ML20215L210
| ML20215L210 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 09/12/1986 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-2433, NUDOCS 8610280530 | |
| Download: ML20215L210 (19) | |
Text
CERTIFIED MINUTES DATE ISSUED: Sept. 12,1986 l
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MINUTES OF THE JUNE 26, 1986 GAS-COOLED REACTOR PLANTS The ACRS Subcommittee on Gas-Cooled Reactor Plants met on June 26, 1986 at 1717 H Street N.W.
in Washington D.C.
The purpose of the meeting was to review the bases used by the NRC Staff in imposing requirements on Fort St. Vrain, particularly with regard to equipment qualification.
The Subcommittee was also to look at the concept of severe accidents at Fort St. Vrain, particularly in light of the recent accident at the Chernobyl reactor in the USSR.
Notice of this meeting was published in the Federal Register on Monday, June 9, 1986 (Attachment A).
A list of persons attending the meeting is attached (Attachment B).
A copy of the tentative schedule of discussion is aise attached (Attachment C).
Material distributed to the Subcommittee during the meeting is listed on Attachment D.
Copies of these documents are on file in the ACRS office.
The entire meeting was open to public attendance.
No written statements were received from members of the public and no member of the public requested an opportunity to make an oral statement.
Two members of the public were present during this meeting.
Attendees AClh HEC _ Staff C.P.
Siess, Gubcom. Chairman K.L. Heitner, NRR/DPWRL-B/SSPD J.C Ebers.le H.N. Berkow, NRR/SSPD D.A. Ward O.D.T. Lynch, Jr. NRR/DPWRL-B/SSPD G.A.
Reed P.M. Williams, NRR/DSRO/SPEB P.G. Shewmar.
N. Wagner, NRR/DPWR-B/PEICSB C.
Mark I. Ahmed, NRR/DPWR-B/PEICSB R.K. Major. Staff J.M. Fehring, INEL/NRC Contractor J.C.
McKinley, Staff C.L. Nalezny, INEL/NRC Contractor J.P. Jaudon, NRC/ Region IV J.A.
Calvo, NRR/PWR-A Eublic_ Service _Co=pany DLCclcruk Eublic R.F Walker. CEn H. Dubroff, The Denver Post H.L Brey D.A. Schlissel, Consultant to the M.H Holmer State of Colorado C.H Fuller M.E Nichof:
P.A DiBenedette, Consultant R.J Burg, Consultant Opening _Sta[Ement Dr. Siess, Subcommittee Chairman, opened the meeting at 1:30 pm on Thursday, June 26, 1986 with a statement regarding the conduct of the meeting, introducing the Subcommittee members present, confirming the proposed schedule and topics to be addressed, and noting that it was sort of a continuatior. of the April meeting, held at Fort St. Vrain, that was cu+ short by a major snow storm.
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h Siess noted that this was an informational meeting and he knew of Dr.
He did not no major licensing actions that required ACRS approval.
intend to take Fort St. Vrain before the full committee.
He also clarified that this meeting did not stem from the Chernobyl incident.
Mr. Ebersole expressed his opinion that the HTGR is a unique type and class of reactor that has a lot of potential that hasn't been realized.
He attributed much of Fort St. Vrain's problems to the water lubricated bearings in the gas circulators and suggested that it might be appropriate for some research oriented ' organization (federal to undertake the development of a more reliable or private)
He recognised the inherent safety characteristics of the circulator.
HTGR and wondered if water cooled reactor logic was being applied to a gas-cooled reactor, perhaps in some cases, improperly.
Siess noted that Fort St. Vrain was a demonstration plant and that Dr.
is has had a number of problems not uncommon to demonstration plants.
The first being the core power oscillations that were not difficult to correct but took a great deal of time.
He noted that the Department of Energy is supporting the development of a modular HTGR concept which will presumably use different circulators (magnetic bearings).
NECZUEB_Erementation 1.
NER, began by defining the design basis accidents for Mr. Heitner, Fort St. Vrain (FSV) and the interaction between these and fire protection and equipment qualification.
