ML20087E615

Rev a to Work Spec WS-93-0004, Reevaluation of Plant Protection Sys Setpoints

ML20087E615
Person / Time
Site:	Fort Saint Vrain
Issue date:	02/17/1984
From:	Flower M PUBLIC SERVICE CO. OF COLORADO
To:
Shared Package
ML20087E598	List: ML20087E594 ML20087E615
References
TAC-47416, WS-93-0004, WS-93-4, NUDOCS 8403160233
Download: ML20087E615 (17)

Text

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PUBLIC SERVICE COMPANY OF COLORADO o,

. FORT ST. VRAIN NUCLEAR GENERATING STAT!ON ,

NO. WS 03-0@4 SPECIFICATION COVER SHEET PAGE 1 OF 17 Work Specification: Reevaluation of Plant Protection System PLANT ITEM NO'S.

Set Points TABLE OF CONTENTS SECTION HEADING PAGE 1.0 PURPOSE 2 2.0 SCOPE OF WORK  ?

3.6 WORK TO BE PERFORitrD 1 4.0 GENERAL INSTRUCTIONS C l

5.0 TERitS AND CONDITIONS 7 i

l PAGE SCHEDULE l i

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ATTACHMENTS Attachment 1 - Plant Protection System Set Points to be Evaluated - Page 9 I Attachment 2 - Surveillance Requirements to be Evaluated - Dage la ISSUE

SUMMARY

l ISSUE PREPARED BY APPROVED BY DATE BASIS FOR REVISION R'I REV EW A Mike Flower Q2 d,bjif(,m M.91,% 2.- 17-84 Initial Issue l, f g -

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PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION NO. US-93-0004 l SPECIFICATION CONTINUATION SHEET PAGE 2 OF 17 I

PLANT PROTECTION SYSTEM SET POINT WORK SPECIFICATION 1.0 PURPOSE 10CFR50 Appendix A (General Design Criteria for Nuclear Power Plants) Criterion 13, " Instrumentation and Control", requires that instrumentation be provided to monitor variables and systems important to safety and that controls be provided to maintain these variables and systems within prescribed operating ranges.

NRC Regulatory Guide 1.105 " Instrument Set Points" describes a method acceptable to the NRC staff for complying with the Commission's regulations to ensure that the instrument set points in systems important to safety initially are in and remain within the specified limits.

In July, 1983, the NRC informed PSC that ISA Stanoard 567.04-1982, " Set Points for Nuclear Safety-Related Instrumentation Used in Nuclear Power Plants" would be endorsed by an NRC Regulatory Guide in the near future and :hould be used as a guide by PSC in reevaluating the Plant Protection System set points for Fort St.

Vrain. Revision 2 to Regulatory Guide 1.105, endorsing ISA Standard 567.04-1982, was issued "for comment" in December, 1981.

This work specification provides guidarce for the derivation of -

Plant Protection System (PPS) " Trip Set Points" and " Allowable Values" to ensure compliance with the intent of applicable regulatory requirements.

PSC desires, additionally, to investigate the drift characteristics of the PPS instrument loops to evaluate the propriety of existing surveillance frequency requirements.

Should instrument drift history indicate that longer surveillance intervals may be acceptable, a statistical evaluation would be performed to justify this position.

2.0 SCOPE OF WORK The scope of work includes the following:

a) Derivation of " Trip Set Poin't" and " Allowable Value" points for the PPS instrument loops shown in Attachment 1.

b) Evaluation of existing PPS instrument surveillance frequency requirements based on historical drift data and other documented surveillance testing. Recommendation for new longer surveillance intervals wherever warranted based upon a statistical evaluation of drift experienced to date. See Attachment 2 for surveillances to be evaluated.

REVISION

, PUBLIC S E FiV I C E COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION

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NO. WS-93-0004 SPECIFICATION CONTINUATION SHEET PAGE 3 OF 17 3.0 WORK TO BE PERFORMED A. " Trip Set Point" and " Allowable Value" Derivation

1. Review system drawings to identify the PPS instrument configuration associated with each PPS set point.

(Piping and Instrumentation Diagrams, Control Logic Diagram, Control and Instrumentation Diagrams, Control Schematics, Wiring Diagrams, etc.)

