ML20036A836
| ML20036A836 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 05/06/1993 |
| From: | CENTERIOR ENERGY |
| To: | |
| Shared Package | |
| ML20036A829 | List: |
| References | |
| 2130, NUDOCS 9305170065 | |
| Download: ML20036A836 (17) | |
Text
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. ' Docket Number 50-346 License Number NPF-3 Serial Wumber 2130 Attachment i Page 5 I
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DEFINITIONS 1.29 Deleted I
PROCESS CONTROL PROGRAM l.30 The PROCESS CONTROL PROGRAM (PCP) shall contain the current fomulas, sampling, analyses, tests, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated
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processing of actual or simulated wet solid wastes will be accomplished in such a way as to asst.re compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.
1.31 Deleted l
OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.32 The OFFSITE DOSE CALCULATION MANUAL (0DCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radio-active gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alim/ Trip Setpoints, and in the conduct of the Environ-mental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Enviromental Monitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating
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I and W i: m al adioactive Effluent Release Reports required by Specifications 6.9.1.10and6.@9.1.11.
I l.33 Deleted 1.34 Deleted l
1.35 Deleted U. l 1.36 Deleted MEMBER (5) 0F THE PUBLIC l.37 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupation-ally associated with the plant. This category does not include employees of the utility, its contractors or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries.
This category does include persons who use portions of the site for recrea-1 tional, occupational or other purposes not associated with the plant.
SITE BOUNDARY 1.38 The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.
DAVIS-BESSE, UNIT 1 1-6a Amendment No. EE.170 9305170065 930506 PDR ADOCK 05000346 P
Dockat Number 50'346
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License Number NPF-3
- Serial Number 2130 Attachment i
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Page 6 ll jj INSTRUMENTATION f
4 WASTE GAS SYSTEM OXYGEN MONITOR LIMITING CONDITION FOR OPERATION 3.3.3.9,The Waste Gas System Oxygen monitor shall be OPERABLE with its alarm setpoints set to ensure that the limits of Specification 3.11.2 are not exceeded.
- . APPLICABILITY
- During additions to the waste gas surge tank.
l ACTION:
With the waste gas system oxygen monitor alarm setpoint less a.
i conservative than required by the above Specifications, declare j
the channel inoperable and comply with ACTION b.
b.
With the waste gas system oxygen monitor inoperable, additions to
!j the waste gas surge tank may continue provided another stethod for lt ascertaining oxygen concentrations, such as grab sample anslysis, j'
is implemented to provide measurements at least once per four (4) j hours during degassing and daily during other operations. Exert best efforts to return the waste gas system oxygen monitor to OPERABLE status within 30 days and, if unsuccessful, explain in the i
Metie.Ja4 Radioactive Effluent Release Report why the inopera-bility was not corrected in a timely manner.
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I The provisions of Specifications 3.0.3 and 3.0.4 are not c.
applicable.
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SURVEILLANCE REQUIREMENTS i
I 4.3.3.9 The waste gas system oxygen monitor shall be demonstrated OPERABLE by:
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a.
Perfomance of a CHANNEL CHECK at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during j
additions to the waste gas surge tank.
I b.
At least once per 92 days by perfomance of a CHANNEL CALIBRATION.
The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
j 1.
One volume percent oxygen. balance nitrogen; and 2.
Four volume percent oxygen, balance nitrogen.
DAVIS-BESSE, UNIT 1 3/4 3-57 Amend ent No. Sf 170
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Doriet-Nutbar 50-346
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License Number NPF-3 RADICA T!'iE EFFLUENTS i
.S2 rial Number 2130 ;
j Attachment i LIOUID HOLDUP TANKS
- jPcgo 7
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- LIMITI';3 CONDITION 'FOR OPERATION '
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l3.11.1 The quantity of radioactive material contained in each of the
, following unprotected outdoor tanks shall be limited to less than or j,
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' ecual to 10 curies, excluding tritium and dissolved or entrained noble i
c gases.
a.
Outside temporary tank.
. APPLICA31LITY: At all times.
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' ACT:0N:
9 With the quantity of radioactive material in any of the above a.
listed tanks exceeding the above limit, imediately suspena all i
additions of radioactive material to the tank and within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit. and descrite the i
event leading to this condition in the nextyrf:=:' Radio-l j '.
active Effluent Release Report.
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The provisions of Specifications 3.0.3 and 3.0.4 are not I
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applicable.
