ML20214F770

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Proposed Tech Specs Reflecting Revised Reporting Requirements of 10CFR50.72 & 73,per Generic Ltr 83-43
ML20214F770
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 11/17/1986
From:
ARKANSAS POWER & LIGHT CO.
To:
Shared Package
ML20214F751 List:
References
(2CAN118605), GL-83-43, NUDOCS 8611250427
Download: ML20214F770 (40)
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Text

ENCLOSURE PROPOSED TECHNICAL SPECIFICATION CHANGES IN THE MATTER OF AMENDING LICENSE NO. NPF-6 ARKANSAS POWER & LIGHT COMPANY ARKANSAS NUCLEAR ONE, UNIT 2 DOCKET NO. 50-368 NOVEMBER 17, 1986 8611250427 861117 8 DR ADOCK 0500 (2CAN118605)

r PROPOSED CHANGES It is proposed that the ANO-2 Technical Specifications be revised to incorporate the conforming changes as indicated on the revised copies attached to this transmittal. These changes are submitted in accordance with Generic Letter 83-43.

DISCUSSION The revisions delineated by Generic Letter 83-43 that licensees should make in the " Administrative Control" and " Definitions" sections of the technical specifications are described as follows:

(a) Specification 1.7 (page 1-2) requires deletion of the term, " REPORTABLE OCCURRENCE" and adding the term, " REPORTABLE EVENT" with a new definition referencing 10CFRS50.73.

(b) Speci fication 6. 5.1. 6. f (page 6-7), Speci fication 6.5.2. 7.g (page 6-10), Specifications 6.6 and 6.6.1 (page 6-12), Specification 6.9 (page 6-14), and Specification 6.10.1.c (page 6-22) are revised as suggested by the Commission in the model technical specifications attached to Generic Letter 83-43.

(c) Specification 6.9.1.7 (page 6-16), Specification 6.9.1.8 (page 6-16, 6-17), and Specification 6.9.1.9.a,.b,.c,.d (page 6-17, 6-18) are deleted as suggested by the Commission in the model technical specifications attached to Generic Letter 83-43. Specification 6.9.1.9.e is preserved but redesignated as Specification 6.9.2.n (see (b)6. below).

(d) Specification 3.11.2.1 Action b (page 3/4 11-7) and Specification 3.11.4.b (page 3/4 11-16) are revised to delete reference to Specifications 6.9.1.7 and 6.9.1.9.b, respectively.

Other conforming changes to the technical specifications required to reflect the revised reporting requirements of 10CFR50.72 and 10CFR50.73 are described as follows, pursuant to Generic Letter 83-43:

(a) Specification 6.7.1 (page 6-13) is revised to conform with the notification requirements of 10CFRS50.72 and reporting requirements of 10CFRS50.73 as required by 10CFRS50.36(c)(6) for Safety Limit Violations.

(b) Specification 6.9.2 (page 6-19) is revised to denote Special Reports for certain events, in lieu of a Licensee Event Report, which were previously Reportable Occurrences pursuant to Specifications 6.9.1.7, 6.9.1.8, or 6.9.1.9 being deleted, conforming to Generic Letter 83-43 for retention as Special Reports:

1) Itea h. is revised to include special reportability for radioactive effluent levels exceeding 40CFR190, in accordance with 10CFRS20.405(c).

2CAN118605 1 l

2) Item j. is added for Steam Generator Tubing Surveillance --

Category C-3 Results.

3) Item k. is added for Maintenance of Spent Fuel Pool Structural Integrity.
4) Item 1. is added for Radiological Environmental Monitoring Sample Analysis.
5) Item m. is added to retain a Radiological Environmental Technical Specification regarding unplanned offsite releases. For clarification, reportability timeframes were added to this specification.

(c) Specification 3.3.3.9.d (page 3/4 3-45), Specification 3.3.3.10.c (page 3/4 3-54), Specification 3.11.1.2 Action b (page 3/4 11-4),

Specification 3.11.1.3.b (page 3/4 11-5), Specification 3.11.2.2.a (page 3/4 11-10), Specification 3.11.2.3.b (page 3/4 11-11),

Specification 3.11.2.4.b (page 3/4 11-11), Specification 3.11.2.5.b (page 3/4 11-13), Specification 3.12.1.d (page 3/4 12-2), Specification 3.12.2.c (page 3/4 12-8), and Specification 3.12.3.b (page 3/4 12-9) are revised to delete reference to Specification 6.9.1.7 which has been deleted pursuant to Generic Letter 83-43.

(d) Specification 4.4.5.5.c (page 3/4 4-10) is revised to reference Table 4.4-2 for the appropriate notification action as a result of Category C-3 steam generator tube inspection results, and to reference Specification 6.9.2 instead of 6.9.1 to specify that a Special Report is required, as denoted in Table 4.4-2. A revision is also made to clarify that notification of the Commission of the applicable inspection results will be conducted prior to resumption of plant operation.

(e) Table 4.4-2 (page 3/4 4-12) is revised to delete references to Specification 6.9.1 since the referenced specification, as revised pursuant to Generic Letter 83-43, would inappropriately cite routine reporting requirements. The action required is changed to a Special Report of inspection results.

(f) Specification 3.7.12.a (page 3/4 7-38) is revised to remove the reference to Specification 6.9.1.8 and the term " Reportable Occurrence" which have been deleted in accordance with Generic Letter 83-43.

