ML19308D652
| ML19308D652 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 12/15/1976 |
| From: | FLORIDA POWER CORP. |
| To: | |
| References | |
| PROC-761215, TAC-04548, TAC-4548, NUDOCS 8003120698 | |
| Download: ML19308D652 (79) | |
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1 SURVEILLANCE AND INSERVICE INSPECTION PROGRAbi FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 o
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INDEX The following list of Plant Systems and their respective surveillance and/or inservice inspection program require-ments are identified:
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Reactivity Control Systems Boration Systems Reactol Coolan: System Safety Valves Pressurizer Steam Generators Safety Class I components - Structural Integrity Emergency Core Cooling Systems Core Flooding Tanks ECCS Subsystems Containment Systems Depressurization and Cooling Systems Containment Isolation Valves Plant Systems Turbine Cycle Closed Cycle Cooling Water Systems Sea Water Systems Hydraulic Snubbers Other Safety-Related System Not Addressed In the CR#{ STS:
Fire System Spent Fuel Cooling System s
INDEX -(continued)
Other Safety-Related System Not Addressed In the CRf 3 STS (cont'd):
Radioactive Waste Disposal System Secondary Services Closed Cycle Cooling System Instrument Air System
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4 Penetration Cooling System
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J REACTIVITY CONTROL SYSTEMS t
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'BORATION SYSTEMS
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Equipment FPC Surveillance Program (1)
- 1) Flow Paths (Valve) - Shutdown Section 4.1.2.'1 a) DHV-5, 6, 11, 12, 14, 25 34, 35, 75, 76 b) MUV-23, 24, 25, 26, 27, 31, 58, 73, 103, 108 c) CAV-57, 60
- 2) Flow Paths (Valve) - Operating Section 4.1.2.2 Same as item (1) above.
Flow Paths (Valve)-Shutdown-Every 18 mths. Section 4.1.2.2 c.
a) DHV-42, 43 b) MUV-64, 90, 91, 96, 97
- 3) Makeup Pump - Shutdown Sections 4.1.2.3 a) MUP-1A, IB, 1C
- 4) Makeup Pump - Operating Sections 4.1.2.4 a) MUP-1A, 1B, 1C
- 5) Decay Heat Removal Pump - Shutdow;n Sections 4.1.2.5 a) DHP-1A, 1B
,6) Boric Acid Pump - Shutdown Sections 4.1.2.6
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a) CAP-1A, 1B 7)-Boric Acid Pump - Operating Sections 4.1.2.7'.
a) CAP-1A, 1B l
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(1) The FPC Surveillance Program requirement sections referenced below were contained in the " Proof 4 Review Copy" of the CR#3 STS received by FPC on March 30, 1976.
As this issue of the CR#3 Technical Specifications was thought to#be at that time the final " Proof 5 Review Copy", the surveillance require-ments contained therein were used as the basis for FPC's sur-veillance program.
(Copies of the referenced sections are attached.)
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4.1.2.1 At least one of the above required flow paths shall be demon-(
strated OPERABLE:
At least once per 7 days by:
Cycl.ing each testable power operated 'or automatic valve in the flou path through at least one complete cycle of ful1 travel, and 4.1.2.2 Each of the above required flow paths shall be demonstrated OPERABLE:
f At least once per 7 days by:
- Cycling each testable power operated or automatic valve in the flow path through at least one complete cycle of full travel, and At least once per 13 months, during) shutdown, by cycling each power operated (excluding automatic valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel.
4.1.2.3 At least the above required makeup pump shall be demonstrated OPERABLE at least once per 31 days 'cv:
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Starting (unless
- ady operating) the pump from the control
- room, Verifying, that on recirculation flow, the pump develops a discharge pressure of > 2500 psig, Verifying pump operation for at least 15 minutes, and 3.
Verifying that the p'ap is aligned to' receive electrical. power from an OPERABLE emergency bus.
4.1.2.4 At least tuo makeup pumps shall be demonstrated OPERABLE at least once per 31 days on a STAGGERED TEST BASIS.by:
Starting (unless already operating) each pump from the control
- room, i
Verifying, that on recirculation flow, each pump develops a discharge pressure of > 2500 psig, c,
Verifyii19 that each pump operates for at least 15 minutes, and Verifying that each pump is aligned to receive electrical y
power from separate OPERADLE emergency busses.
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4.1.2.5 At least the above required DiiR pump shall be demonstrate'd OPERABLE at least once per 31 days by:
Starting (unless already operating) the oump from the control
- room, Verifying, that on recirculation flow, the pump develops a discharge pressure of 1 150 psig, Verifying pump operation
/ at least 15 minutes, and Verifying that the pump is aligned to. receive electrical power from an OPEllABLE emergency bus.
4.1.2.6 At least the above required boric acid pump shall be demonstrated OPEllABLE at least once per 7 days by:
Starting (unless already operating) the pump from the control
- room, Verifying, that on recirculation flow, the pump develops a pressure,of 1 75 psig, 7
Verifying pump operation for at least 15 minutes, and I
Verifying that the pump is aligned to receive electrical power l
from an OPEllABLE emerguncy bus.
4.1.2 7 At least the above required' boric acid pump shall be demonst' rated OPERABLE at least once per 7 days by:
Starting (unless already operating) the pump from the control g
- room, Verifying, thct on recirculation flow, the pump develops a j
discharge pressure of 1 75 psig, j
Verifying pump operation for at least 15' minutes, ad Verifying tiat the pump is aligned to receive electrical power from an OPEMBLE emergency bus.
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" Equipment FPC Surveillance Program. (1) f
- 1) Safety Valves - Shutdown Section 4.4.2 a) RCV-8F, 9F
- 2) Safety Valves - Operating' Section 4.4.3 a) RCV-8F, 9F
- 3) Pressurizer Sections 4.4.4-Sections 4.4.5.1 thru 4.4.5.5
- 4) Steam Generators a) Steam Generator Tubes
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(1) The FT/C Surveillance Program requirement sections referenced below were contained in the " Proof 6 Review Copy" of the CR#3 STS received by FPC on March 30, 1976.
As this issue of the CR#3 Technical Specifications was thougnt to be at that time the final " Proof 4 Review Copy", the surveillance require-ments contained therein were used as the basis for FPC's sur-veillance program.
(Copies of the referenced sections are
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attached.)
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4.4.2 The pressurizer code safety valve shall be demonstrated OPERABLE
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per Surveillance Requiremant 4.4.3.
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4.4.3 !!ach pressurizer code safety valve shall be demonstrated OPERABLE
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with a lif t setting of 2500 PSIG ' 15 in accordance with Section XI 'of the A5hE Boiler and Pressure Vessel Code,1974 Edition.
4.' 4. '4 The pressurizer shall be demonstrated OPERABLE by:
Verifying pressurizer level control instrumentation OPERASLE at least once per; 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by performance of a CHAllt:EL CHECK, and 18 months by performance of a CHAlWEL CALIBRATIOM.
Steam Generator Sanule Selection and Insrection - Each steam 4.4.5.1 generator shall be deterninco diURPdLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1.
Steam Generator Tube Somole Selection and Inspection - The 4.4.5'.2 steam generator tube minimum scmple size, inspection result classifi-Letton, and the correspon:iing action required shall be as specified in The inservice inspection of steam generator tubes shall be Table 4.4-2.
performed at the frequencies specified in' Specification 4.4.5.3 and the
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inspected tubes shall be verified acceptable per the acceptance criteria of. Specifica tion 4.4.5.4 Steam generator tubes shall be examined in acc.ordance with the ASI'E Boiler and Pressure Vessel Code - Section XI.-
" Inservice Inspection of I:uclear Pouer Plant Components".1974 Edition The tubes selected for each inservice inspection shall include at least 3% of the total n' umber of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:
Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at icast 50 of the tubes inspected shall be from these critical areas.
The first inservice inspection (subsequent to-the p#cserv' ice
. inspection) of each steam generator shall include:
All nonplugged tubes that previously had detectable wall penetrations (>20".), and u
Tubes in those arcas where experience has indicated potential problems.
The second and third inserv, ice inspections may be less than a
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full tube inspection by concentrating (selecting at least 50%
of the tubes to the inspected) the inspection on those areas of the tube sheet array and on those portions of the tubes where tubes with ir.'?erfections were previously found.
.c The results of each sample inspection shall be classified.into one of the following three categories:
7-Category Inspection Results C-1 Less than 5'; of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
C-2 One or more tubes, but not more than 1% of
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the total tubes inspected are defective, or between 5% and 107.* of the total tubes inspected are degraded tubes.
C-3.
More than 10% of the total tubes inspected are degraded tubes or more than 15 of the inspected tubes are defective.
Note:
In all inspections, previously degraded tubes must exhibit significant (>105)'further wall penetrations to be included in the above percentage calculations.
4.4.5.3 Inspection Frecuencies - The above required inservice inspections of steam generator tubes snall be performed &t the following frequencies:
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. he first inservice insocction shall be performed af ter 6 E fective Full power Months but within 24 calendar months of
'iitial criticality.
Subsequent inservice inspections shall
.$e performed at intervals of not less than 12 nor more than 24 S
calendar conths after the previous inspection.
If two consecu-tive inspections following service under MT conditions, not
. including the preservice inspection, result in all inspection i
Tbhults falling into the C-l category or if two consecutive inspections demonstrate that previuusly observed degradation has not continued and no additional ~ degradation has occurred, the inspection interval r.iay be extended to a maximum of once
, er 40 months.
p If. the inservice inspection of a steam generator conducted in i
accordance with Table 4.4-2 requires a third sampio inspection whose results fall in Category C-3, the inspection frequency shall be reduced to at'least once per 20 months.
The reduction in inspection frequency shall apply until a subsequent inspectioit i
demonstrates that a third sample inspection is not required.
I Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shtgdown subscquent to any of the follcuing conditions.
Primary-to-sccendary tubes'lcaks (not including.lcahs originating frca tube-to-tube sheet wolds) in excess of the limits of Specification 3.4.6.2, A seismic occurrence greater than the Operating Basis' Earthquake,
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1 A loss-of-coolant accident requiring actuation of the
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enginecred safeguards, or A main steam line or feedwater line break.
4.4.5.4 Acceptance Critoria
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As used in this Specif,ication:
Imperfection means an exception to the dimensions, finish or contour Ef a tube from that rcquired by fabrication drawings or :,pecifications.
Eddy-current testing indications below 20% of the nominal' tube wall thickness, if detectable, may be considered as imperfections.
