IR 05000315/1987026
| ML17334B169 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 12/17/1987 |
| From: | Baer R, Choules N, Cooper T, Correia R, Hawkins F, Hooks K, Reynolds S, Toth A, Weiss S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II), NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III), NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV), NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V), Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17334B168 | List: |
| References | |
| 50-315-87-26, 50-316-87-26, NUDOCS 8801070077 | |
| Download: ML17334B169 (27) | |
Text
U.S.
NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION Report No.:
50-315/87-26 Docket No.:
50-315 50-316/87-26 50-316 License No.:
DPR-58 DPR-74 Licensee:
Faci 1 ity:
Inspection At:
American Electric Power Service Corporation Indiana and Michiaan Electric Company 1 Riverside Plaza Columbus, OH 43216 Donald C.
Cook Nuclear Power Plant, Units 1 and
D.C. Cook, Bridaman, Michiaan, October 19-30, 1987 Inspectors:
R.
P. Correia, qua ity peratians nasneer NRR (Team Leader)
. Baer, Radiatio pecia >st Reaion IV N. C.
Chou es, Reac r Inspector Reaion III ooper, Reacto nspector Reaion II
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1.0 INTRODUCTION This special announced inspection at D.C.
Cook assessed the licensee's quality verification oraanizations'bility to identify, solve, and prevent safety-sianificant deficiencies in functional areas encompassina plant systems and operations.
The inspection also assessed line manaaement's ability to ensure that identified deficiencies are dealt with promptly and completely.
The four functional areas reviewed durina the inspection were ( I) maintenance, (2) plant operations, (3) modifications of plant systems and components, and (4) radiation protection.
The details and results of each review follows.
2.0 MAINTENANCE The inspectors reviewed the involvement of gA, gC, enaineerina personnel, and responsible manaaement in various maintenance activities to assess their roles and effectiveness in identifyina and preventina safety-sianificant technical problems and in seekina effective corrective actions.
The inspectors observed maintenance activities; interviewed plant personnel; reviewed completed job orders, audit and surveillance reports; reviewed corrective actions for selected plant problems; reviewed maintenance procedures and records; and attended plannina meetinas.
2. 1 Pressurizer S ra Valves and Nuclear Instrumentation The inspectors examined and evaluated the recurrina problem with the failure of the pressurizer spray valves (NRV-163 and NRV-164 for both units) to operate as intended.
The function of the pressurizer spray valves is to control the flow of reactor coolant (RC) from the system's cold leas to the pressurizer.
Throuah a nozzle in the pressurizer, RC is sprayed into saturated steam to reduce system pressure.
This function is necessary to limit pressure increases, to prevent liftina of the pressurizer power-operated relief valves and to limit challenaes to the pressurizer heaters.
These components are the essential control devices of the pressurizer pressure control system.
The inspectors discussed this issue with the Plant Manaaer, several Maintenance, Instrumentation and Control ( 18C) personnel, gA personnel, and an enaineer from the licensee's corporate staff.
A review was performed of available documents, includina job orders performed by Maintenance and IRC personnel, memoranda from American Electric Power Service Corporation (AEPSC),
and Nuclear Plant Reliability Data System (NPRDS) information.
The inspectors also observed preparation for work on the pressurizer spray valves in Unit 2, includina personnel trainina usina a mock-up valve.
After an in-depth review and evaluation of the spray valve issue, the inspectors made the followina observations:
(I)
Job order packaaes for the valves did not always contain sufficient information and detail to determine exact root cause of problems, as-found conditions, and actions taken.
(2)
The Maintenance and I&C Departments did not document the valve's failure to operate as intended as a recurrino problem on a Condition Repor (3)
Usina input from the plant staff, the Plant Manaaer took credit for escalatina the issue in an attempt to resolve it.
(4)
AEPSC was aware of the issue and determined that worker trainina on a
mock-up valve was the necessary corrective action to alleviate it.
(5)
The gA Department was not aware of nor involved in the identification, prevention, and correction of the recurrina problem, even after AEPSC had identified the valves as such.
(6)
Corrective action to preclude repetition appeared to be slow in that the issue was initially identified in February 1985 and mock-up trainina did not beain until May 1987.
(7)
An AEPSC memorandum dated October 16, 1986, indicated that the addition of block valves around the existina spray valves was beina evaluated and a report was scheduled to be completed within the week; however, there was no indication that this action had been completed.
(8)
A review of NPRDS information and of the present problem with valve 2-NRV-163 indicated a possible oeneric problem with the valve positioners at other plants; however, there was no indication that the licensee had used this information as a possible means of determinina if the cause of their valve problems had been rectified by the other utilities with similar valves.
Durina the course of the inspection, problems with the Unit 2 pressurizer spray valves necessitated a reduction of reactor power to approximately 205 to allow plant personnel to enter the containment buildina to identify and repair the problem.
