Process Computers) have been followed, including the principles of grouping, ordering and usability. Structuring and organization of displays is logical and consistent with its intended use. 0
Question 3 - Data Validation Describe the methods used to validate data displayed in the SPDS. Also describe how invalid data is defined to the operator.
Response
The plant Computer System will perform validity limit checking. This checking will identify all points which exceed predetermined limits. When these limits are exceeded, the computer will indicate a quality of Suspect or Bad.
Quality tags are used by the Plant Computer System to indicate the quality of all displayed values. The following tags are listed:
Blank = GOOD (Sensor is reliable and on scan)
? = SUSPECT (Sensor exceeds calibration range)
S = SUBSTITUTED (Automatically or manually substituted value)
B = BAD (Sensor off scan, sensor failed, sensor exceeds the transducer rangelimit)
Invalid data (SUSPECT and BAD qualities) are displayed to the user as periods (....)
in the value field.
Invalid data (SUSPECT and BAD qualities) on X, Y plots will be indicated by plotting the last known value and identifying the invalid data on the display.
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Question 4 - Parameter Selection i
The staff has reviewed the licensee's Safety Analysis Report on the TMI-1 SPDS. While we find that the parameters selected do comprise a generally comprehensive-list, we note that the following parameters are not proposed for l
the TMI-1 SPDS:
1.
Source and Intermediate Range Neutron Flux Monitors 2.
DHR Flow 3.
Steam Generator Level e
4.
Containment Isolation 5.
Steamline Radiation l'
Neutron flux is a fundamental parameter for monitoring the status of plant i
reactivity control, and should be monitored for all power ranges.
For core l
conditions in the power range, the TMI-1 SPDS provides monitoring of both power level and the startup rate. For conditions.which 'nclude coverage below the-power range-(intermediate range and source range), the TMI-1 SPDS does not monitor power level, but does provide startup rate instead. The licensee states that startup rate allows for easier and quicker interpretation of the core conditions than does power level. We find this acceptable, but request i
confirmatory discussions to demonstrate that startup rate (in lieu of power level) provides adequate indication for use under all conditions and 1
scenarios, including steamline breal., boron dilution, etc.
During decay heat removal and ECCS modes of cooling, when steam generators are not being used, DHR flow is a key indication to monitor the viability of the
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heat removal system.
I Steam generator level is an indication of the availability and proper control of secondary system heat sink for the " Heat Removal" critical safety function. The licensee states that steam generator level is not included because it does not provide useful unambiguous information concerning heat removal. However, we note that frequent reference to steam generator level indication in TMI-1 guidelines (TDR-517, Section 2.3, 2.5 and 2.10) would i
indicate that steam generatar level does provide essential operator 4
infomation and should be included in the TMI-1 SPDS, or additional justification for its omission should be submitted.
Containment isolation is an important assessment of " Containment Conditions."
In particular, a determination that known process pathways through containment have been secured provides significant additional assurance of containment integrity.
Steamline radiation, in conjunction with containment radiation and reactor stack radiation gives a rapid assessment of radiation status for the most likely radioactive release paths to accomplish the~" Radioactivity Control" safety function. Also the TMI-1 SPDS monitors radiation in effluent from the steam generators when steaming either to the condenser or through the
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atmospheric dump values. But, with a steam generator, these detectors may not indicate radiation status within the steam generator. However, the licensee should consider how radiation in the secondary system (steam generators and L
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steamlinas)' will be monitored when the steam generators and/or their steamlines are. isolated, and verify the timeliness of the monitoring in the validation program for TMI-1 AT0G and the SPDS.
In summary, the above parameters do, for given scenarios, provide unique inputs to determinations of status for their respective critical safety functions, and they have not been discussed by the licensee as being satisfied j
by other parameters in the proposed TMI-1 SPDS list. We recommend that the i
licensee address these parameters and their functions by:
(1) adding these parameters to the.TMI-1 SPDS, (2) providing alternate added parameters along L
with justifications that these alternates accomplish the same safety functions
.for all scenarios, or (3) provide justification that parameters currently on the TMI-1 SPDS do in fact accomplish the same safety functions for all scenarios.
Response
1.
Source and Intermediate Range Neutron Flux Monitors The SPDS design can identify a re-criticality condition by using the startup rate parameters, which are calculated from the source and intermediate range neutron flux monitor data.
