ML20090B571

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Rev 3 to Procedure 1-AP-5.1, Unit 1 Radiation Monitoring Sys
ML20090B571
Person / Time
Site: Ginna Constellation icon.png
Issue date: 10/20/1987
From:
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML20090B375 List:
References
FOIA-91-106 1-AP-5.1, NUDOCS 9203030466
Download: ML20090B571 (36)
v • d • e


Text

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REQU2ST TO CHANGE PROCEDURE' ADM-3.4 i

i NORTH ANNA POWER STATION VIRGINIA POWER Page 1 of 1 07-09-87

../ERVISOR RESPONSIBLE FOR FOLLOVING PROCEDURE:

1 O ABNORMAL O CURVE BOOK -

O OPERATING O VELDING O _ ADMINISTRATIVE O EMERGENCY O PERIODIC TEST O

O ANNUNCIATOR O IN-SERVICE INSPECTION O HEALTH PHYSICS O

O CALIBRATION O MAINTENANCE O SPECIAL TEST O

O CHEMISTRY O NON-DESTRUCTIVE TEST O START-UP TEST O

?ROCEDURE NO:

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,4p - f f 2 UNIT NO:

g 3 REVISION DATE:

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HANGES REQUESTED:

(GIVE STEP NUMBER, EXACT SUGGESTED WORDING, AND LIST REFERENCES. STAPLE 6

COPY OF PROCEDURE WITH SUGGESTED CHANGES MARIED TO THIS FORM.)

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ACTION TAKEN:

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'OES THIS CHANGE THE OPERATING METHODS AS DESCRIBED IN THE UFSARv/

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OES THIS CHANGE INVOLVE A CHANGE TO THE TECH. SPECS /

O YES O NO OES THIS CHANGE INVOLVE A POSSIBLE UNREVIEVED SAFETY QUESTION /

O YES O NO F ALL "NO". NO " SAFETY ANALYSIS" IS REQUIRED.

IF ANY "YES", A " SAFETY ANALYSIS" IS REQUIRED.

10CFR50.39) APPROVED COPY TO BE PROVIDED TO LICENSING COORD. FOR INCLUSION IN ANNUAL REPORT.

ECOMMENDED ACTION:

11

% APPROVED O DISAPPROVED DOES THIS PROCEDURE CREATE A QA DOCUMENT /

YES$

NO-O (COGNIZAN} h W ( nSL7ERJISOR) 12 DATE:

13 SY:

3D A* b REVIEWED BY QUALITY ASSURANCE:

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CHANGES MADE:

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15 DATE 16 REVIEWED BY STATION NUCLEAR SAFETY AND OPERATING COMMITTEE:

17 kAPPROVED O DISAPPROVED O APPROVED AS MODIFIED BY COMMIT ~EE 1.

CHAIRMAN SIGNATURE

  • 18 DATE:

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NEW PROCEDURE REVISION DATE:

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20 9203030466 910019 p//

m ACTION COMPLETED BY:

PDR FOIA 21 DATE:

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22 WILLIAM 91-106 PDR s-n

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1 Ap_3,1 Page 1 of 13 00-00-00 VIRGINIA POWER NORTH AEEA POWER STATIM UNIT 1 UNIT 1 RADIATIN MMITORING SYSTElf (With One Attachment)

REFERENCES:

1.

FSAR Chapter 11 and 12 2.

North Anna Tech Specs 3.

EPIP 1.01 4.

EWR-87-569 REV.:NO.

9 PAGE:

ENTIRE DATE: 00-00-00 APPROVAL:

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RECOMMENDED APPROVAL:

APPROVED BY:

CHAIRMAN STATION NUCLEAR SAFETY AND OPERATING COMMITTEE DATE: 00-00-00 l

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j-AP-5.1 Page 2 of 13

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00-00-00 1.0 Pu rpose 1.1 This procedure provides indications, probable causes, immediate operator actions, and long term operator actions for radiation monitoring alare9 associated with any of the following monitors:

Condenser Air Ejector SV-121 Main Steam Header MS-RMS-193 Containment Cas and Particulate RMS-159, 160 Pbnipulator Crane RMS-162 Discharge Tunnel SW-130 Reactor Coolant Letdown CH-128 Steam Generator Blowdown SS.'22. 123. 124 Containment High Range RMS-161 Reactor Containment Area RMS-163 Incore Instrument Room RMS-164 Recirc Spray Cooler Service Water SW-124, 125, 126 127 This procedure also provides release levels from certain Rad Monitors which monitor environmental release points which will i

be used to determine whether the NRC should be notified within L

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or whether the Emergency Plan should be implemented, l

required by Administrative Procedure 19.6 for any l

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" Accidental, Unplanned, or Uncontrolled Radioactive Release",

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1-AP-5.1 Page 3 of 13 00-00-00 2.0 Indications 2.1 Main Control Board annunciator, visual and audible alarm.