Wording is added to specify that a written Special Report is required when the structural integrity of the spent fuel pool is in a nonconforming condition. A typographical error on the same page has also been corrected in Specification 4.7.12.2.a at line 4. The word

" internal" is corrected to read " interval".

(g) Specification 4.8.1.1.3 (page 3/4 8-4a) is revised to specify annual reportability for diesel generator data under Specification 6.9.1 by adding reference to Specification 6.9.1.5.d.

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r (h) Specification 3.12.1.b (page 3/4 12-1) is revised at lines 6, 13, 19, and 22 to add the term "Special Report" to clarify that the required report is intended to be a Special Report.

(i) Bases 3/4 4.5 (page B 3/4 4-3) is revised for consistency with proposed Specification 4.4.5.5(c) to clarify the notification requirement prior to resumption of plant operation and to denote the need for a Special Report of steam generator tube inspection C-3 results [see change (d) above].

(j) Index (page I and page XVII) is revised for consistency with the above proposed changes pursuant to Generic Letter No. 83-43.

(k) Specification 6.9.1.5.d (page 6-15) is added to clarify that diesel generator surveillance test data shall be the subject of an annual report to the NRC.

DETERMINATION OF SIGNIFICANT HAZARDS Arkansas Power & Light Company has performed an analysis of the proposed change in accordance with 10CFR90.91(a)(1) regarding no significant hazards consideration, using the standards in 10CFR90.92(c).

A discussion of those standards as they relate to this amendment request follows:

Criterion 1 - Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated.

The proposed changes are conforming to revised regulations and are made at the request of the Commission. Since these changes affect reportability requirements only, they do not affect the intended operation of the plant or its design bases. Therefore, the proposed changes would not create the probability of a new or different kind of accident from any previously evaluated.

Criterion 2 - Does Not Create the Possibility of a New or Different Kind of Accident from any Previously Evaluated.

These conforming changes are administrative in nature affecting reportability requirements and do not adversely affect the modes of plant operation or the operability of safety-related equipment, or change the design or configuration of the plant. Therefore, the proposed changes would not create the possibility of a new or different kind of accident from any previously analyzed.

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r Criterion 3 - Does Not Involve a Significant Reduction in a Margin of Safety.

The proposed changes would not involve a significant reduction in the margin of safety since these changes conform to the revised Commission regulations by clarifying the wording of the text but have no effect on any plant safety parameters or accident mitigation capabilities.

The Commission has provided guidance concerning the application of these standards by providing examples of amendments that are considered not likely to involve significant hazards considerations. The proposed amendment is most closely encompassed by example (vii): "A change to conform a license to changes in the regulations where the license change results in very minor changes to facility operations clearly in keeping with the regulations."

Therefore, based on the reasoning presented above and the previous discussion of the amendment request, AP&L has determined that the requested changes do not involve a significant hazards consideration.

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PROPOSED TECHNICAL SPECIFICATION CHANGES 2CAN118605

m INDEX DEFINITIONS SECTION 1.0 DEFINITIONS Defined Terms............................................... 1-1 Thermal Power............................................... 1-1 Rated Thermal Power......................................... 1-1 Operational Mode - Mode..................................... 1-1 Action...................................................... 1-1 Operable - Operability...................................... 1-1 Reportable Event............................................ 1-2 l Containment Integrity....................................... 1-2 Channel Calibration......................................... 1-2 Channel Check............................................... 1-3 Channel Functional Test..................................... 1-3 Core Alteration............................................. 1-3 Shutdown Margin............................................. 1-3 Identified Leakage.......................................... 1-4 Unidentified Leakage........................................ 1-4 P re s s u re Bo u nda ry Lea kage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 Azimuthal Power Ti1t........................................ 1-4 Dose Equivalent I-131....................................... 1-4 E-Average Disintegration Energy............................. 1-5 Staggered Test Basis........................................ 1-5 Frequency Notation.......................................... 1-5 Axial Shape Index........................................... 1-5 Reactor Trip System Response Time........................... 1-6 Engineered Safety Feature Response Time..................... 1-6 Physics Tests............................................... 1-6 Software.................................................... 1-6 Planar Radial Peaking Factor................................ 1-6 ARKANSAS - UNIT 2 I AMENDMENT NO. 24

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1 INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.6 REPORTABLE EVENT ACTI0N..................................... 6-12 l 6.7 SAFETY LIMIT VIOLATION........................ ............. 6-13 6.8 PROCEDURES.................................................. 6-13 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REP 0RTS........................................ 6-14 6.9.2 SPECIAL REP 0RTS........................................ 6-19 6.9.3 SEMIANNUAL RADIOLOGICAL EFFLUENT RELEASE REPORT........ 6-19a l 6.9.4 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT..... 6-21 6.10 RECORD RETENTION................................... ....... 6-22 6.11 RADIATION PROTECTION PR0 GRAM............................... 6-23 6.12 ENVIRONMENTAL QUALIFICATION................................ 6-23 6.13 HIGH RADIATION AREA........................................ 6-24 6.14 0FFSITE COSE CALCULATION MANUAL (0DCM)..................... 6-25 ARKANSAS - UNIT 2 XVII AMENDMENT NO. 21, 60

DEFINITIONS REPORTABLE EVENT

1. 7 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10CFR Part 50.

CONTAINMENT INTEGRITY 1.8 CONTAINMENT INTEGRITY shall exist when:

1.8.1 All penetrations required to be closed during accident condition are either:

a. Capable of being closed by an OPERABLE containment automatic isolation valve system, or
b. Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3.1.