' Degradation,r. cans a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube.
D_cgraded Tube means a tube containing imperfcctions >20%
o of the nominal wall thickness caused by degradation.
%Decradationmeanstilepercentageofthetubewall thickncss aficcted or removed by degradation.
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Defect means an imperfection of such severity that it exceeds the plugging limit.
A tube containing a defect is defective.
Any tube which does not permit the passage of the eddy-current inspection probe shall be deemed a defective tube.
Plugging Limit' means the imperfection depth at or beyond which the tube shall be removed frcm service because it may beccme unserviceable prior to the next inspection and is< equal to 40% of the nominal. tube wal! thickness.
Unserviceable describes the condition of a tube if it leaks or contains a defect larce enouch to affect its structural integrity in the event of an Operating Basis
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Earthquake, a' loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3.c, above.
Tube inspection rneans an inspection of the steam generator tube frcm the point of entry "(hot leg side).corapletely around the U-bend to the. top support of the cold 109 The steam generator shall be dctcrained OPERACLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.4-2.
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4.4.5.5 Reports Following each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Cc.T. mission within 15 days.
The complete results o' the steam generator tube inservice f
inspection shall be included in the Annual Operating Report for the period in which this inspection was completed.
This report shall include:
Number and extent of tubes inspected.
Location and percent of wall-thickness penetration for each indication of an imperfection.
Identification of tubes plugged.
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Results of steam generator tube inspections which fall into Category C-3 and require prcmpt notification of the Commission shall be reported pursuant to Specification 6.9.1 prior to resumption of plant operation.
Tne written follouup of this 7
report shall provide a description of investigations conductco to determine cause of tha tube degradation and corrective r.easures taken to prevent recurrence.
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TABLE 4.4-1 C
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MINIMUM NUMBER OF STEAM GENERATORS TO BE
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INSPECTED DURING INSERVICE INSPECTION T
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w No Yes Preservice inspection Two Three Four Two Three Four No. of Steam Generators per Unit All One Two Two First inservice Inspection l
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3 One One Onc One Second & Subsequent inservice Inspections i
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Table Notation:
- 1. The inservice inspection may be limited to one steam generator on a rotating schedule encompass o
(where N is the number of steam generators in the plant) if the results of.the first or previous inspect Note that under some circumstances, the operating conditions in all steam generators are performing in a like manner.
one or more steam generators may be found to be more severe than those in other steam generators.
stances the sampla sequence shall be modified to inspect the most severe conditions.
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- 2. The otNr steam generator not inspected during the first inservice inspection shall be inspected. The YG.?I Q:J:1 I-6 -5 ins; cctions should follow the instructions described in 1 above.
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- 3. Each of the other two steam generators not inspected during the first inservice inspections shall be in C.
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-.i second and third inspections. The fourth and subseque1t inspections shall follow the instructions desc
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PROOF & P.EVEW CO -Y TABLE 4.4-2 E
vs STEAM GENERATOR TUBE INSPECTION
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c-l 1ST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION p
h Sample Size Result Action Required Result Action Required Result Action Rcquired
- o A minimum of C-1 None N/A N/A N/A N/A e
S Tubes per
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N/A
-4 C-2 Plug defective tubes C-1 None N/A and inspect additional Plug defective tubes C-1 None
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w 25 tubes in this S. G.
C-2 and inspect additional C-2 Plurt defective tubes 4S tubes in this S. G.
O Perform action for
~C-3 C-3 result of first sample
' goj Perform action for h
C-3 C-3 result of first N/A N/A w
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sample 5
Q~ g C-3 inspect all tubes in All other this S. G., plug de-S. G.s are None N/A N/A ai fective tubes and C-1
,i *d impect 2S tubes in a n S. G.s Perform action for N/A N/A e
b cach other S. G-M C-2 but no C-2 result of second eild.tional sample Prompt notification S. G. are to NRC purmant C-3 to specification Adik tional inspect all tubes in 5
09I S. G. is C-3 cach S. G. ard plug gg iteiective tubes Prompt notification N/A N/A 4]
gq to NitC pursuant 25 b;
3 to specification
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N Where N is the number of steam generaturs in the unit, anil n is the number of steam generators inspected S=3.
n during on inspection fj cn U
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SAFETY CLASS I COMPONENTS - STRUCTURAL INTEGRITY Equipment FPC Inservice Inspection Program (l)
- 1) See Table 4.4.6 4.4.10.1.1 (copy attached)
(1) The FPC Surveillance Program requirement sections referenced below were contained in the " Proof 6 Review Copy" of the CR#3 STS received by FPC on March 30, 1976.
As this issue of the CR# 3 Technical Specifications was thought to be at that time the final " Proof 6 Review Copy", the surveillance requirements contained therein were used as the basis for FPC's surveillance program.
(Copies of the referenced sections are attached.)
4.'4.10.1.1 The following inspection program shall be performed during shutdown:
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Inservice Insocctions The structural intcarity of the Safety ffass f components shall be demon. trated by verifying their,
acceptability when inspected per the applicabic rcquirements of Section XI of the ASME Coiler and Pressurc Ve:scl Code, 1972, Edition and Winter 1972 Addenda, as outlined by the inspection program shown in Table 4.4-6.
An initial report of any abnormal degradation of the structural integrity of the Safety Class 1 components detected during the above required inspections shall be made within 10 days after detection and the detailed report shall be' submitted pursuant to Specification 6.9.1 within 90 days after completion of the surveillance requirements of this specification.
The Inservice Inspection Program shall be reviewed every 5 years to assure that the equipment, techniques and procedures being utilized are ' current and applicable.'
The results of these reviews shall be reported in Special Reports to the Coranission pursuant to Specification 6.9.2 within 90 days of completion.
, Inspections Followina'Recairs or Renlacements The structural integrity of the reactor coolant ' system shall be iemonstrated after completion of all repairs and/or replacements to the system by verifying the repairs and/or replacements meet the requirements of Section XI of the ASME Boiler and Pressure Vessel Code, 1972 Edition and Hinter 1972 Addenda.
When repairs and/or replacements are made which involve new strength welds on components greater than 2 inch diameter, the new i
welds shall receive a. surface and 100 percent volumetric examination and meet applicable code requirements.
When repairs and/or eplacements are made tihich involve new strength welds on components 2 inch diameter or smaller, the new welds shall receive a surface examination and meet applicable code requirements.
Inspections Followino System Ooeni,nq The structural integrity of the reactor coolant system shall be demonstrated af ter each closing by perfonning a leak test, with the system pressurized to at least 2205 psig, in accordance with Section XI of the ASME Doiler and Pressure Vessel Code,1972 Edition and Winter 1972 Addenda and the pressure-temperature limits of Specifica-tion 3.4.9.1.
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POSTONER AT10NA L f NSPECTIONS
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C1 ASS 1 SECT 10N A. REACTOR VE5SEl AND C1 OSURE HEAD e
Tentative Inspection Item Enemination Inspectica During During 10-year No.
Category Esam6 nation A rea Method 5.vea e Inte rval let e rval Remarke 1.1 A
Longitudinal and Volumetric Nome 5% of the length of The required esaminations may circumfe re ntial the circumferential be made at or near the end of shell welde la weldet 10% et the the 10. yea r inspection interval.
cars region.
length of the long.
An gnamination can be made from laudinal welds.
the (CD) Outende Surface osthout removing the vessel saternale.
When the longitudaral and circum.
ferential weld have received an esposure to neutron fluence in ex.
case of 1039 avt (E of I Mew or n
above). the length of weld la the high fluence region to be examined shall be increased to, at least 50 percent.
- 1. 2 5
Loesttudinal and Volumetric Noes 5% of the length of The required amount of weld circumferential st rcumfe rential lengths may be esamined at welde la shou welde; 107. of the or near the end of the 10. year (othe r than length of the lon.
Inspection inte tval. An esamination le those of Category situdieal welde, planned from the outside surface without removing ne weeset internase.
A and C).
Escluded is the weld joining the lowe r section of the bottom head to the upper
, section of the bottom head. This weld le inside the eupport skirt and examination le impractical due to extremely 1.mited access C
Ve s s el.to.flass e Volumetric t/3 of the weseel.
Cumulative 100%
Both of t5ese uelds are available and head te.
to. flange and 1/3 of each cLr..n..
for emn.indon during rermal g
flange circumfer.
of the head to.
ferential weld.
refueling ope rations. Eithe r eettal welde, flange circumfe r.
mechanised or =anual ultrasonic estial weld.
techniques will b..
ased.
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Primary noaste to. Volarnetric Nos sle.to.ehell Cumulativi 100%
lt le planned that these componente vessel welds and
' weld and lase r of each notale.
be esamined from the (ODI Outside no s ele.to-ve s e el redina exami.
to.shell weld and Surface. Permanent trache will be teside radiused nations os one sousie taner ettsched te the pipe immediately sec tione talet nosale.
radius.
adjacent to each naasle forging.
ene outlet nos.
ale. and one core flood son.
ale.
- 1. 5 E1 Veeest penstra.
Volumetric 10% of the control 25% of the control The only penetrations in this tions, tecluding rod drive pene.
rod drive p.netration category are the control rod control rod drive tretion welds and welde and 25T. of the drive tubes which are welded penetratione and 10% of the control control rod housing to the upper head with a partial control rod rod housing pree.
pressure boundary penetratien weld. Emaniination housing pressure sure boundary welde will be volume. will be made from the penetration boundary welde.
welde will be trically examined.
ID when drive mechanisme are volumetrically removed for maintenance.
e marnise d.
- 1. 6 E.2 Ve.- el penetra.
Visual None Cumulative, 25% of The vteual esamination for leakage tiene, escluelve the in. core instru.
of the lower head penetratione can of control rod mentation pene.
be made wsthout remmal of trattens will be insulation. The enanunatina will be delve penetra.
ttone and control visually emarrined performeet at or near the end af the rod housing pree.
for leakage by the interval at the same time as the end of the interval.
system hydrostatic test of 15.!20.
sure boundary welde.
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penetrations or control rod Mrasing C) c at e g o ry.
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pressure boundary welds in this L
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Cl. ASS l, I
SECTIO *4 A. REACIOR VESSEL AND CLOSURE HEAD Tentative Ins pection item Esaminatsoo Inspection During During 10-yea r Rema rk s No.
Catezery Esaminatten A rea M ethod
$.vea r Inte rval lat e rval L7 T
Primary measles Yleual and The dissimilar The dissimilar metal OcJy the core nood naastes to safe end surface and metal weld on welde on both of the have dissimilar metal welds, welde.
volumet ric.
one core flood core Coc4 noaales.
