NRC-98-0043, Application for Amend to License NPF-43 to Convert Current TS to Improved Ts.Meeting Is Requested to Discuss Review Schedule & Contents of Amend.Vol 1 Includes Application & Justification.Vols 2-13 Contain Improved TS
| ML20217Q493 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 04/03/1998 |
| From: | Gipson D DETROIT EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20217Q501 | List: |
| References | |
| CON-NRC-98-0043, CON-NRC-98-43, RTR-NUREG-1433 NUDOCS 9804130004 | |
| Download: ML20217Q493 (150) | |
Text
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- I Dougir.s R. Gipson f Senior Vice l' resident, Nuc lear Generation Ferm12 6400 North Dixie !!wy., Newport, Michigan 48166 l jT Tel: 313MC.5201 fax: 313M6.4172 l
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l 10CFR50.92 April 3,1998 l NRC-98-0043 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington D C 20555-0001
References:
- 1) Fenni2 NRC Docket No. 50-341 NRC License No. NPF-43 p 2) NUREG-1433, " Standard Technical Specifications - General Q Electric Plants, BWR/4" Revision 1, dated April 1995.
- 3) NEl 96-06," Improved Technical Specifications Conversion Guidance," dated August 1996.
Subject:
Proposed Technical Specification Change (License Amendment)-
Conversion to Imoroved Standard Technical Specifications Pursuant to 10 CFR 50.90, Detroit Edison hereby proposes to amend the Fenni 2 Plant Operating License NPF-43, Appendix A, Technical Specifications (TS) to g convert the Fermi 2 Current Technical Specifications (CTS) to be consistent with the l
Improved Standard Technical Specifications (ISTS)(Reference 2).
The proposed license amendment request to convert the Fermi 2 CTS to the Fermi 2 improved Technical Specifications (1TS) is enclosed with this letter. The proposed license amendment request was prepared considering the guidance of Nuclear Energy Institute (NEI) NEl 96-06, " Improved Technical Specifications Conversion Guidance," dated August 1996 (Reference 3). Since NEI 96-06 was issued after the start of the Fermi 2 ITS project, not all recommendations contained in NEI 96-06 could be strictly followed. Ilowever, following a series of meetings with the NRC fa' stafT, appropriate chaages were made in the submittal preparation process to address those issues necessary to facilitate a timely and efficient NRC review.
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The detailed description and justification of this proposed license amendment request consists of thirteen (13) volumes. A detailed description of the contents and organization of the 13 volumes is included in Attachments I through 4 of this letter, and are described below:
Attachment 1, " Synopsis of the Proposed License Amendment Request," describes the organization and content of the submittal.
Attachment 2, "Pending License Amendment Requests," provides a listing of all currently docketed license amendment requests and the pages in both the CTS and the ITS that are affected.
The pending License Amendment Requests, listed in Attachment 2, are not incorporated into this license amendment request, except for the request to modify the Emergency Diesel Generator out of service times (TAC Number M94171).
During the period of NRC review of the Fermi 2 ITS, supplemental submittals incorporating pending changes to the CTS will be made where necessary. These supplements will be scheduled with the NRC staff reviewers to ensure that the ITS
,m and CTS review processes are efficiently coordinated.
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V Attachment 3, " Approved ISTS Change Travelers," provides a listing of the NRC approved changes affecting the BWR/4 ISTS (Reference 2) and their incorporation status in this license amendment request.
Attachment 4, "'Beyond Scope' Changes," provides a listing of those changes that represent a change in the Fermi 2 current licensing basis ,vhich is beyond the scope of the conversion to ISTS. Detroit Edison has endeavored to limit these types of changes to those needed to successfully complete the ITS conversion. A brief explanation of the need for the change is also provided, implementation of the Fermi 2 ITS will require the performance of a number of new surveillance requirements. Detroit Edison intends to treat these new requirements as being " met" at the time ofimplementation, with the first performance of these new surveillance requirements scheduled to be completed within the required frequency from the date ofimplementation.
Detroit Edison intends to implement this proposed license amendment request at Fermi 2 in June 1999. This date is based on the time required for procedure revisions, including the development of new programs and training. In addition, this
, date considers the timing ofimplementation with respect to planned refueling, L ( ,) outages and the scheduling of an initial licensed operator examination in October
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NRC-98-0043 Page 3
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k- 1999. This date is also predicated on the NRC review being completed and e Safety Evaluation issued by March 1999.
Detroit Edison requests a meeting be scheduled, at your earliest convenience, to discuss the review schedule and contents of this license amendment request. Should you have any questions or require additional information, please contact either the Fermi 2 Licensing Director, Mr. Norman K. Peterson, at (734) 586-4258 or the Fermi 2 ITS Project Manager, Mr. Glen D. Ohlemacher, at (734) 586-4275.
Sincerely, Enclosures cc: A. B. Beach B. L. Burgess G. A. Ihtrris g
'O) A. J. Kugler Supervisor, Electric Operators, Michigan Public Service Commission
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I, DOUGLAS R. GIPSON, do hereby affirrn that the foregoing statements are based on facts and circumstances which are true and accurate to the best of my knowledge and belief.
h DOUGLAS R. OIPSON Senior Vice President, Nuciear Generation 5
On this O" AJ. day of. _,1998 before me personally appeared Douglas R. Gipson, being first' duly sworn and says that he executed the foregeing as his free act and deed.
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Notary Public ROSAl.lE A. Anygg i
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Fermi 2 Improved Technical Specifications Submittal Cover Letter and Split Report .
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Attachment I to NRC-98-0043 Page1 1
ATTACHMENT 1 Q SYNOPSIS OF THE PROPOSED LICENSE AMENDMENT REQUEST e
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Attachment 1 to 1 NRC-98-0043 Page 2 SYNOPSIS OF THE PROPOSED LICENSE AMENDMENT REQUEST The submittal for the conversion to Improved Technical Specifications (ITS) consists of 13 volumes and related attachments to the transmittal letter. The 13 volumes consist of i the Application of NRC Selection Criteria (Split Report), the ITS Section packages, a l compilation of Fermi 2 Current Technical Specification (CTS) marked up pages in C'IS order, and the complete Fermi 2 ITS and Bases. Below is a brief description of the contrats of the Split Report and each of the Section packages, as well as a brief explanation of how the material was prepared and the designations utilized.
APPLICATION OF NRC SELECTION CRITERIA The Application of NRC Selection Criteria document provides a discussion of how the criteria of 10 CFR 50.36(c)(2)(ii) were applied to the Fermi 2 CTS requirements. Also included is a Summary Disposition Matrix which, for each CTS specification, lists the j
title of the CTS specification, the new ITS specification number (if applicable), whether or not the CTS specification is retained in the ITS, the criterion of 10 CFR 50.36(c)(2)(ii) for retention if retained, the basis for inclusion or exclusion, and the proposed new location for relocated requirements.
A i' For those CTS requirements that do not meet any of the NRC selection criteria, an evaluation of the CTS requirement against the criteria of 10 CFR 50.36(c)(2)(ii) or other appropriate evaluation is provided in Appendix A of the Split Report. For CTS requirements being relocated or deleted for reasons other than failure to meet the 4 Selection Criteria for retention. Appendix B of the Splh Repon provides the evaluation supporting the relocation or deletion. If a relocation or deletion is associated with only a portion of a CTS Specification (as opposed to the entire CTS Specification) the evaluation is located with the associated ITS Specification's CTS Markup and Discussion of Changes (as described below). In both Appcndix A and Appendix B of the Split Report. the marked up CTS pages associated wit l1 the relocation are provided followed by the Discussion of Change (DOC) associated with che relocation or deletion.
SECTION PACKAGES FOR SECTIONS 1.0 TilROUGli 5.0 (15 SECTIONS)
Each of the Section packages corresponds to a Section of the proposed ITS. Each Section package is further subdivided into indnidual Specification packages corresponding to an individual ITS Specification or other appropriate subdivision of the ITS Section. Each Specification package contains the information necessary for review of the ITS Section .
and is organized as follows:
a
Attachment 1 to NRC-98-0043 Page 3 g
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- 3. CTS Markuns Contains annotated copies of the CTS pages which show the disposition of existing requirements into the proposed ITS. The pages are arranged in CTS order. The upper right hand corner of the CTS page is annotated with all the ITS Section numbers in which the CTS page occurs. Items on the CTS page that are addressed in other proposed ITS sections are annotated with the appropriate location. CTS pages are also annotated with a cross-reference to {
the corresponding ITS requirement.
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( Where a proposed ITS requirement differs from a CTS requirement, l
'O individual detai': of the CTS revision are annotated with alphanumeric designators which relate to the appropriate Discussion of Change (DOC). The i DOC provides a concise justification for the change. The DOCS are located l behind the CTS Markups of each Specification package. The a . hanumeric designators aisc to the evaluations supporting a finding of No Significant Hazaras Consideration (NSHC).
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The DOCS are numbered sequentially within each letter category for each ITS Specification. The proposed changes for each CTS requirement are separated into the following categories Desicnator Catecorv A ADMINISTRATIVE - changes to the CTS that ,i result in no additional or reduced restrictions or flexibility. These changes are supported in
,o aggregate by a single NSHC.
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Attachment I to NRC-98-0043 .
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fw) v R RELOCATIONS -items in the CTS do not meet I the selection criteria of 10 CFR 50.36(c)(2)(ii) and thus are being relocated to other licensee-controlled documents. These changes are supported in aggregate by a single NSHC. As discussed above, R DOCS and the associated CTS Markup pages are included in the Application of the NRC Selection Criteria.
M TECHNICAL CHANGES - MORE RESTRICTIVE
- changes to the CTS that result in added restrictions or reduced flexibility. These changes are supported in aggregate by a single NSHC.
L TECHNICAL CHANGES - LESS RESTRICTIVE
- changes to the CTS that result in reduced restrictions or added flexibility. There are two groups of changes " Generic" and " Specific"in this category. Each " Specific" LESS RESTRICTIVE change is supported by a p corresponding unique evaluation supporting a
() finding of NSHC. The " Generic" LESS RESTRICTIVE changes are subdivided into 4 subcategoiies each of which is identified uniquely as either LA, LB, LC or LR changes. Each subcategory of" Generic" LESS RESTRICTIVE CH ANGE is supported in aggregate by a single NSHC. The subcategories and their designation are as follows:
The "LA" changes consist of relocation of details out of the CTS and into the Bases, UFSAR, or other appropriate licensee-controlled document.
1 3pically, this involves details of system design and function. or procedural detail on methods of conducting surveillances.
The "LB" changes consist of changes to eliminate a requirement that a Surveillance Requirement be completed at an interval shorter than the normal .
interval just prior to the start of an activity, such as fuel handling, control rod withdrawal or removal, or control rod drive removal.
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The "LC" changes are related to the extension of an -
instrument Completion Time or Surveillance Frequency in accordance with NRC approved vendor topical reports.
The "LR" changes consist of removal of details from the CTS. These details are those of a nature that they do not need to be relocated to a licensee-controlled program with regulatory program controls (such as 10 CFR 50.59).
The CTS Bases pages are replaced in their entirety. A separate discussion in.
each Specificationjustifies the replacement.
- 5. Soecification Markuos Contains annotated copies of the applicable NUREG-1433, Revision 1, Specifications which show how the proposed ITS Specification differs from the NUREG Specification. Where a proposed ITS Specification requirement differs from the NUREG Specification (other than a change where the O NUREG contains bracketed material) individual details of the change are annotated with alphanumeric designators which relate to the appropriate Justification for Difference (JFD). Bracketed material changed only to
'directly substitute the corresponding CTS material is not annotated with a JFD designator. Any other substantive change to bracketed material is annotated with a JFD designator. The JFD provides a concisejustification for the change. The JFDs are located behind the Bases Markups of each Specification package. The JFDs are numbered sequentially within each letter category for each ITS Specification.
In addition to the JFD annotatmn. the Specification markups include a cross-reference to the corresponding CTS reference.
- 6. Bases Markuns Contains annotated copies of the applicable NUREG-1433, Revision 1, Bases which show how the proposed ITS Bases differs from the NUREG Bases.
Where the proposed ITS Bases differ from the NUREG Bases, other than a change where the NUREG contains bracketed material, individual details of ,
the change are annotated with alphanumeric designators which relate to the appropriate JFD. Bracketed material changed only to directly substitute the p corresponding CTS material is not annotated. Any other change to bracketed Q .
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Attachment I to NRC-98 0043 Page 6 ^
(3 p s material is annotated with a JFD designator. The JFD designation scheme for the Bases is the same as for the ITS Specifications.
- 7. IEDhi Ccntains the JFDs which describe the differences from the NUREG-1433,'
Revision 1, Specifications and Bases. A single JFD may be applicable to both the Specifications and Bases.
The differences are separated into the following categories.
Desienator Category P PLANT SPECIFIC CHANGES -
changes that reflect plant specific requirements and inforination.
C GENERIC CHANGES - changes that reflect generic modifications to NUREG 1433, Revision 1. These O changes have been approved by the NRC. Generic changes that were approved as of December 1,1997 are addressed. Attachment 3 to the submittal letter provides a summary of the generic change dispositions.
- 8. NSHC Contains the evaluations required by 10 CFR 50.91(a) supporting a finding of No Significant liazards Consideration (NSHC). Generic evaluations for a finding of NSilC have been wviten for each category of changes except for the " Specific" TECHNICAL Cil ANGES - LESS RESTRICTIVE category.
These " Specific" NSilC determinations are located together at the end of each Specification package. The " Generic" NSHC determinations are provided once at the end of the Split Report.
COMPILATION OF MARKED UP CTS PAGES The CTS Markups from each of the ITS specification packages .re presented in the order of the ITS. The compilation volume provides an entire ma'%p of the CTS in CTS order to facilitate NRC review efforts and to demonstrate that all CTS requirements are accounted for. In many instances, the same CTS page is used in different ITS sections.
Attachment I to NRC-98-0043 Page 7.
As a result, in the compilation volume, the CTS pages that are included in more than one ITS Specification package will appear with the annotations associated with each ITS Specification package in which the CTS page appears.
COMPLETE ITS SPECIFICATIONS AND ITS BASES The Fermi 2 ITS Specifications and ITS Bases are presented in their complete form to facilitate the review of the document in an integrated manner.
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O APPLICATION ,
OF SELECTION CRITERIA for FERMI-2 O TECHNICAL SPECIFICATIONS O
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CONTENTS
- 1. .
INTRODUCTION..............................................................................................................................1
- 2. S E LE CTlON C RITE RI A .. . .. . . . . .. . . . . .. . . . . . .. . . . . . .. . . . . . . . . . . . .. .. . .. .. .. . . ... . . . .. .. . .. . . .. . . . . .. ... . . . . . . .. . . . .. . . .
- 3. PROBABILISTIC RISK ASSESSMENT INSIGHTS .................................. ........................5
- 4. RESULTS OF APPLICATION OF SELECTION CRITERIA ............ ................................................. 8
- 5. DISCUSSION FOR OTHER SPECIFICATION / PAGE RELOCATION / DELETION......................... 8
- 6. GENERIC NO SIGNIFICANT HAZARDS EVALUATION ................................................................... 8
- 7. REFERENCES...............................................................................................................................9 ATTACHMENT ,
1
SUMMARY
DISPOSITION MATRIX FOR FERMI 2 APPENDICES A.' JUSTIFICATION FOR SPECIFICATION RELOCATION BASED ON ' SPLIT" CRITERIA B. OTHER JUSTIFICATIONS FOR SPECIFICATION / PAGE RELOCATION / DELETION C. GENERIC NO SIGNIFICANT HAZARDS EVALUATION 4
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- 1. INTRODUCTION The purpose of this document is to confirm the results of the BWR Owners Group application of the Technical Specification selection criteria on a plant specific basis for Fermi Unit 2. Detroit Edison Company (DECO) has reviewed the application of the selection criteria to each of the Fermi 2 Technical Specifications utilizing in BWROG report NEDO-31466, " Technical Specification Screening Criteria Application and Risk Assessment,* including Supplement 1 (Reference 1). Additionally, in accordance with the NRC guidance (Reference 3), this confirmation of the application of selection criteria to Fermi-2 includes confirming the risk insights from Probabilistic Risk Assessment (PRA) evaluations, provided in the Reference 1, as applicable to Fermi-2.
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Fermi- Unit 2 1 Revision 0, 04/03/98 L
I) s_-
- 2. SELECTION CRITERIA DECO used the selection criteria provided in 10 CFR 50.36 (c)(2)(ii) Technical Specification Improvements to develop the results contained in the attached matrix. The selection criteria were later incorporated into 10 CFR 50.36. Probab;;istic Risk Assessment (PRA) insights as used in the BWROG submittal were used, confirmed by Detroit Edison, and are discussed in the next section of this report. The selection criteria and discussion provided in the NRC Final Policy statement (Reference 3) are as follows:
Criterion 1: Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary:
Discussion of Criterion 1: A basic concept in the adequate protection of the public health and safety is the prevention of accidents. Instrumentation is installed to detect significant abncrmal degradation of the reactor coolant pressure boundary so as to allow operator actions to either correct the condition or to shut down the plant safe!y, thus reducing the likelihood of a loss-of-coolant accident.
{
This criterion is intended to ensure that Technical Specifications control those instruments specifically installed to detect excessive reactor coolant system leakage. This criterion should not, j however, be interpreted to include instrumentation to detect precursors to reactor coolant pressure i boundary leakage or instrumentation to identify the source of actual leakage (e.g., loose parts monitor, seismic instrumentation, valve position indicators).
i.) l Criterion 2: A process variable, design feature, or operating restriction that is an initial l condition of a Design Basis Accident or Transient analyses that either assumes the failure I of or presents a challenge to the integnty of a fission product barrier:
Discussion of Criterion 2: Another basic concept in the adequate protection of the public j health and safety is that the plant shall be operated within the bounds of the initial i conditions assumed in the existing Design Basis Accident and Transient analyses and that the plant will be operated to preclude unanalyzed transients and accidents. These analyses consist of postulated events. ana!yzed in the FSAR, for which a structure, ;
sys'am, or component must meet specified functional goals. These analyses are contained in Chapters 6 and 15 of the FSAR (or equivalent chapters) and are identified as i Condition 11, Ill, or IV events (ANSI N18 2) (or equrvalent) that either assume the failure of or present a challenge to the integnty of a fission product barrier.
