ML20150B901
| ML20150B901 | |
| Person / Time | |
|---|---|
| Issue date: | 09/18/1978 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1556, NUDOCS 7811100001 | |
| Download: ML20150B901 (73) | |
Text
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bb d MINUTES.0F THE FAST FLUX TEST FACILITY (FFTF)
SUBCOMMITTEF ..EETING JULY 12 ;978 WASHINT u" /g qg' - /$56 efP 4 ///2-/py 5
FFTF Subcommittee of the AC- .1et on July 12,1978 at 1717 H St., NW,
- nashington, DC. The main purpt 2 of the meeting was to review the status of the FFTF Project in connecti n with operation of the reactor. Notice of the meeting was published in the Federal Register on Tuesday, June 27, 1978. Copies of the notice, meeting attendees, and the schedule are in-cluded as Attachments A, B, and C, respectively. The Designated Federal Employee for the meeting was Dr. Andrew Bates. No requests for time to make oral statements were recieved from members of the public and no written statements were received.
INTRODUCTORY STATEMENT BY SUBCOMMITTEE CHAIRMAN Dr. Kerr, Subcommittee Chairman, convened the meeting at 8:30 a.m., in-troduced the ACRS members and consultants who were present, and indicated
-that the meeting was being conducted under provisions of the Federal Advisory Committee Act and the Government in the Sunshine Act.
Dr. T. Speis, NRC Staff, indicated that they would be discussing the NRC review of the FFTF Project. The last meetings with the ACRS took place in 1974 and 1975 and dealt with two open issues from the construction review. These issues, sealing of the reactor head cavity and ex-vessel post accident core retention, were addressed in a July 15, 1975 letter.
.Dr. Speis indicated that the Staff would discuss the bases for their current review of the Project and the areas where they have disagree-ments with the Project.
Dr. Simpson, HEDL', indicated that the Project staff would be reviewing i the FFTF design and the current status of the Project. They would also !
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be reviewing the status of issues that were still to be resolved at the I
-construction authori7.ation stage.
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FFTF 7/12/78 The FFTF is a 400 MWt sodium cooled fast reactor. NRC (AEC) review of the facility began with the submittal of the PSAR in the fall of 1970,
-interim construction approval was given in July of 1971, and submittal of the FSAR occurred in 1976. There have been some 14 meetings netween the NRC and the Project since the FSAR submittal, and the Projecc has responded to about 800 questions. Construction began in Octobr.r,1971 and is now better than 90% complete. Sodium was introduced into one
. secondary loop on July 2,1978 The scope of the NRC review, as requested by DOE, includes NRC advice on the adequacy of the FFTF design to ensure safe operation and the ade-quacy of the FFTF Technical Specifications. The review did not include assessments of the "as built" configuration, construction audits, or evalua-tion cf ecceptance test results to verify that the plant was constructed in accordtnce with the design criteria. Safeguards and security were also excluded from the NRC review. These are taken care of by DOE and its contractors at the Hanford Reservation. Dr. Simpson indicated that the FFTF is a project of the Reactor Research Technology Division of DOE which is administered by the FFTF Project Cffice in Richland, Washington. HEDL (Westinghouse-Hanford) is the project manager. Sub-contractors include WARD for reactor plant design, and HEDL, ANL, AI/LMEC, ORNL, LASL, and GE for technology development. Bechtel, AI, and AMC0 l
had responsibilities for construction, fuel handling equipment,
. and maintenance equipment. Construction is essentially complete
- and operation is being turned ov'er to HEDL by the subcontractors.
.Startup testing is in progress and the fuel loading is scheduled for j zMay of 1979.
The reactor site is on the Hanford Reservation outside Richland, Washing-ton. The design basis earthquake corresponds to an SSE of .259 The Project plans to shut the reactor down for inspection after any earth-quakes whc., produce accelerations greater than .05g. The WPPSS 1, 2, and 4 plants are located on the Hanford Reservation within 2-4 miles and have SSE values of 0.32g.
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e a FFTF 7/12/78 Mr. T. Mangelsdorf review the status of construction for the facility.
The Project is ahead of schedule and is 96% complete. The formal procedures by which the completed systems are turned over to HEDL by Bechtel were reviewed. HEDL accepts responsibility for the systems after construction, and conducts the acceptance testing and startup procedures on the equipment. The turnovers are documented and an in-plant inspection takes place at the time of the turnovers.
Mr. Mangelsdorf described the FTR experiments 1 program giving the numbers of fuel assemblies in the core devoted to advanced fuels, -
reference fuels, material programs, absorber programs, safety programs, industrial participation, and foreign exhange programs. (Attachment D ).
Mr. W. McShane reviewed the design of the reactor plant and refueling system (Attachment E). The 400 MWt (Pu/U)0 reactor is cooled with 2
sodium and has eight positions in the core for conducting tests.
Three heat transport loops remove heat at a design flow of 14,500 GpM per loop. The reactor is housed in a s4 eel containment building 187 ft.
high and 135 ft. in diameter with a design pressure of 10 psi. The fuel assemblies are 13 ft. long and contain 217 fuel pins with stainless steel clad. The reactor s ssel is 304 Stainless steel 43' high, 20 ft.
in diameter, and 2.75" thick. System pressure is low with a 15 psi
- overpressure on the vessel's argon cover gas. There are three instru-
- ment trees loacted in the reactor above the core. These measure the temperature and flow of sodium from each assembly, support the control rod drives, and provide secondary holddown for the core. The core support and restraint systems were described (Attachment F). The heat transport system consists of 3 primary sodium loops, 3 primary pumps and intermediate heat exchangers, and 3 secondary sodium loops. Each secondary sodium loop is connected to four air blast heat exchangers.
One loop of the heat transport system is tornado hardened. All loops are physically separated and isolated in steel lined N inerted chambers.
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The primary system is designed to prevent the core becoming unwvered if a leak occurs in piping, pumps, or' heat exchangers through the
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FFTF 7/12/78 use of guard vessels, piping elevations, and volume of sodium available for draining into the guard vessels. Cecay heat can be removed through pony motors on each of the primary pumps.
In response to a question, Mr. McShane indicated that the control rod mechanisms were identical but that subsystems other than the mechanical portions were designed to provide what can be interpreted as two diverse scram systems.
The primary and secondary system components were extensively tested at HEDL and the Liquid Metal Engineering facilitias before final installation at FFTF. The system fluid dynamics characteristics and flow distribution system were all mocked up and studied with tests carried out on finished companents in water.
Pony motors on the primary pumps supply 7.5% core flow for decay heat removal and receive power from the plant's emergency power supply (not Class lE). Analyses were made of primary pump failure caused by seizure or rupture and the effects on the reactor. The plant was designed to have the capability for natural convection cooling for decay heat. This is to be test <a during the startup program. Preheating of the sodium in the vessel and heat transfer loops is provided by electrical trace heating on pipes and oil burners located near the dump heat exchangers.
The Project has concluded that all offsite power and onsite diesels could fail and they could still maintain core cooling. The reactor system does have a Class lE battery system with DC/AC converters to supply instrumentation and control needed for the natural circulation of sodium in a decay heat removal mode. The batteries are sized to provide power for natural circu.lation control and monitoring for at least a week.
Refueling of the core is conducted with an in-vessel handling machine
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(IVHM) which transfers fuel from the reactor core to an in-vessel storage location. After a period of decay, the fuel assembly is moved into an ex-vessel handling machine (EVHM) and is transferred to an interim decay storage pit in the containment building. After a further
FFTF . -S- 7/12/78 period of decay, the fuel can be removed to the fuel storage facility.
Mr. R. P. Warrick reviewed th'e FFTF instrumentation and control systems (Attachment G). The electrical system is supplied by two independent offsite sources. There are two redundant onsite diesel generators and batteries designed to handle emergency ?oads and there are two redundant Class li. systems (battery supplied) which will power the reactor shutdown system, post accident monitoring equipment, and engineered safety features. The emergency diesels are not Class lE.
Mr. Warrick indicated that an analysis of the last 20-25 yeart showed that the Hanford site had never had a loss of grid power. The reactor facility has two cont ol channels which are redundant and are physically separated from each other. Primary and secondary scram breakers are widely separated. Three channels of plant instrumentation are also physically separated. Plant electrical systems are divided into normal, emergency, and Class 1E sys+9ms and all are separated. The principal differences between the emergency and Class lE system is the lack of seismic and tornado qualification in the emergency system.
l The control room is set up to provide the operators adequate access and visual observation of instruments needed for monitoring the reactor system status. Plant shutdown can be controlled from outside the control room at any one of several local control panels.
The plant has approximately 32,000 instrumentation components which provide signals to the plant protection system, the automatic control systems displays for manual control, and surveillance information.
The control room is protected by an automatic sprinkler system which is armed by a smoke detector. Smoke detection instroduces water into the system. Heat melts fusible links in the sprinkler heads.
The plant protection system includes the reactor shutdown system and the engineered safety features. The reactor shutdown system includes
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, two independent' shutdown systems with reduandancy in each system and I
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- > sFFTF 7/12/78 l diversity in the logic, hardware, and testing of the two systems.
The syst'em meets the requirements of RDT Standard C16-IT and is con-sistent with IEEE 279. One system scrams the six control rods, the other system scrams the three safety rods. Reactor shutdown s. 'em input includes flux, primary loop pressure and flow, intermediate heat-exchanger outlet temperature, reactor vessel level, and re-actor outlet plenum temperature. The systems are diverse in that one uses integrated circuits, the other discrete. components, one uses light pipe isolation, the other relay contact isolation; one uses DC signals, the other AC' signals; one has ITF scram breakers, the other Westinghouse scram breakers.
Engineered safety feature post accident monitoring instrumentation is provided for IHX outlet temperature, reactor outlet temperature, containment pressure and temperatures, and interim decay storage tempera ture . In response to a question, Mr. Warrick indicated there were in-plant dial telephones, a public address system, and a port-able radio available for emergency use; however, none of these meet Class lE standards.