The startup rate is a less ambiguous indication of re-criticality since its indication of re-criticality-(stable startup rate t 0) is not dependent on time after trip. The actual flux indication (source and i
intennediate range) can be misleading for the first 30 minutes after a trip. This ambiguity is caused by the post trip response of this t
indication.-If the operator looks at.the flux level after a trip, he must h
know the time since trip and the detector response, or he must monitor, the flux level to see if it is increasing or decreasing in order to t-determine if a re-criticality has occurred. During the walkthroughs, a
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portion of the TMI-1 control room personnel were questioned concerning how they use the flux level indication. Their response indicated that they monitor the flux level over a short period of time to see if it is
. increasing or decreasing.
Since startup rate is a direct indication of flux level changes versus time, it is the most appropriate indicator to provide on the SPDS.- For either a normal post trip or re-criticality condition, the actual flux level value is not an important parameter unless the power is high enough for the power range indication. When the flux level is below the power range, it does not present a heat removal concern and thus, the exact
-level is not important. Once the flux level has reached the power range indication, it will'become a heat removal concern. Thus, power range indication is provided for the post trip SPDS.
The source range startup rate, intermediate range startup rate and power range level provide the best indication of core reactivity during both power operation and post trip modes. The addition of source range and intermediate range flux level will not provide indication that will enhance the users ability to assess the plant safety status.
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2.
DHR Flow 1
The TMI-1 SPDS was designed for the following modes of operation:
t'
' Power Operation Hot Standby Hot Shutdown Post Trip. Initial Response i
On Page 6 of GPUN Topical Report 018, it is specified that the SPDS has not been designed for the cold shutdown and refueling modes. Since DHR flow cannot occur during the modes for which the TMI-1 SPDS has been designed, DHR flow was not identified as an SPDS parameter.
3.
Steam Generator Level Steam generator level can be an ambiguous indicator of RCS heat removal under various conditions such as EFW cooling and interruption of natural circulation. Other indicators provided within the SPDS provide more reliable indication o# RCS heat removal. As a result, steam generator level was not identified as a primary
- SPDS parameter.-
It has been identified as a secondary
- parameter within the primary side heat removal 4
critical safety function As a secondary parameter, steam generator
. startup range level is provided.when the plant is on low level limits and n
operating range level is provided during natural circulation conditions.
- NOTE: Primary and secondary functions are defined on Page 10 of NUREG-0835 " Human Factors Acceptance Criteria for the Safety Parameter Display System".
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' Containment Isolation 4.-
Containment isolation.is only a concern when the containment atmosphere has adverse radiation levels. The TMI-1 SPDS will alert the user to this condition and to the potential for this condition (e.g. RCS integrity and conditions which could lead to fuel failure). The TMI-1 SPDS will'also alert the users that a primary containment isolation setpoint has been F
- reached. With this information,'the'SPDS user can use_his safety. grade control room indication and actual onsite and offsite radiation readings to properly assess the containment isolation status. By monitoring the radiation control and RCS integrity CSF, the TMI-1 SPDS will indirectly alert the user to breaches in the containment.
5.
Steam Line Radiation
- TMI-1 Abnormal Transient Procedure 1210-5 "0TSG Tube Leak / Rupture"
- requires that both steam generators be steamed during a tube leak unless certain criteria are achieved. The following are the steam generator isolation criteria.
1.-
BWST level is less than 21 feet, or
.0ffsite dose projections approach 2.
=
50 mr/hr whole body or 250 mr/hr thyroid E V
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Once tha steam g:nsrators are isolated, the plant will be on the HPI cooling mode.
For all but the extremely severe tube leak accidents, the TMI-1 steam generators will be steamed.
If the condenser is available, then the tube leak is identified using two radiation monitors in the condenser off-gas system.
If the condenser is unavailable, the tube leak may be identified using the steam relief monitors. The steam relief monitors-are not sensitive to small tube leaks unless the failed fuel ratio is greater than one (1) percent. Once the steam generators have been isolated, the time to return the generators back into service should be of sufficient length to allow for sampling. This will provide the means for determining the radiation levels for an isolated steam generator.
The present TMI-1 SPDS parameters are adequate to meet the objectives of the TMI-1 SPDS and NUREG-0737 Supplement 1.
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Question 5 - Desk Audits The licensee has described the method of deriving and validating the TMI-SPDS parameter selections.. The SPDS parameters and TMI-1 ATOG were applied to the following. events (desk audits, as listed by licensee) to validate the t
4 parameter selections:
1.
Excessive Main Feedwater 2.
Loss of Main Feedwater 3.-
Steam Generator Tube Leak / Rupture 4.
5.
Small Steam Leak 6.-
. Loss of Coolant Accidents 7..
TMI-2 Event 8.
