ML20062H601
| ML20062H601 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 11/30/1990 |
| From: | Feigenbaum T PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NYN-90207, NUDOCS 9012050151 | |
| Download: ML20062H601 (24) | |
Text
_ _
Sew Hampshire Ted C. Feigenbovm Yh
- President and Chie ' Executive Oflicer NYN 90207 November 30, 1990 United States Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Document Control Desk
References:
(a) Facility Operating License No. NPF 86, Docket No. 50 443 (b) PSNH Letter (SBN 1211) dated October 9, 1986, *10CFR50.59 ~
" valuation", G. S. Thomas to V. S. Noonan
Subject:
10CFR50.59 Ouarterly Report Gentlemen:
Enclosed please find the Quarterly Report of 10CFR50.59 Safety Evaluations for Seabrook Station. This report covers the period of July 1,1990, to September 30,1990, and is being submitted pursuant to the reporting requirements outlined in Reference (b).
Should you require further information regarding this matter, please contact hir. James M. Peschel, Regulatory Compliance Manager at (603) 474 9521, extension 3772, Very truly yours,
- l.f.Ab O Ted C. Feigenbaum Enclosure (s)
, TCP:JES/ tad cc: Mr. Thomas T. Martin Regional Administrator United States Nuclear Regulatory Commission Region i 475 Allendale Road King of Prussia, PA 19406 Mr. Noel Dudley NRC Senior Resident inspector t, ' [ 1, P.O. Box 1149 Seabrook, NH 03874-90120S0161 901130 -
'0R ADOCK 05000443 R PDC /
New Hampshire Yonkeo Division of Public Service Company of New Hampshire P.O. Box 300
- Scobrook, NH 03874
- Telephone (603) 474 9521
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New Hampshire Yankee
. November 30, 1990 l-l l
l' ENCLOSURE 1 TO NYN 90207 Seabrook Station 10CFR50.59 Safety Evaluation Quarterly Report July 1,1990 September 30, 1990 l
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New Hampshire Yankee November 30, 1990 ENCLOSURE 1 TO NYN 90207
- 1. Design Channes The below listed design changes were made at Seabrook Station and safety evaluations were performed pursuant to the requirements of 10CFR50.59.
Design Coordination Report: Number 86 0096
Title:
Chemical Addition to Floor Drain Tanks
Description:
This Design Coordination Report (DCR) was initiated to add a chemical injection subsystem to the Liquid Waste System to provide pH adjustment of the liquid stored in the floor drain tanks. This would make the liquid stored in the tanks less corrosive, and therefore, prolong the service life of the liquid waste system. The liquid waste system is designed to collect and concentrate radioactive liquid wastes and transfer the concentrated wastes for further processing in the solid waste system. The liquid waste system is non nuclear safety class and non seismic category I.
This DCR will require changes to the list of system components and the system flow diagram contained in the Final Safety Analysis Report-(FSAR).
This DCR will also require that the system descriptions contained in Chapter 11 of the FSAR to be expanded to include the chemical addition subsystem.
Conclusion:
A 10CFR50.59 safety evaluation was performed for this _ design change 'and it was determined that this change wal not create, an unreviewed safety question. Changes to the Final Safety Analysis Report will be incorporated by means of a future amendment.
Design Coordination Report: Number 86-0628
Title:
Secondary Component Cooling Water System By Pass Filter
Description:
This Design Coordination Report (DCR) incorporates a by pass filter in the secondary component cooling water system. It makes permanent a temporary modification which has demonstrated the filters worth during the life of the plant. The secondary component cooling water system is a non-safety related system and has no safety design basis. This DCR will require 1
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New Hampshire Yankee November 30, 1990 the incorporation of the filter onto the system flow diagram contained in Final Safety Analysis Report (FSAR) Figure 10.4.9, General Arrangement Figure 1,2 and text of the system described in Paragraph 10.4.10.2.
The facility as described in Section 1.2.9.5 of the FSAR is not affected.
Conclusion:
A 10CFR50.59 safety evaluation was performed for this design change and it was determined that this change will not create an unreviewed safety question. Changes to the Final Safety Analysis Report will be incorporated
-- by mcans of a future amendment.
I Design Coordination Report: Number 87 0283
Title:
UPS ED I-4 Reliability Changes
Description:
This Design Coordination Report (DCR) replaces existing non safety related
_ single phase Uninterruptatle Power Supply (UPS) I 4 with a Unit 2 three .
phase UPS. This change and revised loading for the paging system increases the diesel generator loading, it also revises the electrical configuration to improve the availability of the system during a bus E5 outage by adding external transfer equipment to allow for an alternate connection to bus 1. Because UPS I 4 is listed as required for hot and cold shutdown in Final Safety Analysis report (FSAR) Appendix R, the additional equipment is required to be added to the report and changes to
- the existing system reflected.
Specifically, this DCR makes changes to the FSAR Diesel Generator loading tables 8.31 & 8.32, figures 8.32 & 8.316, and Appendix R. These changes do not affect the intent of the requirements but are needed to show the new configuration and the change in loading.
