ML20059J167

Amend 160 to License DPR-40,changes TS to Implement GLs 86-10 & 88-12

ML20059J167
Person / Time
Site:	Fort Calhoun
Issue date:	01/14/1994
From:	Beckner W Office of Nuclear Reactor Regulation
To:
Shared Package
ML20059J156	List: ML20059J153 ML20059J167 ML20059J183
References
GL-86-10, GL-88-12, NUDOCS 9401310384
Download: ML20059J167 (18)

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UNITED STATES j

f.j.kLfl.j NUCLEAR REGULATORY COMMISSION 4

t[pg ;g WASHINGTON, D.C. 20565-0001 s.

a...*

OMAHA PUBLIC POWER DISTRICT DOCKET N0. 50-285

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FORT CALHOUN STATION. UNIT NO._1 i

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AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.160 License No. DPR-40 i

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l 1.

The Nuclear Regulatory Commission (the Commission) has found that:

l A.

The application for amendment by the Omaha Public Power District J

(the licensee) dated September 15, 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 j

Commission; t

i C.

There is reasonable assurance:

(i) 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 j

conducted in compliance with the Commission's regulations; D.

The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the

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public; and E.

The issuance of this amendment is in accordance with-10 CFR Part 51 i

of the Commission's regulations and all applicable requirements have

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been satisfied.

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9401310384 940114 PDR ADOCK 05000285 P

PDR

V

. 2.

Accordingly, Facility Operating License No. DPR-40 is amended 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 Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.160, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

In addition, the license is hereby amended by adding paragraph 3.F of Facility Operating License No. DPR-40 to read as follows:

3.F Fire Protection Prooram Omaha Public Power District shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Updated Safety Analysis Report for the facility and as approved in the SERs dated February 14, and August 23, 1978, November 17, 1980, April 8, and August 12, 1982, July 3, and November 5,1985, July 1,1986, December 20, 1988, November 14, 1990, March 17, 1993, and January

, 1994, subject to the following provision:

Omaha Public Power District may make changes _ to the approved Fire Protection Program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

4.

The license amendment is effective 120-days from its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

&Mw D bl%

William D. Beckner, Director Project Directorate IV-1 Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation Attachments:

1. Page 4a of license *

2. Changes to the Technical Specifications Date of Issuance: January 14, 1994

• Page 4a is attached for convenience, so that the composite license reflects this change.

- 4a -

D.

Soent Fuel Pool Modification The licensee is authorized to modify the spent fuel pool as described in the application dated December 7, 1992, as supplemented by letters dated March 19, April 28, and May 14, 1993. The modification shall consist of removing the existing racks which have a storage capacity of 729 fuel assemblies and inserting new racks which will have a storage capacity of 1083 fuel assemblies.

There shall be 11 free-standing rack modules in a discrete-zone, two-region storage system.

The Appendix A Technical Specifications shall be effective as follows:

o Changes to Specifications 2.8(11) and (12), 4.4.2, and 5.10.3 I

and Figure 2-10, Table 3-5, and the basis for Specification 2.8 shall be effective when the last new rack is installed.

E.

License Term The license amendment is contingent on the limitations of monitoring of the long-term load factor to assure that it does not exceed the assumed value of 0.77, and that a reevaluation of the end of license fluence with ENDF/B-VI cross sections and updated uncertainties will be performed to assure that the value of the RT,in place.

will not exceed py the screening criterion monitoring program being_

F.

Fire Protection Proaram Omaha Public Power District shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Updated Safety Analysis Report for the facility and as approved in the SERs dated February 14, and August 23, 1978, November 17, 1980, April 8, and August 12, 1982, July 3, and November 5, 1985, July 1, 1986, December 20, 1988, November 14, 1990, March 17, 1993, and January 14, 1994, subject to the following provision:

Omaha Public Power District may make changes to the approved Fire Protection Program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

4.

This amended license is effective as of the date of issuance and shall expire at midnight on August 9, 2013.

