ML13310B536

Rev 2 to Chemistry Procedure SO123-III-8.8, Alternate Methods of Post-Accident Parameter Sampling

ML13310B536
Person / Time
Site:	San Onofre
Issue date:	05/21/1984
From:	Mcwey B, Rosenblum R Southern California Edison Co
To:
Shared Package
ML13310B535	List: ML13310B534 ML13310B536
References
SO123-III-8.8, TAC-44478, NUDOCS 8407020323
Download: ML13310B536 (40)

Text

SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE S0123-III-8.8 UNITS 1, 2 AND 3 REVISION 2 PAGE 1 OF 3 COMPLETE REVISION 2 ]8 EFFECTIVE DATE ALTERNATE METHODS OF POST-ACCIDENT PARAMETER SAMPLING TABLE OF CONTENTS SECTION

!"AGE 1.0 OBJECTIVE 2

2.0 REFERENCES

2 3.0 PREREQUISITES MAY 1984 2

4.0 PRECAUTIONS ComSiTre 2

5.0 CHECK-OFF LIST(S) 2 6.0 PROCEDURE 2

7.0 RECORDS 3

8.0 ATTACHMENTS 3

PAGES CHANGED WITH THIS REVISION: ALL PREPARED BY:

&rob A Jqpa 4

AUTHOR DATE APPROVED BY:

T__

MANAGER, TECHNICAL 0279c/bl 4I702O323060 e4~F~

E ;PD _DOC3K50002 06,

SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE S0123-III-8.8 UNITS 1, 2 AND 3 REVISION 2 PAGE 2 OF 3 ALTERNATE METHODS OF POST-ACCIDENT PARAMETER SAMPLING 1.0 OBJECTIVE 1.1 To describe alternate methods of post-accident parameter sampling available for use in the event of a Post-Accident Sampling System (PASS) component malfunction.

2.0 REFERENCES

2.1 Station Procedures 2.1.1 S0123-III-8.3.1 (Unit 1)(to be issued) and S0123-III-8.3.23 (Units 2 and 3), Sampling Procedures and In-Line Analysis for the Post-Accident Sampling System 2.1.2 S0123-III-8.1, Post-Accident Sampling System Routine Surveillances 3.0 PREREQUISITES 3.1 Verify this procedure is the most current revision against a white control file and review all applicable Temporary Change Notices (TCNs).

3.2 A need exists to determine the value of a parameter provided by the PASS which cannot be obtained due to a PASS component malfunction'.

The designated alternate method(s) need not be verified unless the unit is inMode 1 or 2.

3.3. Verify that the PASS area has been monitored for oxygen immediately prior to initial entry.

4.0 PRECAUTIONS 4.1 A permanent Radiation Exposure Permit (REP) has been established for the performance of this procedure during accident conditions.

Check with Health Physics.

5.0 CHECK-OFF LIST(S) 5.1 None 6.0 PROCEDURE 6.1 In the event of a Unit PASS component malfunction during a routine surveillance or an accident condition, Attachment 8.1 (Unit 1) and.4 (Units 2 and 3) provide alternate methods for post-accident sampling and analysis.

SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE SO123-III-8.8 UNITS 1, 2 AND 3 REVISION 2 PAGE 3 OF 3.

6.0 PROCEDURE (Continued) 6.2 All analyses should be performed in the Radiochemistry Laboratory if local radiation levels permit. If local radiation levels prevent access to the affected Radiochemistry Laboratory, perform the analyses.in the nonaffected Radiochemistry Laboratory.

6.3 Record all data obtained by the alternate methods in the applicable section of Reference 2.1.1 during an accident condition. During a routine surveillance, record in Reference 2.1.2 that the specified alternate method is available.

7.0 RECORDS 7.1 All recorded data and calculations shall be reviewed by the Chemistry Foreman and approved by the Unit Chemistry Supervisor upon completion. Forward to CDM after completed review.

8.0 ATTACHMENTS 8.1 Alternate Methods of Post-Accident Parameter Sampling -

Unit 1

'8.2 Correlation of Dose Rate to Reactor Coolant Activity -

Unit 1 8.3 Boron Dilution Calculation - Unit 1 8.4 Alternate Methods of Post-Accident Parameter Sampling - Units 2 and 3 8.5 Correlation of Dose Rate to Reactor Coolant Activity -

Units 2 and 3 8.6 Boron Dilution Calculation -

Units 2 and 3 0279c

SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE S0123-111-8.8 UNITS 1, 2 AND 3 REVISION 2 PAGE 1 OF 1 ATTACHMENT 8.1 UNIT 1 ALTERNATE METHODS OF POST-ACCIDENT PARAMETER SAMPLING PASS FUNCTION ALTERNATE METHOD ANALYSIS/REMARKS

1.

Containment Atmosphere A.

Hydrogen

1.

Obtain H2 concentration from meters

1.

Recorded manually.

Al-H2-2001 and Al-H2-3001 in the Control Room.

2.

Grab sample from PASS or effluent monitor

2.

Spectrum Analysis R1211/1212 using S0123-I-5.3.1, Radioactive Gas Sampling and Analysis.

B.

RadionuclIdes

1.

Correlate the dose rates in Containment as

1.

See Attachment 8.2 read on radiation monitor R1255 and R1257 to reactor coolant activity

2.

Grab sample from PASS or effluent monitor

2.

Spectrum Analysis R1211/1212 using S0123-111-5.3.1,.Radioactive Gas Sampling and Analysis.

II. Reactor Coolant A.

Hydrogen

1.

Grab sample from PASS or Radlochemistry,

1.

Gas Chromatography Sample Room B.

Oxygen

1.

Grab sample from PASS or Radiochemistry

1.

Gas Chromatography Sample Room (not mandatory)

C.

Total Gas

1.

Grab sample from Radiochemistry Sample Room

1.

Gas Chromatography

2.

Determine maximum dissolved gas in Reactor

2.

Pressure and Temperature recorded Coolant using pressure and temperature from on the Fox-3 Computer.

meters PR-425 (PT425) and TR-402 (TE-402C, TE-412C, TE-422C),

TI-2422A, TI-2401A, TI-2412A, TI-3402A, TI-3411A, or TI-3421A.

D.

Boron

1.

Grab sample from PASS or Radiochemistry

1.

Curcumin Method or equivalent Sample Room

2.

Calculate final boron concentration from

2.

See Attachment 8.3.

pre-accident concentration and dilution from safety injection.

E.

pH

1.

Grab sample from PASS or Radlochemistry

1.

pH Meter Sample Room F.

