ML20206J647
| ML20206J647 | |
| Person / Time | |
|---|---|
| Issue date: | 11/21/1988 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Scherer A ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| References | |
| PROJECT-675A NUDOCS 8811290064 | |
| Download: ML20206J647 (8) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION 1
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November 21, 1988 Pr ject No.: 675 Mr. A. E. Scherer, Director Nuclear Licensing Combustion Engineering 1000 Prospect Hill Road Post Office Box 500 Windsor, Connecticut 06095-0500
Dear Mr. Scherer:
SUBJECT:
SCOPE OF DESIGN AND SCOPE OF STAFF REVIEW OF CESSAR-DC SYSTEM 80+
For some time now the staff has been discussing the scope of the design and the scope of the staff review of CESSAR-DC, System 80+ with you at::1 your staff. This letter provides guidance regarding these subjects.
Scope of Design In the proposed regulation 10 CFR Part 52, the Commission has defined the desired scope of designs for standard plants requesting certification. The proposed regulation states that "Ideally, designs for which certification is sought will be for essentially complete plants."
It also states that "the NRC will give priority in allocation of resources to support review and approval of application.- for essentially complete plants."
The current version of CESSAR System 80+ does not cover an "essentially complete" plant. Should you decide to pursue the application for design certification of System 80+, the application should be for a design whose scope is essentially complete and consistent with the proposed 10 CFR Part 52.
A more limited scope of design, as is in your System 80+, would not be likely to receive priority allocation of staff resources and the review schedule could be jeopardized.
Scope of Staff Review The staff does not agree with you that the System 80+ design embodies only minor changes beyond your System 80 design. As a result, the staff plans to perform a detailed review of System 80+.
Specifically, since the application will have to comply with requirements stipulated in proposed 10 CFR Parts 52.45 and 52.47, all portions of the design are subject to review and approval.
In addition, the staff is considering matters that go beyond the current Standard Review Plan but that we expect these advanced r& actor designs to embody. Some of these items are briefly discussed in the! enclosure j {()
to this letter. These issues will be addressed in the Licensing Review Bases t
document currently under consideration, 7
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i Hr. A. E. Scherer November 21, 1988 We request you inform us of your intentions in this matter along with your schedule for submitting your application, including the submittal of the Final Safety Analysis Report for System 80+, within 30 days from receipt of this letter.
Sincerely, h.
e ens. Crutchfie Acting Associate Di ctor for Projects Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/ enc 1:
See next page E.
Mr. A. E. Scherer November 2.1 1988 We request you inform us of your intentions in this matter along with your schedule for submitting your application, including the submittal of the Final Safety Analysis Report for System 80+, within 30 days from receipt of this letter.
Sincerely, Original signed by Dennis M. Crutchfield Acting Associate Director for Projects Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/ encl:
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t We request you inform us of your intentions in this matter along with your schedule for submitting your application, including the submittal of the Final Safety Analysis Report for System 80+, within 30 days from receipt of this letter.
Sincerely, Dennis M. Crutchfield Acting Associate Director for Projects Office of Nuclear Reactor Regulation
Enclosure:
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Project No. 675 Advanced CESSAR cc: Mr. C. B. Brinkman, Manager Washington Nuclear Operations Combustion Engineering Inc.
7910 Woodmont Avenue, Suite 1310 Bethesda, Maryland 20814 Mr. Ernest Kennedy Manager of Licensing Combustion Engineering 1000 Prospect Hill Road Post Office Box 500 Windsor, Connecticut 06095-0500 I
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ENCLOSURE POTENTIAL REVIEW SUBJECTS FOR STANDARD PLANT DESIGNS The staff met to discuss major issues related to the review of advanced light water reactor applications.
Included in the discussions were issues in which the scope of the staff's acceptance criteria may go beyond that of the current Standard Review Plan to ensure improved design, construction, and/or operation of these advanced plants. The staff's positions on these matters will be finalized during the development of the Licensing Review Basis (LRB) for the CE design. The following are brief discussions of the staff's current views on some of these matters.