1.8.2 All equipment hatches are closed and sealed, 1.8.3 Each airlock is OPERABLE pursuant to Specification 3.6.1.3, 1.8.4 The containment leakage rates are within the limits of Specification 3.6.1.2, and 1.8.5 The sealing mechanism associated with each penetration (e.g., welds, bellows or 0-rings) is OPERABLE.

CHANNEL CALIBRATION

1. 9 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIOWAL TEST.

i The CHANNEL CALIBRATION may be performed by any series of sequential,

overlapping or total channel steps such that the entire channel is l calibrated.

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l ARKANSAS - UNIT 2 1-2 Amendment No.

r AQMINISTRATIVE CONTgDLS j

f. Review of all REPORTABLE EVENTS. l
g. Review of facility operations to detect potential nuclear safety hazards.
h. Performance of special reviews, investigations or analyses, and 1 reports thereon as requested by the Director, Site Nuclear  !

Operations or the Safety Review Committee.

i. Review of the Plant Security Plan and implementing procedures and submittal of recommended changes to the Director, Site Nuclear Operations and the Safety Review Committee.
j. Review of the Emergency Plan and implementing procedures and submittal of recommended changes to the Director, Site Nuclear Operations and Safety Review Committee.
k. Review of changes to the Offsite Dose Calculation Manual and Process Control Program.

AUTHORITY 6.5.1.7 The Plant Safety Committee shall:

a. Recommend in writing their approval or disapproval of items considered under 6.5.1.6(a) through (d) above.
b. Render determinations in writing with regard to whether or not each item considered under 6.5.1.6(a) through (e) above constitutes an unreviewed safety question.
c. Provide written notification within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Vice President, Nuclear Operations and the Safety Review Committee of disagreement between the PSC and the Director, Site Nuclear Operations; however, the Director, Site Nuclear Operations shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.

RECORDS 6.5.1.8 The Plant Safety Committee shall maintain written minutes of each PSC meeting that, at a minimum, document the results of all PSC activities performed under the responsibility and authority provisions of these technical specifications. Copies shall be provided to the Director, Site Nuclear Operations and Chairman of the Safety Review Committee.

ARKANSAS-UNIl 2 6-7 Amendment No. 5, 7, 25, 52, 60, 73

ADMINISTRATIVE CONTROLS REVIEW e

6.5.2.7 The SRC shall review:

a. The safety evaluations for 1) changes to procedures, equipment or systems and 2) tests or experiments completed under the provision of Section 50.59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question.
b. Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
c. Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
d. Proposed changes to Technical Specifications or this Operating License.
e. Violations of 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 unit equipment that affect nuclear safety.
g. All REPORTABLE EVENTS pursuant to Specification 6.6.1.
h. All recognized indications of an unanticipated deficiency in some aspect of design or operation of structures, systems, or components that could affect nuclear safey.

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i. Reports and meetings minutes of the Plant Safety Committee.

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) j. Proposed changes to tha ODCM and PCP.

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i ARKANSAS - UNIT 2 6-10 Amendment No. 60

ADMINISTRATIVE CONTROLS RECORDS 6.5.2.10 Records of SRC activities shall be prepared, approved and distributed as indicated below:

a. Minutes of each SRC meeting shall be prepared, approved and forwarded to the Vice President, Nuclear Operations within 14 days following each meeting.
b. Reports of reviews encompassed by Section 6.5.2.7 above, shall be prepared, approved and forwarded to the Vice President, Nuclear Operations within 14 days following completion of the review.
c. Audit reports encompassed by Section 6.5.2.8 above, shall be forwarded to the Vice President, Nuclear Operations and to the management positions responsible for the areas audited within 30 days after completion of the audit.

6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS: I

a. A report shall be submitted to the Commission pursuant to the requirements of Section 50.73 to 10CFR Part 50, and
b. Each REPORTABLE EVENT shall be reviewed by the PSC and, the results of this review shall be submitted to the SRC and the Vice President, Nuclear Operations by the Director, Site Nuclear Operations.

ARKANSAS - UNIT 2 6-12 Amendment No. II, 17, 25, 52, 73

ADMINISTRATIVE CONTROLS 6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a. The unit shall be placed in at least HOT STANDBY within one hour.
b. The Vice President, Nuclear Operations and the SRC shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
c. The Nuclear Regulatory Commission shall be notified pursuant to 10CFR50.72 and a report submitted pursuant to the requirements of 10CFR50.36 and Specification 6.6.

6.8 PROCEDURES 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:

a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February 1978.
b. Refueling operations.
c. Surveillance and test activities of safety related equipment.
d. Security Plan implementation.
e. Emergency Plan implementation.
f. Fire Protection Program implementatien.
g. Modification of Core Protection Calculator (CPC) Addressable Constants. These procedures should include provisions to assure

' that sufficient margin is maintained in CPC Type I addressable constants to avoid excessive operator interaction with the CPCs during reactor operation.

NOTE: Modifications to the CPC software (including changes of algorithms and fuel cycle specific data) shall be performed in accordance with the most recent version of "CPC Protection Algorighm Software Change Procedure," CEN-39(A)-P that has been determined to be applicable to the facility. Additions

! or deletions to CPC addressable constants or changes to j addressable constant software limit values shall not be

implemented without prior NRC approval.

I h. New and spent fuel storage.

i. ODCM and PCP implementation.
j. Postaccident sampling (includes sampling of reactor coolant, radioactive iodines and particulates in plant gaseous effluent, and the containment atmosphere).

