ML20059E901
| ML20059E901 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 12/30/1993 |
| From: | Hopkins J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20059E904 | List: |
| References | |
| NUDOCS 9401130035 | |
| Download: ML20059E901 (23) | |
Text
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,'o UNITED STATES g
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NUCLEAR REGULATORY COMMISSION 9 g:
WASHINGTON. D. C. 20555 U
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TOLED0 EDISON COMPANY CENTERIOR SERVICE COMPANY MD THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DOCKET NO. 50-346 DAVIS-BESSE NUCLEAR POWER STATION. UNIT NO. I MENDMENT TO FACILITY OPERATING LICENSE Amendment No.184 License No. NPF-3 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Toledo Edison Company, Centerior Service Company, and the Cleveland Electric Illuminating Company (the licensees) dated May 6,1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission'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 Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of 3
the public, and (ii) that such activities will be conducted in compliance with the Commission'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 Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-3 is hereby amended to read as follows:
l 9401130035 931230 PDR ADDCK 05000346 l
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. (a) Technical Specifications i
The Technical Specifications contained in Appendix A, as revised through Amendment No.184, are hereby incorporated in the license.
The Toledo Edison Company shall operate the facility in accordance 1
with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented not later than 90 days after issuance.
FOR THE NUCLEAR REGULATORY COMMISSION l
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/ on B. Hopkins, Sr. Project Manager Project Directorate III-3 i
Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation i
Attachment:
Changes to the Technical Specifications Date of issuance:
December 30, 1993 1
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ATTACHMENT TO LICENSE AMENDMENT NO. 184 FACILITY OPERATING LICENSE N0. NPF-3 DOCKET N0. 50-346 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain 1
vertical lines -indicating the area of change. The corresponding overleaf pages are also provided to maintain document completeness.
Egmove Insert l-6a 1-6a 3/4 3-57 3/4 3-57 3/4 4-7 3/4 4-7 3/4 4-10 3/4 4-10 3/4 4-12 3/4 4-12 3/4 11-1 3/4 11-1 B 3/4 4-3 B 3/4 4-3 6-7 6-7 6-17a 6-17a 6-22 6-22 i
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i Op DEFINITIONS 1.29 Deleted PROCESS CONTROL PROGRAM 1.30 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure 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 0FFSITE DOSE CALCULATION MANUAL (ODCM) 1.32 The OFFSITE DOSE CALCULATION MANUAL (ODCM) 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 Alarm / 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 Environmental Monitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Reports required by Specification 6.9.1.10 and 6.9.1.1:1.
1.33 Deleted l.34 Deleted l.3S Deleted I
1.36 Deleted
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i 11 EMBER (5) 0F THE PUBLIC l.37 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupation-ally associated ith the plant. This category does not include employees of the utility, its contractors or vendors. Also excluded from this category i
are persons who enter the site to servic,e equipment or to make deliveries.
This category does include persons who use portions of the site for recrea-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 1
i owned, nor leased, nor otherwise controlled by the licensee.
DAVIS-BESSE, UNIT 1 1-6a Amendment No. M,jpy6,184 i
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9 DEFINITIONS UNRESTRICTED AREA 1.39 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, conmercial, institutional, and/or recreational purposes.
The definition of UNRESTRICTED AREA used in implementing the Radiological. Effluent Technical Specifications has been expanded over that in 10 CFR 20.3(a)(17). The UNRESTRICTED AREA boundary may coincide with the exclusion (fenced) area boundary, as defined in 10 CFR 100.3(a), but the UNRESTRICTED AREA does not include areas over water bodies.
The concept of UNRESTRICTED AREAS, estab-lished at or beyond the SITE BOUNDARY, is utilized in the LIMITING CONDITIONS FOR OPERATION to keep levels of radioactive materials in liquid and gaseous effluents as low as is reasonably achievable, pursuant to 10 CFR 50.36a.
1.40 Deleted CORE OPERATING LIMITS REPORT 1.41 The CORE OPERATING LIMITS REPORT is the unit-specific document that provides core operating limits for the current reload cycle.
