ML20070M317
| ML20070M317 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 04/20/1994 |
| From: | Robert Schaaf Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20070M319 | List: |
| References | |
| NUDOCS 9404260048 | |
| Download: ML20070M317 (15) | |
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7590-01 UNITED STATES NUCLEAR REGULATORY COMMISSION ENTERGY OPERATIONS. INC.
QOCKET NO. 50-458 NOTICE OF CONSIDERATION OF ISSUANCE OF AMENDMENT T0 ffCJJ.ITY OPERATING LICENSE. PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION. AND OPPORTUNITY FOR A HEARING The U.S. Nuclear Regulatory Commission (the Commission) is considering issuance of an amendment to facility Operating License No. NPF-47 issued to Entergy Operations, Inc. (the licensee) for operation of the River Bend Station, Unit 1, located in West Feliciana Parish, Louisiana.
The proposed amendment would revise various instrumentation technical specifications by extending the allowable outage times (A0Ts) of the instruments, and by increasing their channel functional surveillance test intervals (STIs) to quarterly.
The amendment also revises certain technical specification actions to address loss-of-function concerns associated with the A0T and STI changes.
Before issuance of the proposed license amendment, the Commission will have made findings required by the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations.
The Commission has made a proposed determination that the amendment request involves no significant hazards consideration. Under the Commission's regulations in 10 CFR 50.92, this means that operation of the j
facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different 9404260048 940420 POR ADOCK 05000458 P
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kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which'is presented below:
Reactor Protection System fRPS) 1.
The proposed change d:es not involve a significant increase in the probability or consequences of an accident previously-evaluated.
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These proposed changes do not involve a change to the plant design or operation, they simply involve the frequency at which testing of the RPS instrumentation is performed and the allowable outage time (A0T) for instruments.
Failure of the RPS instrumentation itself cannot c.reate an accident'.- As a result, these proposed changes cannot increase the probability of occurrence of any design basis accident previously evaluated <
As identified in NEDC-30851P, these propos,ed changes'inc se the average RPS failure frequey/ year) is considered to be y from 4.6x10' / year to 5.4x10' year. This. increase (8x10' insignificant. As identified in the NRC Staff's Safety" Evaluation Report of NEDC-30851P, this increase in average RPS failu're frequency would contribute to a very small increase in core-melt frequency.. The small increase in average RPS failure-frequency is offset by safety benefits such as a reduction in the number of inadvertent test-induced scrams, a reduction in wear due to excessive equipment test cycling, and better optimization-of plant-personnel resources. Hence, the net change in risk resulting from these proposed changes would be insignificant.
Therefore, these proposed changes do not result in a significant increase in either the accident previously eva. probability or the consequences of any luated.
2.
The proposed change does not create the possibility of a new'or different kind of accident from any previously evaluated.
The proposed changes do not result in any change to the plant.
design or operation, only to the A0T and frequency at which testing of the RPS instrumentation is performed. Since failure of the RPS instrumentation itself cannot create an accident, these proposed changes can at most affect only accidents which have been previously evaluated.
Therefore, these proposed changes cannot create the possibility of a new or different kind of accident from any accident previously evaluated.
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_ 3.
The proposed change does not involve a significant reduction in the margin of safety.
As identified above, these proposed changes increag/ y RPS failure frequency from 4.6x10 / year to 5.4x10' e the average 4
The NRC Staff's Safety Evaluation Report of NEDC-30851P concluded that this small average RPS failure frequency increase would contribute to a very small increase in core-melt frequency.
This small increase in average RPS failure frequency would be offset by safety benefits such as a reduction in the number of inadvertent test-induced scrams, a reduction on wear due to excessive equipment test c plant personnel resources.ycling, and better optimization of Hence, the net change in risk resulting from these proposed changes would be insignificant.
In addition, RBS has confirmed that the proposed changes to the functional test intervals will not result in excessive instrument drift relative to the current established setpoints.
Therefore, these proposed changes do not result in a significant reduction in a margin of safety.
Emeroency Core Coolina System (ECCS) 1.