6.8.2 Each procedure of 6.8.1 above, the changes thereto, shall be reviewed by the PSC and approved by the Director, Site Nuclear Operations or the responsible General Manager prior to implementation and reviewed periodically as set forth in administrative procedures.

i ARKANSAS - UNIT 2 6-13 Amendment No. 24, 28, 47, 52, 60, 63, 73, 77

ADMINISTRATIVE CONTROLS 6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided:

a. The intent of the original procedure is not altered.
b. The change is approved by two members of the plant managemer'.

staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.

c. The change is documented, reviewed by the PSC and approved by the Director, Site Nuclear Operations or the responsible General Manager within 14 days of implementation.

6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Administrator of the Regional Office unless otherwise noted.

STARTUP REPORT 6.9.1.1 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 manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.

6.9.1.2 The startup report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics 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 th1s report.

6.9.1.3 Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial power operation),

supplementary reports shall be submitted at least every three months until all three events have been completed.

ARKANSAS - UNIT 2 6-14 Amendment No. 5, 52, 73

3 ADMINISTRATIVE CONTROLS ANNUAL REPORTS 1 6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to March 1 of the year following initial criticality.

6.9.1.5 Reports required on an annual basis shall include:

a. A tabulation on an annual basis for the numte r of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem / year and their associated man rem exposure according to work and job functions,2/ e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. 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.
b. The complete results of steam generator tube inservice inspections performed during the report period (reference Specification 4.4.5.5.b).
c. Documentation of all challenges to the pressurizer safety valves.
d. A diesel generator data report which provides the number of valid tests and the number of valid failures for each diesel generator.

1/ A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station.

2/ This tabulation supplements the requirements of S20.407 of 10 CFR Part 20.

ARKANSAS - UNIT 2 6-15 Amendment No. 5, 4I

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ADMINISTRATIVE CONTROLS MONTHLY OPERATING REPORT 6.9.1.6 Routine reports of operating statist.ics and shutdo n e$perience shall be submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Office no later than the 15th of each month following the calendar month covered by the report. >-

6.9.1.7 Deleted 6.9.1.8 Deleted t

ARKANSAS - UNIT 2 6-16 Amendment No. 52

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ADMINISTRATIVE CONTROLS (text deleted)

Page left blank intentionally 6.9.1.9 Deleted ARKANSAS - UNIT 2 6-17 Amendment No. $2

ADMINISTRATIVE CONTROLS I

(text deleted)

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ARKANSAS - UNIT 2 6-18 Amendment No. 52, 60

ADMINISTRATIVE CONTROLS SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Administrator of the 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:

a. ECCS Actuation, Specifications 3.5.2 and 3.5.3.
b. Inoperable Seismic Monitoring Instrumentation, Specifica-tion 3.3.3.3.
c. Inoperable Meteorological Monitoring Instrumentation, Specifica-tion 3.3.3.4.
d. Seismic event analysis, Specification 4.3.3.3.2.
e. Inoperable Fire Detection Instrumentation, Specification 3.3.3.8.
f. Inoperable Fire Suppression Systems, Specifications 3.7.10.1 and 3.7.10.2.
g. Primary coolant specific activity, Specification 3.4.8.
h. Radioactive Effluents, Specifications 3.11.1.1, 3.11.1.2, 3.11.1.3, 3.11.2.2, 3.11.2.3, 3.11.2.4, 3.11.2.5, and 3.11.3.

This report shall include the following:

1) Description of occurrence.
2) Identify the cause(s) for exceeding the limit (s)
3) Explain corrective action (s) taken to mitigate occurrence.

l 4) Define action (s) taken to prevent recurrence.

5) Summary of consequence (s) of occurrence.
6) Describe levels exceeding 40CFR190 in accordance with 10CFR20.405(c).
i. Inoperable Containment Radiation Monitors, Specification 3.3.3.1.

l j. Steam Generator Tubing Surveillance -- Category C-3 Results, l Specification 4.4.5.5.

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k. Maintenance of Spent Fuel Pool Structural Integrity, Specification 3.7.12.

ARKANSAS - UNIT 2 6-19 Amendment No. 66, 63

ADMINISTRATIVE CONTROLS

1. Radiological Environmental Monitoring Sample Analysis, Specification 3.12.1.
m. Unplanned Offsite Release during one hour period of 1) more than 1 curie of radioactive material in liquid effluents, 2) more than 150 ciries of noble gas in gaseous effluents, or 3) more than 0.05 curies of radioiodine in gaseous effluents. This report shall be submitted within 30 days of the occurrence of the event and shall include the following information:
1. Description of the occurrence.
2. Identify the cause(s) of exceeding the limit (s).
3. Explain corrective action (s) taken to mitigate occurrence.
4. Define action (s) taken to prevent recurrence.
5. Summary of the consequence (s) of occurrence.

SEMI-ANNUAL RADIOLOGICAL EFFLUENT RELEASE REPORT

  • 6.9.3 Routine radioactive effluent release reports covering the operating of the unit during the previous 6 months of operations shall be submitted within 60 days after January 1 and July 1 of each year.
  • A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste system, the submittal shall specify the releases of radioactive material from each unit.

ARKANSAS - UNIT 2 6-19a Amendment No.

ADMINISTRATIVE CONTROLS 6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

6.10.1 The following records shall be retained for at least five years:

a. Records and logs of unit operation covering time interval at i each power level.

