2CAN079702, Application for Amend to License NPF-6,modifying Actions Associated W/Ts Table 3.3-1 for Reactor Protective Instrumentation & Table 3.3-3 for Esfa Sys Instrumentation
| ML20151J329 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 07/28/1997 |
| From: | Hutchinson C ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20151J335 | List: |
| References | |
| 2CAN079702, 2CAN79702, NUDOCS 9708050135 | |
| Download: ML20151J329 (10) | |
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Enttrgy oper:tions,Inc.
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Td 501-E58 4%9 C. Randy Hutchinson We Pres &nt July 28,1997 2CAN079702 U. S. Nuclear Regulatory Commission Document Control Desk Mail Station PI-137 Washington, DC 20555
Subject:
Arkansas Nuclear One - Unit 2 Docket No. 50-368 License No. NPF-6 Technical Specification Change Request Modifying the Allowance to Place Plant Protection System Parameters in the Trip Condition Gentlemen:
Attached for your review and approval is a proposed amendment to the Arkansas Nuclear One Unit 2 (ANO-2) Technical Specifications. The proposed amendment modifies the actions associated with Technical Specification Table 3.3-1 for the Reactor Protective Instrumentation and Table 3.3-3 for the Engineered Safety Feature Actuation System Instrumentation.
On May 14,1997, it was determined that with one of the four channels of Refueling Water Tank Level - Low in trip, as allowed by the Technical Specifications, the potential for a premature initiation of the Recirculation Actuation System during a Loss Of Coolant Accident could result in conditions outside the plant design basis. This was later reported to the staff by way of Licensee Event Report 97-003 (2CAN069703), dated June 13,1997.
The remaining Plant Protection System input parameters were reviewed as part of the ANO corrective action program to determine if any other parameters need to be restricted when in the tripped condition. This review determined an additional input parameter that would produce unwanted results with one of the input channels in the trip condition. Under these conditions, Steam Generator differential pressure was also found to incorrectly feed the i
faulted Steam Generator under unisolable Main Steam Line Break conditions. The allowance to place a channel of Refueling Water Tank Level - Low and Steam Generator differential pressure in the trip condition for an unlimited period of time has existed from the original ANO-2 Technical Specification issuance. Upon discovery of these deficiencies, administrative controls were implemented to limit the amount of time these parameters may be in the tripped condition.
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U. S. NRC July 28,1997
'2CAN079702 Page 2 The proposed change has been evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92(c) and it has been determined that this change involves no significant hazards considerations. The' bases for these determinations are included in the attached submittal.
Entergy Operations requests that the effective date for this change be within 30 days of issuance. Although this request is neither exigent nor emergency, your prompt review is requested.
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Very truly yo r
4 C I/rde Attachments To the best of my knowledge and belief, the statements contained in this submittal are true.
SUBSCRIBED AND SWORN TO before me, a Notary Public in and for bh County and the State of Arkansas, this M day of Qt
,1997.Y O#
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U. S. NRC -
July 28,1997 2CAN079702 Page 3 cc:
Mr. Ellis W. Merschoff Regional Administrator U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 NRC Senior Resident Inspector Arkansas Nuclear One P.O. Box 310 London, AR72847 Mr. George Kalman NRR Project Manager Regan IV/ANO-1 & 2 U. S. Nuclear Regulatory Co'nmission NRR Mail Stop 13-H-3 One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. David D. Snellings Director, Division ofRadiation Control and Emergency Management Arkansas Department ofHealth 1815 West Markham Street Little Rock, AR 72205 a
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2CAN079702 PROPOSED TECHNICAL SPECIFICATION l
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RESPECTIVE SAFETY ANALYSES I
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LICENSE NO. NPF-6 ENTERGY OPERATIONS. INC.
ARKANSAS NUCLEAR ONE. UNIT TWO I
DOCKET NO. 50-368 l
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Attachment to 2CAN079702
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DESCRIPTION OF PROPOSED CHANGES
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Technical Specification (TS) Table 3.3-1 Actions 2 and 3 were revised by removing the e
words "STARTUP and/or POWER OPERATION" and inserting the words " operation in the applicable MODES."
j TS Table 3.3-3 Actions 10 and 11 were revised by removing the words "STARTUP l
e and/or POWER OPERATION" and inserting the words " operation in the applicable MODES."