The dissimilar metal welde nossle.
of each nosate will be eaam6ned at the same time as the aosale.to.sbell weld.
L8 C.!
Closure stude Volumet ric Cumulative 50%
Cumulative 100% with and aute.
and visual with approzamately approximately 10%
or surface 80'. being esam.
being esamined at ined at each each refueling.
re fuellag.
L9 C.1 1.lgaments between Volumet ric I/3 of the weseel Cumulative 100% of the The lignments wi!! be examined t'.readed stud flange liga.
vessel flange liga during the same outage as the hole s, meets between mente between '
threaded stud threaded stud holes will be holes will be examined.
esamloed.
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C.I Closure essbe re, Visual Cumulative 50%
Cumulative lot,%
beshings.
with approxi.
o pprorsmately 10%
mately 10%
being eaamined at being eaamined each refuellag.
at each refueling.
Iell C.2 Pressure retain.
, Visual None None There is no botting less thaa lag botting.
2 inches in diamete r.
.d '
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!ategrally welded Volumet ric Name None The support le not welded to vessel supporte.
the vessel pressure boundary.
The vessel shirt is integrally forged to the lower bead.
1.13 11 Closure head Visual and Two 6 X 6.la.
Cumulative. 6 cladding.
surface or patches.
patches (each 6 volumet ric X 6.inJ evenly distributed in the i
vessel head will be eaamined.
1.14 11 Vessel cladding.
Yleual Two 6 X 6-in.
Cumulative, 6 patc he s.
patches (each 6 X 6.f a.) evenly distributed is the vessel shell.
i I,15 N
Interior surfaces Visual A critical esam.
De examinatione it le planned to visually esamine and inte rnata laation will be made at the 4th by remote viewing devices. There are and lateg rally made of the in.
refueling cycle will ao integrally welded internal supporte.
welded internal terior surfaces be repeated at the suppo rt e.
and the internal 7th and 10th re!9el.
c omponent e ing cycle Due to made available limited access by normal re.
below the reactor
'O fueling opera.
core during normal j
'J.
- d tiene at the let refueling outages.
i refueling cycle.
the space below the r-p This wtll be re.
reactur core will I
peated at the 4th be esamined at or k
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a refueling cycle uth near the end of the the amount of the ins pection inte rval.
eaamination being dependent upon re.
eults of the let eu.am.
laation and that made on othe r pres surised.
w.te r systeme.
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t C1.A551 SECTION fl. PR E55URJ ZER f
Tentattve laepeceton Ite st Esaminaties Inspection During During 10. year N-Cateroer E.m amination A ree M e t h od S.v a r Inte evat late rv31 Remarke 2.8 3
Longitudinal and Visual and 5% of the length Ey the end of the circ umfe rential volumet ric of the circum.
- 10. yea r pe riod, welde.
ential welde 10% of the length joining the lower of each longitudinal head and of the weld and 5% of the
~ ci rc umfe rential length of each cit.
weld joinang t6e cumferential weld upper head to the will be examined.
barrel section will be esamined.
10% of the length of the long6tudinal welde in the upper and lower shelle will be examined.
- 2. 2 D
Nes sle.co.ves sel Volumet ric The surge time All of the mossle.to.
welde.
mos ale.to.o hell shell welde and laner weld and 6aner radit will be examined.
radius and a relief line aos.
ale.to.shell weld and inner radius will be exa mlaed.
- 2. 3 E.1 Hester comaec.
Yleual and None Nemo There are no category E.1 g
tiene.
surface beater connectione la this Pres.
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suriser and all of the penetratlose 8
et this Preeeuriset meet the eseluanos c rite rim of 15121.
- 2. 4 E.1 Heater connes.
Visual None A cumulative totalof The esamination will be pe rformed at or tiene.
25*. of the heater near the end of the interval at the connections and of same time as the system hydrostatic the 6netrument and test of 15 520.
. sample penetratsoos will be visually esamloed for leak.
age by the end of the late rval.
- 2. 5 C-3 Preneure retato.
Visual and Cumulative 50% by Cumulative 100% by Botting 2 laches and larger tag botting.
volumet ric end of 5. year end of Laterval.
la diameter will be examined late rval.
either in place under tension, or when the botting la removed
'er when the bolting ceanection to disassembled. The bolting to be esamined will include stude and nute. The ligament areas between threaded stud holes will be esamined f rom the face of the flange when the connection is dienseembled.
- 1. 6 C.2 Pressure totala.
Yleual Nome None There to ao pressure retaining lag bottlag.
bottleg less than 2 inches la diame t e r.
- 2. 7 H
Integrally welded Visual and The total length The total length of vessel supporte.
volumet ric of two lug attach.
of all lug attachmen 0
H R 'D ment welde will welde enll be
.);
be examined.
e sa mine d.
i 9
!s 12 Vessel cladd6ag.
Yleual None A 6 X 6.in. patch rih f
I of the Pressuriser L '
- K j
-U cladding near the man =ay will be esamined.
s
1
_ CLASS I e
SECTION C. STEAM CENERATORS J
Te' iative Inepection e
n Itzm Examination inspection During During 10. year N e.
Catenary Examination Area M et h od 5 vea r 1r.te rval Int e rval R ema rk e 3.1 B
Longitudinal and Visual and 5% of the lesagth By the end of the Volumetric examination et sne circumfe re ntial volumet ric of the circum.
interval. 5% of the steam gene rator lowe r head.to.
welds. including ferential weld length of the welde tube ehert weld 6e not feasible tube sheet.to. head joining the upper joining the uppe r by ultrasonic e, and esamination er shell welds on prirna ry head to prima ry heads to of the base metal for one plate the primary side.
the tube sheet the tube sheets thicknes s beyond the edge of the of one steam will be examined.
welds which can be esamined can generator will be only partly covered due to be examined.
the geometry of o r steam gen.
eratore. It le not known if an esamination can be made by radio.
g ra phy.
3,2 D
Primary noaale.
Volumet ric The primary naa.
The prima ry nom.
to. head welds ala.to-head welde sle.to. head welds and moaale.to.
and related inner and related inner heed inside radit gf one steam radii of both steam radiesed sections.
gene rator wsil be generators wsil be e xamined.
esamined.
- 3. 3 F
Primary noeste.
Visual and None None There are no diestmilar metal to. safe end surface and welde in the primary nosales weld e, volumet ric.
of the steam generators.
3d C1 Pres sure retain.
Visual a nd Cumulative t0% by Cumulative 100% by Boltieg 2 inches and la rger ing botting.
volumet ric end of 5. year god of interval, in diameter will be examined inte rval.
either in place under tension.
or when the botting is removed or when the bolting connection is g
disa s s emble d.
The botting to be 1
examined will enclude stude and nute.
~
The ligament a reas between threaded stud holes will be examined from the face of the flange when the connection to disas sembled.
3.5 C.2 Pressure retala.
Visual Comalative Sc% by
- Cumulative 10C% by The bolting below 2 lache e in tag botting.
und n!5. year and of interval.
diamete r will be visually examined.
inte rval.
. either in place if the bolted connection le not disassembled during the inspec.
tion interval, or whenever the botting connectaon is disassembled. The boltang to be examined willinclude stude and mute.
Excluded from examination is botting of a single connection whose failure results in conditions that satisfy the e.aclusion criteria of 15121.
- 3. 6 H
lategrally welded visual and 10% of the length 10% of the length veseal supporte, volumetric of the skirt cir.
of the skirt cir.
cumferential cumferential weld weld on one on each steam steam gen.
generator.
erator.
- 3. 7 12 Vessel cladding.
Vleual None A 6 X 6.in. patch c
D
]D D
near the manway in each of the two i,
primary heads of g
each generator mil be examined
~ T f
'I h by the end of the e
f
,n.,
rg ine,ection inier.
~
val.
7 S Ze*
- A M
e CLASS 1 SECTION D. PIPINC PR ESSURE BOUNDARY _
Tentative Ins pection item Examtmation laspection During During 10. year H +.
C+terocy Examination Area M et hod
- 5. year Interval Interval Rema rk e 4.1 F
Vessel, pump, and Visual and Cumulative 50%
Cumulative 100% of The Pressurise r nosele.to-eale sad valve safe ende-to. surface and of the dissimilar the dissimila r metal welds are included la this item.
primary pipe welde volumetric metal welds will welds will be exam.
and safe ende la he eaamined.
ined by the end of bras th piping welde.
the intervah 4.2 01 Pressure rotata.
Visual and Nome None There le no bolting 2 inches tag botting.
volumet ric and larger in the piping system.
- 4. 3 C.2 Pressure rotata.
YLoual Cumulative 50%
Cumulative 100% by All bolting is below 2 inches la ing bolting.
of the botting by the end of the diameter and will be visually the end of 5. year inte rval.
esamined, either in place if the late rval.
bolted connectica is not die.
assembled during the u.o pection interval, or whenever the bolted connection le disassembled. The botting to be esamined wsil include stude and nute.
Excluded from eammination is botting of a single connection whose failure results in cond.tione that 4
satisfy the esclusion criteria of 15 121.
i
- 4. 4 J1 Circumf e rential visual and Cumulative 10% of Cumulattve 25% of the and longitudinal volumet ric the butt weide, in.
butt welde in th e pip.
pipe welde.
cluding one foot of ing system, includ.
any longitudinal ing one foot of any
' weld on either
' longitudinal weld os side of the butt either side of the weld. will be butt welds, will be examined.
examined by the end of the interval.
Visual and Cumulative 0% of Cumulative 25% of the
- 4. 5 J1 Branch pipe '
volumetric the branch pipe brancie pipe connection aonnection welds, enceed.
connection.
welded jointe.
Lag 4.ia, nom.
welded jotate.
3 taal pipe else.
- 4. 6 J.1 Socket welde Visual and Cumulative 10% of Cumulative 25% of the 0
eu rface the socket welded socket welded joints, l./
U J @ ' 3' joints.
9 D {
q f '.* * ' M k -
I RI' O f LQ -
U U
- 4. 7 J.1 Branch pipe Visual and Cumulative 10% of Cumalative 25% of the Jd!1 d
connection surface the branch pipe branch pipe connection weld e. 4.. a.
connection welded welded joints.
sominal pipe joints.
else and smaller.
o.
