ML20134F991

Summary of ACRS 298th Meeting on 850207-09 in Washington,Dc Re OL Review,Fire Protection & Nuclear Power Plants, Pressurized Thermal Shock,Dhr,Backfitting & NRC Safety Research Program & Budget

ML20134F991
Person / Time
Issue date:	02/07/1985
From:	Advisory Committee on Reactor Safeguards
To:	Advisory Committee on Reactor Safeguards
References
REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR ACRS-2280, NUDOCS 8508220118
Download: ML20134F991 (250)

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298TH ACRS MEETING FEBRUARY 7-9, 1985

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I. Chairman's Report (0 pen) ................................ 1 II. Braidwood Nuclear Station Units 1 and 2 OL Review (0 pen) .................................................. 2 III. Fire Protection and Nuclear Power Plants (0 pen) ......... 9 IV. Pressurized Thermal Shock of Reactor Pressure Vessels (0 pen) .................................................. 13 V. Backfitting of Nuclear Power Plants (0 pen) .............. 16 VI. ACRS Meeting with the NRC Comissioners (0 pen) .......... 19 VII. Decay Heat Removal (0 pen) ............................... 24 VIII. NRC Safety Research Program and Budget (0 pen) ........... 26 IX. Executive Sessions (0 pen) ............................... 27 A. Subcommittee Assignments ........................... 27

1. Subcommittee Comments on High-Level Waste Repository Reports ...................... 27

2. Comission Meeting on Equipment Qualification

- January 6, 1984 ............................. 27

3. Discussion of Issues in Nuclear Safety ........ 28 1

B. ACRS Reports , Letters , and Memorandum . . . . . . . . . . . . . . 28

1. Review and Evaluation of the NRC Safety I

' Research Program for FY 1986 and 1987 ......... 28

2. ACRS Report on the Braidwood Station, Units U s@ 1 and 2 ....................................... 28

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3. Proposed Rule on Backfitting .................. 28 g

o ',\ m E! m 4 ACRS Funding and Staffing Reductions for S

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ou FY 1985 and 1986 .............................. 28

'A g 7 5. ACRS Coments on D. L. Basdekas' Professional a @gg Opirion Concerning the Proposed Final Rule i nord on Pressurized Thermal Shock . . . . . . . . . . . . . . . . . . 29 mard

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6. ACRS Coments on Proposed Rule Changes to 10 CFR 50, Appendix E ......................... 29 L

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C. Ge n e r i c I s s u e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29

1. Consideration of Seismic Events in Emergency Planning ...................................... 29

2. Lack of Redundancy in Reactor Scram Systems ... 29 D. Future Scredule ......................................... 29

1. Future Agenda ................................. 29

2. Future Subcommittee Activities ................ 30 E. INP0 Accrediting Board .................................. 30 F. Attendance at Sixth Symposium on Training of Nuclear Facility Personnel ...................................... 30 G. International Atomic Energy Agency Request for '

Services of Charles J. Wylie ............................ 30 H. Cor.fl i c t of I nte res t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30

1. ACRS Effectiveness Study ................................ 31 J. Requalification of Reactor Operators . . . . . . . . . . . . . . . . . . . . 31

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MINUTES OF THE 298TH ACRS MEETING FEBRUARY 7-9, 1985 WASHINGTON, D.C.

The 298th meeting of the Advisory Comittee on Reactor Safeguards, held at 1717 H Street, N.W., Washington, D.C., was convened by Chairman D. A.

Ward at 8:30 a.m., Thursday, February 7, 1985.

[ Note: For a list of attendees, see Appendix I. H. Etherington did not attend the meeting. P. G. Shewmon attended the meeting only on Thursday, February 7, 1985. G. A. Reed was not in attendance on Saturday, February 9,1985.]

Chairman D. A. Ward noted the existence of the published agenda for this meeting, and identified the items to be discussed. He noted that the meeting was being held in conformance with the Federal Advisory Comittee Act and the Government in the Sunshine Act, Public Laws92-463 and 94-409, respectively. He also noted that a transcript of some of the public portions of the meeting was being taken, and would be available in the NRC's Public Lonment Room at 1717 H Street, N.W. ,

Washington, D.C.

[ Note: Copies of the transcript taken at this meeting are also available for purchase from Ace-Federal Reporters, Inc., 444 North Capitol Street, Washington, D.C. 20001.]

1. Chairman's Report (0 pen to Public)

[ Note: R. F. Fraley was the Designated Federal Official for this portion of the meeting.]

Chairman D. A. Ward solicited member comments regarding ad hoc subcomittees on a "Long-Range Plan for the NRC" and "The State of Nuclear Power Safety." It was noted that the Office of Policy Evaluation (OPE) was also asked by the Commission to prepare a long-range plan for the NRC and it will be directed to interface with the ACRS regarding this matter.

Chairman Ward suggested that Committee members study a memorandum dated February 5,1985 entitlect, " Thoughts on the Need for Safety Research and Human factors" in preparation for a discussion on the topic of human factors research during the session on preparation f

of the ACRS Report to Congress on the NRC Safety Research Program and Budget for fiscal 1986 and 1987.

R. F. Fraley, ACRS Executive Director, requested Comittee coments on the preliminary set of proposed guidelines on the subject of apparent conflicts of interest. He indicated that this matter will be discussed further at a meeting of the Procedures Subcomittee on April 9,1985.

MlNUTES OF THE 298TH ACRS MEETING . FEBRUARY 7-9, 1985 Chairman D. A. Ward asked for ACRS member comments on a proposed revision to the ACRS bylaws as set forth in the January 31, 1985 memorandum from T. G. McCreless, ACRS Assistant Executive Director for Technical Activities, regarding conduct of members as individuals. This matter will be referred fer discussion at an April 9, 1985 meeting of the Procedures Subcommittee and subsequently for full Committee consideration.

Chairman D. A. Ward indicated that H. Denton, Director NRR, has asked that G. A. Reed serve on an advisory group to the Technical Specification Improvement Project, an in-house NRC team examining overhaul and possible simplification of the nuclear power plant technical specifications. The Committee took note of the fact that G. A. Reed would not be representing the ACRS and it did not offer objection to his ;;articipation on this advisory group.

Chairman D. A. Ward mentioned that Duke Power's Oconee Station Unit 2 has recently set a record for the longest period of on-line operation, a record previously held by the Connecticut Yankee .

Nuclear Plant. This record also surpasses any records for on-line performance for foreign reactors.

II. Braidwood Nuclear Station Units 1 and 2 OL Review (0 pen to Public)

[ Note: E. G. Igne was the Designated Federal Official for this portionofthemeeting.]

J. Stevens, NRC Licensing Project Manager for the Braidwood Station, presented a chronological overview of licensing activities on the Braidwood project (see Appendix IV). She indicated that the Byron and Braidwood Stations are duplicate plant designs in accordance with the NRC Statcrent of Standardization of Nuclear Power Plants, dated August 31, 1978. She indicated that the Staff's review of the reference design documented in the Byron SER has five additional supplements which include the Westinghouse nuclear steam supply systems, the balance of plant systems and other auxiliary systems. The Staff's final duplicate design approval delineates topics outside the scope of the duplicate design including site-related characteristics, changes from the Byron station design such as offsite power systems due to the different location on the grid, water systems due to the different sources of water between the two sites, and the fact that the river's screenhouse ventilation system and diesel generator fuel oil systems are not applicable to the Braidwood Station due to the differences in the ultimate heat sink.

J. A. Stevens reviewed the status of all open items. J. C.

Ebersole asked if the Byron and Braidwood incorporate diversity in their scram systems. J. A. Stevens was unable to stipulate whether that issue had been resolved for Byron and Braidwood at this time J and offered to respond at a later time.

J. Stevens indicated that there are two license conditions on the 2

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' F"BRUARY 7-9, 1985 MINUTES OF THE 298TH ACRS MEETING Braidwood plant, the issue of masonry walls and the fire protection program. C. J. Wylie asked what type of fire protection systems the Braidwood plant has in their cable spreading room. D. J.

Kubicki, NRC, explained that there are two separate cable spreading rooms at Braidwood. The upper cable spreading room has a halon automatic fire suppression system with a manual carbon dioxide fire suppression system as a backup. The lower cable spreading room has an automatic carbon dioxide fire suppression system. C. J. Wylie asked if there is any requirement for additional water spray for suppression. D. Kubicki indicated that this is not an issue for the Braidwood Station.

E. G. Greenman, NRC Region III, presented an historical chronology of construction inspection experience at the Braidwood site (see Appendix V). He explained that the systematic appraisal of licensing performance (SALP) is one of the key indicators used in the region to assess and identify licensee performance and make sound decisions regarding allocation of NRC regional resources. He indicated that there are numerous Braidwood construction related problems at present. Problems exist in the areas of piping systems and supports, HVAC systems, and in the electrical area (cable separation). E. Greenman indicated that the problems identified at the site have been handled by the licensee through at least twenty corrective action programs. In addition, a Braidwood Construction Assurance Program (BCAP) was introduced in June 1984. He explained that the BCAP is overall objective is to assure that all design specific problems in the construction effort have been identified and corrected. It consists of three basic elements -- a reinspection program, a review of procedures to specification requirements, and a review of significant corrective action programs. He spoke of NRC initiatives in the inspection area including the NRC's own independent nondestructive examination program conducted by the NRC's mobile van. He noted that there is one current issue that remains outstanding -- the ability of the licensee to manage these multiple reinspection efforts including the BCAP and still do ongoing work at the site. He indicated that the region is optimistic that Comonwealth Edison's corrective .

action programs in place, including the BCAP program, will assure that construction is completed in accordance with regulatory requi rements .

R. F. Heishman, NRC, briefly discussed the inspection appraisal team instruction program recently completad at the Braidwood Station on January IS, 1505. I:e explained that the preliminary results at Braidwood indicated that this plant is not much different from other nuclear plants except for some slight differences. One of these is the greater number at Braidwood of different corrective action programs to address previously identified problems. He stated that the Staff found nothing that would preclude the plant from an operating license.

T. Maiman, Comonwealth Edison Company (CECO), comented on the NRC Staff reports (see Appendix VI). He noted that the four Byron and Braidwood units share a comen NSS supplier (Westinghousc), use the 3

MlNUTES OF THE 298TH ACRS MEETING FEBRUARY 7-9, 1985 same FSAR, incorporate the same design, use the same architect engineer (Sargent and Lundy), and many of the same general arrangement drawings. W. Kerr asked if Commonwealth Edison found the duplicate plant approach productive in terms of efficiency and quality. T. Maiman indicated that CEC 0 found the duplicate plant effort tc be productive as far as resolving NRR issues. There is some question, however, as to whether CECO would do the job exactly the same way again. W. Kerr asked what reservations CEC 0 had with regard to the standard plant approach. T. Maiman thought it would be best to have close to a 100 percent engineered plant before the start of construction and an upfront agreement with the regulator precluding changes if the plant were built the way 1 was designed.

The major causes of delays and perturbations in t'te construction effort proved to be changes regardless of whe:her they were initiated by CECO or required by others. M. Wallace, CEC 0, explained that economies go along with the fact that the design is being done basically one time (that is for B/ron as well as Braidwood) with the drawings used at Byron to a great degree also used at Braidwood. Field problems identified .tt Byron were fed ,

back to the design for Braidwood even before construction activity began for that particular part of the job. G. A. Reed asked if CECO transferred construction personnel from Byron to Braidwood.

Wallace indicated that some Byron personnel are being incorporated into the Braidwood team. A number of the people from the LaSalle County Plants are also at Braidwood now that the LaSalle County Plants are up and running.

D. Okrent asked what protection the Byron and Braidwood plants have against steam generator overfill transients. G. Klopp, Edison Engineering, explained that the Byron and Braidwood main steam piping has been designed for adequate support in the filled condition. D. Okrent asked if there is a trip of any kind for the overfill transient. J. Klopp indicated that there is no trip. J.

Ebersole asked regarding diversity in the scram systems against the Westinghouse DB 50 circuit breaker failures experienced at the Salem plant. K. Ainger, CECO, indicated that an automatic shunt trip modification designed by Westinghouse, and an independent testing circuit for that shunt trip automatic function, is being added to DS 416 scram trip breakers in the Byron design which has been reflected in the Braidwood design. J. Ebersole asked if there is diversity in the MG set exitation circuit or any other physical electrical device diversity outside of the breaker boxes. K.

Ainger indicated that there was no diversity.

T. Maiman, CECO, discussed elements of the Commonwealth Edison corporate structure and, in particular, the nuclear organization.

G. A. Reed briefly mentioned the designated representative system in the FAA. M. Wallace described the dynamics of the Braidwood project team. He explained that CECO's management approach at Braidwood is to maintain flexibility in order to make changes to provide better control of project activities (see Appendix VI). He explained that Comonwealth Edison in 1981 comitted to meet the requirements of the shift technical advisor program set by NUREG-0737 by creating the position of Station Control Room i

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MINUTES OF THE 298TH ACRS MEETING FEBRUARY 7-9, 1985 Engineer or SCRE, a degreed technical graduate who has a SR0 and also meets the requirements of the STA. He further noted that

essentially all SR0 and R0 candidates are involved in a hot I participation program at the Zion station. G. A. Reed asked how far CECO is going with R0 and SR0 licenses. M. Wallace indicated that two of the assistant superintendents will be SR0 licensed, f the operating engineers will be SR0 licensed, and the shift engineers and shift foremen as well as SCREs will also be SR0

licensed. F. J. Remick asked CEC 0's thoughts on the option of i having either a separate STA or a combined 50/STA as a second SR0 on shift. C. Reed indicated that CECO supports having the two options. He thought CECO ought to have that flexibility but he noted that CEC 0 has combined the 50 and STA positions at all of its stations and that is what CEC 0 prefers.

1 i M. Wallace briefly discussed a quality first group in the organization. This is a group which promotes the company's strong

. positive attitude regarding quality among the entire work force at Braidwood and provides an opportunity for individuals to express ,

any concerns they may have regarding the quality of construction.

F. J. Remick asked what CEC 0's experience has been with workers coming forward to the quality first program personnel to express

their concerns. He asked whether these concerns come mostly from the crafts, or engineers and other personnel as well. M. Wallace l

indicated that there have been some concerns identified, none of which were considered significant. Concerns have been expressed from the crafts, the QC workers', as well as others, but these findings are preliminary since the program was just set up a .few months ago.

M. Wallace presented an overview of the construction progress at the Braidwood site, including a review of the installation percentage completion of the major construction bulk quantities (seeAppendixVI). G. A. Reed requested an explanation of the SCRE position on shift. M. Wallace indicated that the Station Control Room Engineer is a degreed engineer with an SR0 license who will be in the control room during operations. He functions as the control room supervisor when he is on shift and, in the event of an abnormal occurrence, is relieved of that responsibility by either the shift foreman or shift engineer. He then serves as a shift technical adviser with no responsibilities for directing the l activities in the control room. D. A. Ward asked if the operating engineer position mentioned was a shift assignment. M. Wallace indicated that the operating engineer is not a shift position but

' is the individual who represents the station superintendent on shift, the " top man" on shift.

G. Fitzpatrick, CECO Assistant Manager of quality assurance discussed the quality assurance organization staffing and methods i of operation at Braidwood (see Appendix VI). He ex)lained that the site quality assurance organization is responsible for assuring that the construction contractors meet all requirements and that i they fully implement the quality assurance programs. This is done i primarily through audits and surveillances of the contractor 5

MlkUTES OF THE 298TH ACRS MEETING FEBRUARY 7-9, 1985 I activities and thorough overviews of completed and in-process work.

He indicated two of the overview programs are performed by the onsite testing agency, the Pittsburgh Testing Laboratory (PTL),

under the direction of the site quality assurance organization. He explained that another major quality assurance effort at Braidwood is the quality assurance overview of the Braidwood Construction and Assessment Program (BCAP). G. A. Reed asked whether the transfer policy, the bidding practice used at Comonweal th Edison where individuals could bid in and out of nuclear stations at will, is still in existence. He also wanted to know what CECO is doing resarding selection criteria for new hires. C. Reed indicated that Commonwealth Edison has the same practice of allowing people on one station to bid fcr a higher level job at another station. He noted that as a result of this, most of the people from Byron and Braidwood have come from Quad Cities. Zion, and Dresden. G. A.

Reed asked if CEC 0 uses selection criteria such as the Edison Electric Institute (EEI) validated tests to determine mechanical comprehension, logical reasoning and IQ. C. Reed indicated that new hires are given a general physics examination which tests their knowledge of math and science. He also indicated that Comonwealth Edison is using the POS tests at Braidwood and all of CECO's stations (tests akin to EE! test batteries).

T. Maiman described the BCAP program as a means to provide an additional level of confidence in the quality of construction at Braidwood. He described the three basic objectives of the program (see Appendix VI). He described how project management controls and the CECO quality assurance organization impact on the plant work quality. He indicated that BCAP activities are coordinated by Commonwealth Edison employees and supported by Sargent and Lundy along with other consultants. Stone and Webster and Daniel Construction employees comprise the majority of the task force with Stone and Webster writing procedures and reviews and Daniel Construction doing the field inspections. He noted that an independent expert, John Hansel, has been hired with a team of experts to provide an independent overview of the BCAP program through the conduct of audits, surveillances and other inspections.

He pointed out that the BCAP effort is a highly structured operation using detailed plans, procedures, checklists and instructicns. F. J. Remick asked if Comonwealth Edison would have done anything differently from a QA standpoint were there not an NRC driving the effort. T. Maiman indicated that the BCAP program might not have been as extensive a system had it not been driven by the need for an operating license for Braidwood.

G. Kloop addressed Braidwood design features and their relationship to the Zion PRA (see Appendix VI). He noted that the Byron /Braidwood core melt frequency is lower than that found for Zion, in large part attributable to the fact that human error associated with Zion's manual ECCS switchover to recirculation has been eliminated. This results f rom a semiautomatic switchover system in the Byron /Braidwood design. He compared the Zion PRA with the nuclear industry's IDCOR program through which the ability of current plant designs to accomodate severe accidents was 6

MINUTES OF THE 298TH ACRS MEETING FEBRUARY 7-9, 1985 extensively studied. He pointed out that the Zion PRA broke new ground in assessing the phenomenology of severe accidents. The IDCOR program, however, Las provided new insights in the area of source term evaluation coupled with a more rigorous assessment of time delay involved with containment failure for dominant sequences. The IDCOR work led to source terms much reduced from those in the Zion study.

D. Okrent asked if Zion and Byron /Braidwood are similar or different in their current fire protection capability. G. Kloop indicated that the Zion fire evaluation was done before Appendix R and the Byron study was done with the realization that Appendix R would be in place. He explained that the possibility of fire contributed to the core melt frequency at Zion but essentially made no contribution to risk. He suggested that on that basis CLCO expects fires to have even a lesser contribution to either the core melt frequency or risk at Byron /Braidwood.

D. Okrent asked if Comonwealth Edison had done a systems i interaction study because of requirements to look at internal flooding for certain kinds of pipe breaks mentioned in the standard review plan. G. Kloop explained that CECO has met the regulatory requirements of the standard review plan and tc the extent that systems interactions fall within the purview of te PRA on Byron, systems interactions have been considered for bob Byron and Braidwood. He indicated that there had not been a walkdown of the Braidwood plant to ferret out interactions between safei.y and non-safety systems. J. C. Ebersole asked if the scenario on the i

complete loss of AC power, with the resultant loss of reactor coolant pump seals leading to the use of feed and bleed cooling, was included in the PRA. G. Kloop indicated that complete loss of AC power as well as loss of pump sealt was considered in both the Byron /Braidwood Plants as well as the Zion PRA. But, he indicated that with total loss of AC power, the Braidwood plant does not have a bleed and feed capability. The end of the sequence involves various recovery scenarios for both onsite and offsite AC power.

G. Kloop indicated that an extensive analysis of seismic margins was performed in the Zion PRA. A vigorous seismic analysis beyond the design basis events has not been performed at Braidwood, but some conclusions can be reached by comparison with Zion. D. Okrent j asked if the issue of liquefaction had been examined with regard to vulnerability of the excavated ultimate heat sink pond which is below grade. G. Kloop mentioned a calculated factor of safety of 2 before the onset of serious liquefaction. D. Okrent noted that this factor of safety of 2 would unfavorably bias the expected core

] melt frequency for internal events. G. Kloop agreed. D. Okrent asked if there was a question of interaction regarding the motion of one building relative to another. G. Kloop acknowledged that the Braidwood plant is only partially on rock, but since there is a common basemat the question does not apply. D. Okrent inquired about the seismic design of steam lines beyond the isolation valve and other lines carrying water connected to the steam line. K. ,

Kostal, Sargent and Lundy, indicated that all of the Class 2 piping 7

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

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[', MINUTES OF THE 298TH ACRS MEETING FEBRUARY 7-9, 1985 has been assessed for an SSE event and supports for all of the l piping systems are designed for loads associated with the SSE l

including the main steam line. D. Okrent asked if there are relays at the Braidwood plant that are subject to chatter in earthquakes and if this is an issue of concern to Braidwood. R. Treece, CECO,  !

indicated that all relays identified as safety- related or Class 1E I

' are qualified to the SSE. He was unaware of the available margin  ;

above the SSE for the relays W. Kerr inquired whether the contribution of human error to risk has been evaluated in the PRA with any level of confidence. G.

Kloop indicated that those responsible for the Zion study had explicitly as well as implicitly made a conscious effort to look at failures attributable to human errors that had occurred at Zion. ,

Specific human factors analyses were done on those human errors  ;

that affected critical plant operations. J. C. Ebersole asked how CECO balanced ATWS consec 2ences against the justification for not putting in a clear diversity for the scram breakers. G. Kloop explained that for Zion, it was found that scram breaker ,

reliability was poor and represented a significant contributor to the ATWS fault tree. hwever, after Westinghouse provided updated maintenance instructions for the scram breakers, their reliability i went up dramatically. In addition, when evaluating ATWS events for <

a Westinghouse plant, it was found that the scram breaker system represented a relatively insignificant contribution to the core ,

I melt frequency and essentially no contribution to public health and safety. j t

B. Shelton, Project Engineering Manager for Comonwealth Edison,  !

discussed the issue of the elimination of pipe whip restraints (see f Appendix VI). P3 indicated that Comonwealth Edison's position is that it woulci iike to minimize the number of pipe whip restraints and delete any whose elimination can be technically justified. He i

explained the basis for that position as primarily hinging upon.

experience showing that pipe failures are minimal and that the stresses in Section III, Piping Systems, are generally well below ASME code allowables. He noted that the improved access for in-service inspection would also reduce radiation exposure and the potential for binding of pipes. He indicated that the NRC has i recently given permission for CECO to remove some arbitrary intermediate pipe whip restraints and that CECO expects to get permission to delete additional 24 ASME Class 1 main loop piping restraints. D. Okrent asked the Staff if this involved an application of the leak-before-break concept to systems other than the primary piping. J. Knight, NRC, indicated that the Staff is stilt faithful to its commitment to come back to the ACRS if the i Staff contemplates applying the leak-before-break hypothesis to i systems other than primary piping. This situation amounts to an agreement between the Staff and Comonwealth Edison that they need not protect against arbitrary intermediate pipe breaks. It is an application of a Branch Technical Position to cases where criteria for avoiding a postulated break such as low stress and low fatigue

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are met. In answer to a question by C. Michelson, J. Knight indicated that no jet impingement is assumed at these intermediate 6

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. MINUTES OF THE 298TH ACRS MEETING FEBRUARY 7-9, 1985 1ccati as. D. Okrent and C. Michelson argued that the Staff should postulate a break anywhere on any pipe. K. Green, Sargent and l

i Lundy, indicated that for temperature, pressure and humidity, i breaks in all areas where high energy pipes run are considered at a i very early stage of plant design. Although pipes with defects have i been eliminated, environmental qualification parameters are

! considered regarding qualification of all equipment.

i K. Ainger, in answer to a question by J. Ebersole concerning hot shorts due to fire, indicated that Consnonwealth Edison had conducted a review of all fire areas in the plant considering -

failure of cables and spurious operation of equipment due to hot i shorts. The results of the review, in the containment and auxiliary building, indicated spurious operation of the pressure 1

1 i operated power relief valves possible with coincident loss of the

ability to close a block valve. As a result, a few reactor vessel j

control cables were routed to eliminate that possibility. In .

addition, a motor operated containment isolation valve was replaced with a check valve to reduce the possibility of fire-induced ..

spurious operation. l J. Golden, CECO, briefly explained the three state emer planning program for the Braidwood station (see Appendix VI)gency . D.

! W. Moeller asked if there would be any benefit from combining the Emergency Operating Facilities (EOF) for the Byron and Quad Cities l plants. J. Golden indicated that this option was considered but l

l that the NRC thought that the EOFs were far removed and would be best left split. He also indicated that the the town of Morrison j will be the EOF for Carol County if Comonwealth Edison continues '

< the new emergency planning program.

F. Willaford, Comonwealth discussed physical plant security for the Braidwood plants. Edison (Note:

this closed session is reported in a proprietary supplement to the minutes.)

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!!1. Fire protection and Nuclear power Plants (0 pen to Public)  ;

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[ Note: H. Alderman was the Designated Federal Official for this t portionofthemeeting.]

j C. Michelson referred to a briefing by R. Vollmer of the Staff at l

the 297th ACRS meeting regarding the activities of the Fire i Protection Steering Committee to examine the state of affairs of l fire protection implementation in nuclear plants and to look at a

! differing professional opinion raised by some fire protection j reviewers and inspectors. He noted that a number of questions were e i

raised during work shops held by the NRC concerning interpretations ,

of Appendix R. He indicated that these were clarified by 1

appropriately prepared responses and a generic letter 83-33 which

• l put forth the Staff's position based on the review of exemption l

requests and inspection results. He noted that the Staff had  !

negotiated a number of outstanding issues with the industry in ,

order to give the industry more latitude and flexibility in solving  !

some of their fire protection problems. Some of these activities,

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MINUTES OF 'HE 298TH ACRS MEETING FEBRUARY 7-9, 1985 of ccLese, led to the establishment of differing professional opintors which are now in the review process. The differing profess ional opinions focus on industry making its own  ;

interpr3tation of Appendix R, shif ting the burden on to the Staff to justify unacceptability of modifications on a safety basis  ;

through the inspection process.

C. Mic'elson indicated uat the conclusions of the Steering Comittce Report include the following: l e extensions will no longer be granted to licensees on their i schedule for completion of fire protection features 6 an expedited fire protection inspection program will be 4 instituted e liensees will be' required to have valid analyses supporting l theia fire protection features and these analyses must be ,

avaliable to NRC inspectors. ,,

! He mentiorld that the February 5,1985 Subcomittee Meeting heard 4 briefings on four principal issues -- Appendix R implementation '

l progress at1 problems to date, the utility experience with Appendix i

R, the statas of generic issue 57, and fire protection research at .

Sandia Naticnal Laboratories. l V. Benaroya, NRR Chemical Engineering Branch, presented a short

history on I,opendix R (see Appendix VI). He indicated that a

Branch Technt

al Position on fire protection guidelines for new l

plants was ist Jed as a result of recomendations which derived from a study of the fire at Browns Ferry (March 1975). In August 1976, i the Staff issued Appendix A, a Branch Technical Position for i operating plants and plants under construction. He noted that most l nuclear power plants being reviewed today come under Appendix A  !

j guidelines (a backfit). C. Wiley mentioned a fire in a nearly ,

l completed nuclet* plant in Italy which destroyed the cable  :

l spreading room anf the control room. V. Benaroya pointed out that the Staff's guideilnes specifically state that the guides are not l

) applicable until fuel is on site. During the construction phase. .

OSHA has jurisdiction for fire protection during the construction l phase. C. Wiley pointed out that the fire in Italy was a massive l'

cable fire which occarred just before startup and was a precursor i i to the event at Browns Ferry.

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V. Benaroya explained that Appendix R refers only to open issues in the SERs for specific nuclear plants. It does not re) resent r guidelines. Appendix R has only 15 items with no definit'ons or i explanations. Of these, the Comission decided to backfit three --

fire protection of safe shutdown capability, emergency lighting.

4 and installation of an oil collection system for the reactor coolant pumps. He noted ttat Appendix R also had a 50.48 section which consisted of a modification completion schedule and very  ;

tight implementation dates. The result was the receipt of more i i than 500 exemption requests for a delay in meeting the minimum l

4 10  ;

MINUTES OF THE 298TH ACRS MEETING FEBRUARY 7-9, 19d5 requirements set by the Comission. He noted that about 225 of these werc denied based upon the fact that they were nearly blanket exemptions with no justification or explanation as to the reason for the exemption. Excmptions that were explicit regarding explanations were considered in more detail.

V. Benaroya explained that according to Appendix R, the only thing that NRC was to review was an alternate shutdown system. Regarding the other required items, utilities were supposed to comply and the Staff would accept a declaration from the licensee. J. C. Ebersole suggested that the Staff ought to make some statement about the relevance of this hazard in a comparative way to that to other challenges to core safety, such as seismic events and other external events. He pointed out that at the Subcomittee meeting, it was noted thct for budgetary and other considerations this has a minor position in comparison to tb efforts made in the seismic program.

V. Benaroya indicited that the industry appealed Generic Letter 83-33 in March 1984 and this caused the differing professional opinions f rom the Staff. The industry's position was that the additional explanations in the generic Ictter were interpretations that went beyond Appendix R and were, therefore, not part of the regulation. They contended that interpretation should be made by the utility and results made available to the NRC inspector at the time of review. V. Benaroya indicated that the differing professional opinion is part of the work of the Fire Protection Steering Comittee and the outcome hinges on how these interpretations are to be handled (whether the interpretations are to be resolved outside the specific boundaries of Appendix R).

V. Benaroya presented a table of completed safe shutdown inspections, noting that to date only the Calvert Cliffs Plant was found to be acceptable. He indicated that, however, the Staff expects that more inspections will be successful in 1985 as there is new a greater understanding on the part of utilities as to w'at n is required.

G. A. Reed pointed out that at the Subcomittee meeting, there was no agreement by fire protection engineers on the best way to institute or implement fire protection or even the equipment to use. He expressed concern regarding the request of the Steering Committee Report for very prompt action in application of fire protection when there is still basic disagreement on the use of gaseous versus water systems and where to use water deluge systems.

He suggested that this would be a difficult subject for the Comittee to handle, deciding on what the ultimate fire protection system should be. D. Okrent pointed out that there has been little money expended on NRC research on methodology to analyze the risks from fires, a very small program compared to almost anything else.

The Comittee discussed differences in the applications of the Branch Technical Position Appendix A to newer plants and the use of the Appendix R regulation on operating plants. V. Benaroya pointed out that from experience, if a fire protection engineer and a 11

MINUTES OF THE 298TH ACRS MEET!NG FEBRUARY 7-9, 1985 systems engineer work together to solve a fire protection problem, correct applications are usually the result. Most problems occur when fire protection decisions are made not by fire protection engineers but by others with little knowledge of the objective of the modification.

C. Michelson regretted that a spokesman for the Calvert Cliffs Station could not address the Comittee. He pointed out that one hallmark of the Calvert Cliffs effort was the fact that they established a team which took a systems approach to the problem by subdividing their plant into various fire zones which were in general bounded (in their arrangement) by three-hour fire walls.

They catalogued the important shutdown equipment, room by room, so that they were able to identify what was needed for shutdown and identify what had to be moved in order to accomplish an appropriate safe shutdown configuration. They did a good job of reanalyzing their plant to verify that they were in compliance with Appendix R.

C. Michelson discussed Generic Issue 57, "The Effects of Fire Protection System Actuation on Safety-Related Equipments." He noted that the Staff is still in the process of determining the priority of this item so that there has not yet been a final resolution. The expectation is that the frequency of this type of event would be extremely small. It is not likely that it will be assigned a high priority.

C. Michelson discussed fire protection research at the Sandia National 1. abo ra tory. He described this research as fire characterization tests, enclosure fire tests, and some equipment damage threshold tests. He suggested that the equipment damage threshold tests are particularly important because they give some

degree of identification of the vulnerability of components to fire protection measures. He mentioned the small program on enclosure cabinet tests, concentrating on fundamental components like switches and relays. He noted that it has already been found that terminal boards are susceptible to moisture condensation processes, and that they should be protected from actuation of the fire system. D. Okrent asked if portions of the fire protection system are seismically Class 1 de.igned. V. Benaraya indicated that there are no requirements for fire protection to be seismically qualified except in California and in new plants. P. G. Shewmon asked what assumptions are made regarding the fire source with respect to Generic Issue 57 involving an inadvertent actuation of fire protection systems and their interaction with safety related components. C. Micholson noted that one of the issues being dealt with in Generic !ssue 57 is identifying the nature of the fire source, a subject of sone research.

C. Michelson mentioned a memorandum he wrote to R. Vollmer concerning the results of the September 1984 Fire Protection Subcomittee meeting in which was solicited a clarification on the regulatory basis for the protection of equipment against adverse fire protection actuation. He indicated that a January 9,1985 Subcommittee letter was answered on January 29 outlining the 12

MIhuTES OF THE 298TH ACRS MEETlNG FEBRUARY 7-9, 1985 regulatory positions NRC expects licensees to follow. H- .ited regulatory guides that the utilities appear to be using without taking exception. He suggested to the Committee that there does not appear to be a basis for writing a Comittee letter at this time, particularly because of the unsettled nature of Generic Issue

57. J. C. Ebersole suggested that the Comittee's involvement in this issue ought to extend to evaluation of perhaps a dozen examples of what are thought to be the worst deviations from compliance within the general rules of the Branch Technical

' Position Appendix A.

IV. Pressurized Thermal Shock of Reactor Pressure Vessels (0 pen to Public)

[Ncte: E. G. Igne was the Designated Federal Official for this portionofthemeeting.]

P. G. Shewton indicated that D. L. Basdekas has urged the CRGR and the E00 not to issue the PTS rule as currently written. He believes that the screening criteria in this rule are set too high l

! and urges that the E00 review this matter. He recomended that the Comission withhold issuance of the rule until this review is complete. He indicated that the Comission has written that they will hold up the issuance of the rule until the ACRS has reviered D. L. Basdekas' concerns. He mentioned D. L. Basdekas' two assertions:

1) The steel pressure vessel is more prone to fracture than assumed by the Staff. This is due to the presence of undetected flaws er due to a higher transition temperature.

2) There is evidence that credible transients exist which would take the pressure vessel to lower temperatures than is censidered credible by the Staff.

P. G. Shewnon indicated that he saw no merit to the first assertion and that he had asked I. Catton, ACRS consultant, to look into the second assertion (see Appendia VI!!). He concluded that the ACRS shculd recomend that the Pressurized Thermal Shock (PTS) rule be issued as it is with the additional recomendation for prompt completion of some calculations regarding the effect of steam generator overfill.

1. Catton described the Basdekas' transient as a sequence of events involving a turbine trip, steam generator overfill, some means of rupturing the steam line either through water hartrer, or the occurrence of dead loads leading to a massive steam generator tube rupture. He indicated that one question involves whether 25 degrees of subcooling will lead to a water hamer event as a consequence of the overfilling of the stean line. He noted that the more probable event is a slow overfeed of auxiliary feedwater heated up by decay heat which will cause 100 or so degrees of subcooling.

13

MINUTES OF THE 298TH ACRS MEETING FEBRUARY 7-9, 1985 I. Catton indicated that he could not answer the question of how much mixing occurs in the top part of the steam generator or whether the cold water is going to get all the way to the bottom of the steam generator. He thought, however, that under certain circumstances there is the possibility of water hammer. He stressed that if there is a main steam line break and tube rupture, one cannot have a PTS event because the requirements for PTS are both low temperature and high pressure. There is an absence of high pressure in this scenario. D. Okrent suggested that the operators might inadvertently repressurize the system. 1. Catton l thought this was quite improbable in the case of such a steam generator tube rupture. D. Okrent suggested that there is the opportunity to depressurize under such conditions but one cannot sty that the situation is forever unisolable and repressurization may take place. I. Catton directed the question to J. Reyes of the l Reactor Systems Research Branch since they had done a calculation I

of a main steam line break under normal circumstances and from an i overfill. He suggested that the question is how fast the system l blows down. ,

l J. Reyes indicated that the main steam line break case was I calculated assuming a previous filling up of the steam lines. He indicated that the blowdown is slightly longer when the steam lines I are filled but the overall temperature difference with a filled steam generator is not very large. D. Okrent still thought one i might be able to isolate after the rupture. 1. Catton thought that J. Reyes would incorporate that assumption.

I. Catton mentioned that B&W has been improving their systems by the addition of various kinds of safety grade equipment to handle the overfill transient. He thought that some nuclear plants already have this equipment installed, some have the modification in process, and all B&W plants will have it eventually. R. Hernan, NRC, corroborated the fact that the Cystal River Plant will be installing an Emergency Feedwater Initiation and Control system (EFIC) later this year and the Arkansas nuclear units will also be implementing this modification which includes safety grade steam generator overfill protection. I. Catton noted that one should not give much credibility to the Basdekas' transients in a probabilistic sense without considering the impact of these proposed changes in the B&W design.

D. L. Basdekas, NRC Office of Nuclear Regulatory Research, nontioned several references which he claimed presented his views on pressurized thernal shock. One article mentioned described the German philosophy in addressing the problem of vessel integrity, a paper from a journal entitled, " Nuclear Engineering International" (see Appendix VI!!). He mentioned a report entitled, "The Basdekas' Transient" by I. D. Catton dated February 4,1%5. Four references were mentioned as offering an historic perspective on the issue:

e NUREG-0153, December 1976 14

i

. 1 MINUTES OF THE 298TH ACRS MEETING FEBRUARY 7-9, 1985 1

e a December 20, 1976 memorandum from D. L. Basdekas to B.

Rusche e a letter dated May 28, 1980 from D. L. Basdekas to Congressman Udall (see Appendix IX) e a licensing board notification transmitted by D. Eisenhut to all licensing boards (this contains the differing professional opinion by D. L. Basdekas)

D. L. Basdekas disagreed with the statement in the report by I.

Catton that multiple steam generator tube ruptures will result in reduced primary steam pressure. He suggested that even if there are multiple tube ruptures, the loss of flow may be small enough that the flow may be made up by the ECCS maintaining high pressure in the primary system. P. G. Shewmon cautioned D. L. Basdekas that a critique of I. Catton's report was not of primary interest to the Comittee unless it related to his basis for nonissuance of the PTS rule. D. L. Basdekas stated two basic reasons for his recommendation that the PTS final rule, as presently formulated, not be issued. He indicated that the screening criteria are unrealistically high and will not be meaningful because flux reduction measures have already been taken or are in progress. The benefits expected from flux reduction programs will not be realized beyond what has been'done up to the present. He suggested that there is no reason to issue the PTS rule since there is no benefit to be realized from its issuance. He referred to the Federal Republic of Germany paper he had released to the Comittee and noted their position against using probabilistic arguments in their approach to PTS. He suggested amending the screening criterion to a more realistic 150*to 200*F and use of the West German method for defining pressure and temperature. P. G. Shewmon asked the basis for the criterion of 150 to 200*F. D. L. Basdekas cited operational experience to date which he had mentioned earlier in the discussion to shcw that the temperature can drop to abcut 200*.

D. Okrent asked if flaws left after the fabrication of the vessel are important to PTS. If these flaws exceed ASME allowable levels, is this an important contributor to the pressure vessel failure likelihood? He also asked if this question is independent of temperature or the location of the flaw. M. Vagins, NRC, indicated that the distribution and size of the flaws are significant, fm if there were no flaws there would not be a PTS issue. You must O ve a flaw to initiate fracture. Nevertheless, even if the utility continued to operate the plant, a small flaw would augment the probability of failure by only about a half an order of magnitude.

F. Schroeder. NRC, indicated that the Staff had outlined before the Comittee on several occasions the basis for selection of the screening criteria in the proposed rule. He saw no reason on the basis of D. L. Basdokas' concern to hold up issuance of that rule.

He acknowledged that more work remains to be done in the area of steam generator overfill transients and work is in process in that area. He noted that independent of the PTS issue, ruptures of the 15

MINUTES OF THE 298TH ACRS MEETENG FEBRUARY 7-9, 1985 main steam lines and multiple steam generator ruptures are in themselves enough of a concern to cause action to reduce the probability of such a transient without raising the specter of whether it is also a PTS problem. He suggested that altering the PTS screening criteria is not a way to fix the steam generator overfill problem. A more appropriate method would be to reduce the frequency of such events.

D. Okrent noted that D. L. Basdekas suggests that surface faults which are important for PTS, are not detectable by current methods.

He asked the Staff if this were so. M. Vagins agreed but indicated that the Staff has assumed that those flaws are there in its calculational distribution with a probability close to 1 anyway.

For very, very small flaws the probability of 1 is assumed. He also added that the problem is not as bad as it was four years ago.

Te',ts are now detecting near surface flaws. It is flaws in the cladding itself that still pose a problem.

V. Backfitting of Nuclear Power Plants (0 pen to Public) ,

[ Note: P. A. Boehnert was the Designated Federal Official for this portionofthemeeting.]

H. W. Lewis mentioned a presentation about the proposed backfit rule that was made by J. Tourtellotte at a Subcomittee Meeting held on February 6,1985. He noted that the backfit rule seems to be moving through the Comission. He mentioned that a briefing on the draft Manual Chapter 0514 indicated that, in its current version, it is not in confonnity with the proposed rule or some other rules of the Comission. He cited certain errors in the draft Manual Chapter that primarily concerned definitions such as, "What constitutes a regulatory requirement"? He mentioned a presentation on the specific issue of the probable raximum precipitation at Beaver Valley noting that he left that discussion in greater confusion than when he entered because of questions about which portions of the issue were dealt with probabilistically and which were handled ceterministically. Techniques used by the Staff were derived from flational heather Service data. There is l question whether the f4RC's position on issues like this should change every time another Federal agency changes its position. If so, should the f4RC Staff then impose a backfit as a consequence of this change. He indicated that he was in favor of issuing the backfitting rule.

