ML17290A965

Proposed Tech Specs Supporting Hydrostatic Testing

ML17290A965
Person / Time
Site:	Columbia
Issue date:	02/17/1994
From:	WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:
Shared Package
ML17290A964	List: ML17290A963 ML17290A965
References
NUDOCS 9402220149
Download: ML17290A965 (23)

Text

LIST OF AFFECTED PAGES

~Pe pectin De cri tion of han e xx(a) INDEX Add Figure 3.4.6.1.c to the List Of Figures xxiv INDEX Delete reference to Table B3/4.4.6-1 1-10 DEFINITIONS Add note referring to Special Test Exception 3/4.10.7 3/4 4-18 P/T Limits Added reference to new Figure 3.4;6.1C in two places 3/4 4-21a P/T Limits Added new Figure 3.4.6.1C to new page 3/4 4-21a 3/4 10-7 Special Test Exceptions Added Special Test Exception 3/4.10.7, Inservice Leak And Hydrostatic Testing Operation B 3/4 4-4 Bases, P/T Limits Deleted reference to Table B 3/4.4.6.1. Added reference to Figure 3.4.6.1C in two places.

B 3/4 4-5 Bases, P/T Limits Added reference to Figure 3.4.6.1C.

B 3/4 4-6 Bases, P/T Limits Deleted Table B 3/4.4.6-1 B 3/4 10-1 Bases, Special Test Added a Bases section, 3/4.10.7, for the new Exceptions Technical Specification 3/4.10.7 9402220149 '740217

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~ I I CONTROLLED COPY INDEX LIST OF FIGURES FIGURE PAGE 3.2.4" 5 LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE GEll LEAD FUEL ASSEHBLIES........................................... De1eted 3.2. 6" 1 OPERATING REGION LIMITS OF SPEC. 3.2.6............... 3/4 2-6 3.2. 7-1 OPERATING REGION LIMITS OF SPEC. 3.2.7............... 3/4 2-8

3. 2. 8-1 OPERATING REGION LIMITS QF SPEC. 3.2.8............... 3/4 2"10 3.4. l. 1-1 OPERATING REGION LIMITS OF SPEC. 3.4.1.1............. 3/4 4-3a 3.4.6.1 MINIMUM REACTOR VESSEL METAL TEMPERATURE VERSUS REACTOR VESSEL PRESSURE....................... 3/4 4-20 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST .......... 3/4 7-15 3.9.7-1 HEIGHT ABOVE SFP WATER LEYEL VS. MAXIMUM LOAD TO BE CARRIED OVER.SFP......... 3/4 9-1O B 3/4 3-1 REACTOR YESSEL WATER LEYEL........................... B 3/4 3-8 B 3/4.4.6-1 FAST NEUTRON FLUENCE (E)lHeY) AT 1/4 T AS A FUNCTION OF SERVICE LIFE............................. B 3/4 4-7

5. 1-1 EXCLUSION AREA BOUNDARY ............ 5-2

5. 1-2 LOW POPULATION ZONE.......

5. 1-3 UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS....,........ 5-4 ed Jg(eke/ ky 662-9-/'~~

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'.4.6.1.C CURVES... 3/4 4-21b TESTING AND NONNUCLEAR HEATING WASHINGTON NUCLEAR - UNIT 2 xx(a} Amendment No. gg, ~O9 e

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INDEX LIST OF TABLES Continued TABLE PAGE 4.4.6.1.3-1 '

D ELETEDo ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ i ~ ~ ~ ~ ~ ~ ~ ~ ~ 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 7-31 4.8.1.1. 2-1 DIESEL GENERATOR TEST SCHEDULE ................... 3/4 8-S

4. 8. 2. 1" 1 BATTERY SURVEILLANCE RE(UIREHENTS ................ 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 pgL ETE8 B3/4.4.6-1 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 4-6 5.7. 1-1 COMPONENT CYCLIC OR TRANSIENT LIMITS .............