Mr. Warrick reviewed the in-plant radiological controls. The plant is broken down into uncontrolled and controlled zones for radiation shielding purposes. Uncontrolled areas have dose rates
- of less than 0.2 mr/hr. Controlled areas range from Zone I at a 0.2 mr/hr limit to Zone V, no access area. There are area monitors in plant work locations, effluents are monitored, and personnel radiation monitors are required. Exposures are limited to as low as practical and the requirements of ERDA manual Chapter 0524.
Solid and liquid radioactive wastes will be disposed of through existing facilities on the Hanford Reservation, gaseous radioactive waste will be collected and held in a cyrogenic gas reprocessing system. Releases of nob % gases are expected to be below 0.0032 Ci/ day and Tritium below 0.088 Ci/ day.
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.Dr. Simpson reviewed the design basis -accidents (DBA) for the plant ..
(Attachment H). The approach to the design looked for various faults and abnormal conditions which could be anticipated during the reactor life or were either unlikely-or extremely unlikely. The Chapter 15 (
accident analysis includes various reactivity addition events, loss of cooling events, and plant events such as fires, earthquakes, and tornadoes. The reactivity addition events are all well within the capability of the PPS to handle. Loss of coolant flow events can be handled by natural circulation cooling. Natural circulation tests will be conducted during startup operations. If the natural circulation tests should prove to be unsuccessful, the Project would have to consider the installation of a Class lE system to provide ,
power to the pony motors for forced circulation cooling.
- Dr. Simpson indicated that the reactor was designed to prevent a core disruptive accident (HCDA) by positive design features. The Project has not identified any HCDA initiators that they consider realistic.
They did study the consequences of loss of flow (LOF) without scram and transient overpower (TOP) without scram events. The Project ,
1 believes that the consequences of this type of event would be well below 10 CFR 100 dose level .
Dr. Simpson reviewed the status of issues that arose during the con- +
struction authorization review. These issues included: (1) the containment margin for an HCDA, (2) fallback design positions for the head compartment and core retention system space, and (3) various natural phenomena. With resp .t to the HCDA the Project indicated that there was no realistic basis for postulated initiating events. They did study the consequences of a LOF or TOP event with-out scram. The MELT and SAS computer codes were used to look at these events. The calculated consequences of these events indica ted that there would be a coolable fuel configuration in the vessel che reactor and piping would remain intact for heat removal, and t .at offsite radioactivity would be below guidelines. Complete sodium 4
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FFTF -8 J/12/78 voiding of the core would provide $1.00 of negative reactivity, selective voiding of the central fuel assemblies could cause a reactivity insertion of about $3.00. In response to a question from Dr. Mark and Dr. Stratton, the Project indicated that the prompt neutron lifetime was 4 x 10 ~7 seconds and the doppler coefficient was negative and has a value of about -0.005 (dK/dt). Dr. Mark and Dr. Stratton (see attached letter from Dr. Stratton) indicated that if the core should void in the most reactive configuration, the heat addition to the core would not be fast enough to allow an explosive disassembly of the fuel .
Dr. Simpson indicated that the most probable result of a TOP or LOF without scram would be a non-energetic core dispersal. Studies showed that even reactivity insertions of up to $100/second were within the containment capabilities. The NRC Staff agreed that the studies were adequately conservative and that the energy produced is not likely to exceed the containment capability, although they did recommend additional research.
The Project studied the need to seal the reactor head cavity area to prevent a common atmosphere with the containment working areas. Study of accident sequences did not show significant safety advatages to sealing the head area; it would be a disadvantage from an operational standpoint. It was decided that sealing the head area was not neces-sary. The NRC Staff agreed and suggested that additional sodium deflection provisions be added to the design to prevent sodium from being expelled into the containment atmosphere.
l During early FFTF construction, a space was left below the reactor cavity to allow the installation of a core catcher if necessary.
Based on studies of HCDA, energetics and postulated melt-through condi-tions, the Project concluded that this space (21' in diameter x 14' deep) should be filled with concrete. The NRC Staff in 1975 recommended addi-tional studies and suggested that the lower cavity be sealed. The Project scaled the top of the lower cavity with a 3' concrete slab.
_ _ - _ -_ _ . _ _ _ - n-RA d
f 6 i ~ FFTF 7/12/78 The sealing of the lower cavity and the non-sealing of the reactor head were agreed to by the ACRS in a July 1975 letter.
Following the lower cavity sealing decision, the Prcject conducted additional studies on the containment margins. Results ..f these studies indicated that provisions for containment venting would add flexibility and would thus be incorporated; that large capacity filters would be effective but that the margins were adequate without them and they were therefore not cost effective; and that containment coolers and purge capability were of marginal value and were not justified.
Dr. Simpson reported on the work they have done on sodium / concrete interactions. Using data from 22 small scale ter,ts and more recent tests at Sandia, the Project has developed a model which they believe shows a maximum penetrations of Na in concrete of about 12 inches.
Calculated radiological exposures at the site boundary were shown for various assumptions of the molten core behavior. The Project believes all of the doses would be below 10 CFR Part 100 guidelines.
Using NRC Staff assumptions for accident calculation results in higher doses, in one instance ibout 1270 rem to the thyroid.
Dr. Simpson concluded that considering the combination of tue extremely unlikely accident with the consequences that were cal-culatei, the containment system is adequate for FFTF.
Dr. Simpson reviewed the question of piping integrity and in-service inspection. The reactor core is most sensitive to an inlet pipe break.
A break in the inlet pipe could lead to, sodium boiling in the core during the flow reduction transient. A piping integrity evaluation was made and it was concluded that there is likely to be negligible crack growth based on fracture mechanics and that through wall cracks would develop and leak before a critical length crack could develop.
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FFTF 7/12/78 Leak detection instruments are to be located near the cold leg to detect any small Na leaks. The NRC recommends that the Project pro-vide prcservice and inservice inspection of selected welds on the primary piping, demonstration of leak detection sensitivity, and l qualification of the leak detection system as seismic Category 1.
l The Project will do the preservice baseline inspection but do not want to do inservice inspection on the primary system. They do not yet have the capability to do UT testing on high temperature systems. When it is developed they want to apply it to the secondary system only. The Project does not consider seismic qualification of the leak detection system necessary.
Dr. Simpson reviewed the question of natural circulation and emergency power. The reactor and heat transport systems were designed to provide natural circulation decay heat cooling. Under normal condi-tions, decay heat can be removed with pony motors on each coolant pump. Emergency diesels (not Class lE) can supply power in the event of loss of offsite power. A Class lE system is available to monitor the plant and control valves if both offsite and onsite emergency diesels are lost. One loop of the reactor system is tor-nado hardened and is designed to provide natural circulation cooling.
.The NRC has agreed that the design is. adequate subject to verifi-cation during startup tests. The NRC wants to be provided with the test results for review.
Dr. Simpson discussed changes that have been made in the FFTF design.
Major changes already reviewed relate to the containment margin, sodium spray deflection, and the dedicated safety power supply system. Other changes include a change from mobiltherm to water-glycol coolant for the plant auxiliary systems and an upgrading of the fire protection system. The use of water-glycol eliminates the problem of nobiltherm fires and increases cooling capacity. The fire protection system provides space isolation and nitrogen flooding, a water supply and sprinkler system, an aqueous foaming system, automatic gas blanketing u , __e...
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FFTF - 7/12/78 systems, and sodium carbonate portable extinguishers. NRC review of the fire protection is not yet complete.
Mr. Additon reviewed the planned testing for natural circulation decay ' neat cooling (Attachment I). The tests are designed to show that the system is capable of removing decay heat with appro-priate allowances for uncertainties and residual margins. Loss of electrical power tests will be conducted at 5%, 75%, and 100% power.
The NRC has asked that a test also be conducted at about 25% power to fill in the gap between 5% and 75% power.
Mr. H. Holz discussed the NRC Staff's review of the FFTF (Attachment J).
DOE requested NRC to provide advice on the adequacy of the FFTF design.
(NRC and DOE letters of August 13 and 20,1976, respectively). The scope of the review involved a full technical review of the FFTF N sign.
Safeguards and Security were not reviewed by the NRC. The closed loop test assembly operation, fire protection, and technical specifications will be addressed in the future. The review was based on existing rules and regulations, the Standard Review Plan, Regulatory Guides, and the safety objectives of the General Design Criteria. The Staff review utilized NRC personnel that review LWRs and have reviewed CRBR also.
In addition, the Advanced Reactors Branch and their consultants re-viewed areas special to LMFBRs. The review was comparable to an LWR power plant review insofar as was possible in the areas actually re-viewed. Mr. Holz listed the problem areas that are not yet completely resolved (Attachment K).
Dr. Kerr identified areas of interest to the Subcommittee and consul-tants for the August 12, 1978 meeting.
Dr. Kerr thanked the meeting attendees and adjourned at 5:12 p.m.
A complete set of slides used at the meeting is attached to the Office Record Copy.
NOIE: For additional details, a complete transcript of the meeting is available in the NRC Public Document Room, 1717 11 St., NW, Washington, Ir 20555 or from Ace-Federal Reporters, Inc., 444 North Capitol St. ,
NW, Washington, DC.
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Copics of the petition nre availabic in topics. The subcommittee may then the Commission's Public Document energy companics, C. Iouis Kincannon,
' Room at 171711 Street NW Washlng. caucus to deterrnine whether the mat. 395-3211.
ton, D.C. Comments should be sent to ters identified in the initial session ,
the Office of the Secretary of the have been adenttately covered and ,
,c,,,w e ,7 o,co ,,c, Commission. U.S. Nuclear Regulatory whether the project is ready for Industry and Trade Administration. Repdrt review by the full committee. on Destruction or Proof of netum to U.a.
