ML20062D948

Summary of 781103-04 Meetings Re Full Pwr Oper of Subj Facil.Purpose of Meeting Was to Provide Local People Opportunity to Observe Tech Discussions & Discuss PSC Plans for Fluctuation Testing Above 70% Pwr.W/Encl Agenda

ML20062D948
Person / Time
Site:	Fort Saint Vrain
Issue date:	11/21/1978
From:	Kuzmycz G Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 7812010157
Download: ML20062D948 (92)

Text

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WASHINGTON, D. C. 20555 i( \\v / NOV 211978 DOCKET NO.: 50-267 LICENSEE: Public Service Company of Colorado (PSCo) FACILITY: FortSt.Vrain(FSV)

SUBJECT:

SUK4ARY OF MEETINGS HELD ON NOVEMBER 3 AND 4,1978 TO DISCUSS FSV ITEMS ( Meetings were held in Denver, Colorado on November 3 and 4,1978, to discuss various items related to full power operation of FSV as per the agenda given in Enclosure 1. The purpose of the meetings was two-fold: (1) to provide local people an opportunity to observe technical ,-discussions normally held in the NRC offices in Bethesda, Maryland, I and (2) to provide PSCo the opportunity of discussi.ng: their plans for fluctuation testing above 70% power, inspection of reactor internals during the refueling outage, plans for further instrumentation and , testing after refueling, a possible remedy fer the fluctuations, and other prerequisites for operating at full pcwer. The open meeting was held on Friday, November 3,1978, from 6:00 p.m. until 9:15 p.m. and Saturday, November 4,1978, from 9:00 a.m. until 4:30 p.m. with a break for lunch between 12:00 noon and 1:00 p.m. On Saturday. There was a maximum of 11 persons from the general public, two press reporters and one TV cameraman on Friday night. After the meeting on Friday, TV stations used the meeting room as background for ( a special newsreport on the meeting. On Saturday there were four members of the public, two press reporters and one TV reporter present at the meeting. There were more persons present at the meetings on Friday and Saturday than the official participants and the public, but the additional people not included in the talley were actually PSCo personnel and their families. The complete list of attendees is pre-sented in Enclosure 2. Roger J. Mattson, Director of the Division of Systems Safety, opened both the Friday night arid Saturday morning sessions of the meeting with several introductory remarks on the purpose of the open meeting and the NRC policy on these meetings. A short background of FSV was then presented followed by a description of the topics to be discussed. Dr. Mattson then introduced the NRC staff members after which Fred Swart introduced the PSCo members and George Wessman introduced the GAC members. 781201 0157 {7

NOV 211978, The following topics were discussed: 1. Plant status. Due to the low power operation of FSV during the last few months, the refueling outage has been delayed until March 1,1979, so that the initial cycle fuel would have 200 EFPD of burnup. Until that time, PSCo intends to perform its fluctuation testing, both below and above 70% power if possible, and return to about 70% power for steady state operation. The plant is now scheduled for refueling from March until May although the exact date is dependent upon system demands. Until the refueling it is PSCo's intent to finish fluctuation testing as described in RT 500 and RT 502. During the refueling outage, instrumented control rod drives and a scratcher pad to detect motion will be installed. After the refueling, tests RT 500 and i RT 502 will be repeated with instrumented control rod drives (CRD). During the second refueling, estimated to take place in January 1980, the instrumented CRD and scratcher pads will be removed, / the test results evaluated and core restraints (Luci locks) f installed to control the fluctuations. Mattson inquired if there is any benefit to going above 70% power before tim refueling outage when installation of the instrumented wCRD<and inspection of the core is performed. Swart (PSCo) answered that testing above 70% power is needed to establish a trend in the fluctuation characteristics and the threshold line for start of fluctuations at higher power levels. PSCo feels that visual observation of Region 35, which is scheduled to be inspected during the refueling outage, will not show anything different from the base line observations since TV observations and visual inspection of Region 34, performed earlier this year, did not show any damage and both regions are next to each other. i 2. New data from testing below 70% Jower. Lou Johnson described the additional instrumentation that 1as been added at FSV, along with data collected during the October 6,1978, fluctuatiori. This data is presented in Enciosure 3. The data on displacement probes shows equal and greater displacement at 0% power than some during ( fluctuations at power. GA's contention is that the correlation between displacement probe data and core motion is not as strong as first believed. The displacement during the reactor trip is probably due to valve closure and flow stoppage. Also, GA believes that the displacement probes might not be 100% reliable. l l l l A

r i . NOV 211978 i 4 The October 6 fluctuation was unintentional and occurred during core exit temperature testing with 3 circulators on-line and a ap of 2.65 psi. The fluctuation was repeated with 4 circulators on-line and a ap = 2.9 psi. The data shows that only seven regions were involved in this fluctuation and not the entire core. The nuclear characteristics and period have remained as before. GA feels that the " cleaner" data is a result of limited motion of blocks in only the seven regions. No attempt was made to terminate i the fluctuations since they stopped by themselves after about 1.5 hours. Public Service Company staff were asked to comment on the safety considerations of the fluctuations. Swart stated that the average l \\ power and temperature remain relatively constant during the fluc-tuation testing and the amplitude is small. The plant technical specifications have not been exceeded during fluctuations. The l plant is not normally operated in the fluctuating mode and all plant operators have been instructed to terminate any unintentional fluctuations immediately. Due to the above reasons PSCo does not consider the fluctuations to be of safety significance. Roger Mattson responded by stating that the NRC considers the fluctuations a safety issue. More data and information is necessary even though

• the fluctuations can be controlled, they are, on the average, small, but they are not expected -- this is the rationale for their safety significance.

) 3. Response to NRC Questions' Dated August 18, 1978. a. Has any further work been done on coupling core hydraulics to ( core motion? This correlation is not as strong as originally thought. b. Have the flow conditions that existed at the low power October 6 fluctuation been observed during other tests? The ~ flow conditions were observed for the first time on October 6. c. Did the data from the October 6 fluctuation fall in the stable operating range? The data was close to the stable range. d. Is dowel brekage considered as a probable cause of the fluctuations? Existence of broken pins is not borne out by the new October 6 data. l i m,

. NOV 21 1973 Prior arguments centered on a statistical probability; if broken dowels do exist, coolant blockage would result in increased fuel failure and fission product release. However, to date, the primary system has been operating below the design limits by a factor of 30. Mattson stated that he now doubts that dowel breakage exists in the core and is the cause of the fluctuations. Phillips stated that the response to Question 3 was sufficient to eliminate concern over broken dowels. However, the NRC is still very interested in the con-dition of the dowels and this should be examined during the refueling inspection. e. Mike Tokar stated that the answers provided for question 5 were at temperatures at which fuel failure would be guaranteed. Would not a 25% increase in fission product release well beyond s the nope band of normal operation and the increase in circu-lating activity definitely be observed? Swart replied that the 25% increase in activity corresponds to 70% flow blockage through a column. Tokar then asked if the planned instrumen-tation would provide a better magnitude of the impact forces invol ved. GA responded that they are not sure. Since the displacement probe is not a strong indication o'f the forces involved, the magnitude of the forces thus calculated was used only as an upper boundary. If one uses a restitution factor either greater or less than that originally used, different values of the magnitude of the forces involved can be derived. So far, there are no mechanical instruments available, however plans include evaluating various instruments and their appli-cability. Simon stated that the best indicator of the forces involved is the rise time of the nuclear channel which corresponds ( to the opening of a gap since this gives consistant results as to the magnitude, f. Phillips requested that the data from other fluctuations be added to Figures 7-2, 7-3 and 7-4 of the response to Question 7. Also, a figure of the region outlet temperature variations during fluctuations as a function of reactor power should be added. Sheron questioned if divergent fluctuations are possible and if automatic scram is provided in the event that they are. It might be possible to trip one of the PPS logics in order to have one-out-of-two prior to fluctuation testing. Swart replied that PSCo would not do this since the loss of protection is not worth the risk.

l l I - 5 NOV 21 1978 i i i g. Phillips felt the response to Question 8 was inadequate. He asked if control rods could be inserted during the fluctuation testing. Swart responded that the fluctuations are normally terminated by control rod insertion and resultant power reduction. Also, insertion of one control rod more than 2 feet into the core would be a violation of the technical [ specifications. When asked if a total scram could be per-formed. Swart replied that it could but PSCo would not want t to do this since it would impose a very severe transient on { the plant. l A question was asked as to what constitutes a successful test above 70% power? Swart answered that the purpose of a successful test is to establish the stable operation threshold line and to under- [ stand the mechanism of the fluctuations. Long range plans based on testing would include determining a "fix" for the fluctuations. y 4. Planned Inspection During Refueling Outage. Fred Swart presented i i PSCo plans for inspecting core internals during the refueling i outage. These plans are shown in Enclosure 4 One of the fuel regions that will be removed during refueling is Region 35, in i

the north-west quadrant. The inspection will check Region 35 core components for indication of fluctuation related motion and any damage. The inspection will be performed in the hot cell and reactor and includes visual and functional inspection of the thermal barrier, core barrel and keys, upper plenum elenents orifice valves and their motion, control rod guide tube, top and bottom reflector blocks, fuel elements dowels, and gaps between blocks.

( 5. Tentative Plans for Remedy of Fluctuations. Don Warembourg showed a model, in quarter scale, of PSCo's proposed core restraint locks (.calledLucilocks). These Luci locks would be fitted into the i handling holes of three adjacent fuel columns and hold these columns j together. See Enclosure 5. A total of 83 Luci locks are required t to hold the entire core together. PSCo plans call for installation of these Luci locks during the refueling outage of 1980. Wessman l stated that detailed plans and a safety evaluation for the Luci i l locks will be presented to the NRC e-srly in 1979, and that they are already being manufactured. Mattson stated that PSCo might incur some risk in beginning manufacture of these devices before i receiving NRC approval on design and safety. George Wessman presented a summary of expected plant conditions l for fluctuation test RT 502 (Enclosure 6 ). i I l i \\ . ~

N0Y E I '"79 ! l 6. Other Prerequisites for Increased Power level. A. Analysis of loss of forced circulation Dick Ireland reported on the status of LOFC analyses, and a brief background. According to independent analyses, it was detennined that the helium purification system is now capable of handling the reactor depressurization. Before granting approval of the analyses, NRC requested to :ee the heat curve used in the GA analyses. B. Accident Reanalysis Several questions were asked on the RATSAM code input to the RECA code. Swart agreed to provide this additional information on the sensitivity of RECA results based on RATSAM input. Then, several questions were asked on the core thermal safety limi t. Sheron inquired whether the consequences of accidents initiated at power levels below 100% but with power to flow ratios greater than 1.05 are worse than already analyzed. Wessman answered that the analysis were performed at 105% power and P/F ratios of 1.0. The technical specification Safety Limits in SL 3.1.1 and figures 3.1-1 and 3.1-2 were discussed. The curve of power to flow number versus thermal power, is set up for an equal maximum fuel temperature along the curve. The decay heat for lower power levels would, by necessity, be lower even after only a few minutes of core cooling. The core can be run at full power and a power to flow ratio of 1.05 by technical specifications and these initial conditions result in less severe consequences than the analyses submitted. Sheron inquired about the results of analyses done at power to flow ratios greater than those given in the core safety limit curves in the technical specification. Wessman responded by saying that those conditions were considered to be degraded i plant conditions which, under regulations, do not require accident analyses. C. The moisture monitor technical specifications are to be modified to include a revised monitor flow rate. 1

, e I NOV 21 1978 0 D. Instrument Surveillance and Calibration After the January 23, 1978, incident, PSCo comitted to survey their instruments on system 21 and establish a surveillance Ij, and calibration program. This program has been submitted to NRC Region IV for review and approval has been granted. E. Post Refueling Comitments Swart presented a status report on the alternate cooling method, the fire protection system, and the fire water booster pumps. Ireland requested additional information on the pumps them-selva, their specifications and electrical hook-ups. This information will be provided by PSCo. 7. Public Questions and Coments A. Cheng Shih - Has the possibility of a resonance condition been considered or thermally induced vibrations? The natural frequency of. vibrations should be considered in evaluating the course of the fluctuations. What causes the core to move in a 10 minute cycle, wait and then move back into place? Mattson replied that no evidence of resonance has been observed 'Ein the data to date. Ross stated that the helium transit time around'the primary circuit is much less than the period of the fluctuations. GA e.<plained that bowing of the fuel columns is due to varying thermal espansion of graphite, this changes the flow path by opening or closing of gaps and subsequent reversal of bowing. ( B. Paul Bendt from thd Solar Energy Research Institute - What is the feedback mechanism for the fluctuations and can it be a nuclear feedback? Lou Johnston (GA) answered that the mechanism is one of thermal feedback; the signal shape recorded by the nuclear channels does not correspond to the shape associated with a nuclear feedback mechanism. Nuclear feedback would be recorded on all channels instead of being isolated on one or two as the fluctuation signals have been observed. C. Kevin Markey, Friends of the Earth - What are the long tem t implications of the fluctuations on the plant reactivity and 4-control, on the deterioration rate of core components, on the primary system activity level and the long term reliability of the plant? Mattson. ced that the long term implication on ' the areas in question are'being investigated by all parties concerned but the NRC feels,that short term testing would have no effect. J 4 1 t, / \\ g, \\ h ,e

