ML20214K424
| ML20214K424 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 08/13/1986 |
| From: | Thadani A Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20214K414 | List: |
| References | |
| TAC-53219, NUDOCS 8608210075 | |
| Download: ML20214K424 (15) | |
Text
[f UNITED STATES
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NUCLEAR REGULATORY COMMISSION 5
j WASHINGTON, D. C. 20555
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OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285 FORT CALHOUN STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 99 License No. DPR-40 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by the Omaha Public Power District (the licensee) dated November 18, 1983, as superseded and clarified in letters dated June 17, 1985 and July 8, 1986, complies with the standards and re as amended (the Act)quirements of the Atomic Energy Act of 1954,
, and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with-10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
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. 2.
Accordingly, Facility Operating License No. DPR-40 is c. mended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B. of Facility Operating License No.
DPR-40 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.gg, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specificatior.3.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION M
g Ashok C. Thada
, Di e or PWR Project Directo a
- 8 Division of PWR Licen ing-B
Attachment:
Changes to the Technical Specifications Date of Issuance:
August 13, 1986 l
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ATTACHMENT TO LICENSE AMENDMENT N0. oo FACILITY OPERATING LICENSE NO. DPR-40 DOCKET NO. 50-285 Revise Appendix "A" Technical Specifications as indicated below. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
Remove Pages Insert Pages iii iii 3-25 3-25 3-28 3-28 3-29 3-29 Fig 5-1 Fig 5-1 5-4 5-4 5-5 5-5 5-7 5-7 5-9 5-9 5-12 5-12 5-13 5-14 5-14a-5-18 5-18 5-19 5-19 l
I' o
TABLE OF CONTENTS (Continued)
- Page, 4.3 Nuclear Steam Supply System (NSSS).......................
4-3 4.3.1 Reactor Coolant System............................
4-3 4.3.2 Reactor Core and Contro1.......................... 4-3 4.3.3 Emergency Core Cooling............................
4-3 4.4 Fu e l S to ra ge.............................................
4-4 4.4.1 New Fuel Storage..................................
4-4 4.4.2 Spent Fuel Storage................................
4-4 4.5 Seismic Design for Class I Systems.......................
4-5 5.0 ADMINISTRATIVE CONTR0LS........................................
5-1 5.1 Responsibility...........................................
5-1 5.2 O rga ni za ti on............................................ 5-1 5.3 Facility Staff Qualifications............................
5-la 5.4 T ra i n i n g................................................. 5 - 3 5.5 Review and Audit.........................................
5-3 5.5.1 -Plant Review Committee (PRC)......................
5-3 5.5.2 Safety Audit and Review Connittee (SARC)..........
5-5 5.5.3 Fire Protection Inspection........................
5-8a 5.6 Reporta bl e Event. Action.................................. 5-9 l
5.7 Safety Limit Violation...................................
5-9 5.8 Procedures...............................................
5-9 l
5.9 Repo rti ng Requi rements...................................
5-10 5.9.1 Rou ti ne Repo rts...................................
5-10 4
I 5.9.2 Repo rta bl e Events................................. 5-12 1
5.9.3 Spec i a l Reports...................................
5-15 5.9.4 Unique Reporting Requirements.....................
5-15 5.10 Records Retention........................................
5-18 5.11 Radiation Protection Program.............................
5-19 5.12 DELETED 5.13 Seconda ry Wa ter Chemi s try................................
5-20 5.14 Sy s tems I n teg ri ty........................................
5-21 5.15 Post-Accident Radiological Sampling and Monitoring.......
5-21 1
6.0 INTERIM SPECIAL TECHNICAL SPECIFICATIONS.......................
6-1 6.1 Limits on Reactor Coolant Pump Operation................. 6-1 o
i 6.2 Use of a Spent Fuel Shipping Cask........................ 6-1 6.3 Auxiliary Feedwater Automatic Initiation Setpoint........
6-1 6.4 Operation With Less Than 75% of Incore Detector i
S tri ngs 0pera bl e......,................................ 6-1 i'
iii Amendment No. 32,34,#3,5# 55,57, 73,89,86,Pl,99
3.0 SURVEILLANCE REQUIREMENTS 3.3 Reactor Coolant System, Steam Generator Tubes, and Other Components Subject to ASME XI Boiler & Pressure Vessel Code Inspection and Testing Surveillance (Continued)
Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.
(ii)
The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 3-8.
e.
Reporting Requirements (i)
Following each in-service inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission within 30 days.'