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' : 5"D.VE!'.'.tNCE :EOUIREMENTS
'l4.11.1 7he cuantity of radioactive material contained in each of the
,, acove listed tanks shall be determined to be within the above limit by
' analyzing a representative sample of the tank contents at least once per
,; 7 days.nen ractoactive materials are being added to the tank.
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' Tanks included in this specification are those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents or that do not have tank overflows and surrounding area drains connected to tne liquid radwaste treatment system.
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CAVIS-5E55E. UNIT 1 3/4 11-1 Amene.ent No. 26.170 s
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Docket Number 50-346 M[ h @ N j!)d ]h.)
ijI' License Number NPF-3 Sf
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- Serial Number 2130
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i Attachment b'
ADMINISTRATIVE CONTROLS t
COMPOSITION The Station Review Board (SRB) shall be composed of at least six
- 6. 5.1. 2 members of the Davis-Besse onsite management organization. The members shall be as a minimum, managers or individuals reporting directly to managers from each of the following disciplines: plant operations, maintenance, planning.
radiological controls, engineering and quality assurance. The members shall
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meet the requirements of ANSI N18.1-1971. Sections 4.2. 4.4. or 4.6 for applicable required experience.
The SRB Chairman shall be drawn from the SRB members and designated in writing by the Plant Manager.
ALTERNATES All alternate members shall be appointed in writing by the SRB I
- 6. 5.1. 3 l
Chairman; however, no more than two alternates shall participate as voting members in SRB activities at any one time.
l MEETING FREQUENCY i'
6.5.1.4 The SRB shall meet at least once per calendar month and as convened by the SRB Chairman or his designee.
000 RUM 6.5.1.5 A quorum of the SRB shall consist of the Chairman or his designee and four memoers including alternates.
RESPONSIBILITIES
- 6. 5.1. 6 The Station Review Board shall be responsible for' i
Review of plant administrative procedures and changes thereto.
a.
Review of the safety evaluation for 1) procedures. 2) changes b.
I to procedures, equipment or systems, and 3) tests or experiments completed under the provisions of 10 CFR 50.59. to verify that such actions do not constitute an unreviewed safety question.
3 I
Review of proposed procedures and changes to procedures and c.
equipment detemined to involve an unreviewed safety question as defined in 10 CFR 50.59.
I
. DAVIS-BESSE. UNIT 1 6-6 Amencment *:o. 72.75.98.72).738.127, if 2.169 i
Dockat Nuxb2r % WJ
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Lic;ns2 Nunber NPF-3
- Shrial Number 2130 Attach::nt Pag'e 9 ADMINISTRATIVE CONTROLS
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d.
Review of proposed tests or experiments detemined to involve an unreviewed safety question as defined in 10 CFR 50.59.
e.
Review of reports of violations of codes, regulations, orders.
Technical Specifications, or Operating License requirements having nuclear safety significance or reports of abnomal degradation of systems designed to contain radioactive material.
f.
Review of all proposed changes to the Technical Specifications or the Operating License.
9 Deleted h.
Review of reports of significant operating abnomalities or devi-ations from nomal and expected perfomance of plant equipment that affect plant safety.
- i. Review of the Industrial Security Plan, the Security Training and Qualification Plan, and the Security Contingency Plan, and changes thereto.
- j. Review of the Davis-Besse Emergency Plan and changes thereto.
k.
Review of items which may constitute potential nuclear safety hazards as identified during review of facility operations.
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Investigations or analyses of special subjects as requested by the Company Nuclear Reniew Board.
m.
Review of all REPORTABLE EVENTS.
n.
Review of all Safety Limit Violation Reports (Section 6.7).,
o.
Review of any unplanned, accidental or uncontrolled radioactive releases, evaluation of the event, ensurance that remedial action is identified to prevent recurrence, revi.ew of a report covering the evaluation and forwarding of the report to the Plant Manager and to the CNRB.
p.
Review of the changes to the OFFSITE DOSE CALCULATION W.ANUAL.
4 Review of the changes to the PROCESS CONTROL PROSPM.
Review of the Annual Radiological Envircrrnental Operaung Report.
r.
Review of thebrudemc(Radioactive Effluent Release Report.
s.
T.
Review of the Fire Protection Pro; ram and changes therete.
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DAVIS-BESSE. U*ili 1 6-7 Amencrent No. 27.8f.73 f 2.729.