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FOR THE NUCLEAR REGULATORY COMMISSION Original signed by:

1 A. Giambusso j

A. Giambusso, Deputy Director i

for Reactor Projects Directorate of Licensing

Enclosures:

Appendices A and B - Technical Specifications Date of Issuance: August 9, 1973 Amendment No. 155,150, 160

i ATTACHMENT TO LICENSE AMENDMENT NO.160 FACILITY OPERATING LICENSE NO. DPR-40 DOCKET N0. 50-285 Revise Appendix "A" Technical Specifications as indicated below.

The revised pages ate identified by amendment number and. contain vertical lines indicating' the area of change.

REMOVE PAGES INSERT PAGES 11 11 iv iv l

vi vi 7

7 2-74 2-74 2-89 i

2-90 2-90a 2-91 2-92 2-93 2-94 2-95 3-20 3-20 3-79c 3-79c 3-80 3-81 3-82 3-83 5-la 5-la 5-3 5-3 5-4 5-4 5-5 5-5 5-9a 5-9a 5-10 5-10 5-15 5-15 i

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.I TABLE OF CONTENTS (Continued) i I

L f.il82

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2.12 ControI Room Systems.....

..... 2-59 j

2.13 Nuclear Detector Cooling system................................... 240 2.14 Engineered Safety Features System initiation Instrumentation Settings....................................... 241 2.15 Instrumentation and Control Systems................................ 2-65 2.16 Ri ver level............................................... 2-71 i

2.17 Miscellaneous Radioactive Material Sources...........................2-72 2.18 Shock Suppressors (Snubbers).................................... 2-73 2.19 DELETED l

2.20 Steam Generatcr Coolant Radioactivity.............................. 2-96 2.21 Post. Accident Monitoring Instrumentation

... 2-97 2.22 Toxic Gas M onitors.......................................... 2-99 l

3.0 SURVEILLANCE REQUIREMENTS.

. 3 0a' l

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3.1 Instrumentation and Control

. 3-1 3.2 Equipment and Sampling Tests

......... 3-17 3.3 Reactor Coolant System and Other Components Subject to ASME XI Boiler and Pressure Vessel Code Inspection and Testing Surveillance...

....... 3 21 3.4 Reactor Coolant System Integrity Testing..........

3-36 i

3.5 Containment Test

......... 3 -3 7 3.6 Safety Injection and Containment Cooling Systems Tests

............. 3 -5 4 3.7 Emergency Power System Periodic Tests............................. 3-58 i

3.8 Main steam Isolation Valves 341' 3.9 Auxiliary Feedwater System

. 342 3.10 Reactor Core Parameters..

...................,.......... 343 3.!!

DELETED

.... 3-64 3.12 Radioactive Waste Disposal System

. 3-69 3.13 Radioactive Material Sources Surveillance............................376 3.14 Shock Suppressors (Snubbers).........

3-77

(

3.15 DELETED l

3.16 Residual Heat Removal System Integrity Testing..........

.......... 3 84 i

3.17 Steam Generator Tubes.....

................ 3-86 4.0 DESIGN FEATURES 4.....

4.1 Site

............................. 4-1 i

4.2 Containment Design Features..

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

... 4-1 4.2.2 Penetrations

. 4-1 4.2.3 Containment Structure Cooling Systems

.4-2 ii Amendment No. Si,S6,93,101,122,236,152,160

i TECIINICAL SPECIFICATIONS - TABLES TABLE OF CONTENTS TABLE DESCRIPTION PAGE l-1 RPS LSSS 1-10 1-10a 2-1 ESFS Initiation Instrumentation Setting Limits

.244 244a d

2-2 Instrument Operating Requirements for RPS 2-67 2-67a 2-3 Instrument Operating Requirements for Engineered Safety Features.

........ 2-6 8

.... 2-68a

. 2-68b 2-4 Instrument Operating Conditions for Isolation Functions...

2-69 2-69a 2-5 Instrumentation Operating Requirements for Other Safety Feature Functions

. 2-70 1

2-9 RCS Pressure Isolation Valves 2 2e 2-10 Post-Accident Monitoring. Instrumentation Operating Limits

. 2-98 2 98a 2-98b 2-11 Toxic Gas Monitors Operating Limits 2-100 3-1 Minimum Frequencies for Checks. Calibrations, and Testing of RPS

.33

.3-4

. 3-5

.3-6 3-2 Minimum Frequencies for Checks, Calibrations and Testing of Engineered Safety Features, lastrumentation and Controls.

