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| {{Adams | | {{Adams |
| | number = ML20202F264 | | | number = ML20211G473 |
| | issue date = 07/10/1986 | | | issue date = 02/17/1987 |
| | title = SALP 6 Board Rept 50-155/86-01 for Nov 1984 - Mar 1986. Category 1 Rating Maintained in Area of Emergency Preparedness.Regional Insp Activities for Emergency Preparedness Will Be Reduced | | | title = Errata to SALP Rept 50-155/86-01,consisting of App & Corrected Pages 11.12,13,22 & 25 |
| | author name = | | | author name = |
| | author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) | | | author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
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| | license number = | | | license number = |
| | contact person = | | | contact person = |
| | document report number = 50-155-86-01, 50-155-86-1, NUDOCS 8607150126 | | | document report number = 50-155-86-01, 50-155-86-1, NUDOCS 8702250397 |
| | package number = ML20202F249 | | | package number = ML20211G425 |
| | document type = SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE, TEXT-INSPECTION & AUDIT & I&E CIRCULARS | | | document type = SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE, TEXT-INSPECTION & AUDIT & I&E CIRCULARS |
| | page count = 39 | | | page count = 13 |
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| SALP 6 SALP BOARD REPORT U. S. NUCLEAR REGULATORY COMMISSION d'
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| ==REGION III==
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| SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE 50-155/86001
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| ==Inspection Report==
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| Consumers Power Company Name of Licensee Big Rock Point Plant Name of Facility November 1, 1984 through March 31, 1986 Assessment Period
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| i 8607150126 860710 PDR ADOCK 05000155 O PDR
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| O I. INTRODUCTION The Systematic Assessment of Licensee Performance (SALP) program is an integrated NRC staff effort to collect available observations and data on a periodic basis and to evaluate licensee performance based upon this information. SALP is supplemental to normal regulatory processes used to ensure compliance to NRC rules and regulations. SALP is intended to be sufficiently diagnostic to provide a rational basis for allocating NRC resources and to provide meaningful guidance to the licensee's management to promote quality and safety of plant construction and operatio A NRC SALP Board, composed of staff members listed below, met on May 23, 1986, to review the collection of performance observations and data to assess the licensee's performance in accordance with the guidance in NRC Manual Chapter 0516, " Systematic Assessment of Licensee Performance." A summary of the guidance and evaluation criteria is provided in Section II of this repor SALP Board, for Big Rock Point:
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| Name Title J. A. Hind Director, Division of Radiological Safety and Safeguards E. G. Greenman Deputy Director, Division of Reactor Projects W. G. Guldemond Chief, Reactor Projects Branch 2 L. R. Greger Chief, Facilities Radiation Protection Section E. R. Schweibinz Chief, Technical Support Staff M. Schumacher Chief, Radiological Effluents and Chemistry Section B. Snell Chief, Emergency Preparedness Section D. H. Danielson Chief, Material and Process Section R. B. Landsman ProjectManager,ReactorProjects Section 2D T. Rotella Big Rock Point Project Manager, NRR S. Guthrie Senior Resident Inspector D. A. Kers Plant Protection Analyst l
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| II. CRITERIA Licensee nerformance is assessed in selected functional areas, depending upon whetner the facility is in a construction, preoperational, or operating phase. Functional areas normally represent areas significant to nuclear safety and the environmen Some functional areas may not be assessed because of little or no licensee activities, or lack of meaningful
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| observations. Special areas may be added to highlight significant j observation One or more of the following evaluation criteria were used to assess each functional are . Management involvement and control in assuring quality Approach to the resolution of technical issues from a safety standpoint Responsiveness to WRC initiatives Enforcement history Operational and Construction events (including response to, analyses of, and corrective actions for) Staffing (including management)
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| However, the SALP Board is not limited to these criteria and others may have been used where appropriat Based upon the SALP Board assessment each functional area evaluated is classified into one of three performance categorie The definitions of these performance categories are:
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| Category 1: Reduced NRC attention inay be appropriat Licensee management attention and involvement are aggressive and oriented toward nuclear safety; licensee resources are ample and effectively used so that a high level of performance with respect to operaticnal safety and construction quality is being achieve Category 2: NRC attention should be maintained at normal levels. Licensee
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| management attention and involvement are evident and are concerned with nuclear safety; licensee resources are adequate and are reasonably effective so that satisfactory performance with respect to operational safety and construction quality is being achieve Category 3: Both NRC and licensee attention should be increased. Licensee management attention and involvement is acceptable and considers nuclear safety, but weaknesses are evident; licensee resources appear to be strained or not effectively used so that minimally satisfactory performance with respect to operational safety or construction quality is being achieved.
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| III. SUMMARY OF RESULTS The overall regulatory performance of the Big Rock Point Plant has continued at a satisfactory level during the assessment perio Improved performance in the area of Licensing Activities is note However, performance in the areas of Plant Operations and Surveillance and Inservice Testing declined from a Category 1 to a Category 2. Performance in the area of Outages is rated a Category 3 this period. This rating is a reflection of the breakdown of administrative controls over the outage process and resulted in a Severity Level III violation during the middle of the SALP perio July 1, 1983- November 1, 1984-Functional Area October 30, 1984 March 31, 1986 Plant Operations 1 2 Radiological Controls 2 2 Maintenance / Modifications 2 2 Surveillance and Inservice Testing 1 2 Fire Protection 2 2 Emergency Preparedness 1 1 Security 1 1 Outages *
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| 3 Quality Programs and Administrative Controls Affecting Quality 2 2 Licensing Activities 2 1 Training and Qualification Effectiveness *
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| *Not Rated (new functional areas for SALP 6)
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| IV. PERFORMANCE ANALYSIS Plant Operations Analysis Portions of eight routine inspections by the resident inspector reviewed plant operation The inspections included observations of control room operations, reviews of logs, discussions of operability of emergency systems, and reviews of reactor building <
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| and turbine building equipment status. During the evaluation period the following violations were identified: Severity Level IV - Failure to perform required surveillance on the Reactor Depressurization System (RDS) (155/84017).
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| . Severity Level V - Delay in notification to NRC of a plant
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| shutdown required by Technical Specifications (155/84017).
| | ; SALP BOARD REPOR p |
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| By not performing the surveillance when required, the unit entered a Limiting Condition for Operation statement and a unit shutdown was necessary. The licensee also failed to recognize the reporting requirements of 10 CFR 50.72 associated with a forced shutdown and was late in making the required notificatio The Big Rock Operations Department is adequately staffed with licensed and non-licensed individuals who are dedicated to safe and efficient operation of the reactor. Observation of operators in the control room and on tour in the plant indicates they are generally conscientious in both routine and off-normal activitie They make regular use of drawings and procedures to plan and perform evolutions. Control room decorum is adequate, and a cooperative, results-oriented attitude is apparent among the operators and in the operator's dealings with maintenance men, radiation protection technicians, and the engineering staf Operators are well trained. Shift-manning is accomplished without excessive use of overtime and the number of individuals in training and requalification programs appears adequate to meet future need The operations staff performs well during startups and shutdowns of the reactor, refueling operations, and performance of surveillances. During surveillances, operators appear to understand the objective of the test and the impact of their actions on plant equipment. Operators appear capable of dealing with abnormal and emergency situations,' indicating adequate training and a functional understanding of plant I
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| equipment and systems interrelationships. Examples include operator action that prevented reactor scrams on two occasions, including a potential scram on high pressure when the Initial Pressure Regulator (IPR) cover was lowered onto the IPR linkage,
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| and two near scrams on low vacuum during installation of a pipe patch on bypass line piping. In another instance, operators took conservative action by scramming the unit manually on indications of a major steam leak. Additionally, during the period of reactor vessel inventory loss following the incorrect disassembly of VRD305, operators on the refueling deck were quick to identify the cause of the decreasing level and took proper corrective actio Of the six Reactor Protection Systems actuations resulting in a scram signal during the evaluation period, two were attributable in part to operator erro On January 7, 1985, the scram was caused by failure to reset the FRV and on December 7, 1985, the scram was precipitated by operators who earlier had left a valve mispositioned prior to startup. The other four scrams resulted from spurious actuations of the RPS system because of electrical noise affecting the circuitry at low power, a known operating characteristic of little safety significanc In spite of the ability of the operations staff to operate the plant reasonably well the department experienced a series of human errors throughout the assessment period which detracted 3' from the safe operation of the facility. Most were attributable to inattentiveness or lack of thorough attention to detai Operator inattentiveness on two occasions resulted in misposi-tioned control rods, though one instance was influenced by inadequate management direction and cumbersome administrative controls over several available rod withdrawal sequence Inattention to detail and an assumption that other plant personnel or administrative systems would compensate for failure to assume personal responsibility for plant safety were at the root of errors associated with tagging and isolation of components involving work on the recirculation pump, and in a separate incident, the incorrect disassembly of a control rod drive system check valve. Inattention to detail and a willingness to circumvent administrative controls (see Section IV.H) resulted in an incorrect pipe being severed during construction of Alternate Shutdown systems. Additionally, errors resulted in the incorrect tagging of an electrical breaker, and in missed surveillances detailed in Section IV.D of this repor Finally, failure to follow local tagging procedures resulted in the repair of Valve VNS143 without tagging or isolation and is believed to be a factor in a major steam leak and subsequent
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| . scram on December 7, 198 Licensee management, in response to increased frequency of errors, has emphasized attention to detail, counseled individuals, retrained personnel, and implemented revised administrative controls on control rod manipulation Management demonstrates a thorough understanding of the plant's ,
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| operation, reflecting extensive experience with this facilit '
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| Management personnel are often present in the control room area and tour the plant regularl Management direction and control, however, was considered to be deficient in several instance _ ___ -_ __ __ . _ ._ - - _ - _ _
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| While the composition of the operations department staff continued in transition from a group of older operators with many years of plant specific experience to a mixed group with many operators who are relatively new to the plant, management showed a reluctance to compensate by incorporating lessons learned into plant operating procedures. For example, the bypass valve has a history of erratic behavior in automatic operation at low steam flows, resulting in two plant scrams in 1984. Management had not provided specific guidance to operators on when to remove the valve from automatic control, leaving it up to the individual operator's discretion, even though a disparity in theories and practices existed among operators because of differences in experience levels. The Reactor Scram on January 7, 1985 was caused by failure to reset instrument air to the Feedwater Regulating Vaive (FRV) following air system maintenance. Older operators surveyed were aware of the valve's characteristic of failing on loss of air, but the less experienced operators performing the startup were not. Newer operators could have benefited from an expanded component identification program throughout the plant. Finally, the licensee's revised admini-strative controls over control rod movement, "hich employed laminated cards and was implemented as corrective action following two instances of mispositioned control rods, went into effect with insufficient management direction. As such, the card system went unused until repeated requests from the resident inspector prompted management to publish guidance requiring consistent and regular use by operators for all rod motio Some reluctance to respond to NRC initiatives was in evidence throughout the assessment period. Examples include responses to inspector inquiries about operability of the acoustic monitor during plant startup and operation, the need for a second Control Rod Drive Pump to meet the requirements of Appendix R, the need to test the availability of one electrical power source prior to removal of another, and the recommendation to label the contain-ment escape lock operating handles. However, the quality and quantity of communication and cooperation with regulators steadily improved over the 17 months. In the closing months of the evaluation period the licensee demonstrated a willingness to respond positively to NRC initiatives and a concern for safety by operating the Diesel Fire Pump continuously when its starting reliability was in questio The licensee also exhibited several instances where a conservative approach to resolution of a technical issue was chosen. Examples include conservative declarations of inoperability on the RDS system because of a detensioned hanger and on one tube bundle of the emergency condenser based on a barely detectable indication of a primary to secondary leak. The circumstances surrounding the event discussed earlier in this section point to a licensee decision to emphasize production over safety, but is an isolated example not representative of the licensee's approach to technical issues throughout the remainder of the perio . - _ _ . -
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| . Conclusions The licensee is rated category 2 in this area with a declining trend based on increasing frequency of human errors and difficulty in implementing administrative controls over practices and procedures important to plant safet . Board Recommendations To avoid future declines in this functional area licensee management should address problems with administrative controls, particularly as they relate to nutage management, and reduce the frequency of human error in plant operation B. Radiological Controls Analysis Evaluation of this functional area is based on routine assessments by the resident inspector during implementation of the resident inspection program and six inspections by Region III specialist These inspections covered radiation protection, radwaste management, disposal of low-level radioactive waste, chemistry and radiochemistry, and confirmatory measurement One violation and one deviation were identified as follows: Severity Level V - Failure to conduct a quality control program to assure compliance with waste classification and waste characteristic requirements (155/85006). Deviation - Failure to implement the Radiation Safety Plan by the date specified in the licensee's August 19, 1982 supplemental response to the Health Physics Appraisal (155/85003).
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| The violation and the deviation were the results of inadequate procedures; the licensee's corrective actions were timel Responsiveness to NRC initiatives has been generally adequat In response to inspector concerns regarding mask-fit testing of BioPak 60-P respirators, the licensee replaced these respirators with open circuit Self Contained Breathing Apparatuses (SCBAs).
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| Also, inspector concerns identified related to laboratory performance are often acted upon by the end of the inspectio However, the licensee was somewhat slow in correcting an error in a 1984 semiannual effluent report brought to their attention by the inspector, and was also slow to complete an evaluation and request for approval concerning retention of contaminated soil onsite following a break in an underground line to the condensate j storage tank. The contaminated soil issue was closed by an -
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| Environmental Assessment and Findings of No Significant Impact published in the Federal Register (May 5, 1986 - 51FR16596). New i RETS technical specifications and the ODCM were implemented 1 during this assessment perio I
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| Staffing in chemistry and radiation protection appears adequate, with no changes in key supervisory personnel. The relatively small technician staff has recently experienced a high turnover, with six of 12 technicians replaced during this assessment period; however, the inspectors have not observed a significant effect on licensee performance. All of the replacement technicians have completed the specified basic training course, and assigned responsibilities appeared to have been commensurate with the level of trainin Supervisory personnel appear to have a good understanding of their areas of responsibilit Management involvement has been adequate to assure acceptable quality in the functional are There is adequate ALARA program support and involvement by all levels of management. Records are generally complete and well maintained. Procedure adherence has been generally adequate, and management policy encourages worker identification of problems to help with timely correction However, inspectors have noted a significant number of instances which indicate the need for more management attention, including persons not frisking at exit points, radioactive materials stored outside posted areas, contaminated area postings with inadequate or confusing instructions to workers, and area monitor calibra- i tion sources carried through office areas without appropriate restrictions to personnel access to the area. Quality Assurance (QA) involvement in the health physics activities during 1984 was marginal. This shortcoming was exacerbated by the fact that the formal plant surveillance program required by the licensee's Radiation Safety Plan (RSP) had not yet been implemented. In February 1985, NRC inspectors noted that the formal reporting system for minor radiological occurrences required by the RSP had also not been implemented. The recent implementation of these RSP programs should improve overall management involvement in this functional are Although the licensee's approach to the resolution of radiological technical issues has generally been technically sound, thorough, and timely during this assessment period, instances of poor performance have occurred. In late 1984, a policy was implemented which established a routine decontamination program; however, no dedicated decontamination workers were assigne Despite this decontamination program, the licensee has experienced problems in contamination control, especially during outage The addition of two contractor decontamination workers following the 1985 outage resulted in a major improvement in j plant cleanliness. Formerly inaccessible areas are now accessible. The ALARA program has shown improvement during
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| , and since the 1985 outage. The licensee has committed to an
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| , ambitious program of person-rem reduction that will stress job
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| preplanning and new fuel pool cleaning equipment. With regular
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| - use of decontamination personnel the licensee intends to reduce annual exposure by approximately one-third. The licensee is generally conservative in resolution of potential safety and environmental concerns. Relocation of a storm drain release path to the lake, necessitated by high lake level, was
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| | SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE |
| accomplished by routing it in a manner to ensure that releases would be monitored by the discharge canal monitor. The licensee has also performed extensive testing during shutdown to locate the source of a minor primary to secondary leak that developed in the emergency condenser during operation. When the leak source could not be identified, an augmented sampling program was instituted upon restart to ensure that regulatory limits are me Corporate management is involved in the station's effort to develop a method of measuring minor airborne releases via this pathwa Due to continuing fuel cladding problems, radioactive gaseous releases during this assessment period were about a factor of six higher than normal but have remained well below regulatory limits even when operating at full power. Licensee efforts to minimize releases and to eventually eliminate the problem included removal of identified fuel leakers and use of a new design replacement fuel. Release rates since the November restart have been running at about three to four times the normal rate. Liquid radioactive releases were below average for U.S. boiling water reactor The activity in liquid releases has apparently stabilized during this assessment period following several years of gradual declin The solid radioactive waste volumes in 1984 and 1985 were significantly less than in recent years due, in part, to the implementation of a segregation program for dry active waste (DAW). No transportation problems were identified during this assessment perio Personal exposures were about 120 and 270 person-rem in 1984 and 1985, respectively. These exposures are below the station average over the previous five years (approximately 300 person-rem).