O, Okrent indicated that he would like to examine the role of uncertainties in the backfit rule and find out J. Tourtellotte's thoughts as to whether uncertainties are considered in the published version that is going out for public coment. He expressed reservations regarding adopting the proposed rule in its current form.

D. Okrent indicated that he certainly would not want to be a roadblock to halting unnecessary backfitting but he was concerned that the f4RC Staff might be shackled by this wording in its 16

i '. MMUTES OF THE 298TH ACRS MEET!NG FEBRUARY 7-9, 1985

- attempts to demonstrate a necessary backfit in the face of uncertainty. W. Kerr thought that the use of bulletins or information notices would be a better way at backfitting without requiring the type of analysis being proposed. J. Tourtellotte r i spoke of a certain discipline in the exercise of regulatory '

authority and that this responsibility lies with the Comission.

W. Kerr thought that one of the problems with current backfitting '

practices is that too many people can make decisions. He thought that the decisions ought to be more nearly centralized and more uniform. J. Tourtellotte agreed and indicated that, per the backfit rule provisions, a backfit is ultimately reviewed by the l EDO. The backfit will not be implemented unless the EDO agrees that it can be implemented.  ;

H. W. Lewis suggested that one should separate two significant  !

elements of the proposed rule. One involves moving the ,

decision-making process to a higher and more coordinated level  :

while the second element is the requirement for an analysis. He .

r suggested that the question is whether one would be significantly improving the climate of backfitting by doing the first while not l

j necessarily doing the second. J. Tourtellotte thought that it was not enough to require the EDO to be responsible without saying that i an analysis must be done using certain general guide. lines.

W. Kerr stated his belief that a number of decis' ions have been based primarily on good engineering judgment since the analysis process is significantly compromised by the magnitude of uncertainties. He thought it counterproductive to spend so much i

I effort on a questionable analysis process. J. Tourte110tte agreed that the reality may be reviewer's rationalizing preconceived notions for a decision and then seeking justification for that 1

decision. D. Okrent suggested that the requirement that there be a

} substantial increase in the overall protection of the public health ,

i and safety may result in abandonment of the defense-in-depth l t concept. Safety improvements in one line of defense against undue

! risk should not be disapproved or approved based solely on the i l presence or absence or another line of defense to cope with the j failure of the first.  :

I J. C. Ebersole expressed concern regarding the statement in the  !

i rule which indicated that the Comission would not ordinarily l I expect that safety ir:provements would be required for backfits

! which result in an insignificant or small benefit to public health I and safety. He indicated that many situations stand for years

! identified as being insignificant or of small benefit until t

triggered by a near catastrophe as in the case of the Salem scram  !

j breaker event. He suggested that the cost would have been

! insignificant to fix the scram breaker problem and the Salem event 1

would never have happened, i

l F. J. Remick pointed out that the rule now gives a definition of  !

backfitting that is tied to regulatory requirements, and industry j coments indicate that that is not adequate. He asked if the  :

Comission has decided to go toward the industry definition of .

I 17

MINUTES OF THE 29BTH ACRS MEETING FEBRUARY 7-9, 1985 backfitting and if the rule original'y had a definition which was somewhat closer to what the industry is currently proposing. J.

Tourtellotte indicated that the rule originally did have a definition of backfitting closer to the industry proposal, but the shift was made from rules, regulations, and orders to regulatory requirements with an explanation in a statement of considerations.

He acknowledged that there is a real problem generated by the fact that the Staff and licensees are confused about what the requirements really are. He indicated that in his view what is  !

needed is to state clearly, once and for all, that requirements are only those that are in the regulations and everything else such as regulatory guides, standard review plans, and Branch Technical Positions are guides er advice as to how one might proceed but are not requirements. Unless that distinction is made, the problems of the past will continue. He agreed with F. J. Remick that he was inclined to a definition sonewhat closer to the industry proposed definition. D. Okrent acknowledged that regulatory guides and standard review plans are not regulations but he expressed concern i in going back to the situation in 1967 where the only regulations , ,

were the vague general design criteria which left designers as well as the industry in a quandary as to whether a Technical Position would be defensib'e to the NRC. J. Tourtellotte suggested that guidance which is essential to safety ought to be a requirement.

The NRC ought to amend the regulations and incorporate such l technical guidance. He thought the Staff's current approach of

" waving" the regulatory guide at a licensee, as a condition for a license, ought not to be accepted procedure. C. P. Siess thought that regulatory guides are really not much of a problem since they usually have a clear implementation section. There is no problem J

with an applicant referencing a regulatory guide in his PSAR and committing to it. J. Tourtellotte agreed that a comitment such as i

this is not a requirement but . omething the licensee agrees to do i

voluntarily. C. P. Siess pointed out an obvious problem with the j attempts of the Staff to backfit an existing regulatory guide,

! which is currently being updated.

, H. W. Lewis mentioned that there were a number of minor substantive

! defects in the draft Manual Chapter. He indicated that it was his  ;

understanding that the Staff's definition of a licensee as a l corporate entity would be changed to conform with what is contained i in the regulations. He asked, however, why Regulatory Guides are l still listed in the proposed draft Manual Chapter as regulatory l l requirements. D. Cox, NRC, indicated that the Staff acknowledges  !

i the " looseness of wording" in the draf t Manual Chapter and intends t i to rewrite the terms regulatory requirements as regulatory 4 positions. He tried to explain that the list, including Regulatory I

! Guide, Standard Review Plan, Branch Technical Position, and IE information notices, was to identify those instruments that are i

considered to place an obligation on a licensee in some way. H. W.

Lewis pointed out that Regulatory Guides do not put an obligation i

on the licensee unless they have been adopted by reference in the PSAR. T. Cox explained that in practice over the years the Staff has been operating to, in effect, bind licensees to instruments i

such as the Stardard Review Plans as an accepted method for meeting NRC regulations. F. J. Remick pointed out that the original i 18

M8NUTES OF THE 298TH ACRS MEET!NG~

FEBRUARY 7-9, 1985 -

1 definition of backfit as proposed by the regulatory reform task

! force was changed to incorporate the terminology of regulatory

! requirements. He suggested that this was an unfortunate choice of words. If the language were changed to more closely fit the industry definition of a backfit, the wording problem would be .

eliminated. J. Tourte11otte agreed that, legally, one can only l

impose a requirement by regulation or order. There is no other way that the NRC can impose a requirement.

l D. Okrent asked about a letter from the NRC Chainnan dated September 29, 1983 to Senator Mitchell in which the Comission indicated that it knew of no instances in which I

7 unnecessary backfits were imposed. He asked how the Comission could make a statement such as that. J. Tourtellotte explained that the problem was that no records have been kept of any of these additional requirements that have been imposed by the Staff. Some of the requirements that have been imposed might have been termed as regulatory duress, but there was no record kept. The fact that it left the impression that there were no backfits was probably ,

misleading to Senator Mitchell.

VI. ACRSMeetingwiththeNRCCommissioners(0pentoPublic)

[ Note: Chairman N. Palladino and Comissioners T. Roberts, J.

Asselstine, F. Bernthal and L. Zech were present.]

Chairman N. Palladino indicated that the ACRS was prepared 'to i discuss the use of the " check operator" concept for licensed reactor operator requalification, experience levels required for i reactor operators, the status of ACRS activities on backfitting for nuclear power plants, and quantitative safety goals. Coninissioner F. Bernthal expressed his pleasure at seeing a more expanded scope than in recent previous meetings. ACRS Chairman D. A. Ward noted that there are no new developments on the quantitative safety goal issue and suggested deferring the discussion to a later date.

F. J. Remick briefly discussed the history of the designated representative or " check operator" concept as it relates to the i Comission. He mentioned that the Comission had authorized a QA initiative to study the designated representative concept particularly in the area of construction of nuclear plants, several ,

years ago. The Ford Amendment to the NRC Authorization Act directed the NRC to conduct a study for improving quality assurance i and quality control in the construction of nuclear power plants.

The designated representative concept was included in the Ford Amendment study. He indicated that limited Staff resources and a ,

perceived lack of interest on the part of the Staff, as well as a i perceived lack of interest on the part of the Comission, resulted in scant attention to the designated representative concept in NUREG 1055, the Staff's report to the Congress under the Ford Amendment. He pointed out that the ACRS latter of March 21, 1984 i regarding review of NUREG-1055, indicated that the concept of using designated representatives (DR) was worthy of further consideration  ;

I 19  ;

. MINUTES OF THE 298TH ACRS MEETING FEBRUARY 7-9, 1985 anc might be a way of stimulating and rewarding professionalism and dedication to quality in the work place.

F. J. Remick explained that a majority of the members of the ACRS Subcommittee on Quality and Quality Assurance in Design and Construction concluded that under current regulations, the designated representative concept had little utility in the design and construction of plants unless there were NRC mandated l

certification hold points as part of the construction and design I requirerents. Subcomittee members did believe, however, that the concept of using designated representatives as check operators for operating reactors would make some sense, and be worthy of further consideration regarding improvement of the operator requalification process. He suggested that the check operator concept, if eventually implemented, could provide a controlled limited pilot program for the testing of the broader use of DR's in nuclear regulation. He stressed that the Comission's consideration of the check operator, concept would not be in conflict with the Comission's position in October 1984 which instructed the Staff ,

not to look at a pilot program for designated representatives (this effort involved only a pilot program for construction). F. J.

Remick thought that SECY-84-167 "Use of the FAA Chock Pilot Approach for Reactor Operator Qualification," was a good background document for an initial discussion between the Staff and the Comission regarding the check operator process. Chairman palladino agreed that a meeting ought to be scheduled with the Staff to discuss accelerating this effort.

H. W. Lewis suggested that the Comission might want to restructure the operating reactor requalification examination to incorporate the check operator system. He noted that it works extrenely well

> in the case of the check pilot system at the FAA. He suggested that one might want to split the examination up into a part that could be done by a designated representative and a part that would be administered by the Staff. Chairman Palladino noted that the SECY document suggested use of check operators for a class of plants so that one can avoid the question of DRs examining individuals with whom they work closely. Comissioner Bernthal agreed that the testing ought to be split up between insider and outsider so that part of the test is administered by an individual familiar with that plant. G. A. Reed indicated that he would like to take the other position in the interest of expediting the requalification process, that the test ought to be administered by the lead SRO at a plant. Therefore, the outstanding person in a plant should be designated the check operator for personnel in that plant.

Comissioner Asselstine asked the ACR$ to define what an acceptable reactor operator requalification program is and to examine the

! policy of performance. oriented exam' nations of requalification programs where NRC regional inspectors are examining the operators themselves (20%) rather than reviewing the programs. He asked whether the ACR5 thought that the check operator was a way of l

20

MINUTES OF THE 298TH ACRS MEETING FEBRUARY 7-9, 1985 minimizing these kinds of problems. He requested that the ACRS specifically consider the following:

e take a general look at the requalification program regarding whether it is working properly e explore the impact of the requalificatian program on regionalization e decide whether the objectives of the requalification program are correct e consider what the state of engineering knowledge is on shif t and how requalification programs can help to encourage a gradual increase in engineering knowledge on shift Chaircan Palladino indicated that one of the ways the check operator concept could help the whole requalification program is by the training of the people who are going to carry out the check ,

operator function. He agreed that studying the entire requalification program was not a bad idea and he thought the Commission could benefit greatly from ACRS input.

F. J. Remick noted that there is a demoralizing effect on reactor operators from having an NRC person administering the requalification examinations. He thought that a check operator, a true peer, administering the examinations, would have a salutary effect since they would be currently licensed on either this facility or another facility. Comissioner Asselstine agreed that the respect and aura that would go with the check pilot position is something that should not be discounted. F. J. Remick thought that it would add and encourage professionalism within the entire operation.

Chairman Palladino thought that if one developed a trial program with volunteer utilities, one could learn a bit more about how it could work. F. J. Remick mentioned negative reaction from the industry about two years ago when this was proposed. At that time the requirement for the SRO was just coming out and thPO was asking the industry to provide two SR0s for each of their evaluation visits. Even though it is rcw a couple of years later, F. J.

Remick was not sure how the industry might react to such a request for a trial check operator program. Comissioner Asselstine noted that industry would have to provide top notch experienced operators if the designated representative question were to be developed and industry conperation and support would be crucial to any attempt at developtrent of a DR program. G. A. Reed thought that industry would have difficulty finding sufficient personnel to implement a program of having licensed $ROS from one plant examine operators at another. He thought that if the check operator activity was within the sano plant, you would get the same professional Itcerted person doing the requalifications and it would be superior to the current situation where NRC is administering the requalification examinations. Comissioner Asselstine cocinented that if the check 21

8 MlNUTES OF THE 298TH ACRS MEETfNG FEBRUARY 7-9, 1985 operator concept works well in the requalification program, should not one consider sometime in the future expanding the effort into initial licensing as well? F. J. Remick suggested that this would require a revamping of the licensing process. G. A. Reed thought that NRC regulatory people should continue the initial licensing evaluations which tend to be somewhat more conceptual and theoretical than the requalification process (involves details).

Cemissioner L. Zech was critical of the training departments in nuclear plants. He thought it extremely important that the senior people in the training department be licensed on the plant as SR0s.

He noted that in many utilities they apparently are not. He was strongly in favor of improved training including operator and requalification training but he was not enthusiastic about the check operator concept. He wanted time for further study of the matter. Chairman Palladino suggested that it would be appropriate to meet with the NRC Staff, using SECY-84-167 as an information document, regarding the possible development by the Staff of an action document for some sort of trial program.

G. A. Reed explained that the ACRS, in its letter to the Commission dated October 16, 1984 concerning experience level for licensed persennel (SR0s and R0s), strongly supported " hot" participation for a period of six months as a worthwhile experience. Licensing at a host similar plant could, however, have negative aspects regarding confusion on the part of the operator over details as the plant moves to initial operation. He stressed that he wanted to correct any misapprehension in an E00 letter of December 28, 1984 that purported to state that the ACRS had encouraged that reactor operator candidates be licensed at host operating plants. F. J.

Remick suggested that there were differing views in the latest Comittee oiscussion of this issue. He indicated that he would not propose that individuals are required to obtain licenses at host facilities. But, it was a plus if they did, in contrast to just cbserving at a facility for six months. Chai rinan palladino indicated that the NRC Staff reconsidered this requirement and i

cecided not to include it in the revision to Regulatory Guide 1.8.

Comissioner Asselstine suggested that for nuclear plants that will not go into operation until far into the future, one might like to see a year rather than just six months of " hot" participation experience. He indicated that it was his sense that the six months was really driven by expediency and the inventory of candidates available. D. A. Ward agreed that six months is probably acceptable at the present time because there is a linited number of SR0s. Put, he agreed that a year is probably more appropriate for these very complex jobs. W. Kerr was somewhat skeptical that a visitor to a plant would get the same esposure and learning experience as a member of the plant staff. He thought that one gets a lot of experience in starting up a plant and he had not seen anything that would justify six months versus a year of " hot" participation experience. Corrnissioner Asselstine acknowledged the weakness of the " hot" participation experience in that a visitor would not get the hands on experience that a plant staff member 22

FEBRUARY 7-9, 1985

'. MlNUTE: OF THE 298TH ACRS MEET!NG  ;

l would get actually operating the plant and managing ongoing maintenance work. G. A. Reed mentioned again that " hot" participation experience can be very important. He stressed that what is important to such a visitor would be concepts and principles and not detailed manipulation.

Comissioner Bernthal speculated on whether it would not be a good i idea to require that all newer plants have plant-specific simulators as a way to train operators. D. A. Ward indicated that the Comission is pretty close to requiring it now. Comissioner Bernthal asked whether it is about time that the Commission look at a " flat" requirement far a simulator. Commissioner Asselstine agreed. Comissioner Zech spoke out against simulator regulation but encouraged the effective use of simulators as a management comitment to excellence in performance. Chairman Palladino cautioned against making any decision at this time but indicated j

that the Ccmission has an interest in exploring the subject further.

H. W. l.ewis indicated that the ACRS had written a letter to the Comission in the fall of 1984 in which it simply expressed continued interest in the subject of backfitting. It also stated that the draf t Manual Chapter the staff was writing seemed to be i out of synchronization with the proposals for a new backfitting l

rule to replace 10 CFR 50.109. He indicated that the Comittee ,

l does not often review draft Manual Chapters which are simply an i implementation of Comission policy. However, in this case there i was concern that the draf t Manual Chapter was not an explicit

implementation of the Comission policy on backfitting. He j

explained that it was his opinion that the Comittee is not

seriously at odds with the current version of the backfitting rule.

j On the other hand, there is real concern about several wording errors that need revision in the draf t Manual Chapter. Among i

these are the definition of a " licensee" as a corporate entity that I has a license and tme consideration of Regulatory Guides as t requirements.

j D. Okrent expressed concern that there is a considerable emphasis ,

4 on the use of cost-benefit analysis in decision-making without any reference to the existence and treatment of uncertainties. He  !

i suggested that the uncertainties' in the risk level will be i

substantial. Potential backfit situations may swing from being 1 clearly beneficial to not clearly beneficial based upon the large uncertainties. Commissioner Bernthal indicated that he assumed that uncertainties are implicit whenever one mentions cost benefit analysis, it is implied that uncertainties are considered. He maintained that the NRC regulates to uncertainties.

Chairman Palladino indicated that the presumption is that a nuclear 1 plant is safe without the backfit unless the Director of Nuclear j Reaction Regulation finds that there is so cor.pelling a reason for making the change that he directs it without following all of this

! procedure. He did not deny that there are uncertainties in cost l

benefit evaluations and thought that D. Okrent's suggestion would 23

,m #.<

- -._-.m .____-.t-_-.4 - . _ . - .2 _ . _ _ _._ ~ - _

HINUTES OF THE 298TH ACRS MEET!NG FEBRUARY 7-9, 1985 I be beneficial in improving the rule. D. Okrent asserted that the Staff will find it virtually impossible both to show a significant benefit to the health and safety of the public and to require a cost beneficial backfit in the face of large uncertainties. A i utility could argue legally forever that the Staff has not I demonstrated both a significant improvement and positive cost  !

benefit. W. Kerr suggested that if one accepts the premise that 4

the plant is already safe enough, it is only prudent that one demonstrate with a reasonable level of confidence that a backfit will provide an improvement and a significant change in the plant.

There is a chance that one will not only not provide improvement but make things worse.

Chairman palladino indicated that the initial decision regarding a backfit does not necessarily rely on all of the uncertainties because engineering judgment is factored into the decision-making process. He did indicate, however, that identification of the l

uncertainty is an important consideration and could be helpful to the rule. He welcomed ACRS comments on that point. ,

D. Okrent expressed concern that approval of the backfitting rule f will have a chilling effect on the NRC Staff's ability to put l through backfits which are worthwhile on some basis unless the  :

Comissioners themselves decide and pass a new rule.

Vll. Decay Heat Removal (0 pen) 4

[ Note: P. A. Boehnert was the Designated Federal Official for this portionofthemeeting.]

l D. A. Ward indicated that a meeting of the Subcommittee on Decay Heat Removal Systems was held on January 22, 1985 in Albuquerque, i hM, to review NRC Staff resolution efforts on USI A-45, " Shutdown Decay Heat Removal (OHR) Requirements." He explained that the i focus of the meeting was Sandia National Laboratory's Shutdown DHR Analysis Plan and, in particular, their analysis methodology for -

1

' external floods, extreme event wind contributors and tornado,(seismic, fire,). internal sabotage He noted and external that the '

Sandia work is a four step program:

e develop criteria for screening plants for OHR capabilities e apply criteria to plants and identify those with potential OHR problems j e assess in more (quantitative and qualitative) detail those

plants with potential weaknesses to determine OHR capabilities I e perform value-impact analyses of modifications designed to

! improve DHR reliability i

l 0. A. Ward indicated that the Subcommittee heard some partial j results of the third step noting that the first two steps involved the selection of nine plants which would represent the entire U.S.

l i 24

t MINUTES OF THE 298TH ACRS MEETING FEBRUARY 7-9, 1985 population of nuclear plants. He explained that the objective of the third step will be to assess contributions to the core melt probability from DHR vulnerabilities using PRA methods and data.

The intention is not to do a full scope PRA or improve PRA methodology. The intent is also not to assess outstanding USIs but to focus only on_ decay heat removal systems, and factor in the effects from extarnal events. C. Michelson pointed out that Sandia's analysis will not include rupture of high energy lines

i outside of the containment but that Sandia analysts will suggest ,

conceptual improvements with their estimated costs. D. A. Ward indicated ' that the fourth part of the Sandia plan will be to develop cost estimates of backfit designs to fit a range of The final ~ output of the program will be 4 conceptual improvements.

i cost beneficial improvements which are backed up by cost estimates 1 and should result in core melt probability reductions.

{ D. Okrent discussed the Research Development Associates (RDA) j approach for development of cost estimates for decay heat removal systems in Combustion Engineering's CESSAR plant. D. A. Ward .

indicated that A. Marchese, NRC, discussed the potential impact of j truncating the original plan for A-45 in order to preserve the a

schulule for resolution. Subcommittee members were unhappy with this proposal. The Subcomittee thought the

program ought to review all nine plants and the program not be l

trurcated to two plants just to keep on schedule.

4 D. A. Ward indicated that a further step in the A-45 program would i be casting the results into a rule. The Staff might develop a ,

trial rule while the last seven plants are being reviewed.

• e Sabotage would be considered in the rule by taking advantage of 4 previous Sandia work as background, as well as PRA work which identifies vulnerabilities. G. A. Reed was concerned that resolution of A-45 would be premature since conclusions would be based upon the first two nuclear plant designs neither of which were a B&W design. He thought consideration of the B&W design was l critical to proper resolution of A-45 and that conclusions should '

l incorporate a B&W design review. He suggested that review of a B&W l plant ought to be at the front of the program rather than relegated

to the position of being one of the last plants reviewed. K.

i Kniel, NRC, assured the Comittee that the Staff is definitely committed to review all nine plant designs. He did note that the Staff was not sure at this time how to reduce all of the technical findings to a regulatory licensing basis.

f W. Kerr asked how the Staff would decide how far to go to reduce

! risks. K. Kniel suggested that this would be determined by a collegial decision by the Staff. G. A. Reed thought that if a B&W 4 plant were reviewed at the front of the program, the Staff would be more likely to come up with an integrated fix rather than piecemeal

! modifications. J. C. Ebersole expressed concern that averaging of the vulnerabilities of the plants may lead to a watered-down rule.

He cautioned against treating DHR vulnerabilities generically and i

i 25 i

MENUTES OF THE 298TH ACRS MEETING FEBRUARY 7-9, 1985 suggested that it might be better to take the six " worst" plants and develop a rule from a study of these.

j VIII.NRC Safety Research Program and Budget (0 pen)

[ Note: S. Duraiswamy was the Designated Federal Official for thisportionofthemeeting.)

C. P. Siess mentioned several documents as background material for discussion of the draft Safety Research Program Report (note the 1 first six entries in the list of Appendix XI). E. Conti, NRC, confirmed that the funding for human factors research in the l research program had been zerced and the money transferred to NRR for their allocation. C. P. Siess suggested that the money was removed from human factors research in order to augment research on advanced reactors. He asked the Staff for some background on the

decision to remove the human factors funding from the Staff's research office, l

E. Conti indicated that one half million dollars was used to 1' augment funding for the damaged fuel code analysis. D. Okrent ,

noted that the Nuclear Safety Informstion Center (560,000 in funding FY 1986 )forwas reduced to zero. He asked what the meaning of that reduction was. C. P. Siess noted that work on the component fragilities of electrical components was still in the budget and had not been zeroed out as previously thought. D.

Okrent proposed funding to explore the impacts of design errors for i establishing PRA methodologies to be taken from the MELCOR computer ,

l model.

1 1 F. Gillaspie, NRC, explained that DOE would make up the $60,000 funding for fiscal 1986 for the Nuclear Safety Information Center that was dropped from the NRC research budget. He indicated, however, that if DOE decides not to assist with this funding, NRC would continue to contribute the $60,000 so that the Center was not 4

phased out. F. Gillespie corroborated that human factors research funds were transferred out of research and will be focused in NRR.

D. Okrent asked how the Staff might restructure the human factors program if it' could. F. Gillespie thought that the current approach of many small studies based upon the human factors program i plan applications would not be the way to go. He indicated that he i would concentrate on certain more vital areas such as maintenance and the man-machine interf ace independent of the human factors i program plan. Simulator and cognitive error research would fit correctly into the research area of reliability and ought to be retained outside of human factors considerations. F. Gillespie i noted that NRR has technical support funding for fiscal year 1986

in the amount of $3,500,000, the same as fiscal 1985 which would be j in direct support of the human factors program plan. D. Okrent

asked if there was any interest, from a research point of view, in i evaluating the variability of utility managements. F. Gillespie I

thought it would be rather difficult to correctly allocate funding.

i l

I 26 4

. MINUTES OF THE 298TH ACRS MEETING FEBRUARY 7-9, 1985 D. Okrent reccmended the setting up of a multi-million dollar human factors program with concentration in areas such as noted by F. Gillespie. G. A. Reed was particularly interested in research on selection of power plant employees taking into consideration their natural aptitude or ability. He thought it was a major flaw in the human factors area. C. P. Siess cautioned that the ACRS ought not to br. ':Jght up in being forced to design a human factors research program before any funding could be allocated. He suggested that the Comittee set aside several hours to discuss possible research in the human factors area and censider drafting a separate letter to the Comission on this subject as well as similar individual research subjects. The Comittee decided that the NRC ought to take a leadership role in setting up a substantial program of human factors research of the order of two to three million dollars per year funded in research in fiscal year 1986 and in ensuing years.

IX. Executive Sessions (0 pen to Public)

[ Note: R. F. Fraley was the Designated Federal Official for this portionofthemeeting.]

A. Subcommittee Assignments

1. Subcomittee Comments on High-Level Waste Repository Reports .

D. W. Moeller presented to the Comittee prepared comments of the ACRS Waste Management Subcomittee on several reports prepared by the NRC Staff relative to their review of the upcoming application of the U.S. Department of Energy (DOE) for a license to construct a high-level waste repository. In addition, the Subcomittee reviewed and prepared coments on Chapter 7, " Comparative Evaluation of Sites Proposed for Nomination," included in the Draft Environmental Assessments recently issued by D0E. The Members concluded that the coments of the Subcomittee should be forwarded to the NRC Staff without coment by the full Comittee.

2. Comission Meet ng on Equipment Qualification - January 6,1984 D. Okrent noted that W. Snyder, Director of Safety Programs, Sandia National Laboratory (SNL),

complained during a January 6, 1984 Comission meeting that pressure has been applied by the NRC Staff to SNL (indirectly) to control or inhibit reports which affect past or present regulatory decisions. D. Okrent suggested that this matter be assigned for consideration by an appropriate subcomittee.

27

MlNUTES OF THE 298TH ACRS MEETING FEBRUARY 7-9, 1985

3. Discussion of Issues in Nuclear Safety H. W. Lewis suggested that time be set aside during the 299th ACRS meeting (March) for an unstructured discussion of the areas of concern to ACRS members regarding important nuclear safety issues. Chaiman D. A. Ward pointed out that this subject was the purview of the new ad hoc Subcomittee on the State I of Nuclear Power Safety and ought to be its initial assignment.

B. ACRS Reports, Letters, and Memoranda l

l 1. Review and Evaluation of the NRC Safety Research Program for FY 1986 and 1987 The Comittee completed its report to the U.S.

Congress regarding the proposed NRC Safety Research Program for Fiscal Years 1986 and 1987. ,

2. ACRS Report on the Braidwood Station, Units 1 and 2 The Committee prepared a report to the Commissioners of its review of the application of Comonwealth Edison Company (the Applicant) for a license to operate the Braidwood Station, Units 1 and 2 at full power.

3. Proposed Rule on Backfitting The Comittee prepared a report to the Comissicners of its review of Comission progress toward the issuance of a revision to NRC rule (10 CFR 50.109) on backfitting of nuclear power plants and toward release by the NRC Staff of a new Manual Chapter 0514 to implement that rule. Changes to the report were suggested after adjournment and some members would not agree to the changes without holding another full Comittee discussion. Therefore, additional discussion on this matter was scheduled for the Comittee's 299th meeting on March 7-9, 1985. l

4. ACRS Funding and Staffing Reductions for FY 1985 and 1986 The Cocynittee prepared a report to the Comissioners regarding the major reductions that have been proposed in funding and staffing of the Advisory Committee on Reactor Safeguards' activities during the remainder of FY 1985 and during FY 1986.

28

MINUTES OF THE 298TH ACRS MEETING FEBRUARY 7-9, 1985

5. ACRS Comments on D. L. Basdekas' Professional Opinion l i Concerning the Proposed Final Rule on Pressurized i Thermal Shock The Committee prepared a letter to the ED0 regarding its consideration of Mr. D. L. Basdekas' technical opinion concerning the proposed final rule on ,

pressurized thermal shock.

6. ACRS Comments on Proposed Rule Changes to 10 CFR 50, Appendix E The Committee agreed not to comment on the proposal by the NRC Staff to delete the unusual event emergency classification from 10 CFR 50, Appendix E.

" Emergency Planning and Preparedness for Production and Utilization Facilities."

A t

C. G:neric Issues

1. Consideration of Seismic Events in Emergency Planning D. W. Moeller's draft letter regarding consideration of seismic events in emargency planning will be carried over from the February meeting. D. W.

Moeller requested that the ACRS staff obtain information regarding emergency evacuation procedures during seismic events from other countries and the studies required by the NRC Staff of the San Onofre and Diablo Canyon Plants.

2. Lack of Redundancy in Reactor Scram Systems J. C. Ebersole presented the draft of a letter to the Commissioners containing his views on the lack of diversity in PWR scram systems, offering the letter as either an independent viewpoint ce one from the ACRS. Time did not permit Committee discussion of this letter at the 298th ACRS meeting.

D. Future Schedule

1. Future Agenda The Committee agreed on tentative agenda items for the 299th ACRS meeting, March 7-9, 1985 (see AppendixII).

f 29

FEBRUARY 7-9, 1985 MihuTES OF THE 29BTH ACRS MEETlNG

2. Future Subcomittee Activities _

A schedule of future Subcomittee Activities was distributed to Members (see Appendix III).

E.

~'

s0 $

00 f,'fWlLB'

>,e #

F.

Attendance at Sixth Symposium on Training of Nuclear Facility persennel F. J. Remick has been invited to participate in the Sixth Symposium 15-18, 1985.

on Training of Nuclear Facility Pers The Comittee agreed Nuclear Society of which he is a member.

to support his attendance.

G.

International Atomic Energy Agency (IAEA) Request f_or Servic of Charles J. Wylie C. J. Wylie has been asked to serve as a member of an IAEA Working Group which will undertake the preparation of a supplement for users of IAEA Safety Guide SG-D7, "EmergencyT Power Systems at Nuclear Power Plants." working group of IAEA.

on this ad hoc participation Arrangements will be made for the IAEA to cover related expenses.

H. Conflict of Interest proposed sumary of briefly discussed a The Comittee financial conflict of interest limitations being prepared by OGC.

Several members took exception to the conservative C. P.

nature of the NRC/0GC interpretation of the rules.E. Plaine, NRC Siess suggested inviting Herzel H.

Counsel, to address the Procedures Subcomittee regarding legal precedent on the subject of conflict of interest and th application of NRC regulations to ACRS members.

)

l 30

T l'

. MlNUTES OF THE 298TH ACRS MEETfNG FEBRUARY 7-9, 1985 l

4 I. ACRS Effectiveness Study 4

} In a meeting of its Subcomittee on Comittee Activities at Harpers Ferry in November 1984, the members concluded that it would be useful to engage an outside panel of experts to make an ad hoc appraisal of the present and anticipated future

' effectiveness of the ACRS and to develop recomendations .for improving the Comittee's performance. Chairman D. A. Ward a indicated that a proposed task description has been briefly discussed with the proposed panel chairman, Manning Muntzing.

The Comittee endorsed proceeding with.the study.

a 1

J. Recualification of Reactor Operators During the Comittee's meeting with the Comissioners on.

February 7,1985, Comissioner J. Asselstine asked the ACRS to j

define what an acceptable reactor operator requalification i program is and to examine the policy of performance-oriented ,

examinations of requalification programs where NRC regional ,

i inspectors are examining the operators themselves (20 percent) i rather than reviewing the programs. He asked whether the ACRS thought that the " check-operator" approach was a way of

minimizing these kinds of problems.

f Comissioner J. Asselstine requested that the ACRS specifically consider the following: .

e Take a general look at the requalification program

' regarding whether it is working properly l

e Explore the requalification program's impact on i regionalization e Decide whether the objectives of the requalification j program are correct e Consider what the state of engineering' knowledge is on l

shift and how requalification programs can help to encourage a gradual increase in engineering knowledge on shift The 298th ACRS Meeting was adjourned at 3:25 p.m., Saturday, February 9, 1985.

i i

31

.z.-...---. . _ . . - . - - - , . - - . - , . . - , . .

o e

l APPENDIXES TO MINUTES OF THE 298TH ACRS MEETING FEBRUARY 7-9, 1985 b SS - D.'lk6 0

O l

u _ _ _ _ _ _ _ _ _ _ _

APPENDIX I LIST OF ATTENDEES I

l

~ ATTENDEES 298TH ACRS MEETING FEBRUARY 7-9, 1985 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS David A. Ward, Chairman

. Harold W. Lewis, Vice-Chairman Robert C. Axtmann Max W. Carbon Jesse C. Ebersole William Kerr Carson Mark Carlyle Michelson Dade W. Moeller David Okrent Glen A. Reed Forrest J. Remick Paul G. Shewmon I Chester P. Siess l

Charles J. Wylie ACRS Staff r)

\~./

Raymond F. Fraley, Executive Director M. Norman Schwartz, Technical Secretary Herman Alderman Paul A. Boehnert Anthony J. Cappucci Robert Cushman Sam Duraiswamy Medhat M. El-Zeftawy John Flack John T. Gilbert Elpidio G. Inga Janet Kotra Morton W. Libarkin Richard K. Major John A. MacEvoy Thomas G. McCreless John C. McKinley Owen S. Merrill Austin Newsom Gary R. Quittschreiber.

Richard Savio -

Stanley Schofer Shivaji Seth Alan D. Wang O

f e

i NRC STAFF ATTENDEES

>0 298TH ACRS MEETING Thrusday, February 7,1985

NUCLEAR REACTOR REGULATION NUCLEAR REGULATORY RESEARCH R. Hernan, PPAS D. L. Basdekas j cDo L "P "* ""

B. Requa, DL Jh J. Reyes F. Schroeder, DST W. S. Harelton, DE R. Klecker, DE OFFICE OF INSPECTION & ENFORCEMENT E. Throm, DSI P. N. Randall, RES R. F. Heishman H.W. Woods, DST G. Vissing, DL OFFICE OF EXECUTIVE LEEAL DIRECTOR N. Anderson K. Kniel, GIB E. Chan J. Stevens, DL A. Szukiewicz, GIB D. R. Nuller REGION III T. M. Novak B. J. Youngblood W. L. Forney O- J. P. Knight R. N. Gardner E. G. Greenman i

\

i 4

t

!O 1

l A-2.

( PUBLIC ATTENDEES 298TH ACRS MEETING l Thursday, February 7,1985 i R. E. Schaffstall, KMC, Inc.

, H. Fontecilla, Virginia Power i

J. Nurmi, EPM E. Desterle, Bechtel Power Corp.

R. Borsum, Babcock & Wilcox M. Ryan, McGraw Hill P. F. Riehm, KMC P. F. Collins, KMC E. T. Eilmann, DLC L. Toth, GASSER C. Hudgin, McGraw Hill M. D. Sollivan, Pacific Gas & Electric J. Reyner, EEC J. Gormly, Tudor Engr.

4 H. Renner, NUS Corp.

, M. D. Patterson, Baltimore Gas & Electric

J. Nurmi, EPM i M. Moaba, Florida Power & Light S. A. Collard, Florida Power & Light t

i l

T 4

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!O A-3

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

INVITED ATTENDEES 298TH ACRS MEETING Thursday, February 7, 1985 COMONWEALTH EDISON SARGENT & LUNDY C. W. Schroeder K. Green

'F. Willaford A. K. Singh J. M. Brennan B. G. Treece W. J. Shewski W. C. Cleff T. R. Tram K. Kostal L. O. DelGeorge W. B. Paschal J. C. Blomgren G. L. Sensmeier J. Golden N. N. Kaushal WESTINGHOUSE ELECTRIC CORPORATION B. R. Shelton M. Oper M. J. Wauau J. L. Tain G. T. Klopp D. Dominicis K. A. Ainger J. McInerney W. Shewski E. E. Fitzpatrick D. O'Brien udae ISHAM, LINCOLN & BEALE O C. McDonough E. E. Fitzpatrick R. Lauer C. W. Schrauder E. D. Swartz T. Mariman D. Knug D. Farrar O

A-9

NRC STAFF ATTENDEES 298TH ACRS MEETING Friday, February 8, 1985 K. Kniel, GIB R. Hernan, NRR L. Crocker, NRR P. Tremiday O

O A-5

J  !-

4 j PUBLIC ATTENDEES 4

r i 298TH ACRS MEETING

, c i

, a- i t

I Friday. February 8. 1985  !

i  :

j P. Docherty, Westinghouse Electric Corp. i j 0. Williams, JRA .

j M. Ryan, McGraw Hill  !

i R. Borsum, Babcock & Wilcox

[

. R. S. Boyd, KMC  !

H. M. Fontelilla Virginia Power I R. W. Huston, Delian i

i S. Watson, Bechtel j J. Nurmi, EPM  !

I L. Connor, DSA (

l A. P. Malinanskas, ORNL *

M. Simons  !

j A. Steele, Brattleboro Reformer  !

M. Fertel Delian I f

~

I j  !

i .

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4 4

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)

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j A-6  :

I APPENDIX !!

FUTURE AGENDA

} APPENDIX A FUTURE AGENDA 1

MARCH ACRS MEETING \

Nine Mile Point, Unit 2--OL review 5 hrs

Recent Events at Operating Reactors--Briefing 1 1/2 hrs regarding events and incidents at nuclear power  ;

- plants 1 v

! Vendor Inspection Program--Briefing by tentative representatives of IE NTSB-type Board for Nuclear Accidents--Report 2 hrs J of ACRS Subcommittee on the Brookhaven Report regarding the setting up of such a Board and the  :

L role of ACRS in nuclear accidents l

i Steam Generator Overfill--Proposed ACRS action I hr l ECCS Information status reports on deletion of 1 hr

upper head injection on several plants and '

research efforts to revise Appendix X O cessaa 11 st n re ai nt oesian arierin.--

Description and identification of safety issues 3 *r-Meeting with Director, AE00--Discuss AE00 1 hr activities of interest / concern Systematic Review of Nuclear Power Plants--

Proposed ACRS action i Chilled Water Systems--ACRS comments regarding deferred l l reliability of chilled water systems  !

! f Emergency Planning--Proposed rule on deferred consideration of seismic events in emergency planning r 1 .

l Maintenance Practices and Procedures-- deferred

< Subcouaittee report on Maintenance Task Action

Plan j Control Room Ventilation--Briefing regarding deferred control room ventilation tests at the Shearon j Harris Nuclear Plant f

i A-1  ;

l

l i

2 1

l 1

! i

! O a erts < acas s 6co-4*t -

Subcomittee on Electrical Systems regarding 1/4 hr ,

j the status of USI A-44 Station Blackout

' (WK/P9tE) '

i Subcomittee on Watts Bar Nuclear Plant 1/4 hr

! regarding discussion of major open items prior

! to fuel load (JCE/AJC) i i

l t .

f

+

1 j

!, 1

!O  !

i I

, I

} '

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A-8 I

.- -,_.___-_.._t .- _._ . . - - . . . . . _ _ . - . - . . - -

i COD O O F ACRS SUBCOMMITTEE HEETINGS SCHEDULE OF ACRS SUBCOMMITTEE MEETINGS l D

v FEBRUARY 13 Watts Bar (Knoxville, TN) (Cappucci) - Ebersole, Michelson, Ward.

Purpose:

To discuss major open items prior to fuel load and to update the Subcommittee concerning modifications to the Model D-3 steam generators, the fire protection program, and the resolution of construction and QA deficiencies.

14 & 15 Joint GESSAR II and Reliability & Probabilistic Assessment (Inglewood, CA) (Major /Savio) - Okrent, Ebersole, Etherington, 51ess, Ward, Wylie.

Purpose:

To continue the Subcomittee's review of GESSAR II for Final Design Approval applicable to future plants. The focus of these discussions will be on seismic risk.

15 & 16 Waste Management (Merrill) - Moeller, Axtmann, Carbon.

Purpose:

To review: (1) 10CFR60, " Disposal of High-Level Radioactive Waste in Geologic Repositories: Amendmeats to Licensing Procedures,"

and (2) various NRC Review Plans and Generic Technical Positions relevant to the NRC Waste Management Staff's efforts on Site Characterization. In addition, the manner in which the Subcomittee and Consultant Task Groups will function in support of the NRC Waste Management Staff will be discussed.

20 & 21 Nine Mile Point Unit 2 (Syracuse, NY) (Schiffgens) - Siess, Ebersole.

Purpose:

To begin review of the Niagara Mohawk Power Corporation's application for an OL for Nine Mile Point Unit 2.

21 ECCS

Purpose:

To7ev(Boehnert)-Ward,Etherington,Michelson, Reed.iew provis 10 CFR 50.46 and to discuss proposed Reg. Guide 1.82,

" Containment Emergency Sump Performance" (tentative).

25 Class 9 Accidents (Wang)-Kerr,Shewmon, Ward.

Purpose:

To discuss the status of the NRC's severe accident codes with the NRC Staff.