Commission, Washint: ton, D.C. 20555, Further information regarding Attention: Docketing and ,Scrylce toples to be discussed, whether the $'M'[n" he',$(("2 0'x e tt rs.Y*
Branch, by July 27,1978. .
meeting has been cancelled or resche- Louts Kincannon. 395-3 311.
It is so ordered' duled, the Chairman's ruling on a Industry and Trade Administration, Month.
quests for the opportunity to presen. ly neport on Itxports of Parts to Service By the Commission. Eautpment Shipped Against a Validated oral statements and the time allotted License. EAR 37G 4(D), monthly,1.200 ex.
Dated at Washington, D.C., this 21st therer r can be obtained by a prepaid porters. C. Louis Kincannon. 395-3211.
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DMUEL J. CultX, Andrew L. Bates (telephone 202-634 eting, r. T [ j ,' " * "'*""
Secretary of the Commission.
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ADVl50RY COMMITTEE ON REACTOR SAFE
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i FLUX TEST FACluTY [FR Doc. 78-17914 f!!ed 6-26-78: 8:45 aral Departmental and Other Biographical In.
formation on Proposed Advisory Commit.
88 teo Members, on occasion. 275 biographl.
The ACRS Subcommittee on the [3110-01] cal and personal information, Clearance Office. 395-3712.
i' Fast Flux Test Facility will hold an OFFICE OF MANAGEMENT AND epen meeting on July 12, 1978 in BUDGET
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- l Room 1046,1(17 H St., NW., Washing. ton, CLEARANCE D.C. 20555 to discuss the status v.s.cmL OF REPORTS of smteE comusstow j
- construction and ' the NRC Safety uo , n,q,,,,, Monthly Report of Federal Civilian Em.
- Review of the Fast Flux Test Facility.
' ployment.113-A & D. mcnthly, Federal Notice of this meeting was published The following is a list of requests for Govemment agencies.1.285 responses. 0 in the FEDERAL REctsTER on May 17 Clearance of reports Intended for use hours. Ofuce of Federal Statistical Fvlicy 1 and June 16.1978. & Standard. 673-7959.
in collecting information from the In accordance with the precedures public received by the Office of Man. "^'** ' **C"#*8 out!!ned in the FEDERAL REctsTER on agement and Budget on June ..0,1978 Food and Nutrition Service, Summer Food Octaber 31,1977, page 56972, oral or (44 U.S.C. 3509). The purpose of pub. Service Program for Children. FNS-80A.
i written statements may be presented l Ilshing this list in the FEDERAI. REcts, i
by members of the public, recordings ' Tzu is to inform the public. on occasion. 42.658 service institutions (s
Will be permitted only during those g nsoy, 42 9 61$2.
portions of the meeting when a tran, quest The list includes the title of each re-received; the name of the agency Evaluat!'o n"of Interim National School
' script is being kept, and questions may , sponsoring the proposed collection of Lunch Program WSI.P) Regulations and be asked only by members of the sub. Infarmation: the agency fortn Demonstration Projects. FNS 1045-47 &
committee, its consultants, and staff. I number (s),if applicable! the frequency 1094 to 2000, single time. 112.620 food Persons desiring to make oral state, with which the information is . pro. service cuanagers, student.s. parents, fac- i ulty. 8, 440 responses, 1.322 hours0.00373 days <br />0.0894 hours <br />5.324074e-4 weeks <br />1.22521e-4 months <br />.
ments should notify the Designated posed to be collected; an Indication of iluman Resources Division. E11ctt, C. A I Federal Employee as far in advance as who will be the respondents to the 395-3532. 1 practicable so that appropriate ar. proposed collection; the estimated rangements can be made to allow the number of responses: the estimated mmwErrr or nEALTH, EDUCAT1oN, AND necessary time during the meeting for burden in reporting hours; and the wEtms such statements. name of the reviewer or reviewing divi
- National Center for Education St.atistics, The agenda for subject meeting sion or office. EMEGIS: Eennial and Fall . Surveys, '
shall be as follows: Requests for extension which appear 2.350,5 through 0.11 through 13. annual.
to raise no significant issues are to be ly, 57 responses. 7.980 hours0.0113 days <br />0.272 hours <br />0.00162 weeks <br />3.7289e-4 months <br />. Office of WEDNESDAY, JULY 12,1978 Federal Statistical Polley & Standard 14-approved after brief notice through veme v. Cottins. 673-7959.
s:30 A.M. UNTIL T!!E CoNCLUsroN oF this release. Food and Drug Administration:
sesmEss Further information about the items Cosmcue Raw Material Composition on this daily list may be obtained from Statement, FD-2513 and 2513A. on occa.
The subcommittee may rneet in ex. the Clearance Office. Office of Man- sion, 600 responses, 300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br />. Richard ecutivo session, with any of its consul. agement and Budget Washington, D.C. E! l 8 *8 tants who may be present, to explore 20503, 202-395-4529, or from the re- Notie f uchntin'uance of Commercial and exchange their preliminary opin. viewer listed. Distributton of Cosmette Product or lons regarding matters which should C93rnetic Raw Material. FD-2514. on oc-be considered during the meeting and y,,pn,,, caslon, 2.600 resoonses. 1.300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br />, to formulate a report and recommen* Cicarance Olnce. 395-3772.
dations'to the full committee
- CENERAL SERV!CES ADu!N!sTRAT!oN Transmittal of Periodic Reporta and Pro-y gle concgon og t e executive Associates of the National Archives Survey. motional Material. FD 2301. on occasion, session, the subcommittee will 1. car single time.1.200 members of asacciates 4.200 pharmaceutical firms, 2.583 re.
program. Clearance Office. 395-3772. sponses.10.332 hours0.00384 days <br />0.0922 hours <br />5.489418e-4 weeks <br />1.26326e-4 months <br />. Illchard Elsinger*
-- presentations by and hold discussions 395-3214 sith representatives of the NRC staff, mmwewT or macy National Ce'nter for Education Statistics, the Departinent of Energy, and their Common Core of Data Survey. 1978-79, consultants, pcrtinctit to the above ,s:ncrry Cornpany P nancial Reporting 2.303, 2 through 4. annually,16.057 re-System. EIA-2h. annually, 30 major sponses. 33.140 hours0.00162 days <br />0.0389 hours <br />2.314815e-4 weeks <br />5.327e-5 months <br />. Office of Federal FEDERAL REGl1TER, YOt. 43, NO.124-TUE$ DAY, JUNE 17,1978 ' '
ATTACHMENT A -
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ATTENCANCE LIST ,
1 FAST FLUX TEST FACILITY (FFTF) SUBCOMMITTEE MEETING i JULY 12, 1978 l 1
WASHINGTON, DC ACRS NRC STAFF W. Kerr, Chairman J. Long H. Isbin T. Speis J. Mark J. Meyer W. Str.atton, ACRS Consultant P. Riehm R. Seale, ACRS Consultant H. Ornstein
. A. Bates, Designateo Federal Employee J. Carter A. Marchese BATTELLE NORTHWEST LABORATORY DOE R. Bari ,
G. Bouchey I ARGONNE NATIONAL LABORATORY A. Rizzo P. Davis L.. Baker, Jr. F. Gavigan J.-Marchaterre H. Alter W. Lehto J. Griffith R. Hunter WESTINGHOUSE-HANFORD W. McShane ,
T. Mangelsdorf ,
'S. Additon '
D. Simpson R. Warrick D. Stepnewski R. Peterson J. Ziff
. ATTACHMENT B 4 g e
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REVISION-
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PDOPOSED PRESENTATION SCIIEDULE. .
ACRS SUDCOtflITTE MEETING ON FPIF ,.
7ULY 12., 1978 PRESENTATION APPEOXIMATE TIME TIME -
1 I. EXECUTIVE SESSION - OPENING COMMEtfrS - ACRS 8:30 am - 8:40 am
. II. REVIEN OF FFTP PECLTECT - DOE ,
- a. : DOE ViewJon What'is Being Requested in 5 mins. 8:40 am - 8:45 am NBC Review.
- b. Construction' Status and Schedule for 10 sins. 8:45 am - 8:55 am r Pre-operational Tests and Startup
- c. Review 'of- Site 15 mins. 8:55 am - 9:10 am
)
- 1. seismicity - new information and ,. .
changes since C...P. Stage -
2.' meteorology- .
-3. population density 1 hr. 50 mins.
d . -. Plant Layout and Design' 9:10 am - 11:00 am
- 1. reactor system - core 15 mins.
- 2. heat removal systems- 15 mins.
refueling systems
- 3. . .15 mins.
- 4. control and instrumentation 20 mins.
- 5. engineered safety features., 15 mins
- 6. radiological controls radwaste systems 15 mins. -
- 7. design basis accidents 15 mins.
- e. ' Disposition of CP Issues .
30 mins. 11:00 am - 11:30 am
- f. Major Design Changes Since PSAR and Safety Implications ,
70 mins. 11:30 am - 12:40 pm i
'1.. core' modifications .
- 2. diesels - non-hardened .
- 3. ' one loop and DIO: hardened
- 4. . add cavity vents. !
4
. ATTACHMENT C. .s
.,.E .
.j . .
y -
2- . .
4 PRESENTATT.CN APPPOXIMATE TIME TIME
- 5. add sodium anti-spray ,
- 6. Mobiltherm - glycoi water ,
- 7. HTS separate primary power 8.. N2 system excontainment fires '
LUNCH 60 mins. 12:40 pn - 1:40 pu
- g. Equignent Tests and Qualifications 10 mins. 1:40 pn - 1:50 pn a
- h. Startup Testing-Verification of Natural Convection 30 min. 1:50 pn - 2:20 pu Questions and Discussion 40 mins. 2:20 pm - 3:00 pn III. NRC REVIai OF PIGECT - NRC , 3:00 pm
- a. Type of Review Requested 5 mins. 3:00 pn - 3:05 pn-
- b. Basis for the Review 30 mins. 3:05 pn - 3:35 pn
- 1. applicable rules, regulations and regulatory guides t
2.' comparison between LWRs and.FFTF
- c. Degree to Which Review and Facility Conform to Power Plant Reviews and Design Requirements 15 mins. 3:35 pn - 3:50 pn
- d. Identification of Areas Where There-Have Been and/or. Are Significant Disagreements Between Staff and Project 30 mins, 3:50 pn - 4:20 pn Questions and Discassions' 40 mins- 4:20 pu - 5:00 pn
' kbecond ' ACRS Subcommittee on Aucjust .10,1978 would review, ,
-in depth,.the areas' identified under III-d)..
e
..n.. . . . . -,,m. v. . .