These spray valves are desianed to fail closed upon loss of air to the valve operator.
One valve, NRV-163, would not stay closed despite operator actions to keep it closed.
To close it, the air supply to the valve operator was isolated.
The other valve, NRV-164, was initially thouaht to be stuck in an indeterminate position; however, plant personnel investiaatina the problem discovered that the air supply isolation valve adjacent to the spray valve operator was closed.
When the air isolation valve was opened, the pressurizer spray valve operated satisfactorily.
Plant manaaement decided to operate the system with NRV-163 isolated.
The inspectors were concerned that licensee personnel were not aware that the air isolation valve was closed and had not checked the valve for correct positionina prior to turnina the system over to Operations Department personnel.
At the conclusion of this inspection, the licensee was investiaatina why the air supply valve was closed.
This issue and the possibility that spray valve problems are aeneric to the industry will be referred to the NRC Headquarters and Reoion III offices for further action.
The inspectors also reviewed information on the recurrina problem with the source ranae nuclear instruments, particularly NI-32.
These instruments monitor neutron flux or reactor power at the lowest shutdown levels and provide indication, alarm, and reactor trips to inform operators and protect the reactor.
Numerous Condition and Problem Reports were written by the licensee that indicated the increased attention was beina paid to the problem.
The inspectors noted that plant manaaement had taken appropriate
corrective actions, includina modification and repairs to ensure system operability.
2.2 Com onent Coolina Mater Pi e Crack The inspectors reviewed records of a weld repair made in January 1987 to a
Unit 2 Component Coolina Water (CCW) pipe.
Because there had been several cracks in other CCW pipina, the licensee decided to cut out a 2-inch by 5-inch rectanaular section of pipina containina the crack for use in deter-minina the cause of the cracks.
To repair the pipe, A-36 structural steel plate material (which was different from the existina pipe material, A-106)
was welded into the pipe where the 2-inch by 5-inch rectanaular section was removed.
The repair was determined to be acceptable by the licensee; however, the inspectors'eview of the job order and supportina documentation identified the followina:
( 1)
Chanaes were made to the weldina technique used and references to the applicable codes, but the weld plan was not chanaed to reflect these chanaes at the time the.work was performed.
After an NRC review of the activities was made in September 1987, this deficiency was noticed and chanaes were subsequently made to the weld plan.
(2)
There was no safety analysis performed as required by 10 CFR 50.59 until September 1987, after the NRC review.
(3)
The completed job order form did not reference all procedures used durina the repair; the controllino instruction was the weld plan, but the job order did not mention it under the scope of work.
(4)
guality verification oraanizations did not identify the deficiencies discussed above durina performance of the weld repair.
After the NRC review, the gA Department performed an audit and issued a report in October 1987 reaardina these issues.
These issues will be referred to the NRC Reaion III office for further action.
2.
~PS Durina the review of records aermane to the January 1987 repair of the cracked component coolina water pipe, the inspectors noted that three pipe supports with deficiencies were discovered by licensee personnel durina repair work on the pipe.
These deficiencies were described in Problem Reports written by the plant staff.
In Problem Reports PR-87-110, 111, and 116, pipe supports 2-GCCW-L294, 2-GCCW-V295 and 2-GRH-V15 were identified as beina in confiaurations different from what was shown on their respective detail drawinas.
Specifically, support 2-GCCW-L294, which was desioned to restrain horizontal-lateral pipe movements durina normal operatina and accident conditions, had an anchor bolt nut that was not properly seated on the support's base plate.
Upon further investiaation durina this inspection, the licensee noted that several of the anchor bolts'ocations on the base plate was different than what was shown on the as-built drawina of record.
Supports 2-GCCW-V295 and 2-GRH-V15 are variable sprina-type supports in which the manufacturer's shippina blocks were found to still be installe k
Shipping blocks for these types of supports prevent the sprino from compressing or elongating, making the spring support rigid.
All three Problem Reports were sent to the licensee's corporate mechanical design section (MDS) for evaluation.
MDS analyzed each of the affected pipina systems for a worst-case scenerio and determined that, in both cases, the affected CCW and RHR pipes had not been overstressed.
These evaluation results were documented on memoranda referenced as File No.
DC-D-0020M dated March 11, 1987, for PR-87-110 and 111 and No.
DC-D-0020M dated March 16, 1987, for PR-87-116.
The memoranda also included a description of the cause and actions to prevent recurrence in which it was stated that all three support deficiencies were installation errors and should be addressed by plant personnel.
Plant staff members stated that a walkdown was performed on similar pipe supports on the other train of Unit 2 CCW and RHR lines, as well as both trains of equivalent supports and lines in Unit 1.
No similar deficiencies were found in either instance.
