ML17290B173
| ML17290B173 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 05/06/1994 |
| From: | WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| To: | |
| Shared Package | |
| ML17290B172 | List: |
| References | |
| NUDOCS 9405160094 | |
| Download: ML17290B173 (14) | |
Text
TABLE 4.3.7.5"1 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS I017'I
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9.
Reactor Vessel Pressure Reactor Vessel Water Level Suppression Chamber Mater Level Suppression Chamber Mater Temperature Suppression Chamber Air Temperature Primary Containment Pressure Drywell Air Temperature Drywell Oxygen Concentration Drywell Hydrogen Concentration CHANNEL CHECK M
M M
M M
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M M
M M
CHANNEL.-
CALIBRATION R
R R
R R
R R
Q APPLICABLE OPERATIONAL CONDITIONS 1, 2 1, 2 1, 2 1, 2 1, 2 1, 2
~ I, 2 1, 2 1, 2 O
10.
11.
12.
13.
14.
15.
16.
17.
SafetylRelief Valve Position Indicators" Suppression Chamber Pressure Condensate Storage Tank Level Main Steam Line Isolation Valve Leakage Control System Pressure Neutron Flux:
APRM IRM SRM RCIC Flow HPCS Flow LPCS Flow M
M M
M M
M 1, 2 1,'2 1, 2 1, 2 1, 2 1,.2 1, 2 1,
2 1, 2 1, 2 "This includes acoustic monitor, valve stem position, and tailpipe temperature instrument channels.
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3/4.4. 2 SAFETY/RELIEF VALVES Attachment 1
Page 3 of 5 LIMITING CONDITION FOR OPERATION ACTION: (Continued}
valve(s) within 2 minutes or if suppression pool average water tempera-ture is 110'F or greater, place the reactor mode switch in the Shut-down position.
c.
kith both the acoustic monitor and valve stem position indicator for one or more safety/relief valve(s) inoperable, restore either the acoustic monitor or valve stem position indicator 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 SHUTDOMN 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 position indicators for each safety/relief valve shall be demon-strated 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 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
+s-adequate to perform the test.
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I WASHINGTON NUCLEAR - UNIT 2 3/4 4-7a Amendment No. 88, N, 105
EMERG NCY CORE COOLING SYSTEMS Attachment 1
Page 4 of 5 e
SURVEILLANCE RE UIREMENTS Continued e.
For the AOS by:
1.
At least once per 31 days by verifying that the accumulator backup compressed gas system pressure in each bottle is
> 2200 psig.
2.
At least once per 31 days, performing a
CHANNEL FUNCTIONAL TEST of the accumulator backup compressed gas system low pressure alarm system.
3.
At least ance per 18 months:
a)
Performing a system functional test which inc1udes simulated automatic actuation of the system throughout its emergency operating
- sequence, but excluding actual valve actuation.
b)
Manually opening each AOS valve when the reactor steam dome pressure is greater than or equal to 100 psig* and observing that either:
1)
The control valve or bypass valve position responds accordingly, or 2)
There is a corresponding change in the measured steam flow.
c)
Performing a
CHANNEL CALIBRATION of the accumulator backup compressed gas system low pressure alarm system and verify-ing an initiation setpoint of > 140 psig on decreasing pressure and an alarm setpoint
> 135 psig on decreasing pressure.
d)
Verifying the nitrogen capacity in at least two accumulator bottles per division within the backup compressed gas system.
.The provisisons of Specification 4.0.4 are riot applicable provided the scrveil lance is per oread within i2 hours arts" reactor ste=-r.; pre sure ae artc( Pous "re.
aceo ate to perform the test.
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REACTOR COOLANT SYSTEM Attachment I Page 5 of 5 BASES g ~(-
3/4.4.2 SAFETY/RELIEF VALVES (Continued) the dual pu'rpose 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 MSIY 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 llOX of design pressure.
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Testing of safety/relief valves is normally performed at lower 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 25K 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 110K of design pressure.
TMI Action Plan Item II.D.3, "Direct Indication of Relief and Safety Valve Position," states that reactor coolant system relief and safety valves shall be provided with a positive indication in the control room derived from a reliable valve-position detection device or a reliable indication of flow in the discharge pipe.