These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.7.
Plant operation within these core operating limits is addressed in individual specifications.
I DAVIS-BESSE, UNIT 1 1-6b Amendment No. EE,J//,170 i
9 INSTRUMENTATION 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.
ACTION:
With the waste gas system oxygen monitor alam setpoint less a.
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conservative than required by the above Specifications, declare the channe,1 inoperable and comply with ACTION b.
b.
With the waste gas system oxygen monitor inoperable, additions to the waste gas surge tank may continue provided another method for ascertaining oxygen concentrations, such as grab sample analysis, is implemented to provide measurements at least once per four (4) 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 next Radioactive Effluent Release Report why the inoperability l
was not corrected in a timely manner.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
c.
SURVEILLANCE REQUIREMENTS 4.3.3.9 The waste gas system ox sen monitor shall be demonstrated OPERABLE by:
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 a.
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additions to the waste gas surge tank.
b.
At least once per 92 days by performance of a CHANNEL CALIBRATION.
' samples containing a nominal:The CHANNEL CALIBRATION shall i.
1.
One volume percent oxygen..' balance nitrogen; and 2.
Four volume percent oxygen, balance nitrogen.
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20% of the nominal wall thickness caused by degradation thaE has not been repaired by sleeving in the affected' area.
5.
% Degradation means the percentage of the tube wall thickness
__affected or removed by degradation.
i 6.
Defect means an, imperfection of such severity that it exceeds the repair limit A defective tube is a tube containing a l
defect that has not been repaired by sleeving in the affected l
area or a sleeved tube that has a defect in the sleeve.
i 7.
Repair Limit means 'the imperfection depth at or beyond which the tube shall be removed from service by plugging or repaired by sleeving in the affected area because it may become unservice-able prior to the next inspection and is equal tr 40% of the.
nominal tube wall thickness. The Babcock and Wilcox frocess described in Topical Report BAW-2120P will be used for sleeving.
8.
Unserviceable describes the condition of a tube. if it leaks or-l contains a defect 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 specified in 4.4.5.3.c, above.
9.
Tube Inspection means an inspection of the steam generator-l tube from the point of entry completely to the point of exit.
DAVIS-BESSE, UNIT 1 3/4 4-9 Amendment No. 21,171 l
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t2 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION l
v RM 1ST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION E
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Sample Size Result Action Required Result Action Required Result Action Required A minimum of C-1 None N/A N/A N/A N/A S Tubes per S.G. (1)
C-2 Plug or repair by C-1 None N/A N/A sleeving defective tubes and inspect C-2 additional 2S h Plug or repair by C-1 None in this S.G.
s% defective tubes and inspect C-2 Plug or repair by additional 4S tubes sleeving de%ctive in this S.G.
tubes Perform action for R
'g C-3 C-3 ra=Jt o* first A
sample A1 C-3 W am W N/A N/A ro C-3 result of first sample C-3 Inspect att tubes in Alt other this S.G., plug or S.GE are None N/A N/A y
repair by sleeving C-1 El defective tubes and inspect 2S tubes in Q
each other S.G; Some S.G.s Perform action for FD Report to the NRC C-2 but no C-2 result of second N/A N/A 5
prior to resumption additional sample 2
of plant cpo.diksi.
S.G. are o,
C-3 y
Additional inspect all tubes in S.G. Is each S.G. and plug or D
C-3 repair by slee4 N/A N/A w
derettve tubes.
'y Report M the NRC a
prior to resumption of plant operation.
J (1) S = 3"% Where N is the number of steem Oenerators in the unit, and n is the number of sleem generators Inspected during an k %i;c6.
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TABLE 4.4-1 z
MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION i
Preservice inspection.
No Yes w}
No. of Steam Generators per Urut Two Three Four Two Three Four
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First inservice Inspection All One Two Two I
I One?