In addition, they stated that because no other similar problems had been found as a result of the walkdowns, the probability was remote that other problems of this nature existed.
The inspectors also asked both plant and corporate licensee staff members if any deficiencies on the three pipe supports were identified during their implementation of the requirements described in NRC-IE Bulletin 79-14,
"Seismic Analysis for As-Built Safety-Related Systems."
A memorandum dated October 28, 1987, from the AEPSC offices, stated that IEB79-14 data could be found only for support 2-GCCW-L294.
However, as-built drawings were found for supports 2-GCCW-L294 and 2-GCCW-V295, but these did not include the recently identified discrepancies.
The inspectors were concerned that other as-built pipe supports may also have unidentified deficiencies and questioned the comprehensiveness of the licensee's IEB-79-14 program.
This issue will be referred to the NRC Region III office for further action.
2.4 Inservice Ins ection Pr oaram ISI The inspectors reviewed the in-service inspection (ISI) program for pipe supports at D.C. Cook.
The licensee has committed to comply with the "Boiler and Pressure Vessel Code" of the American Society of Mechanical Engineers (ASME Code)
1983 edition,Section XI, subsection IWF, article IMF-1000.
Southwest Research Corporation had been contracted to develop the plant's ISI proaram.
During the review, the licensee's staff and the NRC inspector identified that pipe supports 2-GCCW-V295 and 2-GRH-V15 were included in the ISI program, but support 2-GCCW-L294 was not.
Southwest Research was contacted as to why it did not include this support.
Southwest Research replied that support 2-GCCW-L294 was a pipe-whip restraint and all other similar supports were outside the requirements of the ASME code and were, therefore, not included in the ISI program at D.C.
Cook or other facilities for which they have developed ISI programs.
The subject support is part of the component cooling water system, which is a low-temperature, low-pressure system.
The CCW svstem would not be subject to high energy pipe whip analysis which could result in the addition of pipe whip restraints.
The inspectors recommended that the responsible corporate mechanical design section staff contact Southwest Research to discuss these issues and determine if supports, such as 2-GCCW-L294, are required to restrain horizontal-lateral normal operating and seismic pipe forces or are indeed
pipe whip restaints not subject to ASHE Section XI criteria.
This issue will be referred to NRC Headquarters and Reaion III offices for further action.
2.5 Electrical Cable Re lacement The inspectors reviewed the maintenance activities that evolved after the discovery of diesel fuel in a safety-related electrical cable conduit.
The inspectors examined job order No.
71927 and other related repair records and made the followina determinations:
(I)
Condition Report No. 1-9-87-1395, written by a contractor gC inspector, identified that the replacement cable beina installed in the containment spray pump IE conduit was different from the one specified.
However, durina that same day, enaineerina performed a safety evaluation and prepared a temporary modification allowina the substitution.
(2)
The job order did not list some of the cable used, as required.
The inspectors concluded that this event showed that this gC identified deficiency resulted in prompt resolution by enaineerina and prevented a
potential problem with the operability of the safety-related containment spray pump.
2.6 Audits and Surveillances The inspectors reviewed recent gA audit and surveillance reports in the Maintenance and ISC areas and made the followina determinations:
(I)
gA personnel were identifyina sianificant procedural and proaram related problems.
(2)
Work in proaress was audited and several surveillances also included observation of work in proaress.
(3)
gA personnel had performed some technically-oriented audits and indicated that future audits and surveillances would be more technically oriented.
2.7 Observation of Work in Progress The inspectors observed IEC personnel work on problem fire detector circuit No.
66 under job order No. 026914.
Personnel who performed the work were experienced and well aware of administrative requirements for handlina job orders and spare parts.
Durino their evaluation of the circuit, ISC personnel determined that a circuit board was not functionino properly.
When they attempted to withdraw a replacement circuit board from stores, records indicated all spare circuit boards of'hat type had been sent to the IRC shop in 1984 for replacement of capacitors that had exceeded their shelf life.
The circuit boards were located but could not be used until the capacitor replacement was properly documented and resulted in the repairs beina delayed at least one day.
Althouah the availability of replacement parts could be improved, the IEC personnel involved in this activity assured that acceptable circuit boards were installe.8
~Summar In conclusion, the inspectors made the followina aeneral determinations:
( I)
Improvement was needed in the documentation of the scope of work, as-found conditions, and work performed on job orders which could be used as data for a trendina prooram.
This would aid
$ n the prompt identification, prevention, and correction of similar or recurrina problems; enhance reliability-based, predictive, and yreventive maintenance and also aid in ALARA and manpower determinations.
The 18C Department manaaement indicated that they were in the process of developina an equipment problem trendina proaram.
The Maintenance Department had no similar plans, but should consider developina a viable trendina proaram.
(2)
Job orders are written to correct safety-related equipment deficiencies.
