ML17285A831

Proposed Tech Spec 3.4.6.1 Re RCS Pressure/Temp Limits

ML17285A831
Person / Time
Site:	Columbia
Issue date:	10/27/1989
From:	WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:
Shared Package
ML17285A830	List: ML17285A829 ML17285A831
References
NUDOCS 8911090140
Download: ML17285A831 (12)

Text

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CONTROLLED COPY INDEX LIST OF FIGURfS FIGURE PAGE

3. 1. 5" 1 SODIUM PENTABORATE SOLUTION SATURATION TEMPERATURE... 3/4 1-21

3. 1. 5-2 SODIUM PENTABORATE TANK, VOLUME VERSUS CONCENTRATION Rf(UIRfi4lfNTS. 3/4 1-22

3. 2. 1-1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPI HGR) VERSUS AVfRAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPE SCR183. 3/4 2"2

3. 2. 1"2 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE, IHITIAL CORE FUEL TYPE SCR233..... 3/4 2-3

3. 2. 1-3 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE BUNDLE EXPOSURE ANF 8x8 RELOAD FUEL 3/4 2-4

3. 2. 1" 4 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPE SCR183............................. 3/4 2-4A

3. 2. 1-5 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE, INITIAL CORE FUEL-TYPE SCR233. . '....;.............: "'3/4 2-48

3. 2. 1-6 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE BUNDLE EXPOSURE ANF 9x9"IX AND 9x9"9X FUEL.......... 3/4 2-4C 3.2. 3-1 REDUCED FLOW MCPR OPERATING LIMIT...... 3/4 2-8

3. 2. 4-1 LINEAR HEAT GENERATION RATf (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE ANF Sx8 RELOAD FUEL.......... 3/4 2-10

3. 2. 4-2 LINEAR HEAT GEHERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE ANF 9x9-IX FUEL.............. 3/4/2-10A

3. 2. 4-3 LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE ANF 9x9-9X FUEL.............. 3/4 2-108

3. 2. 6-1 OPERATING REGION LIMITS OF SPEC. 3.2.6............... 3/4 2-12 3.2. 7-1 OPERATIHG REGION I.IMITS OF SPEC. 3.2.7............. 3/4 2-14 3.2.8-1 OPERATING REGION LIMITS OF SPEC. 3.2.8.... 3/4 2" 16

3. 4. l. 1-1 3;4.6.1 WASHINGTON NUCLEAR THERMAL POWER E II

- UNIT 2 LIMITS E EL OF SPEC.

XX E~

3.4.1.1-1..............

MINIMUM REACTOR VESSEL METAL TEMPERATURE 3/4 4-3a 3/4 4-20 Amendment No. 71 S9li090i40 S9i027 05000397 PDR ADOCK PDC

CONTROLLED COPY INDEX

'/ST 'OF FIGURES FIGURE PAGE

4. 7-1 SAMPLE PLAN 2)

~PBR~N~ANE&3.........

FOR SNUBBER FUNCTIONAL TEST ......,...

~~

3/4 7-15

3. 9. 7-1 HEIGHT ABOVE SFP WATER LEVEL VS. MAXIMUM LOAD TO BE CARRIED OVER SFP 3/4 9-10 B 3/4 3-1 REACTOR VESSEL WATER LEVEL ..................... B 3/4 3-8 B 3/4.4.6-1 FAST NEUTRON FLUENCE (E>1Mev) 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.. . ........ . ........... 5-3

5. 1-3 UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIqUID EFFLUENTS....,.:...... 5-4 WASHINGTON NUCLEAR - UNIT 2 xx(a) Amendment No. 71

REACTOR COOLANT SYSTEM 3/4.4.6 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION

3. 4. 6.1 The reactor coolant system temperature and pressure shall be limited (1) curve/ A a~ for hydrostatic or leak testing; (2) curvej B ~d-B-'or heatup by non-nuclear means, cooldown following a nuclear shutdown and low power PHYSICS TESTS; and (3) .curve/ C and-C-'or operations with a critical core other than low power PHYSICS TESTS, with:

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

b. 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 I-hour period durin nservice hydrostatic and leak testing operations above the up and cooldown limit curves, and

d. The reactor vessel flanpand head flange temperature greater than or equal to 80~F when rea vessel head bolting studs are under tension.

APPLICABILITY: At all times.

ACTION:

With any of the above limits exceeded, restoP temperature and/or pressure to within the limits within 30 minutes; perfo 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

.hours 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 REQUIREMENTS 4.4 '.1. 1 During system heatup, cooldown, and inservice leak and hydrostatic testing operations, the reactor coolant sys em temperature and pressure shall be determined to be within the above required heatup .and cooldown limits and

~d-A-', B wad-8-', or C and~, as applicable, at least once per 30 minutes.

WASHINGTON NUCLEAR " UNIT 2 3/4 4-18

REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS Continued) 4.4. 6.1.2 The reactor coolant system temperature and pressure shall be deter-mined to be to tha right oi'he criticality limit line of Figure/ 3.4.6.1-1 control rods to bring the reactor to criticality and at least once per 30 minutes during system heatup.

