ML20235V460
| ML20235V460 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 03/01/1989 |
| From: | George Thomas PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NYN-89020, NUDOCS 8903100198 | |
| Download: ML20235V460 (7) | |
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- e 4^ 75l George S. Thomas Vice President Nuclear ProducHon i
NYN-89020.
[ New Hampshire Yankee Division March 1, 1989' United States Nuclear Regulatory Commission Washington, DC '20555-3-
. Attention: Document Control Desk
References:
(a) Facility Operating License No. NPF-56, Docket No. 50-443 (b)
PSNH Letter-(SBN-1211) dated October 9, 1986, "10CFR.50.59 Evaluations" G. S. Thomas to V. S. Noonan
Subject:
10 CFR 50.59 Quarterly Report Gentlemen:
Enclosed please find the Quarterly Report of 10 CFR 50.59 Safety Evaluations 1
for Seabrook Station.- This report covers the~ period of October 1, 1988, to
-December 31, 1988,. and.is'being submitted pursuant to the reporting requirements outlined in Reference (b).
Should you require further information regarding this matter, please contact Mr. Timothy G. Pucko at (603) 474-9574, extension 4428.
Very truly yours,
._ r j- - A +
w George S. Thomas Enclosure
.cci Mr. William T. Russell Regional Administrator United States Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Mr. Victor Nerses, Project Manager Project Directorate I-3 United States Nuclear Regulatory Commission
{
Division of Reactor Projects l~
Washington, DC 20555 Mr. David G. Ruscitto NRC Senior Resident Inspector P.O. Box 1149 II I
Seabrook, NH 03874 8903100198 890301
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l PDR ADOCK 05000443 R
PDC 1
P.O. Box 300. Seabrook, NH 03874. Telephone (603) 474-9574
4 ENCLOSURE TO NYN-89020' SEABROOK STATION 10 CFR 50.59 SAFETY EVALUATIONS E
QUARTERLY REPORT OCTOBER 1, 1988 TO DECEMBER 31, 1988
- 1.
Design Changes The below listed design changes have been made at-Seabrook Station and' safety evaluations have been performed pursuant-to the requirements of 10 CFR 50.59.
Design Coordination Report: -Number 87-082
Title:
Modify Thermal Barrier Cooling System Head' Pipe
==
Description:==
The vent system for the Thermal B:1rrier Cooling System' has a --
head. pipe which is sized to dampen any pressure surges due to a reactor coolant pump thermal barrier coil rupture. The vent j
system consists of two, ten inch nozzles open to Containment atmosphere. The design allows.for a significant amount of oxygen ingress to the system and thus necessitates frequent addition of a corrosion inhibitor.
To eliminate the need for excessive addition of a corrosion
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inhibitor and to support the concepts of the Seabrook Station ALARA program, this Design Coordination Report was initiated.
The design change adds rupture discs and water seal loops on each of the two, ten inch vents to eliminate the ingress ef-
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l oxygen to the system. The rupture discs are designed to rupture at 15-18 psig above normal system operating pressure and are sized to maintain tha same cross cectional area as the vents.
In addition to the rupture discs, the existing chemical
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addition connection has been rerouted to a more accessible location for the addition of chemicals.
The vent modification and the chemical addition. connection were designed and l
installed per ASME Section III Class 3.
==
Conclusion:==
A 10 CFR 50.59 safety evaluation was performed for this design change and it has been determined that this change will not create any unreviewed safety concerns.
Changes to the Final Safety Analysis Report will be incorporated by means of a future amendment.
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7 Design Coordination Report: Nunber 87-240
Title:
Thermal Barrier Corrosion Coupon i
==
Description:==
To provide the ability to monitor corrosion within the Thermal Barrier Cooling System, a corrosion coupon was added to the Thermal Barrier Cooling System flow path.
1 A new 3/4 inch line bypasses the two thermal barrier heat exchangers and provides a slipstream flow of Thermal Barrier a
Cooling to come in contact with the coupon.
The design change was added in compliance with ASME III, Class 3 requirements.
==
Conclusion:==
A 10 CFR 50.59 safety evaluation was performed for this design change and it has been determined that this change will not create any unrevieved safety concerns.