However, deficiencies, such as the pressurizer spray valve operability problems, are not always documented on Condition and Problem Reports.
(3)
gA personnel had identified sianificant proorammatic problems.
However, strenathenina of gA personnel's technical qualifications to enhance technically-oriented audits and surveillances is needed for providina detailed technical assessments that will be most useful to manaoement in ensurina that safety-sianificant deficiencies are beina identified where they exist.
3. 0 OPERATIONS The inspectors examined three recurrina problems in the Operations Depart-ment:
personnel errors, the use of Caution Taas, and operator trainina on plant modifications.
In addition, the inspectors examined the quality verification oroanizations'nvolvement in identifyina and resolvina these issues and what actions the Operations Department had taken to address them.
The inspection of Operations Department activities included observina control room operators durina normal, steady-state operations; while they were per-formina a turbine test surveillance; and while reactor power was reduced for maintenance on the Unit 2 pressurizer spray valves.
Licensed and non-licensed operators, as well as quality verification personnel, were also interviewed.
The inspectors observed that the control room operators were attentive to their duties and conducted themselves in a professional manner; noise and conaestion in the control room were kept at low levels.
To avoid any confusion amona the personnel involved, operators were briefed prior to the start of their assianed tasks.
Staffina was increased for 'off-normal plant maneuvers in an effort to prevent plant transients or to mitiaate the consequences if one occurred.
3.1 Personnel Errors The site Safety and Assessment Department personnel review Problem Reports to identify potential sianificant trends.
In May 1987, they identified that the
number of errors committed by operations personnel in the processina of clearance permits indicated a potential trend.
However, a subsequent investiaation performed by Operations Department personnel determined that there was no trend, because they found there were several different types of problems associated with these errors.
The inspectors noted that the majority of problems identified by the investi-aation involved personnel errors.
In reaard to this issue, the inspectors reviewed Condition Reports, Problem Reports, and Licensee Event Reports (LERs) issued durina January 1986 throuah September 1987.
The inspectors noted recurrina problems resultina from personnel errors, such as valve misalianments, procedure noncompliances, and clearance permit problems.
In one instance, Operations Department personnel had cleared a boron evaporator tank for maintenance, but Maintenance Department personnel could not work on the tank because it had not been properly depressurized and vented.
The inspectors also reviewed the Personnel Error Reports and Lessons Learned Reports aenerated by the Operations Department for January 1986 through June 1987.
The reports identified an increasina trend in personnel errors.
The inspectors then discussed the events with Operations Department personnel.
Each incident had been treated by Operations Department manaaement as an isolated event, with disciplinary action taken for the personnel involved and a written description of the incident aiven to all Operations Department personnel.
The inspectors saw no other action beina taken to improve the error rate and recommended that Operations Department manaGement aaaressively pursue this issue and include operators'nput for a prompt effective resolution.
3.2 Caution Taos The inspectors observed a relatively hiah number of Caution Taas beinG used in the main control rooms:
27 in Unit 1 and 34 in Unit 2.
The inspectors reviewed plant procedure PMI-2110, Revision 12, "Clearance Permit Procedure,"
concernina the use of Caution Taas and compared it to the requirements of plant procedure PMS0.090, Revision 2, "Operator Aids and Plant Labellina."
The inspectors determined that several Caution Taas were beina used outside the definition of a Caution Taa under the clearance permit procedure.
The taos in question appeared to be used more as an operator aid or plant label as defined by the operator aid and plant labellina procedure.
This issue was discussed with plant manaaement and after reviewina the Caution Taa loa, they concurred with the inspectors.
The Operations Department Adminis-trative Compliance Coordinator informed the inspectors that durina a sprinG 1987 inspection, personnel from the Institute for Nuclear Power Operations ( INPO) had a similar concern with the use of Caution Taas and, as a result, Operations Department personnel were scheduled to beain a review of their use.
The site QA supervisor was interviewed concernina QA audits and survei llances of the use of Caution Taas.
The inspectors determined that the QA depart-ment's activities tended to be compliance-oriented in this area, in that they essentially ensured that taa loaaina was properly performed and the taa loa properly maintained.
The inspectors were concerned that QA personnel did not question the quantity of Caution Taas beina used in the control rooms and
address the issue with Operations Department personnel to ensure that the taas were beina properly used.
3.3 0 erator Trainina The inspectors reviewed licensed and non-licensed operator trainina practices relatino to plant modifications.
Based on that review, the inspectors determined that the Operations Department staff, who are responsible for developino the trainina packaaes for the operators were receivina and usina information from documents, such as Desian Chanae Guidelines (DCGs), that were not controlled and in some cases did not provide an accurate description of the implemented modification.
The licensee's gA Audit 87-27, identified several DCGs that were inaccurate and contained incomplete information about the desian chanae description, materials, and dispositionino of displaced equipment.