2.2 Individual channel alarm at radiation monitoring panel.

2.3 Area alarm and red light (LOCAL).

2.4 Main Steam Header N-16 radiation monitor alarms:

2.4.1 HI alarm (yellow light - SI) blinking or solid.

2.4.2 HI-HI alarm (red light - S2) blinking.

2.4.3 HI-llI-HI alarm (red light - SE) reflashes the HI-HI annunciator with 100 gpd from recorder.

3.0 Probable Causes 3.1 As applicable, primary leakage.

3.2 As applicable, primary to secondary leakage.

3.3 Movement of radioacciae materials near the monitor.

3.4 Loss of high voltage to the detector, instrument malfunction or inst rument not in the operate mode, as applicable.

3.5 Maintenance or cleanup in progress in the area of the monitor.

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1-AP-5.1 Page 4 of 13 00-00-00 4.0 Immediate Operator Action 4.1 Determine which monitor is in alarm.

4.2 Notify Shif t Supervisor.

4.3 Go Immediately to Long Tern Operator Actions.

Completed By:

Date:

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1-AP-5.1 Page 5 of 13 00-00-00 1

Initials 5.0' Lona Tern Operator Action 5.1 Verify the alarm is VALID for the affected Rad Monitor (s) as follows:

5.1.1 Verify the following on the Main Steam Header N-16 Radiation Monitor:

5.1.1.1 Jg[ the green (SS) light is lit. THEN the N-16 Rad Monitor is operable and not in alarm. -Go to step 5.1.2.

M 1.1.

_ any of the alarm 11ghts (S1, S2, SE) are lit, THEN immediately notify 'the STA to evaluate the SG 1eakrate trend data AND go to step 5.3.

5.1.1.3 JU[ none of the lights (SS, S1, S2, or SE) are lit, THEN immediate'y notify the STA to evaluate the SG 1eakrate trend data AND go to step 5.14.5.

5.1.2 Verify the follovir; for all other Unit 1 Rad Monitors:

5.1. 2.1 ' High reading on meter.

5.1.2.2 AIL switches in normal position.

5.1.2.3 Njl flow C@ filter alarms present.

5.1.2.4 Increase in associated Rad Monitors activity.

p 5.1.2.5 RM-RR-100 (Trend recorder).

I

5. 2.

Jg[ alarm is due to inoperable Rad Menitor, THEN go to steps 5.14.1 thru 5.14.6.as applicable.

5.3 Notify Health Physics to sample the affected system Ojt area.

1-AP-5.1 Page 6 of 13 00-00-00 5.4 Refer to appropriate step:

5.4.1 Condenner Air Ejector (RM-SV-121)

Go to Step 5.5.

5.4.2 Main Steam Header N-16 (01-MS-RMS-192)

Go to Step J.6.

5.4.3 Discharge Tunnel (RM-SW-130)

Go to Step 5.7.

5.4.4 Containment Gas and Particulate (RM-RMS-159,160)

Go to Step 5.8.

5.4.5 Containment Manipulator Crane (RM-RNS-162)

Go to Step 5.8.

5.4.6 Containment 0]t incore Inst. Room (RM-RMS-163,164)

Go to Step 5.9.

5.4.7 Containment Hi Range (RM-RMS-161)

Go to Step 5.10.

5.4.8 RCS Letdown (RM-CH-128)

Go to Step 5.11.

5.4.9 S/G Blowdown (RM-SS-122, 123, 124)

Go to-Step 5.12.

5.4.10 R.S. HX S.W. Outlet (RM-SW-124,125.126,127)

Go to Step 5.13.

o Environmental Release Pathway

^^

1-AP-5.1 Page 7 of 13 00-00-00 5.5 Condenser Air Ejector (RM-SV-121).

5.5.1 Immediately notify the STA to evaluate the SC leakrate trend data and refer to Standing order.#155.

5.5.2 IF, required. THEN verify Air Ejector discharge swaps to containment.

5.5.2.1 TV-SV-102-1 OPEN I

5.5.2.2 TV-SV-103 OPEN

~-

5.5.2.3 TV-SV-102-2 CLOSED 5.5.3 IF necessary, notify personnel to exit containment.

NOTE:

Utilize the following steps to determine whether the I

release limits have been exceeded until grab samples 2

can be obtained AND analyzed (within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />).

NOTE:

Il[ the Air Ejector discharge has diverted to the containment, steps 5.5.4 thru 5.5.5 may OR may NOT be performed as determined by the Shif t Supervisor.

5.5.4 Utilize the appropriate curve on A*cachment 1 to determine if ef fluent release is greater than 10002

)

of the Tech. Spe:. instantaneous allowable limit.

1 5.5.4.1. Valid reading greater than 1000% for greater than 15 minutes.

5.5.4.2 IJ; the release exceeded 1000% of the T.S.

instantaneous limit. THEN 5.5.4.2.1 Initiate EPIP.

