ML17291A104

Amend 122 to License NPF-21,adding Special Test Exception for Inservice Leak Testing & Hydrostatic Testing & Deleting Table B 3/4.4.6-1 Reactor Vessel Toughness from TS Bases

ML17291A104
Person / Time
Site:	Columbia
Issue date:	05/27/1994
From:	Quay T Office of Nuclear Reactor Regulation
To:
Shared Package
ML17291A105	List: ML17291A104 ML17291A106
References
NUDOCS 9406070338
Download: ML17291A104 (24)

Text

~gP,R AE00

~4 C

P UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055&0001 WASHINGTON PUBLIC POWER SUPPLY SYSTEM DOCKET NO. 50-397 NUCLEAR PROJECT NO.

2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 122 License No.

NPF-21 The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by the Washington Public Power Supply System (licensee) dated February 17,

1994, supplemented by letter dated May 13,

1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; 2.

D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the pub1ic; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to -this license amendment and paragraph 2.C.(2) of Facility Operating License No.

NPF-21 is hereby amended to Tead as follows:

9406070338 ',940527 PDR ADOCK"05000397' PDR I'

3.

(2)

Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 122 and the Environmental Protection Plan contained in Appendix 8, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

This amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

May 27, 1994 Theodore R. quay, Director Project Directorate IV-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation

1 I

ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO.

122 TO FACILITY OPERATING LICENSE NO.

NPF-21 DOCKET NO. 50-397 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.

The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

The corresponding overleaf pages are also provided to maintain document completeness.

REMOVE xx(a)

XX1 V 1-10 3/4 4-18 B 3/4 4-4 B 3/4 4-5 B 3/4 4-6 B 3/4 10-1 INSERT xx(a)

XX1 V 1-10 3/4 4-18 3/4 4-21b 3/4 10-7 B 3/4 4-4 B 3/4 4-5 B 3/4 10-1

/1 ll t

I, l

LIST OF FIGURES INDEX FIGURE 3.2.4-5 3.2.6-1 3.2.7-1 3.2.8-1 3.4.1.1-1 3.4.6.1 3.4.6.1.C 4.7-1 3.9.7-1 B 3/4 3-1 B 3/4.4.6-1 OPERATING REGION LIMITS OF SPEC.3.4.1.1

. 3/4 4-3a MINIMUM REACTOR VESSEL METAL TEMPERATURE VERSUS REACTOR VESSEL PRESSURE

. 3/4 4-20 PRESSURE/TEMPERATURE LIMITS FOR 8 EFPY TESTING AND NONNUCLEAR HEATING CURVES 3/4 4-21b SAMPLE PLAN 2)

FOR SNUBBER FUNCTIONAL TEST

. 3/4 7-15 HEIGHT ABOVE SFP WATER LEVEL VS.

MAXIMUM LOAD CARRIED OVER SFP TO BE

. 3/4 9-10 REACTOR VESSEL WATER LEVEL B 3/4 3-8 FAST NEUTRON FLUENCE (E>IMeV) AT 1/4 T AS FUNCTION OF SERVICE LIFE B 3/4 4-7 PAGE LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE GE11 LEAD FUEL ASSEMBLIES.....................

Deleted OPERATING REGION LIMITS OF SPEC. 3.2.6.......

3/4 2-6 OPERATING REGION LIMITS OF SPEC.3.2.7.......

3/4 2-8 OPERATING REGION LIMITS OF SPEC.3.2.8.......

3/4 2-10

5. 1-1

5. 1-2 5.1-3 EXCLUSION AREA BOUNDARY LOW POPULATION ZONE UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS

~

~

~

~

5 2

~

~

~

~

5 3

~

~

~

~

5 4 WASHINGTON NUCLEAR UNIT 2 xx(a)

Amendment No. ~~ 122

~

~

INBEX LIST OF TABLES Continued TABLE 3.3.7. 5-1 4.3.7. 5-1 3.3.7. 12-1 4.3.7. 12-1 3.3. 9-1 3.3. 9-2 PAGE ACCIDENT MONITORING INSTRUMENTATION.............."..

3/4 3-71

'CCIDENT MONITORING INSTRUMEHTATIOH SURVEILLANCE RE)UIREMEHTS............'......"" ""- -

3/4 3-74 EXPLOSIVE GAS MONITORING INSTRUMENTATIOH.............

3/4 3-80 EXPLOSIVE GAS MONITORING INSTRUMEHTATIOH SURVEILLANCE RE(UIREMENTS............................

