ML20084H873
| ML20084H873 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 04/24/1984 |
| From: | Carey J DUQUESNE LIGHT CO. |
| To: | Varga S Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20084H876 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.F.2, TASK-TM GL-82-28, TAC-45114, NUDOCS 8405080239 | |
| Download: ML20084H873 (26) | |
Text
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D M
V Telephone (412) 393-S000 Nuclear Division P.O. Box 4 shippingport, PA 15077-0004 April 24, 1984 Director of Nuclear Peactor Pegulation United States Nuclear Pegulatory Cmmission Attn:
Mr. Steven A. Varga, Chief Operating Peactors Branch Ib.1 Division of Licensing Washington, DC 20555
Reference:
Beaver Valley Power Station, Unit No. 1 Docket No. 50-334, License No. DPR-66 NUREG-0737, Itcm II.F.2; ICC Instrumentation System (Generic Ictter 82-28)
Gentlemen:
Attached is our response to your request for additional information dated December 16, 1983, concerning Inadequate Core Cooling (ICC) Instrumen-tation for the Beaver Valley Pcwer Station, Unit No. 1.
This information describes the components of the proposed final ICC Instrumentation System.
Our previous enbrnittals described the existing core exit thermocouple (CEIC) system with its associated displays, the existing subcooling margin monitor (SbE) and the Westinghouse designed RVLIS.
Based on an evaluation of the merits of an upgraded system, we have determined that an upgraded system may have advantages over the existing instrumntation in mitigation of events beyond the plant design bases. A description of a proposed upgrade to the ICC Instrumntation System was presented to you and members of the NPC Staff and an apparent agreement reached during a meeting on March 1,1984.
The final system will consist of RVLIS, a new SbM, and an upgraded CEIC system all integrated into an ICC Monitor. This system as presented at that meeting has been evaluated against your specific requests for additional inform tion.
The proposed upgrade, as presented by menbers of my staff, represents an upgrade of existing instrumentation to achieve compliance with NURDG-0737 item II.F.2 guidelines with som exceptions as are identified on the following pages. The elemnts of this upgrade include:
use of existing CEICs in the reactor vessel use of environmentally qualified cables inside containment beginning with the connectors on the reactor vessel head through to the containment side penetration terrunation cabinets use of existing cable raceways use of environnentally qualified reference junction boxes use of redundant microprocessor-based ICC nonitoring cabinets which combine the RVLIS, ShN and the CEIC backup display 0h j[
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f Beaver Valley Power Station, Unit 16.1 Docket No. 50-334, License 16. DPR-66 IUREG-0737, It s II.F.2; ICC Instrumentation Page 2 The cost estimate for this proposed upgrade is approximately $3 million.
'Ihe cost estimate to upgrade our ICC Instrumentation System to the full scope of IUREG-0737 Its II.F.2 is approximately 4.6 million. A further breakdown of the costs associated with a full upgrade is provided in the attachment in direct response to your question on cost.
As previously stated, this upgrade will achieve ccnpliance with the IUREG-0737 guidelines with exceptions. The following is a brief sunmary of the degree of cmpliance to the specific criteria which formed the basis for the IGC cost / benefit study which were contained in SECY-82-407.
Environmental Qualification: New cmponents inside contain-ment will be environmentally qualified.
Seismic Design: The IOC Instrumentation Systs will utilize existing raceways which were originally installed in accor-dance with the seismic Class 1 criteria in effect at the time the plant was originally constructed.
New components installed will be installed in accordance with current seismic installation criteria.
Single Failure: There will be two independent trains capable of monitoring ICC conditions thus providing redundancy. However, there are areas where strict ccnpliance with Pegulatory Guide 1.75 cannot be achieved. The following are exanples:
1.
All thermocouple cables enter the control rom through a ccnmon sleeve.
2.
The required physical separation cannot be met between raceways.
3.
The thermocouple cables do not enter an isolation device prior to branching to additional equignent.
4.
Thermocouple cables are routed in raceways with other non-1E instrument grade cabling.
Class lE Power Source: The ICC Instrumentation System will be powered frm two trains of Class 1E power source.
The schedule for achieving this upgrade is dependent upon cmpletion of all engineering activities and procurement of necessary omponents.
A minimum of 18 months is required for the delivery of the major ccrnponents after an order is placed.
Our best estimate schedule would be to cmplete this upgrade during the sixth refueling outage tentatively set for July 1986.
However, unforseen delays caused by equipnent delivery problems and engineering prob 1ms could delay full implmentation until the seventh refueling outage, currently scheduled to occur in late 1987.
Every effort will be made to expedite cmpletion on the best estimate schedule.
{
l Beaver Valley Power Gtation, Unit No.1 Docket No. 50-334, License No. DPR-66 NUREG-0737, Its II.F.2; IOC Instrumentation Page 3 In the interim, the following will be available to the operator to aid in diagnosing ICC conditions:
1.
W e RVLIS will be c m pleted and placed in service following the upcoming fourth refueling outage scheduled to begin October 1984.
2.
We subcooling margin nonitor will remain inservice, as is, until replaced with the proposed upgrade.
3.
We CE7ICs can presently be read to 1650 F on the subcooling margin monitor display (8 T/Cs) and all thermocouples can be read to 1650'F on the Plant Variable Cmputer System.
Following the fourth refueling outage,these CEICs will also be available on the SPDS (debugging of the SPDS will occur for a short time following startup fran that outage).
4.
%e existing reactor coolant system wide-range pressure trans-mitters and hot leg RIDS can be utilized to determine degrees subcooling.
5.
The operators have been instructed to utilize all available instru-ments when assessing plant conditions and not rely on any one instnrent so as to preclude being mislead by any one indication.
Additional instruments that might be used to detect ICC conditions include containment parameters and reactor coolant punp anmeter indications which are known to respond prior to the 1200'F CE'IC indication which involves the highest level function restoration procedure for core cooling.