, l
~
~
FFTF TURNOVER SEQUENCE e TURNOVER SCHEDULING -
e INFORMAL WALKTHROUGHS .
- PRELIMINARY WALKTHROUGHS e COMPLETION OF CONSTRUCTION AND PUNCHLIST ITEMS e PRESENTATION OF TURNOVER PACKAGE e FINAL WALKTHROUGH ke ,
e TURNOVER ACCEPTANCE -
b d .
HEDL 7807-005.5 D
i t
l l BECHTEL CONSTRUCTION STARTUP SYSTEMS TURNOVERS 200 i i i i i i i , , i i i i i i i i i i i i
.l 180 -
160 -
p 140 - .
'y ,,
5 -
/ '138 FINAL
@ 120 -
/ SYSTEM -
f p
,' // ,,/
TURNOVERS -
- u. 100 -
~
~
y 80 -
/ -
g - ./' , -
60 -
/ -
40 -
/ -
O O O 20 e -
~
0 OlNlD J lF lM l A l M l J lJ lA l5 lOlNlD J l F lM l A l M l J l J l A l5 lOlNlD 9
1976 197i' 1978 HEDL 7805-363.E N
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FFTF ACCEPTANCE TEST PROGRAM INERT AND PREHEAT PHASES 111, IV, AND V T-1 SODIUM FILL BASE PLAN -
T-2, 3 SODIUM FLOW-TESTS T-4, 5 INSTALL AND TEST CONTROL RODS .
T-6, 7, 8, 9, 10 NON-NUCLEAR TESTS NIST) .
T-ll INSPECT WELDS
- T-12 LOAD CORE- -
14 ACTIVE INITIAL CRITICAllTYO T-13 CHARACTERIZATION .
POWER -
'T-15, 16 DEMONSTRATION FIRST FULL POWERO P0 i PdWE t TESTS T-20, 21, 22, 23, 24 CHARACTERIZATIONS
' 8-DAY POWER RUN T-24 FINISH TEST PROGRAM
. T-25, 26, 27 28, 29 AUG SEP OCT NOV DEC JAN FEB MAR APR MAY JUN JUL AUG SEP OCI NOV DEC JAN 1978 1979 1980 ,
HEDL 7807-005.33 :
M
. . ^
~
~
ACCEPTANCE TESTING PROGRESS .
e MAJOR AUXILIARY SYSTEMS TESTED AND IN OPERATION
- ELECTRICAL -
- COOLING WATER .
- chilled WATER ,
- INSTRUMENT AIR
- FIREWATER AND SPRINKLER -.
- HEATING AND VENTILATION
- SECONDARY SODIUM PROCESSINd LOOP
- EMERGENCY DIESEL GENERATOR e ILRT SUCCESSFULLY PERFORMED e SODIUM FILL OF FIRST SECONDARY LOOP e DUMP HEAT EXCHANGERS -- FANS, DAMPERS, PREHEATERS IN OTHER LOOPS TESTED AND READY FOR SODIUM OPERATION
'
- TRACE HEATING CIRCUlTS CHECK 0UT ON OTHER LOOPS IN PROGRESS D v
\
HEDL 7807-005.105 '
~
~
TESTING TO G0 BEFORE CORE LOAD
- HOT SODIUM FLOW
- FUEL HANDLING SYSTEM .
f.
e CELL ATMOSPHERE PROCESSING SYSTEM .
e IEM CELL EQUIPMENT e CONTROL ROD DRIVE MECHANISMS -
~
HEDL 7807-005.106 .
D b
oo
ll
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.L,-,,, - ., , - .s. .
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%.1 . - .; : ,- .. : :.,.- -
} ,
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Equipment Air Lock .
"d
.{j {,9] ,,.J.ontrol Room y,,.;
,* ri
- d L ., s L'. e
"/.1,, ' interim Examination Cell -
_ i::
t, . .en, G.
M. . . . *
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- ;g' l ;
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- 6. .
( <.l Interim Fuel Decay Storage \ {f:1 c.yj ,
pc s -Containment Vessel W I:.Y & -b %- / kb :
m Me- M Core Condit.ionmg Station %
3
- Auxiliary Equipment Buildings Fu.. .
.. . a ,
f
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$ T Secondary Sodium Pump [3]
,x
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I d bh m
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, p N fu.- ;
- ' %, Primary Sodium Pump [3] N M:'d Dump Heat Exchangers (12]
, 7 U@j l E
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'l FFTF PARAMETERS ,
j
~
REACTOR POWER 400 MWT -
DRIVER FUEL (74 ASSEMBLIES) (PulU)0 '-
2 .
CONTROL MATERIAL (9 ASSEMBLIES) BC 4 ,
CLOSED TEST LOOPS 4 DESIGN 2 INITIAL.
i 1 .
TOTAL TEST POSITIONS 8 PEAK; FLUX 7.0 x 10 nicm -SEC i n ;
~
i SODIUM TEMPERATURE - INLET 792 F DESIGN 680 F INITIAL . ;
1050% DESIGN 938 F INITIAL
- OUTLET U
- DIFFERENTIAL 350 F DESIGN 258 F INITIAL HEAT TRANSPORT LOOPS 3 i
^
FLOWILOOP '
14,500 GPM DESIGN 13,442 INITIAL
- I CONTAINMENT BLDG' HElGHT 187 FT M i P'#
~
DIAMETER 135 FT % l
' DESIGN PRESSURE 10 psi 3 '
h HEDL 7807-005.156 -
~
~
g -
i .,
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. . . . s FFTFiFAST TEST REACTOR PMAMETERS 1
. h-i REACTOR POWER 400 MWt
,~^
tg i i
~
680 F
~
~
CORE INLETiTEMPERATURE !
CORE OUTLET. TEMPERATURE 980 F .
CORE AT t- 300 t
FUEL': ASSEMBLY FLOW RATE 14.22 x 10 LB/HR ?
TOTAL REACTOR FLOW RATE 17.28 x 10 LB/HR . ,' . I il i
~
ACTIVE CORE HEIGHT - 36'IN.
i ACTIVE CORE DIAMETER 47.2 IN. -
ACTIVE CORE VOLUME 1034 LITERS :
POWER DENSITY O.39 MWlLITER .!
FUEL PIN LINEAR POWER PEAK 12.7 KWlFT '
FUEL PIN LINEAR POWER AVERAGE 7.3 KWIFT -
DELAYED NEUTRON FRACTION 0.003 15 2 PEAKiFLUX TO CLOSED LOOP 7 x 10 g SEC [
CONTROL ROD REACTIVITY WORTH !
PRIMARY RODS (3) 5.66% AKIK
, l
' h.
SECONDARY RODS (6) 6.38% AKlK -
g s1.035 SEC
.. SCRpM INSERTION TIME 4
'! I HEDL 7807-005.161 ~
a c
I
,: - ,~
~
~
- DRIVER FL8EL ASSEMBLY
.s
~~
~
~
- 12 F00T LENGTF . . .
4 ~ f
- 217 FL EL 3 INS 3El BUNDLE .
e 4
- _ EL 3IN !3 ITCH-TO-JIAE~El RA~::0 0F :..:A i
- TY3E 3:.6 STAINLESS STEE._1CT .
-l HEDL 7711-178.23 i
I M
1 I v
. I -
. 7
e s
~
REACTOR VESSEL.
- 304 STAINLESS STEEL s . . ,
~
-
- 43' x 20' .
e
- 2.75 INCH THICK WALLS
~
- FLEXIBLE ARM SUPPORT . .
-* 570 TONS WITHOUT S0DIUM
~
~
- 872 TONS WITH S0DIUM. geot 7733_178.19 q s t
- \
HEAT TRANSPDRT SYStaE & REACTDR
~
- - R,IE :.33, VW - EA~ ~1AN S 301~ _00 35~'
~
. i
. o ONE ;1,5C:l G3v 31: VARY'3Lv3 3El _00 3
~
~
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3B _1: C3 .
~ ~
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- O'N E :~ -X 3 El LOOP .
y u;t! . ,3 HEDL 7711-178.30 m 4
Il Yl s -
c - -- ---- - - - - - - ----------------------------L-------=
~
~
EAIN FAGLANT NMPS - .-
1 ,
~
- 3 PRIMARY AND 3 SECONDARY .
- SINGLE-STAGE, SINGLE SUCTION DESIGNS .
H L.
T
~
~
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e-1100LRPM ~
i'. I .
~
- S0DIUM FLOWRATE OF 14,500 GPM -
- 2.3:1 S' PEED CONTROL RANGE
- PONYMOT0RREkOVESDECAYHEATAT7.5% FLOW
~
HEDL 7711-178.8 "1 .
. %l
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DUMP HEAT EXCHAN3EE[ MIX)
~
- '
- e . .
g . .
- I- =
s.
~
- 33 MW. - -:
T - .
- TWELVE DHX MODULES REJECT HEAT TO AIR
- .. t .
- EACH CONTAINS TUBE BUNDLE, A 1250 HP . FAN, .
CONTROL VANES, MOVABLE OUTLET DAMPERS, AN OIL-FIRED PREHEATER & CONTROLS .--
- 3250 GPM SODIUM FLOW
- S34,.000 CFM AIR FLOW
.. t o
. ., HEDL 7711-178.28 l
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FFTF CORE SUPPORT STRUCTURE
[_ __.
