ML20138E484

Amend 93 to License DPR-40,changing Titles of Senior Mgt Officials & Deleting Environ Qualification of Electrical Equipment Administrative Requirements

ML20138E484
Person / Time
Site:	Fort Calhoun
Issue date:	12/06/1985
From:	Thadani A Office of Nuclear Reactor Regulation
To:
Shared Package
ML20138E477	List: ML20138E473 ML20138E484 ML20138E490
References
NUDOCS 8512130460
Download: ML20138E484 (12)

Text

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UNITED STATES

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}e NUCLEAR REGULATORY COMMISSION W ASHINGTON, D. C. 20555 p

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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. 93 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 July 11, 1985 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), 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 comon defense and security or to the health and safety of the public; and E. .,e issuance of this amendment is in accordance with 10 CFR Part

!( of the Comission's regulations and all applicable requirements have been satisfied.

, 8512130460 851206 5 Drt ADOCK 0500 l l

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2. Accordingly, Facility Operating License No. DPR-40 is amendell 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. 93, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical

- Specifications.

3. This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION (D6 o dw^

Asho C. Thadani, Director PWR roject Directorate #8 Division of PWR Licensing-B

Attachment:

Changes to the Technical Specifications Date of Issuance: December 6, 1985

1'&

ATTACHMENT TO LICENSE AMENDMENT N0.93 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 11 11 iii iii 2-69 2-69 3-13 3-13 5-5 5-5 5-8 5-8 5-19 5-19 5-20 5-20 Figure 5-1 Figure 5-1

. - - - = _ _ _ . - .

TABLE OF CONTENTS (Continued)

Page 2.12 Co ntrol Room Sys tems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7. . . . . 2- 59 2.13 Nuclear Detector Cooling System. . . . . . . . . . . . . . . . . . . . q

.. . . 2-60 2.14 Engineered Safety Features System Initiation Instrumentation Setti ngs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-61

, 2.15 Instrumentation and Control Systems...................... 2-65 2.16 River Leve1.............................................. 2-71 2.17 Miscellaneous Radioactive Material Sources............... 2-72 l 2.18 Shock Suppressors (Snubbers) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-73 2.19 Fi re P rotection Sys tem. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-89 2.20 Steam Generator Coolant Radioactivity.................... 2-96 i 2.21 Post-Accident Monitoring Instrumentation. . . . . . . . . . . . . . . . . 2-97 2.22 To x i c Ga s Mo n i to rs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2- 99 l

3.0 SURVEILLANCE REQUIREMENTS...................................... 3-1 3.1 Instrumentation and Control.............................. 3-1 3.2 Equipment and Sampling Tests............................. 3-17 3.3 Reactor Coolant System, Steam Generator Tubes, and Other Components Subject to ASME XI Boiler and Pressure Vessel Code Inspection and Testing Surveillance........................................... 3-21 3.4 Reactor Cool ant System Integri ty Testing. . . . . . . . . . . . . . . . . 3-36 i- 3.5 Containment Test......................................... 3-37

3.6 Safety Injection and Containment Cooling Systems i

Tests.................................................. 3-54 3.7 Emergency Power System Periodic Tests. . . . . . . . . . . . . . . . . . . . 3-58 3.8 Main Steam Isolation Va1ves.............................. 3-61 3.9 Auxili ary Feedwater System. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-62 3.10 Reactor Core Parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-63 -

3.11 Radiological Environmental Monitoring Programs........... 3-64 3.12 Radiological Waste Sampling and Monitoring. . . . . . . . . . . . . . . 3-69 3.12.1 Liquid and Gaseous Effluents..................... 3-69 3.12.2 Solid Radioactive Waste.......................... 3-71a 3.13 Radioactive Material Sources Surve111ance. . . . . . . . . . . . . . . . 3-76 3.14 Shock Suppressors (Snubbers)............................. 3-77 3.15 Fire Protection System................................... 3-80

4.0 DESIGN FEATURES................................................ 4-1 4.1 Site.....................................................4-1

. 4.2' Containment Design Features.............................. 4-1 I

4.2.1 Containment Structure............................ 4-1

4.2.2 Penetrations..................................... 4-1 4.2.3 Containment Structure Cooling Systems............ 4-2 11 Amendment No. 78,47,46,54,69.93 l 84, M.

