ML19345F785

Forwards Updated Response to TMI Concerns to Reflect Requirements of NUREG-0737.Info Provided Re Shift Technical Advisor,Shift Crew Composition,Special Training for Mitigating Core Damage & Emergency Operating Procedures

ML19345F785
Person / Time
Site:	Mcguire, McGuire
Issue date:	02/06/1981
From:	Parker W DUKE POWER CO.
To:	Harold Denton, Youngblood B Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, TASK-1.A.1.1, TASK-2.B.4, TASK-TM NUDOCS 8102190296
Download: ML19345F785 (200)

Text

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373-4093 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Mr. B. J. Youngblood, Chief Licensing Projects Branch No. 1

Subject:

McGuire Nuclear Station Docket Nos. 50-369 and 50-370

Dear Mr. Denton:

Enclosed with this letter are forty copies of updated responses for the

'w/ document " Duke Power Company, McGuire Nuclear Station, Response to TMI Concerns." This document was transmitted to the NRC via my letter of May 23, 1980 and updated via my letters of July 18, 1980, August 6, 1980, September 8, 1980, October 10, 1980, and October 29, 1980.

In June of 1980 the NRC consolidated the TMI-related licensing requirements for the near-term operating license (NTOL) plants into the document "TMI-Related Requirements for New Operating Licenses, NUREG-0694." This document in conjunction with:

1. the report "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations, NUREG-0578" (transmitted to NTOL plants via Mr. D. B. Vassallo's letter of September 27, 1979);

2. the " clarifications" to NUREG-0578 titled " Discussion of TMI Lessons Learned Short-Term Requirements" (transmitted to NTOL plants via Mr. D. B. Vassallo's letter of November 9, 1979);

3. an NRC letter establishing the required " Qualifications of Reactor Operators" (transmitted to all licensees and applicants via Mr.

H. R. Denton's letter of March 28, 1980); and

4. an NRC letter establishing the " Interim Criteria for Shift Staffing" (transmitted to all licensees and applicants via

,s Mr. D. G. Eisenhut's letter of July 31, 1980)

's-) provided the licensing criteria upon which subsequent Duke actions were based.

These actions have included designing, ordering and installing additional equipment as well as modifying existing equipment.

f 8102190 qq6 l

Mr. Harold R. Denton, Director

~ February 6, 1981 Page Two on October 31, 1980 the NRC issued the document " Clarification of IMI Action Plan Requirements, NUREG-0737." All of the previously issued TMI-related licensing requirements are incorporated into NUREG-0737 and are further

" clarified." In addition NUREG-0737 issues new TMI-related licensing requirements.

It is important to note that NUREG-0737 does not simply clarify the previously issued TMI-related requirements. In many cases the requirements are revised substantially. Duke has made many commitments and installed much additional equipment at McGuire based on the previous requirements. Mr. D. G. Eisenhut's introduction to NUREG-0737 appears to consider this fact in that he states that NTOL plants will be reviewed on a case-by-case basis. McGuire Nuclear Station clearly falls into that category of plants to be reviewed on a case-by-case basis.

The updated responses enclosed herein reflect Duke's review of NUREG-0737 as well as provide the current status of previous commitments. These responses also address the new requirements issued by NUREG-0737.

Very truly yours,

\~/

s/ William O. Parker, Jr.

William O. Parker, Jr.

TEH:ses Enclosures O

Mr. Harold R. Denton, Director February 6, 1981 Page 3 WILLIAM 0. PARKER, JR. , being duly sworn, states that he is a Vice President of Duke Power Company; and he is authorized on the part of said Company to sign and file with the Nuclear Regulatory Commission this docu-ment, Duke Power Company, McGuire Nuclear Station Response to TMI Concerns, and thar all statements and matters set forth therein are true and correct to the best of his knowledge.

s/ William O. Parker, Jr.

William O. Parker, Jr., Vice President

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Subscribed and sworn to before me this 5th day of February, 1981.

\s s/ Sue C. Sherrill Notary Public My Commission Expires:

September 20, 1984 A

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O DUKE POWER COMPANY MCGUIRE NUCLEAR STATION Response to IMI Concerns Changes and Corrections Remove These Pages Insert These Pages Table of Contents Pg. 1 08/06/80 Table of Contents Fg. 1 02/06/81 Table of Contents Pg. 11 10/10/80 Table of Contents Pg. 11 02/06/81 Table of Contents Pg. iii 10/10/80 Table of Contents Pg. iii 02/06/81 I-1 10/10/80 I-1 02/06/81 I-3 08/06/80 I-3 02/06/81 I-3A 10/29/80 I-3A '02/06/81 I-3B 08/06/80 I-3B 02/06/81 I-3C 08/06/80 I-3C 02/06/81 I-4 07/18/80 I-4 02/06/81 I-9 10/10/80 I-9 02/06/41 I-13 10/10/80 I-13 02/06/81 I-14 10/10/80 1-14 02/06/81 I-15 10/10/80 I-15 02/06/81 O I-16 I-16 02/06/81 C II-3 II-4 10/10/80 10/10/80 II-3 02/06/81 II-4 02/06/81 II-5A 02/06/81 II-5B 02/06/81 II-8 09/08/80 II-8 02/06/81 II-8A 09/08/80 II-8A 02/06/81 II-9 09/08/80 II-9 02/06/81 II-10 07/18/80 II-10 02/06/81 Carryover II-12 10/10/80 II-12 02/06/81 II-13 10/10/80 II-13 02/06/81 II-13A 10/10/80 Il-13A 02/u6/81 II-13B 07/18/80 II-13B 07/18/80 Carryover II-13D 07/18/80 Carryover II-14 II-14 02/06/81 II-16 C9/OS/80 II-16 09/08/80 II-16A 09/08/80 II-16A 02/06/81 II-17C 02/06/81 II-17D 02/06/61 II-18 07/18/80 II-18 02/06/81 II-18A 07/18/80 II-18A 02/06/81 II-19 10/10/80 II-19 02/06/81 II-19A 10/10/80 19A 02/06/81 19B .02/06/81 w/ 19C 02/06/81 II-20 02/06/81

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Remove These Pages Insert These Pages III-1 1c/10/80 III-1 02/06/81 III-2 III-2 02/06/81 III-4 09/08/80 III-4 02/06/81 III-5 02/06/81 Appendix A Appendix A Station Directive 3.1.4 (Rev. 1) Station Directive 3.1.4 (Rev. 2)

Station Directive 3.8.2 (Rev. 0) Station Directive 3.8.2 (Rev. 3)

Station Directive 3.1.9 (Rev. 4) Station Directive 3.1.9 (Rev. 6) l Note: (Do Not Remove PT/1/A/4700/10) l Statian Directive 3.1.31 Station Directive 3.1.31 (11/06/80) 1 Station Directive 3.1.32 Charter of the Station Safety Review Group Appendix C Appendix C Remove all of Current Appendix C Insert All of New Appendix C Appendix D Appendix D f June 30, 1980 Letter From l June 30, 1980 Letter From B. J. Youngblood (08/06/80) B. J. Youngblood(02/06/81)

November 6, 1980 Letter from R. L.

Tedesco (02/06/81)

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l A TABLE OF CONTENTS U

zm Section I. Operational Safety Shift Technical Advisor I-I Shift Supervisor Duties and Responsibilities I-2 Safety Engineering Group I-3 Shift Manning I-4 I-5 Qualifications of McGuire Nuclear Station Personnel Revised Scope and Criteria for Licensing Examinations I-6 Upgrading Operator Training and Qualifications I-7 Administration of Training Programs for Licensed Operators I-8 Training for Mitigating Core Damage I-9 Training During Low Power Testing I-10 Control Room Access 1-11 Shift Relief and Turnover Procedures I-12 NSSS Vendor Review of Procedures I-13 Pilot Monitoring of Selected Emergency Procedures I-14 Accident Analysis and Procedure Revision I-15 Procedures for Verifying Correct Performance of Operating Activities I-16 Control Room Design I-17

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\,_/ General Office Training I-18 Section II. Design Relief and Safe'* Valve Position Indication 11-1 Relief and Safet Valve Testing II-2 Auxiliary Feedwater Initiation and Indication II-3 Auxiliary Feedwater System Reliability Evaluation 11-4 Containment Isolation Provisions II-5 E=ergency Power for Pressurizer Equipment II-6 Reactor Coolant System Vents II-8 Inadequate Core Cooling Instruments II-9 Additional Accident Monitoring Instrumentation II-12 Post-Accident Sa'pling II-14 Plant Shielding II-15 Centrol Room Habitability II-17 IE Bulletins on Measures to Mitigate Small-Break LOCA's and Loss of Feedwater Accidents II-18 Final Recommendations of the Bulletins and Orders Task Force II-19 II-20 Commission Orders on Babcock and Wilcox PlantsSection III. Emergency Preparedness and Radiation Effects Upgraded Emergency Preparedness III-l On-Site Technical Support Center III-2 f)

x,,) On-Site Operational Support Center In-Plant Radiation Monitoring III-3 III-4 Primary Coolant Sources outside Containment III-5 1 02/06/81 e p- y--g - -- * - 4-m-w-+ + -- w y = - - -

Appendix A McGuire Nuclear Station Procedures l j

Station Directive 3.1.4, Conduct of Operations  !

Station Directive 3.8.2, Station Emergency Organization Station Directive 3.1.9, Relief at Duties of Plant Operation Periodic Test PT/1/A/4700/10, Shift Turnover Verification Station Directive 3.1.31., Duties, Responsibilities and Qualifications of Gie Shif t Technical Advisor Charter of the Station Safety Review Group i

l Appendix B Control Room Design August 15, 1980 letter from W. O. Parker to H. R. Denton i

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l Appendix C Leak Rate Procedures for Systems Containing Primary Coolant Outside Containment Periodic Test Leak Rate Determination for NI System Periodic Test Leak Rate Determination for ND System

(~N Periodic Test Leak Rate Determination for NS System t

_) Periodic Test Leak Rate Determination for NM System Periodic Test Leak Rate Determination for NB System Periodic Test Leak Rate Determination for NV System Periodic Test Leak Rate Determination for FW System Periodic Test Liquid Waste System Leakage Check Appendix C NRC Requests for Additional Information June 4, 1980 letter from B. J. Youngblood to W. O. Parker June 30, 1980 letter from B. J. Youngblood to W. O. Parker July 2, 1980 letter from B. J. Youngblood to W. O. Parker July 23, 1980 letter from R. L. Tedesco to W. O. Parker August 25, 1980 letter from R. L. Tedesco to W. O. Parker September 17, 1980 letter from Ralph Birkel to W. O. Parker September 17, 1980 letter from B. J. Youngblood to W. O. Parker November 6, 1980 letter from R. L. Tedesco to W. O. Parker 11 02/06/81

Appendix E Special Low-Power Tests Safety Evaluation l

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l SHIFT TECHNICAL ADVISOR

References:

NUREG-0578 - 2.2.lb Action Plan - I.A.l.1 A technical advisor to the shift supervisor is present on all shifts and  !

available to the control room within ten minutes whenever the unit is in operating modes 1, 2, 3 or 4. The shift technical advisor's primary duty is to provide evaluation and assessment of both normal and unanticipated transients. The shift technical advisor is detached from and independent of the normal line responsibility for plant operation.

The shift technical advisors have been selected from the group of licensed senior reactor operators at McGuire. All of the McGuire SRO applicants have received additional simulator and academic training. The simulator training included functioning as the STA during various transients, and the academic training included instruction in heat transfer, fluid fidw, thermo-dynamics, and plant transients.

In addition eleven McGuire SRO licensees / applicants have completed a special STA training program based on the April 30, 1980 guidelines established by INPO. This program commenced on September 29, 1980 and consisted of four weeks of instruction. The first three weeks consisted of forty hours per week of classroom instruction with the remaining week consisting of twenty hours of classroom instruction and twenty hours of simulator instruction.

This training program included instruction in the following topics: STA responsibilities and accountabilities, management and supervisory skills; transient and accident analysis; seismic monitoring; plant status monitoring; plant metallurgy; plant chemistry; instrumentation and control theory; Pk'R heat transfer; Pk'R thermal and hydrau..c transient response; small break LOCA analysis; and mitigating reactor core d>anage. This training program will be inec,rporated into the standard SRO training program for future McGuire-SRO candidates.

Duke's long-term plan concerning the STA position calls for an eventual phase-out of this position. Before the STA position is phased out, our shift supervisors will meet the requirements and qualifications then in effect for the STA position. To prepare for the STA phase-out, we have developed an academic, college level program in which selected operations personnel will be enrolled. This academic program is designed to upgrade the technical educational level of our operating personnel to a level where the STA posi%on will not be necessary. This program is equivalent to sixty semester hours in both basic engineering and plant applications of engineering principles. INP0's standards for the STA and shift supervisor positional requirements will provide guidance in determining the appropriate educational level of our operating staff and shift personnel. It should be recognized that to accomplish a meaningful educational upgrade and to continue to maintain adequate operating experience on site to support the safe operation of ';he plant will require several years.

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OPERATING EXPERIENCE EVALUATION PROGRAM

Reference:

Action Plan - I.B.1.2 and I.C.5 SCOPE The purpose of the Operating Experience Evaluation Program (OEEP) is to provide a formal mechanism for the systematic evaluation of off-normal events occurring at Duke Power Company nuclear units, as well as at other facilities. This evaluation serves to confirm that plant response was as expected for anticipated transients, and assures that any unexpected behavior is investigated thoroughly.

The results of this evaluation are then used to identify procedural and/or design changes which may mitigate or preclude the recurrence of a similar event.

The evaluation program takes on added importance in assessing any unanticipated transients to assure that they are well understood and that appropriate correc-tive measures are taken. In order to assure that the program is truly effective, information gained from Duke Power Company experience will be disseminated to other organizations, as appropriate.

PROGRAM DESCRIPTION ORGANIZATION O In order to achieve the objectives of the OEEP in an efficient manner, an off-site organization has been established, as well as an on-site organization at each of the nuclear stations.

The on-site organization includes the Station Manager, who is responsible for the operation and safety of the plant; the Shift Technical Advisor, who pe: forms the accident assessment function; the Station Safety Review Group (SSRG), which reviews the investigations of events and the adequacy of proposed cricrective actions; and the supervisors of areas relevent to a significant event, who may perform the initial investigation and must implement corrective actions.

The off-site organization consists of principal engineering support groups, who interface with station personnel in investigating and evaluating events and developing remedial actions; company management, who holds overall responsi-bility for nuclear plant safety; and the Nuclear Safety Review Board (NSRB) which performs an independent review function. An~ engineering group has been assigned the responsibility of coordinating and documenting the off-site activities. This group also serves as the principle interface between Duke and external sources of operating experience information, such as INP0/NSAC, the NRC, the NSSS vendors, and other utilities.

EVALUATION OF DUKL DOWER SYSTEM EXPERIENCE A functional flow chart of the OEEP for events which occur at Duke Power Company j plants is provided as Figure 1. The occurrence of any event is brought to the attention of the Projects and Licensing Engineer who determines whether the event is significant enough to warrant investigation. If investigation is warranted, he d notifies the Station Manager, notifies the General Office Project Coordination and I Licensfog (PC&L) Section, assigns an engineer to investigate the event, and noti-1 fies the NRC if appropriate.

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i i 1 The event investigator prepares a report describing the cause of the event and i l

\ any relevant plant behavior, and outlines proposed corrective actions. The  !

SSRG then reviews each report for acciracy and ce=pleteness, and assesses the adequacy of proposed corrective actions. The SSRG sub=its its report to the Station Manager and the NSRS for review and for approval of corrective actions.

The Station Manager assures that the approved corrective actions are i=ple=ented, are sent to the Shift Technical Advisor for review with regard to operating pro-cedures, and are sent to the Supervisor of Training for inclusion of relevant information in the training program.

The SSRG will be composed of four full ti=e members assigned to the day shif t.

The group will be staffed on a rotating basis frc= a=ong experienced station personnel and will be =ultidisciplined with expertise in the areas of instru=en-tation, =aintenance, operations, and technical services. Additional information on the =e=bership and duties of the SSRG is provided in the SSRG Charter which is included in Appendix A.

Jpon notification of an event, the General Office PC&L Section notifies co=pany

=anage=ent and =ay alert other General Office engineering and scientific sup-port groups. The Station Manager forwards the approved event report to the General Office PC&L Section, where a Licensee Event Report (LER) is prepared and sub=itted, if necessary. Infor=ation is provided to other organizations, including INP0/NSAC, NRC, and the NSSS vendor. Detailed evaluations of plant transients are perfor=ed, and event occurrence data is maintained. As appropriate, other engineering support groups review the LER and station event reports for further reco==endations on corrective actions, and =ay interface with appro-priate equipment vendors. The NSR3 perfor=s an independent review of the event report, the LER, and the effectiveness of any follow-up actions.

l INDUSTRY EXPERIENCE EVALUATION Figure 2 illustrates the flow path for infor=ation received concerning industry operating experience. Significant events will be brought to the attention of Duke Power Cc=pany by INP0/NSAC, NSSS vendors, other utilities, or the NRC.

Information is distributed, as appropriate, by the offsite coordinating group to the SSRG, to General Office engineering support groups for review and develop-

=ent of corrective actions and to the Training Services group for incorporation into the training progra=. The SSRG reviews the infor=ation for applicability to the specific station, and makes recce=endations to the NSR3 and the Station The Station Manager then Manager in areas where action may be nere.sary.

developes and i=ple=ents appropriate r>rrective actions with assistance from and review by the engineering suppor' groups.

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FIGURE 2 Industry Operating Experience Evaluation (Functional Flow Chart)

Industry Operating Experience Data h

INF0/NSAC e Screen data for significant events e Issue significant event report Other Utilities Vendor NRC ua l

GO  ; GO Engineering SSRG : Support Groups s,)

PCSL e Review for applicability e Distribute informa- e Review for tion to SSEG and information e Document review cognizant groups e If applicable, determine e Inform INP0/NSAC e Assist Station and NRC of any Manager to areas where action is necessary and submit action taken develop necessary corrective actions recommendation to Station Manager and Nuclear Safety Review Board

• Training Services

-o- Station Manager e Incorporate e Aoprove and implement relevant informa-corrective actions tion into the training program I-3C 02/06/tsi

SHIFT MANNING

Reference:

Action Plan - I.A.I.3 The shift crew composition for operation of McGuire Unit i vill be it. accor-dance with Section 6.0 of the McGuire Technical Specifications. The minimum shif t manning for Unit i vill be one shif t supervisor (SRO), one senior reac-tor operator in the Control Room (SRO), two reactor operators (R0 License -

one in Control Room at all times), two nuclear equip =ent operators (non-licensed personnel). In addition, a shift technical advisor will be assigned to each shift in a strictly advisory capacity. A licensed SRO will be stationed in the Control Room whenever the unit is operating in Modes 1, 2, 3 or 4. If the shife technical advisor is the only SRO in the Control Room, the shift supervisor or assistant shift supervisor vill be available to the Control Room within ten minutes.

Provisions governing the amount of overtime worked by licensed operators have been incorporated into the McGuire administrative procedure, Station Directive 3.1.4, Conduct of Operations. This procedure, provided in Appendix A, states that licensed operators shall (1) not work more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> straight, (2) not work more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48-hour period, (3) oce work more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any 7-day period, and (4) not work more than 14 consecutive days without having 2 consecutive days off. These limits on working time do not include shift turnover time. Deviation from these limits on working time will be authorized and documented by either the Superintendent of Operations or the h

bl Station Manager.

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f TFAINING FOR MITIGATING CORE DAMAGE

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Reference:

Action Plan - II.B.4 A special training program for mitigating core damage weJ corcacted at __

McGuire by the General Physics Coropration. This program commenced on October 13, 1980 and consisted of forty hours of instruction covering five consecutive days. Lecture topics included the following: accident analysis; PWR heat transfer; PWR thermal and hydraulic transient response; LOCA analysis; incore and excore instrumentation; vital instrumentation; reactor chemistry; radiation monitoring; and gas generation. All of the available McGuire operators participated in this training program. Sele cte.d_ McGui,re operator training instructors and other appropriate Duke personnel including the Superintendent of Operations and certain operating engineers also participated.

Instruction in mitigating core damage will be provided to the following McGuire personnel before power escalation of Unit 1:

1) Station Manager,

2) those operating engineers and licensed operators who did not participate in the OctoE r, 1980 training program, and

3) the top level technicians in the Instrumentation and Electrical, Health Physics, and Chemistry sections.

This instruction will be provided in two parts. The first is a program which O' will primarily address what actions are necessary to appropriately respond to an accident which has occurred. This program will apply to all technical l disciplines and employee levels. A second level of training will be provided to specific station groups so that they can effectively perform their accident mitigation duties.

Training for mitigating core damage will be incorporated into the McGuire operator training and requalification programs. This training will place increased emphasis on the operation and significance of any McGuire systems or instrumentation which could be used to monitor and control accidents in which the core may be severely damaged. Vital instrumentation which supplies the operator with needed information in a' degraded core situation and alter-nate methods of obtaining this information will be identified. Specific instruction in the interpretation of instrument readings in degraded core situations will also be provided.

i i The existing body of knowledge regarding nuclear plant response under degraded core conditions is being enlarged. Duke is participating in this effort in conjunction with other utilities, INPO, and the NSSS vendors. Information resulting from this effort will be incorporated into appropriate training programs for McGuire station personnel soon after it is available. Enclosure 3 to Mr. H. R. Denton's letter of March 28, 1980 to All Power Reactor Applicants

and Licensees is being used as a basis in the development of this information.

In addition the cleanup effort at TMI Unit 2 should provide significant infor-mation in this regard.

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v O NSSS VENDOR REVIEW OF PROCEDUREd

Reference:

Action Plan - I.C.7 On May 15, 1980 a selected group of McGuire emergency procedures were sub-mitted to Westinghouse for their review. On August 4, 5, and 6, 1980 a se=inar was held at McGuire to discuss the Westinghouse Emergency Operating Instructions and selected McGuire emergency procedures with Westinghouse.

Subsequent to this seminar the McGuire emergency procedures were modified.

On October 2, 1980 the following draft emergency procedures were sent to Westinghouse for review:

1) Immediate Actions and Diagnostics for Safety Injection

2) Loss of Reactor Coolant

3) Steam Generator Tube Ruptare

4) Loss of Steam Generator Feedwater

5) Secondary Line Rupture Westinghouse has reviewed the McGuire procedurr.s for initial core loading and low-power physics testing and has found them to be consistent with the Standard Westinghouse recommended proceduras.

The special low-power test procedures were sent to Westinghouse on October 3, O-s- 1980 for review. Westinghouse reviewed these procedures and prepared a safety evaluation of this low-power test program. This safety evaluation is provided in Appendix E. The power ascension test procedures will be reviewed by Westinghouse prior to power escalation of McGuire Unit 1.

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- PILOT MONITORING OF SELECTED EMERGENCY PROCEDURES

Reference:

Action Plan - I.C.8 on May 15, 1980 a selected group of McGuire emergency procedures were submitted to the NRC for review. Based on the pre.liminary NRC review of these and other Westinghouse NTOL emergency procedures and Westinghouse's preliminary review of the McGuire emergency procedures Duke Power Company revised certain McGuire emergency procedures. Draft verr. ions of the revised procedures were submitted -

to the NRC for review via Mr. W. O. Parker's letters of September 16, 1980 and October 2, 1980. A second draft of these procedures was submitted to the NRC on November 3, 1980. On November 11-13, 1980 an NRC review team visited the McGuire site to discuss these procedures wit;. Duke personnel and observe the performance of the procedures on the McGuire simulator. On December 3, 1980 copies of the following revised McGuire emergency procedures were sent to the NRC:

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1) Immediate Actions and Diagnostics

2) Loss of Reactor Coolant

3) Steam Generator Tube Rupture These procedures had been revised based on comments received from Westinghouse and from the NRC Staff.

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xJ ACCIDENT ANALYSIS AND PROCEDURE REVISION

References:

NUREG-0578 - 2.1.9 Lction Plan - I.C.1 Duke is in the process of developing new procedures and training guidelines for controlling and mitigating small break LOCAs, incidents of inadequate core cooling, and certain anticipated transients. Duke's effort is in con-junction with analysis and research being performed by Westinghouse.

The Westinghouse analysie of small break LOCAs in upper head injection plants, WCAP 9600 and WCAP 96'r nas been submitted to the NRC for their review. Duke has reviewed these tre arts and made the necessary modifications to the McGuire emergency procedur? and training program.

The Westinghouse analysis of inadequate core cooling, WCAP 9753, WCAP 9754, and WCAP 9744, has been submitted to the NRC for their review. These reports provide the analytical basis for the Westinghouse guidelines for the detection of and recovery from inadequate core cooling. These guidelines have been sub-mitted for NRC review and approval. Duke will prepare an inadequate core cooling procedure consistent with the approved guidelines. This procedure and its accompanying training will be complete by January 1, 1982.

- The Westinghouse analysis of selected transients and accidents, WCAP 9691, has been submitted to the NRC for reiiew. This report and the ongoing pro-grams associated with it are described in greater detail in the December 15, 1980 letter from the Westinghouse Owners Group to the NRC.

l O

v I-15 02/06/81 l

f PROCEDURES FOR VERIFYING CORRECT PERF0FE WCE OF OPERATING ACTIVITIES

Reference:

Action Plan I.C.6 Operating and periodic test procedures that require valve movement in safety-related systems have been reviewed and revised as necessary to provide assurance that these valves are returned to their correct position. These procedures require verification of the operability of a redundant system prior to the removal of any safety-related system from service, verification of the operability of all safety-related systems when they are returned to service, and notification of and action by the Shif t Supervisor and reactor operators whenever any safety-related system is removed from or returned to service. Formal checklists are used to provide assurance that all valves in these safety-related systems are properly aligned. These procedures will be further revised to require indepen-dent verification of proper valve align =ent. These revisions.will be complete by March 1,1981 or prior to Mode 4 operation of McGuire Unit 1.

A removal and restoration procedure governs the repositioning of valves in safety-related syste=s following maintenance activities or other non-normal activities which require valve movement. A formal checklist provides assurance that all safety-related valves are properly aligned following these activities. This procedure will be revised to require independent verification of proper valve

/~' alignment. This procedure revision will be complete by March 1, 19S1 or prior

\s_-}- to Mode 4 operation of McGuire Unit 1.

Notification of and action by the Shift Supervisor and reactor operators when-ever any safety-related system is removed from or returned to service is acco=plished by the use of the operating and periodic test procedure checklists, red tags and the red tag logbook, white tags and the white tag logbook, out of service stickers, and the 1.47 bypass panel. Log entries denoting the removal and restoration are =ade in the Reactor Operator's Log. All of the above docu=ents are reviewed during shift turnovers.

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k 02/06/81

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1 AUXILIARY FEEDWATER INITIATION AND INDICATION I

References:

NUREG-0578 - 2.1.7a and 2.1.7b Action Plan - II.E.1.2 Automatic Initiation Safety-grade automatic initiation and safety-grade emergency power for the Auxiliary Feedwater System (AFS) are features of the McGuire Nuclear Station design (reference FSAR Ch. 7 and 10).

The automatic initiation circuitry for the AFS meets the single failure criteria.

Additionally, for most failures which could prevent the automatic start of an individual auxiliary feedwater pumo, manual initiation of the affected pump is available from the control room. However, should the auxiliary feedwater pump in one safety train not be available due to any single failure, the redundant safety train is available with no loss of system function.

In the final stages of plant shutdown, the main feedwater pumps must be tripped.

Therefore, the automatic start of the motor-driven auxiliary feedwater pumps upon trip of both main feedwater pumps or stean generator low-low level must be bypassed. This bypass is accomplished manually by means of a bypass switch located in the control room. When the bypass is instated a light is energized on the bypass control switet. Additionally, status light indication of the by-

, pass is provided on the associated status light panel. This bypass is adminis-tratively controlled by use of operating procedures.

l When the plant is in the startup mode, station procedures require that the bypass l

of the above motor-driven auxiliary feedwater pump start signals be removed. In I addition to the station procedures, an automatic means to remove the bypass has been provided. This automatic bypass tamoval will be generated when the P-il set point is reached. The P-ll set point is derived from Reactor Coolant System (RCS) pressure (N 1950 psig) and is the same sigual used to unFlock safety injection actuation. The P-ll set point is considered the appropriate signal to automatically remove the bypass of the above motor-driven auxiliary feedwater pump start signals because the reactor is not brought critical until RCS operating temperature and pressure conditions have been reached.

The turbine-driven auxiliary feedwater pump does not have a bypass feature.

Indication Safety grade indication of auxiliary feedwater flow to each steam ger.erator has been provided in the McGuire control room. Provisions for cali'o ration and testing were incorporated into the design of this instrumentation.

Control grade flow instrumentation in the lines to each steam generator'and in the suction piping to each auxiliary feedwater pump is'also provfded.

This control grade flow instrumentation is powered from the highly rea.able

, battery-backed 120 VAC Auxiliary Control Power System (reference FSAR Ch. 8).

Provisions for calibration and testing are included in the design of this control grade flow instrumentation.

II-3 . Q2/06/81 l

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AUXILIARY FEEDWATER SYSTEM RELIABILITY EVALUATION

Reference:

Action Plan - II.E.1.1 An evaluation of the McGuire auxiliary feedwater system has been performed by Duke and Westinghouse. This evaluation consists of the following items:

1. a simplified auxiliary feedwater system reliability analysis that uses event-tree and fault-tree logic techniques to deter-mine the potential for AFWS failure following a main feedwater transient, with particular emphasis on potential failures resulting from human errors, common causes, single point vulnerability, and test and maintenance outage;

2. a determination of the extent to which the McGuire auxiliary feedwater system meets each requirement in Standard Review Plan 10.4.9 and Branch Technical Position ASB-10-1; and

3. a determination of the design basis for the McGuire auxiliary feedwater system flow requirements and verification that these requirements are met.

Mr. W. O. Parker's letter of August 13, 1980 to Mr. H. R. Denton transmitted this evaluation. It revealed that no modifications to the McGuire auxiliary feedwater system are necessary. A discussion of the Bulletin and Orders Task Force recommendations on auxiliary feedwater systems as they apply to McGuire was transmitted to Mr. H. R. Denton via Mr. W. O. Parker's letter of September 18, 1980.

Additional analyses of the reliability of the McGuire auxiliary feedwater system were performed at the request of the NRC. These analyses show that for the four additional cases analyzed, assuming credit for operator action or using more realistic assumptions, the reliability of the McGuire auxiliary feedwater system is in the "high" category. These reanalyses in conjunction with the previous analyses provide the basis for Duke's conclusion that the current design of this system is acceptable. The results of these reanalyses were transmitted to the NRC via Mr. William O. Parker's letter of February 4, 1981.

i II-4 02/06/81

1 I

The majority of containment fluid line penetrations are closed f theupon follow-receipt of the phase A containment isolation signal which is derived from one o ing parameters:

a. low steamline pressure

b. low pressurizer pressure

' c. high containment pressure (1 psig)

d. manual The remaining penetrations are isolated upon receipt of the phase B containme isolation signal which is derived from the following parameters:

j

a. Containment high-high pressure signal (3 psig)

b. Manual containment spray actuation signal.

The penetrations which are isolated on the phase 3 signal provide the servic listed below:

a.

Reactor Coolant Pump Motor Cooling Water Supplies

' This service is maintained to prevent possible motor damage following events, such as main steam break outside containment, or small isolable LOCA's, which may not result in containment press or even necessary.

b. Reactor Coolant Pump Thermal Barrier Cooling Water Supplv 2

The thermal barriers act as a backup to the seal' If sealwater flow isinjection flow whi cools the main seals of the reactor coolant pumps. d further interrupted for any reason, the thermal barrier, which is ible down on the pump shaft locate (

reactor seal damage.

coolant which will flow up through the seals, thus p B isolation is 3

to maintain thermal barrier cooling flow until the phaserece cited in a.above.

c.

Containment Ventilation Unit Cooling Water gop_lyl, This service is maintained.uncil the phase B signal is received i t tofor prevent containment pressure rise to the Containment Spray actuation set po n t inmen4 events, such as small isolable LOCA's, main steam breaks doutside f thecon a and blackout which would otherwise result in unnecessary spray own o containment.

d. Main Steam Isolation valve Closure Closure of the main steam isolation /alves resultsftin loss of sink, loss of normal pressure control, and actuation of main steam sa e yin o valves, with associated system transients. l i closures, these valves are closed upon receipt of the phase B iso at on g ) signal.

02/06/81 II-5A ,

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i e.

Instrument Air Supply to Containment Isolation of instrument air to Containment causes to the pressurizer power operated relief valves, spray valves, con- i loss of control air tainment ventilation unit cooling water control valves, and var ousIt is d in l

other valve controllers.

these valves following many postulated events, such as those cite a.,

which result in phase A containment isolation, but which may not result in containment pressure increasing to the high-high (3 psig)

' phase B isolation set point.

h The McGuire contal_ ment purge valves 23,satisfy 1979 asthe operability discussed in the criteria set fort in the NRC Staff Interim Position of October McGuire FSAR, response to NRC question 042.63.

The McGuire containment purge and vent isolation valves Gi are section FSAR, closed autom upon receipt of a high radiation signal as indicated in the Mc u re 6.2.4.1, page 6.2-148.

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O 02/06/81 II-5B

REACTOR COOLANT SYSTEM VENTS

Reference:

Action Plan - II.B.1 Duke has installed a reactor vessel head high point vent that is remotely operable from the McGuire control room. A one-inch line has been added to the existing reactor vessel manual vent line with the connection located before the first isolation valve. The new vent line contains two parallel flow paths with redundant fail closed solenoid valves in each flow path.

The valves have been designed to pass non-condensible gases, water, steam, and mixtures thereof. Under normal operation these valves are deenergized.

Valve position is indicated in the control room. Train A emergency power serves both isolation valves in one flow path, and Train B emergency power serves both isolation valves in the parallel flow path. A flow limiting orifice has been installed in the common line downstream of the isolation valves.

The McGuire Reactor Coolant System (RCS) vent system is safety grade, seismically qualified, meets the requirements of IEEE 279-1971, and satisfies the single failure criteria. This vent system is operable.

