IR 05000309/1989008
| ML20244E284 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 06/07/1989 |
| From: | Lester Tripp NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20244E283 | List: |
| References | |
| 50-309-89-08, 50-309-89-8, NUDOCS 8906200248 | |
| Download: ML20244E284 (10) | |
Text
{{#Wiki_filter:___ . ' . - . l: U.S. NUCLEAR REGULATORY COMMISSION j Region I Report No.: 50-309/89-08 License No.: DPR-36 Licensee: Maine Yankee Atomic Power 83 Edison Drive Augusta, Maine 04336 Inspection At: Wiscasset, Maine Conducted: April 8, 1989 to May 17, 1989 Inspectors: Cornelius F. Holden, Senior Resident Inspector Ri ard J Fre enberger, Resident Inspector He g , P oject Manager, NRR Ke hk h1 Approved By: < o#well _.Tripp(fChief D te l-Reactor Projects Section No. 3A ~ Summary: Inspection on April 8,1989 to May 17,1989~ TReportNum_ber 50-309/89-08) Areas Inspected: Routine resident inspections of plant operations including: followup on previous inspection findings, operational safety verification, maintenance, surveillance, physical security, radiation protection and fire protection.
Resuly : The inspectors concluded that the plant continues to be operated safely. Two issues which need to be addressed are the applicable standards for verification and validation of the microprocessor based Primary Inventory Trend System and the uncertainties of the Subcooling Margin Monitor system including how those inaccuracies impact the Emergency Operating Procedure actions.
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. . is l - . L TABLE OF CONTENTS Page 1.
Persons Contacted.................................... ......
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Summary.of Facility Activities................ .............
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Operational Safety' Verification (IP 71707)..................
4.
Plant Maintenance.(IP 62703).............. .................
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Surveillance Testing (IP 61726)..............................
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Inadequate Core Cooling Instrumentation (IP 71707)...........
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Observations of. Physical Security (IP 71707)...............
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Radiological Controls (IP 71707)............................
9.
Nuclear Safety Audit and Review Committee (IP 40500).......
10.
Exit Interview (IP 30703)............................ ......
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Persons Contacted Within this report period, inted~;;: and discussions were conducted with various licensee personnM, including plata. operators, maintenance tech-nicians and the licensee's management staff.
2.
Summary of FacG1ty Activities The plant was at 100 percent power at the start of this inspection period.
On April 29, 1989, the plant reduced power to 90 percent power for turl ie valve ; testing and other surveillance activities.
The plant retur-to 100 percent power and remained there for the rest of the inspection period.
3.
Operational Safety Verification On a daily basis, during routine facility tours the following were checked: mannfng, access control, adherence to procedures and LCO's, instrumentation,. recorder traces, protective systems, control room annun-ciators, radiation monitors, emergency power source operability, operabil-ity of the Safety parameter Display System (SPDS), control room logs, shift supervisor logs, and operating orders. On a weekly basis, selected Engineered Safety Features (ESF) trains were verified to be operable. The condition of the plant equipment, radiological controls, security and safety were assessed On a biweekly frequency the inspector reviewed a safety-related tagout, chemistry sample results, shift turnovers, portions of the containment isolation valve lineup and the posting of notices to workers.
Plant housekeeping ard cleanliness were also evaluated.
The inspector observed selected phases of the plant's operations to deter-mine compliance with the NRC's regulations. The inspector determined that the arcas inspected and the licensee's actions did not constitute a health and safety hazard to the public or plant personn.el.
Inspector review of the Temporary Modification Control Log identified an errer in the master. index associated with request No. 88-91.
The tempor-ary modification was authoriled on October 17, 1988, near the beginning of the refueling outage and involved the replacement of a 15-amp breaker with a 70-amp breaker in a spare cubicle in Metor Control Center (MCC) 11B, a non-safety related MCC.
The order also included the installation of a cable from the breaker cubicle to a temporary power stand in the yard and the steam generator sludge lancing trailer.
The evaluation included on the request fcrm considered operation with the temporary modification in place with the plant shut down only. The request form properly identified - - - _ _ _ - _ - _ _ _ - - -
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l , + that the temporary modification be removed prior to plant startup; how-h ever, this information was incorrectly transferred to the master index.
The master index stated that the tempc rary modification need not be ! restored until the completion of contractor use.
As a result of this l error the temporary modification was in place after plant startup. When i notified of the discrepancy the' licensee took immediate action to remove ! the temporary modification.
Further review by the inspector identified no other examples of temporary l modifications that s; ould have been removed prior to plant startup but I that were still in' place. The safety significance of this et ror was con-l - ' sidered.to be minimal primarily due to the fact that the error involved ! non-safety related equipment.