. NOV 21 '978 Markey then asked questions on ultimate capacity of the plant, usefulness of the plant to the rate payers and stockholders and black out due to.FSV unavailability. Swart responded that 'the meeting is a technical one on fluctuation and not rate structure and black outs and that he is not prepared to respond to questions along those lines. Markey then a*sked why PSCo will not refuel and inspect the core before the 200 EFPD. Mattson suggested that further testing would be required to understand the fluctuation phenomenon and the information gained during testing might be more useful than that obtained from a visual examination. Swart stated that the data obtained during testing might help tc interpret the observations made during the refueling outage inspection. Also, Swart stated, operating the plant to 200 EFPD is a matter of economics, PSCo does not want to pay for the unused uranium. With this comr.ent Markey stated that PSCo might place more emphasis on manetary value than on safety considerations. Mattson repl'ed that by Federal law, safety considerations are of prime importance, above financial and any power needs. 8. NRC Requests i r At the conclusion of the open meeting, Roger Mattson requested formal documentation and submittal of the following items by PSCo. 1 A.- A written explanation of PSCo's intent and strategy for pro-ceeding to full power operation while dealing with the fluctuation problem. B '. A schedule for implementing the above strategy. C. A formal commitment not to operate the reactor in a fluctuating mode except during approved testing. ( l D. Submit a report on the results of the below 70% fluctuation l tests for NRC review. Mattson stated that another meeting would probably be held to discuss these results. Also the NRC would require completion of below 70% testing prior to approving above 70% testing. , E. 'A formal statement of PSCo's comitment to follow the established start-up tests above 70% power. 1 a

! NOV T' "' F. Copies of the viewgraphs used during this meeting. G. A written summary of what has been learned from the displace-ment probe (DP) data. This should include a description of changes which have been made to the DP system and an explanation of why correlations between DP data and in-core motion are discounted more today than they were in ^5e status report, H. Further discussion of the effects of the sack of correlation between the onset of fluctuations and a drop in core ap on the lateral pressure gradient theory fo fluctuations. Provide data points from all fluctuation events for figures 7-2, 7-3 and 7-4 of the response to NRC question #7 for Testing Above 70% Power. Provide plots of similar data of fluctuations in average region outlet temperature and average core power vs power level. These items were requested by Phillips. I. Confirmation of the moisture monitor flow specifications agreed upon. J. Results of sensitivity studies performed by GAC on helium and fuel temperatures vs RATSAM calculated pressure histories. This was requested by Sheron. K. A written explanation of-how PSCo will provide margin sufficient to assure that the reactor will be operated outside the fluctu-ation regime. It should be demonstrated that the proposed margin, which might be a function of power level, is sufficient, Roger Mattson recommended that all of these items be submitted as early as possible in order to expedite the review process. , is a copy of the background material provided by the NRC to the general public at the start of the open meeting. <d.... s George Kuzm'ycz, Project Manager Advanced Reactors Branch Division of Project Management Office of Nuclear Reactor Regulation Enclosures l F

i .}_ - ENCLOSURE 1 U. 5. fluCLE % REGULATORY CO*.4!$510fi NUCLEAR REACTOR REGULATIO*; IfiFORMAT!0h MEETING ON FORT ST. VRAIN ITEf45 Friday, November 3,1978 6:00 - 9:30 p.m. 1. Introduction and Background 2. Discussions of plans for fluctuation testing above 70% power a. Plarlt Status b. Response to NRC Questions dated August 18, 1978 c. Test conditions d. Safety implications of tests e. New data from testing below 70% power f. Summary 3. Public comments and questions (8:45 p.m. - 9:30 p.m.) Sa turday, November 4,1978 9:00 a.m. - 5:30 p.m. 4. Introduction and Background 5. Planned inspection of reactor internals during January refueling outage 6. Plans for instrumentation or further tests after refueling; tentative plans for remedy of fluctuations 7. Public connents and questions (11:15 - 12:00 noon) 8. LUNCH 9. Other prerequisites for increased power level a. Analysis of loss of forced circulation b. Accident reanalysis - response to NRC questions dated May 24, 1978 and September 19, 1978. c. Moisture monitor specifications d. Instrument set points

10. Post refueling commitments a.

Auxiliary Cooling F.ethod .b. Fire protection c. Booster pump

11. Summary,of comitments made during Saturday Sessions g

12. PublJccommentsandquestions(4:45p.m.-5:30p.m.)

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n DATA FROM OCTOBER 6. 1978 TEST BEING PERFORMED TO INVESTIGATE REGION EXIT TEMPERATURE MEASUREMENTS. CORE REGION ORIFICES NEAR CLOSED - CORE RESISTANCE HIGH - MOST CONTROL SYSTEMS IN MANUAL "A" CIRCULATOR SHUTDOWN. FLUCTUATIONS STARTED 0255 AM WHEN REGION 37 AND 26 ORIFICE VALVES WERE ~ OPENED AND CONTINUED TILL 0445 AM.

• MAXIMUM STEAM TEMPERATURE FLUCTUATION WAS LESS THAN 10*F.

e

l e 5.0 - ( / / 4.0 - / S FLUCTUATIONS HAVE 2. OCCURRED IN THIS [ REGION (<2 DAY TOTAL 3.0 - ( WO 8 OPERATION WITH NO,FLUCTUATIO.':S ( 2.0 - 4 57 DAYS 116 ~ DAYS l 1.0 - 63 i MYS 35 29 DAYS 17 DAYS t I DAY,5 0 i 30 40 50 60 70 80 5 POWER i i

l s FLUCTUATION INITIATION POINTS 5.0 4.5 - - t e 4.0 - e e a 2 eeee 3.5 - a.* e E l e 8 e e e e e ee 3.0 i e 10/6/78 e l 2.5 e e 2.0 l 25 30 35 40 45 50 55 60 65 70 % POWER 10/10/78

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A W 4 l NRC HANDOUT FOR GENERAL PUBLIC o' ( l I

7 i 1 UNITED STATES NUCLEAR REGULATORY COMMISSION NOVEMBER 3-4, 1978 MEETING ON THE FORT ST. VRAIN NUCLEAR GENERATING STATION t CONTENTS ( Enclosure A Agenda Enclosure B N;C Participants l Enclosure C General Plant Description Enclosure D Background on Agenda Topics / ( i l I i I l l

ENCLOSURE A U. S. IUCLEA PErdiL ATORY C07.M! $.'10f. NUCLEAR REACTOR REGULATlDi; II.F0FJGi!0N "EEi: fig ON FORT ST. VRAIN ITEMS Friday, November 3,1978 6:00 - 9:30 p.m. 1. Introduction and Background 2. Discussions of plans for fluctuation testing above 70% power a. Pladt Status b. Response to NRC Questions dated August 18, 1978 ~ c. Test conditions d. Safety implications of tests e. New data from testing below 70% power f. Summa ry ( 3. Public comments and questions (8:45 p.m. - 9:30 p.m.) Saturday, November 4,1978 9:00 a.m. - 5:30 p.m. 4. Introduction and Background 5. Planned inspection of reactor internals during January refueling outage 6. Plans for instrumentation or further tests after refueling; tentative plans for remedy of fluctuations 7. Public comments and questions (l'l:15 - 12:00 noon) 8. L LFH (,; 9. prerequisites for increased power level r .nalysis of loss of forced circulation 4ccident reanalysis - response to NRC questions dated May 24, 1978 and September 19, 1978.

1. oisture monitor specifications nstrument set points efueling coanitments t

axiliary Cooling Method b. Fire protection c. Booster pump 11. Summary,of commitments made during Saturday Sessions ' comments and questions (4:45 p.m. - 5:30 p.m.) l 12 Put

ENCLOSURE B NUCLEAR REGULATORY COWiiSSION PARTICIPANTS AT THE NOVEt18ER 3-4,1978 MEETING ON FORT ST. VRAIN Dr. Roger J. Mattson, Director, Division of Systems Safety Dr. Denwood F. Ross, Jr., Deputy Director, Division of Project Management Mr. George Kuzmycz, Project Manager for Fort St. Vrain, Advanceo Reactors Branch, Division of Pr.oject Management Mr. Laurence E. Phillips, Section Leader, Analysis Branch, Division of Systems Safety i Dr. Brian Sheron, Senior Reactor Engineer, Analysis Branch, Division of Systems Safety Dr. Michael Tokar, Reactor Engineer, ruels Section, Core Performance Branch, Division of Systems Safety Mr. James J. Watt, Senior Reactor Engineer, Reactor Systems Branch, Division of Systems Safety Mr. Richard E. Ireland, Technical Advisor to Assistant Director for Reactor Safety, Division of Systems Safety Mr. Maynard W. Dickerson, Reactor Inspector, Office of Inspection and Enforcement, Region IV NRC Consultants: /( Mr. Dwayne Dry, Instrumentation and Controls Division, Oak Ridge National Laboratory Dr. Charles A. Anderson, Leader Group Q-13, Los Alamos Scientific Laboratory i

l ENCLOSURE C GENERAL PLANT DESCRIPTION The Fort St. Vrain (FSV) reactor is a High Temperature Gas-Cooled Reactor (HTGR) of a small consnercial size rated at 841 MWt (330 MWe). It is located at the confluence of St. Vrain Creek and the South Platte River near Platteville, Colorado, approximately 35 miles NNE of Denver. Figure 1 shows a cross sectional view of the plant. General Arrangement i Fort St. Vrain is a thermal reactor employing Uranium-235 as fissile material ( and Thorium-232 as fertile material, graphite as moderator, cladding, structure, and reflector, and helium as the coolant. The active core, a cylinder 19.6 feet diameter and 15.6 feet high, is surrounded by graphite reflector elements which have a thickness of 3.3 feet at the top and 3.9 feet at the bottom and around the sides. The core, contained in a steel core barrel which provides lateral restraint and support for the fuel and reflector elements, rests on two foot thick graphite core support blocks. These are keyed together to maintain uniformity and stability and are raised off of the conc. rete core support floor by means of graphite support posts that are 6 inch diameter and 41 inches long. (Figure 2) At rated power the helium coolant, at a pressure of about 700 psia, flows i downward through vertical passages in each of the fuel elements, where it is 0 heated from 762 F to 1444 F, and then into the plenum area under the core. i The coolant subsequently passes through the stea'm generators, after which it flows to the suction of the helium circulators. The discharge from the circulators is routed to the top of the reactor core ia an annular space provided between the core barrel and the liner. (Figure 3)

Two identical loops are used, each including a six-module steam generator and two helium circulators. Each loop contributes half the total output of the nuclear steam system, which produces steam at 2400 psig and 1000 F with single reheat to 1000 F. The power conversion system is shown in Figure 4. Prestressed Concrete Reactor Vessel A prestressed concrete reactor vessel (PCRV) contains all the major components of the nuclear steam supply system: the reactor core, steam generators, helium circulators. The PCRV contains and also shields the reactor and ( the entire helium primary coolant system; it is constructed of high-strength concrete reinforced by bonded reinforcement steel and prestressed by steel tendons. The main cavity of the PCRV is an upright cylinder about 31 feet diameter and 75 feet high. The flat upper and lower heads are nominally 15 feet thick and the walls are 9 feet thick. The top head contains refueling penetrations (which also house the control rod drives) and an access penetration for installation of reactor components. (Figure 3) The PCRV concrete walls and heads are constructed around a carbon-steel ( liner for the internal cavity and penetrations. This liner is anchored to the concrete and provides a helium-tight membrane seal. In order to control liner and concrete temperatures a system of water coolant tubes is welded to the concrete side of the liner. The PCRV inner cavity liner and the primary closures serve as primary containment for the reactor; the massive 1 PCRV and the secondary closures act as the secondary containment. Fuel The FSV reactor core is composed of 1482 hexagonal graphite fuel elements,

(Figure 5)l4.2 inches across the flats and 31 inches high, stacked in vertical columns, and keyed top and bottom with dowel pins and sockets. The active core is made up of 247 columns of fuel with six individual fuel elements vertically stacked in each column. The top and bottom reflectors on each column contain three hexagonal blocks each. The fuel columns are grouped into 37 individually orificed fuel regions, each containing seven columns; the central columns of each region contain the control rods. The fuel material consists of small particles of fissile thorium and uranium ( carbide (0.0118 inch diameter) and fertile thorium carbide (0.0236 inch diameter), each particle coated with pyrolitic carbon and silicon carbides in four layers. These coatings provide a fission product barrier and impart strength to the particles. The particles are then cast into rod fonn approximately 1.5 inches long and 0.4 diameter. Completed rods are then inserted and sealed into 210 fuel holes in the graphite fuel element. Dowel pins on the top of each fuel element mate with recesses in the fuel element above to insure the continuity of 108 coolant passages in each fuel element. I

