ML20042F913

Responds to Generic Ltr 89-19 Re Resolution of USI A-47 on Overfill Protection for Steam Generators in PWRs & Reactor Vessels in Bwrs.Operability of Overfill Protection Sys Assured by Performing Channel Checks Every 12 H

ML20042F913
Person / Time
Site:	Hope Creek
Issue date:	05/04/1990
From:	Labruna S Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
REF-GTECI-A-47, REF-GTECI-SY, TASK-A-47, TASK-OR GL-89-19, NLR-N90094, NUDOCS 9005100153
Download: ML20042F913 (10)

Text

Pubhc Service Dectnc and Gas Company Stanley LaBruna PutWe Service Doctnc and Gas Company P.O. Box 236, Hancocks Bndge, NJ 08038 609-339-4800 vasw.*nt' u w or++ =

May 4, 1990 NLR-N90094 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlement GENERIC LETTER 89-19 RESOLUTION 0F UNRESOLVED SAFETY ISSUE A-47 HOPE CREEK GENERATING STATION DOCKET No. 50-354 On September 20, 1989, the Nuclear Regulatory Commission (NRC) issued Generic Letter 89-19, concerning overfill protec' ion for steam generators in PWRs and reactor vessels in BWRs.

241blic Service Electric and Gas Company (PSE&G) provided its response for Salem Units 1 and 2 in NLR-N90057, dated March 20, 1990.

PSE&G hereby provides its response to Generic Letter 89-19 for the Hope Creek Generating Station.

The Hope Creek Generating Station meets the requirements for satisfactory reactor vessel overfill protection as delineated in Generic Letter 89-19.

The justification for PSE&G's assessment is contained in Attachment 1.

Should you have any questions regarding this transmittal, please feel free to contact us.

Thank you.

Sincerely, ffy/d/tcrzw Affidavit Attachments (3) 9005100153 900504

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Document Control Desk

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5-4-90 NLR-N90094 C

Mr. J.

C. Stone Licensing Project Manager Mr. C.

Y. Shiraki Licensing Project Manager Mr. T.

P. Johnson Senior Resident Inspector Mr. T. T. Martin Administrator - Region 1 Mr. Kent Tosch Chief - New Jersey Department of Environmental Protection Divicion of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625 i

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r Ref NLR-N90094 STATE OF NEW JERSEY

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COUNTY OF SALEM

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S. LaBruna, being duly sworn according to law deposes and says:

I am Vice President - Nuclear Operations of Public Service Electric and Gas Company, and as such, I find the matters set forth on our letter dated May 4, 1990, concerning the Hope Creek Generating Station, are true to the best of my knowledge, information and belief, hf tw

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Subscribp and Sworn,/, +to before me this _'/_

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1990

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htary Public of New Jersey LARAINE Y, BEARD W

Notory Public of New leney My Commission Expires May 1,19yl My Commission expires on 9

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HLR-N90094 ATTACHMENT 1 JUSTIFICATION FOR HOPE CREEK GENERATING STATION RESPONSE TO GENERIC LETTER 89-19

) of Generic Letter 89-19 stated that existing designs would be acceptable if the criteria listed for the specific j

design are met.

The Hope Creek Generating Station is a GE Boiling-Water-Reactor (BWR-4) plant that is equipped with automatic reactor vessel overfill protection.

The criteria and response for the Hope Creek Generating Station are listed below:

criterion All BWRs provide automatic reactor overfill protection to nitigate main feedwater (MFW) overfeed events.

The design of the overfill-protection system should be sufficiently separate from the MFW control system to ensure that the MFW pumps will trip on-a reactor high-water-level signal when required, even if a loss of power, loss of ventilation, or a g

fire in the control portion of the MFW controls system should occur.

Response

PSE&G concurs with "BWROG Response to NRC GL 89-19,, Hardware Change Recommendation" (Attachment 2) in that the Hope Creek Generating Station is presently equipped with adequate, automatic reactor vessel overfill protection.

Any safety benefit gained by providing additional system redundancy and independence from the existing equipment would not be significant.

1 Criterion All BWRs reassess operating procedures and operator training and, if necessary, modify them to ensure that operators can mitigate reactor vessel overfill events that may occur via I

the condensate booster pumps during reduced pressure operation of the system.