However, the inspector expressed concern ! that the error was not_ detected by licensee personnel in the'approximately four months since plant startup following refueling.
The inspector had no further questions.
4.
Plant Maintenance i l The inspector observed and reviewed maintenance and problem investigation l activities to verify compliance with regulations, administrative and main-i tenance procedures, codes and standards, safety tag use, equipment align- -{ ment, jumper use, personnel qualifications,. retest requirements, and i deportability per Technical Specifications.
Specifically, the following maintenance activity was inspected.
! Deficiency Date Report Number Description i 5/5/89 1163-89 DG-2 exhaust expansion joint replacement Additionally, the inspectors monitored the corrective actions that the ! licensee took when they identified that valve DR-A-6, Drain Header Flow ' Control Valve, c uld not shut. DR-A-6 had been opened earlier in the day to reduce radiation exposure levels for other planned maintenance. When j DR-A-6 was returned to its normal shut position, it did not indicate
closed. The plant declared DR-A-6 inoperable. A Special Plant Operations j > Review Committee meeting was held to discuss the issue. DR-A-6 was i so-j lated in accordance with Technical Specification 3.11.B. Two other isola-l tion valves which were tagged shut were located in a high radiation area i and special precautions were utilized to minimize the radiation exposures l for the personnel performing the tagging.
The DR-A-5 remains out of ser-
vice until maintenance can be performed.
The inspector determined that .l the licensee was conservative in their actions to compensate for the loss j of valve DR-A-6.
l The inspector had no further questions.
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5.
. Surveillance Testing The inspector observed parts of tests to assess performance in accordance with approved procedures and LCO's, test results, removal and restoration of equipment, and deficiency review and resolution.
The following surveillance were reviewed: Date procedures Number Title 5/4/89 3.1.4 Emergency Diesel Generator - Monthly Surveillance Test 5/17/89 3.17.5.1 Surveillance Testing of Balance of Plant Charcoal & HEPA Filters The inspector witnessed the testing of the Emergency Response Facility (ERF) ventilation system in conjunction with procedure 3.17.5.1.
This test was conducted in response to concerns raised in Inspection Report 50-309/89-04.
During that inspection the inspector was unable to deter-mine if the ventilation system would align as required for emergency con-ditions and whether the system could perform its intended function.
In order to resolve this concern, the licensee provided access to the dampers that realign during emergency conditions so that their position could be verified as part of the test.
Additionally, the licensee's contractor mapped the flow conditions for several system lineups to verify that the required flow rates were achieved under different flow conditions.
The l inspector witnessed the damper arrangements and flow mapping. The licen-see is reviewing the test results to determine if further action is ' required-The inspector questioned the adequacy of the overpressure sup-plied by the system in the emergency recirculation mode in light of the fact that some personnel would be stationed in areas not supplied by the ERF ventilation system.
The licensee has agreed to review this area.
The inspector had no further questions.
6.
inadequate Core Cooling Instrumentation A review was conducted of the Inadequate Core Cooling Instrumentation (ICCI) systems and their use at Maine Yankee.
The licensee utilized several systems in order to provide the required information to the oper- ' ators. These systems included the Core Exit Thermocouple (CET's) which measured. exit temperature from the reactor core, the Subcooling Margin Monitor (SMM) which calculated a difference between an input temperature and the saturation setpoint and the Primtry Inventory Trend System (PITS) which monitored the level in the reacte r vessel during accident condi-tions. This inspection centered on three areas: installation, calibration and procedures and training.
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Installation The installed configuration of the CETs, SMMs, and PITS was observed , in the control room during plant operation..The output parameters for these systems were displayed on a mimic of the reactor's primary . coolant system.
The location and labeling of these instruments was ' clear. The coolant levels displayed by PITS were related to a dia-gram of the reactor vessel on the control board. During plant oper- < ation, only the operation of the CETs and SMHs could be observed.
One SMM channel displayed the subcooling margin based on the head temperature, while the second channel showed the subcooling margin based on the core exit temperature.
The procedures generally were specific about which channel-was to be considered at a certain pro-cedural step. The licensee stated that the current goal in revising - the procedures was to eliminate reference to the head temperature and the associated head subcooling margin.
A seal pot was designed and fabricated by the licensee to provide space in the PITS for noncondensable gases without affecting the sys-tem's overall operation.
The seal pot consists of an extended ~ horizontal pipe with sufficient volume to hold the expected amount of gas. Because the volume is horizontal, it causes only a slight error in the system differential pressure, and therefore does not affect-the calculated coolant level.
The seal pot is removed at each-refueling, when the connections to the reactor head are removed.
