ML17329A399
| ML17329A399 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 12/31/1991 |
| From: | Fitzpatrick E, Moran W, Svensson B INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| AEP:NRC:1147B, NUDOCS 9203020147 | |
| Download: ML17329A399 (24) | |
Text
ACCELERATED DISTRIBUTION DEMONSTPWTION SYSTEM REGULATO INFORMATION DISTRIBUTION STEM (RIDS)
ACCESSION NBR:9203020147 DOC.DATE:
NOTARIZED:
NO DOCKET FACIL:50-315 Donald C.
Cook Nuclear'ower Plant, Unit 1,'ndiana M
" 05000315 50-316 Donald C.
Cook Nuclear Power Plant, Unit 2, Indiana M
05000316 AUTH.NAME AUTHOR AFFILIATION FITZPATRICK,E.
Indiana Michigan Power Co. (formerly Indiana a Michigan Ele MORAN,W.R..
Indiana Michigan Power Co. (formerly Indiana
& Michigan Ele SVENSSON,B.A.
Indiana Michigan Power Co. (formerly Indiana a Michigan Ele RECIP.NAME RECIPIENT AFFILIATION MURLEY,T.E.
Document Control Branch (Document Control Desk)
SUBJECT:
"DC Cook Nuclear Plant 1991 Annual Operating Rept."
920226 ltr.
DISTRIBUTION CODE:
IE47D 'OPIES RECEIVED:LTR ENCL SIZE:
TITLE: 50.59 Annual Report of Changes, Tests or Experiments Made W o t Approv NOTES:
RECIPIENT ID CODE/NAME PD3-1 LA STANG,J INTERNAL: AEOD/DOA NRR DLPQ/LHFB11 RE FIL 02 EXTERNAL: NRC PDR COPIES RECIPIENT LTTR ENCL ID CODE/NAME 1
0 ",
PD3-1 PD 1
0 1
1 AEOD/DSP/TPAB 1
1 NRR/DOEA/OEAB11 1
1 RGN3
, FILE 01 1
1 NSIC COPIES LTTR ENCL 5
5 1
1 1
1 1
1 1
1 NOTE TO ALL"RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WASTEl CONTACT THE DOCUMENT CONTROL DESK ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISIS FOR DOCUMENTS YOU DON'T NEED!
TOTAL NUMBER OF COPIES REQUIRED:
LTTR 15 ENCL 13
Indiana I'il'chip~
Power Gompanr P.O Bo~ >66' Coom".o,.C~ ":"
lNDlAMA NlCHIOAR POWER AEP NRC:1147B Donald C.
Cook Nuclear Plant Units 1 and 2
License Nos.
50-315. and 50-316 ANNUAL OPERATING REPORT U.S. Nuclear Regulatory Commission Attn:
Document Control Desk Washington, D.C.
20555 Attn:
T.
E. Murley February 26, 1992
Dear Dr. Murley:
Paragraph 6.9.1.5 of the Donald C.
Cook Nuclear Plant Technical Specifications requires that an annual report be submitted to address personnel
- exposure, steam generator in-service inspection
- results, challenges to power-operated relief and safety valves, and information regarding I-131 activity.
In
- addition, 10 CFR 50.59(b)(2) requires submittal of an annual report describing changes,
- tests, and experiments made in the preceding year.
Consistent with these requirements, attached are two copies of the Cook Nuclear Plant 1991 Annual Operating Report.
This document has been prepared following corporate procedures that incorporate a
reasonable set of controls to ensure its accuracy and completeness prior to signature by the undersigned.
Sincerely, QCc4f)~(<~)
E.
E. Fitzpatrick Vice President edg Attachments cc:
D. H. Williams, Jr.
A. A. Blind - Bridgman J.
R. Padgett G. Charnoff NFEM Section Chief A. B. Davis
- Region III NRC Resident Inspector
- Bridgman ri 9203020147
'rr11231 PDR ADOCK 05000315 R
ATTACHMENT TO AEP:NRC: 1147B DONALD C.
COOK NUCLEAR PLANT 1991 ANNUAL OPERATING REPORT
DONALD C.
COOK NUCLEAR PIANT 1991 ANNUAL OPERATING REPORT February 28, 1992 COMPILED BY:
W. R. Moran REVIEWED BY'.
A. Svensson Executive Staff Assistant APPROVED BY'.
E. Fitzpa ick Vice President-Nuclear Operations
TABLE OF CONTENTS
/ECTION.