- 4. 8 J.2 Circumfe re ntial Vieval Cumulative 50% of Cumulative 100% of a.maminatione shall be performed when.
and longitudinal the welds in piping the welds in piping ever the system boundary is subjected j
pipe welds and (including pipe (including pipe,
to a hydrostatic test prior to earti branch pipe coa.
branch connec.
branch connet.
plant sta rt-up followsag refue! tag and section welds tione) whch are tienen which are at the time of the eya'em hydrostatic escluded f rom eacloded f ram test an' required by 15 520.
l the esaminations the examinatione by 15 121.
by 15 121.
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C1.A551 SECTIO *t D. P!PINC PRESSURE BOUNDARY o
Tentative ins pection 3'.samination inspection During During 10-year Nm Cate re er Eneminatten Area M ethod 5.ves e Inte rval Inte rval R ema rk s Item
- 4. 9 K.1 lategrally welded Visual and Cumulative 10T.
Cumulative 25% of Escluded from examination supporte.
volumetric of the total num.
the total number are supports whose f ailure results ber of integrally of integ rally welded in conditions that sattafy the exclusion c rite ria of 15 121.
welded supporte.
supporta.
4.10 K.2 Piping aupporto Visual 50% of the sup.
Cumulative 100'. of The support members and structures had bangere ports and hangere the supports and subject to examination mill include will be examined.
hangers will be those supports within the system examined by the whose structural notegrity se end of the interval.
relied upon to withstand the design loads and seismic. induced displacemente.
The support settings of constaat and variable spring-type hangere, saubbers, and shock absorbere will be esamined to verify proper distribution of design loads among the associated support composeats.
SECTION E. PUMP PR ESSURE BOUNDARY S.1 L.1 Pump casing visual and None.
If one of the pumps The only feasible method knowa le disassembled foe to date to volumetrically examine welde volumet ric maintenance by the these pump casing welds le end of the interval.
radiog ra phy. It le not known the available welds if such radiography can be i
will be examined, performed in se rvice due to the 11 possible.
design and the radiation level in the component. Il expe rie nc e or a study indicates such radio.
raphy is possible, the esamination will be performed at the frequency LadicateJ.
~
S.2 L.2 Pump casinge visual None.
If one of the pumpe The only pumps involved ta this le dienseembled for program are the reactor coolant maintenance by the pumps.
and of the interval.
the available inner surfaces will be examined.
S. 3 T
Nos sle.to-Vleual and The dissimilar The dissimilar metal safe and welde volumetric metal welds on welds on each of the two reactor four reactor coolant I
coolant pumps pumps will be exam.
will be saam.
taed.
laed.
S. d C1 P re s s. ire Visual and Curnulative 50%
Cumulative 100% of Botting 2 inches and larger la rotatalag volumet ric of the bolting the bolting by the km diameter will be examined by the end of and of the interval.
either in place under tension, or when the bolting is removed botting the 5 year or when the bolting connection Late rval.
le diesesembled.
Th' * * 'a b* * = * *$ a'd "Lil
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include the stude and nuts. A etidy will be pe istmed print to the time of the preoperational s
h esami satioa to dete rmi.e the feasileilitt of uitraso sic esambations
/d.
. ; ; d o,. -)
of the ligament areas between u
threaded stJd hales. If ultra sonic enarmt. stions are determined to be feasible. the esaminations will be performed frem the face of the flange when the connection to dioaaeembled.
g.
- m
W*
a CLASS I
- t. M '
$ECT10N E. PUMP PR ESSURE BOUND ARY e
Testative Ins pection Item Examination Inspection During
,During 10. year Caterere Enamination Area M ethod 5-vee r Inte rvs1 Int e rval P ema rk e
- 5. 5 C.2 Pressure Visual Nome.
-None.
There is no bolting below 2 inches retalaims la diameter.
botting S.6 K1 Integrally Visual and Nome.
None.
The pumps contain no lategrally welded supports volumetric welded supporte.
S.7 K.2 Supports and Visual Cumulative 50%
Cumulative 100'". of The support members and structure bangers of the reactor the reactor coolant subject to ezamination will include coolmat pump pump supports will those supports within the system supports will be examined.
whose structural integrity is relied be eaamined.
upon to withstand the design loads and seismic. induced displacements.
The support settings of constant and va riable spring-type hangere, enubbers, and shock absorbers will be esamined to verify proper diatribution of design loads among the associated support composeats.
SECTION T. VALVE PRESSURE BOUNDARY s
M.1 Valve body Visual and None.
None.
'lliere are no valves with pressure welde volumetric retaining welds in the valve bodies.
6.2 M.2 Valve bodies Visual None.
Cumulative 100%
Examinations will be conducted of the valves re.
provided valves can be disassembled quired by the Code without undue radiation espesure to will be examined persoonel.
by the end of the laspection later.
val
- 4. 3 r
Valve.to. safe Visual and None.
None.
There are no velves in this end welde, valuenetric system with disesmilar metal welds as defined by cetegory T of the Code.
- 6. 4 C.1 Pressure retale.
Visual and None.
None.
There are no valves with botting iag botting.
volumet ric 2 inches and larger in diameter.
6.5 C.2 Preseure rotata.
Visual Cumulative 50%
Cumulative 100%
All botting is below 2 inches la tag botting.
of the botting of the botting will diameter and will be visually examined, will be examined be examined by the either in place, if the botting connection by the end of the end of the intervab to not disassembled during the inspection S. year interval.
laterval. or whenever the botting con.
mection is dseassembled. The botting to be examined will tactude stude and aute.
Excluded from examination to botting of a single connection whose failure results in cohditions that satisfy the cellerton, Pa regraph 15121.
6.6 K.1 Integ rally Vis ual and None.
None.
There a re no valves with integrally welded supporte volumet ric welded supporte.
[
yl ;
.-a s
C1. ASS.I _
SECTION P. VALVE PRES 5URE BOUND ARY Tentative Inspection Esarnisation Inspectiva During During 10-year It:rn Examination Area Method 5-yea r laterval Inte rval Rema rk s N '.
Catetery
- 6. 7 K-2 Supports and Visual Cumulative 50". of Cumulative 100'. et The support members and structures hangers the supports and the supports and subject to examination will include hangers will be hangers will be those supports within the system examined by the examined by the e..d whose structural integ rity is and of the 5-year of the interval.
relied upon to withstand the design loads and eeismic. induced Late rval.
displacements.
The support settings of constant and variable spring. type hangers.
emubbe rs, and shock absorbers will be esamined to verify proper distribution of design loads among the associated e apport components.
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EMERGENCY CORE COOLING SYSTEMS G
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Equipment FPC' Surveillance' Program (1)
- 1) Core Flooding Tanks 4. 5.1.
a) CFV-1, 2, 3, 4
- 4. 5. 2 '
a) DHP-1A, 1B Flow Path -(Valves) - Operating 4.5.2.....
a) CAV-57, 60 b) DHV-5, 6, 11, 12, 34, 35, 75, 76, 110, 111 c) MUV-23, 24, 25, 26, 31, 58, 73, 103, 108 ECCS Subsystem (Electrical Power) 4.5.2
~
Flow Path (Valves) - Shutdown 4.5.2 ~
a) DHV-42, 43 b) MUV-27, 64, 90, 91, 96, 97 t
1 (1) The FPC Surveillance Program requirement sections referenced below were contained in the " Proof 6 keview Copy" of the g
CRt3 STS received by FPC on March 30, 1976.
As this issue of the CR#3 Technical Specifications was thought to be at that time.the final " Proof 6 Review Copy", the surveillance require-ments contained therein were used as the basis fc-FPC's sur-veillance program.
(Copies of the referenced secticas are attached.)
l
4.5.1 Each core flooding tank shall be demonstrated OPERABLE:
Demonstrating check valve operation at least once per 18 months.
4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:
At 'least once per 31 days on a STAGGERED TEST BASIS by:
Verifying that each HPI pump:
Starts (unless already operating) from the control room.
Develops a discharge pressure of y_1500 psig on recirculation flow.
Operates for at least 15 minutes.
Verifying that each LPI pump:
- )
Starts (unless already operating) from the control room.
1 Develops a discharge pressure of 3,200 psig on recirculation ficw.
s-Operates for at least 15 minutes.
Cycling each testable power operated or automatic valve in the fica path througn at least one complete cycle of full travel.
Verifying that each ECCS subsystem is-aligned to receive electrical power from separate OPEP.ABLE emergency busses.
At least once per 18 months, during shutdewn, by:
Cycling each power operated (excluding automatic) valve in the flow path that is not testable during plant operation through at least one ccmplete cycle of full travel, and
=
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CONTAINMENT SYSTEMS t
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9
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4 DEPRESSURIZATION AND COOLING SYSTEMS
(
Equipment FPC Surveillance Procr 1 (1)
- 1) Containment Spray System (Pumps) 4.6.2.1 a) BSP-1A, 1B Containment Spray System (Valves) - 4. 6. 2.1 Flow Path - Operating a) BSV-3, 4, 16, 17
~ Containment 3 pray System (Valves) - 4. 6~. 2.1 Flow Path - Shutdown a) DHV-42, 43 b) MUV-27, 64, 90, 91, 96, 97
- 2) Spray Additive System (Valves) -
4. 6. 2. 2. '
Flow Path - Operating a) DHV-34, 35 b) BSV-3, 4, 11, 12, 16,.17, 36, 37
~
Spray Additive System (Valves) -
4.6.2.2 Flow Path - Shutdown a) All power operated valves
(
in this system are tes able during plant operation.
Y (1) The FPC Surveillance Program requirement sections referenced below were contained in the " Proof 6 Review Copy" of the CR#3 STS received by FPC on March 30, 1976.
As thi.s issue of the CR#3 Technical Specifications was thought to'be at that time the final " Proof G Review Copy'i, the surveillance require-ments contained therein were used as the basis for FPC's sur-veillance program.
(Copies of the referenced sections are attached.)
1y
. -.. ~
~ ~ -.--
o
' - - ' ' ' ~ ~ - - - - - ' ~
. i..
n 4.6.2.1 Each containment spray system shall be demonstrated OPERABLE:
At least once per 31 days on a STAGGERED TEST BASIS by:
Starting each spray pump from the control room, Verifying, that on recirculation flow, each spray pump develops a discharge pressure of > 190 psig at a flow of
..I
> 1500 gpm, Verifying that each spray pump operates for at least 15
- minutes, I
Verifying that each spray pump is aligned to receive elec,trical power from separate OPERABLE emergency busses.
