ML20140E023

Amends 52 & 33 to Licenses NPF-9 & NPF-17,respectively. Amends Change Tech Specs Re Administrative Controls & Reportability Requirements of 10CFR50.72 & 50.73

ML20140E023
Person / Time
Site:	McGuire, Mcguire
Issue date:	03/19/1986
From:	Youngblood B Office of Nuclear Reactor Regulation
To:
Shared Package
ML20140E030	List: ML20140E023 ML20140E042
References
NUDOCS 8603270144
Download: ML20140E023 (39)

Text

.

n rag g'o UNITED STATES v8"

~g NUCLEAR REGULATORY COMMISSION n

E WASHINGTON, D. C. 20555 E

,l DUKE POWER COMPANY DOCKET N0. 50-369 McGUIRE NUCLEAR STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE.

Amendiaent No. 52 License No. NPF-9 1.

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

A.

The application for amendment to the McGuire Nuclear Station.. Unit 1 (the facility) Facility Operating License No. NPF-9 filed by the Duke Power Company.(licensee) dated April 25, 1985, and supplemented November 13 and 26, 1985, and January 28, 1986 complies with the standards and requirements of the Atomic Energy Act of 1954,.as amended (tne Act) and the Comission's regulations as set forth in 10 CFR Ch3pter I; B.

The facility will operate in confonnity with the application, as amended, the provisions of the Act, and the regulations of the Comission; C.

There is reasonable assurance (i) that the activities ~ authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this license amendment will'not be inimical to the comon defense and security or to the health and safety of the public; E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulatio'ns and all applicable requirements have been satisfied.

2.

Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachments to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-9 is~ hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications' contained in Appendix A, as revised through Amendment No.52, are hereby incorporated into the license.-

h3270144860319 ADOCK 05000369 p

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.The licensee shall operate the facility in accordance with the Technical Specif.ications and the Environmental Protection Plan.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION p - //, p

,j6t, B. J. Youngblood, Director PWR Project Directorate #4 Division of PWR Licensing-A

Attachment:

Appendix A Technical Specification Changes

'I Date of Issuance: March 19, 1986 i

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'o UNITED STATES NUCLEAR REGULATORY COMMISSION o

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WASHINGTON. D. C. 20555 E

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DUKE POWER COMPANY D0CKET NO. 50-370 McGUIRE NUCLEAR STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 33 License No. NPF-17 1.

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

A.

The application for amendment to the McGuire Nuclear Station, Unit 2 (the facility) Facility Operating License No. NPF-17 filed by the Duke Power Company (licensee) dated April 25, 1985, and supplemented November 13 and 26, 1985, and January 28, 1986, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended.

(the Act) and the Comission's regulations as set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, as amended, the provisions of the Act, and the regulations ~of the Commission; C.-

There-is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will-not be inimical to the common defense and security or to the h'ealth and safety of the public; E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the-attachments to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-17 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.33, are hereby incorporated into the license.

o

, v The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This. license amendment is effective.as of its-date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Of RL 4 8. J. Youngblood, Director PWR Project Directorate #4 Division ~of PWR Licensing-A

Attachment:

Appendix A Technical-Specification Changes Date of Issuance: March 19, 1986 i

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I ATTACHMENT TO LICENSE AMENDMENT NO. 52 e

FACILITY OPERATING LICENSE NO. NPF-9 DOCKET NO. 50-369 AND TO LICENSE ~ AMENDMENT NO. 33 FACILITY OPERATING LICENSE NO. NPF-17 DOCKET NO. 50-370 Replace the-following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change. The corresponding over-leaf pages are also provided to maintain document completeness.

Amended Overleaf Page Page XXI 6-1 6-3 6-4 6-8 6-7 6-9 6-13 6-14 6-15 6-20 6-23 6-24 6-27 6-28 3/4 3-66 3/4 3-65 3/4 3-71 3/4 3-72 3/4 4-15 3/4 4-16 i

3/4 4-17 3/4 4-18 3/4 4-26 3/4 6-14 3/4 6-15 3/4 6-16 3/4 8-7

.3/4 8-8 1

3/4 11-1 3/4 11-2 3/4 11-9 3/4 11-10 8 3/4 4-4 8 3/4 4-3 m..

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INDEX j

ADMINISTRATIVE CONTROLS-4 1

SECTION IPAGE 6.1 RESPONSIBILITY.................................................

6-1 6.2 ORGANIZATION 6.2.1 0FFSITE......................................................

6-1 6.2.2 UNIT STAFF...................................................

6-1 i

FIGURE 6.2-1 (Deleted)............................................

6-3 1

FIGURE 6.2-2 (Deleted)............................................

6-4 TABLE 6.2-1 MINIMUM. SHIFT CREW COMP 0SITION.......................

6-5 6.2.3' STATION SAFETY REVIEW GROUP (SSRG)

Function..................................................

6-7 r

i

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Composition...............................................

6-7 i

Responsibilities..........................................

6-7 i

Authority.................................................

6-7 Records...................................................

6-7 k

1 i

i 6.2.4 SHIFT TECHNICAL ADVIS0R......................................

6-7 i

6.3 UNIT STAFF QUALIFICATIONS......................................

6-7

.l.

4 6.4 TRAINING.......................................................

6-7 1

I 6.5 REVIEW AND AUDIT 6.5.1 TECHNICAL REVIEW AND CONTROL Activities................................................