2. Obtain necessary information to document each PPS instrument's operating characteristics with regard to all effects addressed in ISA-S67.04-1982. This may include contacting vendors, searching existing PSC files for documentation, locating previous test data for identical or similar equipment, evaleation of historical data, engineering analysis of equipment characteristics, etc.

3. Field verification of all documentation. This includes verification of location within environmental zones, verification of instrument loop configuration, verification of manufacturer model number, verification of any name plate data with specification information, or any other information portinent to set point calculations.

4. Obtain information and document PSC Test Equipment Accuracy for all test equipment. used in PPS set point calibrations. The test equipment used during each surveillance must be determined and manufacturers specifications identified for that test equipment.

5. Prepare PPS Set Point " Allowable Value" and " Trip Set Point" calculations based on guidelines established in NRC Regulatory Guide 1.105, ISA-567.04-1982, and the General Instructions contained within this work specification.

6. Prepare a Set Point Calculation File consisting of the following:

a) Cover Page identifying the set point being evaluated, components in the instrument loop, reference to drawings showing the instrument loop, safety limit or analysis value, allcwable value, trip set point. l REVISION

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NO. ilS-93-000d l ,

SPECIFICATION CONTINUATION SHEET PAGE 4 OF 17 b) Analysis Value Page (to be supplied by GA Technologies). This will identify the analysis value as referenced in the Geaeral Instructions of this work specification. It will contain the analysis value, a brief description of how it was determined, information on all considerations such as transient overshoot or time response that are identified in the ISA Standard and included in the analysis to determine the analysis value, c) Set Point Calculation Pages showing the step-by-step calculations used in deriving the " Allowable Value" and " Trip Set Point" in accordance with NRC Regulatory Guide 1.105, ISA-SP67.04 1982, and the

< General Instructior.s of this work specification.

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d) Independent Review Page documenting independent I verification of set point calculations completed i by a qualified individual, e) Supporting Documentation consisting of the necessary documentation to verify every item of .

information used in the calculation.

NOTE: It is PSC's intention to maintain portions of this file as a design document. For this reason, thoroughness and consistency of format will be required.

7. Inform PSC after the completion of any PPS set point calculation that appears to be vore restrictive than set points presently used (as identified in PSC's Master Set Point List) such that action may be initiated to justify a less restrictive analysis value or other appropriate acticn may be taken. This would be in the form of a letter to PSC explaining the problem set point, a copy o' the calculation, and any recommendations to alleviate the problem.

8. Prepare a detailed work completion schedule for presentation to the NRC as a basis for the program completion commitment date. It may be necessary to accompany PSC to a meeting with the NRC regional office to present the methodology used in the Set Point Reevaluation Program and discuss the schedule for completing the work.

R EVISION ,__

PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION NO. WS-93-0004 SPECIFICATION CONTINUATION SHEET ' PAGE 5 OF 17

9. Recommend a work completion tracking method including bi-weekly progress reports to PSC and periodic progress and problem solving meetings with PSC.

B. " Surveillance Frequency" Evaluation

1) Obtain the actual instrument drift data from PPS surveillances and other appropriate surveillance data over the past 2 years. If a definite trend is not identifiable, a larger sample size will be required.

2) Analyze the drif t trends of the PPS instruments to determine which PPS instruments may display sufficiently low drif t characteristics to justi fy extending their surveillance intervals. Criteria to consider would include amount of drift experienced,.

direction of drift, frequency that allowable value is exceeded, other related instrumentation (i.e. second line of defense), etc. (NUREG-1024 may be of assistance)

3) Prepare a for:nal evaluation based on recognized statistical methods to justify extending the surveillance interval for those PPS Instruments which would be acceptable with a longer interval.

4) Provide a recommended interval for the analyzed surveillances based on your evaluation for submittal to the NRC.

