IR 05000029/1986005

Insp Rept 50-029/86-05 on 860312-0609.No Violation Noted. Inadequacy Noted:Failure to Comply W/Tech Specs Re Nuclear Instrumentation Channels.Major Areas Inspected:Maint & Surveillance Observations

ML20212A298
Person / Time
Site:	Yankee Rowe
Issue date:	07/18/1986
From:	Elsasser T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20212A280	List: IR 05000029/1986005 ML20212A278
References
50-029-86-05, 50-29-86-5, NUDOCS 8607280155
Download: ML20212A298 (17)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /86-05 Docket N Licensee N DPR-3 Licensee: Yankee Atomic Electric Company 1671 Worcester Road Framingham, Massachusetts 01701 Facility Name: Yankee Nuclear Power Station Inspection at: Rowe, Massachusetts

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Inspection Conducted: March 12 - June 9, 1986 Inspectors: H. Eichenhol Senior Resident Inspector W. Raymo , S nior Resident Inspector - Vermont Yankee Approved By: /- "' -

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T. Elsesefr, Chief, Reactor Projects Section 3C Date Inspection Summary: Inspection on March 12 - June 9, 1986 (Report No. 50-29/86-05)

Areas Inspected: Routine onsite regular and backshift inspection by resident in-spectors (196 hours0.00227 days <br />0.0544 hours <br />3.240741e-4 weeks <br />7.4578e-5 months <br />). Areas inspected included review of licensee action on pre-vious findings, operational safety verification reviews, review of radiological controls, reviews of events requiring telephone notification to the NRC, review of plant events, maintenance observations, surveillance observations, review of periodic and special reports, Plant Operations Review Committee activities, annual retraining, Siemens Allis BQ frame circuit breakers and licensee plans for coping with strikes.

, Results: No violations were inspector identified; however, one inadequacy involving failure to comply with a Technical Specifications (TS) action statement involving nuclear instrumentation channels was classified as a licensee identified violation (Section 7). Three areas needing increased licensee attention were (1) handling and protection of Safeguards Information (Section 4), (2) establishment of a formal mechanism to control Radioactive Material Areas (Section 5), and (3) evaluation of the need to increase awareness of TS surveillance requirements and to provide procedures to document their accomplishment (Section 7). Positive trends were ob-served in the development and use of safety evaluations when performing maintenance (Section 8) and performance trending of surveillance results (Section 9).

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DETAILS Persons Contacted Yankee Nuclear Power Station 8. Drawbridge, Assistant Plant Superintendent T. Henderson, Technical Director N. St. Laurent, Plant Superintendent The inspector also interviewed other licensee employees during the inspection, including members of the Operations, Radiation Protection, Chemistry, Instru-ment and Control, Maintenance, Reactor Engineering, Security, Training, Tech-nical Services, and General Office Staff . Summary of Facility Activities At the completion of the last resident inspection period on February 14, 1986, the plant was at 100% of rated power. The plant maintained that power level until March 22, 1986 when a planned load reduction to 70% of rated power was conducted by the licensee to perform maintenance on the Nos. 1 and 3 boiler feedwater pumps (replace inboard and outboard packing) and to perform turbine trip and control valve and non return valve exercises. The plant was returned to 100% of rated power on March 23, 1986, and remained essentially at that power level until May 30, 1986. On this date a plant emergency load reduction to 78% of rated power was initiated by the plant operators when a loss of cooling water to the generator hydrogen cooler occurred. A rapid recovery ensued, and within one hour the plant's power output was increased. The plant was returned to 100% of rated power until a low steam generator level reactor scram and turbine trip occurred on June 1, 1986 due to a loss of both heater drain pumps during a severe lightning storm. The plant trip occurred following 152 days of continuous on-line operations. A plant startup was initiated on June 3, 1986 and the plant operated at 100% of rated power from June 6, 1986 until the end of the inspection perio During the inspection interval the dose equivalent iodine (DEI) level was determined to be within 4-8 percent of the Technical Specifications (TS)

limit. An exception to this condition occurred in June 2, 1986, when it was j determined that the DEI level increased to 135 percent of the TS limit. This

! event is discussed in Section 7 of the inspection repor . Licensee Action on Previous Inspection Findings (Closed) Unresolved Item 84-17-01: Completion of Bus 4-1 Repairs. The in-spector reviewed plant operation and test records, and noted that Bus 4-1 was i repaired and tested prior to the plant startup on August 14, 1984. The com-pletion of this action by the licensee met the requirements of Technical Specification 3.8.2.1 regarding operable electrical power supplies. This item is closed.

