ML20043D672
| ML20043D672 | |
| Person / Time | |
|---|---|
| Issue date: | 01/18/1990 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-2685, NUDOCS 9006110049 | |
| Download: ML20043D672 (13) | |
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55 PJ DATE ISSUED:'
1/18/90 47 ACRS Meeting Minutes / Summary of the Advanced Pressurized. Water Reactors Subcommittee January 10, 1990 Bethesda, Maryland The purpose'of this Subcommittee meeting was to continue discus-sion'and review of the HAPWR (SP/90) design.
Attendees ACRS HEC J. Carroll, Chairman L. Donatell, NRR W. Kerr, Member S. Newberry, NRR C.
Michelson, Member G.
Kelly, NRR P. Shewmon, Meater C. Goodman, NRR C.-Siess, Member H. Pastis, NRR D. Ward, Member D. Notley, NRR C. Wylie, Member M. Marayama, NRR
.M.
El-Zeftawy, Staff R. Architzel, NRR C. Michols, NRR o
L Others
.D.
Shum, NRR T. Van'De Venne, H G..Remley, H J. Easter, H E. Burns, E ll.
W. Schivley, H R. Span, E l'
'A..Cheung, H D. Noonan, Bechtel Meetina hiahliahts.'Aareements, and Reauests l '.
Mr. Carroll, Subcommittee Chairman, stated the purpose of the Subcommittee meeting and introduced the other ACRS members.
2.
Mr. L. Donatell,-NRR/ Project Manager, briefed the Subcommit-tee regarding the PDA st atus for the SP/90 design.
He indicated that the NRC has issued 13 PDAs in the past.
The 900611o049 900110 1,u ct' A ORIGINAL 0
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Advanced'PWRs Mtg Minutes 1 2-January 10, 1990 last one war in November 14, 1978, which was for the M/414 design.
The SP/90 PDA is the first one to include severe accident policy statement in the review.
This has caused a great deal of difficulty with the staff.
The PDA does constitute a reference design, which means that when it is enforced, it can be referenced in applications.
Those applications are construction permits and manufactur-ing permits.-
Mr. Donatell indicated that the staff at the November 3, 1989, ACRS Subcommittee meeting incorrectly stated the current' requirements for a PDA.
The staff now defines the PDA to require design details equivalent to a preliminary safety analysin eeport and as defined in 10 CFP. 50.34.
Mr.
Donatell mentioned that the previous statement from the staff that "the PDA would be. subject to a complete review even after it is being issued," is not strictly true.
It is only subject to 10 CFR 50.109, the backfit rule.
The Commission has approved the June 1990 completion of the PDA.
Due.to the shortage of time and the low priority, Mr.
Donatell is taking draft input from the reviewers on the open items and incorporating this information into the draft final SER.
Mr. Donatell indicated that the USIs, GSIs, TMI open items, and the severe accident issues will not be reviewed at the PDA stage.
The staff has received input from H on the original 107 open items with the exception of 15 items.
The staff has found E's response for 50 open items to be accep-table.
There are 18 open items that will be defcrred to the FDA stage, and 2 open items that will extend beyond the FDA
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Januar} 10, 19?O.
i commitment..These'are the design audit of reactor vessel manufacturer-(not selected at this time) and sample capabil-ities for process systems.
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-.There are 9 minor open items that still remain, and 13 open items that could remain open upon issuance of the PDA.
The staff is planning to issue the draft-final SER in April 1990.
The staff is requesting two additional ACRS Subcom-mittee meetings, and a full ACRS meeting and letter in May 1990.
13.
'Mr.
E.
Burns, Manager of Plant Licensing at H, summarized the review status and interaction with the.NRC staff regard-
'ing the SP/90 design.
He indicated that Chapters 13 (Con-duct of Operations), 14 (Testing and Maintenance), 16 (Technical Specifications), and 17 (Quality Assurance) of the DSER are not anticipated to be part of the ACRS review for the PDA stage.
H expects to meet with the ACRS Subcom-mittee in March 1990 to complete the review of open items and in April 1990 to review the-final draft SER.
4.
Mr. Van de Venne, H, responded to Dr. E5att m's questions (from the November 3, 1989 ACRS Subcombic meeting) that dealt primarily with the pressure boundary for the SP/90 design.
Question 1:
Does the SP/90 design has any welds in the core region?
Answer:
The reactor vessel will have no welds
- .he core region.
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ouestion 2:
What are the standards for a pipe joint design to make UT inspection more reliable?
Answer:
The albows in the primary coolant piping will have a straight section of pipe to facilitatei UT-inspection.
Question 3:
What are the specifications for the cast stainless steel pipe elbows, valve bodies, etc., to make inspection more reliable, and aging less of a problem?
Answer:
The use of stainless steel castings will bc, minimized.
The reactor coolant piping and elbows are forged.
The reactor coolant pump casing consists of a single casting, i.e.,
there are no welds.
