ML17285A958

Proposed Tech Spec 3/4.4.2, Safety/Relief Valves.

ML17285A958
Person / Time
Site:	Columbia
Issue date:	01/09/1990
From:	WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:
Shared Package
ML17285A957	List: ML17285A956 ML17285A958
References
NUDOCS 9001190108
Download: ML17285A958 (11)

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CONTROLLED COPY ".'-.-.:-",.EACTOR COOLANT SYSTEM 3/4.4.2 SAFETY/RELIEF VALVES

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LIMITING CONDITION FOR OPERATION a) 3.4.2 ~The safety valve function of at least 12 of the following reactor coolant system safety/relief valves shall be OPERABLE with the specified code safety valve function lift settings:"

2 safety/relief valves ta 1150 psig +1%/-3X 4 safety/relief safety/relief valves I 1175 psig +3%/-3X 4 valves 8 1185 psig +EX/-3X .

4 safety/relief valves 8 1195 psig +1%/-3X

+MS/ot 4 safety/relief valves I APPLICABILITY: OPERATIONAL CONDITIONS 1 1205 psig am'y~Pg

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a. Mith the safety valve function of one or more of the above required safety/relief valves inoperable, 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 next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. With one or more safety/relief valves stuck open, provided that suppression pool average water temperature is less than 90~F, close the stuck open safety/relief valve(s);

valve(s) within 2 minutes or if if unable to close the open suppression pool average water tempera-ture is 110'F or greater, place the reactor mode switch in the Shut-down position.

c. Mith one or more safety/relief valve acoustic monitors inoperable, restore the inoperable monitor(s) 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 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 UIREMENTS 4.4.2 The acoustic monitor for each safety/relief valve shall be demonstrated OPERABLE by performance of a:

a. CHANNEL CHECK at least once per 31 days, and a

b. CHANNEL CALIBRATION at least once per 18 months.""

"The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.

""The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.

WASHINGTON NUCLEAR - UNIT 2 3/4 4-7 Amendment No. 38

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~h The safety valve function of at least M of the following reactor coolant system safety/relief valves shall be OPERABLE with the specified code safety valve function lift settings:"

2 safety/relief valves 8 1150 psig +1 /-3X 4 safety/relief valves 8 1175 psig +3%/-3X 4 safety/relief valves 8 1185 psig +1K/-3X 4 safety/relief valves 8 1195 psig +1 /-3X 4 safety/relief valves 8 1205 psig +3%/-3X APPLICABILITY: OPERATIONAL CONDITIONS 1, 1 wh~ Teem L )

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s REACTOR COOLANT SYSTEM BASES 3/4.4.2 SAFETY/RELIEF VALVES (Continued) the dual purpose safety/relief valves in their ASME Code qualified mode (spring lift) of safety operation.

The overpressure protection system must accommodate the most severe pres-surization transient. There are two major transients that represent the most severe abnormal operational transient resulting in a nuclear system pressure rise. The evaluation of these events with the final plant configuration has shown that the MSIV closure is slightly more severe when credit is taken only for indirect derived scrams; i.e., a flux scram. Utilizing this worse case transient as the design basis event, a minimum of 12 safety/relief valves are required to assure peak reactor pressure remains within the Code limit of 110K

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3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4. 3. 1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary. These detection systems are consistent with the recommendations of Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.

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

INSERT B Testing of safety/relief valves is normally performed at low power. It is desirable to allow an increased number of valves to be out of service during testing. Therefore, an evaluation of the MSIV closure without direct scram was performed at 25% of RATED THERMAL POWER assuming only 4 safety/relief valves were operable. The results of this evaluation demonstrate that any 4 safety/relief valves have sufficient flow capacity to assure that the peak reactor pressure remains well below the code limit of 110% of design pressure.

