This report sets forth a basic design philosophy with its associated functional criteria and design principles for present-day, hard-wired annunciator systems in the control rooms of nuclear power plants. It also presents a variety of annunciator design features that are either necessary for or useful to the implementation of the design philosophy. The information contained in this report is synthesized from an extensive literature review, from inspection and analysis of control room annunciator systems in the nuclear industry and in related industries, and from discussions with a variety of individuals who are knowledgeable about annunciator systems, nuclear plant control rooms, or both. This information should help licensees and license applicants in improving their hard-wired, control room annunciator systems as outlined by NUREG-0700. i; i

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To address identified problems and provide information from which behavior of large-diameter wire rope could be better understood, efforts in the following areas were undertaken during FY79 and continued in FY80:• large-diameter rope testing • small-diameter rope testing • data analysis and evaluation • wear and failure analysis • load sensor development • technology transfer.Wire ropes 3/4 in., 1-1/2 in., and 3 in. in diameter were tested in bend-over-sheave fatigue. Attempts were made to correlate fatigue life of these ropes. limited field rope data were available to compare with test results. The modes of failure and wear in laboratory ropes were compared with those seen previously in field ropes.A load sensor was designed and ordered in FY79. It will be connected to the drag rope and jewelry of working draglines during the summer of FY80.Technology transfer was achieved through disseminating written materials, conducting seminars, holding a national symposium, and filming of selected field operations • iii • . .

Pacific Northwest laboratory (PNL) was asked to develop and recommend a regulatory position that the Nuclear Regulatory Commission (NRC) should adopt regarding the ability of reactor pressure vessels to withstand the effects of pressurized thermal shock (PTS). Licensees of eight pressurized water reactors provided NRC with estimates of remaining effective full power years before corrective actions would be required to prevent an unsafe operating condition. PNL reviewed these responses and the results of supporting research and concluded that none of the eight reactors would undergo vessel failure from a PTS event before several more years of operation. Operator actions, however, were often required to terminate a PTS event before it deteriorated to the point where failure could occur. Therefore, the near-term (less than one year) recommendation is to upgrade, on a site-specific basis, operational procedures, training, and control room instrumentation. Also, uniform criteria should be developed by NRC for use during future licensee analyses. Finally, it was recommended that NRC upgrade nondestructive inspection techniques used during vessel examinations and become more involved in the evaluation of annealing requirements.Pacific Northwest Laboratory (PNL) was asked by the Nuclear Regulatory Commission (NRC) to develop and recommend a near-term (<1 year) regulatory position that NRC should adopt to avoid or mitigate pressurized thermal shock (PTS) at nuclear power plants. The PNL technical staff and several independent consultants, who provided an overview of the program, evaluated what corrective actions, if any, must be taken before longer-term PTS generic resolution and acceptance criteria are established. Responses to NRC's request for information are still being received from licensees and owners groups. In this regard, the PNL review is limited to information available through May 1982.The responses considered several classes of overcooling scenarios which could lead to a PTS event. For all scenarios, it was concluded that none of the eight reactors under review would undergo vessel failure should a PTS event occur before several more years of operation and, in most cases, before the end of reactor life. However, in many scenarios, operator actions were required to terminate the event before it deteriorated to a state where the conditions necessary for vessel failure were present. The NRC evaluation of PTS procedures and operator training at two of the eight plants indicated deficiencies in these areas. Therefore, it is recommended that procedures, training, and control room instrumentation be changed on a site-specific basis in the nearto long-term period.In addition, the responses differed in terms of event conditions, assumptions, and acceptance criteria beyond what would be expected because of plantspecific situations. It is therefore recommended that uniform criteria be used to evaluate the effective full power years (EFPY) remaining before further corrective actions are required. Adopting these...

The Pacific Northwest Laboratory has studied the safety aspects of monitoring the preoperational status of safety systems in nuclear power plants. The goals of the study were to assess for the NRC the effectiveness of current monitoring systems and procedures, to develop acceptance criteria by which the adequacy of safety status monitoring systems can be evaluated, to develop nearterm guidelines for reducing human errors associated with monitoring safety system status, and to recommend a regulatory position on this issue. A review of safety system status monitoring practices indicated that current systems and procedures do not adequately aid control room operators in monitoring safety system status. This is true even of some systems and procedures installed to meet existing regulatory guidelines (Regulatory Guide 1.47). In consequence, this report suggests acceptance criteria for meeting the functional requirements of an adequate system for monitoring safety system status. Also suggested are near-term guidelines that could reduce the likelihood of human errors in specific, high-priority status monitoring tasks. It is our recommendation that 1) Regulatory Guide 1.47 be revised to address these acceptance criteria, and 2) the revised Regulatory Guide 1.47 be applied to all plants, including those built since the issuance of the original Regulatory Guide.