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(Closed) Unresolved Item 84-17-02: Operation of the Electrical Protection Re-lays. The licensee's engineering evaluation and calculations regarding the protection scheme for the station electrical systems was provided in an in-ternal memorandum to J.D. Haseltine dated August 9, 1984. The evaluation in-cluded a review of the location and settings of the protection relays in the fault current path, and a review of the design and operation of the relay schem In spite of the fact that the relay scheme did not clear the Bus 4-1 fault prior to failure of breaker 448, the licensee concluded that the pro-tective relay scheme operated correctly, according to design, and provided adequate electrical protection for the station. No changes to the protection scheme were deemed warrante The inspector noted that the fault protection scheme did clear the Class IE electrical circuits from the fault and thereby assured the availability and operability of safety class equipment needed to ensure safe shutdown of the plant. No inadequacies were identified in the licensee's evaluation. This item is close (Closed) Unresolved Item 84-17-04: Subsequent Recovery Actions. This item was open pending completion of licensee actions to repair Bus 4-1, complete post-repair testing, identify and correct the bus failure mechanism, evaluate the operation of the protective relaying, verify the operability of low pres-sure safety injection (LPSI) pump No. 2 , verify the adequacy of the fault protection design, and report the event per 10 CFR 50.7 The inspector determined, based on a review of plant records and discussions with the maintenance manager, that the licensee had addressed the above item Following completion of repairs, Bus 4-1 and its associated breakers and equipment were tested per procedure OP 5000.168 on August 13, 1984. All tests were completed satisfactorily to demonstrate that the Bus 4-1 and associated breakers, controls and protection systems functioned properly. LPSI pump N was demonstrated operable by testing per procedure OP 4204 on August 14, 1984. Failure of the finger springs was the most probable cause of the Bus 4-1 faul Actions were completed to replace the finger spring package on breakers Nos. 4548, 548 and 648 with the upgraded design recommended by the vendor. New DB-50 breakers with upgraded finger springs were installed in breakers Nos. 448 and 4548. The 480 volt switchgear and associated components, aside from Bus 4-1, affected by the fault were functionally tested per proce-dure OP 5000.168 on August 13, 1984 and found to be satisfactor The test included a verification that bus fault and undervoltage protection was oper-abl Upon completion of the above actions, Bus 4-1 was returned to service and has operated satisfactorily during subsequent plant operations. The licensee re-ported the event per 10 CFR 50.73(a)(2)(x) as License Event Report (LER) 84-1 The inspector reviewed the LER and found that it accurately described the failure, the impact on plant operation, and subsequent licensee actions to correct the cause of the faul This item is closed.

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4. Operational Safety Verification Reviews Daily Inspection During routine facility tours, the inspector checked the following items:

shift manning, access control, adherence to procedures and limiting con-ditions for operation (LCOs), instrumentation, recorder traces, protec-tive systems, control rod positions, containment temperature and pressure, control room annunciators, radiation monitors, radiation monitoring, emergency power source operability, control room and shift supervisor log, tagout log, and operating orders. No inadequacies were identified except as noted belo In past inspections, the inspector noted the excellent performance of the licensee in maintaining the control room annunciators in as close to a " black-board" status as possible. Past observations of spurious safety injection tank level alarms have been resolved by the installation of a time delay feature on the annunciator, which has satisfactorily precluded the condition from recurring. The last spurious alarm, neutron shield tank temperature, which sometimes actuates when moving a control rod, is being investigated by the licensee for additional corrective actio In general, performance by control room personnel in maintaining required logs and records has been improving. However, a consistent high level of performance has not been achieved. Recognition by control room personnel of applicable Technical Specifications (TS)

action statements must be recorded in the appropriate logs. Con-tinued Operations Department management attention to this matter has been observed by the inspector, as evidenced by the issuance of Special Orders dealing with this issue and expected performanc The status of plant systems and equipment aligned for routine opera-tions at power was reviewed for the period from April 16-18, 1986 to verify that Technical Specifications requirements were met. Shift staffing and standby emergency systems were reviewed and found to meet applicable requirements. The inspector noted that charging pump No. 1 (one of three) was out of service for repairs due to a cracked casing. No LCO action statement applied since the Technical Speci-

, fications require only two of the three pumps to be operable at that tim The status of degraded equipment was reviewed on April 16. 1986 with the duty shift superviso Containment isolation valve, TV-214,

! for which maintenance request 86-587 was written, addressed a prob-lem with the operation of the "B" system air solenoid operato The licensee noted that air leaked past the seat of the solenoid valve when it was actuated. Based on a review of the nature of the problem, and the design and operator of the isolation valve, the inspector concurred with the licensee's determination that the de-l t I