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| The licensee performed generally well in confirmatory measurements with 34 agreements in 36 comparisons with Region III during the assessment period. The disagreements were both for iodine collected on a charcoal cartridge, with the licensee's values about 20% lower than the NRC's. Recalibration following a similar disagreement during the previous SALP period did not resolve the difficulty owing, apparently, to differences of activity distribution between the licensee's standard and plant sample The licensee readily agreed to use a correction factor until another recalibration could be accomplishe . Conclusions The licensee is rated Category 2 in this functional are This is the same rating given the previous SALP perio . Board Recommendations Non _
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| C. Maintenance / Modifications Analysis Portions of eight routine inspections by the Resident Inspector reviewed maintenance activities. One violation discussed in Section IV.H, Outages, reflects on the licensee's ability to conduct maintenance work during outages. In addition, two Regionally based inspections'were performed. The inspections included reviews of normal maintenance and modification activities to ensure that approvals were obtained prior to initiating work, activities were accomplished using approved procedures, post maintenance testing was completed prior to returning components or systems to service, and parts and materials were properly certifie In addition, work planning and scheduling was reviewed as well as the effectiveness of administrative controls to ensure proper priority is assigne No violations or deviations note During the evaluation period the licensee interrupted plant operations for nine unscheduled maintenance outage periods ranging from one to 11 days. Three outages were required to repair Reactor Depressurization System (RDS) valves due to the degraded condition of the system preventing successful performance of quarterly surveillance These included one forced shutdown required by Technical Specifications unidentified leak rate limitations. Two outage periods of one day each were required to successfully repair IA-60B, seal leakage to heat exchanger for Reactor Recirculation Pump No. 2. Also, two outages of three and four days each were required to diagnose and correct steam leakage from the reactor vessel head o-rings. One outage period of four days was used to replace a recirculation pump seal, and a one day outage was required to correct steam leaks associated with the plant scram on December 7, 198 Proper planning and outage control was generally evident for the !
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| nine unscheduled outages. Although unplanned, the licensee in the case of the RDS and recirculation pump outages had sufficient warning to plan activities, prepare parts and procedures, and perform other maintenance work that fell within the scope and time limitations of the forced outag Repair to RDS valve top assemblies have become commonplace to the point that the licensee routinely overhauls spare top assemblies. The licensee did not
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| , overhaul the spare recirculation pump seal in advance of the outage and was still rebuilding the seal as the plant was being shutdown to perform the replacement, even though the pump had been idled for two weeks prior to shutdown. The licensee made extensive use of vendor consultants and pump experts from the
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| General Office for the seal replacement, resulting in a refined and useful procedure for rebuilding and installatio Outages for RDS and recirculation pump repairs were well planned and execute Outages to repair IA-60B represented an operational situation that offered little warning and first attempts at repairs were unsuccessful. The reactor vessel o-ring offered no warning prior to failure, but successful repairs were delayed when the problem was misdiagnose Once the decision was made to perform the vessel head removal and ring replacement the physically demanding job was successfully completed with conservative consideration to ALARA and personnel safet Maintenance work (including mechanical, electrical, and instrument / control) at Big Rock Point is performed by generally competent repairmen who exhibit craftsmanship and a general familiarity with the facility and the equipmen The amount of unsuccessful repair attempts resulting in rework is generally small. Repairmen generally are cognizant of procedural require-ments associated with their assigned task, communicate effec-tively with operators and health physics technicians, a'd reflect concern for ALARA consideration While the input repairmen provide to machinery history is often marginal, communication with co-workers and supervisors indicates genuine interest in continued safe and successful operation of the reactor. The mechanic who performs the work, for example, often participates in post maintenance testing. While the retirement of older,
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| ! in Section IV.H, Outages, the maintenance staff demonstrated flexibility and dedication throughout the evaluation period.
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| l The size of the maintenance staff is generally adequate for all periods other than major refueling outages. A gradually increasing backlog of maintenance orders over the period is explained in part by increased emphasis on skills training which over the short term reduces staff size availabilit Like the Operations Department the loss of older experienced personnel due to retirement or other duties has altered composi-tion of the maintenance staff. While the I & C group remained unchanged, in the mechanical maintenance group of 12 men, five were added during the assessment period. Because hiring and promotion is heavily influenced by Labor Relations agreements that emphasize seniority, newly added staff members generally
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| have little or no experience with nuclear powered generating plants in general or Big Rock Point specifically. Altbough the licensee has long recognized the need for maintenance staff training, no training was provided until February 1986, when a regular program of skills training offsite was initiate The skills training is general in nature and is not nuclear plant specifi No nuclear plant system or concepts training is provide I
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| First line supervision in the maintenance department reflects adequate technical skills and managerial competence. During the 1985 outage, the maintenance department overcame the loss of staff experience, inadequate outage planning, and parts procure-ment to accomplish a relatively large number of modifications, repairs, and preventive maintenance task Throughout the evaluation period several recurring problems were not successfully repaired or adequately addressed. Valve M0-7067, Turbine Bypass Isolation Valve, was not declared operable for much of the evaluation period, based on difficulties with the valve operator. Reactor Depressurization System (RDS) valves exhibit inherent design deficiencies that have resulted in three forced shutdowns during the assessment period and a long history of problems dating back to their installatio Management, however, has not placed a high priority on a comprehensive solution and as a result the RDS system was not improved over the period. Problems with the Emergency Diesel Generator (EDG)
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| fuel pump were allowed to continue and a design change to the pump mounting bracket scheduled for completion during the 1985 refueling outage was deleted in an effort to return the plant to an operable status. Shortly thereafter the pump failed again, placing the EDG in an action statement for the generator's Limiting Condition for Operation. Finally, the licensee made a commitment to verify, prior to startup from the 1985 outage, i
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| Limitorque Switch settings on 18 Limitorque Valves the licensee considered important to safety. As of this date only 15 have
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| been checked. The torque settings for valve M0-7067 have been reset on three different occasions, indicating a lack of decisive
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| direction on problems with Limitorques Operators that goes back to September, 1984, as was addressed in SALP SALP 5 expressed concern that the Prevention Maintenance (PM)
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| program may be inadequate to address aging equipment. At the end of this assessment period the PM program continues to be reactive in nature, relying heavily on visual inspections that do not involve disassembly or physical measurements, and on the obser-vations of operators monitoring noticeable changes in component l operating characteristics. There continues to be no program to analyze for trends in failures or any other measurable parameter other than pump capacity on certain pumps. The licensee has not responded to NRC initiatives to upgrade the PM program to incor-porate vendor recommendations and industry experience. The plant continues to rely on surveillance tests to identify problems that may be in some advanced stage of development due to aging equipmen At the close of the assessment period the licensee assigned an engineer to develop a program of predictive analysis focusing on vibration and lubricating oil analysi Evidence of problems associated with aging of plant equipment during the assessment period included: Several examples of end of service life for solenoid valves on the turbine stop valve, diesel fire pump (DFP), and the exhaust ventilation downstream isolation valv _ _ _ . ._ _
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| . Deterioration of fuel delivery system on the DF Failure of several motor operated valves to operate on demand, including the turbine bypass isolation valve, the recirculation pump suction valve, turbine stop valve, and the shutdown system reactor isolation valv A regionally based inspection performed in response to a
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| , declining performance trend identified in SALP 5, pointed out
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| ! weaknesses in the PM program including failure to update the progri.a based on plant experience, inadequate root cause analysis, and inadequate consideration of the generic implications of maintenance actio The report recommended a more comprehensive method of evaluating potential end-of-service-life failure Another regionally based inspection assessed the adequacy of the licensee's response to Generic Letter 83-28 and determined the licensee was generally meeting the requirements in the areas of vendor interface and post maintenance testing. The report noted the lengthy delays in implementation of the vendor interface program and inadequacy of post maintenance testing instructions and documentatio For the last half of the assessment period the site engineering group has functioned under the Maintenance Department, an organizational move intended to improve coordination between the engineering and maintenance functions. The engineering group seems slightly overburdened, a situation compounded by lack of consistent prioritization of project assignments. Engineers were regularly redirected from one project to another based on manage-ment's sense of urgency over a given engineering project. The licensee, at the end of the assessment period, performed an inventory of all engineering projects and has devised a system of consistent prioritization which should alleviate this proble The quality of modification packages prepared by the site
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| , engineering group is consistently high, reflecting the group's extensive familiarity with the facility and a genuine interest in
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| the safe and successful operation of the plant. Sound engineering judgement that stresses safety and reliability is evident. Some members of the staff do not consistently identify and incorporate into their proposals and designs the quality requirements derived from the various codes and regulations, relying instead on review
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| ! The deficiencies in the Nuclear Operations Department Standards (N0DS) discussed in Section IV.I contribute to this proble A communication barrier exists between members of the engineering
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| and mechanical maintenance staffs, and the knowledge of mechanics
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| ; is not routinely conveyed to engineers or factored into design decisions. A notable example is information gathered by mechanics during disassembly and cleaning of RDS valve top assemblies which never made its way to the engineer in charge of the projec This resulted in the repeated failure of the RDS valve f
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| While licensee management is generally informative and cooperative with NRC inspectors, there is only a marginal level of respon-siveness to NRC initiatives displayed. Compliance with regulatory requirements is generally adequate, but mediocrity or deficiencies in performance or programs is tolerated and often justified by citing budgetary and manpower constraints. Management action in the areas of preventive maintenance, mechanical training upgrading, and resolution of long standing engineering projects is marginal. Management's lack of effective control of the outage process was a major factor in the events during the 1985 outage discussed in Section IV.H. The reorganization of both the maintenance and engineering functions under one Superintendent appears to be too much activity for any one individual to effectively manage, contributing in part to the licensee's commencement of the 1985 outage with incomplete engineering projects, inadequate scheduling of maintenance activities, and deficient material procurement to support planned wor . Conclusions The board rates the licensee Category 2 with a declining trend based primarily on insufficient management control over the maintenance proces . Board Recommendations The board notes that this is the second consecutive assessment period of declining performance and special management attention is needed to offset the effects of aging equipmen .
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| D. Surveillance and Inservice Testing Analysis During this evaluation period the resident inspector regularly observed licensee performance in this area. These inspections included observations of technical specifications required surveillance testing to verify adequate procedures were used, that instruments were calibrated, and that test results conformed with technical specifications and procedure requirements. In addition, all or part of four regional inspections were conducted in this area. These inspections reviewed startup core perfor-mance, Containment Integrated Leak Rate Tests, intergranular stress corrosion cracking, and inservice testin Big Rock Point uses a manual tracking system to schedule performance of operational surveillance of mechanical, electrical, and Instrumentation and Control (I & C) components and system Each surveillance procedure is sponsored by a knowledgeable individual, and the mechanism exists for revision to the procedure based on performance experienc Surveillances are generally taken seriously by those performing the test and not run to simply satisfy a requirement. Two surveillance tests were overlooked during the evaluation period, including daily
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| control rod drive exercises and test of fire detectors in the recirculating pump roo Cumbersome administrative controls over fire detector tests contributed to the pump room detector
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| omission.
| | 50-155/86001 l Inspection Report |
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| During the evaluation period, one unresolved item resulted from a concern over the frequent lack of detail in instructions and documentation of post maintenance testing when work orders and equipment outage requests are used to meet the post maintenance testing requirements of Generic Letter 83-28, Sections 3.1 and An inspection reviewed the licensee's Inservice Inspection (ISI)
| | l Consumers Power Company i |
| program after the corporate ISI group was disbanded in favor of inder endent program administration at each plant. The inspection )
| | i Name of Licensee i |
| | ! Bic Rock Point Plant l' Fame of Facility 1 , |
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| reviewed the Big Rock 1985 ISI Examination Program Plan and the licensee's outage plan and found them to be acceptabl Observations of ISI activities shows plant personnel have an adequate understanding of work practices and adhere to procedures that are generally well define Records are generally complete, and indicate that equipment and material certifications are kept curren This inspection also reviewed the licensee's inspection program to detect intergranular stress corrosion cracking (IGSCC) in i
| | November 1, 1984 through March 31, 1986 |
| large diameter recirculation system piping to verify that the actions set forth in Generic Letter 84-11 were performed. The inspection determined the acceptability of inspection procedures and techniques, documentation, and examiner qualification In reviewing the licensee's containment integrated leak rate test (CILRT), the inspector noted that the activity was adequately staffed with knowledgeable individuals experienced in the Big Rock unit. No specific training of the participants had preceded the event and the licensee's familiarity with Type A testing requirements was wea Licensee management involvement in supplemental verification testing was considered marginally acceptable as evidenced by efforts to complete the Type A test ,
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| | I Assessment Period |
| before acceptable supplemental verification test data was '
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| obtaine In the area of startup and surveillance testing programs subsequent to the refueling outage, the inspector concluded that licensee personnel appeared to understand technical issues and had a genuine interest in plant operations, providing timely and thorough responses to inspector identified concerns. Procedures appeared to be well written and employed a technically sound methodology.
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| l Conclusions The licensee is rated Category 2 in this area. This is a decline in performance from the last assessment period based primarily on the missed surveillances.
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| i . Board Recommendations None.