ENCLOSURE D BACKGROUND INFORMATION ON AGENDA TOPICS FOR FORT ST. VRAIN MEETING NOVEMBER 3-4, 1978 Purpose of Meeting This meeting is being held to exchange technical information between the Nuclear Regulatory Commission (NRC) and the Public Service Company of Colorado (PSCo) in connection with proposed operation of the Fort St. Vrain reactor above 70% of rated thermal power. Before any action is taken, the NRC seeks to ( establish that certain outstanding technical review matters are sufficiently well resolved that there is assurance that such higher power operations can be accomplished safely. 4 Much of the meeting will focus on the core temperature fluctuations which have been observed in the reactor under some operating conditions. Some technical matters gennaine to full power operations, but not related to the temperature fluctuations, will also be discussed. ( Finally, there will be a short discussion on the status of certain plant modifications which are to be completed during the first refueling outage. A brief orientation on the nature of the principal agenda topics isroutlined below. Friday Evening, November 3, 1978 l Temperature Fluctuations l Cyclic temperature fluctuations were first noted at the Fort St. Vrain reactor on October 31,1977 at 58% power during the initial rise in power above the previously authorized 40% level. Subsequent fluctuations have been observed under a variety of core conditions and at power levels between 40% and the m

. present limit 70%. As of August,1978 a total of 20 separate core fluctuation events for a total of about 40 hours of cyclic operation had occurred at Fort St. Vrain. The fluctuations were observed in core cutlet helium temperatures, external themal neutron flux, and steam temperatures. Temperature fluctuation's have remained within design and Technical Specification limits, and have been non-divergent and reproducible. The fluctuations are core-wide and out-of-phase from one refueling region to ( another with a range in period from 5 to 20 minutes. The average core themal power and average helium temperatures remain relatively constant during the fluctuations. Conditions have been established which pennit operation of the reactor for routine power production up to 70% of rated power without fluctuations. The reactor is operated in a non-fluctuating mode except when special tests are to be done under carefully controlled and monitored conditions. The NRC has authorized the conduct of such tests below 70% power. C' The precise cause of the fluctuations is not yet completely understood. However, it is quite clear from available evidence that some degree of core component motion takes place during the fluctuations. This may be caused by small lateral pressure fo'rces which can be developed when the reactor is operated with a "high" pressure difference across the core. ~ Figure 2 shows the general core arrangement. Note that coolant flow goes from top to bottom. Total flow is increased as reactor power is increased by increasing the speed of the helium circulators. This increases the pressure difference across the core from top to bottom. Each one of the 37 " refueling regions" shown in Figure 6 consists of seven fuel columns

. made up by stacking graphite fuel elements on top of one another. Each one of these regions contains a flow control valve at the top which regulates flow to that region. These valves also affect the pressure difference across the core. Most of the coolant flow goes through the fuel columns to remove nuclear heat. However, a small fraction goes down the gaps between the regions as bypass flow. This is in accordance with the design; however what was not evident during the design stage was the possibility that lateral pressure differences could be large enough to push the fuel regions sideways a fraction of an inch to take up the available clearance. Of course there must be some associated differential expansion effect to reverse the motion t to get a cyclic behavior. Figure 7 shows an example of the typical behavior of thermometers which measure temperatures at the exit of the refueling regions. At issue in connection with the meeting are questions related to the structural margins in graphite to accomodate small impacts between components and the precautionary measures that will be used during tests above 70% power. While it is not expected that the characteristics of the fluctuations ( will change from what they have been in the past, and margins should be adequate, the tests are being approached with deliberate caution. This is due to the fact that the reactor will be operated at a power level it has not reached before. Also it is not known whether fluctuations can be avoided at higher power levels. Of course, one of the objectives of 'he t proposed tests is to find this out. Saturday Morning, November 4,1978 (Some of Friday evening's discussion may be continued on Saturday morning l since the scheduled topics will go quickly and Friday's discussions may I not be complete.) l l t

i Planned Inspection During Refueling The reactor is tentatively scheduled to be shutdown for its first refueling in January 1979. New fuel elements will be put in several regions of the reactor at that time. Of particular interest will be the nature of the inspections that will take place in order to find out if there is any sign of the effects of the fluctuations on any of the graphite components that are taken out. A good deal of attention will be placed on region 37 { which is in the core quadrant where the fluctuations appear to originate and the temperature swings have been the greatest. Plans for Instrumentation and Further Tests After Refueling; Tentative Plans for Remedy of Fluctuations It is questionable whether sufficient infonnation can be obtained during fluctuation testing before the refueling outage to fully diagnose the problem. In order to plan its activities, the NRC seeks advance information on the nature of special instrumentation that may be placed in the core while the plant is shutdown. We know for example that pennission may be sought to (' install a special instrumented assembly in the core, and there may be plans for the installation of other instruments. Whether or not there are any tentative plans for remedial measures to control the fluctuations is also of direct interest to overall planning. Saturday Afternoon - November 4,1978 (Other Prerequisites for Increased Power Level) Analysis of Loss of Forced Circulation Although the possibility that the reactor could totally lose the capability l l to circulate coolant is very remote because helium circulators can be driven l by steam (from the reactor or an auxiliary boiler) or by water supplied l

from various sources to a back-up pelt'on wheel drive, such an event is une of the postulated design basis accidents the reactor is designed to acconinodate. If forced circulation capability were lost while the reactor were operating at high power, it would be necessary to remove coolant from the reactor (depressurize) in order to assure that the shutdown heating from fission product decay fs properly distributed to the cooling system of the reactor vessel. To depressurize, the helium is passed through a helium purification system which removes radioactivity released from the fuel as it ( heats up so that only purified helium is vented to the plant stack. The General Atomic Company evaluated the capabilities of the "as-built" purification system and determined that the piping was too small to depressurize the reactor, if it were at full power, before fission product release from the core were so high that the Low Temperature Adsorber (cooled by liquid nitrogen) could no longer retain all the fission products passing into it. Modifications were made and tests performed to verify that depressurizat. ion could be done in about seven hours. The NRC staff k recently completed an independent evaluation of the capabilities of this modified system, and detennined that it is now acceptable. We will discuss the results of this evaluation during the meeting. The purification system depressurization path is shown schematically in Figure 8. Accident Reanalysis In connection with evaluating the capability of the as-built fire water system to supply water to the steam generators and to drive the pelton wheels on the helium circulators, General Atomic noted a discrepancy between the Technical Specifications and the assumptions used in the original analysis. l

l Reanalysis of those events that could be affected was done using assumptions consistent with the Technical Specifications and employing an analysis method different from that originally used. The results Indicated that for two events of particular interest - cooldown by a fire water driven circulator and a design basis rapid depressurization accident - fuel temperatures and insulation cover plate temperautres would be somewhat higher than previously predicted, if the reactor were operating at full power at the time of the event. At 70% power it has been determined that the ( predicted temperatures would not exceed those of the original analysis and are therefore acceptable. For the full power reanalysis, the NRC staff has undertaken a detailed review in order to assure that the new analysis method gives acceptable results and that predicted temperatures would not exceed safety limits. This review is nearing completion, and at this meeting PSCo's response to several staff questions will be discussed. This evaluation pertains principally to plant operation in a steady mode, but some aspects of the ( fluctuation problem may be discussed. Moisture Monitor: Specifications The plant protection system contains moisture monitors which are designed to l automatically detect a large leak in one of the steam generators, identify which one and isolate and dump the water in it to a tank. The reactor.is also automatically shutdown. This action protects the graphite in the l l core from undergoing an undue amount of oxidation by water and prevents a consequential rise-in reactor pressure.

. Certain circuit modifications have been made, but not yet implemented, to allow fixed alarm settings on moisture monitor minimum allowed sample flow for various reactor power ranges, rather than having reactor operators make manual adjustments. This is a distinct improvement and final implenentation waits on1y Commission approval of the proposed alarm settings for the full power range. Staff review is essentially complete. At this meeting we will seek mutual understanding of some minor changes the staff would like to make before final action is taken. { Instrument Setpoints On January 23, 1978 a minor release of radioactivity occurred owing to the failure of a level controller in a tank in the helium circulator auxiliary system. This allowed the tank to overfill. The back-up level controller also failed. Before the plant went back into operation, PSCo was required to take certain corrective actions to prevent a recurrence. PSCo also comitted to complete and implement surveillance ( procedures for the entire helium auxiliary instrumentation system before the plant is permitted to go above 70% power. Implementation is being reviewed by Region IV inspectors and at this meeting we will seek infonnation on the status of this program from both PSCo and the Region IV inspector.

. Saturday Afternoon, November 4, 1978 (Post Refueling Comitments) Auxiliary Cooling Method (ACM) In late 1975, PSCo made a comitment to install a system which would provide a back-up source of power to assure that power could be delivered to vital components in the event a fire in the electrical distribution system were to interrupt the capabihity for delivering power via off-site or on-site emergency sources. ( An interim version of the Auxiliary Cooling Method has been functional at the plant for about two years. The final version (now under review) is to be functional before reactor start-up after the first refueling outage. The Auxiliary Cooling method installation includes a 2500 KW diesel generator which supplies a power source via cables routed outside of potential areas of electrical fires. Transfer switches located near vital components can be used, as necessary, to disconnect interrupted normal power sources and reconnect to the dedicated ACM source. The basic functions are to (1) assure a continued source of cooling water to the reactor vessel liner cooling system (2) assure that cooling water can be supplied to heat exchangers in the helium purification system during reactor depressurization and (3) to assure that the back-up reactor shutdown system can be actuated. At this meeting there may be a limited amount of discussion pertaining to the overall status of the ACM installation and the procedures employed'in its j operation. Fire Protection At the time, the comitment was made to install the ACM, PSCo also under-l l l

.g. took to upgrade fire protection and prevention measures for the electrical 4 dsitribution system. A number of measures have already been taken, and final improvements are scheduled to be installed during the first refueling outage. Details pertaining to these improvements have been supplied by PSCo. At the meeting we will seek information on the overall status. Booster Pump PSCo plans to put a new pump in the fire water system during the first refueling outage to augment the capability of this system to supply water to the helium ( circulator pelton wheel drives under emergency conditions. This will allow l circulator speed to be increased and reduce fuel temperatures for the fire-water cooldown event discussed earlier in the afternoon. We seek infomation on the status of this improvement and on the capabilities of the new pump.

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• Low Temperature Adsorber Cooled by Liquid Nitrogen Scw eriaric Diasnan 09E H Euon Punis icaTios TgaiN

ENCLOSURE 2 .y LICENSING CRITERIA FOR HTGRs IN THE UNITED STATES Robert A. Clark and Peter M. Williams U.S. Nuclear Regulatory Commission Washington, D.C., USA INTR 0'UCTION D The philosophy under which all high temperature gas cooled reactors (riTGRs) are reviewed for licensing in the United States is that a comparable level of safety must be established for all reactor types, with the full recognition that the great majority of licensing criteria were developed for light water reactors. The implementation of this philosophy in the establishment of HTGR criteria has taken the follcwing three forms with respect to the previously existing criteria: cirect adoption, suitable adaptation, and recognition of the need for and develop-ment of specialized HTGR criteria. Fortunately, direct adoption of the existing criteria is possible in the great majority of instances and provides for the HTGR the best means for assuring a comparable level of safety. Examples of ( direct adoption are numerous and range from the IEEE criteria to most of the NRC Regulatory Guides. A~ list of Regulatory Guides applicable to HTGRs was presented at the 1974 Gatlinburg conference on gas cooled reactors'. While this list was limited of course to the Regulatory Guides available at that time, it has been our experience since then that almost all Regulatory Guides except those which deal with specific aspects of the nuclear staam supply systems or accident analyses apply directly to HTGR licensing. Before describing the other two forms of HTGR licensing criteria we provide a brief description of the HTGR concept and recall that essentially three types of HTGRs have been considered in detail for licensing purposes in the United States. All these designs are those of a single manufacturer, the General Atomic Company of San Diego, California. They are: the 40 MW(e) Peach Bottom I reactor,z which was operated for seven years by the Philadelphia Electric Company (' until 1974; the 300 MW(e) Fort St. Vrain reactor,3 which is currently undergoing power ascension testing by the Public Service Compan{ of Colorado; and the "large" HTGR concept of the 700 to 1000 MW(e) class. Licensing of the large HTGR concept was initiated in 1969 with the "1000 MW(e) Study"5 and developed more fully during the 1973 and 1975 period when the reviews of the Sumit and Fulton applications,7 of Delmarva Light and Power, and Philadelphia Electric, 6 respectively, reached the stage where reports from the Advisory Committee on Reactor Safeguards (ACRS) were issued.8,9 Review continues at a lower level of activity for the GASSAR-6 design,10 a standardized nuclear steam supply system based on Fulton. The GASSAR-6 design was recently discussed at a Sub-committee meeting of the ACRS.ll* Table 1 sumarizes the design parameters for these three HTGR types and Figures 1, 2, and 3 illustrate the design features for each. The changes primarily visible in these figures are in the arrangement of the primary coolant system and the introduction of the prestressed concrete

• Presented at the ANS Topical Meeting on Thermal Reactor Safety, Sun Valley, Idaho August 1.1978

p m l TABLE I IITCR DESICN PARAMETERS Peach Bottom Fort St. Vrain CASSAR-6 Net electr***1 output W 40 330 1159 overall station net efficiency I 34.6 39.2 38.6 Containment type Steel Atmospheric confinement Reinforced Concrete / Steel Number of main / emergency cooling loops 2/2 2/2 6/3-Date criticality Comm. op. Hay 1967 January 1974 NA i Reactor core output HW(t) 115 842 3000 27.7/20.8 l Core dimenulons dia/ht.ft 9.16/7.5 19.6/15.6 llellum coolant inlet pressure psig 305 688 725 Avg. coolant temp, reactor inlet