3/4 3-81 FEEDWATER SYSTEM/MAIH TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION...~." ~"~.. ~ ~ ~"~"".".

3/4 3-85 I

FEEDWATER SYSTEM/MAIH TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SETPOINTS."".............

3/4 3-86

¹.3.9. 1-1 FEEDWATER SYSTEM/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE RE)UIREMENTS.........................................

3/4 3-87 3.4.3.2-1 3.4.3. 2-2 3.4. 4-1 4.¹. 5-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES.....

3/4 4-11 REACTOR COOLANT SYSTEM INTERFACE VALVES LEAKAGE PRESSURE MONITORS........... """.."""""""".

3/4 4-11 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS..............

3/4 4-14 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM.....................................

3/4 4-17 WASHINGTON HUCLEAR UNIT 2 Amendment Ho. H, 98

INDEX LIST OF TABLES Continued TABLE 4.4.6.1.3-1 3.6.3-1 3.6.5.2-1 3.7.8-1 4.8.1.1.2-1 4.8.2.1-1 3.8.4.2-1 3.8.4.3-1 B3/4.4.6-1 5.7.1-1 6.2.2-1 DELETED PAGE

. 3/4 4-22 PRIMARY CONTAINMENT ISOLATION VALVES........ 3/4 6-21 SECONDARY CONTAINMENT VENTILATION SYSTEM AUTOMATIC ISOLATION VALVES AREA TEMPERATURE MONITORING DIESEL GENERATOR TEST SCHEDULE BATTERY. SURVEILLANCE RE(UIREMENTS

. 3/4 6-39

. 3/4 7-31 3/4 8-9

. 3/4 8-14 DELETED COMPONENT CYCLIC OR TRANSIENT LIMITS MINIMUM SHIFT CREW COMPOSITION.

SINGLE UNIT FACILITY B 3/4 4-6

~

~

~

~

~

~

5 7

~

~

~

~

~

~

6 6 PRIMARY CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES...........

3/4 8-23 MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION.....................

3/4 8-26 l

I WASHINGTON NUCLEAR UNIT 2 XX1V Amendment No. ~,-98, ~

122 hi

TABLE 1.1 SURVEILLANCE FRE UENCY NOTATION NOTATION SA S/U N,A.

FRE(RUENCY At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

At least once per 7 days.

At least once per 31 days.

At least once per 92 days.

At least once per 184 days.

At least once per 366 days.

At least once per 18 months (550 days).

Prior to each reactor startup.

Prior to each radioactive release.

Not applicable.

WASHINGTON NUCLEAR - UNIT 2 1-9

CONDITION 1.

POWER OPERATION 2.

STARTUP 3.

HOT SHUTDOWN 4.

COLD SHUTDOWN 5.

REFUELING*

TABLE 1.2 OPERATIONAL CONDITIONS MODE SWITCH POSITION Run Startup/Hot Standby Shutdown¹ ***

Shutdown¹ ¹¹ ***

Shutdown or Refuel** ¹ AVERAGE REACTOR COOLANT TEMPERATURE Any temperature Any temperature 200 F****

2PP F****

c 140 F

OThe reactor mode switch may be placed in the Run or Startup/Hot Standby position to test the switch interlock functions provided that the control rods are verified to remain fully inserted by a second licensed operator or other technically qualified member of the unit technical staff.

¹¹The reactor mode switch may be placed in the Refuel position while a single control rod drive is being removed from the reactor pressure vessel per Specification 3.9.10.1.

• Fuel in the reactor vessel with the vessel head c'closure bolts less than fully tensioned or with the head removed.

• • See Special Test Exceptions 3.10.1 and 3.10.3.

• • • The reactor mode switch may be placed in the Refuel position while a single control rod is being recoupled provided that the one-rod-out interlock is OPERABLE.

• See Special Test Exception 3.10.7.

WASHINGTDN NUCLEAR UNIT 2 1-10 Amendment No. 122

1 TABLE 4.4.5-1 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PRONAH CI C

Ifll I

4l IW TYPE OF HEASUREHENT AND ANALYSIS 1.

Gross Beta and Gaaoa Activity Determination 2.

Isotopic Analysis for DOSE EQUIVALENT l-131 Concentration 3.

Radiocheaical for K Determination 4.

Isotopic Analysis for Iodine 5.

Isotopic Analysis of an Off-gas Sample Including guantitative Heasurements for at least Xe-133, Xe-135 and Kr-88 SAHPLE AND ANALYSIS FEEIF Ell Y At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> At least once per 31 days At least once per 6 aonths*

a)

At least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, whenever the specific activity exceeds a limit, as required by ACTION b..

b)

At least one sample, between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following the change in THERMAL PlNER or off-gas level, as required by ACTION c.