DCGs are summaries of required modification activities, such as drawina and procedure chanaes, testino, and job orders issued to implement the chanaes.
The gA oraanization closed the audit findina based upon Revision 10 to Plant Procedure PNI-5040, which deleted the requirement for preparation of DCGs, and other requirements for content and distribution of the data.
The gA auditors did not recoanize that DCGs had been used in the past, and that plant personnel planned to continue to use them.
As an example, the Operations Trainina Coordinator stated that he used the DCGs as (1) his sole source to initially decide the extent of operator trainina required, (2) his sole basis for preparina required readina trainina summaries for operators, and (3) as his basis for attestina that required trainina had been completed, which was a condition for release of the chanaed equipment for operation.
He indicated that he relied on the information in the DCGs in the past, and anticipated its continued use in the future.
He was not aware of any question reaardina the accuracy or completeness of the information.
In addition, the Facility Data Base Supervisor confirmed that data base evaluation involves reference to the DCGs as a key element, but not a sole source of information on a particular modification.
The Reactor Simulator Manaaer confirmed that his staff has been compi lina DCGs for future action when the plant simulator is delivered to the site for installation.
The Plant Manaaer stated that he had not been aware that the DCG was beina used for information in these ways.
He also stated that the DCG was not intended for that purpose, nor was it suitable.
He stated that the plant staff will formulate an information distribution document that will meet the identified needs.
The inspectors also interviewed various licensed operators on several shifts and determined that, in most cases, the operators do not consider the trainina they receive on plant modifications adequate.
Generally, this trainina consists of 'required readina of information that does not include sufficient detail to correctly impart the required level of knowledae to the licensed Operations staff.
Discussions with the quality verification oraanization and with Operations Department line manaaement revealed that they were unaware of this potential inadequacy in plant modifications trainina.
The inspectors saw this as
another opportunity for Operations Department manaaement to work with their operators and pursue this issue to affect prompt resolution thus providino operators with adequate trainina necessary for safe,. reliable plant operation.
3.4 0 erations De artment Audits The inspectors reviewed the results of seven Limitina Condition of Operation (LCO) Verification audits conducted from January 1986 throuah September 1987 by the on-site QA Department.
The audits consisted of a daily review of the shift operations loas, LCO loas, and anv corrective actions taken to resolve LCOs.
The audits appeared complete and comprehensive and were of sufficient technical depth.
Five additional QA audits and seven QA surveillances of other Operations Department functions for the same time period were also reviewed.
These appeared to be primarily compliance-oriented, althouah the reports from the latter part of 1987 did show a shift toward areater technical-orientation.
This shift stemmed primarily from the fact that the on-site QA Department added three former operators who presently hold or formerly held Senior Reactor Operator licenses.
The inspectors noted this as a positive step by QA towards becomina more performance-oriented and encouraaed QA manaoement to continue in this direction.
3. 5
~Sama r The inspectors determined that the Quality Assurance and the Safety Assess-ment oraanizations have the personnel and resources necessary to perform a complete and comprehensive review of operational events.
Recent audits, survei llances, and reviews by these oraanizations show an increase in tech-nical depth and are more performance-oriented than in the past.
However, problems identified by these quality verification oraanizations must be addressed and resolved by line manaaement.
Problems, such as personnel errors, operator traininq in plant modifications, and the use of Caution Taos must be addressed in a timely manner.
Manaaement should also consider workina with operators to aaaressively pursue these issues and correct the apparent deficiencies.
These issues will be referred to the NRC Reaion III office for further action.
4. 0 MODIFICATIONS The inspectors reviewed specific modifications to plant equipment and examined the extent of verification activities for enaineerino work, includ-ina
CFR 50.59 safety reviews; translation of desian packaaes into installation and testina instructions; installation work in proaress; com-pleted work, includina restoration of items temporarily removed or deacti-vated; post-modification testina; and control of temporary modifications.
The licensee utilizes a Request For Chanae (RFC) system for oriainatina and controllina the aeneration of desian chanaes and modifications.
The enaineerina desian work is performed at the licensee's corporate offices; the translation into work instructions and work coordination is performed by a desion chanae coordinator (DCC) at the site.
Field chanaes to the RFC are coordinated by the DCC with the assistance of two on-site field chanae enoineers assianed by the corporate desion oraanization to expedite the as-built drawina proces The inspectors reviewed LERs for 1986 through 1987, the licensee's internal Condition and Problem Reports, and design changes implemented or issued for work during this period.
From these, the inspectors selected and examined
plant modification packages.
Various engineering, Operations Department, and gA personnel were interviewed, and gA audits of the modification program, temporary modification logs and applicable procedures were reviewed.
Also, electrical cable installations in the cable spreading vault associated with one modification were examined and materials stored for pending modifications were inspected.