TMI-2 Loss of One Main Feedwater Pump 9.
'TMI-2 Turbine Trip Test (No Reactor Trip)
From the results of this program the licensee concluded.that TMI-1 critical safety functions and their associated parameters represent a complete list of 3
parameters required to adequately and correctly display the safety status of TMI-I to the user.
5 It is not clear that the above events included certain scenarios of concern such as a. steam generator tube rupture with simultaneous loss of site power, large steamline breaks, and boron dilution events. These events could place demands on the monitoring capability of the SPD5 to assess radiation t-conditions in an isolated steam generator and to essess post-accident neutron i
flux behavior of the reactor.- The licensee should review his desk audit to assure that he had considered these concerns-and address them in the validation program under development.
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Based on discussions with the licensee (telecon F. Orr, NRC) and H. Crawford (GPU) on July 19, 1984, it is our understanding that the licensee ~is l
- developing a program to further. validate the useability of SPDS by applying a representative set of events in control room walk-thrcoghs. The control room
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demonstration will be conducted using the as-installed SPDS and the plant f-process computer with input developed from transient analyses. During this
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program, the parameter selection should be reassessed addressing our comments j
in previous paragraphs.- We conclude that the TMI-1 program to validate the l-SPDS parameter selection can be acceptable by addressing our concerns. stated a bove.-
Response
i 1.
The steam generator tube leak / rupture scenario which was evaluated for the SPDS document was from the ATOG development program. This scenario considered transients with offsite power and without offsite power.
The important SPDS monitoring considerations associated with a large
' steam line break scenario were reviewed and found to be accounted for by the following events:.
EVENT MONITORING REQUIREMENTS Excessive Main Feedwater Overcooling Loss of Coolant Accident Degraded Containment Environment Small Steam Line Break Event Diagnosis While the SPDS desk audit did not specifically cover a large steam line break event, the events that were considered encompass the SPDS monitoring requirements for this event.
The SPDS covers all boron dilution type accidents by providing power range indication and source and intermediate range startup rate during post trip. These indications will alert the SPDS user to re-criticality caused by a boron dilution accident.
2.
The referenced telecon was a discussion of testing methods which were under consideration by GPUN. At that time, a decision had not been made concerning the testing methods to be used for the formal validation of the TMI-1 SPDS. A validation test plan will be issued as part of the verification and validation program which is in progress. This test plan will describe the minimum testing required to complete the formal validation testing of the TMI-1 SPDS.
(Additional testing may be performed outside the formal V&V process to further review the adequacy of the TMI-1 SPDS.)
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Question 6 - Unreviewed Safety Questions Provide conclusions regarding unreviewed safety questions and changes to Technical Specifications.
Response
The development of the SPDS has not uncovered any unreviewed safety questions. Also, there have not been any changes made to the Technical Specification or any pending due to the incorporation of the SPDS.
Appendix 1 SPDS Parameters and Isolation Devices Reactor Trip--The Reactor Trip signal is input to the plant computer for SPDS use via a dry contact provided from the CRD system.
Isolation between channel trip signals at circuit breakers and auxiliary contacts provided to the reactor " tripped" logic is via mechanical linkage, providing inherent isolation.
Core Power (Heat Balance)--Core Power is calculated from the following plant parameters:
Main Feedwater Flow RCS Total Flow Start-Up Feedwater Flow RC Pressure Feedwater Temperature Hot Leg Temperature Cold Leg Temperature RC Pump Status These parameters are addressed below.
Power Range Power--Power Range signals are input to the computer through an 4
isolation amplifier which is part of Bailey 880 System Sum /Diff, amplifier part number 6625202A. This amplifier prevents loading of the previous stages and provides an isolated output. This buffer has proven reliable in operation.
Signals for the following parameters also are input to the computer through an isolation amplifier which is part of the Bailey 880 System.
Power Range Imbalance RCS Total Flow Source Range Start Up Rate Intermediate Range Start Up Rate RCS Wide Range Pressure--Existing computer input is from original plant ES.
This is provided through an isolation amplifier identical to that of the RPS referenced above. This is a IE type parameter and the computer input will be transfered to the IE type display loop (per TMI-1 Regulatory Guide 1.97 comparision, GPUN Letter 5211-84-2252 dated October 1,1984) isolated with Foxboro 2AO-VAI voltage to current converters described in Appendix 2.