Response

The Hope Creek System Engineering, Operations and Operations Training Departments have reviewed current operating procedures and operator training programs pertaining to reactor high level conditions and overfill events and have concluded that no modifications are required.

(continued) i Page 1 of 2 1

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4 Abnormal Operating Procedure OP-AB.ZZ-117 (Q), " Reactor High Level" (Attachment 3) provides guidance to control room operators pertaining to reactor overfill events.

Directions are given to recover level during conditions in which reactor level control is either on the Master Level Controller-(>20% reactor power, normal operating pressure) or the Startup Level Controller (<20% reactor power, zero to normal operating pressure).

Additionally, direction is given to close the MSIVs, terminate all vessel feeds and ensure the reactor has scrammed if reactor level increases to +90 inches; this corresponds to a level that is 36 inches above the high level trip setpoint and 28 inches below the bottom of the main steam line vessel penetrations.

Therefore, the entire spectrum of possible reactor operating pressures is addressed.

Hope Creek Reactor Operators and Senior Reactor Op3rators receive training on OP-AB.ZZ-117(Q) during the Initial Licensed Operator Training Program and periodically thereafter in Licensed Operator Requalification Training.

Criterion Plant procedures and technical specifications for all BWRs with main feedwater overfill protection include provisions to periodically verify the operability of overfill protection and ensure that automatic protection is operable to mitigate main feedwater overfill during power operation.

Response

v Hope Creek Technical Specification 3/4.3.9, "FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION" delineates the limiting condition for operation and associated surveillance requirements for the automatic overfill protection system.

This includes the automatic trip of the Main Turbine and Reactor Feed Pump Turbines caused by a high

. reactor vessel level (+54 inches).

Operability of the overfill protection system is assured by performance of: 1)

Channel Checks every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> in accordance with OP-DL.ZZ-026(Q); monthly Channel Functional Tests in accordance with IC-PT.BB-027(Q), IC-PT.BB-028(Q),

and IC-FT.BB-029(Q); and 18 month Channel Calibration Tests j

in accordance with IC-CC.BB-060(Q), IC-CC.BB-061(Q), and IC-CC.BB-062(Q).

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ATTACHMENT 2 INGt0G RESPONSE TO NRC GL 89-19,

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ENCIDSURE 2, HARDWARE CHANGE RECOBOtENDATIONS i

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=o WROG 9048 clo ca,ohno power 4, Light company e all povetteville street. Aoisigh. NC 27602 April 2. 1990 Office of Nuclear Reactor Regulation U. 5. Nuclear Regulatory Commission Washington, DC 20555 Attention:

James G. Partlow Associate Director for Projects I

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SUBJECT:

SUBMITTAL 0F BWR OWNER $' GROUP RESPONSE TO GENERIC LETTER 89 19 L

Reference:

' Request for Action Related to Resolution of Unresolved i

l; Safety Issue A 47 ' Safety Implication of Control Syttees in LWR Nuclear Power Plants' Pursuant to 10 CFR 50.54(f) -

1 Generic Letter 89 19", September 20, 1990 ThislettersubmitstotheNRCtheBWROwners' Group (SWROG)reportin response to Generic Letter 89-19 reference.

results of a study of automatic ov(erfill pro)tection systems curMntlyThe repo utili7ed by SWRs. The report concludes that the SWRs addressed by the report provide adequate and reliable automatie overfill protection consistent with the NRC requirements for closure of Unresolved Safety Issue A 47.

In WUREG 1217 and NUREG 1218, the NRC recognizes that the safety benefits gained by providing additional protection system redundancy and independence from existing main feedwater control system equipment is not significant, and.that modifications costing in excess of $100,000 are not cost beneficial.

The BWROG report demonstrates that the cost to make plant modifications to provide additional redundancy and independence is substantial and therefore the modifications are not cost beneficial.

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4 BWROG 9044 April 2, 1990

' page t The comments / positions provided in this letter and report have been endorsed by a substantial number of the members of the WROG; however, it should not be interpreted as a commitment of any individual member to a specific course of action. Each member must formally endorse the SWROG position in order for that position to become that member's position.