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graded solenoid valve did not adversely impact the isolation func-tion served by TV-214. Thus, the isolation valve was degraded, but operable. No inadequacies were identifie b. System Alignment Inspection Operating confirmation was made of selected piping system trains. Access-ible valve positions and status were examined. Power supply and breaker alignments were checked. Visual inspections of major components were performed. Operability of instruments essential to system performance was assessed. The following systems were checked:

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Low and high pressure injection systems verified during tours of the safety injection building and during a control board status revie Non-return valves (NRV) verified during tours of the NRV platform and during a control room equipment cabinet revie Emergency diesel generator (EDG) units verified during tours of the EDG rooms and during a control board status revie Charging system verified during control board status revie No inadequacies were identifie c. Biweekly and Other Inspections (1) During plant tours, the inspector observed shift turnovers, compared boric acid tank samples and tank levels to Technical Specifications requirements, and reviewed the use of radiation work permits and radiation protection procedures. Area radiation and air monitor use and operational status were reviewed. Verification of tagouts indi-cated the action was properly conducted. There were no inspector identified deficiencies in this are (2) Observations of Physical Security On April 4, 1986, the inspector was reviewing control room log en-tries and noted that descriptive information, pertaining to security system deficiencies, was being logged by shift personnel that was probably of a Safeguards Information nature. This practice was re-viewed with the licensee's security supervisor, who acknowledged the inspectors comments and concerns. The inspector verified that corrective measures were immediately implemented to resolve the deficiency. Recurring problems involving the handling and protection of Safeguards Information by licensee personnel requires additional management attentio No additional deficiencies were noted by the inspector in this area during the inspection period.

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(3) Fire Protection and Housekeeping

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The inspector reviewed actions taken to meet the requirements of Technical Specification 3.7.11 on April 15-16, 1986 regard-ing fire stops for the switchgear room. A fire watch was es-tablished on April 15, 1986 following discovery by plant workers that no fire stop existed in the bus duct that runs between the two halves of 2400 volt Bus No. 1. Subsequent lic-ensee review determined that the 1/8-inch steel bus duct has a three-hour or greater fire rating, and thus, the fire pro-tection boundary / envelope for the switchgear room was not com-promised without fire stop No inadequacies were identifie Licensee activities in this area during the inspection period included routine conduct of backshift fire drills. On May 19, 1986, at 4:10 p.m., a failed fire detector at the non-return valve platform area resulted in activating the fire brigade to investigate the alarm. Once the nature of the alarm was established, the fire brigade was secured and a once per-hour fire patrol established in accordance with TS requirement No inadequacies were noted regarding licensee housekeeping practices, which continues to reflect a high level of perform-ance by the entire plant staf . Review of Radiological Controls Radiological controls were observed on a routine basis during the reporting period. Standard industry radiological work practices and conformance to radiological contrcl procedures and 10 CFR Part 20 requirements were observe Independent surveys of radiological boundaries and random surveys of nonradio-logical areas throughout the facility were taken by the inspecto Inspector observations and review of an event relating to radiological con-trols are contained below:

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During a review on May 6, 1986 of the radiological conditions in the fan room located in the radiation control area the inspector observed that some of the hoses and electrical cables stored in the room were affixed with radioactive material tags. The entrances to the room were not posted as a radioactive material area (RMA). The status and/or need to post the

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room was discussed with a radiation protection department (RPD) repre-sentative. Because the RPD representative could not readily provide an

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, evaluation as to the nature and extent of existing radioactive material in the area, the room was posted. A subsequent licensee analysis demon-strated that the room did not need to be posted as a RMA. However, this example demonstrates the need for the licensee to establish a more formal mechanism to control RMAs. In addition, the RPD representative acknowl-I h

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edged the inspector's concern that this type of condition is not being identified by his RPD shift technicians on their plant tours. The in-spector had no further questions on this matte No other deficiencies were identified in this are . Review of Events Requiring Telephone Notification to the NRC The circumstances surrounding the following events, which required NRC noti-fication via the dedicated ENS-line, were reviewed. A summary of the inspec-tor's review findings follows or is documented elsewhere as noted below:

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At 10:45 a.m. on April 17, 1986, the NRC was notified that on April 16,

