ML20247B977
| ML20247B977 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 07/14/1989 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20247B803 | List: |
| References | |
| NUDOCS 8907240251 | |
| Download: ML20247B977 (6) | |
Text
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l ATTACHMENT I PROPOSED TECHNICAL SPECIFICATION CHANGE REGARDING APRM ROD BL6CK OPERABluTY DURING START-UP (JPTS-81-012) l New York Power Auth erity i
JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 l
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ATTACHMENT 11
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SAFETY EVALUATION FOR l
PROPOSED TECHNICAL SPECIFICATION 1
CHANGE REGARDING APRM ROD BLOCK OPERABILITY DURING START-UP (JPTS-81-012) i New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50 333 DPR-59
e 3
I 1
' 1 SAFETY EVALUATION Page 1 of 3 1.
DESCRIPTION OF THE PROPOSED CHANGE l
The proposed change to the James A. FitzPatrick Technical Specifications revises Note 1 to Table 3.2-3, " Instrumentation that Initiates Control Rod Blocks," on page 73 by:
- 1) Replacing,
"... the APRM and R3M rod blocks need not be operable in start-up mode."
with
"... the HBM rod block need not be operable in start-up mode.", and
- 2) following the above sentence, add "When the reactor is in the start-up mode, the APRM upscale (start-up mode) rod block shall be operable. When the reactor is in the run mode, the APRM upscale (flow biaseri) and APRM downscale rod blocks shall be operable."
11.
PURPOSE OF THE PROPOSED CHANGE The proposed change corrects Note 1 of Table 3.2-3. The present wording of Note 1 states that the APRM (Average Power Range Monitor) rod blocks are not required during start-up. This contradicts both the body of Table 3.2-3 which shows a separate trip level setting for the APRM in the start-up mode and Table 3.1-1, " Reactor Protection System (SCRAM) Instrumentation Requirement," which requires 2 APRM instrument channels to be operable in the start-up mode.
Note 1 is revised to specify that the APRM " upscale (start-up mode)" rod blocks are required to be operable when the reactor is in the start-up mode. As a further clarification the other two APRM rod block functions, the " upscale (flow biased)" and the "downscale," are identified as being required when the reactor is in the run mode.
This revision to Table 3.2-3 is purely editorial in nature and it makes Note 1 consistent with the plant's design basis, the FSAR description, and the balance of the Technical Specifications.
lil.
IMPACT OF THE PROPOSED CHANGE This change does not alter any setpoints; nor does it alter any administrative controls, procedures or other limitations imposed on tF.e plant. This change is purely editorial in nature and does not involve the modification of any ;xisting structures, equipment, systems, or components. The change can not alter the conclusions of the plant's accident analyses as documented in the FSAR or the NRC staff's SER.
1 SAFETY EVALUATION Page 2 of 3 IV.
EVALUATION OF SIGNIFICANT HAZARDS 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 change is administrative in nature and would not:
1.
Involve a significant increase in the probability or consequences of an accident previously evaluated. The change corrects Note 1 to Table 3.2-3 to clearly identify the l
APRM rod blocks required during start-up operat5ns. This change is purely editorial in nature and does not involve the modification of any oxisting safety structures, systems, or components. There are no changes in probability or consequences to 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 proposed change is purely editorial in nature with no resulting impacts on plant parameters or operational procedures. This change does not create any new failure modes or accident scenarios.
3.
Involve a significant reduction in the margin of safety. The proposed change does not alter any established trip level settings nor does it revise the start-up procedure. This change does not relax any administrative controls or limitations imposed on the existing plant equipment; nor does it involve the modification of any system or component. This change is purely editorial in nature and does not involve a reduction in the margin of safety.
In the April C,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 this change and states:
"A pure!y administrative change to technical specifications: for example, a change to achieve consistency throughout the technicel specifications, correction of an error, or a change in nomenclature."
The proposed change can be classified as not likely to involve significant hazards considerations, i
since the change is purely editorial in nature and does not involve hardware changes to the plant's safety related structures, systems, r omponents. The proposed chenge improves the clarity of the Technical Specifications.
V.
IMPLEMENTATION OF THE PROPOSED CHANGE implementation of the proposed change will not impact the ALARA Program at FitzPatrick, nor wil the change impact the environment.
_ _ = -.
.*- 1 SAFETY EVALUATION Page 3 of 3 1
VI.
CONCLUCION I
This change, as proposed, does not ( anstitute an unreviewed safety question as defined in 10 CFR 50.59. That is,it:
J 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; L.
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.
Vll.
REFERENCES
[
1.
James A. FitzPatrick Nuclear Power Plant Updated Final Safety Analysis Report, Section 7.5.
2.~
James A. FitzPatrick Nuclear Power Plant Safety Evaluation Report (SER), dated November 20,1972 and Supplements.
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