6-8 4

1 i

1 i

l McGUIRE - UNITS 1 and 2 XXI Amendment No. 52 (Unit 1)

Amendment No'. 33 (Unit 2) i l

1 i

6.0 ADMINTSTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Station Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

6.1.2 The Shift Supervisor (or during his absence from the control room, a designated individual) shall be responsible for the control room command function. A management directive to this effect, signed by the Vice-President Nuclear Production shall be reissued to all Nuclear Production Department station personnel on an annual basis.

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

UNIT STAFF 6.2.2 The unit organization shall be as shown in 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 for each unit shall be in the control room when fuel is-in either reactor.

In addition, while either unit is in MODE 1, 2, 3, or 4, at least one licensed Senior Operator shall be in the control room; A Health Physics Technician # shall be on site when fuel is in either c.

reactor; d.

All CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Operator or licensed Senior Operator Limited to Fuel Handling who has no other concurrent responsibil-ities during this operation; A site Fire Brigade gf at least five members shall be maintained e.

onsite at all times.

The Fire Brigade shall not include three 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; and

1. 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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence, provided immediate action is taken to fill the required positions.

McGUIRE - UNITS 1 and 2 6-1 Amendment No.52(Unit 1)

Amendment No.33(Unit 2)

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i McGUIRE - UNITS I and-2 6-3 Amendment No.52(Unit 1)

Amendment No.33(Unit 2)

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McGUIRE - UNITS 1 and 2 6-4 Amendment' No. SXUnit 1) i Amendment No. 3XUnit 2) i i

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ADMINISTRATIVE CONTROLS 9.

6.2.3 STATION SAFETY REVIEW GROUP FUNCTION 6.2.3.1 The Station Safety Review Group (SSRG) shall function to examine plant operating characteristics, NRC issuances, industry advisories, Licensee Event Reports and other sources which may indicate areas for improving plant safety.

COMPOSITION 6.2.3.2 The SSRG shall be composed of at least five dedicated, full-time engineers located onsite.

RESPONSIBILITIES 6.2.3.3 The SSRG shall be responsible for maintaining surveillance of plant activities to provide independent verification

• that these activities are performed correctly and that human errors are reduced as much as practical.

AUTHORITY 6.2.3.4 The SSRG shall make detailed recommendations for revised procedures, equipment modifications, or other means of improving plant safety to the l

Director, Nuclear Safety Review Board.

RECORDS 6.2.3.5 Records of activities performed by the SSRG shall txt prepared, mainta'ned, and forwarded each calendar month to the Director, Nuclear Safety Review Boar'd.

6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shift Technical Advisor shall serve in.an advisory capacity to the Shift Supervisor.

6.3 UNIT STAFF QUALIFICATIONS 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifica-tions of ANSI N18.1-1971 for comparable positions and the supplemental require-ments specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, except for the Radiation Protection Manager who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975.

6.4 TRAINING 6.4.1 A retraining and replacement training program for the unit staff shall be maintained under the direction of the Station Manager and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix A of 10 CFR Part 55 and the supplemental requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, and shall include familiarization with relevant industry operational experience identified by the SSRG.

A Not responsible for sign-off function.

McGUIRE - UNITS I and 2 6-7

ADMINISTRATIVE CONTROLS 6.5 REVIEW AND AUDIT 6.5.1 TECHNICAL REVIEW AND CONTROL ACTIVITIES 6.5.1.1 Each procedure and program required by Specification 6.8 and other procedures which affect nuclear safety, and changes thereto, shall be prepared by a qualified individual / organization.

Each such procedure, and changes thereto, shall be reviewed by an individual / group other than the individual /

group which prepared the procedure, or changes thereto, but who may be f rom the same organization as the individual / group which prepared the procedure, nr changes thereto.

6.5.1.2 Proposed changes to the Appendix A Technical Specifications shall be prepared by a qualified individual / organization.

The preparation of each proposed Technical Specifications change shall be reviewed by an individual /

group other than the individual / group which prepared the proposed change, but who may be from the same organization as the individual / group which prepared the proposed change.

Proposed changes to.the Technical Specifications shall be approved by the Station Manager.

6.5.1.3 Proposed modifications to unit nuclear safety-related structures, systems and components shall be designed by a qualified individual / organization.

Each such modification shall be reviewed by an individual / group other than the individual / group which designed the modification, but who may be from the same organization as the individual / group which designed the modification.

Proposed modifications to nuclear safety related structures, systems, and components shall be approved prior to implementation by the Station Manager; or by the Operating Superintendent, the Technical Services Superintendent, the Superintendent of Integrated Scheduling, or the Maintenance Superintendent, l

as previously designated by the Station Manager.

6.5.1.4 Individuals responsible for reviews performed in accordance with Specifications _ 6.5.1.1, 6.5.1.2, and 6.5.1.3 shall be members of the station supervisory staff, previously designated by the Station Manager'to perform such reviews.

Each such review shall. include a determination of whether or not additional, cross-disciplinary, review is necessary.

If deemed necessary, such review shall be performed by the appropriate designated station review personnel.

6.5.1.5 Proposed tests and experiments which affect station nuclear safety and are not addressed in the FSAR or Technical _ Specifications shall be i

reviewed by the Station Manager; or by the Operating Superintendent, the Technical Services Superintendent the Maintenance Superintendent, or the Superintendent of Integrated Scheduling as previously designated by the Station Manager.

