ML19332D663
| ML19332D663 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 12/31/1988 |
| From: | Fernandez W POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| JAFP-89-0820, JAFP-89-820, NUDOCS 8912050094 | |
| Download: ML19332D663 (184) | |
Text
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, g James A. Fit:Petenck Nucteer Power Plant P6. Box 41 .
l LycommD. Ne'3 Y rk 13093 315 342-3040 William Femander ll ;
Resident Manager i- November 16, 1989 -
p JAFP 89-0820 i- 'I t
l United States Nuclear Regulatory Commission Mail Station PI-137 p Washington, D.C. 20555
! Attention: Document Control Desk p'
SIIBJECT: JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET No. 50-333 ANNUAL SUMMAR'l 0F CHANGES, TESTS, AND EXPERIMENTS FOR 1988
Enclosure:
- 1) Annual Summary of JAFNPP Changes, Tests, and g Experiments for 1988 t( Gentlement Enclosed is a summary of the changes, tests, and experiments implemented at the James A. FitzPatrick Nuclear Power Plant during 1988.
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This report provides the Nuclear Safety Evaluation number (e.g.
(' JAF-SE-88-001) followed by a brief description of the correspond-ing change, test, or exaeriment and safety evaluation. summary as L required by 10CFR50.59('a)(2).
Very truly yours,
/
f WILLIAM FERNAND 2 WF: 1s ENCLOSURE CC: R. Liseno (w/o Enc.)
JAF Resident Inspector (w/ Enc.)
J. Gray, WPO (w/ Enc.)
R. Beedle, WPO (w/ Enc.)
RMS, WP0 (w/ Enc.)
TS File (w/o Enc.) g DCC (w/ Enc.)
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'2050094 891231 ADOCK 05000333 /
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ANNUAL:
SUMMARY
0F ' CHANGES, TESTS, AND EXPERIMENTS COMPLETED AT THE JAMES A. FITZPATRICK' NUCLEAR POWER PLANT DURING 1988-u ,
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p NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET No. 50-333 a iY
r e a Annual Summ ry of JAFNPP Ch;ngas, Tosts, and Experiments for 1988 JAF-SE-83-065 - Miscellaneous Fence and Gate Installation (Minor Modification M1-83-037)
This non-safety related minor modification (M1-83-037) added a gate between the southwest corner of the Waste Surge Tank Shield Wall and the Turbine Building, and relocated the existing chain link fence southeast of the Condensate Storage Tanks.
Safety Evaluation JAF-SE-83-065 contained details concerning this modification which included sketches showing the new location of the fence and the new gate. This work was performed in order to prohibit unauthorized diation areas. personnel access to potential high ra-These changes enhanced the site "ALARA Program and did not involve an unreviewed safety question as defined by 10CFR50,59.
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Annual Summ2ry of JAFF . Ch ngas , Tcsts, cnd Exparim;nts for 1988 JAF-SE-85-118, Rev. 2- Second Level of Undervoltage Protection (Plant Modification F1-84-041)
This safety evaluation provided a detailed review of the possible safety concerns associated with Plant Modification F1-84-041.
This safety-related modification installed a second level of undervoltage protection on the 4160 KV emergency buses. The primary purpose of this modification was to increase the range of protection for safety-related equipment from abnormal voltage conditions.
This modification consisted of (1) the addition of test devices and status indicators to the auxiliary relays of the existing loss of Voltage Protection System which will automatically initiate transfer of the Emergency AC Power System to the on-site Emergency. Diesel Generator (EDG) source, (2) the addition of two channel degraded voltage protection (including voltage relays, !
timers, auxiliary-relays, test devices, alarms and status indica-tors) which in the event of a sustained voltage level below the operating camability of the safety-related loads connected to these buses but above the setting of the loss of voltage pro-tection will automatically transfer the Emergency AC Power System to the EDG sources, (3) the modification of the logic to prevent i the shedding of the emergency bus motor loads when the EDGs are !
supplying the power to the bus, and (4) the adjustment of the primary taps on the reserve station transformers to increase the emergency bus voltage to a nominal value of 4160 volts.-
This evaluation addressed equipment capability, the effects of the modification on the performance of interfacing ecuipment and systems , adherence to the appropriate codes , standarc.s , and-regulations, modification quality assurance and control, and j changes to the Technical Specifications.
The conclusion of this safety evaluation was that an unreviewed l safety question as defined in 10CFR50.59 did not exist as a i result of this modification. The safety analysis for the preop- !
erational tests associated with this modification were contained in Nuclear Safety Evaluations JAF-SE-85-121 and JAF-SE-87-153.
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Annual Summsry of JAFNPP Chnngas, Tcsts, and Experiments for 1988 JAF-SE-85-121 - Second Level of Undervoltage Protection o Modification Checkout and Functional Test
- (Preoperational Test / POT-93A)
This-safety evaluation reviewed the safety issues associated with the implementation of Preoperational Test POT-93A. This test was conducted to verify the operability of the. revised loss of voltage protection scheme and the new degraded voltage protection system. The operability of the Emergency Diesel Generator System was also tested to ensure the entire emergency AC power system functions properly. This test was performed as mart of the safety-related Plant Modification F1-84-041 which was reviewed in Nuclear Safety Evaluation JAF-SE-85-118, Rev. 2.
Specific' conditions were outlined in the procedure in order to ensure compliance with the Technical Specifications. Operations Surveillance Test Procedure No. F-ST-9C, " Emergency AC Power Load Sequencing Test and 4KV Emergency Power System Voltage Relays Instrument-Functional Test", was used as a basis in the develop-ment of POT-93A.
It was concluded that the performance of this test did not constitute an unreviewed safety question as defined in !
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, , .'Annucl_SummaryoffJAFNPPChtnges,Tosts, and Expr.riments for 1988 ;
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tu i JAF-SE-852127,JRev. 1- Standby. Gas Treatment (SGT) System Train A&B Filtration Upgrade (Plant Modifica- s tion.F1-85-045)-
This' safety 1 evaluation'provided the analysis of the~ safety.
concerns associated with Plant Modification F1-85-045. This
' safety-related modification consisted of (1) raising'the setpoint '
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of the air. heater temperature switches ( 01-'125TS-100A&B) to
' eliminate nuisance SGT System trips, (2)-sparing the carbon 7
drying heaters whose' function was no longer required and were deemed to be a. potential fire hazerd, (?) installing remote
." reset" pushbuttons:in the Control-Rocm to eliminate the time delay in:. resetting the SGT air heater control circuits 'after the air heaters trip, (4) installing time delay relays to allow for the-automatic resetting of the heater control circuits following l
a loss of, and subsequent restoration of power, and (5) sparing the' humidity transmitters which-were no longer required.
These changes were performed to help eliminate SGT System' trips E and improve overall operability of_the SGT System. The time L delay' relays were installed to preclude the common-mode failure ,
l discussed in NRC Information Notice 85-63, " Potential for Com- -
mon-Mode Failure of the SGT System on Loss of Off-Site Power".
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, l Included in this safety evaluation was a review of this modifitn-
- tion's. impact on plant systems:and' documents.
It was determined that only Section 5.3 of the FSAR was required to be revised to shoe the deletion of the heaters and control circuits.- There was l
no effect on the Technical Specifications. This modification did not change the system in a manner that can change the existing -
design basis and was determined to not constitute an unreviewed safety question as defined in 10CFR50.59.
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- The_ safety analysis for the preoperational test associated with this modification was contained in Nuclear Safety Evaluation
'JAF-SE-88-I"3.
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. i' Annual Summery of JAFNPP Ch nges, Tosts,.
end Experiments ~for 1988 l
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JAF-SE-85-133, Rev. 1- Steam-Leak Detectiv a Sensor Setpoint
-Change-in the Residual Heat Removal (RHR)_ Heat Exchanger Rooms (Minor.
Modification M1-85-075) ;
This safety evaluation reviewed the safety concerns associated with the raising of the temperature setmoints of the four Resis-tance Temperature' Detectors (RTDs) which function as steam leak Jetection1 sensors on the High Pressure Coolant Injection'(HPCI) h 2nch steam lines. The setpoint was' changed to-account for the actual maximum ambient temperatures associated with the areas where these RTDs are located.
A review of the High Energy Line Break (HELB) analysis determined that this slight setpoint change-would'have a-negligible effect-on the conclusions of the HELB analysis. This modification.wasi designed to eliminate the potential for inadvertent triaping of the Steam Leak Detection System during;the hottest months of the 3 year.
It was concluded that this safety-related modification did not
-have an effect on the FSAR or the Technical Specifications. This modification did not constitute an unreviewed safety question pursuant to 10CFR50.59.
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Annusl Summ ry of JAFNPP Changes, Tests,
.and Experiments for-1988 t'
- JAF-SE-86-060, Rev. 1- Valve Replacement - 46ESW-1A, IB and 10RHR-14A, B, C, and D (Minor Modifica-tion M1-86-035)
-This modification realaced six 12" Velan check valves with'12" Atwood and Morrill check valves. The new check valves were procured to meet or exceed the technical requirements of the original' purchase specifications for these safety-related valves, i
This cafety evaluation determined that_the original pipe stress-analysia for the Service Water and Residual Heat Removal Systems -
would not be-affected by this substitution since the replacement val.ves weigh less.
The FSAR and Technical Specifications were reviewed and it was determined that this. modification did not affect these documents.
The existing design' basis for the affected systems was.not .
changed as a result of this modification. l 1
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~Annu:1 Summary of JAFNPP Chcnges, Tosts, and Experiments-for 1988 '
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JAF-SE-86-125 - Turbine Building Crane Trolley and Main Hoist "
. Speed Reduction (Plant Modification F1-86-005)
, .This safety evaluation provided a review of Plant Modification
.F1-86-005. This non-safety.related modification reduced the Turbine Building crane trolley speed by utilizing a slower.
two-speed motor in place of the existing five-speed motor. In addition, this modification-added a micro-drive in tandem with :;
the existing main hoist drive to provide a low speed control mechanism.- This modification was performed in order to facili-tate maintenance activities.
The FSAR and Technical Specifications were reviewed with regard to this modification and it was determined that this modification did not have an impact on either of these documents. This '
modification did not alter the structure or load. rating of the-crane.nor did it affect any safety-related systems.
The conclusion of this safety evaluation was that an unreviewed-safety question pursuant to 10CFR50.59 did not exist as a result of the changes associated with F1-86-005.
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3 Annusl'Summ ry of JAFNPP Changes, Tests,-
and Experiments for 1988 >
- p JAF-SE-86-130 - Emergency Diesel Generator (EDG) Room Ventilation Capacity Test (Pre-Operational Test No. POT-92A)
The purpose-of this test was to determine the maximum ambient temperature associated with the extended operation of'two adja-cent safety-related EDG's walle fully loaded. The data collected was required in order to perform an analysis of the EDG Ventila-tion System capacity.
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This. safety evaluation addressed.the safety concerns associated with the performance of POT-92A.- Provisions were included in the-test procedures to ensure compliance with JAF Technical Speci- 1 fications concerning-EDG operation and availability. If'an accident had occurred during the performance of this test the- i design and safety functions of the JAF Emergency Pcwer System '
would'not be compromised. The evaluation concluded that the performance of POT-92A did not constitute an unreviewed safety
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and Experiments for 1988 ,
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-JAF-SE-86-141'- Reactor Water Clean Up (RWCU) Containment [
. Isolation Upgrade Provision Check Out and
. Functional Test (Preoperational Test POT-12B) l The purpose of this preoperational test was to ensure the correct operation of;the newly? installed valve 12MOV-69. :In' addition, the containment' isolation signals on high Drywell pressure to valves 12MOV-15, 18, and 80 were tested. Safety-related'Modifi-
- cation F1-87-068, which made these changes, was reviewed in Nuclear Safety Evaluation JAF-SE-87-133.
Functional tes' ting of the control and monitoring circuits was performed. .This test verified the proper operation of the logic for containment isolation by-simulating isolationLsignals and checking valve closure.
POT-12B was scrutinized to' account for testing. errors, and it was
. determined-that-no single error could damage more than one safety division. This was in compliance with Safety Design Bases of-the
-FSAR, Section 8 (Electrical Systems). This test was-conducted when the plant was in the cold condition, the RWCU System was out of service and boundary valves were closed to preclude inadver-tent flow.
This-safety evaluation concluded that an unreviewed safety question was not involved due to the execution of this test.
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AnnuslLSummary of JAFNPP Chingts,-Tests,
- . and Experiments for 1988 JAF-SE-86-146.- ' Electrical Changes to Linearize the Input to 01-107FRIC-108 - Off-Gas Flow Indicator
.(Minor Modification M1-86-022)
The purpose 1of this non-safety related modification was to
. provide for the accurate indication of the off' gas flow indicator at both high and low flow ranges. This consisted of' minor wiring je ' changes and: instrument <recclibration in order to establish a j
' linear. flow ^ monitoring range.. i 1This modification'did not change the d- 1ses for lis system, but did ' improve : system op" bf . e a' roviding more accurate indication'of off-gas.f' r s no change recuired'to the FSAR or Technie1 ons as a result of '
th?s modification.
It was determined that tF m2- .ution did ~ constitu '
an unreviewed safety que 1 in 10CFR' L 4
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- .* a Annual Summary of'JAFNPP Chengos, Tests, .
and Experiments for 1988 JAF-SE-86-148 - Fire Protection System for the Meteorological Modeling and Radiological Assessment System (MMRAS) Computer Room (Preoperational Test POT-76X) t
-The purpose of this safety evaluation was to provide a review of the possible safety issues associated with the implementation of ,
Preoperational-Test POT-76X. This test verified the proper functioning of the Fire Protection System for the MMRAS Computer Room. A complete discharge test was performed to verify the specified Halon concentration was achieved upon actuation. _The correct operation: of the Purging System was. verified to ensure; the. removal of Halon from the Computer Room. ,
-This test id not affect any safety-related equipment or systems.
The release of Halon into the MMRAS Computer Room was not con-3 sidered.to be a personnel safety hazard because of the low initial concentration used and the 30 second period between the
. pre-release alarm and the actual Halon release. Neither the FSAR nor the Technical Specifications were affected by this test.
This evaluation determined that an unreviewed safety question pursuant to 10CFR50.59-did not exist as a result of the perfor-mance of POT-76X.
Safety Evaluation JAF-SE-S6-149' reviewed.the non-safety related modificaticn (F1-86-062) which installed this Fire Protection System.
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.JAF-SE-86-1491- Fire Protection System for the Meteorological Monitoring and Radiological Assessment System (MMRAS) Computer Room (Plant Modification .'
F1-86-062)
This' safety. evaluation reviewed the non-safety related' Plant .
Modification F1-86-062 which installed the MMRAS Computer Room's
-Fira_ Protection System. This-modification installed two Halon cylinders along with the necessary piping, valving, and electri-cal components required to provide _ proper fire suppression in the i RF event of a fire.
This Fire' Protection System is not required for the safe shutdown of the_ plant and the Technical Specifications were not affected.
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'Section 9.8 of the.FSAR required revision to reflect the addition of this fire' protection equipment.
'This evaluation. concluded that this modification enhanced plant safety and that an unreviewed safety question as defined in 10CFR50.59-did not exist as a result of its implementation.
Safety' Evaluation JAF-SE-85-148 evaluated the preoperational test associated with this modification.
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JAF-SE-86-156, Rev. 1- Replacement of Solenoid Pilot Valves on 1
Unit Coolers in Ventilation System 67 (Minor Modification M1-86-075)
This' safety evaluation was written in order to evaluate Minor Modification M1-86-075. This non-safety related modification allowed for the replacement of the' solenoid-operated pilot valves far-the' air-operated dampers on various unitfcoolers. _The replacement valves were determined to meet or exceed the original !
design requirements for the desired application.-
The unit coolers affected pc.rform no safety-related~ functions.
Failure of a scienoid valve to operate would only result in an air discharge damper.not opening fully,: which would impede the air flow and reduce the capacity and efficiency-of the unit.
The' conclusion of the safety evaluation was that this type of valve replacement did not constitute an unreviewed safety ques-tion pursuant to 10CFR50.59. No change to the FSAR was required as a result of this modift:ation.
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- JAF-SE-86-158 - Control Rod Drive (CRD) Suction Piping' Relief valve Replacement (03RV-7A,B)_(Minor Medi-fication M1-86-076) l This. safety evaluation provided a review for the replacement of the CRD' suction piping relief valves. The replacement valves.
were determined to meet or exceed the original purchase speci-
- fications for this valve application. This non-safety related modification allowed for the replacement of these valves on an as-needed basis, l The slight increase in weight of the new valves was determined to 1 be well within the capability of- the existing supaorts. These l components do not' perform a safety function and this replacement does.not' create the possibility of an accident or malfunction of l a type other than previously evaluated in the FSAR.
It was concluded in this evaluation that.an unreviewed safety question pursuant to 10CFR50.59 did not exist as a result of this modification.
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,, and Experiments for 1988' a .
JAF-SE-86-l'651- Automatic Depressurization= System (ADS) i Pneumatic Supply System Upgrade (Preopera-tional Test POT-27T)
The purpose of this Preoperational Test-(POT-27T) was to ensure '
proper operation of the valves and instruments installed for the '
ADS Pneumatic Supply System Up,7,rade (Plant Modification: ;
F1-84-072), prior to placing the system in service.- An analysis
( of the safety-related Modification F1-84-072 was provided in Safety Evaluation JAF-SE-86-166, Rev. 5.
'This functional test verified manual' valve operation ~from the .
'27 CAD panel-and automatic valve operation by? simulating process and containment isolation signals. It waa determined:that no single error-during the performance of this test:could damage-more than one safety division which is in compliance'with the.
Safety Design Bases of FSAR, Section 8. It was' concluded that an unreviewed safety question as defined by 10CFR50.59-did not exist
-due to this test.
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.-Annual Summcry of JAFNPP Chengas, Teets, ik '
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-and Experimants for 1988.
- JAF-SE-S6-166,.Rev.:5 -
Automatic Depressurization' System (ADS)
Pneumatic Supply System Upgrade (Plant Modification F1-8f-072) :
This. safety-related modification was implemented in order to ensure.adequata pneumatic supply to ADS valves for 100 days after 4 an accident as required by NUREG-0737,.Section II.K.3.28. The changes; accomplished included: (1) installation of a new seis-mically designed pneumatic supply line to the pneumatic supply headers (2)Lupgrade of the existing pneumatic supply line to the header,to seismic Category I criteria along with all of.the piping from the header to the air operated components in the
-Drywelli (3) installation of solenoid-operated containment '
isolation valves in each of the two pneumatic supply' lines to the
. headers (4)11nsta11ation of a scienoid-operated isolation valve
-(27SOV-144)-at the bcandary of the seismic'and non-seismic piping of'the Containment Atmosphere Dilution (CAD) Systems (5) removal L of the primary' containment isolation signal from the existing L cross-tie valves since 27S0V-144 now provides the seismic /non-D seismic boundary (6) replacement of components with seismically L and environmentally qualified equipments and (7) installation of valve position' indication for the three new solenoid-ope ated valves.
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All' lines which penetrated containment were designed in accor-dance with General Design Criteria 57 for containment pene-trations. This safety evaluation detailed the actions taken to ensure compliance with the requirements of Appendix R, Regulatory Guide 1.97, and JAF Technical Specifications.
The in-depth review presented in this safety evaluation provided the basis for the conclusion that this modification did not constitute nn unreviewed safety question pursuant to 10CFR50.59. -
The safety analyses for the preoperational tests (POT-27T and POT-27U) associated with this modification were contained.in Nuclear Safety Evaluations JAF-SE-86-165 and JAF-SE-87-110, Rev.
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F <C Annual; Summary of JAFNPP.Chnngos Tests, and-Experiments for 1988
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ee JAF-SE-86-191 - 87P-27 (Glycol Transfer Pump) GE Type SBM '
Control 1 Switch --Equipment Substitution' *
(Minor Modification M1-86-130)'
.The purpose'of this non-safety related minor modification was to replace.the :ontrol-switch on the. glycol transfer pump.- This was
-necessary since the old control switch was no longer manufac- .
tured. !
This safety evaluation addressed this modification's affect on '
plant drawings and procedures. This equipment ~ substitution had no adverse impact on the operation of the-glycol transfer-pump and'did'not affect the design bases for this system. It was determined that this vendor recommended switch substitution was acceptable upon. review cf the switch characteristics and the desired dssign application.-
Thefconclusion'was that an unreviewed' safety question pursuant to
'10CFR50-59 did'not exist as a result of this equipment replace-ment.-
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. Annual Summary;of JAFNPP Chnnges, Tests, and Experiments for 1988 m
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'JAF-SE-86-192, Rev. 2- LControl Room and Relay Room Recorder.
-Replacement (Plant Modification F1-86-115)
Plant Modification F1-86 ~rs orovided for the replacement of h
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- various Control Room and 5 d
- r' Room Recorders in order to in-crease their reliability and performatice.
The safety evaluation provided details of the thirteen recorders p replaced, including their location and the plant parametert:which -
they record. The addition of-these replacements to two Seismic 1
' Class I panels was addressed and was found not to affect the
. l seismic adequacy of these structures. The implementation of this ;
modification did not affect the design bases presented in the ,
Technical Specification and the FSAR. l This non-safety related equipment upgrade was determined to not censtitute an unreviewed safety question as defined by 10CFR50.59.
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JAF-SE-36-197 - Replacement of High Pressure Coolapt!In-jection (HPCI) Turbine Steam Inlet Valve 23MOV-14 (Plant Modification F1-86-039)
This safety evaluation was written in order to evaluste-the replacement of the safety-related HPCI turbine steam.in1.et valve.
The new valve meets or' exceeds all of the requirements of the original valve's purchase specifications, and in addition, satisfies the requirements-of NUREG-0588, " Environmental Qualifi-cations of Safety-Related Equipment". '
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.The-replacementaof this valve did not-change the operation of this system and therefore the probability of an accident or !
malfunction of a differen: type other than previously evaluated ,
was not created. The valve's pressure boundary materials were impact tested to-the requirements of the ASME Code Section III, Articles NC 2300 and NC 2332.1 which' adds to the increased reliability of the replacement valve. The new piping installa .
, tion exceeded the requirements.of the FSAR in that all welds were 200% radiographed and'11guid penetrant inspected. The margin of safety of-the piping-system ,as not reduced as a result of this modification.
A new pipe support was-added and several existing supports were modified to accommodata the new valve's increased. weight. Safe -!
load paths were established for-the handling of-the valves which satisfied the requirements of NUREG-0612,'" Control of Heavy Loads ,
at-Nuclear Power Plants".
This evaluation concluded that this modification did not consti-tute an unreviewed safety question pursuant to 10CFR50.59.
Modification F1-86-039 also replaced valves 23MOV-15 and l 23MOV-16. The replacement of these valves was evalueted in 1
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'JAF-SE-87-055, Rev. 2- High Pressure Coolant Injection (HPCI) j Monorail Extension (Plant Modifiention 1 F1-87-051)
A removable' crane extension with a~4-ton capacity was'added to.
the HPCI-turbine-monorail crane ro facilitate disassembly and c "
reassembly of the HPCI pump and turbine during the 1987 Refueling Outage. It was; subsequently decided to retain the cross-members i as permanent plant structures which will reduce difficult'and i time-consuming assembly / disassembly of the HPCI crane extension. "
' l The cross-members were stress analyzed for their own deadweight I and seismic loading in accordance with the requirements of PSAR ,
Section112. The structural steel used for the monorail er.cension l meets the; applicable-technical and quality requirements. The I original welds for this extension vere removed and rewelded using i approved weld procedores and qualified welders. l It was determined from these stress analysis calculations that '
the baseplate and anchor bolts configurations were insufficient ,
to withstand the 4-ton. design loading condition. Rod hanger assemblies:were added by this modification in order for this extension to withstand all the design basis loading conditions.
This safety evaluation determined that this non-safety related H modification did not constitute an unreviewed safety question.