26 Electrical Systems fEl-Zeftawy/Savio) - Kerr, Ebersole Lewis, Mark,Okrent(tent.),Wylie.

Purpose:

To discuss plant experience with the loss of AC power and the status of NRC actions on USI A-44, " Station Blackout."

MARCH 1

Long Range Plan for NRC (Major) - Carbon, Lewis, Moeller, Remick, 51ess.

Purpose:

To plan Subcomittee activities.

POSTPONED Maintenance Practices & Procedures (Aldennan) - Michelson, O- Ebersole, Reed, Ward, Wylie.

Purpose:

To review Staff Maintenance Program Plan.

A 'l

Page 2 SCHEDULE OF ACRS SUBCOMMITTEE MEETINGS a

MARCH (CONT'D) 5 Reactor Operations (Major) - Ebersole, Kerr, Michelson, Moeller, Okrent, Reed, Remick, Ward, Wylie.

Purpose:

To discuss recent plant operating experience.

6 Regulatory Policies and Practices (Cappucci) - Lewis, Michelson, Moeller, Remick, Wylie.

Purpose:

To continue the review of the NRC report on the need for an "NTSB-like" board in the NRC.

7-9 299th ACRS Meeting 14 Class 9 Accidents (Wang) - Kerr, Moeller, Shewmon, Siess, Ward.

(1:00p-7:00p)

Purpose:

To discuss New York Power Authority's Source Term studies.

15 Joint ATWS and Electrical Systems (Boehnert/El-Zeftawy/Savio) -

Kerr, Ebersole, Michelson, Ward, Wylie.

Purpose:

To review NRC Staff activities associated with the implementation of the ATWS Rule and the status of NRC actions on scram breaker reliability.

19 Reliability Assurance (Major) - Michelson, Ebersole, Kerr, Okrent, Reed Ward, Wylie.

Purpose:

To review concerns arising O from a significant failure of an RCIC steam line isolation valve to open against operating reactor pressure.

20 Electrical Systems (El-Zeftawy/Savio) - Kerr, Ebersole, Lewis, Mark, Michelson, Wylie.

Purpose:

To discuss the status of recent NRC actions related to diesel generator reliability.

21 & 22 Combined Extreme External Phenomena, Structural Engineering, and Diablo Canyon (Los Angeles, CA) (Savio/Igne/McKinley) - Okrent.

Axtmann, Siess, Carbon, Etherington, Lewis, Michelson, Moeller.

Shewmon, Ward.

Purpose:

To discuss the status of the NRC Staff's seismic design margin programs and PG&E's plan for a seismic reevaluation of Diablo Canyon.

27, 28 & 29 Combined GESSAR II, Reliability and Probabilistic Assessment, and Safeguards and Security (Albuquerque, NM)

(Major /Savio/5chiffgens) - Okrent, Carbon, Ebersole, Etherington, Kerr, Lewis, Mark, Michelson, Reed, Ward Wylie.

Purpose:

To continue the Subcommittee's review of GESSAR II for Final Design Approval appitcable to future plants and review design features for protection against sabotage at comercial nuclear power reactors, explore the potential consequences of successful sabotage of nonpower reactors, and to hear how the NRC Staff reviews and evaluates licensees' security plans. The principal topics to be discussed are plant safeguards and the GESSAR II PRA.

A-to 1

Pag; 3 SCHEDULE OF ACRS SUBC0tNITTEE MEETINGS O -

APRIL

! 4 Human Factors (Major) - Ward, Lewis, Michelson, Moeller, Reed, Remick, Wylie.

Purpose:

To discuss NUREG/CR-3737, a method of .

ascertaining management / organization's contribution to safety of operating reactors.

9 Reactor Operations (Major) - Ebersole, Kerr, Michelson, Moeller, i (A.M.) Okrent, Reed, Remick, Ward, Wylie.

Purpose:

To discuss recent plant operating experience, l

i 9 Procedures & Administration (Fraley) - Ward, Remick, Siess.

(P.M.)

Purpose:

To discuss proposed revisions of ACRS Bylaws regarding l activities of members as individuals and possible limitations on ACRS activities to accomodate budgetary and manpower resources cuts.

10(Closed) Safety Research Program (Duraiswamy) - Siess, Carbon, Kerr, Mark, Michelson, Moeller, Remick, Ward, Wylie.

Purpose:

To discuss a draft report on the "NRC Safety Research Program" prepared by RES dealing with the justifications for a base NRC Safety Research Program in the future.

10 Safety Philosophy, Technology, and Criteria (Savio) - Okrent, (3:00p.m.) Ebersole, Kerr, Lewis, Michelson, Remick, Ward, Wylie.

Purpose:

To review the status of the NRC Staff's evaluation of the trial use of the Comission's proposed Safety Goal Policy.

11 - 13 300th ACRS Heeting 17 Joint Reactor Operations and Human Factors (Atlanta, GA) l (Closed) (Major) - Ebersole, Ward, Kerr, Lewis, Michelson, Moeller, Okrent, Reed, Remick, Wylie.

Purpose:

To discuss INP0's eval-uation of nuclear plant operations and incidents / accidents --

briefing by INPO representatives.

MAY 7 Safeguards & Security (Schiffgens) - Mark, Carbon, Ebersole, Michelson, Reed, Siess, Wylie.

Purpose:

To review the potential consequences of sabotage at non-power reactors and to be briefed by NMSS on sabotage protection at power reactors.

8 Safety Research Program (Duraiswag) - Siess, Carbon, Kerr, Mark, Michelson, Moeller, Ukrent, Remick, Shewmon, Ward, Wylie.  !

Purpose:

To discuss the proposed NRC Safety Research Program and budget for FY 1987 and to gather information for use by the ACRS in its preparation of the annual report to the Comission on the NRC Safety Research Program.

O 9 - 11 301st ACRS Meeting A-l\

! Page 4 i l

i SCHEDULE OF ACRS SUBC0PMITTEE MEETINGS j

l j JUNE

^

1 j 4 RegulatoryActivities(Duraiswamy)-Siess, Carbon,(Kerr, Michelson, Ward, Wy11e.

Purpose:

To review the: 1) proposed il t

(tent.) General Revisions to Appendix J to 10 CFR 50, " Leak Tests for l Primary and Secondary Containments of Light-Water Cooled Nuclear Power Plants, and (2) Draft Regulatory Guide on " Containment  !

Leakage Testing."

i 5 SafetyResearchProgram(Duratswamy)-Siess, Carbon,Kerr, Mark,  ;

Michelson, Moeller, otrent Remick, Shewmon, Ward, Wylie.  !

j

Purpose:

To discuss the updated information (possibly the Budget  !

j Review Group mark) on the proposed NRC Safety Research program f j budget for FY 1987. Also to discuss a draft ACRS report to the t

Comission on the NRC Safety Research Program and budget for FY I 1987. ,

e 6-8 302nd ACRS Meeting f f

} '

DATES TO BE i NTERRTIVEU~

f AirSystems(Schiffgens)-Moeller, Mark,Michelson, Ward. i j

(late Mar.)

lQ j

Purpose:

To review the NRC Staff's Supplement to the Control Room Habitability Working Group Report. This Supplement is to

[

4 discuss the Staff's survey of NT0L and OR control rooms.  !

(Mar.) (ualification Program for Safety-Related Equipment (Cappucci) -  !

l 1 (tent.) Flichelson, Ebersole, Reed, 5hewmon, 51ess Ward, Wylie.

Purpose:

To discuss the NRC Staff's resolution of U5! A-46, " Seismic  !

Qualification of Equipment in Operating Plants." Also to discuss  ;

valve operability with the NRC Staff and industry.  !

(April) Joint Metal Comoonents and Seismic Design of Pioine (Igne) - l (tent.) 5hewmon, Axtmann, Etherington, Lewis, Mark, M1cneison. 0krent.

4 Siess, Ward.

Purpose:

To review the NRC Piping Review Commit- I tee's overall recommendations on piping system concerns. j (April) SeismicDesignofPiping(Igne)-Siess,Etherington, Mark. (

(tent.) Okrent 5hemon.

Purpose:

To review draft esports issued by the  ;

! NRC Piping Review Cosmittee on dynamic loads and load combina-i tions and seismic design requirements of piping. l i

(early spring) ECC$ (Palo Alto. CA) (Boehnert) - Ward, Ebersole, Etherington, f Michelson, Reed.

Purpose:

To continue the review of the joint -

l I NRC/B&W0G/EPRI/B&W joint IST Propram. A visit to the EPR!

5tanford Research Institute faci ities supporting this Program is also. planned. i l i l

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Paga 5 j

l l r I, SCHEDULE OF ACRS SUBC0P911TTEE MEETINGS jO i DATES TO BE

! MTERRflRT(CONT'D) l Date To Be Palo Verde (Maricopa County. AZ) (Wang) - Ebersole, Kerr, Lewis.

Determined Wylie.

Purpose:

To review the final reports for various con-l struction deficiencies and the results of the preoperational j

testing as requested in ACRS letter dated December 15, 1981.  ;

1 1 Date To Be Human Factors Tour (Russellville, AR) (Major) Ward, Lewis, j Determined Michelson, Moeller, Reed, Remick, Wylie.

Purpose:

This will 1

(Closed) be a tour and examination of ANO-l's emergency procedures (symptom based) and facilities.

Date To Be Westinghouse Water Reactors (Cappucci) - Ebersole Etherington, Determined Michelson, Okrent, 51ess. Wylie.

Purpose:

To begin the PDA (Closed) review of the Westinghouse Advanced PWR (RESAR SP/90).

f Date To Be Joint Reliability & Probabilistic Assessment and Millstone 3 Determined (location to be determined) (5avf o/Duraf swamy) - Ukrent, Kerr, Ebersole, Lewis, Mark, Michelson, Siess, Ward, Wylie.

Purpose:

To review the probabilistic risk assessment for Millstone 3.

3 Date To Be RiverBend1&2(locationtobedetermined)(Savio)-Okrent, j Determined Ebersole, 5hewmon.

Purpose:

To continue the review of Gulf i

States Utilities' application for an operating license for the River Bend Nuclear Power Plant Units 1 & 2.

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SCHEDULE OF ACRS SU8 COMMITTEE MEETING O SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS DATE FEBRUARY 13, 1985 WATTS BAR (CAPPUCCI ) Ebersole, Michelson, Ward l 1

Cons.: Patton, Epler

! LOCATION: Knoxville TN BACKGROUND:

Who proposed action: J. Ebersole 4

Purpose:

To discuss major open items prior to fuel load and to update the Subcommittee concerning modifications to the Model D-3 steam 9enerators, the fire protection program, and the resolution of significant contruction and QA deficiencies.

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L PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

Project Status Report issued 1/28/85.

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4 SCHEDULE OF ACRS SUBCOMMITTEE MEETING O SUBCOMMITTEE MEETING STAFF ENGR.' & MEMBERS DE (MAJOR /SAVIO) Okrent; Ebersole, FEBRUARY 14 & 15 JOINT GESSAR II AND RELIABILITY & Etherington. Siess. Ward,  ;

PROBABILISTIC ASSESSMENT Wylie Cons.: Bohn. Camp l LOCATION: Inglewood, CA i BACKGROUND:

Who proposed action: Okrent

Purpose:

To continue the Subcommittee review of the GESSAR II design. The principal I topic to be discussed will be seismic risk.

6 J

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SCHEDULE OF ACd5 SU8 COMMITTEE MEETING O

SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS DATE WASTE MANAGEMENT (MERRILL) Moeller, Axtmann, FEBRUARY 15 & 16 Carbon ,,

Cons.: Carter. Donoghue, Foster, Krauskopf, Parker, Steindler, G. Thompson LOCATION: WASHINGTON, DC SACKGROUND:

Who proposed action: ACRS' I

Purpose:

To review: (1) 10CFR60, " Disposal of High-Level Radioactive Waste in Geologic Repositories: Anendments to Licensing Procedures.

I January 10,1985, and (2) various NRC Review Plans and Generic 1 Technical Positions relevant to the NRC Waste Management Staff's ,

efforts on Site Characterization. In addition, the menner in which the Subcomittee and Consultant Task Groups will function in support of the NRC Waste Management Staff will also be discussed.

I 1

! PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. Proposed Sunmary and Minutes of the meeting of the ACRS Subcomittee on Weste Management, Washington, DC, January 17 a 18,1985.

2. NRC Review Plan: Quality Assurance Programs for Site Characterization of High-4 Level Nuclear Waste Repositories (Draft) June 1984.

i

3. 10CFR60, " Disposal of High-level Radioactive Waste in Geologic Repositories: -

Amendments to Licensing Procedures," January 10, 1985.

4. Final Technical Position on Documentation of Computer Codes for High-Level Weste Management (NUREG-0856), June 1983.

5. Draf t Technical Position, Subtask 1.1: Weste Package Perfomance After Repost-tory closure (NUREG/CR-3219. Vol.1), August 1983.

6. Post-Emplacement Monitoring (NUREG/CR-3219. Vol. 2) (Draft) May 1983.

7. Draft copy of Guidelines for Waste Management Subcommittee Task Groups, ,

O r 6r# rx 7 i'85-b- u, e .. , _ _ . , .m . _ . _ . _ , _ _ . . , , . - . - . . . - , , , . . - - .

SCHEDULE OF ACRS SUBCOMMITTEE MEETING i

SUBCOMMITTEE MEETING STAFF ENGR. 8 MENSERS

DATE NINE MILE POINT, UNIT 2 (SCHIFFGENS) Siess Ebersole, FEBRUARY 20 & 21, 1985 l

' I i

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i LOCATION: Syracuse, NY l

i SACKGROUkO:

t Who proposed action: Applicant 3 I

Purpose:

To begin review of OL for Nine Mile Point, Unit 2.  :

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PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:  :

To be provided later. (SERdue2/8/85) f, .

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING SUBCOMMITTEE MEETING STAFF ENGP. & MEMBERS DATE ECCS (BOEHNERT) Ward,Etherington, FEBRUARY 21, 1985 Michelson, Reed i

Cons.: Catton, Schrock, Sullivan, Theofanous.

Tien (tent.)

WASHINGTON, DC J.0 CAT 10N:

I i BACKGROUND:

I Who proposed action: RES

Purpose:

(1) Review provisions of proposed Rule to revise Appendix K of

! 10CFR50.46 and (2) Discuss proposed Reg. Guide 1.82, " Containment

.' Emergency Stsnp Perfonnance" (tentative) i PERTINENT PUBLICATIONS AND THElR AVAILABILITY:

j 1. SECY paper will be provided when available

2. Revision 1, Reg. Guide 1.82 (will be provided when available)

(

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SCHEDULE OF ACRS SU8 COMMITTEE MEETING SU8 COMMITTEE MEETING _

STAFF ENGR. & MEM8ERS DATE Kerr i FEBRUARY 25, 1985 CLASS 9 ACCIDENTS (WANG)Eard,Shewmon. I i

l LOCATION: WASHINGTON, DC 4

l BACKGROUND:

Who proposed action: ACRS 1

To discuss with the NRC Staff the status of the NRC's severe accident codes.

Purpose:

1

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l PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. ORNL/TM 8842, " Review of the Status of Validation of the Computer Codes Used

! in the NRC Severe Accident Source Term Reassessment Study (BMI-2104)."

1 i

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING O DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS FEBRUARY 26, 1985 ELECTRICAL SYSTEMS (EL-ZEFTAWY/SAVIO) Kerr, Ebersole, Lewis, Mark.

Okr'ent(tent.),Wylie I

LOCATION: Washington, DC i

BACKGROUND:

Who proposed action: ACRS

Purpose:

discuss the recent plant experience with loss of AC power and Tget status of NRC actions on USI A-44, " Station Blackout."

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l PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1 To be supplied.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING

.O DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS _

LONG RANGE PLAN FOR THE NRC (MAJOR) Carbon, Lewis, MARCH 1,1985 Moeller, Remick. Siess ,

LOCATION: WASHINGTON, DC BACKGROUNO:

I Who proposed action: Dr. Carbon

Purpose:

To plan Subcomittee activities.

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! PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

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l' SCHEDULE OF ACRS SUBCOMMITTEE MEETING STAFF ENGR. & M ERS DATE SUBCOMMITTEE MEETING (ALDERMAN)M elson_, Ebersole, MARCH 5, 1985 MAINTENANCE PRACTICES AND PROCEDURES Reed, Ward Wylie WASHINGTON, DC h LOCATION:

4 BACKGROUNO:

Who proposed action: C. MICHELSON 1

Purpose:

To review NRC Staff Maintenanj P gr Plan.

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PERTINENT P LICATIONS AND THEIR AVAILABILITY:

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l SCHEDULE OF ACRS SUBCOMMITTEE MEETING O STAFF ENGR. & MEMBERS DATE SUBCOMMITTEE MEETING MARCH 5, 1985 REACTOR OPERATIONS (MAJOR Ebersole, Kerr, Michel) son, Moeller, Okrent.

Reed, Remick, Ward, Wylie .

LOCATION: WASHINGTON, DC BACKGROUND:

Who proposed action: Staff /ACRS

Purpose:

To discuss recent plant operating experience. The office of Inspection and Enforcement and Nuclear Reactor Regulation will identify a number of significant recent operating events and present these events to the Subcomittee for consideration. The Subcomittee will also identify recent events for discussion. A subset of the events presented during

(^ the Subcomittee meeting will be selected for presentation to the full ACRS at the next general meeting.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

A Status Report will be provided prior to the meeting.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING O DATE SUBCOMMITTEE MEETING STAFF E LR. & MEMBERS (CAPPUCCI) Lewis, Michelson, MARCH 6' 198r* REGULATORY POLICIES & PRACTICES Moeller, Remick, Wylie LOCATION: WASHINGTON, DC BACKGROUND:

Who proposed action: H. Lewis l

Purpose:

To continue the review of MRC report on the need for an "NTSB-like" Board in the NRC.

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PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. Final Draft of BNL "NTSB-Report"

2. December 26, 1984 NRC Staff coments on (1) above.

3. December 27, 1984 comments on (1) above from System Safety, Inc. (C. O. Miller)

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING O DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS CLASS 9 ACCIDENTS (WANG) Kerr, Moeller, MARCH 14,1985 Shewmon Siess, Ward (1:00p-7:00p)

LOCATION: WASHINGTON, DC BACKGROUND:

Who proposed action: New York Power Authority (NYPA)

Purpose:

NYPA proposes to elaborate on their source tem studies presented in a previous subcommittee meeting.

O PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. " Source Term Safety Assessment - Indian Point 3," NYPA.

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4 SCHEDULE OF ACRS SUBCOMMITTEE MEETING O DATE SUBCOMMITTEE MEETING STAFF ENGR. 8 MEM8ERS l

MARCH 15,1985 JOINT ATWS and ELECTRICAL (BOEHNERT/EL-ZEFTAWY/SAVIO)

SYSTEMS Kerr, Ebersole, Michelson,  !

Ward, Wylie Cons.

Davis, Lee, Lipinski f LOCATION: WASHINGTON, DC l

BACKGROUND:

Who proposed action: ACRS i

f

Purpose:

To review NRC Staff activities associated with implementation of the ATWS Rule and the status of NRC actions on scram breaker reliability.

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PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

4

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SCHEDULE OF ACRS SUBCOMMITU.E MEETING O DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS MARCH 19,'1985 (MAJOR) Michelson, Ebersole, RELIABILITY ASSURANCE Kerr, Okrent, Reed, Ward, Wylie LOCATION: WASHINGTON, DC BACKGROUND!

Who proposed action: Michelson

Purpose:

To review concerns arising from a significant failure of an RCIC steam line isolation valve to open against operating reactor pressure. Other concerns related to valve operability may be discussed, i.e., the August 25, 1982 event at Hatch 2 when a

' long uncontrolled blowdown occurred outside containment because of a scram discharge volume drain valve failure.

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1 PERTINEhT PUBLICATIONS AND THEIR AVAILABILITY:

1. Memo for J. Ebersole from C. Michelson,

Subject:

Significant Failure of an RCIC Steam Line Isolation Valve to Open Against Operating Reactor Pressure, 11/14/84

2. Memo for J. Ebersole f rom R. Fraley,

Subject:

Significant Failure of an RCIC Steam Line Isolation Valve to Open Against Operating Reactor Pressure, 11/21/84

3. Item Prioritizatioh: Issue 87: Failure of HPCI Steam Line Without Isolation (Draft) e A-r1 .

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

t i

SCHEDULE OF ACRS SUBCOMMITTEE MEETING O SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS DATE MARCH 20, 1985 ELECTRICAL SYSTEMS (EL-ZEFTAWY/SAVIO) Kerr, Ebersole, Lewis, Mark.,

Michelson, Wylie i

LOCATION: WASHINGTON, DC i

f BACKGROUND:

Who proposed action:

1

Purpose:

To discuss the status of recent NRC actions related to diesel generator reliability.

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PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

i f

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING SUBCOMMITTEE MEETING STAFF ENGR. 4 MEMBERS DATE MARCH 21 & 22. 1985 COMBINED EXTREME EXTERNAL PHENOMNA/ (SAVIO/IGNE) Okrent. Siess.

STRUCTURAL ENGINEERING /DIABLO CANTON Carbon. Axtmann. Etherington.

Lewis, Michelson, Moeller. ,

Shewmon. Ward LOCATION: LOS ANGELES. CA BACKGROUND:

Who proposed action: Okrent/Siess

Purpose:

To discuss the status of the NRC Staff seismic design margins programs, and PG&E's plan for a seismic reevaluation of Diablo Canyon.

O PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

To be provided later.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING O DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS _

MARCH 27, 28, 29, 1985 COMBINED GESSAR II, RELIAB:LITY (Major /Savio)Okrent.Ebersole, 8

& PROBABILISTIC ASSESSMENT, AND Etherington. Kerr. Lewis.

SAFEGUARDS & SECURITY Mark. Michelson. Carbon, Reed, Ward. Wylie Cons.: Camp. Catton, First LOCATION: SANDIA (ALBUQUERQUE, NM)

BACKGROUND:

Who proposed action: D. Okrent

Purpose:

To continue the Subcomittee's review of GESSAR II for Final De >ign Approval applicable to future plants, review design features for protection against sabotage at commercial nuclear power reactors, explore the potential consequences of successful sabotage of nonpower reactors, and to hear how the NRC Staff reviews and evaluates licensees' security plans. The principal topics to be discussed are plant safe-O guards and the GESSAR II PRA.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. The SER through Supp. 3 is available. ,

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O A-30

SCHEDULE OF ACRS SUFCOMMITTEE MEETING SUBCOMMITTEE MEETING STAFF ENGR. & MEM8ERS f DATE I APRIL 4, 1985 HUMAN FACTORS (MAJOR) Ward, Lewis,Michelson, Moeller, Reed, Remick, Wylie l 1

I i

I j LOCATION: WASHINGTON, DC BACKGROUND:

f Who proposed action: ACRS A detailed discussion of NUREG/CR-3737, 'a method of ascertaining 1

Purpose:

management / organization's contribution to safety of operating reactors. This effort is attempting to identify and collect existing objective measures of plant safety perfomance and management and organization. l ACRS Fellow, John MacEvoy, has also been attempting to identify 7

4 safety and performance indicators at operating plants. A dis-

cussion of John's work is planned.

i j

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. NUREG/CR-3737, "An Initial Empirical Analysis of Nuclear Power Plant Organization and Its Effects on Safety Perfomance." (currently available) j

2. Memo for J. Ebersole fm J. MacEvoy,

Subject:

" Report on Nuclear Power Plant Safety Measurements " dtd November 7,1984.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING i

O DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS APRIL 9,1985(A.M.) REACTOR OPERATIONS (MAJOR)Ebersole,Kerr, Michelson, Moeller, Okrent,

' Reed, Remick, Ward, Wylie I

1 LOCATION: WASHINGTON, DC BACKGROUND:

Who proposed action: Staff /ACRS

Purpose:

To discuss recent plant operating experience. The office of Inspection and Enforcement and Nuclear Reactor Regulation will identify a number  ;

of significant recent operating events and present these events to the Subcommittee for consideration. The Subcommittee will also identify ,

recent events for discussion. A subset of the events presented during the Subcommittee meeting will be selected for presentation to the full

, O ACRS at the next general meeting.

J l

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! PERTINENT PUBLICATIONS AND THE!R AVAILABILITY:

A Status Report will be provided prior to the meeting.

1 S

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A-32.

SCHEDULE OF ACRS SUBCOMMITTEE MEETING O DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS _

APRIL 9, 1985 PROCEDURES & ADMINISTRATION (FRALEY) Ward,Remick,Siess (1:00P.M.)

LOCATION: WASHINGTON, DC BACXGROUND:

Who proposed action:

Purpose:

To discuss proposed revisions of ACRS Bylaws regarding activities of members as individuals and possible Ifmitations on ACRS activities to accomodate budgetary and manpower resources cuts.

O PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

O A-33

SCHEDULE OF ACRS SUBCOMMITTEE MEETING

.O DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS SAFETY RESEARCH PROGRAM (DURAISWANY) fess. Carbon.

APRIL 10. 1985 e er. Michelson.

i Kerr Mark,d.

Remick, War. Wylie (CLOSED)

LOCATION: WASHINGTON,DC i

BACKGROUND:

Who proposed action: C. SIESS 1

Purpose:

To discuss ths "NRC Safety Research Program" Report prepared by RES which provides justification for a base NRC Safety Research in the future.

I j

4 PERTINENT PUBLICATIONS AND THEIR AVAILABILITY: 21,1984

1. Draft 1 of the NRC Safety Research Program report, dated November (was sent Draft to all members on 11/28/84) 2oftheNRCSafetyResearchProgramreport(willbesenttomembers 2.

assoonaspossible) 1

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING l

O DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS APRIL 10, 1985 SAFETY PHILOSOPHY, TECHNOLOGY, AND (SAVIO)Okrent.Ebersole.

Kerr. Lewis, Michelson, Remick.

(3:00 p.m.) CRITERIA Ward. Wylie ,l 1

LOCATION: WASHINGTON, DC BACKGROUND:

i Who proposed action: NRC Staff /Subcomittee

Purpose:

To review the status of the NRC Staff's evaluation of the trial use of the Commission's proposed Safety Goal Policy. '

i

O PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. NRC Task Force report on the two-year evaluation of the trial use of the Comission's proposed Safety Goal Policy (expected in early February).

2. Swimary of tentative NRC Task Force recomendations on revisions to and future use of the NRC Safety Goal Policy.

I A-35

SCHEDULE OF ACR$ SU8 COMMITTEE MEETING O

• SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS DATE (MAJOR)Ebersole, Ward, APRIL 17, 1985 JOINT REACTOR OPERATIONS Kerr, Lewis, Michelson, (CLOSED) AND HUMAN FACTORS Moeller, Okrent, Reed, Remick,Wylie LOCATION:

INPO Offices,1100 Circle 75 Pkwy, Suite 1500. Atlanta, GA (404)953-3600 INPO contact: Terry Sullivan BACKGROUND:

Who proposed action: ACRS/INPO

Purpose:

To discuss INPO evaluation of nuclear plant operations and incidents / accidents -- briefing by INPO representatives. This briefing was endorsed by the ACRS.

This meeting is expected to be closed due to the privileged and proprietary nature of the information to be discussed.

Other Topics as Devised

- assistance activities

- training and accreditation

- evaluations (simulator, SRO Peer evaluations, corporate, etc.)

- INPO relationship with NRC-NUMARC

- NPROS

- Human performance evaluation

- INPO international participation program PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

Available background will be sent in a status report.

A-36

t SCHEDULE OF ACRS SUBCOMMITTEE MEETING O SUBCOMMITTEE MEETING STAFF ENGR. 4 MEMBERS DATE MAY 7, 1985 3AFEGUARDS & SECURITY (SCHIFFGENS) Mark, Carbon, Ebersole Michelson, Reed, ,

Siess, Wylie tl LOCATION: WASHINGTON, DC SACKGROUND:

]

Who proposed action: Okrent, Michelson, Mark i

Purpose:

To review the potential consequences of sabotage at non-power reactors and to be briefed by NMSS on sabotage protection at power reactors.  ;

O PERTINENT PUBLICATIONS AND THE!R AVAILABILITY:

None 4

O A-M x - _. _ _ - _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _

SCHEDULE OF ACRS SU8 COMMITTEE MEETING 1

)

,p V DATE SUBCOMMITTEE MEETING STAFF ENGR. 4 MEMBERS SAFETY RESEARCH PROGRAM (DURAISWAMY) Stess. Carbon.

MAY 8, 1965 Kerr. Mark. Michelson. Moeller.

Okrent. Remick. Shewmon. Ward.

Wylie

/

! I l

LOC ATION

WASHINGTON,DC 1 8ACKGROUh0:

Who proposed action: Routine process

Purpose:

To discuss the proposed NRC Safety Research Program and budget for FY 1987 and gather infomation for use by the ACRS in its preparation of the annual report to the Consnission on the NRC Safety Research Program and budget. .,

O i

PERTINENT PUBLICATIONS AND THE!R AVAILABILITY:

1. Proposed NRC Safety Research and budget for FY 1987 (expected to be made availabletotheACRSduringApril1985).

1

\

'O A-36 69

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

SCHE 0ulE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENYt. 4 MEMSERS JUNE 4, 1985 REGULATORY ACTIVITIES (DURA 15WAMY)Siess. Carbon.

Kerr Michelson. Ward, Wylie (tentative) i 1

LOCATION: WASHINGTON, DC BACKGROUND:

1 i

Who proposed action:

Purpose:

To review the following: l l  !

1) Proposed General revisions to Appendix J to 10 CFR 50, " Leak Tests l

for Primary and Secondary Containments of Light-Water Cooled l

i Nuclear Power Plants." and

2) Draft Regulatory Guide on " Containment Leakage Testing."

l I

l PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

The following are expected to be made available to the ACR$ during May 1985:  !

1. Proposed revisions to Appendix J to 10 CFR 50

2. Draf t Regulatory guide on " Containment Leakage Testing."

1 1

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. A- M "

i

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1 SCHEDULE OF ACRS SIJ8 COMMITTEE MEETING SU8 COMMITTEE MEETING STAFF ENGR. 4 MEMBERS DATE SAFETY RESEARCH PROGRAM (DURA 15WAMY)51ess, Carbon.

JUNE 5. 1985 Kerr, Mark, Michelson. Moeller.

Okrent. Remick, 5heimon.

Ward,Wylie LOCATION: WASHINGTON, DC SACKGR00hu:_

Who proposed action: Routine process

Purpose:

To discuss the updated infonnation (possibly the Budget Review Group mark) on the proposed NRC Safety Research program budget for FY 1987.

Also to discuss a draft ACRS report to the Consnission on the NRC Safety Research Program and budget for FY 1987.

O l i

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i PERTINENT PUBLICATIONS AND THEIR AVAILABillTY:

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. A-Ho

1 SCHEDULE OF ACR5 $USC(MMITTEE MEETING O STAFF E W . 4 ME E RI DAfg SUSCOMMITTEE MEETING AIR SYSTEMS H FGENS 11er. M e "O BE DETEm!NED ,

(late Her.) <

i l .

1 ,

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LOCATION: WASHINGTON. DC

! BACKGROUND: .

Who proposed action

ACRS (

Purpose To review the NRC Staff's Supplement to the Control Room Habitability i

Working Group Report - June 1984. This Supplement is to discuss the Staff's Survey of NTOL and OR control rooms.

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i PERTINENT PUBLICATIONS AND THEIR AVAILA81LITY:

To be supplied. l 1

i O

l l

)

A-46 1

! SCHE 00LE OF ACR$ 500 COMMITTEE MEETING i O STAFF ENGR. 8 MEMSER5 DATE 500 COMMITTEE MEETING TO BE DETERMINED QUALIFICATION PROGRAtt FOR (CAPPUCC!/QUITTSCHRE!BER)

SAFETY-RELATED EQUIPMENT Michelson. Ebersole. Reed.

(MARCH) 5hewmon. Siess. Ward. Wylie (tentative)

Cons.: to be detemined i

LOCATION: WASHINGTON DC l

i i

SACKGROUND:

Who proposed action: NRC Staff

Purpose:

To discuss the NRC Staff resolution of U5! A 46 " Seismic Qualification of Equipment in Operating Plants." Also to discuss valve operability with the NRC Staff and industry.

O i

l PERTlhEht PullLICATIONS AND THEIR AVAILABILITY:

To be supplied.

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• O i

A-42. ,

SCHEDULE OF ACRS SUBCOMMITTEE MEETING 4

SUBCOMMITTEE MEETING STAFF ENGR. 4 MEMBERS DATE JOINT METAL COMPONENTS (IGNE)Shewmon.Axtmann,

! TO 8F DETERMINED Etherington, Lewis, Michelson, (April) AND SE!SMIC DESIGN OF PIPING Okrent. Ward. 5f ess, Mark ,

(tentative)

Cons.: Sender. Pfckel.

Rodabaugh. B. Thompson, Dillon, Kassner. Bush LOCATION: WASHINGTON, DC BACKGROUND:

l Who proposed action: NRC Staff I

Purpose:

To review the NRC Piping Review Comittee's overall recomendations on piping system concerns.

JO i

l l

PERTIhthT PUBLICA110NS AND THEIR AVAILABILITYt_

1. Report of the NRC Piping Review Comittee on recomendations on piping.

NUREG 1061 Vo. V l

2.

NUREG 03'3, Rev d.(available in early April 1985). implementation repo March).

i O .

A-% . . . .

At

SCHEDULE OF ACRS SU8 COMMITTEE MEETING

O OATE SU8 COMMITTEE MEETING STAFF ENGR. 4 MEMBER 5 (IGNE) i . Etherington.

l TO BE DETERMINED SEISMIC DESIGN OF P! PING

(April) Mark. 0 ren . Shewmon i (tentative)

Cons.: Bender. Pickel . Bush LOCATION: WASHINGTON DC BACKGROUND:_

Who proposed action: NRC/Siess

Purpose:

To review draft reports issued by the NRC Piping Review Committee on dynamic loads and load combinations and seismic design requirements .

of piping.

1 O

1 PERTINENT PUSLICAfl0NS AND THElR AVAILAllLITY:

l

1. NUREG 1061 Vol. IV " Evaluation fOther Dynamic Loads and Load Combination."

(report published and distributed $

2. NUREG 1061. Vol !!. "Neview of Sei mic Design Requirements for Nuclear Power Plant Piping." (should be available by March).

1 I e A-H

SCHEDULE OF ACR$ 508COMMITitt METING O g susCOMMifftt mitTING 5fAFF ENGR. 4 MMstal 70 BE DETEMINED ECC5 (90EHNERT) Ward.Ebersole.

Etherington Michelson, Reed i (earlySpring)

Cons.: Catton, sullivan.

Schrock. Theofanous.

Tien LOCATION _ PALO ALTO. CA SACNGROUhor Who proposed action: EPRl/NRC

Purpose:

To continue the review of the joint NRC/94W0G/EPRl/84W joint IST Program. A visit is planned to the EPRI Stanford Research Institute facilities supporting this Program.

O PERT!hthi PUBLICAfl0NS AND THilR AVAILABILITY:

To be provided.

I A-45 L. ,

l  :

i

> SCHEDULE OF ACR5 $USC(MMITTEE MEETIM

- SURCOMMITTEE MEETI N STAFF ENGR. 4 MDISER5 DATE PALO VERDE (WANG) Ebersole. Narr.

T0 SE DETEMINED Lewis, Wylie -

1

.i LOCATION: Maricopa,AZ l

SACKGROUND: (

l Who proposed action: NRC Staff /ACR$

Purpose To review the final reports for various construct e deficiencies and -

j the results of the preoperational testing as requested in ACR$ letter dated December 15, 1981.

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{ r PtRTINENT PutLICAfl0NS AND THElR AVAILA81LITY: c To be supplied.

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SCHEDULE OF ACRS SU8 COMMITTEE TOUR SUSCOMMITIEE TOUR STAFF ENGR. 4 MEM8ERS DATE HUMAN FACTORS (MAJOR) Ward, Lewis, Michelson, TO BE DETERMINED Moeller.Ed. Remick , Wylie (CLOSED) l l

LOCATION: ANO-1, Russellville, AR (-50 miles outside of Little Rock AR)

I SACKGROUND:

j Who proposed action: Human Factors Subcomittee

Purpose:

This will be a tour and examination of ANO l's emergency, procedures (symptombased)andfacilities. The Subcomittee wants the opportunity to examine procedures at an operating plant and see how the Ttil required backfits such as SPOS interface. Up to a day f

1 and a half is expected. ANO-1 is an 850 MWe, B&W PWR.

!o 4

) '

PERTINENT PUBLICAfl0NS AND THEIR AVAILABILITY:

1. One copy of Arkansas fluclear One, Unit 1 Emergency Operating Procedures l '

is available for your inspection at the ACRS Office (ask ftr. R. Major i

forthem).

l .

O 2

{

A-m

SCHEDULE OF ACRS SUSCOMMITTEE M ETING STAFF ENGR. & M MER$_

SU8 COMMITTEE MEETING _

DATE WESTINGHOUSE WATER (Cappucci/Quittschreiber)

TO BE DETERMINED Ebersole. Etherington. .

REACTORS ,l (CLOSED)

Michelson, Okrent. Siess.

Wylie 4

i LOCATION: Washington, DC BACKGROUND:

1 Who proposed action:

Purpose:

To begin the PDA review of the Westinghouse Advanced PWR (RESAR SP/90).

!O i

4 I

3 i

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

(To be provided later.)

j j

O .

A-4B 0

Q

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING ,

0 DATE SUBCOMMITTEE MEETING STAFF ENGR. A MEMBERS JOINT RELIABILITY PROBABILISTIC (SAVIO/DURAISWAMY)Okrent. l TO BE DETERMINED Kerr, Ebersole Lewis. .

ASSESSMENT AND MILLSTONE 3 Mark,Michelson,Stess, l i

Ward,Wylie t

LOCATION: (to be detemined)

I BACKGROUND:

Who proposed action: Okrent/Kerr i

Purpose:

To review the Millstone 3 PRA.

4 i

O t

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY: ,

j

1. NRC report documenting the results of the NRC/LLNL review of the Millstone 3 PRA.

O 1 l

. A-%9 '

l '

SCHEDULE OF ACRS SUBCOMMITTEE MEETING O DATE SUBCOMr.: TTEE MEETING STAFF ENGR. & MEM8ERS

(SAVIO)Okrent,Ebersole.

1 TO BE DETERMINED RIVER BEND 3 Shewmon i

j LOCATION: (to be detennined)

SACKGROUND:_

i Who proposed action: D. Okrent

Purpose:

To continue the review of Gulf States Utilities'capplication for i

an operating license.

l lO t 1

1 l

PERTIf4ENT PUBLICATIONS AND THEIR AVAILABILITY:

1. NRC Staff to supply SER Supplement 2 in early March to support the April 10,1985 Subcommittee meeting and action at the April 11-13. 1985 full Comittee meeting.

l 1

1 iO p. m88 e

A-50 . . . .

l APPENDIX IV NRC STAFF PRESENTATION ON BRAIDWOOD STATION, UNITS 1 AND 2 O

NRR STAFF PRESENTATION TO THE ACRS FULLCOMMITTEE FOR.

BRAIDWOOD STATION, UNITS 1 AND 2 FEBRUARY 7, 1985 PRESENTED BY O JANICE A. STEVENS LICENSING PROJECT MANAGER O

A-Si

o DISCUSSION TOPICS o OVERVIEW 0F LICENSING ACTIVITIES' o DUPLICATE PLANT CONCEPT o STATUS OF UNRESOLVED ITEMS -

O O

A-5L

O

OVERVIEW 0F LICENSING ACTIVITIES 2

o CONSTRUCTION PERMIT ISSUED 12/31'/75 o APPLICATION FOR OPERATING LICENSE SUBMITTED 6/27/78 o BYRON /BRAIDWOOD FSAR DOCKETED 11/30/78 o BYRON SER PUBLISHED 2/82 o BRAIDWOOD SER PUBLISHED 11/83 o PUBLIC HEARING ON SAFETY MATTERS TO BE SCHEDULED o CONSTRUCTION STATUS: UNIT 1 80%

UNIT 2 547 o APPLICANT FUEL LOAD DATE: UNIT 1 4/1/86 UNIT 2 7/1/87 o BYRON LOW POWER LICENSE ISSUED 10/31/84 o BYRON COMMISSION MEETING FOR FULL POWER LICENSE SCHEDULED FOR 2/12/85 I

O A-ss

a l

O DUPLICATE PLANT CONCEPT THE BYRON AND BRAIDWOOD STATIONS USE A DUPLICATE PLANT DESIGN IN ACCORDANCE WITH THE NRC'S " STATEMENT ON STANDARDIZATION OF NUCLEAR POWER PLANTS" DATED 8/31/78.

O Sim;LTANEOUS REVIEW OF THE DUPLICATE PORTIONS OF A LIMITED NtNBER OF PLANTS TO BE CONSTRUCTED WITHIN A LIMITED TIE SPAN AT MJLTIPLE SITES.

