ML20008D872

Forwards Updated Responses for Duke Power Co,Mcguire Nuclear Station,Response to TMI Concerns. Unit 1 Fuel Loading Is Scheduled for Dec 1980

ML20008D872
Person / Time
Site:	McGuire, Mcguire
Issue date:	10/10/1980
From:	Parker W DUKE POWER CO.
To:	Harold Denton, Youngblood B Office of Nuclear Reactor Regulation
Shared Package
ML20008D873	List: ML20008D872 ML20008D874
References
RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 8010230408
Download: ML20008D872 (32)

Text

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Duxz POWER COMPANY

. Powra Bur.otwo 422 Sourn Caunca Stazzi. CuAntortz. N. C. asa4a

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- w,w- e. pannen.sn. October 13, 1980

%ct P#ESIDENT TELEP=oNC: AaCA 704 Setau Paoouction 373-4083 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 Branch No. 1 Re: McGuire Nuclear Station ~

Docket Nos. 50-369

Dear Mr. Denton:

O Enclosed is the results of an evaluation of compliance of McGuire Nuclear Station with the regulations contained in Title 10, Code of Federal Regulations, Parts 20, 50 and 100. Although the evaluation is generally applicable to both units, it has been prepared specifically to demonstrate compliance of Unit I with the regulations.

If there are questions regarding this matter, please advise.

Very truly yours, s/ William i. Parker, Jr.

William O. ktrker, Jr.

GAC:scs Enclosure n

U col 3410230 i

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Mr. Harold R. Denton, Director October 13, 1980 O Page Two WILLIAM 0. PARKER, JR., being duly sworn, states that he is Vice President or Duke Power Company; that he is authorized on the part of said Company to sign-and file with the Nuclear Regulatory Commission this document entitled " Compliance of McGuire_ Nuclear Station Unit I with the NRC Regula-tions of 10CFR Parts 20, 50 and 100"; and that 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 i

Subscribed and sworu to before me this 13th day of October, 1980. _

i s/ Sue C. Sherrill i

Notary Public My Commission Expires:

September 20, 1984 4

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DUKE POWER COMPANY MCGUIRE NUCLEAR STATION Response to TMI Concerns Changes and Corrections Remove These Pages Insert These Pages Table of Contents pg, 11 09/08/80 Table of Contents pg. 11 10/10/80 Table c. Contents pg. iii 10/10/80 I-1 08/06/80 I-1 10/10/80 I-3A 09/08/80 I-3A 10/10/80 I-6 09/08/80 I-6 10/10/80

I-9 09/08/80 I-9 10/10/80 I-10 09/08/80 I-10 10/10/80 I-13 I-13 10/10/80 I-14 I-14 10/10/80 I-15 09/08/80 I-15 10/10/80 '

II-3 09/08/80 II-3 10/10/80 II-4 09/08/80 II-4 10/10/80 II-12 08/06/80 II-12 10/10/80 II-13 07/18/80 II-13 10/10/80 Carry Over II-13A 07/18/80- II-13A 10/10/80 Carry Over II-13B 07/18/80 II-13B 07/18/80 Carry Over II-13C 07/18/80 II-13C_ 07/18/80 Carry Over II-130 07/18/80 Carry Over II-17 07/18/80 II-17 10/10/80 II-19 08/06/80 II-19 10/10/80 II-19A 10/10/80 III-1 III-1 10/10/80 Appendix A Appendix A Station Directive 3.1.32 Station Directive 3.1.32 (10/10/80)

Appendix D Appendix D July 2, 1980 letter from July 2, 1980 letter from B. J. Youngblood (08/06/80) B. J. Youngblood (10/10/80)

August 25, 1980 letter from August 25, 1980 letter from R. L. Tedesco (09/08/80) R. L. Tedesco (10/10/80)

September 17, 1980 letter from Ralph Birkel (10/10/80) l - September 17, 1960 letter from

( (g) B. J. Youngblood (10/10/80) l  %/

l- Appendix E Special Low-Power Test Safety Evaluation L

Appendix A McGuire Nuclear Station Procedures

[\d ) Station Directive 3.1.4, Conduct of Operations Sta: ion 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 the Shift Technical Advisor  ;

j Station Directive 3.1.32, Station Safety Engineering Group l Appendix B Control Room Design August 15, 1980 letter from Mr. W. O. Parker to  !

Mr. H. R. Denton l Appendix C NRC Requirements for McGuire Nuclear Station September 27, 1979 letter from D. B. Vassallo to all Pending Operating License Applicants November 9, 1979 letter from D. B. Vassallo to all Pending Operating License Applicants s ,) March 28, 1980 letter from H. R. Denton to all Power Reactor Applicants and Licensees NUREG-0694: TMI-Related Requirements for New Operating Licenses Appendix D 4RC 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. S. Youngt- od to W. O. Parker

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Appendix E Special Low-Power Tests Safety Evaluation 1

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SHlFT TECHNICAL ADVISOR -

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References:

NUREG-0578'- 2.2.lb Action Plan - I.A.1.1 p A technical advisor to the shif t supervisor will be present on all shif ts and

} .available to the control room within ten minutes. The shift technical advisor's.

l primary duty will' be to provide evaluatica and assessment of both norac1 and unanticipated transients. The shift technical advisor will be detached from

- and independent of the normal line responsibility for plant operation.