5.5.4.2.2 Declare an " Alert".

5.5.4.2.3 Initiate notifications as per ADM-19.6.

4 1-AP-5.1

^

Page 8 of 13 00-00-00 5.5.5 Utilize the appropriate curve on Attachment 1 to determine if effluent release is greater than 100% of the T.S. limit:

5.5.5.1 Valid reading is greater than 100% T.S.

curve.

5.5.5.2 g the release exceeded 100% of the T.S. Limit, THEN 5.5.5.2.1 Initiate EPIP.

5.5.5.2.2 Initiate " Notification of Unusual Event".

5.5.6 E S/G tube leakage is suspected, THEN refer to l-AP-24.

5.6 Main Steam N-16 Radiation Monitor (01-MS-RMS-193)

s 5.6.1

'E the yellow alarm light (SI) is lit OJ blinking.

,e THEN initiate 1-AP-24.2.

5.6.2 IF,one or both of the"ted alarm lights (S2 or SE) are lit OR blinking, THEN refer to Standing Order #155.

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1 - AP-5,1 Page 9 cf '3 00-00-00 5.7 Dischatge Tunnel (RM-SW-130) 5.7.1 Secure any possible release to the discharge tunnel.

NOTE:

Utilize the following steps to determine whether the release limits have been exceeded until grab samples can be obtained AND analyzed.

(within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) 5.7.2 Determine if ef fluent release is greater than 10 times the T.S. instantaneous limit:

3 5.7.2.1 Valid reading is greater than 6.0 x 10 cpm for greater than 15 minutes.

5.7.2.2 1][ the release exceeded 10 times the T.S.

instantaneous limit. THEN:

5.7.2.2.1 Initiate EPIP.

5.7.2.2.2 Declare an " Alert".

5.7.2.2.3 Initiate notifications as per ADM-19.6.

5.7.3 Determine if ef fluent release is greater than 100% of the T.S. limit:

2 5.7.3.1 Valid reading greater than 6.0 x 10 cp, 5.7.3.2 IF the release exceeded 100% of the T.S.

limit:

5.7.3.2.1 Initiate EPIP.

5.7.3.2.2 Initiate " Notification of Unusual Event".

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l-AP-5.1 Page 10 of 13 00-00-00 5.8 Containment Caseous or Particulate (mi-RMS-159.160)

Manipulator Crane (RMS-162) 5.8.1

)); unit is in Mede 5 or 6 AND alarm is due to Hi Hi radiation, THEN verify the following:

1-HV-F-4A NOT running 1-HV-F-4B NOT running 1-HV-F-5A NOT running 1-HV-F-5B NOT running MOV-HV-100A Closed MOV-HV-100B Closed MOV-HV-100C Closed MOV-HV-100D Closed MOV-HV-101 Closed MOV-HV-102 Closed 5.8.2

)); required refer to 1-AP-30 for Fuel Failure During Handling.

5.8.3 Determine cause for alarm AND take corrective action as required.

5.8.4

))[ unit is in Mode 1-4 AND alarm is due to Hi-Hi radiation perform the following as required:

5.8.4.1 Perforu l-PT-52.2 (RCS Leakrate).

5.8.4.2 Have camples of Containment Sump Discharge analyzed by Chemistry and/or Health Physics.

5.9 Containment (RMS-163) OR Incore Room (RMS-164) 5.9.1 Secure any maintenance in Containment AND determine source of activity.

5.9.2 IF activity is due to Incore Detector Movement, keep all personnel away from Seal Table Area.

5.9.3 Insure alans clears when flux mapping complete OR maintenance clesnup is complete.

!-AP-5.1 Page 11 of 13 00-00-00 5.10 Containment Hi Range (RMS-161) 5.10.1 Determine if activity is from contaminated materials 0]! sources being moved in area C@t if radiography is in progress in vicinity of Rad Monitor.

5.10.2 Instruct H.P. to monitor radiation levels if required.

5.11 RCS Letdown (RM-CH-128)

NOTE:

IF excessive activity levels warrant securing normal letdown.

THEN refer to 1-OP-8.5 for operation of Excess letdown.

5.11.1 consult with H.P. AND Chemistry to determine JJ[ a change in letdown and purification rate is desired.

5.11.2 IF required refer to 1-0P-8.2.

5.11.3 Monit'or Aux Bldg AND Vent Stack Rad Monitors for increasing radiation levels.

5.11.4 Refer to T.S. 3.4.8 for further actions as required.

5.12 S/G Blowdown (RM-SS-122.123.124) 5.12.1 Monitor Condenser Air Ejector (RM-SV-121) AND' Main Steam Header N-16 (MS-RI-193A) radiation monitors for increasing activity.

5.12.2 Determine if Rad Monitor should be backflushed.

5.12.3 IF alarm persists C$ activity level increases refer l

to 1-AP-24.