Mr. Heitner made a point that the design basis accidents for FSV were different in that they The involved multiple failures of redundant safety grade components.
a permanent loss of all forced circulation, requires first accident, multiple failures of the gas circulators or their power supplies since each circulator is provided with two independent and diverse power Even with a total loss of forced circulation and all offsite sources.
and onsite ac power the operators have about two hours to restore circulation before they have to initiate the alternate cooling method Even after two hours there is no consequence to the health and (ACM).
safety of the public.
Mr. Ebersole thought this was a critical issue, that the licensing bases for the HTGR had to go well beyond that of light water reactors (LWR) to find a consequence to the public.
He thought it was unfair.
Dr. Siess reminded him that the ACRS was party to the requirements.
the Mr. Heitner went on to describe the heat up of the core, reactor vessel (PCRV),
depressurication of the prestressed concrete the assumed failure the functioning of the PCRV liner cogling system, of the coated fuel garticles at 2900 F, the eventual fuel center temperature of 5500 F, the small driving force to push fission and whatever dose reduction may occur as the products out of the PCRV, fission products leak through the reactor building.
He pointed out that if all offsite power was lost and the normal onsite diesel generators failed to start there was still the independent ACM diesel Because of the low fission engine to keep temperatures under control.
product leakage rates the consequential doses from this accident are well below 10CFR100 limits.
2
G s-Cooled Reactors June 26, 1986 In response to a question, Mr Ebersole was informed that the main feedwater pumps were safety grade electric pumps backed up with de controlled turbine driven pumps backed up by a diesel engine driven fire water pump.
Mr. Calvo, NRC, agreed that a total blackout of all electrical power should be looked at.
Mr. Ebersole wanted to know how long the plant could maintain cooling with just the station batteries and no battery chargers functioning.
The second design basis accident is the rapid depressurization accident initiated by the failure of both the inner and outer closures on a single PCRV penetration.
Short term radiation doses from this accident come from the circulating fission product inventory and the lift off of any plated out fission products.
Again the doses are below 10CFR100 limits.
Mr.'Heitner noted that these were mechanistic assumptions and it was assumed that the reactor building was ineffective due to louvers opening or a panel blowing off.
Because the the fuel has performed much better than assumed in the analysis, actual doses would be a small fraction of the calculated values.
Continued cooling after the initial blowout depends on nonsafety grade For air to get to the hot graphite the air must diffuse back systems.
through the opening by which the helium exited.
Mr. Holmes, PSC, indicated that the plant operators were to keep two boiler feed pumps running and circulate helium fast enough so that the core cools down In licensing FSV the NRC Staff analyzed the without any fuel damage.
radiological consequences of the first phase of DBA-2 and assumed that core cooling was successful thereafter.
As a result of the Chernobyl accident the NRC asked PSC to consider PSC postulated the failure of both beyond design basis accidents.
closures on the largest PCRV penetrations at the top and bottom of the PCRV in order to get a convective air flow through the core.
Mr.
Ebersole inquired into the effect of one penetration blowing out and removing the driving force to blow out the second.
He was told that this was an arbitrary assumption made in order to get a flow path.
It was as close to incredible as Dr. Siess had heard.
The oxidation (burning) of the graphite was limited by the flow rate of air through the penetrations and the heat input from oxidation was small compared to that from decay heat.
This event can be terminated by flooding the reactor building up to the bottom of the PCRV, This could be done within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and again the 30 day dose is within 10CFR100 limits.
2.
Eire _Erctectionmantend1LB Mr. Heitner next addressed fire protection in terms of the DBAs.
The NRC Staff accepted the fact that certain fires in FSV would require dependency on the ACM.
There is a congested cable area below the control room and in the two rooms below it where cable separation could not be-achieved and thus a fire there would wipe out all PSC has provided this area with automatic fire protective measures.
In Protection systems to extinguish a fire as soon as it is detected.
addition the cable is covered with Flameastic fire retardant and the third channel of instrumentation is run in conduit.
To compensate for this congestion PSC designed and installed the ACM to be completely independent of this area.
The NRC has tried to maximize the fire protection in order to minimize the probability that the ACM 3
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Gas-Cooled Reectors June 26, 1986 would be called upon to function.
