IR 05000344/1988040
| ML20206F534 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 11/03/1988 |
| From: | Rebecca Barr, Crews J, Mendonca M, Suh G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20206F513 | List: |
| References | |
| 50-344-88-40, NUDOCS 8811210187 | |
| Download: ML20206F534 (22) | |
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U. S. NUCLEAR REGULATORY C0 m!SS10N
REGION V
i Report N /88-40
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Docket N License N NPF-1 Licensee: Portland General Electric Company 121 S.W. Salmon Street Portland, Oregon 97204 Facility Name: Trojan Inspection at: Rainter, Oregon Inspection conducted: August 7 - October 1, 1988 l Inspector % . N . hb * - M R. C. Barr "
" /J /O Date Signed Senior Resident Inspector
% . % . % : ,__ h . [vy /d/ff G. Y. Suh Date Signed Resioent Inspector
% ' % ~ % L c+ p '
ss s&fn J. L. Crews Date Signed Senior Reactor Engineer Approved By:
- * ** "///#1 H. H. Mendonca, Chief Date Signed Reactor Projects Section 1 Summary:
Inspection on August 7 - October 1. 1988 (Report 50-344/88-40)
Areas Inspected: Routine inspection of operational safety verification, maintenance, surveillance, event follow-up, system engineering, and open item follow-u Inspection procedures 30703, 37702, 61726, 62703, 71707, 71709, 71710, 71881, 90712, 92700, 92701, 92720, and 93702 were used as guidance during the conduct of the inspectio Results
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This inspection identified areas for improvemer:t in the maintenance area (paragraphs 4 and 6). Observed weaknesses included procedures and work instructions not adequate for the circumstances, engineering involvement and est1210187 AD0cg 831103 0500 PDR
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management review of maintenance evolutions, engineering understanding of the pressurizer instrumentation system, and involvement of quality organization ;
An unresolved item was identified in the review of licensee testing of reactor t coolant system pressure fsolation valves (paragraph 5) in which testing temporarily placed safety related equipment in a position that prevented the automatic accomplishment of the equipment function without the licensee first ,
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declaring the components inoperable. Further, examples of voluntary entry into Technical Specification 3.0.3 for surveillance testing purposes were also L identified for further evaluatio A violation was identified concerning inadequate control of a metal ladder in a vital switchgear room (Paragraph 3).
Another violation was identified concerning a work instruction that did not (
provide adequate detail (Paragraph 6). [
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DETAILS Persons Contacted
- C, A. Olmstead, Plant General Manager
- L. W. Erickson, Manager Nuclear Quality Assurance
- R. P. Schmitt, Manager, Operations and Maintenance
- 0. W. Swan, Mantger, Tect.nical Services
- J. Singh, Manager, Plant Modifications
- J. D. Reid, Manager, Plant Services
- J. W. Lentsch, Manager, Personnel Protection J. M. Anderson, Manager, Material Services
- R. E. Susee, Manager, on assignment D. F. Levin, Supervisor, Plant Modifications
- E. A. Curtis, Procurement Supervisor P. A. Morton, Branch Manager, Plant Systems Engineering R. L. Russell, Operations Supervisor R. H. Budzeck, Assistant Operations Supervisor D. L. Bennett, Maintenance Supervisor R. A. Reinart Instrument and Control Supervisor T. O. Meek, Radiation Protection Supervisor R. W. Ritschard, Security Supervisor C. H. Brown Operations Branch Manager Quality Assurance
- D. L. Nordstrom, Nuclear Engineer, Nuclear Safety and Regulation The inspectors also interviewed and talked with other licensee euployees during the course of the inspectio These included shift supervisors, reactor and auxiliary operators. maintenance personnel, plant technicians and engineers, and quality assurance personne * Denotes those attending the exit intervie . Plant Status The plant was operating at 100% power at the beginning of the inspection period. During the inspection period, numerous failures and replacements of pressurizer pressure transmitters occurre PT-456 failed on August 7, 14, 19, 24, and 2 PT-457 failed on September 7 and 1 And PT-458 failed on September 13. On August 16, a reactor trip from 100% power
- occurred on indicated low flow in reactor coolant system loop "B" during the return to service of flow transmitter FT-424 following calibratio The plant was returned to power operation later that day and attained 100% power conditions on August 17. With PT-458 in a failed condition with its associated protection bistables tripped, a reactor trip occurred on September 16 on overtemperature delta temperature when a technician incorrectly tripped the overtemperature delta temperature bistable associated with PT-455. During the subsequent forced outage, the licensee dealt with numerous issues including testing and maintenance work on pressurizer pressure and level instrumentation, the discovery of significant amounts of Asiatic clams in the service water system and twa locked valves being found out of position. With the plant in Mode 3, Hot Standby, a depressurization transient occurred in the primary system when
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maintenance was performed on pressurizer pressure transmitter PT-457 :
while it was selected as the controlling channel for primary pressur ;
Operator action terminated the transient before reactor coolant system pressure decreased to the reactor trip or safety injection f.etpoint On :
September 22, the plant was returned to power. High vibration alarms on the main turbine halted power ascension at approximately 60% power and (
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necessitated a rapid reduction in reactor power, during which control rod '
F-6 dropped which subsequently necessitated a reactor shutdow The ;
licensee's investigation found the movable gripper coil associated with !
control rod F-6 had reversed polarity which resulted in reduced gripping i force to the dropped control ro Subsequent reactor startup on :
September 23 was delayed when the turbine generator load decrease button f failed and was replaced. The plant attained 100% power on September 2 !
Conder.ser tube leakage was detected later on September 24, necessitating i a power decrease to 55% in order to plug condenser tubes. On September i 25, the plant returned to 100% power and remained at full power through j the remainder of the inspection perio !
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3. Operational Safety Verification (71707. 71709, 71710 and 71881) !
