ML20117H018
| ML20117H018 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/29/1996 |
| From: | James Knubel GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20117H024 | List: |
| References | |
| RTR-NUREG-1430 C311-95-2479, NUDOCS 9609090086 | |
| Download: ML20117H018 (18) | |
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l GPU Nuclear, Inc.
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Route 441 South NUCLEAR Post office Bon *80 Middletown, PA 17057-0480 l
Tel 717-944 7621 l
(717) 948-8005 August 29, 1996 C311-95-2479 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
Dear Sir:
Subject:
Three Mile Island Nuclear Station, Unit 1 (TMI-1)
Operating License No. DPR 50
- Docket No. 50-289 Technical Specification Change Request No. 257 Changes to Incorporate Improvements from the Revised B&W Standard Technical Specifications (STS), NUREG-1430 In accordance with 10 CFR 50.4(b)(1), enclosed is TMI-l Technical Specification Change Request (TSCR) No. 257. The purpose of this TSCR is to incorporate certain improvements from the Standard Technical Specifications (STS) for B&W Plants (NUREG-1430).
Using the standards in 10 CFR 50.92, GPU Nuclear has concluded that these proposed changes do not constitute a significant hazards consideration, as described in the enclosed analysis performed in accordance with 10 CFR 50.91(a)(1). Also enclosed is a Certificate of Service for this request, certifying service to the chief executives of the township and county in which the j
facilities are located, as well as the designated official of the Commonwealth of Pennsylvama, i
Bureau of Radiation Protection.
Sincerely, J. Knubel Vice President and Director, TMI i
JSS b
t 9609090086 960829 l
DR ADOCK 05000287 p
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i knclosures: 1) TMI-1 TSCR 257 Safety Evaluation and No Significant Hazards Consideration
- 2) TMI-1 Technical Specifications Revised Pages
- 3) Certificate of Service for TMI-I TSCR 257 cc:
RegionI Administrator i
TMI-l Senior Project Manager TMI Senior Resident Inspector i
C311-95-2479 il age 3 of 14 i
i METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER AND LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY i
GPU NUCLEAR, Inc.
Three Mile Island Nuclear Station, Unit i Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request No. 257 COMMONWEALTH OF PENNSYLVANIA
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COUNTY OF DAUPHIN
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l This Technical Specification Change Request is submitted in support of Licensee's request to change Appendix A to Operating License No. DPR-50 for Three Mile Island Nuclear Station, Unit 1. As part of this request, proposed replacement pages for Appendix A are also included.
All statements contained in this submittal have been reviewed, and all such statements made and matters set forth therein are true and correct to the best of my knowledge.
4 ce P' resident and Director, TMI Sworn and subscribed before me this Mday of
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dotary Public
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TMI-I TSCR 257 Safety Evaluation and No Significant Hazards Consideration i
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C311-95-2479 Page 1 of 14 I.
TECHNICAL SPECIFICATION CH ANGE REOUEST (TSCR) NO. 257 GPU Nuclear requests that the following change be made to the existing TMI-l Technical Specifications (TS):
A. Replacement pages:
3-10,3-18c,3-18g,3-44,3-45,3-46,3-63,4-5a,4-9,4-54, 4-56,4-59, 5-1, 5-2, 5-4 B.
Deleted pages:
3-1 1, 3-18h, 4-57, 4-60, 4-61, 4-62, 4-63, 4-64, 4-65, 4-66, 4-67, 5-3 II.
R_EASON FOR CHANGE The purpose of this TSCR is to incorporate certain improvements from the Revised Standard Technical Specifications (RSTS) for B&W Plants (NUREG-1430) that would allow deletion or relaxation of selected limiting conditions for operation (LCOs) that do not meet the criteria for technical specifications as set forth in 10 CFR 50.36(c)(2)(ii) and are not reflected in the Revised Standard Technical Specifications (RSTS) for B&W plants delineated in NUREG 1430. The requirements of the deleted LCOs are contained in licensee controlled documents.