Table 3.3-3 Actions 10 and 11 were also modified to restrict the allowance to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> for e
placing a channel of Steam Generator (SG) differential pressure (AP) or the Refueliag Water Storage Tank (RWT) Level - Low channel in the tripped condition.
The Bases for specification 3/43 was modified to include bases information for Actions 10 and 11 for Table 3.3-3.
BACKGROUND 1
The Recirculation Actuation Signal (RAS) is initiated by a 2 out of 4 logic for the RWT low level signal. If this occurs during the injection phase of a Loss of Coolant Accident (LOCA),
i the RAS system will change the mode of operation of the Emergency Core Cooling Systems (ECCS). The mode of operation will be changed from the injection phase to the recirculation phase. At the onset of the recirculation phase, the RAS automatically stops the Low Pressure j
Safety Injection (LPSI) pumps and shifts the High Pressure Safety Injection (HPSI) pumps and the Containment Spray System suction from the RWT to the containment sump. The containment sump then supplies the inventory requirements for the ECCS pump suctions. The RAS is designed to occur when the inventory of the RWT is nearly depleted; therefore, ensuring the containment sump will have adequate inventory for the recirculation phase.
TS Table 3.3-3 Actions 10 and 11 allow the placement of a cht.nnel of RWT Level - Low in the tripped condition. With one channel in the tripped condition the system is susceptible to a i
single failure of a second channel resulting in a RAS actuation. If this condition were to occur post-LOCA, prior to the RWT reaching the low level setpoint, a premature RAS actuation would occur. Without adequate inventory in the containment sump, the HPSI pumps and Containment Spray pumps would have their suctions supplied by an inadequate suction source. This event would be of no consequence ifit were to occur after the RWT has reached the low level setpoint because the ECCS systems would already be aligned for the recirctiation phase. If the fr.ilure occurred while the plant is at power, the event would be undesirable but would not place the plant in a transient.
As reported in Licensee Event Repen 97-003, dated June 13,1997 (2CAN069703), this scenario would allow a single failure of the second channel of RWT Level - Low to remove both trains of ECCS and the Containment Spray System from service. The Waterford 3
Attachment to 2CAN079702 Page 2 of 6 Nuclear Station has submitted a similar TS change request for this same issue by letter dated June 2,1997. This issue was previously identified by the Fort Calhoun Station and ultimately resolved by amendment 173 to its Unit 1 TS.
The remaining Plant Protection System (PPS) input parameters we
.1ewed as part of the ANO corrective action program upon identification of this condition. This review determined an additional PPS input parameter that would produce unwanted results under a single failure scenario with one of the input channels in the tripped condition as allowed by the ANO-2 TS.
Under these conditions, Steam Gererator differential pressure (AP) was found to incorrectly feed the faulted Steam Generator under unisolable Main Steam Line Break conditions.
The following scenario would result in the faulted Steam Generator being supplied emergency feedwater during an unisolable Main Steam Line Break event.
One channel of Steam Generator AP will have to be in the tripped condition. A Main Steam Line Break occurs that i
is unisolable due to the leak being located downstream of the feedwater isolation valves and-upstream of the MSIVs. During this event, one of the remaining Steam Generator AP l
channels fails, resulting in feeding the faulted Steam Generator. If neither Steam Generator is faulted, the event would be of no consequence.
Similar to the inadvenent RAS, an inadvertent EFW actuation at power would be undesirable but would not place the plant in a transient.
j The Emergency Feedwater Actuation System (EFAS) is comprised of two trains (EFAS 1 and EFAS 2) that provide EFW to the Steam Generators. EFAS 1 controls the EFW to Steam l
Generator 1 (A) and EFAS 2 controls the EFW to Steam Generator 2 (B). Each train of EFAS is comprised of four channels that is actuated by a 2-out-of-4 logic. The EFAS performs the following functions:
Starts the EFW pumps; e
Determines whether a steam generator is intad l
e Opens the EFW valves to the intact steam generator; and, Prevents a high level condition in the intact steam generator (s) by closing the EFW valves e
when the water level is re-established above the low level trip setpoint.