Finally, it should be noted that a sustained / irreversible core uncovery will not occur during any IOCA condition as long as at least one train of safeguards equipent is operating.
I would also like to take this opportunity to document our position and understanding of further NRC activity associated with the ICC Instrumentation System as discussed at the March 1 mceting with members of the NRC staff.
We consider the proposed upgrade of the ICC Instrumentation System as being acceptable for satisfying the NRC design considerations identified in Regulatory Guide 1.97.
We do not intend to perform further modifications to the final ICC Instrumentation System as a result of Regulatory Guide 1.97.
When the sunmary report is subnitted in November of 1985 docunenting our review of Regulatory Guide 1.97, the design information and accepbability of the design for ICC instrumentation will be consistent with infomation contained in this sutrtittal.
During the March 1,1984 meeting, members of my staff questioned the NBC's reaction to the work performed for the NRC by BGG of Idaho in their review of CE'ICs in PWRs.
(
References:
NUREG/CR-3386, dated Novmber 1983, and report number DGG-ED-6361, dated October 1983). We expressed our concern
Beaver Valley Power Station, Unit No. 1 Docket No. 50-334, License No. DPR-66 NUREG-0737, Item II.F.2; ICC Instrumentation Page 4 that the NBC may reconsider their position on ICC Instrumentation require-ments based on these reports and undermine our efforts to develop and install the proposed final ICC Instrumentation System. We were informed by the CPB that these subject reports "do not contain new information on the expected operation of CEICs during accident and post-accident situations, therefore, there is not any anticipated change in the NRCs position based on these reports".
We believe this information is consistent with the proposed design described at the meeting with your staff on March 1, 1984 and our engineering will proceed based on the acceptability of that design.
Attached are the detailed responses to your request for additional information.
If you have any questions regarding this subnittal, please contact me or mmbers of my staff.
Very
- yours, hWA
a.
y Vice President, Nuclear Attachments cc:
(w/o drawings)
Mr. W. M. Troskoski, Pesident Inspector U. S. Nuclear Regulatory Ca mission Beaver Valley Power Station Shippingport, PA 15077 U. S. Nuclear Regulatory Cmmission c/o Dmument Management Branch Washington, DC 20555 Director, Safety Evaluation and Control Virginia Electric and Power Company P.O. Box 26666 One James River Plaza Richnend, VA 23261 (with drawings)
Mr. Peter Tam, Project Manager U. S. Nuclear Pegulatory Camission Phillips Building Washington, DC 20555
- Mail Stop 438 -
I
ATT1GMENT letter dated April 24, 1984 Desponse to NBC Request for Additional Information dated December 16, 1983 IOC Instrumentation System The followirg provides the requested information describing the proposed design of the Beaver Valley Unit No. 1 ICC Instrumentation System.
'Ihis description is consistent with the proposed upgrade presented to the hPC at a treeting on March 1,1984.
1.
Provide a description of any deviations frm the generic Westinghouse system described in previous and referenced subnittals, and a descrip-tion of the plant-specific features of the design including displays and their locations.
Pesponse There are no deviations frm the generic Westinghouse RVLIS System.
The Westinghouse design has added two new wide-range RCS pressure transmitters outside containment, tapping into the capillary tubing, in order to keep the indicating accuracy of RVLIS within the generic design specifications. The installation will be completed and RVLIS made operatim al fnllowing the fourth refueling outage, scheduled to begin in October 1984.
Plant-specific features of the design are presently limited to the location of various cmponents. At this time, no plant specific cmponents have been added to the RVLIS, nor have any cmponents been deleted. 'Ihe location of the RVLIS displays were provided in our April 21, 1983 subnittal, Figure 4.
The location is designated as vertical board B.
The proposed upgrade of the IOC Instrumentation System will result in the incorporation of scxte plant-specific features of the RVLIS design. The proposed upgrade will replace the present RVLIS microprocessor cabinet with a new microprocessor cabinet integrating RVLIS with a therroccuple/ core cooling monitor (T/CCN). This upgrade has been proposed by Westinghouse to result in an integrated, redundant, qualified RVLIS/T/CCM in one system. 'Ibgether these will make up our IOC &nitor.
Included in this upgrade are redundant cualified plasma displays to indicate either reactor vessel level, core exit tmperature or margin to saturation.
The new displays will be located in the control roan in a manner consistent with good human engineering principles. 'Ihe above mentioned existing RVLIS displays will be renoved when the upgrade is implemented.
2.
Provide a description of tests planned and results of tests ccrpleted for evaluation, qualification and calibration of RVLIS.
Provide a schedule for tests which have not been cmpleted.
l
Attachment ICC Instrumentation System Page 2
Response
Tests Cortpleted EQ tests,10 reports received high volume sensors hydraulic isolation RIDS Marathon Terminal Blocks Tests Cmpleted EQ tests, EQ reports not yet received remote displays dp transmitters reworked RVLIS cabinet three pen recorder W.R. pressure transmitters Tests Planned Systesn Pill: this includes leak testing the capillary tubing, hydraulic isolator alarm check, vacuum filling the system with deareated demineralized water and calibration of the dp and W.R. pressure trans-mitters.
This will be cmpleted at the end of the fourth refueling outage.
Scaling Test: performed during plant startup frm the fourth refueling outage and will consist of inputing and checking:
RID engineering unit conversion and setpoints high volum sensor engineering unit conversion and setpoints RCS WR pressure input and conversion measuring RCP performance and making corrections to PCP performance curve inpulse line vertical length compensation enter heights of high-volume sensors and RIDS.
3.
Provide a description of the locations of the dp transmitters, the locations of temperature measurements for cmpensation of the vertical runs of the sense lines, and an estimate of the uncertainty in level measuremnt of the system including the contribution of each cmponent and the method of cambining uncertainties.
