ML20195G117

Tech Spec Improvements Based on Risk Significance

ML20195G117
Person / Time
Site:	Comanche Peak
Issue date:	11/17/1988
From:	TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:
Shared Package
ML20195G114	List: ML20195G108 ML20195G117
References
NUDOCS 8811230203
Download: ML20195G117 (35)

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i i CIMANWE PEAX WIT 1

HOWICAL SPECIFICATION DtPICVDtDf!5

, based on RISK SIM IFICAN 2 i.

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TABLE OF CONTENTS Section Zagg, 1.0 PURPOSE 1 2.0 SCOPE 1 3.0 METHOD of ANALYSIS 1 4.0 ANALYSIS RESULTS 3 4.1 Taitiating Events Analysis 3 4.2 Systems Analysis 3 4.3 Risk Evaluation  ?

5.0 CONCLUSION

S 6 REFERENCES 7 APPENDIX A A-1

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1.0 PURPOSE The purpose of this study is to evaluate the risk impact on Comanche l Peak Unit 1 (CP1) of a number of identified systems, components or requirements from a Probabilistic Risk Assessment (FRA) point of view. If the risk impact of these systems is found to be insignificant, their Technical Specifications may then be removed from the Technical Specifications docuaent and relocated to another  !

controlled document.

Since CPI does not currently have a PRA available, a detailed quantitative evaluation cannot be performed. Instead, a qualitative f study is done co enaure that the Westinghouse Owners Group MERITS Program on "Technical Specification Improvements for Nuclear Power P

Reactors" (reference 1) applies to Comanche Peak Unit 1.

1.0 SCOPE

, The following systems, ecmponents or requirements are included in this analysis:

I o Meteorological Monitoring System o Seismic Monitoring System 4

o Loose Parts Monitoring System o Incore Instrumentation System j

o Steam Generators a

l o Sealed Source Contamination Each of these is qualitatively analyzed to determine whether or not it is a significant risk contributor.

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3.0 METHOD OF ANALYSIS Since there is no PRA available for CP1, a qualitative approach had i to be adopted for this study to ensure that the MERITS Program is  ;

applicable to cpl. In general, the systems listed in the previous section are not included in typical PRA models and, therefore, do not appear to significan.ty impact the plant risk. However, they could indirectly affect the operations of other safety systems which are required for safe shutdown of the plant following any design basis accident. For this reason, the following steps are taken to address any potential effects that any of the systems under consideration may have en pl, ant risk.

Sten 1. Each system is examined to daeermine whether or not its function has any impact on initiating event frequencies or could initiate other events not already defined.

Step la Each system is examined to determine whether or l- not its function has any impact on availability of the

, safety systems required for safe shutdown of the plant.

I j If a syrtem failure is shown to have no potential for accident initiation or any impact on the availability of the safety systems,

it is then judged to be an insignificant risk contributor. The risk

! evaluation process is shown in Figure 1.

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4.0 ANALYSIS RESULTS 4.1 INITIATING EVENT ANALYSIS A detailed study was performed to develop a PRA-identified initiating event list for cpl. Initiating events are defined as events (operator actions, system failures, component malfunctions, external conditions) which would interrupt normal plant operations, and which when coupled with subsequent failures could lead to a degraded core condition. Table 1 summ r!ses the initiating events identified for cpl. Since there are a large number of potential initiating events, the events are grouped into categories, each of which encompasses events with similar plant responses.

4.2 SYSTEMS ANALYSIS A study was performed to identify all critical systems that are required to function for the safe shutdown of CP1 following any of the events listed in Table 1.

The CP1 critical systems identified for PRA studies are listed in Table 2. The critical system list was prepared based on the review of CP1 Emergenew Response Cuidelines (ERGS) and Emergency Operating Procedures (EOPs). This list was then compared with other generic PRAs for completeness.

l 4.3 RISK EVALUATION t

The risk evaluation logic diagram (Figure 1 ) was used to evaluate each of the systems described in section 2 and Appendix A. The results of the evaluation are given below.