5.13 R.S. S.W. Outlet (RM-SW-124,125,126,127) 5.13.1 Isolate affected RS EX.

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Page 12 of 13 00-00-00 5.14 Perfore the folicving for any Inoperable Rad Monitor (s) 5.14.1 Containment Caseous and Particulate {gt Manipulator Crane during Refueling.

5.14.1.1 Verify. stopped fyt secure the following:

1-HV-F-5A 1-HV-F-5B 1-HV-F-4A 1-HV-F-4B 5.14.1.2 Verify closed Cgt close the following:

MOV-HV-100A MOV-HV-100B MOV-HV-100C MOV-HV-100D MOV-HV-101 MOV-HV-102 5.14.1.3 IF, necessary advise personnel in containment 5.14.1.4 Refer to T.S. 3.9.9 AND 3.3.3.1 5.14.2 Containment Caseous and Particulate Modes 1 - 4.

(RHS-159,160) 5.14.2.1 Refer to T.S. 3.1.6.1 AND 3.3.3.1 5.14.3 Discharge Tunnel (RM-SW-130) 5.14.3.1 Refr to T.S. 3.3.3.10 5.14.4 Condenser Air Ejector (RM-SV-121) 5.14.4.1 Immediately notify the STA to evaluate the SG 1eakrate trend data.

5.14.4.2 Notify the NRC Resident Inspector within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the inoperable A.E. Rad Monitor.

5.14.4.3 Refer to Standing order #155.

5.14.4.4 Refer to T.S. 3.3.3.11

> -. ~ ~

1-AP-5.1 I.

Page 13 of 13 00-00-00 5.14.5 Main Steam Header N-16 (01-MS-RMS-193) 5.14.5.1 Immediately notify the STA to evaluate the SG leaktate trend data.

5.14.5.2 Perform the following to detetuine operability:

5.14.5.2.1 Depress the "STA" button and verify the display reads "DF.

CPTCE".

5.14.5.2.2 E the display does not read "DF. CPTGE" or the green (SS) light does not come on after a maximum vait period of 10 micuras. THEN declare the N-16 Rad monitor inoperable AND refer to Standing order #155.

5.14.5.2.3 E the green (SS) light comes on after the required cycle time elapses AND the annunciator still does not clear. THEN return to step 5.1.

5.14.6 Submit a Work Request for ANY Inoperable Rad Monitor Completed By:

Date:

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l-AP-5.1 Page 1 of 1 00-00-00 3

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North Anna Power Station Standing Order #155 (Revision 1)

Issue Date: October 9,1987 TITLE: UNIT ONE (1) AND UNIT TWO (2) PRIMARY 'IV SECONDARY LEAKAGE 1.

During power operation the following leak rate limits are in effect. 'Ihese limits are more restrictive than Tech Spec 3.4.6.2.C and are consistent with PT-46.2, Primary-to-Secondary Leak Rate Determination test, Total leakage of 160 GPD increase from one surveillance interval to a.

the next surveillance interval as specified in PT-46.2.

b.

Leakage trom an Individual steam generator of 160 GPD increase from one surveillance interval to the next surveillance interval as specified in PT-46.2.

c.

Totalleakage of <300 GPD.

d.

Leakage from an individual steam generator of <100 GPD.

An increasing trend based on the latest surveillance that indicates >100 e.

GPD would be exceeded on an individual steam generator within ninety (90) minutes-If the limits of (a) or (b) are exceeded, then power is to be reduced to 150%

as soon as possible under normal controlled conditions, but not to exceed ninety (90) minutes.

If the limits of (c) or (d) are exceeded, then initiate a unit shutdown to be in MODE 3, Hot Standby, within six (6) hours. Initiate AP-24.1, Large Steam Generator Tube Leak Requiring Rapid Unit Shutdown.

If the limits of (e) are exceeded, then initiate a unit shutdown to reduce power below 50% within ninety (90) minutes and be in MODE 3, Hot Standby, within six (6) hours of exceeding limit (e). Initiate AP-24.1, Large Steam Generator Tube Leak Requiring Rapid Unit Shutdown.

2.

If the leak rate trend is such that the limits of (l.a) or (1.b) will be exceeded within the next 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, then notify the Superintendent Operations or SRO on-call.

3.

If the limits of (1.a) through (l.e) are exceeded, then carry out station proce-dure ADM-19.6, Notification. When power reduction is commenced a one (1) hour report per 10CFR50.72 (b)(1)(i)(A) is to be made even though a Tech Spec limit has not been exceeded.

4.

In order to perform PT-46.2, Primary-to-Secondary Leak Rate Determination, the following equipment or systems are required to be operable.

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Page 8,of.3 t

Primary sample system to obtain a RCS sample from any point except a.

the pressurizer.

b.

Secondary sample system to obtain a steam generator liquid sample, c.

Air ejector radiation monitor, d.

Air ejector flow rate indication for each air ejector.

e.