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During this inspection period, the inspectors observed and examined ;
activities to verify the operational safety of the licensee's facilit I The observations and examinations of those activities were conducted on a i daily, weekly or biweekly basi ,
I Daily the inspectors observed control room activities to verify the l licensee's adhereate to limiting conditions for operation as prescribed in the facility Technical Specifications. Logs, instrumentation, !
recorder traces, and other operational records were examined to obtain .
information on plant conditions, trends, and compliance with regulatf o l On occasions when a shift turnover was in progress, the turnover of i information on plant status was observed to determine that pertinent l information was relayed to the oncoming shift personnel, i
Each week the inspectors toured the accessible areas of the facility to observe the following items: {
L (a) General plant and equipment condition [
(b) Maintenance requests and repstr ;
(c) Fire hazards and fire fighting equipment, j (d) Ignition sources and flammable material contro l (e) Conduct of activities in accordance with the licensee's administrative controls and approved procedure (f) Irteriors of electrical and control panel (g) Implementation of the licensee's physical security pla (h) Radiation protection control (1) Plant housekeeping and cleanlines (j) Radioactive waste system !
(k) Proper storage of compressed gas bottles, t Weekly, the inspectors examined the licensee's equipment clearance control with respect to removal of equipment from service to determine l '
that tne licensee complied with technical specification limiting conditions for operatio Active clearances were spot-checked to ensure l
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that their issuance was consistent with plant status and maintenance !
evolutions. Logs of jampers, bypasses, caution and test tags were l examined by the inspector Each week the inspectors conversed with operators in the control room, and with other plant personnel. The discussions centered on pertinent ,
i topics relating to general plant conditions, procedures, security, ;
training and other topics related to in progress work activitie ;
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) The inspectors examined the licensee's nonconformance reports (NCRs) to l 4 confirm that deficiencies were identified and tracked by the syste '
Identified nonconformances were being tracked and followed to the !
completion of corrective action, j
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! Routine inspections of the licensee's physical security program were {
J performed in the areas of access control, organization and staffing, and i detection and assessmant systems. The inspectors observed the access !
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cor, trol measures used at the entrance to the protected area, verified the [
integrity of portions of the protected area barrier and vital area i barriers, and observed in several instances the implementation of I compensatory measures upon breach of vital area carrier Portions of
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the isolation zone were verified to be free of obstructions. Functioning
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of central and secondary alarm stations (including the use of CCTV
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monitors) was observed. On a sampling basis, the inspectors verified
] that tha required minimum number of armed guards and individuals authorized to direct recurity activities were on sit The inspectors conducted routine inspections of selected activities of j the licensee's radiological protection program. A sampling of radiation j work permits (RWP) was reviewed for completeness and adequacy of
- information. During the course of inspection activities and periodic l tours of plant areas, the inspectors verified proper use of personnel l monitoring equipment, observed inM viduals leaving the radiation I controlled area and signing out t.. appropriate RWP's, and observed the j posting of radiation araas and tar.inated areas. Posted radiation levels at locations within the vuel and auxiliary buildings were verified using both NRC and licensen portable survey meters. The ferolvement of i health physics sup.rvisors and engineers and their awareness of i significant plant activities was assessed through conversations and review of RWP sign-in records, i
The inspectors ir dfied the operability of selected engineered safety features. This was done by direct visual verification of the correct i position of valves, availability of power, cooling water supply, system
) integrity and general condition of equipment, as appitcable. The j Emergency Diesel Generator System was verified operable during this inspection period.
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On September 26, 1988, the inspectors, while performing routt'te bach shifi.
, inspection in the contro! building vital switchgear room, found an aluminum extension ladder unattended, not tagged, not secured, upright,
. leaning against and supported by switchgear ventilation decting, and was j adjacent to safety related switchgear, Plant Safety procedure PS-8-5,
"Ladders", sections A and B, t.tquires that ladders be tagged with the i
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user's name anC the date the ladder ic first used for.the associated job, and that unattended ladders be securely tied down or laid down.to p. event equipment damage in the event of seismic activity, respectively. This is an apparent violation (50-344/88-40-01).
A subsequent walkdown by the licensee safety represent - t ' hiantified-numerous instances of non-compliance with PS-8-5. As a . -
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this-plant safety procedur One violation was identified, and no deviation was identific . Maintenance (62703, 92Ig11
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Replacement of Pressurizer Pressure Transmitter PT-456 The inspectors observed the replacement of pressurizer pressure transmitter PT-456 after the fourth failure axperienced at this instrument location in the month of August 198 The licensee had experienced similar failures for the same pressure transmitter model at pressurizer pressure transmitter location PT-455 and <T-456 in late 1987 and early 1988. However, these earlier failures did not exhibit the repetitive short term failures experienced for PT-456 in August 198 In all cases, the transmitter failed with a high indication. In response to the earlier failures, the licensee had developed a detailed action plan to understand the cause of the failuret and to work with the manufacturer to presant recurrenc The replacement of the fourth August 1988 failure of PT-456 was controlled by maintenance requests MR 88-6792 and MR 88-6808. Tha work was accomplished by two instrumentation and control technicians, two electricians, a quality control inspector, and a radiation protection technician. The inspectors verified the work was performed in accordance with the requirements of radiological work permit RWP 88-408. The inspectors discussed the RWP and ALARA worksheet with the unit radiatfor protection superviso Based on observations of the work activities, subsequent discussions with licensee personnel and review of the work package, the inspectors concluded that the maintenance activities indicated three areas in need of improvcment: adequacy of procedures end work instructions, engineering understanding of the instrumentation system, and involvement of quality organization The work instructions for the replacement of PT-456 were not adequate for the circumstances. The instructions ' tere written and implemented by the technicians in such a way that upon placing the transmitter in service, a significant pressure pulse was imposed upon the pressure transmitter sensing element. This was contrary to maintenance management and eng'neering expectations. The instructions lacked specific guidance on valve manipulation to prevant the pressure pulse. In addition, the work package lacked piping diagrams for PT-456 associated tubing which given the implementation of various temporary modi'Ications were in a non-standard configuratio Also, the inst. actions were writton such that the technicians unnecessarily cycled the transmitter three times which resulted in extending the work time inside containment when the hand priming pump failed, and necessitated its repair during the third cycling attemp The cycling of the