The diesel generator new total loading is within the rated capacity,
- replacement equipment meets- all the technical requirements, and the changes .do not change any of the- accident scenarios, therefore, this modification does not increase the consequences of an accident previously evaluated by the FSAR. Equipment changes to Appendix R meet the intent and does not reduce the effectiveness to mitigate the consequences caused by a fire in the applicable fire zone.
Conclusion:
A 10CFR50.59 safety evaluation was performed for this design change and it was determined that this change will not create an unreviewed safety question. Changes to the Final Safety Analysis Report will be incorporated by means of a future amendment.
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November 30, 1990 1
1 Design Coordination Report: Number 87 0311
Title:
CBS/RHR Redundant Check Valves ,
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Description:
This Design Coordination Report (DCR) installs additional check valves in the 12' and 16' interconnecting piping between the Containment Building Spray (CBS) and Residual Heat Removal (RHR) systems as a result of NHY commitments to the NRC to install system enhancements following leakage from RHR to CBS during late 1986. This design change also includes new CBS relief valves installed adjacent to the new check valves and l' bypass piping around the existing and new check valves for performance monitoring and inservice testing of the check valves.
Leakage from the RHR system, suction side of the RHR pump, into the CBS system across the interfacing check valves is a concern only during plant hot shutdown operation (Mode 4) when the RHR system is aligned to the Reactor Coolant System (RCS) and RCS temperature is.between 200'F and 350*F, which occurs infrequently and for a short duration.
During operational modes 1,2, and 3 the RHR system is isolated from the l RCS, depressurized and at ambient temperature.
The additional check valves, relief valves and bypass piping and valves all meet the design criteria of the existing systems. The additional check
- valves provide greater assurance against leakage which would overpressurize l-the CBS system. The calculated NPSH available for the RHR pumps during recirculation (the design basis case in compliance with Reg. Guide 1.1) is maintained at the previously identified value.
This design change will require modifications to Final Safety Analysis Report (FSAR) Sections 3, 5, 6 and 9. . Additionally,' a backseat test- not described in the FSAR requires a change to FSAR Table 3.9(B) 23.
Conclusion:
A 10CFR50,59 safety evaluation was performed for this design change and associated back rat test, and it was determined that they will not create an unreviewed safety question. Changes to the Final Safety Analysis Report were incorporated into Amendment 63.
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New Hampshire Yankee November 30, 1990 Design Coordination Report: Number 87 0331
Title:
ATWS Mitigation System
Description:
This Design Coordination Report (DCR) installs an Anticipated T:ansients Without Scram (ATWS) Mitigation System (AMS). Tbc AMS a:ts as a backup to the Reactor Protection System for a loss of feedwator event.
The AMS will trip the turbine and start emergency feedwater any time tt.rbine power is above 40% and narrow range level is below 5% in three steam generators. The AMS is a non safety related system required by 10CFR50.62. The AMS receives six isolated inputs from the Process Protection System (four narrow-range steam generator levels and two turbine impulse pressures). The AMS has been designed to meet the criteria described in the PSAR.
The AMS is an extra measure of redundancy required by 10CFR50.62 to provide mitigating actions in case of a failure which would prevent operation of the Reactor Trip System and does not reduce the margin of safety for any accident or the basis for any Technical Specifications. ;
Conclusion:
A 10CFR50.59 safety evalunda . was pc,-formed for this design change, and it was determined that this change will not create an unreviewed safety question. Changes to the Final Safety Analysis Report will be incorporated by means of a future amendment.
Design Coordination Report: Number 87 0422
Title:
Replacement of RHR Miniflow Valves,1 RH 610 & 1-RH 611
Description:
This Design ' Coordination Report (DCR) revises _ Residual Heat Removal' (RHR) pump miniflow piping, changes miniflow isolation valves 1 RH-FCV 610 and 1 RH FCV 611 from Westinghouse ASME Class 2 motor operated gate valves to Velan ASME Class 2 motor operated globe valves, increases the RHR pump. miniflow. rate from 500 gpm to 600 gpm and changes the flow restricting orifice' diameter (1 RH FE 2473 & 2474) to redistribute the pressure drop in the piping system to reduce piping vibration and increase system reliability.
The Velan valve operator meets the requirements for the harsh environment in the RHR vaults and for an active valve.
The increase in RHR pump miniflow provides additional margin for meeting the RHR pump flow requirements to address NRC Bulletin 88 04, Potential Safety Related Pump Loss.
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New Hampshire Yankee November 30,- 1990 The principle design considerations are that the valve and piping are ANS Safety Class 2, ASME Section III, Class 2; the valve is an active valve I capable of operating against the pump differential pressure; the 1E motor operator is environmentally qualified; and the flow characteristics of the !
I miniflow line are maintained. These considerations were maintained with the application of the Velan valve. l 1
This FCR will affect Final Safety Analysis Report (FSAR) Tables 3.9(N)- ,
11, 3.9(B) 25, 5.417, 6.3 5, and Figure 5.410. I
Conclusion:
A 10CFR50.59 safety evaluation was performed for this design change, and it was determined that this change will not create an unreviewed safety question. Changes to the Final Safety Analysis Report will be incorporated by means of a future amendment.
