ML20045D102
| ML20045D102 | |
| Person / Time | |
|---|---|
| Issue date: | 02/12/1993 |
| From: | Wilkins J Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-2862, NUDOCS 9306250328 | |
| Download: ML20045D102 (19) | |
Text
-
06fS-M4;L
' CERTIFIED BY:
DATE ISSUED:
2/4/93
. ~ J. E. Wilkins, 2/12/93
[pg- [r[A//f 3
SUMMARY
/ MINUTES OF THE ACRS SUBCOMMITTEE-MEETING ON THE ADVANCED REACTOR DESIGNS JANUARY 6, 1993 BETHESDA, MARYLAND l
PURPOSE The purpose of this meeting was to hear presentations by the NRC staff regarding the key policy issues for the MHTGR, PIUS, PRISM, and CANDU 3 advanced nuclear power plant designs.
The meeting I
convened at 8:30 a.m. and adjourned at 2:45 p.m.
The meeting was l
held entirely in open session. No written comments or requests for j
time to make oral statements were received from members of the public.
Dr. El-Zeftawy was the cognizant staff engineer for this meeting.
The principal attendees were as follows:
1 ATTENDEES ACRS J.
E. Wilkins, Chairman P.
Shewmon, Member
{
1 J.
Carroll, Member D. Ward, Consultant P.
Davis, Member M.
El-Zeftawy, Staff T. Kress, Member C. Michelson, Member NRC OTHERS M.
Slosson, NRR C. Reid, PDCO E. Throm, NRR J. Herring, PDCO J. Donoghue, NRR C. Molnar, ABB-CE D.
Scaletti, NRR P. Rohr, ABB-CE R. Meyer, RES L. Rib, AECLT J. Kennedy, NRR P.
Fulford, NUS S.
Sands, NRR N. Grossman, DOE T. Cox, NRR J. Herczeg, DOE D. Roth, NRR H. Thornburg, ABB J. Donohew, NRR T. Anderson, H H. Pastis, NRR V.
San Angelo, Bechtel M. Case, NRR j
T. Ulses, NRR j
D.
Ebert, RES j
D.
Carlson, RES Z. Rosztoczy, RES G. Capponi, RES DESIGNATED ORIGI m 93062's6fas 930212
'~
/p
\\
u ntified By 3
I
)
-QN O(
i
, Advanced Reactor Designs January 6, 1993 Minutes chairman's openina Remarks In his opening remarks, Dr. Wilkins stated that the purpose of this meeting is to review ten specific issues that have been identified by the staff in a draf t SECY paper regarding the PRISM, PIUS, CANDU 3, and the MHTGR reactor designs.
Each of these ten issues may or may not be relevant to each of the four separate reactor designs.
Dr. Wilkins indicated that the Subcommittee will have additional meetings in the future as new information becomes available to the NRC staff.
NRC Staff Presentation Ms. Slosson, NRR, stated that the staff has prepared the draft SECY and has released it to the pre-applicants for their comments.
The staff requested the comments to be received by January 25, 1993.
The staff is also requesting the ACRS review of the subject draft SECY.
Once the staff receives the comments from the pre-applicants and from the ACRS it will finalize the SECY paper and go forward to the Commission for final approval and guidance.
The current schedule for the staff to complete the pre-application reviews-is as follows:
PRISM - December 1993 CANDU 3 - December 1994 PIUS - April 1995 MHTGR - December 1995 i
i
Advanced Reactor Designs January 6, 1993 Minutes A.
Accident Evaluation Issue Mr. Throm, NRR, summarized this issue.
This issue deals with the selection of accidents, associated frequency ranges, and acceptance criteria to assess the safety'of the four proposed designs.
The current regulations (i.e.,
GDC 4) requires the consideration of accidents in the design bdsis.
Also, 10 CFR 52.47 requires the consideration of consequences of both severe accidents (through the PRA) and design basis accidents (DBA) for designs that differ from evolutionary /paasive or other innovative means to accomplish safety functions.
The pre-applicant's approach is to analyze accidents signifi-cantly less probable than the present design range and to assure through their designs that these accidents had accept-able consequences limited to specific dose levels to the public.
Such levels could be the EPA lower level Protective Action Guidelines (PAG) of 1 rem whole body and 5 rem thyroid.
The PIUS guidelines invoke the PAG for accident sequences more probable than.10-6 per reactor year.
The CANDU 3 has excluded analyses of the consequences of events with frequencies of less than 10-6fgy, The PRISM guidelines invoke the PAGs for accident sequences e
more probable than 10-4/RY.
1 q
Advanced Reactor Designs January 6, 1993 Minutes I
e The MHTGR accident guidelines invoke the lower-level PAG dose limit for all sequences more probable than 5X10-7/RY.