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1986 the licensee apparently had violated the TS requirements for oper-able nuclear instrument channels on April 16, 1986. This event is dis-cussed in Section 7 of this report. The licensee made this notification due to a conservative interpretation of the NRC's reporting criteria, however, this event was not required to be reporte At 4:30 p.m. on May 13, 1986, the NRC was notified in accordance with 50.72(b)(2)(iii)(D) that the post accident hydrogen vent and analyzer systems required by TS Sections 3.6.3.2 and 3.6.3.1, respectively, were inoperable due to the discovery during surveillance testing that valve HV-50V-2 would not close. It was the intent of the licensee to report this event as an LER; however, a subsequent determination was made that the valve's performance did not create an inoperability for the afore-mentioned systems. A second ENS call was made at 8:17 a.m. on May 21, 1986 due to removal from service of the HV-50V-2 valve for maintenance on it's position indication features. The inspector reviewed the licen-see's detailed evaluation involving the effects the HV-SOV-2 valve can have on the operability of the systems involved, and identified no de-ficiencie At 9:05 p.m. on June 1, 1986, the NRC was notified in accordance with 50.72(b)(2)(ii) of an automatic reactor scram that occurred at 8:35 as a result of a low steam generator level. This event is discussed in Section 7 of this repor On June 2,1986 at 2:20 p.m. , the licensee notified the NRC via the ENS line that a security event involving loss of the security system's cen-tral and secondary alarm stations had occurred at 8:33 p.m. on June 1, 1986. Compensatory measures were instituted by the licensee. The system was returned fully to service and the compensatory measures terminated by 5:25 a.m. on June 2, 1986. The inspector determined that this event was a moderate loss of physical security effectiveness (i.e., a major loss of physical security effectiveness but properly compensated), with the licensee's appropriate use of the ENS in complying with the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> i

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reporting requirement stipulated in 10 CFR 73.71(c). In addition, the licensee submitted to NRC Region 1 their five day written report describ-ing the even No inadequacies were noted by the inspector in this area during the inspection perio . Review of Plant Events The licensee notified the NRC duty officer via the ENS at 10:45 a.m. on April 17, 1986 regarding the apparent violation of the Technical Speci-fications requirements for operable nuclear instrument channels on April 16, 1986. While making adjustnents to intermediate range (IR) channels Nos. 3 and 4 starting at 9:30 a.m. on the 16th, the requirements of Technical Specifications Table 3.3-1 were not met when technicians did not reset the power range high neutron level scram logic for noncoinci-dence operation. The plant was operating at full power at the time and was apparently in a mode less conservative than the LC0 for two consecu-tive one-hour periods as each channel was removed from service for ad-justment. The error was discovered and reported on April 17, 1986 fol-lowing a review of the operating log by plant management. Based on re-view of the event and plant design features, the inspector determined the following information and finding (i) Plant technicians performed a functional test of IR channels No and 4 per procedure OP 4601 in response to two maintenance re-quests initiated on April 15, 1986. MR 86-583 was submitted for IR channel No. 3 because the startup rate (SUR) meter on the main

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control board was drifting high in comparison to the SUR indication from the same channel on the FN cabinets. MR 86-582 was submitted because of an observed difference of 0.9 decades between IR channels Nos. 3 and 4. Both channels were returned to service following ad-justments on April 16, 198 (ii) The nuclear instrumentation channels are designed such that the com-pensated signal from two ion chambers is used as IR channels Nos.

! 3 and 4, and the uncompensated signal is used for intermediate power range (IPR) channels Nos. 3 and 4. The IR/ IPR channels are designed such that some testing can be done on the IR channel without affect-ing the IPR channel. However, when OP 4601 was completed on IR channels Nos. 3 and 4 on April 16th, the input signal cable from the ion chamber was disconnected per step B.10 of the procedure, which also caused IPR channels Nos. 3 and 4 to be inoperable. The technicians who worked on the IR channels notified the control room operators that the checks per OP 4601 would be completed, but the potential effect on the IPR channels was not discussed. Thus, ac-tions to meet the LCO for an inoperable IPR were not taken. The failure to place the IPR channels into a noncoincidence mode con-

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stituted a failure to meet the requirements of TS Table 3.3-1, Functional Item 2, Action .

