Text

Proposed Tech Specs,Reflecting Recent Util Organizational Changes,Recent 10CFR50 Correspondence Requirements & Conflicts in Tech Spec Reporting Requirements & Revising Section 6.0 Re Audit Requirements
	ML13316B827
Person / Time
Site:	San Onofre
Issue date:	08/27/1987
From:	; SOUTHERN CALIFORNIA EDISON CO.
To:	;
Shared Package
ML13316B828	List: ML13316B827; ML13331A959; ML13331A963;
References
	TAC-66164, NUDOCS 8709020049
	Download: ML13316B827 (65)
	v • d • e

Attachment 1 3.5.8 RADIOACTIVE LIQUID EFFLUENT INSTRUMENTATION APPLICABILITY:

During releases via this pathway.

OBJECTIVE:

Monitor and control radioactive liquid effluent releases.

SPECIFICATION:

A. The radioadtive liquid effluent monitoring instrumenta tion channels shown in Table 3.5.8.1 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of Specification 3.15.1 are not exceeded.

B. Action

1. With a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint less conservative than a value which will ensure that the limits of 3.15.1 are met, without delay suspend the 86 release of radioactive liquid effluents monitored by 1/1/85 the affected channel or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

2. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.5.8.1.

If the inoperable instruments remain inoperable for greater than 30 days, explain in the next Semiannual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.

3. The provisions of Specifications 3.0.3, 3.0.4 and 90 6.9.2.b(2) are not applicable.

8/5/85 BASIS:

The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases.

The alarm/trip setpoints for these instruments are calculated in accordance with methods in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20.

709020049 870827 PDR ADOCK 0500206 6PDR6 R

3-66 Revised:

08/19/85

3.5.9 RADIOACTIVE GASEOUS PROCESS AND EFFLUENT MONITORING INSTRUMENTATION APPLICABILITY:

During releases via this pathway.

OBJECTIVE: Monitor and control radioactive gaseous releases.

SPECIFICATION: A. The radioactive gaseous process and effluent monitoring instrumentation channels shown in Table 3.5.9.1 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of Specification 3.16.1 are not exceeded.

B. ACTION

1. With a radioactive gaseous process or effluent monitoring instrumentation channel alarm/trip setpoint less conservative than a value which will ensure that the limits of 3.16.1 are met, without delay suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

2. With less than the minimum number of radioactive gaseous process or effluent monitoring instrumentation 86 channels OPERABLE, take the ACTION shown in Table 1/1/85 3.5.9.1.

If the inoperable instruments remain inoperable for greater than 30 days, explain in the next Semiannual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner..

90

3. The provisions of Specifications 3.0.3, 3.0.4 and 8/5/85 6.9.2.B(2) are not applicable.

BASIS The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases.

The alarm/trip setpoints for these instruments are calculated in accordance with methods in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20.

3-69 Revised:

08/19/85 Typo Revised 8/30/85

3.15.2 LIQUID EFFLUENT DOSE APPLICABILITY:

At all times.

OBJECTIVE:

Maintain the release of radioactive liquid effluents from the site as low as is reasonably achievable.

SPECIFICATION: A. The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released to UNRESTRICTED AREAS (see Figure 5.1-1) shall be limited:

1. During any calendar quarter to < 1.5 mrem to the total body and to < 5 mrem to any organ, and 86 1/1/85

2. During any calendar year to < 3 mrem to the total body and to < 10 mrem to any organ.

B. Action:

1. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the 91 limit(s) and defines the corrective actions that have 11/14/85 been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

2. The provisions of Specification 3.0.3 and 3.0.4 are 90 not applicable.

8/5/85 BASIS:

This specification is provided to implement the requirements of Section II.A and IV.A of Appendix I, 10 CFR Part 50.

Specification A implements the guides set forth in Section II.A of Appendix I. Specification B provides the required 86 operating flexibility and at the same time implements the 1/1/85 guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable."

3-106 Revised:

11/14/85

3.15.3 LIOUID WASTE TREATMENT APPLICABILITY: At all times.

OBJECTIVE:

Maintain radioactive releases from the site as low as is reasonable achievable by use of the liquid radwaste treatment system.

SPECIFICATION: A. The liquid radwaste treatment system shall be used to 86 reduce the radioactive materials in liquid wastes prior to 1/1/85 their discharge when the projected dose due to the liquid effluent from San Onofre Unit 1, to UNRESTRICTED AREAS (see Figure 5.1-1) would exceed 0.06 mrem to the total body or 0.2 mrem to any organ in a 31 day period.

B. Action:

1. With radioactive liquid waste being discharged without 90 treatment and in excess of the above limits, in lieu 8/5/85 of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specification 91 6.9.2, a Special Report that includes the following 11/14/85 information:

a. Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems and the reason for inoperability.

b. Action(s) taken to restore the inoperable 90 equipment to OPERABLE status.

8/5/85

c. Summary description of action(s) taken to prevent a recurrence.

2. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

BASIS:

The OPERABILITY of the liquid radwaste treatment system ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirements that the appropriate portions of this system be used when specified provides assurance that 86 the releases of radioactive materials in liquid effluents will 1/1/85 be kept uas low as is reasonable achievable.u This specification implements the requirements of 10 CFR Part 50.36a and the design objective given in Section II.0 of Appendix I to 10 CFR Part 50.

The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the guide set forth in Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents.

3-107 Revised:

11/14/85

3.16.2 DOSE, NOBLE GASES APPLICABILITY: At all times.

OBJECTIVE:

Maintain the dose due to noble gases in gaseous effluents as low as is reasonable achievable.

SPECIFICATION: A. The air dose due to noble gases released in gaseous effluents, from San Onofre Unit 1 to areas at and beyond the SITE BOUNDARY (see Figure 5.1-1) shall be limited to the following:

1. During any calendar quarter:

< 5 mrad for gamma 86 radiation and < 10 mrad for beta radiation.

1/1/85

2. During any calendar year: < 10 mrad for gamma radiation and < 20 mrad for beta radiation.

B. Action:

1. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission-within 30 days, pursuant to Specification 6.9.2, a Special Report that 91 identifies the cause(s) for exceeding the limit(s) and 11/14/85 defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

2. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

BASIS:

This specification is provided to implement the requirements 86 of Section II.B and IV.A of Appendix I, 10 CFR Part 50.

1/1/85 Specification A Implements the guides set forth in Section II.B of Appendix I. Specification B provides the required operating flexibility and at the same time implements the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonable achievable."

3-109 Revised:

11/14/85

3.16.3 DOSE, IOOINE-131, IODINE-133, TRITIUM AND RADIONUCLIDES IN PARTICULATE FORM APPLICABILITY: At all times.

OBJECTIVE:

Maintain the dose due to radioiodines, radioactive materials in particulate form and radionuclides other than noble gases in gaseous effluents as low as is reasonable achievable.

SPECIFICATION: A. The dose to a MEMBER OF THE PUBLIC from 1-131, 1-133, from 86 tritium and from all radionuclides in particulate form I/as with half-lives greater than 8 days-in gaseous effluents 1/

released from San Onofre Unit 1 to areas at and beyond the SITE BOUNDARY (see Figure 5.1-1) shall be limited to the following:

1. During any calendar quarter:

< 7.5 mrem to any organ; and

2. During any calendar year:

< 15 mrem to any organ.

B. Action:

1. With the calculated dose from the release of 1-131, 1-133, tritium and radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) 91 for exceeding the limit and defines the corrective 11/14/85 actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

2. The provisions of Specification 3.0.3 and 3.0.4 are 90 not applicable.

8/5/85 BASIS:

This specification is provided to implement the requirements of Sections II.C and IV.A of Appendix I. 10 CFR Part 50.

Specification A Is the guide set forth in Section II.C of Appendix I. Specification B provides the required operating 86 flexibility and at the same time implements the guides set 1/1/85 forth in Section IV.a of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonable achievable."

3-110 Revised:

11/14/85

3.16.4 GASEOUS RADWASTE TREATMENT APPLICABILITY: At all times.

OBJECTIVE:

Maintain radioactive gaseous releases from the site as low as is reasonable achievable by use of the GASEOUS RADWASTE and VENTILATION EXHAUST TREATMENT SYSTEMS.

SPECIFICATION: A. The GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected gaseous effluent air doses due to gaseous effluent releases from San Onofre Unit 1 to areas at and beyond the SITE BOUNDARY (see Figure 5.1-1) would exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiation over 31 days.

The VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases from San Onofre Unit 1 to areas at and beyond the SITE BOUNDARY (see Figure 5.1-1) would exceed 0.3 mrem to any organ over 31 days.

B. Action:

1. With gaseous waste being discharged without treatment 86 and in excess of the above limits, in lieu of a 1/1/85 Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specification 91 6.9.2, a Special Report which includes the following 11/14/85 information:

a. Explanation of why gaseous radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems and the reasons for the inoperability.

b. Action(s) taken to restore the inoperable equipment to OPERABLE status.

c. Summary description of action(s) taken to prevent a recurrence.

2. The provisions of Specification 3.0.3 and 3.0.4 are 90 not applicable.

8/5/85 BASIS:

The OPERABILITY of the GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensurers that the 86 systems will be available for use whenever gaseous effluents 1/1/85 require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used when specified provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable."