. 3-7

.3-8

... 3-9 3-10 3-11 3-12 3-12a I

iv Amendment No. 446rn6,160

TECHNICAL SPECIFICATIONS - TABLES i

TABLE OF CONTENTS (ALPHABETICAL ORDER)

TABLE DESCRIPTION PAGE l

1 2-1 ESFS Initiation lastrumentation Setting Limits.......................... 244

......................... 2 -64 a l

2-4 Instrument Operating Conditions for Isolation Functions................... 249 2 49a 22 Instrument Operating Requirements for RPS........................... 2-67

............... 2 47a 2-3 Instrument Operating Requirements for Engineered Safety Features............. 248 248a

...............248b 2-5 Instrumentation Operating Requirements for Other Safety

....... 2-70 Features Functions.

..................................3-16e 3-3a Minimum Frequency for Checks. Calibrations and Functional Testing of Attemate Shutdown Panels (Al-185 and Al-212) and Emergency Auxiliary Feedwater Panel (Al-179) Instrumentation and Control Circuits...,,.

. 3 16d 3-2 Minimum Frequencies for Checks, Calibrations and Testing of Engineered Safety Features. Instrumentation and Controls.................... 3-7

................. 3-8

. 3-9

......... 3-10

....... 3-11 i

........... 3-12 j

3 - 12a 3-3 Minimum Frequencies for Checks, Calibrations. and Testing of Miscellaneous Instmmentation and Controls

................ 3-13 l

........................... 3-14

........................... 3-15 4

...........................3-16

............... 3 16a

............. 3-16b 3 - 16e l

............................ 34 i

i 3-1 Minimum Frequencies for Checks, Calibrations, and Testing of RPS......................................... 3-3

...................................... 3-4

.................................... 3-5

.......................................... 34 i

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vi Amendment No. ?",125,112.160 i

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DEFINITIONS Azimuthal Power Tilt - T; Azimuthal Power Tilt shall be the maximum difference between the power generated in any core quadrant (upper or lower) and the average power of all quadrants in that axial half (upper or lower) of the core divided by the average power of all quadrants in that axial half (upper or lower) of the core.

1 Unrodded Planer Radial Peaking Factor - F.y The Unrodded Planar Radial Peaking Factor is the maximum ratio of the peak to average power density of the individual fuel rods in any of the unrodded horizontal planes, excluding azimuthal tilt, T,. The maximum F,, limit is provided in the Core Operating Limits Report.

i Unrodded Integrated Radial Peakinc Factor - Fa l

The Unrodded Integrated Radial Peaking Factor is the ratio of the peak pin power to the ave:. age pin power in an unrodded core, excluding azimuthal tilt, T,. The maximum Fa limit is provided in the Core Operating Limits Report.

l Process Control Program (PCP)

The document (s) that contains 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 20, 61, 71, State Regulations, burial ground requirements, and other requirements governing the disposal of solid waste.

Dose Eauivalent I-131 That concentration of I-131 (pCi/gm) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually present. In other l

words, l

7 Amendment No. 32,38,67,86,141,152;160

2.0 LIMITING CONDITIONS FOR OPERATION 2.18 Shock Suppressors (Snubbers)

Basis Snubbers are designed to prevent unrestrained pipe motion under dynamic loads as might occur during an earthquake or severe transient, while allowing normal thermal motion during startup or shutdown. The consequence of an inoperable snubber is an increase in the probability of structural damage to piping as a result of a seismic, or other event, initiating dynamic loads. It is therefore required that all snubbers required to protect the primary coolant system or any other safety system or component be operable during reactor operation.

Because the snubber protection is required only during low probability events, an inoperable period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed for repairs or replacements and an inoperable period of two hours is allowed for surveillance.

2-74 Amendment No. 27,48,59,76,105,160 (next page is 2-96) l

TABLE 3-5 MINIMUM FREOUENCIES FOR EOUIPMENT TESTS FSAR Sation Test Frequency Reference 1.