McGUIRE - UNITS 1 and 2 6-8 Amendment No.52 (Unit 1)

Amendment No.33 (Unit 2) m

ADMINISTRATIVE CONTROLS ACTIVITIES (Continued)

6. 5.1. 6 ALL REPORTABLE EVENTS and all violations of Technical Specifications shall be investigated and a report prepared which evaluates the occurrence and which provides recommendations to prevent recurrence.

Such reports shall be approved by the Station Manager and transmitted to the Vice President, Nuclear Production, and to the Director of the Nuclear Safety Review Board.-

6. 5.1. 7 The Station Manager shall assure the performance of special reviews and investigations, and the preparation and submittal of reports thereon, as requested by the Vice President, Nuclear Production.

6. 5.1. 8 The station security program, and implementing procedures, shall be reviewed at least once per 12 months.

Recommended changes shall be approved by the Station Manager or Superintendent of Station Services and transmitted to the Vice President, Nuclear Production, and to the Director of the Nuclear Safety Review Board.

6.5.1.9 The station emergency plan, and implementing procedures, shall.be reviewed at least once per 12 months.

Recommended changes shall be approved l

by the Station Manager and transmitted to the Vice President, Nuclear Produc-tion, and to the Director of the Nuclear Safety Review Board.

6.5.1.10 The Station Manager shall assure the performance of a review by a qualified individu.11/ organization-of every unplanned onsite release of radio-active material to the environs including the preparation and forwarding of reports covering evaluation, recommendations, and disposition of the corrective ACTION to prevent recurrence to the Vice President, Nuclear Production and to the Nuclear Safety Review Board.

6.5.1.11 The Station Manager shall assure the performance of a review by a qualified individual / organization of changes to the PROCESS CONTROL PROGRAM,,

OFFSITE DOSE CALCULATION MANUAL, and Radwaste Treatment Systems.

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6.5.1.12 Reports documenting each of the activities performed under Specifi-cations 6.5.1.1 through 6.5.1.11 shall be maintained.

Copies shall be provided to the Vice President, Nuclear Production, and the Nuclear Safety Review Board.

6.5.2 NUCLEAR SAFETY REVIEW BOARD (NSRB)

FUNCTION 6.5.2.1 The NSRB 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, McGUIRE - UNITS 1 and 2 6-9 Amendment No. 52(Unit 1)

Amendment No. 33(Unit 2)

ADMINISTRATIVE CONTROLS-RECORDS 6.5.2.11 Records of NSRB activities shall be prepared, approved, and distributed as indicated below:

a.

Minutes of each NSRB meeting shall be prepared,' approved, and forwarded to the Vice President, Nuclear Production, and to the Executive Vice President, Engineering, Construction, and Production, within 14 days following each meeting; b.

Reports of reviews encompassed by Specification 6.5.2.8 above, shall be prepared, approved and forwarded to the Vice President, Nuclear Production, and to the Executive Vice President, Engineering, Construction, and Production, within 14-days following completion of the review; and c.

Audit reports encompassed by Specification 6.5.2.9 above, shall be forwarded to the Vice President, Nuclear Production, and to the Executive Vice President, Engineering, Construction, and Production, and to the management positions responsible for the areas audited within 30 days after completion of the audit by the auditing organization.

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

~

a.

The Commission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and b.

Each REPORTABLE EVENT shall be reviewed by the Station Manager; or by:

(1) the Operating Superintendent, (2) the Technical Services Superintendent, (3) the Maintenance Superintendent, or (4) the Superintendent of Integrated Scheduling, as previously designated.by the Station Manager, and the results of the review shall be submitted to the NSRB and the Vice President, Nuclear Production.

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

a.

The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

The Vice President, Nuclear Production, and the NSRB shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; b.

A Safety Limit Violation Report ~shall be prepared.

The report shall be reviewed by the Operating Superintendent and the Station Manager.

This report shall describe:

(1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems, or structures, and (3) corrective action taken to prevent 1

recurrence; McGUIRE - UNITS 1 and 2 6-13

. Amendment No.52(Unit 1)

Amendment No. 33(Unit 2) i

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ADMINISTRATIVE CONTROLS SAFETY LIMIT VIOLATION (Continued) c.

The Safety Limit Violation Report shall be submitted to the Commission, the NSRB and the Vice President, Nuclear Production, within 14 days of the. violation; and d.

Critical operation of the' unit shall not-be resumed until authorized by the Commission.

6.8 PROCEDURES A'ND 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.

The applicable procedures required to implement the requirements of NUREG-0737; c.

Security Plan implementation;*

d.

Emergency Plan implementation; e.

PROCESS CONTROL PROGRAM implementation; f.

OFFSITE DOSE CALCULATION MANUAL implementation; and g.

Quality Ass'.:rance Program for effluent and environmental monitoring.

6.8.2 Each procedure of Specification 6.8.1 above, and changes thereto, shall be reviewed and approved by the Station Manager;'or by:

(1) the Operating Superintendent,_(2) the Technical Services Superintendent, (3) the Maintenance Superintendent, or (4) the Superintendent of Integrated Scheduling as previously l designated by the Station Manager; prior to implementation and shall be reviewed periodically as set forth in administrative procedures.

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

a.

The intent of the original procedure is not altered; b.

The change is approved by two members of the plant management staff, at least one of whom holds a Senior Operator license on the unit affected; and

• Review and approval may be performed by the Superintendent of Station Services.

McGUIRE - UNITS 1 and 2 6-14 Amendment No.52 (Unit 1)

Amendment No.33 (Unit 2) i 1

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) c.

