ML20117H593
| ML20117H593 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 05/17/1996 |
| From: | OMAHA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML20117H582 | List: |
| References | |
| NUDOCS 9605280272 | |
| Download: ML20117H593 (20) | |
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I TABLE OF CONTENTS (Continued) fBRt 2.12 Control Room Systems........................................ 2 59 2.13 Nuclear Detector Cooling System.................................. 2-60 2.14 Engineered Safety Features System Initiation l
Instmmentation Settings....................................... 2-61 2.15 Instrumentation and Control Systems................................ 2-65 2.16 River Level............................................... 2-71 stenst-(2rt 2 18 DELE'1TD^r^5 S ;;:r tre (S 255x:)
2 73 l
2 M NAuh-Ad 2.20 Steam Generator Coolant Radioactivity.............................. 2-96 l
2.21 Post-Accident Monitoring Instrumentation 2-97 2.22 Toxic G as Monitors.......................................... 2-99 3.0 SURVEILLANCE REQUIREMENTS................................... 3-0a 3.1 Instrumentation and Control.....................................
3-1 3.2 Equipment and Sampling Tests 3-17 1
3.3 Reactor Coolant System and Other Components Subject to ASME XI Boiler and Pressure Vessel Code Inspection and Testing Surveillance....................... 3-21 3.4 Reactor Coolant System Integrity Testing............................. 3-36 3.5 Containment Test............................................ 3-37 3.6 Safety Injection and Containment Cooling Systems Tests 3-54 3.7 Emergency Power System Periodic Tests............................. 3-58 3.8 Main Steam Isolation Valves..................................... 3-61 3.9 Auxiliary Feedwater System..................................... 3-62 3.10 Reactor Core Parameters....................................... 3-63 3.11 DELETED 3-64 j
3.12 Radioactive Waste Disposal System 3-69 l
1 our urveillan
.CC..........N' 3.14 DELETEDShx5 pp 2a (S :htes) 2" 3.16 Residual Heat Removal System Integrity Testing 3-84 3.17 Steam Generator Tubes........................................ 3-86 4.0 DESIG N FEATURES..............................................
4-1 4.1 Site....................................................4-1 4.2 Containment Design Features....................................4-1 4.2.1 Containment Structure...................................
4-1 4.2.2 Penetrations..........................................
4-1 4.2.3 Containment Structure Cooling Systems
......................4-2 l
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PDR ADOCK 0500 5
TECHNICAL SPECIFICATIONS - TABLES TABLE OF CONTENTS TABLE DESCRII'rION PAGE 3-3 Minimum Frequencies for Checks, Calibrations, and Testing of Miscellaneous Instrumentation and Controls 3-13 3-14 3-15 3-16
................ 3-16a
. 3-16b
......... 3-16c 3-3a Minimum Frequency for Checks Calibrations and Functional Testing of Alternate Shutdown Panels (Al-185 and Al-212) and Emergency Auxiliary Feedwater Panel (Al-179) Instrumentation and Control Circuits.........
....... 3-16d
........ 3-16e j
1 3-4 Minimum Frequencies for Sampling Tests 3-18 l
3-19 3-5 Mmimum Frequencies for Equipment Tests 3-20 3-20a
.. 3-20b
. 3-20e 3-20d 3-6 Reactor Coolant Pump Surveillance 3-27 3 13 Steam Generator Tube inspection....................
3-90 3'*
S=5'c m=! ' 7a:.
'-:c=!
3 '9' 5.2-1 Minimum Shift Crew Composition.
. 5-2 i
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v Amendment No. 15,!25,M2,M5, 52 l
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i TECHNICAL SPECIFICATIONS - TABLES l
4 TABLE OF CONTENTS (ALPHABETICAL ORDER) i Continued 1
1 1
TABLE DESCRIITION PAGE i
3-5 Minimum Frequencies for Equipment Tests 3-20
................................3-20a
................................3-20b i
................................3-20e
................................3-20d l
3-4 Minimum Frequencies for Sampling Tests 3-18 3-19 i
5.2-1 Minimum Shift Crew Composition.................................