Design Coordination Report: Number 88-0183
Title:
Additional Sodium Hypochlorite Tank Capacity
Description:
. This Design Coordination Report (DCR) modifies the current Chlorination System configuration to include two additional Hypochlorite Storage Tanks (8000 gallons each) and associated PVC piping it,.* a total storage capacity of 21600 gallons. The Chlorination System is non shmic and. non safety related.
This design change will require modification of Final Safety Analysis Report (FSAR) Figure 10.4 3, " Chlorination System Overview".
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Conclusion:
A loCFR50.59 safety evaluation was performed for this design change, and l
it was determined that this change will not create an unreviewed safety question. Changes to the Final Safety Analysis Report will be incorporated l by means of a future amendment.
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New Hampshire Yankee November 30, 1990 Design Coordination Report: Number 89 0037
Title:
Fuel Storage Building Fire Detector Replacement
Description:
This Design Coordination Report (DCR) provides modifications to the fire detection system in the Fuel Storage Building (FSB). The FSB is provided with a full area fire detection system installed to the requirements of NFPA 72E, " Automatic Fire Detectors'. The area has negligible fire loading, seismically designed walls and arca wide detection.
The existing areas of concern are: Area (1) _over the fuel pool and Area (2) over the west staltway bay. Detectors in these areas are approximately 40 feet above the floor and .cquire scaffolding that is unique in rigging. !
The fire detection system is non safety related. The infra red flame detectors and associated components are non 1E. Since the entire FSB is seismic, the conduit and detectors are scismically supported.
This DCR deletes eleven inaccessible ionization detectors from the two areas noted above and installs four infa red flame detectors in the area.
The existing Control Panel (1 FP CP 454) Zones 2 & 4 will not change with the new detectors. ;
This DCR will require modification of Final Safety Analysis Report (FSAR) .[
Tables 9.5 2 and 16.3 7 and Technical Requirement 12, item 17.
The fuel pool area detectors are approximately 40 feet off the floor and ;
over an area devoid of combustibles or transient combustibles. Since the fire loading in this area is negligible, removal of eleven ionization detectort and installation of four infra red flame - detectors will not add to the consequences of. an accident previously evaluated in the FSAR. f Additionally, since the fuel pool area contains no safe shutdown cables or equipment, and no equipment important to safety, the consequences of a malfunction of equipment important to safety is not increased.
Conclusion:
A 10CFR50.59 safety evaluation was performed for this design change, and :
it was determined that this change will not create an- unreviewed safety- ;
question. Changes to the Final Safety Analysis Report wil'. be incorpoeted by means of a future amendment.
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New Hampshire Yankee November 30, 1990 Design Coordination Report: Numbe 89 0046
Title:
Containment Gaseous backup Radiation Monitor
Description:
This Design Coordination Report (DCR) implements the installation of a containment gaseous radiation backup monitor which will act as a backup to regular monitors 1 RM RM 65261 & 2. The babo monitor 1 RM.
RM 6548 along with the containment sump level channel w!!! provide a minimum of 2 out of 3 channels required per Technical Specification 3,4.6.1, for the Reactor Conlant System (RCS) leakage monitoring system.
This DCR deals with a non nuclear, non s./ety related (NNS) modification to the plant. The radiation monitors for RCs leakage monitoring are not used to mitigate any accident as evaluated in FSAR. The new backup gaseous radiation monitoring skid RM SKD 131 is of the same quality as the regular monitoring oxid RM-SKD 60. The new skid is a seismic Category 1 monite. which is also installed per Seismic Category 1 installatica details to preclude potential missile generation following a seismic event, Additionally, there is no safety grade equipment in its immediate vicinity. The backup gaseous radiation monitor is only for monitoring RCS leakages within containment. It has no control function associated with it. This radiation monitor is isolated by qualified isolators from Class 1B radiation mo nitors which provide control and actuation functions, so a failure of this rionitor will not affect safety related radiation monitors.
Final Safety Analysis Rept,rt (FS AR) Sections 12,3,4,2,b.2a, 5.2.5,3.b2, 5.2.5.5c; Table 12,3-14 and Figure 12.3.19, 5.2 21 and 1.2 3 are affected by this DCR and .will be revised to include monitor 1 RM RM 6548.
Conclusion:
A 10CFR50.59 safety evaluation was performed for this design change, and it was determined that this change will not create an unreviewed safety question. Changes to the Final Safety Analysis Report will be incorporated
- by means of a future amendment, e
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Design Coordination Report: Number 89 0047
Title:
Start up Strainers for Condensate & Heater Drain Pumps
Description:
This Design Coordination Report (DCR) adds new start up strainers to the suction piping of the condensate pumps and heater drain pumps. A total of five strainers will be installed. The strainers are required to protect the pump from debris that brcaks loose frcm the piping systen.s during power ascension testiq. The strainers can be removed followirs cleanup of the respective systems. Flush and cleanout connections will be added to the suction piping to allow strainer cleaning to be performed in place without complete removal of the pipe spool. Differential pressure indicators will )
be installed to monitor the strainer differential pressure and provide a means to determine when a strainer requires cleaning.
This DCR also modifies the seal injection to Heater Drain Pumps 1 HD- 1 P31A/P31B to enhance the control of seal water.