The change is documented, reviewed, and approved by the Station Manager; or by:

(1) the Operating Superintendent, (2) the Technical Services Superintendent, (3) the Maintenance Superintendent, or (4) the Superintendent of Integrated Scheduling, as previously l

designated by the Station Manager, within 14 days of implementation.

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

a, Reactor Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain highly-radioactive fluids during a serious transient or accident to as low as practical levels.

The systems include RHR, Boron Recycle, Refueling Water, Liquid Waste, Waste Gas, Safety Injection, Chemical and Volume Control, Contain-ment Spray, and Nuclear Sampling.

The program shall include the following:

1)

Preventive maintenance and periodic visual inspection requirements, and 2)

Integrated leak test requirements for each system at refueling cycle intervals or less.

l b.

In-Plant Radiation Monitoring A program'which will ensure the capability to accurately determine the airborne iodine concentra^!on in vital areas under accident conditions.

This program shall include the following:

1)

Training of personnel, 2)

Procedures for monitoring, and 3)

Provisions for maintenance of sampling and analysis equipment.

c.

Secondary Water Chemistry 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 variables and control points for these variables, 2)

Identification of the procedures used to measure the values of the critical variables, i

3)

Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in-leakage, l

McGUIRE - UNITS 1 and 2 6-15 Amendment No.52 (Unit 1)

~ Amendment No.33 (Unit 2)

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e ADMINISTRATIVE CONTROLS SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT (Continued)

The Radioactive Effluent Release Reports shall include the following information for each type of solid waste shipped offsite during the report l

period:

a.

Total container volume, in cubic meters, b.

Total Curie quantity (determined by measurement or estimate),

c.'

Principal radionuclides (determined by measurement or estimate),

d.

Type of waste (e.g., dewatered spent resin, compacted dry waste, l

evaporator bottoms),

e.

Number of shipments, and l

f.

Solid _ification agent or absorbent (e.g., cement, or other approved agents (media)).

The Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.

The Radioactive Effluent Release Reports shall include'any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the OFFSITE DOSE CALCULATION MANUAL (00CM), as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the la~nd use census pursuant to Specification 3.12.2.

MONTHLY OPERATING REPORTS 6.9.1.8 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or safety valves, shall be submitted on a monthly basis to the Director, Office of Resource Manage-ment, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the NRC Regional Office, no later than the 15th of_each month.following'the calendar month covered by the report.

McGUIRE - UNITS 1 and 2 6-20 Amendment No.52(Unit 1)

Amendment No 33(Unit 2)

ADMINTSTRATIVE CONTROLS-RECORD RETENTION (Continued) g.

Records of training and qualification for current members of the unit staff; h.

Records of inservice inspections performed pursuant to these Technical Specifications; i.

Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59; j.

Records of meetings of the NSRB and reports required by Specification 6.5.1.12; k.

Records of the service lives of all snubbers listed in Tables 3.7-4a and-3.7-4b including the date at which the service life commences and associated installation and maintenance records; 1.

Records of secondary water sampling and water quality; and m.

Records of analyses required by the Radiological Environmental Monitoring Program that would permit evaluation of the accuracy of the analysis at a later date.

This should include procedures effective at specified times and QA records showing that these procedures were followed.

6.10.3 Records of quality assurance activities required by the QA Manual shall be retained for a period of time required by ANSI N45.2.9-1974.

6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and.

adhered to for all operations involving personnel radiation exposure.

6.12 HIGH RADIATION AREA 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area, as defined in 10 CFR Part 20, in which the intensity of radiation is equal to or less than 1000 mrem /hr at 45 CM (18 in.) from the radiati.on source or from any surface which the radiation penetrates shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP).

Individuals qualified in radiation protection procedures (e.g., Health Physics Technician) or personnel contin-uously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates equal'to or less than 1000 mrem /hr provided they are otherwise following plant radiation protection procedures for entry into high radiation areas.

McGUIRE - UNITS 1 and 2 6-23 Amendment No.52 (Unit 1)

Amendment No.33 (Unit 2) e

..n.

ADMINISTRATIVE CONTROLS HIGH RADIATION AREA (Continued)

Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a.

A radiation monitoring device which continuously indicates the radiation dose rate in the area; or b.

A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received.

Entry.into such areas with this monitoring device may be made after the dose rate level in the area has been

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established and personnel have been made knowledgeable of them; or c.

An individual qualified in radiation protection procedures with a l

radiation dose rate monitoring device who is responsible for providing positive control over the activities ~within the area and shall perform periodic radiation surveillance at the frequency specified by the Station Health Physicist in the RWP.

6.12.2 In addition to the. requirements of Specification 6.12.1, areas acces-sible to personnel with radiation levels greater than 1000 mrem /hr at 45 CM (18 in.) from the radiation source or from any surface which the radiation penetrates shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the Shift Foreman on-duty and/or health physics supervision.

Doors.shall remain locked except during periods of access by personnel under an approved RWP which shall l

specify the dose rate levels in the immediate work area and the maximum allowable stay time for individuals in that area.

In lieu of the stay time specification of the RWP, direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.

For. individual areas accessible to personnel with radiation levels greater than 1000 mrem /hr* that are located within large areas, such as PWR containment, where no enclosure exists for purposes of locking, and no enclosure can be reasonably constructed around the individual areas, that area shall be bar-ricaded, conspicuously posted, and a flashing light shall be activated as a warning device.

6.13 PROCESS CONTROL PROGRAM (PCP) 6.13.1 The PCP shall be approved by the Commission prior.to implementation.

6.13.2 Licensee-initiated changes to the PCP:

a.

Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was 4

made.

This submittal shall contain:

• Measurement made at 18 inches from source of radioactivity.

McGUIRE - UNITS 1 and 2 6-24 Amendment No.52 (Unit 1)

Amendment No.33 (Unit 2)

3 This page deleted.

McGUIRE - UNITS 1 and 2 6-27 Amendment No.52(Unit 1)

Amendment No.33(Unit 2)

This page deleted.

McGUIRE - UNITS I and 2 6-28 Amendment No.52 (Unit 1)

Amendment No.33 (Unit 2)

TABLE 3.3-11 (Continued)

?

g3 FIRE DETECTION INSTRUMENTATION 2

n' TOTAL NO. INSTRUMENTS *

[

DETECTOR DESCRIPTION LOCATION SMOKE FIXED TEMP.

FUNCTION **

z ZONE RATE OF RISE w*

189 Unit 2 RB Annulus 16* - 54*

23 23 8

190 Unit 2 RB Annulus 122* - 180*

16 16 B

k 191 Unit 2 RB Annulus 160* - 256*

13 13 B

ro Y.

Y$

a The fire detection instruments located within containment are not required to be operable during the performance of Type A Containment Leakage Rate Tests.

• • Function A:

Early warning fire detection and notification only.

' Function B: Actuation of fire suppression system and early warning and notification.

INSTRUMENTATION RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.8 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall.be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded.

The Alarm / Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM).

APPLICABILITY:

At all times.

ACTION:

a.

With a radioactive liquid effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above specification, immediately suspend the release of radio-active liquid effluents monitt, red by the affected channel, or declare the channel inoperable.

b.

With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-12.

Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION or, in lieu of a Licensee Event Report, explain in the next Semiannual Radioactive Effluent Release Report why this inoperability was not corrected within the time specified.

c.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

l SURVEILLANCE REQUIREMENTS 4.3.3.8 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST operations at the frequencies shown in Table 4.3-8.

McGUIRE - UNITS 1 and 2 3/4 3-66

. Amendment No.52 (Unit 1)

Amendment No.33 (Unit 2)

4 INSTRUMENTATION RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION LIFLilNG CONDITION FOR OPERATION 3.3.3.9 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-13 'shall.be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of Specification 3.11.2.1 are not exceeded. The Alarm / Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the ODCM.

APPLICABILITY:

A shown in Table 3.3-13.

ACTION:

a.

With a radioactive gaseous effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above specification, immediately suspend the release of radioactive I.

gaseous effluents monitored by the affected channel, or declare the channel inoperable.

b.

With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-13.

Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION or, in lieu of a Licensee Event Report', explain in the next Semiannual Radioactive Effluent Release Report why this inoperability was not corrected within the time specified.

c.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

l SURVEILLANCE REQUIREMENTS 4.3.3.9 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST operations at the frequencies shown in Table 4.3-9.

t McGUIRE - UNITS 1 and 2 3/4 3-71 Amendment No.52(Unit 1)

Amendment No.33(Unit 2)

TABLE 3.3-13

?F RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION E

m MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION hk 1.

WASTE GAS HOLDUP SYSTEM U

Noble Gas Activity Monitor - Providing a.

Alarm and Automatic Termination of Release 1 per station 35 g

(Low Range - EMF-50 or IEMF-36, low-range)

[

b.

Effluent System Flow Rate' Measuring Device 1 per station 36 2.

WASTE GAS HOLOUP' SYSTEM Explosive Gas Monitoring System a.

Hydrogen Monitor 1 per station 41 b.

10xygeti Monitors 2 per station 39 3.

Condenser Evacuation System 3

Noble Gas Activity Monitor (EMF-33) 1 37 92 4.

Vent System S$

a.

Noble Gas Activity Monitor 1

37 (Low Range - EMF-36) b.

Iodine Sampler 1

40 c.

Particulate Sampler 1

40 d.

Flow Rate Monitor 1

36 e.

Sampler Minimum Flow Device 1

36-O

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 9)'

Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing.

This inspection shall be performed after the field hydrostatic test and prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.

b.

The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.4-2.

4.4.5.5 Reports a.

Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2; b.

The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection.

This Special Report shall include:

1)

Number and extent of tubes inspected, 2) location and percent of wall-thickness penetration for each indication of an imperfection, and 3)

Identification of tubes plugged.

Amendment No.52 (Unit 1)

McGUIRE - UNITS 1 and 2 3/4 4-15 Amendment No.33 (Unit 2)

t g

8 TABLE 4.4-1 m

t C5 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION

a h3 Preservice inspection No Yes No. of Steam Generators per Unit Two Three Four Two Th_ree Four First. Inservice Inspection

. All One Two Two l

l 2

3 Second & Subsequent inservice lospections One One One One sy Table Notation:

cn

1. The inservice inspection may be limited to one steam generator on.a rotating schedule encompassing 3 N % of the tubes (where N is the number of steam generators in the plant) if the results of the first or previous inspections indicate that all steam generators are performing in a like manner. Note that under some circumstances; the operating conditions in one or more steam generators may be found to be more severe than those in other steam generators. Unde: such circum-stances the sampla sequence shall be modified to inspect the most severe conditions.

2. The other steam generator not inspected during the first inservice inspection sha!! be inspected. The third and subsequent inspections should follow the instructions described in 1 above.