Question 4:
What will be the composition of the steel in the RPV, and how will it be made?
Answer:
The RPV will consist of forged sections.- The material will be in accordance with the EPRI ALWR Requirements Document.
.ouestion 5:
What will be the materials of construction of the steam generators?
Answer:
The S.G. tube material is Inconel 690 TT.
The tube support plates are manufactured from 405 SS.
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Advanced'PWRs'Mtg Minutes 5
January _10, 1990.
Mr. Van de Venne.provided a general arrangement layout l
configuration for the SP/90 design as per Mr. Michelson's request. -This layout highlighted the reactor building, containment, location of the diesel generators room, the RHR heat exchangers, EFWST, accumulator tanks, reactor vessel, steam generators, pressurizer, fuel handling crane, etc.,
and service building.
5.
Mr. G. Remley, H, presented an overview of the Instrumenta-tion and Control Systems (Chapter 7 of the DSER) for the SP/90 design.
He indicated that the SP/90 design will be using digital technology that includes:
- 32-bit plant computer systems 10-bit microprocessors a
Distributed digital processing architecture Multiplexed communications Fiber optic cabling Sophisticated control and protection algorithms Fault-tolerant design.
E's-objective is to use digital electronic technology to provide improvements in availability, operability, main-tainability, construction schedule, costs, aind flexibility for the future.
Another objective is the integration of total plant instrumentation and control systems.
Mr. Remley indicated the fiber optic signal transmission prevents fault propagation.
The design also provides a clean separation of safety trains and channels.
Unlike prev.ious H designs, the SP/90 design does not have a
. separate reactor trip system or engineered safsty features
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actuation system..
The integr '.ed protection system (IPS) will sense plant cor.ditions tid generate signals to initiate reactor trip and engineered safety features system actua-tion.
'The IPS will use digital processing techniques to perform safety functions.
The protective action system will carry out the protective E
functions such as reactor trip and engineered safety fea-tures system actuation on demand from the IPS.
Safety-LI related display instrumentation will provide safety system operating status and information to enable the operator to perform the required safety actions.
The SP/90 will include a new addition of reactor trip functions for departure from nucleate boiling ratio and linear heat generation rate.
The IPS will consist of four redundant channels, contained in integrated protection cabinets (IPCs), eacn with its own set of process inputs.
Each channel will read the process measurements and perform any required processing.
The results of the processing from the channel will be combined 1
with similar results received as inputs from the remaining three channels in a trip logic computer.
The trip logic computer will generate the reactor trip functions when needed.
The IPS will also contain two engineered safety featuras b
actuation cabinets (ESFACs).
Each ESFAC will receive inputs from the four channels regarding the status of the protec-tive functions and perform the voting logic manipulations "or the engineered safety systems actuation.
Each channel of the IPS will include a manually initiated testing system.
Each channel will transmit u4 formation on m
i Advanced PWRs Mtg, Minutes 7.
January 10, 1990 various parameters'to the safety-related display information system and provide data-to the plant control system.
The IPCs provide signal conditioning and digitizing-for analog sensor inputs.
.These cabinets provide the coin-cidence-logic to generate trip outputs for the reactor trip breakers.
They also provide bistable trip outputs for actuation of the engineored safety features system.
The integrated logic cabinets provide the component level actuation signals developed from the system level actuation signals to actunte the ESF system, final actuation devices, and actuated equipment.
The. standard cabinet design environment is as follows:
- Temperature - (60-105"F for normal range, and 40-120'F for abnormal range).
- Humidity - (0-95% non-condensing and 95'F_ max. wet bulb temperature).
- Seismic - IEEE 344-1975.
- Electromagnetic interference - EMI reduction window glass, screened louvers, cable entrance plates, optical isolation, field wiring shielded from microcomputer wiring, and shielded sensor cables.
- Maintenance bypasses to allow on-line repair without error-induced trips.
- Remote roadout of complete system status.
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January 10, 1990 Microcomputer, printed circuit (input / output) cards.
Modular repair with standardized components.
1 The protective action system will consist of four reactor trip actuation trains and two engineered safety features actuation trains.
The protective action system will also furnish status information to the IPS for interlock? and l
permissives-and for transmission to the safety-related l
l display information system.
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l The reactor trip actuation trains will cor sist of eight circuit breakers (two in each train) interconnected in a two-out-of-four logic.
Each reactor trip actuation train receives a trip signal from the associated channel of the IPS.
s For the IPS, H has not analyzed the effect of common mode
. failures.
However, other items such as functional diver-sity, functional testing, qualification, fail-safe design
- l principles, designed for maintenance, and verification and validation were considered.
The NRC staff will require-H to perform a design analysis of the IPS in accordance with the defense-in-depth guidelines at the FDA stage.
The staff will also require a suitable periodic testing program be conducted to demonstrate that the IPS performs as analyzed.
