ML20113C527

TS Change Request NPF-38-175 to License NPF-38,modifying TS 3/4.3.3.6, Accident Monitoring Intrumentation

ML20113C527
Person / Time
Site:	Waterford
Issue date:	06/27/1996
From:	Sellman M ENTERGY OPERATIONS, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20113C528	List: ML20113C527 ML20113C531
References
W3F1-95-0210, W3F1-95-210, NUDOCS 9607010318
Download: ML20113C527 (12)

Text

Ente gy Operatioris,Inc.

" .. Killona. LA 7006G0751 Tel 504 739 6660 Mike Sellman

c. President, operawns W3F1-95-0210 A4.05 PR June 27,1996 l ' U.S. Nuclear Regulatory Commission Attn: Document Control Desk '

Washington, D.C. 20555 l

Subject:

Waterford 3 SES l Docket No. 50-382 License No. NPF-38 Technical Specification Change Request NPF-38-175 Gentlemen:

The attached description and safety analysis supports a change to the Waterford 3 Technical Specifications (TS). This change will modify specification 3/4.3.3.6,

" Accident Monitoring Instrumentation," based on the Combustion Engineering improved Standard Technical Specifications (STS) approved and issued by the NRC as NUREG 1432. This change will revise the TS to include Accident Monitoring Instrumentation as recommended by Regulatory Guide (RG) 1.97, L Revision 3.

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The proposed change has been evaluated in accordance with 10CFR50.91(a)(1) using criteria in 10CFR50.92(c) and it has been determined that the proposed l change involves no significant hazards considerations. The Plant Operations l Review and Safety Review Committees have reviewed and accepted the proposed change based on the evaluation mentioned above.

l Waterford 3 requests that the implementation date for this change be within 90 days of NRC issuance of the amendment to allow for distribution and procedural revisions necessary to implement this change. Although this request is neither exigent nor emergency, your prompt review is requested.

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Technical Specification Change Request NPF-38-175 l W3F1-95-0210 Page 2 June 27,1996 l

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Should you have any questions or comments concerning this request, please contact Paul Caropino at (504)739-6692.

Very truly yours, l

,/ w M.B. Sellman Vice President, Operations I Waterford 3 l MBS/DFUtjs

Attachment:

Affidavit NPF-38-175 cc: L.J. Callan, NRC Region IV C.P. Patel, NRC-NRR R.B. McGehee N.S. Reynolds NRC Resident inspectors Office l Administrator Radiation Protection Division (State of Louisiana)

American Nuclear Insurers l

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. UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION in the matter of ) l

. ) i l Entergy Operations, incorporated ) Docket No. 50-382 Waterford 3 Steam Electric Station )

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I l AFFIDAVIT l

l M.B. Sellman, being duly sworn, hereby deposes and says that he is Vice President l Operations - Waterford 3 of Entergy Operations, incorporated; that he is duly l authorized to sign and file with the Nuclear Regulatory Commission the attached l Technical Specification Change Request NPF-38-175; that he is familiar with the  ;

I 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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l M.B. Sellman i l Vice President Operations - Waterford 3 l

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? l STATE OF LOUISIANA )

) ss 1 PARISH OF ST. CHARLES )

Subscribed and sworn to before me, a Notary Public in and for the Parish and State above named this 2 fM day of d o w c. 1996.

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d~L1?.Ti(fx Notary Public l

My Commission expires - w.r" c en l

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! DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-175 1

1 The proposed change will modify TS 3/4.3.3.6, " Accident Monitoring Instrumentation,"

by removing instruments from the TS which are not RG 1.97 Type A or Category 1 and i by adding others that are not currently addressed by this TS. It will also extend the allowed outage times for the post accident monitoring instruments and replace the current HOT SHUTDOWN requirement for the number of OPERABLE channels being less than the Required Number of channels with a Special Report requirement. This proposed change is based on the Combustion Engineering improved Standard Technical Specifications (STS) approved and issued by the NRC as NUREG 1432.

Existina Specification l

See Attachment A Proposed Specification See Attachment B Backaround '

l The primary purpose of the accident monitoring instrumentation is to display plant variables that provide information required by the control room operators during accident situations. This information provides the necessary support for the operator to take manual actions, for which no automatic control is provided, that are required for safety systems to accompiish their safety functions for Design Basis Events. The ]

OPERABILITY of post accident monitoring (PAM) instrumentation ensures that there is i sufficient information available on selected plant parameters to monitor and assess ,

plant status and behavior following an accident.