As mentioned above, this head vent system has been designed to single failure criteria. If any single failure prevents a venting operation through one flow path the second flow path is available for venting the RCS. The two

-} isolation valves in each flow path provide a single failure method of isolat-ing RCS venting.

Inadvertent actuation of RCS venting is limited by the use of fail closed solenoid isolation valves. In addition the use of an orifice in the commen line downstream of the valves limits the flow to less than the make up capacity of the RCS charging pumps.

Exhaust from the vent system is directed to the Pressurizer Relief Tank (PRT) and therefore will not impinge upon vital equipment. A path is available to allow venting hydrogen from the reactor vessel head to the Waste Gas System storage tanks, via the pressurizer relief tank, should such an option be judged desirable. The PRT is located in the lower containment which is vent. lated and cooled by four air handling units. In addition, the hydrogen skimmer system has ducts in the lower containment high points to disperse any accumulated hydrogen.

Assuming that 100% hydrogen is being vented from the reactor vessel head, the vent system flow rate is 690 cfm. This allows the venting of a gas volume of one-half the RCS in approximately ten minutes.

The power-operated relief valves (PORV) are us?d to vent the RCS pressurizer.

The PORV's are discussed in the McGuire FSAR Section 5.2.2. The RCS vent is located at the top of the reactor vessel head which is the high point of the reactor vessel and coolant loops. This system in conjunction with the f-ss PORV's provides a venting capability for the entire RCS with the exception of the U-tube steam generators.

()

II-8 02/06/81

p A postu-lated break of the reactor vessel head vent line upstream of the flow Q limiting orifice would result in a small LOCA of not greater than one inch diameter. Such a break is similar to those analyzed in WCAP-9600 for hot leg breaks or pressurizer vapor space breaks. Extensive discussions were provided in WCAP-9600 regarding the applicability of the break flow models employed as well as other specific modeling features employed for small LOCA analysis. System response for this postulated break location would closely parallel that described in Section 3.2 of WCAP-9600. Since the break location in the head vent line would behave similarly to the hot leg break case pre-sented as Case F of that section, the discussion presented in WCAP-9600 for that study applies to this postulated break. As such, this postulated break would result in no calculated core uncovery.

The Westinghouse Owners' Group has approved a program to develop appropriate instructions regarding the conditions under which reactor vessel head vent operations should be conducted, and the manner in which head vent operations would be made. This program is scheduled to be completed by the second quarter of 1981.

It should be noted that the McGuire Nuclear Station already meets the requirements of 10CFR 50.46 and 10CFR 50.44 for design basis accidents.

The head vent system and its operation will not compromise this ability.

The instructions developed as part of the above referenced Westinghouse Owners' Group effort will be developed specifically to enhance the safe recovery of the plant for events beyond the design basis, accounting for combustible gas limit considerations and natural circulation.

A flow diagram of the McGuire RCS including the RCS vent system is provided in the McGuire FSAR Figures 5.1-1 and 5.1-2.

O II-8A 02/06/81

INADEQUATE CORE COOLING INSTRUMENTS

References:

NUREG-0578 - 2.1.3b Action Plan - II.F.2 Subcooling Monitor The margin to saturation is calculated from Reactor Coolant System (RCS) pressure and temperature measurements (wide-range and low-range pressures, wide-range hot leg temperatures, and temperatures from in-core thermocouples).

When RCS pressure is below 800 psig wide-range and low-range pressure inputs are compared, and if the inputs agree within 20 psig the low range pressure inputs are used. The wide range pressure inputs are used for the remaining conditions. The in-core thermocouple readings (65) are averaged and compared with the four wide-range hot leg temperatures (RTD's). The highest of these temperatures and the appropriate pressure are then used to calculate a con-servative margin to saturation. Averaging of the thermocouple readings and calculation of margin to saturation are performed by the plant computer.

The computer output consists of a CRT graphic display of conservative margin to saturation conditions, that is, a plot of plant pressurn and temperature in relation to a computer generated saturation curve. In addition, the follow-ing numerical values are displayed: each RCS hot leg temperature, RCS pressure,

[' ' power level, margin to Psat, each RCS loop margin to Tsat, thermocouple average margin to Psat, and the minimum allowable margins to Psat' and Tsat. Alarm status is indicated by flashing the alarming parameter on the CRT graphic display, the Alarm CRT, and by printout on the Alarm Typer. Two alarm set-points are provided for both Tsatand Psat. The alarm setpoints are dependent on reactor power. Further details on this subcooling monitor are provided in the table which follows.

Normal control board instrumentation for RCS temperature and pressure will be used in conjunction with a control room copy of the steam tables and a written procedure to determine margin to saturation as a backup to the computer calculation.

This system for determining the degree of subcooling is fully operational.

I Reactor Vessel Level Measurement l Duke will install the Westinghouse designed reactor vessel level measurement system in McGuire Unit 1. This system is designed to monitor the l water level in the reactor vessel, or the approximate void content under t

forced circulation conditions, during certain postulated accident conditions.

Included is equipment to monitor both the upper plenum (head) level, as well as the entire height of the reactor vessel.

O b II-9 02/06/81 I

The system instrumentation permits vessel level measurement from the bottom to g

the top of the reactor vessel, utilizing taps off of an existing spare head penetration and a tap off of a thimble tube at the seal table. Two sets of differential pressure transmitters are provided which have differing measure-ment ranges to cover dif ferent flow behavior with and without pump operations.

The narrow range cells indicate water level when zero or one reactor coolant pump is operating. The wide range cells indicate the combined core and internals pressure drop for any combination of operating reactor coolant pumps.

The upper plenum measurement is taken by two differential pressure transmitters between the same spare head penetration, and taps off two hot legs.

To minimize containment post-accident environment effects in measurement accuracy, the system design is based upon locating the transmitters outside the containment. Hydraulic isolators in the impulse lines provide the required double barrier protection between the RCS and outside containment. Reference leg temperature measurements, together with the existing RCS temperature and pressure, are utilized to automatically compensate for difference in coolant and reference leg temperature effects.

A more detailed description of this system was submitted to the NRC by Mr. W. O. Parker's letter of January 16, 1981. Installation and functional testing is scheduled to be completed by January 1,1982.

O V

O II-lo 02/06/81

ADDITIONAL ACCIDENT MONITORING INSTRUMENTATION

References:

NUREG-0578 - 2.1.8b Action Plan - II.F.1 Noble Gas Monitors Vent monitors for noble gases will be provided with a range adequate to cover both normal and postulated accident conditions. The present gas monitors at McGuire cover the range of 10~7 uCi/ccto10}3installednoble uCi/cc. A gross gamma detector will be added to these monitors to extend the range up to 10+5 uCi/cc. This detector will be attached to the outside of the unit vent and shielded to minimize count rate contribution from other possible sources. The detector will be sensitive to the 80 Kev energy range of noble gases and will have a minimum of one decade overlap with the existing noble gas monitor. If an event were to occur to cause the activity being released to be in the range of this additional detector, the noble gas monitor sample will be isolated.

This action will prevent the noble gas monitor from becoming contaminated and rendering erroneous indications when activity starts decreasing.

The additional detectors are scheduled to be installed by March 1, 1981.

Procedures for estimating noble gas release rates if the existing instrumenta-tion goes off scale have been written to cover the interim period between fuel loading and installation of the new detectors. These procedures require the use of portable high range surv'ey instruments to measure the radiation levels

/)

( on lines going to the radiation monitors for the unit vents if the radiation levels are such that personnel exposure could exceed 3 rem /qtr whole body and 18 3/4 rem /qtr to the extremities in the collection of a sample. The contact dose rate (mR/hr) on the lines is used to estimate the concentration (uCi/cc)

. of gas in the line. If the radiation levels do not exceed the above personnel exposure limits the procedures require collection of gas, particulate, and radioiodine samples. Silver zeolite cartridges are used for radioiodine samp-ling when noble gas interference is expected. Counting rooms in the Auxiliary Building and the Technical Training Center, located just outside the McGuire exclusion boundary, are available for sample analysis. In addition, the whole body counting room in the Administrative Building can be made available for sample analysis.

The present radiation monitoring system provides detection of volatile and non-volatile radioactive contamination of the secondary. A condensor air ejector monitor continuously monitors gaseous activity released to the unit vent by the condensor air ejector exhaust. A steam generator sample monitor continuously monitors non-volatile activity in all steam generators. An alarm on either of these monitors provides control room operators with an indication of steam generator tube failure. By cycling steam generator samples individually through the steam generator sample monitor control room operators can identify and if desired isolate the affected steam generator. The condensor air ejector monitor would quantify the level of radioactivity released to the environment prior to isolation of the affected steam generator.

O II-12 02/06/81

i To quantify the level of radioactivity released in the event the af fected stea= generator is not isolated and the atmospheric steam dump valves open.

Duke has a steam radiation monitoring system under design. This system will use an area radiation monitor (GA Model RD-1A) mounted near each steam line as it exits the reactor building. Continuous display of the monitor readout will be provided in the control room. A strip chart recorder will also be provided.

The radioactivity range covered will be 10-2 to 103 R/hr. These monitors will be calibrated every refueling outage using a 10 mci source. In addition an electronic calibration will be performed biannually. This system is scheduled to be installed by January 1, 1982.

Open-closed indication of the atmospheric steam du=p valves is provided in the control room. Procedures will be written to use this indication in conjunction with the design steam flow per valve to estimate the total steam mass released during a dump.

The containment hydrogen purge exhaust discharges through the unit vent and is monitored by the unit vent radiation monitors.

Containment High Range Radiation Monitors Two physically and electrically separated radiation =enitors will be installed inside the McGuire containment. These monitors will be supplied by General Atomics and will feature GA detector model number RD23. Each monitor will utilize an ionizatiop chamber to measure gamma radiation and will cover the range from 10 0 to IT R/hr. No overlapping of ranges is required Monitor to 62 Kev is 9.8X10-12 Amps / Rad /hr and the sensivity to 52 Kev

) sensitivit[2 is 9.0X10- Amps / Rad /hr. Seismic qualification of the monitor is in accord-ance with IEEE344-1975 and environmental qualification is per IEEE323-1971.

One =enitor will be powered from the Train A vital instrument bus, and the other =enitor will be powered from the Train B vital instrument bus. Analog meters (one per train) will continuously indicate monitor output in the control room. A continuous strip chart recorder (one train) will also be located in the control room.

An electronic calibration of the monitors will be performed every refueling outage. In addition a radiation source will be used to perform an in-situ calibration of the monitor range below 10 R/hr.

The monitors will be mounted on the primary shield wall at an elevation of at least 750+2 (10 feet or more above the maximum post-LOCA water level of 0 and 180 in the lower containment). The folleving McGuire General Arrangement drawings show the plan and sectional views with the monitor locations drawn in.

The monitors are scheduled to be installed by March 1,1981.

i Containment Pressure Continuous indication of containment pressure will be provided in the control room. Measurement and indication range will extend from -5 psig to 60 psig.

Each of the redundant differential pressure transmitters will be located in an electrical penetration room and will be equipped with one-half inch tubing impulse lines. Each impulse line will have a fail-closed isolation valve located in the O annulus. These valves will be normally open and will have position indication 02/06/81 II-13

and manual control in the control room. Continuous indication from each trans-O mitter will be provided in the control room. In addition, one channel of con-tainment pressure will be recorded. These instruments will be completely independent of the existing containment pressure transmitters and are scheduled to be installed by March 1,1981.

Containment Water Level Two containment floor and equipment sumps are provided on the floor of the lower containment (El 725') to collect floor drains and .quipment drains.

However, these sumps and their associated pumps and instrumentation serve no safety function.

The containment emergency recirculation sump at McGuire encompasses the entire floor of the lower containment. The two ECCS recirculation lines take suction just inside the Containment wall at elevation 725' and are oriented horizontally.

They are not located in the bottom of a recess or sump in the floor. Redundant safety grade level instrumentation is provided to measure emergency recirculation su=p level. The range of this instrumentation is 0-20 feet (El 725' to El 745')

which is equivalent to a lower containment volume of approximately 1,000,000 gallons. The accuracy of this instrumentation is 10% over the full range.

The redundant differential pressure transmitters utilized in this instrumentation have been relocated to the annulus where a filled capillary system eennects its associated transmitter with bellous sensors located inside containment.

Continuous indication from each transmitter is provided in the control room.

In addition, one channel of containment water level is recorded.

N Containment Hydrogen Monitoring Continuous indication of hydrogen concentration in the containment atmosphere will be provided in the control room. This hydrogen monitoring system will consist of two redundant Comsip, Inc./Delphi Systems Division K-111 analyzer systems with a range of 0 to 30% hydrogen by volume. These analyzers operate independent of the recombiner system and will be powered from redundant Class 1E power supplies. Each analyzer will have its own containment sample and return lines, and will be able to monitor either of two identical containment sampling headers or the calibration gases. Each analyzer will have a local control panel indicator and alarm and a separate control room indicator and alarm. In addition, one channel of containment hydrogen concentration will be recorded.

Each containment sample header will have five inlet samples available for monitoring.

1. Top of containment

2. Operating level

3. Basement 4 Radiation Monitor /Recombiner Inlet header

5. Radiation Monitor /Recombiner Discharge header All sample selection and switching is accomplished manually by the operator from the local analyzer control panel.

This instrumentation is scheduled to be installed during the second quarter of 1981.

II-13A 02/06/81

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II-las 07/18/80 Carry Ovet

POST-ACCIDENT SAMPLING

References:

NUREG-0578 - 2.1.8a Action Plan - II.B.3 l

A new sampling panel has been designed to allow analysis of reactor coolant i samples under accident conditions. A new sampling line will be connected to the present sampling line and routed directly to the sampling panel. In order to minimize personnel access limitations, this routing will be in accordance with the findings of tTe plant shielding review. This system will allow collection of reactor coolant samples under both pressurized and zero pressure conditions at any 1(vel of coolant activity. The design of this system signi-ficantly reduces radiation exposures during sample collection under accident conditions.

In addition to the reactor coolant sample line, a centainment atmosphere sample line will be routed to a new accident level sampling panel. The containment atmosphere sample will be obtained from the hydrogen analyzer sample lines.

This new post-accident sampling system is scheduled to be installed before January 1, 1982. A revised sampling program for the interim period between fuel loading and completion of the new system has been developed.

Procedures for collection and transport of reactor coolant, sump water, and containment air samples under post-accident conditions have been revised to

/N incorporate actions to be taken to minimize radiation exposures. These proce-k, s dures specify the preplanning to be performed as well as modifications and approvals required prior to sample collection. Samples can be collected within one hour in all instances whare personnel exposure does not exceed 3 rem /qtr whole body and 18 3/4 rem /qtr to the extremities. If the predicted personnel exposures exceed the above, samples can still be collected and analyzed but not within one hour. The time required to install additional shielding and allow sample collection while minimizing personnel exposure will be dependent upon the nature of the event. The analytical procedures have been reviewed and determined to be adequate for the expected sample activity levels.

Os_ ,/ II-14 02/06/81

- Plant systems or portions of syste=s which might contain significant levels of radioactivity as a result of a Design Basis Accident were selected for the O- station accessibility review. Included in the review were:

1) those portions of the Residual Heat Removal, Reactor Building Spray, Safety Injection, and Chemical and Volume Control Systems which could be aligned for recirculation of water from the contain=ent su=p to the Reactor Coolant System,

2) those portions of the Liquid Waste Recycle System which would collect and store leakage from the systems mentioned in item no. 1,

3) those portions of the Nuclear Sampling System which would be used in determining radiation levels inside containment or those systems mentioned above, and

!*) those portions of the Chemical and Volume Control System which supply seal water to the Reactor Coolant Pump seals, and

5) the Waste Gas System and those portions of the Chemical and Volume Control Syste= which could be used to degas the primary coolant.

To aid in identifying potential personnel access problems, the station was di% ;ed into post-LOCA radiation zones. Included in the radiation zones were all areas necessary for personnel access in controlling and mitigating a possible accident.

Two types of area access have been designated: 1) continual occupancy and

2) inf requent occupancy.

Areas designated as requiring continual access are: 1) Control Room, 2) On-site Technical Support Center (TSC), 3) Onsite Operational Support Center (OSC),

and 4) Personnel Access Portal (PAP). The Control Room, TSC, and OSC are located in the Contrrl Complex. The PAP is located in the Administration Building and serves as the main personnel access point to the station proper. The Control Room serves as the initial onsite center of emergency control and is designed to evaluate, control, and respond to various accident conditions. A detailed description of Control Room design and functions is presented in Section 7.8 of the McGuire FSAR.

Areas requiring infrequent access are generally located in Auxiliary Building corridors. Two exceptions to this are: 1) Containment Hydrogen Recombiner controls and 2) Emergency Diesel Generators. Redundant hydrogen recombiners are located in the Upper Containment. Power control panels for these recombiners are located in the Electrical Penetration Rooms. Redundant diesel generators are located in a section of the Auxiliary Building designated as the Diesel Generator Area. Various control functions associated with the diesel generators and supporting systems are located in the Diesel Generator Area. Typical func-tions centered in Auxiliary Building corridors are: 1) station radioactive waste control panels, 2) motor control centers, and 3) instrumentation panels for various station systems.

m II-16 09/03/80

() The major emphasis of the McGuire Nuclear Station plant enielding review was to assure that station personnel would be able to carry out their emergency procedures. The review featured the consistent use of the defined source terms in conjunction with the KAP-6 computer code. Listed below are all areas where radiation problems may exist and their preser.t status.

1. Sample Room - Unit No. 1 New sampling panels have been designed to r.11ow enalysis of reactor coolant and containment atmosphere samples under accident conditions.

However, samples from shared process systems are taken from the Unit I sample room. A large cable tray penetration is located in the wall separating the sample room from the Unit 1 penetration area. The locations of the cable tray opening and recirculation piping within the penetration area allow significant radiation steaming into the sample room. Using the radiation sources discussed earlier, the sample room will be inaccessible for slightly more than one week. However, the following samples have been identified as needing to be analyzed before one week:

1. Recycle Holdup Tank

2. Waste Evaporator Holdup Tank

3. Waste Drain Tank

4. Boron Recycle Evaporator Condensate Demineralizer Outlet

5. Waste Evaporetor Condensate Demineralizer Outlet

() Due to the complexity of the geometries involved, shielding of the Unit 1 sample rocm is not a viable solution. Modifications which would allow analysis of these samples in the Unit 2 sampling room could not be completed until mid 1042.

An alternate method for collecting these samples through tell-tale drain lines has been established. A review of tell-tale valve locations has shown that all valves are located in the Auxiliary Building corridors and are therefore accessiole. Procedures for collecting these samples have been written and no design modifications are necessary.

2. Floor Drain Tank Room The RHR sump pumps discharge to the Floor Drain Tank. Manual valves associated with isolating the Floor Drain Tank, and directing its contents elsewhere for storage or processing, are located in this room.

As a result of the location and manual operation of these valves, per-sonnel exposures could exceed GDC 19. Reach rods have been added to these valves.

Duke Power Company is currently conducting a thorough review of the environ-mental qualifications of Class lE equipment at McGuire. This review includes possible radiation environments resulting from the source terms assumed in the plant shielding review.

O G

II-16A 02/06/61

i l

COMPARISON OF THE MCGUIRE O CONTROL ROOM DESIGN WITH ATTACHMENT 1 TO NUREG 0737 Item III.D.3.4 Paragraph Reference / Comment 1 FSAR Section 6.4.1 2a FSAR Section 6.4.1 2b FSAR Section 6.4.1 2c FSAR Section 6.4.2-1, 2, 3 2d FSAR Section 6.4.3 2e Tech. Spec. 3/4.7.6 2f FSAR Section 6.4.2 FSAR Section 1.2 2g FSAR Section 1.2 FSAR Section 6.4.3 2h FSAR Section 12.0 21 FSAR Section 6.4.2 See Summary Description 2j See Summary Description 2k See Summary Description 21 See Summary Description 2m No emergency food or potable water supply is available.

2n Normal - 15 persons Emergency - 20 persons 2o KI for 150 people for 13 days 3a FSAR Section 2.2 See Sunsnary Description 3b See Summary Description O

II-17C 02/06/81 i

. ,-.,..,..,_.-__.,,_,..,....,_-,_,,.,....,_.,_.__.,,,m._

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

r i

Paragraph Reference /Coment j 4a FSAR Section 2.2 4b FSAR Section 2.2 4c FSAR Section 2.2 4d FSAR Section 2.2 Sa No info. in Tech. Spec.

Sb Tec. Spec. 3/4.7.6 l

l i

l II-17D 02/06/81

IE BULLETINS ON MEASURES TO MITIGATE SMALL-BREAK

[ LOCA'S AND LOSS OF FEEDWATER ACCIDENTS

Reference:

Action Plan - II.K.1 C.I.5 During the planning and procedure development stage of tha integrated, Engineered Safety Features (ESF) test a complete review of all valves receiving a safety injection actuation signal and containment isolation signal was conducted. This review primarily evaluated the response time requirements for each of these valves. However, in order to verify the response times, valve positioning requirements were also reviewed. As a result of this review and the successful performance of the integrated ESF test, direct verification of correct valve positioning requirements and valve positions under ESF conditions was obtained.

Correct valve positioning requirements, valve positions, and valve response times are verified during the ESF test via the Operator Aid Computer (OAC).

The correctness of the OAC indication is verified through the use of operating procedures which require visual verification that the valve position indication in the control room and on the OAC is identical to the actual valve position.

These operating procedures are required to be performed on all ESF valves after any maintenance activities which could affect proper operationof the

() valve.

C.1.10 Operating and periodic test procedures that require valve movement in safety-related systems have been reviewed and revised as necessary to provide assurance that these valves are returned to their correct position. These procedures require verification of the operability of a redundant system prior to the removal of any safety-related system from service, verification of the operability of all safety-related systems when they are returned to service, and notification of and action by the Shift Supervisor and reactor operators whenever any safety-related system is removed from or returned to service.

Formal checklists are used to provide assurance that all valves in these safety-related systems are properly aligned. These procedures will be further revised to require independent verification of proper valve aligcment. These revisions will be complete by March 1, 1981 or prior to Mode 4 operation of McGuire Unit 1.

A removal and restoration procedure governs the repositioning of valves in safety-related systems following maintenance activities or other non-normal activities which require valve movement. A formal checklist provides assurance that all safety-related valves are properly aligned following these activities.

This prccrJure will be revised to require independent verification of proper valve alignment. This procedure revision will be complete by March 1, 1981 or prior to Mode 4 operation of McGuire Unit 1. _ _ ___

b)

V II-18 02/06/01

1 Notification of and action by the Shift Supervisor and reactor operators when-ever any safety-related system is removed from or returned to service is accomplished by the use of the operating and periodic test procedure checklists, the removal and restoration procedure checklists, red tags and the red tag log-book, white tags and the white tag logbook, out of service stickers, and the 1.47 bypass panel. Log entries denoting the removal and restoration are made in the Reactor Operator's Log. All of the above documents are reviewed during shift turnovers.

The McGuire 'n'ork Request Program governs all maintenance activities performed at McGuire. These work requests describe the maintenance to be performed and the procedures for performing it. Upon completion of the maintenance all work requests are entered into the plant computer. This program provides for a historical record of all maintenance perforr _ on safety-related systems.

C.1.17 The design of McGuire Nuclear Station does not feature safety injection initiation on coincident pressurizer level and pressure signals. Therefore, no modifications are necessary to assure that safety injection is initiated whenever the low pressurizer pressure trip setpoint is reached.

w J

l II-18A 02/06/81

. . - -- - _ _ . =. = _ _ _ - -

O FINAL RECCMMENDATIONS OF THE BULLETINS AND ORDERS TASK FORCE V

Reference:

Action Plan - II.K.3 C.3.1 and C a.2 The Westinghouse Owners Group is in the process of developing a report which will provide historical PORV failure rate data and document the various actions that have been taken to reduce the probability of a small-break LOCA caused by a stuck-open PORV. This report will provide a basis for a decision on the necessity of incorporating an automatic PORV isolation system. This report is scheduled to be submitted to the NRC by March 1, 1981.

C.3.3 Duke Power Company will promptly report to the NRC any failure of a McGuire PORV or safety valve to close. In addition, all challenges to the PORV's or safety valves will be documented and reported to the NRC.

C.3.5 The Westinghouse Owners Grc p has performed analyses, using the Westinghouse small break evaluation model (WFLASH), which show ample time is available for the operator to trip the reactor coolant pumps following certain size small Os break LOCA's (see WCAP-9584). In addition the Westinghouse owners Group is supporting a best estimate study, using the NOTRUNP computer code, to demon-strate that tripping the reactor coolant pump at the worst trip time after a ,

small break will lead to acceptable results.

For both of these analysis efforts, the Westinghouse owners Group is performing blind post-test predictions of LOFT experiment L3-6. The input data and model to be used with the WFLASH on LOFT L3-6 was subsitted to the NRC on December 1, 1980 (NS-TMA-2348). The information to be used with NOTRUMP cu LOFT L3-6 will be submitted to the NRC prior to performance of the L3-6 test as stated in the Westinghouse Owners Group letter dated December ', 1980 (letter OG-45).

The LOFT prediction from both models will be submitted to the NRC on February 15, 1981 given that the test is performed on schedule. The best estimate study is scheduled for completion by April 1,1981.

Based on these studies, the Westinghouse Owners Group believes that resolution of the automatic reactor coolant pump trip issue will be achieved without any design mc difications. In the event that this is not the case, a schedule will be provided for potential modifications.

C.3.9 Westinghouse has completed its review of the pressure integral derivative (PID) controller installed on the McGuire PORVs. WCAP 8921, the NSSS Control System Setpoint Study, gives a value of "zero" fo,r the pressurizer PID controller rate

[~)

\s / time constant. The McGuire time constant has been adjusted accordingly.

II-19 02/06/81

C. 3. lf, and C.3.12 The design of McGuire Nuclear Station does not feature a reactor trip on turbine trip. This trip was removed from the McGuire design to prevent unnecessary reactor trips, particularly during initial startup. Unnecessary reactor trips should be avoided to minimize reactor coolant system thermal cycles and challenges to the reactor coolant system protective devices. The removal of this anticipatory trip was possible due to the full load rejection capability of McGuire.

The McGuire trip system keeps surveillance on process variables which are directly related to equipment mechanical limitations, suci. as pressure, pressurizer water level (to prevent water discharge through safety valves) and also on variables which directly affect the heat transfer capability of tne reactor (e.g., flow, reactor coolant temperatures). Still other parameters utilized in the reactor trip system are calculated from various process variables. In any event, whenever a direct process or calculated variable exceeds a setpoint, the reactor will be shut down in order to protect against either gross damage to fuel cladding or loss of system integrity which could lead to release of radioactive fission products into the Containment.

An analysis was conducted to determine the potential for pressurizer PORV challenges following a turbine trip from full power both with and without an ic=ediate reactor trip on trubine trip. This analysis considered both normal plant respense and cases assuming the failure of certain central systems that can influence challenges to the pressurizer PORV's. Two types of control system failures were considered: failure of all steam dump valves to open on demand (not O' including the stea= generator PORV's); and complete failure of pressurizer spray to function on demand. Partial failures (for example, failure of half of the steam dump valves) were not considered.

The analysis demonstrated that if all of the steam dump valves failed to open the pressurizer PORV's would be challenged regardless of the presence or absence of an immediate reactor trip on turbine trip at full power. If there was no failure of the steam dump valves the absence of the subject trip would result in challenges to the pressurizer PORV's whereas the presence of such a trip would not challenge the PORV's.

Installation of a direct reactor trip on turbine trip would only protect against PORV challenges initiated by a narrow range of events, that is turbine trips not initiated by a reactor trip or a safety injection and occurring at or near full power. Furthermore, valves identical to the McGuire PORV's and PORV block valves have been subjected to extensive steam flow testing. This testing was conducted at Duke's Marshall Steam Station in conjunction with the EPRI valve testing program. The testing demonstrates that the McGuire PORV's and PORV block valves meet all functional and design requirements and provides added assurance of proper PORV and PORV block valve - ation.

Duke will install a direct reactor trip on turbine trip to provide this additional protection against PORV challenges. This trip will be installed prior to power escalation of McGuire Unit 1.

The reactor trip on turbine trip will be generated by either of the following O signals:

e Four-out-of-four turbine stop valves closed e Two-out-of-three turbine auto-stop oil pressure low II-19A 02/06/81

The four turbine stop valve signals will be developed through the actuation of independent limit switches mountad on the stop valve assemblies. Each of the four turbine stop valve signals can de tested individually from the control room through the digital electrohydraulic (DEH) control panel. The turbine auto-stop oil system is the medium through which a turbine trip is initiated. Turbine auto-stop oil pressure is measured by three independent pressure switches which are mounted in a terminal box located adjacent to the turbine. The pressure switches can be tested from the control room by applying power to three solenoid valves located in the auto-stop oil system. The test circuit is designed to allow one pressure switch test per test actuation.

The limit switches and pressure switches used in this application are similar to those used in other Class 1E applications in the plant. Although the main turbina-generator is not seismic Category I, these limit switches and pressure switches are seismically qualified and the associated cables will be installed in accordance with the McGuire separation criteria.

Each turbine stop valve limit switch and each turbine auto-stop oil pressure switch provide an input to both trains of the solid state protection system (SSPS). If either logic function as described above is satisfied, a reactor trip signal will be generated provided reactor power is greater than approximately 48% (P8). A logic diagram for the reactor trip on turbine trip is provided on the following figure.

It should be noted that the reactor trip on turbine trip was originally part of the McGuire SSPS design with the exception that the P8 interlock will be substituted for the P7 interlock. It is therefore concluded that reinstituting this trip will not degrade the existing protection system since all separation, testing, and reliability considerations are in acccrdance with the original SSPS design.

C.3.30 and C.3.31 Westinghouse feels very strongly that the small break LOCA analysis model currently approved by the NRC for use on Westinghouse designed NSSS is conservative and in conformance with Appendix K of 10CFR 550.46. It is noted that the NRC has not indicated the contrary. However, Westinghouse believes that improvement in the realism of small break calculations is a worthwhile effort. Whenever possible and whenever allowed by Appendix K, more realistic modeling assumptions will be employed. This will hopefully provide more con-sistency among analyses performed for licensing, training and procedure writing.

If the results of the new Westinghouse Model (and subsequent NRC review and approval) indicate that the present small-break LOCA analysis for McGuire are not in conformance with 10CFR 550.46, a new analysis utilizing the new and approved Westinghouse model will be submitted to the NRC.

II-19B 02/06/81

I  !

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Reactor Trip on Turbine Trip I j l

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r TURBINE TURBINE

STOP VALVE POSITION AUTO-STOP OIL PRESSURE i t

LS LS LS LS PS PS 4

PS j  ;

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l A y Ll. _Ld..

4/4 2/3 l

l I 4 a OR

P8

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REACTOR TRIP NOTE: SHOWN FOR ONE TRAIN ONLY; other train is similar l

O 02/06/81 l II-19C s

COMMISSION ORDERS ON BABCOCK A'O WILCOX PLAhTS O

Reference:

Action Plan - II.K.2 C.2.13 A program which completely addresses the NRC requirements of detailed analysis of the thermal-mechanical conditions in the reactor vessel during recovery fro = s=all breaks with an extended loss of all feedvater vill be completed, documented, and submitted to the NRC by January 1,1982. This program vill censist of analyses for generic Westinghouse PWR plant groupings. If required, additional plant specific analyses will be provided following completion of the generic program. A schedule for the plant specific analyses will be determined based on the results of the generic analyses.

C.2.17 The Westinghouse Owners Group is currently addressing the potential for void for=ation in the Reactor Coolant System (RCS) during natural circulation cooldown conditions (as described in Westinghouse's letter NS-TMA-2298). The results of this effort vill fully address the NRC requirement for analysis to deter =ine the potential for voiding in the RCS during anticipated transients.

A report describing the results of this effort will be provided to the NRC by January 1, 1982.

C.2.19 II.K.2.19 - Sequential Auxiliary Feedvater Flow Analysis The transient analysis code, LOFTRAN, and the present small break evaluations analysis code, kTLASH, have both undergone benchmarking against plant informa-tion or experi= ental test facilities. These codes under appropriate condi-tions have also been compared with each other. The Westinghouse Owners Group will provide a report addressing the benchmarking of these codes by January 1, 1982.

l

\ 02/06/81 II-20

._ _ __ . _ . . _ . .__ _ _ _ . _ _ . _ _ _ . _ _ . _ _ ~ _ - - . . . .

UPGRADED DiERGENCY PREPAREDNESS V

Reference:

Action Plan - III.A.1.1 The documents, " Emergency Plan for McGuire Nuclear Station," and " Duke Power Company Crisis Management Plan," describe the actions to be taken in the event

. of a radiological accident where the health and safety of station personnel and the general public may be involved. The station document was reviewed by the NRC and was found to meet the requirements of Appendix E to 10CFR Part 50 (McGuire SER, NUREG-0422) . The State of Forth Carolina, the counties of Mecklenburg, Lincoln, Gaston, Iredell and Catawba, and the State of South Carolina have also developed plans for coping with radiological emergencies.

Duke Power Company's emergency plan for McGuire was revised per NUREG-0654 and submitted to the NRC in a March 20, 1980 letter from Mr. *J. O. Parker to Mr. R. L. Baer. An NRC emergency planning review team visited McGuire on June 16 and 17, 1980 and reviewed the McGuire emergency plan and selected McGuire emergency procedures. Duke evaluated the review team's report and revised the McGuire emergency plan appropriately. On August 25, 1980 the McGuire Nuclear Station Emergency Plan was submitted to the NRC for formal review against NUREG-0654. R. L. Tedesco issued the comments from this review to Mr. William O. Parker, Jr. on November 14, 1980. Duke Power Company will reply to these comments, submit revision 1 to the McGuire Emergency Plan,

<"'s and submit the McGuire Nuclear Station Emergency Plan Implementing Procedures for NRC review by February 16, 1981. Duke's Crisis Management Plan, the state's emergency plan, and the counties' emergency plans were submitted for NRC review on January 7, 1981.

