ML20153B438
| ML20153B438 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 09/18/1998 |
| From: | Dugger C ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20153B441 | List: |
| References | |
| W3F1-98-0168, W3F1-98-168, NUDOCS 9809230121 | |
| Download: ML20153B438 (8) | |
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a Entirgy Operations, Inc.
Killona. LA 70066-0751 Tel 504 739 6660 Charles M. Dugger vic res dent. Operations W3F1-98-0168 A4.05 PR September 18,1998 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555
Subject:
Waterford 3 SES Docket No. 50-382 License No. NPF-38 Emergency Technical Specification Change Request NPF-38-210 Gentlemen:
On September 11,1998, Waterford 3 submitted Technical Specification Change Request NPF-38-210 by Letter W3F1-98-0166. On September 16,1998, Waterford 3 discussed this change with members of your staff. Based on that conversation, Waterford 3 is superceding that Technical Specification Change Request with the attached request. The attached change replaces the September 11,1998 request in its entirety. The No Significant Hazards Evaluation is unchanged from the September 11,1998 submittal. Discussion of Reactor Coolant Flow - Low trip has j
been celeted from this submittal as this function is controlled by another bistable and is not applicable to this change.
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As discussed in the September 11,1998 submittal, this change could become an Emergency change contingent upon plant conditions. Since the September 11,1998 submittal, plant conditions have changed such that Waterford 3 placed the plant in h
Mode 4 on September 18,1998. This forced outage is the result of unexpected leakage from a pressurizer safety relief valve. As degradation of this valve could r.a be anticipated, this is an unplanned outage. The approval of this change will be required for plant restart from this forced outage. Based on this need and the criteria provided in 10 CFR 50.91a.5, Entergy Operations requests that this request be processed as an Emergency change. Entergy Operations requests the effective date for this change be upon receipt.
9809230121 9eo91st PDR ADOCK 05000382 P
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Emergency Technical Specification Change Request NPF-38-210 W3F1-98-0168 Page 2 September 18,1998 The attached description and safety analysis support a change to the Waterford 3 Technical Specifications. The proposed change modifies the Notes in Table 2.2-1 (Reactor Protective Instrumentation Trip Setpoints Limits) and Table 3.3-1 (Reactor Protective Instrumentation). The Reactor Protective System as designed and built is physically not capable of meeting the required TS 2.2.1 and 3.3.1 Logarithmic Power Level, Local Power Density, Departure from Nucleate Boiling Ratio, and Core Protection Calculator bypass removal setpoint simultaneously. The proposed change revises the Technical Specifications such that the existing design will be capable of satisfying the Technical Specification requirements. Without this change, the above Reactor Protective Systems would not meet the entry requirements for Mode 2 upon plant restart.
This proposed change has been evaluated in accordance with 10CFR50.91(a)(1),
using the criteria in 10CFR50.92(c), and it has been determined that this request involves no significant hazards consideration.
Should you have any questions or comments concerning this request, please contact Early Ewing at (504) 739-6242.
Very truly yours, q
C M. Dugger Vice President, Operations Waterford 3 CMD/ CWT /rtk Attachments:
Affidavit NPF-38-210 (w/ Attachments) cc:
E.W. Merschoff, NRC Region IV, C.P. Patel, NRC-NRR, J. Smith, N.S. Reynolds, NRC Resident inspectors Office, Administrator Radiation Protection Division (State of Louisiana), American Nuclear Insurers
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UNITED STATES OF AMERICA i
NUCLEAR REGULATORY COMMISSION j
l In the matter of
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I Entergy Operations, incorporated
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Docket No. 50-382 Waterford 3 Steam Electric Station
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l AFFIDAVIT a
Charles Marshall Dugger, being duly sworn, hereby deposes and says that he is Vice President Operations Waterford 3 of Entergy Operations, incorporated; that he is duly 1
authorized to sign and file with the Nuclear Regulatory Commission the attached Technical Specification Change Request NPF-38-210; that he is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of
' his knowledge, information and belief.