This 3-111 Revised: 11/14/85

Specification implements the requirements of IOCFR Part 86 50.36a, and the design objective given in Section II.D of 1/1/85 Appendix I to 10CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the guide set forth in Section II.B and II.C of Appendix I, IOCFR Part 50, for gaseous effluents.

T h i s P a g d ed.

FOR REFERENCE ONLY Typo Revision:

2/12/85 3-112 Revised: 01/01/85

.3.17 DOSE APPLICABILITY:

At all times.

OBJECTIVE:

Maintain the dose due to the release of radioactive materials within specified limits.

SPECIFICATION: A. The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC, due to releases of radioactivity and to radiation, from uranium fuel cycle sources shall be limited to < 25 mrem to the total body or any organ (except the thyroid,which shall be limited to < 75 mrem).

86 1/1/85 B. Action:

1. With the calculated dose from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specifications 3.15.2.A, 3.16.2.A or 3.16.3.A, calculations should be made to determine whether the above limits of Specification 3.17 have been exceeded. If such is the case, in lieu of a Licensee Event Report, prepare and submit to the 91 Commission within 30 days pursuant to Specification 11/14/85 6.9.2, a Special.Report that defines the corrective action to be taken to reduce subsequent releases, to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits.

The Special Report, as defined in 10 CFR Part 20.405c, shall includean analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release(s) covered by this report. It shall also describe levels of 86 radiation and concentrations of radioactive material 1/1/85 involved, and the cause of the exposure levels or concentrations. If the estimated dose(s) exceeds the above limits, and if the release condition resulting in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

3-115 Revised:

11/14/85

2.

The provisions of Specification 3.0.3 and 3.0.4 are 90 not applicable.

8/5/85 86 BASIS:

This specification is provided to meet the reporting 1/1/85 requirements of 40 CFR 190. In complying with 40 CFR 190, nuclear fuel cycle facilities over five miles away are not considered to contribute to the dose assessment.

This page included FOR REFERENCE ONLY 3-116 Revised:

08/19/85

3.18 RADIOLOGICAL ENVIRONMENTAL MONITORING 3.18.1 MONITORING PROGRAM APPLICABILITY:

At all times.

OBJECTIVE:.

Monitor exposure pathways for radiation and radioactive

-material.

SPECIFICATION: A. The radiological environmental monitoring program shall be conducted as specified in Table 3.18.1.

B. Action:

1. Withathe radiological environmental monitoring program not being conducted as specified in Table 3.18.1, in lieu of a Licensee Event Report, prepare and submit to the Commission, in the Annual Radiological Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.

2. With the level of radioactivity as the result of plant 86 effluents in an environmental sampling medium 1/1/85 exceeding the reporting levels of Table 3.18.2 when averaged over any calendar quarter, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days from the end of the affected calendar quarter a Special Report pursuant to Specification 6.9.2. When more than one of the 91 radionuclides in Table 3.18.2 are detected in the 11/14/85 sampling medium, this report shall be submitted if:

concentration (1)

+ concentration (2)

+... > 1.0 reporting level (1) reporting level (2)

When radionuclides other than those in Table 3.18.2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to a MEMBER OF THE PUBLIC is equal to or greater than the calendar year limits of Specifications 3.15.2, 3.16.2 and 3.16.3. This report is not required 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 Operating Report.

3-117 Revised:

11/14/85

3. With fresh leafy vegetable samples or fleshy vegetable samples unavailable from one or more of the sample locations required by Table 3.18.1, prepare and submit to the Commission within 30 days, pursuant to 91 Specification 6.9.2, a Special Report which identifies 11/14/85 the cause of the unavailability of samples and identifies locations for obtaining replacement samples. The locations from which samples were unavailable may then be deleted from those required by 86 Table 3.18.1, provided the locations from which the 1/1/85 replacement samples were obtained are added to the environmental monitoring program as replacement locations.

4. The provisions of Specification 3.0.3 and 3.0.4 are 90 not applicable.

1/1/85 BASIS:

The radiological monitoring program required by this specification provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides, which lead to the highest potential radiation exposures of a MEMBER OF THE PUBLIC resulting from the station 86 operation. This monitoring program thereby supplements the 1/1/85 radiological effluents monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure pathways. The initially specified monitoring program will be effective for at least the first three years of commercial operation. Following this period, program changes may be initiated based on operational experience.

3-11is page Included FOR REFERENCE ONLY 3-118 Revised:

11/14/85

3.18.2 LAND USE CENSUS APPLICABILITY:

At all times.

OBJECTIVE:

Monitor the UNRESTRICTED AREAS.surrounding the site for potential changes to the radiological monitoring program as necessary.

SPECIFICATION: A. 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 86 square feet producing fresh leafy vegetables in each of 1/1/85 the 16 meteorological sectors within a distance of five miles.

B. Action:

1. With the land use census identifying a location(s) which yields a calculated dose or dose commitment greater than the values currently being calculated in Specification 4.6.3, in lieu of a Licensee Event 91 Report, prepare and submit to the Commrission within 11/14/85 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the new locations. Identify the new locations in the next Semiannual Radioactive Effluent Release Report.

2. With a land use census identifying a location(s) which yields a calculated dose or dose commitment via the same exposure pathway 20 percent greater than at a location fr6m which samples are currently being obtained in accordance with Specification 3.18.1, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies 91 the newlocations. The new location shall be added to 11/14/85 the radiological environmental monitoring program within 30 days. The sampling location, excluding the control station location, having the lowest calculated dose or 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.

Broad leaf vegetation sampling may be performed at the SITE BOUNDARY in the direction section with the highest D/Q in lieu of the garden census.

3-123 Revised:

11/14/85

3. The provisions of Specification 3.0.3 and 3.0.4 are 90 not applicable.

8/5/85 BASIS:

This Specification is provided to ensure that changes in the use of UNRESTRICTED AREAS are identified and that modifications to the monitoring program are made if required by the results of this census.

The best survey information 86 from the door-to-door, aerial or consulting with local 1/1/85 agricultural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10CFR Part 50.

Restricting the census to gardens of greater than 500 square feet provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (25 kg/year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were used, (1) that 20% of the garden was used for growing broad leaf vegetation (i.e.,

similar to lettuce and cabbage), and (2) a vegetation yield of 2 kg/square meter.

314TRhes page08/19/

FOR REFERENCE ONLY) 3-124 Revised:

08/19/85

3.18.3 INTERLABORATORY COMPARISON PROGRAM APPLICABILITY:

At all times.

OBJECTIVE:

To ensure laboratory analysis of radiological environmental monitoring samples is correct and accurate.

SPECIFICATION:

A.

Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program which has been approved by the Commission.

B.

Action:

1. 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 Operating Report.

86 1/1/85

2. The provisions of Specification 3.0.3, 3.0.4 and 90 6.9.2.b(2) are not applicable.

8/5/85 BASIS:

The requirement for participation in an Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid.

3-125 Revised:

08/19/85

3.19 SOLID RADIOACTIVE WASTE APPLICABILITY:

At all times.

OBJECTIVE:

Ensure meeting the requirements for the SOLIDIFICATION and shipment of solid radwaste.

SPECIFICATION:

A.

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.

B.

Action:

86

1. With the provisions of the PROCESS CONTROL PROGRAM 1/1/85 not satisfied suspend shipments of defectively processed or defectively packaged solid radioactive wastes from the site.

2. The provisions of Specification 3.0.3 and 3.0.4 and 90 6.9.2b(2) are not applicable.

8/5/85 BASIS:

This specification implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50.

The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to waste type, waste pH, waste/liquid/

solidification/agent/catalyst ratios, waste oil content, waste principal chemical constituents, mixing and curing times.

3-126 Revised:

08/19/85

This page included FOR REFERENCE ONLY 4.2.3 SAFETY INJECTION SYSTEM HYDRAULIC VALVE TESTING (SURVEILLANCE REQUIREMENT)

An interim surveillance testing program shall be conducted during the remainder of the current fuel cycle which began in June 1981.

At the next refueling outage, the interim program shall be supplanted by a long term surveillance testing program. It is 63 intended that this long term program will be developed and 11/5/81 submitted to the NRC for review and approval at least 60 days prior to the next refueling outage.

The interim surveillance program shall be as follows:

1.

At least once every 92 days, (except when the inverval lapses while in mode 5 or 6, in which case the test may be 67 delayed until a mode 3 or 4 operation prior to the next 9/24/82 entry into mode 2) the unit shall be placed in mode 3 or 4 and a Hot SIS functional test (with the MOV-850 A, B&C valves locked closed) shall be performed. This test shall include a determination of the force required to open valves HV-851 A&B and the margin to available actuation force. This test shall be evaluated on the basis of the following criteria:

a. If the measured actuator force for both the HV-851 A&B valves is less than 10,000 lbf*, the unit may be returned to power.

b. If the measured actuator force of either HV-851 A or B is between 10,000 and 22,000 lbf, the Hot SIS test for both valves shall be repeated to again determine required opening force and available margin. The prediction will assume a straight line extrapolation 63 from the following equation:

11/5/81 (22,000 - F 2)

T =

(F -

F)/T 1

L where F1 =

measured actuator force from the first Hot SIS test during the current surveillance test (lbf)

F2 = measured actuator force from the second Hot SIS test during the current surveillance test (lbf)

Upon receipt of satisfactory data from continuing testing and analysis, the NRC staff will consider a request from Southern California Edison Company to change this number to more accurately reflect existing conditions.