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(iii) The licensee made a one-hour ENS notification about the event per administrative procedure OP-Memo-2A-1, which provides a conservative interpretation of the reporting criteria in 10 CFR 50.72. Failure to meet a Technical Specification action statement must be reported as an LER per 10 CFR 50.73, but is not required to be reported per 10 CFR 50.7 (iv) The Technical Specifications allow continued, unrestricted operation at power with one IR channel inoperable, as was the case on April 16th when IR channels Nos. 3 and 4 were removed from service sequentially to perform OP 4601. During the one-hour periods when 1 of 6 intermediate power range channels were also inoperable, de-sign of the RPS high power trip logic has sufficient redundancy that protection against an overpower condition was assured (remained single failure proof) even though the affected channels were not placed into the noncoincidence mode of operation. Based on the above, there was no safety concern that resulted from a failure to meet the technical specification requirement (v) There is some confusion by the plant staff regarding the proper in-terpretation of the TS Table 3.3-1, Action 2.b requirements. The interpretation used by plant management on April 17th was to enter the Technical Specifications action statement if any one of six power / intermediate power range channels was inoperabl Precaution No. 5 in OP 4601, Revision 16, requires entry into the action statement only if fewer than four power / intermediate power range channels are operable. OP 4601 should be revised to make it con-sistent with Technical Specifications interpretation used by plant managemen On May 16, 1986, the licensee reported the failure to comply with the TS action statement in LER 50-29/86-003. Personnel error was listed as the root cause of the occurrence. Corrective actions consisted of making the appropriate licensee personnel aware of the proper interpretation of the TS requirement and revising the precaution in OP-4601 accordingl The inspector determined that this item is considered a licensee identi-fied violation, in accordance with the criteria in 10 CFR 2, Appendix C and, therefore, no Notice of Violation will be issued. The inspector had no further comment on this ite At approximately 9:20 a.m. on May 30, 1986, the control room operators initiated an emergency load reduction due to a leak in the cooling water supply to the generator hydrogen cooler. A pregressively worsening leak in the supply line was observed earlier, and plant personnel were pre-paring to install an alternate cooling water supply arrangement when a total loss occurred at 10:02 a.m. The inspector was in the control room when the cooling water loss occurred. Procedures OP-3003, Rev. 10,

" Emergency Controlled Plant Load Reduction", and OP 2107, Rev. 12,

" Changing Generator Load" were in use at the time.

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The inspector observed a well-coordinated effort between the operations, i

maintenance, and reactor engineering departments in responding to the

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event. Operator response was excellent, with the flow of information and commands occurring freely between control room personnel. No TS restric-I tions involving thermal or control rod insertion limits were exceede <

l' Because a replacement pipe was available for immediate installation; the loss of cooling lasted for approximately six minutes. The shift super--

i visor maintained an appropriate overview and was ready to initiate a-turbine trip if generator hydrogen temperature conditions warranted such

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l actio One item of concern observed by the inspector regarded the lack of knowl- '

i edge by the subsequent shift's control room personnel of the outcome of

TS 3.4.7, Table 4.4.2 and TS 3.11.2 required surveillances. The isotopic i analysis of the main coolant and analysis of gaseous effluents are re-

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quired following a 15 percent per hour power change, as occurred during the earlier emergency load reduction. The inspector verified that the 1 analyses were performed, but chemistry department technicians did not report the results to control room personnel for inclusion in the control

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room lo The inspector's concern was discussed with the chemistry departr.ent man-

ager, who indicated that a written instruction to his personnel will be i issued requiring them to notify the control room verbally and to record the results on procedure form APF-9003.1, which is to be forwarded to the control room at the completion of the shift. Additionally, he indi-cated that procedure AP-4900, " Chemistry Surveillance Tests", will in-

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clude appropriate notification requirements. The inspector was satisfied j with the licensee's timely responsiveness and had no further questions.

i During a severe lightning storm, an automatic reactor scram occurred from ,

i 100% of rated power at 8:34 p.m. on June 1,1986. An electrical distur-

bance on the 120V a-c motor generator set distribution panel caused the .

! failure of the heater drain tank level control channel which tripped both j heater drain pumps (HOPS). This resulted in the loss of the three boiler

! feed pumps (BFPs) on low suction pressure. The loss of feedwater resulted j in a low steam generator level trip of the Reactor Protection Syste .

! Operator efforts to regain control of the HDPs and restart the BFPs were

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t j Plant equipment damaged by the storm included control room annunciators,

! the heater drain tank level control channel's power supply, a blown fuse i in No. 1 vital bus inverter, and certain security system electronics.

. The licensee will report the details of equipment damage in LER 50-29/

l 86-004, and Physical Security Event Report 86-2. Other than the afore-

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mentioned problems, the plant equipment and operator response to the i event were normal.