When the unit is reinstalled,.it is completely refilled with reactor coolant and any gas is effectively removed.
This is in accordance with the licensee's description of the system.
Components of the SMMs and PITS within the containment are required to be qualified for that environment by Regulatory Guide (RG) 1.97 and 10 CFR 50.49. The licensee stated that all applicable components for these systems were included in the plant's Environmental Qualifi-cation (EQ) program. The only exception was the Resistance Thermal Detectors (RTDs) used for the PITS reference leg temperature compen-sation. At the time of the inspection the preliminary Qualification Document. Review (QDR) packages for these RTD's were under review at Yankee Atomic by the EQ Coordinator.
The SMMs and PITS both are Class IE systems with outputs to control grade systems.
The licensee stated that NRC approved isolators are.
used to perform the isolation function.
The licer also stated that approved isolators are used to provide separatton between the trains which use the common reactor head resistance temperature detectors.
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The inspector reviewed the basic standards of construction applicable to these systems.
Key emphasis was on obtaining appropriate stand-ards for seismic resistance of the installed hardware including the piping added for the PITS.
Approj'riate standards for a plant of Maine Yankee's original construction practices were followed.
Al- + hough RG 1.97 requested that redundant trains t,e separated in ac- . cordance with RG 1.75, the licensee had requested to use the separa-tion criteria contained in the Maine Yankee Final Safety Analysis Report (FSAR) section 7.3.8.
The NRC staff has not objected to this alternative standard.
The insta'llation appears t0 conform to this standard. Photographs of the containment instt11ation were reviewed, and the installation behind the main control board was also reviewed.
To the extent that this inspection examined critical areas of the system that cculd be subject to common mode failure, separation and standards of construction are considered acceptable.
b.. Calibration la the licensing submittals which described the Subcooling Margin Monitor (Si4M) and the Primary Inventory Trend System (PITS), the licensee noted these were microprocessor-based instrumentation sys-tems. These systems are categorized as class IE by the licensee for post accident monitoring. The inspector asked the licensee how these systems were evaluated for correct performance initially and during subsequent surveillance testing. The licensee described a surveil-lance procedure which involved traditional approaches to evaluate the analog portions of the instrumentation system.
The microprocessor portions of the system were checked with a simulation of the analog inputs over the range of expected values and direct validation that the analog or digital outputs were in agreement with expected values.
The inspector agreed that the overall checks were adequate for an-analog system, but noted that the NRC staff's approach to verifica-tion and validation of software under RG 1.152 nnd Institute of Elec- . trical and Electronics Engineers Standard 7432 (IEEE 7432) were more stringent for microprocessor based systems used in class IE applica-tions.
Application Criteria for Programmable Digital Systems in Safety Systems of Nuclear Power Generating Stations, IEEE 7432, describes a process for separately determining hardware, software, and system integration requirements. Each step must be independently verified before the software is written and the system integration performed.
These steps must also in turn be independently validated before the overall system valid? tion takes place.
The verification and validation requirements for Class IE software applied at the time i _ - __ -__ __ _ _ _ _ - _
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. that Maine Yankee purchased the PITS and applied for staff.'approvai of this.' system..(The licensee's. position was that -the staff had not j - requested or that Maine Yankee had not specifically adopted these - standards.) The licensee supplied proprietary vendor documents dur-ing'the course of the inspection which addressed the safety qualifi- . cations of the vendor supplied system.
However, this. material did g not supply a complete evaluation.against the referenced standard.
X This item requires further staff and licensee. action for resolution.
At the inspection exit meeting, the licensee expressed a willingness ". to meet with the' staff and review the system design 'and standards utilizod.
The licensee. stated that PITS had an extensive self test capability - built-in - based on the use of a microprocessor.
Diagnostic self checks of all components were performed at regular intervals to assure overall system performance. Microprocessor errors would cause the system output to go either high or low. Diagnostic error lights were also available on the instrument cards and were accessible to the service technician. Based on the above, the inspector. concludes the PITS has the desired self-test capability.
The. licensee conducts surveillance calibrations for the SMM and PITS in accordance with the plant Technical Specifications.
At the re-fueling interval surveillance, the licensee conducts a primary cali-bration of the pressure sensors against known test pressures.
Tem-- perature sensor calibrations are based on manufacturer's calibration curves.
The licensee is evaluating how to better explain common-error factors for the Core Exit Thermocouple (CET). The balance of the analog signal channels are calibrated as a normal analog system.
The licensee's methods of enlibrating the digital portion of the system is discussed 6bove.