Cycling each testable power operated or automatic valve in the flow path through at least one complete cycle of full travel, At least once per 18 months, during shutdown, by:
Cycling each power operated (excluding automatic) valve in the flow path that is not testable during plant operation through at least one complete cycle of full
- travel, and 3
4.6.2.2.The spray additive system shall be demonstrated OPERABLE:
jj At least once per 31 days by;
~
Cycli y each testable power operated or automatic valve
~
in the flow path through at least one compleN cycle of full travel, and I' ~
~~
t lea:t once per 18 months ~-during shutdown, 69f
~
Cycling each power operated (excluding automatic) valve in the flow path that is not testable during plant operation through at least one complete cycle of full
- travel, and Verifying that each automatic valve in the flow path actuates to its ccrrect position on a containment spray signal.
-*
WSV-6 M,
B.
CONTAINMENT PURGE AND EXHAUST 1.
AHV-1C & 1D iso. pur. sup. system 60 AHV-1B & 1A iso. pur, exhaust system 60 t
C.
MANUAL w
1.
NA IAV-29 e
2.
.LRV-50 iso leak rate test system NA b
LRV-36 from RB NA LRV-51 iso. atmos. vent and RB NA NA LRV-35 & 47 purge exhaust system
'r from RB h.;
LRV-49 iso. atmos, vent from RB NA
{
LRV-38 & 52 u,
NA l
LRV-45 iso. LR test panel from RB NA NA LRV-44 3.
MSV-146#
iso. misc. waste storage NA.
L tank from RCSG-1B i,.
Eb
! m 4.
NGV-62 iso. NG system from NA NGV-81 #
steam generators NA NA NGV-82 iso. NG system f, rom pzr.
CRYSTAL RIVER - UNIT 3 3/4 6-20 8
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TABLE 3.6-1 (Continued)
}
I CONTAINMENT ISOLATION VALVES F,3
[&.;E VALVE NUMBER FUNCTION ISOLATION TIME i.
(seconds) 5.
SAV-24 iso. SA from RB NA NA 6.
SFV-18 iso. SF system NA
.SFV-19
$f;.
s b.
SFV-119#
iso. Fuel Transfer tubes NA 3;e-SFV-120#
from F.T. Canal
~
NA
,p 7.
WSV-1
' containment monitoring NA NA
- i.
WSV-2 system from RB D.
PENETRATIONS REQUIRING TYPE B TESTS Blind Flange 119 iso. RB NA
$7 NA sm- -
Blind Flange 120 NA Blind Flange 116 NA Biind Flange 202 Blind Flange 348 iso, fuel transfer tube from NA Blind Flange 436 Tror.sfer Canal NA M
Equipment Hatch iso. RB NA is ;
Personnel Hatch iso. RB NA 7~r ;
I r..
- Not subject to Type C Leakage Test P
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1 CRYSTAL RIVER - UNIT 3 3/4 6-21 1
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k PLANT SYSTEMS t
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,r TURBINE CYCLE Equipment FPC Surveillance Pro' gram '(1)
- 1) Safety Valves 4.7.1.1 a) MSV-33, 34, 35, 36, 37, 38, 39, 40, 41, 42, 43, 44, 45, 46, 47, 48
- 2) Auxilia ry Feedwater Sys tem (Rmps) 4. 7.1. 2
- - ~ ~ __
a) EFP-1, 2
~
~ -'
-1 Auxiliary Feedwa ter System (Valves)4. 7.1.2. - -
-Flow Path - Operating a) EFV-1, 2,
3, 4,
11, 14, 32, 33, 161, 162
, Auxiliary Feedwater System 4.7.1.2.~
(Electrical Power)
Auxiliary Feedwater System OTalves)4. 7.1. 2
-Flow Path - Shutdown a) All the power operated valves for this system are testable during plant operation.
- 3) Main Steam Line Isolation 4.7.1.5 Valves a) MSV-411, 412, 413, 414 8
(1) The FPC Surveillance Program requirement sections referenced below were contained in the " Proof 5 Review Copy" of the
' CRf3 STS received by FPC on March 30, 1976.
As this issue of the CR#3 Technical Specifications was thought to be at.that time the final " Proof 5 Review Copy", the surveillance require-ments contained therein were used as the basis for FPC's sur-veillance program.
(Copies of the referenced sections are attached.)
/'N.
/'
l 4.7.1.1 Each main steam line code safety valve shall be demonstrated OPERABLE, with lift settings and orifice sizes as shown in Table 4.7-1, in accordance with Section XI cf the ASME Boiler and Pressure Vessel Code, 1974 Edition.
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PROOF L
.EVIEW COPY TAB E 4.7-i STEAM LINE SAFETY VALVES c.,
E 1%) (psigl ORIFICESIZE(inches)
M VALVE NUMBER LIFT SETTING (
E STEAM GENERATOR 3A n3 9
Main steam line Al b
MSV - 34 1050 4.515 i5 MSV - 38 1070 4.51 5 MSV - 43 1090 4.515 MSV - 40 1100 3.750
~
Main steam line A2 MSV - 33 1050 4.51 5 MSV
.37 1070 4.515 MSV - 42 1090 4.515 R MSV - 46 1100 4.515
~
? STEAM GENERATOR 3B W
~
Main steam line B1 MSV - 35 1050 4.515 MSV - 39 1070 4.51 5 Y4E?)MSV - 44 1090 4.51 5 sc L,.1!MSV - 47 1100 4.515 b
g Main steam line B2 a
.o m
g
( MSV - 36 1050 4.515 g
MSV - 41 1070 4.515 MSV - 45 1090 4.515 -
C MSV - 48 1100 3.750 r
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Each auxiliary feedwater. system shall be demonstrated OPERABLE:
F 4.7.1.2 Il At least once per 31 days on 'a STAGGERED TEST BASIS by:
(~
Starting each pump from the control room, t
Verifying that:
The motor driven pump develops a discharge pressure.
of > 1100 psig on rectrculation flou, and Each pump operates for at least 15 minutes.
Cycling cach testable power operated or automatic valve -
in the flow path through at least once complete cycle of full travel.
Verifying that each system is aligned to receive electrical
, power from seperate OPERABLE emergency busses.
At least once por 18 months, during shutdown, by:
power operated (excluding automat.ic) valve Cycling each in the flow path that is not testable during plant operation through at.least once complete cycle of full travel, and Each main steam line isolation. valve that is open shall be 4.7.1.5 demonstrated OPERABLE by performance of the following surveillance requirements:
Part-stroke excercising the valve at least once per.92 days, and Verifying full closure within 5 seconds on any closure act-515*F during uation signal while in HOT STAfIDGY with T eachreactorshutdown,exceptthatverifiEXEionoffullclosure within 5 seconds need not be dctormined more often than once every 92 day's.
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lr CLOSED CYCLE COOLING WATER SYSTEMS
(
/'
1 FPC Surveillance Program (1)
Equipment
- 1) Nuclear Services (Pumps) 4.7.3.1 3
a) SWP-1A, 1B, 1C Nuclear Services Valves -
4.7.3.1 Flow Path - Operating a) SWV-35, 37, 39, 41, 43, 45, 353, 354 Nuclear ServicesValves -
4.7.3.1 Flow Path - Shutdown a) All the power operated valves for this system are testable during plant operation
- 2) De, cay Heat (Pump) 4.7.3.2.
a) DCP-3A, 3B Decay Heat (Valves)-Flow Path-4.7.3.2 Operating a) DCV-17,_18, 177, 178 Deeay Heat (Valves.)-Flow Path-Shpwn
- 4. 7.3. 2 a) All the power Operated valves for this system are testable during plant operation l
(1) The FPC Surveillance Program requirement sections referenced' below were contained in the " Proof 6 Review Copy" of the CR#3 STS received by FPC on March 30, 1976.
As this issue of the CR#3 Technical Specifications was thought to be at that time the final " Proof 6 Review Copy", the surveillance require-ments contained therein were used as the basis for FPC's sur-veillance program.
(Copies of the referenced sections are
(
attached.)
_e 4.7.3.1 The nuclear services closed cycle cooling system shall be
?
demonstrated OPERABLE:
~'
f.
At least once per 31 days by:
Starting each pump (unless already operating) from the Control room, Verifying that each pump develops at least 93% of the discharge pressure for the applicable flow rate as deter-mined from the manufacturer's Pump Performance Curve.
Verifying that each pump operates for at least 15 minutes, Verifying that each pump is aligned to receive electrical power from separate OPERABLE emergency busses,
- (,
Cycling each testabic power operated or automatic valve servicing safety related equipment tifr%TM&OrltW,0,0.0c complete cycle of full travel, and At least once per 18 months, during shutdown, by:
Cycling each power operated (excluding automatic) valve servicing safety related equipment that is not testable during. plant operation through at least one complete cycle of full travel 4.7.3.2 Each decay heat closed cycle cooling water loop shall be demon-i strated OPERABLE:
At least once per 31 days on a STAGGERED TEST BASIS by:
Starting each pump (unless already operating) from the control.roo'm, Verifying that each pump develops at least 93% of the discharge pressure for the applicable flow rate as deter-min'ed from the manufacturer's Pump Performance Curve, Verifying that each pump operates for at least 15 minutes, Verifying that each loop is aligned to receive electrical power from separate OPERABLE emergency busses, Cycling (ach testable power operated or automatic valve I
. servicing safety related equipment through at least one complete cycle of full travel. -
. At.least once per 18 months, during shutdown, by:
Cycling each power operated (excluding au.tomatic) valve
(
servicing safety related equipment that is not testable-during plant operation through at least one complete cycle of full travel 9
r
SEA WATER SYSTEMS
~
r Equipment FPC Surveillance Progran '(1)
- 1) Nuclear Services 4.7.4.1 a) RWP-2A, 2B
- 2) Decay Heat 4.7.4.2
.,i a) RWP-3A, 3B (1) The FPC Surveillance Program requirement sections referenced below were contained in the " Proof 6 Review Copy" of the CRf3 STS received by FPC on March 30, 1976.
As this issue of the CR#3 Technical Specifications was thought to'be at that i
time the final " Proof 4 Review Copy", the surveillance require-
)
ments contained therein were used as the basis for FPC's sur-
~
veillance program.
(Copies of the referenced sections are attached.)
.m a
.e wwm e e
am m e--
4
j i
4.7.4.1 The nuclear service sea water systems shall be demonstrated OPERABLE:
At least once per 31 days on a STAGGERED TEST BASIS by:
' Starting each pump (unless already operating) from the Control room, Verifying that each. pump develops at least 93% of the shutoff head as determined from the manufacturer's Pump Performance Curve.