3. Each of the other two steam generators not inspected during the first inservice inspections shall be inspected during the second' and third inspections. The fourth and subseque.nt inspections shall follow the instructions described in 1 above.

i I

m 3

S TABLE 4.4-2 S

-STEAM GENERATOR TUBE INSPECTION C2 IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION

' 3RD SAMPLE INSPECTION

]

Sample Size Result Action Required Result Action Required Result Action Required A minimum of C-1 None N/A N/A N/A N/A a

S Tubes per

'N S. G.

C-2 Plug defective tubes' C-1 ~

None N/A N/A and inspect additional Plug defective tubes C-1 None 2S tubes in this S. G.

C-2 and inspect additional C-2 Plug defective tubes 4S tubes in this S. G.

Perform action for C-3 C-3 result of first g

sample 1

Perform. action for a

C-3 C-3 result of first N/A N/A 4

sample C-3 Inspect all tubes in All other this S. G., plug de-S. G.s are None N/A N/A fective tubes and C-1 inspect 2S tubes in S me S. G.s each other S. G.

Perform action for r!/A

-N/A C-2 but no C-2 result of ser:ond additional sample S. G. are yy C-3 mo gg Additional Inspect all tubes in aa S. G. is C-3 each S. G. and plug defective tubes.

N/A N/A "w %.

^mgg N

Where N is the number of steam generators in the unit, and n is the number of steam generators inspected i

S=3

• $ n*

n during an inspection OO

- - ~.

t

REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE:

The Containment Atmosphere Gaseous Radioactivity Monitoring System, a.

b.

Either the Containment Floor and Equipment Sump Level System or-the Flow Horitoring System, and c.

Either the Containment Ventilation Condensate Drain Tank Level Monitoring System or a Containment Atmosphere Particulate Radioactivity Monitoring System.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With only two of the above required Leakage Detection Systems OPERABLE, operation may continue for up to 30 days provided grab samples of the contain-(.

ment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the required Gaseous or Particulate Radioactivity Monitoring System is inoperable; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by:

Containment Atmosphere Gaseous and Particulate Radioactivity a.

Monitoring Systems performance of CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequencies specified in Table 4.3-3, b.

Containment Floor and Equipment Sump Level System and Flow Monitoring System performance of CHANNEL CALIBRATION at least once per 18 months, and Containment Ventilation Condensate Drain Tank Level Monitoring c.

System performance of CHANNEL CALIBRATION at least once per 18 months.

.McGUIRE - UNITS I and'2 3/4 4-18

REACTOR COOLANT SYSTEM ACTION:

(Continued)

MODES 1, 2, 3, 4, and 5:

With the specific activity of the reactor coolant greater than 1.0 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/E microcuries per gram of gross specific activity, perform the sampling and analysis requirements of Item 4.a) of Table 4.4-4 until the specific activity of the reactor coolant is restored to within its limits.

In lieu of a Licensee Event Report for this ACTION statement within 30 days, l

prepare and submit a Special Report to the Commission pursuant to Specifica-tion 6.9.2 with a copy to the Director, Nuclear Reactor Regulation, Attention:

Chief, Core Performance Branch, and Chief, Accident Evaluation Branch, U.S.

Nuclear Regulatory Commission, Washington, D.C., 20555. This report shall contain the results of the~ specific activity analyses together with the following information:

1.

Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; 2.

Results of the last isotopic analysis for radioiodines performed prior to exceeding the limit, while limit was exceeded, and one analysis after the radioiodine activity was reduced to less than the

-limit, including for each isotopic analysis, the date and time of sampling and the radioiodine concentrations; 3.

Clean-up flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the'first sample in which the limit was exceeded; 4.

History of degassing operations, if any, starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; and 5.

The time duration when the specific activity of the reactor coolant exceeded 1.0 microcurie per gram DOSE EQUIVALENT. I-131.

t SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the reactor coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.

Amendment No. 52 (Unit 1)

McGUIRE - UNITS 1 and 2 3/4 4-26 Amendment No'. 33 (Unit 2)

CONTAINMENT SYSTEMS CONTAINMENT VESSEL STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.6 The structural integrity of the containment vessel shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.6.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

With the structural integrity of the containment vessel not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing.the Reactor Conlant System temperature above 200 F.

SURVEILLANCE REQUIREMENTS 4.6.1.6 The structural integrity of the containment vessel shall be determined during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.2) by a visual inspection of the exposed accessible interior and exterior surfaces of the vessel.

This inspection shall be performed prior to the Type A containment leakage rate test to verify no apparent changes in appearance of the surfaces or other abnormal degradation.

Any abnormal degra-dation of the containment vessel detected during the above required inspec--

tions shall be reported to the Commis'sion pursuant to 10 CFR Sections 50.72 and 50.73.

McGUIRE - UNITS 1-and 2 3/4 6-14 Amendment No. 33(Unit 2)

Amendment No. 52(Unit 1) l

CONTAINMENT SYSTEMS REACTOR BUILDING STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.7 The structural integrity of the reactor building shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.7.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

With the structural integrity of the reactor building not conforming to the above requirements, restore the structural integrity to within the. limits prior to increasing the Reactor Coolant System temperature above 200 F.

SURVEILLANCE REQUIREMENTS ~

4.6.1.7 The structural integrity of the reactor building shall be determined during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.2) by a visual inspection of'the exposed accessible interior and exterior surfaces of the reactor building and verifyingino apparent changes in appearance of the concrete surfaces or other abnormal-degradation.

Any abnormal degradation of the reactor building detected during the above required inspections shall be reported to the Commission pursuant to 10 CFR Sections 50.72, and 50.73.