1.
60 YEAR LIFE:
For applications proposing a 60-year design life, the staff would review the designs for a 60-year life notwithstanding the fact that a 40-year license term limitation is presently in the regulations.
It is the applicants' responsibility to identify the components and systems which are affected. Applications for design certification will have to provide information and programs to support design life, and the reviews for such issues as fatigue, corrosion and thermal aging.
2.
FIRE PROTECTION:
Improved fire protection criteria are needed in view of the significant contribution of fires to core melt probability. The current Appendix R and BTP 9-5.1 requirements (e.g. 20 ft. separation) should be replaced by a requirement for safe shutdown capability in the event of a complete loss of any fire area.
3.
TECHNICAL SPECIFICATIONS:
The staff considers that (1) proposed Technical Specifications should be developed as early as practicable, but be submitted no later than the FDA application, (2) proposed Technical Specifications representative of the design should be submitted for review and approval by the staff as part of the FDA submittal, and will be included in the Design Certification process, and (3) applicants should identify design features that are necessary for testing and maintenance during operation without challenging safety systems.
The Technical Specifications should be developed, where practicable, based upon risk and reliability considerations.
4.
TESTING AND MAINTENANCE:
Certification of a design will be based in part upon a probabilistic risk assessment of that design.
In that the validity of a PRA is highly dependent on the reliability of systems, structures and components, the staff requires assurance that programs will be implemented which will ensure that the reliability of those systems, structures and components (assumed in analyses) will be maint61ned throughout plant life. There-fore, a program to assure design reliability must be provided as part of
2 the FDA application. This program which will be certified as part of the design should address items such as (1) the Technical Specifications and ISI/IST,(2)theMaintenanceProgram,(3)PlantPrecedures,and(4)
Security.
5.
EDUSTRYUSEOFMAAP:
MELCOR and Source Term Code Package (y since the staff can apply its own Review of the MAAP code is unnecessar STCP) codes in its evaluations.
6.
STATION BLACK 0UT AND ELECTRICAL SYSTEM:
Future ALWRs should adopt improved electrical systems to ensure a safe shutdown of the reactor. These systems should provide, in part, for diverse power sources in order to eliminate the concerns related to station 31ackout. General guidelines will be developed and finalized as part of the Licensing Review Basis documents.
7.
LEAK BEFORE BREAK:
Leak before break can be considered where justified. Where applicable, designs must address good practices in order to maintain steam generator tube integrity. Also, designs should address issues of material embrittle.
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ment associated with current vessel materials and vessel supports.
A new rule and draft SRP Section 3.6.3 have been issued. The EPRI design requirements has adopted i.5ese criteria.
8.
SOURCE TERHS:
The staff is concerned th4t the licensing basis source term "TID 14844" is not consistent with cuirent '.nowledge, therefore, with EPRI input, realistic source terms will be established to be uniformly applied to future ALWRS.
9.
PHYSICAL SECURITY:
Sabotage should be addressed in c11 future ALWRS applications. As a minimum, information should be providcd to demonstrate the existence of adequate physical barriers to protect vital equipment in accordance with 10 CFR 73.55(c) and to identify (access control points to all vital areas in accordance with 10 CFR 73.55 d).
- 10. OBE/ DYNAMIC ANALYSIS METHODS:
The staff agrees that the OBE should not control the design of'safetyThestaffwk11 systems as now required by 10 CFR 100 Appendix A.
this issue under consideration as part of the design certification t
process.
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3-II. TYPE C CONTAINMENT LEAKAGE RATE:
Containment leakage is acknowledged by the stafI as being a function of containment pressure.
- 12. HYDROGEN GENERATION:
10 CFR 50.34(f) related to the issue of a 100% r9tal water reaction will be invoked for ALWRs consistent with Comission Policy and proposed 10 CFR Part 52.
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