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A decision was made by plant management to maintain the plant in Mode 3 (Hot Standby) to conduct hot leak inspection in the vapor container, repack the No. 3 BFP, and repair equipment damaged by the storm. Subse-quently, at 1:10 a.m. on June 2, 1986, the primary coolant Dose Equvalent Iodine 131 (DEI) was determined to be 1.35 microcuries/ gra This condition is in excess of the TS 3.4.7 limit of 1.0 microcuries/ gra The DEI in the main coolant was reduced to approximately 75 percent of the TS limit by increasing purification flow from 26 to 48 GPM. The lic-ensee has attributed the condition to cladding defects in second cycle Core XVIII fuel assemblies. Exceeding the TS DEI limit will be reported by the licensee in LER 50-29/86-00 Plant startup occurred on June 3, 198 During a post-trip review of the control room log, the inspector noted that on June 2, 1986 at approximately 4:25 a.m., the control room opera-tors opened all four main coolant loop (MCL) bypass valves in an attempt to mitigate dropping MCL temperatures. At 8:47 a.m. on June 2, 1986 the MCL bypass valves were close The plant was in Mode 3 and all four main coolant pumps were running at the tim TS 3.4.1.1.2 stipulates that in Mode 3 that all MCLs shall be operable, with all loop isolation valves open and at least one main coolant loop shall be in operation. TS Surveillance 4.4.1.1.2.3 requires that the steam generators associated with the main coolant loops required to be in operation have their respective bypass valves verified closed once per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> The inspector conducted an interview of the licensed personnel who were on duty during the shift in question and noted that (1) they indicated they were unaware of the specific surveillance requirement while in Mode 3, and (2) were unable to direct the inspector to an applicable surveil-lance procedure that addressed the TS surveillance requirement in ques-tio The licensed operator actions in opening all four MCL bypass valves, when at least one should have been closed, raises concerns about TS training and knowledge, and the licensee's Surveillance Testing Program that is supposed to implement the TS requirements. Accordingly, the inspector

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has requested the licensee to provide a safety analysis or evaluation to ascertain if the operators' actions (1) negated any safety analysis assumptions, or (2) resulted in any actual safety concerns. This item remains unresolved pending the NRC review of the licensee analysis (UNR 86-05-01).

8. Maintenance Observations l The inspector observed and reviewed maintenance and problem investigation ac-tivities to verify compliance with regulations, administrative and maintenance procedures, codes and standards, proper QA/QC involvement, safety tag use, i

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equipment alignment, jumper use, personnel qualification, radiological con-trols for worker protection, fire protection, retest requirements and report-ability per Technical Specifications. The following activities were included:

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Maintenance Request (MR)86-434, Waste Gas Oxygen Analyzer Annunciator Power Supply

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MR 86-564, Spare Charging Pump Block Crack In Suction Valve Cover

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MR 86-544, No. 1 Charging Pump No. 3 Ram Pinhole In Block Below External Flange

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MR 86-583, Intermediate Range Channel No. 3 Start Up Rate Meter on MCB

& FN Panel Differ

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MR 86-587, TV-214 Pressurizer Relief Line System B Solenoid-Blows By When Actuated

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MR 86-592, Hydrogen Vent Valve HV-50V-2 Open/Close Indicator Lights Illuminate When Valve Closed

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MR 86-525, Tk 34-2 No. 2 Test Tank Inlet Line Split

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MR 86-717, Hydrogen Vent Valve HV-50V-2 Does Not Close

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MR 86-720, Heater Drain Tank Level Transmitter Troubleshooting Based upon a review of licensee activities in this area the inspector noted the following:

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Regarding MR 86-720, the licensee initiated maintenance activities to troubleshoot erratic operation of the heater drain tank's level trans-mitter on May 14, 1986. This activity necessitated the use of Lifted Lead Request (LLR) No. 86-82 issued on May 15, 1986 to lift the leads from the transmitter. As described in the LLR's Safety Analysis, compensatory actions were taken by operations personnel to compensate for the loss of the transmitters functions described in the FSAR. The inspector noted that the licensee's development of a thoroughly prepared safety evalu-ation that deals with the removal of equipment from service described in the FSAR is indicative of their willingness (particularly the I&C Department) to be responsive to past NRC concerns in this area. The lic-ensee is encouraged to continue this positive tren The inspector reviewed the licens m's repair activities associated with MR 86-564, which covered the weld repair of the spare charging pump bloc The license implemented AP-5116, Rev. 0,"Section XI Repair / Replacement Program of Class 1,2, or 3 Components and Their Supports," to provide the necessary guidelines on how the repair cycle would be performed. Ad-ditionally, procedure OP-2000,149, Rev. 0," Hydrostatic Pressure Test of the Spare Charging Pump Block," was developed to proof test the repai .

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The inspector observed the hydrostatic testing process, verified that the designated test personnel were certified on OP-MEM0 2EE-1, dated October 3, 1985, and were utilizing Yankee Nuclear Services Division Procedure YA-VT-11 to perform the required visual examinatio No violations were identifie . Surveillance Observations The inspector observed tests and parts of tests to assess performance in ac- ,

cordance with approved procedures and LCOs, test results (if completed), re-moval and restoration of equipment, and deficiency review and resolution. The following tests were reviewed: s ..