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b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.
c. ALL REPORTABLE EVENTS.
d. Records of surveillance activities, inspections and calibrations required by these Technical Specifications.
e. Records of changes made to the procedures required by Specifica-tion 6.8.1.
f. Records of radioactive shipments.
g. Records of sealed source and fission detector leak tests and results.
h. Records of annual physical inventory of all sealed source material of record.

6.10.2 The following records shall be retained for the duration of the Facility Operting License:

a. Records and drawing changes reflecting unit design modifications made to systems and equipment described in the Final Safety Analysis Report.
b. Records of new and irradiated fuel inventory, fuel transfer and assembly burnup histories.
c. Records of radiation exposure for all individuals entering radiation control areas.
d. Records of gaseous and liquid radioactive material released to the environs.
e. Records of transient of operational cycles for those unit com-ponents identified in Table 5.7.1.

ARKANSAS - UNIT 2 6-22 Amendment No. 60

INSTRUMENTATION RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.9 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their alarm / trip setpoints set to ensure thet the limits of Specification 3.11.2.1 are not exceeded. j 1

APPLICABILITY: During releases via this pathway.

ACTION:

a. With the following gaseous effluent monitoring in-strumentation channels alarm / trip setpoint less con-servative than required by the above specification, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel.
1. Waste Gas Holdup System Noble Gas Activity Monitor. (during periods of gaseous releases.)
2. Containment Purge and Ventilation System Noble Gas Activity Monitor. (during periods of containment building PURGE.)
b. With less than the minimum number of monitoring instrumentation channels OPERABLE, take the action shown in Table 3.3-12.
c. Return the instruments to OPERABLE status within 30 days or, in lieu of any other report, explain in the next
Semiannual Radioactive Effluent Release Report why the j inoperability was not corrected.
d. The provisions of Specifications 3.0.3, 3.0.4, and 4.0.4 are not applicable. l SURVEILLANCE REOUIREMENTS 4.3.3.9 Each radioactive gaseous effluent monitoring instrumenta-tion channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST at the frequencies shown in Table 4.3-12.

ARKANSAS - UNIT 2 3/4 3-45 Amendment No. 80

INSTRUMENTATION i

RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.10 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their alarm / trip set-points set to ensure that the limits of Specification 3.11.1.1 are not exceeded.

APPLICABILITY: During releases via this pathway.

ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the l above specification, immediately suspended the release of radioactive liquid effluents monitored by the affected channel, until the set point is changed to an acceptable conservative value.
b. With less than the minimum number of monitoring instrumentation channels OPERABLE, take the action shown in Table 3.3-13.
c. Return the instruments to OPERABLE status within 30 days or, in lieu of any other report, explain in the next Semiannual Radioactive Effluent Release Report why the inoperability was i not corrected.

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d. The provisions of Specifications 3.0.3, 3.0.4, and 4.0.4 are not applicable. l SURVEILLANCE REQUIREMENTS 4.3.3.10 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST at the frequencies shown in Table 4.3-13.

ARKANSAS - UNIT 2 3/4 3-54 Amendment No. 60

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) _

9. Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed after the field hydrostatic test and prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
b. 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 4.4-2.

4.4.5.5 Reports

a. Following each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission within 15 days.
b. The complete results of the steam generator tube inservice inspection shall be reported on an annual basis for the period in which the inspection was completed. This report shall include:
1. Number and extent of tubes inspected.
2. Location and percent of wall-thickness penetration for each indication of an imperfection.
3. Identification of tubes plugged.
c. Results of steam generator tube inspections which fall into Category C-3 shall be reported in a Special Report pursuant to Specification 6.9.2 as denoted by Table 4.4-2. Notification of the Commission will be made prior to resumption of plant operation. The written Special Report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

ARKANSAS - UNIT 2 3/4 4-10 Amendment No.

20

x2 N

5 M TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION E

q IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Sample Size Result Action Required Result Action Required Result Action Required A miminum of C-1 None N/A N/A N/A N/A S Tubes per S. G.

C-2 Plug defective tubes C-1 None N/A N/A and inspect additional 25 tubes in this S.G. Plug defective tubes C-1 None C-2 and inspect additional C-2 P' ) defective tubes 45 tubes in this S.G.

Perform action for C-3 C-3 result of first sample Perform action for C-3 C-3 result of first N/A N/A w samnie g C-3 Inspect all tubes in this S.G., plug de-All other

5. G.s are None N/A N/A i a fective tubes and C-1

' inspect 25 tubes in i i ru each other S.G. Some S. G.s Perform action for N/A N/A l C-2 but no C-2 result of second l

additional sample Special Report S.G. are to NRC per C-3 Specification 6.9.2 Additional Inspect all tubes in l l '

S. G. is C-3 each S. G. and plug defective tubes.

Special Report N/A N/A to NRC per Spec. 6.9.2.

g ro 3 S=3N % Where N is the number of steam generators in the unit, and n is the number of steam generators inspected during during*an inspection

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PLANT SYSTEMS 3/4.7.12 SPENT FUEL POOL STRUCTURAL INTEGRITY

[IMITING CONDITION FOR OPERATION 3.7.12 The structural integrity of the spent fuel pcol shall be maintained in accordance with Specification 4.7.12.

APPLICABILITY: Whenever irradiated fuel assemblies are in the spent fuel pool.