Radionuclides

1.

Correlate the dose rates in Containment as

1.

See Attachment 8.2 read on radiation monitor R1255 and R1257 to reactor coolant activity

2.

Grab sample from PASS or Radiochemistry

2.

Spectrum Analysis Sample Room G.

Chlorides

1.

Grab sample from PASS or Radiochemistry

1.

Mercuric Nitrate Titration Sample Room

2.

Ion Chromatography NOTE:

TG or H2 satisfies requirement.

SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE SO123-III-8.8 UNITS 1, 2 AND 3 REVISION 2 PAGE 1 OF 3 ATTACHMENT 8.2 UNIT 1 CORRELATION OF DOSE RATE TO REACTOR COOLANT ACTIVITY The dose rates at the high range in containment area radiation monitors have been correlated to reactor coolant activity levels. Curves of dose rate versus time are provided for the following conditions:

1% Failed Fuel -

Assumes that 1% of total core activity is available for release (page 3 of 3).

10% Failed Fuel -

10 times the 1% case.

Gap Activity - Assumes release of 100% of Gap-in-Fuel inventory.

Extrapolation between these curves will provide an indication of degraded core conditions.

0279c

SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE SO123-III-8.8 UNITS 1, 2 AND 3 REVISION 2 PAGE 2 OF 3 ATTACHMENT 8.2 DOSE RATE VS TIME FOR HIGH RANGE IN-CONTAINMENT MONITORS (1255/1257) 108 105 10% Failed Fuel S 1046 1% Failed Fuel i6 01 4

6 1

10 It 14 16 18 30 2?

24 T1 ME, howrs

SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE S0123-III-8.8 UNITS 1, 2 AND 3 REVISION 2 PAGE 3 OF 3 ATTACHMENT 8.2 Maximum Fission and Corrosion Product Isotopic Inventories of Reactor Coolant for Reference One Percent Fuel Defect Case (Based upon 570aF Coolant)

Coolant.

Specific Activity Inventory, Isotope Microcuries/cc Curies Br-84 0.16 31 Kr-85 4.2 800 I

Kr-85 (m) 1.3 250 I

Kr-87 0.73 140 l

Kr-88 2.1 400 Rb-88 2.1 400 Rb-89 5.Ox10-2 10 Sr-89 2.Ox10-1 0.4 I Sr-90 1.0x10-1 0.02 I Y-90 1.9x10-'

0.04 I Sr-91 1.Ox10-I 0.2 I Y-91 2.8x10" 2 l

5.3 Mo-99 2.5 l

477 l

Te-129 1.6x10-2 3

1 I-129 2.4x10-'

0.5x10-1 I I-131 1.4 l

267 I Te-132 0.13 l

25 I

I 1-132

.0.50 j

95 l

I 1-133 1.9 l

363 l

Xe-133 158 I

30.200 I

Te-134 1.4x10-2 2.7 l

1-134 0.26 50 Cs-134 0.68 130 I1-135 0.99 189 Xe-135 4.4 840 Cs-136 7.xO10- 2 13 l

Cs-137 15.7 3000 lXe-138 0.32 I

61 l

Cs-138 0.52 I

100 I Ba-140 2.1x10-3 l

0.4 La-140.

7.1x10-*

I 0.1 l

Co-60 1.5xlO-'

I 0.29 l

Fe-59 1.9x10-'

l 0.36 l

Co-58 8.5x10-'

I 1.6 l

Mn-56 2.3x10-2 l

4.4 I Mn-54 4.4x10-1 0.84

SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE S0123-111-8.8 UNITS 1, 2 AND 3 REVISION 2 PAGE 1 OF 1 ATTACHMENT 8.3 UNIT 1 BORON DILUTION CALCULATION A

B C

Boron Level Concentration Change Product LOCATION (ppm)

(gallons)

(AX B)

Boric Acid Storage Tank (BAS-D-5) ppm X

gal

=

NOTE:

This calculation assumes Refueling Water complete mixing of reactor Storage Tank (CRS-D-1) ppm X

gal

=

coolant and containment sump after LPSI/HPSI Auxiliary Feedwater as recirculation resulting In Storage Tank (AFW-D-2A) 0 ppm X

gal a conservative value for the calculated final boron Condensate Storage e concentration.

Tank (CND-D-2) 0 ppm X

gal

=

• Pre-Accident Boron Pre-Accident concentration levels RCS ppm X

35550 gal

=

can be obtained from the Reactor Laboratory TOTAL Analysis Book.

The effects of these tanks will only be considered If a steam line ruptures Inside containment.

Final Boron Concentration = Total Column (C) =_

=

ppm Total Column (B) 0279c

SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE S0123-111-8.8 UNITS 1, 2 AND 3 REVISION 2 PAGE 1 OF 1 ATTACHMENT 8.4 UNIT 2 AND 3 ALTERNATE METHODS OF POST-ACCIDENT PARAMETER SAMPLING PASS FUNCTION ALTERNATE METHOD ANALYSIS/REMARKS

1.

Containment Atmosphere A.

Hydrogen

1.

Obtain H2 concentration from meters

1.

Recorded on the plant.monitoring 2A18108 and 2A18118 or 3A18108 and 3A18118 computer.

in the Control Room.

2.

Grab sample from PASS or Effluent Monitor

2.

Gas Chromatography 780 4 using S0123-1ll-5.3.23, Radioactive Gas Sampling and Analysis.

B.

Radionuclides

1.

Correlate the dose rates in Containment as

1.

See Attachment 8.5 read on radiation monitors 2T7820-1 and 2T7820-2 or 3T7820-1 and 3T7820-2 to reactor coolant activity.

2.

Grab sample from PASS or Effluent Monitor

2.

Spectrum Analysis 7804 using S0123-111-5.3.23, Radioactive Gas Sampling and Analysis.

II. Reactor Coolant A.

Hydrogen

1.

Grab sample from PASS or Radlochemistry

1.

Gas Chromatography Laboratory.

B.

Oxygen

1.

Grab sample from PASS or Radiochemistry

1.

Gas Chromatography Laboratory (not mandatory).

C.

Total Gas

1.

Grab sample from Radiochemistry Laboratory.

1.

Gas Chromatography Laboratory.

2.

Determine maximum dissolved gas In Reactor

2.

Pressure and temperature recorded Coolant using pressure and temperature from on plant monitoring computer.

meters 2T1-0911-1 and 2TI-0921-2 or 3TI-0911-1 and 3TI-0921-2.

D.

Boron

1.

Grab sample from PASS or Radiochemistry

1.