5-2 2-10 Post-Accident Monitoring Instrumentation Operating Limits................. 2-98
..................... 2 -9 8 a
..................... 2-9 8 b 2-9 RCS Pressure Isolation Valves.............................-.....
2-2d 3-6 Reactor Coolant Pump Surveillance 3 27 1-1 RPS LS S S................................................ 1-10 1-10a l
..a w..._,r_._-.u_v._.._._,....
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c _..t t _
1 3-13 Steam Generator Tube inspection.................................. 3-90 i
2-11 Toxic Gas Monitoring Operating Limits............................. 2-100 l
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vii Amendment No. !!5.145,152 I
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l 2.0 LIMITING CONDITIONS FOR OPERATION 2.17 MISCELLANEOUS RADIOACTIVE MATERIAL SOURCF3 Applicability Applies to byproduct, source, and special nuclear radioactive material sources.
Obiective To assure that leakage from byproduct, source, and special nuclear radioactive material sources does not exceed allowable limits.
i Specifications Radioactive sources shall be leak tested for contamination. The leakage test shall be capable of detecting the presence of 0.005 microcurie of radioactive material on the test sample. If the test reveals the presence of 0.005 microcurie or more of removable contamination, it shall immediately be withdrawn from use, decontaminated, and repaired, or be disposed ofin accordance with Commission regulations. Those quantities of byproduct material that exceed the quantities listed in 10 CFR Part 30, Section 30.71, Schedule B are to be leak tested in accordance with the schedule shown in Surveillance Requirements. All other sources (including alpha emitters) containing greater than 0.1 microcurie are also to be leak tested in accordance with the Surveillance Requirements.
Basil Ingestion of inhalation of source material may give rise to total body or organ irradiation. This specification assures that leakage from radioactive material sources does not exceed allowable limits. In the unlikely event that those quantities of radioactive byproduct materials of interest to this specification which are exempt from leakage testing are ingested or inhaled, they represent less then one maximum permissible body burden for total body irradiation. The limits for all other sources (including alpha emitters) are based upon 10 CFR Part 70, Section 70. 39(c) limits for plutonium.
l Amendmen: No.11 i
2-72 Dated: February 4,1976 M#11NWR219j
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j 2.
LIMITING CONDITIONS FOR OPERATION j
2.18 hock Suppressors (Snubbers)
A licability N
I Opera ing Modes 1, 2 and 3 (Operating Modes 4 and 5 for snu ers located on sys s required operable.in those Operating Modes).
Specificat ons (1) All snu rs required to protect the reactor c lant and other safety related tems shall be operable except as np ed in 2.18(2) through 2.18(4) b These snubbers shall be ident'ified as safety-related snubbers.
j (2) With one or moeg safety-related snubbers inoperable, within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> I
replace or restke the inoperable snub r(s) to operable status and
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perfonn an engine ing evaluation per pecification 3.14(3) on the
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supported componen or declare the s.pported system inoperable and follow the appropri e actions spe (fied in the Technical Specifica-tions for that syst (3) A safety-related snubbe may b removed for surveillance in accordance l I
l with Section 3.14(3) of ese echnical Specifications, provided the l
following conditions are s
l (a) A given snubber station hall not be without an operable l
snubber for more t n two hours during surveillance of I
attendant snubber A snub r may be replaced by an' operable snubber during su 111ance and repair.
(b) No other snub r station is kn n to be inoperable.
I (c) Only one s bber station shall be removed for testing at l
l a time to ensure that no two snubb stations are without I
an oper le snubber during the same ime interval.
l (4) Safety-rel ted snubbers may be added, chang, or deleted provided an engin ring analysis justifies each change 4
Amendment No. 27,48,H.105 2-73 l
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, 2.0 '
LIMITING CONDITIONS FOR OPERATION
.18 Shock Sunoressors (Snubbers) l
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ubbers are designed to prevent unrestrained pipe motion under dynamic lo ds as t occur during an earthquake or severe transient, while allowing nor thermal mi during startup or shutdown. The consequence of an inoperable s bber is an moti increa in the probability of structural damage to piping as a result of seismic, or other ev t, initiating dynamic loads. It is therefore required that all nubbers required t rotect the primary coolant system or any other safety stem or component operable during reactor operation.