ACTION:

a. With the structural integrity of the spent fuel pool not conforming to the above requirements, in lieu of any other report, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days of a determination of such non-conformity.
b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REOUIREMENTS 4.7.12.1 Inspection Frequencies - The structural integrity of the spent fuel pool shall be determined per the acceptance criteria of Specification 4.7.12.2 at the following frequencies:

1

a. At least once per 92 days after the pool is filled with water. If no abnormal degradation or other indications of structural distress are detected during five consecutive inspections, the inspection interval may be extended to at least once per 18 l months.
b. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following any seismic event which actuates or should have actuated the seismic monitoring instrumentation of Specification 3.3.3.3.

4.7.12.2 Acceptance Criteria - The structural integrity of the spent l

fuel pool shall be determined by a visual inspection of at least the interior and exterior surfaces of the pool, the struts in the tilt pit, the surfaces of the separation walls, and the structural slabs adjoining the pool walls. This visual inspection shall verify no changes in the concrete crack patterns, no abnormal degradation or other signs of structural distress (i.e, cracks, bulges, out of plumbness, leakage, discolorations, efflorescence, etc.).

ARKANSAS - UNIT 2 3/4 7-38 Amendment No.

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

10. Verifying that the auto-connected loads to each diesel generator do not exceed the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> rating of 3135 kw.
11. Verifying the diesel generator's capability to:

a) Synchronize with the offsite power source while the generator is loaded with its emergency loads upon a simulated restoration of offsite power, b) Transfer its loads to the offsite power source, and c) Proceed through its shutdown sequence.

12. Verifying that with the diesel generator operating in a test mode (connected to its bus), a simulated safety injection signal overrides the test mode by (1) returning the diesel generator to standby operation and (2) auto-matically energizes the emergency loads with offsite power.
13. Verifying that the fuel transfer pump transfers fuel from each fuel storage tank to the day tank of each diesel via the installed cross connection lines.
d. At least once per 10 years or after any modifications which could affect diesel generator interdependence by starting both diesel generators simultaneously, during shutdown, and verifying that both diesel generators acce'erate to at least 900 rpm in < 15 seconds.

4.8.1.1.3 Reports - See Specification 6.9.1.5.d.

ARKANSAS - UNIT 2 3/4 8-4a Amendment No.

J t RADIOACTIVE EFFLUENTS I DOSE

! LIMITING CONDITION FOR OPERATION i

3.11.1.2 The dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from ANO-2 to the discharge canal

. shall be limited:

a. During any calendar quarter to less than or equal to 1.5 mrem to the total body and to less than or equal to 5 mrem to any organ, and
b. During any calendar year to less than or equal to 3 mrem to the  !

total body and to less than or equal to 10 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, in lieu of any other report submit a Special Report pursuant to Specification i 6.9.2.h within 30 days.

1

b. The provisions of specifications 3.0.3 and 3.0.4 are not applicable. l SURVEILLANCE REQUIREMENTS 4.11.1.2 Dose Calculations. Cumulative dose contributions from liquid effluents shall be determined in accordance with the ODCM at least once per j

31 days.

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4 ARKANSAS - UNIT 2 3/4 11-4 Amendment No. 60 i

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, . - , - - - - . , - , -,--,_--,,,,,,-,y- ----.e-ve,.m.-, ,

1 RADI0 ACTIVE EFFLUENTS LIQUID RADWASTE TREATMENT LIMITING CONDITION FOR OPERATION 3.11.1.3 The LIQUID RADWASTE TREATMENT SYSTEM shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses due to the liquid effluent, from ANO-2 to the discharge cu.7al, would exceed .18 mrem to the total body or .625 mrem to any organ in any calendar quarter.

APPLICABILITY: At all times.

ACTION:

a. With radioactive liquid waste being discharged without treatment and in excess of the above limits, in lieu of any other report, submit a Special Report pursuant to Specification 6.9.2.h within 30 days.
b. The provisions of Specifications 3.0.3 and 3.0.4 are not l

applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.3.1 Doses due to liquid releases shall be projected at least once per 31 days in accordance with the ODCM.

ARKANSAS - UNIT 2 3/4 11-5 Amendment No. 50

RADIOACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS DOSE RATE LIMITING CONDITION FOR OPERATION 3.11.2.1 The dose rate due to radioactive materials released in gaseous effluents from the site to UNRESTRICTED AREAS (see Figure 5.1-3) shall be limited to the following:

a. For noble gases: Less than or equal to the 500 mrem /yr to the total body and less than or equal to 3000 mrem /yr to the skin.
b. For iodine-131, for tritium and for all radionuclides in particular form with half-lives greater than 8 days: Less than or equal to 1500 mrem /yr to any organ.

During periods of containment purging the dose rate may be averaged over a one hour interval.

APPLICABILITY: At all times.

ACTION:

a. With the dose rate (s) exceeding the above limits, without delay restore the release rate to comply with the above limit (s).
b. The provisions of Specifications 3.0.3 and 3.0.4 are not l applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of the ODCM.

4.11.2.1.2 The dose rate due to iodine-131, tritium, and all radionuclides in particulate form with half-lifes greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 4.11-2.

ARKANSAS - UNIT 2 3/4 11-7 Amendment No. 60

l RADI0 ACTIVE EFFLUENTS DOSE - NOBLE GASES LIMITING CONDITION FOR OPERATION 1

3.11.2.2 The dose due to noble gases released in gaseous effluents from l j ANO-2 to UNRESTRICTED AREAS (See Figure 5.1-3) shall be:

a During any calendar quarter, less than or equal to 5 mrads for gamma radiation and less than or equal to 10 mrads for beta radiation, and

b. During any calendar year, less than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for beta radiation.