O O APPROVED DUPLICATE DESIGN MAY BE INCORPORATED BY REFERENCE IN OL APPLICATIONS, UNLESS SIGNIFICANT NEW INFORMATION EXISTS WHICH SUB-STANTIALLY AFFECTS THE FINDINGS OF THE REFERENCE DESIGN REVIEW OR OTHER GOOD CAUSE.

i O

a-ss

O DUPLICATE PLANT CONCEPT, CONT, STAFF'S RFVIEW 0F REFEPENCE DESIGN DOCUMENTED IN THE BYRON SER (NUREG-0876), INCLUDING FIVE SUPPLEMENTS O N) CLEAR STEAM SUPPLY SYSTEMS 0 BALANCE OF PLANT SYSTEMS 0 ASSOCIATED AUXILIARY SYSTEMS STAFF'S FINAL DUPLICATE DESIGN APPROVAL (FDDA) DATED 6/82 DELINEATES O TOPICS OUTSIDE THE SCOPE OF THE DUPLICATE DESIGN:

O SITE-RELATED CHAPACTERISTICS O CHANGES FROM Tm BYRON STATION DESIGN 0FFSITE POWER SYSTEMS

- WATER SYSTEMS RIVER SCREENHOUSE VENTILATION SYSTEM AND DIESEL GENERATOR FUEL OIL SYSTEM 0 UTILITY-ORIENTED SAFETY-RELATED PATTERS 0 OTHER I T 4S O

A-55

STATUS OF OUTSTANDING ITEMS O

(1) PurP AND VALVE OPERABILITY (2) SEISMIC AND DYN#41C CUALIFICATION OF EQUIPMENT (3) ENVIRCMENTAL QUALIFICATION OF ELECTRICAL AND MECHANICAL EQUIPMENT (14) EMERGENCY PREPAREDNESS PLANS AND FACILITIES (5) PPOCEDURES GENERATION PACKAGE (PGP)

(6) CONTROL ROOM Hlf%N FACTORS REVIEW (7) TURBINE MISSILE EVALUATION (8) IMPROVED WERMAL DESIGN PROCEDURES (9) TMI ACTION ITEM II.F.2: INADEQUATE CORE COOLING INSTRLPENTATION (10) CONFORMANCE OF ESF FILTER SYSTEM TO RG 1.52 (11) FIRE PROTECTION PROGRAM (12) VOLUME REDUCTION SYSTEM ,

1 O

n-se

i O

STATUS OF CONFIRMATORY ITEMS --

(1) APPLICANT COMPLIANCE WITH WE COPNISSION'S REGULATIONS (2) PIPING VIBRATION TEST PROGRAM (3) PPESERVICE INSPECTION PROGRAM (ll) REACTOR VESSEL MATERIALS (5) ELECTRICAL DISTRIBUTION SYSTEM VOLTAGE VERIFICATION O (6) taoerenoe"ce os aeou~oa"r s'ecTarc^' s4FE1Y Eouiem NT (7) RPM QUALIFICATIONS (8) INSERVICE TESTING OF PLN S AND VALVES (9) CHARGING PLW DEADHEADING (10) REMOTE SHUTDOWN CAPABILITY (11) TMI ACTION PLAN ITEM II.D,1

'O A-51

_______.__.___.___________...__s 4

t I

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1 lh i

l I J k i

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l STATUS OF LICENSE CONDITIONS 3

(1) MASONRY WALLS .

4 (2) FIRE PROTECTION PROGRAM

.P f

e

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A-58 <

4:

O IN CONCLUSION o NRR IS PREPARED TO RECOMMEND LICENSING 0F BRAIDWOOD STATION

PENDING FAVORABLE RESOLUTION OF THE OUTSTANDING SER ITEMS BY NRR AND THE CONSTRUCTION ISSUES BY REGION III, o THERE ARE NO DIFFERING PROFESSIONAL OPINIONS RELATING TO THE BRAIDWOOD OPERATING LICENSE REVIEW, O

4

.O A-59

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

I APPENDIX V '

! BRAIDWOOD NRC CONSTRUCTION

, INSPECTION EXPERIENCE I

i i..

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BRAIDWOOD 1

i NRC I'

I l i .

CONSTRUCTION  !

4  !'

i

! INSPECTION  !

l r

4 iO i

EXPERIENCE l l

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A-40 l

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j INSPECTION HISTORY INITIAL INSPECTION OCTOBER 17, 1973

. MANAGEMENT MEETING TO DISCUSS

.. CECO MANAGEMENT

.. QUALITY ASSURANCE PROGRAM

.. PROJECT RESPONSIBILITIES AND STATUS

.. ENGINEERING CONSTRUCTION ACTIVITIES MONITORED s

. SOILS AND FOUNDATIONS

. CONTAINMENT AND OTHER SAFETY-RELATED STRUCTURES j . PIPING SYSTEMS AND SUPPORTS

. SAFETY-RELATED COMPONENTS

. SUPPORT SYSTEMS

() . ELECTRICAL POWER AND DISTRIBUTION

. INSTRUMENTATION AND CONTROL

. QUALITY PROGRAMS / ADMINISTRATIVE CONTROLS

. LICENSING ACTIVITIES

. PREOPERATIONAL TESTING

~

ONSITE INSPECTOR ASSIGNMENTS

. SRI CONSTRUCTION JANUARY 1982

. SRI OPERATIONS JANUARY 1983

. BCAP AUGUST 1984

. RI CONSTRUCTION DECEMBER 1984 O

hi-(el

l l

l

() SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE (SALP)

• PURPOSE EVALUATE AND IDENTIFY LICENSEE PERFORMANCE MAKE SOUND DECISIONS REGARDING REGIONAL RESOURCES 1

~

PREPARED BY REGION III - INPUT FROM O

RESIDENT INSPECTORS A

REGIONAL SPECIALIST INSPECTORS 1

NRR PROJECT MANAGERS -

NMSS PROJECT MANAGERS i l

l l

4 O

A-G

l O SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE (SALP) 1 BRAIDWOOD 1 AND 2 SALP RESULTS I

l CYCLE DATES RATING i

1 07/01/79 - 06/30/80 RATED AS AVERAGE IN '

7 0F 7 AREAS.

l 2 07/01/80 - 12/31/81 RATED CATEGORY 2 IN 7 0F 7 AREAS.

I a

3 01/01/82 - 12/12/82 RATED CATEGORY 1 IN 1 AREA. CATEGORY 2 IN 4 ,

l AREAS, CATEGORY 3 IN 2 AREAS.

4 01/01/83 - 06/30/84 RATED CATEGORY 2 IN 7 AREAS, CATEGORY 3 O IN 3 AREAS.

A -(.3

l i l 1

i 4

CURRENT STATUS

SUMMARY

i

. NUMEROUS CONSTRUCTION RELATED i

i PROBLEMS EXIST.

l . CECO HAS PROGRAMS TO IDENTIFY i

!O 1

AND CORRECT PROBLEMS.

i

. CECO HAS "FIRST" TEAM ASSIGNED 4

TO COMPLETE PLANT. -

i

)

. NRC HAS PLANNED PROGRAM

! TO FOLLOW CECO ACTIONS.

l lO A-64 i

I O PROBLEM AREAS i

i I

i.

)

• PIPING r

. TRACEABILITY j . MINIMUM WALL THICKNESS .

. NAME PLATE APPLICATION

. INSTRUMENTATION l

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t i

• HVAC

'l

. HANGER FITUP

. HANGER WELDING j

i

• ELECTRICAL i

l i .

! . CABLE SEPARATION j . UNAUTHORIZED USE OF BUTT SPLICES

, INCOMPLETE RECORDS l

. CONTROL OF REWORK l i

A-G5

i i

O CECO CORRECTIVE ACTION PROGRAMS

) -

20 CORRECTIVE ACTION PROGRAMS 1

i BCAP s

l 3

. REINSPECTION PROGRAM ,

t

!. . REVIEW OF PROCEDURES TO SPECIFICATION REQUIREMENTS

. REVIEW OF SIGNIFICANT CORRECTIVE ACTION PROGRAMS t i

O ,

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A-GG 1

a L

O ac mmmves .

e

INSPECTION PERSONNEL h

i

. SRI CONSTRUCTION

. RI CONSTRUCTION l i

l 3 . SRI OPERATION s i

. BCAP INSPECTOR l  !

i 1

. REGIONAL SPECIALISTS I 0

i

} MANAGEMENT OVERSIGHT 4

iO

. MONTHLY MANAGEMENT MEETINGS (BCAP)

! . SALP 4

MAJOR INSPECTION ACTIVITY 1

. NDE VAN j . CAT INSPECTION l r I

I

!O

A-M r ,, , . - - - - . , . . , -, .,.~,, .~- . - . . . -- --.-,,,,,r....,,-.w.,,,.,n,. .

,.--_-.e.,--.-a,.. -,,--.,- . . , , - -. , .,..,, , ..-.-...m. ,---.nn-,--- e,,

1 NDE VAN INSPECTION l

! l

MARCH AND APRIL 1984

I 464 HOURS ONSITE AND 240 HOURS OFFSITE l

i PURPOSE: VERIFY THE ADEQUACY OF THE QC PROGRAM FOR ,

NDE THROUGH INDEPENDENT TESTING RESULTS: FOUR VIOLATIONSJ 1

. UNACCEPTABLE RADIOGRAPHS O . OBSOLETE DRAWINGS

. FAILURE TO IDENTIFY NONCONFORMING CONDITIONS

. FAILURE TO IDENTIFY WELD DEFECTS l

SUMMARY

REASONABLE AGREEMENT BETWEEN INDEPENDENT ,

J VERIFICATION AND APPLICANTS DETERMINATION i

1 A-GS

4%

j:.e q.4mbu4A -'AL4 44W" 5.4wmpes a-4 t r I l j

i CURRENT ISSUE ,

f l

l

' i i

i t i j ABILITY TO MANAGE MULTIPLE REINSPECTION EFFORTS l

AND DO ONGOING WORK s 1 I

)

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l -. . ..--- . . . .- . , .. ._- - _ . . - . . _ . . _ _ . _. . . -

I

APPENDIX VI I i COMMONWEALTH EDISON COMPANY  !

I BRAIDWOOD PRESENTATION  !

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i COMMONWEALTH EDISON COMPANY i

! PRESENTATION T0 i ADVISORY COMMITTEE ON REACTOR SAFEGUARDS i

BRAIDWOOD STATION UNITS'l AND 2 (

1

-FEBRUARY 7, 1985 i

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PRC/ACRS WASHItGTON DC P.01

O 298TH ACR$ MEETING BRAIDWOOD STATION UNITS 1 and 2 t TENTATIVE SCHEDULE I

FEBRUARY 7. 1985 (2:30 PM to Cfose of Business.)

Topic Presenter Time *

!. Report by ACHS Braidwood Subconmittes J. Carson Mark 2:30 PM (Subcomittee Chaiman)

II. Report by NRC Staff J. Stevens

A. NRR Discussion

8. Region III Discussion W. Forney

)

C. I&E Discussion R. Heishman III. Presentations by Comonwealth (dison A. CECO Coments on NRC Staff Reports T. Maiman 3:30 PM

5. Organization and Management Over-view

1. Corporate Structure T, Maiman 3:40 PM

2. Braidwood Project M. Wallace 3:50 PM Management and Operations C. Overview of Braidwood Construction M. Wallace 4:10 PM and Startup Schedule D. Quality Assurance Overview E. Fitzpatrick 4:20 PM

• • • 8REAX*** 4:30 PM i

' E. Braidwood Construction Apprefsal T. Maiman 4:45 PM

- Program (BCAP)

F. Braidwood Design Features Klopp 5:00 PM

1. Relationship to 2fon PRA i

2. Savere Accidents O 3. Reiated Issues G. Seismic Margin Klopp 5:30 PM A-11 l - - - - - - - . - - - .-- .. . _. _ _ _ _ _ _ _ _

NE'ACPS WASHIPGTCN DC p,g i

2 4

J i

H. Elimination of Pipe Whip Shelton 5:40 PM

1. CECO Submittals

2. ALARA Considerations ,

i 3. RG 1.45 Leak Detection I

I. E:nergency Planning Golden 6:00 PM i

l

J. Physical Plant Security F. Willaford 6

10 PM (CLOSEDSES$10N)

IV. Public Presentations. if any 6:30 PM V. ACRS Full Connittee Caucus 6:46 PM VI. ADJOURN 7:00 PM i

O

• Approximately 50% of the presentation time for each presentation topic i is to aces.snodate ACRS comments, questions, and discussion.

1 I

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

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COMMONWEALTH EDIS0N COMPANY 1

1

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PRESENTATION lO 4

l l i s 1 iII.A. COMMENTS ON NRC STAFF REPORTS i

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O A-73

7 O COMMONWFALTH EDISON COMPANY COMMENTS ON NRC STAFF REPORTS MY NAME IS TOM MAIMAN. I AM MANAGER OF PROJECTS FOR COMMONWEALTH EDISON COMPANY AND HAVE RESPONSIBILITY FOR THE ENGINEERING, CONSTRUCIION, TESTING, AND STARTUP 0F THE FOUR BYRON AND BRAIDWOOD UNITS.

III.A.-1 THESE FOUR UNITS ARE ESSENTIALLY IDENTICAL, AND UTILIZE O THE SAME FSAR, WITH A FEW SITE SPECIFIC PAGES FOR EACll SiiE. THE DESIGN IS THE SAME, Tile ARCHITECT ENGINEER (SARGENT AND LUNDY) IS THE SAME, AND EVEN MANY OF THE DRAWINGS FOR BOTH SITES ARE THE SAME.

III.A.-2 I AM PROUD TO REPORT THAT JUST THIS PAST WEEK, ON SATURDAY, FEBRUARY 2, 1985, BYRON UNIT 1 ACllIEVED ITS INITIAL CRITICALITY. Tile NRC COMMISSIONERS MEETING FOR THE FULL POWER LICENSE IS SET FOR NEX1 WEEK, TUESDAY, FEBRUARY 12, 1985. Tills MILESTONE OF INITIAL CRITICALITY FOR BYRON UNIT 1 SETS Tile STAGE FOR MY COMMENTS ON THE NRC STAFF REPORTS ON BRAIDWOOD.

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FIRST, WITH RESPECT TO NRR ISSUES. THE OVERALL LIST OF ISSUES REMAINING OPEN IS RELATIVELY SMALL. THIS IS DIRECTLY ATTRIBUTABLE T0 THE DUPLICATE DESIGN OF BYRON AND BRAIDWOOD, WHICH HAS GENERALLY ALLOWED CLOSURE OF

( ISSUES WITH NRR ON ALL FOUR UNITS SIMULTANE0USLY. THE REMAINING ISSUES ARE EXPECTED TO BE CLOSED ON A SCHEDULE WHICH SUPPORTS OUR FUEL LOAD DATE OF MARCH 31, 1986.

I III.A.-4

'O NEXT, I'D LIKE TO DISCUSS THE CONSTRUCTION ASSESSMENT TEAM VISIT TO BRAIDWOOD.

l ALTHOUGH THE NRC CONSTRUCTION ASSESSMENT TEAM (CAT)

INSPECTION REPORT IS NOT YET ISSUED, WE HAVE ALREADY l

TAKEN A NUMBER OF ACTIONS IN RESPONSE TO THEIR PRELIMINARY OBSERVATIONS. BEFORE THE INSPECTION BEGAN WE i

MOBILIZED A TASK FORCE WITH REPRESENTATIVES FROM THE PROJECT ORGANIZATION, FROM QUALITY ASSURANCE, AND FROM i THE MAJOR CONTRACTORS ON-SITE. TASK FORCE MEMBERS l INTERACTED WITH THE CAT DAILY THROUGHOUT THEIR

! INSPECTION, AND MET COLLECTIVELY SEVERAL TIMES A WEEK TO l REVIEW AND UNDERSTAND IN PROCESS CAT OBSERVATIONS. THOSE WHICH WERE JUDGED TO REPRESENT DISCREPANT CONDITIONS WERE DOCUMENTED ON NON-CONFORMANCE REPORTS, AND ARE THEREBY ALREADY PROCEEDING THROUGH THE EVALUATION AND CORRECTIVE

,O ACT10N STAGES.

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O EVERY OBSERVATION MADE BY THE CAT WHICH WE ARE AWARE OF.

HAS BEEN DOCUMENTED IN OUR SITE TRACKING SYSTEM TO ASSURE SATISFACTORY CLOSE0UT. SUBSEQUENT TO THE CAT EXIT MEETING. PROJECT PERSONNEL AND QUALITY ASSURANCE HAVE MET TO FURTHER DISCUSS OBSERVATIONS AND IDENTIFY ADDITIONAL ACTIONS WHICH ARE EITHER PRUDENT OR REQUIRED. MORE0VER, WHEN THE CAT INSPECTION REPORT IS ISSUED WE WILL COMPARE THAT TO THE PRESENT LISTING 0F PRELIMINARY OBSERVATIONS.

IDENTIFY ANY AREAS NOT PRESENTLY ADDRESSED, AND AGGRESSIVELY PURSUE EVALUATION AND DISPOSITION OF NEW ITEMS.

4 1 WE HAVE A QUALITY ASSURANCE PROGRAM AT BRAIDWOOD WHICH HAS BEEN EFFECTIVE IN IDENTIFYING QUALITY ISSUES. WE

()

1 INTEND TO KEEP A HEIGHTENED FOCUS ON QUALITY ISSUES, NOT ONLY THOSE IDENTIFIED BY CAT. NRC AND INP0 BUT ALSO THOSE WE IDENTIFY OURSELVES. WE HAVE ESTABLISHED MANAGEMENT

, TRACKING SYSTEMS FOR ALL QUALITY ISSUES AND G0ALS TO ASSURE PROPER ATTENTION FOR RESOLUTION AND CLOSE OUT.

III.A.-5 LASTLY WE ARE WELL AWARE OF THE NEED TO BOTH PERFORM ONG0ING WORK TO ENSURE THAT IT MEETS QUALITY REQUIREMENTS AND TO DEV0TE NECESSARY MANAGEMENT ATTENTION FOR CLOSE OUT OF CORRECTIVE ACTION PROGRAMS. I AM CONFIDENT THAT WE HAVE MANAGEMENT TALENT AND SUPPORT STAFF SUFFICIENT IN ABILITY, DEDICATION, AND NUMBERS TO MEET THE DEMANDS OF

() BOTH ONG0ING WORK AND THE CORRECTIVE ACTION PROGRAMS.

ARE BOTH CAPABLE AND PREPARED TO PROVIDE THE NECESSARY WE A-w

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III.A.-l A-78

O BYRON STATION STATION LICENSED OCTOBER 31, 1984 FUEL LOADED NOVEMBER 17, 1984 INITIAL CRITICALITY FEBRUARY 2, 1985 EXPECT FULL POWER LICENSE FEBRUARY 12, 1985.-

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FEW REMAINING ISSUES CLOSURE OF BYRON ISSUES ALSO CLOSES BRAIDWOOD ISSUES DUE TO DUPLICATE DESIGN l

REMAINING ISSUES EXPECTED TO BE CLOSED ON SCHEDULE TO SUPPORT A MARCH 31, 1986 FUEL LOAD DATE l

O III.A.-3 A-so

CONSTRUCTION ASSESSMENT TEAM (CAT) ISSUES REPORT NOT YET ISSUED COMMONWEALTH EDIS0N COMPANY ORGANIZED TASK FORCE TO INTERACT WITH CAT MEMBERS ALL CAT OBSERVATIONS HAVE BEEN ENTERED INTO THE SITE TRACKING SYSTEM TO ASSURE SATISFACTORY CLOSE OUT O -

CERTAIN CAT OBSERVATIONS JUDGED TO REPRESENT DISCREPANT CONDITIONS WERE DOCUMENTED ON NON-CONFORMANCE REPORTS S0 EVALUATION AND CORRECTIVE ACTION COULD BEGIN IMMEDIATELY NEW ISSUES, IF ANY IDENTIFIED IN THE CAT FINAL I REPORT WILL BE LIKEWISE HANDLED ALL QUALITY ISSUES IDENTIFIED BY CAT, NRC, INPO OR COMMONWEALTH EDIS0N COMPANY ARE~ INCLUDED IN MANAGEMENT TRACKING SYSTEM TO ASSURE PROPER ATTENTION FOR RESOLUTION AND CLOSE OUT III.A. 84 R-U  !

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ONG0ING WORK MEETS ALL QUALITY REQUIREMENTS ONG0ING WORK IS BALANCED WITH CLOSE OUT OF CORRECTIVE ACTIONS EDIS0N IS RESOLVING ALL ISSUES AND COMMITTING RESOURCES TO ACCOMPLISH THIS TASK NO REASONABLE QUESTIONS ABOUT CONSTRUCTION QUALITY

([) THAT REMAIN UNANSWERED III.A.-5 A-st

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l- III.B.1 ORGANIZATION AND MANAGEMENT OVERVIEW 1

CORPORATE-STRUCTURE i

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O ORGANIZATION AND MANAGEMENT OVERVIEW CORPORATE STRUCTURE MY NAME IS TOM MAIMAN. NEXT, I WOULD LIKE TO DISCUSS CERTAIN ELEMENTS OF COMMONWEALTH EDISON'S CORPORATE STRUCTURE.

III.B.i.-1 THE TOP ECHELON OF EDIS0N'S NUCLEAR ORGANIZATION BEGINS WITH THE CHAIRMAN AND PRESIDENT, JIM O'CONNOR. REPORTING O 10 H1M IS THE exeCuT1vE vICe eaeS1 DENT, BIDE THOMAS AND TO HIM IS THE VICE PRESIDENT OF NUCLEAR OPERATIONS, CORDELL REED. HE HAS RESPONSIBILITY FOR THE NUCLEAR OPERATING STATIONS AS WELL AS LICENSING, AND TECHNICAL SUPPORT.

4 I AM THE MANAGER OF PROJECTS AND ALSO REPORT DIRECTLY TO l THE CHAIRMAN AND PRESIDENT. I HAVE OVERALL i RESPONSIBILITY FOR THE ENGINEERING, CONSTRUCTION, TESTING AND STARTUP OF THE FOUR BYRON AND BRAIDWOOD NUCLEAR UNITS.

REPORTING SEPARATELY TO THE CHAIRMAN AND PRESIDENT IS THE MANAGER OF QUALITY ASSURANCE. WALT SHEWSKI. HE IS RESPONSIBLE FOR QUALITY ASSURANCE AT THE CONSTRUCTION

! STATIONS AS WELL AS THE OPERATING STATIONS.

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O III.B.1.-2 THIS CHART SHOWS THE PROJECT AND QUALITY ASSURANCE ORGANIZATIONS DOWN ONE MORE LEVEL.

IN PARTICULAR IT SHOWS A SEPARATE PROJECT MANAGER FOR BYRON AND BRAIDWOOD AND THAT THE WORKING LEVELS ARE TIED CLOSELY TO THE TOP OFFICER IN THE COMPANY. MIKE WALLACE.

THE PROJECT MANAGER FOR BRAIDWOOD WILL DISCUSS HIS ORGANIZATION FOLLOWING ME.

THE CHART ALSO SHOWS THE ASSISTANT MANAGER OF QUALITY ASSURANCE. GENE FITZPATRICK. HE IS AT THE SAME LEVEL AS O MIKE WALLA'CE AND IS CURRENTLY ASSIGNED ESSENTIALLY FULL TIME TO BRAIDWOOD QUALITY ASSURANCE. GENE FITZPATRICK WILL ALSO BE ADDRESSING THIS COMMITTEE LATER ON.

ONE OF THE STRENGTHS OF EDISON'S ORGANIZATION IS THAT IT IS RELATIVELY FLAT AND ABSENT OF A NUMBER OF INTERMEDIATE LAYERS OF MANAGEMENT. COMMUNICATION IS SWIFT AND SIMPLE RIGHT TO THE TOP S0 DECISIONS CAN BE MADE QUICKLY WHERE NEEDED.

THIS IS THE SAME BASIC ORGANIZATION WITH MANY OF THE SAME PEOPLE THAT RECENTLY COMPLETED LASALLE COUNTY STATION l TWO UNITS, LICENSED AND NOW IN FULL COMMERCIAL OPERATION.

THIS IS THE SAME ORGANIZATION THAT IS FINISHING THE BYRON STATION - RECENTLY LICENSED ON OCTOBER 31, 1984. UNIT 1 ACHIEVED CRITICALLY ON FEBRUARY 2. 1985 AND ON FEBRUARY

,O 12, 1985. WE EXPECT TO RECEIVE A FULL POWER LICENSE.

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O THIS IS THE SAME ORGANIZATION THAT IS WORKING TO COMPLETE BRAIDWOOD.

I HAVE A GREAT DEAL OF CONFIDENCE IN THE PROJECT TEAM AND ITS ABILITY TO PERFORM. THIS INCLUDES OUR ABILITY TO MANAGE THE NECESSARY CORRECTIVE ACTION PROGRAMS IN A TIMELY AND SUCCESSFUL WAY WHILE COMPLETING THE ONGOING CONSTRUCTION ACTIVITIES IN ACCORDANCE WITH REGULATORY REQUIREMENTS. I ALSO HAVE CONFIDENCE IN THE EFFECTIVENESS OF OUR QUALITY ASSURANCE EFFORT AND THUS IN THE PLANTS CONSTRUCTION.

III.8.1.-3 O

I'D LIKE TO CONCLUDE WITH THIS:

1. EDISON MANAGEMENT DOES UNDERSTAND TODAY'S ENVIRONMENT: REGULATORY AND OTHER,

2. WE HAVE MOVED AGGRESSIVELY TO ADDRESS ALL AREAS OF CONCERN.

3. WE ARE MANAGING THE BRAIDWOOD PROJECT EFFECTIVELY TO COMPLETION. -

WHEN THE PLANT IS COMPLETE IT WILL INDEED MEET ALL  :

l REGULATORY REQUIREMENTS. THAT'S A CORPORATE COMMITMENT FROM JIM 0' CONN 0R, WALT SHEWSKI AND FROM ME. -

Q I WOULD NOW LIKE TO INTRODUCE MIKE WALLACE TO DISCUSS BRAIDWOOD'S ORGANIZATION AND STAFFING.

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O O O OMIRMAN & PRESIDENT Jim O' Connor MANAGER W MANAGER OF QUALI1Y ASSLEANE PROJECIS Walt Shewski Tom Maiman DIRECTOR OF PROJECT SOEDULING

& COST Q)NTROL

, @ Ed Peterson i

ASSISTANT MANAGER PROJECT MANAGER PROJECT MANAGER PROJECTS ENGINEERING QUALITY ASSURANCE BRAlre nn STATION BYRON STATION MANAGER Gene Fitzpatrick Mike Wallace 'Vern Schlosser Brent shelton

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1. EDIS0N MANAGEMENT DOES UNDERSTAND TODAY'S ENVIRONMENT: REGULATORY.AND OTHER.

2. WE HAVE MOVED AGGRESSIVELY TO ADDRESS ALL AREAS OF i CONCERN.

3. WE ARE MANAGING THE BRAIDWOOD PROJECT EFFECTIVELY

, TO COMPLETION.

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l III.B.2. ORGANIZATION AND MANAGEMENT OVERVIEW  !

BRAIDWOOD PROJECT  :

MANAGEMENT AND OPERATIONS -

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ORGANIZATION AND MANAGEMENT OVERVIEW O BRAIDWOOD PROJECT MANAGEMENT AND OPERATIONS GOOD AFTERN0ON. MY NAME IS MIKE WALLACE. I AM EMPLOYED BY COMMONWEALTH EDISON AS THE ASSISTANT MANAGER OF PROJECTS. FOR THE PAST TWO (2) YEARS I HAVE ALSO BEEN ASSIGNED AS THE FULL TIME ON-SITE PROJECT MANAGER FOR BRAIDWOOD.

III.B.2.-1 I WOULD LIKE TO TALK ABOUT THE BRAIDWOOD PROJECT TEAM.

SINCE THE PROJECT ITSELF IS VERY DYNAMIC. THE PROJECT O ORGANIZATION UNDERG0ES CHANGE. AS NECESSARY. TO MEET THE NEEDS OF THE TIME. THE MANAGEMENT APPROACH BEING FOLLOWED TODAY AT BRAIDWOOD IS MODIFIED SLIGHTLY FROM THE APPROACHES FOLLOWED ON OUR LA SALLE COUNTY AND BYRON NUCLEAR PROJECTS. OVER THE LAST TWO YEARS, A NUMBER OF CHANGES HAVE BEEN MADE IN OUR ORGANIZATION, IN ORDER TO PROVIDE EVEN BETTER CONTROL OF PROJECT ACTIVITIES.

I WOULD NOW LIKE TO TELL YOU A LITTLE ABOUT THE PROJECT TEAM AND WHAT THEIR JOBS ARE AT BRAIDWOOD. THE MANAGERS AND SUPERINTENDENTS SHOWN ON THIS CHART HAVE AN AVERAGE OF OVER 17 YEARS EXPERIENCE IN THE COMPANY AND NEARLY 14 ,

YEARS NUCLEAR EXPERIENCE. G0ING DOWN ONE RUNG ON THE MANAGEMENT LADDER, THE NEXT 13 LEAD SUPERVISORS HAVE NUCLEAR EXPERIENCE OF OVER 13 YEARS ON THE AVERAGE. AND EVEN MORE SPECIFICALLY, A LOT OF LA SALLE COUNTY AND BYRON EXPERIENCE HAS BEEN TRANSFERRED TO BRAIDWOOD:

O- ADDITIONAL BYRON EXPERIENCED INDIVIDUALS WILL CONTINUE TO BE ASSIGNED TO BRAIDWOOD, AS THEY BECOME AVAILABLE.

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-g-O LET ME NOW SAY A FEW WORDS ABOUT EACH OF THE GROUPS ON THE PROJECT TEAM.

THE PROJECT FIELD ENGINEERING GROUP IS LED BY WARREN VAHLE. FORMERLY SUPERVISING MECHANICAL FIELD ENGINEER ON OUR LA SALLE COUNTY PROJECT. THIS ON-SITE GROUP DIRECTS THE ARCHITECT / ENGINEERING ORGANIZATION LOCATED AT THE SITE. WE HAVE.FOUND THAT MORE TIMELY RESOLUTION OF FIELD l PROBLEMS CAN BE REALIZED BY LOCATING A SIZEABLE FORCE OF ARCHITECT / ENGINEER PEOPLE ON SITE.

l THE LICENSING AND COMPLIANCE GROUP SERVES AS THE PRIMARY INTERFACE BETWEEN THE PROJECT SITE ORGANIZATION AND THE

! NRC RESIDENT INSPECTORS. AS WELL AS ALL OTHER REGULATORY

O AND 1NSeeCTION eERSONNEL WHO VISIT THe SITE. INCLUDING THE RECENT NRC CONSTRUCTION ASSESSMENT TEAM AND THE INPO CONSTRUCTION PROJECT EVALUATION TEAM. IT IS THEIR I RESPONSIBILITY TO ASSURE TIMELY AND COMPLETE RESPONSES TO ANY CONCERNS WHICH ARE RAISED AND TO TRACK THE STATUS OF NRC ISSUES UNTIL THEY ARE SATISFACTORILY CLOSED. THIS GROUP IS HEADED BY CHUCK SCHROEDER. CHUCK BRINGS TO BRAIDWOOD EXTENSIVE LASALLE COUNTY EXPERIENCE WHERE HE

HELD A SENIOR REACTOR OPERATOR'S LICENSE AND LATER SERVED AS THE NUCLEAR LICENSING ADMINISTRATOR DURING THE 5% AND FULL POWER LICENSE PROCESS FOR BOTH LASALLE COUNTY UNITS.

PROJECT CONSTRUCTION IS HEADED UP BY DAN SHAMBLIN FORMER CONSTRUCTION SUPERINTENDENT ON OUR LA SALLE COUNTY PROJECT. HE AND HIS FIELD ENGINEERS MANAGE AND DIRECT THE WORK 0F THE CONTRACTORS UNDER THEIR AREAS OF RESPONSIBILITY. THEY COORDINATE CONTRACTOR ACTIVITIES.

O SET PRIORITIES. AND MONITOR PERFORMANCE.

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O NEXT IS THE PROJECT START-UP GROUP HEADED BY CHUCK TOMASHEK, THE START-UP SUPERINTENDENT WHO SERVED IN A SIMILAR CAPACITY FOR MOST OF THE BYRON PROJECT. ABOUT TWO YEARS BEFORE A UNIT IS READY FOR FUEL LOAD. THERE MUST BE ORDERLY COMPLETION OF SYSTEMS BY CONSTRUCTION PRIOR TO TURNING THEM OVER TO OPERATING. THIS INVOLVES CHECKING THAT ALL PARTS OF THE SYSTEM -- VALVES PIPE.

CABLE. AND INSTRUMENTATION -- ARE INSTALLED AND WORKING PROPERLY. IT MEANS CLEANING AND FLUSHING THE SYSTEMS.

IT ENTAILS COORDINATION AND FOLLOW-UP TO ENSURE THAT ALL PARTIES WORK TOGETHER TO COMPLETE CONSTRUCTION. AND FINALLY. IT MEANS GAINING ACCEPTANCE BY THE STATION AND THE STATION SUPERINTENDENT. THAT IS THE ROLE OF THE STARTUP GROUP. THE STARTUP SUPERINTENDENT ALSO OVERSEES THE PROJECT SCHEDULING AND EQUIPMENT EXPEDITING GROUPS.

Q' SINCE BOTH OF THESE ACTIVITIES ARE ALSO CRUCIAL TO THE.

TIMELY COMPLETION AND TURNOVER OF SYSTEMS.

JOHN GUDAC. FORMERLY ASSISTANT SUPERINTENDENT AT OUR QUAD CITIES NUCLEAR PLANT. IS OUR PROJECT STATION SUPERINTENDENT. JOHN HAS NEARLY 21 YEARS COMMERCIAL NUCLEAR POWER EXPERIENCE. INCLUDING HOLDING RESPONSIBLE POSITIONS AT OUR QUAD CITIES STATION AND BEING INVOLVED IN STARTUP ACTIVITIES AT DRESDEN. QUAD CITIES AND LASALLE COUNTY. THE PROJECT OPERATING GROUP UNDER JOHN HAS THE TASK 0F OPERATING AND ACCEPTING THE SYSTEMS AS THEY ARE COMPLETED AND OPERATING THEM AFTER THEY HAVE BEEN TURNED OVER. IN PARALLEL WITH THIS. THEY HAVE TO BE CONCERNED WITH THE TIMELY MANNING OF THE STATION.

TRAINING OF PERSONNEL, WRITING THE PROCEDURES TO OPERATE THE PLANT. AND GENERALLY DOING ALL THE TASKS NECESSARY TO O READY A STATION FOR OPERATION. BEFORE DISCUSSING THE OTHER GROUPS ON THIS CHART I WOULD NEXT LIKE TO DISCUSS THE EXPERIENCE OF OUR OPERATING GROUP.

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III.B.2.-2 THIS CHART SHOWS THE AVERAGE EXPERIENCE LEVEL OF THE KEY PERSONNEL IN OUR OPERATING ORGANIZATION. HERE ARE SOME EXAMPLES OF THE PRESENT EXPERIENCE LEVELS AT BRAIDWOOD.

THE THREE ASSISTANT SUPERINTENDENTS HAVE AN AVERAGE OF 17 YEARS OF NUCLEAR EXPERIENCE WITH 88% OF THEIR NUCLEAR EXPERIENCE BEING PWR-RELATED. THE THREE OPERATING ENGINEERS HAVE AN AVERAGE OF 10.7 YEARS OF COMMERCIAL NUCLEAR EXPERIENCE AND AN AVERAGE OF 12.8 YEARS TOTAL WORK EXPERIENCE. THE SHIFT ENGINEERS HAVE AN AVERAGE OF 10.2 YEARS OF NUCLEAR EXPERIENCE. THE THREE MAINTENANCE MASTERS HAVE AN AVERAGE OF 12 YEARS OF POWER PLANT Q EXPERIENCE.

THE EDUCATIONAL BACKGROUND OF OUR STATION PERSONNEL COMPLIMENTS THEIR POWER PLANT EXPERIENCE. WE HAVE TEN MASTERS LEVEL TECHNICAL GRADUATES, 60 TECHNICAL GRADUATES EIGHT NON-TECHNICAL GRADUATES AND 57 ASSOCIATE LEVEL GRADUATES.

WE BELIEVE THE NUCLEAR EXPERIENCE LEVELS OF THE BRAIDWOOD STAFF WILL PROVIDE FOR SAFE, RESPONSIBLE AND RELIABLE OPERATION OF THE FACILITY. .

IN 1961 COMMONWEALTH EDIS0N COMMITTED TO MEET THE REQUIlsEMENTS OF THE SHIFT TECHNICAL ADVISOR (STA) PROGRAM SET BY NUREG-0737 BY CREATING THE POSITION OF STATION CONTROL ROOM ENGINEER (SCRE). A SCRE IS A DEGREED TECHNICAL GRADUATE WHO HAS AN SR0 AND WILL ALSO MEET THE O REQUIREMENTS OF THE STA. THIS PUTS ENGINEERING A-H

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EXPERIENCE ON SHIFT AND IN THE CONTROL ROOM AND WE BELIEVE WILL SATISFY THE INITIATIVES NOW BEING DISCUSSED BY THE NRC IN THIS AREA. ESSENTIALLY ALL OF OUR INITIAL BRAIDWOOD SENIOR REACTOR OPERATOR AND REACTOR OPERATOR PERSONNEL ARE COMPLETING A HOT PARTICIPATION PROGRAM AT ZION STATION WHICH MEETS THE REQUIREMENTS FOR ON-SHIFT OPERATING EXPERIENCE DISCUSSED IN GENERIC LETTER 84-16.

l IT IS OUR UNDERSTANDING THAT THE STAFF HAS APPROVED

! BRAIDWOOD'S PROGRAM AND IT WILL PROVIDE THE NECESSARY EXPERIENCE PREVIOUSLY PROVIDED BY SHIFT ADVISORS.

III.B.2.-3 i

O THE THREE OTHER GROUPS SHOWN ON THIS CHART, AS WITH LICENSING AND COMPLIANCE. HAVE JUST BEEN FORMED THIS PAST YEAR. AS DIRECTOR OF THE BRAIDWOOD CONSTRUCTION ASSESSMENT PROGRAM, BCAP, NINU KAUSHAL HAS RESPONSIBILITY FOR IMPLEMENTATION OF THAT EFFORT. TOM MAIMAN WILL BE

, DESCRIBING THE BCAP PROGRAM IN SOME DETAIL IN A FEW MINUTES.

IN ORDER TO ASSURE PROPER COORDINATION OF ALL SUPPORT SERVICES WITHIN THE ORGANIZATION, INCLUDING DATA MANAGEMENT TRAINING, AND CLERICAL SUPPORT, WE HAVE ESTABLISHED AN ADMINISTRATIVE SERVICES GROUP.

THE FINAL GROUP ON OUR TEAM IS THE QUALITY FIRST GROUP.

COMMONWEALTH EDISON MAINTAINS A STRONG COMMITMENT TO THE QUALITY OF CONSTRUCTION AT BRAIDWOOD. TO EMPHASIZE AND O c'e^RLY COMMUNICATE THAT COMMITMENT, WE HAVE ESTABLISHED A-95

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O l A QUALITY FIRST PROGRAM. THE QUALITY FIRST PROGRAM PROMOTES THE COMPANY'S STRONG. POSITIVE ATTITUDE l REGARDING QUALITY AMONG THE ENTIRE WORK FORCE AT BRAIDWOOD AND. MOST IMPORTANTLY PROVIDES AN OPPORTUNITY FOR INDIVIDUALS TO EXPRESS ANY CONCERNS THEY MAY HAVE REGARDING THE QUALITY OF CONSTRUCTION. THROUGH THE -

QUALITY FIRST PROGRAM. INDIVIDUALS ON THE SITE CAN RAISE CONCERNS. HAVE THEM REVIEWED WITH ANONYMITY AND RECEIVE A PERSONAL FOLLOWUP RESPONSE.

THE PROGRAM ITSELF IS MULTIFACETED. IT PROVIDES AN OPPORTUNITY FOR EVERY INDIVIDUAL WHO IS PERMANENTLY ,

LEAVING THE SITE TO DISCUSS ANY CONCERNS THEY MAY HAVE REGARDING THE QUALITY OF CONSTRUCTION WITH A MEMBER OF Q THE QUALITY FIRST TEAM. IT INCLUDES SCHEDULED SESSIONS WITH EACH INDIVIDUAL IN CRITICAL WORK GROUPS PRESENTLY ON THE SITE. AS AN OPPORTUNITY TO RAISE ANY CONCERNS. IT INCLUDES ORIENTATION OF ALL CRAFTS ON SITE AS TO THE PROCESS AVAILABLE TO THEM FOR RAISING CONCERNS. ON AN ONGOING BASIS. AND. IT INCLUDES THE COMMUNICATION OF THE COMPANY'S POLICY AND COMMITMENT REGARDING WORK QUALITY.

ALONG WITH THE NEED. IMPORTANCE. AND RESPONSIBILITY FOR EACH INDIVIDUAL TO TAKE PRIDE IN PERFORMING QUALITY WORK.

IN FORMING OUR PROJECT TEAM AT BRAIDWOOD. WE HAVE CONTINUED TO DRAW ON OUR BROAD CORPORATE BASE OF EXPERIENCE BY REASSIGNING PERSONNEL FROM OUR BYRON AND LASALLE COUNTY NUCLEAR STATIONS. AS THEY BECOME AVAILABLE. THE CORPORATE COMMITMENT TO PROVIDE THE MOST CAPABLE AND EXPERIENCED PERSONNEL AVAILABLE hEMAINS STRONG. AND IS EVIDENCED THROUGH THE DEPTH OF EXPERIENCE lO EXISTING A-%

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THROUGHOUT THE COMMONWEALTH EDISON BRAIDWOOD PROJECT

, ORGANIZATION. FURTHER. WHEN NEEDED. WE HAVE DRAWN

. OUTSIDE EXPERTISE INTO OUR ORGANIZATION. AUGMENTING COMMONWEALTH EDISON PERSONNEL.

AS WE MOVE TOWARD COMPLETION AT BRAIDWOOD WE INTEND TO CONTINUE EVALUATING THE EFFECTIVENESS OF OUR PROJECT i ORGANIZATION. THIS WILL HELP ASSURE THAT OUR EFFORTS PRODUCE A QUALITY PLANT MEETING ALL DESIGN AND LICENSING REQUIREMENTS.