This TSCR also proposes to delete the Surveillance Requirements that are related to the LCOs proposed for removal from Technical Specifications. As with the LCOs, the activities required by the Surveillance Requirements are contained in licensee controlled documents.
In addition, this TSCR proposes to delete from the TMI-l Technical Specifications design features specifications that are contained in other licensing-basis documents, such as the FS AR and the ODCM, and are not contained in the RSTS for B&W plants delineated in NUREG 1430.
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Page 2 of 14
'The requested changes are summarized as follows:
A.
Limiting condition for operation 3.1.5, Chemistry and Surveillance Requirement Item 1(d), Table 4.1.3, Minimum Sampling Frequency:
Deletion of limiting condition for operation (LCO) 3.1.5 and all subsections and surveillance requirement (SR) item 1(d) of Table 4.1.3. Limiting condition of operation 3.1.5 specifies the permissible concentration of oxygen, chloride and fluoride in the Reactor Coolant System and surveillance requirement item 1(d) of Table 4.1.3 specifies the required frequency for sampling the RCS for the concentration of oxygen, chloride and fluoride.
B.
Limiting condition for operation 3.1.11 and Surveillance Requirement 4.16:
Reactor Intemals Vent Valves Deletion oflimiting condition for operation 3.1.11 and surveillance requirement 4.16 in their entirety. Limiting condition for operation 3.1.11 and surveillance requirement 4.16 require verification during each refueling shutdown that the Reactor Internals Vent Valves are not stuck open.
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Limiting condition for operation 3.1.13 and Surveillance Requirement 4.11:
i Reactor Coolant System Vents Deletion oflimiting condition for operation 3.1.13 and surveillance requirement 4.11 in their entirety. Limiting condition for operation 3.1.13 requires the Reactor Coolant System Vent flow paths to be operable when the plant is critical and surveillance requirement 4.11 requires the vent paths be demonstrated operable once per refueling
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cycle.
I D.
Limiting condition for operation 3.8.1 Radiation Monitors in the Reactor Building Refueling Area and the Fuel Handling Building Spent Fuel Storage Area and Surveillance Requirement item 28, Table 4.1-1, Instrument Surveillance Requirements:
Deletion oflimiting condition for operation 3.8.1 and surveillance requirement items 28 (a), (b), and (c), table 4.1-1 in their entirety. Limiting condition for operation 3.8.1 requires radiation levels in the reactor building and the spent fuel storage area be monitored using either installed monitors or portable survey equipment. Surveillance requirement items 28 (a), (b) and (c) table 4.1-1 specify the instrument surveillance frequency for the installed radiation monitoring equipment.
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C311-95-2479 Page 3 of 14 E.
Limiting condition for operation 3.8.5:
Communication during Changes in Core Geometry Deletion in its entirety requirements for direct communication between the control room and the refueling personnel in the Reactor Building whenever changes in core geometry are taking place.
- F.
Limiting condition for operation 3.10 and Surveillance Requirement 4.13 Scaled Source Leak Testing and Inventory of Miscellaneous Radioactive Sources Deletion of limiting condition for operation 3.10 and surveillance requirement 4.13 in their entirety. Limiting condition for operation and surveillance requirement 4.13 require periodic leak testing and the maintenance of an inventory of miscellaneous radioactive materials sources.
G.
Limiting condition for operation 3,16 and Surveillance Requirement 4.17:
Shock Suppressors (Snubbers)
Deletion oflimiting condition for operation 3.16 and the related surveillance requirement 4.17 in their entirety. Limiting condition for operation 3.16 specifies operability requirements for snubbers.
H.
Design Feature 5.1:
Effluent Release Locations Deletion of a portion of Section 5.1.1 that describes the locations of gaseous effluent release points and liquid effluent outfalls (as tabularized on page 5-10), and the meteorologicM tower location (designated as " weather tower" on the figure).
I.
Design Feature 5.2 Containment Deletion of Section 5.2 in its entirety. Section 5.2 describes design fe'atures of the containment system that are described elsewhere in the technical specifications, the -
UFS AR and other licensee controlled documents.