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Fermi- Unit 2 2 Revision 0, 04/03/98 l
(m
() 2. (continued)
As used in Criterion 2, process variables are only those parameters for which specific values or ranges of values have been chosen as reference bounds in the Design Basis Accident or Transient Analyses and which are monitored and controlled during power operation such that process values remain within the analysis bounds. Process variables captured by Criterion 2 are not, however, limited to only those directly monitored and controlled from the control room. These could also include other features or chara:teristics that are specifically assumed in Design Basis Accident or Transient analyses if they cannot be directly observed in the control room (e.g., moderator temperature coefficient and hot channel factors).
The purpose of this crite' ion is to capture those process variables that have initial values assumed in the Design Basis Accident and Transient analyses, and which are monitored and controlled during power operation. As long as these variables are maintained within the established values, risk to the public safety is presumed to be acceptably low. This criterion also includes active design features (e.g., high pressure / low pressure system valves and interlocks) and operating restrictions (pressure / temperature limits) needed to preclude unanalyzed accidents and transients.
Criterion 3: A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product
' barrier:
Discussion of Criterion 3: A third concept in the adequate protection of the public health and safety is that in the event that a postulated Design Basis Accident or Transient should j occur, structures, systems, and components are available to function or to actuate in order l to mitigate the consequences of the Design Basis Accident or Transient. Safety sequence i analyses or their equivalent have been performed in recent years and provide a method of I presenting the plant response to an accident. These can be used to define the primary success paths.
I A safety sequence analysis is a systematic examination of the actions required to mitigate the consequences of events considered in the plant's Design Basis Accident and Transient analyses, as presented in Chapters 6 and 15 of the plant's FSAR (or equivalent chapters). Such a safety sequence analysis considers all applicable events whether explicitly or implicitly presented. The i primary success path of a safety sequence analysis consists of the combination and sequences of equipment needed to operate (including consideration of the single failure criteria), so that the plant response to Design Basis Accidents and Transients limits the consequences of these events to within the appropriate acceptance en%na
/
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-- ______-_____-_-_a
i 4
- 2. (continued)
(G) lt is the intent of this criterion to capture into Technical Specifications only those 4 structures, systems, and components that are part of the primary success path of a safety sequence analysis. Also captured by this criterion are those support and actuation i systems that are necessary for items in the primary success path to successfully function.
The primary success path for a particular mode of operation does not include backup and
. diverse equipment (e.g., rod withdrawal block which is a backup to the average power range monitor high flux trip in the startup mode.
Criterion 4: A structure, system, or component which operating experience or probabilistic ;
safety assessment has shown to be significant to public health and safety:
Discussion of Criterion 4: It is the Commission's policy that licensees retain in their Technical Specifications LCOs, action statements, and Surveillance Requirements for the following systems (as applicable), which operating experience and PSA have generally shown to be significant to public health and safety and any other structures, systems, or components that meet this criterion: 1
. Reactor Core isolation Cooling / isolation Condenser, e Residust Heat Removal, o Standby :.iquid Control, and
, -.3
( ) . Recirculation Pump Trip.
The Commission recognizes that other structures, systems, or components may meet this J
criterion. Plant- and design-specific PSAs have yielded valuable insight to unique plant vulnerabilities not fully recognized in the safety analysis report Design Basis Accident or Transient analyses. It is the intent of this criterion that those requirements that PSA or operating experience exposes as significant to public health and safety, consistent with the Commission's Safety Goal and Severe Accident Policies, be retained or included in the 4 Technical Specifications. j The Commission expects that licensees. in prepanng their Technical Specification related submittals, will utilize any plant specific PSA or nsk survey and any available literature on risk insights and PSAs. This matenal should be employed to strengthen the technical bases for those requirements that remain in Technical Specificatirns, when applicable, and to verify that none of the requirements to be relocated contain constraints of prime importance in limiting the likelihood or seventy of the accident sequences that are r:ommonly found to dominate nsk. Similarly, the NRC staff will also employ risk insights and PSAs in evaluating Technical Specifications related submittals. Further, as a part of the Commissions ongoing program of improving Technical Specifications, it will continue ,
to consider methods to make better use of risk and reliability information for defining future -!
generic Technical Specification requirements.
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Fermi Unit 2 4 Revision 0, 04/03/98
- 3. PROBABILISTIC RISK ASSESSMENT INSIGHTS Introduction and Obiectives The Final Policy Statement includes a statement that NRC expects licensees to utilize the available
. literature on risk insights to verify that none of the requirements to be relocated contain constraints of prime importance in limiting the likelihood or severity of the accident sequences that are commonly found to dominate risk.
- Those Technical Specifications proposed for relocation to other plant controlled documents will be maintained under an appropriate regulatory control mechanism, such as 10 CFR 50.59. These specifications have been compared to a variety of Probabilistic Risk Assessment (PRA) material with two purposes: 1) to identify if a component or variable is addressed by PRA, and 2) to judge if the component
- or variable is risk-important, in addition, in some cases risk was judged independent of any specific PRA rnaterial. The intent of the review was to provide a supplemental screen to the deterministic criteria.
Those Technical Specifications proposed to remain part of the Improved Technical Specifications were not reviewed. This review was documented in Reference 1 except where discussed in Appendix A,
" Justification For Specification Relocation," and has been confirmed by Detroit Edison for those Specifica-tions to be relocated. The Fermi 2 plant-specific Probabilistic Safety Assessment (PSA) was reviewed
- during this process (the Fermi-2 PSA is equivalent to a PRA) Where Reference 1 did not review a Technical Specification against the criteria of Reference 3, Detroit Edison performed a review similar to that described below for Reference 1.
O Assumotions and Acoroach Briefly, the approach used in Refe" nce 1 was the following:
The risk assessment analysis evaluated the loss of function of the system or component whose LCO was being considered for relocation and qualitatively assessed the associated effect on core damage frequency and offsite releases. The assessment was based on available literature on plant risk insights and PRAs. Table 31 lists the PRAs used for making the assessments and is provided at the end of this section. A detailed quantitative calculation of the core damage and offsite release effects was not performed However the analysis did provide an indication of the relative significance of those LCOs proposed for relocation on the likelihood or severity of the accident sequences that are commonly found to dominate plant safety risks. The following analysis steps were performed for each LCO proposed for relocation:
I '
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' Fermi Unit 2- 0 Revision 0, 04/03/98
- 3. PROBABILISTIC RISK ASSESSMENT INSIGHTS (continued)
- a. List the function (s) affected by removal of the LCO item.
- b. Determine the effect of loss of the LCO item on the function (s),
- c. Identify compensating provisions, redundancy, and backups related to the loss of the LCO I item. j
- d. Determine the relative frequency (high, medium, and low) of the loss of the runction(s) assuming the LCO item is removed from Technical Specifications and coatrolled by other procedures or programs. Use information from current PRAs and related analyses to establish the relative frequency.
- e. Determine the relative significance (high, medium, and low) of the loss of the function (s).
Use information from current PRAs and related analyses to establish the relative significance.
- f. Apply risk category criteria to establish the potential risk significance or non-significance of the LCO item. Risk categories were defined as follows:
( RISK CRITERIA Consequence Freouency tiigh Medium law High S S NS t
Medium S S NS Low NS NS NS 1
S = Potential Sgnifcant Risk Contributor l.
NS = Risk Non Signifrant I g. List any comments or caveats that apply to the above assessment.
I
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Fermi Unit 2 6 Revision 0, 04/03/98 L______..._-_____.____________________.______-__.__
TABLE 3-1 O(~~N BWR PRAs USED IN NEDO 31466 (and Supplement 1)
RISKASSESSMENT 4
. BWR/6 Standard Plant. GESSAR 11,238 Ndclear Island, BWR/6 Standard Plant Probabilistic Risk Assessment, Docket No. STN 50-447, March 1982.
. La Salle County Station. NEDO-31085, Probabilistic Safety Analysis, February 1988. !
. Grand Gulf Nuclear Station. IDCOR, Technical Report 86.2GG, Verification of IPE for Grand Gulf, March 1987.
. Limerick. Docket Nos. 50-352, 50-353,1981, 'Probabilistic Risk Assessment, Limerick l Generating Station,' Philadelphia Electric Company. !
. Shoreham. Probabilistic Risk Assessment Shoreham Nuclear Power Station, Long Island Lighting Company, sal 372-83-PA-01, June 24,1983.
. Peach Bottom 2. NUREG 75/0104, ' Reacts Safety Study,' WASH-1400, October 1975. l
. Millstone Point 1. NUREGlCR-3085, ' Interim Reliability Evaluation Program: Analysis of the Millstone Point Unit 1 Nuclear Power Plant,' January 1983.
l
. fgand Gulf. NUREG/CR 1659, ' Reactor Safety Study Methodology Applications Program:
Grand Gulf #1 BWR Power Plant.' October 1981.
. NEDC-30936P. 'BWR Owners' Group Technical Specification Impmvement Methodology (with Demonstration for BWR ECCS Actuation Instrumentation) Part 2,' June 1987.
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Fermi Unit 2 7 Revision 0, 04/03/98
i i
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- 4. RESULTS OF APPLICATION OF SELECTION CRITERIA The selection criteria from Sectior 2 were applied to the Fermi-2 Technical Speedications. The attachment is a summary of that application indicating which Specifications are being retained or 1 relocated. Discussions that document the rationale for the relocation of each Specification which !
failed to meet the selection criteria are provided in Appendix A. No Significant Hazards Considerations (10 CFR 50.92) evaluations for those Specifications relocated are provided with the Discussion of Changes for the specific Technical Specifications. DECO will relocaa those Soecifications identified as not satisfying the criteria to licensee controlled documents whose changes are govemed by an appropriate regulatory control mechanism, such as 10 CFR 50.59.
- 5. DISCUSSION FOR OTHER SPECIFICATION / PAGE RELOCATION / DELETION For Fermi 2 CTS requirements, specifications or pages being relocated or deleted for reasons other than failure to meet the Selection Criteria for retention, Appendix B provides these pages with the evaluation supporting the relocation or deletion.
- 6. GENERIC NO SIGNIFICANT HAZARDS EVALUATION Generic evaluations for a finding of No Significant Hazards Consideration (NSHC) have been provided for each category identified. These Generic NSHC determinations are located in Appendix C.
A V
Fermi- Unit 2 8 Revision 0, 04/03/98
- 7. REFERENCES 1.- NEDO-31466 (and Supplement 1), " Technical Specification Screening Criteria Application i and Risk Assessment,' November 1987.
- 2. NUREG 1433, " Standard Technical Specifications, General Electric Plants BWR/4,"
Revision 1, April 1995.
- 3. - Final Policy Statement on Technical Specifications improvements, July 22,1993 (58 FR 39132).
l O
1
, Fermi- Unit 2 , 9 Revision 0,' 04/03/98 r
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O O O O O O O e F N N N N N N F
O
. AP_PENDIX A JUSTIFICATION FOR SPECIFICATION RELOCATION BASED ON " SPLIT" CRITERIA O
O
kM O -l INSTRUMENTATION SEISMIC MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERAff0N 3.3.7.2 The seismic monitoring instrumentation sh in Table 3.3.7.2-1 shall be OPERABLE.
APPLICABILITY: At all times.
gI1Qt[:
- a. With one or more of the above requi ed sei- .c monitoring instruments inoperable for more th 30 . prepare and submit a Special Report to the Commission rsuat... . specification 6.9.2 within the next 10 days outlinin the cause. ' the malfunction and the plaris for restoring the ins ument(s) to OPERABLE status,
- b. The provisions of Specificatio s 3.0.3 are not applicable. l SURVEILLANCE REOUIREMENTS 4.3.7.2.1 Each of the above reouir seismic monitoring instruments shall be demonstrated OPERABLE by the perfo ance of the CHANNEL CHECK. CHANNEL FUNCTIONAL TEST and CHANNEL CAllB 10N operations at the frecuencies shown in Table 4.3.7.2 1.
4.3.7.2.2 Each of the above re ired seismic monitoring instruments actuated during a seismic event greater han or soual to 0.01 g shall be restored to OPERABLE status within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and a CHANNEL CALIBRATION perforined within 5 O* days following the seismic ev nt. Data shall be retrieved from actuated instruments and analyzed to termine the magnitude of the vibratory around motion. A Special Report s 11 be prepared and submitted to the Commission ]
pursuant to Specification .9.2 within 10 days describing the magnitude, frecuency spectrum, and r ultant effect upon unit features important to safety.
/
~
FERHl. UNIT 2 3/4 3 51 t.menoment No. 83 O
V PAGE / _0F y3
EMM 1ABtE 3.3.7.2 1 ,
SEISMIC MONITORING INSTRUMENTATION MINIMUM MEASUREMENT INSTRUMENTS RANGE OPERABLE INSTRUMENTS AND SENSOR LOCATIONS
- 1. Active Triaxial System 1
- a. Active Triaxial Accelerometers HPCI Room 31 1 1)
- 2) Base of RPV Pedestal, In Drywell + g I f Active Seismic Recording System- l b.
- 1) Relay Room Auxiliary Building NA 1"
- c. Active Seismic Play'back System
- 1) Relay Room, Auxiliary Butidi NA NA
- 2. Passive Triaxial Peak Shock Recorders
"* 1 3
- a. HPCI Room 4 i
- b. Relay Room, Auxiliary Buildi *" I
( Refuel Floor Reactor Build ng *" I c.
yf
- d. Diesel Generator Room. RH Complex '" l ,j
- " I
- e. Pump Room, RHR Complex
- f. Cooling Tower. RHR Cor lex 1
- Incluo wg seismic trigger.
"With reactor control roo annunciation.
"*Each passive accelerome er nas 12 reeos, eacn monitoring a different frequency. The frecue ctes correspono to varying accelerations. The widest range is g 90 .
FERMI UNIT 2 3 4 3 52
. ;q PAGE 2. _0F o_ _
$ELOCA160 D
TABLE 4.3.7.2 1 I
stf EMIC MONITORING INSTPUMENTATION EURV LLANCE REDUfREMENTS CHANNEL C EL FUNCTIONAL CHANNEL INSTRUMENTS AND SENSOR LOCATIONS ECK TEET CALIBRATION
- 1. Active Triaxial System
- a. Active Triaxial Accelerometers HPct Room NA SA R 1)
- b. Active Seismic Recording 5)ste2'
- 1) Relay Room. Auxilia y Building" M(a) SA R
- c. Active Seismic Playba System
- 1) Relay Room. Aux inry tuilding M SA R O 2. Passive Triaxial Peak 5 ock Recorders l l
R I
- a. HPCI Room NA NA l
- b. Relay Room. Au iliary Building NA NA R
- c. Refuel Floor Reactor Building NA NA R j
- d. Diesel Ge rator Room, RHR Complex NA NA R
l 4
" Including sismic trigger.
- r control room annunciation.
(a)Withrest Except s smic trigger.
FERM1 - UNIT 2 3/4 3 53 PAGE 3 -
0F 4 ;
kt1D (d O INSTRUMENTATION o f I
METEnone nCICAL MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION !
3.3.7.3 The meteorological monitorin instrumentation channels shown in Table 3.3.7.3-1 shs11 be OPERABLE.
APPLICABILITY: At all times.
E.Ijati:
- a. With less than the re trad channels OPERABLE in Table 3.3.7.31 for more than 7 days, pr are and submit a Special Report to the Ceaunission pursuant o Specification 6.9.2 within the next 10 days outlining the cause f the malfunction and the plans for restoring the instrumentatic to OPERABLE status.
- b. The provisions of Specifications 3.0.3 are not applicable. !
SURVEfLLANCE RE0tifREME S 4.3.7.3 Each of the bove reoutreo meteorological monitoring instrumentation channels shall be d nstrated OPERABLE by the performance of the CHANNEL CHECK and CHANNEL C IBRATION operations at the frequencies shown in Table 4.3.7.3-1.
O l
F6RMI UNIT 2 3/4 3 54 Amenoment No. 83 n
U~
t' PAGE 4' Or43
b TABLE 3.3.7.3 1 (
METEOROLOGICAL MONITORING INSTRUMENT MINIMUM CHANNELS INSTRUMENT OPERABLE
- a. Wind Spted
- 1. Elev. 10 meters 1
- 2. Elev. 60 meters 1
- b. Wind Direction
- 1. Elev. 10 meters 1
- 2. Elev. 60 meters I
- c. Air Temperature Difference
- 1. Elev. 10/60 meters 1 O
l i
'I FERM1 UNIT 2 3/4 3 55 i
G PAGE 5- 0F a3
E O
TABLE 4.3.7.3-1 1 g'
METFORDt0GICAL MONITORING INSTRtMENTATION RVEILLANCE REO 5 CHANNEL CHANNEL CHECK
- a. Wind Speed
- 1. Elev. 10 meters D SA
- 2. Elev. 60 meters D SA
- b. Wind Direction
- 1. Elev. 10 meters D SA .
- 2. Elev. 60 meters D SA
- c. Air Temperature Differene
- 1. Elev. 10/60 meters O SA I
1 FERMI UNIT 2 3/4 3-56
.(O v ]
PAGE 6 _0F 43
~ 62SfMM INSTRUMENTATION i i TRAVER$1NG IN t0DF PROBE EYETEM LIMITING CONDITION FOR 09tDaT10N 3.3.7.7 The traversing in core probe system sh I be OPERABLE with:
- a. Five movable detectors, drives and adout equipment to map the core, and
- b. Indexing soutoment to allow all f ve detecters to be calibrated in a '
coanon location.