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.. , a n Annuc1 Summary of JAFNPP Chengas, Tests, tud Experiments for 1988 I
J!F-SE-87-057 - Replacement of High Pressure Coolant In-
-jection (HPCI) Valves 23MOV-15 and 23MOV-16' (Plant Modification F1-86-039)
This eafety evaluation addressed the replacement of the safety-related HPCI ~ Steam Valves 23MOV-15 and 23MOV-16. The new valves meet or exceed'all of the requirements of the original valve's purchaseLspecifications, and in addition, the valve actuators satisfy the requirements of 10CFR50.49 for environmental con- j
.ditions inside the containment. The existing pipe supports were j modified to accommodate the valve's increased weight and the a specified seismic loads. The closing times of the valves were, .i' determined to be within the required closing times stated in the High Energy Line1 Break Analysis. I The replacement of theseLyalves did not change the operation or l flow characteristics of the HPCI System and did not involve a-change to the Technical Specifications. The new piping installa- ,
tion exceeded the rea'lirements of the PGAR since all welds were- !
100% radiagraphed and liquid penetrant inspected. The margin of j safety of the piping system was not reduced as a result of this 1 modification.
This-evaluation concluded that the overall. plant's safety was enhanced ter this modification and that an unreviewed safety :
question as defined in 10CFR50.59 did not exist as a result of ,
these valves-being replaced. '
Modification F1-86-039 also replaced 23MOV-14 which evaluated in ,
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?JAF-SE-67-060,:Rev. 1-Emergency-Diesel Generator (1DG)-Service Water Heat Exchanger Channel Flange ;
Repair and Expansion Joint Feplacement s (Plant Modification F1-87-O'3) a Plant Modification F1-87-073- entailed the repair of the discharge Lflange of the- inlet / discharge channel on the EDG jacket water to emergency service water heat exchangers. Also-, the. discharge i side expansion joints were replaced / upgraded in order to elimi-nate: flange 1eakage which developed due to corrosion. Steel reinforced flanges were utilized in the expansion joint replace-'
ment to-provide the necessary clamping force to prevent leak rate.
The discharge flange was machined to remove corrosion on the flange face. 1The reduction in flange thickness was determined to be acceptable based on the requirements of Appendix II of ASME !
Section VIII,:the code to which the EDG heat exchangers were i originally supplied per. design specification AP-9.
The material differences for the replacement parts utilized in ;
this modification'were outlined in this safety evaluation and the '
new materials were determined to be acceptable for the design application. . This modification did not alter this. system in a manner that affected the existing design bases or accident analyses.
This safety-n lated modification resulted in the ~ performance enhancement of the existing equipment which improved its ability "
to perform'its design-functiot.. It was concluded in this safety evaluation that this modification did not involve an unreviewed
- safety question..
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Annusl' Summary of JAFNPP Chengas, Tosts, I and Experimentscfor 1988 f,
JAF-SE-87-061 -- Reactor Protection System (RPS)/ Primary Containment Isolation System (PCIS)'HFA' Relay Computer Contact Change _(Plant Modification F1-86-087)
The' subject safety evaluation provided a review for Plant Modi-F fication F1-86-067. .This modification changed the process computer c.gital inputs from the safety-related RPS and PCIS Type HFA relays'from normally closed to normally open contacts.- These ,
changes were made in order to eliminate spurious signals due to
" contact bounce". associated'with.normally closed contacts. As a
- result,n only one signal will be sent to the Process Computer System when a scram signal is received.
The RPS.and PCIS are nuclear safety-related systems which pros 1de logic input to both safety-related and.non-safety related sys-tems. To maintain separation between the essential (RPS, PCIS) '
and non-essential systems-(process comauter) the coil to contact isolation arrangement was not changed by this modification ensuring. compliance with:IEEE Standard 384-1981 and FSAR Section 8.5.4.2. The-Technical Specifications were not affected by this modification.
It was determined that this modification did not constitute an unreviewed safety question pursuant to 10CFR50.59.
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. 4 Annual' Summary of JAFNPP Chtnges, Tests, and Experiments-for 1988 4
JAF-SE-87-064, Rev. l'- Installation of Conductivity Elements in <
E the Ion Exchange Bed Discharges (Plant Modification-F1-87-056)
.This.non-safety related. modification consisted of replacing three
. conductivity elements in the existing streams of the cation, :
, anion,'and mixed' bed ion exchangers for the Makeup Water Treat-ment System. New conductivity / resistivity analyzers were instal-
~1ed along with new stria chart recorders. In addition, the demineralizer tanks high'1evel trip.setpoint was changed from 95%
. volume.to 98% volume to increase tank-holding capacity and the
- low level interlock was deleted to allow for water processing and tank filling above the 40% level.-
This. modification was performed in order to optimize the opera-tion and maintenance of acid and caustic regenerations of the .
Makeup' Water Demineralizer.Cystem.
The Makeup Water Treatment Acid and Caustic Regenerative System does not suply_any chemicals to any safety-related systems or equipment. The implementation of this modification did not >
affect the Technical Specifications.
L This safety analysis determined that the design basis as stated
.in the FSAR was not affected by this modification and that an unreviewed safety question was not involved as a result of this modification. ,
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'JAF-SE-87-069 - Installation of Personnel Frisking Booths-(Plant Modification F1-87-017) i
-This safety evaluation provided the review for Modification F1-87-017 which' installed six personnel frisking booths at svarious locations in the plant. Due to the fact that safe-tv-related equipment is located near three of these;1ocations, three booths were constructed to assure that Seismic Class II requirements were met. These six booths were irestalled to replace the temporary frisking enclosures that had_ existed and to allow for more effective'whole body frisking by Site personnel. .
The design and fabrication of these booths was accomplished'in accordance withithe AISC Manual of Steel Construction '(Sth Edition). The floor loads due.to these booths were found to be in compliance with the limits presented in Section 12.4 of the .
FSAR.- The heavy load concerns of-NUREG 0612 did not apply since all of the subcomponent-of the booths weigh less than 750 lbs.
These booths are situated so that access to fire protection and
- other plant equipment would not be hampered.
It was concluded that this non-safety related modification did ,
-not involve an unreviewed safety question as defined by 10CFR50.59.
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LJAF-SE-87-072,' Rev.'11- Containment Vacuum Breakers Spare Part-and Material Substitution for 27VB-1 -
through 27VB-7 (Minor Modification M1-87-027) e, This' safety evaluation, reviewed the scope of work associated with Minor Modification'M1-87-027. This modification allowed for the
.useaof certain vendor recommended part substitutions-for vacuum 1 breaker valvecc27VB-1 through'27VB-7.-
The replacement materials for the new disc / stem assembly have:
- significantly higher tensile and yield strengths than the origi- ;
nal material. The new assembly's. increased weight was determined 1 by the maaufacturer to not have any. adverse affect on valve operation and would not affect system design or seismic analysis.
- The replacement disc' pins and roll pins improve the design and I
- function of these valves and the substitute materials were found l to be acceptable for their intended use.- l This modification did not require'the revision of any sections of l
- the FSAR or'the Technical Specifications. It was determined that ;
this medification did not constitute an-unreviewed safety ques-
- tion ns defined by 10CFR50.59.
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EP; Annual Summtry of JAFNPP Changes, Tests, and Experiments for 1988 JAF-SE-87-081, Rev. 1- 600'VAC and 125 VDC Motor Control Center (MCC) Drip Shield and-Gasketing Instal-lation (Plant Modification F1-86-061).
'This safety evaluation was written in. order to provide a review of the possible safety issues associated with-Plant Modification F1-86-061. Drip _ shields were installed to various safety-related-
,MCCs in~ order to protect internal electrical devices from direct water intrusion. This modification also removed existing drip
- shields over the breaker cubicles for switchgear L15 and L16 to provide for better air circulation. These shields were no longer i required since environmental enclosures were installed around each-sw.tchgear under Plant Modification F1-84-005.
. This modification had no impact on-the design and safety basis of the subject electrical equipment.. The. slight added weight was determined to be inconsequential and did not affect the seismic.
adequacy of the.MCCs. No changes to the FSAR or the Technical
. Specifications were required as a result of this modification.
It was determined that this modification did not involve an-
.unreviewed safety question pursuant to 10CFR50.59.
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f Annual Summary of JAFNPP Changes, Tests, s,' and Experiments for 1988-JAF-SE-87-086, Rev. 21- Upgrade of ARM Auxiliary Alarm System (Plant Modification F1-87-097)
This-non-safety related modification replaced the auxiliaryLalarm horns in high noise areas with new high output horns, installed new halogen beacon lights and adjusted the level on existing alarm horns not located in high noise areas. In addition, indicating lights were installed on all auxiliary alarm units which will- indicate that power is available for operation. ,
JThis-modificationL did not alter the existing method of detection or modify any circuitry involved. These changes were designed to. '
conform to the applicable FSAR requirements and to Seismic Class II requirements. The design and safety bases for this system were no altered as a-result of this modification.
1These changes enhance the plant's safety by providing additional warning to workers and operators of. abnormal radiological con-e ditions. This modification did not constitute an unreviewed safety question.
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'JAF -SE-87-089 -
' ~
Recirculation Flow Control Improvement (Plant
. Modification F1-87-043)
.The aurpose-of this safety evaluation was to provide an analysis
- of:the safety concerns associated with Plant Modification F1-87-043. This non-safety.related modification simplified the recirculation pump motor generator set's control system. This was accomplished by eliminating the feedback loop which;had contained the speed controllers and. incorporating the speed controllers into the remaining control loop. This new circuit m allows separate adjustment of " ramp-up" and " ramp-down" rates.
The new variable rate limiter installed allows for the rate of change selected to be fast enough to meet all manual load change requirements, including startup, but slow enough to provide a ,
stable and smooth flow-adjustment.
The removal of-the outer speed loop causes the Recirculation system to be in a full-time manual flow control mode. The recirculation aump trips and runbacks were not affected. The failure-of both manual flow control loops will not result in a l
more severe situation than that which has been evaluated in
- Section 14.5.6.2 of the FSAR. The ability of the Recirculation Flow Control System to control reactor power over a limited range was not affected.
This safety evaluation reviewed the effect of this' modification against further postulated accident scenarios and determined that an unreviewed: safety question as defined in 10CFR50.59 did not exist as a result of this change and no changes to the Technical Specifications was required..
The preoperational test associated with this modification was reviewed by JAF-SE-87-152.
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Annurl' Summary of JAFNPP Changas, Tests, and Experiments for 1988 .
W JAF-SE-87-090 - Equipment Storage Pool Seal Replacement (Plant Modification F1-86-063)
This. safety' evaluation provided a review of the safety concerns
,. associated with the installation of an Equipment Pool Seal System for the Reactor Pressure Vessel Internals Storage Pit. This new system provides an effective seal to keep the dryer-separator pit-completely flooded'when the reactor cavity is drained, thereby reducing. local dose rates.
A detailed analysis of the-system design was presented in this
< evaluation. The types of materials and the support structure utilized were' determined to be acceptable. The heavy load concerns were addressed and the necessary precautions and appli-cable procedures stated. The equipment and its_ supporting elements were reviewed for proper anchorage and' load carrying ca-pability under seismic forces and it was determined that no safety-related equipment would be affected during a seismic event.
L' The evaluation concluded that this non-safety related modifica-tion did not constitute an unreviewed safety question as defined by 10CFR50.59. '
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. Annual = Summary of JAFNPP Chtngas, Tasts, and Experiments for 1988 r: '
s JAF-SE-87-091 -- Additional Residual Heat Removal (RHR) g" . Suction Valve Interlocks to Prevent Reactor Vessel Draining _(Plant Modification 1 F1-86-058) '
The purpose of~this safety evaluation was to provide a review of the safety concerns associated with the implementation of Plant Modification F1-86-053. The RHR suppression pool suction valves (10MOV-13A-D) were interlocked with the RHR Shutdown Cooling System, valves (10MOV-15A-D) to prevent simultaneous opening.
Also, the RHR torus cooling isolation valves-(10MOV-39A&B) were
. interlocked with the RHR Shutdown Cooling System Valves (10MOV-15A-D).'to prevent simultaneous opening. This modification was performed in order to prevent the rapid decrease in reactor coolant inventory which could result from the inadvertent simul- i taneous opening of valves in the same train.
=These interlocks did not change any of the RHR System operating i modes nor did they impact or prevent any of the RHR System safety
- functions from being performed. It was noted that failure of the interlocks could result in lowering the reactor water level.
However, this event is protected against by Reactor Protection System and Emergency Core Cooling System reactor water level instrumentation. The new conduit added was designed.to Seismic Class I criteria and was routed to meet plant separation crite-ria. 'The new. interlocks do not affect the existing control circuits designed to comply with 10CFR50, Appendix R (Safe 1
Shutdown form outside the Main Control Room) requirements. All of the wiring changes and electrical equipment added for this l modification comply with FSAR Section 8.0, 10CFR50.49, and D 10CFR50,-Appendix R requirements. It was concluded that an unre-viewed safety question pursuant to 10CFR50.59 did not exist a e result of this safety-related modification.
Nuclear Safety Evaluation JAF-SE-87-092 provided the evaluation for the preoperational test (POT-10G) associated with this modification.
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' Annual Summary of JAFNPP Changas, Tssts,
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and Exparim*nts for 1988 L JAF-SE-87-092~-- Additional Residual Heat Removal (RHR)
- Suction Valve-Interlocks to Prevent Reactor i i
Vessel-Draining (Preoperational. Test POT-10G) ll
' This' safety evaluation reviewed the impact of implementing Preoperational Test POT-10G which verified the proper operation.
of the limit switch interlocks installed under the safety-related-Plant Modification F1-86-053.. This test was performed-during a plant outage, with'the plant in a cold condition, Precautions were' outlined to prevent inadvertent draining-of the reactor H
vessel due: to the possible failure of the new interlocks during the - .te s t . Only one:RHR loop was tested at a time-in order to-ensure any required shutdown cooling-could be, performed.
1 -The performance of'this test did not change the results of the '
L accident analysis nor reduce the-safety margins defined in the-l Technical Specifications.
l It was concluded that.the performance of this test did not constitute an unreviewed safety question pursuanc to 10CFR50.59.
- Nuclear Safety Evaluation JAF-SE-87-091 provided the-review of Modification-F1-86-053.
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Annual-Summary.of JAFNPP Chrngss, Tests, and Experiments for 1988 JAF-SE-87-097, Rev.'1 - High Pressure Coolant Injection (HPCI)
Gland Seal Condenser Exhauster Replace-ment (Plant Modification F1-86-105)-
.This. safety evaluation provided a review of the safety concerns i associated with the implementation of Plant-Modification F1-86-105. The HPCI gland seal exhauster was replaced by this modification due to severe corrosion on the exhauster's inside
- casing. . The new exhauster has a stainless steel casing to preventithis from. reoccurring. The exhauster motor which was
-installed has a lower full load amperage rating due to design improvements and the overload heater for this motor was resized accordingly.
This modification did not change the design bases for the HPCI System as defined in FSAR Section 6.2, but instead increased the life expectancy of the gland seal exhauster. A' change to.the Technical Specifications was not required as a result of this modification.
It was determined for this safety-related modification that an unreviewed. safety question as defined in 10CFR50.59 did not exist._
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and Experiments for 1988. *
~ JAF-SE-87-099 -~ Cooling Water to Engineered Safety Feature
-(ESF) System Components for RegulatoryLGuide ;[
1 1.97, Rev. 2 (Plant Modification F1-87-044)
~ Plant Modification F1-87-044 was' designed to meet.the-require-. "
- .ments for Regulatory Guide 1.97, Rev.-2, for the monitoring of cooling water ~. flow to ESF: system components.
This safety-related modification was performed.in order-to provide indication-of flow rate for both trains of the Emergency Service Water System. -New flow measurement instrumentation was
' installed along with~ instrument sensing lines, environmentally qualified cable, and seismically. qualified supports for the n'ew local' transmitters.
Safety Evaluation JAF-SE-87-099 reviewed the safety issues associated with this. modification. This review included environ-mental. qualification concerns, seismic qualification require-ments, applicable code adherence, and potential accident scenar-ios during installation. This modification complied with the-Safety Design BasesoSection 8, FSAR (Electrical Systems) require-- '
ments. The Technical Specifications were not-required to be revised as a resultaof this. modification.
It was determined that this modification did not involve an unreviewed safety question as defined by 10CFR50.59.
The~preoperational test associated with this modification was analyzed in Nuclear Safety Evaluation JAF-SE-88-005.
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Annus11Summtry of JAFNPP-Changos, Tasts,. t 4 and Experiments for 1988 q
JAF-SE-87-103', Rev. 5'- Standby Liquid' Control (SLC) Modifica-tion Required to Comply with 10CFR50.62 Requirements (Plant Modification F1-85-055)
This safety evaluation provided a review of the safety issues associated with Plant Modification F1-85-055. This safety-related modification changed the solution of the SLC tank to an enriched sodium.pentaborate solution, replaced the SLC tank level transmitter, and changed the setpoints of the tank heater, heat tracing system,-and level transmitter. 1 The change in the SLC tank liquid was done in order to comply with the-NRC raquirements for Anticipated Transient Without. Scram (ATWS) as stated in 10CFR50, paragraph 50.62 " Requirements for Reduction of Risk from ATWS Events in Light Water Cooled Nuclear Power-Plants".
This= evaluation analyzed the new boron injection capability of the system, the heat tracing installed, seismic design consid-erations, and the mechanical and electrical design of the entire system.- The design bases for this system described in the-Technical Specifications required revision to-reflect the changes performed by this modification. The surveillance requirement for the. checking of Boron-10 enrichment needed to be added along with the new pump's increased minimum flow rate.
It was determined that this new design was' consistent with the original design requirements and that no new or different kinds of. accidents could result from this improvement of the. effective-ness and reliability of the SLC' System.
It'was determined that this modification did not adversely affect the operation of the SLC System and did not constitute an unre-viewed safety question pursuant to 10CFR50.59.
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i JAF-SE-87-105 - Alternate Rod Insertion '(ARI) Modification -
Pre-Outage (Plant Modification F1-S5-053)
This' safety evaluation provided a review of the non-outage related tasks associated with the ARI System modification instal-
- lation. .The total task, which was partially safety-related, was ;
- evaluated in Nuclear Safety Evaluation JAF-SE-87-106. '
- The non-outage tasks included installation of new solenoid valves, a scram header pressure transmitter, relay panels, and logic panel. .No tie-ins were made with existing safety-related. l electrical or mechanical systems during this pre-outage phase nor l was there any impact on adjacent safety-related equipment due to the installation sequence. Therefore,-these tasks did not have any impact from a safety system standpoint.
1 This evaluation reviewed system interfaces, construction con- l cerns,-Appendix R concerns, Human Factors issues, and seismic design. A review of the FSAR and Technical Specifications determined that there were no sections or items which required ,
changes due to this pre-outage work for the ARI System. It was- '
the conclusion of this evaluation that this work did not consti-tute an unreviewed safety _ question pursuant to 10CFR50.59.
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+ and Experiments for 1988 .
L'7 1 JAF-SE-87-106 - Alternate Rod Insertion (Outage) (Plant-Modification F1-85-053)
This modification was implemented to meet the NRC requirements for-Anticipated Transients Without-Scram'(ATWS) as delineated in' ,
10CFR50, paragraph 50.62 " Requirements for Reduction'of Risk from ATWS Events for Light Water Cocied Nuclear Power Plants". This t, rule requires that an alternate means be provided for automatic 1 reactor control rod insertion that is diverse and independent of ;
the Reactor Protection System (RPS). The Alternate Rod Insertion (ARI)_ System provides this means. ,
The ARI System scrams the reactor by venting the common portions of the scram valve pilot header through multiple vent valves.
The ARI vent signal is generated via new relay logic using reacter vessel high pressure and low level signals from the Analog Transistor Trip System.-
This modification added.four solenoid vent valves to vent the scram' valve pilot header, a solenoid valve to isolate the instru- -:
ment air supply from the scram valve pilot air header, an ARI logic panel, two ARI isolation relay panels, Control Room ARI status indicators, a scram valve pilot air header pressure-transmitter, and an ARI valve position input into the Emergency Plant-Information Computer (EPIC). The only safety-related j' equipment installed by this modification were the two ARI iso-lation relay panels.
This safety evaluation analyzed the safety issues associated with the implementation of this modification. This included a review of the electrical and mechanica1 interfaces with existing plant systems, seismic design, ARI ' system design, Appendix R impact, human factors issues, EPIC concerns, and Licensing documents.
Although the ARI is a new system'and was not addressed in either the FSAR or Technical Specifications, additions were required to these plant documents as a licensing commitment to the NRC.
The conclusion of this very detailed analysis was that an unre-viewed safety question as defined by 10CFR50.59 did not exist as l
a result of the implementation of Plant Modification F1-85-053.
, The. partial installation of the hardware associated with this l modification was analyzed in Nuclear Safety Evaluation JAF-SE-87-105. The safety analysis for this modification's l preoperational test (POT-03E) was contained in JAF-SE-87-107.
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"' Annusl Summ:ry of JAFNPP Changos, Tosts,
__and Expnrimsnts for 1988
.JAF-SE-87-107 - Alternate Rod Insertion (ARI) System (Preop-erational Test POT-03E)
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This safety evaluation reviewed the performance of-Preoperational Test POT-03E which verified the operability of the new ARI LSystem. This system, of which portions are safety-related, was installed by Modification-F1-85-053 which was analyzed in safety-evaluations ,JAF-SE-87-105 and JAF-SE-87-106.
l= In addition to verifying ARI System operation, this test estab-
,. lished that the existing pilot air header: depressurizes to less than 16 psig within-the first 15-seconds after ARI actuation and i' completes depressurization in less than 25 seconds.
The test was performed with the reactor in a cold condition with all of the control rods inserted. Under these conditions, it was ;
determined that this test would not impact the safety of the '
plant-and that no sections of the FSAR or Technical'Specifica- l tions required changes due to the performance of POT-03E. An l unreviewed safety question pursuant to 10CFR50.59 did not exist as a result of the performance of this test.
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A'nnual Summtry'of JAFNPP Chengas, Tests, and Experiments for 1988 JAF-SE-87-110, Rev. 1 -- Automatic Depressurization System (ADS).
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Pneumatic Supply System, 27S0V-141, 27S0V-144, and~27S0V-145-Functional Test (Preoperational Test POT-27U)
This' safety evalustion was written to review the safety issues p associated with the performance of-Preoperational Test POT-27U. <
This test was performed to ensure the correct operation of the control, monitoring circuits and solenoid valves installed for
-Modification F1-84-072. The_ safety analysis for this-safe-
-ty-related1 modification was contained in Nuclear' Safety Eval-
, uation JAF-SE-86-166, Rev. 5.
Since the plant was in cold shutdown, it was determined that any l
failure of these solenoid. valves during the performance of this 1 test would not have an adverse-effect on the safety status of the plant.. The ADS valves were not required to be operational during this test since the plant was in cold' shutdown and the reactor u
-pressure _was less than 100 psig. No changes to the Technical '
Specifications or the FSAR were required due to the performance of-this test. The conclusion of this safety evaluation was that j an unreviewed safety question did-not exist as a result of this preoperational test.
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. i Annual ~ Summary of-JAFNPP:Changss, Tosts, and: Experiments for 1988 .
JAF-SE-87-113 - Off-Gas ChillerLCooling Water Flow Indicator (37FI-84) Replacement (Minor Modification M1-87-070)
Minor Modification M1-87-070 replaced the. damaged off-gas chiller-cooling water flow indicator with a similar model from the same manufacturer. In addition, an integral flow:restrictor was installed in the outlet part of the indicator to prevent instru-ment damage due to excessive flow rates.
' This flow indicator allows improving control of.the off-gas moisture content. .The design bases-and accident analyses were not affected by the implementation of this modification, and no change to the FSAR'or Technical Specifications was required.
This safety evaluation determined that-this non-safety related flow indicator replacement did not constitute an unreviewed safety question.
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and Experiments for:1988 JAF-SE-87-116 - Replacement of' Reactor Water Clean-Up (RWCU). I Thermocouples 12TE-117A-F (Plant Modification l F1-86-132) !
i Plant Modification F1-86-132 replaced the safety-related thermo- l couples used for RWCU steam lea'< detection. The new thermo- l couples are environmentally qualified-to.10CFR50.49 and are seismically qualified to IEEE 344-1975.