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transmitter per vendor recommendations had previously been performed outside containmen During the first four failures and replacements of PT-456 in August 1988, minimal involvement from quality organizations was observed for the actual replacement work activities. This involvement consisted of quality control inspection hold points for the environmentally qualified splicing of transmitter electrical connections and the torque values for reinstallation of the transmitter cover plate. A quality control inspector wrote an observation report for the second August 1988 replacement which had no negative findings. During this time personnel from quality assurance and the performance monitoring / event analysis group were involved in following plant engineering efforts to understand the failure root cause and to work with the transmitter manufacturer, however, these personnel did not make observations of the transmitter replacement work activities giver t he repetitive short term failures of PT-456 in August 1988, prior to tne fifth failur Subsequent analytical results and evaluation indicated a lack of full engineering understanding of the instrumentation system associated with the pressurizer pressure and leve Instrumentation and control technicians related to the inspectors that during the replacement and placement on service of the transmitter they observed gas emission with little or no water flow upon opening valves and fitting The inspectors discussed this observation with plant and corporate engineers who stated they were aware of the presence of gas in the sensing line In subsequent sampling of the sensing line following the fifth PT-456 failure in August 1988, the licensee determined that hydrogen gas was present in significantly higher concentrations than expected, with some sensing lines associated with pressurizer pressure and level instrumentation devoid of water. In response, the licensee performed:
additional sampling of the sensing lines, safety evaluations on the effect of the observed conditions on instrument operability, and b6ckfilling of the sensing lines with demineralized wate The inspectors discussed preliminary findings with maintenance supervisory personnel and plant management from the observation of the replacement of the fourth PT-456 failur The inspectors observed increased participation by engineering, maintenance management, corporate support groups, and quality assurance personnel following the fifth PT-456 failure. Increased effort was observed to improve the quality of procedures, work instructions, and pre-work briefings in the replacement and associated maintenance of PT-456 and of PT-457 and PT-458 which clso failed twice and once, respectively, before the end of the inspection perio The licensee has (1) expanded its action plan to address the possible effects of high concentrations of hydrogen gas with little or no liquid water in the sensing lines; (2) assigned a new project manager; and (3) accelerated its efforts to replace the pressuriter pressure transmitters with a different manufacturer's model. The inspectors will continue to follow licensee activities in this are In discussions with plant management, the inspectors also shared the abservation that maintenance tool boxes and storage chests inside containment were left open during the time maintenance was being
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i performe The plant was at full powe The concern of post-LOCA .
recirculation sump blockage was discussed in the event that the containers were not secured after maintenance or.upon containment evacuation.' Plant management stated it would address the concer In
_ subsequent containment entries with the unit at power, the inspectors
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observed that tool boxes and storage chests were immediately secured by l maintenance' personnel after tool retrieva Additionally, the-inspectors noted that an instrumc.ntation drain isolation valve was not in its intended position. The licensee corrected the valve position and is ;
currently addressing valve position problems based on other finding No violations or deviations were identified.
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! Surveillance (61726, 92701)
Emergency Core Cooling System (ECCS) Check Valve Leak Testing :
During routine inspections of engineered ~ safety features, the resident '
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inspectors identified that (1) the facility had apparently been placed in Technical Specification 3.0.3 during the performance of Periodic Operating Test (POT) 2-4, titled "ECCS Pressure Boondary and Accumulator Check Valve Leakage," and (2) the condition of system inoperability had l not been formally declared, such as by' documentation in operators' log The inspection focused on identifying where redundant cafety systems were removed from service, and on the review process for. changes to POT 2- Instances of V_ol,antary Entry Into T.S. 3.0.3
! The inspectors identified that during the 1988 Refueling Outage while
. conducting POT 2-4, the purpose of which was to determine if the leakage rate of reactor coolant system pressure isolation check valves met the j requirements of T.S. 3.4.6.2. f. , it appeared that the licensee voluntarily entered T.S. 3.0.3 nineteen times for periods from five minutes to greater than two hours when the testing was performed in Mode
, 3 Hot Standby. Entry into T.S. 3.0.3 apparently occurred because none of the ECCS safety systems as required by T.S. 3.5.2 were operabl Trojan Technical Specification 3.5.2 requims es a limiting condition for operation that two independent ECCS subsystems De operable when in Modes
- 1, 2 and Each subsystem is comprised of a centrifugal charging pump,
, safety injection pump, RHR heat exchanger, RHR pump, and applicable flow path. Plant evolutions that render om ECCS subsystem inoperable place the facility in a technical specification action statemen Plant evolutions that render both ECCS subsystems inoperable place the plant
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outside T.S. 3. The inspectors also noted that entry into T.S. 3.0.3 could have been avoided eit*1er through test performance in an alternate mode (Mode 4) or through the use of temporary instrumentation or test equipment at piping test connections. Technical Specification 3.5.3 requires as limiting condition for operation that one ECCS subsystem comprised of a centrifugal charging pump, RHR heat exchanger, RHR pump, and applicable flow path be operable in Mode 4, Hot Shutdown. Plant evolutions that render both ECCS subsystems inoperable in Mode 4 would place the plant in
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a-technical specification action statemen With respect to evolutions associated with ECCS check valve leak testing, the applicable action statement in T.S.-3.5.3 would require that the reactor coolant system average temperature be. maintained less than 350 degrees Fahrenheit by use of alternate heat. removal method In this.section of.the report various ECCS related valves will be mentione The function of each of these valves is as follows: ,
MO-8835 Intermediate head safety injection (SIS) cold leg'
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injection isolation valve M0-8802A SIC hot' leg injection (loops 2 and 4) isolation valve t
M0-8802B SIS hot leg injection (loops 1 and 3) isolation valve M0-8703 Residual heat renioval (RHR) hot leg discharge isolation valve ,
h0-8809A RHR cold leg injection (loops 1 and 2) isolation valve e
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MO-8809B RHR cold leg injection (loops 3 and 4) isolation valve MO-8923A,B SIS pump suction valves MO-8821A SIS pump 'A' discharge line isolation valve !
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The following are the instances where testing apparently placed the
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facility in T.S. 3.0.3 in July 1988 during the 1988 Refueling Outage:
Testing of the first off ECCS cold leg check valves (8948 A,8,C and D) '
during the closure of M0-8835 and MO-8809A while l'1 Mode Testing of residual heat removal second off cold leg check valves (8818 ;
A,B,C and 0) during the separate closures of MO-8809A and MO-88098 while '
in Mode '
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Testing of residual heat removal second off hot leg check valves (8736 A l
and B) during the opening of MO-8802A while in Mode Testing of safety injection second off cold leg check valves (8819 A,B,C and D) during the closure of MO-8835 while in Mode 3.