Each WNP"2 SRV has both a valve stem position indication device and an acoustic monitor flow detection device which independently meet the requirements of Item II.D. 3.
Hence failure of one device does not impact compliance to II.D.3 and entry into Limiting Condition for Operation action statement 3.4.2.c is required only for inoperability of both devices associated with a specific SRY.
Demonstration of the safety/relief v'alve lift settings will be performed in accordance with the provisions of Specification 4.0.5.
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.
The primary containment sump flow monitoring system monitors the UNIDENTIFIED LEAKAGE collected in the floor drain sump with a sensitivity such that 1 gpm change within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> can be measured.
Alternatively, other methods for measuring flow to the sump which are capable of detecting a change in UNIDENTIFIED LEAKAGE of 1 gpm within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> with an accuracy of + 2X may be used, for up to 30 days, when the installed system is INOPERABLE.
WASHINGTON NUCLEAR - UNIT 2 B 3/4 4-la Amendment No. 88, XSS, 111
II II I
REQUEST FOR TO TECHNICAL.SPECIFIC IONS 3/4.3.7.5, 3/4 4.2,
-AND 3/4.5.1 FOR MAIN STEAM RELIEF VALVE POSITION INDICATION-SUPPLEMENTAL INFORMATION A.
D RIPTIO OF PR P
ED HAN Page 1 of 5 In accordance with 10CFR50.90, the following changes are being proposed:
1)
Technical Specification 3/4.4.2, "Safety/Relief Valves," and Technical Specification 3/4.5.1, "Emergency Core Cooling Systems," would be revised to identify that the 12-hour time limitation for performing the safety relief valve (SRV) position indicator surveillance testing starts when plant steam pressure and flow are adequate to perform the test.
Notes 4.4.2.b "**"and 4.5.1.e.3.b"*" would be revised by changing the current wording:
"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" to "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 and floware adequate to perform the test."
2)
A note is added to the channel calibration requirements for the Safety/Relief Valve Position Indicators (acoustic monitor, valve stem position, and tailpipe temperature instrument channels) identified in Table 4.3.7.5-1, "Accident Monitoring Instrumentation."
This note qualifies SRV position indicator applicability and surveillance requirements, including the exception provided by Technical Specification 4.0.4, by stating:
"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 and floware adequate to perform the test."
In addition, the Bases for Technical Specification 3/4.4.2 is being revised to clearly identify that plant conditions adequate to perform the testing include having sufficient steam flow as well as adequate reactor pressure.
Attachment 1 contains the proposed changes as reflected on the marked-up copies ofpages from the WNP-2 Technical Specifications and Technical Specification Bases.
d fi
REQUEST FOR AME NT TO TECHNICALSPECIFIC IONS 3/4.3.7.5, 3/4.4.2, AND 3/4.5.1 FOR MAIN STEAM RELIEF VALVE POSITION INDICATION-SU0PLEMENTAL INFORMATION Page 2 of 5 B.
D~SI Technical Specification 3/4.4.2, "Safety/Relief Valves," and 3/4.5.1, "Emergency Core Cooling System," contain tests required to be performed during plant startup for operability verification prior to entry into Operational Conditions 1, 2, and 3.
0 General startup activities and pertinent required startup testing is described below:
1)
Technical Specification 3/4.3.7.5 (Table 3.3.7.5-1) requires either the SRV acoustic monitors or valve stem position indicators (VPI) be operable prior to entry into Operational Condition 1 or 2.
(With approval of this request, this testing will be included with the item 3 testing below.)
2)
Following entry into Operational Condition 2 and achievement of reactor criticality, reactor operation enters the heatup phase and reactor coolant temperature and pressure increase.
This evolution includes placing portions of the balance of plant systems in service.
3)
Technical Specification 4.4.2, requires SRV position indicator testing within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reaching adequate steam pressure to perform the test.
Technical Specification 4.5.1.e.3.b requires manual opening of the Automatic Depressurization System (ADS) valves (SRVs) within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reaching adequate steam pressure to perform the test.
4)
Reactor power is increased.