4.0 GENERAL. INSTRUCTIONS It is necessary to make some clarifications and modifications to the step-by-step procedure outlined in ISA-567.04-1982 for use at FSV. The following general instructions were adopted at a meeting with the NRC:

A. ISA-S67.04-1982, " Set Points For Nuclear Safety-Related Instrumentation Used in Nuclear Power Plants", will be used as guidance in derivation of PPS Instrument Set Points.

Most accident analyses at FSV do not identify a " Safety Limit" as defined in this standard and, therefore, an alternate approach is necessary. Other possible conflicting areas exist where the accident analysis already compensates for items specified within the standard.

B. The following simplified model depicts the method discussed with the NRC as an acceptable method for calculating PPS REVISION

PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION NO. !!S-93-0004 SPECIFICATION CONTINUATION SHEET PAGE 6 OF 17 Instrument Set Points. An explanation of considerations for each area follows:

AREA SAFETY LIMIT

'l ANALYSIS VALUE 2

ALLOWABLE VALUE 3

TRIP SET POINT Area 1 is bounded by the " Safety Limit" and the " Analysis Value". The " Analysis Value" exists which represents the set point value used in the computer code for the accident or the associated equipment limit provided by other testing and has been determined to bg satisfactory. The code may simulate the transient oveMhoot, time response of components, etc. and therefore this " Analysis Value" may not represent the most limiting value the set point parameter may achieve. The accident analysis does, however, determine that the consequences of reaching this most limiting value are acceptable. An additional unquantified margin of safety will most likely exist between the most limiting value reached in the computer analysis and the actual safety limit. Developing a " Safety Limit" for derivation of the set point, as inferred in the ISA Standard, is not required.

GA Technologies will provide the information concerning this analysis value to be added to the set point calculation package.

Area 2 consists of allowances between the " Analysis Value" and the " Allowable Value" for the following items if not already considered in the ac'cident analysis (i.e., transient overshoot, time response of temperature instruments, are often included in the computer code used for the accident l analysis): l

1. Accuracy (including drift) of components not tested when the monthly / quarterly surveillance is performed.

(Same surteillance used to obtain "As Found/As Left" data.)

REVISION

PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GE 9ERATING STATION NO. US-93-0004 SPECIFICATION CONTINUATION SHEET PAGE 7 op 17

2. Test Equipment Accuracy.

3. Process Measurement Accuracy (if not included in computer code for accident analysis).

4. Transient Overshoot (if not included in computer code for accident analysis).

5. Time Response (if not included in computer code for accidentanalysis).

6. Environmental Effects (if not included in computer code for accident analysis).

Area 3 consists of allowances between the " Allowable Value" and the " Trip Set Point" to account for instrument d.ift.

The instrument drift considered here is determined from the s "As Found/As Left" data taken during monthly or quarterly surveillances, vendor specifications, and other test data available. The drift is for the portion of the instrument channel that is tested during the surveillance.

The area below the " Trip Set Point" will be controlled by the surveillance procedures and has no direct implication concerning the safety of any set point providing the set point is always left below the " Trip Set Point" following surveillances.

The methodology outlined in ISA-S67.04-1982, NRC Regulatory Guide 1.105, and the General Instructions of this work specification are intended to give detailed instructions in the derivation of PPS " Trip Set Points" and " Allowable Values". Due to the complexity of some instrumentation involved in the Plant Protection System, it may not be possible to comply verbatim with this methodology. If deviation is unavoidable, it must also be accompanied with a detailed justification for the alternate approach concluding that the intent of the aforementioned guides has been satisfied.

5.0 TERMS AND CONDITIONS A. Work Manager for this project will be Mr. Mike Flower (303) 571-7372.

1 REVISION

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NO. US-93-0004 i SPECIFICATION CONTINUATION SHEET 8

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i l B. Invoices should be addressed to:

Mr. H. L. Brey, Manager Nuclear Engineering Division Public Service Company of Colorado 2420 West 26th Avenue, Suite 100-D Denver, Colorado 80211 .