SECTION T PAGE
~MR 1.0 Introduction 1.1 1.2 Plant Description Report 'Preparation 2.0 3.0 Personnel Radiation Exposure Summary Steam Generator In-Service Inspection 3.1 3.2 Unit 1 Inspection Summary Unit 2 Inspection Summary 4
4 4.0 Changes to Procedures 4.1 4.2 4.3 4.4 Maintenance Procedures Operating Procedures Chemistry Procedures Environmental Monitoring Procedures 5.0 Tests and Experiments
-10 6.0 Challenges to Pressurizer Power Operated Relief Valves and Safety Valves 7.0, Reactor Coolant Specific Activity 12 8.0 Irradiated Fuel Examinations 13 9.0 Changes to Facility 14 9.1 9.2 9.3 9.4 9.5 9.6 9.7 9.8 9.9 Removal of Obsolete Westinghouse Monitor Removal of Axial Power Distribution Monitoring System Replacement of Liquid Effluent Monitor R-18 Removal of Accumulator Backup Nitrogen Supply Replacement of Automatic Gas Analyzer Removal of Failed Fuel Detector Removal of Local Halon Fire Suppression System Replacement of U-1 Moisture Separator Reheater Tube Bundles Install ArtificialLeak By on SI Pump Discharge Check Valve 14 14 14 15 15 15 16 16 16
1.0 INTRODUCTION
1.1 LANT DESCR IO The Donald C.
Cook Nuclear Plant is owned by Indiana Mich'igan Power Company and is located'five miles north of Bridgman, Michigan.
The plant consists of two nuclear
.power
- units, each employing a
Westinghouse pressurized water reactor nuclear steam supply system.
Each reactor unit employs an ice condenser reactor containment system.
The American Electric Power Service Corporation was the architect-engineer and constructor.
Unit 1 and 2 reactor design power output (and licensed rating) are 3250 MWt and 3411 MWt, respectively.
Unit 1 approximate gross and net electrical outputs are 1056 MWe and 1020 MWe, respectively.
Unit 2 approximate gross and net electrical outputs are 1100 MWe and 1060 MWe, respectively.
The main condenser cooling method is open cycle using Lake Michigan water as the cooling source for each unit.
1.2 EPORT PREPARAT 0 This report was compiled by W. R. Moran with the following individuals contributing information as follows:
D. C. Loope C. A. Freer B. A. Svensson
- Personnel Exposure Summary
- Steam Generator ISI Summary
- Changes to Procedures B. A. Svensson
- Challenges to Pressurizer PORVs and Safety Valves B. A. Svensson B. A. Svensson D. H. Malin R.
G. Vasey B. A. Svensson
- Tests and Experiments
- Reactor Coolant Specific Activity
- Results of Irradiated Fuel Inspections
- Changes to Facility - RFCs,
- MMs, PMs
- Changes to Facility - Temporary Modifications to Unit 1 & 2
~
~
2.0 PERSONNEL RADIATION EXPOSURE
SUMMARY
Table 2 provides a summary of the number of station, utility, and contractor (and others) personnel receiving exposures greater than 100 millirem in 1992.
The total record dose for all personnel was 69.087 rem as measured by thermoluminescent dosimetry (TLD) and reported in accordance with Regulatory Guide 1.16.
2
TABLE ANNUAL OPERATING REPORT - RG.1.16 FOR PERSONNEL >100 mR TOTAL MAN-REM STAT. UTIL. CONT.