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Design Features 5.3.1.2 through 5.3.2.3 Reactor Core and Reactor Coolant System i
Deletion of subsections 5.3.1.2 through 5.3.2.3 which describe the reactor core and the reactor coolant system.
C311-95-2479 Page 4 of 14 III.
S AFETY EVALUATION JUSTIFYING CHANGE
Background
The NRC has amended 10 CFR 50.36(c)(2)(ii) for the purpose of allowing licensees to improve Technical Specifications by simplifying them to place emphasis on two classes of technical matters; those that are related to the prevention of accidents and those that are related to the mitigation of accidents. The NRC has identified four criteria against which a Technical Specification limiting condition for operation (LCO) can be compared. Those criteria are:
Criterion 1 Installed instrumentation that is used to detect, and indicce in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
Criterion 2 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.
Criterion 3 A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basir Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
Criterion 4 A structure, syF' m, or component which operating experience or probabilistic safety assessment hu shown to be significant to public health and safety.
If a current Technical Specification limiting condition for operation does not meet any of the above criteria, it may be proposed for deletion ifit is contained in other licensee 3
controlled documents.
This TSCR proposes the deletion oflimiting conditions of operation that are currently contained in the TMI-l Technical Specifications that do not meet the criteria set forth in 10 CFR 50.36 (c)(2)(ii) and that are not contained in NUREG 1430, Revised Standard Technical Specifications for B&W Plants but that are contained in licensee controlled documents. The evaluation of the deletion of each LCO specifies why the I in this context, licensee controlled documents are those documents for which changes thereto must be evaluated in accordance with 10 CFR 50.59 prior to implementation. Examples oflicensee controlled documents are site procedures and the Updated Final Safety Analysis Repon.
,0 C311-95-2479 Page 5 of 14 current LCO does not meet any of the 4 criteria in 10 CFR 50.36 (c)(2)(ii), states that the LCO is not contained in NUREG 1430, RSTS and specifies what licensee controlled document contains the requirements / activities addressed by the LCO.
This TSCR proposes the deletion of the LCO for the reactor vesselinternal vent valves because it is an LCO that is not contained in the RSTS of NUREG 1430 and is part of a licensee-controlled document.
In addition, the TSCR proposes the deletion of the surveillance requirements related to the LCOs proposed for deletion. These surveillance requirements are proposed for deletion from the Technical Specifications because they are not justified in the absence of a supporting LCO.
The TSCR also proposes for deletion Technical Specification design features. The design features proposed for deletion are addressed in other licensing basis documents and are not contained in NUREG 1430, Revised Standard Technical Specifications for B&W Plants. The evaluations for the deletion of the design features state which licensing basis documents contain the design feature proposed for deletion.
The changes are evaluated as follows:
Limitine Condition for Operation 3.1.5 and Surveillance Reauirement Item 1(d) of Table 4.1-3. Minimum Samoline Freauency The current TMI-I Technical Specifications limiting condition for operation (LCO) 3.1.5 specifies acceptable concentrations of oxygen, chloride and fluoride in the primary reactor coolant.
Maintaining the concentration in the Reactor Coolant of oxygen, chloride or fluoride as specified in LCO 3.1.5 is intended to protect the reactor coolant system against potential stress corrosion attack. Maintenance of the concentration of oxygen, chloride and fluoride in the RCS does not involve surveillance of plant parameters to prevent an accident nor is it directly involved in the mitigation of an accident. LCO 3.1.5 therefore does not meet any of the requirements of Criterion 1,2,3 or 4 in 10 CFR 50.36(c)(2)(ii) for retention in Technical Specifications.
l There is no RSTS in NUREG 1430 conceming the concentration of oxygen, chloride l
or fluoride in the reactor coolant system.