RE91,174
~ Docket Nunb:r 50-346
, License Number NPF-3 Serial Number 2130 Attachment lPage10
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ADMINISTRATIVE CONTROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 6.9.1.10 The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted before May 1 of each year. The report shall include surmaries, interpretations, and analysis of trends of the results of the Radiological The material
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Environmental Monitoring Program for the reporting period.
provided shall be consistent with the objectives outlined in (1) the ODCM, and (2) Sections IV.B.2. IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.
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RADIOACTIVE EFFLUENT RELEASE REPORT r
6.9.1.11 The Stri;nnus. Radioactive Effluent Release Report covering the operation of the unit dring the previou; 5 nonths af operatica shall be j
d wts4tted-within E0 d:y: :ft r kne:ry i :nd July i sf cr.;h yc.r, The
- report shall include a sutrnary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be (1) consistent with the objectives outlined in the ODCM and PCP, and (2) in confomance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.
SkAll be. Cobm 4ftd. in Ac. corda.n
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i DAVIS-BESSE, UNIT 1 6-17a Amendment No. EE JJ,170
Dociet Number 5'0-346 Lfcense Number NPF-3 Serial Number 2130
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Attech. ment
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'Page 11 ADMINISTRATIVE CONTROLS b
6.14 FROCESS CONTROL PROGRAM (PCP)
Changes to the PCP:
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a.
Shall be documented and records of reviews perfomed shall be retained as required by Specification 6.10.2.p.
This documentation shall contain:
1)
Sufficient infomation to support the change together with the appropriate analyses or evaluttions justifying the change (s),
and 2)
A determination that the change will maintain the overall contomance of the solidified waste product to existing require-ments of Federal, State, or other applicable regulations.
b.
Shall become effective after review and acceptance by the SP3 and the approval of the Plant Manager.
6.15 0FFSITE DOSE CALCULATION MANUAL (ODCM)
Changes to the ODCM:
a.
Shall be documented and records of reviews perfomed shall be retained as required by Specification 6.10.2.p.
This documentation shall contain:
1)
Sufficient infomation to support the change together with the 1
appropriate analyses or evaluations justifying the change (s),
and 2)
A detemination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.106, 40 CFR Part 190,10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose or setpoint calculations.
b.
Shall become effective after review and acceptance by the SP3 and the approval of the Plant Manager.
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c.
Shall be submitted to the Comission in the fom of a complete, legible copy of the entire ODCM as part of or concurrent with the
/ kinneQRadioactive Effluent Release Report for the period of the report in which any change to the ODCM was made.
Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month / year) the change was implemented.
s DAVIS-BESSE. UNIT 1
' 6-22 Amendment No. EE.170 i
LGcL%t Nu2ber 5@-346 License Numb 2r NPF-3 Sarial Number 2130 g
Ci R E* 3 i
RE*: TOR COOLANT SYSTEM STEU4 GENERATORS i
l LIMITING CONDITION FOR OPERATION
- 3.4.5 Each steam generator shall be OPERABLE with a water level between 18 ano 348 inches.
i APPLICABILITY _: MODES 1, 2. 3 and 4.
.: ACTION:
a.
With one or more steam generators inoperable due to steam generator tube imperfections, restore the inoperable generator (s) to OPERABLE status prior to increasing T,yg above 200*F.
b.
Witn one or more steam generators inoperable due to the water level being outside the limits, be in at least HDT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOW'N within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
i SU:NEILLANCE REOUIREMENTS 4.~.5.0 Each steam generator shall be demonstrated OPERABLE by perfomance of tne following augmented inservice inspection program and the requirements
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of 5;ecification 4.0.5.
- 4..5.1 Steam Generator Sample Selection and Inspection - Each steam
- ererator sns11 oe cetermineo OPERAdLE ourino shutoown by selecting and irs
- :ecting at least the minimum number of steam generators specified in Ta:1e 4-4.1.
.4.5.2 Steam Generator Tube Sample Selection and Inspection - The steam generator tube minimum sample size, inspection result classification, and i
-tne corresponding action required shall be as specified in Table 4.4-2.
The inservice inspection of steam generator tubes shall be perfomed at the
- fre::vencies specified in Specification 4.4.5.3 and the inspected tubes shall
- te verified acceptable per the acceptance criteria of Specification 4.4.5.4.
- Tne tubes selected for each inservice inspection shall include at least 3%
,3 of the total number of tubes in all steam generators; the tubes selected for t tnese inspections shall be selected on a random basis except:
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The first sample inspection during each inservice inspection a.
(subsequent to the baseline inspection) of each steam generator snali include:
1.