Safety Evaluation JAF-SE-87-116 analyzed this thermocouple i replacement for environmental qualification concerns, proper design application, measurement accuracy and. seismic adequacy.
It was determined that this equipment replacement increased the reliability of these components.
The conclusion was that this modification did not constitute an unreviewed safety question as defined in 10CFR50.59..
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Annus1 Sumo:ry of JAFNPP Changes, Tests,-
'and-Experiments for 1988 JAF-SE-87-120 - - Byron-Jackson Valve Seat Material'Substi-tution (Minor Modification M1-87-138) .
The purpose.of this safety evaluation was to provide a technical review of the valve seat material change for safety-related valves,27MOV-122 and 123. This manufacturer-recommended'cubsti-tution (ASTM A-515 Gr. 70 in lieu of A-106 Gr. B) was determined to be acceptable based upon a comparison of the materials' properties. This material change did not impact the valve's operability, and:therefore the valve and Purge System will '
perform its intended safety function. The design bases for this system was not affected by the implementation of this modifica-tion.-
The conclusion of this safety evaluation was that this substi-tution did not involve an unreviewed safety question. ,
t Page Annuc1' Summery of JAFNPP Chengas, Tests,
,. - and Experiments for 1988 JAF-SE-87-1221- Condensate Sample Probe Replacement (Minor ,
~ Modification M1-87-032) 1 This safety evaluation provided the review for Minor Modification M1-87-032. This non-safety related modification involved the replacement of condensate sample probes in the condensate domin- .,
eralizer~ influent and effluent headers. The new probes' improved ~ >
- design eliminates the chance of corrosion-failure which could have led to a probe breaking off and becoming lodged in valves ;
downstream.
The installation of-this modification had no affect on the monitored-Condensate System described in the FSAR, and therefore did not increase the probability of occurrence, or the conse-quences of an accident or malfunction of-equipment _important to maintaining the safety and reliability of the plant. The FSAR and Technical Specifications were not affected by this modifica-tion. This change resulted in the improved reliability of this probe.
The conclusion of'this safety evaluation was that this modifica-tion did not. constitute an unreviewed safety _ question as defined by_10CFR50.59.
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,3 Annual Summary of.JAFNPP Chnnges, T0sts, ,
and Experiments for 1988 JAF-SE-87-125 - Replacement Pnal-Seal International Inc.
VVI-60CN Trunnion Valves (Minor Modification M1-86-088)
' The purpose of this safety evaluation was to evaluate the.re-
, placement of trunnion valves with newer moscls. These non-safety related valves are utilized in the Clean-Up Filter Demineralizer System.; Each valve is remote manually operated and the high
-performance butterfly valves retained the critical dimensions of the units they replaced. The materials utilized were reviewed
, . and were found to be suitable for their intended design applica-tion. These replacements were found to meet or exceed the applicable requirements of the original Purchase Specification APO-87.
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The-conclusion of this safety evaluation was that this valve
- replacement did not involve an unreviewed safety question as defined in 10CFR50.59.
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Annus1 Summtry of JAFNPP Chengss, Tosts, o.
and Experiments for~1988 .
I JAF-SE-87-129 - Modification.of 20A0V-95 Control Circuit (Minor Modification M1-85-056)
- This safety evaluation evaluated the change to the control circuit of 20A0V-95, Drywell Equipment Drain Sump Outboard
' Isolation-Valve. This safety-related modification was performed to facilitate compliance with Technical Specification 4.6D which requires that the reactor coolant leakage rate be monitored'and recorded every four hours.
JIn order to obtain a measurement of equipment leakage,. Operations personnel previously have had to manually byaass the automatic ,
logic for.20A0V-95 in order for the sump to 3e pumped down to lo level.. A new control switch was installed.which allows Op-erations personnel to puma down the sump without having to rely.
on the assistance of another operator out in the-plant. The-primary containment isolation function of 20A0V 95 was not compromised.due to this modification since this' valve will' ,
isolate on an isolation signal'even when the new control switch i is in the " pump down" position. l The surveillance requirements of the Technical Specifications = l were unaffected by this modification. FSAR Section 4.10, which describes the means by which leakage from systems essential for ,
safe plant shutdown are monitored, required no revisions as a !
result of this modification. This new design did Aot change-the safety-design basis of the system, but was an operational en-hancement that did not impact its safety function. t This review concluded that this modification did not adversely affect plant-safety and that an unreviewed safety question pursuant to 10CFR50.59 did not exist, i.
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-JAF-SE-87'-130 - HPCI Gland Seal' Condenser Level Switch (23LS-100) High Limit Setpoint Change (Minor Modification M1-87-147)
This' safety evaluation reviewed Minor. Modification M1-87-147 which changed the.high limit setpoint for non-safety related 23LS-100 from 44 inches to 35 inches above the floor (272' elevation)~. This modification was performed in order to ensure that the High Pressure Coolant Injection'(HPCI) turbine would not accumulate condensate which could create an undesirable condition.
upon system initiation. -This is an improvement to the original system configuration.
This changing of the high level setpoint 23LS-100 did not in-crease the-probability of-occurrence or consequences of an accident or malfunction of equipment important to safety because the turbine exhaust-line condensate will be removed ss originally intended.- The margin of safety as defined in the basis for the Technical Specifications was not reduced since the condensate drain function is not discussed in the Technical Specification bases-.
This modification did not present an unreviewed safety question as-defined in 10CFR50.59.
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gnd Expsrim;nts for 1988 JAF-SE-87-133 - Reactor Water Cleanup (RWCU) Containment Isolation Provision Upgrade (Plant Modifica-tion F1-87-068)
I This~ safety-related modification installed a new motor-operated outboard containment isolation valve, 12MOV-69, in the RWCU return line to the Feedwater System. The new valve and associ- E ated circuitry are classified as Nuclear Safety-Related, Seismic Class I, Electrical Class IE, QA Category I, and are environment-ally qualified.
Upon loss of actuating power, the valve will fail in the as-is position. If the valve fails in the open position, the check valve inside containment (34FWS-28A) will perform the- recuired isolation function. The position status of 12MOV-69 is cisplayed above its new-control switch on panel 09-3 in'the' Control Room as
. well as on the isolation valve mimic display on panel 09-3..
The new isolation valve provides the necessary isolation capabil-
'ity to. satisfy the requirements of NUREG-0737, Item II.E.4.2, and 10CFR50, Appendix A, Criterion 54 and 55. The additional weight of this valve in the piping system was analyzed and it was determined that the system would maintain its structural and operational functions during and after a Design and. Operating Basis Earthquake.
A diverse containment isolation. signal (high Drywell pressure) was added to the existing logic for 12MOV-15, -18, and--80. When 1
.I an isolation signal is received, these pump suction valves will close. The new valve 12MOV-69 will also close causing the RWCU piping to be pressurized to CRD' system pressure. The RWCU piping L
system has been analyzed for this. occurrence and can withstand '
this condition without damage.
L l This safety evaluation determined that since this modification i met the original design requirements and added redundancy to the '
plant design, that an unreviewed safety question did not exist.
l The preoperational test (POT-12B) associated with this modifica-
! tion was reviewed in Nuclear Safety Evaluation JAF-SE-86-141.
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JAF-SE-87-137, Rev. 1- Reactor Water Cleanup ~(RWCU) System
. Precoat Evaluation Program The purpose of this program war. to compare the performance of alternative precoat formulations and precnat techniques on.the non-safetyLrelated RWCU filters. The various formulations'and methods used were evaluated for their potential to improve reactor: water chemistry, reduce spent precoat disposal-volumes, and-improve radwaste processing operations.
o This evaluation provided a safety analysis for the five planned at tests in the RWCU System Precoat Program. All materials utilized t, '
in this program were endorsed by the General Electric Company for RWCU service. Details concerning the performance of the actual
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testing-were provided in Temporary Operating Procedure TOP-89.
Limits.for various RWCU filter /demineralizer effluent' parameters.
were provided in order to ensure that JAF Technical Specifica-tions were.not violated.
This safety evaluation concluded that the RWCU Precoat Evaluation Test Program did not involve any unreviewed safety concerns as defined.by 10CFR50.59.
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.. a' Annual Summ;ry of JAFNPP Chengss, Tests, rnd Exp.orim*;nts for 1988 JAF-SE-87-140 - Re-Trim of the Fire Protection System Viking Flow Control / Deluge Valves (Plant Modifica-tion F1-87-071) g The? primary purpose of Modification F1-87-071 was to allow for then replacement of failing components (three-way alarm test valves) on the Viking Water Flow Control / Deluge Valves and to standardize the trim configuration. In addition, this non-safety related modification provided for the installation of Pressure-Operated Relief. Valves (PORVs) on these flow control / deluge I
. valves.
l4 The valve trim was standardized by ensuring that all the con-trol / deluge valves had the current-trip configuration utilized by V Ning installed.= This new configuration consisted of two ball valves and one check valve instead of the previously used-three-way valve. This standardization alleviated any confusion that may have existed relative to the operation of these valves and simplified training and maintenance.
The additionLof the PORVs prevents the control / deluge valves from
. resetting in-the-event electrical power was lost to the solenoid pilot valve. The purpose of these PORVs is to require deliberate operator action to shutdown a Fire Suppression System.
The FSAR and Technical Specifications were not affected by this modification. This modification provided increased reliability to the overallLFire Protection System and did not involve an unreviewed safety question.
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-M;92 -s Annual Summary of JAFNPP Chrngas, Tasts, ff' .end-Exptrim;nts for 1988 i
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-JAF-SE-87-141, Rev. 1- Shipping Cask Handling for Spent Fuel Pool Cleanup 1
This safety evaluation addressed the safety concerns associated i L with the handling of shipping casks used in the removal of
-various irradiated hardware from the Spent Fuel Pool. In order to ensure compliance with the heavy. load concerns of NUREG 0612, a detailed review of the specific load paths, the required j maneuvering., and the procedures providing the administrative controls were presented. l Precautions were outlined to ensure that the cask would not overturn du -ing a seismic event when resting. on its transport '
' baseplate. The rigging components were all required to have a 10:1 safety factor against ultimate strength or have complete i
i redundancy in the rigging with a 5:1 safety factor. J L
. Restrictions on the allowed cask movement and handling were l provided to minimize the effects of any accident or malfunction. i The evaluation concludedithat this non-safety related task.did- l not constitute an unreviewed' safety question as defined by 10CFR50.59.
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'JAF-SE-87-142 - Automatic Reset of Scoop Tubes on Runback Signal-(Plant Modification F1-87-150)
-The purpose of this safety' evaluation.was to provide a safety analysis for Modification F1-87-150 which provided_for the
. automatic reset of the Reactor Water Recirculation (RWR) System
-Scoop Tube Positioner when a runback signal occurs. This auto-matic runback will occur even with the scoop tube positioner electrically locked.
This non-safety related modification, in order to decrease the possibility of a scram caused by a loss of a feedwater pump, allows for the RWR pumps to slow down as soon as a runback is called for. This allows for the reactor power to be reduced to
-within the capacity of the operating feedwater pump thereby averting'a low level: reactor trip.
Two relays and two switches were the only components installed as part of this modification which eliminated the need for manual reset by the operators.
Potential equipment malfunctions and misoperation were discussed in this evaluation.- It was concluded that this modification resulted ineno degradation of safety as defined'in the FSAR and di'd not affect any design limits expressed in the Technical '
Specifications.
This safety evaluation concluded that an unreviewed safety question did not exist. '
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~JAF-SE-87 144 - Permanent Support Attachment for-Unit Heater 66UH-29C (Minor Modification M1-87-J24)
This safety evaluation provided the review for non-safety related' Minor Modification M1-87-124, which installed a permanent support for Unit Heater 66UH-29C. This support replaced a previously
, installed temporary support which had been attached to a QA~
Category I electrical conduit by means of a threaded rod. '
The new support was designed to Seismic Class II recuirements to ensure that it would not affect plant safety-relatec; equipment during a seismic event. The addition of this, support did not in any way affect the. safety design bases for the plant.
The determination of.this safety evaluation was that this minor modification did not adversely affect plant safety and did not involve an unreviewed safety question.
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Annual Summary of~JAFNPP Chcngcs, Tests, and Experiments for 1988 JAF-SE-87-145, Rev. 1 - Refueling Cavity Inner and Outer Bellows Seal Leak Detection System Improvement.
(Plant Modification F1-87-055)
F The purpose of this non-safety related-modification was to improve the Leak Detection System for the the refueling cavity inner and outer bellows seals so that a gross failure or minor-leakage from the seals can be detected. This modification involved the installation of piping between the " dry side" of the inner bellows. seal and the inner bellows seal. leak detection flow switch, the rerouting of piping from the'" wet side" of the inner bellows seal-to a true drain line, the installation'of new flow indicating switches in both the inner and outer bellows seal leak detection sensing lines, and the installation of the necessary electrical equipment to support the. flow indicating switches and computer points for inner and outer-bellows seal leak detection.
As a result of this modification, early warning to the Control Room in the case of a bellows seal failure, during refueling, was provi ded. The addition of these switches upgraded this system so that the description in Section 9.3.5.2 of the FSAR was correct.
The plant conditions required for the safe implementation of all parts of this modification were outlined within this evaluation.
This modification did not change the power generation design basis for the Fuel Pool Cooling and Cleanup System. It was.
determined that no:unreviewed safety question was presented-by this modification.
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'und Experiments for 1988 JAF-SE-87-146 - Control Room and Relay Room Door Latch Modification (Minor Modification M1-84-101)
Modification.M1-84-101_ installed push bars on the_ Control Room >
and Relay Room doors and added'a time delay relay circuit to the door control circuits. The relay was added in order to facili-tate the unlatching of the door mechanism with the use of the existing pushbutton switch on the secure side of these doors.
-Neither.the FSAR nor the Technical Specifications were affected by this modification. The Security Plan required revision to address ~the changes to-the doors affected by M1-84-101. The seismic qualifications of these doors were not affected by these changes.
An unreviewed safety question did not exist as a result of the minor changes associated with this non-safety related modifica-tion, i
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JAF-SE-87-152.- -Reactor Water Recirculation (RWR) Flow Control Conversion and Scoop Tube Auto Unlock (Preoperational Test' POT-02F)
The purpose of this preoperational test was~to verify proper operation of the non-safety related RWR flow control circuits prior ~to the actual start-up of the plant. This test ensured that the rate limits were set as specified in Plant Modification F1-87-043' and that the scoop tube auto unlock relays installed by Plant Modification F1-87-150 operated as designed. . Safety Evaluation JAF-SE-87-089 provided the review for F1-87-043 and JAF-SE-87-142'provided.the review for F1-87-150. . .
l This test checked each'of the flow control circuits in all four i stages of operation: start-up (before field breaker closes), i start-up- (after, field brea'ker closes), normal, and runback. The up-rates and down-rates of the newly installed rate limiters were checked and the new and reset relays functionally tested.
Preoperational Test POT-02F was conducted with the plant in a cold shutdown condition. The operation of the.RWR System during a cold condition did not create conditions-that exceeded any i safety margins. No change to the FSAR or to the Technical '
Specifications was~ required as a result of the performance of ,
this test. I This' safety evaluation concluded that since this preoperational !
test did not constitute an unreviewed safety question pursuant to 10CFR50.59.
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JAF-SE-87-153 - Bus 10500 and 10600 Degraded Voltage Pro-tection System Modification Checkout and Functional Test (Preoperational Test, POT-93B)
Preoperational Test POT-93B functionally tested the Degraded Voltage Protection System to verify relay operation, breaker trips, breaker closing logic, and computer and annunciator logic.
Engineering Change. Request ECR-F1-84-041-019 revised the alarm logic of this system in order to eliminate a time delay between degraded voltage detection and subsequent alarm annunciation.
The safety-related modification which installed the Degraded Voltage Protection System (F1-84-041) was reviewed for safety concerns in Nuclear Safety Evaluation ~JAF-SE-85-118, Rev. 2.
This test was conducted with the designated emergency bus out of service and with the plant in the cold shutdown condition. The redundant emergency AC bus, as well as the off-site sources remained in service complying with the Technical Specifications.
l The plant conditions necessary.for this test were specified'in order to ensure compliance with Technical Specifications. This evaluation concluded that this test did not constitute an unre-viewed safety question as defined in 10CFR50.59.
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JAF-SE-87-157 - Replacement of Reactor Water Cleanup (RWCU) l Effluent Pneumatic Instruments with Electron- :
ic Instruments This safety evaluation reviewed the replacement of pneumatic instrumentation in the RWCU effluent loops with electronic instrumentation : The flow transmitters, indicating controllers, recorders, indicators, switches, and the pneumatic to voltage converters associated with flow elements (12-4FE-74A,B) were replaced with new electronic instruments which provide more precise measurement and control of the system. This non-safety related modification greatly reduces the constant maintenance
-problems which were experienced with the pneumatic instruments.
This modification did not have a negative impact on any plant systems and did not affect the safety design bases defined in plant licensing documents.
The conclusion of this evaluation was that an unreviewed safety question as defined by 10CFR50.59 did not exist as a result of this modification.
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The. purpose of this safety evaluation was to evaluate the conver-sion of the CRD mechanisms from the BWR/4 to the BWR/6 type design. The BWR/6 CRD incorporates an improved cooling water orifice to reduce plugging and utilizes the redesigned piston l
, tube and hydraulic buffer. Other changes include minor dimen-sional changes, component material upgrades and the redesign of the uncoupling rod to prevent inadvertent installation of the rod in the wrong spud hole which would lead to coupling difficulties.
This conversion provides increased CRD reliability and reduced maintenance.
This evaluation' reviewed the design modifications, system opera-tion, interface concerns, qualification, and surveillance testing 1 associated with this equipment substitution. This replacement resulted in a significant increase in the drive component safety design margin. It was concluded that the implementation of this safety-related modification did not constitute an unreviewed safety question pursuant to 10CFR50.59.
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'JAF-SE-87-159 - Replacement of Three-Valve Manifold with ,
Separate Whitey Valves for 02-3LIS-101B and 02-3LIS-101D (Minor Modification M1-87-171)
This safety evaluation provided the review for Minor Modification M1-87-171 which replaced the three-valve manifold on the high and low pressure connecting tubing for the HPCI reactor water level transmitters 02-3LIS-101B and 101D. This manifold was replaced ,
L with three separate instrument shutoff valves that provide the identical function of the previously installed manifold and are easier to operate.
All new materials were purchased as safety-related per QA Catego-ry I requirements and installed in accordance with the app 3.ieable design and installation criteria. There was no change required -
+ to the Technical Specifications or to the FSAR as a result of-this modification.
This safety-related minor modification did not adversely affect the operation of this system and was determined to not constitute t an unreviewed safety question, s
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JAF-SE-87-161 - Velan Valve Stem Material Substitution (Minor Modification M1-87-144)
This stfety evaluation provided the technical review for the i
vendor recommended material substitution for valve stems for various safety-related and non-safety related manually-operated gate valves. The new material which was approved for use was ASTM A-479, Type 410SS, Class 2, which replaced ASTM A-276, Type 410SS . hardened.
', An engineering review comparing the chemical and mechanical properties of these two materials concluded that the material was acceptable for use. This change did not impair the valve's
, operability and consequently the intended design function was not
': adversely affected.
This material substitution did not constitute an unreviewed safety question.
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and Experiments for 1988 :
JAF-SE-87-164, Rev. 2- Reactor Vessel Water Level Instrumenta-tion (Plant Modification F1-87-014) !
1 This safety evaluation provided a detailed analysis of the safety concerns associated with Plant Modification F1-87-014. The i
purpose of this safety-related modification was to change the wide range reference leg sensing lines and fuel zone instruments in order to reduce indication errors caused by potential high ;
Drywell temperatures under certain accident conditions. These improvements were completed in order to comply with the require-ments of NRC Generic Letter 84-23 and Regulatory Guide 1.97, Rev.
2.
The scope of this modification included the removal of the old reference leg systems for the wide range level instruments, the installation of the new sensing lines, ehe capping of conttinment penetrations no longer in use, the recalibration of fuel zone instrumentation to extend the range to the bottom of active fuel, the replacement of'certain level indicating instruments, and the deletion of instruments that were no longer needed.
This safety evaluation reviewed the issues and concerns involved with this modification. The sections of the FSAR and Technical Specifications which required revision because of this modifica-tion were specified in this evaluation. The requirements of the various applicable FSAR sections, Technical Specifications, Regulatory criteria, industry codes and standards were specified in detail and the steps taken to assure compliance were stated.
The steps required for the safe installation of this modification were also specified in great detail. The conclusion of this very comprehensive evaluation was that an unreviewed safety question pursuant to 10CFR50.59 did not exist.
The preoperational test (POT-02-2D) associated with this modi-fication was reviewed in Nuclear Safety Evaluation JAF-SE-87-165.
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l Page Annual Summary of JAFNPP Ch:ngas, Tcsts, and Experiments for 1988 JAF-SE-87-165 - Reactor Vessel Water Level Instrumentation Calibration and Functional Test (Preopera-tional Test POT-02-2D)
The purpose of this preorerational test was to calibrate and functionally test the instrumentation loops installed and/or affected by Plant Modification F1-87-014. The safety-related modification itself was reviewed in Nuclear Safety Evaluation JAF-SE-87-164, Rev. 2.
This test involved instrument line flow check valve operability verification, loop calibration, and functional testing for the level instrumentation associated with the new reference legs.
The test procedural steps were analyzed to ensure that no single accident occurring during the performance of this test could damage more than one safety division. This ensured compliance with the safety design bases. The test was conducted with the plant in cold shutdown.
The Technical Specifications required no changes as a result of this preoperational test and did not reduce the margin of safety as defined in the bases for Technical Specifications. It was concluded that the performance of this test did not involve an unreviewed safety question as defined by 10CFR50.59.
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'JAF-SE-88-001 - Special Cycling of 23MOV-19 (Temporary -
Operating Procedure TOP-90)
Special cycling of.the safety-related High Pressure Coolant [
Injection Discharge Valve (23MOV-19) was required in-order to ,
respond to NRC Bulletin IEB 85-03, " Motor Operated Valves Common :
Mode Failutes During Plant' Transients Due to Improper Switch 1 Settings". Specifically, TOP-90 addressed the closing of 23MOV-19 against hydrostatic-line pressure in the range of 1131 '
to 1181 psig. a This safety evaluation provided the safety analysis for the- '
performance of TOP-90. Provisions in TOP-90 regarding proper test pressure, thermal overload protection, and proper adminis-trative control were reviewed. All pertinent documents (includ-
- ing JAF Technical Specifications and Final Safety and Analysis i Reports) were reviewed and it was determined that the performance t of TOP-90 would not involve an unreviewed safety question pursu- ?
ant to 10CFR50.59. ,
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i Annual Summ ry of JAFNPP Chcngos, TOsts, I rnd Exp^ rim;nts for 1988 JAF-SE-88-003 - Drywell and Torus Spray Flow for Regulatory Guide 1.97, Rev. 2 (Plant Modification F1-87-048)
This safety evaluation analyzed the safety concerns associated with Modification F1-87-048 which provided low range indication of flow rate in two redundant trains of the Residual Heat Removal (RHR) System supply to Drywell and Torus sprays. Flow rate will be displayed on demand in the Emergency Plant and Information Computer System (EPIC) snd continuously on new flow indicators in the 09-3 panel. New flow measurement instrumentation was in-stalled for flow from RHR pumps 10P-3A,C and 10P-3B,D. The
- addition of this low range indication provides better resolution of flow readings and will aid the operator in mitigation of an accident.
The preoperational test (POT-10J) associated with this modifica-tion was analyzed in Nuclear Safety Evaluation JAF-SE-88-006.
This safety-related modification was designed to meet the re-quirements of Reg. Guide 1.97, Rev. 2, design and qualification criteria for a Category 2 variable Type D. The flow transmitters were environmentally qualified in accordance with Regulatory Guide 1.89 and the methodology described in NUREG-0588. Seismic qualifications were in accordance with the provisions of Regu-latory Guide 1.100. The primary flow elements were designed, fabricated, and tested ~in accordance with the applicable sections of the Power Piping Code ANSI B31.1-1967. The additional weight of the new flow elements and the changes made to the piping system were reviewed for seismic concerns and it was determined that the stresses were within allowable limits. There was no change required to the Technical Specifications as a result of this modification.