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o Meteorological Monitorina System This system provides the meteorological data for estimating potential annual radiation doses to the public as a result of routine or accidental release of radioactive nr.terials to the atmoaphore. Its failure does not impact any of the initiating events listed in Table 1 or any of the critical systems listed in Table 2. Therefore, the meteorological monitoring system is considered to be an insinnificant risk contributor.

o Seismic Monitorine System In case of a severe seismic event, the seismic monitoring system provides input to control room alarms which direct operators to trip the reactor. For this reason, this system is determined to have a potential risk impact on plant operation.

i However, the design La:is ground acceleration levJl at the Comenche Peak dice is 0.12g. At this acceleration level, the safety-related systems and components are not expected to malfunction since they are designed to withstand the earthquakes of much higher magnitudes. The results of other PRAs also show that the risk contributions from earthquakes of 0.12g or less are va:f small. Therefore, the Seismic Monitoring System is considered to be an insinnificant risk contributor, o Loose Parts Monitorine System

! This system is used to detect and communicate unusual noises that indicate a metallic loose part in the Reactor Coolant System (RCS) in order to avoid or mitigate any damage to the RCS. Its failure to function does not affect any of the initiating events listed in Table 1. However, the loose parts may damage the RCS pressure boundary (System No. 6 Table 2).

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O Based en nuciocr plent epsroting experience end tha results of other PRAs, the loose parts are assumed not to damage the RCS.

Therefore, the Loose Parts Monitoring System is considered to be j en insinnificant risk contributor.

o Incore Instrumentation System The system is used to provide information on the neutros. flux l

distribution and fuel assembly outlet temperatures at molected -

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core locations. The system provides a means for acquiring data l

only and performs no operational plant control. This system does  !

not impact any of the initiating events listed in Table 1 or any of the critical systems listed in Table 2. Therefore, the Incore Instrumentation System is considered to be an insinnificant risk contributor.

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o Steam Generators i

The Cteam Cenerators transfer heat from the Reactor Coolant System (RCS) to the secondary side while maintaining a radiological barrier between the two systems. The specification for this  !

system provides inspection requirements for the steam generator [

I tubes.

I Steam generators have the potential for a Steam Generator Tube  ;

I Rupture accident (Initiatir.g Event No.12, Table 1). They may !

also impact the RCS (Critical System No. 6, Table 2). Therefore, f i

following the logic shown in Figure 1. the steam generaters could  !

be significant risk contributors, i The risk contributions from the steam generators are a degradation of either heat removal capabilities or RCS and secondary pressure boundary integrity. The assurance of heat rer. oval capability of the steam generators is already provided by the liutting condition i for operation contained in Technical Specifications 3/4.4.1, 3/4.7.1.1 and 3/4.7.1.7. The primary means of ensuring RCS l

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, pressuro bound::ry integrity is ths cugAnted inservico inspsetion program which contains corrective action requirements and has to be performed in plant mode 5.ar 6 due to physical constraints.

During plant modes 1 through 4 the integrity of the '. team generator tubes is monitore.1 by the surveillance requirements of Technical Specification 3/4.4.6.2. Operation with tube leakage is 1

J restricted by this same Technical Specification. Based on the preceding discussion, the steam generator limiting condition for operation can be removed with no increased threat to the safe i operation of the plant. The surveillance requirements that l describe the augmented inservice inspection program are retained in the Technical Speciffsation but are relocated to Specification 4.0.6. Therefore, this removal of the steam generator limiting condition for operation and the relocation of the surveillance I requirements are considered to have an insinnificant risk contribution.

o Sealed Source Contamination l

i The Sealed Source Contamination is not a system or equipment that prevents any abnormal plant operating condition or mititates the

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effect of any design basis event. This is rather a set of l

requirements which ensures that the leakaga from special nuclear l

material sources will not exceed the limits in 10CFR Section j 70.39. These requiremsnes do not impact any of the initiating i events listed in Table 1 or any of the critical systems listed in Table 2. Therefore, the Sealed Source contamination is considered to be an insinnificant risk contributor.