The monitoring of the concentration of oxygen, chloride and fluoride in the reactor coolant system and the required plant responses to elevated concentrations is addressed in TMI-l Chemistry procedure N1800.2, Chemistry Limits and Requirements.
l LCO 3.1.5 can be deleted from the TMI-l Technical Specifications because it does not meet the criteria of 10 CFR 50.36(c)(2)(ii) for a technical specification, there is no corresponding RSTS in NUREG 1430 and the maintenance of the concentrations of
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C311-95-2479 l
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oxygen, chloride and fluoridein reactor coolant within acceptable limits are addressed in controlled plant procedures.
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The deletion of Surveillance Requirement 1(d) of Table 4.1-3 is justified because its -
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related LCO, LCO 3.1.5 is not justified to be a Technical Specification and is to be i
deleted.
j Limitine Condition for Oneration 3.1.11 and Survain-e Reauirement 4.16 i
l The current TMI-l Technical Specifications limiting condition for operation 3.1.11 i
and surveillance requirement 4.16 require verification during refueling shutdowns that each reactor internal vent valve is not stuck in the open position and that each valve I
continues to exhibit freedom of movement. LCO 3.1.11 is intended to ensure, to the i
extent practical that the reactor internal vent valves are able to function as designed i
during a large break LOCA to ensure adequate core cooling.
l The reactor intemal vent valves are a component that may be called upon to ensure the mitigation of a large-break LOCA, a design basis accident. In this context, the reactor internal vent valves are components which are in the primary success path designed to actuate to mitigate a large-break LOCA. However, a LCO for the reactor intemal vent valves is not in the RSTS contained in NUREG 1430.
Maintaining LCO 3.1.11 as a Technical Specification is inappropriate because it is in essence a surveillance requirement located in Section 3 of the Technical Specifications as a limiting condition of operation. 'Ihere is no associated action statement or bases.
Per 10 CFR 50.36, limiting conditions for operation are to be the lowest functional capability or performance level of equipment required for safe operation of the plant or equipment. When an LCO can not be met, the reactor must be shut down or remedial action taken until the LCO can be met. This directly implies that for something to be an LCO, its condition / status must be observable during the time when a plant is in operation.
The position of the reactor internal vent valves are not monitored during reactor operation because they are located inside the Reactor Vessel. Therefore, it can not be j
determined during the reactor operation if the internal vent valves have the required freedom of motion. This can only be verified during an outage when the reactor is shut down.
The inspection of the reactor intemal vent valves during reactor outages is addressed i
in TMI-l inservice test program and in TMI-l Surveillance Procedure 1301-10.1, Reactor Vessel Internal Vent Valve Inspection and Exercise.
l The deletion of LCO 3.1.11 is justified because it is not contained in the RSTS i
of NUREG 1430 and the surveillance requirement to periodically inspect the reactor intemal vent valves is addressed in another license-basis document and in a controlled plant procedure.
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Page 7 of 14 he deletion of Surveillance Requirement 4.16 is justified because its related LCO, LCO 3.1.11 is not justified to be a Technical Specification and is to be deleted.
Limitine Condition for Oneration 3.1.13 and Surveillance Reauirement 4.11 The current TMI-l Technical Specifications limiting condidon for operation 3.1.13 requires that sufficient reactor coolant system (RCS) vent flow paths are operable to vent non-condensable gases to mitigate accidents beyond the design basis for the plant.
The RCS vents are not instrumentation and therefore do not meet Criterion 1 of 10 CFR 50.36(c)(2)(ii). The RCS vents are not a design feature that is an initial condition of a design basis accident and therefore do not meet Criterion 2 of 10 CFR 50.36(c)(2)(ii). The RCS vents can be used to help restore natural circulation conditions following an event in which natural circulation was lost due to non-condensable gas collection. However, design basis events do not generate sufficient non-condensable gasses to block natural circulation. RCS vents are, therefore, not a system or component that is part of the primary success path and do not function to mitigate a design basis accident. Hus, the RCS vents do not meet Criterion 3 of 10 CFR 50.36(c)(2)(ii). The RCS vents have not been shown in operating experience or the TMI-l PRA to be significant to the public health and safety and therefore LCO 3.1.13 does not meet criterion 4 of 10 CFR 50.36(c)(2)(ii) here is no technical specification that addresses the RCS vent valves in the RSTS contained in NUREG 1430.
