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Reactor Operations informal monthly report September 1994

Description: This paper presents operations at the MRR and HFBR reactors at Brookhaven National Laboratory for September 1994. Reactor run-times, instrumentation, mechanical maintenance, occurrence reports and safety information are listed. Irradiation summaries are included.
Date: September 1, 1994
Creator: Junker, L.
Partner: UNT Libraries Government Documents Department

Los Alamos National Laboratory Omega West Reactor restart

Description: This report is a critical evaluation of the effort for the restart of the Omega West reactor. It is divided into the following areas: progress made; difficulties in restart effort; current needs; and suggested detailed steps for improvement. A brief discussion is given for each area of study.
Date: August 27, 1993
Partner: UNT Libraries Government Documents Department

Preliminary Phenomena Identification and Ranking Tables (PIRT) for SBWR start-up stability

Description: Phenomena Identification and Ranking Tables (PIRT) have been developed for start-up transient for SBWP. The information used for PIRT came from RAMONA-4B and TRACG analyses of the transient and from related small scale tests. The transient was divided into four distinct phases, namely, Subcooled Core Heat-up, Subcooled Chimney, Saturated Chimney and Power Ascension. The assessment criterion selected was Minimum Critical Power Ratio. The SBWR system was divided into ten components. A total of 33 distinct phenomena among the components were identified. The Phase I has 28 ranked phenomena with 17 low, 6 medium and 5 high ranking. The Phase II has 39 ranked phenomena with 18 low, 13 median and 8 high ranking. The Phase III has 47 ranked phenomena with 22 low, 10 medium and 15 high ranking. The Phase IV has 46 ranked phenomena with 16 low, 12 medium and 18 high ranking. 12 refs., 22 figs., 21 tabs.
Date: March 1997
Creator: Rohatgi, U. S.; Cheng, H. S.; Khan, H. J. & Wulff, K. W.
Partner: UNT Libraries Government Documents Department

Simulation of SBWR startup transient and stability

Description: The Simplified Boiling Water Reactor (SBWR) designed by General Electric is a natural circulation reactor with enhanced safety features for potential accidents. It has a strong coupling between power and flow in the reactor core, hence the neutronic coupling with thermal-hydraulics is specially important. The potential geysering instability during the early part of a SBWR startup at low flow, low power and low pressure is of particular concern. The RAMONA-4B computer code developed at Brookhaven National Laboratory (BNL) for the SBWR has been used to simulate a SBWR startup transient and evaluate its stability, using a simplified four-channel representation of the reactor core for the thermal-hydraulics. This transient was run for 20,000 sec (5.56 hrs) in order to cover the essential aspect of the SBWR startup. The simulation showed that the SBWR startup was a very challenging event to analyze as it required accurate modeling of the thermal-hydraulics at low pressures. This analysis did not show any geysering instability during the startup, following the startup procedure as proposed by GE.
Date: June 1, 1998
Creator: Cheng, H.S.; Khan, H.J. & Rohatgi, U.S.
Partner: UNT Libraries Government Documents Department

Determination of maximum reactor power level consistent with the requirement that flow reversal occurs without fuel damage

Description: The High Flux Beam Reactor (HFBR) operated by Brookhaven National Laboratory (BNL) employs forced downflow for heat removal during normal operation. In the event of total loss of forced flow, the reactor will shutdown and the flow reversal valves open. When the downward core flow becomes sufficiently small then the opposing thermal buoyancy induces flow reversal leading to decay heat removal by natural convection. There is some uncertainty as to whether the natural circulation is adequate for decay heat removal after 60 MW operation. BNL- staff carried out a series of calculations to establish the adequacy of flow reversal to remove decay heat. Their calculations are based on a natural convective CHF model. The primary purpose of the present calculations is to review the accuracy and applicability of Fauske`s CHF model for the HFBR, and the assumptions and methodology employed by BNL-staff to determine the heat removal limit in the HFBR during a flow reversal and natural convection situation.
Date: April 19, 1990
Creator: Rao, D.V.; Darby, J.L.; Ross, S.B. & Clark, R.A.
Partner: UNT Libraries Government Documents Department

Special disarmament study

Description: This study investigates the cost of shutting down and later starting up the Hanford production complex. Also included are comments on two aspects, other than cost, which are consider to be of great importance. One is the impact of a complete shutdown on the Tri-Cities community. The other is the effect on Battelle`s Pacific Northwest Laboratory. This study encompasses six special shutdown-startup cases. One set of three cases assumes a shutdown of three years; the other set assumes the reactors have been down for a period of five years (from shutdown of the last reactor to startup of the first conventional reactor in the series). within each of the above mentioned sets there are three cases assuming the return of five, six, or seven reactors to weapons production. The sixth and seventh reactors are D and DR respectively. N Reactor is assumed transferred to WPPSS instead of shutdown, and consequently is available for recapture earlier than the conventional reactors. The shutdown schedule is shown in Table IA. The dates, except for the shut down of D Reactor, were picked arbitrarily and can be moved ahead so long as the intervals are kept the same. There is one exception to this generalization. The sludge program at the B Plant will end in 1973 and then waste management will be on a current basis. Thus, the operation of B Plant depends on the shutdown timing.
Date: April 26, 1967
Creator: Nilson, R.
Partner: UNT Libraries Government Documents Department

HFBR restart activity A2.6: Review of FSAR and 60 MW addendum to assure consistency of operation at 40 MW