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, in lieu of any other report, submit a Special Report pursuant to Specification 6.9.2.h within 30 days
b. The provisions of Specifications 3.0.3 and 3.0.4 are not l applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.2 Dose Calculations. Cumulative dose contributions for noble gases for the current calendar quarter and current calendar year shall be determined in accordance with the ODCM at least once per 31 days.

ARKANSAS - UNIT 2 3/4 11-10 Amendment No. 50

RADIOACTIVE EFFLUENTS DOSE - IODINE-131, TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM l

LIMITING CONDITION FOR OPERATION l

3.11.2.3 The dose to a MEMBER OF THE PUBLIC from iodine-131, from tritium, and from all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from ANO-2 to UNRESTRICTED AREAS (see Figure 5.1-1) shall be:

a. During any calendar quarter, less than or equal to 7.5 mrems to any organ, and
b. During any calendar year, less than or equal to 15 mrens to any organ.

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of iodine-131, tritium, and radionuclides in particulate form with half-lives greater than 8 days, in gasecus effluents exceeding any of the above limits, in lieu of ar.y other report, submit a Special Report pursuant to Specification 6.9.2.h within 30 days.
b. The provisions of Specifications 3.0.3 and 3.0.4 are not l applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.3 Dose Calculations. Cumulative dose contributions for the current calendar quarter and current calendar year for iodine-131, tritium, and radionuclides in particulate form with half-lives greater than 8 days shall be determined in accordance with the ODCM at least once per 31 days.

l ARKANSAS - UNIT 2 3/4 11-11 Amendment No. 60

RADI0 ACTIVE EFFLUENTS GASEOUS RADWASTE TREATMENT LililI1 E _CQFDITION FOR OPERATION 3.11.2.4 The VENTILATION EXHAUST TREATMENT SYSTEMS shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected gaseous effluent doses from ANO-2 to UNRESTRICTED AREAS (see Figure 5.1-1) would exceed .625 mrad for gamma radiation and 1.25 mrad for beta radiation in any calendar quarter; or when the projected doses due to iodine-131, tritium, and radionuclides in particulate form with half-lives greater than 8 days would exceed 1.0 mrem to any organ over a calendar quarter.

APPLICABILITY: At all times.

ACTION:

a. With gaseous waste being discharged without treatment and in excess of the above limits, in lieu of any other report, submit a Special Report pursuant to Specification 6.9.2.h within 30 days,
b. The provisions of Specifications 3.0.3 and 3.0.4 are not l applicable.

SURVEILLANCE REOUIREMENTS 4.11.2.4.1 Doses due to gaseous releases from the site shall be projected at least once per 31 days in accordance with the ODCM.

ARKANSAS - UNIT 2 3/4 11-12 Amendment No. 60 l

_ _ ---___-----____---_____]

RADI0 ACTIVE EFFLUENTS GASEOUS RADWASTE TREATMENT LIMITING CONDITION FOR OPERATION 3.11.2.5 When degasifying the reactor coolant system, the GASE0US RADWASTE TREATMENT SYSTEM shall be used to reduce radioactive material in gaseous waste prior to their discharge when the projected gaseous effluent doses for ANO-2 to UNRESTRICTED AREAS (see Figure 5.1-1) would exceed .625 mrad for gamma radiation and 1.25 mrad for beta radiation in any calendar quarter.

APPLICABILITY: At all times.

ACTION:

a. With gaseous waste being discharged without treatment and in excess of the above limits, in lieu of any other report, submit a Special Report pursuant to Specification 6.9.2.h within 30 days.
b. The provisions of Specifications 3.0.3 and 3.0.4 are not l applicable.

SURVEILLANCE REOUIREMENTS f 4.11.2.5.1 Doses due to gaseous releases from the site shall be projected at least once per 31 days in accordance with the ODCM.

f ARKANSAS - UNIT 2 3/4 11-13 Amendment No. 60

RADI0 ACTIVE EFFLUENTS 3/4.11.4 SOLID RADI0 ACTIVE WASTE LIMITING CONDITION FOR OPERATION 3.11.4 The solid radwaste system shall be used in accordance with a PROCESS CONTROL PROGRAM to process wet radioactive wastes to meet shipping and burial ground requirements.

APPLICABILITY: At all times.

ACTION:

a. With the provisions of the PROCESS CONTROL PROGRAM not satisfied, suspend shipments of defectively processed or defectively packaged solid radicactive wastes from the site.
b. The provisions of Specifications 3.0.3 and 3.0.4 are not l applicable.

SURVEILLANCE REOUIREMENTS 4.11.4 Proper solidification of wet radioactive waste shall be verified in accordance with the surveillance requirements of the Process Control Program.

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ARKANSAS - UNIT 2 3/4 11-16 Amendment No. 60

3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM LIMITING CONDITION FOR OPERATION 3.12.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12-1 and shall be analyzed pursuant to the requirements of Table 3.12-1 and 3.12-2. The sample locations shall be shown in Table 4-1 in the ODCM.

APPLICABILITY: At all times.