At 10-12% power, plant conditions are adequate to perform SRV setpoint, ADS SRV function, and SRV position indicator (acoustic monitor and VPI) surveillance testing.
5)
Following successful completion of the above
- tests, reactor power is increased to approximately 18%, the main turbine is loaded, and plant startup continues.
Opening of SRVs can result in a reactor pressure transient which initiallyintroduces negative reactivity, and subsequent SRV closure can result in the introduction of positive reactivity (through steam void collapse).
Two conditions must be met before a stroke test of the SRVs can be safely performed:
(1) adequate reactor pressure must exist to protect the SRV from damage when stroking, and (2) sufficient steam flow must exist such that the turbine bypass valves maintain pressure control (do not fully close) when an SRV is opened.
The latter is required because a slow depressurization of the reactor vessel willoccur ifthe turbine bypass valves fully shut in an attempt to control pressure when an SRV is opened.
This slow depressurization/cooldown, combined with the subsequent SRV closure, can result in a pressure and reactor power spike and reactor scram.
REQUEST FOR i
T TO TECHNICALSPECIFIC IONS 3/4.3.7.5, 3/4.4.2, AND 3/4.5.1 FOR MAIN STEAM RELIEF VALVE POSITION INDICATION-SUPPLEMENTAL INFORMATION Page 3 of 5 SRV testing cannot be performed until the required steam flow and pressure conditions are achieved.
A review of recent plant startup data shows 10%-12% power, when plant conditions are adequate for SRV testing, is achieved approximately 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after reaching 950 psig.
k The SRVs must also be stroked to complete the Technical Specification surveillance requirements for the associated acoustic monitors and VPI.
Specifically, necessary acoustic monitor gain adjustments are performed under adequate steam flow conditions for each SRV.
The proposed change will not modify the surveillance requirements for completion of testing within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The sole intent of the proposed change is to clarify the point when the 12-hour clock begins to provide assurance that adequate steam pressure, including adequate steam flow to safely perform testing, exists prior to commencing such testing.
Prior to entry into Operational Condition 3, the Supply System performs surveillance testing to ensure that the SRVs, the ADS function of the SRVs, and the SRV position indication will function on demand.
Testing is performed prior to startup to ensure plant safety and to limit economic impact since failure of these components to meet the surveillance requirements could result in plant shutdown and cooldown to effect repairs.
The conditions that ensure plant safety for the limited period oftime prior to verification ofTechnical Specification operability include:
Atleast four SRVs are required to be operable for the safety mode function prior to entry into Operational Condition 1, 2, or 3. WNP-2 meets this requirement.
2.
The ADS logic is surveilled from the sensing devices (pressure and level switches), up to and including the relay contacts that actuate the SRV solenoid valves.
This testing is performed prior to plant startup.
3.
After SRV maintenance that involves disconnecting the solenoid, the solenoid valves are post maintenance tested prior to plant startup to verify that the solenoids stroke (an audible click is heard). This testing provides a high level ofconfidence that the solenoids willfunction on demand.
4.
The SRVs are tested during the power reduction for the refueling outage.
The testing is performed to satisfy ASME Code setpoint testing requirements.
This testing is not required, but is normally performed to verify SRV safety mode setpoints to support plant startup, and to identify SRVs that require setpoint adjustment.
5.
There are three methods ofopening the SRVs: 1) mechanically via steam pressure (safety mode); 2) pneumatically via either of two (the "A"or "B" solenoid) redundant solenoids for each of the seven ADS SRVs; or 3) pneumatically via one solenoid (the "C" solenoid) for each of the 18 SRVs.
Methods 2 and 3 are available to the operators in the control room. Control of three of the SRVs is available at the Remote Shutdown Panel.
Control of an additional three SRVs is available at the Alternate Remote Shutdown Panel.
0 0
REQUESTFORAME NTTO TECHNICALSPECIFIC IONS 3/4.3.7.5,3/4.4.2,
'AND 3/4.5.1 FOR MAIN STEAM RELIEF VALVE POSITION INDICATION-S&PLEMENTALINFORMATION Page 4 of 5 6.
