JPN-89-020, Proposed Tech Specs Re Pertubation of Reactor Water Level Following Instrument Functional Tests
| ML20245D525 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 04/24/1989 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20245D516 | List: |
| References | |
| JPN-89-020, JPN-89-20, JPTS-86-024, JPTS-86-24, NUDOCS 8905010032 | |
| Download: ML20245D525 (8) | |
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ATTACHMENT I JPN-89-020 i
PROPOSED TECHNICAL SPECIFICATION CHANGES BEGARDING PERTUBATION OF REACTOR WATER LEVEL FOLLOWING INSTRUMENT FUNCTIONAL TESTS (JPTB-8 6-0 2 4 )
j New York Power Authority JAMES A.
FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59 8905010032 890424 1
PDR ADOCK 05000333
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ATTACHMENT II-JPN-89-020 i
SAFETY EVALUATION FOR PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING PERTUBATION OF REACTOR WATER LEVEL FOLLOWING INSTRUMENT FUNCTIONAL TESTS - (JPTB-86-02 4 )
4 1
New York Power Authority JAMES A.
FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59
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- 1 Attachment II SAFETY EVALUATION Page 1 of 4 I.
DESCRIPTION OF THE PROPOSED CHANGES The proposed changes to the James A.
FitzPatrick Technical Specifications revise Technical Specification 4.1, Table 4.1-1 on pages 44 and 45a.
The changes delete the requirement to perturb the reactor vessel water level following the monthly functional test on reactor water level scram instruments.
Textual changes are as follows:
Pace 44; Table 4.1-1 Delete "(5)" following " Reactor Low Level" Pace 45a; Table 4.1-1 Delete text associated with Note 5 and replace with " Deleted" II. PURPOSE OF THE PROPOSED CHANGES The proposed Technical Specification changes remove the requirement to perturb the reactor vessel water level as part of the monthly functional test for reactor water level scram instruments.
Perturbing the reactor water level is considered an unnecessary operational inconvenience and is superfluous to the existing channel check requirements. Removing the pertubation requirement is consistent with the recent Staff activities on reducing testing at power (Reference 1).
Since an excessive testing requirement is being deleted, the probability of an accidental plant transient caused by the performance of this surveillance test is decreased.
In addition, the proposed change promotes consistency with the requirements of the Standard Technical Specifications (NUREG-0123).
A similar change was approved for Browns Ferry Units 1,2 and 3 on September 19, 1984.
This change was Amendment 113, 107 and 81 for Units 1,2 and j respectively.
The NRC safety evaluation for the Browns Ferry Technical Specification Amendments noted that the level pertubation was not a regulatory requirement, was an operational inconvenience and was not required by Standard Technical Specifications.
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Attachment II SAFETY EVALUATION Page 2 of 4 III. IMPACT OF THE PROPOSED CHANGES The design basis for the Reactor Protection System (RPS) is in conformance with IEEE-279 " Criteria for Protection Systems for Nuclear Power Generating Stations" as noted in FSAR Section 7.2.3 and Technical Specification Bases Section 3.1.
IEEE-279-1971 Subsection 4.9 requires the capability for checking the operational availability of each system input sensor during reactor operation.
The means for accomplishing this are various and may include any of the following:
- 1) perturbing the monitored variable;
- 2) introducing and varying, as appropriate, a substitute input to the sensor of the same nature as the measured variable; or
- 3) cross-checking between channels that bear a known relationship to each other (channel check).
Deletion of the pertubation requirement does not degrade the RPS design basis because each of the reactor water level sensors are being cross-checked with each other on a daily basis.
This daily channel check is required by Technical Specification 4.1, Table 4.1-1, note (8).
The proposed Technical Specification change to delete the requirements in Table 4.1-1 to perturb the reactor water level after functional tests of water level scram instrurcents is administrative in nature.
The operability of the level sensors and trip channels are being adequately verified by other surveillance requirements which are consistent with the RPS design basis, the Standard Technical Specifications, and the vendor's (GE) recommendations for ATTS components (General Electric Topical Report NEDO-21617-A and accompanying NRC letter of approval dated June 27, 1978).
The proposed changes do not involve modification of any existing equipment, systems, or components; nor do they alter the j
conclusions of the plant's accident analyses or radiological release analyses as documented in the FSAR (Reference 2) or the NRC staff's SER (Reference 3).
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Attachment II SAFETY EVALUATION Page 3 of 4 IV. EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Operation of the FitzPatrick Plant in accordance with the proposed amendment would not involve a significant hazards consideration as stated 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 Technical Specification changes to delete the requirements in Table 4.1-1 to perturb the reactor water level after functional tests of water level scram instruments is administrative in nature.
The operability of the level sensors and trip channels are being adequately verified by other surveillance requirements which are consistent with.the RPS design basis, the Standard Technical Specifications, and the vendor's (GE) recommendations for ATTS components (General Electric Topical Report NEDO-21617-A and accompanying NRC letter of approval dated June 27, 1978).
The proposed changes do not involve modification of any existing: equipment, systems, or components; nor do they alter the' conclusions of the plant's accident analyses or radiological release 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 any accident previously evaluated.
The proposed changes are administrative in nature and do not introduce any new failure modes.
They do not involve modification to any of the plant's equipment, systems, or components; nor do they place the plant in an unanalyzed configuration.
3.
involve a significant reduction in a margin of safety.
The proposed changes affect the capability for checking the operational availability of the sensor inputs to the Reactor Protection System (RPS).
Consistent with the IEEE-279-1971, the deletion of the pertubation requirement does not degrade the RPS design basis i
because each of the reactor water level sensors are I
being cross-checked with each other on a daily basis.
The proposed change deletes a superfluous testing requirement and does not involve modification of the plant's systems, equipment, or components.
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Attachment II SAFETY EVALUATION Page 4 of 4 V.
_ IMPLEMENTATION OF THE PROPOSED CHANGE 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 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 safely 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.
SECY-88-304, Policy Issue regarding Staff Actions To Reduce Testing At Power, dated October 26, 1988.
2.
James A.
FitzPatrick Nuclear Power Plant Updated Final Safety Analysis Report, Section 7.2.3.
3.
James A.
FitzPatrick Nuclear Power Plant Safety Evaluation Report (SER), dated November 20, 1972, and Supplements.
4.
NUREG-0123, Standard Technical Specifications, General Electric Boiling Water Reactors, BWR/4.
5.
NEDO-21617-A, General Electric Topical Report, Analog Transmitter / Trip Unit System For Engineered Safeguard Sensor Trip Inputs.
6.
IEEE Std 279-1971, Criteria for Protection Systems for Nuclear Power Generating Stations.
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