The ACM provides, from a location outside the control room, the ability to insert both the control rods and the reserve shutdown system, continued operation of the liner cooling system, depressurization of the PCRV, operation of the reactor building exhaust system, and associated monitoring equipment.
Heitner reminded the Subcommittee that Appendix R requires plant Mr.
shutdown without radiological consequences no matter where the fire For the DBA-1 scenario using the ACM a small radiological occurred.
consequence is present and the NRC Staff has granted an exemption for All other zones of the plant are being required to meet that event.
These analyses are still being reviewed by the NRC Staff.
Appendix R.
Mr. Ebersole noted that at another meeting earlier in the day he received the impression that the NRC was willing to make concessions with respect to Appendix R if a highly reliable and conservative shutdown cooling system was available.
3.
E2uirment Qualificalisa Mr. Heitner, NRR, explained that, with regard to DBA-1, PSC claims that no harsh environment exists since the only consequence is a slight release of radioactivity, less than the threshold of equipment damage.
main The next consideraticn was a high-energy line break such as; steam, feedwater, cold or hot reheat steam.
PSC has proposed automatic isolation of the high-energy lines if a break is detected.
Since the circulators are steam driven the isolation of the steam lines produces a loss of forced circulation.
It has to be restored by using the water driven pelton wheels.
In addition, water has to be This whole supplied to the steam generators to remove decay heat.
system falls under the scope of equipment qualification and PSC will have to show that the equipment is qualified.
DBA-2 creates a harsh environment in the reactor building where the hot helium comes out.
Any equipment in the reactor building used to mitigate DBA-2 must be qualified.
However, the harsh environment created by DBA-2 is enveloped by the steam line and feedwater line Those temperature profiles will be the ones to which the breaks.
equipment must be qualified.
The equipment in the turbine building is used to mitigate DBA-2 will be operating in a normal that environment and need not be qualified to a more severe environment.
The NRC Staff has accepted this position but it is not yet formalized.
Heitner pointed out that the ACM has no role in the equipment Mr.
qualification issue.
ACM relates only to fire protection.
The harsh environment that the Fort St. Vrain equipment must endure results from the high temperature (1000 F) steam the reactor produces.
Vrain is in a unique position in that its environment is Fort St.
hotter than that for the equipment qualification programs for the light water reactors.
Mr. Heitner emphasized that the licensee elected to use an automatic steam line rupture detection and isolation system (SLRDIS) to limit the environment created by steam line breaks.
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Gaa-Cooled Reactors June 26, 1986 Large breaks will be automatically isolated by +SLRDIS based on ra'te of temperature rise and small ones by the plant operators bases on temperature alarms.
It was noted that in about 1972 PSC made an analysis of high-energy line breaks and that analysis was reviewed by the NRC Staff and accepted.
In about 1973 and again in about 1977 the NRC Staff accepted that the operators would become aware of and be able to isolate steam line breaks by manual action from the control It was assumed that this action could be done in about four room.
It is now the NRC Staff position that no operator actions minutes.
will be accepted in less than 10 mgnutes.
TheSLRDISlimgtsthepeak environment temperature to 300-400 F instead of about 500 F for operator action.
This brings FSV back into the realm of light water reactor qualification.
SLRDIS utilizes automatic actuation of existing hydraulic valves to isolate the steam lines.
SLRDIS is a two train safety system requiring both trains to actuate to cause an isolation.
This is an effort to reduce spurious actuations.
SLRDIS trips the circulators and closes the isolation valves.
For a period of time, it causes a loss of forced circulation until the operators can restore circulation using an alternate mode.
To illustrate the safety significance of the EQ program, Mr. Heitner noted that 350 solenoid valves, 50 transmitters, 50 thermocouples, and 12 electric motors have had to be replaced.
These were components that could not be demonstrated to meet the environment and therefore did not meet the rule.
Mr Holmes, PSC, reminded the Subcommittee that g
in the early '70s a great deal of FSV equipment was tested to 500 F for about 30 minutes and then allowed to cool down but while the temperature was elevated the tests were terminated.
The new environmental qualification regulations require pre-aging of equipment prior to test.