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u Charles Marshall Dugger Vice President Operations - Waterford 3 STATE OF LOUISIANA
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) ss PARISH OF ST. CHARLES
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Subscribed and sworn to before me, a Notary Public inynd for the Parish and State above named this ao day of M s 4,- d h.1998, c/
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Notary Public r
My Commission expires - +e' 4M
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- DESCRIPTION AND NO 'SIGNIFICANT HAZARDS EVALUATION 2
OF PROPOSED CHANGE NPF-38-210 o
i The proposed change modifies Note "1" of Table' 2.2-1 and Note "a" of Table 3.3-1 to
. better describe the performance of the 10-4 % Bistable by specifically addressing its
. hysteresis characteristics.
i Existing Specification j
See' Attachment A-Proposed Specification -
- See Attachment B
Background
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The Reactor Protective System (RPS) consists of sensors, calculators, logic, and other
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equipment necessary to monitor selected Nuclear Steam Supply System (NSSS) and
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containment conditions and to effect reliable and rapid Control Element Assembly (CEA) insertion (reactor trip) if any or a' combination of the monitored conditions approach specified safety system settings. The system's functions are to protect the -
core and Reactor Coolant System (RCS) pressure boundary for defined anticipated q
operational occurrences (AOOs) and also to provide assistance in limiting the consequences for certain postulated accidents. Four measurement channels with electrical and physical separation are provided for each parameter used in the direct generation of trip signals, with the exception of CEA position. A two-out-of-four coincidence of like trip signals is required to generate a reactor trip signal. The fourth channel is provided as an installed spare and allows bypassing _of one channel while a
maintaining a two-out-of-three system. Manual reactor trip is also provided.
Some reactor trip signals are provided with bypasses that are required to allow reactor startup. The High Logarithmic Power (HLP) trip and the Core Protection Calculators
-i (CPC) generated High Local Power Density (LPD) and Low Departure from Nucleate Boiling Ratio (DNBR) trips are bypassed at prescribed power levels since these trips would generate an unnecessary trip signal during reactor startup and power increase.
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I DescriDtion and Safety Considerations The Notes in Table 2.2-1 and Table 3,3-1 are being changed. Specifically, the proposed change modifies Note "1" of Table 2.2-1 and Note "a" of Table 3.3-1 to accurately describe the value for automatic bypass removal to allow the reset value for the 10-4 % Bistable to be below the setpoint by 3.0x10-5 % RATED THERMAL POWER to account for hysteresis.
The high logarithmic power level trip is provided to trip the reactor when indicated -
neutron flux power reaches a preset value. The flux signal used is the logarithmic power signal originating from the excore detectors in each nuclear instrument safety channel. The nominal trip setpoint is < 0.257 % of full power. This trip provides protection against inadvertent CEA withdrawals initiated from subcritical conditions (less than 10-4 % power) as described in FSAR Section 15.4.1.1. The high logarithmic power level bypass is provided to allow reactor power to be increased above 0.257 %
power into Mode 1 during a controlled reactor startup. Without the bypass in place, a reactor trip is generated when the trip setpoint is reached, thereby preventing further power increases. The bypass may be manually inserted above 10-4 % power and is automatically removed when the 10-4 % Bistable is reset at a power below 10-4 %.-For
. example, when power increases during a reactor startup to the 10-4 % Bistable setpoint, a permissive signal is generated to allow the operator to bypass the HLP trip function. The manual bypass by the operator will occur only during a controlled power increase and not if the increase is due to an inadvertent CEA withdrawal. When power decreases to the 10-4 % Bistable reset value, the bypass is automatically removed.
. The automatic bypass removal ensures that the trip will be available in the event of a CEA withdrawal from subcritical conditions (below 10-4 % power).
Additionally, safety analyses described in FSAR section 15.4.1.2 assume that a CEA withdrawal from critical conditions can be initiated from the lowest power level (the most limiting initial condition) at which the HLP trip is not available. In this case, since the HLP trip is not available, a reactor trip is generated by the CPC variable overpower trip function. The initial power level could theoretically be just above the 10-4 % Bistable reset value'during a shutdown if the power decrease were stopped at this power level which is highly improbable. Furthermore, since Waterford 3's standard practice is to trip the reactor between 5% - 10% power during a shutdown, it is extremely unlikely that this condition would occur.
The CPC provides reactor trips on DNBR and LPD when core power is above 10-4 %.
The CPC also generates a reactor trip signal when RCS conditions are outside the range for which CPC is applicable (e.g., all RCPs not running, shutdown CEA banks not fully withdrawn, etc.). The CPC bypass, which bypasses the low DNBR and high LPD trips, is provided to allow reactor trip breakers to be closed in preparation for reactor 2
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l startup before all of the CPC range checks are met. If these range checks are not met, the CPC generates a reactor trip, thereby preventing CEA withdrawal and reactor startup. Safety E.nalyses credit a CPC trip at the 10-4 % Bistable setpoint when conditions do not meet the CPC range checks (e.g., shutdown bank withdrawal). The bypass may be manually inserted if power is below 10-4 % and is automatically l
removed when the power level increases to the 10-4 % Bistable setpoint. For example, when power increases during a reactor startup to the 10-4 % Bistable setpoint, the CPC
. bypass is automatically removed. This ensures that the CPC is available under l
conditions where LPD and DNBR are of concern. If conditions do not meet the CPC range limits, as in the case of a CEA shutdown bank not fully withdrawn, a reactor trip is l
immediately generated. When power decreases to the 10-4 % Bistable reset value, as after a reactor trip, a permissive signal is generated to allow the operator to bypass the CPCs.