This DCR requires a revision to Final Safety Analysis Report (FSAR)
Figure 10.4 4 to incorporate the addition of flush, clean out and differential pressure tap connections and the change in seal injection flow to the Heater Drain Pumps.
The portions of the Condensate System and Heater Drain System affected by this DCR provide no safety function and do not support or impact any safety system.
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Conclusion:
A 10CFR50.59 safety evaluation was performed for this design change, and it was determined that this change will not create an unreviewed safety question. Changes to the Final Safety Analysis Report will be incorporated-by means of a future amendment.
Design Coordination Report: Number 89-0048 i
Title:
East Pipe Chase Heating
Description:
This Design Coordination Report (DCR) replaces the temporary propane heaters in the east main steam feedwater pipe chase with six (6) permanent 3 electric unit heaters. The heaters added by this DCR will maintain the east f pipe chase at the minimum acceptable temperature during winter shutdowns.
This DCR will reduce the possibility of a fire, explosion, or other hazards a related with the use of propane. This modification does not adversely 8
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New Hampshire Yankee November 30, 1990 I
impact any safety related systems or components because the heaters are seismically mounted, and the associated breaker is left open during normal operation.
This DCR changes Final Safety Analysis Report (FSAR) Figures 3.11 (B) 1 and 9.4 7 to reflect the addition of the unit heaters.
Conclusion:
A 10CFR50.59 safety evaluation was performed for this design change, and it was determined that this change will not create an unreviewed safety question. Changes to the Final Safety Analysis Report will be incorporated by means of a future amendment.
Design Coordination Report: Number 89 0055
Title:
Copes Vulcan Valve Modification
Description:
This Design Coordination Report (DCR) modifies the Copes Vulcan valves in the Main Steam (MS), Moisture Separation Drains (MSD), Heater Drains (HD), and Auxiliary Steam (AS) systems as follows: the clearances between the stem / guide bushing, stem / plug bushing, and stem / packing gland will be increased by replacement of the guide bushing, plug bushing, and packing' gland. In addition, the material of the guide bushing will be changed from 410 stainless steel having a hardness of 32 to 38 Rockwell C to 420 stainless steel with malcomizing which has a hardness of 93 to 94 Rockwell 15N. This change in hardness will increase the load carrying capability and reduce the potential far galling the guide bushing / stem interface.
This modification' will enhance the operability of the Copes-Vulcan valves.
Before this mo6fication, an event with all condenser steam dump valves stuck open wo, possible due to the common mode failure of the valve stem
-galling and binding. This event is bounded by a steam line break which
~ is evaluated in Section 15.1.5 of the Final Safety Analysis Report (FSAR).
This event would be classified as an ANS Condition IV event. This modification eliminates this common mode failure. This DCR does not effect the event of inadvertent opening of a condenser steam' dump . valve covered _by PSAR Section 15.1.4 which is still applicable. This event is -
classified as an ANS Condition Il event. The portions of these ' systems which contain these modified valves are non nuclear safety related.
This DCR also modifies the condenser steam dump valve stroke times. This change consists of an increase from three to five seconds in valve stroke time from the. full closed to full open position after receiving a trip /open signal. The modulation mode full stroke time has been changed to 25 1
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j New Hampshire Yankee November 30, 1990 i
seconds. This change ensures the condenser steam dump valves do not overshoot the control stroke position during the modulation mode.
FSAR Section 10.4.4 will be modified to reflect the new valve stroke times, m
Conclusion:
A 10CFR50.59 safety evaluation was performed for this design change, and it was determined that this change will not create an unreviewed safety question. Changes to the Final Safety Analysis Report will be incorporated b by means of a future amendment. I
- 2. Minor Modifications n
The below listed minor modifications were made at Seabrook Station and safety evaluations m were performed pursuant to the requirements of 10CFR50.59, Minor Modification: Number 90 505 h 1
Title:
Sample System Cation Exchanger Sample Points _
Description:
This Minor Modification (MMOD) provides a redesign of conductivity cell l discharge lines which currently drain into a common header within the control panels. The redesign will cap existing connections on the header and re-route discharge lines to the sample sink enclosure. This redesign will be common for control panels 1 SS CP.166B,1 CAS IR 75 .%I l CAS-CP 186; allowing chemistry to check conductivity cell accuracy and pre-condition samples for chemistry analysis, i This MMOD ,dso:
Incorpa'ates the design to extend bubble and reboiler vent
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lines , above SS CP 166B, adds vent and drain valves to file c. (SS F-177, 178, 179 & 180) on steam generator .i c npling lines in SS CP 166B and adds details for maximum -
-tube length at sample points in SS CP 166A.
Incorporates check valves, downstream of CAS-F-87601 thru 6, to prevent back flow of sample streams CAS CP 186; and revises the-location of isolation valves to drain header per as built.
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New Hampshire Yankee November 30, 1990
- Revises vent and drain valve types on filters SS F 177,178, 179 and 180 and adds a tee with tube stub for a drop in thermometer downstream of SS CE 2748 for reboiler discharge temperature.