The staff recommends the following:
e Select events deterministically and supplement with PRA.
insights.
Establish event categories consistent with LWR practices.
The selected range of events will encompass less likely event sequences, e
Establish consequence acceptance limits consistent with commission guidance with appropriate conservatisms to account for uncertainties, Develop methodologies and assumptions consistent with LWR e
practices.
Determine source term in accordance with proposed SECY and Commission guidance.
e Select set of events deterministically to assess safety margins, determine scenarios to mechanistically determine a source term and to identify a containment challenge scenar-lo.
Select external events deterministically, consistent with LWRs practice.
6
Advanced Reactor Designs January 6, 1993 Minutes Evaluate multi-module reactor designs considering scenarios e
allowed by proposed operating practices.
B.
Source Term Mr.
Throm, NRR, described this issue as follows:
Should mechanistic source terms be developed in order to evaluate the subject four designs.
Current regulations such as Appendix I to 10 CFR Part 50, 10 CFR Part 100, and 10 CFR Part 20, all have limitations - on releases related to power plant source terms.
The PRISM designers have proposed the calculation of a source term different from that done for LWRs.
The MHTGR designers have proposed siting source terms for accidents based on the expected fuel integrity.
The PIUS designers have proposed using a mechanistic LWR source term.
The CANDU 3 designer uses a source term for each scenario.
The NRC staff is recommending that the advanced reactor designers-develop source term ' based on mechanistic analysis provided that:
Performance data on fuel during normal and off-normal conditions is well understood, i
.+
Advanced Reactor Designs January 6, 1993 Minutes Fission product transport can be modeled adequately, e
e Events are selected to bound credible severe accidents designs dependent uncertainties.
C.
Containment Performance Mr.
J.
Donoghue, NRR, s':ated that for this issue the staff is concerned with:
Should the subject advanced reactor designs be allowed to employ alternatives to " essentially leak-tight" containment structures?
The current regulation (e.g., GDC 16) requires that LWR reactor i
containments provide an essentially leak-tight barrier against i
any uncontrolled release of radioactivity.
GDC 38-40 set requirements for containment heat removal, GDC 41-43 for containment atmosphere cleanup, and GDC 50-57 for containment design, testing, inspection, and integrity.
Mr. Donoghue st.ated that the MHTGR is not designed with a leak-tight containment barrier. The design relies on high integrity fuel particles and on a below-grade, safety-related concrete reactor building to provide retention.
The PIUS contaminant is above grade and is designed to have low-leakage rate.
The CANDU 3 is designed with a large,
- dry, steel-lined, concrete containment, without containment spray.
i
F Advdnced Reactor Designs January 6, 1993 Minutes The PRISM containment design is a high strength steel, low
- leakage, pressure-retaining
- boundary, consisting of two components, the upper containment dome and lower containment vessel.
The staff believes that new reactor designs with - limited operational experience requires containment system that provides a
substantial level of accident mitigation for defense-in-depth against unforeseen events.
The staff is recommending the following:
Move away from prescriptive containment design criteria and-e utilize a performance standard based on accident evaluation -
criteria.
Containment must be adequate to meet onsite/offsite radio-e nuclide release limits for DBA.
Containment will be evaluated for a deterministically selected severe accident event.
I D.
Emeraency Plannina Mr. Throm, NRR, characterized this issue as:
Should advanced reactors with passive design safety features be able to reduce emergency planning zones and requirements?
The current regulation 10 CFR 50.47 requires that no operating license be issued unless a finding is made by NRC that there is i
'l
. Advanced Reactor Designs January 6, 1993 Minutes reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency.
The proposed PRISM approach to emergency planning (EP) is significantly different from LWRs, particularly in the area of offsite EP.
A design objective of PRISM is to meet the lower level PAG criteria, such that formal of fsite-EP involving early notification and detailed evaluation planning would not be required.
A The MHTGR also proposed reduced offsite-EP for similar reasons as the PRISM.
For the PIUS, ABB-CE expects that due to the passive safety features, oncite and offsite-EP can be considerably simplified.
The staff recommends the following:
Advanced reactor licensees should develop offsite emergency e
plans.
Provisions for periodic emergency exercises should be developed.
Relaxation from existing LWR requirements may be considered based on accident evaluation review.
Approach will be consistent with passive LWR decisions.
e
m.
I 1
-Advanced Reactor Designs January 6, m3
- Minutes i
l E.
Reactivity Control Mr.
D.
Scaletti, NRR, described this issue as:
Should the NRC accept a reactivity control system design that has no control rods?
Current regulations (e.g., GDC 26) requires that two.indepen-dent reactivity control systems be provided.
One of the.
systems shall use control rods.