6.2. 2-1 MINIMUM SHIFT CRBI COMPOSITION-SINGLE UNIT FACILITY ............................ 6-6 WASHINGTON NUCLEAR - UNIT 2 XXiv Amendment No. 87, 98, )07

pPONTROLLED COPY TABLE 1.2 OPERATIONAL CONDITIONS MODE SMITCH AVERAGE REACTOR CONDITION POSITION COOLANT TEMPERATURE

1. POWER OPERATION Run Any temperature

2. STARTUP Startup/Hot Standby Any temperature

3. HOT SHUTDOMN Shutdowns ~"~ 200oF $$ I+

4. COLD SHUTDOMN Shutdowns à 200OF k

5. REFUELING~ Shutdown or Refuel"" 0 < 140 F

~!he reactor mode switch may be placed in the Run or S artup/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.

1 44The reac or 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 closure bol s less than fully tensioned or with the head removed.

"~See Special Test Exceptions 3. 10. 1 and 3. 10.3.

"~"The reactor mode switch mav be placed in the Re uel position while a single cont",ol rod is being recoupled provided that the one-rod-out interlock is OPERABLE.

Sye',cl Qes}- E)ccepkew 8'. la, 7 MASHINGTON NUCLEAR - UNIT 2 1-10

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a ga CONTROLLED COPY REACTOR'COOLANT SYSTEM 3/4.4. 6 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION or A

3. 4.6. 1 The reactor coolant system tempera re ur shall 'oe li ted in accordance with the limit lines shown on Fi ure 3.4.6. {1) curve A for hydrostatic or leak testing; (2) curve B for ea up y n n-nuclear means, cooldown following a nuclear shutdown an low power PHYSICS TESTS; and (3) curve C for operations with a critical co e other than low power PHYSICS TESTS, with: ov'

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

b. A maximum cooldown of 100OF in any 1-hour period,

c. A maximum temperature change of less'han 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 1 times.

ACTION:

Mith 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 SHUTOOWH within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURYEILLANCE RE UIREMEHTS

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4. 6. 1. 1 testing oper be determined to be within the abov t th ight 3 th ii it ii 3 gi

'up During system heatup, coo down, and inservice leak and hydrostatic ations, the reactor cool nt system temperature and pressure shal.l 3.3..3 and cooldown limits and applicabla, at least once per 30 minu as.

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'WASHINGTON HUCLEAR " UNIT 2 3/4 4"18 Amendment No. 87

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WNP-2 PRESSURE/TEMPERATURE LIMITS FOR 8 EFPY TESTING NO NONNUCLM HEATING CURVES A'0, P 1400 Core beltline limits.

1300 Limits after an 1200 assumed 51.1 F core beltline temp shift from an iniPal 1100 RTNDT of 28 F.

1000 A40 000

~ 800 I

700 600 500 400 110 F 312 PSIG F

300 312 PSIG 200 100 Boltup limit 80 F I

0 50 100 150 200 250 300 350 400 450 500 MINIMUM REACTOR METAL TEMPERATURE TEMPERATURE F m

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PECIAL TEST EXCEPTIONS

/4.10.7 INSERVICE LEAK AND HYDR TATIC T TIN LIMITINGCONDITION 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, and the requirements of LCO 3.4.9.2, "Reactor Coolant System - Cold Shutdown," may be suspended, to allow performance of an inservice 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 LCO 3.8.4.3, "Motor-Operated Valves Thermal Overload Protection."

A~PL ABI I: OPERATIONALCONOITIONA

>200'F and (212'F Ig 4 I P

~AT~IN:

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 REQUIREMENTS 4.10.7 Verify applicable OPERATIONAL CONDITION 3 surveillances for specifications listed in 3.10.7 are met.

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p REACTOR COOLANT SYSTEM BASES 3/4.4. 6 PRESSURE/TEHPERATURE LIMITS 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 control-ling 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 o1 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 RTNp To Reactor operation and resultant fast neutron irradiation, E greater than 1 MeV, will cause an increase in the RTNDT. Therefore, an adjusted reference temperature, based u on the fluence nickel co tent and co er co nt of the material in i ure B 3/4. .6-1 and the recommendations ques son, can be pre >cte using Base Haterials." The pressure/ temperature limit curve 'r of Regulatory Guide 1.99, Revision 2, 'Radsat on mbrit ement f Reactor Vessel

.4.6. includes predicted adjustments for this shift in RTNDT for the end of life uence and is effective for 10 EFPY. gad ~.'f.5. I C The actual shift in RTNpT 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 material specimens installed near the inside wall of the reactor vessel in the core area.

