ML20077Q659
| ML20077Q659 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 01/13/1995 |
| From: | Wadley M NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20077Q630 | List: |
| References | |
| NUDOCS 9501190308 | |
| Download: ML20077Q659 (7) | |
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UNITED STATES NUCLEAR REGUIATORY COMMISSION NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR CENERATING PIANT DOCKET NO. 50-282 50-306 REQUEST FOR AMENDMENT TO OPERATING LICENSES DPR-42 & DPR-60 LICENSE AMENDMENT REQUEST DATED JANUARY 13, 1995 Northern States Power Company, a Minnesota corporation, requests authorization for changes to Appendix A of the Prairie Island Operating License as shown on the attachments labeled Exhibits A, B and C.
Exhibit A describes the proposed changes, reasons for the changes, safety evaluation and a significant hazards evaluation.
Exhibits B and C are copies of the Prairie Island Technical Specifications incorporating the proposed changes.
This letter contains no restricted or othe-defense information.
NORTHERN STATES POWER COMPANY By
( k (A c4 L5-4 Mike Wadley Plant Manager Prairie Island Nuclear Cenerating Plant On thisf3 ay of MN efore me a notary public in and for said county, personally hppeared $ ke Wadley, Plant Manager, Prairie Island Nuclear Generating Plant, and being first duly sworn acknowledged that he is authorized to execute this document on behalf of Northern States Power Company, that he knowc the contents thereof, and that to the best of his knowledge, information, and belief the statements made in it are true and that it is not interposed f del An v
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9501190308 950113 PDR ADOCK 050002B2 P
4 Exhibit A Prairie Island Nuclear Generating Plant License Amendment Request Dated January 13, 1995 Evaluation of Proposed Changes to the Technical Specifications Appendix A of Operating License DPR-42 and DPR-60 Pursuant to 10 CFR Part 50, Sections 50.59 and 50.90, the holders of Operating Licenses DPR-42 and DPR-60 hereby propose the following changes to Appendix A, Technical Specifications:
Background
This amendment request proposes a change to Technical Specification (TS) 4.4.D.1 to change the interval for the performance of the RHR system leakage test from once every 12 months to perform the test during each refueling shutdown. The current 12 month test interval results in the test being performed during power operation with the RHR system not operating.
The preferred time to perform the test is during a refueling outage with the RHR system in operation and at 350 psig.
Section 5.7.2.4.b of the NUREC-1431 Revision 0 allows a refueling interval for such cests and the Technical Specifications for the Kewaunee Plant, which is a similar design to Prairie Island, also allows this test to be performed on a refueling interval.
Proposed Channes A brief description of the proposed revisions is provided below. The specific wording changes to the Technical Specifications are shown in Exhibits B and C.
- 1. Proposed Channes to Technical Specification 4.4.D.1 The phrase "at 12 month intervals" is being replaced with "during each refueling outage" as shown in Exhibit B.
- 2. Proposed Chanres to the Bases for Technical Specification 4.4.D The statement which described how the 350 psig test pressure is achieved is being deleted to support changing the RHR leakt.ge test to refueling shutdown.
l Justification The preferred method for performance of the RHR leakage test specified by Technical Specification 4.4.D.1 is to perform the test during a refueling shutdown with the RHR system in service and at a pressure of 350 psig or above. However, the current 12 month test interval, specified by Technical Specification 4.4.D.1, results in the test being performed during power operation with the RHR system not operating. The performance of the test at power is more complex and involves pressurizing the RHR system using reactor coolant system water from the chemical and volume control system.
Since the CVCS system is only connected to one train of the RHR system, the test also involves cracking open a cross-connect valve to pressurize the other RHR train.