Description: The purpose of this task (HFBR Restart Activity A2.6) is to perform a review of the design basis accident (DBA) analyses sections of the 1964 HFBR-Final Safety Analysis Report; Volumes I and II, and the 1982 Addendum to the HFBR-FSAR for 60 MW operation to assure that operation at 40 MW will be consistent with these analyses. Additional documents utilized in the review included the Level 1 PRA for HFBR, HFBR-PDMs and HFBR-OPMs. The review indicates that the 1964 FSAR-DBA analysis in incomplete in the sense that it did not analyze some of the important initiators for 1-loop operation that include: Accidental throttling of primary flow control valves; seizure of primary pump; loss of secondary pump; accidental throttling of secondary flow control valves; rupture of secondary piping. The first three initiators were later studied in the 1982 addendum. The other two initiators have not been examined to-date for 1-loop operation. It is recommended that the impact of these initiators be assessed prior to the restart, if 1-loop operation is chosen for the restart. The review demonstrated that at 40 MW operation there are only a few accident initiators that will culminate in core damage (fuel melting and /or cladding failure) regardless of the availability of mitigating systems. For 1-loop Operation these accidents include: Fuel channel blockage, primary pump seizure, and large-large LOCA (a LOCA with effective break diameter > 2.81 in. is referred to as a large-large LOCA in this document as well as in PRA). Although all these accidents listed above could lead to core damage for 1-loop operation as well, the probability is expected be very low.
Date: February 26, 1990
Creator: Rao, D.V.; Ross, S.B.; Darby, J.L. & Clark, R.A.
Partner: UNT Libraries Government Documents Department

Reactor statistics, April, 1961--April 1962

Description: The primary effort to date in connection with this study has been directed toward obtaining source data which indicates (1) the functions performed during reactor outages and the distribution of time required to accomplish these corrective functions, (2) the groups of crafts associated with each of the recovery functions performed, and (3) the radiation exposures experienced during these activities. The first phase of preliminary analysis has been based on the ``time accountability`` report data originated by the various reactor analysts. The attached computer tabulation is one of the analyses performed considering the time and date a reactor was shut down, the ``cause`` for which it went down and the time and date the reactor was considered back on-line. The report summarizes these accountability data into the following summaries in the order presented below: (1) Total hours down per reactor per cause. (April, 1961 to April, 1962) (2) Number of records indicating experience of outages per reactor per cause. (3) The average and standard deviation; same relationship. (4) Outage summary; total hours down, percentage contribution to the department total outage, and time operating efficiency. (5) Department summary (self explanatory). (6) through (21). Interval between like outages by cause. These reports illustrate which reactor was hit by the various problems during the year, how many times it was involved, and the number of hours lapsed between each like shutdown cause. Department summary of outage interval report; (This again is a preliminary analysis to determine whether there is sufficient data to make sound statistical conclusions regarding outage time-cause relationships).
Date: May 11, 1962
Creator: Burke, R. C.
Partner: UNT Libraries Government Documents Department

Interim report one to Production Test-IP-549-A, half-plant low alum feed water treatment at F Reactor

Description: A half-plant low alum water treatment test began at F Reactor on January 16, 1963 at startup from the scheduled January 3 tube replacement outage. The test, which was prompted by results obtained from a statistical analysis of fuel ledge corrosion attack, will demonstrate whether or not high alum feed is responsible for increasing the frequency of ledge corrosion attack on fuel element surfaces. The effect will be evaluated by comparing visual examination results obtained from normal production fuel irradiated in two different alum treated process waters. This report discusses the results obtained from twenty fuel charges, ten from each side of F Reactor, which were discharged prior to the reduction in alum feed to establish the pre-test corrosion environment.
Date: February 11, 1963
Creator: Clinton, M. A. & Geier, R. G.
Partner: UNT Libraries Government Documents Department

Reactivity and efficiency trends vs operating trends for B, D, DR, and F Reactors, 1955--1959

Description: Changes in operation and corresponding changes in the reactivity status of Hanford reactors are the result of a continuing effort to improve operating efficiency. Trends data related to these changes in operation and reactivity have been published previously for the periods from 1950 through 1958. The purpose of this report is to include trends data for 1959. Bar graphs in the first part of the report show yearly averages of selected data, and tables in the last part of the report show maximum, average, and minimum values. This document presents trends data for B, D, DR, and F reactors while a second document, HW-64932, presents trends data for C, H, KE, and KW reactors. Data included in past years which have not been included in this report are trends in pile power level at shutdown omitted due to a security status change regarding power levels, and number of temporary poison columns per startup omitted due to virtual elimination of temporary poison startups at B, D, DR, and F Reactors; added were potential non-equilibrium gains and potential equilibrium gains. Notice that all reactivity values are listed in the unit per cent excess k.
Date: April 29, 1960
Creator: Clark, D. E.
Partner: UNT Libraries Government Documents Department


Description: The purpose of this report is to provide data for preparation of a NEPA Environmental Impact Statement in support the U. S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives is to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu) -239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept has been proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR is proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu will be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) is being considered for fabrication of WG Pu fuel for the ABR. This report is provided in response to ‘Data Call’ for the construction of startup fuel fabrication facility. It is anticipated that the facility will provide the startup fuel for 10-15 years and will take to 3 to 5 years to construct.
Date: April 1, 2007
Creator: Khericha, S. T.
Partner: UNT Libraries Government Documents Department