ML17289A731

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Proposed Tech Specs 3/4.6.5.3 Re Standby Gas Treatment & 3/4.7.2 Concerning Control Room Emergency Filtration Sys
ML17289A731
Person / Time
Site: Columbia Energy Northwest icon.png
Issue date: 07/13/1992
From:
WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:
Shared Package
ML17289A730 List:
References
NUDOCS 9207200090
Download: ML17289A731 (24)
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Text

The following Technical Specification pages are affected by this amendment request:

REMOVE INSERT X11 X11 X111 X111 X1V X1V XV XV XV111 XV111 XX1V XX1V 3/4 6-41 3/4 6-41 3/4 6-42 3/4 6-42 3/4 7-5 3/4 7-5 3/4 7-6 3/4 7-6 3/4 7-7 3/4 7-7 B 3/4 6-5 B 3/4 6-5 B 3/4 6-6 9207200090,9207i'3 PDR. ADOCK'5000397 P PDR

t' INDEX BASES SECTION PAGE 3/4. 0 APPLICABILITY......................:........... - .. - -. .... B 3/4 0-1.

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN.................................. B 3/4 1"1 3/4.1.2 REACTIVITY ANOMALIES..................'........... B 3/4 1-3.

3/4.1. 3 CONTROL RODS..................................... B 3/4 1-2 3/4.1.4 CONTROL ROD PROGRAM CONTROLS..................... B 3/4 1"3 3/4.1.5 STANDBY LI(UID CONTROL SYSTEM.....,.............. B 3/4 1-4 3/4. 1.6 FEEDWATER TEMPERATURE............................ B 3/4 1"5 3/4.2 POWER DISTRIBUTION LIMITS 3/4 '. 1 AVERAGE PLANAR LINEAR HEAT GENERATION R ATE ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 2-1 3/4.2.2 APRM SETPOINTS................................... B 3/4 2-2 3/4.2.3 MINIMUM CRITICAL POWER RATIO..................... 8 3/4 2-3 3/4.2.4 LINEAR HEAT GENERATION RATE....... B 3/4 2-4 3/4.2.5 (RESERVED FOR FFTR) 3/4.2.6 POWER/FLOW INSTABILITY........................... B 3/4 STABILITY MONITORING - TWO LOOP OPERATION........ B 3/4 2-4'/4.2.7 2 j( (u 3/4.2.8 STABILITY MONITORING - SINGLE LOOP OPERATION. B 3l4 2-$ 6 3/4.3 INSTRUMENTATION 3/4.3. 1 REACTOR PROTECTION SYSTEM INSTRUMENTATION.... B 3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION.............. B 3/4 3 2 3/4. 3. 3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION.................................. B 3/4 3"2 3/4. 3. 4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION.................................. B 3/4 3-3 3/4. 3. 5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION...................... B 3/4 3-4 3/4.3. 6 CONTROL ROD BLOCK INSTRUMENTATION........ ..... ................... B 3/4 3-'4 WASHINGTON NUCLEAR - UNIT 2 X11 Amendment No. 77 hfAR 1 1Sga

INOEX BASES SECTION PAGE INSTRUMENTATION (Continued) 3/4.3. 7 MONITORING INSTRUHENTATION Radiation Monitoring Instr umentation....... B 3/4 3-4 Seismic Monitoring Instrumentation......... B 3/4 3-4 Meteorological Honitoring Instrumentation.. B 3/4 3-5 Remote Shutdown Honitoring Instrumentation. B 3/4 3-5 Accident Monitoring Instrumentation........ B 3/4 3-5 Source Range Monitors...................... B 3/4 3-5 Traversing In-Core Probe System............ B 3/4 3-5 Loose-Part Detection System................ B 3/4 3-6 Explosive Gas Honitoring Instrumentation... B 3/4 3"6 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEH............. B 3/4 3-6 3/4.3. 9 FEEDWATER SYSTEM/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION....................... B 3/4 3"6 3/4.4 REACTOR COOLANT SYSTEM 3/4.4. 1 RECIRCULATION SYSTEM............................ B 3/4 4-1 3/4.4.2 SAFETY/RELIEF VALVES........-................... B 3/4 4-1 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems....................... B 3/4 4-la Operational Leakage.....i....................... B 3/4 4-2 3/4.4. 4 C HEMISTRYo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 4-2 3/4.4. 5 SPECIFIC ACTIVITY........'....................... B 3/4 4-3 3/4.4. 6 PRESSURE/TEHPERATURE LIMITS..............,...... B 3/4 4-4 3/4.4.7 HAIN STEAM LINE ISOLATION VALVES................ B 3/4 4-5 gP+ 8 svE'u~ulOL WT.bGliWP

&.4. 9 QkmQURL HE'S PemovRL.