Response
dp transmitter location: safeguards building (elev. 722-6) in the Stairwell / ventilation area east of the cable vault area (see attached FSAR figure 5.2-29) l
~
Attachment IT Instrurentation System Page 3 3.
Response, (Continued)
R'ID location:
inside containment at each vertical drop of capillary tubing which is greater than 3 inches, in accordance with Westinghouse installation criteria.
Estimate of uncertainty:
this information was previously subnitted in response to the NBC RAI of August 5, 1981, which included 25 questions on RVLIS. Our subnittal was dated Septmber 2,1981 and provided this specific information in response to question number 6.
This same information is provided with this subnittal as Appendix A.
4.
Clarify the DIf position with respect to II.F.2 Attachment 1 requirments and provide a schedule for upgrade.
Areas of concern include:
a.
The primary display for the CET system is the existing plant ccmputer which provides indication only over a range of 0-700F, instead of 200F to 1800F as specified by II.F.2, Attachment 1.
The plant cmputer display includes 51 CEr's with adequate quadrant coverage.
Response
At the time of our response, which identified the existing plant comput-er as the primary display, there did not exist another system capable of monitoring all CEICs. However, it was stated that there were two other systms being designed and installed at BVPS which would provide in-creased indicating capability for monitoring the CEICs.
Since that time, we have placed in cperation the Plant Variable Cmputer System (PVS) which is capable of indicating all thermocouple tmperatures over the range of 0-1650 F on CRTs in various locations at the site including the shift supervisors office. Scme CRT teminals have yet to be in-stalled, however they are scheduled for installation during our fourth refueling outage. hhen a CRT is installed at the operator's console, the operator will be able to call for and rapidly receive a graphics display of all CEICs.
The range of 0-1650*F is considered acceptable since the approach to, existence of, and recovery from ICC occurs at tmperatures well below 1650*F, and there is the potential for large errors above this point.
(Ref: EGG-ED-6361 dated Octcber 1985).
The other system being designed and installed is the Safety Parameter Display System (SPDS). Upon ccupletion of its installation and operator training, the SPDS will be considered the primary display for the CEIC system.
Our current schedule includes cmpleting all ternunations in the plant which serve as inputs to the SPDS during the upcming fourth refueling outage.
After terminating all inputs, and ccupleting loop calibration, a cmplete check-out of the SPDS will be accmplished.
Between this outage and the fi:th refueling outage, our plant operating
I Attachmnt IOC Instrumentation Systent Page 4 4.
Response, (Continued) staff will beccme familiar with this system, receive training on the systs, and learn how to nest effectively make use of this operating aid for responding to emrgency conditions.
'Ihe SPDS will display core exit temperatures under several different formats and will indicate temperature over the range of 200 F to 2500 F.
b.
Operator-display human factor design to provide rapid access to displays and alarm capability consistent with operator procedure requirements are not addressed in the applicant response.
Pesponse The Westinghouse design of the SPDS display included the use of human engineering principles for ease of use. There will be two SPDS displays installed in the control rocrn, one on the operators console and a second above the vertical board facing the operators with its keyboard in-stalled on the operators console These locations will provide the operators rapid access to the displays and keyboards within the area of the control rocxn where the operator will be monitoring other critical plant paraneters.
The SPDS has been designed by the Westinghouse Electric Corporation and has received considerable review by the NPC. A favorable SER was issued by the NRC to Westinghouse on February 2,1984 which documents ac-ceptability of their SPDS design. This system is capable of providing the operator six different types of displays for nonitoring core exit temperature. The first level display is the Westinghouse Iconic display with the maximum core exit thenroccuple being shown graphically. Also graphically shcwn on this display is the core exit temperature high limit. The high limit setpoint equals PCS saturation temperature and is a calculated value based on PCS wide-range pressure.
An alpha-numeric indication of the maximum core exit thermocouple and degrees subcooling is also provided beside the Iconic display.
The second level display includes trending information of the core exit tmperature over a range of 100 to 700 F.
h third level display is a spatially-oriented core map of all CmCs with the PCS wide-range pressure and the average core exit temperature also being indicated. The fourth level display is an alpha-numeric display of all CmCs divided into core quadrants. The operator can also obtain a hard copy of all cmc terperatures as displayed on the SPDS.
Core exit terperatures displayed on the SPDS are in an alpha-numeric format over the range of 200*F to 2500*F. When an alarm condition is reached the SPDS display provides a visual alarm indication to the operator by changing the back-ground field around the pararoter to red.
Alarm setpoints are provided for both the CmCs and degrees subcooling.
The operator can select any display frcm a menu / map within seconds thus satisfying rapid access to displays and alarms for the Cmc primary display.
Attachment ICC Instrumentation System Page 5 4.
Response, (Continued)
'Ihe proposed upgraded system will serve as the backup CE7IC display and will also cambine other indicators of ICC conditions such as subcooling and reactor vessel level and as such will be our ICC monitor.
'Ihis system consists of redundant plasma displays which provide digital readouts of ICC parameters and the presentation format is human engi-neered for ease of use.
In addition, there is a one-line digital display furnished on the front of each microprocessor train of the new ICC monitor.
'Ihe operator controls the display to bring up various pages of data which provides all pertinent information of the thermocouple core cooling functions, such as:
temperature of the tv>-hottest thermocouples per quadrant individual thermocouple and location for each quadrant, four quadrants total reference junction box RID temperatures T-hot leg RID tenperature RCS pressure margins to saturation saturation temperature reactor vessel level The ICC monitor (proposed upgrade) is also provided with two alarm setpoints, the first to indicate the developnent of off-normal con-ditions, the second to indicate the approach to saturation temperature.
Contact closures are provided for interfacing to the existing annunciator system. Alarm setpoints will be set consistent with proce-dure requirements, c.