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OP-4272, Rev. 0," Accident Monitoring Instrumentation Channel Check"

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OP-4601, Rev. 15," Nuclear Instrumentation Channels Functional Test"

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OP-4216, Rev. 19," Testing of the Post Accident Hydrogen Venting System"

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OP-4202, Rev. 9, " Control Rod Operability Check"

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OP-4630, Rev. 8, " Accumulator Time Delay Actuation Verification"

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OP-4716, Rev. 6, " Vapor Container Personnel Hatch Leak Test" Based upon a review of the licensee activities in this area, the inspector noted the following: During the performance of OP-4601 on March 27, 1986, the surveillance identified that the channel No. 4 intermediate range start up rate scram and alarm was found to be 5.25 decades / minute. The administrative and TS limits are 5.0 + 0.1 and less than or equal to 5.2 decades / minute, t respectively. The discrepancy was reported to the I&C supervisor, shift '

supervisor, and plant management. The channel was adjusted to within ac-ceptable limits and tested satisfactorily. The inspector reviewed the equipment performance with the I&C department and determined that there is no history of setpoint drift on this channel. A review of the April '

16, 1986 test results on this channel by the inspector indicated that '

the channel performed satisfactory. At the time the channel was found inoperable, the TS did not require the channel to be operable. The in-spector had no further questions on this item, and noted that the licen-see had appropriately dispositioned the identified discrepanc The inspector reviewed the results of recent performances of OP-4716 with the reactor engineering manager. The inspector verified licensee adher-ence to TS surveillance requirements 4.0.2.a and 4.0.2.b. The procedure stated that a +25 percent time period may be added to the surveillance interval, but the 3.25 combined interval time in three consecutive in-tervals is not specified. The inspector recommended that at the next

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i procedure update the reactor engineering manager should consider adding the additional consideration and make additional documentation improve-

, ments in the procedure to reflect appropriate time and date informatio , The reactor engineering manager accepted for consideration the inspector's suggestions for procedure improvement The March 18, 1986 monthly surveillance testing results of OP-4630 indi-cated that the accumulator time delay relay, TDC 4, actuated in 12.26 seconds. The TS limit i:; 11.85 +/- 0.23 seconds. Immediate corrective action consisted of resetting the relay to within acceptable limits. The inspector reviewed the licensee's performance trending charts maintained by the I&C department for this surveillance, and noted that this was the first indication of exceeding the TS limit since the relay was replaced on July 16, 1985. The surveillance test results of April 15, 1986 did not indicate that there was a repetitive setpoint drift problem. The data presentation of the performance trending charts displayed the monthly results in a manner which made even minor deviations easily recognizabl No violations were identifie . Review Periodic and Special Reports Periodic and special reports submitted to the NRC pursuant to Technical Specification 6.9 were reviewed. The review ascertained: inclusion of infor-mation required by the NRC including test results and/or supporting informa-tion; consistency with design predictions and performance specifications; adequacy of planned corrective action for resolution of problems; determina-tion whether any information should be classified as an abnormal occurrence; and validity of reported information. The following periodic reports were reviewe Monthly Statistical Reports submitted per TS 6.9.3 for the months of October 1985 through May 198 Special report submitted per TSs 4.4.10.5 b and 6.9.6 containing the re-sults of the steam generator tube inservice inspection (ISI) completed in October, 1985, FYR 86-041, dated April 22, 198 Annual Report submitted per TS 6.9.2.b and 10CFR50.59(b), FYR 86-026,

dated February 28, 1986.

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1985 Annual Personnel Exposure Report by Duty Function submitted per TS

! 6.9.2.a, FYR 86-024, dated February 28, 1986 (

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Semiannaul Effluent Release Report submitted per TS 6.9.5.b, FYR 86-025,

dated February 28, 1986

.l In addition, a review was conducted by a region-based specialist inspector j of the licensee's Annual Radiological Environmental Report for 1985 submitted j per TS 6.9.5.a. FYR 86-047, dated April 30, 1986. As a result of this review, i

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the inspector determined that the licensee has complied generally with the Technical Specifications requirements for sampling frequencies, types of measurements, analytical sensitivities, and report schedules. The report in-

cluded summaries of the laboratory quality assurance program and of the land 1 use survey. The analyses of environmental samples indicated that doses to

! humans from radionuclides of station origin were negligible.

. No inadequacies were identifie ,

11. Onsite Review Committee

On May 6, 1986 the inspector observed the meeting of the Yankee NPS onsite

! review committee to ascertain that the provisions of TS 6.5.1 were met.