' The shift technical advisors will-be selected from the group of licensed senior j reactor operators!at McGuire. All of the McGuire SRO applicants have received j additional' simulator and academic training. The simulator training included functioning as the STA during various transients,and the academic training

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included instruction in heat transfer, fluid flow, thermodynamics, and plant

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transients.

l' i- In addition eleven McGuire SRO licensees / applicants are enrolled in a special l STA training program based on the April 30, 1980 guidelines established by l

1NPO. This. program commenced on September 29, 1980 and will consist of four weeks of instruction. The first three weeks will consist 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.

I This training program includes instruction in the following topics: STA

} - responsibilities and accountabilities, management and supervisory skills; i transient and accident analysis; seismic monitoring; plant status monitoring; l plant metallurgy; plant chemistry; instrumentation and control, theory; PWR 4 heat transfer; PWR thermal and hydraulic transient response; small break LOCA analysis; and mitigating reactor core damage. This training program will be

! incorporated into the standard SRO training program for future McGuire SRO l

candidates.

j' Duke Power Company is actively pursuing the development of a program to provide

supplementary technical education for both the STA's and the shift supervisors.

j. This program will be taught at the college level and will be equivalent to j sixty semester hours in both basic engineering and plant applications of i engineering principles. A complete description of this program will be provided

by January 1, 1981.

I t The duties, responsibilities, and qualifications of the shif t technical advisor

! . have been defined in-the 'icGuire administrative procedure, Station Directive

! ' 3.1. 31. - This procedure is provided in Appendix A. l l-

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I The event investigator prepares a report describing the cause of the event and

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(,,j any relevant plant behavior, and outlining proposed corrective actions. The SSEG then reviews each report for accuracy and completeness, and assesses the adequacy of proposed corrective actions. The SSEG submits its report to the Manager of Nuclear Production, the Station Manager and the NSRB for review and for approval of corrective actions, to the Shift Technical Advisor for review  ;

l with regard to operating procedures, and to the Supervisor of Training for  ;

inclusion of relevant information in the training program.  !

i' The SSEG will be composed of four full time engineers assigned to the day shift.

The group will be staffed on a rotating basis from among experienced station ,

personnel and will be multidisciplined with expertise in the areas of instrumen- I tation, maim ce, operations, and technicel services. Additional information on the membership and duties of the SSEG is provided in Station Directive 3.1.32 which is included .a Appendix A.

Upon notification e an event, the General Office PC&L Section notifies company management, and may alert other General Office engineering and scientific support 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-mitted, if necessary. Information is provided to other organizations, including ,

NSAC, NRC, and the NSSS vendor. Detailed evaluations of plant transients are q performed, and event occurrence data is maintained. As appropriate, other

, engineering support groups review the LER and station event reporta for further i recommendations on corrective actions, and may interface with appropriate equipment vendors. The NSRB performs an independent review of the event report,

the LER, and the effectiveness of any follow-up actions.

INDUSTRY EXPERIENCE EVALUATION Figure 2 illustrates the flow path for information received concerning industry operating experience. Significant events will be brought to the attention of Duke Power Company by NSAC, NSSS vendors, other utilities, or the NRC. Informa-tica is distributed, as appropriate, to General Office engineering support gr e en for review and development of corrective actions and to the Training Services group for incorporation into the training program. The SSEG reviews the-informa-tion for applicability to the specific station, and makes recommendations to the Manager of Nuclear Production, the NSRB and the Station Manager in areas where i action may be necessary. The Station Manager then developes and implements

appropriate corrective actions with assistance from and review by the engineering support groups.

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REVISED SCOPE-AND CDITERIA FOR LICENSING EXAMINATIONS i

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Reference:

Action Plan - Plan I.A.3.1 The McGuire reactor operator and senior reactor operator license applicants have taken the NRC pre-revision written examinations. However, these applicants were required to pass the examinations with a grade of 80% over-all and 70% in each category. Those license applicants who must be reexamined will take the new revised examination. The McGuire license applicants have submitted letters to Mr. Paul Collins, NRC Operator Licensing Branch, which request the release of their examination results.