$-16 radiation monitor on the main steam header.

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Multichannel analyzer for counting various samples collected.

g.

Steam geaerator blowdown flow' indication when blowdown is in service.

5.

Action to be taken for inoperable systems or equipment listed in item (4).

If the primary or secondary sample systema are inoperable, then the a.

system must be returned to service within twenty-four (24) hours, provided the N-16 and air ejector radiation monitors are operable, or power reduced to <50% within the next ninety (90) minutes. If the N-16 and air ejector radiation monitors are not operable, then reduce power to <50% under normal controlled conditions, but not to exceed ninety (90fminutes, b.

If the air ejector radiation monitor is inoperable, then carry out the Tech Spec requirt ments every four (4) hours. If the air ejector radiation montfor is not returned to service within the next seven (7) days, then reduce power to150% within the next four (4) hours.

If air ejector flow rate indicators are inoperable, then carry out the c.

Tech Spec requirements, d.

If the N-16 radiation monitor is inoperable, it must be returned to ser-Vice as soon as possible, and the NRC Resident Inspector notified within twenty-four (24) hours. If the N-16 radiation monitor is not returned

-to service within seven (7) days, then reduce power to 150% within the next hour (4) hours, If no Multichannel Analyzer (MCA) is operable, one must be-returned e.

to service as soon as possible. During the time of inoperability, an 7

l RCS leak rate calculation must be performed every eight (8) hours.

If a MCA is not returned to service within seven (7) days, then reduce power tof50% within the next '. ur (4) hours, f.

If blowdown flow indication is inoperable, then estimate flow when required.

l i

m d

m e

P C"

Page 3 of 3 If both the N-16 and air ejector radiation monitors are inoperable, then g.

reduce power to <50% under normal controlled conditions, but not to exceed ninety (90)Tninutes.

h.

If other combinations of required equipment or systems are not available, !

then notify Superintendent Operations (or Assistant Station Marager O&M) and the STA. A JCO will be required as well as discussions with the NRC.

6.

Standing Order #152, U-1 Primary to Secondary Leakage, is superseded and cancelled by this standing order.

7.

No deviations to this Standing Order are allowed without prior approval of the SNSOC and concurrence of the Station Manager or Vice President Nuclear Opera tions. The NRC Resident Inspector will be notified of any approved deviations.

Nh Chairman G

R. O. [;;lfinftr SNSOC Superintendent Operations y

O/

/)>W,)

E. Waynt iffirrell

Station Manager ACTION REQUIRED TO CLEAR STANDING ORDER:

Revise Technical Specifications.

(

1

1K-D4 s

l Page 1 of I 00-00-00

}

a NORTH ANNA POWER STATION UNIT #1 w 15...

i RAD MONITOR SYST HI-HI RAD LEVEL ANNUNCIATOR IK-28 1.0 Probable Cause 1.1 High-High radiation alarm on a Unit 1 or common radiation monitor.

1.2 High-High radiation alarm on the Main Steam Header N-16 Radiation monitor.

1.3 High-High-High radiation alarm on the Main Steam Header N-16 radiation monitor (reflash of the HI-HI alarm).

1.4 Failure of alarm relay 74C.

2.0 Operator Action Response Not Obtained 2.1 Verify Njl alarms on Unit 1 or Initiate 1-AP-5 (radiation monitoring).

common radiation monitors.

2.2 Verify NO red alarm light (s) are Initiate 1-AP-5 (radiation monitoring).

~

lit on the Main Steam Header N-16 radiation monitor (100 gpd from recorder causes the annunciator to reflash).

2.3 Verify NO radiation monitors Notify Instrument department.

have failed.

Refer to T.S. 3.3.3.1.

IJ[ Main Steam Header radiation monitor is inopertble, refer to Standing Order #155.

3.0 References 3.1 ESK-10AAW, llW. and llX 3.2 S&W dwg 11715/12050-1.21-67 3.3 NAPS Instrumentation Manual page RM016 3.4 EWR-87-569 3.5 T.S. 3.3.3.1 (Radiation Monitoring) 4.0 Actuation 4.1 74HHC high-high radiation alarm relay.

4.2 K101 contacts from all Unit I and common radiation monitors.

4.3 Main Steam Header radiation monitor.

(

sg

.,-..L IK-D2 Page 1 of 1 00-00-00 NORTH ANNA POWER STATION UNIT $1-

_ RAD MONITOR SYSTEM HI RAD LEVEL ANNUNCIATOR 1K-12 1.0 Probable Cause

' 1.1 High radiation alarm on a Unit 1 or common radiation monitor.

1.2 High radiation alarm on the Main Steam Header N-16 Radiation monitor.

1.3 Failure of alarm relay 74-HC.

I 2. 'O Operator Act' ion Response Not Obtained 2.1 Verify:NO alarms on Unit 1 or Initiate 1-AP-5 (radiation monitoring).