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Testing of safety injection second off hot leg check valves (8905 A. i and D) during the closure of M0-8923A, M08923B, MO-8835, M0-8821A, the opening of M0-8802A, and placing both safety injection pumps in the l 1 pull-to-lock position while in Mode '
In response to the potential that other systems could have been made #
1 inoperable and T.S. 3.0.3 entered, the licensee perfctmed a revie The
licensee determined that six procedures had the potential to place the
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plant in a 3.0.3 condition: POT 2-4 on ECCS Check Valves; ONI 14 on
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Component Cooling water System; A0 3-21 on Control Room Emergency ,
Ventilation Boundary; Temporary Plant Test 258 on ECC.S Check Valve l
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Leakage; POT 4-2 on Containment Spray System; and' HOP 5-7 on Component Cooling Water Surge Tank hitrogen Regulator Adjustment. The licenrae was completing their assessment of the potential impact of these procedures on equipment operability.and 3.0.3 entry during the inspection perio The ifcensee also reviewed procedures to determine when the plant was placed in action statements, and there was no annotation of the entry in the control room log The licensee found about 15% of their procedures had the potential for entry into action statements without declaration of the inoperable equipment condition in an appropriate lo The licefisee plans to complete resolution of this by the end-of-the yea Periodic Operatina Test (POT) 2-4 Procedure Reviews The inspectors examined revision 2 through 16 of POT 2-4 to evaluate the
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history of the testing methodology.
The inspectors noted the initial conditions for testing of the check valves were specified in terms of Reactor Coolant System (RCS) pressure.
. Normally, initial condition 9 also include the operational mode in which the facility is required to conduct a test. According to the licensee,
- the use of RCS pressure as the initial condition w&s to ensure the test
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was conducted as close to normal operating pressure as possible.
- The following is a discussion of the inspectors' reviews of the specific
revisions to POT 2-4.
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j In Revision 2, approved on June 20, 1980, the licensee added leak testing i
of the residual heat removal second off check valves and intermediate i
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head safety injection system second off check valve This testing resulted in the closure of valves M0-8809A, MO-88098, and MO-8835 which affected redundant trains of the emergency core cool'.ng system (ECCS).
Revision 2 included the following precautions:
" Valves shall be lef t in the abnormal position for a minimum amount of time."
"S. 8 Plant is to be placed in the appropriate mode of operation as limited by plant Technical Specification, Sections 3.5.2 and
3.5.3."
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Revision 5, approved on June 4, 1981, addea additional testing which
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involved opening valves MO-8802A and MO-8703, 1his soultion also
affected redundant trains of ECC Revision 6, approved on August 31, 1981, incorporated testing performed
{ under a temporary plant test, corrected typographical errors, clarified
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Revision 5, and deleted Precaution The inspectors reviewed the completed plant operating manual revision form and attachments for Revision 6. The PGE review, as documented by the completed plant opt.ating manual revision form, concluded that the
- revision did not constitute en unreviewed safety question. However, the
licensee's justification for this conclusion did not specifically address
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the deletion of precaution- Included in the attachments to the'
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revision form were marked up pages of Revision 5 which showed changes to Revision 5 which were incorporated in Revision 6. Next to precaution were the handwritten words, "This is true at all times, not needed here."
The inspectors reviewed the associated Plant Review Board Consideration Form and the Plant Review Board meeting minutes which~ stated that Revision 6 was reviewed. The inspectors did not find any discussion on >
the deletion of Precaution i i
Revision 10 was approved on June 21, 1985, and expanded precaution 5.1 to i the followin I
"Valves shall be left in their abnormal position for a minimum amount of !
time. Shutting M0-8835 isolates both trains of safety injection pump [
discharge to RCS cold legs. Operability is ensured by stationing an :
operator while the valve is shut to open MO-8835 to realign safety injection upon receipt of an automatic safety injection actuation signal or at tha direction of the control operator."
Stationing an operator would not necessarily ensure operability of buth trains of safety injection. Additionally, operator action to return valves from abnormal positions could not be assured in all situations; such as, valve failure or operator oversight during response to actual events. In addition, ECCS safety analyses make the assumption that no operator action is assumed to be taken by plant operators to correct
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problems during the time period immediately following a design basis
accident. The inspectors reviewed the completed plant operating manual revision form and attachments. Aside from the indicated change to ;
Precaution 5.1, the inspectors found no analysis or discussion addressing the change. The inspectors reviewed the associated Plant Review Board Consideration Form and found no discussion concerning the change of
Precaution 5.1. The inspectors reviewed minutes of Plant Review Board 1
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meetings held in the period immediately about June 21, 1985 and found no t discussion concerning the change of Precaution '
Revision 11, approved on March 26, 1986, in part changed testing of check !
- valves 8905 A,B C and D to require that both intermediate head safety [
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injection pumps be placed in pull-to-lock position and resulted in several safety injection system valves being placed in abnormal l position Revision 11 required that this portion of the testing be done !
in Modo 4. As indicated previously, intermediate head safety injection is not required by Technical Specifications while in Mode l
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l' Revision 12, approved on June 6, 1986, altered precaution 5.1 to the following: ;
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"Valves shall be lef t in the abnormal position or pumps in a pull-to-lock j mode for a minimum amount of time (six hours). Several steps in this
. proceduria position valves or place pumps in pull-to-lock that affect two i trains of ECCS flow. Operators shall be prepared to reposition these i valves or pump switches to their normal or auto position upon receipt of !
an automatic safety injection actuation signal or at the direction of the l control operato The PRB has determined that T.S. 3.0.3 has not been !
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entered if these valves or pump switches have been shut and reopened or repositioned in a timely manner (six hours). . Failure to be.able-to reopen or reposition these valves 1or switches constitutes entry into .0.3 with the applicable portions of T.S. 4.0.3 requiring complianc .