Completed RFC records reviewed by the inspectors at the site indicated that there had been independent verification of the modification design process through in depth technical reviews by corporate engineering and Nuclear Safety and Licensing personnel.
The site gA engineers reviewed completed RFC packages, including associated work completion records, to assure that all required documentation is contained in the package.
Field deviations from as-designed RFCs have been identified and controlled on redlined drawings and design change deviation requests (DCDRs), since early 1986.
The gA oraanization appeared to have effectively audited this activity, including assurance that control room operator s have updated drawings which reflect modified plant equipment and systems.
gA has also identified implementation discrepancies and effected corrective actions which included review of prior DCDRs and their safety implications.
The licensee's design control procedures and processes have been upgraded in the past two years to incorporate lessons learned from various audit findings.
Plant procedures prescribe numerous form memoranda and checklists to ensure completion of design, installation, testing, as-built, and verification activities.
A review of audit reports and associated Condition and Problem Reports demonstrated that gA effectively monitored compliance with established procedures.
However, the gA effort did not appear to address the adequacy of the technical content of procedures and records.
gA efforts appeared to be effective in prompting program improvements; however, the findings of this inspection indicate that some basic problems in the work control process were overlooked by this approach.
These are discussed in Paraoraph 4.2 of this Section.
4.1 Procurement and Storage The inspectors reviewed procurement records for various RFC packages.
Procurement procedures have apparently been upgraded since 1986, as reflected by increased controls regarding inclusion of 10 CFR Part 21 requirements in purchase orders and requirements for vendors to supply recommendations for storage and maintenance of purchased products.
These improvements resulted from gA audit findings and the associated plant Problem Reports.
Materials received for RFC work are stored on-site in holding areas separate from the main warehouse facilities.
One area inspected was an unheated building that had been used by construction personnel before the Operations Department took control of the facility in Yiay 1986.
This building had been subject to temperature extremes which can be well below freezing during the winter months.
The building did not meet the Class B storage requirements of ANSI-N45.2.2, Part 6. 1.2, which requires that the building be provided with
0 II
P
~
uni'form heating and temperature control to prevent condensation and corrosion with temperatures no less than 40'
nor more than 140'.
Class B storage is required for instrumentation, motors, generators, motor control centers, switchgear, control panels, and other similar equipment.
The inspectors observed that guality Class I equipment was stored in this building, some of which had been received as early as October 1984.
Typical items included were Foxboro current repeater for the reactor coolant system wide range RDT (RFC-01-2667 and 02-2684),
and numerous Mercoid switches for various safety-related applications (RFC-12-2502).
During the inspection licensee personnel contacted the vendors, who stated that the items required Class B storage.
The inspectors also observed an air-operated valve (AOY) (Tag 5908, for RFC-2448)
and a Limitorque motor-operated valve (MOV) (Tag 9626, for RFC-2903) in storage in the plant's Auxiliary Building RFC material holding area.
The AOV receipt tag was dated December 3, 1982.
The Stores Supervisor stated that parts such as this in the RFC holding area do not receive preventive maintenance.
Further review indicated that the vendor's MOV maintenance procedures recommend inspection, cleaning, and lubrication of these AOYs on at least an 18 month basis, depending on the environment.
Inattention to these items appeared to be inconsistent with the intent of ANSI-N45.2.2, and plant procedure PMI-5030 which requires that the Plant Preventive Maintenance Program will establish the frequency and type of maintenance to be performed on stored equipment, and preventive maintenance schedules should be developed to specify lubrication schedules, inspections of equipment, replacement of items such as filters and strainers.
The inspectors also determined that preventive maintenance of MOVs installed in the plant has not been performed in accordance with vendor recommendations.
The licensee had identified this matter as part of more general MOV concerns in mid-1987, and formed a task force to address the issues.
The inspectors reviewed draft material indicating that this area of concern was being pursued by the task force.
Licensee personnel stated that preventive maintenance in storage, including RFC material in storage, would also be addressed.
guality verification personnel had not identified the material storage and associated preventive maintenance deficiencies.
The receiving guality Control supervisor stated that he had recently expressed concerns over the non-heated RFC storage area, but had not elevated his concerns to a Condition Report status.
The NRC Region III office will follow-up on licensee action to evaluate impact of the storaae environment, the lack of preventive maintenance on RFC materials, and preventive maintenance practices on installed MOVs.
4.2 Modification Installation and A Review Modification installation activities reviewed by the inspectors were well planned by DCCs.
Problems experienced before 1987 involving the control of field deviations from design specifications appear to have been addressed by the assignment of two corporate design engineers to the site to assist the DC The gA organization has clearly been active in monitorino compliance with approved procedures in this area.
During the inspectors'eview of gA final review records, it appeared that ineffective implementation of pre-installation walkdowns had resulted in excessive field chanoes.