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TABLE OF CONTENTS (Continued)

Page 4.3 Nuclear Steam Supply System (NSSS)................n ..... 4-3 4.3.1 Reactor Coolant System............................ 4-3 4.3.2 Reactor Core and Control.......................... 4-3 4.3.3 Emergency Core Cooling............................ 4-3 4.4 Fu el S to ra g e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 - 4 4.4.1 New Fu el Stora ge . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 -4 4.4.2 Spent Fuel Storage................................ 4-4 4.5 Sei smic Design for Cl as s I Systems . . . . . . . . . . . . . . . . . . . . . . . 4-5 l 5.0 ADMINISTRATIVE CONTR0LS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.1 Re s po n s i b i l i ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 1 5.2 O rg a n i za t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.3 Facili ty Staff Quali fications . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-l a 5.4 Training................................................. 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 Reportable Occurrence Action............................. 5-9 5.7 Safety Limit Violation................................... 5-9 5.8 Procedures............................................... 5-9 5.9 Repo rti ng Requi rements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-10 5.9.1 Routi ne Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 10 5.9.2 Reportable Occurrences . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-12 5.9.3 Speci al 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 l 5.13 Secondary Water Chemistry................................ 5-20 5.14 Systems Integrity........................................ 5-21 5.15 Post-Accident Radiological Sampling and Monitoring....... 5-21 6.0 INTERIM SPECIAL TECHNICAL SPECIFICATIONS....................... 6-1 6.1 . Limits on Reactor Coolant Pump Operation. . . . . . . . . . . . . . . . . 6-1  ;

6.2 Use of a Spent Fuel Shipping Cask. . . . . . . . . . . . . . . . . . . . . . . . 6-1 i 6.3 Auxiliary Feedwater Automatic Initiation Setpoint........ 6-1 1 6.4 Operation With Less Than 75% of Incore Detector '

S tri ng s 0pera bl e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 1 iii Amendment No. 32,34,#3,54,55,57,93 73,89,86,

TABLE'2-4 '

INSTRUMENT OPERATING CONDITIONS.FOR ISOLATION FUNCTIONS l

- Test, Maintenance Minimum Minimum Permissible -

and Operable Degree of Bypass Inoperable No. Functional Unit Channels Redundancy Condition Bypass 1 Containment Isolation A Manual 1 None None N/A B Containment High Pressure A 2(a)(e) 1 During Leak (f)

B 2(a)(e) j Test C Pressurizer Low /

Low A 2(a)(e) Reactor Coolant (f) 1 B 2(a)(e) j Pressure Less Than 1700 psia (b) i 2 Steam Line Isolation A Manual 1 None None N/A B Steam Generator Low Pressure A 2/Ste4m 1/ Steam Steam Generator (f)

Gentei Gen Pressure Less B 2/Stegm 1/ Steam Than550 psia (c)

Gentel Gen 3 Ventilation Isolation A Manual 1 None None N/A B Containment High Radiation A 2(d) None If Containment (f)

B 2(d) Ventilation Isola-tion Valves Are Closed a A and B circuits each have 4 channels.

b Auto removal of bypass above 1700 psia, c Auto removal of bypass above 550 psia.

d. A and B circuits are both actuated by any one of the five VIAS initiating channels; RM-050, RM-051, RM-060, RM-061, or RM-062; however, only RM-050 and RM-051 are required for containment ventilation isolation.

e If minimum operable channel conditions are reached, one inoperable channel must be placed in the tripped condition within eight hours from the time of discovery of loss of operability. The remaining-inoperable channel may be by-passed for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from the time of discovery of loss of operability and, if an inoperable channel is not returned to operable status within this time frame, a unit shutdown must be initiated (see Specification (2)).

2-69 Amendment No.49 ,93

TABLE 3-3 ,

MINIMUM FREQUENCIES FOR CHECKS. CALIBRATIONS AND TESTING -

, OF MISCELLANEOUS INSTRUMENTATION AND CONTROLS 2 -

Surveillance .

' Channel Description Function Fre'quency Surveillance Method

- 1. Primary CEA Position a. Check S a. Comparison of output data with secondary CEAPIS.

Indication System i b. Test M b. Test of power dependent insertion limits, devia-tion, and sequence monitoring systems.

c. Calibrate R c. Physically measured CEDM position used to verify

, system accuracy. Calibrate CEA position inter-

locks.

1

2. Secondary CEA Position a. Check S a. Comparison of output data with primary CEAPIS.

Indication System

b. Test M b. Test of power dependent insertion limit, devia-tion, out-of-sequence, and overlap monitoring 3

systems.

w c. Calibrate R c. Calibrate secondary CEA position indication

.L system and CEA interlock alarms.

3. Area, Process, and a. Check D a. Normal readings observed and internal test

, Post-Accident signals used 'to~ verify instrument operation.

Radiation Monitors Except Effluent b. Test M b. Detector exposed to remote operated radiation check source or test signal.

1 g Radiation Monitors (l) k c.. Calibrate R c. RM-063L', M, and H and RM-064 - One time factory calibration is acceptable provided linearity 3

a solid sources are used to check the integrity of 2 the detectors. RM-091A and B - In situ calibra-

P tion by electronic signal substitution i i s accept-g able for all range decades above 10' R/hr. In situ calibration for at least one decade below

$t 10 R/hr shall be by means of calibrated radiation source. All other monitors - Exposure to known I radiation source.