t 5.0 conclusions i

3 The res..its of this qualitative study indicate that the following systems, subsystems or components are not significant risk i contributors:

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O o Meteorlogical Monitoring System o Seismic Monitoring System o Loose Parts Monitoring System o Incore Instrumentation System o Sealed Source Contamination The Steam Generators could have some impact on plant risk. However, their limiting conditions for operation are included in other Technical Specifications, and their surve<ilance requirements will be included in Section 4.0.6.

It is concitded that the MERITS Program applios to CP1 for the above systems. The Technical Specification taquirements for these systems may, therefore, be relocated to another controlled document without affecting the plant risk.

6.0 References 1.0 MERITS Program, Phase II, Tasks, "Methodically Engineered, Restructured and Improved, Technical Specifications", dated November 1987.

2.0 Comanche Peak Steam Electric Station Unit 1, "Draft Technical Specifications", o.ted August 1988. ,

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Tcblo 1. FRA-id:ntifica Initicting Ev:nt Cotogorios for CP1 Init!ating Event Category . Designator

1. Reactor trip RT

2. RCS overpressurination R0

3. Inadvertent SIAS generation IS

4. Main steamline break MB

5. Loss of normal feedwater-MTV available MF

6. Loss of normal feedwater No MTV available NF

7. Main feedline break MB

8. Large Break LOCA LL

9. Medium Break LOCA ML

10. Small Break LOCA SL

11. Reactor Vessel LOCA RL

12. Steam Generator Tube Ruptuts SG

13. Interfacing System LOCA IL

14. Loss of a DC bus DC

15. Loss of the HVAC System HV

16. Loss of offsite power AC

17. Loss of a non-vital AC bus ACN

18. Loss of a safeguards AC bus ACS

19. Loss of Component Cooling Vater System CCV

20. Loss of Station Servico Water System SV

21. Loss of the Instrument Air System IA

22. External events (seismic, fire, vind) EE 8

Tablo 2. CP1 Critical Syst:3 List for PRA Studios System Designator

1. Auxiliary Feedwater System AFV

2. Containment Spray System CS

3. Residual Heat Removal System RHR

4. Chemical & Volume Control System CV

5. Emergency Core cooling System EC (including Safety Injeciton)

6. Reactor Coolant System RCS

7. Component Cooling Vater System CW

8. Service Water System SW

9. Electric Power System EP

10. Engineered Safety Features Actuation ES System

11. Instrument Air System IA

12. HVAC Systems HV i (including Safety Chilled Vater)

13. Main Feedwater System HFV

14. Main Steam System MS 1

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Appendix A SYSTEM DESCRIPTIO!iS O

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• .o A.1 METNOROLOGICALMONITORING. SYSTEM  ;

A c t.1 SYSTEM FUNCTION -

. Meteorological instrumentation is provided within the 1

plant to ensure that sufficient meteorological data are availaide for esticating potential ani. cal radiation doses to the public as a resbIt. Of routine or accidental 3

release of radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent with the recommendations for Regulatory Guide 1.23, "Onsite

, Meteorological Programs," February 1972. The j meteorological monitoring system does not nood to be environmentally or seismically qualiflod.

4 A.1.2 SYSTEli_$GTFIGURATION The meterological monitoring system consists of all the instruments, structures, recorders, and computer inputs used to gather, transmit, display, or store meteorological date gathered from either the primary or backup meteorological systems. The primary system consists of a 200 foot instrument tower and all of its equipment. Similarly, the backup system consists of a 33 foot instrument tower with its equipment. Both systems share an environmentally controlled meteorological instrument building. In addition, motoorological data from both towers is transmitted through tolophone cable to contol room computers. Digital readout of both 15-minuto- and hourly-averaged data are generated by the computers frem the meteorological data. Failure of the aspirator motor, which insures the accuracy of temperaturo paramotors during strong daytime heating conditions, is indicated as an equivalent of a power failure on Panel P2500 in the control room. If an A-2

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. aspirator motor' fails, the monitored temperature data will be suspect since the Resistance Temperature Detectors (RTD) cannot be fully aspirated.