The McGuire Emergency Management Response Exercise took place on December 5 and 6, 1980. NRC and FEMA representatives observed the exercise and, with minor changes to the plans, indicated they would, in their preliminary review, recommend approval of onsite and offsite planning and capabilities.

s III-1 02/06/81

ON-SITE TECHNICAi. SUPPORT CENTER

References:

NUREG-0578 - 2.2.2b Action Plan - III.A.I.2a Duke Power Company has established an on-site Technical Support Center (TSC) at McGuire Nuclear Station to serve both units in an area on the same eleva-tion and 40' south of the control room. Capabilities within this area include:

a. Ready access to as-built plant drawings including general arrangement drawings, flow diagrams, electrical and instrument drawings.

b. A CRT having the capability to access, print and/or display plant para-meters independent from control room actions.

c. Habitability to the same degree as the control room for postulated accident conditions. This included the installation of an iodine filter system.

d. Co::cunications between the control room, the off-sit Grisis Management Center (CMC), the state and county Emergency Operation Centers (EOC),

and the NRC via dedicated telephone circuits. Backup radio communications exist between the TSC, the CMC, the control room, the county EOC's, the EOC in Raleigh (via Lincoln County), and the State Emergency Response O Team (SERT).

e. Monitoring equipment which provides local readout of radiation level and alarms if preset radiation levels are reached. Portable radiation  ;

survey instruments are also available in the Technical Support Center.

The Technical Support Center includes approximately 1200 square feet of space.

l It is in close proximity to the Control Room and has similar environmental control features. The Technical Support Center is established and will meet all applicable requirements.

Staffing of the Technical Support Center includes the Station Manager, Super-intendent of Operations, Superintendent of Technical Services, Superintendent of Maintenance, Superintendent of Administration, Station Health Physicist and staff personnel necessary to support them.

I 1

O v III-2 02/06/81

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

1

'\ IN-PLANT RADIATION MONITORING

References:

NUREG-0578 - 2.1.8e Action Plan - III.D.3.3 Portable air samplers with silver zeolite radioiodine sampling cartridges are used at McGuire for sampling air when the presence of noble gases is suspected. McGuire Health Physics personnel are knowledgeable in the appropriate station procedures and are trained in the equipment required to determine airborne iodine concentrations in the plant under all conditions.

Counting rooms in the Auxiliary Building and the Technical Training Center, located just outside the McGuire exclusion boundary, are available for radioiodine analysis. Included in these counting rooms are shielded GeLi detectors and shielded sample storage areas. In addition the whole body counting room in the Administrative Building can be made available for radioiodine analysis by moving in and setting up a shielded GeLi detector with a culti-channel analyzer and a shielded storage area.

A precedure to determine airborne radiciodine concentrations has Wen established which does not rely on the availability of a counting room. This procedure utilize; portable " survey-type" instrumentation with energy discrimination for iodine to determine a "go" or "no go" iodine concentration for respiratory

) equipment use. The use of silver zeolite radioiodine sampling cartridges minimizes the amount of Xenon interference. The "go" or "no go" count rates are dependent upon the calibration of the instrument used. The results of this analysis will be available within ten minutes. This instrumentation in conjunc-tion with the portable air samplers is a fully adequate method to moniter iodine in-plant and meets the November 19, 1979 " clarifications" to NUREG 0578, Item 2.1.8c.

To reduce counting system saturation, sample sizes will be varied to minimize counting system problems. In addition, nitrogen purging of the counting room GeLi detector shields can be used to reduce airborne activity interferences.

l i

l l

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v III-4 02/06/81 l

t I

PRIMARY COOLANT SOURCES OUTSIDE CONTAINMENT

]

References:

NUREG-0578 - 2.1.6a Action Plan - III.D.l.1 Periodic leak rate test procedures have been written for systems carrying radioactive fluids outside of centainment. These procedures are provided in Appendix C. The following systems are included: Safety Injection, Residual Heat Removal, Containment Spray, Nuclear Sampling, Boron Recycle, Chemical Volume and Control, Refueling Water, Liquid Waste, and Waste Gas.

These tests, to be performed before startup and during each refueling outage, will be accomplished by pressurizing a system or part of a system and checking non-welded pipe joints, penetrations, flanges, valve separations, packing, and pump packing for leakage. Where possible, pumps included in the leak test boundary will be run so that a more accurate determination of the leak rate may be made.

A separate periodic test procedure has been revised to assure that excessive leakage is detected on a timely basis. This procedure will be provided in Appendix C. This test will be run at least weekly and will require that systems carrying radioactive fluids outside of containment be visually in-spected for excessive leakage. Appropriate corrective action will be taken if excessive leakage is detected.

O O

III-5 02/06/81

STATION p!RrCTIY4 3.1.4

,  ? ,., y APPROVAL f' 8'

/m \

DATE Original issued 11/30/79 REVISION 2 _ DATE // 7 //

DUKE POWER COMPANY McGUIRE NUCLEAR STAfl0N CONDUCT OF OPERATIONS O BJECTI'7E The objective of this direc:ive is to o.ucline the respon,1bilities, authority, handling of special orders and rules of practice for licensed operators at Mc-Guire Nuclear Station. This directive also outlines control roo= authority, succession control, access procedure and on-call professional and supervisory personnel require =ents.

IM?LE'iENTATICN C'

Outies of licensed operators and licensed senior operators described in this directive are derived fro = require =ents in 10CFR55, ANSI NIS.1-1971, McGuire FSAR See:1cn 13, McGuire Technical Specifications Section 6, Regulatory Guide 1.11!. , :.'JREC-0578 and ANSI N18.7-1976.

F.ESFCNSI3ILITY A.C AUTHCRITY SUCESSICN A. Station Manager The Station Manager repor:s to the Manager, Nuclear Production and has direct responsibility for operating the station in a safe, reliable and efficient manner. He is responsible for protection of the station staff and the gen-eral public fro = radiation exposure and/or any other consequences of an .

t

( acciden: at the station. He bears the responsibility for conpliance with the facility operating license.

3. Operating Superintendent The Opera:ing Superintendent has the responsibility for directing the actual day-to-day operation of the station. In the event of the absence of the l

Station Manager, the Opera:ing Superintenden:, if so designated, assu=es the p respons bilities and au:hority of the Station Manager.

) C. Operating Engineer The Operating Engineer assistr, the Operating Superintendent in directing I

__ . - _ _ _ _ _ - - _ _ - _ - - _ _ _ _ _ _ _ . - . - - _ _ ~ _ _ - .-

1.

station operation and say assume complete responsibility for the actual day-to-day operation of the station in the absence of the Operating Superintendent.

> He also serves as "On Call" Duty Engineer for Operations.

3. Assistant Operating Engineer The Assistant Operating Engineer assists the Operating Engineer in directing station operation and nay assume complete responsibility for the Operating Engineer in his absence. He also serves as "On Call" Duty Engineer.

"On Call" Duty Engineer An Operations "On Call" Duty Engineer is responsible for directing actual shift and station activities. He will act as a liasion between Operations and other l groups at the station and is available for coordinating and assisting the Shift Supervisor during accident or emergency conditions. He shall hold a current Senior Reactor Operators License.

E. Shif: Supervisor .

A Shif: Supervisor is responsible for the actual operation of the station on his assigned shif t. He has the responsibility to be cogni: ant of all opera-s tional condi: ions affecting the safety of the planc as a matter of highest priority when on duty.

A Shif t Supe: visor directs the activities of the operators on his assigned shift and m st be cognizant of all maintenance activities affecting plant operation beias perfor:ed while he is on duty.

• The Shif t Supervisor on duty has both the authority and the obligation to shutdcwn a uni: if, in his opinion, conditions warran: this action. During accident or e=ergency condi: ions, the Shift Supervisor is responsible for directing activities of control room personnel. He shall remain in the cen-trol roo= at all times during :hese situations and should not become totally involved in any single operation when multiple operations are required during emergencies.

The Shif: Supervisor can only be relieved by another Shif t Supervisor or by a encer of =anagemen: that holds a cur ent Senior Peactor Operators license.

Outside of the nor=al working hours of the station staff, the shift supervisor shall ac: in the position of the Station Manager in matters that concern the safe and efficient operation of the staff.

O

I l

s t F. Assistant Shif t Supervisor An Assistant Shif t Supervisor assists the Shif t Supervisor in operation of the station on his assigned shift. The Assistan: Shif t Supervisor on duty has both the authori:y and the obligation to shut down a unit if, in his I opinion, condi: ions warran: this action. The Assistant Shif t Supervisor i assu=es the responsibilities and authority of the Shif t Supervisor, as de- t

! fined for the Shif t Supervisor, if so designated.

G. Nuclear Control Operator

. A licensed Nuclear Control Operator is responsible for the actual operation of a uni: on his assigned shift. The licensed Nuclear Control Operator has

! authority and obligation to shut d,own a uni: if, in his opinien, conditions varrant this action. He may also serve as " Operator at the Controls" per Stat *.on Direc:ive 3.1.17, if so designated.

H. Assistant Nuclear Control Operator 3

A licenJed Assistan: Nuclear Control Operator is responsible for the actual operation of a unit on his assigned shift as direc:ed by the licensed Nuclear ,

Control Operator. He has both the authori:y and obligation to shut down a l unit if, in his opinion, condi: ions warrant this ac: ion, He may also serve as " Opera:cr at the Con:rols", if so designated, per Station Directive 3.1.17.

LICENSED OPE?ATOR DUTIES

• 1. A licensed Nuclear Con:rol Operator or Assistant Nuclear Con:rol Operator shall i act as " opera:or at the controls" in accordance wi h Station Directive 3.1.17 except in si:uations where a licensed senior operator elec:s to assume these duties.

2. The " operator at :he controls" shall continue in his duties until relirved by a qualified o vera:or in accordance with Station Directive 3.1.19.

3. " Opera:or a: :he controls" duties shall not be delegated to non-licensed 1

persens. Howeier, a non-licensed person =ay =anipulate the controls of the f acility under :he direction and in :he presence of x licensed operator

, or licensed senior opera:or as a part of his training to qualify for an operator license under 1CCTR55.

4 The " operator at the con:rols" shall =oni:or instrumentation displays and alarms to assure safe opera:Ing conditions for his assigned unit.

NOTE: All station personnel performing functions which =4y affect uni:

operation or control roon indications are required to notify the

" operator at :he controls" prior to initiating such action in accordance with McCuire Station Directive 3.1.5.

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5. The " operator at the controls" shall knew and comply with limits and set-points associated with safety-related equipment and systems as specifiled in Technical Specifications and designated in the Operating Procedures.

He shall review routine operating data in order to assure safe operation of his assigned unit. All licensed reactor operators shall participate in the requalifica:ica progras outlined in McGuire FSAR Sectica 13.2.2.

6. The " operator a: the controls" shall be responsible for the manipulation of centrols which directly or indirectly affec: core reactivity. He shall cporate or direct others to operate a:her equipment associated with his assigned unit. He shall direct the activities of persons assigned under him in the operation of this assigned unit. These operations pertain to startup opertion, power operation, shutdown operation, or during a shut-down condicion.

7. The "operater at the centrols" shall acknowledge all alar =s. He shall notify licensed senior operators on duty of unexpected alarms or alarms of unknown cause. He shall initiate prompt corrective action on receipt of any indication (instrument movement or alarm) of an irregular cperating

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\- conditica in accordance with applicable Operating, Alarm, or Emergency Procedures. This corrective action =ay include tripping the reactor should, in his best judgoment, a situation e=1st requiring prompt action and the licensed senior opera:crs on duty are not available for consultation.

The "cperator at :he centrols" shall verify that appropriate automatic action has taken place in case of an alar =. 'a' hen autcmatic action is determined to have failed or to have been ineffective, he shall =anually initiate this action.

8. The " operator at the controls" shall notify the licensed senior operators en duty of all abnor=al operating condi: ions. The " operator at the con:rols" shall notify the area dispatcher of conditiens which could significantly affect station 1 cad and other information which the area dispatcher may request.

9. The " operator a: the centrols" may stop fuel handling operations if, in his best judge:ent, centrol rocs indication or ce=munications show warrant-ing conditions.

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10. ~4 hen two or more licensed operators are assigned to the sa:e unit, one shall act as " operator at the controls" while the others assist in all

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the above duties.

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11. The "cperator at :he controls" shall =aintain the Reactor Operator's Log-back in accordance wi:h Statien Directive 3.1.1-

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12. The " operator at the controls" can authorize the removal of any instrunent or components from service in accordance with Station Directive 3.1.5 or 3.1.19.

13. An operator without recent operational experience shall receive a one (1) week refresher course with a documented oral exam prior to assuming the " operator at the centrols" position.

14 A licensed operator say be assigned to the reactor building operating deck to observe core alterations.

LICENSED SENIOR OPERATOR DUTIES

1. Shif t Supervisors and Assistant Shif t Supervisors responsible for the units shall be licensed senior operators and shall only be relieved by =anagement personnel who hold Senior Reactor Operators license.

2. The Shift Supervisor on duty shall oversee station operations. The Assist-

. ant Shift Supervisor on duty shall oversee operations associated with his assigned unit. Persons in these positions while on duty fullfill the on-site require =ents for licensed senior operators.

\ 3. All licensed senior operators within their area of authori:y and while on b duty shall keep the=selves informed of operating status and shall direct the licensed activi:1es of licensed operators. All licensed senior oper-ators shall within their areas of authority and while on duty organize, direct, and control activities to insure safe, efficient operation in conpliance with ad=inistrative and technical requirenents for operation.

When required a licensed senior operator may perform any duty of a licensed operator.

. The Shift or Assistant Shift Supervisor can authorize the removal of any instru=ent or co=ponents from service in accordance with Station Direct-ive 3.1.5 or 3.1.19.

5. The Shift or Aasistant Shift Supervisor shall maintain the tait Supervisor's Logbook in accordance with Station Directive 3.1.13.

6. An Assistant Shift Supervisor will provide administrative assistance on shift to assist in ti=e tickets, scheduling days off and other associated duties.

7. A Senior Reactor Operator who has no other conct rent responsibilities will be assigned to fuel handling during core alterations. He or another ifcensed operator (R0 or SRO) must be on the reactor building operating deck to observe core alterations.

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d SHIFT COMPOSITION The 21nimu= shift crew composition shall consist of the following positions as required:

A. Shif t Supervisor - One required to be onsite whenever either reactor contains fuel. He must hold a current senior reactor operator (SRO) license on the I

applicable unit (s). In operating modes 1-4 he may relieve the SRO in the Control Room for either or both units. He may be designated to the fire brigade if not designated as the Control Room SRO (see Item G) .

3. Shif t Technical Advisor - One required for to be onsite for each unit operating in modes 1-4 He =ust hold a current senior reactor operator (SRO) license on the applicable unie. See Station Directive 3.1.31 (Duties, Responsibilities, and Qualifications of the Shif t Technical Advisor) for Shift Technical Advisor qualifications.

C. Senior Reactor Operator - One required for each unit operating in modes 1-4.

He must hold a current senior reactor operator (SRO) license on the applica-ble unit (s). He may serve as the SRO in the Control Room. He may be desig-O nated to the fire brigade if not designated as the Control Room SRO (see Item G).

D. (Assistant) Nuclear Control Operator - One required for each unit with fuel

. in the reactor. He must hold a current reactor operator (RO) license on the applicable unit. The R0 =ay relieve the Senior Reactor Operator in the Centrol Room while both units are in modes 5 and 6. In addition, when a unit is operating in modes 1-4, and additional RO for that unit shall be onsite.

and available to serve as relief operator for that unit. The second RO may be assigned to the fire brigade (see Iten G) .

E. Nuclear Equipment Operator - One required for each unit with fuel in the reactor. In addition, when a unit is operating in modes 1-4, an additional nuclear equipment operator (NED) shall be assigned to that unit. The NEO(s) may be assigned to the fire brigade (see Item G).

F. Additionally, three NE0's shall be onsite to be available to the fire brigade.

G. A site fire brigada of 5 me=bers shall be saintained onsite at all times.

A maximus of two (2) of the individuals specified in A, C, D, and E may be as-

) signed to the fire brigade.

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H. For relief of each of the positions described in A thorugh F, a individual of equal or better qualifications is required.

I. The =ini=um shift. crew composition for each unit as specified in items 3, C, D, and E =ay be one (1) less 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 />.

1 This is to acccmodate unexpected absence of on-duty shift crew members pro-vided i==ediate action is taken to restore the required minimum shift crew composition as described above.

J. In operating modes 1-4, one of the individuals specified in A and C must be designated as the Control Room Senior Reactor Operator (SRO). He must be in the Control Rocs at all times except when relieved by an SRO authorized in A and C. .

HOURS OF WCRK A nor:al. tour of duty is eight (S) hours at an assigned duty station. Overtime

=ay be used for unanticipated or unavoidable circumstances. Overtime shall not be routinely scheduled to compensate for an inadequate nu=ber of personnel to IA =eet the shift crew staffing requirements.

Licensed operators shall:*

A. Not work more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> straight.

3. Not work more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period.

C. Not work ore than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any 7 day period.

D. Not werk more than la consecutive days without having 2 consecutive days off.

• Cvertime limits do not include shift turnover etne.

ADJUSTING OF CONTROL RCOM ANNUNCIATOR SOUND LEVEL The audible ~ sound level of all control room annunciators shall be maintianed be-tween 7 and 9 decibels above background sound levels. No adjusement to these levels shall be =ade by the operators. Any adjustment by the Instrument and Electrical Depart =ent must be approved by the Superintendent of Operations.

AN'!U' CIATOR LA.#P REPLACE}ENT

Anytime annunciator la=p replacement is required in an area of the Control Room

where an operator must stand on top of control boards, two (2) operators shall

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(/3) be present. In the event a control is inadvertantly actuated by the operator changing the lamp, the second operator shall take Lacediate action to remedy the situation.

ON CALL OF ?ROFESSIONAL AND SUPERVISORY PERSONNEL Each group superintendent shall ensure that the Operations Shif t Supervisor is supplied with an updated "on call" list for professional and supervisory per-sonnel in their respective group. This "on call" list shall include na=e, tele-phone numbers and beeper nc=bers of these "on call" individuals.

CONTROL ROCM ACCESS Access to the area deemed " Surveillance Area", per Station Directive 3.1.17 (Definition of "At the Controls") shall not be permitted by non-licensed per-sonnel, while fuel is in either reactor, without permission having been given to enter such area by the on-duty licensed " operator at the controls" or the 7-~g senior licensed unit supervisor.

S?ECIAL ORDERS

• Manage =ent written instructions, other than procedures, Station Directives, etc.,

which encompass special operations, housekeeping, personnel actions or other st:ilar instructions shall be kept in an accessable document for Operations per-sonnel.

This docu=ent shall be reviewed on a se=1-annual basis for applicability of in-structions.

Updates to the docu=ent shall be placed in this document, after review by appli-cable personnel, for future reference.

Any cancellation of a vritten order shall be removed from this document and cec-

=unciated to applicable personnel.

USAGE AND TESTING OF EMERCENCY NOTIFICATION SYSTEM (NRC PHONE)

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Any event listed in Attachment i shall be reported as soon as time permits and in all cases within one hour to the NRC Operation Center via the Emergency No-tification System (ENS) - NRC Phone. In the event that the (ENS) is not oper-

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fi V able, the alternate notification methods, in the order to be used, are the NRC regional co=nercial telephone (404-221-4503), and the Health Physics Network (EPN) dedicated line. RPN phone is the orange phone located in the Technical Support Center.

The event shall be reported to the NRC stating that the report is being made oursuant to ICCFR50.72(a). This notification should be made by the Operations Shif t Supervisor or the Operations Unit Supervisor. Documentation of the no-tification should be made in the Uni: Supervisor's log. Prior to notifying the NRC, co=alete Attachment 2, Par: A of this directive with available informa-tion. Attachment 2 is similar to the form that the NRC duty officer will be using when ae is notified.

Any event reported pursuant to this directive should also be reported to the Licensing Engineer, the Projects and Licensing Group during normal work hours or to the ?rojects and Licensing " duty person" if the report is made outside of regular working hours.

The Projects and Licensing Croup is responsible for notifying the NRC resident q_, inspector as soon as practical.

NRC ENS PHONE *ESTING

- Daily, between the hcurs of 0400 and 0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> the NRC will call the station

o obtain the status of each unit. This call will also be used as a check of the dedicated phone line. This call should be answered by either the Operations Shift Supervisor or an Operations Uni: Supervisor. A brief operating status of each unit should be provided.

in addition, the NRC may request that we hang up and return the call to verify operability with the station as the initiator.

Occasionally, a check of each extension at the facility will also'be made. This check will be scheduled in advance by the NRC and should consist of verifying co==unicatica f rom each extension at the station.

If the ENS is found to be inoperable, notify the Duty Officer at the NRC Oper-ations Center at the telephone number given in Section " Usage and Testing of E=ergency Notification System (ENS Phone)".

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Page 1 of 2 J STAT!CN LIRECTIVE 3.1.4

- ATTACHMENT 1 Events Requiring I= mediate NRC Notification Via ENS Phone

• l. Any even: requiring initiation of the station e=ergency plan or any section of that plan.

• 2. The exceeding of any Technical Specifications safety limit.

• 3. Any situation whereby a reactor is not in a controlled or expected condition of operation. A situation such as :his is defined as any unscheduled event involving a reactor which cannot be stabilized by use of nor=al operating procedures or the follovup actions of existing E=ergency Procedures.

• 4 Any ac: that threatens the safety.of the statica, or station personnel, or the security of special nuclear =aterial, including instances of sabotage, and a:te=pted sabotage.

5. Any event requiring ini:iation of shutdown of a Uni: in accordance with Technical Specifica:icn li=1:ing conditions for operations.

6. Any event involving personnel error or procedural inadequancy which, during nor=al operations, anticipated operational occurrences, or accident condi: ions f-s prevents or could prevent, by itself, the fulfill:ent of the safety function (x-of : hose structures, syste=s, and co=ponents i=por: ant to safety : hat are needed to (a) shutdown the reactor safely and =aintain it in a safe shutdown condition, (b) re=ove residual heat following reactor shutdown, and (c) limit the release of radioac:ive =aterial to acceptable levels or reduce the potential for such release.

7. Any event resulting in =anual or auto =atic actuation of engineered safety fea:ures, including the Reactor Protective Syste=.

l S. Any accidental, unplanned, or uncontrolled radioactive release. (Normal or expected releases frc= =aintenance or other operation activities are not included).

o. Any fatality or serious injury occurring at the site and requiring transport to an off-site =edical facility for treat =ent.

10. Any serious personnel radioac:1ve conta=1 nation requiring extensive onsite l decenta=ination or outside assistance.

11. Any event involving licensed =sterial which =ay have caused or threatens to cause:

(1) Exposure of the whole body of an individual :o 5 re=s or more of radi-acion; exposure of the skin of the whole body of an individual to 30 re=s or = ore of radia: ion, or exposure of the feet, ankles, hands, or forear=s to 75 re=s or = ore of radiation; or l i

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Pags 2 of 2 STATION DIFICTIVE 3.1.4 ATTACH!" INT 1 (2) The release of radioactive =aterial in concentrations which if averaged over a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, would exceed 500 times the limits specified for such materials in 10CFR20, Appendix B, Table II; or (3) The loss of cne day or more of the operation of any facilitiec affected; or (4) Damage to property in excess of $2,000.

12. Strikes of operating employees or security guards, or honoring of picket lines by these employees.

• NOTE: klen an event is reported as a result of ite=s (1), (2), (3), or (4),

establish and =aintain an- open, continuous cc::munication channel with l

the NRC Operation Center and close the channel only when notified to do so by the NRC.

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l PAGE 1 OF 4 STATION DIRECTIVE 3.1.4 ATTACliMENT 2 SIGNIFICANI EVENT REPORT i

Date Ti=e Duty Officer's Na=e Facility Na=e Caller's Na=e A. Deter =ine:

1. Descriction of Event:

Reactor Svste=s Status:

Pressure Te=p. Pewer Level l

Flev ECCS Operating (Pu=ps On) Operaole PZR or RX Level Cooling Mede s

Anv Radioactive Release or Increased Release? Pa th Stopped? Release Rate Monitored Stea= Plant Status: S/G Levels Equip. Failures Feedvater Source /Flev i

Electrical Dist. Status: Nor=al Offsite Power Sources Available?

Major 3usses/ Loads Lost Safeguards 3usses Power Source D/G Running? Loaded?

Personnel Casualties /Contanination?

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- STATION DIRECTIVE 3.1.4 ATTACHMENT 2 l

2. Consecuences of Event:

i Actual and Potential Safety Hazards Tech. Spec. Violations?

State Notified? Press Release Planned?

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3. Cause of Event:

4 Licensee Corrective Actions: Taken Planned l

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3. Assign Severitr Level I II III

, Assigned Sy: NRC HQ f Contact IE Management if you have any doubt on Severity Level i

C. Level I or II imnediatelv Notifv IE Management and Region j

Persons Notified: HQ Region ,

Identify 50.72 Reporting Requirement Itam 1 2 3 4 Other t

D. Level III, Perforn the Following:

1. Identify 50.72 Reporting Requirenent I:en 1 2 3 4 Cther O

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STATION DIRECTIVE 3.1.4 ATTACHMENT 2

2. Determine if any of the following exist:

(a) You have concern the licensee is not taking or has not taken adequate corrective action.

(b) You consider that further i= mediate evaluation of the circum-stances of the event is required. Examples include:

(1) Reactor is not in a normal operating or normal shutdown condition.

(2) Any eme.rgency system is still in use (i.e. , HPCI, LPCI, contaia=ent isolated,, emergency diesels operating).

(3) The secondary system heat sink is not the condenser.

(4) In connection with the event, there is continuing radio-activity release from the site ot; a continuing leak of radioactivity frem any pipe or system onsite.

(c) You have any question or doubt on the significance of the event.

If one of the above exists, notify the applicable HQ and Regional

()9 q;- contacts and establish conference call communications as directed.

Otherwise, proceed to the next step.

. 3. Break Cc=sunication With the Caller (a) Inform the licensee that you have the infor=ation needed at this ti=e, you will break communication now, and you will establish ce==unication if additional information is needed.

(b) Request the licensee to report promptly as they occur, items such as:

Degradation in Plant Status New or Increased Releases / Releases Ter=inated Incident Compounded by Other Failures Plant in Nor:al/ Stable Condition 4 Promotly Obtain Regional Confirmation and Notify Ho Contacts Persons Notified:

(a) Region Confirmation of Severity Level Confir=ation of Ita=s 1-4

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Confi. ation of termination of ce==unication

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STATION DIRECTIVE 3.1.4 g ATTACHMENT 2 i

l (b) Headquarters Notify IE Management (1) I:enediately if the situation warrants, otherwise (2) At the end of each watch during nights, weekends, holidays.

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, -o STATIC RFCTIVE 3.8.2 APPROVAL /MMsb

_,) DATE ORIGINAL ISSUED 4/8/m0 REVISION 3 DATED ['v[6r> 1

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DUKE POWER CCMPANY McGUIRE NUCLEAR STATION STATION DfERGENCY ORGANIZATION CBJECTIVE This directive establishes the Station E=ergency Organi stion and the functions it is responsible for in eff ectively supporting :he nor=al operating shif t in the manage =ent of any e=ergency condition at the station. It particularly addresses the aug=entation of the operat,ing shif t resources for accident re-sponse situa ions where the health and safety of station personnel and me=bers of the general public are concerned. It provides a structure by which the nor=al func: ions of the operating shif: are augmented and i==ediately direc:ed to acciden: ter=ination and =itigation, offsite censequence determina: ion, and plant recovery operations.

3 x_ / , GENERAL Initial a::ivities during any e=ergency condicion are directed by the Shift Supervisor fro: the control roo=, in accordance with :he McGuire E=ergency Plan and any !=ple=enting Procedure. The Shif Supervisor shall assu=e the functions of the E=ergency Coordinator until the arrival of the Station Manager or his des-ignee at which time the Station Manager or his derignee assu=es the responsibility .

of the E=ergency Coordina:or. The Shif t Supervisor and the E=ergency Coordina:or will assure that the following e=ergency objectives are achieved during the initial phases of any e=ergency condition described in the appropriate E=ergency Procedure:

1. Early warning and clear instructions to the population-at-risk in the event of a serious emergency condition involving radio-active =aterials or the existing potential for a release of radioactive =aterials affecting station personnel or =e=bers of the general popula: ion.

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2. Notification and activation of the Station, Corporate, North Carolina, and the Nuclear Regulatory Commission I emergency organizations having a response role.

3. Continued assessment of actual or potential consequences i

both onsite and offsite throughout the evolution of the t

e=ergency condition.

4. Ef f ective imple=entation of emergency =easures in the t

environs including protective actions and or evacuation i

of affected areas, inplementation cf emergency monitoring teams and facilities to evaluate the environsental conse-quences of the emergency condi, tion, prompt notification and co=munications w1:h of fsite authorities.

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5. Continued r.aintenance of an adequate state of emergency pre-paredness until the emergency situation has been effectively

=anaged and the station is returned to a normal or safe operating condition.

% IMPLDIENTATION Cnsi:e E=ergency Organization 1 1. The Onsite Emergency Organization shall be : hat organication 4

of normal plant cperating shif t aug=ented with additional personnel as deemed necessary by the Shif t Supervisor /E=ergency

! Coordinator, the Station Manager or as required by any E=ergency Procedure or Station Directive.

2. The Shift Supervisor on duty shall ensure that all actions required by any initiating E=ergency Procedure or by any e=ergency condition have been performed and that all actions necessary for the protection of persons and property are being taken. The Shif t Supervisor upon being relieved of the E=er-gency Coordinator functions shall continue to take all actions necessary to ensure that any emergency situation is brought under control.

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

o 3. The E=ergency Coordinator shall have the authority and responsibility to initiate any emergency actions within :he provisions of the Emer-gency Plan, including the notifications and exchange of infor=ation with those authorities responsible for coordinating offsite emergency measures. The E=ergency Coordinator will work closely with the Shif:

Supervisor, other Station Managemen: and Engineering and Technical suppor personnel at the Technical Support Center (T.S.C.) (See Attachment 1) .

He shall also maintain communications with of fsite personnel at the Crisis Management Center, County Emergency Operations Center (s) and with the North Carolina State Emergency Operations Center initially, then with the North Carolina State Emergency Response Team headquarters as :his organization is activaEed. This function will later be assumed by the Cuke Power Crisis Management Plan.

4 The Control Room is the initial onsite center of emergency control.

It is designed for evaluation and control over :he ini:ial aspec:s s of an emergency and fer those actions necessary for coping with the s,,) e=ergency condition. These actions include but are not limi:ed to:

(a) Continuous evaluation of the magnitude and potential consequences of the e=ergency condition, (b) Initial notifi-2 ca: ions and co==unications with those station personnel and of f site agencies responsible for coordinating ef fective response

=easures for the e=ergency condition. The control room shall be

, staffed with one operating shift, the E=ergency Ccordinator and any other personnel the Shift Supervisor, Station Manager or E=ergency Coordinator may require in response to the emergency condition.

5. The Onsite Technical Suppor: Center (T.S.C) acts in support of

he coc=and and control function of the control room and to disolay -

current plan: status and diagnostic infor=ation to those individuals who are knowledgcable and responsible for engineering, technical, and =anage=en: support of reac:or operation in any emergency condition.

The Technical Support Center is located in offices 911, 912, 913, 914, and 921 in the service building at elevation 767 (See Attach =ent 3, Technical Suppor: Center Layout) and has the capability to house 15

'- - persons, necessary co==unication equipment, diagnostic display information, plant drawings, layouts, =aps, and char:s necessary to support the

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. s e=ergency organization. In the event the Technical Support Center located in offices 911, 912, 913, 914, and 921 becomes environmentally uninhabitable due to radiological or other conditions and the control room remains secure (habitable), Phase I of the Technical Support Center shall move to offices 930 and 931 inside the control room.

Phase II shall evacuate to the administration building, Cowan's Ford Hydro or the Technical Training Center as directed by the Station Manager. In the event the control room also becomes uninhabi-table due to radiological or other conditions, Phasa I of the Technical Support Center will evacuate to the administration building, Cowan's Ford Hydro or the. Technical Training Center as directed by'the Station Manager. -

The Technical Support Center shall be activated by the Emergency Coordinator in accordance with the applicable Emergency Procedure. The Control Room shall notify and activate the members of the Technical Support Center -

by notifying the E=ergency Coordinator and Superintendents in accordance with Attachment 4, who shall be responsible for notifying the personnel

/\ under their direction for implementation of Phase II of the Technical s-Support Center. The Section heads in Phase II shall be responsible for notifying the personnel under their direction assigned to the Technical Support Center and any other personnel that they may deem necessary to support the Emergency Condition.

Phase I of the Technical Support organization shall be operational in o0 =inutes and will be staffed and organi:ed as per Attachment 1, or as deemed necessary by the Station Manager.

NOTE: In the event that radiological emergency conditions exist, the Health Physics section of the T.S.C.

shall be activated with Phas' I of the Technical Support Center organizatica e,a deemed necessary by the Station Manager or the Superintendent of Technical Services.

A. Phase I of the Technical Support Center shall include but not be limited to thu following personnel:

(1) The Station Manager (E=ergency Coordinator) or in his-absence a designated alternate per Attachment 1. The Station Manager

-f s.) shall have complete responsibility for activation

U of :he Technical Support Center and the Corporate Crisis Manage =ent Plan. He shall staff the Technical Support Center with those personnel listed in Attachment 1 or at his descretion with those

! persennel deemed necessary to effectively assess :he

, e=ergency condition. He shall institute those procedures i

necessary to allow the control rocs to gain i==ediate control of the e=ergencv condition. The Station Manager will have direct cc=munications via telephone or radio with the Recovery Manager at the Crisis Manage en: Center, each county Emergency Operating Center, the North Carclina Sta:e Emergency Respo'nse Team and via telephone only :o Nuclear Regula:ory Commf.ssion. He shall saintain lines of cc==unication and censultation with these agencies to ensure that they are infor=ed of :he emergency ccndition at all times in accordance with the E=ergency Plan.

