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Home My WebLink About841141.tiff `s UNITED STATES , NujLEAR REGULATORY COMMISSION b o t � REGION IV �' a,w.f ft: 611 RYAN PLAZA DRIVE, SUITE 1000 ARLINGTON, TEXAS 76011 *•+,a qq, liiPX Z C E34 wEtu wpm, ,7a,1,, Docket: 50-267 0 °^IISJ1Gif;7S MAY 2 2 1984 /1 Mr. 0. R. Lee, Vice President cq,L+_ Electric Production LEy Public Service Company of Colorado P.O. Box 840 Denver, Colorado 80201 Dear Mr. Lee: We have reviewed the information provided in your April 12 and 25, and May 7, 1984, letters related to the prestressed concrete reactor vessel (PCRV) tendon wire corrosion problem at Fort St. Vrain (FSV). Based on that information, supplemental information from supporting documents such as the FSAR, and our independent calculations, we have concluded that the integrity of the PCRV will not be compromised by the existing tendon wire degradation. Therefore, we find it acceptable for FSV to return to power operation. The details of our review are contained in the enclosed safety evaluation. In order to provide assurance that the existing corrosion problem has been abated and that the load carrying capability of the tendons continue to be adequately monitored, we request that you provide us the details of your proposed tendon surveillance program within 30 days of your receipt of this letter. We understand that this expanded tendon surveillance program will be incorporated into the FSV Technial Specifications after it is finalized. If you have any questions on this subject, please contact your project manager. Since this request relates solely to FSV, OMB clearance is not required under P. L. 96-511. Sincerely, -az �uari E. H. Johnson, Chief Reactor Project Branch 1 Enclosure: Safety Evaluation cc: Attached List 841 141 /77©D/e Fort St. Vrain cc list C. K. Millen Chairman, Board of County Commissioners Senior Vice President of Weld County, Colorado Public Service Company Greeley, Colorado 80631 of Colorado P. 0. Box 840 Regional Representative Denver, Colorado 80201 Radiation Programs Environmental Protection Agency James B. Graham, Manager 1860 Lincoln Street Licensing and Regulation Denver, Colorado 80203 East Coast Office General Atomic Company Don Warembourg 2021 K Street, NW, Suite 709 Nuclear Production Manager Washington, DC 20006 Public Service Company of Colorado P. 0. Box 368 J. K. Fuller, Vice President Platteville, Colorado 80651 Public Service Company of Colorado Albert J. Hazle, Director P. 0. Box 840 Radiation Control Division Denver, Colorado 80201 Department of Health 4210 East 11th Avenue Denver, Colorado 80220 G. L. Plumlee NRC Senior Resident Inspector Kelly, Stansfield & O'Donnell P. 0. Box 640 Public Service Company Building Platteville, Colorado 80651 Room 900 550 15th Street Denver, Colorado 80202 Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 ENCLOSURE SAFETY EVALUATION REPORT FOR FORT ST. VRAIN REACTOR TENDONS 1.0 INTRODUCTION During a shutdown for reactor refueling the licensee began the reactor vessel prestressing tendon surveillance in accordance with technical specification SR 5.2.2 a - X, "Tendon Corrosion and Anchor Assemblies Surveillance". The procedure requires a visual inspection for corrosion and a lift-off test to confirm the prestress force of the tendons. This procedure was initiated on March 23, 1984. The first tendon lifted off and examined was VM-17, a vertical tendon which consists of 169 wires that are 1/4 inch in diameter. Several broken wires were found and extensive corrosion was evident in the area of the fracture. The NRC was informed of the broken wires and two reports have been issued by the licensee. One report was by letter dated April 12, 1984 (No. P-84110) and the other, a response to an NRC request for additional information , was dated April 25, 1984 (No. P-84119) . This SE addresses the ability of the reactor vessel to withstand the design internal pressure. • 2.0 DESCRIPTION OF THE REACTOR VESSEL The reactor vessel at Fort St. Vrain is a prestressed reinforced con- crete cylinder with a flat plate upper and lower head; refer to figure 1. The prestressing system for the concrete reactor vessel is composed of ""stem of 90 vertical 'longitudinal ) , 310- circumferential . 2- lower an;i 24 upper head transverse tendons; refer to figure 1 for details. -2- A "tendon" is a bundle of separate specially treated wires, 1/4 in dia- meters. Each of the bundles for £10 circumferential tendons in the reactor vessel wall have 152 wires while all other tendons have 169 wires per bundle. The purpose of these prestressing tendons is to resist the tensile forces generated from the pressure inside the reactor vessel cavity. The FSAR states that the tendons are designed to resist an internal reference pressure of 845 psig without producing tensile forces in the concrete. The prestressing system also has an alarm system incorporated into the design. This alarm system consists of a load cell incorporated into six vertical tendons, 17 circumferential , 2 upper head transverse and 2 lower head transverse tendons. These load cells are connected to an alarm that will alert the operators if the prestressing force falls below a certain load. The tendon wires were coated by the manufacturer with calcium-zinc phosphate (Meta Bond) and overcoated with a water emulsified oil (Rustarest) which reportedly seals the pores of the phosphate and dries to leave a wax-like finish on the wires. The installation procedure of the tendon into the tubes calls for a continuous coating of the tendon with a film of No-Ox-Id CM (tradename of Dearborn chemical) as the final corrosion protection. -3- 3.0 LICENSEE INVESTIGATION The tendon surveillance procedure requires an inspection of 5 tendons: 1 vertical , 3 circumferential , and 1 upper cross head. This inspection was initiated, on March 23, 1984. The first tendon lifted off and examined was VM-17, a vertical tendon. It contained 4 broken wires out of a tendon design of 169 wires that are 1/4 inch in diameter. The broken wires showed evidence of extensive corrosion in the area of the fracture. The wires are protected from corrosion by protective coatings applied at the fabricators and additionally in the field during installation. Missing areas of the protective coating were observed on the remaining wires in the tendon. Based on the conditions observed in the tendon the surveillance was expanded to include all of the tendons in the reactor vessel . The licensee has reported in the April 12, 1984 and April 25, 1984 letters that 89 vertical , 4 top head, 44 bottom head and 33 circumfer- ential anchor assemblies have been visually inspected for corrosion and broken wires. No top cross head tendons contain broken wires while the bottom cross head was found to have 7 tendons with broken wires and 2 of the circumferential tendons had broken wires. Of the 7 bottom head tendons that contained broken wires, it was found that: one tendon (BILU4) contained 17 broken wires, one tendon (BILU3) contained 12 broken wires, one tendon contained 3 broken wires, one tendon contained 2 broken wires and 3 tendons had 1 broken wire. Lift off data for circum- ferential and cross head tendons will not be available until a tendon jack can be repaired and calibrated . The two load cells in the bottom head were reported to read 1155 kips and 1191 kips of prestressing force. The licensee states that the force necessary for reactor operation and to maintain the FSAR safety factors is 980 kips , the low nominal alarm value. The two load cell readings for the top head tendons were reported to be 1003 