ML20211J197
| ML20211J197 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 09/29/1997 |
| From: | Gipson D DETROIT EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20211J201 | List: |
| References | |
| CON-NRC-97-0089, CON-NRC-97-89 GL-95-10, NUDOCS 9710080026 | |
| Download: ML20211J197 (21) | |
Text
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, September 29,1997 NRC 97 0089 E
- U. S. Nuclear Regulatory Commission 4
Attn.: Document Control Desk i Washington, D. C. 20555-0001
References:
- 1) Ferm12 NRC Docket No. 50 341 d
NRC License No. NPF-43
- 2) NRC Generic Letter 9510 " Relocation of Selected Technical Specincations Requirements Related to Instrumentation," dated December 15,1995 Sub, lect
- Proposed Technical Specincation Change (License Amendment) for Relocation of Selected instrumentation Technical Specifh ations Pursuant to GeneHe Letter (OL195 In accordance with the requirements of 10 CFR 50.90, the Detroit Edison Company
. hereby proposes to ' amend the Fermi 2 plant Operating License NPF-43, Appendix A, Technical Specifications (TS), by relocating the following requirements and their associated Bases from the TS to the Updated Final Safety Analysis Report (UFSAR):
TS 3/4.3.7.2 Seismic Monitoring Instrumentation TS 3/4.3.7.3 Meteorological Monitoring Instrumentation TS 3/4.3.7.7 Traversing in Core Probe System TS 3/4.3.7,8 Chlorine Detection System
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" TS 3/4.3.7.10 - Loose Part Detection System
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o e USNRC NRC.97 0089 Page 2 This proposed license amendment also requests that the TS Index and List of Tables be revised to reDect the relocation of these TS and associated Bases.
These TS will be relocated to the Fermi 2 UFSAR. The continued capability of the Seismic Monitoring Instrumentation, Meteorological Monitoring instrumentation, Traversing in Core Probe System, Chlorine Detection System, and Loose-Part Detection System to perform their functions will be ensured through a controlled change process govemed by 10 CFR 50.59," Changes, Tests and Experiments,"
rather than the present process of 10 CFR 50.90, " Application for Amendment of License or Construction Permit." Accordingly, contingent upon NRC approval, Detroit Edison commits to relocate these Ove TS to the UFSAR (except for Special Report requirements as described in the enclosure) and control their future changes pursaant to the requirements of 10 CFR 50.59. The NRC will receive,in accordance with 10 CFR 50.71(c), future changes to these relocated TS in the form of UFSAR revisions.
These Gye TS are speciHcally identified in NRC Generic Letter 9510 " Relocation of Selected Technical Speci0 cations Requirements Related to Instrumentation,"
dated December 15,1995, as suitable for relocation to licensee controlled documents. Furthennore, the relocation of these TS is consistent with the NRC's
" Final Policy Statement on Technical Speci0 cation improvements for Nuclear Power Reactors" as published in the Federal Register, dated July 22,1993,10 CFR 50.36 as amended July 19,1995 and the NRC's NUREG 1433, Revision 1," Standard Technical SpeciDcations, Generai Electric Plants, BWR/4," Revision 1, dated April 1995. Relocation of these five TS is in support of Fe mi 2's ongoing project to convert to the improved Standard Technical Specifications.
Enclosure 1 provides a description and evaluation of the proposed TS changes.
Enclosure 2 provides an analysis of the issue of significant hazards consideration using the standards of 10 CFR 50.92. Enclosure 3 provides marked op pages of the existing TS to show the proposed changes and a typed version of the afTected Technical Specification pages with the proposed changes incorporated.
Detroit Edison has evaluated the proposed TS changes against the criteria of 10 CFR 50.92 and determined that they do not involve a significant hazards consideration.
The Fermi 2 Onsite Review Committee has reviewed and recommended approval of the proposed TS changes. The Nuclear Safety Review Group has reviewed the j proposed TS changes and concurs with the enclosed determinations. In accordance with 10 CFR 50.91, " Notice for Public Comment; Sta'.e Consultation," Detroit Edison is providing a copy of this letter to the State of Michigan.