McGUIRE - UNITS 1 and 2 3/4 6-15

. Amendment No.33 (Unit 2)

Amendment No.52 (Unit 1)

CONTAINMENT SYSTEMS ANNULUS VENTILATION SYSTEM.

t LIMITING CONDITION FOR OPERATION

3. 6.1. 8 Two independent Annulus Ventilation Systems shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3, ar.d 4.

ACTION:

With one Annulus Ventilation System inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the'next

~

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.1.8 Each Annulus Ventilation System shall be demonstrated OPERABLE:

At least once per 31 days on a STAGGERED TEST BASIS, by initiating, a.

from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with the pre-heaters operating;

(

b.

At least once per 18 months, or (1) after any. structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following

. painting, fire,- or chemical release in any ventilation zone communi-cating with the system,.by:

1)

Verifying that tne ventilation system satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 1% and uses the test procedure guidance of Regulatory Positions C.S.a, C.5.c and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system' flow rate is 8000 cfm i 10%;

2)

Verifying within 31 days after removal that a. laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the. laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than js; and I-McGUIRE - UNITS 1 and 2 3/4 6-16

ELECTRICAL' POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 1)

Draining each fuel oil' storage tank, removing the accumulated sediment and cleaning the tank using a sodium hypochlorite solution, and.

2)

Performing a pressure test of those portions of~the diesel fuel oil system designed to Section III, subsection ND of the ASME Code at a test pressure equal to 110% of the system design pressure.

4.8.1.1.3 Reports - All diesel generator failures, valid or non-valid, shall be reported to the Commission pursuant to Specification 6.9.2 within 30 days.

l Reports of diesel generator failures shall include the information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977.

If the number of failures in the last 100 valid tests (on a per' nuclear unit basis) is greater than or equal to 7, the report shall be supplemented to include the additional information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977.

4.8.1.1.4 Diesel Generator Batteries - Each diesel generator 125-volt battery bank and charger shall be demonstrated OPERABLE:

a.

At least once per 7 days by verifying t' hat:

1 -)

The electrolyte level of each battery is above the plates, and 2)

The overall battery voltage is greater than or equal to 125 volts under a float charge.

b.

At least once per 18 months by verifying that:

i 1)

The batteries, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration; l

2)

The battery-to-battery and terminal connections are clear, tight, free of corrosion and coated with anti-corrosion material; arii 3)

The battery capacity is adequate to supply.and maintain in OPERABLE status its emergency loads when subjected to a battery i

service test.

McGUIRE - UNITS 1 and 2 3/4 8-7 Amendment No. 52 (Unit 1)

. Amendment No. 33 (Unit 2)

TABLE 4.8-1

(

DIESEL GENERATOR TEST SCHEDULE NUMBER OF FAILURES IN LAST 100 VALIO TESTS

• TEST FREQUENCY

<1 At least once per 31 days 2

At least once per 14 days 3

At least once per 7 days

>4 At least once per 3 days

(

l I

• Criteria for determining number of failures and number-of valid tests shall be in accordance with Regulatory Position C.2.e of Regulatory Gui.de 1.108, Revision 1, Augu't 1977, where the last s

100' tests are determined on a per nuclear unit basis.

For the purposes of this test schedule, only valid tests conducted after the OL issuance date shall be included in the computation of the

last 100 valid tests." Entry into this. test schedule shall be made at the 31 day test frequenc~y.

'McGUIRE - UNITS 1 and 2 3/4 8-8

3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS CONCENTRATION LIMITING CONDITION FOR OPERATION 3.11.1.1 The concentration cf radioactive material released in liquid effiuents to UNRESTRICTED AREAS (see Figur'e 5.1-4) shall'be limited to.the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for' radionuclides other than dissolved or entrained noble gases.

Fordigsolvedor entrained noble' gases, the concentration shall be limited to 2 x 10 microcurie /ml total activity.

APPLICABILITY:

At all times.

ACTION:

With the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the above limits, immediately restore the concen-tration to within the above limits.

SURVEILLANCE REQUIREMENTS 4.11.1.1.1 Gadioactive liquid. wastes shall be sampled and analyzed according to the sampling and analysis program of Table 4.11-1.

4.11.1.1.2 The results of the radioactivity analyses shall be used in accordance with the methodology and parameters in the ODCM to assure that the concentrations at the point of release are maintained within the limits of Specification 3.11.1.1.

s I

MCGUIRE - UNITS 1 and 2 3/4 11-1 Amer dment No. 52 (Unit 1)

Amendmer.t No. 33 (Unit 2) l

TABLE 4.11-1

~

RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM LOWER LIMIT-MINIMUM 0F DETECTION LIQUID RELEASE SAMPLING ANALYSIS TYPE OF ACTIVITY

.(LLD)

TYPE FREQUENCY FREQUENCY ANALYSIS (pCi/ml)(1)

1. Batch Waste P

P

,7 Release (4)

Each Batch Each Batch Principaf6 gamma 5x10 Tanks Emitters

-6 I-131 1x10 (Waste Monitor

-5 P

M Dissolved and 1x10 ks and One Batch /M Entrained Gases ye (Gamma emitters)

Monitor Tank)

-5 P

M H-3 1x10 Each Batch Composite ( )

-7 Gross Alpha 1x10

-8 P

Q Sr-89, Sr-90 5x10 Each Batch Composite (2)

-6 Fe-55 1x10

2. Continuous (5)'

Continuous (3) Composite (3)

Principaf6 gamma 5x10 Releases Emitters

-6 I-131 1x10 (Containment Ventilation

_S.