Verifying that each pump operates for at 1, east 15 minutes, Verifying that each pump is aligned to receive electrical power from separate OPERABLE emergency busses, 4.7.4.2 Each decay heat sea water loop shall be demonstrated OPERABLE:
At least once per.31 days on a STAGGERED TEST BASIS by:
Starting each pump (unless already operating) 'from the
(
control room, Verifying that each pmp develops at least 93% of the I
shutoff head as determined from the manufacturer's Pump Performance Curve.
Verifying that each pump operates for at least 15 minutes, Verifying that each loop is aligned to receive electrical power from separate OPERABLE emergency busses, e
4 O
4 1
e
=
l l
(
- 1) Hydraulic Snubbers listed 4. 7'. 9.1 in Table 3.7-3 (attached) d t
a (2)
The FPC Surveillance Program requirement sections referenced below are contained in the December 3, 1976 issue of the CR#3 Technical Specifications with the exception to the reference to Section 4.0.5.
(Copies of the referenced sections are attached.)
I.
t==
\\
~
7 4.7.9.1 Hydraulic snubbers will be demonstrated OPERABLE by performance of the following augmented inservice inspection program 6
Each hydraulic snubier with seal material fabricated from
~
ethylene propylene or other materials demonstrated compatible with the operating environment and approved as such by the NRC, shall be determined OPERABLE at least once after not less than 4 months but within 6 months of initial criticality and in accordance with the inspection schedule of Table 4.7-4 thereafter, by a visual inspection of the snubber. Visual inspections of
~
the snubbers shall include, but are not necessarily limited to, inspection of the hydraulic fluid reservoirs, fluid connections, and linkage connections to the piping and anchore.
Initiation of the Table 4.7-4 inspection schedule shall be made assuming the unit was previously at the 6 month inspection e
interval.
Each hydraulic snubber with' seal m&Lerial not fabricated from
~ '
ethylene propylene or other materials demonstrated compatible with the operating environment shall be determined OPERABLE at least once per 31 days by a visual inspection of the srJbber. Visual inspections of the snubbers shall include, but are not necessarily limited to, inspection of the hydraulic fluid reservoirs, fluid connections, and linkage connections to the piping and anchors.
t At least once per 18 months during shutdown a representative sample of at least 10 hydraulic snubbers or at least 10% of all snubbers listed in Table 3.7-3, whichever is less, shall be selected and functionally tested to verify correct piston movement, lock up and bleed. Snubbere greater than 50,000 lbs capacity may be excluded from functional testing requirements.
Snubbers selected for functional testing shall be selected on a rotating basis.
Snubbers identified in Table 3.7-3 as either "Especially Difficult to Remove" or in "High Radiation Zones" may be exempted from functional testing provided these snubbers were demonstrated OPERABLE during previous functional tests. Snubbers found inoperable during functional testing shall be restored to OPERABLE status prior to retuming opera-tion. For each snubber found. inoperable during these func-j.,
tional tests, an additional minimum of 10% of all snubbers or j
10 snubbers, whichever is less, shall also be functionally testefd until no more failures are found or all snubbers have been functionally tested.
g ei. *
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.g
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m
E
~
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Y 9
1/.
E TABLE 3.7-3
's
-l, g
SAFETY RELATED HYDRAULIC SNUBBERS *
=
T '.
l3..
g SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT
[1
[
-INACCESSIBLE ZONE ****
TO REMOVE
- I NO.
ON, LOCATION ** AND ELEVATION (A or I)
(Yes or No)
(Yes or No)
[.
Core Flood System A.
4 CFH-12 116'-7" A
No Yes Y. i.
CFH-13 111'-4" A
No No r-CFH-14 110'-6" A
No No I
CFH-15 120'-9" I
Yes Yes 4e.5,,
I Yes No 3
CFH-16 120'-9" I
Yes No 120'-9"
[
CFH-17
^
7.!
CFil-18 120'-9" I
Yes No
-[.)
w CFil-19 123'-0" I
Yes Yes
- q. -
1'i.
Decay Heat Removal System hV DHil-17 110'-6" A
No No DHH-18 110'-6" A
No.
No DHH-19 109'-0" A
No No
^. -
b.
DHH-20 115'-0" I
'Yes No DHH-21 117'-0" I
Yes Yes
./.."
DHH-22 117'-0" I
Yes Yes
+
DHit-23 110'-6" A
No Yes DHH-24 109'-3""
A No No
- ~
.N.
. DilH-25 110'-6" A
No Yes Dl01-26H 110'-6" A
No No
-g DHil-26V 110'-6" A
No No S
DHil-27 110'-6" A
No Yes L
'?p
, s.
6.
4 :.
I L'
.f
~."2 [' f EI ! - 4,e,f [ T'. ';. ',k
(.l
(.h I
[
..[,i@I
~~
.h E..
~
TABLE 3.7-3 r-SAFETY RELATED HYORAULIC SNUBBERS *
~
- o
.i 2
W
-lT.
SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH 'RADI ATION ESPECIALLY DIFFICULT g
NO.
ON, LOCATION ** AND ELEVATION INACCESSIBLE ZONE ****
TO REMOVE (A or I)
(Yes or No)
(Yes or No) g q
t Emergency Feedwater System EFH-140 115'-0" A
No No EFH-14L 115'-0" A
No No EFil-15U 115'-0" A
No No f
EFil-15L 115'-0" A
~
No No
~
EFil-27 145'-9" I
Yes Yes I
k EFH-28 145'-9" I
Ves Yes
[
EFH-92 140'-0"***
A No No y,
EFH-93 131'-8"***
A No No a
m EFH-94 131'-8"***
A No
.No 7
EFil-95 140'-0"***
A No No
- l:'
l EFH-96 141'-3"***
A No No EFH-106 141'-3"***
A, No No
.~
EFil-107 141'-3"***
A No No 4
EFH-108 141'-3"***
A No,
No EFH-109 133'-0"***
A No No A
No No V
EFH-110 133 ' -0" * * * '
A No No
?.Z EFH-141
'126'-6"***
EFH-143 133'-6"***
A-No No
" 4.'
EFil-144 141'-3"***
A No No Make-up and Purification System Mull-32 115'-11" I
Yes No y
)/
MUH-33 110'-6" A
No Yes K
MUH-34 110'-6" A
No No a
4c i
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.'kp.y QT
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t N'. !n I,
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TABLE 3.7-3 f::
C 6
SAFETY RELATED HYDRAULIC SNUBBERS
- r-i.
B 2
i g
v',
- =
SNUBBER SYSTEM SNL'BBER' INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT N0.
ON, LOCATION ** AND ELEVATION INACCESSIBLE ZONE ****
TO REMOVE ef;~
E E
(A or I)
(Yes or No)
(Yes or Ho)
I Q
Make-up and Purification System w
(Continued)
MUH-35 110'-6" I
Yes No MUH-36 109'-0" A
No Yes ef.'.!
MUH-37 109'-0" A
No No 3; i MUH-38 123'-10" I
Yes Yes f(
Mull-39 112'-0" I
Yes No w
I Yes No 2
Mull-40 112'-0" C.
Mull-41 129'-1" I
Yes No y
.I '.
A, Mull-42 119'-0" I
Yes No f3
=*
11U11-43 114'-0" I
Yes No Mull-44 114'-0" A
No No Jf,
Mull 114'-0" A
No No "d
Mull-46 114'-0" I
Yes No k.
Mull-47 114'-0" A
No Yes Yi Mull-48 114'-0" A
No
.. {
Mull-49 108'-6" A
No Yes Yes j*.,
Mull-50 129'-9" I
Yes Yes
'c Mull-51 117'-0" I
Yes Yes
'/.' '
Mull-52 110'-0" A
No No
/
Mull-53 110'-0" A
No No MUH-80 108'-0" I
Yes Yes Mull-81 108'-7" A
No No
..i Mull-82 108'-7" A
No No Y
Mull-83 108'-7" A
No No
-(
Mull-84 108'-7" A
No No E.
Mull-85 104'-0" A
No No 2.i q.
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Y
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^
(
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I'T 5'
)
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N TABLE 3.7'-3 N
SAFETY RELATED HYDRAULIC SNUBBERS *
.,i.-
m l!
SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGil RADIATION ESPECIALLY DIFFICULT r
=
NO.
ON, LOCATION ** AND ELEVATION INACCESSIBLE ZONE ****
TO REMOVE
' t.
U (A or I)
(Yes or No)
(Yes or Nc) 3.?
w 4
Make-up and Purification System 9.'
(Continued) 3
'/.
Mull-86 102'-3" A
No No MUH-87 102'-3" A
No No
-F-MUH-88 103'-0" I
Yes No
~:.s MUH-89 108'-0" A
No No
[
Decay Heat Removal System
.ij 2
k.
DilH-35 152'-5" I
Yes No DHH-36 152'-5" I
Yes No y
I$
DilH-37 159'-7" I
Yes Yes
',fr$
DHH-38 160'-1" I'
Yes Yes
- 5. ',
DHil-39 165'-9" I
Yes No
- S DiiH-661 86'-6"***
I Yes No
$I DHR-18 84'-7"***
I
,Yes No
.-t DHR-21 103'-6"***
A No No
~$
DilR-24U 129'-6"***
A No No
~
j\\
DilR-24L 129'-6"***
A No No
.i DHR-28 134'-4"***
.A No No DilR-31 84'-9"***
I Yes No
.j_,j.
DHR-37 85'-6"***
I Yes No Y,
DilR-49 85'-6"***
I Yes No E'
M 44...
G y.'
- aJb
% 3 ? ~Y li'Q f' fed
' Wi#
.. : :;> ]!?1E', j \\,.
"Y' 7-
s.';
.9
~
g v
TABLE 3.7-3 5
SAFETY RELATED HYDRAULIC' SNUBBERS
- r-
- o 9
SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT '
l-
- 2 T
i NO.
Ott, LOCATION ** AND ELEVATION INACCESSIBLE ZONE ****
TO REMOVE (A or I)
(Yes or No)
(Yes or No) c Reactor Coolant System RCH-26 171'-1" I
Yes No RCH-27 171'-1" I
Yes No RCH-28 168'-8" I
Yes No 3
RCH-29 168'-8" I
Yes No RCH-30 168'-8" I
Yes No RCH-31 168'-8" I
Yes No R
RCH-32 159'-6" I
Yes No -
RCH-33 159'-6" I
Yes No No
?