Each channel of EFAS is actuated by a monitoring circuit comprised oflow Steam Generator Level, low Steam Generator Pressure, and high Steam Generator AP When applying Table l
3.3-3 Action 10 or 11 reference to placing a channel of" Steam Generator 1 AP (EFAS 1)" in bypass or trip within one hour, the reference is to the entire monitoring circuit.
l EFAS is initiated to Steam Generator 1 either by a low steam generator level coincident with no low pressure trip present on Steam Generator 1 or by a low steam generator level i
coincident with a differential pressu:e between the two steam generators with the higher j
pressure in Steam Generator 1 (the intact Steam Generator). An identical EFAS is generated i
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Attachment to 2CAN079702 Page 3 of 6 for Steam Generator 2.
Safety Analysis Report (SAR) Figure 7.3-2 (b) reflects a typical EFAS logic.
In reviewing Table 3.3-3 Actions 10 and 11, a deficiency was found with these Actions la that they do not clearly account for d the applicable modes. Actions 10 and 11 apply to all the functional units in Table 3.3-3 except for loss of power. The associated functional units include a range of Applicable Modes of 1,2, 3, and 4. Actions 10 and 11 include the words "STARTUP and/or POWER OPERATION" which are defined terms listed in TS Table 1.1 as modes 1 and 2. These Actions could be read as not being applicable to the remainder of the applicable modes. This would exclede modes 3 for all of the functional units except for loss l
of power and mode 4 for EFAS. The Reactor Protective Instrumentation Table 3.3-1 Actions 2 and 3 were also reviewed and found to contain the same deficiency and both have been corrected by this change.
The current Table 3.3-3 Actions 10 and 11 and Table 3.3-1 Actions 2 and 3 in the ANO-2 TS are consistent with a previous revision of the Combustion Engineering Standard TS and do not clearly account for all applicable modes. The latest revision of the improved standard TS (ITS), NUREG-1432, " Standard Technical Specifications for Combustion Engineering Plants," has corrected this deficiency. This change also reflects how the current TS Table 3.3-1 Actions 2 and 3 and Table 3.3-3 Actions 10 and 11 are implemented at ANO-2.
DISCUSSION OF CHANGE I
TS 3/4.3.2 Table 3.3-1 (Actions 2 and 3) and Table 3.3-3 (Actions 10 and 11) were modified by the removal of the words "STARTUP and/or POWER OPERATION" and inserting the words " operation in the applicable MODES." This change was necessary to ensure that d
. the applicable modes were covered by these actions. The proposed allowance for these actions to be applied to all the applicable modes is consistent with the latest revision of the l
ITS.
Table 3.3-3 Action 10 continues to provide the guidance to place the inoperable channel in the bypass or trip condition within one hour. The proposed change to Action 10 will be used when one of the four channels of RWT Level - Low or Steam Generator AP channels are inoperable and desired to be placed in the tripped condition. The current Action 10 allows placing a channel in the tripped condition for as much as an entire operating cycle. An operating cycle for ANO-2 is approximately 18 months. The proposed Action 10 will reduce i
the time a RWT Level - Low and Steam Generator AP channels can be in the tripped i
l condition from approximately 18 months to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
i The current Action 11 will allow placing one channel in bypass and another channel in trip j
until the next required channel functional test. The current TS ESFAS channel functional i
testing frequency is one channel each month. This frequency implies that the current Action i
11 will essentially allow a channel to be in the tripped condition for approximately one month.
l The proposed change to Action 11 will reduce the time a RWT Level - Low or Steam I
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Attachment to 2CAN079702 Page 4 of 6 Generator AP channel can be in the tripped condition from approximately one month to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
The proposed change to Table 3.3-3 Action 11 will be used when two of the four ESFAS instrumentation channels are inoperable for a particular functional unit. Action 11 allows for continued operation in the applicable MODES provided that one of the inoperable channels is bypassed and the other inoperable channel is placed in trip within one hour. This action will allow an inoperable channel to be placed in the tripped condition to allow another channel to be bypassed for testing or repair purposes.
The 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> allowed time was selected since it is the current allowed outage time for other Reactor Protection System (Table 3.3-1 Actions 1 and 8) and Engineered Safety Features Actuation System (ESFAS) instrumentation (Table 3.3-3 Actions 9 and 13). This time period is also consistent with the allowed time for a channel to be in trip for the analog ESFAS instrumentation listed in the ITS for Combustion Engineering plants. Operating experience has demonstrated the very smah probability of a random failure of another RWT Level - Low or Steam Generator AP channelin a given 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period.
DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION An evaluation of the proposed change has been performed in accordance with 10 CFR 50.91(a)(1) regarding no significant hazards considerations using the standards in 10 CFR 50.92(c). A discussion nf these standards as they relate to this amendment request follows:
Criterion 1 - Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated.
The proposed change to the ANO-2 Technical Specifications (TS) modifies the allowed outage time tl.at a channel of the Refueling Water Tank (RWT) Level - Low or Steam Generator differential pressure (AP) can be in the tripped condition from a maximum of approximately 18 months when one channel is inoperable, and 31 days when two channels are l
inoperable, to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> for either of these conditions.
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If a channel of RWT Level Low is in the tripped condition and a single failure occurs that l
results in one of the other three channels of RWT Level - Low to actuate, a Recirculation l
Actuation System (RAS) signal would be generated. This scenario would not be considered severe if the condition occurred as a single event. However, during the injection phase of a Loss of Coolant Accident (LOCA) with a channel of RWT Level - Low in the trip condition with the above single failure, a premature RAS actuation would be the result. The premature RAS actuatbn would prevent the contents of the RWT from being injected into the reactor coolant system and possibly resulting in failure of both trains of Emergency Core Cooling System (ECCS) and the Containment Spray System.
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Attachment to 2CAN079702 Page 5 of 6 With one channel of Steam Generator AP in the tripped condition, as allowed by the TS, the plant is vulnerable to the single failure of a second Steam Generator AP channel under an unisolable Main Steam Line Break condition. The following scenario will result in the faulted Steam Generator being supplied feedwater by the Emergency Feedwater System during an unisolable Main Steam Line Break. One channel of Steam Generator AP is in the tripped condition as allowed by the TS and a Main Steam Line Break occurs that is urjsolable.
l During this event one of the remaining channels of Steam Generator AP fails resulting in incorrectly feeding the faulted Steam Generator. Reducing the time that a channel of RWT Level - Low or Steam Generator AP can be placed in the tripped condition will reduce the probability of these scenarios from occurring.
. The consequences of feeding the faulted Steam Generator during a main steam line break L
event or a premature RAS actuation during a LOCA are both significant. The proposed change reduces the allowed time a channel of RWT Level - Low or Steam Generator AP can be in the tripped condition. Reducing the time the channel can be in the tripped condition and thus, the exposure time to this scenario, would not be an accident initiator or involve an l
increase in the consequences of any accident previously evaluated.
The remaining proposed changes are consistent with NUREG-1432, " Standard Technical l
Specifications for Combustion Engineering Plants" and are intended to correct the actions required by TS Tables 3.3-1 and 3.3-3 to the current NRC approved guidance.
Therefore, this change does nol involve _ a significant increase in the probability or consequences of any accident previously evaluated.
Criterion 2-Does Not Create the Possibility of a New or Different Kind of Accident l
from any Previously Evaluated.
The proposed change does not modify the design or configuration of the plant. The proposed change provides a more conservative time limit for a channel to be in the tripped condition and provides the required actions when a channel is out of service There has been no physical change to plant systems, stmetures or components nor will the proposed change i
reduce the ability of any of the safety related equipment required to mitigate anticipated operational occurrences or accidents. This change will potentially increase the ability of safety related equipment to perform their functions. The configuration allowed by the proposed l
specification is permitted by the existing specification.
i Therefore, this change does not create the possibility of a new or different kind of accident
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from any previously evaluated.
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Attachment to 2CAN079702 Page 6 of 6 Criterion 3 - Does Not Involve a Significant Reduction in the Margin of Safety.
The proposed change provides a more restrictive time limit for a channel of RWT Level Low or Steam Generator AP to be in the tripped condition than is currently allowed by the TS. By reducing the allowed time, the probability is reduced that a single failure of another channel would result in a premature RAS actuation during the injection phase of a LOCA or the feeding of a faulted Steam Generator. By limiting the vulnerability to these events and their consequences, the proposed change will increase the margin of safety.
Therefore, this change does nqt involve a significant reduction in the margin of safety, c
Based upor the reasoning presented above and the previous discussion of the amendment request, Entergy Operations has determined that the requested change does agi involve a significant hazards consideration.
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