The meteorological monitoring system is to be protected by grounding and is provided with surge protection devices on the incoming power lines and signal lines in order to minimize the impact of damage due to direct lightning strike on the system or electrical surge in the power lines. The m'eteor J1ogical distribution panel (CPX-ECDPNC-29) is powered from a non-Class lE, 480 V power system. The primary tower multiplexer (CPS-MTMSPR-01T), the elevator control panel (CPX-MTMSPR-01C), meteorological tower instrument building distribution panel, and the backup tower multiplexer (CPX-MTMSBU-O1T) are powered from a non-Class 1E, 120 V AC powered system. The meteorological receiver cabinet is powered from a Class 1E uninterruptible power cupply system that trips upon receipt of an Safety Injection (SI) signal. The meterological instrument pa r.o1 (CPX-ECCPRCV-05) is powered from two separated sources, a non-Class 1E, 120 V AC power system and a class 1E uninterrruptible power supply system that trips upon receipt of an SI signa 1.,

A.1.3 REOUIRED OPERATOR ACTIONS There are no requirements for automatic actions or inital operator actions. However, subsequent operator actions are specified for two different scenarios:

Igrpado/Hioh Winds observed or reported aonroaching__

the olant site.

a. Announce on the Gaitronics the direction from which the tornado /high wind is approaching.

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b. Refer to EPP-201, "Assessment of Emergency Action Levels and Plant Activation."

c. If time partits, dispatch operators to ensure  !

closure of safeguards structures and rooms.

Plant Facility Struck by a Tornado /Hiah Winds I

a. If a reactor trip occurs, proceed per EOP-0.0,  !

"Reactor Trip or Safety Injection." {

b. Refer to EPP-201, "Assessment of Emergency Action t Levels and Plan Activation."

c. When conditions permit, dispatch personnel to check for damage. Operations personnel should .

check safety-related equipment.

d. Assess plant conditions and proceed as directed by ,

the Shift Supervisor.

Aol.4 TECHNICAL SPECIFICATION REOUIREMENTE r l

Technical Specification 3/4.3.3.4 applies to the I meteorological monitoring synten.

Aeolicability: At all times. l Action j

a. With one or more required meteorological monitoring l channels insperable for.nore than 7 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 [

days outlining the cause of the malfunction and the "

plans for restoring the channel (s) to OPERABLE status, j

b. Tne provisions of Specifications 3.0.3 and 3.0.4 are not applicabic.

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' A.2 SEISMIC MONITORING SYSTEM '

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! A.2.1 SYSTEM FUNCTION q

seismic instrumentation is provided within the plant so that, in the case of an earthquake, sufficient data is i

generated to permit verification of the dynamic analysis of the plant and evaluation of the safety of continued

] operation.

. A.2.2 SYSTEM CONFIGURATION s

The seismic monitoring system is primarily a passive system, designed to function only during testing or in i actual earthquake or blast. The seismic monitoring l instrumentation itself is distributed throughout the plant. For those instruments having an alarm or test function, the associated controls will be on a common seismic monitoring cabinet. The seismic monitoring panel j provides for both audible and visual alarm indications.

In addition, any cabinet alarn will in turn trigger a common control board alarm.

The seismic instrumentation provided is specified in accordance with ANSI N18.5-1974, "Earthquake Instrumentation Critoria fo,r Nuclear Power Plants," as recommended by NRC Regulatory Guide 1.12, Revision 1, "Instrumontation for Earthquakes," and comprises the following instruments:

1. A triaxial time history accolorograph, which consists of triaxial accoloration sensors, a seismic trigger, a magnetic tape recorder and controls, and a magnetic playbac< unit. The function of the triaxial time history accelerograph is to measure and permanently record absolute acceleration as a function of time during an earthquake.