Because the cross-connect valve is normally maintained closed to provide train separation, the valve is opened per the guidance of Section 6.7 (Manual Action in Place of Automatic Action) of Generic Letter 91-18.
e Eddbt A Pgp 2 e# 5 The performance of the test during refueling shutdown has no impact on the RHR system operability. While the RHR system has been considered operable during the performance of the test during power operation, there are minor changes to the normal system configuration during the test, and as described above, manual actions are used in place of automatic actions for a portion of the test. As such, we believe the performance of the RHR leakage test only during refueling shutdown will eliminate the need to alter the system configuration for the test and to utilize manual actions in place of automatic actions, and thus will enhance plant safety.
The performance of the RHR leakage test on a refueling interval is also consistent with the guidance in NUREG-1431, Revision 0 and with the refueling test interval specified in 10 CFR Part 50, Appendix J for containment penetration Type B and C testing.
Safety Evaluation Introduction The RHR system which operates pcstaccident to maintain the reactor core in a safe condition, becomes part of the containment system during the post accident period when the RHR system operation is changed over from the injection phase to the recirculation phase.
The RHR system is designed for pressures well in excess of the peak containment pressure and the system is designed to remain intact during postaccident operation. However, in order to minimize postaccident leakage from the RHR system and therefore assure the protection of the health and safety of the public, leakage from the RHR system is limited by Section 4.4.D of the current Prairie Island Technical Specifications.
Section 4.4.D requires that those portions of the residual heat removal (RHR) system external to the isolation valves at the containment, shall be hydrostatically tested for leakage at 12-month intervals.
Leakage is limited from the RHR system rather than limiting leakage through the RHR containment isolation valves because the isolation valves will be open postaccident.
This test is required to be perforned at a pressure of 350 psig and consists of a visual inspection for excessive leakage from components of the RHR system.
The Technical Specifications specify that any visual leakage that cannot be stopped at test conditions shall be measured by collection and weighing or by another equivalent method.
The acceptance criteria is that the maximum allowable leakage from either train of the RHR system components shall not exceed two gallons per hour when the system is at 350 psig.
Evalu*L32D As discussed above, the performance of the RHR system leakage test at power is more complex than performing the test during refueling shutdown.
During the performance of the test during power operation, there are minor changes to the normal system configuration during the test.
While RHR system operability is not affected by these minor changes, it is preferable, from an RHR system reliability and plant safety standpoint, to perform the test during refueling shutdown when the RHR system is already operating and when no changes to the RHR system configuration are required.
Exhibit A Pgp 3 sd5 It is believed that any possible increase in the risk to the public health and safety incurred by extending the RHR leak test interval from 12 months to refueling shutdown will essentially be off-set by the reduction in risk
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obtained by not performing the RHR system leakage test during power operation.
j The RHR system functionally becomes part of the containment volume during the i
post-accident period when its operation is changed over from the injection phase to the recirculation phase.
This partial role in postaccident containment warrants surveillance of their leak-tightness. The extension of the test interval would mean that possible RHR leakage could exist undetected for a longer period than allowed by the current Technical Specifications.
However, the possible consequences of leakage from the RHR system outside containment are minor relative to those of the design basis accident.
A review of the results of previous RHR system leakage tests found that no significant leakage has been identified during previous testing.
In addition, it is probable that RHR system leakage would be $dentified during the normal quarterly functional testing and inspection of the RHR system. While the quarterly functional testing is performed at a system pressure significantly lower than 350 psig, there is a high potential that any significant leakage from the RHR system would also be apparent at the lower system pressure.
Therefore, because:
- 1) Leakage from the RHR system has a minor effect on offsite dose,
- 2) Previous testing on a 12 month interval has not found significant RHR system leakage, and
- 3) It is probable that quarterly RHR system testing and inspection would identify system leakage, the extension of the test interval to refueling is not expected to significantly impact the offsite dose consequences of an accident.
The performance of the RHR leakage test on a refueling interval is also consistent with the refueling test interval specified in 10 CFR Part 50, Appendix J for containment penetration Type B and C testing.