It was the determination of this evaluation that an unreviewed safety cuestion was not involved due to the implementation of this mocification. The effect of this modification was to increase the plant's capability for safe operation.
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JAF-SE-88-004 - Stack and Standby Gas Treatment Flow for Reg.
Guide 1.97, Rev. 2 (Plant Modification F1-87-046)
Modification F1-87-046 was performed to meet the requirements of Reg. Guide 1.97, Rev. 2, concerning the monitoring of variables required to determine the magnitude of any possible release of rac ioactive material. This modification installed new environ-mentally qualified flow measurement instruments to provide flow .
rate indication for the stack exhaust. Also, the existing instrumentation loop for the Standby Gas Treatment Purge Flow indication was upgraded to meet environmental and seismic crite-ria.
These flow rates will be monitored, on operator demand, in the Emergency Plant and Information Computer System (EPIC) by the newly installed flow instrumentation which is powered from a highly reliable (battery and diesel backed) power supply. The preoperational test associated with this modification was an-alyzed in Nuclear Safety Evaluation JAF-SE-88-011. -
The stack elements were seismically mounted in order to increase their reliability during and after a seismic event. The location of these elements in no way hinders the operation of the isokinetic probes and the velocity profile of the stack exhaust '
was not significantly altered by this modification.
No significant combustible material was added by this modifica- .
tion and there was no change to the Technical Specifications required because of this modification.
The implementation of this non-safety related modification, analyzed in JAF-SE-88-004, did not involve an unreviewed safety question and actually enhanced the plant's capability for safe operation.
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m Annual Summary of JAFNPP Ch ngas, Tcsts, and Experiments for 1988 i JAF-SE-88-005 - Cooling Water Flow to ESF System Components for Regulatory Guide 1.97, Rev. 2 (Preopera-
- tional Test POT-46D)
This preoperational test was performed in order to verify that the new transmitter loops installed by Modification F1-87-044 for the measurement of Emergency Service Water (ESW) flow operate within stated limits of accuracy. A safety analysis of this safety-related modification was contained in Nuclear Safety Evaluation JAF-SE-87-099.
Since the instruments tested were isolated-from the fluid system by the manifold valve during the test, the ESW System was com-pletely functional and capable of being utilized, if required, during the test. Therefore, there was no reduction in plant safety during the performance of this preoperational test. The test involved simulation of process conditions and did not involve running of major plant equipment.
There were no revisions required to the FSAR or the Technical Specifications because of this test and the safety evaluation concluded that the performance of this preoperational test did not involve an unreviewed safety question as defined by 10CFR50.59.
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JAF-SE-88-006 - Drywell and Torus Spray Flow for Reg. Guide i 1.97, Rev. 2 (Preoperational Test POT-10J) l The purpose of this preoperational test'(POT-10J) was to verify that the newly installed instrument loop components for Drywell and Torus spray flow operate within the stated limits of accura-cy. This test was performed by simulating flow at the individual transmitters utilizing a water manometer tester. A five point calibration test was performed on each transmitter loop with the expected readout results checked on the Control Room indicator on panel 09-3 and at EPIC.
This test was merformed while the plant was in the cold condi-tion. Since the instruments tested were isolated from the fluid system by the manifold valve during the test, the RHR System was completely functional and could have been utilized, if required, during the execution of this test. Therefore, there was no reduction in plant safety during the performance of this preoper-ational test and plant process parameters remained unchanged.
This test complied with FSAR Section 8.5.3 because no single error such as shorting of terminals in racks 9-24 or 9-25 could have resulted in the failure of safety-related components of the opposite train.
The safety-related modification (F1-87-044) which installed these new flow instrument loops (10FT-136A,B and 10FT-137A,B) was analyzed in Nuclear Oafety Evaluation JAF-SE-88-003.
It was determined that the performance of this preoperational test did not constitute an unreviewed safety question.
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Annuni Summary of JAFNPP Changos, Tests, and Experiments for 1988 JAF-SE-88-007 - High Pressure Coolant Injection Line Support Modification (Minor Modification M1-88-003)
Minor Modification M1-88-003 modified safety-related pipe support (PFSK-4516) angle steel members above and below trap 23T-3 by making them bolted support members. This modification was performed in order to facilitate disassembly and reassembly during maintenance on this trap.
Bolt loading was determined to meet the FSAR Seismic Class 1 requirements. No reduction in the margin of safety as defined in the basis for Technical Specifications or change to the Technical Specifications resulted from this modification to the HPCI drain line support.
This minor structural modification did not constitute an unre-viewed safety question as defined in 10CFR50.59. t i
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- Annu01 Summary of JAFNPP Chnngos Tosts, and Experiments for 1988 Y JAF-SE-88-009 - Feedwater Runback. Time Delay Relay Replace-ment (2A-K35A,B) (Minor Modification '
M1-88-004) t This safety evaluation provided the safety analysis for non-safety related Minor. Modification M1-88-004 which installed
. vendor recommended relay replacements for the feedwater runback
' time delay relays. This timer's function is to limit to a ';
. minimum or runback the recirculation pump speed if feedwater flow is not greater than 20 percent of the rated value. ,
The logic associated with the recirculation pump control system
.was not affected by this modification. The safety design basis of the= Reactor Recirculation System was not affected by this change and the Technical Specifications did not require revision. ,
The replacement relay was found to meet all technical and quality requirements specified for the original relay. This direct equipment replacement did not alter the functioning of the system ,
in any-way and-did not constitute an unreviewed safety. question, t
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Annual Summ2ry of JAFNPP Ch;ngas, Tosts, and Experiments for 1988 i JAF-SE-88-010 - Installation of Inline Filter in Pressure Sensing Line to 23PCV-50 (Minor Modification M1-88-006)
The purpose of this. safety evaluation was to provide the safety review of Modification M1-88-006 which installed a filter in the 1 pressure sensing line to the safety-related control valve 23PCV-50. This modification was performed in order to prevent the pressure snubber at 23PCV-50 (HPCI booster pump recirculation .
pressure control valve) from becoming plugged.
The added weight of the new filter was determined to not have an significant affect on the piping stresses involved. This modi-fication did not change the system in a manner that could affect the existing design bases and accident analyses.
This minor modification was determined to have no detrimental effects on the HPCI System and did not constitute an unreviewed safety question.
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Annu21 Summ;ry of JAFNPP Changas, Tosts, and Experiments for 1988 JAF-SE-88-011 - Stack and Standby Gas Treatment Flow Rate for Reg. Guide 1.97, Rev. 2 (Preoperational Test POT-26A)
The purpose of this preoperational test (POT-26A) was to verify '
the correct operation of the newly installed stack flow monitor-ing loop and the newly installed loop for the Standby Gas Treat-ment (SBGT) flow measurement. In addition, the readout devices on panel 09-75 and in the Emergency Plant and Information Comput-er System for the SBGT flow were checked for conformity with each other. This test was performed with the reactor in the cold condition.
During the initial test phase, the instrument loops were isolated
.from the system by the manifold valve. Therefore, no system capability was compromised and plant safety was not reduced. ,
This phase consisted of only simulating plant process conditions.
The second phase involved starting and stopping of the dilution fans and SBGT fans and comparing the loop readouts to the previ-l ously measured flow rates. No single error occurring during the execution of POT-26A would have had a negative impact on safe-ty-related equipment of redundant trains which complied with FSAR requirements.
No abnormal or potential conditions adverse to safety were ,
created and no change to the Technical Specifications were required as a result of this modification.
The non-safety related modification (F1-87-046), which this test was performed under, was analyzed in Nuclear Safety Evaluation JAF-SE-88-004.
The conclusion of this safety evaluation was that the performance of this test did not involve an unreviewed safety question.
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Annusi Summary of JAFNPP Ch:ngas, Tosts, and Experiments for 1988 JAF-SE-88-012, Rev. 1 - Cut Drywell Steel Frame to Clear 13MOV-15 Valve Actuator Interferences (Minor Modification M1-88-009)
This safety evaluation provided the review for safety-related Minor Modification M1-88-009 which removed a portion of one flange of a structural framing brace which was interfering with the handwheel operator of Reactor Core Isolation Cooling (RCIC) steam supply isolation valve 13MOV-15.
It was determined that the interference was being caused by the thermal movements of the RCIC steam supply line. The cutting and removal of the angle flange provided adequate clearance for thermal, deadweight, and dynamic movement of this piping.
The interfering drywell structural angle bracing was determined to be a secondary load carrying member providing bracing for the primary load carrying members that support main steam line snubbers. This bracing member was determined to be low stressed under the design loading conditions. This modification did not result in this member's inability to withstand the design loading.
- This minor structural modification did not adversely affect any safety-related systems and was determined to not constitute an unreviewed safety question.
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i Annual Summary.of JAFNPP Chtngos, Tosts, and Experiments for 1988 JAF-SE;,88-014 - Standby Liquid Control (SLC) Demineralized Water Test Tank Isolation Valve (11SLC-29)
Replacement (Minor Modification M1-87-053)
Minor Modification M1-87-053 replaced the 1" isolation valve (11SLC-29) on the demineralized water test tank (11TK-3). The replacement valve is of the same form, fit and function as the original valve and meets or exceeds the original plant design specification requirements detailed in the original Purchase Specification APO-16.
The new valve does not contain any cobalt alloys in order to limit the formation of activated corrosion products in the reactor coolant. The replacement of 11SLC-29 did not affect the safety design bases of the FSAR and no revision is required to the Technical Specifications.
It was the conclusion of this safety evaluation that an unre-viewed safety question was not involved due to the replacement of this non-safety related valve.
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l Page Annual Summ ry of JAFNPP Ch:ngas, Tosts, and Experiments for 1988 JAF-SE-88-016, Rev. 1- Interim Operation of Residual Heat Removal (RHR) Service Water System with Restricting Orifices (10RO-105A,B)
Removed l
l This safety review evaluated the acceptability of operating the RHR Service Water System with the safety-related restricting i
orifices downstream of RHR Heat Exchangers (10E-2A,B) removed.
! Investigation indicated that the orifices were removed in an attempt to gain additional flow during surveillance testing.
The restricting orifices were designed to provide additional system resistance to limit puma flow to between 8,000 and 10,000 l gpm. It was determined that the flow could be maintained in the proper design range by throttling the 10MOV-89 valves. Admini-strative controls to limit this flow to no greater than 8,000 gpm provides assurance that pump runout will not occur with these orifices absent. This condition of the RHR Service Water System was found to not affect the Technical Specifications and did not constitute an unreviewed safety question pursuant to 10CFR50.59.
Modification F1-88-011 was initiated to further investigate this
, situation and to provide for the return to the original system I
configuration if determined to be desirable.
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,' JAF-SE-88-017 -
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Interim Removal of Recirculation Pump Seal Staging / Inner Seal Flow Switch from Service (02-2FIS-26A)
This safety evaluation provided the evaluation for the temporary removal +of 02-2FIS-26A, Reactor Water Recirculation Pump Seal Cavity Flow Indicating Switch. This interim removal was de- .
termined to be acceptable since the seal leakage, which this switch is designed to indicate, can be determined from monitoring the pressure drop across the seals and the cavity temperature of each seal.
It was determined that this modification did not affect any systems in a manner that could change the existing design bases or accident analyses.
Based upon the review presented in this evaluation, it was '
concluded that an unreviewed safety question did not exist as a result of this non-safety related temporary modification.
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! Annu21 Summ0ry of JAFNPP Chtngos, Tosts, and Experiments for 1988 JAF-SE-88-022, Rev. 1- HPCI/RCIC Steam Leak Detection Setpoint Change (Minor Modification M1-88-026)
This safety evaluation provided the review for safety-related Minor Modification M1-88-026 which authorized the setpoint increase for seven steam leak Fesistance Temperature Detectors (RTDs). These RTDs provide steam leak detection for the High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) turbine steam supply lines in the Drywell Entrance Room and the Torus Room. This setpoint change was necessitated by the fact that the normal ambient temperatures in the areas are slightly higher than the temperatures originally calculated. The new setpoint was raised ten degrees from 133 to 143 degrees Fahrenheit in order to account for this higher ambient tempera-ture.
The RTD response time is sufficient to ensure that'the sensors will detect a HPCI or RCIC steam supply pime break in their areas within the maximum allowable time span wit:1 the setpoint raised.
The new setpoint did not invalidate the environmental qualifica-tion status of the equipment in these areas that is in the scope of 10CFR50.59. This raised atpoint is within the allowable limits of Technical Specification Table 3.2-2.
This setpoint change was determined to not affect any of the safety-related functions of plant equipment and did not increase the consequences or the probability of an accident. Modification M1-88-026 did not involve an unreviewed safety question as defined in 10CFR50.59.
Page Annual Summ2ry of JAFNPP Ch;ngos, Tosts, and Experiments for 1988 JAF-SE-88-023 - Primary Containment Isolation Valve Position for Regulatory Guide 1.97, Rev. 2 (Plant Modification F1-87-045) ;
Modification F1-87-045 was performed in order to satisfy the requirements of Reg. Guide 1.97, Rev. 2, concerning primary ,
containment isolation valves. This safety-related modification installed a containment valve position status array on panel 09-4, installed new containment isolation valves to provide for completely redundant loops on the Drywell radiation monitors. .
replaced two aged air-operated containment valves, replaced four solenoid-operated valves due to leakage, and deleted the old Beckman analyzers along with associated valves and tubing.
Interlocks were provided to ensure that the appropriate samales are directed to the Post-Accident Sampling System through the modified sample lines.
This safety analysis specified the Technical Specification and FSAR sections which required revision because of this modifica-tion. This evaluation also specified the various regulatory and .
industry. requirements and standards that were adhered to in this modification. This included environmental and seismic qualifica-tion criteria.
Construction inside the drywell was addressed to help ensure that radiation exposure associa'ted with the implementation of this modification'was kept as low as reasonably achievable.
The margin of safety as defined in the bases of any Technical Specification was not reduced and this modification was de-termined to not involve an unreviewed safety question.
The preoperational test associated with this modification was analyzed in JAF-SE-88-024 Page ,
Annual Summ ry of JAFNPP Chcngas, Tosts, and Experiments for 1988 ,
JAF-SE-88-024 - Primary Containment Isolation Valve In-dication for Regulatory Guide 1.97, Rev. 2 (Preoperational Test POT-27V)
The purpose of this test procedure was to verify valve operabil-ity and position indication for the containment isolation valves installed and/or modified by F1-87-045. This safety-related modification was evaluated in Nuclear Safety Evaluation JAF-SE-88-023.
Under the preoperational test, each valve associated with this modification was stroked to verify correct valve position in-dication at the appropriate status location. In addition, the control and monitoring circuit for 27SOV-137 was tested to check out valve circuit interlocks and logic interfaces.
POT-27V was performed with the plant in the cold shutdown condi-tion. In this condition, the containment sample system is not required to be oparational. Therefore, there was no reduction in plant safety and the plant arocess parameters remained unchanged during the performance of tL11s test.
It was determined that this preoperational test did not involve ,
an unreviewed safety question as described in 10CFR50.59.
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I Annu21 Summary cf JAFNPP Ch ngas, Tosts, and Experiments for 1988 JAF-SE-88-026, Rev. 1- Cleanup Decant Pump (20P-44) Casing Material Change (Minor Modification M1-87-156)
Minor Modification M1-87-156 was performed in order to allow for the use of replacement parts manufactured using different mate-rials than were originally utilized. This evaluation detailed the material differences associated with this modification. The caterial changes were found to be acceptable in form, fit, and function and will not adversely affect the operation of the cleanup decant pump. The cleanup decant pump is in the Radioac-tive Waste System and is not a safety-related component.
This safety evaluation concluded that an unreviewed safety question did not exist as a result of this non-safety related minor modification.
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l Annual Summ;ry of JAFNPP Chtnps , Tosts ,
and Experiments for 1988 ,
L JAF-SE-88-032 - Turbine Building Crane (88CR-1) Radio Upgrade from Analog to Digital Control (Plant Modi-fication F1-87-085)
Plant Modification F1-87-085 replaced the non-safety related Analog Radio Control System for the Turbine Building Crane (88CR-1) with a new digital radio control system. This change ,
improved system reliability and reduced the maintenance that had !
previously been required with the Analog System, i The new radio control incorporates several safety features which .
allow only the desired carrier signals to be received. Also, the transmitter has self-centering switches that always returns to the "0FF" position when released, thereby stopping the crane motion.
This crane control system is not required for the safe shutdown of the plant and the installation of the new system did not affect the safety margin of any Technical Specification design bases.
In conclusion, it was determined in this safety evaluation that this control circuit improvement did not constitute an unreviewed safety question.
f Page Annual Summ ry of JAFNPP Ch;ngos, Tosts, and Experiments for 1988 '
JAF-SE-88-034 - 13MOV-131 Motor Actuator Spring Pack Substi-tution (Minor Modification M1-88-030)
Minor Modification M1-88-030 replaced the spring pack in the motor actuator of valve 13MOV-131. This non-safety related "
modification allows the sprimg pack to be more " finely tuned" to the torque output requirements for this valve configuration.
This spring pack replacement and corresponding torque switch setting change resulted from refinements in actuator design and sizing analysis and actuator testing. This modification enhances the operation of 13MOV-131 and there are no additional failure mechanisms introduced by the incorporation of this design change.
This evaluation discussed the issues concerning the proper torque switch setting for this actuator and the methods used to deter-mine this value. The ability of 13MOV-131 to perform its intend-ed safety function was enhanced by this .nodification and no margin of safety was reduced.
It was the conclusion of this safety evaluation that an unre-viewed safety question as defined in 10CFR50.59 did not exist as a result of the implementation of this modification.
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JAF-SE-88-035 - Temporary Repair of Reactor Protection System (RPS) Fuse Block (Minor Modification M1-88-034)
The purpose of this safety evaluation was to assess the impact on system operability.and nuclear safety resulting from the tempo-rary repair of RPS Fuse Block 5A-F3F in panel 09-17. The repair consisted of attaching a piece of bakelite (Phenaolic) from a new fuse terminal block to the broken fuse block with Hardman Epoxy.
This bakelite isolation barrier was installed to prevent the possibility of an electrical short which could have caused an actuation of the RPS logic for the safety-related Main Steam Line Isolation Valve 29A0V-86B and initiated a half scram signal.
There were no changes to the FSAR or Technical Specifications required due to this modification. It was determined that the panel did not need to be seismically requalified since there was no significant change in weight due to this repair. This modi-fication did not alter the system in a manner that could change the existing design bases or accident analyses.
The minor repair work associated with this modification did not constitute an unreviewed safety question.
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Annual Summ3ry of JAFNPP Chcng s, Tcsts, and Experiments for 1988 JAF-SE-88-037 - Drywell Equipment Drain Sump Valve 20A0V-95 Functional Test (Preoperational Test POT-20Y)
This safety evaluation was written to evaluate the impact of
-POT-20Y on the safe operation of the plant. This preoperational test was performed to functionally test the operation of the new control switch installed for the safety-related Drywell equipment
Valve 20A0V-95 is a Primary Containment Isolation System (PCIS) isolation valve whose circuitry was modified by Minor Modifica-tion M1-85-056 (evaluated in JAF-SE-87-129). This test estab-lished that the PCIS isolation function of this valve was not compromised by this modification. j Since the reactor was in a cold condition when POT-20Y was performed the PCIS function of 20A0V-95 was not required per the Technical Specifications. This test initiated a simulated PCIS isolatinn signal. The Shift Supervisor was notified prior to this instiation to minimize the effect on other system components and plant operations. This test did not alter the safety design basis of any existing systems described in the FSAR. ,
The plant conditions required for this test ensured that no adverse effects on the plant's safety would result from the performance of POT-20Y. It was the conclusion of this safety evaluation that an unreviewed safety question pursuant to
.10CFR50.59 was not involved due to the conduction of this test.
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- . s Annual Summ ry of JAFNPP Chengos, Tests, and Experiments for 1988
, JAF-SE-88-038 - Standby Gas Treatment (SBGT) System Centrifu-gal Exhaust Fan (01-125FN-1A,B) Packing Replacement (Minor Modification M1-88-038)
This safety evaluation provided the safety analysis for Minor Modification M1-88-038 which authorized the use of vendor recom-mended replacement packing for the safety-related SBGT centrifu-gal exhaust fans. The new die-formed grafoil packing replaced the originally supplied packing which had contained asbestos and lead ribbon.
The die-formed graphoil packing has a higher nuclear radiation resistance value and also does not require replacement as fre-quently as conventional packing. This packing replacement did not involve a change in Technical Specifications or in the safety design bases as described in the FSAR.
This material substitution did not adversely affect the SBGT System and did not constitute an unreviewed safety question.
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"J LJAF-SE-88-039 - Replacement of 87AHB-25 Velan> Valve'with l Henry.Vogt Valve'(Minor Modification o t M1-88-047) !
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The purpose of this safety evaluation was to provide the safety l wp review of Modification'M1'-88-047. .This non-safety.related l modification. replaced 87AHB-25 which is a normally closed manual
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i bypass valve for pressure control valva 87PCV-101. The replace-ment valve was procured to specificaticas!which conformed-with the origina1' technical and quality requirements for this applica-tion, j This valve-performs no safety function and its failure would not !
. affect the performance'of'any safety-related system. This valve replacement did not constitute an unreviewed safety question. y 1
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This: safety evaluation provided the. review for Minor Modification >
n M1-88-057 which replaced the 4-way ball valve 27CV-710 with the vendor recommended replacement model.. This valve acts to allow the PASS to select gas semples from the Drywell, suppression pool or secondary containment atmosphere. The new valve.was procured !
.to meet and/ or exceed all the applicable design specifications imposed on the original valve.
This non-safety related valve substitution in no way adversely affected the operation of the system and did not constitute an t unreviewed safety question, n.
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b s Annuc1' Summary of JAFNPP Chtngos, Tcsts, s and Experiments for 1988 e JAF-SE-88-044 - Replacement of Auxiliary Boiler Feedwater l n Valves (87LCV-103A,B and 87AHB-11A,B) (Minor !
Modification M1-87-120) :
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. Minor Modification M1-87-320 involved the replacement of the P non-safety related auxiliary boiler feedwater valves which werc ,
beyond repair. This safety evaluation reviewed the vendor
- recommended realacements and determined ~their design:to be '
consistent with those of the valves being replaced. The increase in weight presented by these new valves was analyzed for its y
- affect on the piping system stresses and it was determined to be '
well within the code allowable. The Auxiliary Boiler System does .-
not: perform any, safety-related function and the failure of any of i the subject valves would not affect any plant safety-related components. !
The conclusion of this safety evaluation was that this valve substitution did not constitute an unreviewed safety question.
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and Experiments for 1988 '
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'-~~ Replacement of Feedwater Pump Inlet Relief 1 Valve 34RV-104A&B (Minor Modification >
M1-87-036)-
This safety evaluation provided the review for non-safety related-Minor Modification M1-87-036 which allowed for the replacement of relief valves 34RV-104A&B. This component upgrade was recom-
These pressure relief valves in.the Feedwater System relieve to the main condenser through a one inch line. The new valve's i
' increased flow capacity was determined to be within the capabil- i ity:of the existing piping configuration. The slight increase.in weight of these valves was determined to have no significant effect on the piping systeni. FSAR Section 10.8.3 and the Techni-cal Specifications were reviewed and no changes were required as !
a result of this modification.
The margin of safety as defined in the basis of Technical Speci-fications was not reduced and it was concluded that this relief valve replacement did not constitute an unreviewed safety ques- '
tion es defined in 10CFR50.59.
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Annucl; Summary of JAFNPP Chcngos, Tosts, and Experiments for 1988
'JAF-SE-88-048 - Site Utilities Expansion Project - Site-Civil Work (Plant Modification F1-86-003C)
This safety evaluation was written to provide a review of the safety' concerns associated with the non-safety related Plant P Modification F1-86-003C.. As a result of the construction of new expansion facilities and to correct existing drainage conditions, the yard storm drainage system and site grading were upgraded.
In addition, the east access road and the parking lot were '
paved / surfaced and lighting was added.
The hydrology of the site, as defined in the FSAR, remained unchanged as a result of this modification and there was no change to the Technical Specifications. This review concluded l
that an unreviewed safety question pursuant to 10CFR50.59 did not exist as a result of this modification.