C. All coordination in this effort involving GA Technologies will be accomplished through PSC personnel.

D. A Quality Assurance Program which complies with 10CFR50 Appendix B and ANSI N45.2 must be in effect and implemented throughout the course of this work. Approval of this QA program is required prior to awarding the contract.

REVISION

PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VR AIN NUCLEAR GENERATING STATION NO. !!S-93-0004

. SPECIFICATION CONTINUATION SHEET PAGE 9 OF 17 FORM 344 4083 ATTACHMENT 1 Plant Protection System Set Points to be Evaluated REACTOR SCRAM FUNCTIONAL UNIT PRESENT TECHNICAL SPECIFICATION TRIP SETTING Linear Channel-High, 1140% power Channels 3, 4, 5 Linear Channel-High, -<140% power Channels 6, 7, 8 Wide Range Channel- 5 DPM Rate of Neutron Flux Rise High Channel 3, 4, 5 Primary Coolant Moist'ure High Level Monitor <67 degrees F Dewpoint Loop Monitor 127degreesFDewpoint Reheat Steam Temperature $1075 degrees F-

-High Primary Coolant Pressure 150 psig below

- Low normal, load programmed Primary Coolant Pressure <7.5% above

- High iiormal rated, load programmed Hot Reheat Header >35 psig Pressure - Low Main Steam Pressure >1500 psig

- Low Plant Electrical (to be System-Loss provided) s High Reactor Building -<326 degrees F Temperature (Pipe Cavity)

REVISION

PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION no,WS-93-0004 SPECIFICATION CONTINUATION SHEET PAGE 10 OF 17 ATTACHMENT 1 Plant Protection System Set Points to be Evaluated LOOP SHUTDOWN FUNCTIONAL UNIT PRESENT TECHNICAL SPECIFICATION TRIP SETTING i

Steam Pipe Rupture Under i PCRV, Loop 1 19 v. dc. l Steam Pipe Rupture Under PCRV, Loop 2 19 v. dc.

Steam Pipe Rupture, North Pipe Cavity Loop 1 19 v. dc.

Steam Pipe Rupture, South Pipe Cavity Loop 1 19 v. dc.

Steam Pipe Rupture, North Pipe Cavity Loop 2 19 v. dc.

Steam Pipe Rupture, South Pipe Cavity Loop 2 1 9 v. dc.

High Pressure, Pipe 12.5" w.g.

Cavity ,

High Temperature, Pipe 1130degreesF Cavity High Pressure, Under -<2.5" w.g.

PCRV High Temperature, Under 1130 degrees F PCRV High Reheat Header <5 mr/hr 'Atove Activity, Loop 1 Tackground High Reheat Header <5 mr/hr Above Activity, Loop 2 Tackground Low Superheat Header ->800 degrecs F Temperature, Loop 1 R EVISION

PUBLIC SERVICE COMPANY OF COLORADO 1 FORT ST. VRAIN NUCLEAR GENERATING STATION NO.WS-93-0004 SPECIFICATION CONTINUATION SHEET PAGE 11 OF 17 ATTACHMENT 1 Plant Protection System Set Points to be Evaluated LOOP SHUTDOWN (continued)

FUNCTIONAL UNIT PRESENT TECHNICAL SPECIFICATION TRIP SETTINGS Low Superheat Header ->800 degrees F Temperature, Loop 2 High Differential Temp. <50 degrees F Between Loop 1 and Loop 2 6

i REVISIOtJ

PUBLIC SE9VICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION NO. US-93-0004 SPECIFICATION CONTINUATION SHEET PAGE 12 OF 17 ATTACHMENT 1 Plant Protection System Set Points to be Evaluated CIRCULATOR TRIP FUNCTIONAL UNIT PRESENT TECHNICAL SPECIFICATION TRIP SETTINGS Circulator Speed-Low 1910 rpm Below Normal as Programmed by FW Flow Loop 1, Fixed Feedwater 20% of Rated Flow-Low (Both Full Load Circulators)