Control Element Drop times of all full-length CEA's Each refueling operation 7.5.3 Assemblies 2.

Control Element Partial movement of all CEA's Every two weeks 7

Assemblies (Minimum of 6 in) 3.

Pressurizer Safety Set Point Once each refueling outage 7

Valves 4.

Main Steam Safety Set Point Each refueling outage 4

Valves 5.

Refueling System Functioning Prior to refueling outage 9.5.6 Interlocks 6.

Raw Water System Functioning Each refueling outage 9.8 Valve Actuation 7.

DELETED l

8.

Reactor Coolant Evaluate Daily

• 4 System Leakage 9.

Diesel Fuel Supply Fuel Inventory Daily 8.4 10a.

Charcoal and IIEPA

1. In-Place Testinc**

9.10 Filters for Control Charcoal adsorbers and HEPA Each refueling shutdown not to exceed 18 Room filter banks shall be leak months or after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system tested and show 199.95 %

operation or after each complete or Freon (R-11 or R-112) and partial replacement of the charcoal cold DOP particulates adsorber/IIEPA filter banks, or after removal, respectively.

any major stnictural maintenance on the system housing and following significant painting, fire or chem-ical releases in a ventilation zone communicating with the system.

• Whenever the system is at or above operating temperature and pressure.

• • Tests shall be performed in accordance with applicable section(s) of ANSI N510-1980.

3-20 Amendment No. !5,21,128,160

TABLE 3-14 (Continucxi)

Note 5: If the number of unacceptable snubbers is equal to or greater than the number in Column C, the next inspection interval shall be two-thirds of the previous interval. However, if the number of unacceptable snubbers is less than the number in Column C but greater than the number in Column B, the next interval shall be reduced by a factor that is one-third of the ratio of.the difference-between the number of unacceptable snubbers found during the previous interval j

and the number in Column B to the difference in the numbers in Columns B and C.

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3-79c Amendment No.' M5 160 T

(next page is 3-84) l l

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i 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization (Continued) j b.

An Operator or Technician qualified in Radiation Protection Procedures shall be onsite when fuel is in the reactor.

c.

All core alterations shall be directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator limited to fuel handling who has no other concurrent responsibilities during the operation.

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d.

Fire protection program responsibilities are assigned to those positions and/or groups designated by asterisks in USAR 12.1-1 through 12.1-4 according to the procedures specified in Section 5.8 of the Technical Specifications, e.

Administrative procedures shall be developed and implemented to limit the working hours of plant staff who perform safety-related functions. Administrative procedures shall reflect the personnel whose working hours will be affected.

Shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work a normal 8-hour day, 40-hour week while the plant is operating.

However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance, or major plant modifications, on a temporary basis, the guidelines identified in the administrative procedures shall be followed.

Deviations from the guidelines shall be authorized by the Department Manager, Plant Manager, or their designated alternates, or higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation. Routine deviation from the administrative guidelines shall not be at'udrized.

f.

The Supervisor - Operations, the Shift Supervisors and Licensed Senior Operators l

shall hold a senior reactor operator license. The Licensed Operators shall hold a reactor operator license.

5-la Amendment No. 38,54,85,115,160

5.0 ADMINISTRATIVE CONTROLS 5.4 Training 5.4.1 A retraining and replacement training program for the plant staff shall be maintained under the direction of the Manager - Training and shall meet or exceed the requirements of Section 5.5 of ANSI N18.1-1971 and 10 CFR Part 55.

5.5 Review and Audit 5.5.1 PJant Review Committee (PRC)

Function j

5.5.1.1 The Plant Review Committee shall function to advise the Manager - Fort Calhoun Station on all matters related to nuclear safety.

Comoosition 5.5.1.2 The official Plant Review Committee shall be composed of the:

Chairman: Manager - Fort Calhoun Station Member: Supervisor - Operations Member: Manager - Training Member: Supervisor - Maintenance Member: Supervisor - System Engineering Member: Reactor Engineer Member: Supervisor - Radiation Protection Member: Supervisor - Chemistry Member: Assistant Plant Manager Alternates 5.5.1.3 Altemate members shall be appointed in writing by the Plant Review Committee Chairman to serve on a temporary basis; however, no more than two alternates shall participate in Plant Review Committee activities at any one time.