APPlitAElt1TY: When the traversing in c re probe is used for:
gJ,13:
l With the traversing in-core probe ystem inoperable. suspend use of the system for the above apolicable monitor 9 or calibration functions. The provisions of Specification 3.0.3 are not appi cable.
f SUovttitautt eteuteturwTs j
4.3.7.7 The traversing in creeprobe system shall be demonstrated OPERABLE by recuired detector outputs within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to i
normalizing esen of the ab use wtion required for the PF>t calibration function.
l I
I l
- 0nly the cettttor(s) in the required measurement location (s) are required to be OPERABLE.
Amendment No. 83 f[RMI UNIT 2 3/4 3 65 O
PAGE 7 op gy
0$
o -
b Z DUMENTATION CHt0RINE DETECTION SYSTEM LIMITING CONDITION FOR OPERATION 3.3.7.8 Two independent chlorine detecto s shall be OPERABLE with their trip setpoints adjusted to actuate at chlorin concentration of less than or equal to 5 ppe.
APPLICABILITY: All OPERATIONAL CONDIT N5. .
ACTION:
- a. With one chlorine detecto inoperable, restore the inoperable detector to OPERABLE sta us within 7 days or within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, initiate and mai sinisolationofallcontrolroomemergency intakes by placing the VAC system in the chlorine mode of operation,
- b. With both chlortne a ectors inocerable, within I hour initiate and maintain isolation all controi room emergency intakes by placing the HVAC system in ne chlorine mooe of operation.
I SURVEftLANCE REOUIREMFNT 4.3.7.B Each of the ab e recutred chlorine detectors shall be demonstrated OPERABLE by performanc of a:
- a. CHANNEL FUN 10NAL TEST at least once per 31 days, and
- b. CHANNEL CA IBRATION at least once oer 18 months.
l FERMI UNIT 2 3/4 3 66 Amenoment ho. 83 y(
PAGE 5? OF
MM
/N .
u- )
(
INSTRUMENTATION toosE-PART DETECTION SYSTEM i llMITING CONOTTION FOR OPERATION k
3.3.7.10 The loose part detection system sh- I be OPERABLE.
APPLICABILITY: OPERATIONAL CONDITIONS 1 an 2.
gI1QH:
- a. idith one or more loose part d ection system channels inoperable for more than 30 days, prepare a submit a Special Report to the Cossetssion pursuant to Speci ication 6.9.2 within the next 10 days outlining the cause of the .alfunction and the plans for restoring the channel (s) to OPERABLE status.
- b. The provisions of Specif cations 3.0.3 are not applicable. l SURVElllANCE RE001REMENTS 4.3.7.10 Each channel of th loose-part detection system shall be demonstrated 00ERABLE by per ormance of a:
- a. CHANNEL CHECK at east once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, CHANNEL FUNCT! 4AL TEST at least once per 31 days, and
/O b.
V c. CHANNEL CAllB T10N at least once per 18 months.
i FERMI - UNIT 2 3/4 3 7C Amenoment No. 83 O
L.) '
PAGE 9 op g
f GLocATED D
J .
INSTRUMENTATION EXPLOSIVE CAS MONITORING INSTRUMENTATIO.
j LIMITING CONDITION FOR OPERATION !
3.3.7,12 The explosive cas monitori g instrumentation channel shown in Table 3.3.7.12 1 shall be OPERABLE with i s alarm etpoint set to ensure that the limits of Specification 3.11.2.6 a not exceeded. ;
APPLICABILITY: As shown in Table 3.3.7.12-1 E1121:
- a. With an explosive ga monitoring inetrumenation channel alarm setpoint less conse ative than root.*ed by the noove Specification, declare the channe inoperable and t 2e ACTION shown in Table 3.3.7.12 1; or ch ge the setpotnt so it is acceptably conservative.
- b. With less than ti. minimum numoer of explosive cas monitoring instrumentation nannels OPERABLE take the ACTION shown in Table 3.3.7.12-1. Re tore the inocerable instrumentation to OPERABLE status within days and. if unsuccessful, prepare and submit a special repor to the Commission pursuant to Spectftcation 6.9.2 to explain why is inoperability was not corrected in a timely manner.
- c. The provisi s of Specifications 3.0.3 are not applicable. [ j SURVE1((ANCE RE001P ENTS instrumentation channel shall be 4.3.7.12 Each exo osive LEgas monitorin! theoCHANNEL CHECK. CHANNEL CALIBRATION by performance C demonstrated OPE and CHANNEL FUNCT ONAL TEST operations at the frequenctes snown in Table 4.3.7.12 1.
3/4 3 76 Amenoment ',o. 24, 22, 83 FERMI UNIT 2 O
PAGE / 0 _0F 43 L
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FERMI UNIT 2 3/4 3 78 Amenorr.ent ho. 82 O
PAGE / 2. OF 43
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TABLE 3.3.7.12 1 (Cont'inued)
~
TABLEh0 TAT 10!S j e l
- During main condenser offgas treatmen system operation.
ACTI N STATEMENTS ACTION 124 - With t numoer of channels OPERABLE less than required by the I i
Minimu Channels OPERABLE recuirement, operation of main conde ser offgas treatment system may continue provided grab samol s are collected at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and analyzed
' with n the following 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Otherwise, suspeno release of
( rad onctive effluents via this pathway.
.i l
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FERMI UNIT 2 3/4 3 60 Amenoment No. 2/, 73. 82 s
PAGE /3 _OF 43
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i FERMI - UNIT 2 3/4 3 E2 Amenoment No. 82 1
O I,
PAGE /5 _OF 43 l
EZ.dCATPp O
TABLE 4.3.7.12-1 (Conti ed)
TABLE NOTATION i
- During main condenser offgas treatmen system operat' n.
l l
(3) The CHANNEL CALI T10N shall include the use of standard gas samples containing a no nal:
I i i
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FERMI UNIT 2 3/4 3 84 Amenoment No. 22 n
U 1
l
/6 PAGE 0F a3
Rnoorreo O
INSTRUMENTATION 3/4.3.11 APPENDIX R ALTERNATIVE SHUTDOWN INSTR NTATION LIMITING CONDITION FOR OPERATION !
3.3.11 The alternative shutdown system nstrumentation channels shown in Table 3.3.11 1 shall be OPERABLE.
APPLICABILITY: OPERATIONAL CONDITION 1, 2, and 3.
g1LQH:
With the number of OPERABLE alter tive shutdown monitoring instrumentation channels less than reoutred, tak the ACTION required by Table 3.3.11 1. The provisions of Specification 3.0 are not applicable.
$URVEILL ANCE DEOUIREMiNTS 4.3.11.1 Each alternati e snutdown instrumentation channel shall be demonstrated OPERABLE performance of the CHANNEL CHECK and CHANNEL CAllBRATION operation at the frecuencies snown in Table 4.3.11.1-1.
o R.7 1 I
i l
FERMI UNIT 2 3?4 3 90 Amenoment No. 59 O
PAGE /7 0F 43
RohT TABLE 3.3.11-1 APPENDIX R ALTERNATIVE SHUTDOWN NSTRUMENTATION INIMUM CHANNELS INSTRUMENT OPERABLE ACTION CTG 11 Unit 1-Volts 1 131 1.
131
- 2. CTG 11 Unit 1-Frecuency 1 131
- 3. CTG 11 Unit 1-Watts 1 131 4 CTG 11 Unit 1-Vars 1 132
- 5. Condensate Storage Tank Level 1 132
- 6. Standby Feedwater Flow 1
- 7. Reactor Water Level 1 133 Reactor Pressure 1 133
- 8. 133
- 9. Torus Water Temperature 1 133
- 10. Torus Water Level 1 133
- 11. Primary Containment Temper ture 1 ACTION STATEMENTS ACTION 131 - Declare C 11 Unit 1 inoperable and take the ACTION required by Specif cation 3.7.11.
ACTION 132 - Declare the SBFW system inoperable and tske the ACTION required by Spe iftention 3.7.11.
ACTION 133 - Rest e the inoperable channel to OPERABLE status within 7 days
-( or in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in C0 SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, l
l FERMI UNIT 2 3/4 3 91 Amenoment ho. 59 O
PAGE /8 0F 43 .
DO&
O TABLE 4.3.11.1-1 )
APPENDIX R ALTERNATIVE SHUTDOWN INS UMENTATION SURVEILLANCE RE001REME S l
ANNEL CHANNEL I i
CHECK CALIBRATION INSTRUMENT M R
- l. CTG 11 Unit 1-Volts
- 2. CTG 11 Unit 1 Frequency M R M R
- 3. CTG 11 Unit 1 Watts R
- 4. CTG 11 Unit 1-Vars M Condensate Storage Tank Level M R
- 5. R
- 6. Standby Feoowater flow M M R
- 7. Reactor Water Level R
- 8. Reactor Pressure M
- 9. Torus Water Temperature M R M R
- 10. Torus Water Level R
- 11. Primary Contatnment Temperat re M 3/4 3-92 Amendment No. 59 FERM! - UNIT 2 PAGE /9 _0F 43
bc CtLIC
( REACTOR COOLANT SYSTEM k.O b(/ 3/4.4.4 CHEM: STRY LIMITING COND TION FOR OPERATION ,
3.4.4 The chemistry of the reactor coolant system shall be maintai d within the limits specified in Table 3.4.4 1.
APPLICABILITY: At all times.
E11Dft:
- a. In OPERATIONAL CONDITION 1:
- 1. With the conductivity, chloride concentr ton or pH exceeding the limit specified in Table 3.4.4-1 for i than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during one continuous time interval and, for onductivity and chloride concentration, for less than 336 ho s per year, but with the conductivity less than 10 pmho/cm 25'C and with the chloride concentration less than 0.5 ppm, his need not be reported to the Commission. l
- 2. With the conductivity, chio de concentration or pH exceeding the limit specified in Table 3 .4-1 for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during one continuous time inte al or with the conductivity and chloride concentration xceeding the limit specified in Table 3.4.4 1 for more than 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> per year, be in at least STARTUP within the next 6 h rs.
- 3. With the conduct 1 ity exceeding 10 pmho/cm at 25'C or chloride concentration e eeding 0.5 ppm. be in at least HOT SHUTDOWN within 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,
- b. In OPERATIONAL C ITIONS 2 and 3 with the conductivity, chloride ph concentration o pH exceeding the limit specified in Table 3.4.41 for more than 48 h urs during one continuous time interval, be in at least HOT SHUTDOWN ithin the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ano in COLD SHUTDOWN within the-following 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
- c. At all o er times:
- 1. ith the:
a) Conductivity or pH exceeding the limit specified in Table 3.4.41. restore the conductivity and pH to within the limit within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or b) Chloride concentration exceeding the limit specified in Table 3.4.41. restore the chloride concentration to g
within the limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or i perform an engineering evaluation to determine the effects of the out-of limit condition on the structural integrity of the reactor l
) coolant system. Determine that the structural integrity of the i reactor coolant system remains acceptable for continued operation prior to proceeding to OPERATIONAL CONDITION 3.
- 2. The provisions of Specification 3.0.3 are not applicable. /
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'( [REACTORCOOLANTSYSTEM [ *b SURVEILLANCE RE0VIREMENTS 4.4.4 The reactor coolant shall be detennined to be within t specified chemistry limit by;
- a. Measurement prior to pressurizing the reactor ring each startup, if not performed within the previous 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> .
- b. Analyztag a sample of the reactor coolan or: ,
l
- 1. Chlorides at least once per: t I
a) 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and b) 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> whenever conductivity is greater than the limit in Tabi .4.4-1.
- 2. Conductivity at lea once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
- 3. pH at least once er:
a) 72 urs, and b) hours whenever conductivity is greater than the limit in Table 3.4.4-1.
- c. Continuous recording the conductivity of the reactor coolant, or, O when the entinuous recording conductivity monitor is inoperable by obtain g an in-line conductivity measurement at least once per:
J f
- 1. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> in OPERATIONAL CONDITIONS 1, 2, and 3, and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at all other times,
- d. Performance of a CHANNEL CHECK of the continuous conductivity
{ monitor with an in-line flow cell at least once per:
- 1. 7 days, and
\ 2. 2A .our; whenever conductivity is greater than the limit in Esble 3.4.4-1.
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O CTOR C00LANT SYSTEM 3/4.4.8 STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.8 The structural integrity of ASME Code Class , 2, and 3 components shall be maintained in accordance with Specifica n 4.4.8.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2,' , 4, and 5.
EIIAti:
- a. With the structural integr y of any ASME Code Class 1 component (s) not conforming to the ab e requirements, restore the structural integrity of the affec d component (s) to within its limit or isolate the affected emponent(s) prior to increasing the reacter coolant system tem rature more than 50*F above the minimum temperature reau ed by NDT considerations.
- b. With the stru ural integrity of any ASME Code Class 2 component (s) not conform' g to the above requirements, restore the structural integrity , the affected component (s) to within its limit or-isolate e affected component (s) prior to increasing the reactor coolan system temperature above 200*F.
O c. Wit the structural integrity of any ASME Code Class 3 coinponent(s) no conforming to the above requirements, restore the structural I
j i egrity of the affected component (s) to within its limit or I solate the affected component (s).from service, t !
i S VEILLANCE RE00fREMENTS
.4.8 No requirements other than Specification 4.0.5.
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[CONTAlletENTSYSTEMS SUPPRFESION POOL AND DRYWELL SPRAY LIMITING CONDITION FOR OPERATION /
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3.6.2.2 The suppression pool ud drywell spray mode of he residual heat removal (RHR) system shall be 07ZRABLE with two indep dent loops, each loop k j I consisting of:
- b. An OPERABLE flow path capable of circulating water from the i suppression chamber through an heat exchanger and the "
l' suppression pool and drywell s ay spargers.
APPLICABILITY: OPERATIONAL CONDITIONS , 2, and 3.
ACTION:
With one suppression prol and/or drywell spray loop inoperable. !
- a. j restore the inoperabl4 loop to OPERABLE status within 7 days or be .
in at least HOT SH 00WN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLO j SHUTDOWN within t following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. l j b. With both suppr sion pool and/or drywell spray loops inoperable, j
' restore at leas't one loop to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD
( I SHUTDOWN
- within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE REOUlpEMI S ~
/
j 4.6.2.2 The suppres/sion pool and drywell foray mode of the RHR system shall l
t be demonstrated 0 RABLE:
- a. At least once per 31 days by verifying that each valve (manual, I power operated, or autocatic) in the flow path that is not locked,
,/ sealed, or otherwise secured in position is in its correct j
, position.
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- b. ' By verifying that each of the reautred RHR pumps develops a flow l
' ' of at least 500 gom on rectrculatten flow through the RHR heat I exchanger and suppression pool spray sparger when tested pursuant i to Specification 4.0.5. ,
1
- c. By performance of an air or smote flow test of the drywell spray no2Zies at least once per 5 years and verifying that each spray nozzle is unobstructed.
- Whenever both RHR subsystems are inoperable, if unable to attain COLD
( SHUTDOWN as required by this ACTION, maintain reactor coolant temperature N as low as practical by use of alternate heat removal methods.
FERH1 - UNIT 2 3/4 6 18 PAGE M' op 43
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PLANT SYSTEMS SURVEfLLANCE RE0VIREMENTS (Continued)
- e. At least once per 18 months by:
- 1. Verifying that the pressure drop across the recirculation train and across the makeup train combined HEPA filters and charcoal Sll adsorber banks is less than 8 inches and 6 inches water gauge gMh respectively while operating the system at a flow rate of 3000 cfm a 10% through the recirculation filter train and 1800 cfm 55 7 10% through the makeup fiiter train.
- 2. j%fying that the system will automatically switch to the S CL Irecirculationmodeofoperationoneachofthebelowactuation test signals and verifying that on any one of the below
[ gat > recirculation mode actuation test signals, the system automatically switches to the recirculation mode of operation, hO'b the isolation valves close within 5 seconds and the control room is maintained at a positive pressure of at least 0.125 inch water gauge relative to the outside atmosphere during system operation at a flow rate less than or equal to 1800 cfm h h the emergency makeup air filter:
SfA '
Sydh4Nm I a) control center inlet radiation montter.
Fual pool ventilation exhaust radiation monitor.
3'b-} ' g b)
O c) Low reactor water level. , h 'j D' d) High drywell pressure. ,
, Verifyi that on the hlorine m e tuat 'o signal, syste utomatically witches t e chio e mode of oper ion, the iso tion valve osc wie n 4 seco , and a mimmum vi 1:00 :. 9="y aci rcul a t' on h a m 4%
I4. Verifying that each of the emergency makeup inlet air heaters dissipate 12.0 a 2.0 kW when tested in accordance with ANSI N510 1980.
- f. After each complete or partial replacement of a train HEPA filter bank by verifying that the train HEPA filter bank satisfies the Su inplace penetration-and bypass ieakage testing acceptance criteria of less than 1.0% in accordance with ANSI N5101980 while operating pcdk4hM the system at a flow rate of 1800 cfm a 10% for the makeup train and gq 3000 cfm 10% for the recirculation train,
- s. After each complete or partial replacement of a train charcoal adsorber bank by verifying that the train charcoal adsorber bank satisfies the inplace penetratton and bypass leakage testing acceptance criteria of less than 1.0% in accordance with ANSI N510 1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 1800 cfm a 10% for the makeup train and 3000 cfm a 10% for the recirculation train. ,
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) FERMI - UNIT 2 3/4 7 10 Amendment No. 7, 81 PAGE 25 0F 43
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PLAN SYS*' EMS 3/4. J.3 SHORE BARRIER PROTECTION LIMITING CONDITION FOR OPERATION 3.7.3 The shore barrier shall be structurally sound and cap e of limiting wave action as intended. The shore barrier shall be maint ed such that the elevation of each survey point listed in Table 3.7.3-1 i ot less than 1.0 foot below the elevation listed in the table.
APPLICABILITY: At all times.