Curcumin Method or equivalent

~

Laboratory.4

2.

Calculate final boron concentration from

2.

See Attachment 8.6 pre-accident concentration and dilution from safety injection.

E.

pH

1.

Grab sample from PASS or Radiochemistry

1.

pH Meter Laboratory.

F.

Radionuclides

1.

Correlate the dose rates in Containment as

1.

See Attachment 8.5 read on radiation monitors 2T7820-1 and 2T7820-2 or 3T7820-1 and 3T7820-2 to reactor

2.

Grab sample from PASS or Radiochemistry

2.

Spectrum Analysis Laboratory.

G.

Chlorides

1.

Grab sample from PASS or Radiochemistry

1.

Mercuric Nitrate Titration Laboratory.

2.

Ion Chromatography NQTE:

TG or H2 satisfies requirement.

SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE S0123-III-8.8 UNITS 1, 2 AND 3 REVISION 2 PAGE 1 OF 5 ATTACHMENT 8.5 UNIT 2 AND 3 CORRELATION OF DOSE RATE TO REACTOR COOLANT ACTIVITY The does rates at the high range in containment area radiation monitors have been correlated to reactor coolant activity levels as shown in FSAR Figures 12.3-68 (page 2 of 5) and 12.3-69 (page 3 of 5).

Curves of dose rate versus time are provided for the following conditions:

RCS Average -

100% of core average activity from FSAR Table 11.1-3 (page 4 of 5) released to containment.

RCS Maximum -

100% of core average activity from FSAR Table 11.1-2 (page 5 of 5) released to containment.

1% Failed Fuel - Assumes that 1% of total core activity is available for release and of that 1%, 100% of the noble gases, 50% of the halogens and 1% of the other isotopes are released.

10% Failed Fuel -

10 times the 1% case.

Gap Activity - Utilizes Regulatory Guide 1.25 assumptions with 10% of the core Xenon and Krypton, 30% of~the-Krypton 85 and 10% of the lodines released.

LOCA - Utilities Regulatory Guide 1.4 assumptions with 100% of the noble gases, 50% of the halogens and 1% of the other isotopes released.

Extrapolation between these curves will provide an indication of degraded core conditions.

0279c

SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE S0123-11I-8.8 UNITS 1, 2 AND 3 REVISION 2 PAGE 2 OF 5 ATTACHMENT 8.5 DOSE RATE VS TIME FOR HIGH RANGE IN-CONTAINMENT MONITOR AT EL. 99'6" ON SECONDARY SHIELD WALL 10 IDS CAP I1 to NFAIRED FU L 0

FIGURE5 104 R CS AVERAGE 103 102 0

2 4

6 1

12 14 18 18 20 22 24 TIME HMRS)

FIGURE 1

SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE S0123-III-8.8 UNITS 1, 2 AND 3 REVISION 2 PAGE 3 OF 5 ATTACHMENT 8.5 DOSE RATE VS TIME FOR HIGH RANGE IN-CONTAINMENT MONITOR AT EL. 94' ABOVE THE ELEVATOR SHAFT 100 10 10

_____._RCS AVERAGE 10 0

2 4

6 8

10 12 14 16 18 20 22 24 RME (HIRS)

FIGURE 2

SA NFENUCLEAR GENERATING STATION CHEMISTRY PROCEDURE S0123-III-8.8 UNITS 1, 2 AND 3 REVISION 2 PAGE 4 OF 5 ATTACHMENT 8.5 AVERAGE REACTOR COOLANT RlADIOISOTOPE CONCENTRATION (a)

(No Gas Stripping)

Nucid I Activity Activity Nucid (v~Ci/cm2 )

Nuclide I

(Pci/cm')(b I H-3 I

9.7E-1 II I Br-84 I

2.65E-3 I

1-134 I

4.88E-2I I Kr-85m I

2.20E-1 I

Cs-134 I

4.03E-2 I Kr-85 I

1.50E-1 I

1-135 I

2.26E-1I I Kr-87 I

6.OOE-2 I

Cs-136 I

2.03E-2I I Kr-88 I

2.OOE-1 I

Cs-137 I

2.90E-2I I Rb-88 I

2.02E-1 I

Xe-131m I

1.1OE-1I I Sr-89 I

5.95E-4 I

Xe-133 I

1.80E1I I Sr-90 I

1.71E-5 I

Xe-135 I

3.50E-1I Y-90 I

2.64E-5 I

Xe-135m I

1.30E-2 I Y-91 I

3.42E-2 I

Xe-138 I

4.40E-2 I Y-91m 3.91E-4 I

Ba-140 I

3.67E-4I I,Sr-91 I

8.06E-4 I

La-140 I

2.22E-4I I MO-99 I

6.98E-1 I

Pr-143 I

8.35E-15 I Ru-103 I

7.70E-5 I

Ce-144 I

5.64E-5 I Ru-106 I

1.71E-5 I

Cr-51 I

3.36E-3I I Te-129 I

1.66E-3 I

Mn-54 I

5.48E-4 I 1-131 I

4.56E-1 I

Zr-95 I

1.03E-4I 1-132 I

1.08E-1 I

Co-60 I

3.54E-3 I Te-132 I

4.24E-2 I

Fe-59 I

1.76E-3 I 1-133 I

5.28E-1 I

Co-58 I

2.82E-2 (a) From draft N237 Standard (b) At 70OF

-SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE SO123-III-8.8 UNITS 1, 2 AND 3 REVISION 2 PAGE 5 OF 5 ATTACHMENT 8.5 MAXIMUM REACTOR COOLANT RADIOISOTOPE CONCENTRATION ONE PERCENT FAILED FUEL, NO GAS STRIPPING l

Activity j

Activity Nuclide (iCi/cm )(a)

Nuclide (pCi/cm')

II I-3I H-3 l

1.8E0 Br-84 l

3.97E-2 1-134 5.73E-1 Kr-85m l

2.35E0 Cs-134 1.53E-1 Kr-85 5.13E0 l

1-135 l

2.58E0 l

Kr-87 l

1.26E0 I

Cs-136 l

2.64E-2 Kr-88 l

4.08E0 l

Cs-137 l

6.20E-1 I

II l

Rb-88 l

4.06E0 l

Xe-131m 2.39E0 I ~

II I

I Sr-89 1.07E-2 Xe-133 3.34E2 Sr-90 5.79E-4 Xe-135m 1.10EO Y-90 1.42E-3 Xe-135 9.24E0 Y-91 4.84E-2 l