Conclusions In conclusion, Northern States Power believes there is reasonable assurance that the health and safety of the public will not be adversely affected by the proposed Technical Specification changes.
l Exhibst A Page 4 of 5 I
Determination of Sienificant Hazards Consideratinas The proposed changes to the Operating License have been evaluated to determine whether they constitute a significant hazards consideration as required by 10 CFR Part 50, Section 50.91 using the standards provided in Section 50.92.
This analysis is provided below:
1.
The proposed amendment will not involve a significant increase in the probability or consecuences of an accident previous 1v evaluated.
The proposed changes to the RHR system leakage test interval only involve the leak-tightness of the RHR system for postaccident operation. As such, the proposed changes will have no impact on the probability of an accident previously evaluated.
The extension of the RHR system leakage test interval could increase the possibility of undetected RHR system leakage outside the containment during post accident conditions. However, the possible consequences of leakage from the RHR system outside containment are minor relative to those of the design basis accident.
Therefore, because leakage from the RHR system has a minor effect on offsite dose, and since previous testing on a 12 month interval has not found significant RHR system leakage, the extension of the test interval to refueling is not expected to significantly impact the offsite dose consequences of an accident.
In addition, it is probable that RHR system leakage would be identified during the normal quarterly functional testing and inspection of the RHR system.
Therefore, for the reasons discussed above, the proposed changes will not significantly affect the probability or consequences of an accident previously evaluated.
2.
The proposed amendment will not create the possibility of a new or different kind of accident from any accident previous 1v analyzed.
There are no new failure modes or mechanisms associated with the proposed changes.
The proposed changes do not involve any modification of plant equipment or any changes in operational limits.
The proposed changes only modify the interval for the performance of the RHR system leakage test.
The performance of the RHR system leakage test on a refueling basis instead of every 12 months cannot create a new or different kind of accident.
Therefore, for the reasons discussed above, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated, and the accident analyses nresented in the Updated Safety Analysis Report will remain bounding.
i Exhibit A
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Page 5 of 5 l
3.
The proposed amendment will not involve a significant reduction in the marnin of safety.
The performance of the RHR system leakage test at power is more complex than performing the test during refueling shutdown.
It is preferable, from an RHR system reliability and plant safety standpoint, to perform the test during refueling shutdown when the RHR system is already operating and when no changes to the RHR system configuration are required. Any possible increase in the risk to the public health and safety incurred by extending the RHR leak test interval from 12 months to refueling shutdown will be off-set by the reduction in risk obtained by not performing the RHR system leakage test during power operation.
The extension of the test interval would mean that pnssible RHR leakage could exist undetected for a longer period than allowed by the current Technical Specifications. However, the possible consequences of leakage from the RHR system outside containment are minor relative to those of the design basis accident.
In addition, it is probable that RHR system leakage would be identified during the normal quarterly functional testing and inspection of the RHR system.
Based on the above, it is concluded that the proposed change does not result in a significant reduction in margin with respect to plant safety as defined in the USAR or the Technical Specification Bases.
Based on the evaluation described above, and pursuant to 10 CFR Part 50, Section 50.91, Northern States Power Company has determined that operation of the Prairie Island Nuclear Generating Plant in accordance with the proposed license amendment request does not involve any significant hazards considerations as defined by NRC regulations in 10 CFR Part 50, Section 50.92.
i Environmental Assessment Northern States Power has evaluated the proposed changes and determined that:
1.
The changes do not involve a significant hazards consideration, 2.
The changes do not involve a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or 3.
The changes do not involve a significant increase in individual or cumulative occupational radiation exposure.
Therefore, the proposed Technical Specification changes would not rew.lt in a significant radiological environmental impact.
Exhibit B Prairie Island Nuclear Generating Plant r
License Amendment Request Dated January 13, 1995 Proposed Changes Marked Up On Existing Technical Specification Pages Exhibit B consists of existing Technical Specification pages with the proposed j
changes highlighted on those pages. The pages affected by this License Amendment Request are listed below:
TS.4.4-4 B.4.4-2 P
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