EllM:
With the elevation of one or more survey poi s different by more than 1 foot from the elevation determined during the 1 4 survey (see Table 3.7.3-1),
prepare and submit to the Tosaission wit n 90 days, pursuant to Specification 6.9.2, a Special Report which includes e following information:
- a. Explanation of how the de dation occurred and if the shore barrier is continuing to degrade *
- b. A planned course to r air the damage and a schedule for accomplishing the re ir; f
- c. Evaluation of and ustification for continued plant operation; and
- d. The current el ation of each survey point shown in Table 3.7.3 1. f SURVEf ttANCE REOUlpEME 5 4.7.3 The shore rrier shall be determined to be structurally sound and capable of limit g wave action by visual inspection and instrument survey:
- a. At ast once per.12 months.
- b. thin 7 days after a severe storm in which the crest elevation of ncident waves at the snore line exceeds the top of the shore barrier (583'0').
l
. Within 7 days after any eartnouake event with intensity greater than an operating basis eartnouane (OBE).
FERMI UHlf 2 3/4 7 11 r
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PAGE 2.6 0F q
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TABLE 3.7.3-1 SURVEY POINTS FOR SHORE BARRIER
- SURVEY LOCATION ** D BER 1984 K NORTH. SOUTH EAST WEST C0 ROL ELEVATION 1A N6807 ES945 580.05 18 N6803 E5957 576.99 1C N6803 E5972 575.10 2A N6824 E5947 581.63 28 N6825 E59 581.01 2C N6826 ES B 579.02 20 N6822 976 577.65 1
3A N6901 E5944 581.52 38 N6898 E5958 579.89 3C N6905 E5972 577.08 4A N7020 E5949 580.92 48 N7023 E5960 580.59 4C N702 E5967 578.58 40 N70 E5974 576.02 SA 17119 E5947 582.09 58 N7122 .E5957 581.45 N7120 E5964 578.72 SC 50 N7121 ES974 575.52 l
6A N7222 ES931 582.55 68 N7223 E5950 582.70 6C .N7215 E5958 581.22 60 N7228 ES966 F.'d.59 6E N7233 E5973 575.59 7A N7328 E5946 582.22 )
78 N7322 E5958 581.18 7C N7317 E5966 578.99 7D N7328 ES974 575.09 SA N7422 ES950' 582.16 1 B N7418 E5957 581.40 )
SC N7429 E5963 578.12 BD N7428 E5974 576.53 m
. FERMI UNIT 2 3/4 7 12
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Relecaded O TARlf 1.7.31 (Continued) [//
SURVEY P0fMTS FOR SHORE RARRIER* ,
SURVEY LOCATION ** . DECEMBER 1984 NORTH SOUTH EAST-WEST CONTROL ILEVATIpN P.QM[1 9A M7529 E5948 583.04 98 N7531 E59 582.10 SC N7531 E 65 579.91 9D N7526 5973 575.13 10A N7612 E5937 583.85 108 N7610 E5950 582.21 4 582.56 j 10C N7618 E5961 '
100 N7616 E5972 576.58 11A N77 E5940 583.15 118 N 1 E5956 582.08 11C 718 E5963 579.82 110 N7722 E5971 576.43 12A M7814 E5949 581.86 128 N7809 E5955 581.11 E5965 578.88 0
1 12C N7814 12D N7815 E5975 577.81 i
- Measuring reference points are anchored into the capstones using ,
center notched self drilling bolts. l "See Figure B 3/4.7.31 for location sketch.
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FERN! - UNIT 2 3/4 7-13 f O
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PLANT SYSTEMS 3/4.7.6 SEALED SOURCE CONTAMINATION f LIMITlHG CONDITION FOR OPERATION 3.7.6 Each sealed source containing radioactive mate al either in excess of
{ 100 microcuries of beta and/or gamma emitting mater 1 or 5 microcuries of alpha emitting material shall be free of greater an or equal to 0.005 microcurie of removable contamination.
APPLICABILITY: At all times.
l ACTION:
- a. With a sealed source having emovable contamination in excess of l the above limit, withdraw he sealed source from use and either: !
- 1. Decontaminate an repair the sealed source, or
- 2. Dispose of the ealed source in accordance with Commission Regulations.
- b. The provisions o Specification 3.0.3 are not applicable, j p SURVEILLANCE RE00fREMENT V
4.7.6.) Test Recuire nts Each sealed source shall be tested for leakage and/or contamtnation y:
l
- a. The 1 ensee. or
- b. Oth persons specifically author 12ed by the Commission or an Ag ement State.
The test met d shall have a detection sensitivity of at least 0.005 ;
microcurie er test sample. l 4.7.6.2 est Frecuencies lacn category of sealed sources, excluding startup ,
sources nd fission detectors previously subjected to core flux, shall be I' tested t the frequency describeo Delow.
- a. Sources in use At least once per 6 months for all sealed sources containing radioactive eatertal:
- 1. With a half life greater than 30 days, excluding Hydrogen 3, and !
- 2. In any form other inan gas.
FERMI UNIT 2 3/4 7-22 Amendment No. 83 I
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[.12 PLANT SYSTEMS SURVEfttANCE REOUTREMENTS (Continued) -
- b. Stored sources not in use - E s aled source and fission detector shall be tested p r to use or transfer to another licensee unless tested in the previous 6 months. Sealed sources and fission ectors transferred without a certificate indicating the las est date shall be tested prior to being placed into use
- c. Startuo so es and fission detectors - Each sealed startup source and fiss n detector shall be tested within 31 days prior to being subje d to core flux or installed in the core and following r r or maintenance to the source.
4.7.6.3 corts - A report shall be prepared and submitted to the Commission on an ual basis if sealed source or fission detector leakage tests reveal the esence of greater than or eat.al to 0.005 microcurie of removable e amination.
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I FERM1 - UNIT 2 3/4 7 23 O
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se PLANT SYSTEMS 3/4.7.11 APPENDfX R ALTERNATIVE SHUTDOWN AUXILI ARY SYSTEMS LIMITING CONDITION FORE 0PERATION 3.7.11 The Appendix R Alternative Shutdown auxiliary stems shall be OPERABLE as described below:
- a. A Standby Feedwater (SBFW) System sisting of two OPERABLE SBFW pumps and an OPERABLE flow p h from the condensate storage tank to the reactor ves .
- b. An OPERABLE CTG 11 Unit I an power train capable of supplying power to the Pea r Bus.
- c. Two 0F ILE Drywell Co ing Units (Units 1 and 2) i consist...g of a fan an cooling coil capable of being supplied with coolin water from EECW system.
[
APPLICABILITY: OPERATIONAL COND IONS 1, 2, and 3
- a. For the S W system:
- 1. W h one SBFW inoperable, within 7 days and at least ce per 31 cays thereafter, perform Surveillance eautrement 4.7.11.1.b. using the OPERABLE SBFW pump.
The OPERABLE SBFW oump need not be tested more i
frecuently than once per 31 days. Otherwise, be in at i least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- 2. With the SBFW system otherwise inoperable, restore the system to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
With an inoperable CiG 11 Unit 1:
{ .
( 1. Verify within I hour, the#120 KV bus is available by performing Surveillance Requirement 4.8.1.1.1, and
- 2. Within 7 days, restore the CTG 11 Unit I to OPERABLE status or provice an alternative source of power to the Alternat.ve Shutoown bus.
. Otherwise, be in ",t least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
\ and in COLD SHU 00WN witnin the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
FERMI UNIT 2 3/4 7 41 Amenoment No. 59 r
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PLANT SYSTEMS 3/4.7.11 APPENDIX R ALTERNATIVE SHUTDOWN AUX 1tIARY SYSTEMS LIMITING CONDITION FOR:0PERATION (Continued) /
- c. For the Drywell Cooling System:
- 1. Ilthonedrywellcoo g unit inoperable, within 7 l
days and at least e per 31 days thereafter, perfore '
Surveillance Re rement 4.7.11.3 using the OPERABLE ' i Drywell Coolin nit. The OPERABLE Drywell Cooling Unit need n be tested more frequently than once per 31 days. herwise, be in at least HOT SHUTDOWN within e next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within ;
the f owing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. g
]
- 2. W n both drywell cooling units it. operable, restore at I f
east one drywell cooling unit to OPERABLE status
.f within.7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- d. With one or more reautred alternative shutdown system control circuits inoperable, restore the inoperable circuits to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN
( within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- e. The provisions of Specif t. cation 3.0.4 are not applicable.
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FERMI - UNIT 2 3/4 7-42 Amenoment No. 59 l
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PAGE 32 _0F 43
d bCd/ded
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[/3 g SYSTEMS SURVEILLANCE RE0UIREMENTS /
4.7.11.1 The SBFW system shall be demonstrated OPERAB .
- a. At least once every 31 days by:
- 1. Verifying by venting at the gh point vents that the system piping from the pump disc ge to the system isolation valves is filled with er, and
- 2. Verifying that en valve (manual, power operated or automatic) in t flow path that is not locked, sealed, or otherwise sec ed in position, is in its correct position.
- b. At least once e ry 92 days on a STAGGERED TEST BASIS by verifyi that each SBF pump develops a flow of a 600 gpm in a test flow path with system nead corresponding to the reactor vessel operatin ressure including injection line losses.
4.7.11.2 CTG 1) Unit I shall be demonstrated OPERABLE by starting and supplying load at least 10 MW to the Peaker Bus at least once every 31 days.
4.7.11.3 Drywell Gooling Units 1 and 2 shall be verified OPERABLE at least p once eve /y 92 days on a STAGGERED TEST BASIS by operating the unit for 72
. hours th the fan in "HIGH" speed.
4.7. 1.4 Each reouireo alternative shutdown system control circuit shall be demonstrated OPERABLE by verifying its capability to perform its intended function (s)atleastonceper18 months. ;
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FERMI UNIT 2 3/4 7 43 Amenoment to. 59
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PAGE 33 0F 43
Raoenen ELECTRICAL POWER SYSTEg x
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Q 3/4.8.4 ELECTRICAL EQUIPME PROTECTIVE DEVICES ,
A.C. CIRCUITS IN5fDE MARY CONTAINMEHI ,/4 LIMITING CONO OR ODERATION 3.8.4 At least the following A.C. circuits ins e primary containment shall )
be energized *:
- a. Circuit Number 6 in p 1 728 20
- b. Circuit Numbers 1. 2 , 4, 5, 15, 16, 17, 18 in pan IR i
APPLICARit1TY: OPERATIONAL CONDIT NS 1, 2, and 3.
EI1QH:
With ary of the above re rec circuits energized, open e associated circuit breaker (s) in the spec led panel (s) within I hour.
SURVEILLANCE RE REMENTS 4.8.4.1 ach of the above reautred A. . circuits shall L: determined to b deener red at least once per 24 he ** by verifying tha
- associate cir t breaket s are in the off p tion.
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{***Exceptd g entry into the crywell.
t (xcep least once per 31 cays if locke sealed, or otherwise secured in (the pped condttton.
FERMI UNIT 2 3/4 8 16 Amencment No. 48 ,
O PAGE M 0F 43
0 bfCTRICALPOWERSYSTEMS, d PRIMARY CONTAINMENT P TRAT10NCONDUCTOROVERCURRdTPROTECTIVEDEVICES LIMITING CONDITIO OPERATION !
3.8.4.2- All imary containment penetratio conductor overcurrent protective h/S i
devices s in Table 3.8.4.2-1 shall be ERABLE.
APPLICA ITY: OPERATIONAL CONDIT!DN , 2, and 3.
With one or more of the imary containment penetration con ctor j overcurrent protective vices shown in Table 3.8.4.2-1 i perable, declare the affected stem or component inoperable and ply the appropriate ACTION atoment for the affected system, d
- 1. For 4.16- circuits, deenergize the 4.16 kV rcuit(s) by tripping the ass lated circuit breaker (s) within 72 ours and verify the circui breaker to be tripped at least on per 7 days thereafter.
l
- 2. Fo 480 volt circuit devices, remove t inoperable device (s) from rvice by racking out or removing t device within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and !
verify the inoperable device (s) to racked out or removed at least once per 7 days thereafter.
therwise, be in at least HOT SHUTD within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the followin 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
l O SURVEILLANCE RE001REMENTS 4.8.4.2 Each of the or ary containment penetration conductor ov current protective devices sh n in Table 3.8.4.2 1 shall be demonstrat OPERABLE:
- a. At le once per 18 months:
1 By verifying that the 4.16 kV circuit are OPERABLE by performing.
a) A CHANNEL Call 8 RAT 10N of e associated protective !
i relays. and l
l b) An integrated system ubctional test which includes simulated automatt actuation of the system and verifying that e relay and associated circuit breakers and o current control circuits function as designed.
1 FERH1 UNIT 2 3/4 8 17 Amendment No. 83 -
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ELECTRICAL POWER SYSTEMS ,
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()i SURVEfttANCE REOUfBERENTS (Continued)
By functionally testing the 48 k.b olt circuit breakers.
Testing of these circuit br ers shall consist of injecting a current in excess of 12 f the breakers nominal setpoint and measuring the respons ime. The measured response time will be compared to the nufacturer's data to insure that it is less than or equ to a value specified by the manufacturer Circuit break found inoperable during functional tes g shall be re ored to OPERABLE status prior to resum operatto f the affected equipment,
- b. At least o e per 60 months by subjecting each ci it breaker to an inspecti and preventive maintenance in accor ce with procedures
{ prepar in conjunction with its manufacture s reconsnendations.
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FERM1 UNIT 2 3/4 8 18 -
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l ICAL POWER SYSTEMS MOTOR-0PERATED VALVES HERMAL OVERLOAD PR07ECTI0tV LIMITING CONDITIO R OPERATION /
3.8.4.3 Th hermal overload protect 1 of each valve used in safet systems j
. shall be 0 RABLE.
APPLIC ILITY: Whenever the moto -operated valve is required t be OPERABLE.
With the thermal overloa protection for one or more of e above required valves inoperable, cont' uously bypass the inoperable ermal overload within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or declare th affected valve (s) inocerable d apply the appropriate ACTION statement (s) or the affected system (s).
SURVEILLANCE OUTREMENTS 4.8.4.3 he thermal overload protection r the above recuired valve shall be de strated OPERABLE at least once r 18 months and following intenance on t motor starter by the performanc of a CHANNEL CAllBRATION o a ;
rep sentative sample of at least 25% f all thermal overloads f the above >
required valves. l 1
i l
l l
i FERMI UNIT 2 3/4 8 20 Amendment No. 102 -
D b
PAGE SF
- OF 43
RELOCAisD
?
POWER SYSTEMS STANDBY L10UID CONTROL SY 1ATED ISOLATION DEVICES LIMITING CONDITION ERATION J7 3.8.4.5 All ircuit breakers shown in Table 3. .5-1 shall be OPERABLE.
APPLICAB ITY: When standby liquid control ystem (SLCS) is required to e OPERA 8 .
ith one or more of the circuit akers shown in Table 3.8.4 1 inoperable either:
- a. Restore the in erable circuit breaker (s) t OPERABLE status within 8 hou , or
- b. Trip the noperable circuit breaker (s , rack out or remove the device rom service within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> d verify the circuit brea r(s) to be racked out or re ved from service at least once pe days thereafter, and deci e the affected SLCS component operable and apply the appr riate ACTION as required by pecification 3.1.5*.
SURVE LANCE RE0VIREMENTS
.8.4.5 Each of the above r cuired circuit breaker (s) shall demonstrated PERABLE:
- a. At least e per 18 months by performing CHANNEL CALIBRATION of the asso ated protective relays and a NNEL FUNCTIONAL TEST of each b aker which includes simulatio of actuation of the system and rifying that each relay and ociated circuit breaker and ov current control circuits fun ons as designed,
- b. t least once per 60 months suDjecting each circuit breaker to an inspection and preventi matntenance in accordance with procedures prepared in c
- unction with its manufacturer's reconenendations.
1he requirement to apply e approortate ACTION as recuired by pecification
}*3.1.5isnotreautredf snoperable SLC tank heater circuit reaker(s)
'kprovidedtheotherre, trements of ACTION 3.8.4.5.b are ce ted with. J FERMI - UNIT 2 3/4 8 26 Amenoment No. 38 .
O v
PAGE 37 0F 43
MOA Wh TABLE 3.B.4. 5 1 11gQBY LIOUID NTRotSYSTEMASSOCIATEDISOLAdONDEVICES 480 V MOTOR CONTROL CENTERV l MCC 728 4C P tion 2AR SLC Pump ,f
^
MC 72C-4A Position SC C Heater A f l MCC 72E-5B Positio 2B SLC Pump B Pos ton 2CR SCL Heater B O
FERMI UHli 2 3/4 8 27 Amendment No. 22, 38 O
ME Vo _0F 43
1C 00 &D On U
EUNG OPERATION 3/4.9.5 COMMUN MATIONS w LIMITING C IONFOROPERATd - /
3.9. Direct comunic ton shall be intained be een the control room and re eling platform o sonnel.
APPLICABILITY: ERATIONAL C0 TION 5 durte CORE ALTERATIONS.*
gl1QB:
When dir t communicat n between the ontrol room and r eling platform perso cannot be intained, i tately suspend CO ALTERATIONS.* ,
l
/ l SURVEtttANCE OUIREMENTS 4.9.5 rect ecmunicatt between the con 1 roor, and refueling platform cer'so el shall be demo trated within one our prior to the start of and at
.eas once per 12 hou during CORE ALTE 10NS.*
- Lucent movement of control ro with tneir core drive system.
FERMI UNIT 2 3/4 9 7 g
O PAGE 6'/ _0F 43
SeloCQ'bcb l i
)
REFUELING OPERATIONS R.Ih 3/4.9.6 A EFUELING M ATFORM _ __
LI C CONDI FOR OPE ON
/
/3.9.6 i refueling assemb es or contr atform shall be rods within the RABLE and used for actor pressure vess .
dling fuel uring handling of .uel assemblies or e trol rods within the l M11111: !
r ctor press vessal.