(ii)
The complete results of the steam generator tube inservice inspection shall be reported to the Commission within six (6) months following completion of the inspection.
This report shall include:
1.
Number and extent of tubes inspected.
2.
Location and percent of wall thickness penetra-tion for each imperfection.
3.
Identification of tubes plugged.
(iii)
Results of steam generator tube inspections which fall into Category C-3 and require prompt notification of the Commission shall be reported pursuant to Section 5.6 of l
the Technical Specifications prior to resumption of plant operation. The written followup of this report shall provide a description of investigations conducted to determine cause of the tube degradation and correc-tive measures taken to prevent recurrence.
(3) 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
- To satisfy ALARA requirements, leakage may be measured indirectly (as froa the perfomance 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 criteria.
3-25 Amendment No. 46, 0/dd/ difdd f/20/EJ, 99
TABLE 3-8 STEAM GENERATOR TUBE INSPECTION IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Samole Size Result Action Required Result Action Required Result Action Required
-1 N ne N/A N/A N/A N/A f
t bes per S.G.
C-2 Plug defective tubes C-1 None N/A N/A and inspect additional Plug defective tubes C-1 None 600 tubes in this S.G.
C-2 and inspect additional C-2 Plug defective tubes 1200 tubes' in this C-3 Perform action for S'G*
C-3 result of first sample w
C-3 Perform action for C-3 g
result of first sample The C-3 Inspect all tubes in this S.G., plug defec-jnd tive tubes and inspect
.is C-1 600 tubes in other S.G.
g m
The Perform action for y
Prompt notification second C-2 result of second N/A N/A to NRC pursuant to S.G.
sample m
E specification S.6 is C-2 l
=
P The Inspect all tubes in 3.
second the second S.G. and
?
S.G.
plug defective tubes.
N/A N/A g
is C-3 Prompt notification to NRC pursuant to i
specification 5.6 l
l N/A = Not Applicable.
3.0 SURVEILLANCE REQUIREMENTS 3.3 Reactor Coolant System Steam Generator Tubes, and Other Components!
Subject to ASME XI Boiler & Pressure Vessel Code Ir.spection and Testing Surveillance (continued)
Basis Undetected prolonged leakage of borated reactor coolant onto carbon steel j
sets up an unusual corrosion mechanism.. Detection of this leakage at an early stage can best be accommodated directly after an outage and before i
startup. The inspection program specified in Specification 3.3(1) places major emphasis on the areas of highest stress concentration as determined by general design evaluation and experience with similar systems. The-inspections will rely on non-destructive analysis methods utilizing up-to-date analyzing equipment and trained personnel.
Volumetric inspection of the reactor pressure vessel is to be performed completely from the outside diameter. The testing techniques and acceptance criteria of Section XI of.
i the ASME B&PV Code will be utilized, except where specific relief is granted by the Comission.
i The. surveillance requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-tained. The program for in-service inspection of the steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1, dated July 1975.
In-service inspection of steam generator tubing is essential in order i
to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or in-service conditions that lead to corrosion.
In-service inspection of steam generator tubing also provides a means of characterizing.the nature and cause of any tube degradation so that correc-tive measures can be taken.
L Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service,- it l
will be found during scheduled in-service steam generator tube examinations.
Plugging will-be required for all tubes with imperfections exceeding the plugging limit of 40% of the tube nominal wall thickness.
Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.
Whenever the results of any steam generator tubing in-service inspection fall into Category C-3, these results will be promptly reported to the Commission pursuant to Section 5.6 of the Technical Specifications prior to the resump-l tion of plant operation.
Such cases will be considered by the Commission on.
a case-by-case basis and may result in a requirement for analysis, laboratory i
examinations, tests, additional eddy-current inspection, and revision of the i
Technical Specifications, if necessary.
References (1) FSAR, Section 4.5.3 3-29 Amendment No. 46, 9fdd/ Adidd 4/29/BJ, gg i
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- OTHER VICE-PRESIDENT POSITIONS EXIST WHICH HAVE NO DIRECT INVOLVEMENT WITH FACILITY MANAGEMENT OR TECHNICAL SUPPORT.
5.0 ADMINISTRATIVE CONTROLS Responsibilities 5.5.1.6 The Plant Review Connittee shall be responsible for:
Review of 1) all procedures required by Specification 5.8 and a.
changes thereto, 2) any other proposed procedures or changes thereto as determined by the Manager - Fort Calhoun Station to affect nuclear safety.
b.