Mr. Remley indicated that H has committed to provide a verification and validation program for the SP/90 design
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which will closely conform to ANSI /IEEE Std. 7-4.3.2-1983,
" Application Criteria for Programmable Digital Computer
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January 10,.1990 T
Systems in Safety Systems of Nuclear Power Generating Stations."
6.
-Mr. J.
Easter, H, briefed the Subcommittee regarding the
, Human Factors Engineering (Chapter 18 of DSER) for the-SP/90 design..He indicated that H has followed the preferred top-down approach for systems / function / task analysis recommended ~
by the NRC staff in NUREG-0700, " Guidelines for Control Room Design Revieus."
In addition, the analyses take advantage of existing research in operator decision-making models to optimize: the man / machine interf ace.
H has adapted a simpli-fled operator decision-making model developed by Mr. Rasmus-sen, of Denmark National Laboratory.
For the control room des'ign process, H's task analysis overlays simplifieu decision-making model onto functional model of plant proces-ses.
For the alarm system, the types of messages are:
- alarm overview messages - display abnormality
- alarm support messages - provide means for operator to query the alarm system
- alarm systems actions messages - messages telling the operator what the automatic control systems are doing
- emergency safeguards status messages - continuous indication of the binary state of the ESF.
For the display system, the displays are structured to match-the. functional organization of the plant in order to support the decision-making activities of plant operation.
- Also, this' functional organization continually reinforces the operator's understanding of the plant design.
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i Ad' v gd PWRs Mtg Minutes 10 January-10,_1990 The physical depicticn of the plant supports ~the planning of
' detailed control actions and feedback on those actions.
Graphic data utilizes the operator's perceptual abilities to portray complex " data behavior" while data on graphics supports planning activities by organizing plant data and conveying its significance, calculated data is also in-cluded in displays and updated regularly.
The control room layout is of a circular design to have easy visual communication with other workstations from super-visory console and individual station:s, substantiated by full-scale mock-up.
In addition, this circular design will result in an easy verbal communications due to effective sound reflections and reduction of extrareous traffic and vistors to the control room.
7.
As a result of the Subcommittee's discussion, some of the Subcommittee's members expressed concerns in regard to the following:
- Dr. Kerr expressed concern that the severe accident issues, which are cor.sidered as principal contribut;.r to reactor risk, would not be reviewed at the PDA stage for the SP/90 design.
- Mr. Michelson expressed concern regarding the implica-tions of the'PDA and the depth of the review at the PDA stage.
Dr. Kerr commented that the staff's criteria of how to use the PRA for the SP/90 design is not well defined.
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Advanced PWRs Mtg Minutes 11 January 10,,1990
- Mr. Carroll ' suggested that the staff and H representa-tives to read the ACRS letter for-the ABWR design-issued on November 24, 1989, and provide any comments that relates to the SP/90 design.
A copy of such letter was provided to the staff and H.
- Dr. Kerr. indicated that in come sections of the DSER (e.g., Chapter 11, some statements were made implying that the staff is relying on H to determine what are the acceptable methods for reviewing the design.
Dr.
Kerr commented that, an acceptable method is something that only the staff can determine.
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- Mr. Michelson expressed concern regarding-the proximity of the diesel generator room to switch gear room, relay room, and the control room in the event of fire or some other external events.
- Mr. Michelson expressed concern due to the. fact that there is no supporting study to determine the effects of valve bonnet or cover plates acting as missiles, or broken components in the diesel generator. room also acting as missiles.
- Dr. Kerr questioned the reliability goals that H is trying to achieve, namely:
10~7 with respect to failures per demand for the IPS, 10-5 with respect to failures per demand for the engineered safeguards, and 10-3 with respect to failures per demand for the control' systems.
He indicated that these reliability goals do not include the common mode failures criteria
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in the analysis, and it is difficult to emphasize the-credibility of such numbers.
- Mr. Carroll commented that ACRS has concerns that the NRC staff lacks the kind of compute-expertise needed to review digital control and protection systems.
He suggested that this issue needs further discussion.
- Mr. Michelson expressed some concern regarding the fire protection system for the SP/90 design and the use of ventilation systems that intertie fire areas.
- Mr. Michelson indicated that additional'information is needed to know more about the enviromental qualifica--
tions and separation of the solid state control systems and the fiber optical cables.
- Mr. Michelson indicated that-the feedwater system control for the SP/90 design'should be elevated to a safety system in the design.
- Mr. Wylie commented that the SP/90 design should provide some protection against station grounding and lightning Future Action
.The Subcommittee Chairman and members have decided to conduct another Subcommittee meeting-in early March 1990, to continue discussion of the subject matter, i
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NOTE:
Additional' meeting details can be obtained from a l
transcript of this meeting available in the NRC Public-Document Room, 2120 L Street, NW, Washington, DC 20006, o
1 (202) 634-3273, or can be purchased from Ann Riley and.
Associates, Ltd., 1612 K Street, NW, Suite 300, Wash-ington, DC-20006, (202) 293-3950.
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