At least once per 31 days OPERATIONAL CONDITIONS IN NICH SAMPLE ANO ANALYSIS RE UIREO 1, 2, 3

1f, 20, 3f, 48 1, 2

• Sample to be taken after a minimum of 2 EFPO and 20 days of POMER OPERATION have elapsed since reacto~

was last subcritical for IS hours or longer.

NUntil the specific activity of the primary coolant system is restored to within its limits.

REACTOR COOLANT SYSTEH 3/4.4.6 PRESSURE/TEHPERATURE LIHITS REACTOR COOLANT SYSTEH LIHITING CONDITION FOR OPERATION 3.4.6.1 The reactor coolant system temperature and pressure shall be limited in accordance with the limit lines shown on Figure 3.4.6. 1 or 3.4.6. l.c*

(1) curve.A or A'or hydrostatic or leak testing; (2) curve 8 or B'or heatup by non-nuclear

means, cooldown following a nuclear shutdown and low power PHYSKCS TESTS; and (3) curve C for operations with a critical core other than low power PHYSICS TESTS, with:

a.

A maximum heatup of 100'F in any 1-hour period, h.

A maximum cooldown of 100'F in any 1-hour period, c.

A maximum temperature change of less than or equal to 20'F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves, and d.

The reactor vessel flange and head flange temperature greater than or equal to 80'F when reactor vessel head bolting studs are under tension.

APPLICABILITY: At al l times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the reactor coolant system; determine that the reactor coolant system remains acceptable for continued operations or be in at least HOT SHUTDOWN within 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 />.

SURVEIlLANCE RE UIREHENTS 4.4.6.1.1 During system heatup,

cooldown, and inservice leak and hydrostatic testing operations, the reactor coolant system temperature and pressure shall be determined to be within the above required heatup and cooldown limits and to the right of the limit lines of Figure 3.4.6. 1 or 3.4.6.1.c curves A, A', B, B', or C, as applicable, at least once per 30 minutes.

• Figue e 3.4.6.1.c A'nd B'urves are effective for less than or equal to 8 EFPY of eperation.

WASH1%GTON NUCLEAR UNIT 2 3/4 4-18 Amendment No.~ 122

WNP-2 PRESSURE/TEMPERATURE LIMITS FOR 8 EFPY TESTING AND NONNUCLEAR HEATING CURVES A'R P 1400 1300 1200 1100 1000 u

II00

~

800 700 600 500 400 300 200 110 F

312 PSIG A40 F

312 PSIG Core beltline limits.

Limits after an assumed 51.1 F core beltline temp shift from an ini)ial RTN0T of 28 F.

100 Boltup limit 80 F

0 50 100 150 200 250 300 350 400 450 500 NININJN REACTOR NETAL TEMPERATURE TEMPERATURE F MASHINGTON NUCLEAR <<UNIT 2 FIGURE 3.4.6.1.c 3/4 4-21b Amendment No. f22

V t

"I

S ECIAL TEST EXCEPTIONS 3/4.10.7 INSERVICE LEAK AND HYDROSTATIC TESTING LIMITING CONDITION FOR OPERATION

3. 10.7 When conducting Reactor Vessel inservice leak or hydrostatic testing, the average reactor coolant temperature specified in Table 1.2 for OPERATIONAL CONDITION 4 may be increased above 200'F, and operation considered not to be in OPERATIONAL CONDITION 3, to allow performance of an in service leak or hydrostatic test provided the maximum reactor coolant temperature does not exceed 212'F and the following OPERATIONAL CONDITION 3 LCO's are met:

a.

LCO 3. 1.3.8, "Control Rod Drive Housing Support";

b.

LCO 3.3.2, "Isolation Actuation Instrumentation,"

Items 2a, 2c, and 2d of Table 3.3.2-1; c.

LCO 3.6.5.1, "Secondary Containment Integrity";

d.

LCO 3.6.5.2, "Secondary Containment Automatic Isolation Valves";

e.

LCO 3.6.5.3, "Standby Gas Treatment";

and f.

LCO 3.8.4.3, "Motor-Operated Valves Thermal Overload Protection."

APPLICABILITY:

OPERATIONAL CONDITION 4 with average reactor coolant temperature

>200'F and <<212'F ACTION:

With the requirements of the above specification not satisfied, immediately enter the applicable condition of the affected specification or immediately suspend activities that could increase the average reactor coolant temperature or pressure and reduce the average reactor coolant temperature to ~ 200'F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE RE UIREMENTS 4.10.7 Verify applicable OPERATIONAL CONDITION 3 surveillances for specifications listed in 3.10.7 are met.