The backup display is a Honeywell Precision Tenperature Indicator which apparently does not have selective reading capability of a minimum of 16 thermocouples, 4 from each quadrant,nor the capabil-ity to selectively read 16 thermocouples in 6 minutes as required.
'Iwo new systems are being designed and installed which may provide additional backup display. These are the Plant Variable Canputer System (PVS) and the Safety Parameter Display System (SPDS). 'Ihe PVS will have an alpha-numeric map display of an unspecified number of CE r's but only up to 1650 F instead of the required 1800F. The SPDS will measure and display up to 2500F for an unspecified number of CE7f's.
Signal and power isolation for the new systems is not mentioned in the submittals.
Response
'Ihe Honeywell Precision Tenperature Indicator does have selective reading capability of all 51 thermocouples and the ability does exist to read 16 within 6 minutes.
Attachment ICC Instrumentation System Page 6 Besponse, (Continued)
The SPDS will be our primary display for the final ICC Instrumentation System and has been previously discussed. The PVC, which is in opera-tion, will also measure and display all CECs and will provide the operators an indication of core exit terrperatures.
A new back-up system has been previously discussed and was proposed as the upgraded IOC Instrumentation system during the March 1, 1984 meet-ing.
'Ihe display capabilities are discussed in response to question 4b. The tertperature readout of CECs for the backup display as provided by the equipnent supplier is to about 2200*F.
Refer to Question 4e for information on the power supplies for the primary Cmc display system and the proposed ICC Monitor.
Signal isolation is discussed in response to Question 41.
d.
Additional operator training will be based on new EOP's which are under develognent following Westinghouse ERG's.
Schedule for cmpletion is not provided.
Response
The new EOPs will be completed in accordance with the schedule subnitted in our supplemental response to Supplement 1 to NUREG-0737 dated July 25, 1983, ht subnittal states that full EOP inplmentation will be accortplished during the fifth refueling outage tentatively scheduled for July of 1986.
c.
The CEr system and the plant cmputer are powered from a diesel backed emergency bus (Class lE not indicated). 'Ihe backup display power (required to be electrically independent) is not identified.
Separation and isolation are not provided as required by II.F.2,.
Pesponse The proposed upgrade of the ICC Instrumentation Systm will consist of a highly reliable power supply to the IT Monitor (back-up CEIC display) and to the SPDS (primary CEC display). h SPDS will be powered frcm the ERP substation which was constructed to support the power require-ments of the new Dnergency Response Facility (ERF). The ERF substation is capable of being powered fran two separate 4160V power sources and-will be backed up by a new separate diesel generator.
(The installation of this diesel generator is scheduled to be carpleted during the upcm-ing fourth refueling outage).
The back-up display, the proposed ICC Monitor will be powered frm a power source separate and independent frm the primary CRC display systs. The power source for the back-up systs will be frm the station's diesel backed mergency power system which was originally constructed as lE.
Attachment ICC Instrumentation Systs Page 7 f.
The requirment for 99% availability to display a minimum of four thertroccuples per quadrant is not addressed. The required Techni-cal Specification on display availability is also not addressed.
Response
The ICC & nitor will be capable of displaying all thermocouple readings upon request.
This systcra will be powered frcnn the stations diesel backed mergency power system which was originally designed to achieve a 99% reliability factor.
When installation of these systes is empleted, plant Technical Specifications will be reviewed against the reccomended Technical Specifications provided in Generic Intter 83-37, NUREG-0737 Technical Specifications, and the appropriate changes will be suhnitted at that time.
'Ihe existing core cooling monitor has eight CEICs as inputs and is presently covered by Technical Specifications.
g.
A human factors analysis is to be performed later - schedule not provided.
Fesponse A human factors analysis of control rom equipnent as it interfaces with the opnrator when using omergency operating procedures will be ccrnpleted in accordance with our subnittals in response to Supplement 1 to NUREG-0737. The Control Rocan Design Review is currently in progress and the sumnary report on this activity is scheduled for subnittal to the NRC by November 1985.
The ICC tenitor will not be installed in time to be evaluated during our CRDR, however, procedures for considering human engineering principles when installing equignent in the control rom will be in place. There-fore, when this eqttipnent is installed, it will be in accordance with acceptable human engineering principles, h.
Integration of the use into the EDPs is to be done later - schedule not provided.
Pesponse The use of ICC instrumentation has been factored into the Westinghouse generic Dnergency Pesponse Guidelines (EPGs) in accordance with NUREG-0737 Its 1.C.1.2.
New plant E0Ps are currently being developed following these EPGs and will follow the mitigation strategy presented in the ERGS. The schedule for implementing the new E0Ps was provided in response to question 4d.
i.
The system does not meet the single failure criterion.
Attachment ICC Instrumentation System Page 8
Response
During the March 1, 1984 meeting, an upgrade of the CEICs was proposed.
The proposed upgrade was not determined to be unacceptable and there-fore, we are proceeding with the upgrade based on our presentation.
It was stated that this upgrade will not fully satisfy all of the design criteria of NUREG-0737.
The following describes the proposed upgrade with respect to single failure criteria.
The CETICs are routed frm the reactor vessel to the control roan as two separate trains. Due to the configuration of the CETIC's inside the reactor vessel, it is not possible to meet single failure criteria. This is a generic issue and has been recognized as an area where single failure criteria cannot be met. Each train consists of a half of the CEICs. Either train will be capable of providing core exit tmperatures across the entire core. As a result of the installation of the SPDS, the CEIC signals are fed directly frm the cable vault side containment penetration cab-inets to the SPDS cmputer. This results in separate and independent instrument channels beginning at this point for the primary and the back-up display channels.