and 1210 kips. These forces are above the licensee stated minimum value of 980 kips. The 17 load cells in the circumferential tendons were read by the licensee. The minimum value for the 169 wire tendons was 1144 kips and the largest value was 1313 kips. The forces read by the licensee for the 152 wire circumferential tendons was 1011 kips minimum and 1072 kips maximum. All of these forces are larger than the forces the licensee states are necessary for reactor operation. The licensee observed that of the tendons that have exhibited some corrosion on the wires, this corrosion was found only on the factory prepared button head end. The licensee reported that no corrosion has been found on the field prepared button head end. Additionally, the corrosion of the tendon wires occurred at the joints between the com- posite washer and the split shims , between the split shims and between the split shims and the base plate. Refer to figure 2 which shows the corrosion locations. The washer that holds the button head of the tendon -5- wire is, supported by several split shims between the composite washer and the base plate. These shims vary in height from approximately 2 to 6 inches. The split shim stack usually consists of two thick shims and several thinner shims, usually 1/8 to 1/4 inch thick. Refer to Figure 2 for more details of the shim stack. A metallurgical examination of the broken tendon wires is being conducted by the licensee to determine the possible cause of the corrosion and wire failure. This report is not available as of the preparation of this SE The report should be of value in developing a corrosion prevention program and preparing an intensified inspection program. The staff will require that the licensee submit a new surveillance program that is designed to detect early signs of corrosion and possible additional wire failures. The staff has also reviewed the tendon surveillance documents from the 1973 and 1979 surveillances. The 1973 inspection showed one tendon (BORL4) had some corrosion on the upper segment of the bearing plate and the area along a sample wire corresponding to the split shims location. The sample wire was analyzed in the laboratory and it was shown the wire strength was essentially the same in the corroded area as in the uncorroded areas. The 1979 surveillance did not inspect the same wires as the 1973 surveillance. No corrosion was found. However, some water was found in Tendon CO-1 .6 cover, but no rust was reported on the tendon itself. The surveillance procedure does not require lift off tests of any tendon, therefore, no lift off forces are available. . 0 STAFF FINDINGS The staff has reviewed the two submittals by the licensee, Public Service Company of Colorado, dated April 12 and April 25, 1984 and the 1973 and 1979 tendons surveillance reports. Based on these documents and a visit to the plant on April 12, 1984 the staff makes the following findings: There is some evidence that the corrosion has been in progress for an extended period of time. The licensee has inspected a large percentage of the vertical tendons and found that the minimum prestress load of 1296 kips is larger than the minimum necessary to maintain the required safety margins for operation of the reactor and is larger than the projected minimum prestress force of above 1175 kips. The licensee has inspected a smaller number of the upper and lower head circumferential tendons and found that the load cells built into these tendons show they have a larger prestress force than necessary to maintain reactor integrity and the FSAR committed safety margins. The licensee has indicated they intend to modify the surveillance requirements to a semi-annual surveillance program to inspect the tendon terminations. The licensee stated that the inspection program -7- will inspect a minimum of 33% of the vertical tendon terminations in each inspection and include some samples of the corroded circumferential , top cross head and bottom cross head tendons. The licensee will monitor the twenty-seven (27) tendon load cells on a monthly basis for trends of relaxation of the prestress values. 5.0 CONCLUSION The staff concludes that some degrading of the tendon system has occurred. The lift off data available to date indicates that currently available prestress is larger than the forces necessary to satisfy the reactor design conditions of no tensile forces in the concrete at the reference pressure of 845 psig. Therefore, the staff concludes that the reactor vessel still meets its required design strength. The staff- requires that the licensee submit the results of its complete surveillance program currently underway and laboratory test tendon wires removed during the inspection. These tests and augmented surveillance program will assist the licensee in developing his program to prevent or mitigate further deterioration of the tendon wires . The staff requires the licensee to submit the details and results of his surveillance program and the results of his laboratory tests as soon as available. 1 zQ _ _ . i f • -. 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Box 840 Denver, Colorado 80201 Dear Mr. Lee: In order to evaluate the adequacy of the Detailed Control Room Design Review (DCRDR) you are performing for the Fort St. Vrain station (FSV), we have reviewed your program plan and conducted an onsite, in-progress audit. This audit was conducted from March 12 to 15, 1984. Although a full evaluation of the DCRDR is not possible because you have not completed several parts of the review, our evaluation thus far indicates that you are pursuing a course of review which should satisify most of the requirements of Supplement 1 to NUREG-0737. In addition to some minor modifications, the areas of task analysis and the subsequent comparision of the results of the analysis with control room inventory need to be reviewed in order for your DCRDR to be found acceptable. The details of our review and audit are contained in the enclosed report. We request that you review this report and factor the conclusions into the DCRDR. Any changes to the ongoing, or committed to, actions should be submitted to us to ensure agreement prior to your Summary Report submission. If you have any questions on this subject, please contact your project. manager-. Sincerely, E. H. Johnson, Chief Reactor Project Branch 1 Enclosure: Audit Report cc: Attached List ] 71I -5/93 ig51 Fort St. Vrain cc list C. K. Millen Chairman, Board of County Commissioners Senior Vicc Presidcnt of Weld County, Colorado Public Service Corpeny Greeley, Colorado 80631 of Colorado P. 0. Box 840 Regional Representative Denver, Colorado 80201 Radiation Programs Environmental Protection Agency James B. Graham, Manager 1860 Lincoln Street Licensing and Regulation Denver, Colorado 80203 East Coast Office General Atomic Company Don Warembourg 2021 K Street, NW, Suite 709 Nuclear Production Manager Washington, DC 20006 Public Service Company of Colorado P. 0. Box 368 J. K. Fuller, Vice President Platteville, Colorado 80651 Public Service Company of Colorado Albert J. Hazle, Director P. 0. Box 840 Radiation Control Division Denver, Colorado 80201 Department of Health 4210 East 11th Avenue Denver, Colorado 80220 G. L. Plumlee NRC Senior Resident Inspector Kelly, Stansfield & O'Donnell P. 0. Box 640 Public Service Company Building Platteville, Colorado 80651 Room 900 550 15th Street Denver, Colorado 80202 Darrell G. L; enhut. Director Division of Licensing Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 NUCLEAR REGULATORY COf4,W.ISSJUN IN-PROGRESS AUDIT OF THE DETAILED