. common radiation monitors.

2.2 _ Verify lio, alarms (yellow light)

Initiate 1-AP-5 (radiation monitoring),

on the Main Steam Header N-16 radiation monitor.

2.3 Verify NO radiation monitors Notify Instrument department.

l have failed.

L Refer to T.S. 3.3.3.1.

l IF Main Steam Header radiation monitor is l

inoperable THEN refer to Standing Order L

  1. 155.

3.0 References

.3.1 ESK-10AAW, 11W, and 11X 3.2 S&W dwg 11715/12050-1.21-67 3.3 NAPS' Instrumentation Manual page RM011 3.4 EWR-87-569 3.5 T.S. 3.3.3.1 (Radiation Monitoring).

~

4.0 Actuation 4.1 ~ 74-HC radiation high alarm relay.

4.2 K102 contacts from all Unit 1 and commen radiation monitors.

4.3 Main Steam Header radiation monitor.

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/

Page 1 of 3 North Anna Power Station Standing Order #155 (Revision 1) lasus Date: October 9,1987 TITLE: UNIT ONE (1) AND UNIT TWO (t) PRIMARY 'fC SECONDARY LEAKAGE

^

I 1.

During power operation the following leak rate Umits are in effect. 'these

!!mits are more restrictive then Tech Spec 3.4.6.2.C and are consistent with PT-46.2, Primary-to-Secondary Leak Rate Datermination test.

Total leakage of 160 GPD increase from one surveillance interval to a.

the next surveillance interval as specified in PT-46.2.

W b.

Leakage from an individual steam generator of 160 GPD increase from one surveillance Interval to the next surveillance interval as specified in PT-46.2.

c.

Totalleakage of <300 GPD.

4 d.

Leakage from an Individual steam generator of <100 OPD.

An increasing trend based on tye latest surveillance that indicates >100 e.

GPD would be exceeded on an' individual steam generator within ninety (90) minutes.

~

If the limits of (a) or (b) are exceeded, then power is to be reduced to <50%

i as soon as possible under normal controlled conditions, but not to exceed ninety (90) minutes.

If the limits of (c) or (d) are exceeded, then initiate a unit shutdown to be in MODE 3, Hot Standby, within six (6) hours. Initiate AP-24.1, Large Steam

( merator Tube Leak Requiring Rapid Unit Shutdown.

If the limits _of (e) are exceeded, then initiate a unit shutdown to reduce power below 50% within ninety (90) minutes and be in MODE 3, Hot Standby, wiudn six (6) hours of exceeding limit (e). Initiate AP-24.1, Large Steam Generator Tube Leak Requiring Rapid Unit Shutdown.

. 2.

. If the leak rate trend is such that the limits of (1.a) or (1.b) will be exceeded within the next 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, then notify the Superintendent Operations or SRO on-call.

3.

If the limits of (l.a) through (i.e) are exceeded, then carry out station proce-dure ADM-19.6, Notification. When power reduction is commenced a one l

(1) hour report per 10CPR50.72 (b)(1)(IXA) is to be made even though a Tech Spec limit has not been exceeded.

1 4.

In order to perform PT-46.2, Primary-to-Gecondary Leak Rate Determination, l

the following equipment or systems are required to be operable.

l g - Pasdv.4 %pckoe Gmed *. W dW:xswib < M% Ocwsee. / bye.w s

(ruwm in dwls raJe &>,'il be ensedere d f> cwm pm m j

sg ry.,.J /+e a, c./ in s WUl /> e {c) Ae A bch 5 el cm 6 it g. % tut /,'f gw ( c/tamis /rfd vN

/d W6 fove dWhewise, {

Pcge I cf 3 Primary sample system to obtain a RCS sample from any point except 8.

the pressurizer.

Secondary sample, system to obtain a steam generator liquid sample.

b.

c.

Air ejector radiation monitor.

d.

Air ejector flow rett. Indication for each air ejector.

N-16 radiation monitor on the main steam header.

e.

f.

Multichannel analyzer for counting various samples collected.

g.

Steam generator blowdown flow Indication when blowdown is in service.

5.

Action to be taken for inoperable systems or equipment listed in item (4).

a.

If the primary or secondary sample systems are Inoperable, then the system must be returned to service within twenty-four (24) hours, provided the N-16 and air ejector radiation monitors are operable, or power reduced to <50% within the next ninety (90) minutes. If the N-16 and att ejector radiation monitors are not operable, then reduce power to <50% under normal controlled conditions, but not to exceed ninety (90Tminutes.

b.

If the air ejector radiation monitor is inoperable, then carry out the Tech Spec requirements every four (4) hours. If the air ejector radiation montfor is not returned to service within the next seven (7) days, then reduce power toiS0% within the next four (4) hours, If air ejector flow rate indicators are Inoperable, then carry out the c.

Tech Spec requirements.

d.

