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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

Douglas United Nuclear, Inc. and the National Defense Program

Description: This report presents the results of an updated study to determine the preparation, schedules and cost for restart of KE Reactor and the subsequent plutonium production and incremental conversion cost of sustained plant operation. The restart of Hanford`s KE Reactor features short startup time, large capacity, low cost, and high purity product. The study shows that KE Reactor can be restarted and achieve full power 7 months after authorization. Prestartup costs are $4,416,000 for labor and material with only $375,000 for facility repair and replacement. Over 375 kg of plutonium are produced in the first years and equilibrium operation is achieved near the end of the second year.
Date: April 21, 1972
Creator: Harrington, C. D.
Partner: UNT Libraries Government Documents Department

Water treatment during C Pile start up

Description: This memorandum details the recommendation of the Technical Section that the C Pile be started up using alum treated water to which has been added sodium dichromate at a concentration of 2 ppm., and that the pile be purged at monthly intervals despite the expected absence of film formation. During the initial months of operation the necessity of adding sodium dichromate and of purging the pile should be evaluated with the hope that these factors can be eliminated from the process. The reasons for this recommendation, and proposals for water treatment in the current operating piles are discussed.
Date: September 22, 1952
Creator: Woods, W. K.
Partner: UNT Libraries Government Documents Department

Start-up of the 100-B pile following 1569-B tube failure

Description: Tube 1569-B ruptured on September 22, 1948 and was replaced. Between the time of failure and final isolation of this tube, considerable water escaped into the graphite packing. As a result, the unit start-up and subsequent operation was quite different from that usually followed. This report has been prepared to summarize the observations and activities related to this operation.
Date: November 8, 1948
Creator: Carlton, G. B.
Partner: UNT Libraries Government Documents Department

History of the reactivation of 100-B pile

Description: This report summarizes the preparations made for reactivation of the 100-B pile and the operational activities associated with the reactivation and subsequent operation up to the time of reaching desired power level. The period covered is from approximately June 1, 1948 until July 16, 1948.
Date: October 29, 1948
Creator: Filip, E. J.
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

Status of IFR fuel cycle demonstration

Description: The next major step in Argonne`s Integral Fast Reactor (IFR) Program is demonstration of the pyroprocess fuel cycle, in conjunction with continued operation of EBR-II. The Fuel Cycle Facility (FCF) is being readied for this mission. This paper will address the status of facility systems and process equipment, the initial startup experience, and plans for the demonstration program.
Date: September 1, 1993
Creator: Lineberry, M. J.; Phipps, R. D. & McFarlane, H. F.
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

HFBR: Review of the technical specifications against the FSAR

Description: The purpose of this review is to determine the adequacy of the High Flux Beam Reactor (HFBR) Technical Specifications for 40 MW operation by comparison with the HFBR Final Safety Analysis Report, particularly the accident analyses chapter. Specifically, the Technical Specifications were compared against the Design Basis Accident (DBA) Analyses presented in the Addendum to the HFBR FSAR for 60 MW Operation. The 60 MW DBA analyses was used since it is more current and complete than the analyses presented in the original FSAR which is considered obsolete. A listing of the required systems and equipment was made for each of the accidents analyzed. Additionally, the Technical Specification instrument setpoints were compared to the DBA analyses parametric values. Also included in this review was a comparison of the Technical Specification Bases against the FSAR and the identification of any differences. The HFBR Operations Procedures Manual (OPM) was also reviewed for any inconsistencies between the FSAR or the Technical Specifications. Upon completion of this review it was determined that the Technical Specifications are well written and the items commented on should not delay the low power restart (40 MW). Additionally, the OPM is also well written and does not require further modification before restart.
Date: January 25, 1990
Creator: Rao, D.V.; Ross, S.B.; Claiborne, E.R.; Darby, J.L. & Clark, R.A.
Partner: UNT Libraries Government Documents Department

Modeling of two-phase flow instabilities during startup transients utilizing RAMONA-4B methodology

Description: RAMONA-4B code is currently under development for simulating thermal hydraulic instabilities that can occur in Boiling Water Reactors (BWRs) and the Simplified Boiling Water Reactor (SBWR). As one of the missions of RAMONA-4B is to simulate SBWR startup transients, where geysering or condensation-induced instability may be encountered, the code needs to be assessed for this application. This paper outlines the results of the assessments of the current version of RAMONA-4B and the modifications necessary for simulating the geysering or condensation-induced instability. The test selected for assessment are the geysering tests performed by Prof Aritomi (1993).
Date: October 1, 1996
Creator: Paniagua, J.; Rohatgi, U.S. & Prasad, V.
Partner: UNT Libraries Government Documents Department