Stephen D. Floyd, Chairman SWR Owners' Group

Attachment:

'8WROG Response to NRC Generic Letter 89 1g Enclosure 2 Hardware Change Reconner.dations" i

cc:

F. J. Miraglia, NRC W. T. Russell, NRC A. C. Thadani, NRC G. J. Beck, BWROG Vice Chairman D. N. Grace, RRG Cl. airman BWROG Executive Oversight Committee BWROG Primary Representatives L

4WROG Control Systems Committee 4

L. S. Gifford, GE

5. J. Stark, GE 1

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1 4-EDE 07-0390 DRF A00 03773 March 30, 1990 l

Revision 0 3

4 BWROG RESPONSE TO NRC GL89 19, ENCLOSURE 2, HARDWARE CHANGE RECOP91ENDAT!0NS i

i BY:

D. E. BENNETT J. Y. FUJITANI APPROVED: T.Q. b S. J. Stark, Manager BWR Owners' Group Programs Regulatory and Analysis Services APPROVED:

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.t Koslow, Manager Regulatory and Design Compliance Application Engineering Electrical Design Engineering (EDE)

BWROGF2.TXT Q(l y fj{p& fl $ N f f

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EDE 07 0390 CONTENTS i

i EAE1 1.

INTRODUCTION 3

2.

SUMMARY

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DESCRIPTION 5

3.1 Methodology and Analysis 5

3.2 Cost of Modifications 6

3.3 Existing Systems Reliability 7

3.4 Conclusions 7

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REFERENCES' 8

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TABLES l

TABLE 1: PARTICIPATING BWROG NUCLEAR POWER PLANTS AND LICENSEES 9

TABLE 2: TYPICAL BWROG PLANT RPV'0VERFILL PROTECTION SYSTEM 10 ll-CONFIGURATIONS TABLE 3: MODIFICATION COST RANGES 11 L

i APPENDICES y

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APPENDIX A - REACTOR PRESSURE VESSEL HIGH WATER LEVEL TRIPS 12 l

APPENDIX B - MODIFICATION COST ESTIMATES 15 2

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l EDE 07 0390 i

1.

INTRODUCTION The U.S Nuclear Regulatory Comission (NRC) has conducted technical

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evaluations of Unresolved Safety Issue (USI) A-47 " Safety Implications of Control Systems" (see references 2 and 3). As part of the resolution of USI A 47, the NRC issued Generic Letter (GL) 89 19 (see reference 1) that summarries these evaluations and makes recommendations to implement changes which address technical evaluation concerns. Specifically, GL89 19,, Section la recomn. ends that all BWR nuclear power plant licensees:

1) provide an automatic reactor pressure vessel (RPV) overfill protection

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system to mitigate main feedwater (MFW) control system overfill events; and that adequate system logic configurations, trip channel

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separation, and separation from main feedwater (MFW) control system equipment be provided to prevent specific MFW control system common l

L mode failures (loss of power or ventilation, or a fire in the MFW control system) from resulting in a RPV overfill event; and if not, at least ensure that a MFW pump trip will occur from such failures.

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2) reassess their operating procedures and operator training and modify

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l-them if necessary to ensure that the operators can mitigate RPV overfill events that may occur via the condensate booster pumps during reduced reactor pressure operation of the system.

This report presents the results of a BWROG study of automatic overfill protection systems (item 1 above) related to the reconnendations of GL8919 and to ths NRC assessments reported in NUREG 1217 and NUREG-1218 (see references 2 and 3).

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EDE 07-0390 2.

SUMMARY

A review of individual BWR plant-specific drawings end high water level trip records confirms that all of the BWR plants listed in Table I currently provide adequate automatic RPV overfill protection. Furthermore, these overfill protection systems are believed to be consistent with the NRC requirements for closure of US! A-47. In references 2 and 3, the NRC reccgnized that the safety benefits gained by providing additional RPV protection system reiundancy and independence frou existing MFW control system equipment is not significant, and that modifications costing in excess of $100,000 are not cost beneficial. What is significant in these references is mainly that some sort of reliable automatic RPV overfill protection be provided. The BWROG concurs with this assessment.

As indicated in reference 2 (NUREG-1217), Page 13, Se tion 3.2.1, a review of BWR plant operating experience did not identify any MFW system RPV overfill event subsequent to the installation of an automatic RPV overfill protection system. A current GE survey of BWR high RPV water level 8 trips supports this conclusion (see Appendix A).