If the N-16 radiation monitor is inoperable, it must be returned to ser-Vice as soon as possible, and the NRC Resident inspector notified within twenty-four (24) hours. If the N-16 radiation monitor is not returned to service within seven (7) days, then reduce power to 150% within the next hour (4) hours, If no Multichannel Analyzer (MCA) is operable, one must be returned e.

to service es soon as possible. During the time of inoperability, an RCS leak rate calculation must be performed every eight (8) hours, if a MCA is not returned to service within seven (7) days, then reduce power tof50% within the next four (4) hours.

f.

If blowdown flow indication is (noperable, then estimate flow when required.

JAE8 3 GI 3 g.

l'f both the N-16 and air ejector radiation monitors are Inoperable, then reduce power to <50% under normal controlled conditions, but not to exceed ninety (90)Ininutes.

h.

If other combinations of required equ!pment or systems are not available, then notify Superintendent Operations (or Assistant Station Manager O&M) and the STA. A JCO will be required as well as discussions with the NRC, 6.

Standing Order #152. U-1 Primary to Secondar) Leakage, is mq>erseded and cancelled by this standing order.

7.

No deviations to this Standing Order are allowed.vithout prior approval of the SNSOC and concurrence of the Station Manager or Ylee President Nuclear.

Operations. The NRC Resident inspector will be notified of any approved deviations.

Nh D~hairman g)

R. O. g#finitt SNSOC 7

Superintendent Operations 6//

"~

E. Wayne Ifkrrell

Station Manager ACTION REQUIRED TO CLEAR STANDING ORDER:

Revise Technical Specifications.

l l

-~

'?

m, awa wn -

CAP-4.0 Page 1 of 3 00-00-00 o

VRCIMIA POWER

)KRH Altta P0m STATIM usarusTRY PRIMARY-TO-66 EAR rafnffAT5:Jd 1.0 References 1.1 Calculational Basis for CAP-4.0, approved March 14, 1987.

1.2 PT-46.2, " Primary-to-Secondary Leak Rate Determination."

i.3 Mg Crders, 1.0 Purpose 2.1 The purpose of this procedure is to provide a method for calculating the primary to secondary leakage through a steam generator.

3.0 Procedure NOTE:

The primary to secondary leak rcte must be perfcrmed prior to 0300 every day.

3.1 Record the most recent RCS (Reactor Coolant System) specific isotopic activities on Attachment 6.1 (Note 5.1).

3.2 Record the most recent specific isotopic activities for each steam generator (S/G) on Attachment 6.1 (Note 5.1).

3.3 From the Control Room Operator, obtain the steam generator blowdown rates at the time the samples in Step 3.2 were pulled.

3.4 Determine which isotopes may be used for leak rate calculation ~by verifying that the respective unit has. been at stable power and S/G blowdown rate has been constant for the period of time listed on.2.

3.5 On Attachment 6.3, record whether each isotope has met the Stability Criteria. If the Stability Criteria is not met, then mark the i

respective leak rate N/D.

l l

C B>

l

+

CAP-4.0 Page 2 of-3 00-00-00

,3. 6 For each steam generator, calculate the leak rate for the isotopes that have. net the Stability Criteria by using the leak rate equation on Attachment 6.1.

Record the results on Attachment 6.3.

3.7 For each steam generator, calculate the Short Lived Iodine Average Leak Rate by averaging the calculated leak rates for I-132 I-133, I-134, and I-135. Record the results on Attachment 6.3.

3.8 For each steam generator, determine which of the following leak rates is the largest:

Na-24, I-131, or Short Lived Iodine Average.

~

l.

3.9 Record.the largest leak rate for each steam generator on Attachment 6.4 (Note 5.2).

3.10 Record the largest leak rate for each steam generator on the Steam Generator Activity Log Sheet.

3.11 Notify the Shift Technical Advisor of the results on Attachment 6.4

'4.0 Discussion 4.1 A problem that. can.occurLwith a nuclear power plant is a primary to secondary leak.

A leak can develop in a steam generator: Lprimary coolant enters and mixes with the bulk water in the steam generator, then 'the radioactive isotopes in the primary coolant enter the steam system. ' The particulate ae:ivity' will stay in the bulk water of 'the steam generator and tend to. concentrate; removal of the radioactive isotopes is accomplished by-natural decay and steam generator blowdown.