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SUPPORT FOR I VESSEL THERMAL LINER
\ INSTRUMENT TREE l
% \
l
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', 4 SillELO AND CORE RESTRAINT P-j
/ IN VESSEL
+ CORE BARREL c STORAGE
- ' RADIAL SillELD SUPPORT LOCATION s'
/ \.
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CORE SUPPORT SKET LOCATION STRUCTURE SKIRT [ q e / p' y!
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~ . . ,
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00ME J
' 7511-60.2
~ .
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[.' STANDOFF \
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ELECTRICAL 4
KEY DESIGN FEATURES INDEPENDENT OFFSITE SUPPLIES _
e MAIN PREFERRED FOR OPERATING LOADS e ALTERNATE PREFERRED FOR SHUTDOWN LOADS
~
REDUNDANT ONSITE DIESEL GENERATORS e MINIMlZE ECONOMIC LOSS FROM OFFSITE POWER FAILURE -
REDUNDANT EMERGENCY BATTERY SYSTEMS --
e CONTROL POWER ,
e UNINTERRUPTABLE POWER FOR EMERGENCY LOADS REDUNDANT CLASS lE SYSTEMS e REACTOR SHUTDOWN SYSTEM p 8 POST ACCIDENT MONITORING
- ENGINEERED SAFETY FEATURES t
S HEDL 7807-004.1
i .
SIMPLIFIED SINGLE LINE DIAGRAM MAIN PREFERRED ALTERNATE PREFERRED REAR BUS POWER , , POWER H1 SWITCHGEAR I I I H2 I FRONT BUS 3 s I I '#
') .
) H3 })
d11 NORMAL LOADS TTA NA PUMPS - DHX FANS /
e e DG1 DG2
,e ,e
~
0% OCT h h 0C0 CO I
B4B ' B4A B10 lB5Al B5B 1 l') ')') 1 ) ) ) 'J')JJJ ') BATTERY BATTERY CHARGER 4 I II Ill l} l } CHARGER EMERG LOADS E MERG LOAb i EMERG LOADS, N
D1 D101 D102 ' D2 l' l> ) lI }l ) N H>;> ;> ;> ;> ;> ;>li;>
CLASS lE LOADS EMERG LOADS EMERG LOADS CLASS lE LOADS
~
~
CLASS lE EQUIPMENT LOCATION CH"B" SAFETY POWER SYSTEM HIS TRIP BREAKERS Ste
/- CONTROL ROOM B -CH'R" SAFETY POWER SYSTEM SEC SCRAM BREAKERS E PRI SCRAM BREAKERS CH"A" CONTAINMENT PENETRATIONS
\N3 b s2 CH'R" DHX CH'B" DHX -
CONTROLS CONTRO,LS \
m h r=1 6/ b o o Ih R si In
,6 -
h .
AH'B" CONTRAINMENT PENETRATIONS
, , - rm -
- p. q p si , . . , Is! si Is.r msi :
CH"A" INST / , p[g] p __ g)u CH'C" CONTAINMENT PENETRATIONS l O
[
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=
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CH'R" DHX CH'B" DHX CH"A" DHX CONTROLS N m M INST CONTROLS CONTROLS
~4 % gj
l=
CONTROLS 5[ 'l
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- 2
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DJ HEOL 7807-004.76 -
~ " . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . __ . _ _ _ _ _ _ _ _ _ _ _ _ _ .
- REACTOR SHUTDOWN SYSTEM KEY DESIGN FEATURES TWO INDEPENDENT SHUTDOWN SYSTEMS '
- e EACH INDEPENDENTLY CAPABLE OF REACTOR SHUTDOWN
- EQUIPMENT DIVERSITY FOR LOGIC, COMPARATORS AND BREAKERS ,
e FUNCTIONAL DIVERSITY FOR INSTRUMENTATION REDUNDANCY PROVIDED BY EACH SHUTDOWN SYSTEM e THREE INSTRUMENT CHANNELS (2 0F 3 TRIP) e TWO LOGIC TRAINS (10F 2 TRIP) _
e TWO SERIES BREAKERS (10F 2 TRIP)
DIVERSE FEATURES OF LOGIC SYSTEMS e HARDWARE IMPLEMENTATION DIFFERENT e TESTING METHODS DIFFERENT ,
MEETS REQUIREMENTS OF RDT STD C16-lT e CONSISTENT WITH IEEE STD 279 n D HEDL 7807-004.43 i
G-r m
REACTOR SHUTDOWN SYSTEM PRIMARY. SCRAM BREAKERS TRANSMITTERS
'-l I' (Rod Power)
COMPARATORS 2/3 1/32 g SENSORS o to >ay >- > _
>g r*-
o c to i
[ >[ >]. m h >Ij HTS BREAKERS
- HTS (HTS Power) CONTROL ROOS U l SECONDARY
>((g+-
Q" J<
9 I I, ;
SCRAM BREAKERS , LOGIC ,
(Rod Power) TRANSMITTERS 1/8 1/4
} ll 2/3 a '
COMPARATORS SENSORS
- LJ: : :
C: O: O
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- n. n_
i _s _ v v O ll l r: _!: _
- O O< C: O CONTROL ROOS. 8457 1 t
_ - - - -- -g-I b
=
PPS DIVERSITY -
PRIMARY SECONDARY INTEGRATED CIRCulTS DISCRETE COMPONENTS ,
LIGHT ISOLATION RELAY COILICONTACT ISOLATION DC SIGNAL LEVELS IN LOGIC AC SIGNAL LEVELS IN LOGIC LOCAL COINCIDENCE LOGIC HYBRID LOCAL-GENERAL COINCIDENCE LOGIC t
AUTOMATIC TEST CIRCUlTS MANUAL TEST CIRCulTS ITE SCRAM BREAKERS WESTINGHOUSE SCRAM BREAKERS
- 4 HEDL 7807-004.41 D
b
RADIATION ZONES FOR SHIELDING. DESIGN -
EXTERNAL DOSE RATE .
ZONE DESCRIPTION (mRemlhr)
UNCONTROLLED LUNCHROOMS, OFFICES, CONTROL LESS THAN ROOM, HALLS, ETC. 0. 2 CONTROLLED -
ZONEI ROUTINELY OCCUPIED, FULL-TIME LESS THAN
. BASIS 0. 2 ZONE 11 ROUTINELY OCCUPIED, 500 HR/YR LESS THAN BASIS 2. 0 ZONE Ill LIMITED ROUTINE ACCESS, 50 HRlYR LESS THAN BASIS 20.0 ZONE IV LIMITEP NONROUTINE ACCESS, 5 LESS THAN HRlYR BASIS 200.0 ZONE V NORMALLY INACCESSIBLE, NO ACCESS BAS!S HEDL 7807-004.12 D 4
- 2'
~
SOLID RADI0 ACTIVE WASTE SOURCE HARDWARE AND REPLACED EQUIPMENT
~
VOLUME 2,300 FT lYR LOW LEVEL: LESS THAN 100 FT lYR HIGH LEVEL ACTIVITY LOW LEVEL - NO SHIELDING REQUIRED FOR 100 mrcmIHR AT SURFACE HIGH LEVEL - SHIELDING REQUIRED -
DISPOSAL EXISTING FACILITIES ON HANFORD RESERVATION
^
23 9 I
HEDL 7807-004.9
e LIQUID RADI0 ACTIVE WASTE SOURCE DECONTAMINATION AND CLEANING PROCESSES .
VOLUME 32,920 GALSlYR EXPECTED: 49,300 GALSlYR MAXIMUM ACTIVITY LESS THAN 1.0pCilml; LESS THAN 5 Ci/YR TOTAL DISPOSAL EXISTING FACILITIES ON HANFORD RESERVATION s'I HEDL 7807-004.7 p i
. 4 g
l .
GASE0US RADI0 ACTIVE WASTE SOURCE FISSION PRODUCT AND ACTIVATION GASES VOLUME LESS THAN 25 SCFM
~ 5. 7 Cild
~
ACTIVITY XENON KRYPTON ~17.5 Cild DISPOSAL RADIOACTIVE ARGON PROCESSING SYSTEM HEDL 7807-004.6 h
$) .
GASE0US RELEASES .
. NOBLE GASES 0 RELEASE 0.0032 Cild (FROM LEAKAGE ANALYSIS)
~
e CONCENTRATION 1.5 x 10 LESS THAN MPC AT 1.5 MILES ,
~
- RELEASE 0.088 Cild (FROM RELEASE ANALYSIS)
-5
- CONCENTRATION 1.4x 10 LESS THAN MPC AT 1.5 MILES I .
COMBINED EXPOSURES - WELL BELOW BACKGROUND I 1 HEDL 7807-004.3 h
k
&-l b MAXIMUM EXPECTED GASEOUS EXPOSURES 3
. 10 .. . , , , , , . , , . . . . , . ,. ,
ERDAM 0524 LIMIT FOR INDIVIDUALS - UNRESTRICTED AREAS MEASURED BACKGROUND AT HANFORD FFTF SITE
, 10 - -
d .
>- r E
a.
f 3 10 I - -
EXPOSURE LIMIT FOR NOBLE GASES PER 10 CFR 50, APPENDIX l h
8 w
0
$ 10 _ _ j a ,
5! l n
0
\\ l MAXIMUM EXPECTED EXPOSURE
\ DUE TO TRITIUM LEAKAGE
$ 10
~I - _
U \
\ MAXIMUM EXPECTED EXPOSURE DUE 9< \ TO NOBLE GAS LEAKAGE Q \
, g TOTAL EXPOSURE DUE TO A MAXIMUM
\.