1 (2) The Suoerintendent of Operations when designated shall assu=e :he duties of the Station Manager. He will pro-G

- vide expertise :o the Station Ma' nager and the Shif t Supervisor regarding solutions :o operational problems.

He shall ensure that each operating shift is =anned wi:h competent personnel trained and prepared to sanage all operation emergency conditions and he shall augment his persennel resources as necessary to accomplish this goal. He shall provide technical expertise to other members of the Technical Support Center and shall work closely with the Superintendent of Maintenance in restoring station equipment to an operational status during and after the emergency condicion.

4 (3) The Suoerintendent of Technical Services when designa:ed shall assu=e the duties of :he Station Manager. He will provide expertise to the S:ation Manager and the Shift

Supervisor regarding solu

ions to operational proble=s.

He shall provide technical expertise to the other = embers of the Technical Support Center in the areas of Health p/

k._- Physics, Chemistry, Perfor=ance, and Reactor Engineering and in Licensing and Engineering support programs.

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N-- He shall ensure that all areas of responsibility under his direction are staffed with competent personnel pro-perly trained and prepared to support any operational emergency conditions.

(4) The Superintendent of Maintenance when designated shall assume the duties of :he Station Manager. He will provide expertise to the Station Manager and the Shif t Supe rvisor regarding solutions to operational proble=s. He shall provide technical expertise to the other members of the

!.S.C. in areas of Mechanical Maintenance, Planning, Ins:rument and Electrical Maintenance, and Materials Support. He will ensure that all areas of responsibility under his direction are staffed with competent personnel properly trained and prepared to support any opera:ional e=ergency conditions.

(5) The Suoerintendent of Administration when designated shall assume the duties of the Station Manager. He will provide Te:hnical Expertise to the Station Manager and the Shif t

( Supervisor regarding solutions to administrative probless associated wi:h emergency conditions a: the station. He shall provide technical expertise to other =e=bers of the Technical Supper: Center in the areas of Contract Services, Administrative Coordination and Training / Safety. He shall ensure that all areas under his directica are s affed and prepared to canage administrative support for any e=ergency condition.

3. Phase !! cf the Technical Support Center organization shall be-cpera:ional in 1-4 hours and will be staffed and organi:ed as per A::ach=en: 1., or as deemed necessary by the Station Manager.

In the event that radiological emergency condi: ions exist, :he Health Physics section of the T.S.C. shall be activated with phase I of the T.S.C. organization as deemed necessary by the Station Manager or the Superintendent of Technical Services.

Phase II of the Technical Support Center shall include as a mini =us the following personnel:

v

N O

O (1) The Operating Engineer shall assume the duties of the Superintendent of Operations when so designated. He will provide technical exper:ise to the Superintendent of Operations and other members of the Technical Support Center as required. He will assist the Superintendent of Operations in coordinating Operation activities during i

the E=ergency condition by developing work schedules, equipment and material procurement, guidance and assistance to the Shift Supervisor, co=munication with the Crisis Management Center incident report preparation, and other support functions as,needed or required to restore the plant status to normal. He shall ensure that all areas under his direction are staffed and prepared to manage operational support for any emergency condition.

(2) The Assistant Operating Engineer shall assume the duties of the Operating Engineer when so designated. He will "x provide technical expertise to the Superintendent of 3

f x,_ / Operations, the Operating Engineer and other = embers of the Technical Support Center as required. He shall assist the Operating Engineer in assessment and evaluation of the emergency condition and in any other areas of expertise dee=ed necessary to the Technical Support Center organiza: ion.

(3) The Health Physics section of the T.S.C. shall consist of the Station Health Physicist or his designated alternate, an Environmental Surveillance Coordinator, a Data Evaluation Specialist and a Radio Operator and other Health Physics personnel as dee=ed necessary by the Station Health Physicist e to support the Health Physics functions during the emergency condition.

NOTE: The Environmental Surveillance teams shall be predesignated in the Station Health Physics Manual.

The Station Health Physicist shall assume the duties of the Superintendent of Technical Services when so designated.

()

He will provide technical expertise to the Superintendent of Technical Services, the Station Manager, and other members of the Technical Support Center as required. The

\ \

l l Health Physics section shall be responsible for gathering 4

and compiling onsite and of fsite radiological monitoring

data from N.R.C., State, Corporate and Station radiolo-gical monitoring and evaluation teams and for providing this information to other members of the Technical Support Center

+

as required. The Station Health Physfcist shall provide for input to and distribution of Offsite Dose Calculations for l Airborne Releases (CDCAR) infor=ation accessable by Health Physics personnel. The Sca: ion Health Physicis shall make recommendations to the Station Manager through the Superin:endent of Technical Services on Protective Actions deemed necessary to ensure that station personnel and me=bers of the general public do not exceed exposure limits to radioactive

sterials. The Station Health Physicist shall also work clcsely with the appropriate = embers of the Corporate Crisis

. M.anage=ent Center to ensure : hat radiological hazards during q any emerger.cy condition are minimi:ed. The Station Health j Physicist shall ensure that all areas under his direction are

\~/ staffed and prepared to =anage health physics support for any

, emergency condition.

) (4) The Station Che:is: shall assume the duties of the Superintendent of Technical Services when so designated. He will provide

echnical expertise to the Superintendent of Technical Services

! and to other me=bers of the Technical Suppor: Center as i

! required. He is responsible for coordinating chemical technical a

support and for initiating necessary action to insure adequate chemical sampling and evaluation to support the emergency condi: ion.  ;

j The Station Che=1st shall ensure that all areas under his i direction are staffed and prepared to =anage chemistry support for any emergency condition.

(3) The Performance Engineer shall assume the duties of the Superintendent of Technical Services when so designated. He

]

will provide technical expertise to the Supe

Intendent of Technical Services and to other members of the Technical. Support i

Center as required. He will assure that adequate-levels of i technical and engineering =anpower are available to:

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d(~ manage test procedure review, carryout special test procedures, ensure control and accountability of special nuclear materials, and evaluate plant and reactor perfor-mance. The Performance Engineer shall ensure that all areas under his direction are staffed and prepared to manage Performance support for any emergency condition.

(6) The Reactor Enginee shall assume the duties of the Performance Engineer or the Superintendent of Technical Services when so designated. He will provide technical expertise to the Perfor-mance Engineer and to other members of the Technical Support Center as required. The Reactor Engineer shall ensure that all

~

areas under his direction are staffed and prepared to manage technical support for any emergency condition.

(7) The Projects and Licensing Engineer shall assu=e the duties of the Superintendent of Technical Services when so designated.

He will provide technical expertise to the Superintendent of g Technical Services and to other members of the Technical s_,) Support Center as required. He is responsible for coordinating

' station activities with regulating agencies, coordinating the reporting and investigation of all incidents and for providing review of appropriate Station technical matters. The Projects and Licensing Engineer shall ensure that all areas under his direction are staffed and prepared to =anage technical support for any emergency condition.

(8) The Instrument and Electrical Engineer shall assume the duties of the Superintendent of Maintenance when so designated. He will provide technical expertise to the Superintendent of Maintenance and to other members of the Technical Support

! Center as required. He is responsible for maintaining all 1

station I6E equipment in an operational state. The Instrument and Electrical Engineer shall ensure that all areas under his direction are staf fed and prepared to manage any _ISE support for any e=ergency condition.

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V (9) The Planning Engineer shall assume the duties of the r

Superintendent of Maintenance when so designated. He will provide technical expertise to the Superintendent of Maintenance and to other members of the Technical Support Center as required. He is responsible for the implementation and evaluation of the maintenance management progra= and for the administration of the =aterials procurement programs. The Planning Engineer shall ensure that all areas under his direction are staffed and prepared to manage planning and materials support for any emergency condition.

(10) The Mec'hanical Maintenance Engineer shall assume the duties of the SuperintendenIt of Maintenance when so designated. He will provide technical expertise to the Superintendent of Maintenance and to other members of the Technical Support Center as required. He is responsible for preventive and actual maintenance for all station mechanical equipment and facilities. The Mechanical Maintenance Engineer shall l,_sh

( ) ensure that all areas under his direction are staf fed and prepared to =anage =aintenance support for any e=ergency condition.

(11) The Contract Services Coordinator shall assume the duties of the Superintendent of Administration when so designated.

He will provide technical expertise to the Superintendent of Administration and to other members of the Technical l

i Support Center as required. He is responsible for coordi-nating Security, Utility Services and Food Vending Service's for the Station. The Contract Services Coordinator shall ensure that all areas under his direction are staffed and prepared to manage Contract Services for any emergency condition.

(12) The Administrative Coordinator shall assume the duties of the Superintendent of Administration when so designated.

She will provide technical expertise to the Superintendent of Administration and to other members of the Technical f Support Center as required. She is responsible for se coordinating and =aintaining general administrative

-11 functions and support for personnel at the station. The Administrative Coordinator shall ensure that all areas under her direction are staffed and prepared to manage administrative functions during any emergency condition.

(13) The Training and Safety Coordinator shall assume the du:les of the Superintendent of Administration when so designated.

He will provide technical expertise to the Superintendent of Administration and to other members of the Technical Suppor: Center as required. He is responsible for coordinating the station training and safety activities Fire Protection and Medical Services in support of the emergency organi:ation.

The Training and Safety Coordiaator shall ensure tha: all areas under his direttien are staffed and prepared to provide needed training and safety evaluations during any emergency condition.

6. The Onsi:e Operational Suppor: Center (0.S.C.) shall be Iccated in Rocs 909 (cperators' ki:chen). In the event the Onsite Operatienal Support Center becomes environmentally uninhabitable, personnel assigned shall scve to offices 930 and 931 inside the control room u

/ ventilation system. It shall be staffed with operators not asnigned to the control reos, and Health Physics persennel in support of the emergency condition as per Attachment 2. The nor=al statien telephone

system with a backup radio system (P&T frequency hand held) shall serve as a line of direct communication. The personnel assigned

o :he Operational Suppor: Center shall be under the direct super-vision of the Shift Supervisor.

The Operational Support Center shall be activated by the Station Manager in accordance with the applicable Emergency Procedure. The Shift Supervisor shall alert the on-shift operations personnel and the Health Physics Supervisor listed on At:achment 5 to :he Emergency conditien. The Health Physics Supervisor shall alert the Health Physics Technicians listed on Attachment 5 to the activation of the Onsite Operational Suppor: Center. The Operational Suppor: Center will be staffed and organized as per Attachment 2, or as deemed necessary by the Shif t Supervisor or Station Manager. The Operational g .Suppor: Center shall include as a mini =um the follcwing personnel:

1 g_,/ Operators: Operators on shift who are not actually assigned to the Cen:rol Room and additional call-out operators as required or deemed necessary by the Shif t Supervisor or -Station Manager.

,. s- I Health Physics: A Health Physics support group consisting of a Health Physics Supervisor and two technicians or additional tech-nicians as deemed necessary by the Station Health Physicist shall be designated to serve in a support function to the Operational  ;

Support Center. The Health Physics group supporting the Operational Support Center shall be physically located in the Health Physics laboratory or in the event this facility is evacuated they shall function from the Auxiliary Count Room in the Administration Building. The Health Physics Supervisor shall ensure that adequate instrumentation, respiratory protective equipment, protective clothing, and any other material needs are provided in support of the Health Physics coverage for the Operational Support Center.

The Shift Supervisor shall request Health Physics coverage thru the Health Physics Supervisor assigned to the Operational Support Center.

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- f STATION DIRECTIVE 3.1.9 APPROVAL Y DATE Original issued 8/26/76 REVISION 6 DATE /!/3 /

DUKE PCWER COMPANY McGUIRE NUCLEAR STATICN RELIEF AT DUTIES OF PLANT CPERATICN O3JECTIVE This directive specifies procedures to precote continuity of safety and efficiency during the process of relief of persons at duties of plan: operation.

DKPlEMENTATION

1. The Shift Supervisor, Assistan: Shif: Supervisor, Nuclear Centrol Operator, Assistan: Nuclear Control Operator, and Nuclear Equipment Operator shall review back to the point where last on duty, the statua of structures, sys-( te=s, equipment, and components under their cognizance prior to being re-11eved and shall insure all conditions are registered in logs and records for which they are responsible as required by Statien Directives. Out-of-normal condi:1cns shall be e=phasized in this review. They shall verbally pass cc their reliefs the results of this review as well as all available infor=atien as expected occurrences which will affect plant operation.

Additionally, Shift Supervisors shall jointly review the Shift Supervisor's Logbcok, the worklist and the periodic test schedule for the oncoming shift, and the Nuclear Control Room Operators shall jointly review the Reactor Operator's Logbook.

2. When the relieving process is completed, the persen relieving shall verbally declare to the person being relieved that the act of assuming assigned re-spensibili:ies is ccepleted. Persons shall not leave their assigned duties of plant operation until properly assured that their responsibilities have 1

< been assu=ed by their relief or until receiving verbal approval frem a Shift Supervisor or Assistan: Shift Supervisor.

ps) s 3. When the opera:ing shift is relieved, opera:ing areas will be toured ,. a r-v scnnel of the on-ccming shift as soon as practical after relieving to verify equipment cperating conditions. The Shift Supervisor or Assistan: Shift Supervisor shall inspect as necessary to determine any out-of-normal conditions.

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, , 4. The Nuclear Equipment Operator assigned to the Auxiliary Building shall fill j out just prior to turnover and f ollow the Nuclear Equipment Operator Turnover Checklist (Attachment No. 3) at shift turnover.

The Nuclear Control Operator shall run the Nuclear Control Operator Turnover Checklist (Attachment No. 2) computer program and complete at shift turnover. l In che event that the computer is unavailable at shift turnover, the Nuclear Control Operator Turnover Checklist (Attachment No. 2) shall be completed.

The Supervisors shall fill out and follow the Supervisor Turnover Checklist (Attach =ent No. 1) during shift turnover. The encoming Shifc Supervisor shall deter =ine the required mini =um shift composition for his shift. He shall then identify the individuars that will man the positions. No re-quired shift crew positions may be unmanned upon shift change due to an one==ing shift crewman being late or absent. See Station Directive 3.1.4 (Conduct of Operations) for minimum shift crew composition.

The relieving Shift Supervisor or Assistant Shift Supervisor shall review the turnover checkli ts (Attachmenta 1, 2, and 3) and route to the Duty Engineer d

• or his designee.

The Duty Engineer or his designee shall review these turnover checklists daily and place in the Master File. The turnover checklists shall remain in the Master File for a =inimum of six years.

A (a) aww-e --.-rw y --r-a i, w -- ,- -w w.- - , - - - r- e-+ - - - -- + -

d Station Directive 3.1.9 Attach =ent 1 Rev. 1

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.v Page 1 of 3 UNIT 1 SUPERVISOR TUP.SOVER CHECKLIST Shift being relieved: A B C D E (circle one)

Date/Ti=e __

I. NCS Leakage GPM Power Level Xenen PCM (INC / Dec) (circle one)

II. Review the general status of each"section of the Main Control Borads and note any abnormal conditions. The Nuclear Control Operator of the shift being re-lieved agrees to the status of each section.

Re= arks:

/~') Nuclear Control Operator Unit Supervisor Unit Supe rvisor

( ,) (Being Relieved) (Relieving)

Initials Initials Initials

.III. Review Unit Supervisor (Being Relieved) Unit Supervisor (Relieving)

1. Re= oval & Restoration Notebook

2. Cut of Nor=al Notebook

3. Red Tag Ecok 4 '4hite Tag Sock

5. Reactor Cperator Log-boek

6. Unit i Supervisor Log-book

7. Alarm Su==ary Print-out S. Stat"- of Monitor L;ghts

9. Unit 1 Sypass Panel s,

G

Station Directive 3.1.9 Attachment 1 Rev. 1 Page 2 of 3 r

IV. Minimum Shift Crew Composition Mode of Operation SHIFT CRE'J POSITION MODE REQUIRED FIRE BRIGADE

• NAME

1. Shift Supervisort 1-6

2. Shift Technical Ad-visor 1-4

3. Senior Reactor Operatort A. Unit 1 . 1-4

3. Unit 2 1-4

• 4 (Assistant) Nuclear Control Operator A. Unit 1 1-6 (1) 1-4 (2)

3. Unit 2 1-6 (1) p 1-4 (2)

5. Nuclear Equipment

\k # Cperator A. Uni: 1 1-6 (1) 1-4 (2)

3. Unit 2 1-6 (1) l-4 (2)

6. Nuclear Equipment Operator All / (1)

All / (2)

All / (3)

(4)

• A 5 me=ber fire brigade must be onsite at all times.

t In modes 1-4, one of these individuals must be designated as the Control Room Senior Reactor Operator and be located within the Control Room.

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Section .setive 3.1.9 Attachment 1 Rev. 4 Page 3 of 3 Remarks:

(Shift Being Relieved) (Relieving Shift)

Shift Supervisor E Asst. Shift Supv.

(Signature) (Signature)

Reviewed By:

(Duty Engineer) l  %'

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Staticn t ..ective 3.1.9 Attachment 2 s Rev. 4

) Page 1 of 5 e i NUCLEAR CONTROL OPERATOR TURNOVER CHECXLIST Shift Being Relieved: A B C D E (circle one)

Mode: 1 2 3 4 5 6 (circle one)

Date/Ti=e A. NCS Leakage __ GPM Power Level Xenen PCM (Inc / Dec) (c.ircle one)

3. Critical Parame;ers (if in Modes 1 or 2)

Tave (557-588) 'F NCS Pressure (2235115) psig P r. Level (25-61.5)

Q:

C. ICCS PLHPS CN/0FF CMIRCL POWER REQ'D IN THE FOLLOWING .M (I) CONTROL POWER 1 2 3 4 5 6 AVAILABLE (/)

LA CA X X X i .

lA ND X X X X* X*

1A NI X X X i 1A NV X X X X* X* X*

1A NS X X X X 1A RN X X X X 1A1 KC X X X X 1A2 KC X X X X 13 CA X X X 13 ND X X X X* X*

13 NI X X X 13 NV X X X X* X* X*

13 NS X X X X 13 P3 X X X X v 131 KC X X X X 132 KC X X X X i

• Control Power Required for only one train l

acatisa Direcs6*. 3.1.9 Attachment ;

Rev. 4 Page 2 of s J

< D. Valve Alignment 0 = Open X = Closed Req'd Position inIthe following Modes Actual 1 2 3 4 5 6 Position 1CA-7 Aux. FDWP No. 1 Suction Isol. 0 0 0 __

ICA-6 Aux. FDWP Suction From AFWCST Isol. 0 0 0 1CA-4 Aux. FDWP Suction from UST Hdr. Isol. 0 0 0 1CA-9 Aux. FDWP 13 Suction Isol. 0 0 _g, ICA-ll Aux. FDWP 1A Suction Isol. 0 0 0 1CA-2 Aux. FDW7 Suction from Hot- .

well Isol. 0 0 0 ICA-56 Aux. FDWP 1A Disch. to S/G 13 Control 0 0 0 1CA-53 Aux. FDWP 1A Disch. to S/G Control Outlet Isol. _0_ 0 0 1CA-44 Aux. FDWP 13 Disch. S/G IC Control 0 0 0 1CA-46 Aux. FDWP 13 Disch. S/G 1C 4? Control Outlet Isol. 0 0 0 1CA-50 Aux. FDWP No. 1 D..ch to 1C Control Outle: Isol. 0 0 0 ICA-48 Aux. FDWP No.1 Disch. to S/G IC Control 0 0 0 1CA-36 Aux. FDWP No. 1 Disch. to S/G ID Control 0 0 0 1CA-38 Aux. FDWP No. 1 Disch. to l S/G 1D Control Outlet Isol. 0 0 0 ,,

ICA-42 Aux. FDWP 13 Disch. to S/G Control Outlet Isol. 0 0 0 i 1CA-40 Aux. FDW7 13 Disch. to S/G l

1D Control 0 0 0 1CA-54 Aux. FDWP No. 1 Disch. to S/G 13 Control Outlet Isol. _0 , 0 0 1CA-52 Aux. FDWP No. 1 Disch. to S/G 13 Control 0 0 0 1CA-64 Aux. FDW? No. 1 Disch. to S/G

/'~' LA Control 0 0 0 k 1CA-66 Aux. FDW? No. 1 Disch. to S/G LA Control Outlet Isol. 0 0 0

Station D .ctive 3.1.)

Attachner. 2 Rev. 4 g

C Page 3 of 5 A

Req'd Position in the following Modes Actual 1 2 3 4 5 6 Position ICA-62 Aux. TDW7 1A Disch. to S/G LA 0 0 0 Control Outlet Isol.

ICA-60 Aux. TDW7 1A Disch. to S/G LA Control 0 0 0 1CS-18 Aux. Feedvater Supply Isol. 0 0 0 1CM-265 Aux. Teedwater Supply Isol. 0 0 0 1N!-54 Accum. lA Disch. Isol. 0 0 0 LNI-65 Accu =. 13 Disch. Isol. 0 0 0 151-76 Accum. IC Disch. Isol. 0 0 0 INI-88 Acce=. 13 Disch. Isol. 0 0 0 INI-242 UHI Accum. Disch. Isol. 0 0 0 151-243 UHI Accu =. Disch. Isol. 0 0 0 1N!-244 UHI Accu =. Disch. Isol. 0 0 0 "N INI-245 CHI Accum. Disch. Isol. 0 0 0

\-' 0 1T7-27 TVS* to ND Pumps Isol. 0 0 0*

1ND-19 NC Loop C to ND Pu=p 1A Cont.

Isol. Outside 0 0 0 0*

IND-33 ND EX 1A Sypass X X X X*

1ND-34 ND EX 1A and 13 Sypass Centrol X X X X*

1ND-30 ND HX 1A Outle: Crossover 31ock 0 0 0 0*

150-29 ND Pu=p 1A Disch. Flow Control 0 0 1 0*

1NI-173 ND Mdr. to NC Cold Legs A & 3 0 0 0 0*

15D-4 NC Loop C to ND Pu=p 13 Cont.

Isol. Outside 0 0 0 0*

1ND-13 ND EX 13 Sypass X X X X*

IVD-15 ND EX 13 Outlet Cressover Block 0 0 0 0*

1ND-14 ND Pump 13 Disch. Flow Centrol 0 0 0 0*

1NI-178 ND Edr. to NC Cold Legs C & D 0 0 0 0*

ND Edr. to NC Hot Legs 3 & C f(, q) INI-lS3 Isol. X X X X*

• In Mode 4 caly one train of ND and NV are req'd to be operable.

Statien Directiva 3.1.9 Attachment 2 Rev. 4

! ,_,) Page 4 of 5 i

Req'd Position in the following Modes Actual 1 2 3 4 5 6 Position 1NI-100 FWST to NI Pumps 0 0 0 LNI-103 NI Pump 1A Suction 0 0 0 1NI-ilS NI Pump 1A Cold Leg Inj. Lines i Isol. 0 0 0 INI-121 NI Pu=p LA Hot Leg Inj . Hdr.

Isol. X _L, X 3

1NI-162 N! Pu=ps Cold Legs Inj. Edr.

Isol. 0 0 0 INI-135 NI Pump 13 Suction O O O

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1NI-150 NI Pump 13 Cold Leg Inj. Line Isol. 0 0 ,1 1NI-151 NI Pump 13 Hot Leg Inj. Hdr.

Isol. X X X 1NS-32 NS Pump 1A Disch. Cont. Isol.

Outside X X X X O NS Pump 1A Disch. Cont. Isol.

Q INS-29 Outside X X X X b

1NS-15 NS Pump 13 Disch. Cont. Isol.

Outside X X X X

. INS-12 NS Pu=p 13 Disch. Cont. Isol.

Outside X X X X

1N5-20 NS Pu=p 1A Suction from FWST 31ock 0 0 0 0 1N5-3 NS Pu=p 13 Suction from FWST 31ock o 0 0 0 E. List the systems and components that are in a degraded mode of operation as per=itted by Technical Specifications and time in the degraded mode.

System / Component Date and Time Length of Time Allowed Length of Time in Degraded in Degraded Mods Remaining in Mode Degraded Mode 1.

2. ,

1 m_/ 4.

5.

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

Sectien Di . tetiva 3.1.9 Attachnent 2 Rev. 4 Page 5 of

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F. Operator Being Relieving Operator Review Relieved (Initial) (Initials)

1. Removal and Restoration Notebook

2. Out of Nor=al Notebook

3. Alarm Su==ar/ Printout 4 Centrol Boards (Switch Alignment /

Statalar=s)

3. Reactor Operator's Notebook

6. Status of Monitor Lights

7. Un-it 1 Sypass Panel

3. Indicating Lights (Surned cut bulbs)

• G. Remarks (operator being relieved)

Reviewed By Relieving Shift Supervisor og Asst. Shif t Supervisor Shift Being Relieved Relieving Shift (Signature) (Signa ture)

Nuclear Control Operator NOTE: Sections A, 3, C, D, E and G are to be co=pleted by the Nuclear Control Operator being relieved.

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Section : cetivo 3.1.9 Attach =en- 3 Rev. 3 l'm \ Page 1 of 2

\v j UNIT 1 h"JC' E.AR EQUI? MENT CPERATOR TL">1;0VER CHECKLIST Date Time Shift Re ng Re leved A. v.n.D e.

v. ..f.r.

3. Electrical Volts lETA (4160 1 20) ~

lETS (4160 2 20) 1ELXA (600 ; 20) 1El'c (600 1 20)

LEU (3 (600 i 20) 1ECO (600 1 20)

ECCS PUMPS Motor Cooling Nor=al 011 On/Off Control Pur. Avail- Skr. Rack In (/)

IJa:er (/) Levels (/) able (/) (Accal)

CA 1A CA 13

O 1A

D 13 US 1A NS 13 NI 1A

. 43-.2

'Il 1A

i'i 13 F5 1A

?S 13 KC 1A1

, KC 1A2

-~

( ) KC 131

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Static .e: ive 3.1.9 Attachme.: 3 Rev. 4 Page 2 of 2 9L #

D. Lis any equip =ent under =aintenance er test that could degrade a safety systes for presen: sede of plant operation.

System /Co=ponent Re= arks 1.

2.

3.

4.

5. -

E. Remarks:

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NEO 3eing Relieved Relieving NEO (Signature) (Signature)

Reviewed by Relieving Shif: Sup v.

cI, Asst. Shift Supe rviscr: (Signature)

/'~q NOTE: Secticns A, 3, C, D, and I are to be completed by Nuclear Equip =en: Opera:ct (m.- )* being relieved.

STATZON DIRECTIVE 3.1.31 *

') f Of/

APPROVAL // #

DATE O![/bO s

REVISION O DATE 11/6/mo DUKE P0k'ER CCM?ANY McGUIPI NUCLEAR STAT!CN DUTIES, RESPCNSIBILITIES AND QUALIFICATICNS CF THE SHIFT TECHNICAL ADV!SCR (STA)

SHIFT TECHNICAL ACVISOR FU'iCTICN The basic function of the STA is to provide cdditional on-shift capability for evaluaticn and assess =snt of off-nor=al events and nor=al transien:. During transient si:ua: ions, the STA must be available (within :en (10) =inutes of the Control Roo=) to advise the Shif: Supervisor of any appropriate action. The STA's pri=ary concern is the safe operation of the plant without jeopardizing the health and safety of the public. The STA shall be detached from and inde-pendant of the normal line function of shif: operation.

N/

CUALIFICAT!CNS The individual =ust =eet Duke Power Ccepany's general require =ents and in addition:

1. Should be a high school graduate with :wo (1) years technical school or equivalent experience.

2. Shall have a =ini=u of two (2) years nuclear power plan: experience acco=-

panied by an overall knowledge of the plant. At least one (1) year shall be at the station at ahich the position is to be filled.

3. Shall hold a Senior Reactor Operators License.

4. Shall have a working knculedge of stea and water properties.

DUTIES AND PISPONS!3ILITIES A. During nor=al operation:

1. Conduct a turnover of plant status wi:h STA being relieved of duties.

2. Review the plant status at star: of assigned shif:.

A. Review Shift Supervisor and Control Roo= legs.

2_

B. Review Cut-of-Nor=al and Re= oval and Restoration logs for plant s:atus.

C. Review turnover sheets of Shift Supervisors, Nuclear Control Op-erators and Nuclear Equip =en: Operators.

D. Make rounds in the Control Roc = to review Control Roo= status.

E. Review unit work lis:s.

3. Review with Shift Supervisors plan: status and activities scheduled for shift.

4. Infor= the Shift Supervisor of his planned activities during the shif:.

5. Assu=es no responsibili:1es that cannot be i==ediately put aside in order to advise the Shift Supervisor during off-nor=al events.

6. Co==unica:e directly with the Chair =an of Station Safety Review Co==ittee on opera:ing experiences at the station as well as c:her operating plants.

7. Investigate the cause(s) of abnor=al or unusual even:s occurring on as-signed shift and assess any adverse effects therefro=. Reco==end changes to procedures or equip =ent as necessary to prevent recurrence.

S. Evalua:e the effec:iveness of operating procedures in ter=s of ter=inating

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cr =1tigating accidents and =ake reco==endations to the Shif: Supe rviso r hen changes are needed.

9. Prepare special reports when requested by the Superintendent of Operations.

3. During Off-Nor=al Opera:icn.

1. Evaluate plant conditions f ro= available infor ation and deter =ine appropriate responses.

2. Advise Shift Supervisor of appropria:e responses to :he si:uation.

3. Obtain overall picture of the situation by analyzing all available infor=ation and prevent shift personnel frc= " locking i'n" on one indication.

4 Se readily available (within ten (10) =inutes) to :he Control Roc = in the event of an off-nor=al or nor=al transien't.

RELATIONSHIPS

1. The STA will report to an Opera:ing Engineer not responsible for the Shifts.

2. The STA will maintain a close working relationship with the operating shifts.

3. The STA will =aintain a good working relationship with other statien groups as well as Stes: Production Depart =ent personnel.

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4 The STA will only advise the Shift Supervisor and vill not direc: the actions of shif t personnel during nor=al and off-nor=al transients.

_ . _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ ._ _ _ _ _ . . __ _ _ _ _ __. ._ . _ ~ ___ __ ___ . . . . , _ _ _ _ _

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! Charter Of The Station Safety Review Group 4 McGuire Nuclear Station t

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Issued: January 21, 1981 l /

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! Issued By: ~

' R. C. Futrell, Director Nuc rSafetyReview] Board -.-

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Approved: / UW a ' -

William O. Parker, Jr. U W

' Vice President Steam Production Department j

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-**=e-w.'e,w---,,-.-<m_.-. ,___ _ , _ , _ _ _ _

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Table of Contents

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Section i

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. 1.0 Purpose t i

2.0 Membership  !

t 2.1 Appointments j

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2.2 Qualifications i 3

3.0 Responsibilities l

4.0 Authority i

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! 5.0 Method of Operation [

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5.1 General

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! 5.2 SSRG Meetings i t

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O i C

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Charter Station Safety Review Group McGuire Nuclear Station 1.0 Purpose The McGuire Nuclear Station Safety Review Group (SSRG) is established as an independent technical review group for the purpose of examining and making detail recommendations to management on plant operating characteristics, NRC issuances, Licensing Information Service advisories, and other appropriate sources of plant design and operating experience information that may indicate areas for improving plant safety.

2.0 Membership 2.1 Appointments The SSRG is a full time group consisting of a permanent Chairman and a minimum of four rotating members appointed by the Director of the Nuclear Safety Review Board (NSRS). The Chairman of the SSRG will report directly to the Director of the NSR3. The other four rotating members will report to the Chairman of the SSRG. The members will be chosen such that appropriate knowledge and expertise O' is maintained in the areas of instrumentation, maintenance, operations, and technical services (e.g. radiation protection, chemistry, etc.). Appointments i

of rotating members will be for a minimum of 6 months. Turnover of rotating personnel shall be conducted in series and with a sufficient time period to maintain continuity of work assignments.

2.2 Oualifications All members of the SSRG shall have at least six years of technical experience with a minimum of two years being nuclear station experience. A maximum of four years of the six years may be fulfilled by academic or related technical training. The Chairman of the SSRG must be at or above a level of an Assistant Operating Engineer, Associate Engineer, or equivalent level in the Steam Produc-tion Department.

3.0 Responsibilities The SSRG shall be functionally responsible for the following:

1. Improving plant safety by examining plant operating characteristics, 'AEC issuances, Licensing Information Service advisories, and other appropriate sources of plant design and operating experience information.

[N 2. Improving plant safety and operational performance by performing independent (m-) reviews and audits of plant activities including maintenance, modificationg operational problems, and operational analysis, and aiding in the establish-ment of programmatic requirements for plant activities.

L.

3. Verifying that plant operations and maintenance activities are perforned O correctly and that human errors are reduced as far as practical by maintaining necessary surveillance.

4. Advising station and corporate management on the overall quality and safety of operations.

5. Investigating station incidents as assigned by the Station Licensing Group, and preparing required reports.

4.0 Authority The SSRG shall report to and advise the Director of the NSR3 in matters relating to nuclear station safety and shall have the authority to have access to all McGuire nuclear facilities and records to perform its assigned functions. On-site inspections of activities may be conducted as necessary.

5.0 Method of operation 5.1 General The SSRG will carry out the responsibilities described in Section 3.0. The Chairman will give work assignments to hi=self and the remaining members of the SSRG as appropriate. Completed assignments shall be reviewed by each 7- s member of the SSRG. All SSRG recommendations will be sent directly to the i '

'-) Director of the NSR3 and to the Station Managers. They in turn will assure that proper action is taken to address the SSRG recommendations.

5.2 SSRG heetings The Chairman will call formal meetings of the SSRG as necessary to review completed assignments as a group prior to forwarding them to the Director of the NSR3 and to the Station Manager.

5.3 Quorum A quorum of the SSRG shall consist of the Chairman, or his designated alternate, and at least three of the four assigned members.

5.4 Reporting Requirements The Chairman shall ensure that a monthly summary report of the SSRG activities, including the minutes of each SSRG meeting, is prepared and forwarded to the Director of the NSRB and to the Station Manager. The monthly report should include SSRG recommendations, staTO of implementation of corrective actions, and any follow-up evaluations. the esport should be issued by the 15th day of the following month.