WASHINGTON NUCLEAR UNIT 2 3/4 10-7 Amendment No. i22

lit II t,

fl,

REACTOR COOLANT SYSTEH BASES 3/4.4. 5 SPECIFIC ACTIVITY The limitations on the specific activ!ty of the primary coolant ensure that the 2.hour thyroid and vhole body doses resulting from a aain steam line failure outside the containment during steady-state operation vill not exceed small fractions of the dose guidelines of 10 CFR Part 100.

The values for the limits on'specific activity represent 5nterim limits based upon a parametric evaluation by the NRC of typical site locations.

These values are conserva-tive in that specific site parameters, such as SITE BOUNDARY location and meteorological conditions, sere not considered 5n this evaluation.

The ACTION statement permitting PAAR OPERATION to continue for limited time periods ~5th the primary coolant's spec)fic activity greater than 0.2 s:$ crocur$ e per gram DOSE EQUIVALENT I-131, but less than or equal to 4.0 aicro-curies per gram DOSE E(UIVALENT I-X3}, accomnodates possible iodine spiking phenomenon which may occur following changes in THERHAL POWER.

Closing the aain steam line 5solation valves prevents the release of activity to the environs should a steam line rupture occur outside containment.

The surveillance requirements provide adequate assurance that excessive specif5c activity levels in the reactor coolant arill be 4etected in sufficient time to take corrective action.

WASHINGTON NUCLEAR - UNIT 2 8 3/4 4-8 Amendment No.

39

REACTOR COOLANT SYSTEH BASES 3/4.4.6 PRESSURE/TEHPERATURE LIHITS All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.

These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations.

The various categories of load cycles used for design purposes are provided in Section 4.9 of the FSAR.

During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall.

These thermal induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure.

Therefore, a pressure-temperature curve based on steady-state conditions, i.e.,

no thermal stresses, represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location.

The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location.

The thermal gradients established during heatup produce tensile stresses which are already present.

The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined.

Subsequently, for the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on an individual basis.

The reactor vessel materials have been tested to determine their initial RTgpT Reactor operati on and resul tant fast neutron irradi ati on, E greater than I HeV, will cause an increase in the RT~>>.

Therefore, an adjusted reference temperature, based upon the fluence, nickel content, and copper content of the material in question, can be predicted using Bases Figure B

3/4.4.6-1 and the recommendations of Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Haterials."

The pressure/temperature limit curves, Figure 3.4.6el and 3.4.6. I.c include predicted adjustments for this shift in RToT for the end of life fluence and are effective for 10 EFPY and 8 EFPY, respectively.

The actual shift in RT~pT of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTH E185-73 and 10 CFR Part 50, Appendix H, irradiated reactor vessel mate-rial specimens installed near the inside wall'f the reactor vessel in the core area.

Thc.irradiated specimens can be used with confidence in predicting reactor vessel material transition temperature shift The.operating limit curves of Figure 3.4.6.1 and 3.4.6.l.c shall be adjusted, as required, on the basis of the specimen data and recommendations of Regulatory Guide 1.99, Revision 2.

WASHINGTON NUCLEAR UNIT 2 B 3/4 4-4 Amendment No.~ 122

REACTOR COOLANT SY EH BASES PRESSURE/TEHPERATURE LIHITS (Continued)

The pressure-temperature limit lines shown in Figures 3.4.6. 1 and 3.4.6. l.c for reactor criticality and for 'inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR Part 50 for reactor criticality and for inservice leak and hydrostatic testing.

3/4.4.7 HAIN STEAN LINE ISOLATION VALVES Double isolation valves are provided on each of the main steam lines to minimize the potential leakage paths from the containment in case of a line break.

Only one valve in each line is required to maintain the integrity of the containment,

however, single failure considerations require that two valves be OPERABLE.

The surveillance requirements are based on the operating history of this type valve.

The maximum'closure time has been selected to contain fission products and to ensure the core is not uncovered following line breaks.

The minimum closure time is consistent with the assumptions in the safety analyses to prevent pressure surges.

3/4.4.8 STRUCTURAL INTEGRITY'he inspection programs for ASHE Code Class 1,

2 and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant.

Access to permit inservice inspections of components of the reactor coolant system is in accordance with Section XI of the ASME Boiler and Pressure Vessel Code 1974 Edition and Addenda through Summer 1975.