Heitner mentioned the cable splice problem and the need to replace Mr.
taped splices with Raychem insulated splices.
He went on to discuss the cable qualification problems.
The key problem is the inability to specifically identify the exact cables at exact locations in the plant.
An attempt is being made to demonstrate that all of the cable is capable of withstanding the worst environment in the plant.
4.
Eublic_ Service _Ccmnanz of colorade Mr. Niehoff discussed the cable qualification matter a little more.
He discussed the temperature profiles seen at various locations and that the peak temperature in the reacgor building would be poigted out371 F and in the turbine building it would be 360 F.
The pressure profiles are less than one psi.
The humidity drops below 95% in 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />.
Steam condensation on equipment is being taken into account.
The 30-day radiation dose is less than 415 rads and is insignificant.
Submergence-and chemical sprays are not applicable.
No known synergisms have been identified.
About 700 splices have been remade using Raychem insulating materials.
There will be about 3000 splices in the program (See Attachment E).
Mr. Niehoff described the maintenance history review that was to identify those items which were not identical to the as procured state and to determine the impact on their qualification.
This effort has 5
Gas-Cooled Reactors June 26, 1986 resulted in the total replacement of 21 components, partial and additional verification on 39 replacement of 7 components, components.
Mr. Walker, PSC, briefly reviewed the status for the PerformanceHe though Enhancement Program.
In PSC continues to have difficulty filling specific positions.
if he response to Mr. Ebersole's question, Mr. Walker indicated that, was doing it again, he would opt for the modular design HTGR now being designed cooperatively with DOE.
Further, he wouldn't want to be the
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first of a kind again.
Mr. Walker announced that PSC had a new Vice President for Nuclear Mr. Williams is a retired executive Operations, Mr. Robert Williams.
from Rockwell International with a nuclear engineering degree from a master's degree in sociology and graduate work at North Carolina, He started out at the Westinghouse Bettis plant then went to Rockwell International where he held a variety of positions including UCLA.
operation of the Rocky Flats plant.
Mr Walker also announced that he had given up his position asHe reduced his span of President and Chief Operating Officer of PSC.the President an COO, the Executive V control to three people; and the Vice President for Nuclear President and General Counsel, Operations Mr Walker readily acknowledged that the water ingress has been the Actions have been taken to reduce the incidence number one problem.
Mr Brey reviewed but he didn't think the problem was eliminated.
various actions taken or proposed to enhance the performance of FSV.
The primary purpose was to make FSV a better electricity producer, not He described the circulator bearing necessarily to improve safety.
water system and the improvements that have been made (See Attachment One, a He then briefly discussed alternate circulators.
would F).
hermetically sealed electric motor drive with magnetic bearings, in the existing 42 inch diameter cavity and operate at have to fit This not only is a technical challenge but speeds up to 11,000 rpm.
also raises many regulatory issues such as what happens to a canned rotor operating at high speed in 700 psi helium when there is a sudden Costs and schedule are other significant problems.
depressurization?
He also mentioned a number of other alternatives that are being investigated with varying levels of support.
Mr Ebersole reiterated his view that FSV is a good concept that He suggested that PSC suffers from a weakness in a major component.
It was should be subsidized to help solve the circulator problem.
suggested that FSV might be used as a test bed for circulators being developed for the modular HTGR.
A rough estimate to replace all of the FSV circulators is about $20 million and five years time.
5.
Conclusion There are neither plans to bring Siess concluded the meeting.
Dr.
Vrain before the full committee nor to have another Fort St.
Subcommittee meeting in the foreseeable future.
6
?
Qas-Cooled Recctors June 26, 1986
/
NOTE:
A transcript of this meeting is available in the NRC Public Document Room, 1717 H Street N.W.
Washington, D.C. or ca.n be purchased from Ace-Federal Reporters, Inc. 444 North Capitol Street, Washington D.C.
20001 (202) 347-3700 i
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Fedir:1 Register / Vol St. No. '110 / Mcnd:y- )una se1980 /' Notices
'30907 0
Dated June 2.1986.
Dated. june 2.1986.
advised of any changea in achedula.4tc s'
Morton W. Libarkin.