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O O O PREMEC1 MANAER Mike Hallace

@nLIlY FIRST DIRECIOR Ray Preston BRAlrimnn CONSTRUCTION ASSESSENT PRDERAM DIRECTOR ADRINISTRATIVE SERVIE S DIRECTOR Ninu Kaushal Terry m ilaren

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l PROJECI FIELD PRIMEC1 LIGNSING PRIMECT CDN51RUCTION PROJECT STARTUP PROJECT STATION ENGIKERING MAftAER AIS CGFLIANE SIFERINTEREIENT 54FERINTEISENT SLFERINTEIEENT SLFERINTENEENT I Herren Wahle Chuck Schroeder Den Shmblin Chuck Toumashek John Gudec i

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(17 t M)

O O O BRAIDWOOD OPERATING ORGANIZATION EXPERIENCE LEVELS IN YEARS l

t AVE. AVE. AVERAGE COMMERCIAL MILITARY TOTAL TOTAL EXP. PER RANGE l JOB TITLE Nuc NuC. OTHER* EXPERIENCE JOB TITLE YEARS

'STATICN SUPT. (1) 20.9 -

5.3 26.2 26.2 N/A l OP. Ass'T. SUPT (1) 11.3 7.0 -

18.3 18.3 N/A l MAIOT. Ass'T. SUPT. (1) 9.4 6.0 -

15.4 15.4 N/A l ADMIN. Ass *T SUPT. (1) 12.0 4.7 -

16.7 16.7 N/A

- 15 OPERATING ens. (3) 10.7 2.1 -

38.4 12.8 10 .

ens. (6) 9.8 -

2.2 60.9 10.2 6 - 15

'HiFT.

ZIrT FOREMAN (17) 7.5 4.5 -

204.7 12.0 6 - 19 ,

SCRE (3) 2.2 1.5 4.0 15.1 5.0 2 -

6 l (NS0) (18) 4.9 -

113.9 6.3 3 - 13 EA (41) 1.3 2.3 6.5 155.4 3.8 .8 - 9 E0 (6) 3.1 -

1.2 19.6 3.3 3 - 13 OP. STAFF (6) 7.4 3.3 5.0 69.5 11.6 9-15 MASTER MAINT. ENG. (3) 12.3 - -

36.9 12.3 9-15 N AD/ CHEM SUPV. (1) 9.7 - -

9.7 9.7 N/A 1 RAD /CHEMMGMT. (15) 6.3 -

11.1 105.4 7.0 .6 - 35 F.AD/ CHEM TECH. (17) 4.3 0.1 13.8 89.8 5.3 .8 - 12 Q.C. SUPV. (1) 11.3 6.0 -

17.3 17.3 -

> 0.C. STAFF (7) 5.7 -

30.0 71.2 10.2 6 - 13 l SECURITY ADMIN. (1) 9.3 5 -

14.3 14.3 N/A l TECH. STAFF SuPV. (1) 19.4 - -

19.7 19.4 N/A

.5 - 17 TECH. STAFF (49) 3.2 1.0 34.3 240.6 4.9 '

TRAINING SUPV. (1) 8.5 1.0 -

9.5 9.5 N/A l NUCLEAR TRAINING STAFF (13) 1.9 8.0 2.5 132.2 10.2 1 - 17 ,

OTHERs** (115) 3.9 0.1 259.7 719.5 6.3 .2 - 19 i TOTALS (328) - - -

2223.4 6.8 o INDICATES POWER PLANT RELATED EXPERIENCE AREAS 00 INDICATES REMAINDER OF PLANT STAFF OTHER THAN CLERICAL STAFF t

(12820) III.B.2-2

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COMMONWEALTil EDIS0N COMPANY PRESENTATION O

III.C. OVERVIEW 0F BRAIDWOOD CONSTRUCTION AND STARTUP SCHEDULE O

l A-ton

4 OVERVIEW OF BRAIDWOOD CONSTRUCTION AND STARTUP SCHEDULE J

MY NAME IS MIKE WALLACE.  ;

BEFORE CONTINUING WITH AN UPDATE OF THE CONSTRUCTION STATUS. I WOULD LIKE TO QUICKLY REVIEW A COUPLE OF SIGNIFICANT PROJECT MILESTONES.

III.C.-1 THE BRAIDWOOD PROJECT WAS AUTHORIZED IN SEPTEMBER, 1972.

O ^ CONSTRUCTION PERMIT WAS ISSUED IN DECEMBER, 1975. AND THE FIRST SAFETY-RELATED CONCRETE WAS SUBSEQUENTLY POURED IN MARCH. 1976 ELECTRICAL BACKFEED FOR UNIT 1 WAS COMPLETED AND THE

! SWITCHYARD WAS LIVENED IN NOVEMBER, 1981.

l THE START OF SYSTEM COMPLETION. OR TURNOVER FOR TESTING.

I BEGAN IN FEBRUARY. 1983.

COLD HYDRO 0F THE UNIT 1 REACTOR VESSEL WAS COMPLETED IN AUGUST. 1983.

a III.C.-2 O

h- Loz.

O WITH TilAT BACKGROUND, LET ME NOW REVIEW THE INSTALLATION PERCENTAGE COMPLETION FOR THE MAJ0R CONSTRUCTION BULK QUANTITIES OF UNIT 1. THIS BAR CHART INDICATES OUR STATUS ON UNIT 1 FOR THE INSTALLATION OF MAJOR COMMODITIES IN THE MECHANICAL AREA.

AS YOU CAN SEE, THE INSTALLATION OF SPOOLS, OR LARGE BORE PIPE IS NOW COMPLETE. Ti1E INSTALLATION OF LARGE HANGERS, SMALL PIPE, AND SMALL HANGERS HAS SEEN SIGNIFICANT PROGRESS IN THE PAST YEAR AS WE HAVE REACHED THE PRESENT PERCENTAGE COMPLETIONS OF 78%, 93% AND 44%.

INSTRUMENTATION WORK ON UNIT 1 IS 82% COMPLETE TODAY.

l O lii.C.-5 f

THE NEXT CHART Sil0WS OUR PROGRESS FOR SEVERAL OTHER COMMODITIES. CONCRETE AND CABLE PAN INSTALLATION FOR UNIT 1 IS COMPLETE. CONDUIT AND CABLE INSTALLATION ARE l

89% AND 72% COMPLETE, RESPECTIVELY. Ti1E INSIALLATION OF CABLES FOLLOWS CONDUIT AND IS A CRITICAL PATH ACTIVITY FOR US. SINCE WE HAVE MOVED AIIEAD WITH CONDUII, WE ARE '

NOW CONCENTRATING OUR EFFORTS ON CABLE INSTALLATION.

CLEARLY THE "DULK CONSTRUCTION" PHASE OF Tile PROJECT IS -

NOW GETTING BEHIND US, AS WE ARC WELL INTO THE SECOND l

PHASE OF THE PROJECT " SYSTEM COMPLETION".

O A-im

!O l

1 III.C.-4 0UR PLANS FOR SYSTEM COMPLETION AND PRE 0PERATIONAL TESTING ARE BASED ON DUPLICATING THE APPROACH USED ON IDENTICAL. SYSTEMS AT BYRON. WE ARE CONTINUING TO DO THIS IN ORDER TO REALIZE THE MAXIMUM BENEFIT FROM OUR EXPERIENCES GAINED AT BYRON. UNIT 1 IS DIVIDED INTO APPR0XIMATELY 224 " SYSTEMS" FOR CONSTRUCTION COMPLETION AND TESTING PURPOSES. OUR PERCENTAGE COMPLETION FOR SYSTEM TURNOVERS IS NOW 58%. AND OUR PERCENTAGE FOR TESTING COMPLETION IS NOW 32%. MORE IMPORTANTLY. OUR RECENT PROGRESS IN THESE AREAS HAS BEEN SIGNIFICANT AS THEY HAVE INCREASED OVER THE LAST TEN MONTHS FROM 22% AND Q 7% RESPECTIVELY. IN GENERAL SYSTEM TURNOVER AND TESTING ALSO CONTINUES TO BE A CRITICAL PATH ON OUR SCHEDULE.

HOWEVER RECENT PROGRESS GIVES US CONFIDENCE THAT OUR MANAGEMENT CONTROL SYSTEMS WILL ASSURE THE TIMELY COMPLETION OF ALL SYSTEMS AS WELL AS THE QUALITY OF THE COMPLETED SYSTEM.

OVERALL UNIT 1 IS NOW 80% COMPLETE. AS CONSTRUCTION ACTIVITES CONTINUE WITH OUR CONSTRUCTION WORK FORCE OF OVER 3200, WHICH IS THE PEAK FORCE LEVEL EXPERIENCED TO DATE ON THE BRAIDWOOD PROJECT.

III.C.-5 1

FINALLY LET ME SUMMARIZE OUR SCHEDULE FOR COMPLETION OF O IsE BRAIDw0OD PR03ECT. IN DECEM8ER. 1989. WE COMeLETED AN EXTENSIVE REVIEW AND EVALUATION OF PROJECT STATUS.

l l A-toy

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WITH PARTICULAR ATTENTION TO VERY RECENT CONSTRUCTION COMPLETION AND TESTING EXPERIENCES AT BYRON, OUR SISTER I PLANT. AS A RESULT OF THAT REVIEW WE HAVE REPLANNED OUR

, PROJECT SCHEDULE, AND IDENTIFIED NEW MILESTONE DATES FOR CRITICAL PROJECT ACTIVITIES PRIOR TO FUEL LOAD. OUR NEXT i

. KEY MILESTONE IS THE START OF THE EMERGENCY CORE COOLING SYSTEM TESTING BY MARCH 17, 1985. CURRENTLY, WE ARE PROJECTING THE START OF THAT TEST BY FEBRUARY 25, 1985.

AN IMPROVEMENT OF ALMOST ONE MONTH. WE THEN PLAN TO

. BEGIN OUR INTEGRATED HOT FUNCTIONAL TESTING IN JULY OF

THIS YEAR. THAT WILL BE FOLLOWED BY COMPLETION OF THE INTEGRATED LEAK RATE TEST IN SEPTEMBER. THOSE MILESTONES ARE FULLY IN SUPPORT OF OUR FUEL LOAD DATE FOR UNIT 1.

WHICH IS MARCH 31, 1986 AND OUR PLANT IN SERVICE DATE OF Q OCTOBER 30, 1986.

FOR THE SAKE OF COMPLETENESS. I HAVE ALSO INDICATED KEY MILESTONE DATES FOR UNIT 2. THAT CONCLUDES MY PRESENTATION ON THIS AGENDA ITEM.

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i (12140)  !

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O BRAIDWOOD PROJECT l

OVERVIFW --

CONSTRUCTION PROGRFSS PROJLCl AUTHORIZED -- SEPTEMBER, 1972 FIRST CONCRETE POURED -- MARCH, 1976 UNIT I BACKF'EED AND SWITCHYARD LIVENED -- NOVEMBER, 1981 O

• START OF SYSTEM TURNOVERS FOR TESTING -- FEBRUARY, 1983 COLD HYDRO 0F THE REACTOR VESSEL -- AUGUST, 1983 I!!.C.-1 (1210D)-

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UNIT 1

! EMERGENCY CORE COOLING SYSTEM TESTING -- MARCH 17, 1985 j

INTEGRATED H0T FUNCTIONAL TESTING -- JULY 15, 1985 INTEGRATED LEAK RATE TEST -- SEPTEMBER 15, 1985 O BEGIN FUEL LOADING -- MARCH 31, 1986 PLANT IN SERVICE -- OCTOBER 30, 1986 UNIT 2 4

COLD HYDRO 0F REACTOR VESSEL -- APRIL 30, 1986 i

i BEGIN FUEL LOADING .- JUNE 30, 1987

' PLANT IN SERVICE .- DECEMBER 31, 1987 III.C.-5 (1210D) A-lIO

O COMMONWEALTH EDISON COMPANY PRESENTATION O

III.D. QUALITY ASSURANCE OVERVIEW l

l l

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QUALITY ASSURANCE OVERVIEW O

I'M GENE FITZPATRICK, ASSISTANT MANAGER OF QUALITY ASSURANCE FOR COMMONWEALTH EDISON.

I'M G0ING TO COVER OUR DEPARTMENT ORGANIZATION, STAFFING, AND METHODS OF OPERATING. WITH SPECIAL EMPHASIS ON OUR BRAIDWOOD EFFORTS.

SLIDE III.D-1 THIS CHART SHOWS OUR OVERALL DEPARTMENT ORGANIZATION. THE MANAGER OF QUALITY ASSURANCE REPORTS TO THE CHAIRMAN AND PRESIDENT, S0 THAT THERE IS TOTAL INDEPENDENCE OF QUALITY ASSURANCE FROM THE

<- ENGINEERING, CONSTRUCTION, AND OPERATIONS FUNCTIONS. THE MANAGER OF QUALITY ASSURANCE HAS A SMALL STAFF AND THREE MAIN DEPARTMENTS - ONE FOR ENGINEERING AND CONSTRUCTION, ONE FOR OPERATIONS, AND ONE FOR MAINTENANCE. AS THE ASSISTANT MANAGER OF QUALITY ASSURANCE. I REPORT TO THE MANAGER AND SPEND PRACTICALLY ALL 0F MY TIME AT BRAIDWOOD, WHERE I HAVE DIRECT CONTROL 0F THE BRAIDWOOD GENERAL SUPERVISOR OF QUALITY ASSURANCE WHO IS RESPONSIBLE FOR THE QUALITY ASSURANCE OVERVIEW 0F THE BRAIDWOOD CONSTRUCTION ASSESSMENT PROGRAM

- BCAP - AND OF THE SITE DESIGN ACTIVITIES, AND THE BRAIDWOOD SITE QUALITY ASSURANCE SUPERINTENDENT, WHO IS RESPONSIBLE FOR ALL OTHER CONSTRUCTION SITE QUALITY ASSURANCE MATTERS.

/O

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A-ilt

PAGE 2 O.

THE ORGANIZATION IS WELL STAFFED WITH TALENTED PEOPLE WHO HAVE RELATED HANDS-0N-EXPERIENCE IN CONSTRUCTION, OPERATIONS, ENGINEERING AND NAVY NUCLEAR WORK. THEY HAVE RECEIVED EXTENSIVE TRAINING AND ARE QUALIFIED AND CERTIFIED, AS APPLICABLE, FOR THE WORK THEY PERFORM.

THE OVERALL QUALITY ASSURANCE OBJECTIVE IS TO ASSESS THE QUALITY OF A MULTITUDE OF ACTIVITIES ASSOCIATED WITH OPERATIONS, CONSTRUCTION AND ENGINEERING. THIS COVERAGE INVOLVES COMPANY WIDE ACTIVITIES, AT OPERATING STATIONS, CONSTRUCTION SITES AND CORPORATE ENGINEERING AND NUCLEAR ASSOCIATED OFFICES, AS WELL AS ACTIVITIES PERFORMED BY CONTRACTORS AND VENDORS PROVIDING MATERIALS AND G

(/ SERVICES.

SLIDE III.D-2 I'D LIKE TO GET A LITTLE MORE SPECIFIC CONCERNING 0. A.

ACTIVITIES AT BRAIDWOOD. THIS CHART SHOWS THE DETAILS OF THE BRAIDWOOD SITE QUALITY ASSURANCE ORGANIZATION. THE SITE QUALITY ASSURANCE GROUP IS TOTALLY INDEPENDENT OF THE BRAIDWOOD PROJECT, INCLUDING CONSTRUCTION, ALTHOUGH WE WORK CLOSELY WITH THE PROJECT IN RESOLVING CONSTRUCTION QUALITY ISSUES. WE HAVE FIVE SUPERVISORS.

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.o FOUR OF THESE SUPERVISORS. PLUS INCLUDING OUR BCAP Q.A. SUPERVISOR.

THE AUDIT COORDINATOR, REPORT TO THE SITE QUALITY ASSURANCE SUPERINTENDENT. THE NUMBERS BELOW THE BOXES INDICATE THE NUMBER OF ,

PERSONNEL ASSIGNED IN EACH AREA, 0VERALL A TOTAL OF 85. IN ADDITION TO THIS SITE QUALITY ASSURANCE GROUP THERE IS A 13 PERSON OPERATIONS QUALITY ASSURANCE GROUP AT THE STATION THAT MONITORS PRE 0PERATIONAL j TEST, STARTUP, AND OTHER STATION ACTIVITIES.

SLIDE III.D-3 i

OBVIOUSLY NUMBERS OF PEOPLE DON'T TELL THE WHOLE STORY. THE EXPERIENCE LEVELS OF OUR CONSTRUCTION SITE QUALITY ASSURANCE PERSONNEL ARE SHOWN ON THIS CHART.

THE SITE QUALITY ASSURANCE ORGANIZATION IS RESPONSIBLE FOR ASSURING THAT THE CONSTRUCTION CONTRACTORS MEET ALL REQUIREMENTS AND THAT THEY FULLY IMPLEMENT THEIR QUALITY ASSURANCE PROGRAMS.

WE DO THIS PRIMARILY THROUGH AUDITS AND SURVEILLANCES OF THE CONTRACT 0RS' ACTIVITIES AND THROUGH OVERVIEWS OF COMPLETED AND IN-PROCESS WORK.

O h- 119

PAGE 4 SLIDE III.D-4 SITE AUDITS AND SURVEILLANCES LOOK AT IN-PROCESS CONSTRUCTION ACTIVITIES. ON-SITE DESIGN EFFORTS CONDUCTED BY THE ARCHITECT ENGINEER AND PLANT CONSTRUCTION AND TESTING ACTIVITIES. THROUGH 1984 AT BRAIDWOOD OUR QUALITY ASSURANCE PEOPLE HAVE CONDUCTED OVER 400 AUDITS AND ALMOST 4000 SURVEIL!ANCES. IN 1984 ALONE WE PERFORMED 78 AUDITS AND 828 SURVEILLANCES. AUDITS AND SURVEILLANCES ARE CONDUCTED TO SCHEDULES APPROVED BY THE MANAGER OF QUALITY ASSURANCE AND AUDITS ARE CONDUCTED TO APPROVED PLANS AND CHECKLISTS. WE PURSUE TIMELY CORRECTIVE ACTION FOR DEFICIENCIES Q FOUND IN THESE AUDITS AND SURVEILLANCES. AND WE FOLLOW-UP M ENSURE THAT CORRECTIVE ACTIONS ARE COMPLETED AND EFFECTIVE IN FIXING THE ORIGINAL CONCERN AND THEN QUARTERLY THEREAFTER FOR AT LEAST A YEAR.

TWO OF OUR OVERVIEW PROGRAMS ARE PERFORMED BY THE ON-SITE TESTING AGENCY - THE PITTSBURGH TESTING LABORATORY - PTL - UNDER THE DIRECTION OF SITE QUALITY ASSURANCE AND PROVIDE US WITH ADDITIONAL ASSURANCE THAT THE CONTRACTORS' INSPECTORS ARE DOING A GOOD JOB AND THAT THE PLANT IS BEING BUILT RIGHT. ,

SLIDE III.D-5 O

V FIRST. WE HAVE AN OVERINSPECTION PROGRAM, WHERE QUALIFIED PTL INSPECTORS PERFORM INSPECTIONS IN SELECTED AREAS THAT HAVE RECEIVED INITIAL INSPECTION.

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OUR SECOND OVERVIEW PROGRAM IS CALLED THE UNIT CONCEPT INSPECTION WHERE A PORTION OF THE PLANT, SUCH AS A SPACE BOUNDED BY FOUR COLUMNS AND THE FLOOR AND CEILING IS SELECTED FOR INSPECTION TO VERIFY ITEMS IN THIS PORTION OF THE PLANT MEET VENDOR AND ARCHITECT ENGINEER DESIGN REQUIREMENTS. S0 FAR ABOUT 85% OF UNIT 1 AND 10% OF UNIT 2 HAS BEEN INSPECTED IN THIS MANNER.

BOTH OF THESE PROGRAMS INDICATE A HIGH RATE OF ACCEPTABILITY OF INITIAL INSPECTION RESULTS. ALSO, ANY DEFICIENCIES IDENTIFIED AS A RESULT OF THESE PROGRAMS ARE DISPOSITIONED AND TRENDED AS ANY OTHER INSPECTION DEFICIENCY WOULD BE.

ANOTHER MAJOR QUALITY ASSURANCE EFFORT AT BRAIDWOOD IS THE QUALITY ASSURANCE OVERVIEW OF THE BRAIDWOOD CONSTRUCTION ASSESSMENT PROGRAM - BCAP. BCAP ITSELF WILL BE DISCUSSED BY THE MANAGER OF PROJECTS, MR. MAIMAN, WHO FOLLOWS ME. I WANT TO MAKE A FEW COMMENTS ABOUT 0.A.'S ROLE IN BCAP.

SLIDE III.D-6 FIRST, BCAP IS BEING IMPLEMENTED UNDER THE APPLICABLE REQUIREMENTS OF THE COMMONWEALTH EDIS0N QUALITY ASSURANCE PROGRAM.

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SLIDE III.D-7 TO ASSURE THE QUALITY AND PROGRAMMATIC REQUIREMENTS OF BCAP ARE

MET, WE HAVE ESTABLISHED A DEDICATED BCAP QUALITY ASSURANCE OVERVIEW GROUP UNDER OUR GENERAL SUPERVISOR OF QUALITY ASSURANCE, WHO REPORTS TO ME. THIS GROUP CONSISTS OF EIGHTEEN (18) INDIVIDUALS AND IS ORGANIZED, AS SHOWN HERE, INTO FOUR SUBGROUPS - ONE FOR EACH OF THE THREE ELEMENTc 0F THE BCAP PROGRAM AND AN OVERINSPECTION GROUP.

SLIDE III.0-8 THIS CHART GIVES YOU AN IDEA 0F THE EXPERIENCE LEVEL OF OUR BCAP QUALITY ASSURANCE OVERVIEW GROUP, INCLUDING OUR INSPECTORS.

OVERALL FROM A QUALITY ASSURANCE STANDPOINT, WE BELIEVE BCAP IS

) ) WELL STRUCTURED, WELL DOCUMENTED, AND IS BEING CONDUCTED IN ACCORDANCE WITH THE APPROVED PLANS AND PROCEDURES.

IN

SUMMARY

, WE BELIEVE WE HAVE AN EFFECTIVE ORGANIZATION WITH A

, STRONG PRESENCE ON SITE. WE'VE HAD SOME PROBLEMS IN THE PAST, i

MAINLY INVOLVING INSPECTION AND DOCUMENTATION COMPLETENESS, BUT THEY ARE BEING SOLVED. REALISTICALLY WE WILL PROBABLY ENC 0UNTER SOME MORE AS WE PROCEED, BUT BOTH THE PROJECT ORGANIZATION AND QUALITY ASSURANCE ARE STRUCTURED AND STAFFED TO DETECT THEM AND ENSURE THEIR

! CORRECTION. IN SHORT, WE BELIEVE OUR QUALITY ASSURANCE PROGRAM IS WORKING PROPERLY AND WE ARE PREPARED JR) ENSURE THAT IT CONTINUES IN THIS MANNER AS WE MOVE THROUGH THE REMAINING CONSTRUCTION AND

([) PRE-0PERATIONALTESTINGINTOTHESTART-UPANDOPERATINGPHASES.

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O BRAIDWOOD QUALITY ASSURANCE REPORTING RELATIONSHIPS f

Manager Quality Assurance I

Assistant Manager Quality Assurance 1  :

l p Braidwood '

L -

Quality Assurance

-]; Braidwood ~'_

i ',__ Superintendent General Supervisor 4

! Quality Assurance Audit Coordinator i

I I I I l Testing Agency Electrical / HVAC/

! Mechanical 8 Procurement Pre-Ops Structural

! Site Design BCAP upeMsor Supervisor Supervisor Supervisor

) Quality Quality j Assurance Group Assurance Group 11 20 16 8 1

1 4 18

! SLIDE III.D-2

BRAIDWOOD SITE QUALITY ASSURANCE PERSONNEL EXPERIENCE (AVERAGE YRS. PER PERSON)  !

JANUARY 1985 4

TOTAL EXPERIENCE 13.1

NUCLEAR EXPERIENCE 7.5 I

(]) CUALITY ASSURANCE / QUALITY 5.1 CONTROL EXPERIENCE DEGRFFS BACHELOR'S 33 i

1

ADVANCED 10 l SLIDE III.D-3 W

i 1

O AUDITS /SURVEILLANCES COVER:

0 IN-PROCESS ACTIVITIES 4

0 ON-SITE DESIGN EFFORTS 0 PLANT CONSTRUCTION / TESTING THROUGH 1984:

O 4 OVER 400 AUD US

- 81 IN 1984  !

j 0 3916 SURVEILLANCES

- 828 IN 1984  :

I O Stroe II1.0-4 A-it.l

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QUALTY ASSURANCE OVERVIEW PROGRAMS  !

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0 OVERINSPECTIONS ,

! 0 UNIT CONCEPT INSPECTIONS 0 BCAP .

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! i l SLIDE III.0 ;

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

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

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.1 BCAP QUALITY ASSURANCE  ;

}

IMPLEMENTED UNDER APPLICABLE REQUIREMENTS OF 4

1 COMMONWEALTH EDISON QUALITY ASSURANCE PROGRAM

{

i 1

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. SLIDE III.D-6 l

A-liLS

O O O

) BRAIDWOOD CONSTRUCTION ASSESSMENT PROGRAM SITE QUALITY ASSURANCE OVERVIEW GROUP Assistant Manager Quality Assurance

. I Site General Supervisor Quality Assurance I I I I Lead Element Overinspection/

Lead Element Lead Element 2 CSR Engineer RPSR Engineer RSCAP Engineer Overview roup 1

I I I i

! Quality Engineers / Quality Engineers / Quality Engineers /

Specialists Specialists inspectors l Specialists l CSR - Construction Sample Reinspection

. RPSR -Reverification of Procedures to Specification Requirements l RSCAP -Review of Significant Corrective Action Programs i

j SLIDh Ill D -7 1 _ _ _ _ _-_ ____

O S

BCAP QUALITY ASSURANCE EXPERIENCE LEVELS (AVERAGE YRS. PER* PERSON)

) ENGINEERS / SPECIALISTS INSPECTORS l

TOTAL EXPERIENCE 21.5 14.9 i

i NUCLEAR EXPERIENCE 14.2 6.7 QUALITY ASSURANCE /0UALITY 5.6 12.1

~

CONTROL EXPERIENCE 5

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) 1 COMMONWEALTH EDISON COMPANY PRESENTATION I

Q- i r

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III.E. BRAIDWOOD CONSTRUCTION ASSESSMENT PROGRAM l

(BCAP)

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BRAIDWOOD CONSTRUCTION ASSESSMENT PROGRAM (BCAP)

O MY NAME IS TOM MAIMAN. I WOULD LIKE TO DISCUSS A NEW PROGRAM WHICH COMMONWEALTH EDIS0N HAS UNDERTAKEN AS A MEANS TO PROVIDE AN ADDITIONAL LEVEL 0F CONFIDENCE IN THE QUALITY OF CONSTRUCTION AT BRAIDWOOD. THE PROGRAM IS CALLED THE BRAIDWOOD CONSTRUCTION ASSESSMENT PROGRAM OR BCAP.

III.E.-1 BCAP IS A PROGRAM 0F SAMPLE REINSPECTIONS AND REVIEWS COVERING SAFETY-RELATED CONSTRUCTION ACTIVITIES AT THE BRAIDWOOD NUCLEAR STATION. IT HAS BEEN UNDERTAKEN AS A

() PRUDENT MEASURE TO ANSWER ANY LEGITIMATE QUESTION CONCERNING QUALITY OF CONSTRUCTION. BCAP IS IN ADDITION TO THE ONGOING VIGILANT IMPLEMENTATION OF CONTRACTORS' QUALITY CONTROL AND EDISON'S QUALITY ASSURANCE PROGRAMS.

III.E.-2 THE BASIC OBJECTIVES OF BCAP ARE THREEF0LD. THESE ARE TO ASSURE:

1. THAT THERE ARE NO PROGRAMMATIC DESIGN SIGNIFICANT PROBLEMS IN THE CONSTRUCTION OF BRAIDWOOD WHICH HAVE NOT BEEN IDENTIFIED AND ADDRESSED.

O A-m

[]

2. THAT THE ON-SITE CONTRACT 0RS' PROCEDURES GOVERNING THE ONG0ING SAFETY RELATED CONSTRUCTION AND QUALITY ASSURANCE ACTIVITIES ADDRESS ALL APPLICABLE DESIGN AND REGULATORY REQUIREMENTS.

3. THAT WHERE PAST CONSTRUCTION PROBLEMS HAVE BEEN IDENTIFIED WHICH RESULTED IN SIGNIFICANT CORRECTIVE ACTIONS, SUCH CORRECTIVE ACTIONS HAVE BEEN ADEQUATELY IMPLEMENTED AND DOCUMENTED.

III.E.-3 TO ACHIEVE THESE THREE OBJECTIVES THERE ARE THREE ELEMENTS TO THE BCAP.

1. THE FIRST ELEMENT IS THE CSR OR " CONSTRUCTION SAMPLE REINSPECTION". UNDER THIS ELEMENT, A SAMPLE OF THE CONSTRUCTION WORK COMPLETED PRIOR TO JUNE 30, 1984 IS REINSPECTED AND THE ASSOCIATED QUALITY DOCUMENTATION IS REVIEWED FOR COMPLETENESS AND ACCURACY.

2. THE SECOND ELEMENT IS THE RP_SR OR " REVERIFICATION 0F PROCEDURES TO SPECIFICATION REQUIREMENTS".

UNDER THIS ELEMENT ALL ON-SITE CONTRACTOR PROCEDURES IN EXISTENCE ON JUNE 30, 1984 AND GOVERNING ON-G0ING SAFETY RELATED CONSTRUCTION AND QUALITY ASSURANCE ACTIVITIES ARE REVIEWED TO ASSURE

(] THAT THESE PROCEDURES ADDRESS ALL APPLICABLE DESIGN AND REGULATORY REQUIREMENTS.

A-t2.s

e#

O

3. THE THIRD ELEMENT IS THE RSCAP OR " REVIEW 0F ,

SIGNIFICANT CORRECTIVE ACTION PROGRAMS". IT IS COMPOSED OF A REVIEW 0F THE IMPLEMENTATION, METHODOLOGIES AND RESULTING DOCUMENTATION ASSOCIATED WITH SPECIFIC SIGNIFICANT CORRECTIVE ACTION PROGRAMS IDENTIFIED PRIOR TO JUNE 30, 1984.

THE COMPLETION OF THIS ELEMENT WILL ASSURE THAT THESE SPECIFIC AREAS OF CONSTRUCTION ARE OF ACCEPTABLE QUALITY.

III.E.-4 O THE CHART SHOWN REPRESENTS A PICTORIAL 0F HOW BCAP FITS INT 0 OUR EFFORTS TO ACHIEVE WORK QUALITY. THE ARROWS INDICATE DIRECT INPUT OR FEEDBACK.

III.E.-S THE BCAP ACTIVITIES ARE CARRIED OUT BY A TASK FORCE ORGANIZED IN JUNE 1984. IT IS AN INfEGRf. FED ORGANIZATION INCLUDING PERSONNEL FROM COMMONWEALTH EDISON, SARGENT &

LUNDY, STONE & WEBSTER ENGINEERING CORPORATION, DANIEL CONSTRUCTION CORPORATION AND VARIOUS OTHER CONSULTANTS.

THE PERSONNEL FROM SARGENT & LUNDY PROVIDE THE BACKGROUND INFORMATION ON DESIGN AND DOCUMENTATION PRACTICES AND

- ALSO THEIR KNOWLEDGE OF THE EDIS0N SYSTEM. MOST OF THE

(_/ PROCEDURE AND REVIEW WORK IS PERFORMED BY THE STONE &

A- tz.9 y w - -

WEBSTER PEOPLE WHILE THE FIELD INSPECTIONS ARE PERFORMED BY INSPECTORS FROM DANIEL CONSTRUCTION CORPORATION. ALL THESE PERSONNEL ARE DEDICATED TOTALLY TO BCAP AND HAVE HAD N0 PRIOR INVOLVEMENT IN THE CONSTRUCTION ACTIVITIES BEING VERIFIED UNDER BCAP.

III.E.-6 THE BCAP TASK FORCE IS HEADED BY A DIRECTOR WHO IS A COMMONWEALTH EDIS0N EMPLOYEE. THIS DIRECTOR REPORTS T0 THE BRAIDWOOD PROJECT MANAGER WHO IN TURN REPORTS TO ME.

(\ /-') I REPORT TO THE CHAIRMAN AND PRESIDENT OF THE COMMONWEALTH EDIS0N COMPANY.

THE ACTIVITIES OF THE BCAP TASK FORCE ARE MONITORED BY A SPECIAL GROUP WITHIN THE QUALITY ASSURANCE DEPARTMENT.

THIS BCAP QA GROUP IS HEADED BY THE SITE QUALITY ASSURANCE GENERAL SUPERVISOR WHO REPORTS TO THE ASSISTANT MANAGER OF QA WHO IN TURN REPORTS TO THE MANAGER OF QA.

THE MANAGER OF QUALITY ASSURANCE REPORTS DIRECTLY TO THE CHAIRMAN AND PRESIDENT OF THE COMPANY. THUS, A QUALITY ASSURANCE OVERVIEW 0F THE BCAP TASK FORCE ACTIVITIES IS PROVIDED BY A SEPARATE AND DEDICATED GROUP REPORTING THROUGH THE 0A CHAIN DIRECTLY TO THE CHAIRMAN.

III.E.-7 A- 130

h IN ADDITON TO THE OVERVIEW BY THE QUALITY ASSURANCE DEPARTMENT, COMMONWEALTH EDIS0N HAS RETAINED THE SERVICES OF JOHN HANSEL TO PROVIDE AN INDEPENDENT EXPERT OVERVIEW 0F BCAP. FOR THIS PURPOSE MR. HANSEL HAS ASSEMBLED A TEAM 0F SENIOR EXPERTS, ASSISTED BY AN ON-SITE STAFF.

MR. HANSEL IS A NATIONALLY RECOGNIZED EXPERT IN QUALITY ASSURANCE AND QUALITY CONTROL. HE IS CURRENTLY THE PRESIDENT OF THE AMERICAN SOCIETY FOR'00ALITY CONTROL.

THE CHARTER OF THE EXPERT OVERVIEW GROUP IS TO PROVIDE ANOTHER LAYER OF INDEPENDENT OVERVIEW BY CARRYING OUT WHATEVER ACTIVITIES IT DEEMS APPROPRIATE TO ASSURE THAT THE BCAP EFFORT WILL MEET ITS OBJECTIVES. IN PARTICULAR,

() THL INDEPENDENT EXPERT OVERVIEW GROUP HAS REVIEWED THE OVERALL PROGRAM DOCUMENT, THAT WAS SUBMITTED TO THE NRC, TO CONFIRM THAT THE PROGRAM AS COMMITTED WILL FULFILL ITS OBJECTIVES. THE INDEPENDENT OVERVIEW GROUP AND ITS STAFF MONITORS THE ACTIVITIES OF THE BCAP TASK FORCE TO ASSURE THAT THE WORK IS PROPERLY FOCUSED AND THAT THE BCAP G0ALS ARE BEING FULFILLED. THE OVERVIEW GROUP DOES THIS BY PERFORMING AUDITS AND SURVEILLANCES AS IT DEEMS NECESSARY AS WELL AS OVERVIEW INSPECTIONS OF THE BCAP FIELD INSPECTIONS.

THE INDEPENDENT EXPERT OVERVIEW GROUP ALSO PERFORMS A REVIEW 0F BCAP INDENTIFIED DISCREPANCIES TO ASSURE THAT THE ENGINEERING EVALUATIONS HAVE BEEN ADEQUATELY PERFORMED AND DOCUMENTED. FINALLY, THE GROUP PREPARES MONTHLY REPORTS OF ITS ACTIVITIES WHICH ARE PROVIDED SIMULTANE0USLY TO THE NRC AND TO COMMONWEALTH EDIS0N g

\->3 MANAGEMENT, MR. HANSEL AND HIS GROUP REPORTS DIRECTLY TO THE MANAGER OF PROJECTS, THAT IS MYSELF, AND IS INDEPENDENT OF THE BCAP TASK FORCE AND THE BRAIDWOOD PROJECT MANAGER.

{} ,

III.E.-8 COMMONWEALTH EDIS0N HAS TAKEN MANY STEPS TO ASSURE CONFIDENCE IN THE QUALITY OF THE BCAP EFFORT. THE BCAP ACTIVITIES ARE BEING CARRIED OUT BY HIGHLY EXPERIENCED PERSONNEL, MOSTLY FROM OUTSIDE OF COMMONWEALTH EDISON.

THE INDIVIDUALS PERFORMING BCAP ACTIVITIES HAVE NOT BEEN ASSOCIATED WITH THE CONSTRUCTION WORK BEING REVIEWED OR REINSPECTED. THE BCAP EFFORT IS HIGHLY STRUCTURED AND ALL QUALITY RELATED ' WORK IS CONTROLLED THROUGH WRITTEN PLANS, PROCEDURES, CHECKLISTS AND INSTRUCTIONS.

(] THE BCAP WORK IS BEING PERFORMED UNDER THE COMMONWEALTH EDIS0N QUALITY ASSURANCE PROGRAM AND IS MONITORED BY A DEDICATED GROUP WITHIN THE COMMONWEALTH EDISON COMPANY OUALITY ASSURANCE DEPARTMENT. ALL BCAP PLANS, PROCEDURES, CHECKLISTS AND INSTRUCTIONS ARE REVIEWED AND CONCURRED IN BY THIS QUALITY ASSURANCE GROUP.

FURTHERM0RE,.THE INDEPENDENT EXPERT OVERVIEW GROUP PROVIDES AN EXTENSIVE OVERVIEW 0F ALL BCAP ACTIVITIES.

IN ADDITION, THE NRC HAS ASSIGNED A RESIDENT INSPECTOR AT THE BRAIDWOOD SITE WITH A FULL TIME AND INCLUSIVE RESPONSIBILITY FOR' MONITORING THE BCAP ACTIVITIES. THIS INSPECTOR HAS ACCESS TO ALL BCAP PROGRESS INFORMATION ON SITE. PROGRESS ON BCAP ACTIVITIES IS REPORTED MONTHLY TO THE NRC IN A MEETING OPEN TO THE PUBLIC. THUS, THE BCAP ACTIVITIES AND BCAP PROGRESS ARE OPEN TO CONTINUAL SCRUTINY IN A PUBLIC FORUM. THESE FACTORS SERVE TO O PROVIDE A HIGH LEVEL OF CONFIDENCE IN THE QUALITY OF THE BCAP EFFORT.

j A- DL

i 7-O 1

III.E.-9 IN

SUMMARY

, BCAP IS A COMPREHENSIVE PROGRAM 0F REVIEWS AND REINSPECTIONS. HEAVY EMPHASIS HAS BEEN PLACED ON THE QUALITY OF THE 8 CAP EFFORT WITH MANY BUILT-IN CHECKS. IT WILL PROVIDE CONFIDENCE IN THE QUALITY OF PAST CONSTRUCTION THROUGH REINSPECTIONS AND REVIEW. IT WILL PROVIDE CONFIDENCE IN ONG0ING CONSTRUCTION ACTIVITIES THROUGH A REVIEW 0F CONSTRUCTION PROCEDURES. IT WILL PROVIDE CONFIDENCE THAT THE SPECIFIC CORRECTIVE ACTION PROGRAMS HAVE BEEN 0R WILL BE IMPLEMENTED PROPERLY AND WILL SATISFY OUR COMMITMENTS. THUS THE COMPLETION OF THE BRAIDWOOD CONSTRUCTION ASSESSMENT PROGRAM WILL PROVIDE AN

(]) ADDITIONAL LEVEL 0F CONFIDENCE THAT THE CONSTRUCTION AT THE BRAIDWOOD NUCLEAR PLANT IS OF ACCEPTABLE QUALITY.

J 1

i

() 12300 A-in

(

o BCAP DEFINITION ,

PROGRAM 0F INSPECTIONS AND REVIEWS UNDERTAKEN AS A PRUDENT MEASURE TO ANSWER ANY LEGITIMATE QUESTIONS CONCERNING THE OVERALL QUALITY OF CONSTRUCTION AT THE BRAIDWOOD STATION BCAP IS IN ADDITION TO THE VIGILANT IMPLEMENTATION OF EXISTING QUALITY CONTROL AND QUALITY ASSURANCE .

l PROGRAMS i

O i

(

O e

III.E.-1 i

A-tM

O OBJFCTIVES OF BCAP TO ASSURE THAT:

THERE ARE N0 PROGRAMMATIC DESIGN SIGNIFICANT PROBLEMS IN THE CONSTRUCTION OF BRAIDWOOD WHICH HAVE NOT BEEN IDENTIFIED AND ADDRESSED THE ON-SITE CONTRACT 0RS' PROCEDURES GOVCRNING THE ONG0ING SAFETY-RELATED CONSTRUCTION AND QUALITY ASSURANCE ACTIVITIES ADDRESS ALL APPLICABLE DESIGN

()' AND REGULATORY REQUIREMENTS i I

WHERE PAST CONSTRUCTION PROBLEMS HAVE BEEN IDENTIFIED WHICH RESULTED IN SIGNIFICANT CORRECTIVE ACTIONS, SUCH CORRECTIVE ACTIONS HAVE BEEN ADEQUATELY IMPLEMENTED AND DOCUMENTED {

III.E.-2 f4- 135

O

\_/

PROGRAM EIFMENTS l

CSR -

CONSTRUCTION SAMPLE REINSPECTION l

- l A SAMPLE OF THE CONSTRUCTION WORK COMPLETED PRIOR TO JUNE 30, 1984 IS REINSPECTED AND THE ASSOCIATED QUALITY DOCUMENTATION IS REVIEWED FOR COMPLETENESS AND ACCURACY i RPSR -

REVERIFICATION OF PROCEDURES TO SPECIFICATION REQUIREMENTS

(

CONTRACTOR PROCEDURES IN EXISTENCE ON JUNE  !