STATION UTILITY Reactor Operations,& Surveillance CONTRACT Maintenance Personnel Operations Personnel Health Physics Personnel Supervisory Personnel Engineering Personnel 0007 0023 0026 0002 0001 0000 0000 0000 0000 0000 0004 0001 0005 0000 '026, 001.308 000.000 005.594 000.336 008.514 000.000 000.494 000.000 000.161 000.000 001.285 002.001 008.108 000.000 000.000 Routine Maintenance Maintenance Personnel Operations Personnel Health Physics Personnel Supervisory Personnel Engineering Personnel 0014 0002 0003 0000 0001 0000 0000 0000 0000 0000 0031 0008 0022 0000 0000 002.796 000.000
.000.299 000.000 000.763 000.000 000.000 '00.000 000.572 000.000 008.659 002.169 005.704 000.000 000.000 In-'Service Inspection Maintenance Personnel Operations Personnel Health Physics Personnel Supervisory Personnel Engineering Personnel 0000 0000 0000 0000 0000 0000 0000 0000 0000 0000 '000
'0000 OOO0 OOOO OOOO 000.000 000.000 000.000
. 000.000 000.000 000.000 000.000 000.000 000 000 000.000 000.000 000.000 000.000 000.000 000.000 Special Maintenance Maintenance Personnel Operations. Personnel Health Physics Personnel Supervisory Personnel Engineering Personnel 0000 0000 0000 0000 0000 0000 0000 0000 0000 0000 0044 0004 0000 0000 0001 000.000 000 000 000.000 000.000 000'000 000.000 000.000 000.000 000.000 000.000 017 183 000.808 000.000 000.000 000.182 Waste Processing Maintenance Personnel Operations Personnel Health Physics Personnel Supervisory Personnel Engineering Personnel 0001 0000 0000 0000 0000 0000 0000 0001.
0002 0000 0004 0000
.0000 0000 0000 000.154 000.000 000.000 000.000 000.000 000.000 000.000 000.000 000.000 000.000 000.512 000.656 000.830 000.000 000.000 Refueling Maintenance Personnel Operations Personnel Health Physics Personnel Supervisory Personnel Engineering Personnel 0000 0000 0000 0000 0000 0000 0000 0000 OOOO O0OO 0000 0000 0000 0000 0000 000.000 000.000 000.000 000.000 O0O.OOO 000.000 000.000 000.000 000.000 000.000 000.000 000.000 000.000 000.000 000.000
-TOTALS Maintenance Personnel Operations Personnel
'ealth Physics Personnel Supervisory Personnel Engineering Personnel GRAND TOTALS 0017 0025 0026 0002 0002 0072 0000 0001 0000 0000 0000 0001 0071 0015 0047 0000 0001 0134 004.258 005.893 009 '77 000.494
, 000.733 020.655 000.000 000.336 000.000 000.000 000.000 000.336 027 '39 005.633 014.642 000.000 000.182 048. 096
3.0 STEAM GENERATOR IN;SERVICE INSPECTION 3.1 UNI NSPECTIO S
During 1991, there-were no steam generator in-service inspections performed for Unit 1.
3.2 UNIT 2 NSPECTIONS
.During 1991, there were no steam generator in-service inspections performed for Unit 2.
4
4.0 CHANGES TO PROCEDURES This section contains a brief description of the procedure changes implemented under the provisions of 10 CFR50.59 and the associated safety evaluations.
4.1 ENANCE PROCEDURES 4.1.1 ocedure o
"Disassemb Va ves" 502 0
00 ev s 6
C S
e ai a d assemb of o v C am seal Description of Change:
This maintenance procedure change was implemented to permit the use of "Chesterton" valve packing on safety related valves in systems described in the UFSAR.
The Chesterton valve packing differs from the valve packing configurations described in the UFSAR for certain safety related valves.
Safety Evaluation Summary:
A technical evaluation determined that the packing system described in the UFSAR has been rendered obsolete by advances in technology.
The Chesterton system of packing currently being installed in the majority of the safety related valves is a direct development of an EPRI study and is technically superior to that described in the UFSAR.
The safety evaluation determined that the changes do not involve an unreviewed safety question as defined in 10 CFR50.59.
Changes to the UFSAR willbe incorporated in the next annual update.
4.1.2 ocedure o
P4030 S
0 evis on 2
C S
16 ea to SPS c an eacto cake a n "A" u
e la ce est o th Description of Change:
The change sheet to the above procedure was written to allow the use of backup manual testing for the solid state protection system (SSPS).
The change was necessary because the semi-automatic
- tester, which is described in the
- UFSAR, appears to be malfunctioning and providing ambiguous test results for some logic circuits.
The change to allow manual testing is a change to a procedure described in the UFSAR.
The SSPS was designed for either automatic or manual testing.
Safety Evaluation Summary:
The changes do not affect the function of any safety related equipment.
The safety evaluation concluded that the changes to the procedure do not constitute an unreviewed safety question as defined in 10 CFR50.59.