Changes to the Final I
Safety Analysis Report will be incorporated by means of a future amendment.
Design Coordination Report Number 87-291 j
Title:
Engineered Safety Features Response Time
==
Description:==
The Design Coordination Report was initiated to incorporate two changes to the engineered safety features response times. The first change incorporated the addition of an additional ona second to the time required for the automatic switchover to the containment sump when the refueling water storage tank (RWST) lo-lo level setpoint activated coincident with a safety injection.
The one second was added to account for the RWST lo-lo level signal instrument processing time.
The second change involves the addition of fifteen seconds to the time required for the RWST and volume control tank (VCT) valves to go to the required position on receipt of a safety injection signal. A recent review by Westinghouse of the Safety Injection System sequence logic has identified that the volume control tank isolation valves do not begin to close until the RWST isolation valves are full open.
This delay will add an additional 15 seconds to the overall time.
The addition of the fifteen seconds has been analyzed to verify that VCT water level drawdown will not occur prior to the RWST alignment for the charging pumps.
==
Conclusion:==
A 10 CFR 50.59 safety evaluation was performed for this design change and it has been determined that this change will not create any unreviewed safety concerns.
Changes to the Final Safety Analysis Report and the Technical Requirements Manual will be incorporated by means of a future amendment.
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Y l Design Coordination Report: Number 87-407
Title:
. Service Water Fix for Secondary Component Cooling Heat Exchangers.
==
Description:==
A recent inspection of the Service Water-system (SW) revealed the existence of corrosion in the sixteen inch inlet lines to the Secondary Component Cooling (SCC) Water heat exchangers.
To correct the effects of the corrosion, the affected piping was replaced. In addition to the replacement, new six inch spare connections were added to the sixteen inch SW supply and discharge lines.. The new connections were installed in anticipation of the addition of additional SCC heat exchangers.
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==
Conclusion:==
A 10 CFR 50.59 safety evaluation was performed for this design change and it has been determined that this change will~not create any unreviewed safety concerns.
Changes to the Final Safety. Analysis Report will be incorporated by means of a future amendment.
1 Design Coordination Report: Number 88-008 j
Title:
Auxiliary Secondary Component Cooling Water Heat Exchangers
==
Description:==
The Secondary Component Cooling (SCC) Water system heat exchangers have experienced tubeside corrosion due to low velocity Service Water (SW) system flow.
The corrosion problem has-resulted in the necessity to plug many of the tubes.
To address the problem of low flow to the SW system during low SCC heat load requirements, two auxiliary SCC heat exchangers will be added to enhance system operations at low heat load requirements. The SW cooling water for the new SCC auxiliary heat exchanges will be supplied by the new six inch spare connections identified in Design Coordination Report Number 87-f 407 above.
l The SCC system is not safety related, and the auxiliary heat exchangers will only be used during outages and other low heat load conditions.
==
Conclusion:==
A 10 CFR 50.59 safety evaluation was performed for this design
)
change and it has been determined that this change will not I
create any unreviewed safety concerns. Changes to the Final Safety Analysis Report will be incorporated by means of a future amendment.
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Design Coordination Report: Number 88-122
Title:
Diesel Generator Cooling Water System
==
Description:==
During recent maintenance testing of the A train emergency diesel, a spurious low pressure signal from the on-skid coolant system pressure switch initiated a start of the off-skid auxiliary coolant pump.
The off-skid auxiliary coolant pump discharge ~ valves aligned for auxiliary coolant pump flow and the off-skid and on-skid coolant pumps ran in parallel. As a result of the parallel operation, the system pressure exceeded the setpoint of auxiliary coolant pump relief valve and coolant was discharged to the building floor.
To preclude a repeat occurrence, this Design Coordination Report was initiated to justify an increase in the relief valve setpoint from 70 psig to 100 psig. A complete review of the system design and the shut-off head of the coolant pumps associated with the system provided sufficient justification to support the increase in the relief valve setpoint. A change to the B train emergency diesel relief valve setpoint was also performed.
==
Conclusion:==
A 10 CFR 50.59 safety evalustion was performed for this design change and it has been determined that this change will not create sny unreviewed safety concerns.
Changes to the Final Safety Analysis Report will be incorporated by means of a future amendment.