WASHINGTON NUCLEAR " UNIT 2 xiii Amendment No. Q, 98

INDEX BASES SECTION PAGE 3/4. 5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 and 3/4.5.2 ECCS - OPERATING and SHUTDOWN.......... 8 3/4 5-1 3/4.5.3 SUPPRESSION CHAMBER.............................. 8 3/4 5-2 C

3/4. 6 CONTAINMENT SYSTEMS 3/4. 6. 1 PRIMARY CONTAINMENT Primary Containment Integrity................. 8 3/4 6-1 Primary Containment Leakage................... 8 3/4 6"1 Primary Containment Air Locks................. 8 3/4 6-1 HSIV Leakage Control System...................... 8 3/4 6"1 Primary Containment Structural Integrity......... 8 3/4 6"2 Drywell and Suppression Chamber Internal P ressure....................................... 8 3/4 6-2 Drywell Average Air Temperature............... 8 3/4 6"2 Drywell and Suppression Chamber Purge System.. 8 3/4 6-2 3/4.6  ? DEPRESSURIZATION SYSTEMS......................... 8 3/4 6 3 3/4.6.3 'PRIMARY CONTAINMENT ISOLATION VALVES............. 8 3/4 6-4 3/4. 6. 4 VACUUM RELIEF.................................... 8 3/4 6-5 3/4.S. 5 SECONDARY CONTAINMENT............................ 8 3/4 6"'5 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL........... 6 3/4 6-g S 3/4.7 PLANT SYSTEMS 3/4.7. 1 SERVICE WATER SYSTEMS............................ 8 3/4 7"1 3/4.7.2 CONTROL ROOM EMERGENCV FILTRATION SYSTEM......... 8 3/4 7-1.

3/4.7.3 REACTOR CORE ISOLATION. COOLING SYSTEM............ 8 3/4 7-1 3/4.7.4 S NUBBERS......................................... 8 3/4 7"2 WASHINGTON NUCLEAR - UNIT 2 Xiv Amendment No. ~OO

INDEX

'ASES SECTION PAGE PLANT SYSTEMS (Continued) 3/4.Z S.SEALED SOURCE CONTAMINATION................... B 3/4 Z g+

3/4.7.8 AREA TEMPERATURE MONITORING..................... B 3/4 7"4 3/4.7.9 MAIN TURBINE BYPASS SYSTEM.................. B 3/4 7-4 3/4.8 ELECTRICAL POWER SYSTEMS 3/4. 8. 3 '.

3/4.8.1, 3/4. 8.2, C.

and SOURCES, DISTRIBUTION D. C. SOURCES', and ONSITE POWER SYSTEMS............................ B 3/4 8-1 3/4.8.4 ELECTRICAL E(UIPHENT PROTECTIVE DEVICES...--..-. B 3/4 8-3 3/4. 9 REFUELING OPERATIONS 3/4. 9.1 REACTOR MODE SWITCH............................. B 3/4 9-1.

3/4. 9. 2 INSTRUMENTATION,........... B 3/4 9-1.

3/4.9. 3 CONTROL ROD POSITION....... B 3/4 9-1 3/4.9.4 DECAY TIHE ~..................................... B 3/4 9"1 3/4. 9. 5 COMMUNICATIONS.................................. B 3/4 9-1 3/4.9. 6 REFUELING PLATFORH.... 3 3/4 9$ 1 3/4.9.7 CRANE TRAVEL " SPENT FUEL STORAGE POOL.......... B 3/4 9-2 3/4. 9. 8 and 3/4. 9. 9 WATER LEVEL - REACTOR VESSEL and WATER LEVEL - SPENT FUEL STORAGE POOL....... B 3/4 9-2 3/4.9. 10 CCNTROL ROD REMOVAL............................. B 3/4 9-2 3/4.9. 11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION... B 3/4 9-'2 WASHINGTON NUCLEAR - UNIT 2 XV Amendment No. 67 MAY g g ig89

ft INDEX ADMINISTRATIVE CONTROLS SECTION PAGE

6. 1 RESPOHSIBI LITY 6-1
6. 2 ORGANIZATION.. ~ ~ ~ ~ ~ o 6 1
6. 2. 1 OFFSITE AHD OHSITE ORGANIZATIONS.
6. 2 2

~ UNIT STAFF. 6-1

6. 2. 3 NUCLEAR SAFETY ASSURANCE GROUP... 6-7 F UNCTION ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-7 COMPOSITION 6-7 RESPONSIBILITIES 6" 7 RECORDS 6-7
6. 2.4 'HIFT TECHNICAL ADVISOR.. 6-7
6. 3 UNIT STAFF UALIFICATIOHS 6-7
6. 4 TRAINING..
6. 5 REVIEM AND AUDIT. 6-8 PLANT OPERATIONS COMMITTEE (POC) 6-8 FUNCTION.