EXPECTED GASEOUS LEAKAGE h 10-2 _
\
g \
COLUMBIA RIVER w s ERDA SITE
\
g s s
WPPSS S'TE BOUNDARY a s HANFORD 12 10~U - s'N -
a_
W
%, %y 10~4 ' ' I' ' ' ' ' ' ' " ' ' ' ' l ' ' Y' ' I O 1 2 3 4 5 DISTANCE FROM FFTF - MILES HEDL 7807-004.34 t'l 1
I
..t I f. ^
~ ~
PROTECTED ACCIDENT ANALYSES
~
. . FSAR CHAPTER 15
~^
REACTI'hlTY ADDITION EVENTS -
O CONTINUOUS CONTROL R0D WITHDRAWAL. ,
O L'OSS OF HYDRAULIC BALANCE TO FUEL ASSEMBLIES -
O RADIAL DISPLACEMENT OF CORE ASSEMBLIES .
~
0 FUEL ASSEMBLY! SODIUM VolDIN'G 0 COLD SODIUM INSERTION O CLOSED LOOP TEST SECTION MELTDOWN -
~'
O SINGLE CONTROL ASSEMBLY MELTDOWN _
8 SINGLE FUEL ASSEMBLY MELTDOWN !
HEDL 7710-202.5 !
l i
e s i !
Q 3- % .
o
. . . t
- g. .' , '
~ ~
PP;0TECTED ACCIDENT ANALYSES -
~ ~
- FSAR CHAPTER 15 -
LOSS OF COOLING EVENTS
.s O PUMP SPEED CONTROLLER FAILURE ,-
~
o LOSS OF POWER. SOURCES . .
~
G PUMP STOPPAGE ,
~
~
0 LOSS OF PUMP. POWER O LOSS OF DHX COOLING -
~
~~
~ '
st a m o 202.4 -
i e
(>
I k.
~
-- - - - - - - - - - - 11
~
s s i .
~
-PROTECTED ACCIDENT MIALYSES -
~
'FSAR t CHAPTER 15 l
.s '
PLANT EVENTS :
' ~
O EARTHQUAKE -
~ .
O TORNADO .
~
~
O RAD WASTE RELEASE
[' ~
O ' FIRES , -
- 0 FUEL HANDLING' ACCIDENTS .-- ,
~
HEDL 7710-202.6 .
O I 1 i .i Q
- I
- 1
, f' i .c
,c'
~
i
~
l
, i
~
[
PREVENTION OF HCDA INITIATION -
s .
I'
- 1. PREVENT EVENTS BEYOND PPS CAPABILITY BY: .
i O POSITIVE DESIGN FEATURES
~
~
e MULTIPLE, INDEPINDENT' PREVENTIVE MEASURES- -
~
o R&D ASSURANCE AGAINST POTENTIAL AUTOCATALYTIC '
~
PROGRESSION .,
- i 1
- 2. PROVIDE RELIABLE PPS PERFORMANCE BY: -
. y e ADEQUATE ASSURED DESIGN CAPABILITY .
i e DESIGN VERIFICATION AND QA H e REDUNDANCY, PARTIAL DIVERSITY l
~
3 t
l 6 f
HEDL 7807-005.89 i
< 8 i
o i
- l
..\
i .
INITIATING MECHANISMS
SUMMARY
,. x NO REALISTIC INITIATION OF HCDA .
STUDIES CARRIED OUT OF PROTECTIVE MARGINS FOR UNEXPECTED AND UNFORESEEN EVENTS -
MECHANISTIC ACCIDENT CLASSES STUDIED
'l l
- 1. LOSS-0F-FLOW (WITHOUT SCRAM) .
%\
hl } HEDL -7807-005.90 7
O
i 1
e THE OVERALL OBJECTIVE 0F THE PLANT NATURAL CIRCULATION TESTING IS TO SAFELY DEMONSTRATE ADEQUATE NATURAL CIRCULATION CAPABILITY IN THE FF1F FOR TWO BOUNDING EVENTS; EACH ASSUMING LOSS OF ELECTRIC POWER (LOEP) TOTHE PUMP PONY MOTORS:
(1) " DESIGN CASE" FROM FULL POWER - RATED CONtBinNS -
AND GOAL FUEL EXPOSURE - SHOW SUFFICIENT CAPABILITY TO MAINTAIN PEAK HOT CHANNEL TEMPERATURE BELOW THE BOILING TEMPERATURE.
~
(2) LOEP AT REFUELING CONDITIONS - LOW DECAY POWER, 400 F ,
SODIUM, COLD DAY - SHOW SUFFICIENT CAPABILITY TO-PREVENT PREMATURE (FIRST HOUR) SECONDARY 1.00P l FREEZING. _
e THESE PROVIDE THE ACCEPTANCE BASIS FOR TESTING NATURAL CIRCULATION CAPABILITY AND THE LIMITING CONDITIONS FOR THESE EVENTS MAY BE TRANSLATED TO ACCEPTANCE TEST LIMITS WITH APPRORIATE ALLOWANCES FOR UNCERTAINTIES AND RESIDUAL MARGINS.
HEDL 7807-005.51
. ~%'
H.
~
TESTSEQUENCE-ACAUTIOUSANDLOGfCALPROGRESSION ,
_TO A PROTOTYPIC LOEP FROM FULL POWER ,
~
~
SUBTESTS: .
1.
SECONDARY-LOOP TEST - STEADY-STATE AND TRANSIENT 2.
TESTS AT LOW FISSION POWER - .
(1) -PRIMARY HEAT ~ TRANSPORT SYSTEM (HTS) STEADY-STATE ~ TEST- .
~
(II) OVERALL PRIMARY / SECONDARY HTS STEADY-STATE TEST
-(III) SHALL POWER STEP TRANSIENT TEST AND SHUTDOWN
~
- 3. 1 L0EP TRANSIENT TESTS FROM POWER OPERATION AT (I) - 5% -
~
(II) 75%
~
(III) 100%
e e
3
~
2- i e.
_ = _ . - _ _ _ . . . .
l
. .~
SUMMARY
OF CONSIDERATIONS l
l
- l.
i e THE TEST SEQUENCE PROVIDES A CAUTIOUS AND LOGICAL PROGRESSION.
~
I e THE SUB-TESTS ARE DESIGNED TO ISOLATE PHENOMENA. -
e STEADY-STATE RESULTS PROVIDE FEEDBACK ON PRE-TEST MODELS.
- PRETEST PREDICTIONS AND ACCEPTANCE CRITERIA FOR ALL TESTS.
O PREREQUISITES SPELLED OUT FOR LATER TESTS.
e EMPHASIS ON MEASUREMENT AND INSTRUMENTATION.
e TESTS CONDUCTED WITHIN NORMAL OPERATING TEMPERATURE RANGE.
e TRANSIENT TESTS PROVIDE SOLID BASE FOR ESTABLISHING DESIGN CASE ACCEPTABILITY - EMPHASIS ON TEST PROTOTYPICALITY.
HEDL 7807-005.12 Q.
a .
p
~
SECONDARY LOOP TEST
SUMMARY
e DHX HEAT LOSSES -
~
f e STEADY-STATE FLOW VS. AT S
- e IS0 LATED MODULES
~
e MISMATCHED MODULES 3
~
e TRANSIENT STARTUP.
P ~
e LOOP THERMAL HEAD DEVELOPMENT H >
COUPLED '
s .
< e DHX HEAT LOSSES . 1 O _
M e LOOP DYNAMICS, INCLUDING HYSTERESIS AND UNBALANCE E BETWEEN MODULES .. !
" e IHX THERMAL CENTER SENSITIVITY PREREQUISITES e DHX HEAT LOSS CHECK e UPDATE PREDICTIONS, IF NECESSARY (MODEL LIMITS) e DETERMINE INITIAL TEMPERATURE FOR TRANSIENT ,i N -
_s x HEDL 7807-004.77
.~
. SECONDARY NATURAL CIRCULATION TEST SEQUENCE AND ACCEPTANCE LIMITS I i i i I i i l i I
~
~
REFUELING CASE .
ACCEPTANCE LIMIT CALCULATED PER MINIMUM g / STARTUP TEMPERATURE
/ OF 2400F '
3: / ( A P @ + 36.5%)
l
$ 2.0 y -
> 4 PECTED HEAT -
E
@ /
/ LOSSES WITH PREHEATERS f
'/ .
o / SHUT DOWNe full POWER S
- / '
/ ACCEPTANCE LIMIT f ESTIMATED PER MAXIMUM HOT
/ / CHANNEL TEMPERATURE OF17050F LO
/ /
/ FOR 3 LOOP LOSS OF POWER -
7 TRANSIENT j
' ( A P g + 186%)
7 j /
//
//
Q l I I I I t i 1 I t N
g 0 20 40 60 80 100 - 120 140 -160 180 200 V -
SECONDARY LOOP oT (OF)
~
b \
HEDL 7806-058.32
~
& L- l rs SECONDARY LOOP TRANSIENT RESPONSE FOR "STARTUP FROM REFUELING" NOMINAL PREDICTION (NOMINAL A P's)
--- ACCEPTANCE LIMIT TRANSIENT ( A P's + 36.5%)
DHX TUBE BUNDLE OUTLET TEMPERATURE 500 , i a"- 450 . ,
u ja 400 _ _
h ,-T_____________._
g 350 -
,/
v ,8 A /
/ t i 30Q LO 2.0 TIME (HOURS)
SECONDARY MASS FLOW 50.0 , i g37.5 -
,/ % ~~~ __.__________.
3 /
' 25.0 -
l\' /
8 l /
. d 12.5 - / ~ _
0.0 ' i 1.0 2.O TIME (HOURS)
HEDt. 7806-058.31 O .
~
- PRIMARY LOOP " STEADY-STATE" TESTS
SUMMARY
e CAREFUL TRANSITION TO NATURAL CIRCULATION e LOOP & FOTA AT VS. REACTOR POWER LEVEL - PRIMARY LOOP -
T e LOOP & FOTA AT VS. REACTOR POWER LEVEL - PRIMARY & SECONDARY E LOOP -
S e POWER STEP ,
5 e IHX SENSITIVITY - COMPLETE FOR SECONDARY & PRIMARY e MISMATCHED LOOPS -
.