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9 No unacceptable conditions were identifie l 12. Annual Retrainina

1 The inspector completed the Yankee General Employee Training program for NRC

,' inspectors on April 16, 1986. The inspector noted that the course presenta-  !

tion and instructional material was very good, and provided adequate coverage

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of site specific information needed for unescorted access in plant radiologi-cal areas. The inspector suggested that the licensee consider modifying the

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l presentation during the next revision of the program to recognize the response  :

role of NRC inspectors during an emergency at the site. The licensee noted the inspector's comments. No inadequacies were identified.

l l 1 Siemens Allis BQ Frame Circuit Breakers The inspecter discussed with the maintenance manager information recently re-ceived by NRC Region I regarding Siemens Allis BQ frame circuit breaker The Yankee plant was identified by the breaker-vendor as a user of BQ breakers

that have operating characteristics potentially different than that specified  !

l by the time-current curves previously supplied by the vendor. Time-current l curves for the circuit breakers produced in a Wilmington, N.C. plant since

_ October, 1985 are different than the curves previously supplied for the same i breaker when it was produced in a Philadelphia plant in the 1979-1985 time period. The different breaker data could cause a coordination problem if the breakers are used in a control circuit in conjunction with other equipmen i

The inspector requested the maintenance manager to de'. ermine whether any of j the subject breakers are in use in safety related applications at Yankee, and,

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if so, to identify what actions are necessary to preclude a problem based on j the information supplied by the vendo The licensee stated that BQ circuit breakers of 15 amp size are used on emer-gency motor control centers Nos. 3 and 4. Further licensee review is needed to determine whether the breakers are also in use on the recently installed

MCC's (panel CDP) in the Appendix R alternate shutdown system buildin i

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m , , , ,w .--~y ---v~, - - . , , - - - . ~ ., - - - - - - - + --v<,,.~,w +e--,--,-- m----n.~---- e+=.-*-+-a-~~-~~------~~r- -*'--r

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The licensee stated that the breakers are only used in the above referenced 120V a-c lighting circuits at the plant, and that any problems with the vendor supplied time current information would not be a problem at Yankee since it was believed that the station breaker coordination studies stop at the 480V level in the electrical distribution system. The licensee stated further that his engineering organization received notification of the potential problems from the vendor, and had also received corrected time-current curve The licensee stated that additional actions would be taken by station person-nel (1) to verify the coordination study does not include 120V a-c components, and (2) to verify that the safe shutdown panels use either.ET or EQ breakers, and not BQ's. The results of the licensee's review will be examined by the inspector during a subsequent routine inspectio . Licensee Plans for Coping with Strikes 3 The inspectors conducted a detailed review of the licensee's preparation for I a possible strike of nonexempt employees at 12:01 a.m. , June 1,1986. The j inspectors reviewed the qualifications of personnel assigned to cover vacan-

cies created in the operations, maintenance, health physics, chemistry, in-strument and control, reactor engineering, and plant management staffs. The licensee had used the requirements of ANSI N18.1-1971, 10 CFR 55 (licensed operator requalification), the Job Description and Qualifications Manual, plant training procedures, and the facility Technical Specifications as quali-

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fication criteria. The licensee documented their intended actions in a de-i tailed Strike Contingency Plan.

The inspectors also verified that the licensee had reviewed and prepared for

the requirements for manning and qualification in the areas of fire brigade, emergency planning response, and security. Following the completion of con-l tract negotiations on June 4, 1986, the nonexempt employees ratified a two

! year contract on June 13, 1986. The inspector observed no inadequacies in the i

licensee's plans for coping with a potential strike situatio Additionally, the inspector conducted a detailed review of the licensee's i security contractor's ability to cope with a strike of plant security person-l nel. This review action was necessitated by the expiration on April 10, 1986 of the contract between the licensee's security contractor - Green Mountain Security Co. and the union representing their security personnel. The inspec-tor determined that strike contingency planning activities were established and identified no inadequacies. At the end of this inspection period, the security contractor / union negotiations were continuin . Management Meetings l' During the inspection period, the following management meetings were conducted or attended by the inspector as noted below:

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Exit meeting held on April 23, by an NRC:NRR audit team review of the !

design, installation, and operating aspects of the licensee's safety parameter display system required by TMI Action Plan Item I. Exit meeting held on May 22, 1986, by a region-based operator licensing examiner at the conclusion of operator licensing examination 50-29/86-07 to discuss results and identify strengths / weaknesses in the licensee's Operator Licensing Training Progra At periodic intervals during the course of the inspection period, meet-ings were held with senior facility management to discuss the inspection scope and preliminary findings of the resident inspecto . _ . .- . _ _ _ _ , - . - - _ _ . . - _ .- .. - - - . - -