4-41 Revised:

9/2/82

TL =

time (in days) since the last surveillance testing F =

the actuator force from the previous surveillance test (lbf)*

If the calculated value of T does not exceed 92 days, 63 the next surveillance test must be performed before T 11/5/81 days had elapsed.

c. If the measured actuator force of either HV-851 A or B is greater than 22,000 lbf, the valve(s) shall be declared inoperable. Test results shall be reported to 91 the NRC pursuant to Specification 6.6 along with 11/14/85 proposed corrective actions and NRC approval obtained prior to returning the unit to service.

2.

The first test shall be performed not less than 14 days nor more than 21 days following return to power from the current outage which began September 3, 1981.

For the first surveillance test, the value of F shall be the average actuator force of HV-851 A&B valves from pre-operation testing (3135 lbf).

All subsequent surveillance testing shall assume the F2 value from the previous surveillance test for each valve. If an P2 was not required during the previous surveillance test, the F1 value for each valve shall be assumed.

4-42 Revised:

11/14/85

3.

Snubber release rate, where required, is within the specified range in compression or tension. For snubbers specifically required not to displace under continuous load, the ability of the snubber to withstand load without displacement shall be verified.

F. Snubber Service Life Monitoring A record of the service life of each snubber, the date at which the designated service life commences and the installation and maintenance records on which the designated service is based shall be maintained as required by Specification 6.10.2.k.

Concurrent with the first in-service visual inspection and at least once per refueling cycle shutdown thereafter, the installation and maintenance records for each snubber shall be reviewed to verify that the indicated service life has not been exceeded or will not be exceeded prior to the next scheduled snubber service life review. If the indicated service life will be exceeded prior to the next scheduled snubber service life review, the snubber service life shall be reevaluated or the snubber shall be replaced or reconditioned so as to extend its service life beyond the date of the next scheduled service life review. This reevaluation, replacement or reconditioning shall be 81 indicated in the records.

10/16/84 BASIS:

Refer to the basis given in 3.13.

4-86 Revised:

10/22/84

4.16 INSERVICE INSPECTION OF STEAM GENERATOR TUBING APPLICABILITY:

Applies to the inservice inspection and sampling selection for steam generator tubing.

OBJECTIVE:

To monitor the integrity of the steam generator tube primary boundary and provide guidance for corrective action when imperfections are observed.

SPECIFICATION: A. GENERAL STEAM GENERATOR TUBE SELECTION The steam generators shall be inspected when shutdown by 44 selecting steam genertor tubes on the following basis:

10/31/78

1.

Tubes for the inpsection shall be selected on a random basis except where experience at San Onofre Unit 1 or experience in similar plants indicates critical areas to be inspected.

2.

Each inspection shall include at least 3% of the total number of tubes in each steam generator to be inspected.

3.

Inservice inspections may be limited to one steam generator on a rotating schedule encompassing 3% of the total tubes of steam generators in the plant if the results of previous inspections indicate that all steam generators are performing in a like manner.

4.

Every inspection shall include all non-plugged tubes in the steam generators to be inspected that previously had detected imperfections greater than 20%, except as specified in Specification C.I.

B. SUPPLEMENTARY INSPECTIONS If the inspections in Specification A indicate imperfections, additional inspections shall be required as follows:

1.

If any of the tubes inspected pursuant to Specification A.3 have imperfections greater than 20%

that were not detected during the previous inspections or have previously detected imperfections that have increased more than 10% wall penetration since their last inspection, inspect 3% of the tubes in one of theuninspected steam generators.

This page included FOR REFERENCE ONLY 4-90 Revised:

12/15/78

2.

If more than 10% of the tubes inspected in a steam generator have imperfections greater than 20% that were not detected during the previous inspections or have previously detected imperfections that have increased more than 10% wall penetration since their last inspection, or one or more of the tubes inspected have an imperfection in excess of.the plugging limit, inspect an additional 3% of the tubes in that steam generator, concentrating on tubes in those areas of the tube sheet array where tubes with imperfections were found and on that length of tube where the imperfections were found.

In addition., the 44 rest of the steam generators shall be inspected in 10/31/78 accordance with Specification A.2.

3.

If the additional inspection in Specification B.2 indicates that more than 10% of the additionally inspected tubes have imperfections greater than 20%

that were not detected during the previous inspections or have previously detected imperfections that have increased more than 10% wall penetration since their last inspection, or one or more of the additionally inspected tubes have an imperfection in excess of the plugging limit, inspect an additional 6% of the tubes in that steam generator in the area of the tubesheet array where tubes with imperfections were found and through that length of tube where the imperfections were found.

C. SPECIAL STEAM GENERATOR TUBE INSPECTIONS In addition to the general steam generator tube inspections performed in Specifications A and B, every inspection shall include the following special inspections:

1.

Every inspection shall include all nonplugged tubes in one of the steam generators that previously had been noted as having discretely quantifiable imperfections greater than 30% at antivibration bar (AVB) intersections, and all non-plugged tubes in that steam generator that previously had been noted as having imperfections at AVB intersections which were not discretely quantifiable but which were identified during previous inspections as being in the 30 to 50% range.

2.

At each steam generator inspection, all previously identified restricted tubes in either steam generator A or C shall be gauged by using eddy current probes to determine restriction sizes.

This page included 4-91 Revised:

12/15/78 OR REFERENCE ONLY

D. INSPECTION FREQUENCY The inspections in Specifications A and B above shall be performed at the following frequencies:

1.

Inservice inspections shall be not less than 10 nor more than 24 calendar months after the previous inspection.

2.

If two consecutive inspections indicate that less than 10% of the tubes inspected have either (a) new imperfections greater than 20% or (b) previous imperfections that have increased more than 10% since their last inspection, the inspections shall be not less than 10 nor more than 40 calendar months after the previous inspection.

3.

Unscheduled inspections shall be conducted in accordance with Specification A in the event of primary-to-secondary leaks exceeding Specification 44 3.1.4.C, a seismic occurrence greater than an 10/31/78 operating.basis earthquake, a loss-of-coolant accident requiring actuation of engineered safeguards, or a major steam line or feedwater line break.

E. ACCEPTANCE CRITERIA

1.

As used in this specification:

a. Imperfection means an exception to the dimensions, finish, or contour required by drawing or specification.

b. Defect means an imperfection of such severity that the tube is unacceptable for continued service.

c. Plugging limit means the imperfection depth at or 60 beyond which plugging of the tube must be 6/8/81 performed. The plugging limit is equal to or greater than 50% of the nominal tube wall thickness, except where sleeves are installed, in which case the plugging limit is equal to or greater than 40% of the nominal sleeve wall thickness.

This page included FOR ReviREd:E12NLY 4-92 Revised:

12/15/78

2. If, in the inspections performed under Specification A,

a. Less than 10% of the total tubes inspected have imperfections greater than 20% that were not detected during the previous inspections or have previously detected imperfections that have increased more than 10% wall penetration, and

b. No tube inspected exceeds the plugging limit, plant operation may resume.

3. If, in the inspections performed under Specification B,

a. Less than 10% of the total tubes inspected have imperfections greater than 20% that were not detected during the previous inspections or have 44 previously detected imperfections that have 10/31/78 increased more than 10% wall penetration, and

b. No more than 3 of the tubes inspected exceed the plugging limit, plant operation may resume after performance of the corrective action in Specification F.

4. If, in the inspections performed under Specification B,

a. More than 10% of the tubes inspected have imperfections greater than 20% that were not detected during the previous inspections or have previously detected imperfections that have increased more than 10% wall penetration, or

b. More than 3 of the tubes inspected exceed the plugging limit, the situation shall be reported to the Commission in accordance with Technical Specification 6.6 for 91 approval of the proposed remedial action.

11/14/85

5. If in the inspections performed under Specification C.1, wear rates are observed at AVB intersections which are inconsistent with the 50% plugging criterion, the situation shall be reported to the Commission in accordance with Technical Specification 6.6 for approval of the proposed 91 remedial action.

11/14/85 4-93 Revised:

11/14/85

6.

If in the inspections performed under Specification C.2 progression of the denting process is observed to be recurring, the situation shall be reported to the Commission in accordance with Technical Specification 6.6 for 91 approval of the proposed remedial action.

11/14/85 F. CORRECTIVE ACTION All leaking tubes, defective tubes, and tubes with imperfections exceeding the plugging limit shall be repaired or plugged.

BASIS:

The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the Reactor Coolant System will be maintained. The program for inservice inspection of steam generator tubes is 44 based on Regulatory Guide 1.83, Revision 1. Inservice 10/31/78 inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes.

If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking.

The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage =.3 gallons per minute per steam generator).

Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents.

Operating plants have demonstrated that primary-to-secondary leakage of.3 gpm per steam generator can readily be detected by radiation monitors of steam generator 44 blowdown.

Leakage in excess of this limit will require 10/31/78 shutdown during which the leaking tubes will be located and plugged and additional inspections performed.

4-94 Revised:

11/14/85

If a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations.

Plugging will be required for all tubes with imperfections 60 exceeding the plugging limit of 50% of the tube nominal wall 6/8/81 thickness, except where sleeves are installed, in which case the plugging limit is 40% of the nominal sleeve wall thickness. A plugging limit of 50% for tubes and 40% for sleeves ensures that defects will not occur between inspection intervals.