Probability and consequences of a rapid boron dilution sequence in a PWR

Description: The reactor restart scenario is one of several beyond-design-basis events in a pressurized water reactor (PWR) which can lead to rapid boron dilution in the core. This in turn can lead to a power excursion and the potential for fuel damage. A probabilistic analysis had been done for this event for a European PWR. The estimated core damage frequency was found to be high partially because of a high frequency for a LOOP and assumptions regarding operator actions. As a result, a program of analysis and experiment was initiated and corrective actions were taken. A system was installed so that the suction of the charging pumps would switch to the highly borated refueling water storage tank (RWST) when there was a trip of the RCPs. This was felt to reduce the estimated core damage frequency to an acceptable level. In the US, this original study prompted the Nuclear Regulatory Commission to issue an information notice to follow work being done in this area and to initiate studies such as the work at BNL reported herein. In order to see if the core damage frequency might be as high in US plants, a probabilistic assessment of this scenario was done for three plants. Two important conservative assumptions in this analysis were that (1) the mixing of the injectant was insignificant and (2) fuel damage occurs when the slug passes through the core. In order to study the first assumption, analysis was carried out for two of the plants using a mixing model. The second assumption was studied by calculating the neutronic response of the core to a slug of deborated water for one of the plants. All three types of analyses are summarized below. More information is available in the original report.
Date: November 1, 1995
Creator: Diamond, D.J.; Kohut, P.; Nourbakhsh, H.; Valtonen, K. & Secker, P.
Partner: UNT Libraries Government Documents Department

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Evaluation of severe accident risk during mid-loop operations. Volume 6, Part 2: Appendices

Description: The objectives are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation. In phase 2, mid-loop operation was selected as the plant configuration to be analyzed. Volume 1 summarizes the results of the study. The scope of the level-1 study includes plant damage state analyses, and uncertainty analysis. The internal event analysis is documented in Volume 2. The internal fire and internal flood analysis are documented in Volumes 3 and 4, respectively. A separate study on seismic analysis, documented in Volume 5, was performed for the NRC by Future Resources Associated, Inc. A phased approach was used in the level 2/3 PRA program, however both phases addressed the risk from only mid-loop operation. The first phase of the level 2/3 PRA was initiated in late 1991 and consisted of an Abridged Risk Study. This study was completed in May 1992 and was focused on accident progression and consequences, conditional on core damage. Phase 2 is a more detailed study in which an evaluation of risk during mid-loop operation was performed. The results of the phase 2 level 2/3 study are the subject of this volume of NUREG/CR-6144, Volume 6. This report, Volume 6, Part 2, consists of five appendices containing supporting information for: the PDS (plant damage state) analysis; the accident progression analysis; the source term analysis; the consequence analysis; and the Melcor analysis. 73 figs., 21 tabs.
Date: May 1, 1995
Creator: Jo, J.; Lin, C.C.; Neymotin, L. & Mubayi, V.
Partner: UNT Libraries Government Documents Department

Tokamak-FED plasma-engineering assessments

Description: A wide range of plasma assumptions and scenarios has been examined for the current US tokamak FED concept, which aims to provide a controlled, long pulse (approx. 100 s) burning plasma with an energy amplification of greater than or equal to 5, a fusion power of 180 MW, and a neutron wall load of greater than or equal to 0.4 MW/m/sup 2/. The results of the assessment suggest that the current FED baseline parameters of R = 5.0 m, B/sub t/ = 3.6 T, a = 1.3 m, b = 2.1 m (D-shape), and I/sub p/ = 5.4 MA are appropriate in reaching the above plasma performance, despite uncertainties in several plasma physics areas, such as confinement scaling, achievable beta, impurity control, etc. To enhance the probability of achieving fusion ignition and to provide some margin against a short fall in our physics projections in FED, a limited operating capability at B/sub t/ = 4.6 T and I/sub p/ = 6.5 MA is incorporated. Various other options and remedies have also been assessed aiming to alleviate the impact of the uncertainties on the FED design concept. These approaches appear promising because they can be studied within the current fusion physics program and may lead to drastically more cost-effective FED concepts.
Date: January 1, 1981
Creator: Peng, Y.K.M.; Lyon, J.F. & Rutherford, P.H.
Partner: UNT Libraries Government Documents Department