The surveillance and operation of RCS vent valves do not meet the criteria for a Technical Specification limiting condition for operation. They are addressed in controlled plant procedures. The positioning of the RCS Vent Valves is addressed in TMI-l Operations Procedure 1103-2, Fill and Vent of the RCS. The in-service testing of the valves that comprise the RCS vent system is addressed in TMI-1 Surveillance Procedure 1300-3R, IST of Valves During Shutdown and Remote Indication Check.
The deletion of limiting condition for operation 3.11.1 from the TMI-l Technical Specifications is justified because the LCO does not meet criteria 1 through 4 of 10 CFR 50.36(c)(2)(ii), is not contained in the RSTS of NUREG 1430 and is addressed in controlled plant procedures.The deletion of Surveillance Requireinent 4.11 is justified because its related LCO,3.1.13 is not justified to be a Technical Specification and is to be deleted.
m C311-95-2479 Page 8 of 14 Limitine Condition for Operation 3.8.1 and Surveillance Reauirement 4.1.
Table 4.1-1. Items 28 a. b and c l
The current TMI-l Technical Specifications limiting condition for operation 3.8.1 and Surveillance Requirement 4.1, Table 4.1-1, Items 28 a, b, and c require radiation monitors RM-G6 (Fuel Handling Bridge #1 Aux), RM-G7 (Fuel Handling Bridge #2 Main) and RM-G9 (Fuel Handling Bridge - FH Building) to be operable during fuel loading and refueling operations.
The intent of LCO 3.8.1 is to provide a means for the early detection of a radiation release that would occur coincident with fuel handling operations. However, RM-G6, J
RM-G7 and RM-G9 are not interlocked to any plant equipment that is relied upon to mitigate the consequences of such an accident.
The radiation monitors RM-G6, RM-G7 and RM-G9 are not instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of I
the reactor coolant pressure boundary. Therefore, RM-G6, RM-G7 and RM-G9 do not meet Criterion 1 of 10 CFR 50.36(c)(2)(ii).
Radiation monitors RM-G6, RM-G7 and RM-G9 are not a process variable or an operating restriction. They are not a design feature that is an initial condition of a design basis accident. RM-G6 and RM-G7 monitor the radiation levels on the fuel j
handling bridges in the Containment Building. There is no interlock between RM-G6 and RM-G7 and the Reactor Building Purge or Containment isolation. Containment Isolation is initiated in response to a high radiation alarm on a different radiation monitor, RM-A9, the Reactor Building Purge Exhaust radiation monitor. RM-G9 is a monitor on the fuel handling bridge in the Fuel Handling Building and its purpose is to detect an accidental criticality. An accidental criticality in the Fuel Handling Building is not a design basis accident. Therefore, RM-G6, RM-G7 and RM-G9 do not meet Criterion 2 of 10 CFR 50.36(c)(2)(ii)
Radiation monitors RM-G6, RM-G7 and RM-G9 are not a structure or a system.
They are not components that are part of the primary success path and which function to mitigate a design basis accident or transient that either assume the failure of or presents a challenge to the integrity of a fission product barrier. RM-G6, RM-G7 and RM-G9 are not the radiation moni* ors which are relied upon to mitigate a fuel handling design basis accident. RM-A9 shuts down the Reactor Building purge in response to high radiation levels in Containment and RM-A4 shuts down the Fuel l
Handling Building Exhaust ventilation in the Fuel Handling Building. Therefore, RM-l G6, RM-G7 and RM-G9 do not meet Criterion 3 of 10 CFR 50.36(c)(2)(ii).
RM-G6, RM-G7 and RM-G9 are not components that operating experience or probabilistic safety assessment has shown to be important to public health and safety.