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| ; E. Fire Protection Analysis I
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| i During this assessment period the resident inspector routinely observed licensee activities in the fire protection functional area, including routine housekeeping. One special inspection was
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| ; conducted by Region III personnel to assess the licensee's compliance with 10 CFR 50, Appendix R, close out previously identified open items, and verify compliance with routine fire
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| protection program requirement The report has been delayed
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| while the staff completes the review of Big Rock Point's '
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| compliance with Appendix R, and will be addressed in SALP '
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| The licensee's ability to respond to fire alarms both in preplanned drills and actual alarms was observed on several occasions with satisfactory results that indicated the effective-ness of fire brigade member training and fire response procedure Licensee personnel received hands-on fire training and training in use of the self contained breathing apparatus. Licensee
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| ; personnel are generally knowledgeable about fire preventio '
| | t i 8702250397 870217 |
| Housekeeping has improved from the previous assessment perio Housekeeping during plant operation is generally of high quality and accessible plant areas of the facility are routinely police Maintenance workers generally clean up after their job is i complete. Cleaning lockers and assigned areas are part of a
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| | PDR ADOCK 05000155 G PDR i |
| housekeeping system that is incorporated into the daily routin Supervisors are regularly in the plant and monitor cleanliness levels, taking action as appropriate. Housekeeping during
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| extended outage periods, as it relates to both fire protection and contamination control, declines noticeably from periods of normal operation. While plant appearance deteriorates during extended outages, post-outage cleanup is generally prompt and thorough and reflects management involvemen During NRR visits to the plant, the staff was impressed with the clean, well ordered appearance of the plant. Even during construction of the Alternate Shutdown Building, cc struction materials and supplies were well controlled. The Control Room appeared very well run and well organized in terms of reference materials and drawing '
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| Throughout the evaluation period the licensee experienced difficulties with the dependability of the diesel fire pump (DFP), including end of service life for a solenoid coil on the fuel supply shutoff valve, sluggish behavior and slow start times that required cleaning in the fuel delivery system, and a leaking fuel filte The greatest cause for concern about the
| | , r-,,, - , , .~.,,,---.,,,_,.,-,_-,-,,~-,,n.--,-- . , - , .,,,cnnnn.---,n,--,n.,,n.,. ,, ,,--,,~,-ur,,n ,,., = . . - , - . . , - |
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| DFP reliability arose during February, 1986, when excessively
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| long start times and erratic starting behavior were corrected by further cleanings, tightening, and adjusting. Efforts to diagnose and correct problems were hampered by the age of the engine and the unavailability of parts and diagnostic instrument These facts combined with a shortage of vendor representatives experienced on older engines make future repairs unlikely. These factors, combined with the DFP's Core Spray function, make replacement of the DFP a high priority for the 1986 outag The licensee has committed to replace the engine at that tim . Conclusions The licensee is rated Category 2 in this are There is improvement noted in housekeepin . Board Recommendations Non F. Emergency Preparedness Analysis i Three inspections were conducted during the assessment period to evaluate the licensee's performance with regard to emergency preparedness. These included two routine inspections of the emergency preparedness program and observation of the licensee's annual emergency preparedness exercise. Two violations were identified as follows: Severity Level V - Failure to evaluate the adequacy of interfaces with State and local governments as part of the annual audit as required by 10 CFR 50.54(t) (155/84014). Severity Level V - Failure to conduct Health Physics drills in 1983 as required by the Site Emergency Plan (155/84014).
| | Big Rock Point Plant |
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| The above violations were the result of isolated administrative breakdowns in the emergency preparedness program and not indica-tive of any major programmatic problem. In both cases the ;
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| licensee took prompt corrective actions to resolve the violations and ensure that they would not reoccu Management involvement in assuring quality is evidenced by the fact that corrective actions are effective as indicated by the lack of repetition of identified weaknesses. Management support is also shown through the significant corporate assistance in the training program and in the planning and conducting of exercise During the emergency preparedness exercise, licensee management demonstrated an above average command and control capability and were effective in carrying out their assigned emergency responsibilitie l
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| The licensee continues to be responsive to NRC concern Violations and weaknesses that are identified are almost always resolved in a timely manner and demonstrate technically sound and thorough approache This is evidenced by the fact that few issues of concern are identified by the NRC, and those that are have generally been resolved by the next inspectio Staffing of key emergency response positions has been adequate with the authorities and responsibilities of personnel identifie The licensee has a Senior Nuclear Emergency Planning Coordinator position at the site, which has been generally adequate to main-tain the daily emergency program activities at an acceptable level of performance. Knowledge and capability of personnel to carry out their assigned emergency response duties and responsi-bilities was demonstrated during both the annual emergency preparedness exercise and through walkthroughs of personnel during the routine inspections. The licensee's performance in these areas is indicative of an effective training program that has adequately prepared personnel to carry out their emergency response assignments. Examination of the training program and observation of several training sessions during the last routine inspection determined that the program was sufficiently thorough and well conducte However, several events during the assessment period indicated awkwardness with interpretation of reporting requirements and emergency event classification. An example of this was the notification to NRC Headquarters on May 25, 1985 of Unit shutdown, which did not advise of the declaration of the Unusual Even During these events the licensee's capability to interpret reporting requirements and classify the events was less than the level of performance demonstrated during drills, exercises, and inspection walkthrough . Conclusions The licensee is rated Category 1 in this area. The licensee was rated a Category 1 in this area in the last two SALP periods which reflects the continued effectiveness of the emergency preparedness progra . Board Recommendations Non G. Security Analysis Two inspections were conducted by region based inspectors during this assessment period. The resident inspector also conducted periodic observations of security activities. No violations were noted during the inspection effort l | | Meeting Summary The findings and conclusions of the SALP Board are documented in Inspection Report No. 50-155/86001. They were discussed with the licensee on July 21, 1986, at the Region III office in Glen Ellyn, Illinois. The licensee's regulatory performance was presented in each functional area. Overall performance and performance in each functional area was found to be acceptabl The performance rating improved in the area of Licensing Activitie Performance in the areas of Plant Operations and Surveillance and Inservice Testing declined based on increased frequency of personnel error and problems encountered in implementing administrative controls for the Plant Operations area, and missed surveillances for the Surveillance and Inservice Testing are While the performance rating in the new area of Outages was given a Category 3 based on the Severity Level III violation received during the middle of the SALP period, there has been no opportunity to evaluate the effectiveness of your corrective measures. The Emergency Preparedness and Security areas continued to have a high level of performanc While this meeting was primarily a discussion between the licensee and the NRC, it was open to members of the public as observer The following licensee and NRC personnel were in attendance on July 21, 198 Consumers Power J. Reynolds, Executive Vice President F. W. Buckman, Vice President, Nuclear Operations G. B. Slade, Executive Director, Nuclear Assurance K. W. Berry, Director, Nuclear Licensing D. Hoffman, Plant Superintendent R. R. Frisch, Senior Licensing Analyst T. C. Bordine, Staff Engineer B. Alexander, Technical Engineer U.S. Nuclear Regulatory Commission A. B. Davis, Deputy Regional Administratcr E. G. Greenman, Deputy Director, Division of Reactor Projects D. C. Boyd, Chief, Reactor Projects Section 20 S. Guthrie, Senior Resident Inspector, Big Rock Point R. B. Landsman, Project Manager, Section 2D J. Bauer, Technical Staff NRC Headquarters T. S. Rotella, NRR Project Manager J. A. Zwolinski, Director, BWR Project Directorate N . |
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| | ERRATA SHEET |
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| Several allegations pertaining to alleged deficiencies with the licensee's security program were received from a member of the public during this evaluation period. The investigation and resolution of the allegations have extended beyond the close of this evaluation period and will be addressed in a future inspection repor The licensee has been generally responsive to resolving NRC concerns. An inspection conducted early in the assessment period (November 1984) identified the need for revision of the security plan and some supporting implementing procedure The most significant concern pertained to training methods for newly hired security force officer These concerns did not constitute violations or enforcement issues and were generally administrative in nature. However, they were indicative of security management's need to more closely monitor the administrative aspects of the security program. All of the concerns were reviewed during a February 1986 inspection, and the licensee's actions were considered adequate to resolve the concerns. The site and corporate security staff have provided timely and sound technical solutions to inspection finding The February 1986 inspection noted that the morale of the security force was low but had not deteriorated to the point where job performance was affected. The primary cause for the morale
| | Facility: Big Rock Point SALP Report N /86001 Page Line Now Reads Should Read 11 39-43 The licensee . . . to Delete shutdow Basis for Change: Additional information provided by licensee subsequent to SALP issuanc no training was provided with the exception of until February 1986, some I&C classes and certain skill training, no training was provided during this SALP period until February 1986 Basis for C(a ge: The phrase incorrectly implied that training was never provided and should have stated only that during most of the SALP it wasn't. |
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| concern was attributed to long-term labor relation concerns
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| ; beyond the immediate control of the license Licensee management
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| : was aware of the concern and was addressing the issue, within existing labor relation constraints. Deterioration of certain security equipment was also noted and the licensee committed to resolve the issue in a timely manner. The licensee needs to | |
| , continue to be sensitive to required maintenance for aging security equipmen Only one security event was reported during the assessment perio The event pertained to degradation of a vital area barrier and did not constitute an enforcement issu Training and performance of the security force continued to be maintained at a high and consistent level during this assessment period as evident by the excellent enforcement history and lack of reportable events caused by personnel error. Supervision of day-to-day operations appears stron Corporate security support appears adequate. Licensing issues are responded to in a timely manner and analysis of such issues are generally thorough and technically sound. Inspection results are closely monitored by the corporate security office and the corporate office responds in a timely manner to help resolve 1 inspection findings and concerns. Audit functions by the corporate security office appear adequat '
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| . Conclusions The licensee is rated Category 1 in this area based on j demonstrated good performance by the uniformed security force
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| members and no violations being cited during this assessment period. In spite of that the trend is declining based on the | | 13 26-27 As of this date only Delete 15 have been checked Basis for Change: Additional information provided by licensee subsequent to SALP issuanc Forced retirement of Untimely retirement of i |
| | several older key several older key members members of the licensee of the licensee staff, staff which was honored by |
| | ; licensee management Basis for Change: The phrase was not meant to mean that the employees were |
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| aging security equipmen and continued low morale of the guard
| | I forced out, only that they were encourage , It is noted however that |
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| forc . Board Recommendations Non Outages
| | * some relief in the form of additional QA personnel i |
| , Analysis The Resident Inspector performed routine inspections during outage periods and one inspection by a Regional Inspector reviewed refueling activities. These inspections included observation of maintenance activities including administrative requirements, review of planning activities, refueling activities, plant modifications, and post outage testing. One violation was issued as follows: | | : from the Palisades plant |
| Severity Level III - this violation combined in the aggregate seven identified violations stemming from three separate examples during the 1985 outage of supervisory personnel, repairmen, and operators circumventing or ignoring administrative requirements and not exercising sufficient care and attention to detail to ensure plant safety. Contributing to the situation was the lack
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| ) of component identification throughout the facility, the absence 1 of a single point supervisory contact to direct the activities of i travel repair crews, inadequate management involvement in i directing maintenance activities during the outage, and evidence of a lackadaisical attitude on the part of certain operators
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| toward adherence to procedural requirement During the assessment period the licensee conducted one refueling
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| outage. Originally scheduled for 53 days, the outage was extended 10 days due to delays associated with repairs to feedwater and poison system valves, turbine alignment troubles, and the dis-assembly of incorrect valves which was the subject of the violation noted above. Despite the delays, a significant number of major outage activities were successfully completed, including ISI/IGSCC inspections, electrical equipment environ-mental qualifications modifications, and installation of the alternate shutdown pane The licensee completed 1100 main-tenance orders, eight facility changes and 18 specification field change !
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| Operations Department personnel performed fuel handling operations for the 1985 refueling outage. Fuel handling was safely conducted by adequately trained individuals in accordance with approved procedural requirements. Staffing on both the reactor deck and in the control room was adequate, and communi-cation between the two areas was effective. Management involve-ment in refueling activities was evident. Tool control and status board maintenance was adequate. Licensee responsiveness to NRC initiative was evident by their prompt action to correct procedural deficiencies in data recording and in relocation of bagged equipment that had obstructed access to the. refueling deck status boar During the 1985 refueling outage several incidents occurred which demonstrated inadequate management control over the outage process. The incidents involved:
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| * Repeated examples of contractors and licensee travel crew j personnel, not normally assigned to Big Rock Point, performing work on the wrong component or system, pointing i to inadequate control over the activities of travel crews and contractor * Repeated examples of supervisors, maintenance, operations, and engineering personnel, and travel crew personnel, circumventing or failing to adhere to administrative requirements, particularly those related to component tagging and isolatio * Repeated examples by individuals, throughout the organization, of inattention to detail and failure to exercise sufficient care in performance of outage related work to ensure plant safet Several factors contributed to the breakdown in the outage
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| management process:
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| * Throughout the facility, components, valves, and syste;ns identification was generally inadequate, with many compo-nents unlabeled. The licensee had not acted upon earlier requests from the Resident Inspector to improve component identification and discounted warnings on the potential for mishap * Forced retirement of several older key members of the licensee staff, including the Operations Superintendent, the coordinator of the ISI program, an experienced Shift Supervisor, and a Maintenance Supervisor who in the past had acted as a coordinator and single contact point for control of travel crew personnel. The impact of the loss
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| of these individuals two months prior to commencement of
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| the outage was exacerbated by a major reorganization of the plant staff with reassignments of functional depart-ments, creation of new departments, and redistribution of duties within the Maintenance and Operations Departments immediately prior to the outage.
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| * In the absence of a single point contact to direct and
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| ; coordinate the activities of travel crew personnel, manage-l ment involvement in directing maintenance work was j inadequate.
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| * Training provided for travel crew and local licer.see
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| personnel on tagging and isolation was inadequat l * Outage planning, including parts procurement and job sequencing of specific work activities was inadequat Design work on many facility changes was incomplete at
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| ; outage commencemen * Licensee travel crews were inadequately supervised and did not display the same level of concern for reactor safety
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| normally in evidence among Big Rock personne * Work crews assigned a particular task often were comprised entirely of travel crew members without the guidance and experience of Big Rock employee ! * Travel crew supervisors invested too little time and effort
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| ; in inspecting and planning a specific job activity and in
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| l * A lackadaisical attitude on the part of certain personnel i toward attention to detail was a major contributing factor l in the events.
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| I' The licensee has enjoyed decades of safe and successful reactor operation resulting primarily from the professional attitude i
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| displayed by talented and experienced individuals. The plant's ,
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| limited staff and small physical size makes the outage process a J manageable activity. The events of the 1985 outage appear to
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| have impressed upon the licensee the need to aggressively manage outages. The corrective actions in response to the 1985 events have been comprehensive and include:
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| l * An expanded component identification progra !
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| * A photograph book of the plant to aid in job planning.
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| * Counseling and disciplinary action for personnel involved in the problem l * Expanded training for Big Rock and travel crew personne !
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| * Procedure upgrading for valve preventive maintenanc * Controls on future outage activities of travel crews, including single point contact and prejob plannin In addition, the licensee has created new groups to handle material procurement and control and outage planning and schedulin The licensee's responses to NRC concerns in response to the incidents showed a desire to communicate and cooperate. The Plant Superintendent made separate visits to regional and NRR management to present licensee programs to correct programmatic and management deficiencie Prior to the close of the evaluation period no outage activities were conducted that would permit evaluation of the licensee's corrective actio . Conclusions The licensee is rated Category 3 in this are . Board Recommendations The licensee should aggressively implement the corrective actions noted previousl ,
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| I. Quvlity Programs and Administrative Control Affecting Quality Analysis Throughout the rating period the Resident Inspector routinely reviewed the activities of the Quality Assurance (QA) and Quality Control (QC) groups. This included administrative controls for maintenance and operations as well as deviation reports and quality control department involvement in accordance with the QA Plan. In addition, this functional area was examined by the region addressing the adequacy of site QA staffing levels and qualifications in light of increased work load and the impact of Nuclear Operation Department of Standards (NODS) deletion at the facilit The QC activities associated with disposal of low level radioactive waste under 10 CFR 20 and 10 CFR 61 were reviewed in Inspection Report 155/85006(DRSS), resulting in one violation discussed in section IV- The violation resulted from the inadequacy of the licensee's procedure governing the shipment
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| & of radioactive materials to provide for determination of the correct waste classification, although worksheets used to classify the three waste shipments made since the regulation became effective resulted in the correct waste classificatio The site QA and QC staffs are comprised of a generally adequate number of qualified individuals with site experience who demon-strate a high degree of professional conduct and integrit .- . . _ . - - - . -.- -
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| During the evaluation period there was eviJence that the site QA staff was in danger of becoming overburdened by assignment of
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| several functions formerly performed by the corporate QA grou Those added duties were subsequently completed or reassigned elsewhere and the site staff appears adequate for the remaining workloa The site QA staff communicates effectively with plant management and is persistent in pressing for management action to resolve audit finding The Plant Review Committee (PRC)
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| considers the quality aspects of technical and safety issue In turn, plant management generally demonstrates their regard for the significance of findings and comments from the QA staf Site QC inspectors are generally thorough and conscientious and draw heavily on their plant experience. Both the QA and QC site i staff are responsive to NRC initiatives and inquirie Licensee corporate management detracted from the effectiveness of Programs and Administrative controls affecting quality. Examples include: Licensee corporate management, by transferring to the site
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| staff several significant Quality Assurance functions with-
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| ; out a corresponding increase in available site resources, placed a burden on the staff which resulted in QA reviews that were less comprehensive, withdrawal of commitments to support audit activities off site, and a virtual elimir.ation of time available to auditors to review and observe activi-ties in the plan Some QA functions were performed by QC inspectors. The reluctance of corporate management to respond to the concerns of the site QA Superintendent in i this regard and their poor response to NRC initiatives to address the issue was noted.