M M

Dissolved and 1x10 Unit Grab Sample Entrained Gases Condensate (Gamma Emitters)

Drain Tank

. Discharge and

,g Conventional M

H-3 1x10 Waste Water Continuous ( ) Composite (3)

_y Treatment Gross Alpha 1x10 System Outlet)

-8 l-Q Sr-89, Sr-90 5x10 Continuous ( ) Composite (3)

-6 Fe-55 1x10 MCGUIRE - UNITS 1 and 2 3/4 11-2 Amendment No. 52 (Unit 1)

Amendment ~ No. 33 (Unit 2)

RADI0 ACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS DOSE RATE LIMITING CONDITION FOR OPERATION

'3.11.2.1 The dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1-3) shall be limited to the following:

a.

For noble gases:

Less than or equal to 500 mrem /yr to the whole body and less than or equal to 3000 mrem /yr to the skin, and b.

For Iodine-131 and 133, for tritium, and for all radioactive materials in particulate form with half-lives greater than 8 days:

Less than or equal to 1500 mrem /yr to any organ.

APPLICABILITY: At all times.

ACTION:

With the dose rate (s) exceeding the above limits, immediately restore the release rate to within the above limit (s).

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

4.11.2.1.2 The dose rate due to radioactive materials, other than noble gases, in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters of the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 4.11-2.

MCGUIRE - UNITS 1 and 2 3/4 11-9 Amendment No. 52 (Unit 1)

Amendment No. 33 (Unit 2)

TABLE 4.11-2 RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM 1

9, MINIMUM LOWER LIMIT OF 5

SAMPLING ANALYSIS TYPE OF DETECTION (g)

A GASEOUS RELEASE TYPE FREQUENCY FREQUENCY ACTIVITY ANALYSIS (pCi/ml)

Principal Gamma Emitters (7) 1x10-4 h

1.

Waste Gas Storage Each Tank Each ank 5

Tank Grab d

Sample 2.

Containment Purge Each Purge (2) Each Purge (2)

Principal Gamma Emitters (7) 1x10-4 Grab

-6 Sample M

H-3 1x10 m

I7)

~4 3.

Unit Vent W(2),(3),(5) y(2)

Principal Gamma Emitters 1x10 Grab

-6 Sample H-3' 1x10 4.

a.

Radwaste W

W Principal Gamma Emmiters (7) 1x10~4 t'

Facility Grab

-6 Vent Sample H-3 1x10 M

b.

Contaminated Materials o

Warehouse 5.

All Release Types Continuous (6)-

0(4)

I-131 1x10-11 as listed in 3.

Charcoal

_g and 4. above.

Sample I-133 1x10 54)

Principal Gamma Emitters (7)

-10 Continuous (6) 0 1x10 Particulate (I-131, Others)

Sample Gross Alpha (8) 1x10'II Continuous (6)

M Composite Particulate Sample

-II Continuous (6)

Q Sr-89, Sr-90 1x10 Composite Particulate Sample n

m

REACTOR COOLANT SYSTEM BASES 3/4.4.4 RELIEF VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump.

Operation of the PORVs minimizes the undesirable opening of the spring-loaded pressurizer code safety valves.

Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable.

3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity-of this portion of the RCS will be main-tained.

The program for inservice inspection of' steam generator tubes is based on a modification of 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 mechanical damage or progressive 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 theLsecondary 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 Reactor Coolant System and the Secondary Coolant System (reactor-to secondary leakage = 500 gallons per day per steam generator).

Cracks having a reactor-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 reactor-to-secondary leakage of 500 gallons per day per-steam generator can readily be detected by radiation monitors of steam l

generator blowdown.

Leakage in excess of this limit will require plant shutdown and an unscheduled inspection,' during which the leaking tubes will be located and plugged.

Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant.

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

Plugging will be required for all tubes with imperfections exceeding the plugging limit of 40% of the tube nominal wall. thickness.

Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.

McGUIRE - UNITS 1 and 2 B 3/4 4-3

REACTOR COOLANT SYSTEM BASES STEAM GENERATORS (Continued)

Whenever the results of any steam generator tubing i.nservice inspection fall into Category C-3, these results will be promptly reported to the Commission pursuant to 10 CFR Sections 50.72 and 50.73 prior to resumption of

{

plant operation.

Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory exami-nations, tests, additional eddy current inspection, and revision of the Technical Specifications, if necessary.

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

3/4.4.6.2 OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 gpm.

This threshold value is sufficiently low to ensure early detection of additional leakage.

The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve. failure and consequent intersystem LOCA.

Leakage from the RCS pressure isolation valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allnwid Tirit.

The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited amount cf leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the Leakage Detection Systems.

The CONTROLLED LEAKAGE limitatien restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 40 gpm with the modulating valve in the supply line fully open at a nominal RCS pressure of 2235 psig.

This limitation ensures that in the event c' a LOCA, the Safety Injection flow will not be less than assumed in the accident analyses.

The total steam generator tube leakage limit of 1 gpm for all steam generators not isolated from the RCS ensures that the dosage contribution from the tube leakage will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the~ event of either a steam generator tube rupture or steam line break. The 1 gpm limit is consistent with the assumptions used in the analysis of these accidents.

The 500 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.

McGUIRE - UNITS 1 and 2 B 3/4 4-4 Amendment No.52(Unit 1)

Amendment No.33(Unit 2)