4.2 OPERATI G
P OCEDURES 4.2.1 ocedure No 1-OHP4021 0
8 01 evisio ocedu e No 40 8'01 ev o
0 - "H h Veloc ush o
Containment Lower Com artment Ve t lation U its and CP oto Ai Coolers" Description of Change:
The above procedures were implemented for the purpose of performing high velocity flushes of the containment lower compartment ventilation units and the RCP motor air coolers.
Chapter 5.5 of the UFSAR states that the maximum non-essential service water flow to the lower compartment ventilation units is 440 gpm.
During the flushing procedure, the flow will be greater than 440 gpm.
Safety Evaluation Summary:
Section 9.8.3.2 of the UFSAR describes the containment lower
~ compartment ventilation units and the RCP motor air coolers as components having no safety-related functions.
The safety evaluation concluded that implementation of these procedures will not involve an unreviewed safety question, as defined in 10 CFR50.59.
4.2.2 roc'edu e No 0
40 05 008 Rev sio 0 - "
e at o
o t e C rculatin Water S stem to crease Foreba Te e atures" Description of Change:
The above procedure was implemented for the purpose of operating the circulating water system in a
manner which increases the forebay temperature above the lake temperature.
The procedure willbe used for the purpose of activating zebra
., mussels so that a mulluscicide treatment can be performed.
Safety, Evaluation Summary:
The review focused on the impact of the change on the essential service water system, since it is the only safety-related system directly impacted by the procedure.
The procedure contains a precaution that the temperature. of the condenser inlet water cannot exceed 87 F.
The unreviewed safety question determination concluded that the procedure does not constitute an unreviewed safety question as defined in 10 CFR50.59.
4.2.3 rocedure o
-0 40 4 102 Revision 4 C
S 6
¹102 Res onse' seel aneous Aieas Fire S stem" Description of Change:
Annunciator ¹102 Response procedure was changed to reflect the removal of the Halon fire, suppression system from the office 6
building fourth floor Q/C vault.
The change in the Halon fire suppression system was implemented under Plant Modification No.
1-PM-744.
The area is no longer being used as a
Q/C vault.
Safety Evaluation Summary:
The portion of the Halon system being removed does not affect nuclear safety.
The safety evaluation concluded that removal of the Q/C vault Halon system doe's not involve an unreviewed safety question as defined in 10 CFR50.59.
rocedure o
2-OHP P 09 evis on 0
" ais e cto ower w t ocedu e
o eedwater Delta-P Si al" 6030 4
Rev sion 0
C S
2 - "Mai eed eed Co t o Ca b ation Al ent" Description of Change:
A problem had developed with steam generator
¹21 feedwater regulating valve, 2-FRV-210, sticking when the valve was open greater than 60%.
A temporary solution was to operate with the regulating valves to all four steam generators in a less open position.
This could be accomplished by increasing the differential pressure across the valves by operating the feedwater pumps at a higher speed.
The total feedwater flow to the steam generators would remain the same.
Special Procedure 2-OHP SP,092, Revision 0 was implemented to allow raising reactor power from approximately 90% to 1008 while operating with the elevated feedwater differential pressure and collecting data for establishing the revised differential pressure program.
The new differential pressure program would be determined based on the feedwater regulating valves operating approximately 50% open at 100% power.
Procedure 2-IHP6030.IMP.422 was changed to implement the recalibration of'the differential pressure program controller to,the new values.
Safety Evaluation Summary:
Operation of the feedwater regulating valves as described above may result in a maximum flow exceeding that'used in the safety analysis for a failed open feedwater regulating valve.
The UFSAR contains an analysis of a feedwater regulating valve malfunction using a flow of 150% of nominal.
Thus operation in this mode will result in a change to the facility as described in the UFSAR.
The safety evaluation was performed assuming a maximum flow of 200% of nominal, a value judged to be bounding.
A feedwater malfunction is evaluated against departure from nucleate boiling (DNB) criteria (a minimum ratio of 1.38 to 1.607, depending on the location).
The evaluation concluded that the 7
~
DNB criteria are met with a flow of 200% of nominal.
The unreviewed safety question determination concluded that implementation of this procedure does not constitute an unreviewed safety question as defined in 10 CFR50.59.
4.3 CHEMISTRY PROCEDURES 4.3.1 ocedu e
60 0 04 ev s on 16 - "Data She t Description of Change:
One of the changes implemented by Revision 16 involves an increase in the upper limit of Lithium-7 (Li-7) concentration in the reactor coolant system (RCS) recommended by Westinghouse which will require a
change to the UFSAR.