COMPOS IT ION. 6-8 ALTERNATES... 6-8 MEETING FREQUENCY ..... 6-8 QUORUM. 6-8 RESPOHSI BI LITIES 6-9 HulHoWTQ RECORDS I

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-10

6. 5.2 CORPORATE NUCLEAR SAFETY REVIEW BOARD (CNSRB)......... 6-10 FUNCTION. 6-10 COMPOSITION. ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-Ã 1/

ALTERNATES 6-N))

CONSULTANTS................... 6-11 WASHINGTON NUCLEAR - UNIT 2 XV111 Amendment Ho.94

INDEX LIST OF TABLES (Continued TABLE PAGE 4.4. 6.1. 3-1 D E LE TED t ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

'o ~ ~ ~

~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 4-22 +

3. 6. 3-1 PRIMARY CONTAINMENT ISOLATION VALVES.............. 3/4 6-21
3. 6. 5. 2-1 SECONDARY CONTAINMENT VENTILATION SYSTEM AUTOMATIC ISOLATION VALVES....................... 3/4 6"39 3.7. 8-1 AREA TEMPERATURE MONITORING ...................... 3/4 7-
4. 8.1.1. 2" 1 DIESEL GENERATOR TEST SCHEDULE ................... 3/4 8-9
4. 8. 2. 1-1 BATTERY SURVEILLANCE RE(UIREMENTS ................ 3/4 8-14 3.8.4. 2-1 PRIMARY CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES ............,...... 3/4 8-23 3.8.4. 3-1 MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION ....................................... 3/4 8-26 B3/4. 4. 6-1 REACTOR VESSEL TOUGHNESS ......................... B 3/4 4-6
5. 7. 1-1 COMPONENT CYCLIC OR TRANSIENT LIMITS .............
6. 2. 2-1 MINIMUM SHIFT CREW COMPOSITION-SINGLE UNIT FACILITY ......

SWNEBE@ ILKaAL ~as'PErVTaa TAT'E<4AL Bj4 7-I't<

WASHINGTON NUCLEAR - UNIT 2 XX1V Amendment No. ll7, 98, 107

ONTROLI EQ COPY CONTAINMENT SYSTEMS STAND BY GAS TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION 3.6.5.3 Two independent standby gas treatment subsystems shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and ".

ACTION:

a. With one standby gas treatment subsystem inoperable, restore the inoperable subsystem to OPERABLE status'within 7 days, or:
1. In OPERATIONAL CONDITION 1, 2, or 3, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
2. In OPERATIONAL CONDITION ", suspend handl ing of irradiated fuel in the secondary containment, CORE ALTERATIONS and operations with a potential for draining the reactor vessel. The provi" sions of Specification 3.0.3 are not applicable.
b. With both standby gas treatment subsystems inoperable in OPERATIONAL CONDITION ", suspend handling of irradiated fuel in the secondary containment, CORE ALTERATIONS or operations with a potential for draining the reactor vessel; The provisions of Specification 3.0.3.

are not applicable.

SURVEILLANCE RE UIREMENTS 4.6.5.3 Each standby gas treatment subsystem shall be demonstrated OPERABLE:

At least once per 31 days by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the IPWIWW dp,g ya. f'Ady, ada.LI "When irradiated fuel is being handled in the secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

WASHINGTON NUCLEAR - UNIT 2 3/4 6-41 Amendment No. 26 I

1

~ CONTROLLED COP CONTAINMENT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued

b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the subsystem by:
1. Verifying that the subsystem satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0 05sa and uses the test procedure guidance in Regulatory Posi-tions C.5; a, C. 5.c, and C.'5.d of Regulatory Guide 1. 52, Revision 2, March 1978, at a system flow rate of 4457 cfm + 10K.
2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, 9i,
3. Verifying a subsystem flow rate of 4457 cfm + 10K during system operation when tested in accordance with ANST N510-1980it sa~eroo <.> I(<)

C. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of .

d. At least once per 18 months by:
1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 8 inches water gauge while operating the filter train at a flow rate of 4457 cfm + 10K.
2. Verifying that the filter train starts and isolation dampers open on each of the following test signals.
a. Manual initiation from the control room, and
b. Simulated automatic initiation signal.
3. Verifying that the filter cooling bypass dampers can be manually opened and the -fan can be manually starters e4e e>>.~hi~) oe
4. - Verifying that the heaters dissipate 20. 7 + 2. 1 kW when tested in accordance with ANSI N510-1980. i(O uelkS WASHINGTON NUCLEAR - UNIT 2 3/4 6-42 Amendment No. 25 I

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INSERT A Page 3/4 6-42, paragraph 4.6.5.3.b.2 ASTH D3803-1989 by showing a methyl iodide penetration of less than 20% when tested at a temperature of 30'C and a relative humidity of 95%"; and (footnote)

  • Carbon removed for disposal need not be sampled or tested.

INSERT B Page 3/4 6-42, paragraph 4.6.5.3.c ASTH D3803-1989 by showing a methyl iodide penetration of less than 20% when tested at a temperature of 30'C and a relative humidity of 95%".

(footnote)

  • Carbon removed for disposal need not be sampled or tested.

ONTROLLED COPY PLANT SYSTEMS 3/4.7.2 CONTROL ROOM EMERGENCY FILTRATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.2 Two independent control room emergency filtration system trains shall be OPERABLE.

APPLICABILITY: Al 1 OPERATIONAL CONDITIONS and ".

ACTION:

a. In OPERATIONAL CONDITION 1, 2, or 3 with one control room emergency filtration train inoperable, restore the inoperable train to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLO SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b, In OPERATIONAL CONDITION 4, 5, or ":

With one control room emergency filtration train inoperable, restore the inoperable train to OPERABLE status within 7 days or initiate and maintain operation of the OPERABLE train in the pressuri zati on mode of operation.

2. With both control room emergency filtration trains inoperable, suspend CORE ALTERATIONS, handling of irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel.
c. The provisions of Specification 3. 0. 3 are not applicable in OPERATIONAL CONDITION *.

SURVEILLANCE RE UIREMENTS

4. 7. 2 Each control room emergency filtration system train shall be demonstrated OPERABLE:

At least once per I2 hours by verifying that the control room air temperature is less than or equal to 85'F.

b. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the train operates for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with the heaters 4RSAQ&

"When irradiated fuel is being handled in the secondary containment.

WASHINGTON NUCL'EAR - UNIT 2 3/4 7-5

~ CONTROLLED COP' PLANT SYSTEMS SURVEILLANCE REOUIREMENTS Continued C. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone, communicating with the train by:

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l. Verifying that the train satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05K M~~ the test procedure guidance in Regulatory Positions C. 5.a, C.5. c, and C.S.d of Regulatory Guide 1. 52, Revision 2, March 1978 when operating at a flow rate of 1000 cfm + 10'.

ac~i

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide
1. 52, Revision 2, March 1978, meets the laboratory testing z<s, < criteria of
3. Verifying a train flow rate of 1000 cfm + 10Ã during train operation when tested in accordance with ANSI N510-1980'akes.~

+pro s4, 8'.s. l (5.),

After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Positon C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of 3 pse~k t b

e. At least once per 18 months by:
1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches water gauge while operating the train at a flow rate of 1000 cfm + 10~.

WASHINGTON NUCLEAR - UNIT 2 3/4 7"6 'mendment No. 36

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INSERT C Page 3/4 7-6, paragraph 4.7.2.c.2 ASTH D3803-1989 by showing a methyl iodide penetration of less than 20% when tested at a temperature of 30'C and a relative humidity of 95%"; or by replac-ing the used carbon with new carbon which meets the laboratory testing criteria of ASTH D3803-1989 by showing a methyl iodide penetration of less than 3% when tested at a temperature of 30'C and a relative humidity of 95%;

and (footnote)

  • Carbon removed for disposal need not be sampled or tested.

INSERT D Page 3/4 7-6, paragraph 4.7.2.d ASTH D3803-1989 by showing a methyl iodide penetration of less than 20% when tested at a temperature of 30'C and a relative humidity of 95%"; or by replac-ing the used carbon with new carbon which meets the laboratory testing criteria of ASTH D3803-1989 by showing a methyl iodide penetration of less than 3% when tested at a temperature of 30'C and a relative humidity of 95%.