The McGuire license requalification program will include academic instruction in heat transfer, fluid flow, thermodynamics, and mitigation of accidents involving a degraded core. The program will include both normal and emergency operation instruction on the McGuire simulator and will meet the requirements of Enclosure 4 to Mr. H. R. Denton's letter of March 28, 1980 to All Power Reactor Applicants and Licensees. This requalification program will include examinations given in multiple segments following the lecture series for that segment. Any operator scoring less than 80 percent overall for the entire series of examinations or less than 70 percent on a segment examination will be removed from licensed duties and placed in an accelerated requalification program. In addition, any operator scoring less than 80 percent on a segment examination will receive remedial instruction in the topics covered in that segment.

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TRAINING FOR MITIGATING CORE DAMAGE M( ,/

Reference:

Action Plan - II.B.4 A special training program for mitigating core damage will be conducted at McGuire by the General Physics Coropration. This program will commence on October 13, 1980 and consist of forty hours of instruction covering five consecutive days. Lecture topics include 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, til of the availa'd e McGuire operators will participate in this training program. Selected 14cGuire operator training instructors and other appropriate Duke personnel will also attend.

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 '\ The existing body of knowledge regarding nuclear plant response under degraded m- l core conditions is being enlarged. Duke is participating in this effort in conjunction with other utilities, E;PO, 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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TRAINING DURING LOW POWER TESTING O

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Reference:

Action Plan - I.G.1 Duke Power Company will conduct a series of special low power tests at McGuire Nuclear Station for the purposes of providing meaningful technical information beyond that obtained in the normal startup test program and to provide supple-mental operator training. These tests will be performed at reactor power levels no greater than five percent of full power.

The' low power tests to be performed at McGuire are listed belcw:

1. Natural Circulation Verification

2. Effect of Steam Generator Isolation on Natural Circulation

3. Natural Circulation with Simulated Loss of Offsite Power

4. Natural Circulation at Reduced Pressure

5. Natural-Circulation with Loss of Pressurizer Heaters

6. Simulated Loss of All Onsite and Offsite AC Power

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Each operating shif t will either observe or participate in tests 1, 4, and 5. .

Tests 2, 3, and 6 are designed to provide plant specific technical information and therefore will only be performed once.

One page abstracts of each of these tests are provided in the pages which follow. The complete test procedures have been developed and will be reviewed j by Westinghouse. These test procedures were sent to the NRC on October 10, 1980.

g A'safaty analysis of this program has been prepared by Westinghouse and is sj provided in Appendix E.

Duke does not intend to perform the "Cooldown Capability of the Charging and Letdown System" (CCCLS) test at McGuire. The purpose cf this teat is to demonstrate the capability of the Chemical and Volume Control cystem (CVCS) to cooldown the reactor coolant system (RCS) with the steam generators isolated while using RCS pump heat as the heat source. This test has very little value from either a technical information collection standpoint or an operator training standpoint.

The energy removal capability of the CVCS has been long recognized by the operators at McGuire. During normal operation, the CVCS removes approximately 3 MW from the primary system. The operators have had many opportunities to l experience the effects of varying letdown rates on the primary system during

, the approximately 10 weeks of Hot Functional Testing performed at McGuire.

Therefore, a significant training benefit will not result from performing this test.

The ability to cool the pi cary system has been verified at McGuire during the CVCS functional test. The heat removal capacity of the McGuire CVCS is less than 5 MW which is extremely small in comparison to the heat loads normally experienced shortly after shutdown. This capacity is approximately equal to the decay heat load normally experienced by McGuire's core after approximately one hundred days. In the unlikely event that the McGuire CVCS would be required for cooldown after this length of time, there would be adequate time to prepare for such a cooldown.

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NSSS VENDOR REVIEW OF PROCEDURES

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Reference:

Action Plan - I.C.7 On May 15, 1980 a seiected group of McGuire emergency procedures were submitted to Westinghouse for their review. On August 4, 5, and 6, 1980 a seminar was held at McGuire to discuss the Westinghouse Emergency Operating Instructions

and selected McGuire emergency procedures with Westinghouse. Subsequent to this

! sem';.or certain McGuire emergency procedures were modified. On October 2, 1980 l 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 Rupture

4) Loss of Steam Generator Feedwater

!) Secondary Line Rupture The spe.c.'.al low-power test procedures were sent to Westinghouse on October 3, 1980 ~

for review. The power ascension teet procedures will be reviewed by Westinghouse prior to power escalation of McGuire Unit 1.