The three radiation monitors are not relied upon to mitigate the release of radioactive l
materialinto the environment. Therefore, LCO 3.8.1 does not meet Criterion 4 of 10 CFR 50.36(c)(2)(ii).
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C311-95-2479 Page 9 of 14 The requirement to use local survey equipment in lieu of RM-G6, RM-G7 and/or RM-G9 does not meet Criterion 1,2,3 or 4 of 10 CFR 50.36(c)(2)(ii) because the three installed radiation monitors do not meet the criterion.
There is no corresponding technical specification in the RSTS contained in NUREG 1430 regarding radiation monitors that are not part of the primary success path in mitigating the consequences of an accident..
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The requirements for the operability and calibration of RM-G6, RM-G7 and RM-G9 are addressed in Refueling Procedure 1505-1, Fuel and Control Component Shuffles and TMI-l Surveillance Procedure 1302-3.1, RMS Calibration.
The deletion of LCO 3.8.1 and Surveillance Requirement 4.1, Table 4.1-1, Items 28 a, b, and c from the TMI-l Technical Specifications is justified because LCO 3.8.1 does not meet criteria 1 through 4 of 10 CFR 50.36(c)(2)(ii). In addition, there is no corresponding technical specification in the RSTS contained in NUREG 1430.
Finally, the requirements regarding the operability and surveillance of RM-G6, RM-G7 and RM-G9 are addressed in controlled plant documents.
The deletion of Surveillance Requirements 4.1, Table 4.1-1, Items 28 a, b and c are justified because the related LCO, LCO 3.8.1 is not justified to be a Technical Specification and is to be deleted.
Limitine Condition for Operation 3.8.5 The current TMI-l Technical Specifications limiting condition for operation 3.8.5 requires direct communication between the Control Room and the refueling personnel in the Reactor Building whenever changes in core geometry are taking place.
The requirement for direct communication between the Control Room and the refueling personnel in the Reactor Building whenever changes in core geometry are taking place does not meet Criterion 1 in 10 CFR 50.36(c)(2)(ii) because the existence of direct communication is not used to detect and indicate, in the Control Room, a significant abnormal degradation of the reactor coolant pressure boundary, because whenever changes in the core geometry are taking place, the reactor coolant pressure boundary is not intact.
The requirement for direct communication between the Control Roo'm and the refueling personnel in the Reactor Building whenever changes in core geometry are taking place does not meet Criterion 2 of 10 CFR 50.36(c)(2)(ii) because the requirement is not a process variable or a design feature. The presence of direct communication is not an assumption or initial condition of a design basis accident or transient analysis that assumes the failure of or presents a challenge to the integrity of a fission product barrier.
The requirement for direct communication between the Control Room and the refueling personnel in the Reactor Building whenever changes in core geometry are
e C311-95-2479 Page 10 of 14 taking place does not meet Criterion 3 because the requirement is not a structure, system or component.
The requirement for direct con.munication between the Control Room and refueling personnel in the Reactor Buildmg whenever changes in core geometry are taking place has not been shown by operating experience or probabilistic safety assessment to be lmportant to public safety and health.
There is no technical specification in the RSTS contained in NUREG 1430 that requires direct communication between the Control Room and the refueling personnel in the Reactor Building during refueling.
The requirement for direct communication between the Control Room and refueling personnel in the Reactor Building is addressed in TMI-l Refueling Procedure 1505-1, Fuel and Control Component Shuffles.
The deletion of LCO 3.8.5 is justified because it does not meet the criteria 1 through 4 of 10 CFR 50.36(c)(2)(ii), is not contained in the RSTS of NUREG 1430 and is addressed in controlled plant documents.
i Limitine Condition for Operation 3.10 and Surveillance Reauirement 4.13 Limiting condition for operation 3.10 requires the maintenance of a complete inventon of licensed radioactive materials at all times and specifies periodic leak testing for sealed sources. TMI-l Technical Specification Surveillance Requirement 4.13 specifies the acceptance criteria for allowable leakage from a sealed source.