One3 Secosul & Subsequent Inservice Inspecticns One One Table Notation:
- f. The inservice inspection may be limited to one steam generator on a rotating schedule encompassing 3 N % of the tubes (where N is the number of steam generators in the plant) if the results of the first,x previous instr 1 sons irxhcate that all stearn generators are performing in a hke snanner. Note that urnier some circumstances, the operatmg corwistions m one or more steam generators may be found to be more severe than those in other steam generators. Under such circum starices the sampla sequence shall be mothfied to inspect the most severe corxletions.
- 2. The other steam generator not inspected during the first inservice inspection shall be inspected. The third and subsequent inspections should follow the instructions described in 1 above.
- 3. Each of the other two steam generators not inspected during the first inservice inspections shall be inspected during the secoexl ami tinird inspections. The fourth arul sirbsequent inspections shall follow the instructions described in 1 above.
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Q RADI0 ACTIVE EFFLUENTS LIQUID HOLDUP TANKS
- LIMITING CONDITION FOR OPERATION 3.11.1 The quantity of radioactive material contained in each of the following unprotected outdoor tanks shall be limited to less than or equal to 10 curies, excluding tritium and dissolved or entrained noble gases.
a.
Outside temporary tank.
APPLICABILITY: At all times.
ACTION:
With the quantity of radioactive material in any of the above a.
listed tanks exceeding the above limit, immediately suspend all 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 describe the event leading to this condition in the next Radioactive Effluent Release ~Re~ ort.
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p b.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.1 The quantity of radioactive material contained in each of the above listed tanks shall be deter 6ined to be within the above limit by analyzing a representative sample of the tank contents at least once per 7 days when radioactive materials are being added to the tank.
- 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 the liquid radwaste treatment system.
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DAVIS-BESSE, UNIT 1 3/4 11-1 Amendment No. EE. Y77.184
O-l RADI0 ACTIVE EFFLUENTS l
EXPLOSIVE GAS MIXTURE (Hydrogen rich systems not designed to withstand a hydrogen explosion)
LIMITING CONDITION FOR OPERATION 3.11.2 The concentration of oxygen in the waste gas system shall be l
limited to less than or equal to 2% by volume whenever the hydrogen concentration exceeds 4% by volume.
APPLICABILITY: At all times.
k ACTION:
a.
With the concentration of oxygen in the waste gas system greater than 2% by volume but less than or equal to 4%
j by volume, reduce the oxygen concentration to the above limits,
~ ithin 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
w b.
With the concentration of oxygen in the waste gas system greater than 4% by volume and the hydrogen concentration greater than 4%
by volume, immediately suspend all additions of waste gases to the system and recuce the concentration of oxygen to less than or equal to 2% by volume without delay.
c.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2 The concentrations of oxygen in the waste gas system shall be l
determined to be within the above limits by monitoring tne waste gases in the waste gas system as required by Specification 3.3.3.9.
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t DAVIS-BESSE, UNIT 1 3/4 11-2 Amendment No. gg, 170 i
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REACTOR COOLANT SYSTEM BASES (Continued) operation and by postulated accidents.
Operating plants have demonstrated that primary-to-secondary leakage of 1 GPM can be detected by monitoring the secondary coolant.
Leakage in excess of this limit will require plant shutdown and an unscheduled inspe: tion, during which the leaking tubes will be located and plugged or repaired by sleeving in the affected areas.
Wastage-type defects are unlikely with proper chemistry treatment of the secondary 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 BAW-2120p. degradation as small as i
l 20% through wall can be detected in all areas of a tube sleeve except for
, the roll expanded areas and the sleeve end, where the limit of detectability i; is 40!; through wall.
Tubes with imperfections exceeding the repair limit of
- 40% of the nominal wall thickness will be plugged or repaired by sleeving
,: the affected areas.
Davis-Besse will evaluate, and as appropriate implement.
', better testing methods which are developed and validated for commercial use
- so as to enable detection of degradation as small as 20% through well without i
exception.
Until such time as 20% penetration can be detected in the roll
- expanded areas and the sleeve end, inspection results will be compared to those j; obtained during the baseline sleeved tube inspection.
' Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results shall be reported to the Commission
, prior to resumption 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 examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.