Because the snu r protection is required only during low p ability events, an inoperable period f 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed for repairs or rept ments and an inoperable period of two hours 's allowed for surveillance.
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i 2-74 Amendment No. 27,48,59,76,105, Q (next page is 2-96)
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l 3.0 SURVEILLANCE REOUIREMENTS L
3.3 Reactor Coolant System and Other Components Subject to ASME XI Boiler &
l Pressure Vessel Code Insoection and Testing Surveillance Applicability i
Applies to in-service surveillance of primary system components and other components subject to inspection and testing according to ASME XI Boiler &
Pressure Vessel Code.
1 Obiective To ensure the integrity of the reactor coolant system and other components subject to inspection and testing according to ASME XI Boiler & Pressure Vessel Code.
Soecifications (1)
Surveillance of the ASME Code Class 1,2 and 3 systems, except the steam generator tubes inspection, should be covered by ASME XI Boiler & Pressure Vessel Code.
a.
In-service' pection of ASME C e Cl Class 2, and Class 3 compone s[i@l6jiijiT[s@5sbl[jiE@@ anp in-service testing of '
ASME lass Cla 2, a Clas 3 ptrmps and valves shall be performed in accordance wit ection XI of the ASME Boiler and Pressure Vessel Code, as required by 10 CFR Part 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50, Section 50.55a (g)(6)(i).
b.
Surveillance of the reactor coolant pump flywheels shall be performed as indicated in Table 3-6.
c.
A surveillance program to monitor radiation-induced changes in the mechanical and impact properties of the reactor vessel materials shall be maintained in accordance with 10 CFR Part 50 Appendix H.m (2)
Surveillance of Reactor Coolant System Pressure Isolation Valves a.
Periodic leakage testing
- on each valve listed in Table 2-9 shall be accomplished prior to entering the power operation mode every time the plant is placed in the cold shutdown l
- To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage critetia.
3-21 Amendment No. 46,76,104,142,157
4 3.0 SURVEILLANCE REOUIREMENTS 3.13 RADIOACTIVE M ATERIAL SOURCES SURVEILLANCE Aoplicability Applies to leakage testing of byproduct, source, and special nuclear radioactive material sources.
Objectiv_c To assure that leakage from byproduct, source, and special nuclear radioactive material sources does not exceed allowable limits.
Soecification Tests for leakage and/or contamination shall be performed by the licensee or by other persons specifically authorized by the NRC or an agreement State, as follows:
1.
Each sealed source, except startup sources subject to core flux, containing radioactive material, other than Hydrogen 3, with a half-life greater than thirty days and in any form other than gas shall be tested for leakage and/or contamination at intervals of six months.
2.
The periodic leak test required does not apply to sealed sources that are stored and not being used. The sources excepted from this test shall be tested for leakage prior to any use or transfer to another user unless they have been leak tested within six months prior to the date of use or transfer, in tne absence of a' certificate from a transferor indicating that a test has been made within six months prior to the transfer, sealed sources shall not be put into use until tested.
3.
Startup sources shall be leak tested prior to and following any repair or maintenance and before being subjected to core flux.
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-76 Amendment No. I1,122,157 (Next Page is 3-84)
l SURVEILLANCE REOUIREMEN'IN i
j-3.14 Shock Suuu-u (Snubbers)
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e noticabilliv
'Ihis
'fication applies to all safety-related snubbers.