EllDti:
With t eautrements for efueling platform RABILITI not sati ed, susp use of any inop able refueling pla rm eautement fro erations invo ing the handlin f control rods an uel assemblies w in the reactor pr ure vessel afte placing the load a safe condition SURVEILLANCE RE IREMENTS 4.9.6 Eac refueling platf . hoist used for ndling of control rods o uel assembli within the reat r pressure vess shall be demonstrated 0 BLE within days prior to t start of such erations with that hois :
O \
'\
- a. Demonstr ing operation o he overload cuteff wn exceeds 200 00unds for e fuel grapple hoist NF-40 mast.1395 pot s for the fuel grappi
'd 1050 pounds for al he load h the Model oist with the ner hoists.
l Mode NF-500 mast. !
- b. monstrating o ration of the uptrav stop when fuel grappl hoist uptrave ano frame mounted an monorail auxiliary ho s uptravel br g the point of attac ent of the fuel asse y or control r to within 6 feet 4 ches or greater belo he top of the ref ing platf orm trac, ' ,
- c. Demo trating operatto the downtravel cut wnen the end of th fuel grapple not cowntravel reaches 5 est 3 inches or 1 s oW the top of t platform tracks and en the end of the ame ounted ano non il austliary noists r en 85 feet or less elow the top of in latform tracts, I d. Demonstra ng operation of the ek' cable cutoff pri to the o less than 40 poun for the fuel 'I hoist c e tenston cecreasin grapp holst.
- e. D onstrating operatton the loaded interloc wnen the load '
xceeds 535 pounds for ne fuel grapple hois with either t
~
{
Model NF 400 mast or ne Model hF 500 mast, and 450 pound or all l 5 other hoists, w -
FERMI UN112 3/4 9 8 Amenoment No. 4, 86 PAGE %2 0F 43
C b 6CQ,- O REFUELING OPERATIONS I3/4.9.7 CRANE TRAVEL-SPdFUEL STOMCFEL LIMITI ONDITION F PERATION , ,
.9.7 Loads excess of 0 pounds shall b rohibited from travel over fuel assemb es in the s nt fuel storage p racks.
APPLIC LITY: Wit uel assemblies in e spent fuel storage poo racks.
ith the re trements of the a e specification not sa fled, place the crane loap in a safe conditt . The provisions of Sp fication 3.0.3 are not applica,bie.
SURVEIllAN RE001REMENTS /
4.9. Loads, other than fu assemblies, shall erified to be 1 s than or O eau to 1100 pounos prior o movement over fuel ssemblies in t htoragepoolrack spent fuel j
)
FERMI . UNIT 2 3/4 9 9
'/9 v
DAGE G 0F 43 t
_-_____ _--_ _ _ a
DISCUSSION OF CHANGES RELOCATED CTS INCLUDED IN APPENDIX A TO THE SPLIT REPORT O RELOCATED SPECIFICATIONS R.1 In the event of an earthquake, seismic instrumentation is required to permit comparison of the measured response to that used in the design basis of the facility to determine if plant shutdown is required pursuant to Appendix "A" of 10 CFR Part 100. Since this is determined
. after the event has occurred. it has no bearing on the mitigation of any DBA.
Comoarison to Deterministic Screenina Criteria:
- 1. These instruments are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a design basis accident (DBA).
- 2. These instruments do not monitor a process variable that is an initial condition of a DBA or transient analyses.
- 3. These instruments are not part of a primary success path in the mitigation of a DBA or transient.
- 4. As discussed in Sections 3.5 and 6. and summarized in Table 41 (item 151) of NED0 31466. the loss of seismic monitoring instrumentation was found to be a non significant risk l contributor to core damage frequency and offsite releases. DECO has reviewed this evaluation, considers it applicable to Fermi- l 2, and concurs with the assessment.
Conclusion-Since the screening criteria have not been satisfied, the Seismic Monitoring LCO and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications.
I
'I O) u FERMI UNIT 2 1 REVISION 0. 04/03/98
4 DISCUSSION OF CHANGES RELOCATED CTS INCLUDED IN APPENDIX A TO THE SPLIT REPORT
'O RELOCATED SPECIFICATIONS R.2 Heteorological instrumentation is used to measure environmental parameters that may affect distribution of fission products and gases following a design basis accident (DBA), but it is not an input assumption for any DBA analvsis and does not mitigate the accident.
Heteorological information is required to evaluate the need for initiating protective measures to protect the health and safety of the public.
Comoarison to Deterministic Screenina Criteria _;.
- 1. These instruments are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. These instruments do not monitor a process variable that is an initial condition of a DBA or transient analyses.
- 3. These instruments are not part of a primary success path in the o mitigation of a DBA or transient, b) As discussed in Sections 3.5 and 6, and summarized in Table 41 4.
(item 152) of NED0 31466, the loss of meteorological monitoring {
instrumentation was found to be a non significant risk contributor to core damage frequency and offsite releases. DECO has reviewed this evaluation, considers it applicable to Fermi-2, and concurs with the assessment.
Conclusion:
Since the screening criteria have not been satisfied, the Meteorological Monitoring Instrumentation LCO and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications, j l
n t
FERMI UNIT 2 2 REVISION 0, 04/03/98
DISCUSSION OF CHANGES RELOCATED CTS INCLUDED IN APPENDIX A TO THE SPLIT REPORT C)
V BELOCATED SPECIFICATIONS R.3 The Traversing In Core Probe (TIP) System is used for calibration of the LPRM detectors. The TIP System is positioned axially and radially throughout the core to calibrate the local power range monitors (LPRMs). When not in use the TIP instruments are retracted into a storage position outside the drywell. The TIP System supports the operability of the LPRMs. With LPRM operability addressed there is no need to address the TIP System in the Technical Specifications.
Comnarison to Screenina Criteria:
- 1. The TIP System is not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a design basis accident (DBA).
- 2. The TIP System does not monitor a process variable that is an l initial condition of a DBA or transient analyses. l
)
- 4. As discussed in Sections 3.5 and 6, and summarized in Table 41 (item 183) of NED0 31466. the loss of the TIP System was found ;
to be a non significant risk contributor to core damage l frequency and offsite releases. DECO has reviewed this evaluation. considers it applicable to Fermi 2, and concurs with 4 the assessment. j
Conclusion:
Since the screening criteria have not been satisfied, the TIP System LCO and Surveillances may be relocated to other plant controlled I documents outside the Technical Specifications. l O
FERMI UNIT 2 3 REVISION 0, 04/03/98
DISCUSSION OF CHANGES RELOCATED CTS INCLUDED IN APPENDIX A TO THE SPLIT REPORT b RELOCATED SPECIFICATIONS R.4 The Operability of the Chlorine Detection System ensures sufficient capability is available to promptly detect and initiate protective action in the event of an accidental chlorine release. This capability is required to protect control room personnel. However, the instruments are not assumed to mitigate a design basis accident (DBA) or transient since an accidental chlorine release is not a DBA or transient.
Comparison to Screenina Criteria:
- 1. The Chlorine Detection System is not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. The Chlorine Detection System does not monitor a process variable that is an initial condition of a DBA or transient analyses.
p 3. The Chlorine Detection System is not part of a primary success
( path in the mitigation of a DBA or transient.
- 4. As discussed in Sections 3.5 and 6. and summarized in Table 41 (item 184) of NED0 31466. the loss of the Chlorine Detection System was found to be a non significant risk contributor to core damage frequency and offsite releases. DECO has reviewed this evaluation, considers it applicable to Fermi 2, and concurs with the assessment.
Conclusion:
Since the screening criteria have not been satisfied, the Chlorine Detection System LC0 and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications.
? A
- i. )
v FERMI UNIT 2 4 REVISION 0, 04/03/98
DISCUSSION OF CHANGES RELOCATED CTS INCLUDED IN APPENDIX A TO THE SPLIT REPORT O
d RELOCATED SPECIFICATIONS R.5 The Loose Part Detection System is used to detect loose parts in the reactor vessel. The instrumentation does not indicate that there is a degradation in the primary pressure boundary but indicates that there might be a remote chance of damage to a component due to a loose part.
In the event of fuel failure due to fuel bundle flow blockage from a loose part, it will be detected by the radiation monitors in the offgas stream.
Comoarison to Screenina Criteria:
- 1. The Loose Part Detection System is not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a design basis accident (DBA).
- 2. The Loose Part Detection System is not used to monitor a process variable that is an initial condition of a DBA or transient analyses.
- 3. The Loose Part Detection System is not part of a primary success path in the mitigation of a DBA or transient.
- 4. As discussed in Sections 3.5 and 6, and summarized in Table 41 (item 187) of NED0 31466, the loss of the Loose Part Detection System was found to be a non significant risk contributor to core damage frequency and offsite releases. DECO has reviewed this evaluation, considers it applicable to Fermi 2, and concurs with the assessment.
Conclusion:
Since the screening criteria have not been satisfied, the Loose Part Detection System LC0 and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications.
O FERMI UNIT 2 5 REVISION 0, 04/03/98
DISCUSSION OF CHANGES RELOCATED CTS INCLUDED IN APPENDIX A TO THE SPLIT REPORT
/^)
V RELOCATED SPECIFICATIONS R.6 The explosive gas monitor Specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the main condenser offgas treatment system is adequately monitored, which will help ensure that the concentration is maintained below the flammability limit of hydrogen. However, the offgas system is designed to contain detonations and will not affect the function of any safety related equipment. The concentration of hydrogen in the offgas stream is not an initial assumption of any design basis accident (DBA) or transient analysis.
Comparison to Screenino Criteria:
- 1. The Explosive Gas Monitoring Instrumentation is not used for.
nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. The Explosive Gas Monitoring Instrumentation does not monitor a process variable that is an initial condition of a DBA or f, transient. Excessive system hydrogen is not an indication of a
- 3. The Explosive Gas Monitoring Instrumentation is not part of a primary success path in the mitigation of a DBA or transient.
Excessive hydrogen discharge is not considered to initiate a primary success path in mitigating a DBA or transient.
1
- 4. As discussed in Sections 3.5 and 6. and summarized in Table 41 (items 189 and 306) of NEDO 31466. the loss of the Explosive Gas Monitoring Instrumentation was found to be a non significant risk contributor to core damage frequency and offsite releases.
DECO has reviewed this evaluation, considers it applicable to Fermi 2, and concurs with the assessment. Moreover, the offgas system, including provisions to maintain H, below explosive limit, while under scope of the Maintenance Rule (MR) has been determined by the HR Expert Panel to be non risk significant.
Conclusion:
i Since the screening criteria have not been satisfied, the Explosive Gas Monitoring Instrumentation LC0 and Surveillances may be relocated (g~) to other plant controlled documents outside the Technical Specifications.
FERMI UNIT 2 6 REVISION 0, 04/03/98 1
DISCUSSION OF CHANGES RELOCATED CTS INCLUDED IN APPENDIX A TO THE SPLIT REPORT RELOCATED SPECIFICATIONS R.7 The Appendix R Alternative Shutdown Instrumentation is used to ensure that a fire will not preclude achieving safe shutdown. This instrumentation is independent of areas where a fire could damage systems normally used to shutdown the reactor. However, the
. instrumentation is not used to detect a degradation of the reactor coolant pressure boundary nor assumed to mitigate a design basis accident (DBA) or transient event. The Appendix R Alternative Shutdown Instrumentation capability is consistent with the requirements of 10 CFR 50, Appendix R.
Comoarison to Screenino Criteria:
- 1. The Appendix R Alternative Shutdown Instrumentation is not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. The Appendix R Alternative Shutdown Instrumentation does not q
Q monitor a process variable that is an initial condition of a DBA or transient analyses.
- 3. The Appendix R Alternative Shutdown Instrumentation is not part of a primary success path in the mitigation of a DBA or transient.
- 4. The instrumentation included in this LCO are part of the Dedicated Shutdown System. This system, while covered under the Haintenance Rule (MR) scope. has been determined by the HR Expert Panel to be non risk significant due principally to the low probability of a fire of sufficient magnitude to require alternate shutdown action. Therefore, DECO found that the loss of the Appendix R Alternative Shutdown Instrumentation to be a non significant risk contributor to core damage frequency and offsite releases.
C.gnclusion:
i Since the screening criteria have not been satisfied, the Appendix R '
Alternative Shutdown Instrumentation LC0 and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications.
FERMI UNIT 2 7 REVISION 0, 04/03/98
_____________.______U
DISCUSSION OF CHANGES RELOCATED CTS INCLUDED IN APPENDIX A TO THE SPLIT REPORT G
V RELOCATED SPECIFICATIONS R.8 Poor reactor coolant water chemistry may contribute to the long term degradation of system materials and thus is not of immediate importance to the plant operator. Reactor coolant water chemistry is monitored for a variety of reasons. One reason is to reduce the possibility of failures in the reactor coolant system pressure boundary caused by corrosion. Severe chemistry transients have resulted in failure of thin walled LPRM instrument dry tubes in a relatively short period of time. However, these LPRM dry tube failures result in loss of the LPRM function and are readily detectable. In summary, the chemistry monitoring activity serves a long term preventative rather than mitigative purpose.
[_omoarison to Screenina Criteria:
- 1. Reactor coolant water chemistry is not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a design basis accident (DBA).
(O/ 2. Reactor coolant water chemistry is not, and does not monitor, a process variable that is an initial condition of a DBA or transient. I
- 3. Reactor coolant water chemistry is not part of a primary success ]
path in the mitigation of a DBA or transient.
- 4. As discussed in Sections 3.5 and 6, and summarized in Table 41 )
(item 211) of NED0 31466, the reactor coolant water chemistry was found to be a nonsignificant risk contributor to core damage frequency and offsite releases. DECO has reviewed this i evaluation, considers it applicable to Fermi 2, and concurs with the assessment.
Conclusion:
Since the screening criteria have not been satisfied, the Reactor Coolant System Chemistry LCO and Surveillances may be relocated te other plant controlled documents outside the Technical Specifications. -
( )
v FERMI UNIT 2 8 REVISION 0, 04/03/98
DISCUSSION OF CHANGES RELOCATED CTS INCLUDED IN APPENDIX A TO THE SPLIT REPORT RELOCATED SPECIFICATIONS R.9 The inspection programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity of these compcnents will be maintained throughout the components life. Other Technical Specifications require important systems to be operable (for example, ECCS 3/4.5.1) and in a ready state for mitigative action. This Technical Specification is more directed toward prevention of component degradation and continued long term maintenance of acceptable structural conditions. Hence it is not necessary to retain this specification to ensure immediate operability of safety systems.
Further, this Technical Specification prescribes inspection requirements which are performed during plant shutdown. It is, therefore, not directly important for responding to design basis accidents (DBA).
Comoarison to Screenina Criteria:
- 1. The inspections stipulated by this Specification are not "instclied instrumentation" capable of detecting a significant O abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. The inspections stipulated by this Specification do not monitor process variables that are initial assumptions in a DBA or transient analyses.
- 3. The ASME Code Class 1. 2. and 3 components inspected per this Specification are assumed to function to mitigate a DBA. Their capability to perform this function is addressed by other Technical Specifications. This Technical Specification, however, only specifies inspection requirements for these components: and these inspections can only be performed when the plant is shutdown. Therefore, Criterion 3 is not satisfied.
O FERMI UNIT 2 9 REVISION 0, 04/03/98
DISCUSSION OF CHANGES RELOCATED CTS IN,mVDED IN APPENDIX A TO THE SPLIT REPORT
\ R9 (continued)
- 4. As discussed in Sections 3.5 and 6. and summarized in Table 41 (item 216) of NED0 31466, the assurance of operability of the entire system as verified in the system operability specification dominates the risk contribution of the system. As such, the lack of a long term assurance of structural integrity as stipulated by this Specification was found to be a non-significant risk contributor to core damage frequency and offsite releases. Furthermore, the requirement is currently covered by 10 CFR 50.55a and the plant's Inservice Inspection Program. Deco has reviewed this evaluation, considers it applicable to Fermi 2, and concurs with the assessment.
Conclusion:
Since the screening criteria have not been satisfied, the Structural Integrity LC0 and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications.
O G l 1
i U
1 FERMI UNIT 2 10 REVISION 0, 04/03/98
DISCUSSION OF CHANGES RELOCATED CTS INCLUDED IN APPENDIX A TO THE SPLIT REPORT {
RELOCATED SPECIFICATIONS R.10 The suppression pool and drywell spray modes of RHR are utilized to condense the steam that is discharged into the drywell and suppression chamber during a LOCA. Emergency operating procedures direct manual initiation of the suppression pool and drywell spray modes of RHR.
However, in the analysis of the bounding event for the containment analysis and the suppression pool pressurization due to bypass leakage, the suppression pool and drywell spray modes of RHR were not utilized for mitigation of the event.
Comoarison to Screenina Criteria:
- 1. The suppression pool and drywell spray modes of RHR are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a design basis accident (DBA).
1
- 2. The suppression pool and drywell spray modes of RHR do not monitor a process variable that is an initial condition of a DBA ,
or transient analyses. 1
- 3. The suppression pool and drywell spray modes of RHR are not part of a primary success path in the mitigation of a DBA or transient.
- 4. As discussed in Section 6 and summarized in Table 41 (Item 368) of NED0 31466. Supplement 1. the loss of the drywell spray mode of RHR was found to be a non significant risk contributor to core damage frequency and offsite releases. DECO has reviewed this evaluation, considers it applicable to Fermi 2, and concurs with the assessment. DECO has also evaluated the loss of the suppression pool spray mode of RHR and determined it to be a non significant risk contributor to core damage frequency and offsite releases.
Conclusion:
Since the screening criteria have not been satisfied, the Suppression Pool and Drywell Spray mode of RHR LC0 and Surveillances, may be -
relocated to other plant controlled documents outside the Technical Specifications.