Specifically, the UFSAR gives a
range of RCS Li-7 concentration of 0.22 to 2.2
- ppm, whereas the new proposed range received from Westinghouse is 0.20 to 3.8 ppm.
Safety Evaluation Summary:
An increase in the Li-7 concentration in the RCS can theoretically increase the amount of tritium produced in the primary system.
The Li-7 is introduced to the RCS in the form of lithium hydroxide, which is used in controlling primary system pH.
An evaluation of the impact on RCS tritium inventory that could result from a higher Li-7 concentration in the RCS was performed.
The evaluation concluded that, from an operational standpoint, the proposed increase in the allowable range of RCS Li-7 concentration would still result in total tritium activity in the RCS well below the Li-7 concentration limit of 2.51Ci/cc given in the UFSAR.
The safety evaluation concluded that increasing the allowable range of RCS Li-7 concentration will not increase the dose consequences of an accident and that the change does not involve an unreviewed,safety question as defined in 10 CFR50.59.
4,4 RO L
0 0
G PRO UR S
OCED E
0 P60 0 ENV 06 REVIS ON 1
"COLLECTION OF RINK NG WATER SAMPL S FROM PUBL C NTAKES" Description of Change:
The above procedure was changed to reflect changes in the Technical Specification requirements for obtaining drinking water samples as part of the routine radiological environmental monitoring program.
'The revision deleted two sample locations (Benton Harbor and New Buffalo) which were still identified in the then current Emergency Plan.
The Emergency Plan and implementing procedures have since been changed to reflect the current NRC approved sample locations.
8
Safety Evaluation Summary:
The procedure changes apply only to the collection of drinking water samples as part of the radiological environmental monitoring program for normal plant operations, and weie implemented after receiving NRC approval via technical specification Amendment 94 for Unit 1
and Amendment 80 for Unit 2.
The safety evaluation concluded that implementation of the changes does not represent an unreviewed safety question as defined in-10 CFR50.59.
5.0 TEST OR EXPERIMENTS NOT DESCRIBED IN THE FSAR This section describes procedures classified as a
"Test and Experiment" implemented under the provisions of 10 CFR50.59 including the associated safety evaluation.
During 1991, there were no procedures classified as a "Test and Experiment" implemented.
10-
6.0 CHALL12IGES TO PRESSURIZER POWER OPERATED RELIEF VALVES AND SAFETY VALVES During 1991, there were no challenges on either Unit 1 or Unit 2 to the pressurizer po~er operated relief valves (PORV's) or the pressurizer safety valves as a result of the valves being called upon to mitigate an actual overpressure condition during 1991..
I 7.0 REACTOR COOIANT SPECIFIC ACTIVITY During 1991, there were no instances on either Unit 1 or Unit 2 in which the reactor coolant I-131 specific activity exceeded the limits of Technical Specification 3.4.8.
12
- 8. 0 IRRADIATED FUEL EXAMINATIONS During 1991, there were no examinations performed on the irradiated fuel due to not having any scheduled refueling outages.
0 9.0 CHANGES TO FACILITY This section contains a brief description of the design changes implemented under the provisions of 10,CFR50.59 and the associated safety evaluations.
9.1 emoval of Obso ete West ouse onitor Description of Change:
RFC-DC-12-4078 removed Westinghouse and'MC radiation monitors which had been replaced by Eberline units, but left in place.
The monitors were R.25/26 (unit vent stacks air particulate),
R.2 (upper containment area),
R.33 (turbine gland steam condenser vent) and R.31/32 (unit vent stack effluent iodines and noble gases).
The change consisted of removing cabinets and disconnecting electrical leads.
Safety Evaluation Summary:
This change was reviewed and it was concluded that it did not represent an unreviewed safety question.
This conclusion was based on the fact that the
~ change simply removes equipment that has been replaced by another brand.
The function of the system has not been changed.
9.2 emova of ia owe Distr bution Monitorin S stem Description of Change:
RFC DC-12-2937 removed the axial power distribution monitoring system components from their control room racks and removed the system annunciators.
Safety Evaluation Summary:
This change was reviewed and it was concluded that it did not represent an unreviewed safety question.
This conclusion is based on the fact that the subject system was only a monitoring system that was not used for any safety related function but which had once been referred to in the Technical Specifications.