In addition, the reviewed records did not identify the occurrence of any common-mode MFW control system failure that might have resulted in RPV overfill.

If such a control system failure had occurred, because of current plant designs (described in Section 3) it is unlikely thct such a failure would have actually resulted in a RPV overfill event (filling the main steam lines with water).

As discussed in this report, to fully implement the recommendations of GL89-19 requires substantial plan' modifications with little safety benefit, therefore the modification; were not considered to be cost beneficial, l

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3. DESCRIPTION i

3.1 Methodoloav and Ana'lynis -

The study first reviewed in detail, c9rrent plant specific documents (mainly plant control system and some plant protection system elementary drawings) to assess and determine the existing plant RPV overfill protection and Main feedwater (MFW) cnntrol system design configurations. These data were then reviewed and tabulated by plant i

into a date matrix format. This mat'rix identified the number of.

sensor lines, sensor trip units and trip relays used as well as the power source for each and the logic channel separation and type.

In L

addition the number and 'Icc4 tion of the inriraent racks.and panels used by each device ~was identified.

Using this matrix, each plant was then grouped by logic configuration, sensor lines, sensors, and the racks and panels used by each, into one of five groupse A through E (see Table 2). These plant groups tvere -

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then used as the. basis:for the analysis and for estimating the cost to L

modify the existihg RPV overfill protection system consistent with the L

GL89-19 recommendations, s

As was expected, each plant RFY overfill p~rotection system accesses l:

two or more independent RPV level sensor lines, using two or more sensors. This configuration supported the associated trip logic, but

'in many cases the device and'the rack and panel served both the RPV overfill protection system and the MFW control system functions. This arrangement is not consistent with the complete channel and system separation recommendations of GL89-19.

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1 3.1 Methodoloav and Antivsis (continued)

Similar to most.non sefety related system arrangements, most of the plants reviewed have mixed alternate channel trip logic devices, commingled with other MFW control system equipment and installed in i

common racks, panels and plant locations. Many of the RPV overfill protection system devices were not only located in MFW control system equipment racks and panen, but in many cases served both RPV overfill protectics system and MFW control system functions (see Table 2).

' f This. study assumed that all control system inter panel connection l

cable and wiring was commingled. To fully comply with GL89-19 '

l recommendations, substantial equipment and wiring rearrangment and additional RPV overfill-protection trip channels, sensors, logic

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devices, wire and cabling, and evien new racks and panels to house this equipment,.would have to be provided.

t 3.2 Cost of Modifications V

Where modifications would be'needed for complete separation, the cost of providing. additional'RPV level sensor lines and multiple sensors-

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would be prohibit ~ively high (see reference 3, page 28).

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A breakdown of the costs related to making the modifications that would be 'needed to comply with the minimum separation aspects of GL 89-19 is provided in Table 3 and Appendix 8.

From these data, a range D

of modification methods were considered to determine if expensive L

modifications could somehow be made at a more reasonable cost.

c Because ofcthe major design.and engineering cost associated with most changes, no cost effective solutions were found (see Table 3 and AppendixB).

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3.3 Existina Systems Reliability A review of the existing RPV overfill protect. ion system and MFW l

control' system power sources and logic configurations indicated that most plants incorporate some type of " fail-safe" design, where power failures, control signal failures and other credible failures would 1

most likely result in actuating RPV overfill protection and MFW control system alarms, MFW pump trips, main turbine valve closure, a reactor scram, and in some cases MFW flow control valve lock up.

In the unlikely event that a MFW control system common mode failure did result in MFW pump overfeed,.any of these occurrences would alert reactor operators to take immediate corrective action.

Appendix-A provides an assessment of plant operational-experience with the existing RPV overfill protection systems. Based on this assessment, and the current system designs, it is concluded that the

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existing systems provide adequate RPV ovctfill-protection and that any L

safety benefits from modifying these systems in' full accordance with I

GL89-19 would not be significant. This appraisal is consistent with I

theNRC'sassessmentinNUREG1218(seereference3).

