IR 05000155/1986001

Errata to SALP Rept 50-155/86-01,consisting of App & Corrected Pages 11.12,13,22 & 25

ML20211G473
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	02/17/1987
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20211G425	List: IR 05000155/1986001 ML20211G421 ML20211G516
References
50-155-86-01, 50-155-86-1, NUDOCS 8702250397
Download: ML20211G473 (13)

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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

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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

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I Assessment Period

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t i 8702250397 870217

PDR ADOCK 05000155 G PDR 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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198 t

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B4 sis for Change: Additianal information provided by licensee subsequent to i SALP issuance.

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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

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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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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.

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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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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

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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.

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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