The results of tube ID gauging and dent detection conducted in San Onofre Unit I steam generators demonstrate that the denting process has been arrested. Continuing assurance of this condition can be provided by performing a program of limited tube ID gauging and dent detection in either steam generator A or C on a refueling outage frequency. Adequate surveillance of denting related tube restrictions can be maintained at refueling intervals by noting any new 44 restrictions during the conduct of general surveillance and 10/31/78 AVB inspections and by gauging tubes which have previously been noted as being restricted. Progression of denting can also be monitored in either steam generator A or C by evaluating third and fourth support plate denting data obtained from the general surveillance and AVB inspections as well as from the ID gauging program and comparing these results with those of previous inspections.

The results of AVB area inspections conducted in San Onofre Unit I steam generators demonstrate that AVB modifications installed during the Cycle VI refueling outage were.successful in eliminating significant growth of tube wall penetration indications at AVB locations.

Continuing assurance of this condition can be provided by performing U-bend inspections at refueling outage intervals of tubes having wall penetration indications in excess of 30% at AVB locations.

Ries Page ledcluded FOR REFERENCE ONLY 4-95 Revised:

6/23/81

4.19 SOLID RADIOACTIVE WASTE APPLICABILITY:

At all times.

OBJECTIVE:

Ensure meeting the requirements for the SOLIDIFICATION and shipment of solid radwaste.

SPECIFICATION:

The PROCESS CONTROL PROGRAM shall be used to verify the SOLIDIFICATION of at least one representative test specimen from at least every-tenth batch of each type of wet radioactive waste (e.g., filter sludges, spent resins, evaporator bottoms, boric acid solutions, and sodium sulfate solutions).

a. If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICATION of the batch under test shall be suspended until such time as additional test specimens can be 86 obtained, alternative SOLIDIFICATION parameters can be 1/1/85 determined in accordance with the PROCESS CONTROL PROGRAM, and a subsequent test verifies SOLIDIFICATION.

SOLIDIFICATION of the batch may then be resumed using the alternative SOLIDIFICATION parameters determined by the PROCESS CONTROL PROGRAM.

b. If the initial test specimen from a batch of waste fails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection and testing of representative test specimens from each consecutive batch of the same type of wet waste until at least three conse cutive initial test specimens demonstrate SOLIDIFICATION. The PROCESS CONTROL PROGRAM shall be modified as required, as provided in Specification 6.12, to assure SOLIDIFICATION of subsequent batches of waste.

BASIS:

This specification implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to waste type, waste pH, waste/liquid/solidifica tion agent/catalyst ratios, waste oil content, waste principal chemical constituents, mixing and curing times.

Typo Revision:

2/12/85 4-103 Revised: 01/01/85

6.2 ORGANIZATION OFFSITE 6.2.1 The offsite organization for unit management technical support shall be as shown on Figure 6.2-1.

UNIT STAFF 6.2.2 The Site organization shall be as shown on Figure 6.2-2 and:

a. Each on-duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1.

b. At least one licensed Operator shall be at the controls when fuel is in the reactor.* During refueling operations, this operator is permitted to step outside the red line to update the refueling status board. In addition, while the unit is in MODE 1, 2, 3 or 4, at least one licensed Senior Reactor Operator shall be in the Control Room Area.**

c. A health physics technician# shall be on-site when fuel is in the reactor.

d. All CORE ALTERATIONS shall be observed and directly supervised by a 91 licensed Senior Reactor Operator or Senior Reactor Operator Limited 11/14/85 to Fuel Handling who has no other concurrent duties during this operation.

e. A Fire Brigade of at least five members shall be maintained on site at all times.#

The Fire Brigade shall not include the Shift Superintendent and the two other members of the minimum shift crew necessary for safe shutdown of the unit and any personnel required for other essential functions during a fire emergency.

"At the controls" means within the area bounded by the three vertical instrumentation boards and the red line on the floor of the control room.

"Control Room Area" is defined by the control room and the Shift Superintendent's office.

The health physics technician and Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence provideO immediate action is taken to fill the required positions.

SAN ONOFRE - UNIT 1 6-2 Revised: 11/14/85

This page included FOR REFERENCE ONLY

f. Administrative procedures shall be developed and implemented to 91 limit the working hours of unit staff in the following job 11/14/85 classifications:

1) Shift Superintendents, Control Room Supervisors, Control Operators, Assistant Control Operators, Nuclear Plant Equipment Operators, Plant Equipment Operators;

2) Electricians and their first line supervisors;

3) I&C Technicians, Computer Technicians, Test Technicians and their first line supervisors;

4) Operational Health Physics Technicians and their first line supervisors;

5) Boiler and Condenser Mechanics, Machinists, Welders, Crane Operators and their first line supervisors;

6) Contractor or other Department personnel performing functions 89 identical to those performed by personnel identified in items 1 3/6/85 through 5 above and within the organizational framework of the Station.(1)

Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel identified above work a normal 8-hour day, 40-hour week (excluding shift turnover and meal time) while the plant is operating (MODES 1, 2, 3 and 4).

However, in the event that overtime which exceeds 25%(11) of normal time is required due to unforseen problems(iii) or during extended outages(iv), on a temporary basis, the following guidelines shall be followed:

(1)

Shift Technical Advisors are exempt from the overtime guidelines specified, since sleeping accommodations are provided.

(1) 25% is established as a level of overtime which will not significantly reduce the effectiveness of personnel, but which requires additional management approval prior to exceeding this level.

(III)

Unforeseen problems are forced shutdowns or power reductions of any unit, equipment failure or unscheduled repair, surveillance, calibration or maintenance, entry into a Technical Specification ACTION Statement or the absence of personnel required to provide normal shsft coverage.

(iv) Extended outages are periods in Modes 5 and/or 6.

SAN ONOFRE - UNIT 1 6-3 Revised:

11/14/85

1) An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight (shift turnover and meal time are not included when calculating hours worked).

2) An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 24-hour period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48-hour period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any 7-day period (shift turnover and meal time are not included when calculating hours worked).

3) A break (the time an individual leaves the work location to the time an individual returns to the work location) of at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months should be allowed between work periods (shift turnover time is not included when calculating the break; meal time is not included when calculating the break, unless it represents an 89 administrative entry on the timesheet and not extra hours spent at 3/6/85 the work location).

4) Except during extended shutdown periods, overtime should be considered on an individual basis and not for the entire staff on a shift.

Any deviation from the above guidelines shall be authorized by the Station Manager, the Deputy Station Manager, the Manager, Operations, the Manager, Maintenance, the Manager of Nuclear Generation Services or higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation.

Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the Station Manager or his designee to assure that excessive hours have not been assigned.

Routine deviation from the above guidelines is not authorized.

SAN ONOFRE -

UNIT 1 6-4 Revised:

3/22/85

Chairman of the Board Executive F

Vice President rn C)

CIT Senior Vice President Vice President a

vice President Vice President (Engineering Vice President

& Site Manager (Nuclear Engineer-Vice President and (Power Supply)

I (Nuclear a

ng, Safety and (Fuel Supply)

Construction)

Generation Site)

Licensing)

Manager manager of of Nuclear

Manager, Manager, Engieerng eneatin a

Nuclear Nuclear Engineering Engineering Division Services

[Headquarters UMaern

Manager, Manager,

& Siter Mana

(

uality Nuclear (Administrative Asafe adfel y

Servicesite)aLcesing)

I I

I

Manager, SNcetr Budgets and GrSafet i

Cost Control aGru 1I

tManager, Neadquarters a

ucNc (D

Ltaicensing TrainingEa StatisOn Site a S~teSalltanager Review a

SManager Committee a

Maeiaad a

Maae.Mngr a

a a

a c,

L---------

FIGURE 6.2-1 OFFSITE ORGANIZATION SAN ONOFRE NUCLEAR GENERATING STATION

-UNIT 1

Vice President a Site Manager Nuclear Generation Site C)

C:)

01 Manager of Station Manager Nuclear Generation Services Depty Station Manager. Operations Assigned Manager and Maintenance Support Proct Managers

Manage, Nuclear Tiraining 4

Manager, Station Emergency Manage,.

Manager, preparedness Station Compliance Station Security Manager.

Budgets and Cost Control Mana*er.

manage,.

Manager Manager.

Manager Health Physics Technical Operations Planning & Control Maintenance Health Physics Supervising Plant I

Planning A Control Assistant Administrative NSS5 Superintendent Supeisnr Maintenance Manager Services Supdrio Entear P er Um1 nineing Unit 1 Unit 1 Engineering and Uonintr Sue visingeric s

Redwst e Supervising Supevisr En ees Poerhift Supervisor MaintenanceMage Superntenents Coordination Deosimetry Suprviin Supervisor Superviso Engineer Cotrol 2

Plant Maintenance Computers RomCoordinatosPang Supervisore Health Physics SuperivisingI I'D Engpneering Engineer Operatorsr o

/1,110 1

I I'D Supervisor of PlanI SupervisorPln Chemistry Maintenance Ptenin Mechanical & lctia 4 -

1. At time of appointment to the position Senior Reactor Operator License required O0 2, Setor Reactor License required U-1

3. Control and Assistant Control Operator ave FIGURE 6.2.2 holde.

fReco Operator License, SITE ORGANIZATION Includes hr protection SAN ON FRE NUCLEAR GENERATING STATION

This page included 6.5.3 NUCLEAR SAFETY GROUP (NSG)

FOR REFERENCE ONLY FUNCTION 6.5.3.1 The Nuclear Safety Group shall function to provide independent review and audit of designated activities in the areas of:

a. nuclear power plant operations

b. nuclear engineering

c. chemistry and radiochemistry

d. metallurgy

e. instrumentation and control

f. radiological safety

q. mechanical and electrical engineering

h. quality assurance practices COMPOSITION 91 6.5.3.2 The NSG shall consist of a Supervisor and at least three staff 11/14/85 specialists.