Xe-138 5.58E-1 I

III Y-91m 3.85E-3 l

Ba-140 I

1.27E-2 Sr-91 6.15E-3 La-140 1.23E-2

'Mo-99 2.24E0 I

Pr-143 1.13E-2 I

I I

I Ru-103 l

8.53E-3 I

Ce-144 l

7.82E-3 II I

I Ru-106 I

5.12E-4 I

Cr-51 l

3.36E-3 I

I Te-129 l

4.98E-2 I

Mn-54

.5.48E-4 I1-131 I

4.67EO Zr-95 1.15E-2 I

I I

I1-132 l

1.32E0 Co-60 l

3.54E-3 Te-132 I

6.48E-1 Fe-59 l

1.76E-3 I1-133 I

5.88E0 Co-58 l

2.82E-2 (a) At 70aF

SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE S0123-111-8.8 UNITS 1, 2 AND 3 REVISION 2 PAGE 1 OF 1 ATTACHMENT 8.6 UNIT 2 AND 3 BORON DILUTION CALCULATION A

B C

Boron Level Concentration Change Product LOCATION (ppm)

(gallons)

(A X B)

Boric Acid Makeup Tank T071 ppm X

gal

=

NOTE:

This calculation assumes Boric Acid complete mixing of reactor Makeup Tank T072 ppm X

gal

=

coolant and containment sump after LPSI/HPSI recirculation resulting In a Safety Injection conservative value for the Tank T007 ppm X

gal

=

calculated final boron concentration.

Safety Injection' Tank TOOB Dpm X

gal

=

Pre-Accident Boron concentration levels can Safety Injection be obtained from the Tank T009 ppm X

gal

=

Reactor Laboratory Analysis Book.

Safety Injection Tank T010 ppm X

gal Level changes can be

-obtained from the Refueling Water plant monitoring Storage Tank T005 ppm X

gal

=

computer.

Refueling Water Storage Tank T006 ppm X

gal

=

initial Containment Sump Level ppm X

ga I

=

Pre-accident RCS ppm X

83,039 gal

=

TOTAL Final Boron Concentration = Total Column (C) =

ppm Total Column (B) 0279c

ENCLOSURE 3 SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE S0123-III-8.8 UNITS 1, 2 AND3 REVISION 2 PAGE 1 OF 3 COMPLETE REVISION my 21 184 EFFECTIVE DATE ALTERNATE METHODS OF POST-ACCIDENT PARAMETER SAMPLING TABLE OF CONTENTS SECTION 1.0 OBJECTIVE 2

2.0 REFERENCES

.2 3.0 PREREQUISITES MAY 1984' 2

4.0 PRECAUTIONS COMSITe 2

5.0 CHECK-OFF LIST(S) 2 6.0 PROCEDURE-2 7.0 RECORDS 3

8.0 ATTACHMENTS 3

PAGES CHANGED WITH THIS REVISION: ALL PREPARED BY: &

Lqo4 AUTHOR APPROVED BY:

Act ROSENBLUM DATE MANAGER, TECHNICAL 0279c/bl

SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE SO123-III-8.8 UNITS 1, 2 AND 3 REVISION 2 PAGE 2 OF 3 ALTERNATE METHODS OF POST-ACCIDENT PARAMETER SAMPLING 1.0 OBJECTIVE 1.1 To describe alternate methods of post-accident parameter sampling available for use in the event of a Post-Accident Sampling System (PASS) component malfunction.

2.0 REFERENCES

2.1 Station Procedures 2.1.1 S0123-III-8.3.1 (Unit 1)(to be issued) and SO123-III-8.3.23 (Units 2 and 3), Sampling Procedures and In-Line Analysis for the Post-Accident Sampling System 2.1.2 S0123-III-8.1, Post-Accident Sampling System Routine Surveillances 3.0 PREREQUISITES 3.1 Verify this procedure is the most current revision against a white control file and review all applicable Temporary Change Notices (TCNs).

3.2 A need exists to determine the value of a parameter provided by the PASS.which cannot be obtained due to a PASS component malfunction.

The designated alternate method(s) need not be verified unless the unit is in Mode 1 or 2.

3.3. Verify that the PASS area has been monitored for oxygen immediately prior to initial entry.

4.0 PRECAUTIONS 4.1 A permanent Radiation Exposure Permit (REP) has been established for the performance of this procedure during accident conditions.

Check with Health Physics.

5.0 CHECK-OFF LIST(S) 5.1 None 6.0 PROCEDURE 6.1 In the event of a Unit PASS component malfunction during a routine surveillance or an accident condition, Attachment 8.1 (Unit 1) and.4 (Units 2 and 3) provide alternate methods for post-accident sampling and analysis.

SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE S0123-III-8.8 UNITS 1, 2 AND 3 REVISION 2 PAGE 3 OF 3 6.0 PROCEDURE (Continued) 6.2 All analyses should be performed in the Radiochemistry Laboratory if local radiation levels permit. If local radiation levels prevent access to the affected Radiochemistry Laboratory, perform the analyses in the nonaffected Radiochemistry Laboratory.

6.3 Record all data obtained by the alternate methods in the applicable section of Reference 2.1.1 during an accident condition. During a routine surveillance, record in Reference 2.1.2 that the specified alternate method is available.

7.0 RECORDS 7.1 All recorded data and calculations shall be reviewed by the Chemistry Foreman and approved by the Unit Chemistry Supervisor upon completion. Forward to CDM after completed review.

8.0 ATTACHMENTS 8.1 Alternate Methods of Post-Accident Parameter Sampling -

Unit 1

'8.2 Correlation of Dose Rate to Reactor Coolant Activity -

Unit 1 8.3 Boron Dilution Calculation - Unit 1 8.4 Alternate Methods of Post-Accident Parameter Sampling -

Units 2 and 3 8.5 Cortelation of Dose Rate to Reactor Coolant Activity -

Units 2 and 3 8.6 Boron Dilution Calculation -

Units 2 and 3 0279c

SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE 80123-111-8.8 UNITS 1, 2 AND 3 REVISION 2 PAGE 1 OF 1 ATTACHMENT 8.1 UNIT 1 ALTERNATE METHODS OF POST-ACCIDENT PARAMETER SAMPLING PASS FUNCTION ALTERNATE METHOD ANALYSIS/REMARKS I. Containment Atmosphere A.

Hydrogen

1.

Obtain H2 concentration from meters

1.

Recorded manually.

Al-H2-2001 and Al-H2-3001 in the Control Room.

2.

Grab sample from PASS or effluent monitor

2.

Spectrum Analysis R1211/1212 using S0123-1ll-5.3.1, Radioactive Gas Sampling and Analysis.