'Ihe CEIC trains are physically separated to the extent that they do not share the same raceway (conduit or cable tray) at any time until just prior to entering the control rom. 'Ihe cables are mixed in a single short piece of cable tray then pass through a single floor penetration prior to entering the thernoccuple indicating cabinet IHoneywell Precision Tarperature Indicator). See Figure 1 for a schmatic diagram of the proposed upgraded ICC Instrumentation Systen.
A review of plant drawings has been completed in order to determine the routing and location of these raceways.
It has been determined that the two trains are physically next to each other throughout containment and physically separated outside containment.
In sme instances the raceways are located side-by-side, and in other instances, they are routed above the other (one foot separation between raceways in either configuration).
Included with this suhaittal are copies of plant drawings indicating raceway routings.
See Appendix B for a listing of these drawings.
The routing indicated on these drawings have been reviewed for the following comon failure modes:
High energy line break: The cmponents to be installed inside containment will be environmentally qualified to the DBA environ-mont; the components outside containtmnt are located in areas where a harsh environment is not expected as determined by the station'c review of areas containing safety-related cmponents and reviewed in accordance with equipnent environmental qualification requirements.
Attachment IOC Instrumentation System-Page 9 Missile protection; raceways in containment are located on the operating deck, incore instrument room and between the crane wall and the containment wall. 'Ihere are no missile sources within any of these imnediate areas.
'Ihere are no missile sources outside containment which could affect these raceways.
Fire: Inside containment - in accordance with our review pursuant to 10CFR50, Appendix R, it has been determined that the fire loading in containment is very low.
In specific areas where a greater potential for a fire does exist, (ie:
residual heat rmoval pumps and the cable penetration areas), fire suppression and detection equignent has been added. As seen frm the drawings, the CEICs pass through containment in two separated areas (east cable vault, west cable vault) thus further reducing the possibility of a comnon failure inside containment. Also, an oil collection system has been installed on the reactor coolant punps.
Outside containment - the two trains enter the cable vault area, one train in each cable vault, and these areas lead to the cable mezzanine. Each of these areas are protected by a CO,, fire suppression systm. 'Ihe routing for the CEIC cabling to tM primary display follows a separate path to the process instrument rom where they are terminated in the SPDS and PVS cabinets.
Once entering the mezzanine, the two back-up display trains separate and follow different paths until rejoining prior to entering the control rom.
One train reains in the mezzanine area for the entire distance, the other does not.
Therefore, one train is protected by a 00 system frm the containnent penetration to the 7
control rocm in its entirity.
Our review has also concluded that there are no power cables sharing raceway with the CEICs at any time. All cabling in these raceways are Non-lE instrument grade, with low ratings, therefore failures are further reduced due to there not being power cables in any of the subject raceways.
The other non-lE instrument cabling in these shared raceways would be considered associated circuits in accordance with the application of Regulatory Guide 1.75.
Isolation devices will not be added to these associated circuits and as such represents an exanple of not meeting separation criteria.
This is considered acceptable since this is instrumut cabling which only carries small currents associated with plant instrumentation and is not a credible ignition source.
Upon entering the control rom, the CEICs terminate in the lioneywell Precision Tenperature Indicator. As part of the proposed upgrade the CEICs will be fed in parallel frcm this lioneywell Inmcator to *he new microprocessor cabinet (IOC tbnitor).
'Ihis cabinet is divided into two trains with each train separated into a field wiring area and a processing electronics area, with a metal barrier separating the two areas.
A parallel feed of the CEICs will be taken frm the input side of the microprocessors cabinet and routed to the plant P-250 cmputer.
The control room is equipped with a fire detection system.
Attachment ICC Instrumentation System Page 10 The design of the proposed upgrade includes a separate 1E power source to each train of the ICC Monitor and its displays (See Figure 1). A description of the primary CEIC indicating system and the back-up systems for monitoring CEICS, as part of the ICC Monitor, has been provided in response to question 4e.
Based on this review, there is no single loss of power which will cause a loss of both the primary and the back-up systems.
'Ihere is not a single power-related failure as described by this review that is considered credible which would remove the ability to monitor core exit tmperatures.
It is therefore concluded that the proposed upgrade of the CEICs will adequately meet single failure criterion.
-i.
The specified @ has not been applied to the system which was ccmpleted before implementation of NUREG-0737.
Response
Equipnent and structures, whether safety related or not, are subject to engineering review, and shop and field inspection to a degree propor-tional to the value of the equipnent and to its contribution to safety, accessibility, reliability and operability. This philosophy of quality assurance was applied during the construction cf Beaver Valley, Unit 1, and is still applied today. The core exit thernoccuple system was orig-inally installed as an instrument grade system, not a protection system, and as such was not defined as a @ Category I system. As per our FSAR, Appendix A, a @ Category I system is defined as follows:
Plant systms, or portions of systems, structures and equip-ment whose failure or malfunction could cause a release of radioactivity that would endanger public safety. 'Ihis cat-egory also includes equipmnt which is vital to a safe shutdown of the plant and the removal of decay and sensible heat, or equignent which is necessary to mitigate consequences to the public of a postulated accident.
There are three quality assurance categories at Beaver Valley.
These categories are not levels of quality per se, but rather denote differ-ences in application of the Quality Assurance Program. Category I and only Category I is intended to include activities affecting structures, systems, portions of systems, and equignent to which the 10CFR50, Appendix B criteria apply.
The formal Quality Assurance Program, incorporating the intent of the eighteen criteria in Appendix B to 10CFR50, applies without exception to all items classified as Category I.
Our @ program for Category I activities has been reviewed against the @ guidelines identified in NUREG-0737, Appendix B, and the equiva-lent of thase guidelines. The applicability of the @ Program for other categories is not determined by the eighteen criteria in Appendix B to 10CFR50, but rather by Duquesne Light Cmpany's policy and is a matter of agreement between the C xpany and our agents and suppliers.