RCH-34 147'-0" A
No RCH-35 142'-6" A
No Yes 4
'l RCH-36 144'-9" A
No Yes RCH-37 144'-9" A
No No RCH-38 144'-9" A
No No
- y RCH-39 147'-0" A
No Yes RCH-40 147'-0" A
No Yes RCH-47tl 121'-9" I
Ye's Yes RCH-47S 121'-9" I
Yes Yes E
RCH-48 121'-9" I
Yes Yes
'i RCH-49 121'-9" I
Yes Yes RCH-50 167'-5" I
Yes No RCH-51 159'-6" I
Yes Yes RCH-52 147'-0" A
No No RCH-53 147'-0" A.
No No 7-RCH-54 147'-0" I
Yes No
-l RCH-55 123'-9" A
No No RCH-56 158'-6" I
Yes No RCH-57 158'-3" I
Yes Yes RCH-58 120'-0" A
No No i'
RCH-59 168'-0" I
Yes
-No
'%.s 5:
U f
}- l.]
~
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Yl-
.l'M;'.'l0+~
. I }',-l '
)
'p
'. T TABLE 3.7-3 q.
Q SAFETY RELATED llYDRAULIC SNUBBEks' 5f_.J i $'}
j.
! jP 2
SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT 9
NO.
ON, LOCATION ** AND ELEVATI0f!
INACCESSIBLE ZONE ****
TO REMOVE (A or I)
(Yes or No)
(Yes or No)
.~..
c5 Reactor Coolant System (Continued)
H
, u RCH-60 167'-5" I
Yes No No No I
RCll-61 145'-6" A
RCll-62 148'-4" I
Yes Yes i
~
l RCil-63 148'-4" I
Yes No 7
RCll-64 148'-3" I
Yes Yes RCll-65 131'-6" I
Yes Yes
'E
.t' RCil-66 140'-0" I
Yes No 6
RCll-67 166'-0" I
Yes No
?'.
Y RCil-68 167'-1" I
Yes No J.
.h;
.M RCil-69 167'-1" I
Yes No I
Yes No RCil 167'-1" 4.
RCil-71U 167'-1" I
Yes No i:I l
RCll-71L 167'-1" I
Yes No RCil-73 167'-1" I
Yes No RCll-74 167'-1" I-Yes No b'<
RCH-76 139'-11" I
Yes Yes i."
[
RCil-77 131'-6" I
Yes Yes RCH-78 150'-10" I
Yes Yes 5
RCll-79 156'-8" I
Yes Yes v.-: *
~.
RCH-60 168'-3" I
Yes No i
RCll-81 168'-4" I
Yes No l ';
RCH-82E 143'-9" A
No Yes b~
RCil-82W 143'-9" A
No Yes RCH-83 145'-11" A
No No
'i RCil-84 123'-9" A
No No RCil-85 147'-0" A
No No
'.*.?
RCll-86 144'-9" A
No No
'g RCil-87 147'-0" A
No No
-a if
.Y RCil-88 147'-0" A
No
-No
?.;
RCil-89 124'-0" A
No No I
RCil-90 120'-0" A
No No 8
f'l-4 f'
I.7F',?Jij$.C[i-
.l' -]i:
}pi
}.
~{i{T ~~~~ p's I
J s y, i . (ABLE 3.7-3 1 o SAFETY RELATED HYDRAULIC SNUBBERS
- 3
.? i W m SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT .w r-h. 3 NO. ON, LOCATION ** AND ELEVATION INACCESSIBLE ZONE **** TO REMOVE i (A or I) __ (Yes or No) (Yes or No) m I 'h Reactor Coolant Pump 3Al f,7.. g g 134'-6" I Yes Yes RCHS-1A /J RCHS-2A 134'-6" I Yes Yes w' m Yes '5 RCHS-3A 134'-6" I Yes~ Yes RCitS-4A 134'-6" I Yes G ~i RCHS-SA 134'-6" I Yes Yes ?- I RCHS-6A 134'-6" I Yes Yes c RCHS-7A 134'-6" I Yes Yes E -] - RCHS-8A 134'-6" I Yes Yes -I'. Reactor Coolant Pump 3A2 t i y 3' l d RCHS-1B 134'-6" I Yes Yes' e i .'ji RCHS-28 134'-6" I Yes Yes -t RCHS-3B 134'-6" I Yes Yes 'I l RCHS-4B 134'-6" I Yes Yes
- 1' RCHS-5B 134'-6"
-I Yes Yes 4. RCHS-6B 134'-6" I ie~s Yes RCHS-78 134'-6" I Yes Yes D I RCHS-8B 134'-6" I Yes Yes. Q- .g< 7.c Reactor Coolant Pump 381 .RCHS-lC 134'-6" I Yes Yes
- Y
.k. ' RCHS-2C. 134'-6" I .Yes Yes RCHS-3C 134'-6" I Yes Yes .RCHS-4C 134'-6" I Yes Yes .J i.7 RCHS-5C 134'-6" I Yes Yes RCHS-6C 134'-6" I Yes Yes '{C. f RCHS-7C 134'-6" Yes Yes j RCHS-8C 134'-6" I Yes Yes g e e g e
- g.,.
- )
e 6
. 7- \\s 'q ' ?. - TABLE 3.7-3 O. n$ SAFETY RELATED llYDRAULIC SNUCBERS* S'l jE: SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT q NO. ON, LOCATION ** AND ELEVATION INACCESSIBLE _ ZONE **** TO REMOVE y 9 (A or I) (Yes or No) (Yes or No) s Reactor Coolant Pump 382 c-5 H RCilS-10 134'-6" I Yes Yes RCilS-2D 134'-6" I. Yes Yes RCHS-3D 134'-6" I Yes Yes RCHS-4D 134'-6" I Yes Yes ") RCilS-5D 134'-6" I Yes Yes ( RCliS-60 134'-6" I Yes Yes~ i~ RCllS-7D 134'-6" I. Yes Yes RCilS-8D 134'-6" I Yes Yes V -i R' Snubbers may be added to safety related systems without prior License Amendment to Table 3.7-3 provided that a revision to Table 3.7-3 is included with,the next License Amendment request. i.r f y w
- All safety related snubbers are located in the Reactor Building except-those with the triple asterisk.
- Modifications to this table due to changes in high radiation, areas shall be submitted to the NRC as part of the next License amendment request.
- All safety related snubbers are located in the Reactor Building except-those with the triple asterisk.
i 's .a Yes -(,- [ i l- -l.
- s ofh\\e'h**.f.=I
^ f'. ~ ^ ?
h 'h.h. .....kg{.k.....,'_ $;.' * '. N 'W. ] j i 4 ( ?,,,
- )
n TABLE 4.7-4
- ,,3\\
y E HYDRAULIC SNUBBER INSPECTION SCHEDULE ,t a 5 5 A HEXT REQUIRED E NUMBER OF SNUBBERS FOUND INOPERABLE DURING INSPECTION OR DURING INSPECTION INTERVA'
- INSPECTION INTERVAi.**
- I t
.Q c-C 0 18 months + 25% 12 months T 25% l 6 months T 25% 2 3 or 4 124 days + 25% 5, 6, or 7 62 days T 25% i >8 31 days T 25% j 'j t g 5-Y f,,' O q ( -}. ,3 s Snubbers may be categorize'd into two groups, " accessible" and " inaccessible". This categorization shall )( be based upon the snubber's accessibility for inspection during reactor operation. These two groups 'f 't may be inspected independently according to the above schedule. 5:- The required inspection interval shall not be ler.gthened more than one step at a time. 1 ~ 'l 7 i V l,,.i. .i., i,'?
O b e OTHER SAFETY-RELATED-SYSTEMS NOT ADDRESSED IN THE CR#3 STS i 9 4 g h
( FIRE SYSTEM Equipment FPC Surveillance Program (3)
- 1) FSP-1, 2A, 2B, 3 6.0
- 2) FSV-24 6.0
) 1 ( i J i (3) Section 6.0 of the surveillance procedure for the Fire System is attached. +- y-y ,,.3.-
( a 6.0 PROCEDURE 6.1 -Notify the Shift Supervisor before performing this Surveillance Procedure. 6.2 Verify that the fire service water system is aligned in accordance with Enclosure 1. t 6.3 Cor.plete Section 1 of Enclosure 2. 6.4 Place fire service makeup pump (FSP-3) local control in the "Off" position. 6.5 Slowly open FSV-24 to decrease system pressure until electric. motor-driven fire service pump (FSP-1) starts. Pump should start between 98-92 psig. 6.6 Record in Section 2 of Enclosure 2 the lowest pressure observed on FS-6-PI before FSP-1 started. 6.7 Slowly _open FSV-24 further. As pressure drops, diesel-d' iven r fire service pump' FSP-2A should start between 88-82 psig., 6.8 Record in Section 2 of Enclosure 2 the lowest pressure observed on,FS-6-PI before FSP-2A started. i 1
'69 Continue to open FSV-24. As pressure drops further, diesel 4 driven fire service pump FSP-2B should start between 78-72 psig. 6.10 Record in Section 2 of Enclosure 2 the lowest pressure observed on-FS-6-PI before FSP-2B started. 6.11 Placc the selector switch for the FSP-2A and FSP-2B controllers in the "Off" position. Open the breaker for FSP-1 at the local controller. 6.12 Close FSV-24. 6.13 Place the local control f or FSP-3 in the " Auto position and observe that system pressure increases to greater than 100 psig on FS-6-PI. 6.14 Place the selector switches for the FSP-2A and FSP-2B controllers in the " Auto" position. 6.15 Place the local control for FSP-3 in the "Off" position. 6.16 Slowly open FSV-24. As system pressuts drops, FSP-2A and FSP-2B should start between 98-92 psig. 6.17 Record in Section 3 of Enclosure 2 the lowest pressure observed on -FS-6-PI before FSP-2A and FSP-23 started. f' 6.18 Place the selector switches for the FSP-2A and FSF-2B controllers / in the "Off" position. / 6.19 Close FSV-24 6.20 Flace the local control for FSP-3 in the " Auto" p.'sition and obcerve that system pressure returns to normal. ( (' 4 l l l r
r" U.dttshunE 2 (Pags 1 of 2) DATA SHEET I s Section 1: l PRE-TEST CHECK LIST CRITERIA VALUE REMARKS INITIALS j FST-1A Level (Water)
- 2. 34' ft.
FST-1B Level (Water)
- 2. 34' ft.