The sample lines for 1 SS CP-166B are addressed in the Final Safety Analysis Report (FSAR) section 9.3.2.2.A.2 and shown on Figure 9.3 5, sht. 4. This modification does not change the sample system design function. Sample lines for CAS IR 75f and CP 186 do not appear in the FSAR.
The probability of a malfunction and its consequences of equipment important to safety will not be increased due to the design changes within this MMOD. This portion of the Sample System and Chemical Analysis System has no emergency or safety function and is not required for reactor safe shutdown, The portions of "SS" and
'CAS' systems affected by this change are classified as non nuclear and non safety and does not alter or affect the designed operation of those systems.
Conclusion:
A 10CFR50.59 safety evaluation was performed for this design change and ;
it was determined that this change will not create an unreviewed safety question. Changes to the Final Safety Analysis Report will be incorporated by means 'of a future amendment.
Minor Modification: Number 90 537
Title:
Service Water Vacuum Breaker SW V173 Replacement
Description:
This Minor; Modification (MMOD) replaces SW V-173, a swing check' valve utilized as a vacuum breaker valve to resolve-a continuing leakage problem.
SW V-173 is installed in a seismically supported non nuclear safety portion of the Service Water (SW) system. It provides pressure transient control for the Secondary Component Cooling Water (SCCW) heat exchanger and its supply piping. Since the open vent area require" by the supporting hydraulic transient calculation is satisfied by the. new valve, the design function is also satisfied. The new: configu' ration has been seismically qualified.
The vacuum breaker check valves are discussed in Section 9.2.1.2 of the-Final Safety Analysis Report (FS AR). The design function and qualification, as described, is satisfied by'the replacement valve and l 11
New Hampshire Yankee November 30, 1990 t
1 arrangement. PSAR Figure 9.21, Sheet 3, will require revision to depict the new vacuum breaker arrangement.
Conclusion:
A 10CFR50.59 safety evaluation was perforrvd for this design change and it was determined that this change will sot create an unreviewed safety question. Changes to the Final Safety Analysis Report will be incorporated by means of a future amendment.
Mioor Modification: Number 89 547
Title:
Service Air Cornpressor Wiring Changes and Appendix R Lighting
Description:
This Minor Modification (MMOD) designates Service Air Compressors SA-C 1A and SA C 1B as manually started loads on the diesel generators after a Loss of Power (LOP) or LOP / Safety injection (SI). After fire water is aligned for cooling water, the local reset button is pushed to start the compressors. - The compressors will not automatically start and load on the diesels, and the potential for a high temperature trip due to a lack of cooling is minimized.
[ Eliminating this period of operation prior to a potential high temperature trip is acceptable as follows. From an Appendix R/ fire viewpoint, cir is not needed to support required actions until after the time period it takes to manually align fire water so delaying compressor start until cooling is aligned is acceptable. . From the viewpoint of coping with a design basis l accident, the compressors are non safety related so credit can not be taken l
for their availability. Where. air is required to operate safety related l equipment, separate bottled air sources are provided. The compressors l were provided with diesel backed power to improve the_ reliability and l availability of service air. Preventing the compressors from starting without cooling also improves reliability and availability by eliminating the potential
. for equipment damage from high temperatures. ,
This MMOD also adds an eight hour lighting pack to resolve an Appendix R Report concern identified in SIR 89 064.
This MMOD requires changes to the Final Safety Analysis (FSAR) Diesel Generator Loading Table 8.3 2 and to the Appendix R Report. .
This MMOD does not reduce the margin of safety as defined in the basis of any Technical Specification because the Service Air System does not form the 12
New Hampshire Yankee November 30, 1990 basis of any Technical Specification and the overall diesel generator loading ,
is within their rated capacity.
Conclusion:
A 10CFR50.59 safety evaluation was performed for this design change and it was deter nined that this change will not create an unreviewed safety -
question. Changes to the Final Safety Analysis Report will be incorporated ;
by means of a future amendment. =
T" Minor Modliication: Number 90 549 2
Title:
Steam Generator Blowdown System Tle in for Portable Demineralizers
Description:
This Minor Modification (MMOD) provides tie ins and valves for the steam generator blowdown system in order to facilitate the plant utilization of b mobile demineralizers to aid in meeting secondary side water quality 1
_ standards prior to and during power ascension.
I Supply to the mobile units will be from an added - flanged removable -
connection downstream of the distillate coolers to be blinded when not a utilized. The treated fluid is returned to the condenser directly upstream of S3 PCV-1925. Installation of the shut off valves in the lines will aid in rcelignment of the system flowpath.
The incorporation of the above connections into the existing piping will have no detrimental effects on the normal system operation or design of
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the steam generator blowdown system. During use of the portable demineralizers, permanent blowdown system demineralizers as described in Final Safety Analysis Report (FSAR) Sections 10.4.8.3g and 10.4.8.6c will be bypassed.
E Conformance to plant water chemistry design will be enhanced by this .
modification. A temporary modification will evaluate and approve the usage of mobile domineralizers in the blowdown system.
Implementation of this MMOD will require modification of FSAR Figure 10.4-7 to depict the location of the tie-ins and shut off valves.