The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
These proposed changes do not involve a change to the plant design or operation, they simply involve the frequency at which testing of the ECCS actuation Instrumentation is performed and the allowable outage time (A0T) for instruments.
Failure of the ECCS actuation instrumentation itself cannot :reate an accident.
As a result, these proposed changes cannot increase the probability of occurrence of any design basis accident previously evaluated.
As identified in NEDC-30936P (Part 2), these proposed changes increase the ca from 1.952x10lculated avepage water injection failure frequency to 1.992x10' per year for Case 58 and from 4
1.386x10 to 1.4 4
increase of 4x10'glx10 per year for Case SC.
T for Case 5B (2.0%) and 1.5x10'gis represents an for Case SC (1.1%), which are well within the acceptance criteria (4%)
provided in NEDC-30936P (Part 2).
The small increase in average water injection failure frequency is offset by safety benefits such as a reduction in the number of inadvertent test-induced scrams, a reduction in wear due to excessive equipment test cycling, and better optimization of plant personnel resources.
Hence, the net changa in risk resulting from these proposed
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changes would be insignificant.
Therefore, these proposed changes do not result in a significant increase in either the
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. probability or the consequences of any accident previously evaluated.
2.
The proposed change does not create the possibility of a new or
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different kind of accident from any previously evaluated.
The proposed changes do not result in any change to the plant design or operation, only to the A0T and frequency at which testing of the ECCS actuation instrumentation is performed.
Since failure of the ECCS actuation instrumentation.itself cannot create an accident, these proposed changes can at most affect only accidents which have bean previously evaluated. Therefore, j
these proposed changes cannot create the possibility of a new or different kind of accident from any accident previously evaluated.
3.
The proposed change does not involve a significant reduction in the margin of safety.
As identified above, these proposed changes increase the calculate,d average wat injection failure frequency from to 1.40lx10'g 1.992x10'gr 1.952x10' t per year for Case 58 and from 1.386x104 per year for Case SC.
This increase is well within the acceptance criteria found acceptable in the NRC Staff's Safety Evaluation Report for NEDC-30936P (Part 2). This small increase in average ECCS actuation failure frequency would be offset by safety benefits such as a reduction in the number of inadvertent test-induced scrams, a_ reduction on wear due to excessive equipment test c plant personnel resources.ycling, and better optimization of Hence, the net change in risk resulting from these proposed changes would be insignificant.
In addition, RBS has confirmed that the proposed changes to the functional test intervals will not result in excessive instrument drift relative to the current, established setpoints.
Therefore, the proposed changes do not result in a.significant reduction in a margin of safety.
Control Rod Block Instrumentation 1.
The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
These proposed changes do not involve a change to the plant design or operation, only the Allowable Outage Time (A0T) and frequency at which testing of the Control Rod Block Instrumentation is performed.
Failure of the Control Rod Block instrumentation itself cannot create an accident. As a result, these proposed changes cannot increase the probability of occurrence of any design basis accident previously evaluated.
! As identifieo in NEDC-30851P, Supplement 1, these proposed changes increase the average Control Rod Block failure frequency less than 0.06%. As provided'in the NRC Staff's Safety Evaluation Report of NEDC-30851P, Supplement 1, this increase is very slight and b offset by the safety benefits associated with the proposed changes to the RPS and Control Rod Block Instrumentation.
As a result, the combined effect of the changes proposed for the RPS and Control Rod Block Instrumentation requirements should result in an overall improvement in plant safety.
Therefore, these prnposed changes do not result in a significant increue in either the probability or the consequences of any accident previously evaluated.
2.
The proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
The proposed changes do not result in any change to the plant design or operation, only to the A0T and frequency at which testing of the Control Rod Block instrumentation is performed.
Since failure ef the Control Rod Block instrumentation itself cannot create an nccident, these proposed changes can at most affect only accidents which have been previously evaluated..
Therefore, these proposed changes cannot create the possibility of a new or different kind of accident from any accident previously evaluated.
3.
The proposed change does not involve a significant reduction in the margin of safety.
As identified above, these proposed changes increase the average Control Rod Block failure frequency less than 0.06%.