!3 o
FERMI UNIT 2 11 REVISION 0. 04/03/98 Y
DISCUSSION OF CHANGES REl0CATED CTS INCLUDED IN APPENDIX A TO THE SPLIT REPORT RELOCATED SPECIFICATION R.11 The purpose of the shore barrier is to protect the site backfill from wave erosion. Category 1 structures are designed to withstand the impact of waves up to 5.4 feet. So long as the backfill is in place, waves greater than 5.4 feet cannot impact Category 1 structures because of the lack of sufficient depth of water to sustain such waves, As noted in Section 2.4.5.7 of the UFSAR, the importance of the shore barrier in protecting Category 1 structures has been greatly reduced from the originally approved concept. The shora barrier can sustain a high degree of damage and still perform its function, protecting the site backfill from erosion. As such, the shore barrier requirement is more directed toward prevention of degradation and continued long term maintenance of acceptable structural conditions.
Hence it is not necessary to retain this specification to ensure immediate operability of safety systems.
Comoarison to Screenino Criteria:
- 1. The shore barrier protection requirement is not i!>,ed for, nor capable of, detecting a significant abnormal degradation of the "O reactor coolant pressure boundary prior to a DBA.
- 2. The shore barrier protection requirement is not, and does not monitor, a process variable that is an initial condition of a DBA or transient analyses.
- 3. The shore barrier protection requirement is not part of a primary success path in the mitigation of a DBA or transient.
- 4. Based on the discussion above and noting that it is not under the scope of the Maintenance Rule (MR) following evaluation by the MR Expert Panel, the shore barrier is found to be non risk significant.
Conclusion:
Since the screening criteria have not been satisfied, the Shore Barrier Protection LC0 and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications. -
l
- ]v j i
FERMI UNIT 2 12 REVISION 0, 04/03/98
DISCUSSION OF CHANGES RELOCATED CTS INCLUDED IN APPENDIX A TO THE SPLIT REPORT R.12 The limitations on sealed' source contamination are intended to ensure that the total body or individual organ irradiation deses does not exceed allowable limits in the event of ingestion or inhalation. This is done by imposing a maximum limitation of s 0.005 microcurier of removable contamination on each sealed source. This requirement and the associated Surveillance Requirements bear no relation to the
. conditions or limitations which are necessary to ensure safe reactor operation.
Comoarison to Screenino Criteria:
- 1. Sealed source contamination is not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a design basis accident (DBA).
- 2. Sealed source contamination is not a process variable that is an initial condition of a DBA or transient.
- 3. Sealed source contamination is not part of a primary success path in the mitigation of a DBA or transient.
- 4. As discussed in Sections 3.5 and 6, and summarized in Table 41 (item 267) of NED0 31466, the sealed source contamination being not within limits was found to be a non significant risk contributor to core damage frequency and offsite releases.
Detroit Edison has reviewed this evaluation, considers it applicable to Fermi 2 and concurs with the assessment.
Conclusion:
Since the screening criteria have not been satisfied, the Sealed Source Contamination LCO and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications.
1 I
[
FERMI UNIT 2 13 REVISION 0, 04/03/98
DISCUSSION OF CHANGES RELOCATED CTS INCLUDED IN APPENDIX A TO THE SPLIT REPORT R.13 The Appendix R Alternative Shutdown Auxiliary Systems identified in this specification (standby feedwater system, combustion turbine-generator, drywell cooling units, and necessary control circuits) are used to ensure that a fire will not preclude achieving safe shutdown.
These auxiliary systems are independent of areas where a fire could damage systems normally used to shutdown the reactor. However, the
, systems are not used to detect a degradation of the reactor coolant pressure boundary nor assumed to mitigate a design basis accident (DBA) or transient event. The Appendix R Alternative Shutdown Auxiliary Systems capabilities are consistent with the requirements of 10 CFR 50, Appendix R.
Comoarison to Screenina Criteria:
- 1. The Appendix R Alternative Shutdown Auxiliary Systems are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. The Appendix R Alternative Shutdown Auxiliary Systems do not monitor a process variable that is an initial condition of a DBA O or transient analyses.
- 3. The Appendix R Alternative Shutdown Auxiliary Systems are not part of a primary success path in the mitigation of a DBA or transient.
- 4. The auxiliary systems covered by this LC0 are used to effect a plant shutdown from outside the control room using the Dedicated Shutdown System. This system, while covered under the Maintenance Rule (HR) scope, has been determined by the HR Expert Panel to be non risk significant due principally to the low probability of a fire of sufficient magnitude to require alternate shutdown action.
Conclusion:
l Since the screening criteria have not been satisfied, the Appendix R )
Alternative Shutdom Auxiliary Systems LC0 and Surveillances may be relocated to other plant controlled documents outside the Technical -!
Specifications.
tO v
FERMI UNIT 2 14 REVISION 0, 04/03/98
1 DISCUSSION OF CHANGES
{
RELOCATED CTS INCLUDED IN APPENDIX A TO THE SPLIT REPORT in b R.14 The circuits involved in this LC0 are kept normally de energized and do not participate in plant safety actions. These circuits are primarily for lighting, utility outlets and convenient power plugs, to be used in the event of plant walkdowns, maintenance and in situ test and/or observations. Therefore, they are of non Class 1E nature.
Comoarison to Screenina Criteria:
- 1. The A.C. circuits listed in this Specification are de energized during operation and are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a design basis accident (DBA).
- 2. The A.C. circuits listed in this Specification do not monitor a process variable'that is an initial condition of a DBA or transient.
- 3. The A.C. circuits listed in this Specification are not part of a primary success path in the mitigation of a DBA or transient.
- 4. As discussed in Sections 3.5 and 6, and summarized in Table 4 1
( (item 275) of NED0 31466. the A.C. circuits inside primary containment governed by thic specification were found to be a non significant risk contributor to core damage frequency and offsite releases. DECO has reviewed this evaluation, considers {
it applicable to Fermi 2, and concurs with the assessment. '
Conclusion:
Since the screening criteria have not been satisfied, the A.C.
Circuits Inside Primary Containment LCO and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications.
1
- 1
-l l
m]
l FERMI UNIT 2 15 REVISION 0, 04/03/98 1
1
DISCUSSION OF CHANGES RELOCATED CTS INCLUDED IN APPENDIX A TO THE SPLIT REPORT O R.15 The primary feature of these protective devices is to open the control and/or power circuit whenever the load conditions exceed the present current demands. This is to protect the circuit conductors against damage or failure due to overcurrent heating effects. The continuous monitoring of the operating status of the overcurrent protection devices is impracticable and not covered as part of the centrol room monitoring, except after trip condition indication.
In the event of failure of this protective device to trip the circuit, the upstream protective device is expected to operate and isolate the faulty circuit. Thus, the upper level (back up) protection will prevent loss of redundant power source. In the worst case fault condition, a single division of protective functions can be lost.
However, this scenario is covered under a single failure criterion.
The overcurrent protection devices ensure the pressure integrity of the containment penetration. With failure of the device it is postulated that the wire insulation will degrade resulting in a containment leak path during a LOCA. However, containment leakage is not a process variable and is not considered as part of the primary success path. Containment penetration degradation will be identified O during the normal containment leak rate tests required by 10 CFR Part 50, Appendix J.
Comoarison to Screenino Criteria:
- 1. The primary containment penetration conductor overcurrent protective devices are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a design basis accident (DBA).
- 2. The primary containment penetration conductor overcurrent protective devices do not monitor a process variable that _is an initial condition of a DBA or transient.
3.- The specific circuits of the primary containment penetration conductor overcurrent protective devices are not part of a
. primary success path in the mitigation of a DBA or transient.
O FERMI UNIT 2 16 REVISION 0. 04/03/98
l l
DISCUSSION OF CHANGES RELOCATED CTS INCLUDED IN APPENDIX A TO THE SPLIT REPORT
\ R.15 (continued)
- 4. ~ As discussed in Sections 3.5 and 6, and summarized in Table 41 i (item 276) of NED0 31466, the loss of the circuits associated I with the primary containment penetration conductor overcurrent protective devices was found to be a non significant risk contributor to core damage frequency and offsite releases. Deco has reviewed this evaluation, considers it applicable to Fermi 2, and concurs with the assessment.
Conclusion:
Since the screening criteria have not been satisfied, the Primary Containment Penetration Conductor Overcurrent Protective Devices LC0 and Surve111ances may be relocated to other plant controlled documents outside the Technical Specifications.
O FERMI UNIT 2 17 REVISION 0, 04/03/98
DISCUSSION OF CHANGES RELOCATED CTS INCLUDED IN APPENDIX A TO THE SPLIT REPORT r
b, R.16 For valves with thermal overload protection (i.e., trip on overload condition), the valve function should be accomplished prior to overload trip. The valve function for these valves is meant to take precedence over the overload protection. If the overload condition occurs during valve operation, in case of failure of overload protection operation to disconnect the load, the safety function of the valve will not be performed. This affects the Operability of the system containing the valve. Accordingly the more appropriate LC0 i would be to address the overall system Operability and not the Operability of a support system. Additionally, the surveillance and maintenance of the devices can be controlled by sources other than the plant Technical Specifications.
Eqmoarison to Screenino Criteria:
- 1. Motor operated valve thermal overload protection is not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a design basis accident (DBA).
p 2. Motor operated valve thermal overload protection is not, and V does not monitor, a process variable that is an initial condition of a DBA or transient.
- 3. Actuation of a motor operated valve's thermal overload I protection is not part of a primary success path in the mitigation of a DBA or transient. The supported system (e.g.,
ECCS) may be part of a success path and is then retained in the Technical Specifications. However, motor operated valve thermal overload protection retention in the Technical Specifications is not necessary as its safety related role is implicitly addressed in the Operability determination of the supported system.
- 4. PRAs address system risk contribution and identify systems which can be significant risk contributors to core damage and offsite l releases. Subcomponent risk contribution from motor operated valve thermal overload protection malfunction is not addressed as the inoperability of the supported system and subsequent risk contribution is inclusive of thermal overload protection failure i potential.
N (O
FERMI UNIT 2 18 REVISION 0, 04/03/98
-l
DISCUSSION OF CHANGES q RELOCATED CTS INCLUDED IN APPENDIX A TO THE SPLIT REPORT i R.16 (continued)
Conclusion:
Since the screening criteria have not been satisfied, the Motor-Operated Valve Thermal Overload Protection LC0 and Surveillances may
. be relocated to other plant controlled documents outside the Technical Specifications.
O O
FERMI UNIT 2 19 REVISION 0, 04/03/98
1 i
( DISCUSSION OF CHANGES RELOCATED CTS INCLUDED IN APPENDIX A TO THE SPLIT REPDRT
'O O R.17 Circuit breakers and thermal overload devices actuated by fault currents are used as isolation devices to protect equipment associated with the Standby Liquid Control (SLC) System. Lack of continuity of any isolation device associated with either the incoming line or the pumps will disable only one train of the SLC System: the other train would still be available. Only one train is needed to be Operable for the SLC System to perform its intended function (i.e., meet 10 CFR 50.62 requirements). The SLC tank heaters are only required when mixing sodium pentaborate and/or water to establish the required solution operating parameters. Normal operation of the SLC System does not depend upon these heaters. Therefore, failure of the circuit breakers associated with the heaters and the heat tracing will not affect SLC System performance.
Comoarison to Screenina Criteria:
- 1. The SLC System associated isolation devices are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. The SLC System associated isolation devices do not monitor a (m) process variable that is an initial condition of a DBA or transient analyses.
- 3. The SLC System associated isolation devices are not part of a primary success path in the mitigation of a DBA or transient.
- 4. The SLC System is judged by DECO to be risk significant.
However, SLCS is retained as ITS 3.1.7 with pump surveillance requirements that adequately assure their operation without the need to separately test the continuity of the isolation devices.
The isolation function itself is not judged to be risk significant nor are the function of the heaters as discussed above. Thus. this LCO provides no additional risk benefits.
Conclusion:
Since the screening criteria have not been satisfied, the SLC System Associated Isolation Devices LCO and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications. '
O FERMI UNIT 2 20 REVISION 0, 04/03/98
DISCUSSION OF CHANGES RELOCATED CTS INCLUDED IN APPENDIX A TO THE SPLIT REPORT (3
U RELOCATED SPECIFICATIONS R.18 Communication between the control room and refueling floor personnel is maintained to ensure that refueling personnel can be promptly informed of significant changes in the plant status or core reactivity condition during refueling. The comunications allow for coordination of activities that require interaction between the control room and refueling floor personnel (such as the insertion of a control rod prior to loading fuel). However, the refueling system design accident or transient response does not take credit for communications.
Comparison to Screenina Criteria:
- 1. Communication during any mode of plant operation is not used for, nor capable 'of, detecting a significant abnormal j degradation of the reactor coolant pressure boundary prior to a design basis accident (DBA).
- 2. Communication during any mode of plant operation is not, and does not monitor, a process variable that is an initial p condition of a DBA or transient. ;
J \
- 3. Communication during any mode of plant operation is not part of a primary success path in the mitigation of a DBA or transient.
- 4. As discussed in Sections 3.5 and 6. and summarized in Table 41 (item 286) of NED0 31466. the loss of communication was found to be a non significant risk contributor to core damage frequency and offsite releases. DECO has reviewed this evaluation, considers it applicable to Fermi 2, and concurs with the assessment.
Conclusion:
Since the screening criteria have not been satisfied, the Communications LC0 and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications.
O FERMI - UNIT 2 21 REVISION 0, 04/03/98
DISCUSSION OF CHANGES RELOCATED CTS INCLUDED IN APPENDIX A TO THE SPLIT REPORT
)
n '
U RELOCATED SPECIFICATIONS I
R.19 OPERABILITY of the refueling platform equipment (main hoist and auxiliary hoists) ensures that: 1) only Operable hoists of the refueling platform will be used to handle fuel within the reactor pressure vessel: 2) hoists have sufficient load capacity for handling fuel assemblies and/or control rods: 3) fuel and control rods will not be brought too close to the surface; and 4) the core internals and pressure vessel are protected from excessive lifting force if they are inadvertently engaged during lifting operations. Although the interlocks designed to provide the above capabilities can prevent damage to the refueling platform equipment and core internals, they are not assu:?ed to function to mitigate the consequences of a design basis accident. Further, in analyzing the control rod withdrawal error during refueling, if any one of the operations involved in initial failure or error is followed by any other single equipment i failure or single operator error, the necessary safety actions are j taken (e.g., rod block or scram) automatically prior to violation of j any limits. Hence the refueling platform interlocks are not part of the primary success path in mitigating the control rod withdrawal o error during refueling.
Q Comparison to Screenina Criteria:
l
- 1. The refueling platform and associated instrumentation is not used for, nor capable of, r;etecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a '
design basis accident (DBA).
- 2. The refueling platform and associated instrumentation does not )
monitor a process variable that is an initial condition of a DBA )
or transient. l
- 3. The refueling platform and associated instrumentation is not part of a primary success path in the mitigation of a DBA or transient.
)
- 4. As discussed in Sections 3.5 and 6. and summarized in Table 4-1 (item 287) of NED0 31466, the refueling platform and associated instrumentation was found to be a non significant risk -)
contributor to core damage frequency and offsite releases. DECO i m has reviewed this evaluation considers it applicable to Fermi- )
- 2. and concurs with the assessment.
(]
FERMI - UNIT 2 22 REVISION 0, 04/03/98 I
DISCUSSION OF CHANGES RELOCATED CTS INCLUDED IN APPENDIX A TO THE SPLIT REPORT O R.19 (continued)-
I
Conclusion:
Since the screening criteria have not been satisfied, the Refueling Platform LC0 and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications.
O ,
O FERMI UNIT 2 23 REVISION 0, 04/03/98
I DISCUSSION OF CHANGES RELOCATED CTS INCLUDED IN APPENDIX A TO THE SPLIT REPORT O
V RELOCATED SPECIFICATIONS I
R.20 The restriction on movement of loads in excess of the nominal weight I of a fuel assembly over other fuel assemblies in the storage pool ensures that in the event the load is dropped, the activity release will be limited to that contained in a single fuel assembly and any possible distortion of the fuel in the storage racks will not result in a critical array. This load control is implemented administratively by Fermi 2 Conduct Manual MMA07 (" Hoisting, Rigging and Load Handling"). The cited chapter, explicitly includes the 1100 pound limitation for load travel over spent fuel storage pool racks.
Comoarison to Screenino Criteria:
- 1. The load limit is not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. The maximum severity assumed for the fuel handling DBA is limited by the limits placed on the crane travel. These crane travel limits are not, however, process variables monitored and
( controlled by the operator: they are administrative controls.
Therefore, Criterion 2 is not satisfied.
- 3. The load limit is not a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA.
- 4. Traditional PRAs do not review risks associated with the spent fuel storage pool. Design basis analyses indicate that the release associated with fuel assembly damage in the spent fuel storage pool due to crane accidents is significantly lower than releases of concern evaluated by PRAs. Moreover, the associated crane and reactor building storage pools, while under the scope of the Maintenance Rule (MR) have been determined by the MR Expert Panel to be non risk significant.
Conclusion:
Since the screening criteria have not been satisfied, the " Crane '
Travel Spent Fuel Storage Pool" LC0 and Surveillances may be
. relocated to other plant controlled documents outside the Technical Specifications.
FERMI - UNIT 2 24 REVISION 0, 04/03/98 J
O
. APPENDIX B OTHER JUSTIFICATIONS FOR SPECIFICATION / PAGE RELOCATION / DELETION O
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O IIRNT UNil 2. 1 2b Amendment No. 64 G
PAGE I 0F 57
REWM6 O
REACTIVITY NTROLSYSTEM/ '}
CONTROL R00 DRIVE H SUPPORT LIMITING CONDITIO R OPERATION ' r 3.1.3.8. The e trol rod drive housing support all be in place.
APPLICABILIT : OPERATIONAL CONDITIONS 1, 2 d 3.
EllDH:
With th control rod drive housing su art not in place, be in at least HOT SHUTD0 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD HUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> SURVElllANCE REOUIREMENTS f 4.1.3.8 The control ro drive housing support shall be verif ed to be in place by a visual insp ction prior to startup any time it h been disassembled or when aintenance has been performed in the ontrol rod drive g support are i
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FtRMI - UNIT 2 3/4 3 85 henoment No. 33, 71 l
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O FERH1 - UNIT 2 3/4 6-7 Amendment No. 46, 108 Corrected August 19. 1996
O 1.t Table 3.6.3 1 has b n deleted.