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!! The steam generator water level limits are consistent with the initial '
!I: assu ptions in the FSAR.
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DAVIS-BESSE, UNIT 1 B 3/4 4-3 Amendment No.171,184
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a.
9 REACTOR COOLANT SYSTIM BASES 3/4.4.6 REACTOR COOLANT SYSTEM LEAGGE 3/4.4.6.1 LEAUGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to detect and eenitor leakage from the Reactor Coolant Pressure Boundary.
These detection systems are consistent with the recocrnendation of Regulatory Guide 1.45. " Reactor Coolant Pressure Boundary Leakage Detection Systems,! May 1973.
3/4.4.6.2 OPERATIONAL LEAKAGE PRESSURE BOUNDARY LEAKAGE of any magnitude is uaacceptable since it may be indicative of an impending gross failure of the pressure boundary.
Therefore, the presence of any PRESSURE BOUNDARY LEMAGE requires the unit to be promptly placed in COLD SHi/TDOWN.
Industry experience has shown that, while a ' limited amount of leakagt is expected from the RCS. the UNIDENTIFIED LEAKAGE portion of this can be reduced to a threshold value of less than 1 GPM. This threshold value is sufficiently low to ensure early detection of additional feakage.
The total steam generator tube leakage limit of 1 GPM for all steam generators ensures that the dosage contribution from tube leakage will be limited to a small fraction of 10 CFR Part 100 limits in the event of eitner a steam generator tube rupture or steam line break. The 1 GPM limit is consistent with the assumptions used in the analysis of these accidents.
The 10 GPM IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.
The CONTROLLED LEAXAGE limit of 10 GPM restricts operation with a total RCS leakage from all RC ptrnp seals in excess of 10 GPM.
The surveillance requirements for RCS Pressure Isolation Valves provide added assurance of valve integrity thereby reducing the prob-ability of gross valve failure and consequent intersystem LOCA.
from the RCS Pressure Isolation Valves is IDENTIFIED LEAKAGE and will beLeakage considered as a portion of the allowed limit.
DMIS-lESSE tMIT 1 g 3/4 4 4 Amendment No.180
f ADMINISTRATIVE CONTROLS d.
Review of proposed tests or experiments determined to involve an unreviewed safety question as defined in 10 CFR 50.59.
Review of reports of violations of codes, regulations, orders, e.
Technical Specifications, or Operating License requirements having nuclear safety significance or reports of abnortnal degradation of I
systems designed to contain radioactive material.
f.
Review of all proposed changes to the Technical Specifications or the Operating License.
g.
Deleted h.
Review of reports of significant operating abnormalities or devi-ations from normal and expected performance 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.
1.
Investigations or analyses of special subjects as requested by the Company Nuclear Review Board.
m.
Review of all REPORTABLE EVENTS.
Review of all Safety LimitMiolation Reports (Section 6.7).
n.
Review of any unplanned,# 5ccidental or uncontrolled radioactive o.
releases, evaluation of the event, ensurance that remedial action is identified to prevent recurrence, review of a report covering the evaluation and forwardirig of the report to the Plant Manager and to the CNRB.
p.
Review of the changes to the OFF61TE DOSE CALCULATION MANUAL.
q.
Review of the changes to the PROCESS CONTROL PROGRAM.
Review of the Annual Radiological Environmental Operating Report.
r.
s.
Review of the Radioactive Effluent Release Report.
T.
Review of the Fire Protection Program and changes thereto.
DAVIS-BESSE, UNIT 1 6-7 Amendment No. 7 E
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a ADMINISTRATIVE CONTROLS AUTHORITY 6.5.1.7 The Station Review Board shall:
Recommend to the Plant Manager written approval or disapproval a.
of items considered under Sections 6.5.1.6 a, c, d, f, i and j.
l b.
Render determinations in writing with regard to whether or not each item considered under Sections 6.5.1.6 a, c, d and f above constitutes an unreviewed safety question as defined in 10 CFR 50.59.
c.
Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.co the Vice l
President, Nuclear and the Company Nuclear Review Board of disagreement between the SRB and the Plant Manager; however, the Plant Manager shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.
d.
Make recommendations in writing to the Plant Manager that actions reviewed under Section 6.5.1.6 b above do not constitute an unreviewed safety question.
RECORDS 6.5.1.8 The Station Review Board shall maintain written ninutes of each
- teeting and copies shall be provided to the Plant Manager Vice President, l
Nuclear and Chairman of the Company Nuclear Review Board.
6.5.2 COMPANY NUCLEAR REVIEW BOARD (CNRB)
FUNCTION 1
6.5.2.1 The Company Nuclear Review Board (CNRB) shall function to
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provide independent review and audit of designated activities in the areas of:
a.
Nuclear power plant operations.
b.
Nuclear engineering, c.
Chemistry and radiochemistry, d.
Metallurgy, e.
Instrumentation and control, f.
Fadiological safety, g.
Mechanical and electrical engineering, and
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h.
Quality assurance practices.
l DAVIS-BESSE, UNIT 1 6-8 Amendment No. 27,ff pg, 109
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ADMINISTRATIVE CONTROLS l
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 summaries, i
interpretations, and analysis of trends of the results of the Radiological-Environmental Monitoring Program for the reporting period. The material 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.
RADIDACTIVE EFFlllENT RELEASF RFPORT 6.9.1.11 The Radioactive Effluent Release Report covering the operation of the unit shall be submitted in accordance with 10 CFR 50.36a. The report shall include a summary 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 conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.
4 a
DAVIS-BESSE, UNIT 1 6-17a Amendment No. E5,73.T70.184
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ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued) 8)
BAW-10119, " Power Peaking Nuclear Reliability Factors," June 1977 The methodology for Rod Program received NRC approval in the Safety Evaluation dated January 11, 1990.
The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
The CORE OPERATING LIMITS REPORT, including any mid-cycle revision or supple-ments thereto, shall be provided upon issuance for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
f DAVIS-BESSE, UNIT 1 6-17 Amendment No.144
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ADMINISTRATIVE CONTROLS 6.14 PROCESS CONTROL PROGRAM (PCP)
Changes to the PCP:
a.
Shall be documented and records of reviews performed shall be retained as required by Specification 6.10.2.p.
This documentation shall contain:
))
Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s),
and 2)
A determination that the change will maintain the overall conformance 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 SRB and the approval of the Plant Manager.
6.15 0FFSITE DOSE CALCULATION MANUAL (0DCM)
Changes to the ODCM:
a.
Shall be documented and records of reviews performed shall be retained as required by Specification 6.10.2.p.
This documentation shall contain:
1)
Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s),
and 2)
A determination tha 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 SRB and the approval of the Plant Mana*ger.
Shall be submitted to the Commission in t'he form of a complete, c.
legible copy of the entire ODCM as part of or concurrent with the Radioactive Effluent Release Reoort for the Deriod of
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l 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 (e.e area of the page that was changed, and shall th indicate the date g., month / year) the change was implemented.
DAVIS-BESSE, UNIT 1 6-22 Amendment No. EE, 110.184
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ADMINISTRATIVE CONTROL 6.13 ENVIRONMENTAL QUALIFICATION 6.13.1 By no later than June 30, 1982 all safety-related electrical equipment in the facility shall be qualified in accordance with the provisions of Division of Operating Reactors " Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors"(DOR guidelines); or, NUREG-0588 " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment",
December 1979.
Copies of these documents are attached to Order for Modi-fication of License NPF-3 dated October 24, 1980.
6.13.2 By no later than December 1,1980, complete and auditible records must be available and maintained at a certain location which describe the environmental qualification method used for all safety-related electrical equipment in sufficient detail to document the degree of compliance with
[
the D0R Guidelines or NUREG-0588.
Thereafter, such records should be updated and maintained current as equipment is replaced, further tested, or otherwise further qualified; 1
l DAVIS-BESSE, UNIT 1 6-21 Order dated October 24, 1980 m