ML20247H570

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Proposed Tech Specs Re Containment Atmosphere Continuous Leak Rate Monitoring
ML20247H570
Person / Time
Site: FitzPatrick Constellation icon.png
Issue date: 05/19/1989
From:
POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20247H494 List:
References
JPTS-89-005, JPTS-89-5, NUDOCS 8905310266
Download: ML20247H570 (7)
v • d • e


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ATTACHMENT I-PROPOSED TECHNICAL SPECIFICATIvW j CRANGES REGARDING CONTAINMENT ATMOSPEERR CONTINUOUS LEAK RATE MONITORING (J PTS-89-00 5 )

New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59 8905310266 89051933 (

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ATTACHMENT II SAFETY EVALUATION FOR PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING CONTAINMENT ATMOSPHERE CONTINUOUS LEAK RATE MONITORING (JPTS-89-00 5 )

i New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT D)cket No. 50-333 DPR-59

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SAFETY EVALUATION.

Page 1 of 3 I. DESCRIPTION OF THE PROPOSED CHANGES The proposed changes to the James A. FitzPatrick Technical Specifications revise Specification 4.7.A.3, " Continuous Leak Rate Monitoring," on pages 176 and 177.

Pace 176 Replace the word, " Monitor" with the word, " Monitoring."

Pace 177

1. Delete the last sentence of Specification 4.7.A.3:

"The monitoring system may be taken out of service for maintenance, but.shall be returned to service as soon as possible."

2. Move the first sentence of Specification 4.7.A.3, "When the primary containment is inerted, it shall be continuously monitored for gross leakage by review of the inerting system makeup requirements." to page 176.

II. PURPOSE OF THE PROPOSED CHANGES These changes remove a misleading reference to a continuous leak rate monitoring system. There is no dedicated monitoring system.

Leakage is monitored by periodically calculating the makeup requirements of the Containment Air Dilution Inerting System and by recording flow integrator readings on the nitrogen makeup trains. A significant increase in the amount of nitrogen required to maintain the differential pressure at the specified 1.7 psid is a direct indication of drywell leakage, vacuum breaker leakage, or low suppression pool level. If abnormal amounts of leakage are noted, an immediate investigation is made.

In addition, review of the drywell to suppression chamber differential pressure instrumentation also provides an indication of drywell integrity and suppression chamber to drywell vacuum breaker status.

This is consistent with the description of the Primary Containment Leakage Monitoring System in FSAR Section 5.2.3.13 and with present plant practices. K -

III. IMPACT OF THE PROPOSED CHANGES These changes are purely administrative in nature. They remove a misleading reference and are consistent with present plant practices. The changes do not involve the modification of any existing equipment, systems, or components; nor do they relax any

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i Attachment II L'c, SAFETY EVALUATION l Page 2 of 3 administrative controls or limitations imposed on existing plant equipment.- The changes do not alter the conclusions of the plant's accident analyses as documented in the FSAR or the NRC ]

staff's SER. Operation of the plant in accordance with the j proposed amendment is not a safety concern. j

- I IV. EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION. l I

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 changes to Specification 4.7.A.3 are purely administrative in nature and remove a misleading reference to a dedicated continuous leak rate monitoring system. There is no dedicated monitoring system.

Containment leakage is determined by periodically calculating the Containment Air Dilution System makeup requirements and by review of the drywell to suppression chamber differential pressure instrumentation. These changes do not involve the modification of any existing equipment, systems, or components; nor do they relax any administrative controls or limitations imposed on existing plant equipment. These changes cannot increase the probability or consequence of a proposed accident previously evaluated.

2. create the possibility of a new or different kind of accident from those previously evaluated. The proposed _

changes are purely administrative in nature. The changes  !

do not alter the conclusions of the plant's accident H analyses as documented in the FSAR or the NRC staff's SER. They do not create any new failure modes; nor do  ;

they place the plant in an unanalyzed condition, j

3. involve a significant reduction in the margin of safety.

The proposed changes improve the consistency of the Technical Specifications and reflect present plant practices. These changes improve the clarity of the Technical Specifications by remove a misleading reference. These changes do not involve a reduction in the margin of safety.

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 these changes and states:

g' , C Attachment II

, SAFETY EVALUATION Page 3 of 3 "A purely administrative. change to a technical specifications: for example, a chhnge to achieve consistency throughout the technical specifications,' correction.of an error, or a change in nomenclature."

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

V. IMPLEMENTATION OF THE PROPOSED CHANGES Implementation of the proposed changes will not impact the ALARA or Fire Protection Programs at FitzPatrick, nor will the changes impact the environment.

VI. CONCLUSION The changes, as proposed, do not constitute an unreviewed safety i 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.

.VII. REFERENCES

1. James A. FitzPatrick Nuclear Power Plant Updated Final Safety Analysis Report, Section 5.2.3.13, " Primary Containment Leakage Monitoring System."

- 2. James A. FitzPatrick Nuclear Power Plant Safety Evaluation Report, dated November 20, 1972 and Supplements.

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