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PILOT MONITORIFG OF SELECTED EMERGENCY PROCEDURES I i

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Reference:

Action Plan - I.C.8 l

On May 15, 1980 a selected group of McGuire emergency procedures were submitted to the NRC for review. Based on the preliminary NRC. review of these and other i Westinghouse NTOL emergency procedures and Westinghouse's preliminary review of I the McGuire emergency procedures Duke Power Company revised certain McGuire emergency procedures. The following draf t procedures were submitted to the NRC for review via Mr. W. O. Parker's letters of September 16, 1980 and October 2,

- 1980:

1) Immediate Actions and Diagnostics for Safety Injection

2) Loss of Reactor Coolant

3) Steam Generator Tube Rupture

4) Loss of Steam Generator Feedwater

5) Secondary Line Rupture t

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ACCIDENT ANALYSIS AND PROCEDURE REVISION

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References:

NUREG-0576 - 2.1.9 o Action 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 analysis of small break LOCAs in upper head injection plants, WCAP 9600 and WCAP 9639, has been submitted to the NRC for their review. Duke has reviewed these reports and made the necessary modifications to the McGuire emergency procedures 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 pro-vide an analytical basis for subsequent Westinghouse development of guidelines for the detection of and recovery from inadequate core cooling. These guidelines will be submitted for NRC review prior to January 1, 1981. Following NRC approval of these analyses and guidelines, scheduled for July 1, 1931, Duke will prepare an inadequate core cooling procedure consistent with the approved guidelines. This procedere and its accompanying training will be complete by December 31, 1981.

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The Westinghouse analysis of selected transients and accidents is continuing on a deliberate schedule. Duke is closely following the development of this analysis and will modify the McGuire emergency procedures and training program as appropriate.

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AUXILIARY FEEDWATER INITIATION AND INDICATION

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References:

NUREG-0578 -'2.1.7a and 2.1.7b Action Plan - II.E.L.2 l

Automatic Initiation 4 -

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).

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-The automatic. initiation circuitry for the Abd meets the single failure criteria.

Additionally, for most failures which could prevent the automatic start of an individual auxiliary feedwater pump, manual initiation of the affecte.1 pump is available from the control room. However. should the auxiliary feedwater-pump in one safety train not be available > ; tu any single failure, the redundant safety train is.available with no loss of stem function.

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

Therefore, the automatic start of the motor-driven auxiliary feedwa :er pumps upon trip of both main feedwater pumps or steam gourator low-low 1 vel 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 energizel l on the bypass control switch. 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.

When the plant is in the startup mode, station procedures require that the bypass of the above motor-driven auxiliary feedwater pump start signals be removed. In 4 addition to the station procedures, an automatic means to remove the bypass will be provided. This automatic bypass removal will be generated when the P-ll set-point is reached. The P-ll setpoint is derived from Reactor Coolant System (RCS) pressure (%-1950 psig) and is the same signal used to unblock safety injection actuation. The P-il setpoint 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 i

pressure conditions have been reached. Installation and functional testing of -

i this automatic bypass removal will be completed by January 1, 1981.

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

Indication i 1 Safety grade indication uf auxiliary feedwater flow to each steam genera *o-has been provided in the McGuire contrv: coom. Provisions for calibration l 4 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 1s also provided.

i. This control grade flow instrumentation is powered from the highly reliable es battery-backed 120 VAC Auxiliary Control Power System (reference FSAR Ch. 8).

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

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AUXILIARY FEEDWATER SYSTEM RELIABILITY EVALUATION

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Reference:

Action Plan - II.E.1 .

An evaluation of the McGuire auxiliary feedwater system has been performed 'oy Duke and Westinghouse. This evaF:acion consists of the following items:

1. a simplified auxiliary feedwater system reliabilicy 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 decermination 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 thesa requirements are met.

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

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- -ADDITIONAL ACCIDENT MONITORING INSTRUMENTATION s-

References:

NUREG-0578 - 2.1.8b 4 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 presently installed noble gas monitors at McGuire cover the range of 10-7 pCi/cc to Ig&3 uCi/cc. A gr gammadetectorwillbeaddedtothesemonitorstoextendtherangeupto10gss pCi/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

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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 i of this additional detector, the noble gas monitor sample will be isolated.

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

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. The adlitional detectors will be shipped on December 30, 1980 and installed by

) February 16, 1981. Procedures for estimating noble gas release rates if the q (xisting instrumentation goes off scale will be written to cover the interim

! - :iod between fuel loading and installation of the new detectors. These pro-i cedures will require the use of portable high range survey instruments to mea-sure 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 will be used to estimate the concentration (p C1/cc) of gas in the line. If the radiation levels do not exceed the above personnel exposure limits the procedures will require collec-tion of gas, particulate, and radioiodine samples. Silver zeolite cartridges will be used for radioiodine sampling when noble gas interference is expected.

-Two independent counting rooms are available onsite at McGuire for analysis of

these samples. One counting room is located in the Administration Building and 4 the other is located in the Auxiliary Building. In addition a third counting room is availaole in the Duke Technical Center located just outside the McGuire exclu-sion boundary.

The present radiation monitoring system provides detection of volatile aid non-volatile radioactive contamination of the seconda?y. A condensor air ejector monitor continuoulsy 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 generatots. An alarm on either of these monitors provides control room operators with an indication of steam L

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

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would quantify the level of radioactivity released to the environment prior to isolation of the affected steam generator.

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To quantify the level of radioactivity released in the'i event the affected '

. steam generator:is not isolated and the atmospheric steam dump valves open,

- Duke has a steam rsdiation monitoring system under design. This system will

use an area radiation monitor mounted near each-steam line as.it exits the reactor building. These monitors will have controg room readout and alarm.