This increase is very slight and is offset by the safety benefits associated with the proposed changes to the RPS and Control Rod Block Instrumentation.
As a result, the combined effect of the changes proposed for the RPS and Control Rod Block Instrumentation requirements should result in an overall improvement in plant safety.
In addition, RBS has confirmed that the proposed changes to the functional test intervals will not result in excessive Instrument drift relative to the current, established setpoints.
Therefore, the proposed changes do not result in a significant reduction in a margin of safety.
Isolation Actuation Instrumentation 1.
The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
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j These proposed changes oJ not involve a change to the plant design or operation, only the Allowable Outage Time (A0T) and l
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. frequency at which testing of the Isolation Actuation instrumentation is performed.
Failure of the Isolation Actuation instrumentation itself cannot create an accident. As a result, these proposed changes cannot increase the probability of occurrence of any design basis accident previously evaluated.
1 As identified in NEDC-30851P, Supplement 2, these proposed changes to the surveillance test interval requirements for the Isolation Actuation instruments which are common to RPS or ECCS have a negligible (less than 1%) impact on the average isolation unavailability when combined with the individual valve failure probability, and that the changes to the A0Ts has [have] less than a 2% impact. The analyses demonstrate that the individual valve failure probabilities dominate the overall isolation failure probability. As provided in the NRC Staff's Safety Evaluation Report of NEDC-30851P, Supplement 2, these proposed
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changes would have a very small impact on plant risk. As a result,_overall plant safety is not reduced by these proposed changes.
As identified in NEDC-31677P, the proposed changes to the requirements for Isolation Actuation instrumentation ot common to RPS or ECCS result 'in a small decrease of 1.97x10'p/ year in the average isolation failure frequency. As identified in the NRC Staff's Safety Evaluation Report of NEDC-31677P, the NRC agreed that these proposed changes are' acceptable because the failure frequency impact is minimal. As a result, overall plant safety is not reduced by these proposed changes.
The small increase in the average failure frequency of the instruments common to RPS or ECCS due to the proposed changes to the. Isolation Actuation instrumentation requirements is offset by safety benefits such as a reduction on the number of inadvertent test-induced scrams and engineered safety feature actuations, a reduction in wear due to excessive test' cycling, and better optimization of plant personnel resources. Hence, the net change in risk resulting from these proposed changes would be insignificant. Therefore, these proposed changes do not result in a significant increase in either the probability or the consequences of any accident previously evaluated.
2.
The proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
The proposed changes do not result in any change to the plant design or operation, only to the A0T and frequency at which testing of the Isolation Actuation instrumentation is performed.
Since failure of the Isolation Actuation instrumentation itself cannot create an accident, these proposed changes can at most affect only accidents which have been previously evaluated.
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Therefore, these proposed changes cannot create'the possibility of a new or different kind of accident from any accident previously evaluatedo n
3.
The proposed change does not involve a significant reduction in tha margin of safety.
As identified above, the proposed changes to the requkrements for
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Isolation Actuation instruments common to:RPS-or ECCS have a negligible impact on the average isolation unavailability when combined withithe individual valve = failure probability. The -
analyses demonstrate that the individual valve failure probabilities dominate the overall isolation failure probability.
The proposed changes to the requirements for-Isolation Actuation instruments not common to RPS or ECCS decrease the,ir average isolation failure frequency approximately 1.97x10' / year.
The small increase-in average Isolation Actuation instrumentation failure frequency of the instruments common to RPS or ECCS are offset bytthe safety benefits such asia < reduction on the number of ir 2dvertent test-induced scrams and engineered safety feature actu dions, a reduction in wear due to excessive test. cycling, and better optimization of plant personnel resources. As a result, the NRC Staff's Safety Evaluation Reports for. these BWROG reports concluded that these proposed changes would have a very small, impact on plant risk.
In addition, RBS has. confirmed that the proposed changes to the functional test intervals will not result in excessive instrument drift relative to the current, established setpoints.
Therefore, the proposed changes do not result in-a significant reduction in a margin of safety.
Other Technical Soecification Instrumentation 1.
The proposed change does not; involve a significant. increase in the probability or consequences of an accident previously evaluated.