3/4 6 22 t ough 3/4 6-47 are "not used."
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'O r<a > u 11 2 ><< e >> a e t . ior MiE 15 op 57 ,
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(PLANTSYSTEMS 3/4.7.5 SNUBBERS JIMITINGCONDITIONFOR:0PERATION 3.7.5 All* hydraulic and mechanical snubbers shall be OPERA .
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. OPE IONAL CONDITIONS 4 and 5 for snubbers located on systems required OPERABLE those OPERATIONAL CONDITIONS.
ACTION:
With one or more snubbers inocerable on any syste , within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> replace or restore the inocerable snubber (s) to OPERABLE s tus and perform an engineering evaluation per Specification 4.7.5 on the attached component or declare the attached system inoperable and fo low the appropriate ACTION statement for that system.
SURVElttANCE RE0VIREMENTS 4.7.5 Each snubber shall be demonst ted OPERABLE by performance of the following augmented inservice inspe ion program in addition to the requirements of Specification 4.0. .
A
( ) a. Inseettien Tvoes
%)
As used in this s ecification, type of snubber shall mean snubbers of the same dest n ano manufacturer, irrespective of capacity.
- b. Visual Inseet ons .
Snubbers ar categorizeo as inaccessible or accessible durtng '
reactor oc ation. Eacn of these categories (inaccessible and accessibl may De inspected inoeoendently according to the schedule etermined by Table 4.7.5-1. The visual inspection interva for each category of snuDDer shall be determined btsed upon t e criteria orovioed in Table 4.7.5 1. The first inspection int'er al determineo using this criteria shall be based upon the I
prev ous inspection interval as established by the requirements in eff ct before Amenoment 84. l l
- A( cescribed in the bases.
FERM! - UNIT 2 3/4 7 16 Amendment No. 84
,A PAGE a6 0F 57 l,
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k&cded
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U.
ELANT SYSTEMS SURVEILLANCE REOUTREMENTS (Continued)
- c. Visual Insoection Accentance Crite a Visual inspections shall verify that: (1) there are no visible indications of damage or impa ed OPERABILITY and (2) attachments to the foundation or supper ng structure are functional, and (3) l f fasteners for attachment o the snubber to the component and to the snubber anenorage ar functional. Snubbers which appear l inoperable as a result visual inspections shall be classified i as unacceptable and m be reclassified acceptable for the purpose of establishing the xt visual inspection interval, provided that: (1) the caus of the rejection is clearly established and remedied for that articular snubber and for other snubbers that j may be genericall susceptible; and (2) the affected snubber is functionally te ed in the as found condition and determined OPERABLE per 5 cification 4.7.5f. For those snubbers common to more than one ystem. the OPERABILITY of such snubbers shall be S considered i assessing the OPERABILITY of each of the related i systems. A eview and evaluation shall be performed and 'I I
d documented 10 justify continueo operatie with an unacceptable f continued operation cannot De justified, the snubber snubDer.
shall De eclared inoperable and tne ACTION requirements shall be met.
- d. Transi.nt Event Inseertien An i specticn shall be performeo of all hydraulic and mechanical snu oers attached to sections of systems that have experienced un xpected, potentially camaging transients as determined from a r iew of operational cata ano a visual inspection of the systems ithin 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for accesstole areas and 6 months for inaccessible areas following sucn an event. In addition to satisfying the visual inspection accettance criteria, freedom of motion of mechanical snuecers shall ce verified using at least one of the following: (1) manually incucea snubber movement; or (2) evaluation of in place snuDeer c1ston setting; or (3) stroking the mechanical snubDer througe its full range of travel.
FERM1 - UNIT 2 3/4 7 17 Amendment No. 84
/
IME 1.7 op
dkoO V /42 l NT SYSTEMS SURVElttANCE RE00fpEMENTS (Continued)
- e. Functiotial . Tests During the first refueling shutdown and at least ce per 18 months thereafter during shutdown, a represent we sample of snubbers shall be tested using one of the fol wing sample plans.
The sample plan shall be selected prior to e test period and cannot be changed during the test period. ne NRC Regional Administrator shall be notified in writ g of the sample plan selected prior to the test period or t sample plan used in the prior test period shall be implement :
- 1) At least 10% of the total of each type of snubber shall be functionally tested either n place or in a bench test. For each snuboer of a type th t does not meet the functional test acceptance r-iteri of Specification 4.7.5f.. an additional 5% of that pe of snubber shall be functionally tested until no more ilures are found or until 11 snubbers of that typ have been functionally tested; or
- 2) A representative s mole of each type of snubber shall be functionally tes d in accordance with Figure 4.7.5 1. "C" is the total nu. er of snubbers of a type found not meeting the acceptance eQuirements of Specification 4.7.5f. The cumulative nu er of snuboers of a type tested is denoted by I
) "N". At the .nd of each day's testing, the new values of
[(o/ "N" and "C" preytous day's total plus current day's increments) shall be plotted on Figure 4.7.5 1. If at any i
time the int plotted falls in the " Reject" region all snubbers f that type snall be functionally tested. If at any tte the coint clotted falls in the " Accept" region, testin of snuocers of that type may De terminated. When the D nt plotted lies in the " Continue Testing" region, ,
addi onal snuocers of that type shall be tested until the poi s falls in tne " Accept" region or the " Reject" region, j or 11 the snubbers of that type have been tested. Testing T e itoment failure curing functional testing may invalidate at day's testing and allow that day's testing to resume new at a later ttme, providing all snubbers tested with the failed eouioment ouring tne oay of equipment failure are retesteo; or
- 3) An initial representative sample of 55 snubbers shall be functionally teste:. fc' each snubber type which does not meet the functional test acceptance criteria, another sample of at least one half tre size of the initial sample shall be tested until tne total numoer tested is eoual to the initial sample size multtolied oy the f actor.1 + C/2. wnere "C" is the numoer of snuceers found which do not meet the '
l functional test acceptance criteria. The results from this sample plan snall De FERMI UNIT 2 3/4 7 lE p ,
V !
l%E 2 9 0F 57
St. Clt hed
,m l
r- PLANT SYSTEMS u.2 SURVEILLANCE REOUTREMENTS (Continued) plotted using an " Accept" line which fo ows the equation N = 55(1 + C/2). Each snubber point ould be plotted as soon as the snubber is tested. If e point plotted f alls on or below the " Accept" line, te ing of that type of snubber may be terminated. If e point plotted falls above the " Accept" line, testing mu continue until the point falls in the " Accept" region r all the snubbers of that type have been tested. 1 The representative samp selected for the functional test sample plans shall be andomly selected from the snubbers of each type and review before beginning the testing. The review shall ensure s far as practical that they are representative of e various configurations, operating environments, ra e of size, and capacity of snubbers of j each type. Snu ers placed in the same locations as ,
snubbers which alled the previous functional test shall be retested at t time of the next functional test but shall not be inclu d in the sample plan. If during the I functional sting, additional sampling is required due to failure of nly one type of snubber, the functional testing results s all be reviewed at the time to determine if e- addition samples should be limited to the type of snubber (x
x f.
which h failed the functional testing.
Funct _nal Test Accectance Criteria The nubDer functional test shall verify that:
1 Activation (restraining action) is achieved within the specifiec range in botn tension and compression;
- 2) Snubber bleed. or release rate where required is present in both tensten and compression, within the speciftea range: l
- 3) for recnanical snubbers, the force required to i intitate or eatntain motion of the snubber is within the specif ted range in both directions of tri, vel; and i 4) For snubbers spectftcally'reoutred not to displace uncer centinuous loa:. the ability of the snubber to withstand 1040 without displacement.
Testing methods may be usec to measure parameters indirectly
. \ or parameters other inan those specified if those results can be correlated to the scecified parameters through establisned trethocs, j
FERMI UNIT 2 3/4 7 19
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suRVE1 LANC grou1REMENTs (continuedi
- g. Functional Test FaWure Analysis An engineering aluation shall be made of each failure to meet the funcWonal test acceptance criteria to det ine th cause of the failure. The results of this evalua on shall used, if a licable, in selecting snubbers to be ested in an q
) effort to etemine the OPERABILITY of other s bers 6 J 1rrespe tve of type which may be subject to e same failure mode.
For e snubbers found inoperable, an en neering evaluation sh 1 be perfomed on the components to ich the inoperable s bbers are attached. The purpose of his engineering valuation shall be to detemine if a components to which the inoperable snubbers are attache were adversely affected by the inoperability of the snub s in order to ensure that
- the component remains capable of eting the designed service.
If any snubber selected for f ctional testing either falls to lock up or fails to move,1. .. frozen-in-place, the cause will be evaluated and if c sed by manufacturer or design deficiency all snubbers o the same type subject to the same defect shall be functio ly tested. This testing requirement shall be independent od the requirements stated in h Specification 4.7.5e. Jor snubbers not meeting the functional /
test acceptance cri ris,
- h. Functional Testin of Renaired and Renlaced Snubbers Snubbers which all the visual inspection or the functional test accepta e criteria shall be repaired or replaced.
Replacement nubbers and snubbers which have repairs whi might affe the functional test result shall be teste to meet the unctional test criteria before installation n.the unit, chanical snubbers shall have met the accept ce criter a subsequent to their most recent service, d the free e of motion test must have been performed w hin 12 nths before being installed in the unit.
- 1. ubber Seal Reelacement Prooram The service life of hydraulic and mechanic snubbers shall be monitored to ensure that the service life s not exceeded between survalliance inspections. The ximum expected service life for various seals, spring , and other critical parts shall be detemined and establi ed based on engineering information and shall be extended or shortened based on monitored test results and failure istory. Critical parts shall be replaced so that the sa sua service 1tfe will not be exceeded during a period when t snubber is required to be OPERABLE. The parts replaceae s shall be documented and the documentation shall be retain d.
FERMI - UNIT 2 3/4 7-20 Amendment No.113 PAGE 30 _0F .57
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TABLE 4.7.5-1 SNUBBER VISUAL INSPECTION INTERVAL NUMBER OF UNACCEpTABL NUBBERS Population Column A Column Column C or Category Extend Interval Repeat I erval Reduce Interval (Notes 1 and 21 (Notes 3 and S) (Notes and 6) (Notes 5 and 6) 1 0 0 1 80 0 0 2 0 1 4 100 150 0 3 8 200 5 13 300 5 12 25 G 36 l 400 8 18 500 12 24 48 750 20 40 78 1000 or gr ter 29 55 109 l l t
I Note 1: The next visual inspection interval for a snubber population or '
category shall be ottermined based upon the previous inspection I interval anc the numoer of unacceptable snubbers found during that I
! interval. Snubeers say be categorized based upon their I
accessibility curing power operation, as accessible or I inaccessible. These categories may be examined separately or jointly. However, the licensee must make and document that
[
decision before any inspection and shall use that decision as the i basis upon which to cetermine tne next inspection interval for that category.
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FERMI UNIT 2 3/4 7 20a Amendment No. 84 kg FAGE 3/ OF 57
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TABLE 4.7.5 1 NOTES (Continued) l l
I Note'2: Interpolation between population or category sizes and e number of unacceptable snubbers is permissible. Use next er integer for the value of the limit for Columns A, B, or f that integer includes a fractional value of unacceptable s pers as determined by interpolation.
Note 3: If the number of unacceptable snubber s equal to or less than the number in Column A, the next i ection interval may be twice the previous interval but not gr er than 48 months.
Note 4: If the number of unacceptab snubbers is equal to or less than the number in Column 8 b rester than the number in Column A, I the next inspection in va shall be the same as the previous !
interval, j
Note 5: If the numoer o unacceptable snubbers is equal to or greater than I I
the number i olumn C, the next inspection interval shall be two thirds the previous interval. However, if the number of I unaccept e snubbers is less than the numoer in Column C but great than the number in Column B. the next interval shall be red d proporttonally by interpolation: that is, the previous i erval shall be reduced by a factor that is one-third of the atto of the difference betweon the number of unacceptable snubbers found during the previous interval and the number in 1
Column 8 to the difference in the numoers in Columns B and C. -
No 6: The provisions of Specification 4.0.2 are applicable for all inspection intervals up to and . including 48 months. ,
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. FERMI - UNIT'2 3/4 8 21 Amendment No. 102 O
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RELocs TED l l
LM REFUELING OPERATIONS
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f3/4.9.4 DECAYJIME LIMITI TION F PERATION
, / .4 The re or shall be s critical for a east 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
APPLICABI : OPERATIO CONDITION 5. ring movement of irradiat uel in the rea pressure v sel.
t W h the reac suberitical r less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, su d all operations nvolving ment of irrad ted fuel in the reactor ssure vessel.
k f
SURV LLANCE RE0VfREMENTS
, 4 The reactor sha be determined to have bee subcritical for t least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifica on of tne cate and time of uberiticality pr r to movement of irradi d fuel in the reactor pres re vessel. )
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FERMI - UNIT 2 3/4 9 6 O
PAGE 3 F 0F 57
DEdE TE D M. ]
E 10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY tIMITING CONDITION FOR OPEPATION 3.10.1 The provisions of Specift' cations 3.6.1.1 3.6.1.3 and 3.9.1 and Table 1.2 may be suspenoed to permit the reactor pre ure vessel closure head and the drywell head to be removed and the primar contain:nent air lock doors to be open when the reactor mode switch is in e Startup position during low power PHYSICS TESTS with THERMAL POWER les than 1% of RATED THERMAL POWEK and reactor coolant temperature less than 20 .
APPLICABillTY: OPERATIONAL CONDITION and S. during low power PHYSICS TESTS.
ACTION:
With THERMAL POWER greater tha or eoual to 1% of RATED THERMAL POWER or with the reactor coolant temoerat e greater than or equal to 200'F, immediately place the reactor mode swi in the Shutdown position.
SURVEfttANCE RE IREMENTS 4.10.1 Th HERMAL POWER and reactor coolant temoerature shall be verified to be within he limits at least once per hour ouring low power PHYSICS TESTS.
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I FERMI. UNIT 2 3/4 10 2 Amenoment No. 67 D FAGE yo QF 57
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t SPECIAL TEET EXCEPTIONS 3/4.10.4 RECIRCULATION LOOPS LIMITING CONDITION FOR ODEpATION 3.10.4 The reautrements of Specifications 3.4.1.1 d 3.4.1.3 that recirculation loops be in operation with matched p o speed may be suspended for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the performance of PHYSICS ESTS, provided that THERMAL POWER does not exceed 5% cf RATED THERMAL POWER jpPLICABILITY: OPERATIONAL CONDITION 2. duri.g PHYSICS TESTS. ,
ACTION:
- a. With the above specified ti limit exceeded, insert all control rods.
- b. With the above specifie THERMAL POWER limit exceeded during PHYSICS TESTS. tmmedia ly place tne reactor mode switch in the Shutoown position.
SURVEfttANCE RE001pfMENTS 4.10.4.1 The time during hich the above soecified reautrement has been O suspended shall be vertf ed to be less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at least once per hour Q during PHYSICS TESTS.
4.10.4.2 THERMAL P0 R shall be determined to be less than 5% of RATED THERMA l, POWER at le st once per hour during PHYSICS TESTS.
FERMI UNIT 2 3/4 10 4 Amenoment No. 53
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FERMI UNIT 2 3/4 10 5 Amendment No. 3, 46 FAuh %1. OF 57
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DE LETE D M.s SPECIAL TEST EXCEPTIONS 3/4.10.6 TRAINING STARTUPS LIMITING CONDITION FOR OPERATION ,
- 3.10.6 The provisions of Specification 3.5.1 m be suspended to permit one RHR subsystem to be aligned in the shutdown co ing mode during training startups provided that the reactor vessel is at pressurtzed. THERMAL POWER is less than or equal to 17. of RATED THERMAL P AER and reactor coolant temperature is less than 200*F.
APPLICABILITY: OPERATIONAL CONDITIO , during training startups.
EllD3:
With the requirements of the aD ve specification not satisfied, immediately place the reactor mode switch n the Shutdown position.
SURVEILLANCE RE001REMENT 4.10.6 The reactor essel shall be verified to be unpressurized and the THERMAL POWER and actor coolant temperature shall be verified to be within j O the limits at les t once per hour during training startups. 4
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FERMI UNIT 2 6 22b Amencment No. 64 l O
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DISCUSSION OF CHANGES RELOCATED AND DELETED CTS INCLUDED IN APPENDIX B TO THE SPLIT REPORT ADMINISTRATIVE A.1 In the conversion of the Fermi 2 current Technical Specifications (CTS) to the proposed plant specific Improved Technical Specifications (ITS), certain wording preferences or conventions are adopted which do not result in technical changes (either actual or interpretational). Editorial changes, reformatting, and revised numbering are adopted to make the ITS consistent with the Boiling Water Reactor (BWR) Standard Technical Specifications NUREG 1433, Rev. 1.
TECHNICAL CHANGES LESS RESTRICTIVE
" Generic" LA.1 CTS 3.1.3.8 requires that the control rod drive housing be verified installed by visual inspection when maintenance has been performed in the control rod drive housing support area prior to reactor operation in Modes 1, 2. or 3. This explicit post maintenance verification is not retained in the ITS. The o requirement for the control rod drive housing support to be installed is included in the physical design requirements, and will be relocated to the Technical Requirements Manual (TRM), )
which requires revisions be controlled by 10 CFR 50.59. Post I maintenance inspections conducted as part of the plant I configuration management control serve the same function as the current Technical Specification requirements. Control of changes to these requirements by Technical Specification amendment is not necessary to provide adequate protection of the public health and safety since there is no change in the design requirement for the CRD housing support. The relocation of this information maintains the consistency with NUREG 1433.