Design Coordination Report Number 88-125
Title:
125 VDC System Load Calculation Update
==
Description:==
While performing a review of the 125 VDC System load calculation, it was identified that the calculation was not current with latest design changes. This Design Coordination Report was implemented to incorporate the latest data and provide the basis used in the computations to bring the calculation current with the design documents. A review was performed to verify that sufficient safety margin exists t
between the actual identified loads and the capacity of the batteries and battery chargers.
The batteries and chargers continue to satisfy design requirements.
==
Conclusion:==
A 10 CFR 50.59 safety evaluation was performed for this design change and it has been determined that this change will not create any unreviewed safety concerns.
Changes to the Final j
Safety Analysis Report will be incorporated by means of a future amendment.
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6 Design Coordination Report: Number 88-158
Title:
Connections for Mobile Demineralizers
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==
Description:==
In preparation for low power testing, it was identified that the capability to maintain the water quality standards for the steam generators within required specifications would require additional demineralizers.
To accommodate the use of mobile demineralizers, this Design Coordination Report was initiated to incorporate blind flange connections into the Condensate and Demineralized Water Systems. The new connections will attach to the temporary mobile demineralized and will be used for cleanup of the secondary plant water.
==
Conclusion:==
A 10 CFR 50.59 safety evaluation was performed for this design change and it has been determined that this change will not create any unreviewed. safety concerns.
Changes to the Final Safety Analysis Report will be incorporated by means of a future amendment.
2.
_T_emporary Modifications The below listed temporary modification has been made at Seabrook Station and a safety evaluation has been performed pursuant to the requirements of 10 CFR 50. 59.
Temporary Modification Request: Number 88-034 Title Waste Liquid Temporary Demineralization / Filtration.
==
Description:==
In response to the anticipated requirements to process potentially radioactive effluents generated during or following low power testing, temporary hose connections were installed to allow for the use of a mobile waste liquid treatment skid.
This temporary modification will allow the liquid effluent to be processed thru a waste liquid demineralization skid prior to the process liquid being directed back to the permanent plant waste test tank.
The normal waste test tank liquid influent radiation monitor RM-RM-6514 will be out of service as a result of this modification, but a radiation monitor installed on the temporary demineralization skid will provide this monitoring function.
System restoration will be upon completion of the need for waste liquid processing services following low power testing.
==
Conclusion:==
A 10 CFR 50.59 safety evaluation was performed and it has been determined that installation of this temporary modification will not create any unreviewed safety concerns.
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3.
Technical Requirements Manual The below listed changes have been made to the Seabrook Station Technical Requirements Manual and evaluations have been performed pursuant to the requirements of 10 CFR 50.59.
Technical Requirements Manual Change Request: Number 87-001
Title:
Technical Requirement Number 12 - Table 16.3-7
==
Description:==
Changes to the Seabrook Station Technical Requirement 12, Table 16.3-7, " Fire Detection Instrumentation" have been made to incorporate the results of field verifications. Where discrepancies were noted between the installed condition and Table 16.3-7. the National Fire Protection Association (NFPA)
Codes were reviewed to identify the proper installations.
Some areas previously identified in Table 16.3-7 have been deleted as they are not required to be operable for safe shutdown.
These areas will maintain-the same level of defense-in-depth fire protection, but will not be surveilled to the requirements of Table 16.3-7.
==
Conclusion:==
A 10 CFR 50.59 safety evaluation was performed for this design change and it has been determined that this change will not create any unreviewed safety concerns. Changes to the Final Safety Analysis Report will be incorporated by means of a future amendment.
Technical Requirement Manual Change Request Number 87-004 has been described in Section 1, Design Coordination report 87-291.
4.
Final Safety Analysis Report No revisions have been submitted for the Final Safety Analysis Report during this reporting period.
5.
Procedure Channes Procedure changes that require review and approval by the Station Operation Review Committee (SORC) are subject to the requirements of 10 CFR 50.59.
No procedure changes have been made at Seabrook Station during this reporting period that would require a change to the Final Safety Analysis Report.
6.
Test or Experiments There were no tests or experiments performed during this reporting period that require evaluations in accordance with 10 CFR 50.59.
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