• F 650 762 606 Avg. coolant temp, reactor outlet
• F 1380 1445 1392 Avg. power density KW(t) liter 8.3 6.3 8.4 Avg. conversion ratio 0.44 0.60 0.65 Fuel material Th/U-235, 951 enriched /U-233, recycle--

Element length / min width in. 144/3.5 31.22/14.7 31.22/14.7 Total quantity of U-235/Th kg. 220/1450 882/19.458 1747/37,487 Average fuel burnup HWJ/ tonne 60,000 100,000 98,000 Reactor Vessel Type Steel pressure vessel Prestressed concrete reactor vessel- ~ Max. external dimensions dia/ht ft 14.5/35.5 49/106 100.5/91.2 i Helium circulator type Centrifugal, electric single-stage axial flow, steam turbine drive drive i Forced recirculation once-through, helical coil with integral rehe.6L Steam generatc, type Reactivity Control Control. rods and emergency shutdown canisters---- Scram method rods hydraulic / electric -gravity Emergency Core Cooling System Pony motors, natural use existing main cires.. 3 indepent cooling loops, 4 convection water turbine electric motor

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reactor ve'ssel. Other principal changes are in the fuel design and in the direction of flow through the cere, with both Fort St. Vrain and the large HTGR having downward flow. Further aspects of the designs will become more apparent later in the context of the discussions of the factors affecting the HTGR licensing criteria. To return to the subject of licensing and licensing actions, Table 2 pro-vides a listing of the principal actions and considerations affecting HTGR licensing where it can F.e seen that the fundamental considerations of the ceramic core design and HTGR hazards for Peach Bottcm I were augmented as the t design progressed toward commercial status. Peach Bottem I continues to con-tribute to HTGR development through the study of fission product transport and plateout being carried out by the Peach Bottom End-of-Life Program.12 It is through a study of these licensing actions that we develop background for L discussion of those HTGR licensing criteria that are not exactly the same as those for light water reactors. TABLE 2 PRINCIPAL ACTIONS AND CONSIDERATIONS AFFECTING l HIGR LICENSING CRITERIA Licensing Actions Licensing Considerations Peach Bottom I 1. Ceramic Core Design 1961 - present 2. Fission Product Transport and Plateout 3. Delineation of HTGR Hazards / Fort St. Vrain 4 Prestressed Concrete Reactor Vessel 1966 - present 5. Retention of Fission Products Within Coated Fuel Particles 6. Detailed Definition of Depressurization i and Core Heat-up Accidents 1000 MWe Study 7. Reactor Cantiinment Requirements 1969 8. Integrated Primary Coolant System Summit and Fulton 9. Containment Backpressure Requirements 1973 - 1975 10. ECCS Perfcrmance Including Air Ingress 11. Testing Requirements of Primary Mechanical Components 12. Steam Generator Design

• 13.

Vendor QA 14. Decay Heat Rate GASSAR 15. Conformance of Application with HTGR ~ Edition of Standard Format 16. Revised Seismic and Structural Analysis 17. Detailed Review of Fission Product 3 Release From Failed Particle Coating i l

For those existing criteria that can not be regarded as unequivocably applicable to HTGRs and suitable adaotation is appropriate, information is j developed to permit the use of tne pnrase, " meets the objectives of" or words to ? this effect. Development of such information is usually a straightforward i process. Our practice is to request the applicant to clarify and justify dis-crepancies from the strict adherence to the criteria cited and then to review the approach to assure that a ccmparable level of safety is met. An example of the adaptive approach is the conformance of HTGR designs to the Ccmmission's i General Design Criteria for nuclear pcwer plants. 1 4 The HTGR reviews since Pgach Bottom I have required substantial conformance j with the General Design Criteria and have established a background for later HTGR designs. The Fort St. Vrain project was the first HTGR reviewed in the j context of the General Design Criteria and also initiated the review of an HTGR configuration substantially different from Peach Bottcm I and more nearly re-lated to tne current configurations of General Atomic's designs. The Fort St. 1 Vrain operating license review found: plant structures, systems and compo-nents important to safety...are in accord with the Commission's General Design ( Criteria...and that any departures from these criteria have been identified and 4 \\ j us ti fied. " In the Sumit and Fulton reviews the staff concluded that the design complied with the objectives of the General Design Criteria subject to 1 the development of "such further technical or design information.as may be i required to complete the safety analysis..." l While in many cases either adoption or adaptation of light water reactor. criteria is suitable and clearly justified, difficulties may be encountered in these approaches, such that it becomes necessary to study the issue to determine if the development of a specific HTGR criterion is warranted. Further it is necessary to recognize the need for specialized criteria that develop from the distinctive features of the HTGR concept and its postulated design basis accidents. This paper deals with specialized HTGR licensing criteria from the standpoint of these considerations. DISTINCTIVE HTGR CHARACTERISTICS k-An organized knowledge of the HTGR's similarities and differences with respect to light water reactors has been one of the basic means for identifying needs for and the requirements of specialized licensing criteria. While it has been our experience that the existence of distinctive characteristics in them-selves may not necessarily result in the need for specialized criteria (e.g., much [ of the nuclear design falls in this category), other distinctive features require a great amount of specialization in criteria (e.g., the. prestressed concrete reactor vessel). The following is a guide to the HTGR's distinguishing l characteristics. l FRINCIPAL DESIGN RELATIONSHIPS l The following equation for thermal output illustrates the interdependence of power density, a measure of economic merit and reactor productivity, and mechanical and thermal design parameters. This relationship, which is generally descriptive of all types of gas cooled reactors, is' nearly the same as given in a paper by M. Troost, P. Fortesque, et all3 where expressions for heat transfer f 1 l ,.1 --p y 7 ..-n. .+ .e,-

r and pressure droo are eliminated by the use of the Reynolds analogy between the friction factor and the heat transfer ccefficient. Development is based on a one dimensional cescripticn of the core cooling loop and other simolifications. Some changes in nemenclature and groupings of terms are given here as an aid to the ensuing discussion. -h ( To-T ) (at) b h b h("[2)3 l G= rn s where G = the average pcwer density of the reactor core R = the universal gas constant M = the molecular weight of the coolant n = the number of degrees of freedcm in the coolant gas molecule. For helium n = 3 T = the ratio of the pumbing pcwer to the gross power output of the core n = the mechanical e'fficiency of the compressor P = average pressure of the coolant L = length of the reactor core ( c = the fractional volume of the core occupied by coolant \\ Ti = the average core inlet temperature in absolute units To = the average core outlet temperature in absolute units T = the average temperature of the coolant in the core in absolute units at = the effective temperature difference between the. coolant channel surface and To l The four bracketed expressions on the right deal with the design variables in the following order: the thermodynamic properties of the coolant; the pumping power fraction with consideration of mechanical efficiency of the ccmpressor; the reactor pressure and the controlling physical dimensions; and the thermal design parameters. The following elementary aspects of gas reactor design are evident. ( l. Helium is a preferred thermodynamic choice of coolant because of its low molecular weight. It is interesting to note that on the basis of ther-modynamics hydrogen would be even better because of the five degrees of freedom in its molecular motion. 2. The pumping power ratio is an important design parameter. In practice, pumping power for HTGRs is about twice that for water cooled reacters. 3. Power density increases linearly with reactor pressure, hence the economic desirability of high reactor pressures. 5 .l I 1 O

4. The power density is a strong functicn of the coolant fraction, c, the variable affecting the heat transfer area alcng with the dimensicns of individual cooling channels. Fe e it can be seen that a reduction in coolant volume to acccmmodate higher thorium loadings, as has been pro-posed, would result in a lcwer pcwer density with other design variables held constant. While not explicit in the equation it also results that coolant channels of minimum r1ze are desirable. This minimum is estab-lished partly by manufacturing c:nsiderations, but princioally by the need to assure that for normal operatien the coolant ficw remains in the turbu-lent regime. 5. Power density is seen to be inversely precortional to core length. This reflects that a short, squ,at ::re is desirable for efficient heat transfer since flow resistance for a given ccolant fraction will increase with core length but not with core diameter. For neutron conservation purposes the core length and diameter are designed to be abcut equal, hcwever. 6. The desirability of high core outlet gas temperatures and low core inlet -s ( and average gas temperatures are indicated by the first expression in the last bracketed term. High outlet temperatures are constrained by both the thermal limits of the core and of essential excore components of the primary coolant system such as the metallic duct liners. Low core inlet temperatures both maximize the numerator and, by lowering the average core temperature, minimize the denominator in the expression. Desirable'lcw values of inlet temperature are limited by economic considerations of heat exchanger design for transfer of reactor heat to steaa. In practice, the gas temperature rise in gas cooled reactors is several times greater than for water cooled reac tors. 7. The power density will rise as the square root of the difference between the coolant channel surface temperature and the average temperature of the helium at the core outlet. Implicit in this relationship is the advantage of the graphite core material which permits Seat transfer to the coolant at I-a temperature substantially above that which can be attained by metal clad fuel. GRAPHITE FUEL Fuel design bases and safety limits are evolving for the HTGR fuel that are not analogcus to those for light water reactors. Analogous quantities are calculated, such as the center line temperature for the fuel rod compact, but these are used only for reference purposes and general design guidance. The proposed bases and limits recognize statistical mechanisms for fission product release from the coated fuel particles and avoid precise thermal thresholds. For normal and upset conditions, circulating activity in the primary coolant would be determined by fission product decay and plateout, and limited to a value that is controlled by the coolant purification system. The circulating activity must be consistent with cperating, maintenance and safety requirements of the primary coolant system. For emergency and inulted conditions, mechanisms t l O I

for release of. fission products frcm fuel failed as a ccnsecuence cf certain postulated accidents will depend in large part on a time-at-temperature rela-tionship and previous ocerating history. Accident models to embody these phenomena can not be directly related to water reactor licensing experience where fixed limits, such as the critical heat flux ratio or maximum cladding temperature, define the potential for fission product release frcm the fuel in the event of postulated accidents. Fuel failure mechanisms anc their relationship to design bases and safety limits were used in the development'of the Fort St. Vrain Technical Specifications and were the subject of a recent NRC study.1+ The application of these studies to HTGR accident analysis is in progress. GRAPHITE STRUCTURES Graphite is a non-metal and brittle material, although not so brittle as typical ceramics. Structural design criteria for graphite have not been as well established as for metals, and may evolve substantially. Graphite in (,; water cooled reactors and is subject to different corrosion considerations.the HTGR is exp The seismic design exoerience and practice for light water reactors is only generally applicable to the HTGR core and core supports. This results from the many differences in the core and core support arrangements, and in the structura' design criteria for graphite. The seismic design experience for Fort St. Vrain does not provide as much guidance for large commercial HTGRs as might be first anticipated as the structural engineering demands for a -large plant are significantly greater due to the effects of scale, the lower plenum arrangement, and a desire to design for higher seismic loads. NUCLEAR DESIGN The Fort St. Vrain startup 'and power ascension program is providing data confirming the nuclear design of the large HTGRs. The nuclear characteristics ( of the graphite moderated core are summarized below: 1. The core power density is about an order of magnitude less than that of light water reactors. In combination with the good thermal coupling between the fuel and moderator, the dynamic effects of pcwer generation transients on fuel perfomance limits are mitigated. It also follows that negative reactivity insertion rates required to terminate limit-ing anticipated operational occurrences and postulated accidents are sufficiently low that gravity insertion is judged adequate for both control rods and the r1tserve shutdown system. 2. The graphite moderator provides a well distributed source of neutrons entering the thermal spectrum. This, in combination with the low power density, provides' comparable reductions in power spiking. The combina-tion of low power density and the more epithermal spectrum in graphite reduces xenon oscillation potentials in comparison to light water reactors. S e

3. Core pcwer distributiens that are acceptable with resoect to the thermal and metallurgical recuirements of the core cenoonents must also be acceptable to the thermal and metallurgical limits of the graphite structural support elements and essential metallic components of the primary coolant system. 4. The principal contribution to the overall negative temperature coef-ficient is the Doppler effect for thorium resonance and epithermal capture. There are essentially no density changes in the core with temperature. The moderator ccefficient is small and may not be always l negative. 5. Special consideratiori is necessary for protactinum-233 which is' an intermediate isotope in the production of uranium-233 by neutron capture in thorium-232. It decays to uranium-233 with a half-life of 27.4 days and it has a sufficient ccmbination of abundance and thermal neutron absorption cross section that, like xenon-135, it must be r (l separately considered in the control system design. THERMAL AND FLUID MECHANICAL DESIGN 3ecause of the high exit temperatures of the helium coolant, it is neces-sary to provide design bases that address the design limits of essential compo-nents throughout the primary coolant system as well as the fuel and core design limits. Hot streaks exiting the core from different fuel regions, either planned as a convenience to load following operations or as a result of certain postulated accidents, must be considered in establishing the design limits of the components affected. Flow transition from turbulent to laminar will occur in low-power operation and in certain postulated accidents. Laminar flows characteristically have high friction factors and reduced heat transfer coefficients. Analyses for' degraded cooling conditions must also consider the potentials for local flow stagnation ( and reversal, and a fully depressurized coolant. PRIMARY COOLANT SYSTEM No portion of the primary coolant system finds direct precedent in light water reactor design. Design practices for the prestressed concrete reactor vessel, heat exchangers, gas circulators, ducts and valves find their earliest precedents in the gas cooled reactors developed abroad. This includes the " pod" type design for enclosure of the primary coolant system within cavities of the vessel. The Peach Bottom I reactor provided the first confirmation in a high temperature helium environment of the fluid mechanical and materials performances of the system components. While most of these ccmponent designs were not pro-totypical of Fort St. Vrain and later HTGR designs, the materials selection generally was. Fort St. Vrain provides the first experience for downward coolant flow in an HTGR but the component design, except for materials selection and equ5ent layout, differs substantially from the larger design.