The ' inspector concluded that the calibration of SMM and PITS is adec;uate.
f.:. Procedures and Training The licensee's current Emergency Operating Procedures (EOP) and Funce tion Restoration Procedures (FRP) were reviewed to determine how the SMM and PITS were utilized by the operators. Examination of the pro-cedures focused on the SMM s.etpoints to be used by the operators for making significant decisions, including termination of safety injec-tion and operation of the reactor coolant pumps. The inspector noted h . ' _ - - _ _ - -. _ _ - _ _.. _. _ _ _ - _ - - - - -
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s, 7' . , that the values used in the procedures were not consistent..In particular, the inspector noted that the values cited.in the proced-ures did not reflect the licensee's latest evaluation of. the overall SMM instrument uncertainty.
For example, in' the March 14, 1986, background; document' for the E0P's the-basis given for securing the - reactor coulant pumps was 25 degrees F of ; core region subcooling margin to allow for instrument uncertainty.
This value is still reflected in the current procedures although the estimate. of instru-ment uncertainty has changed'at least twice.
When discussed with. the licensee staff, there was agreement that the , . terrent E0P's and FRP's were not consistent with latest understanding.
of the SMM ' instrument system uncertainty.
The. inspector observed that the E0Ps and FRPs had to be consistent with both the licensee's R relevant safety analyses and the known characteristics of the instru- ! mentation to' assure that the operator's actions were conservative-The - licensee stated that a ' procedure. setpoint upgrade was already underway, but. would give these findings. immediate attention.
This item is considered unresolved pending licensee actions to update the E0P's and FRP's (UNR 50-309/ 89-08-01) and will.be reviewed in future inspections.
Review of"AOP 2-90-2 revealed that in the event of a fire in tht ton-( tainment or adjacent cable penetration areas, the operator 'would be shut' down the plant in accordance with E0P E-0.
Ec0 required the operator to consult the SMMs, although oc caution 'was noted about their acc'uracy.
By contrast, AOP 2-90-2_ required that the operator - verify the post fire accuracy of other instruments.
The inspectw noted ' that this could lead the operator, to take an inappropriate . action such as securing the reactor coolant pumps, because of. fire damage to the SMMs in the fire area. The licensee s.tated that these procedures would be reevaluated to see if this issue should be corrected.
The current traintag program covered the use of the CEts, SMMS and PITS. Some aspects of the recent upgrades to these systems had not. been implemerited in the plant simulator. However, the plant simul 6-tor could output all aspects of the displays by-utilizing the Safety Parameter Display System. Training covered all aspects cf the use of these ' systems ' as specifically referenced in the plant pr6cedures.
The training staff member interviewed had a good overall understand-ing of these systems and their role in plant operation. The inspec- ' tor found the training area acceptable.
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. 0bservatiord olphysied Spupity .. , Checks were. made to determine whether security conditions met regulatory requirements, the physical security plar., and approved procedures.
Those-checks fncluded security staffing, protected and vital area barriers, L.
. vehicle searches and. personnel identification, access control, badging, i and compensatory m 9 sures when required. The inspectors continue to monitor the licensee's upgrades to the Security program.
B.
Radiclogic61 Couttols Radiological controls were observed on a routine basis during the report - ing period.
Arear -reviewed included exteraal radiation exposure control and contamination control. Standard industry radiological work practices, , conformance to radiological control procedures and 10 CFR Part 20 require-T- ments wert observed.
Independent turveys of reciological boundcries and [' random surveys of nonradiological points throughout the facility were taken by the inspector.
The inspector noted that the Morning Managers Meeting has hen focusing attention on the plant's ALARA goals.
These goals were discussed during saveral meetings, Even though the plant's goal of 8B Man-kem exposure appears as though it will be met, the Managers I Meeting continues to look for other ways to reduce exposures.
Through their discussions, actions have been initiated to control the exposure received auring Poutinr activities. Additionally, emphasis on the control of contaminated areas and the reduction of radiological areas continues to receive upper monaseinent attention.
The inspector had no further comments.
9.
Nuclear Safety Audit and Review Committea De May 9,1989, the inspectors attended a portion of the Nuclear Safety Audit and Review (NSAR) Committse meeting. This was a regularly scheduled meet'ing.
Issues discussed included recent plar.t incidents, status of the security program improvements, followup of issues identified during the recent refueling outage and review of recommendations and action items from previous meetings. The committee members present ensured that pre-sentations were of adequate dettil to allow proper consideration of issues.
Inspector review of the outstanding NSAR issues identified no long term items which appeared to require priority attention. The inspec-tors had no further questions.
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Exit Interview Meetings were periodically held with senior facility management to dhcuss the inspection scope end findings. A summary of findings for the report period was also discussed at the conclusion of the inspection.
The licensee did not identify 2.790 material.
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