30, 1984 AND GOVERNING ON-G0ING SAFETY

({}

RELATED CONSTRUCTION AND QUALITY ASSURANCE ACTIVITIES ARE REVIEWED TO ASSURE THAT THESE PROCEDURES ADDRESS ALL APPLICABLE DESIGN AND REGULATORY REQUIREMENTS RSCAP - REVIEW 0F SIGNIFICANT CORRECTIVE ACTION PROGRAMS THE IMPLEMENTATION, METHODOLOGIES, AND l RESULTING DOCUMENTATION ASSOCIATED WITH SPECIFIC SIGNIFICANT CORRECTIVE ACTION l

PROGRAMS IDENTIFIED PRIOR TO JUNE 30, 1984 ARE REVIEWED TO ASSURE THAT THESE SPECIFIC AREAS OF CONSTRUCTION ARE OF ACCEPTABLE QUALITY.

O III.E.-3 A-136

O Ed1EH1 Contractor PROJECT

• Design / Engineering Craft Skills Specification MANAGEMIDIT Production Procurement Controls Construction CONTROLS Inspections

• Testing / Start-up Construction QA i

1I PLANT BCAP P WORK  % NRC I QUALITY

• Residents

• Region Specialist

• CAT Task Force Inspection

• NDE Van BCAP QA Independent Overview I l l

O Ceco 94

• Field Inspection Audit

• Surveillance

_ -. . - . . .- . - A _1 3 1

BCAP TASK FORCE 1

• BCAP TASK FORCE SPECIFICALLY ORGANIZED IN JUNE, 1984, TO IMPLEMENT BCAP ACTIVITIES.

HEADED BY CECO. AND SUPPORTED BY SARGENT AND LUNDY ALONG WITH OTHER CONSULTANTS.

l l STONE AND WEBSTER AND DANIEL CONSTRUCTION COMPRISE  ;

MAJORITY OF THE TASK FORCE.

i O

• ALL BCAP PERSONNEL DEDICATED TOTALLY TO BCAP.

1 BCAP PERSONNEL HAVE HAD N0 PRIOR RESPONSIBILITY FOR THE CONSTRUCTION ACTIVITIES BEING VERIFIED.

4 l

[

1 III.E.-5

{ r 4 A-138

i O BCAP ORGANIZATIONAL CHART l Commonwealth Edison Company ' Chairman and President j Manager of Manager of Projects

          -Quality Assur.ance O

Assistant Manager Braidwood Independent Quality Assurance Preject Manager Expert l l Overview Group Site OA General Supv. BCAP Director BCAP QA BCAP Task Force Overview Group III.E.-6 (1238d) A -lSi

                                   . _.          - _ - -          _ _ _ . _ . _ .           _._a

INDEPENDENT EXPERT OVERVIEW GROUP (]) COMMONWEALTH EDIS0N COMPANY HAS RETAINED JOHN HANSEL TO HEAD A TEAM 0F EXPERTS TO PROVIDE AN INDEPENDENT OVERVIEW OF BCAP. THE INDEPENDENT EXPERT-0VERVIEW GROUP IS ASSISTED BY ON-SITE STAFF. THE INDEPENDENT EXPERT OVERVIEW GROUP OVERVIEWS ALL BCAP ACTIVITIES TO ASSURE THAT BCAP IS MEETING ITS OBJECTIVES BY: O - REVIEWING OVERALL PLANS, PROCEDURES AND IMPLEMENTING INSTRUCTIONS REVIEWING AND MONITORING BCAP QUALITY ASSURANCE PROGRAM CONDUCTING AUDITS, SURVEILLANCES AND OVERINSPECTIONS MONITORING BCAP PROGRESS AND Rd'IIEWING RESULTS ON AN ONG0ING BASIS. l i ([) III.E.-7 f\-\90

O QUALITY OF BCAP EFFORT BCAP ACTIVITIES ARE BEING CARRIED OUT BY EXPERIENCED PERSONNEL NOT ASSOCIATED WITH THE CONSTRUCTION WORK BEING REVIEWED OR REINSPECTED. BCAP EFFORT IS HIGHLY STRUCTURED USING DETAILED PLANS! PROCEDURES, CHECKLISTS AND INSTRUCTIONS. BCAP WORK IS BEING PERFORMED UNDER COMMONWEALTH EDIS0N COMPANY QUALITY ASSURANCE PROGRAM AND IS MONITORED BY A DEDICATED GROUP WITHIN COMMONWEALTH Q EDIS0N COMPANY QUALITY ASSURANCE DEPARTMENT. ADDITIONALLY AN INDEPENDENT EXPERT OVERVIEW GROUP OVERVIEWS ALL BCAP ACTIVITIES. ALL BCAP PROGRESS INFORMATION IS AVAILABLE TO NRC ON-SITE RESIDENT INSPECTOR. i BCAP PROGRESS IS REPORTED MONTHLY TO NRC IN A MEETING OPEN TO THE PUBLIC. O III.E.-8 n-isi _ ._ . - a

A U

SUMMARY

BCAP IS A COMPREHENSIVE PROGRAM 0F REVIEWS AND REINSPECTIONS. 4 HEAVY EMPHASIS HAS BEEN PLACED ON THE QUALITY OF THE BCAP EFFORT WITH MANY BUILT-IN CHECKS. i UPON COMPLETION, BCAP WILL PROVIDE CONFIDENCE IN: THE QUALITY OF PAST CONSTRUCTION THROUGH {' REINSPECTION AND DOCUMENTION REVIEWS () - THE QUALITY OF ONG0ING CONSTRUCTION THROUGH PROCEDURE REVIEWS THE SATISFACTORY COMPLETION OF SPECIFIC COMMITTED CORRECTIVE ACTIONS THUS BCAP WILL PROVIDE AN ADDITIONAL LEVEL 0F i  ! CONFIDENCE IN THE CONSTRUCTION QUALITY AT BRAIDWOOD 4 () III.E.-9

(1230D) fi-l92.

4

     ,-         -          . . . n . _    -.
                                             . - - - , . , . - _ __ - . , _ , , - , . -. ,,.-.I

r.- f !~~  ! 1- , i i s 4 i (  ; i , i  ! i t i COMMONWEALTH EDIS0N COMPANY  ! i t~ l PRESENTATION' l i S  ! i III.F. BRAIDWOOD' DESIGN FEATURES i i i I I i

                                                                                                      .1.1 e

iG l ! A-in -

BRAIDWOOD DESIGN FEATURES MY NAME IS GEORGE KLOPP. I AM A SECTION ENGINEER IN THE BYRON /BRAIDWOOD PROJECT ENGINEERING DEPARTMENT AND I WILL BE ADDRESSING BRAIDWOOD DESIGN FEATURES. I. R5LATIONSHIPTOZIONPRA A. GENERAL REMARKS

1. EVEN THOUGH NOT REQUIRED, WE HAVE PERFORMED A " MINI-PRA" FOR BYRON.

WITHIN STRICTLY DEFINED LIMITS THE BYRON MINI-PRA APPLIES DIRECTLY TO BRAIDWOOD. THE BYRON WORK IS DIRECTLY O LINKED TO THE ZION PRA SINCE IT WAS DERIVED BY EVALUATING EXCEPTIONS TO THE ZION PRA WHERE DESIGN DIFFERENCES EXIST.

2. THE LIMITATIONS ON APPLYING THE BYRON

                      " MINI-PRA" TO BRAIDWOOD STEM FROM THE SITE RELATED DESIGN DIFFERENCES.

SPECIFICALLY THIS CENTERS AROUND THE SEISMIC FRAGILITY OF THE ULTIMATE HEAT SINK.

3. WE ARE G0ING TO EXAMINE SOME SPECIFIC DESIGN FEATURES FROM THE STANDPOINT OF THEIR EFFECT ON PRA. LET ME POINT OUT THAT THE SPECIFIC SYSTEM EFFECTS TURN OUT TO HAVE A NEGLIGIBLE EFFECT ON O B TT M LINE RISK MEASUREMENT VALUES.

A-t+t

O B. EFFECTS OF SELECTED SPECIFIC DESIGN DIFFERENCES BRAIDWOOD VS ZION.

1. OFFSITE AC POWER SUPPLY RELIABILITY (AS FREQUENCY OF A TOTAL LOSS OF 0FFSITE AC POWER TO ONE UNIT) WAS ASSESSED TO BE ESSENTIALLY THE SAME FOR BYRON AND ZION. RECENT WORK BY EDIS0N CONFIRMS THAT THIS IS THE CASE WITH LESS THAN 10% DIFFERENCE BETWEEN THE NUMBERS DERIVED FOR ALL 3 STATIONS.

2. ON SITE POWER RELATIVE RELIABILITIES ARE MUCH MORE DIFFICULT TO SEPARATE DUE O TO THE NUMEROUS POSSIBLE POWER STATES ' AND PLANT CONDITIONS. HOWEVER A BASIC COMPARISON BETWEEN ZION AND BYRON /BRAIDWOOD CAN BE OBTAINED BY LOOKING AT A SPECIFIC SEQUENCE. THE WASH 1400 TMLB, WHICH INVOLVES A TOTAL LOSS OF ALL AC POWER AND A FAILURE OF AUXILIARY FEEDWATER. THE BYRON AND ZION FREQUENCY VALUES ARE ESSENTIALLY THE SAME (WITHIN A FACTOR OF 3) FOR THIS SEQUENCE. (IGNORING EXTERNAL EVENTS)

3. THE AUXILIARY FEEDWATER SYSTEMS AT ZION AND BYRON /BRAIDWOOD HAVE ALSO BEEN EVALUATED AND, DESPITE A DIFFERENT DESIGN CONCEPT HAVE RELIABILITIES FOR

O 1

A-lMS , _ _ _. _ _ . . _ _ __ __ _ _ _ . _ _ . _ , .. _ J

O , THE MOST PROBABLE POWER STATE THAT ARE ESSENTIALLY THE SAME (WITHIN A FACTOR OF 3).

4. THE ESSENTIAL ASPECTS OF SERVICE WATER

• SYSTEM RELIABILITY HAVE BEEN EVALUATED  !

FOR ZION AND BYRON /BRAIDWOOD. THE l SHARED, CROSS CONNECTED ZION ~ SYSTEM IS MORE RELIABLE (BY A FACTOR OF 103) THAN THE SEPARATED, BUT CROSS CONNECTED, BYRON /BRAIDWOOD SYSTEM. HOWEVER NEITHER VALUE SIGNIFICANTLY IMPACTS THE OVERALL RESULT IN TERMS OF RISK OR CORE MELT FREQUENCY.

5. THE 3 TRAIN ZION CONTAINMENT SPRAY ,

SYSTEM IS MORE RELIABLE THAN THE 2 TRAIN BYRON /BRAIDWOOD SYSTEM BY A FACTOR OF ABOUT 102 FOR THE MOST PROBABLE POWER STATE. WITH THESE FEW EXAMPLES. WE CAN PASS TO SOME

    " BOTTOM LINE" COMPARISONS.

l III.F.-1 , O A-nc.

I i FREQUENCY OF CORE MELT (INTERNAL EVENTS ONLY) ZIQN BYRON /BRAIDWOOD 4.2 x 10-5 9.5 x 10-6 1 FREQUENCY OF RELEASE CATAGORY 2R* (INTERNAL EVENTS ONLY) IIQN BYRON /BRAIDWOOD (]} 2.4 x 10-7 7 x 10-8 DOMINANT RISK CONTRIBUTOR THE BYRON /BRAIDWOOD CORE MELT FREQUENCY IS < LOWER BECAUSE, IN PART, A LARGE PORTION OF THE HUMAN ERROR ASSOCIATED WITH ZION'S MANUAL-ECCS SWITCH 0VER TO RECIRCULATION HAS BEEN ELIMINATED. THIS RESULTED FROM A SEMI-AUTOMATIC SWITCH 0VER SYSTEM IN THE BYRON /BRAIDWOOD DESIGN. SIMILARLY, THE 2R - RELEASE CATAGORY FREQUENCY IS LOWER AT (]) BYRON /BRAIDWOOD BECAUSE. IN PART. THE ELECTRIC POWER SUPPLY FOR THE FAN COOLERS A-M'l

'I O REQUIRES A 1 OUT OF 2 BUS AVAILABILITY WHEREAS ZION REQUIRES A 2 OUT OF 3 BUS -i AVAILABILITY. i l k THIS SERIES OF EXAMPLES HIGHLIGHTS THE FACT

THAT CONSIDERATION OF ISOLATED SYSTEMS RELIABILITY OUTSIDE THE CONTEXT OF AN INTEGRATED EVALUATION CAN BE VERY MISLEADING.

!         II. SEVERE ACCIDENTS A. GENERAL REMARKS

1. THE ZION PRA AND THE INDUSTRY DEGRADED O CORE RULEMAKING PROGAM OR IDCOR "AS IT IS CALLED" HAVE EXTENSIVELY STUDIED THE l ABILITY OF CURRENT PLANT DESIGNS TO ACCOMODATE SEVERE ACCIDENTS. I HAVE

BEEN DIRECTLY INVOLVED IN BOTH THE ZION PRA AND THE BYRON " MINI-PRA". ALSO I HAVE SERVED AS EDISON'S TECHNICAL REPRESENTATIVE ON IDCOR AND WAS THE TEAM LEADER FOR THE MAAP EVALUATION OF ZION.

2. IDCOR IS CURRENTLY WORKING WITH THE NRC STAFF TO DEVELOP AN ACCEPTABLE METHODOLOGY BY WHICH PLANTS CAN DETERMINE-IF THEIR DESIGNS FALL WITHIN THE GENERAL CONCLUSIONS OF IDCOR'S EVALUATION OF THE FOUR REFERENCE PLANTS.

A-l48

l 4 O

3. BASED ON MY BACKGROUND AND EXPERIENCE.

I BELIEVE THE BRAIDWOOD PLANT AS CURRENTLY DESIGNED WILL DEMONSTRATE AN ABILITY TO ACCOMODATE SEVERE ACCIDENTS EQUAL TO OR BETTER THAN ZION. B. ZION PRA AND IDCOR RESULTS. THE ZION PRA BROKE NEW GROUND IN ASSESSING THE PHENOMENOLOGY OF SEVERE ACCIDENTS.

'      HOWEVER, THE IDCOR PROGRAM HAS PROVIDED NEW INSIGHTS IN TWO KEY AREAS.

1. THE FIRST OF THESE AREAS IS SOURCE TERM O EVALUATION. THE ZION STUDY EMPLOYED WASH-1400 SOURCE TERMS OR RELEASE CATAGORIES, ADJUSTED AS NECESSARY TO ACCOUNT FOR DELAYED CONTAINMENT FAILURE. THE ADJUSTMENT RELIED TO SOME EXTENT ON ENGINEERING JUDGEMENT BUT IT 4

WAS CONSERVATIVE. THE IDCOR PROGRAM DEVELOPED A TECHNICAL ASSESSMENT OF i SOURCE TERM MAGNITUDE BASED ON DETAILED EVALUATIONS OF AEROSOL BEHAVIOR AND RELATED PHENOMEN0 LOGY. THIS, COUPLED WITH MORE RIGOROUS ASSESSMENTS OF THE TIME DELAY INVOLVED IN CONTAINMENT FAILURE FOR DOMINANT SEQUENCES, LED TO SOURCE TERMS MUCH REDUCED FROM THOSE IN THE ZION STUDY. A QUICK COMPARISON FOR THE TMLB SOURCE TERM AT ZION IS SHOWN O Bet 0w, A- t 'i'l

                             -7 O

III.F.-2 ISOTOPE ZION PRA ZION IDCOR Cs 7.5 E-2 1.7 E-3 I 1.05 E-1 1.7 E-3 TE-S8 4.5 E-2 2.0 E-5 THE ZION PRA VALUES ARE THE APPR0XIMATE MEAN VALUES FROM THE 2R RELEASE CATAGORY HISTORGRAM. ALL VALUES ARE FRACTIONS OF CORE INVENTORY. THE ZION O. PRA ESTIMATED CONTAINMENT FAILURE AT 12 T0 14 HOURS AFTER ACCIDENT ONSET AND ASSUMED A VERY LARGE FAILURE OPENING. IDCOR RESULTS SHOW A FAILURE TIME NEAR 30 HOURS AND A VERY SMALL FAILURE l OPENING WITH AN EXTENDED RELEASE TIME. THE IDCOR RESULTS FOR ZION SHOW RISK VALUES MUCH LOWER THAN THE MEAN VALUES PREDICTED IN THE ZION PRA.

2. THE SECOND KEY RESULT FROM THE IDCOR WORK INVOLVES THE TIME AND MEANS AVAILABLE FOR SUCCESSFUL OPERATOR INTERVENTION IN SEVERE ACCIDENTS. WITH A FEW NOTABLE EXCEPTIONS, REC 0VERY ACTIONS ARE NOT MODELLED IN THE ZION PRA. THE IDCOR PROGRAM HAS DEV0TED

(]) SIGNIFICANT ATTENTION TO THE A - iso

O IDENTIFICATION AND EVALUATION OF THE POTENTIAL FOR REC 0VERY ACTIONS. THE USE OF IDCOR'S MAAP CODE IN AN INTERACTIVE MODE ALLOWS AN ALMOST UNLIMITED VARIETY OF " SCENARIOS".TO BE ASSESSED. TO DATE. THE WORK DONE SHOWS THAT RECOVERY ACTIONS CAN PLAY A VASTLY MORE SIGNIFICANT ROLE IN RISK REDUCTION THAN HAS'8EEN CREDITED IN ANY EXISTANT PRA. C. APPLICATION OF IDCOR RESULTS TO BRAIDWOOD.

1. WHILE WE HAVE NOT FORMALLY EVALUATED O BRAIDWOOD (0R BYRON) AGAINST THE IDCOR WORK, WE HAVE EVALUATED THE BYRON /BRAIDWOOD BASIC PLANT'S DESIGN AGAINST ZION. THIS EVALUATION WAS PERFORMED BY INDIVIDUALS FULLY CONVERSANT WITH THE PLANT FEATURES FOUND TO BE IMPORTANT AT ZION. BASED ON THESE FACTORS. I CAN STATE THAT I EXPECT ANY FORMAL IDCOR EVALUATION OF BRAIDWOOD WOULD YIELD RESULTS NEARLY IDENTICAL TO THOSE DERIVED FOR ZION.

2. THESE RESULTS WOULD BE EXPECTED TO CLEARLY DEMONSTRATE COMPLIANCE WITH THE PROPOSED NRC SAFETY GOAL. THEY WOULD NOT REVEAL ANY SIGNIFICANT RISK OUTLIERS. ,

A - 15l

, _g_

III. RELATED ISSUES i

WE NOTE THAT THE SUBCOMMITTEE ASKED US ABOUT FLOODS AND INDUSTRIAL ACCIDENTS. BRAIDWOOD IS DESIGNED

SPECIFICALLY TO ACCOMODATE A SPECIFIC FLOOD AND TO ACCOMODATE SPECIFIC INDUSTRIAL ACCIDENTS (CHEMICAL AND EXPLOSIVE). WE ARE IN COMPLIANCE WITH REGULATORY REQUIREMENTS IN THIS REGARD. ALS0; I CAN SAY, BASED ON EXPERIENCE THAT THESE EVENTS ARE NOT EXPECTED TO BE RISK OUTLIERS IN ANY SENSE OF THE WORD.

t O - 12890 i h ~ ISL_- -- - - - - - - .. . ,

i i f

O i

i FREQUENCY OF CORE MELT , (INTERNAL EVENTS ONLY) i HQN BYRON /BRAIDWOOD 4.2 x 10-5 9.5 x 10-6 ! FREQUENCY OF RELEASE CATAGORY 2R* (INTERNAL EVENTS ONLY) j Z1Q8 BYRON /BRAIDWOOD i i 2.4 x 10-7 7 x 10-8 i i , THIS IS FOR THE RISK DOMINANT RELEASE CATEGORY i l I O i l l A - 153

  ,...,...__r-...,_.._.   .w.-_r. . __ -    -

____.-__n_ .s_..._,_g-_- ._,_. . . , _ , _ _ . . . . - , _ . _ , , ,-,_..,.,,,-~__.,,.-,,,y%..- , _ . , .-,-,.

t -i i. t [ a lO 1 l 4- , i i . 4 1 i 1 b i

ZION PRA ZION IDCOR  :

ISOTOPE i i 4 Cs 7.5 E-2 1.7 E-3 i j , 1 i l I 1.05 E-1 1.7 E-3  ; i e i  :

e 4

TE-SB 4.5 E-2 2.0 E-5 " I [ I i i !. j i t ! i t i b i .r i  : i  ! i i i 1 a fC 4 III.F.-2 If A -159

(12900)

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

O 1 1 l COMMONWFALTH EDIS0N COMPANY l PRESENTATION O III.G. SEISMIC MARGIN i l l l0 l i A -Iss

SEISMIC MARGINS O MY NAME IS GEORGE KLOPP. OUTLINE THE FOLLOWING ARE THE SEISMIC DESIGN CRITERIA FOR ZION, BYRON AND BRAIDWOOD. ZION BYRON BRAIDWOOD 0.170 0.20s 0.20s IN THE ZION PRA, WE PERFORMED AN EXTENSIVE ANALYSIS OF SEISMIC MARGINS. ALTHOUGH A VIG0ROUS SEISMIC ANALYSIS O BEYOND THE DESIGN BASIS HAS NOT BEEN PERFORMED AT BRAIDWOOD WE CAN REACH SOME CONCLUSIONS IN COMPARISON WITH ZION. III.G.-1 THE ATTACHED TABLE, FROM THE ZION PRA, SHOWS THE FRAGILITIES OF KEY ZION STUCTURES AND EQUIPMENT IN "G'S". (COLUMN A A). FOR BRAIDWOOD, THE FOURTH ENTRY, SERVICE WATER PUMPS, MAY BE DELETED. AT ZION THESE PUMPS ARE LONG SHAFT, VERTICAL PUMPS. THE BRAIDWOOD PUMPS ARE HORIZONTAL, SINGLE STAGE PUMPS LOCATED IN A SUB-BASEMENT OF THE AUXILIARY BUILDING. AS SUCH, THEIR FRAGILITY , WOULD TYPICALLY BE ON THE ORDER OF 6.0 TO 7.00 (BASED ON SIMILAR PUMPS IN THE ZION PRA). NOTE THAT THESE O FRAGILITIES ARE MEAN VALUES AND THAT WE ACKNOWLEDGE A A-15(.

i SIGNIFICANT UNCERTAINTY IN THESE VALUES. SIMILARLY, WE 1 ACKNOWLEDGE A SIGNIFICANT UNCERTAINTY IN THE FREQUENCY WITH WHICH EARTHQUAKES OF VARIOUS MAGNITUDES OCCUR. IN FACT, ON ZION, IT WAS THE INTERACTION OF THE " TAILS" 0F THESE 2 SETS OF PROBA8ILITY DISTRIBUTIONS THAT YIELDED SEISMIC RISK. AN EVALUATION BASED ON MEAN VALUES WOULD  ! HAVE SHOWN ESSENTIALLY NO RISK. THIS IS, PERHAPS THE MOST POWERFUL POSSIBLE ENDORSEMENT OF OUR SEISMIC DESIGN MARGINS. SINCE WE HAVE NOT CONSIDERED SEISMIC MARGIN ON THE BRAIDWOOD ULTIMATE HEAT SINK, WE CANNOT FULLY ASSESS THE

           " SEISMIC RISK" AT BRAIDWOOD. HOWEVER, ABSENT THAT g3       POTENTIAL CONTRIBUTION, THE SEISMIC MARGINS AT BRAIDWOOD V        ARE JUDGED TO BE EQUAL TO OR BETTER THAN THOSE OF ZION.

4 O (12890) A- 157

1 WM -~~- AEROX TELkCGPitfe u.o J er - c et.

                         .                                                                                                      \
                                                                                                                            . i TABl.E 7.2-3 FRAGILITY OF 1(EY 710N STRUCTURES AND EQUtPMENT                                 .

St%cture/Cout;weent eF sa sg Systel 0.20 0.20

                                 @         Offsite Power Cerenic Insulaters 0.tl (i                                                                                  3.40         0.37       0.90
                                 @         lll VAC 01strthttem Pare 1*                              ,

0.50

                                 $         126 VDC topwort'                                O l.60 O.37 0.43         0.19       0.34
                                 $         Service Watee em:
                                 $.        4.16CV luf tshgeer (crettering)*                 0.71        0 .36      0.47 0.34       0.47
                                 $         4401 Swit:Pseer (caattering!*                    3.72 0.47
                                 @          660V Meter Centrol Centers (chattering)*        0.72        0.s4
                                 $          Aus111ery Dutidirg. Failure of Centrate Skeer well 0.13       0.30       0.14 0.30
                                 $          Refueling water storage tenn                     0.73                  0.28
                                 $          intercernetting Piptns/sett Fatture teneath Reacter kilding 0.73
                                                                                                      '   .         0.33 Impac.t,,fetwee.n,Ros,ette        ord 3.78**      0 .28      0.4n
                                 $          Ceaseneste storage Tset                          0.03        0.ft       0.29
        %                        @          4,16CV Stesel Generators'                        O.86        0.36       0.37
                                 $           Crib House Ce15 apse of krip Ensleeure Reef 0.84        0.24       0.27 0           safet, inseesiea rue..                          0.90        0.20       0.37 0.25
                                 $           Centalment Ventiletten Ouctwort e u se ,ers 0.47                   0.63 O             : Voc setiertei iad nicti                      t.01       0.te       0. 3 0           core see ein                                 . i.:s         0.ts       0.48 O          ne. iter Cosi  i nt sinis sei4.f T.              i.ie        0.20       0.s3 0          4,100r Trentferv                                 : 39        0.25       0.60 I

lervice Water Systre 101.d Dipe 44' 0.87

                                  $                                                        i 1.40       0.?Q 0.17
                                  $           C1T Pfptng 20*                                   1.40       0.20 0.33 0           av aliterr evitata r. i o of C airete Reef ofernrari i.40       0.3t 0           f atture of Mstenry halts                        t.70       0.$0       0.f4 0                                                            t.74       0. .       0.23 l

' esat co.ie .r.iva.ent,versiistion s,sie. 1 i.s0 0.s4 0 Coisseie of freiiuriier tacieiure Stef 0.3,

         %                     *Fr6gttf ty values (Micated are forms etter, relay trip er other interinttlant er selfly reeeverable $3Seltjens, keersteveraelp fs(f ore il esseeled te MCur et seeut three times the inetcated fregilf ty valve.
                               **Applitects only with 4 medien ic=ee 9eune of 0.749 ead da

• 0.M.

A-n58 7.2-1s

          ~

l III.G.-1 l* .

O COMMONWEALTH EDIS0N COMPANY PRESENTATION 4 O III.H. ELIMINATION OF PIPE WHIP RESTRAINTS O  ! A - 159

O ELIMINATION OF PIPE WHIP RESTRAINTS MY NAME IS BRENT SHELTON AND I AM THE PROJECTS ENGINEERING MANAGER: REPORTING TO TOM MAIMAN. COMMONWEALTH EDISON'S POSITION REGARDING ELIMINATION OF PIPE WHIP RESTRAINTS IS AS SHOWN ON THIS SLIDE. III.H.-1 THAT POSITION IS THAT WE DESIRE TO MINIMIZE THE NUMBER OF q PIPE WHIP RESTRAINTS USED AT OUR PLANTS TO AS FEW V RESTRAINTS AS POSSIBLE, AND TO DELETE ANY PIPE WHIP RESTRAINT WHOSE ELIMINATION CAN BE TECHNICALLY JUSTIFIED. III.H.-2 THE BASIS FOR EDIS0N'S POSITION IS AS FOLLOWS: OUR EXPERIENCE AND THAT OF THE INDUSTRY SHOWS THAT PIPING FAILURES ARE MINIMAL. STRESSES IN ASME SECTION III PIPING ARE GENERALLY WELL BELOW ASME CODE ALLOWABLES WHICH MINIMIZES THE POTENTIAL FOR PIPE RUPTURE. ANOTHER ASPECT IS THAT OF COST. PIPE WHIP RESTRAINTS ARE EXPENSIVE TO INSTALL. THE REMAINING WORK FOR THE PRIMARY LOOP PIPE WHIP RESTRAINTS AT BRAIDWOOD IS ESTIMATED TO . COST $300,000 TO $400,000 PER UNIT EVEN THOUGH THESE {} RESTRAINTS ARE PARTIALLY INSTALLED. A- l(oo .

                              -2 O

THE ELIMINATION OF PIPE WHIP RESTRAINTS ALSO IMPR0VES THE THERMAL EFFICIENCY OF THE SYSTEM. DUE TO THE CLEARANCES INVOLVED BETWEEN THE PIPING AND PIPE WHIP RESTRAINT AND THE REQUISITE MOTION BETWEEN HOT AND COLD OPERATING CONDITIONS IT IS VERY DIFFICUL,T TO ADEQUATELY INSULATE AROUND A PIPE WHIP RESTRAINT. THIS RESULTS IN ADDITIONAL HEAT LOSS AT THE RESTRAINT LOCATION. . THE ACCESS FOR IN-SERVICE INSPECTION IS ALSO IMPROVED WITHOUT PIPE WHIP RESTRAINTS. IN MANY CASES SHIMS OR JET DEFLECTORS HAVE TO BE REMOVED FOR ACCESS T0 A WELD TO PERMIT THE REQUIRED INSPECTIONS. ADDITIONALLY, WITHOUT PIPE WHIP RESTRAINTS RADIATION EXPOSURE DURING MAINTENANCE AND IN-SERVICE INSPECTION IS SIGNIFICANTLY O REDUCED. THE FINAL POINT IS THAT THERE IS REDUCED POTENTIAL FOR BINDING OF PIPES WITHOUT THE PIPE WHIP RESTRAINTS. WE HAVE DESIGNED THE RESTRAINTS WITH PROPER CLEARANCE IN BOTH THE COLD AND HOT CONDITIONS AND WILL VERIFY THESE CLEARANCES PRIOR TO START-UP. THE PROBLEM IS THAT THE CLEARANCE IN THE HOT CONDITION IS ONLY ABOUT 1/4 INCH. ALTHOUGH THERE IS NOTHING WRONG WITH THIS TYPE OF DESIGN, OR ANYTHING WE COULD 00 DIFFERENTLY. IN MY JUDGEMENT, I JUST 00 NOT LIKE TO HAVE A PIPE EXPAND TOWARD A MASSIVE FIXED RESTRAINT WITH THIS SMALL OF A RESULTING CLEARANCE. III.H.-3 O O A -l(o l

{

O THE QUANTITY OF PIPE WHIP RESTRAINTS USED AT EACH UNIT AT BRAIDWOOD IS 124. THIS INCLUDES 4 PIPE WHIP RESTRAINTS

! WHICH HAVE INTEGRAL JET DEFLECT 0RS. TWENTY-FOUR (24) ! _ PIPE WHIP RESTRAINTS ARE LOCATED ON THE MAIN LOOP 4 ! PIPING. ANOTHER 35 RESTRAINTS ARE USED ON THE OTHER ASME SECTION III PIPING BECAUSE OF TERMINAL END AND REQUIRED INTERMEDIATE BREAKS. FINALLY. THERE ARE 65 ARBITRARY l INTERMEDIATE BREAK (AIB) PIPE WHIP RESTRAINTS BUT NONE 0F THESE ARE LOCATED ON THE PRIMARY LOOP PIPING. ON l , i JANUARY 7, 1985 WE RECEIVED PERMISSION FROM THE NRC STAFF ! TO ELIMINATE THE 65 ARBITRARY INTERMEDIATE BREAK WHIP l RESTRAINTS. OF THE REMAINING 59 RESTRAINTS OUR CURRENT. ! SUBMITTAL DATED SEPTEMBER 17, 1984 DISCUSSING LEAK BEFORE BREAK ASPECTS REQUEST DELETION OF THE 24 MAIN LOOP

O RESTRAINTS. WE INTEND T0-ASK THE NRC STAFF FOR l PERMISSION TO DELETE AS MANY OF THE REMAINING 35 l RESTRAINTS THAT WE CAN TECHNICALLY JUSTIFY.

) l III.H.-4 l l . WITH RESPECT TO OUR ALARA EVALUATIONS FOR MAINTENANCE CONSIDERATIONS W'E BELIEVE THAT THE ELIMINATION OF PIPE , WHIP RESTRAINTS WILL SIGNIFICANTLY REDUCE OUR IN-PLANT l, RADIATION EXPOSURE. IN OUR ARBITRARY INTERMEDIATE BREAK l . SUBMITTAL WE ESTIMATED THAT WE WOULD SAVE 1000 PERSON REM OVER THE LIFETIME OF THE BYRON AND BRAIDWOOD UNITS. THIS l WAS BASED ON ONLY A 4% SAVING OF THE MISCELLANEOUS PLANT l RADIATION EXPOSURE. AND WAS EXCLUSIVE OF STEAM GENERATOR 4 WORK. OUR ESTIMATE WAS JUST UNDER THE 9 PERSON REM PER !O YEAR eER UNIT SUBMITTED BY THE v0GTtE etANT AND JuST l

A -l(.E

i _q. O AB0VE THE 2 1/2 PERSON REM PER YEAR PER UNIT ESTIMATE BY CATAWBA UNIT II. GENERIC LETTER 84-04 ADDRESSING USI-A2, ASYMMETRIC BLOWDOWN LOADING FOR CERTAIN OPERATING PLANTS, ESTIMATED A MAINTENANCE DOSE SAVINGS WITHOUT PIPE WHIP RESTRAINTS OF 2 PERSON REM PER YEAR PER UNIT WITH A RANGE OF i TO 6 PERSON REM PER YEAR PER UNIT. OUR ESTIMATE FOR THE ELIMINATION OF 24 MAIN LOOP PIPING RESTRAINTS IS 1/2 PERSON REM PER YEAR PER UNIT. PERSON REM IS A USEFUL BUT SOMEWHAT ABSTRACT-0UANTITY. I WOULD LIKE TO PUT THE IMPACT OF PRIMARY LOOP PIPE WHIP RESTRAINTS IN A DIFFERENT PERSPECTIVE. ZION'S MAIN LOOP PIPING RUNS AT ABOUT 500 MR/HR AT CONTACT. THIS MEANS IF WE HAVE 2 MEN SPEND 2 HOURS REMOVING AND INSTALLING SHIMS O NEAR THE PRIMARY PIPING. THOSE WORKERS WILL RECEIVE THEIR QUARTERLY DOSE OF NEAR 1250 M REM IN TWO HOURS AND WOULD NOT BE ALLOWED TO WORK IN A RADIATION AREA FOR THE REMAINDER OF THE CALENDAR QUARTER. THIS TYPE OF WORK REPRESENTS A SIGNIFICANT IMPACT ON OUR OPERATIONS. III.H.-5 NEXT, I'D LIKE TO DISCUSS CONTAINMENT LEAK DETECTION. WE HAVE A SERIES OF METHODS AVAILABLE FOR DETECTING LEAKAGE IN THE CONTAINMENT. THESE METHODS INCLUDE VOLUME CONTROL TANK LEVEL, SUMP DRAINAGE FLOW, REACTOR COOLANT SYSTEM WATER INVENTORY BALANCE TECHNIQUES, AND GASE0US AND PARTICULATE CONTAINMENT ATMOSPHERIC RADIATION DETECTORS. lO l l A -IG3

O III.H.-6 OTHER METHODS INCLUDE A CONTAINMENT ATMOSPHERIC SAMPLING SYSTEM, CONTAINMENT AIR PRESSURE MONITORS FAN COOLER DRY AND WET-BULB TEMPERATURE TO DETERMINE DEWPOINT i CHANGES, AND SUMP PUMP RUN TIME TOTALIZING METERS. THE COMBINATION OF THESE DEVICES MEET THE REQUIREMENTS OF REGULATORY GUIDE 1.45. SOME OF THESE DEVICES ARE NON-IE DEVICES WITH PROVEN POWER PLANT CAPABILITY, AND WE HAVE A SUFFICIENT AND DIVERSE NUMBER SUCH THAT WE HAVE ADEQUATE ABILITY TO DETECT LEAKAGE. IN ADDITION THE CONTAINMENT SUMP MONITORS AND WEIRS ARE SEISMICALLY QUALIFIED. O IN

SUMMARY

, COMMONWEALTH EDISON WOULD LIKE TO MINIMIZE THE NUMBER OF PIPE WHIP RESTRAINTS USED IN OUR PLANTS AND WILL CONTINUE TO PURSUE THAT G0AL FOR THE VARIOUS REASONS l DISCUSSED PREVIOUSLY. l l 12950 A-l69

i e PIPE WHIP RESTRAINTS I CECO POSITION: MINIMIZE THE NUMBER OF PIPE WHIP RESTRAINTS TO O AS FEW RESTRAINTS AS POSSIBLE AND DELETE ANY PIPE WHIP RESTRAINT WHOSE ELIMINATION CAN BE TECHNICALLY JUSTIFIED. 4 III.H.-i A - m5

e --- l BASIS F09 COMMONWEALTH EDIS0N COMPANY POSITION I i EXPERIENCE SHOWS PIPING FAILURES MINIMAL STRESSES GENERALLY WELL BELOW ASME CODE ALLOWABLES COST SAVINGS IMPROVED THERMAL EFFICIENCY IMPROVED ISI ACCESS (]} REDUCED RADIATION EXPOSURE REDUCED POTENTIAL FOR BINDING PIPES i 'l ([) III.H.-2 A- 166

QUANTITY OF PIPE WHIP RESTRAINTS () _. (PER UNIT) l ASSOCIATED BREAK TYPE l TERMINAL END AND ARBITRARY RFOUIRED INTERMEDIATE JNTERMEDIATE ASME CLASS 1 PIPING MAIN LOOP PIPING 24' 0 OTHER PIPING . 23 9 ASME CLASS 2/3 PIPING 12 33 MAIN STEAM PIPING (TUNNEL) 0 23 () SUBT0TALS 59 65 TOTAL RESTRAINTS REQUIRED 124 MEB 3-1 BREAK CRITERIA TOTAL RESTRAINTS REQUIRED WHEN 59 65 ARE DELETED (ARBITRARY INTERMEDIATE BREAKS) TOTAL RESTRAINTS REQUIRED WHEN 35 24 ARE DELETED (LEAK BEFORE BREAK) l

     *4 MAIN LOOP PIPE WHIP RESTRAINTS INCORPORATE JET

(-) v DEFLECTORS IN THE-DESIGN III.H.-3 4 A .t c.7

O MAINTENANCE ALARA EVALUATION B & B AIB DOSE 6 PERSON REM /YR./ UNIT (1000 PERSON REM:4 UNITS: LIFETIME) - V0GTLE AIB DOSE 9 PERSON REM /YR.IUNIT (700 PERSON REM:2 UNITS: LIFETIME) CATAWBA UNIT 2 2 1/2 PERSON REM /YR./ UNIT O AIB DOSE (95 PERSON REM:1 UNIT: LIFETIME) GENERIC SAFETY EVALUATION 2 PERSON REM /YR./ UNIT 84 USI-A2 (RANGE 1-6 PERSON LBB CONCEPT REM /YR./ UNIT) AIB = ARBITRARY INTERMEDIATE BREAK  ; LBB LEAK BEFORE BREAK l O I11.s.-9 A-t G8 i

  ~

l

METHODS AVAILABLE FOR DETECTING LEAKAGE IN THE CONTAINMENT O VOLUME CONTROL TANK LEVEL CONTAINMENT FLOOR DRAIN SUMP WEIR BOX (1 GPM DETECTION CAPABILITY *) REACTOR CAVITY SUMP WEIR B0X (1 GPM DETECTION CAPABILITY *) REACTOR COOLANT SYSTEM WATER INVENTORY BALANCE CONTAINMENT ATMOSPHERE PARTICULATE RADIATION DETECTOR (1 GPM DETECTION CAPABILITY) CONTAINMENT ATMOSPHERE GASEOUS RADIATION DETECTOR O

                                        ~

(1 eeM DETECT 10N CAeAB1LITY) CONTAINMENT ATMOSPHERE SAMPLING SYSTEM (PROVIDES GRAB SAMPLE CAPABILITY TO BACK-UP RADIATION MONITOR) CONTAINMENT AIR PRESSURE  ; REACTOR CONTAINMENT FAN COOLER INLET & OUTLET DRYBULB TEMPERATURES  ; 1 REACTOR CONTAINMENT FAN COOLER INLET & OUTLET DEWPOINT TEMPERATURES l SUMP PUMP RUN TIME TOTALIZING METERS l

• AFTER LIQUID REACHES SUMP WEIR BOX l

III.H.-5 A -tut

O COMMONWEALTH EDISON COMPANY PRESENTATION O III.I. EMERGENCY PLANNING O A-i7o

O EMERGENCY PLANNING MY NAME IS JOHN GOLDEN. I'M SUPERVISOR OF EMERGENCY PLANNING FOR THE COMMONWEALTH EDISON COMPANY. THE PURPOSE OF MY PRESENTATION TODAY IS TO GIVE YOU, THE MEMBERS OF THE ACRS, A CAPSULE LOOK AT THE STATUS OF 0FFSITE EMERGENCY PLANNING AT COMMONWEALTH EDISON, WITH PARTICULAR ATTENTION FOCUSED ON THE PLANNING FOR BRAIDWOOD STATION AND ITS ASSOCIATED EMERGENCY RESPONSE FACILITIES. THE COMMONWEALTH EDIS0N EMERGENCY PLANNING PROGRAM INVOLVES THREE STATES: ILLIN0IS, WISCONSIN, AND IOWA, (PLUS INDIANA AS IT INVOLVES THE 50-MILE INGESTION () PATHWAY) AND TWELVE COUNTIES LOCATED WITHIN SIX 10-MILE EMERGENCY PLANNING ZONES. .SINCE 1980 WE HAVE HAD SEVENTEEN (17) EXERCISES AT FIVE STATIONS. IN 1985 WE WILL PARTICIPATE IN SIX MORE, INCLUDING THE NEAR TERM OPERATING LICENSE (NT0L) EXCERCISE FOR BRAIDWOOD. THIS EXERCISE IS SCHEDULED FOR NOVEMBER 13, 1985. III.I.-1 I ANTICIPATE FEW PROBLEMS AT THIS EXERCISE BECAUSE TWO 0F THREE COUNTIES AT BRAIDWOOD ARE NOW IN THE DRESDEN OFFSITE PROGRAM. IN ADDITION TO THESE EXERCISES, THERE ARE TWO OTHER KEY ASPECTS OF 0FFSITE PLANNING THAT ARE WELL UNDERWAY. THE () FIRST IS THE APPROVAL BY FEMA 0F STATE AND LOCAL PLANS. BOTH ILLIN0IS AND WISCONSIN STATE PLANS HAVE RECEIVED h - \ ~1 \

O CONDITIONAL "350" APPROVALS, AS HAVE THE SITE-SPECIFIC PLANS AT OUR FIVE OTHER NUCLEAR STATIONS. ONLY THE IOWA STATE PLAN AND THE QUAD CITIES SITE-SPECIFIC.FOR TWO IOWA COUNTIES HAVE NOT YET RECEIVED APPROVAL. THESE PLANS ARE STILL BEING REVIEWED BY FEMA REGION VII. THE BRAIDWOOD SITE-SPECIFIC PLAN WILL BE SUBMITTED FOLLOWING RESOLUTION OF ANY FINDING OF THE NT0L EXERCISE. THE SECOND PROGRAM IS THE INSTAL!.ATION OF SIRENS OR OTHER PUBLIC ALERTING EQUIPMENT WHICH WILL NOTIFY THE CITIZENS WITHIN TEN MILES OF A STATION SHOULD AN EMERGENCY OCCUR. THESE SYSTEMS ARE INSTALLED AND OPERATIONAL AT FIVE STATIONS AND WILL BE INSTALLED AT BRAIDWOOD THIS YEAR. APPROVAL BY FEMA 0F THE FIVE SYSTEMS OPERATING TODAY IS () ALL THAT REMAINS TO REMOVE THE CONDITIONAL STATUS FROM OUR "350" APPROVALS. FINAL APPROVAL IS CONTINGENT ON PASSING AN EFFECTIVENESS TEST. THE LASALLE COUNTY SYSTEM WAS TESTED IN 1984 AND WE AWAIT ITS FINDINGS. INFORMAL COMMENTS ON THE RESULTS HAVE BEEN FAVORABLE. MOST OF THE OTHER ALERTING SYSTEMS WILL BE TESTED THIS YEAR OR EARLY NEXT YEAR. A UNIQUE FEATURE OF THE BRAIDWOOD EMERGENCY RESPONSE FACILITIES IS ITS EMERGENCY OPERATION FACILITY (E0F). III.I.-2 THE MAZON E0F, LOCATED APPR0XIMATELY TEN MILES FROM BRAIDWOOD, IS SITED IN THE CENTER OF A TRIANGULAR AREA O' FORMED WITH OUR DRESDEN, LASALLE COUNTY AND BRAIDWOOD b-llE

O STATIONS AT THE CORNERS. THE MAZON E0F SERVES ALL THREE STATIONS. CECO HAS THREE OTHER EOFS, PLUS A BACKUP EOF FOR ZION BECAUSE THE ZION EOF IS WITHIN THE 10-MILE EPZ, AND A CORPORATE EOF IN CHICAGO. TOSUMMARIZE,THEOFFSITEEMERGENCYPLANNINGFOR BRAIDWOOD, NOW UNDERWAY, IS BASED ON PROVEN PRINCIPLES OF , PLANNING. SUCCESSFUL EMERGENCY PLANNING BY COMMITTED PEOPLE IS OUR G0AL. O l l

                                                                               \

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COMMONWFALTH EDISON COMPANY STATUS OF EMERGENCY PLANNING SUBJECT BYRON DRESDEN BRAIDWOOD EXFRCISES

    . NUMBER TO DATE      1                   5        N/A
    . NEXT SCHEDULED     6/85              -4/85      11/85 EPZ COUNTIES          OGLE              WILL       WILL GRUNDY     GRUNDY KENDALL    KANKAKEE 1980 POPULATION
  . . WITHIN 10 MILES     21,622             72,204    27,482
    . WITHIN 50 MILES     955,652         6,301,641  4,580,641 FFMA 3SO            APPROVED         APPROVED   TO BE SUBMITTED

(}}) AP ) ROVALS EARLY 1986 PROMPT NOT SYS.