The NRC had approved Technical Specification changes which had deleted the reference to this system.
9.3 e
a erne t o L
u ue t o
to
-18 Description of Change:
RFC DC-12-2853 replaced an existing monitor with a more sensitive monitor.
The R-18 monitor is used to monitor the release of liquid effluents from the waste disposal system.
Should the discharge exceed a preset limit, R-18 provides a signal which closes the discharge valve.
Safety Evaluation Summary:
This change was reviewed and it was concluded that it did not represent an unreviewed safety question.
This conclusion was based on the fact that the function of R-18 remains the same'and the increased sensitivity of the replacement monitor enhances the system's capabilities.
9.4 emoval of Accumulato acku t
e u
,Description of Change:
PM 12-491 removed a four-bottle backup nitrogen supply from the accumulator nitrogen space.
The backup supply was normally isolated from the accumulator, with nitrogen still supplied to the accumulators (an intermittent process) by the main nitrogen
- system, a six tank bank.
Safety Evaluation Summary:
This change was reviewed and it was concluded that it did not, represent an unreviewed safety question.
This conclusion was based on the fact. that the nitrogen system performs no active function during an accident, and a nitrogen supply to the accumulator continues to exist.
9.5 e
acement of utomatic Gas Anal zer Description of Change:
RFC DC-12-2591 replaced the automatic gas analyzer sampling panel, connected the new panel to the nuclear sampling room ventilation system and modified a water drain line.
The automatic gas analyzer is, used to measure oxygen and hydrogen concentrations in the gas decay tanks.
Its replacement was necessary because major components were obsolete and could not be repaired.
Safety Evaluation Summary:
This change was reviewed and it was concluded that it did not represent an unreviewed safety question.
This conclusion is based on the fact that the change involved the replacement of existing equipment with comparable equipment performing the'same function.
9.6 emoval of Failed el Detecto Description of Change:
RFC DC-12-2911 involved removal of the failed fuel detector (FFD) because of difficulties in maintaining the calibration of the system.
The changes consisted of cutting and capping of lines, removal and disposal of the failed fuel detector and associated instrumentation, disconnecting the power supply to the FFD and removal of computer inputs from the FFD.
Safety Evaluation Summary:
This change was reviewed and it was concluded that it did not represent an unreviewed safety question.
This conclusion was based on the fact that the FFD is a monitoring device and its function can be accomplished by the analysis of periodically obtained grab samples.
The use of grab samples had been approved by the NRC in response to our Regulatory Guide 1.97 deviation request submitted via letter AEP:NRC:0773AB.
9.7 e oval o c
e u
e s
Description of Change:
PM 1-744 removed the Halon tanks from a fire suppression system in a former Quality Control (QC) records vault in the office building.
Because of several concerns with the use of, the fluorocarbon Halon, its use is being minimized at the Cook Nuclear Plant.
The room associated with the bottles being removed was no longer used to store QC records, and the Halon suppression system was no longer required in that area.
Safety Evaluation Summary:
This change was reviewed and it was concluded that it did not represent an unreviewed safety question.
This conclusion was based on the fact that no safety-related equip'ment was impacted by the removal of the Halon system from the office building area.
9.8 e lacement o
-1 oisture Se a ator Reheat be Bu d es Description of Change:
PM 1-803 replaced the Unit 1 reheater tube bundles with an improved
- design, installed perforated plates and improved chevrons in the moisture separator, and modified the moisture separator drain lines.
Safety Evaluation Summary:
This 'change was reviewed and it was concluded that it does not represent an unreviewed safety question.
This conclusion was based on the fact that the components which were modified are non safety related and the changes did not impact any safety related Components or systems.
9.9 stall Art i a
Leak B
on SI Pum Disc a e Check Va ve Description of Change:
MM 2-166 installed a bypass line around the south safety in)ection (SI) pump discharge check valve.
This modification prevents pressure buildup in the SI pump headers (caused by leaking check valves) by allowing a
small quantity of fluid to flow through the SI pump miniflow line to the refueling water storage tank.
16
Safety Evaluation Summary:
This change was reviewed and it was concluded that it does not represent an unreviewed safety question.
This conclusion was based on the fact that'the modification does not impair the capability of the SI pumps to provide adequate cooling water flow in the event of an accident.
- 17