The modifications made by this MMOD have no impact on any safety-related equipment, y_;
A 10CFR50.59 safety evaluation was performed for this design change and
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Conclusion:
it was determined that this change will not create an unreviewed safety question. Changes to the Final Safety Analysis Report will be incorporated by means of a future amendment.
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l New Hampshire Yankee November 30, 1990 Minor Modification: Number 90 551
Title:
Steam Generator Feed Pump (SGFP) Oil Filter Trip Setpoint Changes
Description:
This Minor Modification (MMOD) revises two setpoints associated with the Steam Generator Feed Pumps (FW P 32A & 32B). One change is the setpoint associated with the skids lube Oil Filter high differential pressure alarm. This setpoint was lowered so as to be present prior to filter automatic bypass on high differential pressure. The second change is specific to P 328. The time delay associated with the pumps automatic trip on low NPSH has been increased. This increased time delay will stagger the feed pump trips so that P 32A trips first and if conditions warrant P-32B will follow. This change is intended to climinate abrupt loss of feedwater on loss of pump suction head.
These setpoint changes are to non safety related components of the Feedwater System. Their function or the function of the Steam Generators Feed pumps are not required for safe shutdown of this plant. These MMOD changes have no impact on the Emergency Feedwater system and the consequences of a previously evaluated accident have not been increased.
Setpoint changes, as specifically revised by this MMOD, do not change the verbiage or figures of Final Safety Analysis Report (FSAR) Chapter 10.4.
These changes do, however, revise the - overall operation of the Steam Generator Feed Pumps and, therefore, makes changes to the facilities as described in the FSAR.
Conclusion:
A 10CFR50.59 safety evaluation was performed for this design change and it was determined that this change will not create an unreviewed safety question. Changes to the Final Safety Analysis Report will be incorporated by means of a future amendment.
Minor Modification: Number 90-555 L
Title:
Containment Isolation P&ID Changes
Description:
This Minor Modification (MMOD) revises P&lD SS B/D20518 (FSAR Fig.
9.3 5) to depict Reactor Coolant. System (RCS) sample outside containment isolation valves RC FV 2894 and RC FV-2896 as locked closed (disabled in the closed position). In addition,- Final Safety Analysis report (FSAR)
Table 6.2 83 will be revised to indicate that these valves are locked closed normally and during shutdown and FSAR Figure 6.2 94 will be corrected 14 i
New Ilampshire Yankee November 30, 1990 to depict drain connections on the Reactor Coolant Hot Leg Loop 3 sample line as closed in accordance with Procedure OS1090.05.
This MMOD corrects inconsistencies between the FSAR and plant configuration to show compliance with 10CFR50, Appendix A, General Design Criteria (GDC) 55 and 56. This MMOD does not result in any hardware changes to the facility or alter the design basis for isolation of these lines.
Conclusion:
A 10CFR50.59 safety evaluation was performed for this design change and it was determined that this change will not create an unreviewed safety question. Changes to the Final Safety Analysis Report will be incorporated by mcans of a future amendment.
Minor Modification: Number 90-556
Title:
Feed Pump Seal Water Systems Modifications
Description:
This Minor Modification (MMOD) provides two enhancements for the main and start up feed pumps as follows:
. The start up feed pump lube oil cooling water supply source will be changed to the first stage extraction of the pump, rather than the present high pressure final (sixth) stage. The lower pressure supply combined with an enhanced pressure breakdown orifice will reduce the potential for continued pipe crosion.
. The' Main feed pumps seal water drain cavity will be vented to promote adequate drain capacity.,
This MMOD also provides for rerouting of a 1 1/2" vent (CO 4076 02),
located upstream of CO-LV 4074, to prevent further overflow of the Steam Generator Feed Pump (SGFP) lube -oil berm.
The feedwater system is described in Final Safety Analysis Report (FSAR)
Section 10.4.7 and the start up feedwater system in Section 10.4.12. The
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i modifications made herein affect non safety related portions of the' systems, and have no impact on the feedwater accidents and malfunctions described in FSAR Section 15. -In'the event that these systems are not available, the emergency feedwater system will provide the required emergency supply of' feedwater. The change to the startup feed pump lube oil cooling subsystem will impact FSAR Figure 10.4 4, sheet 6 (Reference FCR 90 042).
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Conclusion:
A 10CFR50.59 safety evaluation was performed for this design change and it was determined that this change will not create an unreviewed safety question. Changes to the Final Safety Analysis Report will be incorporated by means of a future amendment.
Minor Modification: Number 90 574
Title:
Control Orifices for General Electric Control Valves
Description:
This Minor Modification (MMOD) adds restricting orifices in the Electro-Hydraulic Control System (EHC) fluid emergency trip supply (ETS) lint s leading to the solenoid operated Fast Acting valve of General Electric control valves. The orifice acts as a restrictor to fluid flow to prevent !
unwanted LTS pressure drops that may be experienced during valve testing.
There are twenty (20) affected G.E. valves associated with the operation of the non nuclear safety class Turbine Generator. Additionally, by installing the orifice in the ETS supply to the Fast Acting valve there is no affect on valve closure requirements.