Detroit Edison sequests that the NRC approve and issue these changes by I J
USNRC NRC 97 0089 Page 3 hiarch 30,1998 with an Nplementation period of within 60 days following NRC approval.
Should you have any questions or require additional infonnation, please contact hir. Norman K. Peterson, Director Nuclear Licensing, at (313) 586 4258.
Sincerely, 81L *"
Enclosures cc: A.II,lleach O. A. Ilarris hi, J. Jordan A. J. Kugler J. R. Padgett, Supervisor, Electric Operators, hilchigan Public Service Commission e
s J
= USNRC NRC.97 0089 Page 4 I
1, DOUOLAS R OIPSON, do hereby affirm tha: the foregoing statements are based on facts and circumstances which are true and accurate to the best of my knowledge and belief, i
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DOUOLAS R. 6 PSON Senior Vice President On this AY day of 'AZr L -t 1997 before me personally appeared Douglas R. Gipson, being fMy sworna , nd says that he executed the foregoing as his free act and deci.
e!Q Notary Public SHARON A1.u lb W Pubbo, Monroe County.Mi
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Enclosure I to l NRC 97 0089 Page1 i
i ENCLOSUREI FERMI 2 NRC DOCKET NO. 50 341 NRC LICENSE NO. NPF-43 DESCRIPTION AND EVALUATION OF REQUEST TO REVISE TECilNICAL SPECIFICATIONS TO RELOCATE SELECTED INSTRUMENTATION ,
TECilNICAL SPECIFICATIONS PURSUANT TO GENERIC LETTER 95-10
Enclosure I to NRC-97 0089 Page 2 DESCRIPTION AND EVALUATION OF Tile PROPOSJD CilANGES DESCRIPTION:
i The purpose of this proposed license amendment is to modify the Fermi 2 Operating License NPF-43, Appendix A. Technical Specifications (TS) by relocating the following TS and their associated Bases to the Fermi 2 Updated Final Safety Analysis Report (UFSAR):
TS 3/4.3.7.2 Seismic Monitoring Instrumentation TS 3/4.3.7.3 Meteorological Monitoring Instrumentation TS 3/4.3.7.7 - Traversing in Core Probe System TS 3/4.3.7.8 Chlorine Detection System TS 3/4.3.7.10 Loose Part Detection System it is also proposed that the TS Index and List of Tables be revised to reibet the relocation of these TS and their associated Bases.
The proposed relocation of these five instrumentation TS is consistem with the guidance provided in NRC Generic Letter (GL) 95 10, " Relocation of Seected Technical Specifications Requirements Related to Instrumentation," dated l>ccember 15,1995. This line item TS improvement was developed in response to TS amendments proposed by licensees and ongoing NRC TS improvement programs.
The purpose of OL 95 10 is to recommend a means by which to reduce the resources spent by licensees and the NRC stafTin amending requirements related to the above listed instrumentation TS under the 10 CFR 50.90," Application for Amendment of License or Construction Permit" process. Relocating these five TS to the UFSAR will provide for future changes to be made by the Fermi 2 plant staff und.r the controls of 10 CFR 50.59, " Changes, Tests and Experiments."
EVALUATION:
The purpose of the Operating License, Appendix A, Technical Specificationa is to impose those conditions or limitations upon reactor operation necessary to presen e the validity of the results of UFSAR Design Basis Analyses. Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licensees to include TS as pan of the license. The NRC regulatory requirements related to the content of the TS are set forth in 10 CFR 50.36," Technical Specifications." This regulation requires that the TS include items in five specific categories, including:
(1) Safety Limits, Limiting Safety System Settings, and Limiting Control Settings;
. ?
o Enclosure I to NRC 97 0089 Page 3 (2) Limiting Conditions for Operation; (3) Surveillance Requirements; (4) Design Features; and,(5) Administrative Controls. Ilowever,10 CFR 50.36 does not specify the particular requirements to be included in a plant's TS.