The Supervisor shall have a Bachelor's Degree in Engineering or Physical Science and a minimum of 6 years of professional level managerial experience in the power field.

Each staff special le all have a Bachelor's Degree in Engineering or Physical Science and a minimum of 5 years of professional level experience in the field of his specialty.

The NSG shall use specialists from other technical organizations to augment its expertise in the disciplines of 6.5.3.1.

Such specialists shall meet the same qualification requirements as the NSG members.

CONSULTANTS 6.5.3.3 Consultants shall be utilized as determined by the NSG Supervisor

,to provide expert advice to the NSG.

RESPONSIBILITIES 6.5.3.4 The NSG shall review:

a. The safety evaluations for 1) -charges to procedures required by Specification 6.8, 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.

SAN ONOFRE - UNIT 1 6-14 Revised:

11/14/85

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 in 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 deviation from normal and expected performance of unit equipment that affect nuclear safety.

g. All REPORTABLE EVENTS.

h. All recognized indications of an unanticipated deficiency in 91 some aspect of design or operation of safety related structures, systems or components that could affect nuclear 11/14/85 safety.

1. Reports and meeting minutes of the Onsite Review Committee.

O AUDIT 6.5.3.5 Audits of unit activities shall be performed under the cognizance of the NSG. These audits shall encompass:

a. The conformance of unit operation to all provisions contained within the Technical Specifications and applicable license conditions at least once per 12 months.

b. The performance, training and qualifications of the entire unit staff at least once per 12 months.

c. The results of all actions taken to correct deficiencies occurring in unit equipment, structures, systems or method of operation that affect nuclear safety at least once per 6 months.

d. The peformance of all activities required by the Quality Assurance Program to meet the criteria of Appendix "B",

10 CFR 50, at least once per 24 months.

e. Any other area of unit operation considered appropria~e by the Nuclear Safety Group or the Vice President and Site Manager, Nuclear Generation Site.

SAN ONOFRE - UNIT 1 6-15 Revised:

11/14/85

f. The Fire Protection Program and implementing procedures at least once per 24 months.

g. An independent fire protection and loss prevention inspection and audit shall be performed annually utilizing either qualified offsite licensee personnel or an outside fire protection firm.

h. An inspection and audit of the fire protection and loss 91 prevention program shall be performed by a qualified outside 11/14/85 fire consultant at least once per 36 months.

AUTHORITY 6.5.3.6 The NSG shall report to and advise the Manager, Nuclear Safety on those areas of responsibility specified in Sections 6.5.3.4 and 6.5.3.5.

RECORDS 6.5.3.7 Records of NSG activities shall be prepared and maintained. Report of reviews and audits shall be distributed monthly to the Vice President and Site Manager, Nuclear Generation Site, and to the management responsible for the areas audited.

This page included FOR REFERENCE ONLY I

SAN ONOFRE - UNIT 1 6-16 Revised:

11/14/85

6.8 PROCEDURES AND PROGRAMS 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 implementation.

g. PROCESS CONTROL PROGRAM implementation.

h. OFFSITE DOSE CALCULATION MANUAL implementation.

I. Quality Assurance Program for effluent and environmental monitoring, using the guidance in Regulatory Guide 4.15, 91 Revision 1, February 1979.

11/14/85 J. "Nuclear Plant Fire Protection Functional Responsibilities, Administrative Controls and Quality Assurance," June 14, 1977, as specified in Section 6 of the Fire Protection Safety Evaluation Report dated July 19, 1979.

6.8.2 Each procedure of 6.8.1 above, and changes thereto, shall be approved by the Vice President and Site Manager, Nuclear Generation Site; or by (1) the Station Manager; or by (2) the Deputy Station Manager; or by (3) the Manager of Nuclear Generation Services; or by (4) Cognizant Managers reporting directly to them as previously designated by the Vice President and Site Manager, Nuclear Generation Site, prior to implementation and shall be reviewed periodically as set forth in administrative procedures.

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 site/station management staff exercising responsibility in the specific area and unit or units addressed by the change, and at least one of whom holds a Senior Reactor Operator's License on the unit affected.

Procedures and administrative policies shall meet or exceed the requirements and recommendations of Sections 5.1 and 5.3 of ANSI r18.7-1976, Administrative Controls for Nuclear Power Plants.

SAN ON0FRE -UNIT 1

6-19 Revised:

11/14/85

This page included FOR REFERENCE ONLY

c. The change is documented, reviewed and approved by responsible management, as delineated in 6.8.2 above, within 14 days of implementation.

6.8.4 The following programs shall be established, implemented, and maintained:

a. Primary Coolant Sources Outside Containment A program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels.

This program shall include the following:

(1) Preventative maintenance and periodic visual inspection requirements, and (ii) Integrated leak test requirements for each system at refueling cycle intervals or less.

b. In-Plant Radiation Monitoring A program which will ensure the capability to accurately determine 91 the airborne iodine concentration in vital areas under accident 11/14/85 conditions.

This program shall include the following:

(i) Training of personnel (ii) Procedures for monitoring, and (III) Provisions for maintenance of sampling and analysis equipment.

c. Backup Method for Determining Subcooling Margin A program which will ensure the capability to accurately monitor the Reactor Coolant System subcooling margin. This program shall include the following:

(1) Training of personnel, and (11)

Procedures for monitoring

d. Secondary Water Chemistry Monitoring Program A program for monitoring of secondary water chemistry to inhibit steam generator tube degradation. This program shall include:

(1) Identification of a sampling schedule for the critical parameters and control points for these parameters, SAN ONOFRE -

UNIT 1 6-20 Revised:

11/14/85

(ii) Identification of the procedures used to measure the values of the critical parameters, (iii) Identification of process sampling points, (iv) Procedures for the recording and management of data, (v) Procedures defining corrective actions for off control point chemistry conditions, and 91 (iv) A procedure identifying (a) the authority responsible for the 11/14/85 interpretation of the data, and (b) the sequence and timing of administrative events required to initiated corrective action.

e. Post-Accident Sampling A program which will ensure the capability to obtain and analyze reactor coolant, radioactive lodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. The program shall include:

(i) Training of personnel, (ii) Procedures for sampling and analysis, and (III) Provisions for maintenance of sampling and analysis equipment.

ThIs Page Included FOR REFERENCE ONLY OSAN ONOFRE -UNIT 1

6-21 Revised:

11/14/85

MONTHLY OPERATING REPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to pressurizer safety and relief valves, shall be submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory 91 Commission, Washington, D.C., 20555, with a copy to the Regional 11/14/85 Administrator of the Regional Office of the NRC, no later than the 15th of each month following the calendar month covered by the report.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the NRC Regional Administrator, unless otherwise indicated, within the time period specified for each report.

SAN ONOFRE - UNIT 1 6-26 Revised: 11/14/85

This page included 6.16 ENVIRONMENTAL PROTECTION FOR REFERENCE ONLY FACILITY DESIGN AND OPERATION 6.16.1 This section contains a description of facility design features and operating practices which, if changed, could have a significant effect on environmental impact. Any significant change in facility design features or operating practices described here must be reported to the NRC in accordance with the provisions of Section 6.16.2.a prior to the change.

a. Intake System The circulating water system, under normal operating conditions, draws water from the ocean at a point approximately 3,200 feet offshore. The ocean bottom at this point is approximately 27 feet below mean lower low-water level.

The intake structure rests on a foundation located 33 feet beneath the ocean bottom and rises vertically to a point 10 1/2 feet above the ocean floor. The inside horizontal dimensions of the intake structure are 16 to 21 feet. A velocity cap, 1-foot thick, rests on eight columns above 91 the top of the intake structure. The top surface of the velocity 11/14/85 cap is 15 1/2 feet above the ocean bottom and 11 1/2 feet below mean lower low-water.

A 12-foot ID reinforced concrete conduit is connected horizontally to the shoreward side of the intake structure. This conduit is buried beneath the ocean bottom, with a minimum of 4 feet of sand cover over its top and 4 feet of rock cover surrounding the intake structure. All sand cover was placed so as to approximate the local ocean bottom profile.

Water entering the top of the intake structure is accelerated to a design velocity of about 2.5 feet per second and directed into a 12-foot ID reinforced concrete conduit. As the water enters the concrete conduit from the structure, it is accelerated to a design velocity.of 6.9 feet per second.

The circulating water system is designed to deliver 350,000 gpm at this velocity.

The offshore system joins the onshore portion of the circulating water system at the screenwell. The screenwell is located just inside the seawall on the Station property.

Cooling water entering the onshore system passes, through a coarse bar screen, through finer traveling screens, and proceeds to two circulating water pumps designed to operate at 175,000 gpm. Water entering the screenwell structure is decelerated so that the approach velocity, at the screens is approximately 2.0 feet per second.

SAN ONOFRE - UNIT 1 6-35 Revised:

11/14/85

This page included FOR REFERENCE ONLY The circulating water system uses three methods of handling the marine growth and debris associated with the flow of seawater through the plant condensers. These are heat treating, bar and traveling screens, and chlorination.

Heat treatment is used for incrustation control.