    . INSTALLATION        1983               1982       1985
    . FEMA TESTING        1985               1985       1986 E0E
    . LOCATIONS           DIXON              MAZON"    MAZON*
    . RANGE (MI.)          19                  10         10

• ALSO SERVES LASALLE COUNTY STATION

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i i l , I t l I I 2 l l 1  ! 1 i l i i t- i i i i  ! !, COMMONWEALTH EDIS0N COMPANY i PRESENTATION f I l i f@ 3 I l III.J. PHYSICAL PLANT SECURITY i i 4 (CLOSED SESSION)  ! 1, . h I l i i s a > l i- r i r i  ; i A- n(,

W. kma S F ESENTATION ON FIRE PROTECTION g HISTORY MARCH 1975 BROWNS FERRY FIRE l FEBRUARY 1976 RECOMMENDATIONS RELATED TO BROWNS FERRY FIRE (NUREG-0050) MARCH 1976 BRANCH TECHNICAL POSITION APCSB 9.5-1 , FIRE PROTECTION GUIDELINES FOR NEW PLANTS ISSUED JUNE 1976 PROPOSED REG. GUIDE 1.120 (SIMILAR TO THE BTP) ISSUED FOR PUBLIC COMMENT AUGUST 1976 APPENDIX A TO BTP APCSB 9.5-1 FIRE PROTECTION O GUIDELINES FOR OPERATING PLANTS ISSUED (BACKFIT) DECEMBER 1979 SERs FOR ALL OPERATING PLANTS ISSUED (SOME HAD UNRESOLVED ITEMS) (MARCH 1979 TMI-2 ACCIDENT-I MUSTRY AND STAFF RESOURCES DIVERTED TO ASSOCIATED CONCERNS) 4 i 1 e h e l O A- 177

O APPENDIX R MAY 1980 PROPOSED RULE TO RESOLVE OPEN ITEMS ISSUED FOR PUBLIC COMMENT NOVEMBER 1980 FIRE PROTECTION RULE ISSUED RULE CONTAINS 10 CFR 50.48 AND APPENDIX R

                 ,.ALL OPERATING PLA?lTS TO HAVE A FIo.E PROTECTION PROGRAM
                 . APPENDIX R ESTABLISHES REQUIRED FIRE PROTECTION FEATURES FOR CERTAIN ITEMS WHICH WERE UNRESOLVED I:1 SER ISSUED IN 1978
                 . SECTIONS III.G, III.J AND III.0 BACKFIT TO ALL PLANTS
                 . STAFF TO PEVIEW ONLY ALTERNATIVE SHUTDOWN MODIFICATIONS OR EXEMPTION REQUESTS
                 . MODIFICATIONS COMPLETION SCHEDULE (50.48)

FEBRUARY 1981 FIRE PROTECTION RULE EFFECTIVE (72 UNITS LICENSED PRIOR TO 1/1/79) l i O A-ils

O UTIt'ITY GROUP JANUARY 1981 UTILITY GROUP PETITIONED COMMISSIONERS TO STAY- , BACKFIT OF SECTIONS III.6, III.J AND 111.0 PENDING JUDICIAL REVIEW JANUARY 1981 UTILITY GROUP PETITIONED US COURT OF APPEALS FOR I REVIEW 0F FIRE PROTECTION RULE JUNE 1981 UTILITY GROUP'S MOTION TO STAY DENIED BY COMMISSION (CL1-81-11) l , , MARCH 1982 COURT UPHELD NRC's ADOPTION OF THE FIRE PROTECTION ! RULE (C0URT NOTED THAT NRC PROCEDURES GAVE THE - O INDUSTRY THE MINIMUM ACCEPTABLE OPPORTUNITY TO l RESPOND AND PROVIDED LIMITED TECHNICAL GUIDANCE. ! . NEVERTilELESS THE RULE IS TEMPERED BY THE EXEMPTION PROCESS. USING THIS PROCESS, POWER PLANTS WILL BE ABLE TO SHOW THAT ALTERNATIVE FIRE PROTECTION I SYSTEMS PROTECT THE PUBLIC SAFETY. THEIR FAILURE TO MAKE SUCH SHOWING WILL BE FURTHER PROOF THAT PUBLIC SAFETY URGENTLY REQUIRED A STRINGENT FIRE

PROTECTION PROGRAM FOR NUCLEAR POWER PLANTS) i i OCTOBER 1982 SUPREME COURT REJECTS APPEAL BY UTILITY GROUP  !

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O tiCENSiNs Mites 10NES MARCH 1981 SUBMITTALS DUE FOR EXEMPTIONS AND MODIFICATIONS  : FOR ALTERNATIVE SAFE S30TDOWN--60 PERCENT REQUEST EXTENSION 2 PLANTS (FT. ST. VRAIN AND D. C. COOK) CLAIM COMPLIANCE  : JULY 1982 REVISED DATE FOR COMPLETING SUBMITTALS JULY-DEC. 1982 RECEIVED APPROXIMATELY 500 EXEMPTION PEQl'ESTS, ' 10 UNITS REQUESTED NO TECHNICAL EXEMPTIONS AT THIS TIME, 30 UNITS COMPLETED DEC. 1982 APPEAL PROCESS FOR EXEMPTION DENIALS ESTABLISHED 225 EXEMPTIONS DENIED JAN.-SEPT. 1983 RESOLVING 225 EXEMPTION DENIALS THROUGH REVIEW 0F REVISED EXEMPTION REQUESTS AND NEW PROPOSED MODIFICATIONS FOR ALTERNATIVE SAFE SHUTDOWN CAPABILITY OCT 1983 ISSUED GENERIC LETTER 83-33. STAFF POSITIONS BASED ON REVIEW 0F EXEMPTION REQUESTS AND INSPECTION P.ESULTS i NOV. 1983 ED0 MEETING ON INSPECTION RESULTS O . A-iso . -_ _ --

O FEB. 1984 INDUSTRY SEMINAR BASED ON GENERIC LETTER 83-33 AND INSPECTION RESULTS MARCH 1984 INDUSTRY APPEALS GENERIC l.ETTER 83-33 -- STATF INTERPRETATIONS OF APPENDIX R MARCH-APRIL 1984 REGIONAL WORKSHOPS MAY 1984 DIFFERING PROFFESSIONAL 9 PINIONS OF FPE'S, COMMISSION MEETING--COMMISSION TO APPROVE STAFF INTERPRETATIONS AND POLICY JAN.1984-PRESENT SUBMITTALS OF NEW EXEMPTION REQUESTS AND ALTERNATIVE SHUTDOWN MODIFICATIONS BASED ON 83-33, WORKSHOPS AND O iNSeeCTiON RESut1S PRESENT STATUS - ORIGINALLY SUBMITTED UNRES9LVED EXEMPTION REQUESTS: 7 UNITS NEW EXEMPTION REQUESTS - 38 UNITS NEW ALTERNATIVE SAFE SHUTDOWN MODIFICATIONS - . 15 UNITS ABOUT 1/3 ORS HAVE NO EXEMPTION REQUESTS . OR STAFF APPROVALS PENDING NT0L REVIEW FOR ABOUT 16 PLANTS ONG0ING l O - i I A-1st )

O _ COMPLETED SAFE SHllTDOWN INSPECTIONS PRE-79 POST-79 FY 82 D.C. COOK 1/2 ., FY 83 FT. CALHOUN, TROJAN, DAVIS BESSE, Q FR. ST. VRAIN, VERMONT YANKEE FY 84 SALEM, CALVERT CLIFFS 1/2 WATERFORD 3, FERMI 2, il0LF CREEK, BYRON 2 LIERICK, MCGUIRE EL85_. MAINE YANKEE COMANCHE PEAK, BYRON l NINE MILE POINT 1 (REINSPECTION), FERMI j

(REINSPECTION),SHOREHAM l 1

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APPENDIX VIII m PRESSURIZED THERMAL SH0CK OF REACTOR

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i PRESSURE VESSELS o.m n; p \ r .

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Meeting No. Agenda Item Randout No. 298 4 1 1 Title Pressurized Thermal Shock of Reactor Pressure Vessels 7 3

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List of Documents Attached f b} ADDITIONAL DOCUMENTATION: A. Tentative Schedule

B. SHEW)0N's memo to ACRS Members, Subj

BASDEKAS' OBJECTIONS TO ISSUING THE FINAL PTS RULE, dated Feb. 6,1985 C. D. BASDEKAS memo to D. Ward, ACRS, Subj : PROPOSED RULE ON PRESSURIZED THERMAL SHOCK, dated February 5, 1985

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O A  ! 298TH ACRS MEETING Tentative Schedule Basdekas' Professional Opinion on Proposed Rule on Pressurized Thermal Shock February 7, 1985 1:30 - 2:30 p.m. I. Report of P. Shewmon, Chairman, Metal Components 1:30-1:40 p.m.

Subcommittee II. Presentation by D. Basdekas, RES, Electrical 1

40-1:50 p.m.

Engineering Instrumentation and Control III. Commints by F. Schroeder, Asst. Director 1:50-2:00 p.m. () for Generic Projects, DST 2:00-2:30 p.m. IV. Discussion by Committee and I. Catton, ACRS i Consultant i t t l l k~IS _. 2 _ __ ._

[ean UNITED STATES NUCLEAR REGULATORY COMMISSION 8 o 4 .I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20605 February 6, 1985 i MEMORANDUM FOR

ACRS Members FROM: P. G. Shew.non, Chairman, Metal Components Subcommittee

SUBJECT:

D. BASDEKAS' OBJECTIONS TO ISSUING THE FINAL PTS RULE Mr. D. Basdekas has urged CRGR and Mr. W. Dircks that the PTS Rule not be issued. The reasons for this are not developed in the detail that one would like, but he clearly believes that the screening criterion (RT-NDT) in this Rule is set too high. We have been asked by Mr. Dircks to review this matter, and the Commission has written that they will hold up the issuance of the Rule until we have done so. The reasons given by Mr. Basdekas are numerous, but are never developed in sufficient detail to lead to a quantitative number about how the things he alludes to might increase the probability of failure. I feel that his assertions can be grouped into two categories:

1. The steel PV is more prone to fracture than allowed by the Staff through the presence of undetected flaws, or due to a higher transition temperature. Often, but not always, he asserts that this is supported by foreign experience that the Staff has ignored.

I I have looked at this and can find no merit to his arguments. l 2. There is evidence that credible transients exist which would take j the PV to lower temperatures than is considered credible by the Staff. 1 I have asked Ivan Catton to look at this and he has written a . I report which you have in front of you. He traces the history of the transient that he feels Basdekas is mcst concerned about and concludes that "The ACRS should reccmmend that the PTS rule be issued as it is...", with the additional recommendation that we urge the prompt completion of some calculations about the effect of steam generator overfill. Three studies by)the H. B. Robinson-2 have been NRC (Oconee-1, completed whichCalvert considerCliffs-1, the and probability of various pressure, temperature transients that could occur in these plants. Catton feels that these have been well done, and support the Staff position. In view of the above I can find no technical basis for holding up I the issuance of the PTS Rule. l 1 l l

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p e" "'% uNiTso sTAvas R TE C i ' ' 'I O W T V Us.!STTEE ON k p ,7 k NUCLEAR REGULATORY COMMIS$10N .0. JW.YSS, UAN.RA d [" j WASHINGTON. D. C. 2066s e February 5,1985 N eu

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                                                                                                  'A David A. Ward, Chairman                                                      '

MEMORANDUM FOR: Advisory Committee on Reactor Safeguards FROM: Demetrios L. Basdekas Electrical Engineering Instrumentation and Control Branch, DET, RES

SUBJECT:

PROPOSED FINAL RULE ON PRESSURIZED THERMAL SHOCK I understand that as'a result of Mr. Dircks' request to Mr. Ebersole, dated December 19, 1984, the ACRS has scheduled a session to discuss certain concerns I have expressed regarding the subject Rule and related matters for the after-noon of February 7,1985. Based on the wishes of the Committee as conveyed to me by your staff. I plan to attend this session as "an interested observer". I will be available to discuss any questions the Committee may wish me to discuss. Since it may be necessary for me to comment at the end of the session, regardless of whether you have any questions for me, I request that you make a few minutes available Os to me at that time. I am enclosing a copy of a paper by Kussmaul on the Geman Basis Safety Concept, published in the December 1984 issue of Nuclear Engineering International. I thought that you and your colleagues on the Comittee might find it interesting and useful in connection with the upcoming discussion of Pressurized Themal Shock. keh.bek Demetrios L. Basdekas Electrical Engineering Instrumentation and Control Branch, DET, RES

Enclosure:

As stated cc: W. J. Dircks EDO O v A-1%

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es'ca c1c , cnuactre l

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v e r e S German Basis Safety Concept rules out i possibility of catastrophic failure e

  '          By K. Kussmaul The German Basis Safety Concept is an approach to nuclear power plant safety which allows the

[t- possibility of catastrophic failure to be completely excluded, without invoking probabilistic j arguments. The essence is that the quality of manufacturing and design alone is sumcient to assure safety, but that this is reinforced by four redundancies: multiple party testing; the " worst case"

  >          principle; in service monitoring; and validation.

e d technology must therefore be directed to development programmes were laun-

  .e         Nuclear power plants have attained                                                                   ched to demonstrate the safety margin the use of reinforced main component e          more than 2800 reactor 3 ears of sersice bodies with simple weld configurations              even under boundary conditions 24, p ,

o with,in some cases, availabihties of 80 and to the use of base and filler materials rallel to these research activities the o per cent or more. The reliability of which are not sensitive to any degrada- German nuclear industry has accepted it German rwis is outstanding For exam- the challenge of putting the principles ot tion and damage as a consequence of 3- pie. for Obngheim (345MWe), which the Basis Safety Concept into prac-began operation in 1968. bietime availa- fabrication operations, especially weld-ly tice2-s, bihty to date is 83 A per cent, while for ing and'or heat treatment8 . The Basis Safety Concept was de-s Neckar 1(855MWe)which began opera- veloped in Germany to render the tion in 1976, the corresponding figure is Development of the Basis Safety Concept V probabilistic approach unnecessary for 78 A per cent. It is also uorth noting that (BSC). From the beginning of rwa 6-technology, a very cautious approach safety cases relating to nuclear power

    -         in the last three ycers (from 1981 up to has been adopted to the prevention of              plants.The process of evaluation started V          now) the availabibr> of Neckar has been                                                              in 1972, and in 1977 the Basis Safety improved, to a value of SS per cent.              catastrophie failure. In the framework of l-                                                           a German national programme on the                 Concept was adopted in principle by the o                It could thus be shown that the safe                                                            German Reactor Safety Commission and economic operation of such plants is          integrity of components,under the lead-
   .e                                                                                                              (RSK). In 1979 it was officially pub-possible on the basis of existing regula-         ership of the MPA Stuttgart*, and with r-the panicipation of about 30 highly                lished and thus became a legal it          tions and design prinoples. Howeser, quahfied laboratories in Germany and               requirement *. Until then " burst protec-
  ;d           the optimization of nuclear power plant                                                              tion" arrangements for the pressure other countries, extensive research and ty           design and materials technology. includ.                                                             vessels, the steam generators and the 3f           ing quality assurance. control and moni-                    -- asra7         as7s2                   pressurizers of rwas were ,ander discus-toring durir.g operation, and repeated                                                               sion and even licensed for one 680MWe inspection. is an important prerequisite                     r's       'T T if safety ar.d reliabihty are to N guaran-                                       ,,,    I*            rwn. Had the customer not cancelled his
                                                                                             "         len           order, the plant would have been built teed in the future.                                                                                   with these tremendous protective mes-i530'      d'"

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                                                                             "-                                      sures. Only since the Basis Safety Con-Failure experience. Experience in the field of cota entional pressure vessel and                                    -H               so    cept became established, which piovides sufficient redundancies, have all Ger-piping technology proves that catas.               esa s                               ~

E trophic in service failures are possible 78 133o, 12x2 man reactors been licensed without such burst protection measures. and have even occurred recently in spite De improvements in manufactunns of relatively high quality requirements. . and design due to the Basis Safety which are nevertheless r.ot sufficent by lb- e43e4 Concept can clearly be demonstrated in themselves to make failure incredible. the advanced style of vessel adopted for Most of the catastrophic failures of ,,o _250 German rwas (see Fig.1). For example, components have initiated from weld-

                                                                                              '                       the effect of geometry on ultrasonics has ments where embrittlement and'or                            $

cracking are most likely to occur and

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Mj become unimportant and the total us length of welds has been shortened from may create initial conditions for poten- ,

                                                                                         --                            122 to 61m. As a consequence,expendi-tial failure, especially when corrosion                                        a**"*d                ture on non-destructive testing and risks cenanmenes attack cannot be avoidedi.                                                                            in production has been decreased All efforts in design and materials          ns.1. Conventional and advenew eryte et e    -                as M y.

Professor K. Kesseaut es Director, staathcbc A similar design evolution can be seen Materialpr6fungsanstalt (MrA). Univerut) of " state Laborstory for Testing and Matenals (MrA). Univeruty of stungan. FR Germany, with the other vessels and piping systems Stuttgart. Federal Repubbc of German). 41 December land

1- - _ _

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Design and Manuactue credible". It is the provision of such previously used restraints and snubbers. redundancies which constitute the eas-a the primary and secondary circuits'. Dese latter secondary protection de. ence of the Basis Safety Concept (BSC).

vices proved in most cases to be de- De remarks that follow referfor mainly 9 Fig 2 presents companson trimental of consen.tmnal rather than advantageous.and adsanced technology, to rwas up to 1300MWe.Conesponding steam generators, main coolant piping With the Basis Safety Concept,it is t.nd steam lines, in terms of shape and developments have taken place for awns essential to assess and quantify integnty number of ueldments. For primary up to 1300MWe, the 300MWe ran in a mechanistic / deterministic way. coolant piping.the number of welds h'as' Nes ertheless, probabilistic analysis does (SNR.300) and the 300MWe Dorium c.lso been reduced drastical!), from 250 High Temperature Reactor (mra), have a role, primarily as a tool to Safety related components and sys-in earlier plants to 60 in later ones, by advanced design and forging technolo- evaluate areas of weakness in structural tems of all plants either in operation or integrity. However, through stringent under construction in Germany were gy. Full scale verification testing under measures for all safety-related compo- l examined by legal order to see whether operational conditions gave confidence nents,in the choice of optimized mate-  ; in the extent of the impresements and they corresponded with the principles of rial and processing, design, stress analy- BSC and were replaced when they did gnabled primary and secondary piping sis, manufacture, operation, testing and not comply with the BSC philosophy. systems to be built without most of the inspection, it is possible to create the Fig. 2. Esampies of consentional and necessary prerequisites for tedundancies Significance et design. As illustrated in edsanced desigm for primar} and secondary which make catastrophic failures "in-circuit components of a m a. Fig. 3 Basis Safety is provided by optimization of design, material and manufacturing. This implies that strin-sneamma./en u _ n a _ gent and well documented process con-trol and qualification is carried out at 42, cach and every step of the manufactur-i *isoo- - as43 ing process, throughout the history of

     -                                                        e7c  - - +                                                          1       the materials, using in-process monitor-numumuur unummuumpur;

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                                                                                                                          -               through producuon pnnople).

Optimized design, according to the Basis Safety Concept, includes the fol-lowmg features: Primart carendt er e a.4eep si em e.ae.eto, N**""**** G 1.cw nominal and local stresses are cenantons aavanced achieved by using simpic smooth shapes Eto= Ecom j , _ with minimum discontinuities which 3 produce stress concentrations. T

                         "              l----         1 Ecom we s**"

V** 4 Reinforcements are located in the

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                                                                                                                $Ng                         vessel or piping wall instead of in nozzies.

j C# 4 De number of weldmentsis reduced 1763 i 3300 i O" *P'*' by using " integral design". isF - -- f Pume casm fresas.* e Lonptudinal weldments are avoided

                      ,                                                                                     v as far as possible (through use of forged 2no       _     e3443_ -42" v                                                                  rmgs and pipmg instead of formed L                                                                                                                     plates).

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g Q 9 Weldments are positioned'to be re-

                            -                       tas               4                                                  J mote from criticallocations with tespect l

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                                                                                                )                                            to mechanical and thermal loading and irradiation exposure.

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                             -                                                                                                               G Good access is provided for non-( d=:;I                      ,          %:ie, ,,,so ws.,                           geo,                .        destructive examination (defect detee-jr              g\                                                                                          l       tion, sizing and interpretation).

De implementation of these features

                                                                                        ******"**'*"' P'*aae                                   in pnmary circuit components other f                     l tL             3               conwnnons                          *ic4    aanno.o than the reactor pressure vessel is de-monstrated in Fig. 2. De optimum
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__ --h _ ps2 m design solution for the pressure vessel itself consists of combining the flange 6 ll b 5 @\k and nozzle paits into one single , smooth, highly reinforced ring with large ligs-n N _ sooo _ ments between the nozzle penetrations (j l '

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                                                                                          **"*'                  ** *3 and set-on nozzles (see right hand side of Fig.1). Note that the set.on nozzle has prpetically no load bearing function. De coaw .en ,

e c, w ec u es 7 w .w, is Neuwwees M** $ o resultirg monolithic structure is supenor to a composite structure with through-to teaptwo.n,.ees o o t ,3 Non. ,,,,c,,nm. is Neuwwas 54 wall nozzle weldments of conventional 28 ToWawa te'Waeos 21 2n townunece es o design (left band side of Fig.1). NUCLEAR ENGINEERINGINTERNATIONAL a2 A-m

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Design anc Monu aduce ral nozzle shell design has been a conunuous e-mem y,e , w e'pepan=s worsr caw

             . of 20 water reactor pressure                           o m tr evougn D'oduc"ca r"c=                 **9 8"mo*                ""S*            Q C,ue.                   P"**

Qvs supplied by a German vendor. 4 including a giant sessel 7368mm i.d. x l h ve<waterv ) 290:516mm wall thickness. As far as op,y,,,, e,,,, encoue.u anonot

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Amencan nuclear steam supply systems FaAse . w ,,nc, . ,,,cw l are concerned, some integral vessels Desti g ,,,,,y ,nc, mesuga'*" Repeatec NoE es , **'"e ' have been designed by Cocken!! SA for * * * """' Noe Westinghouse Europe and fabricated l  ; accord'ngly'. The advantages compared with the conventional design are clear. l moecenoem reounearces l

[ sas,s save>y l Whilst these sessels hase been built I I for only a few 3-loop rwa plants,i c. up 5"S **v aa***

to 1000MWe.the German nuclear steam ""***'N* I supply vendor. KWU. has chosen the 4 . f integral vessel style right from the *c"D*ty of cataurooNc l beginning of German rw a technology, a '** D'" S* ! quarter of a century ago. Oserall ex-perience has been excellent both tech- l ' nically artd economically. Fig. 3. Summary of the Basis Safety Concept i 220t to 570t in the past 20 years using and the lacredMity.et catastrophic. failure 8 adsanced forging technology . Msterials and manufacturing. A signifi- PMP 8'- cant feature in the deselopment of The use of ferritic steel instead of austenitic matenal for primary piping ris. d. hdestese esaminah eau, l reactor pressure boundary steels and dation centre at MPA. Sturtsart. The p6cten has prosed to be advantagecus because _ welding techniques for reactor compo- inductise bending of seamless ferritic shows an laside vien of the full size reactor nents has been the increase in maximum pressure vessel tBWR typel with central mast l straight pipes followed by longitudinal ingot size for the forged nngs of each manipulator. l austenitte two-lay er sinp cladding of the pressure vessel. The size has gone from N

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M, Design anci Mont>acture

bend can he performed. In piping sys- parts deliberately manufactured with tems fabricated with such advanced artificial defects, but also rejects occur-8 1

technology there is no restriction on non.desin etive examination. Before the adsent of forged piping in this

• ring in the ordmary process of manufac.

ture. MPA Stuttgan has been provided with and has investigated 1200t of such 5 2 application, plates had to be used to d ,gu'y _ material and has conducted an extensive research programme to quantify the cope with the extreme dimensions of a htm o4 [ i hau sosaj main coolant piping system. - - safety margin for lower bound con-ditions. T Another important feature affecting  %,,,, 2 materials for reactor pressure bound- memci f,% h yQ, ln investigations of the averali state of ,

1. - aries. which has accompanied the im- eaws i,i 4 toughness and cracking in base materials a pros ements in steelmakmg practice, has 4 e i e and weldments, the test material ranged M*55 "**" from high to low upper shelf energy and 2 been continuous optimization of quahty . t T

w with regard to homogeneity and iso-tropy, fracture toughness and weld-Q en

                                                                                       ~
                                                                                                     $ "_ r        "g* **  from sound material to material with cracks of vanous sizes. The toughness je          spectrum of the materials, characterized 3                       abihty. By introducing suitable materials               c- -    -

e-- by the upper shelf energy and the Z with balanced allo >ing elements and a M very low content of accompanying and ut scas  ; }' [5"" transition temperature indicating the transition from brittle to tough be-2 trace elements it has for example been i / haviour, covers all realistic possibilities. d possible to control the phenomenon of  ! g c m man A third redundancy in the Basis Safety stress rehef embnttlement and cracking re= ,w , m manwas d7 in the heat-affected zone (HAZ) of  ! - L - 1 -- Concept consists of continuous in-ser-vice verification that design conditions 3 weldments. Narrow gap w elding reduces " _ are not exceeded during plant operation a the sensitive coarse grained areas ir the (" continuous in-service monitoring and

--!                     HAZ of weldments and the weld                           . j,        d                               documentation" principle). This 3-volume. Thus a high degree of purity and a limitation on segregations is                        N  b' p- %
                                                                                 ' ge                                       guarantees that the design is conserva-obtained which is one decisive pre.                                          J       '

tive in relation to the service conditions requisite for the performance of the [ -

                                                                                                     '.)     f;             it encounters.

Because of the special significance materials. To cope withthe growing ambitions of attached to non-destructive testing tech-the nuclear industry in terms of the size, niques, considerable interest is centred J shape and strength of the components. Fig. 5. Data processing end management at the on this subject. Advanced non-destruc-g tive testing techniques such as Alok l the Shape w elding process for all w elded Stuttgart MPA non-destructhe esamination tahdation centre, as used in cordunction with (amplitude time of flight locus curves) 1  : nuc! car and other high quality compo-nents has been successfully introduced. the fuli sine rtactor pressure vessel and central will proside improved flaw ' detection 15 mass manipulator. and sizing in future; holography, Saft j , Regardmg stainless steel, the problem of integranular (stress assisted) corro- (synthetic aperture focussing techique) g sion has been overcome by using stabil- Independent redundancies. He Basis and phased arrays represent the next J i Safety Concept requires the existence of generation of equipment in analytical J ized austenitic material. ne hot cracking phenomenon in enough tedundancies to ensure that any systems. The systematic coIIection. codi- _ l austenitic weld metal is controlled by possible departure from optimization is fication and evaluation of operation

              !           using an optimized chemical composi-          sufficiently unimportant and unhkely for             experience is a pow e rful tool for increas-
    "a '                  tion. The susceptibility of ineonel 600 in    catastrophic failure to remain incredible            ing the availability and safety of nuclear 9I                  steam generator tubes was the reason for      (see Fig. 3). The essence of the BSC is, however, that the manufacturing pro-facilities.

As w ell as the reactor pressure vessel, the change to Incoloy 800 as the steam ] t cess alone suffices to gise assurance the testing of steam generator tubes has a? generator tube material in German against catastrophic failure, and that the become an issue of specialimportance.

   =;$                     nuclear plants. The operating expen-ence with this matenal confirms the          redundancies reinforce that assurance.               The automation of testing and the effectiseness of this choice of material.        The first redundancy is provided by               application of multi frequency eddy cur-The success of that development was       having the initial component integrity                rent techniques have brought about a i

venfied through extensise research and venfied by a number of independent significant reduction in inspection times developmen' programmes" as well as quahty assurance programmes. His and in radiation exp35ures received by through full scale procedure qualifica- constitutes the principle of " multiple personnel. j tion tests. Against that background, party testing".

      ~'

effective quality control becomes feasi- A second and crucial redundancy is Validation. De fourth tedundancy is the ble, practical and economical, because furnished by the demonstration that " validation principle" which relies on remaining small defects are of minor there is a worst state below which the verification and validation of calculation -Pb , importance. condition of the material cannot fall; this codes and mechanistic fracture mechs

=Ej                 ;

is the " worst case" principle. The neces- nics, and probabilistic evaluatic.: Of Nevertheless,in-process inspection at different stages of fabrication (provision sary assurance is provided by ongoing non-destructive examination methuds. _4M> of optimum conditions for inspection research and development work: defect Defects resulting from deviations in from both inside and outside surfaces and failure investigations performed by which production processes are never-before and after cladding) remains im- industry and independent laboratories theless within the presenbed manufac-portant and has to be performed with the to establish the mechanism and possible turing steps may pass undetected during

     --C                    best possible technology. The best way        extent of degradation. To assess the                  manufacture. Derefore the validation worst case performance of materialit is               of non-destructive examination is, with-9                           to accomplish this is to use automated essential to have available, not only                 out doubt, the key issue.

inspection techniques, as used in service. 44 NUCLEAm ENGINEEmlNGINTERNATIONAL A-d O ,

C Ngn and Mcnu octure Fig. 6. View of some of tk test h a d== at mens which reproduce actual wall thick- MP A Stuttaart. The tan mactAe to the right is "ecause of the complex weldmg nesses and the entire range of toughness. etries used in cons entional compo. the le ooot flooMN) salversal testles The specimens are being tested in a high machine. und it is tM upper head of a fus

           . design, inspection procedures re-4 ire the utmost care. There is no better way to salidate and to quantify the pressure loop (see Fig. 8). Further pressurized thermal shock studies are
                                                                                                                                          ',8[8 8*
                                                                                                                                                           *"hP"**{
                                                                                                                                             ,, ,, gg y,, ,g gg ,,,,

f being performed on a full scale sessel capabihty of remote testing techmques under operational conditions to produce for in sersice inspection than by usmg full scale components, see Figs 4 and 5. ,y., Such components must contam all Dp ; ], i ..i , kinds of natural and artificial flaws m i j  ;- f Yi! .I i sufficient number and at representatise

                                                                                                                   +"

91"""""!!!H!!IR Elll locations. The complete automated m- - g"-H * ' M l N Dl T spectinn sptem must be sahdated. ^ " ~ iM r Data gathering, data storage and data .- ii processmg for flaw detection. flaw sizing and flaw interpretation require the most b zu U "I Itw'E i T 3 M,",.

                                                                                                                                                                                        ,<                ~

s aosanced computer technology and a BE g i y- J "*

• deep understanding of all defect phe-nomena related to fabncation and to

                                                                  -                        ,.           - .       i; =.g*ht                                                                                         g         lp kI.

mechanical and enuronmental con- 3.* ~4 y -e dicions encountered durmg plant opera- pF e ' "' s E tion. The full scale sahdation tnals g* i i ' M r y , mentioned abose must melude all the r g electronics and data processmg systems (Fig. 5) as well as the traimng pro-

                                                                 }

94' a==

                                                                                                                      =g .
                                                                                                                                                                                                                              ;4 grammes used for personnel.                                                                                                                                                                                             I There is at MPA Stuttgart a full scale 900MWe swa reactor pressure sessel.

j p -

                                                                                        ]giygg                a;
                                                                                                                                                                                                                               ]

which is the heart of the non-destructise

                                                                                                              ]         (*.                   ,

casmmation sahdation centre. Other sery large facihties at MPA '

         uttgart for sahdation include: a
                                                                            ~...

l .; (100MN) unisersal testing .g O*n.000tachinc (see Fig 61) for the sahdation

• of tracture mechames(includmg pressu- - <

                                                                                              -~

mn 3 rized thermal shock loading), a high rate -*  ; tensile testmg machme (under construc-tion.see Fig 7) with a capacity of 12MN. . , - ,,. ii 14m in length and 400t in weight. - . ..W "i ' l'. operating with gunpowder as a pro- -ME- j pellant to simulate the impacts from g _

                                                                                                                            **                       g'                                                                            :.

hazards such as water hammer, earth- - e

                                                                                                                                                                                                                                   .i quake and aircraft crash on the primary                                       -                                                                                                                                      f coolant circuit and safety protection                                              c=r=ms                 umumac                         6 m mn.J                                                          1             ,

systems as well as energy excursion esents in tsN a pipir g test facihty fo-computer code sahdation (extestial f ~ s

                                                                                                                                                                                                                                 .c Q             -

ma ~ - t , haz.trds). and a set of autoclase systems - to simulate water chemistry and cyche . loading, temperature and pressure. The large scale specimens used for

                                                                                                                                                                                           ' " -                   A studying failure behaviour range up to                                                                                                                                                        MM 910mm thwiness for round bars and CT Sio (compact tertson test) specimens                                                                     -b QQ .,                                                           J i

t respectisely, and 800mm o.d. for hollow

• a cylmders with a 200mm wa!!. Also clawified as large specimens are plates

                                                                                                                                                                                    ----M "E                              q up to 1250mm wide and 250rnm thick.
                                      ,                                                                                                     e                               ~

There are m addition mtermediate _ size vewels, with diameters of about 2000 to 3000mm and walls up g to 150mm thickness. The object of the x tests on the sessels and pipes ts to into .he - - - { .p-}lT.'Tf ' * ,

                                                                                                                                                             ""                           cr g-
!O    givefuture  a fundamental         insightretam-behasiour of pressure i                                                                                                                                                                                                                   $.

thermal shock experp w n -- A e sur e , ment is currently under way usmg speu- 45 Decemoor 1964 Am - . . _ . -_ .. w-n

g Designand Manuaduce l , Acknowledgements.The outhor fully acknow. ledges the sponsorship of the research and developmeni work by the Federal Mmister of Research and Technology and the Federal

  -

• I Mmister of the Intenor of the Federal Republic of Germany as well as by the au-lear l1 1

4 industry. Special acknostedgement is due to Kraftwerk Union AG(KWI)),BrownBoveri 4

                                               ,,'                                                                                    I         Reaktor GmbH (BBR). Brown Boveri Hoch-
  "                                                                                                   #                                        temperatur.Reaktorbau GmbH (HRB). In-
                                                                       .i m             a teratom Internationale Atomreaktorbau b -e

• N MPA GmbH, M.A N. GHH Sterkrade, Mannes.

                                                                    "" %                    J
                                                                                                                     'N mann Anlagenbau GmbH, Klocknerwerke AG. De Japan Steel Works Ltd and
                                      '.i 3

[- @ Thyssen AG for contmuous co-operation and for providing pnvate company information.

               .          .                                       - ,, mW
               .                                               pummmew -
                                                                       '            .o '                   -

a h.--. Thanks are also due to colleagues at the

                                       , p. 7E' J[
                                                                            - '                                                                 German Reactor Safety Commission and in

{1 i the international nuclear community for their 7 i a e  % p column unfashng support and encouragement. tra

                                                     % we.ne       ene              g               s
                            %aste pas            $0eoWn     Pis'en       comDwsbolt aHi&*iper Guide                                             g ,g,,,,c,,

mae A wssei gnp sod 1. K. Kusasreul and D. Sand, Sasis Safety- a challenge to nuCbear technology (peper reed at r IAEA Speciahsts Meetmg Madnd. 54 March I l Fig. 7. High espacity dynamic tensile testing 1979) m R. W. Nichots (ed). Trends 6n reactor prusure wusel and circuit development. Ap. f E

       ~

machine, which uses gunpowder as the pied Scence Publehers. London 1980, pp. propellant. The machine, which is under preesse control vehe 1 13- -

• construction =i!! proside controlled lood-time -M ~

2. Invited papers at the seventh MPA seminarin histories, for use in studies on a ster hammer,  ! Eiserncal sp,,g 100m8 w es.sm ,,,,c, N ares "Sa% of me prusum boundan d earthquake and aircraft crash impacts. \, egee,,sw, heating ig , hgnt water reactors": Elastic-plastic fracture A I N J concepts. Nuclear Engmeenng and Design,72,

                                                                                                        ).        ' CP 1982. No.1, special asue, edited by K. Kusemaut.

thermal fatigue nozzle corner cracking. + n,n- 3. Invited papers at the e ghth MPA commar 6n In addition an acoustic emission trial, i the senes " Safety of trw pmuure boundary of

                                                                                                        \'
                                                                                                           \                4=                    "0"' **'*' '****~; F'*cture behav6our and usinE an intermediate size vesselis under way. This test,in which a loop ts being LowpM'N, pu"V g,,,,,,,,,

nondestructive testmg. Nuclear Engmeeting

                                                                                                                                                  ,na on,on.76.1983,No 3.specelieve.edded used to create reactor pressure vessel                                          evneson numps                              by K. Kuumaut.

service conditions,is part of an interna-

• d '"'"*d D*D*4 *' * *"'h "' A **"""*' '"

tional co operative effort. Acoustic emission has the potentialior adding a Tis. 8. Test loop need for pressurised thermal aback trials. Simulates in operational condi.