Steam Generator Pressure--Existing computer input is from original plant NNI. This input is provided through transfer relay contacts which provide an incidental isolation function. The NNI is not a IE type system. This parameter is IE type and the computer input will be tranferred to the IE type display loop (per TMI-1 Regulatory Guide 1.97 comparision) isolated with Foxboro 2A0-VAI voltage to current converters described in Appendix 2.
Signals from the following parameters will be transferred from the NNI to safety related display loops, 2AO-VAI isolated:
RCS Wide Range Cold Leg Temperature RCS Wide Range Hot Leg Temperature Al-1
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- Reactor Building Prassure--Input to the computer is isolated with Foxboro 2AO-VAI Voltage to current converter.
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- Signals for the following parameters also are isolated with Foxboro 2AO-VA1 Voltage to current converters:
RCS Saturation Temperature Margin Reactor Building Sump Level Reactor Building Flood Level Reactor Vessel Water Level Hot Leg Water Level Pressurizer Level Reactor Building H.ydrogen Monitor Core Exit Thermocouple Temperature--36 of the Core Exit Thermocouples are wired directly to the computer. The remaining 16 which are utilized by the redundant read-out device can be switched directly to the computer via a model 24 Electroswitch described in Appendix 2.
f.
-Reactor Building Temperatures--Four Reactor Building RTD's are wired directly to Rochester Instruments resistance to Current Converter Model SC 1374. The converter output of 0 to 20 ma de is routed through a Sohm, 1/2 watt, 0.1% tolerance wirewound resistor to provide a 0 to 100 my input to the computer. These RTDS are also displayed on a recorder in the Main Control Room.
Isolation is provided (600 VAC/1000 VDC) between input / power.
Heat-Up/Cooldown Rate--This is derived from RCS Wide Range Hot Leg Temperature by the computer.
RCS Letdown Flow--Input to the computer is from original plant NNI. The NNI i
is not a safety system and this parameter is not IE type (per TMI-1 Regulatory Guide 1.97 comparision). Griginal plant NNI non-isolated j
interface to the computer is satisfactory.
Signals for the following NNI parameters are not IE type and are not isolated:
l RCS Make-Up Flow I
Main Feedwater Start-Up Feedwater Flow (Heat Balance)
Feedwater Temperature (Heat Balance) i High Pressure Injection Flow--Existing computer input is from original plant NNI. The NNI is not a IE type system. This input will be provided through an isolation amplifer identical to that of the RPS referenced above. This is a IE type parameter and the computer input will be transferred to the IE type display loop per TMI-1 Regulatory Guide 1.97 comparison) isolated with Foxboro 2AO-VAI voltage to current converters described in Appendix 2.
i' Reactor Coolant Pump Status--Input to the computer is from breaker auxiliary contacts completely separate from the safety grade pump power monitors provided for the RPS.
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Void Fraction--Input to th2 computer is via Rochester Instrument System Model SC-1302 isolation amplifiers in the pump power monitoring system.
These isolation ampliiffers were purchased as qualified devices. GPUN is currently attempting to secure a copy of the Qualification Report.
The following Radiation Monitoring signals are input to the plant computer for use by SPDS:
a) RM-L1 Lo (Letdown) b) RM-L2 (A Loop Decay Heat) c) RM-L3 (B Loop Decay Heat) d) RM-L9 (Intermediate Closed) e) RM-A2(RBAtmosphere) f) RM-A4 Gas (FHBAtmosphere) g) RM-A5(CondenserExhaust) h) RM-A6 Gas ( Auxiliary Bldg. Atmosphere)
- 1) RM-A8 Gas (Plant Stack) j) RM-A9 Gas (RBStack) k) RM-G8 (RB Dome)
- 1) RM-G25(CondenserExhaust) m) RM-G26 (A Loop Steam Relief) n) RM-G27 (B Loop Steam Relief)
The Victoreen 842 series ratemeter provides an auxiliary 0 to 50 MVDC output for use by the plant computer. This signal is derived from the ratemeter auxiliary circuit. This circuit is designed to be tolerant of a 120V fault applied at the computer output terminal such that other portions of the ratemeter circuitry will be unaffected.
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Appendix 2 Qualification Data for Isolation Devices
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Foxboro The 2A0-V21 is a IE qualified isolation device.
This is documented in test report Q0AAB44 Rev. A, enclosed.
Isoltion testing is described in A0AAA20-1, referenced in QIAAB44. The appropriate page (72) of Q0AAA20-1, enclosed, shows application of faults an output.
Please note that this is part of the seismic test.
Electroswitch Series The Series 24 is a IE Qualified 24 Switch isolation device.
Isolation testing is decribed in the enclosed publication ESC-STD-1000.
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