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I JAF-SE-88-049, Rev. 1- Heavy Load Analysis for Lifts in the JAF Fuel Pool This safety evaluation addressed the heavy load concerns associ-l ated with the placement of equipment into the Spent Fuel Pool.
l The equipment was required for the processing and removal of L
irradiated hardware. This evaluation discussed the load lift requirements to ensure that a factor of safety of at least 10:1 was maintained at all times which satisfied NUREG-0612 concerns.
H Site evaluations were performed on critical components in the l load path of the equipment to assure that this factor of safety was maintained. The removal of equipment from.the spent fuel pool was accomplished in accordance with the approved plant
! procedures to ensure that radioactive contamination was not i spread from the '3001 to other areas and to keep dose rates as low as reasonably achievable.
It was concluded that an unreviewed safety question as described i l by 10CFR50.59 was not involved as a result of this non-safety l related activity.
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AnnuS1 Summtry of JAFNPP Chengss, Tosts, and Experiments for 1988 JAF-SE-88-050 - Moving' Fuel. Pool Equipment Low Specific Activity:(LSA) Boxes This safety evaluation was written in order to address the~ safety concerns involved in the moving of LSA boxes between the Refuel-Floor (369' elevation) and the Reactor Building =(272' elevation).
The various LSA boxes contained equipment utilized in the removal of irradiated hardware from the Spent Fuel Pool. The evaluation demonstrated that this non-safety related activity complied with
' NUREG-0612 guidelines concerning the handling of heavy loads.
This activity did not reduce the margin of safety as defined in the basis'for the Technical Specifications and it was concluded that an unreviewed safety question as defined by 10CFR50.59 did not exist.
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Annu21-Summtry of JAFNPP Chingos -Tosts, and Experiments for 1988 JAF-SE-88-051 - Control Rod Hanger in Spent Fuel Pool This safety evaluation addressed the temporary installation of the Control Rod Hanger Assemblies on the wall of the-Spent Fuel Pool. The fuel poof cleanu) project necessitated the removal of the control rod storage raci. The General Electric designed hanger assemblies provided temporary storage of the twenty control rods which were removed from.this rack. This evaluation included a separate General Electric evaluation which provided an in-depth review-of these wall mounted hangers and concluded that design provisions and administrative controls precluded-any degradation of the function, or method of performing the func-tion, of any safety-related structures, systems, or components.
The' conclusion of the overall safety evaluation was that an L unreviewed safety question pursuant to 10CFR50.59 did not exists as a result of this non-safety related task, i
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Annutl Summery of JAFNPP Chengas, Tosts, and Experiments for 1988 JAF-SE-88-052 - James A. FitzPatrick Quality Assurance.
Classification Program Update (Q-List)
This safety. evaluation provided an.in-depth analysis of the-project to review and determine the QA safety classification of JAF systems and equipment as was required by NRC Generic Letter 83-28, " Required Action Based on Generic Implications of Salem ATWS Events". A-result of this project was the establishment of.
- a. Master Equipment List (MEL). Items which were determined to need reclassification, which included safety-related and non-safety. related components, were identified and_ the justifica-tion and reasoning for these changes in classification were presented in detail. ,
The conclusion of this safety evaluation was that an unreviewed i safety question pursuant t'o 10CFR50.59 did not exist as a result of these reclassifications.
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. 4 Annual Summary of JAFNPP Changes, Tosts, and Experiments for 1988 JAF-SE-88-053 - Condensate Demineralizer Resin Ration Change i
This safety evaluation reviewed the implementation of a resin "
ration change for the non-safety related condensate demineralizer beds. The Cation to Anion (C/A) resin ratio of 2.0 by volume was changed to a C/A ratio of 1.2 by volume. The new resin bed ratio has been shown to have a greater capacity for removing ionic impurities specific to JAF. This review provided an overview of' the operations associated with these beds and the benefits gained-by altering the C/A ratio.
Specified in this evaluation were the required changes in the FSAR and the Site Operating Procedures. No sections of the Technical Specifications were affected. This comprehensive review of the Condensate Demineralizer System's vessels and capabilities concluded that this change did not involve an unreviewed safety question.
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-Annual Summery of JAFNPP Changes, Tests,
'- .and Experiments for 1988 JAF-SE-88-054 - Replacement of Valve 13MOV-15 (Plant Modi- -
fication F1-87-016)
This safety evaluation provided the safety review for Modifica-tion F1-87-016. This safety-related modification replaced the containment isolation valve 13MOV-15 in the Reactor Core Isola-tion Cooling (RCIC) turbine steam supply line. The new valve was relocated to a more accessible location to enhance future opera-tion and' maintenance.1 This valve was replaced as part of a continuing effort to improve the leaktightness of-the primary containment; The valve's operator is environmentally qualified for containment isolation service..
An' increase in valve weight required that several supports in the piping run be redesigned. The piping run was evaluated for pressure, deadweight and thermal loads, and was determined to be seismically acce'3 table. The replacement valve was purchased to meet or exceed the original specification requirements of AP-13.
The replacement of.this valve did not change the system in a manner that altered the existing design bases and accident-analyses. '
The conclusion of this safety evaluation was that an unreviewed safety question-as defined.in 10CFR50.59 did not exist as a result of this modification.
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Annus1 Summtry of JAFNPP Changes, Tests, and Experiments-for 1988 i
JAF-SE-88-055 - Replacement-of Valve 12MOV-15 (Plant Modi-fication F1-87-130)- ;
-This safety evaluation provided the safety analysis for Modifica- <
tion F1-8f-130 which replaced safety-related containment isola-tion valve 12MOV-15 in the Reactor Water Cleanup Supply Line.
The-replacement valve was determined to meet or exceed the original specification requirements for this QA Category 1 valves.
The replacement valve's operator is environmentally qualified as required in the plant design bases. The necessary pipe sup3 orts were redesigned to meet the new piping loads presented by tae slightly heavier valve. .As required by the FSAR, this valve was brittle fracture toughness tested to demonstrate a nill ductility.
transition temperature at least 60 degrees Fahrenheit lower than the lowest service temperature of the Primary Reactor Coolant System. The margin'of safety as defined in the basis of any Technical Specification was not reduced as a result of this
. modification.
Based upon the engineering design and quality assurance require-ments set forth during procurement, design, and-installation of this replacement valve, it was determined that the implementation of this modification did not constitute an unreviewed safety question.
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-Annuti Summ:ry of JAFNPP Changes, Tests, and Experiments for 1988 JAF-SE-88-056, Rev. 2- Replacement of Valves 10MOV-26A, -26B,
-31A,-and -31B (Plant Modification F1-87-133) ;
I This safety evaluation was performed in order to evaluate the replacement of the safety-related Residual Heat Removal (RHR) i containment isolation valves on the_ containment spray supply '
lines to the Drywell. This replacement was completed as part of the continuing effort to improve the leaktightness of the primary containment.
The original gate valves 10MOV-31A and 31B-were replaced with ;
globe valves in. order to enhance the operability and performance of the containment spray mode of the RHR System. These valves are used for throttling purposes which the globe valves are designed to do while the gate valves were not. This added control in throttling ability reduces the possibility of an uncontrolled cooldown of the primary containment during some postulated-loss of coolant accidents.
l A change in the logic for these valves was also performed by !
moving the. seal-in circuitry from the 10MOV-31 valves to the 10MOV-26 valves. The logic change for these valves was de- t termined to have no affect on the original design basis for the RHR System.
The 10MOV-26_ valves were replaced with gate valves which utilize l
- a. double disc arrangement which have been known to experience l overpressurization problems. In order to prevent this potential problem from occurring, a small diameter hole was drilled in the ;
downstream disc closest to the containment. As a result of this ;
action, no change is anticipated to the expected penetration !
,- leakage rate and the maximum differential design pressure for l opening the valves shall not be exceeded. :
L p Where required, the supports were redesigned to support the new i piping loads. This modification did not alter the system in a manner that could change the existing design bases and accident ;
analyses.
The conclusion of this safety evaluation was that this valve replacement enhanced plant safety and did not constitute an unreviewed safety question.
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Annual Summary of JAFNPP Chtnges, Tests, and Experimtnts for 1988 LJAF-SE-88-057 - Replacement of Valves 23HPI-11, -12, -65
-(Plant Modification F1-87-134)
Plant Modification F1-87-134 replaced two containment isolation ;
valves (23HPI-65 and 23HPI-12) and one maintenance valve
'(23HPI-11). The two containment valves were replaced with. lift check type valves instead of.the previously installed swing type check valves. 23HPI-11 remained a gate type maintenance valve.
This safety-related modification was performed as part of the continuing effort to improve the leaktightness of the primary containment.
These new QA Category 1 valves were purchased to meet or exceed the original specification requirements for these valves.
The.new lift check valves were chosen in order to improve the check valve operation and improve the performance of the check valve function toward achieving optimum isolation capability.
l: Pipe' supports were modified as required to properly support the new stress loads presented as a result of the increased valve weights.
' Based upon the engineering design and quality assurance require-
.ments set.forth during the design and installation of this modification, it was determined that these replacements enhanced plant safety and did not constitute an unreviewed safety ques-tion.
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Annual Summtry of JAFNPP Changes, Tests, and Experiments for 1988 1
'JAF-SE-88-058 - Replacement of Valves 13RCIC-7-and 13RCIC-8 (Plant Modification F1-87-135)
This safety evaluation provided the safety analysis regarding the implementation of Modification F1-87-135 which replaced two safety-related containment isolation valves (13RCIC-7 and 8).
These new check valves are located on-the discharge of the. .
Reactor Core Isolation Cooling (RCIC) vacuum pump.
The valves were purchased to specifications which-equalled or surpassed those of the original saecification requirements Piping supports were redesigned where required to account for the increased loading due to the. increase in valve weights.
This valve replacement did not alter the system in a manner which affected the existing design bases or accident analyses. It was. '
the conclusion of this safety evaluation that this modification enhanced the plant's safety and did not constitute an unreviewed-safety question as defined in 10CFR50.59.
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. =w Annun1'Summ3ry;of: JAFNPP Chengos, Tests,
,and Experiments'for 1988 JAF-SE-88-059, Rev. 1 - Replacement-of Valves 13RCIC-4 and 13RCIC-5 (Minor Modification M1-87-136)
This safety evaluation provided the review for the activities associated with Modification M1-87-136 which replaced the safe-ty-related swing type check valves located on the Reactor Core Isolation Cooling-(RCIC) turbine exhaust line with new lift-type check valves. This realacement was performed in order to improve :
the leaktightness of the primary containment and to improve the
_ performance of the valves during test conditions.
The new valves were purchased to specifications which equalled or surpassed the requirements for the original valves. Due to the increased weight of these valves, piping supports were redesigned where required in order to ensure that the Seismic Class I requirements for this system were maintained.
This modification also changed the RCIC turbine exhaust pressure setpoint from 25 psig to 50 psig as recommended by General Electric in order to reduce the likelihood of turbine trips on high exhaust pressure. The princiaal consequence of this-setpoint change is steam leaiage through the gland seal' system.
The radiological effects of this leakage was evaluated by GE and ,
were found to be within the limits of 10CFR20 and 10CFR100. J The margin of safety as defined in the basis of any Technical Specification was not reduced by these valve realacemento and the conclusion of this-safety evaluation was that this modification did not constitute an unreviewed safety question.
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Annual Summtry of JAFMPP Chcngos, Tssts, and Experiments for 1988 JAF-SE-88-062 --
Main Low Pressure Turbine LP "A" Rotor Replacement (Minor Modification M1-88-080)
Minor Modification M1-88-080 replaced the "A" low pressure- ,
-turbine rotor with one of integral design. This non-safety :
related replacement was performed due to the fact that tne
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original rotor had developed indications of. stress corrosion
, cracking which in turn increased the probability of generating missiles due to a partial or total failure of a rotor wheel. The inctallation of the integral design. rotor eliminated this poten- '
tial problem.
This evaluation' discussed the industry's experience with this type of rotor and.the testing criteria necessary to verify the required aroperties. Stress-corrosion cracking was addressed
-along with the design considerations taken to minimize this type of problem. It was determined that this rotor replacement did not increase the probability of missile generation and that the margin of safety as defined in the bases of any Technical Speci-fications have not been reduced by this modification.
This rotor replacement reduced the probability of rotor failure >
and did not constitute an unreviewed safety question.
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Annual Summ0ry'of JAFNPP Ch ngss, Tests, rnd Expnrim;nts for 1988 JAF-SE-88-064 - Addition of Permanent Piping Connections and Concrete Core Bores for Chemical Decontamina-tion Activities (Plant Modification F1-87-166)
~
This safety evaluation provided-the safety analysis for Modifica-tion F1-87-166. This safety-related modification consisted of providing the necessary connections for-the injection of_ chemical decontaminants into th'e Reactor Water Recirculation (RWR) System, the jet pump-nozzle drain lines, and the Residual Heat Removal (RHR) service water line between 10MOV-149B and 10RHR-177B. In -
addition, core bores were made between the Standby Gas Treatment :
Room and the Reactor Building to allow for the easy flow of
, process chemicals,. water, air and various electrical and in-L strumentation connections from vendor equipment to injection L points.
Special precautions were taken to ensure that the addition of the necessary fittings to allow process chemical injection did not L adversely affect the systems that were decontaminated. Upon completion of the decontamination process, the core bores were subsequently sealed:in accordance with approved plant procedures to ensure compliance with the fire barrier requirements.
The operation of the systems was not altered in any way by the o addition of these permanent piping branch connections. It was the conclusion of this safety evaluation that this modification did not constitute an unreviewed safety question. pursuant to 10CFR50.59.
A safety analysis of this'chemfcal decontamination process was contained in Nuclear Safety Evalu,ation JAF-SE-88-098, Rev. 1.
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_Annu 1 Summary of JAFNPP Chtngso, Tosts, and Experiments for 1988 i
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JAF-SE-88-066 - Safety Ladder - Administration Building Roof to Administration Building Fan Room Roof (Minor Modification M1-88-072) i I
Minor Modification M1-88-072 installed a safety ladder connecting i the Administration Building roof and the Administration Building Fan Room roof. This non-safety related modification was per-formed to permit the safe access and egress for operators during ,
shift tours and inspections of equipment in the area. l l
This safety ladder was an Occupational Safety and Health Adminis- i tration (OSHA) approved Heavy Industrial Type 1A (300 lb. rat-ing).- The design and installation were in conformance with all i applicable OSHA requirements. Failure of this ladder will impose i no impact upon plant safety-related equipment.
An unreviewed safety question did not exist as a result of this minor modification.
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. Annual Summtry of JAFNPP Changss, Toots, and Experiments for 1988
',- 'JAF-SE-88-067, Rev. 1- Second Level of Undervoltage Protection
- Addition of 45 Second Time Delay Relay (Plant Modification F1-88-001)
The purpose of this safety evaluation was to provide the safety review of Modification F1-88-001. This safety-related modifica-tion consisted of (1) the addition of Loss of Coolant Accident (LOCA) contacts.in series with the existing 27T2 time delay contact to initiate an Emergency Diesel Generator (EDG) start and subsequent' load shedding after 9 seconds of degraded voltage coincident with a LOCA condition, and (2) the addition of a 45 second time delay relay 27T3 to initiate an EDG start after 45 seconds of degraded voltage. A computer alarm to indicate the operation of this 45 second relay was also added.
The purpose of this modification was to allow the operator enough time to transfer loads without adversely affecting the Class 1E buses, but still restore normal bus voltage on short notice under LOCA conditions.
The ability of existing plant equipment connected to the Class 1E bus to withstand a 45 second undervoltage condition without adverse effects was determined by the Time Delay Extension Study performed by the Stone and Webster Engineering Corporation. '
This safety evaluation addressed equipment capability, system interface, quality assurance, code satisfaction, and Technical Specification requirements and determined that an unreviewed safety question was not involved as a result of this modifica-1 tion.
A safety analysis for the preoperational test associated with this modification was contained in Nuclear Safety Evaluation JAF-SE-88-068.
Page -104-
1 Annucl~Summ;ry of JAFNPP Chengos, Tosts, and~ Experiments for-1988 JAF-SE-88-068 - Second Level of Undervoltage Protection -
Addition of 45 Second Time Delay Relay (Preoperational Test POT-93C)
The purpose of Preoperational Test POT-93C was to verify the operability of the new time delay relay and new computer alarm installed by modification F1-88-001. A safety analysis of this safety-related modification was contained in Nuclear Safety Evaluation JAF-SE-88-067, Rev. 1.
The Degraded Voltage Protection System was functionally tested to verify proper relay operation under simulated degraded voltage '
conditions (less than or equc1 to 90.8% of nominal) and Emergency Diesel Generator (EDG) start signals. Core Spray and Residual Heat Removal logic relays used as a permissive for EDG starting after 9 seconds of degraded voltage were also functionally
-tested.
This test was reviewed to account for testing errors and it was determined that no single-error would have had a negative impact on safety-related equipment of redundant trains which complies with FSAR requirements. This preoperational test did not involve any change.to the Technical Specifications or to the FSAR. ,
This safety evaluation determined that this test did not involve cn unreviewed safety question.
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Annu21 Summary of JAFNPP Chzngos, Tests, and Experiments for 1988
-JAF-SE-88-070 - Reactor Feedwater Pump Auxiliary 011 Pump Running Annunciator (Plant Modification F1-87-123) 1 This safety evaluation reviewed the safety concerns associated '
with Plant Modification F1-87-123. This non-safety related modification revised the Reactor Feedwater Pump (RFP) auxiliary oil' pump's annunciator alarm window logic circuitry to allow annunciator activation only when the RFP auxiliary oil pumps are ,
l' started automatically. Under the previous arrangement, this annunciator was activated for both automatic and manual starts.
Since these pumps are manually started as a normal part of plant l operations, it was determined that this annunciation was not L necessary.
L The Reactor Feedwater System is not important for the safe shutdown of the plant. A failure of the RFP auxiliary oil pumps and associated annunciator circuit will not have an adverse affect on any safety systems evaluated in the FSAR. There were no changes required to the FSAR or Technical Specifications as a result of this modification.
The conclusion of this safety evaluation was that this logic change did involve an unreviewed safety question as defined in 10CFR50.59, 1
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J. ; 1 Annual Summ ry of JAFNPP Changes, Tosts, and Experiments for 1988 JAF-SE-88-072 - ' Main Steam Isolation Valve (MSIV) Actuator Rebuild and Upgrade (Plant Modification F1-88-096)
& This safety evaluation provided the safety analysis for safe-ty-related Plant Modification F1-88-096 which upgraded the actuators to the manufacturer's current design, This improvement in actuator design, based upon the vendor's service experience,-
included the modification of several subcomponents and the replacement of'other items no longer manufactured.
The new items procured for use on the MSIVs are replacements in ;
kind and will not affect the form, fit, or function of the i actuators. All items were procured to the requirements of AP-20 and to the applicable environmental qualification criteria for ;
electrical components, !
The actuators seismic qualification was not affected by this modification and there was no change required to the FSAR or to
- i the Technical Specifications.
It was determined that the modified actuator meets all of the design requirements of the original procurement specification and that the design upgrade did not involve an unreviewed safety )
question, ,
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Annucl Summ3ry of JAFNPP Chenges, Tests, and Experiments for 1988 JAF-SE-88-073 - Permanent Capping of Primary Containment Penetrations X-21, X-61, and X-50a (Plant Modification F1-87-066) l The purpose of this modification was to eliminate future local ,
leakage rate testing of containment isolation valves 39 BAS-4, '
=39 BAS-5, 39SAS-9, 39SAS-10 on the Service Air and Breathing Air l Systems. Portions of piping near the primary containment pene- ,
trations X-21 and X-61 were removed and welded pipe caps were '
added inside and outside primary containment. Also, more piping was removed from previously cut and capped piping located out-board of penetration X-50a in order to eliminate a radiation " hot spot". ,
i The Service Air and Breathing Air System headers located inside l the primary containment were retained and equipped with quick
- L disconnect fittings to permit the installation of temporary air l supply for use during plant outages. The welded end-caps were l instviled in accordance with the original Piping Specification AP-23 and were liquid penetrant examined, radiography tested, and pneumatic pressure tested as is required.
?. The FSAR and Technical Specifications required _ updating to ;
o reflect the elimination of containment isolation valves for l- m penetrations X-21 and X-61. It was determined that the ..
L implementation of this modification did'not affect the plant's design bases or accident analyses.
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This safety evaluation concluded that this safety-related taodi-fication did not constitute an unreviewed safety question as defined in 10CFR50.59.
Page -108-
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. Annual 1Summ2ry of JAFNPP Chengss, Tests, and Experiments for 1988 JAF-SE-88-074 - On-Site Storage of Radioactive Low Pressure Rotors This safety evaluation addressed the on-site storage of the two non-safety related low pressure rotors which were replaced with rotors of improved design. The rotors are stored in specifially adapted covered rail cars within the site perimeter fence.-
A non-water-coluble coating was apalied to the rotors for further protection in the case-of the unlikely event of exposure-to the ,
elements. It was determined that this storage complied with the requirements of NRC Generic Letter 81-38, " Storage of Low-Level Radioactive Wastes-at Power Reactor Sites". Compliance with 10CFR190, " Environmental-Dose Standards" were assured by compli-ance with plant Technical Specifications. This evaluation concluded that this on-site storage does not constitute an unreviewed safety question pursuant to 10CFR50.59.
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Annum 1 Summary of JAFNPP Changes, Tests, and Experiments for 1988 JAF-SE-88-075 - Top Head Assembly Replacement of the Off-Gas Dryers 107D-6A&B (Plant Modification-F1-87-086)
Modification F1-87-086 replaced the top head assembly (including heater wells and heater elements) of each drying tower with an improved-design from the original equipment manufacturer. The new design incoraorates expansion' joints in the heater wells to compensate for t3e possibility of occasional over-temperature conditions.-
-The previous off-gas dryer heating elements had a tendency to back out of their heater wells due to thermal distortion. This l- modification corrected this problem and thereby improved system operation and maximized-its availability.
The safety evaluation determined that the off-gas system's safety l design basis was not affected because this modification involved L replacing the dryer top head assembly which is equal in fit, form l and function to the original design with the addition of expan-l sion joints to the heater wells being a design improvement. This l modification did not reduce the margin of safety as defined in L the Technical Specifications because there are no safety limits l associated with this non-safety related component and the sys-tem's operation is not affected.
There was no unreviewed safety question involved as a result of this non-safety related modification.
Page -110-
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. Annual" Summary of~JAFNFP Changes, Tests, and Experiments.for 1988 JAF-SE-88-076, Rev. 1 - Genera 1EElectric HGA Control Relay Replacement Upgrade with General.Elec-tric Type HMA (Minor Modification ,
M1-88-100)_
Minor Modification M1-88-100 provided for the replacement of the-GE supplied HGA control relays with HMA control. This safe-ty-related modification was performed as a result of NRC Informa-tion Notice 88-14 which identified problems with the" seismic-qualifications of the_HGA1 relays. The-new HMA relays are func-tionally equivalent to the original relays, but were specifically designed'for seismic purposes.
L This modification did not introduce any additional components or-
. provide any additional failure modes to the plant systems. The new relays did not change any system in a manner which adversely affected the existing design bases and accident analyses. ,
It was determined in this evaluation-that'this modification enhanced plant safety by providing relays with upgraded seismic capability and that this replacement did not constitute an unreviewed safety question pursuant to 10CFR50.59.
Page -111-
4 Annual Summtry of JAFNPP Changes, Tests, and Experiments for 1988 i
JAF-SE-88-077 - Velan Valve Parts (3/4" thru 2" VOS-150D)-
Material Substitution (Minor Modification M1-88-089) ,
Minor Modification M1-88-089 authorized the use of vendor recom-mended replacement parts made of material different from material-used originally. Theimechanical and chemical properties for the
~
new material used in the retaining ring and back seat for these valves (safety-related and non-safety related) were reviewed and it was determined that these replacement parts were acceptable for their intended'use. >
This material substitution did not degrade the ability of these valves to perform their intended design functions and in no way reduced the margin of safety of the plant's systems. ,
This material substitution did not constitute an unreviewed safety question as defined in 10CFR50.59.