CONTROL RULI', DESIGN REVIEW FOR FORT ST. VRAIN NUCLEAR GENERATING STATION This report documents the findings of the NRC staff and its consultant, Science Applications, Inc. (SA1) , during an in-progress audit of the Detailed Control Room Design Review (DCRDR) of Public Service Company (PSC) of Colorado's Fort St. Vrain (FSV) Nuclear Generating Station conducted March 12-15, 1984. The requirements set forth in NUREG-0737, Supplement 1, "Requirements for Emergency Response Capability," (Reference 5) served as the basis for the in-progress audit. The purpose of the in-progress audit was twofold: (1) to ensure that PSC is not proceeding with an inadequate program, and (2) to provide PSC with feedback on its approach. BACKGROUND Licensees and applicants for operating licenses are required to conduct a Detailed Control Room Design Review (DCRDR) . The objective is to ". . .improve the ability of nuclear power plant control room operators to prevent accidents or cope with accidents if they occur by improving the information provided to them" (NUREG-0660, Item I .D.1) . The need to conduct a DCRDR was confirmed in NUREG-0737 and Supplement 1 to NUREG-0737. DCRDR requirements in Supplement 1 to NUREG-0737 replaced those in earlier documents. Supplement 1 to NUREG-0737 requires each applicant or licensee to conduct a UCRUR on a schedule negotiated with the NRC. Guidelines for conducting a DCRDR are provided in NUREG-0700 and assessment processes for NRC are contained in draft NUREG-0801 (The NUREG -documents cited are listed as References 3 through 7) . A UCRDR is to be conducted according to the licensee's program plan (which has been submitted to the NRC) . NUREG-0700 recommends that the DCRDR include four phases: (1) planning, (2) review, (3) assessment, and (4) reporting. The product of the last phase is a summary report which must include an outline of proposed control room changes, proposed schedules for implementation, and justification for human engineering discrepancies with safety significance to be left uncorrected or partially corrected. Upon receipt cf the licensee' s summary report and prior to impler,entation of proposed chances, the NRC will prepare a Safety Evaluation Report (SER) . PSC has structured its DCRDR into the five phases recommended hi the NUTAC CRUR Guidelines. These phases are: (1) Planning, (2) Execution, (3) Assessment., (4) Correction, and (5) Post Review. The submissicn e the FSV pre r plan was the concluding activity cf the Planning Phase . The program , kelerence 1) was receives by the NRC on July 18, 198 . the time the as ccrcucted, PSC had completed mcst of the Executirr H se. The Phase included data collection H HED idortilicatu ctivities . • - L - the remainder of the DCRDR will consist of assessing and resolving HEDs, imtiementing HEAD corrections in the control room and integrating these changes with the other improvement programs , and submitting the summary report. After the DCRDR is completed, PSC plans to conduct two ongoing activities referred to in the program plan as the "Effectiveness Phase" and the "Design Control Plan" (p. 36 of the program plan) . The objective of the Effectiveness Phase is to assure that changes resulting from the DCRDR. are effective. This objective will be accomplished by "selective validation" and operator interviews. The objective of the Design Control Plan is to ensure that post-DCRDR changes to the control room are evaluated for human factors acceptability. This objective will be accomplished by. evaluating the proposed changes on FSV detailed mockup or a similar tool before the changes are made in the plant control room. The DCRDR in-progress audit was conducted concurrently with a pre-implementation audit of the FSV Safety Parameter Display System (SPDS). The results of the SPDS Audit are reported under separate cover. The DCRDR audit began with an entrance briefing conducted by the NRC audit team and an overview of the FSV HTGR by PSC staff. These initial briefings were followed by in-depth discussions about specific DCRDR program elements, review of • selected portions of the DCRDR documentation, and verification of some PSC findings. The audit was concluded with an exit briefing. Participants in the audit meetings are identified in Exhibit A. Results of the In-Progress Audit are summarized below. PLANNING PHASE • The two areas relevant to the Planning Phase that were addressed during the audit were the DCRDR team composition and the system for documenting DCRDR• materials. The composition of the DCRDR team and the involvement of the various disciplines comprising the team became apparent through discussions about the DCRDR activities whereas the system for dccumcnt,tion was specifically addressed. Qualifications and Structure of the DCRDR Team The concerns raised by the NRC in its comments on the FSV DCRDR. Program Plan (Reference 2) for this requirement were the following: o The lack of a qualified person experienced in nuclear engineering to be involved in the DCRDR on a daily basis, and o The rule of the human factors specialist in the DCRDR. While the lack of involvement of a person with the nuclear engineering qualifications recommended by the NRC does not necessarily preclude the successful performance of a DCRDR, the NRC emphasizes that personnel ',pith c;oalificaticrs ancf experience appropriate to PSC' s LCR.DF, tasks should he aec:. Through di :, un ions with PSC representatives on Rho metheeclopies of F.un activities 1,L 1 cLri: that the Review Teain hcncer, who possesses the • - 3 appropriate nuclear engineering qualifications , is not involved in the daily work of the review team. The Engineering Technician, who possesses nine to ..ten years of nuclear-related experience, is involved in the daily work of the review team, but does not meet the recommended educational qualifications. Despite this , we found no evidence that this arrangement had a detrimental effect upon the conduct of the DCRDR activities. The human factors specialist appeared to be the only position that was not filled at the time of the program plan submittal . Since part of the role of a human factors engineer is to provide guidance in the planning phase of the DCRDR, concerns were raised about the planned level of involvement of the human factors specialist in the rest of the DCRDR activities, especially in the Execution Phase. The NRC audit team had extensive interaction with the individual filling the role of human factors specialist and found his role in the Execution Phase to be primarily in the latter portion of the task analysis (see this report's "Function and Task Analysis" section) . The other activities comprising the Execution Phase were structured in a way which PSC believes precludes the need for participation by the-human factors specialist. While we believe the review activities Completed to date, notably the control room survey, would have benefitted from additional participation by human factors personnel , we find that the minimum requirements'for a multidisciplinary team have been met thus far in the DCRDR. The intended role of the human factors specialist for the remaining DCRDR activities was found to be satisfactory by the NRC audit team. In summary, the concerns raised by the NRC about PSC's staffing plans have been satisfied. Therefore, we believe that PSC will meet the NUREG-0737 Supplement 1 requirement for "'The establishment of a qualified multidisciplinary review team. . ." . Documentation System Although no requirement exists for a specific system for documenting DCRDR materials, our review of the