This method consists of reversing the flow in the intake conduit and adjusting the temperature of the water to approximately 100OF and maintaining this temperature for approximately two hours once every five to six weeks and occasionally once every four weeks, and discharging through the intake conduit. This is accomplished by recirculating a portion of the condenser discharge back through the condenser. Cross-connections between intake and outfall conduits are provided to create the reversal of flow necessary for the treatment of the conduits. Normally only the intake conduit is treated. The water temperature in the outlet conduit can be raised for treatment when necessary. The sudden temperature increase of the cooling water causes incrustations growing in the circulating water system to expire, relax their hold, and be flushed out of the system.

91 Traveling and bar screens are provided to remove marine growth and 11/14/85 debris from the seawater passing through the screenwell.

The materials removed from the seawater are marine growth, shells, fish, driftwood, and other debris present in the ocean.

For chlorination, sufficient sodium hypochlorite is injected into the circulating water upstream of the circulating water pumps three times a day for each condenser half to eliminate slime-forming organisms on condenser internal surfaces.

The traveling screens and bar screens are placed in series, perpendicular to the flow. The screens are cleaned automatically, with the frequency of cleaning being dependent on the 'rate of material buildup on the screens. The bar screens are cleaned by a traveling mechanical rake that deposits accumulated debris, by means of a seawater jet spray washing process, into sluiceways for removal. The traveling screens are motor driven, and are capable of rotating as a unit in continuous sequence when activated by pressure differential due to trash buildup. The debris picked up by the traveling screens is also deposited in a sluiceway by means of a seawater jet spray.

b. Discharge System Under normal operating conditions, the heated cooling water leaves the condenser and is discharged to the ocean through a 12-foot ID 2,600-foot-long concrete conduit. A single point discharge is effected through a discharge structure located in 24 feet of water. The dimension of the structure is the same as the intake; however, there is no velocity cap. The top of the discharge structure is about 11.5 feet below mean lower low-water.

0 SAN ON0FRE -

UNIT 1 6-36 Revised:

11/14/85

A 12-foot ID reinforced concrete conduit is connected horizontally to the shoreward side of the discharge structure. This conduit is buried beneath the ocean bottom, with a minimum of 4 feet of sand cover over its top and 4 feet of rock surrounding the discharge structure. All sand and rock cover was placed so as to approximate the local ocean bottom profile.

The water travels through the discharge conduit with a design velocity of 6.9 feet per second and exits with a vertical velocity of about 2.5 feet per second. The vertical orientation creates a single orifice jet diffuser which entrains surroundng cooler water and assists in rapid diminution of the discharge temperature.

About seven minutes is required for water to travel from the condensers to the end of the discharge.

c. Land Management The facility occupies about 16 acres of the 84 acre site.

No use of herbicides is practiced to manage vegetation along the 91 transmission line except in isolated cases to meet property owners' 11/14/85 requests or permit stipulations from public agencies. Standard erosion control measures are used to minimize erosion at the facility, at tower sites, and along access roads.

REPORTS 6.16.2 The following reports shall be submitted pursuant to Specification 6.9.2, with a copy to the Director, Office of Nuclear Reactor Regulation.

a. A report shall be made to the NRC prior to implementation of a change in plant design, in plant operation, or in procedures described in Section 6.16.1 if the change would have a significant adverse effect on the environment or involves an environmental matter or question not previously reviewed and evaluated by the NRC.

The report shall include a description and evaluation of the change and a supporting benefit-cost analysis.

b. Unusual or Important Environmental Events Any occurrence of an unusual or important event that indicates or could result in significant environmental impact causally related to station operation shall be recorded and promptly reported to the NRC within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months followed by a written report within 30 days.

No routine monitoring programs are required to implement this condition.

SAN ONOFRE -

UNIT 1 6-37 Revised:

11/14/85

The written report shall (a) describe, analyze, and evaluate the event, including extent and magnitude of the impact and plant operating characteristics, (b) describe the probable cause of the event, (c) indicate the action taken to preclude repetition of the event and to prevent similar occurrences involving similar components or systems, and (e) indicate the agencies notified and their preliminary responses.

Events reportable under this subsection which also require reports to other Federal, State or local agencies shall be reported in accordance with those reporting requirements in lieu of the requirements of this subsection. The NRC shall be provided a copy of such report at the same time it is submitted to the other agency.

The following are examples of unusual or important events:

excessive bird impaction events; onsite plant or animal disease outbreaks; mortality or unusual occurrence of any species protected by the Endangered Species Act of 1973; unusual fish kills; increase in nuisance organisms or conditions; and unanticipated or emergency discharge of waste water or chemical substances.

c. Reporting Related to the NPDES Permits and State Certifications 91 11/14/85 Violations of the NPDES Permit or State certification (pursuant to Section 401 of the Clean Water Act) shall be reported to the NRC by submittal of copies of the reports required by the NPDES Permit or certification. The licensee shall also provide the NRC with a copy of the results of the following studies at the same time they are submitted to the permitting agency:

Section 316(b) Demonstration Study Changes and additions to the NPOES Permit or the State certification shall be reported to the NRC within 30 days following the date the change is approved. If a permit or certification, in part or in its entirety, is appealed and stayed, the NRC shall be notified within 30 days following the date the stay is granted.

The NRC shall be notified of changes to the effective NPDES Permit proposed by the licensee by providing NRC with a copy of the proposed change at the same time it is submitted to the permitting agency. The licensee shall provide the NRC a copy of the application for renewal of the NPOES Permit at the same time the application is submitted to the permitting agency.

This page included FOR REFERENCE ONLY SAN ONOFRE - UNIT 1 6-38 Revised:

11/14/85 3.5.8 RADIOACTIVE LIQUID EFFLUENT INSTRUMENTATION APPLICABILITY: During releases via this pathway.

OBJECTIVE: Monitor and control radioactive liquid effluent releases.

SPECIFICATION: A. The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.5.8.1 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of Specification 3.15.1 are not exceeded.

B. Action

1. With a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint less conservative than a value which will ensure that the limits of 3.15.1 are met, without delay suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

2. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.5.8.1.

If the inoperable instruments remain inoperable for greater than 30 days, explain in the next Semiannual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.

3. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

BASIS:

The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases.

The alarm/trip setpoints for these instruments are calculated in accordance with methods in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20.

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3.5.9 RADIOACTIVE GASEOUS PROCESS AND EFFLUENT MONITORING INSTRUMENTATION APPLICABILITY: During releases via this pathway.

OBJECTIVE: Monitor and control radioactive gaseous releases.

SPECIFICATION: A. The radioactive gaseous process and effluent monitoring instrumentation channels shown in Table 3.5.9.1 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of Specification 3.16.1 are not exceeded.

B. ACTION

1. With a radioactive gaseous process or effluent monitoring instrumentation channel alarm/trip setpoint less conservative than a value which will ensure that the limits of 3.16.1 are met, without delay suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

2. With less than the minimum number of radioactive gaseous process or effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.5.9.1. If the inoperable instruments remain inoperable for greater than 30 days, explain in the next Semiannual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.

3. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

BASIS:

The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive material in gaseous effluents during actual or potential releases. The alarm/trip setpoints for these instruments are calculated in accordance with methods in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20.

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3.15.2 LIQUID EFFLUENT DOSE APPLICABILITY: At all times.

OBJECTIVE:

Maintain the release of radioactive liquid effluents from the site as low as is reasonably achievable.

SPECIFICATION: A. The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released to UNRESTRICTED AREAS (see Figure 5.1-1) shall be limited:

1. During any calendar quarter to < 1.5 mrem to the total body and to < 5 mrem to any organ, and

2. During any calendar year to < 3 mrem to the total body and to < 10 mrem to any organ.

B. Action:

1. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

2. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

BASIS:

This specification is provided to implement the requirements of Section II.A and IV.A of Appendix I, 10 CFR Part 50.

Specification A implements the guides set forth in Section II.A of Appendix I. Specification B provides the required operating flexibility and at the same time implements the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable."

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3.15.3 LIOUID WASTE TREATMENT APPLICABILITY: At all times.

OBJECTIVE:

Maintain radioactive releases from the site as low as is reasonable achievable by use of the liquid radwaste treatment system.

SPECIFICATION: A. The liquid radwaste treatment system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected dose due to the liquid effluent from San Onofre Unit 1, to UNRESTRICTED AREAS (see Figure 5.1-1) would exceed 0.06 mrem to the total body or 0.2 mrem to any organ in a 31 day period.

B. Action:

1. With radioactive liquid waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that includes the following information:

a. Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems and the reason for inoperability.

b. Action(s) taken to restore the inoperable equipment to OPERABLE status.

c. Summary description of action(s) taken to prevent a recurrence.

2. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

BASIS:

The OPERABILITY of the liquid radwaste treatment system ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirements that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonable achievable." This specification implements the requirements of 10 CFR Part 50.36a and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the guide set forth in Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents.

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3.16.2 DOSE, NOBLE GASES APPLICABILITY: At all times.

OBJECTIVE:

Maintain the dose due to noble gases in gaseous effluents as low as is reasonable achievable.

SPECIFICATION: A. The air dose due to noble gases released in gaseous effluents, from San Onofre Unit 1 to areas at and beyond the SITE BOUNDARY (see Figure 5.1-1) shall be limited to the following:

1. During any calendar quarter: < 5 mrad for gamma radiation and < 10 mrad for beta radiation.

2. During any calendar year: < 10 mrad for gamma radiation and < 20 mrad for beta radiation.

B. Action:

1. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

2. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

BASIS:

This specification is provided to implement the requirements of Section II.B and IV.A of Appendix I, 10 CFR Part 50.