The radioactivity range covered will be 10-1 to 10 uCi/cc. Equipment installa-tion is anticipated by the middle of 1981.

4 As an inte rin measure, a Ludium model 300 area radiation monitor will be mounted in a protected area near'the atmospheric steam dump valves. The monitor will be shielded from background radiation. All controls, alarms, and readouts will be

,! located locally.- Station procedures will require that this monitor be checked' at least once' every eight hour shif t.s In the event of suspected steam generator tube failure'this monitor will be checked at least once every hour. With veri-i 4

fied steam generatorLtube' failure, this monitor will be checked every 15 minutes as long as the potential for a steam release exists. If a steam release occurs,

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' this monitor will:be checked as frequently as the situation warrants. In no r case will the frequency be less than stated above.

i Open-closed indication of the atmospheric steam dump valves is provided in the i control room. Procedures will be written to use 'this indication in conjunction l  ;

with the design steam flow per valve to estimate the total steam mass released .

? during a dump.

1 The Eontainment 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 monitors will be installed  ;

^inside the McGuire containment. -These monitors will be supplied by General j Atomics and will. feature GA detector model number RD23. Each monitor will

' utilize an iogization chamber to measure gamma radiation and will cover the range from 17 to1108 R/hr. No overlapping of ranges is required. Monitor

! ' sensitivity to 62 Kev is 9.8X10-12 Amps / Rad /hr and the sensivity to 52 Kev

!. is 9.0XF' Amps / Rad /hr. Seismic qualification of the monitor is in accord-l ance witn IEEE344-1975 and environmental qualification is per IEEE323-1971.

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[ 10ne monitor will be powered 'from the Train A vital instrument bus, and the

! other monitor will be powered ~from the Train B vital instrument bus. Analog I 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 contro11 room.  ;

t An electronic calibration of the' monitors will be performed every refueling i outage.- In addition a' radiation source will be used to perform an in-situ-calibration ~ of the monitor , range below 10 R/hr.  !

l The detectors will b'e~ mounted on the primary shield wall at an elevation of at l 1 east 1750+2 (10 feet or more above the maximum post-LOCA water level at 00 and t

1800 in the lower containment. The following McGuire General Arrangement drawings show the plan snd' sectional views with the monitor' locations drawn in.

The current schedule for' implementing this requirement calls for equipment ship-

-mention December 30,'1980 and installation by February'16, 1981.

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Containment Pressure i-O'-

Continuous indication of containment pressure will be provided in the control room. : Measurement and indication range will extend from -5psig to 60 psig, Each of the redundant differential pressure transmitters will be located in an i 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 annulus. These valves will be normally open and will have position indication and manual control in the control room. Continuous indication from each trans-i mitter will be provided in the control. room. In addition, one channel of con-E tainment pressure will be recorded. .These instruments will be completely independent of the existing containment pressure transmitters and will be

, installed by January 1, 1981. l

Containment Water Level t

) Two containment floor and equipment sumps are provided on the floor of the lower j containment (El 725') to collect floor drains and equipment drains. However, j 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.

~

j 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 sump level. The range of this instrumentation is 0-20 feet (El 725' to El 745') t 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.

\,)

i The redundant differential pressure transmitters utilized in this instrumentation have been relocated to the annulus where a filled capillary system will connect '

! its associated transmitter with bellous sensors located inside containment.

j Continuous indication from each transmitter will be provided in the control room.

In addition, one channel of containment water level will be recorded.

Containment Hydrogen Monitoring Continuous. indication of hydrogen concentration in the containment atmosphere I

will be provided in the control room. ^ This hydrogen monitoring system will consist of two redundant Comsip, Inc./Delphi Systems Division K-111 analyzer 1 -systems with a range of 0 to 30% hydrogen by volume. These analyzers operate j independent of the recombiner system and will be powered from redundant Class i

1E power supplys. 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 i - control panel indicator and alarm and a separate control room indicator and alarm. In addition, one channel of containment hydrogen concentration will be j recorded. '

! Each containment sample header will have five inlet samples available for 1 monitoring.

II-13A. 10/10/80 3

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4. Radiation-Monitor /Recombiner Inlet header

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

This instrumentation will be installed by January 1, 1981 contingent upon l timely equipment delivery.

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CONTROL ROOM HABITABILITY

Reference:

Action Plan - III.D.3.4 Duke has conducted an evaluation of the McGuire Control Room to determine the level of protection provided against radioactivity and toxic gas hazards described in the following documents.

1. Standard Review Plan - Sections 2.2.1, 2.2.2, 2.2.3, and 6.4.

2. Regulatory Guides 1.78 and 1.95.

This evaluation revealed that the design of the McGuire Control Room both meets the criteria specified in the above documents and provides adequate protection of the Control Room from radioactivity and toxic gas hazards. A de:. ailed descrip-tion of this protection is provided in the applicable sections of the McGuire FSAR.