S ifi lons i
1 (1)
All draulic snubbers shall be visually inspected. As used in s specificadon, l
j of capacity. 'Ihis inspection shall include, bu not necessarily be i
limited spection of the hydraulic fluid reservoir, id connections, and j
lir.~. age ca ections to the piping and anchor to verify ubber operability. In those locati where snubber movement can be.
ually induced without disconnecting snubber, verify that the snubber
' freedom of movement and is not frozen u Snubbers which appear in le as a result of visual i
inspections shall classified as unseeptable an may be reclassified acceptable i
for the purpose of e lishing the next visual
- spection interval, provided that (1) the cause of the rejection is clearly tablished and remedied for that particular snubber and or other snub inespective of type that may be generically susceptible; (2) the affee snubber is functionally tested in the as-found condition and det ined OP BLE per functional testing acceptance criteria. All snubbers found to an inoperable common hydraulic fluid
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reservoir shall be counted as i = _, table for determining the next i===*iaa interval. A review and evaluati shall be performed and documented to justify continued operation with an unn able snubber. Ifcontinued operation cannot be justified, the snubber 3 hall 'bk declared inoperable and the ACI' ION requirements shall be met. Visual inspections shall be performed in accordance with Table 3-14.
(2)
On a refueling freq cy and subject to the nditions below:
l (a)
A represeptative sample (88) of hydrau
- snubbers shall be functionally tested ei er in-place or in a bench test.
i 3-77 Amendment No. 27,5^,I^5,145 157
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SURVEILLANCE REQUIREMENTS 3.14 Shock Suppressors (Snubbers) (Continued)
(1)
For each hydraulic ~ snubber above 3 which does no meet the functional test acceptar:e criteria, an add fional sample of 22 hydraulic snubbers shall be funct )onally tested.
(ii) For each inoperable hydraulic snubber found during a resample test, an additional 22 hydraulic nubbers will be tested until no additional inopeyable hydraulic snubbers are found within a sample or u til all hydrau-lic snubbers have been functionally t ted; and (b) 10% f the safety-related mechanical snubbers shall be funct nally tested either in-place or 14 a bench test.
(1)
F each mechanical snubber whi does not meet the fun ional test acceptance cri eria, an additional samp t of 10% of the mechani 1 snubbers shall be functf nally tested.
(ii) For eachsinoperable mecha cal snubber found during a resampi test, an addi ional 10% of the mechanical snubbers w 11 be teste until no additional inoperable mechanical ubbers a found within a sample or until all mechanica snub rs have been functionally tested.
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(c) The representative s selected for functional testing shall include the various co the range of size an4 ca(pacity of snubbers.igurations, operating envi At least 25% of -
the snubbers in the tepre' entative sample shall include snubbers from any of the fo owing ree categories:
1.
The first nubber away rom each reactor vessel nozzle.
2.
Snubber within 5 feet o heavy equipment (valve, pump turbin, motor, etc.).
3.
Snub rs within 10 feet of e discharge from a safety rel ef valve.
Snubbers th are especially difficult to ove or in high radiation zones dur g shutdown shall also be includ:.a in the representative sample.
nnanent or other exemptions from ctional testing for individy 1 snubbers in these categories may be ranted by the and/ gion only if a justifiable basis for exem Comis ion is presented snubber life destructive testing was perfo d to gialify er operability for all design conditions at e ther the completion snu of their fabrication or at a subsequent date.
d) In addition to the regular sample, snubbers which failed the previous functional test shall be retested during e next test period.
If a spare snubber has been installed place of a failed snubber, then both the failed snubber (1 it is repaired and installed in another position) and the sp re snubber shall be retested. Test results of these snubb rs may not be included for the resampling.
Amendnent No. 27,59,105 3-78 n
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,.0 SURVEILLANCE REOUIREMENTS j
- 3) 4 Shock Suppre;;ers (Snubbers) (Continued) j j
f If any snubber selected for functional testing either falls to lockup or to 1
move, i.e., is frozen in place, the cause will be evaluated. If the ca is a
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manufacturer or design deficiency, appropriate action shall be taken for ubbers j
of the same design subject to the same defect to determine if any mopf defects i
exist. This testing requirement shall be inhaaahat of the seg
- ts stated j
above for snubbers not meeting the functional test==y= ara cri i
j For y snubber (s) found locked up during normal operation or und inoperable i
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- g a seismic event, an engineering evaluation shall be ormed on the com ts which are supported by the snubber (s).
purpose of this j
engin evaluation shall be to determine if the com ts supported by the i
snubber (s) ere adversely aficcted by the inoperabsity o 'the snubber (s) in order j
to ensure tha the supported component remains capab of meeting the designed i
service. If th engineering evaluation shows the ponents to be capable of meeting the desi ed service without the failed bber, that snubber may be i
deleted from servi per Specification 2.18(4).