8 1 (1) The surveillance requirements for effluent radiation monitors are described under Specification 3.12.1.

Effluent radiation monitors are: RM-054A, RM-054B, RM-055, RM-055A, RM-057, RM-060, RM-061, and RM-062)

RM-050 and RM-051 are considered effluent radiation monitors when monitoring the ventilatter. 5i.ack.

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 unreviewed safety question. ~

c. Provide imediate written notification to the Division Manager -

Nuclear Production and the Safety Audit and Review Comittee of disagreement between the Plant Review Comittee and the Manager -

Fort Calhoun 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 Comittee 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 Comittee.

5.5.2 Safety Audit and Review Comittee (SARC)

Function 5.5.2.1 The Safety Audit and Review Comittee 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 Comittee shall be composed of:

Chairman: Division Manager - Quality Assurance and Regulatory Affairs Member: Vice President - Nuclear Production, Production Operations, Fuels, and QA&RA Member: Senior Vice President - Electric Operations and Engineering Member: Division Manager - Engineering Member: Division Manager - Nuclear Production Member: OPPD Operations, Engineering, and Technical Support staff 4

Member: Qualified Non-District Affiliated Consultants as Required and as Determined by SARC Chairman 5-5 Amendment No. 9,77,69,76,M,93

5.0. ADMINISTRATIVE CONTROLS 5.5.2.8 e. The Fort Calhoun Station Emergency Plan and implementing procedures j at least once every twelve months.

~

f. The Site Security Plan and implementing procedures at 1 east once every twelve months.

i ,

l g. The Safeguards Contingency Plan and implementing procedures at i least once every twelve months.

I h. The Radiological Effluent. Program including the Radiological Environ-

! mental Monitoring Program and the results thereof, the Offsite Dose l Calculation Manual and implementing procedures, and the Process i Control Program for the solidification of radioactive wastes at least once per 2 years.

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1. Any other area of facility operation considered appropriate by the Safety Audit and Review Comittee or the Vice President - Nuclear 1 Production, Production Operations, Fuels, and Quality Assurance &

! Regulatory Affairs.

l Authority 5.5.2.9 The Safety Audit and Review Committee shall report to and advise the 4

Vice President - Nuclear Production, Production Operations, Fuels, and I Quality Assurance & Regulatory Affairs on those areas of responsibility specified in Sections 5.5.2.7-and 5.5.2.8.

Records 5.5.2.10 Records of Safety Audit and Review Comittee activities shall be prepared, approved and distributed as indicated below:

' a. Minutes of_each Safety Audit and Review Committee meeting shall be prepared, approved and rorwarded to the Vice President - Nuclear l Production, Production Operations, Fuels, and Quality Assurance &

Regulatory Affairs within 14 ' days following each meeting.

b. Reports of-reviews encompassed by Section 5.5.2.7e, f, g, h, and i j

.above shall be prepared, approved and forwarded to the Vice Presi- l dent - Nuclear Production, Production Operations, Fuels,-and Quality Assurance & Regulatory Affairs within 14 days following completion of the review.

c.- Audit reports encompassed by Section 5.5.2.8 above shall be forwarded i to the Vice President - Nuclear Production, Production Operations, l Fuels, and Quality Assurance & Regulatory Affairs and to the respon-sible management positions designated by the Safety Audit and Review i Committee within 30 days after completion of the audit.

5-8 Amendment No. 9,J),78,55,75,5A,56,93 1

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5.0 ADMINISTRATIVE CONTROLS 5.10.2 The following records shall be retained for the duration of the Facility Operating License: --

a. Records of drawing changes reflecting facility design modifications made to systems and equipment described in the Final Safety Anal-ysis 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 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.

4

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 commences and associated installation and maintenance records.

n. Records of analyses required by the Radiological Environmental Monitoring Program.

5.11 Radiation Protection Program Procedures for personnel radiation protection shall be prepared consis-tent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radia-tion exposure.

5-19 9/ddf dAgdd Jp/24/ge, Amendment No. 59, pp, 93

. - ~ _ ,_~. _ - _ , _ . . - . - - _ .

5.0 ADMINISTRATIVE CONTROLS 5.12 Environmental Qualification Deleted 5.13 Secondary Water Chemistry A secondary water chemistry monitoring program to inhibit steam generator tube degradation shall be implemented. This program shall be described in the station chemistry manual and shall include:

1. Identification of a sampling schedule for the critical parameters and control points for these parameters;

2. Identification of the procedures used to measure the values of the critical parameters;:

3. Identification of process sampling points;

4. Procedures for the recording and management of data;

5. Procedures defining corrective actions for off control point chemistry conditions; and

6. A procedure identifying (a) the uthority responsible for the interpre-tation of the data, and (b) the aquence and timing of administrative events required to initiate corrective actions.

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Amendment No. 57,93

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