A summary description is provided below. -

The only path of entry of a toxic gas to the Control Room is the outside air intakes. These intakes are located adjacent to the Reactor Building opposite each unit vent. They are at El. 772+0 and are separated by 360 ft. The only toxic material normally used at the station which might be drawn into an air n intake of the Control Room is chlorine. It is stored in standard 150 lb.

( cylinders at two locations on site. The shortest distance from a Control Room air intake to a chlorine storage area is 500 feet. Each air intake has a redundant chlorine detection and isolation system. A single component failure will not impair the function of the system. The chlorine detectors are seismically qualified and have a five second response time to 5 PPM chlorine.

The isolation valves conform to ASME code, section III, class 3. They have electric motor ope ators and a closure time of 8.4 seconds. During normal operation 500 CFM is taken from each intake to maintain the Control Room at 1/8 in. w.g. positive pressure. The outside air plus 1000 CFM of return air is circulated through carbon filters before delivery to the Control Room.

Upon chlorine detection the affected intake is automatically isolated and the full 1000 CFM of pressurization air is brought in through the second intake.

In the unlikely event that both intakes become contaminated the Control Room will be isolated with 2000 CFM circulated through carbon filters. Additional protection includes manual isolation at the discretion of the operator and full face respirators with a six hour air supply for five men which can be replenished from outside the Control Room if necessary.

This habitability review of the McGuire Control Room has revealed that no design modifications are necessary.

(O/ II-17 10/10/80

.c j 1

FINAL RECOMMENDATIONS OF THE BULLETINS AND ORDERS TASK FORCE b

V

Reference:

Action Plan - II.K.3 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, a'l challen?'s to the FORV's or safety valves will be documented and reported to the NRC.

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" for the pressurizer PID controller rate time constant. The McGuire time constant will be adjusted accordingly.

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, such as pressure, pressurizer water level (to prevent water _ discharge through safety valves) and also on variables which directly affect the heat transfer capability of the 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 varial'e 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 immediate reactor trip on trubine trip. This analysis considered both normal plant response 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 includitg.the steam 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.

i (m~')

II-19 10/10/80 i:

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 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 Ste m Station in conjunction with the EPRI valve testing program. The testing demonstrates that the McGuire PORV's and PORY block valves meet all functional and design requirements and provides added assurance of proper PORV and PORV black valve operation.

Duke's position is that installation of the subject trip at McGuire is not warranted. The full load rejection capability and the added assurance of proper PORV and PORV block valve operation coupled with the benefits of mini-mizing reactor coolant system thermal cycles and challenges to the reactor

' coolant system protective devices far outweigh the additional protection against a narrow range of events.

O l

1 O II-19A 10/10/80

UPGRADED EMERGENCY PREPAREDNESS

Reference:

Action Plan - III.A.1.1 The document, " Emergency Plan for McGuire Nuclear Station," describes 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. This 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 North Carolina has also developed plans for coping with radio-logical 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. W. 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 has evaluated the review team's report and has revised the McGuire emergency plan as appropriate.

The State of North Carolina's emergency plan for Mc';uire has been revised i per NUREG-0654 and was submitted for FEMA review in early August 1980.

North Carolina officials are evaluating the FEMA review and will revise the state plan as appropriate.

The McGuire Emergency Management Respoise Exercise is scheduled to be performed on December 5 and e, 1980. A descrip'. ion of this exercise has been submitted to the NRC. North Carolina officia3- have submitted a similar description of the exercise to FEMA.

III-1 10/10/80

STATION DIRECTIVE 3.1.32

' APPROVAL: MN y w -

DATE: m17 INFORMAT!ON

" ANDj0R REVIEW ONLY i

DUKE POWER COMPANY i MCGUIRE NUCLEAR STATION j STATION SAFETY ENGINEERING GROUP OBJECTIVE This directive will define and establish administrative guidelines for function-ing of the on-site Station Safety Engineering Group (SSEG). The responsibilities and distribution of reporting for the SSEG will be outlined below.

APPLICABILITY The SSEG will function as a full-time group consisting of four members and a permanent chairman. The Chairman of the SSEG will be appointed by and will .

report directly to the Manager of Nuclear Production. All SSEG recommenda-tions will be sent directly to the Manager of Nuclear Production, the Director of the NSRB and the Station Manager. They in turn will assure proper action is taken to address SSEG recommendations.

MEMBERSHIP All members of the SSEG 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 calated technical training. The four members of the SSEG will be chosen from the attached list (Attachment 1) with as least one person from each of the four areas listed.

These members will be selected by and will report to the Chairman of the SSEG during their assignment in this area. Personnel will be assigned to this area for at least six months.

RESPONSIBILITIES The SSEG will function as an independent technical review group in the areas out-lined below.