B.

Radionuclides

1.

Correlate the dose rates in Containment as

1.

See Attachment 8.2 read on radiation monitor R1255 and R1257 to reactor coolant activity

2.

Grab sample from PASS or effluent monitor

2.

Spectrum Analysis R1211/1212 using S0123-111-5.3.1, Radioactive Gas Sampling and Analysis.

II. Reactor Coolant A.

Hydrogen

1.

Grab sample from PASS or Radiochemistry

1.

Gas Chromatography Sample Room B.

Oxygen,

1.

Grab sample from PASS or Radiochemistry

1.

Gas Chromatography Sample Room (not mandatory)

C.

Total Gas

1.

Grab sample from Radiochemistry Sample Room

1.

Gas Chromatography

2.

Determine maximum dissolved gas in Reactor

2.

Pressure and Temperature recorded Coolant using pressure and temperature from on the Fox-3 Computer.

meters PR-425 (PT425) and TR-402 (TE-402C, TE-412C, TE-422C),

TI-2422A, TI-2401A, TI-2412A, TI-3402A, TI-3411A, or TI-3421A.

D.

Boron

1.

Grab sample from PASS or Radlochemistry

1.

Curcumin Method or equivalent Sample Room

2.

Calculate final boron concentration from

2.

See Attachment 8.3.

pre-accident concentration and dilution from safety injection.

E.

pH

1.

Grab sample from PASS or Radlochemistry

1.

pH Meter Sample Room F.

Radionuclides

1.

Correlate the dose rates in Containment as

1.

See Attachment 8.2 read on radiation monitor R1255 and R1257 to reactor coolant activity

2.

Grab sample from PASS or Radlochemistry

2.

Spectrum Analysis Sample Room G.

Chlorides

1.

Grab sample from PASS or Radiochemistry

1.

Mercuric Nitrate Titration Sample Room

2.

ion Chromatography NOTE:

TG or H2 satisfies requirement.

SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE SO123-III-8.8 UNITS 1,-2 AND 3 REVISION 2 PAGE 1 OF 3 ATTACHMENT 8.2 UNIT 1 CORRELATION OF DOSE RATE TO REACTOR COOLANT ACTIVITY The dose rates at the high range in containment area radiation monitors have been correlated to reactor coolant activity levels. Curves of dose rate versus time are provided for the following conditions:

1% Failed Fuel - Assumes that 1% of total core activity is available for release (page 3 of 3).

10% Failed Fuel -

10 times the'1% case.

Gap Activity - Assumes release of 100% of Gap-in-Fuel inventory.

Extrapolation between these curves will provide an indication of degraded core conditions.

0279c

SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE 50123-III-8.8 UNITS 1, 2 AND 3 REVISION 2 PAGE 2 OF 3 ATTACHMENT 8.2 DOSE RATE VS TIME FOR HIGH RANGE IN-CONTAINMENT MONITORS (1255/1257) 10% Failed Fuel 104

-GAP__

GAP 1% Failed Fuel 1

2 4 4

1 10 It 14 16 18 30 22 24 TIME, hows

SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE S0123-III-8.8 UNITS 1, 2 AND 3 REVISION 2 PAGE 3 OF 3 ATTACHMENT 8.2 Maximum Fission and Corrosion Product Isotopic Inventories of Reactor Coolant for Reference One Percent Fuel Defect Case (Based upon 570aF Coolant)

Coolant Specific Activity Inventory, Isotope Microcuries/cc Curies Br-84

.0.16 31 Kr-85 4.2 800 Kr-85 (m) 1.3 250 Kr-87 0.73 140 Kr-88 2.1 400 Rb-88 2.1 400 Rb-89 5.0x10-2 10 Sr-89 2.0x10-1 0.4 Sr-90 1.0x10-'

0.02 Y-90 1.9x10-1 0.04 Sr-91 1.0x10-0.2 Y-91 2.8x10-1 5.3 Mo-99 2.5 477 Te-129 1.6x10-*

3 I1-129 2.4x10-0.5x10 -I I1-131 1.4 267 Te-132

.0.13 25 1-132 0.50 95 l

I1-133 1.9 363 I

Xe-133 158 30.200 I

Te-134 1.4x10-1 2.7 l

1-134 0.26 50 Cs-134 0.68 130 I1-135 0.99 189 Xe-135 4.4 840 Cs-136 7.0x102-13 Cs-137 15.7 3000 Xe-138 0.32 61 Cs-138 0.52 100 Ba-140 2.1x10-1 0.4 La-140.

7.1x10-'

0.1 Co-60 1.5x10-1 0.29 Fe-59 1.9x10-0.36 Co-58 8.5x10-3 1.6 Mn-56 2.3x10-2 4.4 I Mn-54 4.4x10-3 0.84

SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE S0123-111-8.8 UNITS 1, 2 AND 3 REVISION 2 PAGE 1 OF 1 ATTACHMENT 8.3 UNIT 1 BORON DILUTION CALCULATION A

B C

Boron Level Concentration Change Product LOCATION (ppm)

(qallons)l (AX B)

Boric Acid Storaqe Tank (BAS-D-5) ppm X

gal NOTE:

This calculation assumes Refueling Water complete mixing of reactor Storaqe Tank ICRS-D-1) ppm X

gal

=

coolant and containment sump after LPSI/HPSI Auxiliary Feedwater we recirculation resulting in Storaqe Tank (AFW-D-2A) 0 ppm X

gal

=

a conservative value for the calculated final boron Condensate Storage ?

concentration.

Tank (CND-D-2) 0 pom X

gal

• Pre-Accident Boron Pre-Accident concentration levels RCS ppm X

35550 qal

=

can be obtained from the Reactor Laboratory TOTAL Analysis Book.

The effects of these tanks will only be considered if a steam line ruptures inside containment.

Final Boron Concentration = Total Column (C) =

ppm Total Column (B) 0279c

SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE S0123-lIl-8.8 UNITS 1, 2 AND 3 REVISION 2 PAGE 1 OF 1 ATTACHMENT 8.4 UNIT 2 AND 3 ALTERNATE METHODS OF POST-ACCIDENT PARAMETER SAMPLING PASS FUNCTION ALTERNATE METHOD ANALYSIS/REMARKS I. Containment Atmosphere A.

Hydrogen

1.

Obtain H2 concentration from meters

1.

Recorded on the plant monitoring 2A18108 and 2A18118 or 3A18108 and 3A18118 computer.

in the Control Room.

2.

Grab sample from PASS or Effluent Monitor

2.