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Annual Summary of JAPNPP Ch nges, Tcsts, and Experiments for 1988 b
JAF-SE-88-079, Rev. 1- Motor Operated Valve (MOV) Torque Switch Bypass and Close Switches Setpoint ;
Change (Plant Modification F1-88-106)
This safety evaluation provided the safety analysis for Plant i Modification F1-88-106. This safety-related modification au- '
thorized the revising of the torque switch bypass setpoint from f 10% to 33% for all Limitorque MOVs so as to meet the intent of ' '
NRC IE Bulletin 85-03 regarding the recalculated actuator thrust -
values. In addition, this modification authorized the revising ;
of the MOV torque switch bypass setpoint for all Limitorque MOVs so as to ensure valve opening under design. differential pressure ficw conditions during normal and abnormal events within the:
plant design bases. The rewiring of MOV valve closed position .
lights to a separate' limit switch rotor on the MOV in order to maintain consistent valve position indication was also au-thorized, o
This modification was determined to not have a negative impact on the FSAR, Technical Specifications, MOV seismic qualifications, MOV environmental qualifications, or any safety-related systems ,
or components.
This modification enhances the valve's ability to perform their intended safety function and it was determined that an unreviewed safety question as defined in 10CFR50.59 does not exist.
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Annual Summary cf-JAFNPP Ch:nges, Tcsts, and Experiments for 1988 i i
JAF-SE-88-081 - Emergency Diesel Generators (EDGs) Lube Oil Modification (Plant Modification F1-87-137) :
I s
The purpose of this safety evaluation was to provide the safety review of Modification F1-87-137 which upgraded the Lube Oil System for the safety-related EDGs. This vendor recommended modification consisted of (1) providing additional 3 gpm pumps to ensure constant oil supply to the turbochargers, (2) adding 25 ,
psig relief check valves to relieve excess pressure when the engines are running, (3) relocating the existing relief valves to prevent backflow when the engines are operating, (4) installing new pressure switches for pump indication, (5; replacing the alarm pressure switches, and (6) adding sight glasses to verify oil level in the oil cooler.
The restart restrictions previously required due to lube oil considerations were eliminated by this modification. INPO Significant Operating Exaerience Report (SOER) 83-1 commitments were also satisfied by the completion of this modification.
This modification did not introduce any additional abnormal engine. conditions, therefore the existing Control Room and local lube oil alarms were not altered. The availability for each EDG was enhanced by these changes and the margin of sa~fety>ns defined in the bases for the Technical Specifications was not reduced.
No change to the FSAR or to the Technical Specifications was required as a result of F1-87-137.
1 The conclusion of this safety review was that an unreviewed safety question as defined in 10CFR50.59 was not involved.
The preoperational test (POT-93D) associated with this modifica-tion and the installation of the cable and raceway were analyzed l in Nuclear Safety Evaluations JAF-SE-88-146 and JAF-SE-88-145,
- respectively. .
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Page -114-m.
.s, Annu21 Summ ry of JAFNFP Chenges, Tosts,
- and Experiments for 1988 JAF-SE-88-082 -
~~ Elimination of Interference Between Feedwater ;
Pipe Break Restraint and Main Steam Safety '
Relief Valve (02RV-710) (Minor Modification M1-88-017)
This safety evaluation addressed the trimming of Feedwater Pipe Break Restraint FW-15 to eliminate an interference which had prohibited the removal of the Pilot Valve of the safety-related Main Steam Relief Valve 02RV-71C during maintenance activities.
- The limited removal of base metal from this restraint was an- ,
alyzed by the Stone and Webster Engineering Corporation and was- '
determined to not affect the restraint's structural integrity.
The trimming of this restraint did not adversely affect any safety-related equipment nor did it change any system in a manner which altered the design bases or accident ana,yses.
l The minor structural changes performed under this modification did not constitute an unreviewed safety question.
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E' Annual Summ2ry of JAFNPP Ch nges, Tcsts, L and Experiments fo,r 1988 JAF-SE-88-083 - Standby Liquid Control (SLC) System Squib Continuity Circuit Upgrade (Plant Modifica-tion F1487-155) !
Modification F1-87-155 was performed in order to ensure that the loss of continuity' circuits for the SLC squib valves function as designed.. The circuits were altered from A.C.,to D.C., the meter relays were moved to the ground side of the continuity circuits, resistors were replaced, filter capacitors and surge capacitors were installed to ensure the correct operation of these circuits.
This safety-related modification, which was reviewed and found
- acceptable by General Electric, increases the reliability of this circuitry.
l-The FSAR states the loss of continuity circuit will alarm if either of the squib circuits open up. This modification did not alter this required function. Failure of an individual component or the continuity circuit as a whole will not result in a config-uration that will degrade the SLC System such that the squib valves will not fire and the pumps will not operate when required. This modification upgraded the SLC System by giving better indication of squib primer continuity and this circuitry does not affect any margin of safety.
This upgrade of the squib monitoring circuits did not constitute l an unreviewed safety question.
l The preoperational test-(POT-11G) associated with this modifica-tion was reviewed in Nuclear Safety Evaluation JAF-SE-88-162 l
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, Annual Summary of JAFNPP Ch:ngas, Tests, ;
and Experiments for 1988 JAF-SE-88-084, Rev. 1- Disc Material Change for Main Steam Isolation Valves (MSIVs) (Minor Modi-
..f fication M1-88-088)
This safety evaluation provided the review for safety-related Minor Modification Mi-88-088. This modification authorized the use of replacement discs made from a different material than the originally supplied MSIV discs. The new material was evaluated 1 for the desired application and it was determined to be accept-able.
This evaluation detailed the material change for these valve discs and the effect that this change had on plant operation and procedures. This change did not degrade the ability of the MSIVs to perform accident mitigating functions and the margin of safety as defined in the bases for the Technical Specifications was not reduced.
This material substitution did not constitute an unrevieked safety question. ,
Page -117-r
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Annual Summary of JAFNPP Ch;ngos, Tosts, ;
and Experiments for 1988 r
JAF-SE-88-085 - Temporary Storage of Packaged Radioactive Material in the Condensate Storage Tank (CST) ;
Area ;
This safety evaluation reviewed the temporary storage of packaged radioactive material pending shipment from the site. The materi-al, which will be in strong Low Specific Activity (LSA) tight boxes -is to be kept in the CST area. This temporary storage is desirable during periods of heavy work activity when such pack- '
aged radioactive material may interfere with work spaces in the plant. Various administrative controls outlined in this document i ensures that the requirements of NRC Generic Letter 81-38,
" Storage of Low-Level Radioactive Wastes at Power Reactor Sites". '
The various regulatory requirements for this storage were ad-dressed in this evaluation and compliance with these was assured by ensuring compliance with the plant's Technical Specifications.
Radiological control was also addressed which included area posting and surveillance requirements.
This evaluation determined that an unreviewed safety question pursuant to 10CFR50.59 does not exist as a result of this non-safety related activity.
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Annual Summary of JAFNPP Chtngas, Tosts, and Experiments for 1988 i
l JAF-SE-88-086 - Exciter Brush Rigging Support (Minor Modi-fication M1-88-092)
Minor Modification M1-88-092 installed a support between the Exciter Brush Rigging and the Turbine Deck in order to reduce brush vibration. This non-safety related modification was implemented to improve brush performance and prevent sparking which degrades the surface of the slip rings.
The Technical Specifications and FSAR were reviewed and it was determined that this change had no impact on either. The failure of this rigging support was addressed an the possible effects-were found to be inconsequential to the operation of the plant.
This support addition did not constitute an unreviewed safety question.
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Annusi Summ ry of JAFNPP Ch ng:s, Tests, and Experiments for 1988 JAF-SE-88-087 - Velan Valve Spare Part Material Substitution (Minor Modification M1-88-020)
Minor Modification M1-88-020 allowed for the use nf vendor recommended replacement parts for various valves 'fety-related and non-safety related). The material substitutic s were an-alyzed-against the requirements applicable to construction of the original valves and were determined to be acceptable for the intended use and did not affect form, fit, or function.
This evaluation detailed the various part changes and provided justification for the acceptance of each. The material changes addressed were determined to not reduce the margin of cafety as defined in the bases for the Technical Specifications as the ability of the valves to perform their design function was not affected.
This material substitution did not involve an unreviewed safety question.
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Annusi Summ2ry of JAFNPP Chcngos, Tosts, and Experiments for 1988 ,
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,' JAF-SE-88-088 - Velan Valve Spare Part Material Substitution:
Mark IVGW-ISA, ISB Carbon Steel Gate Valve *
(Minor Modification M1-88-066)
This safety evaluation provided the technical review for Minor Modification M1-88-066. This modification authorized the use of various vendor recommended realacement valve parts which are made of materials different from tL1ose originally supplied. The '
chemical and mechanical properties of the replacement materials (for both safety-related and non-safety related valves) were analvred and it was determined that the substitutions were accehtable for the design applications.
This evaluation detailed the various part changes and provided justification for the acceptanco of each. The material changes addressed were determined to not degrade the ability of the affected valves to perform their required design functions.
The material substitution evaluated by this safety evaluation did not constitute an unreviewed safety question.
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t Annual Summary of JAFNPP Chengas, Tosts, and Experiments for 1988 t
JAF-SE-88-089, Rev. 1 - Pacific Valve Spare Part Material ,
Substitution: V0W-15N, V0W-15A, and VGW-15A (Minor Modification M1-88-110)
This safety evaluation provided the technical review for Minor Modification M1-88-110. This modification authorized the use of several replacement parts for Pacific valves (safety-related and non-safety related) which are no longer manufactured with the e me.terial originally supplied. The chemical and mechanical ;
properties associated with the replacement metertals were an-alyzed and it was determined that these substitutious were acceptable for-the design applications. The new spare parts meet and/or exceed the original procurement specifications for the valves affected.
The evaluation included a detailed evalurtion of the various part changes and arovided the justification for the acceptance of each. The changes addressed were determined to not reduce the margin of safety as defined in the bases for the Technical .
S aecifications as the ability of the affected valves to perform their design functions was not altered.
The conclusion of this safety evaluation was that an unreviewed safety question did not exist as a result of this material substitution.
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Annu21 Summ2ry of JAFNPP Ch;ngas, Tosts, and Experiments for 1988 JAF-SE-88-091, Rev. 2 - Replacement Materials for Atwood &
Morrill Check Valve Subcomponents (Minor Modification M1-88-109)
This safety evaluation provided the technical review for Minor Modification M1-88-109. This modification authorized the use of several vendor recommended replacement valve parts which are made of materials different from those originally supplied. The chemical and mechanical properties of these replacement materials (for safety-related valves 30A0V-68A&B, 14A0V-13A&B, and 27VB-6&7) were analysed and it was determined that the substi-tutions were acceptable for the design applications.
These material substitutions were specified in detail in this evaluation and justification was given for the acceptance of each. Implementation of these substitutions did not involve a change to the Technical Specifications or the FSAR. Only re-placement of existing parts are involved and system operations -
are not affected.
These material substitutions evaluated by this safety evaluation did not constitute an unreviewed safety question pursuant to 10CFR50.59.
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L-Annual Summary of JAFNPP Chtngos, Tcsts, and Experiments for 1988 JAF-SE-88-092, Rev. 1- Realacement of Tube Side Relief Valve on 5tLi Point Feedwater Heater 35RV-115A&B (Minor Modification M1-87-107)
Minor Modification M1-87-107 replaced the fifth point feedwater heater tube side relief valves due to frequent repair and leakage problems. The new. valves have an 0-ring seat seal which is more tolerant piping vibration and stress than the original metal to >
metal seat.
A comparison of the replecement valve an.t the original valve was presented in this evalcation. For the design application it was determined that the new relief valve would provide increased reliabilitf and reduced maintenance. This modification did not -
involve a change in the Technical Specifications nor did it reduce any margin of safety as defined in the basis for the t Technical Specifications.
The conclusion of this safety evaluation was that an unreviewed j safety question as defined in 10CFR50.59 did not exist as a
, result of this non-safety related minor modification.
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-Annual Summ;ry of JAFNPP Chcngos, Tosts, and Experiments for 1988 JAF-SE-88-094 - Removal of Empty Waste Disposal Liners from the Spent Fuel Pool This safety evaluation addressed the concerns associated with the removal of three empty waste disposal liners from the Spent Fuel Pool. A detailed review of.the lifting procedures and hardware used was provided, and it was determined that a minimum safety factor of 10:1 would be maintained. As a result of this eval-uation, it was established that the requirements of NUREG-0612,
" Control of Heavy Loads at Nuclear Power Plants", were met.
Since safa load paths were utilized and NOREG-0612 was complied with the probability of occurrence or consequentes of an accident evaluated in the FSAR or other safety analysis reports were not increased. The removal of these liners was acrformed in accor- i dance with plattt procedures and guidelines which ensured that the !
plant's margin of safety was not compromised.
It was concluded that this non-safety related activity did not constitute an unreviewed safety question as defined by 10CFR50.59.
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Annual Summ ry of JAFNPP Chengos, Tosts, and Experiments for 1988 r
JAF-SE-88-098, Rev. 2 - Chemical Decontamination of Reactor '
Water Recirculation (RWR) System and Portions of Reactor Vessel, Residual Heat Removal (RHR) and Reactor Water Clean-Up (RWCU) Systems (Temporary Maintenance Procedure MPT-37)
This safety evaluation addrested the non-safety related chemical r decontamination of the RWR System as well as portions of the RHR to RHR Service Water cross-tie piping, the RHR to Fuel Pool !
Cooling (FPC) cross-tie piping, and the reactor vessel annulus region below the slip joints of the jet pumps. This decontamina-tion was indicated due to increased radiation levels in the Drywell which were in large part due to the activated corrosion products in the oxide layer inside the RWR and adjacent system's piping. The process utilized removed this oxide layer without ,
any detrimental effects on the components and piping systems.
- A detailed review of the activities required to implement this decontamination was presented in this safety evaluation.- Special attention was given to flow paths, boundaries, decontamination equipment, leakage precautions, monitoring of changing radio-logical conditions, and control of waste products. This thorough review established that an unreviewed safety cuestion pursuant to 10CFR50.59 did not exist as a result of this c.econtamination effort.
Safety analyses concerning the installation of the necessary system connections and floor loading concerns were contained in Nuclear Safety Evaluations JAF-SE-88-064 and JAF-SE-88-144, Rev.
1.
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-Annual Summary of JAFNPP Ch:ngas, Tosts, and Experiments for 1988 JAF-SE-88-102 - Replacement of Valve Manifold with Separate Whitey Valves for 23LT-201C (Minor Modifica-tion M1-88-105)
Modification M1-88-105 replaced the five-valve manifold installed on the high and low pressure connecting tubing for the High Pressure Coolant Injection (HPCI) suppression pool level trans-mitter 23LT-201C with four separate instrument shutoff valves.
Thic non-safety related modification eliminated the bypass capability for equalizing pressure since this instrument can be calibrated and tested without having to equalize the pressure.
All new materials and components installed per this modification were installed and inspected in acccrdance with the applicable plant design and installation criteria. Weld joints were dye-penetrant examined and the completed installation was leak-tested in accordance with the applicable design code (ANSI E31.1, 1967).
Implementation of this modification did not in any way affect the existing design bases and accident analyses.
It was determined in this safety evaluation that the implementa-tion of Minor Modification M1-88-105 did not constitute an unreviewed safety question.
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f Annuc1 Summary of JAFNPP Chcngos, Tosts, and Experiments for 1988 JAF-SE-88-105 - Exciter Coupling Replacement (Minor Modifica-tion M1-88-124)
Minor Modification M1-88-124 replaced the coupling between the generator and the exciter with one of improved design. The old coupling had the potential to bind and cause high vibration at the exciter bearings. The installation of the new coupling eliminated these problems.
This evaluation determined that there was no change required to the FSAR as-a resolt of the implementation of this modificttion.
Also, no revisions were required to the plant's Technical Speci- i fications.
This non safety related coupling replacement did not constitute an unreviewed safety question as defined in 10CFR50.59, i
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, Annual Summary of JAFNPP Ch:ng:.s, Tcsts,
! and Experiments for 1988 s.
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JAF-SE-88-112 - Installation of DC Milliammeter in Main Generator Field Ground Circuit (Plant Modi-fication F1-88-087)
The purpose of this safety evaluation was to provide the safety review of Modification F1-88-087. This non-safety related modification installed a de milliammeter in the main generator i field ground circuit. This was performed in order to provide Operations personnel information concerning the buildup of cupric oxide on the insulated cooling water heat exchanger counections which is indicated by leakage current. This will enable omera-tors to take the necessary correcrive actions to protect the main generator from initiating unnecessary turbine trips and resultitig reactor trips due to a cupric oxide buildup.
The evaluation determined that this modification enhanced the
- main generator system ar.d plant availability. The FSAR and Technical Specifications were reviewed and no changes were required due to the implementation of this eadification. No -
systems were adversely affected by this modificaticn and the seismic qualification of the panel in which this milliammecer was added was not compromised.
This modification enhanced plant safety and did not constitute an unreviewed safety question pursuant to 10CFR50.59.
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l L JAF-SE-88-114 - Emergency Diesel Generator (EDG) Speed Switch and Crankcase Vacuum Monitoring Upgrade (Plant Modification F1-87-079)
Modification F1-87-079 seismically installed NEMA Type 12 enclo-sures to the floor of the EDG Rooms. New upgraded syncho-start speed switches were mounted inside these enclosures along with new crankcase vacuum gages.
The new speed switches provide increased reliability and a qualified repair program. Seismic qualification of the speed switch was performed by en independent laboratory. The new vacuum gages were installed in order to '3rovide for engine 4
condition trending. It was determined t' Tat a break in the line or gauge that allowa air to bleed into the crankcase vill not degrade the vacuum in the crankcase to the poir.t where the engine would be disabled.
There was no change in the safety design bases of the FSAR and -
the margin of safety as defined in the Technical Specifications was not reduced.
This safety evaluation determined that this improvement to the safety-related EDG Control Syst. ems will reduce the probability and consequences of an accident or safety equipment malfunction.
The conclusion of the safety evaluation was that this improvement did not involve an unreviewed safety question.
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F l i AnnuS1 Summ2ry of JAFNPP Changos, Tests, and Experiments for 1988 f JAF-SE-88-116 - Reactor Water Cleanup (RWCU) Precoat Evaluation Program - Test No. 5 r This safety evaluation was written to evaluate.the change in scope of the fifth and final test in the RWCU Precoat Evaluation Test Program which was analyzed in Nuclear Safety Evaluation JAF-SE-87-137, Rev. 1. The premix precoat material used in this test presented the most potential for resin bleedthrough.
Special precautions regarding the execution of this test were outlined in this review in order to minimize this potential.
This evaluation discussed the various operational aspects of this cost as well as the potential impact on plant parameters. It was '
determined that the design bases for the RWCU System stated in the FSAR usa taot compromised by the use of the 9:1 resin / fiber ,
mix in this test program.
It was concluded that this non-safety related test did not involve any unreviewed safety concerns as defined by 10CFR50.59.
This conclusion was based on the comprehensive review of the precoat materials and the temporary operating procedure followed.
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Annual Summary of'JAFNPP ChCngos, Tosts, and Experiments for 1988 JAF-SE-88-117 - Reactor Building Emergency Ventilation Damper Position Indication for Reg. Guide 1.97, Rev.
2 (Plant Modification F1-88-136)
Modification F1-88-136 mrovided environmentally qualified open-close status for the Reactor Building's emergency ventila-tion dampers 66A0V-100A,B and 66A0V-101A,B. The Emergency Plant and Information Computer System (EPIC) connections at Panel 09-75 were upgraded fona the unqualified damger control circuitry to provide new environmentally qualified cable and power supplies (71ACA2 and 71ACB2) for EPIC status indication directly from the position switches. This safety-related modification was per-formed in order to annure that damper position indication will be available to the operator in the event of an accident.
The position switches and egbling used are environmentally qualified and the cabling is routed through a seismically sup-ported raceway system. Any single short circuit would not eliminate both trains of position indication. This modification did not make any significant changes in the existing fail-safe circuitry and EPIC points were upgtaded to ensure indicatien will be avrilable in the event of an accident. Human ft.ctors -
considerations were addressed in this evaluation.
This modification did not involve any system design change which could cause an accident or malfunction other than previously evaluated in the FSAR.
L It was the conclusion of this safety evaluation that this modi-fication enhanced plant safety and did not constitute an unre-viewed safety question.
i The safety analysis for the preoperational test (POT-66E) associ-ated with this modification was contained in Nuclear Safety h- Evaluation JAF-SE-88-118.
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Annual Summary of JAFNPP Chcngss, Tosts, ;
.and Experiments for 1988 '
JAF-SE-88-118 - Reactor Building Emergency Ventilation Damper Position Indication for Reg. Guide 1.97, Rev.
2 (Preoperational Test POT-66E) 1 The purpose of Preoperational Test POT-66E was to verify damper o' operability and position indication for the emergency ventilation dampers modified under safety-related Plant Modification F1-88-136. A safety analysis for this modification was contained in Nuclear Safety Evaluation JAF-SE-88-117. Each affected damper was stroked and correct damper position indication was verified by comparison at the aparopriate status location, i.e. the indicating lights and the inputs to the Emergency Plant and Information Computer System.
Plant process parameters remained unaffected by this preopera-tional test and no change was required to the Technical Speci-fications.
It was-determined that no single error during the implementation of this test would have resulted in the failure of safety-related components of more than one train as is required by the nafety design baser of the FSAR, Section S.
The conclusion of this safety evaluation was that the execution of this test did not involve an unreviewed safety quertion.
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+ Annual Summary of JAFNPP Chengas, Tcsts, and Experiments for 1988 1
JAF-SE-88-129 - Byron Jackson Bonnet Seal Material Substi-tution for 27MOV-122 and 123 (Minor Modifica-tion M1-87-096)
The purpose of this safety evaluation was to provide a technical review of the bonnet seal material change for the safety-related valves 27MOV-122 and 27MOV-123. This manufacturer recommended substitution (ASTM A743 Gr. CF3M for A276 Type 316) was de-termined to be acceptable based on the similarities in material strengths, composition, and ductility.
It was determined that this bonnet seal material change would not degrade..the ability of the valves to perform their design functions and that this change did not adversely affect any margin of safety as defined in the Technical Specifications.
The conclusion of thir safe:y evaluation was that this substi- ,
tution did not involve an unreviewed safety question.
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Annual Summ3ry of JAFNPP Chcng3s, Tests, and Experiments for 1988 JAF-SE-88-132 -
-~
Feedwater Flow Transmitter (06FT-50A,B) .
Replacements (Minor Modification M1-87-119) ,
i Minor Modification M1-87-119 involved the direct replacement of Rosemount differential pressure transmitters for the previously installed Foxboro models. This non-rafety related replacement was performed in order to increase the component's reliability and to eliminate the need for excessive calibrations which were required with the Foxboro models. ,
The change in loop loading due to this substitution is negligi-ble. Oniv minor changes in the instrument lines were made to account for the configuration differences between the new trans-mitters and the old. This modification did not change the safety design bases as stated in the FSAR and did not reduce any margin of safety.
The cenclusion of this safety evaluation was that this direct substitution of more reliable units did not constitute an unre-viewed safety question.
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t e 4 Annual Summary of JAFNPP Chtngos, Tosts, and Experiments for 1988 JAF-SE-88-137 - Upgrade of Condensate Booster Pump (33P-9A,B,C) Lube Oil System to Reduce Leaks (Plant Modifi cation F1-87-161) the purpose of Modification F1-87-161 was to reduce the oil leaks in the condensate booster pump's lube oil system. This consisted :
of (1) replacing the lube oil supply header with flexible hose, '
(2) adding piae sealant on all threaded plugs and fittings, (3) replacing leaiing gaskets, and (4) modifying the coupling guard to contain oil in leakage collection area.
This modification did not change the existing design bases for this system, but instead eliminated the personnel safety hazard associated with lube oil leaking on the floor in the vicinity of the condensate booster pumps.
This non-safety related modification did t.ot affect any equipment important to s'afety and did not involve an unreviewed safety question. ,
The preoperational test (POT-33B) associated with this modifica-tion was reviewed in Nuclear Safety Evaluation JAF-SE-88-157.
