ML20245A078

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Proposed Tech Spec Table 3.2-6, Surveillance Instrumentation to Correct Total Number of Reactor Pressure Channels.Safety Evaluation for Proposed Change Encl
ML20245A078
Person / Time
Site: FitzPatrick Constellation icon.png
Issue date: 06/14/1989
From:
POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20245A063 List:
References
NUDOCS 8906210061
Download: ML20245A078 (7)
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Text

j ATTACHMENT I PROPOSED TECHNICAL SPECIFICATION CHANGE j REGARDING CLARIFICATION OF THE NUMBER OF i REACTOR PRESSURE CHANNELS (JPTS-89-011)

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New York Power Authority j JAMES A. FITZPATRICK. NUCLEAR POWER PLANT l Docket No. 50-333 j

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ATTACHMENT II l

SAFETY EVALUATION FOR PROPOSED TECHNICAL SPECIFICATION CHANGE REGARDING CLARIFICATION OF THE NUMBER OF REACTOR PRESSURE CHANNELS (J PTB-8 9-011) l New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333

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Attachment II SAFETY EVALUATION Page 1 of 4 I. DESCRIPTION OF THE PROPOSED CHANGE The proposed change to the James A. FitzPatrick Technical' Specifications corrects the total number of reactor. pressure channels shown on Table 3.2-6, " Surveillance Instrumentation,"

on page 76. The column entitled "No. of' Channels Provided by Design" is revised to indicate that there are'3 reactor pressure instrument channels instezd of the 5 currently noted. That is, change the "5" to a "3".

II. PURPOSE OF THE PROPOSED CHANGE Specification 3.2.F and Table 3.2-6, " Surveillance Instrumentation", specify that two reactor pressure channels having a range of 0-1200 psig are required to be operable as a S condition for plant operation. It also states that one of these channels is used for feedwater control. The proposed change 1 to the Technical Specifications corrects Table 3.2-6 to reflect )

the total number of instrument channels, available by design, to ] '

meet these requirements.

The actual plant configuration consists of three independent channels; A, B and C. Each channel includes a reactor pressure sensor, a transmitter, and a control room indicator. Two f 6

recorders (one narrow-range and one wide-range) are also provided. These recorders can display input from any two of the three pressure sensors. Either channel A or B can be used for feedwater control. .

This change is purely administrative in nature and makes the Technical Specifications consistent with the as-built plant configuration as described in FSAR Section 7.8.5.4. It corrects an error introduced in Amendment 48 to the Technical  ;

Specifications. This amendment inadvertently-identified the '

total number of recorders and indicators as the number of channels available by design.

l III. IMPACT OF THE PROPOSED CHANGE T..a proposed change to the Technical Specifications is purely administrative in nature. It corrects Table 3.2-6,

" Surveillance Instrumentation," to accurately reflect the as-built configuration of the plant. The proposed change does not involve modification of any existing equipment, systems, or components; nor does it relax any administrative controls or limitations imposed on existing plant equipment.

The change does not affect the existing limitations on surveillance information circuits. That is, Table 3.2-6 still requires a minimum of two redundant reactor pressure channels to be operable. This is consistent with the licensing design

Attachment II SAFETY EVALUATION Page 2 of 4 basis for the reactor vessel monitoring system (See Response to Question 7.2, Reference 2, which identified two indicators i (06PI90A and 06PI90B) and a recorder (06PR97) as the control l room information circuits having the specified range of 0-1200 psig.)

Operation of the plant in accordance with the proposed amendment is not a safety concern. The conclusions of the plant's accident analyses as documented in the FSAR or the NRC staff's SER are not altered by this change to the Technical Specifications.

IV. EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Operation of the James A. FitzPatrick Nuclear Power Plant in accordance with the proposed amendment would not involve a significant hazards consideration as defined in 10 CFR 50.92, since it would not:

1. involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed change is purely administrative in nature and corrects an error introduced in Amendment 48 to the Technical Specifications. There are no changes to setpoints, safety limits, surveillance requirements, or lilaiting conditions for operation. The change does not impact previously evaluated accidents; nor does it affect safe plant operations.
2. create the possibility of a new or different kind of accident from those previously evaluated. The proposed change is purely administrative in nature and corrects the Technical Specifications. The change does not involve modification to any of the plant's systems, equipment, or components; nor does it allow the plant to operate in an unanalyzed condition.
3. involve a significant reduction in the margin of safety.

The proposed change is purely administrative in nature. The change corrects the number of instrument channels which are available (by design) to satisfy the requirements of having two operable reactor pressure instrument channels. The change does not involve any plant modifications, nor does it affect the FiAR information regarding reactor vessel instrumentation. The number of instrument channels associated with monitoring reactor pressure is not reduced by this proposed amendment.

Attachment II SAFETY EVALUATION Page 3 of 4 In the April 6, 1983 Federal Register (48FR14870), 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 this change and states:

j "A purely administrative change to technical I specifications: for example, a change to achieve consistency throughout the technical specifications, correction of an error, or a change in nomenclature."

The proposed change can be classified as not likely to involve significant hazards considerations, since the change is purely administrative in nature and does not involve hardware changes nor any changes to the plant's safety related structures, systems, or components. The proposed change is designed to improve the quality of the Technical Specifications.

V. IMPLEMENTATION OF THE PROPOSED CHANGE l

Implementation of the proposed change will not impact the ALARA or Fire Protection Programs at FitzPatrick,.nor will the change 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 sa fety 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 Attachment II SAFETY EVALUATION Page 4 of 4 1 l

VII. REFERENCES

1. James A. FitzPatrick Nuclear Power Plant Updated Final Safety Analysis Report, Sections 7.8 and 7.10.
2. James A. FitzPatrick Nuclear Power Plant Final Safety Analysis Report, Response to AEC Questions of 11/29/71, Response to Question 7.2
3. USAEC " Safety Evaluation of the James A. FitzPatrick Nuclear '

Power Plant" (SER), dated November 20, 1972.

4. USAEC " Supplement 1 to the Safety Evaluation of the James A. FitzPatrick Nuclear Power Plant" (SER), dated _ February 1, 1973.

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5. USAEC_" Supplement 2 to the Safety Evaluation of the James I A. FitzPatrick Nuclear Power Plant" (SER), dated October 4, 1974.
6. USNRC Letter, T. A. Ippolito to G. T. Berry (NYPA), dated January 23, 1980, issuing Amendment 48 to the Fitzpatrick Technical Specifications.

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