3.4 Conclusions p

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The cost involved with the design, purchase and installation of-additional RPV-overfill protection and MFW control system logic b.

channel devices and separation modifications, that fully satisfy the L'

GL89-19 recommendations, is high (see Table 3). Therefore, based on the study findings, the BWROG concludes that. any. RPV overfill risk-reduction provided does not justify the substantial additional 1

cost, not to mention the outage time needed to implement these chenges.

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EDE 07-0390 4.

REFERENCES

')) Generic Letter " Request for Action Related to Resolution of GL89 19 Unresolved Safety Issue A-47, ' Safety Implication of Control Systems in LWR Nuclear Power Plants' pursuant.

to 10CFR50.54(f) Generic Letter GL89-19", issued

~ September 20, 1989.

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2) NUREG 1217

" Evaluation of Safety Implication of Control L

' Systems in LWR Nuclear Power Plants - Technical Finding Related to USI A-47", issued June 1989 i

3) NUREG-1218

" Regulatory Analysis for Resolution of USI A-47",

issued July 1989 4

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L EDE 07 0390 Table 1 PARTICIPATlWG BWROG NUCLEAR POWER PLANTS AND LICENSEES gag Licensee' Brunswick 1 & 2 Carolina Power & Light Company Perry 1 Cleveland Electric Illuminating Company t

Dresden 2 & 3 Connonwealth Edison Company Quad Cities 1 & 2-Commonwealth Edison Company LaSalle 1 & 2 Commonwealth Edison Company Enrico Fermi 2 Detroit Edison Company Hatch 1 & 2 Georgia Power Company i

Clinton'1 Illinois' Power Company Duane Arnold Iowa Electric Light & Power Company Cooper Station Nebraska Public Power District James FitzPatrick New York Power Authority Nine Mile' Point l & 2-Niagara Mohawk Power Company Monticello Northern States. Power Company Susquehanna-1 & 2 Pennsylvania Power & Light Company L

Peach Bottom 2 & 3 Philadelphia Electric Company u

Limerick 1 & 2 Philadelphia Electric Company.

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' Hope Creek Public Service Electric & Gas Company L

Grand Gulf 1 Systems Energy Resources Browns Ferry 1,2-& 3 Tennessee Valley Authority y;

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Table 2 TYPICAL'BWROG PLANT RPV OVERFILL PROTECTION SYSTEM CONFIGURATIONS Group A Plants a

Two out of three high RPV level 8 trip logic l

Three or more shared sensors using ind6 pendent sensor lines and one rack

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Two RPV overfill protection and MFW control system panels l

Groun B Plants i

s Two out of three high RPV level 8 trip logic j

-Two or more. shared sensors using a comon-sensor line and rack l

One or two RPV overfill protection and MFW control system panels Group C P1 ants One out of two.twice high RPV level 8 trip logic -

.Two or more separate sensors using a common sensor. line and two racks Two'or more RPV overfill protection and MFW control-system panels

. Group D Plants

.Two out of two high RPV level 8 trip logic

'Two separate se'nsors using independent sensor lines and two racks-One RPV overf111 prote-tion and MFW control system panel Grouc E P1' ants-1 Two:outL of two high RPV level 8 trip logic

. Two shared sensors u:ics s comon sensor line and rack One RPr overf

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1 Table 3 p

MODIFICATION COST RANGES l

General Costs Associated With Modifying Each Table 2 Plant RPV Overfill Protection System For Compliance With GL89-1S'(see Appendix B)

Anolication Estimated Cost Ranae Minimum -- Maximum r

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1. DESIGN ENGINEERING _

$31K -- $155K-h Provide plant. specific design modification drawings, hardware 3

purchase spectfications, vendor selection, delivery. schedules, quality assurances etc.

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- 2. HARDWARE <

16K --

59K 1

Purchase and delivery n

cost +of hardware j

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3. PLANT ENGINEERING _-

45K -- 160K f

Provide site installation guidance;.

generate change documents; update plant design, operating, licensing, maintenance procedures and documents; equipment.and system acceptance tests

4. INSTALLATION 100K -- 700K Equipment installation and testing

.(Craft labor and materials)

TOTAL COSTS TO IMPLEMENT CHANGES 192K - 1074K n;

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EDE 07-0390 APPENDIX A.