-l I
The PIUS design does not have control rods.
The pre-applicant is proposing to comply with the intent of GDC 26 by having two independent liquid boron reactivity control systems.
The staff concludes that a reactivity control system without control rods should not necessarily disqualify a reactor design.
However, the pre-applicant must provide sufficient 1
information to justify that there is an equivalent level of safety in reactor control and protection as compared to a traditional rodded system (e.g., reliability, shutdown margin, operational control).-
F.
Operator Staffino and Function Mr. Throm, NRR, summarized this issue as follows:
Should advanced reactor designs be allowed to operate with a staffing complement that is less than that currently required by LWR regulations?
j i
f Current regulations [e.g.,10 CFR 50.54 (m) (2) (iii) ]. states that' a senior operator must be present in the control room at all
4 Advanced' Reactor Designs January 6, 1993 Minutes times and a licensed operator or senior operator must. be present at the controls of a fueled nuclear power unit.
i The standard MHTGR plant is four reactor-steam generator i
modules and two steam turbine-generator sets.
The plant has three licensed and five non-licensed operators 'for. four reactor-steam generator modules.
The PRISM control room would contain the instrumentation'and controls for all nine reactor modules.
In addition, it is designed for eight licensed operators for nine reactor modules.
The CANDU 3 pre-applicant has not proposed a specific number of licensed operators.
i i
The staff believes that operator staffing may be design-dependent and intends to review the justification for a smaller crew size by evaluating the function and task analyses for
- i normal operation and accident management.
In addition, the adequacy of analyses shall be tested on a control room proto-type.
G.
Mr. Donoghue, NRR, described this issue as follows:
Should advanced reactor designs that rely on a single completely passive, safety-related RHR system be acceptable?
Current regulations (e.g., GDC 34) requires the RHR function to be accomplished using. only safety-grade systems, assuming a l
l
Advanced Reactor Designs January 6, 1993 Minutes loss of either onsite or offsite power, and assuming a single failure within the safety system.
The PRISM design uses the reactor vessel auxiliary cooling system (RVACS) as the safety-grade system for residual heat removal.
Reactor generated heat is transferred through the reactor vessel to the containment vessel outer surface.
RHR is then accomplished through natural circulation heat transfer to the atmosphere.
The MHTGR is designed with only one safety-grade system for removing RHR.
The PIU3 design uses a safety-grade passive closed cooling system (PCCS) RHR from the reactor pool.
The staff recommends the following:
Reliance on a single, completely passive, safety-related RHR system may be acceptable due to unique advanced reactor design features.
Treatment of highly reliable, non-safety related backup systems will be consistent with decisions on passive LWR design requirements.
H.
Positive Void Reactivity Coefficient i
Mr. Throm, NRR, summarized this issue as follows:
Should a design in which the overall inherent reactivity tends to increase under specific conditions or accidents be acceptable?
~-
~.
b
)
Advanced Reactor Designs January 6, 1993
-Minutes Current regulations (e.g.,
GDC 11) requires that the reactor core and coolant system be designed so that in the power operating range the new effect of prompt inherent nuclear feedback characteristics tend to compensate for rapid increases in reactivity.
The overall power coefficient for a CANDU 3 reactor is claimed to be slightly negative and very close to zero.
However, the coolant void reactivity is positive throughout the fuel core lifetime.
The staff is concerned that during a large LOCA at specific locations, void reactivity increases dramatically.
For the PRISM design, should sodium boiling begin on a core-wide basis under failure to scram conditions, the reactor could experience a severe power excursion.
The pre-applicant. claims that for sodium voiding-to occur, multiple failure of redundant and diverse safety-grade systems would be required.
The NRC staff recommends the following:
A positive void coefficient should not necessarily disquali-fy a reactor design.
PRISM and CANDU 3 pre-applicants should analyze the conse-quences of events such as ATWS, Ascrammed LOCAs, and transients that could lead to core damage as a result of.
positive void coefficients.
Consideration of changes in designs to mitigate consequences of these accidents should depend on estimated probability _ of accidents.
~
-1
)
Advanced Reactor. Designs January 6, 1993
' Minutes I.
Control Room and Remote Shutdown Mr.
Donoghue, NRR, described this issue as follows:
Are current control room and remote shutdown requirements fulfilled by a safety grade remote shutdown area and a non-safety grade control room?
Current regulations (for LWR; e.g.,
GDC 19) requires that a control room with adequate radiation habitability be provided to operate the plant safely under normal and accident condi-tions and that there be an ability to shut down the plant from outside the control room.
The control room for PRISM contains the instrumentation and-controls for all nine reactor modules.
The control room structure is not considered safety related and, therefore, the-room is not designed-to seismic Category I design requirements.