1 (3)
Snubber Service Life onitorine j
l A record of the service (life of each snub
, the date at which the designated i
service life commences and ke installati and maintenance records on which the j
designated service life is base shall maintained as required by Specification 5.10.2.m. On a refueling freq(uay j
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e installation and maintenance reconi for l
j each raubber shall be reviewed crify that the indicated service life'has not been exmeded or willnot be ex prior to the next scheduled snubber service
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life review. If the indicated life will be exceeded prior to the next scheduled snubber service ' e revi the snubber service life shall be re-evaluated or the snubber be rept or reconditioned so as to extend its service life beyond the of the next se uled service life review. This re-l evaluation, replacement reconditioning be indicated in the records.
4 Basis i
j All safety snubbers sh be operable to ensu e that the uralintegrity of the reactor i
coolant system and other safety-related systems is main ed during and following j
a seismic or other t initiating dynamic loads. Snubbers excl ed from thisinspection i
program are tho installed on non-safety-related systems and then only if their failure l
or failure of the ystem on which they are installed would have no hdverse effect on any l
safety-related ystem.
I The visua) mspection frequency is based upon maintaining a constant !
I of snubber j
protectig6 to systems. 'Ihe required inspection inLval will be based on able 3-14.
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l 3-79 Amendment No. 27,59,105,M5,157
3.0 ' SURVEILLANCE REOUIREMENTS 14 Shock Suppressors (Snubbers) (Continued)
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When a snubber is found locked up or frozen in place or when a snubber has
'noperable during a seismic event, an engineenng evaluation shall be perform,in dition to the determination of the snubber mode of failure. The purpose of the en 'neering evaluation is to determine if any safety-related component or syste been adve y'ly affected by the inoperability of the snubber. The engineering eval tion shall dete e whether or not the snubber mode of failure has imparted & sig ficant effect or deg tion on the supported component or system.
To provide surance of snubber functional reliability, a represen ve sample of the installed snubgrs will be functionally tested during plant shu wns at 18 month intervals. Sel
'on of a representative sample of hydraulic sn bers according to the expression 35(1+ 2) provides a confidence level of approximately 95% that 90% to 100% of the snubbe in the plant will be operable within tance limits. The District selected the value of be 3. Observed failures of these ple snubbers shall require functional testing of ad 'tional units. For each number snubbers above c which does not meet the functional tes acceptance criteria, an
- 'onal sample selected according to the expression 35(1+c/(2 (2/(c+1))2(a-c) will be etionally tested, where a is the total number of snubbers nd inoperable' d g the functional testing of the representative sample. Functs testing will ntinue according to the expression le snubbers are found within a samp or l b(35(1+c/2)(2/(c+1))2) w* e e is the num r of snubbers found inoperable in the previous resampic until no additiohg1inape until all snubbers have been functio 'ly ted.
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A "10%" criterion is utilized for m ical snubbers be'ause of the considerably smaller number of mechanical snub rs.
Hydraulic snubbers and mechani snubbers 'll each be treated as a different entity for the above surveillance prog The servicelife of a snu is evaluated via raanuf urerinputand information through consideration of the ubber service conditions d associated installation and maintenance records
.g. newly installed snubber, se,Q rglacal, spnng replaced, in high l radiation area, in 1;i tempernture area, etc.). The requuement to monitor the snubber service life is incl ded to enst.re that the snubbers periodically undergo a performance evaluation in vi of their age and v, crating conditions.
records will provide statistical b for future consideration of snubber service li The requirements for the nainte ce of records and the snubber service life revie,v not intended to affect plant ope tion.