The NRC has provided guidance for the content of TS in its " Final Policy Statement on Technical Specifications improvements for Nuclear Power Reactors," dated July 22,1993 (reference: 58 FR 39132)in which the NRC indicated that compliance with the Final Policy Statement satisfies Section 182a of the Atomic Energy Act. In particular, the NRC indicated that certain items may be relocated from the TS to lleensee controlled documents since TS are to be reserved for those matters as to which the imposition of rigid conditions or limitations upon reactor operation is deemed necessary to obylate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety.
The NRC's Final Policy Statement recognized, as had previous statements related to the NRC's TS Improvement Program, that implementation of the policy would result in the relocation of existing TS requirements to licenseesontrolled documents such as the UFS AR. Those items relocated to the UFSAR would, in turn, be controlled in accordance with the requirements of 10 CFR 50.59. This regulation provides criteria to determine when facility or operating changes planned by a licensee require prior NRC approval in the form of a license amendment in order to address any unreviewed safety question.
Four criteria were published by the NRC in the Final Policy Statement. The policy established that any TS requirements which did not meet any of the four criteria could be proposed for relocation to licensee controlled documents such as the UFSAR. These criteria were subsequently incorporated into the regulations by an amendment to 10 CFR 50.36, dated July 19,1995 (reference: 60 FR 36953). The five TS cited above do not meet this criteria for inclusion in the Fermi 2 TS and are, therefore, proposed for relocation. These relocations are also consistent with the improved " Standard Technical Specifications, General Electric Plants, BWR/4,"
NUREG 1433, Revision 1, dated April 1995.
The above TS and their Bases are proposed for relocation to the Fermi 2 UFSAR. In gener? these TS will be incorporated into the UFSAR with the same content they possessed as part of the Operating License, with the exception that the Action statement Special Report requirements for inoperable Seismic Monitoring Instrumentation, Meteorological Monitoring Instrumentation, and Loose-Part Detection System channels will be eliminated. The NRC reporting criteria of 10 CFR 50.72,"Immediate Notification Requirements for Operating Nuclear Reactors," and 10 CFR 50.73," Licensee Event Report Systems" provide appropriate a m _ J
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l
+- Enclosure I to NRC 97 0089 Page 4 l
P requirements for reporting degraded and non confomiing conditions to the NRC.
l This elimination of such Special Reports is an administrative change and consistent L with NUREG 1433.
Each of the five TS proposed for relocation is evaluated below with respect to the four criteria of 10 CFR 50.36(cX2Xil) used in determining wh* ether a particular item ,
is required to be retained in the TS, or may be relocated to other licensee controlled documents. The four criteria that call for retaining a particular item in TS are as -;
follows- i
- 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. ,
As described in the Federal Register notice (Referencei 58 FR 39132) of the NRC's Final Policy Statement, the purpose of this criterion is to ensure that TS
- control those instruments specifically installed to detect excessive Reactor Coolant System leakage. The Federal Register states that this criterion should :
Tot be interpreted to include instrumentation to detect precursors to reactor coolant pressure boundary leakage or instrumentation to identify the source of +
cetual leakage. -
l
- 2. A process variable, design feature, or operating restriction that is en initial -
condition of a Design Basis Accident or Transient Analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
As described in the Final Policy Statement's Federal Register notice, the purpose of this criterion 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
. - (e.g., pressure / temperature limits) needed to preclude unanalyzed accidents and transients.
i
- 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.
4 As described in the Final Policy Statement's Federal Register notice, the purpose I -
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. Encl:sure 1 ta NRC 97 0089 Page 5 of this criterion is to capture only those 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 systems that are necessary for items in the primary success path to successfully function. The primary success path for a panicular mode of operation does not include backup and diverse equipment (e.g., safety valves which are a backup to low temperature overpressure relief valves during cold shutdown).
- 4. A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.
As described in the Final Policy Statement's Federal Register notice, the purpose of this criterion is to retain in TS those requirements that the probabilistic safety assessment or operating experience shows as significant to public health and safety.