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2. A trioxici pOak cccolorogrcph, which 10 d0cigned [

to permanently record peak seismic accelerations i of seismi,c Category I equipment and piping l

3. A passive response spectrum recorder, wnich is  !

designed to permanently record spectral l accelerations corresponding to specified frequencies and located at the foundation of the i Containment Building and the supports of seismic Category I equipment and piping i

4. A response spectrum switch, which is designed to provide a signal for remote, immediate indication l

that any specificd preset spectral acceleration has  !

been exceeded '

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5. A seismic switch, which is designed to provide a i signal for remote immediate indication that a c specified preset acceleration has been exceeded I f

Electrical power is required for the operation of the e

seismic monitoring system. Under normal operating conditions, the seismic cabinet is supplied with AC power .

from a 120 VAC distribution panel. AC power is j transformed to DC in order to provido power to 1 instruments and is available to charge batteries in both j the accelerograph and seismic switch. Installed  !

batteries supply power in the event AC is lost. Battery l

operation can run one accelerograph recorder for 30  !

t minutes. The annunciator battery will supply all lights  !

for about 30 minutes.

k A.2.3 REOUIRED OPERATOR ACTIONS i i

I Control room operatorn are required to take actions in j response to alarm conditions on control Room Annuciator i Panel 1-ALB-2A. The associated alarm procedure, l ALM-0021A, specifics that the operator take tno following f' actions which are precipitated by the probaolo cause of a.. carthquake or soismic disturbance: I j

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.' hpnuciator No. 1.1 (Time History Accelerometer Ocarl Automatic Actions: None.

Initial Operator Actions

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1. Monitor plant parametais to ensure safe operation exists.

Subseauent Operator Actiqh):

1. Refer to ABN-907A, "Acts of Nature", for an earthquake condition.

2. Contact the appropriate department for operation of the seismic recorder instrumentation.

Annunciator No. 2.1 (OBE Seismic Switch Act)

Autenatic Actions:

1. The seismic switch starts operation cf the Time History Accelograph.

Initial Oneration Actions:

1. Monitos' plant parameter (s) to ensure safe operation exists.

Subsecuent Onoration Actions:

1. Refer to ABN-907A, "Acts of Nature," for an earthquake condition.

Annunciator No. 3.1 (75% OBE Seismic Switch Act)

Automatic Actions: None.

Initial Operator Actions:

1. Monitor plant parameters to ensure safe operations exists.

Subsecuent Operator Actions:

1. Refer to ABN-907A, "Acts of Nature," for an earthquake cond!. tion.

2. Contact the appropriate departnent for operation of the seismic recorder instrumentation.

Annunciator No. 4.1 (100% OBE Seismic Switch Act)

Automatic Actions: None.

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Initial Onerator Actional s

1. Monitor plant parameters and place plant in a  !

ao shutdown condition per ABN-907A, "Acts of [

Nature." i

, i Subseauent.Onerator Actionst  ;

1. Contact the appropriate department for operation of the recorder seismic  !

instrumentation. t

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Operator actions specified in ABN-907A, "Acts of Nature," .[

can be summarized as i i

Earthauake monitor alarm, no other indicationst i Automatic actions: None.

Initial Ocarator Actions: None.

Eghteauent Ocarator Actions  !

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a. Refer to Technical Specifications 3/4.3.3.3 Renconse SDectrum Recorder Alert _f amber) Alarm Plus other [

Indications (control room vibration, taDe recorders i runnina, etc.): I Automatic Actionst  !

a. Magnetic type recorders will start when white  :

"Event Indicator" light on SMA-3 is ,

illuminated. '

Initial coerator Actions  !