REACTOR PRESSURE VESSEL HIGH WATER LEVEL TRIPS i

Dackaround Several. data bases were reviewed to' determine the frequency of High Water Level (HWL) events in the Reactor Possure Vessel (RPV) of U.S. BWRs. The RPV HWL trip in BWRs is often referred to as the RPV water level 8 trip, and its purposes are to prevent overfill of the RPV and to prevent the introduction of liquid water into the main steam lines.

In all BWRs listed in Tab 1'e 1, a level 8 trip will trip the main turbine, the HPCI turbine or motor, the RCIC turbine, the FW pump turbines or motors, and on some plants the HPCS pump motor.

In newer plants (BWR 6) the level.8 trip will also directly scram the reactor.

If the plant is operating at power levels above the turbine bypass capacity-(typically 15%

to 35% of full power), a main turbine trip should automatically lead to a reactor scram.

If not, the resulting high reactor neutron flux, high reactor pressure, or main turbine control valve fast closure will scram the i.

reactor.- In some cases the operator will manually scram the reactor in the event of HWL before the automatic trip or scram signals take effect.

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l COMPASS Data Base The most complete data base for BWR scrams is GE's Comprehensive Perform-ance Analysis and Statistics System (COMPASS), which includes all outage events for U.S. BWRs from the start of electric power generation to the present.

Scrams in COMPAS3 that could have resulted from RPV HWL were reviewed to determine how many could be positively identified.

E Three cat'egories of trips were identified:

True HWL trips -

' Water level rose to Level 8 and main turbine trip ll occurred. There were 84 such events in COMPASS.

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EDE 07 0390 1

4 APPENDIX A (continued)

. False HWL trips -

A false HWL signal, due to instrumentation or human error, led to turbine trip. Although this.

1 does not represent a true HWL event, the trip-logic was challenged and successfully performed

.l the trip. There were 15 such events.

Possible HWL trips' -

-Scram occurred, and there was a water level L

transient, but the description of the'cVent'was.

not detailed enough to assure that it resulted E

fron,a HWL trip.

There were 11 such events.

l In all'-three of these categories there were llo events, over.a period of:

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'431 reactor. years of comercial operation. This represents 0.26 high RPV L

water level signals per plant year. This period included the long shutdown j

of several BWR plants. In summary, there were, L

True.HWL trips -

84 False HWL trips -

15 Possible HWL trips -

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- Total HWL events -

110 = 0.26 per plant year-NEWLER Data Base The NEWLER data base, maintained by INPO, reports ci licensee event reports

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I(LERs) from January 1984 to present. Unplanned reactor scrams are reporteo-u as LERs, so a search of-the LERs was made to locate HWL events.

Several categories were identified, as follows:

HWL trips that led to turbine trip or scram -

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HWL trips from false signals -

7 HWLtrips(HPCI/RCIC)_whileshutdown-16

. Possible HWL1 trips -

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'Possible HWL trips from false signals -

1 HWL trips from false signals, while shutdown -

4 Possible HWL trips while shutdown -

2 Possible trips froft false signals, shutdown -

1 Tutal events:

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Trips while shutdown appear here and not in COMPASS. Thus, the events identified here but not in COMPASS number 23, and they cover 162.7 plant years of comercial operation, so the shutdown HWL events occurred at a

' rate of 0.14 per plant year.

SUPetARY i

A total frequency of HWm events is obtained by taking the C0HPASS scram

' data, 0.26 HWL related scrams per plant year,-plus the NEWLER data for HWL-trips while the reactor was shutdown (generally following scram), 0.14 HWL trips per plant year. The total frequency of HWL trips in U.S. BWRs has L

been 0.4 per plant-year over the history of commercial operation.

The NEWLER data indicate that the total HWL trip rate since 1984 is 0.34 per plant year, slightly lower than the rate for all years. This is consistent with the scram frequency-experience that shows a decreasing scram frequency per plant year in recent years.

'The total number of challenges to the overfill protection system is tive sum of HWL trips occurring during reactor power operation plus HWL trips-occurring.during reactor shutdown. The COMPASS' data base reports 110 HWL trips and the NEWLER. data base reports 23 HWL trips occurring while sh'utdown during the 1985 th m 1989 period. Thus, there have been.133 total f

HWL challenges to the currently configured automatic RPV overfill

-protection systems with not a single recorded instance of failure.