The bypass permissive function and automatic bypass removal function are reflected in notes "1" and "5" of Table 2.2-1 and notes "a" and "c" of Table 3.3-1. The setpoint for both the CPC and HLP functions is given as 10-4 % of RATED THERMAL POWER.
However, a sing!e bistable is used to initiate both the permissive and automatic bypass removal for both the CPC and HLP trip functions. A single bistable cannot both energize and de-energize at the same value as required by the Technical Specifications (TS) due to hysteresis. The CPC automatic bypass removal and permissive for the HLP trip bypass occur at the bistable setpoint (nominally 10-4 % power). However, the i
HLP automatic bypass removal and permissive for CPC trip bypass occur at the reset j
value of the bistable, which is slightly below 10-4 % power. The demonstrated hysteresis is within 1.5% of a 0-10 volt range, which is within 3.0x10-5% of RATED THERMAL POWER of the 10-4% Bistable setpoint. Therefore, the reset value will be within a power of 3.0x10-5 % below the bistable setpoint value. Thus, literal compliance with the TS which requires both to occur at 10-4 % power is not possible.
If the bistable is set so that the High Log Power automatic trip bypass removal occurs at 104 % power, the CPC automatic trip bypass removal will be slightly above the required 10-4 % power TS value. If the bistable is set so that the CPC automatic trip bypass removal occurs at 10-4 % power, the High Log Power automatic trip bypass removal will be slightly below the required 10-4 % power TS value. Waterford 3 procedures follow the latter case with the CPC automatic trip bypass removal set at 10-4 % power.
No Significant Hazards Evaluation l
l The proposed change described above shall be deemed to involve a significant j
hazards consideration if there is a positive finding in any of the following areas:
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Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response
No The proposed changes modify the table notations for the 10-4 % Bistable in TS 2.2.1 and 3.3.1. The proposed changes to these trip bypass removal functions do not adversely impact any system, structure, or component design or operation in a manner that would result in a change in the frequency or occurrence of accident initiation. The reactor trip bypass removal functions are not accident initiators. System connections and the trip setpoints themselves are not affected by trip bypass removal setpoint variations.
Since the hysteresis for the 10-4 % Bistable is small, there is a negligible impact on the CEA withdrawal analyses. Revised analyses, accounting for slightly different bypass removal power levels caused by the bistable hysteresis, would result in negligible changes to the calculated peak power and heat flux for the l:
pertinent CEA withdrawal events. Therefore, the consequences of any accident previously evaluated will not significantly change.
L Therefore, the proposed change will not involve a significant increase in the probability or consequences of any accident previously evaluated.
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Will operation of the facility in accordance with this proposed change create the j
possibility of a new or different type of accident from any accident previously evaluated?
Response
No The trip bypass removal functions in question protect against possible reactivity events. The power, criticality levels, and possible bank withdrawals associated with these trip functions have already been evaluated. Therefore, all pertinent reactivity events have previously been considered. Slight differences in the j
p power level at which the automatic trip bypass removal occurs can not cause a different kind of accident.
There has been no changes to any plant system, structure, or component, nor will these changes reduce the ability of any of the safety-related equipment 7
i required to mitigate AOOs.
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Therefore, the proposed change will not create the possibility of a new or j
different kind of accident from any accident previously evaluated, 3.
Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?
Response
No The safety function associated with the CPC and HLP trip functions are maintained. Since the hysteresis for the 10-4 % Bistable is small, there is a negligible impact on the CEA withdrawal analyses. Calculated peak power and heat flux are not significantly changed as a result of the bistable hysteresis. All acceptance criteria are still met for these events. There is no change to any margin of safety as a result of this change.
Therefore, the proposed change will not involve a significant reduction in a margin of safety.
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Safety and Significant Hazards Determination Based on the above No Significant Hazards Evaluation, it is concluded that: (1) the proposed change does not constitute a significant hazards consideration as defined by 10CFR50.92; and (2) there is a reasonable assurance that the health and safety of the public will not be endangered by the proposed change; and (3) this action will not result in a condition which significantly alters the impact of the station on the environment as described in the NRC final environmental statement.
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