The standard MHTGR plant has (for the four modules) a non-safety related central control room to operate the plant and a seismic Category I remote shutdown area.
The CANDU 3 design utilizes a main control room to perform all monitoring ' and control functions for normal and accident conditions, except those events for which the control room becomes unavailable.
The design also has a secondary control area that duplicate some of the capabilities present in the main control room.
.1 l
l
'Advsnced Reactor Designs January 6, 1993 Minutes The central control room for the PIUS design is a seismic-Category I structure.
However, the safety-related systems within this structure are for monitoring only.
The staff recommends the following:
Apply current LWR requirements / guidance until passive LWR e
policy for design requirements of control rooms and remote shutdown facilities is finalized.
J.
Eafety Classification Mr.
J.
Donohew, NRR, described this issue as follows:
What criteria should be applied to advanced reactor designs' to determine safety-related structures, systems, and components.
Current regulations
[e.g.,
10 CFR
- 50. 49 (b) (1) ]
list the criteria to define the safety-related structures, systems, and
-- t components:
e those needed to maintain the integrity' of the reactor coolant pressure boundary (RCPB),
those needed to shut down the reactor, and e
e those needed to prevent or mitigate the consequences of l
accidents.
1 The staff claims that only the MHTGR design is not using the
-i current LWR criteria for safety classification.
The staff recommends the following:
)
i
Advhnced Reactor Designs January 6, 1993
- Minutes Apply the following criteria for determining safety-related e
structures, systems, and components; needed to maintain RCPB integrity,
- needed to shut down the reactor and maintain it in a safe condition, needed to prevent / mitigate accidents that could result in dose consequences comparable to Part 100.
[
i
.=
-Adv6nced Reactor Designs January 6, 1993
' Minutes ACTIO!iS. AGREEMEtiTS, AND FOLLOW-UP ITEMS 1.
The Subcommittee members agreed that the-staff should-have a brief presentation at the January 7, 1993 Full Committee.
The presentation should focus on the draft SECY-paper and general status.
I 2.
The Subcommittee members expressed concern regarding the lack of developing any generic design requirements in the staff proposed program.
3.
Some of the Subcommittee members disagreed with the staff's priority assignment (e.g.,
PRISM. for preapplication safety 1
evaluation report completion date, December 1993).
4.
The Subcommittee members commented that the NRC should state l
what.they.believe is required for adequate advanced reactor designs-instead of asking for more information on what design-1 ers are proposing.
5.
The Subcommittee members expressed concern regarding the lack of sabotage and security requirements and the fact that the staff is not planning to develop any criteria regarding this
~
matter.
i i
6.-
The Subcommittee members believed that particular attention needs to be given to operator staffing of multiple-module designs.
1 Advanced Reactor Designs January 6, 1993
- Minutes 7.
Regarding the criteria for accident selection and evaluation, the Subcommittee members believed that quantitative guidelines should be identified up front and debated by policy makers.
8.
The scope of the draf t SECY paper should be expanded to include other issues such as shutdown risk at low-power operation, the resolution of USIs/GSIs, and external events analysis.
9.
Some of the Subcommittee members believed that phrases such as
" bound credible severe accidents" and " uncertainties" should be quantified and should not be left for the designer's interpre-tation.
- 10. The Subcommittee members indicated that the staff needs a better argument for its position that there must be safety-grade shutdown equipment in the main control room, even if there is such equipment in accessible alternative shutdown areas.
FUTURE ACTIONS Following the receipt of the industry's comments on the draft SECY paper by January 25, 1993, the DOE representatives and the AECL technologies are planning to brief the full Committee at the 394th ACRS (February 11-13, 1993) meeting regarding this matter. The NRC staff will also participate as appropriate.
In addition, the Committee may wish to write its report to the Commission regarding the subject draft SECY paper.
= f. *
'Advsnced Reactor Designs January 6, 1993
' Minutes BACKGROUND MATERIAL PROVIDED TO THE SUBCOMMITTEE 1.
Draft SECY paper:
Issues pertaining to the Advanced Reactor
' (PRISM, MHTGR, and PIUS) and CANDU 3 Designs. and their relation to current regulatory requirements.
2.
SECY-92-393, " Updated Plans and Schedules'for the Preapplica-tion Reviews of the Advanced Reactor (MHTGR, PRISM, and PIUS) and CANDU 3 Designs," dated November 23, 1992.
NOTE:
Additional meeting details can be obtained from a transcript of this meeting available in the NRC Public Document Room, 2120 L Street, NW, Washington, DC 20006, (202).634-3273 or can be. purchased from Ann Riley and Associates, Ltd., 1612-K Street, NW, Suite 300, Washington, DC 20006, (202) 293-3950.
-