E e LOOP AND REACTOR AP VS. FLOW ,
N 4 e REACTOR " STEADY-STATE" FLOW DISTRIBUT10tj M e CHANNEL TRANSIENT DYNAMICS N ,
e IHX TEMPERATURE SENSITIVITY .
e SECONDARY LOOP TESTS AND REVISE ANALYSES, IF NECESSARY
{ ,
E e COASTDOWN MEASUREMENTS I < e CONTROL STABILITY .
- POWER CAllBRATION (THERMAU S
, o MFM CAllBRATION .
I 4 S HEDL 7807-005.19 '7
~
PRIMARY HTS AND OVERALL PRIMARY / SECONDARY HTS TEST MEASUREMENT AND INSTRUMENTATION -
n e MAGNETIC FLOWMETER FLOW (CAllBRATED WITH PULSED NEUTRON ACTIVATION FLOWMETER) - ~ 5% .
e EDDY CURRENT FLOWMETERS ABOVE OPEN TEST ASSEMBLIES (OTA)
AND DRIVER FUEL ~10% GOAL e DIRECT CONTACT IN CORE THERMOCOUPLES IN OTA'-S ~2% ABSOLUTE (TWO FUELED AND ONE ABSORBER ASSEMBLIES)
~
- REACTOR POWER (SECONDARY LINEAR) - ~7.4%
e PRIMARY LOOP RTD'S - 5% ON T e M1SCELLANEOUS THERMOCOUPLES INION I TEMPERATURE AND LIQUID LEVEL MdNITOR PROXIMITY TEST PLUG ,
INSTRUMENT TREE .
VIBRATION OPEN TEST ASSEMBLY ,
REACTOR VESSEL IHX PIPING Rt HEDL 7807-005.9 - 7
w i
f9 POWER AND KEY PLANT FL0?/ RATES .' '
FROM 1007. AFTER 25 HRS OPERATION
- 40 , , , , , , , , , ,
~
30 -
b '
N20 l
REACTOR FLOW ,
10 -
POST SCRAM POWER 0 ' ' ' ' ' ' ' ' ' ' I 0 1 2 3 4 5 6 7 8 9 10 11 TIME (MINUTES)
KEY PLANT TEMPERATURES FROM 100% AFTER 25 HRS OPERATION 1000 i i i i i i i i i REACTOR VESSEL OUTLET
/
. 900 -
i C
c 800 -
AVERAGE CORE -
3 OUTLET e
h700 ' -
REACTOR VESSEL INLET -
/
600 ' ' ' ' ' ' ' ' '- '
0 1 2 3 4 5 6 7 8 9 10 11 TIME (MINUTES)
HEDL 7806-058.21 4
t
~
EXTRAPOLATION OF TEST POINT TO THE ' .
DESIGN CASE TRANSIENT -
~
800 i ; i iiig ,' , i i iii gi i
. p y KEY CASES
- ESTIMATED HOT 5
m 1 POWER o
= 5% TEST CHANNEL BOILING 2 POWER g - 75% TEST
}400 ,
5 3 POWER g - 100% TEST
, h -
4 DESIGN CASE m 200
$- FLOW g - 75%,1 HR. AT POWER f',s
% /
x 100 -
/ POWER / FLOW g g
- 1.0 _
6
~ s' .
b I ,s
@ 50 ' 'I ' ' ' I ' I ' ' ' I g 0.04 E0.06(' 0.'08 0.1 0. 2 0. 4 0.60.81.0 2. 0 4.0
- POWER. AT TIME OF PEAK SUBASSEMBLY TEMPERATURE RISE [%) g HEDL 7807-005.22 . . it C/
SUMMARY
.THE PRINCIPAL OBJECTIVES & THE. TEST FROGRAM WILL BE NEAR PROTOTYPIC DEMONSTRATION & ACCEPTABLE PEAK CORE TEMPERATURES FOR THE DESIGN
~
LOSS & POWER CASE AND TRANSITION T0 NATURAL- CIRCULATION FROM.RE-FUELING WITHOUT DHX FREEZING. ,
TEST PLANS.SPECIFY A CAUTIOUS AND LOGICAL- PROGRESSION TO THE KEY TRANSIENT TESTS.
TEST PREPARATIONS HAVE INCLUDED:
- 1. EMPHASIS ON PHENOMENA AND INSTRUMENTATION
- 11. PRE-TEST M0DELING, PREDICTION, AND SENSITIVITY STUDIES
'i1. LIMITS FOR PLANT ACCEPTANCE, TEST CONTINUATION. AND MODEL ADEQUACY IV. INTERCHANGE WITH THE LMFBR COMMUNITY MODEL CAllBRATION WILL BE SUFFICIENT TO DETERMINE THE ADEQUACY & .
DESIGN UNCERTAINTY ALLOWANCES AND TO CONFIRM ACCEPTABLE POST-TEST MARGINS FOR THE TWO LIMITING CASES.
HEDL 7807-005.20
H. Hg
( C d ..t l 1
- 1 T-s .
l
- v. .
'1 d
NRC REVIEW i
l I. TYPE OF REVIEW REQUESTED 0
USED "0LD" AEC INTERAGENCY AGREEt;1ENT Ad '%
MANUAL CHAPTER 0540 AS BASIS,
- NRC TO PROVIDE ADVICE ON ADEQUACY OF THE l FFTF DESIGN TO ENSURE SAFE OPERATION OF THE PLANT.
- NRC AUGUST 13., 1976 AND ERDA AUGUST 20., 1976 LETTERS DEFINED SCOPE OF REVIEW, i
l
. )
. -.: ;- - 3 7 ,.--- . : - , . =- - - - - - - - ; - - - - -- - f i
p o. ) -
y
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key'#0fNTS"EN'5fd55"O#"55VIW f.
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- FULL TECHNICAL REVIEW '0F THE FFTF DESIGN COMPARABLE [
TO A LICE $ SED PLANT REVIEW IN TECHNICAL DEPTH 1 i!
- LIMITED SITE RELATED REVIEW - SUFFICIENTLY REVIEWED ,, !!
AT CP STAGE.
]J t1
- IN DEPTH REVIEW BY NRC NOT REQUESTED BY ERDA BUT ll 3
l COMMENTS WERE SOLICITED IN THE FOLLOWING AREAS i: I!
!1
?l OPERATIONS l' STARTUP TESTING ,i f I QUALITY ASSURANCE (i TECH SPECS
{'
i i
1
- FSAR RECEIVED 4-2-76; ACCEPTED FOR REVIEW BY NRC ON j 8-13 TW0' YEAR REVIEW SCHEDULE PROPOSED -
ii i
4 4
4 .
e 4
p.n.~ ' ;.~;, ~,
-
- m . ..-.;~:------___-----
V
' i: '
r _ %- .k.
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' 7-3 -
Ii v- . 1 CHANGES FROM INITIAL SCOPE OF REVIEW l
l 8 SAFEGUARDS AND SECURITY TO BE PERFORMED BY ERDA - NRC NOT INVOLVED.
- CLOSEDLOOPDESIGNTOBEAdDRESSEDINFUTURI ,
' i' '
, SUPPLEMENT.
- CONVENTIONAL PLANT FIRE PROTECTION TO BE ADDRESSED ,
IN A FUTURE SUPPLEMENT - FALL OF 78
- TECH SPECS - TO BE ADDRESSED IN FUTURE SUPPLEMENT.
6 b
L 3
G c.*
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7- >' .
v .
BASIS FOR REV! $
[
- EXISTING RULES AND REGULATIONS.
/
e STANDARD REVIEW PLAN
. i
.f* l
- REGULATORY GUIDES
- THE SAFETY OBJECTIVES OF THE GDC ARE USUALLY l APPLICABLE TO OTHER REACTOR TYPES -
i
- WHERE FFTF DESIGN DEVIATED FROM GDC THE SAFETY ,
SIGNIFICANCE OF THE DIFFERENCES WERE EVALUATED l
i
. 4 9
. . . . .. . . . . . . . .._ - z___., _.
p., __. .. . ~ . . _ . . . . _ . - . - - . .
,p .,,.
) .
1
~
. 7f.
. i COMPARIS0N OF REVIEW BETWEEN'LWRs AND FFTF w. ,
, ./
I
- THE REVIEW 0F FFTF WAS PERFORMED USING THE SAME STAFF THAT REVIEWS, LWRs MANY OF WHOM HAVE FOLLOWED BOTH FFTF AND CRBR i
.- .i '
- IN ADDITION IN SPECI ALIZED AREAS- THE ARB WlTH'in. '^...s.
- 15. . x .
.5
- CONSULTANTS. REVIEWED: !
CDA'S l PAHR l
SODIUM FIRES S0DIUM-CONCRETE INTERACTION .!
RADIOLOGICAL ANALYSIS 1 1
REACTOR SYSTEMS FUEL HANDLING I
.:l
- THE QUESTIONS,- RESPONSES AND NUMBER OF MEETINGS PARALLEL THAT OF AN-LWR REVIEW -
4
[
' .f;.
.Q. . . . .
.) - .
l
- y. . l DEGREE TO WHICH REVIEW AND FACILITY CONFORM TO POWER -!
PLANT REQUIREMENTS I. REVIEW ,
. I i i
- THE COMPARABILITY OBJECTIVE FOR REVIEW 0F THE i, ,
1 FFTFDESIGNWASACHlEVEDINAPRACTi'CALSENSE -y
- DIRECT COMPARISONS ARE DIFFICULT DUE TO MOST .
l ISSUES ON LWRs BEING RESOLVED AS'A GENERIC ITEM RATHER THAN PLANT SPECIFIC i
, 1
- THE FFTF FSAR UTILIZED AN UNUSUALLY LARGE i NUMBER OF REFERENCES WHICH WERE NOT HANDLED i AS GENERIC STUDIES BUT AS PLANT SPECIFIC PROBLEMS.