TS 3/4.3.7.2 - Seismie Instrumentation:
The Seismic Monitoring Instrumentation is described in Fermi 2 UFSAR Section 3,7.4,"Scismic Instrumentation Program." This instrumentation is provided in accordance with 10 CFR 100," Reactor Site Criteria," Appendix A,Section VI(a)(3),
" Required Seismic Instmmentation," to promptly detennine the magnitude of a seismic event so the effect on those plant features important to safety may be evaluated. This capability is provided to permit colnparison of the measured response to that used in the design basis for the plant. Comparison of such data is used to determine whether the plant can operate safely, and permit timely action as may be appropriate. The Seismic Monitoring Instrumentation has no nuclear safety related function and does not automatically shut down the plant when an earthquake occurs which exceeds a predetemiined intensity.
An evaluation of the Seismic Monitoring Instrumentation with respect to the four criteria of 10 CFR 50.36(c)(2)(ii) follows:
Criterion 1: Seismic Monitoring Instrumentation is not installed instrumentation that is used to detect degradation of the reactor coolant pressure boundary. The instrumentation is not installed for, or capable of, detecting reactor coolant leakage. The NRC's Final Policy Statement and GL 9510 explicitly identify the Seismic Monitoring Instrumentation as an example of controls that are not required to be retained in the TS. This instmmentation does not meet Criterion I of 10 CFR 50.36(c)(2)(li) for inclusion in TS.
A
Enclosure 1 to NRC 97 0089
(
Page 6 Criterion 2: Seismic Monitoring Instrumentation is not a process variable, de6;n feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient Analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier, it is provided to promptly detect and determine the magnitude of a seismic event so that the effect on those plant features, important to safety, may be evaluated. This instrumentation does not meet Criterion 2 of 10 CFR 50.36(c)(2)(ii) for inclusion in TS.
Criterion 3: The Seismic Monitoring Instrumentation is not a component that is part of the primary success path and it does not provide a function or actuation in order to mitigate the consequences of a Design Basis Accident or Transient. This instrumentation does not meet criterion 3 of 10 CFR 50.36(c)(2)(ii) for inclusion in TS.
Criterion 4: The Seismic Monitoring Instrumentation has not been shown to be significant to public health and safety by either operational experience or probabilistic risk assessment. Seismic Monitoring Instrumentation was not included in the scope of the Fermi 2 Individual Plant Examination or the Individual Plant Examination for External Events, nor is it " risk significant" under the Femii 2 Maintenance Rule Program. This instrumentation does not meet Criterion 4 of 10 CFR 50.36 (c)(2)(ii) for inclusion in TS.
TS 3/4.3.7.3 - Meteorological Monitor 8ne Instrumentationt The Meteorological Monitoring Instrumentation is described in Fermi 2 UFSAR Section 2.3.3.2," Operational Meteorological Monitoring System."
The Meteorological Monitoring Instrumentation is used to measure environmental parameters (wind speed, wind direction and air temperature differences) which may affect the distribution of radioactive ef1luents following a release of radioactive material, in 10 CFR 50.47," Emergency Plans," and 10 CFR 50, Appendix E,
" Emergency Planning and Preparedness for Production and Utilization Facilities,"
the NRC requires that licensees provide reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency.
Timely access to accurate local meteorological data is important for estimating potential radiation doses to the public and for determining appropriate piotective measures.
. d
Encl:sure I to '
NRC.97-00119 Page 7 in 10 CFR 50.36a(aX2)," Technical Specifications on Effluents from Nuclear Power Reactors," the NRC requires licensees to submit annual reports specifying the quantity of each of the principle radionuclides released to unrestricted areas in liquid :'
and airborne effluents, and such other information as may be required by the NRC to estimate maximum potential annual doses to the public. A knowledge of ,
meteorological conditions in the vicinity of the plant is important in providing a basis ;
for estimating annual radiation doses to the public from either routine or accidental releases of radioactive materials to the atmosphere.