These proposed changes do not involve a. change to the plant design or operation, only the Allowable Outage Time (A0T) and frequency at which testing of the associated instrumentation is performed. These instruments are designed to mitigate the consequences of previously analyzed accidents.
Failure of these instruments cannot increase, and is independent of, the probability of occurrence of such accidents. As a result, these proposed changes cannot increase the probability of any accident previously evaluated. As identified in GENE-770-06-01, although not specifically analyzed, these proposed changes are bounded by the results of the analyses discussed in Parts I through IV of this request. Such analyses have shown that the safety function
il failure frequency is not significantly impacted by similar proposed changes.
In addition, any increase in the probability of failure of these instruments to perform their required 4
I functions would be offset by safety benefits such as a reduction in the number of inadvertent test-induced scrams and engineered' safety features actuations, a reduction in wear due to excessive equipment test cyclin personnel resources. 6 and better optimization of plant Therefore, these proposed changes do not result in a significant increase in the probability or the
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J consequences of any accident previously evaluated._
2.
The proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
The proposed changes do not result in any change to the plant design or operation, only to the A0T.and, frequency at which te' sting of the associated instrumentation is performed. As a result, these proposed changes can at most affect.only accidents which have been previously evaldsted. Theref6re, these. proposed changes cannot create the possibility of a new or.different kind of accident from any accident previously evaluated.
3.
The proposed change does not involve a'significant reduction in the margin of safety.
i As identified in GENE-770-06-01, although not specifically analyzed, these proposed changes are bounded by the results of the analyses discussed in Parts I through IV of this raquest.
i Such analyses have shown that the safety function ~ failure frequency is not significantly impacted by similar proposed changes.
In addition, any~ increase in the probsbility of failure of these instruments to perform their required functions would be offset by safety benefits such as a reduction in the number of inadvertent test-induced scrams and engineered safety features actuations, a reduction in wear due'to excessive equipment test cycling, and.better. optimization of plant personnel resources.
As a result, these proposed changes will reduce overall plant risk.
In addition,.RBS has confirmed that the proposed changes to the functional test intervals will not result in excessive instrument drift relative to the current, established setpoints.
Therefore, these proposed changes do not involve a significant reduction in a margin of safety.
. Technical Soecification Chanaes Relatina to Loss-of-Function Issues 1.
The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
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The proposed changes associated with the " loss-of-function" checks ensure a plant configuration which would have the capability to automatically actuate.th,9 respective.
system / valve (s).
These instruments are designed to mitigate the consequences of previously analyzed accidents.
Failureofthese instruments cannot increase, and is independent of, the probability of occurrence of such accidents. As a result, the proposed changes cannot increase the probability of any accident previously evaluated. The proposed. changes do not involve a change to the plant design or operation and do not degrade the i
capability of the system (s) to perform,its required., function.
Further, these functions or tripped channel (s) in an isolation logic are not considered as. initiators for.any accidents previously analyzed. Therefore, these changes do not~
significantly increase the consequences of any, accident.
previously evaluated.
2.
The proposed change does not create the possibility of a new or different kind of, accident from any previously evaluated.
The proposed changes do not result in;any change to the plant-
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design and no new mode of plant operation is introduced. As a result, the proposed changes can at most affect only accidents which have been previously evaluated.
Therefore, the proposed 1
changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
3.
The proposed change does not involve a significant reduction in the margin of safety.
The proposed changes do not involve a significant reduction in a margin of safety since the required safety function 'of the inoperable channel Time (A0T) for seve(s) will be fulfilled. The allowable Outage ral trip functions'have been incr'ased but e
only in conjunction with the incorporation of the loss-of-function check which ensures a plant configuration which would have the capability to automatically actuate the respective system / valve (s). Therefore, the proposed changes do not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.'
3 l The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making-any final determination.
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Normally, the Commission will not issue the amendment until the expiration of the 30-day notice period. However, should circumstances change during the notice period such that failure to.act in a timely way would result, for example, in derating or shutdown of the facility, the Commission may issue the license amendment before the expiration of the 30-day notice period, provided that its final determination is that the-amendment involves no significant hazards consideration.