documentation produced from the DCRDR activities performed up to .the time ,of the audit reflects a practical and well planned approach to record-keeping by PSC. The files and notebook binders appear to document every activity conducted so far. There appears to be an ample amount of cross-references to enhance the accessibility of HEDs, survey data,. and other documentation. A specific example demonstrating the usefulness of the documentation system and the foresight PSC had in establishing this system is the records for the Instrument and Function Tag Survey. These records served as an efficient basis for performing the control room survey work, most notably the labeling survey. Based upon the documentation we have seen, we believe that this documentation system should effectively support the remaining DCRDR activities . • • 4 - REVIEW PHASE The activities recommended in NUREG-0700 in the Review Phase are performed by PSC in its Execution Phase. PSC's activities include.the following: 1. Operating Experience Review 2. Task Analysis 3. Control Room Survey Activities two and three addresss specific DCRDR requirements contained in NUREG-0737 Supplement 1. For reasons discussed in the Control Room Inventory section of this report, an actual inventory of instruments and controls in the control room was not developed. Operating Experience Review PSC's review of operating experience comprises two areas of data collection and analysis: A. ` Review of plant historical documents B. Operator surveys or questionnaires The review of plant historical documents covered various sources of information, including. Reportable Occurrences (R0s) , Plant Trouble Reports (PTRs), EPA violations, and flagged logs. Although FSV has been in commercial operation since July 1979, the scope of PSC's review was limited to the last one-and-a-half years of plant historical documentation. PSC's reasons for limiting the scope of the review to this-time period were: (1) only in this time period has FSV achieved a fairly steady state of operation and consistency in control room design, (2) the benefit of going further back in time was not judged to be substantial enough to justify the effort, and (3) other DCRDR activities , such as task analysis , control room survey, and operator surveys, will pick up HEDs that possibly would have been found in earlier plant historical documents. PSC' s reasoning does not appear to provide a solid rationale for not going back further in the plant historical documentation. We believe that additional valuable information can be found by reviewing the documentation of incidents that occurred in the earlier history of the plant, such as the two loop dump and the loss of all four helium circulators. Although these incidents may have been "resolved," the DCRDR offers the opportunity to review these and other incidents from a new perspective; human factors. The potential benefit of reviewing these incidents from a human factors perspective is the assurance that the operators have the instruments and cont;os necessary to support operators in future similar events and t.h;.t these iinstrurner;ts and controls arc suitable for their intendec - 5 - functions. Additional benefit may be received from this enlarged scope of review by familiarizing the operators and other relevant personnel who were not at FSV at the time the selected events occurred with -these significant events. The operator surveys appeared to provide a useful means of providing operator input to the DCRDR and served as a "warm-up" function for subsequent activities. PSC judged the HEDsidentified from the review of operating experience to be valuable inputs to the task analysis walk throughs by marking areas where special attention was necessary. We suggest that PSCenlarge the scope of the review of operating history to include earlier incidents and events which may provide additional valuable input to the task analysis walk through effort. We believe that the Operating Experience Review has made a valuable contribution to the overall DCRDR and that additional -benefit can ,be derived from expanding its scope. Function and Task Analysis PSC stated that .NUTAC' guidance was followed in the development of the task analysis activities. PSC's methodology for reviewing;FSV functions and performing the task analysis consisted of the following steps: A. Definition of FSV Critical Safety Functions (CSFs) B. Table Top Analysis of Tasks C. Paper and Pencil Analysis of Tasks D. Walk-throughs of Operational Scenarios This methodology was performed by engineering and operations personnel at PSC. A human factors consultant was involved in the .paper and pencil analysis and the walk-throughs. - Before defining the CSFs for FSV, PSC performed a cursory review of FSV's set of event-based -emergency procedures . This set of emergency procedures was divided into "emergency," "catastrophic" and "abnormal" condition categories. The emergency procedures that addressed "emergency" conditions were selected to provide the basis for defining FSV's CSFs. Through lengthy discussion with PSC representatives the NRC audit team concluded that PSC's definition of "emergency" conditions and thus the scope of the FSV task analysis was comprehensive and appropriate for the DCRDR. From an analysis of the set of emergency procedures addressing emergency conditions , the CSFs of FSV were ccfined as (1) reactivity, (2) core cooling, and (3) containment. PSC used each of the three CSFs as bins into which em ryency procedures relevant t e! eh CSF were sorted. In the interest of _ 6 developing a set of emergency procedures that would include more operational contingencies than the present set of emergency procedures, PSC proposed that three symptom-based emergency procedures be developed from the three emergency procedure groupings. These emergency procedures would ;be- (1) Reactor Shutdown, (2) Assure Adequate Core Cooling, and (3) Containment. In the Table Top and Paper and Pencil efforts, PSC developed flowcharts of the stimuli and resultant steps or tasks necessary to ensure reactor shutdown, core cooling, and containment integrity. The stimuli in these flowcharts are indications, alarms, or other cues of plant status. PSC developed these flowcharts as part of the task analysis and as the basis for upgrading emergency procedures. The Table Top analysis of tasks was performed at a general level and produced flowcharts of the "hypothetical tasks" the operator has to perform in order to maintain FSV's CSFs. After diagramming these general tasks, PSC operators then listed the instruments, communications, or other aids required to complete each task. PSC used this list as the basis for conducting two activities: (1) a survey of alarms and (2) a verification of availability and useability of the stimuli instrumentation. The survey. of alarms was essentially a review of all alarm locations, accuracies, and applicability. The verification effort was essentially a verification that the stimuli listed in the flowcharts were represented by actual instrumentation in the control room and that the information presented was useable for its intended function. HEDs were generated from these two activities. The Pencil and Paper analysis of tasks differed from the Table Top effort in that it defined the specific stimuli and the resultant tasks the operator has to perform in order to maintain FSV's CSFs. The flowcharts were refined to reflect this greater specification of stimuli and tasks. PSC also listed the specific instrument numbers on the flow charts and mapped the locations of the