Morton W. Libarkin.
which may have occu,rred.
A ssistant Executive Directorfor Project Assistant Eaecut;nr Directorfor Project gated; gone 2.1988.
3 Reticw.
Review.
Morton W.!Jberkin.
IFR Doc. 86-12922 Filed (MN16. e 45 am)
[FR Doc. as-129 1 Fded 6-6-48; et45 aml Assistant Executim Directorforhwiect sw=c coos rseo-es-as enuwo coos zeeo-ow Review..
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suo cooe mo4w Advisory Committee ori Reactor Advisory Committee on Reactor Safeguards, Subcommittee on Davis-Safeguards Subcommittee on Gas Besse; Meeting Cooled Reactor Plants; Meeting Regulatory Goldes;leeuence and '
l The ACRS Subcommittee on Davis.
The ARCS Subcommittee on Cas Availability Besse will hold a meeting on June 27, Cooled Reactor Plants will hold a 1986. Room 1046.1717 H Street.NW.,
meeting on June 26,1988. Room 1M6.
He Nuclear Regulatory Commission Washington, DC.
1717 H Street, NW., Washington, DC.
has issued a new guide in its Regulatory The entire meeting will be open to ne entire meeting will be open to Guide Series.%is series has been public attendance.
public attendance, developed to desenbe and make The agenda for the subject meeting The agenda for the subject meeting available to the public methods shall be as follows:
acceptable to the NRC staff of shall be as follows:
(i, d i 8 "P parts o T rds une 1
1:00 P.M. until p e"b'sion s Friday, /une 27,19so-o.30 o.m. until the me m
conclusion of business.
cascs, to delineate techniques used by The Subcommittee will review the the staff in evaluating specific problerns The Subcommitfee will review start-8PphcaWy oWRC quimments for or postulated accidents and to provide up activities for Davis-Desse-equipment qualification and cable guidance to applicants concerning Oral statements may be presented by testing and other topics related to Fort certain of the information needed by the members of the pubhc with the St. Vrain, an HTGR.
staffin its review of applications for concurrence of the Subcommittee Oral statements may be presented by perrnits and licenses.
Chairman wntten statements will members of the public with the Regdatory Gqe 156.. General be accepted and made available to the concurrence of the Subcommittee Guidance for Designing. Testing,
Committee. Recordings will be permitted Chairman; written statements will be Operating. and Maintainmg Enussion only during those portions of the accepted and made available to the nm ev ces at Uramum Ms.
meeting when a transcript is being kept.
Committee. Recordings will be permitted MC M pmedurn acceptaWe M k and questions may be asked only by only during those portions of the a
n ng, tesung.
members of the Subcommittee,its meeting when a transcript is being kept'
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consultants, and Staff. Persons desiring and questions may be asked only by emissi n c ntrol devices to ensure the to make oral statements should notify members of the Subcommittee,its reahaWy owr perfonnance.
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views regarding matters to be Comm,ission. Washington, DC 20555.
The Subcommittee will then hear considered during the balance of the Regulatory guides are available for presentations by and hold discussions meeting.
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The Subcommittee will then hear inspection at the Commission's Public Document Room,1717 H Street NW.,
its consultants, and other interested presentations by and hold discussions persons regarding this review.
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before the scheduled meeting to be (5U.SC.5n(W T&
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ACRS SUBCOMMITTEE MEETING ON GAS-COOLED REACTORS LOCATION ROOM 1046, 1717 H ST. NN, WASHINGTON, D.C.
DATE JUNE 26, 1986 ATTENDANCE LIST PLEASE PRINT:NAME AFFILI ATION 2.l, o A. b 3 m e o c-m PSc / DBA hecT I. 'Buee pa,c/TymsaA tlt N R4 D v 6I2c7 F bwNWST MllD A. ScH d556L Schhsel %ks Estu Jesc A. 0 ^ w<
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Rsv. 3 TENTATIVE SCHEDULE GAS-COOLED REACTORS SUBCOMMITTEE JUNE 26, 1986 C.P Siess, ACRS 1:30 pm Oraning R*=+amant 1:45 pm Licensing Bases for Fort St. Vrain K.L. Heitner, NRR Design Basis Accident vs Beyond Design Basis Accident vs Something Else Severe Accident Considerations for FSV vs LWRs (Chernobyl potential)
K.L. Heitner, NRR 2:30 pm Enuipment Qualification Basis for selection of design harsh environment.