The final determination will consider all public and State comments received.
Should the Commission take this action, it will publish in the FEDERAL REGISTER a notice of issuance and provide for opportunity for a hearing after issuance.
The Commission expects that the need to take this action will occur very infrequently.
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Written comments may be submitted by mail to the Rules Review and Directives Branch, Division of Freedom of Information and Publications Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and should cite the publication date and page number of this FEDERAL REGISTER notice. Written comments may also be delivered to Room 6D22 Two White Flint North, 11555 Rockville Pike, Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555.
, The filing of requests for hearing and petitions for leave to intervene is discussed below.
By May 26, 1994
, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating _
license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene.
Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's " Rules of Practice for Domestic Licensing Proceedings" in 10 CFR Part 2.
Interested persons should consult a current copy of 10 CFR 2.714 which is available at the Commission's Public Document Room, the Gelma;i Building, 2120 L Street, NW., Washington, DC 20555 and at the local public document room located at the Government Documents Department, Louisiana State University, Baton Rouge, Louisiana 70803.
If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or an Atomic Safety and Licensing Board, designated by the Commission or by the Chairman of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the designated Atomic Safety and Licensing Board will issue a notice of hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular referencs to the following
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. factors:
(1) the nature of the petitioner's right under the Act to be made j
party to the proceeding; (2) the nature and extent of the petitioner's e
property, financial, or other interest in the proceeding; and (3) the possible effect of any order which may be entered in the proceeding on the petitioner's interest. The petition should also identify the specific aspect (s) of the subject matter of the proceeding as to which petitioner wishes to intervene. Any person who has filed a petition for leave to 4
intervene or who has been admitted as a party may amend the petition without requesting leave of the Board up to 15 days prior to the first prehearing conference scheduled in the proceeding, but such an amended
- petition must satisfy the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference scheduled in the proceeding, a petitioner shall file a supplement to the petition to int.ervene which must include a list of the contentions which are sought to be litigated in the matter. Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted.
In addition, the petitioner shall provide a brief explanation of the bases of the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner intends to rely in proving the contention at the hearing.
The petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner intends to rely to establish those facts or expert opinion.
Petitioner must provide sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact.
Contentions shall
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- be limited to matters within the scope of the amendment under consideration. The contention must be' one which, if proven, would entitle the petitioner to relief. A petitioner who fails to file such a supplement which satisfies these requirements with respect to at least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing, including the opportunity to present evidence and cross-examine witnesses.
If a hearing is requested, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment.
If the final determination is that the amendment request involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must be filed with the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Attention: Docketing and Services Branch, or may be delivered to the Commission's Public Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555, by the above
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date. Where petitions are filed during the last 10 days of the notice period, it is requested that the petitioner promptly so inform the Commission by a toll-free telephone call to Western Union at 1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator should be given Datagram Identification Number N1023 and the following message addressed to William D. Beckner, Director, Project Directorate IV-1:
petitioner's name and telephone number, date petition was mailed, plant name, and publication date and page number of this FEDERAL REGISTER notice.
A copy of the petition should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and to Mark J. Wetterhahn, Esq., Winston & Strawn, 1400 L Street, NW, Washington, DC 20005, attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended petitions, supplemental petitions and/or requests for hearing will not be entertained absent a determination by the Commission, the presiding officer or the presiding Atomic Safety and Licensing Board that the petition and/or request should be granted based upon a balancing of the factors specified in 10 CFR 2.714(a)(1)(1)-(v) and 2.714(d).
For further details with respect to this action, see the application for amendment dated January 14, 1994, which is available for public inspection at the Commission's Public Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555 and at the local public document
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room located.at the Government Documents Department, Louisiana State University, Baton Rouge Louisiana 70803.
Dated at Rockville, Maryland, this 20th day of April, 1994.
FOR THE NUCLEAR REGULATORY. COMMISSION Robert G. Schaaf, Acting Project Manager Project Directorate IV-1 Division of Reactor Projects III/IV.
Office of Nuclear Reactor Regulation J
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