instruments on panel layout drawings. PSC then verified that the instruments and controls needed to complete the tasks'on the flowcharts were available in the control room. This verification effort provided the basis for developing general flowcharts for the walk-through effort and also generated HEDe. The walk-through effort consisted of three sequentialrefforts. i. The derivation of general flowcharts from the flowcharts generated in the Table Top and Paper and Pencil efforts for use in the walk-throughs of operational scenarios, 2. The development of walk-through data sheets based upon the general flowcharts, and 3. Walk-throughs ci the three proposed symptom,-based emergency procedures as repesented in the general flowcharts. - / - The walk-throughs were essentially a validation of operator tasks and of the instruments needed by the operator that were identified in the paper and pencil analysis. One of the objectives of the requirement for a function and task analysis is to identify not only the control room operator tasks during emergency operations, but also the information and control requirements for performing those tasks. To properly identify the information and control requirements, eacli task must be analyzed to determine what information the operator needs to have and what control capabilities the operator must have in order to complete the task. These information and control requirements should be specified in terms ofthe characteristics of the information and control capabilities needed. The analysis of information characteristics includes determining the appropriate type of information that the operators need (e.g., trends, status indication, digital , analog), its units (e.g. , PSIG, inches of H.U, RPM) , the range which would be inclusive of all possible readings, rosolution,_acceptable tolerances and so on. The analysis of control characteristics includes determining the type of control capability needed (e.g. , discrete versus continuous) , rate, gain, response characteristics and lotus of control (e.g. , manual , automatic, operator selectable) , etc. The;analysis of information and control requirements should be done independently of the sources of the information and control in the control room. In this way an objective and valid ;basis can be provided for verifying the (1) availability of information resources and control capabilities in the control room and (2) the suitability of the available instruments and controls for their intended functions. PSC stated that the information requirements to perform emergency tasks were identified during the paper and pencil effort prior to,specification of existing instruments and controls in the control room. Judging from the written and oral presentation of the task analysis methodology given by PSC, we do not believe that the information requirements were (1) identified in most instances, (2) identified down to the characteristics level or (3) identified independently of existing control room instruments. Based on our definition of information requirements, we found no evidence that this analysis was performed in the paper and pencil effort and only partial evidence:'of this analysis in the walk-through effort. In neither effort did we find evidence of a. level of analysis that`,specified the characteristics of the. needed information. Based on PSC's definition of information requirements, an analysis was performed in the paper and pencil effort, although this'-still did not achieve the information characteristics level of specificity. An example of an information requirement by PSC's definition is "Verily :reactor power decreasing. " In our view, this is a task or task requirement. lo perform an analysis of information requirements, cne must ask "What does the operator need to perform this task?" or "What information does the operator need to verify reactor power is decreasinc anc' kowshould this information be presented?' PSC is essentially eefinine what the operator needs to do to maintain the CSF. While this c"iurt is appropriate, it is net complete. - & The general flowcharts developed in the walk-through effort demonstrate that in some, but nut all , instances PSC identified the information requirements, although still not a t'a level of analysis that specified information characteristics: An example_ of this can be found for the task "Verify abnormal flux" which follows the recognition of abnormal flux indications. These indications are the entry conditions for the reactor shutdown procedure. PSC lists four information requirements :under this task: (1) Positive rate of change, (2) Recorder trace increasing, (3) Runback rods inserting, and (4) Reheat temperature increasing. However, PSC does not list the information requirements for any of the other tasks in this procedure. Also, the analysis of information requirements that was performed in the walk-through effort was conducted after the verification of instruments and controls in the paper and pencil effort. The order in which the analysis was performed relative to the verification effort may have-an effect upon the validity of the analysis. In addition to performing an analysis of information requirements, part of the objective of the function and task analysis requirement is to identify control requirements. Judging from the written and oral presentation of the task analysis methodology given by PSC, we do not believe that a thorough analysis of control requirements was performed. Our concern here is the same as that described for the definition of information requirements. That is, little evidence was provided that demonstrated that PSC has identified, independently, of existing controls, the characteristics of controls needed to carry out emergency tasks. Control Room Inventory. NUREG-0737, Supplement I, requires that the DCRDR include "a comparison of the display end control requirements with a control room inventory to identify r,issinc displays and controls." PSC approached this reouirement through uEe of the control room mockup rather than by developing a detailed control room, inventory. While this approach can be acceptable, the problems with the identification of information and control capability requirements in the task analysis make ;it unlikely that the intent of the inventory requiremert is being met. The problem is not with the`use of the control roori mockup; we found it to be well-conceived, detailed, and an excellent tool to • i support the DCRDR. Rather, our concern is with the -definition of information and control requirements against which the inventory (or mockup)-is compared. Control Room Survey The guidelines PSC developed for use in the control room survey originated from various sources of human factors criteria, primarily the NUTAC Control koom Desigr Review Survey Development Guideline. The survey was performed by the CRUR coordinator two engineering technicians, one maintenance technician, and a person with a non-technical background. Although no personnel +:iu` human factors expertise were used in the survey, PSC contends that the d- ign of the survey precluded the need for this expertise. For Et - g least the "static" criteria, the structure of the guidelines and checklists limited the response or judgment of the evaluator to a "yes" or "no" evaluation. Explicit instructions given to personnel on the checklisting methodology precluded ,judgments more complex than the binary type that would require human factors expertise. PSC stated that the criteria that are "dynamic" or that should be evaluated in an operational context were addressed