Loop 2, Fixed Feedwater 20% of Rated Flow-Low (Both Full Load Circulators)

Loss of Circulator >475 psid Bearing Water Circulator Penetration 1810 psig Trouble Circulator Drain -->5 psid Malfunction Circulator Speed-High, 111,000 rpm Steam Circulator Seal >-10"H20, or Malfunction 180"H20

. Circulator Speed-High, 18,800 rpm Water REVISION

PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION NO.HS-93-0004 SPECIFICATION CONTINUATION SHEET PAGE 13 OF 17 a

ATTACHMENT 1 Plant Protecticn System Set Points to be Evaluated R0D WITHDRAWAL PR0HIBIT

. FUNCTIONAL UNIT PRESENT TECHNICAL SPECIFICATION TRIP SETTINGS Startup Channel-Low >2.5 cps count rate Linear Channel-Low power RWP ->5%

(Channels 3, 4 and 5)

I Linear Channel-tow power RWP ~>5% j (Channels 6, 7 and 8) j i

Linear Channel-High power RWP <30% -

(Channels 3, 4 and 5)

Linear Channel-High power RWP <30% -

(Channels 6, 7 and 8) l i

REVISION

. PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION NO. WS-93-0004 SPECIFICATION CONTINUATION SHEET PAGE 14 OF 17 ATTACHMENT 2 Surveillance Requirements to be Evaluated Surveillance Channel Des'cription Method SR 5.4.1.2.lb-m Linear Power Channel Test Internal test signal to verify trips and alarms SR 5.4.1.1.4b-m Linear Power Channel Test Internal test signal to verify trips and alarms SR 5.4.1.1.6e-m Primary Coolant Moisture Test Verify that each of the eight (all channels) monitors will alarm on low and high sample flow SR 5.4.1.1.7a-m Primary Coolant Moisture Test Trip one high level, one low (High Level Channels) level channel, pulse another low level channel SR 5.4.1.1.8b-m Reheat Steam Temperature Test Trip channel, verify alarms and indications. Internal test signal to verify trips and alarms SR 5.4.1.1.9b-m Primary Coolant Pressure Test Trip channel, internal test signal to verify trips and alarms SR 5.4.1.1.10b-m Circulator In.let Temperature Trip channel, internal test l Test signal to verify trips and y alarms SR 5.4.1.1.11a-m Hot Reheat Header Pressure Reduce pressure at sensor to Test trip channel, verify alarms and indications SR 5.4.1.1.12a-m Main Steam Pressure Test Reduce pressure at sensor to trip channel, verify alarms and indications SR 5.4.1.1.13a-m Two Loop Trouble Test Special test module used to trip channel by energizing e' of four appropriate pairs e two-loop trouble relays REVISION

. PUBLIC SERVICE COMPANY OF COLORADO

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, no,US-93-0004 SPECIFICATION CONTINUATION SHEET PAGE 15 OF 17 ATTACHMENT 2 Surveillance Requirements to be Evaluated Surveillance Channel Description Method SR 5.4.1.1.14a-m Plant 480 V Power Loss Test Trip each channel by applying simulated loss of voltage signal, verify alarms and indications SR 5.4.1.1.15b-m High Reactor Building Trip channel, verify alarms Temperature (Pipe Cavity) and indications. Internal test Test signal to verify trips and alarms SR 5.4.1.2.lb-m Steam Pipe Rupture (Pipe Pulse test one temperature and Cavity) Test pressure channel with another temperature and pressure channel tripped, while simul-taneously having two ultrasonic l channals tripped j l

SR 5.4.1.2.1d-m Steam Pipe Ruptu're Pressure switch actuated by ,

(Pipe Cavity) Test pressure applied at sensor ,

I SR 5.4.1.2.le-m Steam Pipe Rupture Temperature swiuch actuated by (Pipe Cavity) Test heat applied at sensor. l SR 5.4.1.2.lf-m Steam Pipe Rupture Internal test signal to adjust (Pipe Cavity) Test ultrasonic trip. ,

i SR 5.4.1.2.lg-m Steam Pipe Rupture Trip test signal solenoid valves ;

