ML20247K010
| ML20247K010 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 05/19/1989 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20247J992 | List: |
| References | |
| JPTS-89-008, JPTS-89-009, JPTS-89-010, JPTS-89-10, JPTS-89-8, JPTS-89-9, NUDOCS 8906010184 | |
| Download: ML20247K010 (11) | |
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ATTACWIENT ' I PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING EDITORIAL ERRORS (JPTB-89-008, JPTS-89-009, JPTS-89-010)
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ATTACHMENT II SAFETY EVALUATION FOR PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING EDITORIAL ERRORS (JPTS-89-008, JPTB-89-00 9, JPTB-89-010)
New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59
, . Attachment II SAFETY EVALUATION Page 1 of 5 l
l I. DESCRIPTION OF THE PROPOSED CHANGES The proposed changes to the James A. FitzPatrick Technical I Specific '; ions revise Tables 4.1-1, 4.1-2, and 4.2-1 on pages 45, 46, 47, and 78. These changes are as follows: !
Reactor Low-Low-Low Water Level On page 78, replace " Reactor Low-Low Water Level" with
" Reactor Low-Low-Low Water Level".
Reactor Pressure Permissive on pages 45 and 47, delete line identified as " Reactor Pressure Permissive".
TIP System Traverse on page 46, replace " Trip System Traverse" with "TIP System Traverse".
II. PURPOSE OF THE PROPOSED CHANGES
'The vurpose of the proposed changes is to correct editorial and typographical errors inadvertently introduced as a result of the amendment process.
Reactor Low-Low-Low Water Level This change corrects Table 4.2-1 which was omitted from Amendment 103's revision of the reactor water level setpoint associated with the Primary Containment Isolation System (PCIS). The change specifically addresses the PCIS calibration and testing specification for the low water level instrument channels.
Amendment 103 reduced the possibility of spurious Main Steam Isolation Valve (MSIV) closure due to variations in water level, reduced challenges to the Safety Relief Valves (SRV's), and reduced suppression pool heatup. This was accomplished by changing the reactor water level setpoint from Low-Low to Low-Low-Low.
This correction to Table 4.2-1 properly identifies the instrument channels requiring calibration and tests as established by Amendment 103 and makes the Technical Specifications consistent with modifications to the plant as approved by the NRC.
l
. . Attachment II L SAFETY EVALUATION Page 2 of 5 Reactor Pressure Permissive This set of changes deletes references to the reactor pressure switches from Tables 4.1-1 and 4.1-2 regarding instrument j
. functional test and instrument calibration, respectively. These four pressure switches, 02-3PS-51 (A-D), were removed from Technical Specification Table 3.1-1 as part of Amendment 122. l They were originally installed to establish the pressure setpoint 1 below which a bypass function is activated in the SCRAM and isolation logics. This installation was based on instability observed in an early European Boiling Water Reactor during its start-up. Subsequent start-up tests at Browns Ferry, a BWR 4 reactor similar to FitzPatrick,'showed that the instability observed in the European reactor did not exist in a BWR 4.
Therefore these switches were unnecessary and subsequently removed.
The two tables (i.e. - 4.1-1 and 4.1-2) retained instrument testing and calibration check frequencies for these removed pressure switches. These two changes should have been submitted as part of JPTS-88-010, which was subsequently approved by the NRC as Amendment 122. They are now submitted here as editorial changes to make the Technical Specifications consistent with previously approved Amendments.
TIP. System Traverse The third change corrects a typographical error, introduced in Amendment 89, which affects the description of the calibration method used for the Local Power Range Monitoring System (LPRMS) in Table 4.1-2. A review of the original Technical Specification and' prior amendments up through Amendment 75 indicates that the proper calibration nethod is the "TIP" system traverse which is consistent with the Technical Specification Basis. This change is purely editorial in nature and makes Table 4.1-2 consistent with previously approved Amendments.
III. IMPACT OF THE PROPOSED CHANGES Reactor Low-Low-Low Water Level The proposed correction to Table 4.2-1 properly identifies the instrument channels requiring calibrations and tests as established by Amendment 103. This change is therefore editorial in nature and improves the consistency of the Technical Specifications with no impact to the plant's administration or operation.