by one or more of the following activities: (1) operational analysis, (2) operator surveys, and (3) task "analysis walk-throughs. The operational analysis is an evaluation of the panel layout and control/display integration. This analysis was performed by a person who is naive in regard to the systems on the control boards. The analysis was not based upon the operations or tasks the operator performs but upon the labeling of the components on the panel layout drawings and Piping and Instrumentation diagrams. So in actuality, the dynamic criteria were not addressed in the operational analysis. The operator surveys, which were discussed in the Operating Experience Review of this report, are not designed to systematically andi specifically address all applicable dynamic criteria. The use of the task analysis walk-throughs as a vehicle to address the dynamic criteria raises two concerns: (1) apparently contradictory statements made by PSC about which activity, either the task analysis walk-through or the control room survey, evaluated the suitability of controls in an operational or "dynamic" context and (2) the task analysis walk-throughs only allow an evaluation of the controls and displays that are involved in emergency operations. If the task analysis walk-throughs were, in fact, the only activity that systematically and specifically addressed controls and displays from an operational perspective, then the controls and displays that are not involved in emergency operations were not evaluated with the dynamic criteria. The dynamic criteria may have been applied to all controls and displays by means of operational expertise or assistance during the checklisting. However, this remains to be verified by PSC. PSC states, that the survey was incomplete and this became apparent during the review of survey documentation when we found that not all guidelines were evaluated. A specific example from our review is PSC's guideline "LMD-12" which directs the checklister to find out if "strings of six' or more components or matrices of greater than 4x5 similar components arc demarcated in functional groups or are mimicked." HWe found the spaces provided for recording checklisting data for this guideline were not filled. PSC did not evaluate the guidelines which addressed torque requirements for controls. These and other guidelines which required manipulation of controls were not measured because of the on-line operational status of the control room during the period of time when the survey was conducted. PSC is relying or: the operator survey effort to identify controls which present operational prel,l€tis. We should note that one NED was identified from the operator survey that was applicable to these guidelines. - 10 - In summary, the content of the survey appeared to be comprehensive it respect to static guidelines and the performance of the survey was systematic. however, two concerns we have must be addressed by PSC in order for the NUREG-0737, Supplement 1 requirement to be met. PSC must assure us that: 1. All applicable ' ynamic" criteria have been applied to all controls and displays' in the control room, and 2. All guidelines that were not,evaluated at the time of the audit are identified and evaluated before the control survey is considered complete. ASSESSMENT AND IMPLEMENTATION PHASE The activities recommended by NUREG-0700 in its Assessment and Implementation Phase are performed by PSC in its Assessment and Correction phases. PSC's activities include the following: 1. Assessment of HEDs 2. Selection of corrections and enhancements 3. Integration:of other NUREG-0737 activities 4. Implementation of final design changes. These activities address specific DCRDR requirements contained in NUREG-0737 , Supplement 1. Due to the interrelatedness of: these activities, activities 1, 2, and 4 will be primarily addressed in the "Assessemnt of HEDs" section of this report. A continuation of the HED resolution discussion follows in the sections entitled, "Selection of Design Improvements" and "Verification that Ii ,i ovements will Provide the Necessary Corrections without Introducing New hEDs." The integration of other NUREG-0737 activities is discussed in the "Coordination of the'DCRDR with other Improvement Programs" section. Assessment of HEDs The -HED assessment and resolution process will be performed by personnel representing management and the disciplines of operations , instrument and control , nuclear engineering, and human factors. Engineering technicians will also participate in this process. Since PSC management will ultimately be responsible for the review and approval of the DCRDR team's recommendations fur correcting HEDs,- management should be familiar with the HED generation, assessment, and resolution process. From discussions held with PSC representatives, we found that the interaction between manaper.,ent and the DCRDR team is sufficient to allow informed decisions to be made on the recommendations for correcting HEDs. • - 11 The HEDs which PSC found in its Execution Phase will be assessed using a set of criteria which determines both the significance and classification of HEDs. At the time of .the audit, PSC had not developed .the assessment criteria to be used in sorting HEDs into the following categories. Category 1 (Highest Priority) - HEDs that are judged likely to adversely affect the management of emergency conditions by control room operators. Category 2 - HEDs that are known to have caused problems during normal operation. Category 3 - HEDs that can be "fixed" with simple and inexpensive, and perhaps, obvious enhancements; the so-called "paint, tape, and label" fixes. Category 4 - (Lowest Priority) - HEDs that do not fit into the other categories. Before assessing HEDs`with a set of criteria, PSC plans to perform an=initial screening of the HEDs to streamline the assessment process. This screening will essentially be an informal or intuitive assessment of whether or not the HEDs can be sorted into Category 3. In other words, the screening will determine which HEDs can be easily fixed so that PSCwill not have to undergo the latter assessment process for those particular HEDs. An area of concern we have about PSC's plans for performing this initial assessment and the categorization of HEDs is that the categories are not mutually exclusive. The potential exists for assigning HEDs to an inappropriate category and implementing RED corrections on a schedule inconsistent with that dictated by the HEDs significance. An MED with safety significance ,(Category 1) could be placed in Category 3 since the reviewer determined in the initial screening that the RED was easy to fix. The importance of that hED could be hidden because the determination made in the initial screening precluded the performance of an assessment of the HED's significance. The effect that an error such as this could have on the subsequent implementation process is to postpone the correction of a safety significant RED beyond the time which a Category,; 1 assignment dictates. In other words, there may be a safety significant RED that has been assigned to Category 3 and will not be corrected along with the other safety significant HEDs in Category 1. This concern was discussed during the audit with PSC and PSC responded by stating that the reviewers will take caution not to place HEDs that qualify for Category 1 or 2 in Category 3. We suggest that PSC makes some procedural arrangement to ensure this potential problem will not occur since the initial screening planned by PSC allows only a superficial review of the