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The availability of PAM instrumentation is important so that responses to corrective actions can be observed and the need for, and magnitude of, further actions can be j determined. These essential instruments have been identified by Waterford 3 per the recommendations of Regulatory Guide 1.97, as required by Supplement 1 to NUREG-0737, "TMI Action items". By letter dated February 28,1991, Waterford 3 changed its commitment from RG 1.97, Revision 2 to Revision 3 and submitted information regarding the implementation of RG 1.97, Revision 3. The NRC staff's safety evaluation dated July 12,1993 accepted the Waterford 3 submittal as being in l conformance with, or justified in deviating from, the guidance of RG 1.97, Revision 3. i 1

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Type A variables are included in the TS because they provide the primary information i required to permit the control room operator to take specific manually controlled

actions, for which no automatic control is provided, that are required for safety systems ,

l to accomplish their safety functions for Design Basis Accidents (DBAs). l Category 1 variables are the key variables deemed risk significant because they are needed to:

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. Determine whether other systems important to safety are performing their intended

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l . Provide information to the operators that will enable them to determine the potential I for causing a gross breach of the barriers to radioactivity release; and 1

. - Provide information regarding the release of radioactive materials to allow for early l l indication of the need to initiate action necessary to protect the public as well as to

! obtain an estimate of the magnitude of any impending threat.

As mentioned above, these key variables are identified by Waterford 3's Regulatory l

Guide 1.97 analysis which has been approved by the NRC.

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Description i

TS 3/4.3.3.6 ensures the OPERABILITY of Regulatory Guide 1.97 Type A variables, so that the control room operating staff can:

. Perform the diagnosis specified in the emergency operating procedures. These l variables are restricted to preplanned actions for the primary success path of DBAs; and l

l Take the specified, preplanned, manually controlled actions, for which no automatic l control is provided, that are required for safety systems to accomplish their safety functions.

l TS 3/4.3.3.6 also ensures OPERABILITY of Category 1, non-Type A variables. This l ensures the control room operating staff can:

. Determine whether systems important to safety are performing their intended functions;

e Determine the potential for causing a gross breach of the barriers to radioactivity l release; 2

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• Determine if a gross breach of a barrier has occurred; and i

j e Initiate action necessary to protect the public as well as to obtain an estimate of the

magnitude of any impending threat.

j Category 1, non-Type A instruments are included in the TS because they assist

, operators in minimizing the consequences of accidents. Therefore, these Category 1, j non-Type A variables are important in reducing public risk.

in general, Category.1 provides for full qualification, redundancy, and continuous real-time display and requires onsite (standby) power. Category 2 provides for qualification but does not require seismic qualification and requires only a high-reliability power source (not necessarily standby power). Category 2 generally applies to instrumentation designated for indicating system operating status, in addition to Type A and Category 1 instruments, the current Waterford 3 TS 3/4.3.3.6 includes the following RG 1.97 Category 2 instruments: Steam Generator Pressure, Refueling Water Storage Pool Water Level, Emergency Feedwater Flow Rate, Reactor  !

Coolant System Saturation Margin Monitor, Safety Valve Position Indicator, and Containment Water Level (Narrow Range). The selected instruments for this TS were  ;

incorporated pursuant to the recommendations of Generic Letter 83-37. l The proposed change will remove the above listed Category 2 instruments from the  !

requirements of TS 3/4.3.3.6. This change is consistent with the CE improved STS and associated safety analyses which require only RG 1.97 Type A and Category 1, non-Type A instruments. All accident monitoring instruments designated by Waterford 3's hG 1.97, Revision 3 analysis as Type A and/or Category 1 are governed by the Waterford 3 TS. The following Type A and/or Category 1 instruments are governed by TS 3/4.3.3.6 and will remain so per this change: Containment Pressure, Reactor Coolant Outlet Temperature (Wide Range), Reactor Coolant inlet Temperature (Wide Range), Reactor Coolant Pressure (Wide Range), Pressurizer Water Level, Steam Generator Water Level (Narrow Range), Steam Generator Water Level (Wide Range),  !

Containment Water Level (Wide Range), Core Exit Thermocouples, Containment isolation Valve Position Indicators, Condensate Storage Pool Level, and Reactor Vessel Level Monitoring System.

The proposed change will add Containment Pressure (Wide-Wide Range) instruments to TS 3/4.3.3.6 to distinguish them from the Containment Pressure (Wide Range)

Instruments. The current Waterford 3 TS require 2 channels for Containment Pressure, but does not specify between Wide Range and Wide-Wide Range instruments. Both variables are required by RG 1.97 and both are designated as Category 1 per the Waterford 3 analysis. The need for this change was identified per the Waterford 3 3

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' Corrective Action Program. This proposed change will appropriately revise TS Tables .