1 V

f PT/1/A/4206/11 DUKE POWER COMPANY J -(fh,[3 --l McGUIRE NUCLEAR STA" ION FOR I.nr..O? M/. i,,e..

LEAK RATE DETERMINATION T.;R NI SYSTDI ANDj0R REVIEW OfU.

1.0 Purpose

< To periodically test the Safety Injection System outside containment for leakage.

2.0 References 2.1 NUREG 0578 2.2 OP/1/A/6200/06, Safety Injection System 3.0 Time Raouired 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> at intervals not to exceed each refueling cycle.

4.0 Prerecuisite Tests _

None 5.0 Test Ecuipment 5.1 Graduated Cylinders 5.2 Stop watch 6.0 Limits and Precautions 6.1 Follow Health Physics procedures when collecting and disposing l

of potentially contaminated leakage during this test.

6.2 Ref er to Safety Injection System operating procedure, OP/1/A/6200/06.

7.0 Recuired Unit Status 7.1 RC system is at normal operating pressure and temperature.

8.0 Prerecuisite System Conditions 8.1 NI system nust be capable of being operated at normal pressure and temperature.

J 9.0 Test Method

( 9.1 While ese Safety Injection System is at normal operating pressures and temperatures, all leakage from the system shall be measured and recorded.

9.2 System leakage measurements shall include the following.

f l 9.2.1 All valves listed in Enclosures 13.3, 13.4, 13.5 and 13.6.

9.2.2 All accessible flanges within the system boundaries, i

9.2.3 All instrument loops that are normally valves in within the system boundaries, 9.2.4 Pump packing / seals.

9.3 When measuring leakage on flanges which are lagged, the lagging shall be pierced down to the flange with an awl or similar tool, at the low point in the flange lagging, and the hole observed s

J for leakage. Where any leakage might also run down the pipe, such as in a vertical run with a flange, the lagging shall also be pierced at the low point of the run and checked for leakage.

If any water is observed, the lagging shall be removed for further inspection.

9.4 Leakage measured on instruments shall be recorded as the total leakage from the following:

9.4.1 The instrument itself.

9.4.2 All valves in the loop.

l 9.4.3 All connections within the loop.

l 9.4.4 All drain and vent points in the loop.

9.5 Leakage measured on valves shall include:

9.5.1 Body to bonnet seal.

9.5.2 Packing / stem leakage.

- 9.5.3 Leakage through capped connections for which the valve is used as an isolation; 1.e., a capped vent or drain valve.

9.5.4 Valve body to system flanges in cases where valves are not welded in.

10.0 Data Recuired All leakage will be recorded on data sheets, Enclosures 13.3, 13.4, 13.5 and 13.6.

11.0 Acceptance Criteria e

This test will be considered satisfactory if the total NI system leakage external to the containment, from valves, flanges, instruments, and pump packing / seals does not exceed 1 gym. Notify the Shift Supervisor if this test fails to meet the acceptance criteria.

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12.0 Procedure l

/ 12.1 Align NI system per OP/1/A/6200/06.

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/ 12.2 Start NI pumps LA and 13.

l / 12.3 Check gauges and flanges on Ene_osure 13.1 for leakage.

4 / 12.4 Check gauges and flanges on Enclosure 13.2 for leakage.

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/ 12.5 Check valves on Enclosures 13.3 - 13.5 for leakage.

/ 12.6 Stop pumps.

/ 12.7 Close valves LNI243A and 1NI245A.

l / 12.8 Open valves 1NI2443 and 1NI2423.

f / 12.9 Check valves on Enclosure 13.6 for leakage.

'; 13.0 Inclosures i

13.1 Gauges, Instruments and Flanges in NI Pump Room 1A 13.2 Gauges, Instruments and Flanges in NI Pump Room 13 13.3 NI Valve List for Leakage 13.4 NI Valve List for Leakage (3oron Inj. Iank) 13.5 NI Valve List for Leakage 13.6 NI Valve List for Leakage (UHI Iank) i j

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f PT/1/A/4200/11 1 Enclosure 13.1 j Gauges, Instruments and Flanges in NI Pump Room 1A i >

j Cauges Leakage: yes/no Initial i

i 1NIPT5450 / t i

1NIPG5230 / t 5

l 1 NIPS $310 /

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1NIPG5310 s

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Flanges Leakage

yes/no Initial Flanges on Suction Side / ,

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Flanges on Disch. Side /  !

i, Other Flanges /

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' Enclosure 13.2 Cauges, Instruments and Flanges in NI Pump Room 13 f

j 1 Cauges Leakage yes/no Initial .

1NIPG5240 / }

l 1NIPT5130 /

f /

1NIFT5120 1 NIPS $320 / r i 1NIPG5320 /

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Flanges Leakage: yes/no Initial i

Flanges on Suction Side /

l Flanges on Disch. Side /

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Other Flanges i

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k PT/1/A/4200/11 Enclosure 13.3 1

NI Valve List for Leakage valve Number /Name Leskage: ves/no Initial f

i 1 nil 003 FWST to Safety Inj. Pumps /

1 nil 03A NI Pump 1A Suction /

1NIlC8 NI Pump 1A Vent / i

1NI110 NI Pump LA Drain Valve Tell Tale /

1NI111 NI Pump 1A Drain Isol. /

1NI112 NI Pump 1A Base Drain /

1NI113 NI Pump 1A Discharge Nozzle Drain /

, 1NI115B NI Pump 1A Miniflow Line Isol. /

1N1117 NI Pump 1A Discharge Isol. /

1NI135B NI Pump 13 Suction /

1NI137 NI Pump 13 Vent /

15!139 NI Pump 13 Drain Valve Tell Tale /

'T 1NI140 NI Pump 13 Drain Isol. /

1NI141 NI Pump 13 Base Drain /

1N1142 NI Pump 13 Disch. Nozzle Drain /

1NI144B NI Pump 13 Miniflow Line Isol. /

! 1N!145 NI Pumps Disch. Sample /

1NI147A NI Pumps Miniflow Hdr. to FW /

1N1149 NI Pump 13 Disch. Isol. /

1N1138 Baron Inj. Surge Tank Drain Tell Tale /

1NI189 Boron Inj. Surge Tank Drain Isol. /

1NI192 Boron Inj. Recire. Pump 1A Flush Water Supply /

1NI193 3cron Inj. Recire. Pump 1A Vent. /

I IN!195 Boron Inj. Recirc. Pump 13 Suction Drain /

1NI197 Reactor Makeup Water Supply to Boron Inj.

1 Surge Tank Tell Tale /

1NI199 Boron Inj. Recire. Pump Flush Supply Tell Tale /

1N1200 Boron Inj. Recire. Pump 13 Flush Supply /

1N!201 Boron Inj. Recire. Pump 13 Vent. /

1N1206 BIT Drain Tell Tale /

i (ah i

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

_,__,,,,.7_._._, _m - _,, , , . _ - , . , ,,,e -

,, - m__ , , _-.

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J PT/1/A/4200/11 Enclosure 13.4 NI Valve List for Leakage (Boron Inj. Tank)

Valve Number /Name Leak 2Ee: yes/no Initial INIl BIT Flush to RET Tell Tale /

1NI2 BIT Flush Line Isolation /

INI3 Boron Inj. Check Valve Flush Line Isol. /

LNI4A BIT Inlet Isol. /

i 1NISB BIT Inlet Isol. /

BIT Inlet sample /

1NI6 1NI7 BIT Drain /

INI9A BIT Discharge Isol. /

1N110B BIT Discharge Isol. /

/

f 1NI11 BIT Discharge Check Valve Test Line Isol.

INI22 BIT Vent /

INI23A BIT Recire. Auto Isol. /

/'~' LNI243 BIT Recire. Auto Isol. /

l /

1NI25A BIT Recire. Auto Isol.

INI26 BIT Recire. Man. Isol. /

)

1NI27 BIT Safety Relief /

1NI23 Reactor Makeup k'acer to Baron Inj. Surge Tank Block /

1NI29 BIT Recire. to Boron Inj. Surge Tank Isol. /

1NI30 BIT Flush Line Isol. /

1 15131 Boron Inj. Surge Tank Recire. Bypass /

1N132 Scron Inj. Surge Tank Drain /

INI33 Boron Inj. Surge Tank Sample _

/

INI34 Boron Inj. Surge Iank Outlet /

l l

INI35 Boron Inj. Recire. Pump 1A Suction /

1NI36 Baron Inj. Recire. Pump 1A Discharge Check /

1NI37 Boron Inj. Recire. Pump LA Discharge __ __/

! 1NI38 Boron Inj. Recire. Pump 1B Suction /

1 INI40 Boron Inj. Recire. Pump 13 Discharge /

INI41A 3oron Inj. Racire. Pump Discharge Auto Block /

)

J

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

i r

PT/1/A/4200/11 Enclosure 13.5 NI Valve List for Leakage leakage: yes/no Initial Valve Number /Name 1NI207 BIT Drain Isol. / l 1NI:08 Isol. Valve 1NI121A Test Vent /

1NI209 Isol. Valve 1NI121A Test Vent /

INI210 Isol. Valve 1NI152B Test Vent / i 1NI211 Isol. Valve IN11525 Test vent /____

INI212 Isol. Valve 1NI162A Test Vent /

1NI213 Isol. Valve 1NI162A Test Vent /

1N1126 Isol. Valve 1NI183B Test Vent /

1NI327 Baron Inj. Cont. Isol. Test Vent /

1N1332A NI Pump Suction Crossover from NV /

, 1NI333B NI Pump Suction Crossover frem NV -/

1N1358 NI Pump 1A to UHI Accum. Fill Line Isol. /

f 1NI364 Accum. 13 Disch. High Point Vent Isol. /

l 181385 Isol. Valve 1NI162A High Point Drain /

\

i 1NI399 NI Pump 1A to UHI Accum. Fill 71ne High i Point Vent /

1NI402 NI Pump 13 Suction High Point Vent /

151403 NI Pump 1A Suction High Point Vent /

l 1NI412 NI Check Valve Test Hdr. High Point Vent /

1NI413 NI Check Valve Test Hdr. High Point Vent /

1NI414 NI Check Valve Iest Edr. High Point Vent /

l 1NI415 NI Check Valve Test Edr. High Point Vent /

1NI421 NI Check Valve Test Edr. High Point Vent /

l l

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PT/1/A/4200/11 Enclosure 13.6 i

NI Valve Lisc for Leakage (WI Tank) i Valve Number /Name Leakate: yes/no Inicial INI237 WI Accum. Till Line Isol. / t 1NI233 WI Accum. Fill Line Check /

f /

l 1NI239 WI Accum. Till Line Isol. r 1NI2423 WI Accum. Disch. Isol. /

l INI243A WI Accum. Disch. Isol. /

1NI2443 WI Accum. Disch. Isol. /

1NI245A WI Accum. Disch. Isol. /

/

1NI246 WI Accum. Disch. Isol. Test Drain l

1NI247 WI Accum. Disch. Isol. Test Drain /

l 1NI350 WI Accum. Disch. Till and Drain Conn. /

I s

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,._.m. - . _ ..__,.._.,_._,_..._m...._ . . , . , _ _ . . . _

r u u ni -sv~ o us DURE POWER COMPANY ).

I McGUIRE NUCLEAR STATION FOP. l*FO:WTCJ:

LEAK RATE DETERMINATION FOR ND SYSTEM g g p g gi; g /V O . M 1.0 Purpose To periodicall, test the Residual Heat Removal System outside of containment for leakage.

2.0 References 2.1 OP/1/A/6200/04, Residual Heat Removal System Operating Procedure 2.2 NUREG 0578 2.3 MC 1561-1.0 3.0 Time Required als not to Ten hours each for two Perfor=ance Technicians exceed each refueling cycle.

4.0 Prerequisite Tests None 5.0 Test Ecuipment 5.1 1 Liter Graduated Cylinder 5.2 One Stop Watch

{~}

' 6.0 Limits and Precautions 6.1 Follow Health Physics procedures and recommendations when collecting and disposing potentially contaninated leakage frem the Residual Heat Re= oval Syste=.

6.2 The Reactor shall not be taken above the Cold Shutdown condition until the leakage outside of containment for the Residual Heat Removal Syste= is reduced to less than gallons per hour.

6.3 Do not open 1NI185A and/or INIl84B (Contai= ment Su=p Line A and B Isolation).

6.4 Do not allow the Residual Heat Re= oval System pressure and tempera-ture to exceed the limits as stated in OP/1/A/6200/04, Residual Heat Removal Syste= Operating Procedure.

7.0 Required Unit Status Initial /Date

/ 7.1 The unit must be in one of the following operational modes:

Hot Shutdown - Operational Mode 4 Cold Shutdown - Operational Mode 5 O

i Refueling - Operational Mode 6 v/

_z.

Initial /Date S.O prerecuisite System Conditions

. /f 8.1 The Residual Heat Re= oval System is filled and vented.

/ 8.2 One and/or both RHR Pumps are in operation as per OP/1/A/6200/04 (Residual Heat Renoval System Operating Procedure).

9.0 Test Method 9.1 During unit operation of Modes 4, 5 or 6 with the Residual Heat Removal System in operation, all valves and non-welded connections shall be checked for leakage. This leakage shall be checked by the use of a graduated cylinder and stop watch. Leakage will be calculated by collecting fluid in the graduated cylinder over a convenient period of time.

9.2 System leakage measurement shall include, but not be limited to, the following:

9.2.1 All valves listed in Enclosure 13.2.

9.2.2 All accessible flanges within the system boundaries.

9.2.3 All instrument loops listed in Enclosure 13.3.

9.3 When measurinE leakage on flanges which are insulated, the es insulation shall be pierced down to the flange with an awl or

\g s

strilar tool at the low point in the flange insulation, and the hole observed for leakage. Where any leakage may also run down the pipe, such as in a vertical run with a flange, the insulation shall be pierced at the low point in the run and checked for leakage. If any water is observed, the insulation shall be removed for fuccher inspection.

9.4 Leakage measured on instrument loops shall be recorded as the total leakage from the following:

! 9.4.1 The instrument itself.

9.4.2 All valves in the instrument loop.

l 9.4.3 All connections within the instrument loop.

9.4.4 All drain and vent points within the instrument loop.

)

~s' l

9.5 Leakage measured on valves shall include the following:

9.5.1 Body to Bonnet Seal leakage.

./ 9.5.2 Packing and/or Stem leakage.

9.5.3 Leakage through capped connections for which the valve is used as an isolation, such as in the case of a capped vent or drain valve.

9.5.4 Valve body to system flanges in cases where valves are not welded in to the system.

10.0 Data Recuired The data required will be the calculated leakage rate for all items which are listed in the enclosures.

1 gallon per hour = 3785 ML per hour 20 drops = 1 ML 11.0 Acceotance Criteria This test will be considered satisfactory if the total external Residual Heat Removal System leakage rate from valves, flanges instrument loops, and pump seals does not excee'd gallons per hour. If in the case that acceptance criteria is not met due to excessive leakage, the Shift Super-visor and/or the Operating Engineer should be notified that the system

\s.) does not meet acceptance criteria and work request (s) should be initiated.

O G

12.0 Procedure Initial /Date

[ 'Y 12.1 The Residual Heat Removal System is filled and vented.

D / 12.2 The Residual Heat Removal System is aligned for normal operation as per OP/1/A/6200/04, Residual Heat Removal System Operating Procedure.

/ 12.3 One and/or both Residual Heat Removal Pumps are in operation.

/ 12.4 The temperature and pressure of the RER system are not exceeding the limits of the system as specified in OP/1/A/6200/04, Residual Hea t Removal System Operating Procedure.

/ 12.5 Record the Reactor Coolant System pressure or Enclosure 13.1.

/ 12.6 Visually inspect all Residual Heat Removal System lines and camponents (as listed in Enclosures 13.2 and 13.3) outside con-tainment for leakage.

/ 12.7 Measure leakage of desired component with graduated cylinder and stop watch and record leak rate on Enclosure 13.2 and 13.3.

/ 12.8 Total all leakage for the Residual Heat Removal System and record this amount on Enclosure 13.1.

/ 12.9 Verify that the total leakage rate of the Residual Heat Removal Syste= does not exceed gallons per hour and the acceptance criteria for this test has been met.

/ 12.10 Infor= the Shift Supervisor and/or the Operating Engineer if the leakage from the RER system is exceeding the acceptance criteria. Ensure the proper Work Request (s) has been initiated to correct the defective components.

13.0 Enclosures 13.1 Sumnary Data Sheet 4

13.2 Valve Leak Rate Data Sheet 13.3 Flange and Instrument ak Rate Data Sheet 1

O d

1 l

l 1

f

PT/1/A/4204/07 Enclosure 13.1 Su==ary Data Sheet Residual Heat Re= oval Syste= Total Leakage Rate Outside Contain=ent GPH Reactor Coolant Syste= Pressure During Testing PSIG l

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

9 Page 1 of 3 PT/1/A/4204/07 Enclosure 13.2 p Valve Leak Rate Data Sheet Q I-Leakage Valve Name Valve GPH NC Loop 3 to ND Pump 1B - Containment Isol. Outside _GPH IND4B __

ND Pump 1B Suction Line Drain GPH INC5 ND Pump 13 Drain GPH IND6 ND Pump 1B Plush Supply _GPH 1ND7 ND Pump 1B Discharge Check _GPH INDS ND Pump 1B Discharge _GPH IND9 ND Pump 1B Discharge Saaple _G. PH IND10 IND11 ND EX 1B inlet GPH 1ND12 ND EX 13 Tube Drain __GPH ND EX 1B Tube Drain GPH IND13 ND Pump 1B Discharge Flow Control _GPH IND14' IND15B ND EX 13 Outlet Crossover Block _GPH ND EX IB Discharge to Letdown HX #1 GPH IND17 ND EX 1B Bypass _GPH INDIS NC Loop 1C to ND Pump 1A - Containment Isol. Outside _GPH 1ND19A ND Pump 1A Suction Line Drain _GPH IND20 ND Pump 1A Drain _GPH IND21 __

ND Pump 1A Flush Supply G

_ PH IND22 ND Pump 1A Discharge Check GPH IND23 ND Pump 1A Discharge GPH IND24 ND Pump 1A Discharge Sample GPH IND25 IND26 ND EX 1A inlet _GPH KD EX 1A Tube Drain GPH IND27 IND28 ND EX 1A Tube Drain _GPH ND Pump 1A Discharge Flow Control GPH IND29 IND30A ND EX 1A Outlet Crossover Block _GPH ND EX 1A Discharge to Letdown HX #1 _GPH IND32 ND EX 1A Bypass _GPH INC33 ND EX 1A and 1B Bypass Control GPH IND34 ND System to WST Isolation _GPH IND35 ND Pump 1B Discharge Sample Drain __GPH IND36 ND Pump 2B Vent IND38

PT/1/A/4208/08 DUKE POWER COMPANY McGUIRE NUCLEAR STATION pg.

L4P. f,

6 7l O LEAK RATE DETERMINATION FOR NS SYSTEM FOP. INFOR

".Ti.?N ANDjCP, REVIEW ONLY 1.0 Purpose To periodically test the Containment Spray System outside containment for leakage.

2.0 References 2.1 NUREG 0578 L 2.2 OP/1/A/6200/07 3.0 Time Required 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> - at intervals not to exceed each refueling cycle.

4.0 Prerequisite Tests None 5.0 Test Equipment 5.1 Graduated Cylinders 5.2 Stop Watch 6.0 Limits and Precautions 6.1 Follow Health Physics procedures when collecting and disposing I of potentially contaminated leakage during this test.

6.2 Refer to Containment Spray operating procedure, OP/1/A/6200/07.

7.0 Required Unit Status 7.1 RC System is at nor=al operating pressure and temperature.

8.0 Prerequisite System Conditions 8.1 NS system must be capable of being operated at normal pressure and temperature.

9.0 Test Method 9.1 While the Containment Spray System is at normal operating pressure and temperature, all leakage from the system shall be measured and l

recorded.

l 9.2 System leakage measurements shall include the following:

9.2.1 All valves listed in Enclosures 13.1 and 13.2.

9.2.2 All accessible flanges within the system boundaries.

9.2.3 All instrument loops that are normally valved in within l

the system boundaries.

9.2.4 Pump packing / seals.

O

9.3 Whst measuring laskage on fienges which cre laggsd, tha lagging shall be pierced down to the flange with an awl or similar tool, at the low point in the flange lagging, and the hole observed for leakage. Where any leakage might also run down the pipe, such as in a vertical run with a flange, the lagging shall also be pierced at the low point of the run and check for leakage.

If any water is observed, the lagging shall be removed for further inspection.

9.4 Leakage measured on instruments shall be recorded as the total leakage from the following:

9.4.1 The instrument itself.

9.4.2 All valves in the loop.

9.4.3 All connections within the loop.

9.4.4 All drain and vent points in the loop.

9.5 Leakage measured on valves shall include:

9.5.1 Body to bonnet seal.

9.5.2 Packing / stem leakage.

9.5.3 Leakage through capped connections for which the valve is used as an isolation; i.e., a capped vent or drain f_

(, ) valve.

9.5.4 Valve body to system flanges in cases where valves are not welded in.

10.0 Data Recuired All 3 :akage will be recorded on data sheets, Enclosures 13.1 and 13.2.

11.0 Acceptance L*tteria This test will be considered satisfactory if the total NS System leakage external to the ntainment, from valves, flanges, instruments and pump packing / seals i not exceed 1 gym. Notify the Shift Supervisor if this test fails to m. e acceptance criteria.

N

--]

Initial /Date 12.0 Procedure

/ 12.1 Align NS System per OP/1/A/6200/07.

/ 12.2 Star: SS Pu=ps lA and 13.

/ 12.3 Check valves and accessible flanges en Enclosures 13.1 and 13.2.

/ 12.4 Check instru=en:s and gauges on Enclosure 13.2.

/ 12.5 Stop NS Pu=ps la and 13.

13.0 Enclosures 13.1 Valve Checklist 13.2 Gauges and Instru=ents O-

PT/1/A/420S/08 Enclosure 13.2 Gauges and Instruments

(

Gauges and Instruments Leakage: yes/no Initial 1NSPG5100 /

INSPX5290 /

1NSPX5330 /

1NSFG5470 /

1NSTH5000 /

1NSPG5080 /

1NSPS5080 /

1NSFE5020 /

1NSFT5020 /

1NSP 5020 /

1NSRD5130 /

INSTT5130 /

1NSF 5130 /

() 1NSPG5110 1NSPX5300

/

/

1NSPX5340 /

! LNSPX5300 /

1NSFG5480 /

1NSTE5010 /

1NSPG5090 /

1NSPS5090 /

1NSFE5030 /

1NSFT5030 /

1NSF 5030 /

1NSRD5140 /

1NSTT5140 _ _ _

/

INGP 5140 /

i O

PT/1/A/4207/01 OP

1. ~~ ' ' W' DUKE POWER COMPANY M NEE EI; McGUIRE NUCLEAR STATION v..'_y LEAK RATE Dr iuci! NATION FOR NM SYSTEM ANDl0R RD/L /

1.0 Purpose The purpose is to periodically test the Pri=ary Sampling System outside the containment for leakage. ,

2.0 References 2.1 NUREG 0578 2.2 OP/1/3/6200/ll, Primary Sampling System 2.3 MC-1572-1.0 MC-1572-1.1 MC-1572-2.0 MC-1572-2.1 MC-1572-3.0 3.0 Time Recuired The test takes hours under conditions ranging from full power operation to Hot Shutdown.

a.0 Prerequisite Tests None 5.0 Test Eauipment 5.1 Graduated Cylinder 5.2 Scopwatch 5.3 Air Tank 5.4 Liquid Leak Detector (SNOOP) 5.5 Test gauge (0-20 PSI) 5.6 Volumetric Leak Rate Monitor 5.7 Test tubirgs 6.0 Limitations and Precautions 6.1 Follow Health Physics procedure when collecting and disposing of potentially contaminated leakage during this test.

6.2 All normally closed valves should be closed when the valves are not in use for testing.

6.3 The flow to the Volume Control Tank and the Waste Evaporator Feed O Tank shall be kept to a minimum.

(O e , ,>. - - - < ,- -w, , < < + , c~,_-ww r--- x , -

rw a r,-< , ,s+ - ~ - - - , -, -y - ,

6.4 Do not exceed PSIG to pressurize VCT Gas Space Sample Line for leakage test.

7.0 Required Unit Status 7.1 All systems which are connected to the Primary Sampling System shall have operational pressure and temperature.

8.0 Prerecuisite System Conditions 8.1 The NM System is in normal operation as per OP/1/B/6200/11 (Enclosure 7.2 through 7.28).

9.0 Test Method 9.1 While the NM system is at normal operating pressures and tempera-tures, all leakage from the system shall be measured using a graduated cylinder and a stopwatch. The VCT Gas Space Sample Line is pressurized to psig and a leak rate test is performed.

9.2 System leakage measurements shall include, but not be limited to the following:

9.2.1 All valves listed in Enclosure 13.1.1 through 3.1.23.

9.2.2 All instrument loops that are normally valved in within the system boundaries.

D 9.3 Leakage measured on instruments shall be recorded as the total leakage from the following:

9.3.1 The instrument itself 9.3.2 All valves in the loop 9.3.3 All connections within the loop 9.3.4 All drain and vent points in the loop 9.4 Leakage measured on valves shall include:

9.4.1 Body to bonnet seal 9.4.2 Packing / stem leakage 9.4.3 Leakage through capped connection for which the valve is used as an isolation; i.e., a capped vent or drain valve.

10.0 Data Recuired All leakage will be recorded on Enclosure 13.1.1 through 13.1.28.

11.0 Acceptance Criteria This test will be considered satisfactory if the total external Primary Sampling System leakage rate from va2res and components does not exceed

. Natify the Shift Supervisor if this test fails to meet the acceptance criteria.

T 12.0 Procedure NOTE: Steps in this procedure may be done out of order at the discretion of the Test Coordinator except Steps 12.1 and 12.2.

i Initial /Date 12.1 NC Reactor Coolant Pressurizer water

/ 12.1.1 Request operators to align the sample line by opening INM3A (PZR Liquid Sampling Line Inside Containment Isol.)

and then opening INM73 (PZR Sample Header Outside Contain-ment Isol.).

/ 12.1.2 To isolate the system, request Operations to close valve INM102.

/ 12.1.3 Position the following valves as indicated.

PZR Sample Edr. Needle INM17 OPEN PZR Atmospheric Sample INM18 CLOSE PZR Sample Vessel Outlet INM16 OPEN PIR Sample Vessel Dissolved Gas App. Isol. INM61 CLOSE PZR Sample Vessel Dissolved Gas

- App. Isol. INM14 OPEN PZR Sample Vessel Dissolved Gas App. Isol. INM13 CLOSE P2R Sample Vessel Dissolved Inlet INM19 OPEN

/ 12.1.4 Perform leakage measurements on the components listed in Enclosure 13.1.1.

/ 12.1.5 After the completion of the above test, close valves:

PZR Sample Hdr. Needle INM17 PZR Sample Vessel Inlet 1NM19 PZR Sample Vessel Outlet INM16 PZR Sample Dissolved Gas App. Isol. INM14 PZR Sample Dissolved Gas App. Isol. 1NM15

/ 12.1.6 Request Operations to close valves:

PZR Liquid Sample Line Inside Cont. Isol. 1NM3A PZR Sample Header Outside Cont. ,

Isol. 1NM7B

~

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--m-- .--r , ,,,,.,-.e .c,,.,,, -m....,--,- .,-,,,w---,,,-- ,-e.--- .-,,--emm.- m-e, v .- -,-r,--~,, ,,c, -, , - - - , - -e--- e , - - ,r-+,-e.

bi V

laitial/Date

/ 12.1.6 Sum the leakages in Enclosure 13.1.1 and record the swa on Enclosure 13.2.

12.2 NC Reactor Coolant Hot Leg Loops 1 and 4 J

/ 12.2.1 Request Operations to align the sample line by opening either 1NM22A (NC Hot Leg #1 Sample Line Inside Con-tainment Isol.) or INM25A (NC Hot Leg #4 Sample Line Inside Containment Isol.).

/ 12.2.2 Position the following valves as indicated. I NC Hot Leg Atmospheric Sample INM37 CLOSE NC Hot Leg Sample Hdr. Needle INM38 OPEN NC Hot Leg Sample Hdr. Dissolved Gas App. Isol. INM32 OPEN NC Hot leg Sample Hdr. Dissolved Gas App. Isol. INM33 CLOSE NC Hot Leg Sample Hdr. Dissolved Gas App. Isol. INM34 OPEN NC Hot Leg Sample Edr. Dissolved 7-sg Cas App. Isol. 1NM63 CLOSE

( } NC Hot Leg Sample Vessel Outlet INM35 OPEN NC Hot Leg Sample vessel Inlet INM62 OPEN

/ 12.2.3 Perform leakage measurement on the components listed in Enclosure 13.1.2.

/ 12.2.4 After the completion of the above test, request Operations to realign the valve, which was opened in Step 12.2.1, and to open valve INM102.

/ 12.2.5 Close the following valves:

NC Hot Leg Sample Edr. Needle INM38 NC Hot Lag Sample Vessel Inlet INM62 NC Hot Leg Sample Vessel Outlet INM35 NC Hot Leg Sample Hdr. Dissolved Gas App. Isol. INM34 NC Hot Leg Sample Edr. Dissolved Gas App. Isol. INM32

/ 12.2.6 Sus the leakages in Enclosure 13.1.2 and record the su= on Enclosure 13.2.

r t

x-

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

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[ )

LJ Initicl/Date 12.3 NI Accu =ulator (A, 3, C, and D)

/ 12.3.1 Request Operations to align the sa=ple line for testing by opening one of the following valves:

J lA Sa=ple Line Inside Cont. Isol, lA-12723 13 Sa=ple Line Inside Cont. Isol. 13-12753 1C Sa=ple Line inside Cont. Isol. 1C-lNM783 1D Sample Line Inside Cont. Isol. 13-1 2 813 and then opening valve 1SM82A (NI Accu =ulator Sa=ple Edr. Outside Cont. Isol.).

/ 12.3.2 To isolate the sa=ple line for testing, request Operations to close valve:

Purge Edr. to WIFT Isol. 12103

/ 12.3.3 Position the following valves as indicated.

1 276 CLOSE Atmospheric Sa=ple 15M73 OPEN Sample Purge to T4En Isol.

/ 12.3.4 Perfor= leakage =easure=ents on the co=ponents listed in Enclosure 13.1.3.

/ 12.3.5 After the co=pletion of the above test, close valve:

Sa=ple Purge to '.T.C ! sol. 15M73

/ 12.3.6 Request Operations to close the valves which were opened in Step 12.3.1 and to ocen ISM 103 (Purge Edr. to WEM Isol.).

12.3.7 Su= the leakages in Enclosure 13.1.3 and record the sum

/

on Enclosure 13.2.

12.4 S/G A, 3, C and D Sample Line to Radiation Monitor - DiF #34

/ 12.4.1 The following valves are checked open to per=it flow to Di? #34 as per OP/1/3/6250/08 (SG 31cwdown) .

S/G Slowdown Sa=ple EX 1A to Rad. Men. 15M274 S/G 3 lowdown Sa=ple EX 13 to Rad. Mon. INM275 S/G 31ovdown Sa=ple EX 1C to Rad. Men. 15M276 S/G Blowdown Sa=ple EX ID to Rad. Mon. 15M277

/ 12.4.2 To align the sa=ple lines for testing, open valves:

NM1 Flush Water Header Isolation

,- S/G Sa=ple Isolation NM244

/

1 }

v

./ 7 (w' )

Initial /Date

/ 12.4.3 Perfor: leakage measurenents on the co=ponents listed in Enclosure 13.1.4

/ 12.4.4 After the completion of the above test, close valve:

Flush k*ater Header Isolation INH 1 S/G Sa=ple Isolation 1NM244

/ 12.4.5 Su= the leakages in Enclosure 13.1.4 and record the su=

on Enclosure 13.2.

12.5 ND Pu=p 1A or 13

/ 12.5.1 The sample line for these points is routed fro: a dead leg recirculation piping arrange =ent. Prior to testing, flow should be verified in this piping by contacting Operations and verifying valve IN36SA (N3 Pu=p 1A and HX 1A Miniflow Stop) for "A" Train g valve IN3673 (N3 Pu=p 13 and HX 13 Miniflow Stop) for "B" Train is open or partially open.

/ 12.5.2 If valve IN368A is open, then request Operations to open valve:

ND Pu=p 1A Disch. Sa=ple Line Isol. INM39 If valve lh3673 is open, then request Operations to open valve:

ND Pu=p 13 Disch. Sa=ple Line Isol. 1NM40

/ 12.5.3 To ali;n the ND Pu=p LA or 13 Sa=ple line for testing, open valves:

RER Sample Edr. Purge INM42 RER Sa=ple Bypass 1NM47 Residual Ha.at Re= oval Sa=ple Headar Isol. INM45

/ 12.5.4 Perfor: leakage =easurements on the ec=ponents listed in Enclosure 13.1.5.

/ 12.5.5 After the completion of the above test, close valve:

Residual Heat Re= oval Sample Header Purge INM42 Residual Heat Re= oval Sa=ple Bypass ISM 47 Residual Heat Re= oval Sample

(h INM45

(_) Hdr. Isol.

Initial /Date

/ 12.5.6 Close the valve which was opened in Step 12.5.2.

/ 12.5.7 Sum the leakages in Enclosure 13.1.5 and record the sum on Enclosure 13.2.

12.6 NV Volume Control Tank Gas Space

/ 12.6.1 Isolation the sample line for testing be closing valves:

Sample Line Isol. 1NM273 WG Decay Tank "E" Sample lWG8

/ 12.6.2 Connect the test equipment to the test connection at 1NM52 (Discharge Sample Line Drain) and supply air for testing.

/ 12.6.3 Position the following valves as inidicated.