Attachrent ICC Instrutentation Systs-Page 11 The Westinghouse Electric Corporation provided the incore instru-nentation system for Beaver Valley Unit 1.
As a NSSS supplier, they have been evaluated by the NRC to determine the adequacy of their quality programs.
The Westinghouse quality programs were determined acceptable by the NPC. The Duquesne Light Cmpany QA Department has also evaluated the Westinghouse quality program and have included them as a qualified supplier for specific parts and services.
Therefore, although the core exit thernoccuples have not been labeled a Category I system, their interface with other Category I emponents (ie:
reactor vessel) and the fact they were supplied by a vendor with an acceptable QA program provides an adequate level of assurance that a quality systs has been installed at Beaver Valley Unit 1.
The proposed upgrade of the ICC Instrumentation System will be perforned as a QA Category II modification and will not be subject to full imple-nentation of our QA Program.
This is consistent with established criteria for determining the QA category for new work.
Although this upgrade is recognized as not meeting the definition of QA Category I, and therefore not subject to the DIC QA Program it should be noted that Quality Assurance requirements can be applied to the supplier for work performed associated with this upgrade. Therefore, in accordance with our agreement with the Westinghouse Electric Corporation, the supplier of the c m ponents for this proposed upgrade, their activities associated with this upgrade will be performed in accordance with HCAP 9245, the Westinghouse Quality Assurance Program Plan, which has been reviewed and determined to be acceptable by both the NBC and the Duquesne Light Cmpany.
This is considered an acceptable application of QA criteria and will provide an ICC Instrumentation System procured with an adequate level of quality assurance.
k.
No periodic testing is provided as required by NUREG-0737 and Pegulatory Guide 1.118.
Posponse NUREG-0737, Appendix B item 18 states that periodic testing should be in accordance with the applicable portions of Regulatory Guide 1.118,
" Periodic Testing of Electric Power and Protection Systems", pertaining to testing of instrument channels.
We have reviewed this Pegulatory Guide for applicability to the ICC Instruruntation System. This Guide describes a method for testing systems during plant operation such that a licensee may cmply with Comission regulations regarding operability of electric power and protection systems.
The ICC Instrumentation System is not a part of the electric power and protection system and is therefore not subject to the testing guidelines provided in Pcgulatory Guide 1.118.
The operability of this system will be determined in accordance with Technical Specifications when this instrumentation systm is added to the Beaver Valley Unit 1 Technical Specification addressing Accident bbnitoring Instrumentation.
Our Technical Specifications will be anended to include this instrumentation as stated in our response to
Attachment IT Instrumentation System-Page 12 question 4f.
By including this instrumentation in our Technical Speci-fications, operability requirments will be established which are consistent with other accident monitoring instrumentation.
This ap-proach is considered acceptable and meeting the intent of Pegulatory Guide 1.118.
1.
No isolation of primary and backup displays or instrumentation is provided.
Desponse An isolation device in a circuit is designed to prevent credible fail-ures at the output of the isolation device frm interfering with a protection system channel frcm meeting the minimum performance require-ments specified in the design basis. Instrument signals associated with protective actions at Beaver Valley Unit 1 are electrically isolated frcm the proposed ICC Tnstrumentation System to preclude any dcwnstream failures frm interfering with their associated protective function.
'Ihe proposed ICC Monitor includes microprocessor-based units designed to perform sigr.a1 conditioning, isolation, data acquisition, processing, and transmitting of the data to the displays.
As such this design provides additional protection for instruments associated with protec-tive functions frcm failures of ccuponents beyond the IOC Monitor microprocessor.
This electrical isolation protects the CEICs from failures downstream of the ICC monitor.
Therefore, isolation of the plasma display which serves as the CEIC back-up display has been provid-ed.
As seen on Figure 1, the CEICs are parallel fed frcm the containment terminal boxes to the ERF cmputer (PVS and SPDS) and to the Honeywell Precision Temperature Indicator before going to the ICC Monitor micro-processor cabinet.
No electrical isolation is provided for the CEICs.
Since the CEICs do not provide protective actions, this design is consistent with the Beaver Valley philosophy on signal isolation, and we consider this acceptable for the CEIC system.
As previously stated, the IOC Monitor microprocessor is equipped with signal conditioning.
The signal conditioning boards provide surge protection capabilities, front end filtering and conmon node isolation.
This design feature will provide an adequate level of protection to the IOC bbnitor frcm upstream failures. The primary CET display, the SPDS, in also equipped with front end surge protection.
In sunmary, the protection system instrument signals are provided with electrical isolation, the SPDS and ICC tbnitor are each equipped ~ with front end surge protection and isolation exists between the plasma displays and the inputs to the ICC Monitor.
This design will provide protection to the ICC Instmmentation System and preclude ccnmon mode failures thus meeting the intent for providing isolation of primary and back-up displays and related instrunentation.
Attachment ICC Instrumentation System-Page 13 m.
None of the CET system is environmentally qualified as specified in Appendix B.
Fesponse During the March 1,1984 meeting with members of the NRC staff, the details of a proposed upgrade to the CEIC system was presented. This proposed upgrade eddressed environmntal and seismic concerns and was considered to have been an acceptable approach. Apparent agreement on the details of this upgrade was obtained.
'Ihc upgrade of this system will result in qualified cabling, thermocouple connectors and reference junction boxes installed inside containment. Also included in the upgrade will be a new ICC Monitoring System which will cmbine CEICs, subcooling, and RVLIS into one qualified cabinet with redundant plasma displays.
This upgrade will meet the qualification criteria contained in Regulatory Guide 1.97 on this basis.
The proposed upgrade considered the original seismic design of Beaver Valley Unit 1 as being adequate.