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| f Licensee Corporate management deleted entirely fifteen N0DS, i the document in which the licensee staff can theoretically j be assured of finding all applicable code and regulatory requirements compiled in one location. The N0DS are the j means by which the licensee's Quality Assurance Program
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| ! Description for Operational Nuclear Power Plants (Topical
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| ! Report CPC-2A) is implemented, and results from a commitment j made in the licensee's Regulatory Performance Improvement Program submitted in response to a March 9, 1981 Confirmatory i 4 Order. Wholesale deletion of the N0DS without a review to
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| ! insure all of the quality requirements contained therein
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| were already addressed in existing administrative procedures resulted in a period when the quality requirements were not
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| ; available to the N0DS user. Inspectors identified at least two examples of cancelled N0DS being referenced in other procedure I
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| 25 i
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| .-, , - . - - - - , - - - ,.,r-__ - .
| | ' was provided in September |
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| . The findings of the licensee's team that N0DS development was incomplete and that an inadequate review and approval process allowed the issuance of N0DS with a significant magnitude of deficiencies relative to CPC-2A basis documents went unacted upon by management.
| | 198 t |
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| During the evaluation period the licensee designed and implemented a program to reduce QA involvement with reviews of procedures in departments where there was long term evidence of high levels of quality performance. The program was implemented late in the period with the licensee's stated goal of redirecting auditor resources into areas of poorer performanc . Conclusions The licensee is rated a Category 2 in this area. The exemplary level of performance by the site QA and QC staff is offset by our concerns with the actions of corporate managemen . Board Recommendations Non Licensing Activities Analysis Methodology
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| The basis of this appraisal was the licensee's performance in support of licensing actions that were either completed i
| | B4 sis for Change: Additianal information provided by licensee subsequent to i SALP issuance. |
| or active during the current rating perio These actions, consisting of license amendment requests, exemption requests, relief requests, responses to generic letters, TMI items, LER's, and other actions, are summarized below:
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| (1) Amendment Requests Technical Specifications (TS) Defining Operability for Safety Systems Containment Pressure and Water Level Monitor TS Reporting Requirements TS
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| ; TS Change Section 6 - Plant Staff Reorganization TS Change for Surveillance Frequencies
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| ; Control Rod Testing Frequency Incorporation of Byproduct License
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| Cycle 21 Reload Il Fuel TS Change Package Administrative TS Gamma Monitor Calibration Frequency Control Rod Withdrawal Rate Limit TS
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| Post Maintenance Testing TS Change - Item 3. PRC Approval Method TS CRD Performance Testing Frequency TS Auto-Isolation (CV-4049) TS Stack Gas Monitoring System TS Organizational TS Appendix I TS Implementation Review Administrative TS Changes Related to RETS Integrated Program Plan (ILS)
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| Appendix "R" Alternate Shutdown System TS (2) Exemption Requests ATWS Recirculation Pump Trip Containment Airlocks Reporting Requirements - Spurious RPS Actuations Fire Protection Equipment Environmental Qualification High Point Coolant System Vents (3) Relief Requests In-Service Testing In-Service Inspection (4) TMI Items I.C.1, Emergency Operating Procedures I.D.1, Detailed Control Room Design Review I.D.2, Safety Parameter Display System II.B.1, Reactor Coolant System Vents II.D.1, RV and SV Testing II.F.1, Accident Monitoring II.F.2.3, Inadequate Core Cooling Instrumentation III.A.1.2, Emergency Response Facilities III.A.2.2, Meteorological Data Upgrade (5) Other Licensing Actions Control of Heavy Loads BWR Pipe Cracking Salem ATWS Follow-up Electrical Equipment Qualifiction Systematic Evaluation Topics Fire Protection Modifications Diesel Generator Reliability Retention of Contaminated Soil Onsite During the SALP period, 67 licensing actions were completed which consisted of 45 plant-specific actions, and 22 multi-plant actions including nine TMI (NUREG-0737) action __ -___ _______ . _ _ _
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| A very important licensing activity completed during the review period was the formalization of the Big Rock Point Integrated Assessment. License Amendment No. 82, " Plan for the Integrated Assessment," issued February 12, 1986, incorporates the requirement to adhere to the " Plan," as documented in License Condition (7) of Big Rock Point Facility Operating License DPR-6. This achievement is noteworthy as Big Rock Point is one of the industry leaders in terms of long-term program implementatio In addition to these licensing activities the project manager and other members of NRR participated in an in progress audit of the licensee's Detailed Control Room Design Review process as well as 10 CFR Part 50, Appendix R related modifications, b. Management Involvement and Control in Assuring Quality Licensing activities for Big Rock Point show consistent evidence of prior planning and assignment of priorities and decision making is almost always done at a level that ensures adequate management review. The cornerstone of the licensee's efforts in this area is the Big Rock Point Integrated Assessment (termed the Plan). The licensee adopted this integrated approach to licensing issues in early 1983. Much of the initial assessment was completed during the last evaluation period; however, the incorpora-tion of the Plan was completed during this evaluation perio As part of an on going process, the licensee makes safety judgements based on the use of the Big Rock Point Proba-bilistic Risk Assessment as well as standard safety assess-ment methods to ensure that plant safety is optimized in a cost-effective manner. The Plan governs the implementation of significant facility change As presented above, there have been a significant number
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| ; of licensing actions processed, and for the most part, the majority were completed requiring little or no additional information or meetings. Adequate management control was not exercised, however, in the handling of the Reactor Depressurization System (RDS) Valve Testing Technical Specification Change Request to reduce surveillance testing frequency. The request showed a lack of prior planning and the technical evaluation was not thorough. This RDS issue has been ranked by the licensee as the most important current facility project as described in Integrated Plan Update No. 4. NRR agrees with the licensee's ranking and believes a continued strong management involvement for assuring quality on this project is neede An area in which Big Rock needs to focus more attention is in their safety evaluations generated to support submittals to NRR involving proposed license amendment Examples of safety evaluations which we found to be less than adequate
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| | - Maintenance / Modifications |
| | - Analysis Portions of eight routine inspections by the Resident In ector reviewed maintenance activities. One violation discuss d in Section IV.H Outages, reflects on the licensee's abi ty to conduct maintenance work during outages. In additio , two Regionally based inspections were performed. The i spections included reviews of normal maintenance and modiff tion activities j to ensure that approvals were obtained prior to itiating work, activities were accomplished using approved to edures, post maintenance testing was completed prior t e rning components or systems to service, and parts and ma were properly certified. In addition, work plannin cheduling was reviewed as well as the effectiveness ministrative controls toensureproperpriorityisassigneg violations or daviations note During the evaluation period the 44, see interrupted plant |
| | ! operations for nine unscheduled tenance outage periods ranging from one to 11 days, outages were required to repair Reactor Depressurizat stem (RDS) valves due to the |
| | ! degraded condition of the preventing successful performance |
| | .. of quarterly surveillance ese included one forced shutdown i required by Technical ecif cations unidentified leak rate |
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| | limitations. Two ou (p iods of one day each were required to successfully repa K DB, seal leakage to heat exchanger |
| | . for Reactor Recir 1 Pump No. 2. Also, two outages of three and four days eac required to diagnose and correct steam leakage from the reac r vessel head o-rings. One outage period of four days was us to replace a recirculation pump seal, and i |
| | a one day outage w required to correct steam leaks associated with the plant sc m un December 7,198 Proper plannin and outage control was generally evident for the |
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| were the application for the incorporation of the byproduct license and the application related to the corporate reorganization. Both applications required extensive NRC efforts to evaluate the impact of the proposed change Also, the depth of explanation of the no significant hazards consideration (NSHC) determinations could be improved. It should be noted that the applications presented above were evaluated during the first half of the evaluation period; and we have noted improvement over the past yea During the last half of the evaluation period, the licensee's evaluations have been well stated, understandable, and
| | nine unschedu ed outages. Although unplanned, the licensee in |
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| applications received have been timely, thorough, and showed decision making consistently at a level that ensures
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| adequate management revie We recognize the strong improving trend; however, Big Rock i must be keenly aware of their unique plant design and as such should strive to fully present complete information to the staf The key point being that the audience to which Big Rock is presenting their SEs, in some cases, is not as familiar with plant-specific design features unique .
| | the case of he RDS and recirculation pump outages had sufficient 1 warning to lan activities, prepare parts and procedures, and perform o er maintenance work that fell within the scope and time li tations of the forced outag Repair to RDS valve top |
| i to Big Rock, and therefore, a conscious effort should be j made to present more information to better understand a
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| c. Approach to Resolution of Technical Issues from a Safety
| | assemb es have become commonplace to the point that the licensee |
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| | routi ely overhauls spare top assemblies. The licensee did not ove aul the spare recirculation pump seal in advance of the ou ge and was still rebuilding the seal as the plant was being |
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| | l utdown to perform the replacement, even though the pump had been died for two weeks prior to shutdown. The licensee made extensive use of vendor consultants and pump experts from the |
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| The licensee generally demonstrates understanding of the technical issues involved in licensing actions and proposes technically sound, thorough, and timely resolutions.
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| , However, there have been issues where the licensee's approach was good, but the licensee did not thoroughly understand NRR staff guidance. Once the staff guidance : | | , |
| was fully explained, the licensee proposed timely resolutions l l which were technically sound and exhibited proper conserva-
| | C. Maintenance / Modifications Analysis |
| : tism. For a few issues, full explanation of the staff guidance required an above average amount of staff effor ,
| | ' Portions of eight routine inspections by the Resident Inspector reviewed maintenance activities. One violation discussed in Section IV.H. Outages, reflects on the licensee's ability to conduct maintenance work during outages. In additional, two Regionally based inspections were performed. The inspections included reviews of normal maintenance and modification activities to ensure that approvals were obtained prior to initiating work, activities were accomplished using approved procedures, post maintenance testing was completed prior to returning components or systems to service, and parts and materials were properly certified. In addition, work planning and scheduling was reviewed as well as the effectiveness of administrative controls to ensure proper priority is assigne No violations or deviations note During the evaluation period, the licensee interrupted plant operations for nine unscheduled maintenance outage periods ranging from one to 11 days. Three outages were required to repair Reactor Depressurization System (RDS) valves due to the degraded condition of the system preventing successful performance of quarterly surveillances. These included one forced shutdown required by Technical Specifications unidentified leak rate limitations. Two outage period of one day each were required to successfully repair IA-60B, seal leakage to heat exchanger for Reactor Recirculation Pump No. 2. Also, two outages of three and four days each were required to diagnose the correct steam leakage from the reactor vessel head o-rings. One outage period of four days was used to replace a recirculation pump seal, and a one day outage was required to correct steam leaks associated with the plant scram on December 7, 198 Proper planning and outage control was generally evident for the nine unscheduled outages. Although unplanned, the licensee in the case of the RDS and recirculation pump outages had sufficient warning to plan activities, prepare parts and procedures, and perform other maintenance work that fell within the scope and time limitations of the forced outage. Repair to RDS valve top assemblies have become commonplace to the point that the licensee routinely overhauls spare top assemblie The licensee made extensive use of vendor consultants and pump experts from the General Office for the seal replacement, resulting in a refined and useful procedure for rebuilding and installation. Outages for RDS and recirculation pump repairs were well planned and executed. Outages to repair IA-60B represented an operational situation that offered little warning and first attempts at repairs were unsuccessful. The |
| Examples of such issues are Incorporation of Byproduct
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| . License, RDS Valve Testing, and Environmental Equipment Qualificatio It should be noted, however, that the issues presented above
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| last half of the evaluation period, the licensee has demonstrated a clear understanding of the issues, appropriate conservatism when the potential for safety significance existed, and generally sound and thorough approache This reflects positively on Big Rock Point's willingness to work closely with the staf l I
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| . | | General Office for the seal replacement, resulting in a refined and useful procedure for rebuilding and installation. Outages for RDS and recirculation pump repairs were well planned and |
| d. Responsiveness to NRC Initiatives The licensee's initial responses to NRC initiatives almost always contain acceptable resolutions, provide for timely resolution of issues, always met deadlines and were generally sound and thorough. Although the assessment for this attribute was determined to be near average for the first half of the evaluation period (due to the Incorporation of the Byproduct License, RDS Valve Testing, and Environmental Qualification of Electrical Equipment), the performance of the licensee for this attribute during the second half of the evaluation period was excellent. We attribute this, in part, to the willingness of the plant manager to take control and ensure mutual goals are attaine e. Enforcement History l | | - |
| | executed. Outages to repair IA-60B represented an operationa situation that offered little warning and first attempts at repairs were unsuccessful. The reactor vessel o-ring offer d no warning prior to failure, but successful repairs were . layed when the problem was misdiagnosed. Once the decision wa made to perform the vessel head removal and ring replacemen the physically demanding job was successfully completed w' h conservative consideration to ALARA and personne sa et Maintenance work (including mechanical, electr < , and instrument / control) at Big Rock Point is per . by generally competent repairmen who exhibit craftsmans d a general familiarity with the facility and the equip . The amount of unsuccessful repair attempts resulting W ework is generally small. Repairmen generally are cognizant o procedural require-ments associated with their assigned , communicate effec-tively with operators and health ph echnicians, and reflect concern for ALARA consideration the input repairmen provide to machinery history is o with co-workers and supervisors 'Ng .arginal, communication tes genuine interest in continued safe and successful on of the reactor. The mechanic who performs the work jo example, often participates in post maintenance testin . W e the retirement of older, experienced maintenance de tm t personnel during the period had a negative impact on ,ance as documented further in Section IV.H, Outage e naintenance staff demonstrated flexibility and dedica i roughout the evaluation perio The size of the maintenon staff is generally adequate for all periods other than major > fueling outages. A gradually increasing backlog of ma ntenance orders over the period is explained in part by i reased emphasis on skills training which over the short term r uces staff size availabilit Like the Operations Department the loss of older experienced l personnel due to etirement or other duties has altered composi-l tion of the main enance staff. While the I & C group remained |
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| | unchanged, in e mechanical maintenance group of 12 men, five were added dur ng the assessment period. Because hiring and promotion is eavily influenced by Labor Relations agreements that emphas'ze seniority, newly added staff members generally l have litt or no experience with nuclear powered generating |
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| | plants i general or Big Rock Point specifically. Although the I license has long recognized the need for maintenance staff l traini g, no training was provided until February 1986, when a i |
| | regu' r program of skills training offsite was initiated. The skiy s training is general in nature and is not nuclear plant splcific. No nuclear plant system or concepts training is provided. |
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| This area is addressed in other functional areas of this repor f. Reporting and Analysis of Reportable Events The Big Rock Point plant operated at power during most of the report period, except for about two months of refueling outage from September 6, 1985 to November 7, 1985, and short periods of shutdown for other causes. In a period of about eight months (from January 1, 1985 to September 6, 1985),
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| the plant operated with a Reactor Service Factor * of 82%.
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| In the 17 months covered by this SALP evaluation, the licensee reported eight** events to the NRC Operations Center as required by 10 CFR 50.72. Three unusual events I concerning mechanical and electrical failures were also I reporte One of the unusual events reported dealt with the shutting down of the unit from 91% power on December 31, 1984 due to failure of the reactor depressurization system (RDS) valves to pass as 'urveillance test. Failure of the RDS valves was noted in the previous SALP report on this plan The repetition of the RDS valve failure suggests that the licensee needs to give more attention to follow-up analyses and action Two of the three unusual events, including RDS valve failure, resulted in entry into limiting condition for operation (LCO) action statements. During this report period, 12 Licensee Event Reports (LERs) per 10 CFR 50.73 were receive * Reactor Service Factor = (Hours of Critical Reactor Operatior./Possible Hours) x 100%.
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| **The number of events reported to the operations center may not be the same as the number of Licensee Event Reports because of different reporting criteria and in some cases an event initially reported to the operations center may be reassessed as not reportabl .