... =.. j i Development of licensing criteria for many aspects of the primary system and its components is being aided by activities of various standards csmittees identified later in this paper. As an example, the inservice inspection require-ments developed for light water reactors offer a guide to the scope of such requirements for HTGRs, but the extent of directly analogous requirements and methods has not yet been determined. ENGINEERED SAFETY FEATURES The HTGR engineered safety features are both unique and analogous in com-parison to those of light water reactors. The unique aspects derive from the many fundamental distinctions in design, transient response characteristics and postulated accidents. - The analogous features result from conformity of the HTGR design to meet the objectives of tne General Design Criteria and the 10 CFR Part 1C0 guidelines for plant siting. The fundamental design requirements for the reactor containment structure j ( ] have been a point 6f discussion for many years. The postulated fission product ( source term for light water reactors is not consistent with the time and l temperature dependent mechanism for fuel particle coating failure and fission product release characteristics of graphite fuel. These aspects of the fuel were considered in the Fort St. Vrain design where the prestressed concrete reactor vessel is engineered as the primary containment structure while the design for the large HTGRs provides for a conventional containment structure with design requirements similar to those for pressurized water reactors. The characteristics of future containment designs will depend on siting philosophy and the application of 10 CFR 100 principles as well as mechanistic accident analyses. Engineered safety features of analogous functions but of differing designs are the structures precluding control rod ejection, the pressure relief system, and an emergency core cooling system. In the large HTGRs the designs for pres-sure relief and for the prevention of control rod ejection are similar to Fort' i ( St. Vrain while the emergency core cooling system is substantially revised and 2 ] is more consistent with current practices in light water reactors. An engineered safety feature unique to the HTGR concept which has its origins in the Peach Bottom I design is the loop dump system which mitigates the entry of steam and water into the reactor should a tube failure occur in a steam generator. INSTRUMENTATION AND CONTROL SYSTEMS l Conventional nuclear grade instrumentation systems are appropriate to monitor reactor variables with the principal exception of instrumentation to monitor low level moisture levels and rapid moisture ingress. In Fort St. Vrain " dew point" moisture monitors are used to continuously measure very low moisture concentrations in the primary coolant system, and can also detect rapidly any substantial increase in the moisture level. The instruments func-tion by use of liquid nitrogen to chill the surface of a mirror to a pre-determined low temperature. A continuous sample of reactor coolant gas that passes over the mirror will cause it to fog and disturb a light beam if the moisture concentration in the sample increases above a set amount.

The control rod position indication system in Fort St. Vrain and the large HTGR differs frcm light water reactors because the high temoeratures in the icwer core region preclude a conventional means of signaling full insertion. Position indication is determined frcm 1.he red drive system and a slack cable device which indicates if the red motien into the core is not being properly fulfilled. ELECTRICAL, AUXILIARY, AND POWER CONVERSION SYSTEMS There are no unique aspects of the electrical systems and many of the auxiliary systems. The principal auxiliary systems where distinction exists are the cooling water system for the liner of the concrete reactor vessel, the fuel handling and storage systems, the helium purification and gaseous waste system. The basic designs for the heluim purification and gaseous radwaste systems were established with Peach Bottcm I and the remaining auxiliaries with Fort St. Vrain. The power conversicn system differs frcm those of light water reactors in that steam conditions (2513 psig and g55*F) supply a conventional high pressure, high temperature turbine. A second distinction is that steam is extracted follow-(~ ing the high pressure stage and returned to the reactor to drive the helium circulators. The extracted steam is reheated in the reactor and returned to the power conversion system. The overall plant control system is thus more ccmplicated than that for light water reactors but the design has been func-tioning satisfactorily at Fort St. Vrain. ACCIDENT DELINEATION AND ANALYSIS In the analysis of HTGR transients and accidents the distinctive character-istics in comparison with light water reactors become manifest. While the guidance value of light water licensing experience becomes useful in illustrat-ing the nature of review procedures and acceptance criteria that are needed to perform a ccmparable analysis, the delineation of HTGR accidents is not primarily founded on light water reactor analogies, although such analogies are evident. Rather, the delineation of HTGR accidents develops from first principles of ( nuclear reactor safety design that include the characteristic features of gas reactor design and operating experience. The principal focus of accident delineations in HTGR licensing actions and studies is on the development of postulated design basis accidents for the reactor and other systems capable of the release of radioactive materials. As for light water reactors, design basis accidents are postulated and then analyzed to represent the public consequence of a wide spectrum of accidents that are considered. In the selection of reactor design basis accidents, four fundamental types of events are considered: reactivity insertion, steam and water ingress, depressurization of the primary coolant, and loss of forced circulation. From the spectrum of accidents that lead to release of radioactivity outside the containment building, the principal events considered are the leakage of reactor coolant into the power generation system via a failed re-heater tube, a failure in the helium purification and gaseous radioactive waste systems, and fuel handling accidents. Table 3 sumarizes this classification for large HTGRs and o e ee - 4 .as en e,

= i IABLE 3 CLAS$1flCAil0N 0/ ACCIDtNIS 60R HICRs Release of Radioactivity Reactivity Steam and toss of forced Outside 5 Classification Insertion Water Ingress wp g urtration Circulation Crm ta inment 9 Accidents of lesser loss of flued injectfun from helium Slow depressurization. Temporary loss of Reheater tube consequences than burnable poison. circulator water Rapid depressurliation main cooling system.

failure, design basis Loss of fission bearing seal.

less than design bests los) of feedwater. Fallure of primary accidents. These product poison. failure in reactor depressurtration coolant instru-are representative Lore component vessel liner cooling accident. and not inclusive. rearrangement, tube. ment piping. Caseous rod waste Holsture ingress. Steam generator leaks. Decrease in systew failure. reactor tempera-fuel handling and ture. storage accidents Control rod motlop, Release of radio-active liquid. Design basis accidents. Spurious rod with-Steam generator tube Rapid depressurlsation Sustained failure Consequences buusul-drawal terminated or header failure, rate determined by of nnraal core ed by other t>y protective 100 square inch area. cooling. design basis action. core accidents. Unreg Engineered safety Reactor protection Steam generator dump Reactor vessel closure Core ausillary cooling features lletting system. and Isolation system. design. system. consequences of Contaffment vessel. Containment vessel. design tasis accidents. Accidents precluded Control rod ejection. Large moisture ingress Depressurtration area Umutricted (ure heatup I,y design pro-Core drop. combined with reactor greater than !OO sq.in, in combination with

visions, depressurization or Depressurtration combined tontainment failure.

core heatup. with con'tairmient failure.

• unrestricted core heatup is the design basis for the maximum hypothetical fission product release used for reactor siting puepuses.

e =~ ' - ~ ~ ' ~

also identifies the engineered safety features provided to cope with the design basis accidents and those accider. s that are precluded by design. As will be described, the large HTGR has design features that often ccpe differently with the postulated design basis accicents than either the Peach Bottcm or Fort St. Vrain designs, although the same fundamental type of hazards detennine the design basis of the safety related systems. REACTIVITY INSERTION ACCIDENTS Studies originating with the Peach Bottem I design considered a broad spectrum of reactivity insertion events including loss of burnable and fission product poisoning, core ccmponent rearrangement, moisture ingress, sudden decrease in reactor temperature and spuricus withdrawal of a centrol rod. The control rod withdrawal event has been consistently found as the bounding reactivity insertion mechanism. For Fort St. Vrain and the large HTGRs it is analyzed on the basis of the withdrawal of the maximum worth rod pair occurring at plant i conditions ranging frem source level to full power, with the withdrawal motion - terminated by reactor trip on signals frem the plant protection system as a last ( ' resort. This trip would follow failure of both operator action and automatic action of the operational protection system. The event results in some fuel particle damage with a potential reactivity release to the coolant which is a small fraction of the design-level circulating activity. The assumptions, analysis and the consequence for this reactivity insertion accident have been consistently found acceptable in all the HTGR licensing actions. It is of interest to note that essentially all of the reactivity insertion mechanisms postulated for the large HTr?. were originally analyzed in the design of the Peach Bottom I reactor. For this reactor, however, spurious red with-drawals were analyzed for both with and without reactor trip. In the case without reactor trip, stabilizaticn of the transient occurred before reaching excessively high core temperatures based on the negative reactivity coefficient. This calculation was indirectly confirmed by an experiment with the AGR, a German reactor of similar core physics parameters. In this experiment the reactor was ( allowed to operate by natural convection cooling without reactor trip or set back. The reactor stabilized at a higher temoerature but at a lower power level on the basis of its negative reactivity feedbacks. STEAM AND WATER INGRESS Ingress of moisture into the reactor is of concern for both chemical and physical reasons. From a chemical standpoint the reaction rate between graphite and water vapor becomes significant at temperatures greater than about 1300"F. Products from this reaction are largely " water-gas", a mixture of hydrogen and carbon monoxide. For slow rates 'of moisture ingress the major concern is with the long term corrosion of the graphite structure and fuel and not with the reaction products which are continuously removed by the helium purification system. For high rates of ingress ccncerns derive frcm the physical increase in reactor pressure and from the rapid generation of the reaction products which i are potentially combustible. The bounding case postulated for the design basis l l l

accident is a steam generator failure of sufficient magnitude such that the rate of moisture ingress would recuire automatic action tn mitigate the course The same essentials of protection are provided for all three of the accident. HTGR designs. When excessive moisture levels are detected by the specially designed moisture monitor instrumentation, the reactor is scrammed in coincidence with a signal to isolate the leaking steam generator frcm its feedwater supply and its out-The content of the isolated st?am generator is then dumped let s team path. outside the reactor vessel to prevent further ingress of moisture into the In the event that the dump system fails to perform its function, reactor. If reactor pressure will rise with the continued introduction of moisture. the rise in pressure is substantial the oressure relief system will function and the primary coolant will be periodically vented by opening and closing of relief valves to prevent excessive pressure loading of the reactor vessel, but the reactor will not be depressurized. The steam generator isolation and dump system dces not have an analogy The key to its performance is rapid detection of in light water reactors. (- excessive moisture with correct identification of the failed steam generator Our evaluation of the steam or water ingress hazard includes consideration of We have reviewed the failure of the isolation and dump system to function. and accepted General Atomic's analysis showing that timely tripping of the reactor with continued core cooling, and functioning of the pressure relief At these system will result in graphite temperatures not greater tha'n 1300*F. temperatures the endothermic water-gas reaction would be minimal. DEPRESSURIZATION OF THE PRDtARY COOLANT The rapid depressurization accident is selected as one of the design basis accidents for the HTGR concept and is effectively analogous to the loss-of-Engineered safety features coolant accident postulated for light water reactors. are required for its mitigation, with the postulated depressurization accident affecting the design basis for the containment system, the emergency core cooling Further, the postulated depressuriza-system, and the primary coolant boundary. tion accident detemines the lim I, the essential reactor internals must be designed. Required functions of the plant protection system associated with rapid depressurization are containment isolation, reactor trip on low pressure, reactor trip on high containment pres-i sure and initiation of emergency core cooling on signals of low primary coolant flow or icw feedwater flow. A conservative value for the flow area through which the reactor is depres-The flow area then surized detennines the maximum depressurization flow rate. t is the governing parameter in developing those design bases determined by the A value of 100 square inches was postulated for the depressurization accident. design basis depressurization accident for Fort St. Vrain as a value which would recognize both mechanistic and non-mechanistic causes for the initiation of the This postulated arer was maintained for the Surmit and Fulton accident. designs.

The Surmit and Fulton reviews illustrated that the use of a ccnventional contair. Tent system together with the core auxiliary cooling system reduces public exposure to an acceptable value in the event of a design basis depresuri-

ati;n accident.

LOSS OF FORCED CIRCULATION The requirements for the reliability of forced circulation in the large HTGR design differ markedly frca light water reactors and from both the Peach Bottom I and Fort St. Vrain designs. For water cooled reactors natural con-vection cooling is capable of removing reactor heat from low pcwer and shutdown conditions. Natural convection cooling was also possible for Peach Bottom I for shutcown cooling'under accident ccnditicns. In Fort St. Vrain, in the event of a sustained loss of forced circulation, decay heat would be transferred to the

nermal carrier by radiation and removed by the reactor vessel liner cooling water system.