(Pipe Cavity) Test to verify loop integrity j i

SR 5.4.1.2.2b-m Steam Pipe Rupture Pulse test one temperature and i (Under DCRV) Test pressure channel with another r temperature and pressure channel j tripped while simultanecusly -

having two ultrasonic channels l tripped SR 5.4.1.2.2d-m Steam Pipe Rupture Pressure switch actuated by (Under PCRV) Test prssure applied at sensor SR 5.4.1.2.2e-m Steam Pipe Rupture Temperature switch actuated by  !

(Under PCRV) Test heat applied at sensor )

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REVISION

. PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION NO)lS-93-0004 SPECIFICATION CONTINUATION SHEET PAGE 16 Op 17 ATTACHMENT 2 Surveillance Requirements to be Evaluated Surveillance Channel Description Method  ;

SR 5.4.1.2.2f-m Steam Pipe Rupture Internal test signal to adjust (Under PCRV) Test ultrasonic trip SR 5.4.1.2.2 9-m Steam Pipe Rupture Trip test signal solenoid valves l (Under PCRV) Test veri fy to loop integrity i SR 5.4.1.2.3a-m Circulator 1A and 1B Pulse test and verify proper Tripped Test indications SR 5.4.1.2.4a-m Circulator 1C and ID Pulse test and verify proper Tripped Test indications I SR 5.4.1.2.5a-m Steam Generator Penetration Pressure switches actuated by i Pressure Test pressure applied SR 5.4.1.2.5b-m Steam Generator Penetration Pulse test each channel with '

Pressure Test another channel tripped and verify proper indications SR 5.4.1.2.6b-m Reheat Header Activity Pulse test each channel with l Test another channel tripped and verify proper indications SR 5.4.1.2.7c-m Superheat Header Temperature Pulse test one channel with Test another channel tripped and verify proper indications SR 5.4.1.2.8a-m Primary Coolant Moisture Trip each channel, verify proper (Low Level Channels) Test- indications SR 5.4.1.2.8b-m Primary Coolant Moisture Trip each channel, pulse test {

(Low Level Channels) Test other loop to check loop identi-fication SR 5.4.1.2.9a-m Primary Coolant Pressure Pulse test one channel with Test another channel tripped and 1 verify proper indications, both '

channels h

SR 5.4.1.3.lb-m Circulator Speed-Steam and Internal test signal to verify [

Water Test trip settings and indicators I

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REVISION

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PUBLIC SERVICE COMPANY OF COLORADO S * -' FORT ST. VRAIN NUCLEAR GENERATING STATION NO. 11S-93-0004 SPECIFICATION CONTINUATION SHEET PAGE 17 OF 17 l

ATTACHMENT 2 Surveillance Requirements to be Evaluated Surveillance Channel Description Method SR 5.4.1.3.1c-m Circulator Speed-Steam and Pulse test one channel with Water Test another channel tripped, and verify proper indications SR 5.4.1.3.2b-m Feedwater Flow Test Internal test signal to verify trip setting and indications SR 5.4.1.3.2c-m Feedwater Flow Test Pulse test one channel with another channel tripped, and verify proper indications ,

SR 5.4.1.3.3b-m Circulator Bearing Water Pulse test one channel with Pressure Test another channel tripped, and verify proper indications SR 5.4.1.3.4a-m Circulator Penetration Pressure Pressure switches actuated by Test pressure applied SR 5.4.1.3.4b-m Circulator Penetration Pressure Pulse test one channel with Test another channel tripped, and verify proper indications SR 6.4.1.3.5b-m Circulator Drain Pressure Pulse test one channel with Test another channel tripped, and verify proper indications SR 5.4.1.3.6b-m Circulator Seal Malfunction Pulse test one channel with another channel tripped, and verify proper indications SR 5.4.1.4.2b-m Linear Channel Test Internal test signal to verify trips and alarms REVISION -