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| Of the eight 50.72 reports, two reports involved reactor scrams which occurred in 198 These scrams were manually performed at 10% and 15% powe This reactor trip frequency of two per year compares favorably with the current national average frequency of 5.9 trips per yea Of the remaining six 50.72 reports, four reports involved ,
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| signal resulting from electrical noise affecting power level instrumentation at low power levels. Two of the spurious RPS actuations involved no rod movement, while a third occurred during control rod drive testing and resulted in the insertion of the single withdrawn control ro The fourth actuation occurred at 0.1% power while shutting down for routine maintenance. One report dealt with the loss of emergency notification sirens. The last of the 50.72 reports pertained to a discovery that a support hanger for the reactor depressurization system had not been preten-sioned after a system hydro several years ago (3-6 years)
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| due to what the licensee called a procedure inadequac Although this incident represented a fourth unusual event, the licensee failed to inform the NRC that an Unusual Event had been declared until securing from that classificatio None of the reportable events was considered individually l significant enough to warrant detailed NRR staff follow-u None of the events reported during the period was discussed
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| at the Operating Reactor Events Briefing Staffing The licensee has a licensing staff which appears to be sufficient to provide adequate and timely response Positions are identified and authorities and responsi-bilities are well define The CPC licensing contacts for the NRR licensing Project Manager at the facility and in the Corporate Office have or once held an SR0 licens Because of the Operations experience of these contacts many technical issues can be' resolved on initial contact with the license Management attention and involvement was generally aggressive l and disciplined. This was evident in both the safe and efficient i operation of the facilit Staffing levels and quality were '
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| adequate. Commurication levels between the operating staff and proper management were established and generally effectiv The i licensee has been, in most cases, effective in dealing with !
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| significant problems and NRC' initiatives. The licensee's ;
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| attention to housekeeping appears to have been excellent. The licensee's efforts in the functional area of Licensing Activities has significantly improved during this evaluation period. This is reflected in the quality of work, attention to NRR concerns and involvement of senior managemen Big Rock was an active l participant at the counterparts meeting of January 30, 1986, and 31 l
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| their plant superintendent has visited Headquarters to give an independent perspective of this concerns, and views regarding major issues confronting Big Rock and the utility industr Thus, we see several trends which have brought this utility upward in our evaluation scale. We note room for improvement and all indications reflect a very positive attitude toward continued improvemen . Conclusions The overall rating for the functional area of licensing activities is a Category 1. During this period, the licensee's performance was found to be above average to excellent overal . Board Recommendations Non Training and Qualification Effectiveness Analysis The resident and regional based inspectors regularly reviewed training and qualifications during inspection of other areas and review of events. No violations were identified in this are During the assessment period, NRC examinations were administered to five Reactor Operator candidate All candidates passed the examinations. This passing rate is significantly above the national passing rate. Based on these results, the operator licensing training program at Big Rock Point is considered satisfactor ", During the evaluation period several instances were identified where specialized training was conducted prior to non-routine operations or maintenance activities. Examples include: I Prior to installation of a Control Rod Drive with a unique l modification the maintenance crew received instructions from an experienced Superintendent using mock-up The licensee's maintenance staff received training in the use of new Control Rod Drive overhaul equipment by the Vendor, General Electri Some training was conducted for Operation Personnel prior to installation of spent fuel pool rack Walkthrough by Operation Personnel on Emergency Operating Procedures under preparation served to familiarize the operators and identify weaknesses in the procedure _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
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| . Extensive training was conducted for Operators, Superinten-dents, On-call Technical Advisors, and Instrument and Control Technicians prior to startup of the Alternate Shutdown System required by Appendix Reviews were conducted of procedures prior to installation of recirculating pump seal Training was conducted for all personnel in use of Scott Air Pak Hands-on Fire Training for all personnel was conducte One Example where training was inadequate was in the preparation l of travel crew workers in the use of local control (equipment tagging) procedure The training department routinely incorporates into system training and requalification cycle training the information i contained in all LER's, IE Notices, GE-SILS, Deficiency Reports, and industry report i Effective April, 1986, the Operations department instituted an On the Job Training (0JT) program aimed at consolidation of five former training programs leading to the SR0 license. The program will use qualification card The program's effectiveness will be evaluated during SALP Maintenance personnel during the period received virtually no l
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| training. In February the licensee began sending maintenance personnel to the Bay City Skills Training Department for General Maintenance Training that is not specific to nuclear application The Training Department has not received a request for Systems Training for Maintenance Personne The Training Department during the SALP period has added to its staff several individ'uals with extensive experience in operations, maintenance, or instrumentation and controls. This in plant, hands-on experience contributes to the quality of lesson plans and presentations. Students seem to exhibit a high degree of respect for the instructor Management involvement was reduced because of the frequent temporary offsite assignments of the Training Administrato There were no licensing actions which provided a clear opportunity to judge this attribut Based on interface with CPC's licensing and operations personnel, it appears that the training and qualification program makes a positive contribution to the understanding of technical issues and adherence to procedures with few personnel errors. Based on first-hand experiences with operations personnel, the NRR licensing Project Manager believes, however, that some improvement could still be achieve __ __. _ _ _ _
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| O 2. Conclusions The licensee is rated Category 1 in this functional are . Board Recommendations Non ;
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| | l reactor vessel o-ring offered no warning prior to failure, but |
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| | successful repairs were delayed when the problem was misdiagnosed. l Once the decision was made to perform the vessel head removal and ' |
| | ring replacement the physically demanding job was successfully completed with conservative consideration to ALARA and personnel safet Maintenance work (including mechanical, electrical, and instrument / control) at Big Rock Point is performed by generally competent repairmen who exhibit craftsmanship and a general familiarity with the facility and the equipment. The amount of unsuccessful repair attempts resulting in rework is generally small. Repairmen generally are cognizant of procedural require-ments associated with their assigned task, communicate effec-tively with operators and health physics technicians, and reflect concern for ALARA considerations. While the input repairmen provide to machinery history is often marginal, communication with co-workers and supervisors indicates genuine interest in continued safe and successful operation of the reactor. The mechanic who performs the work, for example, often participates in post maintenance tasting. While the retirement of older, experienced maintenance department personnel during the period had a negative impact on performance as documented further in Section IV.H, Outages, the maintenance staff demonstrated flexibility and dedication throughout the evaluation perio The size of the maintenance staff is generally adequate for all periods other than major refueling outages. A gradually increasing backlog of maintenance orders over the period is explained in part by increased emphasis on skills training which over the short term reduces staff size availabilit Like the Operations Department the loss of older experienced personnel due to retirement or other duties has altered composi-tion of the maintenance staff. While the I & C group remained unchanced, in the mechanical maintenance group of 12 men, five were added during the assessment period. Because hiring and promotion is heavily influenced by Labor Relations agreements that emphasize seniority, newly added staff members generally have little or no experience with nuclear powered generating plants in general or Big Rock Point specifically. Although the licensee has long recognized the need for maintenance staff training, with the exception of some I&C classes and certain skill training, no training was provided during this SALP period until February 1986, when a regular program of skills training offsite was initiated. The skills training is general in nature and is not nuclear plant specifi No nuclear plant system or concepts training is provide . |
| | First line supervision in the maintenance department reflects adequate technical skills and managerial competence. During the 1985 outage, the maintenance department overcame the loss of staff experience, inadequate outage planning, and parts proc e-ment to accomplish a relatively large number of modificati s, repairs, and preventive maintenance task Throughout the evaluation period several recurring pro ems were not successfully repaired or adequately addressed. V ve M0-7067, Turbine Bypass Isolation Valve, was not declare op able for much of the evaluation period, based on diffic ti s with the valve operator. Reactor Depressurization Sy m RDS) valves exhibit inherent design deficiencies that sulted in three forced shutdowns during the assessment p 1d a long history of problems dating back to their instal . Management, however, has not placed a high priority comprehensive solution and as a result the RDS system w not improved over the period. Problems with the Emer M y iesel Generator (EDG) |
| | fuel pump were allowed to continue nu design change to the pump mounting bracket scheduled pletion during the 1985 refueling outage was deleted i fort to return the plant to an operable status. Shortly ter the pump failed again, placing the EDG in an actio ment for the generator's Limiting Condition for Oped] . Finally, the licensee made a commitment to verify, prior startup from the 1985 outage, Limitorque Switch sett so 18 Limitorque Valves the licensee considered important ty. As of this date only 15 have been checked. The t .ettings for valve M0-7067 have been reset on three di occasions, indicating a lack of decisive direction on probl th Limitorques Operators that goes back to September, 1984, a was addressed in SALP SALP 5 expressed c cern that the Prevention Maintenance (PM) |
| | program may be in dequate to address aging equipment. At the end of this assessm t period the PM program continues to be reactive in nature, rel ng heavily on visual inspections that do not involve disa embly or physical measurements, and on the obser-vations of erators monitoring noticeable changes in component operating aracteristics. There continues to be no program to analyze r trends in failures or any other measurable parameter other an pump capacity on certain pumps. The licensee has not respo ed to NRC initiatives to upgrade the PM program to incor-pora e vendor recommendations and industry experience. The plant co inues to rely on surveillance tests to identify problems that m be in some advanced stage of development due to aging quipment. At the close of the assessment period the licensee assigned an engineer to develop a program of predictive analysis focusing on vibration and lubricating oil analysi Evidence of problems associated with aging of plant equipment during the assessment period included: Several examples of end of service life for solenoid valves on the turbine stop valve, diesel fire pump (DFP), and the exhaust ventilation downstream isolation valv . |
| | First line supervision in the maintenance department reflects adequate technical skills and managerial competence. During the 1985 outage, the maintenance department overcame the loss of staff experience, inadequate outage planning, and parts procure-ment to accomplish a relatively large number of modifications, repairs, and preventive maintenance task Throughout the evaluation period several recurring problems were not successfully repaired or adequately addressed. Valve M0-7067, Turbine Bypass Isolation Valve, was not declared operable for much of the evaluation period, based on difficulties with the valve operator. Reactor Depressurization System (RDS) valves exhibit inherent design deficiencies that have resulted in three forced shutdowns during the assessment period and a long history of problems dating back to their installation. Management, however, has not placed a high priority on a comprehensive solution and as a result the RDS system was not improved over the period. Problems with the Emergency Diesel Generator (EDG) |
| | fuel pump were allowed to continue and a design change to the pump mounting bracket scheduled for completion during the 1985 refueling outage was, deleted in an effort to return the plant to an operable status. Shortly thereafter the pump failed again, placing the EDG in an action statement for the generator's Limiting Condition for Operation. Finally, the licensee made a commitment to verify, prior to startup from the 1985 outage, Limitorque Switch settings on 18 Limitorque Valves the licensee considered important to safety. The torque settings for valve M0-7067 have been reset on three different occasions, indicating a lack of decisive direction on problems with Limitorques Operators that goes back to September,1984, as was addressed in SALP SALP 5 expressed concern that the Prevention Maintenance (PM) |
| | program may be inadequate to address aging equipment. At the end of this assessment period the PM program continues to be reactive in nature, relying heavily on visual inspections that do not involve disassembly or physical measurements, and on the obser-vations of operators monitoring noticeable changes in component operating characteristics. There continues to be no program to analyze for trends in failures or any other measurable parameter other than pump capacity on certain pumps. The licensee has not responded to NRC initiatives to upgrade the PM program to incor-porate vendor recomendations and industry experience. The plant continues to rely on surveillance tests to identify problems that may be in some advanced stage of development due to aging equipment. At the close of the assessment period the licensee assigned an engineer to develop a program of predictive analysis focusing on vibration and lubricating oil analysis. Evidence of problems associated with aging of plant equipment during the assessment period included: Several examples of end of service life for solenoid valves on the turbine stop valve, diesel fire pump (DFP), and the exhaust ventilation downstream isolation valv . |
| | Operations Department personnel performed fuel handling operations for the 1985 refueling outage. Fuel handling was |
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| | safely conducted by adequately trained individuals in accordance with approved procedural requirements. Staffing on both the reactor deck and in the control room was adequate, and communi-cation between the two areas was effective. Management involve-ment in refueling activities was evident. Tool control and status board mainteaance was adequate. Licensee responsive ss to NRC initiative was evident by their prompt action to co rect procedural deficiencies in data recording and in relocat' n of bagged equipment that had obstructed access to the refu ling deck status boar During the 1935 refueling outage several incid s ccurred which demonstrated inadequate management con er the outage process. The incidents involved: |
| | * Repeated examples of contractors and see travel crew personnel, not normally assigned to B Rock Point, performing work on the wrong com nen or system, pointing to inadequate control over the ties of travel crews and contractor * Repeated examples of superv , maintenance, operations, and engineeringorpersonnel circumventing failing t p@a19d |
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| | ere ravel crew personnel, to administrative requirements, particula) ose related to component tagging and isolatio * Repeated examples b 'n ividuals, throughout the organization, o ntion to detail and failure to exercise suffic re in performance of outage related work to ensure p safet Several factors contr' uted to the breakdown in the outage management process: |
| | * Throughout e facility, components, valves, and systems identifica on was generally inadequate, with many compo-nents un1 eled. The licensee had not acted upon earlier request from the Resident Inspector to improve component identi ication and discounted warnings on the potential for mish- * F ced retirement of several older key members of the icensee staff, including the Operations Superintendent, the coordinator of the ISI program, an experienced Shift Supervisor, and a Maintenance Supervisor who in the past had acted as a coordinator and single contact point for control of travel crew personnel. The impact of the loss of these individuals two months prior to commencement of |
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| V. SUPPORTING DATA AND 9JMMARIES Licensee Activities The unit engaged in routine power operation throughout most of SALP 6 except for a scheduled outage for the 20th plant refueling which began on September 6, 1985 and was completed on November 8, 198 The remaining outages throughout the period are summarized below:
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| December 31, 1984 - January 6,1985 Scheduled outage for surveillance on RDS valves l l
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| April 5-17, 1985 Scheduled outage to l repair RDS valves ;
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| May 15-19, 1985 Outage to repair recirculating pump seal May 25-20, 1985 Shutdown to repair leak on heat exchanger on recirculation pump May 26-27, 1985 Shutdown to repair leak l on heat exchanger on recirculation pump November 14-18, 1985 Vessel flange 0-ring leakage November 19-24, 1985 Vessel flange 0 ring leakage December 7-8, 1985 Steam leak February 11-17 RDS valves leaking The plant scrammed six times (four occurred while the plant was less than 0.1% power). In 1985, the two at power scrams were manually initiated. One was caused by a failure to manually reset a feedwater valve prior to plant startup while the other was caused by a minor steam leak in the recirculation pump room. The four remaining scram signals were caused by susceptibility of the picoammeters to electrical noise at low neutron flux levels, a known operating characteristic of the equipment with little safety significanc Inspection Activities An emergency preparedness exercise was conducted during the SALP period by Region III involving observations by nine NRC representatives of key functions and locations during the exercis Violation data for the Big Rock Point Plant is presented in Table 1, which includes Inspection Reports No. 84013-8600 _ _ _ _ _ _
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| Table 1 ENFORCEMENT ACTIVITY FUNCTIONAL N0. OF VIOLATIONS IN EACH SEVERITY LEVEL AREA III IV V Plant Operations 1 1 ! Radiological Controls 1 Maintenance / Modifications Surveillance and Inservice Testing Fire Protection Emergency Preparedness 2 Security Outages 1 Quality Programs and Administrative Controls Affecting Quality Licensee Activities Training and Qualification Effectiveness TOTALS 1 1 4 Investigations and Allegations Review Several allegations pertaining to alleged deficiencies with the licensee's security program were received from a member of the public during this evaluation period. While no immediate safety concerns were identified the investigation and resolution of the allegations have extended beyond the close of t,his evaluation period and will be addressed in a future inspection repor Escalated inforcement Actions A Severity Level III violation was issued early in 1986 for two separate incidents which occurred in 1985 resulting from a failure of supervisory personnel and repairmen to follow procedures. No civil penalty was issued because of prior good performance and extensive and comprehensive corrective action Licensee Conferences Held During Appraisal Period January 29, 1985 (Glen Ellyn, Illinois) i
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| Licensee presentation on history and operation of Reactor Depressurization Syste _ _ _ _
| | operations for the 1985 refueling outage. Fuel handling was safely conducted by adequately trained individuals in accordance with approved procedural requirements. Staffing on both the reactor deck and in the control room was adequate, and communi-cation between the two areas was effective. Management involve-ment in refueling activities was evident. Tool control and status board maintenance was adequate. Licensee responsiveness to NRC initiative was evident by their prompt action to correct procedural deficiencies in data recording and in relocation of bagged equipment that had obstructed access to the refueling deck status boar During the 1985 refueling outage several incidents occurred which demonstrated inadequate management control over the outage process. The incidents involved: |
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| | Repeated examples of contractors and licensee travel crew personnel, not normally assigned to Big Rock Point, performing work on the wrong component or system, pointing to inadequate control over the activities of travel crews and contractor Repeated examples of supervisors, maintenance, operations, and engineering personnel, and travel crew personnel, circumventing or failing to adhere to administrative requirements, particularly those related to component tagging and isolatio Repeated examples by individuals, throughout the organization, of inattention to detail and failure to exercise sufficient care in performance of outage related work to ensure plant safet Several factors contributed to the breakdown in the outage management process: |
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| . March 12, 1985 (Glen Ellyn, Illinois) t Meeting to review Systematic Assessment of Licensee Performance (SALP 5). October 1, 1985 (Glen Ellyn, Illinois)
| | Throughout the facility, components, valves, and systems identification was generally inadequate, with many compo-nents unlabeled. The licensee had not acted upon earlier requests from the Resident Inspector to improve component identification and discounted warnings on the potential for mishap Untimely retirement of several older key members of the licensee staff, which was honored by licensee management, including the Operations Superintendent, the coordinator of the ISI program, an experienced Shift Supervisor, and a Maintenance Supervisor who in the past had acted as a coordinator and single contact point for control of travel crew personnel. The impact of the loss of these individuals two months prior to commencement of |
| Licensee presentation on new reorganizatio . December 5, 1985 (Glen Ellyn, Illinois)
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| l Meeting to discuss the breakdown in management controls of
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| plant work activities.