A. Licensee Event Report (LER) Summaries - Each month the General Office Licensing Group will forward a summary list of all LER's applicable to McGuire to the SSEG. The Chairman will assign these LER's to members'of the SSEG to determine if any specific action needs to be taken at McGuire to prevent or mitigate the consequences Uf a similar event. All recommendations will be t forwarded directly to the responsible station group, the Manager of Nuclear Production, the Station Manager and the Director of

.p the NSRB.

t I v

10/10/80

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B. Effectiveness of Plant Programs - The Chairman will assign members of the SSEG specific station programs to review and

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/} determine if any recommendations to increase the effectiveness of the program should be considered. After review by each member of the SSEG, any recommendations will be forwarded to the responsible station groups, the Manager of Nuclear Produc-tion, the Station Manager and the Director of the NSRB.

C.- Plant Modification Review - All design changes involving struc-tures, systems, or components with QA conditions will be reviewed by the SSEG. This review is to insure all safety concerns are properly addressed.

D. Station Procedures and Changes - Selected station procedures and/or changes to procedures will be reviewed by the SSEG to determine their edequacy.

E. Plant Incident Reports - All incidents involving reportable items as defined by Station Directive 2.8.1 or other investigations as deemed appropriate by the Chairman of the SSEG will be assigned to a member of the SSEG for investigation and preparation of the station incident report as outlined in Station Directive 2.8.1.

The incident report will be reviewed by the SSEG and any additional recommendations included. These reports will be sent to the Manager of Nuclear Production, the Station Manager, the Director of the NSRB, and other groups as outlined by Station Directive 2.8.1.

v I /'

! 10/10/80

( )

i NRC REQUEST FOR INFORMATION TRANSMITTED p)

(

BY LETTER OF JULY 2, 1980 FROM B. J. YOUNCBLOOD, CHIEF, LICENSING BRANCH NO. 1 DIVISION OF LICENSING l

4 II.K.1 IE Bulletins on Measures to Mitigate Small Break LOCAs and Loss of Feedwater Accidents C.I.5 The response provided on the review of all valves involved with Engineered Safety FeaturesL(ESF) operation is qualitative in nature.

The following information is needed to assess the adequacy of the review required by the IE Bulletins.

a) Describe the review process employed by Duke Power to verify that valve positioning requirements and valve positions are met Let the ESF tests. This description should include measured and vic9al

, inspections made for the verification of valve operations.

b) Describe the procedure used to verify valve response time, and the measurements and visual observations made for this verifica- -

tion.

c) Describe the review of positive controls and maintenance procedures performed in response to the IE Bulletins.

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C.I.10 a) Describe the process by which the operability of redundant systems and safety related systems is verified.

b) Describe the process for notification of operators when safety-related systems are removed from, or returned to service.

c) Describe the preparation, contents, and maintenance of historical records of safety-related systems maintenance.

Response

See IE Bulletins on Measures to Mitigate Small-Ereak LOCAs and Loss of Feedwater i

Accidents.

II.K.3 Final Recommendations of the Bulletins and Orders Task Force I

C.3.9 Provide a schedule for completion of your review of the pressure integral derivative controller.

C.3.12 Provide a discussion on analyses performed to assess plant response, and potential for PORV challenges when turbine trip, loss of feed-water flow, or loss of steam generator secondary inventory occurs.

Response

'See Final Recommendations of the Bulletins and Orders Task Force D,

l 10/10/80 I

i

NRC REQUEST FOR INFORMATION TRANSMITTED

(

N BY LETTER OF AUGUST 25, 1980 FROM R. L. TEDESCO, ASSISTANT DIRECTOR FOR LICENSING, DIVISION OF LICENSING

1. The McGuire response to NUREG-0578, Item 2.1.8.c proposes the use of portable survey instruments in conjunction with silver zeolite cartridges to determine radioiodine concentrations for respiratory use. Explain whether this procedure will be used as a substitute for, or in conjunction with standard radioiodine sampling and analysis systems outlined in the

" clarifications" of the NUREG-0578 clarification letter dated November 19, 1979. Also provide a description of the method for evaluation.

Response

See In-Plant Radiation Monitoring

2. Provide a commitment to complete alternate auxiliary building telltale valve sampling procedures by January 1, 1981 in lieu of modifications to the Unit 1 sampling room. .

Response

See Plant Shielding

3. Provide an explanation as to why the Waste Gas Treatment System was not included in the evaluation as a potential highly radioactive system or O' provide the results of the evaluation of this system.

Response

See Plant Shielding

, 4. Verify that the environmental and seismic qualifications of the proposed radiation monitors meets the position of NUREG-0578, Item 2.1.8.b as stipulated in the requirements of Regulatory Guides 1.89, 1.97 Rev. 2 (ANSI N320-1978) and 1.100.