An evaluation of the Meteorological Monitoring Instrumentation with respect to the .
four criteria of 10 CFR 50.36(cX2Xil) follows:
r Criterion 1: The Meteorological Monitoring Instrumentation is not installed
- instrumentation that is used to detect degradation of the reactor ,
coolant pressure boundary. This instrurentation is not inst lled for, i or capable of, detecting reactor coolant leakage. it is used to assess the need for recommending protective measures following an l accident. This instrumentation does not meet Criterion 1 of 10 CFR 50.36(cX2Xil) for inclusion in TS.
i Criterion 2: The Meteorological Monitoring Instrumentation is not a process .
variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient Analysis that either i assumes the failure of or presents a challenge to the integrity of a fission product barrier, This instrumentation does not meet Criterion i 2 of 10 CFR 50.36(cX2Xil) for inclusion in TS. i Criterion 3: _ The Meteorological Monitoring Instrumentation !s not a component that is part of the prims.ry success path and it does not provide a function or actustion in order to mitigate the consequences of a Design Basis Accident or Transient. This instrumentation does not meet Criterion 3 of 10 CFR 50.36(cX2Xii) for inclusion in TS, Criterion 4: Although important for use in estimating potential radiation doses to the public and for determining appropriate protective measures, the Meteorological Monitoring Instrumentation has not been shown to be significant to public health and safety by either operational experience .
or probabilistic safety assessment. This instrumentation cannot be ,
used to predict, preven', or directly mitigate the consequences of a Design Basis Accident. This instrumentation was not included in the scope of the Fermi 2 Individual Plant Examination or the Individual c -r w- wen -
--erw,--+r-ed , 'm-s--n-+- e, wv.* .-e+-- m- *+ ~ + -v-u-'
, o Enclosure 1 to NRC 97 0089 Page 8 Plant Examination for Extemal Events, nor is it " risk significant" under the Fermi 2 Maintenance Rule Program. This instrumentation does not meet Criterion 4 of 10 CFR 50.36(cX2Xil) for inclusion in TS.
i TS 3/4.3.7.7 - Traveral== In-Core Probe Svytml The Traversing in Core Probe System is described in the Fermi 2 UFSAR, Sections 7.1.2.1.4.6," Traversing in Core Probe System," and 7.6.1.13.8," Traversing in Core Probe System."
The Traversing in Core Probe System provides a signal proportional to the axial neutron flux distribution over the regions of the core where the Local Power Range Monitor System detection assemblies are located. This signal is of high precision to -
allow reliable calibration of the Local Power Range Monitor System. These signals 1 do not provide direct input to the Reactor Protection System, tia isolation actuation instrumentation, or the Emergency Core Cooling System actuation instrumentation.
An evaluation of the Traversing in-Core Probe System with respect to the four criteria of 10 CFR 50.36(cX2Xil) follows:
Criterion 1: The Traversing In Core Probe System is not installed instrumentation that is used to detect degradation of the reactor coolant pressure boundary, it is used as a calibration and measurement tool. This system is not installed foi, or capable of, detecting reactor coolant leakage. This system does not meet Criterion 1 of 10 CFR 50.36(cX2Xil) for inclusion in TS.
Criterion 2: This system is not a process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient Analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Although the core power distributions (measured by the in core probe) constitute ea important initial condition to Design Basis Accidents and, therefore, will continue to be addressed by TS 3/4.2, " Power Distribution Limits," the probe itselfis not an active design feature needed to preclude analyzed accidents or transieras. The Local Power Range Monitor System itself will continue to be required to be operable per TS 3/4.3.1, " Reactor Protection System," and calibrated at least once per 1000 :frective full power hours using the Traversing In Core Probe System under TS Table 4.3.1.1 1," Reactor Protection System
i <
a Encl:sure I to j- NRC 97 0089' t'
Page 9-4 Instrumentation Surveillance Requirements." Therefore, this system does not meet Criterion 2 of 10 CFR 50.36 (cX2Xil) for inclusion in 4
TS.
- i Criterion 3: This system is not part of the primary success path and does not -
provide a function or actuation in order to mitigate the consequences of a Design Basis Accident or Transient. This syst:m does not meet i l Criterion 3 of 10 CFR 50.36(cX2Xii) for inclusion in fS. l Criterion 4: This system has not been shown to be significant to public health and 1
- safety by either operational experience or probabilistic safety ;
assessment. This system was not included in the scope of the Fermi 2 !