All tubes or tube sleeves that previously had detectable wall penetrations (> 20%) that have not been plugged or repaired by sleeving in the affected area.
2.
At least 50% of the tubes inspected shall be in those i
areas wnere experience has indicated potential problems.
- .'IS-BESSE - UNIT 1 3/4 4-6 Amendment No. 27J71
Dockat-Nuxb r 50-346' Lic nsa Nuxbar NPF-3 Sarici Numb:r 2130 Attcchm:nt.
REACTOR COOLANT SYSTEM Pagt 13 I
SURVEILLANCE REQUIREMENTS (Continued) fg u.2 c
3.
Atubeinspection(pursuanttoSpecification4.4.5.4.ag shall be performed on each selected tube.
If any seletted tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an i
adjacent tube shall be selected and subjected to a tube
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inspection.
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b.
Tubes in the following groups may be excluded from the first random sample if all tubes in a group in both steam generators are inspected. No credit will be taken for these tubes in
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meeting minimum sample size requirements, r
1.
Group A-1: Tubes within one, two or three rows of the open inspection lane.
2.
Group A-2: Tubes having a dr'illed opening in the 15th support plate.
l 3.
Group A-3: Tubes included in the rectangle bounded by rows 62 and 90 and by tubes 58 and 76, excluding tubes included in Group A-1.*
c.
The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be e
subjected to less than a full tube inspection provided:
1 1.
The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found.
2.
The inspections include those portions of the tubes where imperfections were previously found.
The results of each sample inspection shall be classified into one of the following three categories:
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 the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.
1 C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are j
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defective.
l Tubes in Group A-3 shall not be excluded after completion of the fifth f
refueling outage.
DAVIS-EESSE, UNIT 1 3/44-7 Amendment No. 4+,113
Dock;t Nu:ber 50-346 NVj O l
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}g g; Liccnsa Numb 2r NPF-3 h
S: rial Numbar 2130 h
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REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
Notes:
(1)
In all inspections, previously degraded tubes must exhibit significar,t (> 10%) further wall penetrations to be included in the above percentage calculations.
(2) Where special inspections are performed pursuant to 4.4.5.2.b, defective or degraded tubes found as a result of the inspection shall be included in detemining the Inspection Results Category for that special inspection but need not be included in determining the Inspection Results Category for the general steam generator inspection.
4.4.5.3 Inspection Frecuencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:
The baseline inspection shall be performed to coincide with the a.
first seneduled refueling outage but no later than April 30. 1980.
Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection.
If the results of two consecutive inspec-tions for a given group
- of tubes following service under all volatile treatment (AVT) conditions fall into the C-1 category I
or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval for that group may be extended to a maximum of 40 mont.hs.
b.
If the results of the inservice inspection of a steam generator performed in accordance with Table 4.4-2 at 40 month intervals for a given group
- of tubes fall in Category C-3, subsequent inservice inspections shall be perfomed at intervals of not less than 10 nor more than 20 calendar months after the previous inspection. The increase in inspection frequency shall apply until a subsequent inspection meets the conditions specified in 4.4.5.3a and the interval can be extended to 40 months.
Additional, unscheduled inservice inspections shall be c.
performed on each steam generator in accordance with the first sa.-ple inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:
1.
Primary-to-secondary tube leaks (not including leaks originating from tube-to tube sheet welds) in excess of the limits of Specification 3.4.6.2.
- A group of tubes means:
(a) All tubes inspected pursuant to 4.4.5.2.b, or (b) All tubes in a steam generator less those inspected pursuant to 4.4.5.2.b.
A DAVIS-EESSE, UNIT I 3/4 4-8 Amendment No. 21
i Docket Nurber 50-346
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{. g i Liccnse Number NPF-3 Ser'ial Number 2130
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'5URVEILLANCE REOUIREMENTS (Continued)
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.1 2.
A seismic occurrence greater tnan tne Operating Basis j
j' Earthquake.
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3.
A loss-of-coolant accident requiring actuation of the engineered safeguards.
l 4.
A main steam line or feedwater line break.
I d.
The provisions of Specification 4.0.2 are not applicable.
4.4.5.4 Accectance Criteria a.
As used in this Specification:
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1.
Tubino or Tube means that portion of the tube or tube sleeve wnicn forms tne primary system to secondary system boundary.
2.
I oerfection means an exception to the dimensions. finisn or l
contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20t of the nominal tube wail thickness. if detectable, may be considered as imperfections.
3.