Sample Line Purge Isol. INM49 Open VCT Sample vessel Isol. 1NM56 Open VCT Sample Vessel Isol. 1NM54 Open VCT Sample Vessel Isol. 1NM53 Open VCT Sample Vessel Isol. 1NM55 Open

("'N'

\s/ / 12.6.4 Open valve INM52 (Discharge Sample Line Drain), and slowly pressurize the system to psig with air.

/ 12.6.5 Record the leakage on Enclosure 13.1.6.1 and Enclosure 13.2.

12.6.6 If the leakage is larger than , then bubble test the

/

components listed in Enclosure 13.1.6.2 to locate the leakage.

/ 12.6.7 After completion of the test, position the following valves:

Discharge Sample Line Drain 1NM52 Close Wasta Gas Decay Tank "E" Sample IWG8 Open Sample Line Purge Isol. INM49 Close VCT Sample vessel Isol. 1NM56 Close VCT Sample Vessel Isol. 1NM54 Close VCT Sample Vessel Isol. 1NMS3 Close VCT Sample Vessel Isol. INM55 Close

/ 12.6.8 Close the air supply valve and depressurize the test gauge, and disconnect the test equipment.

s

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. . _ _ . . . . ~ _ .._ , . _ . . . __ , , .__ ...-.-. -.___ _. .._ -.

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laitial/Date 12.7 NV MAxed Bed Demineralizer

/ 12.7.1 To align the sample line for testing, open valve:

Mixed Bed Demin. Sa=ple Line Purge INM94 Isol.

/ 12.7.2 Perform leakage measurements on the components listed in Enclosure 13.1.7.

/ 12.7.3 After the completion of the above tett, close valve:

Mixed Bed Demin. Sa:ple Line Purge INM94 Isol.

/ 12.7.4 Sus the leakages in Enclosure 13.1.7 and record the sum on

~

Enclosure 13.2.

12.8 NC Cation Bed Demineralizer

/ 12.S.1 To align the sample line for testing, open valve:

Cation Bed Demin. Sample Line Purge Isol. INM98

/ 12.8.2 Perfor= leakage measurements on the ccmponents listed in Enclosure 13.1.8.

12.8.3 After the compittion of the above test, close valve:

L_/

Cation Bed Demin. Sample Line "trge Isol. INM98

/ 12.8.4 Sum the leakages in Enclosure 13.1.8 and record the su= on Enclosure 13.2.

12.9 NV Boric Acid Blender

/ 12.9.1 To align the sa=ple line for testing, open valve:

Boric Acid 31 ender Sample Line Purge Isol. INM110 12.9.2 Perform leakage measurements on the ce=penents listed in

/

Enclosure 13.1.9.

/ 12.9.3 Af ter the completion of the above test, close valve:

Soric Acid Blender Sample Line 1NM110 Purge Isol.

/ 12.9.4 Sum the leakages in Enclosure 13.1.9 and record the sum on Enclosure.13.2.

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v Ini:ial/0 !e 12.10 Vole =e Control Tank Outle:

/ 12.10.1 To align the sa=ple line for :esting, open valve:

VCT Cutle: Sample Line Purge Isol. INH 206 J

/ 12.10.2 Perf or= leakage =easure=ents on the components in Enclosure 13.1.10.

/ 12.10.3 Af ter the completion of the above :est, close valve:

VCT Outle: Sa=ple Line Purge Isol. 1NM106

/ 12.10.4 Sus the leakages in Enclosure 13.1.10 and record the su=

on Enclosure 13.2.

12.11 NV Seal Injection Fil:er

/

12.11.1 To align the sample line for testing, open valve:

Seal Water Inj. Filters Sample Line Purge Isol. INM114 Seal Wa:er Inj. Fil:ers Sa=ple Line Needle INM115

/ 12.11.2 Perfor= leakage =easure=ents en the co=ponents lis:ed in Enclosure 13.1.11.

/ 12.11.3 Af:er the completion of the above test, close valve:

Seal Water inj. Fil:ers Sa=ple Li=e Purge Isol. INM114 Seal Water inj . Filters Sa=ple Line Needle 1.W.115

/ 12.11.4 Su= :he leakages in Enclosure 13.1.11 and : hen record the su= on Enclosure 13.2.

12.12 NV Le:down Hea: Exchanger Outle:

/ 12.12.1 To align the sample line for testing, open valve:

Letdown EX Sa=ple Line Purge Isol. INM90

/ 12.12.2 Perform leakage =easure=ents on the componen:s listed in Enclosure 13.1.12.

/ 12.12.3 Af:er the completion of the above test, close valve:

Letdown EX Sample Line Furge Isol. INM90

/ 12.12.4 San the leakages in Enclosure 13.1.13 and then record the sus en Enclosure 13.2.

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Initial /Date 12.13 NR Demineralizer Outlet

/ 12.13.1 To align the sample line for testing, open valve:

NR Demin. Outlet Sample Purge Isol. INM119

/ 12.13.2 Perform leakage measurements on the components listed in Enclosure 13.1.13.

/ 12.13.3 Af ter the completion of the above test, close valve:

NR Demin. Outlet Sa=ple Purge Isol. INM119

/ 12.13.4 Sum the leakages in Enclosure 13.1.13 and record the sum on Enclosure 13.2.

12.14 NB Recycle Evap. Feed Demineralizer "A" Outlet

/ 12.14.1 To align the sample line for testing, open valve INM123.

/ 12.14.2 Perform leakage measurements on the components listed in Enclosure 13.1.14.

/ 12.14.3 Af ter the completion of the above test, close valve:

Recycla Evap. Feed Demin. A Sample Purge Isol. INM123 12.14.4 Sun the leakages in Enclosure 13.1.14 and then record the

__/

sum on Enclosure 13.2.

12.15 NB Recycle Evap. Feed Demineralizer "B"

/ 12.15.1 To align the sample line for testing, open valve:

Recycle Evap. Feed Demin. B Sample Purge Isol. 1NM127

/ 12.15.2 Perform leakage measurements on the components listed in Enclosure 13.1.15.

/ 12.15.3 After the completion of the above test, close valve:

Recycle Evap. Feed Demin. B Sa=ple Purge Isol. INM127

/ 12.15.4 Sum the leakages in Enclosure 13.1.15 and record the sum on Enclosure 13.2.

O

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Initial /Date 12.16 NB Recycle Evap. Feed Pump Outlet (Recycle Holdup Tank)

/ 12.16.1 To align the sample line for testing, open valve:

Recycle Evap. Feed Sample Line Purge Isol. INM131

/ 12.16.2 Perfor= leakage measurements on the components listed in Enclosure 13.1.16.

/ 12.16.3 Af ter the completion of the above steps, close valve:

Recycle Evap. Feed Sample Line Purge Isol. INM131

/ 12.16.4 Su= the leakages in Enclosure 13.1.16 and record the sum on Enclosure 13.2.

12.17 NB Condensate Demineralizer

/ 12.17.1 To align the sample line for testing, open valve:

Recycle Evap. Cond. Demin. Sa=ple Line Purge Isol. INM135

/ 12.17.2 Perform leakage =easurements on the components listed on Enclosure 13.1.17.

12.17.3 After the completion of the above test, close valve:

_/

Recycle Evap. Cond. Demin. Sample Line Purge Isol. 1NM135

/ 12.17.4 Sum the leakages in Enclosure 13.1.17 and record the sum on Enclosure 13.2.

12.18 NB Recycle Evap. Feed De=ineralizer Inlet

/

12.18.1 To align the sample line for testing, open valve:

Recycle Evap. Feed Demin. Inlet Sample Purge Isol. INM139

/ 12.18.2 Perform leakage =easurements on the components listed in Enclosure 13.1.18.

/ 12.18.3 After the completion of the above test, close valve:

l Recycle Evap. Feed Demin. Inlet Sample Purge Isol. 1NM139

/

12.18.4 Sus the leakages in Enclosure 13.1.18 and record the sum on Enclosure 13.2.

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\'I$1tial/Date 12.19 WL Waste Evap. Feed Pump Discharge (Waste Evap. Feed Tank)

/ 12.19.1 To align the sa=ple line for testing, open valve:

Waste Evap. Feed Tank Sample Line

. Purge Isol. 1NM143

/ 12.19.2 Perform leakage measurements on the components listed in Enclosure 13.1.19.

/ 12.19.3 After the completion of the above test, close valve:

Waste Evap. Feed Tank Sample Line Purge Isol. 1NM143

/ 12.19.4 Sum the leakages in Enclosure 13.1.19 and record the sum on Enclosure 13.2.

12.20 Waste Drain Tank Pu=p Discharge (Waste Drain Tank)

/ 12.20.1 To align the sample line for testing, open valve:

Waste Drain Tack Sa=ple Line Purge Isol. INM147

/ 12.20.2 Perform leakage measurements on the components listed in Enclosure 13.1.20.

12.20.3 After the completion of the above test, close valve:

l __/

l Waste Drain Tank Sample Line Purge l

Isol. INM147 l

/ 12.20.4 Sum the leauages in Enclosure 13.1.20 and then record the sum on Enclosure 13.2.

12.21 Waste Evap. Dist. Cooler Outlet

/ 12.21.1 To align the sample line for testing, open valve:

Waste Evap. Cond. Demin. Outlet Purge Isol. INM155

/ 12.21.2 Perform leakage measurements on components listed in Enclosure 13.1.21.

.-/

12.21.3 After the completion of the above test, close valve:

Waste Evap. Cond. Demin. Outlet Purge Isol. INM155

/ 12.21.4 Sus the leakages in Enclosure 13.1.19 and record the su=

on Enclosure 13.2.

l l

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1 Initial /Date 12.22 Waste Evap. Cond. Demin. Outle

/ 12.22.1 To align the sample line for testing, open valve:

Samp1w Line Purge Isol. INM155

/ 12.22.2 Perform leakage measurements on the components in 4 Enclosure 13.1.22.

/ 12.22.3 After the completion of the above test, close valve:

Sample Line Purge Isol. INM155

/ 12.22.4 Sum the leakages in Enclosure 13.1.22 and then record the sum on Enclosure 13.2.

12.23 Recycle Monitor Tank Pump Discharge

/ 12.23.1 To align the sample line for testing, open valve:

Sample Line Purge Isolation INH 159

/ 12.23.2 Perform leakage measurements on the components in Enclosure 13.1.23.

/ 12.23.3 After the completion of the above test, close valve:

Sample Line Purge Isolation 1NM159

/ 12.23.4 Sum the leakages in Enclosure 13.1.23 and then record the sum on Enclosure 13.2.

12.24 RMWST Discharge

/ 12.24.1 To align the sample line for testing, open valve:

Sample Line Purge Isol. 1NM163 I

/ 12.24.2 Perform leakage measurements on the components listed in Enclosure 13.1.24.

/ 12.24.3 After the completion of the above test, close valve:

Sample Line Purge Isol. INM163

/ 12.24.4 Sum the leakages in Enclosure 13.1.24 and record the sum on Enclosure 13.2.

12.25 KF Spent Fuel Pool Cooling Prefilter

/ 12.25.1 To align the sample line for testing, open valve:

Sample Line Purge Isol. INM153

/ 12.25.2 Perform leakage measurements on the components listed in Enclosure 13.1.25.

l

IT'

%s' Initial /Date

/ 12.25.3 After the completion of the above test, clo.e valve:

Sample Line Purge Isol. 1NH153

/ 12.25.4 Sum the leakages in Enclosure 13.1.25 and record the sum on Enclosure 13.2.

12.26 KF Spent Fuel Pool Cooling Post-Filter

/ 12.26.1 To align the sample line for testing, open valve:

Sa=ple Line Purge Isol. INM141

/ 12.26.2 Perform leakage =easure=ents on the components listed in Enclosure 13.1.26.

/ 12.26.3 After the completion of the above test, close valve:

Sample Line Purge Isol. INM141

/ 12.26.4 Sus the leakages in Enclosure 13.1.26 and record the sum on Enclosure 13.2.

12.27 RWST Discharge

/ 12.27.1 To align the sample line for testing, open valve:

Sa=ple Line Purge Isol. 1NM129 12.27.2 Perfor= leakage measurements on the ec=ponents listed

__/

in Enclosure 13.1.27.

/ 12.27.3 After the completion of the above test, close valve:

Sample Line Purge Isol. INM129

/ 12.27.4 Sus the leakages in Enclosure 13.1.27 and record the su=

en Enclosure 13.2.

12.28 Spent Resin Sluice Filter Discharge

/

12.28.1 To align the sa=ple line for testing, open valve:

Sample Line Purge Isol. INM117

/ 12.28.2 Perfors leakage measurements on the components listed in Enclosure 13.1.28.

/ 12.28.3 After the completion of the abeve test, close valve:

Sa=ple Line Purge Isol. INM117

/ 12.23.4 Sum the leakages in Enclosure 13.1.28 and record the sum on Enclosure 13.2.

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( 13.0 Enclosures 13.1 Leak Check Worksheets 13.1.1 Pressurizer Water and Steam Space 13.1.2 Reactor Coolant Loops 1 and 4 13.1.3 Accumulator 13.1.4 S/G Blowdown (A, B, C, and D) 13.1.5 RHR Pump Discharge (A and B) 13.1.6 VCT Gas Space 1331.7 NV Mixed Bed Demineralizer 13.1.8 NV Cation Bed Demineralize:

13.1.9 NV Boric Acid Blender 13.1.10 VCT Outlet 13.1.11 NV Seal Injection Filters 13.1.12 NV Letdown EX Outlet 13.1.13 NR Demineralizer 13.1.14 Recycle Evap. Feed Demin. "A" Outlet 13.1.15 Recycle Evap. Feed Demin. "B" Outlet l

l g 13.1.16 Recycle Evap. Feed Pump Outlet 13.1.17 NB Condensate Demineralizer 13.1.18 NB Evap. Feed Demineralizer Inlet 13.1.19 Waste Evap. Feed Pump Discharge 13.1.20 Waste Drain Tank Pump Discharge 13.1.21 Waste Evaporator Distribution Cooler Outlet 13.1.22 Waste Evap. Condenser Demineralizer Outlet 13.1.23 Recycle Monitor Tank Pump Discharge 13.1.24 RMWST Discharge 12.1.25 KF Prefilter 12.1.26 KF Postfilter 12.1.27 RWST Discharge 12.1.28 Spent Resin Sluice Filter Discharge 13.2 Total NM System Leakage l

t

Pago 1 of 14 PT/1/A/4207/01 Enclosure 13.1 d

Leak Check Worksheets i Enclosure 13.1.1 Pressurizer Water and Steam Space LEAKAGE DESCRIPTION CC/ MIN INITIAL ITEM 1

INM73 PZR Steam Header Outside Cont. Isolation 15M8 Cont. Isolation Valve INM7B Test Vent 1NM9 PZR Sample HX Inlet Isolation 15M23 Flush Line to PZR Sample HX 1 INM64 PZR Sample HX L Tube Vent HX Pressurizer Sample Heat Exchanger l

i 15MIH5040 Pressurizer Sample Temperature 15M263 PZR Sample HX Outlet Needle 15M10 PZR Sample Hdr. Purge 15M11 PZR Sample Edr. Purge Check _

15M12 PZR Sample Vessel Outlet Check s

727 PZR Sample Bdr. Needle 1:0'JS PZR Atmospheric Sample I

15M19 PZR Sample Vessel Inlet 15 MPG 5080 Pressurizer Sampler Vessel Pressure 15MMI5280 PZR Sample Vessel 1 15M13 PZR Sample Vessel Dissolved Cas Apparatus Isolation 1:011 4 PZR Sample Vessel Dissolved Gas Apparatus Isolation INM15 PZR Sample Vessel C1.3 solved Gas Apparatus Isolation 1NM61 PZR Sampic Vessel Dissolved Gas Apparatus Isolation 15M16 PZR Sample Vessel Outlet TOTAL LEAKAGE 1

Pags 2 of 14 O)

(,, PT/1/A/4207/01 Enclosure 13.1 Leak Check Worksheets Enclosure 13.1.2 NC Loops 1 and 4 LEAKAGE DESCRIPTION CC/ MIN INITIAL ITEM INM263 NC Hot Legs Sample Edr. Outside Cont. Isolation 1NM27 Cont . Isol. Valve JNM26B Test Vent 1NM2S NC Hot Leg Sample EX Inlet Isolation 1NMS7 Flush Line to Reactor Coolant Hot Leg Sample EX 1 INM65 NC Hot Leg Sample EX 1 Tube Vent EX Reactor Coolant Hot Leg Sample Heat Exchanger 1NM264 NC IIoc Legs Sample HX Outlet Needle INMTH5050 Reactor Coolant Hot Leg Sample Temperature INM38 NC Hot Leg Sample Edr. Needle INM37 NC Hot Leg Atmospheric Sample NM62 Reactor Coolant Hot Leg Sample Vessel Inlet N Reactor Coolant Hot Leg Sample Pressure iiPG5090 1NHME5290 Reactor Coolant Hot Leg Sample Vessel 1 INM32 NC Hot Leg Sample EX Dissolved Gas Apparatus Isol.

INM33 NC Hot Leg Sample EX Dissolved Gas Apparatus Isol.

1NM34 NC Hot Leg Sample EX Dissolved Gas Apparatus Isol.

INM63 Reactor Coolant Hot Leg Sample Vessel Gas Appar. Isol.

1NM35 NC Hot Leg Sample Vessel Outlet 15M31 NC Hot Leg Sample Vessel Outlet Check

, 1NM29 NC Hot Legs Sample Hdr. to Rad. Monitor 1 EMF 48 1EHF48 NC Hot Legs Radiation Monitor 1NM30 NC Hot Legs Sample Edr. , Rad. Monitor lEMF48 Outlet Check TOTAL LEAKAGE

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• m-- -w yy - w- *> - , -

Page 3 of 14 PT/1/A/4207/01 '

Enclosure 13.1 j

Leak Check Worksheets Enclosure 13.1.3 NI Accumulator (A, B, C, D)

LEAKAGE DESCRIPTION CC/ MIN INITIAL ITEM INM280 NI Accumulatcrs Sample Edr. Vent ,

1NM82A NI Accumulators Sample Edr. Outside Cont. Isolation Cont. Isol. Valve INM82A Test Vent 1:383 1:373 NI Accumulators Sample Purge to wen Isolation INM76 NI Accumulators Acmespheric Sample INM79 NI Accumulators Sample Header Needle i

! 1NM70 NI Accumulators Sample Purge to WEK Check TOTAL LEAKAGE  :

O J

l l

l l

l l

Pass 4 of 14 FT/1/A/4207/01 Enclosure 13.1 Leak Check Worksheets Enclosure 13.1.4 Steam Generator Slowdown "A" LEAKAGE DESCRIPTION CC/ MIN INITIAL 1 ITDi INM191B S/G 1A Sample Edr. Cont. Isol. Outside i

1NM192 Cont. Isot. Valve INM1913 Test Vent l 1NM193 S/G Blowdown Sacple HX 1A Isolation 1:01229 Flush Line to S/G Blowdown Sample RX 1A 120C59 S/G Blowdown Sample MX 1A Vent HX S/G Blowdown Sample Heat Exchanger 1A INMTS$120 S/G Blowdown Sample HX 1A Outlet Temperature ,

1:01274 S/G Blowdown Sample HX 1A to Rad. Hon.

Rad. Mon. 1EFF-34 Disch. to Ground Water Drain Sump "A" 1W.279 1:0C44 S/G Sample Isolation LNM194 S/G 1A, 1B, 1C and ID Shell Side Sample Line Needle i S/G 1A Sample Hdr. to Conv. Sampling System i 01269 STEAM GENERATOR BLOWDOWN "B" 1NM201A S/G 1B Sample Hdr. Cont. Isol. Outside 1:01202 Cont. Isol. Valve 1iM201A Test Vent INM203 S/G Blowdown Sample HI 1B Isolation 1NM233 Flush Line to S/G Blowdown Sample HX 1B INM260 S/G Elowdown Sample HX IB Vent INMHX S/G Blowdown Sample Heat Excharger 1B 1NMTS5130 S/G Blowdown Sample HI 1B Outlet Temperature 3 12270 S/G 1B Sample Edr. to Cotiv. Sampling System j 12275 S/G Blowdown Sample HX 1B to Rad. Hon.

STEAM GENERATOR BLOWDOWN "C" 1:01211B S/G 1C Sample Hdr. Cont. Isol. Outside INM212 Cont. Isol. Valve INM211B Test Vent 1NM213 S/G Blowdown HK IC Isolation 1NM237 Flush Line to S/G Blowdown Sample HX 1X NM261 S/G Blowdown Sample HX IC Vent i

d i

i

Pegs 5 of 14 PT/1/A/4207/01 Enclosure 13.1 Leak Check k'orksheets Enclosure 13.1.4 (Cont.)

Steam Generator Blowdown "C" LEAKAGE DESCRIPTION CC/ MIN INITIAL ITEM EX S/G Blowdewn Sample EX LC 11MIE5140 S/G Blowdown Sample EX C Outlet Temp.

INMTS$140 S/G Blowdown Sample KX iC outlet Temp.

11M271 S/G IC Sample Hdr. to Conv. Sampling System 1W.276 S/G Blowdown Sample HX IC to Rad. Mon.

STEAM GENERATOR BLCWDOWN "D" 1.W.221A S/G 1D Sample Edr. Cont. Isol. Outside INT 50 Cont. Isolation Valve INM221A Test Vent 1:3222 S/G Blowdown Sample HX 1D Isol.

113241 Flush Line to S/G Blowdown Sample HX ID _

v l262 S/G Blowdown Sample HK 1D Vent S/G Blowdown Sample HX 1D 11MIE3150 S/G Blowdown EX D Outlet Temperature 1SMTS5150 S/G Blowdown HX 1D Outlet Temperature 1!3277 S/G Blowdown Sample EX ID to Rad. Mon.

17.272 S/G 1D Sample Hdr. to Conv. Sampling System S/G BLOWDOWN TO BLOWDOWN TANK ISM 226 S/G 1A, 13, 1C & 1D Sampla to Rad. Monitor Needle 1 MFE5160 S/G Sample Flow 1NT G5160 S/G Sample Pressure 1W.227 S/G 1A, 1B, 1C & ID Sample to Rad. Monitor Check 1:3T55270 S/G Sa=ple Hdr. Rad. Monitor Inlet Temp.

f 1!MTE5270 S/G Sample Hdr. Rad. Monitor Inlet Temp.

1 7.267 S/G Sample Header Rad. Monitor Inlet Isolation 1 EMF 34 Radiation Monitor to S/G Blowdown Tank l 15M245 Radiation Monitor to S/G Blowdown Tank Check

1W.246 Radiation Monitor to S/G Blowdown Tank Isol.

113279 Rad. Mon.1DiF-34 Disch. to Ground Water Drainage Sump "A" O *278 Rad. Monitor 1 EMF-34 Disch. to S/G Blowdown Tank Q

i Psgo 6 of 14 PT/1/A/4207/01 Enclosure 13.1 ,

Laak Check Worksheets i

Enclosure 13.1.4 (Cont.)

RMWST Flush Line to S/G Blevdown EX LEAKAGE .

CC/ MIN INITIAL DESCRIPTION ITEM INM1 Flush Water Header Isolation 1NM4 Flush Water Header Telltale INM20 Flush Line to PZR Sample EX 1 Check INM48 Flush Line to Reactor Coolant Hot Leg Sample EX 1 Check 1!M58 Flush Line to Residual Heat Removal Sample HX 1 Check 1NM22S Flush Line to S/G Blowdown Sample EX 1A Check 1NM232 Flush Line to S/G Blowdown Sample EX 1B Check 1NM237 Flush Line to S/G Blowdown Sample EX IC INM241 Flush Line to S/G Blowdown Sample EX ID TOTAL LEAXAGE s

Pego 7 of 14 j

l PT/1/A/4207/01 Enclosure 13.1 Leak Check Worksheets Enclosure 13.1.5 ND Pump Discharge (A & 5)

LEAKAGE ITE DESCRIPTION CC/ MIN INITIAL l

l 1.W39 ND Pump 1A Disch. Sample Line Isolation 1240 ND Pump 13 Disch. Sample Line Isolation ISM 41 Residual Heat Removal Sample HX Isolation 15M59 Flush Line to Residual Heat Removal Sample HX 1 INM66 Residual Heat Removal Loop Sample EX 1 Tube Vent HX RER Sample HI 1hv.265 Residual Heat Removal I. cop HX Outlet Needle 1.W.4 2 Residual Heat Removal Sample Hdr. Purge 15M43 Residual Heat Removal Sample Hdr. Purge Check 15M45 Residual Heat Removal Sample Edr. Isolation 15M46 Residual Heat Removal Sample Line Needle

!  ?.l.7 Residual Heat Removal Sample Bypass ihrd5060 RER Loop Sample Temperature TOTAL LEAKAGE i

! Pcgs 8 of 14 PT/1/A/4207/01 Enclosure 13.1 Leak Check Worksheets VCT Gas Space Enclosure 13.1.6 Enclosure 13.1.6.1 Volumetric Leak Race Monitor i Cal. Due Date:

Test Cauge #

Cal. Due Date VCi Gas Space Sample Line Leakage seem Date Data Recorded By VCT Gas Space Enclosure 13.1.6.2 LEAXAGE DESCRIPTION YES/NO INITIAL ITEM 1 2273 Volume Control Tank Gas Space Sample Isolation 1.T.56 Volume Control Tank Gas Space Sample vessel Isol.

1254 Volume Control Tank Gas Space Sample vessel Isol.

1253 Volume Control Tank Gas Space Sample Vessel Isol. _

1255 Volume Control Tank Gas Space Sample Vessel Isol.

1252 Volume Control Tank Gas Space Disch. Sample Line Drain 1W.51 Volume control Tank Gas Space Sample vessel outlet check 1249 volume Control Tank Gas Space Sample Line Purge 1250 Volume Control Tank Gas Space Sample Line Purge Check 1.71'.E5300 VCT Gas Space Sample vessel IWJI5070 VCT Gas Sample Purge Flow

Pegs 9 of 14 PT/1/A/4207/01 Enclosure 13.1 Leak Check Worksheets Enclosure 13.1.7 NV Mixed Red Domineralizer LEAKAGE CC/ MIN INITIAL DESCRIPTION ITEM 1.v.93 Mixed Bad Domin. Sample Line Purge Check 1NM94 Mixed Bed Domin. Sample Line Purge Isol.

INM95 Mixed Bed Domin. Sample Line Purge Needle TOTAL LEAKAGE Enclosure 13.1.8 NV Cation Bed Domineralizar 1NM97 Cacion Bed Demin. Sample Line Purge Check 1NM93 Catton Bed Desin. Sample Line Purge Isol.

1:019 9 Cacien Bed Demin. Sample Line Needle TOTAL LEAKAGE I

Enclosure 13.1.9 NV Boric Acid Blender INM109 Boric Acid Blender Sample Line Purge Check f

1NM110 Boric Acid Blender Sample Line Purge Isol.

1NM111 Boric Acid 31 ender Outlet Sample Line Needle TOTAL LEAKAGE Enclosure 13.1.10 Volume Control Tank Outlet 1:01105 VCT Outlet Sample 1.ine Purge Check 1NM106 VCT Outlet Sample Line Purge Isol.

INM107 VCT Outlet Sample Line Needle TOTAL LEAKAGE l

Pego 10 of 14 f_,T s /

PT/1/A/4207/01 Enclosure 13.1 Leak Check Worksheets Enclosure 13.1.11 NV Seal Injection Filter LEAKAGE DESCRIPTION CC/ MIN INITIAL ITEM INM266 Seal Water Inj. Filter Sample Line Needle INM113 Seal Water Inj. Filters Sample Line Purge Check ISM 114 Seal Water inj. Filters Sample Line Purge Isol.

1NMil5 Seal Water Inj. Filters Sample Line Needle 151216 Seal Water inj. Filters Sample Line Disch.

TOIAL LEAKAGE Enclosure 13.1.12 NV Letdown Heat Exchanger Outlet 1NMS9 Letdown EX Sample Line Purge Check 1NM90 Letdown EX Sample Line Purge Isolation Letdown EX Sample Line Needle it91

] TOTAL LEAKAGE Enclosure 13.1.13 NR Demineralizer Outlet INMll3 NR Demineralizers Outlet Sample Line Purge Check 1NM119 NR Deminerali:ers Outlet Sample Line Purge Isol.

1NM120 NR Domineraliser Outlet Sample Line Needle INM101 Purge Rdr. to VCT Check 1NM102 Purge Edr. to VCT Isolation INMTE5110 Purge Flow to VCT TOTAL LEAKAGE Enclosure 13.1.14 NB Recycle Evap. Feed Demineralizer "A" Recycle Evap. Feed Demin. "A" Sample Line Purge Check 1NM122 INM123 Recycle Evap. Feed Demin. "A" Sample Line Purge Isolation 1NM124 Recycle Evap. Feed Demin. "A" Sample Line Needle TOTAL LEAKAGE O

Page 11 of 14 O PT/1/A/4207/01 Enclosure 13.1 Leak Check Worksheets Enclosure 13.1.15 NB Recycle Evap. Feed Demineralizer "B" LEAKAGE DESCRIPTION CC/ MIN INITIAL ITEM INM126 Recycle Evap. Feed Demin. "B" Sample Line Purge Check 1NM127 Recycle Evap. Feed Demin. "B" Sample Line Purge Isol.

15M128 Recycle Evap. Feed Domin. "B" Sample Line Needle TOTAL LEAKAGE Enclosure 13.1.16 NB Recycle Evap. Feed Pump Outlet 1NM130 Recycle Evap. Feed Sample Line Purge Check 1NM131 Recycle Evap. Feed Sample Line Purge Isol.

INM132 Recycle Evap. Feed Sample Line Needle TOTAL 1EAKAGE l

Enclosure 13.1.17 N3 Condensate Deminerali:er 1NM134 Recycle Evap. Cond. Domin. Sample Line Purge Check 1NM135 Recycle Fvap. Cond. Domin. Sample Line Purge Isol.

15M136 Recycle Evap. Cond. Demin. Sample Line Needle TOTAL LEAKAGE Enclosure 13.1.18 NB Recycla Evap. Feed Demineralizer Inlet 1NM133 Recycle Evap. Feed Demin. Inlet Sample Purge Check 1NM139 Recycle Evap. Feed Domin. Inlet Sample Purge Isol.

1NM140 Recycle Evap. Feed Demin. Inlet Sample Line Needle TOTAL LEAKAGE O

Pego 12 of 14 PT/1/A/4207/01 Enclosure 13.1 Leak Check Worksheets Enclosure 13.1.19 WL Waste Evap. Feed Pump Discharge LEAKAGE CC/ MIN INITIAL ITEM DESCRIPTION (Waste Evap. Feed Tank)

ISM 142 Waste Evap. Feed Tank Sample Line Purge Check 1W.143 Waste Evap. Feed Tank Sample Line Purge Isol.

1W.144 Waste Evap. Feed Tank Sample Line Needle TOTAL LEAKAGE Enclosure 13.1.20 Waste Drain Tank Pump Discharge 1W.146 Waste Drain Tank Sample Line Purge Check 1W.147 Waste Drain Tank Sample Line Purge Isol.

12.148 Waste Drain Tank Sample Line Needle TOTAL LEAKAGE Enclosure 13.1.21 Waste Evap. Dist. Cooler Outlet 1W.150 Waste Evap. Dist. Cooler Outlet Sample Line Purge Check 12251 Waste Evap. Dist. Cooler outlet Sample Line Purge Isol.

1W252 Waste Evap. Dist. Cooler Outlet Sample Line Needle TOTAL LEAKAGE Enclosure 13.1.22 Waste Evap. Cond. Demin. Outlet 1 W.154 Waste Evap. Cond. Domin. Outlet Sample Line Purge Check 1W155 Waste Evap. Cond. Demin. Outlet Sample Line Purge Isol.

17.156 Waste Evap. Cond. Demin. Outlet Sample Line Needle TOTAL LEAKAGE O

'l Pass 13 of 14 PT/1/A/4207/01 Enclosure 13.1 Leak Check Worksheets Enclosure 13.1.23 Recycle Monitor Tank Pump Discharge ,

LEAKAGE DESCRIPTION CC/ MIN I:11TIAL ITEM INM158 Recycle Monitor Tanks Sample Line Purge Check 1NM159 Recycle Monitor Tanks Sample Line Purga Isolation INM160 Recycle Monitor Tank Sample Line Needle TOTAL LEAKAGE Enclosure 13.1.24 RMWST Discharge INM162 RMWST Sample Line Purge Check 1NM163 RMWST Sample Line Purge Isolation 1!0'.161 RMWST Sample Line Needle TOTAL LEAKAGE s

Ecclosure 13.1.25

( ,,) KF Spent Fuel Pool Cooling Prefilter ISM 157 Spent Fuel Pool Sample Line Purge Check 1NM153 Spent Fuel Pool Sample Line Purge Isolation 1NM149 Spent Fuel Fool Sample Line Needle TOTAL LEAKAGE Enclosure 13.1.26 KF Spent Fuel Cooling Post-Filter 1NM145 KF Post Filter Outlet Sample Line Purge Check 1N>u41 KF Post Filter Outlet Sample Line Purge Isolation INM137 KF Post Filter Outlet Sample Line Needle TOTAL LEAKAGE Enclosure 13.1.27 RWST Discharge INM133 FWST Sample Line Purge Check 1NM129 FWST Sample Line Purge Isolation INM125 FWST Sample Line Needle TOTAL LEAKAGE

l Pags 14 of la 4 .

6 PT/1/A/4207/01

' Enclosure 13.1 Leak Check Worksheets Enclosure 13.1.28 Spent Resin Sluice Filter Discharge LEAKAGE i CC/ MIN INITIAL

' DESCRIPTION ITEM 1.W.121 Spent Resin Sluice Filter Sample Line Purge Check 1NM117 Spent Resin Sluice Filter Sample Line Purge Isol.