The inventory of licensed radioactive material and the leak testing of sealed sources required by LCO 3.10 is intended to control relatively small amounts of radioactivity i
as compared to that contained within the core. Therefore LCO 3.10 does not meet criteria 1,2,3 or 4 of 10 CFR 50.36(c)(2)(ii).
There is no technical specification in the RSTS contained in NUREG 1430 regarding the inventory of licensed radioactive material or the leak testing of sealed sources.
The requirements for the inventoy of licensed radioactive material are addressed in TM1-1 Administrative Procedure 6610-ADM-4420.01, Radioactive Source Accountability System and the leak testing of sealed sources is addressed in TMI-1 Surveillance Procedure 1301-7.2, Radioactive Source or Fission Detector Leak Test.
The deletion of LCO 3.10 and Surveillance Requirement 4.13 arejustified because LCO 3.10 does not meet criteria 1 through 4 of 10 CFR 50.36(c)(2)(ii), there is no 1
corresponding technical specification in the RSTS in NUREG 1430 and because the inventory of licensed radioactive material and the leak testing of sealed sources is addressed in controlled plant procedures d
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C311-95-2479 Page 11 of 14 l
The deletion of Surveillance Requirement is justified because the related LCO, LCO' 3.10 is not justified to be a Technical Specification and is to be deleted.
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Limitine Condition of Operation 3.16 and Surveillance Reauirement 4.17 The TMI-1 Technical Specifications limiting condition for operation 3.16 requires l
safety related snubbers to be operable whenever the system protected by the snubber is operable.
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Snubbers are not installed instmmentation used to detect degradation of the RCS pressure boundary therefore Criterion 1 is not satisfied. Snhbbers are not process variables that are initial condition assumptions in a design basis accident, therefore, Criterion 2 is not satisfied. Snubbers are components that are assumed to function to mitigate the effects of a seismic event and a loss of cooling accident (LOCA).
Technical Specification 3.16 and the related surveillance requirement 'SR) 4.17 define snubber operability in terms of the inspection requirements in SR 4.1 i. The requirements in SR 4.17 vary depending on thei accessibility of the snubber, area radiation levels and the mode of the plant. In addition, snubber (s) may be inoperable if an engineering evaluation determines that the affected system is still sufficiently protected. Therefore, snubbers do not meet Criterion 3. Snubbers have not been shown by PRA or operating experience to be significant to public health and safety therefore LCO 3.16 does not meet Criterion 4.
There is no technical specification in the RSTS contained in NUREG 1430 regarding snubber operability.
Snubber surveillance requirements for TMI-l are addressed in TMI-I Surveillance Procedures 1301-9.9, Hydraulic Snubber Visual Inspection and 1303-9.9, Hydraulic Snubber Functional Test and Seal Replacement In summary, the deletion of LCO 3.16 isjustified because LCO 3.1.6 does not meet criterion I through 4 of 10 CFR 50.36(c)(2)(ii), is not contained in the RSTS of NUREG 1430 and is addressed in controlled plant procedures.
The deletion of Surveillance Requirement 4.17 is justified because its related LCO, LCO 3.16 is not justified to be a Technical Specification and is to be deleted.
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C311-95-2479 Page 12 of 14 Desien Feature 5.1.1 The current TMI-1 Technical Specifications contain a description of the TMI-l Site in 4
Specification 5.1.1. The proposed amendment deletes a portion of Specification 5.1.1 that describes the gaseous and liquid effluent release points and the location of the meteorological tower. In addition, the proposed amendment deletes the associated Figure 5-3 and the tables on page 5-10.
1 There is no design feature regarding the gaseous and liquid effluent release points or the location of the meteorological tower in the RSTS contained in NUREG 1430 The location of the gaseous and liquid effluent release points and the location of the meteorological tower are addressed in the Off-Site Dose Calculation Manual (ODCM) 4
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for Three Mile Island Units 1 and 2, a licensing basis document.