Six hours comes from the fact that the testing is done in either Mode 4 i or Mode 3." *
The inspectors reviewed various licensee documents and found no .
evaluations wh:ch demonstrated that the ECCS trains continued to be !
operable with the valves being in abnormal positions and pumps in a ,
pull-to-lock mode for a time of six hour The reviewed documents i included the plant operating manual revision form for Revision 12 and ,
attachments, the associated Plant Review Board review consideration form, l and minutes of the Plant Review Board meeting which reviewed Revision 1 l The inspectors noted that the NRC had previously identified an inadvertent entry into T.S. 3.0.3 on March 31, 1986, for a residual heat i removal cold leg discharge isolation valve. During the inspectors' *
current follow-up, the licensee provioed Plant Review Board meeting minutes which provide in part the bases for the position described in the POT 2-4 statemen An excerpt of the meeting minutes follow "The Event Report raised the question of whether valves which affect both trains of a system should be cycled periodically as is requirtd by Section XI of the ASME Boiler and Pressure Vessel Cod It was noted that we have relief requests for some of tho valves listed in Technical Specification 4.5.2.a to only cycle them in cold. shutdow Others,
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however are still cycled quarterly. The PRB directed Trojan Plant Engineering to develop a prlicy on this and to ensure that it is consistantly implemeated. The PRB did not feel that cycling one of these valves closed and then open placed the Plant in Technical Specification 3.0.3. Only if the valve would not reopen would the Plant have to enter
- this Technical Specification."
l The PRB also noted that Technical Specification 3.0.3 should not be intentionally entered and that anytime it is entered a Licensee Event Report is required.
- In Revision 13, approved on July 30, 1986, the licensee further changed the testing procedure for check valves 8905 A B.C and 0 (referred to in L the discussion of Revision 11) to allow the testing to be performed not !
only in Mode 4 but also Mode 3. T.S. 3.5.2 requires that the intermediate head safety injection system to be operable in Modes 1,2 and The inspectors reviewed the plant operating manual revision form for :
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Revision 13 and attachments, associated Plant Review Board review l
consideration form, and minutes from the Plant Review Board meeting which -
i reviewed Revision 13. The inspectors also reviewed documents associated with Deviation 086-193 which had initiated this chang Based on the
! review of these documents, the inspectors concluded that the licensee had not documented the impact of testing in Mode 3 vs. Mode ;
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I kevistor 14 was approved on June 29, 1987; Revision 15 was approved on November 19, 1987; and Revision 16 was approved on June 3, 1988. These ,
revisions did not involve changes which would have directly raised e
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y questions of operability of redundant ECCS' trains (as in Revisions 6', 10, 12 and 13).
These items.will remain unresolved pending NRC Staff review of the ,
subject (50-344/88-40-02). Event Follow-up (93702, 62703. 92701)
Reactoa Trip on Low Reactor Codlant System Indicated Flow On August 16, 1988, with the unit at 100% power, a reactor trip occurred on indicated low flow in reactor coolant system loop "B". Instrumentation and control technicians had completed the calibration of reactor coolant system flow transmitter FT-424 and were venting the high pressure side of the differen'.ial pre sure uni The calibration was being performed in response to observed control room indications for reactor coolant system
"B" loop flow approaching channel check limits. The reactor trip bistable
'for low flow associated with FT-424 was in the tripped position to allow perfornance of-the' calibratio The venting of the high pressure side of FT-424 apparently resulted in a pressure transient in the common high pressure line shared by FT-425 and FT .426 on reactor coolant system loop -
"B". Based on review of computer data', FT-425 indicated flow momentarily dropped below the 90% setpoint and thus satisfied the two out of three reactor trip logic with the plant above permissive P-8 (39% nuclear power).
The inspectors observed plant response to tha trip; reviewed control room board indications, chart ecorder data, and co puter printouts; and discussed the event with control room operators, maintenance supervisory personnel, and plant managemen All safety systems appeared to have functioned as required. The inspectors reviewed tne completed post trip reviews and observed that items v.o be addressed prior to restart were contro11e * through the use of a plant recovery review form per administrative order A0-3-25 "Ready for Startup."
Calibration of "B" loop reactor coolant system flow transmitter FT-424 was performed per maintenance request MR-88-6557. The inspectors reviewed the completed work package and discussed the work activities with the instrumentation and control technicians, supervisors, and operations personnel involved in the wor The work instructions associated with MR 88-6557 incorporated lower tier maintenance procedure ICP-27-1, titled "Calibration of Electronic Pressure and Oifferential Pressure Transmitters," which in turn incorporated plant operating manual procedure MP-2-3, titled "Venting and Filling of Field Transmitters."