                                                                                                                                                         , , , "8*M.'.7,
                                                                                                                                                  ,,, m ,n,y,,,p,punn,n.a.

P,'."*,",",  % further important Une of defence to the noas with high pressere injwtina of cold underlying safety philosophy. water. F , = m bar. Tennperature = $. S. Onodera. K. Fupete. M. Teukada, and K. As well l 320"C.is Landed by it0MN temalie seeslag Suzuki. use or a more mregrar type. serger sire currently vohed m.as m. these . activities. projects at four MPA machin. efeet forgmg in pnmary components and car-installationi oli sites other than its own. curts as inreded for hsghe* structural mfegrtry These validation tests, for LwRs, 71BRs in R. W. Nichofs (ed). Trends m ructor preuure vusel end esrcud developnwnt Appled Scence and HTGRs, are going on at four loca- """***"A""*"*"**"" tions: lacredibilit) of failure. Basis Safety and 4 Mannheim - a very large under- the four independent redundancies form e as< oundermes for Prnsunrod Water Mosc-

             -          ground facility to validate the concept of              the Basis Safety Concept which allows                              tors 3rd Edition. October. test (ed,ted by leak before catastrophic failure.                       the possibihty of catastrophic failure to                          gIuschan tur Reaktorsietween faRS) mbH.

O Meppen - an army test facility for be excluded by a deterministic /mecha-high energy burst tests. nistic analysis. This is the principle of 7. K. Ku smout. Developments in avei ar p.J preuure wwwi and circuit technoiogy in ite i G Grosswelzheim - a small decommis- " incredibility of catastrophic failure" 'n-6 pc { sioned light , water power teactor Note that this exclusion does not invoke ,h"*D%l,'("",*" g < (HDR). This ts used for blow dow7 and probabilistic arguments. It ts posstble to Stanikopf and L. H. Larsson (eds). Structurei ' s external hazards investigations on the see, without detailed calculation and integnty of hght water ructor components, despite human errors and human fac. Apphed Semence Put>tstwrs. London 1982. pressure circuit, and for studies on-thermally induced nozzle corner crack tors, that failure probabilities are so low .,,,,,,,,,,,,,,,,,,,,,,,,,,,,g,, a growth and pressurized thermal shock m as to be meaningless. preuvre vesset newes and correspond,ng the reactor pressure vessel. According to the Basis Safety Con- stren swysis Asme Trurd intemetenet con-3 forence on Pressure Vesset Technology, Part i, G Kahl - a small operating Lwa for cept, the safety margin between the Tokyo. aspan u77.

   ]                     investigating the interaction of irradia-               loads encountered, even under severe tion, cyclic loading and coolant corro-                 conditions, and the reduced load car-                              9. K. Kusemeut,0. sturm, w. seeppier, end D.
             .n          sion.                                                   rying capacity which might be experi-                              Moner.gener. rnperimentar dnveer, gar,on en the enced under degraded conditions can be                             crack openme behaviour of cyfmancer vessets g                 All the MPA work is supported by a                                                                                                                                     eo q,           large number of laboratones and spe-                     shown to be large enough to justify the                            Q','8l' *[*y'l*ct
                                                                                                                                                                 ,                    r se p                     j calist research orgaruzattons through-                   assumption of incredibility of catas-j l

(PvP voi s2).the Amencan societr ot Mechan 6c-out the Western World. trophicfailure. O at Engineers. New York,1982. pp 97134. 46 NUCLEAR ENGINEERINGINTERNATIONAL i , l 4 A-N - - - -

              -                                                    O.     ' .S   ! nims  n-   ..v-..-.

APPENDIX IX ( D. L. BASDEKAS PRESENTATION ON PTS

                                                                     -q <                   1 -g. t
                                                                                                ~

h{ .. .~ o, , 4 UNITEt$ STATES g(?,5 N j NUCLE AR REGUL ATORY COMMISSION f 7 i.a y wasniwetos.o. c. rosss y(h (s.g/...* . MAY 2 81980

                                                     \
                                                      \

The Honerable Morris K. Udall Chairman, Subcommittee on the Energy and the Environment Committee on Interior and Insular Affairs United States House of Representatives , Washington, DC 20515

Dear Mr. Chairman:

In response to your letter of February 7, 1980, regarding reactor control system failures, which could lead to accident sequences not previously anticipated in Commission regulations, the Commission transmitted to you its position on this matter on May 14, 1980. O V The purpose of this letter is to briefly comment on the Commission's response, clarify my bases for differing with the Commission's official position on this important matter, and reiterate on the record my long-standing concerns, the reasons for them, and a proposed interim solution that I believe is prudent, practical, and long overdue. I agree with Commissioners Gilinsky and Bradford that the Commission's response of May 14, 1980, down plays a serious problem. I also agree with them that a better discussion of the safety implications of control systems is provided in the October 22, 1979 memorandum from Mr. Denton. However, although Mr. Denton's discussion is better than that given by the Commission, the position he has taken with respect to the need for Failure Modes and Effects Analysis (FMEA), and the interim derating of all operating nuclear power plants is not correct. The Commission's response is in essence that given to the ACRS and the Congress in 1976-772 reflecting a standard position that "We are aware of the problem, we have no fundamental disagreement on the need to learn more about it, we are I working on it, and no regulatory action is necessary." It is largely premised l on the faulty assumption that " safety systems will mitigate control system failures at any power." This illusion appeared to have dissipated quickly in the aftermath of the THI accident and several near accident events (Rancho Seco, Oconee, and Crystal River). Surprisingly enough, however, it is surfacing again as an official Commission position despite the ove mhelming technical evidence that it is wrong. It is very much like someone ahancing the argument l that "You should not be overly concerned about the function.il integrity of the p v steeringmechanism of your automobile because even if it g..is, at any speed, ACRS0mCEC0?Y% - ma Jo Xoi Remove Trom AS .u m ce - A-19A

l-( ( l 2 E 20 The Honorable Morris K. Udall you have, a seat belt to protect you." Does the Commission's statement on the safety implications of control systems imply that Class 9 accident initiator sequences are to be brushed aside again, or be given a cosmetic, prolonged lip service? The Kemeny Commission Technical Staff Analysis Report on the Failure of the Pilot-Operated Relief Valve (PORV), a non-safety grade control system component,2 concluded that:

                 "The TMI-2 accident would probably not have progressed beyond a severe feedwater transient, had the PORV been recognized and treated as a safety-related component."                                            .

Furthermore, on page 35 of the same report it is stated that:

                 "This PORV failure is a clear indication of the need for better configur-ation control and interface coordination. It provides a good opportunity for a failure mode and effects analysis (FMEA)."

The Commission's response cites examples of control system failures in Crystal River, and the dismissed potential of " control system failures leading to unacceptable consequences" in all plants caused by high energy line breaks. O These are just two more examples of the " band-aid," reactive approach to safety, that has been rather characteristic of the regulatory process.8 Even in the traditional " band-aid" approach, the Commission has failed to address, for instance, the effects of control system failures in operating plants due to an earthquake. Control systems are not seismically qualified, and an earthouake could cause massive failures driving the plant to extremely unsafe

;          conditions. Isn't the Commission and its staff concerned about that? The hard reality remains that the most likely way to find out what is wrong with the design, configuration and qualification of control systems is for something wrong to happen. Certainly this is not the correct approach to safety.

It appears that there is a growing tendency within the industry and the NRC to not only forget or down play the deficiencies brought out by the TMI accident, but to view this accident as a confirmation that "The system worked; nobody was killed." It is true that nobody was killed, but it is clear that the system did not work; unless one of its unwritten design bases was to bring us

   -        within 30 minutes of disaster, cause damages and losses exceeding the cost of the plant, and substantially disrupt life in nearby communities. No, the i            system did not work except in the sense of, hopefully, driving the point home that we must recognize and correct the deficiencies in nuclear power plants s   before it is too late.       '

The matter of poorly designed and installed control systems and their arbitrary classification as "non-safety" systems are not the only things that require prompt and resolute attention. The following is a summary of points I have t t

• c P 44 . r O i ** t'. o ia r t $ii tr t ta the problem of deteriorating safety, its reasons, and an interim solution to l l

it. l l

                        -    __                   _ -._twA-              ___              ___

( (_ The Honorable Morris *K. Udall 3 NM 2 8 1900 The safety of operating nuclear power plants has been deteriorating since the The reasons for this irony relate directly to the

• l Three Mile Island accident.

          " corrective actions" taken by the Commission prompted by the TMI accident.

Their net contribution to safety, at an already unacceptable level before the TMI accident, has been negative because of:

1. Fragmented and haphazard changes in design and procedures without proper, if any, consideration of their effects on overall safety.

2. Attempts to compensate for design deficiencies by instituting more and more procedures for the already overburdened operator.

3. Lack of basic understanding of plant dynamic characteristics and the  ;

importance to safety of the so-called "non-safety" control' systems, which continue to go unreviewed.

4. Continuing improper use of reliability and risk assessment methods, and repetition of mistakes made during and after the study reported in WASH-1400.

Sound and tested engineering methods and practices that have served the aerospace and defense industrias well, known as Failure Mode and Effects Analyses (FMEA) are indispensible prerequisites for any meaningful reliability and risk assessment.s.s Nonetheless, NRC shies away from them as if it feared what they might uncover. An example of what an FMEA should uncover O V is given in Reference 7, and it could prevent a catastrophe.a Even a preliminary analysis of this matter may very well show the need to tempor-

~

arily shutdown many plants. The interim solution to these and pre-existing problems is to derate all operating nuclear power plants to about 65% of full power until their dynamic characteristics are established and well understood, proper and meaningful reviews of their control and other so-called "non-safety" systems are performed, and their results evaluated. During this time of two to three years, a thorough review of some 130 unresolved safety issues, along with the " corrective actions," taken since TMI should be performed outside the NRC. The proposed derating will substantially enhance safety, while producing a minimal impact on integral power generation, due to adjustments in the refueling schedules that will be possible (see Figure 1). Power generation may be aug-mented by issuing where appropriate, derated operation licenses to a limited l 4 number of new plants already completed. Not only can we afford to implement this interim solution, we can't afford not Failure to decidedly reverse the present trend will most likely result in a catastrophic nuclear ageidents within the next two years or so; and with it to. a denial of the nuclear option to this country, even as a "last. resort" component of our energy supply. It will not be possible then to undo the hars, and too

• late to save the nuclear option.

 'O                                             .

A- M5

(. 4 UAY 2 8 IE' The Honorable Morris k. Udall el and This letter has not been concurred by any supervisory 0.735-55, Annex A. Commission per it is sub'mitted under the provisions of 10 CFR Part The references cited are \ available*from NRC or subcomitte If I can be of further assistance, please let me know. Respectfully, A L.Bsd L . b Demetrios L. Basdekas ., Reactor Safety Engineer cc: Rep. Stephen Symms , e 1 a J A-1% -___ - _ . _ . .-

(, ( 5 MM 2 8 1980 The Honorable Morris K. Udall Notes and

References:

1. Letter from M. Bender, ACRS to M. Rowden, Iscue No. 22, Report on Selected Safety Issues Related to Light Water Reactors-Issues 16-27, dated February 23, 1977.

2. Technical Staff Analysis Report on Pilot Operated Relief Valve (PORV) Design and Performance to the President's Commission on the Accident at Three Mile Island, dated October 31, 1979. Memorandum from D. L. Basdekas to J. F. Ahearne, Safety Implications of 3. Control Systems and Plant Dyanmics, dated October 25, 1979.

                                                                                                              ~

L. Basdekas to J. F. Ahearne, Safety Implications of

4. Memorandum from D.'

Control Systems and Plant Dynamics Recommendation to Derate Operating Plants to 65 Percent of Rated Power, dated December 20, 1979.

5. Institute of Electrical and Electronics Engineers (IEEE) Standard 352-1972-75

                  " General Principles for Reliability Analysis of Nuclear Power Generating Station Protection Systus."

6. Memorandum from D. L. Basdekas to the Commissioners, Review of Uses of WASH-1400 in Regulatory Decision Making Identification of Unresolved Safety Issues, and Report to the Congress, NUREG-0510, dated February 14, 1979.

7. Memorandum from D. L. Basdekas, to J. Murley, Failure of Main Feedwater Control System Resulting in Unacceptable Overcooling of Reactor Vessel, dated February 27, 1980.

8. It is likely that a loss-of-control sequence of unprotected events will involve the failure of the reactor vesse17 and/or steam generator tubes, resulting from malfunctions in the feedwater/ steam cortrol systems, and/or main steam line breaks.

J s O

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APPENDIX X REPORT TO THE ACRS ON NRC STANDARD REVIEW PLAN FOR DRAFT ENVIRONMENTAL ASSESSMENTS O REPORT TO THE ACRS ON NRC STANDARD REVIEW PLAN FOR DRAFT ENVIRONMENTAL ASSESSMENTS BY ACRS WASTE MANAGEMENT SUBCOMMITTEE FEBRUARY 9, 1985

1. General Coments:

a. Overall, the NRC Staff has done an excellent job in developing a generally well prepared report under severe time constraints. Their organization and leadership are

             , good, and the working plans are sound,

b. We believe the NRC Staff should state clearly and con-cisely(that their primary stage 1) serious goals flaws in any of are to identify the sites at this that would cause them not to be acceptable, and (2) any safety issues that may not have been identified by DOE. We support the Staff's position that they should not become involved in the accuracy of the ranking by 00E of the several sites. NRC's prime position should be (as O

stated) to assure that DOE properly applied its Siting Guidelines in evaluating the acceptability of each parameter involved in making its selections.

c. Quality assurance requirements as given in the NRC Standard Review Plan seem inadequate for supporting the Staff's activities in conducting these reviews. 10 CFR 60, Subpart G, Quality Assurance, paragraph 60.152, requires DOE to implement in its HLW repository activities a quality assurance program based on the criteria of Appendix B of 10 CFR 50, and the SRP should reflect this fact in Section 9.0, page 18. It would also seem appropriate to incorporate the requirements of Regulatory Guide 1.28, " Quality Assurance Program Requirements." Although this Guide was prepared primarily for use in the design and construction of nuclear power plants, it would appear useful in helping to implement the quality assurance criteria of Appendix B of 10 CFR 50, both to DOE site evaluation and to NRC site

' reviews. The Subcommittee acknowledges, however, that we have not reviewed the NRC Staff Technical Positions on QA.

2. Specific Coments:

Page 5, Section 3.1.1, next to last sentence -- the potential v safety issues should be limited to those that DOE had over-looked. A - l*lal _ _ _ -

STANDARD REVIEW PLAN 2 < Page 5, Section 3.1.2 -- the wording of this section appeared to be unduly adversarial. Page 6, Section 3.1.3 -- it would be helpful if the specific subject areas of NRC's "special expertise," presumably not available within DOE, could be identified in this report or elsewhere. Page 7, Section 3.3 -- the second and third paragraphs of this section were not clear to us. Perhaps others reading the document will have similar problems. Page 8, Section 3.5 -- if the main basis under which the NRC Staff plans to offer comments is NEPA, this should be stated. Page 9, upper portion -- another question to be asked is whether the data are adequate and sufficient. Decisions on items (1) and (5) appear to be dependent on input from NRC's "special expertise." Page 9, Section 4.3 last sentence -- it would be better to state that major issues will be given priority. Page 9, Section 5.0 -- considering the work load, we recomend that the NRC Staff consider limiting its reviews to only the 3

or 5 sites considered best by DOE. This is particularly true in the case of the draft EAs.

l Page 10, upper portion -- we recommend that the NRC Staff consider taking advantage of the report on " Waste Isolation Systems," recently published by the National Academy of Sci-t ences. This report includes, for example, the scenarios they , used in analyzing the acceptability of the several proposed sites. Page 12. Section 1.3, first sentence -- we do not understand why the development of a site-related data inventory should be discretionary. Also, under item (2), we suggest that consid-eration be given to the adequacy, as well as the amount, of data. Page 13, Section 6.3, first paragraph -- it would appear that the NRC Staff would already have an " initial perspective on  ! which processes and parameters are most sensitive and there-  ! fore critical to repository performance...." prior to their analyses. Might this~have said that "The objective is to amplify and/or confirm an initial perspective ....."? Page 14, Section 6.6 -- it would appear to us that a scoping would be mandatory. Detailed follow-up reviews might be l performed "as time and resources permit." O Page 16 item 4. -- are the " problems" for which resolutions l are to be suggested of a " procedural" or a " safety" nature. 1 _. _. b _ _EC)O_

STANDARD REVIEW PLAN 3 If they are the latter, and the approach suggested by the NRC (d) Staff does not correct them, might this not lead to major controversies? We recommend that the NRC Staff avoid providing suggested solutions to such problems. Page 18, Section 9.0, third paragraph -- the modus operandi suggested here appear inconsistent with acceptable QA prac-tices. The person responsible for a given item of work should not also sign off on the acceptability of its QA review. Page 22, item #4; page 23; item #7 -- These items appear inconsistent in that they would require NRC review of the DOE ranking and selection of sites. Page 24, item #12 -- in an earlier portion of this report, the statement was made that the NRC Staff would only connent on the repository design impacts and any related safety issues. This item appears to contradict that statement. Page 25, Table 3 -- although not absolutely necessary, it would make this Table easier to follow if the specific Tasks were placed in chronological order. Otherwise, they do not coincide with the outline presented on page 10. Page 27, Table 5 -- we note that Section 3.0 omits Chapter 7 and Appendix B. They should bs included. O A _z.ot

l I l

O V

REPORT TO THE ACRS ON NRC DRAFT GENERIC TECHNICAL POSITION ON LICENSING ASSESSMENT METHODOLOGY FOR HIGH-LEVEL WASTE GEOLOGIC REPOSITORIES BY ACRS WASTE MANAGEMENT SUBCOMMITTEE FEBRUARY 9, 1985 General Comments:

 ;      1. This report can serve as a useful supporting document to the
             " Standard Review Plan for NRC Review of Draft Environmental Assess-ments for High-Level Waste Repositories." It reflects considerable work and effort on the part of the NRC Staff.

2. We note that the subject of occupational doses is not addressed.

As a minimum we believe the report should include a statement that the requiremonts of 10 CFR 20 will apply. We also note (page 5, Section 1.2.2.1) that the definition of systems and components important to safety involves an accident which could cause a dose

             " commitment", greater than 0.5 rem. We believe that the word
             " commitment," should be deleted, as the reference 10 CFR 60.2 is improperly quoted. The statement, as given in Section 3.3.1, page Os        18, is properly worded. In this case, however, we do not understand the emphasis on " radioactive material in air."

3. We are pleased to note (Section 1.2.2.1. part (2), page 6) that the 3

NRC Staff will accept a lower subsystem performance goal, provided that DOE can demonstrate with reasonable assurance that the overall system meets the EPA standard. We concur with this approach. I 4 Although we endorse (Section 1.2.2.2, page 7) the application of Probabilistic Risk Assessment techniques to evaluations of HLW repositories, we believe it is important that the NRC Staff recog-nize that such an approach may favor the use within repositories of materials, components and systems for which PRA input data and models exist. This could have economical as well as technical implications.

5. Section 1.2.3, page 8, implies that the Atomic Safety and Licensing i

Board routinely becomes involved in a full range of licensing issues, and will examine all of the technically defensible material related to a judgment that the public health and safety are being protected. However, the attention of the ASLB is almost always 1 confined to items contested by intervenors.

6. In Section 2.0, page 9, item (1) states that DOE shall " designate component performance requirements for the~ geologic repository O system and establish criteria for demonstrating compliance with the requirements." It was our understanding that the EPA had the i responsibility for establishing standards for the acceptable

                .            . _             . A - Lo2.

LICENSING ASSESSMENT METHODOLOGY 2 p performance of a repository and that the NRC had the responsibility

  \j           for developing criteria which, if complied with, would assure that EPA's standards were met. The same coment applies to item (4) on
                      ~

this page.

7. In view of che low off-site dose rate that represents a failure important to safety, we are not sure we understand how " safety" assessments and " environmental" assessments for a repository are to be separated (Page 13, last paragraph). Perhaps some form of clarification would be helpful.

8. Although Section 3.2, pages 14 and 15, covers "Perfortrance Assessment Methodology," inadequate attention appears to have been directed to assuring that the quality of the data generated will be l adequate as input to the mathematical models being used in repos-itory evaluations. In addition, we note what appears to be a lack of attention to providing details on the assumptions used in developing these models.

Specific Comments: In line 8, page 14, the statement is made that "All existing and 1. future conditions and processes" must be described. Should not this perhaps say "All the relevant, or important existing and i future conditions...."? ) O V

2. In part (2), Section 3.2, page 15, we are not sure what is meant by the word, " environment." Would the phrase, " repository environment," be better?

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O w/ REPORT TO THE ACRS ON NRC DRAFT GENERIC TECHNICAL POSITION ON IN-SITU TESTING DURING SITE CHARACTERIZATION FOR HIGH-LEVEL NUCLEAR WASTE REPOSITORIES BY ACRS WASTE MANAGEMENT SUBCOMMITTEE FEBRUARY 9, 1985 General Comments:

1. This position paper appears to highlight how and where tests are to be done, as contrasted to the parameters to be measured. The Subcommittee believes that the latter should be emphasized.

2. This document develops at length the regulatory background and technical reasons for testing rocks and groundwater in place, as opposed to laboratory tests. It does not attempt to develop the strategy or wisdom of testing programs at particular kinds of sites, and this necessarily limits its usefulness. It points out the need for planning of tests in advance of exploration shaft construction. The Subconnittee believes that flexibility will be required to meet the problems discovered during site charac-terization.

3. Our review suggests that the technical position taken here may seem to require costly, time-consuming and comprehensive testing beyond what is needed and appropriate for a particular site. We hope that attention will be focused on the critical questions to be explored at each site and the best tests to answer these questions.

4 3 situ testing during site characterization is properly part of a continuum of measurements and studies before and after site charac-terization. While critically important, the in situ testing should be integrated with these other studies.

5. Thought should be given to assuring the long term durability of instruments used for the conduct of tests under the conditions of elevated temperature, humidity, corrosiveness, etc. of repository environments.

6. It should be recognized that in situ testing is frequently done under conditions, particularly with respect to oxidation, that will be different from those existing after repository closure. This position paper should include consideration of this fact.

7. The most rapid changes in rock stress, in water levels and in equilibrium conditions, occur when the rock is first penetrated.

Consequently, M situ tests should begin as soon as feasible. A-fao9

IN-SITU TESTING 2 Specific Comments:

1. In line 5, second paragraph, Section 1.0, page 5 -- the definition of "in situ testing" as performed solely from underground openings

' is unnecessarily restrictive. Inserting "mainly" after " performed" would cover surface-to-shaft seismic measurements, for example. Although this is partially explained by the paragraph that follows item 8 at the top of page 12, it would have been helpful to have this statement expressed earlier in the report.

2. In Section 3.2.1, D, pages 13 and 14 -- geochemical tests are mentioned. It should be made clear that chemical analyses of rocks j and groundwater need not all be carried out underground.

i 1 3 4 lO i 4 0 O A-zos

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      )

REPORT TO THE ACRS ON DRAFT TECHNICAL POSITION ON BOREHOLE AND SHAFT SEALING OF HIGH-LEVEL NUCLEAR WASTE REPOSITORIES BY ACRS WASTE MANAGEMENT SUBCOMMITTEE FEBRUARY 9, 1985 General Comments:

1. In some ways, this report is too complete in that specifications are outlined for such a wide variety of factors that the reader is left with no concept as to their relative importance. Although we

' realize that this may be difficult to accomplish in a generic report, we recomend that consideration be given to highlighting the more important considerations.

2. The statement is made in several portions of the report (Section 1.0, page 1, and Section 6.0, page 17) that the NRC will specify minimum design measures for the repository. We thought that the main purpose of the NRC was to specify performance criteria and that the design of the repository would be left to the DOE.

(~ 3. The EPA standards for a repository are referred to in the report in

 \                           several different ways Forexample,theabstract(page1) states that the standard places a limit on the " amounts" of radionuclides that can accumulate at the accessible environment; Section 2.1 (page 2) refers to limits on the " cumulative release" of radionuclides into the environment; the first item at the top of page 17 refers to EPA limits for "radionuclide releases" from a repository. We believe these statements should be made consistent with the EPA standard.

Specific Comments:

1. Page 1, Abstract -- the last sentence of the first paragraph is repeated in the first sentence of the second paragraph. We suggest the former be deleted.

2. Section 2.1, page 2, first paragraph, ifne 3 -- the word " dis-posal," has a typographical error. In the third paragraph of this Section, the introductory phrase to the second sentence is mislead-ing. It is the repository and DOE that must meet the EPA standard; ~

not the NRC.

3. Section 2.2.1, page 3 -- it would be helpful if the phrase, "in non-nuclear facilities," were added at the end of the first sen-tence.

4 Page 3, next to last line -- should this read, "In particular, they refer to the ....."

                                                       . _ _ _ _    _     -204_    _ . _ _ _ _ _ . _   ___

BOREHOLE & SHAFT SEALING 2

5. Page 4, item (c) near the bottom of the page -- this might be

'" )      reworded as follows:
          "(c) Test sections shall be established to evaluate the effective-ness of borehole and shaft seals before they are sealed."

C. Section 2.2.4, page 6 -- if part (d), (1) of 10 CFR 60.10 was omitted, it would be helpful if this were indicated.

7. Section 3.0, page 7 -- we suggest the first sentence be reworded as follows: "The following are issues ....... by the NRC as most critical in terms of repository performances."

8. Section 3.1, page 7 -- the first paragraph could be considerably shortened. The current version is repetitive.

9. Page 8 -- in the first item at the top of the page, the phrase, "Long-term," is defined as being "3 to 5 years." Since this is so different than what is comonly assumed in the report, we recommend that a different phrase be used or clarification be provided.

10. Section 3.2, page 8 -- the second sentence in the first paragraph needs editing.

11. Page.9 -- the second item at the top of the page neglects the fact p) that shafts in some repositories will not have liners.

O

12. Page 9 -- the fourth item at the top of the page might include acknowledgement of the fact that, for a repository in salt, you expect the rock to relax.

13. Page 10, Section 3.5, next to last item at the bottom of the page

           -- should not mention have been made of the potential effects of radiation damage (for example, the radiolysis of water)?

14. Page 11, Section 3.6, first sentence -- are you concerned about creating a preferential pathway for the " radionuclides" in the waste? The same coment applies to item (ii), Section 8.0 first paragraph, page 18. ,

15. Page 13, Section 4.2, fi M item at the top of the page -- is "sorptivity" a chemical property? In addition, has the NRC Staff considered the relative importance of this item. The shaft on borehole seal is but a small part of the total sealing system and it would soon become saturated even if leakage occurred.

16. Page 15, Section 4.5, last item at the bottom of the page -- is this item complete? It appears that it may have ended in mid-sentence. .

17. Page 19, last paragraph, third line -- should this read: " Sections

(, 4, 5, 6 and 7 identi fy information. . . . . . ."? v A 4 0'7

O REPORT TO THE ACRS ON DRAFT GENERIC TECHNICAL POSITION

,                               WASTE PACKAGE RELIABILITY                            :

BY , ACRS WASTE MANAGEMENT SUBCOMMITTEE FEBRUARY 9, 1985 General Comments:

1. The Staff should provide evidence that it has demonstrated the validity of applying Monte Carlo techniques to the waste package performance assessment process before it urges DOE to use this method. Similarly, the use of random variables in performance models should be validated in advance of application (p.12).

2. Section 3.2 concerns reliability analysis addressed in a condition-al way. What options does the DOE have if they do not elect to use reliability analysis?

3. Requirements for predictive equations are unrealistic and not in accord with the " reasonable assurance" provision of the regu-lations. Many phenomena will not be amenable to such analyses and will be simply bounded. Strict adherence to the predictive equation requirement may force the applicant to use only materials that can meet this requirement, resulting in excessive costs and O' less than optimal choices. In addition, the added requirement for PDFs compounds this problem.

4. The Staff position appears to be that at least one method for calculating probabilities of failure by analytical techniques must exist. It should be a requirement to be fulfilled by the Staff to show that for all parts of the waste disposal analysis, at least one such method exists and can be legitimately applied. This demonstration should be completed before regulations are imposed on the applicant.

Specific Coments:

1. p. 3, item d: change to read: "(d)includedinpotential

           ....are:" Also, why are groundwater flow rates mentioned twice?

2. p. 4 Section 3.1.1: add "... fail to reveal new failure modes of significance."

3. Section 3.1.2, line 4: omit " physically"

4. Section 3.1.3, last line: change to read: "

                                                           ...until each of the pertinent failure modes..."

1

5. p. 5, top paragraph, last line: add "...model uncertainties and u assumptions should be included."

A-2.oa

i I WASTE PACKAGE RELIABILITY 2

6. p. 7, line 3: omit "all"

7. p. 7, Section 3.2.3: The applicant should be required to identify the quality of the materials property data, especially as used in model calculations.

8. p. 7, footnote: Not all of a repository may be unsaturated at closure.

9. p. 8, Table 1: Title should include " Example List." Properties under groundwater exclusion function should be part of the footnote. Chemical properties of the packing material should be added to the Table.

10. p. 15, Section 4.2: The technical position should identify the required output and require the applicant to define the .

methodology. The draft is generally successful in outlining these requirements except in Section 4.2 which appears to contain an excessive amount of "how to" information.

11. p. 17, line 7: It is not clear what is meant by " safety factor" when describing a model uncertainty that could, in fact, be highly unconservative.

12. p. 19: The expectation that sensible models will describe several of the phenomena listed is unrealistic. The text should be modified to convey this fact.

13. p. 19 Section 4.2.2 It would be better to call for accuracy i (rather than reproducibility) as the basic criterion for acceptance of numerical data.

14. p. 20, line 5: The requirement to furnish the expected distribution of errors for experimental data having "... uncertainties larger than a few percent..." fails to appreciate the broad range of numerical uncertainties associated with data obtained by state-of-the-art methods. It is inappropriate to specify a generic level of acceptable uncertainties for the multitude of data types associated with the waste package. Further, some of the data used in analytical models need not be very precise owing to the insensitivity of the model to such data. In these instances, there is no need to determine the distribution of errors.

O A- 2.oal

l l O , REPORT TO THE ACRS ON NRC TECHNICAL POSITION ON THE DETERMINATION OF RADIONUCLIDE SOLUBILITY IN GROUNDWATER FOR ASSESSMENT OF HIGH-LEVEL RADIONUCLIDE WASTE ISOLATION BY ACRS WASTE MANAGEMENT SUBCOMMITTEE FEBRUARY 9, 1985 General Comments:

1. The Subcommittee has noted that this Technical Position paper is

!        designated as " Final." Does this mean that it will not be subject to public comments with subsequent review? We believe that revisions are necessary.

2. As a discussion of radionuclide solubility in groundwater, the position paper is complete and generally accurate. Most of the discussion, however, is standard textbook material, and it is not clear why it needs to be repeated at such length here.

Specific Comments:

1. Page 2, 5th line, Section 1.3; We would suggest this be reworded to say: " Fourth, solubility estimates can be used to specify the source. term for far-field performance assessment."

2. Page 3, bottom paragraph, Section 3.0.: Do any site programs not take credit for solubility-limited release rates?

3. Page 5, 4th line, Section 4.0; We suggest you say: "The possibility that complexes or radionuclide-bearing colloids and i particulates..."

4. Page 6, end of 2nd paragraph, Section 4.1: Can radionuclides be eliminated when their low solubilities may subsequently be modified by formation of complexes or colloids? Should not consideration be given to this possibility?

5. Page 9, 3rd paragraph, Section 4.3.2: Consideration should be given to the fact that steady state conditions with metastable forms of a solid may be far from equilibrium and would not give a reliable estimate of equilibrium concentractions.

i

6. Page 10, last 4 lines, Section 4.3.2: Another advantage is that such experiments would provide conservative solubilities.

7. Page 11, Section 4.4.1, lines 6 and 7: We suggest you say:

A-?lO

                                     .                                                                i SOLUBILITY IN GROUNDWATER            2

(}}. i "However, when the concentration of other components and complexes, common to both groundwater and a ....." i

8. Page 12, Section 4.4.2, top paragraph: The estimation of redox potentialsbymeasuringtheconcentrag{onsofoxidizedandreduced l

forms of a couple of ions, such as Fe & Fe , could be mentioned, t

9. Page 13, Section 4.4.5, last sentence, first paragraph: This sentence is wrong. The formation of colloids could increase the effective concentration of a radionuclide, but colloids certainly do not travel faster than materials in solution.

k 1

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'1 4 O f\- 2Lll -- . - -. .- . . -

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                                                                                                               +

l lO j i l REPORT TO THE ACRS ON REVISED MODELING STRATEGY DOCUMENT , FOR HLW PERFORMANCE ASSESSMENT  ; j BY l ACRS WASTE MANAGEMENT SUBC0m lTTEE l FEBRUARY 9, 1985 i General Conments: i 1. The modeling strategy is well thought out and attacks the problem t i in a fundamental way, by asking nine post closure performance , questions. These range from the rate at which water gets to the . l waste form to the rate at which radionuclides are released to the l accessible environment. In addition, there are one preclosure and one retrievability design question. The NRC Staff notes that they 4 are not required to independently develop computer codes nor to do  :

)                    independent numerical analysis, but can proceed from level (1)                            l j                     critical evaluation of DOE's work, to level (2) use of simple                             :

conservative models, to level (3) qualification of the use of DOE's  !

I models and codes for independent analyses, up to level (4) in-  ! dependent development of models and codes for independent analysis , of the performance questions. They note the choice among the four i

strategies in Figure 6. For those processes that are well known, i i they could easily fulfill their obligations by utilizing level (1)  ;

! studies. For those processes that are least well understood -it  ! ) would appear wise to use levels (2) and (4). Yet, in the basin and f I regional groundwater flow analyses and in the far field containment i } transport and groundwater flow, which are the simplest to model and l best understood, they use levels (1) and (4). The need for independently developing models for this comparatively well' understood area is inadequately justified. In the two regions of i , greatest uncertainty and concern, underground facility /near field, j only groundwater flow is treated at levels (1), (2), and (4) and 3 only heat transport and mechanical / structural are treated at levels i j (1),(2)and(3). Yet it would appear that heat transport is ! comparatively well understood and the contaminant transport,  ! 1 geochemistry, and radiation induced effects which are comparatively j less well understood are only treated at levels (1) and (2). 4 2. 4

• The staff is Function Distribution preparing (to evaluate the Complementary CumulativeCCDF)

EPA's final ruling.' This"is wise yet they might proceed slowly on this matter since EPA's Science Advisory Board Subcommittee on High Level Waste had recommended that the probabilistic risk assessment

be used only if it could be shown that sufficient information was

! available to do this. In particular, the EPA Subcommittee was } concerned about the ability to reach consensus on the probability

• i of some of the events.

O 3. The staff notes that there are four cate.ories of technicai

                                                                                                               )

i uncertaint 1) identification of basic phenomena, 2) accuracy of i

models,3)y

accuracy'of data and, 4) calculational uncertainties. l 1

l _ _ __-A g, w _ . _ _ _ _._

MODELING STRATEGY 2 p V They will attempt to resolve these difficulties by using simple and bounding models where appropriate and by closely evaluating the data used. This is a sensible course of action. We concur. Specific Coments:

1. The Section on preclosure analyses notes that only the transport and transformation in the ventilation systems needed further work.

We concur.

2. Retrievability - this section states that the analysis of corrosion rates is based on experimental data rather than on a "first principles" model. This appears to be a wise choice.

3. Water contact to backfill and waste package - will utilize their own model for water flow which is good. However, the use of the TOUGH code for unsaturated flow to be used later in conjunction with FEMWASTE prompts the question why not use FEMWATER which is the coupled code for FEMWASTE.

4. Water contact waste form - this section does not indicate how they plan to handle rapid or discrete failure of the waste package.

5. Radionuclide release from waste form - the NRC Staff recognizes that the technical uncertainties are so great that they will depend primarily on experimental data. There is no discussion of the use O of backfill or the modeling of it.

6. Radionuclide release from waste form - This is possibly the key item, yet the NRC Staff will rely primarily on extrapolation of empirical data. This may not be sufficient; modeling, although difficult, may be necessary.

7. Evaluation of radionuclide release from waste packages takes no credit for the cannister yet some possible cannister materials, such as, T1 Code-12, would have reasonable containment times.

8. Radionuclide release from backfill - this is one of the key items, yet the Technical Position states that independent staff analyses "may be conducted on a limited basis." Is this adequate?

9. Radionuclide release from disturbed zone - we question whether thermal bouyancy effects in groundwater are the major problem.

What about other factors such as biogeomobilization of wiste materials, colloid or complexed material transport, competing fons, and flows in the unsaturated zone? Fracture flow may be particularly important in the region back to the waste package, yet this subject also is not mentioned. 10 Radionuclide release to accessible environment - on1 addresses releases up to 10,000 years. Although this is al t at is currently required, it is almost certain that EPA will require some Q C sort of analysis beyond this time to determine if any drastic l l l l A-2.o

MODELING STRATEGY 3 i events could occur (this has nothing to do with the credibility of O events taking place this far in the future).

11. The NRC Staff appears to concentrate its attention on areas that are best understood (i.e., "far-field groundwater flow and contaminant transport," page 56). The reverse approach should be used.

12. Pre-emplacement groundwater travel time -- again, the staff appears to concentrate on the far field, yet for certain sites, such as Yucca Mountain, the major action may be closer in. Is the staff dismissing these areas with the hope that they can demonstrate that

consideration of the far field alone will provide sufficient data to show that the requirements of EPA and NRC will be met?

l i i

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

l , 1 REPORT TO THE ACRS ON CHAPTER 7. " COMPARATIVE EVALUATION OF SITES PROPOSED FOR NOMINATION," DRAFT i ENVIRONMENTAL ASSESSMENTS, U.S. DEPARTMENT OF ENERGY BY ACRS WASTE MANAGEMENT SUBC0PetITTEE FEBRUARY 9, 1985 At a meeting on January 17-18, 1985, the Waste Management Subcommittee of the Advisory Committee on Reactor Safeguards prepared comments on Chapter 7 " Comparative Evaluation of Sites Proposed for Nomination," of the Draft Environmental Assessments published by the U.S. Department of Energy. Because of time constraints, these comments, which are given below are brief and incomplete. Nonetheless, the Subcomittee hopes they will be useful. General Comments:

1. In general, this is a well prepared report. The DOE Staff has exercised care in their evaluation and comparison of the candidate sites for a HLW repository, and has followed the guidelines that they prepared and with which the NRC has concurred.

2. The comparative evaluation, however, covers only the final five "best" sites and the reader is provided no information on the four sites that were rejected. Although this information is available in the individual EAs, it would have been helpful to have had it summarized in this chapter.

3. The Subcommittee was unable to understand the rationale or justi-fication for the assignment of weighting factors (In Appendix B of each EA) to the various parameters considered in the comparative evaluations of the sites. For example, the reasons for assigning a weighting of "5" to hydrology, while population density was assigned a weighting of "12i", are not readily apparent.

Clarification would be helpful. Specific Coments: Page 7-6, Table 7-1 -- the information presented in this Table is minimalanditdoesnotadequatelysupporttheaccompanyingtext(page 7-14). Although later sections 'of the Chapter include supporting quantitative data, only qualitative information is given sere. We suggest that this section be expanded to include more supporting quantitative data. Page 7-9 -- we were unable to understand the relative significance or to ' 6i 7-t-O i aart c or ta co ditia"> d >cri6 d i" ** <aat"at Would it be possible to develop statements that could be more readily understood, based on Appendix !!! of the Siting Guidelines, 10 CFR 9607 A-2.l5

CHAPTER 7 2 i Page 7-15 -- the first paragraph on this page states that the " dry O conditions likely at Yucca Mountain are thought to almost balance the fact that the site has a shorter time of ground-water travel than j i Richton, Davis Canyon, and Deaf Smith -- ." What is the basis for this 1 statement? What were the key factors used-in this balance? , i Page 7-21 -- in line 6 the statement is made that geochemical processes that could reduce the sorption of radionuclides or degrade the strength : l of the rocks -- is not present at any of the sites." Would not the l presence of saline waters lead to a decrease in sorption? i O .) 4 !. O d

A - 2.t(,

APPENDIX XI ADDITIONAL DOCUMENTS PROVIDED FOR ACRS' USE Q D ADDITIONAL DOCUMENTS PROVIDED FOR ACRS' USE

1. Memorandum, S. Duraiswamy to ACRS Members and Technical Staff, Portion of the Green Book Related to the NRC Safety Research Program Budget for FY 1986 and 1987. February 6,1985

2. Memorandum,'S, Duraiswamy to C. P. Siess, Chairman, Safety Research Program Subcomittee, Changes to the FY 1986 and FY 1987 NRC Safety Research Program Budr'st Since the January 1985 ACRS Meeting, February 4,1985

3. Memorandum, N. J. Palladino to W. J. Dircks, EDO, FY 1986 Budget Submission, January 18, 1985

4. Tables, Nuclear Regulatory Commission, Budget Estimates - Program Support /Sumary by 85 B&R

5. Memorandum, H. R. Denton, Director, NRR, to R. B. Minogue, Director, RES, FY 1986 RES Budget, January 30, 1985

6. Note, M. Riggs to E. F. Conti, ACRS Requests / Questions Raised on 2/7/85 February 7,1985

7. Memorandum, R. F. Fraley to ACRS Members, Proposed Legislation Regarding Nuclear Power, January 7,1985 O 8. Report, Background Information for Region III Presentation to ACRS Evaluation of Construction Quality Braidwood 1 and 2 February 7, 1985

9. Memorandum, M. Bender (ACRS consultant) to E. P. Igne, Comentary on Braidwood Units 1 and 2 Operating License Review Ref., ACRS 2:

BDWD-COM, February 6,1985 r r 1 e f I u)

                              ,           A- Z.n '                              ,}}