Implementation of this MMOD will not affect the Technical Specifications nor will it reduce the margin of safety as defined in any Technical Specification basis.
Conclusion:
A 10CFR50.59 safety evaluation was performed for this design change and it was determined tlat this change will not create an unreviewed safety question, Changes to the Final Safety Analysis Report will be incorporated by means of a future amendment.
Minor Modification: Number 90 577
Title:
Main Steam Drain Line Condenser Isolation
Description:
This Minor Mod!fication_(MMOD) modifies the Main Steam Drain (MSD) j J
system to provide for the decoupling of the drain lines from their heat sink, Condenser C. It incorporates two sets of two in series isolation valves, located adjacent to the condenser. The valves are manually operated and y are normally kept in the open position until the condenser is not available to receive condensate from the drain lines. Also in'orporated into the MSD system are two new drain lines which can remove condensate from the system when the condenser is decoupled. These lines are closed when the plant is in normal operation.
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This MMOD also provides for the installation of a diffuser internal to l condenser "C" and the incorporation of break down orifices in the feed
.mp turbines, high pressure stop valves above the seat drains. Both changes dissipate the high energy entering the condensers, eliminating the potential of serious tube damage caused by jet impingement. Also, piping has been up graded to stainless steel to prevent the potential of erosion l downstream of the orifices.
The affected piping is non safety class and non safety related. Final Safety Analysis Report (FSAR) Figure 10.3 2 reflects this piping configuration and will be revised to indicate the new valves.
Conclusion:
A 10CFR50.59 safety evaluation was performed for this design change and it was determined that this change will not create an unreviewed safety question. Changes to the Final Safety Analysis Report will be incorporated by means of a future amendment.
Minor Modification: Number 90 609
Title:
Additional Support on 1 Inch Main Steam Instrument Sensing Line
Description:
This Minor Modification (MMOD) adds a support to the 1" Main Steam Instrument Sensing Line leading from the 30" Ma!9 Steam line to line 1-MS PI 3081 P and a flex hose to the H" tubing for additional stability and to reduce vibratory movements. The 1" Main Steam Sensing line is non-nuclear, non safety and non seismic.
Implementation of this MMOD will involve- revisions to' the Final Safety .t Analysis Report (FSAR) describing the facility. This change tr the Main !
Steam (MS) System effects FSAR to Figure ;No.10.31, Sheet 5. This change has no affect on the operation or function of the instrument. .
Conclusion:
A 10CFR50.59 safety evaluation was performed.for this design change and it was determined that this change will not create an unreviewed safety question. Changes to the Final Safety Analysis Report will be incorporated by means of a future amendment.
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. New llampshire Yankee November 30, 1990 Minor Modification: N a.ber 90-619
Title:
Rescale EFW Flow Transmitters
Description:
This Minor Modification (MMOD) rescales the Emergency Feedwater (EFW) flow transmitters and amiated indicators and recorders and revises the setpoint for automatic iso.ttion of EFW flow to a faulted Steam Generator.
The effected instrument loops are FW F-4214-2,4; 4224 2,4; 4234-2,4 and 4244 2,4. This MMOD also affects the EFW flow transmitters used for indication on the Remote Shutdown Fanel, instrument channels FW F 4214-5,4224 5,4234 5 and 4244-5. The range of the transmitter has been revised from 0 to 500 gpm to O to 600 gpm. The automatic isolation setpoint for the faulted Steam Generator (SG) has been revised from 425 gpm to 525 gpm.
The reason for this change is to prevent inadvertent isolation of EFW flow when both EFW pumps are running, the previous setpoint was too low and in fact on an EFW initiation on 6/20/90 one steam generator did isolate.
The previous setpoint was 425 gpm.
Following the inadvertent EFW isolation, the calculations for the flow required to isolate a faulted SG have been reviewed. It was determined that the original calculation had conservatively used a large value for system line loss. The revised calculation concluded that the flow to the faulted SG could be as high as 590 gpm and still assure 470 gpm to the intact SGs.
The 470 gpm flow to the three intact SGs is a Condition IV (feed / steam line break) accident analysis f?ow assumption. Thus the new design value for this setpoint is 590 gpm. A setpoint of 525 gpm was selected which is conservative to the design value and accounts for instrument uncertainty including some margin and also provides operating margin. The maximum flow with both EFW pumps running under normal conditions should be between 325 and 474 gpm for a range of SG pressures between 1200 and 700 psig.
The probability of a malfunction of equipment important to safety will not be increased by rescaling the instrument loops and changing the setpoint for high EFW flow isolation. The same instrument flow elements, transmitters and circuit cards are used. The only physical changes required
.are new scales for the indicators. Scales are passive components which will.
not cause a malfunction of the devices.
This MMOD will require revision of Final Safety Analysis (FSAR) Table 7.51, " Accident Monitoring Instrumentation", and Table 7.5 2, " Main Control Board Indicators".
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Conclusion:
A 10CFR50.59 safety evaluation was performed for this design change and it was determined that this change will not create an unreviewed safety '
question. Changes to the Final Safety Analysis Report wd1 be incorporated ,
by means of a future amendment.