The reactor trip on August 16, 1988, occurred during the return to service of FT-424 af ter calibration. Based on discussions with the Instrumentatian and Control technicians who performed the work, the inspectors understand that the techniciant filled the transmitter from the low pressure line with the equalizing valve of the associated three valve manifold ope The equalizing valve was then closed, and the high pressure line isolation valve opened. The technicians then vented the low pressure side of the flow transmitter differential pressure unit by cycling the associated vent valv The reactor trip occurred in the next
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step when ti.e technicians vented the high pressure side of the pr_ essure transmitter. The air venting apparently resulted in a pressure decrease in the common high pressure line shared by flow transmitter FT-425 and
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FT-426 for the "B" reactor coolant system loo As discussed above, this resulted in a reactor trip on indicated low reactor coolant system flo i The work instructions of MR 88-6557 required the venting of air from the ;
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transmitter but did not provide specific instructions on valve manipulations to accomplish the refil1 and venting proces MP-2.-3 i recommended refill of the flow transmitter with demineralized vater using l The work instructione were written
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a specially adapted filling devic and implemented by the technicians in such a way that backfilling was not performed. MP-2-3 also provided guidance on air venting but, similarly to MR 88-6557 work instructions, did not provide specific instruction This is an apparent violation of 10 CFR 50 Appendix B criterion V (50-344/88-40-03). At the end of the inspection period, the. licensee was revising MP-2-3 to improve its instructions on refilling and venting of ,
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transmitter A second area of concern dealt with management and engineering review of non-routine maintenance evolutions. Calibration of reactor coolant systen flow transmitters are normally performed during plant shutdown ,
conditions. In addition during the 1988 refueling outage, design changes '
were made to the reactor coolant system flow transmitters including a change in transmitter manufacturer, relocation of various transmitters resulting in tubing changes, and replacement of threi valve manifolds for various transmitter These indicators, along with the fact that the high pressure sensing line was shared among the three flow transmitters for each reactor coolant system loop, could have indicated the need for increased management and engineering involvement in the calibration of FT-424. At the exit meeting, licensee management acknowledged the i concern and discussed the following corrective actions: (1) communicate t clearly plant management expectations on the work planning process to the involved work groups; (2) provide formal guidance on the conduct of pre- ,
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evolution hriefings; and (3) provide guidance on what constitutes non routine and high risk evolutions. The inspectors also noted that the ,
licensee's event evaluation process did not identify the lack of detail in the maintenance work instructions as a major contributing cause of the ;
reactor tri Reactor Trip Durina Replacement of PT-458 On September 16, 1988, with the plant at 100% power, a reactor trip occurred on overtemperature delta temperature due to personnel error during replacement of pressurizer pressure transmitter PT-458. PT-458 had failed itr channel check on September 13, and operators had tripped its associata bistables including the overtemperature delta temperature reactor trip histable as required by technical specification The maintenance work instructions for replacement of the failed PT-458 transmitter incorrectly directed the instrumentation and control technician to the wrong protection set cabinet. While in this event the detail of the work instructions was adequate, the quality of procedure review was not sufficient to identify incorrect instruction The technicisn then checked and tripped the bistable switch for
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. . 13 overtemperature delta temperature reactor trip associated with PT-45 This action resulted in the two of four logic for overtemperature delta temperature reactor trip being satisfie The intent of the procedure was to check that the associated bistable for PT-458 was in the tripped positio The licensee's evaluation of the event indicated that the
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technician incorrectly read bistable switch label TS-411C as TS-441 During the subsequent ferced outage, the licensee dealt with numerous issues including testing and maintenance work on pressurizer pressure and level instrumentation, the discovery of significant amounts of Asiatic clams in the service water system, two instances of locked valves being found out of position, and a depressurization transient described in the next sub paragrap The inspectors observed that corporate office personnel and plant staff worked closely together to address identified problems. A management team consisting of senior Nuclear Division managers and chaired by the Vice President, Nuclear reviewed the resolution of the significant items which were documented in action plans. NRC concerns with regard to personnei performance, procedural inadequacy, and effectiveness of pre-evolution briefings which apparently led to the September 16 reactor trip and to the control of the locked valve program were discussed in an October 7 management meeting as documented in Inspection Report 50-344/88-4 Depressurization Transient During Performance of TPT-274 On September 19, 1988, a depressurization transient occurred during the performance of temporary plant test TPT-274, titled "PT-457 Backfilling."
TPT-274 involved the back filling with demineralized water of the sensing line associated with pressurizer pressure transmitter PT-457 followed by ultrasonic testing to verify the line was filled with water. These activities were performed as a result of finding a lack of water in the sensing line for pressurizer pressure tra.nsmitter PT-456 as discussed above in paragraph 4. During the evolution the plant was in Mode 3, Hot Standby, at normal operating pressure. TPT-274 incorrectly directed the control room operators to select PT-457 as the reactor coolant system pressure controlling channe This is another example of the quality of procedure reviews being inadequate to identify incorrect vosk instruction. When PT-457 isolation valves were closed as part of TPT-274, its signal and indication failed high resulting in a depressurization transient as spray valves opene The operators responded by taking manual control of the spray valves, closing the valves and directing test technicians to terminate TPT-274 activitie Upon return to service, PT-457 was out of channel check high and was subsequently replace The inspectors reviewed control room records and discussed the transient with the involved operators on the sequence of events, system response, and operator action The first indication of the transient was a low reactor coolant pump seal water flow alarm which resulted from a decrease in charging flow as pressurizer level increased during the transient. The operators were able to terminate the transient before reactor coolant system pressure reached the reactor trip or safety injection setpoint , . _ - -_ _
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The licensee initiated an event report to investigate the causes of the '
event. Preliminary results indicated a procedural error in the selection of PT-457 as the controlling channel, inadequate technical review of e procedure, and operator failure to question the selection of PT-457 the controlling channe The pre-evolution briefing for the conduc f TPT-274 was not attended by the involved oper.' tors. NRC concerns wh h developed in part as a result of this event were shared with licensee ;
management in an October 7, 1988, management meeting as documented in *
Inspection Report 50-344/88-4 f
. Control Rod Droo During Plant Shutdown i
- 1 On September 22, 1988, the plant returned to power operation following a forced outage resulting from the September 16 reactor trip on over- ,
temperature delta temperature. The operators halted power ascension at l approximately 62% power in response to computer alarms for main turbine i bearing high vibration, and subsequently completed a rapid controlled i power reduction to below permissive P-10 at 10 percent power. The main turbine was tripped, and the reactor was placed in Mode 3, Hot Standb ;
During the power reduction, control rod F-6 in control bank "C" dropped :
from about position 215. In addition, nuclear instrumentation source [
range channel N-32 was observed to have failed low for a period of four '
minutes and then with no corrective maint a nce operated as expected.
The licensee initiated an internal event evaluat e . performed corrective ;
maintenance on the N-32 channel, and initiated troub N hooting to l
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determine the cause of the dropped control rod. The cause of the main {
turbine high vibration was determined to be an excessive heat up rate of l the moisture separator reheaters and low pressure turbine. The inspectors reviewed control room records, discussed the event with the ;
involved operations crew, and observed portions of the dropped rod I troublesh" 'an Reversed polarity was found on the F-6 control rod !
movable gr ;er coil. Licensee discussions with the vendor indicated t that rever Jd polarity on this control rod drive mechanism coil would [
1ead to sequence and magnetic field magnitude discrepancies, which in '
turn could result in a dropped control rod. Additionally, in 1978 via !