This cooling mcde requires controlled depressurization of the reactor vessel and could result in substantial core damage. In the large HTGRs, sustained loss of forced circulation is not acceptable, and the redundancy k_,anddiversityofthecoreauxiliarycoolingsystemwhichprovidesforcedcircula-tion independent of the main cooling system are required. Consequences due to the failure of forced circulation will vary in proportion to the time during which forced circulation is interrupted. Because of the low core power density and the relatively close thermal coupling between the fuel and the graphite moderator, the thermal response of the reactor is ccmparatively slow. However, protection systems provide for immediate reactor trip or setback on first indica-tions of total or partial loss of coolant circulation. Both the large HTGR design and Fort St. Vrain have cooling systems designed to restore circulation within time frames well below those needed to cause damage to the core, its support structure, or essential ccmponents of the primary coolant system. MAXIMUM HYPOTHETICAL FISSION PRODUCT RELEASE For the purpose of reactor site evaluation 10 CFR Part 100 requires the determination of radiological consequences of a postulated fission predact ( release accident that would result in potential hazards not exceeded by those from any accident considered credible. Models under development by both the NRC and General Atomic postulate a maximum hypothetical fission product release based on unrestricted heatup of the core. This hypothetical event would result in the failure of fuel particle coatings and transport of fission products from the fuel to the containment atmoschere withcut restraint by the primary coolant system boundary. Postulation of unrestricted Gore heat-up for siting purposes originated with the Sucmit and Fulton applications, and its development continues in the GASSAR-6 review. In Fort St. Vrain the postulated hypothetical release was included in the design basis depressurization accident while for Peach Bottem I core heat-up mitigated by natural convection cooling of the core following a primary system rupture was used as the siting basis accident. l i e

I In the Summit and Fulten reviews a staff model for the t.2 pethetical fission product release was develcped that the staff censidered "...very conservative to allow for the lack of past enerience with HTGRs of this pcwer level." It was concluded that the Sumit and Fult:n plants could be constructed and operated at the proposed locations withcu undue risk to the health and safety of the Recently, the. RC issued a document entitled " Evaluation of High 1 public. Temperature Gas Cooled Reactor Fuel Particle Coating Failure Models and Data".h Information in this document, coupled with additional experimental data, is expected to lead to a recucticn in the conservatism in the model developed for Sumit and fulton. ACCIDENTS PRECLUDED BY DESIGN In Table 3 accident cases' are identified as precluded by design provisions. These accidents potentially ccul: result in greater ccnsequences to the public than calculated for the design basis accidents. As multiple failures in struc-tures and engineered safety features wculd be required for those accidents to occur, we have judged that these accidents are not credible, and that they need not be considered as bases for the reactor safety design. Nevertheless, we (- believe that the licensing of a large HTGR for comercial purposes will require additional confirmation of the icw risk associated with these events. We believe this risk can be shcwn to be extremely low using the methodology of the quantitative risk assessments being developed for reactor safety studies. STANDARDS DEVELOPMENT It is customary for the NRC to adopt, either in full or in part, industry i standards pertinent to nuclear safety whenever justified. In addition, the NRC develops standards as a means of codifying its licensing criteria. However, our activities to develop a standard review plan for HTGRs comparable to that for light water reactors were terminated with the curtailment of licensing efforts for the large HTGR in 1975. We describe in Table 4 our understanding of various standards activities pertinent to the special needs of HTGRs. In cases where forthcoming publications are cited we have not provided an estimate of the i issuance date, although many of these publications are expected to be issued shortly. While most of the standards efforts listed are currently active, the continued level of activity in most cases will depend upon comercialization activities for the large HTGR. It may be.10ted that many of these standards deal with topics where suitable licensing criteria are not yet available (e.g., [ inservice inspection of primary system components and graphite structures). CONCLUSIONS Licensing criteria for HTGRs in the United States has been under evolutionary development since initiation of the Peach Bottem I review in 1961. The construc-tion and operation of both the Peach Bottom I and Fort St. Vrain reactors have provided means for establishing licensing criteria that is specific and generally comparable to those for light water reactors, although some deficiencies remain. During the Summit and Fulton reviews standards development efforts were intensified

r r i TABLE 4 STATUS OF PRINCIPAL HTGR STANDARDS O organtration Standard Title Status herican Nuclear Society, Subconuntttee Nuclear Safety Criteria for the Design of Draf ts issued for consnent starting ANS-53.1 and herican National Stanfards Stationary Gas Cooled Reactor Plants in 1974. Subconmittee now inactive. Institute N18 Couenittee American Society for Testing Materials ASlH Standard C-701 Recommended Practice Standard under developnent. Consnittee C5 - Industrial Graphite for lesting Graphite and Doronated Graphite Subconnittee C505 - Nuclear Appilcations American Society of Mechanical Engineers. Code Cases 1592, 1593, 1594, 1595, and 1596 lhese code cases deal with auterials of interest to HIGAs and are updated Boller and Pressure Vessel Code, Section 111. quarterly. Subgroups on Elevated Temperature Design and Elevated Temperature Construction Anerican Society of Mechanical Engineers, Rules for Inspection and Testing of Components To be issued for trial use and consucnt, llotter and Pressure Vessel Code. Section XI. of Gas Cooled Nuclear Power Plants Ulvision 2, Subgroup on Gas Cooled Systems, Subconmittee on Nuclear Inservice Ir5nection Energy Research and Development Admini:tration ERDA Standard: RDT EG-1. Near Isotropic lhls standard is under develogunent fcr Petroleum Coke Based Graphites for liigh trtal use in ERDA programs. Temperature Gas Cooled Core Components Joint Connittee of the hnerican Concrete ASMC Botier and Pressure Vessel Code. Adopted and published December 1975. fastitute and the American Society of Section Ill Division 2. ACI Standard 359-14 Certificates can be issued after Mechanical Engineers on Concrete Pressure Code for Concrete Reactqr Vessels and January 1976. Couponents for Nuclear Service (Section Ill) - Containment Working Groups on Concrete Reactor Vessels and Concrete Containment Working Group on Core Support Structures Design Requirenents for Graphite Core To be pubitsbed as Subsection CE in Structures above standard. U.S. Nuclear Regulatory Copenission llTGR Edition of the Standard Fonnat and content Issued in July 1973 for trial use and of Safety Analysts Reports for Puclear Power coument. No present NRC/il1GR standards Plants activities.

and many remain active. Within the NRC, development of licensing criteria was curtailed with the termination of the Summit and Fulten reviews, but confirmatory research programs, sponsored by the Office of Nuclear P.egulatory Research, continue to provide information in significant areas. In summary, a substantial amount of experience is available for the licens-ing of commercial HTGRs. It has been the purpose of this paper to report on the development of HTGR licensing criteria and to provide a brief record of HTGR licensing history. REFERENCES 1. R. A. CLARK and G. L. WESSF%N, " Status of Regulatory Guide and Criteria Develecment for Gas-Cooled Reactors", Conference on Gas-Ccoled Reactors - HTGR and GCFBR, Gatlinburg, Tennessee, May 7,1974. 2. PHILADELPHIA ELECTRIC COMPANY, " Final Hazards Summary Report, Peach Bottom Atomic Power Station", February,1964. ( - 3. PUBLIC SERVICE COMPANY OF COLORADO, " Final Safety Analysis Report, Fort St. Vrain Nuclear Generation Station", November,1969. 4. GENERAL ATOMIC COMPANY, " Standard Safety Analysis Report (GASSAR-6)", August, 1974 l 5. JOSEPH M. HENDRIE, Acting Chairman, Advisory Comnittee on Reactor Safeguards, letter to Glenn T. Seaborg, Chairman, U. S. Atomic Energy Commission, " Conceptual Design for Large High Temperature Gas-Cooled Reactor (HTGR)," November 12, 1969. 6. NRC STAFF, " Safety Evaluation of the Summit Power Station, Units 1 and 2", NUREG-75/004, U.S. Nuclear Regulatory Commission, January,1975. 7. NRC STAFF, " Safety Evaluation of the Fulton Generating Station, Units 1 (, and 2", NUREG-75/015, U.S. Nuclear Regulatory Commission March, 1975. 8. WILLIAM KERR, Chairman, Advisory Committee on Reactor Safeguards, letter to William A. Anders, Chairman, U.S. Nuclear Regulatory Comnission, " Report on Summit Power Station, Units 1 and 2", March 12,1975. 9. WILLIAM KERR, Chairman, Advisory Comnittee on Reactor Safeguards, letter to William A. Anders, Chainnan, U.S. Nuclear Regulatory Commission, " Report on Fulton Generating Station, Units 1 and 2", April 8,1975. 10. NRC STAFF, " Interim Safety Evaluation Report Related to the Review of the GASSAR-6 Nuclear Steam Supply System", in preparation by the U.S. Nuclear Regulatory Comnission. ll h

enc.L@SogL3 Presented at the U.S. - Japan Seminar on HTGR Safety November 24, 1978 Fuji, Japan HTGR POSTULATED ACCIDENTS AND SAFETY RESEARCH NEEDS' IN THE UNITED STATES by R. D. Schamberger and P. M. Williams U.S. Nuclear Regulatory Commission Washington, D.C., USA INTRODUCTION In the United States the primary guidelines for testing the reactor safety design in conjunction with a selected site against potential radiation hazards to the health and safety of the public are provided in the Code of Federal Regulations, Title 10 (Energy), Part 100. Although Part 100 was developed from safety analyses of light water reactors it is applicable to all types of commercial nuclear power plants. The delineation process for High Temperature Gas Cooled Reactor (HTGR) accidents must, therefore, take into account the requirements of Part 100, as well as the characteristic features of gas reactor design and experience. In this regard the distinctive charac-teristics of the HTGR in comparison with light water reactors become manifest. While the guidance value of light water licensing experience is basic in determining the review procedures and acceptance criteria that are needed to perform an HTGR safety analysis, the delineation of HTGR accidents includes the characteristic features of gas cooled reactor (GCR) design and pertinent ( operating reactor experience of GCRs. This paper will discuss postulated HTGR accidents first from the perspec-tive of the design basis accidents established in past licensing actions for steam cycle HTGRs in the United States and then from recent events which include NRC licensing and research activities on low probability accidents, and the Fort St. Vrain startup and operating experience. The licensing actions to be cited are those pertaining to the Peach Bottom HTGR, Fort St. Vrain, and the large steam cycle HTGR designs of the General Atomic Company. The research activities will include principally NRC sponsored work. This paper will deal only with reactor accidents as there are presently no pending HTGR licensing actions where our focus would include other potentials for the release of radioactivity such as the radioactive waste system or the spent fuel storage facility. [

DESIGN BASIS ACCIDENTS From the spectrum of iccidents and transients postulated and analyzed " design basis accidents" have been selected that provide bases for the plant i safety design including the engineered safety features. As in the safety I analyses for light water reactors, design basis accidents, for HTGRs are selected to illustrate the consequences of accident events that are considered credible but far from probable. Hypothetical failures in the functional performance of certain engineered safety features are also assumed and a limiting radiation exposure to the public is conservatively estimated and compared with the Part 100 guidelines. Other accidents, customarily identified as " Class 9" accidents, can be postulated that could have consequences substan-tially greater than the design basis accidents but by analysis and judgement are determined to have a substantially lower potential for occurrence than even the design basis accidents and hypothetical failures. These accidents need not be explicitly considered in the safety design of the reactor and power plant. l Deterministic methodology has been and continues to be the principal k basis for selection of the design basis accidents. The selected HTGR design basis accidents are representative of four fundamental reactor hazards; reactivity insertion, steam or water ingress, depressurization of the primary coolant, and loss of forced circulation. Table 1, which was developed from our reviews of the Summit and Fulton projects of Delmarva Light and Power and Philadelphia Electric, respectively, summarizes this classification and also identifies the engineered safety features provided to cope with the design basis accidents and those accidents that w,;recluded by design provisions.1 As described below, the large, steam cyc h 4T21 would often cope differently with postulated accidents than either the Peach Bottom HTGR or Fort St. Vrain, although the same fundamental types of hazards determine the design bases of the safety-related systems. REACTIVITY INSERTION / The safety analysis reports for Peach Bottom, Fort St, Vrain, and the \\ large, steam cycle HTGRs described a broad spectrum of reactivity insertion events including loss of fixed burnable and fission product poisoning, core component rearrangement, moisture ingret.s, sudden decrease in reactor temper-ature and spurious withdrawal of a single control rod pair. The control rod withdrawal event was proposed by General Atomic and accepted by the NRC as,the enveloping reactivity insertion mechanism. It is analyzed on the basis of the withdrawal of the maximum worth rod pair at plant conditions ranging from source level to full power, with the withdrawal motion terminated by reactor trip on signal fram the plant protection system. This trip would follow failure of both ope:ator action and automatic action of the operational pro-tection system. The consequences as estimated by General Atomic for the large HTGR under assumptions stated to be pessimistic, would be gross fuel particle damage of less than 0.5 percent with a potential activity release to the coolant which would be a small fraction of the design-level circulating activity. l Both the postulated event and its consequences are similar to those analyzed for Fort St. Vrain.

l l Reactivity insertion mechanisms attributable to control rod ejection or the dropping away of the core from the control rods are not considered credible on the basis that their occurrence is to be precluded by design. Control rod ejection in the large HTGR is precluded in a manner similar to Fort St. Vrain in that the anchorages for the control rod drive mechanisms in the reactor vessel penetrations are backed up by a cover plate arrangement in the head of the PCRV. Core drop is precluded by the integrity of the core support system. t For licensing of a large HTGR, additional information would be required to confirm that these accidents are precluded by a satisfactory design. For rod ejection, further knowledge and analysis of the penetration anchorage and the holddown mechanism would,be needed. For the core drop accident, the applicant would be expected to discuss potential mechanisms for failure of the core support system and demonstrate that any failures of this type are of very low probability, yet should any credible failures occur the consequences would be bounded by other design basis accidents. STEAM OR WATER INGRESS High rates of steam or water ingress result in the increase in reactor pressure, rapid graphite oxidation and rapid generation of the reaction products which are potentially combustible. For the steam cycle HTGR, three different types of sources in the reactor coolant pressure boundary provide locations for potential water or steam ingress; the steam generator heat transfer surfaces, the circulator bearing water interfaces with helium, and the cooling tubes of the concrete vessel liner. Failure of the liner cooling tubes would introduce water into the reactor only when it is depressurized. Circulator bearing malfunctions and most steam generator leaks would introduce water at rates where detection, followed by corrective action, would prevent development of a significant hazard. The bounding events are steam generator failures which offer breeches of sufficient size that the rates of water or steam are greater than can be accommodated by operation of the helium coolant purification system or normal shutdown of the reactor. For these events engineered safety features are required to mitigate the course of the event. The engineered safety features to cope with the design basis steam or water ingress event are similar in function for all of the HTGR developed in the United States. Briefly, when excessive moisture levels are detected by the safety grade moisture monitor instrumentation, the reactor is scrammed in coincidence with a signal to isolate the leaking steam generator from its feedwater supply and its outlet steam path. The content of the isolated steam generator is then dumped outside the reactor vessel to prevent further ingress of moisture into the reactor. In the event that the moisture monitors and dump systems fail to perform their functions, reactor pressure will rise with the continued introduction of moisture. If the rise in pressure is substantial the reactor will trip on a high pressure signal if the pressure continues to rise the pressure relief system will function. The primary coolant will be periodically vented through opening and closing of relief valves to the primary containment building to prevent excessive pressure loading of the reactor i

i vessel, but the reactor will not be depressurized.