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a. If the reactor has tripped proceed per E0P-0.0,  ;

"Reactor Trip or Safety Injection." i Subsecuent oeerator Actions f i

a. Dispatch operators to perform a visual  !

inspection of safety related equipment. }

b. Refer to EPP-201, "Assessment of Emergency Action Levels and Plan Activation." fi

c. Notify Results Engineering Department personnel {

to perform EST-710 "Data Retrieval, Analysis o and Reporting following a Scismic Event." j l

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l' Response Spectrum Recorder Alert famber) Alara Plus Other j j Indications (control room vibration, taoe recorders i j runnina. etc. ) . l

] Automatic Actions  !

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3. a. Magnetic tape recorders will start when white t 1 "Ever; Indic1 tor" light on SMA-3 is  !

illuminated.

Initial _Ooarator Actions I

a. Trip the reactor and proceed per E0P-0.0,  ;

j "Reactor Trip or Safety Injection," when plant  !

conditions have stabilized, continue with this  :

4 procedure.  !

Subsemuent Ooerator Actionst i ,

a. Dispatch operators to perform a visual [

inspection of estety related equipment.

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Action Levels and Plan Activation." >

l c. After stabilizing plant conditions per the J

Emergency Procecures proceed to cold shutdown f t

l per Ipo-005A, "plant Shutdown from Hot Standby to cold Shutdown." ,

I A.2.4 TECHNICAL SPECIFICATIONS REOUIREMENTS 4 I J  !

} Technical Specification 3/4.3.3.3 applies to the scismic j monitoring system. l i

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Apolicability: At all times.

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! a. With one or more of the above required seismic l

! monitoring instruments inoperable for more than i i 30 dayr, prepare and submit a Special Report to j 1 the commission pursuant to Specification 6.9.2  !

j wit).in the next 10 days outlining the cause of  !

1 the malfunction and the plans for rostoring the 1

instrument (s) to OPERABLE status.

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! b. The provisions of Specifications 3.0.3 and j 3.0.4 are not applicable.

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A.3 LOOSE PARTS MONITORING SYSTEM A.3.1 SYSTEM FUNCTION The Loose Parts Monitoring System (LPMS) is used to detect and annunciate unusual noises that indicate a metallic loose part in the Reactor Coolant System (RCS).

The system monitors at particular locations and alerts the operator to unusual occurrences via the use of accelerometers (sensors). The system will alarm the presence of unusual' noises above the normal background Jevel present in the plant. Baseline data taken during the initial startup of the plant will allow determination of the appropriate alarm settings for cach channel to be used later as a comparison for loose part detection and analysis. This system is required to be functional during modes 1 and 2 of operation.

. 3. 2 SYSTEM CONFIGURATION The LPMS uses an array of active and passive accelerometers to detect metal to metal impacts within the RCS. The acceleromoters are permanently installed at selected locations where a loose part would tend to collect or impact. Two ace,elerometers are located at each of the following locations:

o Reactor Pressure Vessel - Upper Head Region (both active) o Reactor Pressure Vessel - Lower Region (both

! active) i o All Steam Generators - Primary Coolant Inlet ,

Region (active and passive)

One accelerometer is located at each of the following locations:

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.' o All Rtactor CcolLnt Pumpa (pnecivo) o All Steam Generators - Upper Region (passive)

The accelerometers are transducers that convert vibratory accelerations into electrical signals that are amplified, filtered, and conditioned to accentuate the frequency band known by measurement to ccrrespond to metal to metal impacts. The active channels are monitored continuously for detection of loose parts. Tne passive channels are used for diagnostics (location determination) of a loose part that has been detected. An alarm on any one of the active channels will activate an audible alarm in the main control room and activate two four-channel tape recorders and a paper printer located in the control room area.

A.3.3 REOUIRED OPERATOR ACTIONS The required operator action por Alarm Procedure ALM-0062A are as 'ollows:

Autonatic Actions:

1. High alarm starts:

o Tape Recorders o Paper Printer

2. Low alarm:

o None Initial Onorator Actionn:

1. Verify chann61 with alarm.

2. Hi alarms:

o Verify tape recorders have started, o verify printer has printed.