HEDs. PSC plans to detect HEDs contributing to interactive or cumulative effects with other HEDs by taking advantage of certain features of the documentation system. PSC described pictures or cards which are present it the files that - 12 illustrate the panel or area en the panel on which the discrepancy appears . In addition—to taking advantage of documentation available, PSC envisions accounting for these types of HEDs by maintaining an overall consciousness of the potential for interactive or cumulative effects of HEDs. We suggest that PSC explore -methods to incorporate this "consciousness" in its assessment and resolution procedures. -PSC's overall assessment approach appears- not to have been well defined at the time of the audit. We stressed to PSC during the audit that they should provide detailed documentation of the assessment approach, including the criteria used, in the summary report. We believe that if PSC incorporates the feedback and suggestions given in this report and performs a systematic and comprehensive approach to assessing (and resolving) HEDs, they will meet this requirement of NUREG-0737, Supplement 1. Selection of Design Improvements Although PSC did not define a detailed procedure for selecting corrections to HEDs, we believe that the general approach envisioned is practical and should be adequate. The program plan (p.31) lists the criteria to be considered when selecting corrections and making recommendations. The criteria appear realistic and comprehensive of all necessary considerations. PSC stated that recommendations will include one or more options management can agree to act on resolve the HED. PSC also stated that the description of the corrective actions or justifications for not taking corrective action will reflect management's final decision and PSC's commitment to resolving the HEDs. We suggest that the summary report include a detailed description of the approach to selecting HED corrections. In addition, we suggest that PSC include a description of the HEDs found and the selected corrective action or justification for not taking corrective action in enough detail to allow us to perform an adequate evaluation. The overall approach PSC envisions at this time appears to be sufficient to meet this requirement of NUREG-0737, Supplement 1. Verification That Improvements Will Provide the Necessary Corrections Without Introducinq New HEDs PSC plans to use the control room mockup to verify that the selected HED corrections are valid and that they do not-create new HEDs. Before changes will be made to the control room, the changes will be implemented on the mockup and then assessed. Since PSC and we recognize the potential for an initial operator performance decrement with some changes and PSC intends to use the mockup as a' training device, we suggest that.a sufficient period of operator training time be established on the mockup with the changes in effect. PSC plans to handle changes to the control bores vithin the immediate scope of the DCaDR and further in the future by desk d ing- a committee which will eidcress each changewith a systematic list o'. cjc + ians . This review should 13 ensure that changes to the control boards are not arbitrary and will fully consider the impact(s) upon the control room. This "Effectiveness Phase" and the "Design Control Plan" appear to be an excellent means of verifying the adequacy of future control room changes. PSC's plans to perform these verification activities and the relatively high fidelity -of the mockup should contribute to the overall valiuity of the DCRDR and future control room improvements. The plans PSChas made for these activities should successfully meet this requirement of NUREG-0737, Supplement 1. Coordination of the DCRDR with Other Improvement Programs PSC appears to recognize the need to coordinate the various emergency response capability programs. The mechanism or management structure PSC has set up to perform the,coordinating and integrating function is the "Improvement Program::Committee." There is an improvement program committee for each program. The person responsible for each program, such as the CRDR Coordinator, is a member of the other improvement program committees. All improvement program committees are coordinated by a designated individual in management. In addition to the DCRDR, the following improvement programs apparently have committees: o Emergency Operating Procedure (EOP) Upgrade o Safety Parameter Display System (SPDS) o Reg. Guide 1.97 Instrumentation o Training. Of these four improvement programs, only the EOP upgrade was discussed at length with respect to the DCRDR. PSC stated that the task analysis will form the basis for the development of the symptom-based EOPs (see the "Function and Task Analysis" section) . The CRDR Coordinator foresees being involved in the EOP upgrade program when the DCRDR winds down since he has familiarity and expertise with. the task analysis effort. • PSC demonstrated in discussions held during the audit that they are committed to ensuring these programs are properly coordinated and that applicable inputs are integrated among the programs. We suggest that the summary report contain a description of how each program was integrated with the DCRDR. SUMKARY AND CONCLUSIONS Throughout the audit we found PSC to be very helpful in providinc information and assistance when needed by the NRC audit. team. We were impressed with the overall level of commitment by PSC to a cun. lish the DCRDk, to ensure the coordination of the. various imprcvercn , pr crars , and to ensure that future control room modifications receive the proper review. We believe that PSC has demonstrated the necessary level of commitment to perform a successful - DCRDR. With the exception of the task analysis and minor portions of other activities, the activities that have been performed and the plans for the upcoming activities were found to be acceptable. In order for an acceptable DCRDR to be completed, we suggest that PSC incorporate the feedback given in this report into its plans and methodologies. Below is a summary of°-the points made in this report on the areas or activities in PSC's DCRDR. o Qualifications and Structure of the DCRDR Team: The personnel PSC has committed to the DCRDR and their level of involvement in the various activities were found to be adequate. o Documentation System: PSC's documentation system is effective in supporting the DCRDR. o Operating Experience Review: We suggest that PSC enlarge the scope of its review of operating history to include incidents or events which may provide additional valuable input to the task analysis walkthrough effort. o Function and Task Analysis: PSC has not performed an adequate analysis of information and control requirements and thus has not met this requirement of t4UREG-0737, Supplement 1. o Control Room Inventory: PSC's mockup is an adequate substitute for the inventory in the verification activities, however, these activities can not be adequately performed without an adequate task analysis. c Control Room Survey: in order for PSC to meet this requirement of NUREG-0737 , Supplement 1, etidence must be provided that: - All applicable "dynamic" criteria have been applied to all controls and displays in the control room, and - All guidelines that have not yet been evaluated are identified :and evaluated. o Assessment Of HEDs: PSC should provide a detailed description of the assessment approach in the summary report. o Selection of Design Improvements: PSC should describe the following areas in det<.il in the summary report: - The approach to selecting HED corrections, and - Each discrepancy found and PSC' s selected corrective action or thc- , rstificttiun for not takine corrective actions. 