(footnote)

  • Carbon removed for disposal need not be sampled or tested.

<a 4A

CGNTROL'LED'OPY PLANT SYSTEMS SURVEILLANCE REQUIREMENTS Continued)

2. Verifying that on each of the below pressurization mode actuation test signals, the train automatically switches to the pressurization mode of operation and the control room is maintained at a positive pressure of 1/8 inch water gauge relative to the outside atmosphere during train operation at a flow rate 1000 pscf@ ~ to'?o ',

a) Drywell pressure-high, b) Reactor vessel water level-low, and c) Reactor Building exhaust plenum-high ragiation.

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3. Veri fying that the heaters di ssl pate 5. 0 + 0. 5 kW when tested in accordance with ANSI N510-1980.

After each complete ol partial replacement of a HEPA filter bank by verifying that the HEPA filter bank satisfies the inplace penetration and bypass leakage testing acceptance criteria of less than 0.05Ã in accordance with ANSI H510-1980 while operating the train at a flow rate of 1000~cfm + 10~.

g. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the inplace penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with AHSI H510-1980 for a hal.oge0ated hydrocarbon refrigerant test gas while operating the train at a flow rate of 1000 cfm + 10K.

WASHINGTON NUCLEAR - UNIT 2 3/4 7-7 Amendment Ho. 36

CONTAINMENT SYSTEMS 0

CONTR01ILEG t-"OPY BASES 3/4. 6. 4 VACUUM RELIEF Vacuum relief breakers are provided to equalize the pressure between the suppression chamber. and drywell and between the reactor building and suppres-sion chamber. This 'system will maintain the'tructural integrity of the primary containment under conditions of large di'fferential pressures.

The vacuum breakers between the suppression chamber and the drywell must.

not be inoperable in the open position since this would allow bypassing of the suppression pool in case of an accident'. There are nine pairs of valves to provide redundancy and capacity so that operation may continue indefinitely with no more than two pairs of vacuum breakers inoperable in the closed position.

3/4.6.5 SECONDARY CONTAINMENT Secondary containment is designed to minimize any ground level release of radioactive material which may result from an accident. The reactor building and associated structures provide secondary containment during normal opera-tion when the drywell is sealed and in service. At other times the drywell may be open and, when required, secondary containment integrity is specified.

Establishing and maintaining a vacuum in the reactor building with the standby gas treatment system once per 18 months, along with the surveillance of the doors, hatches, dampers, and valves, is adequate to ensure that there are no violations of the integrity of the secondary containment.

The OPERABILITY of the standby gas treatment systems ensures that suf-ficient iodine removal capability will be available in the event of a LOCA.

The reduction in containment iodine inventory reduces the resulting SITE BOUNDARY radiation doses associated with containment leakage. The operation of this system and resultant iodine removal capacity are consistent with the OL dd h LOCA Ly 3/4. 6. 6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL The OPERABILITY of the systems required for the detection and control of hydrogen gas ensures that these systems will be available to maintain the hydrogen concentration within the primary containment below its flammable limit during post-LOCA conditions. Either drywell and suppression chamber hydrogen recombiner system is capable of controlling the expected hydrogen generation associated with (1) zirconium-water reactions, (2) radiolytic decomposition of water, and (3) corrosion of metals within containment. The hydrogen control system is consistent with the recommendations of Regulatory Guide 1.7, "Control of Combustible Gas Concentrations in Containment Following a LOCA," September 1976.

WASHINGTON NUCLEAR - UNIT 2 B 3/4 6-5 Amendment No. 100

INSERT E Bases page 8 3/4 6-5 The standby gas treatment units have thermostatically controlled strip heaters that elevate the charcoal inlet plenum temperature to 90'F. This design feature combined with the naturally low humidity of the area assures that the relative humidity of the charcoal remains below 70%. Accordingly, there is no need for a periodic drying out to remove condensation from the charcoal as these two features continuously ensure moisture does not accumulate in the charcoal beds.

Each unit is operated for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> during each 31 day period as an operability check. An hour provides adequate time for the unit to reach design operating conditions.

A system flow rate of 4457 + 10% indicated cfm is used to simplify the testing.

In the range of the SGT operating conditions, using indicated flow provides a more conservative and restrictive test condition because for a given indicated flow the actual flow through the standby gas treatment unit will be higher.

Therefore, for the purposes of validating the allowable limits for HEPA or carbon bed bypass leakage and total standby gas treatment train pressure drop, a higher flow will present a greater challenge to the acceptance limits.

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