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AnnuS1 Summ ry of JAFNPP Chengos, Tests, i and Experiments for 1988 JAF-SE-88-138 - Operation of the Zine Injection Passivation i System (GEZIP) (Plant Modification F1-88-048) l This safety evaluation provided the safety analysis for the implementation of the non-safety related GEZIP process (General Electric Zine Injection Passivation). This process was develooed to inhibit the corrosion of stainless steel thereby reducing tNe buildup of cobalt-60 on the reactor coolant pressure boundary piping systems. The potential benefit lof zine injection is the '
reduction in radiation levels by a factor of three or more.
The process consists of injecting controlled concentrations of zine oxide into the suction side of the feedwater pumps whenever the reactor is at rated temperature and there is feedwater flow to the reactor. Process controls have been installed to accu-rately monitor the water chemistry and adjust the zine concen-trations accordingly. Periodic cleaning of the injection tank will prevent buildup of solid zine oxide on the tank sides which can flake off and cause clogging problems. The only significant ,
activation product of natural zine is 2n-65 which has a short half-life and a less er.ergetic decay product than the Co-60 isotope whose production GEZIP was designed to inhibit.
The presence of rinc on reactor materials was determined to not be detrimental to the affected structures or components. This ;
process will significantly reduce personnel exposure during maintenance activities and has no effect on reactor characteris-tics or operating variables. The safety evaluacion determined that the implementation of the GEZIP system did not constitute an unreviewed safety question.
The preoperational tests associated with this modification were analyzed in Nuclear Safety Evaluation JAF-SE-88-160 and JAF-SE-88-161. The installation of the hardware for this modi-fication was analyzed in JAF-SE-88-173, Rev. 1.
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-Annual Summ0ry of JAFNPP Ch ngos, Tosts, and Experiments for 1988 JAF-SE-88-139 - 12-4CFD-17A,B Manual Valve Replacement (Minor Modification M1-87-067)
" This non-safety related modification consisted of the direct replacement of two single dise pressure seal gate valves. The new valves eliminated the previous pressure seal leakage problem associated with the original valves.
The replacement of these non-safety related clean-up filter demineralizer effluent isolation valves with functionally identi-cal valves did not change the flow or pressure drop characteris-tics of the system. The new valves weigh less than the ones previously installed so no change in the pipe supports was necessary.- This modification did not change the RWCU design bases and had no effect on any accident analyses.
It was the conclusion of this safety evaluation that the perfor-mance of this modification did not constitute an unreviewed safety question.
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Annual Summary of JAFNPP Ch:ngos, Tosts, and Experiments for 1988 JAF-SE-88-141, Rev. 1- Drywell Personnel Airlock (16X-2A)
Upgrade (Plant Modification F1-26-108)
This safety evaluation provided the safety analysis for Plant Modification F1-86-108. The aurpose of this safety-related modification was to upgrade t'ae Drywell Personnel Airlock in ,
order to increase the reliability and maintainability of the door drive mechanism, door seals, and shaft seals. All drive mecha-nism assembly boxes', associated hardware, shaft seals and door gaskets were replaced. The interior door strongback tie bolt assemblies for the airlock were also replaced due to their degraded condition.
The detailed review presented in this safety evaluation of the design improvements included the effects of weight changes of components, the acceptability of material upgrades, the heavy load concerns and the leak testing requirements. The ability of 4 the Drywell personnel airlock to perform its design basis func-tion was not adversely affected by this modification.
The conclusion of this evaluation was that an unreviewed safety question as defined in 10CFR50.59 did not exist as a result of the implementation of this modification.
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Annual Summ2ry of JAFNPP Chcngos, Tosts, -
and Experiments for 1988 i
JAF-SE-88-142 - Replacement of Moisture Separator Reheater (MSR) Drain Tank Bypass Valves (31LCV-118A&B,
-120A&B, and -122A&B) (Plant Modification F1-86-084)
This safety evaluation addressed the safety concerns associated with the implementation of Modification F1-86-084. This ;
non-safetr related modification replaced the MSR drain tank bypass va.tves in order to increase the plant's power generating '
efficiency, Also, the carbon steel piping downstream of these valves was replaced with a chrome-moly alloy to provide increased erosion resistance.
The detailed review of the replacement valves provided in this evaluation addressed the resizing of these valvas to reflect system flow conditions, the reduction of valve leakage, the effects of the valves' weight changes on the piping system, and ,
the improved erosion resistant piping. This modification did not reduce the margin of safety as defined in the bases for the Technical Specifications because this modification replaced existing valves and piping with an improved design for the application specified.
It was concluded that this modificction did not constitute an unreviewed safety question as defined in ICCFR$0.59.
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. Annual Summtry of JAFNPP Chengr.s, Tosts,
/J g Experiments for 1988 E ,
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- xJAF-SE-88-144, Rev.'1'-
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Standby Gas Treatment Room Floor Loading Due to Placement of. Temporary Chemical- ,
1- ,
. Decontamination Equipment i
. This1 safety evaluation addressed the utilization of the Ultimate
,7 Strength Design (USD) method of the American' Concrete Institute's
-ACI-349-1985 code in lieu of-the Working Stress' Design Method (WSD) of'ACI-318-1963 as specified in the FSAR. The USD method is currently the predominant design' method used for nu~ clear safety-related structures. Utilization of this design method ,
allowed.for the1 temporary placement of chemical decontamination ;
equipment.on the Standby Gas Treatment Room floor.
,1 A, I .'byffL The revised method of floor loading evaluation was.found to meet w, current applicable industry and; regulatory criteria, cThe margin 5 '
of safety as defined in the Technical Specifications was not reduced because-no safety margins were affected by this. change in design code. '
It was'the determination of this safety evaluation that uti-lization of this method-of stress analysis for the temporaryL installation;of the necessary chemical decontamination equipment
. . did not constitute an unreviewed' safety question, r
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-JAF-SE-88-145 - Emergency Diesel Generator (EDG) Lube Oil Modification Raceway and Cable Installations
! (Plant Modification F1-87-137) '
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This-safety evaluation reviewed the activities associated with.
L , the.fnstallation of cables, conduit, and seismic supports for' l
-cafety-related_ Plant Modification F1-87-137. The~ safety analysis L for the entire' lube oil modification was' contained in-
! JAF-SE-88-081, y
V' The implementation of this partial modification was accomplished h during normal plant operation in order to' reduce the amount of work required to be performed during the outage. Safety issues addressed in this review included (1) structural integrity of the
< raceway and conduit supports, (2) fire barrier concerns associ-F ated with the opening of existing wall sleeves, and (3) the l
u loading of the conduittand raceways.
Since this partial modification did not tie.into any safe- ,a g .
ty-related equipment and adequate precautions were implemented to tg < prevent any impact on adjacent equipment, this: safety evaluation concluded that an unreviewed safety question did not exist, 1
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Annual Summary:of JAFNPP Chengos, T3sts, f
o)4 and Experiments for 1988 c.
- JAF-SE-83-l'46 - Emergency-Diesel Generator (EDG) Lube.011 Modification Test (Preoperational Test '
, POT-93D)
The purpose of Preoperational Test POT-93D was to verify that the .
vendor recommended lube oil modification functioned as designed.
A safety analysis of this safety-related modification (F1-87-137) was presented in Nuclear Safety Evaluation JAF-SE-88-081.
To ensure proper system operation, this test was performed with the engine control. switch in the " maintenance" mode for two, n conditions: -first, with both lube oil pum heater. turned en and the engine shut down,psand andsecond, the immersion, with the.
engine at rated speed with no load and both pumps on. The conduct of this test verified that1the gallery suaply line is sized properly. This test was conducted within the parameters provided in the Technical Specifications. Therefore, the~ safety margins were not reduced.
It was determined that the conduct of this test did not adversely affect any associated safety-related equipment and did not constitute an unreviewed safety question pursuant to 10CFR50.59.
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Annual Summ ry of JAFNPP Chengas, Tests, !
nd Experiments for 1988 l
- JAF-SE-88-147,~Rev. 1 - Material Substitution to Byron Jackson Pump Parts-(10P-3A,B,C,D and 14P-1A,B) r (Minor Modification M1-88-112) 1'
-This modification, M1-88-112,- involved the material substitution l by Byron Jackson Pump Division of the shaft keys and the impeller '
nuts on the safety-related Core Spray Pumps-14P-1A,B,C,D and
- Residual Heat Removal Pumps 10P-3A,B. The chemical and mechan- j
'ical' properties of the replacement materials were analyzed and-it was determined that the substitutions were acceptable for the !
' design applications.
. This evaluation contained a review of each of the part material changes and provided the justification for the acceptability of these substitutions. It was determined that these material changes did not affect pump operability, and therefore did not I affect existing system integrity or operability.
An unreviewed safety question did not exist as a result of this l material substitution. '
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Annual Summsry of'JAFNPP Chtnges, Tcsts, and Experiments for 1988-JAF-SE-88-152 - Reactor Core Isolation Cooling (RCIC) Gover-nor Lube Oil Reservoir Support Clamp Instal-11ation (Minor Modification M1-88-163) s ' Modification M1-88-163- installed support clamps for the RCIC turbine governor _ lube oil sump in order to ensure the capability p' of the turbine:to; function as intended during a design basis event. This charige ensured that the sump will be adequately supported to withstand the applicable-seismic and operational forces.
This evaluation' discussed the' original installation of the oil-sump as was recommended in General Electric's Field Disposition Instruction FDI 83/88595 and in Terry Corporation's Design Improvement DI-6. The type of clamps originally installed as L part of the implementation of this change exhibited poor perfor- !
I mance in other plant applications. It was decided to replace i J.
these clamps with a new support configuration. j This configuration consisted of new stainless steel tie-wraps attached to a 1/4-inch thick support bracket on the RCIC turbine. !
This non-safety related modification improved the reliability of !
the RCIC turbine and was determined to not involve an unreviewed '
! safety question.
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Annucl Summcry of JAFNPP'Chenges, Tests, and Experiments for 1988 JAF-SE-88-153, Rev. 1- Setpoint Change for 72TS-122A&B -
Battery Room Low Temperature Alarm (Minor Modification M1-88-161)
The setpoints for temperature switches 72TS-122A&B were raised from 45 degrees to 60 degrees Fahrenheit by Modification .
-M1-88-161. This safety-related modification was implemented in order to alert operators of the fact that the batteries need to be monitored on a more frequent basis when the temperature drops below 60 degrees. This is due to the fact that battery tempera-tures lower than 60 degrees results in decreased battery capaci-ty.
The incorporation of this modification did not adversely' affect any safety-related systems and did not alter the existing design bases or accident analysis. -
l This safety evaluation determined that this modification enhanced
, battery performance and did not involve an unreviewed safety '
L question pursuant-to 10CFR50.59.
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Annucl'Summtry of JAFNPP Chengas, Tests, and Experiments for 1988 JAF-SE-88-154 - Velan Valve Gasket Material Substitution (Minor Modification M1-88-122) 4 This safety evaluation provided the technical review for Minor !
Modification M1-88-122. This modification authorized.the use of 1
- the new pressure seal gaskets for safety-related_ valve 23MOV-21 and non-safety related valves 12MOV-68, 12MOV-74, 13MOV-30, and i 29MOV-79. The replacement gaskets were furnished in a material different from the originally supplied gaskets. The chemical and ;
mechanical properties.of the replacement material was analyzed !
and it was determined that the substitution was acceptable for !
the part's design application. 1 No' activation droblems will exist with the substitution of this material and the ability of these valves to perform their design function was not compromised as a result of this modification..
This gasket material substitution did not involve an unreviewed i safety question, i
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i 1 ' e. j Annuzl Summary of JAFNPP Changes, Tests, j and Experiments for 1988 l
- JAF-SE-88-155 - Outer Bellows Seal Drain Lines L 19-4"-W19-151-61A,B,C,D and 19-2"-W19-151-52A&B Modification (Minor. ,
Modification M1-88-164)
Minor Modification M1-88-164 consisted of the installation of high strength pipe fittings on the outer bellows seal rupture 1
' lines and liner drain-lines . This non-safety related modifica- ,
tion was implemented in order to allow access to the piping _for inspection to ensure that the. lines are free of obstruction and are operable.as designed.- This was done in-response to NRC j l Generic Letter 87-05, " Request for Additional Information - '
Assessment of Licensee Measures to Mitigate and/or Identify Potential. Degradation of Mark I Drywells". ,
1 Precautions were outlined in this evaluation to ensure the safe !
installation of these pipe fittings. The ability of the drain I lines to perform their normal-drainage function was not degraded-by this modification since the' original piping design criteria of i L AP-23 was-utilized. No changes were required to the Technical l Specifications and the margin of safety as defined in the bases I of the Technical Specifications was not reduced.
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- and did not involve an~unreviewed safety question.
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Annual Summtry of JAFNPP Changss, Tests, and Experiments for 1988 T
JAF-SE-88-156 -- 19FPC-31 Manual Valve Replacement (Minor- ,
-Modification M1-88-165)L i Minor Modification M1-88-165 replaced the Residual Heat Removali (RHR)_ Assist to Fuel Pool Cooling System discharge valve-19FPC-31 with a new globe valve and added a cavitation control plate. The installation of this valve and plate configuration was intended to significantly increase-the life of 19FPC-31 in its-service conditions. The newly-installed valve was procured'to specifica-
-tions which were determined to meet or exceed the original-procurement requirements.
Analysis of flow conditions and operating experience demonstrated that cavitation was responsible for the loss of.19FPC-31 seat to disc-tightness. -Calculations were referenced in this evaluation which determined that this new valve and plate configuration would not change the RHR assist to fuel pool i cooling flow rate.
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There was no significant change in piping or support loads because of this_ modification and the design bases for the system was not adversely affected. .
This non-safety related modification increased the re11 ability of.
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valve-19FPC-31 and-it was determined in this safety evaluation that the implementation of this modification did not constitute
.an unreviewed safety question.
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. 1 Annus1 Summiry.of JAFNPP Changas, Tests, and Experiments for 1988- *
~~ Upgrade of Condensate Booster Pump (33P-9A,B,C)-Lube Oil. System to Reduce Oil Leaks (Preoperational test POT-33B)
.The purpose of Preoperational Test POT-33B was to verify that the !
non-safety related Modification F1-87-161~ to the condensate booster pump's lube oil system did not adversely affect the '
operation of the pumps. A safety analysis for F1-87-161 was -
. presented'in Nuclear Safety Evaluation JAF-SE-88-137.
Specifically, this test verified that (1) the lube oil pressure was within the allowed operating : range, (2) the flow to the pumps
'and drive motor. bearings was sufficient to prevent overheating, (3) the auxiliary oil pump and system relief valves were set at L the correct pressure, and (4) the pressure switch logic was correct.
The majority of this test was performed ~with the plant shutdown
) and the pumps out of service. The remaining portions of the test involved only monitoring the pump operating parameters using existing equipment. Neither part of'this areoperational test t L affected the plant design bases, reduced the margin of-safety as L defined in the Technical Specifications, created the possibility c L of_an accident other than those evaluated previously in the FSAR or involved an unreviewed safetyLquestion.
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Annusi Summary of JAFNPP Changos, Tests, and Experiments for 1988-JAF-SE-88-158 - . Replacement of Control Rod Blades (Minor
' Modification M1-88-167)
This safety evaluation provided the safety analysis for safe-ty-related Minor Modification M1-88-167 which replaced 25 control rods which had reached the end of their service life. The new control rods incorporated minor dimensional changes and changes to poison geometry. The replacement control rods are in all aspects functionally equivalent to those that were replaced. In addition to these replacements, this modification also relocated four control-rods within the core in order to reduce the number of replacements required at the end of the next operating cycle.
This evaluation discussed each of the minor changes to the blades lLn detail and determined that these design alterations were acceptable and complied with the design bases as stated in the FSAR. It was determined that the small increase in cold reactiv-ity worth did not require changing lattice physics calculations or the operating plant process computer constants.
These-replacement control blades perform the.same design function .
as the original blades and it was determined in this safety evaluation that this minor modification did not involve an unreviewed safety question pursuant to 10CFR50.59.
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Annual Summ2ry of'JAFNPP Chenges, Tests, and-Experiments for 1988 i
-JAF-SE-88-160'- Zine Injection Passivation (GEZIP) System (Preoperational Test POT-34A) 1 L
1The purpose of this preoperational test was to verify the op-erability of the zine injection skid which-was' installed under '
Modification F1-88-048.- Safety analyses for this non-safety
- related modification were contained in Nuclear Safety Evaluations JAF-SE-88-138 and JAF-SE-88-173,-Rev. 1.
The scope of the test consisted of testing the pumps, mixers, valves, instruments, controls, interlocks, and equipment-compo-nents associated with the GEZIP System. .This test was ' performed 1 with the skid isolated from the Feedwater System, and therefore had no impact on plant safety.
The skid preoperational procedure was analyzed to assure that damage.to. plant-equipment would not occur. The execution of this preoperational test for the GEZIP System had no affect on the operation of any of the plant systems and did not constitute an unreviewed safety question.
Nuclear. Safety Evaluation JAF-SE-88-161 provided the safety analysis for Preoperational Test POT-95C which was performed-to determine the amount of zine injection required.
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!k n- . 3 Annual-Summary of JAFNPP Changes, Tosts, and Experiments for 1988 JAF-SE-88-161 - Zinc Injection Passivation System Operation -
(Preoperational Test POT-95C)
The purpose of this preoperational test was to establish the controls and hold points on the-amount of zine added to the Feedwater System. This safety evaluation addressed the monitor-
.ing, testing, and controls used during this initial injection.
The implementation of this non-safety related modification was analyzed in Nuclear Safety Evaluations JAF-SE-88-138 and. ,
-JAF-SE-88-173, Rev.-l.
This safety evaluation addressed the zinc concentrations required at various stages of the testing'and also the precautions taken to ensure the safe execution of POT-95C. The control of the production of the only significant activation product of concern was addressed and the factors involved in limiting this produc-tion-were discussed. The conduction of this test did not ad-versely affect any plant systems and was determined to not constitute an unreviewed-safety question as defined in 110CFR50 '. 59.
The safety analysis for the preoperational test for the hardware l
installed for this modification was contained in Nuclear Safety L Evaluation JAF-SE-88-160.
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-AnnualiSummary of_JAFNPP Chengas, Tests, and Experiments for 1988 '
JAF-SE-88-162 - Standby Liquid Control (SLC) System Squib !
Continuity Circuit Upgrade (Preoperational Test POT-11G)
This test was performed in order to verify that the newly mod-ified SLC squib continuity circuit functioned as designed. The ;
safety review for this safety-related circuit modification (F1-87-155) was contained in Nuclear Safety Evaluation JAF-SE-88-083.
This test was performed when the reactor was defueled. The
, conduction of this preoperational test with the reactor defueled l did not degrade any special safety function the SLC System was L' designed to perform. The plant Technical Specifications were not affected by this modification and no safety margins were reduced.
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This test did not involve an unreviewed safety question.
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'Annuel Summnry of JAFNPP-Chenges, Tests, and Experiments for 1988 JAF-SE-88-164 - Auxiliary Reactor Cavity Cleanup System This-safety evaluation reviewed the use of~the Auxiliary Reactor Cavity' Cleanup System (ARCCU). This non-safety related system, '
utilized during refueling outages, greatly-reduces the time required to obtain adequate water clarity for in-vessel activ-It is composed of two aump assemblies which are suspended ities.
into the reactor well and discL1arge to the skimmers of the Fuel.
Pool Cooling and Cleanup (FPCC) System. .
Since the entire ARCCU assembly.is contained within the reactor well, most credible mechanical failures would be entirely con- '
tained and result in no loss of pool water. The worst case system rupture would result in water-loss that would be within the capacity of the FPCC aump. -Several other possible failure modes were discussed in t'ais' evaluation and all were determined to not increase the potential for the loss of water from the Spent Fuel Pool. This evaluation concluded that an unreviewed safety question' pursuant to 10CFR50.59 did not exist as a result of the'use of the ARCCU system. ;
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and Experiments for 1988 i
i JAF-SE-88-166 - Vacuum Sipping of Fuel (Radiation' Activity Procedure RAP-7.1.3.2)
This evaluation provided a review of possible safety issues associated with-the set-up and operation of vacuum sipping equipment for testing of irradiated fuel assemblies on the Refuel Floor. -Vacuum _ sipping identifies fuel assemblies that contain a fuel rod with perforated cladding.
The vacuum sipping containers were designed so they could be installed in a control rod blade storage rack in the spent fuel pool. This ensured that cooling of the fuel assemblies was maintained throughout the testing precedure. After the fuel assembly was oneumatically sealed in the test canister, an air space above the assembly was evacuated to about 20" Hg absolute.
Any fission gases released due to fuel cladding perforations would then be: detected in this space.
~
This review addressed the heavy load concerns of the sipping i equipment, the placement of the chambers into the storage racks, '
the-cooling required, and the radiological concerns. It was determined that adherence to the plant's approved procedures for this work would ensure the safe implementation of this test.
This evaluation concluded that an unreviewed safety question as ,
defined by 10CFR50.59 did not exist as a result of this I non-safety related. activity.
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- m> .s Annusl Summary of JAFNPP Ch ngos, Tosts, and. Experiments for 1988-JAF-SE-88-169 - Bypass of SRM and IRM Detector Not Full in i Rod Block During 1988 Refueling Outage This safety evaluation provided a review of the safety concerns involved with plant Jumper No.88-148. This safety-related temporary modification effectively. bypassed the signals which cause a rod block whenever the Source Range Monitor (SRM):- or-Intermediate Range Monitor (IRM) detectors are not fully insert-ed. Due to outage activities, relay panel 25-14 did not have power to it which_ caused an erroneous "not fully inserted"
-signal.
Work Activity Control Procedure WACP 10.1.3 was utilized to
-administrative 1y control this jumper. Before this jumper was implemented, it was-verified that the SRM and IRM detectors were inserted to their normal operating level. The measures outlined in this evaluation provided adequate assurance that this jumper would not degrade-the safety objective of the Neutron Monitoring System or the Reactor Manual Control System.
It was concluded that an unreviewed safety question pursuant to .
10CFR50.59 did not exist as a result of this temporary modifica-tion.
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. e Annual Summary of JAFNPP/Chingas, TGsts, and Experiments for 1988 JAF-SE-88-170 - Circulating Water Pumps Lube Oil Sampling Line Addition (Minor Modification M1-88-149)
This: safety evaluation provided the safety analysis for non-safety related Modification-M1-88-149 which extended the circulating water pumps lube oil reservoir level sightglass line through the pump motor pedestal and added a valve to accommodate pump lube oil sampling. This-modification did not have any adverse affects on the Circulating Water. System. ;
This evaluation referenced calculations which demonstrated that the new hole drilled into the motor medestal of each pump to provide for the extension of the sightglass lines did not result in unacceptable maximum stress values. There.was no change to the design bases for the Condenser Circulating Water System as a result of this modification.
An unreviewed-safety question did not exist as a result of the implementation of this minor modification.
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.,, -i Annus1 Summary of-JAFNPP Changes, Tosts, and Experiments for'1988 JAF-SE-88-172 - Feedpump Turbine Exhaust Duct Drainline A and B Flange Installation (Minor Modification M1-88-187)
This safety evaluation was written in order to provide the safety analysis for Modification M1-88-187 which installed flanged connections in the-A and B reactor feedpump. turbine exhaust duct drain lines. This non-safety related modification was implement-ed in order to facilitate maintenance on the turbines which in the past had required the cutting and subsequent rewelding of drain lines whenever maintenance was performed.
The flange and bolting materials used are consistent with the requirements for these non-seismic drain lines and the gaskets
. utilized will ensure the lines remain air and watertight ~. The additional weight to the lines was determined to have negligible effect on the piping and support stresses.
This flange addition did not adversely affect the design bases for this system and it was determined not to constitute an unreviewed safety question.