Gas Chromatography 7804 using S0123-1II-5.3.23, Radioactive Gas Sampling and Analysis.

B.

Radionuclides

.1.

Correlate the dose rates in Containment as

1.

See Attachment 8.5 read on radiation monitors 2T7820-1 and 2T7820-2 or 3T7820-1 and 3T7820-2 to reactor coolant activity.

2.

Grab sample from PASS or Effluent Monitor

2.

Spectrum Analysis 7804 using SO123-1I1-5.3.23, Radioactive Gas Sampling and Analysis.

II. Reactor Coolant A.

Hydrogen

1.

Grab sample from PASS or Radlochemistry

1.

Gas Chromatography Laboratory.

8.

Oxygen

1.

Grab sample from PASS or Radiochemistry

1.

Gas Chromatography Laboratory (not mandatory).

C.

Total Gas

1.

Grab sample from Radlochemistry Laboratory.

1.

Gas Chromatography Laboratory.

2.

Determine maximum dissolved gas in Reactor

2.

Pressure and temperature recorded Coolant using pressure and temperature from on plant monitoring computer.

meters 2TI-0911-1 and 2TI-0921-2 or 3TI-0911-1 and 3TI-0921-2.

D.

Boron

1.

Grab sample from PASS or Radiochemistry

1.

Curcumin Method or equivalent Laboratory.

2.

Calculate final boron concentration from

2.

see Attachment 8.6 pre-accident concentration and dilution from safety injection.

E.

pH

1.

Grab sample from PASS or Radiochemistry

1.

pH Meter Laboratory.

F.

Radionuclides

1.

Correlate the dose rates in Containment as

1.

See Attachment 8.5 read on radiation monitors 2T7820-1 and 2T7820-2 or 3T7820-1 and 3T7820-2 to reactor

2.

Grab sample from PASS or Radlochemistry

2.

Spectrum Analysis Laboratory.

G.

Chlorides

1.

Grab sample from PASS or Radiochemistry

1.

Mercuric Nitrate Titration Laboratory.

2.

Ion Chromatography NQTE:

TG or H2 satisfies requirementA e

SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE SO123-III-8.8 UNITS 1, 2 AND 3 REVISION 2 PAGE 1 OF 5 ATTACHMENT 8.5 UNIT 2 AND 3 CORRELATION OF DOSE RATE TO REACTOR COOLANT ACTIVITY The does rates at the high range in containment area radiation monitors have been correlated to reactor coolant activity levels as shown in FSAR Figures 12.3-68 (page 2 of 5) and 12.3-69 (page 3 of 5).

Curves of dose rate versus time are provided for the following conditions:

RCS Average -

100% of core average activity from FSAR Table 11.1-3 (page 4 of 5) released to containment.

RCS Maximum -

100% of core average activity from FSAR Table 11.1-2 (page 5 of 5) released to containment.

1% Failed Fuel - Assumes that 1% of total core activity is available for release and of that 1%, 100% of the noble gases, 50% of the halogens and 1% of the other isotopes are released.

10% Failed Fuel -

10 times the 1% case.

Gap Activity -

Utilizes Regulatory Guide 1.25 assumptions with 10% of-the core Xenon and Krypton, 30% of the -Krypton 85 and 10% of the lodines released.

LOCA - Utilities Regulatory Guide 1.4 assumptions with 100% of the noble gases, 50% of the halogens and 1% of the other isotopes released.

Extrapolatton between these curves will provide an indication of degraded core conditions.

0279c

SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE S0123-III-8.8 UNITS 1,.2 AND 3 REVISION 2 PAGE 2 OF 5 ATTACHMENT 8.5 DOSE RATE VS TIME FOR HIGH RANGE IN-CONTAINMENT MONITOR AT EL. 99'S" ON SECONDARY SHIELD WALL 10 toe' E 10-CA 107 9FAILED FUEL 106 Cc 10

-CS MAXIMUM 104 CS AVERA E 102 0

2 4

6 S

j 12 14 18 18 20 22 24 TIME (HRS)

FIGURE 1

SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE S0123-III-8.8 UNITS 1, 2 AND 3 REVISION 2 PAGE 3 OF 5 ATTACHMENT 8.5 DOSE RATE VS TIME FOR HIGH RANGE IN-CONTAINMENT MONITOR AT EL. 94' ABOVE THE ELEVATOR SHAFT 10 LOCA 10~

~____

GAP 0FAILED FUE 10

• 4..

ACSMAXIMUM 10 10RCSAVERAG 0

2 6

8 li 12 14 16 18 20 22 24 WIME (MRS)

FT.IIRF 7

SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE SO123-III-8.8 UNITS 1, 2 AND 3 REVISION 2 PAGE 4 OF 5 ATTACHMENT 8.5 (a)

AVERAGE REACTOR COOLANT RADIOISOTOPE CONCENTRATION (No Gas Stripping)

Activity Activity Nuclide (uCi/cm 3)(b)

Nuclide (vCi/cm2)(b)

H-3 9.7E-1 Br-84 I

2.65E-3 I-134 4.88E-2 Kr-85m 2.20E-1 Cs-134 4.03E-2 Kr-85 1.50E-1 I-135 2.26E-1I Kr-87 6.00E-2 Cs-136 2.03E-2 Kr-88 2.OOE-1 Cs-137 2.90E-2I

-,Rb-88 l

2.02E-1 Xe-131m 1.10E-1 l

Sr-89 l

5.95E-4 Xe-133 1.80E1 I

I I

I Sr-90 l

1.71E-5 Xe-135 3.50E-1 I

IIII l

Y-90 2.64E-5 Xe-135m 1.30E-2 Y-91 3.42E-2 Xe-138 4.40E-2 Y-91m l

3.91E-4 Ba-140 3.67E-4 Sr-91 I

8.06E-4 La-140 2.22E-4 Mo-99 I

6.98E-1 I

Pr-143 8.35E-5 I

I I

II Ru-103 l

7.70E-5 I

Ce-144 I

5.64E-5 I

I II Ru-106 I

1.71E-5 I

Cr-51 3.36E-3 Te-129 l

1.66E-3 Mn-54 5.48E-4 I.I I

I I

l 1-131 I

4.56E-1 I

Zr-95 1.03E-4 I

• I I

I I

I1-132 l

1.08E-1 I

Co-60 I

3.54E-3 II I

I I

Te-132 l

4.24E-2 I

Fe-59 l

1.76E-3 I

I I

lI I

I I

1-133 I

5.28E-1 I

Co-58 l

2.82E-2 I

(a)

From draft N237 Standard (b)

At 70*F

SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE S0123-III-8.8 UNITS 1, 2 AND,3 REVISION 2 PAGE.5 OF 5 ATTACHMENT 8.5.