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| l F. Confirmation of Action Letters (CALs) l l There were no CALs issued during this SALP assessmen G. Review of Licensee Event Reports, Construction Deficiency Reports, I and 10 CFR 21 Reports Submitted by the Licensee Licensee Event Reports (LERs)
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| LERs issued during the 17 month SALP G period are presented below:
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| LERs N through 85-09 86-01 through 86-02 Proximate Cause Code * Number During SALP 6 Personnel Error (A) Design Deficiency (8) -#
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| -0 l External Cause (C) 5 Defective Procedure (D) 1 Management / Quality Assurance Deficiency (E) 0 Others (X) 1 No Cause Code Marked ** 2 Total 12
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| * Proximate cause is the cause assigned by the licensee according to NUREG-1022, " Licensee Event Report System."
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| **NUREG-1022 only requires a cause code for component failure In the SALP 5 period, the licensee issued 27 LERs in 16 months l for an issue rate of 1.7 per mont In the SALP 6 period the l licensee issued 12 LERs in 17 months for an issue rate of per month. Four of the LERS were submitted for RPS activation known as " nuisance trips" resulting from electrical noise which j gives an upscale /downscale trip signal at less than 1% powe ;
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| - - - -- _ - - = - . _ . . . _. =_ _ -. .... __
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| | I During the evaluation period there was evidence that the site |
| | - |
| | staff was in danger of becoming overburdened by assignment of several functions formerly performed by the corporate QA grc Those added duties were subsequently completed or reassigned elsewhere and the site staff appears adequate for the rem (ning I workload. The site QA staff communicates effectively wi i plant management and is persistent in pressing for managemen action , |
| | to resolve audit findings. The Plant Review Committe (PRC) ' |
| | considers the quality aspects of technical and safet issue In turn, plant management generally demonstrate th r regard for the significance of findings and comments the QA staf Site QC inspectors are generally thorougt, and cientious and draw heavily on their plant experience. Botgj QA and QC site staff are responsive to NRC initiatives a W irie Licensee corporate management detracted the effectiveness of Programs and Administrative controls affe ng quality. Examples include: Licensee corporate management, ransferring to the site staff several significant Qu Assurance functions with-out a corresponding increas vailable site resources, placed a burden on the sta ch resulted in QA reviews |
| | .. thatwerelesscomprehensQa., ithdrawal of commitments to support audit activitie ff site, and a virtual elimination of time available to au s to review and observe activi-ties in the plant. Some functions were performed by QC inspectors. The rel ta e of corporate management to respond to the co of the site QA Superintendent in this regard and t oor response to NRC initiatives to address the iss note Licensee Corporate anagement deleted entirely fifteen NODS, the document in w ich the licensee staff can theoretically be assured of fi ding all applicable code and regulatory requirements c piled in one location. The NODS are the means by whic the licensee's Quality Assurance Program Description or Operational Nuclear Power Plants (Topical Report CPC- A) is implemented, and results from a commitment l made in t licensee's Regulatory Performance Improvement Program bmitted in response to a March 9, 1981 Confirmatory Orde holesale deletion of the NODS without a review to |
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| | insur all of the quality requirements contained therein |
| The licensee submitted a request to be exempt from this reporting
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| ' | | ' |
| . requiremen This request was denied because the requirement
| | were 1 ready addressed in existing administrative procedures j res ted in a period when the quality requirements were not |
| ! will be revised to address this problem. The reduction in | | ; av lable to the NODS user. Inspectors identified at least |
| !
| | ; o examples of cancelled NODS being referenced in other , |
| overall LERs is indicative of an improving tren The office for Analysis and Evaluation of Operational Data (AE00)
| | rocedures. |
| *
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| reviewed the LERs for this period and concluded that, in general i the LERs are of above average quality based on the requirements
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| contained in 10 CFR 50.73. However, they identified some minor i deficiencies. A copy of the AE0D report has been provided to the licensee so that the specific deficiencies noted can be corrected i in future reports.
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| l Construction Deficiency Reports i f
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| i No construction deficiency reports were submitted during the
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| assessment perio , CFR 21 Reports No 10 CFR 21 reports were submitted during the assessment period.
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| j H. Licensing Activities i
| | e |
| ) NRR/ Licensee Meetings (at NRC)
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| !
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| ! SALP 5 Region III 03/12/85 i Licensing Action Prioritizations 08/14/85
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| : Maintenance Practice Discussions 10/01/85 j Enforcement Conference 12/05/85
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| . Counterparts Meeting 01/30/86 l Fire Protection 03/31/86 i
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| ! NRR Site Visits / Meetings
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| : Plant /0rientation 11/07-08/84
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| ! Plant Orientation for PM/PD 07/07-12/85 i Licensing Action Prioritization 10/02/85
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| { Fire Protection 12/20/85 i DCRDR In-Progress Audit 1/27/-31/86 i Commission Meetings l
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| ! None
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| i Schedule Extensions Granted i
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| { Equipment Qualification 03/27/85 i !
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| s e Reliefs Granted ISI Relief Requests (Revision 3) 11/01/85 ISI Relief Requests 12/12/85 Exemptions Granted Appendix R, III. /26/85 RCS High Point Vents 07/17/85 Containment Airlocks Testing 01/08/86 ATWS Recirculation Pump Trip 03/20/86 License Amendments Issued Amendment Title Date 71 Plant Review Committee Review Process 12/10/84 72 Incorporation of Byproduct License 04/18/85 73 Control Rod Drive Performance Testing Frequency 05/01/85 l
| | During the evaluation period there was evidence that the site |
| 74 Containment Isolation Valve CV-4049 06/07/85 75 Stack Gas Monitoring System 06/10/85 76 Administrative Controls 07/01/85 77 Radiological Effluent Technical Specifications 08/26/85 78 Definition of Operability & Associated LC0 10/02/85 79 Surveillance Frequencies 10/22/85 80 Containment Pressure & Water Level Monitor 10/29/85 81 Reload Il Fuel MAPLHGR Limits 11/01/85 82 Plan for the Integrated Assessment 02/12/86 83 Plant Staff Reorganization and Administrative Changes 03/10/86 Emergency Technical Specifications None Orders Issued None 1 NRR/ Licensee Managment Conferences None 39 .
| | * QA staff was in danger in becoming overburden by assignment of several functions formerly performed by the corporate QA grou Those added duties were subsequently completed or reassigned elsewhere and the site staff appears adequate for the remaining workload. The site QA staff communicates effectively with plant management and is persistent in pressing for management action to resolve audit findings. The Plant Review Committee (PRC) |
| .
| | considers the quality aspects of technical and safety issue . In turn, plant management generally demonstrates their regard for significance of findings and coninents from the QA staf Site QC inspecto s are generally thorough and conscientious and draw heavily on their plant experience. Both the QA and QC site staff are respontible to NRC initiatives and inquirie Licensee corporate management detracted from the effectiveness of Programs and Administrative controls affecting qualit Examples include: Licensee corporate management, by transferring to the site staff several significant Quality Assurance functions without a correspor. ding increase in available site resources, placed a burden on the staff which resulted in QA reviews that were less comprehensive, withdrawal of commitments to support audit activities off site, and a virtual elimination of time available to auditors to review and observe activities in the plant. Some QA functions were performed by QC inspectors. It is noted however that some relief in the form of additional QA personnel from the Palisades plant was provided in September 1985. The reluctance of corporate management to respond to the concerns of the site QA Superintendent in this regard and their poor response to NRC initiatives to address the issue was noted, Licensee Corporate management deleted entirely fifteen N0DS, the document in which the licensee staff can theoretically be assured of finding all applicable code and regulatory requirements complied in one location. The N0DS are the means by which the licensee's Quality Assurance Program Description for Operational Nuclear Power Plants (Topical Report CPC-2A) is implemented, and results from a commitment made in the licensee's Regulatory Performance Improvement Prograa submitted in response to a March 9, 1981 Confirmatory Order. Wholesale deletion of the N0DS without a review to insure all of the quality requirements contained therein were already addressed in existing administrative procedures resulted in a period when the quality requirements were not available to the N0DS use Inspectors identified at least two examples of cancelled N0DS referenced in other procedure |
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Category:SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE
MONTHYEARIR 05000155/19930011993-04-0808 April 1993
Final SALP 11 Rept 50-155/93-01 for Period Oct 1991 - Dec 1992
IR 05000155/19900011990-08-17017 August 1990
Revised Pages to SALP Rept 50-155/90-01 for Jan 1989 - Apr 1990
ML20056B5211990-08-17017 August 1990
Final SALP Rept 50-155/90-01 for Jan 1989 - Apr 1990, Consisting of Minutes from 900716 Meeting W/Util
ML20055F1681990-07-0606 July 1990
SALP Rept 50-155/90-01 for Period Jan 1989 - Apr 1990
IR 05000155/19890011989-05-24024 May 1989
App to SALP Rept 50-155/89-01 for Sept 1987 - Dec 1988 Covering Diesel Generator as ESF Sys & Sipping of Fuel & Removal of Bad Fuel
ML20236C2671988-12-31031 December 1988
SALP Rept 50-155/89-01 for Sept 1987 - Dec 1988
ML20151A9151988-03-30030 March 1988
Errata to SALP 7 Board Rept 50-155/87-01
ML20150F8581988-03-30030 March 1988
App to SALP 7 Board Rept 50-155/87-01 for Apr 1986 - Aug 1987
IR 05000155/19870011987-08-31031 August 1987
SALP Rept 50-155/87-01 for Apr 1986 - Aug 1987
IR 05000155/19860011987-02-17017 February 1987
Errata to SALP Rept 50-155/86-01,consisting of App & Corrected Pages 11.12,13,22 & 25
ML20202F2641986-07-10010 July 1986
SALP 6 Board Rept 50-155/86-01 for Nov 1984 - Mar 1986. Category 1 Rating Maintained in Area of Emergency Preparedness.Regional Insp Activities for Emergency Preparedness Will Be Reduced
IR 05000155/19850011985-07-24024 July 1985
SALP Board Rept 50-155/85-01 for Jul 1983 - Oct 1984
1993-04-08
[Table view]
Category:TEXT-INSPECTION & AUDIT & I&E CIRCULARS
MONTHYEARIR 05000155/19990051999-09-30030 September 1999
Insp Rept 50-155/99-05 on 990731-0921,site Insp & 990929 Public Meeting.No Violations Noted.Major Areas Inspected: Routine Decommissioning Insp Covered Aspects of Facility Mgt & Control & Decommissioning Support Activities
IR 05000155/19990041999-08-12012 August 1999
Insp Rept 50-155/99-04 on 990609-0730.No Violations Noted. Major Areas Inspected:Aspects of Facility Mgt & Control, Decommissioning Support Activities Including Security, Spent Fuel Safety & Radiological Safety
IR 05000155/19990031999-06-21021 June 1999
Insp Rept 50-155/99-03 on 990416-0608.No Violations Noted. Major Areas Inspected:Facility Mgt & Control,Decommissioning Support Activities,Spent Fuel Safety & Radiological Safety
IR 05000155/19990021999-04-26026 April 1999
Insp Rept 50-155/99-02 on 990226-0415.No Violations Noted. Major Areas Inspected:Aspects of Facility Mgt & Control, Decommissioning Support Activities,Spent Fuel Safety & Radiological Safety
IR 05000155/19990011999-03-19019 March 1999
Insp Rept 50-155/99-01 on 990113-0225.One non-cited Violation Noted.Major Areas Inspected:Aspects of Facility Mgt & Control,Decommissioning Support Activities,Spent Fuel Safety & Radiological Safety
IR 05000155/19980091999-01-15015 January 1999
Insp Rept 50-155/98-09 on 981201-990112.No Violations Noted. Major Areas Inspected:Licensee Management & Control, Decommissioning Support Activities,Sf Safety & Radiological Safety
IR 05000155/19980081998-12-0303 December 1998
Insp Rept 50-155/98-08 on 980827-1130.No Violations Noted. Major Areas Inspected:Radiological Safety
IR 05000155/19980071998-10-14014 October 1998
Insp Rept 50-155/98-07 on 980827-1007.No Violations Noted. Major Areas Inspected:Facility Mgt & Control,Physical Security,Spent Fuel Safety & Radiological Safety
IR 05000155/19980061998-09-0303 September 1998
Insp Rept 50-155/98-06 on 980716-0826.No Violations Noted. Major Areas Inspected:Aspects of Licensee Mgt & Control, Decommissioning Support Activities,Spent Fuel Safety & Radiological Safety
ML20195C6211998-08-18018 August 1998
EN-98-059,informing Commission That Staff Intends to Exercise Broad Discretion Pursuant to Section VII.B.6 of Enforcement Policy & Not Issue NOV or Propose Civil Penalty for Severity Level II Violation Identified by Licensee
IR 05000155/19980051998-07-24024 July 1998
Insp Rept 50-155/98-05 on 980528-0715.No Violations Noted. Major Areas Inspected:Licensee Mgt & Control,Decommissioning Support Activities,Spent Fuel Safety & Radiological Safety
IR 05000155/19980031998-06-16016 June 1998
Insp Rept 50-155/98-03 on 980407-0527.Violation Noted.Major Areas Inspected:Licensee Mgt & Control,Decommissioning Support Activities,Spent Fuel Safety & Radiological Safety
IR 05000155/19980021998-04-21021 April 1998
Insp Rept 50-155/98-02 on 980203-0406.No Violations Noted. Major Areas Inspected:Licensee Mgt & Control,Decommissioning Support Activities,Spent Fuel Safety & Radiological Safety
ML20202J8471998-02-14014 February 1998
Notice of Violation from Insp on 971205-980202.Violation Noted:Chemical Decontamination Worker Working Under RWP B982035 in High Radiation Area for Approx 55 Minutes & Did Not Monitor Dosimetry Time
ML20203A0141998-02-14014 February 1998
Insp Rept 50-155/97-15 on 971205-980202.Violations Noted. Major Areas Inspected:Facility Management & Control, Decommissioning Support Activities,Spent Fuel Safety & Radiological Safety
ML20199A2571997-12-30030 December 1997
EN-97-117:on 980105,informs Commission That Order Prohibiting Involvement in NRC-licensed Activities Issued to Dt Rich.Action Based on Investigation Conducted by Oi,Which Determined on 960721 & 0915,DT Rich Falsified Records
IR 07100008/20120041997-12-17017 December 1997
Notice of Violation from Insp on 971008-1204.Violation Noted:On 971029,items Hanging from Spent Fuel Pool Rail Span 19 Had Combined Weight of 497 Pounds.Length of Time That Weight on Rail Exceeded Weight Restriction Not Determined
ML20197E5901997-12-17017 December 1997
Insp Rept 50-155/97-13 on 971008-1204.Violations Noted. Major Areas Inspected:Major Aspects of Licensee Decommissioning Activities,Including Facility Mgt & Control, Decommissioning Support & Spent Fuel Safety Controls
IR 05000155/19970101997-11-26026 November 1997
Insp Rept 50-155/97-10 on 970802-1007.No Violations Noted. Major Areas Inspected:Operations,Engineering,Maintenance & Plant Support
ML20199J4571997-11-19019 November 1997
Notice of Violation from Insp on 961216-971107.Violations Noted:Certain Radiation Surveys Required by Station Procedure RP-29,not Completed.Also Certain Radiation Survey Records Required to Be Maintained,Not Accurate
IR 05000155/19970121997-10-17017 October 1997
Insp Rept 50-155/97-12 on 970812-1007.Violations Noted.Major Areas Inspected:Miscellaneous Security & Safeguards Issues
IR 05000155/19970091997-10-0909 October 1997
Insp Rept 50-155/97-09 on 970617-0804.No Violations Noted. Major Areas Inspected:Operations,Maint,Engineering
ML20211J7341997-10-0606 October 1997
Notice of Violation from Insp on 970902-05.Violation Noted: on 970902,maintenance Worker Preparing to Participate in Reactor Head Detensioning Operations Laid Down in Contaminated Area for Several Minutes
IR 05000155/19970111997-10-0606 October 1997
Insp Rept 50-155/97-11 on 970902-05.Violations Noted. Major Areas Inspected:Various Aspects of Licensee Chemistry Radiation Protection Programs
ML20217J3391997-08-12012 August 1997
Notice of Violation from Insp on 970303-0428.Violation Noted:On 970202 Rj Flowers,Employee of Consumers Power Co, Did Not Exit High Radiation Area After Receiving Alarm on Electronic Dosimeter as Required
ML20217K5521997-08-12012 August 1997
Notice of Violation from Insp on 970303-0428.Violations Noted:Licensee Failed to Properly Evaluate Potential Radiological Hazards Associated W/Transfer of Highly Contaminated Filters
ML20151J7721997-07-22022 July 1997
Notice of Violation from Insp on 970430-0616.Violation Noted:Inspector Identified That on 960403 & 970502,seven & Five CRDs Respectively Were Left After Insertion Time Testing W/Insert Speeds Greater than 28
IR 05000155/19970081997-07-22022 July 1997
Insp Rept 50-155/97-08 on 970430-0616.Violations Noted. Major Areas Inspected:Operations,Maint,Engineering & Plant Support
PNO-III-97-055, on 970611,decision to Permanently Shut Down Plant Three Yrs Before Expiration of Operating License in Yr 2000.Capacity of 67 Megawatts & Future Needed Improvements Would Be Too Expensive.State of Mi Will Be Informed1997-06-11011 June 1997