The inservice inspection program for ASHE Code Class 1,

2 and 3 compo-nents will be performed in accordance with Section XI of the ASHE Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the NRC pursuant to 10 CFR 50.55a(g) (6) (i).

3/4.4.9 RESIDUAL HEAT REHOVAL A single shutdown cooling mode loop provides sufficient heat removal capability for removing core decay heat and mixing to assure accurate tempera-ture indication, however, single failure considerations require that two loops be OPERABLE or that alternate methods capable of decay heat removal be demonstrated and that an alternate method of coolant mixing be in operation.

WASHINGTON NUCLEAR UNIT 2 B 3/4 4-5 Amendment No.~ 522

BASES TABLE B 3/4.4.6-1 REACTOR VESSEL TOUGHNESS C

I I

C Ob I

C71

~ e I'

COHPONENT BELTLINE Ring 1 Plate Ring 2 Plate Girthweld Girthweld NON-8ELTLINE Ring 3 Plate Ring 4 Plate Vessel Flange Top Head Flange Top Head Dollar Plate Top Head Side Plates Bottom Head Dollar Plates Bottom Head Radial Plates Nozzles Flange Bolt Studs "Regulatory Guide 1.

HATERIAL TYPE SA-533, GRB, CL1 SA-533, GRB, CL1 E8018QI RAC01N%

0. 15

0. 6

-10

0. 15

0. 5

-30

0. 03

1. 01 N.A.

0.08 0.8 N.A.

+28

-8

-50 41 33 36 15

>100

>100 SA-533, GRB, CL1 SA-533, GRB, CL1 SA-508, CL2 SA-508, CL2 SA-533, GRB, CL1 SA-533, GRB, CL1 SA-533, GRB, CL1 SA-533, GRB, CLl SA-508, CL2 SA-540, B23 99, Revision 2, calculated hRTNDT HIGHEST NXINN STARTING 50 FT-LB/35 h

HIN. UPPER SHELF

~

~

'F ~

'F CU Ni RT HIL TEMP F RT

• 'T-LB

3/4

~ 10 SPECIAL TEST EXCEPTIONS I

BASES 3/4. 10. 1 PRIMARY CONTAINMENT INTEGRITY The requirement for PRIMARY CONTAINMENT INTEGRITY is not applicable during the period when open vessel tests are being performed during the low power PHYSICS TESTS.

3/4. 10.2 ROD SE UENCE CONTROL SYSTEM In order to perform the tests required in the technical specifications it is necessary to bypass the sequence restraints on control rod movement.

The additional surveillance requirements ensure that the specifications on heat generation rates and shutdown margin requirements are not exceeded during the period when these tests are being performed and that individual rod worths do not exceed the values assumed in the safety analysis.

3/4. 10.3 SHUTDOWN MARGIN DEMONSTRATIONS Performance of shutdown margin demonstrations with the vessel head removed requires additional restrictions in order to ensure that criticality does not occur.

These additional restrictions are specified in this LCO.

3/4. 10.4 RECIRCULATION LOOPS This special test exception permits reactor criticality under no flow conditions and is required to perform certain startup and PHYSICS TESTS while at low THERMAL POWER levels.

3/4. 10.5 OXYGEN CONCENTRATION Relief from the oxygen concentration specifications is necessary in order to provide access to the primary containment during the initial startup and testing phase of operation.

Without this access the startup and test program could be restricted and delayed.

3/4. 10. 6 TRAINING STARTUPS This special test exception permits training startups to be performed with the reactor vessel depressurized at low THERMAL POWER and temperature while control-ling RCS temperature with one RHR subsystem aligned in the shutdown cooling mode in order to minimize contaminated water discharge to the radioactive waste disposal system.

3/4.10.7 INSERVICE LEAK AND HYDROSTATIC TESTING OPERATION This special test exception allows reactor vessel inservice leak and hydrostatic testing to be performed in OPERATIONAL CONDITION 4 with the maximum reactor coolant temperature not exceeding 212'F.

The additionally imposed OPERATIONAL CONDITION 3 requirement for secondary containment operability provides conserva-tism in the response of the unit to an operational event. This allows flexibility since temperatures of the reactor vessel metal will be ~ 180'F during the testing and a higher reactor coolant temperature will be necessary to sustain the vessel metal temperature.

The flexibilityis provided so that there is margin to allow temperature drift due to decay and mechanical heat.

WASHINGTON NUCLEAR UNIT 2 B 3/4 10-1 Amendment No.

f22

'i4 t

k'I Il I

r