3.3-10 and 4.3-7 to include both sets of Containment Pressure instruments. '

i The proposed change will revise the number of required channels of Steam Generator l Water Level (Wide Range) from 1/ steam generator to 2/ steam generator. The current l requirement of one instrument per steam generator is based on the assumption that other parameters, such as EFW Flow, can be used to assess steam generator level if direct indication is lost. This philosophy is depicted by a note (") in the current TS and is further explained in the FSAR. As mentioned earlier, the EFW Flow Rate .i instruments are being removed from TS 3/4.3.3.6 based on the fact that they are not Category 1 components. Taus the note (") which allows the substitution of EFW Flow Rate for Steam Generator Water Level (Wide Range) will also be removed. In place of this previous philosophy, TS 3/4.3.3.6 will be changed to require two channels of SG

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Water Level (Wide Range) per steam generator, since two channels are available and both are designated as Category 1. The BASES of the CE revised STS state that two channels are required to be OPERABLE for all but one function (Containment isolation Valve Positions) to preserve single failure criteria. The proposed change is consistent with this requirement.

It should be noted that several Type A and/or Category 1 instruments are not included in TS 3/4.3.3.6 but are governed by other TS. The following parameters and associated instrumentation ere govemed by TS other than TS 3/4.3.3.6 and will remain so per this proposed change: Containment Area High Range Radiation (TS 3/4.3.3.1),

Containment Hydrogen Concentration (TS 3/4.6.4.1), and Radioactivity Concentration (Gamma Spectrum)(TS 6.8.4.d). The required Actions for TS 3.3.3.1 and 3.6.4.1 are consistent or more restrictive than those proposed for TS 3.3.3.6 per this change. TS .

6.8.4.d governs the Post Accident Sampling System (PASS) program which provides for grab sampling reactor coolant during post-accident conditions for radioisotope l analysis. The ability to obtain these samples is ensured by periodic testing of the system as required by the Waterford 3 FSAR.

Four channels of Log Power Indication (Neutron Flux), which is a RG 1.97 Category 1 )'

variable, are currently governed only by TS 3/4.3.1, " Reactor Protective Instrumentation." Only two of the channels, C and D, are credited for meeting the requirements of RG 1.97. TS 4.3.1.1 does not distinguish between the two RG 1.97 channels and the two non-RG 1.97 channels. Therefore this proposed change will add channels C and D of Log Power indication to TS 3/4.3.3.6 to ensure the LCO and related Actions for these channels are consistent with the other accident monitoring instrumentation. As a result, both TS will apply to these instruments. The proposed change also adds a note (***) to TS 3.3.3.6 to clarify this.

This change will also revise the Action requirements and associated Allowed Outage Tirr,es (AOT) for TS 3/4.3.3.6 to be consistent with those outlined in the CE improved STS. The applicable safety analysis for each of these requirements is included in the 4

' BASES of the CE improved STS and have been approved by the NRC as NUREG 1432. The changes are as follows:

. The AOT for the number of OPERABLE channels being less than the Required Number of channels will be changed from 7 days to 30 days. The 30 day AOT is based on operating experience and takes into account the remaining OPERABLE channel, the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation during this interval. This 30 day AOT will be incorporated into Action 29 of TS 3/4.3.3.6 per this change.

j . The current HOT SHUTDOWN requirement for the number of OPERABLE channels i being less than the Required Number of channels will be changed to a 14 day

Special Report requirement. This report discusses the results of the root cause

{ evaluation of the inoperability and identifies proposed restorative actions as well as i alternate means of monitoring the function. This Action is appropriate in lieu of a shutdown requirement, given the likelihood of plant conditions that would require

information provided by this instrumentation. Also, alternative actions are identified before a loss of functional capability condition occurs. This Special Report i requirement will be incorporated into Action 29 of TS 3/4.3.3.6 per this change.

l . The AOT for the number of OPERABLE channels being less than the Minimum j Channels OPERABLE will be changed from 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to 7 days. The 7 day AOT is i ,

based on the relatively low probability of an event requiring PAM instrumentation

! operation and the availability of alternate means to obtain the required information.

l~ Continuous operation with less than the Minimum Channels OPERABLE for a j function is not acceptable because the alternate indications may not fully meet all performance qualification requirements applied to the PAM instrumentation.

! Therefore, requiring restoration of one inoperable channel of the function limits the j risk that the function will be in a degraded condition should an accident occur. This

7 day AOT will be incorporated into Action 30 of TS 3/4.3.3.6 per this change. The

, 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> HOT SHUTDOWN requirement of Action 30 will remain the same. ,

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. The current 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> AOT for the Reactor Vessel Level Monitoring System (RVLMS) will be changed to a 7 day AOT for the same reasons described above. The i Special Reporting requirement currently in TS for the RVLMS (less than the i Minimum Channels OPERABLE) will remain essentially the same except that it will j be changed to a 14-day report and will require a discussion of what alternate means of monitoring are being implemented as well as a schedule for restering the normal channels. This 7 day AOT and 14 day Special Report requirement will be incorporated as Action 31 of TS 3/4.3.3.6 per this change.