One RFC package reviewed durina the inspection exemplified this point:
Design chanae RFC-DC-02-2686 involved the relocation of eiaht control switches for the Unit 2 essential service water system (ESW).
The pre-installation walkdown was performed by the DCC on September 6, 1984, without the involvement of the lead desian engineer.
The Design Change Preproduction Walkdown Checklist for the RFC package states that
"No addition of conduit" was involved and that the "Design Change can be installed as desioned."
The completed RFC package, however, included several field change records and Design Chanae Deviation Requests (DCDRs) because of congestion and inaccessability which did not permit installation of the modification as orioinally designed.
Pre-installation walkdowns require additional time for engineerina evaluation.
If poorly performed, they have areat potential to cause work delays, and can also impact and result in safety-related systems and components remaining out of service for extended periods.
Much of the cable and conduit installation work in this example was in the congested cable spreadino vault below the plant control rooms.
The failure to involve the responsible designer in the walkdown and the resulting problems encountered during the modification installation were not identified during the l}A review of the final RFC package.
Records examined showed that the DCC aenerally specified testing of completed modifications; however special test requirements provided by the design enoineer were not included in the RFCs examined by the inspectors.
Reviews of completed RFCs by gA personnel had not identified that test requirements and associated test results were not always included in completed RFCs.
For example, RFC-DC-02-2686, discussed previously, did not include test results that would confirm the operability of the effected ESW components.
The inspectors determined that gA reviewers should recognize these types of deficiencies and use their reviews not only to verify that certain documents are included in the RFC packages but also, and more importantly, that these documents contain correct and applicable data.
Desi'gn change RFC-DC-12-2739 was also reviewed.
It was made in response to an INPO notification of potential problems with large ITE circuit breakers.
The change reflected vendor recommendations that a small spring be installed in each of the 125 5KV-breaker units.
The completed RFC packaae included a
licensee memorandum dated March 19, 1987, which stated that,
"The components supplied by the vendor, once installed, do not require additional maintenance, therefore, the vendor maintenance manual does not require revision."
Licensee personnel that prepared the RFC package and gA personnel that reviewed it failed to identify that unplanned corrective maintenance or preparation of future purchase orders may use the vendor manual that lacks reference to the required additional spring.
This again appears to demonstrate that, the nature of the gA review was compliance-oriented and, therefore, did not evaluate the contents of documents included in the RFC packag. 3
~Summa r The inspectors determined that quality verification activities in the plant modification area requires more attention to details.
Line management and guality Assurance personnel must pay close attention to all aspects of a modification:
engineering analyses, drawing accuracy, field conditions, installation instructions, proper equipment storage and maintenance, and testino.
Prompt resolution of identified deficiencies during the modification process ensures that changes made to systems and components will ensure their safe and reliable operation and prevent unnecessary delays and plant down time.
5.0 RADIATION PROTECTION PROGRAM The inspectors reviewed the quality verification organizations'ontribution to the licensee's radiation protection prooram.
Included in this review were audits and surveillance of radioactive waste, activities aimed at keeping radiation releases as low as reasonably achievable (ALARA), and chemistry/
radiochemistry and operational radiation protection activities.
The licensee had increased the gA staff to include three auditors with previous experience in radiation protection, chemistry/radiochemistry, and radwaste activities.
These gA staff were provided the opportunity to attend technical professional training courses to maintain proficiency in their field of expertise.
5.1 m tl The on-site gA Department is assigned the responsibility to review station procedures to ensure that the applicable codes, standards, guides, and specifications are addressed.
The licensee maintains a code and standard matrix that lists applicable codes and standards for each Plant Manager Instruction (PMI) which is the top tier procedure for each activity.
The inspectors determined that the PMI code and standard matrix did not contain all the regulatory documents and applicable references:
CFR 61.55 and 61.56, ("Waste Classification and Characteristics" ); Inspection and Enforcement (IE)Bulletins 79-19, 79-20 ("Radioactive Waste for Transport and Burial") and 80-10 ("Contamination of Non-radioactive Systems" ); and Technical Specification 3/4.7.7 for Unit 1 and 3/4.7.8 for Unit 2.
In addition, the inspectors noted that procedures were not always consistent with reoulatory requirements and did not always include information contained in applicable IE notices.
For example, procedure 12 THP 6010.RAD.600,
"Personnel Decontamination Incident Reporting," Revision 3, did not address the recommendations included in IE Notices 86-23,
"Excessive Skin Exposure Due to Contamination with Hot Particles,"
and 87-39, "Control of Hot Particle Contamination at Nuclear Power Plants."
The inspectors also noted that the licensee's General Employee Trainina (N-GET) proaram had not been updated to include the hot particle contamination information to meet the requirements of 10 CFR Part 19.12, "Instructions to Workers."
