ML20247K505
| ML20247K505 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 05/24/1989 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20247K474 | List: |
| References | |
| JPTS-89-004, JPTS-89-4, NUDOCS 8906010313 | |
| Download: ML20247K505 (7) | |
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ATTACHMENT I TO JPN-89-XXX PROPOSED TECHNICAL SPECIFICATION CHANGE REGARDING CLARIFICATION OF CONTAINMENT DRYWELL SUMP FLOW INTEGRATOR (JPTS-8 9-0 0 4 )
New York Power Authority JAMES A.
FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59 8906010313 890524 PDR ADOCK 05000333 PDC p
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ATTACHMENT II TO JPN-89-XXX l-RAFETY EVALUATION FOR PROPOSED TECHNICAL SPECIFICATION CHANGE REGARDING CLARIFICATION OF CONTAINMENT DRYWELL SUMP FLOW INTEGRATOR (JPTS-89-004) l l
New York Power Authority JAMES A.
FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59
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Attachment II l
SAFETY EVALUATION Page 1 of 4 I. DESCRIPTION OF THE PROPOSED CHANGE The proposed change'to the James A. FitzPatrick Technical Specifications clarifies the description of the flow integrators associated with the containment drywell sumps by revising Bases Section 3.2 on page 59 as follows:
Replace the sentence:
"The alarm unit in each integrator is set to annunciate before the values specified in Specification 3.6.D are exceeded."
with "The leak rate is calculated by dividing the integrated volume pumped out of the sumps by the time between sump pump operations.
The resultant leak rate value, which is expressed in gallons per minute, is compared to the acceptance criterion specified in Specification 3.6.D."
II. PURPOSE OF THE PROPOSED CHANGE The change clarifies Bases Section 3.2 regarding the description of the flow integrators associated with the containment drywell sumps.
In referring to the flow integrators, Section 3.2 (page
- 59) erroneously states "The alarm' unit in each integrator is set to annunciate before the values specified in Section 3.6.D are exceeded." (emphasis added).
(The values of Section 3.6.D are the maximum permissible reactor coolant leakage rates.)
The flow integrators do not have an alarm unit nor do they annunciate when the leak rate limits are exceeded.
The leak rate.is calculated by pumping out the sumps, and dividing the total gallons of leakage pumped out of the sump, as noted on the flow integrators, by the time between sump pump operations.
The resultant number is the average leak rate expressed in gallons per minute.
Control room operators also monitor the leak rate by comparin s trend lines on the sump level strip chart recorders to a predrawn line on the face of the recorder.
This change is purely administrative in nature and makes the Technical Specifications Bases consistent with the as-built plant configuration and the as-licensed configuration as described in FSAR Section 4.10.
2 Attachment II SAFETY EVALUATION Page 2 of 4 l
i III. IMPACT OF THE PROPOSED CHANGE The proposed change to the Technical Specifications is purely administrative in nature.
It clarifies the Bases sec. tion to accurately reflect the as-built and as-licensed 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 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 proposed amendment is not considered a safety concern.
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 clarifies the Bases of the Technical Specifications.
There are no changes to setpoints, safety limits, surveillance requirements, or limiting 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 is intended to clarify and improve the quality of 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 and clarifies the Bases description of the containment sump flow integrators.
The proposed change does not involve any plant modifications, nor does it affect the FSAR information regarding reactor coolant pressure boundary (RCPB) leakage detection systems.
The margin of safety associated with monitoring and detecting leakage from the RCPB is not impacted by this proposed amendment.
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Attachment II-BAFETY 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:
"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."
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 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 change, as proposed, does not constitute an unreviewed safety question as defined in 10 CFR 50.59.
That is, it:
- 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.
0 Attachment II SAFETY EVALUATION Page 4 of 4 l
VII. REFERENCES 1
- 1. James A.~ FitzPatrick Nuclear Power. Plant Final Safety Analysis Report,-Section 4.10.
- 2..USAEC " Safety. Evaluation of the James A.
FitzPatrick Nuclear Power Plant" (SER), dated November 20, 1972.
- 3. USAEC " Supplement-1 to the Safety Evaluation of the James A.
FitzPatrick Nuclear Power. Plant" (SER), dated February 1, 1973.
- 4. USAEC " Supplement 2 to the Safety Evaluation of the James A.
FitzPatrick Nuclear Power Plant" (SER), dated October 4, 1974.
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