Safety vs compliance with written word.
Basis for 35%
power limitation.
Relationship between Alternate Cooling Method, Appendix R requirements, and the Equipment Qualification requirements.
Issue of the consequences of a fire in a vital area vs a fire in a nonvital area Genesis of Steam Line Rupture Detection /
Isolation System, basis for the requirement, effects on plant operation and forced circulation 3:15 pm BREAK M.E. Niehoff, PSC 3:30 pm Status _of_Enuirment_9aalification Effort Approach to qualifying electrical cable.
Splice replacement Component Qualification / Maintenance effects Current Schedule 4:15 pm Eerformance_ Enhancement Program H.L. Brey, PSC Total Responsibility Management Groups, status, what has been learned?
what changes have been made?
what results?
Has a new VP been named? How will this R.F. Walker, PSC affect CEO's relationship with FSV?
4:45 pm Technical Matters of Interest H.L. Brey, PSC Helium Circulator Auxiliaries C.H. Fuller, PSC Exceeding Regulatory Power Limit M.E. Niehoff, PSC Are vital electrical loads normally carried by offsite power or from the station generator?
Protective rqlaying (April 2-3, 1986 events)
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DOCUMENTS PROVIDED TO ACRS SUBCOMMITTEE ON GAS-COOLED REACTOR PLANTS JUNE 26, 1986 1.
Visuals ured by Mr K.
Heitner about 30 pages U.S.
NhC/NRR 2.
Visuals used by Public Service about 24 pages Con.pany of Color +do ATTACHMENT D
4
SUMMARY
OF ENVIRONMENTAL CHALLENGES
- - TEMPERATURE (QUALIFICATION BY TEST / ANALYSIS)
RX BLDG. PEAK 371 DEGREES F TB BLDG. PEAK 360 DEGREES F
- - PRESSURE (N/A FOR FSV)
LESS THAN 1 PSI FOR LESS THAN 1 SECOND
- - HUMIDITY (QUALIFICATION BY IPCEA ACCEPTANCE TESTING)
DROPS BELOW 95% RH BY 15 HOURS RETURNS TO BACKGROUND IN ABOUT 3 DAYS
- - RADIATION (N/A FOR FSV)'
DBA-2 30 DAY TID = 414 RADS HELB 30 DAY TID = 12 RADS
- - SUBMERGENCE (N/A FOR FSV)
LOCATED ABOVE OR PROTECTED FRDM SUBMERGENCE
- - CHEMICAL EFFECTS (N/A FOR FSV)
NO CHEMICAL SPRAYS O
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4 CONSIDERATION OF OTHER ENVIRONMENTAL PARAMETERS
- - SYNERGISTIC EFFECTS (N/A FOR FSV)
NONE KNOWN AT FSV RADIATION LEVELS
- - AGING (QUALIFICATION BY TEST / ANALYSIS)
THERMAL AGING ONLY
- - OPERABILITY (QUALIFICATION BY TEST / ANALYSIS)
FSV EQ PROGRAM UTILIZES 30 DAY OPERABILITY TIME
- - CABLE DESCRIPTION / CLASSIFICATION FORT ST. VRAIN CABLES ARE GROUPED INTO THREE CLASSIFICATIONS CLASS I (6QUALIFICATIONBINDERS)
CLASS II (13QUALIFICATIONBINDERS)
CLASS III (4 QUALIFICATION BINDERS) k77W&MMEW E 1
RAYCHEM SPLICES INSTRUMENTATION CIRCUITS APPLICATIONS:
(MILLIAMP & LOW VOLTAGE)
MOTORTERMINATIONS(480V)
STATUS:
APPR0XIMATELY 700 SPLICES INSTALLED APPR0XIMATELY 3,000 SPLICES IN PROGRAM REVIEW EFFORTS:
KIT SIZE SELECTION INVOLVES MEASUREMENTS BY Q.C.
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