The irradiated specimens can be used with confidence in predicting reactor ves-s material transition temperature shift. The operating limit curves of Fi ur .4.6. shall be adjusted, as required, on the basis of the specimen data and recomm ndations of Regulatory Guide 1.99, Revision 2.

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

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~~~-qg- /80 3a eo BASES ~~ly'f ('ff PRESSURE/TEMPERATURE LIMITS (Continued) ,g z,ec,lC The pressure-temperature limit lines shown in Fi ure .3.4.6.1 for reactor criticality and for inservice leak and hydrostatic tes sng have een 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 MAIN STEAM 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 The inspection programs for ASME 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 components 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 REMOVAL 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.

MASHINGTON NUCLEAR " UNIT 2 B 3/4 4-5 Amendment No. ~>

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BASES TABLE B 3/4.4.6-1 REACTOR VESSEL TOUGHNESS COMPONENT BELTLINE TERIAL E

CU K X'F Ni HIGHEST STARTING RT

=F*

50 FT" LB/35 MIL TEMP F MAXIMUM h

RT p MIN. UPPER SHELF FT-LB Ring 1 Plate SA-533, GRB, C 0.15 0.6 -10 +28 41 >100 Ring 2 Plate SA"533, GRB, CLl .15 0.5 "30 33 >100 Girthwel d E8018NM 0. 1. 01 N.A. "50 36 Girthweld RACOlNOl 0;08 8 N.A. 15 NON-BELTLINE Ring 3 Plate SA"533, GRB, CL1 Ring 4 Plate SA-533, GRB, CL1 Vessel Flange SA-508, CL2 Top Head Flange SA-508, CL2 Top Head Dollar SA"533, GRB, CLl Plate Top Head Side SA"533, GRB, CL1 Plates Bottom Head Dollar SA-533, GRB, CL1 Plates Bottom Head Radial SA-533, GRB, CL1 Plates Nozzles SA-508, CL2 Flange Bolt Studs SA-540, B23

• Regulatory Guide 1 .99, Revision 2, calculated hRTNDT

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~ 3 3/4. 10 SPECIAL TEST EXCEPTIONS BASES 3/4. 10. 1 PRIMARY CONTAINMENT INTEGR1 t Y The requirement for PRIMARY CONTAINMENT INTEGRITY is not applicable dur ing the period when open vessel tests are being performed dur ing the low power PHYSICS TESTS.

3/4.10.2 R00 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 requirments 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 nalysis.

3/4.10. 3 SHUTOQliH MARGIH OEHOHSTRATIOHS

~9 Performance ol shutdown margin d o trations with the vessel head removed requires additional restrictions in r to ensure that cri icality does not occur. These additional restrictisHY re specified in this LCO.

3/4. IQ. S RECIRCULATIQH LOOPS This special test. exce~ permits reactor criticality under ro flow conditions and is required ~wperrorm certain startup and pHYSICS TESTS while at low THERMAL PQHER 1eve+

3/4. 19. S OXYGEN CONCENTPATION Relief from the oxygen concentration specificaxions is necessary in order to provide access to the primary containment during the ini ial startup and testing phase of ooeration. Mithout th-is access the startup and test program could be restricted and delayed.

3/4. 10. 6 TRAINING STARTUPS This special test exception permits -raining startups to be performed Nith the reactor vessel depressurized at low THERl4NL POMER and temperature while controlling RCS temperature with one RHR subsystem alioned in the shmdown cooling mode in order to minimize contaminated water discharge to the rad:oactive waste disposal system.

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'PtASHINGTON NUCLEAR - UNIT 2 B -/4 10-1

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Insert to page B 3/4 10-1:

/4.10 7 INSERVICE LEAK AND HYDROSTATIC TESTING OPERATI N 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 conservatism in the response of the unit to an operational event. This allows flexibilitysince temperatures of the reactor vessel metal will be

) 180'F during the testing and a higher reactor coolant temperature willbe necessary to sustain the vessel metal temperature. The flexibility is provided so that there is margin to allow temperature drift due to decay and mechanical heat.

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