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3-79a Amendment No. 59,105, 145
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TABLE 3-14
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SNUBBER VISUAL INSPECTION INTERVAL NUMBER OF UNACCEPTART F SNUBBERS I
Population Column A Column B Column C Extend Interval Repeat Interval Reduce InW (Notes 1 and 2)
(Note 3)
(Note 4)
(Note 5 )
1 0
0 1
0 0
2 100 0
1 4
150 0
3 8
200 2
5 13 300 5
12 25 400 8
18 36 5M 1
2 48 750 20 40 78 I
Note 1: The next visual inspectioginterval for a snubber population shall be determined specf'on interval and the number of Ur=-+p-i-le based upon the previous i
snubbers found during that He first inspection interval determined using this criteria shall b
upon the previous inspection interval as established by the rer^:.ents effect before amendment 145. Snubbers may be categorized, basco pon the accessibility during power operation, as accessible or inac
- e. D ese tegories may be examined separately or jointly. However, licensee must and document that decision before any inspection an 1 use that decisiogas the basis upon which to determine the next inspectio interval for that catag Note 2: Interpolation between population or ry sizes and the number of unacceptab)e snubbers is permissible. Use ne lower integer for the value of l
the limit or Columns A, B, or C if that integer
- ludes a fractional value of l
unacce ble snubbers as determined by interpola Note 3: If the number of unacceptable snubbers is equal to or ess than the ' number in Cdlumn A, the next inspection interval may be twice previous interval up j
to a 48 month interval.
Note : If the number of unacceptable snubbers is equal to or less than the number in
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Column B but greater that the number in Coluidn A, the hxt inspection interval shall be the same as the previous interval.
3-79b AeMment No.145 1
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p TABLE 3-14 (Continued) i.
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te 5: If the number of unacceptable snubbers is equal to or greater than the num l;
in Column C, the next inspection interval shall be two-thirds of the previ s l
interval. However, if the number of unacceptable snubbers is less th the l
number in Column C but greater than the number in Column B, the nex rval I be reduced by a factor that is one-third of the ratio of the rence een the number of unacceptable snubbers found during the previ interval an the number in Column B to the difference in the numbers in C umns B and C.
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3-79c Amendment No.
160 l
(next page is 3-84) l
I 5.0 ADMINISTRATIVE CONTROLS 5.10.2 The following records shall be retained for the duration of the Facility i
Operating License:
a.
Records of drawing changes reflecting facility design modifications made to systems and equipment described in the Final Safety Analysis Report.
b.
Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.
c.
Records of facility radiation and contamination surveys.
d.
Records of radiation exposure for all individuals entering radiation control areas.
e.
Records of gaseous and liquid radioactive material released to the environs.
f.
Records of transient or operational cycles for those facility components i
designed for a limited number of transients or cycles.
g.
Records of training and qualification for current members of the plant staff.
h.
Records of in-service inspections performed pursuant to these Technical Specifications.
j i.
Records of Quality Assurance activities required by the QA Manual.
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Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.
k.
Records of meetings of the Plant Review Committee and the Safety Audit and Review Committee.
1.
Records of Environmental Qualification of Electric Equipment pursuant to 10 CFR 50.49.
qcordsjL1he seprvice,11 veto _f_glLhydpandjneclianicappbbergwhich m.
rc covered under th-rc.i:;ica:; cf S_ tion 2.18 cf the Technical
__ s Specific ^:!cJa which theM6 ice life commences and
iiss'ociated installation and maintenance records.
i n.
Records of analyses required by the Radiological Environmental Monitoring Program.
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Records of reviews performed for changes made to the Offsite Dose l
Calculation Manual and the Process Control Program.
5.10.3 A complete record of the analysis employed in the selection of any fuel assembly to be placed in Region 2 of the spent fuel racks will be retained as long as that assembly remains in Region 2 (reference Technical Specifications 2.8 and 4.4) 5.11 Radiation Protection Program Procedures for personnel radiation protection shall be prepared consistent with the r
requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for l
all operations involving personnel radiation exposure.