Individual Plant Examination or the Individual Plant Examination for i
- External Events, nor is it " risk significant" under the Fermi 2
~
j Maintenance Rule Program. This system does not meet Criterion 4 of 10 CFR 50.36(cX2Xil) for inclusion in TS.
) Relocation of the Traversing in Core Probe System requirements from the TS to the UFSAR does not imply any reduction in its importance in confirming that core . .
power distributions are bounded by safety analysis limits. Changes in the Traversing In-Core Probe System will be evaluated in accordance with 10 CFR 50.59. i TS 3/4.3.7.8 - Chlorine Detection System:
The Chlorine Detection System is described in Fermi 2 UFSAR Sections 6.4.2.3,-
[
" Air Conditioning System," and 6.4.3.4, " Air Conditioning System - Control of Main Control Room Chemical Environment." The Chlorine Detector System is
. consistent with the recommendations of Regulatory Guide 1.95," Protection of Nuclear Power Plant Control Room Operators against an Accidental Chlorine Release,' Revision 1, January 1977. Quick response chlorine detectors are located in the normal air intake to the Control Center Air Conditioning System (CCACS). On detection of chlorine concentrations greater than or equal to 5 ppm, the detectors s-automatically initiate complete isolat_ ion of the main control room. Therefore, the Chlorine Detection System ensures that an accidental chlorine release will be v promptly detected and the protective actions wi!! be automatically initiated to -
provide protection for control room personnel.
7 - An evaluation of the Chlorine Detection System with respect to the four criteria of
,10 CFR 50.36(cX2Xii) follows:
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. Enclosure 1 to NRC 97 0089 j Page 10
- J
,l Criterion 1: The Chlorine Detection System is not installed instrumentation that is - !
4- used to detect degradation of the reactor coolant pressure boundary. !
This system is not installed for, or capable of, detecting reactor ,
coolant leakage. It is used to automatically initiate it.olation of the I CCACS in the event of an accidental release offsite of chlorine. This
[ system does not rueet Criterion 1 of 10 CFR 50.36(cK2)(ii) for inclusion in TS. .
d ;
4 Criterion 2: The Chlorine Detection System is not a process variable, design ;
~
feature, or operating restriction that is an initial condition of a Dt sign Basis Accident or Transient Analysis that either assumes the failure of ,
i or presents a challenge to the integrity of a fission product barrier. )
This system does not meet Criterion 2 of 10 CFR 50.36(cX2Xii) for
- inclusion in TS.
1 i~ Criterion 3: The Chlorine Detection System is not a component that is part of the i
primary success path and it does not provide a function or actuation in order to mitigate the consequences of a Design Basis Accident or Transient. The Control Center Air Conditioning System will continue
- to be required by TS 3/4.7.2," Control Room Emergency Filtration .
System," Surveillance Requirement 4.7.2.1.c.3 to automatically 4
l switch to the chlorine mode of operation upon a chlorine actuation 1
signal. However, this system does not meet Criterion 3 of 10 CFR -
50.36(cX2Xii) for inclusion in TS.
4 Criterion 4: The Chlorine Detection System has not been shown to be significant !
, to the public health and safety by either operational experience or :
i probabilistic safety assessment. This system was not included in the scope of the Fermi 2 Individual Plant Examination or the Individual F
Plant Examination for External Events, nor is it " risk significant" ,
L under the Fami 2 Maintenance Rule Program. This system does not ,
j meet Criterion 4 of 10 CFR 50.36(cX2Xil) for inclusion in TS. ;
- IS 3/4.3.7.10 - Loose-Part Detection System:
E - The Loose Part Detection System is described in UFSAR Section 4,4.6.2, " Loose.
Part Monitoring System." This system is provided in response to the recommendations of Regulatory Guide 1.133," Loose Part Detection Program for the
- - Primary System of Light Water-Cooled Reactors", May 1981. The purpose of this system is to identify the existence of possible loose parts in the Reactor Coolant 4
j.,
. Enclosure I to j NRC.97-0089 l
Pagei1 '
, System. Early detection can provide operators time to take corrective actions and to avoid or mitigate damage to or malfunctions of primary system components.