Review of all proposed tests and experiments that affect nuclear safety.
Review of all proposed changes to the Technical Specifications, c.
d.
Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety.
e.
Investigation of all' violations of the Technical Specifications and shall prepare and forward a report covering evaluation and recommendations to prevent recurrence to the Division Manager -
Nuclear Production and to the Chairman of the Safety Audit and Review Connittee.
f.
Review of facility operations to detect potential safety hazards.
g.
Performance of special reviews and investigations and reports thereon as requested by the Chairman of the Safety Audit and Review Connittee.
h.
Review of the Site Security Plan and implementing procedures and shall submit recommended changes to the Chairman of the Safety Audit and Review Committee.
i.
Review of the Site Emergency Plan and implementing procedures and shall submit recommended changes to the Chairman of the Safety Audit and Review Canmittee.
j.
Review of all Reportable Events.
Authority 5.5.1.7 The Plant Review Committee shall:
a.
Recommend in writing to the Manager - Fort Calhoun Station approval or disapproval of items considered under 5.5.1.6.(a) through (d) above.
5-4 Amendment No. 9,JS,$A, 99
5.0 ADMINISTRATIVE CONTROLS 5.5.1.7 b.
Render determinations in writing with regard to whether or not each item considered under 5.5.1.6(a) through (e) above constitutes an unraviewed safety question.
c.
Provie
'mmediate written notification to the Division Manager -
Nucle
.roduction and the Safety Audit and Review Committee of disa9. ement between the Plant Review Committee and the Manager -
Furt. Celhoun Station; however, the Manager - Fort Calhoun Station shall have responsibility for resolution of such disagreements pursuant to 5.1.1 above.
Records 5.5.1.8 The Plant Review Committee shall maintain written minutes of each meeting and copies shall be provided to the Division Manager - Nuclear Production and Chairman of the Safety Audit and Review Committee.
5.5.2 Safety Audit and Review Committee (SARC)
Function 5,JJ.4 The Safety Audit and Review Committee shall function to provide the independent review and audit of designated activities in the areas of:
a.
nuclear power plant operation b.
nuclear engineering c.
chemistry and radiochemistry d.
metallurgy e.
instrumentation and control f.
radiological safety g.
mechanical and electrical engineering h.
quality assurance Composition 5.5.2.2 The Safety Audit and Review Committee shall be composed of:
Chairman: Division Manager - Quality Assurance and Regulatory Affairs Member:
Vice President - Nuclear Production, Production Operations, Fuels, and QA&RA Member:
Vice-President,- Engineering and General Services l
Member:
Division Manager - Engineering Member:
Division Manager - Nuclear Production Member:
OPPD Operations, Engineering, and Technical Support staff Member:
Qualified Non-District Affiliated Consultants as Required and as Determined by SARC Chairman 5-5 Amendment No. 7,77,60,75,M,M,99 i
5.0 ADMINISTRATIVE CONTROLS 5.5.2.7 c.
Proposed tests or experiments which involve an unreviewed safety question as defined in 10 CFR 50.59.
d.
Proposed changes in Technical Specifications or licenses.
e.
Violations of applicable statutes, codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.
f.
Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety.
g.
All Reportable Events.
h.
Any indication of an unanticipated deficiency in some aspect of design or operation of safety related structures, systems, or components.
1.
Reports and meeting minutes of the Plant Review Committee.
The Chairraan of the Safety Audit and Review Committee (SARC) may designate subgroups, special working comittees, or audit teams as he deems necessary in order to carry out the responsibilities of the SARC. These subgroups, comittees, or audit teams will perform the SARC responsi-bilities and report on their activities for review at the next regularly scheduled SARC meeting following any group's action.
Audit 5.5.2.8 Audits of facility activities shall be performed under the cognizance of the Safety Audit and Review Committee. These audits shall encompass:
a.
The conformance of facility operation to all provisions contained within the Technical Specifications and applicable license condi-tions at least once per year.
b.
The performance, training and qualifications of the entire facility staff at least once per year.
c.
The results of all actions taken to correct deficiencies occurring in facility equipment, structures, systems or method of operation that affect nuclear safety at least once per six months.
d.
The performance of all activities required by the Quality Assurance Program to meet the criteria of Appendix B,10 CFR Part 50, at least I
once per two years.
5-7 Amendment No. 50,99
5.6 Reportable Event ~ActionT -
l 5.6.1 The following actions shall be taken in the event of a REPORTABLE EVENT:
[
a.