Response

See Additional Accident Monitoring Instrumentation

5. Provide a commitment to calibrate the proposed high radiation monitors at each refueling outage (at least every 18 months) as a minimum in accordance with the staff position on NUREG-0578, Item 2.1.8.b and according to the Technical Specifications for all Radiation Monitoring Instruments (3/4 3.3, Table 4.3-3).

~

Response

See Additional Accident Monitoring _ Instrumentation (s-,-

10/10/80

NRC Request for Information Transmitted n by letter of September 17, 1980 from Falph Birkel, Project Manager, Division of Licensing Your response regarding the McGuire Health Physics organization submitted by letter of August 6, 1980, has been reviewed. As a result of this review it is our position that the McGuire Station Health Physicist report directly to the Station Manager in accordance with the "Draf t Criteria for Utility Manage-ment and Technical Competence" (Section II.A.1/II.A.2) and Regulatory Guide 8.8 (Section C.l.b.(3)) as required in Item I.B.1.2 of NUREG-0660/0694. The NRC management review team audit cited in your August 6,1980 submittal did not include the McGuire Health Physics Organization in its review of the McGuire Station organization. Provide a commitment to a radiation protection organiza-tion which meets our position in accordance with NUREGS-0660/0694, " Draft Criteria . . .", and Regulatory Guide 8.8, or provide c detailed explanation of an alternate McGuire Health Physics Organization which teets the positions and criteria of NUREGS-0660/0694, " Draft Criteria . . ." ard Regulatory Guide 8.8.

! Response Duke Power's radiation protection policies are established in a System Health Physics Manual that is approved by the Vice-President, Steam Production and is issued by the Manager of Nuclear Production. As stated in this manual, "The T Duke Power Company System Health Physicist and his staff are responsible for establishing the Health Physics Program; for providing technical assistance for conducting the program; for assuring Management that personnel exposures are maintained "as low as reasonably achievable"; for auditing the efficacy of the program in complying with NRC and other governmental regulations and Regulatory Guides; and for modifying the program as required by experience and by regula-tory and technical changes."

The Station Health Physicist is responsible for insuring these established policies are implemented at the station level. Additionally, the current organizational position of the Station Health Physicist reduces his adminis-trative burden and allows dedication to technical management of the Health Physics staff. Likewise the reduced administrative burden of the Station Mant.ger facilitates the safe operation of the plant.

It is Duke Power's position that this arrangement with a System Health Physicist establishing policy implemented by a Station Health Physics Staff fully satisfies l the intent of insuring radiation protection is fully supported by top management.

i In addition to direct access to the Station Manager, the Station Health Physicist can insure any problems or concerns are expressed to the System Health Physicist.

The periodic audits conducted by the System Health Physics Staff insure any con-l cerns or problems are promptly reported to the Station Manager and the Manager of l ' Nuclear Production.

l l The " Draft Criteria for Utility Management and Technical Competence" has been revised by the NRC and issued as NUREG-0731, " Guidelines for Utility Management j A Structure and Technical Resources." NUREG-0731 Section 11.41 states " utilities Q with large commitments to nuclear power may centralize overall management for l 10/10/80 l I

a 1

areas such as training and radiation protection with an offsite organization and reduce the onsite function to that of coordinator or program implementation supervisor."- As discussed above and in Duke's' response of August 6, 1980 (as revised on September 8, 1980) the current McGuire health physics organization satisfies the requirements of NUREG-0694, Item I.B.1.2.

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10/10/80

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NRC Request for Information Transmitted by letter of September 17, 1980 from B. J. Youngblood, Chief, Licensing Branch No. 1 Division of Licensing In response to IMI concerns, you submitted information pertaining to the automatic initiation of the auxiliary feedwater system (Action Plan Item II.E.1.2; NUREG-0578, Section 2.1.7,a). Based on our review of this infor-mation, we have concluded that the McGuire Units 1 and 2 AFWS automatic initiation meets all safety grade requirements with the exception of operat-ing bypasses. The McGuire FSAR states that, during plant shutdown, the operator can manually defeat the-automatic start of the auxiliary feedwater motor-driven pumps on either low-low level in any steam generator or trip of j both main feedwater pumps. The design does not include an automatic removal of this manual bypass feature. This design is not in accordance with IEEE 279-1971 requirement 4.12 (i.e., operating bypasses) which states that, where operating requirements necessitate automatic or manual bypass of a protective

- function, the design shall be such that the bypass will be removed automatically 4

whenever permissive conditions are not met.

1 Therefore, it is our position that automatic removal of this manual bypass be l

provided in accordance with IEEE 279-1971 since possible failure to remove this operating bypass could prevent the AFWS from performing its automatic safety function. Please provide sufficient information (i.e., logic diagram, electrical schematics, etc.)-to demonstrate compliance with this position including identifi-l cation and justification of any exceptions.

Response

See Auxiliary Feedwater Initiation and Indication i

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l O 10/10/80 l-

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