The gA Department maintains a reference library that auditors use for procedure review and preparation of audit checklists.
The inspectors determined that the reference library did not contain a complete set of the applicable references, such as IE Bulletins and the licensee's response, IE Notices, and INPO Good Practices.
These are needed to perform an adequate review of station procedures to ensure that all regulatory requirements and
industry concerns are properly addressed.
The licensee stated in response to the inspectors'oncerns that radiation protection procedures were being reveiwed and revised to include all regulatory requirements and the latest industry practices.
5.
~S The inspectors determined after observing radiation protection personnel perform work activities and in discussions with them, that management oversight of day-to-day operation of radiation protection activities (which includes observation of technician peformance and plant radiological conditions)
should be increased by all levels of supervisory and technical management.
Supervisors must pay particular attention to problem areas, such as the waste drumming logs, which were found deficient during gA Audit 86-19-11.
The same condition was also identified during gA Audit 87-11, conducted April 13 through June 1, 1987, and was addressed in Condition Report 12-6-87-809.
Licensee management stated during the inspection that supervisory oversight would be increased.
5.3
~Summar In summary, the on-site gA Department is staffed with personnel with appropriate knowledge and training to perform effective verification activities in the radiation protection area.
The inspectors determined that gA survei llances are performance oriented; however, gA audits in this area did not contain a good level of detail.
Specifically, the audit scope, items examined, and details of identified problems on the audit reports appeared to be a summary of audit findings rather than a detailed report.
Sufficient details are necessary to ensure that the audited organization can fully understand what the identified deficiencies entail, determine accurate root causes, and implement effective corrective actions.
6.0 INSPECTION RESULTS SUMMARY The following issues addressed in this inspection report will be referred to NRC Headquarters or Region III offices for further action.
Report Section 2.0 Report Section 3.0 Pressurizer Spray Valve operability problems (RIII)
CFR 50.59 safety analysis for repairs to Unit 2 component cooling water pipe (RIII)
As-built configuration of pipe supports and IEB-79-14 program (RIII)
Inservice Inspection Proaram for pipe supports (Hg)
Personnel errors in Operating Department (RIII)
Use of Caution Tags in control rooms (RIII)
Inadequate operator training in plant modifications (RIII)
Report Section 4.0 7.0 EXIT INTERVIEW Storaae and maintenance of materials used in plant modifications (RIII)
The inspectors met with the licensee's representatives (included in the list in Appendix A) on October 30, 1987.
The purpose, scope, and results of the inspection were discusse APPENDIX A Persons Contacted American Electric Power Service Cor oration and Indiana and Michioan Electric om an
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NRC Alexich, Vice President Nuclear Operations, AEPSC Smith, Jr
, D.C.
Cook Plant Manaaer Svensson, Licensina Activities Coordinator Gibson, Assistant Plant Manaaer - Technical Support Rutkowski, Assistant Plant Manaaer - Production Barfelz, Safety and Assessment Caple, Safety and Assessment Ackerman, Nuclear Safety and Licensina Kroeaer, Manager - guality Assurance Horvath, D.C.
Cook guality Assurance Supervisor Hunsicker, Maintenance Supervisor Rickman, ISI/NPRDS Supervisor Leonard, Administrative/Operations Trainina Wizner, Maintenance/Production Control Ruff, Maintenance Supervisor Jones, Maintenance General Supervisor Loope, Technical Physical Science, Radiation Protection Droste, Maintenance Supervisor Mathias, Administration Superintendent Pisarsky, Maintenance Production Supervisor Baker, Operations Superintendent Kriesel, Technical Superintendent
- Physical Science Erikson, Assistant Manaaer - Nuclear Operations McElliaott, Site guality Assurance Huerter, Site guality Assurance Worthinaton, Site guality Assurance Brewer, Manaaer - Radiation Support Klementowicz - Health Physicist Kauffman, Construction Manaaer Waaoner, Trainina Support Supervisor Yount, Office Administrator Ross, Computer Sciences Superintendent DeLona, ISC Production Control General Supervisor Bei lman, IRC Superintendent Arent, Operations Krause, ISC Production Control Supervision Sampson, Safety and Assessment Superintendent Terry, Administrative Compliance Coordinator Veach, Stores Supervisor Wyckoff, Production Control Assistant Section Supervisor Younablood, Stores equality Control Level II Inspector
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- B Weiss, guality Assurance Branch Chief, NRR Hawkins, guality Operations Section Chief, NRR Joraensen, Senior Resident Inspector, D.C.
Cook
J. Heller, Resident Inspector, D.C.
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- L. Greaer, Reaion III
- M. Shafer, Reaion III Other licensee employees contacted included operators, enaineers, auditors, technicians, mechanics, and office personnel.
- Denotes those attendina the exit meetina on October 30, 198 I
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American Electric Power Service Corp.
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