I l

l

CAF-4.0

..1 Page 1 of 1 00-00-00 PRIMARY TO SECONDARY LEAK ltATg mn1 tan 0N MTA SHRET UNIT RCS Spec.ific Activities (uci/ graal Sample Location:

Date: _

Time:

Na-24 =

g I-131 -

I-132 =

I-133 -

I-134 =

M I-135 =

S/G Specific Activities (uci/ gram)

S/G Blowdown Rates (gpa)

A=

B=

C=

Date:

Time:

S/G Na-24 I-131 I-132 I-133 I-134 I-135 A

B C

M hte (#)

(S/G Activity) 0.676 + 1 + (Blowdown) (0.0997)]

=

(RCS Activity) (9.19 x 10 )

Where:

Isotope A (day ~)

Na-24 1.11 I-131 0.0862 I-132 7.23 I-133 0.80 I-134 19.00 I-135 2.49

_ _ _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' - - - - - - - ~ - - - - - - - - - - - - - - - ~ ~

e CAP-4.0

,.2 Page 1 of 1 00-00-00 STABILITY CRITERIA ISOTOPE

  • STABILITY CRITERIA Na-24 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of Stable Power Level & 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Constant S/G Blowdown Rate I-131 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of Stable Power Level & 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> Constant S/G Blowdown Rate 1-132 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> of Stable Power Level & 18 boars Constant S/G Blowdown Rate I-133 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> = of Stable Power Level & 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> Constant S/G Blowdown Rate I-134
    • 0 hours of Stable Power Level 6
      • 0 hours Constant S/G Blowdown Rate I-135 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of Stable Power Level & 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Constant S/G Blowdown Rate Stability Criteria is 2 equivalent half-lives for each isotope, assuming 60 gpa letdown and 10 spa blowdown, and a spiking factor of 70.
    • Actual Stability criteria is 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of Stable Power Level, but use 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> for calculations.
      • Actual Stability Criteria is 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of Constant S/G Blowdown Rate, but use O hours for calculations.

i I

l

CAP-4.0

,.3 Page 1 of 1 00-00-00 PE.IMARY-TO-SumamaaY LEAK EATE CAIEUIATION imK3REET S/C 14ak Este Cale=1=ttee (ass)

  • A" STEAM CumutA'rtR ISOTOPE SIABILITY CRITERIA MET (7 or N)

LFAK RATE (gpd)

Na-24 I-131 I-132 I-133 I-34 1-135 SHORT LIVED IODINE AVERAGE LEAK RATE

  • n" STEAM CENERAT(E ISOTOPE STABILITY CRITERIA MET (Y or N)

LEAK RATE (apd)

Na-24 I-131 1-132 I-133 I-134 I-135 SHORT LIVED IODINE AVERACE LEAK RATE "C" STIAM CENERAT(R ISOTOPE STABILITY CRITERIA MET (Y or N)

LEAK RATE (gpd)

Na-24 I-131 I-132 1-133 I-134 I-135 t

SHORT LIVED 10DIhT AVERAGE LEAK RATE

~. -

i CAP-4.0

,'.4 Page 1 of 1 00-00-00 PRDWLY TO SEcosmalLY LEAK RATE MTS "A" s/G Leak Rate (g%)

"B" S/G Leak Rate (gpd)

"c" s/G Leak Rate (gpd) i

Basis for Primary to Secondary Leakage Limits Virginia Power has placed in effect five administrative primary-to-secondary leak rate Jimits and corresponding actions. These are:

1.

If leakage from an individus) steam generator increases by 50 gpd or more in approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or less, reduce reactor power to 50% or less within 90 minutes or less.

2.

If total leakage from all three steam generators increases by 60 spd or more in approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or less, reduce reactor power to 50%

or less within 90 minutes or less.

3.

If leakage from an individual steam generator is projected to exceed 100 gpd within 90 minutes or less, reduce reactor power to 50% or less in 90 minutes or less and be in Mode 3 (Hot Standby) in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or less.

4.

If total leakage f rom all three steam generators exceeds 300 gpd, be in Mode 3 (Hot Standby) in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or less.

5.

If leakage from an individual. steam generator exceeds 100 spd, be in Mode 3 (Hot Standby) in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or less.

These leakage limits are designed to provide effective response for three different tube degradation mechanisms. These three mechanisms are:

1.

The rapid propogation of a fatigue crack due to the presence of high alternating stresses (i.e. stresses greater than 7 ksi).

2.

The less rapid propogation of a f atigue crack dr to the presence of moderate to lov alternating stresses (i.e.,

stresses of 7 kai or less).

3.

The slow propogation of a corrosion crack.

A f atigue crack can be expected to initially penetrate thru-vall, then turn and propogate around the vall.

This would result in an step initiation of leakage followed by continually increasing leakage.

Leakage limits 1 and 2 above are designed to provide timely, effective response to initiation and propogation of a fatigue crack where very. high alternating stresses are present.

It should be noted that the modifiestions made to the North Anna steam generators make it extremely improbable that a tube _ vill be subjected to these levels of _ stress.

However, it is prudent that the leakage limits address this unlikely case.

Leakage limits 3, 4 and 5 above are designed to provide effective, timely response to the initiation and propogation of fatigue cracks where lower levels of iternating stresses are present and to corrosion cracks.

If the I

stresses are 7 KS1 or less, the absolute limit of 100 gpd allows sufficient time to take corrective action prior to a_ rupture, e

JWo/jlm/023/2PROJ t

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