f Westinghouse technical bulletin TB-78-2, the licensee was notified that reversing the leads of the control rod magnetic coils would result in ;
reversed polarity and weakened magnetic fields and could result in i control rod problem The licensee postulated that the control rod drive i shaft for F-6 could have become worn over time at the threads associated j i with rod positions 210 to 21 During troubleshooting, control rod F-6 ;
was observed to drop from approximately the== positions when being driven
- to *Jttom. At the time of inspection, the licensee's evaluation of the <
event was continuing to investigate the cause of the dropped rod, !
including the cause of the reversed polarity on the movable gripper coil, i The licensee plans to examine the control rod lead screws for abnormal ,
q wear during the next refueling outag ;
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! Prior to plant startup, the licensee reversed '.he wires at the control j rod drive cabinets per a temporary modification to offset the reversed !
polarity condition, cleaned connector contacts for N-32, verified proper [
polarity on a number of other control rod drive mechanisms, performed a
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hot rod drop test on F-6, and instituted additional controls to prevent turbine high vibratio One violation was identifie . Follow-up of Licensee Event Reportr (92700)
LER 88-18. Revision 1 (Closed) Unfiltered In-Leakage in the Control Room , t Emergency Ventilation System: As discussed in Inspection Report 50-344/88-29, NRC review of Revision 0 to Licensee Event Report 88-18 3 concluded that the report could have more fully described the testing history and requirements for inleakage into the control room emergency ventilation system and the relationship of the event to related 1986'
enforcement action. The licensee committed to and subsequently submitted Revision 1 to LER 88-16. Review of Revision 1 by the inspectors and regional personnel indicated significant improvement over Revision 0 but rcsulted in additional questions on the evaluation of the event regarding the accuracy of smoke testing for ventilation system inleakage, the bases for using smoke testing inr,tead of vacuum testing in 1986, and the definition of system integrity as described in LER 88-18 Revision The inspectors discussed these concerns with the utility's licensing representatives who committed to another revisio Based on the above, this item is considered close Revision 2 will be reviewed upon receip . Uncontrolled Entry into Posted Radiation Protection Area (71709)
On August 17, 1988, the licensee identified an instance of failure to 1 follow the requirements of a radiation protection posting on the 45 foot airlock vestibule doo Door 205 was posted as a high radiation contaminated area. A radiation protection technician observed that a plant worker and security guard had opened the door and were working on the inside card reader. The worker and security guard were not wearing anti-contamination clothing, were not signed in on a Radiological Work Permit (RWP) and did not follow associated RWP requirements, e.g.,
dosimetry or RP technician cos . age. The worker and the guard were l
evacuated from the area and frisked. Neither person was contaminate A survey of the area showed no contamination and a general area radfation i level of less than two mrem /hr.
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The licensee initiated an internal event report to evaluate the incident, ,
Preliminary results of the event indicated that although personnel error !
to observe the radiation protection posting was the primary cause, the l applicable maintenance request had not received radiation protection j department revie Further, shift supervisor control of keys to high l radiation areas as specified in Radiation Protection Manual Procedure l RPMP-20, "Key Control for High Radiation Areas" had not prevented this j uncontrolled entr Immediate corrective action included revision of the !
high radiation area listing to specifically state that RPHP-20 applied to l the airlock vestibule door key l The inspectors' review of the event found that an emergency exit door in the immediate vicinity of Door 285 was fully ope In response, t5e licensee covered the emergency exit sign as the door served that purpose
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E only during outage In addition, the inspectors observed that a radiation protection entry control sign located in the adjacent trailer had been altered through the use of masking tape to reflect current plant operating conditions. The entry control sign applied only during plant ;
outages when the trailer functioned as a radiological controlled area '
access control poin In response, the licensee removed the sign. The -l inspectors also discussed security aspects of the incident with the
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regional specialist inspector. Based on follow-up with security personnel, the regional security inspector concluded that work activities were performed in a manner consistent with security procedures and
, requirement At the exit meeting, the inspectors shared concerns on the level of awareness and respect for safety, particularly radiation protection i signs, and postings and on the adequacy of key control for access to high '
radiation areas as required by technical specification 6.1 ;
Licensee mcnagement stated that the event was a serious violation of radiation protection requirements. With regard to safety sign compliance [
in general, licensee management considered this to be an important issue l and had ensured that safety signs had been altered as necessary to ;
reflect changing plant conditions during a recent forced outage. Licensee ;
management expressed the position that this event appeared to be an isolated instance of failure to adhere to posting requirement The licensee's evaluation of the event was not completed at the time of the ,
inspectio The inspectors discussed the event with the regional l radiation protection inspector. The licensee's completed evaluation will !
be reviewed as part of routine inspection ectivitie . Review of Systems Engineering Proaram (37702)
A review of the status and effectiveness of the licensee's System Engineering program was conducted during the current inspection perio Discussions were held with representatives of the Plant Systems Engineering Br.nch (including management, supervision and selected system :
engineers); reiresentatives of the Plant Maintenance Branch; and representatives of the Plant Operations Branch. Facility recrfds r relating to the organization and staffing of the Plant Systems ;
Engineering Branch were examined, as were procedures relating to the duties, responsibilities, and training of System Engineer The l
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following findings resulte Organization and Staffing The Plant Systems Engineering Branch is headed by a Branch Manager who reports to the Manager of the plant Technical Services depar ment, who in turn reports to the Plant General Manager. System Engineers are assigned within the Systems Engineering Branch to one of three Unit Supervisors responsible for Primary Plant, Control &
Electrical, and Balance of Plant system The current (1988)
authorized staffing level provides for 25 permanent System Engineer positions, assigned essentially equally among the three Unit Supervisor This staffing level represents an increase of four positions over the 21 positions authorized for the year 198 . .
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l According to plant management representatives, it is anticipated 1 that two to four additional positions will be authorized for the l calendar year 198 .)