The capacity of each of the two redundant trains of the pressure relief system in the design is based on the failure of the dump system in conjunction with a postulated steam generator failure, with a reactor scram on high moisture. The steam generator isolation and dump system does not have an analogy in light water reactors. The key to its performance is rapid detection of exces-sive moisture with correct identification of the failed steam generator. We are currently evaluating the performance of the moisture monitor and the dump system during the power ascension testing of Fort St. Vrain. Because of the limited experience with the steam generator isolation and dump system, the design basis moisture ingress accident is postulated on the basis of failure of moisture monitor detection to dump the proper steam gener-ator. The safety analysis reports for the large HTGRs analyzed the case where steam leaks into the reactor at a rate of 90 pounds per second, representative of a steam generator tube sheet failure, which is greater than the failure of several tubes. In our construction permit review to the Summit and Fulton applications a finding was made that "H O inleakage accidents will not endanger i 2 the public and that the design features provided to prevent or to mitigate the consequences of these accidents are acceptable." This finding wes made con-tingent on a subsequent review of General Atomic's methodology for calculating t the amount of the graphite oxidized during the postulated event. These findings in the Summit and Fulton reviews were based on an event sequence judged as a reasonable and conservative representation of the potential for a large failure in the steam generator followed by a failure of one of the engineered safety features. Different event sequences were considered by Oak Ridge National Laboratory in a report which considered consequences to the core support posts.2 Several cases of event sequences of low probability for design are postulated including a tube burst coincident with the design bas.is depressurization accident. Other postulated low probability cases could derive from added and combined failures involving failure of the relief valves to close, failure of the moisture monitors to provide a reactor trip, and degradations in the forced core cooling capability. These low probability cases might also produce a combustion hazard within the containment building. ~ We are aware that some countries consider rapid injection of steam or water coupled with reactor depressurization a credible event sequence and consider it in the safety design. We would plan in any future licensing study ~ of an HTGR to re evaluate our Summit and Fulton positions in this regard by l taking into account developing information in steam generator oesign and experience, inservice inspection provisions and techniques, and the potentials l the proposed reactor design would offer for the occurrence of accidental de.ressurization under any circumstances. As noted in a later section, we are s currently supporting research dealing with the potentials for flammability and detonation hazards if combustible gases are introduced into the atmosphere of the containment building. 1

. DEPRESSURIZATION OF THE PRIMARY COOLANT In part licensing reviews we have studied a spectrum of primary system depressurization events ranging from the off-set rupture of a small instrument i line to a postulated leak area as large as 100 square inches. Both sloa and rapid depressurizations of the reactor result in releases into the conta!nment building of the inventory of radioactivity circulating with the helium coolant. With rapid depressurization, the release could also include significant quanti-ties of adsorbed and plated-out fission products. Furthermore, with rapid + depressurization the structural integrity of vital primary system components f subjected to the transient differential pressure, must also be reviewed. Both l rapid and slow depressurization rates must also be considered in the plant safety design with respect'to habitability and operational aspects that would affect safe shutdown of the plant. The rapid depressurization accident bounds other depressurization events and is selected as one of the design basis accidents for the HTGR concept. In many respects it is analogous to the loss-of-coolant accident postulated for light water reactors. Engineered safety features are required for its mitiga-tion, with the postulated depressurization accident affecting the design bases for the containment system, the emergency core cooling system, and the pene-tration closure system. Required functions of the plant protection system associated with rapic deprescurization are containment isolation, reactor trip on low primary pressure, reactor trip on high containment pressure and initiation of emergency core cooling on signals of low primary coolant flow or low feedwater flow. [ A conservative upper design value for the flow area through which the i i reactor is depressurized determines the maximum depressurization flow rate. The flow area then is the governing quantity in developing those design bases determined by the depressurization accident. A value of 100 square inches was postulated for tLe design basis depressurization accident for Fort St. Vrain as an upper bound which would recognize both mechanistic ahd non-mechanistic causes for the initiation of the accident. This' postulated area size was maintained for the large HTGR designs. It shouid be noted that the depres-surization rates for the large HTGRs would be lower in correspondence with their larger volumes of helium coolant. i A consideration in the Summit and Fulton analyses that differed from Fort St. Vrain was the postulation of the depressurization area in a configuration that would be conducive to the ingress of air during the latter course of the accident. The postulated annulus is mechanistically described by the failure of a steam generator closure with the flow restrictor limiting the area of the annulus to 100 square inches. Our consultants at Oak Ridge developed a con-vective model using the eight-foot depth of the penetration cavity for the driving head for the admission of air, with an arbitrarily equal division of l the annulus for inflow and outflow. In the Summit and Fulton reviews this model was considered together with General Atomic's estimates for air ingress, and the effect of air ingress both on the performance of the auxiliary i

__- circulators and on the oxidation of graphite was analyzed. It was concluded far Summit and Fulton that air ingress would occur following the blowdown stage of the design basis depressurization accident but that it would not have a significant effect on the offsite doses. The Summit and Fulton reviews illustrated that the use of a conventional containment system together with the safety grade auxiliary cooling system reduces public exposure to a negligible value in the event of a design basis depressurization accident. For vented type containments knowledge of the amounts and locations of the plated-out fission products and the lift-off fraction anticipated during a depressurization event is an important parameter in developing the containment design bases. Current research activities in this topic will be described later. LOSS OF FORCED CIRCULATION The requirements for the reliability of forced circulation in the large HTGR differ markedly from both light water reactors and from the Fort St. Vrain and the Peach Bottom designs. For water cooled reactors, natural con-vection cooling is capable of removing reactor heat from low power and shutdown conditions, and from Peach Bottom for shutdown cooling under accident conditions. In Fort St. Vrain decay heat can be removed by radiation to the reactor vessel liner if all forced circulation fails. It was determined for large HTCRs, however, that decay heat removal by the liner cooling systen was not practical and that a level of emergency core cooling reliability that would preclude the sustained loss of forced circulation was required. As the large HTGRs, and Fort St. Vrain, are designed for downward core flow natural convection means for emergency core cooling are not provided. Rather, highly reliable, though potentially interruptible, forced circulation is provided for all plant condi-tions including the shutdown state. Consequences due to the failure of forced circulation will vary in proportion to the time during which forced circulation is interr'upted. Because of the low core power density and the relatively close thermal coupling between the fuel and the graphite moderator which provides a large heat sink, the thermal response of the reactor is comparatively / slow. However, protection systems provide for immediate reactor trip or setback on first indications of total or partial loss of coolant circulation. Both the large HTGR design and Fort St. Vrain have emergency core cooling systems designed to restore circulation within time frames well below those needed to produce damage to the core, its support structure or essential components of the primary coolant system. Our review process for a large commercial HTGR application would quantitatively review proposed coolant flow restoration times in terms of various postulated causes for loss of forced circulation. We are currently reviewing this area in connection with the present restriction of 70 percent power on the Fort St. Vrain reactor. A complete description of the response of core and essential primary system components following the various cases for loss and recovery of forced circulation would be developed in a review for a large HTGR application. Knowledge of the core and primary system response for both pressurized and 1 depressurized conditions is needed together with a spectrum of degraded con-ditions for both main and auxiliary circulators and the loop isolation valves. For accident analysis and scoping purposes limiting and enveloping cases would be selected from this spectrum. HYPOTHETICAL EVENTS In order to test the containment design and evaluate the selection of the plant site with respect to the health and safety of the public a hypothetical event involving substantial fission product release from the fuel is postulated, even though such an event is inconsistent with the established performance of the engineered safety features. For the large HTGR this event has been taken as an unrestricted, adiabat'ic heat up of the core by decay heat. Recent work in this area is described in a following section. LOW PROBABILITY ACCIDENTS The term " low probability accidents" is used here in the context that such accidents have a probability of occurrence which is so low that they do not need to be cons.idered in the basis for the reactor design. For purposes of discussion the low probability accidents are categorized by four types of origins. A. Very Rare Natural Events: Accidents in this category would be caused by natural phenomina having an occurrence potential so low that they are not considered credible in the reactor siting analysis. This would include, for example, seismic events of gredter magnitude than the safe shutdown earthquake. B. Non-Controlled Events: This would include human actions beyond the control of any reactor cperator, licensing body or local governments. Examples are war, large terrorist actions and aircraft collisions at the sites remote from air traffic. [ C. Improbable Structural Failures: Reactor structures are designed with \\ well established techniques, large safety factors, and extensive provisions for inservice inspection such that their failure is not considered credible. Gross failure of the concrete reactor vessel would fall into this category. D. Multiple Failures of Engineered Safety Features: Multiple failures of engineered safety features are deemed remote because of functional design provisions for diversity, redundance and independence coupled with safety grade criteria for their design, construction, inspection and operation. The provisions to preclude control rod ejection from an HTGR is an example of how accidents which could be caused by multiple failures are removed to a low probability category. The treatment of the low probability accidents in categories A and B is the same for HTGRs and light water reactors and will not be discussed here.

i i For hypothetical events and the remaining two categories, activities have been underway improving our knowledge of these accidents by licensing activities in the Office of Nuclear Reactor Regulation, research programs sponsored by the Office of Nuclear Regulatory Research, and by safety related research sponsored by the Department of Energy at the Oak Ridge National Laboratory and at the General Atomic Company. There has been substantial communication between these groups and the efforts and activities of each group reflects awareness of the goals and objectives of the other groups. LICENSING ACTIVITIES Licensing activities for HTGRs in the Office of Nuc!r ir Reactor Regulation consist of an on going review of the operation of the Fort St. Vrain reactor and a low-level effort dealing with licensing issues facing commercialization of the large HTGR including advanced applications. The Fort St. Vrain activity is carried out in cooperation with the Office of Inspection and Enforcement and aided by the Office of Nuclear Regulatory Research. The Research Office also aids the licensing studies for the large HTGR. ,r t c t The development of postulated accidents for Fort St. Vrain used the Peach Bottom HTGR experience and occurred during the construction permit review, extending from October 1966 to September 1968. The plant has been upgraded to meet our modern requirements for fire protection and to meet deficiencies found in the routing and separation of electrical and instrumentation cables. Additional equipment has been or is being added to improve helium circulator reliability, the capacity of the emergency core cooling system, and the moisture detection monitors. Since issuance of the operating licensing in December 1973 studies and evaluations have been made or are being made of the seismic and environmental qualification program, core outlet temperature fluctuations and the design basis accidents for rapid depressurization, reactor cool down using seismically qualified systems and sustained loss of forced circulation. i Also being evaluated are the effects of flow reversal in the core following a temporary loss of forced circulation, the potential' strength loss of structural graphite due to oxidation by coolant impurities, the use of isotropic graphite for the fuel blocks, and the inservice inspection and testing program. ( The reactor is currently authorized to operate up to 70 percent of rated thermal power. We anticipate that this restriction will be lifted in the near future with the culmination of certain of the evaluations still in progress. Shutdown for the first refueling operation is scheduled for the January to March period in 1979. Our studies thus far have sustained the selection of design basis and other postulated accidents established prior to operation although certain operating modifications have been incorporated into the Technical Specifica-tions and operating procedures and, as mentioned, additional safety related equipment has been added. The licensing reviews for the Summit and Fulton projects were terminated in late 1975 with the announced withdrawal of General Atomic from the commercial marP*. The staff safety evaluation reports and letter reports from the Advisory Committee on Reactor Safeguards were favorable to construction of