Minu Modification: Number 90-637
Title:
Lag for Steam Flow / Feed Flow Mismatch in Steam Generator Level Control
Description:
This Minor Modification (MMOD) makes enhancements to the steam generator level control system. The specified changes involve the addition of a lag period on the output of the steam flow / feed flow summer in the steam generator level control system. .This change along with slowing down the modulation time of the feed reg valves via the installation of a needle -
valve in the inlet to the valve positioner will provide for more stable steam generator and feed reg valve operation.
The changes to the feedwater system are only for smoother operation, greater system stability and improved system reliability. These changes do not effect the overall feedwater steam generator level control system or feed reg valve control system as described in Final Safety Analysis Report (FSAR) sections 10.4.7, 7.7.1.7, or Fig 7.7 6 and do not change the assumptions used in the Chapter 15 analysis, The addition of these filters (lag cards) do however effect the functional drawing and block diagram as shown in Figure 7.21, SH.13 & 14.
These changes are to nonsafety related equipment. The new cards being added to the control cabinets are being added in spare locations in the card frames. The addition of the cards to cabinets CP 5, 6, 7 and 8 will not affect the seismic qualification of the cabinets.
These changes do not affect the assumed response of the control system as an initial condition for the plant safety analysis or the response of the ,
control system in an accident.
Conclusion:
A 10CFR50.59 safety evaluation was performed.for this design change and it was determined that this change will not create an unreviewed safety question. Changes to the Final Safety Analysis Report will be incorporated by means of a future amendment.
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- 3. Temporary Modifications The below listed temporary modifications were made at Seabrook Station and safety evaluations were performed pursuant to the requirements of 10CFR50.59.
h ,,,orary Modification: Number 90 TMOD 026
Title:
Oxygen Control for the Recovery Test Tanks l
Description:
This Temporary Modification (TMOD) provides oxygen control for . the l Recovery Test Tanks (RTT's) by providing the temporary equipment and j connections necessary to establish a nitrogen blanket. The nitrogen gas is supplied to the recovery test tanks at low pressure by a temporary hose and valving from nitrogen gas cylinders. Small water seal devices are added at the tank vents and overflow pipes. These seals prevent oxygen ingress while at the same time, do not interfere with the tanks ability to vent or overflow. Also, additional connectiens are provided to add water to the seals from the demineralized water system.
Implementation of this TMOD will not increase the probability of an
- accident previously evaluated in the FSAR. The Boron Recovery System (BRS) has no emergency function and is not required for safe shutdown.
The RTT's are located in the tank farm between the waste process building and the primary. auxiliary building in a diked area that does not contain any safety related equipment. The RTT's are not specifically designed to withstand a seismic event. The temporary hoses, valves and gas cylinders are located in areas remote from safety related components.
Conclusion:
A 10CFR50.59 safety evaluation was performed for this- temporary modification and it was determined that this change will not create and unreviewed safety question.
4, Technical Reouirements Manual No changes were made to the Technical Requirements Manual during this reporting period.
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- 5. Final Safety Analysis Report l
l The below listed Final Safety Analysis Report change requests were issued and safety I evaluations were performed pursuant to the requirements of 10CFR50.59. !
Final Safety Analysis Report Change Request: Number 90 012
Title:
Radiation Monitoring System
Description:
This Final Safety Analysis Report (FSAR) Change Request (FC2) modifies FSAR Sections 11.5 and 12.3 to conform it to the as built and designed plant.
The changes to the Radiation Monitoring System description are document changes only to reflect the as built and as designed plant. The changes do not affect the function of the radiation monitors described above.
Conclusion:
A 10CFR50.59 safety evaluation was performed for this FSAR change and it was determined that this change will not create an unreviewed safety question. Changes to the Final Safety Analysis Report will be incorporated by means of a future amendment.
Final Safety Analysis Report Change Request: Number 90 031
Title:
Fiberglass Ladder Additions
Description:
This Final Safety Analysis Report (FSAR) Change Request (FCR) revises
" Fire Protection Program Evaluation and Comparison to BTP APCSB 9.5-1, Appendix A" and "Scabrook Station Fire Protection of. Safe Shutdown Capability (10CFR50 Appendix R)" fire loading analysis due to the addition of combastible fiberglass ladders in various fire areas / zones.
l The addition of these ladders required a new design basis fire to be l
postulated for some fire areas / zones. The new design basis fires and minor changes to combustible loads doe 4 not affect the existing fire protection systems, manual firefighting ctpability, and .does not result in recommendations for new fire protection systems.
Conclusion:
A 10CFR50.59 safety evaluation was performed for this FSAR change and it was determined that this change will not create an unreviewed safety question. Changes to (Le Final Safety Analysis Report will be incorporated by means of a future amendment.
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- 6. Procedures
, No procedt.res which required 10CFR50.59 Safety Evaluations were developed during this reporting period.
- 7. Procedure Changes l
Procedure changes that require review and approval by the Station Operation Review l Committee (SORC) are subject to the requirements of 10CFR50.59. No procedure changes ;
have been made during this reporting period that would require a change to the Final Safety Analysis Report.
- 8. Test or Experiments There were no tests or experiments performed during this reporting period that require evaluations in accordance with 10CFR50.59.
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