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3. I4 Altras o See Subsequent Operator Actions.

Subsecuent Operator Actions:

1.0 HI Alarm:

1.1 Align left and right speakers of an audio monitoring device to alarming channel (s).

1.2 Determine if loose part is audible.

1.3 IE loose part is audible and/or HIGH alarm continues, THEN notify Engineering Support personnel.

1.4 Record plant conditions which may have caused alarm.

1.5 Check meter doflection on LPMS to ensure the reading is less than 100%. If not, contact I&C personnel to adjust "FULL SCALE RANGE-g" switch to obtain highest reading but less than 100%.

1.6 If loose part is not audible after several seconds, THEN stop recorders.

2.0 LO Alarm 2.1 Align left or right spanker to alarming channol(s).

2.2 Dotormino if any sound is audible on the speaker.

2.3 Contact I&C personnel to adjust "Full r Scalo Rango-g". switch to clear low alarm.

2.4 If low alarm does not clear or no cound is-audible, notify I&C personnel.

A.3.4 TECHNICAL SPECIFICATION REOUIREMENTS Technical Specification 3/4.3.3.8 applion to the System.

Annlicabilitc: MODES 1 and 2.

Action:

a. With one or moro Looso-Part Detection Syntom

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chtnnolo insporcblo for core than 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the channel (s) to OPERABLE status.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

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A.4 INCORE INSTRUMENTATION SYSTEM A.4.1 SYSTEM FUNCTION The Incore Instrumentation System provides information on the neutron flux distribution and fuel assembly outlet temperatures at selected core locations. Using the information thus obtained, it is possible to confirm the reactor core design parameters and calculate the hot channel factors. The system provides a means for acquiring data only and performs no operational plant control.

The experimental data obtained from the incore temperature and flux distribution instrumentation system, in conjunction with previously determined analytical information, can be used to determine the three dimensional fission power distribution in the coro at any time throughout core life. Once the fission power distribution has been established, the maximum power  !

output is primarily determined by thermal power distribution and the thermal and hydraulic limitations which determino the maximum core capability.

The incore instrumentation also proyidos information which is used to calculato the coolant enthalpy and fuel burnup distribution, to estimato the coolant flow distribution, and to calibrato the Excore Nuclear Instrumentation System for axial offset.

The incore instrumentation system is used by the operator only for acquiring data. Adequate information from the safety instrumentation has boon provided to the operator so that he does not have to rely upon the incore data A-14

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• t recOivcd for cefo operstien of the plcnt. The dato generated is used only in the analysis of core conditions.

A.4.2 SYSTEM CONFIGURATION The incore instrumentation consists of thermocouples positioned to measure fuel assembly coolant outlet temperature and incore flux thimbles to permit insertion of movable detectors for measuremont of the neutron flux distribution within the reactor core. Movable miniature neutron flux detectors are available to scan the active length of selected fuel assemblies to provide remote reading of the relative three dimensional flux distribution. Figure A4-1 shows the routes taken by the incore flux thimblos.

Minature fission chamber detectors can be remotely positioned in the retractable guide thimbles to provida flux mapping of the core. The retractable thimbles, into which the miniature detectors are driven, are pushed into the reactor coco through conduits which extend from the bottom of the reactor vossol, down through the concreto shield area, and then up to a thimble seal table. Their distribution over the core is nearly uniform with about the same number of thimbios, located in each quadrant.

During reactor operation, the retractable thimbles are stationary.

The drivo system for the insortion of the minaturo detectors consists basically of drive assemblies, five-path rotary transfer assemblics, and 10-path rotary transfer assemblics.

The Control and Roadout System providos means of insorting the minature neutron detectors into the reactor A-15

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core and withdrawing the detectors while plotting neutron flux versus detector position. The control system is located in the Control Room. Limit swit:hes in each transfer device provide feedback of each path selection operation. Each gear box drives an encoder for position feedback. One five-path operation selector is provided for each drive unit to insert the detector in one or five functional modes of operation. One 10-path operation selector is also provided for each drive unit that is then used to route a detector into any one of up to 30 selectable paths. A common path is provided to permit cross calibration of the detectors.