5 19 Order 70/24/80 Amendment No. 59,85,93,99,105,152,155
U.S. Nuclear Regulatory Commission LIC-96-0071 ATTACHMENT B i
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DISCUSSION, JUSTIFICATION AND NO SIGNIFICANT HAZARDS CONSIDERATION DISCUSSION AND JUSTIFICATION l
The Omaha Public Power District (OPPD) proposes to revise the Fort Calhoun Station (FCS) Unit No.1 Technical Specifications (TS) to relocate the operability requirements for shock suppressors (snubbers) from the T5 to the Updated Safety Analysis Report (USAR), and to incorporate snubber examination I
and testing requirements into TS 3.3 which governs components subject to testing in accordance with Section XI of the ASME Boiler & Pressure Vessel Code.
l Snubbers are designed to prevent unrestrained pipe motion under dynamic loads l
that may occur during an earthquake or design basis accident, while allowing normal thermal motion during plant / system startup or shutdown. The consequence of an inoperable snubber is an increase in the probability of structural damage to piping as a result of a seismic, or other event, initiating dynamic loads.
TS 2.18 provides operability requirements and allowed outage times for snubbers. TS 3.14 provides requirements on visual inspection, functional testing, and service life monitoring of snubbers.
i On July 16, 1993, the NRC issued a Final Policy Statement on Technical Specification Improvements for Nuclear Power Reactors. The Final Policy Statement contains four criteria which can be used to determine which constraints on the design and operation of nuclear power plants are appropriate for inclusion in TS. The NRC has incorporated these criteria into 10 CFR 50.36, " Technical specifications." Snubbers do not meet any of the four criteria for inclusion as a Limiting Condition for Operation within the TS, and therefore it is proposed that these requirements be relocated from the TS.
l Specification 2.18 l
It is proposed to relocate the operability requirements for snubbers contained in TS 2.18 to the USAR and/or plant procedures. With the deletion of this TS, it is proposed that the Table of Contents be revised accordingly. Any changes to the operability requirements for snubbers can be made without NRC approval only when the changes meet the criteria of 10 CFR 50.59.
Changes to the i
snubber operability requirements that do not meet the criteria of 10 CFR 50.59 I
must be approved by the NRC by license amendment.
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l DISCUSSION AND JUSTIFICATION (Continued) l Specification 3.3 l
It is proposed to revise TS 3.3(1)a to include applicable supports (snubbers) in the inservice inspection of ASME Code Class 1, Class 2, and Class 3 components. Snubbers are required to be examined and tested in accordance with ASME Section XI by 10 CFR 50.55a except where specific written relief is granted by the NRC. The FCS Inservice Inspection (ISI) Program is in accordance with the 1989 Edition of Section XI and requires that inspections be performed in accordance with the first Addenda to the ASME/ ANSI OM-1987, Part 4.
OPPD was granted approval by the NRC to perform snubber examinations and testing in accordance with the ASME OMb Code-1992, Subsection ISTD, in lieu of the first Addenda of the ASME/ ANSI OM-1987, Part 4.
Because examinations and testing must be in accordance with both the TS and ASME Section XI, there are redundancies or differences between visual inspection requirements, visual inspection frequencies, functional testing sample size and content, functional testing expanded scope requirements, and failure evaluation.
However, where differences between the deleted sections of the TS and ASME Section XI requirements exist, the Section XI requirements are either l
similar or more conservative than the TS requirements.
The proposed revision to TS 3.3(1)a will ensure that snubber surveillance is performed in accordance with ASME Section XI as required by 10 CFR 50.55a.
Specification 3.14 It is proposed that TS 3.14 be deleted. With the addition of snubbers to TS 3.3(1), the visual and functional testing required by TS 3.14(1) and 3.14(2) will be conducted in accordance with ASME Section XI as required by 10 CFR 50.55a. Therefore, TS 3.14(1) and 3.14(2) are redundant and unnecessary. The requirements of TS 3.14(3) aridressing snubber service life monitoring, which is not addressed by the ASME Code, will be relocated to the USAR and/or plant procedures.