An evaluation of the Loose Parts Detection System with respect to the four criteria of 10 CFR 50.36(c)(2)(ii) follows:
]
Criterion 1: The Loose-Parts Detection System is not installed instrumentation
, that is used to detect a significant abnormal degradation of the mactor coolant pressure boundary. This system is not installed for, or ,
a capable of, detecting reactor coolant leakage. The NRC's Final Policy statement and GL 95 10 explicitly identify this system as an example of controls that are not required to be retained in the TS.
This system does not meet criterion 1 of 10 CFR 50.36(c)(2)(ii) for inclusion in the TS.
[ Criterion 2: Thh system is not a process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or
! Transient Analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. This system i
does not meet Criterion 2 of 10 CFR 50.36(c)(2)(ii) for inclusion in TS. ,
Criterion 3: This system is not a component that is part of the primary success path and it does not provide a function or actuation in order to
- mitigate the consequences of a Design Basis Accident or Transient.
This system does not meet Criterion 3 of 10 CFR 50.36(c)(2)(ii) for
- inclusion in TS. <
Criterion 4: This system has not been shown to be significant to public health and safety by either operational experience or probabilistic safety assessment. This system was not included in the scope of the Fermi 2 Individual Plant Examination or the Individual Plant Examination for External Events, nor is it " risk significant" under the Fermi 2 Maintenance Rule Program. This system does not meet Criterion 4 of 10 CFR 50.36(c)(2)(ii) for inclusion in TS.
Summam The relocation of TS 3/4.3.7.2,3/4.3.7.3,3/4.3.7.7,3/4.3.7.8, and 3/4.3.7.10 to the o UFSAR will not impact their system's capabilities. The proposed relocation will maintain similar operation, maintenance, testing and system operability controls for 4
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. Enclosure I to NRCo97 0089 Page 12 the systems. Furthennore, any future proposed changes to these systems and relocated TS will be evaluated as required by 10 CFR $0.59.
These TS changes are administrative and consistent with guidance provided by the NRC. These TS do not meet any criteria under the NRC's Final Policy Statement on TS Improvements for retention as a TS and, therefore, qualify for relocation to the UFSAR. The NRC has previously approved TS relocations under this policy, e.g.,
Amendment No.122 to Facility Operating License No. NPF 12. Virgil C. Summer Nuclear Station, dated February 15,1995 (TAC No. M90765); Amendment No. 201 to Facility Operating License No. NPF 3, Davis Desse Nuclear Power S'-tion, dated November 14,1995 (TAC No. M92805); and Amendmer.t Nos.147 ana 86 to Facility Operating Licenses DPR-67 and NPF-16 St. Lucie 1 and 2, dated August 20,1996 (TAC Nos, M93340 and M93341, respectively).
The elimination of the Special Reports requirements for TS being relocated has been previously approved by the NRC in the aforementioned Operating License Amendments for the Davis Besse Nuclear Power Station and the Virgil C. Summer Nuclear Station.
SIGNIFICANT ll A7,ARDS CONSIDERATION:
In accordance with 10 CFR $0.92, Detroit Edison has made a determination that the proposed amendment involves no significant hazards consideration. To make this determination, Detroit Edison has established that operation in accordance with the proposed TS amendment would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or.
- 2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or
- 3. Involve a significant reduction in a margin of safety.
The significant hazards consideration assessment is provided in Enclosure 2.
ENVIRONMENTAL,lMPACT; Detroit Edison has reviewed the proposed TS changes against the criteria of 10 CFR 51.22 for environmental considerations. The proposed changes do not involve a significant hazards consideration, nor significantly change the types or
Enclosure 1 to NitC 9744189 !
Page 13 I
significantly increase the amounts of effluents that rnay be released offsite, nor l
significantly increase individual or cumulative occupational radiation exposures. l Based on the foregoing, Detroit Edison concludes that the proposed TS changes meet {
the criteria provided in 10 CFR Sl.22(c)(9) for a categorical exclusion from the requirements for an Environmental Impact Statement and an Environmental 1 Assessment. I CONCLUSION:
Ilased on the evaluation above: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the NitC's regulations and the proposed amendment will no' be inimical to the common defense and security, or to the health and safety of the pi blic.