The Commission shall be notified pursuant to the requirements of 10 CFR 50.72, if applicable.
-b.
Each Reportable Event shall be reviewed by the Plant Review Committee and submitted to the Chairman of the Safety Audit and Review Committee and the Division Manager - Nuclear Production.
c.
Submit reports of Reportable Events pursuant to the requirements l
of Specification 5.9.2.
I 5.7 Safety Limit Violation 5.7.1 The following actions shall be taken in the event a Safety Limit is violated:
a.
The unit shall.be placed in at least HOT SHUTDOWN within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
l b.
The Safety Limit Violation shall be reported to the Division Manager - Nuclear Production and to the Chairman of the Safety Audit and Revi2w Committee (SARC) within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
c.
A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the Plant Review Committee. This report shall describe (1) applicable circumstances preceding the viola-tion, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.
d.
The Safety Limit Violation Report shall be submitted to the Chainnan Of the Safety Audit and Review Committee and the Division Manager - Nuclear Production within 14 days of the violation.
5.8 Procedures 5.8.1 Written procedures and administrative policies shall be established, implemented and maintained that meet or exceed the minimum requirements of sections 5.1 and 5.3 of ANSI N18.7-1972 and Appendix A of USNRC Regulatory Guide 1.33 except as provided in 5.8.2 and 5.8.3 below.
5.8.2 Each procedure and administrative policy of 5.8.1 above, and changes thereto, shall be reviewed by the Plant Review Committee and approved by the Manager - Fort Calhoun Station prior to implementation and periodically as set forth in each document.
5.8.3 Temporary changes to procedures of 5.8.1 above may be made provided:
5-9 Amendment No. 9,J9,38,$A,99
5.9.1 Continued work and job functions,E e.g., reactor operations and surveil-lance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling outages. The dose assignment to various duty functions may be estimates based on pocket dosimeter TLD, or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for.
In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions.
c.
Monthly Operating Report. Routine reports of operating statis-tics and shutdown experience shall be submitted on a monthly basis to the Director, Office of Inspection and Enforcement, l
U. S. Nuclear Regulatory Commission, Washington, D. C. 20555, with a copy to the appropriate Regional Office, to arrive no later than the fifteenth of each month following the calendar month covered by the report. This monthly report shall also include a statement regarding any challenges or failures to'the pressurizer power operated relief valves or safety valves occurring during the subject month.
5.9.2 Reportable Events l
A Licensee Event Report (LER) shall be submitted to the U.S. Nuclear Regulatory Consnission, Document Control Desk, Washington, D. C. 20555 with a copy to Region IV of the NRC, within 30 days after discovery of any event meeting the requirements of 10 CFR 50.73.
3/ This tabulation supplements the requirements of 5 20.407 of 10 CFR Part 20.
5-12 Amendment No. 9,U,5,5,99 (Next page is 5-15) wr-m-~
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5.10 Record Retention 5.10.1 The following records shall be retained for at least five years:
Records and logs of facility operation covering time a.
interval at each power level, b.
Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.
c.
Licensee Event Reports (LER).
l d.
Records of surveillance activities, inspections and calibrations required by these Technical Specifications.
e.
Records of reactor tests and experiments.
f.
Records of changes made to Operating Procedures.
I g.
Records of radioactive shipments, l
h.
Records of annual physical inventory of all source material of record.
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5-18 Amendment No. 75, gg l
i 5.0 ADMINISTRATIVE CONTROLS 5.10.2 The following records shall be retained for the duration of the Facility Operating License:
4 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 radia-tion 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 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.
1.
Records of Quality Assurance activities required by the QA Manual.
j.
Records of reviews performed for changes made to procedures or equip-ment or reviews of tests and experiments pursuant to 10 CFR 50.59.
k.
Records of meetings of the Plant Review Comittee and the Safety Audit and Review Committee.
1.
Records of Environmental Qualification of Electric Equipment pursuant to 10 CFR 50.49.
m.
Records of.the service lives of all hydraulic and mechanical snubbers listed on Table 2-6(a) and (b) including the date at which the service life comences and associated installation and maintenance records.
.n.
Records of analyses required by the Radiological Environmental Monitoring Program.
J 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 bundle remains in Region 2 (reference Technical Specifications 2.8(12) and 5.8.4).
3 5.11 Radiation Protection Program i
~
Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.
5-19 9/dd/ difdd 19/ S/SS, Amendment No.
59,86.93, 99 4
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