MAXIMUM REACTOR COOLANT RADIOISOTOPE CONCENTRATION ONE PERCENT FAILED FUEL, NO GAS STRIPPING I

I Activity I

Activity I Nuclide I

(iiCi/cm2 ) (a)

Nuclide U

lCi/CM3)(a)

I H-3 I

1.8E0 Br-84

[

3.97E-2 I

1-134 I

5.73E-1 I Kr-85m I

2.35E0 Cs-134 I

1.53E-1I I Kr-85 I

5.13E0 1-135 I

2.58E0 I Kr-87 I

1.26E0 Cs-136 I

2.64E-2 Kr-88 I

4-.08E0 Cs-137 I

6.20E-1 I Rb-88 I

4.06E0 Xe-131m I

2.39E0 Sr-89 I

1.07E-2 I

Xe-133 3.34E2 I S-90 I

.9-4 I

X-135m I

1.10EOI r

5.79EIXe.

I Y-90

.1.42E-3 I

Xe-135 I

9.24E0 I Y-91 I

4.84E-2 I

Xe-138 I

5.58E-1I I Y-91m I

3.85E-3.I Ba-140 I

1.27E-2 I Sr-91 I

6.15E-3 I

La-140 1.23E -2I I Mo-99 I

2.24E0 Pr-143 I

1.13E-2 I

I Ru-103 I

8.53E-3 I

Ce-144 I

7.82E-3 I Ru-1.06 I

5.12E-4 *I Cr-51 I

3.36E-3I I Te-129 I

4.98E-2 I

Mn-54 I

5.48E-4I 1-I131 j

4.67E0 Zr-95 I

1.15E-2 1-132 I

1.32E0 Co-60 I

3.54E-3I I Te-132 I

6.48E-1 I

Fe-59 I

1.76E-3I I 1-133 I

5.88E0 Co-58 I

2.82E-2I

• (a) At 70*F

SAN ONOFRE NUCLEAR GENERATING STATION CHEMISTRY PROCEDURE S0123-111-8.8 UNITS 1, 2 AND 3 REVISION 2 PAGE I OF 1 ATTACHMENT 8.6 UNIT 2 AND 3 BORON DILUTION CALCULATION A

B C

Boron Level Concentration Change Product LOCATION (Dm)

(qallons)

(A X B)

Boric Acid Makeup Tank T071 ppm X

qal

=

NOTE:

This calculation assumes Boric Acid complete mixing of reactor Makeup Tank T072 ppm X

qal

=

coolant and containment sump after LPSI/HPSI recirculation resulting In a Safety Injection conservative value for the Tank TO07 Oem X

qal

=

calculated final boron concentration.

Safety Injection Tank T008 ppm X

qal

• Pre-Accident Boron concentration levels can Safety Injection be obtained from the Tank T009 ppm X

4al

=

Reactor Laboratory Analysis Book.

Safety Injection Tank TOIO ppm X

qal

=

• • Level changes can be obtained from the Refueling Water plant monitoring Storage Tank T005 ppm X

gal

=

computer.

Refueling Water Storaqe Tank T006 ppm X

al

=

Initial Containment Sump Level ppm X

al Pre-accident RCS ppm X

83.039 9al TOTAL Final Boron Concentration = Total Column (C) =

=

ppm Total Column (8) 0279c

ENCLOSURE 4 ALTERNATIVE POST ACCIDENT SAMPLING CAPABILITY PASS FUNCTION ALTERNATIVE CAPABILITY I. Containment Atmosphere A.

Hydrogen An 'alternate method, independent of PASS, for determining containment atmosphere hydrogen levels is by means of hydrogen monitors inside containment. The containment hydrogen monitors are Seismic Category A, Quality Class II and environmentally qualified Class 1E instruments.

These two instruments provide redundant, channelized continuous readout in the control room of hydrogen levels inside containment from 0 to 10 Volume percent.

The hydrogen monitors are located inside containment at elevation 64' on the outside of two steam generator shield enclosures.

This location provides the hydrogen.sensor

. with access to the upper regions of the containment building and allows optimum

.monitoring of the containment atmosphere.

The limiting conditions for operation of these monitors is delineated in Technical Specification Section 3.6.3, submitted on July 9, 1984 as Proposed Change No. 125.

Buildup of hydrogen inside the containment following an accident is caused primarily by the reaction between the stainless steel fuel cladding and the reactor coolant. The hydrogen monitor system tracks and records this hydrogen buildup described in. The operator uses the system to energize the hydrogen recombiners at 1/2 volume percent hydrogen in containment approximately 1 day into an accident that involves a release to the containment.

However, the hydrogen recombiners will not begin to remove hydrogen until concentrations reach 3-1/2 volume percent which occurs approximately 12 days into the accident.

The system alarms in the control room when air volume percent hydrogen is reached inside the containment after approximately 16 days.

The

-2 PASS FUNCTION ALTERNATIVE CAPABILITY rate of hydrogen buildup for a PWR causes the hydrogen lower combustible limit (4 volume percent) to be reached in approximately 16 days.

This is significantly less rapid than the hydrogen buildup from a BWR whose drywell can experience up to a 6 volume percent level within one and one half. minutes following the accident.

B. Radionuclide The normal method of obtaining a containment grab sample could be used for most postulated accidents because radiation.levels would not preclude these samples from being taken or processed in the normal sample laboratory.

The PASS provides two methods of determining containment atmosphere radionuclide concentrations. An intrinsic germanium multi-channel analyzer is provided to measure post accident containment airborne samples.

In addition, the PASS is provided with the capability to take a diluted containment.

airborne sample. This diluted sample could be measured using onsite laboratory equipment.

An alternate method, independent of PASS, for determining containment atmosphere radiation levels is by means of high range in-containment detectors. These detectors are Seismic Category A, Quality Class II and environmentally qualified Class 1E instruments.

These two instruments provide redundant, channelized continuous readout, immediately adjacent to the main control room panels, of radiation levels inside containment from 100 to 108 R/hr.

As. shown on Attachment 2, the high range in-containment monitors are located at

..elevation 46'-6" on the outer steam generator enclosure wall.s for steam generator E-lA and!

E-1C. These locations permit the detector to "see" the containment atmosphere without being obstructed by concrete shield walls.