4.4.6. 1.3 The reactor vessel material surveillance specimens shall be removed and examined, to determine changes in reactor pressure vessel material properties as required by 10 CFR Part 50, Appendix H in accordance with the schedule in Table

o 4.4.6. 1.3-1. The results of these examinations shall be used to update the 4.4.6.1.4 The reactor vessel flange and head flange temperature shall be verified to be greater than or equal to 80~F:

a. In OPERATIONAL CONOITION 4 when reactor coolant system temperature is:

1. < 1000F, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

2. < 90 F, at least once per 30 minutes.

b. Mithin 30 minutes prior to and at least once per 30 minutes during tensioning of the reactor vessel head bolting studs.

MASHINGTON NUCLEAR - UNIT 2 3/4 4"19

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WNP-Pressure/Temperature limits I~ ~

A B C 1400 CORE BELTLINE 1300 LIMITS 1200 Ar Br C ABC LIMITS AFTER AN ASSUMED 75'F CORE BELTLINE TEMP SHIFT 1100 FROM AN INITIAL RTNDT OF 28'F

, 1000

- 9QQ 800

~ 700 600 500 4QP uo F 312 PSIG 180'F 300 312 PSIG 140'F 200 312 PSIG 100 BOLTUP FEEDWATER NOZZLE NCN ONLY LIMIT FOR CURVE LIMIT 80 F 0 50 100 150 200 250 300 350 400 450 500 MINIMUM REACTOR VESSEL METAL TEMPERATURE Temperature F

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CONTROLLED COPY.

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REACTOR COOLANT SYSTEM BASES 3/4.4.6 PRESSURE/TEMPERATURE 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 4e9 of the FSAR. During startup and shutdown, the r ates of temperature and pressui e 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 str esses 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 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 RTNDT The results of these tests are shown in Table B 3/4 '. 6-1. Reactor operation and resultant fast neutron irradiation, E greater than 1 MeV, will cause an increase in the RTNDT. Therefore, an adjusted reference temperature, yudcksl based upon the fluence, yhosphceus content, and copper content of the material. in question, can be predicted using Bases Figure B 3/4.4.6-1 and the recommendations 4amage-4o Reactor Vessel Materials." The pressure/ temperature limit 'adiatio gl, Ifl for this shift in life fluence/ Curves~

adjustments

~neo-fRg dsower-year s-(BAPY~niy-and-our RTN for ves-A ', 9 the end of re-to-Imh-ef feet-h e-f~4e f

', and-Gs

~t 3-ef feet+ve-fM-1 afMigur~~~r~ e4 .'s 4e effective for 10 EFPY.

The actual shift in RTNDT of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTM 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-sel material transition temperature shift. The operating limit curves of of the specimen data and recommendations of Regulatory Guide 1.99, Revision+ g, WASHINGTON NUCLEAR - UNIT 2 B 3/4 4-4

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REACTOR COOLANT SYSTEM BASES PRESSURE/TEMPERATURE LIMITS (Continued)

The pressure-temperature limit lines shown in Figure 3.4.6.1-l-and-c 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 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

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fission products and to ends e the core is not uncovered following line breaks, The minimum closure time +~sistent with the assumptions in the safety analyses to prevent pressdrtvgrgas.

3/4.4.8 STRUCTURAL INTEGRITY The inspection programs for 1lhtg Code Class 3, 2 and 3 components ensure, that the structural integrity of tMs+components will be maintained at an acceptable level throughout the life~ ha p lant'ccess to permit inservice inspec of components of the reactor coolant system is in accordance with,Sec 'XI of the ASME Boiler and Pressure Vessel Code 1974 Edition and Adde hrough Summer 1975.

The inservice inspection program for ASME Code Class 1, 2 and 3 components will be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a(g) except

~here 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.

WASHINGTON NUCLEAR - UNIT 2 8 3/4 4-5

BASES TABLE B 3/4.4.6"1 REACTOR VESSEL TOUGHNESS NIGflEST HAXIHUH STARTING 50 FT-LB/35 h HIN. UPPER SHELF HATERIAL CU + RT HIL TEHP F RT FT-LB COHPONENT TYPE x x BELTLINE Ring 1 Plate SA-533, GRB, CL1 ~44 O4

-10 +28 ~ +I >100

0. 15 O.S

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-30 Ring 2 Plate Girth@aid Girthweld SA-533, GRB, CL1 E8018NH RAC01NHH 0.15 0.

448;

'.a Ak 020- N.A.

LA.

-50

-44

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59 9&

>100 NON-BELTLINE Og~

Ring 3 Plate SA-533, GRB, CLl Pg Ring 4 Plate Vessel Flange SA-533, GRB, CL1 SA-508, CL2 by~

Top llead Flange SA-508, CL2 Og Top Head Dollar SA-5330 GRB, CLl Plate Top Head Side SA"533, GRB, CL1 Plates Bottom I!cad Dollar SA 533 0 GRB ~ CLl Plates Bottom llead Radial SA-533, GRB, CL1 Plates Nozzles SA-500, CL2 Flange Bolt Studs SA-540, 823 c~ l 4 ~ l~) e K al c Y

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