At the time of the inspection 22 of the 25 authorized System Engineer positions were fille Approximately half of these positions were occupied by individuals who had been assigned to the position for periods ranging f.'om approximately one to nine month Licensee management stated that i. hey are actively involved in recruiting efforts to fill the three vacant positions which currently exist, and that five contre t engineers are also currently assigned to the Branc Licensee rept esentatives also expressed the view that, although for a variety of reasons difficulties have been experienced in the recruiting and retention of System Engineers over the past approximately two years (the Systems Engineering program was initially authorized during 1986 - See Inspection Report N /86-48), efforts over the past several months have been successful in recruiting qualified and experienced engineers to-staff the Plant Systems Engineering Branch. An examination of biographical records and discussions with a number of recently hired System Engineers by the inspector revealed this to be the case, Qualification and Training of System Engineers Discussions with systems engineering personnel revealed that nine individuals within the Plant Systems Engineering Branch, including the Branch Manager and two Unit Supervisors, have certification and/or qualification as licensed plant operators or Shift Tech.1ical Advisors (STA). An additional two individuals are soon to complete an approximately year long licensed operator training program. Two individuals are scheduled to participate in the same training program commencing fn early 198 Facility records indicated that' essentially all of the staff of the Systems Engineering Branch are currently participating in a Technical Manager / Technical Staff Training Program administered by the site Training Branc Plant Engineering Procedure PEP 10-6, "System Engineer Checkouts",
dated November 2, 1987, described the method for System Engineers to attain a base level of knowledge of plant systems to which they are assigned. The inttial phase of the program described by this procedure, Familiarization and System Overview, involves the self-study of documents applicable to assigned systems (FSAR Sections, training system descriptions, technical specifications sections, plant operating manual procedures, etc.) and system training walkdowns. This phase of the program culminates in the satisfactory completion of a "training checkout" from a representative of the plant Operatiuns Branc During discussions with selected System Engineers the status of thef t completion of self-study and training checkouts in accordance with PEP 10-6, was discussed. It was observed that checkout forms from PEP 10-6 were utilized to record the completion of self-study
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elements of this phase of tk.e program as well as system training walkdowns and training checkouts by the Operations Branch. Procedure
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PEP.10-6 states that portions of System Engineer checkouts may be deleted by the Unit Supervisor based on previous system experience or licensed operator /STA trainin During discussions with the inspector, the Manager of the Plant Systems Engineering Branch indicated that completion of the program described in PEP 10-6 had been assigned high priority for System Engineers following. completion of the recent refueling / maintenance outag This prioritj* assignment had been made in recognition of the fact that several individuals had been assigned as System Engineers over the past several months, and that assignments in support of outage work had been made to these individuals on a priority based upon their previous work experience and '
qualification The inspector did observe that Plant Engineering Procedure PEP 40-3, "System Engineer Program", dated June 20, 1987, does state that system engineers should complete the program ,
described in PEP 10-6 for "...the major systems assigned to them, '
however this is not required prior to performing system engineering functions." Discussions with individuals, including selected System Engineers within the Plant Systems Engineering Branch, revealed no instances where this provision of PEP 40-3 had been abused in the assignment of work tasks to System Engineers. The inspector did, ,
however, emphasize the need for licensee management to continue to
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place high priority on the completion of the program described in PEP 10- l System Engineering Functions and Effectiveness The duties and responsibilities of System Engineers are described in Plant Engineering Procedure PEP-40- This procedure lists the responsibilities of System Engineers in the following "approximate" ,
order of priorit ) Support of Plant Operations and Maintenance, 2) Monitor and Evaluate System Performance, !
3) Review and Revise Documents Associated With Assigned Systems, and -
4) System Support Activities such as assisting in the !
identification of procurement requirements and the review of *
spare parts inventor The inspector met with selected System Engineers to discuss with them examples of activities in which they have been involved in those areas listed above. Olscussions were also hela with managers !
and supervisors from tne plant Operations and Maintenance Branches in an effort to assess the quality and effectiveness of system engineering support to these organization These discussions resulted in the following finding _- __ ___________ __ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _
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Representative of the plant Operations organization described what they viewed.as an increasing level of technical support by the Systems Engineering B .ich over the past year. Areas of particular support which thry discussed included recommendations for procedure enhancements and the resolution of system operating problems. The coordination of trouble shooting of plant system problems was described as an area where "major contributions" have been made by plant System Engineers. These individuals also expressed what they felt to be a healthy working relationship between the Operations and System Engineering Branche This relationship, they felt, had grown from an increasingly strong respect for the competency and knowledge of System Engineers in those systems to which they are assigne The latter relationships has grown stronger in the past year as stability within the Systems Engineering Branch has improved, lessening the burden on the Operations organization to
"train" newly assigned System Engineer Representatives of the plant Maintenance organization expressed the i
view, similar to representntives of the Operations Branch, that an increasingly healthy and effective relationship between the Mainterance and Plant Systems Engineering branches has evolved over the past yea They too indicated that stability (lack of attrition) within the Systems Engineering Branch had been an important contributing factor to improved performance by the Branch over the past yea In general, there appeared to be a consistent perception by both the representatives of the Plant Systems Engineering and Maintenance branches as to the nature of problems for which each would have lead responsibilit The perception was that problems at the component level were generally the responsibility of the Maintenance organization, with technical support provided by engineers within the Paintenance Support Unit. Likewise, it was generally agreed that lead responsibility rested with the Systems Engineering Branch where problems were of a system performance natur Representatives of both branches were quick to point out that this general perception (or rule) was by no means rigi They cited recent recurring problems experienced with Barton transmitters as an example where maintenance assumed the lead initially. As the problems recurred and were recognized to be of a generic nature lead responsibility was assumed by the Systems Engineering Branc Licensee representatives stated that formal Action Plans have become a more frequently utilized mechanism to identify and pinpoint individual and organizational responsibility for significant problem resolutio Discussions with individual System Engineers revealed instances of a close working relationship between the Systems Engineering Branch snd the Nuclear Plant Engineering (kPE) organization at the corporate offic Two examples discussed involved instances where plant System Engineers were assigned essentially full time for periods of three to four months to work in the offices of NP These instances involved the preparation of plant design changes or
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-'t modifications for which the individuals were assigned as lead plant 4 enginee . Unresolved Item An unresolved item is a matter about which more information is required to ascertain whether it 10 an acceptable item, a deviation, or a
. violatio An unresolved item'is documented in paragraph . Exit Interview (30703)
The inspectors met with the licensee representatives denoted in paragraph 1 on September 30, 1988, and with licensee management thro 4hout the inspection perio In these meetings the inspectors suumarized the scope and findings of the inspectiun activitie I l
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