The detector is designed to minimize attenuation of low energy gamma. The limiting condition for operation of these monitors is delineated by Technical Specification Table.3.5.10-1 in Proposed Change No. 125 submitted on July 9, 1984.

-3 PASS FUNCTION ALTERNATIVE CAPABILITY Correlations between the radiation readings for the high range incontainment monitor and accident releases have been developed. The basis of the correlation is provided in.2 of Enclosure 3.. This provides the accident definitions and associated releases into the containment free volume.. provides a graph showing t.he projected relationship between monitor dose rate, time after shutdown and accident release. These curves in conjunction with the radiation levels indicated on the high range in-containment detector provide an indication of degraded core conditions.

C. Diluted samples Normal sampling is available as discussed for Radionuclides in I.B above for most of the spectrum of accidents. However, for accidents with TMI type source terms there is no backup capability. In that this capability is a backup, no further backup capabilities are required.

II. RCS.Analysis A. Gas

1. Hydrogen The capability to measure the hydrogen concentration of the containment atmosphere is described in discussion I.A above. The hydrogen concentration in the containment atmosphere is considered representative of the reactor coolant hydrogen content.

Reactor coolant hydrogen is produced as a result of degraded core stainless steel-water oxidation reactions. This hydrogen would not be produced without a significant displacement of water in the reactor vessel to a level below the core. Such a displacement of water volume could occur only if an open path existed to the containment thereby releasing the hydrogen to the atmosphere. This process is enhanced by the presence of the Reactor Coolant Gas Vent System to release to the atmosphere any gas which remains in the vessel.

-4 PASS FUNCTION ALTERNATIVE CAPABILITY Station Chemistry Procedure S0123-III-8.8 describes a second alternate procedurelusing analysis of grab samples obtained from the PASS or Radiochemistry Laboratory. This procedure identifies the specific laboratory instrument to be used.

2. Oxygen Oxygen gas measurements are not a requirement of NUREG-0737.

3. Radionuclide Use the high-range incontainment monitbrs.

See the discussion of I.B above.

4. Total Dissolved The subcooled margin monitor and the core Gas exit thermocouple system will provide temperature and pressure parameters necessary

-to determine the amount of dissolved gas which reactor coolant can retain.

Station Chemistry Procedure S0123-11I-8.8 identifies specific instrumentation which is available under post-accident conditions to monitor reactor coolant temperature and pressure. With known values of temperature and pressure the maximum dissolved gas' content in the coolant may be determined using the principal of Henry's Law. The Reactor Coolant Gas Vent System is available to remove such gas from the vessel.

Station Chemistry Procedure S0123-III-8.8 describes a second alternate method for total gas determination which employs grab samples from the Radiochemistry Laboratory. This alternate method Is available for most of the spectrum accidents.

The procedure iddntifies the instrument to be used in the analysis.

5. Diluted Grab In that this capability is a backup, no Sample (Backup further backup capabilities are required.

to inline instrumentation)

'-5 PASS FUNCTION ALTERNATIVE CAPABILITY B. Liquid,

1. Boron, pH, Boron concentration is measured as a means to Radionuclide verify reactor shutdown. The immediate means for operator verification of this safety function is to verify the open position on the control rod scram breakers. The backup to that action Is the measure of neutron flux using either the incore or excore neutron detectors. Therefore, measurement of boron concentration is in itself a third level alternate method to verify reactor shutdown.

Boron concentration may be calculated by correcting the initial reactor coolant concentration prior to the accident by the amount of spray and safety injection water, added during the accident. The amount of water injected and the RCS inventory can be determined from refueling water storage tank boric acid storage tank level indications.

Station Chemistry Procedure SO123-III-8.8 describes the method to calculate the final boron concentration and dilution for safety injection.

pH is an indication of the potential for long term corrosion. The potential for corrosion would be assessed through pH analysis of that sample which is obtained through undiluted grab sample methods.

2. Chloride Chlorides are monitored in response to the concern for long term stress. corrosion. As discussed above for pH backup undiluted grab sampling analysis is provided as backup.

Station Chemistry Procedures S0123-III-1.3 and S0123-III-4.7 describe the procedures for the chloride measurement of the diluted sample using water chemistry techniques and as an alternate using ion chromatography.,

-6 PASS FUNCTION ALTERNATIVE CAPABILITY

3. Diluted grab In that this capability is a backup, no sample (backup further backup capabilities are required.

to inline instrumentation)

III. Containment Sump The analysis of containment sump liquids may.

be correlated with the RCS grab sample, corrected by the amount of spray and safety injection water. The amount of water.

injected and the RCS inventory are available from refueling water storage tank level indications.

LAB:2186F

E10 X F10 TO THE CENTIMETER 1o X 25 CM.

461510

• E KEUFFEL & ESSEFR CO. M..E u.h FIGURE 3 -

TOTAL SCE POST-LOCA HYDROGE11 BUILDUP IR II I!ii I,

0 d0 20 30 40 50 60' A A TLOCAi

.1 11 DAYS AFER LOC

ATTACHMENT 2 R 7 4efAC70q COOAir PI-A7FO*M RA11.0

~'law REA/4fN C' 7A-'

RIMOVA.46 VTC RfMOVPAT/A PLANHso Iv. 75.

4 A.: OR EEVA TION (A*

ATTACHfIENT 3 DOSE RATE VS TIME FOR HIGH RANGE IN-CONTAINMENT MONITORS (1255/1257) lo

1.

-Ao r

E.

10% Failed Fuel 104GP 1% Failed Fuel

)

4 1 0I14 16 1S M0 22 24, TMEW hows

ENCLOSURE 5 DEFINITION OF OPERABILITY OF THE POST-ACCIDENT SAMPLING SYSTEM

1. Routine surveillances described in Surveillance Procedure (S0123-III-8.1) are conducted at the prescribed intervals when plant conditions permit.

2. In the event of a PASS component malfunction, the specific alternate method of sampling listed.in the "Alternate Methods of Post-Accident

.Parameter Sampling" procedure (S0123-III-8.8) is available and measures are being taken to effect repairs to the component that has malfunctioned.

3. Calibration of PASS Instruments is current.

BASIS FOR DEFINITION

1. Surveillances prove that the system will provide reliable, repeatable and accurate PASS parameter information. In addition, routine surveillances provide training opportunities.

2. The alternate methods procedure refers to techniques available to demonstrate the ability to obtain PASS parameter information when PASS components malfunction. The procedure also indicates the need to effect repairs on the PASS component that has malfunctioned.

3. A routine check of the PASS instrument calibration program ensures instruments are calibrated and avilable when PASS is needed under accident conditions.

LAB:2186F