FST-2A Level (Fuel 011) 3/4 - full FST-2B Level (Fuel 011) 3/4 - full FSP-2A Engine Oil Level full FSP-2B Engine Oil Level' ful'1 r... 'Section 2: I I PARA. ~ POINT CHECKED CRITERIA VALUE INITIALS' 6.6 FSP-1 Auto-Start 92-98 psig psig 6.8 FSP-2A Auto-Start 82-88 psig psig 6.10 FSP-2B Auto-Start 72-78 psig psig Section 3: FARA. POINT CHECKED C.RITERIA VALUE , INITIALS 6.17 FSP-2A Auto-Start 92-t'8 psig psig 6.17 FSP-2B Auto-Start 92-98 psig psin w l ~
t.:.u.auuaN 1 (Paga 2 of 2) ~ DATA S!!EET I (Cont'd) Section 4: Carbon Dioxide System Surveillance Valve Lincup Check REQUIRED VALVE NO. DESCRIPTION POSITION INITIALS COV-CO2 to Fire Prot. Nozz1cs Main Tank Iso. Viv Open COV-PI & PS Inst. Iso. From Tank Vapor Space Open COV-LI Inst. Iso. From Tank Bottom Open Iso. of Pilot CO2 Press. to COV-Master Pilot Control Box Open a Performed By Date
i ~~' ' ~ ~ ~ ' ~ r SPENT FUEL COOLING SYSTEM FPC Preventative Maintenance Program Equipment Monthly
- 1) All pump motors.
- 1) Visually inspect for:
a) Cleanliness, clean as necessary b) Signs of overheating, discoloration, odors, or cracked insulation c) Foreign object damage d) Deterioration and rust Annually -
- 1) Perform winding resistance checks
- 2) All pumps Monthly
- 1) Visually inspect for:
a) Oil and water leaks b) Foreign object damage c) Deterioration and rust
- 2) Check and adjust packing as necessary
- 3) If pump has been disassembled for corrective maintenance, check the alignment of the unit on re-assemble
- 4) Open and adjust speed converters as applicable Semi-Annually
- 1) Visually inspect the spent fuel coolant pumps lubricating oil at the following points:
a) Electric Motor Bearing - Replacement Oil - Gulferest 44 b) Pump Bearings - Replacement Oi] Gulferest 44 4
RADIOACTIVE WASTE DISPOSAL SYSTEM Equipment FPC Preventative Maintenance Program
- 1) Waste Gas Compressor Monthly
- 1) Check strainer screens and gauge glasses
- 2) Check level alarm in separator
- 3) Inspect check valves hinge pins, pivots, springs and clapper nuts.
- 4) Visually inspect motors for:
a) Cleanliness, clean as necessary b) Signs of overheating, dis-coloration, odors, or cracked insulation c) Foreign object damage d) Deterioration and rust
- 5) Visually inspect compressor for:
a) Oil and water leaks b) Foreign object damage c) Deterioration and rust
- 6) If compressor has been disassembled for corrective maintenance, check the alignment of the unit on re-assemble Semi-Annually
- 1) Visually inspect the Compressor Inboard Bearing lubricating oil.
Change as necessary with Gulf Harmony 53
- 2) Perform internal inspection to Air Receivers Annual.
- 1) Visually inspect tanks externally for the following:
a) Penetrations for evidence of leakage b) Manways for evidence of leakage c) Foreign object damage
f RADIOACTIVE WASTE DISPOSAL SYSTEM (cont td) Equipment FPC Preventative Maintenance Program
- 2) Waste Disposal Pumps Each Refueling Shutdown a) See attached list of pumps
- 1) One third of the waste disposal pumps will have the following performed:
- a. Visually inspect for:
- 1. Oil and water leaks
- 2. Foreign object damage
- 3. Deterioration and rust
- b. Disassemble and inspect:
1., Bearings
- 2. Rotor assembly
- 3. Automatic thrust balance and end play
- c. Check alignment of unit on reassemble.
d'. Check and adjust packing as necessary
- e. Open and adj ust speed converters as applicable
- f. Visually inspect pump motors for
- 1. Cleanliness, clean as necessary
- 2. Signs of overheating, dis-coloration, odors, or cracked.
insulation '3. Foreign object damage
- 4. Deterioration and rust
- g. Perform pump motor winding resistance check
RADIOACTIVE WASTE DISPOSAL 1. Reactor Building Sump Pump (WDP-2A) 2. Waste Transfer Pump (WDP-5A) 3. Misc. Radioactive Waste Transfer Pump (WDP-6A) 4. Reactor Coolant Drain Tank Pump (WDP-7) 5. Decant Pump (WDP-10) 6. Concentrated Radioactive Liquid Waste Pump (WDP-12B) 7. Evaporator Condensate Pump (WDP-14A) 8. Reactor Coolant Evaporator Feed Tank Pump (WDP-16A) 9. Reactor Coolant Evaporator Distillate Pump (WDP-17B) 10. Waste Evaporator Feed Tank Pump (WDP-19A) 11. Waste Evaporator Distillate Pump (NDP-20B) 12. Decay Heat Pit Sump Pump (WDP-3B) 13. Reactor Building Sump Pump (WDP-2B) 14. Auxiliary Building Sump Pump (WDP-4A) 15. Waste Transfer Pump (WDP-5B) 16. Misc. Radioactive Waste Transfer Pump (WDP-6B) 17. Neutralizer Pump (WDP-9A) 18. Slurry Pump (WDP-ll) 19. Boric Acid Recycle Pump (WDP-13A)
pm RADIOACTIVE WASTE DISPOSAL (WD) (Continued) 20. Evaporator Condensate Pump (WDP-14B) 21. Reactor -Coolant Evaporator Feed Tank Pump (WDP-16B) 22. Reactor Coolant Evaporator Vacuum Pump (WDP-18A) 23. Waste Evaporator Feed Tank Pump (WDP-19B) 24. Waste Evaporator Vacuum Pump (WDP-21A) 25. Decay lleat Pit Sump Pump (WDP-3A) 26. Auxiliary Building Sump Pump (WDP-4B) 27. Waste Transfer Pump (WDP-5C) 28. Reactor Coolant Drain Pump (WDP-7) 29. Neutralizer Pump (WDP-9B) 30. Concentrated Radioactive Liquid Waste Pump (WDP-12A) 31. Boric Acid Recycle Pump (WDP-13B) 32. Laundry and Shower Sump Pump (WDP-15) 33. Reactor Coolant Evaporator Distillate Pump (WDP-17A) 34. Reactor Coolant Evaporator Vacuum Pump (WDP-18B) 35. Waste Evaporator Distillate Pump (WDP-20A) 36. Waste Evaporator Vacuum Pump (WDP-21B) 4 9
,m SECONDARY SERVICES CLOSED CYCLE COOLING SYSTEM Equipment FPC Preventative Maintenance Program
- 1) All pump motors Monthly
- 1) Visually inspect for:
a) Cleanliness, clean as necessary b) Signs of overheating, dis-coloration, odors, or cracked insulation c) Foreign object damage d) Deterioration and rust Semi-Annually
- 1) Visually inspect the lubricating oil in the motor bearings of SCP-1A and SCP-1B An,nually
- 1) Perform winding resistance checks
- 2) All pumps Monthly
- 1) Visually inspect for:
a) Oil and water leaks b) Foreign object damage c) Deterioration and rust
- 2) Check and adjust packing as necessary
- 3) If pump has been disassembled for correction maintenance, check alignment of the unit on re-assemble
- 4) Open and adjust speed converters i
as applicable Semi-Annually
- 1) Visually inspect the lubrications oil in the pump bearings of SCP-1A and 1B and in the worn gear housing of SCP-2
INSTRUMENT AIR SYSTEM Equipment FPC Preventative Maintenance Program
- 1) Air Compressors Monthly
- 1) Visually inspect motor for:
a) Cleanliness, clean as necessary b) Signs of overheating, dis-coloration, odors, or cracked insulation c) Foreign object damage d) Deterioration and rust
- 2) Visually inspect compressor for:
a) Oil,and water leaks b) Foreign object damage c) Deterioration and rust 3. If compressor has been disassemblet' for corrective maintenance, check the alignment of the unit on re-assemble. Quarterly
- 1) Inspect and clean air compressor feather valves, and visually inspec-cylinder bore
- 2) Check wear.on rider rings
- 3) Perform the following to the unloaders:
a) Clean b) Bench test c) Check contact of unloader fingers with Valve Strip d) Check operation and control elements Semi-Annually
- 1) Clean and inspect cylinder and head water jackets
- 2) Visually inspect piston rings for wear or breakage and any signs of leakage s
INSTRUMENT AIR SYSTEM (Continued) = Equipment FPC Preventative Maintenance Program Semi-Annually (contd)
- 3) Inspect cross-head and pin for wear in pin and pin bore
- 4) Inspect connecting rod bearing for wear or races, retainers, rollers, etc.
- 5) Visually inspect oil in air compressor crankcase.
Change as necessary with Gulf Harmony 69
- 6) Perform internal insepetion to air redeivers Annually
- 1) Inspect the main bearings for wear of rollers, races, retainers, etc.
- 2) Disassemble and inspect air compressor pistons
- 3) Visually inspect tanks externally in accordance with (PMP-24) for the following:
a) Penetrations for evidence of leakage b) Manways for evidence of leakage c) Foreign object damage Each Refueling Shutdown
- 1) Change / clean the following filters: i a) Instrument air compressor inlet filters b) Instrument air dryer prefilter*
c) Instrument air dryer after-filter
- oNOTE:
Instrumen't air dryer prefilter and after-filter should also be changed if the A p across the filter exceeds 6 psia l l l L.
,~ [ PENETRATION COOLING SYSTEM Equipment FPC Preventative Maintenance Program
- 1) ' Fans Monthly a) AHF-9A, 9B
- 1) Visually inspect for:
a) Cleanliness, clean as necessary b) Signs of overheating, dis-coloration, odors, or cracked insulation c) Foreign object damage ad) Deterioration and rust Annually Perform winding insulation resistance checks Yearly or When Fans Accessable Perform the following: 1)"Run each fan' motor and listen for any abnormal bearing noises
- 2) Check the blade settings cf each fan if applicable (Refer to the initial acceptance test for comparison settings)
- 2) Cooling coils Annual a) AHHE-13A, 13B Clean coil bank faces of all cooling coils with a vacuum cleaner, air hose, or water hose.
DO NOT exceed 50 psig water pressure as fin damage may occur. [ Special Preventative Maintenance i If the temperature across the coil ' xceeds 100% of norma 1 AT, chemically e clean the coils I i' i I s_- [ 0 f .}}