PNO-III-97-055:on 970611,decision to Permanently Shut Down Plant Three Yrs Before Expiration of Operating License in Yr 2000.Capacity of 67 Megawatts & Future Needed Improvements Would Be Too Expensive.State of Mi Will Be Informed
ML20140G8931997-06-10010 June 1997
Insp Rept 50-155/97-04 on 970313-0429.Violations Noted. Major Areas Inspected:Operations,Maint & Engineering
ML20140G8851997-06-10010 June 1997
Notice of Violation from Insp on 970313-0429.Violation Noted:Procedures Were Not Followed,In That Required Working Clearances & Protective Tagging Were Not Obtained by Workers Prior to Performing Work Activities
IR 05000155/19970071997-06-0606 June 1997
Insp Rept 50-155/97-07 on 970426-0501.No Violations Noted. Major Areas Inspected:Licensee Fire Protection Program
IR 05000155/19970061997-05-23023 May 1997
Insp Rept 50-155/97-06 on 970428-0502.No Violations Noted. Major Areas Inspected:Review of Vehicles Barrier Sys,Audits & Self Assessment,Testing,Maint,Security Contingency Capability & Protected Area Detection
ML20141J9951997-05-19019 May 1997
Insp Rept 50-155/97-05 on 970303-0428.Violations Noted. Major Areas Inspected:Event in Which Workers Entered Power Block While Reactor at 9% Power Level Prior to Performance of Adequate Evaluation of Area Radiological Conditions
IR 05000155/19970031997-05-13013 May 1997
Insp Rept 50-155/97-03 on 970331-0404.No Violations Noted. Major Areas Inspected:Emergency Preparedness Program
PNO-III-97-012, on 970224,licensee Notified NRC Resident Inspector Ofc That Loose Radioactive Surface Contamination Spread Throughout Turbine Bldg,During Transfer of Radwaste Processing Filters from Temporary Storage Barrel1997-02-26026 February 1997
PNO-III-97-012:on 970224,licensee Notified NRC Resident Inspector Ofc That Loose Radioactive Surface Contamination Spread Throughout Turbine Bldg,During Transfer of Radwaste Processing Filters from Temporary Storage Barrel
IR 05000155/19960121997-02-24024 February 1997
Insp Rept 50-155/96-12 on 961130-970117.Violations Noted. Major Areas Inspected:Operations,Engineering,Maintenance & Plant Support
ML20135D4431997-02-24024 February 1997
Notice of Violation from Insp on 961130-970117.Violation Noted:Use of Inappropriate Surveillance Procedure Causing Uncontrolled Leak & Inline Relief Valve Lifting Due to Failure to Remove Installed Blank Flange
IR 05000155/19960011997-01-29029 January 1997
Insp Rept 50-155/96-01 on 961019-1129.Violations Noted.Major Areas Inspected:Operations,Maint,Engineering & Plant Support
ML20134C9421997-01-29029 January 1997
Notice of Violation from Insp on 961019-1129.Violations noted:961108 Operators Closed EC Number 1 Loop Inlet Valve (MO-7062) Making Number 1 EC Loop Inoperable & Operability Test Number 2 EC Loop Outlet Valve Not Performed within 1 H
ML20133C5231997-01-0202 January 1997
Insp Rept 50-155/96-11 on 961202-06.Violations Noted.Major Areas Inspected:Radiological Protection & Chemistry Controls,Status of Rp&C Facilities & Equipment
ML20133C5111997-01-0202 January 1997
Notice of Violation from Insp on 961202-06.Violation Noted: High Radiation & Contaminated Areas Was Not Posted W/Signs Bearing Radiation Symbol
IR 05000155/19960071996-12-0606 December 1996
Insp Rept 50-155/96-07 on 960909-12.No Violations Noted. Major Areas Inspected:Emergency Plan & Procedures
IR 05000155/19960061996-10-28028 October 1996
Insp Rept 50-155/96-06 on 960727-0906.No Violations Noted. Major Areas Inspected:Operations,Engineering,Maint & Plant Support
PNO-III-96-064, on 961023,shutdown for Evaluation of Environ Qualification Data Made.Licensee Performed Preliminary Assessment of Effect of Higher than Anticipated Temp Profile on safety-related Electrical Equipment Located1996-10-24024 October 1996
PNO-III-96-064:on 961023,shutdown for Evaluation of Environ Qualification Data Made.Licensee Performed Preliminary Assessment of Effect of Higher than Anticipated Temp Profile on safety-related Electrical Equipment Located
IR 05000155/19960051996-09-20020 September 1996
Insp Rept 50-155/96-05 on 960608-0726.No Violations Noted. Continued Attention Needed on Maint Work on Plant TS Required Equipment.Major Areas Inspected:Engineering,Maint, & Plant Support
PNO-III-94-039, on 940517,two Persons,Russian Visitor & Interpreter Were Killed in Traffic Accident Near Charlevoix,Mi.Cause of Accident Under Investigation1994-05-18018 May 1994
PNO-III-94-039:on 940517,two Persons,Russian Visitor & Interpreter Were Killed in Traffic Accident Near Charlevoix,Mi.Cause of Accident Under Investigation
IR 05000155/19930191993-12-21021 December 1993
Insp Rept 50-155/93-19 on 931020-1214.Violations Noted.Major Areas Inspected:Operational Safety Verification,Engineered Safety Feature Sys Walkdowns & Maint & Surveillance Activities
ML20059C1031993-12-21021 December 1993
Notice of Violation from Insp on 931020-1214.Violation Noted:Results of Static Votes Testing at Final Asleft Torque Switch Setting for MO-7053 Were Not Evaluated Prior to Return to Svc of Valve
IR 05000155/19930201993-11-10010 November 1993
Insp Rept 50-155/93-20 on 931025-29.No Violations Noted. Major Areas Inspected:Followup Insp of Previously Identified Edsfi Findings
1999-09-30
[Table view] |
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SALP 6
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APPENDIX
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- SALP BOARD REPOR p
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U.S. NUCLEAR REGULATORY COPWISSION
REGION III
s..
{
SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE
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50-155/86001 l Inspection Report
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l Consumers Power Company i
i Name of Licensee i
! Bic Rock Point Plant l' Fame of Facility 1 ,
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November 1, 1984 through March 31, 1986
!
I Assessment Period
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I
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i f
t i 8702250397 870217
PDR ADOCK 05000155 G PDR i
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i
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.
Big Rock Point Plant
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Meeting Summary The findings and conclusions of the SALP Board are documented in Inspection Report No. 50-155/86001. They were discussed with the licensee on July 21, 1986, at the Region III office in Glen Ellyn, Illinois. The licensee's regulatory performance was presented in each functional area. Overall performance and performance in each functional area was found to be acceptabl The performance rating improved in the area of Licensing Activitie Performance in the areas of Plant Operations and Surveillance and Inservice Testing declined based on increased frequency of personnel error and problems encountered in implementing administrative controls for the Plant Operations area, and missed surveillances for the Surveillance and Inservice Testing are While the performance rating in the new area of Outages was given a Category 3 based on the Severity Level III violation received during the middle of the SALP period, there has been no opportunity to evaluate the effectiveness of your corrective measures. The Emergency Preparedness and Security areas continued to have a high level of performanc While this meeting was primarily a discussion between the licensee and the NRC, it was open to members of the public as observer The following licensee and NRC personnel were in attendance on July 21, 198 Consumers Power J. Reynolds, Executive Vice President F. W. Buckman, Vice President, Nuclear Operations G. B. Slade, Executive Director, Nuclear Assurance K. W. Berry, Director, Nuclear Licensing D. Hoffman, Plant Superintendent R. R. Frisch, Senior Licensing Analyst T. C. Bordine, Staff Engineer B. Alexander, Technical Engineer U.S. Nuclear Regulatory Commission A. B. Davis, Deputy Regional Administratcr E. G. Greenman, Deputy Director, Division of Reactor Projects D. C. Boyd, Chief, Reactor Projects Section 20 S. Guthrie, Senior Resident Inspector, Big Rock Point R. B. Landsman, Project Manager, Section 2D J. Bauer, Technical Staff NRC Headquarters T. S. Rotella, NRR Project Manager J. A. Zwolinski, Director, BWR Project Directorate N .
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ERRATA SHEET
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Facility: Big Rock Point SALP Report N /86001 Page Line Now Reads Should Read 11 39-43 The licensee . . . to Delete shutdow Basis for Change: Additional information provided by licensee subsequent to SALP issuanc no training was provided with the exception of until February 1986, some I&C classes and certain skill training, no training was provided during this SALP period until February 1986 Basis for C(a ge: The phrase incorrectly implied that training was never provided and should have stated only that during most of the SALP it wasn't.
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13 26-27 As of this date only Delete 15 have been checked Basis for Change: Additional information provided by licensee subsequent to SALP issuanc Forced retirement of Untimely retirement of i
several older key several older key members members of the licensee of the licensee staff, staff which was honored by
- licensee management Basis for Change
- The phrase was not meant to mean that the employees were
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I forced out, only that they were encourage , It is noted however that
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- some relief in the form of additional QA personnel i
- from the Palisades plant
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' was provided in September
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198 t
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B4 sis for Change: Additianal information provided by licensee subsequent to i SALP issuance.
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., - - - -
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- Maintenance / Modifications
- Analysis Portions of eight routine inspections by the Resident In ector reviewed maintenance activities. One violation discuss d in Section IV.H Outages, reflects on the licensee's abi ty to conduct maintenance work during outages. In additio , two Regionally based inspections were performed. The i spections included reviews of normal maintenance and modiff tion activities j to ensure that approvals were obtained prior to itiating work, activities were accomplished using approved to edures, post maintenance testing was completed prior t e rning components or systems to service, and parts and ma were properly certified. In addition, work plannin cheduling was reviewed as well as the effectiveness ministrative controls toensureproperpriorityisassigneg violations or daviations note During the evaluation period the 44, see interrupted plant
! operations for nine unscheduled tenance outage periods ranging from one to 11 days, outages were required to repair Reactor Depressurizat stem (RDS) valves due to the
! degraded condition of the preventing successful performance
.. of quarterly surveillance ese included one forced shutdown i required by Technical ecif cations unidentified leak rate
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limitations. Two ou (p iods of one day each were required to successfully repa K DB, seal leakage to heat exchanger
. for Reactor Recir 1 Pump No. 2. Also, two outages of three and four days eac required to diagnose and correct steam leakage from the reac r vessel head o-rings. One outage period of four days was us to replace a recirculation pump seal, and i
a one day outage w required to correct steam leaks associated with the plant sc m un December 7,198 Proper plannin and outage control was generally evident for the
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nine unschedu ed outages. Although unplanned, the licensee in
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the case of he RDS and recirculation pump outages had sufficient 1 warning to lan activities, prepare parts and procedures, and perform o er maintenance work that fell within the scope and time li tations of the forced outag Repair to RDS valve top
!
assemb es have become commonplace to the point that the licensee
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routi ely overhauls spare top assemblies. The licensee did not ove aul the spare recirculation pump seal in advance of the ou ge and was still rebuilding the seal as the plant was being
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l utdown to perform the replacement, even though the pump had been died for two weeks prior to shutdown. The licensee made extensive use of vendor consultants and pump experts from the
<
11
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C. Maintenance / Modifications Analysis
' Portions of eight routine inspections by the Resident Inspector reviewed maintenance activities. One violation discussed in Section IV.H. Outages, reflects on the licensee's ability to conduct maintenance work during outages. In additional, two Regionally based inspections were performed. The inspections included reviews of normal maintenance and modification activities to ensure that approvals were obtained prior to initiating work, activities were accomplished using approved procedures, post maintenance testing was completed prior to returning components or systems to service, and parts and materials were properly certified. In addition, work planning and scheduling was reviewed as well as the effectiveness of administrative controls to ensure proper priority is assigne No violations or deviations note During the evaluation period, the licensee interrupted plant operations for nine unscheduled maintenance outage periods ranging from one to 11 days. Three outages were required to repair Reactor Depressurization System (RDS) valves due to the degraded condition of the system preventing successful performance of quarterly surveillances. These included one forced shutdown required by Technical Specifications unidentified leak rate limitations. Two outage period of one day each were required to successfully repair IA-60B, seal leakage to heat exchanger for Reactor Recirculation Pump No. 2. Also, two outages of three and four days each were required to diagnose the correct steam leakage from the reactor vessel head o-rings. One outage period of four days was used to replace a recirculation pump seal, and a one day outage was required to correct steam leaks associated with the plant scram on December 7, 198 Proper planning and outage control was generally evident for the nine unscheduled outages. Although unplanned, the licensee in the case of the RDS and recirculation pump outages had sufficient warning to plan activities, prepare parts and procedures, and perform other maintenance work that fell within the scope and time limitations of the forced outage. Repair to RDS valve top assemblies have become commonplace to the point that the licensee routinely overhauls spare top assemblie The licensee made extensive use of vendor consultants and pump experts from the General Office for the seal replacement, resulting in a refined and useful procedure for rebuilding and installation. Outages for RDS and recirculation pump repairs were well planned and executed. Outages to repair IA-60B represented an operational situation that offered little warning and first attempts at repairs were unsuccessful. The
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General Office for the seal replacement, resulting in a refined and useful procedure for rebuilding and installation. Outages for RDS and recirculation pump repairs were well planned and
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executed. Outages to repair IA-60B represented an operationa situation that offered little warning and first attempts at repairs were unsuccessful. The reactor vessel o-ring offer d no warning prior to failure, but successful repairs were . layed when the problem was misdiagnosed. Once the decision wa made to perform the vessel head removal and ring replacemen the physically demanding job was successfully completed w' h conservative consideration to ALARA and personne sa et Maintenance work (including mechanical, electr < , and instrument / control) at Big Rock Point is per . by generally competent repairmen who exhibit craftsmans d a general familiarity with the facility and the equip . The amount of unsuccessful repair attempts resulting W ework is generally small. Repairmen generally are cognizant o procedural require-ments associated with their assigned , communicate effec-tively with operators and health ph echnicians, and reflect concern for ALARA consideration the input repairmen provide to machinery history is o with co-workers and supervisors 'Ng .arginal, communication tes genuine interest in continued safe and successful on of the reactor. The mechanic who performs the work jo example, often participates in post maintenance testin . W e the retirement of older, experienced maintenance de tm t personnel during the period had a negative impact on ,ance as documented further in Section IV.H, Outage e naintenance staff demonstrated flexibility and dedica i roughout the evaluation perio The size of the maintenon staff is generally adequate for all periods other than major > fueling outages. A gradually increasing backlog of ma ntenance orders over the period is explained in part by i reased emphasis on skills training which over the short term r uces staff size availabilit Like the Operations Department the loss of older experienced l personnel due to etirement or other duties has altered composi-l tion of the main enance staff. While the I & C group remained
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unchanged, in e mechanical maintenance group of 12 men, five were added dur ng the assessment period. Because hiring and promotion is eavily influenced by Labor Relations agreements that emphas'ze seniority, newly added staff members generally l have litt or no experience with nuclear powered generating
plants i general or Big Rock Point specifically. Although the I license has long recognized the need for maintenance staff l traini g, no training was provided until February 1986, when a i
regu' r program of skills training offsite was initiated. The skiy s training is general in nature and is not nuclear plant splcific. No nuclear plant system or concepts training is provided.