Meeting Freauency 5.5.1.4 The Plant Review Committee shall meet at least once per calendar month and as convened by the Plant Review Committee Chairman.

Ouorum 5.5.1.5 A quorum of the Plant Review Committee shall consist of the Chairman and four members including alternates.

5-3 Amendment No. 9,19,38,84,115,134,160

' 5.0 ADMINISTRATWE CONTROLS Responsibilities -

5.5.1.6 The Plant Review Committee shall be responsible for:

a.

Review of (1) Administrative Controls Standing Orders and changes thereto, (2) procedures required by Specification 5.8 and requiring a 10 CFR 50.59 safety evaluation, and (3) proposed changes to procedures required by Specification 5.8 and requiring a 10 CFR 50.59 safety -

evaluation; 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 to the Core Operating Limits Report.

Review of all proposed changes or modifications to plant systems or e.

equipment that affect nuclear safety, f.

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 _ Operations and to the Chairperson of the Safety Audit and Review Committee.

g.

Review of facility operations to detect potential safety hazards.

h.

Performance of special reviews and investigations and reports thereon as requested by the Chairperson of the Safety Audit and Review Committee.

i.

Review of the Site Security Plan and implementing procedures'and shall submit recommended changes to the Chairperson of the Safety Audit and Review Committee.

I j.

Review of the Site Emergency Plan and implementing procedures and shall submit recommended changes to the Chairperson of the Safety Audit.

and Review Committee, k.

Review of the Fire Protection Program Plan and shall submit changes to the Chairman of the Safety Audit and Review Committee.

1.

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 (e) above.

5-4 Amendment No. 9,19,S',99,115,141, 449,160

5.0 ADMINISTRATIVE CONTROLS l

5.5.1.7 b.

Render determinations in writing with regard to whether or not each item considered under 5.5.1.6(b) through (f) above constitutes an unreviewed safety question.

c.

Provide immediate written notification to the Division Manager - Nuclear Operations and the Safety Audit and Review Committee of disagreement between the Plant Review Committee 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 Committee shall maintain written minutes of each meeting and copies shall be provided to the Division Manager - Nuclear Operations and Chairperson of the Safety Audit and Review Committee.

i 5.5.2 Safety Audit and Review Committee (SARC)

Function 5.5.2.1 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 i.

fire protection l

Comoosition 5.5.2.2 The Safety Audit and Review Committee shall be composed of:

Chairperson: Division Manager - Nuclear Services Member:

Senior Vice President Member:

Vice President Member:

Division Manager - Nuclear Operations Member:

Division Manager - Production Engineering Member:

Manager - Fort Calhoun Station Member:

Other qualified OPPD personnel or consultants as required and as determined by the SARC Chairperson 5-5 Amendment No. 86,93,99,101, 115,119,132,141,149,157, 160

5.0 ADMINISTRATIVE CONTROLS I

5.8.2.1 Each procedure, or change thereto, shall be reviewed by a Qualified Reviewer (QR) who is knowledgeable in the functional area affected but is not the individual preparer. The QR may be from the same line-organization as the l

preparer. The QR shall render a determination in writing of whether or not cross-disciplinary review of a procedure, or change thereto is necessary. If necessary, such review shall be performed by appropriate personnel.

5.8.2.2 Each procedure, or change thereto, shall be reviewed by the Department Head designated by Administrative Controls Standing Orders as the responsible Department Head for that procedure, and the review shall include a determination of whether or not a 10 CFR 50.59 safety evaluation is required. If a 10 CFR 50.59 safety evaluation is not required, the procedure, or change thereto, shall be approved by the responsible Department Head or the Manager-Fort Calhoun Station, prior to implementation. Administrative Controls Standing Orders, the j

Site Security Plan and Implementing Procedures, the Emergency Plan and Implementing Procedures, and the Fire Protection Program Plan shall be reviewed in accordance with Specification 5.5.1.6 and approved by the Manager-Fort j

Calhoun Station.