17.112 Spent Resin Sluice Filter Outlet Sample Line Needle 12.100 Purge Hdr. to WEFT Check 1NM103 Purge Bdr. to WEFT Isolation  :

TOTAL LEAKAGE 4

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+ v_, ---,c -w,wr,ww,--#..v- , -w.---y-r--,--,www,%.,. ,--,w.-..,..w-e

. , , n m e .--e...,, ---,,s----.-,--.---- .m , , - - , , ,----w,-ww.r,.-ms--,,--,- ,w %-r

PT/1/A/4207/01 Enclosure 13.2

!E System Leakage Liquid Gas Leakage Leakage ec/ min SCCM Initial Data _

i Enclosure 13.1.1 Enclosure 13.1.2

• Enclosure 13.1.3 Enclosure 13.1.4 Enclosure 13.1.5 Enclosure 13.1.6 N/A Enclosure 13.1.7 Enclosure 13.1.8 Enclosure 13.1.9 Enclosure 13.1.10 Enclosure 13.1.11 Enclosure 13.1.12 Enclosure 13.1.13 Enclosure 13.1.14 Enclosure 13.1.15

! Enclosure 13.1.16 Enclosure 13.1.17 Enclosure 13.1.18 Enclosure 13.1.19 Enclosure 13.1.20 Enclosure 13.1.21 4

Enclosure 13.1.22

• Enclosure 13.1.23 _

I Enclosure 13.1.24 J

l Enclosure 13.1.25 Enclosure 13.1.26 l

Enclosure 13.1.27 Enclosure 13.1.2S TOTAL NM SYSTEM LEAXAGE I

4

- - , . - . - . . . . , _ , - _ _ , , . . _ , _,,,,-..~.-.-.,..~,_.--...-.--,,-_-.,._.-__~....-._,_,4- ,c. -w.- , .. . . - . , - *

- DWT Fon it;F0F..vTt0N D E PO D CO N b

v .cGUIRE M NUCLEAR STATION At D/OR REVii"# GT.LY

! LEAK RAIE DETERMINATION FOR N3 SYSTEM t

1.0 Purpose ~4 h To periodically test the Eoron Rec 'lhys' tem (NB)forleakage.

j 2.0 References 2.1 NUREG 0578 2.2 Flow Diagrams MC-1556-1.0 MC-1556-1.1 MC-1556-2.0 MC-1556-2.1 MC-1556-2.2 2.3 NB Operating Procedure, OP/0/A/6200/25.

2.4 NB System Description MC-1223.00-06.

3.0 Time Required 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> at intervals not to exceed each refueling cycle.

O 4.0 Prerequisite Tests None 5.0 Test Equipment 5.1 Graduated cylinders 5.2 Stop watch 6.0 Li=its and Precautions Exercise care and follow Health Physics procedures when collecting and disposing of potentially contaminated leakage.

7.0 Required Unit Status None 8.0 Prerequisite System Conditions Initial /Date

/ 8.1 Recycle Holdup Tanks are capable of being placed in recirculation through the feed filters and feed domineralizers.

O

9.0 Test Method The Boron Recycle (NB) System vill be tested by placing the Recycle HoJdup Tanks in recirculation via the feed filters and feed domin-eralizers. Syste= leakage measurements shall include, but not be limited to the Zollowing.

i

1. All valves listed on enclosures.

2. All accessible flanges within the system boundaries.

3. All instruments loops that are normally valved in within the system boundaries.

4. Pump / packing seals.

Leakage measured on instruments shall be recorded as the total leakage of the following:

1. Instrument itself.

2. All valves within the loop.

3. All connections within the loop.

4. All drain and vent points within the loop.

Leakage measured on valves shall include:

1. Body to bonnett seal.

2. Packing /ste= leakage.

3. Leakage through capped connections.

l 4. Valve body to syster. flanges in cases where valves are not velded in.

Leakage, if found, vill be measured by the use of a graduated cylincer and stop watch. Individual leaks vill be recorded and total leakage computed.

f 10.0 Data Recuired 10.1 Leakage in ec/ min. per component.

10.2 Total leakage of NB System.

11.0 Acceptance Criteria This test vill be considered acceptable when all leaks have been iden-tified on Enclosure 13.1 and scheduled for repair by a work request.

O

12.0 Procedure Irial/Date

-/ 12.1 Place the Recycle Boldup Tanks in recirculation via the N3 feed filters and f eed demineraliters per OP/0/A/6200/25, Section 4.4.

/ 12.2 Inspect the valves listed on Enclosure 13.1 for leakage. If leakage is observed, measure it using a graduated cylinder and stop watch. Record leakage on Enclosure 13.1.

/ 12.3 Inspect the recycle evaporator feed pump A or,B (whichever is running) suction and discharge flanges for leakage. If leakage is observed, measure it and record leakage rate on Enclosure 13.2.

/ 12.4 Swap evaporator f eed pumps .ind repeat Step 12.3.

/ 12.5 Observe the instrumentation listed on Enclosure 13.2 for any leak-age. If any is cbserved, measure and record the leakage on Enclosure 13.2.

/ 12.6 Note, measure and record any leak observed within the NB System while performing this test on components not listed on the enclosures which =ay be encountered during this test.

/ 12.7 Calculate total leakage on Enclosure 13.1 and enter on Enclosure 13.3.

./ 12.8 Calculate total leakage on Enclosure 13.2 and enter on Enclosure 13.3.

/ 12.9 Calculate total system leakage on Enclosure 13.3.

13.0 Enclosures 13.1 Leak Check Valves 13.2 Leak Check Components 13.3 Leak check Summary l

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Pcge 1 of 2 PT/1/A/4203/03 Enclosure 13.1 Leak Check Valves Leakage Valve Name ec/ min. Initial Valve No.

Recycle Evap. Feed Pump B Vent Overflow Isol. / .

INB119 Recycle Evap. Feed Pump B Vent /

1NB120 Recycle Evap. Feed Pump A Vent Overflow Isol. /

1NB109 Recycle Evap. Feed Pump A & B Flush Tell Tale /

1NB106 Recycle Evap. Feed Pump A Vent /

1NB110 Recycle Evap. Feed Pump A & B Sample /

1NB123 Recycle Evap. Feed Pump A & B Drain Tell Tale /

1NB117 Recycle Evap. Feed Pump A & B Flush /

1NB107 Recycle Holdup Tank A Drain Tell Tale /

1NB68 Recycle Holdup Tank A Drain Isol. /

1NB69 Recycle Holdup Tank A Overflow /

1NB70 Recycle Holdup Tank A Drain /

1h371 Recycle Holdup Tank A Loop A High Pt. Vent /

INB361 Recycle Holdup Tank A VA Air Supply Isol. /

1NB73 Recycle Holdup Tank A Overflow VAC Bk. /

1NB290 Recycle Holdup Tank A Condensate Drain /

1NB65 Recycle Holdup Tank Sample Vessel Inlet Isol. /

1NB71.

Recycle Holdup Tank Sample Vessel Outlet Isol. /

1NB76 INB354 Waste Drain Tank to WEFT Sump A /

Waste Drain Tank to WEFI Sump A Isol. /

1NB345 l /

1NB93 Recycle Holdup Tank Sample Vessel Inlet Isol.

Recycle Holdup Tank Sample Vessel Outlet Isol. /

1NB97 Recycle Holdup Tank Eductor Suction Control /

1NB98 Recycle Holdup Tank Sample Line Drain /

1NB355 Recycle Holdup Tank Sample Vessel Outlet Isol. /

1NB100 Recycle Holdup Tank Sample Vessel Inlet Isol. / __

1NB102 RET A & B Makeup from KF Unit 2 Isol. /

INB286 Recycle Holdup Tank B Inlet High Point Vent /

1NB359 RET A & B Makeup from KF Unit 1 Isol. /

1NB318 Recycle Holdup Tank A Flush Tell Tale /

1NB63

/

1NB360 Recycle Holdup Tank B Loop Seal High Point Vent

PT/1/A/4203/03 Enclosure 13.1 Leak Check Valves Leakage Valve Name ec/ min. Initial Valve No.

Recycle Evap. Feed Demin. A Flush Isol. /

1NB9 NB Supply Sample Block /

1NB7

/

1NBS Recycle Evap. Feed Demic. A. Inlet 1NB6 NB System Supply Header Safety Relief /

1NB3 Unit 2 NI to Recycle Holdup Tank B Isol. / _

1NB288 VA Air Supply to Recycle Holdup Tank B Isol. /

INB91 Recycle Holdup Tank B Flush Tell Tale /

1NBSO

/

1NB347 Recycle Holdup Tank Sample Vessel Line Drain 1NB367 Recycle Holdup Tank B Inlet High Pt. Vent /

Recycle Evap. Feed Filter A & B Flush Tell Tale /

1NB50 1NB55 Recycle Evap. Feed A & B Flush Tell Tale /

1NB358 NB Feed Filter Outlet High Point Vent /

Recycle Evap. Feed Filter A & B Flush Isol. /

1NB40 Recycle Evap. Feed Domin. B Drain Tell Tale /

1NB32

/

1NB22 Recycle Evap. Feed Demin. A Outlet Sample Recycle Evap. Feed Domin. A Drain Tell Tale /

1NB18 Recycle Evap. Feed Domin. B Flush Isol. /

1NB10 Recycle Evap. Feed Demin. A & B Flush Tell Tale /

1NBli Unit i VCT Valve Hdr. High Point Vent /

1NB365 Unit i VCT Valve Rdr. High Point Vent /

INB366

/

1NB328 Recycle Evap. Feed Domin. A & B Bypass ,

1NB327 Recycle Evap. Feed Domin. A & B Isol. /_

Recycle Evap. Feed Domin. A & B Recire. Isol. /

1NB136

/

LNB36 Recycle Evap. Feed Demin. B Outlet Sample TOTAL LEAKAGE l

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PT/1/A/- 03/03 Enclosure 13.2 O Laak Check Components Laakage Cot::ponents Description cc/ min. Initial Suction Flange Recycle Evap. Feed Pump A /

Discharge Flange Recycle Evaporator Feed Pump A /

Suction Flange Recycle Evaporator Feed Pump B /

Discharge Flange Recycle Evaporator Feed Pump B /

ONBRG5450 Recycle Evap. Feed Pump B Vent Flow /

ONBPG5020 Recycle Evap. Feed Pump B Discharge /

ONBFG5440 Recycle Evap. Feed Pump A Vent Flow / _

ONBPG5010 Recycle Evap. Feed Pump A Discharge /

NBFG5110 Recycle Holdup Tank B Level /

Recycle Holdup Tank A Level /

ONBPG5000 Recycle Holdup Tanks Vent Flow /

ONBFI5100 Recycle Feed Filter B D/P /

ONBPG5130 Recycle Feed Filter A D/P /

ONBPG5040 TOTAL LEAKAGE O

t

~

. ,-.,--.,.,,_...-.._,_-_._.._.,,..,,,..,.__._-,._.._..-___,.,..--,,,....,,-._:.,27_,_..~

, _ . . ~,

PT/1/A/4203/03 O, Enclosure 13.3 Leak Check Summary Total Leakage from 13.1 Total Leakage from 13.2 +

=

Tofal NB System Leakage 0

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PT/1/A/4209/07 DUKE POWER COMPANY l McGUIRE NUCLEAR STATION FOR INTORMATION l LEAK RATE DETERMINATION FOR NV SYSTEM AND/OR REVIEW ONLY l

1.0 Purpose To periodically test the chemical and volume control system outside containment for leakage.

2.0 References .

.f 7%

2.1 NUREG 0578 s' /g[ t 3.0 Time Requf. red [

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at intervals not to exceed each refueling cycle.

4.0 Prerequisite Tests None 5.0 Test Equipment 5.1 Graduated cylinders 5.2 Stop watch 5.3 Awl 6.0 Limits and Precautions 6.1 Follow Eealth Physics procedures when collecting and disposing of potentially contaminated leakage during this test.

7.0 Required Unit Status None 8.0 Prerequisite System Conditions Stated in each section.

9.0 Test Method 9.1 While systems are operating in the recirculation mode, all leakage from the system shall be measured and recorded.

9.2 System lealrage measurements shall include but not be limited to the following:

9.2.1 All valves listed in Enclosure 13.1.

9.2.2 All accessible flanges within the system boundaries.

9.2.3 All instrument loops that are normally valved in within

the system boundaries.

l 9.2.4 Pump packing / seals.

l

(

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_2

'N When seasuring leakage on flanges which are lagged, the lagging

).3 shall be pierced down to the flange with an awl or similar tool, at the low point in the flange lagging, and the hole observed for leakage. Where any leakage might also run down the pipe, such as in a veritical run with a flange, the lagging shall also be pierced at ths low point of the run and checked for leakage.

If any water is observed, the lagging shall be removed for further inspection.

9.4 Leakage measured on instruments shall be recorded as the total leakage from the following:

9.4.1 The instrument itself.

9.4.2 All valves in the loop.

9.4.3 All connections within the loop.

9.4.4 All drain and vent points in the loop.

9.5 Leakage measured en valves shall include:

9.5.1 Body to bonnet seal.

9.5.2 Packing / stem leakage.

() 9.5.3 Leakage through capped connections for which the valve is used as an isolation; i.e., a capped vent or drain valve.

9.5.4 Valve body to system flanges in cases where valves are note welded in.

10.0 Data Recuired l

All leakage will be recorded on data sheets, Enclosures 13.1 and 13.2.

11.0 Acceptance Criteria This test will be considered satisfacec.y if the total NV System leakage, external to the containment, from valves, system flanges and pump packing /

seals does not exceed 1 gpm. Notify the Duty Operating Engineer if this test fails to meet the acceptance criteria.

P:so 1 of 9 PT/1/A/4209/07 Enclosure 13.1 O Leak Check Worksheets Laskage Component Description Yes/No Initial SEAL WATER Ex ROOM 733' 1NV210 SW Ex #1 Tube Overflow 1NV215 SW Hx #1 Tube Drain to WEFT INV213 SW Hz #1 Tube Drain 1NV212 SW Ex #1 Tube Drain 1NV211 SW Ex #1 Tube Drain 1NV126 SW Hz #1 Tube Drain to WDT INV207 SW Ex #1 Tube Vent 1NV208 SW Ex #1 Tube Flush and Vent Isolation 1NV153 SW Hz #1 Tube Outlet Isol.

1NV152 SW Ex #1 Tube Inlet Isol.

1NV209 SW Ex #1 Tube Vent Tell Tale 1NV214 SW Hx #1 Tuba Drain Tell Tale 1NV154 SW Ex #1 Bypass 1NV206 Seal Water Ex #1 Tube Flush Supply 1NV155 Seal Water Ex #1 Tube Inlet Safety Relief HALLWAY OUTSIDE NB FEED PLMP ROOM 716' 1NV865 Recycle Evap. Feed Pump to Charging Pumps Vent Isol.

INV885 Recycle Evap. Feed Pump to Charging Pumps Vent Isol.

WET LAY-UP PLMP ARIA INV438 Seal Inlet Line Containment Isol. Press. Test l

1NV856 Charging Pumps Suction Hdr. Low Pt. Drain 1Nv275 Charging Pumps Suction Edr. Press. Test Point 1NV469 Seal Water Inj . Filters Outlet Sample 1NV28 NC Pump 1A Seal Water Manual Control 1NV5300 RCP D Seal Water Flow 1NV229 CCP Suction Hdr. Safety Relief

P 33 2 of 9 PT/1/A/4209/07 O Enclosure 13.1 Laak Check Worksheets Leskage Co conent Description Yes/No Initial BIT AREA 4

1NV203 Seal Water Filter #1 Drain Tell Tale INV197 Seal Water #1 Flush Supply Tell Tale 1NV305 Seal Water and Seal Water Inj. Filters Flush Supply Tell Tale 1NV308 Seal Water and Seal Water Inj. Filters Drain Tell Tale INV324 NC Filters Flush Supply Tell Tale 1NV328 NC Filters Drain Tell Tale 1NV44 NC Pump 13 Seal Water Manual Control 1NV60 NC Pump 1C Seal Water Manual Control 1NV149 Seal Water Filter Bypass CCP's Recirculation

{'"150 L__151 CCP's Recirculation 1NV5310 RCP C Seal Water Flow LETDOWN Ex ROOM 750' 1NV8 Letdown Reheat Ex (Tubeside) Back Pressure Control Isol.

INV9 Letdown Reheat Ex (Tubeside) Back Pressure Control Isol.

INV850 NC LD High Point Vent 1NV10 Letdown Reheat Hz (Tubeside) Back Pressure Control Isol.

INV126 Low Pressure LD Control Bypass _

INV11 Letdown Reheat Ex (Tubeside) Back Pressure Control Bypass 1NV852 NC Letdown Line High Pt. Vent 1NV851 Letdown Ex #1 Inlet High Point Vent 1NV447 NC Letdown Line Containment Isol. Pressure Test P'"i77 Low Pressure LD Control Outlet Isolation

( )

LJ30 LD Ex #1 Tube Vent Tell Tale 1NV183 LD Ex #1 Tube Drain Tell Tale __

Peas 5 of 9 PT/1/A/4209/07 O Enclosure 13.1 Leak Check Worksheets Leakage Description Yes/No Initial g

733' VCT AREA (Cont.)

VCT #1 Inlet Check Boric Acid Blender Outlet Safety Relief Boric Acid Blender Discharge Flow RX MAL'".UP WATER PL'MP AREA 716' SW Ex #1 Tube Outlet to Charging Pumps Suct.

SW Ex #1 Tube Outlet to VCT #1 VCT #1 Outlet Isolation VCT #1 Outlet

.PE CHASE BEIVEEN NR DEMIN. AND MIXED BED DEMIN.

RX Makeup Water to Mixed and Cation Bad Domin.

Tell Tale Mixed and Cation Bad Domins. Outlet Line Drain .

Tell Tale Mixed Bed Demim 1A Backflush Outlet Tell Tale 716' PENETRATION ROOM COLL %N EH-52 Regenerative Ex #1 Tube Inlet Control Isol.

Regenerative Ex #1 Tube Inlet Control Isol.

Regenerative Ex #1 Tube Inlet Control Bypass _

NV Charging Line Containment Isol. Press.

Test Drain Charging Line Cont. Isol. Outside Charging Line Cont. Isol. Outside Regen. Ex #1 Tube Inlet Control O

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

i

- - Ptg3 8 of 9 l

t PT/1/A/4209/07 '

Enclosure 13.1 Laak Check Worksheets i

Leakage i I Component Description Yes/No Initial t

I RECIPROCATING CHARGING PUMP ROOM LW279 PDP Flush and Tell Tale LW274 PDP Makaup from RMWST Tell Tale LW276 PDP Drain LW272 PDP #1 Drain to WDT INV271 PDP #1 Vent Tell Tale LW830 PDP #1 Drain 1NV270 PDP #1 Flush and Vent Isolation LW273 PDP #1 Vent i

LW219 PDP #1 Discharge Isolation 1NV834 PDP #1 Discharge to UHI Isolation 4

1NV803 PDP #1 Outlet Isolation 1NV497 PDP #1 Suction i 1NV217 PDP #1 Suction 1NV867 UHI to PDP Suction Vent Isolation LW237 CCP Discharge Control Isolation j

1NV239 CCP Discharge Control Isolation

! 1NV835 NV Pump 1 Suction Dampening Bottle Drain 1NV839 NV Pump 1 Suction Dampening Bottle 32 I'#1' 1NV909 NV Pump 1 Suction Dampening Atmosphere Venc 1NV278 Reciprocating Charging Pump #1 Flush Supply 1NV238 CCP's Discharge Control 1NV5780 Reciprocatin;; Charging Pump Discharge Pressure STP.AINER FLANGES BETWEEN INV279 AND INVC35 PDP FLANGES PDP ACCUMULATOR FLANGES INV220 Reciprocating Charging Pump #1 Disch. Line Safety Relief O1NV218 Reciprocating Charging Pump #1 Discharge Check 1NC5540 Reciprocating Charging Pump Discharge Pressure

Pega 9 of 9 PT/1/A/4209/07 O Enclosure 13.1 Leak Check Worksheets Leakage Component Description Yes/No Initial BETWEEN UNIT 1 AND 2 CCP 1NV298 Charging Pump 1B Drain Tell Tale INV293 Charging Pump 13 Vent Tell Tale PIPE CHASE WEST OF SW Ex ROOM INV868 SW Ex #1 Tube Outlet Vent Isol. _

PIPE CHASE NORTH OF PDP 716' 1NV863 Recycle Evap. Feed Pump to Charing Pumps Vent Isol.

716' PIPE CHASE COLUMN LINE GG INV858 CCP Discharge Header Vent 1NV859 CCP Discharge Header Vent 1NV860 SW Injection Filters Inlet Vent Isol.

1NV921 CCP Discharge Header Low Point Drain INV907 CCP Discharge Header Hi Point Vent 1NVSSS CCP Suction Header High Point Vent 1NV866 CCP Suction Header High Point Vent l 1NVS57 ND to NI Pump Suction High Point Vent 1NV906 CCP to Seal Injection Filter Vent 1NV5630 Charging Line Flow Control NI PUMP ARIA INV282 Charging Pumps flA Vent Tell Tale i

INV287 Charging Pump #1A Drain Tell Tale O

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PT/1/A/4209/07 Enclosure 13.2 Leak Check Summary NV System Leakage From:

cc/ min. cc/hr.

Enclosure 13.1 Total Lesuge l

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

PT/1/A/4201/03 m DUKE POWER COMPANY McGUIRE NUCLEAR STATION DRAHT LEAK RATE DETERMINATION FOR W SYSTEM FOR INFORMATION AND/OR REVIEW ONLY 1.0 Purpose To test the (W) Refueling Water System outside of containment for leakage.

2.0 References 2.1 NUREG 0578 [

2.2 2.3 Drawing MC 1571-1.0-c022 hh' OP/1/A/6200/14 *%d 3.0 Time Required 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> for two people 4.0 Prerequisite Tasts None 5.0 Test Equipment 5.1 Graduated cylinder 5.2 Stop watch

(/

t i 6.0 Limits and Precautions 6.1 Follow Health Physics procedures when collecting and disposing of potentially contaminated leakage during this test.

6.2 Refer to Refueling Water System operating procedure for other precautions and limits.

7.0 Required Unit Status 7.1 The system should be tested before or after refueling.

7.2 The system should be at normal operating temperature and pressure.

8.0 Prerequisite System Conditions The system must be aligned according to Enclosure 13.1.

9.0 Test Method 9.1 While running the Refueling Water System, any leakage vill be measured and recorded.

9.2 System leakage shall include, but not be limited to, the following.

9.2.1 All valves within W System.

9.2.2 All accessible flanges within the system boundaries.

9.2.3 All instruments and flanges that couple flow instruments, v

s .

l 9.3 When measuring leakage on flanges which are lagged, the lagging shall be pierced down to the flange with an awl or similar tool, at the low point in the flange lagging, and the hole observed for leakage. Where any leakage night also run down the pipe, such as in a vertical run with a flange, the lagging shall also be pierced at the low point for the run and checked for leakage.

If any water is observed, the lagging shall be removed for further inspection.

9.4 Leaks seasured on instruments shall be recorded as the total leakage from the following.

9.4.1 The instrument itself.

9.4.2 All valves in the loop.

9.4.3 All connections in the loop.

9.4.4 All drain and vent points in the loop.

9.5 Leakage measured on valves shall include:

9.5.1 Body to bonnet seal.

9.5.2 Packing / stem leakage.

9.5.3 Leakage through capped connections for which the valve

()

is used as an isolation; i.e. , a capped vent or drain valve.

9.5.4 Valve body system flange in cases where valves are not welded in.

10.0 Data Recuired Leakage should be measured and recorded on Enclosures 13.2, 13.4 and 13.6.

Leakage shall be measured in ec/ min. or L/ min.

11.0 Acceptance Criteria Test will be completed when all valves, flange and instrument leakage have been recorded.

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Initicl/Date 12.0 Procedug 12.1 Running Pump #1, Recire. through WST

/ 12.1.1 Verify (W) aligned per Enclosure 13.1.

CAUTION: Extreme care must always be used when aligning (W) system to prevent draining WST.

/ 12.1.2 Start refueling water pump #1 and bring system up to normal pressure.

/ 12.1.3 Record leakage on Enclosure 13.2. Enclosure 13.2 valves and instruments are pressurized during #1 pump operation.

Valves must be pressurized before testing.

/ 12.1.4 Secure #1 pump.

/ 12.1.5 Close WST #1 drain (1W44), close #1 discharge direct (1W66) .

12.2 Run Recire. Pump 1A through WST

/ 12.2.1 Verify aligned per Enclosure 13.3. Use caution to prevent draining WST.

/ 12.2.2 Start (W) Recire. Pump 1A and bring system up to normal operating pressure and temperature.

l O 12.2.3 Record data on Enclosure 13.4. Enclosure 13.4 valves and I instruments are pressurized during Recire. Pump 1A operation.

I

/ 12.2.4 Secure Recire. Pump 1A.

/ 12.2.5 Close WST #1 Drain (lW44) .

/ 12.2.6 Close W Recire. Pump 1A discharge (lW42) .

12.3 Run W Recire. Pump 13 through WST.

/ 12.3.1 Verify W aligned per Enclosure 13.5. Use caution to prevent draining WST.

12.3.2 Start refueling water Recire. Pump 1B and bring system up

/_

to normal operating pressure and temperature.

/ 12.3.3 Record data on Enclosure 13.6 of any leakage on instruments or valves. Enclosure 13.6 valves are pressurized during Pump 1B operation.

/ 12.3.4 Secure Recire. Pump 13.

/ 12.3.5 Close W Recire. Pump 13 discharge (lW38) .

/ 12.3.6 Close WST #1 drain (1W44) .

12.3.7 Return all valves to normal operation position (OP/1/A/6200/14).

_/

(J)

4_

13.0 Enclosures 13.1 Valve Alignment List Pump #1 13.2 Valve Data Sheet Pump fl Operation 13.3 Valve Alignment List Recire. Pump 1A  !

13.4 Valve Data Sheet Recire. Pump "1A" Operation

' 13.5 Valve Alignment List Recire. Pump 1B 13.6 Valve Data Sheet Recire. Pump "13" Operation i l

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CCMPANY PT/1/A/4201/03 PAGE 1 0F 2 KI NUCLEAR STATION Enclosure 13.1 II 1 VALVE CHECKLIST 1.VE NO. VALVE NAu.E POSITTON INITIAL lW18 W Pump #1 Drain CLOSED. l 1W45 W Pump #1 Drain Header CLOSED lW15 W Pump #1 Drain Tell Tale CLOSED lW17 W Pu=p #1 Overflow CLOSED I lW16 W Pump #1 Vent CLOSED l lW21 W Pump #1 Discharge Direct to WST CLOSED 1W4 WST to Refueling Cavity #1 Cont. Isol. Outside CLOSED l I

lW19 W Pump 11 Discharge Throttle OPEN OPEN  ;

IW20 f W Pump #1 to KF Isolation i

1713 l Refueling Cavity #1 to W Pump #1 CLOSED

1. _ - N3 to W Isol. CLOSED l

lW24 l Refueling Cavity #1 to FWST Isolation OPEN l

lW53  !, W Pumo 41 to WST Check e OPEN IW56 WST #1 Recirculation Line Isol. OPEN i

i lW61 l ND Pump Supply High Point Vent  ! CLOSED lW27A WST to ND Pump Isol.  ! OPEN IFw29 WST to ND Pumps Check Test Connection CLOSED  !

1FW65 f NS Pump Suction Low Point Drain  ; CLOSED ,

1 1FW64 NS Pump Suction Low Point Drain CLOSED 1W323 Refueling Water Isol. Loop  ! OPEN IWlA Refueling Water Loop Isolation  ! OPEN l lW2 WST to Refueling Cavity #1 Fill  ! CLOSED  !

^

1 W Pump #1 Inlet Isolation  ! OPEN 1rb4 Refueling Cavity #1 to W Pump #1 Low Point Drain CLOSED l IW57 W Pumo #1 Suction OPEN t

PT/1/A/4201/03 2 Enclosure 13.1 PAGE 2 0F JKE tER COMPANY

NUCLEAR STATICN

(! i VALVE CHECKLIST VALVE NAME POSITION INITIAL U.7E NO.

WST Recirculation Loop Isolation CLOSED N33A CZ,0 SED N3 WST to Refueling Cavity #1 Low Point W Pump #1 Suction Strainer Drain OPEN N58 Pump Discharge Nozzles CLOSED iS38 NS Suction from RWST CLOSED iS20A ND System to RWST Isolation CLOSED  !

iD35 22A,C fNCLoop3Disch.toNDContainmentIsol.Inside l CLOSED i CLOSED ND19A }NC Loop 1C to ND Pu=p 1A Contai:: ment Isol. Outside CLOSED ND48 ;ND Pu=p 1A Flush Supply Isol. l

!WST Safety Inj . Pumps l CLOSED nil 003 l Alternate Reciprocating Pump 1 Suction l CLOSED NI NV2223 'Cantrifugal Charging Pump Suction from W f CLOSED CLOSED NV221A lCantrifugal Charging Pump Suction from W f

! OPEN W44 lWST #1 Drain W66 !W Pump #1 Discharge Direct OPEN i

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PT/1/A/4201/03 Pags 1 of 2 Enclosura 13.2 Data Sheet Leakage r o i ,,, yn_ v.1 . . ve . ect.4 , v,4-4s1 lWlA Refueling Water Loop Isolation lW2 WST to Refueling Cavity #1 Fill & Drain lW3 WST to Refueling Cavity #1 Low Point Drair WST to Refueling Cavity #1 Cont. Isol.

1W4 Outside Refueling Cavity #1 to W Pump #1 Centain.

1W13 Isol. Outside Refueling Cavity #1 to W Pump #1 Low IWi*. om ., % ., 4 .,

1W15 W Pump #1 Drain Tell Tale lWie W Pump #1 Vent 1W17 W Pump #1 overflow 1W19 W Pump #1 Discharge Throttle IW20 W Pump #1 to KF Isolation 1W21 W Pump #1 Discharge Direct to WST lW22 NB to W Isolation 1W23 W Pump #1 Inlet Isolation 1W24 Refueling Cavity #1 to WST Isolation lW27A WST to ND Pump Isolation 1 w'a wsv en Nn o,~,s check 1W29 WST to ND Pumos Check Test Connection lW31 W Pump #1 Discharge Check .

lW323 Refueling Water Loop Isolation lW38 W Recirculation Pump "lB" Discharge lW44 WST #1 Drain  !

IW45 W Pump #1 Drain Header lW52 W Test Header Check 1W53 W Pu=p #1 to WST Check

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PT/1/A/4201/03 Page 2 of 2 Enclosure 13.2 Data Sheet Leakage Valve No. Valve Name cc/ min Initial lW54 W Pume to KF Check IW55 WST to WL Overflow Check 1W56 WST #1 Recirculation Line Isolation 1FW57 W Pump #1 Suction lW58 W Pump #1 Suction Strainer Drain 1W61 ND Pump Supply High Point Vent lW64 NS Pump Suction Low Point Drain 1W65 NS Pump Suction Low Point Drain lW66 W Pump #1 Disch. Direct to WST Throttle lW18 W Pump #1 Drain FLANGE FLANGE NAME

1. 1 Downstream from Valve 1W52

1. 2 Upstream from Valve 1FW56 i

N Signature Date O

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PT/1/A/4201/03 Enclosure 13.2 Instrument Data Sheet Instrument Leakage

90. vnser-ear iime ce/ min Tnicia 1WLT5340 W Stor. TK Level lWPX5320 W Recire. Pump "3" Suct. Press.

1WG5290 W Pump #1 overflow Indication in1T5000 W Storage Tank Level 1NLT5010 W Stor. TK Level lh1T5020 W Stor. TK Level lWPG3090 W Purification Pump Disch.

1 NPG5100 W Purification Pump Disch.

lWTE5110 W Purification Pump Disch. Flow lWPG5110 W Purification Pump Disch. Flow l

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PT/1/A/4201/03 Enclosure 13.3 PAGE 1 0F 1

, CCMPANY

NUCLF.AR STA!!CN Valve Alignment List Recire. Pump 1A II . . 1 VAL 7E CHECKLIST 7AL7E NAME POSITION INITIAL (L7E NO.

1FW34 FW Recire. Pu=p Drain Header Tell Tale CLOSED ,

I FW Loop Isol. CLOSED 1FW323 1

1FW33A FWST Recire. Loop Isol. OPEN l i

FWST Recire. Loop Isol. OPEN 1FW49B 17448 F4 to NM Isol.  ! CLOSED FW Recire. Punp 13 CLOSED 1FW37 1

CLOSED 1FW60 l F4ST Pu=p #1 Suction Strainer Header 1F441 FW Recire. Pu=p LA Suction OPEN fFWRecire.Pu=p1AVent l CLOSED 1FW43 1

• FW Recire. Pump 11. Discharge  ! OPEN 1FW42

-- t l OPEN

1. _, lFWST#1Recire.LineIsol.

1FW40 lFWRecire.Pu=p1ADrain i OPEN FW Recire. Pu=p 13 Drain i CLOSED IFW36 lFWRecire.PumpDrainHeader i CLOSED 1FW35 OPEN 1FW44 fFWST#1 Drain i

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PT/1/A/4201/03 Enclosure 13.4 Data Sheet Leakage Valves cc/ min. Initial 1W33A WST Recirculation Loop Isolation /

lW40 W Recirculation Pump "lA" Drain /

1W493 WST Recirculation Loop Isolation /

lW43 W Recirculation Pump "lA" Vent /

lW60 WST to W Recire. Pump "1A" Test Conn. /

lW48 W to NM Isolation /

lW50 W Recire. Pump "lA" Discharge Check /

lW34 W Recire. Pump Drain Edr. Tell Tale /

lW41 W Recirculation Pump "lA" Suction /

lW42 W Recirculation Pump "1A" Discharge /

lW35 W Recirculation Pump Drain Header /

Instruments lWPG5060 W Recirculation Pump "A" Discharge /

l 1NPI5300 W Recirculation Pump "A" Suct. Press. /

lWPX5310 W Recirculation Pump "A" Suet. Press. /

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PT/1/A/4201/03

- Enclosure 13.5 PAGE 1 0F 1 WER COMPANY c NUCLEAR STATICN Valve Alig::=ent List Recire. Pump 13 NI. 1 VALVE CHECKLIST 7ALVE NAME POSITION INITIAL M 7E NO.