15 - o Verification That Improvements Will Provide the Necessary Corrections Without introducing New HECIs: PSC 's plans for this effort should meet this requirement of NUREG-0737, Supplement 1. o Coordination of the DCRDR With Other Improvement Programs: PSC should provide a description of how each program was integrated with the DGRDR and what exchanges were made among them. REFERENCES • 1 . Fort St . Vrain Nuclear Generating Station Control Reom Design Review Plan, attachment to Letter front ft L. Brey, PSC, to J.T. Collins , NRC, dated July 13, 1983. 2. NRC Staff Comments on the Fort St. Frain Detailed Control Room Design Review Program Plan, attachment to Memorandum from W. T. Russell , NRC, to G. C. Lainas, NRC, dated November 22, 1983. 3. NUREG-0660, Vol . "NRC Action Plan Developed as a Result of the TMI-2 Accident," US NRC, Washington, D.C. , May 1980, Rev. 1, August 1980. 4. NUREG-0737, "Requirements for Emergency Response Capability," US NRC, Washington, 0./C. , November 1980. 5. NUREG-0737, Supplement 1, "Requirements for Emergency Response Capability," US NRC, Washington, D.C. , December 1982, transmitted to reactor licensees via Generic Letter 82-33, December 17, 1982. 6. NUREG-0700, "Guidelines for Control Room Design Reviews," US NRC, Washington, D.C. , September 1981. 7. NUREG-0801, "Evaluation Criteria for Detailed Control Room Design Reviews," US NRC, Washington, D.C. , October 1982, draft report. EXHIBIT A FORT ST VRAIN DCRDR IN-PROGRESS AUDIT ATTENDEES AT ENTRANCE AND EXIT BRIEFINGS PSC Representatives Don Warembourg, Manager, Nuclear Production ** N.L. Brey, Manager, Nuclear Engineering Division Mike Niehoff, Site Eng. Manager ** D.: Glenn, Nuclear Engineering Division ** Mike Annon, Proto Power ** Mike Maddox, Search Technology * Jack Gahn, QA Manager * Lee Singleton, Supt. , QA Operations * Chuck Fuller, Tech/Admin Manager NRC Representatives Val Pezoldt, Human Factors Engineering Branch * Leo Beltracchi , " " " Dick Ireland, Region IV Tim O'Donoghue, Science Applications, Inc. G.L. Plumlee, Senior Resident Inspector Pete Williams, Division of Safety Technology * Entrance Briefing Only ** Exit Briefing Only i..' . s UNITED STATES ', NUCLEAR REGULATORY COMMISSION a p .�t REGION IV o 7? ..1 1;`of h 611 RYAN PLAZA DRIVE. SUITE 1000 io •0 ARLINGTON,TEXAS 76011 MMs 1904 Docket: 50-267 WELD COUNTY CD�•ia�^^.,.,Pn¢ Mr. 0. R. Lee, Vice President MAY s , Electric Production 2 1 1984 Public Service Company of Colorado P.O. Box 840 GREeLL ,. y Denver, Colorado 80201 . Dear Mr. Lee: We have completed our review of the information you have provided in response to NUREG-0737 "Clarification of TMI Action Plan Requirements" Item II . F.1.(3) "Containment High Range Radiation Monitor." The documentation of our review is contained in the enclosed safety evaluation. We have concluded that the installed monitors fulfill these requirements and consider the technical evaluation of this issue to be complete. Acceptable Technical Specification operability and testing requirements are still required (as stated in earlier correspondence) to close this issue. If you have any questions on this subject, please contact the NRC project manager. Sincerely, �.11. 301.1.1ice , E. H. Johnson, Chief Reactor Project Branch 1 Enclosure: Safety Evatuation cc: Attached List Del_ rr( eta. 41;_31 -41 Fort St. Vrain cc list C. K. Viller Chairman, board of County Commissioners Senior Vice President of Weld County, Colorado Public Service Company Greeley, Colorado 80631 of Colorado P. 0. Box 840 Regional Representative Denver, Colorado 80201 Radiation Programs Environmental Protection Agency James B. Graham, Manager 1860 Lincoln Street Licensing and Regulation Denver, Colorado 80203 East Coast Office' General Atomic Company Don Warembourg 2021 K Street, NW, Suite 709 Nuclear Production Manager Washington, DC 20006 Public Service Company of Colorado P. 0. Box 368 J. K. Fuller, Vice President , Platteville, Colorado 80651 Public Service Company of Colorado Albert J. Hazle, Director P. 0. Box 840 Radiation Control Division Denver, Colorado 80201 Department of Health 4210 East 11th Avenue Denver, Colorado 80220 G. L. Plumlee NRC Senior Resident Inspector Kelly, Stansfield E O'Donnell P. 0. Box 640 Public Service Company Building Platteville, Colorado 80651 Room 900 550 15th Street D€rver, Colorado 60202 Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation U. S. Nuclear -Regulatory Commission Washington, D.C. 20555 SAFETY EVALUATION REPORT FOR TMI ITEM II.F.1(3), NUREG-0737 "CONTAINMENT HIGH-RANGE RADIATION MONITORS" Fort ST. Vrain, Unit No. 1 Public Service Company of Colorado Docket No. 50-267 Position As required in NUREG-0737, II.F.1 (3), in containment, radiation level moni- tors with a maximum range of 107 R/hr shall be installed. A minimum of two such monitors that are physically separated shall be provided. Monitors shall be developed and qualified to function in an accident environment. Introduction Fort St. Vrain, Unit No. 1, (FSV) is a helium, gas-cooled reactor whose core power density and fuel configuration differs from those found in a water- cooled reactor. FSV core power density is lower and the fuel used in the reactor is encapsulated with a multi-layered ceramic coating having a high temperature capability. This coating will delay the release of highly radio- active fission products after a reactor scram. In addition, the Pre-stress- ed Concrete Reactor Vessel (PCRV) has an approximate thickness of nine feet of concrete. Consequently, the post-accident radiation levels in the reactor building should be considerably lower and the environmental conditions should be less severe than those of a water cooled reactor. Thus, modification of the requirements specified for containment monitors in II.F.1(3) (107 R/hr) is appropriate for FSV. Discussion and Conclusion The staff has evaluated the requirements of monitoring the radiation levels of the containment (reactor building) for FSV, Unit No. 1, following an ac- cident, against the criteria of NUREG-0737, Item II.F.1(3). Specifically, the staff evaluated the appropriate instrument range, energy response,redun- dancy requirement and special instrument calibration requirements for the FSV reactor. The source terms and expected radiation levels used by the staff were those provided by the licensee on December 20, 1980 as part of their shielding design review required in Item II.B.2, NUREG-0737. This review determined that the maximum credible dose-rate in the reactor build- ing was less than 2 R/hr. Public Service Company of Colorado has proposed in their updated FSAR for FSV to meet the requirement of II.F.1(3) by using a portable high range area monitor to measure the reactor building radiation levels. The high range monitor will be calibrated annually using standard plant procedures. The exposure rate,:error, as a function of photon energy will be less than 20% over the range of 100kev to 2.5 mev. This area monitor is easily replace- able, since the dose rates are low, making a redundant system unnecessary. The staff concludes that no additional instrumentation is required to meet the intent of NUREG-0737, II.F.1.(3). Subject to the licensee submitting formal technical specification changes for Item II.F.1 (3), the staff con- siders IMI Item II .F.1. (3) , for FSV closed. Dated: w2c _ •: The following NRC personnel contributed to this evaluation; M. Lamastia . !JRR Hello