9 i
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, .. __ . . . . . . u _ g 7
.. . ; ; . .;_ _ _ , . ._ ,2. a ,,,,,u m. :. ==,-_a., .u . - . . .
. 1 l
- , .e ,
2- . .
@7.
i DEGREE TO WHICH REVIEW AND FACILITY CONFORM TO POWER y .
- ],.
PLANT REQUIREME"TS II. FACILITY
- TORNADO - MISSILE DESIGN
III. OVERALL ,
IT IS THE STAFF'S VIEW THAT AREAS IMPORTANT TO SAFETY HAVE BEEN REVIEWED AND ARE BEING ADDRESSED IN THE SER. WITH RESPECT TO DESIGN REQUIREMENTS, THE STAFF HAS USED THE GUIDANCE PROVIDED BY THE GDC, STD REVIEW PLAN AND REG GUIDES. WHERE THE FFTF DEVIATED THE STAFF HAS EVALUATED THE FUNCTIONAL SIGNIFICANCE. WHERE THE APPLIED TECHNOLOGY IS THE STATE-OF-THE-ART A REASONABLY CONSERVATIVE POSITION HAS BEEN RECOMMENDED. .
7 9
6
- ..f..._.__...._._._m..
}
~
PRBENM5555 ,
y .
1 AREAS WHERE THE STAFF HAS CONCERNS WHICH ARE NOT COMPLETELY RESOLVED ,
l'. NATURAL GONVECTION VERIFICATION (A) PREPREDICTION OF RESULTS (s) IANUS/FLODISC CODE VERIFICATION ,,
(c) CONFIRMATION OF SAFETY ANALY, SIS .
'i
- 2. - PIPING INTEGRITY (A) AEROSOL LEAK DETECTION-SEISMIC DESIGN AND 1E POWER (s) PRESERVICE AND INSERVICE INSPECTION OF HOT CROSSOVER PIPING WELDS (c) MATERIAL SURVEILLANCE PROGRAM
}
- 3. CELL LINER - ESF 0 SUBMIT A PLAN TO SHOW LINERS MAY BE RELIED UPON OVER LIFE OF PLANT .
- 4. LOOSE PARTS MONITORING ,
~
ALL LWRs-(SECTION 4'4 0F SRP) HAVE THEM',' WE
.RECOMM EN D THT A FFTF INCLUDE ONE IN ITS PLANNING AND INCORPORATE IT AS EXPEDITIOUSLY AS PRACTICAL.
. 9 4
.,v.
- 4. . j < == m , ' ~
4 am ..+.w~s-e.., , u. =---r,-.- - -- - .-
ilk ' *s , *
. .-TJ-7 2- -
,- v .
- 5. SEISMIC RESTRAINTS l FURTHER CLARIFICATION NEEDED ON SYSTEM SAFETY CLASSIFICATI0N'; THE C'LASSIFICATION CHANGE OCCURS
~
BEYONDF[RSTSEISMICRESTRAINTONTHECIS. THIS SHOULD BE RESOLVED AND DOCUMENTED'. . .
6 '. TORNADO CONDITIONS . T RECOMMEND PLANT SHUTDOWN UNDER IMPENDING TORNADO CONDITIONS AS A REASONABLE AND PRUDENT THING TO
' ~
D0 RATHER THAN G0 INTO AN UNPLANNED SHUTDOWN.,
7 '. MAX FUEL CHANNEL EXIT TEMPERATURE LIMIT TO 16700F RATHER THAN 17050F (A) UNCERTAINTIES IN HCF (a) UNCERTAINTY ON NC FLOW -
- 8. CONTROLS OUTSIDE OF CONTROL ROOM -
NEED CLARIFICATION OF SAFETY RELATED MONITORING FUNCTION TO BE PERFORMED OUTSIDE OF CONTROL ROOM i
~
- 9. REACTOR POST-ACCIDENT MONITORING (i) CONTROL R0D POSITION (a) FLUX MEASUREMENTS .
- -----~=-----m. - - -- .- . I ,
-. m =-- -:..
~
1 ..
Lb
. v .
- 10. CIS VALVE POSITION INDICATORS (A) DISPLAY IN CONTROL ROOM (s) CIRCUITS AND INDICATOR TRANSMITIEF. T0 BE MISSILE PROTECTED AND ON 1E POWER 11'. - CONTROL ROOM ISOLATION-MISSILE PROTECT INTAKES, INSTRUMENTATION AND CIRCUITS OUTSIDE, ACCIDENTAL RELEASES AFFECTING' CONTROL ROOM RABITABILITY (A)
RADIATION - IODINE (s) NA02 - DHX FIRE (c) CHLORINE - TRUCK SPILL -
- 12. CONTAINMENT MARGIN DEVELOP. PROGRAM (E.a.', CONTR01. VENT RATE) FOR CONTROLLING RADIOLOGICAL RELEASE FROM LOW PROBABILITY ACCIDENT
\
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DE#$R"F0((dWfN50P5N'I'S50s5ONTI[~#((~1575 i e CL SYSTEM (PPS) :
i I
FIRE HAZARD ANALYSIS B
6
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i
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- 9. - _, ,
Or _
~
~ UNIVI'.ItSITY OF CAI.lFO!!NI A
- l.OS Al.A.\lUS SCIE.VIIt'l(* l.AllOltATORY ICONTH A(~r W.74n.<.E.st' %)
P.o.nox 503 "
7 ., , g.. . l ' * * * , . LOS A1.AMOS. .NEW MExIr0 87545
- ':.- .. .s v . . .
f% RI'Pl.Y ms a To: TD-7 + ^> . . . . l si su. .sTor: 231 , '
19 July 1978 Dr. Thomas McCreless l Advisory Conmittee on Reactor Safeguards ,
l Nuclear Regulatory Commission l Washington, DC 20555 . l 1
Dear Dr. McCreless:
This note is concerned with the FFTF r'eview and the information given to us today (July 12). -
Dr. Mark and I discussed briefly. the point he raised about the maximum possible reactivity should the core be voided in a selective manner. Mr. Simpson denied the possibility of such selective voiding, but should it be done, by special experiment for example, the maximum excess reactivity would be about'three dollars, or two (2$ dollars over promp If the delayed neutron fra 3x10gcritical.
and the neutron lifetime, f., is 4x10 gtion, sec, .then 8, isthe about 0.003 =
maximum ret.iprocal period a = 1/T is given by
, , A k prompt _ $ prompt x 8 1 1
, 2.0x3x10-3 ,6 x 10 4 4x10~7
= 1.5 x 10 /second -
Given this value of a, the reactor power would change as:
dE dE - at Power = g = (' g g e-Du' ring the meeting we estimated this "a" to be a bit more than 4
2x10 /sec. This was incorrect. In any case:
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s An Affirmative Action /[qualOpportunity [mployer .
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/ i Dr. Thomas McCreless
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Dr. Mark stated that this period would not cause a nuclear explosion. Quenching motions would occur too fast for pressures -
to build up to a significant value.
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'2). Mr. Simpson stated clearly that such preferential voiding of the core could not occur and any voi. ding would occur slowly, act instantaneously; '
Thus, it is apparent that the maximum reciprocal period, given sodium '
voiding will be very small (very long period) and non-damaging in the e
sense of an explosion, Given Dr. Kerr's and Dr. Mark's concurrence, I request this clari- l fication be incorporated into the meeting]ninutes.
My questions re the capability of the decay heat removal system I wished to were formulated as below and given to the HEDL people.
determine the fraction of full power that the system could remove if:
a) Only convective motion is allowed, ,
b) only pony motor pumping is allowed, .
c) and, both are allowed in the calculation of sodium motion.
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. I will allow the assumptions of sodium boiling, fuel melting, clad .
cracking and leaking, fission product (and even fuel) leaking into the coolant, and DNB if necessary. .I ask only the requirement of cooling of fuel pins and otherwise a best-estimate calculation.
Any experiment in thisarea, of course, would be modest and designed I to be non-damaging. I believe that the HEDL personnel should have access to the final experiments in the Downneay Fast Reactor I have seen nowhich reportexperiments of these ,
were closely related to my question.
experiments.
Another question about which time did not permit discussion is the following; it is safety related. .
I assume full power (or some reasonable fraction there a year.
site power, cmergency power for pony motors, and, finally an interruption of the flow path for sodium such that the core-(and piping in the immediate' neighborhood)isisolated. The questions relate to the loss of sodium:
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19 July 1978 "f
Dr. ThomasiMcCreless
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- 1) Taking full credit for heat . conduction, and condensation of sodium vapor, how long before the sodium no longer covers the ,
core? o(The sodium must go somehwere, so a mechanism must be 3
assumed).
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'7) The sequense of _ events subsequent to the liquid sodium surface j reaching the top of the core is.of interest. For examplei l i
a) liow long can sodium vapor cooling maintain physical integrity I of the top of the fuel pins? i b)- If fuel and fuel-pin claddi,ng commence .to disintegrate above the level of boiling sodium, what is the best estimate for i the destination of the fuel and in.what physical form? i l
- 3) llow long_ before the sodium level reaches the bottom of the core? '
(Assuming full heat conductivity and condensation effects). !
, ~4).. At 'this time, what is the most probable location and physical !
condi. tion for the major portions of the core.
- 5) If the fuel is assumed to' be under sodium, how long to evaporate IEe remainder of the sodium? (Assuming, of course, full con- 4
-ductivityandcondensationeffects). .
- 6) Given this degraded condition, can sodium be added to the system 3 to alleviate the problem?
Given concurrence of the subcommittee chairman, I suggest that these i questions be given to the HEDL to be answered during the next meeting.
Sincerely, f
[ Y W.-R. Stratton l
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