W Recire. Pu=p 1A Suction CLOSED l lW41 li W Recire. Pu=p 13 Suction i OPEN 1W37 W Recire. Pump 13 Drain OPEN l 1W36 l l W Recire. Pu=p 1A Drain I CLOSED 1W40 8

W Recire. Pu=p Drain Header CLOSED IW35 W Recire. Pump 13 Discht.rge CLOSED 1W38 WST #1 Drain l OPEN lW44 I

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PT/1/A/4201/03 Enclosure 13.6 Data Sheet Leakage Valves cc/ min. Initial lW38 W Recire. Pump "13" Discharge /

1W51 W Recire. Pump "lB" Discharge Check /

lW39 W Recire. Pump "13" Vent /

lW37 W Recire. Pump "lB" Suct. /

lW36 W Recire. Pump "lB" Drain / ,

Instruments lWPG5070 W Recire. Pump "B" Disch. /

1 WPX5330 W Recire. Pump "B" Suet. Press. /

lWPX5320 W Recire. Pump "B" Suct. Press. /

ISTS5080 W Recire. Flow /

InTE5080 W Recire. Flow /

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PT ..

FOR INFORMATION DUKE POWER COMPAhi McGUIRE NUCLEAR STATION LIQUID WASTE SYSTEM LEAKAGE CHECK PROCEDURE 1.0 Procedure for To periodically test the Liquid Waste System outside containment leakage.

2.0 References 2.1 NUREG 0578 2.2 OP/0/A/6500/01 2.3 OP/0/B/6200/26 2.4 Liquid Waste (WL) System drawings:

MC-1565-1.0 MC-1565-1.1 MC-1565-2.0

> MC-1565-2.1 MC-1565-3.0 MC-1565-7.0 MC-1565-8.0 MC-1565-8.1 3.0 Time Required 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> at intervals not to exceed each refueling cycle.

4.0 Prerequisite Tests None 5.0 Test Equipment 5.1 Graduated cylinders 5.2 Stopwatch 5.3 Hose and connections for 3/4" test vent for Section 12.1.

6.0 Limits and Precautions 6.1 Follow Health Physics procedures when collecting and disposing of potentially contaminated leakage during this test.

7.0 Required Unit Status None I 8.0 Prerequisite System Conditions i

Stated in each section.

~- >

l

s 9.0 Test Method 9.1 While systems are operating in the recirculation mode, all leakage from the system shall be measured and recorded.

9.2 System leakage measurements shall include, but not be limited to, the following:

9.2.1 All valves listed in Endlosure 13.1 through 13.5.

9.2.2 All accessible flanges within the system boundaries.

9.2.3 All instrument loops that are normally valved in within the system boundaries.

9.2.4 Pump packing / seals.

9.3 When measuring leakage on flanges which are lagged, the lagging shall be pierced down to the flange with an awl or similar tool, at the low point in the flange lagging, and the hole observed for leakage. Where any leakage might also run down the pipe, such as in a vertical run with a flange, the lagging shall also be pierced at the low point of the run and checked for leakage. If any water is observed, the lagging shall be removed for further inspection.

~ 9.4 Leakage measured on instruments shall be recorded as the total leakage from the following:

9.4.1 The instrument itself.

9.4.2 All valves in the loop.

9.4.3 All connections within the loop.

9.4.4 All drain and vent points in the loop.

9.5 Leakage measured on valves shall include:

9.5.1 Body to bonnet seal.

9.5.2 Packing / steam leakage.

9.5.3 Leakage through capped connections for which the valve is used as an isolation; i.e. , a capped vent or drain I

! valve.

9.5.4 Valve body to system flanges in cases where valves are not welded in.

! 10.0 Data Required .

All leakage will be recorded on data sheets, Enclosures 13.2 through 13.11. Total System Leakage will be calculated and recorded on Enclo-sure 13.1.

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' 11.0 Acceptance Criteria This test will be considered satisfactory if the total CS system i leakage, external to the containment, from valves, system flanges, and pump packing / seals does not exceed 1 gpm. Notify the Duty ,

Operating Engineer if this test fails to meet the acceptance criteria.

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/ 12.0 Procedure U' 12.1 Mechanical Penetration M360 of WL system 12.1.1 Reactor Coolant Drain Tank is accessible.

12.1.2 Assure that lWL38 is closed.

12.1.3 Connect temporary hose to pressurize Reactor Coolant Drain Tent vent line at test vent downstream from lWL40.

12.1.4 Perform leakage measurements on components listed in Enclosure 13.2 and record.

12.1.5 Disconnect temporary hose from test vent form Section 12.1.3.

12.1.6 Realign lWL38 as required for normal operation.

12.2 Mechanical Penetration M375 of WL system 12.2.1 Place Reactor Coolant Drain Tank Pumps in normal opera-tion mode per OP/0/A/6500/01, Section 4.0.

12.2.2 Perform leakage measurements on components listed in Enclosure 13.3 and record.

12.2.3 Realign Reactor Coolant Drain Tank Pumps as required.

12.3 Mechanical Penetration M374 of WL System 12.3.1 Assure sufficient level in Containment Floor and Equipment Sump.

s, ,

12.3.2 Pressurize lines by starting Containment Floor and Equipment Sump Pumps referencing OP/0/A/6500/01, Section 6.0 lineup.

Put CF and ES pumps in manual operation.

12.3.3 Perform leakage measurements on components listed in Enclosure 13.4 and record.

12.3.4 Realign Containment Floor and Equipment Sump Pumps as required.

12.4 Mechanical Penetration M221 of WL System 12.4.1 Assure lWL393 is closed.

12.4.2 Place Ventilation Unit Condensate Drain Tank in normal operation mode per OP/0/A/6500/01, Section 5.0.

12.4.3 Perform leakage measurements on components listed in Enclosure 13.5 and record.

12.4.4 Realign lWL393 as required.

12.4.5 Realign Ventilation Unit Condensate Drain Tank as required.

/"'

V)

+wr - -

ytv - -'

-ee -+y-9a~ r rp

N 12.5 Waste Drain Tank of WL Sys6em 12.5.1 Place Waste Drain Tank in Recirculation Operation per OP/0/B/6200/26, Section 5.0.

12.5.2 Perform leakage measurements on components listed in Enclosure 13.6 and record.

12.5.3 Realign Waste Drain Tank as required.

12.6 Recycle Monitor Tank A of WL System 12.6.1 Recirculation of Recycle Monitor Tank A via the WL Evap. Cond. Filter with the W1 Etap. Cond. Demin. using RMT Pump A per OP/0/B/6200/26, Section 6.7.

12.6.2 Perform leakage measurements on components listed in Enclosure 13.7 and record.

12.6.3 Realign Recycle Monitor Tank A as required.

12.7 Recycle Monitor Tank B of WL System 12.7.1 Recirculation of Recycle Monitor Tank B using the RMT Pump B per OP/0/B/6200/26, Section 6.4.

12.7.2 Perform leakage measurements on components listed in s Enclosure 13.8 and record.

m- 12.7.3 Realign Recycle Monitor Tank B as required.

12.8 Waste Evaporator Feed Tank Sumps and Residual Heat Re= oval and Containment Spray Rooms Sump of WL System 12.8.1 Assure sufficient level exists in the Waste Evaporator . _ _ _ .

Feed Tank Sump Pumps and Residual Heat Removal and Containment Spray Rooms Sump Pumps.

12.8.2 Complete Section 3.0 down to Section 3.1.8 of OP/0/B/6200/26.

12.8.3 Check last date Section 3.2.1 of OP/0/B/6200/26 was completed and if still valid.

12.8.4 Place Waste Evaporator Feed Tank Sump Pumps in manual and start.

12.8.6 Perform leakage measurements on components listed in Enclosure 13.9 and record.

12.8'.7 Realign Waste Evaporator Feed Tank Sump Pumps and Contain-ment Spray Rooms Sump Pumps as required.

O_-

\~

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() 12.9 Auxiliary Waste Evaporator Feed Tank of WL System 12.9.1 Align Aux. Waste Evaporator Feed Tank Pump for operation per Section 7.0 of OP/0/B/6200/26.

12.9.2 Perform leakage measurements on components listed in Enclosure 13.10 and record.

12.9.3 Realign Auxiliary Waste Evaporator Feed Tank Pump as required.

12.10 Auxiliary Floor Drain Tank of WL System 12.10.1 Align Auxiliary Floor Drain Tank Pump for operation per Section 7.5 of OP/0/B/6200/26.

12.10.2 Perform leakage measurements on components listed in Enclosure 13.11 and record.

12.10.2 Realign Auxiliary Floor Drain Tank Pump as required.

12.11 Liquid Waste Total System Leakage 12.11.1 Record Total Leakage from enclosures.

13.2 - 13.11 on Enclosure 13.1.

12.11.2 Calculate Total System Leakage.

12.11.3 Determine shether Total System Leakage for the Liquid

-} '

s_ / Waste System meets Acceptance Criteria.

12.11.3.1 If Acceptance Criteria is not met, take correc-tive action as stated in Section 11.0 of this procedure. --

13.0 Enclosures 13.1 Total System Leakage 13.2 Leak Check Worksheet 13.3 Leak Check Worksheet 13.4 Leak Check Worksheet 13.5 Leak Check Worksheet 13.6 Leak Check Worksheet 13.7 Leak Check Worksheet 13.8 Leak Check Worksheet 13.9 Leak Check Worksheet 13.10 Leak Check Worksheet 13.11 Leak Check Worksheet O .

Page 1 of 1 Liquid k'aste System Leakage Enclosure 13.1 PT/

Liquid Waste System Leakage from:

cc/ min. ec/hr.

{,

Enclosure 13.1 _

4 ,

13.3 ,

4 13.4 __

13.5 _

13.6 j 13.7 13.8

i 13.9

. i 13.10 13.11 TOTAL WL SYSTEM LEAKAGE:

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,,.w. ....ya-... ---_:,,,,-.,,,,y,1 m i -.erm... - - - .y - - . r-

- r e.e -*,,-,s-+e.-m-..ww.--..+ . , , _ _ _ .,-,--,.,,.,.--,.w,,,-._ .

DUKE P0kER CDMP.dT  ?).GE 1 CF 1 MCOUIRE NUCLEAR STATION UNIT #1 ENCLOSURE 13.2 LEAK CEECK WORKSHEET

\

LEAKAGE YES NO CC /MT'T . cc/uu TC'

• CCMONUT 1k138 NCDT Vent Isol.

lWL40 NCDT Vent Cont. Isol. Test Vent lb139A NCDT Vent Inside Cont. Isol.

lWL389 NCDT Vent Cont. Isol. Outside Test Vent 1WL41B NCDT Vent Outside Cont. Isol.

lWL42 NCDT Vent Lotr Point Drain ik148 SCDT Vent Gas Sample Inlet lWIME 5820 - Waste cas Vent Rdr. Samole l I i

1WL51 NCDT Vent' Gas Sample Outlet 1bl405 NCDI Vent Auto Drain Isol.

lWLLS 5890 - WG Vent Drain Line Level Control  ! .

lWL404 WL Evap. Condensate Trap Inlet TOTAL LEAKAGE ENCLOSURE 13.2 CALCULAIED BY DATE

.--.--._,-p. .e_.,.-., ~,.,,.,,,v.7 ., .-_.,_,_..__..r- , , _ , _ _ . . .

?AGI 1 G7 1 DUr2 FCk & COMP.uT MCOUIRE .31 CLEAR STATION w.IT #1 ENCLOSURE 13.3 LEAK CHECK k'0PISHEET PT/ f

\

tEAxAcE l vrs NO

~

cc /MT'7. cc/un. TS ev aggn-Ik1859 Mich Point Vent & Drain for Finch & uvam lb113 NCDT Pumps Disch. Outside Cont. Isol.

r5 -. v,-,. * *,-

iet?A? yerve o...,e s n4, a _ <--- < - - . -

ik128 NCDT Pumos Disch. from Cont. Low Point Drain iblFT 5710 urn? on-r ne ca. Y a 4,7. r's ib1261 SLDT Pumps Disch. from Cont. Flow Metee Bvoass ik1260 NCDT Pumps Disch. from Cont. Flow Meter Outlet 1k1860 High Point Vent Drain for Flush & Hydro 1k131 NCDT Pumps. Disch. to WEET l'a1861 Hihg Point Vent & Drain for Flush & Hydro lk'L30 NCDT Pumps Disch. to WST l

l l

TOTAL LEAKAGE ENCLOSURE 13.3 __

CALCULATED BY:

DATE

.-- r ~ . - , -

?.gGI 1 0; i DUKE P0k'ER COMPANY MCGUIRE NUCLEAR STATION DIT J1 ENCLOSURE 13.4 LEAK CHECK WORKSHEET PT/

LEAKAGE ves NO CC/ MIN. cclu?. TTS ccvc0NTN-1WL852 High Point Vent & Drain for Flush & Hydro lWL390 RB Sump Pumps Cont. Isol. Outside Test Vent ik165 B-RB Sump Pump Disch. Outside Cont. Isol.

1k166 RB Sump Pump Disch to WEFT ik1353 High Point Vent & Drain for Flush & Hydro 1h174 RB Sump Pu ps Disch. to FDT Low Point Drain 1h1854 High Point Vent & Drain for Flush & Hydro 8

"~*  % -..

W em TOTAL LEAKAGE ENCLOSURE 13.4 CALCULATED BY DATE t

• * ' ' ~--w+ -

l

UKE PO*1ER CCMP.GT  ?).GZ 1 07 2 ,

MCOUIRE NUCLEAR STATION JJ'11T fl ENCLCSURE 13.5 LEAK CHECK '40RKSHEET PT LEAKAGE VES NO

~

cc/Mr7 cc/uy. TO COWmini ik1800 Drain for Flush Hydro l'a13223 Cent. Vent Unit Drains outside Cont. Isol.

1k1387 Cont. Leak Rate Test Inst. Isol.

1k1371 Cont. Leak Rate Test Inst. Calibration Needle TURSINE FLO'4 METER FLANGES 1k1881 High Point Vent & Drain for Flush & Hydro lk1415 AB Air Handling Units Cond. Drain to VUCDT 1k1880 High Point Vent & Drain for Flush & Hydro ik1361 VUCDT Inlet 1L'4308 VUCDT Tell Tale lk'L310 VUCDT Tell Tale .

Ik1LS5590 VUCDT A Level ik1LT5590 VUCDT Level lk1P5590 VUCDT Level ik1LS5591 VUCDT Level 1k1302 VUCDT Pu=p 1A Suction Isol.

Ik1305 VUCDT Pump 13 Suction Isol.

Ik' LPG 5610 VUCDT Pump A Disch.

Ik1PG5620 VUCDT Pump B Disch.

1k1303 VUCDT Pump 1A Discharge Check 1k1304 VUCDT Pump 1A Discharge Isol.  !

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? IGE 2 07 2 DUKE P0k*ER CCP?AW l

MCGUIRE WC'.E.G STATION l

1) SIT ill ENCLOSURE 13.5 LEAK CHECK WORKSHEET PT/

l LEAKAGE M

NO cc/wTN. cc rut T. 3 ,

cev-;ogy:-

ikL306' VUCDT Pump 13 Discharge Check Ik'L307 VUCDT Pu=p 1B Discharge Isol.

1WL323 VUCDT Rad. Monitor Inlet Isol.

1k1324 VUCDT Rad. Monitor Bypass lWLRT3650 Radiation Transmitter lWLP5650 Radiation Transmitter 12325 VUCDT Rad. Monitor Outlet Isol.

Ik1320 VUCDT Rad. Monitor Outlet 1WLFE5900 VUCDT Flow lEFT5900 VUCDT Flow 1WLP5900 VUCDT Flow IWLCR5900 VUCDT Flow IWL877 Figh Point Vent & Drain for Flush & Hydro lWL871 High Point Vent & Drain for Flush & Hydro lWL882 High Point Vent & Drain for Flush & Hydro lWL872 High Point Vent & Drain for Flush & Hydro , , ,

i lb1870 High Point Vent & Drain for Flush & Hydro j ILW890 High Point Vent & Drain for Flush & Hydro O .

TOTAL LEAKAGE ENCLOSURE 13.5 CALCULATED BY DATE

?J.GI 1_ 07 2 DUKE POWER COMPANY '

MCOUIRE NUCLE.G STATION ITNIT #1 ENCLOSURE 13.6 I.EAK CHECK WORKSHEET PT/

LEAD GE YES NO CCIMIN- CC#"" - ~

CO'#20NEC lWLl47' WDT Pumps Disch. to WEFT 1WLl42 WDT Pumps Disch. to Sample Sink 1k1144 b1T Pumps Recire. Throttle IWL338 WDT Pump B Vent & Flush Isol.

Ik1139 WDT Pumps Vent & Fill Overflow lk1341 WDT Pump A Vent & Flush Isol.

1WL140 WDT Pumps Vent to Atmos.

lWLl41 WDT Pu=ps Flush Supply ik1138 WDT Pu=p's' Drain lWL337 WDT Pump B Drain 1k1137 WDT Pump A Drain 1WL136 WDT Pumps Drain Tell Tale 1k1134 WDT Outlet Isol.

lWL135 WDT Pumps Drain to b1T 1k1133 WDT Drain to WEFT Sump A 1k1131 WDT Diaphragm Drain 1k1329 WDT Overflow Siphon Break lk1128 WDT Diaphragm Sample Vessel Inlet i IWL129 WDT Diaphragm Vent Outlet ik1127 WDT Diaphragm Sample Vessel Return i

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.=- -<-r.-- - -_

?!.GZ 2 CF 2 DUKE POWER COM?NiY MCGUIRE ::UCLEAR STATION P L11T di ENCLOSURE 13.6 LEAK CHECK WORKSHEET PT/

LEAXAGE vig NO

~

cc/Myy, cc/uo. IM cegogy 1WL874' High Point lWL130 WDT Diaphragm Vent Inlet ..

WASTE D?AIN TANK PUMP FLANGES A & B C'alPG5790 WDT Pu=p B Disch.

0'alPG5070 WDT Pump A Disch.

0'alPG5061 WDT Level Ok1PS5060 k1T Level ik1289 Unit 2 Pri. Sa=ple Sink to WIFT Sump B ik1300 Unit 2 Pri. Sample Sink to WEFT Sump A ik1412 Floor Drain from Annulus to WEFT Sump A ---

Ik1LT5940 Annulus Floor Drain Loop Level ,

ik1PS5941 Annulus Floor Drain Loop Level i

TOTAL LEAKAGE ENCLOSURE 13.6 CALCULAIED BY DATE ,

6

- - - - - - - , , , . , - , ,w., ,n , ,---- n, .y., ,- --.- .. ..-

1. g# O TEST TARGET (MT-3)

+

1.0 ls a tu

. GigingpL=-

l,l p

• bb l.8 1.25 1.4 '

l.6 l

= c ,

1. 4 <> <4 Wh?

wi,  %>vW 4

,p

%+%9 p A

%,+Q@ ---

%,f.y, TEST TARGET (MT-3) 1.0 l# EH EM d 8" LE

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OUKE POWER COMPANY ?f.GZ l 07 2 XCOU KE NUC' EAR STATICN

, UN!! 11

! ENCLOSURE 13.7 LEAK CHECK WORKSHEET PT/

LEAKAGE  !

YES NO cc /MT'i . cc/uy. TO t ev:T: y-ik1381 RMT Pu=es Disch. to MIT's lWL'.19 RMT Outlet Header Isol.

Ik197 WL Evap. Condensate Filter to 21T's ik1110 RMT Pu=ps Disch. to 'a'L Evap.-C ndensate Drain ik1106 RMI A AbcVe Diaphragm Vent ik1107 RMT A Above Diaphragm Drain 1k1108 RMT A Vent 1k1101 RMT A Outle: Isol.

l'a1330 RMT Pu=p A Suct. Isol.

RMT PUMP A FLANGES Ik'I . 5 High Point Flush & Hydro i

1*a1332 RMT Pu=p A Drain 1k1333 RMT Pu=p A Vent & Fill Overflow ILW334 RMT Pu=p A Vent at Atmos.

Ok1PG5690 RMT Pu=p B Disch. .

Ok1LT5290 RMT A Level _

OkiPS5291 RMT A Level IWL386 RMT Pu=p A Disch. to Sample Sink l'a1420 RMT Outlet Header Future Connection ik1109 RMT A Drain to WEFT Su=p B -

D e

e- --- - - -w-w w-, ,,,,-n, --

IJ.GZ  ? 07 7 DUKE PCb'ER CCSANY

! .v.CGUIRE NT.* CLEAR STATION I

CCI #1 ENC 1,05URE 13.7 LEAK CHECK '40RC HEET PT/

LEAKAGE YES NO cc/MTN. cc/un. IS.'

CCMONE.-

l'.1124 RMT Pu=ps Disch. to NB Evap. Feed Demin.

ik1123 FJiT Pu=ps Disch. to Mf4ST 1 ik1122 MIT Pu=ps Disch. to Mr4ST Z ik13S2 F2:I ?u=ps Suction Cressover 4

1k1365 High Point Flush & Hydro Ik1422 kl Evap. Dist. Cooler Future Connection 1*.136 kL Evap. Cond. Ce=in. Drain to VEFT Tell Tale I

l'.195 k1 Evap. Condensate Filter D' rain Tell Tale 1

l c a pr,50?.0 L'3ste rveo. C.? n d . riir.r D/o ll 4

I _

l TOTAL LEAKAGE ENCLOSURE 13.7 I CALCI.1ATED BY DATE b

pggg j cy ~y

• f.<E ?OkER COMPANY MC'iUIRE NUCLEAR STATION "N!! 11 ENCLOSURE 13.8 LEAK CHECK WORKSHEET PT/

LEAKAGE YES NO rcIMT't. rrlu9. KK1* -

C McND'~

lk1112 RMT's Drain to WEFT Sump B _

1WLill RMT's Drain Tell Tale ik1120 MT Pumps Drain to WEFT Sump ik1119 RMT Pumps Drain Tell Tale

(

1WL102 RMT B Above Diaphrap Vent y lWL103 ET B Above Diaphrap Drain i

i 1WL104 RMT B Vent e

~ .

WL349 NB Evap. Distillate to MT A l l lbl96 NB Evap. Distillate to RMT B 1%1116 RMT P me B Disch. to Samole Sink l ,

lWLil5 RMT Pump B Vent to Atmos.

l 1b1113 RMT Pump B Suct. Isol.

lWL99 RMT B Outlet Isol, lWL375 RMT Pumps Recire. Crossover QMT o, - B Dis,b. *o ci 'l e c 91r 1L1116 ik1117 RMT Pump B Vent & Fill Overflow lWL118 RMT Pump B Drain ik"L105 RMT B Drain to WEFT Sump B RMT PUMP B FLANGES Ok1LT5810 RMT B Level l OWLPS5811 RMT B Level , , , _, ,

OWLPG5300 RMT Pump A Disch. -- _ _ _ _ .

TOTAL LEAKAGE ENCLOSURE 13.8 CALCUl\TED B,Y l

! DATE l . _ _ - -_

?AGF. 1 07 2 DUKE PCWER COMP /C;Y MCSUIRE NL'CLE.G STA!!ON

]' NIT ill ENCLOSIRE 13.9 LEAK CHECK WORKSEEET PT/

! LEAKAGE VES 30

~

CC /MI?i . ccluu. MCC cavagy y-FLANGES UPSTREAM OF k1392 1kL392 ND & NS Su=p Pumps to RC Disch. Crossover lWL366 Flush & Hydro l'a1198 XD & NS Rec =s Su=p Pu=ps to FDT Ik1379 Flush & Hydro ik1197 ND & NS Rec =s Su=p Pu=ps to WEFT Ik1194 ND & NS Rooms Su=p Pu=p 2A Disch. Isol.

1k1196 ND & NS Rooms Su=p Fump 2B D' isch. Isol.

s'p s sg u---e o,- e..--e i nmek ve1 1%t107 1k1190 ND & NS Roc =s Sumo Pu=o 1A Disch. Isol.

1k1188 WEFT Su=p Pump 3 Disch. Isol. .

Ik1903 Drain ik1902 Drain 1k1186 WEFT Sump Fump A Disch. Isol.

Ik1156 WEFT Vent Outlet ik1154 k' EFT Vent Inlet ik1378 Flush & Hydro Ok1LS5350 ND & NS Rooms Sump Level I

1 Ok1LS5360 ND & NS Rooms Sump Level Ok1LT5950 ND & NS Rooms Sump Level I

l 2 c ~- 2 n2

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CUKE FC'E COM?XiY ~ i

':CC'JI?l : JCI. EAR STATION NIT 41 ENCLOSURE 13.9 LEAK CHECK '40RKSHEET PT/

LEAKAGE YES NO CC/ MIN. ec/up. TCC C O V ?O N E :""

O'.1.LS5330 SD & 55 Rooms Su=p Level Da L55340 53 & SS Roc =s Su=p Level

?

0'4LLS5320 'aT.FT Sump 3 Level __

I OnLS5310 tiEFT Su=p A Level i _

l I

M I

l w) TOTAL LEAKAGE ENCLOSURE 13.9 CALCULATED BY DATE

yj,GI 1 07 2 OL*KE PO%T2 CCMP.C;Y MCG'JIRE ::L'CI. EAR STATION E l! 41 E::CI.05URE 13.10 LEAK CHECK %CF5 SHEET PT/

I LEAKAGE i YE9 NO cc /MT'i . cc ruw . TO_*

cev:cyn:-

l'a1418 'a'L Evap. Feed Filters Outlet Future Connectior lh*L1104 Flush & H/d ro High Point 11.11103 Flush & Hydro High Point 1511102 Flush & Hydro High Point 1k11101 Flush & Hydro High Point l'a11100 Flush & Hydro High Point l'111'3 Low Point 1FLil23 Low Point 1511128 Flush & Hydro High Point i ILW446 Flush & Hydro High Point i

1k11013 Low Point Drain ,

1ku;012 Low Point Drain ik1482 Low Point Drain l'a1481 Low Point Drain ik1908 Low Point Drain ik1483 Su=p Low Point 11L479 Sunp Low Point iL'L1014 Sump Low Point l 1k1477 Sunp Low Point Ok1LS6100 Int. Radwaste Pipe Trench Susp A Level x

j l

l l

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---

,--r--. - ,. _.---._,- - , . _ , - - . - , -- , . - - - - - - . - -- -,-- r- - --

?).GT. 2 C7 2 DUKE PCWER C2:2A';T MCO'JIRE :CCLEAR STA!!CN

!! 11 E';CLOSL*RE 13.10 LIAK CHECK WCRKS' DIET PT/

LEAKAGE I YTC NO cc IvI't . cctuo. TC j

cav:cy v-Ok1LS6130 Int. Radwaste Pipe Trench Su=p A Level Ch1LS6110 Int. Radwaste Pipe Trench Surp 3 Level CLIP 6020 Aux. WEFT Level 0k1?6021 Aux. WIFT Level e

I?u447 Flush & Hydro ik1443 Flush & Hydro AUX. WEFT ? CMP FLA::GES l'ali4 9 Flush & Hydro i

I Ok1PG6030 Aux. WEFT Pu=p d103 Disch. Press.

l'a1450 Flush & Hydro Ik1452 Flush & Hydro Ok1FE6070 Aux. WIFT Recire. Flow 1k1455 Flush & Hydro i

IWL1126 Low Point Drain t

l'a11122 Low Point Drain .

Ik11114 Flush & Hydro Hizh Point

! IVL1115 Flush & Hydro High Point lWLill6 Flush & Hydro High Poinc ik11117 Flush & Hydro High Point Ik11118 Flush & Hydro High Point lWL416 kT Evap. Liquid 'a'aste Inlet Future Connection TOTAL LEAKAGE ENCLOSL~dE 13.10 CALCL1ATED BY r

. Di TE

- , - - - , - - - - - , .- - - y- - - -

C7 2

" MG7 1 DGI PCkE CO.  ?.CiY

. , , . . . v ,e ;_u -: ..A.. , .0...

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.c -.,

" NIT di ENCLOSURE 13.11 LF_E CHECK k'ORKSHEET PT/

l LFAKAGE t VT_ C_

liO cc NT'i . cc /uw . T_ . _ a. -

e e:c .7 ,--

ik11108 Flush & Hydro Hi:;h Poin:

1111107 Flush & Hydro High Point 17 1106 Flush & Hydro High Point D11105 Flush & Hydro High Point ik11120 Low Point Drain ik11124 Low Poin: Drain ik1431 Low Point Drain .

1 .

1ki'. 32 Low Foin: Drain O'.1?6000 Aux. FDT Level 1

0'.1P6001 Aux. FDT Level Ok1LS6060 In. Radwaste Fac'.li: V Su=o Level Ok1LS6040 In. Radwaste Su=1 Pump Control 17,464 Drain lE433 Drain D7.434 Drain ,

AUX. FDT POIP FLANCES Ik1435 Drain Ok1?G6010 Aux. LT.FT Punp #102 Disch. Press, i

ik1436 Drain ik1462 Drain

- - - -- - .em.- ._,--,- - - __ _

yg 2 Cy 2 3GE TUBER C0"PAliY StCOL'!RE NUCLEAR STATION E'iCLOS*3E 13.11 0 ;11! 1 1 LEAK CHECK WORK 3HEET PT/

LEAKAGE  !

veg i NO CC /vTN . ec luo . I-A 70e!,T:-

lh1433 Drain Ok1FE60SO Aux. FDT Recire. Flow ._

l iki'61 Drain i

EAD'. ASTE CHD11 CAL FEED SYSTD1 FLANGES l OblFE5090 Aux. ' EFT Pu=p 102 Flow to L&HST Filt.

1k1444 Drain i

1W'.460 Drain l .

1%1465 Drain f i l l Ok1LG6050 Radwaste Chem. Feed Tank Level ik11125 Lcw Point Drain 1b11121 Low Point Drain l

1k11127 Flush & Hydro High Point 1%11109 Flush & Hydro High Point 1k11110 Flush & Hydro High Point 1WL1111 Flush & Hydro High Point ,

1%11112 Flush & Hydro High Point 1%11113 Flush & Hydro High Point i

I. .

TOTAL LEAKAGE ENCLOSURE 13.11 CALCULATED BY DATE

- - - - . - , , - - --gn,--- - . . . .,,.n-c,-- ,...,-,-, , - , _ - ~ , - - - - , - - - - - - - - - - --

/N NRC REQUEST FOR INFORMATION TRANSMITTED BY LETTER OF JUNE 30, 1980 FROM B. J. YOUNGBLOOD CHIEF, LICENSING BRANCH NO. 1 DIVISION OF LICENSING

1) Additional Accident Monitoring Instrumentation a) Before fuel loading, an interim method is required when the high range noble gas effluent monitors are not yet installed and operable. You should describe the interim methcd, addressing item 2.1.8.b enclosed in our letter dated November 9, 1979, pages 31 to 36, providing infor-mation required in 1.A.l.a and 1.A.l.b for noble gas effluents and 2.A.1 and 2.A.2 for particulate and radioiodine effluents. Your response should contain a descriptive summary of the interim procedures for quantifying high level accidental radioactivity releases to meet the requirement in the Action Plan NUREG-0660, Appendix A, Table A.1, item (17) for II.F.1.(a).

b) By January 1, 1981, complete the installation of the high range noble gas effluent monitors II.F.1.(f) and provide the information required in item 2.1.8.b sections 1.A and 2.B given in the November 9, 1979 letter. Clarify that the steam dump / safety and containment hydrogen purge exhaust will have high range noble gas effluent monitors.

O(,) Response See Additional Accident Monitoring Instrumentation

2) Primarv Coolant Sources Outside Containment Before full power operation, provide a description of the method to be used during refueling outage leak rate tests and the weekly leak test procedure.

Discuss the test method to be used for each system or subsystem, such es hydraulic, mass spectrometer, freon, etc., and the acceptance criteria for the test. Compare the leak test criteria to area and effluent radiation monitor levels. Indicate the steps to be taken to minimize occupational radiation exposure, maintain test results, repair leaks and assure system completeness. Specify the staffing and training requirements.

Response

See Primary Coolant Sources Outside Containment

3) Post Accident Sampling Before full power operation prior to January 1, 1981, provide a descriptive summary of the interim provisions and procedures for sampling and. analyzing the reactor coolant and the containment atmosphere. Consider the modifica-r'] tions needed for the physical, chemical, as well as the radiological analysis

(, y steps. By January 1, 1981, provide a description and final system design of 02/06/81

_... . - . - . - _ . - . - . . - _ _ . . - .. . . - . . ._.. . - .-..-.- - . _.. _ _ _._. .... - _= ._

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the new accident level sampling panel, and modifications to the sample i

i handling and counting facilities to achieve analysis within the time specified in item 2.1.8.a given in the November 9, 1979 letter.

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See Post Accident Sampling

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NRC REQUEST FOR INFORMATION TRANSMITTED BY LETTER OF NOVEMBER 6, 1980 FROM R. L. TEDESCO, ASSISTANT DIRECTOR FOR LICEN31NG, DIVISION OF LICENSING TMI related action plan item II.K.3, Final Recommendations of B&O Task Force, sub item C.3.12 recommends that for Westinghouse designed reactors confirma-tion be obta. that there is an anticipatory reactor trip on turbine trip.

By letter of October 10, 1980 you provided us with your present position on this item. The primary objective of the B&O Task Force recommendation was the reduction of PORV challenges to the extent feasible by design and operational procedures. Retention of the presently proposed full load rejection capability without the anticipatory reactor trip on turbine trip would not conform to this objective.

Based on our review of your design we have concluded that operation without the anticipatory trip action is not justified in view of the PORV challenges that would occur for turbine trips while operating near and at full power without the anticipatory trip action. Consequently, we require that you commit prior to fuel loading to install this trip function and have it in operation prior to issuance of a full power license.

Regarding item C.3.10 on anticipatory trip by-pass above 10 percent nominal power, we require a response to this item if by-pass above 10 percent is O planned.

Response

See Final Recommendations of the Bulletins and Orders Task Force.

02/06/81

-.- . - - _ _ .