!
i l 12 l
!
[
l l
i l
l reactor vessel o-ring offered no warning prior to failure, but
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successful repairs were delayed when the problem was misdiagnosed. l Once the decision was made to perform the vessel head removal and '
ring replacement the physically demanding job was successfully completed with conservative consideration to ALARA and personnel safet Maintenance work (including mechanical, electrical, and instrument / control) at Big Rock Point is performed by generally competent repairmen who exhibit craftsmanship and a general familiarity with the facility and the equipment. The amount of unsuccessful repair attempts resulting in rework is generally small. Repairmen generally are cognizant of procedural require-ments associated with their assigned task, communicate effec-tively with operators and health physics technicians, and reflect concern for ALARA considerations. While the input repairmen provide to machinery history is often marginal, communication with co-workers and supervisors indicates genuine interest in continued safe and successful operation of the reactor. The mechanic who performs the work, for example, often participates in post maintenance tasting. While the retirement of older, experienced maintenance department personnel during the period had a negative impact on performance as documented further in Section IV.H, Outages, the maintenance staff demonstrated flexibility and dedication throughout the evaluation perio The size of the maintenance staff is generally adequate for all periods other than major refueling outages. A gradually increasing backlog of maintenance orders over the period is explained in part by increased emphasis on skills training which over the short term reduces staff size availabilit Like the Operations Department the loss of older experienced personnel due to retirement or other duties has altered composi-tion of the maintenance staff. While the I & C group remained unchanced, in the mechanical maintenance group of 12 men, five were added during the assessment period. Because hiring and promotion is heavily influenced by Labor Relations agreements that emphasize seniority, newly added staff members generally have little or no experience with nuclear powered generating plants in general or Big Rock Point specifically. Although the licensee has long recognized the need for maintenance staff training, with the exception of some I&C classes and certain skill training, no training was provided during this SALP period until February 1986, when a regular program of skills training offsite was initiated. The skills training is general in nature and is not nuclear plant specifi No nuclear plant system or concepts training is provide .
First line supervision in the maintenance department reflects adequate technical skills and managerial competence. During the 1985 outage, the maintenance department overcame the loss of staff experience, inadequate outage planning, and parts proc e-ment to accomplish a relatively large number of modificati s, repairs, and preventive maintenance task Throughout the evaluation period several recurring pro ems were not successfully repaired or adequately addressed. V ve M0-7067, Turbine Bypass Isolation Valve, was not declare op able for much of the evaluation period, based on diffic ti s with the valve operator. Reactor Depressurization Sy m RDS) valves exhibit inherent design deficiencies that sulted in three forced shutdowns during the assessment p 1d a long history of problems dating back to their instal . Management, however, has not placed a high priority comprehensive solution and as a result the RDS system w not improved over the period. Problems with the Emer M y iesel Generator (EDG)
fuel pump were allowed to continue nu design change to the pump mounting bracket scheduled pletion during the 1985 refueling outage was deleted i fort to return the plant to an operable status. Shortly ter the pump failed again, placing the EDG in an actio ment for the generator's Limiting Condition for Oped] . Finally, the licensee made a commitment to verify, prior startup from the 1985 outage, Limitorque Switch sett so 18 Limitorque Valves the licensee considered important ty. As of this date only 15 have been checked. The t .ettings for valve M0-7067 have been reset on three di occasions, indicating a lack of decisive direction on probl th Limitorques Operators that goes back to September, 1984, a was addressed in SALP SALP 5 expressed c cern that the Prevention Maintenance (PM)
program may be in dequate to address aging equipment. At the end of this assessm t period the PM program continues to be reactive in nature, rel ng heavily on visual inspections that do not involve disa embly or physical measurements, and on the obser-vations of erators monitoring noticeable changes in component operating aracteristics. There continues to be no program to analyze r trends in failures or any other measurable parameter other an pump capacity on certain pumps. The licensee has not respo ed to NRC initiatives to upgrade the PM program to incor-pora e vendor recommendations and industry experience. The plant co inues to rely on surveillance tests to identify problems that m be in some advanced stage of development due to aging quipment. At the close of the assessment period the licensee assigned an engineer to develop a program of predictive analysis focusing on vibration and lubricating oil analysi Evidence of problems associated with aging of plant equipment during the assessment period included: Several examples of end of service life for solenoid valves on the turbine stop valve, diesel fire pump (DFP), and the exhaust ventilation downstream isolation valv .
First line supervision in the maintenance department reflects adequate technical skills and managerial competence. During the 1985 outage, the maintenance department overcame the loss of staff experience, inadequate outage planning, and parts procure-ment to accomplish a relatively large number of modifications, repairs, and preventive maintenance task Throughout the evaluation period several recurring problems were not successfully repaired or adequately addressed. Valve M0-7067, Turbine Bypass Isolation Valve, was not declared operable for much of the evaluation period, based on difficulties with the valve operator. Reactor Depressurization System (RDS) valves exhibit inherent design deficiencies that have resulted in three forced shutdowns during the assessment period and a long history of problems dating back to their installation. Management, however, has not placed a high priority on a comprehensive solution and as a result the RDS system was not improved over the period. Problems with the Emergency Diesel Generator (EDG)
fuel pump were allowed to continue and a design change to the pump mounting bracket scheduled for completion during the 1985 refueling outage was, deleted in an effort to return the plant to an operable status. Shortly thereafter the pump failed again, placing the EDG in an action statement for the generator's Limiting Condition for Operation. Finally, the licensee made a commitment to verify, prior to startup from the 1985 outage, Limitorque Switch settings on 18 Limitorque Valves the licensee considered important to safety. The torque settings for valve M0-7067 have been reset on three different occasions, indicating a lack of decisive direction on problems with Limitorques Operators that goes back to September,1984, as was addressed in SALP SALP 5 expressed concern that the Prevention Maintenance (PM)
program may be inadequate to address aging equipment. At the end of this assessment period the PM program continues to be reactive in nature, relying heavily on visual inspections that do not involve disassembly or physical measurements, and on the obser-vations of operators monitoring noticeable changes in component operating characteristics. There continues to be no program to analyze for trends in failures or any other measurable parameter other than pump capacity on certain pumps. The licensee has not responded to NRC initiatives to upgrade the PM program to incor-porate vendor recomendations and industry experience. The plant continues to rely on surveillance tests to identify problems that may be in some advanced stage of development due to aging equipment. At the close of the assessment period the licensee assigned an engineer to develop a program of predictive analysis focusing on vibration and lubricating oil analysis. Evidence of problems associated with aging of plant equipment during the assessment period included: Several examples of end of service life for solenoid valves on the turbine stop valve, diesel fire pump (DFP), and the exhaust ventilation downstream isolation valv .
Operations Department personnel performed fuel handling operations for the 1985 refueling outage. Fuel handling was
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safely conducted by adequately trained individuals in accordance with approved procedural requirements. Staffing on both the reactor deck and in the control room was adequate, and communi-cation between the two areas was effective. Management involve-ment in refueling activities was evident. Tool control and status board mainteaance was adequate. Licensee responsive ss to NRC initiative was evident by their prompt action to co rect procedural deficiencies in data recording and in relocat' n of bagged equipment that had obstructed access to the refu ling deck status boar During the 1935 refueling outage several incid s ccurred which demonstrated inadequate management con er the outage process. The incidents involved:
- Repeated examples of contractors and see travel crew personnel, not normally assigned to B Rock Point, performing work on the wrong com nen or system, pointing to inadequate control over the ties of travel crews and contractor * Repeated examples of superv , maintenance, operations, and engineeringorpersonnel circumventing failing t p@a19d
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ere ravel crew personnel, to administrative requirements, particula) ose related to component tagging and isolatio * Repeated examples b 'n ividuals, throughout the organization, o ntion to detail and failure to exercise suffic re in performance of outage related work to ensure p safet Several factors contr' uted to the breakdown in the outage management process:
- Throughout e facility, components, valves, and systems identifica on was generally inadequate, with many compo-nents un1 eled. The licensee had not acted upon earlier request from the Resident Inspector to improve component identi ication and discounted warnings on the potential for mish- * F ced retirement of several older key members of the icensee staff, including the Operations Superintendent, the coordinator of the ISI program, an experienced Shift Supervisor, and a Maintenance Supervisor who in the past had acted as a coordinator and single contact point for control of travel crew personnel. The impact of the loss of these individuals two months prior to commencement of
[
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Operations Department personnel performed fuel handling
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operations for the 1985 refueling outage. Fuel handling was safely conducted by adequately trained individuals in accordance with approved procedural requirements. Staffing on both the reactor deck and in the control room was adequate, and communi-cation between the two areas was effective. Management involve-ment in refueling activities was evident. Tool control and status board maintenance was adequate. Licensee responsiveness to NRC initiative was evident by their prompt action to correct procedural deficiencies in data recording and in relocation of bagged equipment that had obstructed access to the refueling deck status boar During the 1985 refueling outage several incidents occurred which demonstrated inadequate management control over the outage process. The incidents involved:
Repeated examples of contractors and licensee travel crew personnel, not normally assigned to Big Rock Point, performing work on the wrong component or system, pointing to inadequate control over the activities of travel crews and contractor Repeated examples of supervisors, maintenance, operations, and engineering personnel, and travel crew personnel, circumventing or failing to adhere to administrative requirements, particularly those related to component tagging and isolatio Repeated examples by individuals, throughout the organization, of inattention to detail and failure to exercise sufficient care in performance of outage related work to ensure plant safet Several factors contributed to the breakdown in the outage management process:
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Throughout the facility, components, valves, and systems identification was generally inadequate, with many compo-nents unlabeled. The licensee had not acted upon earlier requests from the Resident Inspector to improve component identification and discounted warnings on the potential for mishap Untimely retirement of several older key members of the licensee staff, which was honored by licensee management, including the Operations Superintendent, the coordinator of the ISI program, an experienced Shift Supervisor, and a Maintenance Supervisor who in the past had acted as a coordinator and single contact point for control of travel crew personnel. The impact of the loss of these individuals two months prior to commencement of
F
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I During the evaluation period there was evidence that the site
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staff was in danger of becoming overburdened by assignment of several functions formerly performed by the corporate QA grc Those added duties were subsequently completed or reassigned elsewhere and the site staff appears adequate for the rem (ning I workload. The site QA staff communicates effectively wi i plant management and is persistent in pressing for managemen action ,
to resolve audit findings. The Plant Review Committe (PRC) '
considers the quality aspects of technical and safet issue In turn, plant management generally demonstrate th r regard for the significance of findings and comments the QA staf Site QC inspectors are generally thorougt, and cientious and draw heavily on their plant experience. Botgj QA and QC site staff are responsive to NRC initiatives a W irie Licensee corporate management detracted the effectiveness of Programs and Administrative controls affe ng quality. Examples include: Licensee corporate management, ransferring to the site staff several significant Qu Assurance functions with-out a corresponding increas vailable site resources, placed a burden on the sta ch resulted in QA reviews
.. thatwerelesscomprehensQa., ithdrawal of commitments to support audit activitie ff site, and a virtual elimination of time available to au s to review and observe activi-ties in the plant. Some functions were performed by QC inspectors. The rel ta e of corporate management to respond to the co of the site QA Superintendent in this regard and t oor response to NRC initiatives to address the iss note Licensee Corporate anagement deleted entirely fifteen NODS, the document in w ich the licensee staff can theoretically be assured of fi ding all applicable code and regulatory requirements c piled in one location. The NODS are the means by whic the licensee's Quality Assurance Program Description or Operational Nuclear Power Plants (Topical Report CPC- A) is implemented, and results from a commitment l made in t licensee's Regulatory Performance Improvement Program bmitted in response to a March 9, 1981 Confirmatory Orde holesale deletion of the NODS without a review to
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insur all of the quality requirements contained therein
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were 1 ready addressed in existing administrative procedures j res ted in a period when the quality requirements were not
- av lable to the NODS user. Inspectors identified at least
- o examples of cancelled NODS being referenced in other ,
rocedures.
.
.
e
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During the evaluation period there was evidence that the site
- QA staff was in danger in becoming overburden by assignment of several functions formerly performed by the corporate QA grou Those added duties were subsequently completed or reassigned elsewhere and the site staff appears adequate for the remaining workload. The site QA staff communicates effectively with plant management and is persistent in pressing for management action to resolve audit findings. The Plant Review Committee (PRC)
considers the quality aspects of technical and safety issue . In turn, plant management generally demonstrates their regard for significance of findings and coninents from the QA staf Site QC inspecto s are generally thorough and conscientious and draw heavily on their plant experience. Both the QA and QC site staff are respontible to NRC initiatives and inquirie Licensee corporate management detracted from the effectiveness of Programs and Administrative controls affecting qualit Examples include: Licensee corporate management, by transferring to the site staff several significant Quality Assurance functions without a correspor. ding increase in available site resources, placed a burden on the staff which resulted in QA reviews that were less comprehensive, withdrawal of commitments to support audit activities off site, and a virtual elimination of time available to auditors to review and observe activities in the plant. Some QA functions were performed by QC inspectors. It is noted however that some relief in the form of additional QA personnel from the Palisades plant was provided in September 1985. The reluctance of corporate management to respond to the concerns of the site QA Superintendent in this regard and their poor response to NRC initiatives to address the issue was noted, Licensee Corporate management deleted entirely fifteen N0DS, the document in which the licensee staff can theoretically be assured of finding all applicable code and regulatory requirements complied in one location. The N0DS are the means by which the licensee's Quality Assurance Program Description for Operational Nuclear Power Plants (Topical Report CPC-2A) is implemented, and results from a commitment made in the licensee's Regulatory Performance Improvement Prograa submitted in response to a March 9, 1981 Confirmatory Order. Wholesale deletion of the N0DS without a review to insure all of the quality requirements contained therein were already addressed in existing administrative procedures resulted in a period when the quality requirements were not available to the N0DS use Inspectors identified at least two examples of cancelled N0DS referenced in other procedure