Supply System calculations have determined that the maximum number ofSRVs required for successful ADS operation is five. This calculation was based on the decay heat load present after an extended run at 100% power. The decay heat load present prior to ADS functional testing as permitted by the requested changes would be significantly lower than this.
The decay heat is lower based on:
- 1) the installation of new fuel during the refueling outage; 2) the significant reduction of decay heat that occurs during the 30 days or longer ofrefueling outage time; and 3) the proportionately lower decay heat generated by a short period of operation at M10% power versus 100% power.
7.
The High Pressure Core Spray (HPCS) and Reactor Core Isolation Cooling (RCIC) systems are required to be operable prior to entry into Operational Conditions 1, 2, or 3 for HPCS and prior to exceeding 150 psig reactor pressure for RCIC.
Thus, both systems are verified operable shortly before exceeding 950 psig.
This provides a high level of confidence that they are operable and willbe available ifneeded.
Therefore, additional time for operator response and alternate systems are available compared to the assumptions made in the accident analyses.
The proposed time period prior to operability verification is short.
Should the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> time limitexpire prior to the completion of required testing, the equipment that has not been surveilled willbe declared inoperable and the required actions taken.
C.
Bases for No Si nificant Hazards onsideration The Supply System has evaluated the proposed changes against the above standards as required by 10CFR50.91(a) and concluded that the change does not:
(1)
Involve a significant increase in the probability or consequences of any accident previously evaluated.
The potential delay in confirming safety relief valve (SRV), SRV position indication (acoustic monitor, valve stem position indicators, and tailpipe temperature instruments),
and ADS operability during plant startup should not result in any change to the expected satisfactory completion of the required surveillance tests.
Surveillance testing that is conducted during the plant shutdown sequence, and during shutdown, provides reasonable assurance that the SRVs willfunction when required.
Under the proposed change, plant test conditions would not be different than in the past since testing was not begun until adequate pressure for the duration of the test was achieved.
Therefore, this change does not involve a significant increase in the probability or consequences of any accident previously evaluated.
REQUEST FOR AME NT TO TECHNICALSPECIFIC IONS 3/4.3.7.5, 3/4.4.2, AND 3/4.5.1 FOR MAIN STEAM RELIEF VALVE POSITION INDICATION-SU0PLEMENTAL INFORMATION Page 5 of 5 (2)
Create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed changes do not involve a design change nor do they involve changes outside the scope of the existing test requirements.
No new failure modes are introduced as a result of the proposed changes.
The acoustic monitors merely provide indication that an SRV is open.
They do not provide an actuation signal.
Alternate mechanisms of SRV position indication exist, i.e., reactor water level changes, reactor pressure changes, main turbine bypass valve position, SRV tail pipe temperature, suppression pool level, and suppression pool temperature.
The time delay prior to operability verification willnot affect Technical Specification requirements.
Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
(3)
Involve a significant reduction in a margin of safety.
The proposed changes have no impact on the operability or performance requirements of the SRVs, including the ADS function, as they do not change the liftsetpoints or minimum number of valves required to be operable.
The effect of delaying the starting point of the time clock is not expected to affect completion of the required tests.
ADS/SRV position indication availability will not be significantly affected by the proposed change since the additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> per refueling cycle of not verifying SRV, SRV position indication, and ADS SRV operability occurs at low power with the High Pressure Core Spray and Reactor Core Isolation Cooling systems available. Additionally, there is a high probability that the SRVs would perform their intended function if required even though they have not been declared operable.
Therefore, the proposed changes do not involve a significant reduction in a margin of safety.
Based on this review, the Supply System has determined that the three standards of50.92(c) are satisfied.
Accordingly, the Supply System has determined that this amendment request involves no significant hazards consideration.
D.
ENVIR NMENTAL N IDERATI NS The proposed changes do not involve a significant hazards consideration, or significantly increase the amounts or change the types of effluents that may be released off-site, nor do they significantly increase individual or cumulative occupational radiation exposures.
Based on the foregoing, WNP-2 concludes the proposed changes meet the criteria given in 10CFR51.22(c)(9) for a categorical exclusion from the requirement for an Environmental Impact Statement.
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