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. , i Annual Summary of JAFNPP Chenges, Tosts, and Experiments for 1988 1
JAF-SE-88-173, Rev. Zine Injection Passivation System (GEZIP) Installation of Hardware (Plant Modification F1-88-048)
This safety evaluation addressed the installation of th~e hardware for the General Electric Zinc Injection Passivation (GEZIP) -
System. Ecuipment installed by this non-safety related modifica-tion incluc ed the injection skid, ion chromatograph analyzer '
panel, and the sample filter analyzer panel. The injection skid equipment installed: included the supply tank, mixers, injection
' pumps, valve and control-instruments. This evaluation also addressed the tie-ins to permanent plant equipment.
Piping was installed to carry feedwater to the skid and then to carry the zinc-oxide laden feedwater to the suction side of the feedwater pumps. The installation of the GEZIP hardware and tie-ins was determined to have no impact on plantJsafety because no safety-related equipment was affected and the plant was in cold shutdown during the installation. Administrative controls ensured that the injection of zine oxide into the Feedwater System was not a possibility during this hardware installation.
The margin of safety as defined in the bases of the Technical
' Specifications was not reduced because the Feedwater System operation was not altered. ,
It was the conclusion of this safety evaluation that an unre-viewed ~ safety question did not exist as a result of the installa-tion of this hardware for the GEZIP System.
The safety analysis for the GEZIP process was contained in 4, Nuclear Safety Evaluation JAF-SE-88-136. The preoperational tests associated with this modification were analyzed in Nuclear Safety Evaluations JAF-SE-88-160 and JAF-SE-88-161.
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- . s Annuel Summary of-JAFNPP Chnngos,. Tests, and Experiments for 1988 JAF-SE-88-175 - Standby Gas Treatment (SGT) Trains A & B Functional Test (Preoperational Test POT-01-125D)'
This~ test was conducted to verify the operability of the SGT System Trains A &'B following the completion of safety-related Modification:F1-85-045 (analyzed in Safety Evaluation i JAF-SE-85:127, Rev. 1). This test was performed with the plant-in cold shutdown and testing of the second SGT train did not commence until the first SGT train's' testing had been satisfac-torily completed.
There wasuno reduction in plant safety during the performance of this test. This test involved verifying proper circuit response, :
to simulated electrical conditions by observing the status of-
< relays, indicating lights,-and annunciators on various appropri-ate panels. This test' simulated all the circuit conditions necessary to. verify proper operability of each SGT train and fire protection circuit associated with each of the disconnected carbon drying heaters.
This test did not alter the safety design basis of the existing system and was determined to not constitute an unreviewed safety question as defined in 10CFR50.59.
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'JAF-SE-88-176 - ITT Hannel Dahl Valve Pin Material Substi-
,tution (Minor Modification M1-88-180)
-This safety evaluation was written to provide the technical review associated with Minor Modification M1-88-180 which au-thorized the replacement of the plug stem pins for the Reactor Building Closed Loop Cooling Water System's safety-related containment isolation air operated valves. The new plug stem pins are made of material different from those originally sup-plied. The chemical and mechanical properties of the replacement ,
material was analyzed and it was determined that-the substitution was acceptable.
This evaluation provided a listing of the affected valves and E provided the justification that the valves' operability was not impaired by this material substitution. Since the performance of the subject valves was not affected by this modification, their-ability to perform this design function was not reduced.
The conclusion of this safety evaluation was that this material substitution did not constitute an unreviewed safety question.
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Annun1 Summtry of JAFNPP Changes, Tests, and' Experiments for 1988 1
n JAF-SE-88-177 -
Material Substitutions for William Powell Valves (Minor Modification M1-88-123)'
This safety evaluation provided the technical review for Minor Modification M1-88-123. This modification authorized the use of various vendor recommended replacement valve parts which are made of materials'different from those originally supplied. The chemical and mechanical properties of the-replacement materials '
were analyzed and it was determined that the substitutions were acceptable for the design applications.
. This evaluation discussed the valves.and materials in question y and provided the justification for the.new material's acceptance.
The ability of the affected valves to perform their design function.was not compromised by this modification and there was >
no impact on the plant's design bases.
The material substitutions (for-both safety-related and non-safety related valves) associated with this modification were-determined not to constitute an unreviewed safety question.
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_(
. -1 Annual Summary of~JAFNPP Changes, Tests, a
and Experiments for 1988 JAF-SE-88-183 - Replacement of Emergency Diesel Generator (EDG) Air Start System Pressure Switches 93PS-9A,B,C,D and 93PS-10A,B,C,D (Minor Modification M1-88-194)
, LThis safety evaluation evaluated the replacement, on an as-needed
- basis, of the safety-related EDG air start pressure switches.
L This modification was necessary due to the fact that-the orig-inally installed switches are no longer manufactured. It was determined that the installation of these new switches would-require a setpoint change plus minor tubing and support changes.
The reset setpoint was required to change due to design differ-ences between the original switch and theLrealacement switch.
This lowering of the reset setpoint causes tae compressor to cycle more often, but it was determined that this would not have any adverse affects on the Air Start System.
This evaluation discussed the differences.in switch design and presented the justifications for the acceptance of these differ-ences. The new switches satisfy the environmental and seismic criteria for these QA Category I instruments. All of the techni-cal and quality requirements of the original purchase order
-(AP-9) were also met.
There.were no changes-required to either the FSAR or Technical Specifications and this upgraded replacement was determined to enhance the reliability of the Air Start System.
This safety evaluation concluded that an unreviewed safety t question did not exist as a result of this modification.
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Annusi Summary of JAFNPP Chenges, Tests,
.and Experiments for 1988 LJAF-SE-88-185 - Additional Drywell and Torus Pressure In-dication (Minor Modification M1-82-002)
This' safety evaluation was written to provide an analysis of the safety concerns associated with the connecting of two pressure indicating transmitters (16-1 PIT-101, 103) in parallel with existing transmitters in order to provide redundant Drywell.and torus pressure indication. These changes were initially imple-mented as temporary modifications in accordance with Work Activi-ty Control Procedure WACP 10.1.3 and were made permanent by Minor Modification M1-82-002.
These transmitters were originally connected to the Drywell and Torus reference pressure vessels which were retired under a previous modification. The placement of these two pressure indicating transmitters in marallel with the existing instrumen-tation does not change the basic design of containment integrated leaktge rate test-instrumentation. It was concluded that the addition of this non-safety related redundant-instrumentation did not constitute an unreviewed safety question pursuant to 10CFR50.59. '
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c . r Annu21'Summzry of JAFNPP Chengas, Tests, and Experiments for 1988 l
JAF-SE-88-186 - 10MOV-18 Motor Actuator Regearing (Minor Modification M1-88-198)
This safety evaluation provided the safety analysis for modifica-tion M1-88-198 which replaced the internal Limitorque gearing of-10MOV-18 with components which arovide_approximately the same overall actuator gear ratio with the self-locking worm gear set.
This safety-related modification eliminated the non-locking worm-and worm gear set which.could have contributed to " hammering" which is the rapid cycling arf a closed valve. This " hammering" could have resulted in damage to the valve and to the actuator.
Although the valve's stroke times are affected slightly as a ,.
result-of this modification, the values are within the limits-
, s7ecified in the Technical Specifications. This change decreased l . the chance of equipment malfunction without degrading equipment reliability in other areas. ;
It was determined in this safety evaluation that this modifica-tion did not adversely affect the' operation of this valve or the-Residual Heat Removal System and that these changes did not l constitute an unreviewed safety question.
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-yw .6 Annual Summary of JAFNPP Changos, Tests, _,
and Experiments for 1988 l
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This. safety evaluation ~ addressed the replacement of two welded Ljoints on_the safety-related Main Steam Pipe Support PFSK-1880 with bolted connections to facilitate removal and installation of the-support structure, as required, in order to perform routine maintenance on valve 10A0V-68B which is located directly under ;
the support. This minor change in support. design did not ad- i versely affect-the integrity of this support.
An engineering evaluation was performed by Stone =end Webster-Engineering Corporation determining that this-bolted connection was acceptable and did not alter the intended design function of the pipe support. -Suitable protection was arovided to ensure adjacent equipment was not damaged during the implementation of
-this modification.
The structural change associated.with this minor modification did not constitute an unreviewed safety question.
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, v Annu21ISummiry of,JAFNPP Changos, Tosts, and Experiments for 1988
-s JAF-SE-88-190,-Rev. 2 - Repair of In-Vessel Core Spray Line
~
Using a Welded Clamshell Sleeve (Plant Modification F1-88-199)
This safety evaluation addressed the two-part welded clamshell repair which.was performed on the safety-related "B" loop of the core: spray piping inside.the reactor vessel. This repair was-accomplished under Modification F1-88-199. The crack which was
' located in the core spray piping between the core spray nozzle and the shroud in the reactor vessel, was discovered-duringLthe in-vessel visual inspections for Intergranular Stress Corrosion
-Cracking-(IGSCC).
This-evaluation addressed the material, welding, inspection and structural requirements necessary which ensured compliance with ;
all the codes which were applicable for this repair work. Stress '
analyses were performed which verified that the repair design :
l satisfied all of the' design bases criteria for the Core Spray-Piping System.
The work was accomplished with the reactor fueled utilizing a shielded work bucket lowered by the Refuel Floor crane. The-safety analysis for this work bucket was contained in Nuclear Safety Evaluation JAF-SE-88-192, p This~ modification did not affect the design and safety bases.of the Core Spray System and did not constitute an unreviewed safety question as defined in 10CFR50.59.
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Annual Summary of JAFFPP Chnngos, TGsts, and Experiments for 1988 X'
JAF-SE-88-192 - Evaluation of Shielded Work Bucket for In Vessel Core Spray Pi Mod.fication F1-88-199)pe Repair (Plant This safety. evaluation addressed the safety concerns associated "
with the use of shielded work bucket to grovide access for the ,
repair of a crack in'the safety-related B" core spray pipe located in the reactor vessel. A safety analysis of the crack '
repair was contained in Nuclear Safety Evaluation JAF-SE-88-190, Rev. 2.
The work bucket was equipped with various amounts of lead shield-ing to provide the required radiation shielding. An opening was provided in the lower section of the structure to permit access for repair of cracked piping. These and other changes to the bucket were performed to provide a safe work platform for the completion of the required task. It was determined that these changes did not affect the structural integrity of the work bucket. .
i Special rigging precautions were taken in order to ensure that at l least a minimum safety factor of 10:1 was always maintained.
Other special precautions were taken to ensure that debris would not fall into'the fueled reactor and that the reactor vessel itself would not be damaged during this repair operation.
This safety evaluation concluded that an unreviewed safety question did not exist as a result of this activity.
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s s, Annus1 Summ;ry of JAFNPP Ch ngos, Tosts,
,and Experiments for 1988 r
JAF-SE-88-193 - Replacement of Reactor Water Cleanup (RWCU)
Manual Valves 12RWC-83A,B (Minor Modification M1-88-203)
Non-safety related Minor Modification M1-88-203 consisted of the i replacement of the, Control Rod Drive (CRD) mini murge first manual isolation val've (12RWC-83A,B) to both RUCJ recirculation pumps (12P-1A,B). The new globe valves replaced gate valves which had been installed previously. The new valves are easier to maintain and meet the requirements of Stone and Webster Engineering Cor? oration's piping specification AP-23 for Class 1504 pipe. Although the new valves contain stellite seats, discs, and backseats, the installation of these valves was necessary due to time and schedule considerations.
This modification did not change the desi n bases for the RWCU
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System, but instead improved system relia 111ty by replacing the worn valves with new ones. This installation of more reliable valves did not adversely affect the operation of the RWCU System and was determined not to constitute an unreviewed safety ques-tion as defined in 10CFR50.59.
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34E-2A and 2B Drain Valve Replacement (Minor Modification M1-88-205)
Minor Modification M1-88-205 consisted of re. placing the drain valve configuration of the seal water coulers 34E-2A and 2B. The replacement configuration meets the requirements of the original drain configuration design. The new valves were procured to specifications which meet or exceed the original design require-ments for their intended application. This change corrected a vibration problem associated with the previous configuration.
Since the weight of the new valve is less than the originally installed valve and the cantilever length was reduced, the new drain configuration was determined to be less susceptible to 4 vibration failure. This change in the non-seismic drain config-uration did not affect the operation of the seal water cooler and no changes to the Technical Specifications were required.
This non-safety related modification had no impact on any safe-ty-related equipment or structures and was determined not to involve an unreviewed safety question.
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Annual Summary of JAFNPP Changes, Tcsts, and Experiments for 1988 t JAF-SE-88-196 - Removal of Check Valve Internals - 46SWS-9 (Minor Modification M1-88-206)
This safety evaluation provided the safety analysis for Minor :
Modification M1-88-206. This non-safety related modification '
evaluated the as-found condition of the wafer-type check valve 46SWS-9. This valve was found to have its disc loose in the piping downstream of the valve. Check valve 46SWS-9 was de-termined to be unnecessary to prevent backflow in an operating system where the main system branches are fed from a single header. Valve 46SWS-10 will provide the isolation of the Reactor Building service water supply if required. .
The margin of safety as defined in the bases for the Technical Specifications was not reduced and no plant safety systems or functions were impacted. It war determined in this safety evaluation that an unreviewed safety question did not exist as a result of this valve's condition.
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- e. e Annual Summary of JAFNPP Ch;ngos, Tosts, and Experiments for 1988 JAF-SE-88-202, Rev. 1- Pressure Test Relief For Minor Repairs and Replacements This safety evaluation addressed the safety issues concerning the adoption of system hydrostatic pressure tests exemptions pursuant to ASME Boiler and Pressure Vessel Code,Section XI. The types of minor repairs and replacements for which exemptions shall apply were presented in detail for both safety-related and non-safety related systems.
This evaluation discussed the reasons for hydrostatic tests and the technical arguments used in the justification of the ex-emptions. Part of the reasoning for allowing exemptions is that
! numerous system and/or component tests at 1.5 times the design pressure could actually initiate or lengthen imperfections due to low cycle fatigue. ,
These exemptions were determined to not reduce the margin of safety as defined in the bases for the Technical Specifications and no changes to either the FSAR er Technical Specifications was required. The evaluation concluded that an unreviewed safety question pursuant to 10CFR50.59 did not exist as a result of this type of pressure test relief.
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Annual Summ3ry of JAFNPP Ch;ngos, Tests, and Experiments for 1988 JAF-SE-88-203 - Operation with Reported Flaws in Reactor Recirculation System Weld Numbers 12-4, 28-53,28-112 and 28-33.
This safety evaluation analyzed the continued plant operation with unrepaired flaws in certain safety-related Reactor Water Recirculation System welds. A detailed analysis of each of the flaws was presented in this evaluation. Crack growth evaluations were done in accordance with NUREG-0313 Rev. 2, and conservative values and assummtions were use; throughout. The predicted time required to reac'1 the Section XI allowable flaw size limits were given for each and it was determined that these flaw depths would
, remain below the allowable limit for at least the next operating cycle.
The postulated worst case failure for this piping has been analyzed in the FSAR under the design basis accident. Any failure of these piping joints would have less severe effects than the design basis accident. This evaluation determined that continued operation with these unrepaired flaws would not consti-tute an unreviewed safety question pursuant to 10CFR50.59.
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Annual Summary of JAFNPP Ch:ngos, Tosts, and Experiments for 1988 JAF-SE-88-205 - Core Spray Motor Vibration Support (Plant Modification F1-88-223)
This safety evaluation provided the safety analysis of Modifica-tion F1-88-223 which added a vibration dampening support assembly at the upper end shield of the A & B core spray pump motors.
This struccure installed at the top of the motor provides enough stiffness to shift the natural frequency of the pump and motor assembly away from the forcing frequencies, thereby reducing the amplitude of vibration.
With these reduced vibration levels, it was determined that the resulting stress, displacements and forces within the pump will be well within the design allowable values. Analyses showed that the combination of seismic.plus operating stresses are less with the addition of the support.
This change did not affect the heat rise in the motor nor did it degrade the motor's reliability due to thermal effects.
It was the conclusion of this safety evaluation that this safe-tv-related modification enhanced the structural integrity and r'eliability of the core spray pumps and that this modification did not constitute an unreviewed safety question pursuant to 10CFR50.59.
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- - r d 7 Annual Summ ry of'JAFNPP Changas, Tosts, and Experiments for 1988 JAF-SE-88-206 - Replacement 3/4 Inch Y-Globe Valves for 29MST-703A,B,C,D (Minor Modification M1-88-229)
This safety evaluation provided the safety analysis for Minor Modification M1-88-229 which replaced the four original plant Y-globe valves used for the safety-related Main Steam Isolation Valves' stem leakoff isolation. The replacement valves were procured to meet or exceed all of the applicable technical and quality requirements specified in the original purchase speci-fications for these leakoff isolation valves.
This evaluation compared the differences between the replacement valves and the originally installed valves. The weight differ-ence did not adversely affect the piping or support stresses and the minor material differences were found to be acceptable.
Tlue margin of safety as defined in the Technical Specifications was not reduced by this modification. The replacement of these valves did not adversely affect the operation of the MSIVs and was determined to not constitute an unreviewed safety question.
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r a Annual Summary of JAFNPP Chtng:s, Tosts, and Experiments for 1988
! JAF-SE-88-210 - Feedwater Insulation Seal at Penetration 16X-9A and 9B (Minor Modification M1-88-234)
The purpose of this safety evaluation was to provide a safety analysis of Minor Modification M1-88-234 which added a pressure seal boot to the main steam pipe tunnel side of feedwater pene-trations 16X-9A and 16X-9B This non-safety related mcdification was installed because an o en gap had developed in these pene-trations due to the therma insulation blanket being disturbed by plant personnel. The boots improved the barrier between the air gap surrounding the Drywell shell and the main steam pipe tunnel.
It was determined that those pressure seal boots will assist in maintaining secondary containment by reducing air leakage, thereby increasing the ability of the Standby Gas Treatment System to reduce Reactor Building aressure to 1/4" water vacuum.
The addition of the pressure seal boot did not degrade the function or integrity of the piping penetrations and no change in-the Technical Specifications was required.
This safety evaluation concluded that the installation of these j pressure seal boots was acceptable and that an unreviewed safety question did not exist as a result of the implementation of this modification.
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Annual Summ ry of JAFNPP Chcngos, Tosts,
,and Experiments for 1988 1 JAF-SE-88-214 - 5th Point Heater 33E-5A Inlet Drain Valve Replacement (Minor Modification M1-88-240) l 1
4 This safety evaluation provided the safety analysis for Minor Modification M1-88-240 which replaced 31XST-104, the non-safety ;
related manually operated inlet drain valve for the 5th point heater 33E-5A. The new replacement valve was procured to meet or ;
exceed all of the applicable technical and quality requirements specified for the original valve. The slight increase in weight was determined to have negligible effect on the piping stresses.
The installation of this modification did not affect any safe- j ty-related structures or components and did not change the system in a mannar that can change the existing design basis or accident '
analysis. l The replacement of this valve did not involve an unreviewed safety question.
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r a Annu21 Summ0ry of JAFNPP Ch;nges, Tosts, and Experiments for 1988
~JAF-SE-88-215 - 70 Degree Full Open Position for Containment Exhaust Bypass Valves 27MOV-113 and 27MOV-117 (Minor Modification M1-88-239)
This safety evalur. tion was written in order to provide the justification for the installed condition of two safety-related containment isolation valves (27MOV-113 and 117) in the Drywell Inerting, Containment Air Dilution, and Purge System. The installed condition of these valves limits the full open position to 70 degrees instead of the usual 90 degrees for this type of butterfly valve.
It was determined that the reason for the partial open limit of 70 degrees was to ensure a closing time of 5 seconds in the event of a containment isolation signal to these valves as required by the Technical Specifications. The net reduction in overall flow capacity of the subject valves was determined to be minor and did not have an affect on the safety function of these valves. This evaluation reviewed the original design basis and it was de-termined that the system's original design was not altered.
It was concluded in this safety analysis that the installed condition of these valves did constitute an unreviewed safety question.
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Annual Summary of JAFNPP Chengos, Tests, and Experiments for 1988
-JAF-SE-88-216 - Interim Justification for Core Spray Pump Minimum Flow Orifice Size (14RO-27A,B)
This safety _ evaluation was performed in order to evaluate and justify the installed bore dimensions of two safety-related core spray pump minimum flow orifices in the core spray pump minimum flow lines. As part of the response to NRC IE Bulletin 88-04,
" Potential Safety-Related Puma Loss", it was determined that the correct bore diameter should'ae 1.23" instead of the installed 1.19" bore diameter. This small bore size-difference represents a flow reduction of less than 7% of the total minimum flow required. It was determined that this relatively small minimum flow reduction did not adversely affect pump low flow operation, i This evaluation discussed the design considerations associated with these orifices. Flow analyses, vendor recommendations, and
- plant operating experience were addressed in detail and the conclusion was that the slight reduction in minimum flow attrib-utable to the installed orifices has not, and will not, affect the continued reliability of the core spray pumps. .
Since the reliable operation of the core saray pumps was de-termined to not be adversely affected by this reduction in flow, .
it was concluded that an unreviewed safety question did not exist.
Minor Modification M1-88-251 eventually replaced these orifices with those having correctly sized diameters. An analysis for this modification was contained in Nuclear Safety Evaluation JAF-SE-88-220. ,
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! Annual Summ0ry of JAFNPP Ch ng:s, Tosts, and Experiments for 1988 ,
JAF-SE-88-217 - Interim Plant Operation with Degraded Crescent Area Unit Cooler Capacity (66UC-22A through -22K)
The purpose of this safety evaluation was to assess the safe interim operation of the plant with less than the original design basis heat removal capacity of the safety-related crescent area unit coolers. Due to degraded unit cooler heat removal capabil-ity, safe plant operation was established by requiring that service water (la).e) temperatures be less than or equal to 66*F.
This evaluation discussed the original design bases, effect of degraded unit coolers, unit cooler design efficiency, environ -
mental qualification concerns and impact on plant licensing documents. It was determined that the unit coolers would perform their design function when the limitations discussed in this evaluation were complied with. .This review concluded that all technical specification recuirements would be met by this limita-tion and that an unreviewec, safety question as defined by 10CFR50.59 did not exist, i
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w + l Annual Summ ry of JAFNPP Changas, Tcsts, and Experiments for 1988 JAF-SE-88-218 - Rerouting of Service Water Supply and Return Piping for East Crescent Unit Cooler 66UC-22D. Gasket Substitution for Unit Coolers 66UC-22A through 22K (Minor Modifica- i tion M1-88-248) ,
This safety evaluation provided the safety analysis associated with Minor Modification M1-88-248 which restored the service water supply and return piping to the originally intended design configuration for 66UC-22D. In addition, the use of neoprene as the gasket material for all of the east and west crescent unit cooler channel covers was authorized by this modification.
The arrangement of the inlet and outlet piping for 66UC-22D was -
found to be reversed which significantly reduced the original design heat removal capacity of the unit cooler. The restoration of the piping to the original design configuration improved the performance of this safety-related unit cooler. The restoration of this piping configuration required no further seismic analysis based on comparison with similar acceptable piping configurations on the identical unit coolers and a Stone and Webster report on safety-related piping systems.
The new gasket material for the channel covers was determined to have greater compressability than the original material supplied which aids in obtaining a leak tight seal.
This modification did not change the system in a manner that altered the-existing design bases and accident analyses.
It was concluded in this safety evaluation that the implementa-tion of these changes did not constitute an unreviewed safety question as defined in 10CFR50.59.
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Annun1 Summary of JAFNPP Changos, Tests, and Experiments for 1988 JAF-SE-88-220 - Replacement of Core Spray Pump Minimum Flow Orifices 14RO-27A&B (Minor Modification M1-88-251)
This safety evaluation evaluated the replacement of the core spray pump minimum flow orifices with increased bore diameter-orifices. This safety-related modification was performed in order to obtain the vendor recommended minimum flow of 475 gpm.
This installation ensures that the core spray pumps will not be operated below the manufacturers recommended pump minimum flow.
The replacement plate is identical in design and material to the
'areviously installed plate with the exception of the increase in
) ore diameter. This modification did not reduce the margin of safety as defined in the bases for the Technical Specifications because it increased the reliability of the core spray pumps by eliminating op_eration outside the manufacturer's recommended range.
This modification increased the reliability of_the core spray pumps and was determined to not constitute an unreviewed safety question as defined.in 10CFR50.59, l i l
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