LA.2 Details of the snubber inspection requirements located in the Fermi CTS 3.7.5 are being removed. These details will be defined in the Fermi Inservice Inspection (ISI) Program consistent with i the CTS requirements. The ISI program requirements, and thus the snubber inspection requirements. are controlled by 10 CFR 50.55a, I which requires that applicable ASME code requirements be met l unless relief is granted by the NRC. ISI requiremnts beyond those required by the ASME Code are controlled under the -)'
provisions of 10 CFR 50.59. I FERMI - UNIT 2 1 REVISION 0, 04/03/98 I I
i DISCUSSION OF CHANGES .
RELOCATED AND DELETED CTS INCLUDED IN APPENDIX B TO THE SPLIT REPORT iO V LA.3 CTS 3/4.9.4 Decay Time requires the reactor to be subcritical for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to movement of irradiated fuel. This requirement is being relocated to the UFSAR. Although this LC0 satisfies Criterion 2 (refer to TS Screening Criteria Application and Risk Assessment), the activities necessary prior to commencing movement of irradiated fuel ensure that there will always be 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of subtriticality before movement of any irradiated fuel.
Therefore, these requirements can be adequately defined and controlled in the UFSAR, which requires change control in accordance with 10 CFR 50.59.
TECHNICAL CHANGES HORE RESTRICTIVE H.1 CTS 3/4.10.1. Primary Containment Integrity (during Physics Tests) is deleted. This exception is no longer needed because low power physics tests have been completed. This change represents an additional restriction on plant operations because an allowed exception to the Limiting Conditions for Operation is deleted, n H.2 CTS 3/4.10.4, Recirculation Loops (during Physics Tests and the V Startup Test Program), is deleted. This exception is no longer needed because low power physics tests and the Startup Test Program have been completed. This change represents an additional restriction on plant operations because an allowed exception to the Limiting Conditions for Operation is deleted.
H.3 CTS 3/4.10.6. Training Startups, is deleted. This exception is no longer needed because provisions for training startups are not needed at an operating plant. This change represents an additional restriction on plant operations because an allowed exception to the Limiting Conditions for Operation is deleted.
,i FERMI UNIT 2 2 REVISION 0, 04/03/98
O APPENDIX C GENERIC NO SIGNIFICANT HAZARDS EVALUATION O
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GENERIC NO SIGNIFICANT HAZARDS EVALUATION ADMINISTRATIVE CHANGES (q) ("A" Labeled Comments / Discussions)
Detroit Edison has evaluated each of the proposed Technical Specification changes identified as " Administrative Changes" in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration. The bases for the determination that the proposed changes do not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92. The criteria and the conclusions of the evaluation are presented below.
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed changes consist of reformatting, renumbering, rewording, and elimination of expired provisions of the existing Technical Specifications to establish consistency with Boiling Water Reactor (BWR)
Standard Technical Specifications (STS), NUREG 1433, Rev 1. This reformatting, renumbering, and rewording involves no technical changes to the Fermi 2 Current Technical Specifications (CTSn The changes are
_ administrative because they do not have any effect on the initiators of
(") analyzed events and do not affect any assumptions associated with the mitigation of accident or transient events. Therefore, these changes do not involve any increase in the probability or consequences of any accident previously evaluated.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed changes will not involve any physical changes to plant systems, structures, or components (SSC), or the manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements.
Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
V FERMI - UNIT 2 1 REVISION 0, 04/03/98
GENERIC NO SIGNIFICANT HAZARDS EVALUATION ADMINISTRATIVE CHANGES
("A" Labeled Comments / Discussions)
- 3. Does this change involve a significant reduction in a margin of safety?
The proposed changes will not reduce a margin of safety because the changes identified as Administrative were deliberately limited to those changes that have no impact on any safety analyses assumptions.
Therefore, no question of safety is involved and these changes do not involve any reduction in a margin of safety.
O V
FERMI UNIT 2 2 REVISION 0,'04/03/98 m
b_m_ -
GENERIC NO SIGNIFICANT HAZARDS EVALUATION TECHNICAL CHANGES MORE RESTRICTIVE
("M" Labeled Comments / Discussions)
Detroit Edison has evaluated each of the proposed Technical Specification changes identified as "More Restrictive" in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration. The bases for the determination that the proposed changes do not involve a significant hazards consideration is an evaluation of these . changes against each of the criteria in 10 CFR 50.92. The criteria and the conclusions of the evaluation are presented below.
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed changes provide requirements determined to be more conservative than the existing requirements for operation of the facility. Therefore, these changes establish or maintain adequate assuran;e that components are operable when necessary for the prevention or mittgation of accidents or transients and that plant variables are maintained within limits necessary to satisfy the assumptions for initial conditions in the safety analysis. Therefore, these changes do
( not involve any increase in the probability or consequences of an accident previously evaluated.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed changes will not involve any physical changes to plant systems, structures, or components (SSC). The changes in normal plant operation are consistent with the current safety analysis assumptions.
Therefore, these changes'will not create the possibility of a new or different kind of accident from any accident previously evaluated.
O i i
( [ERMI UNIT 2 3 REVISION 0, 04/03/98 r
GENERIC NO SIGNIFICANT HAZARDS EVALUATION TECHNICAL CHANGES:- MORE RESTRICTIVE
("M" Labeled Comments / Discussions)
- 3. Does this change involve a significant reduction in a margin of safety?
The imposition of more restrictive requirements either has no impact on or increases the margin of plant safety. As provided in the discussion of each of the changes, each change in this category provides additional requirements designed to enhance plant safety. .Each of the changes maintains requirements.within the safety analyses and licensing basis.
Therefore . these changes do not involve a reduction in a margin of safety.
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FERMI UNIT 2. 4 REVISION 0, 04/03/98
GENERIC N0 SIGNIFICANT HAZARDS EVALUATION RELOCATED SPECIFICATIONS-O ("LA" Labeled Comments / Discussions)
Detroit Edison has evaluated each of the proposed Technical Specification changes identified as "Less Restrictive Administrative" in accordance with
~
the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration. The bases for the determination that the proposed changes do not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92. The criteria and the conclusions of the evaluation are presented below.
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed changes remove from the Fermi 2 Technical Specifications items that are informational or implementing details that are adequately and more appropriately controlled within other documents that are ,
maintained by programs defined in 10 CFR 50.54(a), 10 CFR 50.55a 10 CFR 50.59, the ITS ODCH Program, the COLR, and/or the ITS Bases Control Program. Additionally, the proposed changes remove from the Fermi 2 Technical Specifications items that are contained in the Code of Federal Regulations or other regulatory documents and, therefore, do not need to be repeated in the Fermi 2 ITS. The requirements being moved from Fermi 2 Technical Specifications to another controlled document are not being deleted or changed. These requirements are relocated to an appropriate document, program, the COLR, or ITS Bases that is controlled administratively in accordance with programs defined in 10 CFR 50.54(a),
10 CFR 50.55a. 10 CFR 50.59. the ITS ODCH Program, and/or the ITS Bases Control Program. Therefore. these changes will not result in any changes to the requirements specified in the Fermi 2 CTS, but will reduce the level of regulatory control on the identified requirements.
The level of regulatory control has no impact on the probability or the consequences of an accident previously evaluated, therefore, these changes have no impact on the probability or consequences of an accident previously evaluated.
FERMI UNIT 2 5 REVISION 0, 04/03/98
.1 GENERIC N0 SIGNIFICANT HAZARDS EVALUATION RELOCATED SPECIFICATIONS-O ("LA" Labeled Comments / Discussions)
- 2. Doe's the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed changes will not-involve any physical changes to plant' systems, structures, or components (SSC), or the manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements.
Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does this change involve a significant reduction in a margin of safety?
The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being moved from Technical Specifications to another controlled. document (e.g., TRM, UFSAR, or ITS Bases) remain the same c stated in the CTS. Future changes to these requirements will be evaluated in accordance with the requirements of 10 CFR 50.54(a), 10 CFR 50.55a. 10 CFR 50.59, the ITS ODCM Program, or the ITS Bases Control Program, or in the case of regulations, the change process governing changes to the regulations.
Therefore, no reduction in a margin of safety will be permitted.
Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of safety that can be evaluated. However. the proposed changes continue to provide a level of detail that is consistent with the BWR Standard Technical Specification, NUREG 1433, Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail entails no significant reduction ~in the margin of safety.
W
- FERMI UNIT 2- 6 REVISION 0, 04/03/98 i
GENERIC NO SIGNIFICANT HAZARDS EVALUATION p TECHNICAL CHANGES l'.ESS RESTRICTIVE V ("LB" Labeled Comments / Discussions) f l
Detroit Edison has evaluated each of the proposed Technical Specification changes identified as less restrictive "LB" category change in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration. The bases for the determination that the proposed changes do not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92. The criteria and the conclusions of the evaluation are presented below.
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed changes eliminate the requirement that SRs be completed at ,i an interval shorter than the normal Surveillance Frequency just prior to the start of an activity, such as fuel handling, control rod withdrawal or removal, or control rod drive removal. These changes will not result in a significant increase in the probability or consequences of an accident previously evaluated because there is no change to the requirement that the Surveillance be met throughout the time that the associated function is required. The Surveillance continues to be performed at the normal Frequency and the normal Surveillance Frequency has been shown, based on operating experience, to be adequate for assuring that required conditions are established and maintained. This change is consistent with both CTS 4.0.4 and ITS SR 3.0.4 which require I that a Surveillance be performed within the required Frequency prior to entering the applicable Mode or condition. ITS still requires that if any Surveillance has not been performed within this interval or is determined not to meet the Surveillance Requirement during the Applicability, the LCO is not met. and the associated Actions are entered. This ensures the Surveillance Requirements are adequately checked prior to and during these activities.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed changes will not involve any physical changes to plant systems, structures, or components (SSC), or changes in normal plant operation. Therefore, these changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.
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FERMI UNIT 2 7 REVISION 0, 04/03/98 l
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GENERIC NO SIGNIFICANT HAZARDS EVALUATION l TECHNICAL CHANGES LESS RESTRICTIVE
( ("LB" Labeled Comments / Discussions)
- 3. Does this change involve a significant reduction in a margin of safety?
This change does not involve a significant reduction in a margin of safety because there is no change to the requirement that the Surveillance be met throughout the time that the associated function is )
required. Additionally, the Surveillance continues to be performed at its normal Frequency and the normal Surveillance Frequency has been shown, based on operating experience, to be adequate for assuring that conditions are established or that equipment is available and capable of performing its intended safety function. ITS still requires that if any Surveillance has not been performed within this interval, or is determined not to meet the Surveillance Requirement during the Applicability, the LC0 is not met, and the associated Actions are entered. This ensures the Surveillance Requirements are adequately checked prior to and during these activities.
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FERMI UNIT 2- 8 REVISION 0, 04/03/98
l I
GENERIC NO SIGNIFICANT HAZARDS EVALUATION TECHNICAL CHANGES LESS. RESTRICTIVE O
("LC" Labeled Comments / Discussions)
Detroit Edison has evaluated each of the proposed Technical Specification changes identified as lesa restrictive "LC" category change in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration. The bases for the determination that the proposed changes do not involve a significant hazards consideration is an evalua. tion of these changes against each of the criteria in 10 CFR 50.92. The criteria and the conclusions of the evaluation are presented below.
The proposed changes are based on system reliability analysis documented in:
- 1) NED0 30851 P A, " Technical Specification Improvement Analyses for BWR Reactor Protection System," March 1988:
- 2) NED0 30851 P A, Supplement 2. " Technical Specification Improvement Analyses for BWR Isolation Instrumentation Common To RPS And ECCS Instrumentation," July 1986: and
- 3) NEDC 30936P A, "BWR Owner's Group Technical Specification O Improvement Methodology (With Demonstration for BWR ECCS Actuation Instrumentation)," December 1988.
Fermi has confirmed the plant specific application of these NRC reviewed and approved Topical Reports. The proposed increased Completion Times and increased Surveillance Frequencies are consistent with the allowed outage times and Surveillance intervals reviewed and approved in these Topical Reports, These changes are considered appropriate based on the remaining capability to trip, the diversity of the sensors available to provide the trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the function.
FERMI - UNIT 2 9 REVISION 0, 04/03/98 ,
GENERIC N0 SIGNIFICANT HAZARDS EVALUATION TECHNICAL CHANGES LESS RESTRICTIVE
("LC" Labeled Comments / Discussions)
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed change does not involve a significant increase in the probability of an accident previously evaluated because the allowed out of service time and Surveillance intervals are not a parameter or system assumed to be an initiator for any accidents previously analyzed. The proposed change does not involve a significant increase in the consequences of an accident previously evaluated because this change does not further degrade the capability of the instrumentation to perform its required function under these circumstances (only the out of service time or test interval is affected by this change).
Additionally, the increased times allowed will not adversely affect the performance of any other credited equipment. As such, the consequences remain unchanged from those that would apply utilizing the existing CTS requirements.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed changes will not involve any physical changes to plant systems, structures, or components (SSC), or changes in normal plant operation. Therefore, these changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does this change involve a significant reduction in a margin of safety?
The proposed changes are based on system reliability analysis performed by General Electric, reviewed and approved by the NRC, with Fermi confirmation of the plant specific application of these Topical Reports.
These changes are considered appropriate based on the remaining capability to trip, the diversity of the sensors available to provide the trip signals., the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the function. Therefore, there is no significant .
L reduction in a margin of safety.
FERMI UNIT.2 10 REVISION 0, 04/03/98
GENERIC NO SIGNIFICANT HAZARDS EVALUATION RELOCATED SPECIFICATIONS-O. < 'a t 8eieo ce ent 'o4 ces 4 ens)
Detroit Edison has evaluated each of the proposed Technical Specification
. changes identified as "Less Restrictive Removed" in accordance with the criteria specified by 10 CFR 50.92 and 'ias determined that the proposed changes do not involve a significant t :ards consideration. The bases for'the determination that the proposed changes do not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92. The criteria and the conclusions of the evaluation are presented below.
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed changes remove from the Fermi 2 Technical Specifications items that are informational or implementing details that are not required to be under regulatory control (e.g., Technical Specification amendment or 10 CFR 50.59). The requirements being removed from Fermi 2 Technical Specifications are not being deleted or changed at this time, but are simply eliminated from 10 CFR 50.59 or 10 CFR 50.92 change controls; therefore, these changes will not result in any changes to the requirements specified in the Fermi 2 CTS, but will reduce the level of regulatory control on the identified requirements. The level of regulatory control has no impact on the probability or the consequences of an accident previously evaluated, therefore, these changes have no impact on the probability or consequences of an accident previously evaluated.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed changes will not involve any physical changes to plant ,
systems, structures, or components (SSC). or the manner in which these SSC are operated, maintained modified, tested, or inspected. The .f proposed changes will not impose or eliminate any requirements. I Therefore these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
h FERMI' UNIT 2- 11 REVISION 0. 04/03/98 L m
GENERIC NO SIGNIFICANT HAZARDS EVALUATION k RELOCATED SPECIFICATIONS-(j ("LR" Labeled Comments / Discussions)
- 3. Does this change involve a significant reduction in a margin of safety?
The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being removed from Technical Specifications, eliminate regulatory control of changes to this requirement (e.g., Technical Specification amendment or 10 CFR 50.59) that are not necessary to provide adequate protection of the public health and safety since the ITS continue to impose the appropriate requirements and limitations.
Removal of these items from Technical Specifications eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of this admir.istrative process does not have a margin of safety that can be evaluated. However, the proposed changes continue to provide a level of detail that is consistent with the BWR Standard Technical Specification NUREG 1433 Rev 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail entails no significant reduction in the margin of safety.
i FERMI UNIT 2 12 REVISION 0, 04/03/98
GENERICNOSIGNIFICANTHAZARDSEVAldATION m RELOCATED SPECIFICATIONS-lv ) ("R" Labeled Comments / Discussions)
Detroit Edison has evaluated each of the proposed Technical Specification changes identified as " Relocated" in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration. The bases for the determination that the .
proposed changes do not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92. The criteria and the conclusions of the evaluation are presented below.
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed change relocates requirements and surveillances for structures, systems. components or variables which did not meet the criteria for inclusion in Technical Specifications as identified in the Application of Selection Criteria to the Fermi 2 Technical Specifications. The affected structures, systems, components or variables are not assumed to be initiators of analyzed events and are not assumed to mitigate accident or transient events. The requirements and (D
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surveillances for these affected structures, systems, components or variables will be relocated from the Technical Specifications to an appropriate administratively controlled document which will continue to be maintained pursuant to 10 CFR 50.59. In addition, the affected structures, systems, components or variables are addressed in existing surveillance procedures (as applicable) which are also controlled by 10CFR 50.59 and subject to the change control provisions imposed by plant Administrative Procedures, which endorse applicable regulations and standards. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation. The proposed change will not impose or eliminate any requirements and adequate control of existing -
requirements will be maintained. Thus this change does not create the possibility of a new or different kind of accident from any accident (7 previously evaluated.
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FERMI UNIT 2 13 REVISION 0, 04/03/98
GENERIC NO SIGNIFICANT HAZARDS EVALUATION I RELOCATED SPECIFICATIONS
("R" Labeled Comments / Discussions)
- 3. Does this change involve a significant reduction in a margin of safety?
The proposed change will not reduce a margin of safety because it has no impact on any safety analysis assumptions. In addition, the relocated requirements and surveillances for the affected structure, system, component or variable remain the same as the existing Technical Specifications. Since any future changes to these requirements or the surveillance procedures will bevevaluated per the requirements of 10 CFR 50.59, no reduction (significant or insignificant) in a margin of safety will be allowed.
l The existing requirement for NRC review and approval of revisions, in accordance with 10 CFR 50.92, to these details proposed for relocation, does not have a specific margin of safety upon which to evaluate.
However, since the proposed change is consistent with the BWR Standard Technical Specification, NUREG 1433 approved by the NRC Staff, revising the Technical Specifications to reflect the approved level of detail ensures no significant reduction in the margin of safety.
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FERMI UNIT 2 14 REVISION 0, 04/03/98 I