It was not a part of the proposed upgrade to install ceismically qualified raceways, however, it was pointed out that raceways in the plant were originally installed following the sam installation criteria stiether the original specification required a seismic installation or not. This resulted in raceways installed in accordance with the seismic Class 1 criteria in effeet at the time the plant was originally constructed. Based on the seismic adequacy of the present design in cmbination with the new cmponents emprising the proposed upgrade, we consider the seismic design criteria of the ICC Instrumentation System as being adequately addressed.
The adequacy of the outside containment cabling is based on our review of plant spaces.
Also, the sam type cabling has been used in other plant instrument capacities and is currently being tested for our containment environment in accordance with our equignent qualification effort. The outside containment environment for the CEICs is enveloped by the existing tests which have been performed on this cable.
Therefore, this cabling is adequate for its operating and expected post-accident er.vironment.
Pending the results of the ongoing testing, the adequacy of the cable inside containment can be determined.
n.
Exception from the requirements of Attachmnt 1 and Appendix B is pleaded on the basis that upgrade cost of $4M is unwarranted for the amount of safety enhancemnt achieved. This cost appears too high for those sixteen Cer's used for the backup display. Justi-fication for this high cost estimate should be provided.
Response
The costs associated with upgrading the CEIC system was discussed during the March 1,1984 meeting with nunbers of the NBC Staff. A presentation was made which described the existing CEIC system and the work involved 1
Attachment IOC Instnmentation System' Page 14 Response, (Continued) in upgrading to meet all the design criteria in NUREG-0737. As indicat-ed at that meeting, Beaver Valley has been constructed in a very cmpact fashion due to limited land availability. Our containtrent is subatac-spheric and by its nature small in ccmparison to others. As a result, the most cost efficient routing of raceways is not always available due to interference with other equipment.
This cmbination of design restrictions results in increased difficulty in performing station nodifications and this translates into higher costs.
The cost of performing plant nodifications has been studied to determine the factors influencing the final estimated project cost.
This has resulted in procedures which are used to estimate nodification costs.
New estimates were prepared prior to the March 1 meeting. %e estimate for upgrading to meet all the design criteria of NUREG-0737 was presented as option B to the NBC staff at a cost greater than $4 million.
The cost estimates were obtained by determining labor costs, engineering costs, and material costs. %is breakdown is as follows:
l Material Costs:
$ 800,000 Direct Engineering $ 100,000 1
Costs:
Labor Costs:
$3,700,000 Wis upgrade would have affected all 51 thernoccuple channels in accor-dance with NUREG-0737. There is no differentiation made in the NUREG to state only 16 thermocouples versus all must be upgraded in order to satisfy this NRC design criteria. to this NUREG item states:
thermocouples shall be of sufficient number to provide indi-cation of radial distribution of the coolant enthalpy and that distribution symmetry should be considered when deter-mining the specific nunber and location of thermocouples to be provided for diagnosis of local core problems.
there should be a primary display available on demand indicating tenperature across the core at each core exit thernucouple location.
a back-up display should be provided capable of reading a minimum of 16 thernoccuples, 4 frcm each core quadrant..
the instrumentation must be evaluated for conformance to Appendix B " Design and Qualification Criteria for Accident Ibnitoring Instrumentation".
the primary and back-up display channels should be electrically independent, energized frcm independent station Class lE power sources, and physically separated in accordance with Hegulatory Guide 1.75, with exceptions given beyond the isolation device.
Attachment ICC Instrumentation System Page 15 Response, (Continued) the instrumentation should be environmentally qualified as described in Appendix B, item 1 except for seismic qualifi-cation to the primary display and hardware beyond the isolator.
In addition, the work and mterials involved with installing raceways and detemining routing would be approximately the same for 16 CEICs as it would for 51 CEICs since two trains would have to be installed to meet the above criteria. We have considered our approach at estimating the cost for a full upgrade of all 51 CEICs as consistent with the above criteria.
In order to address other aspects of the ICC Instrumentation System, (RVLIS and SbM), we have elected to integrate these functions into one ICC bbnitor and upgrade the system in accordance with our proposed upgrade presented during the March 1 meeting and further documented in this subnittal.
We believe this to be a cost effective approach in meeting the intent of a system consisting of a reactor vessel level indicator, subcooling margin nonitor and core exit thernoccuples capable i
of detecting the approach to, existance of and recovery frm an I
inadequate core cooling condition.
5.
The subcooling margin nonitor was installed following guidance provided in NUREG-0578 and does not confom to all requirements of NUREG-0737, Appendix B.
The licensee indicates that further review of the SMM will be done as part of his response to Regulatory Guide 1.97 on a schedule subnitted in response to Generic Intter 82-33, dated April 15, 1983.
Provide a review of conformance or non-conformance as requested in Generic letter 82-28.
Response
The proposed upgrade which has been described in detail includes a subcooling margin nonitor as a part of the ICC bbnitor. The existing SbM will be removed when the proposed upgrade is emplete.
The following provides additional information on SbM operation and design.
The reactor coolant system pressure is utilized by the ICC Monitor microprocessor to calculate the core saturation temperature for that reactor coolant system pressure.
By subtracting the auctioncrred high incore thernoccuple signal frm the calculated saturation tmperature, the nargin to saturation is determined. This information is displayed on the plasma displays which have been previously discussed. 'Ihe margin to saturation indication frcxn loop hot leg RIDS is also available on this display. Alarms for this ICC monitor were discussed in response to Question 4b, Tha hot leg RIDS are scheduled to be replaced during the upacming fourth refueling outage in order to satisfy equipment cualification require-ments. The RCS wide-range pressure transmitters have accuracy problems which have been identified by the NBC which resulted in Information
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Attachment ICC Ins *mtation System-Page 16 Response, (Continued)
Notice 82-11 being issued. This Notice addresses potential inaccuracies in wide-range pressure instnments used in Westinghouse designed plants.
A resolution to this probim exists through the Westinghouse RVLIS system.