Specification 5.10.2 It is proposed to revise TS 5.10.2 to delete the reference to TS 2.18 since it is proposed that TS 2.18 be deleted.
ADMINISTRATIVE CHANGES It is proposed to add amendment numbers 11, 27, 32, 38, 43, 46, 54, 60, and 97 to the bottom of Page ii of the Table of Contents as amendments that have revised this page. These amendment numbers were inadvertently deleted in a previous amendment.
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BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION:
The proposed changes do not involve significant hazards consideration because operation of. Fort Calhoun Station Unit No.1 in accordance with these changes would not:
(1)
Involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed change will relocate operability requirements for shock suppressors (snubbers) from the Technical Specifications (TS) to the Updated Safety Analysis Report (USAR) and/or plant procedures. On July 16, 1993, the NRC issued a Final Policy Statement on Technical Specification Improvements for Nuclear Power Reactors. The Final Policy Statement contains four criteria which can be used to determine which constraints on the design and operation of nuclear power plants are appropriate for inclusion in TS. The NRC has incorporated these criteria into 10 CFR 50.36, " Technical specifications." Snubbers do not meet any of the four criteria for inclusion as a Limiting Condition for Operations within the TS, and therefore it is proposed that these requirements be relocated from the TS.
The proposed change would not reduce or revise any of the current requirements for snubber operability, only relocate the requirements. Any changes to the requirements contained in the USAR and/or plant procedures can be made without NRC approval only when the changes meet the criteria of 10 CFR 50.59.
Changes to the snubber operability requirements that do not meet the criteria of 10 CFR 50.59 must be approved by the NRC by license amendment. Therefore, the relocation of the requirements on snubber i
operability from the TS to the USAR does not increase the probability or consequences of any accident previously analyzed.
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l The proposed change also deletes sections of the TS which are redundant l
or in conflict with the American Society of Mechanical Engineers (ASME) l Boiler and Pressure Vessel Code.
Snubbers are required to be examined and tested in accordance with ASME Section XI by 10 CFR 50.55a. The proposed change will ensure that the TS implement ASME Section XI examination and testing requirements for snubbers in accordance with 10 CFR 50.55a. Where differences between the deleted sections of the TS and ASME Section XI requirements exist, the Section XI requirements are j
similar or more conservative than the TS.
For example, although the functional test sample size differs between the methodologies, both ensure that a very high percentage of the snubbers in the plant are operable within acceptance limits.
Therefore, the proposed revision does not reduce the effectiveness of snubber examination and testing.
The proposed change would not reduce the operability requirements, acceptance criteria, or examination and testing of snubbers. Therefore, the proposed change would not increase the probability or consequences of an accident previously evaluated.
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BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION (Continued):
(2)
Crev.e zhe possibility of a new or different kind of accident from any accident previously evaluated.
1 There will be no physical alterations to the plant configuration, changes to setpoint values, or changes to the implementation of setpoints or limits as a result of this proposed change.
The proposed change deletes duplicate or conflicting requirements between the TS and the ASME Section XI.
In these areas, the proposed deletions would remove the TS requirements and testing would be conducted in accordance with ASME Section XI as directed by 10 CFR 50.55a. Although the requirements of ASME Section XI differ from the TS in some cases, the differences do not decrease the effectiveness of testing and examination as compared to the TS requirements. Other areas, such as snubber operability requirements and service life monitoring, which are presently addressed by TS, but are not covered under ASME Section XI, will be maintained in the USAR so that these requirements cannot be deleted without NRC approval.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
(3)
Involve a significant reduction in a margin of safety.
The proposed change does not reduce the operability, examination, or testing requirements for snubbers.
Snubbers will still be required to meet the requirements of ASME Section XI and 10 CFR 50.55a except where specific written relief has been granted by the NRC. Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Therefore based on the above considerations, it is OPPD's position that this proposed amendment does not involve significant hazards considerations as defined by 10 CFR 50.92 and the proposed changes will not result in a condition which significantly alters the impact of the Station on the environment. Thus, the proposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and pursuant to 10 CFR 51.22(b) no environmental assessment need be prepared.
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