A.4.3 REOUIRED OPERATOR ACTIONS operator action of the system is required only when the instrumentation is to be utilized to perform its design functions (i.e. core flux mapping, calculate hot channel factors, calibrate Excore Nuclear Instrumentation System, etc.). System Operating Procedure (SOP)-710 outlines the steps that are needed to be taken to ensure successful performance of the system.

A.4.4 TECHNICAL SPECIFICATION REOUIREMENTS Technical Specification 3/4.3.3.2 applies to the Movable Incore Detectors.

Annlicability When the Movable Incore Detection System is used for:

a. Recalibration of the Excore Neutron Flux Detection System, or

b. Monitoring the QUADRANT POWER TILT RATIO, or

c. P.casurement of FN g, F (Z) and xy*

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.* v Actient With tho Movcble Incora Detection System inoperable, do not use the system for the above applicable monitoring or calibration functions.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

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e r A.5 STEAM GENERATORS A.5.1 SYSTEM FUNCTION .

The Steam Generators allow heat produced in the RCS to be transferred to the secondary side while maintaining a radiological barrier between the two systems. The Steam Generators are functional during Modes 1, 2, 3 and 4 of operation.

A.S.2 SYSTEM CONFIGURATION There are four reactor coolant loops in the RCS with one steam generator per loop.

A.5.3 REOUIRED OPERATOR ACTIONS Each Steam Generator must be verified as operational prior to a successful completion of plant hoatup from Cold Shutdown (Mode 5) to Hot Standby (Mode 3). The following steps were taken from the Integrated Plant Operating Proceduro (I PO) -0 01 A', "Plant Heatup From Cold Shutdown to Hot Standby:"

5.2.22 When RCS temperaturo reachos 200 F, log timo Mode 4 is ontorod. , ,

i 5.2.23 Lino up the Steam Gonorator Atmospherics for heatup as follows:

WHEN steam is verified oxiting the Steam Generator Atmospherica, THEN CLOSE Steam Generator Atmosphoric Relief Valves.

T3ereforo, beforo stops nocosnary for heatup and A-19

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proccurization to Mods 3 cro teksn, tha etsca vsrified as exiting the Atmospheric Relief Valves ensures that the Steam Generators are properly transferring heat from ths'nCS to the accondary side.

A.5.4 TECHNICAL SPECIFICATIQN REOUIkEMENTS Technical Specification 3/4.4.5 applies to the Stesa Generators.

Aeolicability: MODES 1, 2, 3, and 4.

Action: With one or more steam generators inoperable, restore the inoperable generator (s) to OPERABLE status prior to increasing T avg above 200 F.

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4 A.6 SEALED SOURCE CONTAMINATION A.

6.1 BACKGROUND

Within the code of Federal Regulations (10CFR), Part 70 pertains to the domestic licensing of special nuclear material and Section 70.39 deals with specific licenses for the manufacture or initial transfer of calibration or reference sources. The requiremento set forth in this section are primarily determined with respect to any source containing more than 0.005 microcurie of plutonium.

The commission mandates that the five following prototype tests be conducted on such a source so as to measure any radiation leakage and/or contamination that may be present:

o Initial measurement by direct counting of the sourCo.

o Dry wipe test.

o Wet wipe test.

o Water soak test.

o Dry wipe test repeated.

As a result of such testing, removal of more than 0.005 microcurie of radioactivity in any one of the tests is considered cause for re]cction of the source design.

Technical Specification 3/4.7.10 applies to Scaled Source Contamination.

Apnlicability: At all times.

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Q r' f Action: W!.th a' sealed source having removable contamination in excess of the above limits,  !

immediately withdraw the sealed source from use and either:

1. Decontaminate and repair the sealed source, or i

2. Dispose of the sealed source in accordance with Commission Regulations.

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