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NIR-M90094 ATTAC". MENT 3

- OP-AB.ZF-117(Q)

REACTOR HIGH LEVEL

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REACTOR HIGR LEVEL

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- 1.O SIMP 2Dits NORKIN'd t i*ia 1.1 Alarms

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RPV LEVEL 7 VAUDM M ' '

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RPV LEVEL L l

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-j 1.2 Increasing reactor vessel level Mio atter me n >t etu ca is:a c'-

1.2 Increasing reactor power 1.4 Reactor feed flow greater than steam flow 1.5 Controlling level signal fails' low i

1 1.6 Turbine Driven Reactor.Feedwater Pump lock-up 2.O AUTDilATIC ACTIDils

'2.1 Reactor Feed Pump trips

(+54 inches) 2.2 ' Main Turbine trips i+54-inches) 2.3 HPCI/RCIC-Turbine trips,

(+54 inches)

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3.0 IMMEDIATE OPERATOR ACTIONS L

3.1.

the level controller or RFP Turbine Controller nual and' restore vessel level'to between Level 4

Level 7.

3. 2 -

re all appropriate automatic actions are complete.

i 3.3 g.the Unit Scrams implement procedure OP-EO.22-100.

4.O SIIBSIQDERIT DPERAT0]LACTIDItR 4.1 Ensure that all appropriate immediate' operator actions are complete-LOP-AB.22-ll7(O) 1 Rev. O t

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OP-AB o t t-117 ( Q)

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n 4.2' SELECT the alternata level channel (CHAN:A SELECT or CRAN t~ SELECT) if the inservice MASTCR LVL CONT level

' signal failed low, and return level control to auto.

4.3 _If the START-UP LEVEL CONTROL fails transfer valve and pump control to manual.

4.4 If a feedwater input signal fails downscale transfer Tevel-control to the START UP LEVEL CONTROL (single element) with either A or B feedpump selected for_ auto Control.

NOTE 4.5 Main Steam Line flooding occurs at

+118 inches as indicated.on the upset range.-

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4.5 In the event that the. RPV level increases to +90 inches-close the MSIVs, terminate all RPV feeds and ENSURE tne reactor has scrammed.

4.6 If during the transient RPV level reaches +118, ENSURE.

'tKat the steam. lines for the Main Turbine, RFP,.HPCI, il and RCIC Turbines are drained. prior to operation of these_ components.

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CAUTION 4.7 If the MSLs were flooded,1 delay the start:of.'MPCI-and RCICEuntil RPV level decreases to between Level 2 and' Level 3 to maximise the draining-of i

.the steam supply lines.

4.7' t HPCI and RCIC, as necessary, to maintain vessel 7

between Level 4 and Level 7.

5.0 DISCUSSIOR 5.1 A loss of the control signal to the reactor feed pump

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turoine will lock the reactor f eed pump at the speed level' demand prior to the control signal failure.

With restoration of the control signal a manual reset on-C651C is necessary to restore automatic operation.

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LOP-AB.23-117(Q)-

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o OP-AB.23-117(Q) i

5. 2 - Loss of teedwater flow signal in. Master Level Control.

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.a.

The loss:of a single-feedwater-flow input will result in an increase in the reactor vessel water 1evel which may-cause the Main Turbine and RFP 1

Turbines to trip.

b.

The total loss of the feedwater flow signal-input would result in an increase in the reactor vessel water level which would cause the Main Turbine and RFP Turbines to trip.

5.3 Loss nf the reactor vessel water level signal input to the mcster level controller would result. in a reactor water level. increase which will~ trip the Main Turbine and RFP Turbines.

5,4 A highLlevel condition in the RCIC and HPCI steam supply drain pots will cause the respective overhead turbine trouble alarms to annunciate in the control room.

When these alarms clear the steam supply lines should be free of condensate if vessel level exceeded

+118 inches during the high level transient.

0 5.5 Following'a reactor scram the SETPCINT SETDOWN logic will automatically lower the level setpoint to prevent a vessel overfeed.

The logic can be reset when L

L the scram signal is cleared.

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OP-AB.22-Il7(Q)-

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