Attachment II SAFETY EVALUATION Page 3 of 5 Reactor Pressure Permissive The proposed deletions to Tables 4.1-1 and 4.1-2 regard instrument functional test and instrument calibration, respectively, for instruments which are no longer required.
The deletion of the two test requirements in the Technical Specifications will therefore have no impact on plant administration or operation since they are editorial in nature and improve the consistency of the Technical Specifications.
TIP System Traverse This proposed correction to Table 4.1-2 properly identifies the calibration methodology for the LPRMS. There are no impacts to the plant's administration or operation as a result of this modification to the Technical Specifications. This change is purely typographical in nature and makes Table 4.1-2 consistent ,
with previously approved Amendments.
IV. EVALUATION OF SIGNIFICANT HAZA.RDS CONSIDERATION !
Operation of the James A. FitzPatrick Nuclear Power Plant in accordance with this proposed amendment would not involve a significant hazards consideration, as defined in 10 CFR 50.92, since the proposed changes are purely editorial / typographical in nature and would not:
- 1. involve a significant increase in the probability or consequences of an accident previously evaluated. The changes correct errors of the Technical Specifications in regard to calibration and testing requirements for the PCIS, reactor pressure permissive, and proper calibration methodology for the LPRMS. They do not involve the modification of any existing equipment, systems or 1 components nor do they alter the conclusions of the plant's accident analyses as documented in the FSAR or the NRC staff's SER.
- 2. create the possibility of a new or different kind of accident from those previously evaluated. The changes are editorial in nature, update the Technical Specifications and improve consistency.
- 3. involve a significant reduction in the margin of safety.
The proposed changes do not alter any established instrument calibration / test frequencies, methods of calibration, or instrument channel designations. The changes do not relax any administrative controls or
~
Attachment II SAFETY EVALUATION Page 4 of 5 limitations imposed on existing plant equipment-nor do they involve the modification of any system or component.
In the April 6, 1983 Federal Register (48FR14870), the NRC published examples of license amendments that are not likely to involve a significant hazards consideration.' Example (i) from this Federal Register is applicable to these changes and states:
"A purely administrative change to technical specifications: for example, a change to achieve consistency throughout the technical specifications, correction of an error, or a change in nomenclature."
V. IMPLEMENTATION OF THE PROPOSED CHANGES Implementation of the proposed changes will not impact the ALARA Program at FitzPatrick, nor will thx changes impact the environment.
VI. CONCLUSION The changes, as proposed, do not constitute an unreviewed safety question as defined in 10 CFR 50.59. That is, they:
- a. will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report;
- b. will not increase the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report;
- c. will not reduce the margin of safety as defined in the basis for any technical specification; and
- d. involves no significant hazards consideration, as defined in 10 CFR 50.92.
l l
VII. REFERENCEB
- 1. James A. FitzPatrick Nuclear Power Plant Updated Final Safety Analysis Report, Section 7.
l
- 2. James A. FitzPatrick Nuclear Power Plant Safety l Evaluation Report (SER), dated November 20, 1972 and Supplements.
l l - - -- _ - - - - - - _ _ _ _ - - - - _ _ _ _ _ _ _ _ _
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Attachment II 1 SAFETY EVALUATION Page.5 of 5
- 3. NRC ' letter - (JAF-83-3 00) , J . D. Hegner to J.P. Bayne, dated August ~26, 1983, transmits Amendment 75
- 4. NRC letter ([[::JAF-85-153|JAF-85-153]]), H.I. Abelson to J.P. Bayne, dated July 7, 1985, transmits Amendment 89
- 5. NRC letter ([[::JAF-87-001|JAF-87-001]]) , H.I. Abelson to J.C. Brons, dated December 19, 1986, transmits Amendment 103
- 6. NRC letter ([[::JAF-89-074|JAF-89-074]]), D.E. LaBarge to J.C. Brons, dated February 7, 1989, transmits Amendment 122 i
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