Demonstration of the safety/relief valve lift settings accordance with the provisions of specification 4.0.5.

will be performed in

ATTACHMENT 2 lt

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LOW POWER/FLOW ASME OVERPRESSURIZATION CALCULATION RESULTS This analysis provides support for a proposed WNP-2 Technical Specification modification to increase the number of safety/relief valves that can be out of service when the plant is at a low power/flow condition. The Technical Specification change is desirable for checkout of safety/relief valves after valve maintenance has been performed.

ANF was requested to perform an ASHE overpressurization transient calculation for WNP-2 for the conditions in Table 1. For this event the maximum system pressure is calculated for a containment isolation which is the rapid closure of all main steam isolation valves (MSIYs). The analysis results show that for WNP-2, at Table 1 conditions, four safety/relief valves in service have sufficient capacity to prevent the reactor vessel pressure from reaching the established transient pressure safety limit of 1375 psig (110% of the design pressure) specified by the ASHE Pressure Vessel Code.

This overpressurization calculation was performed with the ANF advanced plant simulator code COTRANSA, which includes an axial one-dimensional neutronics model. Neutronics data which represent WNP-2 for Cycle 5 with a 136 assembly reload batch size with Table 1 conditions were used in this calculation. The most critical active component (scram on HSIV closure) was assumed to fail during the transient.

The power rise is terminated by the increased heat transfer and voiding of the core. At about 4.8 seconds, the reactor scram is initiated by reaching the high vessel pressure trip setpoint (1071 psig). Pressures reach the KI gfh I recirculation pump trip setpoint (1170 psig) before the pressurization has

~ P P1 been reversed. Loss of core flow leads to enhanced steam production as less subcooled water is available to absorb 'core thermal power. The calculated maximum pressure in the steam lines, which was 1221 psig, occurred near the vessel at about 11.4 seconds. The maximum vessel pressure was 1226 psig, and it occurred in the lower plenum at about 10.7 seconds. These results and other significant parameters are presented in Table 2, and Figures 1 through 3 show key calculated parameters.

ANF results for a number of boiling water reactors have shown that ASHE overpressurization results vary little due to cycle specific neutronic effects, and this variation is well within the margin to the ASHE criteria calculated for this low-power case. The conclusion regarding the adequacy of four safety relief valves is applicable to current and future WNP-2 operating cycles with the ANF 8x8 fuel design.

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TABLE 1 Initial Parameter Condition Core Thermal Power 25% of rated Core Flow 33% of rated Steam/Feedwater Flow Rate 3 04 x 106 lb/hr Core Exposure EOC 5 (all rods out)

Steam Dome Pressure 1020 psia Safety/Relief Valve Valves With Highest Pressure Pressure Set Points Set Points Assumed Operable Number of Safety/Relief Four Valves in Service Jet Pump M-Ratio 4.7 Feedwater Temperature 276 'F TABLE 2 Calculated Parameter Result Peak Neutronic Power 67.3% of Rated 9 4 sec Peak Steamline Pressure 1236 psia (1221 psig) 9 11.4 sec Peak Dome Pressure 1231 psia (1216 psig:) 9 10.7 sec Peak Vessel Pressure 1241 psia (1226 psig) 8 10.7 sec Peak Heat Flux 32.5% of Rated 9 5e9 sec

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CO CC CL lA CI REFERENCEl RUHTD: SBJA DATE: 89/03/27 0.0 2.0 4.0 6.0 8,0 i0.0 i2.0 i4.0 TIME (SECONOSj SCRAH SPEED FIGURE 2 LOW POWER/FLOW ASHE OVERPRESSURIZATIOH RESULTS, HORtNL ~ 'r SBKO 89/03/27.

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2. VESSEL HATER LEVEL (IH) o0 io i2 i4 i6 i8 20 TIME, SEC SPEED LOW POWER/FLOW ASHE OVERPRESSURIZATIOH RESULTS, HORHAL SCRAM FIGURE 3 SEO. SSOK 09/03/28. 09.37.07.