Detroit Edison requests that the proposed license amendment be approved by the NRC with implementation to occur within the following 60 days.
Enclosure 2 to NRC 97 0089 Pagei ENCLOSURE 2 FERMI 2 NRC DOCKET NO. 50 341 NRC LICENSE NO. NPF-43 REQUEST TO REVISE TECilNICAL SPECIFICATIONS:
10 CFR 50.92 EVALUATION
- Enclosure 2 to 1 NRC 97 0089 l Page 2 l 10 CFR 50.92 EVALUATION BASIS FOR SIGNIFICANT llAZARDS DETERMINATION The proposed Technical Specincation (TS) changes described in Enclosure 1 do not involve a significant hazards consideration for the following reasons:
- 1. The changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed changes would relocate TS 3/4.3.7.2 - Seismic Monitoring Instrumentation, TS 3/4. 3.7.3 - Meteorological Monitoring Instrumentation, TS 3/4.3.7.7 Traversing in Core Probe System, TS 3/4.3.7.8 - Chlorine Detection System, and TS 3/4.3.7.10 - Loose Part Detection System and their associated Bases to the Fermi 2 Updated Final Safety Analysis Report (UFSAR). They would also delete the special reporting requirements from the aforementioned TS which contain such requirements. The proposed changes would revise the TS Index and List of Tables to reflect the relocation of these TS and associated Bases. The relocated TS changes would be controlled in accordance with the requirements of 10 CFR 50.59.
The proposed changes affect TS that do not meet the NRC's " Final Policy Statement on Technical Specification Improvements for Nuclear Power Reactors" or 10 CFR 50.36(c)(2)(ii) criteria for inclusion in TS. These TS relocations are consistent with NUREG 1433 " Standard Technical Specifications, General Electric Plants, BWR/4," Revision 1, April 1995.
Furthermore, these five TS are specifically identified in NRC Generic Letter 95 10," Relocation of Selected Technical Specifications Requirements Related to Instrumentation," dated December 15,1995, as suitable for relocation to licensee-controlled documents.
The Special Report requirements of TS 3/4.3.7.2, TS 3/4.3.7.3, and TS 3/4.3.7.10 would be deleted as part of their relocation to the UFSAR. The NRC reporting criteria of 10 CFR 50.72,"Immediate Notification Requirements for Operating Nuclear Reactors," and 10 CFR 50.73," Licensee Event Report Systems" provide appropriate requirements for reporting degraded and non-conforming conditions to the NRC.'
These proposed TS changes do not invo;ve a'significant increase in the probability of an accident previously evaluated because no changes are being il
-
- Enclosure 2 to 4
NRC 97 0089 Page 3 made to any accident initiator, No previously analyzed accident scenario is changed, ano initiating conditions and assumptions remain as previously analyzed.
These proposd TS changes do not involve a significant increase in the consequences of an accident previously evaluated because the proposed changes do not affect ac'.ident sequences or assumptions used in evaluating the radiological consequences of an accident. The proposed changes do not alter the source term, containment isolation or allowable radiological releases.
- 2. The changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed changes do noi change the way in which the plant is operated and -;
no new or different failure modes have been dermed for any plant system or component. No limiting single failure has been identified as a result of the -
proposed changes, No new or different types of failures or accident initiators are introduced by the proposed changes.
- 3. The changes do not involve a significant reduction in the margin of safety.
The proposed changes involve instrumentation and systems which are not inputs in the calculation of any safety margin with regard to Technical Specification -
Safety Litrits, Limiting Safety System Settings, Limiting Control Settings or
- Limiting Conditions for Operation, or other previously defined margins for any structure, system, or component.
/
- Enclosure 3 to NRC-97 0089 Page1 ENCLOSURE 3 FERMI 2 NRC DOCKET NO. 50-341 NRC LICENSE NO. NPF43 REQUEST TO REVISE TECilNICAL SPECIFICATIONS TECHNICAL SPECIFICATION CHANGES Attached is a mark up of the existing Technical Specifications, including the proposed changes, and a typed version of the Technical Specifications incorporating the proposed changes.
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