The deletion of the description of the locations of the gaseous and liquid effluent j
release points and the location of the meteorological tower from Specification 5.1 is justified because there is no corresponding design feature in the RSTS of NUREG 4
1430 and they are described in another licensing basis document, the ODCM.
i Desien Feature 5.2 The current TMI-l Technical Specifications contain Specification 5.2, Containment, i
that describes on a summary level the TMI-1 Containment Building. Specification 5.2 mentions the structural properties, systems and components that comprise the i
Containment Building and how they function together to provide a barrier between 4
the Reactor Building atmosphere and the environment. However, the Specification does not address the performance standards or surveillance tests for the structure, system or components.
The design features addressed in Technical Specification 5.2 are discussed in equal and greater detail in other licensing basis documents. Technical Specifications 3.6, 3.19,4.4.1,4.4.2, and 4.4.4 specifically discuss the performance standards and j
surveillance tests for those structures, systems and components that function together to provide a barrier between the Reactor Building atmosphere and the environment.
Specification 5.2 is also redundant to the description of the Containment in Chapter 5 of the TMI-l UFSAR.
i There is no design feature in the RSTS of NUREG 1430 that address the Containment Building.
The deletion of Specification 5.2 is justified because it is redundant to other licensing basis documents and there is no corresponding design feature in the RSTS of NUREG 1430..
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Page 13 of 14 Demien Feature 5.3.1.2 throneh 5.3.1.6 i
i The current TMl-1 Technical Specifications Design Features 5.3.1.2 through 5.3.1.6 -
j describe the TMI-l reactor core, fuel and control rod assemblies. The proposed amendment deletes all of the information except one sentence in Specification 5.3.1.6 which limits the maximum enrichment to a nominal 5.0 weight percent of U n 2
'Ihe deletion of Specifications 5.3.1.2 through 5.3.1.5 and the revision of Specification -
5.3.1.6 is justified because the sections proposed for deletion are repetitive of information presented in the Tif.1-1 UFSAR and the Specifications 5.3.1.2 through 5.3.1.6 reference Chapter 3 of the UFSAR for detailed information. The Specifications to be deleted do not impose requirements or restrictions on TMI-l that j
are not contained in other licensing basis documents such as the COLR and the i
The Specifications to be deleted are not contained in the RSTS of NUREG 1430.
l The deletion of Technical Specifications Design Features 5.3.1.2 through 5.3.1.5 and the modification of Design Feature 5.3.1.6 is justified because they are redundant to information and requirements contained in other licensing basis documents and they are not contained in the RSTS of NUREG 1430.
C311-95-2479 j
Page 14 of 14
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IV.
NO SIGNIFICANT HAZARDS CONSIDERATION GPU Nuclear has determined that this Technical Specification Change Request involves no significant hazards consideration as defined in 10 CFR 50.92 because:
- 1. Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability of occurrence or the consequences of an accident j
previously evaluated. The proposed amendment deletes limiting conditions for operation l
(LCOs) from the TMI-l Technical Specifications that are no longer required to be addressed in Technical Specifications per 10 CFR 50.36(c)(2)(ii). The proposed l
amendment deletes Surveillance Requirements from the TMI-l Technical Specifications l
that are related to the LCOs to be deleted. These items are addressed in licensee controlled documents. Certain design feature specifications are also to be deleted consistent with the i
RSTS for B&W plants. The proposed changes do not modify the operation, limits or controls of systems, structures or components relied upon to prevent or mitigate the consequences or accidents previously evaluated. Also, the reliability of systems and
?
components relied upon to prevent or mitigate the consequences of accidents previously
- i evaluated is not degraded by the proposed changes. Therefore, this change does not involve a significant increase in the probability of occurrence or the consequences of an accident previously evaluated.
- 2. Operation of the faciliy in accordance with the proposed amendment would not create the
{
possibility of a new or different kind of accident from any accident previously evaluated l
because no new failure modes are created by the proposed changes.
l
- 3. Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety because the proposed amendment does not change any operating limits for reactor operation.
j V.
IMPLEMENTATION It is requested that the amendment authorizing this change become effective upon issuance.
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