Cegradation means a service-induced cracking, wastage, wear 1
ll or general corrosion occurring on either inside or outside i
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of a tube.
- f 4.
Cecraded Tube means a tube containing imperfections > 20'; of tne nominal wall thickness caused by degradation that has not been repaired by sleeving in the affected area.
5.
% Deoradation means the percentage of the tube wall thickness l
af fecte0 or removed by degracation.
l
.i 6.
efect means an imperfection of such severity that it exceeds tne repair limit. A defective tube is a tube centaining a I
!.l defect that has not been repaired by sleeving in the affected j'
area or a sleeved tube that has a defect in the sleeve.
7.
Repair Limit means the imperfection depth at or beyond which tne tube snall be removed from service by plugging or repaireo by sleeving in the affected area cecause it may beco e unservice-able prior to the next inspection and is equal to 40% of the nominal tube wall thickness. The Babcock and Wilcox process described in Topical Report BAW-2120P will be used for sleeving.
B.
Unserviceable describes the condition of a tube if it leaks or l
contains a cefect large enough to affect its structural integ-rity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as g
specified in 4.4.5.3.c. above.
,i I
9.
Tube Insoection means an inspection of the steam generator l
l tube f rom tne point of entry completely. to the point of exit.
DAVI5-BE5SE. UNIT 1 3/4 4-9 Amendment No. 21.171 i
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Docket Nuxber 50-346 Licensa. Numb:r NPF-3 Seri'al Number 2130 7
, Attachment lPage16 :EACTOR COOLANT SYSTEM i
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5'JRVEILLANCE REOUTREMENTS^(Continued) g 10.
preservice insoection means an inspection of the full
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1engtn of eacn tuce in each steam generator perfomed by eddy current techniques prior to service to establish l
a baseline condition of the tubing. This inspection shall be perfomed prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
The steam generator shall be detennined OPERABLE af ter completing b.
the corresponding actions (plug or repair by sleeving in the affectec areas all tubes exceeding the repair limit and "all tubes containing through-wall cracks) required by Table 4.4-2.
2.4.5.5 Reports Following each inservice inspection of steam generator tubes.
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a.
the number of tubes plugged in each steam generator shall be s reported to the commission within 15 days.
The complete results of the steam generator tube inservice j
c.
inspec* ion shall be submitted on an annual basis in a report for the period in which this inspection was completed. This report shall include:
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1.
Number ano extent of tubes inspected.
2.
Location and percent of wall-thickness penetration for s
i each indication of an imperfection.
3.
Identification of tubes plugged or sleeved.
l Results of steam generator tube inspections which fall into t
c.
Category C-3 and require pr--t notification of the Comission i
shall be reported pur:u:nt 1: Sp:;ific tica 0.0.1 prior to resumption of plant operation. TM Mtten f:!b..; :f $1s report shall provide a description of investigations conducted to detemine cause of the tube degradation and corrective I
measures taken to prevent recurrence.
The steam oenerator'shall be demonstrated OPEPJ.BLE by verifying 4.4.5.6 steam generator level to be within limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.4.5.7 When steam generator tute inspection is perfomed as per Section 4.4.5.2, an additional but totally separate inspection shall be performed on special interest peripheral tubes in theThis vicinity of the secured internal auxiliary feedwater header.
testing snall only be recuired on the steam generator selected for inspection and the test shall require inspection only between
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A enement No. 8.27.f2.171 MV:5-BESSE. UNIT 1 3/4 a-10 m
! LGcLbt WIdh2r 30-346 m,
License Number NPF-3 ds A
hmen j
f Page 17.
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SURVEILCANCE REQUIREMENTS (Continued)
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the upper tube sheet and the 15th tube support plate. The tubes i
selected for inspection shall represent the entire cirenference of the steam generator and shall total at least 150 peripheral j
tubes.
4.4.5.8 Visual inspections of the secured internal auxiliary feedwater i
header, header to shroud attachment welas, and the external header thermal sleeves shall be performed on each steam generator through the auxiliary feedwater injection penetrations.
l t
These inspections shall be performed during the third and fourth i
refueling outages and at the ten-year 151.