ACTION:

a. With the radiological environmental monitoring program not being conducted as specified in Table 3.12-1, prepare and submit to the Commissioned in the Annual Radiological Environmental Report a description of the reasons for not conducting the program as required and the plans for preventing a recurrence. (Deviations are permitted from the required sampling schedule if specimens are not obtainable due to hazardous conditions, seasonal unavailability, or to malfunction of sampling equipment. If the latter, every effort shall be made to complete corrective action prior to the end of the next sampling period).
b. With the level of radioactivity as the result of plant effluents in an environmental sampling medium at one or more of the locations specified in Table 3.12-1 exceeding the limits of Table 3.12-3 when averaged over any calendar quarter, prepare and submit to the Commission, within 30 days from the end of the affected quarter, a Special Report which includes an evaluation of any release conditions, environmental factors or other aspects which caused the l limits of Table 3.12-3 to be exceeded, and defines the actions taken to reduce radioactive effluents so that the potential annual dose to a member of the public is less than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2 and 3.11.2.3. When more than one of the radionuclides in Table 3.12-3 are detected in the sampling medium, this Special Report shall be submitted if:

l Concentration (1) , Concentration (2) . . . . . . . > 1. 0 reporting level (1) reporting level (2)

When radionuclides other than those in Table 3.12-3 are detected and are the result of plant effluents, this Special Report shall be j submitted if the potential annual dose to a MEMBER OF THE PUBLIC is equal to or greater than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2 and 3.11.2.3. This Special Report is not required l if the measured level of radioactivity was not the result of plant effluents, however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Report.

ARKANSAS - UNIT 2 3/4 12-1 Amendment No. 60

3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING LIMITING CONDITION FOR OPERATION (Continued)

c. With milk or fresh leafy vegetable samples unavailable from any of the sample locations required by Table 3.12-1, identify locations for obtaining replacement samples and add them to the radiological environmental monitoring program within 30 days.

The specific locations from which samples were unavailable may then be deleted from the monitoring program. Identify the causes of the unavailability of samples and identify the new location (s) for obtaining replacement samples in the next Semiannual Radioactive Effluent Release Report and also include in the j report a revised table for the ODCM reflecting the new location (s).

d. The provisions of Specifications 3.0.3 and 3.0.4 are not l applicable.

SURVEILLANCE REQUIREMENTS 4.12.1 The results of analyses performed on the radiological environmental monitoring samples shall be summarized in the Annual Radiological Environ-mental Report.

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i ARKANSAS - UNIT 2 3/4 12-2 Amendment No. 60 i

l

RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS LIMITING CONDITION FOR OPERATION 3.12.2 A land use census shall be conducted and shall identify the location of the nearest milk animal, the nearest residence, and the nearest garden

  • of greater than 500 square feet producing fresh leafy vegetables in each of the 16 meteorological sectors within a distance of five miles.

APPLICABILITY: At all times.

ACTION:

a. With a land use census identifying a location (s) which yields a calculated dose commitment due to I-131, tritium, and radio-nuclides in particulate form greater than the values currently being calculated in Unit 2 Specification 4.11.2.3, submit lo-cation description in the Semiannual Radioactive Effluent Release Report per Specification 6.9.3.
b. With a land use census identifying a location (s) which yields a calculated dose commitment (via the sample exposure pathway) greater than at a location from which samples are currently being obtained in accordance with the Specification 3.12.1, identify the new location in the Semiannual Radioactive Effluent Release Report per Specification 6.9.3. The new location shall be added to the radiological environmental monitoring program within 30 days, if possible. The sampling location having the lowest calculated dose commitment (via the same exposure pathway) may be deleted from this monitoring program after October 31 of the year in which this land use census was conducted.
c. The provisions of specifications 3.0.3 and 3.0.4 are not l

applicable.

SURVEILLANCE REOUIREMENTS 4.12.2 The land use census shall be conducted at least once per 12 months between the dates of June 1 and October 1 by door-to-door survey, aerial survey, or by consulting local agriculatural authorities. The results of the land use census shall be included in the Annnual Radiological Environmental Report.

ARKANSAS - UNIT 2 3/4 12-8 Amendment No. 60

RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM kJMITING CONDITION FOR OPERATION 3.12.3 Analyses shall be performed on radioactive materials supplied as part of the Interlaboratory Comparison Program which has been approved by NRC.

APPLICABILITY: At all times.

ACTION:

a. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Report.
b. The provisions of Specifications 3.0.3 and 3.0.4 are not I applicable.

SURVEILLANCE REOUIREMENTS 4.12.3 The results of analyses performed as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Report.

l ARKANSAS - UNIT 2 3/4 12-9 Amendment No. 60 l

I REACTOR COOLANT SYSTEM BASES Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in sevice, it will be found during scheduled inservice steam generator tubes examinations.

Plugging will be required for all tubes with imperfections exceeding the plugging limit of 40% of the tube nominal wall thickness. Steam generator 4 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 inservice inspection fall into Category C-3 certain results will be reported in a Special Report to the Commission pursuant to Specification 6.9.2 as denoted by Table 4.2-2. Notification of the Commission will be made 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.

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE l 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These detection systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems" May 1973.

3/4.4.6.2 REACTOR COOLANT SYSTEM LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 GPM. This threshold value is sufficiently low to ensure early detection of additional leakage.

The 10 GPM IDENTIFIED LEAKAGE limitation provides allowances for a limited amount of leakage from known soruces whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.

The Surveillance Requirements for RCS Pressure Isolation Valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS Pressure Isolation Valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

ARKANSAS - UNIT 2 B 3/4 4-3 Amendment No.

Ofddf ddtid M/20/81

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