5.8.2.3 If the responsible Department Head determines that a procedure, or change thereto, requires a 10 CFR 50.59 safety evaluation, the responsible Department Head shall render a determination in writing of whether or not the procedure, or change thereto, involves an Unreviewed Safety Question (USQ) and shall forward the procedure, or change thereto with the associated safety evaluation to the PRC for review in accordance with Specification 5.5.1.6.a. If a USO is involved, NRC approval is required prior to implementation of the procedure, or change.

5.8.2.4 Qualified Reviewers shall meet or exceed the respective qualifications for either Supervisors Requiring an AEC License, Professional-Technical Personnel, or Technical Support Personnel, as specified in ANSI N18.1 - 1971. Personnel recommended to be QRs shall be reviewed by the PRC and approved and designated as such by the PRC Chairman. The responsible Department Head shall ensure that a sufficient complement of QRs for their functional area is maintained in accordance with Administrative Controls Standing Orders.

5.8.2.5 Each procedure of Specification 5.8.1 shall be reviewed periodically as set forth in Administrative Controls Standing Orders.

5.8.2.6 Records documenting the activities performed under Specifications 5.8.2.1 through 5.8.2.4 shall be maintained in accordance with Specification 5.10.

5-9a Amendment No. 449-160

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.0 ADMINISTRATIVE CONTROLS 5

5.8.3 Temporary changes to procedures of 5.8.1 above may be made provided:

4 a.

The intent of the original procedure is not altered.

b.

The change is approved by two members of the plant supervisory staff, at least one of whom holds a Senior Reactor Operator's License.

The change is documented, reviewed by a Qualified Reviewer and c.

I approved by either the Manager - Fort Calhoun Station or the Department Head designated by Administrative Controls Standing Orders as the responsible Department Head for that procedure within 14 days of implementation.

j 5.8.4 Written procedures approved per 5.8.2 above shall be implemented which govern the selection of fuel assemblies to be placed in Region 2 of the spent fuel racks (Technical Specification 2.8). These procedures shall require an independent verification of initial enrichment requirements and fuel burnup calculations for a fuel bundle to assure the " acceptance" criteria for placement in Region 2 are met.

This independent verification shall be performed by individuals or groups other than those who performed the initial acceptance criteria assessment, but who may be from the same organization.

5.8.5 Written procedures shall be established and maintained for implementation of the Fire Protection Program.

5.9 Reporting Requirements In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following identified reports shall be submitted to the Director of the appropriate Regional Office of Inspection and Enforcement unless otherwi

noted, i

5.9.1 Routine Reports a.

Startup Report. A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufacture by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant. The report shall address each of the tests identified in the USAR and shall'in general include a description of the measured values of the operating conditions or chameteristics obtained during the test program and a comparison of these values with design predictions and specifications.

Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.

5-10 Amendment No. 9,19,35,75,149,160

5.0 ADMINISTRATIVE CONTROLS 5.9.3 Special Reports Special reports shall be submitted to the Regional Administrator of the appropriate NRC Regional Office within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification where appropriate:

a.

In-service inspection report, reference 3.3.

b.

Tendon surveillance, reference 3.5.

c.

Containment structural tests, reference 3.5.

d.

Special maintenance reports.

e.

Containment leak rate tests, reference 3.5.

f.

DELETED g.

Materials radiation surveillance specimens reports, reference 3.3.

h.

DELETED l

i.

Post-accident monitoring instrumentation, reference 2.21 j.

Electrical systems, reference 2.7(2).

5.9.4 Uniaue Reoortine Requirements a.

Annual Radioactive Effluent Release Report The Annual Radioactive Effluent Release Report covering the operation of the unit during the previous 12 months of operation shall be submitted each year.

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 50.

b.

Annual Radiological Environmental Operating Report 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, 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)Section IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR 50.

c.

Fire Protection Program Deficiency Report Deficiencies in the Fire Protection Program described in the Updated Safety Analysis Report which meet the reportability criteria of 10 CFR 50.73 shall be reported pursuant to Section 5.9.2 of the Technical Specifications.

5-15 Amendment No. 9,24,38,46,86,110,

!!3,133,147,152, 160