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DAVIS-Bd5E, UNIT 1 3/44104 Amendment ho. 62 I
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'Q D*Tl4 H @.J M O Dockot Number 50-346 I
E tldL 5 51U W Wbw License Number NPF-3 l
!Q Serial Number 2130 OT"'
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DAY 15-BE55E. UNIT 1 3/4 4-11 i.,.i,.m..-
Docket Number 50-346 License Number NPF-3 i
f Serial Number 2130 Attachment
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I Assendment No. A'd.171 OAVIS-BESSE. UNIT 1 3/6 6-12 i
Docket Nurber'50-3I.6
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License Number NPF-3 g
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Serfal Number 2130 REACTOR COOLANT SYSTEM i
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3/4.4.4 PRESSURIZER l
A steam bubble in the pressurizer ensures that the RCS is not a nydraulically
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solic system and is capaole of accomodating pressure surges during operation.
j The steam buoble also protects tne pressurizer code safety valves and pilot i
c;erated relief valve against water relief.
The low level limit is based on providing enough water volu.e to prevent a
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reactor coolant system low pressure condition that would actuate the Reactor Protection System or tre Safety Feature Actuation System. The high level limit is cases on providing enough steam volume to prevent a pressurizer high level 3
j as a result of any transient.
i The oilot c:erated relief valve and steam bubble function to relieve RCS l
- ressure during all ces);n transients. Operation of the pilot o;erated relief l
valve minimizes the uncesiracle opening of the spring-loaded pressurizer code safety valves.
3/4.0.5 STEAM GENERATORS 1
1 The Surveillance Require ents for inspection of the steam generator tubes ensure i
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- trat the structural integrity of this portion of the RCS will be maintained.
Tne program for inservice ins;ection of steam generator tubes is based on a i
mocification of Regulatory Guide 1.83. Revision 1.
Inservice inspection of steam generator tuoing is essential in order to maintain surveillance of the j
concitions of the tubes in the event that there is evidence of mechanical damage j
or progressive degracation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator i
tuoing also provices a means of characterizing the nature and cause of any tube cegradation so that corrective measures can be taken. A process equivalent to
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the inspecticn methoc cescribed in Topical Report BAW-2120P will be used for i
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inservice inspection of steam generator tube sleeves. This inspection will provide ensurance of RC5 integrity.
i The plant is expected to be operated in a manner such that the secondary 4
j coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary j
coolant enemistry is not maintained within these chemistry limits, localized corrosion may likely result in stress corrosion cracking. The extent of i
, cracking during plant operation would be limited by the limitation of steam l
j generator tube leakage between the primary coolant system and the secondary l
coolant system (primary-to-secondary leakage = 1 GPM). Cracks having a i
primary-to-secondary leakage less than this limit during operation will have
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- an adequate margin of safety to withstand the loads imposed during normal y
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i CAVI5-BESSE. UNIT 1 B 3/4 4-2 Amenement No. 125.171 l
Docket Number 50-346 License Number NPF-3 Serial Number 2130
- Attachment l^Page21 EEACTOR COOLANT SYSTEM
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BASES (Continvec) operation and by postulated accidents. Operating plants have demonstrated that primary-to-seconcary leakage of 1 GPM can be detected by monitoring the
'a seconcary coolant. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will te located and plugged or repaired by sleeving in the affected areas.
Wastage-type defects are unlikely with proper chemistry treatment of the seconcary coolant. However, even if a defect should develop in service.
- it will be found during scheduled inservice steam generator tube examina-tions. As described in Topical Report SAW-2120P, degracation as small as 20, through wall can be detected in all areas of a tube sleeve except for tne roll expanded areas and the sleeve end, where the limit of detectability is 405 througn mall. Tubes with imperfections exceeding the repair limit of 405 of tne nominal wall thickness will be plugged or reoaired by sleeving
- ne affected areas. Davis-Besse will evaluato, and as appropriate implement, better testing metnocs wnich are developed and validated for comercial use so as to enable cetection of degradation as small as 20t through wall without exception.
Until sucn time as 20'; penetration can be detected in the roll expanced areas and the sleeve end, inspection results will te compared to those cotained during the baseline s ube inspection.
Wnenever the results of any steam 4 generator tubing inservice inspection fall into Category C-3, these results -,44-be p= ;tly reported to the Comission
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- _Pt te S;;;i'ic:ti;r. 5.0.1 prior to resumption of plant operation. Such cases will be consicered by the Comission en a case-by-case basis and may result in a requirenent for analysis, laboratory examinatiens, tests, i
aceitional eddy-current inspection, and revision of the Technical Specifica-tions, if necessary.
ine steam generator water level limits are consistent with the initial assumptiens in tne FSAR.
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OAVIS-EESSE, UNIT 1 B 3/4 4-3 Amencment No.171
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