~ ~ I these reactors but identified several safety issues that should be further l l investigated as an adjunat to the construction of these reactors. These i 4 generic issues, together with other issues that later became apparent are l summarized under the following headings: Graphites as Structural Materials, t Core Seismic Response, Fuel Transient Response, Inservice Inspection, Low i Probability Accidents, Containment Requirements, Primary System Integrity and Emergency Core Cooling Provisions. While the topic of low probability accidents is identified here as a separate topic, it contains elements of the licensing and research studies for the seven other topics. l l Following the termination of the Summit and Fulton reviews the generic issue reviews were continued either under the topical report program or the review of the General Atomii: Standard Safety Analysis Report (GASSAR). Although GASSAR is no longer an active review project, our plan is to continue selected generic issue reviews under the NRC's topical report program as these may support licensing needs for advanced HTGR applications. i One of the topics addressed during the GASSAR review was General Atomic's proposed model for fuel failure and fission product release as a consequence of.a hypothetical core heatup caused by sustained loss of forced circulation. An NRC fission product release model for both TRISO and BISO fuel was published l-that was more restrictive than General Atomic's proposed model but nevertheless considered available data that supports a statistical and thermal relationship between fuel particle coating failures thus recognizing a delay mechanism for fission product release.8. Table 2 is reproduced from a publication which shows that the two hour dose at the exclusion distance would be about an order .of magnitude less for an HTGR in comparison with a typical PWR although the 30 day doses in the low population zone are roughly comparable.4 The NRC model j was published with the restriction that it was available for scoping purpose but additional fuel failure data on reference fuel would be necessary before it could be considered a licensing guidelin.e. It is conservative with respect i to the General Atomic's model in that no credit is given for delay in the release of fission products from the fuel kernel or transport through the graphite matrix. 1 THE SAFETY RESEARCH FUNCTION AT THE NRC Safety research has traditionally served both to delineate potential reactor accidents and to provide improved understanding of accidents post-ulated by design or licensing activities. The Office of Nuclear Regulatory l Research through its predecessor organization in the AEC established in early 1974 research programs in gas cooled reactor safety. Originally these programs were oriented to confirm the safety design of General Atomic's large HTGR concept. These programs took into account licensing needs developing from the Office of Nuclear Reactor Regulation and the Advisory Committee on i Reactor Safeguards, safety design and research activities at General Atomic, safety and developmental research underway at Oak Ridge National Laboratory and research, development and licensing activities in England, France and West Germany. More recently the program has been broadened to include specific ~..,, .m s n., application to the Fort St. Vrain reactor and a more intensive examination of Class 9 accidents while reducing the level of effort devoted to large steam cycle HTGR confirmatory research. Table 3 identifies the NRC research activities that will be underway in fiscal 1979 that pertain to both the generic HTGR safety issues and the Fort St. Vrain reactor. These safety issues were developed from the broad back-ground of HTGR information and experience available both in the United States and abroad. Identification of these activities follows below for the fiscal year which began in October 1, 1978. STRUCTURAL GRAPHITE CORROSION This would continue the oxidation studies on structural graphites which recently have been investigating the use of Type PGX graphite in the Fort St. Vrain reactor. We are seeking funds to initiate the construction of a high pressure test loop for the exposure of relatively large graphite specimens under closely controlled conditions of temperatures and environment. CONVECTIVE FLOW MIXING This activity addresses the problem of the effects of hot plumes in the upper plenum and hot streaks in the lower plenum of the reactor vessel under conditions where forced circulation has been lost or restarted after a temporary loss. The principal focus of this activity is on the Fort St. Vrain reactor although the study could be applicable to any large HTGR with similar core cooling geometry. GRAPHITE INSERVICE INSPECTION Changes in ultrasonic velocity can be used as a non-destructive measure of changes of graphite strength at least for some structural graphites. This activity is expected to refine the techniques involved in this measurement so that a non-destructive means will be available for monitoring the strength of ( the core support graphites. CORE HEATUP A high temperature induction furnace has been used to investigate the behavior of uranium and fission product migration under conditions approaching graphite sublimation temperatures. Work continues to gather more data in this area which is pertinent to the hypothetical unrestricted core heatup accident. CORE SEISMIC RESPONSE This activity has two pcrtions. The first is the verification of computer programs against scaled experiments. The second is the attainment of additional experimental modeling data taking into account improved knowledge of scaling parameters. l

. PCRV FAILURE MARGINS In this activity, a multi-dimensional, finite element computer code (NONSAP-C) is being used to investigate PCRV failure data being obtained by programs funded by DOE. HIGH TEMPERATURE PROPERTIES OF METALS Creep-rupture, fatigue and other information pertaining to the design and performance of the alloys Incoloy 800H, Hastelloy X and Type 422 stainless steel (fatigue only) are being obtained in a representative helium environment. FISSION PRODUCT TRANSPORT This activity considers fission product transport within the primary coolant system and through hypothetical breeches of the primary system boundary. Mechanisms of deposition and removal from metal and ceramic surfaces are being studied in terms of the chemical species involved, t DIRECT CYCLE ACCIDENT DELINEATION The scope of this activity is being defined in conjunction with DOE's developing plans for a direct cycle HTGR. COMBUSTION IN CONTAINMENT Postulated low probability accidents can result in release to the contain-ment from the PCRV of quantities of gases such as carbon monoxide and hydrogen that under certain circumstances may be flammable or detonateable. The actual combustion potential will depend on the mixing characteristic of these gases within the containment vessel and it is this phase of the issue which is receiving principal attention. / TRANSIENT ANALYSIS \\ This activity includes the development of a generalized computer code that will simulate dynamic characteristic for Fort St. Vrain, and eventually advanced HTGRs. The code consists of an executive rcutine which operates in conjunction with selected modules describing individual reactor system components. ROLE OF PROBABILISTICS RISK ASSESSMENT The use of probabilistic risk assessment techniques in the nuclear regu-latory process is an often discussed topic with the expectation that this methodology will become of increasing importance. As Saul Levine, Director of l the Office of Nuclear Regulatory Research pointed out, the questions that remain on these techniques are: "Should their use continue to increase, and how can their usefuln9ss be increased?"s The use of risk assessments would of G.*

course be expected to recognize the limited design and operational experience available for HTGR relative to light water systems. Our position in this regard was recently stated as follows: The staff has utilized in the past and is continuing to use a.yi apply probabilistic risk assessment techniques to specific areas and situat!ons in our safety review mostly as a supplement to and not a substitute for NRC regulations and the guidaace embodied in regulatory guides, standard review plans and branch technical positions. We feel that such studies extend and add to the bases provided from NRC's deterministic evaluations. The important + point is that systematic and disciplined evaluations of a plant design (to identify potential causes and pathways for serious accidents which could result in significant releases of radioactive material to the environment) provide additional insight and added assurance that important safety considera-tions are identified.6 It is in this spirit that the NRC welcomed General Atomic's continuing " Accident Initiation and Progress Analysis" (AIPA) study.7 The AIPA study was t' initiated in early 1974 and has produced nine volumes of material on the l probabilistic approach to HTGR accident analysis. The first seven volumes were reviewed by many individuals and organizations. The NRC staff found in its review that on balance the study was beneficial but that "... revisions, some of which are rather fundamental, are required to advance the cause of the study's credibility and its general utility". General Atomic provided in Volume 8 a discussion of all the comments received and in the most recent volume additional studies that reflected the impact of the comments.8 The study concluded that the HTGR reactor is "at least as safe as light water reactors". Reference 8 also contained a description of the unrestrictive core heat-up accident carried out to 240 hours. The NRC staff is developing a plan for both the continuing review of the AIPA study and the development of its own approach to the use of risk assessment techniques in advanced reactor licensing and research. Details of these plans should be forthcoming in the not-to-distant future. { k CONCLUSIONS In this survey we have seen that there is much direct and indirect activity pertaining to the postulation of HTGR accidents, particularly those of low probability. The changes in the safety design of future HTGRs which will result from these activities are not clear; however it is expected that there will be changes. Since both the AIPA study and the Reactor Safety Study l (WASH-1400) have indicated that the public risk from reactor accidents is 1 controlled by Class 9 events, efforts to increase the understanding of their phenomenology and to improve the quantification of their consequences and the mitigation of those consequences will be pursued. We believe the role of probabilistics risk assessments will become stronger and an increasing aid to our determinations but for HTGRs, at least, we do not foracast that decisions strongly affecting the HTGR's reactor safety design will be dominated by this methodology. i l

!' REFERENCES 1. ROBERT A. CLARK AND PETER M. WILLIAMS, " Licensing Criteria for HTGR; in the United States," Proceeding of Topical Meeting on Thc.aal Reactor Safety, Sun Valley, Id, August 1, 1977, CONF-770708, American Nuclear Society, La Grange Park, Illinois. 2. R. P. WICHNER, "Effect of Steam Oxidation Corrosion on HTGR Core Support ~ Posts Strength Loss," Part II. Consequences of Steam Generator Tube Rupture, Oak Ridge National Laboratory, ORNL/TM-5550, January 1977. 3. M. TOKAR, " Evaluation of High Temperature Gas Cooled Reactor Fuel Particle Coating Failure Models and Data," NUREG-0111, USNRC, November 1976. 4. M. TOKAR, W. F. PASEDAG, and P. M. WILLIAMS, "The Relationship of HTGR Fuel Failure Models to Potential Off Ste Doses, Presented at the Annual Meeting of AAAS, Washington, D.C., February 13, 1978. ( '". 5. SAUL LEVINE, "The Role of Risk Assessment in the Nuclear Regulatory Process," Nuclear Safety, Vol 19, No. 5, Sept.-Oct. 1978. 6. W. P. GAMMILL, letter to Gas Cooled Reactor Associates, June 22, 1978, available NRC Public Document Room, Washington, D.C. 7. GENERAL ATOMIC COMPANY, "HTGR Accident Initiation and Progression Analysis Status Report," GA-A13617, Vols 1-8, January 1976 - January 1977. 8. GENERAL ATOMIC COMPANY, "HTGR Accident Initiation and Progression Analysis Status Report, Phase II Risk Assessment," GA-A15000, April 1978. .I 6 I

~ 4 TABLE 1 CLASSIFICATION OF ACCIDENTS FOR HTGRs Classification Reactivity Steam or Water Depressurization Loss of Forced Insertion Ingress Circulation Accidents of lesser Loss of fixed burnable injection from helium Slow depressurlistion. Temporary loss of main consequences than poison. circulator water Rapid depressurtration cooling system. design basis Loss of fission product bearing seal. less than design basis Loss of feedwater. accidents. T hese poison. Failure in reacter depressurization are representativf Core component vessel liner cooling accident. and not inclusive. rearrangement. t ube. Moisture ingress. Steam generator leaks. Decrease in reactor tempe ra tu re. Control rod notion. Design basis accidents. Sparious rod with-Steam generator tube Rapid depressurization Sustained failure drawal terminated or header failure, rate determined by of normal core cooling, by protective actica. 100 square inch area. Osrestricted core heatup.* Engineered safety Reactor protection Steam generator dump Reactor vessel closure Core au illary cooling features limiting system. and isolation system. design. system. consequences of Containment vessel. Containment vessel. design basis accidents. s Accidents precluded Control rod ejection. Large moisture ingress Depressurization area Unrestricted core featup by design pro-Core drop. combined with reactor greater than 100 sq. in, in combination with visions. depressurization or Depressurization combined containment failure. core heatas. with containment failure.

• Unrestricted core heatup is the design basis for the maximum hypothetical fission product release used for reactne siting purposes.

w-w e

TABLE 2 i EFFECT OF FUEL FAILURE MODEL ON MHFPR DOSES

• 2 HR DOSE AT 30 DAY DOSE AT EXCLUSION DISTANCE LOW POPULATION ZONE THYROID WHOLE BODY THYROID WHOLE BODY This Calculation 16

<0.1 72 0.6 Summit HTGR Model 250 1.8 174 1.9 GA Calculation (Ref. 15) 4.2 <0.1 13.7 0.3 Typical PWR 285 5.5 28 1 10 CFR 100 Guidelines 300 25 300 25

• Doses (in rem) were calculated for a 3000 MWt plant with the assumption f

of a single barrier, low-leakage (0.1% per day) containment located on an arbitrary site with an exclusion radius of 900 m. and low population zone distance of 3000 m. R.G. 1.3/1.4 meterology without any wake factor corrections was assumed. ( TABLE 3 NRC RESEARCH ACTIVITIES FERTAINING TO HTGR SAFETY ISSUES E 3 i. 3 I E 2 T D t E e I E' 3 & 3 E R 8 I E E : y ex a E s u 8 a A t E = x V a E @5 a w 3 3 3 a e' 'a E i & x ( c T, t 3 2 I e T x e 2 e T 8 y r 7 3 2 a t 2 E T 2 g i .2 b 3 d' 3 3 8 A w Activity Structural Graphite Corrosion x* x x x Convective riow Hixing x x x x x x Grarbite Inservice Inspection x x x x x X Core Heatup x x x X Core Seismic Response x x x x PCPV Failure Margins x x x Hinh Tamp. Properties of Metals x x x x x Flesinn Prottuct Transport x x x x x x Direct Cycle Accident Delineation x x Ccnbustion in Containsnent x x Transient Analysis u x x s o}}