Specification A implements the guides set forth in Section II.B of Appendix I. Specification B provides the required operating flexibility and at the same time implements the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonable achievable."

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3.16.3 DOSE, IODINE-131, IODINE-133. TRITIUM AND RADIONUCLIDES IN PARTICULATE FORM APPLICABILITY: At all times.

OBJECTIVE:

Maintain the dose due to radioiodines, radioactive materials in particulate form and radionuclides other than noble gases in gaseous effluents as low as is reasonable achievable.

SPECIFICATION: A. The dose to a MEMBER OF THE PUBLIC from 1-131, 1-133, from tritium and from all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from San Onofre Unit 1 to areas at and beyond the SITE BOUNDARY (see Figure 5.1-1) shall be limited to the following:

1. During any calendar quarter: < 7.5 mrem to any organ; and

2. During any calendar year: < 15 mrem to any organ.

B. Action:

1. With the calculated dose from the release of 1-131, 1-133, tritium and radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

2. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

BASIS:

This specification is provided to implement the requirements of Sections II.C and IV.A of Appendix I, 10 CFR Part 50.

Specification A is the guide set forth in Section II.C of Appendix I. Specification B provides the required operating flexibility and at the same time implements the guides set forth in Section IV.a of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonable achievable."

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3.16.4 GASEOUS RADWASTE TREATMENT APPLICABILITY: At all times.

OBJECTIVE:

Maintain radioactive gaseous releases from the site as low as is reasonable achievable by use of the GASEOUS RADWASTE and VENTILATION EXHAUST TREATMENT SYSTEMS.

SPECIFICATION: A. The GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior,to their discharge when the projected gaseous effluent air doses due to gaseous effluent releases from San Onofre Unit I to areas at and beyond the SITE BOUNDARY (see Figure 5.1-1) would exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiation over 31 days. The VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases from San Onofre Unit 1 to areas at and beyond the SITE BOUNDARY (see Figure 5.1-1) would exceed 0.3 mrem to any organ over 31 days.

B. Action:

1. With gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which includes the following information:

a. Explanation of why gaseous radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems and the reasons for the inoperability.

b. Action(s) taken to restore the inoperable equipment to OPERABLE status.

c. Summary description of action(s) taken to prevent a recurrence.

2. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

BASIS:

The OPERABILITY of the GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensurers that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used when specified provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable."

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3.17 DOSE APPLICABILITY: At all times.

OBJECTIVE:

Maintain the dose due to the release of radioactive materials within specified limits.

SPECIFICATION: A. The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC, due to releases of radioactivity and to radiation, from uranium fuel cycle sources shall be limited to < 25 mrem to the total body or any organ (except the thyroid,which shall be limited to < 75 mrem).

B. Action:

1. With the calculated dose from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specifications 3.15.2.A, 3.16.2.A or 3.16.3.A, calculations should be made to determine whether the above limits of Specification 3.17 have been exceeded. If such is the case, prepare and submit to the Commission within 30 days pursuant to Specification 6.9.2, a Special Report that defines the corrective action to be taken to reduce subsequent releases, to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits.

The Special Report, as defined in 10 CFR Part 20.405c, shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release(s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations.

If the estimated dose(s) exceeds the above limits, and if the release condition resulting in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190.

Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

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3.18 RADIOLOGICAL ENVIRONMENTAL MONITORING 3.18.1 MONITORING PROGRAM APPLICABILITY: At all times.

OBJECTIVE:

Monitor exposure pathways for radiation and radioactive material.

SPECIFICATION: A. The radiological environmental monitoring program shall be conducted as specified in Table 3.18.1.

B. Action:

1. With the radiological environmental monitoring program not being conducted as specified in Table 3.18.1, prepare and submit to the Commission, in the Annual Radiological Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.

2. With the level of radioactivity as the result of plant effluents in an environmental sampling medium exceeding the reporting levels of Table 3.18.2 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days from the end of the affected calendar quarter a Special Report pursuant to Specification 6.9.2. When more than one of the radionuclides in Table 3.18.2 are detected in the sampling medium, this report shall be submitted if:

concentration (1)

+ concentration (2)

+... > 1.0 reporting level (1) reporting level (2)

When radionuclides other than those in Table 3.18.2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to a MEMBER OF THE PUBLIC is equal to or greater than the calendar year limits of Specifications 3.15.2, 3.16.2 and 3.16.3. This report is not required 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 Operating Report.

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3.18.2 LAND USE CENSUS APPLICABILITY: At all times.

OBJECTIVE:

Monitor the UNRESTRICTED AREAS surrounding the site for potential changes to the radiological monitoring program as necessary.

SPECIFICATION: A. 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.

B. Action:

1. With the land use census identifying a location(s) which yields a calculated dose or dose commitment greater than the values currently being calculated in Specification 4.6.3, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the new locations. Identify the new locations in the next Semiannual Radioactive Effluent Release Report.

2. With a land use census identifying a location(s) which yields a calculated dose or dose commitment via the same exposure pathway 20 percent greater than at a location from which samples are currently being obtained in accordance with Specification 3.18.1, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the new locations. The new location shall be added to the radiological environmental monitoring program within 30 days. The sampling location, excluding the control station location, having the lowest calculated dose or 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.

Broad leaf vegetation sampling may be performed at the SITE BOUNDARY in the direction section with the highest D/Q in lieu of the garden census.

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3.18.3 INTERLABORATORY COMPARISON PROGRAM APPLICABILITY: At all times.

OBJECTIVE:

To ensure laboratory analysis of radiological environmental monitoring samples is correct and accurate.

SPECIFICATION: A. Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program which has been approved by the Commission.

B. Action:

1. 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 Operating Report.

2. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

BASIS:

The requirement for participation in an Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid.

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3.19 SOLID RADIOACTIVE WASTE APPLICABILITY: At all times.

OBJECTIVE:

Ensure meeting the requirements for the SOLIDIFICATION and shipment of solid radwaste.

SPECIFICATION: A. The solid radwaste system shall be used in accordance with a PROCESS CONTROL PROGRAM to process wet radioactive wastes to meet shipping burial ground requirements.

B. Action:

1. With the provisions of the PROCESS CONTROL PROGRAM not satisfied, suspend shipments of defectively processed or defectively packaged solid radioactive wastes from the site.

2. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

BASIS:

This specification implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to waste type, waste pH, waste/liquid/

solidification/agent/catalyst ratios, waste oil content, waste principal chemical.constituents, mixing and curing times.

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TL =

time (in days) since the last surveillance testing F =

the actuator force from the previous surveillance test (lbf)*

If the calculated value of T does not exceed 92 days, the next surveillance test must be performed before T days had elapsed.

c. If the measured actuator force of either HV-851 A or B is greater than 22,000 lbf, the valve(s) shall be declared inoperable. The test results shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2 along with proposed corrective actions and NRC approval obtained prior to returning the unit to service.

2.

The first test shall be performed not less than 14 days nor more than 21 days following return to power from the current outage which began September 3, 1981.

For the first surveillance test, the value of F shall be the average actuator force of HV-851 A&B valves from pre-operation testing (3135 lbf).

All subsequent surveillance testing shall assume the F2 value from the previous surveillance test for each valve. If an F2 was not required during the previous surveillance test, the F1 value for each valve shall be assumed.

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3. Snubber release rate, where required, is within the specified range in compression or tension. For snubbers specifically required not to displace under continuous load, the ability of the snubber to withstand load without displacement shall be verified.

F. Snubber Service Life Monitoring A record of the service life of each snubber, the date at which the designated service life commences and the installation and maintenance records on which the designated service is based shall be maintained as required by Specification 6.10.2. Concurrent with the first in-service visual inspection and at least once per refueling cycle shutdown thereafter, the installation and maintenance records for each snubber shall be reviewed to verify that the indicated service life has not been exceeded or will not be exceeded prior to the next scheduled snubber service life review. If the indicated service life will be exceeded prior to the next scheduled snubber service life review, the snubber service life shall be reevaluated or the snubber shall be replaced or reconditioned so as to extend its service life beyond the date of the next scheduled service life review. This reevaluation, replacement or reconditioning shall be indicated in the records.

BASIS:

Refer to the basis given in 3.13.

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2. If, in the inspections performed under Specification A,

a. Less than 10% of the total tubes inspected have imperfections greater than 20% that were not detected during the previous inspections or have previously detected imperfections that have increased more than 10% wall penetration, and

b. No tube inspected exceeds the plugging limit, plant operation may resume.

3. If, in the inspections performed under Specification B,

a. Less than 10% of the total tubes inspected have imperfections greater than 20% that were not detected during the previous inspections or have previously detected imperfections that have increased more than 10% wall penetration, and

b. No more than 3 of the tubes inspected exceed the plugging limit, plant operation may resume after performance of the corrective action in Specification F.

4. If, in the inspections performed under Specification B,

a. More than 10% of the tubes inspected have imperfections greater than 20% that were not detected during the previous inspections or have previously detected imperfections that have increased more than 10% wall penetration, or

b. More than 3 of the tubes inspected exceed the plugging limit, prior to resumption of plant operation the situation shall be reported in a Special Report to the Commission in accordance with Technical Specification 6.9.2 for approval of the proposed remedial action.