As mentioned previously, this change removes the Table 3.3-10 notes associated with EFW Flowrate (**) and Containment Water Level (Narrow Range) (***) due to these 5

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' nstruments i being removed from the table. This change also removes the Table 3.3-10 note (*) associated with Containment Isolation Valve Position Indicators. This note is unnecessary because the bases clearly identifies that TS 3/4.3.3.6 applies only to RG 1.97 Class A and Category 1 instruments. For clarification purposes, this change reassigns the note symbols as follows: (*) replaces (#), (**) replaces (****), and (***) is ,

used for the new note associated with Log Power Indication.  !

Finally, this change will revise the BASES section of TS 3/4.3.3.6 in accordance with

, the above description. The proposed BASES are consistent with those provided in NUREG 1432. 1 Safety Analysis  ;

The proposed change described above shall be deemed to involve a significant hazards consideration if there is a positive finding in any of the following areas:

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1. 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 change deletes all non-Type A and non-Category 1 instruments from the requirements of TS 3/4.3.3.6, " Accident Monitoring Instrumentation."

Type A variables provide the primary information required to permit the control room operators to take specific manually controlled actions, for which no automatic control is provided, that are required for safety systems to accomplish their safety functions during a DBA. Category 1, non-Type A variables are important in reducing public risk and are retained in TS because they are intended to assist operators in minimizing the consequences of accidents. ,

Category 2 instruments are generally designated for indicating system operatirig '

status and are not designated as essential key variables necessary for the safe i shutdown of the plant. The proposed change preserves the safety requirements of RG 1.97, Revision 3, and will not adversely affect any material condition of the plant that could directly contribute to causing or mitigating the affects of an accident.

The proposed change also adds two parameters to TS 3/4.3.3.6 which were previously controlled administratively or per another TS. Containment Pressure (Wide Wide Range) is being added because it is a Category 1 parameter

required in addition to Containment Pressure (Wide Range), which is currently in l the TS. Neutron Flux is being added to distinguish the RG 1.97 channels from the non-RG 1.97 channels and to provide action and surveillance requirements j 6 l

l consistent with the other accident monitoring instrumentation. These additions l to TS 3/4.3.3.6 contribute to the overall safety of the plant and therefore in no l

way increase the probability or consequences of an accident previously 1 evaluated.

I Additionally, the proposed change also extends the AOTs for TS 3/4.3.3.6 and replaces the HOT SHUTDOWN requirement for the number of OPERABLE l channels being less than the Required Number of channels with a Special l Report requirement. These changes are based on the relatively low probability of an accident occurring which would require these instruments, the passive nature of these instruments, and alternate means of monitoring available. This i is consistent with the CE improved STS and associated safety analyses which

have been approved and issued by the NRC as NUREG 1432.

The remainder of the proposed change provides enhancements and l'

clarifications to TS 3.4.3.3.6 which have no potential to impact plant operations.

No previous accident scenario is changed, and initiating conditions and assumptions remain as previously analyzed. Therefore, the proposed change l will not involve a significant increase in the probability or consequences of any I accident previously evaluated.

2. Will operation of the facility in accordance with this proposed change create the possibility of a new or different type of accident from any accident l l previously evaluated? l l \

l Response: No. l l

l l The proposed change will not alter the op ration of the plant or the manner in l

which the plant is operated. No new or different failure modes have been

' introduced. TS 3/4.3.3.6 ensures the OPERABILITY of essential Post Accident MonitorinD Instrumentation. This instrumentation provides information to the control room operators durina an accident so that appropriate actions can be taken to mitigate the consequences of the accident. These instruments are passive in nature in that no critical automatic action is assumed to occur from these instruments. Therefore, the proposed change will not create the possibility I of a new or different kind of accident from any accident previously evaluated.

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' 3. Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?

Response: No 1

l The proposed change revises TS 3/4.3.3.6 based on the information provided in CE improved STS, NUREG 1432. The deletion and addition of specific components from the TS per this change is commensurate with the safety l significance of their associated parameters. The proposed change ensures the  !

' operability of the post accident monitoring instrumentation which has been  ;

designated, by RG 1.97 and Waterford 3's associated analysis, as essential for I availability during and following a DBA. The proposed change preserves the ,

single failure criteria required for this instrumentation and maintains the level of 1 i safety currently established in the Technical Specifications. The proposed l change will not affect any physical protective boundary. Therefore, the proposed change will not involve a significant reduction in a margin of safety.

I Safety and Sianificant Hazards Determination Based on the above safety analysis, 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 i 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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i NPF-38-175 ATTACHMENT A ,

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