In order to achieve instrument accuracies required to satisfy their generic design specification, Westinghouse has rtodified its original design to provide two wide-range pressure transmitters outside containment which tap into the RVLIS capillary tubing.
These new transmitters monitor RCS wide-range pressure and provide inputs to the RVLIS microprocessor.
These will be installed as part of the RVLIS system when RVLIS is declared operable following our fourth refueling outage.
Because these transmitters were designed to interface with RVLIS, the capability for them to interface with other cmponents is not possible unless scme modifications are made. We have considered this an a design input to the proposed upgrade to install the new IOC Monitor.
This rtodification will result in IE power being supplied to the new RCS wide-range pressure transmitters and electrical isolation being provided.
The new ICC rionitor will replace the existing RVLIS microprocessor and RVLIS inputs will be transferred to the ICC Monitor.
As a result the StH will have inputs frm qualified systms, as described by this subnittal, and will meet the intent of the tac design criteria for an IOC Instrumentation System.
MILES'IONFS FOR IMPLEMENTATION OF ICC INSTRUMENTATION Pesponse Your request for additional informtion included Appendix B which provided milestones relating to inplerentation of our ICC Instrumentation System which should be incorporated into our scledule.
Each of these milestones have been reviewed to assure each has been addressed.
As ICC instrumentation is made operational, m will subnit an inplerentation letter providing the information requested. All otimr milestones have been considered addressed either in our subnittals or i
schedule.
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APPENDIX A Nem 6 Page 1 Explain how the value of the system accuracy (given as +/- 6% was derived.
How were the uncertainties from the individual components of the system combined? What were the random and systematic errors assumed for each component? What were the sources of these estimates?
I Resoonse 6.
The system accuracy of 16% water level was a target value established during the conceptual design and was related to the dimensions of the reactor vessel (12% from nozzles to top of core) and core (30%), and the usefulness of the measurement during an accident.
Subsequent analyses have established a system accuracy based on the uncertainties introduced by each component in the instrument system.
The individual uncertainties, result 1ng from random effects, were combined statistically to obtain the overall instrument systen accuracy.
Some of the individual uncertainties vary with conditions such as system pressure. The following table identifies the individual uncertainties for the narrow range measurement while at a system pressure of 1200 psia.
Uncertainty Comoonent and Uncertainty Definition
% Level l
l a.
Differential pressure transmitter i 2.1 l
calibration and drift allowance, I
(11.5% of span) multiplied by the f
ratio of ambient to operating water l
density.
l b.
Differential pressure transmitter 10.7 allowance for change in calibration due to ambient temperature change (10.5% of span for i 50aF) multiplied by the density ratio.
APPENDIX A c.
Differential prossure transmitter
[6$3k allowance for change in ' calibration due to change in system pressure (10.27. of span per 1000 psi change) multiplied by the density ratio.
d.
Differential pressure transmitter 1 0.7 allowance for change in calibration due to exposure to long-term overrang,e (10.5% of span) multiplied by the density ratio.
e.
Reference leg temperature instrument 1 0.64 (RTD) uncertainty of 150F and or allowance of 1 50F for the difference bstween the measurement and the true average temperature of the reference leg, applied to each vertical section of the reference leg where a measurement is made.
Stated uncertainty is based on a maximum containment temperature of 4200F, and a typical reference leg installation.
f.
Reactor coolant density based on auc-1 2.3 tioneering for highest water density obtained from hot leg temperature (1 60F) or system pressure (+ 60 psi).
Magnitude of uncertainty varies with system pressure and water level, with largest uncertainty occurring when the reactor vessel is full.
t'.'
APPENDIX A Page 3 g.
Sensor and hydraulic iso.lator bellows i 1.46 displacements due to system pressure changes or reference leg temperature changes will introduce mince errors in the level measurement due to the small volumes and small bellows spring constants.
The changes, such as pressure or temperature, tend to cancel, i.e., the bellows associated with each measurement move in the same direction.
Maximum expected error due to differences in capillary line volume and local tempera-tures is equivalent to a level change of about 5 inches, multiplied by the density ratio.
^
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h.
Density function generator output mis-1 0.50 match with ASME Steam Tables limited to a maximum of:
1.
Electronics system calibration, overall 1 1.0 uncertainty limited to less than:
j.
Control board indicator resolution; 1 0.5 microprocessor digital readout to nearest percent of level span.
The statistical combination of (square root of the sum of the squares) of the individual uncertainties described above results in an overall system instrumentation uncertainty of 13.9% of the level span.
For the narrow range indication of approximately 40 feet, or i 1.5 feet, at a system pressure of 1200 psia.
Examples of the uncertainty at other system pressures are:
Uncertainty = 13.6% at 400 psia Uncertainty = 1 4.2% at 2000 psia
e 1
b e'
s APPENDIX B List of Drawings Drawing No.
Title ll700-RE-57V-5 Instrumentation Conduit Reactor Containment El. 738-10 ll700-RE-34AS-5 Cable Tray Designations Peactor Containment El 738-10 11700-RE-34AY-7 Cable Tray Section Designations Sh. 1 Peactor Contalment ll700-RE-34AR Cable Tray Designations Reactor Containment El. 767-10 ll700-RE-34AF Cable Tray Designations Pcactor Containment El. 767-10 ll700-RE-34AK-9 Cable Tray Section Designations Sh. 1 Peactor Containtent ll700-RE-34AX-7 Cable Tray Designation Cont. & Cable Vault El. 735-6 ll700-RE-34AN-9 Cable Tray Designations Cable Vault Area 8700-RE-34AC Cable Tray Designations Control and Switchgear Area Sheet tb.1 ll700-RE-34AD-18 Cable Tray Designation Cont. & Switchgear Area - Sh. 2 ll700-RE-37DC Sleeve Designations Control and Computer Roms 11700-RE-35A-5 Arrgt. - Electrical Penetration Peactor Containment