5. If in the inspections performed under Specification C.1, wear rates are observed at AVB intersections which are inconsistent with.the 50% plugging criterion, prior to resumption of plant operation the situation shall be reported in a Special Report to the Commission in accordance with Technical Specification 6.9.2 for approvalof the proposed remedial action.

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6. If in the inspections performed under Specification C.2 progression of the denting process is observed to be recurring, prior to resumption of plant operation the situation shall be reported in a Special Report to the Commission in accordance with Technical Specification 6.9.2 for approval of the proposed remedial action.

F. CORRECTIVE ACTION All leaking tubes, defective tubes, and tubes with imperfections exceeding the plugging limit shall be repaired or plugged.

BASIS:

The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the Reactor Coolant System will be maintained. The program for inservice inspection of steam generator tubes is based on Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes.

If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking.

The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage =.3 gallons per minute per steam generator). Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents.

Operating plants have demonstrated that primary-to-secondary leakage of.3 gpm per steam generator can readily be detected by radiation monitors of steam generator blowdown. Leakage in excess of this limit will require shutdown during which the leaking tubes will be located and plugged and additional inspections performed.

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4.19 SOLID RADIOACTIVE WASTE APPLICABILITY: At all times.

OBJECTIVE:

Ensure meeting the requirements for the SOLIDIFICATION and shipment of solid radwaste.

SPECIFICATION: The PROCESS CONTROL PROGRAM shall be used to verify the SOLIDIFICATION of at least one representative test specimen from at least every tenth batch of each type of wet radioactive waste (e.g., filter sludges, spent resins, evaporator bottoms, boric acid solutions, and sodium sulfate solutions).

ACTION:

a. If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICATION of the batch under test shall be suspended until such time as additional test specimens can be obtained, alternative SOLIDIFICATION parameters can be determined in accordance with the PROCESS CONTROL PROGRAM, and a subsequent test verifies SOLIDIFICATION.

SOLIDIFICATION of the batch may then be resumed using the alternative SOLIDIFICATION parameters determined by the PROCESS CONTROL PROGRAM.

b. If the initial test specimen from a batch of waste fails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection and testing of representative test specimens from each consecutive batch of the same type of wet waste until at least three conse cutive initial test specimens demonstrate SOLIDIFICATION. The PROCESS CONTROL PROGRAM shall be modified as required, as provided in Specification 6.13, to assure SOLIDIFICATION of subsequent batches of waste.

BASIS:

This specification implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to waste type, waste pH, waste/liquid/solidifica tion agent/catalyst ratios, waste oil content, waste principal chemical constituents, mixing and curing times.

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6.2 ORGANIZATION OFFSITE 6.2.1 The offsite organization for unit management and technical support shall be as shown in Chapter 13 of the San Onofre Units 2 and 3 Final Safety Analysis Report.

UNIT STAFF 6.2.2 The Site organization shall be as shown in Chapter 13 of the San Onofre Units 2 and 3 Final'Safety Analysis Report.

a. Each on-duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1.

b. At least one licensed Operator shall be at the controls when fuel is in the reactor.*

During refueling operations, this operator is permitted to step outside the red line to update the refueling status board. In addition, while the unit is in MODE 1, 2, 3 or 4, at least one licensed Senior Reactor Operator shall be in the Control Room Area.**

c. A health physics technician# shall be on-site when fuel is in the reactor.

d. All CORE ALTERATIONS shall be observed and directly supervised by a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent duties during this operation.

e. A Fire Brigade of at least five members shall be maintained on site at all times.#

The Fire Brigade shall not include the Shift Superintendent and the two other members of the minimum shift crew necessary for safe shutdown of the unit and any personnel required for other essential functions during a fire emergency.

"At the controls" means within the area bounded by the three vertical instrumentation boards and the red line on the floor of the control room.

- "Control Room Area" is defined by the control room and the Shift Superintendent's office.

The health physics technician and Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence provided immediate action is taken to fill the required positions.

SAN ONOFRE - UNIT 1 6-2

1) An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight (shift turnover and meal time are not included when calculating hours worked).

2) An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 24-hour period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48-hour period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any 7-day period (shift turnover and meal time are not included when calculating hours worked).

3) A break (the time an individual leaves the work location to the time an individual returns to the work location) of at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months should be allowed between work periods (shift turnover time is not included when calculating the break; meal time is not included when calculating the break, unless it represents an administrative entry on the timesheet and not extra hours spent at the work location).

4) Except during extended shutdown periods, overtime should be considered on an individual basis and not for the entire staff on a shift.

Any deviation from the above guidelines shall be authorized by the Station Manager, or by the Responsible Station Division Manager, or higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation.

Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the Station Manager or his designee to assure that excessive hours have not been assigned.

Routine deviation from the above guidelines is not authorized.

SAN ONOFRE - UNIT 1 6-4

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 in 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 deviation from normal and expected performance of unit equipment that affect nuclear safety.

g. All REPORTABLE EVENTS.

h. All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety related structures, systems or components that could affect nuclear safety.

i. Reports and meeting minutes of the Onsite Review Committee.

AUDIT 6.5.3.5 Audits of unit activities shall be performed under the cognizance of the NSG. These audits shall encompass:

a. The conformance of unit operation to the provisions contained within the Technical Specifications and applicable license conditions at least once per 12 months.

b. The performance, training and qualifications of the entire unit staff at least once per 12 months.

c. The results of the actions taken to correct deficiencies occurring in unit equipment, structures, systems or method of operation that affect nuclear safety at least once per 6 months.

d. The performance of the activities required by the Quality Assurance Program to meet the criteria of Appendix "B",

10 CFR 50, at least once per 24 months.

e. Any other area of unit operation considered appropriate by the Nuclear Safety Group or the Vice President and Site Manager, Nuclear Generation Site.

SAN ONOFRE -

UNIT 1 6-15

6.8 PROCEDURES AND PROGRAMS 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 implementation.

g. PROCESS CONTROL PROGRAM implementation.

h. OFFSITE DOSE CALCULATION MANUAL implementation.

i.

Quality Assurance Program for effluent and environmental monitoring, using the guidance in Regulatory Guide 4.15, Revision 1, February 1979.

j. "Nuclear Plant Fire Protection Functional Responsibilities, Administrative Controls and Quality Assurance," June 14, 1977, as specified in Section 6 of the Fire Protection Safety Evaluation Report dated July 19, 1979.

6.8.2 Each procedure of 6.8.1 above, and changes thereto, shall be approved by the Vice President and Site Manager, Nuclear Generation Site; or by (1) the Station Manager; or by (2) the Responsible Station Division Manager; or by (3) Congnizant Managers reporting directly to them as previously designated by the Vice President and Site Manager, Nuclear Generation Site, prior to implementation and shall be reviewed periodically as set forth in administrative procedures.

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 site/station management staff exercising responsibility in the specific area and unit or units addressed by the change, and at least one of whom holds a Senior Reactor Operator's License on the unit affected.

Procedures and administrative policies shall meet or exceed the requirements and recommendations of Sections 5.1 and 5.3 of ANSI N18.7-1976, Admisistrative Controls for Nuclear Power Plants.

SAN ONOFRE - UNIT 1 6-19

MONTHLY OPERATING REPORT O

6.9.1.10 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to pressurizer safety and relief valves, shall be submitted to the Nuclear Regulatory Commission on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Nuclear Regulatory Commission within the time period specified for each report.

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UNIT 1 6-26

A 12-foot ID reinforced concrete conduit is connected horizontally to the shoreward side of the discharge structure. This conduit is buried beneath the ocean bottom, with a minimum of 4 feet of sand cover over its top and 4 feet of rock surrounding the discharge structure. All sand and rock cover was placed so as to approximate the local ocean bottom profile.

The water travels through the discharge conduit with a design velocity of 6.9 feet per second and exits with a vertical velocity of about 2.5 feet per second. The vertical orientation creates a single orifice jet diffuser which entrains surrounding cooler water and assists in rapid diminution of the discharge temperature.

About seven minutes is required for water to travel from the condensers to the end of the discharge.

c. Land Management The facility occupies about 16 acres of the 84 acre site.

No use of herbicides is practiced to manage vegetation along the transmission line except in isolated cases to meet property owners' requests or permit stipulations.from public agencies. Standard erosion control measures are used to minimize erosion at the facility, at tower sites, and along access roads.

REPORTS 6.16.2 The following reports shall be submitted pursuant to Specification 6.9.2.

a. A report shall be made to the NRC prior to implementation of a change in plant design, in plant operation, or in procedures described in Section 6.16.1 if the change would have a significant adverse effect on the environment or involves an environmental matter or question not previously reviewed and evaluated by the NRC. The report shall include a description and evaluation of the change and a supporting benefit-cost analysis.

b. Unusual or Important Environmental Events Any occurrence of an unusual or important event that indicates or could result in significant environmental impact causally related to station operation shall be recorded and promptly reported to the NRC within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months followed by a written report within 30 days.

No routine monitoring programs are required to implement this condition.

SAN ONOFRE -

UNIT 1 6-37

v t e Letter Sequence Other
TAC:66164, (Approved, Closed)
Initiation Request, Request Acceptance... Administration Questions Answers Results Approval, Approval, Approval Other: ML13316B055, ML13316B827, ML13331A252
MONTHYEAR1988-04-01 [Table View]

ML13316B827

Text

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